ML081400330
| ML081400330 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| 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) | |
Text
2008 FitzPatrick NRC Exam Rev1 and SRO-only outlines (i.e.., except for one category in Tier 3 the SRO-only outline, the Tier Totals in each KIA category shall not be less than two).
- 2. The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by 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.
- 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 WAS 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.
- 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
... . .a. . . . .. . ---
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.
Summary
Year Facility NRC Exam Rev 0 I ElAPE # I Name I Safety Function A
1 A
2 G
2 KIA Topic(s) Imp. #
Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR 295001 Partial or Complete Loss of Forced Core Flo COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :
Circulation I 1 & 4 (CFR: 41.8 to 41 .IO) AK1.02 Power/flow distribution 3.3 1 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following:
295003 Partial or Complete Loss of A.C. Power I 6 2.02 (CFR: 41.7 I45.8) AK2.02 Emergency generators 4.1 2 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C.
POWER :
295004 Partial or Complete Loss of D.C. Power I6 3( (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 :
295005 Main Turbine Generator Trip I 3 I ! ] 1 04 (CFR: 41.7 / 45.6) AA1.04 Main generator controls Ability to determine and/or interpret the following as they apply to SCRAM :
2.7 4 295006 SCRAM I 1 2 03 (CFR 41.10 I43.5 145.13) AA2.03 Reactor water level 4.0 5 2.4.1 2 Knowledge of general operating crew responsibilitiesduring emergency operations.
295016 Control Room Abandonment I 7 24.12 (CFR:41.10/45.12) 3.4 6 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR 295018 Partial or Complete Loss of Component COMPLETE LOSS OF COMPONENT COOLING WATER :
Cooling Water I8 1.01 (CFR: 41.8 to 41 .IO) AK1.01 Effects on componentkystern operations 3.5 7 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:
295019 Partial or Complete Loss of Instrument Air 18 2 09 (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 295021 Loss of Shutdown Cooling I21 3c alternate heat removal flow paths 3.6 9 Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS :
295023 Refueling Accidents I8 1 02 (CFR: 41.7 I45.6) AA1.02 Fuel pool cooling and cleanup system 2.9 10 Ability to determine andlor interpret the following as they apply to HIGH DRYWELL PRESSURE:
295024 High Drywell Pressure I5 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 295025 High Reactor Pressure I 3 2.1.7 interpretation. (CFR: 43.5 I45.12 I45.13) 3.7 EAPEsT1 G1
Year Facility NRC Exam Rev 0 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:
295030 Low Suppression Pool Water Level I 5 3 06 (CFR: 41.5 145.6) EK3.06 Reactor SCRAM 3.6 15 Abtldy to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL :
295031 Reactor Low Water Level I 2 113 (CFR: 41.7 145.6) EA1.I3 Reactor water level control 4.3 16 Ability to determine and/or interpret the following as 295037 SCRAM Condition Present and Reactor they apply to SCRAM CONDITION PRESENT AND REACTOR Power Above APRM Downscale POWER ABOVE APRM DOWNSCALE OR UNKNOWN :
or Unknown I 1 2 07 (CFR: 41 . I O 143.5 145.13) EA2.07 Containment conditionsfisolations 4.0 17 295038 High Off-Site Release Rate I 9 2 4 18 2.4.18 Knowledqe of the specific bases for EOPs. (CFR: 41.10 I45.13) 27 ia 600000 Plant Fire On Site I 8 1 02 II I l Il I Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site:
lAKl.02 Fire Fighting IKnowledqe of the interrelations between PARTIAL OR
-r I I 19 COMPLETE LOSS OF COMPONENT COOLING WATER and the 295018 Partial or Complete Loss of Component following:
(CFR: 41.7 I45.8) AK2.01 System loads 3.3 20 KIACategoryTotals: 4 1 EAPEs T1 G1
Year Facility NRC Exam 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 :
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 Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS:
295022 Loss of CRD Pumps I 1 1 02 (CFR: 41.8 to 41 .lo) AK1.02 Reactivity control 3.6 24 Knowledge of the reasons for the following responses 295032 High Secondary Containment Area as they apply to HIGH SECONDARY CONTAINMENT AREA Temperature I5 3 01 TEMPERATURE : EK3.01 Emergencyhormal depressurization 3.5 25 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH 295034Secondary Containment Ventilation High RADIATION :
Radiation I9 2 01 (CFR: 41.10 143.5 I 45.13) EA2.01 Ventilation radiation levels 3.8 26 Knowledge of the operational implications of the following concepts as they 295036 Secondary Containment High SumpIArea apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL 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 nditions or operations:
Shutdown Cooling Mode) 206000 High Pressure Coolant injection ge tank level: BWR-System 207000 Isolation (Emergency) Condenser I I I I 209001 Low Pressure Core Spray System (CFR: 41.7145.5to 45.8)A4.02 Suction valves 209002 High Pressure Core Spray System (HPCS) N/A JAF 3
211000 Standby Liquid Control System Knowledge of the physical connections and/or causeeffect relationships between REACTOR PROTECTION SYSTEM and the following:
212000 Reactor Protection System 1.04 (CFR: 41.2to 41.9/ 45.7to 45.8)K1.04A.C. electrical distribution Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM) 215003 Intermediate Range Monitor (IRM) SYSTEM and the following:
(CFR 41.2to 41.9 I45.7to 45.8)K1.01 RPS 215004 Source Range Monitor (SRM)
System Knowledge of the effect that a loss or malfunction of the SOURCE RANGE MONITOR (SRM) SYSTEM will have on following:
(CFR: 41.7I45.4)K3.02Reactor manual control: Plant-SDecific Knowledge of AVERAGE POWER RANGE MONITOFULOCAL I I 3.4 35 215005 Average Power Range POWER RANGE MONITOR SYSTEM design feature(s) and/or MonitorlLocal Power Range Monitor interlocks which provide for the following:
Svstem (CFR: 41.7)K4.07Flow biased trip setpoints 3.7 36
~~~ ~~
SYSTEMST2 G1
Year Facility NRC Exam Rev 0 I I I I COOLING SYSTEM (RCIC) :
(CFR: 41.5 145.3) K5.06 Turbine operation 2.7 37 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 3.9 38 Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEMINUCLEAR STEAM SUPPLY SHUT-OFF controls including:
(CFR: 41.5 145.5) Al.02 Valve closures 3.7 39 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 3.1 40 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 3.2 41 Ability to manually operate andor monitor in the control room:
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 status and operation of system Iand understand how operator action s and directives affect plant and system conditions. (CFR:
2 4 48 43.5 145.12) 3.5 43 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:
Plant-Specific 2.8 44 Knowledge of electrical power supplies to the following:
(CFR: 41.7) K2.01 Major D.C. loads 3.1 45 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:
(CFR: 41.7) K4.03 Securing of IAS upon loss of cooling water 2.8 47 2.4.1 1 Knowledge of abnormal condition procedures. (CFR:
2 411 41.10143.5145.13) 2.8 48 SYSTEMS T2 G1
Rev 0 Year Facility NRC Exam SYSTEMST2 G1
Year Facility NRC Exam Rev 0 SYSTEMST2 G1
Year Facility NRC Exam Rev 0 I System # I Name A
A I G
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 201003 Control Rod and Drive conditions or operations:
Mechanism (CFR 41.5 145.6) A2.02 Uncoupled rod 3.7 54 P
202001 Recirculation System 202002 Recirculation Flow Control System 2 17 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristicsI 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 Knowledge of the effect that a loss or malfunction of 3.7 3.1 55 56 the following will have on the NUCLEAR BOILER INSTRUMENTATION :
216000 Nuclear Boiler Instrumentation (CFR: 41.7 145.7) K6.02 D.C.electrical distribution 2.8 57 Knowledge of electrical power supplies to the 219000 RHRILPCI: ToruslSuppression following:
Pool Cooling Mode (CFR: 41.7) K2.02 Pumps 3.1 58 Ability to manually operate and/or monitor in the control room:
230000 RHRILPCI: ToruslSuppression (CFR: 41.7 I45.5 to 45.8) A4.06 Valve logic reset following Pool Spray Mode automatic initiation of LPCIlRHR in iniection mode 4.0 59 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 1 04 system: Plant-Specific 3.3 60 Knowledge of the effect that a loss or malfunction of the REACTOF VESSEL INTERNALS will have on following: (CFR: 41.7 I45.6) 290002 Reactor Vessel lnternais 3 03 K3.03 Reactor power Knowledge of REACTORRURBINE PRESSURE REGULATING SYSTEM design feature(s) and/or interlocks which provide for the following:
241000 Reactorflurbine Pressure (CFR: 41.7) 4 01 K4.01 Reactor pressure control SYSTEMST2 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:
271000 Offgas System 3.05 (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 SYSTEMST2 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 SYSTEMST2 G2
Year Facility NRC Exam Rev 0 Category KIA #
2.1 .I ToDic 1 Knowledge of less than one hour technical specification action statements for Imp. #
systems.
(CFR: 43.2 / 45.13) 2.1 .I 1 IMPORTANCE RO 3.0 SRO 3.8 3.c 66 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)
I 2.1.33 IMPORTANCE RO 3.4 SRO 4.0 67 2.1 . I 8 Ability to make accurate / clear and concise logs / records / status boards / and reports.
Conduct of (CFR: 45.12 / 45.13)
Operations 2.1 . I 8 IMPORTANCE RO 2.9 SRO 3.0 68
-3 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) 2 2.2.30 IMPORTANCE RO 3.5 SRO 3.3 3.5 69 2.2.22 Knowledge of limiting conditions for operations and safety limits.
Equipment (CFR: 43.2 / 45.2)
Control 2.2.22 IMPORTANCE RO 3.4SRO 4.1 - 3.L 70 A 2 2.3.1 I2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.
(CFR: 41.12 / 43.4. 45.9 / 45.10)
IMPORTANCE RO 2.6 SRO 3.0 2.f 71 Generics
Year Facility NRC Exam Rev 0 2.3.1 1 Ability to control radiation releases.
(CFR: 45.9 / 45.1 0) 3 2.3.1 1 IMPORTANCE RO 2.7 SRO 3.2 2 72 2.3.1 0 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
Radiation (CFR: 43.4 / 45.10)
Protection 2.3.10 IMPORTANCE RO 2.9 SRO 3.3 2 73 2.4.29 Knowledge of the emergency plan.
(CFR: 43.5 / 45.1 1) 4 2.4.29 IMPORTANCE RO 2.6 SRO 4.0 2.6 74 Emergency 2.4.18 Knowledge of the specific bases for EOPs.
Procedures (CFR: 41.10 / 45.13) and Plan 2.4.1 8 IMPORTANCE RO 2.7 SRO 3.6 2.7 75 Subtotal Tier 3 Point Total 10 Generics
Year Facility NRC Exam Rev 0 Catenorv KIA # Topic Imp. #
2.1 . I 2 Ability to apply technical specifications for a system.
(CFR: 43.2 / 43.5 / 45.3) 2.1 . I 2 IMPORTANCE RO 2.9 SRO 4.0 1
Conduct of 2.1.10 Knowledge of conditions and limitations in the facility license. (CFR: 43.1 / 45.13)
Operations 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 2 experiment involves an unreviewed safety question.
Equipment (CFR: 43.3 / 45.1 3)
Control 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) 2.3.3 IMPORTANCE RO 1.8SRO 2.9 2.9 98 3 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements.
Radiation (CFR: 41 .I 2 / 43.4. 45.9 / 45.1 0)
Protection 2.3.1 IMPORTANCE RO 2.6 SRO 3.0 3 99 I Subtotall 21 2 Generics
Year Facility NRC Exam Rev 0 4
Emergency I I 1 Procedures L
and Plan I I I I I I Tier 3 Point Total1 7 I 7 Generics
Year Facility NRC Exam Rev 0 Tierand I I 1 I I Group T I GI I
Randomly Selected KIA 295038 G2.4.10 Reason for Rejection Generic dealt with use of ARPs for high off-site release rate, randomly selected another generic i
T2 G2 272000 A4.03 Unable to write discriminating question for this WA, randomly selected another Tier 2 system T2 G I 400000 G2.4.43 Unable to write discriminating question for this WA, randomly selected another generic 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 Type Describe activity to be performed (see Note) Codeff Conduct of Operations N Manually Compute Average Drywell Air Temperature Conduct of Operations N Work Hour Restrictions Equipment Control M Complete a Tagout
~
Radiation Control N/A Emergency Plan 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)odifiedfrom bank (21)
(P)revious 2 exams (I1; randomly selected)
(S)imulator
ES-301, Rev. 9 Administrative Topics Outline Form ES-301-1 Facility: FitzPatrick Date of Examination: 3/31/08 to 4/10/08 Examination Level (circle one): RO / SRO Operating Test Number: 1 Administrative Topic Type Describe activity to be performed (see Note) Code*
Conduct of Operations N Manually Compute Average Drywell Air Temperature Conduct of Operations N Work Hour Restrictions Equipment Control M Complete a Tagout I I l
Radiation Control Canal Discharge Approval
~
I Emergency Plan 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.
(N)ew or (M)odified from bank (2 1)
(P)revious 2 exams (I1; randomly selected)
(S)imulator
ES-301, Rev. 9 Control Room/ln-Plant Systems Outline Form ES-301-2 1
1 Facility: FitzPatrick Exam Level (circle one): RO / SRO-I / SRO-U Date of Examination: 3/31/08 to 4/10/08 Operating Test No.: 1 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
- a. Component Cooling Water / Change In-Service RBCLC Pumps A, N, S 8
- b. Recirculation / Start Reactor Recirc Pump A, N, S 1 1 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
- e. Primary Containment System / Respond to High Drywell Temperature A, N, s 5
- f. Low Pressure Core Spray / Core Spray Surveillance Test N, L, s 4
- g. Emergency Generators / Load Test of B Emergency Diesel D, L, s 6
- h. APRM/LPRM / Restore Inoperable LPRM N, s 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 N, E 6 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 I SRO-U (A)lternate path 4-6 14-6/ 2-3 (C)ontrol room (D)irect from bank I ~ I I ~ I G (E)mergency or abnormal in-plant 21I 2 1I 2 1 (L)ow-Power / Shutdown 21/21/21 (N)ew or (M)odified from bank including 1(A) 22/22/21 (P)revious 2 exams I3 I I 3 I I 2 (randomly selected)
(R)CA 2 1 I 2 1I 2 1 (S)imulator
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
- b. Recirculation/ Start Reactor Recirc Pump I A,N,S 1 1
- c. High Pressure Coolant Injection / Shutdown HPCl A, D, S 2
- d. Main and Reheat Steam / MSlV SurveillanceTest N, s 3
- e. Primary Containment System / Respondto High Drywell Temperature A, N, S 5
- f. Low Pressure Core Spray / Core Spray SurveillanceTest
- g. Emergency Generators/ Load Test of B Emergency Diesel
- h. N/A
- i. Safety Relief Valves / Pull Fuses for Stuck Open SRV D, R 3
- j. CRD Hydraulics / Change In-Service Flow Control Valves D, E 1
- 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/ 4-612-3 (C)ontrol room (D)irect from bank 19/58/14 (E)mergency or abnormal in-plant 2 1 / 2 1I 2 1 (L)ow-PowerJ Shutdown 21 I 2 1/21 (N)ew or (M)odified from bank including 1(A) 22/12/21 (P)revious 2 exams I 3 / I 3 / I 2 (randomly selected)
(WA 1 1 / > I 111 (S)imulator
Appendix D Scenario Outline Form ES-D-I Facility: -FitzPatrick Scenario No.: 1 Op-Test No.:
Examiners: Fish Fuller (UI) Operators:
Johnson Presby Initial Conditions: -87% power -(MOL)
Turnover: -Operating normally at 87%, after assuming shift will continue t o 65% in preparation for cleaning condenser water boxes; Perform a swap of CRD pumps t o allow maintenance t o record vibration data on the 'B' CRD pump Event Malf. Event Event No. No. Type* Description 1 Trigger 1 N(R0) Swap in service CRD pumps.
Trigger 2 2 C (BOP) HPCl Aux Oil Pump breaker failure.
Trigger 3 3 R (.R 0 ). 5% power reduction.
Trigger 4 4 c (BOP) 'A' RFP trip and failure of Recirc runback.
Trigger 5 5 ALL) Loss of all CRD pumps 7
1 Trigger 1 C (BOP) I Main turbine trip, loss of Auxiliary busses 8
I Trigger 8 I c (BOP) I RCIC controller failure.
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 Facility: -FitzPatrick Scenario No.: 3 Op-Test No.:
Examiners: Fish Fuller (UI) Operators:
Johnson Presby Initial Conditions: 50% (MOL Turnover: -The plant at 50% power. Power ascension was suspended to allow Electrical Maintenanceto 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 Event No. -1 Malf.
No.
Trigger 1
~
Event Type*
I (RO)
Description APRM instrument failure.
Event 2
Trigger 2 N (ALL) 1 Restoration of busses 10200 and 10400 t o T4.
3 Trigger 3 RGO) I Raises power with control rods Trigger 4 4
~ ~~
Trigger 5 5 C(BOP)pn a d ve rte nt HPCI initiat io n (SRO)
Trigger 6 6 c (ALL)
~ ~~
Unisolable leak on Instrument Air header M (ALL) SBLOCA on recirc line, containment spray required ALL) Spray torus, both loop of RHR fail M (ALL) Spray Torus/Drywell with RHR SW
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-I Facility: -Fi tzPatrick Scenario No.: 4 Op-Test No.:
Examiners: Fish Fuller (U/) Operators:
Johnson Presby 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 Event Malf. Event Event No. No. Type* Description 1 Trigger 1 C (BOP) Trip of 'B' Service Water pump. Manual start of 'A' SW pump.
2 I Trigger2 I R(RO) I Raise power by withdrawing control rods Trigger 4 4 N (BOP) Swap TBCLC pumps.
Trigger 5 5 C(B0P) Loss of level control 6A FW heater.
Trigger 6 6 C(ALL) Both Feed pumps trip.
Trigger 7 7 C (ALL) Condensate pumps trip.
Trigger 8 8 C (BOP) RClC starts then trips after t w o minutes.
I I I HPCI fail to auto start.
9 I Trigger8 I M (ALL)
I Recirc loop break in drywell. Containment sprays required.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
Appendix D Scenario Outline Form ES-D-1 Facility: -FitzPatrick Scenario No.: 5 Op-Test No.:
Examiners: Fish Fuller (U/) Operators:
Johnson Presby Initial Conditions: 50% (MOL)
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.
Event Malf. Event Event No. No. Type* Description 1 Trigger 1 C (BOP) 'A' Standby Gas Treatment fan trip.
2 Trigger2 1 N(B0P) I Normal shutdown of '6' condensate pump.
3 Trigger 3 Trigger 4 I I(R0) I Failure of 'B' NR level instrument upscale, level transient.
4 C (ALL) Fuel clad failure Trigger 5 5 C(ALL) ATWS due t o RPS 'A' failure.
Trigger 6 6 C (BOP) FW startup level control valve failure Trigger 7 C(B0P) RWCU failure t o isolate Trigger 8 M(ALL) Steam leak in Reactor Building resulting in rad release.
Trigger 9 M (ALL) Reactor depressurization.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor