ML101720553

From kanterella
Revision as of 11:18, 22 August 2018 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
R.E. Ginna - Draft - Outlines (Folder 2)
ML101720553
Person / Time
Site: Ginna Constellation icon.png
Issue date: 04/06/2010
From:
Constellation Energy Nuclear Group
To: Fish T H
Operations Branch I
HANSELL S L
Shared Package
ML092470059 List:
References
TAC U01766
Download: ML101720553 (31)
v • d • e


Text

3 Written Examination Outline Form ES-401-2 Facility:

R.E. Date of Exam: 06/21/10 SRO-Only Points Tier Group K G Total A2 G* 2 1. 3 3 3 18 Emergency

& 2 2 9 2 Plant Evaluations Tier 4 4 5 27 5 2 2 3 2 3 2 2 :3 3 3 3 28 3 2. 0 10 0 Systems 3 3 4 3 4 3 3 4 4 4 3 38 5 Totals 2 3 2 3 3. Generic Knowledge

&3 3 2 2 2 2 Ensure that at least two topics from every applicabl,e KIA category are sampled within each tier of the RO and SRO-only outlines (Le., except for one cate1gory in Tier 3 of the SRO-only outline, the Tier Totals in each KIA category shall not be less than two). 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. 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. 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. 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 FlO and SRO-only portions, respectively. Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories. The generic (G) KlAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable K/A's 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. For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3.

Limit SRO selections to KlAs that are linked to Total ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group AA2.23 -Ability to determine 056/ Loss of Off-site Power 16 and interpret the following as they apply to the Loss of Offsite Power: Turbine 3.9 76 reactor button and indicator EA2.04 -Ability to determine or interpret the following as 055 I Station Blackout / 6 they apply to a Station Blackout:

Instruments and 4.1 77 controls operable with only de available AA2.06 -Ability to determine and interpret the following as they apply to the Loss of 026 / Loss of Component Cooling Water 18 Component Cooling Water: The length of time after the loss of CCW flow to a 3.1 78 component before that component may be dam 2.4.34 -Emergency Procedures I Plan: 062/ Loss of Nuclear Service Water / 4 Knowledge of RO tasks performed outside the main control room during an 4.1

  • 79 i emergency and the resultant effects. 2.2.22 -Equipment Control: E12/ Steam Line Rupture Excessive Heat Transfer /4 Knowledge of limiting conditions for operations and 4.7 80. limits. 2.4.3 -Emergency 025 / Loss of Residual Heat Removal System / 4 Procedures

/ Plan: Ability to identify post-accident 3.9 81 instrumentation.

EK1.02 -Knowledge of the operational implications of 009 / Small Break LOCA / 3 X the following concepts as 3.5 39 they apply to the small break LOCA: Use of steam tables ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function 007 / Reactor Trip Stabilization

-Recovery /1 x E05/lnadequate Heat Transfer -Loss of x Secondary Heat Sink / 4 E11 / Loss of Emergency x Coolant Recirculation / 4 E04 / LOCA Outside x Containment

/3 077 / Generator Voltage and Electric Grid x Disturbances KiA Topic(s) EK1.04 -Knowledge of the operational implications of the following concepts as they apply to the reactor trip: Decrease in reactor power following reactor trip (prompt and su uent EK1.1 -Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) Components, capacity, and function of emergency

s. EK2.2 -Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following:

Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facil . EK2.1 -Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and features.

AK2.07 -Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following:

Turbine / erator control 3.6 40 3.8 41 3.9 42 3.5 43 3.6 44 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal PlanI Evolutions

-Tier 1 Group EAPE#/Name Safety A2 G KIA Topic(s) limp. I Q# I I I AK3.03 -Knowledge of the reasons for the following responses as they apply to 022 / Loss of the Loss of Reactor Coolant 3.1 45 Coolant Makeup / Makeup: Performance of lineup to establish excess letdown after determining need AK3.03 -Knowledge of the reasons for the following responses as they apply to 026 / Loss of the Loss of Component 4.0 46 Cooling Water / Cooling Water: Guidance actions contained in EOP for Loss of CCW/nuclear service water AK3.02 -Knowledge of the reasons for the following 040/ Steam Line Rupture / responses as they apply to 4.4 47 4 the Steam Line Rupture: ESFAS initiation AA 1.24 -Ability to operate and / or monitor the following as they apply to the Loss of 056/ Loss of Off-site Power X Offsite Power: Plant 2.9 48 6 computer, to call up in-core temperature monitoring AA 1.21 -Ability to operate and / or monitor the following as they apply to the Reactor 015 / 17 / Reactor Coolant X Coolant Pump Malfunctions 4.4 49 Pump Malfunctions / 4 (Loss of RC Flow): Development of natural circulation How EA1.01 -Ability to operate and monitor the following as 055 / Station Blackout / 6 they apply to a Station 3.7 50 Blackout:

In-core thermocou eratures Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety KIA Topic(s) AA2.17 -Ability to determine and interpret the following as 1057 /Loss of Vilal AC they apply to the Loss of *Electrical Instrument Bus I Vital AC I nstrument Bus: 3.1 51 System and component status, using local or remote controls AA2.12 -Ability to determine 027 I Pressurizer Pressure and interpret the following as Control System Malfunction they apply to the 3.7 52 /3 Pressurizer Pressure Control Malfunctions:

PZR level EA2.01 -Ability to determine

  • 038/ Steam Generator or interpret the following as 4.1 53 Tube Rupture / they apply to a SGTR: to isolate one or 2.1.23 -Conduct of Operations:

Ability to perform specific system and 011 / Large Break LOCA / 4.3 54 integrated plant procedures during all modes of plant 054/ Loss of Main 4.2 55 Feedwater / 4 008/ Pressurizer Vapor 3.8 56 Space Accident / 3 KIA Category Totals 3 3 18/6 ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#fName Safety Fun KIA Topic(s) AA2.03 -Ability to determine and interpret the following as 036 / Fuel Handling they apply to the Fuel Incidents

/8 Handling Incidents:

Magnitude of potential radioactive release AA2.12 -Ability to determine and interpret the following as 033/ Loss of Intermediate they apply to the Loss of Range Nuclear Intermediate Range Nuclear Instrumentation

/7 Instrumentation:

Maximum allowable channel reement 2.4.30 -Emergency Procedures I Plan; Knowledge of events related to system operation I status E03 / LOCA Cooldown and that must be reported to Depressurization

/4 internal organizations or external agencies, such as the state, the NRC, or the transmission system rator. 2.4.6 -Emergency E14/ High Containment Procedures

/ Plan: Pressure /5 Knowledge of EOP m on ies. AK1.01 -Knowledge of the operational implications of 024 / Emergency Boration / 1 X the following concepts as they apply to Emergency Boration:

Relationship between boron addition and in T-ave EK2.08 -Knowledge of the 074/lnadequate Core Cooling /4 x interrelations between the following Inadequate Core Cooling: Sensors and detectors 4.2 82 3.1 83 4.1 84 4.7 85 3.4 57 2.5 58 ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group 061 / Area Radiation Monitoring (ARM) System Alarms / 7 X E02 / SI Termination / 3 E13 / Steam Generator Overpressure / 4 069 / Loss of Containment Integrity / 5 076 I High Reactor Coolant Activity /9 X E10/ Natural Circulation Steam Void in Vessel RVLlS/4 AK3.02 -Knowledge of the reasons for the following responses as they apply to the Area Radiation 3.4 59 Monitoring (ARM) System Alarms: Guidance contained in alarm response for ARM EA 1.3 -Ability to operate and / or monitor the following as they apply to the (SI Termination)

Desired 3.8 60 operating results during abnormal and emergency situations.

EA2.2 -Ability to determine and interpret the following as they apply to the (Steam Generator Overpressure)

Adherence to appropriate 3.0 61 procedures and operation within the limitations in the facility's license and amendments.

2.2.39 -Equipment Control: Knowledge of less than or equal to one hour technical 3.9 62 specification action statements for AK3.05 -Knowledge of the reasons for the following responses as they apply to the High Reactor Coolant 2.9 63 Activity : Corrective actions as a result of high fission-product radioactivity level in the RCS 2.2.37 -Equipment Control: Ability to determine operability and I or 3.6 64 availability of safety related ment.

Form ES-401-2 R.E. 2010 NRC Written Examination Emergency and Abnormal Plant Evolutions

-Tier 1 Group EAPE#/Name Safety Function I K1 I K2 I K3 I A1 I A2 I G I KIA Topic(s) AA 1.01 -Ability to operate and I or monitor the following 003 I Dropped Control as they apply to the2.9 65Dropped Control Demand position and KIA Category 2 2 ES-401 Form ES-401-2 System #/Name 026 Containment Spray 059 Main Feedwater 010 Pressurizer Pressure Control 004 Chemical and Volume Control R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) A2.04 -Ability to (a) predict the impacts of the following malfunctions or operations on the CSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Failure of A2.04 -Ability to (a) predict the impacts of the following malfunctions or operations on the Main Feedwater System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Feeding a SG 2.2.38 -Equipment Control: Knowledge of conditions and limitations in the facil license. 2.4.9 -Emergency Procedures

/ Plan: Knowledge of low power / shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) ation ies. 4.2 86 3.4 87 4.5 88 4.2 89 Form ES-401-2 R.E. 2010 NRC Written Examination Plant Systems -Tier 2 Group I System #/Name I K1 I K21 K31 K41 K51 K61 A 1 I A2 I A31 A41 G KIA Topic(s) I Imp. I Q# I A2.11 -Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those 006 predictions, use 4.4 90 Core procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Rupture of ECCS header K1.01 -Knowledge of the physical connections and/or cause-effect 022 Containment relationships between 3.5 1 Cooling the CCS and the following systems: K1.02 -Knowledge of the physical connections and/or cause-effect 008 Component relationships between 3.3 2 Cooling Water the CCWS and the 064 Emergency 2.8 3 Diesel Generator K2.02 -Knowledge of bus power supplies to 078 Instrument Air 3.3 4 the following Emergency 026 Containment 3.9 5 Spray 076 Service Water 3.7 6 Form ES-401-2 R.E. 2010 NRC Written Examination Plant Systems -Tier 2 Group System #/Name IK1 IK21K31K41K51K61A1 I A2 IA3 A4 G KIA Topic(s) K4.06 -Knowledge of MRSS design feature(s) 039 Main and and/or interlock(s) which X 7 Reheat provide for the following:

Prevent reverse steam flow on steam line break K4.08 -Knowledge of 013 Engineered ESFAS design feature(s)

Safety Features X and/or interlock(s) which 3.1 8 Actuation provide for the following Redu KS.44 -Knowledge of the operational implications of the 004 Chemical following concepts as 3.2 9 Volume they apply to the CVCS: Pressure response in PZR during in-and-out su KS.01 -Knowledge of the operational implications of the OOG Emergency following concepts as 2.8 10 Core Cooling they apply to ECCS: Effects of temperatures on water level indications K6.03 -Knowledge of the effect of a loss or 012 Reactor malfunction of the 3.1 11 Protection following will have on the RPS: Tri ic circuits K6.14 -Knowledge of the effect of a loss or 003 Reactor malfunction on the 2.6 12following will have on the RCPS: Starting uirements A 1.02 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) 007 Pressurizer associated with 2.7 13 Relief/Quench Tank operating the PRTS controls including:

Maintaining quench tank 11 ES-401 Form ES-401-2 C System #/Name 063 DC 005 Residual 010 Pressurizer Pressure Control 059 Main Feedwater 103 Containment R.E. Ginna 2010 NRC Written Examination Outline Plant Systems -Tier 2 Group 1 KIA Topic(s) A 1.01 -Ability to predict and/or monitor changes in parameters associated with operating the dc x 2.5 14 electrical system controls including:

Battery capacity as it is affected rate A2.01 -Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the 2.7 15 consequences of those malfunctions or operations:

Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation A2.01 -Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS; and (b) based on those predictions, use 3.3 16 procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Heater failures A3.06 -Ability to monitor automatic operation of 3.2 17 the MFW, including:

Feedwater isolation A3.01 -Ability to monitor automatic operation of the containment system, 3.9 18 including:

Containment isolation Form ES-401-2 R.E. 2010 !\IRC Written Examination Plant Systems -Tier 2 Group System KIA Topic(s) A4.04 -Ability to manually operate and/or 062 AC Electrical monitor in the control 2.6 19 Distribution room: Local operation of breakers A4.02 -Ability to manually operate and/or 073 monitor in the control 3.7 20 Radiation room Radiation monitoring system control I 2.4.34 -Emergency Procedures

/ Plan: Knowledge of RO tasks 061 performed outside the 4.2 21 main control room during Feedwater an emergency and the resultant operational effects. 2.1.27 -Conduct of Operations:

Knowledge 076 Service 3.9 22 of system purpose and / or function.

K3.01 -Knowledge of 013 Engineered the effect that a loss or Safety Features x malfunction of the 4.4 23 Actuation ESFAS will have on the followi : Fuel 2.4.20 -Emergency Procedures

/ Plan; Knowledge of 010 Pressurizer operational 3.8 24 Pressure Control implications of EOP warnings, cautions, and notes. A3.01 -Ability to monitor automatic operation of 007 Pressurizer the PRTS, including:

2.7 25 Relief/Quench Tank Components which disch to the PRT ES-401 Form ES-401-2 R.E. 2010 NRC Written Examination Plant Systems -Tier 2 Group 039 Main and Reheat Steam 063 DC Electrical Distribution 003 Reactor Coolant x Pump KiA Category Totals A2.05 -Ability to (a) predict the impacts of the following mal-functions or operations on the MRSS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Increasing steam demand, its relationship to increases in reactor r A4.01 -Ability to manually operate and/or monitor in the control room: Major breakers and control 'fuses K5.02 -Knowledge of the operational implications of the following concepts as they apply to the RCPS: Effects of RCP coastdown on RCS 3.3 26 2.8 27 2.8 28 28/5 14 ES-401 Form ES-401-2 068 Liquid Radwaste 011 Pressurizer Level Control 001 Control Rod Drive R.E. Ginna 2010 Written Examination Outline Plant Systems -Tier 2 Group 2 KJA Topic(s) A2.02 -Ability to (a) predict the impacts of the following malfunctions or operations on the Liquid Radwaste System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Lack of tank recirculation prior to release 2.4.21 -Emergency Procedures I Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control etc. A2.19 -Ability to (a) predict the impacts of the following malfunction or operations on the and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Axial flux distribution 2.8 91 4.6 92 4.0 93 15 ES-401 Form ES-401-2 R.E. Ginna 2010 Written Examination Outline Plant Systems -Tier 2 Group 2 System #!Name KIA Topic(s) K5.12 -Knowledge of the operational implications of the 011 Pressurizer Level Control x following concepts as they apply to the PZR 2.7 29 LCS Criteria and purpose of PZR level ram A4.02 -Ability to 015 Nuclear Instrumentation manually operate and/or monitor in the control 3.9 30 I room: NIS indicators A3.01 -Ability to monitor automatic operation of 016 Non-nuclear I nstru mentation the NNIS, including:

Automatic selection of 2.9 31 NNIS inputs to control K2.05 -Knowledge of 001 Control Rod Drive x bus power supplies to the following:

M/G 3.1 32 sets ..... K6.04 -Knowledge of the effect or a loss or 002 Reactor Coolant x malfunction on the following RCS 2.5 33 components:

RCS vent valves K4.01 -Knowledge of design feature(s) and/or 071 Waste Gas Disposal x interlock(s) which provide for the following:

Pressure capability of 2.6 34 the waste gas decay tank K3.02 -Knowledge of the effect that a loss or 072 Area Radiation Monitoring malfunction of the ARM system will have on the 3.1 35 following:

Fuel handling n<u-'"ns ES-401 Form ES-401-2 041 Steam Dump!Turbine Bypass Control 055 Condenser Air x Removal 086 Fire Protection R.E. Ginna 2010 Written Examination Outline Plant Systems -Tier 2 Group 2 KIA Topic(s) A2.02 -Ability to (a) predict the impacts of the following malfunctions or operations on the SDS; and (b) based on those 3.6 36 predictions or mitigate the consequences of those malfunctions or operations:

Steam valve stuck 0 K1.06 -Knowledge of the physical connections and/or cause effect relationships between 2.6 37 the CARS and the following systems: PRM A 1.04 -Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with 2.7 i 38 operating the Fire Protection System controls including:

Fire dam Group Point Total: 17 I ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 1::'. :...o.l"flllY R.E. Ginna r. y KA# 2.1.5 2.1.8 2.1.3 1. Conduct of Operations 2.1.25 2.1.42 Subtotal 2.2.37 2.2.35 2.2.7 2. EqUipment Control 2.2.44 Subtotal Date: 6/21/2010 Topic Ability to use procedures related to shift staffing, such as minimum crew COmfJfClllclit, overtime limitations, etc. Ability to coordinate personnel activities outside the control room. Knowledge of shift or short-term relief turnover Ability to interpret referemce materials, such as graphs, curves tables etc. Knowledge of new and spent fuel movement procedures.

Ability to determine opeirability and / or availability of safety related equipment.

Ability to determine Technical Mode of OperdLlUII.

Knowledge of the process for conduction or infrequent tests .1 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and II conditions.

IR 2.9 3.4 3.7 .. 3.6 3.6 2.9 .;.:.;,';

';. RO SRO-Only Q# IR Q# 66 67 75 4.2 94 3.4 100 3

... 2 68 69 4.4 95 3 r. 1 18 ES-401 Generic Knowledge and Abilities Outline (Tier 3) Form ES-401-3 Knowledge of radiation or I contamination hazards that may arise 3.4 70 during normal, abnormal, or emergency conditions or activities.

Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment 3.4 71 entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc. r 3. Radiation Control Ability to use radiation monitoring systems, such as 'fixed radiation monitors and alarms, portable survey 2.9 96 instruments, personnel monitoring equipment, etc. Knowledge of radiation exposure limits 3.7 98 under normal or emergency conditions.

Subtotal f .. " " 2 '.;::;,t.:1'1:

2 Knowledge of events rellated to system operation

/ status that must be reported to internal organizations or external 2.7 72 agencies, such as the state, the NRC, or the transmission system operator.

Knowledge of the RO's responsibilities 3.9 73 in emergency plan implementation.

4. Emergency I Procedures

/ Knowledge of the emergency action Plan 4.6 97 level thresholds and classifications.

Ability to take actions called for in the facility emergency plan, including 4.4 99 supporting or acting as emergency coordinator if required.

.. 2 2 Tier 3 Point 10 7 i""""" ..,..... "",* 19 I ES-401 Tier / 2/ 1 / 2/ Randomly Selected KA G3 / 2.3.4 replaced by G3 / 2.3.14 G2 / 2.2.35 replaced by G2 / 2.2.44 028/ K6.01 replaced by 002 / K6.04 007/ A2.01 replaced by 059 / A2.04 012 /G2.4.9 replaced by 004 / G2.4.9 040/ AK3.03 replaced by 040 / AK3.02 010/ G2.4.30 replaced by 010/ G2.4.20 G2 / 2.2.15 replaced by G2 / 2.2.7 028/ A2.01 replaced by 001 / A2.19 Record of RejBcted KIA's Form ES-401-4 Reason for Rejection This topic is duplicated on the SRO Examination Duplicated on 1=10 examination and is more suitable to RO examination System is no longer used or referenced by any procedures at the facility Excessive overlap for system. Questions 7 and 47 also test same topic. Reactor Protection has already performed its function in lower mode, making it impossible to develop a test item to evaluate knowledge of this KA topic for this system Direct overlap with question 7. Selected this item for removal because it has the lower KA importance value of the 2 items For this system, this generic KA will not yield an RO test item. This information is confined to responsibility of SRO. Direct overlap with topic on Operating Examination.

System is no longer used or referenced by any procedures at the facility 20 II Administrative Form

-1 Facility:

Ginna Date of Examination:

6/21/10 Examination Level: RO Operating Test Number: NRC ! Administrative Topic Type Code* Describe activity to be performed (see Note) 2.1.25 (3.9) Ability to interpret reference material, such as Conduct of Operations graphs, curves, tables, etc. M,R JPM: Given a Set of Conditions, Perform a Critical Rod Position Calculation In Accordance With 1.2.2, Critical Rod Position Calculation.

2.1.29(4.1)

Knowledge of how to conduct system lineups, I Conduct of Operations such as valves, breakers, switches, etc. N,R JPM: Perform an RHR System Valve and Breaker Position Verification on the Control Board Prior to Mode 3 Entry i 2.2.13(4.1)

Knowledge of tagging and clearance Equipment Control procedures.

M,R JPM: Perform the RO Review of a Tagout For V-3968, 4B Condensate Heater Discharge Check Valve. 2.3.15 (2.9) Knowledge of radiation monitoring systems, Radiation Control such as fixed radiation monitors and alarms, portable survey instruments, personnel N,R monitoring equipment, etc. JPM: Determine Steam Generator Tube Leakage Using Radiation Monitor R-47, Air Ejector Radiation Monitor Low Range Ii Emergency Procedure/Plan N/A CATEGORY NOT SELECTED All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & (C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::; 3 for ROs; ::; 4 for SROs & RO retakes) (0) (N)ew or (M)odified from bank (> 1) (4) (P}revious 2 exams 1; randomly selected)

(0)

Facility:

Ginna Examination Level: SRO Administrative Topic (see Note) Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Plan Type Code* M,R 2.1.25 (4.2) JPM: 2.1.18 (3.8) M,R JPM: I I D,R 2.2.37 (4.6) JPM: 2.3.14 (3.8) P,S M,R JPM: , 2.4.41 (4.6) JPM: Administrative Topics Form ES-301-1 Date of Examination:

6/21/10 Operating Test Number: NRC Describe activity to be performed Ability to interpret reference material, such as graphs, curves, tables, etc. Review a Critical Rod Position Calculation Ability to make accurate, clear, and concise logs, records, status board, and reports. Perform the Shift Manager review of the 0-6.13, DAILY SURVEILLANCE LOG Ability to determine operability and/or availability of safety related equipment.

Perform a Safety Function Determination Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Respond to a Contaminated Injured Person Knowledge of the emergency action level thresholds and classifications.

Determine the EAL For an Event All items (5 total are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when 5 are required.

  • Type Codes & (C)ontrol room, (S)imulator, or Class(R}oom (D)irect from bank (::; 3 for ROs; ::; 4 for SROs & RO retakes) (1) (N)ew or (M)odified from bank (> 1) (3) (P)revious 2 exams (::; 1; randomly selected)

(1)

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

Ginna Date of Examination: Exam Level (circle one): RO Operating Test No.: ntrol Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 Type Code* System I JPM Title Function 064 Emergency Diesel Generator i S,A,D,E,EN 6 Start the Emergency Diesel Generator during a Station Blackout 006 Emergency Core Cooling System S,A,E.EN,N 2 Isolate the Safety Injection Accumulators during inadequate core cooling : c. 003 Reactor Coolant Pump s. L. M 4P Start a RCP during a plant startup Ii I I d. APE 068 Control Room Evacuation S,A,E,P 8 Perform the immediate actions for a Control Room Evacuation

e. 010 Pressurizer Pressure Control System I S.P,L 3 Place L TOP in service i 004 Chemical and Volume Control System S,L, N Perform a Rapid Boration during a plant shutdown I . g. 026 Containment Spray S,A,E,EN,N 5 Secure Containment Spray following an automatic actuation. 061 Auxiliary I Emergency Feedwater S.D,E Place the Standby Auxiliary Feedwater System in service during a loss heat I In-Plant Systems@ (3 for RO; 3 for SRO-I: 3 or 2 for SRO-U) , I I 071 Waste Gas Disposal System II i. D,R 9 Release the A Gas Decay Tank APE 067 Plant Fire On Site D,E,R 8 Ventilation System During a Fire I k. 039 Main and Reheat Steam System P,E 4S Locally Operate the ARVs Control Room/In-Plant Systems Outline Form ES-301-2 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
  • Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1 (A) (P)revious 2 exams (R)CA (S)imulator Criteria for RO 1 SRO-II SRO-U 4-6 (4) 14-6/2-3 S 9 (4) 1s 8 1 s 4 1 (7) 11 11 -I -I 1 (control room system) 1 (3) / 1 I 1 2 (4) I 2 I 1 s 3 (3) I s 3 I s 2 (randomly selected) 1 (2) I 1 I 1 6 Control Room/In-Plant Systems Outline Form ES-301-2 . Facility:

Ginna Date of Examination:

6/21/10 Exam Level: SRO(I) I SRO (U) Operating Test No.: NRC Control Room Systems@ (8 for RO; 7 for SRO-I; 2 or 3 for SRO-U, including 1 ESF) Type Code* Safety System 1 JPM Title Function a. 064 Emergency Diesel Generator S,A,D,E,EN Start the Emergency Diesel Generator during a Station Blackout 006 Emergency Core Cooling System S,A,E,EN,N 2 Isolate the Safety Injection Accumulators during inadequate core cooling I c. 003 Reactor Coolant Pump S,L,M 4P Start a RCP during a plant startup d. APE 068 Control Room Evacuation S,A,E,P 8 Perform the immediate actions for a Control Room Evacuation

e. 010 Pressurizer Pressure Control System S,P,L 3 Place LTOP in service f. 004 Chemical and Volume Control System S,L, N Perform a Rapid Boration during a plant shutdown I g. 026 Containment Spray S,A, E, EN, N 5 Secure Containment Spray following an automatic actuation.

I In-Plant Systems@ (3 for RO; 3 for SRO-I; 3 or 2 for SRO-U) I i. 071 Waste Gas Disposal System Release the A Gas Decay Tank J D,R 9

  • j. APE 067 Plant Fire On Site Secure Ventilation System During a Fire D,E,R 8 i k. 039 Main and Reheat Steam System P,E I 4S Locally Operate the ARVs All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room. 1 ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2
  • Type Codes Criteria for RO I SRO-II SRO-U (A)lternate path 4-6 / 4-6 (4)/2-3 (2) (C)ontrol room {D)irect from bank ::;; 9/ 8 (3)1 4 (3) (E)mergency or abnormal in-plant 1 / ';? 1 (6)/ 1 (3) (EN)gineered safety feature I -1 1 (2)(control room system) (L)ow-Power I Shutdown ';? 1 I 1 (3)/ 1 (1) (N)ew or (M)odified from bank including 1 (A) 21 ';? 2 (4)/ 1 (2) (P)revious 2 exams 3 I 3 (3)1 2 (O)(randomly selected) i (R)CA ';? 1 / ';? 1 (2)/ 1 (2) . {S)imulator Appendix Scenario Outline Form ES-D-1 Facility: Scenario No.: 1 Op Test No.: N10-1 Operators: (SRO) (RO) (BOP) Initial The plant is at 49% power (BOl). The plant power was reduced several days ago due to a malfunction on the A MFW Pump . Corrective Maintenance has been
  • completed, and the pump is to be restarted.

RG&E Energy Control Center has requested that the electriic plant be aligned to a 0/100 configuration on circuit 7T to allow the RG&E to perform an insulator inspection on the 767 Line. Per Chemistry direction Normal letdown is at 60 gpm.

The following eqUipment is Out-Of-Service:

The B AFW Pump is OOS for Bearing Replacement.

Event Malf. Event Type* No. No.

1 NA N-BOP Shift Electric loss of B Instrument 2 Pressurizer Pressure (PT -449) fails High 3 PZR02D Main Turbine High Vibration/EHC control 4 TUR05E TUR09D Feed Line Hupture Inside Containment 5 6 NA Main Turbine Failure to Auto Trip A AFW Pump Fails after start 7 C-BOP TDAFW Pump overspeed trip 8 REM-Standby AFW fails to 9 OVR-C-RO FDW42A FDW15B , * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix Scenario Outline Form ES-D-1 Facility: Scenario No.: 2 Op Test No.: N10-1 Examiners:

Operators: Initial The Plant is at 100% power (EOl). Per the daily work schedule, CROl-7, Swapping Service Water Pumps, is to be performed this shift, swapping to A and D Service Water pumps.

The following equipment is Out-Of-Service:

The B SI Pump is OOS for Bearing

  • Replacement.

Event Malf. No. Event Type* Event No. Description 1 ClG09C N-BOP Swap Service Water PumpsJD Service Water Pump Trip ClG01D C(TS)-SRO 2 CVC10A I-RO VCT level '112 Fails HIGH I-SRO 3 CVC12C C-RO C Charging Pump trips C-SRO 4 TUR16B I-BOP Turbine Impulse Pressure (PT -486) fails High I-SRO 5 A-FDW30 R-RO A MFW Pump Oil Sump HI-lO level/Rapid DownpowerlStuck Rod ROD03* C-BOP G11 C(TS)-SRO 6 SGN04B M-RO SGTR M-BOP M-SRO 7 RPS07A C-RO A SI Pump fails to start in AUTO RPS07C C SI Pump fails to start in AUTO RPS07D * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix Scenario Outline Form ES-D-1 Facility: Scenario No.: 3 Op Test No.: N10-1 Operators: (SRO) (BOP) Initial The plant is at 70% power (MOL). The plant was taken to 50% due to a failure of the B MFW Pump. Corrective Maintenance was performed and plant power raised to 70% four days ago. The plant has been holding at this power level to observe the B MFW Pump operating characteristics.

The plant power is now to be raised to 100% power using 0-5.2, Load Ascension, Step 6.2.33, Placing Condensate

  • Booster Pumps in service. Turnover:

I The following equipment is Out-Ot-Service:

The B SI Pump is OOS for Bearing

  • Replacement.

Event Malf. No. Event Type* No.

I I 1 R-RO Load Ascension/Start Condensate Booster 2 GEN02 Generator Hydrogen Temperature (Cold Gas) Instrument Fails High C-SRO 3 I-RO BLoop Thot RTD fails 4 C-BOP FRV B Controller Fails Low i C-SRO ROD02-F2 C-RO Dropped Rod/2 nd Dropped ROD02-B8 6 RCS03D M-RO Large Break LOCA on BLoop Tcold I7 C-RO Failure of A Train SIS in Auto 8 C-RO Failure of the B RHR Pump to start in AUTO * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Appendix Scenario Outline Form ES-D-1 Facility: Scenario No.: 4 Op Test No.: N10-1 Examiners:

Operators: Initial The Plant is at 100% power (MOL). It is expected that immediately after turnover the crew will swap Condensate Pumps per Maintenance Dept Work Order, and conduct routine Rod Control exercises on Control Bank D.

The following equipment is Out-Of-Service:

The B AFW Pump is OOS for Bearing Replacement.

Event Malf. No. Event Event No. Type* Description 1 CND04B Swap Condensate Pumps (Start B, Secure C)/B Condensate Pump Trip C-SRO 2 ROD13C-N-RO Rod Control Exercise/MRPI Failure K7 C(TS)-SRO Pressurizer Level (LT -428) Fails HIGH 3 I-RO I(TS)-SRO , 4 HTR02A R-RO C-BOP Heater Drain Pump A trips/Rapid Downpower C-SRO 5 STM04C C-BOP C-SRO B SG ARV Controller (AOV-3410) fails in AUTO 6 STM05A M-RO MSIVs Close STM05B M-BOP M-SRO 7 STM09A STM09B NA SG Safeties lift/fail OPEN (1 per SG) 8 RPS07M C-BOP TDAFW Pump Steam Supply Valves Fail to Open in AUTO RPS07N 9 FDW11A NA A AFW Pump trips on Start * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1 Scenario Outline Form ES-D-1 Facility:

Ginna Scenario No.: 5 Op Test No.: N10-1 Examiners:

Operators: (SRO) I (RO) (BOP) Initial The plant is at 1 x 10-8 amps (BOL). The plant ran at 100% power for 12 days, and*I , then tripped four days ago due to a MFW Pump failure. The repairs have been made and the plant is ready to be started back up. The crew will be directed to pull rods to the point of adding (POAH), and start the B MFW Pump, in accordance with 0-1.2, Plant Startup From Hot Shutdown to Full Power Load, Step 6.3.4 and beyond.

The following equipment is Out-Of-Service:

The B Condensate Pump is OOS for Bearing Replacement.

Event Malt. Event Type* Event No. No. Description 1 R-RO Raise power to POAH N-SRO Testing of the MFW Oil Pumps 2 N-BOP N-SRO 3 CLG10 480VAC Ground/A CCW Pump trips w/B CCW Pump fails to start in AUTO CLG02A C(TS)-SRO A-EDS33 Loss of Compensating VoltClge to Intermediate Range N35. 4 C-BOP C(TS)-SRO Fault on 480V Bus 18/SW Pump D fails to start 5 C-BOP CLG01D C(TS)-SRO 6 C-RO PCV-135 fails Closed C-SRO ! 7 RCS02D M-RO Loop B Cold Leg Break (1000 gpm) M-BOP M-SRO RPS05A C-RO Failure of R'eactor to Trip in AUTO and MANUAL RPS05B * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor -1

Retrieved from "https://kanterella.com/w/index.php?title=ML101720553&oldid=966009"