ML12059A366

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Draft - Outlines (Folder 2)
ML12059A366
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 02/13/2012
From:
Division of Reactor Safety I
To:
Entergy Nuclear Indian Point 2
Jackson D
Shared Package
ML112350689 List:
References
U01844, 50-247/12-301
Download: ML12059A366 (35)
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Text

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Indian Point Unit 2 Date of Examination: February 6 2012 Examination Level: RO =:J SRO X Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Review an ECP by Hand Conduct of Operations M,R 1940012137 - Conduct of Operations - Knowledge of proceduf1as, guidelines, or limitations associated with i reactivity management. SRO-4.6 Determine Isolation Boundaries for CCW Leak Using Plant Print Conduct of Operations N,R 1940012'125 - Conduct of Operations - Ability to interpret reference materials such as graphs, curves, tables etc.

SRO-4.2 Reviewal Surveillance Test Equipment Control N,R 1940012212 - Equipment Control - Knowledge of surveillance procedures. SRO - 4.1 Review a Release Permit for Containment Pressure Relief Radiation Control M,R 1940012306 - Radiological Controls - Ability to approve release permits. SRO - 3.8 Classify an Emergency Event Emergency Procedures/Plan M,R 1940012441 - Emergency Procedures/Plan Knowledg!e of the emergency action level thresholds and classifications. SRO - 4.6 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator. or Class(R)oom (D)irect from bank (S; 3 for ROs; s; 4 for SROs & RO retakes)

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

(P)revious 2 exams (S; 'I; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Indian Point Unit 2 Date of Examination: Februa~ 6 2012 Exam Level: RO D SRO-I X SRO-U Operating Test No.:

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

System I J PM Title Type Code* Safety Function

a. Align CVCS Makeup after Chemistry Sample(004007 A4.07 N 1 RO-3.9 SRO 3.7)
b. Align Hot Leg Recirculation 23 SIP fails to Start A,M,EN 2 (000011EA1.13 R0-4.1 SRO-4.2)
c. Depressurize RCS during Natural Circ Cooldown and Block A,N,L 3 Low Pressure SI (WE09EA1.1 RO-3.5 SRO-3.5) i
d. NASROs NA NA
e. Respond to 22 SG "B" Level Channel failure High D 7 (059000A4.08 RO-3.0 SRO-2.9)
f. Perform the Required Action to Isolate the SI Accumulators during a Loss of Coolant Accident with Failure of MOV A,EN,D 2 894B to Isolate (Alternate Path) (006000A3.01 RO-4.0 SRO-3.9)
g. Start 21 RCP during FR-C.1 (WE06EA1.01 R03.8 SRO E,D 4P 3.8)

. h. Verify Containment Phase A Isolation Manually Close EN,A,D 5 Valves (103000A3.01 RO-3.9 SRO-4.2)

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

i. Align 23 Charging Pump to 12FD3 (000068AA1.06 RO-4.1 D,E 6 SRO-4.2)
j. Lineup Alternate Cooling to SIS and RHR Pumps (005000 D,E, R 8 2.4.34 RO-4.2 SRO-4.1)
k. Align 24 Waste Gas Decay Tank for Discharge (071000 R,D 9 2.3.11 RO-3.8 SRO-4.1)

@ 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 diffen:mt safety functions; in-plant systems and functions may i

overlap those tested in the control room.

i

  • Type Codes I Criteria for RO I SRO-II SRO-U

(A}lternate path 4-6 14-6 / 2-3 (C)ontrol room (D)irect from bank s9/s8/=s;4 (E)mergency or abnormal in-plant  ;;:1/;;:1/;;:1 (EN)gineered safety feature 1 ;;:1 (control room system)

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

(R)CA  ;;:1/;;:1/;::1 (S)imulator

ES-301 Administrative Topics Outline Form ES-301-1 Facility: Indian Point Unit 2 Date of Examination: February 6 2012 Examination Level: RO X SRO Operating Test Number:

Administrative Topic Type Describe activity to be performed (see Note) Code*

Perform an ECP by Hand Conduct of Operations M,R 1940012137 - Conduct of Operations - Knowledge of procedures, guidelines, or limitations associated with reactivity management. RO-4.3 Conduct of Operations NA for ROs Review a Surveillance Test Equipment Control N,R 1940012212 - Equipment Control - Knowledge of surveillance procedures. RO - 3.7 Calculate a Release Permit for Containment Pressure! Relief Radiation Control M,R 1940012311 - Radiological Controls - Ability to control radiation releases. RO - 3.8 Perform Initial UNUSUAL EVENT Notifications Emergency Procedures/Plan D,S 1940012439 Emergency Procedures/Plan - Knowledge of the RO's responsibilities in emergency plan implementation. RO - 3.9 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

  • Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::;; 3 'for ROs; ::;; 4 for SROs & RO retakes)

(N)ew or (M)odified from bank (<:! 1)

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

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Indian Point Unit 2 Date of Examination: February 6 2012 Exam Level: RO X SRo-ID SRO-U Operating Test No.:

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

System I JPM Title Type Code* Safety Function

a. Align CVCS Makeup after Chemistry Sample (004007A4.07 N 1 RO-3.9 SRO 3.7)
b. Align Hot Leg Recirculation 23 SIP fails to Start A, M, E 2 (000011 EA1.13 R0-4.1 SRO-4.2)
c. Depressurize RCS during Natural Circ Cooldown and Block I A,N,L 3 Low Pressure SI (WE09EA1.1 RO-3.5 SRO-3.5)
d. Fill the PRT (007000A4.01 RO-2.7 SRO-2.7) N 5
e. Respond to 22 SG "B" Level Channel failure High D 7 (059000A4.08 RO-3.0 SRO-2.9)
f. Perform the Required Action to Isolate the SI Accumulators during a Loss of Coolant Accident with Failure of MOV A,EN,D 2 894B to Isolate (Alternate Path) (006000A3.01 RO-4.0 SRO-3.9)
g. Start 21 RCP during FR-C.1 (WE06EA1.01 R03.8 SRO E,D 4P 3.8)
h. Verify Containment Phase A Isolation Manually Close E,A, D 5 Valves (103000A3.01 RO-3.9 SRO-4.2)

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

i. Align 23 Charging Pump to 12FD3 (000068AA1.06 RO-4.1 D 6 SRO-4.2)
j. Lineup Alternate Cooling to SIS and RHR Pumps (005000 D, E,R 8 2.4.34 RO-4.2 SRO-4.1)
k. Align 24 Waste Gas Decay Tank for Discharge (071000 R,D 9 2.3.11 RO-3.8 SRO-4.1)

I

@ 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 I Criteria for RO I SRO-II SRO-U 1

(A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S9/S8/S4 (E)mergency or abnormal in-plant <!:1/<!:1/;;;:1 (EN)gineered safety feature - / - I ;;;:1 (control room system)

(L)ow-Power I Shutdown  ;;;:1/;;;:1/;;;:1 (N)ew or (M)odified from bank including 1(A) ;;;:2/;;;:2/;;;:1 (P)revious 2 exams S 3 Is 31 S 2 (randomly selected)

(R)CA  ;;;:1/;;;:1/;;;:1 (S)imulator

ES-401 PWR Examination Outline Form ES-401-2 Facilitv: Indian Point Unit 2 Printed: 08/16/2011 Date Of Exam: 02/06/2012 RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 P.,3 A4 G* Total A2 G* Total

1. 1 3 3 3 3 3 3 18 3 3 6 Emergency 2 2 2 I 2 2 0 9 2 2 4

& N/A N/A Abnormal Tier Plant Totals 5 5 4 5 5 3 27 5 5 10 Evolutions 1 3 2 3 3 3 2 2 3 2 2 3 28 3 2 5 2.

Plant 2 I 1 I I I I I I I I 0 10 I I 2 0 3 Systems Tier 6 2 8 4 3 4 4 4 3 3 4 3 3 3 38 Totals 1 2 3 4 1 2 3 4

3. Generic Knowledge And 10 7 Abilities Categories 3 2 3 2 2 2 I 2 Note:
1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO a

and SRO-only outlines (Le., except for one category in Tier 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/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate 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 KlAs 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) KlAs 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 KlAs.

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling eqUipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #1 does not apply), Use duplicate pages for RO and SRO-only exams.
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SRO :selections to KlAs that are linked to 10 CFR 55.43.

NUREG 1021

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 ES-401 1 EtAPE # I Name t Safety Function 000007 Reactor Trip - Stabilization - x Knowledge of the reasons for the Recovery 11 following responses as they apply to the reactor trip: - Actions contained in EOP for reactor 000008 Pressurizer Vapor Space Accident x Knowledge of the reasons for the 13 following responses as they apply to the Pressurizer Vapor Space Accident: - Why PORV or code safety exit temperature is

_ _~,,..below RCS or PZR

-06~00:-:0-:-::9-::S:-m-a-::ll--B-r-ea~k--L--:O-::C:-A-/-::3::-----t---+- , Ability to operate and/or monitor the following as they apply to a small break 1--_ _-11. LOCA: - AFW/MFW .

0011\01\0'-:Il;-L;-ar--ge-B;:;"r-e-ak;--'-:-L-':::O-=C-A-I-3---~ x 1-----.

Knowledge ofthe interrelations between the Large Break LOCA and the following:

1 - - - - \- P.:umps I 000015/000017 RCP Malfunctions I 4 Conduct of Operations - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument inte retation.

000022 Loss of R¥: Coo)~nt Makeupl2 1--~--+IAbuitytodetennineandinterpretthe 3.71 *76 I followingasth~y appJy tome Losiof Reactor Coolant ;Makeup: - Charging 000022KIOI Knowledge of the operational 7 Loss of Reactor Coolant Makeup I 2 implications of the following concepts as they apply to Loss of Reactor Coolant Makeup: - Consequences of thermal shock to RCP seals 000025 Loss ofRHR System 14 x Knowledge of the interrelations between 2.6 the Loss of Residual Heat Removal NUREG 1021 2

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline ES-401 EtAPE # t Name I Safety Function #

000027 Pressurizer Pressure Control Ability to operate and/or monitor the I 4.0 TI3 '

System Malfunction / 3 following as they apply to the Pressurizer Pressure Control Malfunctions: - PZR heaters, sprays, and PORVs 000029 ATWS I 1 x Knowledge of the operational 2.8 8 implications of the following concepts as they apply to the ATWS: - definition of negative temperature coefficient as to lame PWR coolant 000038 Steam Gen. Tube Rupture 13 x Knowledge of the reasons for the 3.9 9 following responses as they apply to the SGTR: - Automatic actions provided by each PRM 000038 Steam Gen. Tu~ Rupture! l 4.6 I 77 000055 Station Blackout / 6 000057 Loss of Vital AC Inst. Bus / 6 Conduct of Operations - Ability to ntl"rnrl"t and execute nrC\('Prillrp 000058 Loss of DC Power 16 Ability to determine and interpret the I 3.5 following as they apply to the Loss of DC Power: - DC loads lost; impact on to

' - - -_ _ _..1..1-,-o,-pe_r-,-at,--e and monitor plant systems NUREG 1021 3

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline ES-40I Emerf!encv and Abnormal Plant Evolutions - Tier 1/ GrouD 1 Form ES-401-2 I E/APE # I Name I Safety Function I KI I K2 I K3 I Al I A2 I G Number I KIA Topic I Imp. IQrt

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

000065 Loss of Instrument Air / 8 to operate and/or monitor the 2.6 14 following as they apply to the Loss of Instrument Air: - Components served by instrument air to minimize drain on

'Cond~ct ofOperatlons'!' Ability to I 4.7 I ~ 79 eyahiateplant'petfonnance and make operational judgmentsbll$OOonopemiqg ch~ri8tic~reactor be~viQf,and 000077 Generator Voltage and Electric Knowledge of the interrelations between 3.1 15 Grid Disturbances / 6 Generator Voltage and Electrical Grid 1--_ _ _-+I~D~istLJrbances and the following: - Motors W IE04 LOCA Outside Containment I 3 Ability to determine and interpret the 3.6 16 following as they apply to the LOCA Outside Containment: - Adherence to appropriate procedures and operation within the limitations in the facility's I--_ _ _-+I~l--'icen~~ and amendments W/E05 Inadequate Heat Transfer - Loss x Knowledge of the operational 3.8 17 of Secondary Heat Sink I 4 implications of the following concepts as they apply to the Loss of Secondary Heat Sink: - Components, capacity, and function of W IE 11 Loss of Emergency Coolant Conduct of Operations - Knowledge of 3.9 18 Recirc. / 4 system purpose and or function.

W/EllLos~ Ability to detenrtineiandinteroretthe I 'lA I 'so Recirc.l4 fullqwingas they E:tnergencyPtlp}antIteclrc\1lfltion:

Ad,hefence to'l.lpprllpfiate proeed~esl.lDd

'operatl. within, theUmitations~Jl the

.f.l'lL'iHkIQ Iicense.1.lIld NUREG 1021 4

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 ES-401 Emerf!encv and Abnormal Plant Evolutions - Tier 11 GrouD 1 Form ES-401-2

[ EtAPE # / Name / Safety Function I KI I K2 I K3 I Al I A2 I G Number I KIA Topic I Imp. I Q#_

W!E12 Steam Lille Rupture ~Exc~ive Abiijtyto determine IJIld interpret the Hea:t:TransfefI4 following as they apply tQ,the Un90ntr()1l~*. ~p.sslJ~tion*.ofall StCatILQener;ators: .: Facilityc9nd~tions.

. lec1;i()tlof~ppropri. procedures during aqnormal*lJIld emergency rations NUREG 1021 5

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline I :~-:~1# I A

NS arne I Ii F' a etv unctIOn E~erge?cy a?d A~nor~al PI~n~ Erolutions - 'fie_r 11 Group 2 FO~1!:S-:~

000001 Continuous Rod Withdrawal / 1 x 3.6 19 82 Knowledge of the interrelations between I 2.6 I 20 the Pressurizer Level Control Malfunctions and the following: - Sensors and detectors 000032 Loss of Source Range NI / 7 X AA1.01 I Ability to operate and/or monitor the 3.1 I 21 following as they apply to the Loss of Source Range Nuclear Instrumentation:

Manual restoration of 0.00037 Steam.Generator Tube Leak 13 Emergency Procedure~lan -Knowledge 4.5 I I S4 of how abnonnal o~ting procedures are used incon'~nctionw:ith 000067 Plant Fire On-site / 9 Ability to operate and/or monitor the 22 following as they apply to the Plant Fire on Site: - Fire alarm 000074 Inadequate Core Cooling / 4 x Knowledge of the interrelations between I 3.6 I 23 the Inadequate Core Cooling and the follo\Ving: - RCP 000076 High Reactor Coolant Activity / 9 Ability to determine and interpret the I 2.8 I 24 following as they apply to the High Reactor Coolant Activity: - Corrective actions required for high fission product inRCS NUREG 1021 6

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline ES-401 2 Form ES-401-2 I --1 E/APE # / Name / Safetv Function WIE03 LOCA Cool down - Depress. 14 XI EK3.2 I Knowledge of the reasons for the 1--3.4 I 25 following responses as they apply to the LOCA Cooldown and Depressurization: -

Normal, abnormal and emergency operating procedures associated with LOCA Cooldown and W/E09Natural eire 14 4.1 85 roRtniti** tion 5 'es WIE 10 Natural Circ. 14 Ability to determine and interpret the 26 following as they apply to the Natural Circulation Operations: - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments I

~-.

W IE 14 Loss of Containment Integrity I 5 X Knowledge of the operational 3.3 I 27 implications of the following concepts as they apply to Loss of Containment Integrity: - Annunciators and conditions indicating signals, and remedial actions associated with the High Containment Pressure NUREG 1021 7

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 Form ES-401-2 003 Reactor Coolant X K2.01 Knowledge of bus power supplies to the I 3.1 I 28 Pump following: - RCPS 003 Reactor Coolant A4.04 Ability to manually operate and/or I 3.1 I 29 Pump monitor in the control room: - RCP seal differential I!ressure instrumentation 004 Chemical and A3.11 Ability to monitor automatic operation of I 3.6 I 30 Volume Control the CVCS, including: - Charging/letdown 004 Chemical and A2.35 Ability to (a) predict the impacts of the I 3.3 I 31 Volume Control following malfunctions or operations on the CVCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Reactor

. I Knowledge of bus power supplies to the I Ix I 005 Residual Heat

  • K2.03 I 2.7 I 32 Removal I I I I I following: - RCS pressure boundary valves 005 Residual Heat Equiptn~ntControl ;. Ability to analyze I 4.2186 Removal tbeeffec1 of maintenance activities, 'such as degnJ;ded power sourc~s, on the status oflintiting QQIlditions. for 006 Emergency Core X Knowledge of the operational I 2.8 I 33 Cooling implications of the following concepts as they apply to the ECCS: - Effects of temperatures on water level indications NUREG 1021 8

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline ES-401 . Pla~ms - Tier 21 Group 1 .. Form ES-:401-2 I KI I K2 I K3 I K4 I K5 lu!<-61ALill.J\~ Number u_ ** u_u I Svstem #lName [KIA Topic lImn ott 006Emerg~ncy.C()re Ability to.(a) predict tbe impacts ofthe 3,0 189 Cooling folIowingmaJfun9tion~ or operati<)BS on th~ECeS and(b)baseqonth~

~*p~ctioO$~ u~ procedwcs to 9()~et.

~{'\ntrol. Or niitigate the consequeQ,(Jes.of rrutlfunctions*or QP~ons:*-;

007 Pressurizer Relief/Quench Tank 007 Pressurizer x Ability to predict and/or monitor changes 35 Relief/Quench Tank in parameters (to prevent exceeding design limits) associated with operating the PRTS controls including:

Maintaining quench tank 008 Component x Knowledge ofeeWS design feature(s) 2.7 1-36 Cooling Water and/or intcrlock(s) which provide for the following: - The "standby" feature for the CeWoumos oI 0 Pressurizer Ability to (a) predict the impacts ofthe 3.9 I 37 Pressure Control following malfunctions or operations on the PZR pes and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Spray valve failures Control- Knowledge of limiting,~nd~n<lO$ fQJ:oper.¢ions and 012 Reactor Protection Knowledge of the effect of a loss or malfunction of the following will have on the RPS: - Bvoass-block circuits 012 Reactor Protection x Knowledge of the operational I 3.3 I 41 implications ofthe following concepts as to the RPS: - DNB NUREG 1021 9

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline ES-401 ... Plant~ier 2/ Group1 . _____ _

I Svstem #!Name I KI I K2 I K3 I K4 I K5 I K61A1_A~u~~~url!~er I KIA Topic i 012 Reactor Protection Ability tp(a} predict the ,itnp8Cts *of the 88 foUpwing m~nc:t;ionsor operatiol)s.on 1;hemand (b) basedonthose .

predictiolls, use pr0ce4~,to correct*,

control.:()rmitipt~ th'Q~uences of tbo~malfunetionsor operatiprui~-Faulty Knowledge of the effect that a loss or 39 malfunction of the ESF AS wi II have on the followimr: - RCS o13 EnginceredSaf~. AbiH~to (ft}pr~ict theimp~softlt~ 3.7 90 Features Actuation* following malfunctions or operationson the ESFASand(b) base<ion th{)se predictions. *U$e .procedures to*correct, control.orrnitip"" the consequences of thosemalmnctions or operations:, ..i. Loss of decontrol 022 Containment x 3.5 40 Cooling 026 Containment Emergency Procedures/Plan - Knowledge 3.8 42 Spray of general guidelines for EOP usage.

039 Main and Reheat Ability to (a) predict the impacts ofthe I 3.4 I 43 Steam following malfunctions or operations on the MRSS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

. . steam 059 Main Feedwater x Knowledge of the physical connections I 3. 1 I 44 and/or cause-effect relationships between the MFW System and the following

-RCS NUREG 1021 10

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 I

ES-401 System #lName I K 1 I K2 I K3 I K4 I K5 I K6 ~

Plant .rs-NRC Written Examination Outline Tier2! Grou:::lp=--:l=--~ ~_~~~.

A31A~~l\tumber I ~--"'---"--'- ________---'-_

~~wledge of the effect of a loss or 2~ I 46

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

061 Auxiliaryl X Emergency Feedwater malfunction of the following will have on the AFW System components:

Controllers and positi 062 AC Electrical Emergency Procedures/Plan - Knowledge I 3.4 I 45 Distribution of fire in the plant procedure.

062 AC Electrical X Knowledge of the effect that a loss or I 3.7 I 47 Distribution malfunction of the A.C. Distribution System will have on the following: - DC 063 DC Electrical X Knowledge of D.C. Electrical System I 2.9 I 48 Distribution design feature(s) and/or interlock(s) which provide for the following: -

Breaker interlocks, permissives, bypasses and cross-ties 064 Emergency Diesel Emergency Procedures/Plan Knowledge 4.2 49 Generator of RO tasks performed outside the main control room during an emergency and 1 - - - - - 4 1 u!he resultant operational effects.

064 Emergency Diesel X Ability to predict and/or monitor changes I 3.1 I 50 Generator in parameters (to prevent exceeding design limits) associated with operating the ED/G System controls including:

Maintaining minimum load on ED/G (to nr.." ..nt reverse 073 Process Radiation X Knowledge of the operational I 2.5 I 51 Monitoring implications ofthe following concepts as they apply to the PRM System:

Radiation theory, including sources,

. and effects 076 Service Water Ability to monitor automatic operation of I 3.7 I 52 the S WS, including:

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 NRC Written Examination Outline

~ ---- -- - Plant1ystejS - rier21 Group 1 ... n __ ** Form E~401-2 .

I KI I K21 K3 I K41 K5 I K6 I Al A3 A4~ Number J KIA Topic I Imp I Q#u Service Water 078 Instrument IX I I Ix lu-uu I

-.. . . . K3.02

~ ..

I Knowledge of the effect that a loss or I

malfunction of the SWS will have on the following: - Secondary closed cooling water Knowledge of lAS design feature(s)

I 2.5 I 53 I 3.2 I 54 and/or interlock(s) which provide for the following: - Cross-over to other air I s~stems._

103 Containment Ix I lu_ul I I luu. I

  • Kl.08 Knowledge of the physical connections I 3.6 I 55 and/or cause-effect relationships between the Containment System and the following systems: - SIS, including action of safety iniection reset NUREG 1021 12

PWR RO/SRO Examination Outline Facility: Indian Pont Unit 2 002 Reactor Coolant

- . . A2.02 Ability to (a) predict the impacts ofthe following malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of I 4.2 I 57 those malfunctions or operations: - Loss of coolant 011 Pressurizer Level IX I

  • K6.01 I -~"-----

Knowledge of the effect of a loss or I 2.8 I 58 Control malfunction of the following will have on the PZR LCS: - Reasons for starting charging pump while increasing letdown 014 Rod Position Indication System (RPIS)

X I I K3.02 flow rate Knowledge of the effect that a loss or malfunction of the RPIS will have on the following: - Plant computer I 2.5 I 59 015 Nuclear Instrumentation 016 Non-nuclear Instrumentation IX I I

  • .X I I

K1.01 A3.01 I Knowledge of the physical connections I

and/or cause-effect relationships between the NIS and the followin s stems: - RPS Ability to monitor automatic operation of the NN IS, including: - Automatic I 4.1 I 60 2.9 61 selection ofNNIS inputs to control 017 In-core IX I . . . K4.03 I Knowled!!e of ITM Svstem desi!ln I 3.1 I 62 Temperature Monitoring NUREG 1021 13

PWR RO/SRO Examination Outline Facility: Indian Pont 2 NRC Written Examination Outline ES-401 .. PlaniJiiifS - Tier 2 / Group 2 ._

[System #lName I K 1 K2 I I K3 I K4 I-K5li61i\TAUA.~~umber I ~- .. Ht" I m-r= I 0Temperature

. . .17.-.In.*.-c..*.o Monit(}ring

...r. . e.. . . , .'. '.. '. ,... .......... *

.1 ..**.*.l. --.......*****.1.*. . . . . .*..*.* * .1.*.*.* ...*........*.1 . . *.. . . . *. . 1.**.**...*.* .*.*. *.*.*. .*.*.1 A

tbe ~ity, fOnOWl~ mal:functloos or operations on

  • . . . ,.b.*. . *i* . . . ;. . t. . .o. (ft.)..p. . rOO.l.*c.*.t*'.*. th ITM System and(b} bas~ -on..tho~
  • . . .s. . . .o*. . *f. .,. t.*h.e.. . ,.

... e..* i. m. . .p.*ac,t. 91

, - - - *prea~eti()~s, -*~*JJfOGe4pres-~ co.Q'eC~

r ormiti~.the oons.eqJ,lences of Iftlfunctioos or.ratims:~ CON*.

93

~ SYSte:m.iri~ludi9g:.,

1 Ability to manually operate and/or 3.7 63 monitor in the control room: - Shift of S/G controls between manual and 045 Main Turbine Generator (MT/G)

System X

I I K5.23 automatic control, b~ bumQless transfer Knowledge of the operational implications of the following concepts as they apply to the MT/G System:

I 2.7 I 56 Relationship between rod control and ReS boron concentration during T/G load increases 072 Area Radiation Monitoring I I I I I IX

  • I
  • Al.OI I Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ARM system controls including:

! 3.4 I 64 075 Circulating Water NUREG 102l I IX I

- - 14

_ K2.03 I

Radiation levels I Knowledge of bus power supplies to the following: - Emergency/essential SWS QumQs I 2.6 I 65

Facility Indian Point Unit 2 Date of Exam 711212010 Category i KIA # Topic RO SRO-Onl IR Q# IR Q# i

\2.1.3 Knowledge of shift or short-term 3.7 66 relief turnover ractices.

2.1.30 Ability to locate and operate 4.4 67

, components, including lm;al I controls 2.1.42

  • Knowledge of new and spent fuel 2.5 68 i movement procedures.
1. Conduct of Operations 2.1.32 Con4uct of ()perations -Ability to .

expwn.and applyallsystemIimits' an(i.t~autions.

2.1.45 A~ilitjit0 identify ~4 intelWet diverse indicationsto.validate the res' 'nse of'another'indication'.:

Subtotal 3 2 2.2.2 Ability to manipulate the console 4.6 69 controls as required to operate the facility between shutdown and designated power levels 2.2.39 Knowledge of less than or equal to 3.9 70 one hour Technical Specification

2. Equipment action statements for syst(:ms Control 2.2.6 Subtotal 2 2 NUREG 1021 15

Facility I Indian Point Unit 2 I Date of Exam 7/1212010 Category KIA # Topic RO SRO-Only IR Q# IR Q#

2.3.4 Radiological Controls - Knowledge of 3.2 71 radiation exposure limits under normal and emergency conditions.

2.3.11 Ability to control radiation releases 3.8 72 I

i 2.3.15 Knowledge of radiation monitoring 2.9 73 i 3.

systems, such as fixed radiation Radiological monitors and alarms, portable survey Controls instruments, personnel monitoring equipment, etc.

i I 2.3.11 Ability to control radiation releases '<

4.3 98  !

.... ' .. I Subtotal 3 1 2.4.45 Ability to prioritize and interpret the 4.1 74 significance of each annunciator or alarm.

2.4.50 Ability to verifY system alarm setpoints 4.2 75 and operate controls identified in the alarm response manual 2.4.16 Knowled~eotEQP iPlplem~~ion 4.4 99 . .

4. Emergency hierarclty.and.coordinatiQllwithotber I.' I ..

Procedures/plan support proceauresotguidetmes. ~llCh  ;. ......

,....*... 1'.( ....

~~ffoperating procedures, .a5~9rmal

()~ing procedures aJ1d~ven~t.*

aqcident.o:umagement gllkielines.. t

"'I~'

I>

2.4.25 Knowledge of fire protection "3;1 --"

100 Pr0i::edures. . ....

Subtotal 2 2 Tier 3 Point Totals 10 7 NUREG to21 16

ES-401 Record of Rejected KJAs Form ES-401-4

-~--~

Tier / Randomly Selected KIA Reason for Rejection

~uQ ~-~

0000402439 Emergency Procedures/Plan - Knowledge of RO Generic KA not applicable to Steam Line R-1/1 Steam Line responsibilities in emergency plan implementation Rupture Event Rupture

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

0000542432 Emergency Procedures/Plan - Knowledge of Generic KA not applicable to Loss of MFW R-1/1 Loss of Min operator response to loss of all annunciators event Feedwater (MFW)

-~

~-~

Knowledge of the reasons for the following 00WE11K304 responses as they apply to the Loss of Emergency Loss of Coolant Recirculation: - RO or SRO function within This KA is evaluated during the R-1/1 Emergency the control room team as appropriate to the SimulatorlWalkthrough portion of the Exam Coolant assigned position, in such a way that procedures are Recirculation adhered to and the limitations in the facilities license and amendments are not violated r----~~-~ -~

~-~

0000262140 Loss of This KA rejected due to inability to write a Conduct of Operations - Knowledge of refueling R-1/1 Component discriminatory question for refueling administrative requirements.

Cooling Water administrative requirements for Loss of CCW.

Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite This KA was rejected due to overlap with R-1/1 000056K103 Power: - Definition of subcooling: use of steam Question 10.

tables to determine it

~-~

0000052449 Emergency Procedures/Plan - Ability to perform Generic KA not applicable to without reference to procedures those actions that Inoperable/Stuck Control Rod. There are no R-1/2 Inoperable/Stuck require immediate operation of system components procedures with immediate operator actions Control Rod and controls related to this condition.

000037AA101 Ability to operate and/or monitor the following as This KA was rejected because, at IP2 for a they apply to the Steam Generator Tube Leak: Steam Generator Tube Leak, a normal R-1/2 Steam Generator Maximum controlled depressurization rate for cooldown and depressurization is performed Tube Leak affected S/G NOT a maximum rate depressurization.

This KA was rejected because, at IP2 for a Ability to operate and/or monitor the following as Control Room Evacuation, the charging pump R-1/2 000068AA111 they apply to the Control Room Evacuation: is aligned to the RWST. Emergency Boration Emergency borate valve controls and indicators is not identified in the procedure 2-AOP-SSD 1.

012000K608 Knowledge of the effect of a loss or malfunction of R-2/1 Reactor Equipment (COLSS) not applicable to IPEC.

the following will have on the RPS: COLSS Protection System 006000K509 Knowledge of the operational implications of the following concepts as they apply to the ECCS: This KA was rejected due to overlap with R-211 Emergency Core Thermodynamics of water and steam, including question 2.

Cooling System subcooled margin, superheat, and saturation Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) This KA was rejected due to overlap with R-2/1 026000A102 associated with operating the CSS controls question 33.

including: - Containment temperature This KA was rejected because Chemistry I 0590002134 Conduct of Operations - Knowledge of primary and limits are no longer in any operations R-2/1 Main Feedwater secondary plant chemistry limits. procedure. Unable to write a discriminatory RO level question for this KA.

I 2

This KA is rejected because it is a Generic Knowledge of the operational implications of the Fundamentals concept with limited impact on following concepts as they apply to the CROS: - the Control Rod Drive System. The change in R-2/2 001000K520 Effects of RCS temperature on boron reactivity temperature will have an impact on the worth CROS. Unable to write a discriminatory RO level question for this KA.

0290002126 I Conduct of Operations - Knowledge of industrial Containment safety procedures (such as rotating equipment, Generic KA not applicable to Containment R-2/2 Purge System electrical, high temperature, high pressure, caustic, Purge System (CPS) chlorine, oxygen and hydrogen).

Generic KA not applicable to written Conduct of Operations - Ability to make accurate, 3 2.1.17 examinations. This KA is evaluated during clear and concise verbal reports.

Simulator Evaluation.

~.

0000082206  !

Pressurizer (PZR) Equipment Control - Knowledge of the process for Generic KA not applicable to Emergency S-1/1 Vapor Space making changes to procedures. Plant Evolutions Accident (Relief Valve stuck open) 0000512225 Equipment Control - Knowledge of bases in Loss of Generic KA not applicable to Loss of S-1/2 technical specifications fo rlimitint conditions for Condenser Condenser Vacuum.

operations and safety limits Vacuum 0260002401 Generic KA not applicable to Containment Containment Emergency Procedures/Plan - Knowledge of EOP S-2/1 Spray. There are no Immediate Operator Spray System entry conditions and immediate actions steps Actions for the Containment Spray System.

(CSS)

. -~ ....... -.-~ ........... __ __ _._ __ _ - - -

3

-~

r-~~

061000A206 Ability to (a) predict the impacts of the following malfunctions or operations on the AFW System and This KA was rejected because IPEC has no Auxiliary 1 S-2/1 (b) based on those predictions, use procedures to procedure to address this condition. Unable to Emergency correct, control, or mitigate the consequences of write a discriminatory SRO level question Feedwater (AFW) those malfunctions or operations: - Back leakage of System

'--~

MFW ~-

0140002113 Rod Position Conduct of Operations - Knowledge of facility Generic KA not applicable to Rod Position S-2/2 Indication System requirements for controlling vital/controlled access. Indication System.

(RPIS)

'---~

0270002235 Containment Equipment Control- Ability to determine Technical Generic KA unable to write a valid SRO Only S-2/2 Iodine Removal Specification Mode of Operation. question.

System (CIRS)

-~

-~

Conduct of Operations - Knowledge of criteria or Generic KA not applicable to written conditions that require plant-wide announcements, 3 2.1.14 examinations. This KA is evaluated during such as pump starts, reactor trips, mode changes, Simulator Evaluation.

etc.

r---~

c- -

-~ -

~~ ~-

-~ ~-~

~~~~

-~

~~

4

I I

i i i i

Facility: Indian Point 2 Scenario No.: 1 Op-Test No.: _1 Examiners: Operators:

Initial Conditions:

Reset simulator to IC-114 Load Simulator Schedule-Scenari01 The Plant is at 16% power. 23 EDG is OOS due to a malfunctioning governor.

Turnover:

Return plant to 100% power.

Event Malf. No. Event Event No. Type'" Description R (ATC) 1 N/A N (CRS) Power Escalation N (BOP) 2 XMT I(ATC)

SGN008A I(BOP) 23 SG Controlling Steam Flow Transmitter Fails High I TS(CRS) 3 MAL C (BOP)

EPSOO8L C (CRS) MCC-28 will trip on overcurrent.

TS (CRS) 4 MAL- Loss of offsite power due to Loss of Station Aux Transformer.

C (ALL)

EPSOO1 The running charging pump (21) will trip.

5 MAL M(ALL) Complete loss of off site power resulting in a Reactor Trip SWDOO3A 6 MAL DSGOO7A M(ALL) 2'1 EDG will trip and team will enter ECA-O.O.

7 N/A 22 EDG will be repaired and started. Team must start a C(BOP) service water pump to cool the EDG before it overheats and tips.

8 N/A C{ATC) Prior to starting a charging pump, RCP Seal Injection must be C(CRS) isolated.

  • {N)ormal, (R)eactivity, (l)nstrument,(C)omponent. (M)ajor U2 NRC 2012 Scenario I: Power Escalation from 15%, FT-4398 fails high, MCC-28 trips, Loss of offsite power, Loss of all EDG to ECA-O.O.

Page 1 of 19

_A---L-pL..pe-"-n_dc-i_x_D_ _____R_e_9.>-u_i_re_d_O--'perator Actions Form ES-D-2 Session Outline:

The evaluation begins with the plant at 15% power steady state operation. 23 EDG is out of service due to a malfunctioning governor. Post maintenance testing will be performed this shift.

After completion of testing, the diesel will be declared operable. The team is currently raising power.

After taking the watch, 23 SG Steam Flow Transmitter fails high. The team will take actions in accordance with 2-AOP-INST -1, Instrument and Controller Failures.

After failed channel is removed from service, MCC-28 will trip on overcurrent. The team will need to restart Control Rod Drive Cooling Fans on MCC-28A.

Following restart on the CRD cooling fans, a loss of offsite power will occur due to a widespread blackout. The running charging pump (21) will trip and EDG 22 will fail to start.

About two minutes after the loss of offsite power, a loss of grid will occur resulting in a unit trip.

The team will perform actions ofE-O, Reactor Trip or Safety Injection. Only Bus SA will be energized. Because both motor driven AFW Pumps are de-energized, the team will take manual action to align AFW now from the turbine driven AFW pump.

Soon after AFW flow is established, 21 EDG will trip. The team will recognize a loss of all AC and enter EOP ECA-O.O, Loss of All AC Power. After equipment is placed in pullout per ECA 0.0,22 EDG will be repaired and started. However, 22 Service Water Pump will not auto start.

The team will manually start the Service Water pump to provide cooling to the EDG before the diesel overheats and trips. The team will proceed through ECA-O.O and transition to 1) ECA-O.l, Loss of All AC Recovery without SI Required, and then transition to ES-O.2, Natural Circulation Cool down OR 2) ECA-O.2, Loss of All AC Recovery with SI Required, and then to E-l, Loss of Reactor or Secondary Coolant. The scenario will be terminated after transition to ES-0.2, E-l, or at the lead evaluator's discretion.

Procedure flow path: AOP-INST-I, 2-AOP-138 KV-IECA-O.O, ECA-O.l or ECA-0.2, ES-O.2 E-l U2 NRC 2012 Scenario I: Power Escalation from 15%, FT -4398 fails high, MCC-28 trips, Loss of ofJsite power, Loss of all EDG to ECA-O.O.

Page 2 of 19

Facility: Indian Point 2 Scenario No.: _2_ Op-Test No.: _1 Examiners: Operators:

Initial Conditions:

The Plant is in a 100% normal full power lineup.

Turnover:

21 Charging Pump and 21 CCW pump are out of service. The team will assume the shift and begin a rapid shutdown in accordance with AOP-RSO-1.

I Perform a Rapid Plant Shutdown in accordance with 2-AOP-RSO-1 due to excessive packing leak on Event Malf. No. Event Event No. Type* Description 1 XMT , (ALL)

TS (CRS) RCS Loop 23 T-Hot fails high I RCS043A 2 N/A R (ATC)

N (CRS) Rapid Load Shutdown N (BOP) 3 MAL C (ATC)

CRF002AV C (CRS) Control Rod P-6 "ratchets in" during rod motion.

TSJCRS) 4 MOT C (ATC)

CVC004A C (CRS) 22 Charging Pump trips.

TS (CRS) 5 MAL M (ALL)

Steam Break down stream of 21 MSIV & Check Valve in Aus SGN004A I Boiler Feed Pump Building.

6 RLY PPL487 C (CRS)

C (BOP) Safety Injection fails to Auto Actuate requiring Manual Actuation.

RLY PPL488 7 MOV RHR011 C (CRS)

C (BOP) RHR valve 746 will fail to auto open requiring Manual Action 8 PLP RHR033 M (ALL) LOCA outside Containment in Primary Auxiliary Building (PAB)

PLP RHR022

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor U2 NRC 2012 Scenario 2: Dropped Rod, Charging Pump trip, Steam Break no Auto SI, LOCA outside Containment Page I of22

Session Outline:

The evaluation begins with the plant at 100% power steady state operation.

21 Charging Pump and 21 CCW pump are out of service.

Shortly after the crew assumes the watch 23 Loop T-hot instrument fails high. The team will take actions in accordance with 2-AOP-INST-1, Instrument and Controller Failures.

The Shift Manager will call the control room and report that 24 SG Feed Regulating Valve air line has been damaged by a maintenance crew. The shift manager will direct the team to begin a rapid shutdown in accordance with 2-AOP-RSD-1.

After adequate power reduction has taken place, Control Rod P-6 will ratchet into the core due to a movable gripper failure, requiring the load reduction to be stopped and the condition evaluated per 2-AOP-ROD-1.

After the load reduction is resumed, 22 Charging pump will trip. The crew will respond using 2-AOP-CVCS-1 and isolate letdown. Chalrging and letdown then will be re established.

Subsequently a Main Steam Rupture will occur downstream of 21 MSIV and check valve. The team may use 2-AOP-UC-1 to trip the reactor and close the MSIVs.

Simultaneously with the reactor trip, a rupture will occur on the RHR discharge header outside of containment. Automatic SI will fail to actuate when demanded, requiring manual actuation.

The team will progress through E-O and may determine that RCP trip criteria is met.

(This depends on the magnitude of the cooldown during the steam break.) The BOP will manually open MOV-746 while performing E-O Attachment 1. the team will continue in E-O until a transition to ECA-1.2 is directed. In ECA-1.2, the source of the LOCA outside containment will be identified and isolated. The scenario is terminated when the team has determined a transition to E-1 is required.

Procedural flow path: 2- AOP-INST-1, 2-AOP-RSD-1, 2-AOP-ROD-1, 2-AOP-CVCS-1, 2-AOP-UC-1, 2-E-0, 2-ECA-1.2, 2-E-1.

U2 NRC 2012 Scenario 2: Dropped Rod, Charging Pump trip, Steam Break no Auto S[, LOCA outside Containment Page 2 of22

Facility: Indian Point 2 Scenario No.: 3 Op-Test No.:_1 Examiners: Operators:

Initial Conditions:

Reset simulator to IC-115 Load Simulator Schedule-Scenari04 The Plant is at 30% power. 21 EDG is OOS for major PM.

Turnover:

Return plant to 100% power.

Event Malf. No. Event Event No. Type* Description 1 N/A R (ATC)

N (CRS) Power Escalation N (BOP) 2 XMT PT-419C (31 SG C Channel Pressure) fails high causing 21 I (ALL)

SGN037A ADV to fail open requiring manual closure.

TS (CRS) 3 MOC C(CRS)

CCW003A C (BOP) 23 CCW Pump trips and 21 and 22 CCW Pump fail to auto start MOC TS (CRS)

  • CCW001/2 4 MAL C (ALL) 35 gpm RCS leak.

RCS002A TS (CRS) 5 MAL M (ALL) Large Breal< RCS LOCA RCS001A 6 MOC C (CRS)

C (BOP) RHR pumps will not auto start and need to be started manually.

RHROO3/4 7 RLY PPL085/09 C (BOP) Failure of Containment Phase A requiring manual initiation.

0

  • (N)ormat, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor U2 NRC 2010 Scenario 3: Power Escalation from 30%, SG Pressure Failure causing ADV to open, CCW Pump trip with auto start failure, RCS leakage, LBLOCA with multiple actuation failures.

Page I of 18

Session Outline:

The evaluation begins with the plant at 30% power steady state operation. The team is instructed to raise power to return to full load. The following equipment is out of service:

  • 21 EDG is out of service for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for malfunctioning governor. Maintenance is in progress with expected return to service this shift.

After starting the power ascension, a SG pressure channel (PT -419C) fails high. The team will take actions in accordance with 2-AOP-INST-1, Instrument/Controller Failures.

Prior to completion of the Subsequent Actions of 2-AOP-INST-1, 23 CCW Pump will trip. 21 and 22 CCW Pumps will fail to auto start requiring the BOP to manually start 21 or 22 CCW Pump before RCP trip is required.

After CCW is restored, a 35 gpm RCS leak will occur. The crew will diagnose RCS leakage, quantify the leak rate and take actions per 2-AOP-LEAK-1.

Large Break LOCA will occur. The team will perform actions of E-O, Reactor Trip or Safety Injection. Both RHR Pumps will fail to Auto start and must be started manually.

Containment Isolation Phase A will fail to auto actuate requiring manual actuation by the operator. Fan Cooler Units 23 and 25 will trip due to bearing failures and will remain out of service for the remainder of the scenario. The team will subsequently transition to E 1, Loss of Reactor or Secondary Coolant.

When RWST level decreases to 9.24 feet, the team will transition to ES-1.3, Transfer to Cold Leg Recirculation. The team will take the appropriate action to place a train of recirculation in service. The scenario is terminated when recirculation is established to one train and SI pumps are secured.

Procedure flow path: 2-AOP-INST-1, 2-AOP-CCW-1, 2-AOP-LEAK-1, 2-E-0, 2-E-1, 2 ES-1.3 U2 NRC 20 I 0 Scenario 3: Power Escalation from 30%, SG Pressure Failure causing ADV to open, CCW Pump trip with auto start failure, Res leakage, LBLOCA with multiple actuation failures.

Page 2 of 18

Facility: Indian Point 2 Scenario No.: _4_ Op-Test No.: _1 Examiners: Operators:

Initial Conditions:

The Plant is in a 100% normal full power lineup.

Turnover:

Event Malf. No. Event Event No. Type* Description 1 XMT I (ALL)

VCT Level Transmitter LT-112 fails low CVC019A 2 MAL C (ALL) 6 gpm SG Tube Leak 24 SG RCS014D TS (CRS)

NA R (ATC) 3 N (CRS) Rapid Load Reduction/Shutdown N (SOP) 4 MAL M (ALL) Steam Generator Tube Rupture RCS014D 5 MAL Station Auxiliary Transformer Fault resulting in a loss of off-site C (ALL)

EPSOO1 power on reactor trip 6 MAL C (ALL) Sus 6A fault after Safety Injection EPS007D 7 MAL C (SOP) Safety Injection Pump 21 Fails to Auto Start SIS001 8 AOV RCSOO3A SWI C (CRS)

PORV 456 loss of control power when attempted to open RCSOO6S C (ATC)

SWI RCSOO6C 9 AOV C (CRS) Auxiliary Spray Valve 212 fails to open resulting in a transition to CVCOO8A C (SOP) ECA-3.3 SGTR With Loss of Pressure Control

  • (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor U2 NRC 2012 Scenario 4: LT-112 failure, SGTL, SGTR, Loss off site power, bus 6A fault, Loss of Pressure Control Page I of22

Session Outline:

The evaluation begins with the plant at 100% power steady state operation.

21 AFW Pump of out of service for scheduled maintenance and is expected back within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Shortly after the crew takes the shift, VCT level transmitter 112 will fail low. The crew will take actions in accordance with 2-AOP-CVCS-2 to restore a normal charging lineup.

VCT level will be maintained by maintaining VeT pressure above the pre-failed value.

A 6 gpm Steam Generator Tube leak will occur in 24 SG. The team will take actions in accordance with 2-AOP-SG-1 and determine that a shutdown must commence per TS 3.4.13. After the magnitude of the leak is determined, the crew will initiate a power reduction using either 2-AOP-RLR-1 or 2-AOP-RSD-1. (The crew may determine that the leakrate will not be adequately reduced at 50% power and perform 2-AOP-RSD-1 to shutdown the unit.)

During the power reduction the tube leak will increase to a Steam Generator Tube Rupture requiring a Reactor Trip and Safety Injection. When the Main Generator output breakers open, the Station Auxiliary Transformer will fault resulting in a loss of offsite power. Approximately 45 seconds after Safety Injection is actuated, bus 6A will fault.

With 6A faulted and 21 AFW pump out of service, 22 Auxiliary Boiler Feed Pump will have to be manually aligned to supply water to the SGs. 21 Safety Injection Pump will fail to auto start and must be manually started.

The team will transition to E-3. Pressurizer Spray will not be available due to loss of Reps. PORV 456 control power will fail when the valve is placed to open. Auxiliary Spray Valve AOV-212 will not open when the crew attempts to align Aux Spray. The crew will transition to ECA-3.3. The scenario is terminated when SI pumps are secured.

Procedure Flow Path: 2-AOP-CVCS-1, 2-AOP-SG-1, 2-AOP-RSD-1, E-O, E-3, ECA-3.3.

U2 NRC 2012 Scenario 4: LT-112 failure, SGTL, SGTR, Loss off site power, bus 6A fault, Loss of Pressure Control Page 2 of22

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