BW160033, 2016 Braidwood Initial License Examination; Outline Submittal: Forms 401-2, 401-3; 301-1; 301-2; ES-D-1 Sheets
| ML16330A520 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 09/21/2016 |
| From: | Marchionda-Palmer M Exelon Generation Co |
| To: | Region 3 Administrator |
| Shared Package | |
| ML16306A156 | List: |
| References | |
| BW160033, ES-401 | |
| Download: ML16330A520 (22) | |
Text
ES-401 PWR Examination Outline FORM ES-401-2 Facility Name:Braidwood Date of Exam:9/05/2016 RO KIA Category Points SRO-Only Points Tier Group K
K K
K K
K A
A A
A G
Total A2 G*
Total 1
2 3
4 5
6 1
2 3
4
- 1. Emergency 1
3 3
3 3
3 3
18 3
3 6
Abnormal 2
1 2
1 NIA 1
2 N/A 2
9 2
2 4
Plant Evolutions Tier Totals 4
5 4
4 5
5 27 5
5 10 1
3 3
3 3
2 2
2 2
3 2
3 28 3
2 5
2.
2 1
1 1
1 1
1 1
0 1
1 1
10 1
1 1
3 Plant Systems Tier Totals 4
4 4
4 3
3 3
2 4
3 4
38 5
3 8
1 2
3 4
1 2
3 4
- 3. Generic Knowledge and Abilities 10 7
Categories 2
3 2
3 2
2 1
2 Note: 1.
Ensure that at least two topics from every applicable KJA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KJA category shall not be less than two). (One Tier 3 Radiation Control KJA is allowed if the KJA is replaced by a K/A from another Tier 3 Category).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted and justified; operationally important, site-specific systems that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KJA 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 K/As 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 KJA categories.
7.*
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the KJA 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/As.
- 8.
On the following pages, enter the KJA 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-only exams.
- 9.
For Tier 3, select topics from Section 2 of the KJA catalog, and enter the KJA numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
G*
Generic K/As
ES-401 2
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (RO)
Q#
E/APE #I Name I Safety Function K
K K
A A
G KIA Topic(s)
IR 1
2 3
1 2
48 000007 Reactor Trip - Stabilization - Recovery / 1 0
3 Reactor trip status panel 3.5 1
49 I000008 Pressurizer Vapor Space Accident I 3
- 02. Knowledge of conditions and limitations in the facility license.
3.6 1
38 50 I000009 Small Break LOCA I 3 3
SOM 3.2 1
2 51 000011 Large Break LOCA I 3 0
2 Pumps 2.6 1
52 I000015 RCP Malfunctions I 4 0
CCW lineup and flow paths to RCP oil coolers I000017 RCP Malfunctions (Loss of RC Flow) I 4 2
3.0 1
53 000022 Loss of Rx Coolant Makeup / 2 0
Actions contained in SOPs and EOPs for RCPs. loss of makeup, 2
loss of charging, and abnormal charging 3.5 1
54 000025 Loss of RHR System I 4 0
1 RCS/RHRS cooldown rate 3.6 1
000026 Loss of Component Cooling Water I 8 0
55 000027 Pressurizer Pressure Control System 0
Malfunction I 3 3
Latent heat of vaporization/condensation 2.6 1
56 000029 A TWS / 1
- 02. Knowledge of the bases in Technical Specifications for limiting 3.2 25 conditions for operations and safety limits.
1 57 000038 Steam Gen. Tube Rupture I 3 0
Causes and consequences of shrink and swell in S/Gs 5
2.8 1
58 000040 Steam Line Rupture - Excessive Heat Transfer 0
4 1
Consequences of PTS 4.1 1
WE 12 Uncontrolled Depressurization of all Steam Generators I 4 59 000054 (CE/E06) Loss of Main Feedwater I 4 0
4
60 000055 Station Blackout I 6 0
2 Actions contained in EOP for loss of offsite and onsite power 4.3 1
61 000056 Loss of Off-site Power I 6 8
1 SIG level meter scale and pressure gauge 3.7 1
000057 Loss of Wal AC Inst. Bus I 6 0
000058 Loss of DC Power I 6 0
62 io00062 Loss of Nuclear Svc Water I 4
- 02.
12 Knowledge of surveillance procedures.
3.7 1
I000065 Loss of Instrument Air I 8 0
63 r.t'J!E04 LOCA Outside Containment I 3 0
Normal, abnormal and emergency operating procedures 2
associated with LOCA Outside Containment 3.5 1
64 r.'VIE11 Loss of Emergency Coolant Recirc. / 4 0
Desired operating results during abnormal and emergency 3
situations 3.7 1
BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0
Secondary Heat Sink I 4 65 000077 Generator Voltage and Electric 0
Prid Disturbances I 6 7
Turbine I generator control 3.6 1
Totals:
3 3
3 3
3 3 Group Point Total:
18 ES-401, 39 of 50
ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (RO)
Q#
E/APE #I Name I Safety Function K
K K
A A
G KIA Topic(s)
IR 1
2 3
1 2
I000001 Continuous Rod Withdrawal / 1 0
40 f000003 Dropped Control Rod I 1 05 Control rod drive power supplies and logic circuits 2.5 1
IOOOOOS Inoperable/Stuck Control Rod / 1 0
I000024 Emergency Boration I 1 0
39 000028 Pressurizer Level Malfunction I 2 01 PZR level indicators and alarms 3.4 1
000032 Loss of Source Range NI / 7 0
000033 Loss of Intermediate Range NI I 7 0
I000036 Fuel Handling Accident I 8 0
I000037 Steam Generator Tube Leak I 3 0
I000051 Loss of Condenser Vacuum/ 4 0
I000059 Accidental Liquid RadWaste Rel. I 9 0
41 I000060 Accidental Gaseous Radwaste Rel. / 9
- 04. Knowledge of annunciator alarms, indications, or response 4.2 1
31 procedures.
f000061 ARM System Alarms I 7 0
42 f000067 Plant Fire On-site I 8 02 Steps called out in the site fire protection plan. FPS manual, and 2.5 1
fire zone manual I000068 Control Room Evac. I 8 0
43 I000069 Loss of CTMT Integrity I 5 01 Isolation valves, dampers, and electropneumatic devices 3.5 1
Wi E14 High Containment Pressure I 5 44 000074 lnad. Core Cooling / 4
- 02. Ability to determine operability and/or availability of safety related 3.6 37 equipment.
Wi E06 Degraded Core Cooling I 4 1
WIED? Saturated Core Cooling I 4 000076 High Reactor Coolant Activity I 9 0
W/E01 Rediagnosis I 3 0
W/E02 SI Termination / 3 W/E13 Steam Generator Over-pressure 14 0
W/E15 Containment Flooding / 5 0
45 W/E16 High Containment Radiation / 9 02 Adherence to appropriate procedures and operation within the 3.0 1
limitations in the facility's license and amendments
.,,; ll,,oll *-*-*-*
-* *- --*,_,.,_,,_ II *-*~*.i l!:l ;:ll):lf IC:U;:i 1 Clll IU,..,,,,..,...,..,,
46 W/E03 LOCA Cooldown - Depress. I 4 03 actions associated with the LOCA Cooldown and 3.5 1
--****J <> * *- -
- -***-* **-*--*!:! ~*"**-*1 ---*-* '
47 W/E09 Natural Circulation Operations I 4 02 emergency coolant, the decay heat removal systems, and 3.6 1
W/E10 Natural Circulation with Steam Voide in Vessel with/without RVLIS. I 4 W/E08 RCS Overcooling - PTS I 4 0
KIA Category Totals:
1 2
1 1
2 2 Group Point Total:
9 ES-401, 40 of 50
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 1 (RO)
K K K K K K A A A A G
KIA Topic(s)
IR Q# System #I Name 1 2 3 4 5 6 1 2 3 4 11 003 Reactor Coolant Pump 04 Containment isolation valves affecting RCP operation 2.8 1
004 Chemical and Volume Control 0
1 MOVs; Boric acid storage tank/boron injection tank 2.7; 2
12,3<
5 0
recirculation flow path 2.7 13, 3~ 005 Residual Heat Removal 0
01. Modes of operation; Ability to use plant computers to 3.2; 2
2 19 evaluate system or component status.
3.9 14,3* 006 Emergency Core Cooling 2
0 Bypassing/blocking ESF channels; Transfer of ECCS 4.1; 2
1 5
flowpaths prior to recirculation 3.9 15 007 Pressurizer Relief/Quench Tank 0
Maintaining quench tank water level within limits 2.9 1
1 0
0 SWS; Setpoints on instrument signal levels for normal 3.1; 16,3! 008 Component Cooling Water 1
1 operations, warnings. and trips that are applicable to the 3.2 2
CCWS 17 01 O Pressurizer Pressure Control 0
RCS 3.6 1
3 18 012 Reactor Protection 0
125V de system 3.4 1
2 19 013 Engineered Safety Features 0
Safety system logic and reliability 2.9 1
Actuation 2
0 01. Correlation of fan speed and flowpath changes with 3.1; 20,3E 022 Containment Cooling 2
23 containment pressure; Ability to perform specific system 4.3 2
and inteorated nlant nrocedures durino all modes of plant 025 Ice Condenser 21 026 Containment Spray 0
Failure of automatic recirculation transfer 4.2 1
2 22 039 Main and Reheat Steam 0
2 Isolation of the MRSS 3.1 1
23 059 Main Feedwater 0
Power level restrictions for operation of M FW pumps and 2.7 1
3 valves 0
- 02. Feed line voiding and water hammer; Ability to interpret 2.7; 24,31 061 Auxiliary/Emergency Feedwater 5
44 control room indications to verify the status and operation 4.2 2
of a svstem and understand how ooerator actions and 25 062 AC Electrical Distribution 0
ED/G 4.1 1
2 26,3f 063 DC Electrical Distribution 0
0 2.9; 2
1 1
Major DC loads; Major breakers and control power fuses 2.8 27 064 Emergency Diesel Generator 0
Control power 3.2 1
3 28 073 Process Radiation Monitoring 0
1.
Radioactive effluent releases 3.6 1
29 076 Service Water 0
ESF loads 3.7 1
I 30 078 Instrument Air 0
1 Air pressure 3.1 1
31 103 Containment 0
Phase A and B isolation 3.5 1
3 0
KIA Category Totals:
3 3 3 3 2 2 2 2 3 2 3 Group Point Total:
28 ES-401, Page 41 of 50
ES-401 5
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (RO)
Q# System #I Name K
K K
K K
K A
A A
A G
K/A Topic(s)
IR 1
2 3
4 5
6 1
2 3
4 5
I001 Control Rod Drive 0
Reactor power 4.1 1
6 1
loo2 Reactor Coolant 1
Relationship between reactor power and RCS differential 3.6 1
0 temperature
!011 Pressurizer Level Control 0
I014 Rod Position Indication 0
2 lo15 Nuclear Instrumentation 0
RPS 3.9 1
1 3
l<J16 Non-nuclear Instrumentation 0
Recorders 2.7 1
2
!017 In-core Temperature Monitor 0
~
!027 Containment Iodine Removal 0
lo2B Hydrogen Recombiner and Purge 0
!Control 029 Containment Purge 0
4 lo33 Spent Fuel Pool Cooling 0
Maintenance of spent fuel pool radiation 2.7 4
1 I034 Fuel Handling Equipment 0
6 lo35 Steam Generator 1
PRM system 3.1 1
1 7
Kl41 Steam Dump/Turbine Bypass Control 0
Main steam pressure 2.9 5
1 lo45 Main Turbine Generator 0
IOSS Condenser Air Removal 0
056 Condensate 0
I068 Liquid Radwaste 0
!071 Waste Gas Disposal 0
lo72 Area Radiation Monitoring 0
9 lo75 Circulating Water 0
Emergency/essential SWS pumps 2.6 3
1 8
lo79 Station Air 01. Ability to locate and operate components. including local 4.4 30 controls.
1 10 IOB6 Fire Protection 0
Fire, smoke, and heat detectors 4
2.6 1
KIA Category Totals:
1 1
1 1
1 1
1 0
1 1
1 Group Point Total:
10 ES-401, Page 42 of 50
ES-401 2
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 1 (SRO)
Q#
E/APE # I Name I Safety Function K
K K
A A
G KIA Topic(s)
IR 1
2 3
1 2
000007 Reactor Trip - Stabilization - Recovery / 1 0
P00008 Pressurizer Vapor Space Accident I 3 0
P00009 Small Break LOCA I 3 0
000011 Large Break LOCA I 3 0
P00015 RCP Malfunctions I 4 0
P00017 RCP Ma~unctions (Loss of RC Flow) I 4 pooo22 Loss of Rx Coolant Makeup I 2 0
P00025 Loss of RHR System I 4 0
- 02. Ability to interpret control room indications to verify the status 88 000026 Loss of Component Cooling Water I 8 44 and operation of a system, and understand how operator actions 4.4 1
ond ni<octives offoct olant and svstem conditions.
P00027 Pressurizer Pressure Control System 0
Ma~unct ion I 3 P00029 A TWS I 1 0
000038 Steam Gen. Tube Rupture I 3 0
000040 Steam Line Rupture - Excessive Heat Transfer 4
1 89 WE12 Uncontrolled Depressurization of all Steam
- 04. Knowledge of local auxiliary operator tasks during an emergency 4.0 Generators I 4 35 and the resultant operational effects.
000054 (CE/E06) Loss of Main Feedwater I 4 0
000055 Station Blackout I 6 0
000056 Loss of Off-s~e Power I 6 0
90 000057 Loss of Wal AC Inst. Bus / 6 1
The plant automatic actions that will occur on the loss of a vital 4.3 1
9 ac electrical instrument bus 91 000058 Loss of DC Power I 6 0
DC loads lost; impact on to operate and monitor plant systems 3.9 1
3 000062 Loss of Nuclear Svc Water I 4 0
92 000065 Loss of Instrument Air I 8
- 04. Knowledge of abnormal condition procedures.
4.2 1
11 W/E04 LOCA Outside Containment I 3 0
W/E1 1 Loss of Emergency Coolant Recirc. / 4 0
93 BW/E04; W/E05 Inadequate Heat Transfer - Loss of 0
Adherence to appropriate procedures and operation within the 4.3 1
Secondary Heat Sink I 4 2
limitations in the facility's license and amendments 000077 Generator Voltage and Electric 0
Grid Disturbances I 6 KIA Category Totals:
0 0
0 0
3 3 Group Point Total:
6 ES-401, 39 of 50
ES-401 3
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions - Tier 1/Group 2 (SRO)
Q#
E/APE #I Name I Safety Function K
K K
A A
G KIA Topic(s)
IR 1
2 3
1 2
looooo1 Continuous Rod Withdrawal/ 1 0
io00003 Dropped Control Rod / 1 0
84 I000005 Inoperable/Stuck Control Rod / 1
- 04. Knowledge of how abnormal operating procedures are used in 4.5 1
08 conjunction..;th EOPs.
000024 Emergency Boration I 1 0
000028 Pressurizer Level Malfunction I 2 0
000032 Loss of Source Range NI I 7 0
000033 Loss of Intermediate Range NI I 7 0
000036 Fuel Handling Accident I 8 0
85 000037 Steam Generator Tube Leak I 3
- 04. Ability to verify that the alarms are consistent with the plant 4.2 1
AR conditions.
000051 Loss of Condenser Vacuum I 4 0
000059 Accidental Liquid RadWaste Rel. I 9 0
000060 Accidental Gaseous Radwaste Rel. I 9 0
000061 ARM System Alarms 17 0
000067 Plant Fire On-site I 8 0
86 000068 Control Room Evac. I 8 02 Local boric acid flow 4.2 1
000069 Loss of CTMT Integrity I 5 0
W/E14 High Containment Pressure I 5 000074 lnad. Core Cooling / 4 W/E06 Degraded Core Cooling I 4 1
87 W/E07 Saturated Core Cooling I 4 02 Adherence to appropriate procedures and operation within the 3.9 limitations in the facility's license and amendments 000076 High Reactor Coolant Activity I 9 0
W/E01 Rediagnosis I 3 0
W/E02 SI Termination I 3 W/E13 Steam Generator Over-pressure I 4 0
W/E15 Containment Flooding / 5 0
W/E16 High Containment Radiation / 9 0
IW/E03 LOCA Cooldown - Depress. I 4 0
W/E09 Natural Circulation Operations / 4 W/E10 Natural Circulation with Steam Voide in 0
Vessel with/without RVLIS. I 4 W/E08 RCS Overcooling - PTS I 4 0
ory Totals:
0 0
0 0
2 2 Group Point Total:
4 ES-401, 40 of 50
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401 -2 Plant Systems - Tier 2/Group 1 (SRO)
Q# System #I Name K K K K K K A A A A G
KIA Topic(s)
IR 1 2 3
4 5
6 1 2 3 4
0 Problems associated with RCP motors, including faulty 79 003 Reactor Coolant Pump 3
motors and current, and winding and bearing temperature 3.1 1
oroblems 004 Chemical and Volume Control 0
005 Residual Heat Removal 0
006 Emergency Core Cooling 0
007 Pressurizer Relief/Quench Tank 0
008 Component Cooling Water 0
80 010 Pressurizer Pressure Control 0
Spray valve failures 3.9 1
2 012 Reactor Protection 0
013 Engineered Safety Features 0
Actuation 022 Containment Cooling 0
025 Ice Condenser 0
026 Containment Spray 0
81 039 Main and Reheat Steam 0
Indications and alarms for main steam and area radiation 3.7 1
3 monitors (during SGTR) 059 Main Feedwater 0
061 Auxiliary/Emergency Feedwater 0
062 AC Electrical Distribution 0
063 DC Electrical Distribution 0
83 064 Emergency Diesel Generator
- 01. Ability to explain and apply system limits and precautions.
4.0 1
32
- 04. Knowledge of the parameters and logic used to assess the 82 073 Process Radiation Monitoring 21 status of safety functions, such as reactivity control, core 4.6 1
coolina and heat removal reactor coolant svstem intearitv.
076 Service Water 0
078 Instrument Air 0
103 Containment 0
0 KIA Category Totals:
0 0
0 0
0 0
0 3
0 0
2 Group Point Total:
5 ES-401, Page 41 of 50
ES-401 5
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant Systems - Tier 2/Group 2 (SRO)
Q# System # I Name K
K K
K K
K A
A A
A G
KIA Topic(s)
IR 1
2 3
4 5
6 1
2 3
4 001 Control Rod Drive 0
002 Reactor Coolant 0
011 Pressurizer Level Control 0
014 Rod Position Indication 0
015 Nuclear Instrumentation 0
!016 Non-nuclear Instrumentation 0
017 In-core Temperature Monitor 0
027 Containment Iodine Removal 0
!028 Hydrogen Recombiner and Purge 0
!Control k>29 Containment Purge 0
!033 Spent Fuel Pool Cooling 0
77 k>34 Fuel Handling Equipment 0
Containment ventilation 2.9 1
1 76 l<J35 Steam Generator
- 02. Knowledge of limiting conditions for operations and safety 4.7 1
22 limits.
78 k>41 Steam Dump/Turbine Bypass Control 0
Steam valve stuck open 3.9 1
2
!045 Main Turbine Generator 0
I055 Condenser Air Removal 0
k>56 Condensate 0
k>S8 Liquid Radwaste 0
I071 Waste Gas Disposal 0
!072 Area Radiation Monitoring 0
k>75 Circulating Water 0
!079 Station Air 0
k>as Fire Protection 0
KIA Category Totals:
0 0
1 0
0 0
0 1
0 0
1 Group Point Total:
3 ES-401, Page 42 of 50
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401 -3 Facility Name:Braidwood Date of Exam:9/05/2016 Category KIA#
Topic RO SRO-Only IR IR Q#
66 2.1. 25 Ability to interpret reference materials, such as graphs, curves, tables, etc.
3.9 1
4.2 67 2.1. 32 Ability to explain and apply system limits and precautions.
3.8 1
4.0 94
- 1.
2.1. 34 Knowledge of primary and secondary plant chemistry limits.
2.7 3.5 1
Conduct of 95 Operations 2.1. 36 Knowledge of procedures and limitations involved in core alterations.
3.0 4.1 1
2.1.
2.1.
Subtotal 2
2 68 2.2. 36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of
- 3. 1 1
4.2 limiting conditions for operations.
69 2.2. 07 Knowledge of the process for conducting special or infrequent tests.
2.9 1
3.6 70 2.2. 44 Ability to interpret control room indications to verify the status and operation of a system, and understand 4.2 1
4.4
- 2.
how operator actions and directives affect plant and system conditions.
96 Equipment 2.2. 21 Knowledge of pre-and post-maintenance operability requirements.
2.9 4.1 1
Control 97 2.2. 39 Knowledge of less than or equal to one hour Technical Specification action statements for systems.
3.9 4.5 1
2.2.
Subtotal 3
2 71 2.3. 14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency 3.4 1
3.8 conditions or activities.
72 2.3. 05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey 2.9 1
2.9 instruments, personnel monitoring equipment, etc.
98 2.3. 06 Ability to approve release permits.
3.8 1
- 3.
Radiation 2.3.
Control 2.3.
2.3.
Subtotal 2
1 73 2.4. 13 Knowledge of crew roles and responsibilities during EOP usage.
4.0 1
4.6 74
- 2.4. 49 Ability to perform without reference to procedures those actions that require immediate operation of 4.6 1
4.4 system components and controls.
75
- 4.
2.4. 05 Knowledge of the organization of the operating procedures network for normal, abnormal, and 3.7 1
4.3 emergency evolutions.
Emergency 99 Procedures I 2.4. 04 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions f 4.5 4.7 1
Plan 100 2.4. 28 Knowledge of procedures relating to a security event.
3.2 4.1 1
2.4.
Subtotal 3
2 Tier 3 Point Total 10 7
ES-401, Page 43 of 50
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Braidwood Date of Examination: 09/06/2016 Examination Level: RO 0 SRO [8]
Operating Test Number: 15-1 NRC Administrative Topic Type Describe activity to be performed (See Note)
Code*
Conduct of Operations S,R,D Determine venting time for RX Vessel Head SIM-108 KIA 2.1.25 SRO 4.2 Conduct of Operations S,R,N Determine if Reactor Startup should continue. SIM 114 KIA 2.1.37 SRO 4.6 Review a surveillance to determine battery operability S-Equipment Control S,R,D 204 KIA 2.2.40 SRO 4.7 Radiation Control S,R,D Review a gas decay tank release S-301 KIA 2.3.6 SRO 3.8 Emergency Plan S,R,M Classify an Event, cold EAL S-413 KIA 2.4.41 SRO 4.6 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(.:::_ 3 for ROs;.:::. 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (.:::_ 1; randomly selected)
ES 301, Page 22 of 27
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Braidwood Date of Examination: 09/06/2016 Exam Level: RO D SRO-I [8J SRO-U D Operating Test No.: 151 NRC Control Room Systems: *8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System I JPM Title Type Code*
Safety Function
- a. Perform Emergency Boration SIM-113 A,S,M,L 1
KIA 0024AA 1.17 RO 3.9 SRO 3.9
- b. Align RHR for cold leg injection Sim-301 D,S,L 2
- c. Raise SI Accumulator Level Sim-203 D,EN,S 3
KIA 006A1.13A1.17 RO 4.1 SRO 4.2
- d. Swap CC pumps Sim-800 A,D,S 8
- f. Start a reactor coolant pump SIM-409P N,S,L 4P KIA 003A4.02 RO 2.9 SRO 2.9
- g. Parallel the 1 B DG to bus 142 with a failure of the voltage controller high SIM-612 A,M,S 6
- h.
In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i.
- j.
Respond to a fire detection alarm/28 DG, IP-801 A,D 8
KIA 086A2.04 RO 3.3 SR03.9
- k. Locally control of aux feed flow IP-411 A,E,N,R 4S KIA APE068AA 1.26 RO 3.6 SRO 3.8 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five 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-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank
~91 ~8 I ~ 4 (E)mergency or abnormal in-plant
~ 1 I
~1 I ~ 1 (EN)gineered safety feature
~ 1 I
~ 1 I
~ 1 (control room system)
(L)ow-Power I Shutdown
~ 1 I
~1 I ~1 (N)ew or (M)odified from bank including 1 (A)
~21 ~2 I ~1 (P)revious 2 exams
~ 3 I
~3 I ~ 2 (randomly selected)
(R)CA
~ 1 I
~1 I ~1 (S)imulator ES-301, Page 23 of 27
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Braidwood Date of Examination: 09/06/2016 Examination Level: RO iZI SRO 0 Operating Test Number: 15-1 NRC Administrative Topic Type Describe activity to be performed (See Note)
Code*
Perform 1 BwOS RF-1. R-113 Conduct of Operations S,R,N KIA2.1.19R03.9 Conduct of Operations S,D Perform QPTR without the plant process computer R-102 KIA 2.1.7 RO 4.4 Identify leak isolation point from station mechanical Equipment Control S,R,D drawings. R-204 KIA 2.2.41 RO 3.5 Radiation Control Perform NARS transmittal for a UE R-401 Emergency Plan S,D KIA 2.4.43 RO 3.2 NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(~ 3 for ROs; ~ 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (~ 1; randomly selected)
ES 301, Page 22 of 27
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Braidwood Date of Examination: 09/06/2016 Exam Level: RO [8J SRO-I D SRO-U D Operating Test No.: 151 NRC Control Room Systems: *8 for RO; 7 for SRO-I; 2 or 3 for SRO-U System I JPM Title Type Code*
Safety Function
- a. Perform Emergency Boration SIM-113 A,S,M,L 1
- b. Align RHR for cold leg injection Sim-301 D,S,L 2
- c. Raise SI Accumulator Level Sim-203 D,EN,S 3
KIA 006A1.13A1.17 RO 4.1 SRO 4.2
- d. Swap CC pumps Sim-800 A,D,S 8
- f. Start a reactor coolant pump SIM-409P N,S,L 4P KIA 003A4.02 RO 2.9 SRO 2.9
- g. Parallel the 1 B DG to bus 142 with a failure of the voltage regulator high SIM-612 A,M,S 6
- h. Respond to 1 PR03J high radiation alarm SIM-706 D,S 7
KIA 073A4.02 RO 3.7 SRO 3.7 In-Plant Systems* (3 for RO); (3 for SRO-I); (3 or 2 for SRO-U)
- i.
- j.
Respond to a fire detection alarm/28 DG, IP-801 A,D 8
KIA 086A2.04 RO 3.3 SR03.9
- k. Locally control of aux feed flow IP-411 A,E,N,R 4S KIA APE068AA 1.26 RO 3.6 SRO 3.8 All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions; all five 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-I I SRO-U (A)lternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank
_:::9/
_:::8 I _::: 4 (E)mergency or abnormal in-plant
~ 1 I
~1 I ~ 1 (EN)gineered safety feature
~ 1 I
~ 1 I
~ 1 (control room system)
(L)ow-Power I Shutdown
~ 1 I
> 1 I ~1 (N)ew or (M)odified from bank including 1 (A)
~2/ ~2 I ~1 (P)revious 2 exams
_::: 3 I
_:::3 I.::: 2 (randomly selected)
(R)CA
~ 1 I
~1 I ~1 (S)imulator ES-301, Page 23 of 27
Simulation Facility Braidwood Scenario Operating Test No.: 15-1 NRC No.:
NRC 1 Examiners:
Applicant:
SRO ATC BOP Initial Conditions:
IC-21 Turnover:
Unit 1 is at 100% power, steady state, equilibrium xenon, BOL. Following completion of turnover, the Shift Manager requests that the BOP swap the OA & OB WS pumps per BwOP WS-1 and WS-3 in preparation for surveillance testing on Unit 2.
Event Malf. No.
Event Event No.
Type*
Description Pre load Trgset 1 "ZL01 HSMP025(3)== O" Trg 1 "IMF ED04A" Offsite power is lost on the main qenerator trip 1
None N-BOP, Swap WS pumps us 2
IMF NI08H 500 10 I-ATC, US PR NI N-44 fails high TS-US 3
IMF RX03C 4.8 I-BOP, US 1 B SG steam flow, 1 FT-522, fails high 4
IMF CV05 600 5 C-ATC, Letdown line pressure controller 1 PK-131 output us fails low 5
IOR ZDl1W001PA TRIP C-BOP, 1A Containment Chilled Water Pump trip us 6
TH03B 5 60 TS-US 1 B SG Tube Leak 7
None R-ATC, 1 BwOA SEC-8 Fast Ramp us 8
IMF TH03B 550 M-ALL 1B SGTR 9
Pre load C-ALL Offsite power is lost on the main generator trip (N)ormal, (R)eactivity (l)nstrument, (C)omponent, (M)ajor Transient
SCENARIO OVERVIEW Unit 1 is at 100% power, steady state, equilibrium xenon, BOL. Following completion of turnover, the Shift Manager requests that the BOP swap the OA & OB WS pumps per BwOP WS-1 and WS-3 in preparation for surveillance testing on Unit 2.
After completing shift turnover and relief, the BOP will swap the OA & OB WS pumps in accordance with BwOP WS-1 and WS-3. The OB WS pump will be started and the OA WS pump will be placed in standby.
After swapping WS pumps, a failure of power range N-44 lower detector will occur. The crew will take actions per hard card 1BwPR 1-10-RD & 1BwOA INST-1 "NUCLEAR INSTRUMENTATION MALFUNCTION" including defeating the channel functions. Technical Specifications 3.3.1 Conditions A, D & E apply.
After the Power Range failure is addressed, a failure of 1 FT-522, 1 B SG steam flow channel, will occur. The crew will take actions per hard card 1BwPR1-15-SG & 1 BwOA INST-2 "OPERATION WITH FAILED INSTRUMENT CHANNEL" and perform required actions. The crew will swap to an operable steam flow channel.
After completion of steam flow failure actions, letdown pressure controller, 1 PK-131, output will fail low. The letdown pressure control valve, 1CV131, will close and letdown pressure will rise lifting the letdown line relief valve. The ATC will take manual control of letdown pressure controller per hard card 1BwPR1-9-LD and restore letdown pressure. The crew may isolate letdown due to the lifting letdown relief valve. If letdown is isolated, it will be restored per BwOP CV-17.
After the 1 PK-131 controller failure is addressed, the 1A Containment Chilled Water pump will trip resulting in a trip of the 1A Containment Chiller. The crew will start the standby 1 B Chilled Water pump and Chiller per BwOP VP-1 after receiving a report that the 1A Chilled Water pump motor appears damaged. Containment pressure and temperature will slowly rise as a result of the loss of Containment Cooling. The US should reference Tech Spec 3.6.4 and 3.6.5 and place a priority to start the standby Cnmt Chiller.
After the Containment Chilled Water malfunction has been addressed, a 1 B Steam Generator Tube Leak will be initiated. The crew will implement 1 BwOA SEC-8 "STEAM GENERATOR TUBE LEAK." The crew should determine from the estimated leak size that a fast ramp to remove the unit from power is required and make preparations to ramp the unit off line using 1 BwGP 100-4, POWER DESCENSION or 1 BwOA PWR-1 "POWER REDUCTION." Technical Specifications 3.4.13 Condition B applies.
After the SGTL has been addressed and the crew has ramped the unit sufficiently for a reactivity manipulation, the 1 B SGTL will degrade to a 550 gpm Steam Generator Tube Rupture.
The crew will implement 1 BwEP-0 "REACTOR TRIP OR SAFETY INJECTION." The crew will transition to 1 BwEP-3 "SGTR."
The scenario is complete when the crew has terminated high head injection and established normal charging flow in 1 BwEP-3.
Critical Tasks
- 1. Identify the 1 B SG as the ruptured SG and isolate prior to a transition to 1 BwCA-3.1 is required.
(Westinghouse - CT-18) (KIA number - EPE038EA1.32 importance 4.6/4.7)
(Westinghouse - CT-20) (KIA number - EPE038EA 1.04 importance 4.3/4.1)
Simulation Facility Braidwood Scenario Operating Test No.: 15-1 NRC No.:
NRC 2 Examiners:
Applicant:
SRO RO BOP Initial Conditions:
IC-31 Turnover: Unit 1 is operating at 90% power, steady state, equilibrium xenon, BOL. Online risk is green.
Following completion of turnover, coordinate with Unit 2 and lower total station reactive load by a total of 2 KV in accordance with BwOP MP-23. The 1 B HD pump was taken out of service 3 days ago and will NOT be returned to service for an additional 5 days.
Event Malf. No.
Event Event No.
Type*
Description Pre load IMF MS01A 100 1A MSIV failed open IMF MS01 B 100 1 B MSIV failed open IMF MS01C 100 1 C MSIV failed open IMF MS01 D 100 1 D MSIV failed open IMF CS01A 1A CS pump fails to start IRF RP63 OUT K643 1 B CS slave relay fails to actuate 1
None N-BOP, US Lower reactive load 1 KV.
2 IMF RX10A 0 15 I-ATC, US Turbine impulse pressure channel, 1 PT-505, TS-US fails low 3
IMF CV10 0 30 I-ATC, US 1CV121 controller failure low 4
IMF RX06K 0 15 I-BOP, US 1 C SG NR level transmitter, 1 L T-539, fails low TS-US 5
IMF FW35A C-BOP, US 1A Heater Drain Pump trip 6
None R-ATC, US Turbine runback (only 1 HD pump running) 7 IMF MS07C 4 M-ALL Uncontrolled depressurization of all SGs 8
Pre load C-ALL Failure of both CS trains to start automatically
- (N)ormal, (R)eactivity (l)nstrument, (C)omponent, (M)ajor Transient
SCENARIO OVERVIEW Unit 1 is operating at 90% power, steady state, equilibrium xenon, BOL. Online risk is green. Following completion of turnover, coordinate with Unit 2 and lower total station reactive load by a total of 2 KV in accordance with BwOP MP-23. The 1 B HD pump was taken out of service 3 days ago and will NOT be returned to service for an additional 5 days.
After completing shift turnover and relief, the BOP will lower Unit 1 reactive load 1 KV in accordance with BwOP MP-23.
After lowering reactive load, First Stage Turbine Impulse Pressure channel, 1 PT-505, will fail low. After the crew diagnoses the failure of 1 PT-505, the A TC will take manual control of rods after verifying turbine load stable. Initial ATC response is per hard card 1BwPR 1-10-RD, followed by entry into 1BwOA INST-2 "OPERATION WITH A FAILED INSTRUMENT CHANNEL, ATTACHMENT D." Tech Spec 3.3.1 Conditions A and P will be entered. The ATC will return rod control to automatic after verifying Tave and Tref are stable and within 1°F.
After the 1 PT-505 failure is addressed, 1CV121, Charging Pump Flow Control Valve Controller 1 FK-121, will fail to 0% demand. The 1CV121 valve will fully close and pressurizer level will drop. The crew will take actions to stabilize the plant by taking manual control of the 1 FK-121 controller and restore charging flow.
Afterthe 1FK-121 failure has been addressed, 1C SG NR level transmitter, 1LT-539, will fail low.
1 FW530, Feedwater Regulating Valve, will open and 1 C SG level will rise. The BOP will take manual control of 1C SG level per hard card 1BwPR 1-15-SG and stabilize 1C SG level. 1BwOA INST-2 "OPERATION WITH A FAILED INSTRUMENT CHANNEL," ATTACHMENT E, will be entered. The BOP will restore 1 C SG level control to automatic after 1 C SG level is restored to normal and an operable 1 C SG NR level controlling channel is selected. Tech Specs 3.3.1 Conditions A and E and 3.3.2 Conditions A and Dare applicable.
After the 1LT-539 failure has been addressed, 1A Heater Drain Pump will trip. 1 BwOA SEC-1 "SECONDARY PUMP TRIP," ATTACHMENT C, will be entered. The BOP will initiate a turbine load reduction to 780 MW at 20 MW/minute. The ATC will borate the RCS as necessary during the load reduction.
As the crew is reducing load to address the 1A HD pump trip, a large steam break will occur on the 1C MS line. When containment pressure reaches 20 psig, Phase B actuates but the CS pumps do not start.
The crew should manually realign train B CS valves which will start the 1 B CS pump. Operators should transition to 1 BwEP-2 "FAULTED STEAM GENERATOR ISOLATION" and recognize that the MSIVs have failed to close and that an uncontrolled depressurization of all SGs is in progress. The crew will transition to 1 BwCA-2.1 "UNCONTROLLED DEPRESSURIZA TION OF ALL STEAM GENERA TORS" where they will throttle AF flow to the SGs. Entry into 1 BwFR-H.1 "RESPONSE TO LOSS OF SECONDARY HEAT SINK" will be required when the crew throttles AF flow to 45 gpm per SG; however, a note in 1 BwFR-H.1 directs that the procedure should NOT be performed. The crew will continue in 1 BwCA-2.1.
Completion criteria is stopping the RH pumps in 1 BwCA-2.1.
Critical Tasks
- 1. Manually actuate one train of containment spray prior to transition out of 1 BwEP-0.
(Westinghouse - CT-3) (KIA number - 013000A4.01 importance - 4.5/4.8)
(Westinghouse - CT-33) (KIA number - EPEE12EA 1.3 importance - 3.4/3.9 151 NRC 2 Page 2
Simulation Facility Braidwood Scenario Operating Test No.: 151 NRC No.:
NRC3 Examiners:
Applicant:
SRO ATC BOP Initial Conditions:
IC-21 Turnover: Unit 1 is operating at 100% power, steady state, equilibrium xenon, BOL. 1 B CW pump is OOS for motor inspection. Automatic Rod Control is failed (rod speed programmer failure).
Rods are in MANUAL. IMO troubleshooting will commence in 3-4 hours. Following completion of turnover, the crew is to perform 1 BwOS EH-M1 "UNIT 1 PUMP OPERABILITY SURVEILLANCE," using the preferred method of depressing and holding the MCB pushbutton. An Equipment Operator has been briefed and is standing by at the Unit 1 EH skid.
Event Malf. No.
Event Event No.
Type Description Pre load IMF RX17 0 Rod speed programmer failure IOR ZDl1CW01PB PTL 1 B CW pump OOS IMF FW44 1A AF pump fails to start IMF FW43 1 B AF pump fails to start IMF Sl12B 1 B SI pump fails to automatically start IRF RP30 OUT Phase A slave failure IRF RP56 OUT Phase A slave failure trgset 1 "ZL052BRKA(2)== 1" IMF FW19C 3.5 (1 0) 1 C feed line break inside Cnmt when reactor trips IMF CV01 B (1 0) 1 B CV pump trips when reactor trips 1
None N-BOP, Perform 1 BwOS EH-M1 us 2
IMF RX04E 0 120 I-BOP, Feed Flow channel, 1 FT530, fails low us 3
IMF CV29A 100 C-ATC, 1A CV pump shaft shear us TS-US 4
IMF SW01A C-BOP, 1A SX Pump Trip us TS-US 5
IMF RX15 2195 I-ATC, Master Pressurizer Pressure Controller output us fails to 2195 psig (0% demand) 6 IMF FW36 500 scfm to 120 R-ATC, Condenser vacuum leak requiring load reduction scfm over 600 seconds us 7
IMF CV03 C-ALL Boric acid transfer pump trip 8
IMF RP09A M-ALL Inadvertent FWI Feed line break inside containment 9
Pre load C-ALL 1 B CV pump trip when reactor trips 10 Pre load M-ALL Loss of heat sink (1A/1 B AF pumps tripped)
- (N)ormal, (R)eactivity (l)nstrument, (C)omponent, (M)ajor Transient
SCENARIO OVERVIEW Unit 1 is operating at 100% power, steady state, equilibrium xenon, SOL. 1 B CW pump is OOS for motor inspection. Automatic Rod Control is failed (rod speed programmer failure). Rods are in MANUAL. IMO troubleshooting will commence in 3-4 hours. Following completion of turnover, the crew is to perform 1 BwOS EH-M1 "UNIT 1 PUMP OPERABILITY SURVEILLANCE," using the preferred method of depressing and holding the MCB pushbutton. An Equipment Operator has been briefed and is standing by at the Unit 1 EH skid.
After completing shift turnover and relief, the BOP will perform 1 BwOS EH-M1 "UNIT 1 PUMP OPERABILITY SURVEILLANCE."
After completing the surveillance, feed flow channel, 1 FT-530, will fail low resulting in rising feed flow and opening of 1 FW530, 1 C SIG Feed Reg Valve, to match feed flow with steam flow. The BOP will diagnose the failure and take manual control of 1 FW530 to restore 1 C SG level. The US will enter 1 BwOA INST-2 "OPERATION WITH A FAILED INSTRUMENT CHANNEL -ATTACHMENT G." The crew will take actions per hard card 1BwPR1-15-SG & 1 BwOA INST-2 and perform appropriate actions. The crew will swap to an operable feed flow channel.
After addressing the failed feed flow channel, the 1A CV pump shaft shear will occur. The crew will implement 1BwOA PRl-15 "LOSS OF NORMAL CHARGING." The crew will start the 18 CV pump to restore normal charging. Technical Specifications 3.5.2 Condition A and TRM 3.1.d Condition A apply.
After the charging pump failure is addressed, the 1A Essential Service Water (SX) pump will trip due to an overcurrent condition. The crew will take actions per hard card 1 BwPR 1-2-A 1 to start the 1 B SX pump. The crew MAY enter 1BwOA PRl-8, ESSENTIAL SERVICE WATER MALFUNCTION. Tech Specs 3.7.8 Condition A applies. The 1A SX pump will remain unavailable for the rest of the scenario.
After the 1A SX pump trip is addressed, the master pressurizer pressure controller output will fail to 2195 psig (0% demand) in automatic. The RO will identify the failure and take manual control to restore pressurizer pressure.
After the master pressurizer pressure controller failure has been addressed, an air leak will develop in the main condenser. The crew will enter 1 BwOA SEC-3 "LOSS OF CONDENSER VACUUM," dispatch operators to locate the leak and contact Engineering to evaluate the leakage. With the 1 B CW pump OOS; the crew will be required to commence a ramp down of Unit 1 to stabilize the loss of vacuum.
During the ramp, the boric acid transfer pump will trip during the first boration. The common boric acid pump will not be available and the ATC will have to continue the ramp with rods only.
After a sufficient reactivity manipulation has occurred, an inadvertent FWI occurs. When the crew manually trips the reactor, a feed line break on the 1 C SG inside containment will occur. The 1 B CV pump trips when the reactor trips. The crew will enter 1 BwEP-0 to stabilize the plant and SI will actuate. The 1 B SI pump fails to auto start and must be manually started. The 1A and 1 B AF pumps will not start resulting in a transition to 1BwFR-H.1 "RESPONSE TO LOSS OF SECONDARY HEAT SINK" at step 15of1BwEP-O. Bleed and feed will be required due to NO CV pumps running.
Completion criteria is establishing bleed and feed during the loss of heat sink.
Critical Tasks
- 1. Close containment isolation valves such that at least one valve is closed on each Phase A penetration before the end of the scenario. (Westinghouse CT-11) (KIA number APE069AA 1.01 importance 3.5/3.7)
- 2. Initiate RCS bleed and feed so that the RCS depressurizes sufficiently for SI pump injection to occur.
(Westinghouse CT-46) (KIA number EPEE05EA 1.1 importance 4.1/4.0) 151 NRC 3 Page 2
Simulation Facility Braidwood Scenario Operating Test No.: 151 NRC No.:
NRC4 Examiners:
Applicant:
SRO ATC BOP Initial Conditions:
IC-16 Turnover:
Unit 1 is at 53% power, steady state, equilibrium xenon, MOL. Online risk is green. Following completion of turnover, the crew will perform 1 BwOSR 3.7.4.1 "MAIN STEAM SYSTEM ISOLATION 1MS018NB/C/D VALVE TRAVEL AND INDICATION 18 MONTH SURVEILLANCE." Once the surveillance is complete, the SM has directed ramping Unit 1 to 938 MW at 3 MW/min.
II Event Malf. No.
Event Event No.
Type*
Description IMF CV32B 1 B CV pump auto start failure TRGSET 1 1A CV pump trip on 1A SI pump start "ZL01 SI01 PA(3) = = 1" IMF CV01A (1 0) 1A CV pump trips on 1A SI pump start TRGSET 2 "ZA01 Pl524A < 0.46" IMF TH03B (2 10) 600 60 1 B SGTR when 1 B SG pressure < 600 psig 1
TRGSET 3 N-BOP, US 1 BwOSR 3. 7.4.1, 1 MS018NB/C/D Surv with "ZA01 PKMS042A > 0.90" TS-US 1 MS018B valve failure.
IMF PB2411 (3 0) ON IMF PB2412 (3 0) ON IOR ZL01 MS018B2 (3 0) ON IOR ZL01 MS018B1 (3 0) ON IOR ZA01ZIMS010 (3 0) 90 TRGSET 2 "ZDl1MS018B== O" TRG 2 "DOR ZA01ZIMS010" 2
None R-ATC, US Raise power from 53% to 938 MW 3
IMF TH10B 100 15 C-ATC, US 1RY455C spray valve fails open in auto 4
IMF RX13A 0 C-ATC, US PZR level channel, 1 L T-459, fails low TS-US 5
IMF RX29B 100 30 I-BOP, US 1 FW520 controller fails high in auto 6
IMF FW17 10 30 I-BOP, US HD Tank level controller failure in auto 7
IOR ZDl1 MS001 B CLS M-ALL 1 B MSIV fails closed causing 1 B SG safety IMF MS03B 100 valves to stick open IMF MS03F 100 IMF MS03J 100 8
Preload C-ALL 1A CV pump trips/1 B CV pump fails to auto start 9
Preload M-ALL 1 B SGTR (600 gpm) (faulted and ruptured SG)
- (N)ormal, (R)eactivity (l)nstrument, (C)omponent, (M)ajor Transient
SCENARIO OVERVIEW Unit 1 is at 53% power, steady state, equilibrium xenon, MOL. Online risk is green. Following completion of turnover, the crew will perform 1BwOSR 3.7.4.1 "MAIN STEAM SYSTEM ISOLATION 1MS018A/B/C/D VALVE TRAVEL AND INDICATION 18 MONTH SURVEILLANCE." Once the surveillance is complete, the SM has directed ramping Unit 1 to 938 MW at 3 MW/min.
After completing shift turnover and relief, the BOP performs 1 BwOSR 3. 7.4.1. The 1 B SG PORV, 1 MS018B, will fail the test due to a broken hydraulic line when the valve is stroked open. The Unit Supervisor will enter Tech Spec 3.7.4 Condition A and Tech Spec 3.6.3 Condition C. 1MS019B will remain closed to comply with TS 3.6.3 Condition C. 1 MS018B will remain unavailable for the remainder of the scenario. The crew should inform the SM following the 1 MS018B failure.
After the 1MS0188 failure has been addressed, the crew will commence a ramp to 938 MW at 3 MW/min.
After a measurable change in power, 1PK-455C, Pressurizer Spray Valve controller, will fail to 100%
demand position. 1RY455C will fail full open and pressurizer pressure will drop. The ATC will take manual control of 1 PK-455C per hard card 1 BwPR 1-12-RY and lower demand to close the pressurizer spray valve.
Tech Spec 3.4.1 Condition A will apply if pressurizer pressure drops below 2209 psig.
After the 1 PK-455C failure has been addressed, the controlling PZR level channel, 1 L T-459, will fail low.
The crew will implement 1 BwOA INST-2 and take actions to restore PZR level control and stabilize plant conditions. Technical Specification 3.3.1 Conditions A and K apply (Tech Spec 3.3.3 and 3.3.4 for Accident Monitoring and Remote Shutdown Panel minimum channel requirements are still met).
After the failed PZR level channel has been addressed, 1 FK-520, Feed Reg Valve 1 FW520 Controller, will fail to 100% demand. 1 FW520, 1 B FRV, will fully open and 1 B SG level will rise. The crew will take actions per hard card 1 BwPR 1-15-SG to stabilize the plant by taking manual control of the 1 B FRV controller.
After the 1 FK-520 failure is addressed, Heater Drain Tank (HOT) Level Controller, 1 LK-HD009A, will fail to 0% demand. The 1 HD046A/B valves will throttle close and HOT level will rise. The BOP will take actions to stabilize the plant by taking manual control of the 1 LK-HD009A controller.
After the HOT Level Controller failure is addressed, the 1 B MSIV fails closed causing three SG safety valves on the 1 B SG to stick open resulting in a faulted SG. SG pressure will drop and a manual reactor trip will be required. The crew will implement 1 BwEP-0 "REACTOR TRIP OR SAFETY INJECTION." When safety injection is actuated, the 1A CV pump will trip. The 1 B CV pump must be manually started to establish high head ECCS flow. After determining 1 B SG secondary pressure boundary is not intact, the crew will transition to 1 BwEP-2 "FAUL TED STEAM GENERA TOR ISOLATION." When 1 B SG pressure drops to 600 psig, a SGTR will occur on the 1 B SG, causing a faulted/ruptured SG. The crew will complete isolation of 1B SG and transition to 1BwEP-3 "STEAM GENERATOR TUBE RUPTURE" based on secondary radiation trends on the 1 B SG. In addition, the crew will recognize 1 B SG pressure does not drop to zero and lowering pressurizer level/pressure will indicate a SGTR (alternate indications). After determining the ruptured SG pressure is less than 320 psig, the crew will transition 1 BwCA-3.1 "SGTR WITH LOSS OF REACTOR COOLANT-SUBCOOLED RECOVERY DESIRED."
Completion criteria is completion of step 8 to stop the RH pumps in 1 BwCA-3.1.
Critical Tasks
- 1. Manually start the 1 B CV pump prior to completion of step 6 of 1 BwEP-0.
(Westinghouse - CT-6) (KIA number - 013000A4.01 importance 4.5/4.8)
- 2. Isolate 1 B Steam Generator prior to completing step 4 of 1 BwEP-2.
(Westinghouse - CT-18) (KIA number - APE040AA 1.10 importance 4.1 /4.1) 151 NRC 4 Page 2