ML031710800
| ML031710800 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| 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 Iacility: Three Mile Island Unit 1 Exam Level (circle one): RO / SRO(I) i SRO(U)
Control Room Systems and Facility Walk-Through Test Outline Form ES-301-2 Date of Examination: May 12,2003 Operating Test No,:
- c. Emergency Core Cooling System (OOG)/Respond to a High Pressure Injection (HPI) initiation (Alt. Path - MU-V-14A fails).
B.l Control Room Systems N, A, S SystemNPM Title
- e. Main Steam System (039)//
pvLdrnfJi 5 Toid vn 1 v t >
T C J r
- f.
Containment Cooling System (022)iReturn Reactor Building (RB)
Emergency Cooling to Engineered Safeguards Standby.
Emergency Diesel Generator (EDG) System (064)iEDG Operation (Alt. Path - EDG Fails to Auto Load).
- g.
Type Code*
I I
Function safety D, S 4 Secondary N, S 5
D, A, s 6
I I
- a. Chemical and Volume Control (004)iPerform an Emergency Boration (Alt. Path - Backup Emergency Boration Required).
- a. Chemical and Volume Control System (004)IManually Open RCP N, R Seal Injection Isolation Valve (MU-V-26).
1 2
Emergency I
I
- b. Engineered Safety Feature Actuation Systems (01 3)IRespond to inadvertent ES Actuation.
- c. Emergency Feedwater System (061)/Local Reset of Emergency Feedwater Pump (EF-P-1).
2 D
4 Secondary Emergency 3
- d. Residual Heat Removal System (0OS)iRespond to a failure of 1
N,A, S, L 1 4~rimary Low Pressure Injection (Alt. Path - DHV-6 Fails to Open).
- b. Pressurizer Pressure Control System (01 O)/Transfer Pressurizer Heater Group 8 or 9 to an Engineered Safeguards Bus.
D l
3 I
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 (71 7) 948-2074/2093.
Respectfully, Dennis May I
Training 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-2 1 NRC Written Exam Review by NRC Attendees:
Dennis May Dave Atherholt Joseph D Antonio (NRC)
Dell McNeil (NRC)
Question #
Initial Comment Resolution Official NRC status 3
5 8
14 17 18 20 24 27 30 31 34 42 44 45 54 55 58 68 69 71 78 80 85 88 89 90 96 97 too easy too easy too easy SRO, but no mark one distractor not plausible Q logic error KA: not appl at TMI too easy RO, but 55.43 checked not meeting KA 2 ans. parts, but need only 1 NOT comprehensive needs reference supplied NOT memory two reviewers disagreed marked RO, should be SRO marked SRO, should be RO too easy KA: not appl at TMI too easy KA interpretation differs too easy NOT comprehensive marked SRO, should be RO needs rewording marked RO, should be SRO SRO, but 55.43 checked not meeting KA marked SRO, should be RO keep as is ok REWRITTEN NEW rewrite new keep as is ok marked 55.43 box NC keep as is (system explained)
NC keep as is (reading error)
NC changed to similar KIA ok KIA per NRC keep as is ok agreed with our KA interpret.
NC keep as is (misunderstood Q)
NC rewrite distractors with like parts changed to MEMORY ok already marked in database NC change to COMPREHENSIVE ok no comment NC partial rewrite changed to SRO ok changed to RO ok keep as is ok applied well, but scrub KA later ok keep as is ok Keep Q, assign appl KA per NRC ok KIA per NRC rewrite new rewrite new changed to MEMORY ok changed to RO ok rewritten with NRC help rewritten & OKd Kept SR0,reordered 2 distracters NC checked 55.43 box ok keep as is (misread Q)
NC Kept SRO NC
- NC = No Comment
3.4*
Group #
ROlSRO Importance Rating 2.5 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.
101 3, Temporary Modifications and Bypass of Safety Functions, pages 1-4, Rev 51.
L New TMlBank TMI Question #
- 5 SR021 AUDIT 0
Modified TMI Bank
&?I Memory or Fundamental Knowledge 3 Comprehension or Analysis
&?I 55.41
. I O d 55.43.3 lid 55.45.I3 Parent Question #
A Incorrect answer. This is a temporary installation that is excluded from 1013 control in section 2.2.3.4, page 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
- 4.
swer. Excluded in section 2.2.3.2, page 3.
TMI SRO Exam - May 2003 Thursday, April 17,2003
Page # 2-7 Tier #
3 SYSiEP# Gen KA# 2.2.19 Group #
ROlSRO Importance Rating 2.1 3.1 Knowledge of maintenance work order requirements. (Equipment Control)
Plant conditions:
- A maintenance work order requires a pneumatic operated valve to be closed as
- The valve fails OPEN on loss of air or loss of power to the solenoids controlling part of the clearance order safety boundary.
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 E Memory or Fundamental Knowledge Parent Question #
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 3.7 Group #
ROlSRO Importance Rating 2.5 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 21 1.
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 21 9.
Technical Specification 6.2.2.2.d, page 6-1, Amendment 21 9.
Technical Specification 6.2.2.2.a, page 6-1, Amendment 21 9.
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.I2 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 3.5 Group #
2 ROISRO Importance Rating 3.2 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:
Based on these conditions, assuming no operator action, complete the statement below that predicts the effect of this event.
- Fuel pin failure occurs.
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 2l Memory or Fundamental Knowledge Parent Question #
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 1
3.6 Group #
ROlSRO Importance Rating 3.4 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 Operating Procedure
- Title I
105-8 Revision No.
Radiation Monitoring System 2.2 Administrative 67 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 1302-1 7.3, 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 Title
-Y 1 105-8 Revision No.
Radiation Monitoring System
- 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.
67 C.
- d.
- e.
- f.
RM-G-9 RM-G-I 8 RM-G-I6 RM-G-20 RM-G-17 RM-G-21 Ensure the rotary switch on each individual control panel module is selected to the "ALL" position Verify that the green "FAIL" light comes to indicate power is on.
Ensure the electronics warm up for 30 minutes for guaranteed accuracy.
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 1 302-3,l IC-1 77 OR verify procedure and/or surveillance are on file for:
0 1302-3.1 0
IC-I 77 Complete the "Satisfactory Alert Setpoint Check" AND "Satisfactory Alarm Setpoint Check" columns in Table 4.
I.
7
TMI - Unit 1 Operating Procedure Title Radiation Monitoring System 11 05-8 Revision No.
67 MU Demin Area AB 305 RM-G-12 Instrument RM-G-11 I None Tech Spec or ODCM Requirements None Solid Radwaste Processing Area AB 305 RM-G-13 Aux Bldg Entrance AB 281 RM-G-14 Waste Evap Area AB 281 RM-G-15 I None None None Hx Vault AB 271 OTSG A sample line RCDT return RM-G-21 RB sump return RM-G-22 RB Inside A D-ring RM-G-23 RB Inside B I None None TS table 3.5-3 TS table 3.5-3 None I
RM-G-I8 RCS sample line RM-G-19 I None RCP seal return line RM-G-20 I None D-ring RM-G-26 I TS table 3.5-3 OTSG A steam line RM-G-27 1 TS table 3.5-3 OTSG B steam line Compensatory Action None None None None None
- Note 1
'*Note 1
- "Note 1 Verify RM-L-1 Lo is operable
- Note 1
- Note 1 Per Tech Spec Per Tech Spec Per Tech Spec Per Tech Spec Interlock None None I
I None I
Switch in DEFEAT (PRF)
I Interlock Closes CA-V-4A & 5A Switch in DEFEAT (PRF)
I Interlock Closes CA-V-48 & 58 Switch in DEFEAT,
i I
(PRF)
Interlock Closes CA-V-I
,2,3 & 13 None Switch in DEFEAT (PRF)
Interlock Closes WDL-V-303 & 304 Closes WDG-V-3 & 4 Switch in DEFEAT
( P W Interlock Closes WDL-V-534 & 535 None None None
- NOTE 1:
NOTE 2:
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.
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 Title
- -1 Radiation Monitoring System I
105-8 Revision No.
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.
Tech Spec or ODCM Requirements None 3.5.1 Review TS & ODCM requirements (refer to Table 3.5).
Compensatory Action Interlock None None 3.5.2 Perform compensatory action required per Table 3.5.
None None 3.5.3 Make a log entry which identifies the inoperable instruments and compliance with Tech Spec or ODCM requirements.
None 3.5.4 Place interlock mode switch in DEFEAT. Refer to Table 3.5 for switch location and interlock description.
None None 3.5.5 Place monitor in OFF None None None None 3.5.6 When appropriate, return monitor to service IAW Section 3.1 None None TS 3.8.1 Instrument ALC-RMI-10 None None None None Per Tech SDec None CCB Area (between CC-T-1 &
CC-T-2)
ALC-RMI-11 CCB Area (above CCB sump)
RM-G-1 TS 3.8.1 TS 3.8.1 Control Room RM-G-2 Per Tech Spec None
- NOTE 2 Per Tech Spec Switch in DEFEAT (PRF)
Rad Chem Lab Nuc Sampling Room Hot Machine Shop RB Personnel Access Door RB Aux FH Bridge RB Main FH Bridge FHB FH Bridge RM-G-3 RM-G-4 RM-G-5 RM-G-6 RM-G-7 RM-G-9 None RM-G-10 Aux Bldg Entrance AB 305 Interlock Trips AH-E-IO Closes AH-D-120,121 8,122 None None None I None I None
Form ES-401-6 Q # 071 Page # 3.6-6 Tier #
2 SYSIEP# 063 K A # K3.01 I
ROlSRO Importance Rating 3.7 4.1 Group #
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 1AllCII E Trouble
- A-3-7 Inverter 1N1 C/1 E 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 IA.
B. (1 ) 1202-9A, Loss of "A' DC Distribution System.
C. (1) Alarm response for AA-3-2, 7KV Bus Trouble.
D. (1) Alarm response for "-3-1, 230 KV Substation Trouble.
(2) Notify Auxiliary Operator to verify MU-P-1 C is ready for start (2) Notify Transmission System Operator (TSO) and trip the reactor (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 3 Memory or Fundamental Knowledge Parent Question #
Comprehension or Analysis I
VI 55.43.5 d 55.45.6 A Correct answer.
B Incorrect answer. Part #I is correct 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 2.8 Group #
2 ROISRO Importance Rating 2.6 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 cons tan t.
- 6. (I) Makeup valve MU-V-17 OPENS to loo%, Pressurizer heaters trip.
C. (I) Makeup valve MU-V-17 CLOSES to 0%, Pressurizer heaters energize in (2) Raise RCS Letdown flow to maximum (140 gpm).
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
2 -
Page # 2-7 Tier #
SYSIEP# 008 K A # 2.2.22 4.1 Group #
3 ROlSRO Importance Rating 3.4 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)
(b)
A. (a) Two (b) Two B. (a) Two (b) One C. (a)One (b) Two D. (a) One (b) One NSCC pump(s) islare required for normal operation heat loads; NSCC purnp(s) islare required for ECCS support during a LOCA.
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 2 55.45.2
- 55.41 k 55.43.2 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
1 ROlSRO Importance Rating 3.7 4.2 Group #
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.
not reclose.
- CRD power supply breaker associated with 'B' RPS cabinet is open, and will
- 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 I
ROlSRO Importance Rating 4.0 4.0 Group #
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:
- LOSS Of BUS ATA.
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
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
1 Page # 4.1-17 Tier #
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.
ROlSRO Importance Rating 4.0 4.6 Group #
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
- At the end of the pressure reduction, Pressurizer level rose rapidly to 1660 psig.
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.
Solid Operations.
(2) Exit EOP-001 and GO TO OP-TM-EOP-009 HPI Cooling - Recovery From C. (1) Expansion of RCS loop water due to depressurization..
D. (1) Expansion of RCS loop water due to depressurization.
(2) Continue with EOP-001 VSSVs.
(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.
II l.c.07.04 -
4 New - TMIBank TMI Question ##
- Modified TMI Bank
~
Memory or Fundamental Knowledge I
3 Comprehension or Analysis C 55.41 4 55.43.5 V 55.45.I3 Parent Question ##
TMI SRO Exam - May 2003 Thursday, April 17,20W
Q # 031 Form ES-401-6 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
1 1
Page ## 4.1-9 Tier #
3.3 Group #
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.
~
A k
Sequence of events:
- Reactor power was initially 1 OO%, with ICs in full automatic
- No maintenance or surveillance tests in progress.
- 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
- 55.43 Q-55.45
.5/.6 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 I4 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 1
Page # 4.2-11 Tier #
1 3.7 Group #
RO/SRO Importance Rating 3.7 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-1 B 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 11 02-4, 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 Responsibilities of 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
1 Page # 4.1-2 Tier #
3.6 Group #
2 ROlSRO Importance Rating 3.5 Knowledge of the interrelations between Reactor Trip and the following: Reactor trip status A T v
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
-1 55.43 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 A
7.4 RPS Test Circuits and Trip Relay Operation (refer to Figures 61, 62)
CB 10 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.
B 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.
CB 11 I RPSCABMET I BKR(s) Tested from that Cabinet C
CB1, CB2 D
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.
CB3, CB4 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
c 1)
,/ --&$+-
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 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 A channel 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.
I 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 to 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 RZLAYS PROGRAMMER I R A N S F E R RELAVS (NOT SHOWN)
I I
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( P I. TRAVEL L I M I T AND I I C S I G N A L S )
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 Y P I C A L OF ALL)
FOR GROUP 4 B EACH GROUP)
POWER SUPPLY OlSCONNEClS TRANSFER -
-t
( T Y P I C A L Of 8 I
E I C H CROUP)
I v
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TO GROUP 5 IO GROUP 6 TO GROUP 7 TO GROUP 8 v
T O GROUP 4
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.
009 088 054 Technical Specifications COLR Rev 0, abstract section.
(Graph of COLR rod insertion limits must be given as part of question)
Technical Specifications page 3-24, Amendment 227.
Thursday, April 17,2003 Page 1 of 1
L
ES-40 1 PWR SRO Examination Outline Cat 1 4
Facility:
Three Mile Island - 1 Cat 2 Cat 3 Cat 4 5
4 4
Exam Date: 05/12/2003 Printed: 04/17/2003 Form ES-40 1-3 Exam Level: SRO Tier 1.
Etnergencq Abnormal Plant Evolutions
- 2.
Plant Systems
- 3. Generic Knowledge K/A Category Points Group And Abilities Point Total 24 16 3
43 19 17 4
40 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
- 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.
o or three K/A topics from a given s) stem unless thq relate to plant-specific priorities.
1
PWR SRO Examination Outline Printed:
0411 7/2003 Facility:
Three Mile Island - 1 SS - 401 YAPE #
00 1 003 003 005 01 1 01 1 015 017 026 026 029 Emerge1 EIAPE Name I Safety Function Continuous Rod Withdrawal I 1 Dropped Control Rod I I Dropped Control Rod / 1 Inoperable/Stuck Control Rod / 1 Large Break LOCA 1 3 Large Break LOCA / 3 Reactor Coolant Pump (RCP) Malfunctions / 4 Reactor Coolant Pump (RCP) Malfunctions (Loss of RC Flow) / 4 Loss of Component Cooling Water (CCW) I 8 Loss of Component Cooling Water (CCW) / 8 Anticipated Transient Without Scram (ATWS) / I y and A KA AA2.04 AK2.05 AA 1.06 2.1.33 EKI.01 2.4.6 AA2.10 AK3.03 AA 1.07 inormal Plant Evolutions - Tier 1 I Group 1 KA Topic Reactor power and its trend Control rod drive power supplies and logic circuits RCS pressure and temperature Axial power imbalance Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.
Natural circulation and cooling, including reflux boiling Knowledge symptom based EOP mitigation strategies.
When to secure RCPs on loss of cooling or seal injection Guidance actions contained in EOP for Loss of ccw Flow rates to the components and systems that are serviced by the CCWS; interactions among the components Breakers, relays, and disconnects Form ES-401-.
Comment 1
PWR SRO Examination Outline Printed:
0411 712003 Facility:
Three Mile Island - I 3s - 401 VAPE #
029 05 I 055 067 069 069 074 A03 A03 A06 Emerge]
E/APE Name / Safety Function Anticipated Transient Without Scram (ATWS) / 1 Loss of Condenser Vacuum 14 Loss of Offsite and Onsite Power (Station Blackout) 16 Plant Fire on Site I 9 Loss of Containment Integrity I 5 Loss of Containment Integrity I 5 Inadequate Core Cooling I 4 Loss of "I-Y I 7 LOSS of NNI-Y I 7 Shutdown Outside Control Room I 8 y and P KA EA1.02 AK3.O 1 m
2.1.32 AA2.01 AKI.OI EA2.06 AK2.2 AAI.1 AKI.3 normal Plant Evolutions - Tier 1 /Group 1 KA Topic Charging pump suction valves from RWST operating switch Loss of steam dump capability upon loss of condenser vacuum Length of time for which battery capacity is designed Ability to explain and apply all system limits and precautions.
Loss of containment integrity Effect of pressure on leak rate Changes in PZR level due to PZR steam bubble transfer to the RCS during inadequate core cooling Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Annunciators and conditions indicating signals, and remedial actions associated with the (Shutdown Ouside Control Room)
Form ES-40 1 -.
Comment 20 24 25 28 29 SRO 30 31 SRO 33 34 35 2
PWR SRO Examination Outline KA EK2.1 2.4.30 EK3.2 Facility:
Three Mile Island - 1 KATopic Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Knowledge of which events related to system operations/status should be reported to outside agencies.
Normal, abnormal and emergency operating procedures associated with (Natural Circulation ES - 401
$/APE ##
E05 E09 E09 Emerge]
EIAPE Name / Safety Function Excessive Heat Transfer / 4 Natural Circulation Operations / 4 Natural Circulation Operations / 4 Printed:
04/17/2003 Form ES-401-3 Comment 37 40 SRO 41 3
PWR SRO Examination Outline Printed:
0411 712003 Facility:
Three Mile Island - 1 ES - 401 EIAPE ##
007 007 008 008 009 009 022 027 033 03 8 03 8 Emerge1 EIAPE Name I Safety Function Reactor Trip / 1 Reactor Trip I 1 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) / 3 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) / 3 Small Break LOCA 13 Small Break LOCA 1 3 Loss of Reactor Coolant Makeup 12 Pressurizer Pressure Control (PZR PCS)
Malfunction / 3 Loss of Intermediate Range Nuclear Instrumentation I 7 Steam Generator Tube Rupture (SGTR) / 3 Steam Generator Tube Rupture (SGTR) / 3 y and P KA EK2.03 EA 1.03 AK2.01 m
2.2.25 EK1.O1 AK 1.02 AK2.03 AK3.01 EA 1.OS EA2.09 lnormal Plant Evolutions - Tier 1 / Group 2 KA Topic Reactor trip status panel RCS pressure and temperature Valves Actions contained in EOP for PZR vapor space accident1LOCA Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
Natural circulation and cooling, including reflux boiling Relationship of charging flow to pressure differential between charging and RCS Controllers and positioners Termination of startup following loss of intermediate-range instrumentation Core cooling monitor Existence of natural circulation, using plant parameters Form ES-401-2 Comment 5
6 7
98 9
I O 15 18 21 22 23 I
PWR SRO Examination Outline EIAPE Name I Safety Function Area Radiation Monitoring (ARM) System Alarms 1 Printed:
0411 712003 KA AA2.01 Facility:
Three Mile Island - 1 Plant Runback I 1 ES - 401 EIAPE #
06 1 06 1 AAI.1 A0 1 LOCA Cooldown 14 EO8 EKl.3 E08 7
I Area Radiation Monitoring (ARM) System Alarms I 12.1.32 7
normal Plant Evolutions - Tier 1 / Group 2 KA Topic ARM panel displays Ability to explain and apply all system limits and precautions.
Components, and hnctions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments Annunciators and conditions indicating signals, and remedial actions associated with the (LOCA Cooldown)
Form ES-401-:
Comment 26 SRO 27
PWR SRO Examination Outline Facility:
Three Mile Island - I E/APE #
A08 E13 E13 Printed:
04/17/2003 KA KATopic Comment E/APE Name / Safety Function Refuel Canal Level Decrease / 8 AA2.1 Facility conditions and selection of appropriate 36 SRO procedures during abnormal and emergency operations EOP Rules 2.2.22 Knowledge of limiting conditions for operations 42 and safety limits.
EOP Rules EKI.2 Normal, abnormal and emergency operating 43 procedures associated with (EOP Rules)
PWR SRO Examination Outline Printed: 0411 712003 Facility:
ES - 401 Three Mile Island - 1 Sys/Ev I#
00 1 003 003 004 004 015 015 022 026 026 06 1 06 I System / Evolution Name Control Rod Drive System / 1 Reactor Coolant Pump System (RCPS) I 4 Reactor Coolant Pump System (RCPS) 1 4 Chemical and Volume Control System (CVCS) I 1 Chemical and Volume Control System (CVCS) I 1 Nuclear Instrumentation System / 7 Nuclear Instrumentation System 17 Containment Cooling System (CCS) / 5 Containment Spray System (CSS) I 5
Containment Spray System (CSS) I 5
Auxiliary I Emergency Feedwater (AFW) System / 4 Auxiliary / Emergency Feedwater JAFW) Svstem 14 KA K6.03 K3.03 A2.02 2.4.4 A4.18 K4.04 K6.04 K2.0 1 K1.O1 A3.01 K5.02 K6.0 1 Plant Systems - Tier 2 / Group 1 KA Topic Reactor trip breakers, including controls Feedwater and emergency feedwater Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures.
Emergency borate valve Slow response time of SPNDs Bistables and logic circuits Containment cooling fans ECCS Pump starts and correct MOV positioning Decay heat sources and magnitude Controllers and positioners Form ES-401-;
Comment 44 47 48 SRO 49 SRO 50 58 59 60 61 62 69 70 1
PWR SRO Examination Outline 063 068 068 07 1 072 Printed: 0411 712003 System I 6 D.C. Electrical Distribution K4.04 Trips 72 System I 6 Liquid Radwaste System (LRS) I 9 K1.07 Sources of liquid wastes for LRS 73 Liquid Radwaste System (LRS) 19 A3.02 Automatic isolation 75 Waste Gas Disposal System A2.02 Use of waste gas release monitors, radiation, gas 76 SRO (WGDS) I 9 Area Radiation Monitoring (ARM)
K3.02 Fuel handling operations 77 SRO flow rate, and totalizer 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 I System I 7 I
I I
I I
I 072 I Area Radiation Monitoring (ARM) I Al.01 I Radiation levels 1 78SRO 2
PWR SRO Examination Outline Printed: 0411 712003
- SysIEv #
002 Facility:
Three Mile Island - 1 System I Evolution Name KA Reactor Coolant System (RCS) 12 K3.02 ES - 401 Plant Systems - Tier 2 /Group 2 I
I I
I 002 Reactor Coolant System (RCS) / 2 K4.05 01 1 Spurious trip protection 012 012 Reactor Protection System 17 Pressurizer Level Control System (PZR LCS) / 2 Reactor Protection System 17 K6.11 KA Topic Fuel 03 3 034 03 5 Spent Fuel Pool Cooling System (SFPCS) 1 8 Fuel Handling Equipment System (FHES) / 8 Steam Generator System (SICS) I 055 I
029 I Containment Purge System (CPS) /
Condenser Air Removal System (CARS) 14 14 Main and Reheat Steam System (MRSS) 14 Main and Reheat Steam System (MRSS) / 4 K1.O1 A3.01 A2.02 2.4.49 2.4.6 A4.07 A3.03 Trip setpoint calculators
~~
Gaseous radiation release monitors Temperature control valves Dropped cask Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
Knowledge symptom based EOP mitigation strategies.
Steam dump valves Automatic diversion of CARS exhaust Form ES-401-Comment 45 46 SRO 55
~ 56 63 64 65 SRO 81 SRO 66 SRO 67 68 1
Facility:
Three Mile Island - 1 ES - 401 KA Topic Control power Comment 73 Sys/Ev #
364 373 373 375 PWR SRO Examination Outline System / Evolution Name KA Emergency Diesel Generator K2.03 (EDIG) System / 6 Process Radiation Monitoring K4.01 (PRM) System / 7 Process Radiation Monitoring A1.O1 (PRM) System 17 Circulating Water System I 8 K3.07 Printed: 0411 712003 Release termination when radiation exceeds setpoint 79 Radiation levels ESFAS Knowledge of which events related to system oaerationslstatus should be reported to outside 80 SRO 82 83 SRO 2
I03 Containment System I 5 2.4.30
PWR SRO Examination Outline Printed: 0411 712003 KA Topic RHR heat exchanger Heatuplcooldown rates Quench tank cooling Knowledge of limiting conditions for operations and safety limits.
Facility:
Three Mile Island - 1 Comment 51 52 53 54 SRO ES - 401 Sys/J%v #
005 Residual Heat Removal System System / Evolution Name (RHRS) 14 Residual Heat Removal System (RHRS) I 4 Pressurizer Relief TankIQuench K4.0 1 Tank System (PRTS) / 5
Facility:
Three Mile Island - I Generic Category Conduct of Operations Radiation Control Equipment Control 2.3.1 KA 2.1.5 2.1.7 2.1.10 2.1.34 2.2.1 2.2.1 1 2.2.19 2.2.26 2.2.27 2.3.3 2.3.8 2.3.10 Generic Knowledge and Abilities Outline (Tier 3)
PWR SRO Examination Outline KA Topic Ability to locate and use procedures and directives related to shift staffing and activities.
Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Knowledge of conditions and limitations in the facility license.
Ability to maintain primary and secondary plant chemistry within allowable limits.
Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.
Knowledge of the process for controlling temporary changes.
Knowledge of maintenance work order requirements.
Knowledge of refueling administrative requirements.
Knowledge of the reheling process.
Knowledge of I O CFR: 20 and related facility radiation control requirements.
Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems).
Knowledge of the process for performing a planned gaseous radioactive release.
Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
Printed: 0411 712003 Form ES-401-5 Comment 85 86 SRO 87 SRO 84 Category Total:
4
~~~
92 90 SRO 89 SRO 88 SRO 91 SRO Category Total:
5 95 94 SRO 93 96 SRO Category Total:
4 1
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L a3... 5 P
B L7
ES-401 PWR SRO Examination Outline Printed: 0411 712003 Facility:
Three Mile Island - 1 Form ES-40 1-3 Tier Exam Date: 05/12/2003 Exam Level:
WA Category Points Group SRO 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
- 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.
they relate to plant-specific priorities.
I 1
ES-401 PWR SRO Examination Outline Printed: 0411 712003 Facility:
Three Mile Island - 1 Form ES-40 1-3 2
ES-401 PWR SRO Examination Outline Facility:
Three Mile Island - 1 Exam Date: 05/12/2003 Tier Printed: 04/17/2003 Group Form ES-40 1-3 Exam Level: SRO IUA Category Points Point I
cat 1 1
c a t 2 I
c a t 3 I c a t 4 I
- 3. Generic Knowledge And Abilities 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
- 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.
they relate to plant-specific priorities.
1
PWR SRO Examination Outline Printed:
04/17/2003 Facility:
Three Mile Island - I ES - 401
</APE #
00 1 003 003 005 01 1 01 1 01 5 017 026 026 Emei E/APE Name / Safety Function Continuous Rod Withdrawal / 1 Dropped Control Rod / 1 Dropped Control Rod / 1 Inoperable/Stuck Control Rod / 1 Large Break LOCA / 3 Large Break LOCA / 3 Reactor Coolant Pump (RCP) Malfunctions / 4 Reactor Coolant Pump (RCP) Malfunctions (Loss of RC Flow) / 4 Loss of Component Cooling Water (CCW) / 8 Loss of Component Cooling Water (CCW) / 8 ency and Abnormal Plant Evolutions - Tier 1 / Group 1 KA Topic AA2.04 - Reactor power and its trend AK2.05 - Control rod drive power supplies and logic circuits AA 1.06 - RCS pressure and temperature AK 1.O 1 - Axial power imbalance 2.1.33 - Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications.
EK1.01 - Natural circulation and cooling, including reflux boiling 2.4.6 - Knowledge symptom based EOP mitigation strategies.
AA2.10 - When to secure RCPs on loss of cooling or seal injection AK3.03 - Guidance actions contained in EOP for Loss of ccw AAI.07 - Flow rates to the components and systems that are serviced by the CCWS; interactions among the components Form ES-401-3 Imp.
4.3 2.8 4.1 3.8 4.0 4.4 4.0 3.7 4.2 3.O Points 1
1 1
1 1
I I
1 1
1 1
PWR SRO Examination Outline Printed:
0411 712003 Facility:
Three Mile Island - 1 ES - 401 Eme E/APE #
029 029 05 1 055 067 069 069 074 E/APE Name I Safety Function Anticipated Transient Without Scram (ATWS) / 1 Anticipated Transient Without Scram (ATWS) I 1 Loss of Condenser Vacuum 14 Loss of Offsite and Onsite Power (Station Blackout) /
6 Plant Fire on Site / 9 Loss of Containment Integrity I 5 Loss of Containment Integrity I 5 Inadequate Core Cooling 14 rnd Abnormal Plant Evolutions - Tier 1 / Group 1 Form ES-401-3 KA Topic EK2.06 - Breakers, relays, and disconnects EAl.02 - Charging pump suction valves from RWST operating switch AK3.01 - Loss of steam dump capability upon loss of condenser vacuum EK3.O 1 - Length of time for which battery capacity is designed 2.1.32 - Ability to explain and apply all system limits and precautions.
AA2.0 1 - Loss of containment integrity AK I.O 1 - Effect of pressure on leak rate EA2.06 - Changes in PZR level due to PZR steam bubble transfer to the RCS during inadequate core mp.
3.1*
3.3 3.1*
3.4 3.8 4.3 -
3.1 4.6 Points 1
1 1
1 1
cooling 2
PWR SRO Examination Outline Facility:
Three Mile Island - 1 42 ES - 401 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 G
x Imp.
3.3 4.0 3.4 4.0 3.6 3.8 Points 1
1 1
1 1
I KIA Category Totals:
4 4
4 4
E05 E09 Excessive Heat Transfer 14 X
Natural Circulation Operations 1 4 Printed:
0411 712003 E09 KA Topic AK2.2 - Facility's heat removal systems, including 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 safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features AKI.3 - Annunciators and conditions indicating signals, and remedial actions associated with the (Shutdown Ouside Control Room)
EK2.1 - Components, and hnctions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Natural Circulation Operations 14 X
2.4.30 - Knowledge of which events related to system operationslstatus should be reported to outside agencies.
EK3.2 - Normal, abnormal and emergency operating procedures associated with (Natural Circulation Operations)
Form ES-401-3 4
4 Group Point Total:
24 3
Facility:
ES - 401 KA Topic EK2.03 - Reactor trip status panel Three Mile Island - 1 Emer Imp.
3.6 E/APE #
007 EAI.03 - RCS pressure and temperature AK2.01 - Valves AK3.03 - Actions contained in EOP for PZR vapor space accidentILOCA 2.2.25 - Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.
EKI.01 - Natural circulation and cooling, including reflux boiling AKI.02 - Relationship of charging flow to pressure differential between charging and RCS 007 4.1 2.7 4.6 3.7 4.7 3.1 008 008 AK2.03 - Controllers and positioners AK3.01 - Termination of startup following loss of intermediate-range instrumentation 009 009 2.8 3.6 022 027 033 E/APE Name / Safety Function Reactor Trip I 1 Reactor Trip I 1 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) 13 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) I 3 Small Break LOCA 13 Small Break LOCA 13 Loss of Reactor Coolant Makeup 12 Pressurizer Pressure Control (PZR PCS) Malfunction 1 3 Loss of Intermediate Range Nuclear Instrumentation I 7
PWR SRO Examination Outline and Abnormal Plant Evolutions - Tier 1 / Group 2 Printed:
0411 712003 Form ES-401-3 Points 1 -
1 1
1 1
1 1
1 1
1
Facility:
ES - 401 KA Topic EA1.OS - Core cooling monitor Three Mile Island - 1 Emergency Imp.
Points 3.8*
1 E/APE #
03 8 038 06 1 06 1 A0 1 E08 EO8 EA2.09 - Existence of natural circulation, using plant parameters AA2.01 - ARM panel displays Steam Generator Tube Rupture (SGTR) / 3 4.2 1
3.7 1
Area Radiation Monitoring (ARM) System Alarms / 7 Area Radiation Monitoring (ARM) System Alarms / 7 Plant Runback / 1 2.1.32 - Ability to explain and apply all system limits and precautions.
AAI.1 - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features EA2.2 - Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments WA Category Totals:
3 3.8 1
3.7 1
4.0 1
PWR SRO Examination Outline ind Abnormal Plant Evolutions - Tier 1 / Group 2 Printed:
0411 712003 Form ES-401-3 I
I EKI.3 - Annunciators and conditions indicating signals, I 3.5 I 1
and remedial actions associated with the (LOCA -
I I
3 2
3 3
2 Group Point Total:
16 2
Facility:
Three Mile Island - 1 ES - 401 E/APE #
A08 E13 E13 PWR SRO Examination Outline Emergency and Abnormal Plant Evolutions - Tier 1 / Group 3 EIAPE Name / Safety Function K1 K2 K3 A1 A2 G KATopic Imp.
Points Refbel Canal Level Decrease / 8 X
AA2.1 - Facility conditions and selection of appropriate 4.0 1
procedures during abnormal and emergency operations EOP Rules X 2.2.22 - Knowledge of limiting conditions for operations 4.1 1
EOP Rules X
EK 1.2 - Normal, abnormal and emergency operating 3.6 1
and safety limits.
procedures associated with (EOP Rules)
Printed:
0411 712003 Form ES-401-3 WA Category Totals:
1 0
0 0
1 1
Group Point Total:
3 1
PWR SRO Examination Outline Printed: 04/17/2003 2
Y Facility:
KATopic K6.03 - Reactor trip breakers, including controls K3.03 - Feedwater and emergency feedwater A2.02 - Conditions which exist for an abnormal shutdown of an RCP in comparison to a normal shutdown of an RCP 2.4.4 - Ability to recognize abnormal ES - 401 iys/Ev #
00 1 003 003 004 004 015 015 022 026 026 06 1 Three Mile Island - 1 Reactor Coolant Pump System (RCPS) / 4 (RCPS) / 4 I
Chemical and Volume Control System I (CVCS) / 1 Chemical and Volume Control System (CVCS) / 1 Nuclear Instrumentation System / 7 Nuclear Instrumentation System / 7 Containment Cooling System (CCS) /
5 Containment Spray System (CSS) / 5 X
Containment Spray System (CSS) I 5 Auxiliary / Emergency Feedwater (AFW) System 14 indications for system operating parameters I which are entry-level conditions for emergency and abnormal operating procedures.
A4.18 - Emergency borate valve t K4.04 - Slow response time of SPNDs K6.04 - Bistables and logic circuits K2.01 - Containment cooling fans K1.O1 - ECCS A3.01 - Pump starts and correct MOV positioning K5.02 - Decay heat sources and magnitude Form Imp.
4.2 3.1 -
3.9 4.3 4.1 3.6? -
3.2 3.1 4.2 -
4.5 3.6 -
S-401-2 Points 1
1 1
1 1
1 1
1
Facility:
K4 X
ES - 401 SysIEv #
06 1 063 063 068 068 07 1 072 072 K5 Three Mile Island - 1 K1 X
System I Evolution Name Auxiliary I Emergency Feedwater K2 (AFW) System I 4 D.C. Electrical Distribution System I 6 D.C. Electrical Distribution System 1 6 Liquid Radwaste System (LRS) 19 Liquid Radwaste System (LRS) 19 Waste Gas Disposal System (WGDS) 19 Area Radiation Monitoring (ARM)
System I 7 Area Radiation Monitoring (ARM)
Svstem I 7 KIA Category Totals:
2 1
PWR SRO Examination Outline Printed 0411 712003 2
1 X
X 3
1 2
2 1
1
- roup 1 KA Topic K6.0 1 - Controllers and positioners K3.01 - ED/G K4.04 - Trips K1.07 - Sources of liquid wastes for LRS A3.02 - Automatic isolation A2.02 - Use of waste gas release monitors, radiation, gas flow rate, and totalizer K3.02 - Fuel handling operations A1.O1 - Radiation levels Form
[mp.
2.8*
4.1 -
2.9?
2.9 -
3.6 3.6 3.5 -
3.6 S-4011 Points 1
1 Group Point Total: 19 2
PWR SRO Examination Outline Printed: 0411 712003 G
Facility:
KATopic K3.02 - Fuel ES - 401 s y s m ##
002 002 X
X 01 1 K4.05 - Spurious trip protection K6.11 - Trip setpoint calculators K1.01 - Gaseous radiation release monitors A3.01 - Temperature control valves A2.02 - Dropped cask 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.
2.4.6 - Knowledge symptom based EOP mitigation strategies.
A4.07 - Steam dump valves 012 012 029 03 3 034 035 039 039 Three Mile Island - 1 Reactor Coolant System (RCS) / 2 I I
Pressurizer Level Control System I
(PZR LCS) / 2 Reactor Protection System / 7 Reactor Protection System / 7 Containment Purge System (CPS) / 8 X
Spent Fuel Pool Cooling System (SFPCS) / 8 I
Fuel Handling Equipment System I
(FHES) I 8 Steam Generator System (S/GS) / 4 Main and Reheat Steam System (MRSS) / 4 Main and Reheat Steam System (MRSS) / 4 ystemA Form I
K5.18 - Brittle fracture ?
I A2.08 - Loss of level compensation Imp.
4.5 -
3.6 2.8 2.9 -
2.9 3.7 2.7*
3.9 4.0 4.0 -
2.9
- S-401-:
Points 1
1 1
1
PWR SRO Examination Outline Printed: 04/17/2003 ns - Tier 44 Facility:
2 /
G X
CS - 401
$ys/Ev #
05 5 K4 X
064 K5 K6 073 073 ant 11 X
075 Sysl A2 Three Mile Island - 1 K2 X
System / Evolution Name K3 X
Condenser Air Removal System (CARS) / 4 3.6 4.3 3.5 3.5*
3.6 Emergency Diesel Generator (ED/G)
System / 6 1
1 1
1 1
Process Radiation Monitoring (PRM)
System I 7 Process Radiation Monitoring (PRM)
System 17 Circulating Water System / 8 Containment System / 5 103 K/A Category Totals:
1 2
2 1
1 1
2
- roup 2 KA Topic A3.03 - Automatic diversion of CARS exhaust K2.03 - Control power K4.0 1 - Release termination when radiation exceeds setpoint A 1.O 1 - Radiation levels K3.07 - ESFAS 2.4.30 - Knowledge of which events related to system operations/status should be reported to Form ES-401-2 outside agencies.
1 3
Group Point Total: 17 2
Facility:
44 ES - 401 G KATopic K6.03 - RHR heat exchanger 005 System / Evolution Name Residual Heat Removal System 007 K1 K2 K3 008 X
Three Mile Island - 1 2.2.22 - Knowledge of limiting conditions for operations and safety limits.
I I
I (RHRS) / 4 Residual Heat Removal System (RHRS) / 4 I l l Pressurizer Relief TarWQuench Tank System (PRTS) / 5 Component Cooling Water System (CCWS) / 8
- A Category Totals:
0 0
0 1
PWR SRO Examination Outline Printed: 0411 712003 Al.01 - Heatup/cooldown rates 1 K4.01 - Quench tank cooling Imp.
2.6 -
3.6 2.9 4.1 -
- S-401-2 Points 1
1 0
1 Group Point Total:
4 1
Facilitv:
Three Mile Island - 1 Generic Category Conduct of Operations 2.1.5 Equipment Control Ability to locate and use procedures and directives related to shift staffing and activities.
3.4 1
Generic Knowledge and Abilities Outline (Tier 3)
PWR SRO Examination Outline 2.1.7 2.1.10 Printed: 0411 712003 Form ES-401-5 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.
Knowledge of conditions and limitations in the facility license.
KA KATopic Imp.
Points Knowledge of 10 CFR: 20 and related facility radiation control requirements.
3.O 1
Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure.
4.4 3.9 3.3 1
I I
I Category Total:
4 2.2.1 2.2.1 1 2.2.19 2.2.26 2.2.27 Radiation Control I 2.3.1 2.3.3 2.3.8 I 2.3.10 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.
Knowledge of the process for controlling temporary changes.
Knowledge of maintenance work order requirements.
Knowledge of refueling administrative requirements.
Knowledge of the refueling process.
3.6 3.4*
3.1 3.7 3.5 1
1 1
1 1
Category Total:
5 Knowledge of SRO responsibilities for auxiliary systems that are outside the control room (e.g., waste disposal and handling systems).
Knowledge of the process for performing a planned gaseous radioactive release.
2.9 3.2 1
Generic Knowledge and Abilities Outline (Tier 3)
PWR SRO Examination Outline Knowledge of annunciator response procedures.
Knowledge of abnormal condition procedures.
Knowledge of the process used track inoperable alarms.
Knowledge of emergency plan protective action recommendations.
Facilitv:
Three Mile Island - 1 3.1 1
3.6 1
2.8 1
4.0 1
Generic Category Emergency ProceduredPlan r KA 2.4.10 2.4. I I 2.4.33 2.4.44 KA Topic Printed: 0411 712003 Form ES-401-5 Imp.
Points 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 1106-1 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.
(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 1 106-1 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
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.
Examinee reviews procedure precautions and steps.
Procedure reviewed by examinee CUE: As examinee reviews procedure, respond as needed for clarification, i.e. ENDPOINT on testing is step 2.2.2.1 2.
From Main Display screen, EXIT to Main Menu Page #I.
Examinee gets Main Menu Page #I on the screen.
From Main Menu page #I, select screen
- 23 Valve Stroke Test Prereqs.
Select FULL STROKE test and execute command on screen #23.
Exit to Main Menu Page #I and select VALVE STROKE TESTING under #23.
Select the desired CIV (Full Stroke) to be tested on the VALVE STROKE TESTING screen. (A plotigraph appears)
Examinee gets #23 Valve Stroke Test Prereqs on Main Menu Page #I.
Examinee executes FULL STROKE test.
Examinee selects VALVE STROKE TESTING on Main Menu Page #I.
~~
Examinee selects CIV-I on the Valve Stroke Testing screen.
On the TRIGGERED PLOT display, verify FULL-STROKE is a logic (1).
Examinee verifies Full Stroke logic (I) on the TRIGGERED PLOT display.
~
Review steps 7 - 11 prior to performance.(PER PROCEDURE NOTE)
Examinee reviews procedure steps 7 - 11.
Initiate test by selecting START and execute command.
Examinee executes START command.
Observe plot for smooth stem stroking and FAST CLOSING of both valves (CIVis made up of lV intercept Valve and ISV intercept Stop Valve) 0 note sharp drop in position indication step.
Examinee observes closing and FAST CLOSURE of CIV on plot.
NOTE: Examinee MUST wait until both valves (IV and ISV) stroke before performing next 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 STOP to terminate test.
Examinee selects STOP (after IV and ISV close).
Page 5 of 7
STANDARD SIU NOTE: Examinee may request local observer report on CIV open position. IC0 roleplay as required.
STEP
~
~~~~
~~
Examinee selects MORE OPTIONS, then SAVE IMAGE.
13 I
NOTE: Examine should terminate this JPM at this time.
Once valves reopen, select MORE OPTIONS 0
THEN select SAVE IMAGE END TASK Page 6 of 7
B.1.e 11.2.05.NEW Revision 0 OW1 2/2003 JPM CHANGE HISTORY PAGE REVISION r
0 DATE 0511 2/03 REFERENCE DESCRIPTION Include AI # if A Section 2.0 Page 7 of 7
TMI - Unit 1 Page 6 of 23 2.0 This test verifies free valve stem dump valves and records stroke transients are induced in the OT nt, fast closure feature of disc es. Because steam flow this test, its performance at I high power levels is by authorization of Plant Operations Director only.
I 2.1 Precautions 2.1.3 2.1.4 d#f 2.1.5 2.1.6 fl/,
2.1.7
? 44 2.1.8 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 o 95 Percent or less prior to and during testing of CROs should be prepared to start MO-P-1s if needed to control level ift turbine header pressure to opposite side. SV-1 and 2 ing turbine valve testing. (Computer point TAOM) be running during valve testing. Terminate test logic prevents CV testing when valve position limit ssure limit MSPL are in effect.
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 II 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.)