ML14059A056
| ML14059A056 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 02/14/2014 |
| From: | Mcneil D Operations Branch III |
| To: | Stewart B Entergy Nuclear Vermont Yankee |
| Jackson D | |
| Shared Package | |
| ML13204A095 | List: |
| References | |
| TAC U01897 | |
| Download: ML14059A056 (34) | |
Text
ES-401 BWR Examination Outline Form ES-401-1 Facility: Vermont Yankee Date of Exam:
2/10/14 RO K/ A 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.
1 4
4 3
3 3
3 20 4
3 7
Emergency 2
1 1
1 1
1 2
7 1
2 3
Plant Evaluations Tier 5
5 4
4 4
5 27 5
5 10 Totals 1
3 2
2 3
2 2
3 2
3 3
1 26 3
2 5
- 2.
2 1
1 1
1 2
1 1
1 1
1 1
12 0
2 1
3 Plant Systems Tier 4
3 3
4 4
3 4
3 4
4 2
38 5
3 8
Totals
- 3. Generic Knowledge & Abilities 1
2 3
4 10 1
2 3
4 7
3 2
3 2
1 2
2 2
Note
- 1.
Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SAO-only outlines (i.e., except for one category in Tier 3 of the SAO-only outline, the Tier Totals in each KIA category shall not be less than two).
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by 1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points and the SAO-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 elimination of inappropriate KIA statements.
- 4.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant specific priority, only those KAs having an importance rating (lA) of 2.5 or higher shall be selected. Use the RO and SAO ratings for the RO and SAO-only portions, respectively.
- 6.
Select SAO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
7.*
The generic (G) KIAs in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIA's
- 8.
On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (IR) for the applicable license level, and the point totals (#) for each system and category. Enter the group and tier totals for each category in the table above. If fuel handling equipment is sampled in other than Category A2 or G* on the SAO-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 SAO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the KIA Catalog, and enter the KIA numbers, descriptions, IRs, and point totals (#) on Form ES-401-3. Limit SAO selections to KIAs that are linked to 10CFR55.43 VY RO 1 0/25/13
ES-401 Vermont Yankee BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KJA Topic(s)
AA2.03 - Ability to determine and/or interpret the following 295023 Refueling X
as they apply to 3.8 76 Accidents I 8 REFUELING ACCIDENTS:
Airborne contamination levels AA2.05 -Ability to determine and/or interpret the following 295018 Partial or Total X
as they apply to PARTIAL 2.9 77 Loss of CCW I 8 OR COMPLETE LOSS OF COMPONENT COOLING WATER : System pressure AA2.03 - Ability to determine 600000 Plant Fire On-site I X
and interpret the following as 3.2 78 8
they apply to PLANT FIRE ON SITE: Fire alarm 2.4.20 - Emergency Procedures I Plan:
295005 Main Turbine X
Knowledge of operational 4.3 79 Generator Trip I 3 implications of EOP warnings, cautions, and notes.
2.4.34 - Emergency Procedures I Plan:
295024 High Drywell Knowledge of RO tasks X
performed outside the main 4.1 80 Pressure I 5 control room during an emergency and the resultant operational effects.
2.4.31 - Emergency 295028 High Drywell Procedures I Plan:
X Knowledge of annunciator 4.1 81 Temperature I 5 alarms, indications, or response procedures.
EA2.01 - Ability to determine and/or interpret the following 295037 SCRAM Conditions as they apply to SCRAM Present and Reactor Power X
CONDITION PRESENT 4.3 82 Above APRM Downscale or AND REACTOR POWER Unknown I 1 ABOVEAPRM DOWNSCALE OR UNKNOWN : Reactor power VY RO 1 0/25/13
ES-401 Vermont Yankee BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KJA Topic(s)
AK1.01 -Knowledge of the operational implications of 295021 Loss of Shutdown X
the following concepts as 3.6 39 Cooling I 4 they apply to LOSS OF SHUTDOWN COOLING :
Decay heat AK1.05-Knowledge of the operational implications of 295003 Partial or Complete the following concepts as X
they apply to PARTIAL OR 2.6 40 Loss of AC I 6 COMPLETE LOSS OF A.C.
POWER : Failsafe component design EK1.02 - Knowledge of the operational implications of the following concepts as 295026 Suppression Pool X
they apply to 3.5 41 High Water Temp. I 5 SUPPRESSION POOL HIGH WATER TEMPERATURE : Steam condensation EK2.04 - Knowledge of the interrelations between 295037 SCRAM Conditions SCRAM CONDITION Present and Reactor Power X
PRESENT AND REACTOR 4.4 42 Above APRM Downscale or POWER ABOVE APRM Unknown I 1 DOWNSCALE OR UNKNOWN and the following: SBLC svstem EK2.11 - Knowledge of the 295025 High Reactor interrelations between HIGH X
REACTOR PRESSURE and 3.5 43 Pressure I 3 the following: Reactor water level AK2.07 - Knowledge of the interrelations between 700000 Generator Voltage GENERATOR VOLTAGE and Electric Grid X
AND ELECTRIC GRID 3.6 44 Disturbances DISTURBANCES and the following: Turbine/generator control.
AK3.01 - Knowledge of the reasons for the following 295006 SCRAM I 1 X
responses as they apply to 3.8 45 SCRAM : Reactor water level response VY RO 1 0/25/13
ES-401 Vermont Yankee BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KJA Topic(s)
EK3.06 - Knowledge of the reasons for the following 295028 High Drywell X
responses as they apply to 3.4 46 Temperature I 5 HIGH DRYWELL TEMPERATURE:ADS AK3.02 - Knowledge of the reasons for the following 295016 Control Room X
responses as they apply to 3.7 47 Abandonment I 7 CONTROL ROOM ABANDONMENT : Turbine trip AA 1.02 - Ability to operate and/or monitor the following 295018 Partial or Total X
as they apply to PARTIAL 3.3 48 Loss of CCW I 8 OR COMPLETE LOSS OF COMPONENT COOLING WATER : System loads EA 1.13 - Ability to operate and/or monitor the following 295031 Reactor Low Water X
as they apply to REACTOR 4.3 49 Level I 2 LOW WATER LEVEL :
Reactor water level control AA 1.07 - Ability to operate and/or monitor the following 295001 Partial or Complete as they apply to PARTIAL OR COMPLETE LOSS OF 3.1 50 Loss of Forced Core Flow X
FORCED CORE FLOW Circulation I 1 & 4 CIRCULATION : Nuclear boiler instrumentation system EA2.02 - Ability to determine and/or interpret the following 295030 Low Suppression as they apply to LOW X
SUPPRESSION POOL 3.9 51 Pool Water Level I 5 WATER LEVEL:
Suppression pool temperature AA2.07 - Ability to determine 295005 Main Turbine and/or interpret the following X
as they apply to MAIN 3.5 52 Generator Trip I 3 TURBINE GENERATOR TRIP : Reactor water level VY RO 1 0/25/13
ES-401 Vermont Yankee BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE#/Name Safety Function KIA Topic(s)
AA2.01 - Ability to determine and/or interpret the following 295019 Partial or Total as they apply to PARTIAL X
OR COMPLETE LOSS OF 3.5 53 Loss of lnst. Air I 8 INSTRUMENT AIR:
Instrument air system
_Qressure 2.2.42 - Equipment Control::
295004 Partial or Total Ability to recognize system X
parameters that are entry-3.9 54 Loss of DC Pwr I 6 level conditions for Technical Specifications.
2.2.38 - Equipment Control:
295038 High Off-site X
Knowledge of conditions and 3.6 55 Release Rate I 9 limitations in the facility license.
AK2.04 - Knowledge of the interrelations between 600000 Plant Fire On-site I X
PLANT FIRE ON SITE and 2.5 56 8
the following: Breakers, relays, and disconnects EK1.01 - Knowledge of the operational implications of 295024 High Drywell the following concepts as X
they apply to HIGH 4.1 57 Pressure I 5 DRYWELL PRESSURE :
Drywell integrity: Plant-Specific 2.4.45 - Emergency Procedures I Plan: Ability to 295023 Refueling X
prioritize and interpret the 4.1 58 Accidents I 8 significance of each annunciator or alarm.
KIA CategoryTotals 4
4 3
3 314 313 Group Point Total:
ES-401 Vermont Yankee BWR Examination Outline Form ES-401 -1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KJA Topic(s)
AA2.01 -Ability to determine and/or interpret the following 295012 High Drywell X
as they apply to HIGH 3.9 83 Temperature I 5 DRYWELL TEMPERATURE
- Drywell temperature 2.4.21 - Emergency Procedures I Plan:
Knowledge of the parameters and logic used to assess the status of 295035 Secondary X
safety functions, such as 4.6 84 Containment High reactivity control, core Differential Pressure I 5 cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
2.2.37 - Equipment Control:
295020 Inadvertent Cont.
Ability to determine X
operability and I or 4.6 85 Isolation I 5 & 7 availability of safety related equipment.
AK1.02-Knowledge of the operational implications of the following concepts as 295017 High Off-site X
they apply to HIGH OFF-3.8 59 Release Rate I 9 SITE RELEASE RATE :
Protection of the general public EK2.02-Knowledge of the interrelations between HIGH 295032 High Secondary SECONDARY Containment Area X
CONTAINMENT AREA 3.6 60 Temperature I 5 TEMPERATURE and the following: Secondary containment ventilation AK3.01 - Knowledge of the reasons for the following 295013 High Suppression responses as they apply to 61 X
HIGH SUPPRESSION 3.6 Pool Temperature I 5 POOL TEMPERATURE:
Suppression pool cooling operation VY RO 1 0/25/13
ES-401 Vermont Yankee BWR Examination Outline Form ES-401-1 Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE#/Name Safety Function KIA Topic(s)
AA 1.03 - Ability to operate 295007 High Reactor and/or monitor the following Pressure /3 X
as they apply to HIGH 3.4 62 REACTOR PRESSURE :
RCIC: Plant-Specific EA2.02 - Ability to determine and/or interpret the following 295033 High Secondary as they apply to HIGH Containment Area X
SECONDARY 3.1 63 Radiation Levels I 9 CONTAINMENT AREA RADIATION LEVELS :
Equipment operability 2.4.31 - Emergency 295012 High Drywell Procedures I Plan:
X Knowledge of annunciator 4.2 64 Temperature /5 alarms, indications, or response procedures.
2.4.1 - Emergency 295020 Inadvertent Cont.
Procedures I Plan:
Isolation /5 & 7 X
Knowledge of EOP entry 4.6 65 conditions and immediate action steps.
KIA CategoryTotals 1
1 1
1 1/1 212 Group Point Total:
ES-401 System #/Name 223002 PC IS/Nuclear Steam Supply Shutoff 261000 SGTS 205000 Shutdown Cooling 206000 HPCI VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KJA Topic(s)
A2.01 - Ability to (a) predict the impacts of the following on the PRIMARY CO NT AI NMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-X OFF ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C.
electrical distribution failures A2.1 0 - Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ;
and (b) based on those X
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low reactor water level: Plant-Specific 2.1.27 - Conduct of X
Operations: Knowledge of system purpose and I or function.
2.4.2 - Emergency Procedures I Plan:
Knowledge of system set X
points, interlocks and automatic actions associated with EOP entry conditions.
3.5 86 3.2 87 4.0 88 4.6 89
ES-401 System #/Name 400000 Component Cooling Water 261000 SGTS X
206000 HPCI X
400000 Component X
Cooling Water 212000 RPS X
262002 UPS (AC/DC)
X VY RO 1 0/25/13 Form ES-401 -1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)
A2.04 - Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use X
procedures to correct, control, or mitigate the consequences of those abnormal operation:
Radiation monitoring system alarm K1.07 - Knowledge of the physical connections and/or cause-effect relationships between STANDBY GAS TREATMENT SYSTEM and the following:
Elevated release stack K1.06-Knowledge of the physical connections I"
and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: Suppression chamber: BWR-2,3,4 K2.01 - Knowledge of electrical power supplies to the following: CCW pumps K2.02 - Knowledge of electrical power supplies to the following: Analog trip system logic cabinets K3.02 - Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Recirculation pump speed: Plant-Specific 3.0 90 3.1 1
3.7 2 2.9 3 2.7 4 2.9 5
ES-401 System #/Name 223002 PC IS/Nuclear X
Steam Supply Shutoff 217000 RCIC 215005 APRM I LPRM 300000 Instrument Air VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KJA Topic(s)
K3.21 - Knowledge of the effect that a loss or r
malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Traversing in-core probe system K4.06 - Knowledge of REACTOR CORE ISOLATION COOLING X
SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Manual initiation K4.08 - Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or X
interlocks which provide for the following:
Sampling of overall core power in each APRM (accomplished through LPRM assignments and symmetrical rod patterns)
K5.13 - Knowledge of the operational implications of the X
following concepts as they apply to the INSTRUMENT AIR SYSTEM: Filters 2.6 6 3.5 7 2.7 8 2.9 9
ES-401 System #/Name 211000 SLC 259002 Reactor Water Level Control 239002 SRVs 264000 EDGs VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 K!A Topic(s)
K5.06-Knowledge of the operational implications of the X
following concepts as they apply to STAND BY LIQUID CONTROL SYSTEM : Tank level measurement K6.05 - Knowledge of the effect that a loss or malfunction of the X
following will have on the REACTOR WATER LEVEL CONTROL SYSTEM : Reactor water level input K6.02 - Knowledge of the effect that a loss or malfunction of the X
following will have on the RELIEF/SAFETY VALVES : Air (Nitrogen) supply: Plant-Soecific A 1.09 - Ability to predict and/or monitor changes in parameters associated with operating the X
EMERGENCY GENERATORS (DIESEUJET) controls including: Maintaining minimum load on emergency generator (to prevent reverse power) 3.0 10 3.5 11 3.4 12 3.0 13
ES-401 System #/Name 205000 Shutdown Cooling 203000 RHR/LPCI:
Injection Mode 262001 AC Electrical Distribution 21 8000 ADS VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)
A 1.06 - Ability to predict and/or monitor changes in parameters associated with operating the SHUTDOWN COOLING X
SYSTEM(RHR SHUTDOWN COOLING MODE) controls including: Reactor temperatures (moderator, vessel, flange)
A2.11 - Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC); and (b)
X based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Motor operated valve failures A2.09-Ability to (a) predict the impacts of the following on the A. C.
ELECTRICAL DISTRIBUTION; and (b) based on those X
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Exceeding voltage limitations A3.05 - Ability to monitor automatic operations of X
the AUTOMATIC DEPRESSURIZATION SYSTEM including:
Suppression pool level 3.7 14 3.4 15 3.1 16 3.6 17
ES-401 System #/Name 263000 DC Electrical Distribution 2150031RM 215004 Source Range Monitor 209001 LPCS 223002 PC IS/Nuclear Steam Supply Shutoff 211000 SLC 263000 DC Electrical Distribution VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KJA Topic(s)
A3.01 - Ability to monitor automatic operations of the D.C. ELECTRICAL X
DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating liahts A4.06 - Ability to X
manually operate and/or monitor in the control room: Detector drives A4.07 - Ability to manually operate and/or X
monitor in the control room: Verification of proper functioning/
operability A3.05 - Ability to monitor automatic operations of X
the LOW PRESSURE CORE SPRAY SYSTEM including: Reactor water level 2.4.11 - Emergency X
Procedures I Plan:
Knowledge of abnormal condition procedures.
A 1.02 - Ability to predict and/or monitor changes in parameters associated with X
operating the STAND BY LIQUID CONTROL SYSTEM controls including: Explosive valve indication K4.02 - Knowledge of D.C. ELECTRICAL DISTRIBUTION design feature(s) and/or X
interlocks which provide for the following: Breaker interlocks, permissives, bypasses and cross ties:
Plant-Specific 3.2 18 3.0 19 3.4 20 3.9 21 4.0 22 3.8 23 3.1 24
ES-401 System #/Name 262001 AC Electrical X
Distribution 262002 UPS (AC/DC)
KIA Category Totals 3
2 2
VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 1 KIA Topic(s)
K1.02 - Knowledge of the physical connections and/or cause-effect relationships between A.C. ELECTRICAL DISTRIBUTION and the following: D.C. electrical distribution A4.01 - Ability to manually operate and/or X
monitor in the control room: Transfer from alternative source to preferred source 3
2 2
3 2/3 3
3 1/2 Group Point Total:
3.3 25 2.8 26 1 26/5
ES-401 System #/Name 201002 RMCS 239001 Main and Reheat Steam 219000 RHR/LPCI:
Torus/Pool Cooling Mode 226001 RHR/LPCI:
X CTMT Spray Mode VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 2 KJA Topic(s)
A2.01 - Ability to (a) predict the impacts of the following on the REACTOR MANUAL CONTROL SYSTEM ;
and (b) based on those X
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Rod movement sequence timer malfunctions 2.1.23 - Conduct of Operations: Ability to perform specific system X
and integrated plant procedures during all modes of plant operation.
A2.06 - Ability to (a) predict the impacts of the following on the RHR/LPCI:
TORUS/SUPPRESSION POOL COOLING MODE X
- and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations
- D.C.
electrical failures K1.12 - Knowledge of the physical connections and/or cause-effect relationships between RHR/LPCI:
CONTAINMENT SPRAY SYSTEM MODE and the following: Suppression pool (spray penetration):
Plant-Specific.
2.8 91 4.4 92 2.9 93 3.0 27
ES-401 System #/Name 256000 Reactor X
Condensate 201002 RMCS X
204000 RWCU 268000 Radwaste 288000 Plant Ventilation 234000 Fuel Handling Equipment VY RO 1 0/25/13 Form ES-401 -1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 2 KJA Topic(s) 1<2.01 - Knowledge of electrical power supplies to the following: System pumps K3.02 - Knowledge of the effect that a loss or malfunction of the REACTOR MANUAL CONTROL SYSTEM will have on following: Rod block monitor: Plant-Specific K4.03 - Knowledge of REACTOR WATER CLEANUP SYSTEM X
design feature(s) and/or interlocks which provide for the following: Over temperature protection for system components K5.02 - Knowledge of the operational implications of the X
following concepts as they apply to RADW ASTE : Radiation hazards and ALARA concept K6.02 - Knowledge of I'
the effect that a loss or malfunction of the X
following will have on the PLANT VENTILATION SYSTEMS : Applicable component cooling water system: Plant-Soecific A 1.03 - Ability to predict and/or monitor changes in parameters associated with X
operating the FUEL HANDLING EQUIPMENT controls including: core reactivity level 2.7 28 2.9 29 2.9 30 3.1 31 2.5 32 3.4 33
ES-401 System #/Name 230000 RHR/LPCI:
Torus/Pool Spray Mode 215002 RBM 290003 Control Room HVAC 290001 Secondary CTMT 201001 CRD Hydraulic KIA Category Totals 1
1 1
VY RO 1 0/25/13 Form ES-401-1 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 2 KIA Topic(s)
A2.12 - Ability to (a) predict the impacts of the following on the RHRILPCI:
TORUS/SUPPRESSION POOL SPRAY MODE ;
X and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve logic failure A3.04 - Ability to monitor automatic operations of the ROD BLOCK X
MONITOR SYSTEM including: Verification or proper functioning/
operability: BWR-3,4,5 A4.01 - Ability to manually operate and/or X
monitor in the control room: Initiate/reset system 2.2.44 - Equipment Control: Ability to interpret control room indications to verify the X
status and operation of a system, and understand how operator actions and directives effect plant and system conditions.
K5.02 - Knowledge of the operational implications of the X
following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM :
Flow indication 1
2 1
1 1/2 1
1 1/1 Group Point Total:
3.2 34 3.6 35 3.2 36 4.2 37 2.6 38 1 12/3
ES-401 System #/Name VY RO 1 0/25/13 Vermont Yankee BWR Examination Outline Plant Systems - Tier 2 Group 2 Form ES-401-1 KIA Topic(s)
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
Vermont Yankee Date:
2/10/14 Category KA#
Topic RO SRO-Only IR Q#
IR Q#
2.1.1 Knowledge of conduct of operations 3.8 66 requirements.
2.1.25 Ability to interpret reference materials, 3.9 67 such as graphs, curves, tables, etc.
Ability to identify and interpret diverse
- 1. Conduct of 2.1.45 indications to validate the response of 4.3 75 another indicator.
Operations 2.1.6 Ability to manage the control room crew 4.8 94 during plant transients.
Subtotal 3
1 Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated 4.5 68 with plant equipment that could affect reactivity.
Knowledge of less than or equal to one 2.2.39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> technical specification action 3.9 69 statements for systems.
- 2. Equipment Control 2.2.23 Ability to track Technical Specification 4.6 95 limiting conditions for operations.
2.2.19 Knowledge of maintenance work order 3.4 99 requirements.
Subtotal 2
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to 2.3.13 radiation monitor alarms, containment 3.4 70 entry requirements, fuel handling responsibilities, access to locked high radiation areas, ali~ning filters, etc.
Ability to comply with radiation work 2.3.7 permit requirements during normal or 3.5 71 abnormal conditions.
2.3.11 Ability to control radiation releases.
3.8 74
- 3. Radiation Ability to use radiation monitoring Control systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey 2.9 96 instruments, personnel monitoring equipment, etc.
Knowledge of Radiological Safety Principles pertaining to licensed 2.3.12 operator duties, such as containment 3.7 100 entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.
Subtotal 3
2 Knowledge of system set points, 2.4.2 interlocks and automatic actions 4.5 72 associated with EOP entry conditions.
2.4.13 Knowledge of crew roles and 4.0 73 responsibilities during EOP usage.
- 4. Emergency Procedures I Knowledge of the specific bases for Plan 2.4.18 EOPs.
4.0 97 2.4.46 Ability to verify that the alarms are 4.2 98 consistent with the plant conditions.
Subtotal 2
2 Tier 3 Point Total:
10 7
ES-401 Record of Rejected Kl A's Form ES-401-4 Tier I Group Randomly Selected Reason for Rejection KA 2950241 EK1.02 The subject KIA isn't relevant at the subject facility. Mk Ill 111 replaced by 2950241 EK1.01 containment not relevant.
313 G312.3.6 replaced Facility is a zero release plant and SROs are not required to by G312.3.12 approve radioactive releases 312 G212.2.39 replaced Same KA as RO #69. Replaced to avoid context overlap for by G2 12.2.23 SROs 313 G312.3.13 replaced Same KA selected for RO #70. Replaced to avoid context bvG312.3.5 overlap for SROs 234000 I A 1.02 212 replaced by 234000 I Unacceptable overlap with SAO #76.
A1.03 295037 I EK2. 11 The subject KIA isn't relevant at the subject facility. System is not 1 11 replaced by 295037 I EK2.04 used in relation to EPE topic selected.
209001 I 2.4.41 It isn't possible to prepare a psychometrically sound question 211 replaced by 209001 I related to the subject KIA. No emergency classification directly A3.05 related to svstem components.
261000 I A2.08 Unacceptable topic overlap with SAO #93; also excessive 211 replaced by 261000 I number of items referring to DC electrical. (5 topics originally A2.10 selected) 600000 12.4.49 There are no immediate actions or actions taken without 111 replaced by 600000 I references for the selected topic.
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Facility:
Vermont Yankee Date:
2/10/14 Category KA#
Topic RO SRO-Only IR Q#
IR Q#
2.1.1 Knowledge of conduct of operations 3.8 66 requirements.
2.1.25 Ability to interpret reference materials, 3.9 67 such as graphs, curves, tables, etc.
Ability to identify and interpret diverse
- 1. Conduct of 2.1.45 indications to validate the response of 4.3 75 another indicator.
Operations 2.1.6 Ability to manage the control room crew 4.8 94 durinQ plant transients.
Subtotal 3
1 Ability to perform pre-startup procedures for the facility, including 2.2.1 operating those controls associated 4.5 68 with plant equipment that could affect reactivity.
Knowledge of less than or equal to one 2.2.39 hour4.513889e-4 days <br />0.0108 hours <br />6.448413e-5 weeks <br />1.48395e-5 months <br /> technical specification action 3.9 69 statements for systems.
- 2. Equipment Control 2.2.23 Ability to track Technical Specification 4.6 95 limitinQ conditions for operations.
2.2.19 Knowledge of maintenance work order 3.4 99 requirements.
Subtotal 2
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 Knowledge of Radiological Safety Procedures pertaining to licensed operator duties, such as response to 2.3.13 radiation monitor alarms, containment 3.4 70 entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
Ability to comply with radiation work 2.3.7 permit requirements during normal or 3.5 71 abnormal conditions.
2.3.11 Ability to control radiation releases.
3.8 74
- 3. Radiation Ability to use radiation monitoring Control systems, such as fixed radiation 2.3.5 monitors and alarms, portable survey 2.9 96 instruments, personnel monitoring equipment, etc.
Knowledge of Radiological Safety Principles pertaining to licensed 2.3.12 operator duties, such as containment 3.7 100 entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning_ filters, etc.
Subtotal 3
2 Knowledge of system set points, 2.4.2 interlocks and automatic actions 4.5 72 associated with EOP entry conditions.
2.4.13 Knowledge of crew roles and 4.0 73 responsibilities during EOP usage.
- 4. Emergency Procedures I Knowledge of the specific bases for Plan 2.4.18 EOPs.
4.0 97 2.4.46 Ability to verify that the alarms are 4.2 98 consistent with the plant conditions.
Subtotal 2
2 Tier 3 Point Total:
10 7
ES-401 Record of Rejected KIA's Form ES-401-4 Tier I Group Randomly Selected Reason for Rejection KA 295024 I EK1.02 The subject KIA isn't relevant at the subject facility. Mk Ill 1 11 replaced by 295024 I EK1.01 containment not relevant.
3/3 G3 I 2.3.6 replaced Facility is a zero release plant and SROs are not required to by G3 I 2.3.12 approve radioactive releases 312 G2 I 2.2.39 replaced Same KA as RO #69. Replaced to avoid context overlap for by G2 I 2.2.23 SROs 313 8312.3.13 replaced Same KA selected for RO #70. Replaced to avoid context by G312.3.5 overlap for SROs 234000 I A1.02 212 replaced by 234000 I Unacceptable overlap with SRO #76.
A1.03 295037 I EK2.11 The subject KIA isn't relevant at the subject facility. System is not 1 11 replaced by 295037 I EK2.04 used in relation to EPE topic selected.
209001 I 2.4.41 It isn't possible to prepare a psychometrically sound question 211 replaced by 209001 I related to the subject KIA. No emergency classification directly A3.05 related to system components.
261000 I A2.08 Unacceptable topic overlap with SRO #93; also excessive 211 replaced by 261000 I number of items referring to DC electrical. (5 topics originally A2.10 selected) 600000 I 2.4.49 There are no immediate actions or actions taken without 111 replaced by 600000 I references for the selected topic.
AK2.04
--,------**--~-*-"-------------
ES-401 Record of Rejected K/A's Form ES-401-4 Tier / Group Randomly Selected KA Reason for Rejection 1 / 1 295024 / EK1.02 replaced by 295024 /
EK1.01 The subject K/A isn't relevant at the subject facility.
3 / 3 G3 / 2.3.6 replaced by G3 / 2.3.12 Facility is a zero release plant and SROs are not required to approve radioactive releases 3 / 2 G2 / 2.2.39 replaced by G2 / 2.2.23 Same KA as RO #69.
3 / 3 G3 / 2.3.13 replaced by G3 / 2.3.5 Same KA selected for RO #70 2 / 2 234000 / A1.02 replaced by 234000 /
A1.03 Unacceptable overlap with SRO #76 1 / 1 295037 / EK2.11 replaced by 295037 /
EK2.04 The subject K/A isn't relevant at the subject facility.
2 / 1 209001 / 2.4.41 replaced by 209001 /
A3.05 It isn't possible to prepare a psychometrically sound question related to the subject K/A.
2 / 1 261000 / A2.08 replaced by 261000 /
A2.10 Unacceptable topic overlap with SRO #93, and excessive number of items referring to DC electrical 1 / 1 600000 / 2.4.49 replaced by 600000 /
AK2.04 There are no immediate actions or actions taken without references for the selected topic 1 / 1 295028 / EK3.06 replaced by 295028 /
EK3.01 It isn't possible to prepare a psychometrically sound question related to the subject K/A.
1 / 1 295004 / 2.2.42 replaced by 295004 /
2.1.27 It isn't possible to prepare a psychometrically sound question related to the subject K/A.
1 / 1 295038 / 2.2.38 replaced by 295038 /
2.1.30 It isn't possible to prepare a psychometrically sound question related to the subject K/A.
2 / 2 256000 / K2.01 replaced by 259001 /
K2.01 Topic would only yield test item at LOD = 1 1 / 1 295025 / EK2.11 replaced by 295025 /
EK3.02 Changed KA to accurately reflect question developed 1 / 1 295005 / AA2.07 replaced by 295005 AK3.04 Changed KA to accurately reflect question developed 1 / 1 295005 / 2.4.20 replaced by 295024 /
2.4.4 Could not develop psychometrically sound SRO test item with original topic
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Vermont Yankee Examination Level: ROD SAO X Date of Examination: 02/10/2014 Operating Test Number:
NRC Administrative Topic Type (see Note)
Code*
N,S Conduct of Operations Conduct of Operations M, R 2.1.31 (4.3)
JPM:
2.1.36 (4.1)
Describe activity to be performed Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.
Perform a system walkdown following a HPCI pump and valve operability tests.
Knowledge of procedures and limitations involved in core alterations.
JPM:
Review Daily SAM Response Check prior to ConductinQ Core Alteration {29921}
D, R 2.2.12 (4.1)
Surveillance Requirements Equipment Control Radiation Control Emergency Procedures/Plan JPM:
Review Completed Surveillance and Take Action for Out of Spec Data (29904)
P, D, 2.3.7 (3.6)
Ability to comply with radiation work permit requirements during normal or abnormal conditions.
R JPM:
Determine the Radiological Protection Requirements for Entering a Locked High Radiation Area. (201 0 SAO NRC JPM-A3 SAO)
N, R 2.4.29 (4.4)
Knowledge of the emergency plan.
JPM:
Classify a loss of all offsite and onsite AC event while in mode 5. (AP 3125)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when ailS are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (:S 3 for ROs; s 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (<:: 1)
(P)revious 2 exams (s 1; randomly selected)
SRO NRC Admin JPM Summary A 1 a This is a new JPM. The operator will be given a set of initial conditions and told to perform a system walkdown following HPCI pump and valve operability tests. The operator will be provided with the completed surveillance as a reference. There will be a valve out of position and indications of a blown fuse. The operator will then be asked to determine any Technical Specifications and reportability requirements.
A1b This is a modified JPM using the 20012 SAO audit exam. This JPM was modified by changing the readings as well as changing the quadrant. The operator will be given a set of initial conditions and asked to review the results of the Daily Neutron Response Check provided and determine which SRMs are operable and whether the scheduled Core Alteration can commence. The operator will be provided with the appropriate procedures and completed SAM Response Check. The operator will evaluate the data and then inform the Shift Manager of the results.
A2 This is a direct bank JPM (29904) not used in the previous two NRC exams. The operator will be given a completed RHR system surveillance and asked to review the provided surveillance data and sign as Shift Supervisor. The Operator will be given a completed VYOPF 4124.01 form for RHR Loop "A". The operator will review the data as well as technical specifications, and determine the results.
A3 This JPM was used in the 2012 SAO NRC exam (SAO A3) and was randomly selected.
(Refer to the random selection order/JPM matrix). The operator will be given a set of initial conditions and asked to determine the conditions where work will be performed.
The operator will be provided the procedure and the survey map to determine how much dose will be received. The operator evaluates the information and provides the conclusions to the evaluator.
A4 This is a new JPM. This JPM evaluates AC power supplies while in Mode 5. The initial conditions will address Local Intense Precipitation and a area wide blackout, flooding in the switchgear rooms, and fault conditions on Emergency Buses. The operator will assess conditions and classify the event.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Vermont Yankee Date of Examination: 02/10/2014 Exam Level: RO D SR0-1 D SRO-U X Operating Test No.: NRC 2014 Control Room Systems0 (8 for RO); (7 for SR0-1); (2 or 3 for SRO-U, including 1 ESF)
System I JPM Title Type Code*
Safety Function
- a. 245000 Main turbine generator and Auxiliary Systems S,M,A 4
Startup the Turbine to Rated Speed (HI vibration) (24506)
- b. 239001 Main and Reheat Steam System S,EN,D 3
Open MSIVs After a Group I Isolation (2001 0)
- c. 206000 High Pressure Coolant Injection S,D,EN,A 2
Manually start HPCI and Inject to vessel
- d.
- e.
- f.
- g.
- h.
In-Plant Systems0 (3 for RO); (3 for SR0-1); (3 or 2 for SRO-U)
I. 211000 Standby Liquid Control System D,A,E,R 1
Perform Local Firing of squib Valves (Valve falls to fire)
(20017F)
- j. 212000 Reactor Protection System D
7 Startup the "A" RPS MG set (21202, A0-010)
- k.
All RO and SR0-1 control room (and in-plant) systems must be different and serve different safety functions; all 5 SRO-U systems must serve different safety functions; in-plant systems and functions may overlap those tested in the control room.
- Type Codes Criteria for RO I SR0-1 I SRO-U (A)Iternate path 4-6 I 4-6 I 2-3 (C)ontrol room (D)irect from bank S91:S81S4 (E)mergency or abnormal in-plant
<!:11<:11<!:1 (EN)gineered safety feature
- I - I <:1 (control room system)
(L}ow-Power I Shutdown
<!:11<!:11<!:1 (N)ew or (M)odified from bank including 1 (A)
<!:21<!:21<!:1 (P)revious 2 exams s 3 Is 3 Is 2 (randomly selected)
(R)CA
<!:11<::11<!:1 (S)imulator SAO NRC CR-IP JPM Summary A{S)
This is a modified JPM using bank JPM (24506) as its basis. The bank JPM was modified by adding turbine vibrations requiring a manual turbine trip. The operator will be given a set of initial conditions with a startup is in progress and the turbine is on the turning gear, and told to startup the Main Turbine to rated speed. During the startup, the turbine will experience high vibrations requiring the operator to stop the startup, and ultimately manually tripping the turbine.
B(S)
This is a direct bank JPM (2001 0) and has not used in the previous two NRC exams. The operator will be placed in a condition where a group 1 isolation has occurred due to condenser backpressure. The signals have cleared I bypassed and the operator will have to open the MSIVs.
C(S)
This is a direct bank JPM (20601 F) and has not been used in the previous two NRC Exams. The operator will manually start HPCI and inject to the vessel. A failure of the HPCI flow controller will require the operator to take manual control and re-establish flow.
I(IP)
This is a direct bank JPM (20017F) and has not used in the previous two NRC exams. The operator will be given a set of initial conditions where a scram condition exists, but the reactor is not shutdown. Torus temperature is approaching 110°F and normal squib valve firing did not function. The operator will then have to fire the squib valve locally. The first attempt will fail, and the operator will have to attempt local firing on a different terminal.
J(IP)
This is a direct bank JPM (21202) not used in the previous two NRC exams. The operator will be given a set of initial conditions where the "A" RPS MG set is being returned to service after brush replacement. The operator will then be asked to start up the "A" RPS MG set and place it in service.
I Appendix D Scenario Outline Form ES-D-1 Facility:
Vermont Yankee Scenario No.:
1 Op Test No.:
2014 NRC Examiners:
Operators:
Initial Conditions:
100% MOL Steady State conditions. SBO EDG is out of service (tagged out) for Electrical Maintenance trouble shooting of the voltage regulator. A & F APRM's are bypassed. House Heating Boilers in service.
Turnover:
Perform Turbine Oil Pump Test in accordance with OP 4160 section 8 Critical Tasks: See scenario summary Event Malt. No.
Event No.
Type*
1 mfRD12A 35%
2 mfRD12B 40%
3 mfRP _09 A&B Partial SCRAM Partial SCRAM Event Description PCIS Group V Isolation Failure ll------1--------------~f.--------------------~--*--------ll 4
N/A N-BOP Perform Turbine Oil Pump Test lAW OP 4160 N-CRS r-----***------*-----t----*--+------------*---------------11 5
mfRD __ 15 1-RO Control Rod Drive flow controller fails low 0%
1-CRS f-**--*--+-------+-----lf--------------*----------*----------!1 6
mfAD __ 09B TS-SRO RV-?1 B de-energized r----------1--------- -----r-----------*-----------------*----------
7 mfRR_07A C-BOP A Recirc Pump# 1 Seal Failure 50% over C-SRO 300 sec.
-*----1-*------- '-----*- -*-----*----------------.. --*---*-----*------
8 rnfRR.08A C-BOP A Recirc Pump# 2 Seal Failure i 5% over I C-SRO
-9~-t::
5~~-lR-ALL __.. -Rapid -~~-;d--;:-~;;~;;t;-;~--single L~~P-Op~;*ati~n (TS} ____________ _
~~t<i::_~. c<TJrRfi:.O~s L~~~=.~ir"':~t~P=~~ Tni:~-==*=-~-~==-~=:~==-~=~=~
Appendix D NUREG 1 Revision 9
I Appendix D Scenario Outline Form ES-D-1 I 11 mfSL_01 A Operating SLC Pump Trips 60 seconds after start orB (N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario Summary The crew assumes the watch with the unit at 1 00% power. SBO EDG is out of service (tagged out) for Electrical Maintenance trouble shooting of the voltage regulator. A & F APRM's are bypassed. House Heating Boilers in service. The SAO will direct the BOP to Perform Turbine Oil Pump Test in accordance with OP 4160 section B.
After the performance of the Turbine Oil Pump test, the Control Rod Drive flow controller will fail low resulting in the in-service FCV throttling close. The RO will respond to typer alarms. Review CAD indications and diagnose the controller failure. The SAO will enter OPON-3145-01 Loss of CAD Regulating Function and direct the RO to place the CAD flow controller in manual and restore CAD flow.
Once CAD flow has been stabilized, the BOP will respond to annunciator 3-A-4, ADS Power Failure and determine that the position indicating lights for RV-71B are extinguished. The SAO will review Technical Specifications, determine that RV-71 B is inoperable and enter TS LCO 3.5.F.2.
As a result of the CAD flow controller failure, the A Recirc Pump #1 Seal will fail. The BOP will respond to seal indications and alarms and diagnose the failure of the #1 seal.
The SAO will enter OPON-3142-01 and direct the BOP to increase monitoring of seal parameters. After the commencement of monitoring, the #2 seal will fail requiring the A Recirc Pump to be shut down and the crew to conduct a rapid load reduction lAW OPOT-3118-01 Recirculation Pump Trip. The RO will insert control rods to lower reactor power to approximately 45%. The SAO will review Technical Specifications and enter TS LCO 3.6.G.1 for single loop operation.
After reactor power is reduced to between 40 to 45% and the B Recirculation pump speed is being lowered, the B Recirculation Pump will trip. The crew will recognize the need to manually scram the reactor. Manual scram will result in partial rod insertion and the SAO will direct actions from EOP-1 and EOP-2. With the reactor not shutdown, the SAO will direct the RO to inject boron using the standby liquid control system. The operating SLC pump will trip after 60 seconds and the RO will diagnose failure and start the remaining SLC Pump to re-establish boron injection (Critical Task). PCIS Group V valves (RWCU-15, 18, and 68) will fail to isolate. The RO will diagnose the failure and manually close the valves.
During the ATWS, conditions will be met to perform power/level control TERMINATE AND PREVENT INJECTION into the RPV using appendix GG, until conditions are met to re-establish injection (Critical Task). The SAO will direct the implementation of OE 3107, Appendix BB to insert control rods using cooling water differential pressure.
The scenario will be terminated when all rods are inserted and RPV level is restored.
Appendix D NUREG 1021 Revision 9
[Appendix D Scenario Outline Form ES-D-1 Facility:
Vermont Yankee Scenario No.:
2 Op Test No.:
2014 NRC Examiners:
Operators:
Initial Conditions:
80% reactor power, MOL. Power was reduced to allow maintenance to evaluate A Main Feedwater Pump seal leakage. A main feedwater pump is in service. SBO EDG is out of service (tagged out) for Electrical Maintenance trouble shooting of the voltage regulator. A & F APRM's are bypassed. House Heating Boilers in service.
Turnover:
Place Torus Cooling in service in accordance with OPOP-RHR-2124 Residual Heat Removal System.
Section 7.3 Torus Cooling during normal operations.
Critical Tasks: See scenario summary Event Malf. No.
Event Event No.
Type*
Description 1
mfSW_18B Standby Service Water Pump(B) auto start failure preinsert 2
mfPC __ 1HP15 HPCI Steam Supply valves fail to isolate (automatic & manual) mfPC_1HP16 preinsert
--~~-----*--w....,.._, __,,,_, __
3 mfRP_08A PCIS Group 3 failure mfRP __ 08B preinsert 4
N/A N-BOP Establish Torus Cooling N-SRO
1-------
~~-------*-~----~--~---~---
5 mfNM __ 05C 1-RO C APRM Fails downscale 0%
1-SRO TS-SRO r----------[----------------
6 mfSW __ 07A C-BOP Service Water Pump trips; standby pump (B) fails to auto start mfSW_.!8B C-SRO TS-SRO r--------- ------
7
. rnfED __ 06C C-ALL of DC Bus 3 I
TS-SRO AppHndix D NUREG *1 021 ~ievision 9
I Appendix D Scenario Outline F_orm ES-D-1 I 8
mfHP_09 C-ALL HPCI Steam Leak in Reactor Building - small 5% over 300 sec 9
mfHP_09 M-ALL Post Scram - HPCI Steam Leak in Reactor Building - large 50% over 900 sec
{N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Scenario Summary The crew assumes the watch with the unit at 80% power. Reactor power was reduced to allow Maintenance to evaluate A Main Feedwater Pump seal leakage. House Heating Boilers are in service. The SAO will direct the BOP to place Torus Cooling in service in accordance with OPOP-RHR-2124 Residual Heat Removal System. Section 7.3 Torus Cooling during normal operations to reduce torus temperature.
After tours cooling has been established, C APRM will fail downscale. The RO will respond to APRM Downscale and Rod Withdrawal Block alarms and diagnose the failure of C APRM. The SRO will direct the RO to remove A APRM from bypass and place C APRM in bypass. The SAO will determine the impact with 2 APRM's inoperable in the same channel (TS Table 3.1.1 Action 1 ). Once C APRM is bypassed the SRO will exit the APRM TS LCO.
After the Technical Specifications for C ARPM have been addressed, B Service Water Pump will trip and the standby Service Water Pump will fail to auto start. The BOP will diagnose the failures. The SAO will direct the BOP to start the standby Service Water Pump, address actions in ON 3148 'Loss of Service Water', declare the B Service water Pump inoperable and enter TS LCO 3.5.D.2.
A loss of DC Bus 3 will result in a loss of the majority of control room annunciators and various control room indicators. The SRO will enter ON 3161 'Loss of DC-3' and direct the transfer of DC-3A to DC-1 (restores annunciators) and the BOP will place the Auxiliary Oil Pump in pull-to-lock. The SAO will review Technical Specifications, declare DC Bus1 inoperable and enter TS LCO 3.10. B.2.b.
A small steam leak will develop from the HPCI steam supply line in the reactor building.
The fire detection Zone 11 Reactor Building Panel will alarm. The SRO will initially enter OP 3020 'Control Room Response to a Fire' and exit the procedure when the report of a steam leak is received. The SRO will enter ON-3158 'Reactor Building High Area Temperature I Water Level' and direct the BOP to isolate HPCI. The HPCI steam supply valves will fail to isolate automatically and wiH not close manually from the control room.
As area temperatures rise, the SRO will enter EOP-4 due to area temperature and will direct a reactor scram prior to any area in the reactor building exceeding the maximum safe operating level (Critical Task) and enter EOP-*1.
D NUREG 1021 Revision 9
I Appendix D Scenario Outline Form ES-D-1]
The HPCI steam leak will increase after the reactor scram. PCIS Group 3 will fail to actuate requiring the crew to align and start standby gas (Critical Task). Additional attempts to isolate the HPCI steam leak will be unsuccessful requiring the crew to enter EOP-5 and perform an emergency depressurization of the reactor pressure vessel when two reactor building area temperatures exceed the maximum safe operating limits (Critical Tack).
The scenario will be terminated when Reactor Vessel Emergency Depressurization has been achieved and vessel level is being controlled between 127 to 177 inches.
Appendix D NUREG *1 021 Revision 9