ML081400330

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Draft - Outlines (Folder 2)
ML081400330
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 02/28/2008
From: Todd Fish
Operations Branch I
To: Peter Dietrich
Entergy Nuclear Northeast
Hansell S
Shared Package
ML072850856 List:
References
50-333/08-301 50-333/08-301
Download: ML081400330 (27)
v • d • e


Text

2008 FitzPatrick NRC Exam Rev1 and SRO-only outlines (i.e.., except for one category in Tier 3 each KIA category shall not be less than two).

The final point total for each group and tier may deviate by f l 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. Systemslevolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to ES-401, Attachment 2, for guidance regarding the elimination of inappropriate KIA statements.

group before selecting a second topic for any system or evolution.

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

the SRO-only outline, the Tier Totals in

2. The point total for each group and tier in the proposed outline must match that specified in the table.
4. Select topics from as many systems and evolutions as possible; sample every system or evolution in the
5. Absent a plant-specific priority, only those WAS having an importance rating (IR) of 2.5 or higher shall be
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.
8. On the following pages, enter the KIA numbers, a brief description of each topic, the topics importance ratings (IRs) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above; if fuel handling equipment is sampled in other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 for Tier 2, Group 2
9. For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, IRs, ant point totals (#) on Form ES-401-3. Limit SRO selections to WAS that are linked to 10 CFR 55.43.

.a..

Summary

Year Facility NRC Exam Rev 0 II ElAPE # I Name I Safety Function 295001 Partial or Complete Loss of Forced Core Flo Circulation I 1 & 4 295003 Partial or Complete Loss of A.C. Power I 6 2.02 295004 Partial or Complete Loss of D.C. Power I6 3(

295005 Main Turbine Generator Trip I 3 I ! ]

295006 SCRAM I 1 295016 Control Room Abandonment I 7 295018 Partial or Complete Loss of Component Cooling Water I8 1.01 295019 Partial or Complete Loss of Instrument Air 18 2 09 295021 Loss of Shutdown Cooling I21 3c 295023 Refueling Accidents I8 295024 High Drywell Pressure I5 295025 High Reactor Pressure I 3 KIA Topic(s)

Imp.

A A

G 1

2 2

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following:

(CFR: 41.7 I45.8) AK2.02 Emergency generators 4.1 2

(CFR: 41.8 to 41.IO) AK1.02 Power/flow distribution 3.3 1

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C.

POWER :

(CFR: 41.5 I45.6) AK3.03 Reactor SCRAM: Plant-Specific 3.1 3

Ability to operate andlor monitor the following as they apply to MAIN TURBINE GENERATOR TRIP :

Ability to determine and/or interpret the following as they apply to SCRAM :

2.4.1 2 Knowledge of general operating crew responsibilities during emergency operations.

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

(CFR: 41.7 / 45.8) AK2.09 Containment 3.3 8

Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.5 / 45.6) AK3.05 Establishing alternate heat removal flow paths 3.6 9

Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS :

Ability to determine andlor interpret the following as they apply to HIGH DRYWELL PRESSURE:

1 04 (CFR: 41.7 / 45.6) AA1.04 Main generator controls 2.7 4

2 03 (CFR 41.10 I43.5 145.13) AA2.03 Reactor water level 4.0 5

24.12 (CFR:41.10/45.12) 3.4 6

(CFR: 41.8 to 41.IO) AK1.01 Effects on componentkystern operations 3.5 7

1 02 (CFR: 41.7 I 45.6) AA1.02 Fuel pool cooling and cleanup system 2.9 10 2 02 I(CFR: 41:lO / 43.5 / 45.13) EA2.02 Drywell temperature I

3.9 I 11 I

I 12.1.7 Ability to evaluate plant performance and make operational judgments I I

based on operating characteristics / reactor behavior / and instrument 2.1.7 interpretation. (CFR: 43.5 I45.12 I45.13) 3.7 EAPEsT1 G1

Year Facility NRC Exam Rev 0 295030 Low Suppression Pool Water Level I 5 295031 Reactor Low Water Level I 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown I 1 295038 High Off-Site Release Rate I 9 600000 Plant Fire On Site I 8 1 02 295018 Partial or Complete Loss of Component KIACategoryTotals:

4 1

Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE :

(CFR: 41.8 to 41.10) EKI.O1 Pump NPSH 3.0 13 NIA JAF Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:

(CFR: 41.7 145.8) EK2.03 Reactor water level indication 3.6 14 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL:

Abtldy to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL :

Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

3 06 (CFR: 41.5 145.6) EK3.06 Reactor SCRAM 3.6 15 113 (CFR: 41.7 145.6) EA1.I3 Reactor water level control 4.3 16 2 07 (CFR: 41. I O 143.5 145.13) EA2.07 Containment conditionsfisolations 4.0 17 2 4 18 2.4.18 Knowledqe of the specific bases for EOPs. (CFR: 41.10 I45.13) 2 7 i a Knowledge of the operation applications of the I

following concepts as they apply to Plant Fire On Site:

I l l

-r I

I I

lAKl.02 Fire Fighting 19 IKnowledqe of the interrelations between PARTIAL OR I

I COMPLETE LOSS OF COMPONENT COOLING WATER and the following:

(CFR: 41.7 I 45.8) AK2.01 System loads 3.3 20 EAPEs T1 G1

Facility NRC Exam Year Rev 0 EAPEsT1 G1

Year Facility NRC Exam Rev 0 295013 High Suppression Pool Temperature I 1 3,

uppression pool cooling Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE :

Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS:

Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION :

Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL 295017 High Off-Site Release Rate I9 2 04 (CFR: 41.10 143.5 / 45.13) AA2.04 $Source of off-site release 3.6 23 295022 Loss of CRD Pumps I 1 1 02 (CFR: 41.8 to 41.lo) AK1.02 Reactivity control 3.6 24 295032 High Secondary Containment Area Temperature I5 3 01 TEMPERATURE : EK3.01 Emergencyhormal depressurization 3.5 25 295034 Secondary Containment Ventilation High Radiation I9 2 01 (CFR: 41.10 143.5 I 45.13) EA2.01 Ventilation radiation levels 3.8 26 295036 Secondary Containment High SumpIArea Water Level I5 101

(CFR: 41.8 to 41.10) EK1.01 Radiation releases 2.9 27

~~

EAPEs T1 G2

0

> 2

Year Facility NRC Exam Rev 0 EAPEs T1 G2

Year Facillty NRC Exam Rev 0 I

System # I Name H

203000 RHWLPCI: Injection Mode 205000 Shutdown Cooling System (RHR Shutdown Cooling Mode) 206000 High Pressure Coolant injection System 207000 Isolation (Emergency) Condenser I I I

I 209001 Low Pressure Core Spray System 209002 High Pressure Core Spray System (HPCS) 211000 Standby Liquid Control System 3 212000 Reactor Protection System 1.04 215003 Intermediate Range Monitor (IRM) 215004 Source Range Monitor (SRM)

System 215005 Average Power Range MonitorlLocal Power Range Monitor Svstem nditions or operations:

ge tank level: BWR-(CFR: 41.7 145.5 to 45.8) A4.02 Suction valves N/A JAF Knowledge of the physical connections and/or causeeffect relationships between REACTOR PROTECTION SYSTEM and the following:

(CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.04 A.C. electrical distribution Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM)

SYSTEM and the following:

(CFR 41.2 to 41.9 I45.7 to 45.8) K1.01 RPS Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following:

(CFR: 41.7 I45.4) K3.02 Reactor manual control: Plant-SDecific I 3.4 I 35 Knowledge of AVERAGE POWER RANGE MONITOFULOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7) K4.07 Flow biased trip setpoints 3.7 36

~~~

~~

SYSTEMS T2 G1

Year Facility NRC Exam COOLING SYSTEM (RCIC) :

(CFR: 41.5 145.3) K5.06 Turbine operation Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM :

(CFR: 41.7 145.7) K6.01 RHFULPCI system pressure: Plant-Specific Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEMINUCLEAR STEAM SUPPLY SHUT-OFF controls including:

Ability to (a) predict the impacts of the following on the RELlEFlSAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR: 41.5 145.6) A2.02 Leaky SRV Ability to monitor automatic operations of the REACTOR WATER LEVEL CONTROL SYSTEM including:

(CFR: 41.7 145.7) A3.04 Changes in reactor feedwater flow 3.2 3.2 Ability to manually operate andor monitor in the control room:

(CFR: 41.5 145.5) Al.02 Valve closures Rev 0 2.7 37 3.9 38 3.7 39 3.1 40 3.2 41 I

I I

I 2 4 48 2 411 status and operation of system I and understand how operator action s and directives affect plant and system conditions. (CFR:

Knowledge of the physical connections and/or cause effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C.D.C.) and the following:

(CFR: 41.2 to 41.9 I 45.7 to 45.8) K1.01 Feedwater level control:

Knowledge of electrical power supplies to the following:

Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEUJET) will have on following:

(CFR: 41.7 I 45.4) K3.02 A.C. electrical distribution 3.9 46 Knowledge of (INSTRUMENT AIR SYSTEM) design feature(s) and or interlocks which provide for the following:

2.8 47 2.4.1 1 Knowledge of abnormal condition procedures. (CFR:

41.10143.5145.13) 2.8 48 43.5 145.12) 3.5 43 Plant-Specific 2.8 44 (CFR: 41.7) K2.01 Major D.C. loads 3.1 45 (CFR: 41.7) K4.03 Securing of IAS upon loss of cooling water I07 l(CFR: 41.7 145.5 to 45.8) A4.07 System flow 1

3.1 I 42 I

12.4.48 Ability to interpret control room indications to verify the I

I SYSTEMS T2 G1

Year Facility NRC Exam Rev 0 SYSTEMS T2 G1

Year Facility NRC Exam Rev 0 SYSTEMS T2 G1

Year Facility NRC Exam Rev 0 I

System # I Name 201003 Control Rod and Drive Mechanism 202001 Recirculation System P 202002 Recirculation Flow Control System 216000 Nuclear Boiler Instrumentation 219000 RHRILPCI: ToruslSuppression Pool Cooling Mode 230000 RHRILPCI: ToruslSuppression Pool Spray Mode 290002 Reactor Vessel lnternais 241000 Reactorflurbine Pressure A

G A

I 3 I WA Topic($

Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

(CFR 41.5 145.6) A2.02 Uncoupled rod 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics I reactor behavior /

and instrument interpretation.(CFR: 43.5 I45.12 I45.13)

Knowledge of the physical connections andlor cause effect relationships between RECIRCULATION FLOW CONTROL SYSTEM and the following:

(CFR: 41.2 to 41.9 I45.7 to 45.8) K1.08 Feedwater flow 2 1 7 Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION :

(CFR: 41.7 145.7) K6.02 D.C. electrical distribution Knowledge of electrical power supplies to the following:

(CFR: 41.7) K2.02 Pumps Ability to manually operate and/or monitor in the control room:

(CFR: 41.7 I45.5 to 45.8) A4.06 Valve logic reset following automatic initiation of LPCIlRHR in iniection mode Knowledge of the physical connections and/or cause effect relationships between FUEL HANDLING EQUIPMENT and the following:

(CFR: 41.2 to 41.9 145.7 to 45.8) K1.04 ?Reactor manual control Knowledge of the effect that a loss or malfunction of the REACTOF VESSEL INTERNALS will have on following: (CFR: 41.7 I45.6)

Knowledge of REACTORRURBINE PRESSURE REGULATING SYSTEM design feature(s) and/or interlocks which provide for the following:

(CFR: 41.7)

K4.01 Reactor pressure control 1 04 system: Plant-Specific 3 03 K3.03 Reactor power 4 01 3.7 54 3.7 55 3.1 56 2.8 57 3.1 58 4.0 59 3.3 60 SYSTEMS T2 G2

Year Facility NRC Exam Rev 0 Ability to predict andlor monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including:

(CFR: 41.5 I45.5) Al.05 RFP turbine control valve position: Turbine 259001 Reactor Feedwater System 1 05 Driven-Only 2.8 63 +

Ability to monitor automatic operations of the OFFGAS SYSTEM including:

(CFR: 41.7 145.7) A3.05 System indicating lights and alarms Ability to manually operate andlor monitor in the control room:

(CFR: 41.7 I45.5 to 45.8) A4.01 Svstem alarms and indicatinq 271000 Offgas System 3.05 SYSTEMS T2 G2

Year Facility NRC Exam Rev 0 Ability to (a) predict the impacts of the following on the FUEL HANDLING EQUIPMENT : and (bl based on those Dredictions. use 234000 Fuel Handling Equipment I

,IVA Category Totals:

0 0

0 0

0 0

0 2

0 0

1 Group Point Total:

3 3

SYSTEMS T2 G2

Year Facility NRC Exam Rev 0 Cat ego ry I

Conduct of Operations 2

Equipment Control ToDic KIA #

2.1.I 1 Knowledge of less than one hour technical specification action statements for systems.

(CFR: 43.2 / 45.13)

IMPORTANCE RO 3.0 SRO 3.8 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.

(CFR: 43.2 / 43.3 / 45.3)

IMPORTANCE RO 3.4 SRO 4.0 2.1.I8 Ability to make accurate / clear and concise logs / records / status boards / and reports.

(CFR: 45.12 / 45.13)

IMPORTANCE RO 2.9 SRO 3.0 2.1.I 1

2.1.33 2.1.I8 2.2.30 Knowledge of RO duties in the control room during fuel handling such as alarms from fuel handling area / communication with fuel storage facility / systems operated fron the control room in support of fueling operations / and supporting instrumentation.

(CFR: 45.12)

IMPORTANCE RO 3.5 SRO 3.3 2.2.22 Knowledge of limiting conditions for operations and safety limits.

(CFR: 43.2 / 45.2) 2.2.30 2.2.22 IMPORTANCE RO 3.4SRO 4.1 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

(CFR: 41.12 / 43.4. 45.9 / 45.10) 2.3.1 I IMPORTANCE RO 2.6 SRO 3.0 Imp. -

3.c 3.5 3.L -

A 2.f 66 67 68 3 -

69 70 2

71 Generics

Year 2.3.1 1 2.3.10 Facility NRC Exam Rev 0 2.3.1 1 Ability to control radiation releases.

(CFR: 45.9 / 45.1 0)

IMPORTANCE RO 2.7 SRO 3.2 2.3.1 0 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.

(CFR: 43.4 / 45.10)

IMPORTANCE RO 2.9 SRO 3.3 2

2 3

4 Emergency Procedures and Plan Radiation Protection 2.4.29 Knowledge of the emergency plan.

(CFR: 43.5 / 45.1 1) 2.4.18 Knowledge of the specific bases for EOPs.

(CFR: 41.10 / 45.13) 2.4.29 IMPORTANCE RO 2.6 SRO 4.0 2.6 74 2.4.1 8 IMPORTANCE RO 2.7 SRO 3.6 2.7 75 Subtotal Tier 3 Point Total 10 72 73 Generics

Year Facility NRC Exam Rev 0 Catenorv 1

Conduct of Operations 2

Equipment Control 3

Radiation Protection KIA #

Topic Imp.

2.1.I2 Ability to apply technical specifications for a system.

(CFR: 43.2 / 43.5 / 45.3)

IMPORTANCE RO 2.9 SRO 4.0 2.1.10 Knowledge of conditions and limitations in the facility license. (CFR: 43.1 / 45.13) 2.1.I2 2.1.10 IMPORTANCE RO 2.7 SRO 3.9 3.9 95 Subtotal 2

2 2.2.8 Knowledge of the process for determining if the proposed change / test / or experiment involves an unreviewed safety question.

(CFR: 43.3 / 45.1 3) 2.2.8 IMPORTANCE RO 1.8SRO 3.3 3.3 96

.. 4 Ability to analyze the aitect ot maintenance activities on LCO status.(U-K'.

43 2 /

45.13) 2.2.24 IMPORTANCE RO 2.6SRO 3.8 3.8 97 Subtotal 2

2 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.9. / waste disposal and handling systems).

(CFR: 43.4 / 45.1 0) 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.

(CFR: 41.I 2 / 43.4. 45.9 / 45.1 0) 2.3.3 IMPORTANCE RO 1.8SRO 2.9 2.9 98 2.3.1 IMPORTANCE RO 2.6 SRO 3.0 3

99 I

Subtotall 21 2

Generics

Year 4

Emergency 11 Procedures and Plan L I

I Facility NRC Exam Rev 0 I

I I

I I

I I

7 Tier 3 Point Total1 7

Generics

Year T I GI T2 G2 T2 G I Facility NRC Exam 295038 G2.4.10 272000 A4.03 400000 G2.4.43 Generic dealt with use of ARPs for high off-site release rate, randomly selected another generic Unable to write discriminating question for this WA, randomly selected another Tier 2 system Unable to write discriminating question for this WA, randomly selected another generic Rev 0 Tierand I

I 1

Group I

Randomly Selected KIA I

Reason for Rejection I

i I

I 1

1 ES 401-4

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1 Facility:

FitzPatrick Date of Examination: 3/31/08 to 4/10/08 Administrative Topic (see Note)

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

N M

Describe activity to be performed Manually Compute Average Drywell Air Temperature Work Hour Restrictions Complete a Tagout

~

N/A ERO CallOut 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 (D)irect from bank (5 3 for ROs; I 4 for SROs & RO retakes)

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

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

(S)imulator

ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1 Conduct of Operations Conduct of Operations Equipment Control I

I Radiation Control ll Emergency Plan I

Type Code*

Facility: FitzPatrick Date of Examination: 3/31/08 to 4/10/08 Examination Level (circle one): RO / SRO Operating Test Number:

1 Administrative Topic (see Note)

N N

M Describe activity to be performed Manually Compute Average Drywell Air Temperature Work Hour Restrictions Complete a Tagout Canal Discharge Approval

~

Post Scenario Event Classification 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 (D)irect from bank (I 3 for ROs; _< 4 for SROs & RO retakes)

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

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

(S)imulator

ES-301, Rev. 9 Control Room/ln-Plant Systems Outline Form ES-301-2

a. Component Cooling Water / Change In-Service RBCLC Pumps
b. Recirculation / Start Reactor Recirc Pump 1

A, N, S A, N, S Facility:

FitzPatrick 11 Exam Level (circle one): RO / SRO-I / SRO-U

e. Primary Containment System / Respond to High Drywell Temperature
f. Low Pressure Core Spray / Core Spray Surveillance Test
g. Emergency Generators / Load Test of B Emergency Diesel
h. APRM/LPRM / Restore Inoperable LPRM Date of Examination: 3/31/08 to 4/10/08 Operating Test No.:

1 A, N, s N, L, s D, L, s N, s I 1) Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, includina 1 ESF)

System / JPM Title Type Code*

Safety Function 8

1 11 c. High Pressure Coolant Injection / Shutdown HPCI I

A,D,S 1

2

1) d. Main and Reheat Steam / MSlV Surveillance Test I

N,S I

3 5

4 6

7

)I i. Safety Relief Valves / Pull Fuses for Stuck Open SRV

)I j. CRD Hydraulics / Change In-Service Flow Control Valves 1

k. Emergency Generators / SBO Start of EDG Q

N, E 6

All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

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

(P)revious 2 exams (R)CA (S)imulator 4-6 14-6 / 2-3 I ~ I I ~ I G 2 1 I 2 1 I 2 1 2 1 / 2 1 / 2 1 2 2 / 2 2 / 2 1 I

3 I I 3 I I 2 (randomly selected) 2 1 I 2 1 I 2 1

ES-301, Rev. 9 Control Roodln-Plant Systems Outline Form ES-301-2 System / JPM Title Type Code*

Safety Function

a. Component Cooling Water / Change In-Service RBCLC Pumps 1

A,N,S I

8

c. High Pressure Coolant Injection / Shutdown HPCl
d. Main and Reheat Steam / MSlV Surveillance Test
e. Primary Containment System / Respond to High Drywell Temperature
b. Recirculation / Start Reactor Recirc Pump I

A,N,S 1

1 A, D, S 2

N, s 3

A, N, S 5

i. Safety Relief Valves / Pull Fuses for Stuck Open SRV
j. CRD Hydraulics / Change In-Service Flow Control Valves
f. Low Pressure Core Spray / Core Spray Surveillance Test D, R 3

D, E 1

g. Emergency Generators / Load Test of B Emergency Diesel
h. N/A
k. Emergency Generators / SBO Start of EDG Q

All control room (and in-plant) systems must be different and serve different safety functions; in-plant systems and functions may overlap those tested in the control room.

Type Codes Criteria for RO / SRO-I / SRO-U (A)lternate path 4-6 / 4-6 12-3 (C)ontrol room (D)irect from bank 19/58/14 (E)mergency or abnormal in-plant (L)ow-Power J Shutdown (N)ew or (M)odified from bank including 1(A)

( W A 1 1 / > I 111 2 1 / 2 1 I 2 1 21 I 2 1/21 22/12/21 I3 / I 3 / I2 (randomly selected)

(P)revious 2 exams (S)imulator

Appendix D Scenario Outline Form ES-D-I Event No.

1 2

3 Facility: -FitzPatrick Scenario No.:

1 Malf.

Event Event No.

Type*

Description Trigger 1 N(R0)

Swap in service CRD pumps.

Trigger 2 Trigger 3 C (BOP)

R(R0) 5% power reduction.

HPCl Aux Oil Pump breaker failure.

Op-Test No.:

c (BOP)

ALL)

Trigger 4 Trigger 5 4

5 Examiners: Fish Fuller (UI)

Operators:

Johnson Presby

'A' RFP trip and failure of Recirc runback.

Loss of all CRD pumps Initial Conditions: -87% power -(MOL)

Turnover: -Operating normally at 87%, after assuming shift will continue to 65% in preparation for cleaning condenser water boxes; Perform a swap of CRD pumps to allow maintenance to record vibration data on the 'B' CRD pump 7

1 Trigger 1 C (BOP) I Main turbine trip, loss of Auxiliary busses I c (BOP) I RCIC controller failure.

I Trigger 8 8

9 1 Trigger 1 M (ALL) 1 Emergency Depressurization (N)ormal, (R)eactivity, (I)nstrurnent, (C)ornponent, (M)ajor

Appendix D Scenario Outline Form ES-D-1 Event I (RO)

Type*

~

Facility: -FitzPatrick Scenario No.:

3 Op-Test No.:

Event Description APRM instrument failure.

Examiners: Fish Fuller (UI)

Operators:

Johnson Presby (SRO) c (ALL)

Initial Conditions: 50% (MOL Unisolable leak on Instrument Air header Turnover: -The plant at 50% power. Power ascension was suspended to allow Electrical Maintenance to perform an inspection of T4 following a report of a cooling fan problem. Busses 10200 and 10400 are currently fed from Reserve. Electrical Maintenance has successfully completed the inspection of T4. The shift will restore the electric plant lineup to normal by transferring busses 10200 and 10400 from Reserve to Normal.

Raise reactor power with control rods. Start at Step 46 with rod 30-19. Pull from notch 12 to 24.

Continue power ascension per OP-65

~

~~

M (ALL)

ALL)

M (ALL)

Event Malf.

-1 No.

No.

Trigger 1 SBLOCA on recirc line, containment spray required Spray torus, both loop of RHR fail Spray Torus/Drywell with RHR SW Trigger 2 Trigger 3 Trigger 4 Trigger 5 2

3 4

5 Trigger 6 6

N (ALL) 1 Restoration of busses 10200 and 10400 to T4.

RGO) I Raises power with control rods

~

~~

C(BOP)pn a d ve rt e n t H PC I in it i at i o n (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor

Appendix D Scenario Outline Form ES-D-I Event Malf.

Event No.

No.

Type*

1 Trigger 1 C (BOP)

Facility: -Fi tzPatrick Scenario No.:

4 Event Description Trip of 'B' Service Water pump. Manual start of 'A' SW pump.

Op-Test No.:

4 5

6 7

8 Examiners: Fish Fuller (U/)

Operators:

Johnson Presby N (BOP)

Swap TBCLC pumps.

C(B0P)

C(ALL)

Both Feed pumps trip.

C (ALL)

Condensate pumps trip.

C (BOP)

Trigger 4 Trigger 5 Trigger 6 Trigger 7 Trigger 8 Loss of level control 6A FW heater.

RClC starts then trips after t w o minutes.

Initial Conditions: 92% (MOL Turnover: 92% CTP, pulling rods to 103% rod line and 96% CTP, then hold for one hour prior to proceeding to 100% CTP using recirc flow.

Rod pull sheet step 65. Next rod will be rod 18-15. rod pull from 12 - 16 I R(RO)

I Raise power by withdrawing control rods I Trigger2 2

I I

I HPCI fail to auto start.

M (ALL)

I Trigger8 I I required.

Recirc loop break in drywell. Containment sprays 9

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

Appendix D Scenario Outline Form ES-D-1 Event No.

1 Facility: -FitzPatrick Scenario No.:

5 Op-Test No.:

Malf.

Event Event No.

Type*

Description Trigger 1 C (BOP)

'A' Standby Gas Treatment fan trip.

Examiners: Fish Fuller (U/)

Operators:

Johnson Presby C (ALL)

C(ALL)

Trigger 4 Trigger 5 Initial Conditions: 50% (MOL)

Fuel clad failure ATWS due to RPS 'A' failure.

Turnover: Drywell is being purged IAW OP-37 D.6.11. SBGT 'A' is running to vent torus. 'C' Condensate pump is running in preparation for securing 'B' condensate pump for maintenance Control rods are at Step 47.

Trigger 6 Trigger 7 Trigger 8 Trigger 9 C (BOP)

C(B0P)

M(ALL)

M (ALL)

FW startup level control valve failure RWCU failure to isolate Steam leak in Reactor Building resulting in rad release.

Reactor depressurization.

2 3

4 5

6 Trigger2 1 N(B0P) I Normal shutdown of '6' condensate pump.

Trigger 3 I I(R0)

I Failure of 'B' NR level instrument upscale, level transient.

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

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