ML13325A855

301 Initial Exam Draft Administrative Documents

ML13325A855
Person / Time
Site:	Harris
Issue date:	11/21/2013
From:	NRC/RGN-II
To:	Duke Energy Progress
References
50-400/13-301
Download: ML13325A855 (76)

Text

ES-401 PWR Examination Outline Form ES-4O1 Facility: SHEARON HARRIS Date of Exam: SEPTEMBER 2013 RO K/A CatoyPts_ SRO-Only_Points Tier Group 1 1 KIK K K KKAAAAG A2 I G Total 112 3 4 5 6 1 2 3 4

• Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency &

Abnormal 2 L__ .L N/A 2 2 N/A 1 9 2 2 4 Evolutkns Tier Totals 4 4 5 5 5 4 27 5 5 10 1 22332333232 28 3 2 5 2.

Plant 2 1 ii 111 011 11 10 0 2 1 3 Systems Tier Totals 3 3 4 4 3 4 3 3 4 3

+/- 38 5 3 8

3. Generic Knowledge and Abilities 1 2 3 4 10 1 2 3 4 7 Categories 3 2 2 3 1 2 2 2 Note: 1. Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each K/A category shall not be less than two).

2. The point total for each group and tier in the proposed outline must match that specified in the table.

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

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

3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate K/A statements.

4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.

5. Absent a plant-specific priority, only those K/As having an importance rating (lR) of 2.5 or higher shall be selected.

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

6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.

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

8. On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.

9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES-401 -3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.

ES-401 2 Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergenr and Abncsrm& Plant Evolutions Tier 1/Group 1 (RO

- /()

E/APE #1 Name / Safety Function K K K A A G K/A Topic(s) =lR #

12312 000007 (BW/E02&E10; CE/E02) Reactor R 0D7 EI<2.O2 Trip - Stabilization - Recovery / 1 K

008 MCLO.

000008 Pressurizer Vapor Space Accident / 3 k R0091(2.03 000009 Small Break LOCA / 3 P . Oil EG2.4.47 000011 Large Break LOCA/3 w

R Q5 Xl.a.

000015/17 RCP Malfunctions / 4 P. O22AA1.Q 000022 Loss of Rx Coolant Makeup / 2 000025 Loss of RHR System/4 S O2 AA2.oc 000026 Loss of Component Cooling O22 AAI. as-Water/8 000027 Pressurizer Pressure Control System Malfunction / 3 S P. G29Ej.i.OI 5 oI iA 000029 A1WS/1 C3S 4-.3c 000038 SteamGen.Tube Rupture/3 000040 (BW/E05; CE/E05;W/E12) o4v M3.O3 Steam Line Rupture Excessive Heat Transfer/4 000054(CE/E06)LossofMain g S R a54APa.OL.

54At 4.47 Feedwater/4 S 5 A.ø4 000055 Station Blackout/6 QsAa.+.45 000056 Lossof Off-site Power/6 000057 Loss of Vital AC Inst. Bus / 6 R R 05P AA2. .01 S OS 000058 Loss of DC Power/6 P. P OP. AA2.ol 000062 Loss ofNuclearSvcWater/4 000065 Loss of Instrument Air/8 P. Q(.5 AK3.Og WE.o4 El4..2.

W/E04 LOCA Outside Containment / 3 P.

WEll EAt .3 W/E1 I Loss of Emergency Coolant Recirc. / 4 BW/EO4nadequate Heat P. WEDS £141.2.

Transfer Loss of Secondary Heat Sink / 4 000077 Generator Voltage and Electric R P. 077 A 2.4.4 Grid Disturbances /6 K/A Category Totals: jj j3 S 3 3 Group Point Total: = 61

ES-401, REV 9 a

T1G1 PWR EXAMINATION OUTLINE FORM ES-401 -2 KA NAME I SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 007EK2.02 Reactor Trip Stabilization

- - Recovery 2.6 2.8 Breakers, relays and disconnects

/1 008AK2.02 Pressurizer Vapor Space Accident / 3 2.7 2.7 Sensors and detectors 009EK2.03 Small Break LOCA /3 3 3.3 S/Gs 011 EG2.4.47 Large Break LOCA / 3 4.2 4.2 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

01 5AK1 .02 RCP Malfunctions / 4 3.7 4.1 Consequences of an RCPS failure 022AA1 .08 Loss of Rx Coolant Makeup / 2 3.4 3.3 VCT level 026AA1 .05 Loss of Component Cooling Water / 8 3.1 3.1 The CCWS surge tank, including level control and level alarms and radiation alarm 029EK1 .01 ATWS / 1 2.8 3.1 Reactor nucleonics and thermo-hydraulics behavior f/et ,f

/46UV1 i W (fr 040AK3.03 Steam Line Rupture - Excessive Heat 3.2 3.5 Steam line non-return valves Transfer / 4 054AA2.02 Loss of Main Feedwater / 4 4.1 4.4 Ditferentiation between loss of all MEW and trip of one MEW pump 056AG2.4.45 Loss of Oft-site Power / 6 4.1 4.3 E LI LI Ability to prioritize and interpret the significance of each annunciator or alarm.

Page 1 of 2 12/18/2012 2:20PM

ES-401, REV 9 T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 058AA2.0l Loss of DC Power / 6 3.7 4.1 [] El El That a loss of dc power has occurred; verification that substitute power sources have come on line 062AA2.0l Loss of Nuclear Svc Water! 4 2.9 3.5 Location of a leak in the SWS 065AK3.08 Loss of Instrument Air / 8 3.7 3.9 El El El El Actions contained in EOP for loss of instrument air 077AG2.4.4 Generator Voltage and Electric Grid 4.5 El El El El El El El El El El Ability to recognize abnormal indications for system Disturbances / 6 operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

WE04EK3.2 LOCA Outside Containment! 3 3.4 4.0 El El El El El El El El El El Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).

WEO5EK1 .2 Inadequate Heat Transfer Loss of

- 3.9 4.5 Normal, abnormal and emergency operating procedures Secondary Heat Sink / 4 associated with (Loss of Secondary Heat Sink).

3.7 LI El El El Desired operating results during abnormal and WE1 1 EA1 .3 Loss of Emergency Coolant Recirc. / 4 4.2 El El El El El El emergency situations.

Page 2 of 2 12/18/2012 2:20PM

ES-401, REV 9 SRO T1G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETYFUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 025AA2.05 Loss of RHR System / 4 3.1 3.5 El El El El El El El ] El El El Limitations on LPI flow and temperature rates of change 029EA2.09 ATWS / 1 4.4 4.5 ] [j El El Occurrence of a main turbine/reactor trip 0382G2.4.30 Steam Gen. Tube Rupture / 3 2.7 4.1 Knowledge of events related to system operations/status that must be reported to internal orginizations or outside agencies.

054AG2.4.47 Loss of Main Feedwater! 4 4.2 4.2 El El El El El El El El El El Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

055EA2.04 Station Blackout / 6 3.7 4.1 ElElElElElElElElElEl Instruments and controls operable with only dc battery power available 058AG2.4.3 Loss of DC Power! 6 ElElElElElEl Ability to identify post-accident instrumentation.

Page 1 of 1 12/18/2012 12:58PM

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ES4O1, REV 9 T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRQ OO1AA2.0l Continuous Rod Withdrawal / 1 4.2 4.2 El El El El El El El El El El Reactor tripped breaker indicator 036AK2.0l Fuel Handling Accident / 8 2.9 El El El El El El El El El El Fuel handling equipment 037AG2.4.34 Steam Generator Tube Leak / 3 4.2 4.1 El El El El El El El El El El Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects 051 AA1 .04 Loss of Condenser Vacuum / 4 2.5 2.5 El El El El El El [] El El El El Rod position 059AK3.04 Accidental Liquid Rad Waste Rel. / 9 3.8 4.3 Actions contained in EOP for accidental liquid radioactive-waste release 074EA2.03 Inad. Core Cooling! 4 3.8 4.1 El El El El El El El El El El Availability of turbine bypass valves for cooldown WEO3EK1 .1 LOCA Cooldown - Depress. /4 3.4 4.0 El El El El El El El El El El Components, capacity, and function of emergency systems.

WE14EA1.l LossofCTMTlntegrity/5 3.7 3.7 El El El El El El El El El El Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

WE15EK3.l Containment Flooding!5 2.7 2.9 EEJElElElElElEElEl Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure and reactivity changes and operating limitations and reasons for these operating characteristics.

Page 1 of 1 12118/2012 12:58PM

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ES-401, REV 9 SRO T1G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

no SRO 005AA2.03 Inoperable/Stuck Control Rod / 1 3.5 4.4 Required actions if more than one rod is stuck or inoperable 060AG2.2.37 Accidental Gaseous Radwaste Rel. / 9 3.6 4.6 LZ E D Ability to determine operability and/or availability of safety related equipment 061 AA2.02 ARM System Alarms / 7 2.9 3.2 LZ D El El El El Normal radiation intensity for each ARM system channel 069AG2.2.25 Loss of CTMT Integrity / 5 3.2 4.2 El El El El El El El El El El Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Page 1 of 1 12/18/2012 12:58PM

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ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IA Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 003A2.02 Reactor Coolant Pump 3.7 3.9 Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP 004A2.13 Chemical and Volume Control 3.6 3.9 Low RWST 004K5.26 Chemical and Volume Control 3.1 3.2 Relationship bOtween VCT pressure and NPSH for charging pumps 005A1 .05 Residual Heat Removal 3.3 3.3 Detection of and response to presence of water in RHR emergency sump 006G2.1 .30 Emergency Core Cooling 4.4 4.0 Ability to locate and operate components, including local controls.

007G2.l .20 Pressurizer Relief/Quench Tank 4.6 4.6 Ability to execute procedure steps.

007K4.0l Pressurizer Relief/Quench Tank 2.6 2.9 Quench tank cooling 008A3.08 Component Cooling Water 3.6 3.7 Automatic actions associated with the CCWS that occur as a result of a safety injection signal 010K6.02 Pressurizer Pressure Control 3.2 3.5 PZR 010K6.03 Pressurizer Pressure Control 3.2 3.6 PZR sprays and heaters 012K4.02 Reactor Protection 3.9 4.3 fl D D El El El Automatic reactor trip when FtPS setpoints are exceeded for each RPS function; basis for each Page 1 of 3 12/18/2012 12:58PM

ES-401, REV 9 Ti PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 013A4.01 Engineered Safety Features Actuation 4.5 4.8 ESFAS-initiated equipment which fails to actuate 013K201 Engineered Safety Features Actuation 3.6 3.8 ESFAS/safeguards equipment control 022A4.03 Containment Cooling 3.2 3.2 Dampers in the CCS 026A1.04 Containment Spray 3.1 3.3 Containment humidity 026K3.02 Containment Spray 4.2 4.3 Recirculation spray system 039A2.01 Main and Reheat Steam 3.1 3.2 Flow paths of steam during a LOCA 059K4.02 Main Feedwater 3.3 3.5 Automatic turbine/reactor trip runback 061A 1.01 Auxiliary/Emergency Feedwater 3.9 4.2 ED DDE LID LI LI LI S/G level 061 K5.05 Auxiliary/Emergency Feedwater 2.7 3.2 DEDDDDDDDD Feed line voiding and water hammer 062K2.01 AC Electrical Distribution 3.3 3.4 DDDDDDD Major system loads 063A3.01 DC Electrical Distribution 2.7 3.1 LI LI LI LI LI LI LI LI LI LI Meters, annunciators, dials, recorders and indicating lights Page 2 of 3 12/18/2012 12:58PM

n ES-401, REV 9 T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 063K3.02 DC Electrical Distribution 3.5 3.7 Components using DC control power 064K6.08 Emergency Diesel Generator 3.2 3.3 Fuel oil storage tanks 073A4.02 Process Radiation Monitoring 3.7 3.7 U U U U U U U Radiation monitoring system control panel 076K1 .01 Service Water 3.4 3.3 U U U U U U U U U U CCW system 078K3.0l Instrument Air 3.1 3.4 U U U U U U U U U U Containment air system 103K1.03 Containment 3.1 3.5 U U U U U U U U U U Shield building vent system Page3of3 12/18/2012 12:58PM

ES-401, REV 9 SRO T2G1 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME/SAFETY FUNCTION: IR Ki K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 039A2.03 Main and Reheat Steam 3.4 3.7 Indications and alarms for main steam and area radiation monitors (during SGTR) 062A2.ll AC Electrical Distribution 3.7 4.1 Aligning standby equipment with correct emergency power source (DIG) 063G2.2.40 DC Electrical Distribution 3.4 4.7 Ability to apply technical specifications for a system.

076A2.Ol Service Water 3.5 3.7 LI LI LI LI LI Loss of SWS 103G2.2.22 Containment 4.0 4.7 LI LI LI LI LI LI LI LI LI LI Knowledge of limiting conditions for operations and safety limits.

Page 1 of 1 12118/2012 12:58PM

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ES-401, REV 9 T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IA Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO 011 K2.02 Pressurizer Level Control 3.1 3.2 PZR heaters 015K3.0l Nuclear Instrumentation 3.9 4.3 RPS 016A2.02 Non-nuclear Instrumentation 2.9 3.2 D D D E E Loss of power supply 028K6.0l Hydrogen Recombiner and Purge 2.6 3.1 Hydrogen recombiners Control 034G2.4.31 Fuel Handling Equipment 4.2 4.1 j Knowledge of annunciators alarms, indications or response procedures 071 K5.04 Waste Gas Disposal 2.5 3.1 Relationship of hydrogen/oxygen concentrations to flammability 072A3.01 Area Radiation Monitoring 2.9 3.1 E Changes in ventilation alignment 075K1 .01 Circulating Water 2.5 2.5 SWS 079A4.0l Station Air 2.7 2.7 El El El El El El El El Cross-tie valves with lAS 086K4.02 Fire Protection 3.0 3.4 El El El El El El El El El El Maintenance of fire header pressure Page 1 of 1 12/18/2012 12:58PM

n fl *fl ES-401, REV 9 SRO T2G2 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SAC 01 7G2.l .7 In-core Temperature Monitor 4.4 4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior and instrument interpretation.

028A2.02 Hydrogen Recombiner and Purge 3.5 3.9 D E LOCA condition and related concern over hydrogen Control 068A2.04 Liquid Radwaste 3.3 3.3 Failure of automatic isolation Pagelofi 12/18/2012 12:58PM

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 L Facility: -rrV1FL4 Date of Exam: Sep.k...J4L a 013 Category K/A # Topic IR #

2.1. 1. 1 Fj A2j4S VfO4/cQ&t44 1 2. 1. 15 t z. q Conduct 2.1.19 L(o*L9. 3.9 of Operations Subtotal 2) 2.2. 2D Pr 2.2.39 r1a Equipment

° Subtotal 2.3. fl Cr4,.. rz4i4t 2.3.??.. i?z* prn( ..r+ 1kt4P1 c4cL1e4- 3.2 Radiation

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ES-401, REV 9 T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME I SAFETY FUNCTION: IR Kl K2 K3 K4 K5 K6 Al A2 A3 A4 G TOPIC:

RO SRO

. I I I I I I I I I I G2.l.13 Conduct of operations 2.5 3.2 Knowledge of facility requirements forcontrolling vital!

controlled access.

G2.l.15 Conductof operations 2.7 3.4 J U U U U U U U U Knowledgeof administrative requirementsfortemporary management directives such as standing orders, night orders, Operations memos, etc.

G2.l .19 Conduct of operations 3.9 3.8 Ability to use plant computer to evaluate system or component status.

G2.2.20 Equipment Control 2.6 3.8 Knowledge of the process for managing troubleshooting activities.

G2.2.39 Equipment Control 3.9 4.5 Knowledge of less than one hour technical specification action statements for systems.

G2.3.1 1 Radiation Control 3.8 4.3 Ability to control radiation releases.

G2.3.12 Radiation Control 3.2 3.7 U Knowledge of radiological safety principles pertaining to licensed operator duties G2.4.22 Emergency Procedures/Plans 3.6 4.4 Knowledge of the bases for prioritizing safety functions during abnormal/emergency operations.

G2.4.27 Emergency Procedures/Plans 3.4 3.9 U U U U U U U U U U Knowledge of fire in the plant procedures.

G2.4.29 Emergency Procedures/Plans 3.1 4.4 U U U U U U U U U U Knowledge of the emergency plan.

Page 1 of 1 12/18/2012 12:58PM

ES-401, REV 9 SRO T3 PWR EXAMINATION OUTLINE FORM ES-401-2 KA NAME / SAFETY FUNCTION: IR K1 K2 K3 K4 K5 k6 Al A2 A3 A4 G TOPIC:

RO sR0 G2.l .41 Conduct of operations 2.8 3.7 Knowledge of the refueling processes G2.2.13 Equipment Control 4.1 4.3 Knowledge of tagging and clearance procedures.

G2.2.40 Equipment Control 3.4 Ability to apply technical specifications for a system.

G2.3.14 Radiation Control 3.4 3.8 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities G2.3.4 Radiation Control 3.2 3.7 Knowledge of radiation exposure limits under normal and emergency conditions G2.4.40 Emergency Procedures/Plans 2.7 4.5 Knowledge of the SROs responsibilities in emergency plan implementation.

G2.4.46 Emergency Procedures/Plans 4.2 4.2 Ability to verify that the alarms are consistent with the plant conditions.

Page 1 of 1 12/18/2012 12:58PM

ES-301 Administrative Topics Outline Form ES-301-l Facility: Harris Nuclear Plant Date of Examination: September 9, 2013 Examination Level: RO SRO LI Operating Test Number: 05000400/2013301 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine Rod Misalignment Using Thermocouples (JPM ADM-062-c)

Conduct of Operations R

, K/AG21.7 2013 NRC RO I SRO Al-I Determine Average RCS Boron Concentration per EOP-ECA-0. 1 (JPM ADM-020-a) Common Conduct of Operations D, R K/A G 2.120 2013 NRC RO I SRO AI-2 Perform a Quadrant Power Tilt Ratio (QPTR) calculation with a control rod misaligned.

(JPM ADM-010-e)

Equipment Control M,R K/AG2.2.12 2013 NRC RO A2 Using Valve Maps And HP Room Survey Maps Determine Stay Time During Refueling.

(JPM ADM-065-a) Common Radiation Control N, R K/A G2.3.4 2013 NRC RO I SRO A3 NOT SELECTED FOR RO Emergency Procedures/Plan N/A 2013 NRC RO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (4)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (2)

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

(P)revious 2 exams ( 1; randomly selected) (1) 03/25/2013 Rev. 0

2013 NRC RO Admin JPM Summary 2013 NRC RO Al-I Determine Rod Misalignment Using Thermocouples (JPM ADM-062-c) Previous 2011 NRC Exam JPM *randomly selected from bank K/A G2. 1.7- Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5 / 45.12 / 45.13) RO 4.4 SRO 4.7 The plant is at 90% power with a load reduction in progress when a control rod is observed indicating 24 steps higher than group demand. The candidate must perform Attachment 2 of AOP-001, Malfunction of Rod Control and Indication System, to calculate the temperature difference between the affected thermocouple and its symmetric thermocouples. For ROs this JPM requires the candidate to identify the control rod is misaligned and notifies the CRS.

2013 NRC RO AI-2 (Common)- - Determine Average RCS Boron Concentration per EOP-ECA-0. I (JPM ADM-020-b)

K/A G2. 1.20- Ability to interpret and execute procedure steps.

(CFR: 41.10/43.5/45.12) RO 4.6/SRO 4.6 The candidate must perform a calculation to determine average RCS boron concentration in order to complete a Shutdown Margin calculation as required by EOP-ECA-0.1, Loss Of All AC Power Recovery Without SI Required. The candidate is provided a list of plant conditions and is required to calculate the average RCS boron concentration for these conditions lAW EOP-ECA-0.1, Attachment 1.

2013 NRC RO A2 Perform a Quadrant Power Tilt Ratio (QPTR) calculation with a control rod misaligned.

(JPM ADM-010-e) MODIFIED K/A G2. 2.12 Knowledge of surveillance procedures.

(CFR: 41.10/45.13) RO 3.7 SRO 4.1 The candidate must perform a QPTR calculation in accordance with surveillance procedure OST-1039, Calculation of Quadrant power Tilt Ratio, Weekly Interval and as required by the AOP-001, Malfunction of Rod Control and Indication System for a misaligned rod at 95%

power. For SROs this JPM requires the candidate to identify applicable Tech Spec LCOs.

NOTE: This JPM will be modified by changing the initial reactor power, the control rod that is dropped into the reactor, and the values of the PRNI upper and lower detectors. These changes result in the QPTR value that exceeds 1.09. The Tech Spec action is now different due to the value exceeding 1.09.

I 2

03/25/2013 Rev. 0

2013 NRC RO Admin JPM Summary (continued) 2013 NRC RO A3 (Common) Using Valve Maps And HP Room Survey Maps Determine Stay Time During Refueling.

(JPM-ADM-065-a) NEW K/A G2. 3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12/43.4/45.10) RO 3.2/SRO 3.7 The candidate will be supplied a survey map of a location in the RAB containing various radiation levels and several hot spots. In this area work must be performed by a refueling team shared resource AO. The AD will be required to have continuous HP coverage. The candidates will also have a copy of HP administrative procedures to use during this JPM.

The candidate must determine the individual stay times for work in the area. The candidate should determine that the HP Technician cannot stay long enough to complete the task without either exceeding the RWP maximum radiation dose or the administrative yearly dose limit. Since the HP coverage cannot be provided for the work to be completed the job will have to be stopped before it is complete.

2013 NRC RO A4 Not selected 3

03/25/20 13 Rev. 0

ES-301 Administrative Topics Outline Form ES-301-l Facility: Harris Nuclear Plant Date of Examination: September 9, 2013 Examination Level: RO SRO Operating Test Number: 05000400/2013301 Administrative Topic Type Describe activity to be performed (see Note) Code*

Determine Rod Misalignment Using Thermocouples and Evaluate Tech Specs

. (JPM ADM-062-d)

Conduct of Operations p, R K/A G2.1.7 2013 NRC ROISRO Al-I Determine Average RCS Boron Concentration per EOP-ECA-0. 1 (JPM ADM-020-b) Common Conduct of Operations D R

, K/AG2.1.20 2013 NRC ROISROAI-2 Perform a Quadrant Power Tilt Ratio (QPTR) calculation with a control rod misaligned and Evaluate Tech Specs.

(JPM ADM-010-f)

Equipment Control M,R K/AG2.2.12 2013 NRC SRO A2 Using Valve Maps And HP Room Survey Maps Determine Stay Time During Refueling.

(JPM ADM-065-a) Common Radiation Control N, R K/A G2.3.4 2013 NRC RO I SRO A3 Given a Set of Plant Conditions, Classify an Event.

(JPM ADM-064-a)

Emergency Procedures/Plan N, R K/A G2.4.41 2013 NRC SRO A4 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (5)

(D)irect from bank ( 3 for ROs; 4 for SROs & RO retakes) (2)

(N)ew or (M)odified from bank ( 1) (3)

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

I 03/25/2013 Rev.O

\)

2013 NRC SRO Admin JPM Summary 2013 NRC SRO Al-I Determine Rod Misalignment Using Thermocouples and Evaluate Tech Specs (JPM ADM-062-d) Previous 2011 NRC Exam JPM *randomly selected from bank K/A G2. 1.7- Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(CFR: 41.5 / 43.5/45.12/45.13) RO 4.4 SRO 4.7 The plant is at 90% power with a load reduction in progress when a control rod is observed indicating 24 steps higher than group demand. The candidate must perform Attachment 2 of AOP-001 Malfunction of Rod Control and Indication System, to calculate the temperature difference between the affected thermocouple and its symmetric thermocouples. For SROs this JPM requires the candidate to identify applicable Tech Spec LCOs.

2013 NRC SRO A1-2 (Common) - - Determine Average RCS Boron Concentration per EOP-ECA-0. 1 (JPM ADM-020-b) DIRECT K/A G2. 1.20 Ability to interpret and execute procedure steps.

(CFR: 41.10/43.5/45.12) RO 4.6/SRO 4.6 The candidate must perform a calculation to determine average RCS boron concentration in order to complete a Shutdown Margin calculation as required by EOP-ECA-0.1, Loss Of All AC Power Recovery Without SI Required. The candidate is provided a list of plant conditions and is required to calculate the average RCS boron concentration for these conditions lAW EOP-ECA-0.1, Attachment 1.

2013 NRC SRO A2 Perform a Quadrant Power Tilt Ratio (QPTR) calculation with a control rod misaligned and Evaluate Tech Specs (JPM ADM-01 0-f) MODIFIED K/A G2.2. 12 Knowledge of surveillance procedures.

(CFR: 41.10/45.13) RO 3.7 SRO 4.1 The candidate must perform a QPTR calculation in accordance with surveillance procedure OST-1 039, Calculation of Quadrant power Tilt Ratio, Weekly Interval and as required by the AOP-001, Malfunction of Rod Control and Indication System for a misaligned rod at 95%

power. For SROs this JPM requires the candidate to identify applicable Tech Spec LCOs.

NOTE: This JPM will be modified by changing the initial reactor power, the control rod that is dropped into the reactor, and the values of the PRNI upper and lower detectors. These changes result in the QPTR value that exceeds 1.09. The Tech Spec action is now different due to the value exceeding 1.09.

2 03/25/2013 Rev 0

2013 NRC SRO Admin JPM Summary (continued) 2013 NRC SRO A3 (Common) Using Valve Maps And HP Room Survey Maps Determine Stay Time During Refueling.

(JPM-ADM-065-a) NEW K/A G2. 3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12/43.4/45.10) RO 3.2 SRO 3.7 The candidate will be supplied a survey map of a location in the RAB containing various radiation levels and several hot spots. In this area work must be performed by a refueling team shared resource AC. The AC will be required to have continuous HP coverage. The candidates will also have a copy of HP adniirr[strative procedures to use during this JPM.

The candidate must determine the individual stay times for work in the area. The candidate should determine that the HP Technician cannot stay long enough to complete the task without either exceeding the RWP maximum radiation dose or the administrative yearly dose limit. Since the HP ooverage cannot be provided for the work to be completed the job will have to be stopped before it is complete.

2013 NRC SRO A4 Given a set of conditions, Classify an Event (JPM-ADM-064-a) NEW K/A G2. 4.41 Knowledge of the emergency action level thresholds and classifications (CFR: 41.10/43.5/45.1 1) RO 2.9 SRO 4.6 Given a set of initial conditions and the EAL Flow Path, the candidate must classify the appropriate Emergency Action Level for the event in progress.

3 03/25/2013 Rev. 0 N

ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2 Facility: Harris Nuclear Plant Date of Examination: 09/09/2013 Exam Level: RO SRO-l SRO-U (bold) Operating Test No.: 05000400/2013301 Control Room Systems@ (8 for RO); (7 for SRO-l); (2 or 3 for SRO-U, including I ESF bold)-

System / JPM Title Type Code* Safety Function

a. Perform Control Rod and Rod Position Indicator Exercise per OST-I 005 (JPM-CR-256-d) A, N, S I KIA 001 A2.1I

b. Respond to the loss of the running CSIP (JPM-CR-038-a)

A D S 2 K/A APE 022 AA1.01

c. Pressurizer Pressure Master Controller Failure (AOP-019)

(JPM-CR-252-a)

A, P, 5 3 K/A APE 027 AAI.03

d. Loss of Power to the TDAFW Pump control system (E-0 A, EN, L, N, S 4S and OP-I 37)

(JPM-CR-28I -a)

K/A 054 AA2.04

e. Return the Containment Fan Coolers to normal following an SI actuation. (OP-I 69)

(JPM CR-260-a) RO Only D, EN, L, S 5 k/A 022 A4.01

f. Restore Off-site Power to an Emergency Bus (OP-156.02)

(JPM-CR-027-b)

A, 0, EN, S 6 K/A 062 A4.01

g. Restore an Excore NI Channel to service (at power, NI failed low) (OWP-RP-25)

(JPM-CR-278-a) N, S 7 K/A 015 A4.03

h. Align CCW to Support RHR System (OP-145)

(JPM CR-085-a)

K/A 008 A4.01 D, L, S 8 03/25/2013 Rev 0 1 /

In-Plant Systems@ (3 for RO); (3 for SRO-I); (2 or 3 for SRO-U - BOLD)

i. Place the ASI System in Standby Alignment (OP-I 85)

(JPM-IP-277-a)

L, N, R 2 K/A 004 A4.11

j. Local Inspection of Annunciator Cabinets (AOP-037)

(JPM IP-273-a)

D E 7 K/A 016A2.02

k. Perform an Instrument Air System Leak Isolation Locally (Turbine Bldg I Yard)

(JPM-IP-161-a) D, E, L 8 K/A APE 065 AA2.03

@ 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-l / SRO-U (A)lternate path 4-6 / 4-6 I 2-3 (5, 5, 3)

(C)ontrol room (D)irect from bank 2.

9 / 8 Is 4 (7, 6, 2)

(E)mergency or abnormal in-plant I / 1 I 1 (2, 2, 1)

(EN)gineered safety feature - I - / 1 (3, 2, 2)

(L)ow-Power I Shutdown 1 / 1 / 1 (5, 4, 3)

(N)ew or (M)odified from bank including 1 (A) 2I 2/ 1 (4, 4, 3)

(P)revious2exams 3/3/s2 (1,1,1)

(R)CA l/1/l (1,1,1)

(S)imulator 03/25/2013 Rev. 0 2

2013 NRC Control Room/In-Plant JPM Summary JPM a Perform Control Rod and Rod Position Indicator Exercise per OST-1 005 (JPM-CR-256-d) SRO Upgrade NEW K/A 001 A2. 11 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Situations requiring a reactor trip (CFR. 41.5/43.5/45.3/45.13) RO 4.4/SRO 4.7 The candidate will assume the watch with the unit operating at 100% power and will be directed to perform OST-1005 commencing with Control Bank D in section 7.2. The candidate will insert and withdraw Control Bank D 10 steps as required. The candidate will continue OST-1 005 and select the next Control Bank and insert the Control Bank 10 steps as required. Once the candidate begins to insert the next selected Control Bank, the Alternate Path will begin and a malfunction of the rod control system will result in th Control Rods continuing to insert once the demand for rod motion has stopped. This will cause RCS Tavg, and Reactor Power will lower in response to the control rods inserting and the Control Rod step counter will continue to lower. The candidate should recognize the failure of the rod control system and perform AOP-001 immediate actions to place Rod Control in manual. The candidate may or may not select the manual position. Rod Control is considered to be in manual as long as the Auto position is not selected and being in Control Bank A satisfies this step. Rod motion will continue in either case requiring the candidate to perform the RNO action and initiate a manual reactor trip. The candidate will announce the Reactor is tripped and begin to perform the immediate actions of E-0. Once the candidate announces entry into E-0, evaluation on this JPM is complete.

JPM b Respond to the loss of the running CSIP (JPM-CR-038-a)

K/A APE 022 AA 1.01 Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: CVCS letdown and charging (CFR: 41.7 / 45.5 / 45.6) RO 3.4 / SRO 3.3 With the plant at 100% power and the ASI system OOS for planned maintenance the candidate will assume the Operator at the Controls (OAC) responsibilities. The A CSIP will trip requiring the candidate to enter AOP-018. AOP-018 will direct the candidate to isolate letdown in response to tEie. loss of charging flow. While the candidate is assessing letdown 1 RCP Thermal Barrier Flow Control valve (1 CC-252) will shut. Following the isolation of letdown the candidate will evaluate thestatus of component cooling water to the RCP Thermal Barrier and determine that flow is isolated. The Alternate Path of this JPM will begin and the candidate should evaluate Attachment I for RCP trip limits and determine trip limit 4 Loss of all RCP seal injection (including ASI) is met due to the loss of CCW flow to the RCP Thermal Barrier Hx and trip the reactor. Once the immediate actions of EOP-E-0 are complete the candidate will return to AOP-018 to stop all RCPs and shut the PRZ Spray controllers for RCS loops A and B. Once the candidate has stopped all RCPs and shut the PRZ Spray controllers for RCS loops A and B, evaluation on this JPM is complete.

03/25/2013 Rev. 0 3

2013 NRC Control Roomlln-Plant JPM Summary JPM c Pressurizer Pressure Master Controller Failure (AOP-019)

(JPM-CR-252-a) Previous NRC Exam 2012* randomly selected from bank K/A APE 027 AA 1.03 Ability to operate and / or monitor the following as they apply to the Pressurizer Pressure Control Malfunctions: Pressure control when on a steam bubble (CFR 41.7/45.5/45.6) RO 3.6/SRO 3.5 The candidate will assume the Operator at the Controls (OAC) responsibilities and be directed to maintain current plant conditions of 100% steady state power. Soon after assuming the watch the Pressurizer Pressure Master Controller PK-444B will begin to fail in Automatic to 100%. This will cause BOTH Pressurizer Spray valves to go from full closed to the full open position. The candidate should identify the failure and enter AOP-01 9. While performing the immediate actions the candidate should complete the Alternate Path (Take manual control of the Pressurizer Master Controller and lower the output to close the Pressurizer Spray Valves.) lEthe candidate takes manual of control of BOTH Pressurizer Spray valves and NOT PK-444B then the master controller will continue to fail and Pressurizer PORV 444B will go full open. Whenthe RCS pressure is < 2000 psig an auto shut signal will be sentW PORV 444B but by this time the pressure excursion will be so great that it will most likely cause an automatic Reactor Trip on OTAT and Safety Injection on Low Pressurizer Pressure (at 1850 psig). Once the candidate places the Pressurizer Master Controller is in manual OR both Pressurizer Spray Valves are manually shut AND PORV 444B is shut, evaluation on this JPM is complete.

JPM d Loss of Power to the TDAFW Pump control system (E-0 and OP-I 37)

(JPM-CR-281-a) SRO Upgrade NEW K/A APE 054 AA2. 04 Ability to operate and / or monitor the following as they apply to the Loss of Main Feedwater (MFW): Proper operation of AFW pumps and regulating valves (CFR: 43.5/45.13) RO 4.2/SRO 4.3 The candidate is informed that a Small Break LOCA has occurred, subsequently a leak developed in the Condensate Storage Tank (CST) and the CST level has decreased to less than 10%. The candidate is directed to supply ESW from the A Header to both the A AFW Pump and the Turbine Driven AFW pumps. This will require shutting down the A Train of Containment Fan Coolers. ESW will be aligned to the A AFW Pump. While aligning service water to the Turbine Drive AFW Pump the Aux Feedwater Pump Turbine Gov Control Power Failure alarm (ALB-017-7/3) will annunciate. The Alternate Path of this JPM will begin, as the TDAFW Pump speed will begin to increase, TDAFW pump flow and pump discharge pressure will increase. These indications will require the candidate to shut the steam supply valves per the alarm response procedure. Once the A AFW Pump is supplied from the ESW system and both TDAFW pump steam supply valves are shut, evaluation on this JPM is complete.

j  ;\

03/25/2013 Rev. 0 4 L

2013 NRC Control Roomlln-Plant JPM Summary JPM e Return the Containment Fan Coolers to normal following an SI actuation. (OP-169)

(JPM CR-.260-a) RO Only K/A 026 A4.01 Ability to manually operate and/or monitor in the control room: CCS fans (CFR: 41.7/45.5 to 45.8) RO 3.6/SRO 3.6 The candidate is informed an inadvertent SI initiation has occurred and the control room staff has entered EOP-E-O and EOP-ES-i .1. Attachment 1 of EOP-ES-1 .1 is being performed to realign plant systems. The candidate is directed to realign containment fan coolers lAW Attachment I step 6.a using OP-i 69, Containment Cooling And Ventilation, Section 8.4. The candidate will be directed to align the A Train of CNMT Fan Coolers for normal service. The candidate will secure both A Train CNMT Fan Coolers and verify proper damper alignment for the secured fans. The candidate will restart the A Train Fans per section 5.1 of OP-169. To minimize the starting current required for Hi-Speed operation the fansare initially started in Lo-Speed, then stopped and restarted in Hi-Speed. The candidate will return to section 8.4 to secure the B Train of CNMT Fan Coolers. Once the B Train of CNMT Fan Coblers are in standby and the determination is made that Maximum Cooling Mode is NOT required, evaluation on this JPM is complete.

JPM f Restore Off-site Power to an Emergency Bus (OP-I 56.02)

(JPM-CR-027-b) SRO Upgrade K/A 062 A4. 01 Ability to manually operate and/or monitor in the control room: All breakers (including available switchyard)

(CFR: 41.7/45.5/to 45.8) RO 3.3/SRO 3.1 The candidate is informed that the plant has tripped due to a LOOP and both EDGs have energized IA-SA and i B-SB 6.9kV Emergency Busses. The candidate will be directed by the CRS to restore power to the iA-SA 6.9kV Emergency Bus lAW OP-156.02, AC Electrical Distribution. The candidate will re-energize theAux Bus Dfrbm Start Up transformer A. With Aux Bus D energized the candidate will close the first Aux Bus D supply breaker to the iA-SA 6.9kV Emergency BUs. The Alternate Path of this JPM will begin after the candidate determines the status of the EDG output breaker. Because the EDG output breaker is shut the candidate must transition to OP-i 55 to synchronize and transfer the IA-SA 6.9kV Emergency Bus to offsite power. The candidate will operate the voltage and governor controls to parallel the EDG and offsite power. Once parallel operations have been achieved and the candidate transitions to OP-i 55 section 7.1 and start to unload the EDG, evaluation on this JPM is complete.

03/25/2013 Rev. 0 5 1)

2013 NRC Control Roomlln-Plant JPM Summary JPM g Restore an Excore NI Channel to service (at power, NI failed) (OWP-RP-25)

(J PM-CR-278-a)

NEW K/A 015 A4. 03 Ability to manually operate and/or monitor in the control room: Trip bypasses (CFR: 41.7/45.5 to 45.8) RO 3.8/SRO 3.9 New JPM to restore previously repaired failed Nl-43 to service.

The candidate will assume the watch with the plant at 100% steady state power and the PRNI channel NI-43 which failed downscale earlier repaired. The candidate will be required to return Nl-43 to service lAW OWP-RP-25. OWP-RP-25 ensures the components that have NI-43 as an input, Rod Control and SG Feedwater regulating bypass valves,are in nanual control to prevent spurious movement or uncontrolled changes in level. The candidate will verify the controllers are in manual.The OWP will require the candidate to contact maintenance (l&C personnel) to return the two previously trip bistables for the Channel Ill OTAT signals to normal in the Process Instrument Cabinet 3 (PIC-3).

The candidate will return the following items to NORMAL

• At the Detector Current Comparator Drawer: Both upper and lower sections of Nl-43

• At the Comparator and Rate Drawer: Comparator Channel Defeat switch The candidate will return the following items to OPERATE

• At the Miscellaneous Control and Indication Panel: Power Mismatch Bypass switch and the Rod Stop Bypass switch.

The candidate will have to contact maintenance (I&C personnel) a second time and direct them to re-connect the NI-43 power supply leads to the NI drawer. After the l&C personnel re-connect the N 1-43 power supply leads the candidate will verify proper bi-stable and annunciator configuration for the restoration of Nl-43 to service. Finally the candidate will have to restore the plant computer (ERFIS) point to processing anddocument the position of MCB components for the current plant conditions with NI-43. Once the candidate reports that OWP-RP-25 is complete to the CRS, evaluation on this JPM is complete.

JPM hAlign CCWto Support RHR System (OP-145)

(JPM CR-085-a)

K/A 008 A4. 10 Ability to manually operate and/or monitor in the control room: Conditions that require the operation of two CCW coolers (CFR: 41.7/45.5) RO 3.3/SRO 3.1 The plant is in Mode 4 and a cool down is in progress. The CRS directs the candidate to align CCW to support RHR operation lAW OP-145 section 8.9. After reviewing section 8.9 the candidate determines a second CCW pump is required to be started and transitions to section 5.2. The candidate starts the B CCW pump lAW section 5.2 and returns to section 8.9 and isolates the A train essential header of the CCW from the non essential header.

The candidate will align the B train essential header to supply RHR HX B. The candidate will verify both trains of the CCW system operating parameters are within the required band on the MCB indicators. The candidate will contact a non license operator (NLO) to locally verify the CCW flow to the Gross Failed Fuel Detector is within the required band. Once the candidate contacts the NLO to verify CCW flow locally then evaluation on this JPM is complete.

03/25/2013 Rev. 0 6

2013 NRC Control Roomlln-Plant JPM Summary JPM i Place the ASI System in Standby Alignment (OP-I 85)

(JPM-IP-277-a) SRO Upgrade NEW K/A 004 A4. 11 Ability to manually operate and/or monitor in the control room: RCP Seal injection flow (CFR: 41.7/45.5 to 45.8) RO 3.4 /SRO 3.3 NOTE: This JPM is inside the RCA.

The plant is in Mode 4 and a heat up is in progress. The CRS directs the candidate to place the ASI system in automatic standby alignment lAW OP-185 section 5.1. The candidate will verify the ASI supply header isolation valves are open and the status of the ASI system control panel. The candidate will realign the ASI pump to automatic and return the Squib valve bypass control switches to normal alignment on the ASI control panel. The candidate will turn on the ASI system control panel feeder supply breaker and the ASI pump power supply breaker. The candidate will recheck the indications on the ASI system control panel for the proper standby alignment of the system. Once the candidate proceeds to section 5.1.3, Automatic Standby alignment configuration control closeout then evaluation on this 1

JPM is complete.

JPM i Local Inspection of Annunciator Cabinets (AOP-037)

(JPM IP-273-a)

K/A 016 A2. 02 Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of power supply (CFR: 41.5/43.5/45.3/45.5) RO 2.9/SRO 3.2 The candidate is informed that the control room annunciator System 2 power failure alarm has been received and the CRS has entered AOP-037. The CRS will direct the candidate to check the status of System 2 annunciator power supplies per AOP-037 Attachment 2. The candidate will perform Attachment 2 and obtain the annunciator cabinet key. The JPM cues include information of the proper status of the power supply light indications. The candidate will initial for the indications that remain lit. The candidate will determine based on the cues that one of the System 2, Bay 1, 12 VDC power supplies, one of the System 2, Bay 3, 12 VDC power supplies and the System 2, Bay 5, 24 VDC power supplies are de-energized.

Once the CRS is notified that AOP-037, Attachment 2 is complete and the correct de energized power supplies have been identified then evaluation on this JPM is complete.

JPM k Perform an Instrument Air System Leak Isolation Locally (Turbine Bldg I Yard) (AOP-017)

(JPM-IP-16I-a) SRO Upgrade K/A APE 065 AA2. 03 Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Location and isolation of leaks (CFR: 43.5/45.13) RO 2.6/SRO 2.9 The candidate is informed that the plant was operating at 100% when the plant was tripped due to lowering instrument air pressure. AOP-017 is being performed. The CRS directs the operator to perform Attachment 3 of AOP-017 to reduce instrument air header loads. They will be required to isolate individual sections of the instrument air system within the Turbine Building and contact the Main Control room staff following the completion of each action to determine if the prior actions have successfully isolated the instrument air leak. The JPM cues include information of the proper sequence of actions that must be taken in order reposition the valves and due to the valve locations a description of the nearest ladder location is given to simulate climbing to the valve location. Once notified by the Main Control room that the instrument air header pressure has stabilized then evaluation on this JPM is complete.

03/25/2013 Rev. 0 7

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 Facility: SHEARON-HARRIS Scenario No.; 1 Op Test No.: 05000400/2013301 Examiners: Operators: SRO:

RO:

BOP:

Initial Conditions:

• IC-5, BOL, 55% power

. RHR pump A-SA is under clearance for pump seal replacement

• lSl-4, Boron Injection Tank Outlet valve is under clearance for breaker repairs

. B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• Boric Acid Transfer Pump A-SA is under clearance for motor replacement Turnover:

• Plant is at approximately 55% power. Plant startup is in progress lAW GP-005 step 1 35.e. After taking shift continue plant startup at 4 DEH Units/mm.

Critical Tasks: Isolate AFW flow to C Steam Generator prior to exiting EOP E-2

• Shut BIT Outlet valve 1 Sl-3 prior to reaching water relief from PZR SRVs Event No. MaIf. No. Event Type* Event Description 1 N/A R RO/SRO Continue plant startup to 100% power NBOP/SRO 2 pt:495 I BOP!SRO Failure of the C SQ Pressure Transmitter PT-495 to 0%

TS SRO 3 lt:1 12 I RO/SRO VCT LT-1 12 fails high, letdown full divert to RHT (AOP-003) 4 rmsOO7 I BOP/SRO Radiation Monitor 3502A fails high and alarms, Containment zcr744 TS SRO- Purge fails to isolate automatically (AOP-005) 5 prsl4b I - RO/SRO Pressurizer Spray Valve, PCV-444D, fails Open (with Manual Control available) (AOP-019) 6 mss0lc M-All Steam line Break on C SG inside Containment 7 zrpk6l7a I BOP/SRO Failure of Auto AFW Isolation on C SG zrpk6l7b 8 nis06b I RO/SRO SR Nuclear Instruments fail to energize post trip due to IR Nl-36 undercompensated (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

\: \\J Harris 2013 NRC Exam Scenario 1 Rev. 0 Page lof 9 t7\

Appendix D Scenario Outline Form ES-D-T L HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I Turnover provided to the crew is The plant is operating at 55% power in BOL. Criticality was achieved 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ago, 74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> after a trip from 100% power. A plant startup lAW GP-005 is in progress. After initial turbine loading, the load increase has been performed at 4 Units/mm.

The unit is currently at 55% power 82 hours9.490741e-4 days <br />0.0228 hours <br />1.35582e-4 weeks <br />3.1201e-5 months <br /> post-trip.

A startup has commenced and the crew has been directed to continue the power increase using GP-005, Power Operation Mode 2 to Mode 1, to 100% power at a ramp rate of 4 DEH Units/Minute.

The following equipment is under clearance:

RHR Pump A-SA is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and is expected to be repaired in the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The pump must be restored to operable status within the next 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. Tech Spec 3.5.2 LCO Action a. and Tech Spec 3.3.3.5.b Action c. applies. OWP-RH-01 has been completed.

!_1iiE C,ARE 3)*4S.2 ECC SUSS1HS -

GRcAYR WA[1 EOW.L TID 35OF LIHITfl+11 COOiTION FOR OPERTJON 3S,2 io i:ndepetx3ent Ernerency Core Coo1irg System (ECCS) subsystems shall be OPER,AU i th ech sihsybem ccopricd f:

e. Oee OE1,E Chorghyqtfety injectic pup.

b. Oee OPEAE1E FHP heat ethan9er, c, One OfERABLE FJ1 pp end

d. An CPERAI3LE fio path capable oI ting wction Trcoi the reThiehng vater tarae tank on a Sarety injection cignal an& upon being manu fly 1igne& transferrtrig suction ft the conta ient sur during the recirculatton phase ef operition.

MWE 1, 2 and 3 ACT a, L4ith one CS subsystem inoperable, restore the inoperable Subi$teafl tO OPERAL tti within 72 o1rs r be in e least. 14T SN()B ithin the next 6 1OU5 ard in [101 SF*JTDON wihin tle f&1oin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 2 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Tech Specs associated with inoperable RHR Pump A-SA continued INSTRUMENTATION REMOTE SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.5.a The Remote Shutdown System monitoring Instrumentation channels shown in Table 3.39 shall be OPERABLE.

3.3.3.5.b AlT transfer switches, Auxiliary Control Panel Controls and Auxiliary Transfer Panel Controls for the OPERABILITY of those components required by the SHMPP Safe Shutdown Analysis to (1) remove decay heat via auxiliary feedwater flow and steam en.rator poweroperated relief valve flaw from steam generators A and 8, (2) control RCS Inventory through the normal charging flow path, (3) control RCS pressure, (4) control reactivity, and (5) remove decay heat via the RHR system shall be OPERABLE.

APPLICABILITY: MODES 1. 2, and 3.

AC1!ON:

c. With one or more inoperable Remote Shutdown System transfer switches, power, or control circuits required by 3.3.3.5.b, restore the inoper able switch(s)/circuit(s) to OPERABLE status withIn 7 days. or be in HOT STANOY wi tM ii the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• £B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• 1 S 1-4, Boron Injection Tank Outlet valve is under clearance for breaker repairs.

Tech Spec 3.5.2 Action a applies. OWP-Sl-01 has been completed.

JYE CODL1N11k 3/4.5 2 EGGS SUBSYSTEMS 6RTATE:P. THAI OR EQUAL 10 350F LIMTTTNf3 CQflD)1TO FOR OPFRATTON 3,5 I inc1epenent Linergency Core COoMri System (1ECCS) subsystens shaH bo OPERABLE with each suhsystefn ciipri se of

o. Ono OPERA&E Chorgini/ofcty 1ctiO pønp.

h. Or OPEMI F RHR hmat nxchangor.

c One PE.RADLE RHR ii and

d. An CPERABLE flow path capie 01 takIng suction Trcra the reitiel log water storage taik on a 3oft.y Injection signal cid, tn behig muay-a1riod transftrring suction to th containmont r.ump du:ring the racirculation phase of operation, EIianLux: MODES I 2. and 3.

1) tON:

a frth oi FF05 subsystem inoperable, restore the i noperabie subsyst.e.n to OPEIABL tatus w]thin 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least cTANO8( within th rxt hours anti in HOT S4JTDMJ within the lflrtLdntn $ hnnr-

• Boron Injection Pump A-SA is under clearance for motor replacement. Tech Spec 3 1 2 2 applies (tracking only) OWP-CS-04 has been completed Harris 2013 NRC Exam Scenario 1 Rev 0 Page 3 of9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Event 1: Continue plant startup to 100% power Crew performs a power increase of approximately 5%-i 0% power (Lead Examiners discretion). For this reactivity manipulation it is expected that the SRO will conduct a reactivity brief, the RO will dilute and monitor auto rod withdrawal per the reactivity plan and the BOP will operate the DEH Controls as necessary to raise Main Turbine power.

Event 2: Failure of the C SG Pressure Transmitter PT-495 to 0%. This event will require the BOP to place the C SG level control to manual and control SG level within Reactor trip limits.

The SRO should provide level band and trip guidance lAW OMM-001. The crew will take the channel out of service using OWP-ESF-04, Protection channel Ill Steam flow. The SRO should evaluate Tech Spec 3.3.1 action 6, Tech Spec 3.3.3.6 Accident Monitoring Instrumentation Action a, and Tech Spec 3.3.2 I.e Steam Line Press Low Action 19 applies.

T.S. 3.3.1: As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

14. Steam Generator Water Level--Low 2 stm. yen. 1 stm. yen. 1 stm. yen. level 1, 2 6 Coincident With Steam! level and level coincident and 2 stm./feed Feedwater Flow Mismatch 2 stm./feed- with 1 water flow water flow stm./feedwater mismatch in same mismatch in flow mismatch in stm. yen. or 2 each stm. yen. same stm. yen. stm. gen. level and 1 stm./feedwater flow mismatch in same stm. yen.

ACTION 6 With the number of OPERABLE channels one Less than the Total.

Number of Channel.s, STARTUP and/or POWER OPERATION nay proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Minimum Channels OPERAELE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

TS 3.3.3.6 The Accident Monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE

• ACTION a. With the number of OPERABLE accident monitoring instrumentation channels except In Core Thermocouples and Reactor Vessel Level less than the Total Required Number of Channels requirements shown in Table 3.3-10 restore the inoperable channel(s) to OPERABLE status within 7 days. or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 4 of 9

[endix D Scenario Outline Form ES-D-1 L HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Tech Spec 3.3.2 INSTRLIMENIAROh 3/4 3 2 E4G NEERED SAFETY FEATURES ACTUATION SYSTEM 11STRUMENTATIO LIMITING CONDJTION FO OPERATION 3.3.? The Encjirieored SaFiy Features tuatien Syt:eeFSFAS) instrumentation chpnno d intor1oc.s hown in rable 939 srell be OPERAIII with the r Trio Setomnts set consistent wsth .he vlues shown irs the Trip Setpoint c kjrr,s f rabe 3.3.4 PJ,jL1Y: As shown ri Table 3.3.3.

NEEREe SAFElY FEATURES T1JTIOF SYSTEM ThS1RU1EiTATION

MUM mEAl MO CHANNELS CHANNELS APPI TF.l F UUICiIftNAL LIIT 1W (flANNEl S T1 TRTP OPERARLE .I5OFIFS I. Siet injection Reaccor Trip.

1eoc1maicr isolatsai. Control Roo iso laticri, Start Diesel neretcrs Ccntalnirarit ventilation Isolation Phase A Cormtalraert Isolation, Start mary Feettvater SysLemim rsotorDrien PumTps, Start Conla innert Fan Coolers. Start J.çs (nercncy 5ervio water Start Erierency Seivc aL.r 13outer Ptos) p Steae line Pros urct mn# 3istein 2/steam 2/steam line 2, 34 line 1mm an steac.

Tine ACTiON $ With the rn.unber of OPERS&E channels one less than the Total Number of Channels, operation imay prooee provided the foiicwlnD conditions are satisfsed;

a. The inoperable channel is placed iii the Lii pte<i rmsrsditir,r; within G hoors, arid b, The Minimum Channels OPERAGLE roqinremmient 15 net; however, time inoocraimle channel may be bypassed for imp to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for urseillancr to iing of ut9r cnannel per Specification i,3.2.i.

The SRO should also prepare OMM-001, Attachment 5 Equipment Problem Checklist for the failure.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 5 of 9

[ Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Event 3: LT-1 12 fails HIGH full divert to RHT. Enter AOP-003. This will require entry into AOP-003, Malfunction of Reactor Makeup Control (no immediate actions). A failure of LT-112 high will cause I CS-I 20, Letdown VCT/Hold Up Tank valve to shift to the Hold Up Tank. The RO will have to return the MCB switch to the VCT position. Since VCT level has failed HIGH auto CSIP suction switch over on 5% VCT level to the RWST will not occur until Maintenance has lifted the leads associated with LT-1 12. The operator will have to monitor VCT level and communicate with Maintenance to resolve this failure.

The SRO should also prepare OMM-001, Attachment 5 Equipment Problem Checklist for the failure.

Event 4: Radiation Monitor 3502A high alarm, Containment Purge fails to isolate automatically.

This failure will cause the radiation monitor output to immediately fail high and RM-1 1 to go into High Alarm. The automatic response to isolate Normal Containment Purge fails to occur due to a failed relay. The crew should respond to the alarms and enter AOP-005, Radiation Monitoring (no immediate actions). AOP-005 Attachment 1 will direct verifying that the automatic response for this alarm has occurred (other procedure options are available and detailed in exercise guide). This will also require the SRO to evaluate Tech Spec 3.3.2 for the failed Containment Isolation and Tech Spec 3.4.6.1, Leakage Detection Systems.

Tech Spec 3.4.6.1, Leakage Detection Systems, Action a REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Airborne Gaseous Radioactivity Monitoring System.

b. The Reactor Cavity Surnp Level and Flow Monitoring System, and

c. The Containment Airborne Particulate Radioactivity Monitoring System.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With a. or c. of the above required Leakage Detection Systems INOPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed for airborne gaseous and particulate radioactivity at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required Airborne Gaseous or Particulate Radioactivity Monitoring System is inoperable: otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 3D hours.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 6 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Tech Spec 3.3.3.1 Table 3.3-6 item 1.b.1) Airborne Gaseous Radioactivity RCS leakage Detection Actions 26 and 27 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS MINIMUM CHANNELS CHANNELS APPLICABLE ALARM/TRIP INSTRUMENT TO TRIP D.ELBABE 1QS IPOINT

1. Containment Radioactivity-

b. Airborne Gaseous Radioactivity

1) RCS Leakage Detection 1 I 1, 2. 3. 4 3 pCi/mi 1.OxlO

2) Pre-entry Purge 1 1 26 27 2.0x10 1iCiJml 3 30 ACTION 26 - Must satisfy the ACTION requirement for Speeification 3.4.6.1.

ACTION 27 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge makeup and exhaust isolation valves are maintained closed.

The SRO should also prepare 0MM-CO 1, Attachment 5 Equipment Problem Checklist for the failure.

Event 5: Pressurizer Spray Valve, PCV-444D, fails to maximum output in Auto (with manual control available). This failure will cause one of the Pressurizer spray valve to fail to 100% open while the other valve closes to 0% open. The crew should respond to multiple alarms and enter AOP-019, Malfunction of RCS Pressure Control. The RO should complete the immediate actions by gaining control of the Pressurizer Spray Valves. If RCS pressure decreases to <

2202 psig during the event the SRO will have to evaluate Tech Spec 3.2.5 POWER DISTRIBUTIOM LIMITS 314 2 5 DNB APFTS LIMITING CONDITiON FOR DPERATIQ9 125 Th f1iini [)ND-ilated tes shal I he iaiitaed wthiri th following limits a Reactor Coolant System T ter addition for intrmgnt 1ncvrainty, and b Pressuriae Pressure 2185 gsig after suhtrct1on for instrument uncertarity, and

c. CS total flow rate 23,:5D gpin after subtraction for trstruiei1t urlcertalnty PttCABILITY MODE 1.

ACTIOk:

With any f the above pararters not witIio its specified 11t, testore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THFRIt& PCLJER to less than 5 of RTED THERftAL POWER within the rext hours The SRO should also prepare 0MM-CO 1, Attachment 5 Equipment Problem Checklist for the failure.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 7 of 9 -\ V

Appendix D Scenario Outline Form ES-D-1

[ HARRIS 2013 NRC Exam SCENARIO 1 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO I (continued)

Event 6: MAJOR Steam line Break on C SG inside Containment. Once RCS pressure control has been established to the satisfaction of the Lead Examiner a Steam Line Break inside Containment on the C SG will occur. The crew should enter and carry out the immediate actions of EOP E-0.

The crew should diagnose that a LOCA is NOT in progress and transition from EOP E-0 to EOP E-2, Faulted Steam Generator Isolation.

While the crew is performing actions of E-2 the Containment pressure will continue to rise beyond 10 psig which will actuate a Containment isolation Phase B signal and require ALL RCPs to be secured.

Event 7: Failure of Auto AFW Isolation on C SG. The crew should identify that an actuation signal for AFW Auto Isolation has failed on the C SG and manually isolate AFW flow to the C SG.

Event 8: Source Range channels will fail to energize post trip due to IR N 1-36 under compensation. The crew will need to identify the failure of the SR instrumentation MCB indication and audible counts. They will then manually energize the SR channels to establish an audio count rate.

The scenario is ended when Safety Injection has been terminated and the crew transitions to EOP-ES-1.1, SI Termination.

Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 8 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC Exam SCENARIO 1 CRITICAL TASK JUSTIFICATION:

1. Isolate AFW flow to C Steam Generator prior to exiting EOP E-2.

Failure to isolate a faulted Steam Generator that can be isolated causes challenges to the Critical Safety Functions beyond those irreparably introduced by the postulated conditions. Also, depending upon the plant conditions, it could constitute a demonstrated inability by the crew to recognize a failure of the automatic actuation of an ESF system or component. This critical task requires the crew to recognize an automatic actuation should have occurred of an ESF system or component but has not and then take manual operator actions to perform the isolation.

2. Shutting BIT outlet valves I SI-3 prior to water relief through the PZR Safety Relief Valves (SRVs).

FSAR Section 15.1.5.2 (page 15.1.5-7) states the operator will secure one of the two CSIPs to facilitate PZR level indication remaining on scale and controllable.

Additionally, shutting the BIT outlet valves is the first steps in realigning normal charging to the RCS. %orItirioed flow through the BIT will cause the PZR level to reach solid conditions(otential!, causing the SRVs to lift. At low fluid temperature (like thosépreserIt in the PZR at this time) the SRVs may not reset after fluid operation. This operator action is critical to prevent a challenge to plant safety

/1 Harris 2013 NRC Exam Scenario 1 Rev. 0 Page 9 of 9 V)

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 Facility: SHEARON-HARRIS Scenario No.: 2 Op Test No.: 05000400/2013301 Examiners: Operators: SRO:

RO:

BOP:

Initial Conditions:

• lC-26, MOL, 88% power

• RHR pump A-SA is under clearance for pump seal replacement

. ISI-4, Boron Injection Tank Outlet valve is under clearance for breaker repairs

• B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

. Boric Acid Transfer Pump A-SA is under clearance for motor replacement Turnover:

• A shutdown is in progress due to problems encountered during the repairs on the A RHR pump. Repairs will not be able to be completed prior to the LCO expiring. The plant is operating at 88% power in MOL. When turnover is complete continue the power reduction at 4 DEH units/mm to support being in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. GP-006 is currently in progress with step 8 completed. 300 gallons of boric acid have been injected lAW the reactivity plan. Control Bank D Rods have just stepped IN. All required notifications have been made to individuals concerning the reason for the shutdown.

Critical Tasks:

• Insert negative reactivity into the core by inserting control rods during an ATVVS before completing the immediate action steps of FR-S.1

. Trip RCPs once RCP Trip Foldout Criteria is met and prior to exiting E-1 Event No. MaIf. No. Event Type* Event Description 1 N/A R RO/SRO Continue plant shutdown at 4 DEH Units/mm N BOPISRO 2 crfl4b C RO/SRO Rods fail to operate in AUTO (manual control works- setting up for ATWS) Enter AOP-0O1 3 ft:497 I BOP/SRO Feed flow transmitter on C SG FT-497 Channel IV (selected TS- SRO for 1 C SG) fails low- OMM-0O1 4 idi xdlil42 C BOP/SRO

- Reactor Primary Shield Fan S2-A-SA Failure ilo xdlol42w ian_xn27e05 5 cvcO5a C RO/SRO CSIP Trip 1 available, requiring AOP-018 entry TS-SRO ASI Pump start / Respond to Boron addition to RCS from ASI 6 N/A N RO/SRO Restore letdown lAW OP-I 07 7 rcsl8b R RO/SRO 30 gpm RCS leak to Containment. Enter AOP-016 increase N BOP/SRO ramp rate lAW AOP-038 -

8 rcsl8b M ALL RCS leakage exceeds VCT makeup capability E-0 manual Rx rps0l b ATWS Trip with ATWS Reactor Trip Breakers fail to open in auto or manual k

Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 1 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO 2 continued Event No.

9 MaIf. No.

rcsl8b

[ Event Type*

M Event Description ALL RCS leakage increases to SBLOCA -650 gpm 10 zdsq2:6b C ROISRO B ESW pump fails to auto start from the Sequencer or from jpb9lOlb low pressure (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 Turnover provided to the crew is A shutdown is in progress due to problems encountered during the repairs on the A RHR pump. Repairs will not be able to be completed prior to the LCD expiring. The plant is operating at 88% power in MOL. Control Bank D is at 200 steps.

The latest RCS Boron sample was 1067 ppm. GP-006 step 9 is in progress. When turnover is complete continue the power reduction at 4 DEH units/mm to support being in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. All required notifications have been made to individuals concerning the reason for the shutdown.

The following equipment is under clearance:

RHR Pump A-SA is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and must be restored to operable status within the next 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />. Tech Spec 3.5.2 LCO Action a and Tech Spec 3.3.3.5.b Action c applies. OWP RH-01 has been completed.

EKEEJCY CQUE CODLING SY1E 3/4..2 LOCS SURSYST(NS . GREATER Ti{4 OR EQUAL TO SAF LINITENG CONDITEON FOR OPERATION 2 Iwo independenL Enrerericy Ccre Co1 I n Systei cfcCS) subystecr shd be OPEPAOtE with çh hsyten ccprie of;

. One OPERABLE Chacqng/sfety injection prinp,

b. One OPERA[E RHP heat echaner, c, One OPERLE RHR pp, end

d. An OPERABLE flow path capable of taking suction frce the refuelinG water torae tank on a Safety Injection signa and. epon etn rnanuaUy aligned, transferring suction to the contairent sun durin9 the recrculation phase of operation.

EiIILJTY DOES 2. 2, end 3.

ACTiO

a. Wth one ECCS subsystee irperab1e. restore the inoperabe w OPRABL M,afus within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or Le in at. lel. hOT STNOI3? rflthln the nest 6 1wnrs and in HOT SLItlXZN iithi the foflcwurg 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 2 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 (continued)

Tech Specs associated with inoperable RHR Pump A-SA continued 14SThU4ENTATION RE)TE 5TO1I YSTJ4 LI11T1fG cPMn!T1ON F QPERATIM 33.3.Bb Alt trwsfer s1.tchqs ALlxfllary Cortol Pa iCntri Trans1er Pin.t Crtio1s M Auxfltarj the PERAL1TY r thou ,mnuntc ft.d by Sefe Shutdown Malysis to (1) rovi dacay heat ila aux1lary fatth flow and staam en.ratr p.r.o#.rat re1f valve flow fr taa ne3atrs A and , (2) ontrl RCS nt3r thrcuçh te noreal harin flaw path, (3) corrral RC ureur, (4) cutrol reatiity, aM (5 remoa daca heat ia the R1R sytaz al I be OPERABLE.

AP!UCABIUTY DE5 1. Z 4n 3.

c. With inc or aor b:r*Ts ta Skztdown Syta transfer ew1tnes.

pcr, or CotroT CfrUts r.qured y 3.3.3.LS, rei-tor* the inoper able sIth(i)/e1ruft() , OPERABL! status within 1 days. or be in HOT STA4BY within tue next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• 1 Sl-4, Boron Injection Tank Outlet valve is under clearance for breaker repairs. Tech Spec 3.5.2 Action a applies. OWP-SI-01 has been completed.

LcOJ 3!4E.2 £CCS SJ8YSTEHS . T GREftTER T3/4M OR EOUL TC 3&OF LIMITING CONITIDN FOR OPERT[O1 3B, Two indordcnt Eccrqcncy Cor Cod ig Sm CECCC sbsystcs h11 be OPERAEU i t each wbsysteto cepri of a Cne OPERABLE Chri&safety injection pwiip,

b. Cne OPER3LE NFf neat exflancer, c Coc ORsLE RN puop,

d. n :WFRMLE fic path Ca2ablE of :air suctin frcoi the refuelinQ atr storg tarx on n a ety Injetion si9nI r4 upon being ffanaiiy. aiinO, trarst9mrg suction to the coritaiment surrp turlriy i1i r& rcu1 dl. un phm 1 o.end LiOn APtIC&RIL:T MGES , 2, id 3.

/T[ON:

a. tn Ex subsystEm lneperabl:e., restore the 1noratte Ub)LU Lu OPERB1.E toLii l.hifl 72 huw on be in t. 1tL HOl ST936y with rn t nc>t B no r ord in HOT 5l-tTX6N wi h n th

• Boron Injection Pump A-SA is under clearance for motor replacement. Tech Spec 3.1.2.2 applies (tracking only). OWP-CS-04 has been completed.

Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 3 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 continued Event 1: Continue plant shutdown at 4 DEH Units/mm Crew performs a power decrease of approximately 5%-i 0% power (Lead Examiners discretion). For this reactivity manipulation it is expected that the SRO will conduct a reactivity brief, the RO will borate and monitor auto rod insertion per the reactivity plan and the BOP will operate the DEH Controls as necessary to reduce Main Turbine load.

Event 2: Rods do not move in AUTO (manual control works- setting up for ATWS) Enter AOP 001. During the downpower the rod control system will fail to operate in AUTO and the crew will be required to enter AOP-001, Malfunction of Rod Control and Indication System. The crew must take immediate actions and place rods in MANUAL. The SRO will have a TS check of 3.1.3.1, 3.1.3.5 and determine that all rods are trippable per Attachment 5 of AOP-00i. System Engineering will be contacted. The crew will be expected to continue with the power reduction.

If necessary, to get the crew moving, prompting by the Manager of Operations can be used.

TS 3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within

+ 12 steps (indicated position) of their group step counter demand position.

Applicability: Modes 1 and 2 TS 3.1.3.4 All shutdown rods shall be fully withdrawn as specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106.

Applicability: Modes 1 and 2 AOP-001 Attachment 5 I MALFUNCTION OF ROD CONTROL AND INDICATION SYSTEM I

Attachment 5 Determination of Control Rod Trippability Sheet I of 1 The fofowing guidance is provided far makino the determination of control rod trippability:

A control rod may be considered trippable under any of the folloing circumstances:

• Rod Control System URGENT FAILURE alarm exists

• Inspection of the affected system cal)nets reveals obvious electrical problems (for example, blown fuses)

• All rods of a particular group or bank are simultaneously affected

• NO control rod motion is possible If none of the four conditions exist the rod must he considered untriopable until proven otherNise Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 4 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 continued Event 3: Feed flow transmitter on C SG FT-497 Channel IV (selected for IC SG) fails low OMM-001. When the plant is in a stable condition, the Lead Evaluator can cue the SG C Feed Flow channel failure. The crew should respond in accordance with the alarm response procedure and OMM-001. The BOP will be controlling SG C level with the FRV in MANUAL and may switch controlling FF channels to restore control to AUTO. The crew will take the channel out of service using OWP-RP-1 0, SF/FF Loop 3. The SRO should evaluate the following Tech Specs for failure of FT-497:

T.S. 3.3.1: As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

14. Steam Generator Water Level--Low 2 stm. gen. 1 stm. gen. 1 stm. gen. level 1, 2 6 Coincident With Steam! level and level coincident and 2 strn./feed Feedwater Flow Mismatch 2 strn./feed- with 1 water flow water flow stm./feedwater mismatch in same mismatch in flow mismatch in stm. gen. or 2 each stm. gen. same stm. gen. stm. gen. level and 1 stm./feedwater flow mismatch in same stm. gen.

ACTtON 6 With the nuthbe of OPERABLE channels one less than the Total.

Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Minimum Channels OPERABLE requirement is net; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3,1.1.

The crew should implement OWP-RP-1 0 for this failure.

Event 4: Reactor Primary Shield Fan S2-A-SA Failure. When event 3 concludes to the satisfaction of the Lead Examiner then event 4 Reactor Primary Shield Cooling Fan S2-A-SA failure can be inserted. The crew will respond to ALB 027-05-5, Reactor Primary Shield CIg Fans S2 Low Flow - O/L and evaluate the condition. The crew should identify that the running fan has tripped, start the standby fan lAW OP-169, Containment Cooling and Ventilation, and dispatch an operator to check the status of the breaker. If flow cannot be established by starting the standby fan then APP-ALB-027-05-5 will direct a Reactor Shutdown to be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and cooldown to <350°F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> would be required. OP-169 P&L 4.0.3 states One Reactor Primary Shield Fan is required to be in operation anytime RCS temperature is greater than 140°F.

LI Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 5 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 continued Event 5: Trip of the running A Charging Pump breaker. The crew will enter AOP-018 and carry out the immediate actions. The crew should isolate letdown and then implement actions to place the B Charging Pump in service. The crew will have to secure the ASI pump after the CSIP is started and evaluate the boration caused by the ASI pump running. The efficiency that the crew has in progressing through AOP-01 8 to the point of securing the ASI pump will determine the amount of boric acid added to the RCS through the RCP seals. This could require the SRO to direct the RO and BOP to coordinate Reactor and Turbine controls (dilute, rod movement and / or Turbine reduction) to accommodate the boron addition for Tavg/Tref stabilization.

The SRO should evaluate the loss of the CSIP in accordance with Tech Specs 3.1.2.2, 3.1.2.4 and 3.5.2 TS 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

b. Two flow paths from the refueling water storage tank via charging/safety injection pumps to the RCS.

ACTION: With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-I06 at 200°F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

TS 3.1.2 4 With only one Charging/safety injection pump OPERABLE restore at least two charging/safety injection pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN as specified in the CORE OPERATING LIMITS REPORT (COLR) plant procedure PLP-106 at 200°F within t he next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s: restore at least two charging/safety injection pumps to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

TS 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

One OPERABLE Charging/safety injection pump One OPERABLE RHR heat exchanger One OPERABLE RHR pump and An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and. upon being manually aligned transferring suction to the containment sump during the recirculation phase of operation.

ACTION: a. With one ECCS subsystem inoperable restore the inoperable subsystem to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The SRO should also prepare OMM-QOl, Attachment 5 Equipment Problem Checklist for the failure Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 6 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 (continued)

Event 6: Restore letdown lAW OP-i 07 once the B Charging Pump is in service, the crew will restore letdown lAW OP-I 07 Section 5.4 to establish inventory control. Once letdown has been restored, Tech Specs have been evaluated for the loss of the A CSIP, and the crew response to the boron addition from the ASI system have been addressed to the satisfaction of the Lead Examiner the next event can be initiated.

Event 7: -30 gpm RCS leak to Containment. Enter AOP-0i6 increase ramp rate lAW AOP 038. This will be the initiating event for the Major event. The crew will identify that leakage is present by radiation monitor alarms associated with increasing Containment sump leakage.

Pressurizer level will decrease and Charging flow will increase. The crew should enter AOP 016, Excessive Primary Plant Leakage (no immediate actions) based on the indications of unidentified RCS leakage. The crew should determine that RCS leakage is now exceeding TS 3.4.6.2 leakage of 1 gpm Unidentified Leakage but RCS leakage is within VCT makeup capability <120 gpm). They shouldeteiinioe that Attachment 7 for leakage inside Containment is the applicable attachment and pContairi iitirge. Since a shutdown is already in progress lAW GP-006, the crew shouiailUa ihtIThre rapid means of plant shutdown is required and enter AOP-038 The SRO should implement AOP-038, Rapid Downpower and direct an increase of the ramp rate to some value> 5 DEH Units/mm in an attempt to rapidly remove the unit from service.

Event 8: RCS leakage exceeds VCT makeup capability E-0 manual Rx Trip with ATWS Reactor Trip Breakers fail to open in auto or manual. The crew will continue actions of AOP 038 and attempt a manual Reactor trip when leakage exceeds VCT makeup capability lAW AOP-0i6 continuous action step 4. When the RO attempts to perform a Manual Reactor trip he/she will find that the Reactor will not trip from either of the MCB trip switches. The crew will enter FR-S.i, Response to Nuclear Power Generation/ATWS and perform the immediate actions of manually inserting control rods (auto rod insertion is not available due to event 2 malfunction), tripping the main Turbine, starting AFW and directing an AC to locally trip the Reactor. The Booth Operator acting as the field operator will locally trip the Reactor trip breakers when the crew initiates emergency boration of the RCS. The crew will then continue with FR-S.1 until completing step 10 or when the foldout for Reactor Subcriticality Criteria is met. At that time they will transition from FR-S.i to E-0.

Event 9: RCS leakage increases to SBLOCA -650 gpm. After the crew transitions from FR-S.i to E-0 they will verify the E-0 immediate actions and actuate SI if automatic SI initiation (PRZ Pressure less than 1850 psig) did not occur after the Reactor trp breakers were opened.

The RCS leakage will increase in magnitude. RCS pressure will decrease to less than 1400 psig which will require the crew to trip the RCPs based on E-0 foldout criteria of RCS pressure less than 1400 psig and SI flow greater than 200 gpm.

Harris2Ol3 NRCExamScenario2 Rev.O Page7of9 U

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 2 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 2 (continued)

Event 10: B ESW pump fails to start from the sequencer or from low pressure. This failure requires the crew to manually start the B ESW pump. The crew should identify that the B ESW pump did not start from the B sequencer. Either the RO should identify that the B ESW pump is not running by observation of the MCB OR the BOP should identify that the B sequencer has skipped the B ESW pump start. IF neither operator identifies that the B ESW pump did not start then the BOP should identify that it is NOT running when performing EOP E 0 Attachment 3. Step 4 of the attachment has the operator verify that ALL ESW AND ESW Booster Pumps are running. By the time the operator performs this step in attachment 4 the actions of AOP-022 to secure both the B CSIP and the B EDG will have been met. The reason for securing the CSIP and EDG can be found in AOP-022 Basis Document. It states that both the CSIPs and EDGs are considered an essential load and requires the component to be stopped to protect against equipment damage due to overheating.

Securing the I B SB EDG will remove the emergency power supply to the B RHR pump. The B RHR pump will continue to run since it was started by the B sequencer but will only have the normal power source available to it.

The scenario can be terminated when directed to initiate RCS cooldown in ES1 .2.

Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 8 of 9

Appendix D Scenario Outline Form ESbfl HARRIS 2013 NRC SCENARIO 2 CRITICAL TASK JUSTIFICATION:

1. Insert negative reactivity into the core by inserting control rods during an ATWS before completing the immediate action steps of FR-Si.

There are four (4) immediate actions steps of FR-S.1 at HNP with step 4 directing an operator to contact OR report to the MCR to receive instructions to locally trip the reactor. Control rod insertion to add negative reactivity to the core attempting to bring the reactor core subcritical is crucial to prevent the possibility of core damage. Not performing rod insertion prior to completing the immediate actions of FR-S. 1 is demonstrating the lack of ability to complete a required operator action during a function restoration procedure.

2. Trip RCPs once RCP Trip Foldout Criteria is met and prior to exiting E-1.

Securing RCPs during a SB LOCA event will prevent depleting the RCS to a critical inventory by pumping more mass through the break than would occur if the RCP operation were ceased. (Critical inventory is defined as the amount of inventory remaining in the RCS when the break completely uncovers and the break flow changes from a mixture of liquid and steam to all steam.) Both E-0 and E-1 foldout criteria requires RCPs to be secured when SI flow of> 200 gpm is established and when RCS pressure is < 1400 psig. IF the crew continues to allow the RCPs to operate then RCS inventory will continue to deplete. Manually tripping the RCPs before depletion below the critical inventory conseriatively ensures that Peak Clad Temperature remains below 2200°F. If the crew does NOT secure the RCPs prior to transition from E-1 then RCP trip criteria will no longer exist and the RCPs could continue to run.

Harris 2013 NRC Exam Scenario 2 Rev. 0 Page 9 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 Facility: SHEARON-HARRIS Scenario No.: 3 Op Test No.: 05000400/201 3301 Examiners: Operators:

Initial Conditions: IC-19, MOL, 100% power

• B MD AFW Pump under clearance 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago for pump packing replacement, due back in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, TS 3.7.1.2 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> LCO

. 1SI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs

. B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

. Boric Acid Transfer Pump B-SB is under clearance for motor replacement Turnover: Plant is operating at 100% steady state power. Maintain present conditions.

. Isolate AFW flow to the ruptured C SG prior to entering ECA-3.1, SGTR with Loss of Reactor Coolant: Subcooled Recovery Critical Tasks:

• Isolate 1 MS-72 Prior to entering ECA-3. 1, SGTR with Loss of Reactor Coolant: Subcooled Recovery

. Shut A and B MSIVs prior to exiting E-2, Faulted SG Isolation Event No. MaIf. No.

[ Event Type*

CRO/SRO Event Description 1 prso6a Pressurizer PORV 445A Leakage 2 nisO3 a N 1-31 high voltage block failure and energization at power I BOP/SRO (OP-i 05) eps05a C BOP/SRO Loss of 1A-SA Emergency Bus with failure of A CSIP Room zdsq94:6a TS SRO HVAC to restart (AOP-025) ccw0l a C RO/SRO Trip of A CCW Pump on 0/C during A EDG sequencer start with ccwO47 TS SRO standby CCW pump failure to auto start (AOP-014) pt3O8c R RO/SRO SG C PORV Pressure Instrument fails high and the PORV stays 5 open requiring the crew to reduce power below 100%. (AOP-042)

N BOP/SRO 6 sgno5c M ALL C Steam Generator tube rupture 420 gpm (AOP-016) 7 mssii Main Steam Header break outside Containment (downstream of MALL MSIVs) zrpk5o4a 8 zrpk5O4a C BOP/SRO Main Steam Line Isolation Signal Fails, C MSIV fails to shut mss05c (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 1 of 10 ,V

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 The plant is operating at -1 00% power in MOL. When turnover is complete the crew will be directed to maintain the current plant conditions.

The following equipment is under clearance:

B MDAFW Pump is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and will be restored to operable status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Tech Spec 3.7.1.2 LCD Action a and Tech Spec 3.3.3.5.b Action c applies.

SYSTEM L[1ITi V3 f(4j)]1 4 O?A!1 JN 3 I I c cr e i n1 .. Wfl lE(P I <1 t T - r prn sscci atec flow atis h: 1 be iPEFiRi F with:

Iwo Totcr-dri?r wx.i 1iiry 1eweiter pulIps, :ch capthio ot e.ing pcweied frar. spaats enerency buses. ird h, Ore steani turbrodrveri aui1iary fedater pur capib1e of being plwered frcm 9fl OPER4BLE 5tEiW SUPPlY system.

APP[ICAEIIITY: OULS 1. 2. iro J, ACT J{N

a. Lith ne auiiar: featr pjino inoperbe, restore the reoired aui Ii ar feedtv urns LPERBLE sttas w ti 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> in at 1et 4J1 S7ANEE3 wLh1n Ih n>.t 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> rd n 4OT sHJTCMN wth1 n LFe 101 1rrirrg 3 hcur.

TE 5M(SThOa4 v1tM UITTNG CCUFONRQP4flai4 L33.5.b Al! trinsier ifthe, ,uxfliy Córttil Panal C4ntio1s ad Azxi1iay Tra1eF Pan1 ct-1 f fl at qt1ed by t*

SHNPP 5fq 5 toawn Ana1y!is t (1) rvi dicay hut da a1ary fu*.ata tlo and staa naratr p.,o.ratad reliaf aTve 9ow fi stea geator A (2 iottrol R inventory thro4* t ni-a1 rin f1w peto (3) cntrl RCS pr.swe, (4) cQntral reatvity, nd (5) re decay the R1R tn ha1i e QPRAaLE.

pLcAsnrr; !COES !. 2. and C. With n. ar r* lncparale R.ti Sut4awn yst.m triMfar switches power, or eazitrol uft required ty 33.35.b, rtIto1* the noper able %wftch(s)fgfruft(s) to OPERAaL! stats withIn 7 day or t. ir HOT ST8Y within the next U hours.

Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 2 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued

• B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• 1SI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs.

Tech Spec 3.5.2 Action a applies. OWP-SI-01 has been completed.

f.pjy CcE LII3 SYSUM J45.2 ECOS WBSYSTEMS T GREATER TW (JR ECUAL TO 3&PF LIMITINIi C0)D]TJON FCf OPERATION 352to independent Ernerncy Ccre tooling System (EcCS) subsystes shaH b OP[R5FLE with each subytem ccpris:ed of:

a. On OPERLE CharginIsafety injecti pnp.

b. On UPERI4LE MH heat Exchanger,

c. On OPEMLE RHR pwnp, and An OPERABLE flo path capable of taking suc:io frm tie refueling water stcrage tik on a Safety Injection signal cr4. tipon being mauayaligned transferring Suction to the COfltaiPfflt SUF Jur riy the rec rculdl ui p1ise uP uperdt ion.

ELi23ILII: MODES L 2 and 3.

ACT 1ON:

a. Wth one ECCS subsystem inoperable, restore the inDperable

,ujytvn to OPERAOLE ttus within 2 hui or be ri at let rOT STNO8Y within the nex 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and n HOT SJTDOW with, the

- rtr

• Boron Injection Pump B-SB is under clearance for motor replacement. Tech Spec 3.1.2.2 applies (tracking only). OWP-CS-05 has been completed.

\\

Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 3 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 1: Pressurizer PORV 445A Leakage: Pressunzer PORV 445A leakage. This failure will cause PRZ PORV 445A to leak, resulting in rising PRT pressure and level. PORV Line Temp indicator TI-463 will increase as observed on the MCB and the crew will respond lAW ALB 009-8-2, PRESSURIZER RELIEF DISCHARGE HIGH TEMP. The crew may utilize AOP-016 to determine which PORV is leaking. The SRO will evaluate Tech Spec 3.4.4, Reactor Coolant System Relief Valves.

TS 3.4.4 applicable LCO is Action a, 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to restore REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: MODES 1. 2. and 3 AGII0i:

a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

\U Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 4 of 10

Appendix D Scenario Outline Form ES-D-i

. HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 2: Nl-31 high voltage block failure and energization at power (OP-105). The high voltage block on NI-31 will fail and cause the Source Range to suddenly energize and the audio count rate to become audible. The crew should respond to the failure by implementing OP-i 05, Excore Nuclear Instrumentation, Section 8.2 Inadvertent Source Range Detector Energization at Power. They will promptly de-energize Nl-3i by removing the il8V 5A instrument power fuse. The SRO should refer to TS 3.3.1 (no action required above P-6) and direct the implementation of OWP-RP-1 9.

3/4.3 INTRUMENTATIUN CTR TRIP SYSTEM INSTRUMEFTATiO LIM[TING CONDITION FOR OPER\TION 33.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 33-i shaH be OPFRABLE.

APPLICABILITY: As shown in Table 33-I.

ACTION: s shoin in Table 33-i.

MINIMUM TOTAL N. CHAELS CmANNLs APPL1CBLE FUNCTIO1JAL iNI1 Ft4tJNEL Iijjj QLE _MOOES ACTION

6. Saure Rng, Nutrrn Flux

a. Startup 2 1 2 2 4

b. 5hutdwn 2 1 2 3, 4, 5 5

1. 1. Below the P-6 (1ntermeiate Range fleutron Flux Interlock) Setpoint.

L Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 5 of 10

• Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 3: Loss of 1A-SA Emergency Bus with failure of CSIP A Room HVAC to restart (AOP-025). Emergency Bus A SA to Aux Bus D Tie Breaker 105 SA trips open causing a loss of power to the Emergency Bus A. The Emergency Diesel Generator A starts and loads the bus. The crew will enter AOP-025 and perform the immediate action of checking any CSIP running. The crew should NOT perform the RNO action to isolate letdown since guidance is provided in AOP-025 Basis Document stating that this should only be done if the sequencer does not start the CSIP. IF the letdown flow is secured then the crew will have to restore letdown lAW OP-i 07. During the recovery steps of AOP-025 the BOP should identify that the CSIP A room HVAC is NOT running. Since this is a Verify step the BOP should start the fan.

The SRO will be directed by AOP-025 to review multiple Tech Specs due to the loss of power to Tech Spec related systems. Tech Spec 3.0.3 is the most limiting action due to the isolation of the CNMT vacuum relief valves caused by 2/4 Radiation monitors failing high after losing power.

Q4. REFER TO the following Tech Specs:

• 30.3 (Due to loss of 214

• 34.61 RCS Leak Detection containment fad (Due to RM-3502A monitors and CVIS mop) affect on CNMT

• 3.6.5 Vacuum RelieT System vacuum reliefs)

• 38L1 AC Sources Operating

• 3&t2 AC Sources Shutdown

• 33.3 1 Radiation Monitoring for

• 3.8.2,1 DC Source.s Operating Plant Operations

• 3.831 Onste Power (Due to inoperable Distribution Operating Control Room Outside

• 3&32 Ons[te Power Air Intake Monitors) Distribution Shutdown Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 6of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 4: Trip of A CCW Pump on 0/C during EDG A sequencer start with standby CCW pump failure to auto start (AOP-014). During the loss of power to IA-SA Emergency Bus the A CCW Pump will be started by the sequencer and after 20 secondsjt will trip on overcurrent.

The standby B CCW Pump fails to Auto Start due to a pressure transmitter failure (instrument is isolated therefore pressure decrease is not sensed). The crew should recognize the loss and enter AOP-014, Loss of Component Cooling Water in conjunction with AOP-025. AOP-025 additionally, has direction to recover the CCW pump lAW AOP-014. AOP-014 will direct the restoration of the CCW system. The RO will be directed by the SRO to manually start the B CCW (or will have started it lAW OPS-NGGC-1 000 when it did not auto start). The SRO should also evaluate Tech Spec 3.7.3, Component Cooling Water System.

PLANT SYSTEMS

/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3,7,3 At least two component cooling water (CCW) pumps, heat exchangers and essential flow paths shall be OPERA3LE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

i4ith only one component cooling water flow path OPERABLE, restore at least two flow paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within, the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

• The breaker for ccw pump iCSAB shall not be racked into either power source (SA or SB) unless the breaker from the applicable CCIJ pump (IA-SA or lB-SB) is racked out.

i Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 7 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 5: SG C PORV Pressure Instrument fails high and the PORV stays open requiring the crew to reduce power below 100% (AOP-042). A transmitter failure will cause the C SG PORV to fail 100% open. Steam Generator C PORV Pressure transmitter PT-308c fails high causing the C SG PORV to open in automatic. The crew should identify this failure by annunciator ALB-014-8-5, Computer Alarrn-SteamGenerators alarming and status light indications for the C SG PORV. Note:, The PT-308c does not have MCB indicatio)The BOP will be directed by the SRO to take manual contrFTfhe PORV idshutitThevaIve control will NOT respond and an Aux Operator will be dispatched to locally shut the isolation valve to the PORV. During this time Reactor Power will exceed 100%. The combination of-the steam leak and Reactor power will cause the crew to enter AOP-042, Secondary Steam Leak/Efficiency Loss. AOP-042 is used to rapidly reduce Reactor power to < 100% at a ramp rate of up-to 45 MW/mm. When Reactor power has been stabilized below 100% the AC that was dispatched to locally isolate the PORV will report back that the valve has been shut. When power is stabilized to the satisfaction of the Lead Examiner eventcan be introduced.

The SRO should evaluate Tech Specs 3.6.3, Containment Isolation Valves and PLP-106 Technical Specification Equipment List Program and Core Operating Limits Report. If the Tech Specs are not referred to during the scenario then if required ask a follow up question at the end of the scenario dealing with the LCO.

TS 3.6.3 Action c, isolate the affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The redundant manual isolation valve per PLP-106 is Containment Isolation valve 1MS-63.

JjNfET$(STEj 3/8. .3 CONT NENT ISOLATION VALV LINITING C DITWU FOR 0PERTI0N 3...3 Each co7itanment soiation ale apecffie In the Technical Specfcait1Gn Equiprnenl List rograri plant proc.sdure PLP-106 shall be OPERABLE with 1solt1oi tires 1es than er equal t required tsulatiu times..

APPLICBILITY FDDES 1, 2, 3, and 4.

ACT TON With one r more of the containment isolatmcn vaiva(s) inoperable, maintain at least one .olation va1v OP[iABLE n each affected pmnntraten th.at is open arid:

a. Restore the inoperabe valve(s) to ERA8IE status ithini 4 hours, or L !solae each affected pe trton within 4 hous y use F at least one deactivated auteestic valve secured in the isolatofl positiUm, er

c. lscjlate each affected penetration ithi:n 4 hauis by use of at least one closed manmal lve or liji flange. r d.. 8e lii t leest HOT ST.ANI)BY w th h t1t, i*. 6 ai ii COLD SHUTDOWN within the following 30 )ours, Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 8 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 3 continued Event 6: C Steam Generator tube rupture 420 gpm (AOP-01 6). C Steam Generator (SGTR) one tube sheared. Break flow of 420 gpm. The crew should recognize the presence of a large leak in the primary and announce entry into AOP-01 6. Due to the leak size the crew will promptly recognize that the leak is beyond CVCS makeup capability and the RCS pressure is being rapidly reduced. Prior to reaching the Reactor trip setpoint on Pressurizer Low Pressure (1960 psig) they should manually trip the Reactor, carry out the immediate actions of E-0 and time permitting manually initiatfety Injection (an automatic Safety Injection may occur if actions are not promptly taken).

Event 7: Main Steam Line Header break outside Containment. Five minutes after the Reactor is tripped a Main Steam line break on the main steam header outside Containment will occur. It is expected that the crew transition from E-0 to E-3 to address the ruptured Steam Generator.

While in E-3 the faulted Steam Generator will become apparent by the rapid reduction in Steam Generator pressure.

Event 8: Main Steam Line Isolation Signal Fails, C MSIV fails to shut. The crew should attempt to manually actuate the Main Steam Line Isolation due to approaching the criteria Any SG Pressure Less Than or Equal to 601 psig. With this failure the crew will attempt to shut the MSIVs manually on the MCB, but only A and B MSIV can be manually shut. C MSIV will not shut from the MCB or locally. The crew should use the Secondary Integrity Foldout Criteria to address the faulted C Steam Generator and transition to E-2, Faulted Steam Generator Isolation, (if isolation attempts were not performed during the implementation of E-3). After entry into E-2 for C Steam Generator isolation the crew will return to E-3.

The scenario will end after the crew transitions to ECA-3. 1, SGTR with Loss of Reactor Coolant:

Subcooled Recovery and initiates an RCS Cooldown.

Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 9 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 3 CRITICAL TASK JUSTIFICATION:

1. Isolate AFW flow to the ruptured C Steam Generator prior to entering ECA-3.1, SGTR with Loss of Reactor Coolant: Subcooled Recovery.

Failure to isolate the ruptured SG causes a loss of differential pressure between the ruptured SG and the intact SGs. Upon a loss of differential pressure, the crew must transition to a contingency procedure that constitutes an incorrect performance that necessitates the crew taking compensating action which complicates the event mitigation strategy. This critical task requires the crew to isolate the feedwater flow to the ruptured SG. Any delay in the isolation of feedwater adds additional inventory along with the primary to secondary leakage.

If the primary to secondary leakage is not stopped, SG inventory increase leads to water release through the PORV or safety valve which has the potential for an unmonitored radiation release. Continued filling of the SG could lead to SG overfill which could fill the SG steam lines with water and potentially cause a steam line pipe failure due to the additional weight of the water in the steam lines.

2. Isolate 1 MS-72 prior to entering ECA-3. 1, SGTR with Loss of Reactor Coolant:

Subcooled Recovery.

Failure to isolate the ruptured SG causes a loss of differential pressure between the ruptured SG and the intact SGs. Upon a loss of differential pressure, the crew must transition to a contingency procedure that constitutes an incorrect performance that necessitates the crew taking compensating action which complicates the event mitigation strategy. This critical task requires the crew to isolate the steam flow from the ruptured SG. Isolation of the ruptured SG is necessary because the crew cannot start RCS cooldown to establish RCS subcooling margin. Delaying isolation of the ruptured SG steam flow further depressurizes the ruptured SG increasing the differential pressure between the RCS and ruptured SG requiring the RCS to be depressurized to a lower value.

This delay increases the duration of the primary to secondary leakage.

3. Shut A and B MSIVs prior to exiting E-2, Faulted SG Isolation.

Failure to isolate a faulted Steam Generator that can be isolated causes challenges to the Critical Safety Functions beyond those irreparably introduced by the postulated conditions. Also, depending upon the plant conditions, it could constitute a demonstrated inability by the crew to recognize a failure of the automatic actuation of an ESF system or component. This critical task requires the crew to recognize an automatic actuation should have occurred of an ESF system or component but has not and then take manual operator actions to perform the isolation.

Harris 2013 NRC Exam Scenario 3 Rev. 0 Page 10 of 10

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 4 Facility: SHEARON-HARRIS Scenario No.: 4 Op Test No.: 05000400/201 3301 Examiners: Operators: SRO:

RO:

BOP:

initial Conditions:

• IC-19, MOL, 100% power

• B MD AFW Pump is under clearance for pump packing replacement

. 1SI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs

. B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• Boric Acid Transfer Pump B-SB is under clearance for motor replacement Turnover:

• A plant shutdown is required due to problems encountered during the repairs on the B MDAFW pump. Repairs will not be able to be completed prior to the LCD expiring.

The plant is operating at 100% power in MDL. When turnover is complete a power reduction at 4 DEH units/mm must be started to support being in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. All required notifications have been made to individuals concerning the reason for the shutdown.

Critical Task:

• Open 1 MS-70 or 1 MS-72 to establish a minimum of 210 KPPH AFW flow to the Steam Generators prior to exiting ECA-0.0

• Energize B AC emergency bus when offsite power becomes available prior to aligning equipment for extended power loss (step 11 of ECA-0.0)

Event No. Malf. No. Event Type* Event Description 1 N/A Plant Shutdown (GP-006) lRO/SRO 2 nis08b PR NIS Channel N-42 fails HIGH (AOP-001) 3 genOl C BOP/SRO Generator Voltage Regulator Failure (APP-ALB-022)

I RO/SRO Controlling Pressurizer Level Channel, LT-459, fails HIGH 4 lt459

. TS SRO (APP-ALB-009) 5 hva04 C BOP/SRO A Emergency Services Chilled Water Pump Trip (AOP-026) 6 cfwl 7b C MnFeedwater Pump 1 B Recirculation Valve (1 FW-39) fails RO/SRO 7 epsOl a M ALL Loss of Offsite Power, Reactor Trip EDG A output breaker trips prior to Load Block 9 8 dsg38, dsgo5b C BOP/SRO EDG B fails to start Loss of ALL power I restoration possible with offsite power zl974tdi 1MS-70 and 1MS-72 fail to auto open C BOP/SRO zi 975tdi (Loss of all AFW until operator opens 1 MS-70 or 72)

(N)ormai, (R)eactivity, (i)nstrument, (C)omponent, (M)ajor Harris 2013 NRC Exam Scenario 4 Rev 0 Page 1 of 9 1

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 A plant shutdown is required due to problems encountered during the repairs on the B MDAFW Pump. Repairs will not be able to be completed prior to the LCO expiring. The plant is operating at 100% power in MDL. When turnover is complete a power reduction at 4 DEH units/mm must be started to support being in Mode 3 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. All required notifications have been made to individuals concerning the reason for the shutdown.

The following equipment is under clearance:

• B MDAFW Pump is under clearance for pump packing repairs. The pump has been inoperable for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and will be restored to operable status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Tech Spec 3.7.1.2 LCD Action a and Tech Spec 3.3.3.5.b Action c applies.

LA SYS L huT [NG COld) I WI TR OPERA Hit) 3.L1 2 At 1est thi-ae 1ndeeTr1oet. ctear genrat a: ey fater punos ano associaTed fc paths sha I be jPERAB[E with:

i. i:o rroter*-driven auxilidny tutor puops, ech capable o po.r-d I roe srp rate enierger icy bires a h flr steae turhind-rJr eui 1 iy ter )u b1e Of being powCred i:roin an IWRABLF steaiu supply systen.

APPI ICABILIJI: MODIS 1. 2. arid 3 ACTION a, With one auxiliary ft ater pui inoperable, restore the reouired uai1tary edater puo l OPFA& tatlis win 72 huns a 1 at least -1J1 5ANDBY nIhin ih r>t b hours ei n IV)1 SHUTIXY-JN wthin the fol iOwiruj i hur -

INS UEN1ATtON E?rE siesrooim STt74 IR OPERATIGN -

.3.3.5.b All tiinsfer swftli.s, Auxiliary Cntr1 aet Contri w4 Auilia Tranrer Piri.i Cto1 fr ta GPE LLXTV f tios* czpon4IJ requf red y eJe SHNPP Safe Sut4cwn Analyss (1) rov. decay hut da awlltar eadwiter flow and .ta rator pa roaratad rillif vele flow f st Uan.rat,ors A and a, (2) Otr1 thuh th. nra1 ch*ln a (3) control RCS resuri, (4) trol reactivity, and (5) remove decay Irnet via the )iR ta safl be ERA1LE.

APPLICABIliTY: - lCQS 1, Z, and 1.

C. With i or aor* ncpeiWli Ri,te S)gt4awn Systi tiensfai 1wItCha or control Circuits required by 3,. 3.3.. i.b, reltore the moper able switch()/cfrcujaCe) ts FEA3U status within I days,. or b -in HOT ST,C1Y within the next U hours Harris 2013 NRC Exam Scenario 4 Rev 0 Page 2 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued The following equipment is under clearance (continued):

• B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

• I SI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs.

Tech Spec 3.5.2 Action a applies. OWP-SI-01 has been completed.

EiIENCY W coni r NJES 3I452 ECS SUBSYSTEMS - T GREATER INAM OR EQUAL TO 350cF L INITING CODJTJON FOR OPERATION 352 T dependent Energency Core ccolirg System (EcCS) subsystems shail be OPERABLE with each subsystem conpris

. One OPRMLE Chrgingfety injcctiei paip.

b One OPERM3LE RUR heat echarer,

c. One OPERABtE RHR ptp, and d Mi OPEPBLE flow path capable of taking suction frci the refu&ing ter storoge twik on o Sofety 1nection ignal end. upon ben nanuay alignd transferring suction to the cota nrient sur during the red railatian phase of cperati on.

fjBIlIT: MOOES 1. 2. and 3.

ACHOII

a. Uth on EtCS subsystem inoperable, restore the noperabie subssWri to OERABt E statLs Mthlr 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in a least HOT STANDB( within the next 6 heurs anc in HOT SHtTEXMN within the irrn t hnr.

• Boron Injection Pump B-SB is under clearance for motor replacement. Tech Spec 3.1.2.2 applies (tracking only). OWP-CS-05 has been completed.

Harris 2013 NRC Exam Scenario 4 Rev. 0 Page3of9

Appendix D Scenario Outline Form ES-D-1

. HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued Event 1: Plant Shutdown (GP-006). Crew performs a power reduction lAW GP-006. For this reactivity manipulation it is expected that the SRO will conduct a reactivity brief, the RO will borate per the reactivity plan and the BOP will operate the DEH Controls as necessary to lower power.

Event 2: PRNIS Channel N-42 fails HIGH (AOP-00i). This malfunction will cause rods to start stepping in at maximum speed (72 steps per minute). The crew should respond by entering AOP-00i, Malfunction of Rod Control and Indication System and perform the immediate actions which will be placing the Rod Control selector switch to MANUAL. The crew will then perform the follow up actions of AOP-00i, implement OWP-RP-24 and OP-i 04 Section 5.5 in order to restore Rod Control to automatic. The SRO will evaluate Tech. Spec 3.3.1 for any impact due to the failed instrument.

HtNIMUM TOTAL NO. CHANNELS CHANNELS APR I CABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABI _..1tO.QEL ACTION I. Nanual Reactor [rip 2 2 1 2 1 2 2 3,4,5 9

2. Power Range, Neutron flux

a. High Setpoint 4 2 3 1, 2 2

b. Low Setpoint 4 2 3 I##W, 2 2

3. Power Range, Neutron Flux 4 2 3 1, 2 2 High Positive Rate

4. Power Range, Neutron Flux, 4 2 3 1,2 2 High Negative Rate ACIIDI4 2 With the number of OPERABLE channels one less than the 10t&

Number of Channels, STARTUP and/cr PcJER OPERATION may proceed provided the fol1oin conditions are satisfied;

a. The inoperable channel is placed in the trpped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,

b. The Minirmini Channels OPERABLE reuirewent is rnet; hever the iroprble rhanncl may be bjpasscd for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveitThnce testing of other channels per Specification 4.3.1.1, and

c. Either, ThERMAL POWER is restricted to less than or equal to 7St of RATED THERMAL POWER and the Power Pane Nejtrcjn Flux Trip Setpont is redue to kss than or rqui1 Ia B5 of RATED THERMAL PaVER wthr 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; :OC, the AERANT POWER TILT RATIO is monitored at least once per 12 ncu. per Specification 4.242, Event 3: Generator Voltage Regulator Failure (APP-ALB-022). The voltage regulator failure will cause generator MVARS to rise above the normal control band. ALB-22-9-4 Computer Alarm Gen/Exciter Systems, ALB-22-4-5 Generator Exciter Field Forcing and ERFIS indications will alert the operators to this condition if not detected earlier by changes in generator MVARS.

Annunciator guidance will have the BOP operator attempting to control voltage with the voltage regulator in MANUAL, but attempts for this type of control will fail requiring the base adjuster to be used to reduce MVARs to a value within normal operational limits (75 to 175 MVAR5). This failure will also require the crew to notify the Load Dispatcher that the voltage regulator is in Manual control within 30 minutes, Harris 2013 NRC Exam Scenario 4 Rev 0 Page4of9 \ -

Appendix D Scenario Outline Form ES-D-1

. HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued Event 4: Controlling Pressurizer Level Channel, LT-459, fails HIGH (APP-ALB-009). The crew should respond in accordance with alarm response procedure APP-ALB-09-4-2 and window 2-

2. The crew should take Charging FCV-122 to Manual and maintain pressurizer level within the control bands and trip limits of OMM-001 Attachment 13. They will shift level control to an alternate channel. The SRO will evaluate Tech. Spec 3.3.1 for any impact due to the failed instrument.

TS 3.3.1 As a minimum the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE Table 3.3-1 1ALE 3.3-1 RtCTOR TRIP SSTE iIBiPJ1QF Il N I NUfl WIAL TL) CHANNLS CHfNELS  ?PPL CtBLL F1NCTICNL U1I NLS 0 TRIP fLPA5E lIXES ACTrOh4 II. Pressurizer Water Level--High 3 2 2 1 5 (Above P4)

ACTIO! 6 with the mbr f OPERA8L ei 1si than tha tit ufnber o annaa, STARTUP ndir POJE OPERATtON may proceed provi4ed the £GiioIing cnd I.óm arE a he perala dtna1. is pLeed iti the tripped con4iin

- withifl 6 hcurs and Te Lnimu. C aina* OPERABLE eqLrement is met; wwever, the LoptabL cheneL viay e 1,ypaaied for a 4 os fG e&rv ince testimg of ote rnets per SpdflcarJon 4 L1+

Harris 2013 NRC Exam Scenario 4 Rev. 0 Page 5 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued Event 5: A Emergency Services Chilled Water Pump Trip (AOP-026). The crew will respond to various alarms on ALB-023, diagnose the event, and enter AOP-026, Loss of Essential Chill Water System (no immediate actions). The SRO will direct the BOP to start the B Train ESCWS Chiller lAW OP-148, Essential Service Chilled Water System. The crew should implement OWP-ECW-01 for the ESCW Chiller IA-SA failure. The SRO should evaluate Tech Spec 3.7.13, Essential Services Chilled Water System and PLP-114, Relocated Technical Specifications and Design Basis Requirements Attachment 4 for Area Temperature Monitoring. Note that the A Chiller will be inoperable for the remainder of the scenario and this will impact plant response during the Major Event in that this failure will prevent Load Block 9 on sequencer Train A from energizing.

TS 3.7.13

/13 ct at ti ir1ont Fsrt:i c>os oi Wtcr SvEi Tos shn H be C P. L

LitAIiL VL3OL 1.

. th on hi 1o h Sts 1c Jp[J to 4 çt t r IflOh 1Ifc5j hour or in n Iei PflT i n V i i! 13 iL 1ITI ih r sO hüurs.

PLP-1 14 Attachment 4 4

0 1 Th e-h 1i ict tx3 ozz

i 1iL zc Q .ayz on.:r feed th -c t ttu, Tth1 A b- r.or tb aO a reir by

a. arid. 4 t.her rt

,it.hi pet Cr aftc.e Harris 2013 NRC Exam Scenario 4 Rev 0 Page 6 of 9

Appendix D Scenario Outline Form ES-D-1

. HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued Event 6: Main Feedwater Pump 1 B Recirculation Valve (1 FW-39) fails OPEN The crew -

should identify that the lB FW pump recirc valve has faild open by MCB light changes from green to red, FW discharge pressure changes, SG Feedflow/Steam flow changes, SG level trends on the ERFIS computer screen displays and by level trends on the WR and NR level recorders. The BOP may attempt to close the valve when the incorrect position is observed but the valve will not close from the MCB. The crew may dispatch the Turbine Building AO immediately or when directed by AOP-01 0, Feedwater Malfunctions. When the crew enters AOP-010 they will initially perform the immediate action to verify that a FW pump has not tripped. The SRO work through procedure steps to have the recirc valve manually closed. The AO will not be successful with shutting the recirc valve and all SG levels will reach OMM-001 trip limits of 30% within approximately 5 minutes. When the Reactor trip is activated event 7 will be automatically inserted.

Event 7 (Major): Loss of Offsite Power, Reactor Trip Once the crew has activated the Reactor trip switch and the Reactor trip breakers open a loss of Offsite Power will occur. The crew will identify that Offsite power has occurred which leads to a loss of all AC when Event 8 occurs.

Event 8: EDG A output breaker trips prior to Load Block 9, EDG B fails to start, Loss of ALL power I restoration possible with offsite power. A EDG will start but the output breaker will trip 30 seconds after energizing the bus. Additionally, Load Block 9 would not have been reached due to Event 5 when the A Chiller tripped.

The crew should enter ECA-0.0, Loss of All AC Power and perform the immediate actions of verifying a Reactor and Turbine trip. The crew will continue efforts to restore power to the station. The Load Dispatcher will inform the crew the fault that has caused the Offsite power failure has been isolated and power has been restored to the switch yard. The Load Dispatcher call is critical and should be performed while the crew is evaluating the status of the diesel generators so they may make a decision to restore offsite power in step 9. If the call for offsite power restoration is delayed the crew could continue with ECA-0.0 extended power loss recovery. They should be allowedjpmak or line up systems for extended power lossclthe scond decision would no!p pjpriaJ The crew should determine that the B Emergency to restore poAer to since the A diesel generator output breaker has tripped and an evaluation of the cause has not been completed. The crew should restore offsite power to the B Emergency Bus lAW EOP-ECA-0.0 Attachment 1.

\O \

Harris 2013 NRC Exam Scenario 4 Rev. 0 Page 7 of 9

Appendix D Scenario Outline Form ES-D-1

. HARRIS 2013 NRC SCENARIO 4 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 4 continued Event 9: 1 MS-70 and 1 MS-72 fail to auto open (Loss of all AFW until operator opens 1 MS-70 or 72). B MD AFW Pump is under clearance and A MD AFW Pump will not have power. The Turbine Driven AFW pump will not start due to failures on both 1 MS-70 and 1 MS-72 to automatically open. The loss of all FW to the Steam Generators will create a RED path on Heat Sink (FR-H.1). With a loss of AC Power ECA-0.0 is in effect. A caution prior to step 1 of ECA

0.0 states

Critical Safety Function Status Trees should be monitored for information only.

Function Restoration Procedures should NOT be implemented unless directed by this procedure. The crew should remain in ECA-0.0 and NOT transition to FR-H.1 when a RED path exists. The crew should identify that there is no Feedwater flow to the SGs and open either 1 MS-70 or 1 MS-72 to establish a motive force to run the Turbine Driven AFW pump. Once flow has been established the TDAFW pump speed controller should be manually adjusted to obtain a minimum of 210 KPPH AFW flow to the Steam Generators.

The scenario is terminated when the crew transitions from ECA-0.0 to E-0, Reactor Trip or Safety Injection then transition from E-0 to ES-0. 1, Reactor Trip Response.

CRITICAL TASK JUSTIFICATION:

1. Open 1 MS-70 or 1 MS-72 to establish a minimum of 210 KPPH AFW flow to the Steam Generators prior to exiting ECA-0.0 Failure to establish the minimum required AFW flow results in adverse consequences or significant degradation in the mitigative capability of the plant. This critical task requires the crew to recognize an automatic actuation of an ESF system or component should have occurred but has not and then take manual operator actions to restore the required flow.

2. Energize B AC emergency bus when offsite power becomes available prior to aligning equipment for extended power loss (step 11 of ECA-0.0).

Failure to energize an AC emergency bus constitutes mis-operation or incorrect crew performance which leads to degraded emergency power capacity. Failure to perform this task also results in the needless degradation of a barrier to fission product release via the RCP seals. Energize at least one AC emergency bus before transition out of E-0, unless the transition is to ECA-0.0, in which case the critical task must be performed before placing safeguards equipment hand switches in the pull-to-lock position. For Harris station safeguards equipment cannot be placed in pull-to-lock so the task would be to energize the emergency bus prior to aligning equipment for extended power loss and locally de-energizing control power to the ESF pumps.

Harris 2013 NRC Exam Scenario 4 Rev. 0 Page 8 of 9

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 Facility: SHEARON-HARRIS Scenario No.: 5 Op Test No.: 05000400/2013301 Examiners: Operators: SRO:

RO:

BOP:

Initial Conditions: IC-27, MOL, -4% power

. Plant startup to full power on HOLD until B Condensate Booster Pump is in service

. B Condenser Vacuum Pump is under clearance for makeup water supply valve problems

. 1SI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs

. Boric Acid Transfer Pump B-SB is under clearance for motor replacement

. GP-005, Power Operation, step 95.c

. Power ascension is on hold for B Condensate Booster pump oil system repairs.

Turnover: Repairs are now completed and the pump is ready for service.

. Start the Second B Condensate Booster Pump AW OP-I 34 Section 5.6.

. With RCS pressure < 1400 psig, establish SI flow of >200 gpm using alternate Critical Tasks: high head safety injection to cold legs prior to securing RCPs

. Manually actuate Main Steam Line Isolation prior to Containment pressure exceeding 10 psig Event No. MaIf. No. Event Type* Event Description 1 N/A N BOP/SRO Start the B Condensate Booster Pump 2 tt:144 I RO/SRO Letdown Temperature Controller fails LD/Diversion Valve fails to jtbl43b bypass demineralizers 3 lt:496 C BOP/SRO Controlling C Steam Generator Level Transmitter, LT-496, fails low TS SRO 4 jfb7579 C-BOP/SRO AH-39 Containment Fan Coil Unit fan trip with back up auto start z27l5tic TS SRO failure (C RCP cooling fan) 5 ccw08a C ROISRO Component Cooling Water system leak requiring AOP-014 entry and manual makeup to maintain level 6 rcs09a C ROISRO RCP A rising vibration requires manual Reactor trip and securing A RCP and associated PRZ spray valve after E-0 immediate actions are completed 7 rcsl8a M- ALL SBLOCA inside containment (100% severity) 8 sisOl 7 C RO/SRO Failure of BIT outlet valve I SI-4 to open requiring alternate high head sisOl8 injection flow path use 9 N/A C RO/SRO Manually trip B and C RCP when RCP trip criteria are met lAW E-0 foldout 10 zrpk504a C BOP/SRO Failure of automatic Main Steam Line Isolation to occur when zrpk504b Containment pressure exceeds 3 psig (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 1 of 8

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 Low power scenario Turnover to crew is the unit startup on hold. The plant is in Mode 2 with Reactor power less than 5%. Power ascension was on hold while B Condensate Booster pump oil system leak repairs are completed. The repairs have just been completed and the pump checkout is completed and ready to be started. When the crew takes the shift the expectation is to start the B Condensate Booster pump lAW OP-i 34, Condensate System, Section 5.6. After the pump is running they will continue with GP-005, Power Operation, to obtain rated power conditions.

The following equipment is under clearance:

• B Condenser Vacuum Pump is under clearance for makeup water supply valve problems.

• iSI-3, Boron Injection Tank Outlet valve is under clearance for breaker repairs.

Tech Spec 3.5.2 Action a applies. OWP-Sl-Oi has been completed.

i[RbEN:Y CTJF. WULINUYf 3/45.2 ECCS SUBSYSTEMS - T GREATER TM OR EO(JAL TO 35OF UMITING COMOITION FOR OPERATION 352 T, hdependnt Energency Core Cxlirig System :(CCS) subsystems sh1l be OPRAE3LE with each subsystem ccnpri sed of

, One OPERLE Chrgnq/safety Injection ptip.

b. One OPERAC RIR heat excharer,

c. One OPERABLE RHR pump, and d An 1FPRI F f1iw sith riphl nf tkin snrtinn frnm tw rpfw1in water storage [nk on a Safety Injection tgnal and, trpon being nanually- aligned, transferring suction to the coataiment sur during the redreulation ph5e of operation, ARPUC,ABILIP: MO&ES I, 2. and 3.

ACT1O

a. Eh one ECCS subsystem inoperable, restore the noperable subsystem to OERABL status withir 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOEW within the nt G hurs eiid in HOT S*JTON within the

• Boron Injection Pump B-SB is under clearance for motor replacement. Tech Spec 3.1.2.2 applies (tracking only). OWP-CS-05 has been completed.

Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 2 of 8

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 continued Event 1: Start the B Condensate Booster Pump. Upon turnover and assuming the shift the crew will start the B Condensate Booster pump lAW OP-I 34, Condensate System, Section 5.6 Second Condensate Booster Pump Start up. After the pump is in operation the crew will discuss raising power lAW GP-005 to prepare to place the Main Feedwater Regulating valves in service. Prior to the power increase event 2 will occur.

Event 2: Letdown Temperature Controller fails LD/Diversion Valve fails to bypass demineralizers. This failure will cause temperature controller TK-144 output to decrease to zero. Without cooling to the letdown heat exchanger, temperatures observed on TI-143 will increase. At 135°F annunciator ALB-07-3-2, Demin Flow Diversion High Temp will alarm. The crew should respond lAW the alarm procedure. The RO should identify that the divert valve to the VCT has failed to respond. The RO should report the failure to the SRO. The SRO should direct manually bypassing the CVCS Demineralizers, and should also provide directions to the RO to restore letdown temperature to normal utilizing MANUAL control of TK-144. The SRO should provide a temperature band to the RO lAW OMM-OO1, Conduct of Operations, for operation of components in manual. The SRO can find this temperature band guidance in OP 107. With TK-144 controller not in auto the temperature band should be from 110 120°F.

The CVCS Demineralizers should remain bypassed pending an evaluation for continued resin use. Soon after stabilizing from this temperature controller failure event 3 will occur.

Event 3: Controlling C Steam Generator Level Transmitter, LT-496, fails low. The BOP should respond to multiple C Steam Generator alarms on ALB-014 and take manual control of the C FRV Bypass valve in accordance with the alarm response procedures and OMM-OO1, Conduct of Operations. The SRO should evaluate the following Tech Specs for failure of LT-496:

T.S. 3.3.1: As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONALJII OF CHANNELS TO TRIP OPERABLE MODES ACTION

13. Steam Generator Water 3/stm. gen. 2/st. gen. in 2/stm. gen. each 1, 2 6(1)

Level--Low-Low any operating operating stm.

stm. gen. gen.

14. Steam Generator Water Level--Low 2 stm. gen. 1 ste. gen. 1 ste. gen. level 1, 2 6 Coincident With Steam! level and level coincident and 2 stm.,feed Feedwater Flow Mismatch 2 stm.!feed- with I water flow water flow stm.!feedwater mismatch in same mismatch in flow mismatch in ste. gen. or 2 each stm. gen. same stm. gen. stm. gen. level and 1 stm./feedwater flow mismatch in same ste. gen.

(1)The applicable MODES for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 3 of 8

Appendix D Scenario Outline Form ES-D-1 L HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 continued ACTION 6 With the number of OPERABLE channel.s one less than the Total.

Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and

b. The Hinimum Channels OPERABLE requirement is met; however, the inoperable channel nay be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1.

T.S. 3.3.2: The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. Turbine Trip and Feedwater Isolation

b. Steam Generator Water 4/stm. gen. 2/ste. gen. 3/ste. gen. 1, 2 19 Level--High-High (P-14) in any ste. in each nen. ste. cien,

6. Auxiliary Feedwater

c. Steam Generator Water Level--Low-Low

1) Start Motor- 3/stm. gen. 2/ste. gen, 2/ste. 1, 2. 3 19 Driven Pumps in any stm, gen. in gen. each stm.

gen.

2) Start Turbine- 3/ste. gen. 2/ste. gen. 2/stm. 1. 2. 3 19 Driven Pump in any 2 gen. in ste. gen. each stm.

gen.

ACTION STATEMENTS (Continued)

ACTION 19 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following I conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and I

b. The Minimum Channels OPERABLE requirement is met; however.

the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> I for surveillance testing of other channels per Specification 4.3.2.1.

The OWP is not required to be implemented in order to continue with the scenario. If the crew allows SG levels to decrease to <30% they will be required to perform a manual Reactor Trip.

Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 4 of 8

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 (continued)

Event 4: AH-39 Containment Fan Coil Unit fan trip with back up auto start failure (C RCP cooling fan). The fan failure will cause annunciator ALB-029 4-5 Containment Fan Coolers AH 39 Low Flow-O/L to alarm. The crew should identify that the standby fan did not auto start and start the standby fan. There are no Tech Spec actions are required for SRO evaluation for this failure.

Event 5: Component Cooling Water system leak requiring AOP-014 entry and manual makeup to maintain level. A CCW leak in the running pump suction header will develop. The leak will be within CCW Surge Tank makeup capability. The crew should identify the leak by observation of MCB indications for CCW Surge Tank level or MCB annunciators based on CCW Surge Tank low level. The crew should respond to the CCW Surge Tank level change and/or alarm and enter AOP-014, LOSS OF COMPONENT COOLING WATER. The RAB RO will be dispatched to investigate the leak. The crew will maintain CCW Surge Tank level in the normal operating range by opening the demin water make up valve I DW-1 5, on the MCB. Shortly after being dispatched the leak will be identified as a leak in the suction header near the pump. The leak can be manually isolated by closing local isolation valves. The crew will then be required to start the standby B CCW pump and secure the running A CCW pump lAW OP-145. The SRO should evaluate TS 3.7.3.

PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two component cooling water (CCW) pumps*, heat exchangers and essential flow paths shall be OPERABLE.

APPLICABILITY: MODES 1. 2, 3, and 4.

ACTION:

With only one component cooling water flow path OPERABLE, restore at least two flow paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• The breaker for CCW pump 1C-SAB shall not be racked into either power source (SA or SB) unless the breaker from the applicable CCW pump (1A-SA or IS-SB) is racked out.

A Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 5 of 8

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 (continued)

Event 6: RCP A rising vibration requires manual Reactor trip and securing A RCP and associated PRZ spray valve after E-0 immediate actions are completed. During this event the A RCP vibrations will begin to increase and over 3 minutes peak at 28 mils shaft. Note: the shaft vibration instrumentation reads up to 30 mils. The crew will respond to the A RCP malfunction by either identifying rising vibrations or when ALB-01 0-3-1, RCP-A Trouble alarms.

The crew should see the A RCP vibration probe readings are increasing. The crew should enter AOP-01 8, Reactor Coolant Pump Abnormal Conditions and perform the immediate actions of checking any CSIP running (YES). Vibrations will continue to increase and exceed AOP-018 Attachment 1 RCP trip criteria of 20 mils shaft. Since the RCP is NOT operating within the trip limits and the Reactor is NOT tripped, the crew will have to Trip the Reactor, GO TO EOP-E-0, perform the immediate actions of E-0 and return to AOP-018 follow up actions of steps 5-8 when time permits. The SRO should address steps 5-8, stopping the affected RCP and shutting the associated PZR spray valve prior to the manual Reactor trip.

Note: AOP-01 8 recent revisions now direct Tripping the Reactor prior to tripping a running RCP. ALL RCPs must be operating whenever the Reactor trip breakers are closed. Previously two loop power operation was allowed after securing one RCP if the initial power level was <49%.

The crew will then transition from EOP E-0 to ES-0.1, Reactor Trip Response. The Lead Examiner can allow the crew to stabilize the plant then insert the major event.

Event 7: Major SBLOCA inside containment (100% severity). The major event is a SBLOCA (100% severity) on A Loop. The crew should recognize a rapid decrease in Pressurizer level and RCS pressure. If the crew responds quickly to the event they may manually actuate a Safety Injection based on ES-0.1 foldout criteria of not being able to maintain Pressurizer level>

5% or RCS subcooling < 10°F. If they do not respond quickly an Automatic Safety Injection will occur. The crew will then transition from ES-0.1 back to E-0, Reactor Trip or Safety Injection.

They will again carry out immediate actions of E-0.

Event 8: Failure of BIT outlet valve ISI-4 to open requiring alternate high head injection flow path use. 1SI-4 will fail to automatically open with the Safety Injection signal and cannot be manually opened from the MCB switch. Additionally, 1 SI-3 was under clearance and cannot be opened from the MCB due to control power being removed from the breaker. In order to obtain Safety Injection flow the crew will have to use the alternate high head injection flow path as directed by E-0 RNO actions. They should OPEN alternate high head Safety Injection to cold legs valve lSl-52 SA and then identify Safety Injection flow exceeding 200 gpm.

Event 9: Manually trip B and C RCP when RCP trip criteria are met lAW E-0 foldout. Shortly after entering E-0, the crew should recognize that the RCS pressure is low enough to meet Foldout Criteria for securing all RCPs but there is no flow indicated on FI-943 (normal SI flow indication). The crew will have to establish SI flow by opening the alternate high head Safety Injection to cold legs valve 1SI-52 SA. After opening I SI-52A adequate flow (> 200 gpm) will be indicated on FI-940 (alternate SI flow indication) to STOP the B and C RCPs.

/L)

Harris 2013 NRC Exam Scenario 5 Rev. 0 Page 6 of 8 \

Appendix D Scenario Outline Form ES-D-1 HARRIS 2013 NRC SCENARIO 5 SCENARIO

SUMMARY

2013 NRC EXAM SCENARIO 5 (continued)

Event 10: Failure of automatic Main Steam Line Isolation to occur when Containment pressure exceeds 3 psig. As the Small Break LOCA continues to flow RCS to the Containment the pressure in the Containment will continue to rise. An automatic Main Steam Isolation signal is generated when Containment pressure is 3.0 psig. The failure of this signal will require the crew to manually actuate Main Steam Line Isolation. The MCB switch will NOT function requiring the crew to shut ALL MSIVs.

The crew will transition from E-0, Reactor Trip or Safety Injection at step 30 when Containment pressure is checked and found to be NOT normal to E-i, Loss of Reactor or Secondary Coolant step 1. The crew will progress through E-1 based on crew performance they will reach a decision point at step 13.

They will transition from in E-i to ES-i .2, Post LOCA Cooldown and Depressurization, based on RCS pressure> 230 psig and RHR HX header flow < 1000 gpm.

While in ES-i .2 based on RCS cooldown rate exceeding i 00°F/HR they will have to wait prior to reducing RCS temperature further.

The scenario ends when the crew has determined that the 100°F/HR cooldown rate has been exceed.

Harris 20i 3 NRC Exam Scenario 5 Rev. 0 Page 7 of 8

Appendix D Scenario Outline Form ES-D-1 L HARRIS 2013 NRC SCENARIO 5 CRITICAL TASK JUSTIFICATION:

1. With RCS pressure < 1400 psig, establish SI flow of> 200 gpm using alternate high head safety injection to cold legs prior to securing RCPs Securing RCPs during a SB LOCA event will prevent depleting the RCS to a critical inventory by pumping more mass through the break than would occur if the RCP operation were ceased. (Critical inventory is defined as the amount of inventory remaining in the RCS when the break completely uncovers and the break flow changes from a mixture of liquid and steam to all steam.) The PRZ Steam Space LOCA event in this scenario is a SB LOCA that requires the RCPs to be secured when E-0 foldout conditions are met. IF the crew continues to allow the RCPs to operate due to lack of establishment of SI flow of> 200 gpm then RCS inventory will continue to deplete.

Manually tripping the RCPs before depletion below the critical inventory conservatively ensures that Peak Clad Temperature remains below 2200°F. NOT establishing SI flow prior to RCS reaching 230 psig (RHR injection pressure) was chosen as an applicable plant parameter to use for grading criteria for the task of securing RCPs.

2. Manually actuate Main Steam Line Isolation prior to Containment pressure exceeding 10 psig Containment pressure increasing to 3 psig should have caused an automatic Main Steam Line Isolation to occur. Since it has not the crew should manually actuate this isolation to prevent the potential release of contamination outside of Containment from the event. At 10 psig a Containment Phase B signal is actuated which will cause the remaining Containment Isolation valves to automatically close thus preventing a release of potential contamination. Since the 10 psig Containment Phase B isolation signal is the last signal to automatically isolate potential Containment release pathways this pressure was chosen as an applicable plant parameter to use for grading criteria for the task of Manual Steam Line Isolation.

Harris 2013 NRC Exam ScenarioS Rev. 0 Page 8 of 8