ML112700886
| ML112700886 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/26/2011 |
| From: | NRC Region 4 |
| To: | |
| References | |
| 50-482/11-08 | |
| Download: ML112700886 (38) | |
Text
Printed: ES-401 Facility: Wolf Creek PWR Examination Outline Form ES-401-2 3 3 3 3 3 3 1 2 2 1 2 1 4 5 5 4 5 4 3 2 3 3 2 2 3 3 2 2 3 0 1 1 1 1 1 1 1 1 1 1 3 3 4 4 3 3 4 4 3 3 4 3 2 2 3 RO K/A Category Points Tier Group Total K1 K2 K3 K4 K5 A1 A2 A3 A4 K6 G* 1. Emergency & Abnormal Plant Evolutions 1 1 2 2 Tier Totals Tier Totals 2. Plant Systems 1 2 3 4 3. Generic Knowledge And Abilities Categories
- 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.
- 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.
- 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 K/A statements.
- 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.
Note: Date of Exam: 08/29/2011 28 18 9 27 10 38 10 1. Ensure that at least two topics from every applicable K/A 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 K/A category shall not be less than two).
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A 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 K/A 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 lef t side of Column A2 for Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
SRO-Only Points A2 G* 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 1 2 3 4 0 0 0 0 0 0 0 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES
-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
N/A N/A Total 0 0 1 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
ES - 401 Facility: Wolf Creek E/APE # / Name / Safety Function Emergency and Abnormal Plant Evolutions
- Tier 1 / Group 1 PWR RO Examination Outline Form ES-401-2 EK2.03 - Reactor trip status panel 3.5 1 X 000007 Reactor Trip
- Stabilization
- Recovery / 1 AK2.02 - Sensors and detectors 2.7* 1 X 000008 Pressurizer Vapor Space Accident / 3 EK2.03 - S/Gs 3.0 1 X 000009 Small Break LOCA / 3 EA2.04 - Significance of PZR readings 3.7 1 X 000011 Large Break LOCA / 3 AA2.01 - Cause of RCP failure 3.0 1 X 000015/000017 RCP Malfunctions / 4 AK1.0 2 - Relationship of charging flow to press. diff. between charging and RCS 2.7 1 X 000022 Loss of Rx Coolant Makeup / 2 AK3.02 - Isolation of RHR low
-pressure piping prior to pressure increase above specified level 3.3 1 X 000025 Loss of RHR System / 4 2.4.2 - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions
. 4.5 1 X 000026 Loss of Component Cooling Water / 8 2.2.22 - Knowledge of limiting conditions for operations and safety limits.
4.0 1 X 000027 Pressurizer Pressure Control System Malfunction / 3 EA1.15 - AFW source level and capacity (chart) 3.9 1 X 000038 Steam Gen. Tube Rupture / 3 AK1.03 - RCS shrink and consequent depressurization 3.8 1 X 000040 Steam Line Rupture
- Excessive Heat Transfer / 4 AK1.02 - Effects of feedwater introduction on dry S/G 3.6 1 X 000054 Loss of Main Feedwater / 4 EA1.01 - In-core thermocouple temperatures 3.7 1 X 000055 Station Blackout / 6 2.1.2 8 - Knowledge of the purpose and function of major system components and controls. 4.1 1 X 000062 Loss of Nuclear Svc Water / 4 AA2.0 9 - Operational status of emergency diesel generators 3.9 1 X 000077 Generator Voltage and Electric Grid Disturbances / 6 EA1.2 - Operating behavior characteristics of the facility 3.6 1 X W/E04 LOCA Outside Containment / 3 EK3.2 - Normal, abnormal and emergency operating procedures associated with Loss of Secondary Heat Sink 3.7 1 X W/E05 Loss of Secondary Heat Sink / 4 2
K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
ES - 401 Facility: Wolf Creek E/APE # / Name / Safety Function Emergency and Abnormal Plant Evolutions
- Tier 1 / Group 1 PWR RO Examination Outline Form ES-401-2 EK3.2 - Normal, abnormal and emergency operating procedures associated with Loss of Emergency Coolant Recirculation 3.5 1 X W/E11 Loss of Emergency Coolant Recirc. / 4 18 3 3 3 3 3 3 K/A Category Totals:
Group Point Total: 3 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
ES - 401 Facility: Wolf Creek E/APE # / Name / Safety Function Emergency and Abnormal Plant Evolutions
- Tier 1 / Group 2 PWR RO Examination Outline Form ES-401-2 AK2.06 - T-ave./ref. deviation meter 3.0* 1 X 000001 Continuous Rod Withdrawal / 1 AA2.02 - Signal inputs to rod control system 2.7 1 X 000003 Dropped Control Rod / 1 AK3.01 - Boration and emergency boration in the event of a stuck rod during trip or normal evolutions 4.0 1 X 000005 Inoperable/Stuck Control Rod / 1 AK2.03 - Controllers and positioners 2.6 1 X 000028 Pressurizer Level Malfunction / 2 AA1.0 4 - Condensate air ejector exhaust radiation monitor and failure indicator 3.6 1 X 000037 Steam Generator Tube Leak / 3 2.4.31 - Knowledge of annunciator alarms, indications, or response procedures.
4.2 1 X 00005 1 Loss of Condenser Vacuum / 4 AK3.02 - Guidance contained in alarm response for ARM system 3.4 1 X 000061 ARM System Alarms / 7 AK1.01 - Effect of pressure on leak rate 2.6 1 X 000069 Loss of CTMT Integrity / 5 EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments 3.3 1 X W/E01 Rediagnosis / 3 9 1 2 2 1 2 1 K/A Category Totals:
Group Point Total:
4 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
K4 K5 A3 K6 A4 ES - 401 Sys/Evol # / Name Facility:
Wolf Creek PWR RO Examination Outline Plant Systems
- Tier 2 / Group 1 Form ES-401-2 K2.01 - RCPS 3.1 1 X 003 Reactor Coolant Pump A4.05 - RCP seal leakage detection instrumentation 3.1 1 X 003 Reactor Coolant Pump K3.0 7 - PZR level and pressure 3.8 1 X 004 Chemical and Volume Control K2.03 - RCS pressure boundary motor
-operated valves 2.7* 1 X 005 Residual Heat Removal A4.0 1 - Controls and indication for RHR pumps 3.6* 1 X 005 Residual Heat Removal K3.01 - RCS 4.1* 1 X 006 Emergency Core Cooling 2.1.3 2 - Ability to explain and apply system limits and precautions
. 3.8 1 X 007 Pressurizer Relief/Quench Tank A2.01 - Stuck-open PORV or code safety 3.9 1 X 007 Pressurizer Relief/Quench Tank K1.02 - Loads cooled by CCWS 3.3 1 X 008 Component Cooling Water K5.01 - Determination of condition of fluid in PZR, using steam tables 3.5 1 X 010 Pressurizer Pressure Control K4.08 - Logic matrix testing 2.8* 1 X 012 Reactor Protection K4.09 - Spurious trip protection 2.7 1 X 013 Engineered Safety Features Actuation A1.04 - Cooling water flow 3.2 1 X 022 Containment Cooling A3.01 - Initiation of safeguards mode of operation 4.1 1 X 022 Containment Cooling A2.07 - Loss of Ctmt Spray pump suction when in recirc. mode 3.6 1 X 026 Containment Spray K5.08 - Effect of steam removal on reactivity 3.6 1 X 039 Main and Reheat Steam K4.17 - Increased feedwater flow following a reactor trip 2.5* 1 X 059 Main Feedwater K6.01 - Controllers and positioners 2.5 1 X 061 Auxiliary/Emergency Feedwater A1.03 - Effect on instrumentation and controls of switching power supplies 2.5 1 X 062 AC Electrical Distribution K3.02 - Components using DC control power 3.5 1 X 063 DC Electrical Distribution 2.1.31 - Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup
. 4.6 1 X 063 DC Electrical Distribution 5
K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
K4 K5 A3 K6 A4 ES - 401 Sys/Evol # / Name Facility:
Wolf Creek PWR RO Examination Outline Plant Systems
- Tier 2 / Group 1 Form ES-401-2 K6.08 - Fuel oil storage tanks 3.2 1 X 064 Emergency Diesel Generator 2.1.30 - Ability to locate and operate components, including local controls
. 4.4 1 X 073 Process Radiation Monitoring K1.01 - Those systems served by PRMs 3.6 1 X 073 Process Radiation Monitoring A1.02 - Reactor and turbin e building closed cooling water temperatures 2.6* 1 X 076 Service Water A3.01 - Air pressure 3.1 1 X 078 Instrument Air K1.04 - Cooling wtr to comp.
2.6 1 X 078 Instrument Air A2.04 - Containment evacuation (including recognition of the alarm) 3.5* 1 X 103 Containment 28 3 2 3 3 3 3 K/A Category Totals:
3 2 2 2 2 Group Point Total:
6 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
K4 K5 A3 K6 A4 ES - 401 Sys/Evol # / Name Facility:
Wolf Creek PWR RO Examination Outline Plant Systems
- Tier 2 / Group 2 Form ES-401-2 K4.01 - Filling and draining the RCS 2.7 1 X 002 Reactor Coolant K6.04 - Operation of PZR level controllers 3.1 1 X 011 Pressurizer Level Control A1.0 4 - Axial and radial power distribution 3.5 1 X 014 Rod Position Indication K3.02 - CRDS 3.3* 1 X 015 Nuclear Instrumentation K5.01 - Separation of control and protection circuits 2.7* 1 X 016 Non-nuclear Instrumentation K2.01 - Fans 3.1* 1 X 027 Containment Iodine Removal A2.01 - Hydrogen recombiner power setting, determined by using plant data book 3.4* 1 X 028 Hydrogen Recombiner and Purge Control A4.01 - Containment purge flow rate 2.5 1 X 029 Containment Purge 2.2.1 2 - Knowledge of surveillance procedures
. 3.7 1 X 034 Fuel Handling Equipment A3.01 - S/G water level control 4.0 1 X 035 Steam Generator 10 0 1 1 1 1 1 K/A Category Totals:
1 1 1 1 1 Group Point Total:
7 Facility: Wolf Creek Generic Category KA KA Topic Imp. Points Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Printed: PWR RO Examination Outline 2.1.18 Ability to make accurate, clear, and concise logs, records, status boards, and reports.
3.6 1 Conduct of Operations 2.1.36 Knowledge of procedures and limitations involved in core alterations.
3.0 1 2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.
4.3 1 3 Category Total:
2.2.41 Ability to obtain and interpret station electrical and mechanical drawings
. 3.5 1 Equipment Control 2.2.4 3 Knowledge of the process used to track inoperable alarms. 3.0 1 2 Category Total:
2.3.7 Ability
to comply with radiation work permit requirements during normal or abnormal conditions.
3.5 1 Radiation Control 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high
-radiation areas, aligning filters, etc.
3.2 1 2 Category Total:
2.4.9 Knowledge
of low power /shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.
3.8 1 Emergency Procedures/Plan 2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
4.0 1 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.
4.2 1 3 Category Total:
10 Generic Total:
8 Printed: ES-401 Facility: Wolf Creek PWR Examination Outline Form ES-401-2 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 RO K/A Category Points Tier Group Total K1 K2 K3 K4 K5 A1 A2 A3 A4 K6 G* 1. Emergency & Abnormal Plant Evolutions 1 1 2 2 Tier Totals Tier Totals 2. Plant Systems 1 2 3 4 3. Generic Knowledge And Abilities Categories
- 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.
- 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.
- 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 K/A statements.
- 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.
Note: Date Of Exam:
08/29/2011 0 0 0 0 0 0 0 1. Ensure that at least two topics from every applicable K/A 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 K/A category shall not be less than two).
7.* The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A 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 K/A 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.
SRO-Only Points A2 G* 3 3 2 2 5 5 3 2 2 1 3 2 1 2 2 1 2 3 4 10 4 6 5 3 8 7 9. For Tier 3, select topics from Section 2 of the K/A catalog, and enter the K/A numbers, descriptions, IRs, and point totals (#) on Form ES
-401-3. Limit SRO selections to K/As that are linked to 10 CFR 55.43.
- 6. Select SRO topics for Tiers 1 and 2 from the shaded systems and K/A categories.
N/A N/A Total 0 5 9 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
ES - 401 Facility: Wolf Creek E/APE # / Name / Safety Function Emergency and Abnormal Plant Evolutions
- Tier 1 / Group 1 PWR SRO Examination Outline Form ES-401-2 EA2.0 1 - Reactor nuclear instrumentation 4.7 1 X 000029 ATWS / 1 AA2.0 3 - Operational status of safety injection pump 3.9 1 X 000056 Loss of Off
-site Power / 6 2.4.8 - Knowledge of how abnormal operating procedures are used in conjunction with EOP's
. 4.5 1 X 000057 Loss of Vital AC Inst. Bus / 6 2.4.3 - Ability to identify post
-accident instrumentation
. 3.9 1 X 000058 Loss of DC Power / 6 AA2.01 - Cause and effect of low
-pressure instrument air alarm 3.2 1 X 000065 Loss of Instrument Air / 8 2.4.18 - Knowledge of the specific bases for EOP's. 4.0 1 X W/E12 - Uncontrolled Depressurization of all Steam Generators
/ 4 6 0 0 0 0 3 3 K/A Category Totals:
Group Point Total:
10 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
ES - 401 Facility: Wolf Creek E/APE # / Name / Safety Function Emergency and Abnormal Plant Evolutions
- Tier 1 / Group 2 PWR SRO Examination Outline Form ES-401-2 2.2.25 - Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. 4.2 1 X 000033 Loss of Intermediate Range NI / 7 EA2.01 - Subcooling Margin 4.9 1 X 000074 Inad. Core Cooling / 4 E A2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.
3.4 1 X W/E 13 Steam Generator Over
-pressure / 4 2.4.1 - Knowledge of EOP entry conditions and immediate action steps.
4.8 1 X W/E1 5 Containment Flooding
/ 5 4 0 0 0 0 2 2 K/A Category Totals:
Group Point Total:
1 1 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
K4 K5 A3 K6 A4 ES - 401 Sys/Evol # / Name Facility:
Wolf Creek PWR SRO Examination Outline Plant Systems
- Tier 2 / Group 1 Form ES-401-2 2.4.6 - Knowledge of EOP mitigation strategies.
4.7 1 X 006 Emergency Core Cooling A2.0 5 - Effect of loss of inst. and cont. air on the position of the CCW valves 3.5 1 X 008 Component Cooling Water A2.03 - PORV failures 4.2 1 X 010 Pressurizer Pressure Control 2.1.25 - Ability to interpret reference materials, such as graphs, curves, tables, etc.
4.2 1 X 062 AC Electrical Distribution A2.06 - Operating unloaded, lightly loaded, and highly loaded time limit 3.3 1 X 064 Emergency Diesel Generator 5 0 0 0 0 3 2 K/A Category Totals:
0 0 0 0 0 Group Point Total:
1 2 K1 KA Topic Imp. K2 K3 A1 A2 G Points Printed:
K4 K5 A3 K6 A4 ES - 401 Sys/Evol # / Name Facility:
Wolf Creek PWR SRO Examination Outline Plant Systems
- Tier 2 / Group 2 Form ES-401-2 A2.1 7 - Malfunction of electrohydraulic control 2.9* 1 X 045 Main Turbine Generator 2.1.20 - Ability to interpret and execute procedure steps
. 4.6 1 X 068 Liquid Radwaste A2.0 4 - Positioning of axial shaping rods and their effect on SDM 3.8* 1 X 001 Rod Control 3 0 0 0 0 2 1 K/A Category Totals:
0 0 0 0 0 Group Point Total:
1 3 Facility: Wolf Creek Generic Category KA KA Topic Imp. Points Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Printed: PWR SRO Examination Outline
2.1.4 Knowledge
of individual licensed operator responsibilities related to shift staffing, such as medical requirements, "no
-solo" operation, maintenance of active license status, 10CFR55, etc.
3.8 1 Conduct of Operations 2.1.35 Knowledge of the fuel
-handling responsibilities of SROs. 3.9 1 2 Category Total:
2.2.38 Knowledge of conditions and limitations in the facility license.
4.5 1 Equipment Control 1 Category Total:
2.3.5 Ability
to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring equipment, etc.
2.9 1 Radiation Control 2.3.13 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high
-radiation areas, aligning filters, etc.
3.8 1 2 Category Total:
2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.
4.4 1 Emergency Procedures/Plan 2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm.
4.3 1 2 Category Total:
7 Generic Total:
1 4 ES-401 Record of Rejected K/As Form ES-401-4 Tier / Group Randomly Selected K/A Reason for Rejection S 1 / 1 056 AA2.02 Components not installed at WC. Randomly selected AA 2.03 2 / 1 005 A4.05 Component not utilized at WCNOC. Randomly selected A4.01 S 2 / 1 008 A2.01 No real actions for SRO Question. Randomly selected A2.05 2 / 1 026 A2.02 Not applicable to WC, replaced with A2.0 7 2 / 1 078 K1.05 Not applicable to WC, replaced with K1.04 1 / 2 059 2.4.31 Due to overlap with other RMS K/A's replaced topic with 051, kept same generic S 1 / 2 076 AA2.01 Due to overlap with other RMS K/A's replaced with E13 EA2.2 S 1 / 2 E16 2.4.1 Due to overlap with other RMS K/A's replaced topic with E15, kept same generic 1 / 1 022 AK1.01 Due to overlap with other RCP K/A's, replaced with AK1.02 2 / 1 004 K3.08 Due to overlap with other RCP K/A's, replaced with K3.07 1 / 1 077 AA2.04 Replaced with AA2.09, topic too similar to SRO topic 0 62 1 / 2 037 AA1.05 Replaced with AA1.04, lack of components to meet K/A S 1 / 1 029 EA2.02 Replaced with EA2.01, unable to get to appropriate SRO level with limited distractors S 2 / 2 079 A2.01 Replaced with 001 A2.04, topic too similar to RO 078 topi c
FINAL 1 of 3 ES-301 Administrative Topics Outline Form ES-301-1 Facility: Wolf Creek Date of Examination:
Aug.- Sept. 2011 Examination Level: RO SRO Operating Test Number:
Administrative Topic (see Note) Type Code* Describe activity to be performed Conduct of Operations R.A.1.a S.A.1.a N, R M, R R.A.1.a Refuel/ Reduced Inventory: Perform the time to core uncovery estimation using the OFN EJ
-015, LOSS OF RHR COOLING, step 31. Requires use of Figures 5 (time to boil) and 6 (time to uncovery).
2.1.25 Ability to interpret reference materials, such as graphs, curves tables, etc. (CFR 41.10/43.5/45.12 RO = 3.9 SRO = 4.2)
S.A.1.a Review/Approve/Evaluate the Reactor Operator's completed manual calculation of RTP; STS SE-002, MANUAL CALCULATION OF REACTOR THERMAL POWER. Requires discovery of errors made by Reactor Operator.
2.1.20 Ability to interpret and execute procedure steps. (CFR 41.10/43.5/45.12 RO = 4.6 SRO = 4.6)
Conduct of Operations R A.1.b S.A.1.b N, R N, R R.A.1.b Determine the shutdown margin using STS RE-004, SHUTDOWN MARGIN DETERINATION, Attachment A, Shutdown Margin Calculation Short form. 2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management (CFR 41.1/43.6/45.6 RO = 4.3 SRO = 4.6)
S.A.1.b Review/Approve/Verify the Reactor Operator's completed manual calculation of the shutdown margi n per STS RE
-004, SHUTDOWN MARGIN DETERINATION, Attachment A, Shutdown Margin Calculation Short form.
2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management (CFR 41.1/43.6/45.6 RO = 4.3 SRO = 4.6)
FINAL 2 of 3 Equipment Control R.A.2 S.A.2 N, R N, R R.A.2 Complete STS AL
-211, TURB DRIVEN AUX FDWTR SYS FLOW PATH VERIFICATION & INSERVICE CHEC VALVE TEST, Attachment A Data Sheet.
2.2.12, Knowledge of surveillance procedures (CFR 41.10/45.13 RO = 3.7 SRO = 4.1)
S.A.2 Review/Approve/Evaluate the Reactor Operator's completed STS EF-100A, ESW SYSTEM INSERVICE PUMP A & ESW A DISCHARGE CHECK VALVE TEST
, Attachment A Data Sheet.
2.2.12, Knowledge of surveillance procedures (CFR 41.10/45.13 RO = 3.7 SRO = 4.1)
Radiation Control S.A.3 N, R S.A.3 The Containment Purge permit that was in progress was stopped. Determine/Authorize the restart for the Containment Purge Permit. (AP 07B
-001, Radioactive Releases, see section 6.2.4.6)
2.3.6 Ability
to approve release permits (CFR 41.13/43.4/45.9 RO = 2.0 SRO = 3.8) and/or 2.3.11 Ability to control radiation releases (CFR 41.11/43.4/45.10 RO = 3.8 SRO = 4.3)
FINAL 3 of 3 Emergency Procedures/Plan R.A.4 S.A.4 N, R D, R R.A.4 Determine percentage of Control Room annunciator loss using OFN PK
-029, LOSS OF NON
-VITAL 125VDC BUS PK01, PK02, PK03, PK04, AND ANNUNCIATORS.
2.4.32 Knowledge of operator response to loss of all annunciators. (CFR 41.10/43.5/45.13 RO = 3.6 SRO = 4.0)
S.A.4 (In the classroom setting) Determine the E
-Plan classification and Protective action recommendations, if any.
2.4.41 Knowledge of the emergency action level thresholds and classifications. (CFR 41.10/43.5/45.11 RO = 2.9 SRO = 4.6) and 2.4.44 Knowledge of emergency plan protective actio n recommendations. (CFR 41.10/41.12/43.5/45.11 RO =
2.4 SRO = 4.4)
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 ( (N)ew or (M)odified from bank ( (P)revious 2 exams (
FINAL 1 of 6 ES-301 Control Room/In
-Plant Systems Outline Form ES-301-2 Facility: Wolf Creek Date of Examination:
Aug. -Sept. 2011 Examination Level: RO SRO Operating Test Number:
Control Room Systems
-I); (2 or 3 for SRO
-U, including 1 ESF)
Bolded is an Alternate Success Path JPM.
System / JPM Title Type Code*
Safety Function a. S1: 001 - Control Rod Drive System Perform the actions of STS SF
-001, CONTROL AND SHUTDOWN ROD OPERABILITY VERIFICATION, for Control Bank A. 001 2.2.12 Knowledge of surveillance procedures. (3.7/4.1)
RO/SRO-I N, S 1 b. S2: 013
- Engineered Safety Features Actuation System (ESFAS)
Perform actions to ensure CRVIS actuation using ALR 00
-062D, FBIS and ALR 00
-063A, CRVIS.
PRA: ESFAS is a Risk Significant System at Wolf Creek.
013 A4.01 Ability to manually operate and/or monitor in the control room ESFAS
-initiated equipment which fails to actuate. (4.5/4.8)
RO/SRO-I/SRO-U N, EN, A, S 2
FINAL 2 of 6 c. S3: 006
- Emergency Core Cooling System (ECCS)
Perform actions to increase level in an Accumulator using a Safety Injection Pump per procedure SYS EP
-200, SAFETY INJECTION ACCUMULATOR OPERATIONS (see sections 6.1, 6.2, 6.3 or 6.4), however, gas voiding is diagnosed due to SIP oscillations and OFN BG
-045, GAS BINDING OF CCPS OR SI PUMPS, is entered and performed.
SOER 97-1, Potential Loss of High Pressure Injection and Charging Capability from Gas Intrusion 006 A2.04 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (2) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: improper discharge pressure. (3.4/3.8) 006 A2.05 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (2) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: improper amperage to the pump motor. (3.4/3.5) 006 A4.01 Ability to manually operate and/or monitor in the control room: pumps. (4.1/3.9)
RO/SRO-I/SRO-U D, A, S 3 d. S4: 041
- Steam Dump System and Turbine Bypass Control Perform actions to establish a maximum rate cooldown using the ARV's per EMG E
-3, STEAM GENERATOR TUBE RUPTURE. 041 A4.06 Ability to manually operate and/or monitor in the control room: Atmospheric relief valve controllers. (2.9/3.1)
RO/SRO-I M, L, S 4S FINAL 3 of 6 e. S5: 003
- Reactor Coolant Pumps System Align alternate seal injection and place excess letdown into service per OFN KA
-019, LOSS OF INSTRUMENT A IR. 003 A4.01 Ability to manually operate and/or monitor in the control room: Seal injection (3.3/3.2) RO/SRO-I/SRO-U N, L, A, S 4P f. S6: 103
- Containment Systems Perform actions to startup the Containment Purge System per SYS GT-120, CONTAINMENT MINI PURGE SYSTEM OPERATIONS, sections 6.1 and 6.2.
103 A1.01 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the containment system controls including: Containment pressure, temperature, and humidity. (3.7/4.1)
RO D, S 5 g. S7: 015
- Nuclear Instrumentation Perform actions to bypass a failed Power Range nuclear instrumentation channel using OFN SB
-008, INSTRUMENT MALFUNCTIONS, Attachment R (see step R4).
015 A4.03 Ability to manually operate and/or monitor in the control room: Trip bypasses. (3.8/3.9)
RO/SRO-I D, S 7 FINAL 4 of 6 h. S8: 008 - Component Cooling Water System (CCW)
Perform actions of ALR 00
-052A, CCW TO RCP FLOW LO , to respond to a loss of a CCW pump.
A4.01 Ability to operate and/or monitor in the control room: CCW indications and controls. (3.3/3.1)
A2.01 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of a CCW pump. (3.3/3.6)
PRA: Component Cooling Water is a Risk Significant System at Wolf Creek.
RO/SRO-I M, A, S 8 In-Plant Systems
-I); (3 or 2 for SRO
-U) i. P1: 004
- Chemical and Volume Control System Perform local actions to borate the Reactor Coolant System. (See OFN BG
-009, EMERGENCY BORATION , Attachment A, Establishing Alternate Boration Flowpath.)
004 A2.14 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations. (3.8/3.9)
APE 024 AA1.04 Ability to operate and/or monitor the following as they apply to Emergency Boration: Manual boration valve. (3.6/3.7)
RO/SRO-I/SRO-U D, A, R, E 1
FINAL 5 of 6 j. P2: 061
- Auxiliary/Emergency Feedwater System Perform actions of STN FC
-002, AUX FEEDWATER TURBINE OVERSPEED TEST section 8.1.6.
061 2.1.20 Ability to interpret and execute procedure steps. (4.4/4.6))
PRA: Auxiliary Feedwater (AL) is a Risk Significant System at Wolf Creek.
RO/SRO-I N 4S k. P3: 064
- Emergency Diesel Generators Perform actions of ALR 00
-020D, DG NE01 TROUBLE alarm. Local alarm response procedure ALR 501, STANDBY DIESEL ENGINE SYSTEM CONTROL PANEL KJ-121, Attachment A, Fuel Oil Press Low and Attachment C, Fuel Strain Diff Press High, are performed.
064 K1.03 Knowledge of the physical connections and/or cause-effect relationship between the ED/G system and the following systems: Diesel fuel oil supply system.
(3.6/4.0) PRA: Diesel Fuel Oil (JE) is a Risk Significant System at Wolf Creek.
RO/SRO-I/SRO-U D, A 6 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.
-I / SRO-U FINAL 6 of 6 (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 (S)imulator 4-6 / 4-6 / 2-3 - / - / d)
FINAL 1 of 5 Appendix D Scenario Outline Form ES-D-1 Facility: ___Wolf Creek_______________ Scenario No.: ___1_____
Op-Test No.: _______
Examiners: ____________________________ Operators:
_____________________________
____________________________
_____________________________
____________________________
_____________________________
Initial Conditions: MOL, 100%
Turnover: Red train CCW (pumps A/C secured due to leakage). TS 3.7.7 Cond A entered (72 hrs to restore). Welding on CCW A Surge tank outlet. Expected return in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TS 3.5.2 Cond A entered (72 hrs to restore). (ESFAS alarms are illuminated). Red train ECCS pumps are DNO'd or have a TEST/CAUTION (TC) tag and pumps are in Pull
-to-Lock (PTL). This includes: CCW "A" (DNO), CCW "C" (DNO), CCP "A" (TC), SIP "A" (TC) and RHR "A" (TC). DNO tags are on EG HV
-11 and 13, EG HIS
-1 and EG ZL
-15 and 53. Perform a power reduction and turbine load decrease to 900 MWE NET using OFN MA
-038, RAPID PLANT SHUTDOWN at a rate of 1%/minute.
Event No. Malf. No. Event Type* Event Description 1 R - ATC, SRO N - BOP The Crew commences a power decrease and turbine load reduction to 900 MWE NET (945 MWE GROSS) per OFN MA-038, RAPID PLANT SHUTDOWN at a rate of 1%/minute
. 2 mAB01D 2 I - BOP, SRO Steam Generator "D" pressure channel AB PT
-545 fails low TS determined & entered. TS 3.3.2, Table 3.3.2
-1, Fu 1e and 4e. Cond A (Immediately) and Cond D (72 hrs to trip bistables) are entered. 3 mBB21B I - ATC, SRO Pressurizer pressure channel BB PI
-456 fails high TS determined & entered. TS 3.3.1, Table 3.3.1
-1, Fu 6, 8.a and 8.b. Cond A (Immediately), Cond E (72 hrs to trip bistables) and Cond M (72 hrs to trip bistables) are entered.
TS 3.3.2, Table 3.3.2
-1, Fu 1.d, 3.a.3, 5.d, 6.e and 8.b. Cond A (Immediately) and Cond D (72 hrs to trip bistables), and Cond L (1 hr to verify interlock (P
-11)). 4 mBB06C M - CREW Large Break LOCA: cold leg break on Loop "C" 5 mEJ13B C - ATC, SRO Post trip malfunction #1: Autostart failure of RHR "B" pump. Manual start is available.
6 mSA27E C02 C - ATC, SRO Post trip malfunction #2: Auto closure of EC HIS
-12, SFP HX B CCW OUTLET VLV, failure to close. Manual closure available. * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
FINAL 2 of 5 Scenario summary:
The unit is at 100% power, middle of life. Turnover items include CCW pumps "A" and "C" (Red train) are secured due to leakage. Welding on CCW "A" Surge tank outlet is ongoing. Technical Specification 3.7.7 Condition A was entered (72 hrs to restore). Expected return to service is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Red train ECCS pumps are DNO'd or have a TEST/CAUTION (TC) tag and pumps are in Pull-to-Lock (PTL). This includes: CCW "A" (DNO), CCW "C" (DNO), CCP "A" (TC), SIP "A" (TC) and RHR "A" (TC). DNO tags are on EG HV
-11 and 13, EG HIS
-1 and EG ZL
-15 and 53.
Topeka Dispatch/System Operator called to inform Wolf Creek that 345
-50 KV Benton line will be removed from service in 20 minutes for four hours. Directive #300 was performed. Per Directive #300 Wolf Creek will be divorced from the Athens line (also opening 69
-14 Breaker). Reduce power and decrease turbine load to less than 900 MWE NET.
The Call Superintendent has directed the crew to use OFN MA
-038, RAPID PLANT SHUTDOWN to maneuver the unit at a rate of 1%/minute.
Event 1: The Crew commences a power reduction and turbine load reduction to 900 MWE NET (945 MWE GROSS) per OFN MA
-038, RAPID PLANT SHUTDOWN at a rate of 1%/minute. Event 2: Steam Generator "D" pressure channel AB PT
-545 fails low. Meter indications change, and Main Control Board alarms annunciate. ALRs 00
-111C, SG D FLOW MISMATCH or 00
-111B SG D LEV DEV, may be entered and performed. OFN SB
-008, INSTRUMENT MALFUNCTIONS, is entered and Attachment C performed. These procedures diagnose and mitigate the instrument failure.
The Control Room Supervisor determines Technical Specifications.
Event 3: Pressurizer (PZR) pressure channel BB PI
-456 fails high. The PZR spray valves close, meter indications change and various Main Control Board alarms annunciate. ALRs 00
-034B, PZR PRESS HI, 00
-034C, PZR PORV BLOCK; 00
-034E, PRT PRESS HI; 00
-035B, PORV OPEN; 00-035D, PZR PORV DISCH TEMP HI; 00
-083C, RX PARTIAL TRIP annunciate. OFN SB-008, INSTRUMENT MALFUNCTIONS, is entered and Attachment K performed. These procedures diagnose and mitigate the instrument failure.
The Control Room Supervisor determines Technical Specifications.
Event 4: The Main Event is a Large Break Loss of Coolant Accident.
Diagnostics include: PZR level decreases and RCS pressure decreases. OFN BB
-007, SG/RCS LEAKAGE HIGH, may be entered & performed. A Reactor trip and Safety Injection occur. EMG E-0, REACTOR TRIP OR SAFETY INJECTION, is entered & performed.
RCP's are tripped per EMG E
-0 Foldout page criteria.
EMG-E-1, LOSS OF REACTOR OR SECONDARY COOLANT is entered & performed.
Eventually 36% Refueling Water Storage Tank (RWST) level is achieved and Main Control Board alarm ALR 00
-047C, RWST LEV LOLO 1 AUTO XFR actuates. ALR 00
-047C directs performance of EMG ES
-12, TRANSFER TO COLD LEG RECIRCULATION.
The crew transitions to EMG ES
-12, TRANSFER TO COLD LEG RECIRCULATION. The procedure is performed through step 10 to establish cold leg recirculation/ECCS recirculation.
FINAL 3 of 5 Post trip malfunctions:
Event 5: Autostart failure of RHR "B" pump. Manual start is available. This component failure is procedurally addressed in Attachment F of EMG E
-0, REACTOR TRIP OR SAFETY INJECTION. However, the pump can be started after the Immediate Actions of EMG E
-0, REACTOR TRIP OR SAFETY INJECTION, are performed and concurrence of the CRS is obtained.
Event 6: Auto closure of EC HIS
-12, SFP HX B CCW OUTLET VLV, fails to close. Manual closure is available. This component failure is procedurally addressed in EMG ES
-12, TRANSFER TO COLD LEG RECIRCULATION, at step 3.
Scenario Critical Tasks (CT):
Event 2: CT: take manual control, select alternate controlling channel prior to actuation of the Reactor Protection System Event 4: CT: using EMG ES
-12, steps 1 through 10, transfer to cold leg recirculation to establish ECCS recirculation Event 5: CT: start RHR "B" pump, as this is the only low head injection pump available for decay heat removal for a Large Break LOCA.
FINAL 4 of 5 Probabilistic Risk Analysis for this scenario includes:
Core Damage Frequency by Initiating Event Initiating Event Initiating Event Frequency (/yr)
Core Damage Frequency (/yr)
CDF Percent Contribution Loss of Offsite Power 2.88E-02 6.59E-06 36.51% Small LOCA 3.00E-03 5.35E-06 29.63% Interfacing Systems LOCA 1.93E-06 10.69% Very Small LOCA 6.20E-03 1.27E-06 7.03% Transients With Power Conversion Systems Available 1.05E+00 9.88E-07 5.47% Steam Generator Tube Rupture 3.67 E-03 8.77E-07 4.86% Reactor Vessel Failure 3.00E-07 3.00E-07 1.66% Steamline Break 1.13E-02 1.88E-07 1.04% Transients Without Power Conversion Systems Available 1.15E-01 1.71E-07 0.95% Medium LOCA 6.10E-05 1.46E-07 0.81% Loss of All Service Water 6.86 E-06 8.30E-08 0.46% Loss of Component Cooling Water 2.14E-04 5.79E-08 0.32% Loss of Vital DC Bus NK04 2.64E-03 4.32E-08 0.24% Large LOCA 7.20E-06 2.80E-08 0.16% Feedwater Line Break 3.17E-03 2.06E-08 0.11% Loss of Vital DC Bus NK01 2.64E-03 1.12E-08 0.06% Top Risk Significant Systems EF Essential Service Water KJ/NE Onsite Emergency Power EG Component Cooling Water AL Aux Feedwater EJ Residual Heat Removal JE Diesel Fuel Oil NB Lower Medium Voltage NK 125 V DC BB Reactor Coolant System GM Diesel Building HVAC GD ESW HVAC GL Aux Building HVAC BN Refueling Water Storage Tank SA/SB ESFAS/Reactor Protection
FINAL 5 of 5 Technical Specifications exercised:
Event 2: TS determined & entered. TS 3.3.2, Table 3.3.2
-1, Fu 1e and 4e. Cond A (Immediately) and Cond D (72 hrs to trip bistables) are entered.
Event 3: TS determined & entered. TS 3.3.1, Table 3.3.1
-1, Fu 6, 8.a and 8.b. Cond A (Immediately), Cond E (72 hrs to trip bistables) and Cond M (72 hrs to trip bistables) are entered.
TS 3.3.2, Table 3.3.2-1, Fu 1.d, 3.a.3, 5.d, 6.e and 8.b. Cond A (Immediately) and Cond D (72 hrs to trip bistables), and Cond L (1 hr to verify interlock (P
-11)).
FINAL 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: _________Wolf Creek_________ Scenario No.: ____2____
Op-Test No.: _______
Examiners: ____________________________ Operators:
_____________________________
____________________________
_____________________________
____________________________
_____________________________
Initial Conditions: Middle Of Life, ~74%
Turnover: Monitor MFP "B" vibration. Started the downpower and are currently on HOLD at ~74% waiting an Engineering Evaluation. Annunciator 00
-058B, VCT VLV NOT IN VCT POS, due to recent 200-gallon dilution to hold power. Diluting ~100 gallons every 10
-15 minutes. No equipment is out of service.
Event No. Malf. No. Event Type* Event Description 1 mBB01E I - ATC, SRO Loop "A," BB TI
-411, Tcold fails high TS determined and entered. TS 3.3.1, Table 3.3.1
-1, Fu 6 and 7, Cond A (Immediately) and Cond E (72 hrs to trip bistables) 2 mAE15C 4 I - BOP, SRO Steam Generator "C" controlling level channel AE LI
-553 failure high TS determined and entered. TS 3.3.1, Table 3.3.1
-1, Fu 14, Cond A (Immediately) and Cond E (72 hrs to trip bistables)
TS 3.3.2, Table 3.3.2
-1, Fu 5.c and 6.d, Cond A (Immediately), Cond I (72 hrs to trip bistable) and Cond D (72 hrs to trip bistable) 3 msovBBPCV455 A C - ATC, SRO PORV BB PCV
-455A fails to 25% open due to control circuitry problems, PZR pressure begins to decrease TS determined and entered. TS 3.4.11 Cond. B.1 (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> close seal valve) and B.2 (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to de
-energize seal valve) and B.3 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair PORV) 4 mAB03A M - CREW Steam line break inside Containment (Steam Generator "A")
Adverse Containment 5 mSNF01 A C - ATC, SRO Malfunction post Reactor Trip and Safety Injection: LOCA Sequencer "A" failure at five second time.
6 mNF01A C - BOP, SRO Malfunction post Reactor Trip and Safety Injection: Main Generator and Exciter breakers fail to automatically trip.
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor
FINAL 2 of 4 Scenario Summary:
The unit is at ~74% power, middle of life. Monitor MFP "B" vibration. Started the downpower and are currently on HOLD at ~74% waiting an Engineering Evaluation. Annunciator 00
-058B, VCT VLV NOT IN VCT POS, due to recent 200
-gallon dilution to hold power. Diluting ~100 gallons every 10-15 minutes.
No equipment is out of service.
Event 1: RCS Loop "A" BB TI
-411 Tcold fails high. Meter indication changes and the Control Rods insert
- the Reactor Operator (RO) places control rods in MANUAL, stopping the insertion. Many Main Control Board alarms annunciate: 00
-065C, 00-065E, 00-066B, 00-067D, 00-068D, 00-069D, 00-082B and 00
-083C. OFN SB
-008, INSTRUMENT MALFUNCTIONS, is entered and Attachment L performed. This procedure will diagnose and mitigate the instrument failure.
The Control Room Supervisor determines Technical Specifications.
Event 2: Steam Generator "C" controlling level channel AE LI
-553 fails high. Meter indications change and Main Control Board alarms, 00
-110A, SG C LEV HI/LO and 00
-110B, SG C LEV DEV, annunciate. ALR 00
-110A, SG C LEV HI/LO, or 00
-110B, SG C LEV DEV, may be entered and performed. OFN SB
-008, INSTRUMENT MALFUNCTIONS, is entered and Attachment F is performed. These procedures diagnose and mitigate the instrument failure.
The Control Room Supervisor determines Technical Specifications.
Event 3: Pressurizer Pilot Operated Relief Valve (PORV) BB PCV
-455A fails to 25% open due to control circuitry problems. Diagnostic parameters include dual indication on hand indicating switch BB HIS
-455A, and alarms 00
-035B, PORV OPEN, 00
-035C, PZR SFTY DISCH TEMP HI, 00-035D, PZR PORV DISCH TEMP HI, 00
-034E, PRT PRESS HI annunciating. ALR 00
-035B may be entered and performed to close the PZR Seal Iso Valve using BB HIS
-8000A. This action mitigates the event.
The Control Room Supervisor determines Technical Specifications.
Event 4: The Main Event is a Steam line break inside Containment (Steam Generator "A"). Diagnostic parameters include Secondary steam flow to feed flow meters mismatch, increasing SG steam flow, Containment pressure and humidity while it decreases Main Turbine load and RCS pressure and temperature. OFN AB
-041, STEAMLINE OR FEEDLINE LEAK may be entered. A Reactor trip and Safety Injection occurs. EMG E
-0, REACTOR TRIP OR SAFETY INJECTION, is entered and performed. The faulted SG is identified and isolated (EMG E
-0 foldout page criteria). Adverse Containment is identified and setpoints for various parameters are used. The Crew transitions to EMG E
-2, FAULTED STEAM GENERATOR ISOLATION.
Eventually the Crew transitions to EMG ES
-03, SI TERMINATION, to mitigate PZR overfill and RCS high pressure.
Post trip malfunctions:
- 1. Event 5: LOCA Sequencer "A" failure at five second time interval frame. This component failure requires the Crew to start ECCS equipment per EMG E
-0 Attachment F.
- 2. Event 6: Main Generator and Exciter breakers fail to automatically trip. This component failure requires the BOP to "permit" MA HS
-5, SWYD 345
-50/60 MAN TRIP PERMIT switch, BEFORE opening the breakers per EMG step 6RNO. (NOTE: MA HS
-5, SWYD 345-50/60 MAN TRIP PERMIT is a new switch added to Panel RL005 during Refuel 18).
FINAL 3 of 4 Scenario Critical Tasks (CT)
Event 1: CT
- place rods to manual prior to actuation of the Reactor Protection System Event 2 - CT - take manual control, select alternate controlling channel prior to actuation of the Reactor Protection System Event 4 - CT - isolate the faulted Steam Generator before an Orange path integrity challenge develops FINAL 4 of 4 Probabilistic Risk Analysis for this scenario includes:
Core Damage Frequency by Initiating Event Initiating Event Initiating Event Frequency (/yr)
Core Damage Frequency (/yr)
CDF Percent Contribution Loss of Offsite Power 2.88E-02 6.59E-06 36.51% Small LOCA 3.00E-03 5.35E-06 29.63% Interfacing Systems LOCA 1.93E-06 10.69% Very Small LOCA 6.20E-03 1.27E-06 7.03% Transients With Power Conversion Systems Available 1.05E+00 9.88E-07 5.47% Steam Generator Tube Rupture 3.67E-03 8.77E-07 4.86% Reactor Vessel Failure 3.00E-07 3.00E-07 1.66% Steamline Break 1.13E-02 1.88E-07 1.04% Transients Without Power Conversion Systems Available 1.15E-01 1.71E-07 0.95% Medium LOCA 6.10E-05 1.46E-07 0.81% Loss of All Service Water 6.86E-06 8.30E-08 0.46% Loss of Component Cooling Water 2.14E-04 5.79E-08 0.32% Loss of Vital DC Bus NK04 2.64E-03 4.32E-08 0.24% Large LOCA 7.20E-06 2.80E-08 0.16% Feedwater Line Break 3.17E-03 2.06E-08 0.11% Loss of Vital DC Bus NK01 2.6 4E-03 1.12E-08 0.06% Technical Specifications exercised:
Event 1 - TS determined and entered. TS 3.3.1, Table 3.3.1
-1, Fu 6 and 7, Cond A (Immediately) and Cond E (72 hrs to trip bistables)
Event 2 - TS determined and entered. TS 3.3.1, Table 3.3.1
-1, Fu 14, Cond A (Immediately) and Cond E (72 hrs to trip bistables)
TS 3.3.2, Table 3.3.2
-1, Fu 5.c and 6.d, Cond A (Immediately), Cond I (72 hrs to trip bistable) and Cond D (72 hrs to trip bistable)
Event 3 - TS determined and entered. TS 3.4.11 Cond. B.1 (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> close seal valve) and B.2 (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to de
-energize seal valve) and B.3 (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to repair PORV)
FINAL 1 of 4 Appendix D Scenario Outline Form ES-D-1 Facility: ______Wolf Creek____________ Scenario No.: ____3____
Op-Test No.: _______
Examiners: ____________________________ Operators:
_____________________________
____________________________
_____________________________
____________________________
_____________________________
Initial Conditions: BOL -~10% power Turnover: Power ascension in progress, negative MTC. Perform step 6.40 through step 6.46 of GEN 00-003, HOT STANDBY TO MINIMUM LOAD. Use SYS AC
-120, MAIN TURBINE GENERATOR STARTUP to synchronize the Main Generator to the grid. Increase power to ~15% immediately after synchronizing Main Turbine Generator to the grid.
Event No. Malf. No. Event Type* Event Description 1 N - CREW Per GEN 00
-003, HOT STANDBY TO MINIMUM LOAD, from step 6.40 through step 6.46. Step 6.40 directs SYS AC
-120, MAIN TURBINE GENERATOR STARTUP (synchronize Main Generator to grid).
GEN 00-003 steps 6.41 through 6.46: valve alignments, increase turbine load using load potentiometer, verify Permissive states etc.
2 mNN02 C - CREW Loss of NN02 (White train)
TS determined and entered. TS 3.3.1
- Protective Interlocks i n correct state; Table 3.3.1
-1, Fu 18; Cond A (Immediately); Cond T for P-7, P-8, P-9 and P-13 (Verify interlock in required state within one hour); Cond S for P
-10 (Verify interlock in required state within one hour).
TS 3.8.7, Cond A (restore to operable within twenty four hours)
TS 3.8.9, Cond C (restore to operable status within two hours) 3 mAB07 G C - BOP, SRO Atmospheric Relief Valve (ARV) "C" fails PARTIALLY open; manual control unavailable TS determined and entered. TS 3.7.4 Cond A (restore to operable within seven days) 4 Precursor: Seismic event Main Feed Pump trip Reactor trip 5 mSF17A mSF17B M - CREW Reactor fails to trip in automatic or manual. Anticipated Transient Without Trip (ATWT) 6 mAC02B C - BOP, SRO Post trip malfunction #1: Turbine will not manually trip.
7 p01024 C C - ATC, SRO Post trip malfunction #2: BG HV
-8104 does not open (see step 6 of EMG FR-S1) RNO performed: aligns RWST to charging pump suction * (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor FINAL 2 of 4 Scenario summary:
Unit is at ~ 10 % power, beginning of life. Power ascension in progress, negative MTC. Perform step 6.40 through step 6.46 of GEN 00
-003, HOT STANDBY TO MINIMUM LOAD. Use SYS AC
-120, MAIN TURBINE GENERATOR STARTUP to synchronize the Main Generator to the grid. Increase power to ~15% immediately after synchronizing Main Turbine Generator to the grid.
Event 1: The Crew, using GEN 00-003, HOT STANDBY TO MINIMUM LOAD, from step 6.40 through step 6.46 will synchronize Main Generator to the grid, verify valve alignments, increase turbine load using load potentiometer, and verify Permissive states etc.
Event 2: Loss of NN02 occurs. White train meter indications change and many Main Control Board alarms annunciate aid in diagnosing the component failure. The Crew may enter ALR 00
-026A, NN02 INST BUS UV. The Crew enters OFN NN
-021, LOSS OF VITAL 120VAC INSTRUMENT BUS, and performs Attachment B to restore power.
The Control Room Supervisor determines Technical Specifications.
Event 3: Atmospheric Relief Valve (ARV) "C" fails PARTIALLY open and manual control is unavailable. The Crew enters OFN AB
-041, STEAMLINE OR FEEDLINE BREAK to mitigate the component failure. An Operator is dispatched to locally close the valve.
The Control Room Supervisor determines Technical Specifications.
Event 4: A Seismic event occurs. Main Control Board alarms 00
-098D, OBE and 00
-098E, SEISMIC RECORDER ON, annunciate. OFN SG
-003, NATURAL EVENTS, is entered. The only running Main Feed Pump trips three minutes later. Main Control Board alarm 00
-123A, MFP B TRIP, annunciates. The Crew determines a Reactor trip is necessary. A Reactor trip condition occurs; only the reactor fails to trip.
Event 5: The Main Event is an Anticipated Transient Without Trip (ATWT).
The Crew enters EMG E
-0, REACTOR TRIP OR SAFETY INJECTION, and from step 1RNO transitions to EMG FR
-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS.
Event 6: The turbine will not trip manually
- the BOP must manually trip the turbine within thirty seconds to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require (2RNO EMG FR
-S1 and EMG E
-0).
Event 7: When aligning emergency boration, BG HV
-8104 does not open. The ATC aligns Refueling Water Storage Tank to charging pump suction instead (6RNO of EMG FR
-S1).
Successful mitigation strategy requires the Crew continues performance of EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS.
FINAL 3 of 4 Post trip malfunction:
- 1. Event 6: Post trip malfunction #1: The turbine will not trip manually.
As part of Immediate Actions of EMG FR
-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, step 2RNO, the BOP must trip the turbine.
- 2. Event 7: Post trip malfunction #2: EMER BORATE TO CHG PUMP SUCT BG HIS
-8104 does not open (see step 6 of EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS). RNO performed: aligns RWST to charging pump suction.
Scenario Critical Tasks (CT):
Event 5: CT: Insert negative reactivity into the core by at least one of the following methods before the Steam Generators dry
-out: De-energize the control rod drive MG sets Manually insert control rods
Event 6: CT: Manually trip the turbine to prevent an uncontrolled cooldown of the RCS due to steam flow that the turbine would require.
FINAL 4 of 4 Probabilistic Risk Analysis for this scenario includes:
Core Damage Frequency by Event Tree Event Tree Core Damage Frequency (/yr)
Percent Contribution Station Blackout 6.46E-06 35.79% Small LOCA 5.35E-06 29.65% Interfacing Systems LOCA 1.93E-06 10.68% Very Small LOCA 1.27E-06 7.05% Steam Generator Tube Rupture 8.77E-07 4.86% Loss of Reactor Coolant Pump Seal Cooling Following a Transient Initiator 5.91E-07 3.28% Transients With Power Conversion Systems Available 3.30E-07 1.83% Reactor Vessel Failure 3.00E-0 7 1.66% Steamline Break 1.88E-07 1.40% Transients Without Power Conversion Systems Available 1.71E-07 0.95% Medium LOCA 1.46E-07 0.81% Loss of All Service Water 8.30E-08 0.46% Anticipated Transient Without Scram 6.67E-08 0.37% Loss of Component Cooling Water 5.79E-08 0.32% Loss of Offsite Power 4.98E-08 0.28% Loss of Reactor Coolant Pump Seal Cooling With At Least One CCW Train Available 5.03E-08 0.28% Loss of Vital DC Bus NK04 4.32E-08 0.24% Large LOCA 2.80E-08 0.16% Feedwater Line Break 2.06E-0 8 0.11% Stuck Open Pressurizer PORV Following a Transient Initiator 3.14E-08 0.17% Loss of Vital DC Bus NK01 1.12E-08 0.06% Technical Specifications exercised:
Event 2: TS determined and entered. TS 3.3.1
- Protective Interlocks in correct state; Table 3.3.1-1, Fu 18; Cond A (Immediately); Cond T for P
-7, P-8, P-9 and P-13 (Verify interlock in required state within one hour); Cond S for P
-10 (Verify interlock in required state within one hour). TS 3.8.7, Cond A (restore to operable within twenty four hours) TS 3.8.9, Cond C (restore to operable status within two hours)
Event 3: TS determined and entered. TS 3.7.4 Cond A (restore to operable within seven days)