ML14119A395

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Draft Outlines (Folder 2)
ML14119A395
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 02/18/2014
From: Hooper G
Entergy Corp
To: David Silk
Operations Branch I
Jackson D
Shared Package
ML13330B159 List:
References
TAC U01887
Download: ML14119A395 (37)
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Text

ES-401 BWR Examination Outline Form ES-401-1 Facility: James A Fitzpatrick Date of Exam: April2014 RO KIA Category Points SRO-Only Points Tier Group K K K K K K A A A A G Total A2 G* Total 1 2 3 4 5 6 1 .2 3 4

  • 1 4 3 3 3 4 3 20 3 4 7 1.

Emergency

& 2 1 1 2 1 1 1 7 2 1 3 Plant Evolutions Tier Totals 5 4 5 4 5 4 27 5 5 10 1 3 2 2 3 2 2 1 3 3 3 2 26 2 3 5 2.

Plant 2 1 1 1 1 1 1 1 :2 1 1 1 12 0 1 2 3 Systems Tier Totals 4 3 3 4 3 3 2 5 4 4 3 38 3 5 8 1 2 3 4 1 2 3 4

3. Generic Knowledge & Abilities 10 7 Categories 2 3 3 2 2 2 2 1 Note 1. Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not b,e less than two).
2. The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
3. Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to section D.1.b of ES-401, for guidance regarding elimination of inappropriate KIA statements.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
5. Absent a plant specific priority, only those 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) KIAs in Tiers 1 and 2 shall b1a selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIA's

8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals(#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2 (Note #~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 Fom1 ES-401-3. Limit SRO selections to KIAs that are linked to 10CFR55.43

ES-401 2 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I Name Safety Function KJA Topic(s)

EA2.03 -Ability to determine and/or interpret the following as 295030 Low Suppression they apply to LOW X 3.9 76 Pool Water Level/ 5 SUPPRESSION POOL WATER LEVEL: Reactor pressure AA2.02 -Ability to determine and/or interpret the following as they apply to PARTIAL OR 295019 Partial or COMPLETE LOSS OF Complete Loss of X 3.7 77 INSTRUMENT AIR: Status of Instrument Air I 8 safety-related instrument air system loads (see AK2.1 -

AK2.19)

AA2.06 -Ability to determine and/or interpret the following as 295005 Main Turbine X they apply to MAIN TURBINE 2.7 78 Generator Trip I 3 GENERATOR TRIP:

Feedwater tem_Qerature 2.4.41 - Emergency 295021 Loss of Procedures I Plan: Knowledge X 4.6 79 Shutdown Cooling I 4 of the emergency action level thresholds and classifications.

2.4.18 - Emergency 295023 Refueling X Procedures I Plan: Knowledge 4.0 80 Accidents I 8 of the specific bases for EOPs.

2.2.37 - Equipment Control:

295038 High Off-site Ability to determine operability Release Rate I 9 X 4.6 81 and I or availability of safety related ~uipment.

2.4.35 - Emergency Procedures I Plan: Knowledge 295031 Reactor Low Water Level/ 2 X of local auxiliary operator tasks 4.0 82 during emergency and the resultant oj>_erational effects.

AK1.01 -Knowledge of the operational implications of the 600000 Plant Fire On Site /8 X following concepts as they 2.5 39 apply to Plant Fire On Site: Fire Classifications by type AK1.02 - Knowledge of the operational implications of the 295006 SCRAM /1 X following concepts as they 3.4 40 apply to SCRAM: Shutdown margin

ES-401 2 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1

~ EAPE # 1 Name Safety Function I I I I I I I K1 K2 K3 A1 A2 G KIA Topic(s) limp. I Q# ~

EK1.02 - Knowledge of the

' ,,/:'"/*': operational implications of the 295037 SCRAM following concepts as they Condition Present and  ;'<Y* apply to SCRAM CONDITION Reactor Power Above X PRESENT AND REACTOR 4.1 41 APRM Downscale or POWER ABOVE APRM Unknown I 1 DOWNSCALE OR UNKNOWN:

'.'* Reactor water level effects on

. "'i reactor power
AK2.02 - Knowledge of the interrelations between
  • < . **. ,~.~.

295016 Control Room CONTROL ROOM Abandonment I 7 X .i. 4.0 42 ABANDONMENT and the

<' ,...* following: Local control

.,,. i.* ** *.:: stations: Plant-Specific AK2.03- Knowledge of the

.  ::;: :: and the following:

RHR/shutdown cooling

  • ':<.***:., EK2.15 - Knowledge of the

.: interrelations between 295031 Reactor Low X i

        • ,:';;:;. 3.2

.,, REACTOR LOW WATER 44 Water Level I 2 LEVEL and the following: A. C.

' ~ . ,.,, ';;,:':ii: distribution: Plant-S_Q_ecific

. EK3.01 - Knowledge of the reasons for the following 295028 High Drywell i):':~ :, responses as they apply to X 3.6 45 Temperature I 5 HIGH DRYWELL

... TEMPERATURE: Emergency

  • .. :>:::::::;:*:' depressurization

/',:i. AK3.03 - Knowledge of the

'I reasons for the following 295003 Partial or .:;:********). : responses as they apply to X

Complete Loss of AC I 6 i

>,[::~i:' PARTIAL OR COMPLETE 3.5 46 LOSS OF A.C. POWER: Load

.: t shedding

.**. *::*.* 1.!/i*T

EK3.02- Knowledge of the i'

reasons for the following 295024 High Drywell '::<:::* responses as they apply to X 3.1 47 Pressure I 5 ,:f;;:i HIGH DRYWELL PRESSURE:

.:[i Auxiliary building isolation:

i' .*.: ~~~;,*' ;j,c Plant-Specific EA 1.05 -Ability to operate

.':: .. and/or monitor the following as

  • i ii; 295030 Low Suppression they apply to LOW Pool Water Level I 5 X **,,,: SUPPRESSION POOL 3.5 48
+***'

.*: WATER LEVEL: HPCI

ES-401 2 Form ES-401-1 JAF 14-1 NR.C Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 1 EAPE #I Name Safety Function G I KJA Topic(s)

,' EA1.06- Ability to operate c'ccc, I' and/or monitor the following as 295038 High Off-Site c ','.

X ,* they apply to HIGH OFF-SITE 3.5 49 Release Rate I 9 ,;<li* RELEASE RATE: Plant y

,,,, ,))j' ventilation EA1.07- Ability to operate IY,,,,,<<Y and/or monitor the following as 295025 High Reactor '

X ,,,,,,** V'*,,.*><<' they apply to HIGH REACTOR 4.1 50 Pressure I 3

,;;, PRESSURE: ARI/RPT/ATWS:

l**:';i'!:,. Plant-Specific AA2.07- Ability to determine 295005 Main Turbine i'

X

\, *t~:!;i::~':.* and/or interpret the following as they apply to MAIN TURBINE 3.5 51 Generator Trip I 3 GENERATOR TRIP: Reactor

', ' , ,* ,,,,;1'* li\:~ :;,,? water level AA2.01 -Ability to determine

',)ci','.C li{iikc* and/or interpret the following as 295019 Partial or they apply to PARTIAL OR Complete Loss of X COMPLETE LOSS OF 3.5 52 Instrument Air I 8

,:' INSTRUMENT AIR: Instrument

( ' c

,,,, air system pressure AA2.03- Ability to determine

<<'** and/or interpret the following as 295001 Partial or they apply to PARTIAL OR Complete Loss of Forced

,~'i'i I*<<Jc,,,,,; (' COMPLETE LOSS OF 3.3 53 Core Flow Circulation I 1 FORCED CORE FLOW

&4

,;)'< CIRCULATION: Actual core

';i) lc:**,;;;;~;;;,., flow

  • ,.~~:: ::*

700000 Generator L Cc 2.2.12 - Equipment Control:

Voltage and Electric Grid Knowledge of surveillance 3.7 54 Disturbances procedures.

r* I <*< 2.4.47- Emergency cc),

l',.,:,;,::,i<i Procedures I Plan: Ability to 295018 Partial or diagnose and recognize trends Complete Loss of Component Cooling x,,.'", in an accurate and timely 4.2 55 manner utilizing the appropriate Water I 8 c \'<;; control room reference

,J,;

material.

2.2.44 - Equipment Control:

Ability to interpret control room indications to verify the status 295023 Refueling Accidents I 8 X and operation of a system, and 4.2 56 understand how operator

'') actions and directives affect plant and system conditions.

ES-401 2 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1

~ EAPE #I Name Safety Function I I I I I I I K1 K2 K3 A1 A2 G KIA Topic(s} limp. I Q# ~

EA2.02- Ability to determine and/or interpret the following as 295026 Suppression they apply to SUPPRESSION Pool High Water X 3.8 57 POOL HIGH WATER Temperature I 5 TEMPERATURE: Suppression pool level AK1.05- Knowledge of the operational implications of the 295004 Partial or following concepts as they Complete Loss of D.C. X apply to PARTIAL OR 3.3 58 Power/6 COMPLETE LOSS OF D.C.

POWER: Loss of breaker protection KIA Category Totals: 4 3 3 3 4/3 3/4 Group Point Total: 1 20/7

ES-401 3 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions -Tier 1 Group 2 EAPE #I Name Safety Function KJA Topic(s)

AA2.05 -Ability to determine 2950201nadvertent and/or interpret the following as Containment Isolation I 5 X they apply to INADVERTENT 3.6 83

&7 CONTAINMENT ISOLATION:

Reactor water level 2.4.45 - Emergency Procedures I Plan: Ability to 295022 Loss of CRD Pumps I 1 X prioritize and interpret the 4.3 84 significance of each annunciator or alarm.

AA2.02 -Ability to determine 2950151ncomplete and/or interpret the following as SCRAM I 1 X they apply to INCOMPLETE 4.2 85 SCRAM: Control rod position AK1.03- Knowledge of the operational implications of the 295010 High Drywell following concepts as they X 3.2 59 Pressure I 5 apply to HIGH DRYWELL PRESSURE: Temperature increases EK2.03- Knowledge of the interrelations between HIGH 500000 High CONTAINMENT HYDROGEN Containment Hydrogen X 3.3 60 CONCENTRATIONS the Concentration I 5 following: Containment Atmosphere Control System AK3.01 - Knowledge of the reasons for the following 295002 Loss of Main responses as they apply to X 3.7 61 Condenser Vacuum I 3 LOSS OF MAIN CONDENSER VACUUM: Reactor SCRAM:

Plant-Specific AA 1.04 -Ability to operate and/or monitor the following as 295007 High Reactor Pressure I 3 X they apply to HIGH REACTOR 3.9 62 PRESSURE: Safety/relief valve operation: Plant-Specific EA2.01 -Ability to determine and/or interpret the following as 295036 Secondary they apply to SECONDARY Containment High Sump/Area Water Level I X CONTAINMENT HIGH 3.0 63 SUMP/AREA WATER LEVEL:

5 Operability of components within the affected area 2.4.45 - Emergency 295032 High Secondary Procedures I Plan: Ability to Containment Area X prioritize and interpret the 4.1 64 Temperature I 5 significance of each annunciator or alarm.

ES-401 3 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE # I Name Safety Function G I KIA Topic(s) limp. I Q# ~

EK3.02 - Knowledge of the reasons for the following 295034 Secondary responses as they apply to Containment Ventilation X SECONDARY CONTAINMENT 4.1 65 High Radiation I 9 VENTILATION HIGH RADIATION: Starting SBGT/FRVS: Plant-Specific 1/ 1/

KIA Category Totals: 1 1 2 1 2 1 Group Point Total:

I 7/3

ES-401 4 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp.

System #I Name A2 G Q#

1 2 3 4 5 6 1 ~~ 4 A2.11 -Ability to (a) predict the impacts of the following on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR 223002 PCIS/Nuclear STEAM SUPPLY SHUT-X 3.9 86 Steam Supply Shutoff OFF; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Standby liquid initiation A2.05 -Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use 239002 SRVs X 3.4 87 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor pressure 2.4.3 - Emergency Procedures I Plan: Ability 218000 ADS X 3.9 88 to identify post-accident instrumentation.

2.4.4 - Emergency Procedures I Plan: Ability to recognize abnormal indications for system 206000 HPCI X operating parameters 4.7 89 which are entry-level conditions for emergency and abnormal operating procedures.

2.1.31 -Conduct of Operations: Ability to locate control room switches, controls, and 2150031RM X 4.3 90 indications, and to determine that they correctly reflect the desired plant lineup.

ES-401 4 Form ES-401-1 JAF 14-1 NR.C Exam Written Examination Outline Plant Systems -Tier 2 Group 1 K K K K K K A f!., A System # I Name A2 G Imp. Q#

1 2 3 4 5 6 1 3 4

    • . K1.03- Knowledge of the

'** physical connections and/or cause- effect 262001 AC Electrical relationships between X 3.4 1 Distribution A.C. ELECTRICAL DISTRIBUTION and the

' following: Off-site power sources K1.19- Knowledge of the physical connections and/or cause- effect

..Y,i relationships between PRIMARY 223002 PCIS/Nuclear CONTAINMENT X 'i  ::.:**.< 2.7 2 Steam Supply Shutoff ISOLATION

<>** SYSTEM/NUCLEAR

* .... STEAM SUPPLY SHUT-OFF and the following:

Component cooling water

.* systems K2.02- Knowledge of 203000 RHR/LPCI:

X electrical power supplies 2.5 3 Injection Mode L** 1:, to the following: Valves

  • }<.*<* K2.01 -Knowledge of electrical power supplies 212000 RPS X 3.2 4 i***u:***** to the following: RPS motor-generator sets K3.02 - Knowledge of the

/. <; effect that a loss or

.... malfunction of the 215004 Source Range SOURCE RANGE X 3.4 5 Monitor MONITOR (SRM)

SYSTEM will have on

... following: Reactor manual

... **. control: Plant-Specific i i* K3.05- Knowledge of the

..*... effect that a loss or malfunction of the 259002 Reactor Water REACTOR WATER X 2.8 6 Level Control LEVEL CONTROL SYSTEM will have on following: Recirculation

,.. flow control system

ES-401 4 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 System # I Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2

  • I Gl l,m, II Q#

K4.01- Knowledge of

Automatic system

>::.; initiation

'"'> K4.02- Knowledge of

< EMERGENCY GENERATORS (DIESEL/JET) design 264000 EDGs X feature(s) and/or 4.0 8

'*'*> interlocks which provide

' for the following:

Emergency generator

  • .,,'". trips (emergency/LOCA)

K5.01 -Knowledge of the operational implications of the following concepts as 263000 DC Electrical they apply to D.C.

X 2.6 9 Distribution ELECTRICAL i.ii. DISTRIBUTION:

. Hydrogen generation

,;* during battery charging .

. ; K5.01 -Knowledge of the

<" ... operational implications of the following concepts as they apply to 218000 ADS X J 3.8 10 AUTOMATIC

.....,.,,. DEPRESSURIZATION SYSTEM: ADS logic

  • .* ... * ~,.j operation

./ **,',1:1 K6.05- Knowledge of the

    • '; , .. effect that a loss or 400000 Component X malfunction of the 2.8 11 Cooling Water i following will have on the

.. ,,,, CCWS: Motors

,.. , i"jC;*.

K6.03- Knowledge of the effect that a loss or

< malfunction of the 211000 SLC X following will have on the 3.2 12 STANDBY LIQUID

,. ,/, CONTROL SYSTEM:

.'**' A.C. power

ES-401 4 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 System # I Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2

  • IGI limp II Q#

K4.01 -Knowledge of

!!: UNINTERRUPTABLE POWER SUPPLY I< \; (A.C./D.C.) design X

262002 UPS (AC/DC) feature(s) and/or 3.1 13 r*y.... , . . :.*

interlocks which provide for the following: Transfer

' from preferred power to alternate power supplies A1.09- Ability to predict and/or monitor changes in parameters associated with operating the 239002 SRVs X RELIEF/SAFETY 3.1 14

.*, VALVES controls including: Indicated vs.

1,,,':.'" actual steam flow: Plant-

  • .* '* . Specific A2.08 -Ability to (a) li predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those 209001 LPCS X l;'i'.' predictions, use procedures to correct, 3.1 15 li:., ,*: control, or mitigate the consequences of those abnormal conditions or

.. \

operations: Valve

!'*. .*. . ,~;. openings f:.,' . :' A2.17 -Ability to (a)

,,,,,;:< predict the impacts of the following on the HIGH PRESSURE COOLANT

'* INJECTION SYSTEM;

~: ~

r :. : . . . !' .

and (b) based on those predictions, use 206000 HPCI X 3.9 16

  • V I' *. . . procedures to correct, control, or mitigate the

[ .. *. consequences of those I abnormal conditions or operations: HPCI 1:' '::*: inadvertent initiation:

lv .'

  • BWR-2,3,4

ES-401 4 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K K K K K K A A A Imp.

System # I Name A2 G Q#

1 2 3 4 5 6 1 :1 4

... , A3.03 -Ability to monitor automatic operations of the SHUTDOWN 205000 Shutdown COOLING SYSTEM X 3.5 17 Cooling ,' (RHR SHUTDOWN COOLING MODE) including: Lights and

[ .*>,:.':** alarms

    • > A3.02- Ability to monitor automatic operations of the AVERAGE POWER RANGE 215005 APRM I LPRM ,,,,X MONITOR/LOCAL 3.5 18 1.*******'*+ POWER RANGE MONITOR SYSTEM including: Full core

,....

  • dis olav r*.,:,,. A4.01 -Ability to manually operate and/or monitor in 300000 Instrument Air X 2.6 19 the control room:

Pressure qauqes A4.06 -Ability to manually operate and/or monitor in 215003 IRM X 3.0 20 the control room: Detector

',i'.,,, * **** drives 7

2.4.20 - Emergency Procedures I Plan:

~,,','"

Knowledge of operational 217000 RCIC ' 3.8 21 implications of EOP warnings, cautions, and J notes.

I) 2.1. 7 - Conduct of Operations: Ability to evaluate plant

< performance and make 205000 Shutdown Cooling x* operational judgments 4.4 22 based on operating

(',,,,

characteristics, reactor behavior, and instrument

'/***',',..' interpretation .

A3.01 -Ability to monitor automatic operations of

+*i the D.C. ELECTRICAL 263000 DC Electrical X DISTRIBUTION including: 3.2 23 Distribution Meters, dials, recorders,

,;',;:,' alarms, and indicating

,,; lights

ES-401 4 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 System# I Name K

1 K

2 K

3 K

4 K

5 K

6 A

1 A2

  • IGI I'm* I Q# I K1.01 -Knowledge of the physical connections and/or cause- effect relationships between 206000 HPCI X HIGH PRESSURE 3.8 24 COOLANT INJECTION SYSTEM and the following: Reactor vessel:

BWR-2,3,4 A2.04 -Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based 218000 ADS X on those predictions, use 4.1 25 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS failure to initiate A4.09 -Ability to manually operate and/or monitor in 212000 RPS X 3.9 26 the control room: SCRAM instrument volume level 3/ 2/

KIA Category Totals: 3 2 2 3 2 2 1 2

3 3 3

Group Point Total:

I 26/5

ES-401 5 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 System # I Name K

1 K

2 K

3 4 K K 5

K 6

A 1

A2

  • IGI l,m, II Q#

A2.04 -Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those 290001 Secondary X predictions, use 3.7 91 Containment procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High airborne radiation 2.4.6 - Emergency Procedures I Plan:

202001 Recirculation X 4.7 92 Knowledge of EOP mitigation strategies.

2.4.8 - Emergency Procedures I Plan:

201003 Control Rod Knowledge of how X 4.5 93 and Drive Mechanism abnormal operating procedures are used in conjunction with EOPs.

K1.06- Knowledge of the physical connections and/or cause- effect 259001 Reactor relationships between X 2.9 27 Feedwater REACTOR FEEDWATER SYSTEM and the following: Plant air systems K2.03 - Knowledge of 201001 CRD electrical power supplies X 3.5 28 Hydraulic to the following: Backup SCRAM valve solenoids K3.03- Knowledge of the effect that a loss or malfunction of the PRIMARY 223001 Primary CONTAINMENT Containment and X 3.4 29 SYSTEM AND Auxiliaries AUXILIARIES will have on following:

Containment/drywell pressure: Plant-Specific

ES-401 5 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems -Tier 2 Group 2 System #I Name K K K K K K A 1 2 3 4 5 6 1 A2

  • I Gl l,m. II Q#

K4.01 -Knowledge of i!i},','" TRAVERSING IN-CORE PROBE design feature(s) 215001 Traversing In- and/or interlocks which X 3.4 30 core Probe provide for the following:

, , <: , Primary containment isolation: Mark-I&II(Not-

< BWR1}

K5.06- Knowledge of the operational implications

,, of the following concepts 245000 Main Turbine as they apply to MAl N Generator and X >,,'!,. 2.5 31 TURBINE GENERATOR Auxiliary Systems

,, AND AUXILIARY SYSTEMS: Turbine shaft

' '>ii!ii, sealing

); K6.02 - Knowledge of the

,, effect that a loss or

,,,,, malfunction of the following will have on the 239003 MSIV X ii MSIV LEAKAGE 2.8 32 Leakage Control i H ,.,,

CONTROL SYSTEM:

Standby gas treatment system: BWR-4,5,6(P-I' ',: Spec)

,,,,,.,!';':'iii A 1.1 0 - Ability to predict and/or monitor changes

'ii in parameters associated 226001 RHR/LPCI: with operating the Containment Spray X '::::,;:', RHR/LPCI: 3.0 33 Mode ,,,

,!< CONTAINMENT SPRAY SYSTEM MODE controls including: Emergency

', Qenerator loading A2.03- Ability to (a) predict the impacts of the I ,,, ,

,, following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT I SPECIFIC); and (b) based on those 201006 RWM X,' predictions, use 3.0 34 procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Rod drift: P-

'ii ,, Spec(Not-BWR6)

ES-401 5 Form ES-401-1 JAF 14-1 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 System # I Name K

1 K

2 K

3 4 K K 5

K 6

A 1

A2

  • IGI limp II Q#

A3.05 -Ability to monitor automatic operations of the REACTOR WATER 204000 RWCU )( 2.8 35 CLEANUP SYSTEM including: Reactor water temperature A4.11 - Ability to manually operate and/or 233000 Fuel Pool X monitor in the control 2.5 36 Cooling/Cleanup room: Closed cooling water temperature 2.1.32 -Ability to explain 216000 Nuclear Boiler X and apply system limits 3.8 37 Instrumentation and precautions.

A2.06 -Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM; and (b) based on those 201003 Control Rod X predictions, use 3.0 38 and Drive Mechanism procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of CRD cooling water flow 21 1/

KIA Category Totals: 1 1 1 1 1 1 1 1 1 Group Point Total: 1 12/3 1 2

ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Facility: James A Fitzpatrick Date: April2014 RO SRO-Only Category KIA# Topic IR Q# IR Q#

Ability to use procedures related to shift 2.1.5 staffing, such as minimum crew complement, 3.9 94 overtime limitations, etc.

Knowledge of conservative decision making 2.1.39 4.3 98 practices.

1. Knowledge of criteria or conditions that Conduct 2.1.14 require plant-wide announcements, such as 3.1 66 of Operations _pump starts, reactor trips, mode chan_ges, etc.

Ability to use procedures related to shift 2.1.5 staffing, such as minimum crew complement, 2.9 67 overtime limitations, etc.

Subtotal 2 2 Ability to track Technical Specification limiting 2.2.. 23 4.6 95 conditions for operations.

Knowledge of the process for managing maintenance activities during shutdown 2.2..18 3.9 100 operations, such as risk assessments, work prioritization, etc.

Ability to manipulate the console controls as

2. 2.2.2 required to operate the facility between 4.6 68 Equipment shutdown and designated power levels.

Control Ability to perform pre-startup procedures for the facility, including operating those controls 2.2.1 4.5 69 associated with plant equipment that could affect reactivity.

Ability to determine the expected plant configuration using design and configuration 2.2.15 3.9 74 control documentation, such as drawings, line-ups, tag-outs, etc.

Subtotal 3 2 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, 2.3.5 2.9 96 portable survey instruments, personnel

3. monitoring equipment, etc.

Radiation Knowledge of radiation monitoring systems, Control such as fixed radiation monitors and alarms, 2.3.15 3.1 99 portable survey instruments, personnel monitoring equipment, etc.

Knowledge of radiological safety principles pertaining to licensed operator duties, such as 2.3.12 containment entry requirements, fuel handling 3.2 70 responsibilities, access to locked high-radiation areas, aligning filters, etc.

Knowledge of radiation or contamination hazards that may arise cluring normal, 2.3.14 3.4 71 abnormal, or emergency conditions or activities.

Knowledge of radiation exposure limits under 2.3.4 3.2 75 normal or emergency conditions.

Subtotal 3 2 2.4.29 Knowledge of the emergency plan. 4.4 97 4.

Emergency Knowledge of procedures relating to a 2.4.28 3.2 72 Procedures I security event (non-safeguards information).

Plan Knowledge of EOP entry conditions and 2.4.1 4.6 73 immediate action steps.

Subtotal 2 1 Tier 3 Point Total 10 7

ES-401 Record of Rejected K/As Form ES-401-4 Tier I Randomly Selected KIA Reason for Rejection 295027 High Containment This topic applies to plants with Mark Ill Temperature containments only. The facility has a Mark I 1 /1 containment.

295011 High Containment This topic applies to plants with Mark Ill Temperature containments only. The facility has a Mark I 1 /2 containment.

207000 Isolation (Emergency) This system is not installed at the facility.

2/1 Condenser 209002 HPCS This system is not installed at the facility.

2/1 201004 RSCS This system is no longer installed at the facility.

2/2 201005 RCIS This system is not installed at the facility.

2/2 2.2.3 Knowledge of the design, This KIA applies to multi-unit facilities only.

procedural, and operational G differences between units.

2.2.4 Ability to explain the Thi:s KIA applies to multi-unit facilities only.

variations in control board/control room layouts, G systems, instrumentation, and procedural actions between units at a facility.

HPCS is not installed at the facility.

Question 48 295030 Low Suppression Pool Randomly re-sampled 295030 Low Water Level Suppression Pool Water Level EA1.05- Ability EA1.03- Ability to operate to operate and/or monitor the following as they 1 /1 and/or monitor the following as apply to LOW SUPPRESSION POOL WATER they apply to LOW LEVEL: HPCI.

SUPPRESSION POOL WATER LEVEL: HPCS: Plant-Specific Question 50 RRCS is not installed at the facility.

295025 High Reactor Pressure EA 1.08 -Ability to operate Randomly re-sampled 295025 High Reactor 1 /1 and/or monitor the following as Pressure EA 1.07 -Ability to operate and/or they apply to HIGH REACTOR monitor the following as they apply to HIGH PRESSURE: RRCS: Plant- REACTOR PRESSURE: ARI/RPT/ATWS:

Specific Plant-Specific.

Question 52 Randomly selected KIA is identical to the KIA for Question 77. Re-sampling this KIA to limit 295019 Partial or Complete overlap.

Loss of Instrument Air AA2.02 -Ability to determine and/or interpret the following as Randomly re-sampled 295019 Partial or 1 /1 they apply to PARTIAL OR Complete Loss of Instrument Air AA2.01 -

COMPLETE LOSS OF Ability to determine and/or interpret the INSTRUMENT AIR: Status of following as they apply to PARTIAL OR safety-related instrument air COMPLETE LOSS OF INSTRUMENT AIR:

system loads (see AK2.1 - Instrument air system pressure.

AK2.19)

Question 62 Isolation Condensers are not installed at the faclility.

295007 High Reactor Pressure AA1.01 -Ability to operate and/or monitor the following as Randomly re-sampled 295007 High Reactor 1 /2 they apply to HIGH REACTOR Pressure AA 1.04- Ability to operate and/or PRESSURE: Isolation monitor the following as they apply to HIGH condenser: Plant-Specific REACTOR PRESSURE: Safety/relief valve operation: Plant-Specific.

Question 2 High pressure core spray (HPCS) is not installed at the facility.

223002 PCIS/Nuclear Steam Supply Shutoff K1.15 - Knowledge of the Randomly re-sampled 223002 PCIS/Nuclear physical connections and/or Steam Supply Shutoff K 1.19 - Knowledge of the cause- effect relationships physical connections and/or cause- effect 2/1 between PRIMARY relationships between PRIMARY CONTAINMENT ISOLATION CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SYSTEM/NUCLEAR STEAM SUPPLY SHUT-SUPPLY SHUT-OFF and the OFF and the following: Component cooling following: High pressure core water systems.

spray: Plant-Specific Question 5 Rod control and information system (RCIS) is not installed at the facility.

215004 Source Range Monitor K3.03- Knowledge of the effect that a loss or malfunction of the Randomly re-sampled 215004 Source Range 2/1 SOURCE RANGE MONITOR Monitor K3.02- Knowledge of the effect that a (SRM) SYSTEM will have on loss or malfunction of the SOURCE RANGE following: Rod control and MONITOR (SRM) SYSTEM will have on information system: Plant- following: Reactor manual control: Plant-Specific Specific.

Question 94 Randomly selected KIA is identical to the KIA for Question 66. Re-sampling this KIA to limit 2.1.14- Knowledge of criteria or overlap.

conditions that require plant-wide announcements, such as G pump starts, reactor trips, mode changes, etc. Randomly re-sampled 2.1.19 -Ability to use plant computers to evaluate system or component status.

Question 68 Randomly selected KIA is identical to the KIA for Question 95. Re-sampling this KIA to limit 2.2.23 -Ability to track overlap.

Technical Specification limiting conditions for operations.

G Randomly re-sampled 2.2.2 - Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Question 27 There is no direct connection or relationship between the Reactor Feedwater and RHR 259001 Reactor Feedwater systems.

K 1. 15 - Knowledge of the physical connections and/or 2/2 cause- effect relationships Randomly re-sampled 259001 Reactor between REACTOR Feedwater K1.06- Knowledge ofthe physical FEEDWATER SYSTEM and the connections and/or cause- effect relationships following: RHR: Plant-Specific between REACTOR FEEDWATER SYSTEM and the following: Plant air systems Question 37 There is no less than one hour technical specification related to Nuclear Boiler 216000 Nuclear Boiler Instrumentation.

Instrumentation 2/2 2.2.39- Equipment Control:

Knowledge of less than one Randomly re-sampled 216000 Nuclear Boiler hour technical specification Instrumentation 2.1.32- Ability to explain and action statements for systems. apply system limits and precautions.

Question 46 A discriminating question at the appropriate license level was unable to be developed with 295003 Partial or Complete the plant reference material and the randomly Loss of AC selected KIA AK3.04- Knowledge of the reasons for the following 1 /1 responses as they apply to Randomly re-sampled 295003 Partial or PARTIAL OR COMPLETE Complete Loss of AC AK3.03- Knowledge of LOSS OF A.C. POWER: the reasons for the following responses as they Ground isolation apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Load shedding.

Question 67 This KIA is tested extensively during the operating portion of the examination. A

2. 1.17 - Ability to make discriminating question was unable to be accurate, clear and concise developed without significant overlap with the verbal reports.

operating examination.

3 Randomly re-sampled 2.1.5 -Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Question 72 This generic KIA is also tested on Question 64, Question 84, and the operating portion of the 2.4.45 -Ability to prioritize and examination. A discriminating question was interpret the significance of unable to be developed without significant each annunciator or alarm.

overlap with these other areas.

3 Randomly re-sampled 2.4.28- Knowledge of procedures relating to a security event (non-safeguards information).

Question 92 This system does not have a significant impact on EOP mitigation strategies. This system is 201006 RWM also tested on Question 34, Question 69, and 2.4.6 - Emergency Procedures I September 2012 NRC Question 23. A valid Plan: Knowledge of EOP question was unable to be developed without 2/2 mitigation strategies. significant overlap with these other areas.

Randomly re-sampled new system 202001 Recirculation.

Question 94 This generic KIA is tested extensively during the operating portion of the exam and does not 2.1.19 lend itself to construction of an adequate SRO-Ability to use plant computers to level question.

evaluate system or component 3 status.

Randomly re-sampled 2.1.5- Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Question 13 The facility has replaced motor generators with static inverters.

262002 UPS (AC/DC)

A1.02- Ability to predict and/or monitor changes in parameters Randomly re-sampled 262002 UPS (AC/DC) 2/1 associated with operating the K4.01 -Knowledge of UNINTERRUPTABLE UNINTERRUPTABLE POWER POWER SUPPLY (A.C./D.C.) design feature(s)

SUPPLY (A. C./D.C.) controls and/or interlocks which provide for the including: Motor generator following: Transfer from preferred power to outputs alternate power supplies.

Question 29 Since the facility does not have Hydrogen Recombiners, a discriminating question was 223001 Primary Containment unable to be developed with the randomly and Auxiliaries selected KIA.

K3.04- Knowledge of the effect Randomly re-sampled 223001 Primary that a loss or malfunction of the Containment and Auxiliaries K3.03- Knowledge 2/2 PRIMARY CONTAINMENT of the effect that a loss or malfunction of the SYSTEM AND AUXILIARIES PRIMARY CONTAINMENT SYSTEM AND will have on following:

AUXILIARIES will have on following:

Containment/drywall hydrogen Containment/drywall pressure: Plant-Specific.

gas concentration

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. Fitzpatrick Date of Examination: April 2014 Examination Level: RO Operating Test Number: 14-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Verification of Core Thermal Power Conduct of Operations M,S KIA2.1.7 (4.4), RAP-7.3.03 Perform RHR Lineup Verification Conduct of Operations N,S KIA2.1.29 (4.1), ST-2AN Explain RPS Operation Using Electrical Drawings Equipment Control D,R KIA 2.2.41 (3.5/3.9), 1.67-99, 1.67-101 Emergency Announcement and Protected Area Emergency Procedures/Plan P, D, S Evacuation 2012-2 NRC KIA 2.4.43 (3.2), OP-63 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 1; randomly selected)

ES-301 Administrative Topics Outline Form ES-301-1 Facility: James A. FitzQatrick Date of Examination: AQril 2014 Examination Level: SRO Operating Test Number: 14-1 Administrative Topic Type Describe activity to be performed (see Note) Code*

Verification of Core Thermal Power Conduct of Operations M,S KIA 2.1.7 (4.7), RAP-7.3.3 Determine Reportability Requirements- Scram with HPCI and RCIC Injection Conduct of Operations N,R KIA2.1.18 (3.8), NUREG 1022, EN-LI-108 Explain RPS Operation Using Electrical Drawings, Determine Technical Specification Impact of Failed Equipment Control D,R Component KIA2.2.41 (3.9), 1.67-99,1.67-101 Determine Release Rates and ODCM Actions Radiation Control D,R KIA 2.3.11 (4.3), ISP-27-2, ODCM Determine Emergency Classification and Initiate Event Notification Emergency Procedures/Plan M,R KIA 2.4.40 (4.5), IAP-2, IAP-1 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 1; randomly selected)

ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: James A. Fitzpatrick Date of Examination: April 2014 Exam Level: RO/SR0-1 Operating Test No.: 14-1 Control Room Systems (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)

System I JPM Title Type Code* Safety Function

a. Reset RPS Scram with Scram Valve Fail to Close P, D, A, L, EN,S 7 KIA 212000 A4.14 (3.8/3.8), AOP-1 2012-1 NRC P, D, A, L,
b. Initiate RCIC in Pressure Control Mode with Speed Failure EN,S 4 KIA 217000 A4.01 (3.7/3.7}, OP-19 2012-2 NRC
c. Restore CRD to Normal Alignment Following ATWS, CRD Pump Trips M,A, L, S 1 KIA 201001 A2.07 (3.2/3.1 }, EP-3, OP-25, AOP-69
d. Feedwater Pump Restoration Following High Level Trip N, L, S 2 KIA 259001 A4.02 (3.9/3.7), OP-2A
e. Perform Area Radiation Monitor Functional Test D,S 9 KIA 272000 A4.02 (3.0/3.0), OP-32
f. Initiate Alternate Containment Spray from RHRSW D, A, L, S 5 KIA 226001 A4.08 (3.2/3.1), EP-14
g. Transfer Buses 10100 and 10200 from Normal to Reserve, Breaker 10212 Fails to Close M,A,S 6 KIA 262001 A4.04 (3.6/3.7), OP-46A
h. Isolate RBCLC Supply to the Drywell (RO Only)

D,L,S 8 KIA 400000 A4.01 (3.113.0), EP-12 In-Plant Systems (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)

i. Startup Main Steam Leakage Collection System D,L, E 9 KIA 239003 A4.01 (3.2/3.2), AOP-40
j. Perform In-Plant Actions for Station Blackout N, L, E 6 KIA 295003 AA1.04 (3.6/3.7}, AOP-49
k. Swap CRD Pump Suction Filter D,R 1 KIA 201001 A2.06 (2.9/2.9), OP-25

@ All RO and SR0-1 control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

  • Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S91S81S4 (E)mergency or abnormal in-plant <!:11<!:11<!:1 (EN)gineered safety feature - I - I <!:1 (control room system)

(l)ow-Power I Shutdown <!:11<!:11<!:1 (N)ew or (M)odified from bank including 1(A) <!:21<!:21<!:1 (P)revious 2 exams  ::; 3 I ::; 3 I ::; 2 (randomly selected)

(R)CA <!:11<!:11<!:1 (S)imulator

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NHC-2 Op-Test No.: 14-1 Examiners: Operators:

Initial Conditions: Reactor power is approximately 85% during a power reduction. Service Water Pump B is out of service for maintenance. SRV J is inoperable. L44 is cross-connected to L43.

Turnover: Restore L44 to the normal power source. Then continue power reduction to 70% with Recirculation flow.

Event Malf. Event Event No. No. Type* Description N- Restore L44 to Normal Power Source 1 N/A BOP, SRO OP-46A R- Reduce Power to ~r5% with Recirculation Flow 2 N/A ATC, SRO OP-65, RAP-7.3.16 Remote HPCI Aux Oil Pump Power Supply Loss 3 C-SRO HP15 ARP 09-3-3-38, Technical Specifications C- MCC 162 Electrical Fault and Drywell Cooling Fan 4A Fails to Auto-ED25:B 4 BOP, Start Override SRO ARP 09-8-3-29, Technical Specifications FW05:B Feedwater Pump B High Vibration and Pump Trip 5 C-All FW01:B ARP 09-6-4-18, AOP-41, AOP-8 FW05:A Feedwater Pump A High Vibration and Pump Trip 6 C-All FW01:A ARP-09-6-4-31, AOP-1, EOP-2 Coolant Leak in Drywell 7 RR15:A M-All EOP-2, EOP-4 C- First Torus Spray Valve Fails to Open 8 Override BOP, SRO EOP-4 C- Trip of All Condensate Pumps FW19:

9 ATC, (A-C) EOP-2 SRO

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

Facility: James A. Fitzpatrick Scenario No.: NRC-2 Op-Test No.: 14-1

1. Total malfunctions (5-8) 7 Events 3, 4, 5, 6, 7, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8, 9
3. Abnormal events (2-4) 3 Events 4, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-4
6. EOP contingencies requiring substantive actions (0-2) 2 EOP-2 Alternate Level Control Leg EOP-2 Emergency Depressurization Leg
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a coolant leak inside the Containment, the crew will spray the Drywell, in accordance with EOP-4.

CT-2: Given a coolant leak, a loss of high pressure injection systems, and the inability to restore and maintain Reactor water level above the Top of Active Fuel (TAF), the crew will initiate actions for an Emergency RPV Depressurization before Reactor water level lowers below -19", in accordance with EOP-2.

Scenario Summary:

The scenario will begin at approximately 85% power with Service Water pump B out of service for maintenance and SRV J inoperable. The crew will begin the shift by restoring Bus L44 to the normal power source per OP-46A. The crew will then continue by lowering Reactor power to 75% with Recirculation flow.

Power will be lost to the HPCI Auxiliary Oil pump. This will render HPCI inoperable for the remainder of the scenario. The SRO will determine the Technical Specification impact.

MCC 162 will de-energize due to an electrical fault and Drywell Cooling fan 4A will fail to auto-start. The crew will start Drywell Cooling fan per the ARP. Multiple safety-related electrical loads will lose power, including a Core Spray B injection valve. The SRO will determine the Technical Specification impact.

Feedwater pump B will develop a high vibration. The crew may lower Reactor power in anticipation of tripping the pump. When the vibration exceeds 6 mils, the crew will trip the pump. If the crew does not trip the pump by 9 mils, the pump will spuriously trip. A Recirculation run back is likely to occur following the pump trip. The crew will respond to stabilize the plant within the capacity of a single Feedwater pump.

Feedwater pump A will also develop a high vibration. The crew will scram the Reactor as the Feedwater pump vibration approaches the level requiring pump trip. If the crew has not already tripped Feedwater pump A, it will spuriously trip 30 seconds after the scram. With HPCI and both high pressure Feedwater pumps unavailable, the crew will utilize RCIC to control Reactor water level.

A coolant leak will develop in the Drywell. The crew will attempt to maintain Reactor water level with RCIC and will eventually need to lower Reactor pressure to inject with Condensate Booster pumps. The crew will place Torus spray in service. This will be complicated by the first attempted Torus spray valve failing to open. The crew will be able to place the other division of RHR in Torus spray. When Torus pressure exceeds 15 psig, the crew will also place Drywell spray in service. Drywell spray will lower Drywell pressure such that the Pressure Suppression Pressure curve is not challenged.

All Condensate pumps will sequentially trip and the coolant leak will get worse. With no Condensate injection and the coolant leak exceeding the capacity of RCIC, CRD, and SLC, Reactor water level will lower to the top of active fuel. The crew will execute an Emergency RPV Depressurization and then control low pressure injection systems to restore Reactor water level.

The scenario will be terminated when all control rods are inserted, the Emergency RPV Depressurization is in progress, and Reactor water level is controlled above 0".

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 14-1 Examiners: Operators:

Initial Conditions: Reactor power is approximately 90%. EHC pump A is out of service for maintenance.

Turnover: Perform ST-3PA, Core Spray Loop A Quarterly Operability Test. The procedure is in progress up to step 8.4.1.

Event Malf. Event Event No. No. Type* Description N- Perform ST-3PA, Core Spray Loop A Quarterly Operability Test 1 N/A BOP, SRO ST-3PA Core Spray A Outboard Injection Valve Fails to Open 2 Override C-SRO ST-3PA, Technical Specifications C- SRV A Inadvertently Opens 3 AD06:A BOP, SRO AOP-36, Technical Specifications R-ED19:E ATC, Fault on 10700 Bus and Service Water Pump A Fails to Auto-Start 4 SRO SW15 AOP-20 C-BOP Remote C- Degraded Stator Water Cooling Pump A Flow with Delayed Trip 5 BOP, Override SRO ARP 09-7-3-08, OP-11 B Remote C- Degraded Stator Water Cooling Pump B Flow with Delayed Trip 6 ATC, Override SRO ARP 09-7-3-08, AOP-1 RD10 Hydraulic ATWS 7 RD13 M-All EOP-2, EOP-3 RD22 CRD Flow Control Valve Fails Closed and ATWS/RPT Fails to RD03:A 1-ATC, 8 Initiate RR13 SRO EOP-3, EP-3 I-BOP, MSIVs Isolate 9 RP03 SRO EOP-3

  • (N)ormal, (R)eactivitv, (l)nstrument, (C)omoonent, (M)aior

Facility: James A. Fitzpatrick Scenario No.: NRC-3 Op-Test No.: 14-1

1. Total malfunctions (5-8) 8 Events 2, 3, 4, 5, 6, 7, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8, 9
3. Abnormal events (2-4) 4 Events 3, 4, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs entered/requiring substantive actions (1-2) 1 EOP-2
6. EOP contingencies requiring substantive actions (0-2) 1 EOP-3
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given a failure to scram with Reactor power above 2.5%, the crew will lower Reactor power by one or more of the following methods, in accordance with EOP-3:

  • Terminating and preventing all RPV injection except SLC, RCIC, and CRD,
  • Tripping Recirculation pumps,
  • Injecting SLC .

CT-2: Given a failure to scram, the crew will initiate Control Rod insertion, in accordance with EOP-3.

Scenario Summary:

The scenario will begin at approximately 90% power with EHC pump A out of service for maintenance. The crew will begin the shift by performing Core Spray A valve stroke timing per ST-3PA. Core Spray A Outboard Injection Valve 14MOV-11A will close and then fail to re-open.

The SRO will declare Core Spray A inoperable and determine the Technical Specification impact.

SRV A will inadvertently open. The crew will execute AOP-36 and close the SRV by pulling control power fuses. The crew will place Torus Cooling in service. The SRO will determine the Technical Specification impact.

An electrical fault will cause Bus 10700 to de-energize. The crew will execute AOP-20. Major loads lost include an operating Circulating Water pump, Service Water pump, Condensate pump, and Condensate Booster pump. The standby Service Water pump will fail to auto-start.

The crew will manually start the standby Service Water pump, lowering Reactor power to less than 65% using Recirculation flow and CRAM rods, and normalize Control Room controls.

Stator Water Cooling pump A flow will degrade. The crew will execute ARP 09-7-3-08, swap to Stator Water Cooling pump B, and restore flow. If Stator Water Cooling pump A is not secured, it will eventually trip. Stator Water Cooling pump B flow will also begin to degrade after being started. With an impending loss of both Stator Water Cooling pumps, the crew will enter AOP-1 and attempt a manual Reactor scram. If Stator Water Cooling pump B is not secured, it will eventually trip.

A hydraulic failure to scram will occur. The crew will enter EOP-2 and transition to EOP-3.

Recirculation pumps will fail to trip on an ATWS/RPT signal, requiring the crew to manually trip Recirculation pumps. The crew will also intentionally lower Reactor water level and inject boron.

The in-service CRD flow control valve will fail closed, requiring the crew to place the standby flow control valve in service to drive control rods with RMCS. The crew may also insert control rods using repeated manual scrams. The majority of control rods will be able to be inserted using repeated manual scrams. Approximately 15 control rods will only be able to be inserted using RMCS and approximately 10 control rods will be completely stuck and unable to be inserted.

The scenario will be terminated when control rod insertion is in progress or completed, Reactor power is downscale on APRMs, and Reactor water level is controlled above 0".

Appendix D Scenario Outline Form ES-D-1 Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 14-1 Examiners: Operators:

Initial Conditions: Reactor power is approximately?% during a startup. The Rod Worth Minimizer is bypassed. IRM F is bypassed Turnover: Return IRM F to service per OP-16. Then continue power ascension with control rod withdrawals.

Event Malf. Event Event No. No. Type* Description N- Return IRM F to Service 1 N/A BOP, SRO OP-16 R- Withdraw Control Hods 2 N/A ATC, SRO OP-65, OP-26 C- Stuck Control Rod 3 RD10 ATC, SRO AOP-25 RPS Level Transmitter Fails High 4 RR22:B 1-SRO ARP 09-5-2-60, Technical Specifications CU07 Reactor Water Cleanup Pump Seal Failure, Reactor Water C-5 CU10 BOP, Cleanup Fails to Automatically Isolate SRO ARP-09-3-3-2(12), EOP-5, Technical Specifications CU12 RC06 C- RCIC Inadvertent Initiation and High Bearing Temperature 6 BOP, Override SRO EN-OP-115, ARP-09-4-1-15, Technical Specifications RC09 RCIC Steam Leak and Failure to Isolate 7 RC12 M-All EOP-5, AOP-1, EOP-2 Remote RP01A 1-ATC, RPS Fails to Scram, ARI Inserts Control Rods 8

RP01B SRO AOP-1, EOP-2 C- Multiple SRVs Fail to Open 9 ADO? BOP, SRO EOP-2

  • (N_lormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor

Facility: James A. Fitzpatrick Scenario No.: NRC-4 Op-Test No.: 14-1

1. Total malfunctions (5-8) 7 Events 3, 4, 5, 6, 7, 8, 9
2. Malfunctions after EOP entry (1-2) 2 Events 8, 9
3. Abnormal events (2-4) 3 Events 3, 5, 6
4. Major transients (1-2) 1 Event 7
5. EOPs entered/requiring substantive actions (1-2) 2 EOP-2, EOP-5
6. EOP contingencies requiring substantive actions (0-2) 1 EOP-2 Emergency Depressurization Leg
7. Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:

CT-1: Given the plant operating at power with an un-isolable primary system discharging into Secondary Containment, the crew will insert a manual Reactor scram, in accordance with EOP-5.

CT-2: Given an un-isolable primary system discharging into Secondary Containment and two areas exceeding Maximum Safe Temperatures, the crew will perform an emergency RPV depressurization, in accordance with EOP-5.

Scenario Summary:

The scenario will begin at approximately 7% power during a startup. The Rod Worth Minimizer is bypassed. IRM F is bypassed and ready to be returned to service following I&C repairs. The crew will begin by restoring IRM F to service per OP-16 Section E.8.1. Then the crew will continue to raise Reactor power by withdrawing control rods.

The third control rod to be moved, 38-15, will be stuck at its initial position of 04. The crew will enter AOP-24, Stuck Control Rod. The control rod will move once CRD drive water pressure is raised above 300 psig. Then the crew will continue with control rod withdrawals per the startup sequence.

Reactor water level transmitter, 02-3LT-101 B, will fail upscale. This transmitter is one of the inputs to the RPS scram function. With the transmitter failed high, the SRO will determine that it cannot perform its scram function, declare that function inoperable, and determine the Technical Specification impact.

Reactor Water Cleanup pump A develops a seal failure. This will cause at least one Reactor Building area temperature to exceed max normal and require entry into EOP-5, Secondary Containment Control. The Reactor Water Cleanup pumps will eventually trip on high room temperature, but the suction line will fail to automatically isolate. The crew can isolate the leak by manually closing 12MOV-15 and/or 12MOV-18. The SRO will determine the Technical Specification impact of the PC IV automatic isolation failure.

RCIC will inadvertently initiate. Once RCIC comes up to near rated flow, the initiation signal will clear, allowing full control of the system. Shortly thereafter, a high bearing temperature will occur. The crew will secure RCIC. The SRO will determine the Technical Specification impact.

RCIC will develop a steam leak. This will cause high area temperatures near RCIC and HPCI.

HPCI will automatically isolate, however RCIC will fail to automatically isolate. The crew will be able to close 13MOV-16, however the breaker for 13MOV-15 trip open, preventing isolation of the steam leak. The crew will execute EOP-5. The crew will enter AOP-1 and insert a manual Reactor scram. On the scram attempt, the RPS push buttons and Mode Switch will fail to work.

The crew will insert control rods by initiating ARI. The crew will enter EOP-2, RPV Control. As Reactor Building area temperatures approach max safe levels, the crew will anticipate Emergency Depressurization by rapidly lowering Reactor pressure with Turbine Bypass Valves.

Once two max safe temperatures are exceeded, the crew will perform an Emergency Depressurization. Two of the ADS valves will fail to open. The SRO will direct two other SRVs open, but only one other SRV will open. The SRO will determine that Emergency Depressurization requirements are still met with only six SRVs open.

The scenario will be terminated when all control rods are inserted, the Emergency RPV Depressurization is in progress, and Reactor water level is controlled above 0".

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