ML19217A271
| ML19217A271 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/01/2019 |
| From: | Exelon Generation Co |
| To: | Operations Branch I |
| Shared Package | |
| ML18151A269 | List: |
| References | |
| CAC000500, EPID L-2019-OLL-0014 | |
| Download: ML19217A271 (31) | |
Text
Exelon Generation Three Mile Island Unit 1 Telephone 717-948-8000 Route 441 South, P. 0. Box 480 Middletown, PA 17057 January 11, 2019 TMl-18-110 USNRC, Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT 1 (TMl-1)
RENEWED OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 SUBMITIAL OF INITIAL OPERATOR LICENSING EXAMINATION OUTLINES Enclosed are the examination outlines, supporting the Initial License Examination scheduled for the week of June 10, 2019, at Three Mile Island Unit 1.
This submittal includes all appropriate Examination Standard forms and outlines in accordance with NU REG 1021, "Operator Licensing Examination Standards for Power Reactors," Revision 11.
In accordance with NUREG 1021, Revision 11, Section ES-201, "Initial Operator Licensing Examination Process," please ensure that these materials are withheld from public disclosure until after the examinations are complete.
Should you have any questions concerning this letter, please contact Mike Fitzwater of Regulatory Assurance at (717) 948-8228. For questions concerning examination materials, please contact Todd Beaver, Exam Author, at (717) 948-2080.
Respectfully, Edward W. Callan Site Vice President, Three Mile Island Unit 1 Exelon Generation Co., LLC
Enclosures:
(Mailed to Peter Presby, Chief Examiner, NRC Region I)
Examination Security Agreement (Form ES-201-3)
Administrative Topics Outline (Form ES-301-1)
Control Room/In-Plant Systems Outline (Form ES-301-2)
PWR Examination Outline (Form ES-401-2)
SUBMITIAL OF INITIAL OPERATOR LICENSING EXAMINATION OUTLINES TMl-18-110 Page 2 Generic Knowledge and Abilities Outline (Tier 3) (Form ES-401-3)
Scenario Outline (Form ES-D-1)
Record of Rejected K/As (Form ES-401-4)
Completed Checklists:
Examination Outline Quality Checklist (Form ES-201-2)
Transient and Event Checklist (Form ES-301-5) cc:
(without attachments)
Chief, NRC Operator Licensing Branch NRC Senior Resident Inspector-TM! Unit 1
bee:
Regulatory Assurance Manager-TM I Unit 1 Training Manager - TMI Unit 1
ES-201 Examination Outline Quality Checklist Form ES-201-2 Facility: Three Mile Island Date of Examination: 6/10/19 Initials Item Task Description a
b*
c**
- 1.
- a.
Verify that the outline(s) fit(s) the appropriate model in accordance with ES-401 or ES-401 N.
- rl3 ~ ~
w
- b.
Assess whether the outline was systematically and randomly prepared in accordance with
~
R Section D.1 of ES-401 or ES-401 N and whether all K/A categories are appropriately sampled.
[13
~
I T
C.
Assess whether the outline overemphasizes any systems, evolutions, or generic topics.
11?, ~ L T
E
- d. Assess whether the justifications for deselected or rejected K/A statements are appropriate.
~ h N
-10
- 2.
- a.
Using Form ES-301-5, verify that the proposed scenario sets cover the required number of normal evolutions, instrument and component failures, technical specifications, and major re
~ b s
I M
- b.
Assess whether there are enough scenario sets (and spares) to test the projected number and u
mix of applicants in accordance with the expected crew composition and rotation schedule L
without compromising exam integrity, and ensure that each applicant can be tested using at
'tH- ~
A least one new or significantly modified scenario, that no scenarios are duplicated from the
~
T applicants' audit test(s), and that scenarios will not be repeated on subsequent days.
0 To the extent possible, assess whether the outline(s) conforms with the qualitative and R
C.
quantitative criteria specified on Form ES-301-4 and described in Appendix D and in
'fl3 ? fr Section D.5, "Specific Instructions for the 'Simulator Operating Test,'" of ES-301 (including overlap).
- 3.
- a. Verify that the systems walkthrough outline meets the criteria specified on Form ES-301-2:
(1) The outline(s) contains the required number of control room and in-plant tasks distributed w
among the safety functions as specified on the form.
A (2)
Task repetition from the last two NRC examinations is within the limits specified on the form.
L (3)
No tasks are duplicated from the applicant's audit test(s).
~ lib K
(4)
The number of new or modified tasks meets or exceeds the minimums specified on the form.
TS T
(5)
The number of alternate-path, low-power, emergency, and radiologically controlled area H
tasks meets the criteria on the form.
R 0
- b. Verify that the administrative outline meets the criteria specified on Form ES-301-1:
u (1)
The tasks are distributed among the topics as specified on the form.
'f'1 ~
,lb G
(2) At least one task is new or significantly modified.
H (3)
No more than one task is repeated from the last two NRC licensing examinations.
- c.
Determine whether there are enough different outlines to test the projected number and mix of T'43 *
~
applicants and ensure that no items are duplicated on subsequent days.
- 4.
- a.
Assess whether plant-specific priorities (including probabilistic risk assessment and individual 1*-~
plant examination insights) are covered in the appropriate exam sections.
~
G TP ~
b E
- b. Assess whether the 10 CFR 55.41, 55.43, and 55.45 sampling is appropriate.
N Ensure that K/A importance ratings (except for plant-specific priorities) are at least 2.5.
~ ~
~
E C.
R
- d. Check for duplication and overlap among exam sections and the last two NRC exams.
hO ~ I~
A 11f ~
L
- e. Check the entire exam for balance of coverage.
'Ti)
- f.
Assess whether the exam fits the appropriate job level (RO or SRO).
~ ~
~
rM/\\ ~.1111.
Printed Name/Signature ti
- a.
Author "1Wvi?ir'-I
~'L.
Facility Reviewer(*)
-rt; 1-!- ' \\.L, l"° \\
~
J. ~-r;r,1,/,/Jf\\
0
- b.
- c.
NRC Chief Examiner (#)
l't>UV 'PH5../ot,(
,;,-2,.DfJIY _, v
)
I
- d.
NRC Supervisor r-....-61,J L... ~tol / rJUuv eL, lt1
\\,... -
"- _/
- Not applicable for NRG-prepared examination outlines.
- The independent NRC reviewer initials items in column "c"; the chief examiner's concurrence is required.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Three Mile Island Date of Examination:
06/10/19 Examination Level: RO
[Zl SRO D Operating Test Number:
TMl2019 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Perform a Reactivity Balance at Power Conduct of Operations N,R KIA: 2.1.25 (3.9)
Complete RB Average Air Temperature Conduct of Operations D,R Calculation KIA: 2.1.7(4.4)
Station Print Reading - Isolate Instrument Air Equipment Control Leak N,R KIA: 2.2.41 (3.9)
Radiation Control Emergency Plan Perform State and Local Event Notification D,S KIA: 2.4.43 (3.2)
NOTE: All items (five total} are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(::;; 3 for ROs;::;; 4 for SROs and 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 JPM RA1 The examinee will have to perform a Reactivity Balance at Power for the given parameters. The examinee will determine that we are within correct bands.
JPM RA1 The examinee will be given a picture of the RB Air Temperature Yokogawa recorder and must complete the shift and daily checks procedure 1301-1. The examinee must perform a calculation and identify any out-of-specification reading.
JPM RA2 - The examinee must identify isolation points for a leaking instrument air valve and determine the effect on plant components.
JPM RA4 - The examinee must perform a state and local event notification for the declared EAL.
ES-301 Administrative Topics Outline Form ES-301-1 Facility:
Three Mile Island Date of Examination:
06/10/19 Examination Level: RO D SRO
[ZJ Operating Test Number:
TMl2019 Administrative Topic (see Note)
Type Describe activity to be performed Code*
Issue a Controlled Key Conduct of Operations M,R KIA: 2.1.13 (3.2)
Calculate and Approve an ECB Conduct of Operations D,R KIA: 2.1.37 (4.6)
Evaluate completed surveillance and perform Equipment Control D,R actions KIA: 2.2.37 (4.6)
Authorize emergency personnel exposure in Radiation Control D,R excess of 5 REM KIA: 2.3.4 (3.7)
EAL and PAR Emergency Plan N,R KIA: 2.4.44 (4.4)
NOTE: All items (five total) are required for SROs. RO applicants require only four items unless they are retaking only the administrative topics (which would require all five items).
- Type Codes and Criteria:
(C)ontrol room, (S)imulator, or Class(R)oom (D)irect from bank(:;; 3 for ROs; :;; 4 for SROs and RO retakes)
(N)ew or (M)odified from bank (~ 1)
(P)revious 2 exams (::, 1, randomly selected)
ES-301 Administrative Topics Outline Form ES-301-1 JPM SA1-1 -The examines must issue a key for a Locked High Radiation Area and a CDA key. The examines must review all the requirements and ensure that the examiner possesses the correct paperwork and qualifications to be issued the key. A CDA key can only be issued by shift management.
JPM SA1 The examines will be given a shutdown boron calculation with errors. The examines will have to find the errors and calculate the correct shutdown boron.
JPM SA2 - The examines will be given a surveillance procedure with some parameters exceeding a threshold. The examines must identify the out-of-specification parameters and determine any technical specification required actions.
JPM SA3 - The examines must authorize dose in excess of 5 REM for emergency personnel.
The examines will have to determine which personnel meet the requirements for the dose.
JPM SA4 - The examines will have to classify an EAL and make a PAR based on plant conditions.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Three Mile Island Date of Examination:
06/10/19 Exam Level: RO
~ SRO-I D SRO-U D Operating Test Number:
TMl2019 Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 001 / Respond to a dropped control rod - ICS fails to complete D,A,S 1
runback 003AA1.02 - Control Rod System
L
- c. 006 I Lower CFT level and pressure from the Control Room - 006 D,S 3
A4.02 - Core Flood System
- e. 003 I Restore SI with a loss of ICCW 003A3.01 - Reactor Coolant D,A,S,P 4P Pump
- f. 007 I Pump RCDT to MWST 007A1.01 - Pressurizer Relief Tank D,S 5
Diesel Generators
- h. 072 / Respond IAW OP-TM-MAP-C0101 Fuel Handling Incident in N,A,S 7
the Spent Fuel Pool - Radiation Monitors In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U i.008 I Loss of Instrument Air - 008A2.05 - Intermediate Closed D, E, R 8
Cooling Water System
- j. 071 / Purge of the Waste Gas System Radiation Monitor (RM-A-7)
D,R 9
071 A4.09 - Waste Gas Disposal System
- k. 061 / Respond to a failure of EF-P-2A and EF-V-30D 061 A2.04 -
D,E 4S Emergency Feedwater All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes I
Criteria for R /SR0-1/SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature (L )ow-Power/Shutdown (N)ew or (M)odified from bank including 1 (A)
(P)revious 2 exams (R)CA (S )imulator 4-6/4-6 /2-3 S 9/S 8/:54
.::1/.::1/.::1
.:: 1/.:: 1/.:: 1 (control room system)
.::1/.::1/.::1
.:: 2/.:: 2/.:: 1 s 3/S 3/S 2 (randomly selected)
.::1/.::1/.::1 Random Selection: JPM E was selected by assigning the JPMs of the previous 2 years a number then using a random number generator to select the JPM.
JPM A - A rod is dropped into the core, which initiates a plant runback. The examinee will observe the plant runback but must recognize the plant did not runback to the appropriate power level. The examinee must run the plant back in manual at the ULD or SG/RX demand station. This JPM is similar to Scenario 2, Event 4. This JPM is different because of the power level in which it starts. In this JPM, the power will drop then the plant will runback. In the scenario, the power is low and control rods will pull to maintain power.
JPM B - The reactor is in a tripped state, with the loss of the 1 D 4kV bus when an RCS leak occurs. The examinee must initiate the 'B' ES, but the manual pushbuttons will fail. The examinee must manually start the 'B' ES equipment.
JPM C - The core flood tanks are above their admin and technical specification limits. The examinee will have to lower level and pressure to within band.
JPM D - The examinee will respond to an RCS leak that is large enough to lose Subcooling Margin. The examinee will perform Rule 1, to secure reactor coolant pumps, initiate a 1600#
ES, and initiate Emergency Feedwater (EFW). Emergency Feedwater will fail to feed to the desired level (50% in the operating range), the examinee will have to take manual control of EFW and begin feeding to the desired level. This JPM is similar to Scenario 4, Event 7. This JPM provides at least 2 significant actions (critical tasks) that the simulator event does not perform. The simulator event does not have any failures with Emergency Feedwater, whereas the standard for the JPM requires the examinee to identify and mitigate an Emergency Feedwater failure.
JPM E - The examinee will be directed to restore seal injection due to a makeup pump trip.
Once seal injection is restored, Intermediate Closed Cooling Water (ICCW) pumps will trip and not be able to be restarted. The examinee must recognize that this should have tripped all reactor coolant pumps on interlock. The examinee must trip the reactor and then trip the reactor coolant pumps.
JPM F - The examinee will be directed to pump the Reactor Coolant Drain Tank and maintain level to above the technical specification limit.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 JPM G - A loss of offsite power has just occurred. The 1 E 4kV bus is powered from the EG-Y-1 B. When the examinee takes the watch, EG-Y-18 will trip and the examinee will have to load the SBO diesel on the 1 E 4kV bus.
JPM H - Fuel is being handled in the spent fuel pool. A fuel assembly is dropped, which cause RM-G-9 and RM-A-4 counts to rise to the alarm setpoint. The interlock fails, the examinee will have to secure ventilation in the spent fuel pool and the combined ventilation exhaust.
JPM I - A loss of instrument air occurs. The examinee will be dispatched to IC-V-4, Letdown Cooler I Reactor Coolant Pump Intermediate Closed Cooling Water Valve, to ensure it is open.
The examinee will then have to block it open.
JPM J -The examinee will have to purge RM-A-7, Waste Gas System Radiation Monitor.
JPM K - The examinee will be directed to investigate steam binding of EF-P-2A. The examinee will find that the pump is steam bound and take steps in accordance with the procedure to fix the pump.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 Facility:
Three Mile Island Date of Examination:
06/10/19 Exam Level: RO D SRO-I
~ SRO-U D Operating Test Number:
TMl2019 Control Room Systems: 8 for RO, 7 for SRO-I, and 2 or 3 for SRO-U System/JPM Title Type Code*
Safety Function
- a. 001 / Respond to a dropped control rod - ICS fails to complete D,A,S 1
run back 003M 1.02 - Control Rod System
L
- c. 006 I Lower CFT level and pressure from the Control Room - 006 D,S 3
A4.02 - Core Flood System
- e. 003 I Restore SI with a loss of ICCW 003A3.01 - Reactor Coolant D,A,S,P 4P Pump
- f. N/A
Diesel Generators
- h. 072 / Respond IAW OP-TM-MAP-C0101 Fuel Handling Incident in N,A,S 7
the Spent Fuel Pool - Radiation Monitors In-Plant Systems: 3 for RO, 3 for SRO-I, and 3 or 2 for SRO-U i.008 I Loss of Instrument Air - 008A2.05 - Intermediate Closed D,E,R 8
Cooling Water System
- j. 071 / Purge of the Waste Gas System Radiation Monitor (RM-A-7)
D,R 9
071A4.09 -Waste Gas Disposal System
- k. 061 / Respond to a failure of EF-P-2A and EF-V-30D 061 A2.04 -
D,E 4S Emergency Feedwater All RO and SRO-I control room (and in-plant) systems must be different and serve different safety functions, all five SRO-U systems must serve different safety functions, and in-plant systems and functions may overlap those tested in the control room.
- Type Codes I
Criteria for R /SR0-1/SRO-U
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 (A)lternate path (C)ontrol room (D)irect from bank (E)mergency or abnormal in-plant (EN)gineered safety feature
( L )ow-Power/Shutdown (N)ew or (M}odified from bank including 1(A)
(P)revious 2 exams (R)CA (S)imulator 4-6/4-6 /2-3
- 5 9/:5 8/:5 4
~ 1/~ 1/~ 1
~ 1/~ 1/~ 1 (control room system)
~ 1/~ 1/~ 1
~ 2/~ 2/~ 1
- ,; 3/s; 3/s; 2 (randomly selected)
~ 1/~ 1/~ 1 Random Selection: JPM E was selected by assigning the JPMs of the previous 2 years a number then using a random number generator to select the JPM.
JPM A - A rod is dropped into the core, which initiates a plant runback. The examinee will observe the plant runback but must recognize the plant did not runback to the appropriate power level. The examinee must run the plant back in manual at the ULD or SG/RX demand station. This JPM is similar to Scenario 2, Event 4. This JPM is different because of the power level in which it starts. In this JPM, the power will drop then the plant will runback. In the scenario, the power is low and control rods will pull to maintain power.
JPM B - The reactor is in a tripped state, with the loss of the 1 D 4kV bus when an RCS leak occurs. The examinee must initiate the 'B' ES, but the manual pushbuttons will fail. The examinee must manually start the 'B' ES equipment.
-JPM C - The core flood tanks are above their admin and technical specification limits. The examinee will have to lower level and pressure to within band.
JPM D - The examinee will respond to an RCS leak that is large enough to lose Subcooling Margin. The examinee will perform Rule 1, to secure reactor coolant pumps, initiate a 1600#
ES, and initiate Emergency Feedwater (EFW). Emergency Feedwater will fail to feed to the desired level (50% in the operating range), the examinee will have to take manual control of EFW and begin feeding to the desired level. This JPM is similar to Scenario 4, Event 7. This JPM provides at least 2 significant actions (critical tasks) that the simulator event does not perform. The simulator event does not have any failures with Emergency Feedwater, whereas the standard for the JPM requires the examinee to identify and mitigate an Emergency Feedwater failure.
JPM E - The examinee will be directed to restore seal injection due to a makeup pump trip.
Once seal injection is restored, Intermediate Closed Cooling Water (ICCW) pumps will trip and not be able to be restarted. The examinee must recognize that this should have tripped all reactor coolant pumps on interlock. The examinee must trip the reactor and then trip the reactor coolant pumps.
JPM G -A loss of offsite power has just occurred. The 1 E 4kV bus is powered from the EG-Y-1 B. When the examinee takes the watch, EG-Y-1 B will trip and the examinee will have to load the SBO diesel on the 1 E 4kV bus.
ES-301 Control Room/In-Plant Systems Outline Form ES-301-2 JPM H - Fuel is being handled in the spent fuel pool. A fuel assembly is dropped, which cause RM-G-9 and RM-A-4 counts to rise to the alarm setpoint. The interlock fails, the examinee will have to secure ventilation in the spent fuel pool and the combined ventilation exhaust.
JPM I -A loss of instrument air occurs. The examinee will be dispatched to IC-V-4, Letdown Cooler I Reactor Coolant Pump Intermediate Closed Cooling Water Valve, to ensure it is open.
The examinee will then have to block it open.
JPM J -The examinee will have to purge RM-A-7, Waste Gas System Radiation Monitor.
JPM K - The examinee will be directed to investigate steam binding of EF-P-2A. The examinee will find that the pump is steam bound and take steps in accordance with the procedure to fix the pump.
Appendix D Scenario Outline Form ES-D-1 ILT 18-01 NRC EXAM MATERIAL Facility:
Three Mile Island Scenario No.:
1 Op Test No.:
TM12019 Examiners:
Operators:
Initial Conditions:
85% power, MOL as ordered by the load dispatcher.
AH-E-18B is running for a surveillance (1303-5.5B)
EF-P-1 is OOS for the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Turnover:
Maintain 85% power Critical Tasks:
Shutdown reactor-ATWS (CT-24)
Restore feed to a dry OTSG (CT-26)
Event Malf. No.
Event Event Description No.
Type*
1 CH630TCRC TSCRS AH-E-18 trip CARO (ARO: Re-aligns ventilation, CRS: TS call).
2 RD10B I CRS Uncontrolled inward rod motion, entry into OP-TM-AOP-070 IURO (ATC/BOP: Manual control of ICS)
IARO 3
MU06 TSCRS MU-V-18 fails partially closed CURO (ATC: Controls pzr level with HPI) 4 FW16A CCRS
'A' MFP Trips, manual runback required RURO CARO (ATC/BOP: manual runback) 5 FW15B MCRS
'B' MFP trips, Reactor Trip with an ATWS RD28 MURO RD32 MARO 6
FW18A CCRS Sequential loss of all EFW pumps. Entry into OP-TM-EOP-004, FW18B CURO Lack of Heat Transfer.
CARO (URO: Secures RCP, ARO: Condensate Booster pump cooling)
CCRS (If required) HPI-PORV cooling, entry into OP-TM-EOP-009, HPI CURO Cooling (URO: Opens PORV)
(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility:
Three Mile Island Scenario No.:
2 Op Test No.:
TMl2019 Examiners:
Operators:
Initial Conditions:
28% power RC-P-1 B ready to start Nl-8 OOS due to a failed power supply RPS Channel 'D' is in manual bypass, RPS logic is 2 out of 3 to trip Turnover:
Start RC-P-1 B Critical Tasks:
Control HPI (CT-5)
Event Malf. No.
Event Type*
Event Description No.
1 NCRS Start RC-P-1 B IAW OP-TM-226-102 NARO (ARO: Start RCP) 2 RCR42 ICRS Pressurizer Spray Valve Failure RCR43 IURO (URO: Closes spray block valve) 3 Nl15B TSCRS Nl-6 failure (fails low)
CARO (ARO: Places RPS channel 'B' in tripped state) 4 RD0117 TSCRS Dropped rod group 7 CURO (URO: Recovers dropped rod) 5 MS02A CCRS Steam leak in RB entry into OP-TM-AOP-051 and 1102-4 RURO CARO (URO: Lower power, ARO: RB Emergency Cooling) 6 MS02A MCRS Steam line rupture in RB, Reactor Trip, OP-TM-EOP-003, XHT MURO entry.
MARO 7
FW19A CCRS EF-V-30A fails open, entry into OP-TM-424-901 CARO (ARO: Closes EF-V-2A, secures EF-P-2A)
(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor Appendix D Scenario Outline Form ES-D-1 Facility:
Three Mile Island Scenario No.:
3 Op Test No.:
TMl2019 Examiners:
Operators:
Initial Conditions:
2% power, MOL, ICS is in manual with reactivity control at the diamond Turbine Reset and all 6 Circulating Water Pumps are running for a PMT.
FW-P-18 is operating with control on the MSC Engineers are doing systems walkdowns in the control tower and turbine building.
Turnover:
Raise reactor power to 10%, initiate a bleed to the 'B' RCBT Critical Tasks:
Establish and Maintain Reactor Shutdown Requirements (CT-23)
Control HPI (CT-5)
Event No.
Malf. No.
Event Type*
Event Description 1
NCRS Raise reactor power from 3% to 10%
RURO (URO: Power ascension with ICS in Manual, ARO: Bleeds to 'B' NARO RCBT) 2 RM0323 TSCRS Reactor Building Hi Range Radiation Monitor, RM-G-23, Failure 3
RC04A I CRS Pressurizer Level Transmitter fails, entry into OP-TM-MAP-G0105, IURO OP-TM-MAP-G0205 (URO: Controls MU-V-17 in HAND) 4 ED40A TSCRS Loss of the 'D' 4kv Bus, EG-Y-1 A fails to auto start EG21A CARO (ARO: Starts EG-Y-1A) 5 CCRS Cavitating Circ Water Pump CARO (ARO: Secure cavitiating circ water pump) 6 MU07 I CRS Seal Flow Instrument Fails, RCP Seal flow High IURO (URO: Normalizes Seal Injection) 7 PLA-4-9 MCRS Circ Water Rupture, Loss of Vacuum, Reactor Trip, Entry into EOP-PLB-8-3 MURO 001, Stuck Rods MARO 8
TH06 CCRS RCS leak, PZR Level Cannot be maintained without HPI, Entry into CURO EOP-006 CARO (URO: Initiate HPI, ARO: Initiate EFW)
(N)ormal, (R)eactivity, (l)nstrument, (C)omponent, (M)ajor ES-401 PWR Examination Outline Form ES-401-2
~**1 Date of Exam: June 2019 RO KIA Category Points SRO-Only Points Tier Group K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
Total A2 G*
Total
- 1.
1 3
1 5
3 4
2 18 2
4 6
Emergency and Abnormal Plant 2
2 1
1 N/A 2
2 N/A 1
9 2
2 4
Evolutions Tier Totals 5
2 6
5 6
3 27 4
6 10 1
3 2
2 3
1 2
3 3
4 2
3 28 2
3 5
- 2.
Plant 2
1 0
1 2
1 2
2 0
0 1
0 10 2
0 1
3 Systems Tier Totals 4
2 3
5 2
4 5
3 4
3 3
38 4
4 8
- 3. Generic Knowledge and Abilities 1
2 3
4 10 1
2 3
4 7
Categories 3
3 1
3 2
2 1
2 Note: 1.
Ensure that at least two topics from every applicable K/A category are sampled within each tier of the RO and SRO-only outline sections (i.e., except for one category in Tier 3 of the SRO-only section, the "Tier Totals" in each K/A category shall not be less than two). (One Tier 3 radiation control KIA is allowed if it is replaced by a K/A from another Tier 3 category.)
- 2.
The point total for each group and tier in the proposed outline must match that specified in the table. The final point total for each group and tier may deviate by +/-1 from that specified in the table based on NRC revisions. The final RO exam must total 75 points, and the SRO-only exam must total 25 points.
- 3.
Systems/evolutions within each group are identified on the 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 0.1.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 K/A categories.
- 7.
The generic (G) K/As in Tiers 1 and 2 shall be selected from Section 2 of the K/A catalog, but the topics must be relevant to the applicable evolution or system. Refer to Section 0.1.b of ES-401 for the applicable Kl As.
- 8.
On the following pages, enter the KIA numbers, a brief description of each topic, the topics' 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 for Tier 2, Group 2. (Note 1 does not apply). Use duplicate pages for RO and SRO-only exams.
- 9.
For Tier 3, select topics from Section 2 of the K/A catalog and enter the K/A numbers, descriptions, I Rs, and point totals(#) on Form ES-401-3. Limit SRO selections to KIAs that are linked to 10 CFR 55.43.
G* Generic KIAs These systems/evolutions must be included as part of the sample (as applicable to the facility) when Revision 3 of the KIA catalog is used to develop the sample plan. They are not required to be included when using earlier revisions of the K/A catalog.
These systems/evolutions may be eliminated from the sample (as applicable to the facility) when Revision 3 of the K/A catalog is used to develop the sample plan.
ES-401 2
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emergency and Abnormal Plant Evolutions-Tier 1/Group 1 (RO/SRO)
E/APE # / Name I Safetv Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000007 (EPE 7; BW E02&E10; CE E02)
X K3.01 Knowledge of the reasons for the following as 4.0 1
Reactor Trip, Stabilization, Recovery/ 1 the apply to a reactor trip:
000008 (APE 8) Pressurizer Vapor Space X
AA2.25 Ability to determine and interpret the 2.8 2
Accident/ 3 following as they apply to the Pressurizer Vapor Space Accident: Expected leak rate from open PORV or code safety 000009 (EPE 9) Small Break LOCA / 3 X
2.4.18 Knowledqe of the specific bases for EOPs.
3 000011 (EPE 11) Large Break LOCA / 3 X
EK1.01 Knowledge of the operational implications 4.1 4
of the following concepts as they apply to the Large Break LOCA : Natural circulation and cooling, includinq reflux boilinq 000015 (APE 15) Reactor Coolant Pump X
AK3.07 Knowledge of the reasons for the following 4.1 5
Malfunctions / 4 responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) : Ensuring that SIG levels are controlled properly for natural circulation enhancement 000022 (APE 22) Loss of Reactor Coolant X
AA2.02 Ability to determine and interpret the 3.2 6
following as they apply to the Loss of Reactor Makeup/ 2 Coolant Makeup: Charqinq pump problems 000025 (APE 25) Loss of Residual Heat X
AK3.01 Knowledge of the reasons for the following 3.1 7
Removal System / 4 responses as they apply to the Loss of Residual Heat Removal System: Shift to alternate flowpath 2.1.25 Ability to interpret reference materials, such X
as graphs, curves, tables, etc.
4.2 76 000026 (APE 26) Loss of Component X
AA 1.03 Ability to operate and I or monitor the 3.6 8
Cooling Water/ 8 following as they apply to the Loss of Component Cooling Water: SWS as a backup to the CCWS 000027 (APE 27) Pressurizer Pressure X
AK2.03 Knowledge of the interrelations between the 2.6 9
Control System Malfunction / 3 Pressurizer Pressure Control Malfunctions and the followinq: Controllers and positioners 000029 (EPE 29) Anticipated Transient X
EK3.02 Knowledge of the reasons for the following 3.1 10.
Without Scram / 1 responses as the apply to the ATWS: Starting a specific charging pump X
2.4.41 Knowledge of the emergency action level 4.6 77 thresholds and classifications.
000038 (EPE 38) Steam Generator Tube X
EA 1.36 Ability to operate and monitor the following 4.3 11 Rupture/ 3 as they apply to a SGTR: Cooldown of RCS to specified temperature 000040 (APE 40; BW E05; CE E05; W E12) X AK1.04 Knowledge of the operational implications 3.2 12 Steam Line Rupture-Excessive Heat of the following concepts as they apply to Steam Transfer/ 4 Line Rupture: Nil ductility temperature X
2.4.21 Knowledge of the parameters and logic used 4.0 78 to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivitv release control, etc 000054 (APE 54; CE E06) Loss of Main X
AA2.02 Ability to determine and interpret the 4.1 13 Feedwater /4 following as they apply to the Loss of Main Feedwater (MFW): Differentiation between loss of all MFW and trip of one MFW pump
ES-401 3
Form ES-401-2 000055 (EPE 55) Station Blackout/ 6 X
EA2.01Ability to determine or interpret the following 3.7 79 as they apply to a Station Blackout: Existing valve positionino on a loss of instrument air system.
000056 (APE 56) Loss of Offsite Power I 6 X
AK1.01 Knowledge of the operational implications 3.7 14 of the following concepts as they apply to Loss of Offsite Power: Principle of cooling by natural convection.
000057 (APE 57) Loss of Vital AC X
AA2.20 Ability to determine and interpret the 3.6 15 Instrument Bus/ 6 following as they apply to the Loss of Vital AC Instrument Bus: Interlocks in effect on loss of ac vital electrical instrument bus that must be bypassed to restore normal equipment operation 000058 (APE 58) Loss of DC Power I 6 X
2.2.44 Ability to interpret control room indications to 4.4 80 verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.
000062 (APE 62) Loss of Nuclear Service X
AK3.01 Knowledge of the reasons for the following 3.2 16 Water I 4 responses as they apply to the Loss of Nuclear Service Water: The conditions that will initiate the automatic opening and closing of the SWS isolation valves to the nuclear service water coolers 000065 (APE 65) Loss of Instrument Air/ 8 X
AA 1.05 Ability to operate and / or monitor the 3.3 17 following as they apply to the Loss of Instrument Air: RPS 000077 (APE 77) Generator Voltage and X
AA2.09 Ability to determine and interpret the 3.9 81 Electric Grid Disturbances / 6 following as they apply to Generator Voltage and Electric Grid Disturbances: Operational status of emerqency diesel generators (W E04) LOCA Outside Containment/ 3 (W E11) Loss of Emergency Coolant Recirculation / 4 (BW E04; W E05) Inadequate Heat X
2.1.31 Ability to locate control room switches, 4.6 18 Transfer-Loss of Secondary Heat Sink/ 4 controls and indications, and to determine that they correctly reflect the desired plant lineup.
KIA Category Totals:
3 1
5 3
4/2 2/4 Group Point Total:
18/6 II
ES-401 4
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Emerqencv and Abnormal Plant Evolutions-Tier 1/Grouo 2 (RO/SRO)
E/APE # / Name I Safety Function K1 K2 K3 A1 A2 G*
KIA Topic(s)
IR 000001 (APE 1) Continuous Rod Withdrawal / 1 X
AA2.04 Ability to determine and 4.2 19 interpret the following as they apply to the Continuous Rod Withdrawal : Reactor power and its trend 000003 (APE 3) Dropped Control Rod / 1 X
AA 1.01 Ability to operate and / or 2.9 20 monitor the following as they apply to the Dropped Control Rod: Demand position counter and pulse/analog converter 000005 (APE 5) Inoperable/Stuck Control Rod / 1 X
AK 1. 06 Knowledge of the 2.9 21 operational implications of the following concepts as they apply to Inoperable/ Stuck Control Rod: Bases for power limit, for rod misalignment 000024 (APE 24) Emergency Boration / 1 X
2.4.30 Knowledge of events 2.7 82 related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission operator 000028 (APE 28) Pressurizer (PZR) Level Control Malfunction / 2 000032 (APE 32) Loss of Source Range Nuclear X
AK2.01 Knowledge of the 2.7 22 Instrumentation/ 7 interrelations between the Loss of Source Range Nuclear Instrumentation and the following:
Power supplies, including proper switch positions 000033 (APE 33) Loss of Intermediate Range Nuclear X
AA2.12 Ability to determine and 2.5 23 Instrumentation/ 7 interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:
Maximum allowable channel disaoreement 000036 (APE 36; BW/A08) Fuel-Handling Incidents/ 8 X
AA2.02 Ability to determine and 4.1 83 interpret the following as they apply to the Fuel Handling Incidents: Occurrence of a fuel handlino incident 000037 (APE 37) Steam Generator Tube Leak/ 3 000051 (APE 51) Loss of Condenser Vacuum / 4 000059 (APE 59) Accidental Liquid Radwaste Release / 9 000060 (APE 60) Accidental Gaseous Radwaste Release/ 9 000061 (APE 61) Area Radiation Monitoring System Alarms 17 000067 (APE 67) Plant Fire On Site / 8 000068 (APE 68; BW A06) Control Room Evacuation / 8 000069 (APE 69; W E14) Loss of Containment Integrity/ 5 X
AK1.01 Knowledge of the 2.6 24 operational implications of the following concepts as they apply to Loss of Containment Integrity:
Effect of pressure on leak rate
ES-401 5
Form ES-401-2 000074 (EPE 74; W E06 & E07) Inadequate Core Cooling/
X 2.2.49 Ability to perform without 4.6 84 4
reference to procedures those actions that require immediate operation of system components and controls 000076 (APE 76) High Reactor Coolant Activity/ 9 000078 (APE 78*) RCS Leak / 3 (W E01 & E02) Rediagnosis & SI Termination/ 3 (W E13) Steam Generator Overpressure/ 4 (W E15) Containment Flooding/ 5 (W E16) High Containment Radiation /9 (BW A01) Plant Runback / 1 X
AA2.1 Ability to determine and 3.7 85 interpret the following as they apply to the (Plant Runback):
Facility conditions and selection of appropriate procedures during abnormal and emergency operations.
(BW A02 & A03) Loss of NNI-X/Y/7 (BW A04) Turbine Trip/ 4 (BW A05) Emergency Diesel Actuation / 6 X
AA 1.3 Ability to operate and / or 3.7 25 monitor the following as they apply to the (Emergency Diesel Actuation): Desired operating results during abnormal and emeraencv situations.
(BW A07) Flooding / 8 X
2.2.22 Knowledge of limiting 4.0 26 conditions for operations and safety limits.
(BW E03) Inadequate Subcooling Margin / 4 (BW E08; W E03) LOCA Cooldown-Depressurization / 4 (BW E09; CE A13**; W E09 & E10) Natural Circulation/4 (BW E13 & E14) EOP Rules and Enclosures X
EK3.02 Knowledge of the reasons 3.2 27 for the following responses as they apply to the (EOP Rules):
Normal, abnormal and emergency operating procedures associated with (EOP Rules).
(CE A 11 **; W E08) RCS Overcooling-Pressurized Thermal Shock/ 4 (CE A16) Excess RCS Leakage/ 2 (CE E09) Functional Recovery (CE E13*) Loss of Forced Circulation/LOOP/Blackout/ 4 Point Totals:
2 1
1 2
2/2 1/2 Group Point Total:
9/4
ES-401 6
Form ES-401-2 ES-401 PWR Examination Outline Form ES-401-2 Plant S stems-Tier 2/Grouo 1 (RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Topic(s)
IR 003 (SF4P RCP) Reactor Coolant X
K4.03 Knowledge of RCPS design feature(s) 2.5 28 Pump and/or interlock(s) which provide for the following: Adequate lubrication of the RCP 2.1.31 Ability to locate control room switches, X
controls, and indications, and to determine that 4.6 54 they correctly reflect the desired plant lineup.
004 (SF1; SF2 CVCS) Chemical and X
K2.05 Knowledge of bus power supplies to the 2.7 29 Volume Control following: MOVs X
2.1.30 Ability to locate and operate 4.0 86 components, including local controls.
005 (SF4P RHR) Residual Heat X
K6.03 Knowledge of the effect of a loss or 2.5 30 Removal malfunction on the following will have on the RHRS: RHR heat exchanaer 006 (SF2; SF3 ECCS) Emergency X
A 1.09 Ability to predict and/or monitor changes 2.8 31 Core Cooling in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: Pump amperage, including start, normal and locked X
A4.02 Ability to manually operate and/or 4.0 51 monitor in the control room: Valves 007 (SF5 PRTS} Pressurizer X
K3.01 Knowledge of the effect that a loss or 3.3 32 Relief/Quench Tank malfunction of the PRTS will have on the followina:
008 (SF8 CCW) Component Cooling X
A2.04 Ability to (a) predict the impacts of the 3.3 33 Water following malfunctions or operations on the CCWS, and (b} based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: PRMS alarm X
A2.03 Ability to (a) predict the impacts of the following malfunctions or operations on the 3.2 87 CCWS, and (b} based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Hiah/low CCW temperature 010 (SF3 PZR PCS) Pressurizer X
K1.08 Knowledge of the physical connections 3.2 34 Pressure Control and/or cause-effect relationships between the PZR PCS and the following systems: PZR LCS X
K5.02 Knowledge of the operational 2.6 49 implications of the following concepts as the apply to the PZR PCS: Constant enthalpy exoansion throuah a valve
ES-401 7
Form ES-401-2 012 (SF7 RPS) Reactor Protection X
A2.07 Ability to (a) predict the impacts of the 3.2 35 following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of de control power X
K1.02 Knowledge of the physical connections 3.4 50 and/or cause effect relationships between the RPS and the following systems: 125V de svstem 013 (SF2 ESFAS) Engineered X
K3.01 Knowledge of the effect that a loss or 4.4 36 Safety Features Actuation malfunction of the ESFAS will have on the followinq: Fuel 022 (SF5 CCS) Containment Cooling X
A3.01 Ability to monitor automatic operation of 4.1 37 the CCS, including: Initiation of safeguards mode of operation X
2.2.12 Knowledge of surveillance procedures.
4.1 88 025 (SF5 ICE) Ice Condenser 026 (SF5 CSS) Containment Spray X
K2.01 Knowledge of bus power supplies to the 3.4 38 followinq: Containment spray pumps 039 (SF4S MSS) Main and Reheat X
A 1.05 Ability to predict and/or monitor changes 3.2 39 Steam in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: RCS T-ave X
A2.03 Ability to (a) predict the impacts of the 3.7 89 following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR) 059 (SF4S MFW) Main Feedwater X
K1.07 Knowledge of the physical connections 3.2 40 and/or cause effect relationships between the MFW and the following systems: ICS X
A3.03 Ability to monitor automatic operation of 2.5 53 the MFW, including: Feedwater pump suction flow pressure 061 (SF4S AFW)
X K4.06 Knowledge of AFW design feature(s) 4.0 41 Auxiliary/Emergency Feedwater and/or interlock(s) which provide for the following: AFW startup permissives X
A3.02 Ability to monitor automatic operation of 4.0 52 the AFW, including: RCS cooldown during AFW operations
ES-401 8
Form ES-401-2 062 (SF6 ED AC) AC Electrical X
K4.03 Knowledge of ac distribution system 2.8 42 Distribution design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers X
A 1.03 Ability to predict and/or monitor changes 2.5 55 in parameters (to prevent exceeding design limits) associated with operating the ac distribution system controls including: Effect on instrumentation and controls of switching power supplies 063 (SF6 ED DC) DC Electrical X
A3. 01 Ability to monitor automatic operation of 2.7 43 Distribution the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating liqhts 064 (SF6 EOG) Emergency Diesel X
K6.08 Knowledge of the effect of a loss or 3.2 44 Generator malfunction of the following will have on the ED/G system: Fuel oil storaqe tanks 073 (SF7 PRM) Process Radiation X
2.1. 7 Ability to evaluate plant performance and 4.4 45 Monitoring make operational judgments based on operating characteristics, reactor behavior, and instrument interoretation.
076 /SF4S SW) Service Water X
2.2.12 Knowledqe of surveillance procedures.
3.7 46 078 (SF8 IAS) Instrument Air X
A4.01 Ability to manually operate and/or 3.1 47 monitor in the control room: Pressure gauges X
4.2 90 2.2.25 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.
103 (SF5 CNT) Containment X
A2.03 Ability to (a) predict the impacts of the 48 following malfunctions or operations on the containment system and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation 053 (SF1; SF4P ICS*) Integrated Control
_ory Point Totals:
3 2
2 3
1 2
3 3/ 4 2
3/ Group Point Total:
28/5 2
3 ES-401 PWR Examination Outline Form ES-401-2 Plant S,stems-Tier 2/Group 2 /RO/SRO)
System # / Name K1 K2 K3 K4 K5 K6 A1 A2 A3 A4 G*
KIA Topic/s)
IR 001 (SF1 CRDS) Control Rod Drive 002 (SF2; SF4P RCS) Reactor Coolant
ES-401 9
Form ES-401-2 011 (SF2 PZR LCS) Pressurizer X
A2.03 Ability to (a) predict the impacts of the 3.8 91 Level Control following malfunctions or operations on the PZR LCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:: Loss of PZR level 014 (SF1 RPI) Rod Position X
K3.02 Knowledge of the effect that a loss or 2.5 56 Indication malfunction of the RPIS will have on the following: Plant computer 015 (SF7 NI) Nuclear X
K6.01 Knowledge of the effect of a loss or 2.9 57 Instrumentation malfunction on the following will have on the NIS: Sensors, detectors, and indicators 016 (SF7 NNI) Nonnuclear X
K4.03 Knowledge of NNIS design feature(s) 2.8 58 Instrumentation and/or interlock(s) which provide for the followinq: Input to control systems 017 (SF7 ITM) In-Core Temperature Monitor 027 (SF5 CIRS) Containment Iodine Removal 028 (SF5 HRPS) Hydrogen X
A 1.01 Ability to predict and/or monitor changes 3.4 59 Recombiner and Purge Control in parameter (to prevent exceeding design limits) associated with operating the HRPS controls includinq: Hydroqen concentration 029 (SF8 CPS) Containment Purge X
A4.01 Ability to manually operate and/or 2.5 60 monitor in the control room: Containment purge flow rate 033 (SF8 SFPCS) Spent Fuel Pool Cooling 034 (SF8 FHS) Fuel-Handling Equipment 035 /SF 4P SG) Steam Generator 041 (SF4S SOS) Steam X
K4.16 Knowledge of SOS design feature(s) 2.6 61 Dump/Turbine Bypass Control and/or interlock(s) which provide for the followina: Low main steam pressure 045 (SF 4S MTG) Main Turbine X
A 1.05 Ability to predict and/or monitor changes 3.8 62 Generator in parameters (to prevent exceeding design limits) associated with operating the MT/G system controls including: Expected response of primary plant parameters (temperature and pressure) following T/G trip 055 (SF4S CARS) Condenser Air X
2.4.8 Knowledge of how abnormal operating 4.5 92 Removal procedures are used in conjunction with EOPs.
056 (SF4S CDS) Condensate X
K1.03 Knowledge of the physical connections 2.6 63 and/or cause-effect relationships between the Condensate System and the following systems: MFW 068 /SF9 LRS) Liauid Radwaste 071 (SF9 WGS) Waste Gas X
A2.02 Ability to (a) predict the impacts of the 2.9 93 Disposal following malfunctions or operations on the Waste Gas Disposal System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Use of waste gas release monitors, radiation, gas flow rate and totalizer 072 (SF7 ARM) Area Radiation X
K5.01 Knowledge of the operational 2.7 64 Monitoring implications of the following concepts as they apply to the ARM system: Radiation theory, including sources, types, units, and effects 075 (SF8 CW) Circulatinq Water 079 /SF8 SAS**) Station Air
ES-401 10 Form ES-401-2 086 Fire Protection X
K6. Knowledge of the effect of a loss or 2.6 65 malfunction on the Fire Protection System following will have on the : Fire, smoke, and heat detectors 050 (SF 9 CRV*) Control Room Ventilation
,~int Totals:
1 0
1 2
1 2
2 Of 0 1
Of Group Point Total:
10f3 2
1
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 II Facility: TMI Date of Exam: June 2019 Category KIA#
Topic RO SRO-only IR IR 2.1.34 Knowledge of primary and secondary plant chemistry 2.7 66 limits.
2.1.36 Knowledge of procedures and limitations involved in 3.0 67 core alterations.
2.1.43 Ability to use procedures to determine the effects on 4.1 68
- 1. Conduct of reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.
Operations 2.1.5 Ability to use procedures related to shift staffing, such as 3.9 94 minimum crew complement, overtime limitations, etc.
2.1.8 Ability to coordinate personnel activities outside the 4.1 95 control room.
Subtotal 3
2 2.2.17 Knowledge of the process for managing maintenance 2.6 69 activities during power operations, such as risk assessments, work prioritizations, and coordination with the transmission system operator 2.2.42 Ability to recognize system parameters that are entry-3.9 70 level conditions for Technical Specifications.
- 2. Equipment 2.2.14 Knowledge of the process for controlling equipment 3.9 71 Control configuration or status.
2.2.6 Knowledge of the process for making changes to 3.6 96 procedures.
2.2.19 Knowledqe of maintenance work order requirements.
3.4 97 Subtotal 3
2 2.3.15 Knowledge of radiation monitoring systems, such as 2.9 72 fixed radiation monitors and alarms, portable survey instruments, personnel monitorinq equipment, etc.
2.3.12 Knowledge of radiological safety principles pertaining to 3.7 98 licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to
- 3. Radiation locked hiah-radiation areas, alianinq filters, etc.
Control Subtotal 1
1 2.4.3 Ability to identify post-accident instrumentation.
3.7 73
- 4. Emergency 2.4.50 Ability to verify system alarm setpoints and operate 4.2 74 Procedures/Plan controls identified in I the alarm response manual.
2.4.2 Knowledge of system set points, interlocks and 4.5 75 automatic actions associated with EOP entry conditions
ES-401 Generic Knowledge and Abilities Outline (Tier 3)
Form ES-401-3 2.4.37 Knowledge of the lines of authority during 4.1 99 implementation of the emerqencv plan.
2.4.9 Knowledge of low power/shutdown implications in 4.2 100 accident (e.g., loss of coolant accident or loss of residual heat removal) mitiqation strateqies.
Subtotal 3
2 Tier 3 Point Total 10 7
ES-401 Record of Rejected K/ As Form ES-401-4 Tier/
Randomly Reason for Rejection Group Selected KIA 1 / 1 056 I AK 1.03 Q14: Calculation of SCM does not change with or without offsite power based on safety grade instrument usage. No requirement for the use of steam tables exists. Cannot make an operationally valid question to meet the KA Replaced with 056 AKl.01 1 / 1 BW E04 I 2.4.49 Q18: No immediate actions are required on a Lack of Heat Transfer.
Cannot make an operationally valid question to meet the KA Replaced with KA BW E04 2.1.31.
1/ 2 033 I AA2.03 Q23: No fuse exists in the Intermediate Range Nuclear Instrument strings.
Cannot make technically accurate question and meet the KA Replaced with033 AA2.12.
2 I l 004 I K2.07 029: Power suooly to heat tracing is minutia. Replaced with 004 KA 2.05 2 I l 061 /K4.13 Q41: Cooling water and lube oil to the EFW pumps at Three Mile Island are not active systems. There is no start signal to initiate cooling water or lube oil to the pumps. Replaced with 061 K4.06.
3 2.2.37 Q69: Operability determinations are SRO only knowledge. Replaced with 2.2.17.
3 2.4.22 Q75: Safety functions are SRO only knowledge. Replaced with 2.4.2.
1/ 1 025 I 2.4.34 Q76: No RO task outside of the control room exists that is exclusively SRO knowledge on a loss ofDHR. Replaced with 025 2.1.25.
1 / 1 040 I 2.4.40 Q78: Could not make a question with operational validity that applies system technical specifications during a steam line rupture. Replaced with 040 KA 2.4.21 1 / 1 077 I AA2.0l QS 1: Could not make a question with appropriate level of difficulty for an initial license test. Replaced with 077 AA2.09.
1/2 024 I 2.4.31 Q82: Could not make an SRO-only question based on this generic K/ A Replaced with 024 KA 2.4.30 1/ 2 074 I 2.2.36 Q84: Could not make a question with operation validity that requires analyzing maintenance activities and Inadequate Core Cooling. Replaced with 074 2.4.49.
2/2 011 / A2.06 Q91: Overlap with Scenario 2 event. Replaced with O 11 A2.03 2/2 071 / A2.07 Q93: Could not find an adequate tie between a waste gas release and a loss of the meteorological tower. Replaced with 071 A2.02