ML032170626
| ML032170626 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/19/2003 |
| From: | Gumbert R AmerGen Energy Co |
| To: | Conte R NRC/RGN-I/DRS/OSB |
| Conte R | |
| References | |
| 050289/OL-03-301 050289/OL-03-301 | |
| Download: ML032170626 (14) | |
Text
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AmerCen Energy Company, LLC Three Mile Island Unit i An Exelon/British Energy Company Route 441 South, P 0 Box 480 Middletown, PA 17057 203 !?WY 27 AW 11: I5 May 23,2003 U.S.
NRC Region I Administrator 475 Allendale Road King of Prussia, PA 19406
Subject:
Submittal of Senior Reactor Operator Final Examination Materials Three Mile Island Unit 1 (Docket # 50-289)
In accordance with NUREG 1021, Revision 8, Operating Licensing Examination Standards for Power Reactors, AmerGen is submitting the final Senior Reactor Operator examination materials for Three Mile Island Unit 1.
This submittal provides final examination documentation for the simulator and JPM examinations administered by the NRC on May 12, 13, and 14, as well as the written examination administered by AmerGen for the NRC on May 19, 2003.
Should you have any questions concerning this letter or the examination materials, please contact Dennis Bok at (717) 948-2006.
Res pectfu II y,
Exam Author Three Mile Island Unit 1
Enclosures:
(Sent directly to Joseph DAntonio, Chief Examiner, NRC Region I)
Written Examination:
Master Examination and Answer Key.
Examination Answer Sheets - originals and ungraded.
Exam Comment Sheet - describing proctor clarifications during the written Examinee Seating Arrangement.
Exam/Answer Key Adjustments, with supporting documentation.
AmerGen Written Exam Grading Results.
Written Examination Performance Analysis.
ES-403-1, Written Examination Grading Quality Checklist.
ES-201-3 Examination Security Agreement.
Day 1 and Day 2.
Administrative, Facility Walkthrough, and In-Plant.
exam.
Dynamic Simulator Scenarios JPMs:
Exelan,.
(QuestionlJPMlScenario, etc.)
TQ-AA-211-0108 Revision 0 Page 1 of 1 Concern or Problem Nuclear Q-018 (2-075 EXAM COMMENT SHEET DATE:
Mav 19,2003 SUBMITTED BY:
Dennis J. Boltz Examinee asked if intent of question was to identify the immediate automatic response.
Proctor noticed typographical error in answer A - RM-Ad should be RMA4.
Test Item Q-005 clarification of location of Breaker Trip Lights.
Recommended Resolution No change for NRC Exam.
Consideiediting this question prior to entering this question into the TMI exam bank.
No change for NRC Exam.
Consider editing the question prior to entering this question into the TMI exam bank.
No change for NRC Exam.
Correct typographical error prior to entering this question into the TMI exam bank.
Reference NIA.
NIA.
NA R O I W Remarks Clarification provided:
Breaker trip lights addressed in the question stem are located on outside front of the RPS Cabinets, at the top right corner.
Proctor response:
Yes, the intent of the question is to address the immediate automatic response.
Announced and noted the typographical error and the correction on the white board in front of the exam room.
Additional comments:
None Exam Analyzer comments:
None - Dennis J. Boltz Final Resolution:
Reviewed by:
Dennis J. Boltz Approved by:
Dennis J. Boltz Facility Author Facility Representative
May 19,2003 NRC SRO Initial License Written Exam Ex Comment Answer Kev identifies D as the correct answer, stating that Pressurizer system failure has occurred. Answer 5 can also be a credible answer if loss of SVMU occurs.
In order to select answer D, the examinee must assume the Pressurizer system failure. Given that the Loss of MU/SI procedure is included in 3 of the 4 answer choices, assuming loss of MUlSl has occurred is not unreasonable.
As written, there is no correct answer provided.
Based on the OS-24 definition for "Available as a Heat Sink," Rule 4, FW Control, does not apply.
Answer Key is not correct.
mlAnswer Key Adjustments Resolution Accept both B and D as correct answers.
With loss of RCP seal injection, seal
- 2 leak-off would provide a transport flow path to carry high activity RCS water to the RC Drain Tank, even in the absence of a Pressurizer system failure.
Even though this bank question was previously approved and used in the 2001 TMI NRC SRO written exam as written, it will be edited in the TMI question bank to preclude 2 correct answers for future use.
Delete this question IAW ES403 Section D.l.b, since there is no correct answer provided.
The intent of question was to have two Pressurizer level channels inoperable in excess of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, to require a Tech Spec shutdown. As written, LI-777 is still operable. Therefore there is no correct answer given. This question will be edited'prior to entry into the TMI question bank to represent conditions with both level channels inoperable.
Chanae correct answer from A to D.
Since 0524 states that a dry OTSG is not considered available as a heat sink, Rule 4, FW Control, does not apply to the conditions in the question stem. Therefore, Guide 13, Dry OTSG, applies. Edit this question prior to entry into TMI question bank: stem should state that both OTSGs are NOT dry.
Chanae Answer Kev from B to C.
This question was apparently edited, and then answers B and C were interchanged for psychometric considerations. This change brings the Answer Key into conformance with the NRC approved question and discriminant validity statements.
Reference RCS P&ID 302-650 shows flow path from Seal #2 leak off to the RC Drain Tank.
TS 3.5.5, Accident Monitoring Instrumentation, and bases, pages 34Oa, 3-40b, and 340c.
Page 340b states LT-1 and LT-3 are common to ONE channel, and that LT-777 is the other channel.
OS-24, Conduct of Operations During Abnormal and Emergency Events, Rev 7, Section 3.9.
Abnormal Transients Rules, Guides and Graphs, Rev. I - Rule 4, FW Control, and Guide 13, Dry OTSG.
Form ES-401-6 for Question #74, as approved by the NRC during week of April 7, 2003, Examiner prep week.
OP-TM-EOP-OI 0,
I
To assure operability o? key iTIstzum&Urn useful in diagnosing
. situations w h i c h could represent. or lead to inadequate core.cooling or evaluate2. and. predict. *e,.course of. eenidents-beyond the desfg, basis.
. 4-
4 --oy 2-3.5.5 ACCIDENT MONITORING INSTRUMENTATION (Continued)
The Emergency feedwater System ( E m is provided with two channels of flow instrumentation on each of the two discharge lines, Local flow indication is also available for the EFW System.
Although the pressurizer has multiple level indications, the separate indications are selectable via a switch for display on a single display. Pressurizer level, however, can also be determined via the patch panel and the computer log. In addition, a second channel of pressurizer level indication is available independent of the "1.
Although the instruments identified in Table 3.5-2 are signlficant in diagnosing situations which could lead to inadequate core cooling, loss of any one of the instruments in Table 3.5-2 would not prevent continued, safe, reactor operation. Therefore, operation is justified for up to 7 days (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for pressurizer level). Alternate indications are available for Saturation Margin Monitors using hand calculations, the PORVlSafety Valve position monitors using discharge line thermocouple and Reactor Coolant Drain Tank indications, and for EFW flow using Steam Generator level and EFW Pump discharge pressure. Pressurizer level has two channels, one channel from NNI (2 D/P instrument strings through a single indicator} and one channel independent of the NNI. Operation with the above pressurizer level channels out of service is permitted for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
Alternate indication would be available through the plant computer.
I The operabijity of design basis accident monitoring instrumentation as identified in Table 3-53, ensures that sufficient information is available on setected plant parameters to monitor and assess the variables following an accident. (This capability is consistent with the recommendations of Regulatory Guide 4 -97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," Rev. 3, May 1983.) These instruments will be maintained for that purpose.
3-4Qb Amendment No. -,
242
w r
N M
v) co c
0, 5
5
'5 a
340c Amendment No. w, 242
1 TMI -Unit 1 Operations Department Administrative Procedure Title Number OS-24 Revision No.
3.7 3.8 3.9 3.10 3.1 1 3.12 3.1 3 Conduct of Operations During Abnormal and Emergency Events LACK OF PRIMARY-TOSECONDARY HEAT TRANSFER (LOHT) is the inability of either OTSG to remove sensible heat from the RCS. LOHT can be confirmed if :
7 b
Neither OTSG has water level control and pressure control.
AND e
Core exit temperatures are rising MINIMIZE SCM: An intentional reduction of the reactor coolant pressure temperature relationship as close as practical to the 25°F subcooling margin or emergency RCP NPSH limit. (Recommended band 30-70°F)
OTSG AVAILABLE AS A HEAT SINK:
A physical condition where the OTSG demonstrates level and pressure control, used to detemine if primary to secondary heat transfer is possible. (Le. heat sink) Primary to secondary heat transfer need not be demonstrated to determine this availability. Primary to secondary leakage should not be considered a means of OTSG level control. A dry OTSG is not considered available as a heat sink.
OVERSIGHT The independent monitoring of plant and crew performance and anv subsequent intervention, as needed, to ensure the appropriate mitigation strategy is being pursued for the current plant conditions. Refer to Attachment B, SM Oversight Management Guidelines.
"PLANNED" REACTOR TRIP A scheduled shutdown, where a reactor trip, is directed by an approved procedure.
PRIMARY-TO-SECONDARY HEAT TRANSFER (PSHT) is the removal of sensible heat from the RCS to one or both OTSG(s). PSHT can be confirmed if:
e Either OTSG has water level control and pressure control.
AND b
RCS T, is approximately the same as secondary Tmt and responds to changes in OTSG pressure.
AND e
RCS forced or verified natural circulation is present.
RCP AVAILABLE -An available RC Pump is one which can be operated without extraordinary efforts. Pump service functions (motor cooling, seal cooling, etc.) are operable (redundancy not requited) and all interlocks can be satisfied. Strict compliance with administrative shutdown criteria (vibration, seal leakoff flow, etc) is not expected when the operation of the pump is more important to safe plant operation.
4
OP-TM-EOP-01 0 Revision 1 Page 8 of 49 ACTIONIEXPECTED RESPONSE Rule 4 Feedwater Control RESPONSE NOT OBTAINED 1
ACTIONiEXPECTED RESPONSE VERIFY SCM > 25°F.
RESPONSE NOT OBTAINED MAINTAIN OTSG level 75 - 85%
OPERATING Range Level.
If RCPs are OFF, then FEED OTSG at maximum available using E W,
within RCS Cooldown rate limit, VERIFY at least I RCP operating.
If EFW is not available, then FEED
> 1.O MIbm/hr using MFW.
MAINTAIN OTSG level 2 50% OPERATING Range Level.
MAlNTAlN OTSG level 2 25" STARTUP Range Level.
I B. If Level < minimum, then MAINTAIN the following MINIMUM required flow:
~
If SCM 25OF and OTSGs are available, then FEED > 215 gpd0TSG using EFW.
If EFW is not available, then FEED
> I
.O Mlbm/hr using MFW.
r ---
If SCM e 25OF and onlv one OTSG is available, then FEED > 430 GPM to' the good OTSG using EFW.
If EFW is not available, then FEED 1.0 Mlbmhr using MFW.
1-There is no minimum required flow rate.
J
~
OP-TM-EOP-010 Revision 1 Page 26 of 49 Guide 13 Dw OTSG I.
VERIFY OTSG SU Level c 6 and OTSG pressure at least 200 psi below Psat for Tc.
- 2. MONITOR Tube to Shell Differential Temperature (TSDT) and REVIEW Guide 14.
- 3. VERIFY the other OTSG is available.
- 4. VERIFY all RCPs are OFF or TSDT Limits are being challenged NOTE Automatic ERN actualion is not restricted by this guidance. Limit feedwater flow to the Dry OTSG until OTSG pressure has been restored. RCP operation is desired.
- 5. VERIFY the DRY OTSG pressure boundary is INTACT.
- 6. If TSDT tensile limit is being challenged,
- then, I
) If OTSG pressure boundary is not intact, then VERIFY an RCP is operating.
- 7. If TSDT compressive limit is being challenged, then,.
I)
If at least one RCP is ON, then FEED the DRY OTSG at a maximum of 435 GPM using EFW.
~ -.-- -__. - L-VERIFY the OTSG pressure boundary failure is in the Intermediate or Reactor Building.
If RCPs are OFF, then FEED the DRY OTSG at a maximum of 185 GPM using EFW.
If RCPs are OFF, then FEED the DRY OTSG at a maximum of 185 GPM using EFW.
KA# K1.07 Group #
- i Rating 2.7 2.9 Knowledge of the physical connections and/or cause-effect relationships between Liquid Radwaste System (LRS) and the following systems: Sources of liquid wastes for L 7 5
Plant conditions:
- Reactor power is loo%, with ICs in full automatic.
Based on these conditions, identify the ONE selection below that describes a NORMAL source of water to the Liquid Waste Disposal System.
A. PORV pilot valve leakoff.
B. Leakoff from between the reactor vessel flange O-Rings.
C. Intermittent drain flow from the Waste Gas Compressor Separator.
D. Valve packing leakage from Letdown isolation valves MU-V-I A and MU-V-1 B.
302-696, Waste Gas Compressors Flow Diagram, Rev. I.
@I New 0 TMI Bank TMI Question #
Parent Question ##
El Modified TMI Bank PI Memory or Fundamental Knowledge 0
Comprehension or Analysis 55.41.2t0.9 55.43 a 55.45.71.8 A Incorrect answer.
B Incorrect answer. This line is normally isolated.
C Correct answer. Level switch operates solenoid operated valve to drain excess water from the separator to D Incorrect answer. Packing leakoff lines are capped.
the Auxiliary Building Sump.
None.
TMI SRO Exam - May 2003 Friday, March 28,2003
TMI May 2003 SRO NRC Written Examination Examination Review Individual Remediation Examinee Individual and Group Performance Analysis Immediately following completion of the written examination by all the examinees, every test question was reviewed with all the examinees to obtain comments and feedback regarding question and answer key validity, and to accomplish immediate individual remediation for all incorrect examinee responses. Refer to the documentation provided that describes answer key adjustments with supporting justifications.
In addition, a written examination error analysis spreadsheet was developed to facilitate reviews of individual and group performance on the examination. All incorrect answers were evaluated independently by three SRO certified people with extensive training experience, knowledgeable in TMI systems, operating, administrative, abnormal and emergency procedures, and station operating license requirements.
The results of these reviews are as follows:
0 No trends were identified that suggest the existence of a common group weakness that requires training material upgrades or more emphasis in future programs.
No areas of individual weakness were identified that require remediation training in addition to that already conducted for individual examinees.
0 Please refer to the attached TMI Written Examination Error Analysis spreadsheet for a detailed presentation of all incorrect answers.
TMI Written Exam Error Analysis Resu Its B
CD TMI SRO Initial Exam May 19,2003 KA Description Knowledge of the operational implications of axial power imbalance as applicable to Inoperable/Stuck Control Rod.
Knowledge of the interrelations between Reactor Trip and the Reactor trip status panel.
Knowledge of the interrelations between Pressurizer Vapor Space Accident (Relief Valve D
BD C
Stuck Open) and Valves.
Knowledge of the operational implications of natural circulation cooling, including reflux A
D CC B
CCC D
boiling, as applied to Small Break LOCA.
Knowledge of the reasons for termination of startup following loss of IR Nls.
Knowledge of the reasons for length of time battery capacity is designed as applied to Station Blackout.
D -
C C
Ability to determine and interpret ARM panel displays.
Knowledge of the operational implications of RB pressure on leakrate as applied to loss of B -
A CC B
B -
containment integrity.
Ability to determine and interpret changes in Pressurizer level due to steam bubble transfe to the RCS during inadequate core cooling.
Knowledge of the operational implications of annunciators and condition indicating signals and remedial actions associated with a LOCA Cooldown.
Knowledge of the reasons for the normal, abnormal and emergency operating procedures Tomposite I AA CD BC A
M B D
DD D
procedures associated with EOP Rules.
Knowledge of the effect of a loss or malfunction of reactor trip breakers, including controls on the Control Rod Drive System.
Knowledge of the operational implications of brittle fracture as applied to the RCS.
Knowledge of the effect of a loss or malfunction of trip setpoint calculators on RPS.
Knowledge of Nuclear Instrumentation System design feature(s) andlor interlock(s) which provide for slow response time of SPNDs.
Knowledge of bus power supplies to the following: Containment Cooling Fans.
Ability to monitor automatic operation of the Spent Fuel Pool Cooling System.
Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling Equipment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:
dropped cask.
BCIassociated with natural circulation cooldown.
I Knowledge of the operational implications of normal, abnormal and emergency operating Page 1 of 2
TMI Written Exam Error Analysis C
CC TMI SRO Initial Exam May 19,2003 Ability to monitor automatic operation of the Condenser Air Removal system, including automatic diversion of exhaust.
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.
Knowledge of which events related to system operations/status should be reported to B
AD outside agencies: Containment System:
Knowledge of the process for controlling temporary changes. (Equipment Control).
C DD BB BC Page 2 of 2 Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity. (Equipment Control).
Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. (Radiation Control).
Knowledge of the process used track inoperable alarms. (Emergency ProcedureslPlan).
Knowledge of annunciator response procedures.(Emergency Procedures/Plan).