ML031710800
| ML031710800 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 05/14/2003 |
| From: | Gumbert R AmerGen Energy Co |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 50-289/03-301 50-289/03-301 | |
| Download: ML031710800 (70) | |
Text
S-301 Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Iacility: Three Mile Island Unit 1 Date of Examination: May 12,2003 Exam Level (circle one): RO / SRO(I) i SRO(U) Operating Test No,:
B . l Control Room Systems SystemNPM Title I Type Code*
I safety Function
- a. Chemical and Volume Control (004)iPerform an Emergency Boration (Alt. Path - Backup Emergency Boration Required). I I 1
- b. Engineered Safety Feature Actuation Systems (013)IRespond to inadvertent ES Actuation. I I 2 3
- c. Emergency Core Cooling System (OOG)/Respond to a High N, A, S Pressure Injection (HPI) initiation (Alt. Path - MU-V-14A fails).
- d. Residual Heat Removal System (0OS)iRespond to a failure of Low Pressure Injection (Alt. Path - DHV-6 Fails to Open).
1 N,A, S, L 1 4~rimary
- e. Main Steam System ( 03 9
)
/ D, S 4 Secondary pvLdrnfJi 5 Toid vn 1 vt>
TCJ r
- f. Containment Cooling System (022)iReturn Reactor Building (RB) N,S 5 Emergency Cooling to Engineered Safeguards Standby.
- g. Emergency Diesel Generator (EDG) System (064)iEDG D, A, s 6 Operation (Alt. Path - EDG Fails to Auto Load).
- a. Chemical and Volume Control System (004)IManually Open RCP N, R 2 Seal Injection Isolation Valve (MU-V-26). Emergency
- b. Pressurizer Pressure Control System (01O)/Transfer Pressurizer Heater Group 8 or 9 to an Engineered Safeguards Bus.
D lI 3 Emergency
- c. Emergency Feedwater System (061)/Local Reset of Emergency D 4 Secondary Feedwater Pump (EF-P-1). Emergency
April 17, 2003 U.S. NRC Region I Administrator ATTN: Joseph DAntonio 475 Allendale Road King of Prussia, PA 19406
Subject:
Post NRC Validation Visit Submittal of Senior Reactor Operator Examination Materials Three Mile Island Unit 1 (Docket ## 50-289)
This submittal supports the initial license examination scheduled for the week of May 12, 2003.
In accordance with NUREG 1021, Revision 8, Section ES-201, please ensure that these materials are withheld from public disclosure until after the examinations are complete.
Should you have any questions concerning this letter or the examination materials, please contact Dennis May at (717) 948-2074/2093.
Respectfully, Dennis May ITraining Center
Enclosures:
(Sent by overnight delivery directly to Joseph DAntonio, Chief Examiner, NRC Region I) 0 Revised exam questions (with technical references for NEW questions) 0 Exam open reference materials (list)
Revised sample plan 0 Revised Simulator Scenarios 0 Replacement Simulator JPM 0 Simulator Scenario EPs/APs
4/9/2003 SRO-21 NRC Written Exam Review by NRC Attendees:
Dennis May Dave Atherholt Joseph DAntonio (NRC)
Dell McNeil (NRC)
Question # Initial Comment Resolution Official NRC status 3 too easy keep as is ok 5 too easy REWRITTEN NEW rewrite new 8 too easy keep as is ok 14 SRO, but no mark marked 55.43 box NC 17 one distractor not plausible keep as is (system explained) NC 18 Q logic error keep as is (reading error) NC 20 KA: not appl at TMI changed to similar KIA ok KIA per NRC 24 too easy keep as is ok 27 RO, but 55.43 checked agreed with our KA interpret. NC 30 not meeting KA keep as is (misunderstood Q) NC 31 2 ans. parts, but need only 1 rewrite distractors with like parts partial rewrite 34 NOT comprehensive changed to MEMORY ok 42 needs reference supplied already marked in database NC 44 NOT memory change to COMPREHENSIVE ok 45 two reviewers disagreed no comment NC 54 marked RO, should be SRO changed to SRO ok 55 marked SRO, should be RO changed to RO ok 58 too easy keep as is ok 68 KA: not appl at TMI applied well, but scrub KA later ok 69 too easy keep as is ok 71 KA interpretation differs Keep Q, assign appl KA per NRC ok KIA per NRC 78 too easy rewrite new rewrite new 80 NOT comprehensive changed to MEMORY ok 85 marked SRO, should be RO changed to RO ok 88 needs rewording rewritten with NRC help rewritten & OKd 89 marked RO, should be SRO Kept SR0,reordered 2 distracters NC 90 SRO, but 55.43 checked checked 55.43 box ok 96 not meeting KA keep as is (misread Q) NC 97 marked SRO, should be RO Kept SRO NC
- NC = No Comment
ROlSRO Importance Rating 2.5 3.4*
- Group #
Knowledge of the process for controlling temporary changes. (Equipment Control)
Identify the ONE operation below that is subject to the requirements of AP 1013, Temporary Modifications and Bypass of Safety Functions.
A. Using a power buggy to energize a portable sump pump in the Amertap pit.
B. Installation of a pipe cap on the outlet of a drain valve to stop leakage to the floor at 'A' 12th stage heater.
C. Bolting a stainless steel collar around a valve stem to prevent vibration induced damage.
D. Installation of a calibrated test gauge to support performance of a Tech Spec surveillance on a temporary basis.
1013, Temporary Modifications and Bypass of Safety Functions, pages 1-4, Rev 51.
L New TMlBank TMI Question # #5 SR021 AUDIT 0 Modified TMI Bank Parent Question #
&?IMemory or Fundamental Knowledge 3 Comprehension or Analysis
& ?I . I O 55.41 d 55.43 .3 lid 55.45 . I 3 A Incorrect answer. This is a temporary installation that is excluded from 1013 control in section 2.2.3.4, page 4.
B Incorrect answer. This is a temporary installation that is excluded from 1013 control in section 2.2.3.6, page 4.
C Correct answer. This is a temporary installation that is NOT excluded from control of 1013 swer. Excluded in section 2.2.3.2, page 3.
TMI SRO Exam - May 2003 Thursday,April 17,2003
SYSiEP# Gen KA# 2.2.19 Page # 2-7 Tier # -3 ROlSRO Importance Rating 2.1 -
3.1 Group #
Knowledge of maintenance work order requirements. (Equipment Control)
Plant conditions:
- A maintenance work order requires a pneumatic operated valve to be closed as part of the clearance order safety boundary.
- The valve fails OPEN on loss of air or loss of power to the solenoids controlling the air.
Based on these conditions, identify the ONE selection below that completes the following phrase:
This valve may be used as part of the safety boundary if . ..
A. the power supply to the solenoids is tagged to ensure a continuous power supply.
B. the fluid controlled by the valve is less than 200°F and less than 500 psig.
C. a temporary air bottle is installed to ensure a continuous air supply.
D. an appropriate gag is used on the valve operator.
OP-MA-109-101, Clearance and Tagging, page 22, Rev. 1.
None.
C New 2l TMI Bank TMl Question # XQR5A01 Q02 C Modified TMI Bank Parent Question #
E Memory or Fundamental Knowledge C Comprehension or Analysis C 55.41 3 55.43 .5 & 55.45 .I3 A Incorrect answer. This does not ensure a fail-safe mode for ensuring the integrity of the pressure boundary.
B Incorrect answer. This is a mis-application of double valve isolation criteria described in OP-MA-109-101.
C Incorrect answer. This approach is not allowed by procedure OP-MA-109-101.
D Correct answer. In accordance with 1002.1 section 4.8.5, application of an appropriate gag is satisfactory if another valve can not be used for the pressure boundary to ensure personnel and plant safety.
None.
TMI SRO Exam May 2003 Thursday,ApriI 17,2003
Q # 088 Page ## 2-7 Tier # -3 ROlSRO Importance Rating 2.5 3.7
- Group #
Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (Equipment Control) 3 3 El Choose the one consequence of being inside of the RESTRICTED REGION of the COLR rod insertion limits at power. (see attached graph)
A. Potential for inadequate shutdown margin.
B. Potential inadequate reactivity insertion during a runback.
C. Potential for exceeding CHF during low probability transients.
D. Potential ejected rod worth greater than that assumed by analyses.
Technical Specifications COLR Rev 0, abstract section.
(Graph of COLR rod insertion limits must be given as part of question)
E New E TMlBank TMI Question #
P Modified TMI Bank Parent Question #
Memory or Fundamental Knowledge C Comprehension or Analysis a 55.43 .2 iZ 55.45 .I3 A Incorrect answer. Plausible distracter since this is a function of the Not Allowed Region rod insertion limit.
B Incorrect answer. Plausible distracter since this is a function of negative reactivity availability.
C Incorrect answer. Plausible distracter for candidate who assumes CHF is part of bases.
wer. Refer to TS page 3-35a, Amendment 211.
question rewritten per NRC review, and re-approved.
TMI SRO Exam - May 2003 Thursday, April 17,2003
SYS/EP# Gen KA# 2.1.5 Page # 2-1 Tier # -
3 ROlSRO Importance Rating 2.3 -
3.4 Group # -1 Ability to locate and use procedures and directives related to shift staffing and activities.
(Conduct of Ops)
Identify the ONE selection below that describes a condition when it is permissible for a Reactor Operator to be the ONLY NRC LICENSED person in the Control Room.
A. Reactor power is 100%;
There is one other person, a CRO trainee (NOT LICENSED) in the Control Room.
B. Plant is in Hot Shutdown condition; The Shift Technical Advisor (NOT LICENSED) is also in the Control Room.
C. RCS temperature is 210°F; The duty Shift Manager (LICENSED) is in the Shift Manager's Office.
D. RCS temperature is 189°F; The duty Shift Manager (LICENSED) is in the Operations Office Building.
Technical Specification 6.2.2.2.d, page 6-1, Amendment 219.
Technical Specification 6.2.2.2.a, page 6-1, Amendment 219.
Technical Specification Table 6.2-1, page 6-2, Amendment 219.
Technical Specification 6.2.2.2.d, page 6-1, Amendment 219.
Technical Specification 6.2.2.2.a, page 6-1, Amendment 219.
Technical Specification Table 6.2-1, page 6-2, Amendment 219.
V.A. 10.08
. New 3' TMlBank TMI Question # June2001 SRO Audit #84 E Modified TMI Bank Parent Question #
d Memory or Fundamental Knowledge C Comprehension or Analysis
& 55.41 .IO E 55.43 .I E 55.45 . I 2 A Incorrect answer. Plausible misconception that requirement is for other licensed personnel to be ON SITE rather than in the Control Room.
B Incorrect answer. Plausible misconception since reactor is shutdown, and STA is in the Control Room.
C Incorrect answer. Plant is above 200"F, and licensed SRO is required to be in the Control Room.
D Correct answer. Plant is <2OO0F.
None.
TMI SRO Exam - May 2003 Thumduy,April 17,2003
Form ES-401-6 Page # 3.7-16 Tier # -
2 ROISRO Importance Rating 3.2 -
3.5 Group # 2 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Process Radiation Monitoring (PRM) System controls including:
Radiation levels.
Plant conditions:
- Reactor power is loo%, with ICs in full automatic.
- RM-L-1, RCS Letdown High Range Monitor, is out of service for calibration testing.
Event:
- Fuel pin failure occurs.
Based on these conditions, assuming no operator action, complete the statement below that predicts the effect of this event.
Auxiliary Building general area radiation levels...
A. RISE until letdown automatically isolates.
B. RISE with NO automatic letdown isolation.
C. DO NOT CHANGE since MU-V-1NB are required to be closed before any RM-L-1 testing.
D. DO NOT CHANGE since MU-V-2NB are required to be closed before any RM-L-1 testing.
EP 1202-12, High Radiation Levels, page 11, Rev. 50.
None.
IV.E.06.04 b New 3 TMlBank TMI Question #
3 Modified TMI Bank Parent Question #
2l Memory or Fundamental Knowledge 3 Comprehension or Analysis g 55.43 .4 c 55.45 A Incorrect answer. Plauseible that the Examinee might think letdown isolation on high radiation is initiated by RM-L-1 Lo, which is still in service.
B Correct answer.
C Incorrect answer. Plausible distracter since the monitor that initiates automatic letdown isolation is out of service during this test.
D Incorrect answer. Plausible distracter since the monitor that initiates automatic letdown isolation is out of service during this test.
None.
TMI SRO Exam - May 2003 Thursday,April 17,2003
Form ES-401-6 SYS/EP# 072 KA# A1.O1 Page # 3.7-14 ROlSRO Importance Rating 3.4 3.6
- Group # -1 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Area Radiation Monitoring (ARM) System controls including:
Radiation levels.
Plant conditions:
- Plant is in HOT SHUTDOWN, initial post-refueling plant startup is in progress.
- RM-G-16 (OTSG A Sample line monitor) is TURNED OFF for maintenance.
- Compensatory actions for RM-G-16 have been taken IAW procedures.
- OTSG tube leakage is at baseline value.
- OTSG A sample is in progress.
Sequence of events:
- I and C technician takes RM-G-16 control switch to ALL position IAW the gamma monitor startup section of 1105-8 Radiation Monitoring System.
Based on these conditions, predict the system response.
A. (1) RM-G-16 alarms due to process flow radiation levels.
(2) AI OTSG sample valves AUTO-CLOSE.
B. (1) RM-G-16 alarms due to process flow radiation levels.
(2) A OTSG sample valves REMAIN OPEN.
C. (1) RM-G-16 alarms due to electronic power spike.
(2) A OTSG sample valves AUTO-CLOSE.
D. (1) RM-G-16 alarms due to electronic power spike.
(2) AI OTSG sample valves REMAIN OPEN.
1105-8 Rev 67, pages 6 and 7,38 and 3 9 1 None.
IV.B.01.06
@ New 2 TMI Bank TMI Question #
C Modified TMI Bank Parent Question #
C Memory or Fundamental Knowledge M Comprehension or Analysis
< 55.41 .5 $ 55.43 .4 hk 55.45 .5 A Incorrect: Both parts wrong. (see D)
B Incorrect: First part wrong, second part correct (see D)
C Incorrect: First part right, second part wrong (see D)
D Correct: Initial power-up of area rad monitors causes electronic spiking into alarm state, and the sample valves will NOT close due to interlocks in BYPASS per procedure. (note: no RCS fuel damage expected immediately post refuel - no process rad levels) rewritten new per NRC review.
TMI SRO Exam - May 2003 Thursday,Aprii 17,2003
Number TMI - Unit 1
- Title Operating Procedure I105-8 Revision No.
Radiation Monitoring System 67 2.2 Administrative 2.2.1 The radiation monitoring equipment has power supplies which produce high internal and external voltages. (Up to 2500 volts) Care should be taken when working near this equipment to avoid electrical shock.
2.2.2 A Radiation work permit must be issued when using radioactive sources to calibrate the detectors as per Rad Con Procedures.
3.0 OPERATING PROCEDURE 3.1 Place Area Gamma Monitors In Service - Level I 3.1.I Controls All controls necessary to operate the area gamma monitoring system are located on the Vertical Panel PRF in the Control Room (except ALC-RMI-10 and ALC-RMI-11 that are located in the CCB Radwaste Panel, CC-CP-1 in the Chemical Cleaning Building). In addition, all area gamma monitors possess one or more local readout and/or alarm modules near the vicinity of the detector.
3.1.2 Prerequisites
- a. Verify the area gamma monitors shall have been calibrated in accordance with Surveillance Procedures 1302-3.1, 1302-15, 1302-17.2, IC-I77 and/or 1302-17.3, as confirmed by contact with I&C Supervisor/Foreman, or Surveillance file verification.
Name Date 3.1.3 Procedure
- a. Ensure power to the Vertical Panel PRF by verifying closed or closing the following breakers:
VBA Breaker No. 2 VBB Breaker No. 2 VBC Breaker No. 2 VBD Breaker No. 2 AB-E Breaker No. 15 (Recorder Power)
Misc Power Panel in CCB Control Room Breaker No. 15 6
Number TMI - Unit 1 Operating Procedure 1105-8 Title Revision No.
-Y Radiation Monitoring System 67
- b. Ensure the bypass switch on the control panel "PRF" is in the defeat position to preclude automatic interlock actuation while placing the monitor in service for the following monitors.
RM-G-9 RM-G-I 8 RM-G-I6 RM-G-20 RM-G-17 RM-G-21 C. Ensure the rotary switch on each individual control panel module is selected to the "ALL" position
- d. Verify that the green "FAIL" light comes to indicate power is on.
- e. Ensure the electronics warm up for 30 minutes for guaranteed accuracy.
- f. Verify proper operation by checking "Power Available green light on" 0 Note it in the first column on Table 4.
Verify proper channel operation by performing source check for each monitor per Surveillance Procedure 1301-4.1 Note "Satisfactory Check Source Check" on Table 4 CAUTION Reading Alarm setpoints from the meter face by pressing the alarm pushbutton does not give an accurate indication of actual setpoint. This feature is not calibrated and should not be relied upon for accurate setpoint information. ~~~~~~
- h. Verify the alert and high alarm setpoints by directing I&C Department to:
REFER TO applicable sections of 1302-3,l IC-177 OR verify procedure and/or surveillance are on file for:
0 1302-3.1 0 IC-I 77 I. Complete the "Satisfactory Alert Setpoint Check" AND "Satisfactory Alarm Setpoint Check" columns in Table 4.
7
TMI - Unit 1 Operating Procedure 1105-8 Title Revision No.
Radiation Monitoring System 67 Tech Spec or ODCM Instrument Requirements Compensatory Action Interlock RM-G-11 None None None MU Demin Area AB 305 RM-G-12 I None None None I Solid Radwaste Processing Area AB 305 RM-G-13 None None None Aux Bldg Entrance AB 281 RM-G-14 Waste Evap Area AB 281 None None None I
RM-G-15 I None None None II Hx Vault AB 271
Interlock I Closes CA-V-4A & 5A I None
'*Note 1 Switch in DEFEAT (PRF) I RM-G-I8 I None *"Note 1 Interlock Closes CA-V-48 & 5 8 Switch in DEFEAT ,
Interlock Closes CA-V-I ,2,3 & 13 i RM-G-19 I None Verify RM-L-1 Lo is operable None I RCP seal return line RM-G-20 I None **Note 1 Switch in DEFEAT RCDT return (PRF)
Interlock Closes WDL-V-303 & 304 Closes WDG-V-3 & 4 RM-G-21 None **Note 1 Switch in DEFEAT RB sump return (PW Interlock Closes WDL-V-534 & 535 RM-G-22 TS table 3.5-3 Per Tech Spec None RB Inside A D-ring RM-G-23 TS table 3.5-3 Per Tech Spec RB Inside B D-ring RM-G-26 I TS table 3.5-3 Per Tech Spec None OTSG
_ _ _ _ A steam line RM-G-27 1 TS table 3.5-3 Per Tech Spec None OTSG B steam line
- NOTE 1: None, if ESAS closure on RB pressure and reactor trip are operable.
Otherwise a dedicated operator will be assigned to the valves whenever they are opened.
NOTE 2: Ensure alternate [RM-G-6 or 71 is operable prior to planned monitor outage. If RM-G-6 and 7 are out of service, then NOTIFY TSC Technical Director to ensure core damage assessment team is prepared to use alternate methods.
39
Number TMI - Unit 1 Operating Procedure I105-8 Title Revision No.
- -1 Radiation Monitoring System 67 3.5 Removing an Area Monitor From Service Level 1-NOTE If an instrument fails, perform this section. For a failed instrument, compensatory actions are not a prerequisite but must be completed.
3.5.1 Review TS & ODCM requirements (refer to Table 3.5).
3.5.2 Perform compensatory action required per Table 3.5.
3.5.3 Make a log entry which identifies the inoperable instruments and compliance with Tech Spec or ODCM requirements.
3.5.4 Place interlock mode switch in DEFEAT. Refer to Table 3.5 for switch location and interlock description.
3.5.5 Place monitor in OFF 3.5.6 When appropriate, return monitor to service IAW Section 3.1 Tech Spec or ODCM Instrument Requirements Compensatory Action Interlock ALC-RMI-10 None None None CCB Area (between CC-T-1 &
CC-T-2)
ALC-RMI-11 CCB Area (above CCB sump)
None I None I None RM-G-1 None None None Control Room RM-G-2 None None None Rad Chem Lab RM-G-3 None None None Nuc Sampling Room RM-G-4 None None None Hot Machine Shop RM-G-5 None None None RB Personnel Access Door RM-G-6 TS 3.8.1 Per Tech SDec None RB Aux FH Bridge RM-G-7 TS 3.8.1 Per Tech Spec None RB Main FH Bridge
Interlock Trips AH-E-IO Closes AH-D-120,121 8,122 RM-G-10 None None None Aux Bldg Entrance AB 305
Form ES-401-6 Q # 071 SYSIEP# 063 KA# K3.01 Page # 3.6-6 Tier # -
2 ROlSRO Importance Rating 3.7 4.1 Group # -
I Knowledge of the effect that a loss or malfunction of the DC Electrical Distribution System will have on: ED/G Plant conditions:
- Reactor power is loo%, with ICs in full automatic.
- MU-P-1A (Makeup Pump 1A) is supplying normal makeup/seal injection.
Sequence of events:
- The following alarms, actuate simultaneously:
- A-1-7 Battery 1A Discharging
- A-2-7 Batt Charger 1AllCIIE Trouble
- A-3-7 Inverter 1N1C/1E System Trouble
- PRFI-1-1 CRDM Breaker Test Trouble
- "-3-1 230 KV Substation Trouble
- AA-3-2 7KV Bus Trouble
- AA-3-5 480V BOP Bus Trouble Based on these conditions identify the ONE selection below that describes (1) the controlling procedure and (2) required actions.
A. (1) 1202-9A, Loss of "A' DC Distribution System (2) Notify Auxiliary Operator to close EG-V-I5A, air start isolation for Emergency Diesel Generator I A .
B. (1) 1202-9A, Loss of "A' DC Distribution System.
(2) Notify Auxiliary Operator to verify MU-P-1C is ready for start C. (1) Alarm response for AA-3-2, 7KV Bus Trouble.
(2) Notify Transmission System Operator (TSO) and trip the reactor D. (1) Alarm response for "-3-1, 230 KV Substation Trouble.
(2) Notify Transmission System Operator (TSO) and trip the reactor EP 1202-9A, Loss of "A" DC Distribution, page 4, Rev. 43.
g New c-1 TMlBank TMI Question #
-2 Modified TMI Bank Parent Question #
3 Memory or Fundamental Knowledge Comprehension or Analysis VI I 55.43 .5 d 55.45 .6 A Correct answer.
B Incorrect answer. Part #is Icorrect procedure, but part #2 is not correct action.
C Incorrect answer. Incorrect procedure, and incorrect actions.
D Incorrect answer. Incorrect procedure, and incorrect actions.
TMI SRO Exam - May 2003 Thursday, April 17,2003
Form ES-401-6 Q # 071 Kept question and reassigned more applicable KA per NRC (validation)
TMI SRO Exam - May 2003 Thursduy, April 17, 2003
SYSIEP# 011 KA# A2.08 Page # 3.2-23 Tier # 2 ROISRO Importance Rating 2.6 2.8
- Group # -
2 Ability to (a) predict the impacts of the following malfunctions or operations on the Pressurizer Level Control System (PZR LCS) and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of level compensation.
Initial conditions:
- Reactor power loo%, with ICs in full automatic.
- Pressurizer level and temperature are normal.
Sequence of events:
- Temperature compensation to the selected Pressurizer level instrument fails LOW.
Based on these conditions, identify the ONE statement below that describes:
( I ) Automatic response of the Pressurizer level control system; (2) Immediate manual actions required.
A. (1) Makeup valve MU-V-17 OPENS to loo%, Pressurizer heaters DO NOT trip.
(2) Transfer MU-V-17 to manual and adjust to maintain Makeup Tank level constant.
- 6. ( I ) Makeup valve MU-V-17 OPENS to loo%, Pressurizer heaters trip.
(2) Raise RCS Letdown flow to maximum (140 gpm).
C. ( I ) Makeup valve MU-V-17 CLOSES to 0%, Pressurizer heaters energize in response to actual Pressurizer level reduction.
(2) Isolate RCS Letdown flow.
D. (1) Makeup valve MU-V-17 CLOSES to 0%, Pressurizer heaters energize in response to actual Pressurizer level reduction.
(2) Raise RCP Seal injection flow to compensate for reduced Makeup flow.
EP 1202-29, Pressurizer System Failure, pages 14 and 19, Rev. 59.
None.
V.D.11.01 C New ?I TMlBank TMI Question # QR5D11-03-QOI
- Modified TMI Bank Parent Question #
7 Memory or Fundamental Knowledge 3 Comprehension or Analysis g 55.41 .5 Y5 55.43 .5 g 55.45 .3/.13 A Correct answer, IAW EP 1202-29, Pressurizer System Failure.
B Incorrect answer. Impact of temperature compensation at these conditions will not reduce indication below 80-inch low level cut-off interlock.
C Incorrect answer. Failed value will be less than normal 220-inch level, therefore valve will open. Action described is not in accordance with EP 1202-29.
D Incorrect answer. Failed value will be less than normal 220-inch level, therefore valve will open. Action described is not in accordance with EP 1202-29.
TMI SRO Exam - May 2003 Thursday, Aprii 17,2003
SYSIEP# 008 KA# 2.2.22 Page # 2-7 Tier # 2 ROlSRO Importance Rating 3.4 -4.1 Group # -
3 Knowledge of limiting conditions for operations and safety limits: Component Cooling Water System (CCWS) 2l B Identify the ONE selection below that completes the description of the basis for the limiting conditions for operation (LCOs) for Nuclear Services Closed Cooling (NSCC).
(a) NSCC pump(s) islare required for normal operation heat loads; (b) NSCC purnp(s) islare required for ECCS support during a LOCA.
A. (a) Two (b) Two B. (a) Two (b) One C. (a)One (b) Two D. (a) One (b) One Technical Specifications page 3-24, Amendment 227.
- 2. New TMlBank TMI Question ##
C Modified TMI Bank Parent Question #
9 Memory or Fundamental Knowledge C Comprehension or Analysis
- 55.41 k 55.43 .2 2 55.45 .2 A Incorrect answer. First part is correct that two are required for normal operations, but only one is needed for ECCS support (incorrect second part).
B Correct answer, in accordance with Tech Spec bases..
C incorrect answer. First part is incorrect (two are required for normal operations), but only one is needed for ECCS support (correct second part).
D Incorrect answer. First part is incorrect (two are required for normal operations), but only one is needed for ECCS support (correct second part).
None.
TMI SRO Exam - May 2003 Thursday,April 17,2003
ROlSRO Importance Rating 3.7 4.2 Group # -1 Knowledge of the effect of a loss or malfunction of the following will have on the Control Rod Drive System: Reactor trip breakers, including controls Plant conditions:
- Reactor power is 100%.
- RPS surveillance testing in progress.
- ICs stations in manual:
- FW Loop masters A and B.
- Delta TC.
- Reactor Master control station.
- Steam Generator Reactor Master.
- Diamond Rod Control panel.
- CRD power supply breaker associated with 'B' RPS cabinet is open, and will not reclose.
- Repair parts will take two days to arrive.
Based on these conditions, AUTOMATIC CRD Diamond Panel control can A. be established, with normal CRD IN/OUT motion control.
B. be established, however CRD OUT motion will be inhibited.
C. NOT be established, due to MOTOR FAULT condition existing.
D. NOT be established, due to SYSTEM POWER FAULT condition existing.
OP 1105-9, Control Rod Drive System, Section 4.4.3 Auto Inhibit, page 74, Rev. 61.
C Modified TMI Bank Parent Question ##
Memory or Fundamental Knowledge E! Comprehension or Analysis v 55.41 .7
- c 55.43 2-55.45 .7 A Correct answer - although only one side of CRD is powered, normal ops is possible.
B Incorrect answer - no inhibit condition exists.
C Incorrect answer - no MOTOR FAULT condition exists.
D Incorrect answer - although power fault exists, it does not prevent normal auto motion.
None.
TMI SRO Exam - May 2003 Thursday, April 17, 2003
KA# AA1.l Page # 4.3-31 ROlSRO Importance Rating 4.0 4.0 Group # -I Ability to operate and/or monitor the following as they apply to Loss of NNI-Y: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.
v D Initial Conditions:
- Reactor power is loo%, with ICs in full automatic.
- ICs Power Supplies are in their normal line-up.
Event:
- LOSSOf BUSATA.
Identify the ONE statement below that describes automatic equipment response to this event, and the reason for the response.
A. MU-V-5 controller fails to the mid position due to loss of ICs HAND Power.
B. MU-V-5 controller fails to the mid position due to loss of ICs AUTO Power.
C. MS-V4NB control transfers to the Back-up Loaders due to loss of ICs HAND Power.
D. MS-V4NB control transfers to the Back-up Loaders due to loss of ICs AUTO Power.
EP 1202-42, Total or Partial Loss of ICSlNNl Auto Power, Pages 2 and 3, Rev. 38.
None.
V. D.22.04 New 2 TMI Bank TMI Question ## #20 7/2001 SRO L- Modified TMI Bank Parent Question #
Memory or Fundamental Knowledge Comprehension or Analysis 9 55.41 .? r_ 55.43 2 55.45 .5/.6 A Incorrect - MU-V-5 fails to mid position on loss of HAND power. AUTO power was lost.
B Incorrect - MU-V-5 fails to mid position on loss of HAND power. AUTO power was lost.
C Incorrect - Plausible, controllers swap to BU loader, but only on loss of AUTO power.
D Correct answer - Requires manual control via BU loader.
None.
TMI SRO Exam - May 2003 Thursrhy, April 17, 2003
Page # 4.1-17 Tier # 1 ROlSRO Importance Rating 4.0 4.6 Group # 1 Ability to determine and interpret the following as they apply to Inadequate Core Cooling:
Changes in PZR level due to PZR steam bubble transfer to the RCS during inadequate core cooling.
Initial plant conditions:
Reactor tripped from 100% power due to loss of off-site power (LOOP).
Emergency Feedwater Pump EF-P-2B tripped.
One Makeup Pump operating.
Pressurizer level = 100 inches, controlled in automatic.
Core exit thermocouple temperature is steady at 570°F.
RCS pressure steady at 2100 psig.
OTSG pressures = 1000 psig.
OTSG levels at 12% Operating Range, slowly rising.
OP-TM-EOP-001, Reactor Trip, Immediate Actions complete.
Initial post trip Symptom Check has been completed.
Sequence of events:
- PORV opened unexpectedly, and failed to reclose.
- RC-V-2 control power fuse failed during attempt to close PORV block.
- RCS pressure rapidly reduced to 1680 psig, and now is slowly lowering to 1660 psig.
- At the end of the pressure reduction, Pressurizer level rose rapidly to 300 inches, and is now rising slowly.
HPI has NOT been actuated at this time. Based on these conditions, identify the ONE set of statements below that describes (1) the reason for the Pressurizer insurge and (2) required actions.
A. (1) Displacement of water from under the RV head due to steam bubble formation.
(2) Continue with EOP-001 VSSVs.
B. (1) Displacement of water from under the RV head due to steam bubble formation.
(2) Exit EOP-001 and GO TO OP-TM-EOP-009 HPI Cooling - Recovery From Solid Operations.
C. (1) Expansion of RCS loop water due to depressurization..
(2) Continue with EOP-001 VSSVs.
D. (1) Expansion of RCS loop water due to depressurization.
(2) Exit EOP-001 and GO TO OP-TM-EOP-009 HPI Cooling - Recovery From Solid Operations.
Analysis of Three Mile Island - Unit 2 Accident OP-TM-EOP-001, Reactor Trip, page 3, Rev. 3.
None.
IIl.c.07.04 4 New - TMIBank TMI Question ##
- Modified TMI Bank Parent Question ##
__ Memory or Fundamental Knowledge
~
3 Comprehension or Analysis I
C 55.41 4 55.43 .5 V 55.45 .I3 TMI SRO Exam - May 2003 Thursday, April 17,20W
Form ES-401-6 Q # 031 A Correct answer. Depressurization has resulted in formation of a steam bubble due to the hot metal temperature and absence of forced flow under then head. The water displacement is manifested in a Pressurizer insurge. Since the RCS is still subcooled and adequate heat transfer exists, no symptom based criteria exist to exit EOP-001. As a side note, EOP-001 follow-up actions will direct the operators to GO TO EOP-006, LOCA Cooldown.
B Incorrect answer. Right first part, wrong second part.(See A)
C Incorrect answer. Wrong first part, right second part. (See A)
D Incorrect answer. Both parts wrong.
None.
TMI SRO Exam - May 2003 Thursday, April 17, 2003
Page ## 4.1-9 Tier # 1 3.3 Group # 1 ROISRO Importance Rating c Ability to operate and/or monitor the following as they apply to Anticipated Transient Without Scram (ATWS): Charging pump suction valves from RWST operating switch.
~
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A Sequence of events:
- Reactor power was initially 1OO%, with ICs in full automatic
- No maintenance or surveillance tests in progress.
- Automatic reactor trip.
- CRD safety groups 1-4 fail to drop into the core.
Based on these conditions, select the ONE statement below that describes required response to this event A. Emergency borate from the BWST in acccordance with Rule 5, EB.
B. Manually insert CRD Groups 1-4 from the Diamond Control Panel.
C. Locally open the CRD DC Hold power supply breakers.
D. Trip all four RCPs.
OP-TM-EOP-001, Reactor Trip, Step 3.3, page 3, Rev. 3.
4 New - TMIBank TMI Question #
- Modified TMI Bank Parent Question #
C- Memory or Fundamental Knowledge hE Comprehension or Analysis
- Q- 55.45 .5/.6 55.43 -
A Correct answer. BWST is the preferred source of emergency boration.
B Incorrect answer. Rods are already de-energized by the reactor trip. This action would actully require trip reset (blocked since Groups I 4 are not at their inlimit.
C Incorrect answer. Although this would de-energize the rods, they are already de-energized by the upstream breakers.
D Incorrect answer. This is an action for loss of SCM, not stuck rods.
Replaced WA with similar KIA per NRC exam review. NOTE: should SCRUB all "BIT" referenced WAS from EPE 029 post exam.
TMI SRO Exam - May 2003 Thunriuy, April 17, 2003
Q # 014 Page # 4.2-11 Tier # 1 RO/SRO Importance Rating 3.7 -
3.7 Group # 1 Ability to determine and interpret the following as they apply to Reactor Coolant Pump (RCP)
Malfunctions (Loss of RC Flow): When to secure RCPs on loss of cooling or seal injection Plant conditions:
- Reactor power loo%, with ICs in full automatic.
- Intermediate Closed Cooling Pump is IC-P-1B 00s for motor replacement.
- Total RCP seal injection flow is 18 gpm, controlled locally in Makeup Valve Alley.
Event:
- IC-P-1A trips.
- Plant remains steady at 100% power.
Based on these conditions, identify the ONE statement below that identifies the applicable procedure and required action(s) to be implemented.
A. Initiate plant shutdown in accordance with OP 1102-4, Power Operations.
B. Increase RCP seal injection flow in accordance with OP 1104-2, Makeup and Purification System.
C. Attempt to restart IC-P-1A one time in accordance with OP-AA-103-103, Operation of Plant Equipment.
D. Trip the reactor AND then trip all 4 RCPs in accordance with OP-AA-101-111, Roles and Responsibilitiesof On-Shift Personnel.
OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel, Rev. 0, Section 4.6.4.7, Page 6.
2 New II] TMlBank TMI Question #
7 Modified TMI Bank Parent Question ##
71 Memory or Fundamental Knowledge 4 Comprehension or Analysis
-.j 55.41 3 55.43 .5 3 55.45 .I3 A Incorrect answer - Reactor/RCPs are required to be tripped due to failure of automatic RCP trip interlock.
B Incorrect answer - plausible, since seal injection flow is low, but reactor/RCPs are required to be tripped due to failure of the automatic RCP trip interlock.
C Incorrect answer - plausible, since this is historical guidance, but reactor/RCPs are required to be tripped due to failure of the automatic RCP trip interlock. OP-AA-103-103, Operation of Plant Equipment does not address this issue.
D Correct answer.
None.
TMI SRO Exam - May 2003 Thursday, April 17,2003
Page # 4.1-2 Tier # -1 ROlSRO Importance Rating 3.5 3.6
- Group # 2 Knowledge of the interrelations between Reactor Trip and the following: Reactor trip status v - -
AT ,
Initial conditions:
- Reactor is at 100%1 power
- All RPS cabinet lights are normal
- Electricians are doing breaker checks in plant Sequence of events:
- RPS cabinet B - Breaker Trip light goes BRIGHT
- RPS cabinet C - Breaker Trip light goes BRIGHT
- (ALL other RPS lights remain unchanged)
Assuming NO operator action, choose the plant status associated with the Control Rods that is a DIRECT result of these conditions.
A. Groups 1 - 4 are DROPPED into core, groups 5 - 7 remain OUT.
B. Groups 5 - 7 are DROPPED into core, groups 1 - 4 remain OUT.
C. Groups 1 - 7 are DROPPED into core.
D. Groups 1- 7 remain OUT.
\ <.&
OPM Section F-02, Reactor Protection System, pages 116 and 117, Rev. 8 (IO).,> &mfl OPM section F-01 Rev 7 Figurel, RPS lesson plan 11.2.01.466 slide 5 A /
None.
1V.E.14.10 4 New - TMIBank TMI Question #
- Modified TMI Bank Parent Question #
- Memory or Fundamental Knowledge fl Comprehension or Analysis
- 55.41 -155.43 4
- 55.45 .I3 A Correct answer B Incorrect - group 5 - 7 still have power from 10 (A) breaker C Incorrect - group 5 - 7 still have power from 10 (A) breaker D Incorrect - groups 1 - 4 are deenergized TMI SRO Exam - May 2003 Thursday, April 17,2003
SECTION F-02 REVISION 8 7.4 RPS Test Circuits and Trip Relay Operation (refer to Figures 61, 62)
Each RPS main trip relay controls a single contact in the supply and return power lines for the Control Rod Drive (CRD) power supply breakers undervoltage coils (Refer to CED section of training manual). The trip scheme for the CRD undervoltage coils is set up in such a way (refer to Figure 62) that when 2 out of 4 RPS master trip relays are de-energized the contacts associated with those RPS cabinets open in each CRD undervoltage coil supply and return power lines. The CRD undervoltage coils then deenergize tripping all CRD power supply breakers, the segment arms in the Control Rod Drives release, the roller nuts disengage from the lead screw and control rods fall into the core.
Test circuitry is set up in the RPS cabinets so a single CRD breaker or breaker set (in the case of the CRD DC hold breakers) can be test tripped from its associated RPS cabinet through the use of simulate trip toggle switches on the Rx trip module in each RPS cabinet (see Figure 62). The simulate trip toggles switches will only open contacts for the CRD breaker or breaker sets W coil associated with that RPS cabinet and will not cause a reactor trip. The manual trip scheme (use of 2 or more our of 4 simulate trip toggle switches) is set up so the CRD breakers can be tested via following scheme.
I RPSCABMET I BKR(s) Tested from that Cabinet A CB 10 B CB 11 C CB1, CB2 D CB3, CB4 When any of the test signals are applied, (i.e., a module taken to the test position) the master trip relay is de-energized and that cabinets contact opens in the power lines of the CRD undervoltage coils. The test lamp on the test module in the RPS cabinet becomes bright and the protective subsystem lights for that cabinet will also become bright.
This will also occur if any modules shown in Figure 62 are withdrawn from the cabinet. This will place the RPS in a 1 out of 3 trip scheme. That is to say if one more of the remaining three u n ~ p p e d channels were to trip, the contact scheme for de-energizing the CRD W coils would be satisfied and a reactor trip would be initiated.
116
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SECTION F-02 REVISION8 %
@s Each RPS cabinet has indicating lights on the outside top to tell the operator that status of each cabinet (see Figure 61). The turbine trip and FW trip bypass lights are bright when the bypass bistables are energized (trip function bypassed). Fan failure lights will be bright when there is a loss of power to the cabinet fans or low flow as sensed by a AF switch. The Protective Subsystem lights come on bright when a channel trips due to a setpoint being exceeded, a test trip signal is inserted or a module I
withdrawn or taken to test, or when a simulate test toggle switch in that RPS cabinet is taken to trip.
The Protective Subsystems lights are set up such that when a channel trips light # 1 signifies A channel tripped, light #2 signifies B channel tripped, light #3 signifies channel C and light #4 signifies channel D. When a test signal is inserted the respective light for that cabinet will become bright in the numbering sequence described above on both outside top of cabinet and inside on the test module.
When a channel is tripped or in test the respective Protective Subsystem light for that channel will be bright on all RPS cabinets (i.e., if Achannel trips then light # 1 on all RPS cabinets becomes bright).
However, when the simulate trip toggle switches are used during CRD breaker testing the only lights that become bright are the lights above the toggle switch and the lights on the outside top of that RPS cabinet. The Manual Bypass light becomes bright when the channel is placed in manual or shutdown bypass and when the channel is placed in manuai bypass a light on the test module (Figure 6 1) also becomes bright. When the shutdown bypass keyswitch is actuated not only does the outside top light become bright but a module in the right side of the RPS cabinet will have bright lights indicating the relay has picked up, the operators overhead alarm has actuated and the light above the cabinet has become bright. The breaker trip light will become bright when the CRD breaker associated with that cabinet is tripped.
When an RPS channel trips due to exceeding a setpoint or putting a module in test the channel is reset by first clearing the trip condition (or taking the test module selector switch to operate), then resetting the initiating module by using a reset spring return toggle switch on the relay status module shown in Figure 6 1. When the master trip relay is de-energized due to exceeding a limiting safep system setpoint the subsystem trip light on the relay module becomes bright.
The transfer circuitry for the non-nuclear instrument inputs t o the ICs has been removed from the A RPS cabinet. Transfer of neutron power, RCS flow, and RCS pressure from A to B W S cabinet input is accomplished by the SASS pushbuttons on console center. SASS operation is explained in section F-3 of the Operation Plant Manual.
117
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SECTION F!
REVISION 7 FIGURE 1 - POWER AND INSTRUMENTATION DIAGRAM PLANT CoYPuiEn 575 30'120 60 CONTACTOR A N 0 TRANSFER R Z L A Y S PROGRAMMER POWER SUPPLY OlSCONNEClS I R A N S F E R RELAVS (NOT SHOWN)
I I I I 10 A L L 6 9 CR D R I V E S
( P I . TRAVEL L I M I T AND I I C SIGNALS) TRANSFER -
IO R E L A T I V E P I ( T Y P I C A L OF 5 ( I V P I C A L OF -t ( T Y P I C A L Of 8
( T Y P I C A L OF ALL) FOR GROUP 4 B EACH GROUP) I E I C H CROUP) I v v v v T O GROUP 4 TO GROUP 5 I O GROUP 6 TO GROUP 7 TO GROUP 8
Tech Spec Reference Purge Requirements Nuclear Regulatory Commission SR021 Licensing Examination Three Mile Island Nuclear Station May 2003 DocumenUPage to be Removed Q#
Technical Specification 3.3 bases, page 3-23, Amendment 229. 009 Technical Specifications COLR Rev 0, abstract section. 088 (Graph of COLR rod insertion limits must be given as part of question)
Technical Specifications page 3-24, Amendment 227. 054 Thursday, April 17,2003 Page 1 of 1
L ES-401 PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - 1 Form ES-40 1-3 Exam Date: 05/12/2003 Exam Level: SRO K/A Category Points Tier Group Point Total 24
- 1. 16 Etnergencq
& 3 Abnormal Plant Evolutions 43 19 2.
17 Plant Systems 4
40 Cat 1 Cat 2 Cat 3 Cat 4
- 3. Generic Knowledge And Abilities 4 5 4 4 17 Note: I . Ensure that at least two topics from every K/A category are sampled within each teir (i.e., the Tier Totals in each K/A category shall not be less than two).
- 2. Actual point totals must match those specified in the table.
- 3. Select topics from many sqstems; avoid selecting more than o or three K/A topics from a given s) stem unless t h q relate to plant-specific priorities.
- 4. Systemdevolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the categoryhier.
- 6. The generic 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.
7 . On the following pages, enter the KIA numbers. a brief description of each topic, the topics importance ratings for the RO license level, and the point totals for each system and category. K/As beloit 2.5 should be justified on the basisof plant-specific priorites. Enter the tier totals for each category in the table aboLe.
1
PWR SRO Examination Outline Printed: 04117/2003 Facility: Three Mile Island - 1 SS - 401 Emerge1 y and A inormal Plant Evolutions - Tier 1 I Group 1 Form ES-401-.
YAPE # EIAPE Name I Safety Function KA KA Topic Comment 00 1 Continuous Rod Withdrawal I 1 AA2.04 Reactor power and its trend 1 003 Dropped Control Rod I I AK2.05 Control rod drive power supplies and logic circuits 2 003 Dropped Control Rod / 1 AA 1.06 RCS pressure and temperature 005 Inoperable/Stuck Control Rod / 1 Axial power imbalance 01 1 Large Break LOCA 1 3 2.1.33 Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.
01 1 Large Break LOCA / 3 EKI.01 Natural circulation and cooling, including reflux 12 boiling 015 Reactor Coolant Pump (RCP) Malfunctions / 4 2.4.6 Knowledge symptom based EOP mitigation 13 SRO strategies.
017 Reactor Coolant Pump (RCP) Malfunctions (Loss AA2.10 When to secure RCPs on loss of cooling or seal 14 SRO of RC Flow) / 4 injection 026 Loss of Component Cooling Water (CCW) I 8 AK3.03 Guidance actions contained in EOP for Loss of 16 ccw 026 Loss of Component Cooling Water (CCW) / 8 AA 1.07 Flow rates to the components and systems that are serviced by the CCWS; interactions among the components 029 Anticipated Transient Without Scram (ATWS) / I Breakers, relays, and disconnects 19 1
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - I 3s 401 -
Emerge] y and P normal Plant Evolutions Tier 1 /Group 1 Form ES-40 1-.
VAPE # E/APE Name / Safety Function KA KA Topic Comment 029 Anticipated Transient Without Scram (ATWS) / 1 EA1.02 Charging pump suction valves from RWST 20 operating switch 05 I Loss of Condenser Vacuum 1 4 AK3.O 1 Loss of steam dump capability upon loss of 24 condenser vacuum 055 Loss of Offsite and Onsite Power (Station Blackout) m Length of time for which battery capacity is 25 16 designed 067 Plant Fire on Site I 9 2.1.32 Ability to explain and apply all system limits and 28 precautions.
069 Loss of Containment Integrity I 5 AA2.01 Loss of containment integrity 29 SRO 069 Loss of Containment Integrity I 5 AKI.OI Effect of pressure on leak rate 30 074 Inadequate Core Cooling I 4 EA2.06 Changes in PZR level due to PZR steam bubble 31 SRO transfer to the RCS during inadequate core cooling A03 Loss of "I-Y I7 AK2.2 Facility's heat removal systems, including primary 33 coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility A03 LOSSof NNI-Y I 7 AAI.1 Components, and functions of control and safety 34 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features A06 Shutdown Outside Control Room I 8 AKI .3 Annunciators and conditions indicating signals, 35 and remedial actions associated with the (Shutdown Ouside Control Room) 2
PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - 1 ES - 401 Emerge] Form ES-401-3
$/APE ## EIAPE Name / Safety Function KA KATopic Comment E05 Excessive Heat Transfer / 4 EK2.1 Components, and functions of control and safety 37 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E09 Natural Circulation Operations / 4 2.4.30 Knowledge of which events related to system 40 SRO operations/status should be reported to outside agencies.
E09 Natural Circulation Operations / 4 EK3.2 Normal, abnormal and emergency operating 41 procedures associated with (Natural Circulation 3
PWR SRO Examination Outline Printed: 0411 712003 Facility: Three Mile Island - 1 ES - 401 Emerge1 y and P lnormal Plant Evolutions - Tier 1 / Group 2 Form ES-401-2 EIAPE ## EIAPE Name I Safety Function KA KA Topic Comment 007 Reactor Trip / 1 EK2.03 Reactor trip status panel 5 007 Reactor Trip I 1 EA 1.03 RCS pressure and temperature 6 008 Pressurizer (PZR) Vapor Space Accident (Relief AK2.01 Valves 7 Valve Stuck Open) / 3 008 Pressurizer (PZR) Vapor Space Accident (Relief m Actions contained in EOP for PZR vapor space 98 Valve Stuck Open) / 3 accident1LOCA 009 Small Break LOCA 13 2.2.25 Knowledge of bases in technical specifications for 9 limiting conditions for operations and safety limits.
009 Small Break LOCA 1 3 EK1.O1 Natural circulation and cooling, including reflux IO boiling 022 Loss of Reactor Coolant Makeup 1 2 AK 1.02 Relationship of charging flow to pressure 15 differential between charging and RCS 027 Pressurizer Pressure Control (PZR PCS) AK2.03 Controllers and positioners 18 Malfunction / 3 033 Loss of Intermediate Range Nuclear Instrumentation AK3.01 Termination of startup following loss of 21 I7 intermediate-range instrumentation 03 8 Steam Generator Tube Rupture (SGTR) / 3 EA 1 .OS Core cooling monitor 22 03 8 Steam Generator Tube Rupture (SGTR) / 3 EA2.09 Existence of natural circulation, using plant 23 parameters I
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 normal Plant Evolutions - Tier 1 / Group 2 Form ES-401-:
EIAPE # EIAPE Name I Safety Function KA KA Topic Comment 06 1 Area Radiation Monitoring (ARM) System Alarms 1 AA2.01 ARM panel displays 26 SRO 7
I 06 1 Area Radiation Monitoring (ARM) System Alarms I 12.1.32 Ability to explain and apply all system limits and 27 7 precautions.
~
A0 1 Plant Runback I 1 AAI.1 Components, and hnctions of control and safety 32 systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EO8 Adherence to appropriate procedures and 38 SRO operation within the limitations in the facility's license and amendments E08 LOCA Cooldown 1 4 EKl.3 Annunciators and conditions indicating signals, 39 SRO and remedial actions associated with the (LOCA Cooldown) 2
PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - I E/APE # E/APE Name / Safety Function KA KATopic Comment A08 Refuel Canal Level Decrease / 8 AA2.1 Facility conditions and selection of appropriate 36 SRO procedures during abnormal and emergency operations E13 EOP Rules 2.2.22 Knowledge of limiting conditions for operations 42 and safety limits.
E13 EOP Rules EKI .2 Normal, abnormal and emergency operating 43 procedures associated with (EOP Rules)
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Plant Systems - Tier 2 / Group 1 Form ES-401-;
Sys/Ev I# System / Evolution Name KA KA Topic Comment 00 1 Control Rod Drive System / 1 K6.03 Reactor trip breakers, including controls 44 003 Reactor Coolant Pump System K3.03 Feedwater and emergency feedwater 47 (RCPS) I 4 003 Reactor Coolant Pump System A2.02 Conditions which exist for an abnormal shutdown 48 SRO (RCPS) 1 4 of an RCP in comparison to a normal shutdown of an RCP 004 Chemical and Volume Control 2.4.4 Ability to recognize abnormal indications for 49 SRO System (CVCS) I 1 system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
004 Chemical and Volume Control A4.18 Emergency borate valve 50 System (CVCS) I 1 015 Nuclear Instrumentation System / 7 K4.04 Slow response time of SPNDs 58 015 Nuclear Instrumentation System 1 7 K6.04 Bistables and logic circuits 59 022 Containment Cooling System K2.0 1 Containment cooling fans 60 (CCS) / 5 026 Containment Spray System (CSS) I K1.O1 ECCS 61 5
026 Containment Spray System (CSS) I A3.01 Pump starts and correct MOV positioning 62 5
06 1 Auxiliary I Emergency Feedwater K5.02 Decay heat sources and magnitude 69 (AFW) System / 4 06 I Auxiliary / Emergency Feedwater K6.0 1 Controllers and positioners 70 JAFW) Svstem 1 4 1
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 Sys/Ev # I System / Evolution Name I K A I KA Topic I Comment 063 I D.C. Electrical Distribution I K3.01 I ED/G I 7 1 SRO System I 6 063 D.C. Electrical Distribution K4.04 Trips 72 System I 6 068 Liquid Radwaste System (LRS) I 9 K1.07 Sources of liquid wastes for LRS 73 068 Liquid Radwaste System (LRS) 19 A3.02 Automatic isolation 75 07 1 Waste Gas Disposal System A2.02 Use of waste gas release monitors, radiation, gas 76 SRO (WGDS) I 9 flow rate, and totalizer 072 Area Radiation Monitoring (ARM) K3.02 Fuel handling operations 77 SRO ISystem I 7 I
I I
I I I 072 IArea Radiation Monitoring (ARM) I Al.01 I Radiation levels 1 78SRO 2
PWR SRO Examination Outline Printed: 0411 712003 Facility: Three Mile Island - 1 ES - 401 Plant Systems - Tier 2 /Group 2 Form ES-401-I I I I
-SysIEv # System I Evolution Name KA KA Topic Comment 002 Reactor Coolant System (RCS) 1 2 K3.02 Fuel 45 002 Reactor Coolant System (RCS) / 2 46 SRO 01 1 Pressurizer Level Control System 55 (PZR LCS) / 2
~
012 Reactor Protection System 1 7 K4.05 Spurious trip protection 56 012 Reactor Protection System 1 7 K6.11 Trip setpoint calculators I
IContainment Purge System (CPS) /
~~
029 K1.O1 Gaseous radiation release monitors 63 033 Spent Fuel Pool Cooling System A3.01 Temperature control valves 64 (SFPCS) 1 8 034 Fuel Handling Equipment System A2.02 Dropped cask 65 SRO (FHES) / 8 035 Steam Generator System (SICS) I 2.4.49 Ability to perform without reference to procedures 81 SRO those actions that require immediate operation of system components and controls.
14 Main and Reheat Steam System 2.4.6 Knowledge symptom based EOP mitigation 66 SRO (MRSS) 1 4 strategies.
Main and Reheat Steam System A4.07 Steam dump valves 67 (MRSS) / 4 055 Condenser Air Removal System A3.03 Automatic diversion of CARS exhaust 68 (CARS) 1 4 1
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Sys/Ev # System / Evolution Name KA KA Topic Comment 364 Emergency Diesel Generator K2.03 Control power 73 (EDIG) System / 6 373 Process Radiation Monitoring K4.01 Release termination when radiation exceeds 79 (PRM) System / 7 setpoint 373 Process Radiation Monitoring A1.O1 Radiation levels 80 SRO (PRM) System 1 7 375 Circulating Water System I 8 K3.07 ESFAS 82 I03 Containment System I 5 2.4.30 Knowledge of which events related to system 83 SRO oaerationslstatusshould be reported to outside 2
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Sys/J%v# System / Evolution Name KA Topic Comment 005 Residual Heat Removal System RHR heat exchanger 51 (RHRS) 1 4 Residual Heat Removal System Heatuplcooldown rates 52 (RHRS) I 4 Pressurizer Relief TankIQuench K4.0 1 Quench tank cooling 53 Tank System (PRTS) / 5 Knowledge of limiting conditions for operations 54 SRO and safety limits.
Generic Knowledge and Abilities Outline (Tier 3) Printed: 0411 712003 PWR SRO Examination Outline Form ES-401-5 Facility: Three Mile Island - I Generic Category KA KA Topic Comment Conduct of Operations 2.1.5 Ability to locate and use procedures and directives related to shift 85 staffing and activities.
2.1.7 Ability to evaluate plant performance and make operational 86 SRO judgments based on operating characteristics,reactor behavior, and instrument interpretation.
2.1.10 Knowledge of conditions and limitations in the facility license. 87 SRO 2.1.34 Ability to maintain primary and secondary plant chemistry within 84 allowable limits.
Category Total: 4
~~~
Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including 92 operating those controls associated with plant equipment that could affect reactivity.
2.2.1 1 Knowledge of the process for controlling temporary changes. 90 SRO 2.2.19 Knowledge of maintenance work order requirements. 89 SRO 2.2.26 Knowledge of refueling administrative requirements. 88 SRO 2.2.27 Knowledge of the reheling process. 91 SRO Category Total: 5 Radiation Control 2.3.1 Knowledge of I O CFR: 20 and related facility radiation control 95 requirements.
2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are 94 SRO outside the control room (e.g., waste disposal and handling systems).
2.3.8 Knowledge of the process for performing a planned gaseous 93 radioactive release.
2.3.10 Ability to perform procedures to reduce excessive levels of radiation 96 SRO and guard against personnel exposure.
Category Total: 4 1
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ES-401 PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 Form ES-40 1-3 Exam Date: 05/12/2003 Exam Level: SRO WA Category Points Tier Group Point Total 24 16 3
43 19 17 4
40 17 Note: 1. Ensure that at least two topics from every WA category are sampled within each teir (i.e., the "Tier Totals" in each KIA category shall not be less than two).
- 2. Actual point totals must match those specified in the table.
- 3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4. Systems/evolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the categoryhier.
- 6. The generic K/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.
- 7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basis of plant-specific priorites. Enter the tier totals for each category in the table above.
I 1
ES-401 PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 Form ES-40 1-3 2
ES-401 PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - 1 Form ES-40 1-3 Exam Date: 05/12/2003 Exam Level: SRO IUA Category Points Tier Group Point
- 3. Generic Knowledge And Abilities I cat 1 1 cat2 I cat3 I cat4 I Note: 1. Ensure that at least two topics from every K/A category are sampled within each teir (i.e., the "Tier Totals" in each K/A category shall not be less than two).
- 2. Actual point totals must match those specified in the table.
- 3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.
- 4. Systems/evolutions within each group are identified on the associated outline.
- 5. The shaded areas are not applicable to the categorykier.
- 6. The generic 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.
- 7. On the following pages, enter the K/A numbers, a brief description of each topic, the topics' importance ratings for the RO license level, and the point totals for each system and category. K/As below 2.5 should be justified on the basisof plant-specific priorites. Enter the tier totals for each category in the table above.
1
PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - I ES - 401 Emei ency and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-3
</APE# E/APE Name / Safety Function KA Topic Imp. Points 00 1 Continuous Rod Withdrawal / 1 AA2.04 - Reactor power and its trend 4.3 1 003 Dropped Control Rod / 1 AK2.05 - Control rod drive power supplies and logic
- 2.8 1 circuits 003 Dropped Control Rod / 1 AA 1.06 - RCS pressure and temperature 4.1 1 005 Inoperable/Stuck Control Rod / 1 AK 1 .O 1 - Axial power imbalance 3.8 1 01 1 Large Break LOCA / 3 2.1.33 - Ability to recognize indications for system 4.0 1 operating parameters which are entry-level conditions for technical specifications.
01 1 Large Break LOCA / 3 EK1.01 - Natural circulation and cooling, including 4.4 I reflux boiling 01 5 Reactor Coolant Pump (RCP) Malfunctions / 4 2.4.6 - Knowledge symptom based EOP mitigation
- 4.0 I strategies.
017 Reactor Coolant Pump (RCP) Malfunctions (Loss of AA2.10 - When to secure RCPs on loss of cooling or 3.7 1 RC Flow) / 4 seal injection 026 Loss of Component Cooling Water (CCW) / 8 AK3.03 - Guidance actions contained in EOP for Loss 4.2 1 of ccw 026 Loss of Component Cooling Water (CCW) / 8 AAI .07 - Flow rates to the components and systems that 3.O 1 are serviced by the CCWS; interactions among the components 1
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Eme rnd Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-3 E/APE # E/APE Name I Safety Function KA Topic mp. Points 029 Anticipated Transient Without Scram (ATWS) / 1 EK2.06 - Breakers, relays, and disconnects 3.1* 1 029 Anticipated Transient Without Scram (ATWS) I 1 EAl.02 - Charging pump suction valves from RWST 3.3 1 operating switch 05 1 Loss of Condenser Vacuum 1 4 AK3.01 - Loss of steam dump capability upon loss of 3.1* 1 condenser vacuum 055 Loss of Offsite and Onsite Power (Station Blackout) / EK3 .O1 - Length of time for which battery capacity is 3.4 1 6 designed 067 Plant Fire on Site / 9 2.1.32 - Ability to explain and apply all system limits 3.8 1 and precautions.
069 Loss of Containment Integrity I 5 AA2.0 1 - Loss of containment integrity 4.3 069 Loss of Containment Integrity I 5 AK I .O 1 - Effect of pressure on leak rate 3.1 074 Inadequate Core Cooling 1 4 EA2.06 - Changes in PZR level due to PZR steam 4.6 bubble transfer to the RCS during inadequate core cooling 2
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-3 4 2 G KA Topic Imp. Points AK2.2 - Facility's heat removal systems, including 3.3 1 primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility AA 1.1 - Components, and functions of control and 4.0 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features AKI .3 - Annunciators and conditions indicating signals, 3.4 1 and remedial actions associated with the (Shutdown Ouside Control Room)
E05 Excessive Heat Transfer 1 4 X EK2.1 - Components, and hnctions of control and 4.0 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E09 Natural Circulation Operations 1 4 x 2.4.30 - Knowledge of which events related to system 3.6 1 operationslstatus should be reported to outside agencies.
E09 Natural Circulation Operations 1 4 X EK3.2 - Normal, abnormal and emergency operating 3.8 I procedures associated with (Natural Circulation Operations)
KIA Category Totals: 4 4 4 4 4 4 Group Point Total: 24 3
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Emer -
and Abnormal Plant Evolutions Tier 1 / Group 2 Form ES-401-3 E/APE #
007 E/APE Name / Safety Function Reactor Trip I 1 KA Topic EK2.03 - Reactor trip status panel Imp.
3.6 Points 1
007 Reactor Trip I 1 EAI.03 - RCS pressure and temperature 4.1 1 008 Pressurizer (PZR) Vapor Space Accident (Relief AK2.01 - Valves 2.7 1 Valve Stuck Open) 1 3 008 Pressurizer (PZR) Vapor Space Accident (Relief AK3.03 - Actions contained in EOP for PZR vapor 4.6 1 Valve Stuck Open) I 3 space accidentILOCA 009 Small Break LOCA 1 3 2.2.25 - Knowledge of bases in technical specifications 3.7 1 for limiting conditions for operations and safety limits.
009 Small Break LOCA 1 3 -
EKI.01 Natural circulation and cooling, including 4.7 1 reflux boiling 022 Loss of Reactor Coolant Makeup 12 AKI .02 - Relationship of charging flow to pressure 3.1 1 differential between charging and RCS 027 Pressurizer Pressure Control (PZR PCS) Malfunction AK2.03 - Controllers and positioners 2.8 1 13 033 Loss of Intermediate Range Nuclear Instrumentation I AK3.01 - Termination of startup following loss of 3.6 1 7 intermediate-range instrumentation 1
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES 401 -
Emergency ind Abnormal Plant Evolutions Tier 1 / Group 2 Form ES-401-3 E/APE # KA Topic Imp. Points 03 8 EA1.OS - Core cooling monitor 3.8* 1 038 Steam Generator Tube Rupture (SGTR) / 3 EA2.09 - Existence of natural circulation, using plant 4.2 1 parameters 06 1 Area Radiation Monitoring (ARM) System Alarms / 7 AA2.01 - ARM panel displays 3.7 1 06 1 Area Radiation Monitoring (ARM) System Alarms / 7 2.1.32 - Ability to explain and apply all system limits 3.8 1 and precautions.
A0 1 Plant Runback / 1 AAI .1 - Components, and functions of control and 3.7 1 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features E08 EA2.2 - Adherence to appropriate procedures and 4.0 1 operation within the limitations in the facility's license and amendments I I
EO8 EKI .3 Annunciators and conditions indicating signals, I 3.5 I 1 and remedial actions associated with the (LOCA - I I WA Category Totals: 3 3 2 3 3 2 Group Point Total: 16 2
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 Emergency
- and Abnormal Plant Evolutions Tier 1 / Group 3 Form ES-401-3 E/APE # EIAPE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp. Points A08 Refbel Canal Level Decrease / 8 X AA2.1 - Facility conditions and selection of appropriate 4.0 1 procedures during abnormal and emergency operations E13 EOP Rules -
X 2.2.22 Knowledge of limiting conditions for operations 4.1 1 and safety limits.
E13 EOP Rules X EK 1.2 - Normal, abnormal and emergency operating 3.6 1 procedures associated with (EOP Rules)
WA Category Totals: 1 0 0 0 1 1 Group Point Total: 3 1
PWR SRO Examination Outline Printed: 04/17/2003 Facility: Three Mile Island - 1 ES - 401 -
Form S-401-2 iys/Ev #
00 1 2 KATopic K6.03 Reactor trip breakers, including Imp.
4.2 Points 1
controls 003 Reactor Coolant Pump System K3.03 - Feedwater and emergency feedwater
- 3.1 1 (RCPS) / 4 003 -
A2.02 Conditions which exist for an abnormal
- 3.9 1 (RCPS) / 4 shutdown of an RCP in comparison to a normal shutdown of an RCP 004 Chemical and Volume Control System I
I Y 2.4.4 - Ability to recognize abnormal
-4.3 1 (CVCS) / 1 Iindications for system operating parameters which are entry-level conditions for emergency 004 015 Chemical and Volume Control System (CVCS) / 1 Nuclear Instrumentation System / 7 t and abnormal operating procedures.
A4.18 Emergency borate valve K4.04 - Slow response time of SPNDs 4.1 3.6?
1 1
015 Nuclear Instrumentation System / 7 K6.04 - Bistables and logic circuits 3.2 1 022 Containment Cooling System (CCS) / -
K2.01 Containment cooling fans 3.1 5
026 Containment Spray System (CSS) / 5 X K1.O1 - ECCS 4.2 026 Containment Spray System (CSS) I 5 A3.01 - Pump starts and correct MOV 4.5 positioning 06 1 Auxiliary / Emergency Feedwater -
K5.02 Decay heat sources and magnitude 3.6 (AFW) System 1 4 -
1
PWR SRO Examination Outline Printed 0411712003 Facility: Three Mile Island - 1 ES 401 :roup 1 Form S-4011 SysIEv # System I Evolution Name K1 K2 K4 K5 KA Topic [mp. Points 06 1 Auxiliary I Emergency Feedwater K6.0 1 - Controllers and positioners 2.8* 1 (AFW) System I 4 063 D.C. Electrical Distribution System I 6 K3.01 - ED/G
-4.1 1 063 D.C. Electrical Distribution System 1 6 X K4.04 - Trips 2.9?
068 Liquid Radwaste System (LRS) 19 X -
K1.07 Sources of liquid wastes for LRS 2.9 068 Liquid Radwaste System (LRS) 19 A3.02 - Automatic isolation
-3.6 07 1 Waste Gas Disposal System (WGDS) X A2.02 - Use of waste gas release monitors, 3.6 19 radiation, gas flow rate, and totalizer 072 Area Radiation Monitoring (ARM) -
K3.02 Fuel handling operations
-3.5 System I 7 072 Area Radiation Monitoring (ARM) X A1.O1 - Radiation levels 3.6 Svstem I 7 KIA Category Totals: 2 1 2 1 3 1 2 2 1 1 Group Point Total: 19 2
PWR SRO Examination Outline Printed: 0411712003 Facility: Three Mile Island - 1 ES - 401 ystemA Form :S-401-:
I s y s m ## G KATopic K3.02 Fuel Imp.
4.5 Points 1
002 I ?
002 Reactor Coolant System (RCS) / 2 K5.18 - Brittle fracture 3.6 1 I I -
01 1 Pressurizer Level Control System I A2.08 - Loss of level compensation 2.8 (PZR LCS) / 2 012 Reactor Protection System / 7 K4.05 - Spurious trip protection
- 2.9 K6.11 - Trip setpoint calculators
- 2.9 012 Reactor Protection System / 7 029 Containment Purge System (CPS) / 8 X -
K1.01 Gaseous radiation release monitors 3.7 033 Spent Fuel Pool Cooling System A3.01 - Temperature control valves
-2.7*
(SFPCS) / 8 034 Fuel Handling Equipment System I
I -
A2.02 Dropped cask 3.9 (FHES) I 8 035 Steam Generator System (S/GS)/ 4 X 2.4.49 - Ability to perform without reference to 4.0 procedures those actions that require immediate operation of system components and controls.
039 Main and Reheat Steam System X 2.4.6 - Knowledge symptom based EOP
-4.0 1 (MRSS) / 4 mitigation strategies.
039 Main and Reheat Steam System A4.07 - Steam dump valves 2.9 (MRSS) / 4 1
PWR SRO Examination Outline Printed: 04/17/2003 Facility: -
Three Mile Island 1 CS - 401 ! ant Sysl ns - Tier 2 / ;roup 2 Form ES-401-2
$ys/Ev# System / Evolution Name K2 K3 K4 K5 K6 11 A2 44 G KA Topic 055 Condenser Air Removal System A3.03 - Automatic diversion of CARS exhaust (CARS) / 4 064 Emergency Diesel Generator (ED/G) X -
K2.03 Control power 3.6 1 System / 6 073 Process Radiation Monitoring (PRM) X K4.0 1 - Release termination when radiation 4.3 1 System I 7 exceeds setpoint 073 Process Radiation Monitoring (PRM) X A 1.O1 - Radiation levels 3.5 1 System 1 7 075 Circulating Water System / 8 X K3.07 - ESFAS 3.5* 1 103 Containment System / 5 X 2.4.30 - Knowledge of which events related to 3.6 1 system operations/status should be reported to outside agencies.
K/A Category Totals: 1 2 2 1 1 1 2 1 3 Group Point Total: 17 2
PWR SRO Examination Outline Printed: 0411 712003 Facility: Three Mile Island - 1 ES - 401 :S-401-2 I I I System / Evolution Name Residual Heat Removal System K1 K2 K3 44 G KATopic K6.03 RHR heat exchanger Imp. Points 2.6 1 1
(RHRS) / 4 005 Residual Heat Removal System -
Al.01 Heatup/cooldown rates
-3.6 007 (RHRS) / 4 Pressurizer Relief TarWQuench Tank I l l K4.01 - Quench tank cooling 2.9 System (PRTS) / 5 008 Component Cooling Water System X 2.2.22 - Knowledge of limiting conditions for
-4.1 1 (CCWS) / 8 operations and safety limits. -
- A Category Totals: 0 0 0 1 0 1 Group Point Total: 4 1
Generic Knowledge and Abilities Outline (Tier 3)
Printed: 0411712003 PWR SRO Examination Outline Form ES-401-5 Facilitv: -
Three Mile Island 1 Generic Category KA KATopic Imp. Points Conduct of Operations 2.1.5 Ability to locate and use procedures and directives related to shift staffing and activities. 3.4 1 2.1.7 Ability to evaluate plant performance and make operationaljudgments based on operating 4.4 characteristics, reactor behavior, and instrument interpretation.
2.1.10 Knowledge of conditions and limitations in the facility license. 3.9 I I I Category Total: 4 Equipment Control 2.2.1 Ability to perform pre-startup procedures for the facility, including operating those
- 3.6
- 1 controls associated with plant equipment that could affect reactivity.
2.2.1 1 Knowledge of the process for controlling temporary changes. 3.4* 1 2.2.19 Knowledge of maintenance work order requirements. 3.1 1 2.2.26 Knowledge of refueling administrative requirements. 3.7 1 2.2.27 Knowledge of the refueling process. 3.5 1 Category Total: 5 Radiation Control I 2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements. 3.O 1 2.3.3 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room 2.9 (e.g., waste disposal and handling systems).
2.3.8 Knowledge of the process for performing a planned gaseous radioactive release. 3.2 I 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
3.3 1 1
Generic Knowledge and Abilities Outline (Tier 3)
Printed: 0411712003 PWR SRO Examination Outline Form ES-401-5 Facilitv: Three Mile Island - 1 Generic Category KA KA Topic Imp. Points Emergency ProceduredPlan 2.4.10 Knowledge of annunciator response procedures. 3.1 1 r
2.4. I I Knowledge of abnormal condition procedures. 3.6 1 2.4.33 Knowledge of the process used track inoperable alarms. 2.8 1 2.4.44 Knowledge of emergency plan protective action recommendations. 4.0 1 Category Total: 4 Generic Total: 17 2
TMI SRO License Exam OW1 2/03 TMI-I OPERATOR TRAINING JOB PERFORMANCE MEASURE B.1.e (new)
Perform Turbine Valve Testing on a CIV (Combined Intermediate Valve)
Page 1 of 7
B.1.e 11.2.05.NEW Revision 0 05/12/2003 TASK TITLE: Perform Turbine Valve Testing on a CIV (Combined Intermediate Valve).
TASK NUMBER: 0450040201 During power ops, perform main turbine valve testing.
TIF: 2.90 KIA
REFERENCE:
System: Steam Generator System (035)
WA: 2. I.23 Ability to perform specific system and integrated plant procedures during all modes of plant operation.
Rating(ROISRO): 3.914.0 POSITION:
EVALUATION METHOD: PERFORM SIMULATE EVALUATION LOCATION: SIMULATOR IN-PLANT 0 CONTROL ROOM 0 OTHER 0 TASK STANDARDS: TG-CIV-1 tested satisfactorily IAW procedure.
APPROXIMATE COMPLETION TIME: 15 minutes.
TIME-CRITICAL TASK COMPLETION TIME: NA minutes REQUIRED TOOLS OR MATERIALS: 1106-1 Rev 109 Appendix C section 2.0 with steps 2.1.1, 2.1.2,2.1.5, 2.1.7, 2.1.8, 2.2.1.1, 2.2.1.2, and 2.2.1.5 NIAd; steps 2.2.1.3 and 2.2.1.4 signed off.
REFERENCES:
1106-1 Rev 109, Appendix C, section 2.1 and 2.2.
ALTERNATE PATH JPM? NO SIMULATOR SETUP:
INITIALIZATION:
Initialize the Trainer to IC16 100% power, ICs in automatic, Xenon equilibrium, BOC.
Reduce power to 90%
Start Second EHC pump (both running)
When stable, place ICs SGlRx Demand to HAND.
Make Snapshot after plant stabilizes.
EVENT TRIGGERS: N/A MALFUNCTIONS: None REMOTE FUNCTIONS: N/A OVERRIDES: N/A MONITOR: N/A Page 2 of 7
B.1.e 11.2.05.NEW Revision 0 05/12/2003 READ TO STUDENT When I tell you to begin, you are to PERFORM Turbine Valve Testing on Combined Intermediate Valve TG-CIV-1 ONLY IAW 1106-1 Appendix C. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.
INITIAL CONDITIONS:
Reactor power is 90%, with ICs in Manual on SG/Rx Demand only.
There are NO maintenance activities in progress.
Turbine valve testing preparations have been made IAW 1106-1Appendix C.
INITIATING CUE:
The Unit Supervisor directs you to PERFORM Turbine Valve Testing on Combined Intermediate Valve TG-CIV-1 ONLY IAW 1106-1 Appendix C, section 2.2. Do NOT test any other valves.
(Hand examinee copy of 1106-1 Rev 109 AppendixC section 2.0 with steps 2.1.1, 2.1.2, 2.1.5, 2.1.7, 2.1.8, 2.2.1.I, 2.2.1.2, and 2.2.1.5 N/Ad; steps 2.2.1.3 and 2.2.1.4 signed off.)
ARE THERE ANY QUESTIONS?
TIME CRITICAL: NO Page 3 of 7
JPM INSTRUCTION SHEET DIRECTIONS TO STUDENT:
When I tell you to begin, you are to PERFORM Turbine Valve Testing on Combined Intermediate Valve TG-CIV-I ONLY IAW 1106-1 Appendix C. Before you start, I will describe the general plant conditions, state the initiating cues, and answer any questions. Perform procedure steps as if you were actually performing the task.
INITIAL CONDITIONS:
Reactor power is 90%, with ICs in Manual on SG/Rx Demand only.
There are NO maintenance activities in progress.
Turbine valve testing preparations have been made IAW 1106-1 Appendix C.
INITIATING CUE:
The Unit Supervisor directs you to PERFORM Turbine Valve Testing on Combined Intermediate Valve TG-CIV-1 ONLY IAW 1106-1 Appendix C, section 2.2. Do NOT test any other valves.
(Obtain copy of 1106-1 Rev 109 Appendix C section 2.0 with steps 2. I. 1, 2.1.2, 2.1 5,2. I.7, 2. I.8, 2.2.1.1, 2.2.1.2, and2.2.1.5 N/Ad; steps2.2.1.3and2.2.1.4signedoff.)
ARE THERE ANY QUESTIONS?
TIME CRITICAL: NO Page 4 of 7
B.1.e 11.2.05.NEW Revision 0 05/12/2003
- STEP STANDARD SIU 1 Examinee reviews procedure precautions Procedure reviewed by examinee and steps.
CUE: As examinee reviews procedure, respond as needed for clarification, i.e. ENDPOINT on testing is step 2.2.2.1 2.
NOTE: The following steps are performed from the Digital Turbine Control Station DTCS on console left.
Manipulations may be done via touch screen, rollerlmouse pad, or keyboard, OR ANY combination of same. Delays between selection and execution of commands may cause no action, but is recoverable via re-selection and timely execution.
From Main Display screen, EXIT to Main Examinee gets Main Menu Page #I on the screen.
Menu Page #I.
From Main Menu page # I , select screen Examinee gets #23 Valve Stroke Test Prereqs on
- 23 Valve Stroke Test Prereqs. Main Menu Page #I.
Select FULL STROKE test and execute Examinee executes FULL STROKE test.
command on screen #23.
Exit to Main Menu Page #and I select Examinee selects VALVE STROKE TESTING on VALVE STROKE TESTING under #23. Main Menu Page #I.
~~
Select the desired CIV (Full Stroke) to be Examinee selects CIV-I on the Valve Stroke tested on the VALVE STROKE TESTING Testing screen.
screen. (A plotigraph appears)
On the TRIGGERED PLOT display, verify Examinee verifies Full Stroke logic ( I ) on the FULL-STROKE is a logic (1). TRIGGERED PLOT display.
~
Review steps 7 - 11 prior to Examinee reviews procedure steps 7 - 11.
performance.(PER PROCEDURE NOTE)
Initiate test by selecting START and Examinee executes START command.
execute command.
Observe plot for smooth stem stroking and Examinee observes closing and FAST CLOSURE FAST CLOSING of both valves (CIVis of CIV on plot.
made up of lV intercept Valve and ISV intercept Stop Valve) NOTE: Examinee MUST wait until both valves (IV and ISV) stroke before performing next 0 note sharp drop in position indication step.
in last 10% of travel.
NOTE: Examinee may request local observer report on CIV positionlfast closure. IC0 roleplay as required.
As soon as both valves are closed, select Examinee selects STOP (after IV and ISV close).
STOP to terminate test.
Page 5 of 7
- STEP STANDARD SIU NOTE: Examinee may request local observer report on CIV open position. IC0 roleplay as required.
~ ~~~~ ~~
13 Once valves reopen, select MORE Examinee selects MORE OPTIONS, then SAVE OPTIONS IMAGE.
0 THEN select SAVE IMAGE I NOTE: Examine should terminate this JPM at this time.
END TASK Page 6 of 7
B.1.e 11.2.05.NEW Revision 0 O W 1 2/2003 JPM CHANGE HISTORY PAGE REVISION DATE REFERENCE DESCRIPTION Include AI # if A r
0 05112/03 Section 2.0 Page 7 of 7
TMI Unit 1 Page 6 of 23 2.0 This test verifies free valve stem nt, fast closure feature of disc dump valves and records stroke es. Because steam flow transients are induced in the OT this test, its performance at I high power levels is by authorization of Plant Operations Director only. I 2.1 Precautions r to 5 90% prior to and during testing of control valves main stop valves (SV-1, 2, 3, 4).
K is automatically placed in operation 2, 3 and 4). Turbine control will be 2.1.3 o 95 Percent or less prior to and during testing of 2.1.4 CROs should be prepared to start MO-P-1s if needed to control level d#f 2.1.5 ift turbine header pressure to opposite side. SV-1 and 2 2.1.6 ing turbine valve testing. (Computer point TAOM) be running during valve testing. Terminate test f l / ,2.1.7 logic prevents CV testing when valve position limit ssure limit MSPL are in effect.
? 44 2.1.8 Hand/Auto Control Stations in HAND for full stroke testing of valves when 86
Page 7 of 23 2.2 Testing of Combined Int Valves and Turbine Stop Valves (TG-CIV-1,2,3,4,5,6 and TG-SV-1,2,3,4)
NOTE 2.2.1 ransmission System Operator of intent to reduce power for Turbine Valve testing. (N/A if already at desired power level) ctor Power to I 90% at a rate specified by the Supervisor. (N/A if already at desired power level)
If performing this test at powe 75%, be sure to place ICs I CAUTION Be sure to wait 5 10 minutes after placing ICs in HAND to allow plant to stabilize, especially if plant is at reduced power levels.
- 3. START the standby EHC pump.
in HAND. This step is N/A if power is > 75%.
2.2.2 Valves (TG-CIV-1,2,3,4,5,6)
IN DISPLAY screen, EXIT to MAIN MENU Page # I .
IN MENU Page #I, SELECT screen #23 VALVE TEST PREREQS.
Page 8 of 23
he desired CIV (Full Stroke) to be tested on the VALVE NOTE will close followed by the valves to fully close before
GGERED PLOT display, VERIFY FULL-STROKE is a dicating full stroke testing has been properly selected.
I NOTE test by SELECTING START and EXECUTE
ot for smooth stem stroking and FAST CLOSING in about the last IO% of valve travel as indicated p to < 0% indicated position.
the valves then return to their previous open positions.
ile for later retrieval.
plant to stabilize before further testing.
Steps 4 through 12 for remaining CIV's.
is to be done, proceed to Section 1.2.2.3. (This step testing will not be performed.)