ML103090017
| ML103090017 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 09/17/2010 |
| From: | Operations Branch I |
| To: | Entergy Nuclear Operations |
| Hansell S | |
| Shared Package | |
| ML100980645 | List: |
| References | |
| 50-286/10-401, TAC U01789 | |
| Download: ML103090017 (39) | |
Text
ES-401 PWR Examination Outline Form ES-401-2 Facilitll:
Indian Point Unit 3 Printed: 06/28/2010 Date Of Exam:
10/04/2010 RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G* Total A2 G"
- 1.
1 3
3 3
3 3
3 18 3
3 Emergency 2
2 2
I 2
I 1
9 2
2 N/A N/A Abnormal Tier Plant Evolutions Totals 5
5 4
5 4
4 27 5
5
- 2.
1 3
2 3
3 3
2 3
3 2
2 2
28 3
2 Plant 2
1 I
0 I
1 1
1 I
i I
1 1
10 1 I 1
1 Systems Tier 4
3 3
4 4
3 4
4 3
3 3
38 5
3 Totals
- 3. Generic Knowledge And 1
2 3
4 1
2 4
10 Abilities Categories 2
3 2
3 2
1 2
2 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 (Le., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KiA category shall not be less than two).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table.
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/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 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 KiAs 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 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 KiAs.
- 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 (Note #1 does not apply). Use duplicate pages for RO and SRO-onlyexams.
- 9.
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 KiAs that are linked to 10 CFR 55.43.
Total 6
4 10 5
3 8
1 PWR RO/SRO Examh.....ton Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline ES-401 E/APE # I Name I Safety Function Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 I Kl I K2 I K3 I At I Ai I G I Number KIA Topic Form ES-401-2 I Imp.T Q#
000007 Reactor
- Stabilization - Recovery 1 000009 Small Break LOCA 1 3 0000 I 1 Break LOCA 1 3 000015/000017 RCP Malfunctions ( 4 to determine and interpret the I 4.3 following as they apply to a reactor trip:
Proper actions to be taken if the automatic safety functions have not taken
'-,:-----,---11 u n_n Conduct of Operations - Ability to I 4.3 identifY and interpret diverse indications to validate the response ofanother indication.
Ability to operate and/or monitor the I 4.0 following as they apply to a Large Break 1---__--;1 LOCA:
of RCPs 4.3 2
4 3
WaterI$
4.7 Emergency ProcedureslPlan - Knowledge of crew roles and responsibilities during EOPtise.
4.0 Knowledge of the reasons following responses as they apply to the SGTR: - Criteria for securing/throttling 4.1 5
77 7
8 000038 Steam Gen. Tube Rupture 13 x
PWR RO/SRO Examh...Lion Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline ES-401 1
E/APE # / Name I Safety Function 000040 Steam Line Rupture - Excessive Heat Transfer 14 000056 Loss of Off-site Power / 6 000057 Loss of Vital AC Inst. Bus / 6 000058 Loss of DC Power 16 x
x x
Knowledge ofthe reasons for the following responses as they apply to the Steam Line Rupture: - Operation of steam line isolation valves I...... P..'..'
..... h I 4.2 6
Ability to determine and interpret the following as they apply to a Station Blackout: - Existing valve positioning on a loss of instrument air Knowledge ofthe operational implications ofthe following concepts as they apply to Loss ofOffsite Power:
Definition of subcooling: use of steam tables to determine it I 3.1 10 x
Ability to operate andlor monitor the I 3.2 11 following as they apply to the Loss of Vital AC Instrument Bus: - Backup instrument indications Knowledge ofthe operational 2.8 12 implications ofthe following concepts as they apply to Loss ofDC Power: - Battery and instrumentation 000062 Loss ofNuclcar Svc Water / 4 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: - Location of a leak in the SWS NUREG 1
3
PWR RO/SRO Examib.. don Outline Facility: Indian Pont Unit 3 ES-401 1
Form ES-401-2 E/APE # I Name I Safety Function Number KIA Topic Imp~-~
000077 Generator Voltage and Electric Grid x
Disturbances I 6 3.1 14 15 W/E04 LO~AOlitside Knowledge ofthe interrelations between the Loss of Secondary Heat Sink and the following: - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features W/E 11 Loss of Emergency Coolant Recirc. 14 x
Knowledge of the interrelations between I 3.6 the Loss of Emergency Coolant Recirculation and the following:
Components, and functions ofcontrol and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features NUREG 1021 4
17
PWR RO/SRO Examh... don Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline ES-401 1
E/APE # 1Name 1Safety Function Number WIE I 2 - Steam Line Transfer 14
- Excessive Heat x
EKI.!
Knowledge ofthe operational implications of the following concepts as they apply to the Uncontrolled Depressurization of all Steam Generators:
- Components:, capacity, and function of 3.4 NUREG 1021 5
PWR RO/SRO Examh...don Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline 2
ES-401 E/APE # I Name / Safety Function 000001 Continuous Rod Withdrawal / 1 x
2.9 19 on I 21 000036 000068 Control Room Evac. / 8 000033 Loss of Intennediate Rang NI / 7 I X I
_IAK3.02 I Kno;ledge;f the r~asons for the I 3.6 following responses as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Guidance contained in EOP for loss of intermediate-range instrumentation Ability to operate and/or monitor the following as they apply to the Fuel Incidents: - ARM 2.1.37 I Conduct of Operations - Knowledge of procedures, guidelines, or limitations associated with Ability to determine and interoretthe following as Reactor Coolant I 3.1 I 4.3 3.2 I
I j
22 23 84 NUREG 1021 6
PWR RO/SRO Examill.. ~lon Outline NRC Written Examination Outline Facility: Indian Pont Unit 3 ES-401 E/APE # I Name I Safety Function 2
W/E02 SI Tennination I 3 x
W/E03 LOCA Cooldown - Depress. /4 W/E06 Inad. Core Cooling / 4 Knowledge of the operational implications of the following concepts as they apply to the SI Termination: -
Normal, abnormal and emergency operating procedures associated with SI Termination Ability to determine and interpret the following as they apply to the LOCA Cool down and Depressurization:
Facility conditions and selection of appropriate procedures during abnormal and 3.4 3.4 24 25 Ability to operate and/or monitor the following as they apply to the Degraded Core Cooling: - Desired operating results during abnormal and emergency situations ij~to 3.7 26 x
Knowledge of the interrelations between the High Containment Pressure and the following: - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features NUREG 1021 7
PWR RO/SRO Exahullation Outline Facility: Indian Pont Unit 3 003 Reactor Coolant Pump I X I Kl.Ol I Knowledge of the physical connections I 2.6 I 28 andlor cause-effect relationships between the RCPS and the following systems: -
003 Reactor 43 malfunction of the RCPS will have on the following: - RCS I 4.0 I 004 Chemical and Volume Knowledge of limiting conditions for 30 Control operations and safety limits 005 Residual Heat X
Knowledge of bus power supplies to the 3.0 I 31 Removal following: - RHR pumps 005 Heat Ability to (a) predict the impacts ofthe 2.9 I 32 Removal following malfunctions or operations on the RHRS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - RHR malfunction 006 Emergency Core X
K3.03 Knowledge of the that a loss or I 4.2 I 33 Cooling malfunction ofthe ECCS will have on the followi 006 EmerllerievCore NUREG 1021 8
PWR RO/SRO Exahdflation Outline Facility: Indian Pont Unit 3 007 Pressurizer Relief/Quench Tank 008 Component Cooling Water 008 Component Cooling Water 010 Pressurizer Pressure Control 012 Reactor Protection 013 Engineered Safety Features Actuation I X I A2.04 Kl.OI X
A4.08 I X I B\\.
K5.02 x
IAbility to (a) predict the impacts ofthe I 2.5 I 34 following malfunctions or operations on the PRTS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: -
Overpressurization of the waste gas vent header IKnowledge of the physical connections I 3.1 I 35 and/or cause-effect relationships between the CCWS and the following systems: -
SWS IAbility to manually operate and/or I 3.1
- I 36 monitor in thp control room: - CCW
- witch I 2.6 I I Knowledl!e of the ooerational 37 3.9 I 38 Ability to monitor automatic operation of I 3.8 the RPS. including: - Individual channel Knowledge of the effect ofa loss or I 2.7*
malfunction ofthe following will have on the ESF AS: - Sensors and detectors 29 39 NUREG 1021 9
PWR RO/SRO Exa.....aation Outline Facility: Indian Pont Unit 3 Form ES-40.:;-1-::::....2-------i 1iil,,_.u~_.
! --- - -~t"-
lJIll}) I Q#
013 Engineered Safety Features Actuation 022 Containment Cooling 059 Main Feedwater x
x x
Ability to manually operate and/or monitor in the control room: - CSS controls Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MFW System controls including:
Power level restrictions for operation of 3.6 3.6 1MFW pu111ps a~dvalvejs
.'. O~r~tr~ps - Abilityt<>
4.4 "ifi"';ij(l'\\,,,tp'",, andilltegrated all m~~esofplapt Knowledge ofthe operational implications of the following concepts as they apply to the AFW System:
Relationship between AFW flow and
'-___--L. RCS heat transfer 3.6 40 41 44 89 45 NUREG 1021 10
Facility: Indian Pont Unit 3 ES-401 06 I Auxiliary/Emergency Feedwater 062 AC Electrical Distribution 063 DC Electrical Distribution 064 Emergency Diesel Generator PWR RO/SRO Exa..llDation Outline x
x Form ES-401-2 I~n -------r-Knowledge of AFW System design I 4. I feature(s) and/or interlock(s) which provide for the following: - Water sources and nrioritv of use Ability to (a) predict the impacts ofthe I 3.2 following malfunctions or operations on the A.C. Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences ofthose malfunctions or operations: - Restoration ofpower to a systel1l~ith a fault on it Knowledge ofD.C. Electrical System I 2.7 design feature(s) and/or interlock(s) which provide for the following: -
Manual/automatic transfers of con~ohl I __~I te Knowledge ofthe effect of a loss or malfunction of the following will have on the ED/G Svstem: - Air receivers Ability to predict and/or monitor changes I 3.1 in parameters (to prevent exceeding design limits) associated with operating the ED/G System controls including: -
Maintaining minimum load on ED/G (to nrpvpnt reverse 46 47 48 50 NUREG 1021 11
PWR RO/SRO Exa***mation Outline Facility: Indian Pont Unit 3 073 Process Radiation KJ.Ol IKnowledge ofthe physical connections
) 3.6 )
51 I X I Monitoring and/or cause-effect relationships between the PRM System and the following
- - Those s stems served by PRMs 073 Process Radiation Knowledge ofthe operational I 2.5 I 52 Monitoring 11_' A1.02 implications of the following concepts as they apply to the PRM System:
Radiation theory, including sources, units, and effects
- 1 076 Service Water X_
1 Ability to predict and/or monitor changes 12.6* I 53 in parameters (to prevent exceeding design limits) associated with operating the SWS controls including: - Reactor and turbine building closed cooling water ItemEeratures 078 Instrument Air 3.1
.xl
- A3.0J
. Ability to monitor automatic operation of 54 103 Containment 4.3 to 55 identifY and interpret diverse indications to validate the response of another indication.
NUREG 1021 12
PWR RO/SRO Exa.... nation Outline Facility: Indian Pont Unit 3 I J;/A Topic 002 Reactor Coolant 01 1 Pressurizer Level Control 014 Rod Position Indication x
x x
3.2 Knowledge of the physical connections 56 and/or cause-effect relationships between Res and the following systems:
PRT 3.3 57 3.1 63 Instrumentation 017 In-core Temperature Monitor x
2.7 59 NUREG 1021 13
PWR RO/SRO Exa..... nation Outline Facility: Indian Pont Unit 3 NRC Written Examination Outline Pilant ~~stems - ~ier 2.L~~,oup 2 035 Steam Generator A2.06 I Ability to (a) predict the impacts ofthe I 4.5 I 60 following malfunctions or operations on the S/GS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Small break LOCA 041 Steam Dump/Turbine A3.03 Ability to monitor automatic operation of I 2.7 I 61 Bypass Control the SDS, including: - Steam flow 045 Main Turbine 2.1.43 Conduct of Operations - Ability to use I 4.1 I 62 Generator procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, etc.
086 Fire Protection 3.3 65 NUREG 1021 14
Generic Knowledge and Abilities Outline (Tier 3)
PWR RO/SRO Examination Outline
. Facility: Indian Pont Unit 3 Form ES-401-3 Facility Indian Point Unit 3 Date of Exam 10/412010
. Category KIA #
Topic RO SRO-Only
~--""---"------i
- 1. Conduct of IR Q#
IR Q#
Operations 2.1.39 Knowledge of conservative decision making 3.6 67 practices.
Equipment Control - Knowledge ofbases in technical speeifications for limiting
- 2. Equipment 2.2.25 Control conditions for 0 erations and safet limits.
2.2.37 Equipment Control - Ability to determine operability and/or availability of safety
- related e UI ment.
2.2.39 Knowledge of less than or equal to one hour 2.1.40 Subtotal Technical Specification action statements for
! systems.
68 4.8 94 95 3.2 69 3.6 66 3.9 70 4.3 96 3
1 NUREG 1021 15
Generic Knowledge and Abilities Outline (Tier 3)
PWR RO/SRO Examination Outline I Facility: Indian Pont Unit 3 Form ES-401-3 Facility Indian Point Unit 3 Date of Exam 10/412010 Category
- K/A #
Topic RO SRO-Only IR Q#
IR activities.
Q#
2.3.12 Knowledge of radiological safety principles 3.2 71 pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
2.3.14 Knowledge of radiation or contamination 3.4 72 hazards that may arise during normal, abnormal, or emergency conditions or
- 3.
Radiological 97 Controls 98 Subtotal 2
2 2.4.20 Knowledge of operational implications of 3.8 73 EOP warnings, cautions, and notes.
2.4.34 Knowledge ofRO tasks performed outside 4.2 74 the main control room during an emergency and the resultant operational effects.
2.4.45 Ability to prioritize and interpret the 4.1 75 significance of each annunciator or alarm.
- 4. Emergency Procedures/plan 2.4.11 9f<l~~~!1~f;g~~Jtion 4.2 99
- 2.4.32 4.0 100 Subtotal 3
2 Tier 3 Point Totals 10 7
NUREG 1021 16
Description of program used to generate IPEC Unit 3 July 2010 Written Exam KlAs Generated the RO and SRO sample plan using the "NKEG" Database Program, version 1.1, developed by Westinghouse Electric Company. This program will automatically produce a Random Sarnple Plan based on NUREG 1122, Rev. 2, Supplement 1 KlAs.
KfAs were suppressed prior to the outline generation process as provided for in the examiner standard, the list of suppressed KfAs is provided as required by the examiners standard.
Inappropriate and inapplicable KfAs were discarded during the outline development process and are included in the record of rejected KfAs. The replacement KfAs were replaced using the random sample function of the NKEG database program.
NUREG 1021 17
r~~
Tier /
Group R-1/1 000008 AK1.02 R-1/1 000009 G 2.2.40 R-1/1 000011 EA2.09 R-1/1 0000221.08 R-1/1 000027 G2.2.15 R-1/1 000027 AK3.03 S-1/1 0000402.4.6 R-1/1 000062 G2.1.3 R-1/2 000076 G2.2.41 Randomly Selected KIA Knowledge of the operational implications of the following concepts as they apply to a Pressurizer Vapor Space Accident: - Change in leak rate with change in pressure Equipment Control - Ability to apply technical specifications fora system.
Ability to determine and interpret the following as they apply to a Large Break LOCA: - Existence of adequate natural circulation Ability to operate and/or monitor the following as they apply to the Loss of Reactor Coolant Pump Makeup:-:YCT level Equipment Control - Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line ups, tag outs, etc.
Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: - Actions contained in EOP for PZR PCS malfunction Emergency Procedures/Plan - Knowledge of EOP mitigation strategies.
Conduct of Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Ability to obtain and interpret station electrical and mechanical drawings.
Reason for Rejection Rejected due to similarities to QUestion # 2 This Generic KA is not applicable to off~
normal procedures KA is not applicable. Natural Circulation does not exist during a Large Break LOCA.
KA is not applicable. Makeup to RCP standpipe if from primary water not CVCS This Generic KA is not applicable for APE/EPE written examination. This KA is evaluated during Simulator and/or JPM Examinations.
KA Rejected due to similarity with Simulator Examination Rejected due to inability to develop an SRO level question for mitigation strategies for Steam LineRu(:Jtur~.
This Generic KA is evaluated during Simulator Exam This Generic KA is not applicable for APE/EPE written examination. This KA is evaluated during Simulator and/or JPM Examinations.
R-1/2 000024 AK3.02 S-1/2 000051 G2.3.14 R-1/2 000067 AA 1.08 S-1/2 000068 G2.3.14 S-1/2 000068 G2.2.37 R-1/2 W/E16 G2.4.47 I---~~~~ ~
~ ~ ~~~~-~~~~
R-2/1 004000 K4.13 R-2/1 004000 A4.04
~~~~~~-
R-2/1 006000 K3.02 R-2/1 007000 G2.2.42 Knowledge of the reasons for the following responses as they apply to the Emergency Boration:
- Actions contained in EOP for emerQencv boration Radiological Controls - Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emerQency conditions or activities.
Ability to operate and/or monitor the following as they apply to the Plant Fire on Site: - Fire fighting equipment used on each class of fire Radiological Controls - Knowledge of radiation or contamination hazards that may arise during normal, a bn()rma I, or emerQency conditions or activities.
Equipment Control - Ability to determine operability and/or availability of safety related equipment.
Emergency Procedures/Plan - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
Knowledge of CVCS design feature(s) and/or interlock(s) which provide for the following:
Interlock between letdown isolation valve and flow control valve Ability to manually operate and/or monitor in the control room: - Calculation of boron concentration changes Knowledge of the effect that a loss or malfunction of the ECCS will have on the foliowinQ: - Fuel Equipment Control - Ability to recognize system parameters that are entry level conditions for Technical Specifications.
This concept is evaluated on JPM exam This Generic KA is not applicable to Loss of Condenser Vacuum off-normal procedure.
Unable to write a discriminatory RO level question for this KA Unable to write a discriminatory SRO level question for this KA This generic KA is not applicable to Control Room Evacuation APE.
This Generic KA is evaluated during Simulator Exam NA at IPEC - No Interlock exists.
Rejected due to similarity with JPM KA Rejected due to similarity with question 26 NA at IPEC - No Tech Spec or TRM exist for the PRT
R-2/1 010000 G2.4.47 R-2/1 039000 K3.06 012000 A2.06 S-2/1 R-2/1 062000 A3.01 R-2/2 028000 K2.01 R-2/2 034000 K4.02 R-2/2 072000 K4.02 S-2/2 075000 2.1.15 R-2/2 079000 A4.01 R-3 2.1.19 2.2.7 S-3 Emergency Procedures/Plan - Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
Knowledge of the effect that a loss or malfunction of the MRSS will have on the following: - SOS Ability to (a) predict the impacts of the following malfunctions or operations on the RPS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Failure of RPS signal to trip the reactor Ability to monitor automatic operation of the A.C.
Distribution System, including: - Vital ac bus amperage Knowledge of bus power supplies to the following:
Hydrogen recombiners Knowledge of Fuel Handling System design feature(s) and/or interlock(s) which provide for the following: - Fuel movement Knowledge of ARM system design feature(s) and/or interlock(s) which provide for the following: - Fuel building isolation Conduct of Operations - Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, Operations memos, etc.
Ability to manually operate and/or monitor in the control room: - Cross-tie valves with lAS Ability to use plant computers to evaluate system or component status.
Knowledge of the process for conducting special or infrequent tests.
This KA is evaluated during Simulator Exam Rejected due to similarity with question 61 Rejected due to similarities with Question 38 NA at IPEC - Not able to monitor Vital AC bus amp in the control room Unable to write a discriminatory RO question for this KA Rejected RO are not directly involved in Fuel Handling activities at IPEC.
KA rejected due to similarity with question 22 This Generic KA is not applicable to the Circulating Water System.
KA rejected due to similarity with question 14 This is evaluated during Simulator Exam Unable to write a discriminatory SRO question for this KA
R-3 2.2.13 Knowledge of tagging and clearance procedures.
This concept is evaluated during JPM Exam R-3 2.2.20 Knowledge of the process for managing troubleshooting activities.
KA rejected because ROs do not manage troubleshooting activities.
~~~~~~~~~~~~~~~~~~~~~
ES-301 Administrative Topics Outline Form ES-301-1 Date of Examination: October 4,2010 Operating Test Number:
Describe activity to be performed Review a gross leakrate determination calculation 2.1.7 (4.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretations Review Reactor Vessel Venting Time Per FR-1.3, 2.1.25 (4.2) Ability to interpret reference materials, such as graphs, curves, tables, etc.
Review a Safety Function Determination 2.2.37 (4.6) Ability to determine operability and/or availability of safety related equipment Review a Manual Gaseous Rad Waste Release Permit 2.3.6 (3.8) Ability to approve release permits Classify E-Plan Event and Complete Part 1 Form 2.4.38 (4.4) Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.
Facility:
Indian Point Unit 3 Examination Level: ROO Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan SRO-U X Type Code*
N,R M,R N,R N,R M,R,P I
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:
Indian Point Unit 3 Date of Examination: October 4, 2010 Exam Level: RO D SRO-I D SRO-U X Operating Test No.:
System I JPM Title Type Code" Safety Function
- a.
NA for SRO-U na na lb.
Transfer to Cold Leg Recirculation D, A. EN, L 2
- c.
NA for SRO-U na na
- d.
NA for SRO-U na na
- e.
NA forSRO-U na na
- f.
Remove a Power Range Nuclear Instrument from Service D
7
- g.
Reset R-18 Alarm Setpoint Using RM-23A N
9
- h.
NA for SRO-U na na In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i.
Locally Start 32 Auxiliary Boiler Feed Pump D,A,E, L 4-S
- j.
NA for SRO-U na na
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-11 steps 1-5)
N, EN, L, R 8
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 Criteria for RO I SRO-II SRO-U (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power 1Shutdown (N)ew or (M)odified from bank including 1 (A)
(P)revious 2 exams (R)CA (S)imulator 4-6 1 4-6 1 2-3 s9/s8/s4 2!1/2!1/2!1
- I - 1 2!1 (control room system) 2!1/2!1/2!1 2!2/2!2/2!1 s 31 s 31 s 2 (randomly selected) 2!1/2!1/2!1
,i
ES-301 Form ES-301-1 Administrative TO~~3 9utline INDIAN POINT UNIT 3 NRC &EXAMINATION CONDUCT OF OPERATIONS: Review a gross leakrate determination calculation. The candidate will be given a set of conditions and appropriate indications and a manualleakrate calculation. The candidate will review the leakrate calculation and locate errors.
This is a New JPM SROOnly CONDUCT OF OPERATIONS: Review Reactor Vessel Venting Time Per FR-I.3, Attachment 4 The candidate will be given a set of plant conditions with a void in the reactor vessel head.
The candidate will be given a reactor vessel venting time calculation and directions to review calculation. The calculation is incorrect and the candidate must locate the errors.
EQUIPMENT CONTROL: Review a Safety Function Determination. The candidate will be given a set of plant/equipment conditions including degraded power sources and equipment failures, The candidate will be directed to perform a Safety Function Determination in accordance with OAP-034.
RADIATION CONTROL: Review a Manual Gaseous Rad Waste Release Permit. The candidate will be given a set of plant conditions for the VC Pressure Relief and a manually prepared release permit The candidate will be directed to review the release permit in accordance with 3-S0P-WDS-013, The candidate must locate errors on the permit.
This is a Modified Bank JPM SRO Only.
EMERGENCY PROCEDURES/PLAN: Classify E-Plan Event and Complete Part 1 Form. The candidate will be given a set of plant conditions. The candidate must evaluate the conditions, determine the emergency plan classification and complete the New York State Part 1 form This is a Modified Bank JPM.
This is a Time Critical,IPM SRO Only.
ES-301 Control Roomlln-Plant Systems Outline Form ES-301-2
- a.
NA for SRO-U
- b.
Transfer to Cold Leg Recirculation: The candidate will enter the simulator following a large break LOCA. The RWST will be at the level requiring transfer to cold leg recirculation. The candidate will be directed to establish cold leg recirculation using 3 ES-1.3. The recirculation pumps will not start and the candidate will be required to establish recirculation using the RHR pumps This is a bank JPM.
Failure to properly perform this JPM will result in inadequate cooling of the core and possible core damage.
- c.
NA for SRO-U
- d.
NA for SRO-U
- e.
NA for SRO-U
- f.
Remove a Power Range Nuclear Instrument from Service: The candidate will enter the simulator at any power level. The candidate will be told that one Power Range NI channel is indicating erratically and the Shift Manager has declared the channel inoperable. The candidate will be directed to remove the Power Range Channel from service in accordance with 3-S0P-NI-001.
This is a bank JPM.
Failure to properly perform this task will result in a Tech Spec violation!
- g.
Reset R-18 Alarm Setpoint Using RM-23A: The candidate will enter the simulator at any power level. The candidate will be directed to reset the Alarm setpoint for R-18, Liquid Rad Waste Release Monitor. The candidate will be told that the Bantam 11 computer system is OOS and the setpoint must be changed using the RM-23A for R-18.
This is a New JPM.
Failure to properly perform this task will result in possible release of radioactive liquid to the river beyond allowable limits.
- h.
NA for SRO-U In Plant JPMs
- i.
Locally Start 32 Auxiliary Boiler Feed Pump: This JPM locally starts the Steam Driven Aux Feedwater Pump. During the startup the steam pressure control valve will not maintain pressure at 600 psig. The candidate will be required to control steam pressure using PCV-1139 Auto!Manual station.
In Plant JPM This is an Alternate Path JPM This is a Bank JPM.
Failure to properly perform this task will result in inability to control SG level during a control room evacuation.
- j.
NA for SRO-U.
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-11 steps 1-5).
This JPM isolates lines that penetrate containment when equipment is shutdown during a post-accident condition. The candidate will be required to locate and simulate opening/closing valves and circuit breakers for containment isolation valves.
Failure to properly perform this task will result in inability to maintain ReS inventory and possible core damage.
i ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Indian Point Unit 3 Examination Level: RO D Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan SRO-I X Type Code*
N,R M,R, N,R N,R M,R,P Date of Examination: October 4, 2010 Operating Test Number:
Describe activity to be performed Review a gross leakrate determination calculation 2.1.7 (4.7) Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretations Review Reactor Vessel Venting Time Per FR-I.3, 2.1.25 (4.2) Ability to interpret reference materials, such as graphs, curves, tables, etc.
Review a Safety Function Determination 2.2.37 (4.6) Ability to determine operability and/or availability of safety related equipment Review a Manual Gaseous Rad Waste Release Permit 2.3.6 (3.8) Ability to approve release permits Classify E-Plan Event and Complete Part 1 Form 2.4.38 (4.4) Ability to take actions called for in the facility emergency plan, including supporting or acting i as emergency coordinator if required.
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:
Indian Point Unit 3 Date of Examination: October 4, 2010 Exam Level: RO 0 SRO-I X SRO-U D Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title
- a.
Emergency Borate
- b.
Transfer to Cold Leg Recirculation
- c.
Respond to a Pressurizer Controlling Channel Fail High
- d.
Initiate Bleed and Feed of the RCS
- e.
Start the Hydrogen Recombiner
- f.
Remove a Power Range Nuclear Instrument from Service
- g.
Reset R-18 Alarm Setpoint USing RM-23A
- h.
NA for SRO-I In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i.
Locally Start 32 Auxiliary Boiler Feed Pump
- j.
Start a CCW Pump from MCC 312A
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-ll steps 1-5)
Type Code*
Safety Function M,A, P,S,L 1
D, A, E, EN, L 2
D 3
D,A 4-P N, E, L 5
D 7
N 9
D,A,E, L 4-S D,E,L 6
N, EN, L, R 8
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 / Shutdown (N)ew or (M)odified from bank including 1 (A)
(P)revious 2 exams (R)CA II (S)imulator Criteria for RO 1SRO-II SRO-U 4-6 1 4-6/2-3 s9/s8/s4
- 1/:::1/;::1 / :::1 (control room system)
- 1/;::1/;::1
- 2/:::2/;::1 s 31 s 3 / s 2 (randomly selected)
- 1/
- ::1/;::1 i
i
ES-301 Administrative Topics Outline Form ES-301-1 9--u-r..
INDIAN POINT UNIT 3 NRC Rq'EXAMINATION CONDUCT OF OPERATIONS: Review a gross leakrate determination calculation. The candidate will be given a set of conditions and appropriate indications and a manualleakrate calculation. The candidate will review the leakrate calculation and locate errors.
This is a New,IPM SRO Only CONDUCT OF OPERATIONS: Review Reactor Vessel Venting Time Per FR-1.3, Attachment 4 The candidate will be given a set of plant conditions with a void in the reactor vessel head.
The candidate will be given a reactor vessel venting time calculation and directions to review calculation. The calculation is incorrect and the candidate must locate the errors.
EQUIPMENT CONTROL: Review a Safety Function Determination. The candidate will be given a set of plant/equipment conditions including degraded power sources and equipment failures. The candidate will be directed to Review a Safety Function Determination in accordance with OAP-034.
RADIATION CONTROL: Review a Manual Gaseous Rad Waste Release Permit. The candidate will be given a set of plant conditions for the VC Pressure Relief and a manually prepared release permit. The candidate will be directed to review the release permit in accordance with 3-S0P-WDS-013. The candidate must locate errors on the permit.
This is a Modified Bank JPM SRO Only.
EMERGENCY PROCEDURES/PLAN: Classify E-Plan Event and Complete Part 1 Form. The candidate will be given a set of plant conditions. The candidate must evaluate the conditions, determine the emergency plan classification and complete the New York State Part 1 form This is a Modified Bank JPM.
This is a Time Critical JPM SRO Only.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- a.
Emergency Borate: The candidate will enter the simulator following a reactor trip with 2 stuck rods. The turbine failed to automatically trip on the reactor trip resulting in a cooldown below 540°F. Emergency boration valve MOV-333 will not open. The candidate must use Attachment 1 of ONOP-CVCS-3 to establish emergency boration Using Normal Boration. The candidate must first borate for temperature and then borate for the additional stuck rod.
This is an Alternate Path JPM.
This is a Modified Bank.IPM Failure to properly perform this task will result in possible inadequate shutdown margin.
- b.
Transfer to Cold Leg Recirculation: The candidate will enter the simulator following a large break LOCA. The RWST will be at the level requiring transfer to cold leg recirculation. The candidate will be directed to establish cold leg recirculation using 3 ES-1.3. The recirculation pumps will not start and the candidate will be required to establish recirculation using the RHR pumps This is a bank JPM.
Failure to properly perform this JPM will result in inadequate cooling of the core and possible core damage.
- c.
Respond to a Pressurizer Controlling Pressure Channel Failure High: The candidate will enter the simulator at normal full power lineup. The controlling pressurizer pressure channel will fail resulting in maximum spray flow and pressurizer pressure decreasing rapidly. The candidate must take manual control of the master pressurizer pressure controller and stabilize pressure. The candidate must then take appropriate actions to trip bistables associated with pressurizer pressure.
This is a bank JPM.
Failure to properly perform this task will result in a reactor trip.
- d.
Initiate Bleed and Feed of the ReS: The candidate will enter the simulator with the plant tripped. The candidate will be informed that a transition to 3-FR-H.1, Loss of Secondary Heat Sink, has been entered and conditions are met to establish Bleed and Feed cooling of the RCS. The candidate will be directed to establish Bleed and Feed Cooling. One PORV will not open and the candidate will be required to install the head vent fuses and open all head vents.
This is an Alternate Path JPM.
This JPM directly from the JPM bank; however, it has not been used on the previous 2 NRC Exams.
Failure to properly perform this task will result in inadequate core cooling and fuel damage.
- e.
Start a Hydrogen Recombiner: The candidate will enter the simulator following simulated a large break LOCA. The plant will be lined up for cold leg recirculation near the end of 3-ES-1. 3, Transfer to Cold Leg Recirculation. The hydrogen concentration has been determined to be approximately 2%. The candidate will be directed to place the hydrogen recombiner in service using 3-S0P-CB-007. (NOTE: The hydrogen recombiner panel is a separate panel and it is not modeled in the simulator computer.
This.IPM will be simUlated).
This is a New JPM.
Failure to properly perform this task will result in excessive hydrogen buildup in the containment building.
- f.
Remove a Power Range Nuclear Instrument from Service: The candidate will enter the simulator at any power level. The candidate will be told that one Power Range NI channel is indicating erratically and the Shift Manager has declared the channel inoperable. The candidate will be directed to remove the Power Range Channel from service in accordance with 3-S0P-NI-001.
This is a bank JPM.
Failure to properly perform this task will result in a Tech Spec violation.
- g.
Reset R-18 Alarm Setpoint Using RM-23A: The candidate will enter the simulator at any power level. The candidate will be directed to reset the Alarm setpoint for R-18, Liquid Rad Waste Release Monitor. The candidate will be told that the Bantam 11 computer system is OOS and the setpoint must be changed using the RM-23A for R-18.
This is a New JPM.
Failure to properly perform this task will result in possible release of radioactive liquid to the river beyond allowable limits.
- h.
- i.
Locally Start 32 Auxiliary Boiler Feed Pump: This JPM locally starts the Steam Driven Aux Feedwater Pump. During the startup the steam pressure control valve will not maintain pressure at 600 psig. The candidate will be required to control steam pressure using PCV-1139 Auto/Manual station.
In Plant JPM This is an Alternate Path JPM This is a Bank JPM.
Failure to properly perform this task will result in inability to control SG level during a control room evacuation.
- j.
Start a CCW Pump from MCC 312A: This JPM locally starts a CCW pump from outside the control room due to a control room evacuation. Loads must be stripped from the MCC that supplies the Appendix R MCC.
In Plant JPM This is a Bank JPM.
Failure to properly perform this task will result in loss of cooling to the RCP seals
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-11 steps 1-5).
This JPM isolates lines that penetrate containment when equipment is shutdown during a post-accident condition. The candidate will be required to locate and simulate opening/closing valves and circuit breakers for containment isolation valves.
Failure to properly perform this task will result in inability to maintain RCS inventory and possible core damage.
/I ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Indian Point Unit 3 Examination Level: RO X SRO Administrative Topic (see Note)
Conduct of Operations Conduct of Operations Equipment Control Radiation Control Emergency Procedures/Plan Type Code,o N,R M,R M,R N,R na Date of Examination: October 4, 2010 Operating Test Number:
Describe activity to be performed Calculate a Power Reduction Reactivity Plan per POP-2.1 2.1.43 (4.1) Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plan, fuel depletion, etc..
Determine Reactor Vessel Venting Time Per FR-I.3, 2.1.25 (3.9)Ability to interpret reference materials, such as graphs, curves, tables, etc.
Perform a Peer Review of a Surveillance Test 2.2.21 (2.9) Knowledge of pre-and post-maintenance operability requirements Perform a SG Tube Leakrate Determination using 3 AOP-SG-1 2.3.5 (2.9)Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, ets.
NA for ROs 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:
Indian Point Unit 3 Date of Examination: October 4,2010 Exam Level: RO X SRO-I D SRO-U D Operating Test No.:
Control Room Systems@ (8 for RO); (7 for SRO-I); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title Ia.
Emergency Borate
- b.
Transfer to Cold Leg Recirculation
- c.
Respond to a Pressurizer Controlling Pressure Channel Failure High
- d.
Initiate Bleed and Feed of the RCS Start the Hydrogen Recombiner
- e.
- f.
Remove a Power Range Nuclear Instrument from Service
- g.
Reset R-18 Alarm Setpoint Using RM-23A
- h.
Transfer buses 1 - 4 to the Station Aux Transformer
- In-Plant Systems@ (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i.
Locally Start 32 Auxiliary Boiler Feed Pump j,
Start a CCW Pump from MCC 312A
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-11 steps 1-5)
Type Code" M,A, S, L D,A, EN, L D
Safety Function 1
2 3
D,A N, L D
N N
4-P 5
7 9
6 D,A, E, L D, E, L N, EN, L, R 4-S 6
8 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 I SRO-II SRO-U 4-6/4-6/2-3 s9/s8/s4
~1/~1/~1 I <1:1 (control room system)
~1/~1/~1
<1:2/~2/<1: 1 s 3 I s 3/ s 2 (randomly selected)
~1/~1/~1
ES-301 Administrative Topics Outline Form ES-301-1 INDIAN POINT UNIT 3 NRC RO EXAMINATION CONDUCT OF OPERATIONS: Calculate a Power Reduction Reactivity Plan per POP 2.1.
The candidate will be given appropriate graphs and tables and directed to calculate a reactivity plan for a power reduction from 100% to 75% power.
This is a New JPM RO Only CONDUCT OF OPERATIONS: Determine Reactor Vessel Venting Time Per FR-I.3,. The candidate will be given a set of plant conditions with a void in the reactor vessel head. The candidate will be given directions to calculate the reactor vessel venting time in accordance with Functional Restoration Procedure FR-1.3 This is a bank JPM EQUIPMENT CONTROL: Perform a Peer Review of a Surveillance Test. The candidate will be given a completed Surveillance Test and directed to perform a Peer Review. The Surveillance Test will have inaccurate information This is a New JPM RO Only RADIATION CONTROL: Perform a SG Tube Leakrate Determination using 3-AOP-SG-1.
The candidate will be given Radiation Monitor and the worksheet from 3-AOP-SG-1 and directed to calculate SG Tube leakage.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2
- a.
Emergency Borate: The candidate will enter the simulator following a reactor trip with 2 stuck rods. The turbine failed to automatically trip on the reactor trip resulting in a cooldown below 540°F. Emergency boration valve MOV-333 will not open. The candidate must use Attachment 1 of ONOP-CVCS-3 to establish emergency boration Using Normal Boration. The candidate must first borate for temperature and then borate for the additional stuck rod.
This is an Alternate Path JPM.
This is a Modified Bank JPM Failure to properly perform this task will result in possible inadequate shutdown margin.
- b.
Transfer to Cold Leg Recirculation: The candidate will enter the simulator following a large break LOCA. The RWST will be at the level requiring transfer to cold leg recirculation. The candidate will be directed to establish cold leg recirculation using 3 ES-1.3. The recirculation pumps will not start and the candidate will be required to establish recirculation using the RHR pumps This is a bank JPM.
Failure to properly perform this JPM will result in inadequate cooling of the core and possible core damage.
- c.
Respond to a Pressurizer Controlling Pressure Channel Failure High: The candidate will enter the simulator at normal full power lineup. The controlling pressurizer pressure channel will fail resulting in maximum spray flow and pressurizer pressure decreasing rapidly. The candidate must take manual control of the master pressurizer pressure controller and stabilize pressure. The candidate must then take appropriate actions to trip bistables associated with pressurizer pressure.
This is a bank JPM.
Failure to properly perform this task will result in a reactor trip.
- d.
Initiate Bleed and Feed ofthe RCS: The candidate will enter the simulator with the plant tripped. The candidate will be informed that a transition to 3-FR-H.1, Loss of Secondary Heat Sink, has been entered and conditions are met to establish Bleed and Feed cooling of the RCS. The candidate will be directed to establish Bleed and Feed Cooling. One PORV will not open and the candidate will be required to install the head vent fuses and open all head vents.
This is an Alternate Path JPM.
This,IPM directly from the,IPM bank; however, it has not been used on the previous 2 NRC Exams.
Failure to properly perform this task will result in inadequate core cooling and fuel damage.
- e.
Start a Hydrogen Recombiner: The candidate will enter the simulator following a large break LOCA. The plant will be lined up for cold leg recirculation near the end of 3-ES 1.3, Transfer to Cold Leg Recirculation. The hydrogen concentration has been determined to be approximately 2%. The candidate will be directed to place the hydrogen recombiner in service using 3-S0P-CB-007. (NOTE: The hydrogen recombiner panel is a separate panel and the actual simulator setup conditions are not required).
This is a New JPM.
Failure to properly perform this task will result in excessive hydrogen buildup in the containment building.
- f.
Remove a Power Range Nuclear Instrument from Service: The candidate will enter the simulator at any power level. The candidate will be told that one Power Range NI channel is indicating erratically and the Shift Manager has declared the channel inoperable. The candidate will be directed to remove the Power Range Channel from service in accordance with 3-S0P-NI-001.
This is a bank JPM.
Failure to properly perform this task will result in a Tech Spec violation.
- g.
Reset R-18 Alarm Setpoint Using RM-23A: The candidate will enter the simulator at any power level. The candidate will be directed to reset the Alarm setpoint for R-18, Liquid Rad Waste Release Monitor. The candidate will be told that the Bantam 11 computer system is OOS and the setpoint must be changed using the RM-23A for R-18.
This is a New JPM.
Failure to properly perform this task will result in possible release of radioactive liquid to the river beyond allowable limits.
- h.
Transfer buses 1 - 4 to the Station Aux Transformer.
The candidate will enter the simulator at low power during a plant shutdown. Turbine load will be less than 40 MWe.
The candidate will be directed to transfer 6.9 kV buses 1 - 4 from the Unit Aux Transformer to the Station Aux Transformer in preparation for a turbine shutdown.
This is a New JPM.
Failure to properly perform this task may result in the loss of 6.9 kV buses 1 - 4.
In Plant JPMs
- i.
Locally Start 32 Auxiliary Boiler Feed Pump: This JPM locally starts the Steam Driven Aux Feedwater Pump. During the startup the steam pressure control valve will not maintain pressure at 600 psig. The candidate will be required to control steam pressure using PCV-1139 Auto/Manual station.
In Plant JPM This is an Alternate Path JPM This is a Bank JPM.
Failure to properly perform this task will result in inability to control SG level during a control room evacuation.
- j.
Start a CCW Pump from MCC 312A: This JPM locally starts a CCW pump from outside the control room due to a control room evacuation. Loads must be stripped from the MCC that supplies the Appendix R MCC.
In Plant JPM This is a Bank JPM.
Failure to properly perform this task will result in loss of cooling to the RCP seals
- k.
Perform Local Containment Isolation Valve Lineup IVSW (SOP-CB-11 steps 1-5).
This JPM isolates lines that penetrate containment when equipment is shutdown during a post-accident condition. The candidate will be required to locate and simulate opening/closing valves and circuit breakers for containment isolation valves.
Failure to properly perform this task will result in inability to maintain RCS inventory and possible core damage.
Appendix 0 Scenario Outline Form ES-O-1 Facility:
IPEC Unit 3 Scenario No.:
1 Op-Test No.:
1 Examiners:
Operators:
Initial Conditions:
Plant is at 46%. 31 EDG is OOS for an oil system modification.
Turnover:
Return power to 100%. 31 EDG is OOS and not expected back on this shift.
Event No.
Malt. No.
Event Type*
Event Description 1
N/A R(ATC)
N(BOP)
N(CRS}
Perform power ascension.
2 MAL PRS006 B
I(ALL)
TS(CRS)
Controlling PZR level instrument fails low causing letdown to isolate. Channel is defeated and letdown is restored.
3 MAL CVC003 C(ALL)
TS(CRS)
Letdown leak inside containment. Leak will be isolated.
4 MAL EPS005 D
C(ALL) 480V Bus 6A fault.
5 MAL ATS004 AlB PRS003 D
M(ALL)
Sequential loss of Main Feedwater Pumps leading to plant trip.
PORV fails open, cannot be isolated.
6 MAL SIS004A IB C(CRS) 31 and 32 Safety Injection Pumps will not auto-start. (Neither board operator credited since either one may start pump.)
7 MAL EPS001 C(ALL)
Station Blackout when Containment Phase A is reset.
(N)ormal, (R)eactivity, (I )nstrument.
(C)omponent, (M)ajor 2010 IP3 NRC Exam Scenario 1 Page 10f 13
Appendix 0 Scenario Outline Form ES-O-1 Facility:
IPEC Unit 3 Scenario No.:
2 Op-Test No.:.2 Examiners:
Operators:
Initial Conditions:
Plant is at 5%. 31 Charging Pump is OOS for maintenance.
Turnover:
Come up to 10% power, them warm up the MTG and place it in service. 31 Charging Pump is OOS and not expected back on this shift.
Event Malt.
Event Event No.
No.
Type*
Description 1
N/A R(ATC)
Perform power ascension.
N(CRS)
N(BOP) 2 3
4 5
6 7
XMT 040A MAL CCWOO 1 AlB MAL CCWOO 5A MAL MSSOO9 A
MAL RPSOO2 AlB MAL CNMOO1 12 I(ALL)
TS{CRS)
C(CRS)
C(BOP)
TS(CRS)
TS(CRS)
C(CRS)
C(ATC)
M(ALL)
C(BOP) 32 SG Pressure Channel Failure causing ADV to open.
31 CCW Pump trips and 32 CCW fails to auto-start.
CCW leak at 31 CCW Pump. Can be isolated Main Steam header break with no MSIVs auto-closing and 1 MSIV failing open.
Failure of the Reactor to trip both Auto and Manual. Team will emergency borate and manually insert control rods.
Containment Sump Pump Isolation Valves fail to close requiring manual action.
(N)ormal, (R)eactivity, (I)nstrument (C)omponent, (M)ajor 2010 IP3 NRC Exam Scenario 2 Page 10t 16
Appendix 0 Scenario Outline Form ES-O-1 Facility:
IPEC Unit 3 Scenario No.:
3 Op~Test No.: -.1 Examiners:
Operators:
Initial Conditions:
Plant is at 100%.31 AFW Pump is OOS due to high vibrations. PORV 455C is inoperable due to blowing fuses.
Turnover:
Maintain 100% power conditions. 31 AFW Pump is OOS due to high vibrations. PORV 455C is inoperable due to blowing fuses.
Event No.
Malf. No.
Event Type*
Event Description 1
XMT CVC049 I(ALL)
VCT level instrument fails low.
2 MAL~
CVC005C C(ALL)
TS(CRS) 33 Charging Pump Trip.
3 MAL SGN005C C(CRS)
C(BOP)
TS(CRS) 900 gpd SGTL on 33 SG.
4 R(ATC)
N(CRS)
N(BOP}
Tech Spec required shutdown.
5 MAL SGN005C MAL PRS003D SWI AIR002C M(ALL)
SGTL becomes SGTR. PORVand Instrument Air to VC Valve failures will lead to loss of pressure control.
6 MAL CFW001C C(CRS)
C(ATC) 33 AFW Pump fails to auto-start.
7 MAL SIS004A C(BOP)
'3,1.Safety Injection Pump fails to auto-start.
(N)ormal, (R)ea~tivity.
(I)nstrument, (C)omponent.
(M)ajor 2010 IP3 NRC Exam Scenario 3 Page 10f 16