ML071360136

Draft - Outlines (Folder 2)

ML071360136
Person / Time
Site:	Crane
Issue date:	03/09/2007
From:	AmerGen Energy Co
To:	D'Antonio J Operations Branch I
Sykes, Marvin D.
Shared Package
ML060800145	List: IR 05000289/2007301 ML071090575 ML071350626 ML071360136 ML071360144 ML071360151 ML071360155 ML071450423 ML071760022 ML071760030 ML071760031 ML071760138 ML071770540 ML071770553
References
50-289/07-301
Download: ML071360136 (46)

Text

TMI 2007 NRC Initial License Written Examination Written Examination Outline Methodology The written examination outline was developed using a proprietary electronic random outline generator developed by Western Technical Services, Inc.

The software was designed to provide a written examination outline in accordance with the criteria contained in NUREG 1021, Revision 9.

The application was developed using Visual Basic code, relying on a true random function based on the PC system clock. The random generator selects topics in a Microsoft Access Database containing Revision 2 of the PWR K&A catalogue. The selected data is then written to a separate data table. The process for selection of topics is similar to the guidance in ES-401, Attachment 1.

The attached outline report and plant specific suppression profile (not used for TMI; ONLY System 025, Ice Condenser, was pre-suppressed) report are written directly from the data tables created by the software. Electronic copies of the data tables are on file.

The process used to develop the outlines is as follows:

For Tier 1 and Tier 2 generic items, only the items required to be included in accordance with ES-401, Attachment 2 are included in the generation process Outline is generated for all topics with KA importance 22.5.

25 SRO topics are randomly selected from Tier 1 AA2 and required generic items, Tier 2 A2 and required generic items, (including all System 034 topics) and Tier 3 generic items (All with ties to 10CFR55.43). 75 RO topics are randomly selected to complete the outline, 100 topics total.

The exam report generated lists the topic (Question) number in the far right column. RO topics are numbered 1-75, and SRO topics are numbered 76-100.

The SRO topics are written in red ink for ease of identification.

Items that are rejected after the initial generation process are placed on the rejected items page. The software tracks which items are added, and a report of outline modification may be generated.

Disposition of any item randomly selected but not included in the outline is documented and included.

000003 ES-401 PWR Examination Outline Form ES-401-2 RO K/A Category Points K K K K K K A A A A G T o t a l 1

2 3

4 5

6 1

2 3

4

=acility:

TMI Unit 1 Date of Exam:

4/22/2007 SRO-Only Points A2 G*

Total Tier

1.

I 1

Group Emergency 2 Abnormal Plant Evolutions Tier Totals Plant Systems Tier Totals

3. Generic Knowledge and Abilities Categories

8.

/9 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 WA category shall not be less than two).

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 k1 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.

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 that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding elimination of inappropriate K/A statements.

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.

Absent a plant specific priority, only those KAs having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.

Select SRO topics for Tiers 1 and 2 from the shaded systems and WA categories.

The generic (G) WAS in Tiers 1 and 2 shall be selected from Section 2 of the WA Catalog, but the topics must be relevant to the applicable evolution or system.

On the following pages, enter the K/A numbers, a brief description of each topic, the topics importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. Use duplicate pages for RO and SRO-only exams.

For Tier 3, select topics from Section 2 of the WA 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 10CFR55.43 NUREG-I021 1

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup:

Charging pump problems Ability to operate and I or monitor the following as they apply to the Loss of Residual Heat Removal System:

RCSIRHRS cooldown rate Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water:

Guidance actions contained in EOP for Loss of CCWInuclear service water 015 I Reactor Coolant Pump Malfunctions1 4 027 i Pressurizer Pressure Control System Malfunction 1 3 029 /Anticipated Transient Without Scram (ATWS) / 1 X

038 I Steam Generator Tube Rupture / 3 3.2 43 3.6 44 4.0 45 054 i Loss of Main Feedwater / 4 I x 065 I Loss of Instrument Air I 8 Ability to determine and interpret the following as they X

AA2 01 apply to the Reactor Coolant Pump Malfunctions (Loss 3 5 76 of RC Flow) Cause of RCP failure Conduct of Operations Ability to explain and apply all system limits and precautions 32 3 8 77 Emergency Procedures I Plan Ability to verify system alarm setpornts and operate controls identified in the alarm remonse manual Ability to determine or interpret the following as they apply to a SGTR: Pressure at which to maintain RCS during SIG cooldown 007 I Reactor Trip I 1 008 / Pressurizer Vapor Space Accident / 3 009 I Small Break LOCA I 3 015 I Reactor Coolant Pump Malfunctionsl 4 022 I Loss of Reactor Coolant Makeup / 2 325 I Loss of Residual Heat Removal System I 4 326 I Loss of Component Cooling Water I 8 I

X X

X X

AA2.06

~x I

2.1.32 EK2.02 +

AA2.17 2.1.2 AK2.07 x I AA2.02 AA1.01 AK3.03 NUREG-I 02 1 2

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 46 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thenno-hydraulics behavior Knowledge of the reasons for the following responses as the apply to the SGTR: Automatic actions associated with hiah radioactivitv in SIG samDle lines EKI.01 029 I Anticipated Transient Without Scram (ATWS) I 1 038 I Steam Generator Tube Rupture I 3 054 I Loss of Main Feedwater I 4 X

X X

X X

X

~

EK3.03 X -

X 47 Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW): I 3.8 Manual control of AFW flow control valves AK3.03 48 -

49 Knowledge of the operational implications of the following concepts as they apply to the Station Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: Definition of saturation conditions, implication for the svstems EK1.02 4.1 055 /Station Blackout 16 AK1.04 3.1 50 056 I Loss of Off-site Power f 6 057 I Loss of Vital AC Electrical Instrument Bus I 6

058 I Loss of DC Power 16 Emergency Procedures I Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.

2.4.4 4.0 51 -

52 -

53

~

2.8 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation Knowledge of the reasons for the following responses as they apply to the Loss of Nuclear Service Water The automatic actions (alignments) within the nuclear service water resulting from the actuation of the ESFAS Knowledge of the interrelations between the (Inadequate Heat Transfer) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

AKI.01 X -

AK3.02 3.6 062 I Loss of Nuclear Service Water I 4 EK2.2 4.2 54 55 E04 I hadequate Heat Transfer I 4

~

Knowledge of the reasons for the following responses as they apply to the (Excessive Heat Transfer) Normal, abnormal and emergency operating procedures associated with (Excessive Heat Transfer).

3 5 X

EK3.2 E05 I Excessive Heat Transfer I 4 NUREG-I 021 3

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 E10 I Post-Trip Stabilization I 1 K/A Category Point Totals:

Knowledge of the reasons for the following responses as they apply to the (Post-Trip Stabilization) RO or SRO function as a within the control room team as appropriate to the assigned position, in such a way that procedures are adhered to and the limitations in the facilities license and amendments are not violated.

4.0 56 EK34 X

Group Point 1816 213 4

3 6

1 213 Total:

NUREG-I 021 4

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 3.6 3.4 3.7 4.0 82 83 84 85 2.4.30 T Emergency Procedures / Plan Knowledge of which events related to system operabonslstatus should be reported to outside agencies Ability to determine and interpret the following as they apply to the Accidental Gaseous Radwaste The effects on the power plant of isolating a given radioactive-aas leak 037 / Steam Generator Tube Leak i 3 X

X X

X 060 I Accidental Gaseous RadWaste Release i 9 EA2.04 2.2.25 067 / Plant Fire On-site I 8 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Ability to determine and interpret the following as they apply to the (Natural Circulation Cooldownj Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

ED9 I Natural Circulation Operations I4 EA22 AK3.02 AA1.04 e

3.3 Knowledge of the reasons for the following responses as they apply to the Steam Generator Tube Leak:

Comparison of makeup flow and letdown flow for various modes of operation Ability to operate and I or monitor the following as they apply to the Loss of Condenser Vacuum: Rod position 037 I Steam Generator Tube Leak I 3 X -

051 I Loss of Condenser Vacuum I 4 AA1.02 t AK2.07 Ability to operate and I or monitor the following as they apply to the Accidental Gaseous Radwaste:

Ventilation system 060 I Accidental Gaseous RadWaste Release I 9 X -

~~

~

Knowledge of the interrelations between the Control Room Evacuation and the following: EDlG 068 I Control Room Evacuation I 8 t'

AK3.2 Ability to operate and I or monitor the following as they apply to the (Plant Runback) Operating behavior characteristics of the facility.

A01 / Plant Runback I 1 Knowledge of the reasons for the following responses as they apply to the (Loss of NNI-Y) Normal, abnormal and emergency operating procedures associated with (LOSS Of NNI-Y).

A03 / LOSS Of NNI-Y 17

~~

~~

Knowledge of the interrelations between the (Turbine Trip) and the following. Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

A04 / Turbine Trip / 4 5

NUREG-IO21

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 E03 I hadequate Subcooling Margin I4 EO9 I Natural Circulation Operations I4 KIA Category Point Total:

I 012 NUREG-I 021 Knowledge of the interrelations between the (Inadequate Subcooling Margin) and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Cooldown)

Manipulation of controls required to obtain desired operating results during abnormal and emergency situations.

64 65 0 -

6

I ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 I G I K1 I K2 I K3 I K4 I K5 I K6 I A I I A2 I A3 I A4 I Number I System #/Name KIA Topics I Imp. I Q#

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS, and (b) based on those predictions. use procedures to correct control. or mitigate the consequences of those malfunctions or operations Spray valve failures A2.02 3.9 86 01 0 Pressurizer Pressure Control Conduct of Operations: Ability to locate and operate components. including local controls.

2.1.30 I 3.4 87 -

88 012 Reactor Protection 026 Containment Spray Conduct of Operations Ability to explain and apply all system limits and precautions 2 1 3 2 I 3 8 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system and (b) based on those predictions use procedures to correct control or mitigate the consequences of those malfunctions or operations Restoration of power to a system with a fault on it Conduct of Operations Knowledge of system purpose and or function X

A2.12 3.6 89 062 AC Electrical Distribution X -

2.1.27 2 9 90 103 Containment 003 Reactor Coolant Pump Knowledge of the effect of a loss or malfunction on the followina will have on the RCPS: Containment I 2.8 K6.04 1

isolation vailes affecting RCP operation Knowledge of the physical connections andlor cause-effect relationships between the CVCS and the following systems: CRDS operation in automatic mode control 004 Chemical and Volume Control X

2.7 2

K1.10 Knowledge of the effect of a loss or malfunction on the following will have on the RHRS: RHR heat exchanger 1

2.5 K6.03 3

005 Residual Heat Removal 006 Emergency Core Cooling Ability to manually operate andlor monitor in the control room: Transfer of ECCS flow paths prior to recirculation Knowledge of the effect that a loss or malfunction of the PRTS will have on the following: Containment A4.05 4 -

5 X

K3.01 007 Pressurizer RelieflQuench 008 Component Cooling Water 2.1.2 I

3.0 Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation.

6 Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PORV failures X

A2.03 4.1 7

010 Pressurizer Pressure Control N UREG-102 1 7

NRC Written Examination Outline Plant Systems - Tier 2 Group 1 I

l l

1 010 Pressurizer Pressure Control Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: Pressure detection systems Knowledge of the effect of a loss or malfunction of the following will have on the RPS: Permissive circuits K6.01 K6.10 01 2 Reactor Protection 013 Engineered Safety Features X

Actuation 013 Engineered Safety Features Actuation K2.01 ESFAS/safeguards equipment control 3.6 10 Knowledge of bus power supplies to the following:

Conduct of Operations: Ability to explain and apply 2.1.32 all system limits and precautions.

3.4 11 12 Ability to monitor automatic operation of the CCS,

-1:

Initiation of safeguards mode of operation I 4'1 I 022 Containment Cooling 012 Reactor Protection 026 Containment Spray 2.8 13 -

14 Conduct of Operations: Knowledge of system purpose and or function.

Ability to manually operate andlor monitor in the control room: css controls 2.1.27 A4.01 4.5 K3.01 3.9 15 -

16 Knowledge of the effect that a loss or malfunction of the CSS will have on the following: CCS Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: Main steam pressure 026 Containment Spray X

039 Main and Reheat Steam 059 Main Feedwater 059 Main Feedwater X

AI.06 3.0 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Tripping of MFW pump turbine Knowledge of the effect that a loss or malfunction of the MFW will have on the following: AFW system A2.07 17 18 3.0 3.6 K3.02 Knowledge of the effect of a loss or malfunction of Ability to monitor automatic operation of the ac 061 Auxillary/Ernergency Feedwater 062 AC Electrical Distribution Ability to monitor automatic operation of the dc electrical system, including: Meters, annunciators, dials, recorders, and indicating lights 2.7 21 A3.01 063 DC Electrical Distribution NUREG-I 021 8

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 1 064 Emergency Diesel Generator IL 103 Containment X

103 Containment K/A Category Point Totals:

5/3 3

1 3

X 4

2 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ED/G system controls including: Operating voltages, currents, and temperatures Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Detector failure Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation.

Knowledge of the physical connections andlor cause-effect relationships between the IAS and the following systems: Containment air Conduct of Operations: Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

Ability to monitor automatic operation of the contain-ment system, including: Containment isolation A I.03 A2.02 K1 2'1 '2 K1.03 2.1.33 A3'01 Group Point Total:

I 3.2 22 2.7 23 3.6 24 3.0 25 3.9 I 28 I 2815 NUREG-1021 9

TMI Unit 1 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 Form ES-401-2 ES-401 I

2.1.32 3.8 91 -

92 Conduct of Operations Ability to explain and apply all system limits and precautions Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System and (b) based on those predictions, use procedures to correct control, or mitigate the consequences of those malfunctions or operations Meteorological changes 01 1 Pressurizer Level Control 071 Waste Gas Disposal A2.08 2.8 1x 086 Fire Protection 4-i Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Knowledge of the following operational implications as they apply to the CRDS: Xenon production and removal process 4.0 2.1.23 K5.33 3.2 001 Control Rod Drive Knowledge of the operational implications of the following concepts as they apply to the PZR LCS Criteria and purpose of PZR level program K5.12 2.7 30 01 1 Pressurizer Level Control 017 In-core Temperature Monitor Knowledge of ITM system design feature(s) and/or subcooling monitors 3.4 31 3.1 32 027 Containment Iodine Removal 029 Containment Purge Knowledge of design feature(s) andlor interlock(s) which provide for the following: Negative pressure in containment Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System ; and (b) based those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Abnormal Spent Fuel Pool water level or loss of water level.

K4.02 A2.03 033 Spent Fuel Pool Cooling I 045 Main Turbine Generator 071 Waste Gas Disposal Ability to manually operate andlor monitor in the control room: Turbine valve indicators (throttle, governor, control, stop, intercept), alarms, and annunciators.

35 X

A4.01 3.1 I

I A3.01 I Ability to monitor automatic operation of the W a r r -,, y l y Gas Disposal System including: HRPS X

072 Area Radiation Monitoring Knowledge of the effect that a loss or malfunction of Containment ventilation isolation 1

K3.01 1 the ARM system will have on the following:

I 3.2 37 I

I I

I I

I O NUREG-I021

ES-401 TMI Unit 1 Form ES-401-2 NRC Written Examination Outline Plant Systems - Tier 2 Group 2 WA Category Point Totals:

0/2 0

1 1

2 2

0 0

1/1 1

2 Group Point Total:

1013 NUREG-I 021 11

Facility: I TMI Unit 1 I

Date of Exam:

Topic Knowledge of conditions and limitations in the facility license.

Ability to explain and apply all system limits 2'1'10 2'1.32 and Drecautions.

1.

Conduct of Operations Ability to make accurate, clear and concise I 2'1'17 I verbal reports.

Knowledge of system purpose and or 12.1.27 I unction Subtotal

2.

1 Equipment Control Knowledge of new and spentfiel movement procedures Knowledge of control rod programming.

2'2'28 2.2.33 Knowledge of the process for determining Knowledge of refueling administrative A

Subtotal

3.

Radiation Control 2.3.2 Knowledge of facility ALARA program.

Knowledge of 10 CFR: 20 and related 2'3'1 facility radiation control requirements Knowledge of radiation exposure limits and 2.3.4 contamination control, including permissible levels in excess of those authorized.

Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

Ability to control radiation releases.

2.3.1 0 2.3.1 1

4.

Emergency Procedures I Plan Subtotal 2'4.4' 1 thresholds and classifications.

Knowledge of the emergency action level Knowledge of general guidelines for EOP flowchart use.

Knowledge of the RO's responsibilities in emergency plan implementation.

2'4.14 2.4.39 Knowlc -

emergency plan implementation.

2.4.39 Knowledge of crew roles and responsibilities during EOP flowchart use.

2'4'1 I

Subtotal 4/22/2007 RO I SRO-Only I

3.9 94 3.5 96 2.9 97 2.8 68 2.5 I 69 I NUREG-I021 12

5 1

I 1

P Tier I Randomly Grow Selected WA Reason for Rejection 2 I 1 025 G2.1.27 #I3 System does not exist at TMI. Reselected system 012 Facility does not perform function for the selected topic. Randomly selected AA2.1

~ 2. 0 1 for same topic.

Excessive overlap with audit examination. Impossible to develop significantly different question. Randomly selected EK2.02 for same topic Facility does not have air ejectors, and Off-Gas monitors do not have a reset function. Randomly selected AK3.03 No pneumatic valves in DHR system at facility. Randomly selected K1.05 for topic area Facility does not have Containment Spray Reset switches. Randomly selected A4.01 111 055 EK1 037 AK3.02 Oo4 K1.lo 026 A4.05 211 211 Waste Gas system has no rupture discs at facility. Randomly selected A2.08 for the same topic.

No procedural support for Inadequate SDM in SFP. Randomly selected A2.03

. for the same topic.

07' A2'03 212 033 A2'01 212 NUREG-1021 13

6300003 ES-301 Administrative Topics Outline Form ES-301-1

-acility:

TMI Unit 1 Date of Examination:

April 2007 Examination Level (circle one):

/ SRO Operating Test Number:

NRC Administrative Topic (see Note)

Code*

Zonduct of Operations N

Sonduct of Operations Equipment Control Radiation Control Describe activity to be performed Perform a batch calculation in accordance with OP CONTROL, Enclosure 1.

2.1.25 (2.8): Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data.

Perform a transient RCS leak rate calculation in accordance with OS-24, CONDUCT OF OPERATIONS DURING ABNORMAL AND EMERGENCY EVENTS, ATTACHMENT F.

2.1.19 (3.0): Ability to use plant computer to obtain and evaluate parametric information on system or component status.

Perform Shiftly Checks of Decay Heat Removal Capability in accordance with Surveillance Procedure 1301-1, SHIFT AND DAILY CHECKS, Data Sheet 3 -

Section C.2.

2.2.12 (3.0): Knowledge of surveillance procedures.

Given a set of conditions, determine and apply the facility dose limits.

2.3.1 (2.6): Knowledge of 10 CFR: 20 and related facility radiation control requirements.

11 03-4, SOLUBLE BORON CONCENTRATION N/A - not selected for RO NOTE:

All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria:

(C)ontrol room (D)irect from bank (I 3 for ROs; I for SROs & RO retakes)

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

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

(S)imulator NUREG-1021, Revision 9

0 0 0 0 ~

ES-301 Administrative Topics Outline Form ES-301-1 THREE MILE ISLAND 2007 NRC RO EXAMINATION CONDUCT OF OPERATIONS (Al-I): Perform a batch calculation in accordance with OP 1103-4, SOLUBLE BORON CONCENTRATION CONTROL, Enclosure 1. Given a situation and conditions, determine the applicable section of 1103-4, Enclosure I and utilize the referenced figures and tables to perform a batch calculation. New JPM. RO/SRO Common.

CONDUCT OF OPERATIONS (AI -2): Perform a transient RCS leak rate calculation in accordance with OS-24, CONDUCT OF OPERATIONS DURING ABNORMAL AND EMERGENCY EVENTS, ATTACHMENT F. Given a frozen simulator with an RCS leak in progress, use the plant computer to extract the necessary data and then calculate the leak rate.

The calculation is ROISRO Common. Modify Bank JPM 11205057 by changing the initial conditions and requiring the applicant to retrieve the data from the plant computer on a frozen simulator.

EQUIPMENT CONTROL (A2): Perform Shiftly Checks of Decay Heat (DH) Removal Capability in accordance with Surveillance Procedure 1301-1, SHIFT AND DAILY CHECKS, Data Sheet 3 - Section C.2. Evaluate DH Pump performance, system alignment and the adequacy of support instrumentation with the simulator in cold shutdown on DH cooling; identifying all (two, or more) errors. RO only. New JPM RADIATION CONTROL (A3): Given an emergency situation and survey maps, determine the area dose rate near a specified plant component and apply the applicable facility limit to determine the stay time for the situation. RO/SRO Common. New JPM.

NUREG-1021 Revision 9

ES-301 Administrative Topics Out I i ne Form ES-301-1 Facility:

TMI Unit 1 Date of Examination:

April 2007 Examination Level (circle one):

RO / ml Operating Test Number:

NRC Administrative Topic (see Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan N

N N

P/M Describe activity to be performed Perform a batch calculation in accordance with CONTROL, Enclosure 1.

2.1.25 (2.W3.1): Ability to obtain and interpret station reference materials such as graphs, monographs, and tables which contain performance data.

OP 1 103-4, SOLUBLE BORON CONCENTRATION Perform a transient RCS leak rate calculation in accordance with OS-24, CONDUCT OF OPERATIONS DURING ABNORMAL AND EMERGENCY EVENTS, ATTACHMENT F, and apply the Technical Specification/procedural requirements to the calculation.

2.1.I2 (4.0): Ability to apply technical specifications for a system.

Evaluate a proposed temporary procedure change.

2.2.6 (3.3): Knowledge of the process for making changes in procedures as described in the safety analysis report.

Given a set of conditions, determine and apply the facility dose limits.

2.3.1 (2.6/3.0): Knowledge of 10 CFR: 20 and related facility radiation control requirements.

Given a set of conditions, determine the Emergency Action Level (EAL) and make a Protective Action Recommendation (PAR).

2.4.44 (4.0): Knowledge of emergency plan protective action recommendations.

NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

• Type Codes & Criteria:

(C)ontrol room (D)irect from bank (I 3 for ROs; I for SROs & RO retakes)

(N)ew or (M)odified from bank (> I )

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

(S) i m u la tor NUREG-1021, Revision 9

000003 ES-301 Administrative Topics Out I i ne Form ES-301-1 THREE MILE ISLAND 2007 NRC SRO EXAMINATION CONDUCT OF OPERATIONS (AI -1): Perform a batch calculation in accordance with 1 103-4, SOLUBLE BORON CONCENTRATION CONTROL, Enclosure 1. Given a situation and conditions, determine the applicable section of OP 1103-4, Enclosure I and utilize the referenced figures and tables to perform a batch calculation. New JPM. RO/SRO Common.

CONDUCT OF OPERATIONS (Al-2): Perform a transient RCS leak rate calculation in accordance with OS-24, CONDUCT OF OPERATIONS DURING ABNORMAL AND EMERGENCY EVENTS, ATTACHMENT F, and apply the Technical Specification/procedural requirements to the calculation. Given a frozen simulator with an RCS leak in progress, use the plant computer to extract the necessary data and then calculate the leak rate. The calculation is RO/SRO Common. The SRO applicants will apply the Technical Specification/procedural requirements based on their calculations. Modify Bank JPM 11205057 by changing the initial conditions, requiring the applicant to retrieve the data from the plant computer on a frozen simulator and having the SRO applicant determine and apply the facility requirements.

EQUIPMENT CONTROL (A2): Evaluate a proposed temporary procedure change. Apply facility requirements to the review of a temporary procedure change containing two (or more) administrative and/or technical errors. New JPM. SRO only.

RADIATION CONTROL (A3): Given an emergency situation and survey maps, determine the area dose rate near a specified plant component and apply the applicable facility limit to determine the stay time for the situation. RO/SRO Common. New JPM.

EMERGENCY PLAN (A4): Given a set of conditions, determine the Emergency Action Level (EAL) and make a Protective Action Recommendation (PAR). Randomly selected task from one of the previous two NRC examinations. Modify by changing the conditions. SRO only.

NUREG-1021, Revision 9

8000643 ES-301 Control Room/ln-Plant Systems Outline Form ES-301-2 Facility:

TMI Unit 1 Date of Examination:

April 2007 Exam Level (circle one):

RO / SRO( I) / SRO (U)

Operating Test No.:

NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF)

System / JPM Title Type Code*

Safety Function M. L 1

a, Take corrective action for a Control Rod Drive Sequence Fault in accordance with (IAW) OP-TM-MAP-GOZOZ.

System: 001 2

b.

Take corrective action for a low pressure injection failure during a large break LOCA IAW OP-TM-211-901.

System: 006 Respond to an RCS narrow range pressure instrument failure IAW alarm response procedure MAP G-1-6.

System: 010 Respond to an RCP Seal problem IAW AOP-040.

System: 003 Cross-connect Secondary River Water to Nuclear River Water IAW System: 076

c.

d.

e.

EP 1202-38. (RO Ody) 3 D

4P D

4s 5

f.

Return RB Emergency Cooling to standby following a manual actuation IAW OP-TM-534-901.

System: 022 6

g.

Energize a Vital AC Bus during a loss of off-site power using the SBO Diesel IAW AOP-020.

System: 062 7

h.

Respond to an alarm on Control Room RMS Channel RM-A1 IAW System: 073 MAP C, C-1-1.

11 Implant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) 4s

i.

Reset the steam-driven EFW Pump overspeed trip IAW EOP-10, GUIDE 16.1.

System: 061 NUREG-1021, Revision 9

~00003 ES-30 1 Control Room/ln-Plant Systems Outline Form ES-301-2

j.

Initiate emergency boration IAW EOP-020.

System: APE 068 M, R, E 8

k.

Manually operate RR-V-6, Reactor Building Emergency Cooler Pressure Control Valve, IAW OP-1104-38.

All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

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

(P)revious 2 exams

( W A (S)imulator Criteria for RO I SRO-I I SRO-U 4-6 14-6 I 2-3 9 1 81 4 1 1 11 1

. I / 11 1 s21 2 1 S I

3 I 3 I 2 (randomly selected) 1 1 1 1 1 THREE MILE ISLAND 301-2 JPM

SUMMARY

STATEMENTS

a. Take corrective action for a Control Rod Drive Sequence Fault in accordance with (IAW)

OP-TM-MAP-G0202. Group 7 rods withdraw out-of-sequence when power is being raised from applicant will respond IAW OP-TM-G0202, CRD SEQUENCE FAULT, and correct the overlap problem IAW OP-TM-622-412, RECOVERING FROM A SEQUENCE INHIBIT CAUSED BY EXCESSIVE OVERLAP. Modify Bank JPM 1 1.2.05.159 by changing the initial conditions and therefore the procedure path. To be performed by: RO, SROI, SROU. Failure to properly perform the task will result in a violation of technical specifications and operation outside of accident analysis assumptions.

5% in preparation for synchronizing the main turbine generator. The

b. Take corrective action for a low pressure injection failure IAW OP-TM-EOP-006.

Applicant will assume the watch with directions to perform EOP-006, LOCA COOLDOWN. DH-P-1 B has failed to start, requiring alternative actions. The alternate path is via OP-TM-211-901, EMERGENCY INJECTION. Bank JPM 11.2.05.195. To be performed by: RO, SROI, SROU. Failure to properly perform the task will result in the possibility that all LPI flow is out the break.

NUREG-1021, Revision 9

ES-301 Control RoomAn-Plant Systems Outline Form ES-301-2

c. Respond to an RCS narrow range pressure instrument failure IAW alarm response procedure MAP G-1-6. Shortly after assuming the watch, an RCS narrow range pressure instrument fails HI and SASS fails to block the signal. The applicant will respond IAW the alarm response procedure to place equipment in the proper alignment/position. Bank JPM TQ-TM-104-220-J001. To be performed by: RO, SROI.

Failure to properly perform the task will result in unnecessarily creating saturated conditions in the RCS.

d. Respond to an RCP Seal problem IAW AOP-040. Shortly after assuming the watch, an RCP # I Seal problem will develop, requiring entry into AOP-040, RCP # I SEAL FAILURE. The alternate path is a requirement to initiate a reactor trip, stop the affected pump and close the return isolation valve. This JPM is identified as NEW because the JPM was removed from the facility bank prior to 2005 and was not based on AOP-040.

To be performed by: RO, SROI, SROU. Failure to properly perform the task could result in a LOCA caused by catastrophic seal failure.

e. Cross-connect Secondary River (SR) Water to Nuclear River (NR) Water IAW EP 1202-

38. The only available NR Pump will trip shortly after the applicant assumes the watch with the unit shutdown. SR is then cross-connected to NR IAW EP 1202-38, NUCLEAR SERVICES RIVER WATER FAILURE. Bank JPM 11.2.05.150. To be performed by:

RO Only. Failure to properly perform the task will result in loss of cooling water to emergency safeguards equipment.

f. Return RB Emergency Cooling to standby IAW OP-TM-534-901, RB EMERGENCY COOLING OPERATIONS, following a manual actuation. The applicant assumes the watch with RB Emergency Cooling in service due to a small steam leak that has been isolated and is directed to restore it to ES Standby alignment. Randomly selected repeat from the 2003 NRC Exam (B.1.f). To be performed by: RO, SROI. Failure to properly perform the task could result in inadequate flow to Reactor Building coolers and possible operation outside of accident analysis assumptions.

Energize a Vital AC Bus during a loss of off-site power using the SBO Diesel IAW AOP-020. The applicant assumes the watch after a loss of off-site power and AOP-020, LOSS OF STATION POWER, just entered. The alternate path is to energize ID or 1 E Vital Bus IAW OP-TM-864-901, SBO DIESEL GENERATOR (EG-Y-4) OPERATIONS.

Randomly selected repeat from the 2005 NRC Exam (JPM F). The JPM will be modified in that it will begin in AOP-020 vice OP-TM-864-901 and a different bus will be re-energized. To be performed by: RO, SROI. Failure to properly perform the task will result in loss of DC power to the respective bus or a station blackout if power is lost to the available bus.

h. Respond to an alarm for RMS Channel RM-A1, Control Room Monitor, IAW MAP C, C-1-1. Shortly after the applicant assumes the watch, alarm C-1-1 (HI ALARM) will actuate. The alternate path is to align components IAW OP-TM-826-901, CONTROL BUILDING VENTILATION SYSTEM RADIOLOGICAL RESPONSE OPERATIONS. The task is similar to Bank JPM TQ-TM-104-826-J001 but will be modified to start from an alarm condition; with failed interlock actuations. To be performed by: RO, SROI. Failure to properly perform the task will result in possible radiation intrusion into the control room environment.

NUREG-1021, Revision 9

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

i.

Reset the steam-driven EFW Pump overspeed trip IAW OP-TM-EOP-010, Guide 16.1, EFW FAILURE. Randomly selected repeat from the 2003 NRC Exam (B.2.c). To be performed by: RO, SROI. Failure to properly perform the task will result in loss of all feedwater flow when the operating pump fails.

j.

Initiate emergency boration IAW EOP-020, COOLDOWN FROM OUTSIDE OF CONTROL ROOM, Step 3.1.19. Perform the steps necessary to initiate boration from outside the control room. Modify Bank JPM TQ-TM-105-211-JOOI to have the applicant perform all sub-steps rather than opening just one manual valve (MU-V-51). To be performed by: RO, SROI, SROU. Failure to properly perform the task will result in inadequate SDM during the cooldown.

k. Manually operate RR-V-6, Reactor Building Emergency Cooler Pressure Control Valve, IAW OP-I 104-38. Applicant is directed to raise Reactor Building Emergency Cooling coil pressure with a LOCA in progress. Bank JPM TQ-TM-105-534-JOOI. To be performed by: RO, SROI, SROU. Failure to properly perform the task could result in an uncontrolled leakage path from the Reactor Building atmosphere to the cooling system to the local environment.

NUREG-1021, Revision 9

I CUR0 Appendix D Scenario Outline Form ES-D-1 11 Facility:

Three Mile Island Scenario No.:

1 OpTestNo.:

NRC Examiners:

Operators:

Initial Conditions:

100% power, MOC.

0 0

EF-P-2B is 00s for bearing replacement.

DR-P-1A is running for effluent flow 11 Turnover:

Maintain 100% power operations.

Critical Tasks:

Initiate HPI Cooling (CT-14)

Initiate HPI (CT-2)

L Event

~

Malf.

No.

~

Event Type*

Event Description RW04A I CRS Decay Heat River Water Pump DR-P-1A trips (TS)

CRS NLO Reports an excessive oil leak from Emergency Feedwater Pump EF-P-1 (TS).

M

+

Turbine Header Pressure instrument fails high slowly 3

4 5

6 7

8 I ARO CC04A C CRS Intermediate Closed Cooling Water Pump IC-P-1A trips and IC-P-I B fails to auto start MS19A Steam Leak in the Turbine Building N CRS N ARO R URO M CRS M URO M ARO FW-15A FW-15B TC076 Feedwater Pump 1A trip Feedwater Pump 1 B trip Turbine stop valve fails open Emergency Feedwater Pump EF-P-1 trips on start Emergency Feedwater Pump EF-P-2A does not start. (CT-

14)

FW17 FW18A C CRS C ARO MU086 C CRS C URO High Pressure Injection Valve (HPI) MU-V-16B fails to open B ESAS Manual Actuation Failure (CT-2)

Scenario Event Description NRC Scenario 1 Three Mile Island NRC Scenario #I The crew will take the watch with reactor power at 100% and ICs in Full Automatic. Emergency Feedwater (EFW) Pump EF-P2B is 00s for a bearing replacement. Decay Heat River Water Pump DR-P-1A is running in preparation for a liquid release.

When the crew has accepted the watch the Lead Examiner can cue the initiation of the trip of Decay Heat River Water Pump DR-P-1A. The Crew should respond in accordance with MAP Alarm B-1-5, 480V ES Motor Trip and 8-24, 480V ES Motor Overload. The CRS should review Tech Spec 3.3.1.4.d and declare a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> timeclock based on Tech Spec 3.3.2. When the Tech Spec call has been made the scenario can continue.

After the TS call is made the Lead Examiner can cue the initiation of the NLO report of a large oil leak from EF-P-1 bearing with an empty bearing oil indicator. The CRS should review Tech Spec 3.4.1.1.a.4 initiate immediate action to restore at least two EFW pumps, and Note 1 and recognize all reactor operating mode changes are suspended until a second EFW Pump is operable. The CRS may initiate action to prevent EF-P-1 from starting by manually closing MS-V-I 3A and MS-V-13B. EF-P-1 will not be returned to operable status during this scenario. The scenario can continue when the Tech Spec timeclock is suspended.

After the Tech Spec call is made and on cue from the Lead Evaluator, the Turbine Header Pressure instrument failure can be initiated. The crew should diagnose the failing pressure and take hand control of the turbine and the SG/Rx Master Integrated Control System station. The CRS should implement OP-TM-301-471, Manual Control of the Main Turbine and OP-TM-621-471, ICs Manual Control. The backup pressure instrument may be selected in accordance with OP-TM-621-451, Selecting Alternate Instrument Inputs to ICs. The backup instrument does not have to be selected for the scenario to continue.

After the plant has been stabilized the Lead Examiner can cue the trip of Intermediate Closed Cooling Water Pump IC-P-1A. The crew should diagnose the loss of IC-P-1A and the failure of IC-P-1 B to auto start using MAP alarms AA-1-6,480V BOP Motor Trip, C-1-2, IC CRD Flow Lo and C-2-2, IC System Flow Lo. The CRS should implement 1202-17, Loss of Intermediate Closed Cooling Water and the URO should manually start IC-P-1 B. Letdown flow may be isolated by CRD high temperature closing MU-V-1A and MU-V-1 B. If this occurs the CRS should initiate OP-TM-211-950, Restoration of Letdown Flow. Letdown flow does not have to be re-established to continue.

After IC-P-1 B has been started the Lead Evaluator can cue initiation of the steam leak in the Turbine Building. The crew should diagnose the steam leak in the Turbine Building and the CRS should implement 1203-24, Steam Leak. An NLO will provide a report of severity enough to begin a plant shutdown. The CRS should order a plant shutdown in accordance with 1102-4, Power Operations.

At the cue of the Lead Examiner or after the first Feedwater Pump (FW-P-1A or FW-P-1 B) is secured initiate the trip of the remaining pump to trip the reactor. If the crew trips the reactor due to the steam leak the trip of both Feedwater Pumps will be initiated at e40 % power. One turbine stop valve will fail to close and the contingency steps of OP-TM-EOP-001, Reactor Trip will have to be taken for the turbine. The steam leak will be isolated when the turbine trips. The CRS will initiate OP-TM-EOP-001, Reactor Trip.

000603 Scenario Event Description NRC Scenario 1 Three Mile Island NRC Scenario #I Continued Following the loss of both Feedwater Pumps, EF-P-I will trip (if not previously isolated by the CRS) and EF-P-2A fails to start. The crew should diagnose the lack of heat transfer and respond in accordance with OP-TM-EOP-004, Lack of Primary to Secondary Heat Transfer.

Without primary to secondary heat transfer PORVlHPl cooling is the only method to maintain adequate core cooling which makes this a critical task. PORVIHPI Cooling will be initiated in accordance with OP-TM-EOP-009, HPI Cooling (CT-14). The ARO will initiate OP-TM-EOP-01 0, Guide 16, EFW Failure for the Emergency Feedwater Pump failures.

When HPI is manually initiated HPI Valve MU-V-16B will fail to open and the B Train HPI manual ESAS actuation does not actuate. This will result in inadequate HPI since there will not be one full train of HPI in operation. Without adequate HPI the mitigation strategy is changed for the event, which makes obtaining adequate HPI a critical task. The B Train components will have to be started at the component level in accordance with OP-TM-211-901, Emergency Injection (HPI/LPI) (CT-2).

The CRS will subsequently transition to OP-TM-EOP-006, LOCA Cooldown from EOP-009.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

OOOGQ3 Scenario Event Description NRC Scenario 1 Description Initial Set-up.

DR-P-1 A Trips EF-P-1 Bearing Oil Leak Turbine Header Pressure instrument fails high Intermediate Closed Cooling Water Pump IC-P-1A trips and IC-P-1 B fails to auto start Steam Leak in the Turbine Building Feedwater Pump 1A (or 1 B) trip Turbine stop valve fails to close Emergency Feedwater Pump EF-P-1 trips on start Emergency Feedwater Pump EF-P-2A does not start.

High Pressure Injection Valve (HPI) MU-V-16B fails to open B ESAS Manual Actuation Failure Procedure Support EF-P-26 00s for bearing replacement 3R-P-1A Running in preparation for a liquid release TS 3.3.1.4.d and 3.3.2, ECCS Systems MAP Alarm 8-1-5, 480V ES Motor Trip MAP Alarm 8-2-5, 480V ES ES Motor Overload TS 3.4.1.1.a.4, Decay Heat Removal Capability 3P-TM-MAP-H0302, Sass Mismatch 3P-TM-MAP-H0203, Mn Turb Hdr Press Hi/Lo 3P-TM-301-471, Manual Control of the Main Turbine 3P-TM-621-471, ICs Manual Control 3P-TM-621-451, Selecting Alternate Instrument Inputs to ICs 1202-1 7, Loss of Intermediate Cooling System 4A-1-6, 480V BOP Motor Trip OP-TM-MAP-CO102, IC CRD Flow LO OP-TM-MAP-C0202, IC System Flow Lo 1203-24 Steam Leak 1 102-4 Power Operations OP-TM-EOP-001, Reactor Trip OP-TM-EOP-010, Guide 16, EFW Failure OP-TM-EOP-004, Lack of Primary to Secondary Heat Transfer OP-TM-EOP-009, HPI Cooling OP-TM-211-901, Emergency Injection (HPIILPI).

080683 Scenario Event Description NRC Scenario 1 Initialization IC-16 Console Left Remote Function FWRl3 Main Console EF-P-2B PTL Console Center Malfunction RW04A DR-P-1A NAS Malfunction MSOl B I/O Override IC-P-1 B NAT Malfunction CC04A Malfunction MS19B Malfunction FW15A Malfunction FWI 58 Malfunction TC07B Malfunction FW17 Malfunction FW18A ZDIEFP2A(5) ON MU08B 01A6S28-03A4S02-ZDIPBI RBB 03A4SOI-ZDIPBl RCB 100% HFP, ICs Full AUTO EF-P-28 Tagged 00s Value:

OUT When:

Immediately Robust Barriers applied IAW Risk Document DR-P-1A Running EF-P-2B PTUEF-P-2B BKR OPEN Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Insert Event 1 Insert Sev. 100%

Event 3 RAMP 300 sec Insert OFF Immediately Insert Event 4 Insert Sev. 10%

Event 5 RAMP 300 sec Insert Event 6 ratpwc40 Insert Event 6 ratpwc40 Insert Immediately Insert Event 7 fwnefp1>0.1 Insert Event 8 ratpw40 Insert Event 9 muvmuvl6b>0.1 Insert OFF Immediately Insert OFF Immediately Scenario Support Scenario Support Scenario Support Scenario Support DR-P-1 A Trips Turbine Header Pressure instrument fails high SPIOA-PT2 IC-P-1 B Fails to Auto Start IC-P-1A trip Steam Leak in the Turbine Building Feedwater Pump 1A trip Feedwater Pump 1B trip Turbine Stop Valve B Fails Open Emergency Feedwater Pump EF-P-1 trip Emergency Feedwater Pump EF-P-2A trip High Pressure Injection Valve (HPI) MU-V-16B fails as is B ESAS 4 psig Manual Actuation Failure B ESAS 4 psig 1600 psig Actuation Failure

Appendix D Scenario Outline Form ES-D-1 Facility:

Three Mile Island Scenario No.:

2 OpTestNo.:

NRC Examiners:

Operators:

Initial Conditions:

0 Reactor Startup is in Progress

//

0 NI-3 is 00s due to a detector failure Turnover:

Take the Reactor Critical Critical Tasks:

0 Initiate HPI (CT-2) 0 Trip all RCPs (CT-1)

II 0

Reduce Steaming/lsolate Affected SGs (CT-22)

II 0

Limit Uncontrolled Radiation Release (CT-21)

Event TY Pe*

CRS N CRS R URO N ARO CRS M CRS M URO M ARO CRS C CRS C URO C ARO C CRS C URO C CRS C URO C CRS C URO C ARO C CRS C URO C CRS C ARO Event Description Reactor Buildina Hi Ranae Radiation Monitor Failure (TS)

Reactor Startup Intermediate Range Instrument NI-4 Fails Low (TS)

Turbine Bypass Valve MS-V3D Fails Open OTSG Tube Leak in OTSG A OTSG Tube Rupture in A OTSG requiring HPI initiation High Capacity Makeup Flow Valve MU-V-217 Breaker Trips Decay Heat Closed Cooling Water Pump Trips on ESAS Actuation OTSG Tube Rupture in A OTSG resulting in loss of subcooling margin (CT-1, CT-22)

Both Trains of HPI Fail to Actuate at 1600 psig (CT-2)

Emergency Feedwater Pump Steam Supply Valve MS-V-13A Fails Open. (CT-21)

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

Scenario Event Description NRC Scenario 2 Three Mile Island NRC Scenario #2 The crew will take the watch with reactor startup in progress pulling toward criticality. NI-3 is 00s due to a detector failure. The crew will continue the reactor startup.

As soon as the crew accepts the watch the lead Examiner can cue the failure of RB Hi Range monitor RM-G-22. MAP Alarm C-1-1, Radiation Level High will be received and the crew should diagnose the failure of the instrument. The CRS should review TS 3.5.5.2 and declare a 7 day timeclock based on the failure. As soon as the TS call is made the crew can be directed to continue the startup.

When the crew has progressed in the startup to satisfy the reactivity manipulation the Lead Examiner can cue the initiation of the Intermediate Range NI-4 detector failure. The CRS should terminate the startup and review TS 3.5.1.1 and Table 3.5-1 and declare a one hour timeclock based on zero Intermediate Range Instruments operable.

After the CRS has made the Tech Spec call the Lead Examiner can cue the initiation of the Turbine Bypass Valve MS-V-3D failure. This will cause an RCS Cooldown and cause the reactor to go critical if it was not already critical. The crew should trip the reactor and enter OP-TM-EOP-001, Reactor Trip at this point. The reactor must be tripped prior to going below 525°F.

After the plant has been stabilized the Lead Examiner can cue the initiation of the OTSG A tube leak. The crew should diagnose the tube leak at approximately 12 gpm and the CRS should review TS 3.1.6.3 and declare a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> timeclock to reach cold shutdown. The CRS should go to OP-TM-EOP-005, OTSG Tube Leakage to mitigate the event.

After the Tech Spec call has been made the Lead Examiner can cue the OTSG Tube Rupture.

The crew should diagnose the leak size increase and continue in EOP-005. The High Capacity Makeup Valve breaker will trip when it is opened and the URO will have to use Makeup Valve MU-V-16B to augment makeup flow. The leak will require initiation of HPI using OP-TM-EOP-010 Guide 9, RCS Inventory Control (CT-2). This is a critical task because the leak size is greater than normal makeup capability which would cause Pressurizer level to lower and eventually empty since HPI will not automatically actuate at 1600 psig.

Decay Heat Closed Cooling Water Pump DC-P-1B will trip when ESAS is actuated and the CRO will have to implement OP-TM-211-901, Emergency Injection (HPVLPI) Contingency Actions to transfer Makeup Pump MU-P-1 C cooling to Nuclear Services Closed Cooling Water (NSCCW). This will required securing MU-P-1 C.

After MU-P-IC cooling has been transferred to NSCCW the Lead Examiner can cue the increase. The leak size will result in a loss of subcooled margin. HPI will not automatically actuate at 1600 psig RCS pressure and will have to be manually actuated in accordance with OP-TM-642-901, 1600 PSlG ESAS Actuation. The CRS should initiate OP-TM-EOP-002, Loss of 25°F Subcooled Margin. All four Reactor Coolant Pumps will be tripped within one minute of the loss of subcooled margin (CT-I) resulting in Emergency Feedwater actuation. This is a critical task since failure to trip the RCPs could result in the core not being adequately covered and raise the potential for fuel clad failure.

The CRS will transition back to EOP-005 and when EF-P-1 is secured the crew should diagnose the Steam Supply Valve MS-V-13A failure to close and close it manually to stop the

O O O G 8 3 Scenario Event Description NRC Scenario 2 Three Mile Island NRC Scenario #2 Continued unmonitored release (CT-21). This is a critical task in that failure to isolate EF-P-1 would result in an uncontrolled radiation release from a non-essential load.

The A OTSG will subsequently be isolated in accordance with EOP-005 (CT-22). This OTSG is not required for cooldown and overfill can not be prevented by steaming, requiring the OTSG to be isolated reducing the radiological consequences of the event.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

OOOG103 Scenario Event Description NRC Scenario 2 Description Initial Set-up.

Reactor Building Hi Range Radiation Monitor Failure (TS)

Reactor Startup Intermediate Range Instrument NI-4 Fails Low (TS)

Turbine Bypass Valve MS-V-3D Fails Open OTSG Tube Leak in OTSG A OTSG Tube Rupture in A OTSG requiring HPI initiation High Capacity Makeup Flow Valve MU-V-217 Breaker Trips Decay Heat Closed Cooling Water Pump Trips on ESAS Actuation OTSG Tube Rupture in A OTSG Loss of Subcooled Margin Both Trains of HPI Fail to Actuate at 1600 psig Emergency Feedwater Pump Steam Supply Valve MS-V-13A Fails Open Procedure Support Reactor Startup in progress Intermediate Range NI-3 is 00s Tech Spec 3.5.5.2 Accident Monitoring Instrumentation OP 1 103-8, Approach to Criticality TS 3.5.1.1, Operational Safety Instrumentation Table 3.5-1 TM-EOP-001, Reactor Trip OP-TM-EOP-005, OTSG Tube Leakage

~

OP-TM-EOP-005, OTSG Tube Leakage OP-TM-EOP-010, Guide 9 RCS Inventory Control OP-TM-EOP-010, Guide 9 RCS Inventory Control OP-TM-211-901, Emergency Injection (HPIILPI)

OP-TM-543-440, Swapping MU-P-1 C Cooling to NS OP-TM-EOP-002, Loss of 25°F Subcooled Margin OP-TM-EOP-005, OTSG Tube Leakage OP-TM-211-901, Emergency Injection (HPVLPI)

OP-TM-642-901, 1600 PSlG ESAS Actuation OP-TM-EOP-005, OTSG Tube Leakage

Scenario Event Description NRC Scenario 2 Initialization IC-..

Remote Function RPS Cabinet C RPS Cabinet C Console Center Malfunction RM0322 NI-3 Malfunction NI 1 1 B Malfunction MSO9D Malfunction THI 5A Malfunction TH16A Malfunction MU09 Malfunction ESOIA Malfunction ESOI B Malfunction CC02B Malfunction THI 7A Malfunction FW45A Reactor Startup in Progress Value:

OUT When:

Immediately Turn NI-3 Detector Power Supply OFF Place EDT on Switch Turn NI-3 Aux Power Supply OFF Place EDT on Switch Place EDT on NI-3 indication Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Insert Event 1 Insert Event 3 Insert Sev. 100%

Event 4 Insert Sev. 0.05%

Event 5 RAMP 100 sec Insert 2.3%

Event 6 RAMP 300 sec Insert Immediately Insert I m med ia te I y Insert Immediately Insert Event 8 rrvdrvl b>0.1 Insert 20%

Event 9 RAMP 10 sec Insert Event 11 fwvmsvl3a=-0.9 Scenario Support Scenario Support Scenario Support Scenario Support RM-G-22 Fails High NI-4 Power Supply Fails OFF MS-V-3D Fails Open OTSG A Tube Leak OTSG A Tube Rupture High Capacity Makeup Valve Fails as is ESAS Failure to Actuate at HPI Setpoint (1600 PSIG)

Train A ESAS Failure to Actuate at HPI Setpoint (1600 PSIG)

Train B Decay Closed Cooling Water Pump 1B Trip OTSG A Tube Rupture MS-V-13A Fails as is (open)

o m 0 3 Appendix D Scenario Outline Form ES-D-1 Facility:

Three Mile Island Scenario No.:

3 OpTestNo.:

NRC Examiners:

Operators:

Critical Tasks:

0 Trip all RCPs (CT-1)

Event p 1

2 3

4 5

6 7

8 (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Initial Conditions:

85% power, MOC 0

Power Escalation Following a trip and maintenance outage MU-P-1B 00s Malf.

No.

Turnover:

Raise Power to 100%.

Event Type*

0 Establish FW Flow and Feed SGs (CT-10)

THO6 M CRS M URO M ARO I CCRS RD0216 CCRS I CURO T H 1 3 D ] NCRS I RURO I N URO C CRS C ARO C URO RWOIOB CCRS I CURO Event Description

~~

~~

Power Operated Reref Valve (PORV) Block Valve RC-V-2 Breaker Trips (TS)

Stuck Rod in Group 7 (TS)

Reactor Coolant Pump RC-P-1 D High Vibration Feedwater Flow Fails to re-ratio after securing RC-P-1 D An RCS leak occurs requiring a Reactor Trip Reactor Building Emergency Cooling Pump RR-P-1 B does not start on ESAS actuation RC-P-1A does not trip when control switch is rotated to the stop position (CT-1)

Emergency Feedwater Valves to the A OTSG EF-V30A and EF-V-30D do not control in automatic due to a level setpoint failure (CT-10)

Scenario Event Description NRC Scenario 3 Three Mile island NRC Scenario #3 The crew will take the watch with reactor power at 85% and ICs in Full Automatic. The plant is in a power escalation following an automatic trip several days ago and a subsequent maintenance period.

When the crew has accepted the watch the Lead Examiner can cue the initiation of the Report from the NLO that he was in the area of I C ES Valves MCC and heard a breaker trip. It was the Power Operated Relief Valve (PORV) Block Valve breaker. The CRS should review TS 3.1.12.4 for PORV Block Valve Operability and declare a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> timeclock is in effect to close the PORV.

The PORV Block Valve breaker will not be returned to service and the scenario can continue when the TS call is made.

After the TS call is made the crew will continue the power escalation. As rods are withdrawn a stuck rod occurs in Group 7 requiring the initiation of procedure OP-TM-AOP-062, Inoperable Rod. The CRS should review TS 4.7.1.2 and 3.5.2.2 and declare the rod inoperable. Tech Spec actions for the inoperable rod are included in AOP-062. Within one hour the other rods in the group must be trimmed so the inoperable rod remains within the group average. The rod will not be returned to operable status during the scenario.

After the Tech Spec call is made and AOP-062 has been implemented to the examiners satisfaction, the Lead Evaluator can cue the initiation of the RC-P-1 D high vibration malfunction.

The crew should diagnose increasing vibration on RC-P-1 D and initiate action to reduce power to ~ 7 5 %

to secure RC-P-1 D in accordance with Computer Alarm Response L3125, RC-P-1 D High Vibration.

When RC-P-1 D is secured feedwater flow will not re-ratio and the crew will have to take hand control of the Feedwater Loop masters in accordance with OP-TM-621-471, ICs Manual Control to manually re-ratio feedwater flow. The ICs does not have to be returned to auto to continue.

When the plant has been stabilized with feedwater flow re-ratioed, the Lead Examiner can cue initiation of the RCS leak. The crew should diagnose the leak based on RB pressure rise and RCS inventory change. Reactor Building pressure will rise rapidly and RCS pressure will lower.

The reactor should be tripped by the crew and the CRS should initiate OP-TM-EOP-001, Reactor Trip. Reactor Building Emergency Cooling Pump RR-P-1 B will not start automatically on the ESAS signal and must be started manually per OS-24, Conduct of Operations during Abnormal and Emergency Conditions.

When subcooling margin is lost RC-P-1A control switch will not trip the breaker and the URO will have to open the breakers for the 1A 7KV bus to trip the pump (CT-1). If the pump is not tripped within one minute the crew will have to keep RC-P-1A running in accordance with OP-TM-EOP-010, Rule 1 Loss of Subcooling Margin. This is a critical task since failure to trip the RCPs could result in the core not being adequately covered and raise the potential for fuel clad failure. The CRS will initiate OP-TM-EOP-002, Loss of Subcooled Margin.

When Emergency Feedwater is actuated EF-V-30A and EF-V-30D will not operate in automatic and will have to be taken to hand to feed the A OTSG (CT-10). This is a critical task in that failure to take manual control would result in the A OTSG going dry and becoming unavailable to maintain or initiate primary to secondary heat transfer.

Scenario Event Description NRC Scenario 3 Three Mile island NRC Scenario #3 Continued The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

806)003 Scenario Event Description NRC Scenario 3 Description Procedure Support Initial Set-up.

3P 11 02-2 Plant Startup Power Operated Relief Valve (PORV) Block Valve RC-V-2 Breaker Trips (TS).

TS 3.1.12.4, PORV Block Valve Operability Stuck Rod in Group 7 (TS)

TS 4.7.1.2, Rod Misalignment TS 3.5.2.2 Operation With Inoperable Rods OP-TM-AOP-062 Inoperable Rod Reactor Coolant Pump RC-P-I D High Vibration Computer Alarm L3125, RC-P-I D High Vibration OP 1 102-4, Power Operation Feedwater Flow Fails to re-ratio after securing RC-P-1 D OP-TM-MAP-H0204, Reactor Inlet ATc HI OP-TM-621-471, ICs Manual Control RCS Leak in the RB OP-TM-EOP-001, Reactor Trip OP-TM-EOP-010, Rule 1, Loss of Subcooling Margin OP-TM-EOP-002, Loss of Subcooled Margin OP-TM-EOP-006, LOCA Cooldown Reactor Building Emergency Cooling Pump RR-P-1B does not start on ESAS actuation OP-TM-534-901, RB Emergency Cooling Operations OS-24, Conduct of Operations During Abnormal and Emergency Events RC-P-1A does not trip when control switch is rotated to the stop position OS-24, Conduct of Operations During Abnormal and Emergency Events OP-TM-EOP-010, Rule 1, Loss of Subcooling Margin OP-TM-EOP-010, Guide 16, EFW Failure Emergency Feedwater Valves to the A OTSG EF-V-30A and EF-VSOD do not control in automatic

~

Scenario Event Description NRC Scenario 3 Initialization IC-I 6 85% Power, ICs Full AUTO Remote Function RCRl9 Malfunction RD0216 Malfunction THI 3D Monitor ICK314B (normal value 0.0101)

Malfunction THO6 Malfunction RW1 OB I/O Override 02A3S11-ZDlCSRCPlA(4) NAS 110 Override 02A3S11-ZDlCSRCPlA(2) STP I/O Override 02A3S11-ZDICSRCPIA(1) PTL Remote Function ICR02 Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Insert OUT Event 1 Insert Immediately Insert Sev. 50%

Event 2 RAMP 300 sec Insert 10.0 Immediately Insert Sev. 0.5%

Event 5 RAMP 100 sec Insert Immediately ON Immediately OFF I m med ia te I y OFF Immediately Insert Sev. 0%

Immediately Scenario Support RC-V-2 Breaker trips Stuck rod in Group 7 RC-P-1 D High Vibration FW Flow Fails to Re-ratio RCS Leak at RCP Discharge RR-P-1 B fails to start on ESAS signal RC-P-1A Breaker Fails to Trip RC-P-?A Breaker Fails to Trip RC-P-1A Breaker Fails to Trip OTSG Operate Level Setpoint for EFW Control A

I M Q 3 Appendix D Scenario Outline Form ES-D-1 Facility:

Three Mile Island Scenario No.:

4 OpTestNo.:

NRC Examiners:

Operators:

Initial Conditions:

100% power, MOC.

I 0

NS-P-1A 00s for Maintenance II 0

NS-P-1 B running on the 1 P 480V bus 11 Turnover:

Maintain 100% power operations.

Critical Tasks:

0 Control RCS Inventory (CT-30) 0 Establish and Maintain Reactor Shutdown Requirements (CT-23)

Isolate Overcooling SGs (CT-17)

Event No.

Malf No.

Event Type*

Event Description 1

ED03D C CRS C URO C ARO Aux Transformer B Fault Pressure (TS) 2 ES07A Inadvertent ESAS Actuation Train A C CRS C URO C CRS C URO N CRS R URO N ARO 3

MU23A Makeup Pump MU-P-1A Fails to Start on ESAS Actuation (TS)

FW35 Main Condenser Vacuum Leak 4

5 MS04A MS03A lCRl3 FWI 1A M CRS Main Steam Safety Valve fails to reseat. (MS-V-17A) and a steam leak in the Intermediate Building SG A Lo Press Isolation Setpoint is at zero psig Startup Feedwater Valve FW-V-16A Fails at 100% Open (CT-17 and CT-30)

M URO M ARO 6

RD02010 RD02056 C CRS C URO Two Control rods fail to fully insert (CT-23)

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

080G03 Scenario Event Description NRC Scenario 4 Three Mile Island Audit Scenario #4 The crew will take the watch with reactor power at 100% and ICs in Full Automatic. NS-P-1A is 00s for maintenance and will be out for two days. NS-P-1 B is running on the 1 P 480 Volt ES Bus and is selected for ES.

After the crew has accepted the watch and on cue from the Lead Examiner the Auxiliary Transformer B trip can be initiated. The I C 4160V bus will fast transfer to the A Auxiliary Transformer and the A Diesel Generator will start and load the 1 D 4160 V ES bus. The crew should respond in accordance with the electrical MAP alarm responses (B-1-1, B-1-5, AA-1-8),

OP-TM-AOP-013, Loss of 1 D 41 60V Bus, and EP 1203-20 Nuclear Services Closed Cooling Water Failure to restart NS-P-1 B on the A Diesel. The CRS should review TS 3.7.2.b and declare a 30 day timeclock due only one Auxiliary Transformer being operable and the diesel generator is already loaded on the bus.

After the plant has been stabilized and the Tech Spec call is made the Lead Evaluator can cue the initiation of the Inadvertent ESAS Actuation. The crew should diagnose the ESAS Train A actuation as inadvertent and use terminate HPI flow in accordance with OP 11 05-3, Safeguards Actuation System. The Crew should also diagnose the failure of Makeup Pump MU-P-1A to start and the CRS should review TS 3.3.1. I.b and TS 3.3.2 and declare a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> timeclock.

MU-P-1A will not be returned to service during this scenario. (Note: the CRS may also review the RCS Leakage Tech Spec 3.1.6.8 due to Reactor Building Radiation Monitor RM-A-2 being out of service)

The crew will have to insert control rods to compensate for the boron addition and stabilize reactor power. The crew will also have to re-establish normal letdown and makeup.

After the plant has been stabilized with rods inserted to compensate for the negative reactivity due to the boron injection and letdown and makeup have been restored the Lead Examiner can cue the initiation of the Main Condenser Vacuum Leak. The crew should diagnose the vacuum leak and insert rods to maintain power less than 100% as plant efficiency is lost. The CRS should initiate MAP Alarm Responses OP-TM-MAP-MO206, Aux Cond Hotwell Level Hi and OP-TM-MAP-NOIOG, Mn Cond VACUUM LO. The crew may elect to manually trip the reactor and turbine as vacuum degrades to the turbine trip setpoint. The CRS should initiate OP-TM-EOP-001, Reactor Trip. If the reactor is not tripped before, the turbine will trip if reactor power is reduced below 95%.

When the reactor trips a Main Steam Safety Valve will stick open on the A OTSG and a large steam leak will occur in the Intermediate Building from the A OTSG. The crew should subsequently diagnose the leak as being from the A OTSG and isolate the OTSG in accordance with OP-TM-EOP-010, Rule 3, Excessive Heat Transfer. In addition, the crew should diagnose the failure of the A OTSG Isolation to occur at 600 psig and FW-V-16A failed 100% open causing an overfeed. The CRS will transition to OP-TM-EOP-003, Excessive Primary to Secondary Heat Transfer. The ARO will have to close FW-V-92A to isolate Feedwater flow to the A OTSG due to FW-V-16A being failed open (CT-17). This is a critical task in that continued feeding of an OTSG with a steam break will continue to overcool the RCS, which could result in emptying the Pressurizer and causing a loss of subcooling margin. This would significantly change the mitigation strategy of the event.

When the OTSG is isolated and empty the URO will have to terminate HPI using Rule 2, HPVLPI THROTTLING CRITERIA and OP-TM-211-901, Attachment 7.3, THROTTLING HPI

8808303 Scenario Event Description NRC Scenario 4 Three Mile Island NRC Scenario #I4 Continued (CT-30). This is a critical task in that failure to throttle/terminate HPI flow will result in a rapid rise in Pressurizer level and pressure eventually challenging the PORV setpoint.

The crew should diagnose the failure of two control rods to fully insert and will have to Emergency Borate in accordance with OP-TM-EOP-01 0, Rule 5, Emergency Boration. The URO will have to initiate letdown if >50 gpm of injection is not achieved CT-23). This is a critical task in that adequate shutdown margin may not exist due to the two stuck rods and the RCS cooldown caused by the steam leak.

The Lead Evaluator can terminate the scenario when all high level activities have been completed and the evaluators agree the crew can be properly evaluated.

0006i03 Scenario Event Description NRC Scenario 4 Description Initial Set-up.

Auxiliary Transformer B Fault Pressure Trip Inadvertent ESAS Actuation Train A Makeup Pump MU-P-1A Fails to Start on ESAS Actuation Main Condenser Vacuum Leak Main Steam Safety Valve fails to reseat. (MS-V-17A) and Steam Leak in the Intermediate Building SG A Lo Press Isolation Setpoint is at zero psig Startup Feedwater Valve FW-V-16A Fails at 100% Open (CT-17 and CT-30)

Two Control rods fail to fully insert Procedure Support 100% Power MOC, NS-P-1A OOS, NS-P-1 B Running on the I P 480V Bus and selected for ES MAP Alarm B-1-5, 480 Volt ES Motor Trip OP-TM-AOP-013, LOSS Of 1 D 41 60V BUS TS 3.7.2.b OP 1 105-3, Safeguards Actuation System OP-TM-211-950, Restoration of Letdown Flow TS 3.3.1. I.b and TS 3.3.2, ECCS Equipment OP-TM-EOP-001, Reactor Trip OP-TM-MAP-M0206, Aux Cond Hotwell Level Hi OP-TM-MAP-NOIO6, Mn Cond VACUUM LO OP-TM-EOP-003, Excessive Primary to Secondary Heat Transfer OP-TM-EOP-010, Rule 3, Excessive Heat Transfer OP-TM-EOP-010, Rule 5, Emergency Boration

\\

b 000G03 Scenario Event Description NRC Scenario 4 Initialization IC-I 6 Console Center Remote Function CCRl9 Console Right Remote Function CCR21 Console Center Remote Function CCRl8 Main Console NS-P-1B NAS NS-P-1 B PTL NS-P-1A PTL Malfunction ED03D Malfunction ES07A Malfunction MU23A Malfunction FW35 Malfunction TCOl Malfunction MS04A Malfunction MS03A Malfunction RD0223 Malfunction RD0239 Malfunction FW1 ?A Remote Function lCRl3 100% HFP, ICs Full AUTO NS-P-1 B Running NS-P-1 B-1P Bkr CLOSED Value:

1P When:

Immediately NS-P-1 B-1 S PTL NS-P-1 B-I S Bkr OPEN When:

Immediately NS-P-1A Tagged 00s NS-P-1A Bkr OPEN Value:

OUT When:

Immediately Robust Barriers applied IAW Risk Value:

NS-PI B Document Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Value:

When:

Insert Event 1 Insert Event 2 Insert Immediately Severity 15%

RAMP 300 Sec Event 4 Insert Event 5 ratpw<95%

Insert Sev. 100%

Event 6 ratpwc5%

Insert Sev. 100%

Event 6 ratpw<5%

Insert Immediately Insert I m med ia te I y Insert Immediately Insert Immediately Scenario Support Scenario Support Scenario Support Scenario Support Scenario Support Scenario Support Main Console Auxiliary Transformer B Fault Pressure Trip Inadvertent ESAS Actuation (1 600 psig) Train A MU-P-1A ES Start Failure Vacuum Leak Turbine Trip Main Steam Safety Valve Leaks/Fails to Reseat (MS-V-17A)

Main Steam Leak Outside the RB Stuck Rod Stuck Rod Startup Feedwater Valve Fails as is (FW-V-16A)

SG A Lo Press Isolation setpoint

000G03 ES-301 Transient and Event Checklist Form ES-301-5 Facility:

A P

P L

I C

A N

T SR016 R02 SROll TM I Date of Exam:

April 2007 ODeratina Test No.:

NRC E

v N

T T

Y P

E 1

2 CREW CREW POSITION POSITION Scenarios POSITION POSITION T

M 0

I T

N A

I L

M U

M(*)

R I

U 1

1 1

0 1

1 1

1 2

Instructions:

1.

Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must service in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; instant SROs must do one scenario, including at least two instrument or component (IC) malfunctions and one major transient, in the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

SRO Instant and Upgrade candidates TOTALS reflect the ATC and SRO positions only.

NUREG 1021 Revision 9

080003 ES-301 Transient and Event Checklist Form ES-301-5 Facility:

TM I Date of Exam:

ADril2007 ODeratina Test No.:

NRC SR015 RO 1 SROU E

v E

N T

T Y

P E

RX NOR IIC MAJ TS Rx NOR IIC MAJ TS RX NOR IIC MAJ TS -

Scenarios 1

2 3

4 T

M 0

I T

N POSITION POSITION POSITION POSITION A

I L

M U

CREW CREW CREW CREW Instructions:

1.

Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must service in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must do one scenario, including at least two instrument or component (I/C) malfunctions and one major transient, in the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-? basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

SRO Instant and Upgrade candidates TOTALS reflect the ATC and SRO positions only.

NUREG 1021 Revision 9

ES-301 Transient and Event Checklist Form ES-301-5

=acility:

TM I Date of Exam:

April 2007 Operating Test No.:

NRC SR017 R03 SR012 Rx NOR -

I/C MAJ TS RX NOR MAJ TS RX NOR I/C MAJ Scenarios Instructions:

1.

Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must service in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must do one scenario, including at least two instrument or component ( K ) malfunctions and one major transient, in the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

SRO Instant and Upgrade candidates TOTALs reflect the ATC and SRO positions only.

NUREG 1021 Revision 9

WbUUJ ES-301 Transient and Event Checklist Form ES-301-5 TMI Date of Exam:

April 2007 Operating Test No.:

NRC

-acility:

A P

P L

I C

A N

T P

SR018 E

v E

N T

T Y

P E -

RX Scenarios POSITION POSITION POSITION POSITION S

R 0

NOR I/C -

1 10 2

MAJ TS RX 2

1 SR014 SR013 NOR 3

1 I/C 6

MAJ 5

2 TS RX -

NOR I/C 9

MAJ TS -

2 2 -

Instructions:

1.

Circle the applicant level and enter the operating test number and Form ES-D-1 event numbers for each event type; TS are not applicable for RO applicants. ROs must service in both the at-the-controls (ATC) and balance-of-plant (BOP) positions; Instant SROs must do one scenario, including at least two instrument or component (VC) malfunctions and one major transient, in the ATC position.

2.

Reactivity manipulations may be conducted under normal or controlled abnormal conditions (refer to Section D.5.d) but must be significant per Section C.2.a of Appendix D. (*) Reactivity and normal evolutions may be replaced with additional instrument or component malfunctions on a 1 -for-I basis.

3.

Whenever practical, both instrument and component malfunctions should be included; only those that require verifiable actions that provide insight to the applicants competence count toward the minimum requirements specified for the applicants license level in the right-hand columns.

4.

SRO Instant and Upgrade candidates TOTALs reflect the ATC and SRO positions only.

NUREG 1021 Revision 9