ML12107A160

Nine Mile Point Unit 2 - Final Outlines (Folder 3)

ML12107A160
Person / Time
Site:	Nine Mile Point
Issue date:	04/12/2012
From:	D'Antonio J M Operations Branch I
To:	Constellation Energy Nuclear Group
Jackson D E
Shared Package
ML110030686	List: IR 05000410/2012301 ML111820284 ML113260312 ML120760402 ML12102A100 ML12102A102 ML12102A114 ML12102A124 ML12107A160 ML12107A166 ML12107A169 ML12117A007
References
TAC U01841, 50-410/12-301, ES-401, ES-401-1
Download: ML12107A160 (35)

Text

BWR Examination Outline Form ES-401-1 Facility:

NMP Unit 2 NRC 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 #1 does not apply). Use duplicate pages for RO and SRO-only exams. Date of Exam: SRO-Only Points Tier Group K K A2 G* 1 2 1. Emergency 1 4 3 4 20 4 & 2 2 1 1 7 2 Plant Evaluations Tier 6 4 5 27 5 5 Totals 1 2 2 2 26 2 3 2. 2 1 1 12 o I 1 2 Plant Systems Tier 3 3 3 3 3 3 4 4 5 4 38 3 5 Totals 3. Generic Knowledge

& Abilities i 2 3 4 10 1 2 3 Note 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the 2 3 2 3 2 2 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. 4 2 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 KAs having an importance rating (IR) of 2.5 or higher shall be selected.

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

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 KIA's Total 7 3 10 5 3 8 7 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

__________________________________________

__

i ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 EAPE#lName Safety Function 295028 High Drywell Temperature / 5 295005 Main Turbine Generator Trip / 3 700000 Generator Voltage i . and Electric Grid Disturbances

.295003 Partial or Complete Loss of AC / 6 295037 SCRAM Conditions Present and Reactor Power Above APRM Downscale or Unknown / 1 295016 Control Room Abandonment I 7 295025 High Reactor Pressure /3 . KiA Topic(s) EA2.04 -Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:

Drywell AA2.05 -Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP: Reactor AA2.09 -Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

Operational status of emergency diesel nerators.

2.4.41 -Emergency Procedures

/ Plan: Knowledge of the emergency action level thresholds and classifications.

2.4.18 -Emergency Procedures

/ Plan: Knowledge of the specific bases for EOPs. 2.4.11 -Emergency Procedures

/ Plan: Knowledge of abnormal condition ures. EA2.04 -Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE:

level 4.2 76 i ! 77 3.9 4.3 78 79 4.6 : 4.0 80 4.2 81 3.9 82 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 EK 1 .06 -Knowledge of the"1, operational implications of I .. :., the following concepts as I " they apply to SCRAM 29S037 SCRAM Conditions

',' CONDITION PRESENT Present and Reactor Power i .....*.. 4.0 39 X AND REACTOR POWER Above APRM Downscale or , " ABOVEAPRM Unknown 11 '. DOWNSCALE OR ,'.; UNKNOWN: Cooldown effects on reactor power I AK 1.01 -Knowledge of the operational implications of ,c':, J.:. the following concepts as :::.29S001 Partial or Complete '.' they apply to PARTIAL OR Xi 3.S i 40*Loss of Forced Core Flow COMPLETE LOSS OF Circulation 11 &4 FORCED CORE FLOW CIRCULATION:

Natural '.,

'< circulation

.........,AK1.03 -Knowledge of operational implications 29S006 SCRAM I 41 they apply to SCRAM: the following concepts as 3.7 X ',. :: Reactivity control .' AK2.06 -Knowledge of the 1 interrelations between 700000 Generator Voltage GENERATOR VOLTAGE 42 3.9 and Electric Grid X f>. AND ELECTRIC GRID Disturbances DISTURBANCES and the following:

Reactor power. I I AK2.02 -Knowledge of interrelations 29S018 Partial or ":'. PARTIAL OR COMPLETE .... 3.4 ,43 ,X ," iLoss of CCW 18 , . LOSS OF COMPONENT COOLING WATER and the following:

Plant qperations I I I EK2.02 -Knowledge of the ! interrelations between SUPPRESSION POOL *29S026 Suppression Pool I HIGH WATER

• 3.6 44 X High Water Temp. IS TEMPERATURE and the I following:

Suppression pool i Plant

.....I

1.01 -Ability to operate ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 [ EAPE#fName Safety Function 1295016 Control Room . Abandonment / 7 ! .295025 High Reactor Pressure /3 Xi AK3.03 -Knowledge of the reasons for the following responses as they apply to CONTROL ROOM ABANDONMENT:

Disabling control room controls EK3.08 -Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE:

Reactor/turbine pressure 3.5 45 3.5 46 re ulating system 0=. f--

..Knowledge of the AA 1.03 -Ability to operate and/or monitor the following as they apply to LOSS OF SHUTDOWN COOLING: Component cooling water s stems: Plant-Specific AA 1.02 -Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Systems necessary to assure safe lant shutdown EA2.01 -Ability to determine and/or interpret the following.

as they apply to REACTOR LOW WATER LEVEL: Reactor water level 295019 Partial or Complete Loss of Inst. Air / 8 295028 High Drywell Temperature / 5 .295021 Loss of Shutdown Cooling /4 295004 Partial or Complete Loss of DC Pwr I 6 295031 Reactor Low Water Level 12 and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE:

Drywell Mark-I&II

____+-_-t---i reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Service air isolations:

Plant-SRecific 3.2 47 3.8 48 3.1 49 i 3.8 4.6 51* 50 ES-401 BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions

-Tier 1 Group 1 AA2.01 -Ability to determine I I I I andlor interpret the following

• .295023 Refueling X as they apply to 3.6 52 [tidents 18 REFUELING ACCIDENTS:

I Area radiation levels I EA2.04 -Ability to determine I 295038 Off-site andlor interpret the following X as they apply to HIGH .4.1

• 53 Release Rate I 9 SITE RELEASE RATE:

• Source of off-site release 2.4.46 -Emergency I I I Procedures I Plan: Ability to 295030 Low Suppression X verify that the alarms are 14.2 54.. Pool Water Levell 5 1 I consistent with the plant I conditions.

I 2.4.31 -Emergency I Procedures I Plan: 295005 Main Turbine X Knowledge of annunciator 4.2 55* Generator Trip 13 . alarms, indications, or response procedures i I I 2.1.23 -Conduct of Operations:

Ability to 295024 High Drywell X perform specific system and ! 4.3 56 Pressure 15 integrated plant procedures

.. during all modes of plant operation.

_ ! AK3.04 -Knowledge of the reasons for the following 600000 Plant Fire On-site /

• responses as they apply to 2.8 57 X* . PLANT FIRE ON SITE: 8 Actions contained in the I abnormal procedure for plant I fire on site j I AK 1.0 1 -Knowledge of the i operational implications of 1295003 Partial or Complete the following concepts as X they apply to PARTIAL OR 2.7 .58 Loss of AC 16 COMPLETE LOSS OF A.C. I i POWER: Effect of battery i dischargerate on capacity 2017 I I 4 ! Group Point Total: KJA Category Totals 3

• 4 3 314 3/3 ES-401 Form ES-401-1 Written Examination Outline Emergency and Abnormal Plant Evolutions

-Tier 1 Group 2 EAPE#fName Safety Function KIA Topic(s) Steam flow/feedflow mismatch 2.1.20 -Conduct of 295020 Inadvertent Cont. Operations:

Ability to 4.6 84* Isolation I 5 &7 interpret and execute re 1295009 Low Reactor Water 2.1.23 -Conduct of Operations:

Ability to perform specific system and 4.4 85 Level 12 integrated plant procedures during all modes of plant AK1.01 -Knowledge of the operational implications of 295022 Loss of CRD X the following concepts as 3.3 59 Pumps 11 they apply to LOSS OF CRD PUMPS: Reactor pressure vs. rod insertion ca AK2.01 -Knowledge of the 295012 High Drywell interrelations between HIGH X DRYWELL TEMPERATURE 3.4 60 Temperature 15 and the following:

Drywell ventilation AK3.02, Knowledge of the reasons for the following 295009 Low Reactor Water responses as they apply to I 2.7 Level 12 i X! LOW REACTOR WATER 61 LEVEL: Reactor feedpump runout flow control: Plant-EA 1.01 -Ability to operate and/or monitor the following as they apply to 295035 Secondary SECONDARY Containment High CONTAINMENT HIGH 3.6 62 Differential Pressure I 5 DIFFERENTIAL PRESSURE:

Secondary containment ventilation AA2.02 -Ability to determine and/or interpret the following 295008 High Reactor as they apply to HIGH Water Levell 2 REACTOR WATER LEVEL:

ES-401 Form ES-401-1 Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function 295033 High Secondary Containment Area Radiation Levels I 9 295008 High Reactor Water Levell 2 295002 Loss of Main Condenser Vac I 3 x . EA2.01 -Ability to determine and/or interpret the following as they apply to HIGH SECONDARY 3.8 63* CONTAINMENT AREA RADIATION LEVELS: Area radiation levels 2.4.34 -Emergency Procedures I Plan: Knowledge of RO tasks performed outside the main control room during an emergency and the resultant o erational effects. AK 1.03 -Knowledge of the operational implications of the following concepts as they apply to LOSS OF MAIN CONDENSER

.4.2 64 13.6 i 65

• VACUUM: Loss of heat sink -----------------+---+---+--+---+---.-"'iI--+...,....\----

KiA CategoryTotals Group Point Total: L-..__________________

-'-------'--

___

_________________'--___-"

ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 System #/Name 209001 LPCS 261000 SGTS KIA Topic(s) A2.04 -Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

D.C. failures A2.05 -Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or o erations:

Fantri s 2,1,7, Conduct of Operations:

Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation, 2,2.44 -Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives effect plant and system conditions, 3,0 86 3,1 87 4,7 88 4.4 89 209002 HPCS 262001 AC Electrical Distribution ES-401 Form ES-401-1 Written Examination Plant Systems -Tier 2 Group System #/Name 239002 SRVs 400000 Component X Cooling Water 2150031RM X 203000 RHR/LPCI:

Injection Mode 262001 AC Electrical Distribution 211000 SLC 223002 PCIS/Nuclear Steam Supply Shutoff X X X X KIA Topic(s) 2.4.9 -Emergency Procedures

/ Plan: Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) m" ation ies. K1.01 -Knowledge of the physical connections and / or cause-effect relationships between CCWS and the following:

Service water K1.05 -Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (lRM) SYSTEM and the following:

Display control :P K2.01 -Knowledge of electrical power supplies to the following:

Off-site sources of K3.01, Ability to shutdown the reactor in certain conditions K3.20 -Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY OFF will have on following:

Standby gas treatment 4.2 90 3.2 1 3.3 2 2.7 3 3.3 4 4.3 5 3.3 6 ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) System #fName ___----l_--L..----II 264000 EDGs _ ........ 217000 RCIC 205000 Shutdown Cooling I i I I 209001 LPCS I I 300000 Instrument Air I K4.02 -Knowledge of I , EMERGENCY GENERATORS t:' (DIESEUJET) design X . , feature(s) and/or 4.0 7 interlocks which provide I I for the following: . '. Emergency generator , trips (emerQency/LOCA) 17 K4.06 -Knowledge of '.<.y REACTOR CORE .,. ISOLATION COOLING J. SYSTEM (RCIC) design X V feature(s) and/or 3.5 8 I*. interlocks which provide for the following:

Manual . .',' initiation K5.02 -Knowledge of the operational

< implications of the following concepts as X they apply to 2.8 9 SHUTDOWN COOLING .... .' SYSTEM (RHR SHUTDOWN COOLING MODE): Valve operation I,.. K5.05 -Knowledge of I the operational ... : implications of the X } following concepts as 2.5 10 they apply to LOW PRESSURE CORE SPRAY SYSTEM: ... System ventinQ K6.13 -Knowledge of the effect that a loss or X malfunction of the 2.8 11 following will have on the I I I INSTRUMENT AIR I SYSTEM: Filters ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) System #/Name 263000 DC Electrical Distribution 261000 SGTS 262002 UPS (AC/DC) 209002 HPCS 1":,,,1;,:

X ;,;,, " ;, X I:: " I;' ',*. I X ,,;.', X lr K6.02 -Knowledge of rr the effect that a loss or malfunction of the following will have on the D.C. ELECTRICAL DISTRIBUTION:

Battery "i";;" ventilation A 1.02 -Ability to predict , and/or monitor changes in parameters , associated with operating the STANDBY GAS TREATMENT SYSTEM controls including:

Primary containment pressure A3.01, Ability to monitor automatic operations of ' the UNINTERRUPTABLE POWER SUPPLY II. (A.C.lD.C.)

including:

I, Transfer from preferred to alternate source. [,1/", A2.11 -Ability to (a) predict the impacts of the following on the " HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) ; and (b) based on those predictions, use " procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low suppression pool level: BWR-5,6 2.5 3.1 2.8 3.3 12 13 14 15 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 System KIA Topic(s) A2.05 -Ability to (a)

• I predict the impacts of .' the following on the RELIEF/SAFETY VALVES; and (b) based on those predictions, use 3.2 16 X 239002 SRVs procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Low reactor ':

A3.02 -Ability to monitor 'J automatic operations of '; the SOURCE RANGE 215004 Source MONITOR (SRM) 3.4 17 X Range Monitor SYSTEM including:

Annunciator and alarm '. signals'.'"'. A3.01 -Ability to monitor automatic operations of the AUTOMATIC 4.2 18 ,," X 218000 ADS DEPRESSURIZATION SYSTEM including:

ADS valve operation

", A4.03 -Ability to manually operate and/or 215005 APRM / monitor in the control 3.2 19 X room: APRM back panel LPRM I' switches, meters and indicating lights A4.08 -Ability to manually operate and/or monitor in the control 3.4 20 212000 RPS X room: Individual system relay status: Plant-Specific 2.4.4, Ability to recognize abnormal indications for system I' 259002 Reactor operating parameters 4.5 21 X " that are entry level conditions for emergency and abnormal procedures.

Water Level Control

• ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) 2.4.49 -Emergency Procedures

/ Plan: ;" Ability to perform without reference to procedures " 262002 UPS those actions that 4.6 22 X (AC/DC) require immediate 1:\ ::;"' operation of system components and controls.

A2.04 -Ability to (a) predict the impacts of" the following on the "; ." SOURCE RANGE !: MONITOR (SRM) SYSTEM; and (b) based 215004 Source on those predictions, use 3.5 23 Range Monitor procedures to correct, ".' control, or mitigate the ':'; consequences of those abnormal conditions or ,::" operations:

Up scale and "; ';;Ii downscale trips A 1.02 -Ability to predict and/or monitor changes in parameters associated with ,:.' operating the ,", 24 INTERMEDIATE

3.7 2150031RM

X " RANGE MONITOR '" (lRM) SYSTEM controls including:

Reactor power indication response to rod position changes A4.05 -Ability to manually operate and/or 239002 SRVs X 4.3 25 monitor in the control room: Reactor pressure ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 1 211000 SLC KIA Category Totals 2 2

• 2.1.7 -Ability to evaluate plant performance and make X operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. 2 2 2 2 2 3fi 3 3 313 Group Point Total: 4.4 26 , I 26/5 I ES-401 Form ES-401-1 Written Examination Plant Systems -Tier 2 Group System #/Name 201006 RWM 202001 Recirculation 290002 Reactor Vessel Internals 201003 Control Rod and Drive Mechanism x 215002 RBM x KiA Topic(s) A2.04 -Ability to (a) predict the impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWH) (PLANT SPECIFIC)

and (b) based on those predictions, use procedures to correct, contro I, or mitigate the quences of those mal conditions or tions

Stuck rod: c(Not-BWR6)

-Emergency dures / Plan: ledge of EOP tioQ§)trategies.

-Emergency dures / Plan; ledge of events d to system tion / status that e reported to I organizations or al agencies, such state, the NRC, or nsmission system operat or. K1.02 the ph -Knowledge of ysical connections

• 4.7 92 4.1 93 and/or cause-effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following:

Reactor water K2.03 -Knowledge of . electrical power supplies to the following:

APRM channels:

BWR-3,4,5 3.3 2.9 27 2.8 28 91 ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 2 System #/Name KJA TOPiC(S_)__---LI_'m_p:li!J 230000 RHR/LPCI:

Torus/Pool Spray X ,Mode 201001 CRD Hydraulic 290002 Reactor !Vessellnternals

! 219000 RHR/LPCI:

Torus/Pool Cooling Mode I I 233000 Fuel Pool i Cooling/Cleanup

_ .... I I L K3.01 -Knowledge of I the effect that a loss or malfunction of the '. "c. RHR/LPCI:<.... : 'J' TORUS/SUPPRESSION 3.7 29 I' POOL SPRAY MODE .. will have on following:

Suppression chamber !pressure K4.10 -Knowledge of CONTROL ROD DRIVE HYDRAULIC SYSTEM design feature(s) and/or X i interlocks which provide 3.1 30 for the following:

Control of rod movement (HCU directional control .. valves) i K5.01 -Knowledge of the operational

'. implications of the X :.: following concepts as 3.5 31 they apply to REACTOR VESSEL INTERNALS:

.. Thermal limits K6.08 -Knowledge of the effect that a loss or malfunction of the following will have on the X RHR/LPCI:

2.7 32 TORUS/SUPPRESSION POOL COOLING MODE: ECCS room cooling ". A 1.06 -Ability to predict and/or monitor changes in parameters X associated with 2.5 : 33 operating the FUEL . POOL COOLING AND I CLEAN-UP controls i I including:

System flow 1 ES-401 Form ES-401-1 Written Examination Outline Plant Systems -Tier 2 Group 2 System #fName KIA Topic(s) 286000 Fire Protection 216000 Nuclear Boiler Inst. 204000 RWCU 256000 Reactor Condensate 215001 Traversing In-core Probe KIA Category Totals 1 1 1 I I A2.08 -Ability to (a) predict the impacts of the following on the FIRE PROTECTION 0f' SYSTEM; and (b) based , on those predictions, use X procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Failure to , actuate when required "/ ' C": A3.01 -Ability to monitor -automatic operations of .i::. the NUCLEAR BOILER :. Instrumentation X including:

Relationship between meter/recorder readings and actual parameter values: . Plant-Specific

"' " A4.08 -Ability to manually operate and/or X monitor in the control ,"' room: Reactor water level 2.1.7 -Conduct of Operations:

Ability to evaluate plant performance and make X operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

A4.03 -Ability to manually operate and/or X monitor in the control room: Isolation valves: Mark-I&II(Not-BWR1 ) 1 1 1 1 1/1 1 2 112 i Group Point Total: 3.2 34 3.4 35 3.4 36 4.4 37 3.0 38 I I 12f3 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility:

Date: Category KA# Topic RO SRO-Only IR Q# IR Q# i 2.1.42 Knowledge of new and spent fuel 2.5 66 movement procedures. . Knowledge of individual licensed ! I operator responsibilities related to shift 2.1.4 staffing, such as medical requirements, 3.3 67 I"no-solo" operation, maintenance of active license status, 10CFR55, etc. 1. Conduct of Operations Knowledge of facility requirements for ".113 controlling vitali controlled access. 3.2 94 Knowledge of criteria or conditions that 2.1.14 require plant-wide announcements, 3.1 99 such as pump starts, reactor trips, mode changes, etc. Subtotal 2 2 2.2.7 I Knowledge of the process for 2.9 68 conducting special or infrequent tests. 2.2.40 Ability to apply technical speCifications 3.4 69 for a system. 2.2.37 iAbility to determine operability and I or 3.6 75 I of safety equipment.

2. Equipment . Control I 2.2.22 Knowledge of limiting conditions for i 4.7 95 operations and safety limits. ! Subtotal 3 1

--ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of radiation monitoring systems, such as fixed radiation 2.3.15 monitors and alarms, portable survey 2.9 70 per.sonnel monitoring e ulpment, etc. m.--c------!---t----l----+---

I Abilit y to comply with radiation ! 2.3.7 perm it requirements during normal 3.5 71 I*abno rmal requirements.

i 3. Radiation Control 4.3 96 Know ledge of radiological safety princ iples pertaining to licensed opera tor duties, such as containment Abilit y to control radiation releases.

2.3.11 37 100 2312.. *entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligninq filters, etc. I Ii 2 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, 2 2.4.21 core cooling and heat removal, reactor 4.0 72 *coolant system integrity, containment

! conditions, radioactivity release Knowledge of the implications of EOP warnings, 2.4.20 3.8 73 and notes. 4. Emergency Knowledge of EOP mitigation

3.7 Procedures

I 2.4.6 strategies.

Plan Knowledge of EOP entry conditions and 4.8 2.4.1 97 immediate action steps. m !Knowledge of low power I shutdown implications in accident (e.g., loss of 2.4.9 4.2 i 98 coolant accident or loss of residual heat removal Illitigation strate ies. Subtotal 3 2 3 Point Total: 10 7 Record of Rejected KIA's Form ES-401-4 Randomly Selected Reason for Rejection Tier / Group KA Question 81, 2.4.31 -Emergency Procedures

/ Plan: Knowledge 295016/24 31 of annunciator alarms, indications, or response procedures.

replaced with '2950161 Thesreopis nOdlinhk to 50.43 because. control is Tier 1 1 Group 1 124 11 *an an t ere are no annunciators assocla e WI I .. . . Randomly selected 2.4.11, Knowledge of abnormal condition rocedures.

Question 39, EK1.04 Knowledge of the operational implications 2950371 EK1 04 .of the following concepts as they apply to SCRAM CONDITION . . 1 PRESENT AND REACTOR POWER ABOVE APRM Tier 11 Group 11 DOWNSCALE OR UNKNOWN: Hot shutdown boron weight: Plant-Specific.

NMP 2 does not use Hot shutdown boron weight. 1 Randomly selected EK1 06 , Cooldown effects on reactor power I Question 44, EK2.05 -Knowledge of the interrelations between 295026 1 EK2.05 SUPPRESSION POOL HIGH WATER TEMPERATURE and the i . Tier 1 1 Group 1 replaced with 295026 following:

Containment pressure:

Mark-III.

This KIA does not 1 EK2.02 apply to NMP 2 which has a Mark II Containment.

Randomly EK2.02, pool spray: Plant Specific Question 53, EA2.02 -Ability to determine and/or interpret the 2950381 EA2.02 following as they apply to HIGH OFF-SITE RELEASE Tier 1 1 Group replaced with 295038 Total number of curies released.

This is not a function of the RO 1 EA2.04 licensed position at NMP 2. Randomly selected EA2.04, Source of off-site release. Question 55, 2.4.3 -Emergency Procedures 1 Plan: Ability to 295005/2.4.3 identify post-accident instrumentation.

There are no Tier 1 1 Group replaced with 295005 accident instruments associated with a main turbine generator 12.4.31 trip at NMP 2. Randomly selected 2.4.31, Knowledge of annunciator alarms, indic;i3tions, or response procegures.

! 1 Question 84, 2.1.30 -Conduct of Operations:

Ability to locate and operate components, including local controls.

There is an 295020 1 2.1.30 SOP 83 for primary containment isolations but there are no Tier 1 1 Group 2

• replaced with operations for an SRO to direct. Additionally there are no local 12.1.20 operations in response to inadvertent isolations in the Unable to write an effective question, randomly selected Ability to interpret and execute procedure Question 85, 2,1.28 -Conduct of Operations:

Knowledge of purpose and function of major system components and 295009 I 2.1.28 Could not write an SRO question to meet 10CFR50.43 Tier 1 1 Group 2 replaced with 295020 requirements.

Randomly replaced with 2.1.23, Ability to perform 12.1,23 specific system and integrated plant procedures during all modes of plant operation.

I Question 86, A2, 11 -Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and 209001 1 A2.11 ,(b) based on those predictions, use procedures to correct, Tier 21 Group 1 replaced with 209001 1 control, or mitigate the consequences of those abnormal . . . .1 A2.04 conditions or operations, Loss of fire protection.

BWR-1, NMP 2 1 is not a BWR-I this KIA does not apply, randomly selected . A2,04, D.C. failures I T 2/ G 1 ler roup T 2/ G 1 ler roup rep ace,

• =r 300000' K6.04 T 2 , G 1 I d

'th 300000 ler roup rep ace,

• 262002 / A 1 02 Tier 2 / Group 1 Tier 2 , Group Tier Category Record of Rejected KIA's Form ES-401-4 I Question 5, K3.03 Knowledge of the effect that a

!malfunction of.the STANDBY will \ 211000 / K3 03 have on followmg:

Core plate differential pressure mdlcatlon.

• I d *th 211000 NMP 2 SLC discharges into the HPCS spray ring in the top of rep ace/ 01 the core there is no relationship between a loss or malfunction of

• SLC and Core plate differential measurement.

Randomly selected K3.01, Ability to shutdown the reactor in certain conditions Question 6, K3.13 -Knowledge of the effect that a loss or 223002/ K3.13 malfunction of the PRIMARY CONTAINMENT ISOLA.TION I d 'th 223002 SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

Isolation Condenser:

Plant-Specific.

NMP2 does not have isolation condensers, randomly selected K3.20 Standby as treatment s stem Question 11, K6.04 -Knowledge of the effect that a loss or malfunction of will have on the INSTRUMENT'AIR SYSTEM: Service air refusal valve. NMP 2 does not have a refusal valve, additionally if this KIA was used to write a question on an isolation valve it would be very similar to another KIA on the NRC exam. Randomly K6.13-,-,_F_ilt..;..e.....crs..-------:-

____-1 Question 14, A1.02 -Ability to predict and/or monitor changes in parameters associated with operating the UNINTERRUPTABLE POWER SUPPLY (AC.lD.C,)

controls including:

Motor generator outputs. NMP 2 does not have any motor generators la d *th 262002 that are UPS. All NMP 2s UPS units have inverters.

Could not rep ce, 01 select another A1 KIA because the remaining importance was

• less than 2.5. Randomly selected A3.01, Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C.lD.C.)

including:

Transfer from preferred to alternate source, 259002 , 2.4.30 replaced with 259002 '2.4.4 2.3.11 replaced with 2,3.7

..Question 21,2.4.30

-Emergency Procedures' Plan; Knowledge of events related to system operation

/ status that must be reported to internal organizations or external agencies, such as the state, the NRC, or the transmission system operator.

There are no RO responsibilities regarding Reactor Water Level Control and reporting to internal organizations or external agencies.

Randomly selected 2.4.4, Ability to recognize abnormal indications for system operating parameters that are ent level conditions for emer enc and abnormal rocedures.

Question 71,2.3.11

-Ability to control radiation releases.

This same KIA is used for Question 96. To prevent a double jeopardy i question randomly selected 2.3.7, Ability to comply with radiation . work permit requirements during normal or abnormal re uirements.

Question 73,2.4.9 -Knowledge of low power' shutdown implications in accident (e.g., loss of coolant accident or loss of Tier 3/ 2.4.9 replaced with

• residual heat removal) mitigation strategies.

This same KIA is Category 4 2.4.20 i* used for Question 98. To prevent a double jeopardy question randomly selected 2.4.20, Knowledge of the operational

"---_____--'-________-'--im....plications of EOP warnings, cautions, and notes.

Record of Rejected KiA's Form ES-401-4 Question 100, 2.3.15 -Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. This Tier 31 2.3.15 replaced with ! same KIA is used for Question 70. To prevent a double jeopardy Category 3 iquestion randomly selected 2.3.12, Knowledge of radiological . safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, I access to locked high-radiation areas, aligning filters, etc.

I Qduest'fytion 88 , 2.4:d 3 -E l.plan:HApbcilisty too t t t 209002/243 I en I pos -acci en inS rumen a Ion. ere IS no p s I d 'th 20'9002 accident instru mentation.

Randomly replaced with 2.1.7, Tier 2 1 Group 1 rep ace 7 Conduct of Operations:

Ability to evaluate plant performance and " . . make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. . 2.2.42 -Equipment Control: Ability to recognize system parameters that are entry-level conditions for Technical 211000/2.2.42 Specifications.

Had difficulty writing an appropriate RO question, . Tier 21 Group 1 replaced with 211000. randomly replaced with 2.1.7 Ability to evaluate plant 12 1 7 .. . perf rman e nd m ke pe ronal' dgments based n 0 c a a 0 ra I JU 0 operating characteristics, reactor behavior, and instrument i interpretation.

! i ES-301 Administrative Topics Outline Form ES-301-1 Facility:

NMP2-NRC Examination Level: RO Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Type Code* M,R N,R M,R Date of Examination:

March 2012 Operating Test Number: NRC Describe activity to be performed Perform Jet Pump Flow Mismatch Checks lAW LOG-DOO1 Attachment 10 The candidate will calculate Jet Pump loop flow mismatch and identify a Jet Pump DP out-of-spec.

2.1.18 (3.6) Ability to make accurate, clear, and concise logs, records, status boards and reports. N2-0SP-LOG-D001, Attachment 4 Determine Heatup Rate During Startup Given turnover conditions during a plant startup, the candidate will interpret heatup data, calculate the heatup rate and take the appropriate actions. 2.1.43 (4.1) Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc. N2-0SP-RC8-@OO1 Determine Adequacy of Clearance for SLC The candidate will review a proposed clearance for SLC and determine the adequacy of the clearance boundaries.

2.2.14 (3.9) Knowledge of the process for controlling equipment configuration or status. CNG-OP-1.01-1007 Emergency Plan D,S Perform RO Actions for an Injured and Contaminated Person The candidate will perform EPIP-EPP-04 actions for a contaminated and injured person to be transported offsite. 2.4.12 (4.0) Knowledge of general operating crew responsibilities during emergency operations.

EPI P-EPP-04 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 Gs.3 for ROs; 4 for SROs & RO retakes) (N)ew or (M)odified from bank (P)revious 2 exams randomly selected)

ES-301 Administrative Topics Outline Form E5-301-1 Facility:

NMP2-NRC Examination Level: SRO Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Type Code'" N,R D,R D,R Date of Examination:

March 2012 Operating Test Number: NRC Describe activity to be performed Detennine Plant Impact for Inoperable Unit Cooler Given a failed closed service water inlet valve to 2HVC"UC107, determine the effect on the unit cooler and UPS2D operability per N2-0P-53E and Tech Specs. 2.1.32 (4.0) Ability to explain and apply system limits and precautions.

N2-0P-53E and Technical Specifications Detennine Personnel Overtime Availability Given a list of personnel and their previous work hours, determine who is available for overtime and why others are not available, then complete a waiver for an individual to perform work beyond administrative work requirements based on administrative requirements.

2.1.5 (3.9) Ability to use procedures related to shift staffing, such as minimum crew requirements, overtime limitations, etc. CNG-SE-1.01-1002 Offsite Dose Calculation Manual (ODCM) Assessment for Inoperable Equipment . Given conditions requiring removing Offgas radiation monitors OFG"RE13A and OFG"RE13B from service. the candidate will determine the actions required by N2-0P-42.

Offgas System and the ODeM. 2.2.38 (4.5) Knowledge of conditions and limitations in the facility license.

Radiation Control D,R Radiological Requirements Related to Operator Inspection of High Radiation Areas Given radiological conditions r.elated to an area where work is to be performed as shown on a survey map, and other applicable conditions such as the RWP, ensure the appropriate radiological aspects of the job are met prior to sending the operator into the area. 2.3.12 (3.7) Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc GAP-RPP-01, 02, 07 and 08; S-RAP-RPP-0703 Emergency Plan M,R Classify Emergency Event and Determine Protective Action Recommendations The candidate will classify an emergency event and notify offsite agencies with Protective Action Recommendations.

2.4.44 (4.4) Knowledge of emergency plan protective action recommendations NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.

• Type Codes & Criteria: (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (,:5.3 for ROs;,:5. 4 for SROs & RO retakes) (N)ew or (M)odified from bank (P)revious 2 exams (.:5.1; randomly selected)

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

Exam Level: RO/SRO Nine Mile Point Unit 2 NRC Date of Examination:

Operating Test No.: Control Room Systems@ (8 for RO; 2 or 3 for SRO-U, including 1 ESF) March 2012 NRC S-1 System 1 JPM Title Align SBGTS Train "A" to reduce Drywell pressure The candidate will place Standby Gas Train "A" on the Drywell in accordance with N2-0P-61A, Section H.1.0. S-2 KIA 295024 EA 1.20 3.5/3.6 N2-0P-61A, H.1.0 Start RCIC in Reject to CST Mode (Tank to Tank) The candidate will start RCIC and place it in tank to tank mode for RPV pressure control lAW N2-EOP-HC.

After RCIC is in service a high RCIC turbine exhaust pressure condition will occur and RCIC will fail to trip. The candidate will take manual action to trip the RCIC turbine. S-3 KIA 217000, A4.04, 3.6/3.6 N2-EOP-HC Swap Instrument Air Compressors The candidate will perform a swap of the Instrument Air Compressors per N2-0P-19.

After the compressors are swapped, the lead air compressor will trip requiring the operator to realign the system per N2-S0P-19.

'S4 KIA 295019 AA2.01 3.5/3.6 N2-0P-19 F.2.0; N2-S0P-19 Startup a Feedwater Pump following a Scram IT cd*lsftyF unclon ype o e ae f D,S I I , 5 CONTAIMENT INTEGRITY I , M,A,L,S I I 4 D,A,S D,L,S HEAT REMOVAL FROM RXCORE 8 PLANT SERVICE SYSTEMS 2 The candidate will restart Reactor Feed Pump A and raise RPV water level above 159 inches following a plant scram lAW N2-S0P-101C.

KIA 259001 A4.023.9/3.7 N2-S0P-101C , REACTOR WATER INVENTORY CONTROL , S-5 Energizing 2ENS*SWG103 from Division II EDG & Energize NNS-SWG15 from SWG 103 i D,S 6 ELECTRICAL The candidate will energize SWG 103 from the Div II EDG and SWG 15 from SWG 103 lAW N2-S0P-3 Sections 8.4 & 9.3 i i KIA 262001 A4.01 3.4/3.7 N2-S0P-3 Sects. 8.4 &9.3 I II I i Control Roomlln-Plant Systems Outline Form ES-301-2 Place the Standby Loop of SFC in Service.S-6 N,E,S 9 RO RADIOACTIVITY The candidate will respond to a loss of fuel pool cooling by RELEASE ONLY starting the standby train lAW N2-S0P-38.

KiA 233000 A2.04 Transfer Recirculation Pump from Low Speed to S-7 N,A,S 1 Speed REACTIVITY CONTROL The candidate will transfer RCS pump A from low to high speed operation lAW N2-0P-29.

When the pump is shifted to high speed the FCV will slowly drift open requiring the candidate to lock up the FCV lAW N2-S0P-08 KiA 202002 A4.08, N2-0P-29, Venting the RPV to the D,A,L,S 3 REACTOR The candidate will line up to vent the Reactor to the Main S-8 PRESSURE CONTROL Condenser through the MSIVs however the MSIVs will fail to open and they must lineup the MSIV Drains lAW Attachment KiA 239001 A4.01 4.2/4.0, 239001, A4.02, N2-EOP-6 Attachment 18 In-Plant Systems@ (3 for RO; 3 or 2 for SRO-U) P-1 Aligning Service Water to SFC Heat Exchanger 1A D,E,R I 8 PLANT SERVICE SYSTEMS The candidate will align Service Water to Spent Fuel Heat Exchanger 1A, lAW N2-S0P-38 Attachment KiA N2-S0P-38 Vent the Control Rod Overpiston D,E,L,R 1 REACTIVITY P-2 The candidate will insert control rod 26-59 to notch 00 by CONTROL locally venting its overpiston area lAW N2-EOP-6 Attachment 14 KA: 295015 AA1.01 N2-EOP-6, Attachment 14 Place Battery Charger 2BYS-CHGR1A1 in service. 6 P-3 D ELECTRICAL The candidate will place battery charger into service lAW KiA 263000 A 1.01 2.5/2.8 N2-0P-73A 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.

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

• Type Codes Criteria for RO I SRO-II SRO-U (A)ltemate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power

/ Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator 4-6 /4-6/ 2-3 S,91s,81!:4

1/;

:1/;::1 1 ;::1 (control room system) ;::1/;::1/;::1

2/;

:2/;::1 s, 3 1 s, 3 1:5 2 (randomly selected)

1/;

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

Nine Mile Point 2 Scenario No.: NRC-1 Op-Test No: March 2012 Examiners:

________ Operators:

__________ Initial Conditions:

Simulator IC-153 1. Reactor Power -100% 2. RHR Band LPCI COOS for Division II Workweek Turnover:

1. Swap Recirc Pump HPU Subloops 1 2 7 Malf. No. AD08A AD08C C (RO) C (SRO) Event .nr"-",",,.,in

... Pool rupture results in loss of inventory in the suppression pool, requires blowdown.

N2-EOP-RC N2-EOP-PC

Failure of the ADS pushbutton or key lock switches to actuate 7 ADS valves N2-EOP-C2

• (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor NRC Scenario 1 March 2012 Facility:

Nine Mile Point 2 Scenario No: NRC-1 Op-Test No: March 2012 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.5.d) ACTUAL ATTRIBUTES

1. Total malfunctions (5-8) Events 2,3,4,5,6,7 6 2. Malfunctions after EOP entry (1-2) Event 7 1 3. Abnormal events (2-4) Events 2, 4 2 4. Major transients (1-2) Event 6 1 5. EOPs entered/requiring substantive actions (1-2) Events 4,6 EOP-RPV, EOP-PC 2 6. EOP contingencies requiring substantive actions (0-2) Event 7 EOP-C2 1 7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

CT-1.0 Given a failure of 2RCS*P18 pump seals, the crew will take action to trip and isolate 2RCS*P18 lAW N2-50P-29.1 This task is identified as critical because without operator action to trip and isolate the Recirc pump, Drywell pressure would continue to rise until the reactor automatically scrams. CT-2.0 Given a lowering suppression pool level, the crew will open 7 SRVs per N2-EOP-C2 prior to suppression pool level reaching 192 feet This task is identified as critical because without operator action to blowdown the RPV prior level reaching 192 feet, the primary containment pressure limit could be exceeded due to a loss of pressure suppression capability concurrent with pressure control via SRVs. NRC Scenario 1 March 2012 Appendix 0 Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 2 Scenario No: NRC-2 Op-Test No: March 2012 Examiners:

Operators:

__________ Initial Conditions:

Simulator IC-154 1. Reactor Power -92% 2. RCIC is out of service 3. A momentary loss of power signal caused lockup of LV-10A last shift. Turnover:

1. Reset LV-10A Lockup lAW N2-S0P-06, Attachment

1. 2. After the valve has been reset and back in automatic, restore power to 100%. Malf. No. *' {N)ormal, {R)eactivity, NRC Scenario 2 Lowering service water bay level with ure 2SWP*'M0V77 AlB to automatically open on the lowering level. Crew will be required to manually open M0V77 AlB to restore level. fault starts, Liquid Poison pump fails to start. Loss of adequate high pressure injection sources requires RPV blow down to restore adequate core cooling. N2-EOP-C2 (I)nstrument, {C)omponent, (M)ajor March 2012 Facility:

Nine Mile Point 2 Scenario No.: NRC*2 Op-Test No.: March 2012 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.S.d) 1. Total malfunctions (5-8) I Events 3,4,5,6,7,8 i 2. Malfunctions after EOP entry (1-2) Events 8 I 3. Abnormal events (2-4) , Events 3, 4, 5, 6, 4. Major transients (1-2) 1 Event 7 I 5. EOPs entered/requiring substantive actions (1-2) Events 7, S EOP-RPV, EOP-PC I 2 6. EOP contingencies requiring substantive actions (0-2) Event S EOP-C2 I 1 I 7. Critical tasks (2-3) 2 ACTUAL ATTRIBUTES 6 1 4 CRITICAL TASK DESCRIPTIONS:

CRITICAL TASK JUSTIFICATION:

I CT-1.0 Given service water intake bay level less than 234 ft and a This task is identified as critical because without I failure of 2SWP"M0V77A

&77B to automatically open, the crew will operator action the plant will lose its ultimate take action to manually open 2SWP"M0V77 A & 77B per N2-S0P-11 heat sink. CT-2.0 Given RPV level at or below the TAF but above the MSCWL, the This task is identified as critical because without crew will open 7 ADS valves lAW N2-EOP..c2 operator action, RPV level will continue to lower until the fuel is no longer adequately cooled. NRC Scenario 2 March 2012 Appendix D Scenario Outline Form ES-D-1 Facility:

Nine Mile Point 2 Scenario No.: NRC-3 Op-Test No.: March 2012 Examiners:

Operators:

__________ Initial Conditions:

Simulator IC-155 1. Reactor Power -63% 2. Reactor shutdown is in progress Turnover:

1. Place the "C n heater drain pump in recirculation mode lAW N2-0P-8, Section G.1. 2. Lower power using recirculation flow to 58%. Event No. 1 Malt. No. Event Type* N/A N heater drain pumps N (SRO) Event I"UICUIUI mode. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, {M)ajor NRC Scenario 3 March Facility:

Nine Mile Point 2 Scenario No.: NRC-3 Op-Test No.: March 2012 TARGET QUANTITATIVE ATTRIBUTES (PER SCENARIO; SEE SECTION D.S.d) 1. Total malfunctions (5-8) Events 3,4,5,6,7,8

2. Malfunctions after EOP entry (1-2) Events 8 3. Abnormal events (2-4) Events 3, 4, 5, 6 4. Major transients (1-2) Event 7 I 5. EOPs enteredlrequiring substantive actions (1-2) Events 7 EOP-RPV I 6. EOP contingencies requiring substantive actions (0-2) ! Event 7 EOP-C5 .. I ACTUAL ATTRIBUTES 6 1 4 1 1 1 3 i I I i i I 7. Critical tasks CRITICAL TASK CRITICAL TASK JUSTIFICATION:

CT-1.0 Given a failure of the reactor to SCRAM the crew will inhibit ADS per N2-EOP-C5 This task is identified as critical because without operator action to inhibit ADS prior to manually lowering RPV level, the reactor could experience a rapid and uncontrolled cooldown and subsequent injection of cold water which will dilute boron concentrations and add positive reactivity to the reactor if level were lowered below Level 1. CT -2.0 Given a failure of the reactor to SCRAM, power above 4%, and This task is identified as critical because without RPV water level above 100 inChes, the crew will tenninate and prevent operator action to terminate and prevent all injection except SLS, CRD and RCIC per N2-EOP-C5 injection, the reactor could experience large irregular neutron flux oscillations induced by neutroniclthermal-hydraulic instabilities.

CT -3.0 Given a failure of the reactor to SCRAM, the crew will insert This task is identified as critical because without control rods per N2-EOP-6, Attachment 14 operator action to insen control rods, the reactor will remain susceptible to inadvertent power generation due to potential boron dilution or displacement NRC Scenario 3 March 2012