ML18088A717
| ML18088A717 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/29/2018 |
| From: | Brian Fuller Operations Branch I |
| To: | Exelon Generation Co |
| Shared Package | |
| ML16210A438 | List: |
| References | |
| CAC U01941 | |
| Download: ML18088A717 (32) | |
Text
Appendix D Scenario Outline Form ES-0-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: LC1 15-1 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 1 % in 5 loop operation during a startup.
ESW Pump 12 is out of service for maintenance.
Turnover: The reactor is critical at approximately 1% power. Perform N1-ST-W15, Manual and Automatic Scram Instrument Channel Test. Continue the startup IAW N1-0P-43A, starting with control rod withdrawals.
Event Malf.
Event Event No.
No.
Type*
Description N-BOP, Perform N1-ST-W15, Sections 6.2 and 6.3 (partial), Channel 11 1
N/A Auto and Manual Scram Tests SRO N1-ST-W15 R-ATC, Withdraw control rods 2
N/A SRO N1-0P-43A, N1-0P-5 C-ATC, Stuck control rod 3
RD04 SRO N1-0P-5, Technical Specifications SRO-TS I-ATC, SRM Fails Upscale 4
NM01C SRO SRO-TS N1-0P-38A, Technical Specifications CW05A Loss of all TBCLC requires scram and MSIV isolation 5
C-AII CW05B N1-SOP-24.1, N1-SOP-1 RP05A I-BOP, RPS fails to scram the reactor, ARI is successful 6
RP05B SRO N1-EOP-2, N1-SOP-1 7
EC02 M-AII EC 11 Steam Leak in the Reactor Building ARP K1-1-1, K1-3-4, N1-SOP-1, N1-EOP-2, N1-EOP-5 EC07A EC 11 will not isolate, requiring a RPV Slowdown 8
EC08A C-All N1-EOP-8 EC08B (N)ormal,
( R )eactivitv, (l)nstrument, (C)omoonent, (Mlaior
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-1 Op-Test No.: LC115-1
- 1. Total malfunctions (5-8) 6 Events 3, 4, 5, 6, 7, 8
- 2. Malfunctions after EOP entry (1-2) 3 Event 6, 7,8
- 3. Abnormal events (2-4) 4 Events 3, 4, 5, 6
- 4. Major transients (1-2) 1 Event7
- 5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
- 6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-8
- 7. EOP Based Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given a complete loss of TBCLC, the crew will insert a manual TBCLC provides cooling to major heat reactor scram in accordance with N1-SOP-24.1.
loads in the plant, including Recirc Pump MG sets and Instrument Air Compressors. Either of which, when lost, greatly reduces the level of safety of the reactor.
CT-2.0: Given conditions requiring a scram and failure of an RPS channel Inserting control rods during a transient to trip, the crew will manually initiate Alternate Rod Insertion (ARI) to lowers reactor power, which reduces shutdown the reactor, in accordance with N1-SOP-1 and/or N1-EOP-3.
challenges to the plant during the transient. With RPS failing to trip, the crew must rely on backup shutdown methods to ensure control rods are inserted to provide long term, stable, core shutdown.
CT-3.0: Given an un-isolable Emergency Condenser leak outside Primary An un-isolable primary system Containment and two general area temperatures above the maximum safe discharging outside of Primary limit, execute N1-EOP-8, RPV Blowdown, in accordance with N1-EOP-5.
Containment resulting in two general area temperatures above the maximum safe limit indicates a wide-spread problem posing a direct and immediate threat to Secondary Containment. A blowdown minimizes flow through the break, rejects heat to the suppression pool in preference to outside the containment, and places the primary system in the lowest possible energy state.
SCENARIO
SUMMARY
The scenario begins at approximately 1% power with a startup in progress. ESW pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking. Control rod withdrawal has been halted for the turnover and to permit the crew to perform N1-ST-W15, Manual Scram Instrument Channel Test, sections 6.1 to 6.3 for Channel 11 only. The surveillance test results will be satisfactory.
After completion of the surveillance test the crew will resume withdrawing control rods in approach to criticality. The second control rod to be moved will be stuck. The crew will enter N1-0P-5, Section H.13 and raise drive water pressure to move the control rod. The control rod will withdraw with increased drive water pressure.
Next, an SRM upscale trip will occur. The crew will respond per ARP F3-4-1 and N1-0P-38A to bypass the SRM.
Then, a complete loss of TBCLC occurs requiring a manual reactor Scram (Critical Task). The Mode Switch will fail to scram the Reactor, however manual ARI actuation will result in successful control rod insertion (Critical Task).
Next, a steam leak will develop from Emergency Condenser 11. The crew will attempt to isolate the leak, however the Emergency Condenser will fail to isolate both automatically and manually.
Two General Areas of the Reactor Building will exceed the maximum safe temperatures. The crew will blowdown the Reactor per N1-EOP-8 (Critical Task).
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC1 15-1 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 100% power. ESW Pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking. PB 11 is aligned to reserve power in preparation for cross-tying PB 16.
Turnover: Lower reactor power to 98% using recirc flow. Cross-tie PB 16A to PB 16B with PB 16B supplying. Then, return PB-11 to normal.
Event Malf.
Event Event No.
No.
Type*
Description NIA Lower reactor power with recirc.
1 R-ATC, SRO N1-0P-1 N/A Cross-tie PB 16A to PB 16B 2
N-BOP, SRO N1-0P-30 3
C-ATC, Control Rod 26-35 Drifts Out RD02 SRO N1-SOP-5.2 FW02A C-Feedwater Booster Pump 11 Trips with Failure of Feedwater 4
RP25 C-AII Respond to trip of Reactor Protection System (RPS) UPS 172 Technical Specification N1-SOP-40.1 6
CU11 M-AII RWCU break in the Secondary Containment requiring scram N1-EOP-2, N1-EOP-5 Failure of the RWCU Isolation Valves to isolate 7
CU14 C-AII N1-EOP-5, N1-EOP-8 C-Mode Switch Fails to Scram 8
Overrides
( R )eactivity, (l)nstrument, (C)omponent, (M)aior
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-2 Op-Test No.: LC115-1
- 1. Total malfunctions (5-8) 6 Events 3, 4, 5, 6, 7, 8
- 2. Malfunctions after EOP entry (1-2) 2 Events 7, 8
- 3. Abnormal events (2-4) 3 Events 3, 4, 5
- 4. Major transients (1-2) 1 Event6
- 5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-5
- 6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-8
- 7. EOP Based Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given an un*isolable RWCU leak outside primary containment and With an un-isolable primary system one general area temperature above the maximum safe limit, the crew will discharging outside of Primary insert a manual reactor scram, in accordance with N1-EOP-5.
Containment resulting in general area temperature above the maximum safe limit, the Reactor must be scrammed.
This reduces the rate of energy production and thus the heat input, radioactivity release, and break flow into the Secondary Containment. This also ensures the Reactor is shutdown prior to need for a blowdown.
CT-2.0: Given an un-isolable RWCU leak outside primary containment and An un-isolable primary system two general area temperatures above the maximum safe limit, the crew will discharging outside of Primary execute N1-EOP-8, RPV Blowdown, in accordance with N1-EOP-5.
Containment resulting in two general area temperatures above the maximum safe limit indicates a wide-spread problem posing a direct and immediate threat to Secondary Containment. A blowdown minimizes flow through the break, rejects heat to the suppression pool in preference to outside the containment, and places the primary system in the lowest possible energy state.
SCENARIO
SUMMARY
The scenario begins at approximately 100% power. ESW pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking. The crew will start by lowering Reactor power to approximately 98% with Recirculation flow. Then the crew will cross-tie PB 16A to PB 168.
Following the power board transfer, a control rod will begin to drift out. The crew will select the drifting control rod and drive it full in. The crew will dispatch an operator to valve out the affected Hydraulic Control Unit to prevent the control rod from continuing to drift.
Then, Feedwater Booster pump 11 will trip. The standby Feedwater Booster pump will fail to auto-start. The crew will manually start the standby Feedwater Booster pump to restore normal system pressures. The SRO will determine the Tech Spec impact for loss of a redundant HPCI component.
RPS UPS 172 will develop an internal fault and drop out the #12 RPS system and RPS Bus 12.
The crew will respond to the trip of UPS per N1-SOP-40.1. The SRO will direct the bus be repowered from l&C Bus 130A and will determine the most limiting Tech Spec condition. The BOP and the RO will reset % scram and % isolations and perform recovery actions after the bus is repowered. The SRO will determine Tech Spec 3.1.2, 3.6.11 and 3.4.4 are the limiting 7 day LCO's applicable with the RPS 12 Bus tripped.
A Reactor Water Cleanup system line break will occur in the Secondary Containment downstream of the Supply Isolation Valves. Reactor Water Cleanup will fail to isolate on high area temperature. The crew will attempt to isolate the system, but the valves will fail to fully close. This break will require a scram (Critical Task) and RPV blowdown (Critical Task) due to exceeding the Maximum Safe Value for general area temperatures. The Mode Switch will fail to scram the Reactor, however either RPS pushbuttons or manual ARI actuation will result in successful control rod insertion.
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC1 15-1 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 100% power. ESW Pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.
Turnover: Remove Line 4 from service per N1-0P-33A to support National Grid maintenance.
Event Malt.
Event Event No.
No.
Type*
Description N-Remove Line 4 from Service 1
N/A
- BOP, ERV Inadvertently Opens SRO N1-SOP-1.4, N1-SOP-1.1, Technical Specifications R-ATC Powerboard 11 Electrical Fault 3
ED04 C-AII N1-SOP-30.1, N1-SOP-1.3, N1-SOP-1.1, Technical Specifications 4
EC01 M-AII Steam Leak in Primary Containment N1-SOP-1, N1-EOP-2, N1-EOP-4, N1-EOP-8 PC10A Torus to Drywell Vacuum Breaker Inadvertently Opens 5
C-AII PC10C N1-EOP-4 FW28A FW28B C-HPCI Fails to Auto-Initiate, Feedwater Pump 13 Disengages, and 6
- BOP, Core Spray Valves Fail to Auto-Open FW06 SRO N1-EOP-2 CS07 C-Partial Primary Containment Isolation Failure 7
Overrides
CT01A C-AII Containment Spray Pump 111 Trips N1-EOP-4 (N)ormal,
( R )eactivity, (l)nstrument, (C)omponent (M)ajor
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-3 Op-Test No.: LC115-1
- 1. Total malfunctions (5-8) 7 Events 2, 3, 4, 5, 6, 7, 8
- 2. Malfunctions after EOP entry (1-2) 4 Events 5, 6, 7, 8
- 3. Abnormal events (2-4) 3 Events 2, 3, 7
- 4. Major transients (1-2) 1 Event4
- 5. EOPs entered/requiring substantive actions (1-2) 2 N1-EOP-2, N1-EOP-4
- 6. EOP contingencies requiring substantive actions (0-2) 1 N1-EOP-8
- 7. EOP Based Critical tasks (2-3) 2 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given an inadvertently open ERV at power, close the ERV or insert a A manual Reactor scram is required manual scram prior to Torus temperature exceeding 110°F, in accordance before Torus temperature exceeds with N1-SOP-1.4 110°F. This reduces the rate of energy production and thus heat input to the Torus. Additionally, this allows evaluating the success of the Reactor scram before boron injection would be required due to Torus temperature in the event of a failure to scram. Closing the ERV prior to the need for the scram avoids the need for these more substantial actions, prevents challenging the plant with a scram, and stops heat input to the Torus.
CT-2.0: Given a LOCA in the Drywell and a failure of HPCI to initiate, inject Maintaining Reactor water level above -
with preferred and alternate injection systems to restore and maintain RPV 84 inches ensures adequate core cooling water level above -84 inches, in accordance with N1-EOP-2.
through the preferred method of core submergence. This protects the integrity of the fuel cladding.
CT-3.0: Given a LOCA in the Drywell and degraded Containment Spray A Slowdown is required to limit further capability, execute N1-EOP-8, RPV Blowdown, when it is determined Torus release of energy into the Primary pressure cannot be maintained below the Pressure Suppression Pressure Containment and to ensure that the RPV limit, in accordance with N1-EOP-4.
is depressurized while pressure suppression capability is still available.
This protects the integrity of the Primary Containment.
SCENARIO
SUMMARY
The scenario begins at approximately 100% power. ESW pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking. The crew will remove Line 4 from service for maintenance. The SRO will determine the Tech Spec impact.
Next, ERV 111 will inadvertently open. The crew will enter N1-SOP-1.4, Stuck Open ERV. The crew will perform an emergency power reduction to approximately 85% power, then take actions to close ERV 111 (Critical Task). These actions will close the ERV, but leave it inoperable.
The SRO will determine the Tech Spec impact.
Next, Powerboard 11 will de-energize due to an electrical fault. This will cause loss of multiple major loads, including a second Recirculation pump, a Service Water pump, and a Circulating Water pump. The crew will respond per N1-SOP-30.1. This will include lowering Reactor power to restore the plant within single Circulating Water pump operating limitations. The SRO will determine the Tech Spec impact of this power loss.
Next, a steam leak will develop inside Primary Containment. The crew will scram the Reactor and execute N1-EOP-2, RPV Control, and N1-EOP-4, Primary Containment Control. After the scram, Feedwater pump 13 will dis-engage early and Feedwater will fail to automatically shift to the HPCI flow-control mode on low Reactor water level. At lower Reactor pressure, Core Spray Isolation Valves will also fail to automatically open. The crew will be able to restore and maintain Reactor water level by manually injecting with preferred and alternate systems (Critical Task). Multiple primary containment isolation valves will fail to close on either manual or automatic containment isolation. The crew will be able to manually close these valves. Two Torus-to-Drywell vacuum breakers will fail open, resulting in some steam escaping from the Drywell directly into the Torus airspace. When the crew initiates Containment Spray, Containment Spray pump 111 will trip. These failures will further degrade Primary Containment pressure control. The Pressure Suppression Pressure (PSP) will be exceeded. The crew will blowdown the Reactor per N1-EOP-8 (Critical Task).
Appendix D Scenario Outline Form ES-D-1 Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 15-1 Examiners:
Operators:
Initial Conditions: The plant is operating at approximately 95% power. ESW Pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking.
Turnover: Start TBCLC Pump 12 and secure TBCLC Pump 11. Then, raise Reactor power to 98%
with Recirculation flow.
Event Malt.
Event Event No.
No.
Type*
Description N-Swap Running TBCLC Pumps 1
N/A
N/A
NM19A I-ATC, APRM #13 fails upscale, half scram, bypass.
SRO ARP Tech Spec RD36A C-BOP, Control Rod Drive Flow Control Valve 44-151 fails closed, requiring 4
SRO shifting to the alternate FCV N1-SOP-5.1, Tech Soec 5
CW16A C-BOP, Service Water Adams Strainer 11 High DIP SRO N1-0P-18, ARP H1-3-2 C-ATC, Loss of main condenser vacuum 6
MC01 SRO N1-SOP-25.1, N1-SOP-1.1, N1-SOP-1 7
ATWS RD33 M-ALL N1-EOP-2, N1-EOP-3 8
TC12 C-ALL All Turbine Bypass Valves Fail Closed N1-EOP-8 (N)ormal, (R)eactivity, (l)nstrument,
( C )om ponent, (M)ajor
Facility: Nine Mile Point Unit 1 Scenario No.: NRC-4 Op-Test No.: LC1 15-1
- 1. Total malfunctions (5-8) 6 Events 3, 4, 5, 6, 7, 8
- 2. Malfunctions after EOP entry (1-2) 2 Events
- 3. Abnormal events (2-4) 3 Events 3, 4, 5
- 4. Major transients (1-2) 1 Event 6
- 5. EOPs entered/requiring substantive actions (1-2) 1 N1-EOP-2
- 6. EOP contingencies requiring substantive actions (0-2) 2 N1-EOP-3, N1-EOP-8
- 7. EOP Based Critical tasks (2-3) 3 CRITICAL TASK DESCRIPTIONS:
CRITICAL TASK JUSTIFICATION:
CT-1.0: Given a failure of the reactor to scram with power above 6% and High Reactor power after a scram is RPV water level above -41 inches, the crew will terminate and prevent all attempted indicates a challenge to injection except boron and CRD, in accordance with N1 -EOP-3.
nuclear fuel and to plant heat sinks. In the event of a loss of the normal heat sink, this may result in adding heat to the Torus and challenging the Primary Containment. Lowering Reactor power reduces these challenges.
CT-2.0: Given a failure of the reactor to scram with power above 6%, the Inserting control rods lowers Reactor crew will lower reactor power by inserting control rods or injecting boron, in power, which reduces challenges to the accordance with N1-EOP-3.
plant during a failure to scram.
Additionally, inserting control rods ultimately provides a long-term, stable core shutdown. Boron injection will lower power, however, alone may not provide a stable shutdown condition.
CT-3.0: Given a failure to scram and the inability to restore and maintain Reactor water level must be maintained RPV water level above -109 inches with the preferred ATWS injection above limits to ensure adequate core systems, the crew will execute N1-EOP-8, RPV Blowdown and restore and cooling. With only low pressure systems maintain RPV water level above -109 inches with preferred and alternate available to inject and Reactor pressure ATWS injection systems, in accordance with N1-EOP-3.
above the pressure limits of these systems, Reactor pressure must be quickly lowered to allow injection. This protects the integrity of the fuel cladding.
SCENARIO
SUMMARY
The scenario begins at approximately 95% power. ESW pump 12 is out of service for maintenance. IRM 11 is bypassed due to spiking. The crew is to start TBCLC pump 12 and secure TBCLC pump 11 per N1-0P-24 section F.1. After the TBCLC pumps are successfully shifted, the crew will raise reactor power with recirc.
Then, APRM 13 will fail upscale causing a half scram. The crew will bypass the APRM and reset the half scram. The CRS will make a Tech Spec call. When the half scram is reset, the Control Rod Drive Flow Control Valve fails closed, requiring shifting to the alternate FCV. After CRD flow is returned to normal, Service Water Adams Strainer 11 will clog, resulting in a high D/P annunciator in the control room. The crew will dispatch an Operator to manually backwash the strainer. When this is unsuccessful at clearing the alarm, the crew will place Service Water pump 12 in service.
Main condenser vacuum will begin to degrade as a result of air in-leakage. The crew will enter N1-SOP-25.1. Power will be lowered per N1-SOP-1.1 and the reactor will be scrammed before vacuum lowers to 22.1" Hg. When the scram is inserted, control rods will only partially insert.
The crew will enter N1-EOP-3 to respond to the failure to scram. The crew will terminate and prevent all injection except boron and CRD (Critical Task). The ATWS will be complicated with the failure of TBVs to open. The crew will initiate liquid poison injection and/or take actions in accordance with N1-EOP-3.1 to insert control rods (Critical Task).
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: February 2017 Examination Level: RO Operating Test Number: LC1 15-1 Administrative Topic Type Describe activity to be performed (see Note)
Code*
Perform Control Rod Position Verification and Conduct of Operations D,S Determine Reactivity Event Severity N1-0P-42, OP-AA-300, N1-0P-5, KIA 2.1.37 (4.3)
Conduct of Oper~tions P,D,R Perform Daily Thermal Limit Surveillance Equipment Control 2015 NRC N1-RESP-1A, KIA2.2.12 (3.7)
Application of Radiation Exposure Limits IAW RP-AA-Radiation Control N,R 203 - SOC Room RP-AA-203, KIA 2.3.4 (3.2)
Actions For External Security Threats Emergency Procedures/Plan D,R OP-NM-106-104, 2.4.28 (3.2)
NOTE: All items (5 total) are required for SROs. RO applicants require only 4 items unless they are retaking only the administrative topics, when all 5 are required.
- Type Codes & Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank (::, 3 for ROs; :::; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (2: 1)
(P)revious 2 exams (::, 1; randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 Facility: Nine Mile Point Unit 1 Date of Examination: February 2017 Examination Level: SRO Operating Test Number: LC1 15-1 Administrative Topic Type Describe activity to be performed (see Note)
Code*
Perform Control Rod Position Verification and Determine Reactivity Event Severity and Notification Conduct of Operations D, S, Requirements OP-AA-300, N1-0P-5, N1-0P-42, K/A 2.1.37 (4.6)
Determine Operator Qualifications Using LMS Conduct of Operations N,R KIA 2.1.8 (4.1)
Perform Daily Thermal Limit Surveillance and Equipment Control P,D,R Determine Corrective Actions 2015 NRC N1-RESP-1A, K/A 2.2.12 (4.1)
Determine Actions for Inoperable Service Water Radiation Control D,R Radiation Monitor N1-ARP-H1, ODCM, KIA2.3.15 (3.1)
Emergency Event Classification and PARs Based on Dose Assessment Emergency Procedures/Plan D,R EP-CE-111, EPIP-EPP-01 EAL Flowchart, KIA 2.4.29 (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 (:5 3 for ROs; ::; 4 for SROs & RO retakes)
(N)ew or (M)odified from bank (?: 1)
(P)revious 2 exams (:5 1; randomly selected)
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility: Nine Mile Point Unit 1 Date of Examination: February 2017 Exam Level: RO/SRO-I Operating Test No.: LC1 15-1 NRC Control Room Systems* (8 for RO); (7 for SRO-I)
System I JPM Title Type Code*
Safety Function
- a. Synchronize Main Generator to the Grid, Main Generator Locks Out D,A,S 4
KIA 245000 A4.02 (3.1/2.9), N1-0P-32, N1-SOP-31.1
- b. Startup Control Room Ventilation (RO Only)
D,S 9
KIA 290003 A4.01 (3.2/3.2), N1-0P-49
- d. Vent the Drywell Prior to Personnel Entry <212 D,P,S,L 5
KIA 223001 A4.03 (3.4/3.4), N1-0P-9 (2015 NRC)
- e. Sequential Loss of Service Water D,A,S 8
KIA 295018 AA1.01 (3.3/3.4) N1-SOP-18.1
- f. Secure a Reactor Recirculation Pump, Pump Trips M,A,S 1
KIA202001 A4.01 (3.7/3.7), N1-0P-1, N1-SOP-1.3
- g. Core Spray Pump Quarterly ST, Suction Strainer Clogging D,A,S,EN 2
KIA 209001 A4.01 (3.8/3.6), N1-ST-Q1C
- h. EOG 103 Control Room Start Following Station Blackout D,S, EN 6
KIA 295003 AA1.02 (4.2/4.3), N1-0P-45 In-Plant Systems* (3 for RO); (3 for SRO-I)
- i. Remove ERV Fuses in the Plant D,E,R 3
KIA 239002 A2.03 (4.1/4.2), N1-SOP-1.4
D,E,R 1
KIA 295037 EA1.10 (3.7/3.9), N1-EOP-3.2
- k. Transfer Reactor Trip Bus 141 to l&C Bus 130A KIA 295003 AA1.01 (3.7/3.8), N1-SOP-30.2 Att 3 D,R 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.
- Type Codes (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L)ow-Power I Shutdown (N)ew or (M)odified from bank including 1 (A)
(P)revious 2 exams (R)CA (S)imulator Pairings:
A then B C and D Criteria for RO / SRO-I / SRO-U 4-6 I 4-6 I 2-3
- 59/:58/:54
~1/~1/~1
~ 1 / ~1 / ~1 (control room system)
~1/~1/~1
~2/~2/~1
- 5 3 I :5 3 / :5 2 (randomly selected)
~1/~1/~1
ES-401 Written Examination Outline Form ES-401-1 Facility:
Nine Mile Point Unit 1 Date of Exam:
February 2017 RO KIA Category Points SRO-Only Points Tier Group K
K K
K K
K A
A A
A G
1 2
3 4
5 6
1 2
3 4
Total A2 G*
Total
- 1.
1 4
3 3
4 3
3 20 3
4 7
Emergency 2
2 1
1 1
1 1
7 2
1 3
Plant Tier Evolutions Totals 6
4 4
5 4
4 27 5
5 10 1
3 2
2 3
2 3
2 2
2 2
3 26 3
2 5
- 2.
Plant 2
1 1
1 1
1 1
2 1
1 1
1 12 0
2 1
3 Systems Tier Totals 4
3 3
4 3
4 4
3 3
3 4
38 5
3 8
- 3. Generic Knowledge & Abilities 1
2 3
4 1
2 3
4 10 7
Categories 3
3 2
2 2
2 2
1 Note:
- 1.
Ensure that at least two topics from every applicable KIA category are sampled within each tier of the RO and SRO-only outlines (i.e., except for one category in Tier 3 of the SRO-only outline, the "Tier Totals" in each KIA category shall not be less than two). (One Tier 3 Radiation Control KIA is allowed if the KIA is replaced by a KIA from another Tier 3 Category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table.
The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRG revisions. The final RO exam must total 75 points and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the associated outline; systems or evolutions that do not apply at the facility should be deleted with justification; operationally important, site-specific systems/evolutions that are not included on the outline should be added. Refer to Section D.1.b of ES-401 for guidance regarding the elimination of inappropriate KIA statements.
- 4.
Select topics from as many systems and evolutions as possible; sample every system or evolution in the group before selecting a second topic for any system or evolution.
- 5.
Absent a plant-specific priority, only those Kl As having an importance rating (IR) of 2.5 or higher shall be selected. Use the RO and SRO ratings for the RO and SRO-only portions, respectively.
- 6.
Select SRO topics for Tiers 1 and 2 from the shaded systems and KIA categories.
- 7.
The generic (G) Kl As in Tiers 1 and 2 shall be selected from Section 2 of the KIA Catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section D.1.b of ES-401 for the applicable KIAs.
- 8.
On the following pages, enter the KIA numbers, a brief description of each topic, the topics' importance ratings (I Rs) 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 a category other than Category A2 or G* on the SRO-only exam, enter it on the left side of Column A2 tor Tier 2, Group 2 (Note #1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the KIA catalog, and enter the KIA numbers, descriptions, I Rs, and point totals (#) on Form ES-401-3. Limit SRO selections to Kl As that are linked to 10 CFR 55.43.
G*
Generic KIAs
ES-401 Form ES-401-1 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE #/Name Safety Function K1 K2 K3 A1 A2 G
KIA Topic(s)
EA2.03 - Ability to determine and/or 295028 High Drywall X
interpret the following as they apply to Temperature I 5 HIGH DRYWELL TEMPERATURE:
Reactor water level AA2.06 - Ability to determine and/or 295001 Partial or Complete interpret the following as they apply to Loss of Forced Core Flow X
PARTIAL OR COMPLETE LOSS OF Circulation / 1 & 4 FORCED CORE FLOW CIRCULATION:
Nuclear boiler instrumentation EA2.01 - Ability to determine and/or 295030 Low Suppression Pool X
interpret the following as they apply to Water Level/ 5 LOW SUPPRESSION POOL WATER LEVEL: Suppression pool level 2.1.23 - Conduct of Operations: Ability 295038 High Off-site Release X
to perform specific system and Rate I 9 integrated plant procedures during all modes of plant operation.
2.4.8 - Emergency Procedures I Plan:
295025 High Reactor Pressure /
X Knowledge of how abnormal operating 3
procedures are used in conjunction with EOPs.
2.2.25 - Equipment Control: Knowledge 295026 Suppression Pool High X
of bases in technical specifications for Water Temperature I 5 limiting conditions for operations and safety limits.
2.2.36 - Equipment Control: Ability to 295004 Partial or Complete analyze the effect of maintenance Loss of DC Power / 6 X
activities, such as degraded power sources, on the status of limiting conditions for operations.
AK1.02 - Knowledge of the operational 600000 Plant Fire On-site / 8 X
implications of the following concepts as they apply to Plant Fire On Site: Fire Fightina EK1.03 - Knowledge of the operational 295038 High Off-site Release implications of the following concepts as X
they apply to HIGH OFF-SITE Rate /9 RELEASE RATE: Meteorological effects on off-site release EK1.02 - Knowledge of the operational 295028 High Drywell implications of the following concepts as X
they apply to HIGH DRYWELL Temperature I 5 TEMPERATURE: Equipment environmental qualification EK2.09 - Knowledge of the interrelations 295031 Reactor Low Water X
between REACTOR LOW WATER Level/ 2 LEVEL and the following: Recirculation system: Plant-Specific AK2.11 - Knowledge of the interrelations 295019 Partial or Complete X
between PARTIAL OR COMPLETE Loss of Instrument Air/ 8 LOSS OF INSTRUMENT AIR and the following: Radwaste AK2.01 - Knowledge of the interrelations 295016 Control Room X
between CONTROL ROOM Abandonment I 7 ABANDONMENT and the following:
Remote shutdown panel: Plant-Specific AK3.01 - Knowledge of the reasons for 295004 Partial or Complete the following responses as they apply to X
PARTIAL OR COMPLETE LOSS OF Loss of DC Power / 6 D.C. POWER: Load shedding: Plant-Specific Imp.
Q#
3.9 76 3.3 77 4.2 78 4.4 79 4.5 80 4.2 81 4.2 82 2.9 39 2.8 40 2.9 41 3.3 42 2.5 43 4.4 44 2.6 45
ES-401 Form ES-401-1 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 1 EAPE #/Name Safety Function K1 K2 K3 A1 A2 G
KIA Topic(s)
AK3.04 - Knowledge of the reasons for 295021 Loss of Shutdown X
the following responses as they apply to Cooling/ 4 LOSS OF SHUTDOWN COOLING:
MaximizinQ reactor water cleanup flow AK3.06 - Knowledge of the reasons for 295005 Main Turbine Generator X
the following responses as they apply to Trip/ 3 MAIN TURBINE GENERATOR TRIP:
RealiQnment of electrical distribution AA 1.01 - Ability to operate and/or 700000 Generator Voltage and monitor the following as they apply to X
GENERATOR VOLTAGE AND Electric Grid Disturbances ELECTRIC GRID DISTURBANCES:
Grid frequency and voltaqe AA 1.02 - Ability to operate and/or 295006 SCRAM / 1 X
monitor the following as they apply to SCRAM: Reactor water level control system EA 1.01 - Ability to operate and/or 295030 Low Suppression Pool monitor the following as they apply to X
LOW SUPPRESSION POOL WATER Water Level/ 5 LEVEL: EGGS systems (NPSH considerations): Plant-Soecific EA2.02 - Ability to determine and/or 295025 High Reactor Pressure/
X interpret the following as they apply to 3
HIGH REACTOR PRESSURE: Reactor cower AA2.04 - Ability to determine and/or 295023 Refueling Accidents / 8 X
interpret the following as they apply to REFUELING ACCIDENTS: Occurrence of fuel handlinq accident EA2.01 - Ability to determine and/or 295026 Suppression Pool High interpret the following as they apply to X
SUPPRESSION POOL HIGH WATER Water Temperature I 5 TEMPERATURE: Suppression pool water temperature 295001 Partial or Complete 2.2.40 - Equipment Control: Ability to Loss of Forced Core Flow X
apply technical specifications for a Circulation / 1 & 4 system.
2.2.44 - Equipment Control: Ability to interpret control room indications to 295018 Partial or Complete X
verify the status and operation of a Loss of CCW / 8 system, and understand how operator actions and directives affect plant and system conditions.
2.2.36 - Equipment Control: Ability to 295003 Partial or Complete analyze the effect of maintenance Loss of AC/ 6 X
activities, such as degraded power sources, on the status of limiting conditions for operations.
EA 1.11 - Ability to operate and/or 295024 High Drywell Pressure I X
monitor the following as they apply to 5
HIGH DRYWELL PRESSURE: Drywell spray: Mark-1&11 EK1.07 - Knowledge of the operational 295037 SCRAM Condition implications of the following concepts as Present and Reactor Power X
they apply to SCRAM CONDITION Above APRM Downscale or PRESENT AND REACTOR POWER Unknown/ 1 ABOVE APRM DOWNSCALE OR UNKNOWN: Shutdown marQin KIA Category Totals:
4 3
3 4
3/3 3/4 Group Point Total:
Imp.
Q#
3.3 46 3.3 47 3.6 48 3.9 49 3.6 50 4.2 51 3.4 52 4.1 53 3.4 54 4.2 55 3.1 56 4.2 57 3.4 58 I 20/7
ES-401 Form ES-401-1 Nine Mile Point Unit 1 2017 NRG Exam Written Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 Group 2 EAPE #/Name Safety Function K1 K2 K3 A1 A2 G
KIA Topic(s)
EA2.02 - Ability to determine and/or 295035 Secondary interpret the following as they apply to Containment High Differential X
SECONDARY CONTAINMENT HIGH Pressure/ 5 DIFFERENTIAL PRESSURE: Off-site release rate: Plant-Specific 2.4.35 - Emergency Procedures I Plan:
295015 Incomplete SCRAM/ 1 X
Knowledge of local auxiliary operator tasks during emergency and the resultant operational effects.
AA2.01 - Ability to determine and/or 295009 Low Reactor Water X
interpret the following as they apply to Level LOW REACTOR WATER LEVEL:
Reactor water level EK1.02 - Knowledge of the operational 295036 Secondary implications of the following concepts as Containment High Sump/Area X
they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA Water Level/ 5 WATER LEVEL: Electrical ground/
circuit malfunction AK2.07 - Knowledge of the interrelations 295002 Loss of Main X
between LOSS OF MAIN CONDENSER Condenser Vacuum/ 3 VACUUM and the following: Offgas system EK3.06 - Knowledge of the reasons for 500000 High Containment the following responses as they apply to X
HIGH PRIMARY CONTAINMENT Hydrogen Concentration / 5 HYDROGEN CONCENTRATIONS:
Operation of wet well vent AA 1.08 - Ability to operate and/or 295008 High Reactor Water X
monitor the following as they apply to Level/ 2 HIGH REACTOR WATER LEVEL:
Feedwater system AA2.01 - Ability to determine and/or 295007 High Reactor Pressure X
interpret the following as they apply to
/3 HIGH REACTOR PRESSURE: Reactor pressure 2.4.45 - Emergency Procedures I Plan:
295022 Loss of CAD Pumps / 1 X
Ability to prioritize and interpret the significance of each annunciator or alarm.
EK1.01 - Knowledge of the operational 295035 Secondary implications of the following concepts as Containment High Differential X
they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL Pressure I 5 PRESSURE: Secondary containment intearity KIA Category Totals:
2 1
1 1
1/2 1/1 Group Point Total:
Imp.
Q#
4.1 83 4.0 84 4.2 85 2.6 59 3.1 60 3.1 61 3.5 62 4.1 63 4.1 64 3.9 65 I 7/3
ES-401 System # I Name K
1 300000 Instrument Air 259002 Reactor Water Level Control 400000 Component Cooling Water 239002 SRVs 218000 ADS 212000 RPS X
264000 EDGs X
2 Nine Mile Point Unit 1 2017 NRG Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K
K K
K A A2 A A G
3 4
5 6
1 3
4 A2.01 - Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM; and (b) based on those X
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Air dryer and filter malfunctions A2.03 - Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b)
X based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of reactor water level input 2.1.28 - Conduct of Operations:
X Knowledge of the purpose and function of major system components and controls.
2.4.41 - Emergency Procedures X
/ Plan: Knowledge of the emergency action level thresholds and classifications.
A2.04 - Ability to (a) predict the impacts of the following on the AUTOMATIC DE PRESSURIZATION SYSTEM; and (b) based on X
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: ADS failure to initiate K1.12 - Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following: Reactor/turbine pressure control system: Plant-Specific K1.01 - Knowledge of the physical connections and/or cause-effect relationships between EMERGENCY GENERATORS (DIESEUJET) and the following: A.C. electrical distribution K2.03 - Knowledge of electrical X
power supplies to the following:
Initiation looic: BWR-2,3,4 Imp Q#
2.8 86 3.7 87 4.1 88 4.6 89 4.2 90 3.4 1
3.8 2
2.8 3
ES-401 System # I Name K
1 400000 Component Cooling Water 263000 DC Electrical Distribution 300000 Instrument Air 239002 SRVs 215005 APRM / LPRM 215003 IRM 205000 Shutdown Cooling 218000 ADS 259002 Reactor Water Level Control Form ES-401-1 K
2 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K
K K
K A A2 A A G
3 4
5 6
1 3
4 K2.02 - Knowledge of electrical X
power supplies to the following:
CCWvalves K3.02 - Knowledge of the effect that a loss or malfunction of the D.C. ELECTRICAL X
DISTRIBUTION will have on following: Components using D.C. control power (i.e.
breakers)
K3.02 - Knowledge of the effect that a loss or malfunction of the X
INSTRUMENT AIR SYSTEM will have on the following: Systems having pneumatic valves and controls K4.04 - Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or X
interlocks which provide for the following: Ensures even distribution of heat load to suppression pool, and adequate steam condensina K4.06 - Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER X
RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Effects of detector aaina on LPRM/APRM readinas K5.01 - Knowledge of the operational implications of the X
following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM:
Detector oceration K5.02 - Knowledge of the operational implications of the following concepts as they apply X
to SHUTDOWN COOLING SYSTEM (AHR SHUTDOWN COOLING MODE): Valve oceration K6.03 - Knowledge of the effect that a loss or malfunction of the following will have on the X
AUTOMATIC DE PRESSURIZATION SYSTEM: Nuclear boiler instrument system (level indication)
K6.01 - Knowledge of the effect that a loss or malfunction of the X
following will have on the REACTOR WATER LEVEL CONTROL SYSTEM: Plant air svstems Imp Q#
2.9 4
3.5 5
3.3 6
3.4 7
2.6 8
2.6 9
2.8 10 3.8 11 3.2 12
ES-401 System # I Name K
K 1
2 207000 Isolation (Emergency) Condenser 223002 PCIS/Nuclear Steam Supply Shutoff 209001 LPCS 215004 Source Range Monitor 262002 UPS (AC/DC) 261000 SGTS 262001 AC Electrical Distribution 211000 SLC 212000 RPS Form ES-401-1 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K
K K
K A A2 A A G
3 4
5 6
1 3
4 A 1.1 O - Ability to predict and/or monitor changes in parameters associated with operating the X
ISOLATION (EMERGENCY)
CONDENSER controls including: Primary side temperature: BWR-2,3 A 1.04 - Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT X
ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: Individual system relav status A2.03 - Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based X
on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: AC. failures A2.01 - Ability to (a) predict the impacts of the following on the SOURCE RANGE MONITOR (SRM) SYSTEM; and (b) based X
on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power supply deqraded A3.01 - Ability to monitor automatic operations of the X
UNINTERRUPTABLE POWER SUPPLY (AC./D.C.) including:
Transfer from preferred to alternate source A3.04 - Ability to monitor automatic operations of the X
STANDBY GAS TREATMENT SYSTEM including: System temperature A4.05 - Ability to manually operate and/or monitor in the X
control room: Voltage, current, power, and frequency on AC.
buses A4.07 - Ability to manually X
operate and/or monitor in the control room: Lights and alarms 2.4.34 - Emergency Procedures
/ Plan: Knowledge of RO tasks X
performed outside the main control room during an emergency and the resultant operational effects.
Imp Q#
3.2 13 2.6 14 3.4 15 2.7 16 2.8 17 3.0 18 3.3 19 3.6 20 4.2 21
ES-401 System # / Name K
1 218000 ADS 400000 Component Cooling Water 205000 Shutdown Cooling 215004 Source Range Monitor 206000 HPCI X
KIA Category Totals:
3 Form ES-401-1 K
2 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 1 K
K K
K A A2 A A G
3 4
5 6
1 3
4 2.4.11 - Emergency Procedures X
/ Plan: Knowledge of abnonnal condition procedures.
K6.07 - Knowledge of the effect that a loss or malfunction of the X
following will have on the CCWS: Breakers, relays, and disconnects K4.03 - Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN X
COOLING MODE) design feature(s) and/or interlocks which provide for the following:
Low reactor water level: Plant-Specific 2.1.20 - Conduct of Operations:
X Ability to interpret and execute procedure stePs.
K1.04 - Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following:
Reactor feedwater system:
BWR-2,3,4 2
2 3
2 3
2 2/3 2
2 3/2 Group Point Total:
Imp Q#
4.0 22 2.7 23 3.8 24 4.6 25 3.6 26 I 26/5
ES-401 K
K System # I Name 1
2 201006 RWM 271000 Off-gas 286000 Fire Protection 215001 Traversing In-core X
Probe 202001 Recirculation X
233000 Fuel Pool Cooling/Cleanup 241000 Reactor/Turbine Pressure Regulator 226001 RHR/LPCI:
Containment Spray Mode 290001 Secondary Containment Form ES-401-1 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K
K K
K A
A2 A
A G
3 4
5 6
1 3
4 A2.02 - Ability to (a) predict the impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on X
those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of steam flow inout: P-Soec/Not-BWR6) 2.2.22 - Equipment Control:
X Knowledge of limiting conditions for ooerations and safetv limits.
A2.11 - Ability to (a) predict the impacts of the following on the FIRE PROTECTION SYSTEM; and (b) based on those X
predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trips: Plant-Soecific K1.01 - Knowledge of the physical connections and/or cause-effect relationships between TRAVERSING IN-CORE PROBE and the following: Local power range monitors /Not-BWR1 \\
K2.02 - Knowledge of electrical power supplies to the following:
MG sets: Plant-Soecific K3.08 - Knowledge of the effect that a loss or malfunction of the X
FUEL POOL COOLING AND CLEAN-UP will have on followina: Refuelino operations K4.19 - Knowledge of REACTOR/TURBINE PRESSURE REGULATING X
SYSTEM design feature(s) and/or interlocks which provide for the following: Steam bypass valve control K5.06 - Knowledge of the operational implications of the following concepts as they apply X
to RHR/LPCI:
CONTAINMENT SPRAY SYSTEM MODE: Vacuum breaker ooeration K6.01 - Knowledge of the effect that a loss or malfunction of X
the following will have on the SECONDARY CONTAINMENT:
Reactor building ventilation:
Plant-Soecific Imp.
Q 2.9 91 4.7 92 3.2 93 2.5 27 3.2 28 2.9 29 3.6 30 2.6 31 3.5 32
ES-401 System # I Name K
K 1
2 272000 Radiation Monitoring 290003 Control Room HVAC 201003 Control Rod and Drive Mechanism 201002 RMCS 201001 CAD Hydraulic 245000 Main Turbine Generator and Auxiliaries KIA Category Totals:
1 1
Form ES-401-1 Nine Mile Point Unit 1 2017 NRC Exam Written Examination Outline Plant Systems - Tier 2 Group 2 K
K K
K A A2 A A G
3 4
5 6
1 3
4 A 1.02 - Ability to predict and/or monitor changes in parameters associated with operating the X
RADIATION MONITORING SYSTEM controls including:
Lights, alarms, and indications associated with surveillance testinQ A2.01 - Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC; and (b) based on those predictions, X
use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:
lnitiation/reconfiQuration A3.01 - Ability to monitor automatic operations of the X
CONTROL ROD AND DRIVE MECHANISM including: Control rod position A4.06 - Ability to manually X
operate and/or monitor in the control room: Rod select matrix power switch 2.1. 7 - Conduct of Operations:
Ability to evaluate plant performance and make X
operational judgements based on operating characteristics, reactor behavior, and instrument interpretation.
A 1.01 - Ability to predict and/or monitor changes in parameters associated with operating the X
MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS controls including: Generator meQawatts 1
1 1
1 2
1/2 1
1 1/1 Group Point Total:
Imp.
Q 2.9 33 3.1 34 3.7 35 2.8 36 4.4 37 2.7 38 I 12/3
ES-401 Generic Knowledge and Abilities Outline {Tier 3)
Form ES-401-3 Facility:
Nine Mile Point Unit 1 Date:
February 2017 RO SRO-Only Category KIA#
Topic IR Q#
IR Q#
2.1.32 Ability to explain and apply all system limits 4.0 94 and precautions.
Knowledge of individual licensed operator responsibilities related to shift staffing, such 2.1.4 as medical requirements, "no-solo" operation, 3.8 99 maintenance of active license status, 1 OCFR55, etc.
- 1.
Conduct Knowledge of industrial safety procedures of Operations 2.1.26 (such as rotating equipment, electrical, high 3.4 66 temperature, high pressure, caustic, chlorine, oxygen and hydrogen).
2.1.1 Knowledge of conduct of operations 3.8 67 requirements.
2.1.8 Ability to coordinate personnel activities 3.4 74 outside the control room.
Subtotal 3
2 2.2.40 Ability to apply technical specifications for a 4.5 95 system.
2.2.23 Ability to track Technical Specification limiting 4.6 100 conditions for operations.
Ability to perform pre-startup procedures for 2.2.1 the facility, including operating those controls 4.5 68
- 2.
associated with plant equipment that could Equipment affect reactivity.
Control Ability to recognize system parameters that 2.2.42 are entry-level conditions for Technical 3.9 69 Specifications.
Knowledge of the process for managing 2.2.18 maintenance activities during shutdown 2.6 75 operations, such as risk assessments, work prioritization, etc.
Subtotal 3
2 Knowledge of Radiological Safety Procedures
- 3.
pertaining to licensed operator duties, such as Radiation 2.3.13 response to radiation monitor alarms, 3.8 96 Control containment entry requirements, fuel handling responsibilities, access to locked high radiation areas, aligning filters, etc.
ES-401 Generic Knowledge and Abilities Outline {Tier 3)
Form ES-401-3 Knowledge of radiation or containment 2.3.14 hazards that may arise during normal, 3.8 98 abnormal, or emergency conditions or activities.
2.3.11 Ability to control radiation releases.
3.8 70 Ability to use radiation monitoring systems, 2.3.5 such as fixed radiation monitors and alarms, 2.9 71 portable survey instruments, personnel monitorinq equipment, etc.
Subtotal 2
2 Knowledge of events related to system operation/status that must be reported to 2.4.30 internal organizations or external agencies, 4.1 97 such as the State, the NRC, or the transmission system operator.
- 4.
Emergency 2.4.25 Knowledge of fire protection procedures.
3.3 72 Procedures/
Knowledge of the parameters and logic used Plan to assess the status of safety functions, such 2.4.21 as reactivity control, core cooling and heat 4.0 73 removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.
Subtotal 2
1 Tier 3 Point Total 10 7
ES-401 Tier/
Group Record of Rejected Kl As Form ES-401-4 Randomly Selected KIA Reason for Rejection
,/"'>:;;)'/,:>',I;~"<},,,,,,,,,,.,.,,
,"\\':\\:\\,,'\\{','
,,,"2"?'~\\J,,,;
\\
The fofk>wiflO K/Ag.~ reteoted following the svstematic and random aamoJino orocess:
1 / 1 1 / 2 3
Question 80 295025 High Reactor Pressure 2.4.3 - Emergency Procedures/
Plan: Ability to identify post-accident instrumentation.
Question 85 295013 High Suppression Pool Temperature AA2.01 - Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool temperature Question 95 2.2.15 - Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.
An acceptable question could not be developed at the SRO level for with randomly sampled generic KIA with this evolution. Reactor pressure instrumentation is not required for post-accident monitoring per Technical Specifications.
Randomly re-selected KIA 295025 High Reactor Pressure 2.4.8 - Emergency Procedures/ Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.
An acceptable question could not be developed for the randomly sampled KIA without overlapping concepts with Questions 53 and
- 81.
Randomly re-selected KIA 295009 Low Reactor Water Level AA2.01 - Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Reactor water level.
An acceptable question could not be developed at the SRO level for with randomly sampled generic KIA. Additionally, this generic KIA is better tested on the operating exam.
Randomly re-selected KIA 2.2.40 - Ability to apply Technical Specifications for a system.
Question 97 The randomly sampled generic KIA is better 2.4.47 - Ability to diagnose and suited for, and extensively tested on, the recognize trends in an accurate operating exam.
and timely manner utilizing the Randomly re-selected KIA 2.4.30 - Knowledge 3
appropriate control room of events related to system operation/status reference material.
that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.
Question 12 An acceptable question could not be developed 259002 Reactor Water Level for the randomly sampled KIA without Control overlapping Question 87.
K6.05 - Knowledge of the effect Randomly re-selected KIA 259002 Reactor Water Level Control K6.01 - Knowledge of the 2 I 1 that a loss or malfunction of the effect that a loss or malfunction of the following following will have on the will have on the REACTOR WATER LEVEL REACTOR WATER LEVEL CONTROL SYSTEM: Plant air systems.
CONTROL SYSTEM: Reactor water level input Question 15 An acceptable question could not be developed 209001 LPCS for the randomly sampled KIA because the facility's Core Spray pumps are not in separate A2.07 - Ability to (a) predict the rooms with dedicated room cooling, as in some impacts of the following on the other plants.
LOW PRESSURE CORE Randomly re-selected KIA 209001 LPCS A2.03 2 I 1 SPRAY SYSTEM; and (b)
- Ability to (a) predict the impacts of the based on those predictions, use following on the LOW PRESSURE CORE procedures to correct, control, SPRAY SYSTEM; and (b) based on those or mitigate the consequences of predictions, use procedures to correct, control, those abnormal conditions or or mitigate the consequences of those operations: Loss of room abnormal conditions or operations: A.C.
cooling failures.
Question 21 An acceptable question could not be developed 212000 RPS for the randomly sampled generic Kl A due to lack of applicable reference material (graphs, 2.1.25 - Ability to interpret curves, tables, etc.) for RPS.
2 I 1 reference materials, such as Randomly re-selected KIA 212000 RPS 2.4.34 -
graphs, curves, tables, etc.
Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.
Question 25 An acceptable question could not be developed 215004 Source Range Monitor for the randomly sampled generic Kl A due to lack of use of SRMs in EOP mitigation 2.4.6 - Knowledge of EOP strategies.
2 I 1 mitigation strategies.
Randomly re-selected KIA 215004 Source Range Monitor 2.1.20 - Ability to interpret and execute procedure steps.
Question 41 An acceptable question could not be developed 295028 High Drywell for the randomly sampled KIA without overlapping Question 76.
Temperature Randomly re-selected KIA 295028 High Drywell EK1.01 - Knowledge of the Temperature EK1.02 - Knowledge of the 1 / 1 operational implications of the operational implications of the following following concepts as they concepts as they apply to HIGH DRYWELL apply to HIGH DRYWELL TEMPERATURE: Equipment environmental TEMPERATURE: Reactor q ual if ication.
water level measurement Question 55 An acceptable question could not be developed 295018 Partial or Complete for the randomly sampled generic KIA due to lack of a less than or equal to one hour Loss of CCW Technical Specification related to Loss of CCW.
2.2.39 - Knowledge of less than Randomly re-selected KIA 295018 Partial or 1 / 1 or equal to one hour technical Complete Loss of CCW 2.2.44 - Ability to specification action statements interpret control room indications to verify the for systems.
status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
Question 65 An acceptable question could not be developed 295035 Secondary for the randomly sampled KIA without overlapping Question 83.
Containment High Differential Pressure Randomly re-selected KIA 295035 Secondary EK1.02 - Knowledge of the Containment High Differential Pressure EK1.01
- Knowledge of the operational implications of 1 / 2 operational implications of the the following concepts as they apply to following concepts as they SECONDARY CONTAINMENT HIGH apply to SECONDARY DIFFERENTIAL PRESSURE: Secondary CONTAINMENT HIGH containment integrity.
DIFFERENTIAL PRESSURE:
Radiation release
Question 67 The randomly sampled generic KIA is better 2.1.6 - Ability to manage the suited for, and extensively tested on, the operating exam.
3 control room crew during plant transients.
Randomly re-selected KIA 2.1.1 - Knowledge of conduct of operations requirements.
Question 6 An acceptable question could not be developed 300000 Instrument Air for the randomly sampled Kl A because there is no feature to cross-tie Instrument Air between K3.03 - Knowledge of the effect Nine Mile Point Units 1 and 2, as there is at that a loss or malfunction of the some other multiple unit sites.
2/1 INSTRUMENT AIR SYSTEM Randomly re-selected KIA 300000 Instrument will have on the following:
Air K3.02 - Knowledge of the effect that a loss Cross-tied units or malfunction of the INSTRUMENT AIR SYSTEM will have on the following: Systems having pneumatic valves and controls.
Question 66 A discriminating question could not be 2.1.19 - Ability to use plant developed at the license level for the randomly sampled KIA. Additionally, the KIA is tested computers to evaluate system extensively on the operating portion of the or component status.
exam.
3 Randomly re-selected KIA 2.1.26 - Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).
Question 92 An acceptable question could not be written for 271000 Off-gas the randomly sampled KIA due to lack of safety related equipment in the Off-gas system.
2/2 2.2.37 - Ability to determine Randomly re-selected KIA 271000 Off-gas operability and / or availability of 2.2.22 - Knowledge of limiting conditions for safety related equipment.
operations and safety limits.