TMI-11-034, Draft - Outlines (Folder 2)

Draft - Outlines (Folder 2)

ML112620088
Person / Time
Site:	Crane
Issue date:	07/01/2011
From:	Chick G Exelon Nuclear
To:	Caruso J Operations Branch I
JACKSON D RGN-I/DRS/OB/610-337-5306
Shared Package
ML110030671	List: IR 05000289/2011301 ML110490206 ML112311224 ML112590501 ML112590507 ML112620074 ML112620096 ML112620102 ML112620105 ML112620125 ML113350223 TMI-11-034, Draft - Outlines (Folder 2)
References
TAC U01829, TMI-11-034
Download: ML112620088 (48)

Text

Exelon Nuclear Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 July 1,2011 TMI-11-034 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit I Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Submittal of Integrated Initial License Training Examination Materials Enclosed are the examination materials, which TMI Unit 1 is submitting in support of the Initial License Examination scheduled for the week of August 29,2011, at TMI Unit 1.

This submittal includes the Reactor Operator Written Examinations, Job Performance Measures, and Integrated Plant Operation Scenario Guides. This submittal also includes the Senior Reactor Operator Job Performance Measures, and Integrated Plant Operation Scenario Guides.

These examination materials have been developed in accordance with NUREG-1021, Revision 9, Supplement 1 "Operator Licensing Examination Standards". Please note that reference materials are attached to each individual examination question or item.

Some minor modifications have been made to the Integrated Examination Outline with regards to the operational scenarios in order to improve balance and content. These changes improve examination quality and are in compliance with NUREG-1021, Revision 9, Supplement 1, "Operator Licensing Examination Standards."

Some modifications or adjustments to the examination material may be required due to procedural changes.

In accordance with NUREG 1021, Revision 9, Supplement 1, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning this letter, please contact Mike Fitzwater at 717-948-8228. For questions concerning examination materials, please contact Greg Hoek at 717-948-2027.

Respectfully

'":7 I ~

'UAA-J,,

Glen Earl Chick

Site Vice President, Three Mile Island Unit I GEC/mdf

Enclosures:

(Fed Ex to John Caruso, Chief Examiner, NRC Region I)

RO/SRO Composite Examination with references attached

Control Room Systems and Facility Walk-Through Job Performance Measures with references attached Administrative Topic Job Performance Measures with references attached Integrated Plant Operation Scenario Guides Completed Checklists:

Operating Test Quality Checklist (Form ES-301-3)

Simulator Scenario Quality Checklist (Form ES-301-4)

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

Competencies Checklist (Form ES-301-6)

Written Exam Quality Checklist (Form ES-401-6)

Examination Security Agreements (Form ES-201-3)

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

(without attachments)

Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector - TMI-1 Operations Training Manager

bcc:

(without attachments)

TMI Unit 1 Project Manager, NRR Site Vice President - TMI Unit 1 Regulatory Assurance Manager - TMI Unit 1

ES-402 Written Examination Outline Form ES-401-2 Facility: TMI Date of Exam:

08/29/11 SRO-Only Points Tier A2 G*

T

1.

Emergency Plant Evaluations 3

2 5

3 2

5 4

10 2

3 5

2.

Plant Systems 2

Tier Totals 3

3 1

4 1

4 4

0 2

1 4

1 3

4 3

1 4

10 38 0

2 4

1 4

3. Generic Knowledge &Abilities 1

2 3

4 10 1

2 3

4 7

2 2

3 3

2 2

1 2

Note

1.

Ensure that at least two topics from every applicable KiA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the Tier Totals in each KiA category shall not be less than two).

2.

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

The final point total for each group and tier may deviate by 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 that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KiA statements.

4.

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

5.

Absent a plant specific priority, only those !<As 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.

6.

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

7.*

The generic (G) KiAs in Tiers 1 and 2 shall be selected from Section 2 of the KiA 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 KiA's

8.

On the following pages, enter the KiA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note H1 does not apply). Use duplicate pages for RO and SRO-only exams.

9.

For Tier 3, select topics from Section 2 of the KJA Catalog, and enter the KiA numbers, descriptions, and totals (H) on Form ES-401-3. Limit SRO selections to KiAs that are linked to

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function_IiiI K21 077 ! Generator Voltage and Electric Grid Disturbances E02 ! Reactor Trip - Stabilization IR""VB", 11

• 055 / Station Blackout i 6 038/ Steam Generator Tube Rupture

/3 015/17 i Reactor Coolant Pump Malfunctions / 4 056 I Loss of Off-site Power I 6 E10 I Reactor Trip - Stabilization x

Recovery 11 008 I Pressurizer Vapor Space x

Accident/3 i 054/ Loss of Main Feedwater I 4 025 I Loss of Residual Heat Removal x

System /4 E05 I Steam Una Rupture - Excessive X

i Heat Transfer /4 077 / Generator Voltage and Electric Grid Disturbances I G KIA Topic(s)

Imp.

K31 A11 A2

--'.i_____,______,__-'-_

1 Q# 1 AA2.01 - Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid I 3.6 76 Disturbances: Operating point on the EA2.1 - Ability to determine and interpret the following as they apply to the (Vital System Status Verification):

4.0 77 Facility conditions and selection of procedures during 3.7 78 i

4.6 79 4.2 80 4.3 81 4.0 39 3.2 40 4.1 41 2,7 42 4.2

,43 44

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1

,--_E_A_P_E_#/_N_am_e_sa_fe~ty_Fu_n_ct_io_n_-LI_K_1-,-1_K2----LIK3 I A 1 I A2 I G I

KIA Topic(s)

Imp.

062 I Loss of Nuclear Service. Water /

4 x

029 1 Anticipated Transient Without Scram (ATWS) 11

058 / Loss of DC Power / 6 009 I Sma" Break lOCA / 3 055 / Station Blackout /6 057 I Loss of Vital AC Electrical Instrument Bus 16 027 I Pressurizer Pressure Control System Malfunction 13 3.4 4.3 2.8 E04/1nadequate Heat Transfer Loss of Secondary Heat Sink / 4 3.2 011 I large Break LOCA /3 4.2 065/ Loss of Instrument Air I 8 038/ Steam Generator Tube Rupture 13 4.2 4.1 056/ Loss of Off-site Power /6 KIA CategoryTotals 3

3 Group Point Total:

I Q#.

46 52 53 54 55 56 18/6

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KIA Topic(s) 005/ Inoperable/Stuck Control Rod I 4.1 82 1

A08 ! Fuel Handling Accident I 8 4.0 037 I Steam Generator Tube Leak 13 4.0 AD6/ Control Room Evac. / 8 4.0 85

.2 implications of the following concepts as they apply to the (Natural E09 / Natural Cire. I 4 X

Circulation Cooldown) Normal, 3.7 57 abnormal and emergency operating procedures associated with (Natural Circulation AK2.1 Knowledge of between the (Shutdown Outside Control Room) and the following:

Components, and functions of control A06 I Control Room Evac. / 8 X

3.8 58 and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

AK3.02 Knowledge of the reasons for the following responses as they apply 067 / Plant Fire On-site I 8 X

to the Plant Fire on Site Steps called 2.5 59 out in the site fire protection plan, FPS and fire zone manual AA 1.22 - Ability to operate and I or monitor the following as they apply to 024 I Emergency Boration 11 the Emergency Boration: Safety 3.2 injection valves, switches, flow meters, E14/ EOP Rules and Enclosures 15 4.0 61 A041 Turbine Trip 14 3.8 62 059 I Accidental Uquid RadWaste 3.6 63 Release / 9 032 I Loss of Source Range Nuclear 3.1 64 Instrumentation 17

ES-402 Form ES-401-2 TMI Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KIA Topic(s)

A03 / Loss NNI-X/Y / 7 x

KIA CategoryTotals 2

System #/Name 026 Containment Spray 061 Auxiliary/Emergency Feedwater 006 Emergency Core Cooling 064 Emergency Diesel Generator 003 Reactor Coolant Pump 003 Reactor Coolant Pump x 061 Auxiliary/Emergency x

Feedwater 005 Residual Heat Removal 004 Chemical and Volume x

Control 007 Pressurizer x

Relief/Quench Tank 026 Containment Spray x

012 Reactor Protection ES-402 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s) predict the impacts of the following malfunctions or operations on the ess; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Reflux boiling pressure spike when first A2.08 - Ability to (a) predict the impacts of the following malfunctions or operations on the APN; and (b) based on those predictions. use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Flow rates expected from various combinations of APN pump tii",,~h"'m" valves 2.2.37 - Equipment to determine operability and / or availability of safety related Imp.' Q# I feature(s) and/or interlock(s) which provide for the following:

Automatic or manual ofRPS 3.0 2.9 4.6 2.8 4.2 3.0 2.9 3.3 3.9 3.2 86 87 88 89 90 3

4 5

6 7

ES-402 FDrm ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 K4.04 - Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the 039 Main and Reheat following: Utilization of steam X

Steam pressure program control when steam dumping through relief/dump valves, limits 013 Engineered Safety X

Features Actuation 006 Emergency Core X

Cooling 064 Emergency Diesel X

Generator 010 Pressurizer Pressure X

Control 076 Service Water 059 Main Feedwater 062 AC Electrical Distribution 008 Component Cooling Water 2.9 2.9 3.2 2.6 2.5 2.8 2.5 8

9 10 11 12 13 14 15 16

ES-402 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 1 System #/Name KIA Topic(s}

063 DC Electrical Distribution 078 Instrument Air 073 Process Radiation Monitoring 1 03 Containment 022 Containment Cooling 062 AC Electrical Distribution 004 Chemical and Volume Control 059 Main Feedwater 003 Reactor Coolant Pump 005 Residual Heat Removal 039 Main and Reheat Steam 063 DC Electrical Distribution KIA Category Totals x

x 2

2 3

2.4.2 - Emergency Procedures Plan: Knowledge of system set points, interlocks and automatic actions associated wlth EOP entry conditions.

2.4.47 - Emergency Procedures /

Plan: Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room 2.4.46 - Emergency Procedures 1 Plan: Ability to verify that the alarms are consistent with the conditions.

K4.01 - Knowledge of de electrical system design feature(s} and/or interlock(s} which provide for the following: Manua1!automatie transfem of control Group Point Total:

2.7 3.7 4.5 4.2 3.1 3.5 3.2 2.5 4.2 2.7 17 18 19 21 22 23 24 25 26 27 28 28/5

ES-401 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name A2 G

KIA Topic(s)

I K11 K21 K31 K41 K51 K61 A1 I IA31A41 A2.02 - Ability to (a) predict the impacts of the following malfunctions or operations on the Waste Gas Disposal System; and (b) based on those predictions, 071 Waste Gas Disposal use procedures to correct, control.

or mitigate the consequences of those malfunctions or operations:

Use of waste gas release monitors, radiation, gas flow rate, and totalizer 2.2.12 - Equipment Control:

086 Fire Protection Knowledge of surveillance res.

A2.03 - Ability to (a) predict the impacts of the following mal functions or operations on the Containment Purge System; and (b) based on those predictions, 029 Containment Purge use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Startup operations and the associated uired valve lineu K5.01 - Knowledge of the operational implications of the following concepts as they apply 014 Rod Position Indication X

to the RPIS: Reasons for differences between RPIS and counter A 1.05 - Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the 045 Main Turbine Generator X

MT/G system controls including:

Expected response of primary plant parameters (temperature and to m 072 Area Radiation automatic operation of the ARM Monitoring system, including: Changes in ventilation K3.01 - Knowledge of the that a loss or malfunction of the 017 In-core Temperature X

ITM system will have on the Monitor following: Natural circulation 027 Containment Iodine X

Removal 011 Pressurizer Level Control 035 Steam Generator 029 Containment Purge X

Imp.

3.6 3.4 3.1 2.7 3.B 2.9 3.5 3.4 3.8 3.8 3.2 I Q# I 91 92 93 29 30 31 32 33 34 35 36

ES-401 Form ES-401-2 TMI Written Examination Outline Plant Systems - Tier 2 Group 2 System #/Name

~______

-L~~~_L~__~-L__L 001 Control Rod Drive x

075 Circulating Water KJA Category Totals KJA Topic(s)

A2.0S - Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or 2.5 38 mitigate the consequences of those maHunctions or operations:

Safety features and relationship between condenser vacuum,

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

Form ES-401-3 Facility:

TMI Date:

(;:,...

y KA#

Topic RO SRO-Only IR Q#

IR Q#

i~bi.lity to locate control room switches, controls, and 2.1.31 1".dh.....,"VIli>,

to determine that they correctly reflect 4.6 66 Ithe desired plant lineup.

iAbility to use procedures to determine the effects on 2.1.43

,wau"v'ity of plant changes, such as RCS temperature, 4.1 67 I ""'VII"", ' plant, fuel U"'I.I,,,,tiVII, etc.

1. Conduct of Operations 2.1.35 I r\\fI

-<r ISRO's.

of the fuel-handling responsibilities of 3.9 94 Ability to use procedures related to Shlft_S~~~~~~such as minimum crew complement, overtime ::;."'u,,.~,

3.9 98 letc.

iSubtotal 2

r:~.:*i,~KR::

2 Ability to manipulate the console controls as required 2.2.2 to operate the facility between shutdown and 4.6 68 Idesignated power levels.

2.2.14 I Kn"UI,lorino of the I configu;crtlon or status.

for controlling equipment 3.9 69

2. Equipment Control 2.2.5 2.2.23

~~erati~g~,~:,:~::;~~~!~~~rli~aking design or

~ I ecnrJIcal SpeCification limiting v 1.1":'''''''

'¥ 4.6 Subtotal 2

)~~ifu.\\'I(k\\'

!/'S)"14't~~

2 2.3.7 lAbility to comply with radiation per':flit IreQuirQ'""'til'lnlOl effects.

4.2 72 3.2 73 4.Emergency!------t~~~~~~~~~~~~--~--_1----1_--_i-----!

Procedures I Plan Knowledge of low power I shutdown implications in 2.4.9 accident (e.g., loss of coolant accident or loss of residual heat rArr.m".1l

!Tier 3 Point Total:

ES-401 Record of Rejected K/A's Form ES-401-4 Randomly Selected Tier / Group KA Reason for Rejection 039/ K4.02 replaced The Facility does NOT have Tave control(MS and reheat is not

• 2/1 by 039 / K4.04 an input to Tave control) i 028/ K2.01 replaced The Hydrogen Recombiners have been re-classified as having 2/2 I

bv 001 I K2.01 no safety related im~act at TMI.

1028/ A2.03 replaced The Hydrogen Recombiners have been re-classified as having 2/2 by 029/ A2.03 no safety: related impact at TMI.

001 / K2.01 replaced 2/2 The subject KIA isn't relevant at the subject facility.

by 001 / K2.02 064 I 2.1.19 replaced It isn't possible to prepare a psychometrically sound question 2/1 i

bv 064 / 2.1.32 related to the subject KIA.

071 / A2.03 replaced It isn't possible to prepare a psychometrically sound question 2/2 by 071 / A2.02 related to the subject KIA.

010 / K6.04 replaced RCDT operations were oversampled on the RO Exam. Question.

2/1 bv 010 I K6.03 5 and 40 also focus on RCDT operations.

i 007 / K5.02 replaced 2/1 KAs involving RCDT Operations have been over-sampled.

by 004 I K5.26 I

063/ K2.01 replaced NRC Examiner Feedback indicating that this KA would not yield 2/1 bv 063 / K4.01 a discriminatinQ question, and the a~~ro~riate level of difficulty:.

034/ A3.01 replaced NRC Examiner requires re-select based on Refueling activities 2/2 by 072 / A3.01 beina performed bv contracted personnel rather than ROs.

NRC Examiner Feedback indicating that this KA may not allow G1 /2.1.45 replaced 3/1 the construction of a written exam question at the level required by G1 /2.1.35 by 10CFR55.43(b)

NRC Examiner Feedback indicating that this KA may not allow G1 /2.1.25 replaced 3/1 the construction of a written exam question at the level required by G1/2.1.5 bv 10CFR55.43(b) 062/ AK3.04 1 /1 replaced by 062 I Effect of a Loss of CCW on NR Flow is non existent at TMI.

AK3.02 056/2.4.46 replaced NRC Examiner Feedback indicated a potential to over sample 1 /1 by 056/2.2.44 between written and operational evaluated; and replaced KIA.

G4/2.4.40 replaced NRC Examiner feedback indicated this may not yield a greater 3/4 by G4 / 2.4.43 than LOD1 Question if asked at the RO level.

079/2.2.12 replaced NRC Examiner Feedback discussion agreed that Station Air at 2/2 by 086 / 2.2.12 TMI is very low safety significance re-picked system.

i

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: August 2011 Examination Level: ROD SRO ~

Operating Test Number: 289-2011-301 Administrative Topic I

Type Describe activity to be performed (See Note)

Code*

Maintain Minimum Shift Staffing, Control Overtime Conduct of Operations N/R 2.1.5 (3.9)

Review an Estimated Critical Rod Position Calculation for Conduct of Operations N/R Iapproval, identify any errors.

2.1.37 (4.6)

Approve Isolation Points for a Component for Maintenance Equipment Control N/R 2.2.41 (3.9)

Authorize emergency exposure in excess of 5 REM.

Radiation Control N/R 2.3.4 (3.7)

Identify and Declare an Emergency Classification and Associated PAR.

Emergency Procedures/Plan N/R 2.4.44 (4.4)

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

• Type Codes &Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (.::;. 3 for RDs;.::;. 4 for SRDs &RD retakes)

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

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

ES 301, Page 22 of 27

ES-301 Administrative Topics Outline Form ES-301-1 THREE MILE ISLAND 2011 NRC SRO EXAMINATION CONDUCT OF OPERATIONS (A1-1): Verify watch standing requirements - Work-hour Rules.

Given plant conditions and references OP-TM-101-111-1001, Shift Manning Requirements, LS-AA-119, qvef!:ime 90ntrol~, a prepared Shift Staffi!l9 Report, L.MS Qual Matrix Report, a!1d a prepared Over time list, Identify required actions to restore minimum staffing, and select personnel In accordance with requirements to control overtime. This JPM is significantly modified in that it incorporates Work-Hour rules lAW LS-AA-119, changing the requirements of whether personnel can report to work or not.

License is evaluated against properly manning a shift lAW 10CFR and LS-AA-119. Several opportunities for error exist in the fact that they must review 5 different documents to determine if personnef can stand watch.

Safety Significance, failure to select the correct personnel will lead to Work-hour rule violations and possible fatigue issues. Failure to properly staff a Control Room crew is also a violation of 10CFR.

CONDUCT OF OPERATIONS (A1-2): Review an Estimated Critical Rod Position Calculation for approval, identify any errors.

Given an Estimated Critical Rod Position previously calculated and references OP-TM-300-000, "Reactivity and power distribution calculations" and OP-TM-300-403, "Estimated Critical Rod Positions" review for SRO approval and identify errors if any.

This is a New ~IPM updated for the current core load and also using a new series of procedures. The old Bank JPM used one procedure with all attached graphs, the new operations procedures uses a calculation procedure and a reference procedure and additionally the numbers have changed to cause the new answer to be outside the range of the old answer.

License is evaluated against a job done by senior reactor operators with measurements of using correct graphs and doing matn properly. Several 0RPortunities for error exist in similar named graphs and applying curve data in the wrong direction. SRO should recognize that errors were made, Identify errors and return for recalculation.

Safety significance, failure to calculate an estimated critical rod position properly could lead to reactor being taken critical at an unexpected rod height.

EQUIPMENT CONTROL (A2): Approve Isolation Points for a Component for Maintenance.

Given a Worker Tag Out Clearance Form, review for approval. A new JPM created for ILT 10-1. The submitted Tag Out Form omits required pOints and leaves a group of breakers in the Off pOSition on restoration.

Operations Authorization would come from an SRO. Safety significance is two valves do not have their remote controls indicated with info tags nor their breakers open this could result in operation that would drain the fuel handling canal if operated. The restoration points are also wrong and could result in not being able to appropriately respond to an excessive seal feak if the breakers were not closed.

RADIATION CONTROL (A3): Authorize emergency exposure in excess of 5 REM.

Conditions are given for a non-contaminated injured individual in a high radiation area, rescue personnel must go through a highly contaminated area. Six volunteers to carry stretcher, candidate must recognize one individual can not volunteer as they have alreaay received a once in a life time exposure, a second volunteer is not respirator qualified.

Safety significance failure to properly remove these volunteers could result in excessive exposure.

EMERGENCY PROCEDURES/PLAN (A4): Identify and Declare an Emergency Classification Given a set of plant conditions declare an Emergency Classification. This new JPM deals with declaring an emergency classification (Time Critical) based on given plant conditions, and declaring the associated PAR also time critical.

License level is SRO as the Emergency Director is the position to declare an event.

Safety significance events must be recognized and declared in a timely fashion to get the required notifications and support, declaration of the PAR in a timely manor allows for the Slate to make appropriate recommendations for public protection.

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Three Mile Island Date of Examination: August 2011 Examination Level: RO L8l SRO 0 Operating Test Number: 289*2011-301 Administrative Topic (See Note)

Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Type Code*

Describe activity to be performed N/R Verify watch standing requirements - Work-hour Rules 2.1.5 (2.9)

Calculate an Estimated Critical Rod Position.

N/R

. 2.1.37 (4.3)

N/R Isolate a component for maintenance.

2.2.41 (3.5)

N/A Category not selected for RO applicants Perform State and Local Event Notification D/S 2.4.43 (3.2)

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

• Type Codes &Criteria:

(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (.$. 3 for RDs;.:$.4 for SROs & RO retakes)

(N)ew or (M)odified from bank (?: 1)

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

ES 301, Page 22 of 27

ES-301 Administrative Outline Form ES-301-1 THREE MILE ISLAND 2011 NRC RO EXAMINATION CONDUCT OF OPERArlONS (A1-1): Verify watch standing requirements - Work-hour rules.

Given a shift schedule for the cycle and reference LS-AA-119, Overtime Controls, identify which requested days the candidate can work while not violating Work-hour rules. This JPM is a new JPM.

License is evaluated against properly identifying the allowable and forbidden overtime shifts lAW LS-AA 119.

Safety significance, failure to select the correct shifts will lead to Work-hour rule violations and possible fatigue issues. Failure to properly staff a Control Room crew is also a violation of 10CFR.

CONDUCT OF OPERATIONS (Al-2): Calculate an Estimated Critical Position.

Given plant conditions and references OP-TM-300-000/ "Reactivity and power distribution calculations" and OP-TM-300-403, "Estimated Critical Rod Positions' calculate an estimated critical rod condition.

This is a New JPM updated for the current core load and also using a new series of procedures. The old JPM used one procedure with all attached graphs, the new operations procedures uses a calculation procedure and a reference procedure additionally the Calculated point IS out side the tolerance band for the old calculation.

License is evaluated against a job done by reactor operators with measurements of using correct graphs and doing math properTy. Several opportunities for error exist in similar named graphs and applying curve data in the wrong directIOn.

Safety significance, failure to calculate an estimated critical rod position properly could lead to reactor being taken critical at an unexpected rod height.

EQUIPMENT CONTROL (A2): Isolate a component for maintenance.

This is a new JPM. The candidate must determine the isolation pOints for removal of a relief valve to repair a flange leak.

Safety significance This failure to adequately isolate the leak prior to removing the relief or opening vent and I or drain prior to isolating the relief would result in RCS water leaking into containment and a reduction in shielding water above the core.

RADIATION CONTROL (A3): Category not selected for RO Candidates.

EMERGENCY PROCEDURES/PLAN (A4): ERO Notification with a faulted sheet. This JPM is from the. LOR.~ ban.!<. RO Ca.ndidat~ will be given a $t?~e,and Local Notification FOfl!l..-.:ith Errors that need to be Identified. rhe Candidate Will then simulate initiating a State and Local notificatIOn on the NARS line, This is a time critical JPM.

Failure to make timely and accurate notification to State and Local emergency management agencies co~ld lead to. a degraaation of safety of public if agencies can not plan and execute tneir responsibilities in a timely fashion.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: August 2011 Exam Level: RO ~ SRO-I D SRO-U D Operating Test Number: 289-2011-301

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

Safety System / JPM Title Type Code*

Function

a. ATWS with Faulted EOP-001 IMA's (007 EA2.04)

N/S/A 1

b. Respond to Pressurizer Level Controller Failure (011 A2.03)

N/S/A 2

c. Transfer to Reactor Building Sump Recirculation (011 EA1.05)

D/S/UNEN 3

d. Restore TBVs and ADVs to ICS Auto Control (041 A4.06)

D/S/L 4S

e. RCP #1 Seal Failure (003 A2.01)

N/S/A 4P

f. Restore 1 D BUS from SBO Operations (064 A2.09)

N/S 6

g. Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure (072 M/S/A 7

A3.01)

h. Cross Connect the Secondary River Water System to the Nuclear River M/SINL 8

Water System (026 AA2.02)

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

i. Manually Operate MU-V-20 and IC-V-4 (015/017 M1.07)

D/E/R 4P

j. Local Start of EG-Y-1 B and Loading of 1 E 4160V bus (068 M 1.10)

N/E 8

k. Respond to a failure of EF-P-2A to start, and EF-V-30D to close (061 M/E 4S A2.04)

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

• Type Codes Criteria for RO I SRO-I / SRO-U (A)lternate path 4-6/ 4-6 / 2-3 (C)ontrol room

{D)irect from bank

~9/ ~8 1 ~4 (E)mergency or abnormal in-plant 2: 1 /

2:1

/ 2: 1 (EN)gineered safety feature

- /

/ 2: 1 (control room system (L)ow-Power 1Shutdown 2: 1 1 2:1 1 2: 1 (N)ew or (M)odified from bank including 1 (A) 2: 2 1 2:2 1 2: 1 (P)revious 2 exams

~31 ~3 / ~ 2 (randomly selected)

(R)CA 2: 1 /

2: "\\ 1 2: 1 (S)imulator I

ES-301, Page 23 of 27

ES-301 Control Room/ln*Plant Systems Outline Form ES-301-2 THREE MILE ISLAND 2011 NRC RO EXAMINATION JPM A - ATWS with Faulted EOP-001 IMA's. New Alternate path.

Safety significance failure to follow the alternate path will result in the reactor not being shutdown when it shoula be.

JPM B - Respond to Pressurizer Level Controller Failure. New Alternate path.

Safety significance failure to complete JPM will result in a loss of RCS inventory.

,IPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA.

JPM D - Restore TBVs and ADVs to ICS Auto Control. Bank JPM.

Safety significance failure to complete the task would result in unnecessary steaming to atmosphere resultmg m a loss of secondary inventory used for heat sink.

JPM E - RCP #1 Seal Failure. New Alternate path ~IPM.

Safety significance failure to properly address excessive seal leakoff could result in Seal LOCA.

JPM F - Restore 1 D BUS from SBO Operations. New Involves restoring the "A" train ES Bus from Station Blackout Diesel power to normal Off-Site power.

Safety significance is improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.

JPM G Respond lAW OP-TM-MAP-C01 01 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety Significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H - Cross Connect the Secondary River Water System to the Nuclear River Water System. Modified JPM, significantly updated to current procedures, change from 2007 NRC is the alternate path.

Safety significance failure to complete the task would result in loss of adequate cooling for Nuclear Services components.

JPM I - Manually Operate MU-V-20 and IC-V-4. Bank JPM In Plant Engineered Safety Features JPM.

Safety Significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J Local Start of EG-Y -1 B and Loading of 1 E 4160V bus. New JPM supplies power to the protected busses on Control Room Evacuation. Safety significance Remote shutdown capability only exists if "B" train power is available.

JPM K - Respond to a failure of EF-P-2A to start. and EF-V-30D to close. Modified Bank JPM. old JPM involved steam bound pumR. when restarted had to be done from switchgear. Switchgear start is no longer in procedure, added failed open discharge control valve on start.

Safety significance faiIMre to complete 1 $I part of task could lead to inadequate secondary feed for loss of steam driven pump. 2 part failure could fead to excessive heat transfer.

Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: August 2011 Exam Level: RO 0 SRO-I I2J SRO-U 0 Operating Test Number: 289-2011-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

I Safety i

System I JPM Title

• Type Code*

Function Ia. ATWS with Faulted EOP-001 IMA's (007 EA2.04)

N/S/A 1

i

b. Respond to Pressurizer Level Controller Failure (011 A2.03)

N/S/A 2

I c. Transfer to Reactor Building Sump Recirculation (011 EA 1.05)

D/S/L/NEN 3

e. RCP #1 Seal Failure (003 A2.01)

N/S/A 4P

f. Restore 10 BUS from SBO Operations (064 A2.09)

N/S 6

g. Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure (072 M/S/A 7

A3.0i)

h. Cross Connect the Secondary River Water System to the Nuclear River M/S/NL 8

Water System (026 AA2.02)

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

i. Manually Operate MU-V-20 and IC-V-4 (015/017 AA1.07)

O/E/R 4P

j. Local Start of EG-Y-1 B and Loading of 1 E 4160V bus (068 AA1.10)

N/E 8

I

k. Respond to a failure of EF-P-2A to start, and EF-V-300 to close (061 M/E 4S A2.04)

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

• Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6/ 4-6 I 2-3 (C)ontrol room (O)irect from bank

$.9/ $.8 I$.4 (E)mergency or abnormal in-plant

.2;: 1 /

.2;:1

/.2;: 1 (EN)gineered safety feature

- /

/.2;: 1 (control room system (L)ow-Power / Shutdown

.2;: 1 /

.2;:1

/.2;: 1 (N)ew or (M)odified from bank including 1 (A)

.2;:2/

.2;:2 /.2;: 1 (P)revious 2 exams

$. 3/ $.3 / $. 2 (randomly selected)

(R)CA

.2;: 1 /

.2;:1

/ > 1 (S)imulator I

ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 THREE MILE ISLAND 2011 NRC SROI EXAMINATION JPM A ATWS with Faulted EOP-001 IMA's. New Alternate path.

Safety significance failure to follow the alternate path will result in the reactor not being shutdown when it shoula be.

JPM B - Respond to Pressurizer Level Controller Failure. New Alternate path.

Safety significance failure to complete JPM will result in a loss of RCS inventory.

JPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available-LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA.

JPM D - Not used for Instant SRO.

JPM E - RCP #1 Seal Failure. New Alternate path JPM.

Safety significance failure to properly address excessive seal leakoff could result in Seal LOCA.

JPM F - Restore 1 D BUS from SBO Operations. New Involves restoring the "An train ES Bus from Station Blackout Diesel power to normal Off-Site power.

Safety significance is improper operation could lead to loss of 1 train of ES components through loss of ES bus, causing significant degradation in coping capability.

JPM G - Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H - Cross Connect the Secondary River Water System to the Nuclear River Water System. Modified JPM, significantly updated to current procedures, change from 2007 NRC is the alternate path.

Safety significance failure to complete the task would result in loss of adequate cooling for Nuclear Services components.

JPM I - Manually Operate MU-V-20 and IC-V-4. Bank JPM In Plant Engineered Safety Features JPM.

Safety significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J Local Start of EG-Y-1 B and Loading of 1 E 4160V bus. New JPM supplies power to the protected busses on Control Room Evacuation. Safety significance Remote shutdown capaoility only exists if "B" train power is available.

JPM K Respond to a failure of EF-P-2A to start, and EF-V-30D to close. Modified Bank JPM, old JPM involved steam bound pumR, when restarted had to be done from switchgear. SWitchgear start is no longer in procedure, added failed open discharge control valve on start.

Safety significance faikelre to complete 1st part of task could lead to inadequate secondary feed for loss of steam dnven pump, 2 part failure could lead to excessive heat transfer.

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Three Mile Island Date of Examination: August 2011 Exam Level: RO D SRO-I D SRO-U [2j Operating Test Number: 289-2011-301 Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)

Safety System I JPM Title Type Code*

Function

. a. ATWS with Faulted EOP-001 IMA's (007 EA2.04)

N/S/A 1

i

c. Transfer to Reactor Building Sump Recirculation (011 EA 1.05)

DIS/LINEN 3

I g. Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure (072 MIS/A 7

A3.01)

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

i. Manually Operate MU-V-20 and IC-V-4 (015/017 M 1.07)

D/E/R 4P

j. Local Start of EG-Y-1 B and Loading of 1 E 4160V bus (068 M 1.10)

N/E 8

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 1SRO-II SRO-U (A)Itemate path 4-6/ 4-6 1 2-3 (C)ontrol room (D)irect from bank 5:.9/ 5:.8 / 5:. 4 (E)mergency or abnormal in-plant

~ 1 /

~1 / ~ 1 (EN)gineered safety feature

- /

/ ~ 1 (control room system

{L)ow-Power 1Shutdown

~ 1 /

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

~21 ~2 1 ~ 1 (P)revious 2 exams 5:. 31 5:.3 1 5:. 2 (randomly selected)

(R)CA

~ 1 1 ~1 1 ~ 1

. (S)jmulator ES-301, Page 23 of 27

ES-301 Control Room/In-Plant Systems Outline Form ES':'301-2 THREE MILE ISLAND 2011 NRC SROU EXAMINATION JPM A - ATWS with Faulted EOP-001 IMA's. New Alternate path.

Safety significance failure to follow the alternate path will result in the reactor not being shutdown when it shoula be.

JPM B - Not used for SRO Upgrade.

JPM C - Transfer to Reactor Building Sump Recirculation, Bank Alternate path involving Engineered Safety Features equipment.

Safety significance failure to recognize DH-V-6A failure to open will result in damage to one of two available~LPI trains during post accident conditions, seriously degrading safety margin. PRA SAHSR2 HSROA switch over to sump following LOCA.

JPM D - Not used for SRO Upgrade.

JPM E - Not used for SRO Upgrade.

JPM F - Not used for SRO Upgrade.

JPM G - Respond lAW OP-TM-MAP-C0101 Alarm Response with Failure. Modified Alternate path Similar to 2007 NRC JPM however this involves alternate path.

Safety significance Failure to place control tower on "recirculation" following high airborne contamination in the Control Room may result in unnecessary dose for the personnel that must remain to operate the plant.

JPM H - Not used for SRO Upgrade.

JPM I - Manually Operate MU-V-20 and IC-V-4. Bank JPM In Plant Engineered Safety Features JPM.

Safety significance failure to complete the task could result in a loss of Reactor Coolant Pump seal cooling, and subsequent LOCA.

JPM J - Local Start of EG-Y-1 B and Loading of 1 E 4160V bus. New JPM supplies power to the protected busses on Control Room Evacuation. Safety significance Remote shutdown capability only exists if "B" train power is available.

JPM K - Not used for SRO Upgrade.

Appendix 0 Scenario Outline Form ES-D-1

~------------~------------------------------------------------

Facility:

Three Mile Island Scenario No.:

2 Op Test No.:

289-2011-301 Examiners:

Operators:

Initial Conditions:

(Temporary IC-52) 100% Power, MOL EF-P-1 ODS for bearing replacement, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock T.S. 3.4.1.1 (2)

.-~..

Turnover:

Maintain 100% Power Operations Critical Tasks:

Establish FW Flow and Feed SG(s) (CT-10)

I Electrical Power Alignment (CT-8)

Event No.

Malt. No.

Event Type*

Event Description I

1 MU01B CCRS MU-P-1 B Trips (TS), entry into OP*TM-AOP-041 CURO (URO: ensures MU-V-32 is in HAND and closed, ARO: Starts MU P-1A)

CARO TSCRS 2

ED18B CCRS Loss of the 8 Bus (TS) with EG-Y-1 B failing to start, entry into OP TM-AOP-014 CARO

• (ARO: Starts the SBO)

TSCRS 3

NI27A ICRS

.. Narrow Range Pressure Instrument Fails high with SASS failure to actuate, entry into OP-TM-MAP-G0308 IC48 IURO (URO: Closes Spray Valve, selects Alternate Pressure Instrument, IARO i ARO: Manual control of Pressurizer Heaters) 4 FW15B NCRS Loss of FW-P-1B, entry into OP-TM-MAP-M0107, Runback Fails to Occur, Power Reduction Performed, entry into OP-TM-MAP R URO H0101 and 1102-4 NARO 5

. RD0230 TSCRS

• Stuck Rod (TS), entry into OP-TM-AOP-062 I

I 6

ED18A MCRS Loss of Offsite Power (tripping SBO output breaker), entry into OP TM-AOP-020, OP-TM-EOP-001 MURO MARO 7

EG01A CCRS

• "AI! EDG fails to start, manual start required. (URO)

CURO 8

ICR02 CCRS EFW Valves for "AI! OTSG fail to 0% in Auto, manual control required. (ARO)

ICR04 CARO (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

- 1

Appendix D Scenario Outline Form ES-D-1

~--------.----------------------

Facility:

Three Mile Island Scenario No.:

3 Op Test No.:

28~-2011-301 Examiners:

Operators:

Initial Conditions:

(Temporary IC-54) 100% Power, MOL EF-P-1 OOS for bearing replacement, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> into a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> clock T.S. 3.4.1.1 (2)

Turnover:

Maintain 100% Power Operations Critical Tasks:

Reduce Steaming/Isolate Affected SGs (includes use of SG drains) (CT-22)

Minimize SCM (CT-7)

I~

Event Type*

Event Description 1

03A4S01 ICRS Inadvertent ES Actuation, "S" Train (TS), entry into OP-TM-AOP 046 IURO ZDIPS1R (URO: Defeats signal, ARO: Opens MU-V-2A/S)

CBON

• IARO i

I i

• TS CRS 2

RC08B ICRS Tc Instrument Fails High, SASS Fails to Actuate, entry into OP lC51 TM-AOP-070 IURO (LlRO: Manual control of Control Rods, ARO: Manual control of IARO Feedwater) 3 TH17S CCRS

-30 gpm "S" OTSG Tube Leak (TS), entry into OP-TM-EOP-005 i CURO (URO: Guide 9) i

• TSCRS 4

N/A NCRS Power Reduction lAW 1102-4 i RURO

• NARO 5

CC04A CCRS Loss of ICCW, entry into OP-TM-AOP-032, OP-TM-EOP-001 CC04B i

CURO (URO: Trips Reactor) 6 THi6B MCRS

-800 gpm "B" OTSG Tube Rupture, entry into OP-TM-EOP-005 MURO MARO 7

MS09A CCRS "S" TBV's fail closed, entry into OP-TM-421-451 MS09B CARO (ARO: Places ADV on Backup Loader)

I MS09C (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor il

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

Three Mile Island Scenario No.:

4 Op Test No.:

NRC Examiners:

Initial Conditions:

nover:

ical Tasks:

Malt. No.

I~o.

1 N/A 2

IC37A IC41A 3

IA07 IA01C 4

MS02B 5

TH08 6

R003C 7

MS02B 8

MUR67 MUR94 Operators:

(Temporary IC-55) 5% Power, MOL EF-P-1 OOS for bearing replacement Continue with Power escalation Control HPI (CT-5)

Isolate Overcooling SGs (CT-17)

Event Type*

NCRS RURO CCRS CARO TSCRS CCRS CARO CCRS CURO CARO TSCRS CCRS CURO CCRS CURO MCRS M URO MARO CCRS CURO Event Description Raise Reactor Power from 5% to 10%

Invalid "A" OTSG Low Level (TS), "An EFW actuation, entry into OP-TM-424-901 (ARO: defeats invalid signal, secures EF-P-2A)

Loss of Instrument Air, entry into OP-TM-AOP-028 (ARO: Starts IA-P-1 A or B)

Steam Leak into the Reactor Building. entry into OP-TM-AOP-051 RR-P-1 B Fails to start (TS)

(URO: Initiate Plant Shutdown, ARO: Initiate RB Emergency Cooling)

PORV fails open, entry into OP-TM-MAP-G0106 (URO: Closes PORV Block Valve)

Uncontrolled outward rod motion, group 7, entry into OP-TM-AOP 064 (URO: Selects Sequence Override)

Steam Rupture in Reactor Building. entry into OP-TM-EOP-001, Excessive Heat Transfer, isolate uB" OTSG, entry into OP-TM EOP-003, OP-TM-EOP-01 0, Rule 3 MU-V-36 breaker opens, Valve fails closed. Alternative minimum flow path for Makeup Pump established.

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

- 1

Exelon Nuclear Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 April 29, 2011 TMI-11-033 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406 Three Mile Island Unit 1 Facility Operating License DPR -50 NRC Docket No. 50-289

Subject:

Submittal of Initial Operator Licensing Ex.amination Outlines

~

Enclosed are the examination outlines, supporting the Initial License Examination scheduled for ~ust 29, 2011, at Three Mile Island Unit 1.

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

In accordance with NUREG 1021, Revision 9 Supplement 1, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.

Should you have any questions concerning this letter, please contact Mike Fitzwater of Regulatory Assurance at (717) 948-8228. For questions concerning examination materials, please contact Greg Hoek, Exam Author, at(717) 948-2027.

Respectfully (2vL~,htv William Noll

Vice President TMI, Unit I WN/mdf

Enclosures:

(Mailed to John Caruso, Chief Examiner, NRC Region I)

Examination Security Agreements (Form ES-201-3)

Administrative Topics Outlines (Form ES-301-1)

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

PWR Examination Outline (Form ES-401-2)

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

Statement detailing method of Written Outline generation Scenario Outlines (Form ES-D-1)

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

Completed Checklists:

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

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

Nuclear Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P.O. Box 480 Middletown, PA 17057 April 29, 2011 TMI-11-032 U.S. NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406

-C!');o

-.~-:

, i rTl Three Mile Island Unit 1

,n Facility Operating License DPR -50

)

NRC Docket No. 50-289

~

~~

1--'0

Subject:

Submittal of Knowledge and Abilities (KIA) statements that will be suppressed frorq;.the random exam generation process w

w It is our intent to develop the upcoming initial license exam scheduled for August 29, 2011 in accordance with NUREG-1021, Revision 9, "Operator Licensing Examination Standards for Power Reactors".

In accordance with NUREG 1021, "Operator Licensing Examination Standards", Three Mile Island Unit 1 is submitting for your review the list of KIA statements that will be suppressed from the random exam generation process in support of our August 29, 2011 license exam.

Should you have any questions concerning this letter, please contact Mike Fitzwater of Regulatory Assurance at (717) 948-8228. For questions concerning examination materials, please contact Greg Hoek, Exam Author, at (717) 948-2027.

Respectfully

~~

WilHam Noll Vice President TMI, Unit I WN/mdf

Enclosures:

Three Mile Island Unit 1 Suppressed KIA statements cc:

(without attachments)

Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector - TMI Unit 1

TQ-AA-151 Revision 3 Page 23 of 81 ATTACHMENT 4 Knowledge and Abilities Suppression Checklist Page 1 of 3 Initial NOTE:

Initial license exams are required by NUREG-1021, ES-401 to be developed from the knowledge and ability (KIA) statements contained within NUREG 1122 (for PWRs) or -1123 (for BWRs). Experience has shown that pre screening the applicable catalog to eliminate (suppress) inapplicable KIA statements before development is the most efficient process.

However, failure to train on a KIA statement is not an acceptable basis for suppressing the statement.

NOTE:

If the station has previously performed a suppression of the KIA manual and had it approved by the NRC, then Steps 1 and 2 may be marked N/A.

NOTE:

Generic KlAs associated with emergency and abnormal plant evolutions (E/APE) and plant systems for both RO and SRO examinations should be randomly selected from the following: 2.1.7, 2.1.19, 2.1.20, 2.1.23, 2.1.25, 2.1.27, 2.1.28, 2.1.30, 2.1.31, 2.1.32, 2.2.3, 2.2.4, 2.2.12, 2.2.22, 2.2.25, 2.2.36, 2.2.37, 2.2.38, 2.2.39, 2.2.40, 2.2.42, 2.2.44, 2.4.1, 2.4.2, 2.4.3, 2.4.4, 2.4.6, 2.4.8, 2.4.9, 2.4.11, 2.4.18, 2.4.20, 2.4.21, 2.4.30, 2.4.31, 2.4.34, 2.4.35, 2.4.41, 2.4.45, 2.4.46, 2.4.47, 2.4.49, and 2.4.50. All other generic KlAs for systems and evolutions may be suppressed.

The only generic KlAs that can be suppressed for the generic section of the exam (Tier 3) are KlAs 2.2.3 and 2.2.4, but only at single-unit facilities.

1.

REVIEW the NRC list of GENERIC KlAs from NUREG-'1122 or NUREG 1123, and SUPPRESS non-applicable KlAs in accordance with NUREG 1021, ES-401, Section D, Examination Preparation.

2.

REVIEW the NRC list of Tier 1 (E/APEs) and Tier 2 (Plant Systems) KlAs and SUPPRESS non-applicable E/APEs, Plant Systems, or KlAs in accordance with NUREG-1021, ES-401, Section 0, Examination Preparation.

3.

VERIFY the list of suppressed knowledge and abilities (KlAs), NUREG-1122 for PWRs or NUREG-1123 for BWRs, has been reviewed for the following:

- Inapplicable KlAs identified during the last NRC exam development.

- Changes incorporated into the facility since the last NRC exam.

4.

VERIFY the list of suppressed KlAs has been approved by both an Operations department representative and a Training department representative.

SRRS: 3D.100;Retain this document until all regulatory actions associated with the exam are complete, at which time it may be discarded.

TQ-AA-1S1 Revision 3 Page 24 of 81 ATTACHMENT 4 Knowledge and Abilities Suppression Checklist Page 2 of 3 Initial

5.

If the site had previously submitted a list of suppressed KJAs AND !!.2 changes were made to the list from the site's last initial license exam then NOTIFY the NRC chief examiner that the list of suppressed KJAs has not changed from the last submittal and enter "N/A" in step 6. For Certification Exams, notify the NRC exam author of 'No change'.

6.

If the site had not previously submitted a list of suppressed KJAs OR changes were made to the list from the site's last initial license exam then contact the applicable NRC chief examiner and schedule a submittal of the suppressed KJAs OR the revisions to the list of suppressed KJAs (per NUREG-1021, ES-401, Section D) to occur before the NRC license exam outline submittal.

Name of NRC chief examiner contacted Scheduled Date 6.1 REQUEST Regulatory Assurance / Licensing to draft a suppressed knowledge and abilities submittal letter. The template shown on page 3 of this attachment may be utilized to help draft a submittal letter.

(NIA for Certification Exam) 6.2 SUBMIT the approved list of suppressed KJAs to the applicable NRC region for review per the agreed date. (NIA for Certification Exam)

SRRS: 30.1 OO;Retain this document until all regulatory actions associated with the exam are complete, at which time it may be discarded.

TQ-AA-1S1 Revision 3 Page 25 of 81 ATTACHMENT 4 Knowledge and Abilities Suppression Checklist Page 3 of 3 Suppression Submittal Letter Template

<Date>

<Letter No>

<U.S. NRC Region I Administrator U.S. NRC Region 11/ Administrator>

<Current Address Current Address>

<Site Name, Units # and #>

Facility Operating License <Nos. NPF-## and NPF-##>

NRC Docket <Nos. ##-### and ##-###>

Subject:

Submittal of Knowledge and Abilities (KIA) statements that will be suppressed from the random exam generation process It is our intent to develop the upcoming initial license exam scheduled for <dates> in accordance with NUREG-1021, Revision <current revision>, "Operator Licensing Examination Standards for Power Reactors".

In accordance with NUREG-1021, Revision <current revision>, "Operator Licensing Examination Standards for Power Reactors", <station> is submitting for your review the list of KIA statements that will be suppressed from the random exam generation process in support of our <date> license exam.

Should you have any questions concerning this letter, please contact <name>, Regulatory Assurance Manager <or current title> at <outside phone number>,

Respectfully,

<name>

Site Vice President

<name> Station

Enclosures:

<station> Suppressed KIA statements cc:

(without attachments)

Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector - <Name> Station SRRS: 3D.100;Retain this document until all regulatory actions associated with the exam are complete, at which time it may be discarded.

Facility: TMII Suppressed KlAs 001 Continuous Rod Withdrawal AK2.0 1 Rod bank step counters AK2.06 T-ave./ref. deviation meter

,~,,~~,.,. Control Rod AK:~ _ -_ JM,rro",op, I

7\\.06

' Reset of demand position rAk i counter to zero

~

1.01 Demand position counter and pulse/analog,-VII velIe' i

r005

, Inoperable/Stuck Control Rod Metroscope Metroscope Trip - Stabilization Printed 01(03/2011 IMPORTANCE BO/SRO no rod bank step counters 2.9/3.2 atTMI no T -ave./ref. deviation 3.0*/3.1 meter at TMI _._------

at 3.1*/3.2*

i no demand position 2.7*/3.0 ~

I counter at TMI no demand position 2.9*/2.9 counter at TMI I

3.4/3.9

009 Break LOCA

.~...--

EK3.01 I CCW System automatic no automatic 3.1 */3.6*

isolation on high delta isolation at TMI i flow/temperature to RCP i

thermal barrier I

J EK3.02 I Opening excess letdown 2.8*/3.2*

i JO eX',,, [etdo"",

' isolation valve atTMI i

I rEK3.09 Closing CCW surge tank vent

' not performed at

3. 1*/3.4*l TMI I

.,__J

.~ ~.

I EA2.07 CCWS surge tank vent isolation no such component 2.7*/3.1

• I valve indication at TMI

..~

I

~ EA'i:18 I CCW temperature indication for no such component 2.3/2.6*

I I

RCP oil coolers atTMI r-EA2.22 I Charging flow trend recorder no such component 3.0*/3.3*

L atTMI I

~A1:08----i~~~~pre~u;~~~~;oo~t-----------------t~.c~tcomPoI;errtl~****22::;.7*/2.6*

KellctCtr Coolant Pump (RCP) Malfunctions RCP run component 2.5/2.5 indicators I

no such equipment i..

2.9*/3.0**-j'.

at TMI

~~:I1~--------11~~~~~~~~~;~~;---t-----------'----------------------~n.~o~S~UdCh~eqWU~i~p~m~e 3.5.*/3..6*-

.. n:J*

.a_tTMI __....

_... ____.. __J

~------.-:--.--.-+~-==='::~~=::::--;-;---;---t---------------.-.-

I no such equipment' 3.2*/3.5*

I atTMI

..._.~!

no such equipment 2.9* /3.0*

at TMI c----::-l:-:l-**--*----L-;::-;,;--*~---:--**-;;-;:::;;:;----;--;-----;~~:;-*-l----*----------------1-:d;to~n;:o~t;.jo;:g--:-RO'CPS at 3.4" /3.8

• TMI

022 Loss of Reactor Coolant Makeup AK.'.03 Performance of lineup to no excess letdown establish excess letdown after path at TMI i

determining need AA1.04 Speed demand controller and no poshive 3.3/3.2*

running indicators (positive displacement displacement pump) pumps used for reactor coolant makeupat TMI AAl.07 Excess letdown containment no equipment 2.8*/2.7*

isolation valve switches and indicators 024 Emergency Horation AAL08 2.7*/3.0*

I-=L~o~ss~':::.:of:c..'C=om="-"o:::::n=en::.t:c..C~()olin Water (CCW) l AA2.03 The valve lineups necessary to 2.6/2.9 restart the CCWS while bypassing the portion of the system causing the abnormal

______________-L~co~n~d~i~ti~on~_________________L _________________________________~_______________-L_~________~

10

-I i 027

• Pressurizer Pressure Control (PZRPCS) M.~a~l~fu~n~c~ti:::o.:::n.,..._______________

~.

AA2.17

--*----rAIlo~able ReStemperature I

I not applicable to 3.1/i3--i I

I difference vs. reactor power TMI I

i i

1____

_ ____.1-1_______

Level Control Malfunction

~~~~:==:~=.~=.i~~~

..------..-.------------,.-~__=:==__:___:__::___-_r-::_==-_::-_____1 PZR level reactor protection no PZR level input bistables to RPS at TMI Regenerative heat exchanger no such component and limits atTMI exc~letdQwn per not.pplicable at 2.8/2.9 theCVCS TMI

EAl.09 Manual rod control Driving of control rods into the core Positive dis,olalcelmeJnt charging pumps 032 Loss of Source Range Nuclear Instrumentation automatically removed Effect improper HV Loss of Intermediate Range Nuclear Instrumentation trip bypass valves at TMI not performed at 4.0/3.6

. TMI not performed at 4.2/3.9 TMI

'1 no positive~3.1*13.4*

displacement

. charging pumps at TMI

_'--__~J does not apply to 3.0/3.1 TMI does not apply to 3.4*/3.7*

TMI

Steam Generator (S/G) Tube Leak AK3.01 i Collection of Condensate in air not applicable to 2.3/2.6 i ejector monitor due to its failure TMI AK3.02 Reset and check of Condensate not applicable to 3.2/3.5 i

air ejector exhaust monitor TMI AK3.04 Use of "feed and bleed" process not applicable to 2.512.9 TMI AA2.07

. Flowpath for dilution of ejector not applicable to 3.1/3.6

! exhaust air TMI i Steam Generator Tube Rupture v,"ve'"

IEKJ.07

_-_-~_-~4_R-C-S~l..o..-op... ~.>o.-.~..~,~-va-lu--e-s----__+ __.~ _________. __. ______________ -r______lO~O_P____~__~~_.J.

, BALl3 Steam flow LUU1'w<<'CVH, no steam flow 3.7*/3.6 I',

indicators at TMI i------.... -.-.-.----~--f___--- --...--.----::-:-=-=-:---,---f----.---- --------------1...--

Interlock between MSIV and no such component b ass valve at TMI Reactor trip breaker and safety no such component injection interlock at TMI f___------------~-~--~~~--~~~---.--.~----

---

r-~~~------i--*----~-~--~

EA1.39 Drawing S/G into not performed at using the "feed and bleed" TMI method not applicable Loss of Condenser Vacuum ejector steam of Main Feedwater (MFW) no steam flow 3.4*/3.7*

indicators at TMI no first-out 3.4*/3.9 indication at TMI no such component 2.9/3.3*

atTMI

055 Loss of Offsite and Onsite Power (Station Blackout)

---

~~~~~--------,-.-------~------------~----------~---

EAl.03 Manual MTjacking 056 Loss of Offsite Power AA1.15 Service water booster pump AA1.19 Battery room ventilation exhaust fan AA1.20 Speed switch room ventilation fan

'AA2.11 Operational status of service water booster pump AA2.28 Auxiliary building gas treatment indicator 1----- --- --

AA2.29 Service water booster pump ammeter and flowmeter AA2.62 Breaker for feedwater pumps 067 Plant Fire on Site AA2.1O I Time limit of long-term breathing air system for control room 068 Control Room Evacuation r----------;:=:

AIG.04 Filling the feedwater system and closing the AFW pump discharge valve

~i(3.05 Repositioning valves to isolate and drain the AFW pump turbine and steam supply header Safety injection setpoint of main I AK3.14 steam line pressure AK3.16 Fail-open of the control room doors for personnel evacuation AA1.20 Indicators for operation of startu~ transformer not performed at 1.9*/1.9*

TMI J

no such components 2.7*/2.9*

atTMI no such components 2.4*/2.4

• atTMI no such components 3.0*/3.0*

atTMI no such components 2.9*/2.9*

I atTMI no such components 2.2*/2.6* ~

I atTMI no such components 3.0* /3.2*

atTMI no such components 1.7/1.9*

atTMI I

no such system at T

2.9*/3.6*

TMI l

"I not performed for 3.0*/3.2*

this event at TMI not performed for 2.5*/3.0*

this event at TMI I

no SI setpoint on 3.2*/3.4 " i MS pressure at TMI

.----1 not applicable at 2.8*/3.3*

I TMI no SU transformers 3.2*/3.2*

j atTMI

High Reactor Coolant Activity AK3.02 Increased CCW flow 00]

Control Rod Drive System K2.01 One-line diagram of power supply to MIG sets K2.04 Control rod lift coil K2.05 MIG

~~-.-

KS.43 Definition of T -ref KS.60 Reason for using MIG sets to po~er rod control system KS.61 Operational theory for MIG sets K.'1.70 Method used to parallel the rod control MiG sets

.~----

KS.71 Reason for maintaining cross-tie breaker between rod drive MIG sets; reliability of control rod drive trip breakers during o!,eration of one MIG set KS.97 Relationship of T-ave. to T-ref 1-.

KS.98 Effect of adding high or low boron concentration to maintain T-ave. egual to T-ref Location and operation of rod I K6.1O control MIG sets and control panel, including tri~

T-ref L~l.02 "Prepower dependent insertion limit" and power dependent insertion limit, determined metroscope Fractured split pins

~~:

• A2.10 Loss of power to MIG sets I A2.20 Isolation of lift coil on Iffected rod to prevent coil burnout A4.12 Stopping T/G load changes; only make minor adjustments to prevent coil burnout I

..~-.----

no CCW flow control at TMI no MIG sets at TMI no such component atTMI no MIG sets at TMI not applicable to TMI no MIG sets at TMI no MIG sets at TMI no MIG sets at TMI no MIG sets at TMI J no T-ref at TMI I no T-ref at TMI M/G

. at TMI no T-ref at TMI no metroscope at TMI Not at TMI no MIG sets at TMI no lift coils at TMI no lift coils at TMI 2.4/2.6 3.5/3.6 2.1 */2.7 3.l*/3.5 3.2/3.4 I:

1.912.4 1.5/1.7 2.1/2.6 2.4/2.9 3.3/3.6 3.4/3.8 3.1 */3.3 3.1/3.4 4.0'?/4.2?

1.9?/1.9

.......----.-.~

3.4/3.9

---~

2.6*/3.6*

2.9*/2.9

Chemical and Volume Control System (CVCS)

KS.33 Use of a boronometer I

no boronometer at 2.3*/2.6 TMI K6.33 Principles of boronometer 1

lO boronometer at 1.9*/2.1 I'MI I M,22 Boronometer chart H;A,vIU",

10 boronometer at 2.5*/2.5*

'MI OIl Pressurizer l..evel Control Sy~tem (PZR LCS)

K4.05 PZR level inputs to RPS I no PZR Level inputs 3.7*/4.1 I

to RPS at TMI K4.06 Letdown isolation no interlock at TMI 3.3/3.7 L----.--......

Reasons for starting charging no interlock at TMI 2.8* /3.2*

lK601 pump while increasing letdown flow rate I

i Rea(':tor Protection System I K4.07 First-out indication no first-out 3.0/3.2*

indication at TMI K6 07 Core protection calculator no such component 2.9*/3.2*

.r I

I atTMI K6.08 COLSS no such component 3.6*/3.7*

at TMI K6.09

! no such component 3.6*/3.7*

I CEAC atTMI M.07 MIG set breakers

=-= no MIG set at TMI 3.9*/3.9*

. no metroscope at 1.9*/2.2

TMI no metroscope at 2.5*/2.7*

TMI I

no metroscope at 2.5*/2.7*

TMI I

no metroscope at 2.1 */2.6 TMI Metroscope reed switch display no metroscope at TMI Loss of L VDT no L VDT at TMI 2.6*/3.0*

Primary voltage not applicable Lo 2.6*/2.7*

measurement TMI

~~6:=-':t Spray System (eSS) i

,--l

K4.08 i Automatic swapover to no automatic 4.1 */4.3*

'containment sump suction for swapover at TMI recirculation phase after LOCA f--------

RWST low-low level alarm K4'()9 Prevention of path for escape of no such interlock at 3.7*/4.1

• radioactivity from containment TMI

to the outside (interlock on f--_______-+-"'Rc.:.W'-'S::..T.::....::is::::o:.:.:::lation after swa ver

--.......,---------.----t-A4.02 Prevention of path for escape of no such components 2.3*/2.6*

radioactivity from containment atTMI to the outside (interlock on RWST isolation after swapover)

The remote location and use of spool pieces and other equipment to set up portable recirculation pump for additive

.~____+-='t:::an:.::k;;).,c.:.i:.:-nc::::l.:::u.::dl:.:*n:<;;guP::::o::...:w.:..;e:..:r...::.s:..:u:.t:p:.t:p::"ly'-_t-_____.___________+-_______+ ____......._

A4.03 The remote location and use of no such components 2.2*/2.5*

tank needed for atTMI Recombiner and Purge Control System (HRPS) such equipment TMI 039 Main and Reheat SteamSystem (MRSS)

K4.08

!5terlocks 00 MSIV and bypass I

. interlocks 3.3/3.4 valves

~8 Reheater steam pressure no interlocks at 1.8/1.9 TM!

i A3.0l Moisture separator reheater no interlocks at 1.9*/1.7 steam supply TM!

A4.05 Moisture separator reheater, no interlocks at 1.8/1.6 checking its temperatures and TMI steam pressures relative to heatup limits and operating L-_.

limits i

041 Steam Dump System (SOS) and Turbine Bypass Control K4.09 Relationship of 10wi1ow T-ave.

setpoint in SOS to primary cooldown K4.11 T -ave.IT -ref. program K4.14 Operation of loss-of-load bistable laps upon turbine load loss Al.OI T-ave., verification above lowi1ow setpoint A4.01 ICS voltage inverter I A4.03 T

mode 045 Main Turbine Generator (MT/G) System K410 Programmed controller for T ref. signal generation from first stage (impulse) pressure in turbine 15 Steam blanketing (atmospheric pressure) moisture separator reheater to drive out air and non condensables rior to starlin u K4.44 Impulse pressure mode control oJ steam dumps

-~---~-+-

K4.46 Defeat of reactor trip by overspeed trip test lever Generator amplidyne balance system 055 Condenser Air Removal System (CARS)

Kl.07 WGOS K6.01 Air ejectors A2.03 Loss of air ejector cooling water

~.-

not annlicahle to 3.0/3.3*

TMI not applicable to 2.8/3.1 TMI not applicable to 2.5*/2.8 TMI not applicable to 2.9*/2.9 TMI not applicable to 2.9*/3.1

• TMI not applicable to 2.4*/2.5*

TMI not rlP,*'*"'*.....F*" at TMI 2.5*/2.8*

no interface with 1.9/1.9 WGOSatTMI no air ejectors at 1.7/1.7 TMI no air ejectors at 1.8*/2.0*

TMI

056 Condensate S K4.21 Operation of pump no equipment air ejector recirculation line at TMI isolation valve to maintain K5. 14 Purpose of valve between upper no such equipment surge tank and hotwell at TMI 1A3_.0~5_________,,~M~o_nl_'to_r.i_n~g__--:~.__.__a_i_r_____-r___________.__~____~__________-+____~___~___~___1_.7_/l._._8_*__~

1--;...

! ejector air flow

A3.10 Upper surge tank Valve between upper and 6l

~

AU.xiliary / Emergenc. y Feedwater (AFW) System Main Feedwater (MFW) System Kl.l 0 I Diesel fuel oil no diesel driven-2.6*/2.7~

1 EFWpumps j

r--------..-.----I-----::-.--:.-:-.:--.--.--.-..--.-.---.---+----....--..--..--

driven pump no diesel driven 4 0* /3 8*-*

L ________

.___'--E_~FW p_u_m_p_s__--'-__

.J

~______~_________

... _~~_

---

..~--- -------------------.---.----.---------

l 1 064

!-",E;.:;;m=ec:.r....e:c;;n:.;;c",-"D;.::i.:..esti.Generator (ED/G) System J

M.IO

~1~~nO~f~E~D~/~G~J~m~e=g=~-~-t--~-----------------~I~oo--~-C~h-c-o-m-~-n-e-n-t~I--2~.~~-:2~.8~**~-1 load controller

! / operation at TMI I

'---T Need for set*-ti-n-g-o-f-fs-it-e-p-o-w-e-r--+--------------

---

1I'-n-o such. component

'1-,--3:-.-1-*/'"":2-,9:-*"---1 A3.11 e-.-._______. __.___-t-b,.-r_ea_k_e_r_Io_al:l!omatic:.._____t_________.___________-+..../.::o.=er;.::a:::li:::o,.;.;n..:a:.:.'..:TM~I=---1'_--:-_-:-_--1 Remote operation of the air no such com~nent 3.2* /3.2 l

compressor switch (different I operation at TMI modes

'---_.----_._--_._---'---'---'------- ---.-~.---.---.~--~

--..~.~-.

.*..----.-.-~-

075 Circulating W~ter Sfstem I Kl.02 Liquid radwaste discharge*

no interface at TMI 2.9/3.1 i KI07_

R,e""",';"n 'pmy 'y,rem no such component 2.2*/2.1

• I at TMI I Kl.09 Vacuum priming no vacuum priming I.S/1.4 for Circ Water at +

-+-I_____..__~.________--I_T:;.:MI_

IK204--**~'

I nosuch component 1.4*11.4*

_j~Ube oil pumps.

I at TMI I....J K3.0S Recirculation spray system I no such componei1tT 2.1*/2.3*

I

-..--.-------------i :~-:~"re <Doling f 17'/21;--1 K4.03 Interlocks between circulating water system pumps and cooling i tower pumps at I

I TMI

-1 K404 Automatic pickup of backup i

no such component 1 7*/1 9 I

f--=-----

lube oil pumps (ac and dc) ----t----..- __.....____.._______-+.:::at:..T:;.:M::..:;,.I____--4-. _

---~i r*_K_4_.0_6____..___

__ --t------.--------------..,.....:~::.~...:;..:.:~~c:~_co_m_p_o_n_e_n_t--+1_.__1_._6/1_.8

-+_T_f.a_:--v_e_lin:.:c_r_e_en_o_p_er_a_ti_on

___ 1 K5.07 Relationship of seawater I no seawater at TMI I 1.4*/1.6*

I--.-----.--.~------f..--temperature tQ marine growth K5.08 Purpose of the vacuum priming I no vacuum-p*-r-:-im-in-g--r---:-1.-6/-1-.6---I I system I ~~Tirc Water at AL08 Circulating water makeup pump such component 1.6*11.6*

molor current (within limits) at TMI A2.01 Loss of intake structure no intake structure 3.0*/3.2 for Circ Water at TMI I__._______.___-I_:---_~--~_:_-_--+-_-**-*-**----*****--***-~-~-..~~.---__I~~._:_-....-----+------.--1 A2.08 Ice buildup on intake structure no intake structure 2.0*/2.0*

for Circ Water at

,---_._-+-______._.__..~___.___.____*t~T~MI _____-1-__.. __.__ -~~

ALlO I Autom-atic startup mode of water no priming pumps 1.S*/1.6*

I

box priming pumps relative to at TMI

~~___.. ____+lspeci.!i~minimum va:.::c.::.uu:::m:::-_-+____~.___~.._~...____. --+--.,.".,.--::-:-----r---~**c**=-:-_:*-::-:--_1 A2.11 Time required for fill of piping not filled by 1.5*/1.6*

by induction of water into induction at TMI circulating system using vacuum system A4.04 Air eductor system

---

.---l-:~;;:;'lh LB'Jl-:j compo",",

Water box vacuum riming

-~---~.

'::n;;"0':"v;;"ac::::u*-u-m-p-ri-m-:-in-g-rl-l~.-=-8*-/1.7*

f-A4.06 isolation valves, control for Cire Water at switches, and il!.':'.dl,,:*c,,:,at,,:o~rs~-_~_-+________________----ji--T~M:.:I_____..:---:---1--;~-::-::*:~:--::-:::-- -,

r-A -4

.-cO ---

Vacuum priming lank/priming no vacuum priming 1.7*/1.6*

compressor controller for Circ Water at TMI A4.08 Gland seal water supply system such component 1.6/1.6 at TMI A4.14 Lube oil pumps for circulating such component 1.S*11. 7*

!_..~'.._--:-__"____--ir-~w~ater pump A4.1S Operation of the vacuum priming system atTMI vacuum priming Cire Water at TMI 1.4/1.S 1.6/1.6 A4.16 Traveling screens in manual operatio~

preventers 076 Service Water System (SWS)

Relationship of SWS to water filtration (RWF) and location of SWS no RWF system at TMI 1.9*/1.9*

2.2*/2.2*

no heat sink pond at TMI 2.2*/2.5*

086 FPS)

KL02 Raw service water

! C~mtainment§~t~e_mT_~__'__"_"_'__"___"____ '_'-'---____~-"'~-'--~--"-----'-~-"'--,-'--------'--"--'--"-'----

Ll.03 ________

Shield building vent system 3.1 */3.5*

. KL06 Subsurface drain system 2.4*/2.7*

A4.02 Excess letdown divert valves 10 2.1*/2.2*

reactor coolant drain tank 2.7*/2.7*

controller 2.4*/2.2*

TMI A4.07 of the air lock rate lest panel no operated or monitored from the control room at TMI A4.08 Operation of refueling drain no operated or 1.9/2.2 valves (for draining refueling monitored from the canal to lower containment control room at

~-----------.------+------.-------------r-:--=----::---:-.-:--.------+--"-=-~:~::-_:-:-:::::---I A4.09 Containment vacuum system

The above KJAs are the pre-suppressed KIA's at TMI in addition to those allowed by D.l.b of ES 401 and all of system 25 Ice Condenser system as we have no Ice Condensers.

Evolution 037 AAl.13 will need to be unsuppressed as we now have S/G blowdown.

Evolution 038 EAl.l8 will need to be unsuppressed as we now have S/G blowdown.