ML20155J235
| ML20155J235 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/11/1998 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | Bissett P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20155J226 | List: |
| References | |
| RTR-NUREG-1021 1920-98-20314, NUDOCS 9811120064 | |
| Download: ML20155J235 (430) | |
Text
. . - _ _ . . . . . .
a t
l . GPU Nuclear, Inc.
-( Route 441 south g(Jggg Post Off ce Box 480 Middletown. PA 17057 0480 Tei 717 s44 7s21 June 11, 1998 1920-98-20314 Paul H. Bissett U. S. Nuclear Regulatory Commission
.475 Allendale Road King of Prussia, PA 19406 Gentlemen:
Subject:
Three Mile Island Nuclear Station, Unit I (TMI-1)
Operating License No. DPR-50
. Docket No. 50-289 Initial Operator Licensing Examination Outline for Three Mile Island Unit 1 The purpose of this letter is to submit the Initial Operator Licensing Examination Outline for Three Mile Island Unit 1 pursuant to the guidance ofNUREG 102L Examination Standard 201.
Examination materials shall be withheld from public disclosure until after the examinations are completed.
If you have any questions regarding this submittal, please contact William Heysek at the TMI-l Nuclear Safety & Licensing Department at 717 948-8191.
Sincerely, 1F)V. N. 4 James W. Langenbach' ,,
! Vice President and Director, TM1 l WGH /
j . Attachment cc w/o att: Regional Administrator, Region I- Hubert J. Miller
' TMI Senior Resident Inspector - Wayne L. Schmidt File 98003
^
9811120064 981104 PDR ADOCK 05000289
$l' V PDR l v
, _ _ _ . _ _ _ _ _ _ _ _ _ m _ . _ . _ _ . . _ _ . . . . _ . _ . _ _ _ _ _ . _ _ . _ . _ . _ _ _ . _ _ . _ . _ _ _ _ . _ . _ _
ES-401 PWR SRO Examination Outline - Form ES-401-3
}. W ,
'y y;_ -
,. s:
Facility: . , ,
Date of.. Exam: '
Exam Levelhd$bhMW.m ,eg Q d@- .
a;;,;.
K/A Category Points Tier Group Point K K K K K K A A A A G Total 1 2 3 4 5 6 1 2- 3 4
- 1. 1 2 2 6 - - -
7 7 - -
24 Emergency 2 2 4 3 - - -
2 5 - -
16
& Abnormal Plant 3 1 0 2 - - -
0 0 - -
3 Evolutions Tier 5 6 11 - - -
9 12 43 Totals 1 19 2 2 1 3 2 1. 0 4 2 2 Plant 2 4 0 0 17 2 1 2 1 3 0 4 Systemi; ;
4 1 0 0 1 0 0 1 1 00 l Tier -7 2 1 6 3 3 2 8 2 6 40 Totals l
- 3. Generic Knowledge and Cat Cat Cat Cat Abilities 1 2 3 4 17 i 5 3 4 5 )
l Note:
- Attempt to distribute topics among all K/A i categories; select at least one topic from I every K/A category within each tier. I
- Actual point totals must match those specified in the tabl e .
- Select topics from many systems; avoid selecting more than two or three K/A topics from a given system
- 2. c l
1
e ES-401 PWR SRO EmominoHon ounineForm ES-401-3 Emergency and Abnormal Plant EvoeuMons - Tier 1/ Group 1 ,
E/ APE # / Nome / Sofety Function K1 K2 K3 Al A2 G 1(/A Topicts) lim. Points 000001 Connnuous Rod WHhdrawal/I 06 Actions for UncontroNed Rod Mollon 2.9 1.0 _
000003 Dropped Control Rod /I 03 Plant Response to Dropped Rods 3.6 1.0 '
000005 Inoperable!$ fuck Control Rod / I 05 Power Operonon with Minolloned Rod 4.2 1.0
.=
000011 Large Break LOCA / lit M Lono Term Coonna of Core 4.2 1A 000015/17 RCP Molfunctions / N 22 RCP Seal Failure ' d.2 1'.0 000015/17 RCP Monunctions / N 03 Actions for idah RCP VibroNons -
40 1.0 BWM09; CE/A13: W/E09&E10 Natural Cire. I N 02 R4$6tYeortsh p b d Edf#n MV d ral Q g f M ak h yM 4.0 1A
~
000024 Emergency Boration /1 01 Required actions when Rods FW to insert on itb '
4.4 - 1 A -
000026 Loss of Component Cooung Water / Vill 03 Loss of CCW to RCP 's a 4.2 "1.0 4 000029 Anticipated Transient w/o Scram / I 12 Required Actions for ATWS 4.7 1.0 >
j 7WU40 (BW/E05; CE/E05; W/El 2) Steam Line
- - Rupture - Excessive Heat ironster / N e
04 Porometers tbr MonMoring Stoom Line Rupture 4.7 - 1.0 ,
CE/Al l: W/EOS RCS 0%wm .y - PTS / N ;tM -
000051 Loss of Condenser Vocuum / N 02 Requirements for turbine Trip on toes of condoneer Vocuum 4.1 1.0 k 4N Desion Opero4cn of $80 Diesef n
000065 Station Blockout / VI 02
^,u 1.0 w[!h n ;
" ~
3A %, ..1.0 ;
000067 Loss of Vilal AC Elec. Inst. Bus /VI 05 Olockoot Aftsets on vilal instrument Bus ~ ' '
i 000059 Accidental Liquid RodWaste Rei. / DC 06 AUTO Actions on Hi@ Liquid Reloose Rod Monbor Alarm #
h.h 1.0 k 2 f 000062 Loss of Nuclear Service Water / N 01 Requirements for Trippino RCPs on Loss of teCCW 3.1 1.0 N 000067 Plant Fire On-sNe / IX 01 Fire ClasstAcoMon -
3.9 1.0 000068 (8W/A06) Control Room Evoc. / VIlt 18 Required Action for Control Room Evocuation 4.5 1.0 ,
000069 (WA141 Loss of CTMT Inteartty / V 03 01 Containment isoloNon RCP Restort Criterio 5'
'. 4.3 4.9 -
1.0 1.0 000074 (W/E06&E07]Inod. Core Coounc / N i
.: i BWK03 Inadequate SubcooNno Morain / N 01 Reason for Securina RCPs V 3.8 1.0
- f'- ,
BWK03 Inadequale SubcooNng Morain / N 02 Immediate Actions forloss of Sei~x.A;.T. 4.0 1.0 ;
000076 Hlah Reactor Coolant ActtvMy / DC 02 ActMiy Limit Requirina Plant Shuidown 3.4 1.0 BW/M)2&A03 toss of fes-X/y/VR 1 Plant Response to Pressure Pressure Channel Fabure 4.0 1.0 .
1 K/A Co'agory Totals: 2 2 6 7 7 Group Point Total: 24 i
t ,
t
ES-401 PWR SRO E-.J,ars OuttineForm ES-401-3 Emergency and Abnormal Plant Evolutions - Tier 1/ Group 2 ,
E/ APE # / Nome / Safety Function K1 K2 K3 Al A2 G KfA'Topicts) Imp. Points 000007 (BW/E02&ElO; CE/E02) Reactor Trip -
Stabiltration - Recovery / I BW/A01 Plant Runbock / I 01 Cause for Runbock of itie Unit load Demand Subsystem 3.5 1.0 BW/A04 Turbine Trip / N 02 Reason for Vertfying Turbine Trto in EOPs 3.5 1.0 000008 Pressurizer Vapor Space Accident / m 01 PORY Actuation 4.0 1.0 000009 Smott Break LOCA / m 23 RCP Trip Criteria During LOCA 4.3 1.0 BW!E08; W/E03 LOCA Cooldown - Depress. / N 2 Requirement for Cooldown <40*F 4.0 1.0 W!Eli Loss of Emergency Coolont Recirc. / N 000022 Loss of Reoctor Coolont Makeup / II 01 b 4k Oi-w for Leck In Makeup Line 3.8 1.0 000025 Loss of RHR System / N 03 DH Vortex Operoflon 4.1 1.0 000027 Pressurizer Pressure Control System Moffunctkm / m 000032 Loss of Source Range NI / Vil 000033 Loss of intermedlote Ranga NI / Vil 10 Loss of Both Intermediate Range Detectors 3.8 1.0 Reason for Minimizing Subcooiing Margin During Steam Generator Tube 000037 Steam Generator Tube Leck / m 02 Rupture 3,9 1.0 000038 Steam Generator tube Rupture / m 01 Reason for Isolation Criterio for SGTR
- 4.7 1.0 000054 [CE/E061 toes of Moln Feedwater / N 05 Affects of Spurious CRD Trip Slanoi to ICS 3.7 1.0 05 Inadequate Heat ironster - Loss of . 02 Reason for Ropid Depressurtzotion During inodoquoto Core Cooling 4.4 1.0 S Heat Str* / N 000058 Loss of DC Power / VI O2 immediate Actions for loss of DC Power 4.2 1.0 ~
000060 Accidentot Goseous Rodwoste Rei / IX 05 Auto Ac*lons to Prevent Release to Pubtle
- 4.2 1.0 000061 ARM System Alarms / VII 01 MAP-5 lodine Sampler Operation 3.6 1.0 W/E16 High Containment Rodlotion / IX 000065 Loss of instrument Air / Vm 05 Operctor Actions for During loss ofIA 3.3 1.0 CE/E09 Functional Recovery M
K!A Co'egory Point Totots: 2 4 3 2 5 Group Point fotot: 16 e
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ES-401 G3neric Knowledge and AbilitiGS Outline (Tier,, .3) Form ES-401-5 s.n , - . ~ .
Facility: M Date of Exam:
l Exam Level:
"" Points CateQory K/A # TODiC' ImD.
2.1.3 Review of Logs and Rhc'ords fot*, shift 2 . . 3.4' 1.0 .2-Turnover
+n' 2.1.12' core Flood Tank Technical specification 4.0 1.0 conduct of 2.1.12 contain ent coois w recanicai soec. l -4.0 1.0 ,;
operations 2.1.29 E Non "7 YA- > 3.F % 150 A i f. :,'4 2.1.4' shift Stannino Recuirements 3.4 1.0 2.1. . .
v 2.2.11 teodification to V&lve Lineups 3.4 1.0 l
gng conaition or usue iaggeo
, 2.2.21 In-service Testino Requirements 3.5 1.0 n o 2.2.
2.2.
2.2.
Total 2.3.1 10CFR20 Limits 3.0 1.0
. 2.3.4 Requirements to exceed 2000 mrem /Qtr 3.1 1.0 t U 2.3.2 Implestation of ALARA for work Practices 2.9 1.0 o I 2.3.10 Postino Requirements for Radiation Areas 3.3 1.0 l
I 2.3.
2.3.
Total 2.4.37 g g 9tgo g or deviation from 3.5 1.0 Emerg ncy 2.4.42 Purpose of Tsc 3.7 1.0 PrQCe ures em r cu9dggncys eian erotection action 2.4.44 in 4.0 1.0 and P an tvent erioritization auring Leergency l 2.4.6 Situations 4.0 1.0 2.4.38 NkfNa N 4.0 1.0 2.4.
Total Tier 1 Target Point Total (SRO) 17 i
, i l
l
l l
Nf% Wel(Acc 3 Revisit, if can't do anything with it, let it oo. Let Paul know.
12 D. Add
- start second Makeup Pump and . .'
20 Insert ' Plant" at beginning of question stem.
28 Typo in DC-V 2A. Also look at Answer D - anything better?
33 insett into stem that valve is in a high radiatum area. Ensure all answers are still okay.
37 Delete Bullet #3 of the question stem.
l 40 Typo in D (ardno) 48 C. Change to "Stop 3 RCPs.* )
49 SELECT ONE. Review answers to avoid ability to drde 2 half-answers. i 54 a. Typo "cannot."
56 Plausibility statement D typo (does - dose).
60 Which one statement " describes the comparison between .. "
62 C and D Remove (only)(twice).
l 63 Delete Bullet 5 in the question stem.
64 Typo in second line of question stem - termination 'of' the..
65 Change Trip to " Trips
- 4 times.
74 8 Chance ' required
- to "needed.*
76 Delete LEVEL CUE ' based upon above conditions."
78 Chance critical to criticality in Bullet #1.
89 insert " position
- and 'open." " Typo *cammand.'
91 Typo in A - WDL-V-535. C Change to WDG-V-3/4.
92 Delete last sentence in question stem. A. Fix location of 'A."
93 3" Bullet HAND POWER (DH AUTO SUBFEED) has occurred.'
94 3" Bullet .. actuations, a trip of .
96 Flip flop B and D 97 Delete 4 Initiate HPI's. Rai e needs to be lower case (twice).
. . 1 Administrative Exam Section
, Provide applicant with a completed ECP prepared by the STA. Have them
. complete one, compare results to identify / analyze cause for differences between the two calculations.
l Set-up preparation: ,
- Copy of latest procedure.
. Clean calculation forms / graphs from procedure for use by the applicants. 1
. JPM needs to identify STA ECP problems. ;
i '
i A.1.2 l Admin JPM (Surveillance Documentation Review) l Provide applicant with completed surveillance. Applicant must perform test documentation review, identify problems and demonstrate performance. corrective actions. Applicant must decide whether ;
component / system can be considered operable or whether test must be re-run.
Possible documentation problems to be included in prop:
e Readings outside o_f tolerance band for acceptance.
. Missing initials / dates / signatures.
. Missing readings. ,
e Calculation errors.
. Graph reading errors.
. First set of readings taken prior to starting componont.
Termination point would be at determination of whethor surveillance must be performed again.
Set-up preparation:
. Dummy completed surveillance with problems.
- JPM must identify problems, corrective actions, etc.
C:\DJB\ ADMIN \ Admin Exam Sectiortdoc Last printed 08/13/98 8:42 AM
.~ -
Administrative Exam Section
'A.2 ;
Admin JPM (Perform Reauired actions for Valve Out-of-Position) i "I am an AO, you are SS. I just found valve in a throttled position. I was peiforming a system tagout to support maintenance activities on
. The switching order required me to close valve . This !
valve was originally in mid-position, but is required to be fully open." l Applicant is required to prform SS actions in response to report of valve out of required position.
- Evaluate conditions and determine actual required position.
. Direct action to return valve to required position. (Critical Step).
- Evaluate system operability during time valve was out of required position l
to determine reporting requirements, if any.
. Determine if operability test is required at this time. I e Investigate cause for occurrence. .
. Perform required Log entry.
e identify action to prevent recurrence, e Notify Duty Supt. If deemed deliberate. l e Fill out a CAP Form for critical components.
JPM information:
Verbal Cue - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> earlier AO fell /was injured while performing valve re-alignment from surveillance testing. He was transported off-site for medical attention. His paperwork was signed off as having fully re-opened the valve.
Point of Termination:
When CAP Form is called up on Computer.
A.3 Admin JPM (RWP) (Integrate with IC-V-4 Manual Operation?)
e Sign-in on RWP.
- Discuss RWP requirements - dosimetry, protective clothing, exposure limits, etc.
. Explain Survey Map Data.
. Identify component location.
. RCA exit procedures are evaluated.
. Sign off RWP.
No pre-scripted questions are needed for this JPM.
JPM will guide examiner to discuss items bulleted above. ,
i C:\DJB\ ADMIN \ Admin Exam Section. doc Last printed 08/13/98 8:42 AM !
1 i
. l Administrative Exam Section A.4 Take credit for E-Plan Event Classifications at end of simulator scenarios.
. identification of specific events classifications must be added to end of each i scenario. !
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2 C:\DJB\ ADMIN \ Admin Exam Section doc Last printed 08/13/98 8:42 AM
l Simulator JPM Review Notes
'#1 l
JPM #1 (.013) e Replace both 2 questions - there is overlap with written /sim scenario.
Prepare 4 copies of OP 1105-9 Section 3.2.3 for use by applicants.
' y , e .
Prepare 4 copies of the ouestions to be handed to the applicants.
Prepare Examinee Preview Sheets to be handed to applicants.
l
" diagnosis" as well as " task performance." Paul Bissett will check with Conte. not back to us on this issue, e Re-write this so applicant is Sec RO. " Directing activities" is not allowed, surrogate actions (errors) could negate evaluation if applicant performance, e Add note regarding local operator actions required to clear annunciator L-1-7 (Gen Stator Liquid Cooling Trouble Alarm.)
e Replace Q#1 with Turbine Loading Rate Limit question based upon OP 1106-1 Figure B-5, Step 3.4.10 plus the note at this section. ;
e Replace Q#2 with Generator Load Limit question, using Hydrogen pressure, l MVAR values to be provided in the question.
. Q#1 is okay.
- Q#2 is okay. !
i Mayhue (Instant) cannot be in simulator scenario with Pressurizer Level Transmitter Failure.
C:\DJBUPMs\Sim JPM Review Notes. doc Last printed 08/13/98 9:18 AM
f-I Simulator JPM Review Notes i
L -g4 Perform Emergency Boration JPM e JPM is okay.
. Replace Q#1 with question requiring use of graph in Figure 3 of OP 1103-4 to correlate gallons to stroke setting, number of strokes required.
'. Replace Q#2 with question giving MU Tank level / pressure conditions (20#,
- 10") which place conditions inside the lower left area of the curve. Applicant i
is asked if it is okay to operate here, if not - what problems could occur?
- 5 Establish Long Term Cooling JPM l
. Re-write the JPM - no surrogate operators allowed.
L e Questions were not reviewed in the simulator.
- 6 l Energize 1C 4KV,1J 480V Bus Using SBO Diesel JPM e Tried to lower auto voltage regulator setpoint (EGR24) to require operator to l go to manual or to have remote setpoint increased in order to get another critical task. Status of response with auto voltage regulator setpoint reduced
- to zero? Should breaker be able to be closed? Is any action reautred? ,
. Questions were not reviewed in the simulator. !
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- e Questions were not reviewed in the simulator.
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I C:\DJBVPMs\Sim JPM Review Notes. doc Last printed 08/13/98 9:18 AM 1
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. TMl LicIn:Ing Ex2 min tien Answer Key QUESTION: 001 (1.00)
. Selected Pressurizer level signal indicates off-scale low. An evaluation must be conducted '
to determine if the transmitter has failed low or if it is ac "'--'~daa a low oressurizer level condition. Which ONE (1) parameter can be 'use determination?
A. . Pressunzer spray valve position (
B. RCS pressure .
C. Makeup tank level D. Pressurizerwater temperature Answer. B j l
- K/A
- 000028 AA2.01 3.6 Page 4.2-23 l
l Objective: IV.D.11.03 i
Reference (s): EP 1202-29, Pressurizer System Failures, Rev. 52, Page 12 History: New NRC Cognitive Level Rating 3 Measurement #1 Ability to evaluate, analyze, operate, and monitor pressurizer level indicators during pressurizer level control malfunctions. ;
i A Discriminant Validity Plausible distracter since increasing pressure would open the spray valve (in Auto)ifleft unattended.
B Discriminant Validity Cormet answer.
C Discriminant Validity Plausible distracter since a change in Makeup Tank level is expected for both a level sensor failure and an actual change in pressurizer level..
D Discriminant Validity Plausible since pressurtzer water temperature fluduates with in-surgas and out-surDes.
N
TMI Licensing Exrminrti:n An';wer Kcy QUESTlON: 002 (1.00) .
Current plant conditions are:
- The reactor is tripped '
- RCS subcooled margin is zero.
1 Which ONE (1) action results in increasing RCS =L*=3 ing margin? 1 l I
A. Decrease RCS pressurizerlevel l
B.- Decrease RCS hot leg flow C. _ increase RCS loop pressure D. Increase RCS hot leg temperature l 2 Answer: C K/A: BW/E03 EK3.1 3.8 Page 4.3-5 Objective: V.E.02.02 l Reference (s): Steam Tables l
l . History- New -
l
- NRC Cognitive Level Rating 2 l Measurement #2 Demonstration of understanding of operationalimplications of remedial i actions related to inadequate core cooling.
l l A Discriminant Validity Plausible misconception since indications of increasing pressurizer level '
during these conditions may be an indication of increasing void formation.
8 Discriminant Validity Plausible distracter since this is the opposite of another correct answer.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter since this is the opposite of another correct answer related to saturated coolant conditions.
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TMI Lic:nxing Eximinatien An:wir Key QUESTION: 003 (1.00)
Reactor power is at 100% when the controlling RCS pressure channel, RC3A-PT1, instantaneously fails LOW, With NO operator action, which ONE (1) statement describes the SHORT-TERM plant response?
A. Reactor trip occurs on high RCS pressure B. SASS shifts control to RC3B-PT1 RCS pressure channel to stabilize the plant C. Pressunzer heaters energize from RC38-PT1 to maintain pressure in normal band with the spray valve open D. Reactor trips and Safety injection is actuated on low RCS pressure Answer. B K/A: BW/A02 & A03 AK2.1 4.0 Pages 4.3-28 & 4.3-31 Objective: IV.E.09.02 Reference (s): OPM F-05 Non-Nuclear Instrumentation System, Rev 11, Page 9 l
History: New l
NRC Cognitive Level Rating 3 Measurement #3 Comprehension of interrelationships between loss of NNis and Components, l fundions of control and safety systems, failure modes. '
A Discriminant Validity Plausible misconception that failure of one channel (high) will cause automatic RPS trip. This is incorrect since only one RPS channel will trip.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible distracter since heaters would, in fact control normally after SASS aduation to the unselected channel, however the spray valve would not be open since its control would also transfer to the unaffected channel.
D Discriminant Validity Plausible distracter since this is the correct answer if SASS failed to transfer to the unselected channel.
TMl Lirn:ing Exrmin tian An;wer K y QUESTION: 004 (1.00)
The reactor is super critical and just entering the intermediate range (4000 cps an the Source Range) when detector compensating voltage to Ni-3 is lost. Which ONE (1) statament explains the effect this loss of compensating voltage will have on NI-3 indication?
L A. - NI-3 would be unaffected at this low power level.
B. NI-3 would indicate higher than NI-4.
C. NI-3 would come on scale some time after NI-4.
D. NI-3 would go off scale low before NI-4 if a reactor trip occurred while at 1E-8 amps.
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K/A- 000015 K5.02 2.9 Page 3.7-6
. Objective: IV.E.11.09 Reference (s): OPM F-04 Nuclear instrumentation System, Rev 9, Figure 4, Page 22
- History- TMI Exam Bank Question SR4E11-09-Q05 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 3 Measurement M Demonstration of knowledge regarding effects of compensation operation of Nuclear Instrumentation.
A Discriminant Vahdety Plausible misconception since this is a correct response for higher reactor power conditions.
8 Discriminant validity Correct answer.
C Discriminant Validity This is a plausible distracter since this is a possible correct effect for overcompensation rather than undercompensation..
D Discriminant Validity This is a plausible distracter since it describes the effects of overcompensation during a reactor trip situation.
I
TMI Lirnring Exrmin tien Answer Ksy QUESTION: 005 (1.00)
The plant is at 100% power when control room indications reveal the following:
Reactor power is DECREASING RCS pressure is INCREASING
- Main Steam safety valves are OPEN
- MS-V-3s and MS-V-4s are OPEN
- Indicating lights on Panel SS-1 are GREEN for the breakers for the Middletown 1092, Jackson 1051 and 500 kV tie lines
- - Indicating lights on Panel SS-1 for the Middletown 1091 breaker switches are GREEN BDd YELLOW Indicating lights for both main generator breakers are RED
- Main generator electrical megawatts are 56 MW Which ONE (1) event is described by these symptoms?
A. Loss of 230 kV substation DC B. Load rejection C. Auxiliary transformer fault D. Loss of station power Answer: B K/A: BW/A01 AK2.1 3.5 Page 4.3-26 Objective: V.C.01.01 Reference (s): Abnormal Procedure 1203-1, Load Rejection, Page 2.0, Rev 26 History: TMI Exam Bank Question QR5C01-010-001 Verified not used on audit exam or in SRO Program quizzes. i NRC Cognitive Level Rating 3 Measurement #6 Knowledge of interrelationships between plant runback conditions, automatic component operation and fundions of controls and safety systems.
A Discriminant Validity Plausible since loss of DC would could possibly cause protective relay operation in the switchyard.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible since this would cause protective relay operation to occur in the switchyard to isolate the auxiliary transformer from its 230 kV bus.
D Discriminant Validity Plausible misconception between loss of offsite power (LOOP), loss of station power, and separation from the grid.
TMI Lic:nsing Excmin:tian !
An:wcr K y '
QUESTION: 006 (1.00) j Which ONE (1) statement describes the requirements for an individual to be allowed to receive a TEDE dose greater than 4000 mrem per year, excluding a planned special exposure?
. A.- A special RWP is written covering the individual to be permitted to exceed 4000 ;
mrem '
B. Approval from RadCon/ Safety Director and Site Director C. Approval from the President, GPU Nuclear D. Notification of the NRC 1
r Answer: B
! K/A: Generic 2.3.4 3.1 Page 2-9 3
Objective: lil.F.02.01 i GET Radiation Worker Training Handout Rev 5, Page 14 Reference (s):
History: -New NRC Cognitive Level Rating 1
- - Measurement #6 Knowledge of radiation exposure limits, including permissible levels in excess of those authorized.
A Discnmenant Validsty Plausible misconception that a special RWP would permit additional
^
exposure.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since this individual is required to pre-approve 3- exposures in excess of 4500 mrem.
D Discriminant Validity Plausible misconception that the NRC could authorize additional exposure.
TMl Lse:n:Ing ExEminstien An:wcr KSy QUESTION: 007 (1.00)
Given a Switching and Tagging Request to remove a fire system heat detector in the EG-Y-1 A' diesel room from service, identify ONE (1) action is required to compensate for this detector being removed from service.
A. Establish a fire watch PATROL within one hour to inspect the diesel room at least ONCE PER HOUR.
B. Station a CONTINUOUS fire WATCH in EG-Y-1 A diesel room WITHIN ONE HOUR.
C. START EG.Y-1 A to perform the one-hour surveillance to verify OPERABILITY.
D Restore detector to operable status WITHIN 14 DAYS or commence plant shutdown to hot shutdown.
Answer: A K/A: 000006 A2.01 3.1 Page 3.8-24 Objective: V.A.15.05 "Given Exhibit 2 of AP 1038 and the status of fire protection system, determine what action, if any, is required."
Reference (s): AP 1038 Administrative Controls - Fire Protection Program Rev 45, Pages E2-1 and E2-3.
History: New l NRC Cognitive Level Rating 3 Measurement #7 Use of procedures to mitigate consequences of shutdown (in this case, l planned removal from service) of fire protection system equipment i A Discriminant Validity Correct answer.
B Discriminant Validity Plausible distracter - but the Fire Watch is not required to be continuous.
C Discriminant Validity Plausible distracter since EG-Y 1 A operability is still contingent upon establishment of fire watch patrol.
D Discriminant Validity Plausible distracter since there is a 14-day time clock for retum to service or reportability to NRC it required.
TMI Llicn:ing Exrminr_ti:n l An:w:r Kcy l
QUESTION: 008 (1.00)
Which ONE (1) statement describes the purpose of the protective action guidelines?
A. Protect plant workers from receiving excessive radiation exposures in excess of 10CFR20 limits B. Prevent radioactive releases from exceeding 10CFR20 limits C. Recommend sheltering or evacuation for the general population D. Determine if potassium iodide tablets should be administered to reduce thyroid
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Answer: C K/A: Generic K2.4.44 4.0 Page 2-15 Objective: V.H.01.1 Reference (s): EPIP-TMI .02, Emergency Direction, Rev 11, Page E8-1 History: TMI Exam Bank Question SO5H01-06-QO2 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #8 Demonstration of Emergency Plan Protective Action Recommendations as applied by accident scenario Emergency Diredor.
A Discriminant Validity Plausible misconception since this is a valid function of the Radiation Protection Plan rather than the Emergency Plan.
B Discriminant Validity Plausible misconception since the Emergency Plan doesnt give guidelines for releases but provides recommended actions with regards to the public health and safety if releases are occurring.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter since decision to administer Ki is a potential action during severe accident conditions.
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TMI Llacn:ing Examinatian
, Answ r Ksy QUESTION: 009 (1.00)
Wdh the plant operating at 50% power, BOTH intermediate Range Nl detectors fail LOW.
Which ONE (1) statement describes the required action (s).
A. Continue power operations but limit power to 50%.
t l B. Continue power operations, power may be increased.
1 C. Immediately take action to place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Take action within one hour to restore at least one Intermediate Range channel to operable status or place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l Answer: B K/A: 000033 AA2.10 3.8 Page 4.2-27 l Objective: IV.E.11.15 l
Reference (s): Technical Specification 3.5.1 Amendment 189, Pages 3-29,3-30 l History: New l
l NRC Cognitive Level Rating 3 l
Measurement #9 Ability to determine and interpret Tec Spec limits as they apply to the loss of intermediate range instrumentation and facility requirements A Discriminant Validity Plausible misconception since power operations may be continued, however, power levelis not restrided.
8 Discriminant Validity Correct answer.
C Discriminant validity Plausible misconception because it is an incorred Tec Spec action. There is no Tec Spec action above 10% FP for the intermediate range deledors.
, D Discriminant Validity Plausible because it is a correct Tec Spec action if power was less than 10%.
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TMI Lirnsing Eximin' tion Answer K:y.
QUESTION: 010 (1.00) 10 CFR 50.54 (x) specdically allows " reasonable action that departs from a license condition ,
or a technical specification in an emergency when this action is immediately needed to i L protect the public health and safety." Select the MINIMUM position that may approve 10 ,
l' - CFR 50.54 (x) actions. !
A. Director Operations & Maintenance L B. - Plant Operations Director C. Licensed Senior Reactor Operator D. Licensed Reactor Operator Answer: C i L K/A: Generic K2.4.37 3.5 Page 2-15 l' Objective: V.H.01.18 h
Reference (s): EPlP -TMI .02, Rev 11 Page 3.0 History: TMI Exam Bank Question SOSH01-18-QO1 L Verified not used on audit exam or in SRO Program quizzes.
I NRC Cognitive Level Rating 1 l
Measurement #10 Knowledge of the lines of authority during an emergency in accordance with the EPIP and 10CFR50 requirements.
A Discriminant Vahdsty Plausible misconception as the Operations & Maintenance Diredoris a senior management ofrdal.
l B Discriminant Validity Plausible misconception as the Plant Operations Diredor is a senior mansgement ofrdal.
- C Discriminant Validity Correct answer.
l D Discriminant Validity Plausible misconception since the Reactor Operator is licensed by the i
Nuclear Regulatory Commission.
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TMI Lirnsing Examinition An;wer K y
' QUESTION: 011 (1.00) in addition to the person having the clearance, which ONE (1) of the following must grant l permission to change the condition of BLUE tagged ES equipment? i i
A. Director Operations and Maintenance
' B. Plant Operations Director C. Duty Shift Supervisor / Shift Foreman D. Licensed Control Room Operator I
Answer:- C K/A: Generic K2.2.14 - 3.0 Page 2-6 Objective: . V. A.01.02 .
Reference (s): Administrative Procedure,1002, Rev 82, Page 7.0 History: TMI Exam Bank Question.AL5A01-02-Q04 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 j Measurement #11 Knowledge of the process of making contguration changes in regards to the TMI Switching and Tagging procedure.
A Discriminant Vahdity A plausible misconception as this is a senior management omcial.
8 Discriminant Validity A plausible misconception as this is a senior management omcial.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception since this individual is licensed by the NRC.
TMI Lic:nsing Ex min ti:n An:w:r K;y QUESTION: 012 (1.00) '
Following a reactor trip, the following conditions exist:
- A OTSG level is 87" and decreasing slowly
- B OTSG level is 82' and decreasing slowly
- MFWflowis O gpm
- MFW valve D/P is O psig
- RCS pressure is 1725 psig and stable
- MUT levelis 62 inches
- PZR levelis 35 inches
- MU flowis 60 ppm Which ONE (1) statement describes the action required per AP 1210-1, Reactor Trip?
A increase MFW pump speed B. Open MU-V-14A or MU-V-148 as necessary C. Initiate HPI D. Open MU-V-217 Answer: D BW/E13 K/A: EK1.2 3.6 Page 4,3-21 Objective: V.E.01.02 Reference (s): Abnormal Transient Procedure,1210-1, Reactor Trip, Rev 37 Page 3.
History: TMI Exam Bank Question QR5E01-03-005 Verified not used on audit exam.
Used in SRO Program 3/12/98 Quiz.
NRC Cognitive Level Rating 3 Measurement #12 Demonstration of ability to interpret and relate operationalimplications, abnormal operating procedures, accident mitigation strategies and rules.
A Discriminant Validity Plausible distracter since question stem establishes zero flow and valve delta-P.
8 Discriminant Validity Plausible distracter since this is a valid action contained within the immediate actions of ATP 12101.
C Discriminant Validity Plausible distracter since this action is required under conditions other than specified in the question stem.
D Discriminant Validity Correct answer.
mm ___-._ _ _ _
TMl Lir:nsing Excmination An:wer K';y QUESTION: 013 (1.00)
Current plant conditions are:
- Reactor is operating at 80% power.
- RM-L-1 (RC Letdown) has increased to the ALERT setpoint.
- Chemistry analysis indicates dose equivalent 1-131 concentration is 0.28 uci/gm.
- RCS specific activity is 220 uci/gm.
- E-BAR is 0.5.
Which ONE (1) statement describes the required Tech Spec actions?
A. Reduce RCS activity to less than the Tech Spec limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. Reduce RCS activity to less than Tech Spec limit within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
C. Initiate actions to place the unit in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Reduce RCS cctivity lo less than Tech Spec limit within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or place the reactorin cold shutdown.
Answer: C K/A: 000073 AA2.02 3.4 Page 4.2-59 Objective: V.D.06.02 Reference (s): Technical Specification 3.1.4.1, Amendment 204, Page 3-8 History: New NRC Cognitive Level Rating 3 Measurement M3 Ability to interpret conditions and direct actions required for high concentrations of fission product activity in the reactor coolant, in accordance with plant technical specifications and facility requirements.
A Discriminant Validity Plausible distracter since this is a valid technical specification action - but for a different set of plant conditions.
B l>6scriminant Validity Pieusible distracter since this is a valid technical specirmation action - but for a different set of plant conditions. ,.
C Discriminant Validity Correct answer. ,
D Discriminant Validity Plausible distracter sinrd this is a modification of a valid technical specification adion - but for a different set of plant conditions.
TMI Lic:nsing Extminctlen Answer Ksy QUESTION: 014 (1.00)
Current plant conditions are:
- Reactor is operating at 100% power.
RCS Tavg is constant at 579'F.
Make up tank level is decreasing slowly - MU-V-17 is in MANUAL control.
Letdown flow has been constant at 45 ev.s.
RCP total seal injection flow is 38 gpna (normal) - MU-V-32 is in AUTO.
- RCP labyrinth seal D/P indicators show low off-scale (negative).
- Auxiliary Building airborne activity is increasing.
Which ONE (1) statement describes the cause for the abnormal conditions?
A. RCP seal #1 leak-off flow is aligned to the Auxiliary Building sump.
B. RCP total seal injection flow transmitter has failed.
C. RCP seal injection flow is not reaching the RCPs.
D. RCP seal #1 leak-off flow has been isolated by closure of MU-V-26.
Answer: C K/A: 000003 A3.10 3.2 Page 3.4-9 Objective: IV.A.05.04 Reference (s): Abnormal Procedure 1203-15, Loss of R.C. Makeup / Seal Injection, Rev 22, Page 2 History: New NRC Cognitive Level Rating 3 Measurement #14 C ion of ability to monitor and evaluate proper and improper autumatic operation of the Reactor Coolant Pump seat injechon support system.
A Discriminant Valkhty Plausible misconception since the question stem provides leak indications in the same plant building.
8 Discriminant Validity Plausible misconception since the question stem provides leak indications in the same plant building.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception since the question stem provides leak indications in the same plant building.
TMl Lirnsing Excmin:ti::n An:wcr K!y QUESTION: 015 (1.00)
Some reactor trip situations require large volumes of makeup water for RCS inventory control simultaneous with the need to emergency borate the core. Which ONE (1) source should be used for this condition?
A. 4% BAMT (CA-T-8) ;
B. Concentrated Waste Storage Tank (WDT-6A/B)
C. RC Bleed Tank 1C (WDL-T-1C)
D. BWST (DH-T-1) 1 l
l Answer: D K/A: 000004 K4.07 3.3 Page 3.2-8 Objective: IV.A.09.25 ;
I Reference (s): ATP 1210-10 Abnorma! Transient Rules, Guides, and Graphs, Rev. 34, Page 7 History: New NRC Cognitive Level Rating 1 Measurement #15 Demonstration of knowledge conceming Makeup system design features that provide for water supplies.
A Discriminant Validity Plausible misconception since this tank is used for addition of concentrated Boric Acid to the Core Flood Tanks.
B Discriminant Validity Plausible misconception since this tank is used to capture evaporator bottoms which also contain high concentration boric acid, and could be mistaken for the Redaimed Boric Acid Storage Tank.
C Discriminant Validity Plausible misconception since this tank is large, its contents can be pumped into the makeup system, and its boron concentration is the highest of all 3 RC Bleed Tanks.
D Discriminant Validity Correct answer.
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TMl Licensing Ex: min; tion !
Answer K y QUESTION: 016 (1.00)
It is necessary to evacuate the control room due to a senous fire in the relay room. Which required action must be performed prior to exiting the control room?
j A. Perform notifications for an ALERT. !
B. Start EG-Y-1 A and EG-Y-18.
C. Trip the MFW pumps.
D. Close MU-V-3.
Answer: C K/A: 000068 AK3.18 4.5 Page 4.2-55 Objective: V.D.18.02 )
Reference (s): EP 1202-37,' Cooldown from Outside the Control Room, Rev. 50, Page 2.0 History- New NRC Cognitive Level Rating 1 Measurement #16 Demonstration of ability to direct adivities required for control room i evacuation in accordance with facility procedures and requirements.
A Discriminant Validity Plausible misconception , =ince declaration of an Alert is the Srst follow-up action to be performed after evacuating the Control Room (Follow-up Step
- 1).
B Discriminant Validity Plausible misconception since this action is included in the follow-up actions to be performed after evacuating the Control Room.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception since this action is included in the follow-up edions to be performed after evacuating the Cor. trol Room.
TMI Lic::n:ing Examin:ti:n Answ;r KCy QUESTION: 017 (1.00)
RCPs are bumped during inadequate coro heat removal conditions. Which ONE statement desenbos the reason for this action? .
I A. Decrease RCS pressure B. Induce OTSG heat transfer C. Prevent RCS inw.,ntory loss D. -increase OTSG pressure l
Answer: B i i
K/A- 000074 EK1.03 4.9 Page 4.1-15 Objective: V.E.06.03
)
Reference (s): Lesson Plan, 11.2.01.219, Lack of Primary to Secondary Heat Transfer, Rev 12 Page 8 History: New NRC Cognitive Level Rating 2 4
Measuremsnt #17 Knowledge of operationalimplications of the process for removing decay heat from the core.
A Discriminant Validity Plausible distracter since RCS pressure may decrease dramatically when RCP is bumped.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since operation of RCPs in this situation actually increases rate of inventory loss from the RCS.
D Discriminant Validity Plausible distracter since OTSG pressure may change significantly when RCPs are bumped under inadequate core cooling conditions.
-_ . . . - - .- - . .. - - -- .- - _ - . . - - . = . . .
TMl Lic:nsing Ex minati:n An:wcr Kcy Question: 018 (1.00)
A small break LOCA is in progress. Which ONE (1) set of conditions requires tripping all RCPs?
RCS TEMP RCS PRESS A. 579'F 1800 psig B. 537'F 1300 psig C. 525'F: 1000 psig D.~ 473*F 800 psig Answer: C K/A: 000009 EK3.23 4.3 4.1-4 Objective: V.E.02.03 Reference (s): ATP 1210-10 Abnormal Transients, Rules, Guides, and Graphs, Rev 34, Page5 History: New
. NRC Cognitive Leve' Rating 3 Measurement #18 Demonstration of knowledge of reasons for responses of RCP tripping requirements during small break LOCA conditions.
A Discriminant Validity Plausible answer since examinee must calculate subcooled margin corredly and then apply the RCP trip rule (and correct limit) to detennine if action is required.
8 Discriminant Validity Plausible answer since examinee must calculate subcooled margin correctly and then apply the RCP trip rule (and correct limit) to determine if action is required.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible answer since examinee must calculate subcooled margin corredly and then apply the RCP trip rule (and correct limit) to determine if action is required.
TMI Lirnsing Examin: tion Answer K;y QUESTION: 019 (1.00)
Which ONE condition requires a Core Flood Tank to be declared inoperable?
A.- Boron is 2290 ppm.
B. Pressure is 620 psig.
C. Levelis 14 ft.
D. Temperature is 100 degrees F.
l Answer: C i
K/A . Generic K2.1.12 4.0 Page 2-2 l
Objective: IV.A.13.05 IV.A.13.08 1
Reference (s): Technical Specifications 3.3.1.2, Core Flood Tanks, Amendment 203, Page 3- i l
History: TMI Exam Bank Question AL4A13-08-QO2
- Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #19 Demonstration of ability to interpret and apply technical specifications for the Core Flood System.
A Discriminant Validity Plausible distracter since the BWST boron concentration requirement has been changed to 2500 ppm.
B Discriminant Validity Plausible distracter since this value is just under the Tech Spec upper limit of 625 psia.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter since shroud heaters are installed (although not needed) to maintain minimum tank temperature 5
- - - - - - y. ,% - - yw__. - 9 -..
l TMI Lic:nring Ex: min:ti::n Answer Kcy QUESTION: 020 (1.00) l Conditions require manual reactor trip. Upon depressing the Trip AND DSS pushbuttons,
' the reactor does not trip (reactor power remains at 100%). Which ONE statement describes the required action?
A. Place the EHC pump control switches in P-T-L and open EHC-FV-1.
I l
B. Place the diamond rod control station in manual and reduce reactor power. '
C. Initiate HPl and maintain primary to secondary heat transfer until power level is l less than 10%. l D. Transfer FW to manual to control OTSG levels.
Answer: C K/A: 000029 EK3.12 4.'7 Page 4.1-8 Objective: V.E.01.06 Reference (s): ATP 1210-1 Reactor Trip Rev 37, Page 2.0 History: New 1 NRC Cognitive Level Rating 3 Measurement #20 Demonstration of knowledge conceming responses and actions required for A1YVS emergency situation.
A Discriminant Validity Plausible distrader since this is a remedial action for failure of the turbine to trip.
B Discriminant Validity Plausible misconception since this action would reduce reador power.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distrader since this is a remedial action in the ReactorTrip ATP.
1
TMI Lic:nsing ExImin:ti:n An;w:r Kcy QUESTION: 021 (1.00)
BS-P-1B is running for surveillance. Which ONE condition would result in automatic trip of tho' BS-P-1B breaker?
A. 1 A Aux Transformer fault with Auto Transfer of loads to 18 Aux Transformer B. 1B Aux Transformer fault with Auto Transfer of loads to 1 A Aux Transformer C. Fault downstream of 1P 480v Bus low side feeder breaker 1
D. Fault downstream of 1S 480v Bus low side feeder breaker ;
1 l
Answer. A l K/A: 000026 K2.01 3.6 Page 3.5-10 1
Objective: IV.A.15.08 Reference (s): OPM A-01 B'OP and IE Electrical Distribution, Rev 16, Pages 36 and 37 History: New NRC Cognitive Level Rating 2 Measurement #21 Application of knowledge regarding physical connections and cause-effect relationships between power distribution system and Reactor Building Spray System.
A Discriminant Validity Correct answer. ;
B Discrim.nant Validity Plausible misconception that automatic transfer occurs on 4KV ES Switchgear, coupled with incorred power supply. The ES bus does not transfer to attemate Auxiliary Transformer, but rather to the backup emergency diesel.
C Discriminant Validity Plausible misconception that 480 V Bus fault results in loss of 4KV feeder Bus coupled with incorrect power supply for BS-P-1B.
D Discriminant Validity Plausible misconception that 480 V Bus fault results in loss of 4KV feeder Bus.
. .. . -- - -. - - - - . _ ~ - . . - - . - - - . . _ .
I TMl Liccnsing Ex:minatien l- Answer K y QUESTION: 022 (1.00)
Current plant conditions are:
l -
Pump down of RCS is in progress.
DH-P-1Ais in service I - Low DH Pump Flow annunciator C-1-7 is actuated.
RCS temperature is increasing.
Discharge pressure for DH-P-1 A is unstable Motor amperage indication for DH-P-1 A is unstable.
7 Noise is reported in DH-P-1 A vault.
, Which ONE statement desenbos the required operator actions?
A. Place Loop B Decay Heat Removal System in service, and trip DH-P-1 A.
B. Trip DH-P-1 A, and do NOT restart this pump until appropriate actions have been completed.
C. Place Loop B Decay Heat Removal System in service, and leave DH-P-1 A running until conditions stabilize.
D. Start one Makeup Pump and do NOT trip this pump until incore Thermocouple temperatures stabilize.
Answer: B K/A- 000025 AK3.03 4.1 Page 4.2-16 Objective: V.D.16.02 Reference (s): EP 1202-35 Loss of Decay heat Removal System, Rev 31, Page 4.0 History: TMI Exam Bank Ouestion SR5D16-02-QO2 (Modified)
Unmodified version was used on audit exam.
NRC Cognitive Level Rating 3 Measurement #22 Knowledge of reasons for responses that apply to loss of reactor core decay heat removal, and ability to dired immediate actions required for response to loss of Decay Heat Removal.
l _
A Discriminant Validity Plausible misconception to start attemate cooling system.
8 Discriminant Validity _ Correct answer.
. C Discriminant Validity "ausible misconception to start attemate cooling system.
l D Discriminant Va:!dity , vlausible misconception to initiate attemate cooling method.
l
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TMl Licen:Ing Examin:ti n Answer K y QUESTION: 023 (1.00)
Current plant conditions are:
- Reactor is operating at 100% power.
- At 0715 a regulating rod became stuck and misaligned by 12" from the group average.
- At 0750 (current time) the problem causing the rod to be stuck is corrected and the rod is ready to be realigned.
What is the maximum permissible power level at which the control rod may be realigned with the group?
A. 45%
B. 55%
C. 60 %
D. 100 %
Answer: D K/A: 000005 EK3.05 4.2 Page 4.24 Objective: V.D.03.04 Reference (s): EP 1202-8 CRD Equipment Failure, Rev 50, Page 4.0 History: New NRC Cognitive Level Rating 3 Measurement #23 Knowledge of requirements for reasons for core power limits on control rod misalignment.
A Discriminant Validity Plausible misconception since this power level (45%) is related to the three-pump limit for operation with an asymmetric rod.
8 Discriminant Validity Plausible misconception since 55% is the ICS runback setpoint for a (dropped) asymmetric rod condition.
C Discriminant Validity Plausible distracter since the procedure limits power to 60% after one hour.
This is also the Tech Spec limit.
D Discriminant Validity Correct answer.
TMI Licensing Ex min tion Answer K:y QUESTION: 024 (1.00)
Current RC-P-1 A conditions:
Number one seal leak-off flow indication is 5.8 gpm.
Periodic RCP shaft vibration ALERT alarms are actuating - alarms can be reset without immediate re-actuation.
Bentley-Nevada vibration readings are ranging between 14 and 18 mils.
Number one Seal leak-off temperature indication is 197'F.
Radial Bearing temperature indication is 170*F.
High Standpipe level alarm is clear.
Which ONE failure could cause the above indications?
A. Seal #1 B. Seal #2 C. Seal #3 l
D. Labyrinth seal Answer: A K/A: 000015/17 AA1.22 4.2 Page 4.2-11 Objective: V.C.06.01 Reference (s): Abnormal Procedure 1203-16, Reactor Coolant Pump and Motor Malfunction, Rev 38, Page 2 History- TMI Exam Bank Question SRSC06-01-001 (Modified)
Verified not used on audit exam.
Original version was used in SRO Program quiz 4/23/98.
NRC Cognitive Level Rating 2 Measurement #24 Ability to operate and monitor operations and responses to RCP seal failure malfundion while adhering to facility requirements.
A Discriminant Validity Correct answer. l B Discriminant Validity Plausible misconception regarding symptoms of seal #2 failure.
C Discriminant Validity Plausible misconception regarding symptoms of seal #3 failure.
D Discriminant Validity Plausible misconception regarding symptoms of labyrinth seat failure.
l
t-
- . TMI Licen2ing Ex'_mination l- Answer K
- y QUESTION: 025 (1.00) '
Current plant conditions are:
Reactor is operating at 95% power.
RCS pressure is 2150 psig and decreasing slowly.
Pzr level is 200 inches and decreasing slowly.
MU TANK LEVEL LO alarm is actuated.
L -
SEAL INJECTION FLOW LOW alarm is actuated.
l - Total RCP seal injection flow indication has decreased to 22 gpm.
Whidi ONE abnormal condition could result in these indications?
A. RCP seal failure B. MU-V-17 failed open L C. Sealinjection line leak l - D. Makeup line leak Answer: D K/A: 000022 AA2.01 3.8 Page 4.2 d3 .
Objective: V.C.05.01 l
L Reference (s): C-302-661, Rev 49, Makeup and Purification System i= History:- New
'NRC Cognitive Level Rating 3 Measurement #25 l Ability to evaluate and respond to a charging line leak.
A Discriminant Validity Plausible misconception of symptoms of RCP seal failure.
B Discriminant Validity Plausible misconception since the question stem shows the MU Tank level decreasing with MU Tank low level alarm aduatino. l C Discriminant Validity Plausible misconception since seal injection flow has decreased in the question stem.
D D6scriminant Validity Correct answer.
4 r- - - - ----
TMI Lic:n:ing Ex minitlen Answer K y QUESTION: 026 (1.00)
Current plant conditions are:
Reactoris' operating at 100% power.
INST AIR PRESS LOW TURBINE AREA alarm is actuated.
INSTRUMENT AIR PRESS LOW AUX BLDG AREA alarm is aduated Instrument Air pressure is 58 psig.
Secondary plant is stable.
Which ONE statement describes the required actions?
A. Manually trip the reactor and perform the immediate Manual Actions of 1210-1, Reactor Trip. l 1
B. Dispatch operators to start backup instrument air compressors.
C. Maintain power at present level and make plant page and radio announcement ;
to all personnel using instrument air to stop use immediately.
D. Cross connect instrument and service air headers until cause for low header pressure is determined and corrected.
Answer: A K/A: 000065 AA1.05 3.3 Page 4.2-50 Objective: V.D.17.03 Reference (s): EP 1202-36, Loss of Instrument Air, Rev 27, Page 3.0 '
History: New NRC Cognitive Level Rating 3 Measurement #26 Ability to operate and monitor the reactor protection system as it applies to a loss of instrument air.
A Discriminant Validity Correct answer.
8 Discriminant Validity Plausible misconception since there are specially installed backup instrument air compressors for continued operation of seleded plant components.
C Discriminant Validity Plausible misconception for remedial action.
D Discriminant Validity Plausible misconception for remedial action.
l TMI Lic:n:Ing Excmin' tion An:w r K;y QUESTION: 027 (1.00) ;
Current plant conditions are: !
- ESAS actuation (A & B) occurred 1 minute ago. I Which ONE statement describes the operation of the RB fans and coolers (AH-E-1N1B/1C)? !
A. Fans run in slow speed with river water flowing through the emergency cooling coils.
B. Fans run in fast speed with river water flowing through the emergency cooling i coils. l l
C. Fans run in slow speed with river water flowing through the normal and emergency cooling coils. j l
D. Fans run in fast speed with river water flowing through the normal and l emergency cooling coils.
Answer: A l 1
K/A: 000022 A3.01 4.3 Page 3.54 Objective: IV.A.17.08 Reference (s): OPM L-01 Reactor Building Cooling and Ventilation System, Rev 8, Page 7 l
History: TMI Exam Bank Question AL4A17-08-QO1 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 2 Measurement #27 Ability to monitor and evaluation automatic operation of the containment cooling system follow:ng initiation of safeguard mode of operation.
A. Discriminant Validity Correct answer.
B Discrimina:J Validity Plausible misconception conceming containment heat removal requirements during LOCA conditions.
C Discriminant Validity Plausible misconception conceming containment heat removal requirements during LOCA conditions.
D Discriminant Validity Plausible misconception conceming containment heat removal requirements during LOCA conditions.
t j TMl Lic:nzing Ex min ticn Answ:r K y QUESTION: 028 (1.00)
. Initial plant conditions are:
- Reactoris shutdown.
- RCS temperature is 250*F and stable.
- RCS pressure is 250 psig and stable.
- - Pressurizer level is 200 inches and stable.
- DR-P-1 A, DC-P-1 A, and DH-P-1 A are operating.
4
- DC-V2A is closed, and DC-V-65A is open.
- Operators have just initiated DHR flow through the "A" DHR Cooler.
The following parameters are now changing:
I j - Pressurizer level is slowly decreasing. ;
- LT-109 indicates DC-T-1 A is increasing. i
- RM-L-2 count rate is increasing.
l Identify the ONE cause for the above conditions:
A. DHCCW temperature is increasing due to energy transfer from the DHRS.
B. DC-T-1 A fill valve (DC-V-19A) is failing open.
C. DHR cooler is leaking into the DHCCWS. I D. LT-109 is failing high.
Answer: C K/A: 000073 K1.01 3.9 Page 3.7 Objective: IV.B.01.06 Reference (s):C-302 645, Rev. 32, Decay Heat Closed Cooling System History- New NRC. Cognitive Level Rating 3 l
l i
i i
I l
TMI Lic:n:ing Examin ti::n An:w:r Kcy Measurement #28 Ability to evaluate hymptoms (induding Radiation Monitoring System responses) and mitigate consequences for closed cooling system heat exchanger tube leakage from contaminated RCS (Decay Heat Removal System).
A Discriminant Validity plausible distrador since surge tank level will increase due to expansion of water as DCCS temperature is elevated.
8 Discriminant Validity Plausible distracter since this surDe tank makeup valve DC-V-19A fails open on loss Instrument Air.
C Discriminant Validity Correct answer.
D Discriminant Validity PlausiUe distracter for surge tank levelincrease.
TMl Lic:nsing Excmin tion An*_ wor K y QUESTION: 029 (1.00)
Initial plant conditions are:
- Reector is operating at 100% power.
- CO-P-1 A and CO-P-1B are running, CO-P-1 C is in Normal-After-Stop.
- CO-P-2A and CO-P-2B are running, CO-P-2C is in Normal-After-Stop.
- CO-P-1B trips on an electrical fault.
- After a period of two (2) seconds, CO-P-1C automatically starts.
Which ONE statement describes the response for these conditions?
A. One condensate booster pump will trip, both main feed pumps remain running, with an ICS runback.
B. One condensate booster pump and one main feed pump will trip, with an ICS runback.
C. One main feed pump will trip, both condonsate booster pumps remain running, with an ICS runback.
D. Both main feed pumps trip, one condensate booster pump trips and the reactor trips.
1 Answer: B j KIA- 000054 EA2.05 3.7 Page 4.2-36 Objective: IV.C.02.10 i
Reference (s): OPM G-01 Condensate System, Rev 12, Page 7 History: NRC Exam Bank NRC Cognitive Level Rating 3 Measurement #29 Demonstration of ability to evaluate and interpret the status of the main feedwater pumps as they apply to a (partial) loss of main feedwater.
A Discriminant Validity Plausible misconception regarding extremely complicated condensate-condensate booster pump counting circuit.
8 Discriminant Validity correct answer.
, C Discriminant Validity Plausible misconception regarding extremely complicated condensate-condensate booster pump counting circuit.
D Discriminant Validity Plausible misconceptior regarding extremely complicated condensate-condensate booster pump counting circuit.
l
l TMl Licensing Extmination l Anawer K;y l QUESTION: 030 (1.00)
Currort plant conditions are:
)
- The reactor is operating at 95% power. )
- Control Rod exercising is in progress. l
- Rod Control, Feedwater, and Reactor-Steam Generator Master controls are in l
MANUAL.
- An "OUT-INHIBir condition is illuminated on the Diamond rod control panel.
- A rapid reduction in power level has just occurred with fluctuations in RCS temperature, l pressure and pressurizerlevel. !
- Asymmetric Rod alarm is actuated.
- Current NI readings are as follows:
NI-5 NI-6 NI-7 NI-8 90% 96 % 97 % 96 %
Identify the ONE cause for the above conditions:
A. Ni detector power supply fault B. Partial insertion of rods during exercising C. Azimuthal xenon oscillation 1
D. Dropped rod Answer: D K/A: 000003 AA2.03 3.8 Page 4.2-5
History: New NRC Cognitive Level Rating 3 l Measurement #30 Ability to interpret symptoms of dropped rod using incore/excore instruments as applied to a dropped control rod.
A Discriminant Validity Plausible misconception regarding ion chamber failure modes.
B Discriminant Validity. Plausible misconception regarding effect of partialinsertion of safety rod
- groups during exercise testing.
1' C Discriminant Validity Plausible misconception regarding initiation of Xenon oscillations due to control rod insertion and withdrawal.
D Discriminant Validity Correct answer, t
. - - - . .._ - -- . --. - - - ~ -. . - -. .
l
( TMI LI:enzing Excminitirn An:wer K y QUESTION: 031 (1.00)
Current plant conditions are:
- Reactor is operating at 100% power.
l - RCS pressure is 2155 psig.
- - CRO closed RC-V-2 one minute ago due to suspected PORV leakage.
!- - RC Drain Tank pressure is 5.0 poig.
L l If the PORV is leaking, what is the expected tailpipe temperature?
n A. 162 degrees F B. 212 degrees F C. 228 degrees F D. 267 degrees F Answer. C K/A- 000008 AA1.01 4.0 Page 4.2-8 l
Objective: lli.C.02.12 i Reference (s): Steam Tables I
! History- New I
NRC Cognitive Level Rating 3 l Measurement #31 Ability to evaluate plant conditions to properly monitor and operate the l PORV/ Block Valve during pressurizer steam space leak (leaking PORV).
A Discriminant Validity Plausible distracter since this is the corred answer for 5 psia rather than 5 psig in case the examinee forgets to convert.
l B Discriminant Validity Plausible distracter since this is the corred answer for atmospheric pressure
- in case the examinee fails to consider the positive pressure in the Reador Coolant Drain Tank.
l C Discriminant Validity Correct answer.
D Discriminant Validity This is a plausible distracter since this is the correct answer for 20 psig rather than 20 psia.
I
TMI Lic:nsing Extmin;ti:n Answ r Kcy QUESTION: 032 (1.00)
Current plant conditions are:
Reactor is at cold shutdown condition.
- "A" Decay Heat Removal string is operating.
- Decay Heat Closed Cooling flow through the Decay Heat Removal cooler is throttled to maintain the RCS at 130*F.
- Total loss of instrument Air (0 psig) occurred.
Which ONE statement describes the response of the cooling system and subsequent effect on RCS temperature for this situation?
A. Closure of DC-V-65A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in RCS heatup.
B. Opening of DC-V-65A (Cooler bypass) AND DC-V-2A (Cooler inlet) results in RCS cooldown.
C. Closure of DC-V-2A (Cooler inlet) results in RCS heatup.
D. Opening of DC-V-2A (Cooler inlet) results in RCS cooldown.
Answer: D K/A: 000008 A2.05 3.5 Page 3.8-4 Objective: IV.B.01.13 1
Reference (s): EP 1202-36, Loss of instrument Air, Rev 27, Page 6 and Ei-1 History: . TMl Question AL4B01-13-QO1 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 3 Measurement #32 Knowledge of failure modes applied to predict impad of loss of instrument air on Decay Heat Closed Cooling System.
1 A Discriminant Validity Plausible distracter since closure of both valves 0",,ared response forloss of Instrument Air) will result in loss of cooling ano therefore RCS heatup.
8 Discriminant Validity Plausible distracter since opening both valves (incorred response for loss of
- instrument Air) could possibly result in RCS cooldown.
I C Discriminant Validity Plausible distracter since closure of DC-V 2A (incorred response for loss of Instrument Alr) would result in RCS heatup.
i D Discriminant Validity Correct answer.
TMl LI::enzing Extmin tien ;
Answ:r Kty i 1
QUESTION: 033
~
(1.00) l Which ONE statement describes the requirements for performing an Independent Verification l for a valve that is required to be in the closed position?
l A. One individual closes the valve, SECOND individual venfies from an independent remote position DEMAND indicator that the valve closed B. One individual closes the valve, SAME individual independently uses a remote position DEMAND indicator to verify valve is closed.
C. One individual closes valve, SECOND individual independently venfies the valve is closed (locally).
D. One individual closes valve (locally), SAME individual verifies from an independent remote POSITION indicator that the valve is closed.
Answer: C K/A: Generic 2.1.29 3.3 Page 2-4 Objective: V.J.01.02 i
Reference (s): Administrative Procedure 1067, independent Verification, Rev 28 History: New l NRC Cognitive Level Rating 1 l
l Measurement #33 Ability to direct and evaluate conduct and verification of plant valve lineups in accordance with facility requirements and procedures.
A Discriminant Validity Plausible distracter since two individuals are used but includes misconception that use of DEMAND Indication is an acceptable means of verification.
B Discriminant Validity Plausible distracter since two independent means of position verification (misconception that use of DEMAND indication is acceptable)is employed -
even though performed by the same individual.
i C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter since two valid independent means of position verification (misconception that this is acceptable - even though performed by the same individual).
l l
1 TMl Licen:ing Exrminiti n Answer Ksy l QUESTION: 034 (1.00)
Current plant conditions are:
Reactor is tripped.
- Large break LOCA is in progress.
- RCS pressure is 540 psig.
- Reactor Building pressure is 35 psig.
When ESAS actuates properly, which ONE statement describes the expected lineup for the listed support systems?
A. Seal injection is isolated to the RCPs since they should be tripped.
B. NSCC is isolated to the RCP motors since they should be tripped C. NSCC is aligned to the RCP motors to support pump operation if needed.
I D. ICCW is aligned to the RCDT cooler to prevent flashing in the tank.
Answer: B K/A: 000026 AK3.02 4.2 Page 4.2-18 Objective: IV.E.24.06 Reference (s): OPM F-06, ESAS, Rev 8, Page 14 l History: New l NRC Cognitive Level Rating 3
! Measurement #34 Demonstration of knowledge of operationalimplications ofisolation of dosed cooling system as a result of actuation of engineered safeguards during LOCA conditions.
A Discriminant Validity Plausible misconception, since the RCPs would be tripped by the operators in this situation (loss of subcooled margin).
l B Discriminant Validity Correct answer.
i C Discriminant Validity Plausible misconception since RCPs are restarted and/or bumped under specified accident conditions.
D Discriminant Validity Plausible distracter since RC Drain Tank is located inside containment, and receives discharge of Pressurizer relief valves and PORV.
i
.. . - . . - -. - - -~ - - .. - - - . - - - - .- -
TMl Licensing Ex: min:ti::n An;wer Kcy QUESTION: 035 (1.00)
Dunna a radioactive gaseous release from a Wasta Gas Decay Tank, RM-A-7 HIGH alarm '
actuated. Whidi ONE statement describes required automatic actions for this condition?
A. Trips AH-E-11
. B. Trips AH-E-10 & 11, closes WDG-V 47, and starts MAP-5 iodine Sampler C. Closes WDG-V-47 D. Closes WDG-V-47 and starts MAP-5 iodine sampler Answer. C KIA: 000071 K1.06 3.1 Page 3.9-5 Objective: IV.B.08.08 -
Reference (s): OPM E-04, WDG System, Rev 8, Page 26 History: New
'NRC Cognitive Level Rating 3 i Measurement #36 Demonstration of ability to evaluate proper automatic operation and interrelationship between Waste Gas Disposal System and the Radiation i' Monitoring System.
A Discriminant Validity Plausible misconception since AH-E-11 provides ventilation flow to the Auxiliary Building, and the release is routed into the building exhaust ventilation flow.
8 Discriminant Validity Plausible misconception since AH-E-10 and AH-E-11 provide ventilation supply air flow to the Auxiliary and Fuel Handling buildings, and the release is routed into the combined building exhaust ventilation flow. Also includes plausible misconception that a MAP sampler is located on gas release flowpath as it is on the adjacent RB PurDe exhaust duct. j C Discriminant Validity Correct answer.
D Discriminant Validity Correct answer with p!ausible misconception that a MAP lodine sampler is installed on gas release flowpath as it is on the adjacent RB Purge Exhaust duct.
TMl Litensing Exrmin tirn Answer Key -
QUESTION: 036 (1.00)
Which ONE statement is the basis for the power-imbalance RPS trip envelop for nuclear overpower?
A. Assure acceptable power distnbution is maintained for control rod misalignment ,
analysis. I B. Assure Nuclear Peaking Factors are within limits in the event of a cold water accident.
C. Assure transient protedion (minimum DNBR) is maintained for loss of coolant flow events.
D. Assure uniform fuel burn 42p over core life.
Answer: C K/A: 000015 K5.11 . 3.7 Page 3.7-6 Objective: IV.E.14.06 Reference (s): COLR, Rev 1, Page 31 History: New NRC Cognitive Level Rating 1 Measurement #36 Demonstration of ability to evaluate and predict operationalimplications and effects of operation with reactor core axial flux Imbalance.
A Discriminant Validity Plausible misconception since misalignment produces local power distribution perturbations. !
R Discriminant Validity Plausible distracter, since power peaking factors are affected by operation l with core axialimbalance. i C Discriminant Validity Correct answer. l D Discriminant Validity This is a true statement, but it is not the correct reason for the basis of tne power-imbalance trip envelope.
l TMI Ll:ensing Ex min . tion i An:wer Key l
l QUESTION: 037 (1.00)
Current plant conditions are:
l
- Large break LOCA is in progress.
- RB pressure and temperature are elevated.
- RCS is experiencing extreme depressurization.
Which ONE statement describes the cause and effect for erroneous pressurizer level indication during LOCA conditions?
A. Level indicates high due to RB depressurization effects by RB spray.
l B. Level indicates low due to RB depressurization effects by RB spray. 1 C. Level indicates high due to reference leg boiling.
D. Level indicates low due to reference leg boiling.
Answer: C l
K/A: 000011 K6.04 3.1 Page 3.2-21 l
Objective: IV.E.09.02 l l
' Reference (s): ATP 1210-10 Abnormal Transient Rules, Guides, and Graphs, Rev. 34, Page l 10 '
History- New NRC Cognitive Level Rating 2 Measurement #37 Demonstration of ability to predict and evaluate effects of post LOCA containment environment on RCS pressurizer levelindication.
A Discriminant Validity Plausible misconception regarding effect of RB depressurization after a LOCA.
8 Discriminant Validity Plausible misconception regarding effect of RB depressurtzation after a LOCA.
! C Discriminant Validity Correct answer.
i D Discriminant Validity Plausible misconception (opposite effect) of reference leg boiling on level indication.
i I
l
TMI Licen ing Examin ti n
. An:wer K y QUESTION: 038 (1.00)
Emergency feedwater pump suction can be lined up directly from the hotwell. After vacuum is broken, the Emergency Alignment pushbutton is depressed to realign the following valves:
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- Condensate Reject Valve CO-V-6
- Normal Makeup Valve to the Hotwell CO-V-7
- Emergency Makeup Valve to the Hotwell CO-V-8 Which ONE statement describes valve response to operation of this pushbutton?
A. CO-V 6 and CO-V-7 open, CO-V 8 closes.
B. CO-V-6 and CO-V-7 close, CO-V 8 opens.
C. CO-V-6 opens, CO-V-7 and CO-V-8 closes.
D. CO-V-6 and CO-V-8 open, and CO-V-7 closes.
Answer: B K/A: 000061 A2.05 3.4 Page 3.4-47 Objective: IV.C.02.08 Reference (s) OPM G-01 Condensate System, Rev 12, Page 13 History: TMI Question AL4C02-08-QO2 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 3
- Measurement #38 Demonstration of ability to predid the impact of automatic control ,
malfundions on Emergency Feedwater, and to direct corrective actions for j mitigation of the effects.
I A Discriminant Validity Plausible misconception regarding emergency alignment configuration ;
and/or control dreuitry. !
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception regarding emergency alignment confgurittkin and/or control drcuitry.
D Discriminant Validity Plausible misconception regarding emergency alignment configuration and/or control drcuitry.
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TMl Licensing Examination Answer K y
. QUESTION: 039 (1.00)
Which ONE accident situation would result in a direct Main Feedwater isolation by HSPS?
A. Feedwater line break outside RB at FW-V-17NB B. Steam line rupture outside RB C. Large break LOCA D. OTSG tube rupture
/ Answer B K/A: 000040 AA2.04 4.7 Page 4.2-32 l
Objective: IV.E.05.03 Reference (s): OPM F-10, Heat Sink Protection System, Rev 9, Page 6 History: New NRC Cognitive Level Rating 1 Measurement .Y39 Demonstration of ability to determine and interpret conditions which require Heat Sink Protection System aduation during steam line rupture conditions.
A Discriminant Validity Plausible misconception since examinee may not recall FW line check valves just outside the containment building wall will prevent dired depressurization of the OTSG during this situation.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since the RCS undergoes rapid depressurization and cold water is injeded into the RCS Cold Legs.
D Discriminant Validity Piausible misconception regarding OTSG pressure response during abnormal conditions (high tube rupture flows).
TMI Lic:n:ing Ex mination Answer K:y QUESTION: 040 (1.00)
Which ONE statement describes the restrictions on operation of TMl-1230KV switchyard auxiliary transformer disconnect switches?
A. Opening operations limited to normal load current interruption.
B. Opening operations limited to isolation of energized transformers (unloaded).
C. Opening operations limited to isolation of transformers currently de-energized.
D. Closing operations ONLY after synchronization is completed due possible arcing.
Answer: B K/A: 000062 A2.07 3.4 Page 3.6-4 Objective: IV.G.01.07 Reference (s): OPM A-01 BOP and IE Electrical Distribution, Rev 16, Page 62 History: TMI Question AL4G01-07-QO1 (Modified)
Verified not used on audit exam or in SRO Prog!am quizzes.
NRC Cognitive Level Rating 1 Measurement #40 Ability to use procedures to control or mitigate the consequences of opening disconnects under loaded conditions.
A Discriminant Validity Plausible misconception that these devices can be safely operated under normalload (current) conditions.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception that these devices can only be operated safely under no-voltage conditions rather than no-current conditions.
D Discriminant Validity Plausible misconception since synchronization is required for some breaker closure operations.
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TMI Lic:nzing Extmin tion Answer K0y QUESTION: 041 (1.00)
Current plant conditions are:
- Reactor poweris 12%.
- Main turbine is on line in manual control.
- Unit load demand is 12%.
- Turbine Bypass Valves are fully closed in automatic control.
- Steam header pressure is 885 psig.
Without operator actions, which ONE statement describes the response of the turbine bypass valves to a pressure increase in both OTSGs to 960 psig?
A. Valves remain closed, since the pressure is less than setpoint plus selected bias.
B. Valves open to reduce pressure to setpoint plus 10 psig.
C. Valves open to reduce pressure to setpoint plus 75 psig.
D. Valves open to reduce pressure to setpoint plus 125 psig.
Answer: B K/A: 000041 A1.02 3.2 Page 3.4-25 Objective: IV.E.27.17 Reference (s): OPM G-03 Main Steam System, Rev 12, Page 9.0 & 10.0 History: New NRC Cognitive Level Rcting 2 Measurement #41 Ability to predict changes in parameters associated with Main Steam Turbine Bypass Control associated with steam pressure.
A Discriminant Validity Plausible discriminator if examinee does not know which bias should be selected by the automatic control system.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible distrader since ICS uses this bias - but only after greater than 15% Unit Load Demand if the turbine bypass valves are dosed.
D Discriminant Validity Plausible distracter since ICS does use this bias - but only under reador trip conditions.
TMI Lic:nPg Ex:mination An:wer Kcy QUESTION: 042 (1.00)
An anticipatory reactor trip due to a turbine trip, or a trip of both feedwater pumps is designed to prevent which ONE condition?
A. Challenges to steam generator tube integrity B. Exceeding core thermal limits (KW#t limits)
C. Challenges to the PORV and pressurizer code safeties D. Exceeding core DNBR limits Answer: C 1 K/A: 000045 K1.18 3.7 Page 3.4-27 '
Objective: IV.E.14.06 Reference (s): OPM F-02 R'eactor Protection, Page 19, Rev 7 l
History: NRC Exam Bank l NRC Cognitive Level Rating 1 Measurement M2 Knowledge of cause and effect relationships between the Main Turtnne Generator system and reactor protedion system.
A Discriminant Validity Plausible misconception since OTSG tubes experience increased stress during conditions of loss of MFW and Turbine trip.
B Discriminant Validity Plausible distracter, since reactor trip limits are being addressed in this question.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter, since reactor trip limits are being addressed in this question.
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i TMI Lic:nsing Excminstion l l Answ r Ksy l
QUESTION: 043 (1.00)
Current plant conditions are: l Plant has experienced a reactor trip and loss of offsite power.
EG-Y-1B failed to automatically START. 1 Which ONE condition will prevent autcmatic startup of the diesel generator?
A. The exciter Auto Manual switch in the control room is in the Manual position.
B. The Emergency Bypass selector switch at the EDG breaker cubicle is in the Emergency position.
C. The Unit / Parallel switch is in the Parallel position.
D. 1 B Diesel Auto-Standby / Manual-Exercise switch in the control room is in the Manual Exercise position.
Answer: D K/A: BW/A05 K3.3 3.8 Page 4.3-35 ;
Objective: IV.G.08.14 Reference (s): OPM A-04 Diesel Generator and Auxiliary Equipment System, Rev 11, Pages 5 and 6 History: New NRC Cognitive Level Rating 3 Measurement #43 Ability to analyze and predict consequences of manipulation of emergency diesel generator controls required to obtain desired results during abnormal and emergency situations.
A Discriminant Validity Plausible misconception that having this switch in Manual would prevent automatic diesel starting.
B Discriminant Validity Plausible misconception that having this switch in EmerDency rather than in normal position would prevent automatic diesel starting.
C Discriminant Validity Plausible misconception that having this switch in Parallel would prevent automatic diesel starting.
D Discriminant Validity Correct answer.
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l TMI Licensing Examination Answer K y QUESTION: 044 (1.00) l Which ONE statement describes the purpose of the emergency diesel generator governor speed droop adjustment.
A. Limits voltage changes during load changes when running in Unit.
B. Adjusts engine response to load changes. ;
C. Limits maximum engine load.
D. Prevents engine overspeed during initial start.
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- 1 Answer:. B KIA: 000064 K4.06 2.7 Page 3.6-9 Objective: IV.G.08.02 Reference (s): OPM A44 Diesel Generator and Auxiliary Equipment System, Rev 11, Pages 8 and 14 History: TMI Question AL4G08-02-QO1 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating i l
I Measurement #44 Explains operating responses of safety-related emergency equipment control settings.
A Discriminant Validity Plausible misconception since this is the function of the generator Voltage Droop control.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since speed droop affeds govemor operational response.
D Discriminant Validity Plausible misconception since speed droop affeds govemor operational response.
. TMI Lic:nsing Ex:mination Answer Kcy QUESTION: 045 '(1.00)
Current plant conditions are:
-Unit is operating at 100% power. t
, -Total loss of Nuclear Services Closed Cooling Water occurred 5 minutes ago.
-Attempts to start NS-P-1 A/B/C were unsuccessful.
Which ONE condition requires the operator to trip Reactor Coolant Pumps?
A. Motor stator temperature indication is 140 degrees C.
- B. Motor bearing temperature indication is 180 degrees F.
' 1 C. Motor thrust bearing temperature indication is 200 degrees F. i D. Seal #1 leak-off temperature indication is 220 degrees F. ;
! Answer: C 4 K/A: 000062 AA1.01 3.1 Page 4.2-49 i Objective: IV.A.05.24
.l l Reference (s): OP 1103-6 Reactor Coolant Pump Operation, Rev 63, Page 9 History: New NRC Cognitive Level Rating 2 Measurement #45 Ability to monitor temperatures of components cooled by Nuclear Services Closed Cooling System following totalloss of the system.
A Discriminant Validity Plausible distracter since this parameter does have an upper limit that requires the RCPs to be shutdown. This value does not yet violate the maximum limit established in procedures. l B Discriminant Validity Plausible distracter since this parameter does have an upper limit that '
requires the RCPs to be shutdown. This value does not yet violate the maximum limit established in procedures.
C Discriminant Validity Correct answer. l D Discriminant Validity Plausible distracter since this parameter does have an upper limit that requires the RCPs to be shutdown. This value does not yet violate the maximum limit established in procedures.
TMI Lirnsing Ex:minatirn
- Answer K;y i QUESTION: 046 (1.00)
- ' An unexpected illness of a CRO has occurred. With action being taken to mrrect the situation,
' whis ONE condition describes the maximum time the shift crew may remain below minimum staffing requirements?
A. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with RCS Taw < 200*F B. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at all RCS temperatures C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RCS Taw > 200*F D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if during shift relief regardless of RCS temperature i
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Answer: B i 1
K/A- Generic 2.1.4 3.4 Page 2-1 Objective: V.F.01.08 "Given specific plant conditions and a copy of Tech Specs, identify any LCOs which are challenged by such conditions, and determine remedial actions permitted or required by those LCOs."
Reference (s): Technical Specifications, Table 6.2.1, Page 6-2, Amendment No.149. Note iii History: New l
NRC Cognitive Level Rating 1 Measurement N6 Application of Plant's Administrative Technical Specircations durire abnormal situation related to minimum shift manning requirements.
A Disedminant Validity Plausible distracter since it does allow the condition, but for an incorrect time limit.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible since it allows a time limit, but is incorrectly restricted by RCS temperature, which is the basis for changing minimum manning requirements.
D Discriminant Validity Plausible since it does not disallow the situation, but incorrectly states the time restriction. This distracter also incorrectly allows the crew to go below minimum staf> ag requirements during shift tumover- specifically disallowed by Tech Specs in note lii.
I TMI Lixnsing Extminatien An2wer K y QUESTION: 047 (1.00)
Current plant conditions are:
Reactor is at hot shutdown condition.
All RCPs are operating.
Plant electrical configuration is normal for power operating conditions.
MU-P-1B is operating. j CW-P 1 A and CW-P-1C are not operating.
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A tagging request requires 1E 125/250 VDC Bus to be removed from service. Which ONE l statement describes impact of de-energizing 1E 125/250 VDC Bus?
A. MU-P-1 A will not start manually (from the Control Room) or automatically.
B. MU-P-1C will not start manually (from the Control Room) or automatically.
C. CW-P-1 A will not start manually (from the Control Room).
D. CW-P-1C will not start manually (from the Control Room).
Answer: A K/A: 000063 K2.01 3.1 Page 3.6-6 Objective: IV.G.10.02 Reference (s): E-206-051 Rev 21 DC Distribution One-Line Diagram History- New l
NRC Cognitive Level Rating i j Measurement M7 Demonstration of knowledge conceming bus power supplies to major DC loads.
l A Discriminant Validity Correct answer.
B Discriminant Validity Plausible distracter since this pump would not be capable of starting from the control room or automatically by ESAS actuation due to loss of breaker DC control power. DC control power to this pump orlainates from 1F DC Bus.
C Discriminant Validity Plausible distracter since this pump would not be capable of starting from the control room due to loss of breaker DC control power. DC control power to this pump originates from a different BOP DC Bus. This pump has no automatic start features.
D Discriminant Validity Plausible distracter since this pump would not be capable of starting from the ;
control room due to loss of breaker DC control power. DC control power to this pump originates from a different BOP DC Bus. This pump has no automatic start features.
TMI Lic:n:ing Ex min tien l An:wer K:y QUESTION 048 (1.00) ,
Current plant conditions are:
- Time is ten minutes after reactor trip due to loss of both Main Feedwater Pumps.
- EF-P-2A is operating, EF-P-1 and EF-P-2B are not operating.
.TH is 585*F and slowly increasing.
- RCS pressure is 2300 psig and slowly increasing.
. All RCPs are operating.
- OTSG 1 A level is 25 inches.
- OTSG 18 levelis 0 inches.
- OTSG 1 A pressure is stable at 1010 psig.
- OTSG 1B pressure is 800 psig and decreasing.
- RCS heat up rate is +75 F/Hr.
Whch ONE action is required concerning operation of the RCPs?
A. Stop 1 RCP per loop.
B. Continue to operate 4 RCPs.
C. Continue to operate only 1 RCP.
D. Stop 4 RCPs.
.' Answer: A K/A- BW/EO4 EK2.2 4.2 Page 4.3-7 1
Objective: V.E.04.07 Reference (s): Abnormal Transient Procedure,1210-4, Lack of Primary to Secondary Heat Transfer, Page 2.0, Rev 10 History: TMI Exam Bank Question OR5E04-07-Q-11 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 3
TMI Licen:Ing Ex2 min 2 tion An2w r K2y Measurement MS Demonstration of ability to apply knowledge of interrelationships between heat removal systems during conditions involving inadequate core heat transfer.
$ A Descdminant Validity Correct answer.
8 Discriminant Validity Plausible distrader since this maintains maximum possible flow through the reactor core during conditions of inadequate heat transfer.
C Discriminant Validity Plausible distrader since this reduces heat input (by the RCPs) into the RCS
- during conditions of inadequate heat transfer while still maintaining forced l flow through the reador core - which is the ' preferred method of core cooling.'
D Discriminant Validity Plausible distrader since this maximizes the reduction of heat input (by the RCPs) into the RCS during conditions of inadequate heat transfer.
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TMI Lic n:ing Ex minati:n i
Answer Kcy QUESTION: 049 (1.00)
To support plant start-up, the valve line-up of a system is required to be modifiexi until work is completed on that system. Which position (s) are required to approve this change in the
- valve line up?
. A. Shift Supervisor !
l B. One CRO and one Shift Supervisor / Shift Foreman with an SRO license !
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C. Two CROs 4
.: D. Plant Operations Director !
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.t Answer- B K/A: Generic K2.2.11 3.4 Page 2-6
. Objective: V.A.02.04 l 1
Reference (s): Administrative Procedure 1001 A, Procedure Review and Approval, Rev 36, Page 18.0 History: TMI Exam Bank Question AL5A02-04-QO1 Verified not used on audit exam or in SRO Program quizzes.
Very similar question was used on TMl License Exam dated 8/31/93 (Q #10).
.NRC Cognitive Level Rating 1 Measurement #49 Application of station administrative requirements during abnormal operations.
A Discriminant Validity Plausible since the SS is the senior shift management representative responsible for all operations.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since this incorred answer involves hetg individuals concurring independentiv.
D Discriminant Validity Piausible distracter since this senior management individual is SRO licensed by the Nuclear Regulatory Commission.
.TMl Licen:ing Examin_tien Answer Kcy QUESTION: 050 (1.00)
A Tech Spec Surveillance is satisfactorily completed upon retuming a plant component to service that has a Regulatory Retest tag. It is now permissible to remove the Regulatory Retest tag. Who is responsible in ensure the Retest Tag book is properly closed out?
A.~ CRO who completed the test. l B. Resoonsible System Engineer witnessing the test. 1 C. IST Coordinator D. Shift Foreman Answer: D K/A: Generic K2.2.21 3.5 Page 2-7 l
Objective: V.A.16.01 Reference (s): AP 1001J Technical Specification Surveillance Testing Rev 18, Page 12 History: TMI Exam Bank Question QR5A16-01-QO1 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #60 Knowledge of pre- and post maintenance operability requirements.
A Discriminant Validity Plausible misconception since the CRO is involved in the testin0, and is licensed by the Nuclear Regulatory Commission.
B Discriminant Validity Plausible misconception conceming duties, responsibilities and authority of the System Engineer.
C Discriminant Validity Plausible misconception conceming duties, responsibilities and authority of the IST Coordinator.
D Discriminant Validity Correct answer.
TMl Liscnsing Extminsti n ;
Answer KOy j QUESTION: 051- (1.00)
Which ONE Abnormal Transient procedure has the highest pnonty during an emergency situation? !
l A. Excessive Primary to Secondary Heat Transfer !
l B. Lack of Primary to Secondary Heat Transfer l
C. Loss of 25'F Subcooling Margin D. Steam GeneratorTube Leak Answer: C K/A: Generic 2.4,6 4.0 Page 2-11 Objective: .V.E.11.13 Reference (s): Abnormal Transient Procedure,1210-1, Reactor Trip, Rev 37 Page 4.0 History.. New l NRC Cognitive Level Rating 1 Measurement #61 Evaluation and prioritization of emergency conditions and mitigation strategies as applied in symptom based emerDency procedures.
A Discriminant Validity Plausible misconception regarding priorttizabon of accident miteation strategies. This is number 2 behind loss of subcooling.
B Discriminant Validity Plausible misconception regarding prioritization of accident megation strategies. This is #3.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception regarding prioritization of accident mitigation strategies This is #4.
TMl Lic nsing Ex min; tion-Answer K0y QUESTION: 052 -(1.00)
Which ONE statement describes the reason for manually tripping or venfying turbine trip in the immediale Adions of ATP 1210-1 Reactor Trip?
A. Ensures OTSGs are no longer cross connected through the Main Steam lines.
B. Reduces Main Feedwater flow requirements.
C. Minimizes steam generator tube-to-shell delta-T.
D. Prevents an uncontrolled RCS cooldown l
i Answer- ' D l
K/A: BW/A04 AK2.2 3.5 Page 4.3-32 i I
Objective: V.E.01.07 Reference (s): EOP Basis Document History: .New NRC Cognitive Level Rating 2 Measurement #52 Analysis erd 5)athesis of relations of proper operations of heat removal systems t ; proper operation of the facility.
A Discriminant Validity Plausible misconception since this is a true statement since Main Turbine Stop Valves close on a Turbine trip.
8 Discriminant Validity Plausible, since tripping the turbine reduces steam flow and therefore Feedwater flow requirements.
C Discriminant Validity Plausible since tripping turbine results in steam temperature reduction towards RCS Tave.
D Discriminant Validity Correct answer.
TMI Licensing Examination Answer K0y QUESTION: 053 (1.00)
Current plant conditions are:
- The reactor is operating at 50% power with rod control in automatic.
- Uncontrolled withdrawal of group 7 control rods is occurring.
- Control rod withdrawal command does not exist.
Which ONE statement describes required action for these conditions?
A. Select group 7and turn the Single Select switch to ALL.
B. Dispatch an operator to pull the fuses on group 7 programmer motor.
C. Depress the In-Limit-Bypass pushbutton and attempt to insert group 7.
- D. Select Sequence Override at the CRD operator's console.
l Answer: D l
K/A: 000001 AA1.06 2.9 Page 4.2-3 I i - Objective: V.D.03.02 Reference (s): Emergency Procedure,1202-8, CRD Equipment Failure, Rev 50, Page 22 History: .TMl Exam Bank Question SR5D03-02-QO3 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 3 Measurement #63 l Ability to determine actions for uncontrolled rod withdrawal.
A Discriminant Validity Piausible misconception resulting from inverting immediate Manual Action number b.2 in EP 1202-8.
8 Discriminant Validity Piausible since this action would stop rod motion by de-energizing the control rods in that specific Group.
l C Discriminant Validity Piausible since the action to bypass the Group Inlimit is required to be taken
! to insert control rods durino response to a dropped control rod.
l D Discriminant Validity Correct answer.
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TMl Licen:ing Examin: tion Answer Kcy s QUESTION: 054 (1.00)
Which ONE condition requires initiation of HPl cooling according to the Abnormal Transient procedures?
A. Pressunzer level can not be maintained greater than 200 inches with the reactor-at 100% power.
B. - RCS subcooling is 30*F.
C. Post trip RCS pressure is 1750 psig with incore exit thermocouples irdcating 560 degrees.
D. Neither OTSG is available as a heat sink.
Answer D K/A: 000006 K1.07 3.3 Page 3.2-16 Objective: V.E.04.08 Reference (s): Abnormal Transient Procedure 1210-10, Abnormal Transients Rules Guides and Graphs, Rev 34, Page 4.0 History: New NRC Cognitive Level Rating 1 Measurement s54 Knowledge of physical cause-effect relationships between the Main Feedwater system and the Emergency Core Cooling. Systems.
A Discriminant Validity Plausible, since 200* is the Pressurizer low level alarm setpoint used during operatino conditions.
8 Discriminant Validity Plausible since loss of subcoolin0 margin (<25 degrees) is a correct application of this rule.
C Discriminant Validity Plausible since reduction of RCS pressure reduces Subcooling Margin.
However, subcooling margin is approximately 58 degrees under these conditions. This is above the HPl Initiation Criteria of less than 25 doorees.
D Discriminant Validity Correct answer.
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TMI Licensing Ex! min: tion Answ r K0y QUESTION: 055 (1.00)
Completion of the Containment integrity Checklist is in progress in preparation to begin refueling operations Which ONE condit'on prevents start of refueling operations?
A. RB purge is in progress.
B. MU-V-25 and MU-V-26 are physically removed for rebuilding.
C. Service Airis in use inside the RB.
D. One door of the RB equipment hatch is open Answer B K/A- 000069 AA2.01 4.3 Page 4.2-57 Objective: V.B.01.05 Reference (s): Tech Spec 3.8.7 Amendment 198, Page 3-44 History: New NRC Cognitive Level Rating 1 Measurement #55 Knowledge ofInte:Telations between t.oss of Containment integrity and Valves.
A Discriminant Validity Plausible misconception that purge valves need to be closed in order to meet containment integrity requirements. Tech Specs allows RB Purge during conditions requiring Containment Integrity as long as specified operability requirements are met.
B Discriminant Validity Correct answer. l C Discriminant Validity Plausible misconception that this violates containment integrity requirements. l D Discriminant Validity Plausible misconception that this violates containment integrity requirements.
l TMl Lic:n:ing Excminction i Answer Kcy QUESTION: 056 (1.00)
A point source in the Auxiliary Building is reading 500 mrem /hr at 2 feet. Two options exist to complete a mandatory task near this radiation source:
Option 1: Operator X can complete the task in 30 minutes working at a distance of l 4 feet from the point source l Option 2: Operators Y and Z, using a special extension tool can complete the same task in 75 minutes at a distance of 8 feet from the point source Which ONE statement describes personrul exposure for completion of this task in accordance with ALARA guidance?
A. Option 1 will expose Operator X to 62.5 mrem.
- 8. Option 1 will expose Operator X to 125 mrem.
C. Option 2 will expose each operator Y and Z to 39 mrem.
D. Option 2 will expose each operator Y and Z to 156 mrem.
l Answer: A l K/A: Generic 2.3.2 2.9 Page 2-9 Objective: lil.F.01.08 Ill.F.01.09 Reference (s): Rad Worker Training Handout Pages 15 and 16 History: New I NRC Cognitive Level Rating 2 Measurement #56 Demonstration and application of facility Al. ARA program implementation for work practices.
A Discriminant Validity Correct answer with accurate calculation and proper comparison of total exposures for both options.
B Discriminant Validity Plausible when 30 minute stay time is not considered in integrated exposure calculation.
C Discriminant Validity Plausible if two-person exposure is not considered.
D Discriminant Validity Plausible if does reduction is calculated linearly rather than using inverse square law.
i TMI Licen:Ing Excminiti:n An:wer Kiy
- QUESTION: 057 '(1.00) l Current plant conditions are:
- Reactoris operating at 20% power.
- CO-P-1 A and CO-P-2A are operating.
- One Main Feedwater Pump is operating.
- The following alarms are actuated simultaneously:
- A-1-8 Battery 1B Discharging
- A-2-8 Battery Charger 1B/1D/1F Trouble
- A-3-8 inverter 1B/1D Inverter System Trouble
- PRF 1-1-1 CRDM Bkr Test Trouble
- H&V A3-2-3 Cont. Bldg. Batt. Chargers B Damper Tbl. Fire-Smoke
- AA-3-2 7 KV Bus Trouble l
- AA-3-5 480V BOP Bus Trouble Which ONE action is required for this condition?
A. Close DC tie switches to provide an attemate source of DC power.
B. Open suction valve (VA-V-5B) for VA-P-18.
C. Reduce power to.within the reduced capability of the condensate system.
D. Transfer Alterex Excitation System DC power supply to B side DC power.
Answer. D K/A: 000058 AK3.02 4.2 Page 4.2-43 Objective: V.D.04.02 Reference (s): Emergency Procedure 1202-9A, Loss of "A" DC Distribution System, Rev 39 Page 4 History: TMl Question SR5D0442-QO2 (Medified)
Original was used on audit exam (#91).
Verified not used on in SRO Program quizzes.
NRC Cognitive Level Rating 3
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TMl Licensing Examination l An:wer Ksy l l
Measurement #67 Interpretation of plant conditions and applying knowledge of schons contained in emeroency operating procedures for loss of DC power.
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A Discriminant Validity Plausible since this action would restore power to A DC Distribution System. I i
However this improper adion would increase plant vulnerability to a total loss 1 i of all DC (A and B Destribution Systems).
B Discriminant Validity Plausible distrader since this valve automatically doses on loss of B DC
! Distribution (reference EP 1202-98 Rev 37 automatic action number 7.)
C Discriminant Validity Plausible since this is the first immediate manual adion for Loss of B DC i Distribution (reference EP 1202 9B Rev 37).
D Discriminant Validity Correct answer.
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TMl Licen:ing Excminttion Answer K y QUESTION: 058 (1.00) l An emergency event has been declared. Which ONE statement describes the maximum time limits for inetsal notification of the NRC, state, and local agences?
A. 15 minutes for NRC notification,15 minutes for state and local notifcations i i
B. 30 minutes for NRC notification,15 minutes for state and local notifcations C. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC notification,15 minutes for state and local notifications l D. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for NRC notification, 30 minutes for state and local notifications i
Answer: C t K/A: Generic K2.4.38 4.0 Page 2-15 Objective: Vll.D.06.07 Reference (s): EPIP-TMI .03, Emergency Notifications and Call Outs Rev 24, Page 5
, AP 1044 Event Review and Reporting Requirements Rev 38, Page 5 History: New j NRC Cognitive Level Rating 1 Measurement #58 Demonstration of ability to take adions called for in the fadlity emergency plan, induding acting as Emergency Diredor (time requirements for emerDency notifications).
A Discriminant Validity Plausible since it indudes correct time limit for State / Local notifications, but erroneously uses same time limit for NRC limit.
8 Discriminant Validity Plausible since it indudes correct time limit for State / Local nobfications - but not NRC.
j C Discriminant Validity Correct answer.
D Discriminant Validity Plausible since it includes correct time limit for NRC notification.
i l.
o
l TMI Litensing Examinsti:n Answ:r K:y QUESTION: 059 (1.00) j Which ONE condition will actuate the CRD PATTERN ASYMMETRIC annunciator?
A. Safety rods greater than 7 inches (5%) from group average position as
, determined by absolute position indication.
l l B. Safety or regulating rods greater than 9 inches (6.5%) from group average l position as determined by relative position indication.
C. Safety or regulating rods greater than 7 inches (5%) from group average position as determined by absolute positiori indication.
D. Regulating rods greater than 9 inches (6.5%) from group average position as l determined by relative position indication l
l Answer: C l K/A: 000014 K1.01 , 3.6 Page 3.1-21 Objective: IV.E.13.18 L Reference (s): Alarm Response Procedure, Main Annunciator Panel G, Rev 50, G-2-1 l
l History: New i
l NRC Cognitive Level Rating 1 Measurement #59 Knowledge of interrelationships between control rod position indication systems and the CRD system.
A Discriminant Validity Plausible since the actual setpoint is 7* from group average - but is not
- limited to Safety Rods.
l B Discriminant Validity Plausible since the TS definition of an asymmetric rod is at 9' deviation from l the group average. Also, a common misconception is that Relative Position l Ind; cation System is used for this function.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception that this alarm applies only to Regulating Rods.
Also common misconception is that Relative Position Indication system is used for this function.
l
[l d
1 1
TMI Lisensing Examination i Answer Key QUESTION: 061 (1.00)
ATP 1210-5, OTSG Tube La=kaga. Which ON 4
i-minimizing whing margin?
. A.
i Minimize RCS leakage through the leaking OTSG tube.
'B.
i' Minimize time required for cooldown of the RCS.
C.
Minimize potential of lifting Main Steam safety valvc 1
D.
4 Minimize tensile stresses on affected OTSG tubes.
i
- Answer
- A K/A- 000037 AK1.02 3.9 Page 4.2-29
- - Objective
- V.E.05.05 Reference (s):
Abnormal page 2 Transient Procedure,1210-5, OTSG Tube Leakage,
. History-
!. TMl Question SR5E054QO3 (Modified)
Original was used on audit exam (#96).
Verified not used on in SRO Program quizzes.
j 1 NRC Cognitive Level Rating 2 1
Measurement #61 Application of knowledge of operationalimplications re Denerator tube leak rate and tube differential pressure.garding steam A Discriminant Validity Correct answer.
8 Discriminant Validity C Discriminant Validity Plausible since depressurization of RCS is required during cooldown.
Plausible since lowering RCS pressure reduces maximum OTSG pres i
D Discriminant Validity affeded OTSG is filled to solid condition.
i Plausible since RCS pressure reduction reduces stress on affeded OTSG tubes 4
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l TMI Ll:ensing Excmination Answer Key QUESTION: 062 (1.00)
- Data contained on survey sheet:
- 2,500 DPM/100 cm 2. beta-gamma
- 10 DPM/100 cm2 alpha
- 450 mRom/hr general area
- 470 mrem /hr on the surface of a tank
- Airbome activity < 10% cf all DACs
!~ Which ONE posting is required for these conditions?
A.. Contaminated /High Radiation Area B. Contaminated / Radiation Area
~C. High Radiation Area (only) i D. Airborne Radiation Area (only) i l
I Answer: A K/A: Generic 2.3.1 3.0 Page 2-9 l
Objective: Ill.F.02.03.
Reference (s): RWP Worker Training Handout Pages 57 and 58 History:- New l l
l NRC Cognitive Level Rating 1 l^ Measurement #62 Knowledge of 10 CFR 20and facility radiation control requirements.
- A Discriminant Validity Correct answer.
j- B Discriminant Validity Plausible distracter since this is a Contaminated Area, but also a High l
- l. Radiation Area.
l C Discriminant Validity Plausible distracter due to misconception on contamination limits.
D Discriminant Validity Plausible distracter due to misconception on airbome limits or definition of DAC.
l l
TMI Lic:nsing Ex:mination An:w:.r K y QUESTION: 063 (1.00)
Current plant conditions are:
- Numerous fire alarms are actuated on panels HVB, PLA, and PLB,
- Fire dampers AH-D-4 and AH-D-5 have automatically closed.
- Air tunnel deluge systems have actuated.
- Air tunnel halon system has actuated.
- Fire system indicated water pressure is 50 psig and decreasing.
- Aux and Fuel Handling Building ventilation supply fans have tripped.
Which ONE statement describes the action (s) required for this condition?
A. Trip Aux and Fuel Handling Building ventilation exhaust fans.
B. Actuate Relay Room CO2 system manually.
C. Trip reactor.
~
D. Start three fire pumps.
4 s
a i
Answer: D l
- K/A
- 000067 AK3.04 4.1 Page 4.2-52 Objective: V.D.13.03 Reference (s): EP 1202-31 Fire, Rev 54, Pages 3 and 4 History: New
. NRC Cognitive Level Rating 3
[ Measurement #63 Applying actions contained in emergency operating procedures for fire on
- site.
J A Discriminant Validity Plausible distracter since this would terminate ventilation effeds on the fire.
] B Discriminant Validity Plausible distracter since this is a corred action for another fire scenario.
j C Discriminant Validity Plausible distracter since this is a correct action for a relay room fire scenario, and since Air intake Tunnel dampers have actuated..
- D Discriminant Validity Correct answer.
(
i 4
TMI Lic:nsing Exrmination An:wer Kcy QUESTION: 064 (1.00)
Auxiliary Operator erroneously initiates a liquid release from A WECST. The Release Permit is
- for B WECST. Wb':h ONE condition causes automatic termination the accidental release of radioactive liquid?
I A. RM-L-7 in ALERT alarm QB loss of sample flow through RM-L-6 B. RM-L 6 HIGH alarm QB high tank release rate C.' High MDCT emuent flow QB low tank release rate D. RM-L-6 ALERT alarm QB.RM-L-7 ALERT alarm l
i Answer: B K/A: 000059 - AA2.05 3.9 Page 4.2-45 Objective: IV.B.09.04 Reference (s): OPM, Waste Disposal Liquid System, E-02, Rev 11, Page 23 l History: TMI Exam Bank Question AL4809-04-QO2 (Modified) l Original was used on audit exam (#6).
l Verified not used on in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #64 Demonstration of ability to determine and interpret the occurrence of
- automatic safety functions as they apply to accidenta! radioactive liquid releases.
A Discriminant Validity Plausible distracter since RM-L-6 and RM-L 7 monitor liquid release flow paths. Misconception that loss of detector sample flow will terminate the radioactive liquid release or that an ALERT alarm will conservatively terminate a liquid release prior to reaching the HIGH alarm.
I ^8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible distracter which reverses high and low flow interlocks between tank release rate and effluent (dilution) flow rate.
I D Discriminant Validity Plausible misconception that an ALERT alarm will conservatively terminate a l liquid release prior to reaching the HIGH alarm.
i TMI Licensing Examin: tion Anawer K0y QUESTION: 065 (1.00)
{
Waste Gas Decay Tank relief valve WDG-V-36 has opened due to high tank pressure. This valve is now failed open. Which ONE statement desenbes automatic action (s) initiated by the Radiation Monitoring system related to this accidental gaseous release?
A. Trip AH-E-14NC (B/D)
B. Trip AH-E-10 AND AH-E-11 C. Trip AH-E-10 ONLY D. Trip AH-E-11 ONLY Answer: B K/A: 000060 EA2.05 4.2 Page 4.2-47 Objective: IV.E.06.02 Reference (s): C-302-841 Rev 46 Auxiliary and Fuel Handling Building Ventilation EP 1202-12 Excessive Radiation Levels Rev 42, Page 11.
History: New NRC Cognitive Lovel Rating 3 Measurement #65 Determine and interpret automatic actions required from high ARM system signal in order to limit accidental radioactive gas release to the public.
A Discriminant Validity Plausible misconception that reduction of ventilation supply flow will terminate or reduce caseous radwaste releases.
B Discriminant Validity Correct answer.
C Discriminant Validity Plausible distracter since this is the supply fan for the Fuel Handlina Building.
D Discriminant Validity Plausible distracter since this is the supply fan for the Auxiliary Building, where this accidental release originates.
- l. .
TMI Licensing Examination Answer Kcy QUESTION: 066 (1.00)
Which ONE condition does NOT meet the evaluation criteria for verification of Natural Circulation as stated in ATP-1210-10?
A. RCS Delta-T is +25*F and Tu is M.
B. Incore thermocouple temperatures are stable and are tracking Ts.
C. Co!d leg temperatures are approaching saturation temperature for secondary sioe pressure.
D. Heat removal from OTSGs has been verifieo by the presence of steam flow and feed flow indication.
Answer: A K/A: BW/EO9 EK21. 4.0 Page 4.3-15 Objective: V.E.10.01 Reference (s): ATP 1210-10, Abnormal Transients Rules, Guides, and Graphs, Rev. 34, Page 12 History: New NRC Cognitive Level Rating 1 Measurement #66 Knowledge of interrelationships between natural circulation and proper operation of the primary coolant system.
Note: This question is constructed as a negative question (containing three correct and one incorrect answer) to evaluais ability to analyze and identify plant conditions which do NOT meet procedural requirements.
A Discriminant Validity Correct answer. This statement does D21 meet the criteria contained in ATP 1210-10.
i B Discriminant Validity This statement meets one section (there are three) of ATP 1210-10 requirements for verifying RCS natural circulation heat removal.
C Discriminant Validity This statement meets one section (there are three) of ATP 1210-10 requirements for verifying RCS natural circulation heat removal.
D Discriminant Validity This statement meets one section (there are three) of ATP 1210-10 requirements for verifying RCS natural circulation heat removal.
TMI Lirnsing ExaminItion Answ;r Key QUESTION: 067 (1.00)
Sequence of events:
- Reactor is operating at 35% power.
- FW-P-1 A is in operation; FW-P-1B trip has not been reset.
- Steam line rupture occurs upstream of Main Steam isolation Valve MS-V1 A.
- The reactor and turbine trip.
- Main turbine Stop Valves 1-4 fail to close.
Which ONE statement describes why only one OTSG depressurizes as a direct result of the steam line break?
A. HSPS isolates Main Feedwater to OTSG 1 A.
B. MS-V-1 A and MS-V-1 B are stop check valves that prevent back flow from OTSG 18.
C. Closure of turbino Control Valves 1-4 results in separation of the two steam generators.
D. Automatic open command for EF-P-1 steam supply valve MS-V-138 is delayed for 40 seconds following MS-V-13A automatic operation.
Answer: B K/A- 000039 K1.01 3.2 Page 3.4-19 Objective: IV.C.01.03 Reference (s): C-302-011, Rev. 56, Main Steam History: New NRC Cognitive Level Rating 3 Measurement #67 Knowledge of physical connections and cause and effect relationships between the Main Steam System and the OTSGs.
A Discriminant Validity Plausible distracter since this is a true statement for this situation.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception that closure of Main Turbine Control Valves separates the two OTSGs.
D Discriminant Validity Plausible distracter since this is a true statement. MS-V-138 opening is delayed to allow MS-V-6 control prevent EF-P-1 overspeed, and to prevent steam pipe over-pressurization.
l
TMI Lic:nzing Ex:mination
!- Answer K y l QUESTlON: 060 . (1.00) j Current plant conditions are:
- A reactor trip has occurred.
1 - Three control rods failed to fully insert into the core.
- Power Range Nis are off scale low.
Which ONE statement describes the required action (s) for this condition?
3-A. Emergency borate the RCS.
- B. Maintain primary to secondary heat transfer.
4 C. Deenergize 1G and il480 volt buses.
l D. Perform immediate actions of EP 1202-8, CRD Equipment Failure for stuck rods.
4
- Answer: A K/A- 000024 AK3.01 4.4 Page 4.2-14
. Objective: V.E.01.03 Reference (s): ATP 1210-1 Reactor Trip Rev 37 Page 2
)
History: New NRC Cognitive Level Rating 1 Measurement #68 Demonstration of knowledge of requirements for emergency boration of the reactor core when rods fail to insert on a reactor trip.
A Discriminant Validity Correct answer.
8 Discriminant Validity Plausible since this is the corted action if reactor power is greater than 10%.
C Discriminant Validity Plausible misconception, since this W.ag a valid action for an ATWS prior to installation of the Diverse Scram System.
D Discriminant Validity Plausible since EP 1202-8 has a special section devoted to responding to stuck control rods.
1
TMl Lic:nsing Examination Answer Key
[ QUESTION: 069 (1.00)
Current plant conditions are:
Reactoris operating at 100% power.
- 6 CW Pumps are operahng
- Winter conditions exist.
Which ONE statement desenbes when the CRO is required to manually trip the turbine?
- Assume no automatic reactor trip.
A. Loss of one CW Pump B. Condenser pressure is 8.7 inches Hg absolute C. Loss of the Gland Steam Exhauster D. Condenscr pressure is 7.7 inches Hg absolute Answer: B K/A: 000051 AA2.02 4.1 Page 4.2-34 Objective: IV.D.04.12 Reference (s): OPM, Main Turbine, H-01, Page 46, Rev 12 History: New NRC Cognitive Level Rating 1 Measurement #69 Ability to determine and interpret conditions which require reador/ turbine trip under loss of condenser vacuum conditions.
A Discriminant Validity Plausible distracter considering summer time operations. Question stem establishes winter conditions, when we often run with 4 pumps. +
B Die:riminant Validity Correct answer. Automatic setpoint is 8.5" Hg Absolute if >75% power,7.5" Hg Absolute if <75% power.
C Discriminant Validity Plausible distracter due to connections with turbine Gland Seal system.
D Discriminant Validity Plausible misconception that setpoint is 7.5" rather than 8.5" Hg Absolute (powerlevel dependent).
_ _ . . . - .- . -.. _ .. - .- - . - - . . . = - _ . . -
TMl Licensing Ex mination Answ r K;y QUESTION: 070 (1.00)
Current plant conditions are: l 1
Reactor is operating at 75% power.
RC-P-1C is shutdown due to high motor winding temperatures.
Entry into the RB is required to venfy the NSCCW valve line up for RC-P-1C. )
- RB purge must be initiated prior to the entry.
- D-Rings will not be entered.
N ONE position is authonzed to approve this entry into the RB7 A. Director of Operations and Maintenance B. Plant Operations Director C. Shift Supervisor D. Group Rad Con Supervisor Answer. B K/A: Generic 2.3.10 3.3 Page 2-10 Objective: V.A.21.02 Reference (s): Radiological Controls Procedure 6610-ADM-4110.09, Page 6.0, Rev.8
' History: New NRC Cognitive Level Rating 3 Measurement #70 Ability to perform procedures to reduce excessive levels of radiation and to guard against personnel exposure.
A Discriminant Validity Plausible misconception since the Director of Operations and Maintenance is a senior management official.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible misconception since the Shift Supervisor has responsibility for overall safe operation of the plant and personnel.
D Discriminant Validity Plausible rmsconception since entry into the reador building involves personnel exposure.
TMl Lictnzing ExcminIti::n I Answ r Ksy QUESTION: 071 i Which ONE statement does NOT satisfy the Superheat Determination / Limit Rule as defined in i 4
ATP-1210-107 '
)'
A. f6 F d superheat as determined by the most conservative d the two subcooling aargin meters on panel PCL i
B 25 F of superheat as determined by the plant computer l C. 25 F of superheat as determined by the average of 5 highest operable incore )
thermocouples and RCS wide range pressure D. 25 F of superheat as determined by the highest operable BIRO incore thermocouple and RCS wide range pressure j
Answer A K/A: 000017 A4.02 4.1 Page 3.7-12 Objective: V.E.10.01 l Reference (s): ATP 1210-10 Abnormal Transient Rules, Guides, and Graphs, Rev. 34, Page 6
History: New NRC Cognitive Level Rating 1 Measurement #71 Use procedures to direct, control or mitigate the consequences of an accident by monitoring the incore temperature monitoring systems to determine the existence ofinadequate core cooling.
Note: This question is constructed as a negative question (containing three correct and one incorrect 1 answer) to evaluate ability to analyze and identify plant conditions which do NOT meet procedural l requirements.
A Discriminant Validity Correct answer. This answer does not meet requirements of ATP 1210-10.
B Discriminant Validity This answer meets requirements for one of three methods to determine existence of superheat in the RCS IAW ATP 1210-10. 1 C Discriminant Validity This answer meets requirements for one of three methods to determine !
existence of superheat in the RCS IAW ATP 1210-10. !
D Discriminant Validity This answer meets requirements for one of three methods to determine '
existence of superheat in the RCS LAW ATP 1210-10.
TMl LI:ensing Ex mination An:wer Key QUESTION: 072 (1.00)
Current plant conditions are: !
Refueling operations are in progress.
While lowering a spent fuel assembly into the upender, the assembly drops into the upender !
basket. The fuel handling bridge operator suspects the fuel assembly is damaged.
Which ONE set of statements describes Refueling Supervisor actions required for this event?
A. Stop all fuel movement. Further handling of the damaged assembly is allowed only under your supervision with Core Load Engineer concurrence B. Direct the dropped fuel assembly to remain in the basket. Continue refueling l operations using the altemate transfer mechanism.
C. Stop all fuel movement. Further handling of any fuel is allowed after obtaining approval from the Director O&M. l D. Direct the dropped fuel assembly to be transferred out of the RB to the Spent Fuel Pool. Continue refueling operations using the alternate transfer mechanism.
Answer: A ,
l K/A: 000034 A2.01 4.4 Page 3.8-13 Objective: V.B.16.04 Reference (s): RP 1503-2, Damaged Fuel and Control Components, Rev. 9, Page 9 History: New NRC Cognitive Level Rating 1 Measurement #72 Use procedures to correct, control or mitigate the consequences of a dropped fuel element.
A Discriminant Validity Correct answer.
B Discriminant Validity Plausible distrader uses attemais cystems while damaged assembly remains in place. Question stem does not address rad conditions to require RB evacuation.
C Discriminant Validity Plausible distracter since the Director of Operations and Maintenance is a senior management official.
D Discriminant Validity Plausible distracter since this action would remove the damaged assembly from the containment building to a secure stora0e location in the Spent Fuel l Pool. j
TMI Liscnzing Excminsticn Answer Kcy
- QUESTION
- 073 (1.00) l Initial Plant Conditions Plant is in the Cold Shutdown condition.
l RB purge is in progress.
- A radioactive liquid release is in progress from A WECST.
- RM-A-4 (FH Building exhaust monitor) interlock control switch is in Defeat for l&C calibration surveillance i Twenty gallons of highly radioactive liquid is accidentally spilled in the Fuel Handling Building.
Which ONE statement desenbes required automatic action (s) related directly to this spill incident?
A. AH-V-1NB/C/D close to terminate RB purge.
B. AH-E-14A/C (B/D) trip to terminate exhaust ventilation flow.
C. AH-E-10 and AH-E-11 trip to terminate supply ventilation flow.
D. WDL-V-257 closes to terminate liquid release.
Answer: C K/A: 000072 K4.02 3.4 Page 3.7-13 Objective: IV.F.04.04 Reference (s): EP 1202-12 High Radiation Levels Rev 42 Page 11 History- New NRC Cognitive Level Rating 3 Measurement #73 Demonstration of ability to apply operational knowledge of plant design and Radiation Monitoring System features and interlocks that provide Fuel Handling Building isolation.
A Discriminant Validity Plausible distracter since RB PurDe is in progress in the question stem.
B Discriminant Validity Plausible misconception, since this would decrease ventilation flow in the l area of the spill.
! C Discriminant Validity Correct answer.
D Discriminant Validity Plausible distracter since liquid release is in progress in the question stem.
i l
l
k TMILic:nzing Ex min tirn An:wer Kcy QUESTION: 074 (1.00)
Current plant conditions are:
- Tube leak exists on OTSG 1 A.
- Plant cooldown is in progress per ATP 1210-5.
2
- BWST levol is 48 feet.
- Tw is 500*F.
- RCS pressure is 980 psig.
Which ONE statement describes the reason why OTSG 1 A should NOT be isolated?
1 A. Buildup of radioactive water in OTSG 1 A t
B. Extension of time required to reach Cold Shutdown C. Increased probability of lifting Main Steam safety valve D. Isolation of one source of steam to EF-P-1 l
Answer: B I K/A- 000038 EA2.01 4.7 Page 4.1-11 Objective: V.E.05.04 Reference (s): EOP Tech Basis History: TMI Question QR5E05-04-QO1 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 2 Measurement #74 Evaluation and application of criteria pertaining to OTSG isolation during tube leak conditions.
A Discriminant Validity Plausible since A OTSG secondary mass inventory would increase due to continued P-S leakage following termination of steaming operations.
B Discriminant Validity Correct answer.
C Discriminant Validity. Plausible distracter since OTSG pressure would increase towards 950 ps%.
First MSSV would open at 1040 psig, assuming proper operation..
D Discriminant Validity Plausible distrader since isolation of OTSG 1 A closes one steam supply to EF-P-1.
l
. TMI Lis nsing Extminati::n
- An
- wcr K0y i 1
- QUESTION
- 075 (1.00) i 5 Current plant conditions are:
Refueling operations are in progress.
FTC water level is 24 feet above the fuel assemblies.
BWST is drained to 5 feet.
e -
RCS temperature is 90 degrees.
2 Loop A DHR is in service.
j -
Repair work is in progress on OTSG 1 A and OTSG 18.
l Which ONE statement describes Tech Spec requirements for the DHR System during these conditions?
A. One DHR string is required to be operable; it is required to be operating.
B. Two DHR strings are required to be operable; one is required to be operating.
C. One DHR string is required to be operable; it is not required to be operating.
D. Two DHR strings are required to be operable; neither is required to be operating.
1 Answer: C K/A: 000005 K4.02 3.5 Page 3.4-11 Objective: IV.A.11.17 Reference (s): Technical Specification, Section 3.4.2.1 & 3.4.2.3.a, Amendment 133, Pages 3-26 & 3-26A History: New NRC Cognitive Level Rating 1 Measurement #75 Knowledge of Decay Heat Removal System required modes of operation as described in plant Technical Specifications.
A Discriminant Validity Plausible misconception that the system must be in operation under these conditions. Although restrided by Tech Specs, there are conditions when system shutdown is acceptable.
8 Discriminant Validity Plausible misconceptions regarding requirement for redundant system and operational status requirements.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconceptions regarding requirement for redundant system.
TMI Lic:nxing Ex:minsti:n Answer K0y QUESTION: 076 (1.00)
Current plant conditions are:
- Alargo break LOCA has occurred.
- RCS pressure is 30 psig.
- Operator actions have been performed up to this point in accordance with ATPs.
- BWST levelis 6 ft. 4 inches.
Which ONE statement desa1bes the required sequence of operator actions to be performed at this BWST level?
A. Open the suction of the DH and BS pumps from the RB Sump (DH-V-62).
B. Close the suction of the DH pumps from the NaOH Tank (BS-V-2NB), then open DH pump suction from the RB Sump (DH-V-6A/B).
C. Open DH and BS pump suction from the RB Sump (DH-V 6NB), then close BWST outlet valves (DH-V-5NB), then close NaOH tank Outlet Valves (BS-V-2NB).
D. Verify open DH and BS pump suctions from the RB sump (DH-V-6A/B), then close BWST outlet valves (DH-V-5NB), then close NaOH Tank Outlet Valves (BS-V-2NB.
Answer: D K/A: 000011 EA1.11 4.2 Page 4.1-7 Objective: IV.A.11.15 Reference (s): Abnormal Transient Procedure,1210-7, Large Break LOCA, Rev 25 ;
Page 4 >
History: New NRC Cognitive Level Rating 1 Measurement #76 Ability to operate and monitor long term cooling of core during LOCA Cooldown. Adual performance of this task is time and sequence critical to the protedion of safety equipment and the adual core.
A Discriminant Validity Plausible distracter, since these are the correct actions to be taken at 9'6*,
These are incorrect steps for 6'4' as induded in the question stem.
8 Discriminant Validity Plausible distrader since steps are correct but out of sequence.
C Discriminant Validity. Plausible distrader, since this is the correct sequence, but DH-V-6A/B will be open already as specified in the procedure when BWST level drops to 9.5 feet.
D Discriminant Validity l Correct answer.
l TMI Lic:nsing Ex mination
! An:wer K y l QUESTION: 077 (1.00) l The reactor is operating at 100% power. Which ONE statement explains the reason for entering a Tech Spec time clock? i A. BS-V-1 A does not open automatically during ESAS testing. !
l B. Sodium Hydroxide Storage Tank NaOH concentration is 10.4%.
C. BS-V-49A (BS-V-2A inlet isolation valve) is not locked closed.
D. Sodium Hydroxide Storage Tank level is 8 ft. 2 inches lower than BWST level.
Answer: A K/A: Generic 2.1.12 4.0 Page 2-2 Objective: IV.A.15.07 "Given Tech Specs, determine limits placed on BS pump o,eration.
Reference (s): Technical Specifications, Section 3.3.1.3a, Amendment 190, Page 3-22 i History- New NRC Cognitive Level Rating 2 L
l l Measurement #77 l Ability to apply Technical Specifications for a system.
I A Discriminant Validity Correct answer.
l B Discriminant Validity Plausible distracter since this tank has Tech Spec NaOH concentration limit.
i C Discriminant Validity Plausible misconception that these valves are normally locked closed.
D Discriminant Validity Plausible distracter since there are Tech Spec restrictions on level l differential between these two tanks.
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TMI Lic:nzing Ex: min:ticn Antw:r K y ;
QUESTION: 078 (1.00)
Current plant conditions are:
- Reactor startup (approach to critical) is in progress, i
- Control rods reach upper ECP limit prior to achieving criticality. I Which ONE statement describes requirements for this condition?
A. Commence immediate boration to achieve 1% Ak/k subcritical condition; reactor l startup may not be continued until authorized by Nuclear Engineering.
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B. Stop rod withdrawal and recalculate entical boron concentration; commence <
I l slow controlled dilution to achieve criticality.
C. Trip the reactor; re-commence reactor startup after authorized by Plant Operations Director.
D. Insert rods to achieve at least 1% Ak/k subcritical condition; notify Nuclear Engineering to evaluate ECP conditions poor to re-commencing startup.
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Answer: D K/A: 000001 A2.12 4.2 Page 3.1-9 Objective: V.C.04.03 Reference (s): OP 1103-8, Approach to Criticality, Rev 44, page 8 History- New l NRC Cognitive Level Rating 1 l Measurement #78 Ability to mitigate consequences of erroneous reactor startup ECP 1 calculation.
A Discriminant Validity Plausible since this is conservative action, but incorrect, and not in accordance with established procedures.
8 Discriminant Validity Plausible since rod withdrawal limit would not be exceeded, but this is non-conservative and constitutes a violation of procedures. /
C Discriminant Validity Plausible but incorrect action. This is conservative action, but not prescribed procedurally for this situation.
D Discriminant Validity Correct answer.
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TMI LI:ensing Excminati:n Answer K;y QUESTION: 079 (1.00)
Which ONE mndition will DIRECTLY close the letdown isolation valve, MU-V-37 A. High letdown temperature B. 30# RB pressure ESAS signal C. Reactor Trip Isolation D. High Makeup demineralizer D/P Answer: A K/A: 000004 K4.09 3.1 Page 3.1-13 1 l
Objective: IV.A.09.11 Reference (s): OPM B-05, Makeup and Purification, Page 13, Rev 15 History: TMl Exam Bank Question AL4A09-11-QO3 Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #7s Knowledge of RCS letdown high temperature interlock designed to protect ion exchange resins.
A Discriminant Validity Correct answer.
B Discriminant Validity Plausible misconception that this valve closes on containment building Hi-Hi pressure isolation signal.
C Discriminant Validity Plausible misconception since this was true in the cast.
D Discriminant Validity Plausible distracter since the Makeup demineralizer is downstream of this valve.
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TMI Licensing Examination Answer K:y QUESTION: 080 (1.00)
Which ONE document is NOT required to be reviewed prior to assuming shift duties as the licensed Shift Foreman?
A. ESAS Checklist B. Locked Valve Book C. TCN/STP Book !
i D. Revision Review Book !
l Answer. B K/A- Generic - K2.1.3 3.4 Page 2-1 Objective: V.A.06.04 Reference (s): AP 1012, Shift Relief and Log Entries, Rev 43, Pages 9 and 10 History- TMl Exam Bank Question QRSA06-04-QO3 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #80 l Knowledge of shift tumover practices.
A Discriminant Validity - Plausible misconception - this document is required to be reviewed onor to assuming the duty. .
8 Discriminant Validity Correct answer. l C Discriminant Validity Plausible misconception - this document is required to be reviewed prior to assumhg the duty.
D Dsscriminant Validity Plausible misconception - this document is required to be reviewed gigtig assuming the duty.
TMl Licensing Examination
, Answer K:y i
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i QUESTION: 081 (1.00)
Which ONE statement describes a function of the Technical Support Center (TSC)?
A. Perform off-site dose prediction calculations.
B. Approve official press releases.
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! C. Notify off-site agencies for event re<:iassification.
D. Perform backup RCS leakrate calculations.
Answer. D l 'K/A: Generic K2.4.42 3.7 Page 2-15 l
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. Objective: Vll.D.06.06 Reference (s): EPIP-TMI .28 Activation of the Technical Support Center Rev 8, Page 2 History: New NRC Cognitive Level Rating 1 Measurement #41 l Knowledge of emergency response facilities.
A Discriminant validity Pleusible misconception since this is an engineering group.
I B Discriminant Validity Plausible misconception that this function can be performed by this group since they are located outside the Control Room.
! C Discriminant Validity Plausible misconception that this function can be performed by this group
- since they are located outside the Control Room.
l D Discriminant Validity Correct answer.
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TMI Lic:n::Ing Extmin tien An:;wIr K y QUESTION: 082 (1.00)
Which ONE statement explains the reason (s) EFW is injected into the OTSG near the top of the tube bundle?
A. Reduce the thermal stress on the lower tubesheet since the upper tubesheet can withstand a higher thermal stress than the lower tube sheet.
B. Reduce the thermal stress on the lower tubesheet and elevate the thermal center of the OTSG.
C. Reduce the thermal stress on the upper shell (steam exit region) and elevate the thermal center of the OTSG D. Reduce the thermal stress on the upper tubesheet and MFW nozzles.
Answer: B K/A- 000035 A4.05 4.0 Page 3.4-16 Objective: IV.A.08.11 Reference (s): OPM, Once through Steam Generator, B-04, Rev 8, Page 10 History: New NRC Cognitive Level Rating 2 Measurement #82 Steam Generator System. Ability to monitor and operate steam generator level and control to enhance natural circulation coolino of the reactor core.
A Discriminant Validity This correctly addresses reduction of stress on the lower tube sheet, but it includes the plausible misconception that the upper tube sheet can withstand Greater stress than the lower tube sheet.
B Discriminant Validity Correct answer.
Includes considerations for natural circulation heat removal and stress considerations due to cold water effects on lower tubesheet.
C Discriminant Validity This is the correct reason for EFW injection high in the tube bundle, but it reduces stresses in the lower shell ra*her than the upper OTSG shell.
D Discriminant Validity Plausible distracter since MFW nozzles are not used for cold EFW (and i therefore stress is reduced). Reduction in upper tube sheet stress is also plausible misconception.
TMl LI:ensing Exrmination Answer K y QUESTION: 083 (1,00)
The OTSG maximum allowable secondary pressure shall be limited to less than when OTSG shell temperature is below .
A. 100 psig,100*F B. 100 psig,200*F C. 200 psig,100*F D. 200 psig,200*F Answer: C K/A- 000035 K4.05 3.4 Page 3.4-15 Objective: IV.A.08.19 "Using Tech Specs identify and explain the bas!s for the low temperature OTSG pressure limit.
Reference (s): Tech Spec 3.1.2.2. & Bases Page 3-3 History- New l NRC Cognitive Level Rating 1 Measurement #83 l Knowledge of OTSG pressure design features and limits while cold..
A Discriminant Validity Plausible since the corted temperature limit is used, however the pressure limit used is incorrect. The pressure limit stated is adually the value for the temperature limit B Discriminant Validity Plausible distracter since the numbers contained are correct - but are reversed.
C Discriminant Validity Correct Answer.
D Discriminant Validity Correct pressure limit, however temperature limit is incorred. This distracter includes a plausible misconception since 200 degrees is the maximum
, (RCS) temperature limit for cold shutdown conditions as defined in TS 1.2.1.
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- l. TMI Licensing Examination l- Answer Kcy L- QUESTION: 084 (1.00)
The SBO diesel has the capability to energize ~
4160V bus through cross-tie breakers via start and loading.
1 A. Only C or D, manual B. C , D or E, auto
.C. Only D or E, auto D. C, D or E, manual i
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Answer: D i K/A- 000055 EA1.06 4.5 Page 4.1-13 ,
l Objective: IV.G.09.01 )
IV.G.09.06 Reference (s): OPM, Site ~ Blackout and Auxiliary Equipment, A-12, Rev 5, Page 9
- History
- New NRC Cognitive Level Rating 2
- l Measurement #84 Ability to direct, operate and monitor restoration of power with one EDG during Blackout conditions. (SBO diesel operation) l A D6scriminant Validity Plausible, includes correct starting mode but incomplete listing of buses capable of being energized.
I B Discriminant Validity Plausible, since answer includes complete list of buses capable of bein0 energized - but includes incorrect start mode (auto).
- C Discriminant Validity Plausible, but answer includes an incomplete listing of buses capable of
- beino energized. Also manual start is required rather than automatic start. ,
D Discriminant Validity Correct answer. j i
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-. . ._ - ._ . -- -. - . - - - . - . - - - . . - . . ~ -
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TMI Lic:nsing Excmin:ti n An:wer K y i QUESTION: 085 (1.00)
An RCS sample is being drawn when RM-G-18 goes into HIGH alarm Which ONE statement desenbos the automatic actions associated with RM-G-18 for this situation.
- A. - CA-V-4A and CA-V-5A close.
L B. CA-V4B and CA-V-58 close.
l-C. CA-V-2 and CA-V-13 close.
D. The Control Building ventilation system is shifted to emergency recirculation mode.
Answer: C K/A: 000061 AA1.01 3.6 Page 4.2-48
- . Objective
- IV.E.06.04 i Reference (s): EP-1202-12, Excessive Radiation Levels, Rev. 42, Page 12 History: New l
l NRC Cognitive Level Rating 1 Measurement #85 Ability to operate and monitor automatic aduation of the Radiation Monitoring System as applied to alarms.
A Discriminar:t Validity Plausible distracter since the valves listed are containment isolation sampling valves.
l 8 Discriminant Validity Plausible distracter since the valves licted are containment isolation sampling valves.
j C Discriminant Validity Correct answer.
l D Discriminant Validity Plausible misconception since the sampling room is located on the first floor of the Control Building.
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Tgg Ltc n-n wer KCY ditionwill JESTION: 086 (1.00) d in each condition listed. Which ONE conSA war signal failure rates are describeRbythe Smart Auto S muse an automaticTRANSFE i trumentdriftstoproduce adifference '
peration?
Selected RCS narrow id range pressure ns A. error of 4% over afive minute per o . h t a rate fof 10%/sec.
Selected"B" OTSG pressure fails hig a ent drift produces B.
Selected Loop A Feedwater Flow instrum C.
5% over 10 minutes. l low at a rate of 10%/30 secon Selected NIPower Range Channelfai s .
D.
B Page 3.7-10 Answer: 2.8 A4.01 K/A: 000016 tem, F-05, Rev 11, Page 14 IV.E.09.07 Objective:
OPM, Non-Nuclear Instrumentation sys Reference (s):
New History: 3 NRC Cognitive Level Rating eration of non-nuclearinstrum Ability to monitor and ensure proper op controls.
tch, Measurement #86channelinstrumentation selection tion would produce a SAS Plausible since f ailure rate is addresse ,
j A DiscriminantValidity cause a SASS actuation. This situaignal selection.
but not an aduationdofbut automatic is at a values atch, that would B DiscriminantValidity Plausible since failure Correct answer.
rate is addresse ignalselection.
t atch,
, tion C DiscriminantValidity cause a SASS actuation. sed, butis This at a valuei thatwould no situa but not an aduation of automatic sa SASS m sm tion would produce D DiscriminantValidity Plausible since f ailure rate is addrescau but not an aduation of automatic s
TMl Lic:nring Excminttion Answer K y QUESTION: 087 (1.00)
A reactor trip from 100% power has occurred.
- All RCPs were tripped 60 minutes ago.
- Steam flow and feedwater flow have been verifed.
- Incore thermocouples are tracking Ts.
Which ONE set of plant conditions indicates natural circulation is occurring?
OTSG RCS Cold Leg RCS Hot Leg Pressure Temperature Temperature i l
A. 700 psig, decreasing 540 deg F, increasing 550 deg F, stable B. 750 psig, stable 513 deg F, stable 545 deg F, decreasing C. 800 psig, stable 550 deg F, decreasing 570 deg F, increasing
. D. 940 psig, increasing 540 deg F, increasing 600 deg F, stable Answer: 8 K/A: BW/E08 EK2.2 4.0 Page 4.3-12 Objective: V.E.10.01 Reference (s)- ATP 1210-10 ATOG Transient Rules, Guides and Graphs, Rev 34 Page 12 History: TMl Question QR4E27-49-001 (Modified)
Original version was used on audit exam (Q #36).
Verified not used SRO Program quizzes.
NRC Cognitive Level Rating 3 Measurement #87 Demonstration of ability to interpret and apply proper inter relationships between natural circulation cooldown and primary system heat removal.
A Discriminant Validity Plausible distracter since conditions presented indicate natural circulation cooling is not occuning per ATP 1210-10 Natural Circulation Vertfication guidelines.
8 Discriminant Validity Correct answer.
C Discriminant Validity Plausible distracter since conditions presented indicate natural circulation cooling may be starting, but do not yet meet ATP 1210-10 Natural Circulation Verification guidelines.
D Discriminant Validity Plausible distracter since conditions presented indicate natural circulation cooling has not yet been established. Loop delta T is excessive in accordance with ATP 1210-10 Natural Circulation Verification guidelines.
i TMI Licensing Examination Answer Koy l
QUESTION: 088 (1.00)
Which ONE statement describes the function of the PORV NDTT Key Lock Switch on Control Room panel PCR7 A. In the AUTO position, setpoint is 2450 psig ONLY if RCS temperature is l BELOW 275'F.
l B. In the OFF position, setpoint is 485 psig at ANY RCS temperature.
C. In the AUTO position, setpoint is 485 psig ONLY if RCS temperature is ABOVE 275'F.
D. In the OFF position, setpoint is 2450 psig at ANY RCS temperature .
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Answer- D l K/A: 000002 K5.18 3.6 Page 3.2-4 Objective: IV.A.01.08 Reference (s): OPM, Reactor Coolant System, B-01, Rev 10, Page 19 History: New NRC Cognitive Level Rating 1 Measurement #88 Demonstration of knowledge of operationalimplications of NDTT and brittle fracture protection as applied to operation of the reactor coolant system.
A Discriminant Validity Plausible but incorred because in Auto the setpoint is 2450 psig if >275 degrees.
B Discriminant Validity Plausible but incorrect because in Oif, the setpoint is always 2450 psig, regardless of RCS temperature.
C Discriminant Validity Plausible but incorrect - setpoint changes to 485 psig only when in Auto and
< 275 degrees.
D Discriminant Validity Correct answer.
i TMl Lic:n:Ing Ex mination ,
Answer Kcy QUESTION: 089 (1.00) l Which ONE statement describes the operation of RC-V-1 Pressunzer Spray Valve?
A. In Manual control, the JOG circuit directs movement of the valve using 1 momentary contact pushbuttons; valve position is limited to 40% open.
B. In Manual control, the JOG circuit directs movement of the valve using I momentary contact pushbuttons; valve can be opened to 100% position.
C. In Automatic control, valve operation is derected by existing RCS pressure -
conditions; valve is opened to 60% open in response to an open cammand.
i D. In Automatic control, valve operation is controlled by existing RCS pressure conditions; valve is opened to 100% in response to an open command 1
Answer: B j l l
- K/A
- 000010 A4.01 3.5 Page 3.3-8 l l
Objective: IV.A.01.06 Reference (s): OPM, Reactor Coolant System, B-01, Rev 10, Page 17 History: New l
l NRC Cognitive Level Rating 1 l Measurement #89 Demonstration of ability to monitor and control RCS pressure through automaticinterlocks and manual operation of the Pressurizer Spray valve.
A Discriminant Validity Plausible misconception since automatic control limits valve travel to 40%
open.
I B Discriminant Validity Correct answer.
l C Discriminant Validity Plausible misconception since it correctly describes automatic opening of the
- spray valve but to an incorrect valve position.
D Discriminant Validity Plausible misconception that the spray valve will open to 100% on an automatic open signal.
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TMI LI:ensing Ex:mination Answ:r K y QUESTION: 090 (1.00)
Which ONE condition actuates the control rod withdrawal Out-inhibit?
A. Sequence Fault condition due to 30% overlap between 2 successive regulating rod groups
- 8. ICS neutron error signal of -5%
C. Reactor power at 56% with Asymmetric Rod Fault D. Stadup rate at +3.2 DPM on one Intermediate Range Ni channel Answer: D K/A: 000001 K4.07 3.8 Page 3.1-4
. Objective: IV.E.13.22 Reference (s): OPM, Control Rod Drive System, F-01, Rev 6, Page 74 History- New NRC Cognitive Level Rating 1 Measurement #90 Application of knowledge conceming CRO interiock features, rod control logic, and rod stops.
A Discriminant Validity Plausible misconception that a Sequence Fault initiates an Out inhibit.
B Discriminant Validity Plausible misconception that 5% neutron error (the setpoint to initiate ICS Neutron Cross Limits) initiates a CRD Out inhibit.
C Discriminant Validity Plausible misconception -this is a true statement if reactor power is greater than 60%.
D Discriminant Validity correct answer.
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(. _ , . . _ . - . . - . . . - - ._ . . . - . -
TMI Lirnsing Examination Answer K y QUESTION: 091 (1.00) i Current plant conditions are:
Plant is shutdown in preparation for a Refueling Outage.
RB purge is in progress for one hour.
RB sump liquid is being gravity drained to the Auxiliary Building sump.
Which ONE statement describes required actions for actuation if RM-A-9G HIGH alarm?
1 A. WDL-V-534 and WDL-535 RB e, ump drain line isolation valves close.
B. Kidney Filter Fan (AH-E-101) trips.
C. RC Drain Tank vent isolation valves (WDG-V-3 and WDG- V-4) close.
D. RCS letdown isolation valves (MU-V-2A/28) close.
Answer: A j K/A: 000068 A4.04 3.7 Page 3.9-4 Objective: V.D.07.02 Reference (s): EP 1202-12, Excessive Radiation Levels, Rev. 42, Page 11 (Enclosure 1) l History: New I NRC Cognitive Level Rating 1 ;
Measurement #91 Ability to manually operate / monitor automatic isolatinn of liquid rad waste system from the Control Room.
A Discriminant Validity correct answer.
8 Discriminant Validity Plausible misconception since the Kidney Filter system (inside containment building) includes charcoal filters which become heated when absorbing lodine. This highly radioadive water being drained from the RB Sump would also cause RB airbome activity to be elevated.
C Discr6minant Validity Plausible distrader since this is another flow path for radioactive materials to exit the containment building.
D Discriminant Validity Plausible distracter since this is another flow path for radioactive materials to exit the containment building.
TMl Licensing Excmination An2wer K y l QUESTION: 092 (1.00)
The reactor is operating at 100% power. Whidi ONE condition causes automatic start of ALL Emergency Feedwater Pumps? Both parts of the condition are required where "M" is used.
A. FW-P-1 A hydraulic oil pressure at 68 psig, M loss of two RCPs in RCS Loop A
1 B. Loss of two RCPs in RCS Loop B E 24 inches OTSG 1B Startup Range level l
l C. RB pressure at 5 psig, QB 9 inches OTSG 18 Startup Range level l
D. FW-P 1B hydraulic oil pressure at 74 psig, QB 24 inches 1 B OTSG Startup Range level l
Answer: C K/A- 000061 A3.01 4.2 Page 3.4-47 Objective: IV.C.05.11 i
Reference (s): OPM, Emergency Feedwater System,1-01, Rev 11, Page 5 History: New NRC Cognitive Level Rating 1 l
Measurement #92 Demonstration of knowledge pertaining to automatic operation and startup of emerDency systems required for core cooling during abnormal operations.
A Discriminant Validity Since loss of two RCPs in same 1000 initiates an automatic reactor trip, j plausible misconception is that this condition would also automatically initiate EFW. Also,68 psig is less than the actual 75 psig setpoint for EFW
! actuation on loss of feedpumps.
8 Discriminant Validity Since loss of two RCPs in same 1000 initiates an automatic reactor trip, plausible misconception is that this condition would also automatically initiate EFW. Also,25 inches is the setpoint for EFW level control under forced l RCS flow conditions, but is not the actuation setpoint (10') for low OTSG l _
level.
! C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconcepuon that EFW is auto started at 25' (EFW level control setpoint). Also,74 psig is less than the actual 75 psig setpoint for EFW actuation on loss of Feedwater pumps.
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1 TMl Licen:Ing Ex min: tion Answer K y QUESTION: 093 (1.00)
Current plant conditions are:
Reactor is at cold shutdown condition.
- DHR Loop A is operating.
- PARTIAL loss of ICS/NNI Hand power has occurred:
- The DH AUTO SUBFEED circuit is de-energized due to an electrical malfunction.
Which ONE statement describes information the operator can use to quickly determine if any or all Control Room DHR Loop A analog meters are inoperable due to LOSS OF DH AUTO POWER?
I A. Analog meters Fall LOW when instrument power is lost.
B. Flow instrument fluctuations (flow noise) STOP when instrument power is lost.
C. EP 1202-40 Loss cf ICS Hand and Auto Power enclosures identify specific meters affected by loss of DH AUTO POWER.
D. Plastic labels glued onto the control console identify DHR analog meter power supplies.
Answer: D K/A: 000057 AA1.05 3.4 Page 4.2-41 Objective: V.D.21.02 Reference (s): Annunciator Response Procedure H-1-8 Rev 11 History: New NRC Cognitive Level Rating 3 Measurement #93 Ability to operate / monitor back up indications as applied to loss of vital instrument bus.
A Discriminant Validity Plausible distracter since this is a possible failure mode for analog meters, but not for all. Many meters at our facility fail to mid-scale on loss of power.
B Discriminant Validity Plausible distracter since this is a true statement, but not the corred answer.
C Discriminant Validity Plausible distracter since this is this procedure does address loss of power effects, but does not delineate which failures are from Hand Power Vs Auto Power. The correct procedure to use specifically addresses loss of Hand Power.
D Discriminant Validity Correct answer.
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TMI Li; nsing Ex: min tion Answer Kcy QUESTION: 094 (1.00)
Current plant conditions are:
Reactor is operating at 100% power.
ICS is in full Automatic.
Due to relay actuations, trip of ALL 230 KV LINE Breakers EXCEPT those associated with the 1091 line (Middletown Junction) occurs.
Which ONE statement describes the expected plant design response to these conditions?
A. Plant remains at 100% load,1 A & 1B Aux Transformers remain energized.
B. Plant remains at 100% load,1 A Aux Transformers is de-energized,18 Aux Transformer remains energized, carrying all station loads.
C. Automatic load reduction to 50% due to inadequate line capacity,1 A & 1 B Aux Transformers remain energized.
D. Reactor Trip, loss of off-site power.
Answer: A K/A: 000062 K1.04 4.2 Page 3.6-2 Objective: IV.G.01.08 Reference (s): E-229-001, Electrical Substation, Rev. 25 History: New NRC Cognitive Level Rating 3 Measurement #94 Knowledge of physicalinterconnections between off-site power sources and the plant AC distribution system
- A Discriminant Validity Correct answer.
B Discriminant Validity Plausible misconception conceming the breaker-and-a-half scheme utilized in the 230 kV switchyard.
C Discriminant Validity Plausible misconception that this line is not capable of sustained operation carrying 100% of TMI output. It is true that one line is capable of carrying only 50%.
D Discriminant Validity Plausible misconception conceming the breaker-and-a-half scheme utilized in the 230 kV switchyard.
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TMI Licen2ing Excmin ti::n Answer K;y QUESTION: 095 (1.00)
During refueling operations RM-G-6, the Auxiliary Fuel Handling Bridge monitor, becomes inoperable. Which ONE condition allows Refueling Operations to continue?
A. Install a suitable portable instrument of comparable range and sensitivity.
B. Verify RM-G-7 is operable on the adjacent Fuel Handiing Bridge.
C. Obtain the approval of the Fuel Handling Supervisor.
D. Obtain the approval of the Radiological Controls GRCS.
Answer: A K/A: 000034 K6.02 3.3 Page 3.8-13 Objective: V.B.16.09 Reference (s)- Technical Specification Section 3.8.1, Amendment 198, Page 3-44 History- TMI Question AL5B16-09-QO3 (Modified)
Verified not used on audit exam or in SRO Program quizzes.
NRC Cognitive Level Rating 1 Measurement #95 Knowledge of Tech Spec and procedural requirements regarding of loss of or malfunction of RMS during Refueling Operations.
A Discriminant Validity Correct answer.
B Discriminant Validity Plausible distracter, since RM-G-7 is mourited on the adjacent fuel handling bridge inside the Containment Building.
C Discriminant Validity Plausible distracter since Fuel Handling Supervisor is senior management representative responsible for Fuel Har'dling operations.
D Discriminant Validity Plausible since GRCS can issue stop work order based upon radiological conditions concems.
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TMl Lincnsing Ex min: tion Answer K y QUESTION: 096 (1.00) l Current plant conditions are:
- A LOCA has occurred. i 4
- Off-site power is being supplied.
- Hydrogen Recombiner is in service due to elevated RB hydrogen concentration. !
- Plant electrical systems are aligned normally. 1
- No equipment failures have occurred.
Control Room staff has just been informed that the Hydrogen Recombiner reaction chamber temperature has steadily decreased over the last half-hour from 1300 F down to 1280 F.
Which ONE statement describes the reason for the slowly decreasing Reaction Chamber temperature?
?
A. RB Pressure decreasing.
B. Hydrogen concentration is increasing.
4 2 C. Hydrogen concentration is decreasing.
D. RB Pressure is increasing. ;
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- Answer C K/A: 000028 A1.01 3.8 Page 3.5-17
}
) Objective: IV.B.12.07
! Reference (s): OP-1104-62, Hydrogen Recombiner, Rev 27, Page 8 i
History: New NRC Cognitive Level Rating 2 Measurement #96 Ability to monitor and predict changes in hydrogen concentration associated with operation of the hydrogen recombiner system.
A Discriminant Validity Plausible misconception that this would decrease recombiner system flow, and therefore reduce reachon rate (and chamber temperature).
8 Discriminant Validity Plausible misconception regarding theory of operation of recombiner system.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception that this would increase recombiner system flow, and therefore increase reaction rate (and chamber temperature).
.- - -.. -=- .- -. -- - .. . - . - _ - _ - - . - ,
TMI Licensing Exrminati:n Answ:r K y QUESTION: 097 (1.00) l Current plant conditions are: '
Reactorisinpped from full power.
- HPI has been manually actuated due to LOCA.
RCS Subcooling Margin is 3 degrees F.
- 4 RCPs are running.
After stopping one RCP in each loop, the primary CRO was distracted. No other control manipulations have taken place for the last 4 minutes. Which ONE statement describes required actions for this situation?
A. Trip remaining RCPs, initiate HPI, Verify EFW, Raise OTSG levels to 75-85%.
B. Trip remaining RCPs, initiate HPI, Verify EFW, Raise OTSG levels to 50%
C. Trip one RCP, initiate HPI, start EFW, raise OTSG levels to 50%
D. Continue operation of both RCPs, initiate HPI, start EFW, raise OTSG levels to 75 % 85%
Answer: D K/A: BW/E03 EA2.2 4.0 Page 4.3-6 Objective: V.E.10.08 Reference (s): ATP 1210-2, Loss of 25 *F Subcooled Margin, Rev 17, Page 2.0 ATP 1210-10 ATOG Rules, Guidelines and Graphs, Rev 34 Page 5 History: New NRC Cognitive Level Rating 3 Measurement #97 Demonstration of ability to adhere to appropriate procedures and operation within limitations of facility's license s,nd amendments. Also evaluates resolution of procedural hierarchy (ATOG Rule VS Procedure step).
A Discriminant Validity Plausible distrader since this is the corted action for loss of subcooling margin if RCPs are tripped within 2 minutes.
8 Discriminant Validity Plausible distracter since tripping all RCPs (within 2 minutes) is expected response to loss of subcooling margin. Distrader also uses EFW 50% level setpoint normally used during natural circulation conditions.
C Discriminant Validity Plausible distracter since this statement includes portion of correct ATOG response, uses EFW level control setpoint.
D Discriminant Validity Correct answer.
TMl Licensing Ex: min tion Answer K y QUESTION:~ 098 (1.00)
Current plant conditions are:
- Reactoris operating at 90% power
- All RCPs are operating.
- High Vibration alarm actuates on RC-P-1 A (15 mils and slowly increasing).
- Reset of vibration alarm has failed to clear the alarm.
Which ONE statement describes the required actions required for this situation?
A. Start the oil lift pump, then if vibration does not decrease trip the RCP.
B. Trip RC-P-1 A, then reduce power as necessary to stabilize plant.
C. Reduce power to 50% - 75%, then trip RC-P-1 A.
D. Close #1 Seal Leakoff valve (MU-V-33A), within 5 minutes, reduce power to 50%
- 75%, and trip the affected pump within 30 minutes.
Answer: C K/A: 000015/17 AK3.03 4.0 Page 4.2-10 l Objective: V.C.06.03 i
Reference (s): Abnormal Procedure,1203-16, Reactor Coolant Pump and Motor Malfunction, Rev 38, Page 5.0 History- New NRC Cognitive Level Rating 1 Measurement #98 Demonstration of ability to sequence proper response to RCP malfunctions in accordance with facility procedures and requirements.
A Discriminant Validity PlausitAe since this is related to coted adion for lube oil emerDency, found in same abnormal procedure (1203-16). RCP should not be secured prior to power reduction below 75%.
B Discriminant Validity Plausible since this indudes correct actions - but the order is reversed.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible since this is the corred adion for Seal No.1 failure, found in same abnormal procedure (1203-16).
TMl Lic2nring Extminitien An:wcr Ksy
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QUESTION: 099 (1.00) 1 Current plant conditions are:
- Readoristripped
- Manual HPl and Manual 4 PSI ES actuation on BOTH trains.
j - Automatic ESAS actuation has NOT occurred.
1 Refer to the attached control panel drawing. The control room operator momentarily
- ~ depresses the following pushbuttons
- 2 of 3 Enable \ Defeat pushbuttons on the 4 PSI Manual actuations (ONCE) for both trains
- All 3 Enable / Defeat pushbuttons on the HPl Manual actuations (ONCE) for both trains Which ONE statement describes the status of the Manual ES signals to affected plant components (pumps, valves, etc.)?
A. BOTH the Manual 4 PSI and the Manual HPI ES actuation signals ARE DEFEATED for ALL affected plant components.
B. The Manual 4 PSI ES actuation signals ARE DEFEATED for ALL components affected by the 4 PSI actuation - but the Manual HPI actuation signals ARE NOT DEFEATED for ALL components affected by the HPI ES actuation.
C. The Manual HPl ES actuation signals ARE DEFEATED for ALL components affected by the HPl actuation - but the Manual 4 PSI actuation signals ARE NOT DEFEATED for ALL components affected by the 4 PSI ES actuation. ,
1 D. BOTH the Manual 4 PSI AND the Manual HPl signals ARE NOT DEFEATED for ALL affected plant components. I l
Answer C K/A: 000013 KS.02 3.3 Page 3.2-24 Objective: IV.E.24.21 Reference (s): OPM F-06, ESAS, Rev. 8, Page 18 History: TMI Question QR4E24-21-QO1 Verified not used on audit exam.
Used in SRO Program 2/18/98 quiz.
NRC Cognitive Level Rcting 3
TMI Lic:n2ing Ex mination Answer Key Measurement #99 Knowledge of operationalimplications as they apply to ESAS safety system logic and reliability.
A Discriminant Validity Plausible misconception regarding logic operation and reset sequences.
8 Discriminant Validity Plausible misconception regarding logic operation and reset sequences.
C Discriminant Validity Correct answer.
D Discriminant Validity Plausible misconception regarding logic operation and reset sequences.
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! TMl Licensing Examination Answer Kcy QUESTION: 100 (1.00)
Current plant conditions are:
Reactoris operating at 90% power EG-Y-1 A is being started for surveillance.
Syndvoscope indicator is at 12:00 and is not moving in edher forward or reverse
. direction.
Indications at time of EG-Y-1 A breaker closure
! - Generator load is 0.15 MW.
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- Generator reactive load is - 0.3 MVars.
After closing the output breaker, the operator pauses for five seconds to monitor generator output conditions. Which ONE statement desenbes the status of the output breaker and generator FlVE SECONDS after the breaker was initially closed?
A. Output breaker is closed with generator load slowly increasing above 0.15 MW.
B. Output breaker is open due to over-current relay trip protection.
C. Output breaker is closed with generator otdput at 0.15 MW.
D. Output breaker is open due to reverse power trip relay protection.
l Answer. D K/A: 000064 A4.06 3.9 Page 3.6-11 Objective: IV.G.08.09 l
Reference (s): MAP A-1-3, Rev 5 History: New NRC Cognitive Level Rating 2 Measurement #100 l Ability to manually start, load, and stop the EDG from the Control Room.
_A Discriminant Validity Plausible distracter since initial load is established at 0.15 MW.
8 Discriminant Validity Plausible misconception regarding normal parameters on Control Room indications during initial generator loading.
C Discriminant Validity Plausible distracter since this matches indications at the time of breaker closure.
D Discriminant Validity Correct answer.
Sc:ntris Outlins Facility: TMl Scenario No.: 1 Op-Test No.: 1 t
Examiners: Operators:
Objectives: NI-5 will fail high and SASS fails to actuate, causing continuous rod insertion. Operators will not be able to select NI-6 attemate ICS Ni signal. Crew must respond to stop the rod insertion to prevent a reactor trip and stabilize the plant. CO-P-2A breaker trips and CO-P-2B fails to auto start.
This causes FW-P-1B to trip and requires manual plant runback. A Tube Rupture (approximately 800 gpm) will occur on B OTSG. The crew should initiate a plant shutdown then reactor trip in accordance with ATP 1210-5 and ATP 1210-1. 25 F subcooled margin may be lost, and the crew must respond in accordance with ATP-1210-2. RC Pumps must be secured. When 25'F subcooled margin is regained, RC Pumps may be restarted. HPI must be throttled to minimize subcooling margin and control RCS pressure within the allowable region of Figure 1 for the current plant conditions. If the crew isolates the B OTSG, RCS pressure must be maintained less than 1000#. The plant should be stabilized and then a cooldown commenced in accordance with ATP 1210-5.
Initial Conditions: IC 17,100% power, steady state, equilibrium Xenon, EOL. Following initialization:
(1) Shift MU-P-1 A cooling source to NSCC using remote function CC12. (2) Start MU-P1 A, (3) Secure MU-P-1B and red both extension controls in pull-to lock, (4) Rack out MU-P-1B breaker using remote function MUR21, (5) Start IC-P-1B and secure IC-P-1 A.
{ Turnover: Plant is at 100% power, steady state, equilibrium Xenon,640 EFPD. RCS boron is 36 PPM and the BAMT is the Tech. Spec. Tank. MU-P-1B is out of service or an oil change. It is expected to be retumed to service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. MU-P-1 A is in service and being cooled by NSCCW.
No other major maintenance or surveillance is in progress. There are no releases in progress and none scheduled o this shift. Continue 100% power operations.
Event Malf. No. Event Event No. Type
- Description
- 1. NI20A (I), (R), NI-5 fails high, CO-V-51 may open due to taking feedwater in hand.
(N), Continuous rod insertion controlled by the PCRO by taking hand IC47 control of the diamond rod control station.
- 2. FW22A (C),(N), CO-P-2A trip, feedwater pump trip, operators must manually control (R) rods and feedwater.
- 3. IA01C (C) lA-P-4 trip, operators respond to loss of IA-P-4
- 4. TH15B (M) (R) B-OTSG Tube Rupture,4% severity, recognize leak, reduce power, possible automatic reactor trip, may manually trip the reactor, possible ESAS actuation, possible loss of SCM
i * (N)ormal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor Page 1of7
4 Operat:r Actions i
Op-Test No.: 1 Scenario No.: 1 Event No.: 1 Page jiL,of 7
(. Event
Description:
NI-5 fails high, CO-V-51 may open due to taking feedwater in hand. Continuous rod insertion controlled by the PCRO by taking hand control of the diamond rod control station.
Time Position Applicant's Actions or Behavior TEAM /PCRO Diagnose and announce NI-5 failing high.
, - Diagnose events and conditions TEAM Diagnose SASS actuation failure with continuous rod insertion.
- Diagnose events and conditions PCRO/SCRO Take manual control of the Diamond Rod Control and Feedwater to stabilize the plant. (CT)
- Manual reactivity control for SRO-1
- Feedwater in manual TEAM Determine that the ICS will have to remain in manual due to the inability to select an operable N1.
- NI-6 selector pushbutton failure SSISF Determine than the minimum degree of redundancy required for power range instrument channels per T.S. Table 3.5-1 is satisfied.
- Comply with and use Tech. Specs
- Direct "A" RPS to be placed in manual bypass i
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Operattr Acti:ns I Op-Test No.: 1 Scenario No.: 1 Event No.:,_2_ Page _J_ of _,2_,
( Event
Description:
CO-P-2A trip, feedwater pump trip, operators must manually control rods and feedwater.
Time Position Applicant's Actions or Behavior RAM Diagnose CO-P 2A and FW-P-1B trip by alarms and panelIndications.
Note that CO-P-28 did agi auto start.
- Understand and Interpret Annunciators and Alarms
- Recognize higher than normal amps on CO-P-2C.
- Perform manual plant runback due to FW Pump trip
- Understand Plant and System Response
- Operate Control Boards
- Manual Reactivity Control for SRO-l IEAM Make Plant announcement for FW Pump trip and power reduction.
- Dispatch personnel to CO-P 2A and its breaker.
- CO-V-51 opens l 1
SCRO Start CO-P-2B from CL
- Diagnose CO-V-51 open on high AP by alarms.
TEAM Stabilize Plant following power reduction 1
- Initiate action to reciose CO-V-51
- Initiate action to FW-P-1B.
l g When plant conditions allow or when directed:
- Reset and reclose CO-V-51 per ARP PAP-18. )
- Restart FW-P-1B per OP 1106-3.
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j Op-Test No.: 1 Scenario No.: 1 Event No.: _3 Page _4. of ,,1._
Event
Description:
lA-P 4 trip, operators respond to loss of IA-P-4 l Time . Position Applicant's Actions or Behavior g Recognize failure of IA-P-4 from alarm PLB-1-6.
- Understand and Interpret Annunciators and Alarms
- Communicate and interact with crew SCRO hecognize failure of IA P-4 from alarm PLB-1-6.
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- Contact AO to open IA V-2104A/B and place IA-P-1 A and/or IA-P-1B l
in Hand per ARP PLB-1-6.
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Operat:r Acti:ns Op-Test No.: 1 Scenario No.: 1 Event No.: d Page jLof l
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k Event
Description:
B-OTSG Tube Rupture,4% severity, recognize leak, reduce power, possible automatic reactor trip, may manually trip the reactor, possib!s ESAS actuation, possible loss of SCM i '
Time Position Applicant's Actions or Behavior 1
ham Diagnose that based upon RCS inventory loss and RMS indications that an OTSG tube leak / rupture has developed.
! - Understand and Interpret Annunciators and Alarms
- Diagnose Events and Conditions E implement ATP-1210-5.
! EfcBQ increase makeup and isolate letdown to attempt to maintain Pressurizer level, I
a PCRO - May announce his intent to trip the reactor to the Team due to:
'f.
i- - < 150" Pressurizer level while reactor power is > 25%,
- Then trip the reactor and initiate HPl based on the ATP 1210-10 .
F Pressurizer Level Guide.
May manually trip the reactor due to low RCS pressure.
TEAM
- Respond to the reactor trip per the immediate actions of ATP 1210-1. I PCRO initiate HPl or verify ESAS actuation as necessary
- Full HPI is considered to be one HPI pump running in each train (2), l l
/ and full flow through all four HPl lines (MU-V-16A,B, C & D full open) 4 k
TEAM If 25'F SCM is lost, perform immediate actions of ATP 1210-2: (CT) j Verify / direct ATP 1210-1 immediate actions. I f
BE BE . Verify / direct ATP 1210-2 immediate actions for Loss of 25'F SCM If 2
applicable; otherwise verify / direct necessary actions for increased loss of i .
RCS inventory (initiation of HPI) into B OTSG.
31 Declare the emergency event as necessary for the SG tube leakage and/or loss of RCS subcooled margin per the Emergency Plan.
- SF may be allowed to classify the event at the end of the scenario at examiner's discretion.
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- IEAM If/when 25'F primary to subcooled margin was lost and subsequently recovered, restart RC Pumps and control HPl in accordance with ATP 1' 1210-2 and 1210-10, and transition back to ATP 1210-5 to commence RCS cooldown. (CT)
- B OTSG will eventually fill due to secondary leakrate which may cause ,
concem for potential equipment damage. Isolation of 8 OTSG for this reason is a subjective call.
31&F., Evaluate rate of B OTSG fill and possibility of B OTSG isolation due to potential equipment damage per ATP 1210-5.
- B OTSG lsolation during this scenario is not considered critical; j
however, it is critical that RCS and B OTSG pressure be mair.tained j less than =1000 psig to prevent MSSV lift.
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Oper;t:r Actions TEAM Ensure RCS temperature is less than 540*F (T-hot or incore) and RCS pressure less than 1000 psig prior to isolation of B OTSG per ATP 1210-5.
k IE8M Control RCS inventory / pressure by throttling HPI, and control B OTSG pressure using TBVs, to prevent MSSV lift on B OTSG (=1040 psig), which would cause an unparthloned and unmonitored radioactive release to atmosphere. (CT)
- B OTSG isolation during this scenario is not required prior to termination.
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Operator Actions Op Test No.: 1 Scenario No.: 1 Event No.: ._b_, Page 7 of 7
(
Event
Description:
DR-P-1 A ES start failure during the OTSG Tube Rupture Time Position Applicant's Actions or Behavior l Recognize that DR-P-1B did not start on ESAS.
M/.PCRO
- Start DR-P-1 A l
- Comply with Tech Specs )
- Comply wit use of procedures l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
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SCENARIO NUMBER: ~1 1
EXAMINERS: l f
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l OPERATORS:
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION i
1.0 - GENERAL DESCRIPTION OF SCENARIO Plant is at 100% power, steady state, EOL, with MU-P-1B out of service NI-5 f ails high, requiring actions to stop continuous inward rod motion and reactor trip on low pressure.
CO-P-1B will trip with taibre of CO P-10 to Auto Start resulting in Main Feed Pump trip and ICS runback.
Instrument Air Compressor IA-P-4 trips resulting in a reduction in Instrument Air pressure.
During plant shutdown a Steam Generator Tube Rupture will occur resulting in a reactor trip and HPl.
Estimated scenario time - 65 Minutes I
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THREE MILE ISLAND UNIT 1 l SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION k ~ 2.0 '- REFERENCES A. 10 CFR 55.45 Operating Test, (a) Content 1
B. Procedures Plant Shutdown
- 1. 1102 10 j
- 2. 1210-1 Reactor Trip
- 3. 1210-2 Loss of 25'F Subcooled Margin
- 4. - .1210 OTSG Tube Leakage ;
- 5. 1210-10 Abnormal Transients Rules, Guides and Graphs C. TECHNICAL SPECIFICATIONS i
- 1. Section 3.1.6, RCS Leakage ;
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l 3.0 SCENARIO INITIAllZATION A. IC-17,100% power, steady state, equilibrium Xenon, EOL Following the initialization:
- 1. Shift MU-P-1 A cooling source to NSCC using Remote Function CC12.
- 2. Start MU-P-1 A
- 3. Secure MU P-1B and red tag both extension controls in Pull-to Lock
- 4. Rackout MU-P-1B breaker using Remote Function MUR21
- 5. Start IC-P-1B and secure IC-P-1 A B. MALFUNCTIONS
- 1. IC47 NI-5'6 SASS channel f ailure activate immediately
- 2. N120A NI-5 fails high; assign to Remote #1 !
- 3. FW22A CO P-2A trip; assign to Remote #2
- 4. IA01C lA-P-4 trip; assign to Remote #3
- 5. TH15B B-OTSG Tube Rupture; assign at 4% severity with a 180 second ramp to Remote #4 6 RWO9A DR-P-1 A ES start failure; insert immediately f
.q C. OVERRIDES
- 1. 01 A4S28-DI, NAT, OFF; Activate immediately to prevent CO-P-2B auto start when CO-P-2A trips
- 2. 02A2SO9-DI, , OFF; Activate immediately to prevent NI-6 selection D. MONITOR
- 1. MSK2609A: Set to 14.7 i
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
!- ~ 4.0 SCENARIO PREVIEW A. ' EXAMINER PREVIEW
- 1. Plant is at 100% power, steady state, with MU-P-1B out of service for an oil change.
- 2. NI-5 will fail high and SASS fails to actuate, causing continuous rod insertion. Operators will not be able to select NI-6 alternate ICS Ni signal. Crew must respond to stop the rod insertion to prevent a reactor trip and stabilize the plant. (CT) CO-V-51 may open. ,
- 3. After crew has taken the required action for the Nl f ailure CO-P-2A breaker trips and CO-P-2B fails to auto start. This causes FW-P-1B to trip and requires manua! plant runback. CO-P-2B may be manually started, and if so, CO-V-51 will open.
- 4. lA-P-4 trips requiring action to restore Instrument Air pressure.
- 5. A Tube Rupture (approximately 800 gpm) will occur on B OTSG. The crew should initiate a plant shutdown then reactor trip in accordance with ATP 1210-5 and ATP 1210-1. (CT)
- 6. 25 F subcooled margin may be lost, and the crew must respond in accordance with ATP-1210-2.
RC Pumps must be secured (CT).
- 7. When 25*F subcooled margin is regained RC Pumps may be restarted.
- 8. HPl must be throttled to minimize subcooling margin and control RCS pressure within the allowable region of Figure 1 for the current plant conditions. (CT)
( 9. If the crew isolates the B OTSG, RCS pressure must be maintained less than 1000# (CT)
, 10. The plant should be stabilized and then a cooldown commenced in accordance with ATP 1210-5.
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1 l THREE MILE ISLAND UNIT 1 - )
SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION !
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'i . B, SHIFT BRIEFING (SEE PLANT STATUS BOARD)- .
- 3. MU-P-1B is out of service for oil change. It is expected to be returned to service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- 4. MU-P 1 A is in service and being cooled by NSCCW. l
- 4. No othe major maintenance or surveillance is in progress.
- 5. There are no releases in progress and none scheduled for this shift.
]
- 6. Continue 100% power operation.
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THREE MILE ISLAND UNIT 1 l SENIOR REACTOR OPERATOR {
SIMULATOR EXAMINATION i k 5.0 SEQUENCE OF EVENTS Examiner Notes and Actions Exnected Operator Actions l
- 1. Initialize the simulator in accordance with Section 3.0. l
- 2. Assign team positions and conduct the SHIFT 2.1 Assume assigned team positions.
BRIEFING per Section 4.B.
2.2 Take turnover and review plant status.
NOTE: Allow crew 3-5 minutes to take turnover and .
l assume the watch. 2.3 Assume the watch. '
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- 3. ICO-MALFUNCTION: When directed by 3.1 TEAM /PCRO - Diagnose and announce NI-5 examiner, activate Remote #1 to cause NI-5 t f all high. failing high.
i Understand and Interpret Annunciators and Alarms NOTE: CO-V-51 may open 3.2 TE AM - Diagnose SASS actuation f ailure with i i !
l t continuous rod insertion. !
Diagnose Events and Conditions l
l Operate ControlBoards 3.3 PCRO/SCRO - Take manual control of the Diamond Rod Control and Feedwater to ManualReactivity Controlfor SRO-I stabilize the plant. (CT)
NOTE: The operator will be unable to select NI-6 3.4 TEAM - Determine that the ICS will have to because of a failure of the NI-6 selector remain in manual due to the inability to select I
pushbutton. an operable Nl.
Demonstrate Supervisory Ability 3.5 SSISF - Contact l&C for troubleshooting.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
! NOTE: The minimum degree of redundancy required 3.6 $$$E - Determine than the minimum degree of for all RPS trip channels is ONE, With NI-5 redundancy required for power range failed high, the actual degree of redundancy = 3 instrument channels per T.S. Table 3.5-1 is operable channels minus 1 channel required to satisfied. l trip = TWO.
Comply With and Use Tech. Specs.
Demonstrate Supervisory Ability 3.7 gjdE - Direct "A" RPS to be placed in Manual Bypass after plant stabilization.
- 4. ICO-MALFUNCTION: After the plant has been 4.1 TEAM - Diagnose CO-P-2A and FW-P 1B trip stabilized at normal plant operating conditions by alarms and panel indications. Note that or when directed by examiner, activate Remote CO-P-2B did ngt auto start.
- 2 to cause CO-P-2A to trip.
Recognize higher than normal amps on l Understand and Interpret Annunciators and Alarms CO-P-2C.
k Diagnose Events and Conditions Understand Plant and System Response Perform manual plant runback due to FW Operate ControlBoards Pump trip.
l ManualReactivity Controlfor SRO-l EXAMINER /ICO: If reactor trips for any reason, continue with Steam Generator Tube Rupture.
NOTE: CO-P-2B did not auto start as designed, 4.2 TE AM - Make Plant announcement for FW l resulting in FW-P-1B trip and the need for an Pump trip and power reduction. Dispatch l ICS runback. CO-P-2B may be manually personnel to CO-P 2A and its breaker.
started anytime, and if so, CO-V-51 will open on high D/P if not already open.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION i ICO-ROLE PLAY: If directed to check CO P-2A and/or 4.3 ffaQ - Start CO-P-2B from CL and diagnose it's breaker, provide Crew with the following CO-V 51 open on high AP by alarms, reports after a time delay (3 to 5 minutes)-
CO-P-2A breaker overload relays are tripped, 4.4 TEAM- Stabilize Plant following power CO-P-2A motor smells burnt. reduction, and initiate action to reclose CO-V-51 and reset FW-P-18.
ICO/ REMOTE FUNCTION: When directed to reset 4.5 SQRQ - When plant conditions allow or when CO-V-51 at Powdex panel, use FWR40 directed, reset and reclose CO-V-51 per ARP PAP-1-8.
Operate ControlBoards NOTE: Examiner may allow time for restarting 4.6 EfRQ- Restart FW-P-1B per OP 1106-3.
FW P-18, but it is not required Operate Control Boards
( 5. ICO-MALFUNCTION: After plant is stable and 5.1 ff RQ- Recognize failure of IA P-4 from alarm CO-V-51 is closed, or as directed by an PLB-1-6. I examiner, activate Remoto #3 to trip IA-P-4.
Understand and Interpret Annunciators and Alarms Communicate and Interact With the Crew l lCO/ REMOTE FUNCTION: Use lAR08 to OPEN 5.2 SCRO - Contact AO to open IA-V 2104A/B and !
lA-V-2104A/B and IAR01 and/or IAR02 to place place IA-P-1 A and'or IA-P-1B in Hand per ARP I lA-P-1 A and/or IA-P-1B in HAND as directed by PLB-1-6.
the SCRO.
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. - _ - _ ._ _ = _ _ _ _ _ . _ _ . . . . . _ _ _ _
i THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION I 6. ICO-MALFUNCTION: When directed by an 6.1 TEAM- Diagnose that based upon RCS examiner, activate Remote #4 to initiate OTSG inventory loss and RMS indications that an l 1B tube rupture on a 3 minute ramp. OTSG tube leak / rupture has developed.
Understand and Interpret Annunciators and Alarms Diagnose Events and Conditions _
1 NOTE: A reactor trip may be initiated prior to 6.2 SF -Implement ATP-1210-5.
implementing ATP-1210-5, in which case ATP-1210-1 would first be implemented.
i Comply With and Use Procedures Demonstrate Supervisory Ability i
Comply With and Use Procedures 6.3 .P_QIlQ - Increase makeup and isolate letdown Operate ControlBoards to attempt to maintain Pressurizer level.
7 3
Comply With and Use Procedures 6.4 ,P_QRQ- May announce his intent to trip the Operate ControlBoards reactor to the Team due to < 150" Pressurizer Communicate andInteract With the Crew level while reactor pcwer is > 25%, then trip the l
reactor and initiate HPl based on the ATP NOTE: Reactor trip will occur either manually or 1210-10 Pressurizer Level Guide.
automatically OR j TEAM - May manually trip the reactor due to !
low RCS pressure.
6.5 TE AM - Respond to the reactor trip per the immediate actions of ATP 12101.
NOTE: Full HPI is considered to be one HPl pump 6.6 PCRO -Initiate HPI or verify ESAS actuation running in each train (2), and full flow through as necessary all four HPl lines (MU-V-16A.B. C & D full open) l l
NOTE: If SCM is lost, it will most hkely be quickly 6.7 TEAM -If 25*F SCM is lost, perform 9
regained. immediate actions of ATP 1210-2: (CT)
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION k Examiner Notes and Actions Exnected Operator Actions I
Comply With and Use Procedures 6.8 EE-Verify / direct ATP 1210-1 immediate Demonstrate Supervisory Ability actions.
Understand Plant and System Response 6.9 TEAM - Recognize that DR-P-1B did not start on ESAS.
Comply With and Use Tech. Specs. 6.10 PCRO - Start DR-P-1 A Comply With and Use Procedures 6.11 SF - Verify / direct ATP 1210-2 immediate Demonstrate Supervisory Ability actions for Loss of 25*F SCM if applicable; otherwise verify / direct necessary actions for increased loss of RCS inventory (initiation of HPI) into B OTSG.
NOTE: SF may be allowed to classify the event at the 6.12 SE Declare the emergency event as f
1 end of the scenario at exarniner's discretion. necessary for the SG tube leakage and/or loss of RCS subcooled margin per the Emergency Demonstrate Supervisory Ability Plan.
ICO/ EXAMINER: ROLEPLAY as GRCS (RAC) and if requested by SS/ED, supply offsite dose projections which are less than the limits for SG isolation per ATP 1210-5. Also inform him that offsite doses are minimal.
NOTE: B OTSG will eventually fill due to secondary 6.13 TE AM - If/when 25cF primary to subcooled leakrate which may cause concern for potential margin was lost and subsequently recovered, equipment damage. Isolation of B OTSG for restart RC Pumps and control HPl in this reason is a subjective call. accordance with ATP 1210-2 and 1210-10, and transition back to ATP 1210-5 to commence Comply With and Use Procedures RCS cooldown. (CT)
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION k NOTE: B OTSG isolation during this scenario is not 6.8 33@F - Evaluate rate of B OTSG fill and considered critical; however, it is critical that possibility of B OTSG isolation due to potential RCS and B OTSG pressure be rnaintained less equipment damage per ATP 1210-5.
than -1000 psig to prevent MSSV lift.
Understand Plant and System Response Comply With and Use Procedures 6.9 TEAM -- Ensure RCS temperature is less than 540'F (T-hot or incore) and RCS pressure less than 1000 psig prior to isolation of B OTSG per i ATP-1210-5.
Note: B OTSG iselation during this scenario is not 6.10 Ig&M - Control RCS inventory / pressure by
. required prior to termination. throttling HPI, and control B OTSG pressure using TBVs, to prevent MSSV lift on B OTSG Operate ControlBoards (-1040 psig), which would cause an unpartitioned and unmonitored radioactive
( release to atmosphere. (CT)
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l THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
'I 6. TERMINATION POINT When all of the following conditions exist:
- 1. Crew has had sufficient opportunity to perform all applicable critical tasks.
- 2. ATP 1210-5 follow-up actions are in progress.
- 3. Subcooling margin has been minimized (30 - 70 degrees).
- 4. A plant cooldown is in progress.
- 5. . All examiners agree on termination.
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Sc:naria Outline Facility: TMI Scenario No.: 2 Op-Test No.: 2
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Examiners: Operators:
4 Ot$ectives: Reactor critical at 10 amps during xenon-free plant startup with CRG-6 at 64%. Reactor power is increased per OP 1102-2, and NI-3 startup rate amplifier fails to respond. The crew must address Tech. Specs. for the failed Ni and determine no action required. When reactor power reaches
=1%, a dropped control rod will occur. The crew will have to establish 1% Ak/k subcriticality. (CT)
During establishment of shutdown reactivity IC-P-1 A will trip with auto / manual start failure for IC-P-18.
The crew will have to manually trip the reador in accordance with 1202-17 Loss of Irtermediate Closed Cooling System. When the reactor is tripped, to MSSVs on the B-SG will fail 100% open. Also B OTSG will suffer an over-feed due to FW-V-16A failing 100% open, resulting in excessive primary to secondary heat transfer. The crew must isolate the B OTSG in accordance with ATP 1210-3 to
< mitigate the excessive primary to secondary heat transfer. (CT) HPl will NOT automatically actuate due to ESAS 1600# channels 2 and 3 bistable failures, requiring manual HPl initiation. MU-V-16A will fail to open on ESAS. (CT) HPl must be throttled to comply with RCS pressure requirementu of 1210-10, Figure 1/1 A. (CT) During the scenario the appropriate E-Plan declaration will be made.
Initial Conditions: IC-9, Hot Zero Power,104amps, Xenon free,640 EFPD. Use Remote Function MSR21 to OPEN AS-V-8. Close GS-V-4 on PL Turnover: Reactor is critical at 10 4amps during a xenon-free startup at 640 EFPD. RCS boron is 543 ppm. Plant was previously at 1,00% power for 65 days and was shutdown for main turbine bearing replacement. It has been 7 days since the plant shutdown. Operating main feedwater pump and gland steam are both on auxiliary steam. No maintenance or surveillance is in progress. There are no
( releases in progress and none scheduled forthis shift. Continue plant startup Event Malf. No. Event Event No. Type
- Description
- 1. N112A (I), (N) NI-3 SUR Amplifier Failure
- 2. RD0115 (C),(R) Dropped Rod (6-1) .
- 3. CC04A (C) IC-P-1 A trip -IC-P-1B will not start CC04B
- 4. MS05A (M)(C) MS-V-18C, MS-V-18D, and FW-V-168 fait MS05B FW11B
- 5. ES01A (C), (1) ESAS failure to actuate (A and B), HSPS low pressure isolation ES01B MUO8A MS10SGB, TRN
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor i
Page 1 of 6
1 Operat:r Acti:ns I
i Op Test No.: 2 Scenario No.: 2 Event No.: 1 Page _2_ of D_
is Event
Description:
NI-3 SUR Amplifier failure !
l Time Position Applicant's Actions or Behavior PCRO Resume plant startup by pulling control rods in manual-sequence mode, to increase neutron flux level to point of adding heat. !
- Reactivity Manipulation for SRO-l !
PCRO Diagnose failure of NI-3 startup rate to respond to addition of positive reactivity, and report to SF.
SS/SF Declare NI-3 inoperable, evaluate Tech. Spec, implications of NI-3 channel failure, and determine that plant startup may resume as long as NI-4 remains operable.
- Per T.S.3.5.1.1, Table 3.5-1 only one Intermediate Range detector is required.
PCRO If stopped for NI 3 failure, resume plant startup by withdrawal of control rods.
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Operat:r Actions Op-Test No.: _2 Scenario No.: 2 Event No.: _.2_ Page _Q_ of 6 Event
Description:
Dropped Rod (6-1)
Time Position Applicant's Actions or Behavior TEAM Recognize asymmetric and probable dropped rod in CRG-6.
SSISF Direct action to insert the remaining rods to achieve at least 1% Ak/k shutdown per 1202-8.
31 Determine inoperable rod per Tech. Spec. Section 4.7.1.2 l
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Operater Actions Op Test No.: 2_ Scenario No.: _2_ Event No.: 3 Pa0e _4_ of 6
.k Event
Description:
IC-P-1 A trip - IC-P-1 B will not start Time Position - Applicant's Actions or Behav!or EAM Diagnose IC-P-1 A trip and note failure of IC-P-1B to automatically start.
- Dispatch AO to IC-P-1 A/B pumps.
PfaQ Attempt to start IC-P-1B,
- IC-P-1 B will not start, ff implement EP 1202-17 and direct attempts to restore ICCW, while team (STA) monitors CRDM temperatures and pressurizer level.
PfdQ Diagnose loss of sealinjection due to closure of MU-V-1 A/B at CRD Coolant Outlet Hi Temp. of 160*F.
133E Prepare team for a manual reactor trip required by EP 1202-8 due to high CRDM temperatures.
- it will take approximately 5 minutes for CRDM temperatures to exceed 180*F.
Eggf Direct team to manually trip reactor when more than one CRDM stator temperature exceeds 180 F, and perform ATP 1210-1.
TEAM - Manually trip the reactor and perform 1210-1 immediate actions.
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4 Operat:r Actions ]
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Op Test No.: 2 Scenario No.: 2 Event No.: _4__ Page _5_, of 6 l I
Event
Description:
MS-V 18C, MS-V 18D, and FW-V-168 fail Time Position Applicant's Actions or Behavior )
SfeEQ Diagnose / report undesired MFW flow to B-0TSG, caused by failure of FW V-16B at 100% and stuck open MSSVs, and take action to control MFW to B-OTSG.
IEA_M Diagnose excessive primary to secondary heat transfer of the RCS due to stuck MSSVs on B OTSGs, and isolate the B-OTSG in accordance with ATP 1210-3.
- The excessive primary to secondary heat transfer is caused primarily by B-0TSG. Also FW-V-928 will NOT automatically close when HSPS
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actuates at less than 600 psig in B-OTSG, but it will close by push-button on CC.
SCRO Close FW-V-92B to isolate FW-V-168 and eventually close EF-V-30A and 30D in manual to isolate EFW. (CT)
PCRO Operate makeup valves and pumps as necessary to maintain pressurizer level after trip. Open or verify open MU-V-14B prior to starting MU-P-1C for additional makeup flow.
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l Operattr Acti:ns Op Test No.: 2 _ Scenario No.: 2 Event No.: 5 Page 6 of 6 Event
Description:
ESAS failure to actuate (A and B), HSPS low pressure isolation failure Time Position Applicant's Actions or Behavior EfaQ When RCS pressure decreases below 1600 psig, diagnoses failure of ESAS to automatically aduate HPI (or if pressurizer level decreases below 20 inches), and manually initiate HPl - two pumps, full flow, ES alignment per 1210 prior to loss of 25'F subcooled margin. (CT)
HPl may be actuated simply by depressing 1600# ESAS Train-A and Train-B manual aduation pushbuttons on CC and CR; or by initiating HPl at the component level in accordance with 1210-10.
- If sufficient HPl is not initiated to compensate for RCS shrink,25'F subcooled margin may be lost.
8F/ TEAM Lead crew in verification of 1210-1 and then 1210-3 immediate actions for reactor trip and isolation of feedwater to at least the A-OTSG.
SF/ TEAM Commence 1210-3 follow-up actions after alarm review.
SS/SF Identify / declare an UNUSUAL EVENT due to the MSSV failure causing OTSG depresserization below 600 psig.
- May wait until end of scenario to complete 3 98 Recognize when the excessive primary to secondary cooling has stopped, requiring control of RCS reheat and repressurization per 1210-3.
A-OTSG pressure will stabilize after B-0TSG goes dry and stops the excessive primary to secondary heat transfer.
11 If actuated, authorize defeat / enable of manual 1600# ESAS channels to ;
throttle HPi and control RCS pressure.
jf - Direct CROs to defeat / enable manual ESAS channels and throttle HPI to stop RCS repressurization. !
PCRO - Operate HPl pumps and/or valves as necessary to throttle HPl flow and !
control RCS pressure within allowable region of 1210-10. (CT) l l
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l THREE MILE ISLAND UNIT 1 i SENIOR REACTOR OPERATOR l l- SIMULATOR EXAMINATION o ,
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SCENARIO NUMBER: 2 !
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~ THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l k 1.0 GENERAL DESCRIPTION OF SCENARIO 4
Plant is at 10 amps during a plant startup and crew continues power increase into point of adding heat.
One channel of Intermediate Range SUR f ails during power increase.
A group 6 control rod drops requiring a shutdown to 1Mk/k.
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During establishment of shutdown reactivity. IC-P-1 A trips with failure of auto / manual start of IC P 18.
- l. The total loss of Intermediate Closed Cooling Water will require a manual reactor trip due to high CRD stator temperatures, i
- l. When the unit is tripped, two main steam safety valves on the B OTSG f ail full open and FW V 16B will f ail full l open with an HSPS low pressure isolation f ailure resulting in excessive primary to secondary heat transfer.
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1600# ESAS actuation on channels 2 and 3 will not occur, requiring manual HPl actuation.
l 25'F Subcooled margin may be lost if HPl is not initiated, i
ll The B OTSG must be isolated to terminate the excessive primary to secondary heat transfer. ;
If HPl was actuated, it will require throttling / termination to control RCS pressure when the B OTSG blows dry, i
Plant events and conditions will warrant declaration of the Emergency Plan.
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j Estimated scenario time 60 Minutes i
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THREE MILE ISLAND UNIT 1 SENIOR REACTCR OPERATOR ,
l SIMULATOR EXAMINATION
2.0 REFERENCES
A. 10 CFR 55.45 Operating Test, (a) Content i B, Procedures l
- 1. 1102-2 Plant Startup
- 2. 1202-8 CRD Equipment Failure
- 3. 1202-17' Loss of intermediate Cooling System
- 4. 1210 1 Reactor Trip ;
- 5. .1203-24 Steam Leak l
- 6. 1210-3 Excessive Primary to Secondary Heat Transfer l l
- 7. -1210-10 Abnormal Transients Rules, Guides and Graphs C. TECHNICAL SPECIFICATIONS
- 1. Section 3.5.1.1, instrumentation System, Table 3.5-1 i
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION i 3.0 SCENARIO INITIAllZATION A. IC-9, Hot Zero Power,10 amps, Xenon free,640 EFPD
- 1. Use Remote Function MSR21 to OPEN AS-V-8
- 2. Close GS V-4 on PL.
B. EVENT
- 1. Event #1 RD:CFTRP C. MALFUNCTIONS
- 1. N112A NI-3 SUR Amplifier Failure; insert at 50% severity insert immediately
- 2. RD0115 Dropped Rod (6-1); assign to Remote #1
- 3. CC04A IC-P-1 A trip; assign to Remote #2 i
- 4. CC04B IC-P-1B trip; assign to Remote #3
- 5. MS05A MS-V-180 f ailure; assign at 100% severity to Event #1
- 6. MS05B MS-V-18D f ailure; assign at 100% severity to Event #1 f 7. FW11B FW-V 160 failure; assign at 100% severity to Event #1
- 8. ES01A ESAS failure to actuate A (1600#); insert immediately
- 9. ES01B ESAS failure to actuate B (1600#); inser1 immediately
.10. MUO8A MU V-16A f ails to open on ESAS; insert immed.iately
- 11. MS10SGB, TRNA: HSPS low pressure isolation f ailure; insert immediately l
- 12. MS10SGB, TRNA: HSPS low pressure isolation f ailure; insert immediately l ' D. OVERRIDES None i
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l SCENARIO PREVIEW 4.0 A. EXAMINER PREVIEW
- 1. Reactor critical at 10* amps during xenon-free plant startup with CRG-6 at 64%.
- 2. Reactor power is increased per OP 1102 2, and NI-3 startup rate amplifier fails to respond. The crew must address Tech. Specs. for the failed NI and determine no action required.
- 3. When reactor power reaches -1%, a dropped control rod will occur. The crew will have to establish 1% AVk subcriticality. (CT) l i
- 4. During establishment of shutdown reactivity IC-P-1 A will trip with auto / manual start f ailure for i IC-P-18. The crew will have to manually trip the reactor in accordance with 1202-17 Loss of Intermediate Closed Cooling System.
- 5. When the reactor is tripped, two MSSVs on the B SG will f ail 100% open. Also B OTSG will suffer an over feed due to FW V-16A failing 100% open, resulting in excessive primary to secondary tH?at transfer.
- 6. The crew must isolate the B OTSG in accordance with ATP 1210-3 to mitigate the excessive primary to secondary heat transfer. (CT)
- 7. HPl will NOT automatically actuate due to ESAS 1600# channels 2 and 3 bistable failures, requiring manual HPi initiation. MU-V-16A will fail to open on ESAS (CT)
- 9. During the scenario the appropriate E-Plan declaration will be made.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION I
B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. Reactor is critical at 10* amps during a xenon-free startup at 640 EFPD.
- 3. Plant was previously at 100% power for 65 days aad was shutdown for main turbine bearing replacement. It has been 7 days since the plant shutdown.
- 4. Operating main feedwater pump and gland steam are both on auxiliary steam.
- 5. No maintenance or surveillance is in progress.
- 6. There are no releases in progress and none scheduled for this shift.
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- 7. Continue plant startup. I l
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l THREE MILE ISLAND UN:71 SENIOR REAC1OR OPERATOR SIMULATOR EXAMINATION a
k 5.0 SEQUENCE OF EVENTS Examiner Notes and Actions Expected Operator Actions l
- 1. Initialize the simulator in accordance with Section 3.0.
! 2. Assign team positions and conduct the SHIFT 2.1 Assume assigned team positions.
BRIEFING per Section 4.B.
l 2.2 Take turnover and review plant status.
Provide SF with sign-off copy of OP 1102-2 signed
' off up to including step where reactor is taken 2.3 Assume the watch and inform examiner.
critical per 1103-8.
NOTE: Allow crew 3-5 minutes to take turnover and 2.4 PCRO - Resume plant startup by pulling control assume the watch. rods in manual-sequence mode, to increase neutron flux level to point of adding heat.
Operate ControlBoards Reactivity Manipulation for SRO-I NOTE: NI 3 Startup rate amplifier is f ailed and will 3.1 PCRO - Diagnose f ailure of NI-3 startup rate to indicate 0 on allindications. respond to addition of positive reactivity, and report to SF.
Understand Plant and System Response NOTE: Per T.S.3.5.1.1, Table 3.5-1 only one 3.2 S3]SF - Declare NI-3 inoperable, evaluate Tech.
Intermediate Range detector is required. Spec. implications of NI 3 channel failure, and determine that plant startup may resume as long Comply With and Use Tech. Specs. as NI-4 remains operable.
3.3 PCRO -If stopped for NI 3 failure, resume plant startup by withdrawal of control rods.
- 4. ICO MALFUNCTION: At -1% power and when 4.1 TEAM - Recognize asymmetric and probabie directed an examiner, activate Remote #1 to cause dropped rod in CRG-6.
dropped rod in CRG-6.
Understand and Interpret Annunciators and Alarms l l
Demonstrate Supervisory Ability 4.2 SS/SF - Direct action to insert the remaining Comply With and Use Procedures rods to achieve at least 1% Ak/k shutdown per 1202-8.
Comply with and Use Tech. Specs. 4.3 SS - Determine inoperable rod per Tech. Spec.
Section 4.7.1.2
5.0 ICO-MALFUNCTION
During the insertion of control 5.1 TEAM - Diagnose IC P 1 A trip and note failure
- rods to establish shutdown reactivity, activate of IC-P-1B to automatically start. Dispatch AO to L Remote #2 and #3 to cause a trip of IC-P-1 A and IC-P-1 A/B pumps.
IC-P-18.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION k Operate ControlBoards 5.2 P_QB2- Attempt to start IC-P 1B, report to SF Communicate and Interact With the Crew that IC-P 1B will not start.
Demonstrate Supervisory Ability 5.3 SE -Implement EP 1202-17 and direct attempts to restore ICCW, while team (STA) monitors CRDM temperatures and pressurizer level.
NOTE: It will take approximately 5 minutes for CRDM 5.4 P_QBQ - Diagnose loss of seal injection due to temperatures to exceed 180 F. closure of MU V-1 A/B at CRD Coolant Outlet Hi Temp. of 160*F.
ICO-ROLE PLAY as AO: 11 directed to check IC-P-1 A 5.5 $$/EE - Prepare team for a manual reactor trip .
report that the monitor is too hot to touch required by EP 1202-8 due to high CRDM If directed to check IC-P-1 A breaker, report the temperatures. .
i breaker is tripped.
If directed to check IC-P-1B or its breaker, report nothing obvious is wrong.
Demonstrate Supervisory Ability Comply With and Use Procedures 5.6 $$/SE - Direct team to manually trip reactor when more than one CRDM stator temperature exceeds 180*F, and perform ATP 1210-1.
Operate the ControlB;ards 5.7 TEAM - Manually trip the reactor and perform
- f. Comply With and Use Procedures 1210-1 immediate actions.
6.0 NOTE: When the reactor is tripped. FW-V-16B f ails 6.1 SCRO - Diagnose / report undesired MFW flow to 100% open and two MSSVs stick 100% open. B-OTSG, caused by f ailure of FW V 168 at 100% and stuck open MSSVs. and take action to Diagnose Events and Conditions control MFW to B-0TSG.
Understand Plant and System Response Operate ControlBoards NOTE: The excessive primary to secondary heat 6.2 TE AM - Diagnose excessive primary to transfer is caused primarily by B-OTSG. secondary heat transfer of the RCS due to stuck Also FW-V 928 will NOT automatically close when MSSVs on B OTSGs. and isolate the B OTSG in HSPS actuates at less than 600 psig in B-OTSG. but it accordance with ATP 1210-3 including the will close by push-button on CC. following:
Diagnose Events and Conditions Understand Plant and System Response ICO-MALFUNCTION: If at any time you are directed as 6.3 ggRQ Close FW-V-92B to isolate FW-V-16B AO to isolate FW-V 16B and to locally close FW V- and eventually close EF-V-30A and 30D in 168, after -2 minute time delay change severity of manual to isolate EFW. (CT)
FW11B to 0% over 60 seconds to simulate FW V-16B localisolation.
Operate ControlBoards a
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION i
i k NOTE: When RCS pressure decreases below 1640 6.4 EfaQ - Operate makeup valves and pumps as psig, ESAS (HPI) will NOT automatically actuate necessary to maintain pressurizer level after trip.
due to f ailed bistables for channels 2 and 3. Open or verify open MU V-14B prior to starting MU-P 1C for additional makeup flow.
Understand Plant and System Response
- Operate ControlBoards NOTE: HPl may be actuated simply by depressing 6.5 EfaQ-When RCS pressure decreases below 1600# ESAS Train-A and Train-B manual actuation 1600 psig, diagnoses failure of ESAS to pushbuttons on CC and CR; or by initiating HPI at the automatically actuate HPl (or if pressurizer level l component level in accordance with 1210-10. decreases below 20 inches), and manually l initiate HPl - two pumps, full flow, ES alignment l NOTE: 11 sufficient HPI is not initiated to compensate for per 1210 orior to loss of 25'F subcooled j RCS shrink,25'F subcooled margin may be lost. margin.
[_ CriticalTask l Comply With and Use Procedures .
Operate ControlBoards l
Demonstrate Supervisory Ability 6.6 SF/ TEAM - Lead crew in verification of 1210-1 and then 1210-3 immediate actions for reactor trip and isolation of feedwater to at least the A-OTSG.
6.7 SF/ TEAM - Commence 1210-3 follow-up actions after alarm review.
f NOTE: Examiner may wait for scenario termination. 6.8 33!.EF._-Identify / declare an UNUSUAL EVENT due to the MSSV failure causing OTSG Comply With and Use Procedures depressurization below 600 psig.
NOTE: A-OT,SG pressure will stabilize after B-0TSG 6.9 TEM!- Recognize when the excessive primary goes dry and stops the excessive primary to to secondary cooling has stopped, requiring secondary heat transfer. control of RCS reheat and repressurization per 1210-3.
- Understand Plant and System Response Comply With and Use Procedures
[ Operate ControlBoards 6.10 $$-If actuated, authorize defeat / enable of manual 1600# ESAS channels to throttle HPI l and control RCS pressure.
Demonstrate Supervisory Ability 6.11 SF - Direct CRos to deteat/ enable manual ESAS channels and throttle HPl to stop RCS repressurization.
Understand Plant and System Response 6.12 PCRO - Operate HPI pumps and!or valves as Operate ControlBoards . necessary to throttle HPI flow and control RCS pressure within allowable region of 1210-10.
(CT) 3 Page 9 of 10
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR I
SIMULATOR EXAMINATION
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- 6. - TERMINATION POINT:
1 When all the following conditions exist' I
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- Crew has had sufficient opportunity to perform all applicable critical tasks. l
- 2. . ATP 1210-3 follow-up actions are in progress. !
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- 3. The plant is stabilized or being cooled down with HPl terminated.
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- 4. All examiners agree on termination.
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Sc narioOutline Facility: TMI Scenario No.: 3 Op-Test No.: 3 Examiners: Operators:
Objectives: Plant is operating at or near 100% power, steady state condition. Pressurizer level transmitter RC-1-LT-1 will fait downscale due to a cracked weld causing MU 17 to go full open and creating a small RCS leak in the RB. The crew will need to take manual control of MU V-17 and swap the controlling channel to RC1 LT-3. A 15 gpm leak will deveiop in "B' letdown cooler causing an increase in ICCW surge tank and increase in RM-L-9 indication. The crew will be required to isolate the "B" letdown cooler. A large feed water leak (1.2E6 lbm/hr), inside containment will develop over 15 minutes. The crew will need to take adion in accordance with 1203-24, Steam Leak, to commence a reactor shutdown or manually trip the reactor due to the size of the break. When the reactor trips the "B" OTSG Atmospheric Dump Valve will fail open, which will require manual isolation. The crew should identify Excessive Primary to Secondary Heat Transfer and perform actions of 1210-3 to isolate "A" OTSG. MU-V 18 will fail open on ESAS and will require to be isolated by closing MU-V-17 and MU-V-217. The crew will need to bypass / defeat ESAS and terminate HPI to stop RCS repressurization. An E-Plan declaration should be made.
Initial Conditions: IC-17,100% power, equilibrium Xenon,640 EPFD Red tag EF-P-2B in pull-to-lock and rack out breaker using Remote Function FWR13 Turnover: Plant is at 100% power, steady state conditions,640 EFPD. RCS boron is 36 ppm and the BAMT is the Tech. Spec. Tank. No maintenance or surveillance is in progress. There are no releases in progress and none scheduled for this shift. FW-P 1 A is to be removed from service for a bearing inspection. You are to reduce power to 60% in preparation for removing FW-P-1 A.
k' Event Malf. No. Event Type
- Event No. Description
- 1. RC04A (I), (N), Pressurizer LT-1 failure,
- 2. TH09 (C) Pressurizer steam space leak
- 3. MU188 (C) B let down cooler leak
- 4. FWO98 (M) (C) FWline break inside containment
- 5. MS07B (C), (1) Atmospheric Dump Valve failure, MS V 48, MU isolation valve fails as is, MU V 18
- (N)ormal, (R)eactivity, (I)nstrument, (C)omponent, (M)ajor Page 1 of 6
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Operater Actions Op Test No.: 3 Scenario No.: 3 Event No.: 1 Page _2_ of _0_
Event
Description:
Pressurizer level transmitter failure Time Position Applicant's Actions or Behavior fgBEH Diagnose Pressurizer LT-1 failure by alarms and indications and diagnose increased RB radiation levels by alarms and indications.
g_ Perform immediate Actions for the LT-1 failure per 1202 29; l
l - Transfer MU-V 17 to Hand and adjust MU flow, l
- Select the altemate Pressurizer level and temperature instruments.
- MU-V-17 will open in auto due to drop in indicated Pressurizer level and should be adjusted in Hand to stabilize actuallevel.
RE Verify or direct immediate Actions of 1202-29, Section D, for the failed Pressurizer levelinstrument.
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Operat:r Actions Op-Test No.: 3 Scenario No.: 3 Event No.: _2_ Page )., of 6 j k Event
Description:
Pressurizer steam space leak Time Position Applicant's Actions or Behavior CREW Diagnose probable RCS leak into RB by increased Radiation levels.
Implement ARP C-1-1 and EP 1202-12 for the excessive Radiation levels.
31/RIN Evaluate RCS leakage for rate implications. Refer to Tech. Spec. 3.1.6 limits and required actions.
p This size leak does not warrant a rapid Shutdown or trip, but the SS may elect to begin a Plant Shutdown.
CREW Commence a one hour leak rate calculation by computer.
CREW If directed by SS, commence a Plant Shutdown.
- This size leak would only warrant an Unusual Event Per EAL US.1 of EPIP-TMI .01.
11 Evaluate EALs for possible E-Plan Event Declaration.
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Operatar Actions Op-Test No.: 3 Scenario No.: 3 Event No.: 3 Page _L, of 6 Event
Description:
B let down cooler leak Time Position Applicant's Actions or Behavior l
IE8M Respond to high ICCW surge tank level alarm, check RM-L 9, and determine that a letdown cooler leak probably exists.
Bf Direct corrective actions to isolate letdown cooler with the leak in accordance with alarm response procedures; should direct isolation of one
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l cooler at a time.
- ICCW surge tank level indication on CR will go off-scale high, and RM-L-9 count rate will tise but may level off prior to alarm setpoint. !
PCRO Close MU-V 1B and MU-V-2B to isolate RCS leak. (CT) l l
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Operat:r Actions Op-Test No.: 3 Scenario No.: 3 Evert No.: ._4_ Pagelof 6 k Event
Description:
FW line break inside containment Time Position Applicant's Actions or Behavior Igg 1 Diagnose indications of a Steam / Feed leak inside containment.
Af Refers to the Immediate actions of AP 1203-24, and instructs pane! .
operators to monitor RB pressure and temperature.
133E Directs an operator to check RB video monitors for evidence / location of leak. Instructs PCRO to place Reactor Building Emergency Cooling in service.
SCRO Diagnoses that leak appears to be a FW leak inside Containment due to increased "A" loop FW flow, lower A OTSG level, increased ATc and no large decrease in Generated MWE.
PCRO Places Reactor Building Emergency Cooling in service.
Bj!/ff With RB pressure stillincreasing, directs Crew to begin a Plant Shutdown at an appropriate rate. Critical Task
- Crew may elect to trip the reactor due to the pressure >2 psig, FW le&%.
SF/PCROI Begin a Plant Shutdown at the requested rate per OP 1102-10.
SCRO - As FW Leak increases, RB pressure will increase to >2 psig resulting in an ES Actuation, and an Auto Reactor Trip if not manually tripped SS/SF RB pressure increases to 4 psig ES Actuation setpoint.
- Direct that a Plant announcement be made, and Reactor manually tripped.
PCRO Manually trips Reactor as directed by the SS/SF.
TEAM Respond to the Reactor Trip per ATP 1210-1.
SCRO Verifies HSPS EFW actuation on high RB pressure and dispatches an AO to EFW Pumps.
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Operat:r Actions Op Test No.: 3 Scenario No.: 3 Event No.: 5 Page 6 of 6 l
.k Event Descnption: Atmospheric Dump Valve failure, MS-V-48, MU isolation valve fails as is, MU-V 18 Time Position Applicant's Actions or Behavior g Diagnose Excessive Primary to Secondary Heat Transfer condition existing in A OTSG, and shtgg SCRO to isolate the A OTSG.
SCRO Isolates A OTSG:
(1) Closes FW V 5A,92A,16A and 17A (2) Closes MS-V 3D,3E,3F,1B,1 A and 4A.
(3) Closes EF V 30A,30D and MS-V-2A.
Critical Task EFW will actuate on HSPS high RB pressure and possibly low OTSG i level.
TEAM Recognizes MS-V 4B is still open and directs the EFW AO to isolate it by closing MS V-15. Critical Task TEAM Complete immediate manual actions of ATP 1210-1 SS/SF/STA Recognize that Excessive Primary to Secondary Heat Traasfer Event is terminated and direct PCRO to bypass / defeat ESAS and terminate HPI to stop the RCS repressurization.
PCRO Bypasses / defeats both Trains of ESAS, and then terminates HPl per OP l
( 1105-3:
(1) Secure MU-P 1C (2) Close MU V-16C and MU-V 16D (3) Throttle close MU V 16A/16B l (4) Open MU V 36/37 when required (5) Close MU-V 14A and MU-V 148 as required.
Critical Task g Verify / direct crew performance of ATPs 1210-1/1210-3 immediate manual actions.
RS Declare an Emergency Plan Event within 15 minutes of verification of EAL.
TEAM Perform the required follow-up actions of ATPs 1210-3 and 1210-1.
SS/SF/STA Provide PCRO/SCRO with adequate data to control *B* OTSG level / pressure to stabilize RCS parameters and initiate a Cooldown.
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, i THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l
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1 SCENARIO NUMBER: 3 i
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EXAMINERS: j l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1.0 - GENERAL DESCRIPTION OF SCENARIO Plant is at 100% power EOL, Equilibrium Xenon, with EF-P-2B is out of service.
Crew will commence a power reduction remove a Feedwater pump from service.
Pressurizer level transmitter LT 1 f allure with a small (2 gpm) leak to the RB causing MU-V-17 to go full open in auto.
After the power reduction, a leak develops in "B" letdown cooler which requires isolation.
' After letdown cooler has been isolated, a large FW leak on "A" OTSG occurs inside containment. j This leak results in a 4 psig RB ES actuation and a manual or automatic Reactor trip.
l MU V-18 will fail to close on ESAS.
The "A" OTSG must be isolated.
MS-V-4B will f ail open. This leak is isolated by closing MS-V-15B.
The ESAS Actuation is bypassed and HPI secured to prevent repressurization.
A Plant Cooldown is commenced a "A" OTSG using the Turbine Bypass Valves.
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( Estimated scenario time - 60 Minutes t
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION k 2.0 ' REFERENCES A. 10 CFR 55.45 Operating Test, (a) Content B. Procedures
-1. --1102 ' Plant Shutdown
- 2. 1202-29 Pressurizer System Failures
- 3. 1104-2 Makeup and Purification i
- 4. 1210-1 Reactor Trip 5, 1203-24 Steam Leak
- 6. '1210-3 Excessive Primary to Secondary Heat Transfer
- 7. 1105-03 Engineered Safeguards ,\ctuation System !
- 8. 1210-10 Abnormal Transients Rules, Guides and Graphs
- 9. EPIP-TMI. 01 Emergency Classification and Basis
.f C. TECHNICAL SPECIFICATIONS t !
- 1. Section 3.1.6, Leakage i
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION I 3.0 - ~ SCENARIO INITIAllZATION l A. 10-17,100% power, equilibrium Xenon,640 EPFD
- 1. Red tag EF-P-28 in pull-to-lock and rack out breaker using Remote Function FWR13 l B. EVENT
- 1. NINIC<10, Assign to Event #4 C. MALFUNCTIONS
- 1. RC04A PZR LT-1 f ailure; assign at 100% over 60 seconds to Remote #1 l
- 2. TH09 PZR steam space leak; assign at 0.1% severity (2 gpm) over 60 seconds *o Remote #2
- 3. MUO6 MU isolation valve f ails as is, MU-V-18, set at 100% severity and activate immediately.
- 4. MU18B B-Letdown cooler leak (15 gpm); assign at 100% severity to Remote #3.
- 5. FWO9B FW line break inside containment, "A" OTSG, set at 20% severity over 900 seconds to Remote #4
- 6. MS078 Atmospheric Dump Valve failure, MS-V-48, set at 100% severity, and place on Event #1.
D. OVERRIDES i
NONE E, MONITOR
- 1. Set MSK2609A to 14.7; Aux. Boiler pressure l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l 4.0 SCENARIO PREVIEW A. EXAMINER PREVIEW l
- 1. Plant is operating at or near 100% power, steady state condition.
- 2. Pressurizer level transmitter RC 1-LT-1 will fait downscale due to a cracked weld causing MU-17 to go full open and creating a small RCS leak in the RB.
- 3. . The crew will need to take manual control of MU V 17 and swap the controlling channel to RC1-LT-3.
- 4. A 15 gpm leak will develop in "B" letdown cooler causing an increase in ICCW surge tank r.nd increase in RM-L-9 indication.
- 5. The crew will be required to isolate the "B" letdown cooler.
- 6. ' A large feed water leak (1.2E6 lbm/hr), inside containment will develop over 15 minutes.
- 7. The crew will need to take action in accordance with 1203-24, Steam Leak, to commence a reactor shutdown or manually trip the reactor due to the size of the break.
- 8. When the reactor trips the "B" OTSG Atmospheric Dump Valve will fail open, which will require manual isolation.
- f ' 9. The crew should identify Excessive Primary to Secondary Heat Transfer and perform actions of 1210-3 to isolate "A" OTSG.
- 10. MU-V 18 will fail open on ESAS and will require to be isolated by closing MU-V-17 and MU V-217.
- 12. An E-Plan declaration should be made.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. Plant is at 100% power, steady state conditions,640 EFPD.
- 3. No maintenance or surveillance is in prog ress.
- 4. There are no releases in progress and none scheduled for this shift.
' 5. FW-P-1 A is to be removed from service for a bearing inspection. You are to reduce power to 60% in preparation for removing FW P 1 A.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
! 5.0 SEQUENCE OF EVENTS Examiner Notes and Actions 'Exnected Operator Actions
- 1. Initialize the simulator in accordance with Section 3.0.
- 2. Assign team positions and conduct the SHIFT 2.1 Assume assigned team positions. ,
l BRIEFING per Section 4.B. l l
2.2 Take turnover and review plant status.
NOTE: Allow crew 3 5 minutes to take turnover and l assume the watch. 2.3 Assume the watch.
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- 3. ICO-MALFUNCTION: After power reduction is ' 3.1 .Q.RjW - Diagnose Pressurizer LT 1 f ailure by '
commenced and when directed by examiner, alarms and indications and diagnose increased activate Remote #1 and 2, to cause Pressurizer RB radiation levels by alarms and indications.
LT 1 fallure and level tap leak of about 2 gpm into cor;tainment.
Understand and Interpret Annunciators Diagnose Events and Conditions Understand Plant and System Response NOTE: MU-V-17 will open in auto due to drop in 3.2 PCRO - Perform Immediate Actions for the indicated Pressurizer level and should be LT-1 failure per 1202-29:
adjusted in Hand to stabilize actual level. A. Transfer MU-V-17 to Hand and adjust MU ,
flow.
Operate ControlBoards B. Select the alteraate Pressurizer level and Comply With and Use Procedures temperature instruments.
Demonstrate Supervisory Ability 3.3 SF - Verify or direct immediate Actions of 1202-29, Section D, for the failed Pressurizer i level instrument.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
'l Examiner Notes and Actions Expected Operator Actions NOTE: This very small leak will not cause excessive 3.4 GBEE - Diagnose probable RCS leak into RB RB pressure or temperature increases. by increased Radiation levels. Implement ARP-C-1-1 and EP 1202-12 for the excessive Diagnose Events and Conditions Radiation levels.
NOTE: This size leak does not warrant a rapid 3.5 SS/STA/ CREW - Evaluate RCS leakage for Shutdown or trip, but the SS may elect to begin rate implications. Refer to Tech. Spec. 3.1.6 a Plant Shutdown. limits and required actions.
Comply With and Use Tech. Specs.
3.6 QBg.W- Commence a one hour leak rate calculation by computer.
NOTE: This size leak would only warrant an Unusual 3.7 CREW If directed by SS, commence a Plant
( Event Per EAL US.1 of EPIP-TMI .01. Shutdown.
3.8 ff Evaluate EALs for possible E-Plan Event Declaration.
4.0 'ICO-MALFUNCTION
When directed by 4.1 M- Respond to high ICCW surge tank examiner, activate Remote Key #3, to cause level alarm, check RM-L-9, and determine that leak in "B" letdown cooler. a letdown cooler leak probably exists.
UnderstandandInterpret Alarms Diagnose Events and Conditions NOTE: ICCW surge tank level indication on CR will go 4.2 SE - Direct corrective actions to isolate letdown off-scale high, and RM-L-9 countrate will rise cooler with the leak in accordance with alarm but may level off prior to alarm setpoint. response procedures; should direct isolation o
- l. one cooler at a time.
. Operate ControlBoards
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION I Examiner Notes and Actions Exnected Operator Actions ICO: Monitor ISD point for ICCW Surge Tank 4.3 EG.R_QR - Close MU-V 1B and MU-V-2B to
' Level, CCLICST.
ICO: ROLE PLAY as AO or Rad-Con Tech. if requested to check ICCW surge tank level, and ICCW surge tank level on ISD is 232 inches, report:"a steady stream of water is coming from ICCW surge tank overflow to floor drain", until leak is isolated (ISD point MUXFMF18 (2) indicates 0).
- 5. ICO-MALFUNCTION: When directed by 5.1 Tggnl- Diagnose indications of a Steam / Feed examiner, activate Remote #4, to initiate a leak inside containment.
1.2E6 LM/HR FW leak on "A" OTSG inside containment over 15 minute period.
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Understand and Interpret Annunciators and Alarms Diagnose Events and Conditions Comply With and Use Procedures 5.2 SE- Refers to the immediate actions of AP 1203-24, and instructs panel operators to monitor RB pressure and temperature.
19.QiIf questioned about steam leak inside 5.3 EH@E - Directs an operator to check RB video l Containment, inform RO that there is evidence monitors for evidence / location of leak.
of fog in both of the D Rings, although, it Instructs PCRO to place Reactor Building appears to be thicker inside "A" D-Ring. Emergency Cooling in service.
Diagnose Events and Conditions 5.4 BCRQ Q - Diagnoses that leak appears to be a Understand Plant and System Response FW leak inside Containment due to increased
, "A" loop FW flow, lower A OTSG level, increased ATc and no large decrease in Generated MWE.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l
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Examiner Notes and Actions Expected Operator Actions I I
Operate ControlBoards 5.5 R EC_R_Q- Places Reactor Building Emergency Cooling in service.
l NOTE: Crew may elect to trip the reactor due to the 5.6 $$!$E- With RB pressure still increasing, pressure >2 psig FW leak. directs Crew to begin a Plant Shutdown at an l
appropriate rate. j Demonstrate Supervisory Ability Critical Task Operate ControlBoards
- 6. NOTE: As FW Leak increases, RB pressure will 6.1 SF/PCRO/SCRO - Begin a Plant Shutdown at increase to >2 psig resulting in an ES Actuation, the requested rate per OP 1102-10.
and an Auto Reactor Trip if not manually tripped.
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- 7. RB pressure increases to 4 psig ES Actuation 7.1 ESLEE - Direct that a Plant announcement be setpoint. made, and Reactor manually tripped.
Demonstrate Supervisory Ability Comply With and Use Procedures I
NOTE: When Reactor trips, the "B" Atmospheric Dump 7.2 PQ,R_Q - Manually trips Reactor as directed by !
Valve will fail open and require Auxiliary the SS/SF.
Operator action to control / isolate. Verify that MALFUNCTION MS07A activates on Event #1.
7.3 EE/EE- Direct the use of Global Alarm
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l lCO: As an AO, MS-V-4A may be throttled 7.4 TEAM - Respond to the Reactor Trip per ATP
!~ locally by using Remote Functions MSR25 and 1210-1.
l MSR26 or isolated with Remote Function MSR14.
Comply With and Use Procedures
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THREE MILE ISLAND UNIT 1 l
SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l
I Examiner Notes and Actions Expected Operator Actions i
7.5 SfRQ Verifies HSPS EFW actuation on high RB pressure and dispatches an AO to EFW l Pumps.
1 Diagnose Events and Conditions 7.6 SF - Diagnose Excessive Primary to Understand Plant and System Response Secondary Heat Transfer condition existing in A j OTSG, and glirst SCRO to isolate the A I OTSG. I NOTE: EFW will actuate on HSPS high RB pressure 7.7 SCRO -Isolates A OTSG:
and possibly low OTSG level. (1) Closes FW-V-5A,92A,16A and 17A 1 (2) Closes MS-V-3D,3E,3F,1B,1 A and 4A. I Comply With and Use Procedures (3) Closes EF-V-30A,30D and MS-V-2A.
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Operate ControlBoards Critical Task
{ 100: MS-V-4B may be isolated by closing MS-V-15B. If requested set Remote Function 7.8 TEAh!- Recognizec MS-V-48 is still open and directs the EFW AO to isolate it by closing MSR14 to close. MS-V 15.
Critical Task Communicate andinteract With Crew ICO: If crew requests to throttle MS-V-4B, 7.9 TE AM - Complete immediate manual actions ensure that Remote Function MSR14 is open, of ATP 1210-1 and then set Remote Function MSR27 and MSR28 to the requested value.
7.10 PCRO -Informs SS/SF of MU-V-18 f ailing open and closes pr verifies close MU-V-17 and M U-V-217.
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THREE MILE ISLAND UNIT 1 !
l SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l 1
I-Examiner Notes and Actions Eve =d Onorator Actions i
Diagnose Events and Conditions 7.11 SS/SF/STA - Recognize that Excessive Understand Plant and System Response Primary to Secondary Heat Transfer Event is Operate Contro/ Boards terminated and direct PCRO to bypass / defeat -I
-Demonstrate Supervisory Ability ESAS and terminate HPl to stop the RCS
! repressurization. I l
l Operate ControlBoards 7.12 EGRQ - Bypasses / defeats both Trains of I I
ESAS, and then terminates HPl per OP 1105-3: i l
(1) Secure MU-P-1C i (2) Close MU-V-160 and MU-V 16D (3) Throttle close MU-V-16A/16B (4) Open MU-V-36/37 when required I (5) Close MU V 14A and MU-V-14B as required.
Critical Task Demonstrate Supervisory Ability 7.13 SE-Verify / direct crew performance of ATPs 1210-1/1210-3 immediate manual actions.
l NOTE: Declaration will probably be made at end of 7.14 SS - Declare an Emergency Plan Event within scenario. 15 minutes of verification of EAL.
Comply With and Use Procedures 7.15 TE AM - Perform the required follow-up actions of ATPs 1210-3 and 1210-1.
Operate ControIBoards 7.16 SS/SF/STA - Provide PCRO/SCRO with adequate data to control "B" OTSG level / pressure to stabilize RCS parameters and initiate a Cooldown.
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r 1-THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR i
SIMULATOR EXAMINATION
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k 6. TERMINATION POINT 1
When all of the following conditions exist: ;
,'- 1. Crew has had sufficient opportunity to perform all applicable Critical Tasks.
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- 3. HPI is terminated.
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- 4. Plant is stabilized or being cooled down with the A OTSG isolated. I i
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- 5. All examiners agree on termination. l i
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Sctnrris Outline Facility: TMI Scenario No.: 5 Op-Test No.: 5 Examiners: Operators:
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Objectives: Plant at 100% power, steady state, with EG-Y-1 A out of service for an oil change. EHC P-1 A trips. EHC-P-1B fails to automatically start requiring the team to promptly start the standby pump to prevent turbine trip. (CT) After the plant has been stabilized, the selected FW temperature transmitter will fail to 300*F over a period of time resulting in a SASS mismatch. (CT) A fault develops of 230KV Bus #8, causing a loss of Bus #8, undervoltage on the 1E 4KV bus, and loss of MU-P-18. After the team responds to the loss of Bus #8, an excitation fault on the generator occurs resulting in a trip of the generator, the ma in turbine and the reactor. Shortly after the reactor trip, a grid fault causes a loss of 230KV bus #4, re suiting in a loss of offsite power. ED-Y-1A trips when it starts resulting in a station blackout. Emergency Feedwater HPSP signal failures require manual initiation of EFW flow to the OTSGs in orderin maintain primary to secondary heat removal. (CT) The SBO dieselis always available for use. One 230KV bus and normal AC power may be restored as time permits.
Initial Condition 3: IC-17,100% power, steady state, equilibrium Xenon,640 EFPD. Following initialization perfcrm the following: Red tag EG-Y-1B out of service. Transfer EG-Y 1B Start Switch to manual. Place red Sticker tags on START push button on CR. Red tag G11-02 breaker controlin pull-to-lock and rack out breaker using Remote Function EGR02. Trip fuel racks using EGR29. Close air start isolation valve EG-V-15B using EGR31.
Turnover: The plant is at 100% power, steady state,640 EFPD. RCS boron is 36 ppm and the BAMT is the Tech. Spec. tank. EG-Y-1B is out of service for oil change. It has been out of service for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and is expected to be retumed to service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. No other maintenance or surveillance is in
( progress. Thers are no releases in progress and none scheduled for this shift. Continue 100% power operation.
Event Malf. No. Event Type
- Event 4
No. Description
- 1. TC10A (C) (tJ) EHC-P-1 A trip
- 3. ED18B (C) 230 KV Bus #8 Fault
- 4. EG03 (M) (C) Main Generator Excitation failure
- 5. ED18A (M) (C) 230 KV Bus #4 Fault EG07A EG Y 1 A trip
- (N)orrnal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor Page 1 of 7
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! Op-Test No.: 5 Scenario No.: 5 Event No.: 1 Page 2_ of _I k Event
Description:
EHC-P-1 A trip Time Position Applicant's Actions or Behavior i IEAM Announce trip of EHC-P-1A from computer alarm.
AfgBQ Recognize auto-start failure of EHC-P-1B and perform manual start to
! prevent a turbine trip on low EHC pressure.
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Operatir Actions Op Test No.: 5 Scenario No.: 5 Event No.: 2 Pagelof 7 b Event
Description:
Feedwater temperature failure, SP-5-TE2 Time Position Applicant's Actions or Behavior l
SCRO Diagnose a FW temperature failure.
SCRO Place ICS FW Demands to Hand to correct FW flow.
Af Verify / direct SCRO to select attemate FW temperature instrument.
j SCRO Retum ICS to automatic.
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Operat r Acti ns Op-Test No.: 5 Scenario No.: 5 Event No.: 3 Pagelof 7 k Event
Description:
230 KV Bus #8 Fault Time Position Applicant's Actions or Behavior IEAM Diagnose loss of 230KV Bus 8 and 1E 4160V causing a loss of MU-P-18.
PCRO Respond to the loss of RC makeup and seat injection per 1203-15 as follows:
- 1. Close MU-V-32
- 2. Start MU-P-1 A, DR-P-1 A and DC-P-1 A. j
- 3. Slowly reesu.blish sealinjection.
gf Verify / direct 1203-15 manual actions.
- 1202-38,1203-19 and 1203-30 should also be addressed. l l
gg!/gf Consult Tech. Specs. And determine that with an Aux. Transformer and ]
EG-Y-1B out of service, the plant must be in Hot Shutdown within 12 I
hours.
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Operatrr Actions Op-Test No.: 5 Scenario No.: 5 Event No.: !_ Page,J_ of 7 I Event
Description:
Main Generator Excitation failure Time Position Applicant's Actions or Behavior IEAM Diagnose and respond to the reactor trip in accordance with ATP 1210-1 IEAM Perfomi ATP-1210-1 immediate actions except that a second make pump (MU-P-1C) will not have power for start, unless the SBO diesel was already used to power 1E 4160V Bus TEAM Perform necessary actions to begin to stabilize the plant on natural circulation.
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Operatar Actirns Op-Test No.: 5 Scenario No.: 5 Event No.: 5 Page 6 ofJ k Event
Description:
230 KV Bus #4 Fault, EG-Y-1 A trip Time Position Applicant's Actions or Behavior TEAM Diagnose the loss of station power and respond in accordance with EP-1202-2.
TEAM Diagnose trip of EG-Y-1 A.
PCRO Attempt to start EG-Y-1 A per 1202-2.
EG-Y-1 A will trip upon start, but if correct actions are taken per EP 1202-2, the diesel may be recovered.
SCRO Verify EFW actuation loss of all RC pumps with at least EF-P 1 running, but note no EFW to OTSGs.
HSPS setpoints for OTSG level control on loss of RC pumps will fail at 0% requiring manualinitiation of EFW flow.
j Operate EF-V-30A/B/C/D as necessary to establish EFW flow to both EQBO OTSGs, and increase levels towards 50% to establish RCS natural j circulation per AT 1210-10. (CT) l -
Sufficient EFW flows must be maintained until OTSG level setpoints (50%) are reached, after which OTSG pressure must be controlled to
{ maintain heat sinks for natural circulation.
g Identify / declare the appropriate event (EAL) and implement the Emergency Plan.
l SF/SCRO Verify RCS natural circulation core cooling in accordance with ATP 1210-10 and EP 1202-2.
Th and incore T/C's may increase until core Delta-T reaches its maximum point as natural circulation flow is established, but should then be stable or decreasing SCRO Throttle EFW flows to OTSGs in accordance with ATP 1210-10 as necessary to prevent or stop RCS excessive cooling, which would result in RCS depressurization, loss of pressurizer level and RCS subcooled margin. (CT)
- EFW should be throttled by EF V 30A/B/C/D, to prevent stable l
cooldown rate from exceeding 50'F/hr on RCS natural circulation, but l
OTSG levels must be increased toward 50% at a rate which will not
. cause RCS excessive cooling and depressurization. This is most criticalif neither 4160V ES bus is re-energized for RCS makeup capability to prevent loss of RCS subcooled margin i
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' Operat:r Acti:ns g Operate (dose) valves required by EP 1202 2 to conserve RCS inventory and pressurizer level.
H Contact dispatcher to investigate return of off-site power.
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'THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
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SCENARIO NUMBER: 5 ,
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.THREE MILE ISLAND UNIT 1 !
SENIOR REACTOR OPERATOR !
SIMULATOR EXAMINATION i L
I 1.0 - GENERAL DESCRIPTION OF SCENARIO
. Plant at 100% power, steady state, EOL', with EG-Y-1B out of service, i
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- A power reduction a 90% is started per dispatcher order.
1 A loss of EHC-P-1A with iailure of EHC-P-1B to automatically start resulting in decreasing EHC pressure until EHC P 1B is started manually.
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~ A SASS mismatch occurs for FW temperature requiring the alternate transmitter to be selected.
t 1 A fault develops on 230KV #8, causing loss of Bus #8, undervoltage on the 1E 4KV bus, and loss of MU-P-1B.
- An excitation fault occurs resulting in a main generator / turbine and reactor trip.
Shortly after the reactor trip, a grid fault causes a loss of 230KV Bus #4, resulting in loss of offsite power.
- ED-Y 1 A trips when it starts resulting in a station blackout.
Emergency Feedwater HSPS signal failures require manual initiation of EFW flow to the OTSGs in order to maintain primary to secondary heat removal.
The SBO diesel is always available for use.
l One 230KV bus and normal AC power may be restored. (As time permits, optional) l
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- Estimated scenario time - 60 minutes l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 2.0 . REFERENCES
. A. 10 CFR 55.45 (a) Content B. PROCEDURES
- 1. 1210-1 Reactor Trip
- 2. _. 1202-2 Loss of Station Power
- 3. 1210-10 Abnormal Transients Rules, Guides and Graphs
- 4. EPIP-TMI .01 Emergency Classification and Bases C. TECHNICAL SPECIFICATIONS
- 1. Section 3.7.2 Electrical Power 1
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- ~ _ -.- . - - . - -. .
l THREE MILE ISLAND UNIT 1 l
SENIOR REACTOR OPERATOR i SIMULATOR EXAMINATION l 1
3.0 SCENARIO INITIALIZATION A. 10-17 100% power, steady state, equilibrium Xenon,640 EFPD Following initialization perform the following:
- 1. Red tag EG-Y 1B out of service by the following
- 2. Transfer EG-Y 1B Start Switch to manual
- 3. Place red Sticker tags on START push button on CR
- 4. Red tag G11-02 breaker control n pull-to-lock and rack out breaker using Remote Function EGR02.
- 5. Trip fuel racks using EGR29.
- 6. Close air start isolation valve EG-V-15B using EGR31.
B. MALFUNCTIONS
- 1. TC10A Fault trip of EHC-P-1 A; assign to Remote #1 !
- 2. FWO4B FW temperature failure, SP-5-TE2, set at 50% severity, over 60 seconds; assign to l Remote #2
- 3. ED18B 230KV Bus #8 Fault; assi 0n to Remote #3.
- 4. EG03 Main Generator Excitation Failure; assign to Remote #4
- 5. ED18A 230KV Bus #4 Fault; ast.ign to Remote #5
- 6. EG07A EG-Y-1 A trip; assign to Event #1
{ C. REMOTE FUNCTIONS:
~ 1. ICR02 EFW level setpoint (A-SG); change to 0% immediately l
- 2. ICR04 EFW level setpoint (B-SG); change to 0% immediately D. EVENT TRIGGERS I
- 1. Event #1 EGNDGRPM(1) >850 E. 10 OVERRIDE
- 1. 05A7S10 ZDIEHCP1B(3) NAP OFF Prevent auto start of EHC-P-1B activate immediately.
F. MONITOR
- 1. MSK2609A; Set to 14.7 l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
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4.0 SCENARIO PREVIEW A. EXAMINER PREVIEW l
- 1. Plant at 100% power, steaay state, with EG-Y-1 A out of service for an oil change. )
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- 2. A power reduction a 90% is started per dispatcher order.
- 3. EHC-P-1 A trips. EHC-P 1B fails to automatically start requiring the team to promptly start the standby pump to prevent turbine trip. (CT)
- 4. After the plant has been stabilized, the selected FW temperature transmitter will f ail to 300EF over a period of time resulting in a SASS mismatch. (CT)
- 5. A fault develops of 230KV Bus #8, causing a loss of Bus #8, undervoltage on the 1E 4KV bus, and loss of MU-P-1B.
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- 6. After the team responds to the loss of Bus #8, an excitation fault on the generator occurs i resulting in a trip of the generator, the main turbine and the reactor.
- 7. Shortly after the reactor trip, a grid f ault causes a loss of 230KV bus #4, resulting in a loss of offsite power.
- 8. ED-Y-1 A trips when it starts resulting in a station blackout.
( in order to maintain primary to secondary heat removal. (CT)
- 10. The SBO diesel is always available for use.
- 11. One 230KV bus and normal AC power may be restored as time permits.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR
- SIMULATOR EXAMINATION B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. The plant is at 100% power, steady state, 640 EFPD.
- 3. EG-Y-1B is out of service for oil change. it has been out of service for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and is expected to be returned to service within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- 4. No other maintenance or surveillance is in progress.
- 5. There are no mieases in progress and none scheduled for this shift.
- 6. - Reduce power to 90% per dispatcher's order.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 5.0 SEQUENCE OF EVENTS Examiner Notes and Actions Expected Operator Actions 1, initialize the simulator in accordance with Section 3.0.
- 2. Assign team positions and conduct the SHIFT 2.1 Assume assigned tearn positions.
BRIEFING per Section 4.B.
2.2 Take tumover and review plant status.
NOTE: Allow crew 3-5 minutes to take turnover and assume the watch. 2.3 Assume the watch.
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- 3. ICO-MALFUNCTION: After the power reduction 3.1 TEAM - Announce trip of EHC-P-1 A from and when directed by Examiner, activate Remote #1 to cause a in,p of EHC-P-1 A. computer alarm.
Understand and Interpret Annunciators and Alarms
. Operate ControlBoards 3.2 S.QEQ - Recognize auto-start fallure of EHC-P-1B and perform manual start to prevent
.[ a turbine trip on low EHC pressure.
ICO-ROLEPLAY: If dispatched to EHC-P 1 A, report motor very hot and a smell of burnt insulation. l l
- 4. ICO-MALFUNCTION: After response to 4.1 SCRO - Diagnose a FW temperature failure.
EHC-P-1 A, when directed by Examiner, activate Remote #2 to cause indicated FW 4.2 SCRO - Place ICS FW Demands to Hand to temperature to decrease and a SASS correct FW flow.
mismatch.
4.3 SF - Verify / direct SCRO to select alternate FW Diagnose Events and Conditions temperature instrument.
Understand Plant and System Response Operate ControlBoards 4.4 SCRO - Return ICS to automatic.
Page 7 of 11
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR !
SIMULATOR EXAMINATION I
Examinor Notes and Actions Expected Operator Actions
- 5. ICO-MALFUNCTION: After the plant has been 5.1 TEAM - Diagnose loss of 230KV Bus 8 and 1E stabilized or when directed by the Examiner, 4160V causing a loss of MU-P-1B.
activate Remote #3 to cause a 230KV Bus 8 fault and loss of 1E 4160V bus. 5.2 PCRO - Respond to the loss of RC makeup and sealinjection per 1203-15 as follows:
Diagnose Events and Conditions l
- 1. Close MU-V-32 Comply With and Use Procedures 2. Start MU P 1 A, DR-P-1 A and DC-P-1 A. i Operate ControlBoards 3. Slowly reestablish sealinjection.
l Demonstrate Supervisory Ability 5.3 SF - Verify / direct 1203-15 manual actions.
l NOTE: 1202-38.1203-19 and 1203-30 should also be addressed.
Comply With and Use Tech. Specs. 5.4 SS/SF - Consult Tech. Specs. And determine f that with an Aux Transformer and EG-Y 1B out i
of service, the plant must be in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 6. ICO-MALFUNCTION: After Tech. Specs. have 6.1 TEAM - Diagnose and respond to the reactor
. been addressed and when directed by the trip in accordance with ATP 1210-1 Examiner, activate Remote #4 to cause a generator / turbine and reactor trip 6.2 TEAM - Perform ATP-1210-1 immediate actions except that a second make pump Comply With and Use Procedures (MU-P-1C) will not have power for start, unless Operate Contro/ Boards the SBO diesel was already used to power 1E 4160V Bus Understand Plant and System Response 6.3 TEAM - Perform necessary actions to begin to stabilize the plant on natural circulation.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
! 7. ICO-MALFUNCTION: Following completion of 7.1 TEAM- Diagnose the loss of station power and 1210-1 and when directed by the Examiner, _ respond in accordance with EP-1202-2.
activate Remote #5, to cause a loss of station power.
NOTE: EG-Y-1 A will trip upon start, but if correct 7.2 IEAM - Diagnose trip of EG-Y-1 A.
actions are taken per EP 1202-2, the diesel may be recovered. 7.3 PCRO - Attempt to start EG-Y-1 A per 1202-2.
ICOIROLE PLAY / MALFUNCTION: As AO -
dispatched to EG-Y-1 A, report overspeed trip alarm present and acknowledge alarm. If/when proper actions are directed by crew to recover EG-Y 1 A, clear malfunction EG07A to allow EG-Y-1 A reset and operation. I Operate ControlBoards
( NOTE: HSPS setpoints for OTSG level control on loss 7.4 SCRO - Verify EFW actuation loss of all RC i
of RC pumps will f ail at 0% requiring manual pumps with at least EF-P-1 running, but note no initiation of EFW flow.
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I Diagnose Events and Conditions NOTE: Sufficient EFW flows must be maintained until 7.5 PCRO - Operate EF-V-30A/B/C/D as necessary OTSG level setpoints (50%) are reached, after to establish EFW flow to both OTSGs, and which OTSG pressure must be controlled to increase levels towards 50% to establish RCS maintain heat sinks for natural circulation. natural circulation per AT 1210-10. (CT)
NOTE: Examiner may ask SF after the scenario. 7.6 33 - Identify / declare the appropriate event (EAL) and implement the Emergency Plan.
NOTE: Th and incore T/C's may increase until core 7.6 SF/SCRO - Verify RCS natural circulation core Delta-T reaches its maximum point as natural cooling in accordance with ATP 1210-10 and
! circulation flow is established, but should then EP 1202-2.
be stable or decreasing Page 9 of 11 l
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION ExaminerNotesand Actions Exnected Operator Actions NOTE: EFW should be throttled by EF-V-30A/B/C/D, to 7.7 SQRQ-Throttle EFW flows to OTSGs in prevent stable cooldown rate from exceeding accordance with ATP 1210-10 as necessary to 50*F/hr on RCS natural circulation, but OTSG prevent or stop RCS excessive cooling, which levels must be increased toward 50% at a rate would result in RCS depressurization, loss of which will not cause RCS excessive cooling and pressurizer level and RCS subcooled margin.
depressurization. Thisis most criticalif neither (CT) 4160V ES bus is re-energized for RCS makeup capability to prevent loss of RCS subcooled margin.
Understand Plant and System Response Operate ControlBoards 7.8 PQRQ - Operate (close) valves required by EP 1202-2 to conserve RCS inventory and pressurizer level.
NOTE: Return of offsite power is not required. 7.9 33 - Contact dispatcher to investigate return of l
off-site power.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
- 6. TERMINATION POINT When all of the following conditions exist:
- 1. Crew has had sufficient opportunity to perform all applicable critical tasks.
- 2. 1202-2 follow-up actions are in progress.
- 3. EG-Y-1 A or the SBO diesel is in operation.
- 4. The plant is stable on natural circulation or a slow cooldown is in progress.
- 5. All examiners agree on termination.
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THREE MILE ISLAND UNIT 1 -
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? INITIAL SENIOR REACTOR OPERATN MXAMINATION .
. JOB PERFORMANCE MEA.
.4x 1 l: TITLE: Operate-the-Station Blackout Diesel Generator TASK NUMBER: '0648000101-(~ TIME: '13 Minutes.
-3
[ EXAMINEE
' REFERENCE : EP 1202-2,-Loss of Station Power '
EVALUATION !
i METHOD: PERFORM: X SIMULATE:
1 EVALUATION LOCATION: -SIMULATOR: -X IN-PLANT: CONTROL ROOM: '
5 OPERATOR PERFORMING JPM:
EVALUATOR: / /
DATE-i K/A: 064 A4.06 l
IMPORTANCE: 3.9 '
10CFR55.45: (a) (3), (6), (8)
COMMENTS: (If results are ansatisfactory, record required data on sheet provided in back of this JPM.)
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Page 1 of 11 i
. FileM48000101. doe .
Task 0648000101 SIMULATOR
'( CONDITIONS: Initialize the Trainer at IC-21 Insert the following Malfunctions to control EFW:
FW19A - EF-V-30A, Severity 19%
FW19B - EF-V-30B, Severity 19%
FW19C - EF-V-30C, Severity 0%
FW19D - EF-V-30D, Severity 0%
Activate Malfunction ED01, Electrical Blackout.
Start MU-P-1A, DC-P-1A and DR-P-1A.
Initiate Global silence.
Freeze the Simulator.
EXAMINER PREVIEW: This JPM deals with responding to a loss of offsite power and loading the SBO Diesel on the 1C 4160v Bus.
The plant will be at a post trip condition and a loss of offsite power will occur. The Examinee will be instructed to load the SBO Diesel onto the 1C 4160v Bus, and restore components on the "J" Bus.
The Examiner will initiate Global silence, after which the ICO will maintain silence from the instructors' booth.
EFW will be throttled by the use of FW Malfunctions to prevent ESAS Initiation.
The sequence of events which'should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the 1:: formation contained in the " EXAMINEE E ~EW" section of this JPM.
- 2. Intoz:n the Examinee of the TASK CONDITIONS.
- 3. Allow +he Examinee 2-3 minutes to scan the Console.
- 4. Go to run on the simulator.
S.Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued)
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Task 0648000101 (Continued) k EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK STANDARDS: The SBO Diesel Generator has energized the 1C 4160v Bus, and GN-P-1/3/4, IA-P-4, SC-P-1B, FW-Y-1B and a Secondary River Water pump are running.
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Page 3 of 11 File 0648000101. doc
Task 0648(m0101 EXAMINEE
( PREVIEW: For this event you are assigned the duties of the 3rd CRO. The instructor / examiner will act as the Console CRO's and the SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
When you are told to begin, you will respond to the cues and/or indications which the Examiner will provide to you either verbally or via the simulator.
You are to respond as you would in the plant for :
a real condition. 1 TASK CONDITIONS: The plant was experiencing problems with the grid.
The Turbine tripped due to the power load unbalance relay actuating.
The Turbine trip resulted in a Reactor trip.
k ATP 1210-1 has been completed.
A loss of offsite power occurred 5 minutes ago.
It has been 30 minutes since the Reactor tripped.
I Page 4 of 11 File O648000101. doe
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Task 0648000101 i
l EXAMINER CUES EXAMINEE ACTIONS STANDARD Go'to Run on'the Simulator.
NOTE:
EFW will be auto throttled.
ICO:
After Global Silence is initiated silence clarms and acknowledge alarms using the instructor '
i station. Continue te acknowledge alarms.
Instruct the Examinee to *1. Examinee places the 1. The following i energize the 1C 4160v following components on Bus using-the SBO Diesel components in the Console Right in and to restore Pull-to-Lock Pull-to-Lock, i components on the "J" position by j 480v Bus in accordance rotating the NOTE:
( with EP 1202-2. Extension control Only 1J-02, 1K-02, counter-clockwise IL-02 and IM-02 and pulling are Critical.
upwards.
ISA-C2 1SA-C2 ISB-C2 ISB-C2 1J -02 1J -02 SAT 1K -02 IK -02 SAT IL -02 1L -02 SAT l 1M -02 1M -02 SAT CW-P-1C CW-P-1C CW-P-1F CW-P-1F CO-P-1C CO-P-1C CO-P-2C CO-P-2C HD-P-1C HD-P-1C l
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l Task 0648000101 EXAMINER CUES EXAMINEE ACTIONS STANDARD
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- 2. Examinee verifies or 2. Fire pump running starts a diesel as indicated by driven fire pump on PI-371 > 150#.
PL.
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- 3. Examinee starts the 3. SB0 Diesel started SdO Diesel by as indicated by i
depressing START Running and Ready pushbutton on CR To Load indicators and verifies ready lit on CR.
to load light.
- 4. Examinee verifies 4. T1-C2 on PR is T1-C2 is Open. Open.
- 5. Examinee Closes 5. G2-12 Closed on CR G2-12 by rotating as indicated by the extension red light
( control in the illuminated, i clockwise direction on CR and verifies that the 1Y SAT Substation is energized.
- 6. Examinee Closes 6. 1C 4160v Bus T1-C2 by rotating energized as the extension indicated by control in the overhead alarms clockwise direction clearing.
on PR and verifies the 1C 4160v Bus SAT energizes.
ROLE-PLAY:
If requested, inform 7. Examinee requests A0 7. AH-E-197A and Examinee that AH-E-197A to verify AH-E-197A AH-E-198 verified and AH-E-198 are and AH-E-198 running running running.
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l4 I Task 0648000101 j l
l l EXAMINER CUES EXAMINEE ACTIONS STANDARD A
+8, Examinee places the 8. The following following components on components in_the Console Right in l Pull-to-Lock Pull-to-Lock.
l position by
- j. rotating the NOTE:
Extension control counter-clockwise Only FW-Y-1B, and pulling VA-P-2B and upwards. EHC-P-1B are Critical.
GN-E-1B GN-E-1B GS-E-1B GS-E-1B SA-P-1B SA-P-1B SC-P-1B SC-P-1B FW-Y-1B FW-Y-1B SAT l l EHC-P-1B EHC-P-1B SAT I l
t VA-P-1B VA-P-1B 1 VA-P-2B VA-P-2B SAT
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f (, k *9. Examinee Closes 9. lJ-02 Closed on PR lJ-02 by rotating as indicated by
- the extension voltage indicated control in the on lJ Bus.
clockwise direction l on PR and verifies SAT l that the 1J l Substation is l enerqized.
- 10. Examinee restarts 10. Following the following components j components by restarted as rotating the indicated by Red extension controls lights illuminated in the clockwise on PLF:
L direction on PLF:
GN-P-1 GN-P-1 SAT GN-P-3 GN-P-3 SAT GN-P-4 GN-P-4 SAT
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Task 0648cDolol
, l EXAMINER CUES EXAMINEE ACTIONS STANDARD l t
) 11. Examinee Secures 11. GN-P-2 Secured as j GN-P-2 by rotating indicated by Green 5 )
the extension control Light illuminated '
in the counter- on PLF.
clockwise direction
- on PLF.
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- i. *12. Examinee starts 12. SC-P-1B started as SC-P-1B by rotating indicated by red the extension light illuminated.
control in the-clockwise SAT direction.
- 13. Examinee verifies 13. Secondary River or starts a Water pump is Secondary River started as Water pump. indicated by red light illuminated.
( SAT l
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- 14. Examinee Restarts 14. Following the following components components by Restarted as rotating the indicated by Red l extension controls Lights illuminated.
in the clockwise direction:
IA-P-4 IA-P-4 SAT FW-Y-1B FW-Y-1B SAT JPM may be terminated at this time.
(*) Denotes Critical Element.
EVALUATION:
l SAT: UNSAT:
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1 Page 8 of 11 Fde 06480001014w
Task 0648(M)olol k TITLE: Operate the Station Blackout Diesel Generator JPM NUMBER: 11.2.05.012 l TASK NUMBER: 0648000101 EXAMINEE:
l EVALUATOR: DATE:
EVALUATION OF EXAMINEE JPM: SAT: UNSAT:
COMMENTS:
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EVALUATOR
, SIGNATURE:
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Page 9 of 11 Fileo648000101 da-
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Task 0648000101 1
During a Loss of Offsite Power describe what automatic actions
( occur to supply AC power to the Substation.
l Include any required operator action.
l ANSWER:
Loss of power will start the Substation Diesels and power up ,
-PM-1 and PM-2. (0.75 point)
No action is required by the operator. (.25 point) k l
c I PEDIGREE INFORMATION:
K&A NUMBER AND VALUE: 055 EA2.03 RO:3.9 SRO:4.7 CFR: 43.5/45.13 TIME ESTIMATE: 4 MIN.
OBJECTIVE / RATING: IV.G.01.13 SRO: 3.2 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
EP 1202-2 HISTORY: QJ4G01-13-001 L
Page 10 of 11 Fileo648000101 doe
f Task 0648000101 l
A. What must the operator do at the "C" 480v Bus to restart
( .VA-P-1A after the breaker has tripped on overcurrent?
B. Why must the operator take the action described above?
kl0 ANSWER:
A. The operator must go to the breaker, and manually trip and re-close the breaker. (0.5)
B. When the breaker is manually tripped this will reset the Bell Alarm switch. (0.5)
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l PEDIGREE INFORMATION:
TIME ESTIMATE: 4 MIN.
K&A NUMBER AND VALUE: 062 K4.02 RO:2.5 SRO:2.7 CFR: 41.7 OBJECTIVE / RATING: IV.G.03.04 SRO: 3.0 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
None HISTORY: QJ4G03-04-001 Page 11 of 11 File 4648000101 Aw
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' THREE MILE ISLAND UNIT 1
( ;
_ INITIAL SENIOR REACTOR OPERATOR EXAMINATION l l
l l TITLE: Perform an Emergency Boration 1 l
TASK I
NUMBER:. 0000240501 !
l-TIME: 4 Minutes EXAMINEE l l REFERENCE : OP 1103-4, Soluble Poison Concentration Control '
ATP 1210-1, Reactor Trip. t ESAS Checklist ; l i
l EVALUATION METHOD: PERFORM: X SIMULATE:
EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: !
OPERATOR PERFORMING JPM:
l EVALUATOR: / /
DATE K/A: 000 024 AA1.17 IMPORTANCE: 3.9 10CFR55.45: (a) (3), (4), (6), (7), (8), (12)
COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.)
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Task 0000240501 SIMULATOR
( CONDITIONS: Initialize the Trainer at IC-17.
Insert Malfunction RD0201. Group 4 Rod 4 Stuck Out.
- Trip the Reactor Perform all ATP 1210-1 Immediate Manual Actions, except Emergency Boration.
Ensure the BAMT is listed on the status board as the Tech. Spec. Tank, and the Boron concentration is 17,500 ppm.
EXAMINER PREVIEW: This JPM deals with performing an Emergency Boration from the BAMT. The plant will be in a post trip state with one control red stuck out.
The Examiner will have the Examinee borate from the BAMT.
DO NOT inform the Examinee of this fact.
The sequence of events which should be followed in the conduct of this JPM is as follows:
k
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
)
1 (Continued)
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Page 2 of 8 Files 0000240501. doe
Task 0000240501 l (Continued) l l EXAMINER l PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK STANDARDS: The BAMT is lined up supplying injection to the makeup tank through MU-V-51.
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Task 0000240501 EXAMINEE
(' PREVIEW: For this event you are assigned the duties of the Primary CRO. The instructor / examiner will act as the Secondary CRO and the SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
You are expected to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to verify proper completion.
When you are told to begin, you will respond to the cues and/or indications which the Examiner will provide to you either verbally or by the simulator.
You are to respond as you would in the plant for i a real condition.
TASK CONDITIONS: Stable post trip with 1 control rod stuck full out.
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I Task 0000240501 l
i l EXAMINER CUES EXAMINEE ACTIONS STANDARD
( As the SF, inform the 1. Examinee checks the 1. ESAS Checklist Examinee to initiate. boric acid pumps (CA-P- verified or AO emergency boration from 1A/B) set at maximum contacted by page r Tech. Spec. Tank. Inform stroke. The ESAS or radio system.
the Examinee to set the checklist has this Stroke Counters on CA-P- information. Or the 1A/B to 27921, to inject Examinee may contact 4814 gallons of 17,500 ppm the primary AO for this boron. information. The required position for the stroke settings of CA-P-1A/B is MAX.
- 2. Examinee checks the 2. Stroke counters stroke counters set at have > 027921 27921. displayed in the windows on the LWDS panel.
- 3. Examinee positions The 3. " Local / Remote"
" Local / Remote" Switches switches for CA-P-for CA-P-1A/B to the lA/B on the LWDS "LWDS" position. panel are in the "LWDS" position.
- 4. Exarainee opens 4. MU-V-51 OPEN as MU-V-51. indicated on the
( LWDS panel by Red Light illuminated.
- 5. Examinee starts the 5. CA-P-1A and B are boric acid pumps, STARTED as CA-P-1A/B. indicated on the LWDS panel by Red Light illuminated.
- 6. Examinee checks that 6. Examinee may use the level in the BAMT local tank level, decreases. or level indication on the Computer, Point A0475.
JPM may be terminated at this time.
(*) . Denotes Critical Element. !
l EVALUATION: SAT: UNSAT:
1
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Page 5 of 8 File 0000240501 Aw
Task 0000240501 4
( TITLE: Perform an Emergency Boration .
s JPM NUMBER: 11.2.05.047 l TASK NUMBER: 0000240501 i
EXAMINEE:
EVALUATOR: DATE:
l 4
EVALUATION OF EXAMINEE JPM: SAT: UNSAT:
i COMMENTS:
I s
(
EVALUATOR SIGNATURE:
Page 6 of 6 File.ouo(c40501 Ax
Task ( 000240501 What-protection does the high pressure limit curve for MU tank j~
level vs. pressure provide? Include in your discussion a scenario where violation of this curve causes a problem.
l ANSWER:
If the pressure is too high for a given level (0.2) and a LOCA l 1
occurs (0.2), when the BWST level drops (0.2) the gas bubble in ^
the MU tank can enter the MU pump section (0.2) causing possible MU pump damage. (0.2) l
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PEDIGREE INFORMATION:
TIME ESTIMATE: 3 MIN.
K&A NUMBER AND VALUE: 004 K4.05 RO:3.3 SRO:3.2 CFR: 41.7,45.6 OBJECTIVE / RATING: IV.A.09.35 SRO: 3.4 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
CURVE 1 FROM OP 1104-2 (MU)
HISTORY: QJ4A09-35-QO1
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Page 7 of 8 nie 0000240501.du
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Task 0000240501 The unit trips as a result of a loss of offsite power. Both D/G
( have started & are supplying their busses.
Post trip it is discovered that 1 control rod-is stuck out.
Describe TWO methods of Emergency Boration.
Include:
Sources Pumps Major valves that must be opened and Injection point into the RCS.
ANSWER:
- 1. BWST (0.15) via MU-V-14A (0.1) to the running MU pump (s)
(0.1) injection to the RCS "B" cold leg (0,1).
- 2. BAMT (0.15) via CA-P-1A/B_(0.1) and
{
MU-V-51 USING THE HANDWHEEL (0,1) !
MU pump (s) (0.1) inject to the RCS "B" cold leg (0.1)
)
Prompt student to check power by telling him the lights on MU-V-51 are out if he attempts to go through the valve without using the handwheel.
(lack of power to G switchgear means MU-V-51 has no power.)
i 1
PEDIGREE INFORMATION:
TIME ESTIMATE: 5 MIN.
K&A NUMBER AND VALUE: 024 AA2.02 RO:3.9 SRO:4.4 CFR: 41.7/45.5/45.6 OBJECTIVE / RATING: IV.A.09.25 RO:3.4 SRO:5.0 COGNITIVE LEVEL: 300
- STUDENT
REFERENCES:
Control Room, Simulator or flow prints,
!_ 1107-4&5, 209 Prints l HISTORY: OJ4A09-25-001
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i Page 8 of 8 File 0000240301.dx
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- THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERFORMANCE MEASURE TITLE: Respond to.a Dropped Control Rod TASK NUMBER: 0000030501 TIME: 12 Minutes EXAMINEE
REFERENCE:
EP 1202-8, CRD Equipment Failure EVALUATION i
METHOD: PERFORM: X SIMULATE:
. EVALUATION '
LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: !
OPERATOR PERFORMING JPM:
EVALUATOR: / /
DATE K/A: 000 003 AA1.05 IMPORTANCE: 4.1 10CFR55.45: (a) (3), (4), (5), (6)
COMMENTS: -(If results are unsatisfactory, record required data on-sheet provided in back of this JPM.)
7 I
Page 1 of 7 File \0000(00501. doc I: . ~ , .--
Task 0000030501 SIMULATOR
( CONDITIONS: Initialize the trainer at IC-17 Change constant ICK106B to 0.1, changes the setpoint for the ICS asymmetric control rod runback to 550 megawatts as opposed to the normal setpoint of 482 megawatts.
Assign Malfunction RD101 on Remote Key #1, Dropped Rod.
, EXAMINER PREVIEW: This JPM deals with responding to a dropped full-length control rod.
DO NOT inform the Examinee of this fact The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK STANDARDS: Reactor power as indicated by PPC point C1708, l or NAS Display #1 or #2 is less than 60% power.
l Page 2 of 7
[ FileWm0030501. doc l
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Task 0000030501 EXAMINEE
( PREVIEW: For this event you are assigned the duties of the Primary CRO. The instructor / examiner will act as the Secondary CR0 and SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
You are expected to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to verify proper completion.
When you are told to begin, you will respond to the cues and/or indications which the Examiner will provide to you either verbally or by the simulator.
You are to respond as you would in the plant for a real condition.
TASK CONDITIONS: Reactor is at power, normal operations.
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I Task 00C3030501 l l EXAMINER CUES EXAMINEE ACTIONS STANDARD l' ,
Activate Malfunction 1. Examinee should 1. Examinee verifies !
RD0101 on Remote Key #1 recognize the dropped that reactor power, i to drop the control Rod, and verify the rod, feedwater flow and plant runback to 482 megawatts MWE or less. electrical are all '
NOTE:.
decreasing as seen Due.to the modification on console of the constant indications.
ICK106B, the plant will stabilize at greater j than 60% thermal power.
l
- 2. Examinee realizes 2. Examinee reduces If the Examinee
~
that the plant has reactor power to inquires, allow him to stabilized at use the ULD, which is less than 60%
greater than 60% thermal power, as still operable, to power on the power indicated by PPC l reduce Reactor power to range NI's and point C1708 or on less than 60%. reduces reactor NAS display No. 1 1
However, the Examinee power either by or 2 at less than !
may take any station (s) decreasing the 60% of 2568 MWt.
j to manual to continue output of the ULD, j
to reduce Reactor or by taking SAT
{
power. control of ICS station (s) in HAND, NOTE: and reducing The examiner shall Reactor Power with )
j provide the examinee the ICS in track. !
with updates of NAS indication of reactor thermal power, JPM may be. terminated at this time.
(*) Denotes Critical Element.
NOTE: Return Constant ICK106B to -0.04.
EVALUATION: SAT: UNSAT:
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Page 4 of 7 Fileio000030501. doc
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Task 0000030501 l -
l ( TITLE: Respond to a Dropped Control Rod l
JPM NUMBER: 11.2.05.014 l TASK NUMBER: 0000030501
! EXAMINEE:
EVALUATOR: DATE:
EVALUATION OF i EXAMINEE JPM: SAT: UNSAT:
COMMENTS: i l
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I l EVALUATOR l SIGNATURE:
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l Page 5 of 7 Filc\0000030501. doc
t-Task 0000030501 Given the following plant conditions:
Control Rod Group 6 consist of 8 individual rods, their individual positions are as follows; Rod 6-3 93%
Rod 6-6 96%
Rod 6-8 95%
The remaining rods are 100%
A. Which asymmetric (condition / Fault) light (s) should be lit?
B. Explain your answer in Part A for the conditions above.
.o 1
ANSWER:
A. The Asymmetric condition light for Rod 6-3 will be lit on the PI panel only (0.5) 1 l
B. Rod 6-3 deviates from the group average by 5% I
.(> 7-inch) (0.25)
All the other rods are within the individual to group average differential setpoint, therefore all the associated lights are out (0.25)
PEDIGREE INFORMATION:
3 TIME ESTIMATE: 5 MIN. i K&A NUMBER AND VALUE: 014 K5.02 RO:2.8 SRO: 3.3-CFR: 41.5/45.7 OBJECTIVE RATING: IV.E.13.19 SRO:3.0 COGNITIVE LEVEL: 300 l STUDENT
REFERENCES:
ARPS HISTORY: OJ4E13-19-001
, i I
Page 6 of 7 File \0(KKK130501. doc
Task 0000030501 The plant is at 800 MW with the Reactor and Feedwater Subsystems in Hand when:
Group 7 Rod 3 drops; No rod motion is observed when the operator attempts to insert CRDM's at the Diamond Panel Shim switch.
A. What must the operator do to manually run the plant back to less than-60% power?
B. Explain WHY this is necessary to complete the manual runback.
1 1
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ANSWER:
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A. Depress the In-Limit Bypass (Latch) pushbutton (0.5)
B. .The GROUP IN-LIMIT from the dropped rod stops rod in motion in that group. (0.5) i l
I PEDIGREE INFORMATION:
TIME ESTIMATE: 3 MIN.
K&A NUMBER AND RATE: 003 AK3.01 RO:3.5 SRO: 3.9 CFR: 41.5,41.10,45.6,45.13 OBJECTIVE RATING: IV.E.13.24 SRO: 3.0 COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
CRD PROCEDURE HISTORY: QJ4E13-24-QO1
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Page 7 of 7 Fileio(XK1030501. doc
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l THREE MILE ISLAND UNIT 1 l A
INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERFORMANCE MEASURE TITLE:-
~
Respond to a Malfunction in Pressurizer Level Indication or Control 1
. TASK NUMBER: '0008140401 1 l
TIME: 10 Minutes ]
EXAMINEE MAP G, G-2-5
REFERENCE:
EP 1202-29, Pressurizer System Failure EVALUATION
- METHOD: PERFORM: X SIMULATE: 'l EVALUATION-LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM:
I
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- OPERATOR PERFORMING JPM:
l EVALUATOR: / /
DATE K/A: . 000 028 AA2.10 IMPORTANCE: 3.4 10CFR55.45: (a) (3) , (4), (6), (7)
COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.)
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a Page 1 of 8
- j. - [*11+ \ ue'Je l 4 04 01. oe l.
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- l. Task 0008140401 SIMULATOR
{ CONDIT10NS: Initialize the trainer at IC-17 l Assign MALFUNCTION RC04A to Remote Key #1 at 100%
severity over 60 seconds. (Fails selected pressurizer level)
EXAMINER PREVIEW: This JPM deals with responding to a failure in pressurizer level indication low.
The plant will be at 1001 power when a failure of the selected pressurizer level instrument will occur.
DO NOT infcrm the Examinee of this fact.
The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
~
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued) l i
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l Page 2 of 6 t . . .+ eum. 2-
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I Task 0008140401 l (Continued)
(
EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK STANDARDS: RC-1 LT-3 is selected to provide correct controlling pressurizer level indication,
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Page 3 of 8 ro-s m s. . i
Task 0008140401 i EXAMINEE l ' PREVIEW: For this event you are assigned the duties of l the Primary CRO. The instructor / examiner will act as the Secondary CRO and the SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
You are expected to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to verify proper completion.
When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you either verbally or by the simulator.
You are to respond as you would in the plant for a real condition. 1 TASK CONDITIONS: The Plant is at 100% power.
No surveillances or maintenance is in progress.
1 Page 4 of 8 r w a u n...t 1
Task 0008140401 EXAMINER CUES EXAMINEE ACTIONS STANDARD l Active Malfunction RC04A 1. Examinee places 1. Makeup flow on Remote Key #1.to cause MU-V-17 into manual adjusted as a failure of pressurizer control and adjusts indicated by level indication low. makeup flow to MU-24 A/B FI on equal letdown flow CC.
minus seal injection by manipulating the control switch on CC.
- 2 Examinee selects 2. Alternate the alternate pressurizer level pressurizer level instrument RC-1 and temperature LT-3 selected, ;
instruments to indicated by i determine the white light failed instrument illuminated.
on CC.
- 3. Examinee verifies 3. Pressurizer that pressurizer heaters verified, heaters are
{ energized on CR.
The JPM may be terminated after either RC-1 LT-3 selection, or
.after RC-1 LT-3 is selected and the Examinee displays successfully manipulation of pressurizer level.
(Examiner perrogative.
(*) Denotes Critical Element.
EVALUATION: SAT: UNSAT:
l Page 5 of 8 n uem m. 2.
-. -. .. _. - - . - - . - . --- . . . = . _ - . . .- ,
Task 0008140401 1 TITLE: Respond to a malfunction in Pressurizer Level Indication of Control JPM NUMBER: l TASK NUMBER: 0008140401 EXAMINEE:
EVALUATOR: DATE:
EVALUATION OF EXAMINEE JPM: SAT: UNSAT:
COMMENTS:
k l l
EVALUATOR SIGNATLRE:
Page 6 of 8
.ru-sea e1404ci.A.
. Task 0008140401 i i
Given the following condition: !
(
Compensated Presssurizer Level is NOT available RCS temperature is 300 F Current Pressurizer Level Differential Pressure is 120" on PPC. point A0503 Based on these conditions, what is actual Pressurizer Level i using the differential pressure? t ANSWER:
1 300" ( + 5") actual level i
PEDIGREE INFORMATION:
TIME ESTIMATE: 5 MIN.
K&A NUMBER AND VALUE: 011 A1.02 SRO: 3.6 OBJECTIVE / RATING: V.D.11.06 SRO: 2.8 COGNITIVE LEVEL: 300 STUDENT
REFERENCE:
EP 1202-29, FIGURE 1 HISTORY: NEW t
Page 7 of 8 r u - w oce14cedo1. e
Task (XX)8140401 Given the following conditions:
(
Pressurizer Level channel RC-1-LT3 had been removed from service due to electrical problems, i Pressurizer Level channel RC-1-LT1 has just failed LOW l RC-1-LT1~had been indicating 225" prior to its failure RCS temperature is 579*F '
- Reactor power is 100%
Both level channels will be inoperable for 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> A. For the given conditions, what actions are required?
B. If response to A above is correct, ask how much Feed Volume would be required to go to Hot Shutdown at 532*F?
l ANSWER:
l A. Return at least 1 to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be j in at least HOT SHUTDOWN within the next six hours. l (Time for Cold Shutdown not required.) !
I B.
6353 gallons From Table 1:
579*F - 576*F = 653 gals 576*F - 573*F = 700 gals 573 F - 532 F = 5000 gals PEDIGREE INFORMATION:
1 TIME ESTIMATE: 5 MIN.
K&A NUMBER AND VALUE: 011 A2.03 RO: 3.8 SRO: 3.9 l Generic 2.1.12 RO: 2.9 SRO: 4.0 OBJECTIVE / RATING: V.D.11.05 SRO: 3.0 ;
COGNITIVE LEVEL: 300 STUDENT
REFERENCE:
EP 1202-29, TABLE 1 Tech Spec 3.5.5 HISTORY: NEW Page 8 of 8 ruoc< e:404c: :
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I Task 000C100501
! i THREE MILE ISLAND UNIT 1 '
.p
-INITIAL' SENIOR REACTOR OPERATOR EXAMINATION JOB PERFORMANCE MEASURE TITLE: _ Respond to Inadvertent Closure of a Main Steam '.
' Isolation Valve
.l I
TASK--
NUMBER: 000C100501- .
-TIME: 6 Minutes
,i EXAMINEE '
REFERENCE:
OP 1105-4, Integrated Control' System. _
AP 1203-42, Inadvertent Closure of a Main Steam Isolation V lve / ;
EVALUATION- ~/' l i
METHOD: . PERFORM: X Y SIMULATE:
EVALUATION /
. LOCATION: SIMULATOR: 'IN-PLANT: CONTROL ROOM.:
)
i OPERATOR PERFORMING JPM-l EVALUATOR: / /
DATE K/A: 039 A3.02
/
IMPORTANCE: 3.p 10CFR55.45: ( ) (3), (4), (6) ..
~ COMMENTS: If results are unsatisfactory, record required jdata on sheet provided in back of this JPM.)
I p- .,
l-Page 1 of 8 l
Task 000C100501 SIMULATOR
( CONDITIONS: Initialize the Trainer at IC-17.
Assign MALFUNCTION MS08A to Remote Key #1, Inadvertent closure of MS-V-1A.
EXAMINER PREVIEW: This JPM deals with an inadvertent closure of a Main Steam Isolation Valve. The plant will be at l power ops, MS-V-1A will have an inadvertent l closure which will require the Examinee to reduce I power to 90% per the EP.
l DO NOT inform the Exsminee of this fact.
The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the l information contained in the " EXAMINEE i PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Inform the Examinee of the INITIATING CUES.
- 4. Allow the Examinee 2-3 minutes to scan the console.
- 5. Go to RUN on the Simulator.
- 6. Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued) l l
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Task 000C100501 (Continued) k EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK STANDARDS: Reactor power is < 901.
1.
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Page 3 of 8 File \000cI00501
Task C30C100501 EXAMINEE
( PREVIEW: For this event you are assigned the duties of the Secondary CRO. The instructor / examiner will act as the Primary CRO and the SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
You are expected to perform the Immediate Manual Actions of the appropriate EF/AP/ATP from memory, but then use the procedure to verify proper completion.
When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you either verbally and via the simulator.
You are to respond as you would in the plant for a real condition.
TASK CONDITIONS: The Plant is at 100% power.
g, There are no surveillances in progress, and no
% major maintenance is scheduled for this shift. j INITIATING CUES: Cues will be provided by the Examiner or by the l Simulator.
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Page 4 of 8 FileiOOOcl00501
Task 000C100501 l EXAMINER CUES EXAMINEE ACTIONS STANDARD
-I Activate MALF ECTION MS08A *1. Examinee diagnoses 1. MS-V-1A diagnosed on Remote Key #1 to close that MS-V-1A has as being closed.
MS-V-1A. closed.
- 2. The Examinee 2. Output Demand of reduces the output the UNIT LOAD of the UNIT LOAD DEMAND ICS DEMAND ICS station station reduced to achieve a to achieve 90%
desired effect of Reactor Power, as reducing Reactor indicated on CC.
Power to 901 SAT
- 3. Exa alnee reduces 3. Reactor Power is Reattcr Power to less than or
< 90t. equal to 90%, as indicated on CC on the power range NI's.
{ SAT
(*) Denotes Critical Element.
EVALUATION: SAT: UNSAT:
(
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Pace 5 of 8 i FileiOOOcl00501 l
Task 000C100501
( TITLE: Respond to Inadvertent Closure of a Main Steam Isolation Valve JPM NUMBER: 076 l TASK NUMBER: ;
EXAMINEE:
1 EVALUATOR: DATE:
I EVALUA1 ION OF I EXAMINEE JPM: SAT: UNSAT:
)
1 COMMENTS: l l
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EVALUATOR SIGNATURE:
Page 6 of 8 File \000c!00501
1 Task 000C100501 If a Reactor trip occurred following a loss of ICS auto power,
-(
A. How would the Atmospheric Dump Valves respond. i B. Explain why.
r 1
lI ANSWER:
A. They would remain closed. (.5) l B. Because they transfer to the Manual Backup Loaders on loss of ICS Auto power. (.5)
PEDIGREE INFORMATION:
TIME ESTIMATE: 3 MIN.
K&A NUMBER AND VALUE: 041 K6.03 RO:2.7 SRO:2.9 CFR: 41.7/45.7 OBJECTIVE / RATING: IV.C.01.05 SRO: 4.2 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
EP 1202-42 HISTORY: OJ4C01-05-002 Page 7 of 8 File \000cl00501
i Task (XX)Cl(X)501 :
Abnormal Procedure 1202-24, Steam Leak, says to reduce power to
.k <45%'and' manually trip the turbine if the steam leak is downstream of the stop valves and time permits.
If the turbine is tripped at 43% power:
A. How will the main steam system respond to control Main Steam Pressure?
B. What will be controlling Main Steam pressure when the plant stabilizes and at what setpoint.
ANSWER:
A. The main steam safeties will open (.2)
The atmospheric dump valves will open (MS-V-4A/B) (.2)
The turbine bypass valves will open (.2)
B. The turbine bypass valves (.2) will be controlling header pressure at 895 psig (setpoint plus bias) (.2) !
PEDIGREE INFORMATION:
TIME ESTIMATE: 4 MIN.
K&A NUMBER AND VALUE: 041 K4.18 RO:3.4 SRO:3.6 CFR: 41.7 OBJECTIVE RATING: IV.C.01.04 SRO: 4.0 COGNITIVE' LEVEL: 400 STUDENT
REFERENCES:
ABN 1203-24, OP 1105-4 HISTORY: QJ4C01-04-QO1 s
Page 8 of 8 FileWOOcl00501
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Task 0538010101 j l
.THREE MILE ISLAND UNIT 1
. t
- INITIAL ~ SENIOR REACTOR OPERATOR EXAMINATION JOB PERFORMANCE MEASURE TITLE:_ Perform the Required Actions for Loss of Stator Coolant Pump with Failure of the Standby Pump to Auto Start TASK
-NUMBER:- 0538010101 TIME: 15 Minutes EXP.MINEE
REFERENCE:
ARP MAP L, L-1 '? !
EVALUATION-METHOD: PERFORM: X SIMULATE:
EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM:
(< .
OPERATOR PERFORMING JPM:
L EVALUATOR: / /
DATE i K/A: .BW/An1 an, . ,-
(j L/ 7 fl/, / L /tol7,)3 ,y y, c IMPORTANCE; # l
, n ,,- ,e r , e._ C [/Z 1 4 [,7 i l 1 l COMMENTS: (If results are unsatisfactory, record required l data on sheet provided in back of this JPM.)
( j i
Page 1 of 8 w --r -- . - -c. -
Task 0538010101 SIMULATOR
( CONDITIONS: Initialize the Trainer at IC-17.
IO OVERRIDE:
05A6S25-ZDICSGNP5B(3), NAP, to prevent auto start GN-P-5B)
MALFUNCTION:
EG04A assign to Remote Key #1 EXAMINER PREVIEW: This JPM deals with performing the required actions to stabilize the plant following the loss of the running stator coolant pump with failure of the standby pump to auto start.
DO NOT inform the Examinee of this fact.
The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
l Page 2 of 8 filc 0538010101. doe
Task 0538010101 A certain amount of liberty -(questioning or
( cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly.
TASK !
STANDARDS: The standby stator coolant pump running, terminating the turbine runback prior to load l being reduced to below 300MWe.
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Task 0538010101 EXAMINEE
{ PREVIEW: For this event you are assigned the duties of the Shift Forman. The instructor / examiner will act as the Primary and Secondary CRO. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
You are expected to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to verify proper completion.
When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you either verbally or by the simulator.
You are to respond as you would in the plant for a real condition.
TASK CONDITIONS: The plant is at 100% power, equilibrium Xenon, EOL.
The Dispatcher has requested that load not be reduced below 300Mwe or there will be a possibility of loss of the grid due to insufficient generation.
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Page 4 of 8 File 0538010101.dm
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Task 0538010101 EXAMINER CUES EXAMINEE ACTIONS STANDARD k l Activate MALFUNCTION 1. Determine Loss of 1. Examinee Remote Key #1 GN-P-5A has occurred. diagnoses by alarms and indications that a GN-P-5A has tripped and a plant runback is in progress Note: *2. Determine GN-P-5B *2. Direct SCRO to Runback needs to be did not auto start. start GN-P-5B.
terminated prior to' reducing load below SAT 300MWe
- 3. Verify runback 3. Load reduction terminated and plant stopped prior to stabilizes at > 300 reducing below MWe. 300 MWe and plant is stable.
SAT JPM may be terminated at this time.
{;(*) Denotes Critical Element.
EVALUATION: SAT: UNSAT:
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2 Page 5 of 8 File 053x010101 doc
Task 0538010101
(. TITLE: Perform the Required Actions for Loss of Stator Coolant Pump with failure of Standby pump to Auto Start JPM NUMBER: 11.2.05. l TASK NUMBER:
EXAMINEE:
EVALUATOR: DATE:
EVALUATION OF EXAMINEE JPM: SAT: UNSAT:
COMMENTS:
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EVALUATOR SIGNATURE:
Page 6 of 8 File 0538010101 doe
l Task 0538010101
- The plant is operating at 100% power.
- Stator Cooling temperature control. valve GSC-V-YO7 has been oscillating.
I& C has informed.you-they need to make adjustments to the controls for GSC-V-YO7. i
- They also warn you that a reduction in stator cooling ;
may result during the adjustments. i What' action, if any, should you take prior to allowing the adjustment to be made?
l ANSWER:
The generator load should be reduced to the "no-liquid flow" load (140 MWe)
I PEDIGREE INFORMATION:
TIME ESTIMATE: 10 MIN.
. K&A NUMBER AND VALUE: ~7. , / , @ RO: SRO: 3. I) l OBJECTIVE / RATING: N/A / 3.4 i
COGNITIVE LEVEL:
i STUDENT
REFERENCES:
OP 1106-7, Page 3.0, Rev 33 I- HISTORY: NEW 1
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4 Page 7 of 8 File 0538010101. doe
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Task 0538010101 Prior to startup of the stator coolant system, the hydrogen gas
( pressure should be greater than 40 psig.
What is the basis for the hydrogen pressure requirement in the generator even when the unit is shutdown and not producing any heat?
ANSWER:
1 Pressure greater than 40 psig will ensure that if a small leak develops in the stator coolant system inside the generator that water will not leak into the generator when the stator coolant pump is started. i l
PEDIGREE INFORMATION:
TIME ESTIMATE: 5 MIN. 1 OBJECTIVE RATING:
1 COGNITIVE LEVEL: 300 1 STUDENT
REFERENCES:
OP 1106-7, Pages 3.0 & 4. 0, Rev. 33
' HISTORY: NEW l
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Page 8 of 8 File 05.18010101 doe
i V
THREE MILE ISLAND UNIT 1 l i: I INITIAL' SENIOR REACTOR OPERATOR EXAMINATION !
l JOB PERFORMANCE MEASURE i I
. TITLE: Direct Establishment.of'Long Term Core Circulation To Prevent Boron Concentration- 1 TASK NUMBER: 3448050303
. i
- TIME: 12 Minutes
)
EXAMINEE
REFERENCE:
OP 1104-4, Decay Heat removal System l
EVALUATION METHOD: PERFORM: X- SIMULATE:
EVALUATION j LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: X-i OPERATOR PERFORMING JPM:
EVALUATOR: / /
DATE K/A: M 011 EA1.11 IMPORTANCE: 4.2 10CFR55.45: (a) (12)
' COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.)
Page 1 of 9 File 3448050303. doe
l Task 3448050303 SIMULATOR
.( CONDITIONS: Initialize the Trainer at IC-22. j Set BWST <6' 4" by setting DHMBWST = 3.3E6 ;
Open DH-V-64 Open DH-V-6A/B on CC/CR j Close DH-V-5A/B on CC/CR l Close RC-V-3 on CC/CR Stop BS-P-1A/B l Close BS-V-1A/B l Close BS-V-2A/B Close BS-V-3A/B Stop MU-P-1B Throttle DH-V-19A/B to < 2800 Gal./ Min.
I EXAMINER PREVIEW: This JPM deals with placing the RCS in long term recirculation as described in OP 1104-4 (Decay Heat Removal)
The RCS will in a post LOCA condition. The l Examinee will direct establishment of long term recirculation.
The sequence of events which should be followed in the conduct of this'JPM is as follows: i k'
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued)
> 1 Page 2 of 9 Fde 3448050303.da:
Task 3448050303 (Continued) k EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed l incorrectly, would result in a failure to meet j the " TASK STANDARD") correctly.
TASK STANDARDS: The Active method of long term recirculation has been established. I i
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i Page 3 of 9 File 344805030Ldoe
Task 3448050303 EXAMINEE
( PREVIEW: For this event you are assigned the duties of the SS/SF. The instructor / examiner will act as the Primary and Secondary CRO. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page system.
When you are told to begin, you will respond to the cues and/or indications which the Examiner will provide to you either verbally or by the simulator.
You are to respond as you would in the plant for a real condition.
TASK CONDITIONS: The reactor tripped on low pressure approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> ago.
A LB LOCA 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> ago has caused RCS depressurization to the current value.
The core is being cooled with low pressure injection flow.
HPI flow has been secured.
LPI pumps are taking a suction from the RB sump.
The Immediate manual Actions and the applicable followup actions of ATP 1210-1, ATP 1210-2, and ATP 1210-7 have been performed.
ESAS has been BYPASSED / DEFEATED / RESET to gain ES Component control.
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DH-V-64 is Open.
DH-V-2 breaker has been closed.
DH-V-19A/B have been throttled for LPI flow.
Adequate core cooling is being maintained by "A" and "B" LPI flow.
Page 4 of 9 l'ili3448050.103 dw
I
, Task 3448050303
, l EXAMINER CUES EXAMINEE ACTIONS STANDARD
-l Inform the examinee to 1. Examinee directs 1. Following steps !
direct' establishment of the Primary CRO to performed. j long term recirculation establish long term l cooling per OP 1104-4. recirculation I cooling by ;
NOTE: performing the '
Examiner should assume the following steps. ;
role.of the Primary CRO and performs the steps as directed by the Examinee..
- 2. Examinee directs 2. DH-P-1A motor the Examiner to current & red verify DH-P-1A is breaker closed running. light are checked.
- 3. Examinee directs 3. RC-V-3 closed.
the Examiner to verify RC-V-3 closed.
- 4. Examinee directs 4. RC-V-4 open, the Examiner to open RC-V-4 to SAT establish aux.
{ spray flow.
- 5. Examinee directs 5. Low range the Examiner to pressure verify RCS pressure indicator is is in equilibrium reading with RB pressure. approximately O.
- 6. Examinee directs 6. Logs verified. i Examiner to verify it has been 12
l 7. Examinee directs 7 DH-P-1B motor l Examiner to verify current & red ;
( DH-P-1B operating. breaker closed j light are l checked. 1
(*) Denotes Critical Element.
Page 5 of 9 File 3448050303. doe
i Task 3448050303 l EXAMINER CUES EXAMINEE ACTIONS STANDARD
( NOTE: The "B" Channel is- *8. Examinee directs 8. "A" channel 400#
not simulated. Inform the Examiner to reset bistable reset as l examinee that the "B" the 400# DH indicated by channel has been reset and bistable in the white OUTPUT
- to reset the 400# bistable "A" ESAS cabinet STATE indicator for "A" channel only. module 1-8-4 by becoming dim.
depressing the toggle switches SAT below the OUTPUT STATE and OUTPUT MEMORY lights.
l 9. Examinee directs 9. MAP C-1-6 clear Examiner to verify I that MAP alarm C-1-6 is clear.
- 10. Examinee directs 10. DH-V-3 partially ;
Examiner to open, partially open DH-V-3 by SAT ;
depressing the open pushbutton )
l 5.5 Seconds on l CC.
{': Inform examinee that motor *11. Examinee directs 11. DH-V-2 open currents will be Examiner to open monitored. DH-V-2 by SAT depressing the open pushbutton on CC.
Inform the examinee that *12 Examinee directs 12. DH-V-1 open motor currents will be Examiner to open monitored. DH-V-1 by SAT depressing the open pushbutton on CC.
JPM may be terminated at this time.
(*) Denotes Critical Element.
EVALUATION: SAT: UNSAT:
\
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1 Task 3448050303 l
1
( TITLE: Direct establishment of Long Term Core Circulation to '
prevent boron concentration ,
JPM NUMBER: 11.2.05.1.56 l TASK NUMBER: 3448050303 I EXAMINEE: !
1 EVALUATOR: DATE:
l EVALUATION OF EXAMINEE JPM: SAT: UNSAT:
i COMMENTS:
l 1
i EVALUATOR j SIGNATURE: ,
l Page 7 of 9 l File 3448050303. doe i
. Task 34480$0303 Plant Conditions
(
3 large break LOCA occurred and the operators have established Reactor Building sump recirculation.
Only DH-P1-B is operable.
Describe the flow path for this mode.
Include major components in the flow path.
ANSWER:
Take suction on the RB sump (0.2) via DH-V-6B (0.2), using DH-P-1B (given), discharge through the DC cooler (0.2), through open cross connect valves (DH-V-38A&B) (0.2) feeding both LPI lines via DH-V-4A & B (0.2) r y#
PEDIGREE INFORMATION:
TIME ESTIMATE: 4 MIN.
K&A NUMBER AND VALUE: 011 EK3.08 RO:3.9 SRO:4.1 CFR: 41.5/41.10/45.6/45.13 OBJECTIVE / RATING: IV.A.11.15 SRO: 3.4 COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
ATP 1210-7 HISTORY: QJ4A11-15-001 c.
Page 8 of 9 I
l Task 3448050303 EXPLAIN the piggyback operation flowpath and WHEN it is
( required?
ANSWER:
WHAT:
The DH pump (0.15) takes suction on the RB sump (0.15) and supplies suction to the MU pump (0.15) which provides HPI to the core (0.15)
WHEN:
Prior to the BWST reaching < 6' 4" (BWST LO LO Level alarm).
(0.4) lD ri.;jf/f41( (, y/7f[' (pf d V PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 011 EK3.08 RO:3.9 SRO:4.1 CFR: 41.5/41.10/45.6/45.13 i OBJECTIVE / RATING: IV.A.11.03 RO:2.2 SRO: 3.8 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
1210-6 and 1210-7 HISTORY: QJ4A11-03-001 Page 9 of 9 Fuc0448050303 doe
Task 0410040104 THREE MILE ISLAND UNIT 1 l-
-INITIAL SENIOR REACTOR OPERATOR EXAMINATION 4
JOB PERFORMANCE MEASURE TITLE: Operate the Turbine Bypass Control. Valves Locally TASK NUMBER: 0410040104 TIME: 10 Minutes EXAMINEE
REFERENCE:
Local Operator Aid EVALUATION METHOD: PERFORM: SIMULATE: X EVALUATION LOCATION: SIMULATOR: IN-PLANT: X CONTROL ROOM: (' I-OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 041 A4.08
.IMPORTANCE: 3.1 10CFR55.45: (a) (3), (4), (6), (7), (8)
COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.)
'l Page 1 of 9
Task 0410040104
- SIMULATOR
( CONDITIONS: N/A L EXAMINER l PREVIEW: This JPM deals with operating the Turbine Bypass l Valves locally. l l The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions about the JPM.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
l A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the f Examinee, however, care must be taken to ensure l 1 that you do not COACH the Examinee. To complete this task successfully, the Examinee must complete each critical element (an element i of the task which, if omitted or performed incorrectly, would result in a failure to meet l the " TASK STANDARD") correctly. For In-Plant JPMs additional safety gear may be required. The Examinee should inform the Examiner that additional gear is required. The Examiner will then specify if the additional safety gear is to be donned, or to be simulated after the Examinee specifies what equipment is required and where it would be located. l Failure to use the proper safety equipment is immediate grounds for failure. TASK STANDARDS: Local manual control of MS-V-3C has been established. I Page 2 of 9 File \0410040104 l
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Task 0410040104 i EXAMINEE l i ( PREVIEW: For this event you are assigned the duties of l the Secondary AO. The instructor / examiner will act as the Console CRO and SF. When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you verbally. Unless otherwise informed, you are to SIMULATE all actions taken. You are expected to maintain proper i communications with the Control Room as you would in the plant for a real condition. It is very important that you describe to the instructor / examiner the physical details of the actions that you would be taking if you were actually performing the manipulations required to complete this task. Failure to adequately describe your actions could result in a failure to meet the task standard. k You are to respond as you would in the plant for a real condition. You are expected to wear all required safety gear in accordance with plant procedures and policies. If during a simulated task additional safety gear is required you should notify the Examiner. Failure to use the proper safety equipment is immediate grounds for failure. TASK CONDITIONS: The Reactor has been tripped, and a "Cooldown from Outside the Control Room" has commenced in accordance with EP 1202-37. r l Page 3 of 9 File \0410040104
i l l Task 0410040104 i i l l l I EXAMINER CUES EXAMINEE ACTIONS STANDARD l 'The SF' instructs you to 1. Valve.is located on 1. Examinee proceeds ! take local manual the west' side of the to MS-V-3C and { control of MS-V-3C'and' Main Condenser on the establishes l l position the valve under 322' elevation, headphone
']
the direction of the CRO Examinee establishes communications ' at the RSD Panel who is communications with with the RSD currently using the. the RSD Panel, Panel. ) headset communication Headphones are , system. available. l ! ROLE-PLAY: l If the Examinee j establishes communication at this time or later, as the CRO, inform the Examinee that communications are established. NOTE: +2. Examinee turns the 2. Examinee aligns Steps 2 through 5 are valve handwheel the holes. sequence critical, clockwise to align the holes in the stem SAT and the manual
.{ operator. 1 POSITIVE CUE:
Inform the Examinee that j the holes are aligned.
*3. Examinee inserts the 3. Examinee inserts pin into the holes of the pin into the stem and the collar holes.
for the manual SAT l operator. POSITIVE CUE: Inform the Examinee that j the pin is inserted if ! they correctly describe l steps 2.and 3. NEGATIVE CUE: Inform the Examinee that the pin can not be inserted into the holes if they incorrectly ' describe steps 2 and 3. i Page 4 of 9 File \0410040104 l L
Task 0410040104 I l EXAMINER CUES EXAMINEE ACTIONS STANDARD ( *4 Examinee depresses 4. AUTO / MANUAL the AUTO / MANUAL switch positioned switch located'on to the MANUAL the controller box position. ' and then turns the switch SAT counterclockwise
)
90 to the MANUAL l position. I POSITIVE CUE: Inform the Examinee that the switch is in MANUAL. 1 NEGATIVE CUE: Inform the-Examinee that the switch will not turn if they do not depress the switch.
*5. Examinee opens the 5. Examinee opens the equalizing valve by equalizing valve j turning it located on the counterclockwise to outside of the
(,; the fully open diaphragm. position. ; SAT POSITIVE CUE: Inform the Examinee that the equalizing valve is fully open. I l
' ,i Page 5 of 9 File \0410040104
4 i Task 0410040104 1 l l
' IEXAMINER CUES EXAMINEE ACTIONS STANDARD-(1 6. Examinee turns the 6. Examinee positions ROLE PLAY: valve handwheel the valve as As the CRO, have the clockwise to close directed by the CRO Examinee position MS-V- the valve two at the RSD Panel.
3C two turns closed and turns. standby for further-instructions. l I l POSITIVE CUE: 1 Inform the Examinee that the. valve is positioned ! as desired if they correctly describe steps i 5 and 6. ! NEGATIVE CUE: Inform the Examinee that it is very difficult to l turn the handwheel and it becomes progressively difficult as the valve is moved away from the initial position if the { Examinee incorrectly describes steps 5 and 6. JPM may be terminated at this-time. (*) Denotes Critical Element. I EVALUATION: SAT: UNSAT: l l l I I l-L l Page 6 of 9 I FileV41W40104 l
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i . Task 0410040104 i TITLE: Operate the Turbine Bypass Valves Locally JPM NUMBER: 11.2.05.031 l TASK NUMBER: 0410040104 l EXAMINEE: l l l l EVALUATOR: DATE: EVALUATION OF EXAMINEE JPM: SAT: UNSAT: l i l COMMENTS: i l i t i ( l l l 6 I i l ! l EVALUATOR SIGNATURE: i l Page 7 of 9 File \0410040104
t * ! Task 0410040104 L . What TWO conditions cause ICS to control the Turbine Bypass ( Valves using a 75# bias? i ANSWER: The turbine is reset (.2 point) AND all TBV are closed (.2 point) AND turbine header pressure is within 10 psig of setpoint (.2 point) (with Unit Load Demand < 15%) OR I Unit' Load Demand > 15% (.4 point) j
' PEDIGREE INFORMATION:
TIME ESTIMATE: K&A NUMBER AND VALUE: 045 K4.42 RO:2.8 SRO:3.0 CFR: 41.7 OBJECTIVE / RATING: IV.C.01.04 RO:3.0 SRO: 4.0 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
OPM SEC G-3 HISTORY: QJ4C01-04-Q01 i Page 8 of 9 file \0410040104
Task 0410040104 ; Plant Conditions: I I - Plant is at normal hot shutdown with the Reactor reset and the Turbine reset, l I
- An internal ICS fault results in the +75 psig bias being selected for Turbine Bypass Valve control.
Explain the plant response to this ICS failure. ; Include the response of Thot, Tcold, OTSG pressure, and RCS l _ pressure, and Turbine Bypass Valves. l l l
)
j l l i ANSWER: Turbine Bypass Valves will close and control header pressure at (. 960 psig. (0.25 point) , l OTSG pressure will increase (0.25 r int ) ' T-hot and T-cold will increase (0.25 point) RCS pressure will increase as the RCS heats up (0.25 point)
-PEDIGREE INFORMATION:
TIME ESTIMATE: 2 MIN. K&A NUMBER AND VALUE: 045 K4.42 RO:2.8 SRO:3.0 CFR: 41.7 OBJECTIVE / RATING: IV.E.27.17 RO: SRO:3.4 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
HISTORY: QJ4E27-17-QO3 i Page 9 of 9 File \0410040104
Task 0610130104 THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION
-JOB PERFORMANCE MEASURE TITLE: ' Locally Operate mergency Feedwater Block Valves and EF-V-30 Valv s 4 TASK NUMBER: 0610130104 TIME: 10 Minutes EXAMINEE \L
REFERENCE:
LocalBakelite\ Tag EVALUATION METHOD: PERFORM:
\ SIMULATE: X EVALUATION LOCATION: SIMULATOR: N- LANT: X CONTROL ROOM:
1 OPERATOR PERFORMING JPM: e EVALUATOR: / / ;
.N- DATE I K/A: 061 A2.07 IMPORTANCE: 3.5 10CFR55.45: (a) (3), (4), (6), (7), (8)
COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 7
Task 0610130104 SIMULATOR ( CONDITIONS: N/A l EXAMINER PREVIEW: This JPM deals with operating EF-V-30A locally. The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions in reference to the JPM.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
This JPM will require that the examiner provide information (verbal cues) to the examinee concerning the effects of his simulated actions. These verbal cues may include information such as valve position, switch position, indicating light status, etc. The examiner must be careful < (: to provide the examinee with only the indications that he should be readily expected to observe. The examiner must provide the examinee with a negative cue whenever the simulated action that he describes is incomplete or not correct. A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee. To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. TASK STANDARDS: EF-V-30A has been opened locally. t l Page 2 of 7 F11 \0610130104 l
Task 0610130104 EXAMINEE ( PREVIEW: For this event you are assigned the duties of the Secondary AO. The instructor / examiner will act as the Console CRO and the SF. When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you verbally. Unless otherwise informed, you are to SIMULATE all actions taken. You are expected to maintain proper communications with the Control Room as you would in the plant for a real condition.
'It is very important that you describe to the instructor / examiner the physical details of the actions that you would be taking if you were actually performing the manipulations required to complete this task. Failure to adequately describe your actions could result in a failure to meet the task standard.
i
- s. TASK j CONDITIONS: The Reactor has been tripped due to a Total Loss l of Main Feedwater.
Emergency Feedwater has actuated but there is a failure in the control for EF-V-30A, 1 l t Page 3 of 7 f File \0610130104 I I. ,-. .
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1 Task 0610130104 1 l 1 L l l EXAMINER CUES EXAMINEE ACTIONS STANDARD 1 ! k The SF instructs you to 1. Valve is located in 1. Examinee proceeds ; I manually initiate EFW the Intermediate to EF-V-30A, and flow through EF-V-30A Building on the 291' establishes l by Opening EF-V-30A. elevation. communications ' Examinee establishes with the Control communications with Room, the Control Room. 1 4 i ROLEPLAY: If the Examinee establishes ; communications at this time.or^1ater, as the CRO, inform the Examinee that communications are L established.
*2. Examil. e initiates 2. EF-V-30A Opened.
EFW through EF-V-30A by opening EF-V-30A. SAT The Examinee opens the valve by turning ( the handwheel in the counter-clockwise direction. : POSITIVE CUE: l As the CRO, Inform the ! Examinee that the EFW flow is established if- l the correctly describe performance of Step 2. NEGATIVE CUE: Inform the Examinee that EFW flow is inadequate if they incorrectly describe step 2 or if they do not describe it in sufficient detail. JPM may be terminated at this time.
-(*) Denotes Critical Element.
- EVALUATION
- SAT: UNSAT:
f ?
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Task 0610130104 ( TITLE: Locally Operate Emergency Feedwater Block Valves and EF-V-30 Valves - JPM NUMBER: 11.2.05.007 l TASK NUMBER: 0610130104 EXAMINEE: EVALUATOR: DATE: EVALUATION OF EXAMINEE JPM: SAT: UNSAT: COMMENTS: l l t l
- EVALUATOR
! SIGNATURE: 1. I ( ^ [ l Page 5 of 7 l File \u6:0130104
_ _ - _ _ _ _ _ ._ m - _ Task 0610130104 Plant Conditions: (
- An overcooling event has occurred, caused by a stuck open Main Steam Safety Valve on the "A" OTSG.
The Main Steam Safety Valve reseated when OTSG pressure reached ] 250 psig and "A" OTSG pressure is stable at 250 psig 1 There is no indicated level in the "A" OTSG j RCS Tcold is 532 F A. What are the restrictions, if any, in establishing Main AND Emergency Feedwater to the "A" OTSG? l l B. Why do these restrictions exist? l t ANSWER: I A. The maximum feedrate for MFW is 0.05E6 lbm/hr (0.4) and for EFW 100 gpm (0.4) B. To minimize thermal stresses on the lower OTSG tubesheet (0.2) PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: E05 EK3.2 RO:3.5 SRO:4.0
- CFR
- 41.5/41.10/45.6/45.13 l OBJECTIVE / RATING: V.E.03.08 SRO: 4.2 COGNITIVE LEVEL: 400 STUDENT
REFERENCES:
ATP 1210-10, Steam Tables HISTORY: QJ5E03-08-001 i l l Page 6 of 7 F11e \ 0t.1013 al u 4
Task 0610130104 When Subcooling Margin < 25 deg. F, OTSG levels are to be raised ( to 75-85% Operating Range Level, yet for Natural Circulation only 50% is required. EXPLAIN the basis for the higher levels. ANSWER: This encourages the Boiler / Condenser Mode of core cooling. k~. PEDIGREE INFORMATION: TIME ESTIMATE: 2 MIN. K&A NUMBER AND VALUE: E03 EK1.2 RO:3.8 SRO:4.0 CFR: 41.8/41.10/45.3 i- OBJECTIVE / RATING: V.E.10.01 SRO: 3.6 l COGNITIVE LEVEL: 200 L STUDENT
REFERENCES:
1210-10 ( HISTORI: QJ5E10-01-QO2
\
l Page 7 of 7 F11-\04 01 M104
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j l THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION
/
JOB PERFORMANCE MEASURE TITLE: Change the spectacle flange from closed to open between EF-V-4 and.EF-V-5. / TASK j NUMBER: 0618010504 TIME: 10 Minutes j EXAMINEE
REFERENCE:
ATP 1210-10 EVALUAT wN METHOD: PERFORM: SIMULATE: X EVALUATION LOCATION: SIMULATOR: IN-PLANT: X CONTROL ROOM:- OPERATOR PERFORMING JPM: , 1 l
. EVALUATOR: / /
DATE K/A: 061 A2.04 IMPORTANCE:' 3.8 10CFR55.45: (a) (8), (12) COMMENTS: (If results are unsatisfactory, record required ! data on sheet provided in back of this JPM.) i l i i l Page 1 of 8 !. File \0618010504. doc i l I
_ -_ - ~ _ . _ _.. .._ _ _ . _ . . . _ l l Task 0618010$04 SIMULATOR ( CONDITIONS: N/A EXAMINER l PREVIEW: This JPM deals with swapping the spectacle flange l between EF-V-4 and EF-V-5 to the thru position. The sequence of events which should be followed i in the conduct of this JPM is as follows: 1
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions about the JPM.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
l- A certain amount of liberty (questioning or i cueing beyond the bounds of the "EKNMINER CUES") l may be taken by the Examiner in trying to extract the requisite information from the l ~ Examinee, however, care must be taken to ensure j k that you do not COACH the Examinee. l To complete this task successfully, the Examinee l must complete each critical element (an element i of the task which, if omitted or performed I incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. TASK STANDARDS: Flange between EF-V-4 and EF-V-5 placed in thru position. i-l l I Page 2 of 8 File'0618010504. doc
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Task 0618010504 EXAMINEE (' PREVIEW: For this event you are assigned the duties of the Secondary AO. The instructor / examiner will act as the Console CRO and SF. When you are told to begin, you will respond to the cues or indications which the Examiner will provide to you verbally. Unless otherwise informed, you are to SIMULATE all actions taken. You are expected to maintain proper communications with the Control Room as you would in the plant for a real condition. It is very important that you describe to the instructor / examiner the physical details of the actions that you would be taking if you were actually performing the manipulations required to complete this task. Failure to adequately I describe your actions could result in a failure to meet the task standard. You are to respond as you would in the plant for a real condition. You are expected to wear all required safety gear in accordance with plant procedures and policies. If during a simulated task additional safety gear is required you should notify the Examiner. Failure to use the proper safety equipment is immediate grounds for failure. TASK CONDITIONS: Emergency Feedwater suction sources are being depleted. A suction supply must be lined up from river water. I J Page 3 of 8 File \0618010504. doc I l
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Task 0618010504 EXAMINEE ACTIONS STANDARD klEXAMINERCUES Inform Examinee to swap 1. Examinee proceeds to 1. Valves are located spectacle flange between the intermediate in the l EF-V-4'and EF-V-5 to the building basement intermediate thru position. 281' hallway. building basement l 281' hallway. l
- 2. Examinee contacts 2. Permission granted l the Control for parmission to ,
perform job task. l
- 3. Examinee visually 3.EF-V-4 and EF-V-5 l checks to ensure are closed.
EF-V-4 and EF-V-5 are closed. POSITIVE CUE: If a visual check for valve position is made, inform Examinee that EF-V-4 and EF-are closed. NOTE: *4.Using the 4. Flange nuts and Examiner ensures that appropriate bolts are enough bolts are wrenches, the removed. removed / loosened to allow Examinee removes the spectacle flange to be the nuts and bolts SAT ( either pulled and from the flange. reinstalled are rotated. NOTE: *5. Examinee removes 5. Spectacle flange If Examinee requests new and rotates the rotated to the gaskets, inform them new spectacle flange thru position gaskets are not required. to the thru with the two position and flexitallic places flange gaskets and between two placed between flexitallic the two pipe gaskets, and into ends. position between the two pipe ends. SAT
*6. Examinee reinstalls 6. Flange nuts and flange bolts and bolts tightens nuts. reinstalled.
SAT 4 Page 4 of 8 File \0618010504. doc
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Task 0618010504 i l EXAMINER CUES EXAMINEE ACTIONS STANDARD POSITIVE CUE: i If Examinee performs steps 4, l 5 and 6 correctly, inform them that the spectacle flange is installed properly.
]
l NEGATIVE CUE: 1 If Examinee performs steps 4, ! 5 and 6 incorrectly, inform ) c them that the spectacle j flange is not installed j properly. l 1 JPM.may be terminated at this time. (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: 4 1 l-l Page 5 of 8 File \0618010504. doc 4
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i 1 Task 0618010504 j
. \
( TITLE: Change the spectacle flange from closed to open between EF-V-4 and EF-V-5. 1 1 JPM NUMBER: l TASK NUMBER: 0618010504 EXAMINEE: ) EVALUATOR: DATE: EVALUATION OF l EXAMINEE JPM: SAT: UNSAT: COMMENTS: I. t i l l 1 . EVALUATOR t SIGNATURE: 4 1 < Page 6 of 8 File \0618010504. doc i
Task 0618010504 PLANT CONDITIONS k The Plant was initially at 100% power with EF-P-2A O.O.S. A Trip of BOTH FW Pumps causes the Reactor & Turbine to Trip. l The Sec. CRO notes:
- EFW flow is low
- Discharge pressure on EF-P-1 is 980 psig
- EF-P-1 speed is 3400 rpm Assuming the Intermediate Building is inaccessible, what action could be taken to increase EFW flow to the OTSGs?
k-ANSWER: j CRO can lower OTSG pressures l l 1 l I l l PEDIGREE INFORMATION: i TIME ESTIMATE: 4 MIN. K&A NUMBER AND VALUE: 007 EA1.08 RO:4.4 SRO:4.3 CFR: 41.7/45.5/45.6 OBJECTIVE / RATING: IV.C.05.14 SRO: 4.8 COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
1210-10 HISTORY: QJ4C05-14-QO1 Page 7 of 8 File \0618010504 doc h i
,~
e Task 0618010504 PLANT CONDITIONS The unit was initially at 100% power when a Loss of Offsite Power concurrent with an ESAS actuation occurs. How long after this event do the motor-driven Emergency l Feedwater Pumps get a start signal? ) Include in your explanation the individual time elements that make up the EFW start delay. i ANSWER: With ESAS 10 sec. for D/G start (0.1) 5 sec. Block 2 (0.1) , 5 sec. Block 3 (0.1) l
- k. 5 sec. Block 4 (0.1) 5 sec. EFW start delay relay (0.1) 30 sec. Total (0.5) l l
l PEDIGREE INFORMATION: TIME ESTIMATE: 3 MIN. K&A NUMBER AND VALUE: 061 K4.02 RO:4.5 SRO:4.6 CFR: 41.7/45.6 OBJECTIVE / RATING: IV.C.05.15 SRO:3.8 j COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
None HISTORY: JJ4C05-15-QO1 l l Page 8 of 8 File \0618010504. doc
l ES-301 Administrative Topics Outline Form ES-301-1 Facility: TMI Date of Exa tion: Examination Level (circle one): RO / SRO Operating Test Number: Administrative-c Describe method of evaluation: Topic / Subject 1. ONE Administrative JPM, OR Description 2. TWO Administrative Questions l A.1 Event ,
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ES-301 Individual Walk-Through Test Outline Form ES-301-2 (. Facility: T/MI of Examination: 8/M/97 l Exam Level (circle one): RO/ / RO(U) perating Test No.: System / .1PM Title / Type Safety Planned Follow-up Questions: l' Ccdes* Function . K/A/G - Importance - Description 1.0000240501 @y/c. a. 004K4.05 3.2 MUT P/T limits ! DSL b. 024AA2.02 4.4 Emer Boration
- 2. 0000300501 gg,'g a 014K5.02 3'.3 IRPI VS Demand MSL (I) b. 003AK3.013.9 Rod Cont Interlocks 3.0008140401 ** a 011A1.02 3.6 PRZ Level Control NS S b. 011A2.03 3.9 PRZ Level Control l
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- 4. 000C100501 a. 041K6.03 2.9 Loss of ICS Power l DS b.041K4.18 3.6 Steam Leak Response 5.0538010101 ** a. 2, /. D 7O h * ** '*
(. ' Y l NS b. 7.l.14 1, T Nt CrP /' M 6.0648000101 a. 055EA2.03 4.7 Loss of OS Power I b DSL b. 062K4.02 2.7 480V Breaker Reset 7.3448050303 ** gg/g a. 011EK3.08 3.9 LBLOCA Recirc l DSL h) b. 011EK3.08 3.9 Piggyback Operation 8.0410040104 ** a. 045K$.42 3.0 Turbine Bypass Control DP b 045K4.42 3.0 Turbine Bypass Control l 9.0610130104 a. E05EK3.2 4.0 Overcooling Event DPL b. E03EK1.2 4.0 OTSG Level Control 10.0618010504 ** a.007EA1.08 4.3 EFW Flow g Requirements DPL b.061K4.02 4.6 EFW Start Time Delay
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- Type Codes: (D)lrect from bank, (M)odified from bank, (N)ew, (A)lternate path,
* (C)ontrol room, (S)imulator, (L)ow-Power, (P)lant, (R)CA f ** SRO Upgrades
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2.3. Total - 3 a. 4 2.4.37 Authorization for dedation from Technscal Spec. 3.5 10 i l i 2.4.42 Purpose of TSC 3.7 81 3 i Emergency emergency Ron Protec50n h Guiestas 1 Procedures- 2444 +0 e i l ever Priore, anon una c.m m l
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(. /s.sSitioi-f in, I NUREG-1021 21 of 26 Interim Rev. 8, January 1997
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l NUREG-1021 22 of 26 Interim Rev. 8, January 1997
I l l f THREE MILE ISLAND UNIT 1 f SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1 l l l i I i l l SCENARIO NUMBER: 1 l l ! l 1 1 l l EXAMINERS: 1 ! I i l l OPERATORS: 1 l l r I l l l l l l I e i j- Page 1 of 13 l l i
. . . - - . . . - . . _ . - - . . ~ _ . . - - - . . . . . - . _-. - . . - . .
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1.0 GENERAL DESCRIPTION OF SCENARI'i , l Plant is at 100% power, steady state, EOL, with MU-P-1B out of service I NI-5 fails high, requiring actions to stop continuous inward rod motion and reactor trip on low pressure. CO-P-1B will trip with failure of CO-P-1C to Auto Start resulting in Main Feed Pump trip and ICS runback. l During plant shutdown a Steam Generator Tube Rupture will occur resulting in a reactor trip and HPl. l Estimated scenario time - 65 Minutes l l l l t l Page 2 of 13
. - . . . - . . . . . - . - _ . - . .. - ~ . . - . _ - - . - -
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l-
2.0 REFERENCES
A. 10 CFR 55.45 Operating Test, (a) Content l l B. Procedures
- 1. 1102-10 Plant Shutdown
- 2. 1210-1 Reactor Trip
- 3. 1210-2 Loss of 25'F Subcooled Margin
- 4. 1210-5 OTSG Tube Leakage l 5. 1210-10 Abnormal Transients Rules, Guides and Graphs l
C. TECHNICAL SPECIFICATIONS
- 1. Section 3.1.6, RCS LeakaQe i
l. I Page 3 of 13
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 4 3.0 SCENARIO INITIALIZATION A. IC-17,100% power, steady state, equilitwium Xenon, EOL Following the initialization:
- 1. Shift MU-P-1 A cooling source to NSCC using Remote Function CC12.
f
- 2. Start MU-P-1 A
- 3. Secure MU-P-1B and red tag both extension controls in Pull-to Lock
- 4. Rackout MU-P-1B breaker using Remote Function MUR21
- 5. Start IC-P 1B and secure IC-P-1 A B. MALFUNCTIONS
- 1. IC47 NI 5/6 SASS channel failure activate immediately
- 2. N120A NI-5 fails high; assign to Remote #1
- 3. FW22A CO-P-2A trip; assign to Remote #2
- 4. IA01C IA-P-4 trip; assign to Remote #3
- 5. THISB B-OTSG Tube Rupture; assign at 4% severity with a 180 second ramp to Remote #4 6 RWO9A DR-P-1 A ES start failure; insert immediately C. OVERRIDES
- 1. 01 A4S28-DI, NAT, OFF; Activate immediately to prevent CO-P-28 auto start when CO-P-2A trips
- 2. 02A2SO9-DI, , OFF; Activate immediately to prevent NI-6 selection D. MONITOR
- 1. MSK2609A; Set to 14.7 Page 4 of 13
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION d 4.0 SCENARIO PREVIEW
< A. EXAMINER PREVIEW
- 1. Plant is at 100% power, steady state, with MU-P 1B out of service for an oil change.
- 2. NI-5 will fail high and SASS fails to actuate, causing continuous rod insertion. Operators will not i be able to select NI-6 attemate ICS NI signal. Crew must respond to stop the rod insertion to prevent a reactor trip and stabilize the plant. (CT) CO-V-51 may open.
- 3. After crew has taken the required action for the Nl failure CO-P-2A breaker trips and CO-P 2B fails to auto start. This causes FW-P 1B to trip and requires manual plant runback. CO-P-2B
- may be manually started at any time, and if so, CO-V-51 will open, j 4. CO-P-2B must be restarted and FW-P-1B reset and brought up in speed
- 5. lA-P-4 trips and actions must be taken to supply Instrument Air from lA-P-1 A and/or B.
- 6. A Tube Rupture (approximately 800 gpm) will occur on B OTSG. The crew should initiate a plant shutdown then reactor trip in accordance with ATP 1210-5 and ATP 1210-1. (CT)
- 7. 25'F subcooled margin may be lost, and the crew must respond in accordance with ATP-1210-2.
! RC Pumps must be secured (CT).
- 8. When 25'F subcooled margin is regained RC Pumps may be restarted.
l 9. HPl must be throttled to minimize subcooling margin and control RCS pressure within the allowable region of Figure 1 for the current plant conditions. (CT) , 1
- 10. If the crew isolates the B OTSG, RCS pressure must be maintained less than 1000#. (CT)
, 11. The plant should be stabilized and then a cooldown commenced in accordance with l l ATP-1210-5. 1 i l ) i Page 5 of 13
m -_ ._.m .. _ _ .. . _ . . _ _ _ _ _ _ _ - . _ _ _ THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION j B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. Plant is at 100% power, steady state, equilibrium Xenon,640 EFPD.
- 2. RCS boron is 36 ppm and the BAMT is the Tech. Spec. Tank.
- 3. MU-P-1B is out of service for oil change. It is expeded to be retumed to service within 2 hours.
- 4. MU-P 1 A is in service and being cooled by NSCCW.
- 4. No oth9r major maintenance or surveillance is in progress.
- 5. There are no releases in progress and none scheduled for this shift.
- 6. Continue 100% power operation.
Page 6 of 13
THREE MILE ISLAND UNIT 1 !' SENIOR REACTOR OPERATOR 1 SIMULATOR EXAMINATION l 5.0 SEQUENCE OF EVENTS Examiner Notes and Actions Expected Operator Actions j- 1. Initialize the simulator in accordance with Section 3.0. I
- 2. Assign team positions and condud the SHIFT 2.1 Assume assigned team positions.
BRIEFING per Section 4.B. 2.2 Take tumover and review plant status. l NOTE: Allow crew 3 5 minutes to take tumover and I I assume the watch. 2.3 Assume the watch'.
- 3. ICO-MALFUNCTION: When directed by 3.1 TEAM /PCRO - Diagnose and announce NI 5 examiner, activate Remote #1 to cause NI 5 t failing high-fail high. j Understand and Interpret Annunctators and Alarms
!~ NOTE: CO V-51 may open 3.2 Ig&M - Diagnose SASS actuation failure with continuous rod insertion. Diagno.se Events and Conditions Operate ControlBoards 3.3 PCRO/SCRO - Take manual control of the Diamond Rod Control and Feedwater to ManualReactivity Controlfor SRO-l stabilize the plant. (CT) NOTE: The operator will be unable to select Ni-6 3.4 IEAM - Determine that the ICS will have to because of a failure of the NI-6 selector remain in manual due to the inability to select l pushbutton. an operable N!. Demonstrate Supervisory Absty 3.5 IjaE Contact lac for troubleshooting. Page 7 of 13
- THREE MILE ISt.AND UNIT 1 i- SENIOR REACTOR OPERATOR i SIMUt.ATOR EXAMINATION 1 Examiner Notes and Actions Expected Ocerator Actions i NOTE: The minimum degree of redundancy required 3.6 3,33E - Determine than the minimum degree of for all RPS trip channels is ONE. With Ni-5 redundancy required for power range failed high, the actual degree of redundancy = 3 instrument channels per T.S. Table 3.51 is operable channels minus 1 channel required to satisfied. trip = TWO. Comply with and Use Tedn. Specs i Demonstrate Supon# cry Abdty 3.7 315E - Direct 'A' RPS to be placed in Manual Bypass after plant stabilization.
- 4. ICO MALFUNCTION: After the plant has been 4.1 IE&M- Diagnose CO-P 2A and FW-P 1B trip stabilized at normal plant operating conditions by alarms (MAP M-17) and panel indications.
or when directed by examiner, activate Remote Note that CO-P 2B did pgl auto start.
#2 to cause CO-P-2A to trip.
Recognize higher than normal amps on Understand and Interpret Annuncdors and Alarms CO-P 2C. Diagnose Events and Condhons Understand Plant and System Response Perform manual plant runback due to FW Operato ControlBoards Pump trip. (MAP H-1-1) ManualReactivity Controltbr SRO-l EXAMINER /lCO: If reactor trips for any reason, ! continue with Steam Generator Tube Rupture. i NOTE: CO-P 2B did not auto start as designed, 4.2 IE8M- Make Plant announcement for FW resulting in FW-P-1B trip and the need for an Pump trip and power reduction. Dispatch ICS runback. CO-P-28 may be manually personnel to CO-P 2A and its breaker. started anytime, and if so, CO-V 51 will open on high D/P if not already open. Page 8 or 13
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Expected Operator Actions ICO-ROLE PLAY: If diteded to check CO-P-2A and/or 4.3 3.QEQ - Start Co-P-28 from CL and diagnose it's breaker, provide Crew with the following CO V-51 open on high AP by alarms. , reports after a time delay (3 to 5 minutes)- l CO-P-2A breaker overload relays are tripped. 4.4 IJAM - Stabilize Plant following power i CO-P-2A motor smells bumt. reduction, and initiate action to raciose CO V-51 and reset FW-P-18. ICO/ REMOTE FUNCTION: When ditaded to reset 4.5 3GBQ - When plant conditions allow or when i CO-V 51 at Powdex panel, use FWR40 direded, reset and reclose CO-V-51 per ARP ] PAP 1-8. Operate ControlBoards . 1
- NOTE: Restarting FW-P-1B is required as a normal 4.6 3.QEQ - Restart FW-P-1B per OP 1106-3.
{ evolution, ]. j-i Operate Control Boards
- 5. . ICO-MALFUNCTION: After plant is stable and 5.1 $28Q - Recognize failure of IA-P 4 from alarm CO-V 51 is closed, or as directed by an PLB-1-6.
examiner, activate Remote #3 to trip lA-P-4. Understand and Interpret Annurociators and Alarms Communicate and Interact With the Crew ICO/ REMOTE FUNCTION: Use LAR04 to OPEN 5.2 3.Q39 - Contact AO to open lA V 2104A/B and IA-V-2104A/B and IAR01 and/or IAR02 to place place IA-P-1 A and/or IA P-1B in Hand per ARP IA-P 1 A and/or IA-P-1B in HAND as directed by PLB 1-6. the SCRO. Page 9 of 13 x
l THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION , l Examiner Notes and Actions Expected Operator Actions
- 6. ICO-MALFUNCTION: When directed by an 6.1 IEAM- Diagnose that based upon RCS I examiner, activate Remote #4 to initiate OTSG inventory loss and RMS indications that an 18 tube rupture on a 3 minute ramp. OTSG tube leak / rupture has developed.
Understand and Interpret Annunciators and Alarms Diagnose Events and Conditions NOTE: A reactor trip may be initiated prior to 6.2 EE-Implement ATP-1210-5. implementing ATP 1210-5,in which case ATP-1210-1 would first be implemented. Comply With and Use Procedures ' Demonstrate Supervisory Ability Comply With and Use Procedures 6.3 EQEQ - Increase makeup and isolate letdown ' Operate ContmlBoards to attempt to maintain Pressurizer level. I Comply With and Use Procedures 6.4 PCRO - May announce his intent to trip the ! Operate ControlBoards reactor to the Team due to < 150' Pressurizer Communicate and Interact Wth the Crew level while reactor power is > 25%, then trip the reactor and initiate HPI based on the ATP NOTE: Reactor trip will occur either manually or 1210-10 Pressurizer Level Guide, automatically OR TEAM - May manually trip the reactor due to low RCS pressure. l 6.5 TEAM - Respond to the reactor trip per the immediate actions of ATP 1210-1. NOTE: Full HPI is considered to be one HPI pump 6.6 PCRO -Initiate HPl or verify ESAS actuation running in each train (2), and full flow through as necessary all four HPl lines (MU-V-16A,B, C & D full open) Page io of 13
l I
. THREE MILE ISLAND UNIT 1
!. SENIOR REACTOR OPERATOR l SIMULATOR EXAMINATION s Examiner Notes and Actions Expected Operator Actions NOTE: If SCM is lost, it will most likely be quickly 6.7 M -If 25*F SCM is lost, perform regained. immediate actions of ATP 1210 2: (CT) i ComfWy With and Use Procedures 6.8 Af- Verify /dited ATP 12101 immediate Demonsente Supervtaary Abary actions. 6 I understNxi Pllant and System Response 6.9 IE8M - Recognize that DR-P-1/ did not start l on ESAS. Com$dy With and Use Tech. Specs. 6.10 E.QBQ ~ Start DR P 1 A Comply With and Use Procedwes 6.11 BE - Verify / direct ATP 1210 2 immediate Demonstrate Supervisory Ability actions for Loss of 25*F SCM if applicable; otherwise verify / direct necessary actions for increased loss of RCS inventory (initiation of HPI) into B OTSG. NOTE: SF may be allowed to classify the event at the 6.12 31- Declare the emergency event as end of the scenario at examiner's discretion. necessay for the SG tube leakage and/or loss of RCS subcooled margin per the Emergency Demonstrate Supervisory Abikty Plan. ICO/ EXAMINER: ROLEPL.AY as GRCS (RAC) and if requested by SS/ED, supply offsite dose projections of 1 mrem TEDE and 5 mrem CDE. NOTE: B OTSG will eventually fill due to secondary 6.13 IE8M - If/when 25'F primary to subcooled leskrate which may cause concem for potential margin was lost and subsequently recovered, equipment damage. Isolation of B OTSG for restart RC Pumps and control HPIin this reason is a subjective call. accordance with ATP 1210-2 and 1210-10, and transition back to ATP 1210-5 to commence Comply With and Use Procedures RCS cooldown. (CT) Page 11 of 13
. - .~... - . - ._. - - - -~
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Expected Operator Actions NOTE: B OTSG isolation during this scenario is not 6.8 133E- Evaluate rate of 8 OTSG fill and considered critical; however, it is critical that possibility of B OTSG isolation due to potential RCS and B OTSG pressure be maintained less equipment damago per ATP 1210 5. than =1000 psig to prevent MSSV lift. 1
- Understand Pfant and System Response \
Comply WNh and Use Rocedures 6.9 M - Ensure RCS temperature is less than 540*F (T hot or incore) and RCS pressure less than 1000 psig prior to isolation of B OTSG per ATP-1210-5. Note: B OTSG isolation during this scenario is not 6.10 M - Control RCS inventory / pressure by required prior to termination. throttling HPI for a 30 70'subcooling margin, and control B OTSG pressure using TBVs, to Operate ControlBoards prevent MSSV lift on B OTSG (=1040 psig), which woulo cause an unpartitioned and unmonitored radioactive release to atmosphere. (CT) Page 12 or 13
I THREE MILE ISLAND UNIT 1 I
' SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
- 6. TERMINATION POINT When all of the following conditions exist:
- 1. Crew has had sufficient opportunity to perform all applicable critical tasks.
- 2. ATP 1210 5 follow up actions are in progress.
l
- 3. Subcooling margin has been minimized (30 - 70 degrees).
- 4. A plant cooldown is in progress.
- 5. All examiners agree on termination.
- 7. EMERGENCY PLAN DECLARATION (Admin Section A.4)
If 25'F Subcooling Margin is maintained:
~
Alert- EAL AS.1 (RCS/OTSG Leakage) J'- A _. t if 25'F Subcooling Margin is lost: Site- EAL S5.1 (RCS/OTSG Leakage) or Site - EAL SS.2.1 (Loss of Subcooling) , 1 l l 4 Page 13 or 13
Sc: naris Outline Facility: TMI Scenario No.: 1 Op Test No.: 1 i Examiners: Operators: Objectives: NL5 will fail high and SASS fails to aduate, causing continuous rod insertion. Operators will not be able to select NL6 allemate ICS Ni signal. Crew must respond to stop the rod insertion to prevent a reador trip and stabilize the plant. CO-P-2A breaker trips and CO-P-28 fails to auto start. This causes FW-P-1B to trip and requires manual plant runback. A Tube Rupture (approximately 800 gpm) will occur on B OTSG. The crew should initiate a plant shutdown then reactor tiip in accordance with ATP 1210-5 and ATP 1210-1. 25'F subcooled margin may be lost, and the crew must respond in accordance with ATP-1210-2. RC Pumps must be secured. When 25'F subcooled margin is regained, RC Pumps may be restarted. HPI must be throttled to minimize subcooling margin and control RCS pressure within the allowable region of Figure i for the current plant conditions. If the crew isolates the B OTSG, RCS pressure must be maintained less than 1000#. The plant should be l stabilized and then a cooldown commenced in accordance with ATP 1210-5. Initial Conditions: IC-17,100% power, steady state, equilibrium Xenon, EOL Following initialization: l (1) Shift MU-P-1 A cooling source to NSCC using remote function CC12 (2) Start MU-P1 A, (3) Secure l MU-P-18 and red both extension controls in puikto lock, (4) Rack out MU-P-1B breaker using remote l function MUR21, (5) Start IC-P-1B and secure IC P-1A. Turnover: Plant is at 100% power, steady state, equilibrium Xenon,640 EFPD. RCS boron is 36 PPM and the BAMT is the Tech. Spec. Tank. MU-P-1B is out of service or an oil change. it is expected to be retumed to service within 2 hours. MU-P-1 A is in service and being cooled by NSCCW. No other major maintenance or surveillance is in progress. There are no releases in progress and none scheduled o this shift. Continue 100% power operations. Event Malf. No. Event Event No. Type
- Description
- 1. N120A (I), (R) Nb5 fails high, CO-V 51 may open due to taking feedwater in hand.
Continuous rod insertion controlled by the PCRO by taking hand IC47 control of the diamond rod control station.
- 2. FW22A (C),(R) CO-P-2A trip, feedwater pump trip, operators must manually control rods and feedwater.
(N) Restart FW-P-18.
- 3. lA01C (C) lA P-4 trip, operators respond to loss of IA-P-4
- 4. THISB (M) (R) B-OTSG Tube Rupture,4% severity, recognize leak, reduce power, possible automatic reactor trip, may manually trip the reactor, possible ESAS actuation, possible loss of SCM
- 5. RWOgA (C) DR-P-1 A ES start failure during the OTSG Tube Rupture (N)orrnal, (R)eactivity, (1)nstrument, (C)omponent. (M)ajor Page i of 7
Operat:r Actions Op Test No.: 1 Scenario No.: 1 Event No.: 1 Page ,,2.,, of I., Event
Description:
NI-5 fails high, CO V 51 may open due to taking feedwater in hand. Continuous rod insertion controlled by the PCRO by taldng hand control of the diamond rod control station. Time Position Applicant's Actions or Behavior TEAM /PCRO Diagnose and announce Nl 5 failing high.
- Diagnose events and conditions M Diagnose SASS actuation failure with continuous rod ir.ssrtion.
- Diagnose events and conditions PCRO/SCRO Take manual contml of the Diamond Rod Control and Feedwater to l stabilize the plant. (CT) ;
- Manual reactivity control for SRO 1
- Feedwater in manual M Determine that the ICS will have to remain in manual due to the inability to select an operable Nl.
- NI-6 selector pushbutton failure 3,33f Determine than the minimum degree of redundancy required for power range instrument channels per T.S. Table 3.5-1 is satisfied.
- Comply with and use Tech. Specs
- Direct 'A' RPS to be placed in manual bypass Page 2 of 7
Cperator Actions Op-Test No.: 1 Scenario No.: 1 Event No.: _ ,2_ Page ,, }_ of,,Z_ Event
Description:
CO-P-2A trip, feedwater pump trip, operators must manually control rods and feedwater. Time Position Applicant's Actions or Behavior RAM Diagnose Co-P 2A and FW-P-1B trip by alarms and panelindications. Note that CO-P-2B did agt auto start.
- Understand and Interpret Annunciators and Alarms
- Recognize higher than normal amps on CO-P 2C.
- Perform manual plant runback due to FW Pump trip
- Understand Plant and System Response
- Operate Control Boards l
- Manual Readivity Control for SRO-l IFAM Make Plant announcement for FW Pump trip and power reduction.
- Dispatch personnel to CO-P-2A and its breaker.
- CO-V-51 opens SCRO Start Co-P-2B from CL
- Diagnose CO-V 51 open on high AP by alarms.
IFAM Stabilize Plant following power reduction
-Initiate action to reciose CO-V 51
- Initiate action to FW-P-18.
g When plant conditions allow or when directed:
- Reset and reclose CO-V-51 per ARP PAP-1-8.
Restart FW-P-1B per OP 1106-3. I l f I l Page 3 of 7
Operator Actions Op Test No.: 1 Scenario No.: 1 Event No.: 3 Page 4,_ of J.,,, Event
Description:
lA-P-4 trip, operators respond to loss of IA-P-4 Time Position Applicant's Actions or Behavior g Recognize failure of IA-P-4 from alarm PLB 16. Understand and Interpret Annunciators and Alarms Communicate and interad with crew g Recogniza failure of IA-P-4 from alarm PLB 1-6. Contact AO to open IA-V 2104A/B and place IA-P-1 A and/or IA-P-1B in Hand per ARP PLB-1-6. Page 4 of 7
Operatrr Acti:ns Op-Test No.: 1 Scenario No.: 1 Event No.: f_ Page _5_ of _Z_ Event
Description:
B-OTSG Tube Rupture,4% seventy, recognize leak, reduce power, possible automatic reador trip, may manually trip the reactor, possible ESAS aduation, possible loss of SCM l Time Position Applicant's Actions or Behavior IE8M Diagnose that based upon RCS inventory loss and RMS indications that an OTSG tube leak / rupture has developed. i
- Understand and Irterpret Annunciators and Alarms
- Diagnose Events and Conditions RE Implement ATP-1210-5.
P.qB.Q Increase makeup and isolate letdown to attempt to maintain Pressurizer level. 4 f.qB.Q- May announce his intent to trip the reactor to the Team due to:
< 150' Pressurizer level while reactor power is > 25%,
Lien trip the reactor and initiate HPl based on the ATP 1210-10 i Pressurizer Level Guide. TEAM May manually trip the reactor due to low RCS pressure.
- Respond to the reactor trip per the immediate actions of ATP 1210-1.
PCRO initiate HPl or verify ESAS actuation as necessary Full HPl is considered to be one HPl pump running in each train (2), y and full flow through all four HPllines (MU V 16A,B, C & D full open) TEAM If 25'F SCM is lost, perform immediate actions of ATP 1210-2: (CT) 5 Ef Verify / direct ATP 1210-1 immediate actions. af Verify / direct ATP 1210-2 immediate actions for Loss of 25'F SCM If applicable; otherwise verify / direct necessary actions forincreased loss of RCS inventory (initiation of HPI)into B OTSG. 3.1 Declare the emergency event as necessary for the SG tube leakage and/or loss of RCS subcooled margin per the Emergency Plan. SF may be allowed to classify the event at the end of the scenario at examiner's discretion. TEAM If/when 25'F primary to subcooled margin was lost and subsequently recovered, restart RC Pumps and control HPIin accordance with ATP 1210 2 and 121010, and transition back to ATP 1210-5 to commence RCS cooldown. (CT) B OTSG will eventually fill due to secondary leakrate which may cause concem for potential equipment damage. Isolation of B OTSG forthis j reason is a subjective call. SS/SF Evaluate rate of B OTSG fill and possibility of 8 OTSG isolation due to potential equipment damage per ATP 1210-5.
- B OTSG isolation during this scenario is not considered critical; however, it is critical that RCS and B OTSG pressure be maintained
, less than =1000 psig to prevent MSSV lift. Page 5 of 7
Operator Actions
.TI.AM Ensure RCS temperature is less than 540*F (T-hot or incors) and RCS pressure less than 1000 psig prior to isolation of B OTSG per ATP-1210 5.
IE&M Control RCS inventory / pressure by throttling HPI, and control B OTSG pressure using TBVs, to prevent MSSV lift on B OTSG (=1040 psig), which would cause an unpartitioned and unmonitored radioactive release to atmosphere. (cT) B OTSG isolation during this scenario is not required prior to termination. Page 6 of 7
Operator Actions Op Test No.: 1 Scenario No.: 1 _ Event No.: 5 Page 7 of 7 Event
Description:
DR-P 1 A ES start failure during the OTSG Tube Rupture Time Position Applicant's Actions or Behavior gg Recognize that DR-P-18 did not start on ESAS.
- Start DR-P-1 A
- Comply with Tech Specs
- Comply wit use of procedures Page 7 of 7
THREE MILE ISLAND UNIT 1 s SENIOR REACTOR OPERATOR ! SIMULATOR EXAMINATION j j SCENARIO NUMBER: 2 EXAMINERS: l l OPERATORS: e Pa081of11
l l THREE MILE ISLAND UNIT 1 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1.0 GENERAL DESCRIPTION OF SCENARIO 4 Plant is at 10 amps during a plant startup and crew continues power increase into point of adding heat. One channel of intermediate Range SUR fails during power increase. A group 6 control rod drops requiring a shutdown to 1%Alvk. 1 During establishment of shutdown reactivity, IC-P-1 A trips with failure of auto / manual start of IC-P 18. The total loss of intermediate Closed Cooling Water will require a manual reactor trip due to high CRD stator , temperatures. ' When the unit is tripped, two main steam safety valves on the B OTSG fail full open and FW-V-168 will fall full open with an HSPS low pressure isolation failure resulting in excessive primary to secondary heat transfer. l 1600# ESAS actuation on channels 2 and 3 will not occur, requiring manual HPI actuation. Subcooled margin may be lost if HPl is not initiated. 4 oe B OTSG must be isolated to terminate the excessive primary to secondary heat transfer, if HPl was actuated, it will require throttling /tennination to control RCS pressure when the B OTSG blows dry. Plant events and conditions will warrant declaration of the Emergenc) Plan. l 1 Estimated scenario time - 60 Minutes l I l l Page 2 or 11
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1 l THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
2.0 REFERENCES
A. 10 CFR 55.45 Operating Test, (a) Content B. Procedures
- 1. 1102-2 Plant Startup
- 2. 1202-8 CRD Equipment Failure l
- l. 3. 1202-17 Loss ofIntermediate Cooling System
- 4. 1210-1 ReactorTrip
- 5. 1203-24 Steam Leak 1
l
- 6. 1210-3 Excessive Primary to Secondary Heat Transfer
- 7. 1210-10 Abnormal Transients Rules, Guides and Graphs C. TECHNICAL SPECIFICATIONS
- 1. Section 3.5.1.1, instrumentation System, Table 3.51 t
i i i l l I l l Page 3 of 11
-. . - - - - - . - .- _-- ~ _ . - - _ - . . - - _ - - - - . - . - . . . -~- ~ _ . -
i THREE MILE ISt.AND UNIT 1 ' @ENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1
- 3.0 SCENARIO INITIALIZATION l A. IC-0, Hot Zero Power,104amps, Xenon free,640 EFPD
- 1. Use Remote Fundion MSR21 to OPEN AS-V-8 j
- 2. Close GS-V-4 on PL B. EVENT
- 1. Event #1 RD:CFTRP C. MALFUNCTIONS l
- 1. N112A NI-3 SUR Ampidier Failure; insert at 50% severity insert immediately
} i 2. RD0115 Dropped Rod (6-1); assign to Remote #1
- 3. CC04A IC-P-1 A trip; assign to Remote #2
- 4. CC04B IC-P-1B trip; assign to Remote #3
- 5. MS05A MS-V-18C failure; assign at 100% severity to Event #1
- 6. MS058 MS-V-18D failure; assign at 100% severity to Event #1
- 7. FW118 FW-V-168 failure; assign at 100% severity to Event #1
- 8. ES01A ESAS failure to actuate A (1600#); insert immediately
- 9. ES01B ESAS failure to actuate B (1600#); insert immediately l
- 10. MUO8A MU-V-16A fails to open on ESAS; insert immediately
- 11. MS10SGB, TRNA; HSPS low pressure isolation failure; insert immediately
- 12. MS10SGB, TRNB; HSPS low pressure isolation failure; insert immediately D. OVERRIDES None Pa084of11 u__ _ , _
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 4.0 SCENARIO PREVIEW A. EXAMINER PREVIEW 1, 4 Reactor critical at 10 amps during xenon-free plant startup with CRG-6 at 64%.
- 2. Reactor power is increased per OP 1102-2, and NI 3 startup rate ar plifier fails to respond. The crew must address Tech. Specs. for the failed NI and determine no action required.
- 3. When reactor power reaches =1%, a dropped control rod will occur. The crew will have to establish 1% Alvk subcriticality. (CT)
- 4. During establishment of shutdown reactivity IC-P-1 A will trip with auto / manual start failure for IC-P-18. The crew will have to manually trip the reador in accordance with 1202-17 Loss of Intermediate Closed Cooling System.
- 5. When the reactor is tripped, two MSSVs on the B-SG will fail 100% open. Also B OTSG will suffer an over feed due to FW-V-16A failing 100% open, resulting in excessive primary to secondary heat transfer.
- 6. The crew must isolate the B OTSG in accordrence with ATP 1210-3 to mitigate the excessive primary to secondary heat l'ansfer. (CT)
- 7. HPl will NOT automatically atAuate due to ESAS 1600# channels 2 and 3 bistable failures, requiring manual HPl Initiation. MU-V-16A will fail to open on ESAS. (CT)
- 8. HPl must be throttled to comply with RCS pressure requirements of 1210-10, Figure 1/1 A. (CT)
- 9. During the scenario th i appropriate E-Plan declaration will be made.
1-l 5 l Page 5 of 11
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR l SIMULATOR EXAMINATION t l B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. 4 Reactor is critical at 10 amps during a xenon-free startup at 640 EFPD.
- 2. RCS boron is 543 ppm.
- 3. Plant was previously at 100% power for 65 days and was shutdown for main turbine bearing replacement. It has been 7 days since the plant shutdown.
- 4. Operating main feedwater pump and gland steam are both on auxiliary steam.
L
- 5. No maintenance or survedance is in progress.
- 6. There are no releases in progress and reone scheduled for this shift.
- 7. Continue plant startup.
l l l I i l r l Page 6 of 11 l
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 5.0 SEQUENCE QF EVENTS Examiner Notes and Actions Expected Operator Actions
- 1. Initialize the simulatorin accordance with Sedion 3.0.
- 2. Asse0n team posstions and condud the SHIFT 2.1 Assume assigned team positions BRIEFING per Section 4.B.
2.2 Take tumover and review plant status. Pmvide SF with sign-off copy of OP 1102-2 signed ~ off up to including step where reactor is taken 2.3 Assume the watch and inform exarniner. i critical per 1103-8. i NOTE: Allow crew 3 5 minutes to take tumover and 2.4 PCRO - Resume plant startup by pulling control assume the watch. rods in manual-sequence mode, to increase neutron flux level to point of adding heat. Operate ControlBoanis Reactmty Afanipulation for SRO-I NOTE: Ni-3 Startup rate amplifier is failed and will 3.1 E.QBQ - Diagnose failure of NI-3 startup rate to indicate 0 on allindications, respond to addition of positive reactivity, and report to SF. Understand Plant and System Response NOTE: Per T.S.3.5.1.1, Table 3.5-1 only one 3.2 133E - Dedare N1-3 inoperable, evaluate Tech. Intermediate Rance detector is required. Spec implications of N13 channel failure, and determine that plant startup may resume as long Comply With and Use Tech. Specs. as NI-4 remains operable. l 3.3 EQBQ -If stopped for NI 3 failure, resume plant startup by withdrawal of control rods.
- 4. ICO MALFUNCTION: At =1% power and when 4.1 Ig&M - Recognize dropped rod in CRG-8. l directed an examiner, activate Remote #1 to cause dropped rod in CRG-8. Must occur before 2%
power. Understand and Interpret Annunciators and Alarms Demonsfrate Supervisory Ability 4.2 113f - Dired action to insert the remaining Comply With and Use FYocedures rods to achieve at least 1% Ak/k shutdown per 1202-8. Comply wilth and Use Tech Specs 4.3 31- Determine inoperable rod per Tech. Spec. Section 4.7.1.2 Pa0e 7 of 11
4 THREE MILE ISt.AND UNIT 1
- j. SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Expected Operator Actions
- 5.0 ICO MALFUNCTION
- During the insertion of control 5.1 H&M - Diagnose IC-P 1 A trip and note failure i
rods to establish shutdown reactivity, a v.vate of IC-P 1B to automatically start. Dispatch AO to Remote #2 and #3 to cause a trip of IG-P-1 A and IC-P-1A/B pumps. I. IC-P-18. j Operate Con 6clBoards 5.2 EGBQ - Attempt to start IC-P-1 B, report to SF , Communicate and kitaract With the Crew that IC P 1B will not start. 1 1 Demonsdratt SupeMaory Abillry 5.3 RE-Implement EP 1202-17 and dired attempts ) 4-to restore ICCW, whits team (STA) monitors 1 j CROM temperatures and pressurizer level. l
- NOTE
- It will take approximately 5 minutes for CRDM 5.4 PCRO - Diagnose loss of seal injection due to
! temperatures to exceed 180*F. closure of MU-V 1 A/B at CRD Coolant Outlet Hi
' Temp. of 160*F.
ICO-ROLE PLAY as AO: if direded to check IC-P-1 A 5.5 333E - Prepare team for a manual reador trip i report that the motor is too hot to tooch required by EP 1202-8 due to high CRDM j If directed to check IC P-1 A breaker, report the temperatures.
- breaker is tripped.
j if direded to check IC-P-1B or its breaker, report 4 nothing obvious is wrong. 4 Demonstrate Supervisory Ability i Comply With and Use Pmcedures 1 4 5.6 333E - Dired team to manually trip reador when more than one CRDM stator temperature ] exceeds 180*F, and perform ATP 1210-1. Operate the ContolSoards 5.7 TEAM - Manually trip the reactor and perform Comply With and Use Procedures 1210-1 immediate actions. . 6.0 NOTE: When the reador is tripped, FW-V-168 fails 6.1 SCRO - Diagnose / report undesired MFW flow to 100% open and two MSSVs stick 100% open. B-0TSG, caused by failure of FW V 168 at 100% and stuck open MSSVs, and take adlon to . Diagnose Events and Conditions controi MFW to B-0TSG. l- Understand Plant and System Response l Operate ControlBoards NOTE: The excessive primary to secondary heat 6.2 HAM-Diagnose excessive primary to transfer is caused pnmarily by B-OTSG. secondary heat transfer of the RCS due to stuck Also FW-V-g2B will NOT automatically close when MSSVs on B OTSGs, and isolate the B OTSG in HSPS actuates at less than 600 psig in B-OTSG, but it accordance with ATP 1210 3 including the will close by push txAton on CC. following: Diagnose Events and Condihons Understand Plant and System Response Pape 8 or 11
- . - . .-. -- - - - - . . - -- - . . . . ~ - _ _ - . - - . - - - . --.
i THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1 Examiner Notes and Actions Ernected Operator Actions ICO-MALFUNCTION: If at any time you are direded as 6.3 gnBQ Close FW-V-928 to isolate FW-V 168 AO to isolate FW-V-16B and to locally dose FW-V- and eventually dose EF-V-30A and 300 in 168, aner =2 minute time delay change severity of manualtoisolate EFW (CT) FW118 to 0% over 60 seconds to simulate FW-V-168 localisolation. l Operate ContmlBoards NOTE: When RCS pressure decmases below 1640 6.4 EGBQ - Operate makeup valves and pumps as i psi 0, ESAS (HPI) wa NOT automatically actuate necessary to mairtain pressurizer level after trip. due to failed bistables for channels 2 and 3. Open or verify open MU V-148 prior to starting ! MU-P-1C for additional makeup flow. i Understand Plant and System Response Operate ControlBoards ! l NOTE: HPl may be aduated simply by depressing 6.5 ESEQ -When RCS pressure decreases below 1600# ESAS Train-A and Train-B manual aduation 1600 psig, diagnoses failure of ESAS to pushbuttons on CC and CR; or by initiating HPl at the automatically actuate HPl (or if pressurizer level component level in accordance with 'd1010. decreases below 20 inches), and manually initiate HPl- two pumps, full flow, ES alignment NOTE: If sufficient HPl is not initiated to compensate for per 1210 mior to loss of 25Y sut: cooled RCS shrink,251 subcooled margin may be lost. margin. Critical Task l Comply With and Use Procedures Operate ControlBoards l
- i Demonstrate Supevisory Ability 6.6 SF/ TEAM - Lead crew in verification of 1210-1 and then 1210-3 immediate adions for reador trip and isolation of feedwater to at least the A.
OTSG. 6.7 SF/ TEAM - Commence 1210-3 follow-up actions after alarm review. NOTE: - Examiner may wait for scenario termination. 6.8 SS/SF -Identify / declare an UNUSUAL EVENT i ' due to the MSSV failure causing OTSG Comply With and Use Procedures depressurization below 600 psig. L NOTE: A-OTSG pressure will stabilize after B-OTSG 6.9 IEAM - Recognize when the excessive primary goes dry and stops the excessive primary to to secondary cooling has stopped, requiring l secondary heat transfer, control of RCS reheat and repressurization per 1210-3. Understand Plant and System Response Comply With and Use Procedures Operate ControlBoattis l 1 6.10 31-If aduated, authorize defeat / enable of manual 1600# ESAS channels to throttle HPl and control RCS pressure. e Page 9 of 11 i
i l THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l l Examiner Notes and Actions Expec.ted Operator Actions l l l^ ! Demonstrate Supervisory Ab&ty 6.11 RE - Dired CRos to defeavenable manual ESAS channels and throttle HPl to stop RCS , repressurization. I (Jnderstand Pllant and System Response 6.12 Operate Control 8oards EGBQ - Operate HPl pumps and/or valves as ! necessary to throttle HPl flow and control RCS l pressure within allowable re0i on of 121010. l (CT) ' < l l l I l l 1 t t l l l I. i e t PaBe 10 of 11 l.
l' , THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
- 6. TERMINATION POINT:
l When all the following conditions exist:
- 1. Crew has had suffldent opportunity to perform all appbcable critical tasks.
- 2. ATP 12103 follow-up actions are in progress.
- 3. The plant is stabilized or being cooled down with HPl terminated.
- 4. All examiners a0ree on termination.
- 7. EMERGENCY PLAN DECLARATION (Admin Section A.4)
Usual Event - EAL US.1 (Depressurization/ Overcooling) l
)
l l l l~ 1 i i> I i Page 11 of 11 (
- - - . - . . = - - - - - -
l Scenarin Outline Facility;,,_ TMI Scenario No.: 2 Op Test No.: 2 Examiners: Operators: l l 4 Objectives: Reador critical at 10 amps during xenon free plant startup with CRG-6 at 64%. Reador power is increased per OP 1102 2, and NI-3 startup rate amplifier fails to respond. The crew must address Tech. Specs. for the failed NI and determine no action required. When reador power reaches
=1%, a dropped control rod will occur. The crew will have to establish 1% Alvk subcriticality. (CT)
During establishment of shutdown readMty IC-P-1 A will trip with auto / manual start failure for IC-P-1 B. The crew will have to manually trip the reador in accordance with 120217 Loss of Intermediate Closed Cooling System. When the reactoris tripped, two MSSVs on the B-SG will fail 100% open. Also B OTSG will suffer an over-feed due to FW V-16A failing 100% open, resulting in excessive primary to secondary heat transfer. The crew must isolate the B OTSG in accordance with ATP 1210-3 to mitigate the excesssve pnmary to secondary heat transfer. (CT) HPl will NOT automatically aduate due to ESAS 16001 channels 2 and 3 bistable failures, requiring manual HPl initiation. MU-V 16A will i fail to open on ESAS. (CT) HPl must be throttled to comply with RCS pressure requirements of 1210- l 10, Figure 1/1 A. (CT) During the scenario the appropriate E-Plan declaration will be made. l Initial Conditions: 10 9, Hot Zero Power,10 4amps, Xenon free,640 EFPD. Use Remote i Function MSR21 to OPEN AS-V-8. Close GS-V-4 on PL. 4 Turnover: Reactoris critical at 10 amps during a xenon-free startup at 640 EFPD. RCS boron is 543 ppm. Plant was previously at 100% power for 65 days and was shutdown for main turtaine bearing replacement. It has been 7 days since the plant shutdown. Operating main feedwater pump and gland steam are both on auxiliary steam. No maintenance or surveillance is in progress. There are no releases in progress and none scheduled forthis shift. Continue plant startup. Event Malf. No. Event Event No. Type
- Description (N) Plant Startup
- 1. N112A (l) NI 3 SUR Amplifier Failure
- 2. RD0115 (C),(R) Dropped Rod (61)
- 3. CC04A (C) IC-P 1 A trip -IC-P-1B will not start CC048
- 4. MS05A (M) (C) MS-V 18C, MS-V 18D, and FW-V 16B fail MS05B FW11B
- 5. ES01A (C), (1) ESAS failure to actuate (A and B), HSPS low pressure isolation ES01B MUO8A MS10SGB, TRN (N)ormal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor Page 1 of 6
1 Operator Action 3 I Op Test No.: _2 Scenario No.:1 Event No.: 1 Page , 2.,, of _,g Event
Description:
NI-3 SUR Amplifier failure Time Position Applicant's Actions or Behavior l Efr.8Q Resume plant startup by pulling control rods in manual-sequence mode, to j in::rease neutron flux level to point of adding heat. l - Reactivity Manipulation for SRO-l l EraQ Diagnose failure of NI-3 startup rate to respond to addition of positive reactivity, and report to SF. R$.3E Declare NI-3 inoperable, evaluate Tech. Spec. implications of Ni-3 channel ! failure, and detennine that plant startup may resume as long as NI 4 remains operable.
- Per T.S.3.5.1.1. Table 3.51 only one Intennediate Range detector is required.
EfRQ If stopped for NI 3 failure, resume plant startup by withdrawal of control rods. l l I I l 1 l l t l Page 2 of 6
_ . - - . . . __~ ._. . . -. .- OperatIr Actions OpTest No.:_2 Scenario No.: _2_. Event No.: _2_ Page _L of 6 Event
Description:
Dropped Rod (6-1) Time Position Applicant's Actions or Behavior TEAM Recognize asymmetric and probable dropped rod in CRG 6. 3.R!$f Direct action to insert the remaining rods to achieve at least 1% A10k shutdown per 1202 8. 31 Determine inoperable rod per Tech. Spec. Section 4.7.1.2 I l l 1 Page 3 of 6
_ __ - . _ . - _ -.- .- . =. - -._ - - -. . - - - Operator Actions l Op-Test No.: 2 Sc w o No.: 2 Event No.: 3 Page _!_ of 6 I Event
Description:
IC-P 1 A trip - IC-P 1B win not start Time Position Applicant's Actions or Behavior HAM Diagnose IC-P-1 A trip and note failure of IC-P-1B to automatically start.
- Dispatch AO to IC-P 1 A/S pumps.
Ef,EQ Attempt to start IC-P-1B, 4
- IC-P-1B will not start.
BE Implement EP 1202-17 and direct attempts to restore ICCW, while team (STA) monitors CRDM temperatures and pressurizer level. P_qEQ Diagnose loss of seal irijection due to closure of MU-V-1 A/B at CRD Coolant Outlet Hi Temp. of 160*F. SS/SF Prepare team for a manual reactor trip required by EP 1202-8 due to high CRDM temperatures.
- it will take approximately 5 minutes for CRDM temperatures to exceed 180*F.
Bjsf Direct team to manually trip reactor when more than one CRDM stator temperature exceeds 180'F, and perform ATP 1210-1. TEAM - Manually trip the reactor and perform 1210-1 immediate adions. l l l Page 4 of 6
Operat:r Actions Op Test No.: 2 Scenario No.: 1 Event No.:.J_ Page i_ of 6 Event
Description:
MS V 18C, MS-V 180, and FW-V 188 fail l Time Position Applicant's Adions or Behavior ! 3.GILQ Dia0 nose / report undesired MFW flow to B-0TSG, caused by failure of FW-V 168 at 100% and stuck open MSSVs, and take action to control l MFW to B.OTSG. l ! TEAM Diagnose excessive primary to secondary heat transfer of the RCS due to stuck MSSVs on B OTSGs, and isolate the B-OTSG in accordance with ATP 1210-3. The excessive primary to secondary heat transfer is caused primarily l by B-OTSG. Also FW-V-928 will NOT automatically dose when HSPS l aduates at less than 600 psig in B-OTSG, but it will dose by push- l button on CC. l SCRO Close FW-V-928 to isolate FW-V 168 and eventually dose EF-V 30A and 30D in manual to isolate EFW. (CV) PCRO Operate makeup valves and pumpc as necessary to maintain pressurizer l level after trip. Open or verify open ML'-V-14B prior to starting MU-P-1C for additional makeup flow. l l e i l l l l l 4 i Page 5 of 6
I Operat:r Actions Op Test No.:,,_2_,, Scenario No.: __2 _ Event No.: A Page 6 of 6 Event
Description:
ESAS failure to aduste (A and B), HSPS low pressure isolation failure Time Position Applicant's Adions or Behavior PfaQ When RCS pressure decreases below 1600 psig, diagnoses failure of ESAS to automatically aduate HPl (or if pressurtzer level decreases below 20 inches), and manually initiate HPl -two pumps, full flow ES alignment per 1210 gigt.ig loss of 259 subcooled margin. (CT) HPl may be actuated simply by depressing 1600# ESAS Train-A and Train-8 manual aduation pushbuttons on CC and CR; or by initiating HPl at the component level in accordance with 1210-10. If sufficient HPl is not initiated to compensate for RCS shrink,25'F subcooled margin may be lost. SF/ TEAM Lead crew in verification of 12101 and theit 1210 3 immediate adions for reador trip and isolation of feedwater to at least the A-OTSG. SF/ TEAM Commence 1210 3 fobow-up adions after alarm review. BASE Identify /dedare an UNUSUAL EVENT due to the MSSV failure causing OTSG depressurization below 600 psig. May wait until end of scenario to complete H&M Recognize when the excessive primary to secondary cooling has stopped, requiring control of RCS reheat and repressurization per 1210-3. A 0TSG pressure will stabilize after B-OTSG goes dry and stops the excessive primary to secondary heat transfer, ja If aduated, authorize defeat / enable of manual 1600# ESAS channels to throttle HPl and control RCS pressure. SE - Direct cROs to defeat / enable manual ESAS channels and throttle HPl to stop RCS repressurization. Pf RQ - Operate HPl pumps and/or valves as necessary to throttle HPl flow and control RCS pressure within allowable region of 1210-10. (CT) I
~
Page 6 of 6
I. THREE MILE ISLAND UNIT 1
- j. SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION SCENARIO NUMBER: 3 EXAMINERS:
OPERATORS: 1 ! i l' 1 1 Page 1 of 13 i
.. . . . _ _. . . . - . -_ ._ _ ___ _ . __ . _ ~ _ . . . . - - -
i .
\
! THREE MILE ISI.AND UNIT 1 SENIOR REACTOR OPERATOR l GIMULATOR EXAMINATION l 1.0 GENERAL DESCRIPTION OF SCENARIO Plant is at 100% power EOL, Equilibrium Xenon. Crew will commence a power reduction remove a Feedwater pump from service. Pressurizer level transmitter LT 1 failure with a small (2 gpm) leak to the RB causing MU-V-17 to go full open in ado. A large FW leak on 'A' OTSG occurs inside containment. This leak results in a 4 psig RB ES actuation and a manual or automatic Reactor trip. l MU-V 18 will fail to close on ESAS. l The 'A' OTSG must be isolated. MS-V-4B will fail open. This leak is isolated by closing MS-V 158. The ESAS Actuation is bypassed and HPl secured to prevent repressurization. A Plant Cooldown is commenced a 'A' OTSG using the Turbine Bypass Valves. Estimated scenario time - 60 Minutes i l l l l l l l i t Page 2 of 13 l
n_ . .= . - . -.. - . . . . - .. . -- - . - _ . - -. . _. .- THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
2.0 REFERENCES
A. 10 CFR 55.45 Operating Test (a) Content
- 8. Procedures
- 1. 1102-10 Plant Shutdown
- 2. 1202 29 Pressurizer System Failures
- 3. 1104-2 Makeup and Purification I
- 4. 1210 1 Reactor Trip
- 5. 1203-24 Steam Leak
- 6. 1210 3 Excessive Primary to Secondary Heat Transfer
- 7. 1105 03 En0i neered Safeguards Actuation System
- 8. 1210-10 Abnormal Transients Rules, Guides and Graphs
- 9. EPIP-TMI .01 Emergency Classification and Basis I C. TECHNICAL SPECIFICATIONS
- 1. Section 3.1.6, Leakage l
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- THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION i
- 3.0 SCENARIO INITIALIZATION l A. IC-17,100% power, equilibrium Xenon,640 EPFD l
B. EVENT
- 1. NINIC<10, Assign to Event #1 C. MALFUNCTIONS
- 1. RC04A PZR LT-1 failure; assion at 100% over 60 seconds to Remote #1 l 2. TH09 PZR steam space leak; assign at 0.1% severity (2 gpm) over 60 seconds to Remote #2
- 3. MUO6 MU isolation valve falls as is, MU-V-18, set at 100% severity and activate immediately.
- 4. FW09A FW line break inside containment, 'A' OTSG, set at 20% severity over 900 seconds to <
Remote #4 l l
- 5. MS07B Atmospheric Dump Valve failure, MS-V-48, set at 100% severity, and place on Event #1.
l l D. OVERRIDES NONE E. MONITOR
- 1. Set MSK2609A to 14.7; Aux. Boller pressure Page 4 of 13 l
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 4.0 SCENARIO PREVIEW A. EXAMINER PREVIEW
- 1. Plant is operating at or near 100% power, steady state condition.
- 2. Crew will begin a power reduction to remove FW-P-1 A from service.
- 3. Pressurizer level transmitter RC 1 LT-1 will fall downscale due to a cracked weld causing MU V 17 to go full open and creating a small RCS leak in the RB.
- 4. The crew will need to take manual control of MU-V-17 and swap the controlling channel to RC1 LT-3.
)
- 5. A large feed water leak (1.2E6 lbm/hr), inside containment will develop over 15 minutes.
- 6. The crew will need to take action in accordance with 1203-24, Steam Leak, to commence a reactor shutdown or manually trip the reactor due to the size of the break.
- 7. When the reactor trips the "B* OTSG Atmospheric Dump Valve will fail open, which will require manual isolation.
- 8. The crew should identify Excessive Primary to Secondary Heat Transfer and perform actions of i 1210-3 to isolate "A" OTSG. i
- 9. MU-V-18 will fail open on ESAS and will require to be isolated by closing MU V 17 and MU V 217. i
- 10. The crew will need to bypass / defeat ESAS and terminate HPl to stop RCS repressurization.
- 11. An E Plan declaration should be made.
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i THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR l SIMULATOR EXAMINATION B. SHIFT BRIEFING (SEE PLANT STATUS BOARD)
- 1. Plant is at 100% power, steady state conditions,640 EFPD.
- 2. RCS boron is 36 ppm and the BAMT is the Tech. Spec. Tank.
l 3. No maintenance or surveillance is in progress.
- 4. There are no releases in progress and none scheduled for this shift.
l 5. ' FW-P 1 A is to be rernoved from service for a bearing inspection. You are to reduce power to 60% in preparation for removing FW-P 1 A. I 1 l i l l l l l Page W ia l _
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION I 5.0 EQUENCE OF EVENTS Examiner Notes and Actions Expected Operator Actions
- 1. Initialize the simulator in accordance with Section 3.0.
- 2. Assign team positions and conduct the SHIFT 2.1 Assume assigned team positions, BRIEFING per Section 4.B.
2.2 Take tumover and review plant status. NOTE: Allow crew 3-5 minutes to take tumover and assume the watch. 2.3 Assume the watch.
- 3. ICO-MALFUNCTION: After at least a 5% power 3.1 M - Diagnose Pressurizer LT-1 failure by reduction is commenced and when directed by alarms and indications and diagnose increased examiner, activate Remote #1 and 2, to cause RB radiation levels by alarms and indications.
Pressurizer LT-1 failure and level tap leak of about 2 Opm into containment. Understand and Interpret Annunciators Diagnose Events and Conditions Understand Plant and System Response NOTE: MU V 17 will open in auto due to drop in 3.2 E,Q.BQ - Perform immediate Actions for the indicated Pressurizer level and should be LT-1 failure per 1202 29: adjusted in Hand to stabilize actual level. A. Transfer MU-V 17 to Hand and adjust MU flow. Operate ControlBoards B. Select the attemate Pressurizer level and Comply \Mth and Use Procedures temperature instruments. Demonsfrate SupeMsory AbWy 3.3 ff - Verify or direct immediate Actions of 1202 29, Section D, for the failed Pressurizer level instrument. I
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i' THREE MILE ISLAND UNIT 1
- SENIOR REACTOR OPERATOR j SIMULATOR EXAMINATION i
!. Examiner Notes and Actions Exnected Operator Actions a l NOTE: This very small leak will not cause excessive 3.4 M- Diagnose probable RCS leak into R8 I
- R8 pressure or temperature increatas. by increased Radiation levels. Implement j j ARP-C-1-1 and EP 120212 for the excessive
- Diagnose Events and Condtons Radiation levels.
i NOTE: This size leak does not warrant a rapid 3.5 88/STA/ CREW - Evaluate RCS leakage for Shutdown or trip, but the SS may elect to begin rate implications. Refer to Tech. Spec. 3.1.6 1 a Plant Shutdown, limits and required actions.
~ '
i
- Comply Hath and Use Tech. Specs.
3.6 M- Commence a one hourleak rate calculation by computer. ICO: if asked to look at RB cameras, only a very 3.7 GREW If directed by SS, commence a Plant small amount of steam is seen coming from the Shutdown. 1
~ A D-ring.
NOTE: This size leak would only warrant an unusual 3.8 31 Evaluate EALs for possible E Plan Event Per EAL US.1 of EPIP-TMI. 01. . Event Declaration.
- 4. ICO-MALFUNCTION: When directed by 4.1 Itam - Diagnose indications of a Steam / Feed examiner, activate Remote #4, to initiate a leak inside containment.
1.2E6 LM/HR RN leak on 'A' OTSG inside containment over 15 minute period. Understand and interpret Annunciators and Marms Diagnose Events and Condnons Comply With and Use Procedures 4.2 ff - Refers to the immediate actions of AP 1203-24, and instructs panel operators to monitor RB pressure and temperature. Page 8 of 13
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Expected Operator Actions 1GQi If questioned about steam leak inside 4.3 333f - Directs an operator to check RB video Containment, inform RO that there is evidence monitors for evidence / location of leak, of fog in both of the D-Rings, although, it instructs PCRO to place Reactor Building appears to be thicker inside *A* D Ring. Ememency Cooling in service. Diagnose Events and Conditions 4.4 3.QEQ - Diagnoses that leak appears to be a Understand Plant and System Response FW leak inside Containment due to increased
'A' loop FW flow, lower A OTSG level, increased ATc and no large decrease in Generated MWE.
Operate ControlBoards 4.5 P9EQ - Places Reactor Building Emergency Cooling in service. NOTE: Crew may etset to trip the reactor due to the 4.6 3331 - With RB pressure still increasing, pressure >2 psig, FW leak. directs Crew to begin a Plant Shutdown at an appropriate rate. Demonstrate Supervisory Ability Critical Task Operate ControlBoards
)
l NOTE: As FW Leak increases, RB pressure will 4.7 SF/PCRO/SCRO - Begin a Plant Shutdown at increase to >2 psig resulting in an ES Actuation, the requested rate per OP 110210. and an Auto Reactor Tdp if not manually idpped.
- 5. RB pressure increases to 4 psig ES Actuation 5.1 SSISF - Direct that a Plant announcement be setpoint. made, and Reactor manually tripped.
Demonstrate Supervisory Ability Comply %1th and Use Procedures Page 9 of 13
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examirter Notes and Actions Expected Operator Actions NOTE: When Reador trips, the 'B' Atmospheric Dump 5.2 Pf,8Q - Manually trips Reactor as directed by Valve will fail open and require Auxiliary the SS/SF. Operator action to control / isolate. Verify that MALFUNCTION M807A activates on Event #1, !. 5.3 3,3/3E- Direct ti.9 use of Global Alarm Silence. I ICO: As an AO, MS-V-4A may be throttled 5.4 IE&M - Respond to the Reactor Trip per ATP locally by using Remote Functions MSR25 and 1210 1. MSR26 or isolated with Remote Function MSR14. ( s Comply With and Use Procedures Operate ControlBoards 5.5 32RQ - Verifies HSPS EFW actuation on high j' RB pressure and dispatches an AO to EFW l Pumps. Diagnose Ewnts and Conditions 5.6 gg-Diagnose Excessive Primary to Understand Plant and System Response Secondary Heat Transfer condition existing in A OTSG, and ditgqi SCRO to isolate the A OTSG. i NOTE: EFW will actuate on HSPS high RB pressure 5.7 3 SEQ-Isolates A OTSG: and possibly low OTSG level. (1) Closes FW V 5A,92A,16A and 17A (2) Closes MS V 3D.3E,3F,1B,1 A and 4A. Comply With and Use Procedures (3) Closes EF V 30A,300 and MS V 2A.
- Operate ControlBoards critical Task i
4 h- Page 10 of 13
l THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Exnected Operator Actions ICO: MS V 48 may be isolated by closing MS- 5.8 IE8M - Recognizes MS-V-48 is still open and V 158. If requested set Remote Function directs the EFW AO to isolate it by dosing M8R14 to dose. MS-V-15. Critical Task Communicate andIntered %1th Crew ICO: If crew requests to throttle MS-V-48, 5.9 IE8M - Complete immediate manual actions ensure that Remote Function MSR14 is open, of ATP 1210-1 and then set Remote Function MSR27 and MSR28 to the requested value. 5.10 PfaQ-Infnrms SS/SF of MU-V-18 failing open and closes g verifies close MU-V 17 and MU-V 217. Olagnose Events and Conditions 5.11 SS/SF/STA - Recognize that Excessive Understand Plant and System Response Primary to Secondary Heat Transfer Event is Operate ControlBoards terminated and direct PCRO to bypass / defeat Demonstrate Supervisory Ability ESAS and terminate HPl to stop the RCS repressurization. Operate ControlBoards 5.12 Pf RQ - Bypasses / defeats both Trains of ESAS, and then terminates HPl per OP 1105 3: (1) Secure MU P-1C (2) Close MU-V-16C and MU-V 16D (3) Throttle close MU-V-16A/16B (4) Open MU V-36/37 when required (5) Close MU V-14A and MU V 148 as required. Critical Task Demonstrate Supervisory Ability 5.13 jf-Verify / direct crew performance of ATPs 12101/1210-3 immediate manual actions. Page 11 of 13
l THREE MILE ISI.AND UNIT 1 i SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION Examiner Notes and Actions Expected Operator Actions NOTE: Declaration will probably be made at end of 5.14 3.R.- Dedare an Emergency Plan Event within scenario. 15 minutes of verirmatiori of EAL. Comply With and Use Procedures 5.15 IE&M - Pe:Torm the req &ed follow-up actions of ATPS 1210-3 and 1210-1. Ope /afe CortoISoards 5.16 SS/8F/STA- Provide PCRO/SCRO with adoquate data to control"B" OTSG level / pressure to stabilize RCS parameters and initiate a Cooldown, t i Page 12 of 13
_ . _ . _ _ _ . _ _ - - . ~ . - -- - -~ THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
- 6. TERMINATION POINT When all of the following conditions exist:
- 1. Crew has had sufficient opportunity to perform all applicable Critical Tasks.
- 2. ATP 1210 3 follow-up mulons are in progress.
. 3.' HPl is terminated.
- 4. Plant is stabilized or being cooled down with the A OTSG isoleted.
- 5. All examiners agree on termination.
- 7. EMERGENCY PLAN DECLARATION (Adm:n Section A.4)
Alert - EAL A2.1 (Containment Integrity) 4 I Page 13 of 13
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1 Sc:n:ris Outline Facility: TMl Scena.io No.: 3 Op-Test No.: 3 f Examiners: Operators: 4 l Ot&in: Plant is operating at or near 100% power, steady state condition. Pressurizer level transmitter RC-1-LT-1 will fail downscale due to a cracked weld causing MU-17 to go full open and creating a small RCS leak in the RB. The crew will need to take manual control of MU-V-17 and swap l the controlling channel to RC1 LT-3. A large feed water leak (1.2E6 lbm/hr), inside containment will ' < develop over 15 minutes. The crew win need to take action in accordance with 1203 24, Steam Leak, D commence a reactor shutdown or manually trip the reactor due to the size of the break. When the rt.ccior trips the *B' OTSG Atmospheric Dump Valve will fail open, which will require manual isolation. The crew should identify Excessive Primary to Secondary Heat Transfer and perfonn actions of 1210 3 to IF. late 'A' OTSG. MU-V-18 will fail open on ESAS and will require to be isolated by closing MU-V-l 17 and MU V 217. The crew will need to bypass / defeat ESAS and terminate HPl to stop RCS j repressurization. An E-Plan declaration should be made. i j initial Conditions: IC-17,100% power, equilibrium Xenon,640 i Turnover: Plant is at 100% power, steady state conditions,640 EFPD. RCS boron is 36 ppm and the i BAMT is the Tech. Spec. Tank. No maintenance or surveillance is in progress. There are no releases in progress and none scheduled for this shift. FW-P-1 A is to be removed from service for a bearing i inspection. You are to reduce power to 60% in preparation for removing FW-P-1 A. 1 l Event Malf. No. Event Type
- Event
; No. Description (N) Plant power reduction i 1. RC04A (I), (C) Pressurizer LT-1 failure, Pressurizer steam space leak TH09 j 2. FWO98 (M) (C) FWline break inside containment j
- 3. MS0753 (C), (1) Atmospheric Dump Valve failure, MS V-48, MU isolation valve a s as is, E-W8 l MUO6 l
l * (N)ormal, (R)eactivity, (1)nstrument, (C)omponent, (M)ajor t Page i of 4
Operator Actions Op Test No.: 3 Scenario No.: 3 Event No.: 1 Page 2.,, of 4_,, Event
Description:
Pressurizer level transmitter failure, Pressurizer steam space leak Time Position Applicant's Actions or Behavior
.QBEW Diagnose Pressurizer LT-1 failure by alarms and indications and diagnose increased RB radiation levels by alarms and indications.
4 g Perform immediate Actions for the LT-1 failure per 1202 29:
- Transfer MU-V 17 to Hand and adjust MU flow.
Select the altemate Pressurizer level and temperature instruments. MU-V-17 will open in auto due to drop in indicated Pressurizer level and should be adjusted in Hand to stabilize actuallevel, a Bf Verify or direct immediate Actions of 1202 29, Section D, for the failed Pressurizerlevelinstrument. CREW Diagnose probable RCS leak into RB by increased Radiation levels. implement ARP-C-1 1 and EP 1202-12 for the excessive Radiation levels. SS/STA/ Evaluate RCS leakage for rate implications. Refer to Tech. Spec. 3.1.6 j lim ts and required actions. REO 4 - This size leak does not warrant a rapid Shutdown or trip, but the SS may elect to begin a P! ant Shutdown. j CREW Commence a one hour leak rate calculation by computer.
.Q. HEW If directed by SS, commence a Plant Shutdown.
This size leak would only warrant an Unusual Event Per EAL US.1 of EPIP TMI .01. _S_1 Evaluate EALs for possible E Plan Event Declaration. Page 2 of 4
Operat:r Actions Op-Test No.: 3 Scenario No.: 3 Event No.: _2 Page 3_ of 4 Event
Description:
FW line break inside containment Time Position Applicant's Actions or Behavior 13301 Diagnose indications of a Steam / Feed leak inside containment. Af Refers to the immediate actions of AP 1203-24, and instructs panel operators to monitor RB pressure and temperature. 333E Directs an operator to check RB video monitors for evidence / location of leak. Instructs PCRO to place Reactor Building Emergency Cooling in service. AgBQ Diagnoses that leak appears to be a FW leak inside Containment due to increased 'A' loop FW flow, lower A OTSG level, increased ATc and no large decrease in Generated MWE. PCRO Places Reactor Building Emergency Cooling in service. 3,1@f With RB pressure still increasing, directs Crew to begin a Plant Shutdown at an appropriate rate. Critical Task Crew may elect to trip the reactor due to the pressure >2 psig FW leak. SF/PCROI Begin a Plant Shutdown at the requested rate per OP 1102-10. 3SRQ - As FW Leak increases, RB pressure will increase to >2 psig resulting in an ES Actuation, and an Auto Reactor Trip if not manually tripped SSISF RB pressure increases to 4 ps'g ES Aduation setpoint. Direct that a Plant announcement be made, and Reactor manually tripped, l PCRO Manually trips Reactor as directed by the SS/SF. TEAM Respond to the Reactor Trip per ATP 1210-1. 32R.Q Verifies HSPS EFW aduation on high RB pressure and dispatches an AO to EFW Pumps. Page 3 of 4
Operator Actions Op Test No.: 3 Scenario No.: 3 Event No.: 3 Page 4 of 4 Event
Description:
Atmospheric Dump Valve faHure, MS V 48, MU isolation valve fails as is, MU V 18 Time Position Applicant's Actions or Behavior f g Diagnose Excessiva Primary to Secondary Heat Transfer condition
- existing in A OTSG, and dugg SCRO to isolate the A OTSG.
l 3.GBQ lsolates A OTSG: i (1) Closes FW-V 5A,g2A,16A and 17A 3 (2) Closas MS V-3D.3E,3F,1B,1A and 4A.
- (3) Closes EF-V-30A,300 and MS-V-2A.
CriticalTask l EFW win adaste on HSPS high RB pressure and possibly low OTSG level. 4 TEAM Recognizes MS-V-4B is still open and direds the EFW AO to isolate it by j closing MS-V-15. Critical Task ) HAM Complete immediate manual adlons of ATP 1210-1 $ SS/SF/STA Recognizc that Excessive Primary to Secondary Heat Transfer Ever:t is , terminated and dired PCRO to bypass / defeat ESAS and terminate HPl to . stop the RCS repressurization. g Bypasses / defeats both Trains of ESAS, and then terminates HPI per OP
; 1105-3:
(1) Secure MU-P-1C (2) Close MU-V-16C and MU-V-160 (3) Throttle dose MU-V-16A/168 (4) Open MU-V 36/37 when required (5) Close MU-V 14A and MU-V-14B as required. Critical Task g Venfy/ direct crew performance of ATPs 1210-1/1210-3 immediate msnual actions. El Dedare an Emergency Plan Evr.:nt within 15 mirmtes of verification of EAL IgAM Perform the required follow-up actions of ATPs 1210-3 and 1210-1. SS/SF/STA Provide PCRO/SCRO with adequate data to cwtrol *B' OTSG level / pressure to stabilize RCS parameters and initiate a Cooldown. Page 4 of 4
.t THREE MILE ISLAND UNIT 1 t SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
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x .i p .2 ,.;a a . # , 3 qu g,9 g. g ;, r f.4 g g g g p;rt q : g i s: p g ; ; , g g .7.e SCNNARIO NUMBER: 5 44 a :a,,
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I EXAMINERS: OPERATORS: i l I i r l t 'k 1 4 Page 1 of 11
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1.0 GENERAL DESCRIPTION OF SCENARIO 1 Plant at 90% power, steady stata, EOL, with EG-Y-1B out of service.
~
A power escalation to 100% is stated. A loss of EHC-P-1 A with failure of EHC-P-1B to automatically start resulting in decreasing EHC pressure until EHC-P-1B is started manually. A SASS mismatch occurs for FW temperature requiring the alternate transmitter to be selected. A fault develops on 230KV #6, causing loss of Bus #6, undervoltag
- on the 1E 4KV bus, and loss of MU-P 18.
An excitation fault occurs resulting in a main generator / turbine ark.I reador trip. Shortly after the reactor trip, a grid fault causes a loss of 230KV Bus #4, resulting in loss of offsite power. ED-Y-1 A trips when it starts resulting in a station blackout. Emergency Feedwater HSPS signal failures require manual initiation of EFW flow to the OTSGs in order to maintain primary to secondary heat removal. The SBO diesel is always available for use. One 230KV bus and normal AC power may be restored. (As time permits, optional) Estimated scenario time - 60 minutes l Page 2 or 11
1 l t f THREE MILE ISLAND UNIT 1 , SENIOR REACTOR OPERATOR j SIMULATOR EXAMINATION
- j. ,
1 l
2.0 REFERENCES
,1
. A. 10 CFR 55.46 (a) Content . v nn ~
m . B. PROCEDURES' # #* ' ~""
- 1. 12101 '> ReactorTrip i
l 2. 2 1202-2 Loss of Station Power J ? -
- 3. 1210-10 Abnormal Transients Rules, Guides and Graphs
- 4. EPlP-TMI .01 Emergency Classification and Bases C. TECHNICAL SPECIFICATIONS i
l 1. Section 3.7.2 Electrical Power l l \ l l l l l l 1 4
)
j f ] Page 3 of 11 l-
). 4 THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
- 3.0 SCENARIO INITIAllZATION i
. A. IC-17 100% power, steady state, equilitulum Xenon,640 EFPO l
Following initialization perform the followin0* I l 1. Reduce powerto 90% i 2. Red tag EG-Y-1B out of service by'the following
- 3. Transfer EG-Y-1B Starting Switch to manual
- 4. Place red Sticker tags on START push button on CF.
- 5. Red tag G1102 breaker control in pull-to-lock and rack out breaker using Remote Function j EGR01
- 6. Trip fuel rocks using EGR29.
- 7. Close air start isolation valve EG-V-158 using EGR31. ?je r p ., j e B. MALFUNCTIONS
, .. ',f _ 1
'j u !f*
- 1. TC10A Fault trip of EHC-P-1 A; assign to Remote #1
- 2. FWO48 FW temperature failure, SP 5-TE2, set at 50% severity, over 60 seconds; assign to Remote #2
- 3. ED188 230KV Bus #6 Fault; assign to Remota #3.
- 4. EG03 Main Generator Excitation Failure; assign to Remote #4
- 5. ED18A 230KV Bus #4 Fault; assign to Remote #8
- 6. EG07A EG-Y-1 A trip; assi0n to Event #1 C. REMOTE FUNCTIONS:
- 1. ICR02 EFW level setpoint (A-SG); change to 0% immediately
- 2. ICR04 EFW level setpoint (B-SG); cannge to 0% immediately D. EVENT TRIGGERS
- 1. Event #1 EGNDGRPM(1) >850 E. 10 OVERRIDE
- 1. 05A7S10 ZDIEHCP1B(3) NAP OFF Prevent auto start of EHC-P-1B activate immediately.
F. MONITOR
- 1. MSK2609A; Set to 14.7 Page 4 of 11
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THREE MILE ISt.AND UNIT 1 SENIOR REACTOR OPERATOR
- SIMUt.ATOR EXAMINATION 4.0 SCENARIO PREVIEW A. EXAMINER PREVIEW Iu ghesu, .
f; , , . l
- 1. Plant at 90% power, steady sisle, with EG-Y-15 out of service for an oil change.
- ;A x I 1
2
- 2. ; A power ancalation to 100% is started.
'1 g i'
- 3. EHC-P-1 A trips. EHC-P-1B fails to automatically start requiring the team to promptly start the standby pump to prevent tuttiine trip. (CT)
- 4. After the pierd has been stabilized, the soledad FW temperature transmitter will fail to 300*F over a period of time resulting in a SASS mismatch. (CT) a
- 5. A fault develops of 230KV Bus #6, causing a los,s of Bus #8, undervoltage on the 1E 4KV bus, and loss of MU P-1B.
- 6. After the team responds to the loss of Bus #6, en excitation fault on the generator occurs resulting in a trip of the generator, the main turbine and the reactor.
- 7. Shortly after the reactor trip, a grid fault causes a loss of 230KV Bus #4, resulting in a loss of offsite power.
- 6. EG-Y-1 A trips when it starts resulting in a station blackout. j
- 9. Emergency Feedwater HPSP signal failures require manual initiation of EFW flow to the OTSGs in order to maintain primary to secondary heat removal. (CT) 1
- 10. The SBO diesel is always available for use.
- 11. One 230KV bus and normal AC power may be restored as time permits.
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THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR j SIMULATOR EXAMINATION j B. SHIFT BRIEFING (SEE Pt. ANT STATUS BOARD)
- 1. The piant is at 90% power, steady state,640 EFPD.
, 2. RCS boron is 36 ppm and the BAMT is the Ted. Spec. tank. m+ ,
- 3. EG-Y 1B is out of servios for oil change. R has been out of servios for4 hours and is expected j to be retumed to service within 2 hours.
- 3
- 4. No other maintenance or surveillance is in progrees. .,
- 5. . There are no releases in progrees and none scheduled for this shifL ...
4 g ,/ '. , ,;y .wp.' , , T gk
- 6. Turbine valve testing has been completed. Retum power to 100%.
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- THREE MILE ISLAND UNIT 1 i
SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 5.0 SEQUENCE OF EVENTS
- 4 .
Examiner Notes and Actions -' Exoscted Operator Actierry 1
- 1. Initialize the simulator in acconianos wth .fA 8ection 3.0. O E ' MMb ' '
l j- 1
- 2. Assign team positions and conduct the SHIFT ' 2.1 Assume assigned team positions.
i BRIEFING per Section 4.8.
- a. w y,.
<,, z a, . v#
4 2.2 Take tumover and review plant status, i NOTE: Allow crew 3-5 minutes to take tumover and assume the watch. 2.3 Assume the watch. t i 3. ICO-MALFUNCTION: After the power reduction 3.1 Ig&M- Announce trip of EHC-P-1 A from and when directed by Examiner, activate Remote #1 to cause a trip of EHC-P 1 A. com#er alann. f Understand and katerpret Annunciators and Nerms Operate ControlBoants ** M ~ ***** * *#*A '*""* # j EHC-P-1B and perform manual start to preverd j- a turbine trip on low EHC pressure. ICO ROLEPLAY: If dispatciwd to EHC-P-1A, report motor very hot and a smell of bumt insulation.
- 4. ICO MALFUNCTION: After response to 4.1 3.9.8Q- Diagnose a FW temperature failure.
EHC-P-1 A, when directed by Examinor, activate Remote #2 to cause indicated FW 4.2 3QBQ-Place ICS FW Demands to Hand to temperature to decrease and a SASS correct FW flow. mismatch. 4.3 3E-Verify / direct SCRO to select attemate FW Dagnose Events and Cond6cos temperature instrument. Understand Plient and System Response Operate ControIBoards 4.4 3GEQ- Retum ICS to automatic. Page 7 or 11
. ._ t
! ' THREE MILE ISLAND UNIT 1 l SENIOR REACTOR OPERATOR s SIMULATOR EXAMINATION - Examiner Notes and Actions Exnected Operator Actions I
- 5. ICO-MALFUNCTION: After the plant has been 5.1 TEAM - Diagnose loss of 230KV Bus 8 and 1E stabilized or when directed by the Examiner, 4160V causing a loss of MU-P-18.
f ' activate Remoto'#3 to cause a 230KV Bus'8 7; s
<MW n ,
j fault and loss of 1E 4160V bus. 5.2 PCRO- Respond to the loss of RC makeup 3 +r > n m ams _g + y, e syvwand seal injection per 1203-15 as follows:
- riagnose Events and Conditions 1.' Close MU-V-32 l Comply WGfh and Use Procedwas 2. Start MU-P-1 A, DR-P-1A and DC P-1A.
Operate ControlBoards 3. Slowly reestablish seal!njedion. Demonstrate Supervisory Ability 5.3 Ef - Verify / direct 1203-15 manual actions. NOTE: 1202-38,1203-19 and 1203 30 should also be addressed. I Comply With and Use Tech. Specs. 5.4 3.3 FEE- Consult Tech. Specs. And determine that with an Aux Transformer and EG-Y-1B out of service, the plant must be in Hot Shutdown within 12 hours. (T.S. 3.7.2.d)
- 6. ICO-MALFUNCTION: After Tech. Specs. have 6.1 IE6M - Diagnose and respond to the reactor '
been addressed and when direded by the trip in accordance with ATP 1210-1 Examiner, activate Remote #4 to cause a generator / turbine and reactor trip 6.2 TEAM- Perform ATP-1210-1 Immediate actions except that a second make pump Comply With and Use Procedures (MU-P-1C) will not have power for start, unless Operate Contro/ Boards the SBO diesel was already used to power 1E 4160V Bus Understand Plant and System Response 6.3 IEAM - Perform necessary actions to begin to stabilize the plant on natural circulation. Page 6 of 11
THREE MILE ISLAWD UNIT 1 i SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION l
- 7. ICO-MALFUNCTION: Following completion of ' 7.1 M - Diagnose the loss of station power and 1
1210-1 and when direded by the Examiner, respond in accordance with EP-1202 2. '
- activate Remote #5, to cause a loss of stahon Power. n s,. ht tv ' 4W( ~ , w ,u; ~ ~"
i '
- r. , .
NOTE: EG Y-1 A will trip upon start, but if correct - e 7.2 Ig&M-Diagnose trip of.EG Y-1A. actions are taken per Eri 602-2, the diesel may be recovered. l 7.3 Ef,BQ - Attempt to start EG-Y 1 A per 1202 2. 4 ICO/ ROLE PLAY / MALFUNCTION: As AO dispatched to EG-Y-1 A, report overspeed trip alarm present and acknowledge alarm. If/when i proper actions are directed by crew to recover EG-Y-1A, clear malfunction EG07A to allow i EG Y-1 A reset and operation. f Operate ControlBoards 4 NOTE: HSPS setpoints for OTSG level control on loss 7.4 $_QEQ - Verify EFW actuation loss of all RC of RC pumps will fait at 0% requiring manual pumps with at least EF P 1 running. but note no initiation of EFW f'c ? EFW to OTSGs. Diagnose Events and Conditions } NOTE: Sufficient EFW flows must be maintained until 7.5 P_QRQ - Operate EF-V-30A/B/C/D as necessary OTSG level setpoints (50%) are reached, after to establish EFW flow to both OTSGs, and which OTSG pressure must be controlled to increase levels towan$s 50% to establish RCS maintain heat sinks for natural circulation. natural circe'ition per AT 121010. (CT) NOTE: Examiner may ask SF after the scenario. 7.6 31-Identify / declare the appropriate event (EAL) and implement the Emergency Plan. NOTE: Th arvi incore T/C's may increase until core 7.6 SF/SCRO - Verify RCS natural circulation core Delta-T reaches its maximum point as natural cooling in accordance with ATP 1210-10 and circulation flow is established, but should then EP 1202-2. be stable or decreasing Page 9 or 11
. .- - . . . . . ._~
, e l
4 THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION 1 Examiner Notes and Actions Excected Ocorator Actions I i l
- NOTE
- EFW should be throttled by EF V-30A45/C/D, to 7.7 3219-Throttle EFW flows to OTSGs in prevent stable W $ from anoseding
~ ~ ~
s accontance with ATP 121010'as necessary to ! 50*F/hr on RCS naturel M, but OTSG ~
' prevent or stop RCS excesolve cooling, which I
levels must be increased toward 50% at a rate would result in RCS depressurization, loss of which will not cause RCS excessive cooling and ' pressurizer lov'el and RCOM margin. depressurization. This is most critical if neither (CT) 4160V ES bus is re-energized for RCS makeup capableity to prevent loss of RCS subcooled margin. Underatand P\\snt and System Response Operate ControlBoards 7.8 P.GB2 - Operate (close) valves required by EP 1202 2 to conserve RCS inventovy ard pressurizerlevel. NOTE: Retum of offsite power is not required. 7.9 31 - Contact dispatcher to investigate retum of off-site power. l l Page 10 or 11
THREE MILE ISLAND UNIT 1 SENIOR REACTOR OPERATOR SIMULATOR EXAMINATION
, ,a 9,... . I
- 6. TERMINATION POINT ~
, When all of the following conditions exist:
- 1. Crew has had sufficient opportunity to perform all applicable critical tasks.
...(
- 2. '
1202-2 foNow up actions are in pm0mes.
- 3. EG Y-1A or the 880 dieselis in .
- 4. .The plant is stable on natural circulation or a slow cooldown is in progress.
,, r: ~r
- 5. All examiners agree on termination.
- 7. EMERGENCY PLAN DECLARATION (Admin Section A.4)
If the D and E Buses are de-energized < 15 minutes: Alert - EAL A6.1 (AC) If the D and E Buses are de-energized > 15 minutes: SNe - EAL S6.1 (AC) \ , l l l Pe0811 of 11
Scenari3 Outtine Facility: TMl Scenario No.: 5 Op Test No.: 5 I ' Examiners: Operators: Ot$ectives: Plant at 90% power, steady state, with EG-Y-1 A cut of service for an oil change.l A' power increase to 100% is started. EHC-P-1 A trips. EHC-P-1B fails to automatically start requiring the team to prompuy staat the standby pump to prevent turbine trip. (CT) ARet the plant has been stabilized, the selected FW temperature transmater wHl fall to 300*F over a period of time resulting in a SASS mismatch. (CT) A fault develops of 230KV Bus #8, causing a loss of Bus #8, undervoltage on the 1E 4KV bus, and loss of MU-P-18. After the team responds to the loss of Bus #8, an excitation fault on the generator occurs resulting in a trip of the generator, the main turbine and the reactor. Shortly after the reactor trip, a grid faut causes a loss of 230KV bus M, resulting in a loss of offsite power. ED Y-1 A trips when it starts resulting in a station blackout. Emergency Feedwater HPSP signal failures require manual initiation of EFW flow to the OTSGs in order to maintain primary to secondary heat removal. (CT) The SBO diesel is always available for use. One 230KV bus and normal AC power may be restored as time permits. Initial Conditions: IC 17,90% power, steady state, equilibrium Xenon,640 EFPD. Following initialization perform the following: Red tag EG-Y-1B out of service. Transfer EG-Y 1B Start Switch to manual. Place red Stickertags on START push button on CR. Red tag G11-02 breaker controlin pull-to-lock and rack out breaker using Remote Function EGR02. Trip fuel racks using EGR29. Close air start isolation valve EG-V 1SB using EGR31. Turnover: The plant is at 90% power, steady state,640 EFPD RCS boron is 36 ppm and the BAMT is the Tech. Spec. tank. EG-Y-1B is out of service for oil change. It has been out of service for 4 hours and is expeded to be retumed to service within 2 hours. No other maintenance or surveillance is in progress. There are no releases in progress and none scheduled for this shift. Continue 100% power operation. Retum to 100% power. Event Malf. No. Event Type' Event No. Description (N) Plant power increase.
- 1. lTC10A (C) l EHC-P 1 A trip
- 2. FWO4B (I) Feedwater temperature failure, SP-5-TE2
- 3. ED18B (C) 230 KV Bus #8 Fauk
- 4. EG03 (M)(C) Main Generator Excitat'on failure
- 5. ED18A (M)(C) 230 KV Bus #4 Fault EG07A EG-Y-1 A trip (N)ormal, (R)eadivity, (1)nstrument, (C)omponent, (M)ajor i
Page 1 of 7
_ _ . . _ . - , _ __ _ _ _ . _ . . _- - . _ _ _ . _ _ . __m._ _ __ l l Operator Actions l Op Test No.: 5 Scenario No.: 5 Event No.: Page J., ofl 1 l Event
Description:
EHC-P-1 A trip Time Position Applicant's Adions or Behavior H&M Announce trip of EHC-P 1 A from computer alarm. RCRO' ' Recognias auto start fature of EHC#-15 and W manual start to prevent a tuttilne trip on low EHC pmesure, ra < p 43 . c. ;3R m 4 cT e % a * . 1 i l 1 l l 1 Page 2 of 7 1 l 1
l
.i )
1 Operator Actions i Op Test No.: 5 Scenario No.: 5 Everd No.: ._2_ Page _;L, of 7 Event
Description:
Feedwater temperature failure, SP 5-17.2 Time Position Applicant's Adions or Behavior AGRQ Diagnose a FWtemperature fature. EGBQ Place IC8 FW Demands to Hand to coned FW flow. ' RE verwyAsred SCRO to select altamate FW temperature instrument. AGBQ Retum ICS to automatic.
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. . = _ . - . -. .- - --- - ~ . . - - . _ _ . - . - . - - . - .-- -
Operator Actions 2 Op Test No.: 5 Scenasio No.: 5 Event No.: _) Page 1,,of 7 Event
Description:
230 KV Bus #8 Fault Time Position Applicant's Actions or Behavior
.3&M Diagnose loss of 230KV Bus 8 and 1E 4160V causing a loss of MU-P-18.
EGBQ v Respond to the loss of RC makeup and seal istection por 120k15 as 3 3 ,
+
ser -
- 1. Close MU-V-32
, 2. Stast MU-P-1 A, DR-P-1 A and DC P-1 A.
- 3. Slowly tw=Nieh sealinjection RE vertfy/ direct 120315 manual adions.
- 1202-38,1203-19 and 1203 30 should also be addressed, i
133E Consult Todi. Specs. And determine that with an Aux. Transformer and i EG-Y-1B out of service, the piard must be in Hot Shutdown within 12 hours, i i i l I i I l PaSe 4 of 7
Operator Actions i Op Test No.: 5 Scenado No.: 5 Event No.: (_ Page _f , of 7 Event
Description:
Mein Generator Exdtation fo8ure Time Position Applicant's Actions or Behavior UAM Diagnose and respond to the roedor trip in accon$ence with ATP 12101 H8N Perform ATP-12101 immodate adions except that a second make pump (MU-P-1C) will not have power for start, unless the 880 diesel was r s already used to power 1E 4160V Bus :i G vi NAM Perform necessary adlons to begin to stabilize the plant on natural ' circulabon. L i 4 h Page 5 of 7
a } - Operator Actions , l Op Test No.: 5 Scenario No.: 5' Event No.: 5 Page 6 of1 1 ! Event
Description:
230 KV Bus M Fault, EG Y 1 A trip
- Time Posstion Applicant's Adions or Behavior g Diagnose the loss of station power and respond in accordance with EP-
+ , x, ,weg# 1202 4;g h % s g a w ma m.a3 . u
\
g Otagnose trip of EG.Y-1A.' -m *" g 3- Attempt to start EG Y-1A per 1202-2. . e. t EG-Y-1 A wit trip upon start, but if correct actions are taken per EP 1202 2, the diesel may be recovered. - 1 l g Vertry EFW actuallon loss of aN RC pumps with at least EF-P-1 running, but rde no EFWto OTSGs.
- HSPS setpoints for OTSG level control on loss of RC pumps wig feu at 1 0% requiring manualinitiation of EFW flow.
EG&Q Operate EF V-30A/B/C/D as necessary to establish EFW flow to both OTSGs, and increase levels towards 50% to establish RCS natural circulation per AT 1210-10. (CT) l 1
- Sufficient EFW flows must be maintained until OTSG level setpoints
]
(50%) are reached, after which OTSG pressure must be controlled to maintain heat sinks for natural circulation. g identify /dedare the appropriate event (EAL) and implement the Emergency Plan. SF/SCRO Verify RCS natural circulation core cooling in accordance with ATP 1210-10 and EP 1202-2.
- Th and incore T/C's may increase until core Delta-T reaches its maximum point as natural circulation flow is established, but should
)
then be stable or decreasing l ELEQ Throttle EFW flows to OTSGs in accordance with ATP 121010 as necessary to prevent or stop RCS excessive cooling, which would result in RCS depressurization, loss of pressurizer level and RCS subcooled margin. (CT)
- EFW should be throttled by EF-V 30A/B/C/D, to prevent stable cooidown rate from exceeding 50*F/hr on RCS natural circulation, but OTSG levels must be increased toward 50% at a rate which will not cause RCS excessive cooling and depressurization. This is most critical if neither 4160V ES bus is re-energized for RCS makeup capability to prevent loss of RCS subcooled margin Page 6 of 7
. ;?p!M;?sy~ 1 g Operate (close) vahms required tiy EP 1202-2 to conserve RCS inventory and pressurtzerlevel.
H Contad dispatcher to investigate retum of off-site power. 5 5~
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r e THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERFORMANCE HEASURE
'<^'? <
, Q. 7 TITLE:
Operate the Station Blackout' Diesel Generator TASK NUMBER: 0648000101 TIME: 13 Minutes EXAMINEE
REFERENCE:
EP 1202-2, Loss of Station Power EVALUATION METHOD:- PERFORM: X SIMULATE: EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 064 A4.06 IMPORTANCE: 3.9 10CFR55.45: (a) (3), (6), (8) COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 13 FileV448000101. doc
Task 0648000101 SIMULATOR CONDITIONS: Initialize the Trainer at IC-21 Set Remote T;;nction EGR24 to 25%. Insert the following Malfunctions to control EFW: FW19A - EF-V-30A, Severity 19% N19B - EF-V 308, Severity 19% N19C - EF-V-30C, Severity 0% FW19D - EF-V-30D, Severity 0% Activate Malfunction ED01, Electrical Blackout. Start MU-P-1A, DC-P-1A and DR-P-1A. Initiate Global silence. Freeze the Simulator. EXAMINER PREVIEW: This JPM deals with responding to a loss of offsite power and loading the SBO Diesel on the 1C 4160v Bus. The plant will be at a post trip condition and a loss of offsite power will occur. The Examinee will be instructed to load the SBO Diesel onto the 1C 4160v Bus, and restore components on the "J" Bus. The Examiner will initiate Global silence, after which the ICO will maintain silence from the instructors' booth. EFW will be throttled by the use of FW Malfunctions to prevent ESAS Initiation. The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the informa?. ion contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Go to run on the simulator.
- 5. Begin the cueing sequence as described in the EXAMINER CUES ser. tion.
(Continued) Page 2 of 13 FueW640000101. doc J
. _ . . -- . -. .~ _ . - _ - . - . .
A Task 0648000101 (Continued) EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the
- EXAMINER CUES")
may be~taken by the' Examine'r in trying to extract , the requisite information from the' Examinee, however, care'aust be taken to ensure'that you do not COACH the Examinee. 4 To' complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. . TASK STANDARDS: The SBO Diesel Generator has energized the 1C 4160v Bus, and GN-P-1/3/4, IA-P-4, SC-P-1B, FW-Y-5 1B and a Secondary River Water pump are running. I Page 3 of 13 Re@640000101. doc
Task 0648000101 EXAMINEE PREVIEW: For this event you are assigned the duties of the 3rd CRO. The instructor / examiner will act as the Console CRO's and the SF. The auxiliary instructor is available to act as Auxiliary l Operators on~the radio or'page system. When you are told to begin, you will respond to the cues and/or indications which the Examiner ; will provide to you either verbally or via the l
- simulator.
You are to respond as you would in the plant for a real condition. 1 TASK CONDITIONS: The plant was experiencing problems with the
- grid, i
The Turbine tripped due to the power load unbalance relay actuating. The Turbine trip resulted in a Reactor trip. . 4 ATP 1210-1 has been completed. 2 A loss of offsite power occurred 5 minutes ago. I It has been 30 minutes since the Reactor tripped. l Page 4 of 13 Fdee648000101. doc
Task 0648000101 EXAMINER CUES EXAMINEE ACTIONS STANDARD Go to Run on the Simulator.
' (( ,, . . ,, -
~
NOTE:" ' . ! EIN will be 'aut6" l
- throttled.- ,,,
ICO: 1 After Global Silence is
- initiated silence alarms cnd acknowledge alarms using-the instructor station.- Continue to ccknowledge alarms.
Instruct the Examinee to *1. Examinee places the 1. The following energize the 1C 4160v following components on Bus using the SBO Diesel components in the Console Right in and to restore Pull-to-Lock Pull-to-Lock, components on the "J" position by 480v Bus in accordance rotating the NOTE: with EP 1202-2. Extension control Only 1J-02, 1K-02, l l counter-clockwise IL-02 and 1M-02 and pulling are Critical. upwards. l 1SA-C2 1SA-C2 l ISB-C2 ISB-C2 IJ -02 IJ -02 SAT l 1K -02 1K -02 SAT , IL -02 IL -02 SAT 1M -02 1M -02 SAT CW-P-lC CW-P-1C CW-P-1F CW-P-1F CO-P-1C CO-P-1C CO-P-2C CO-P-2C HD-P-1C HD-P-1C Page 5 of 13 Fileml01. doc
Te1 0648000101 EXAMINER CUES EXAMINEE ACTIONS STANDARD
- 2. Examinee' verifies or 2. Fire pump running starts a diesel as indicated by driven fire pump on PI-371 > 150#.
PL. 1-NOTE: Voltage setpoint *3. Examinee starts the 3. SBO Diesel started is too low. Operator SBO Diesel by as indicated by l will have to have the depressing START Running and Ready cetpoint adjusted pushbutton on CR To Load indicators locally or use the and verifies ready lit on CR. manual voltage control. to load light. SAT l 3 4. Examinee verifies 4. T1-C2 on PR is f T1-C2 is Open. Open.
*5. Examinee Closes 5. G2-12 Closed on CR G2-12 by rotating as indicated by the extension red light control in the illuminated.
clockwise direction on CR and verifies that the 1Y SAT Substation is energized.
*6. Examinee Closes 6. 1C 4160v Bus T1-C2 by rotating energized as the extension indicated by control in the overhead alarms clockwise direction clearing.
on PR and verifies the 1C 4160v Bus SAT l energizes. 1 ROLE-PLAY: If requested, inform 7. Examinee requests AO 7. AH-E-197A and Examinee that AH-E-197A to verify AH-E-197A AH-E-198 verified l and AH-E-198 are and AH-E-198 running running ! running. I Page 6 of 13 FileM48000101. doc i
Task 0648000101 r .s-EXAMINER CUES EXAMINEE ACTTONS STANDARD
*8. Exam n'ee places the 8. The following following components on components in the . Console Right in Pu113Ehleci jE2[
.. Pull-to-Lock.
' position by '
rotating.the NOTE: Extension control counter-clockwise Only FN-Y-1B, and pulling VA P-2B and upwards. EHC-P-1B are
, Critical.
GN-E-1B GN-E-1B GS-E-1B GS-E-1B SA-P-1B SA-P-1B SC-P-1B SC-P-1B W-Y-1B W-Y-1B SAT EHC-P-1B EHC-P-1B SAT VA-P-1B VA-P-1B VA-P-2B VA-P-2B SAT
*9. E;:aminee Closes 9. 1J-02 Closed on PR 1J-02 by rotating as indicated by the extension voltage indicated control in the on IJ Bus.
clockwise direction on PR and verifies SAT that the 1J 0"i s s a a a s IO u i s i s s 100 iRIIHHI
*10. Examinee restarts 10. Following the following components components by restarted as rotating the indicated by Red extension controls lights illuminated in the clockwise on PLF:
direction on PLF: GN-P-1 GN-P-1 SAT GN-P-3 GN-P-3 SAT GN-P-4 GN-P-4 SAT Page 7 of 13 Filem!01. doc i
Task 0648000101 EXAMINER CUES EXAMINEE ACTIONS STANDARD
- 11. Examinee Secures 11. GN-P-2 Secured as GN-P-2 by rotating indicated by Green the extension control Light illuminated in the counter- ,
_ on PLF. clockwise direction on PLF.
~
*12. Examinee starts 12.SC-P-kBstartedas 3 SC-P-1B by_ rotating indicated by red '
the extension light illuminated. control in the , clockwise SAT l direction. I
*13. Examinee verifies 13. Secondary River or starts a Water pump is Secondary River started as Water pump. indicated by red light illuminated.
SAT
*14. Examinee Restarts 14. Following the following components components by Restarted as rotating the indicated by Red extension controls Lights illuminated.
in the clockwise direction: IA-P-4 IA-P-4 SAT FW-Y-1B FW-Y-1B SAT JPM may be terminated , at this time. (*) Denotes Critical
)
Element, j EVALUATION: SAT: UNSAT: i 4
)
Page 8 of 13 FiWeWmM101.hc l
i Task 0648000101 ) TITLE: Operate the Station Blackout Diesel Generator JPM NUMBER: 11.2.05.012 ITASK NUMBER: 0648000101 EXAMINEE: , - - ( EVALUATOR: LATE: EVALUATION OF EXAMINEE JPM: SAT: UNSAT: COMMENTS: EVALUATOR SIGNATURE: Page 9 of 13 FileW-101Acc
4 Task 0648000101 ) . Describe the consequences of operating equipment powered from-
- the Emergency Diesel. Generator with_ voltage controls set too low.
i. 1 ANSWER: Low voltage causes elevated motor current (amps) and can lead to overheating and motor damage. PEDIGREE INFORMATION: K&A NUMBER AND VALUE: 064 A2.07 RO 2.5 SRO 2.7 CFR: 41.5 43.5 45.3 45.13 TIME ESTIMATE: 2 MIN. OBJECTIVE / RATING: COGNITIVE LEVEL: 100 STUDENT
REFERENCES:
NONE HISTORY: NEW Page 10 of 13 FiWO640000101. doc
FF . , Task 0648000101 A. What must the operator do at the "C" 480v Bus to restart VA-P-1A after the breaker:has tripped on overcurrent? B.- Why must the operator teke the action described above?
.v:
&'5 t a
l l l l l I ANSWER: A. The operator must go to the breaker, and manually trip and re-close the breaker. (0.5) B. When the breaker is manually tripped this will reset the Bell Alarm swit(:h. (0.5) Ti:DIGREE INFORMATION: TIME ESTIMATE: 4 MIN. K&A NUMBER AND VALUE: 062 K4.02 RO:2.5 SRO:2.7 CFR: 41.7 OBJECTIVE / RATING: IV.G.03.04 SRO: 3.0 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
None HISTORY: QJ4G03-04-001 Page 11 of 13 FileW648000101. doc t
. , ..- . .. . . . . . . . _ . . . . - . . .. .... . ~ ..- . . . . . - . . . . .
i Task 0648000101 ! Describe the consequences of operating equipment powered from the Emergency Diesel Generator with voltage controls set too
' low.
i I
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i l' 1 I I l i l l l I' l Page 12 of 13 Fde9648000101. doc y -- -- -
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l Task 0648000101 l A. What-must the operator do at the "C" 480v Bus to restart VA-P-1A after the breaker has tripped on overcurrent?. B. Why must the~ operator take the action described above? b f ' .Q . i 4 .y, I Page 13 of 13 FileW48000101. doc
1 l l THREE MILE ISLAND UNIT 1 l INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERIORMANCE MEASURE , l TITLE: Respond to an Inadverterit ES Actuation TASK l NUMBER: 0000300501 1 TIME: 3 Minutes EXAMINEE
REFERENCE:
OP 1105-3, Safeguards Actuation System Alarm Response Procedure E-1-1 EVALUATION METHOD: PERFORM: X SIMULATE: EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 013 A2.06 IMPORTANCE: 4.0 10CFR55.45: (a) (3), (4), (6), (7) COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 10
Task 0000300501 P SIMULA10R CONDITIONS: Initialize the Trainer at IC-06.
' Place Malf. ES07B on Remote Key 1 EXAMINER ' ..
PREVIEW: i*This N N dea $ with', responding to an inadvertent ESAS actuation.'The plant will be at Hot" Shutdown.3 -An_ inadvertent actuation of.1600# ESAS will~ occur on the B Train. The Examinee will be f required to respond in accordance with' OP 1105-3. i DO NOT inform the Examinee of this fact. The sequence of events which should be followed a in the conduct of this JPM is as follows: e
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE l PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the
- Console.
- 4. Begin the cueing sequence as described. in the
] EXAMINER CUES section. i EXAMINER
- PREVIEW
- A certain amount of liberty (questionin< or
, cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the recuisite information from the Examinee, however, care must be taken to' ensure that you do not COACH the Examinee. To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly, i TASK STANDARDS: ESAS is bypassed, MU-P-lC is secured, and MU-V-37 is opened. l I R Page 2 of 10 File \00000300501. doc
Task 0000300501 EXAMINEE PREVIEW: For this event you are assigned the duties of the Primary CRO. ;The instructor / examiner will act as the Secondary CRO and the SS/SF. The auxiliary instructor is available to act as Auxiliary Operators.on the radio.or page system. When you are told to begin, you will, respond to the cues and/or indications which the Examiner will provide to you either verbally or by the si?aulator.
~
Youaretorespondasyouwobidintheplantfor l a real condition. ' TASK-CONDITIONS: The Reactor is at Hot Shutdown. The I&C Technicians are performing a surveillance on the "A" Train of the ESAS Actuation System. l i l l 1 l 1 Page 3 of 10 F11e\00000300501. doc
Task 0000300501 EXAMINER CUES EXAMINEE ACTIONS STANDARD Actuate Remote Key 1 to 1. Examinee. recognizes 1. Examinee l cause an Inadvertent ESAS that an ES recognizes that ! on "B" side. After block actuation has an ES actuation 3 has actuated delete occurred on the *B" has occurred on Malf..ES07B. ,., side, and that.it; -the' *Bi-side by , isTinadvertent. observing alarm and indications
- associated with
, .the; actuation, and, by verifying that l RCS. pressure and l
R.B. Pressure are normal, that it is inadvertent.
- 2. Examinee obtains 2. Permission permission to reset Requested.
the "B" train RCS 1600 PSIG ESAS I actuation. ROLEPLAY: *3. Examinee Resets 3. "B" Train ESAS l As SS, Grant permission to *B" Train ESAS Actuation RESET Reset "B" train ESAS Actuation by on CR, as Actuation, pressing the indicated by the
" ENABLE AND FULLY ENABLED
- TE
- CHANNEL RESET" Green Light St 3 is seque ce pushbuttons for illuminated on crit al for ep 5. Use channels RC1B, Channels RC1B, of the ti- ump feature RC2B and RC3B. RC2B, and RC3B.
to secure Makeup pump by placing e extension SAT control in St prior to bypas ng ESAS s grounds NOTE: for ailure. Only Two of the Three Channels must be RESET to allow component control.
- 4. Examinee obtains 4. Permission permission to requested, secure MU-P-1C.
l l l l l Page 4 of 10 F11e\00000300501. doc
l Task 0000M0501 l EXAMINER CUES EXAMINEE ACTIONS STANDARD ROLEPLAY: *5 Examinee secures 5. MU-P-1C STOPPED As SS, Grant permission to MU-P-1C by turning on CR, as i cecure MU-P-1C. the extension indicated by control to the Stop Green Light position.
- m. . .
illuminated. SAT NOTE: 6. Examinee opens 6. MU-V-37 OPEN on Must be opened prior to MU-V-37 by CC as indicated l closing MU-V-16C/D if depressing the OPEN by Red Light MU-P-lC has not been pushbutton on CC. Illuminated. secured first. SAT
- 7. Examinee verifies 7. MU-V-12 verified MU-V-12 Open OPEN on CC, as
, indicated by Red Light illuminated.
- 8. Examinee closes the 8. MU-V-14B/16C/16D following valves by CLOSED on CC/CR, depressing the as indicated by CLOSED pushbutton Green Lights on CC and CR for: illuminated.
A. MU-V-14B B. MU-V-16C l C. MU-V-16D JEM may be terminated at this time. (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: 1 Page 5 of 10 F11e\00000300$01. doc
Task 0000300$01 i l d TITLE: Respond to an Inadvertent ES Actuation JPM NUMBER: 11.2.05.157 l TASK NUMBER: 0000300501 EXAMINEE: y ,
,.v , s EVALUATOR: DATE:
l l , EVALUATION OF EXAMINEE JPM: SAT: UNSAT: l i' l l 4 l COMMENTS: ' 1 i 1 ! l I l i !- l i k l 1 4 11 4 ) i.
- EVALUATOR i SIGNATURE
1 a i i Page 6 of 10 F11e\00000300501. doc
Task 0000300$01 Plant Conditions - 07:00 1600 psig ESAS actuation occurred 07:10 1600 psig ESAS actuation was bypassed 07:20 a loss of offsite power occurs,
. z.,a + a .. , , ,_ .... = L ,
Describe"the' complete res'onse of"th$ Emergency Diesels and
~
p Block Loading to this event, starting at 07:00 and up:through the loss of offsite power. ANSWER:
- 1. Diesels start on ES signal but do not load. (.25 point)
- 2. Block Loading Sequence is initiated by the ES signal. (.25 point)
- 3. When the loss of offsite power occurs, the Diesel Generator Breaker will close (.25 point) (2.5 sec time delay) {
and only previously running block 1 loads will start (.25 point) (will not sequence through block loading) I PEDIGREE INFORMATION: ! K&A NUMBER AND VALUE: 064 K4,10 RO:3.5 SRO:4.0 l 064 K4.11 RO:3.5 SRO:4.0 l CFR: 41.7 ! TIME ESTIMATE: 4 MIN. l OBJECTIVE RATING: N/A / 4.2 COGNITIVE LEVEL: 400 STUDENT
REFERENCES:
ESAS OPM ; OBJECTIVE: IV.E.24.10 RO: SRO: ' HISTORY: QJ4E24-10-QO1 l l Page 7 of 10 l File \00000300501. doc j
. . .~ - . .-- -. . _ . - . . .-. ..
Task 0000300501 If the Vital Bus D (VBD)~ losses. power, describe the effect on the F.SAS system: , wg A. Bistable Cabinets
- 1
.B. . Relay, Cabinets.* _ ,
'~
;g{i
, , ~ . . ,
Actuation Cabinets
~
C. . - i < l l l l ANSWER: A. No effect (.2 point) l l B. 1-Relay cabinet (.2 point) in the B train would become ; deenergized. (.2 point) i C. 1 Actuation cabinet (.2 point) would cause 1 channel in the B train to actuate. (.2 point) i l PEDIGREE INFORMATION: K&A NUMBER AND VALUE: 013 A2.04 RO:3.6 SRO:4.2 CFR: 41.5/43.5/45.3/45.13 TIME ESTIMATE: 4 MIN. OBJECTIVE RATING: N/A / 3.52 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
OP 1107-2 OBJECTIVE: IV.E.24.057 RO: SRO: HISTORY: QJ4E24057-001 Page 8 of 10 F11e\00000300501. doc
l Task 0000300501 Plant Conditions
- 07:00 1600 psig ESAS actuation occurred 07:10 1600 psig ESAS actuation wa's bypassed
- ' 07:20.a loss of offsite power occurs.
, , ;,y ,,
Describe ~the complete response ~,of the Emergency Diesels'and
. Block Loading to this event, starting at 07:00 and up through the loss of offsite. power. _ ;
4 i Page 9 of 10 File \00000300501. doc
. . ~ . .
[' l Task 0000300501 1 !~ If the Vital Bus D -(VBD) losses power, describe the effect on the ESAS system: A.- Bistable Cabinets
~ B.. Relay Cabinets _
7: l C. Actuation' Cabinets ,
.Z l
{ r [ .- i i. i 'P l Page 10 of 10 i File \00000300501. doc l l
- s Task 0410040104 THREE MILE ISLAND UNIT 1
. INITIAL SENIOR REACTORLOPERATOR EXAMINATION JOB PERFORMANCE MEASURE
. , . . _ , y. ~
TITLE: Operate the Turbine Bypas's~ Control Valves Locally 4 TASK NUMBER: 0410040104 - TIME: 10 Minutes EXAMINEE
REFERENCE:
Local Operator Aid EVALUATION METHOD: PERFORM: SIMULATE: X 1 } EVALUATION I LOCATION: SIMULATOR: IN-PLANT: X CONTROL ROOM: 1 OPERATOR PERFORMING JPM: EVALUATOR: / / DATE , K/A: 041 A4.08 j IMPORTANCE: 3.1 10CFR55.45: (a) (3), (4), (6), (7), (8) l COMMENTS: (If results are unsatisfactory, record required i I data on sheet provided in back of this JPM.) l l Page 1 of 11
w Task 0410040104 SIMULATOR CONDITIONS:' .N/A - # EXAMINER PREVIEW: This JPN deals with operating the Turbine Bypass Valves locally %;
,.m
', # ~j , _
o, m. w,w nw . ~ 7 The. sequence of events which should-be~followed ~ in the conduct of this J14( is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE 1 PREVIEN"1section of this.JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions about the i
JPM.
- 4. Begin the cueing sequence as described in the
- EXAMINER CUES section.
A certain amount of liberty (questioning or cueing beyond the bounds of the "EXNHINER CUES") , I may be taken by'the Examiner in trying to . extract the requisite information from the l
- Examinee, however, care must be taken to ensure l that you do not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed l . incorrectly, would result in a failure to meet ! the
- TASK STANDARD") correctly.
4 For In-Plant JPMs additional safety gear may be required. The Examince should inform the Examiner that additional gear is required. The Examiner will then specify if the additional safety gear is to be donned, or to be simulated after the Examinee specifies what equipment is required and where it would be located. Failure to use the proper safety equipment is i immediate grounds for failure. TASK STANDARDS: Local manual control of MS-V-3C has been established. 4 4 Page 2 of 11 FiWO410040104 y g. ,.-, aw .e - - - -
Task 0410040104 EXAMINEE PREVIEW: For this event you are assigned the duties of the Secondary AO. The instructor / examiner will act as the Console CRO and SF. When'you are told to begin,iyou,will respond'to the cues'or indications which the Examiner will provide to you verbally. Unless otherwise informed, you are to SIMULATE all actions taken. You are expected to maintain proper i communications with the Control Room as you would in the plant for a real condition. It is very important that you describe to the instructor / examiner the physical details of the actions that you would be taking if you were actually performing the manipulations required to complete this task. Failure to adequately describe your actions could result in a failure to meet the task standard. i You are to respond as you would in the plant for i a real condition. You are expected to wear all required safety
- gear in accordance with plant procedures and policies. If during a simulated task additional safety gear is required you should notify the Examiner.
Failure to use the proper safety equipment is immediate grounds for failure. i TASK l CONDITIONS: The Reactor has been tripped, and a "Cooldown
- from Outside the Control Room" has commenced in I
accordance with EP 1202-37. t l t Page 3 of 11 l
Task 0410040104 EXAMINER CUES EXAMINEE ACTIONS STANDARD The'SF instructs you to 1. Valve is located on 1. Examinee proceeds tske local manual the west side of the to MS-V-3C and control of MS-V-3C and Main Condenser on the establishes position the valve under 322' elevation. , . ,, headphone the direction of the CRO Examinee establishes- .
^ communications et the RSD Panel who is communications'with with the RSD currently using the the RSD Panel, Panel.,
h adset communication Headphones are 3 cystem. available. ROLE PLAY: If the Examinee e catablishes communication at this
- time or later, as the ,
- CRO, inform the '
Examinee that communications are ; catablished. ' Examinee turns the NOTE: *2. 2. Examinee aligns Steps 2 through 5 are valve handwheel the holes. 4 coquence critical. clockwise to align the holes in the stem SAT and the manual i operator. POSITIVE CUE: 1 Inform the Examinee that the holes are aligned.
*3. Examinee inserts the 3. Examinee inserts pin into the holes of the pin into the l stem and the collar holes.
for the manual SAT operator. POSITIVE CUE: Inform the Examinee that the pin is inserted if they correctly describe steps 2 and 3. NEGATIVE CUE: Inform the Examinee that the pin can not be inserted into the holes if they incorrectly describe steps 2 and 3. Page 4 of 11 FileW410040104
Task 0410040104 EXAMINER CUES EXAMINEE ACTIONS STANDARD
*4. Examinee depresses 4. AUTO / MANUAL the AUTO / MANUAL switch positioned switch located on to the MANUAL the controller box position.
and then turns the switch SAT counterclockwise 90* to the MANUAL position. POSITIVE CUE: Inform the Examinee that - m the switch is in MANUAL. NEGATIVE CUE: , Inform the Examinee that the switch will not turn if they do not depress
,the switch.
*S. Examinee opens the 5. Examinee opens the equalizing valve by equalizing valve turning it located on the counterclockwise to outside of the the fully open diaphragm.
, position.
SAT POSITIVE CUE: Inform the Examinee that the equalizing valve is fully open. Page 5 of 11 FileW410040104
~ .. . . _ - . _ _ . -. .. - - . - _ - , . . - - ..
Task 0410040104 ) l EXAMINER CUES EXAMINEE ACTIONS STANDARD l
- 6. Examinee turns the 6. Examinee positions ROLE PLAY: valve handwheel the valve as l As the'CRO, have the clockwise to close directed by the CRO Examinee position ME-Vy the valve two at the RSD Panel.
3C;two turns closed and turns. l ctandby for further instructions. - POSITIVE CUE: w' Inform the Examinee that the valve is positioned - co desired if they correctly describe steps 5 and 6. NEGATIVE CUE: Inform the Examinee that it is very difficult-to turn the handwheel and it becomes progressively j difficult as the valve la moved away from the initial. position if the Examinee incorrectly describes steps 5 and 6. JPM may be terminated at this time. l (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: i l Page 6 of 11 File)D410040104
l l Task 0410040104 l l TITLE: Operate the Turbine Bypass Valves Locally l JPM NUMBER: 11.2.05.031 l TASK NUMBER: 0410040104 1 l 1 EXAMINEE: 'r - ,J l EVALUATOR: DATE: . . . 1 ! ) l EVALUATION OF ' l EXAMINEE JPM: SAT: UNSAT: l COMMENTS: l l l l l l l l i EVALUATOR SIGNATURE: i l 4 l Page 7 of 11 FileW410040104
. \
l
- Task 0410040104
. Examiner Note: No references allowed.
I l Describe how the Main Feedwater Pumps respond during a plant startup when Extraction Steam becomes available.
.c %.
l e + l r. I i i. l l l l 4 i ANSWER: The low pressure Extraction Steam will begin to enter the low pressure control valves. This will cause turbine speed to begin to increase. The governor will begin to close the control valves to maintain the current speed. The high pressure control valve is the first to close so as more low pressure steam
~becomes available, the high pressure steam will eventually g [ e_
shut off. PEDIGREE INFORMATION: TIME ESTIMATE: 2 minutes K&A NUMBER AND VALUE: 059 A1.07 RO 2. 5 SRO 2.6 CFR: 41.5 45.5 OBJECTIVE / RATING: - COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
OPM SEC G-3 HISTORY: NEW Page 8 of 11 File;0410040104
1 Task 0410040104 l l l Plant Conditions: Plant is at normal hot shutdown with the Reactor reset l l and the Turbine reset. l 3 An internal ICS fault results in the +75 psig bias
.being selected for Turbine Bypass., Valve Control.
a*- i c 4.1 . :p.i , : Jy n. - - Explain the plant response to this ICS failure.
' Include the response of Thot, Tcold, OTSG pressure, and RCS pressure,.and Turbine Bypass Valves.
,r .,
I l I
\
, i l ANSWER: Turbine Bypass Valves will close and control header pressure at 960 psig. (0.25 point) OTSG pressure will increase (0.2S point ) T-hot and T-cold will increase (0.25 point) RCS pressure will increase as the RCS heats up (0.25 point) PEDIGREE INFORMATION:
-TIME ESTIMATE: 2 MIN.
K&A NUMBER AND VALUE: 045 K4.42 RO:2.8 SRO:3.0 CFR: 41.7 OBJECTIVE / RATING: IV.E.27.17 RO: SRO:3.4 COGNITIVE. LEVEL: 200 STUDENT
REFERENCES:
HISTORY: QJ4E27-17-003 Page 9 of 11 FileW410040104
. _ _ . ~ . . . . . .. . . - - . . . - . . . _ . . - - . - - -. - . . - . . . . _ .- = . . .
i Task 0410080104
.w Describe how the Main Feedwater Pumps respond during a plant startup when Extraction Steam beconnes available.
1
.y - , . .,
l I l
. e ,, .n I
1 Io J
l l I 1
.1 J
l l l l l Il l 1 l i- I l i 1 f l Page 10 of 11 FikM410040104
. ~ . - - - . - - . - - . = _ .._ - - - .
1 Task 0410040104 1 l Plant Conditions: Plant is at normal hot shutdown with the Reactor reset ! and the Turbine reset. An internal-ICS fault results in the +75 psig bias i being selected for Turbine Bypass Valve Control. Explain the plant' response'to this ICS failure.
-Include.the response of Thot, Tcold, OTSG pressure, and RCS pressure, and Turbine Bypass Valves.
1 .5 i 4 f i I l l l l Page 11 of 11 FileW410040104
l. l l THREE MILE ISLAND UNIT 1
~
> ;+
INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERIORMANCE HEASURE
., .;17 ~ cg.
TITLE: Change the spectacle'~ flange from closed to open l between EF-V-4 and EF-V-5. TASK NUMBER: 0618010504
- v. -
TIME: 10 Minutes- U l EXAMINEE
REFERENCE:
ATP 1210-10 l EVALUATION ! METHOD: PERFORM: SIMULATE: X 1 1 EVALUATION l LOCATION: SIMULATOR: IN-PLANT: X CONTROL ROOM: OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 061 A2.04 IMPORTANCE: 3.8 10CFR55.45: (a) (8), (12) COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 10 FileW618010504. doc
l Task 0618010504 SIMULATOR CONDITIONS: N/A l EXAMINER PREVIEW: ~ This JPM deals _with swapping the spectacle flange
' ~
between EFMV-4. and EF-V-5 to the(thru position.
- m. ## a The sequence of events which should be followed 1
in the conduct of this JPM is as follows:
~
- 1. Read or otherwise inform the Examinee of the q, '
information contained inethe " EXAMINEE
~ ^
' PREVIEW" section of this JPM.'
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions about the JPM.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
2 A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure { that you do not COACH the Examinee. l' To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. TASK STANDARDS: Flange between EF-V-4 and EF-V-5 placed in thru position. l l l l Page 2 of 10 File'0618010504. doc
,. . . -. - - . - - . - - - . . . - ~ - - - _ _ - - . - . - - . . . . - I Task 0618010504 ! l [ EXAMINEE PREVIEW: For this event you are assigned the duties of the Secondary AO. The instructor / examiner will act as the Console CRO and SF. l t . s- . . . When,you are told to'begin,-you will respond to the cues or indications which the Examiner will l provide to you verbally. Unless otherwise informed, you are to SIMULATE all actions taken. l You are expected to maintain proper ,
- communications with the Control Room as you I would in the plant for a real condition.
It is very important that you describe to the instructor / examiner the physical details of the actions that you would be taking if you were actually performing the manipulations required ' to complete this task. Failure to adequately describe your actions could result'in a failure to meet the task standard. You are to respond as you would in the plant for a real condition. You are expected to wear all required safety gear in accordance with plant
- procedures and policies. If during a simulated task additional safety gear is required you should notify the Examiner.
Failure to use the proper safety equipment is immediate grounds for failure. TASK CONDITIONS: Emergency Feedwater suction sources are being depleted. A suction supply must be lined up frca river water. l l 7 Page 3 of 10 [. FileW618010504. doc
Task 0618010$04 EXAMINER CUES EXAMINEE ACTIONS STANDARD Inform Examinee to swap 1. Examinee proceeds to 1. Valves are located spectacle flange between the intermediate in the EF-V-4 and EF-V-5 to the building basement intermediate thru position. 281' hallway.; building basement a ' ' 281' hallway.
- 2. Examinee contacts 2. Permission granted the Control for permission to perform job task.
- 3. Examinee visually 3.EF-V-4 and EF-V-5
' checks to ensure are closed.
EF-V-4 and EF-V-5 are closed. POSITIVE CUE: If a visual check for valve position is made, inform Examinee that EF-V-4 and EF
- are closed.
NOTE: *4.Using the 4. Flange nuts and Examiner ensures that appropriate bolts are enough bolts are wrenches, the removed. removed / loosened to allow Examinee removes a the spectacle flange to be the nuts and bolts SAT either pulled and from the flange. reinstalled are rotated. NOTE: *S. Examinee removes 5. Spectacle flange If Examinee requests new and rotates the rotated to the gaskets, inform them new spectacle flange thru position gaskets are not required. to the thru with the two position and flexitallic places flange gaskets and 4 between two placed between flexitallic the two pipe j gaskets, and into ends, position between the two pipe ends. SAT
*6. Examinee reinstalls 6. Flange nuts and l flange bolts and bolts tightens nuts. reinstalled, l
SAT Page 4 of 10 FileW18010504. doc
... -. . - . - --. _ _ _ _ . - - .- - . _ . ~_ _. .-
i Task 0618010$04 EXAMINER CUES EXAMINEE ACTIONS STANDARD i POSITIVE CUE: If Examinee performs steps.4,.. 5 and 6 correctly, inform .
~ u -
them that the spectacle ~ l flange is installed properly. I J NEGATIVE CUE: . If Examinee performs steps 4, 5 and 6 incorrectly, inform them that the spectacle ; flange is not installed ' properly. , JPM may be terminated at this time. l (0)- Denotes Critical Element. EVALUATION: SAT: UNSAT: I l l l Page 5 of 10 FileVM18010504. doc
1 l Task 0618010504 l l l l TITLE: Change the spectacle flange from closed to open between EF-l V-4 and EF-V-5.
- JPN NUMBER
- l TASK NUMBER: 0618010504
-- e .. , . , ,
EXAMINEE: l l EVALUATOR: DATE: > l l EVALUATION OF EXAMINEE JPM: SAT: UNSAT: COMMENTS: l EVALUATOR SIGNATURE: i i Page 6 of 10 FileW618010504. doc
Task 0618010504 Examiner Note: No reference allowed. A Main Steam line has ruptured on the "A" OTSG in the Reactor Building. A. What design feature ensures Emergency Feedwater flow to'the intact"OTSG7 B. Whateth[erfunctionsdoesthisdesignfeatureperform? Y l l N ANSWER: j A. Emergency Feedwater cavitating venturis. 4 i B. Limit mass and energy release to the Reactor Building. Limit flow to the OTSGs to reduce excessive RCS overcooling. l 4 a l PEDIGREE INFORMATION: TIME ESTIMATE: 2 MIN. K&A NUMBER AND VALUE: 061 K4.04 RO 3.1 SRO 3.4 CFR: 41.7 OBJECTIVE / RATING: COGNITIVE LEVEL: 100 STUDENT
REFERENCES:
OPM Section I-01 p. 12 HISTORY: NEW Page 7 of 10 FileW618010504. doc
Task 0618010504 PLANT CONDITIONS The unit was initially at 100% power when a Loss of Offsite l Power concurrent with an ESAS actuation occurs. l
~
. How long after;this event dojthe motor-driven Emergency i Feedwater Pumps get aistart sigrihl?
Include inipour explanatibn the individual time elements that make up the EFW start ~ delay. .
- i. - a v;
., n n a ,
t ANSWER: With ESAS
- 10 sec. for D/G start (0.1) 5 sec. Block 2 (0.1)
! 5 sec. Block 3 (0.1) 5 sec. Block 4 (0,1) 5 sec. EFW start delay relay (0.1) J 30 sec. Total (0,5) PEDIGREE INFORMATION:
- TIME ESTIMATE
- 3 MIN.
061 K4.02 RO:4.5
~
K&A NUMBER AND VALUE: SRO:4.6 CFR: 41.7/45.6 OBJECTIVE / RATING: IV.C.05.15 SRO:3.8 COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
None HISTORY: QJ4C05-15-001 4 Page 8 of 10 FileW618010504. doc
i Task 0618010$04 OTSG in the Reactor A Main Steam line has ruptured on the "A" l Building. .
-A. What design feature ensures Emergency Feedwater flow to the !
intact OTSG? !
"'*[-r' '
*M ,
+ .&h ,,
i B. What" other functions does' this desigr. feature perform?, , ,
, .r a l 3
l l l 1
, 1 i
r 1 ( f i i e i I l l l Page 9 of 10 Fikn0618010504. doc 1
. . . . _. . - . - . . - . _ - - . - . . _ - - ~ - . . = _ - _ . . . -.
1 Task 0610010$04 i PLANT CONDITIONS , The unit was initially at 100% power when a Loss of offsite Power concurrent with an ESAS actuation occurs.
.....4..- ,.
How long after this event do the motor-driven Emergency .
, , eN,.,
4~ Feedwater Pumps'g'et a start' signal? "'""*""'" #'
~
Include in your explanation the 1ndividual time elements that make up the EFW start delay. 4 l l 1 l l Page 10 of 10 FileW618010504. doc
v Task 0000650504 THREE MILE ISLAND UNIT 1 l INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERIORMANCE NEASURE
,n #
TITLE: Respond 'to a Loss of InstirumerSt LAE (IC-V-4) i TASK NUMBER: 0000650504 TIME: 7 Minutes . . . EXAMINEE
REFERENCE:
Local Operator Aid EVALUATION
. METHOD: PERFORM: SIMULATE: X EVALUATION LOCATION: SIMULATOR: IN-PLANT: X CONTROL ROOM:
OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 008 A2.05 RO 3.3 SRO 3.5 IMPORTANCE: 10CFR55.45: 41.5 43.5 45.3 45.13 COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 12
Task 0000650504 SIMULATOR CONDITIONS: N/A EXAMINER l PREVIEW: This JPM deals _with establishing local manual control of IC-V-4.following a loss of Instrument Air. The sequence of events which should be followed in the conduct of this JPN is as follows:
- 1. Read or otherwise -inform the Examinee of the l information contained in the " EXAMINEE l PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee to ask questions about the JPM.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
l A certain amount of liberty (questioning or l cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee. To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. 1 Page 2 of 12 FileW410040104
Task 0000650504 For In-Plant JPMs additional safety gear may be required. The Examinee should inform the Examiner that additional gear is required. The Examiner will then specify if the additional safety. gear is to be donned, or to be simulated after the Examinee specifies what equipment-is required and where.it would be located. , 1 Failure to use the proper safety equipment is immediate grounds for failure. i l TASK l STANDARDS: Local manual control of IC-V-4 has been i established. ' l l 1 l Page 3 of 12 FileW410040104
, . . - . . - ~. . , _ _ - - - - - . . - - . . - . - . - - . - . . - -
1 Task 0000650504 i ~ EXAMINEE
. PREVIEW: For this event you are assigned the duties of the Pritsary AO. The instructor / examiner will j act as the Console CRO and SF.
4 i When you are',tblM td beg E [ you will. respond to "the cues or* indications which*the* Examiner'will
' ~
provide to you. verbally.. ,
.p s,,, - ,
Unless otherwile~ s informed, youzare to, SIMULATE b all actions taken
- You are expected to maintain proper communications with the Control Room as you would in the plant for a real condition.
It is very important that you describe to the l instructor / examiner the physical details of the l actions that you would be taking if you were actually performing the manipulations required to complete this task. Failure to adequately I 4 describe your actions could result in a failure to meet the task standard. You are to respond as you would in the plant for ,
- a real condition.
i. ! You are expected to wear all required safety
; gear in accordance with plant procedures and policies. If during a simulated task additional
, safety gear is required you should notify the
- Examiner.
I, Failure to use the proper safety equipment is j immediate grounds for failure. 4
- TASK CONDITIONS
- A loss of Instrument Air has occurred.
The reactor has been tripped and Reactor Coolant Pumps are still operating. It is desired to keep the Intermediate Closed Cooling 'ater supply to the Reactor Coolant Pumps. Intermediate Closed valves IC-V-3, 4 and 6 are in the CLOSED position. Page 4 of 12 FileV)410040104
Taskmea504 1 EXAMINER CUES EXAMINEE ACTIONS STANDARD The SF instructs you to 1. Examinee proceeds to 1. Examinee proceeds go to IV-V-4, establish IC-V-4. Valve is to IC-V-4 and communications with the located on the first -establishes Control: Room, take , . . m w4 floorxof the Aux. communications , manual; control of IC-V-47 s, . [ Bldg. Near the f.g3 ; with the Control ) cnd open it.. 'IC-P-1s, and Room., establishes. - c - communication with the' Control-Room. ~ *
- ROLE PIAY:
If the Examinee " catablishes communication at this time or later, as the CRO, inform the Examinee that communications are cotablished. l NOTE: *2. Examinee rotates the 2. "T" handle Steps 2 through 3 are "T" handle under the aligned with ! ccquence critical. clutch arm to align groove under the with the groove under "T" handle. l the "T" handle. SAT l POSITIVE CUE: Inform the Examinee that the "T" handle is aligned with the groove if they correctly describe step 2. Page 5 of 12 FileW410040104
1 Task 0000650504 , EXAMINER CUES EXAMINEE ACTIONS STANDARD
*3. Examinee rotates 3. Actuator aligned, handwheel to align clutch arm held the manual' actuator down and the "T" and pulls down and handle is in the holds the clutch ara: "g-groove.
!' J';d , "'^ to giushithe "T"~ "" ' handle 71:sto the SAT groove. POSITIVE CUE: . Inform the Examinee that j the "T" handle.is in the 4 groove if they correctly 3 describe steps 2 and 3. NEGATIVE CUE: Inform the Examinee that the "T" handle is not in the groove if they incorrectly describe oteps 2 and 3 or do not j d3 scribe them in i cufficient detail. . NOTE: 4. Examinee places the 4. Valve placed in If the valve is valve in the "OPEN" the "OPEN" correctly positioned position by turning position. i then the "T" handle will the handwheel in the rcmain in the groove if counter-clockwise
, the manual actuator is direction.
epplying force against the automatic actuator. d 1 i 1 Page 6 of 12 FikA0410040104 1
Task 0000650$04 EXAMINER CUES EXAMINEE ACTIONS STANDARD POSITIVE CUE: ' l Inform the Examinee that !*'i the valve indicates open if they. correctly. , . . . y . ., describe steps 1 through, jj , g: j . -w <= r 7, 7, g ,
- 4. l NEGATIVE CUE: ,
Inform the Examinee that - the valve is not correctly positioned if they incorrectly I I describe steps 1 through 4 or do not describe them in sufficient detail'.
- 5. Examinee releases the 5. Clutch arm is I clutch arm and released and ;
verifies that IC-V-4 valve remains remains open. open. JPM may be terminated at this time. l (*) Denotes Critical Element. l EVALUATION: SAT: UNSAT: Page 7 of 12 FileW410040104
1 Task 0000650504 i
- TITLE
- ' Respond to a Loss of Instrument Air (IC-V 4) 4
) JPM NUMBER: 11.2.05.037 ITASK NUMBER: 0000650504 ' 'i EXAMINEE: .I i i EVALUATOR: DATE: i EVALUATION OF EXAMINEE JPM: -SAT: UNSAT: i COMMENTS: i i i l i 4 l { EVALUATOR SIGNATURE: Page 8 of 12 FileW410040104
Task 0000650504 EXAMINER NOTE: No references allowed. What signals will automatically close IC-V-47 T g ,3 3. ; 1
)
; c j
4 1 ANSWER:
- 1. 30 psig Reactor Building pressure ESAS signal
- 2. Line Break Isolation actuated by Low Intermediate Closed Surge Tank level concurrent with Any ESAS signal r
PEDIGREE INFORMATION: TIME ESTIMATE: 1 minute KEA NUMBER AND VALUE: 004 A3.08 RO 3.6 SRO 3.8 CFR:- 41.7 45.5 OBJECTIVE / RATING: COGNITIVE LEVEL: 100 STUDENT
REFERENCES:
OPM Section B-10 p. 6 HISTORY:- { Page 9 of 12 FileW410040104
Task 0000650504 P EXAMINER NOTE: No references allowed. , I What design feature protects the Intermediate Closed Cooling ' Pumps from low flow conditions?
- ~. . .. ...
1
~ j l
i i i l I ANSWER: IC-V-74 (minimum flow valve) opens if j IC-V-2 or IC-V-3 close i or IC-V-4 and IC-V-6 close. , i PEDIGREE INFORMATION: TIME ESTIMATE: 1 MIN. K&A NUMBER AND VALUE: 008 A1.01 RO 2.8 SRO 2.9 : CFR: ! OBJECTIVE / RATING: i COGNITIVE LEVEL: 100 , STUDENT
REFERENCES:
OPM Section B-10 p. 6 HISTORY: Page 10 of 12 4 FileW410040104
Task 0000650504 What signals will automatically close IC-y-47 l i 1 l l l l l l l l l I l Page 11 of 12 FileW10040104 i I
- . -. . . . - - . .. - - - ,.._. ..- _.-~- - .~--..-- -. ....-. -....-...
Task 00006M24
'What design feature prot'ects the Intermediate Closed Cooling Pumps from low flow conditions?
' - a
- g. it,
' I '* f N
Page 12 of 12 p g g g ,g
~
i l 3 THREE MILE ISLAND UNIT 1 l i INITIAL SENIOR REACTOR OPERATOR EXAMINATION i JOB PERFORMANCE MEASURE
~
TITLE: ' Perform 'IR$gulating Group Transfer Operations-to/from the Auxiliary Power Supplies TASK NUMBER: 0010120101 1 TIME: 5 Minutes l EXAMINEE
REFERENCE:
OP 1105-9, Control Rod Drive EVALUATION METHOD: PERFORM: X SIMULATE: - EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: OPERATOR PERFORMING JPM-EVALUATOR: / / , DATE K/A: 001 A4.08 y i IMPORTANCE: 3.4 { 10CFR55.45: 41.7 45.5 to 45.8 ! t COMMENTS: (If results are unsatisfactory, record required " data on sheet provided in back of this JPM.)
)
Page 1 of 12 0010120101. doc !
I l Task 0010120101 SIMULA'A GR CONDITIONS: Initialize the trainer at IC-17. ce Place A and B ICS Feedwater Loop Masters, the ICS Reactor Demand, the Diamond Rod Control Panel and the ICS SG/Rx Master to Hand. Place Group 7 control rods on the auxiliary power supply and return the Diamond Rod Control Panel to Manual Sequenced operation. ,. l-i EXAMINER-l PREVIEW: This JPM deals with transferring the power supply for Group 7 control rods from the auxiliary supply to the normal supply. The examinee will take the turnover with A and B ICS Feedwater Loop Masters, the ICS Reactor Demand, the Diamond Rod Control Panel and the ICS SG/Rx Master to Hand. The examinee will be given a cue in the form of a directive from the examiner acting as the Shift Foreman to return Group 7 control rods to the normal power supply l- and then return the Diamond rod control station l to the Manual Sequenced mode of operation. l l The sequence of events which should be followed j in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
! 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
j 4. Begin the cueing sequence as described in the l EXAMINER CUES section. l l l l Page 2 of 12 0010120101. doc
l Task 0010120101 A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure ! that you do not COACH the Examinee. ( 3 + , - , To complete this task successfully, the Examinee must complete each critical element (an-element of the task which, if omitted or performed incorrectly, would result in a failure to meet the
- TASK STANDARD") correctly.
TASK STANDARDS: Group 7 control rods are on the normal power supply, the Diamond rod control station is in the Manual Sequenced operation in Run speed. 1 1 l l
)
l 1 l l 1 i Page 3 of 12 0010120101. doc
Task 0010120101
- EXAMINEE PREVIEN: For this event you,,are assigned the duties of the Primary CRO. The instructor / examiner will act as the Secondary CRO and SF. The auxiliary instructor is available,to act as Auxiliary Operators onithe radio;.or_page system. . . _
i o w . w w a#.o You are expected to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to. verify proper completion. When you are told to begin, you will respond to the cues and/or indications which the Examiner J will provide to you either verbally or by the 1 simulator. You are to respond as you would in the plant for a real condition. i ^ TASK CONDITIONS: Reactor is at 100% power, ICS in track, with A and B ICS Feedwater Loop Masters, the ICS Reactor Demand, the Diamond Rod Control Panel and the ICS SG/Rx Master in Hand. Group 7 control rods are on the auxiliary power supply. 4 W i s l l l Page 4 of 12 0010120101. doc t _ - -
Task 0010120101 i EXAMINER CUES EXAMINEE ACTIONS STANDARD Direct the examinee to *l. Examinee selects 1. ' SEQ / SEQ OR" pushbutton r0 turn Group 7 to its normal' sequence override on the depressed, ' SEQ OR" power supply, and the Diamond control panel, lamp illuminated and Diamond control panel to the ' SEQ" lamp remains Manual Sequenced mode of illuminet.et on the cperation. , Diamond c trol panel. N 5 %. 3m j SAT i NOTE: Steps 1 through 5 may *2. Examinee selects 2. " GROUP /AUXIL" be performed in any auxiliary on the Diamond pushbutton depressed, esquence. control panel. ' GROUP" lamp extinguished and
'AUXIL" lag l illuminated on the Diamond control panel.
SAT
*3. Examinee selects jog on 3. *RUN/ JOG" selector the Diamond control switch rotated to the panel. ' JOG" position on the Diamond control panel.
t SAT l j *4. Examinee selects the 4. " GROUP SELECT" selector l l group to be transferred switch rotated to the ! (Group 7) on the Diamond '7" position, " MAN j i control panel. TRANS" lamp ' l illuminated, *SY" lamp , j illuminated after a l i short time delay, all l on the Diamond control panel. SAT
*5. Examinee selects the 5.
- SINGLE SELECT" switch control rods to be rotated to the 'ALL" transferred on the position on the Diamond Diamond control panel. control panel.
l SAT
- 6. Examinee verifies that 6. *SY" lamp illuminated the "SY" lamp has on the Diamond control illuminated. panel.
Page 5 of 12 0010120101. doc l l
l ! 1 l l Task 0010120101 l l EXAMINER CUES EXAMINEE ACTIONS STANDARD l t NOTE: The order in which *7. Examinee selects clamp 7. ' CLAMP / CLAMP REL" l ctops 7 and 8 are performed on the Diamond control pushbutton depressed, i is not critical, however panel. ' CLAMP" lasp they cannot be performed illuminated and " CLAMP until after steps 1 through REL" lag extinguished ( 5 have been performed.
- zon the Diamond control l ' panel.
! SAT l *8. Examinee transfers Group 8. ' MAN TRANS" pushbutton l
- 7 control rods to their depressed, 'TR CF" lamp i normal power supply. on the Diamond control l panel and 8 individual 1
- CONTROL ON" lasps for l Group 7 control rods on the PI panel ,
extinguish. l SAT l
~
- 9. Examinee selects clamp 9. " CLAMP / CLAMP REL" release on the Diamond pushbutton depressed, control panel. " CLAMP REL" lamp l l illuminated and " CLAMP" {
- lamp extinguished on i )
the Diamond control l panel. NOTE: Step 10 cannot be *10. Examinee selects group 10. " GROUP /AUXIL" performed until after steps on the Diamond control pushbutton depressed, l 7 and 8 have been performed. panel. *AUXIL", " MAN TRANS" and "SY" lamps extinguished, and
- GROUP" lamp
- illuminated on the Diamond control panel.
l SAT NOTE: Steps 11 and 12 can 11. Examinee selects "OFF" 11. " GROUP SELECT" selector be performed at any time and on the group select switch rotated to the in any sequence as long as switch. "OFF" position. step 8 has been performed. l 12. Examinee selects "OFF" 12. "Sli4GLE SELECT" switch I on the single select rotated to the *0FF" switch, position. Page 6 of 12 0010120101. doc
Task 0010120101 EXAMINER CUES EXAMINEE ACTIONS STANDARD NOTE: Step 13 cannot be *13. Examinee resets transfer 13. 'TRANS RESET" performed until after step logic on the Diamond pushbutton depressed, 10 has been performed. control panel. 'TRANS RESET" lamp illuminated, and Group 7
- CONTROL Qel" lasp
" estinguished on the
' EI$$4,f'-, .
' Diamond control panel.
SAT NOTE: Steps 14 and 15 can *14. Examinee selects 14. ' SEQ / SEQ OR* pushbutton be performed at any time and sequence on the Diamond depressed, ' SEQ OR" in any sequence as long as control panel. lamp extinguished, step 8 has been performed. ' SEQ" lamp remains illuminated, Group 7
- CONTROL ON" 1 asp illuminated (all on the Diamond control panel),
and 8 individual
- CONTROL ON" lamps for Group 7 control rods on the PI panel illuminated.
[ SAT
*15. Examinee selects run on 15. "RUN/ JOG" selector the Diamond control switch rotated to the panel. "RUN" position on the Diamond control panel.
SAT l JPM may be terminated at this time. (0) Denotes Critical Element. EVALUATION: SAT: UNSAT: Page 7 of 12 0010120101. doc
I Task 0010120101 TITLE: Perform Regulating Group Transfer Operations to/from the Auxiliary Power Supplies JPM NUMBER:- 11.2.05.013 l TASK NUMBER: 0010120101 EXAMINEE: I EVALUATOR: DATE: EVALUATION OF ; EXAMINEE JPM: SAT: UNSAT: COMMENTS: 1 1 i l EVALUATOR SIGNATURE: ; Page 8 of 12 0010120101. doc
t-I Task 0010120101 Given the following. plant conditions: Control RodLGroup 6 consists of 8 individual rods, their l individual positions are as follows: 1 U./f i Rod 6-3. 82% Rod 6-697%" Rod 6-8 91% _ I [ The remaining rods are 90% Which rod (s) are inoperable due to misalignment? Explain your answer. l ANSWER: l Rod 6-3 4 Group Average = ((5
- 90) + 82 + 97 + 91) / 8 = 90 Rod 6-3 misaligned the most by 90 - 82 = 8%
8% of 139" = 11.12" Inoperable by T.S. 4.7.1.2 New Group Ave. = ((5
- 90) + 97 + 91) / 7 = 91.14%
Rod 6-6 now misaligned by the most by 97 - 91.14 = 5.86% 5.86% of 139" = 8.15" Not inoperable by T.S. 4.7.1.2 PEDIGREE INFORMATION: i TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 014 K5.02 RO:2.8 SRO: 3.3 CFR: 41.5/45.7 OBJECTIVE RATING: IV.E.13.19 SRO:3.0
- COGNITIVE LEVEL
- 300 STUDENT
REFERENCES:
Tech. Spec. 4.7.1.2 HISTORY: NEW l Page 9 of 12 0010120101. doc i
.- ~ _ - -_
Task 0010120101 Plant Conditions: 73% Power 400 EFPD RC-P-1D secured due to #1 Seal failure Full Incore Instrumentationfsystem operable . s y , _, J _ If quadrant tilt is 8.5%, what~is the maximum allowable power level? n 3 a ANSWER: 66.66% Power 1 Quadrant Tilt 8.50% . FIS Limit 4.33% Excess Tilt 4.17% 1 4.17%
- 2 = 8.34%
75% - 8.34% = 66.66% I PEDIGREE INFORHATION: I TIME ESTIMATE: 5 MIN. K&A NUMBER AND RATE: Generic 2.1.12 4.0 CFR: 43.2 43.5 45.3 OBJECTIVE RATING: COGNITIVE LEVEL: 300 STUDENT
REFERENCES:
COLR and Tech. Spec. 3.5.2.4d HISTORY: NEW Page 10 of 12 0010120101. doc
. .- . - . . . . . - -- . - . . = ~ - .. - - - - -
l Task 0010120101 l
= ,g . ,
Given the following plant conditions: l 1 Control Rod Group 6 consists of 8 individual rods, their l . individual positions are as follows: !~ Rod 6-3~~ 82%. ~ v' #?* Rod 6-6' 97% M' " Rod 6-8 91% The remaining rods are 90% Which rod (s) are inoperable'due to misalignment? t-l Explain your answer. l l
)
i l l l Page 11 of 12 0010120101. doc
Task 0010120101 Plant Conditions: l 73% Power 400 EFPD
.RC-P-lD secured due to #1 Seal. failure "
Full Incore.In'strumentation.systemt operable If quadrant tilt is 8.5%, what is the maximum allowable power level?- l I I l Page 12 of 12 0010120101. doc l
l Task 0538010101 THREE MILE ISLAND UNIT 1 , INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERNRMANCE MEASURE TITLE:
'Phrform the~ Required ~ Actions" .ner- for Loss of Stator Coolant Pump with Failure of the Standby Pump to Auto Start e
TASK j NUM8ER: 0538010101 , TIME: 15 Minutes EXAMINEE
REFERENCE:
ARP MAP L, L-1-7, L-3-7 EVALUATION l METHOD: PERFORM: X SIMULATE: EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: BW/A01 AA1.3 IMPORTANCE: 3.7 10CFR55.45: COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) l i l 1 Page 1 of 10
Task 0538010101 SIMULATOR CONDITIONS: Initialize the Trainer at IC-17. IO OVERRIDE: 05A6S25-ZDICSGNP5B(3), NAP, OFF to prevent auto start GN-P-5B)'
,ae r< . "
HALFUNCTION: EG04A assign to Remote Key #1 ) I EXAMINER ; PREVIEW: This JPM deals with performing the required l actions to stabilize the plant followingLthe loss - of the running stator coolant pump with failure i of the standby pump to auto start. l DO NOT inform the Examinee of this fact. ) The sequence of events which should be followed , in the conduct of this JPM is as follows: l 4 i
- 1. Read or otherwise inform the Examinee of the 1 information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Ovnsole. )
- 4. Begin the cueing sequence as described in the i 4
EXAMINER CUES section. l l l l i l l Page 2 of 10 File 0538010101. doc
Task 0538010101 A certain' amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract the requisite information from the Examinee, however, care must be taken to ensure that you do not COACH the Examinee'._ . To complete this task successfully, the Examinee must complete each critical element (an element of the task which, if omitted or performed incorrectly, would result in a failure to meet the " TASK STANDARD") correctly. TASK STANDARDS: The standby stator coolant pump running, terminating the turbine runback prior to load being reduced to below 300MWe. i Page 3 of 10 File 05380101014ac
Task 0538010101 EXAMINEE , PREVIEW: For this event you are assigned the duties of the Secondary CRO. The instructor / examiner will act as the Primary CRO and Shift Foreman. The auxiliary instructor is,available to act as Auxiliary operators.onithe radio or p' age system. You are expected.tc perform the Immediate Manual Actions of the appropriate EP/AP/ATP,from
- ,, memory,.but then use the procedure to' verify proper completion.
When you are told to begin, you will respond to the cues or indications which the Examiner will , provide to you either verbally or by the simulator. You are to respond as you would in the plant for i a real condition. TASK
- CONDITIONS
- The plant is at 100% power, equilibrium Xenon, EOL.
The Dispatcher has requested that load not be reduced below 300Mwe or there will be a possibility of loss of the grid due to insufficient generation. 4 4 4 t Page 4 of 10 i File 0538010101. doc
Task 0538010101 EXAMINER CUES FXAMINEE ACTIONS STNJDARD l Activate MALFUNCTION 1. Determine Loss of 1. Examinee
- Remote Key #1 GN-P-5A has occurred. diagnoses by l alarms and indications that
. .a.GN-P-5A has tripped and a
., b plant runback is in progress Note: , *2. Determine GN-P-5B *2. Examinee %
Runback needs to be did not auto start. starts GN-P-5B. terminated prior to , reducing load below SAT 300Mwe l Note: 3. Verify runback 3. Load reduction MAP L-3-7 would have to terminated and plant stopped prior to be reset at the local stabilizes at > 300 reducing below alarm panel by an MWe. 300 MWe and plant Auxiliary Operator. is stable. SAT l JPM may be terminated at this time. (o) Denotes Critical Element. EVALUATION: SAT: UNSAT: l Page 5 of 10 File 0$38010101. doc l
Task 0538010101 1 TITLE: Perform the Required Actions for Loss of Stator Coolan l Pump with failure of Standby pump to Auto Start l JPM NUMBER: 11.2.05.NEW [ TASK NUMBER: 0538010101 ri+ r' EXAMINEE: l l EVALUATOR: DATE:
~
i j EVALUATION OF EXAMINEE JPM: SAT: UNSAT: l COMMENTS: l 4 i i 1 i i l i 4 i j i e i i EVALUATOR SIGNATURE: i 4 Page 6 of 10 m 0538010101. doc
Task 0538010101 Plant Conditions: J l Main Generator Hydrogen Pressure = 45 psig Dispatcher requires 200 MVARs lagging What is the'imaximum Megawatts for which the generator.is rated under the conditions'above? + j ANSWER: " l 915. Megawatts' maximum l (allow 870'- 915) i PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 045 K4.09 RO: 1.8 SRO: 2.2 OBJECTIVE / RATING: COGNITIVE LEVEL: STUDENT
REFERENCES:
OP 1106-1, Figure B-2B HISTORY: NEW . Page 7 of 10 Mis C D8010101 Aoc
Task 0538010101 J When loading the Main Turbine, the rate'of change of first-stage
-shell inner metal temperature must be limited to 150*F/ hour.
What-is the minimum time required to prevent exceeding this limit when changing load from 10% to 80%?
. >;t e
l J t t ANSWER: 50 minutes 4 i PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. ; K&A NUMBER AND VALUE: 045 K4.05 RO: 1.7 SRO: 2.2 i OBJECTIVE RATING: ) COGNITIVE LEVEL: 300 i STUDENT
REFERENCES:
OP 1106-1, Figure B-5 HISTORY: NEW Page 8 of 10 Fue 0$38010101. doc
. . -.. . . . - . . . - ~ . - . - . . .
Task 0538010101 Plant Conditions: Main Generator Hydrogen Pressure = 45 psig Dispatcher requires 200 MVARs lagging ! l What is the maximum Megawatts for which the generator is rated under^the conditions'above? '" V;fg . 3. l l l I l l I I i 1. 1 i l l 1 I l l-t I l r r l Page 9 of 10 l File 0538010101. doc
Task 0538010101 When loading the Main Turbine, the rate of change of first-stage shell inner metal temperature must be limited to 150*F/ hour. What is the minimum time required to prevent exceeding this limit when changing load from 10% to 80%?
- w . , wh > ; . 2 1
i ' 3 ]- I. I d 3 1 4 4 t i e Page 10 of 10 File 0538010101. doc
l
- THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERIORMANCE MEASURE 6, ,
"^'4 *
..e .,,z. fi*:[Q ', r%
Respond'toaMalfun,ctiod"p'inPressurizerLevel
- s. .
TITLE: Indication or Control TASK . NUMBER: 0008140401 m ' d TIME: 10 Minutes EXAMINEE MAP G, G-2-5
REFERENCE:
EP 1202-29, Pressurizer System Failure . EVALUATION METHOD: PERFORM: X SIMULATE: EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: 1 OPERATOR PERFORMING JPM: 4 a EVALUATOR: / / DATE K/A: 000 028 AA2.10 IMPORTANCE: 3.4 10CFR55.45: (a) (3) , (4), (6), (7) [ COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) 1 I Page 1 of 11 File \0008140401. doc
. i Task 0008140401 J
SIMULATOR , -. CONDITIONS: Initialize the trainer at IC-17. Assign MALFUNCTION.RC04A to Remote Key #1 at 100% severity over 60 seconds. (Fails selected
. pressurizer level) .% , y EXAMINER - - -,/ , - w-w, PREVIEW: This 'JPN" deals 'with resp $ riding to'a failure in
, pressurizer; level indication -low. 7-The plant will be at 100% power when a failure of the selected pressurizer level instrument will occur. .,.
. . . :, yg vpy ,,q g Do NOT inform the Examinee of this fact.
The sequence of events which should be followed i in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as describeci in ';he EXAMINER CUES section.
1 (Continued) i f 8 i Page 2 of 11 F11e\000814 0401. doc
.. _ . . - . = . - . .- . . _ _ _ - . . .- _. - .
Task 0008140401 (Continued) EXAMINER l PREVIEW: A certain amount of liberty (questioning or
~
.., cueing beyond the bounds of the 'l EXAMINER CUES")
7may be"taken by ths; Examiner inftrying to extract
~
.2s"+" the requisite'information from the-Examinee, however, care must be taken to ensure that you do not COACH the Examinee.
Tocompletethisiasksuccessfully, the Examinee
,must complete each critical element (an element of"the^ task which, if omitted or performed incorrectly, would result in a failure to meet 1
the " TASK STANDARD") correctly. I TASK STANDARDS: RC-1 LT-3 is selected to provide correct , controlling pressurizer level indication. Page 3 of 11 ktie\0008140401. doc
Task 0008140401 EXAMINEE PREVIEW: For this event you are assigned the duties of 3 the Primary CRO. The instructor / examiner will act as the Secondary CRO and the SF. The auxiliary instructor is available to act as Auxiliary Operators on the radio or page' system. You are expected.to perform the Immediate Manual Actions of the appropriate EP/AP/ATP from memory, but then use the procedure to verify proper completion.
)
..When you are told to begin, you will respond to ,
the cues or indications which the Examiner will provide to you either verbally or by the i simulator. I l You are to respond as you would in the plant for a real condition. TASK l CONDITIONS: The Plant is at 100% power. l No surveillances or maintenance is in progress. l Page 4 of 11 F11e\000814 0401. doc
Task 0008140401 ) I l EXAMINER CUES EXAMINEE ACTIONS STANDARD l Active Malfunction RC04A 1. Examinee' places . 1. Makeup flow l on Remote Key #1 to cause MU-V-17 into manual adjusted as a failure of pressurizer control and. adjusts indicated by
- level indication low, makeup flow to MU-24 A/B FI on equal letdown flow CC. i minus seal injection by manipulating the control switch on CC.
4
*2 Examinee selects- 2. Alternate the alternate pressurizer level !
4 pressurizer level instrument RC-1 and temperature LT-3 selected, instruments to indicated by determine the white light failed instrument illuminated. on CC. SAT
- 3. Examinee verifies 3. Pressurizer 1
that pressurizer heaters verified. heaters are energized on CR. l l The JPM may be terminated after either RC-1 LT-3 selection, or after RC-1 LT-3 is selected and the Examinee displays successfully manipulation of pressurizer level. . (Examiner perrogative. (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: 4 Page 5 of 11 F11e\0008140401. doc l
- -. . = . . . . . . . . . - . - . _ . . . _ . - - . _ . .
Task 0008140401 TITLE: Respond to a malfunction in Pressurizer Level Indication of Control JPM NUMBER: l TASK NUMBER: 0008140401 y 4 EXAMINEE: -o -.
- ,- ,o
.+. . , , :.y ;
EVALUATOR: DATE: EVALUATION OF - L' + EXAMINEE JPM: SAT: UNSAT: COMMENTS: EVALUATOR SIGNATURE: Page 6 of 11 F11e\0008140401. doc
Task 0008140401 Given the following condition:
- Compensated Presssurizer Level is NOT available RCS temperature is 300*F Current Pressurizer Level Differential Pressure is 120" on
, PPC point ~A0503 -
Based on these conditions, what is actual Pressurizer Level using the differential pressure? ANSWER: 300" (+5") actual level PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 011 A1.02 SRO: 3.6 OBJECTIVE / RATING: V.D.11.06 SRO: 2.8 COGNITIVE LEVEL: 300 STUDENT
REFERENCE:
EP 1202-29, FIGURE 1 HISTORY: NEW Page 7 of 11 File \0008140401. doc
Task 0008140401 l Given the following conditions: I Pressurizer Level channel RC-1-LT3 had been removed from service due to electrical problems. Pressurizer Level channel RC-1-LTl has just failed LOW l RC-1-LTl had been indicating 225" prior to its failure
- RCS temperature is 579'F l -
Reactor power is 100% Both level channels will be inoperable for 60 hours A. or the given conditions, what actions are required? B. \ If response to A above is correct, ask how much Feed Volume i would be required to go to Hot Shutdown at 532*F? ANSWER: A. Return at least 1 to operable status within 48 hours or be in at least HOT SHUTDOWN within the next six hours. (Time for Cold Shutdown not required.) B. 6353 gallons From Table 1: l 579*F - 576*F = 653 gals 576*F - 573*F = 700 gals 573*F - 532*F = 5000 gals PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 011 A2.03 RO: 3.8 SRO: 3.9 l Generic 2.1.12 RO: 2.9 SRO: 4.0 OBJECTIVE / RATING: V.D.11.05 SRO: 3.0 l COGNITIVE LEVEL: 300 STUDENT
REFERENCE:
EP 1202-29, TABLE 1 . Tech Spec 3.5.5 l HISTORY: NEW i l Page 8 of 11 i F11e\0008140401. doc
, _ - _ = - ~ . - - . . - - . . .. - ~ . . . . . - - . ~ - . .- . .. . . - - - . - . - . - - . - . - .-.
Task 0008140401 l I Given the following condition:
- Compensated Presssurizer Level is NOT available
- RCS temperature is 300*F Current Pressurizer Level: Differential Pressure is.120".on PPC point.A0503- ,
,74 ., , , . , ..
Based on these conditions, what is actual Pressurizer Level using tho sdifferential pressure? , L l t i i l l i l-l' l-Page 9 of 11 F11e\0008140401. doc f I _,_ _.
l j Task 000s140401 Given the following conditions: Pressurizer Level channel RC-1-LT3 had been removed from service due to electrical problems. Pressurizer Level channel RC-1-LT1 has just failed LOW RC-1-LT1. had been indicating ,225", prior to its failt re RCS temperature is 579'F l - Reactor power is 100% Both level channels will be inoperable for 60 hours For the given conditions, what actions are required? f i i l l 1 l l I l l l l l l l t. l l Page 10 of 11 File \000814 0401. doc i l
Task 000s1404o1 How much Feed Volume would be required to go to Hot Shutdown at l-7 532*F? t l l' e ,j w.jg a,.. 5',? - 7":,
._ ,/).'
c , ' , O D o Q UT Cv/oG 7Na f5 W N 0"FST 0 (hk./hlhhh' / p /s ty7msaED I
),
l l l l l l l-I -, Page 11 of 11 File \0008140401. doc
. i, THREE MILE ISLAND UNIT 1 INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERNOIINANCE NEASURE TITLE: Perforr an, EmergencyBoration[4 TASK .
NUMBER: 0000240LM , TIME: 4 Minutes > , s. a ;o . . 7, ' r-EXAMINEE
REFERENCE:
OP 1103-4, Soluble Poison Concentration Control ATP 1210-1, Reactor Trip. ESAS Checklist EVALUATION METHOD: PERFORM: X SIMULATE: , EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: OPERATOR PERFORMING JPM: i EVALUATOR: / / DATE K/A: 000 024 AA1.17 IMPORTANCE: 3.9 10CFR55.45: (a) (3), (4), (6), (7), (8), (12) COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) , Page 1 of 10
Task 0000240501 SIMULATOR CONDITIONS: Initialize the Trainer at IC-17 Insert Malfunction RD0201. Group 4 Rod 4 Stuck Out. Trip the Reactor ! Perform all ATP 1210-1 Immediate Manual Actions, except Emergency Boration. Ensure the BAMT is listed on the status board as the Tech. Spec. Tank, and the Boton concentration is 17,500 ppa. EXAMINER ~ PREVIEW: This JPM deals.with performing an Emergency Boration from the BAMT. The plant will be in a post trip state with one control rod stuck out. The Examiner will have the Examinee borate from the BAMT. DO NOT inform the Examinee of this fact. The sequence of events which should be followed in the conduct of this JPM is as follows: 1 1
- 1. Read or otherwise inform the Examinee of the '
information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued) I 1 Page 2 of 10 FileW000240$01. doc
Task 0000240501 i , (Continued) 1 EXAMINER l PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES") may be taken by the Examiner in trying to extract i the requisite information from the Examinee, l i however, car'e'must be-taken to ensure that you do
~
1 i not COACH the' Examinee. To complete this task successfully, the Examinee
. must complete each critical element (an element
.of-the task which, if omitted or performed -
incorrectly, would result in a failure to meet the
- TASK STANDARD") correctly.
TASK STANDARDS: The BAMT is lined up supplying injection to the makeup tank through MU-V-51. i l Page 3 of 10 FiW0000240501. doc
a . 1 Task 0000240501 EXAMINEE I PREVIEW: ,For this evpt;.you;are assigned'the duties of c the Primary CRO.M The instructor / examiner will I act as the Secondary CRO and the SF. The
. auxiliary instructor is available to act as j Auxiliary Operators onlthe radio or page system.
Mf# ^
; g . h i.
You are. expected to perform the ImmediAte Manual , Actions of the appropriate EP/AP/ATP from memory, but then use the pro,cedure to, verify proper completion.
)
w
. e
^
j "
- - . . . z.'
When you are.' told to begin, you will respond to the' cues and/or indications'which the Examiner will provide to you either verbally or by the simulator. i You are to respond as you would in the plant for I a real condition. ! TASK ! CONDITIONS: Stable post trip with 1 control rod stuck full 3 out, i l Page 4 of 10 Fil#000240501. doc -
Task 0000240501 EXAMINER CUES EXAMINEE ACTIONS STANDARD As the SF, inform the 1. Examinee checks.the 1. AO contacted by Examinee to initiate boric acid ptmps (CA-P- page or radio cmergency boration from 1A/B) set at maximum system. Tech. Spec. Tank. Inform stroke. The ESAS the Examinee to set the checklist has this Stroke Counters on CA-P- information. Or the 1A/B to 28000. Examinee may contact the primary AO for this information. The required position for the stroke settings of CA-P-1A/B is MAX.
,7 ,
- 2. Examinee checks the 2. Stroke counters Stroke counters 1,et at have > 028000-28000, displayed in the windows on the LWDS panel.
- 3. Examinee positions The 3. " Local / Remote"
" Local / Remote" Switches switches for CA-P-for CA-P-1A/B to the 1A/B on the LWDS "LWDS" position, panel are in the "LWDS" position.
*4. Examinee opens 4. MU-V-51 OPEN as HU-V-51. indicated on the LWDS panel by Red Light illuminated. l SAT
*5. Examinee starts the 5. CA-P-1A and B are boric acid pumps, STARTED as CA-P-1A/B. indicated on the LWDS panel by Red Light illuminated.
CA-P-1A SAT CA-P-1B SAT
- 6. Examinee checks that 6. Examinee may use the level in the BAMT local tank level, decreases. or level indication on the Computer, Point A0475.
JPM may be terminated at this time. (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: ___ Page 5 of 10 FileWOO240501. doc
-. . _ . . . - . - - . _ . - - . - ~ - - - _ _ - . . . . - . . - . _ . . . . . - . _ . - . . ..-.._ . . _. .. ._- - - -
k Task 0000240501 i 4 TITLE: Perform an Emergency Boration JPM NUMBER: 11.2.05.047 l TASK NUMBER: 0000240501
- EXAMINEE
i EVALUATOR: DATE:
~
~
h % i
- EVALUATION OF -
EXAMINEE JPM:
. SAT : - UNSAT:
4 i 1 ! COMMENTS: l 1 l l 1 i t 1 i ) i e EVALUATOR SIGNATURE: Page 6 of 10 FileW240501. doc
_. . q \ l ! Task 0000240501 i-l' I ! Due to a plant transient the following Makeup Tank conditions l l exist.-
%, - ' f' i
i ! Level 20" Pressure 10 psig l . . A. Does operation with these~ Makeup Tank conditions pose any . problems? - y i ! l B. If yes, what problem or problems? l .i ANSWER: A. Yes B. HPI pump (Makeup Pump) gas entrainment during a large break LOCA. Insufficient Makeup Pump NPSH PEDIGREE INFORMATION: TIME ESTIMATE: 3 MIN. K&A NUMBER AND VALUE: 004 K5.26 CFR: 41.7 OBJECTIVE / RATING: COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
1104-2 Figure 1 HISTORY: NEW Page 7 of 10 r w oooo240 son.4=
I Task 0000240501 During the JPM, both stroke, counters.were set to 28000. !
- p: gg
- v l
Approximately how many gallons would be pumped from the Boric Acid Mix Tank t'o the RCS using both pumps set to their maximum stroke? -
,. l% w ~ ~ '
l i ,,
'" ! : t', .
ANSWER: 5180 - 5600 gallons Work: j Approximately 925 - 1000 gallons at a stroke counte.r setting of 10000. 925 1000 x 2.8 x 2.8 2590 2800 x 2 2 5180 5600 PEDIGREE INFORMATION: TIME ESTIMATE: 5 MIN. K&A NUMBER AND VALUE: 004 A1.04 4.1 CFR: OBJECTIVE / RATING: COGNITIVE LEVEL: 300
' STUDENT
REFERENCES:
1103-4 Figure 3
. HISTORY: NEW Page 8 of 10 n.cooo2*5ouse
1 Task 0000240501 Due to a plant transient the following Makeup Tank conditions exiat. 'W 4% li'. - Level 20" Pressure 10 psig ,
.. s7er . .
A. Does operation 'with these Makeup Tank conditions pose any problems? , B. If yes, what problem or problems? s
. 1
, . 1
,%; :,- x + l l
l Page 9 of 10 rd.cono2moian
Task 0000240501 During the JPM, both stroke _ counters were set to 28000, rr, : .a;@yiyiO
+ :;
Approximately how many gallons"would be pumped from the Boric Acid Mix Tank to the RCS using both pumps set to their maximum stroke? ,j . pp;g.y , a u. .. i- r _ I l l l l I l i
'1 Page 10 of 10 !
N eoco mosot a 4
1 . \ l l
. .THREE MILE ISLAND UNIT 1 V -
. g.y ,
INITIAL SENIOR REACTOR OPERATOR EXAMINATION JOB PERfCBMANCE REAERE ' TITLE: Establish Long Term Recirculation after a LOCA TASK NUMBER: 0058020101 - TIME: 12 Minutes. ,- J
" tt ,;f . y ,,j - j, EXAMINEE
REFERENCE:
OP 1104-4, Decay Heat removal System EVALUATION METHOD: PERFORM: X SIMULATE: EVALUATION LOCATION: SIMULATOR: X IN-PLANT: CONTROL ROOM: X OPERATOR PERFORMING JPM: EVALUATOR: / / DATE K/A: 000 011 EA1.11 IMPORTANCE: 4.2 10CFR55.45: (a) (12) COMMENTS: (If results are unsatisfactory, record required data on sheet provided in back of this JPM.) Page 1 of 11 FileU4400MD03. doc
E Task 344a050303 a SIMUIATOR . CONDITIONS: Initialize the Trainer;at IC-22. Set BWST <6' 4" by setting DHMBWST = 3.3E6 Open DH-V-64
'Open DH-V-6A/B on CC/CR Close DH-V-5A/B on CC/CR Close'RC-V-3 on CC/CR s Stop'BS-P-1A/B ,
Close BS-V-1A/B Close BS-V-2A/B Close BS-V-3A/B
,Stop MU-P-1B Throttle.DH-V-19A/B to <c2800 Gal./ Min.
EXAMINER PREVIEW: This JPM deals with placing the RCS in long term recirculation as described in OP 1104-4 (Decay Heat Removal) The RCS will in a post LOCA condition. The Examinee will direct establishment of long term recirculation. The sequence of events which should be followed in the conduct of this JPM is as follows:
- 1. Read or otherwise inform the Examinee of the information contained in the " EXAMINEE PREVIEW" section of this JPM.
- 2. Inform the Examinee of the TASK CONDITIONS.
- 3. Allow the Examinee 2-3 minutes to scan the i Console.
- 4. Begin the cueing sequence as described in the EXAMINER CUES section.
(Continued) Page 2 of 11 rde\3448050303. doc
Task 3448050303 l (Continued) EXAMINER PREVIEW: A certain amount of liberty (questioning or cueing beyond the bounds of the " EXAMINER CUES")
- 7 ,"imay beLtaken_by the Examiner in trying.to extract the requisite"information from the Examinee, however, care must be taken to ensure that you do
, not COACH the Examinee.
To complete this task successfully, the Examinee must complete each critical element (an element of the task *which,.if omitted or performed incorrectly,'hould result in a failure to meet the " TASK STANDARD") correctly. TASK STANDARDS: The Active method of long term recirculation has been established. l l Page 3 of 11 FueG440050303Anc
Task 3448050303 I EXAMINEE PREVIEW: For this event.you are assigned the duties of the Primary CRO. The instructor / examiner will act as the Shift Foreman and Secondary CRO. The 4 auxiliary instructor _is available to act as Auxiliary Operators on_the
,,.c radio or page' system.
_ u u.
;When you are, told to begin [you will1 respond to
~
the cues and/or indications which the Examiner will provide to youLeither verbally;gr.by the simulator. You are to respond as you would in the plant for l a real condition. TASK CONDITIONS: The reactor tripped on low pressure , approximately 13 hours ago. 1 l A LB LOCA 13 hours ago has caused RCS depressurization to the current value. The core is being cooled with low pressure injection flow. HPI flow has been secured. LPI pumps are taking a suction from the RB sump. ! The Immediate manual Actions and the applicable followup actions of ATP 1210-1, ATP 1210-2, and ATP 1210-7 have been performed. ESAS has been BYPASSED / DEFEATED / RESET to gain ES Component control, d DH-V-64 is Open. DH-V-2 breaker has been closed. DH-V-19A/B have been throttled for LPI flow. Adequate core cooling is being maintained by "A" and "B" LPI flow. Page 4 of 11 File \34480$0M3. doc
Task 3448050303 EXAMINER CUES EXAMINEE ACTIONS STANDARD
- Direct the examinee to 1. Examinee verifies 1. DH-P-1A motor ostablish long term DH-P-1A is running. current & red recirculation cooling per breaker closed OP 1104-4. .
< j . +,, . , , light,are
- ; -- , ' checked.
- 2. Examinee verifies 2. RC-V-3 closed.
RC-V-3 closed.- - 4 *3. Examinee opens 3. RC-V-4 open. RC-V-4 to
- establish aux. SAT spray flow;
- 4. Examinee verifies 4. Low range RCS pressure is in pressure equilibrium with RB indicator is pressure. reading approximately O.
Note: Examiner provided 5. Examinee verifies 5. Logs verified. this information. it has been 12 hours since the LB LOCA. i *6. Examinee verifies 6. DH-P-1B motor DH-P-1B operating. current & red breaker closed
- light are l checked.
(*) Denotes Critical Element. l l Page 5 of 11 FileG44s0MD03Jac
. .. . - - - - . - . . - = - - . - .-.-._.-. _ _ - _ - . - - - . - - .. .
Task 34480$0303
~
l EXAMINER CUES EXAMINEE ACTIONS STANDARD NOTE: The "B" Channel is *7. Examinee resets 8. "A" channel 400# not simulated. Inform the the 400# DH bistable reset as l cxaminee that the "B" bistable in the indicated by channel has been reset and "A" ESAS cabinet white OUTPUT-to reset the 400# bistable module 1-8-4 by STATE indicator for "A" channel'only. ~ ^ ' ' ' depreasing the. bec'oming dim'. toggle' switches below'the OUTPUT SAT STATE and OUTPUT MEMORY lights.
- 8. Examinee verifies 8. MAP C-1-6 clear that MAP alarm C-1-6 is clear.
*9. Examinee partially 9. DH-V-3 partially opens DH-V-3 by open.
depressing the open pushbutton SAT 5.5 Seconds on CC. Inform examinee that motor *10. Examinee opens 10. DH-V-2 open currents will be DH-V-2 by monitored. depressing the SAT open pushbutton on CC. Inform the examinee that *11 Examinee opens 11. DH-V-1 open motor currents will be DH-V-1 by monitored. depressing the SAT open pushbutton on CC. JPM may be terminated at this time. (*) Denotes Critical Element. EVALUATION: SAT: UNSAT: Page 6 of 11 rua44aomo3Ac .
. . - .. -. . . - . .. _ . - - -- - . - - .- - = - - - -.- . . -
! Task 34480$0303 TITLE: Direct establishment of Long Term Core Circulation to
- prevent boron concentrati~on 4
JPM NUMBER: 11.2.05.156 l TASK NUMBER: 3448050303 EXAMINEE: l l 1 l 4 EVALUATOR: DATE: . e
+ .- c. ;
i l 4 l EVALUATION OF 3 EXAMINEE JPM: SAT: UNSAT: l l COMMENTS: l i 4 i 3 1 4 i i 4 I I e 5 EVALUATOR
- SIGNATURE
4 i I J Page 7 of 11 i Fileu448050303. doc
Task 3448030303 Plant Conditions A large break LOCA occurred and the operators have established Reactor Building sump recirculation. Only DH-P1-B is operable. '
. 's f, , 4 .. . <
m q s %i4R p ;g.,]; , . c Q. " Point out ths flow path fdr this n'odi N a'iflow-printi. Include major components in the flow path.. jpy[ n . '. m t, 1 ANSWER: Take suction on the RB sump (0.2) via DH-V-6B (0.2), using DH-P-1B (given), discharge through the DC cooler (0.2), through open cross connect valves (DH-V-38A&B) (0.2) feeding both LPI lines via DH-V-4A & B (0.2) PEDIGREE INFORMATION: TIME ESTIMATE: 4 MIN. ! K&A NUMBER AND VALUE: 011 EK3.08 RO:3.9 SRO:4.1 l CFR: 41.5/41.10/45.6/45.13 OBJECTIVE / RATING: IV.A.11.15 SRO: 3.4 COGNITIVE LEVEL: 200 STUDENT
REFERENCES:
ATP 1210-7 I HISTORY: NEW Page 8 of 11 mmk c i
l Task 34480$0303 You are preparing to start Decay Heat Removal during a plant I cooldown. Plant Conditions: RCS Temperature ~ 210*F - RC-P-1A islthe.only;RCP.;operating g JAO---
~.
e A. What is the' minimum and maxisiini' pressures for simultaneous operation of DH-P-1A and RC-P-1A? l B. What RCS pressure instrinsent is,to be used? _ .y z .- .4 l l l ANSWER: A. Minimum 307 psig (11) i Maximium. 319 psig (11) B. RC3A-PT5/P12 1 i i PEDIGREE INFORMATION: l TIME ESTIMATE: 3 MIN. K&A NUMBER AND VALUE: 005 K4.02 RO 3.2 SRO 3.5 CFR: 41.7 OBJECTIVE / RATING: COGNITIVE LEVEL: 200 ! STUDENT
REFERENCES:
1104-4 HISTORY: NEW i Page 9 of 11 l rd o44eomos.d :
Task 3444050303 Plant Conditions l A large break LOCA occurred and the operators have established
- Reactor Building sump recirculation. %
only DH-Pl-B is operable.' 4
..,gg .~,
Point out'the.fl W b path'forlthis mode on a. flow-print. Include major components in'the flow path.
- s. , ., .
. 1. . .
4
* ~ r s 4:e n .9 Liu .> wc . .
1 t Page 10 of 11 FileU448050303. doc
Task 3448050303 You are preparing to start Decay Heat Removal during a plant cooldown. Plant Conditions: RCS Temperature 210*F , . RC-P-1Atis the only RCP operating '" A. Wh'at is the minimum and maximum pressures for simultaneous operation of DH-P-1A and RC-P-1A7 B. What RCS pressure instrument is to be used? Page 11 of 11 raeo44mso3034x
J r, 4 4 k l 1 Administrative JPM A.1.1 1 4 5
6* ExaminerInfonnation: Examinee is provided with the followin0' e Examinee infonnehon peGe e List of plant conditions imm which to perform an ECP
. Copy ofistest procedure
. . STA ECP with errors (wrong Xenon value from Nuclear Engineers)
Examinee Actions: The examinee is to perform an independent ECP from the data provided The examinee will then compare their ECP to the STA's ECP. The examinee is to identify and resolve any discrepencies found. Examinee Response:
. See correct ECP answer key.
Examinee should identify the following errors on the STA ECP:
. Line 8-incorrect Xenon value
. Line 10-Incorrect value
. Line 11 -Inconed values
. Line 12a -Incorrect tolerance band (due to incorrect Xenon value)
. Line 12b-Incorrect values e Line 12c-incorrect values
i d Examinee information: I i l Plant Conditions: l l Hot Shuklown 1 40 hours post-hip imm 100% power for 280 days T-evg. 532 dooroes F. i . RCS Somn 1387 ppm Somn depletion factor 0.98 Group 8 at 30% Use Xenon from curves Task: You are the Shift Foreman preparing for an approach to cribcality. The STA has already performed an ECP.
- You are to
. Perform an independent ECP from the data provided.
.
- Compare your ECP to the STA's ECP.
- e identify and resolve any discrepancies found.
d 4 9 i h
, $og2Ecr [CP (Ga N t?c) 1103 15]
Revision 27 ENCLOSURE 1 Page 1 of 1
- Estimated Critical Rod Position l- CALCULATION IS FOR AN ECP AT 532 i 2*F ON
- DATE TIME
- 1. CYCLE BURNUP M@ EFPD
~2.a. FINAL MEASURED BORON CONCENTRATION ' /038 opmB 2h. BORON DEPLETION CORRECTION FACTOR O 78 (NAS Display 10, Control Room Log, Nuclear Engineenng) 3 4 2.c. FINAL CORRECTED BORON CONCENTRATION (2a)X(2b) = /d/7 opm8
- 3. CRG 8 POSITION AT CRITICALITY 30 % WD i b 5 AM
~
- 4. FUEL EXCESS REACTIVITY (FIG 1)
- 5. CRG 8 REACTIVITY WORTH (FIG 2) -0 // %AM
- 6. INVERSE BORON WORTH (FlG 3) /N- opmB/% Ak/k
- 7. BORON REACTIVITY WORTH l ppmB / Inverse Boron x ( 1) = -7 595 % Ak/k L #2c / #6 _
- 8. XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR., FIG 4) -O 'k % Ak/k ,
- 9. SAMARIUM AND PLUTONIUM BUILDUP (FIG 5) 9a. TIME SINCE LAST SHUTDOWN 40 HRS 9b. REACTIVITY DUE TO BUILDUP -0 0// % Ak/k i
- 10. INSERTED CRG 5-7 WORTH REQUIRED FOR CRITICALITY I
'(FUEL) (CRG 8) (BORON) (XENON) (SM)~ , --/ M %M
#4 #5 #7 #8 #9b , l
- 11. ESTIMATED CRITICAL ROD POSITION (FIG 6) 7b_% WD on b CRG '
$ 12. CRITICAL ROD POSITION TOLERANCE BAND (FIG 6) 12a. CHECK ONE 0.5% Ak/k if l Xenon (#6)l is 5 0.5%Ak/k)-
0.8% Ak/k if IXenon (#8)l is > 0.5%Ak/k) 12b. MINIMUM ROD WITHDRAWAL LIMIT (#10 - #12a) N % WD on CRG . 5 - 12c. MAXIMUM ROD WITHDRAWAL LIMIT (#10 + #12a) 70 % WD on CRG 7 ! CALCULATED BY: DATE/ TIME l APPROVED BY (SRO): DATE/ TIME [ Seno copy of this Enclosure to Shift Engineenng Send onginal to Operations for filing I ! 9 l
/
CONTROL i Fn COPY Number TMI NUCLEAN Operating Procedure 1103-15B l ( Tce Reymon No. Estimated Critical Conditions 27 Apphcatminy/ Scope Responsde Offee Effectve Date l TMl Division Plant Operations Dir. This document is within QA plan scope 07/2 ales X Yes No Safety Rewows Required X Yes No THAlNING DEPl. . List of Effective Pages :e Ea2n Revision Eagg Revision Eagg . Rewston Eagg Revision 1 27 2 27 3 27 4 27 5 27 6 27 7 27 8 27 9 27 10 27 . 11 27 12 27 13 27 (- 14 15 27 27 16 27 17 27 c 1 7,
, , sI .'s{ s&'s - ' , 1s <" >l ;JJ e ^ : ;'t i ^ .,. Gi
$.! . %: ' ' Idgnatuft s < >
s s., 2 3 2 ', i Dalef}fjikis 7- 1
..-].;:-
Originator
/ /g 73 Procedure Owner
((#[f8
, 7- C /~ f T
( Approver g y 7 - / { - f f' 4 1 l l
_. . . - . . . . . _ . . . .~. . . _ = _ - . _ - - - - - i . Examinee Information: l Plant Conditions: ! Hot Shutdown mod NO ! 40 hours post, trip from 100% power for 280 dsys~' ~' ' .
.. "y.
"^
l T-avg. 532 degrees F. 7 RCS Boron n"" ppm /038 ppm g, ._ ,g Boron depletion factor 0.98 -
' - o ,
i Group 8 at 30% l Use Xenon from curves - Task: You are the Shift Foreman preparing for an approach to ciiticality. The STA has ! already performed an ECP. ! You are to: t
.. Perform an independent ECP from the data provided.
{, . Compare your ECP to the STA's ECP.
. Identify and resolve any discrepancies found.
t l [ I i l l { h
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l l STrA CcP bWf h IleA 1103-158 ! Revision 27 ENCLOSURE 1 Page 1 of 1 Estimated Critical Rod Position l l. CALCULATION IS FOR AN ECP AT 532 2'F ON: DATE TIME 1
- 1. CYCLE BURNUP 400 EFPD
! 2.a. FINAL MEASURED BORON CONCENTRATION /030 ppmB 2.b. BORON DEPLETION CORRECTION FACTOR O88 (NAS Display 10 Control Room Log Nuclear Engineering) 2.c. FINAL CORRECTED BORON CONCENTRATION (2a)X(2b) = /o/7 opm8
- 3. CRG 8 POSITION'AT CRITICALITY 30 % WD l
- 4. FUEL EXCESS REACTIVITY (FIG 1) / 6 h i % ak/k
- 5. CRG 8 REACTIVITY WORTH (FIG 2) -d l/ % ak/k '
- 6. INVERSE BORON WORTH (FlG 3) /M ppmB/% ak/k
- 7. BORON REACTIVITY WORTH
' ppm 8 / Inverse Boron 1 -- i. 5 9 5 ,,, , l
- #2c / #6 j
- 8. XENON REACTIVITY WORTH (PPC, NUCLEAR ENGR.. FIG 4) -OW N % ak/k
(' ,
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- 9. SAMARIUM AND PLUTONIUM BUILDUP (FIG 5) r 9a. TIME SINCE LAST SHUTDOWN HRS l( - 9b. REACTIVITY DUE TO BUILDUP -0 dll % ak/k I
l
- 10. INSERTED CRG 5-7 WORTH REQUIRED FOR CRITICALITY
~(FUEL) (CRG 8) (BORON) , (XENON) (SM)~
, , /, 7[]~ auk
#4 #5 #7 #8 ~b -- -
- 11. ESTIMATED CRITICAL ROD POSITION (F 3
% wD on CRd {
- 12. CRITICAL ROD POSITION TOLERANCE BAND (FIG 6) 12a. CHECK ONE F
OJ% ak/k if IXenon (#8)l is 5 0.5%Ak/k) 0.8% ak/k If IXenon (#8)l is > 0.5%Ak/k) 12b. MINIMUM ROD WITHDRAWAL LIMIT (#10 - #12a) 72 % WD on CRG8 12c. MAXIMUM ROD WITHDRAWAL LIMIT (#10 + #12a) 75 % WD on CRG 6 CALCULATED BY: (2 DATE/ TIME APPROVED BY (SRO): DATETTIME j Send copy of this Endosure to Shift Engineenng ( Send onginal to Operations for filing 9
I-9 1 Administrative JPM 1 l A.1.2 l l l
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- - i .
t i 4 Examinerinformation: Examinee is to be provided with the following j
- Examinee information pape e Completed Tech. Spec. Surveillance with enors.
i-Examinee Actions: The examinee is to review the surveillance. The examinee is to identify and discuss any descrepencies found. The examinee is to identdy any addihonal adions required resulting from his review. Examinee Response: A. MU-P-1B Pump outboard bearing vibration is in the alert range. This win require that the cover sheet be stamped *lSi Review Required" and a IST Coordinator / System Engineer resolution and si0nature is required. B. MU-V-16C Opening time is in excess of the maximum stroke time. MU-V-16C fails the Post Maintenance Test
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MU-V-16C remains out of serv 6ce and the T.S. timeciock will continue. This will require that the cover sheet be stamped *lSI Review Required" and a IST Coordinator / System En0ineer resolution and signature is required. 1 I i
- O l Examinee information Task Conditions:
You are the Shift Foreman propering to perform your review of a completed Tech. Spec. Surveillance. d 1300-3H was perfoimed as scheduled. MU V-16C has been out of service for the past 12 hours for motor replacement. I The plant is on a 72 hour T.S. 3.3.2 hmeciock for MU-V-16C. Post Maintenance Testing of MU-%16C is to be satisfied be the surveillance. Task: You are to: 1 e Review the surveillance datasheets. l
- Identify and discuss any discrepancies found.
! e identify any additional adions required resulting from his review. 4 i e f
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1 Examinerinfoemation: l i
- Examinee is to be provided with the following.
- Examinee information page
?1 . <
Examinee Actions: Reference. Slaps 4.21.2 throu0h 4.21.5 of AP 1029
. Evaluate plant / system conditions to determine the appropriate position for valve MU V-77A. !
(Applicant should consult with operating procedure OP 1104-2 and/or Print C-302 461 to determine valve is required to be open.) e Dired adion to retum valve to required position. (Crtical Step.) l
. Evaluate system operability during time valve was out of required position to determine reporting requirements, r any. .
(System was operable.) l e Determine if operability test is required at this time. (Testing of redundard train is not required.) e investgate cause for occunence. ExaminerVerbal Cue:
'Six hours earlier an Auxiliary Operator fell, and was injured, while re-possbonin0 this valve at the conclusion of a surveillance test. He was transported off-site for medical attention. His
! surveiliance test documentation indicates this valve had been opened prior to the mishap." l . Perfonn required Log ordry in unified control room log. l (Discuss only - adual 100 entry should not be performed.) e identify edion to prevent recurrence (Applicant should discuss how independent verification process or other administrative controls should prevent recunence.) !
- Notify Duty Supt. If deemed deliberate.
(Not required.)
- Fill out a CAP Form for cribcal components.
(Terminate when applicant pulls up a CAP Form on the PC.) l I I l i t
.- .. - ~ _ ~ - -..- - - . . . . - . . _ - _ - . . - - _ - . - - . - - - . _ - . . . - . 's l Examinee information 4 .
, Task Conditions: You are the Shit Supervisor. You receive the fotowing ce5 tem an AO in the gdnt: " -
- "I am the Primary Aungiary Operator While perfonning a system tag out for 4 maintenance, I just found valve MU V-77A throttled to 50% cdosed ph. I think this valve is required to be funy open during normal run power operations.-
Plant io at ID076 lb r diihig8 58> r.,u. tuo makt. orporv. In aferpicytb6. You are to penorm the required saions vor this situation I 3 e 4 1 1 J 1 i 3 J f i 4 4 y - - --
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i i i i Examiner information: n Examiner will have the examinee perform the following:
,..i
- 3,
- Sign in on the blanket Operations RWP ! , r -
- 4 +H' -
. Discuss RWP requirements - dosimetry, protective clothing, exposure limits,
, - , .x i etc;~ .
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. Show examiner survey map for IC-V-4 area I
e Explain survey map data - A .- l1_ e identify component location
. Exit RCA properly -c
- Sign off RWP l
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Ravicinn K ~~ # ' 06/19/98 ~ ( - y W MpA . Module'Nine: Radiation Work Permit ~(RWP) I LO: . - State the function of the RadiatiogWoirkl Permit (RWP)? 2 A Radiation Work Permit (RWP) contains details concerning a radiological area and has three major functions: ., r . An RWP authorizes entry into a RWP Required Area. ' k An RWP details the radiological requirements necessary for the work being performed, including dosimetry requirements, protective clothing requirements, and special precautions. , t
> An RWP provides information concerning radiological conditions in the work area.
LO: Given a completed RWP, be able to state the following informatign:
> RWP number > limiting conditions
( > RWP revision number RWP status - worker description standardized protective
> need and closeout dates clothing set requirements
> required available exposure > dosimetry requirements
> if automated access is > ESRD alarm setpoint permitted > pre-job briefing /
> work description . discussion requirements
> job location > special precautions and
> radiological data instructions
> Rad Con monitoring requirements The type of RWP used at TMI and Oyster Creek is a computerized, electronically generated process that is part of the GMS-2 Maintenance Job Order System. RWPs become a permanent part of the plant's machinery history. The GMS 2 System provides the worker with the ability to review RWPs from various computers located throughout the plan't.
It is up to each worker to read and understand the RWP before signing on to it. Any questions concerning the RWP should always be directed to the Radiological Controls Department. Rad Con personnel can also provide the worker with a detailed briefing of the current radiological conditions in the work area. Any worker who is unable to access RWPs via GMS-2 should contact Rad Con for assistance. ( Alternate means of accessing RWP information may be available, including hard copy RWPs and/or the use of a dedicated computer system for RWP access. At TMI," easy access" computers are provided so that the general work force can readily access RWP information.
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T b kl}$O ' y;6 s Part I of the RWP contains the followine informationi - " letROO3 RO3 UADIATICIE WORK FERatIT PART 1 ' "[ J00/RB&O 110001 RIf7 RE70 1 womK TTFE PM F33 N UNIT 1 Rwr nasp DATE 1/04/95 mwr CIosID DATE / / STATUS APVD MARA REVIEwt C95001 ces CAfeeORY 03 Rag AVAIL mar 300 MREN NfW 9106000 LEAD CRAFT 6211 INSPECTOR AUTO ACCESS Y CORIPOWEIFF MU-V-1 VALVE 14C AUI. BLDG. 281 - 0 j .
)- ..
l WORK MarWTFTICII INSPECT SYSTEM FOR LEAKAGE AND ENSURE AREA IS PREPARED FOR PREVErtIVE MAINTENANCE WORK ON VALVE. SPECIAL PRECAUTICGIS JOB LOCATZQat Atri. BLDG. / 281* ELEV. / MUV ALLEY AND InssTADCT20ess Y oaxeranton SMITH EmTE' 12/25/94 total RAD COEl AFFROYAL NS690 JOHNSON , ACT DOSE EfffER UPDATE F1-RWP PART2 F2-RNP PART3 F4-TEMPLATES F5-DRESSOUT F7-NEXT RWP9 F8-RWP REQUEST F9-SPEC PREC F10-RWP CLOSE F11-PRIlff RWP Pt F12-JO At*TACHMZtfrS F13-RWP REV F20-RE'IURN F21-RADCON KDi ' v22-MAIN F23-SIGNOFF F24-HELP l k The Job Order Number (JO#) and the RWP Number: These numbers are the same ( and are consecutively generated with each newjob order. r The RWP Revision Number
'r PSE ( Y or N ): Noted if the evolution is a Planned Special Exposure. > RWP Need and Closcout Dates: The Need Date is the date the job is scheduled to start.
The Closecut Date is the date the job order is closed. r Required Available Exposure: Will indicate amount of dose a worker must be able to receive (without exceeding applicable dose limits) to work under this RWP.
'r Auto Access ( Y or N ): Will designate if the RWP is setup for auto access of workers.
l l 'r Work
Description:
If RWP is " Job Specific", details on work will be provided as part of the RWP. " Blanket RWP" is listed for otherwise noted evolution. If the planned work differs in scope from what is described on the RWP, immediately report the differences
- to Rad Con.
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~" Rwision S 06/19/98
( Part 2 of the RWP describes the following Radiological Data: _ p '; p . t-WMROO4 RO4 RADIATIcel WORK FEltIEIT PART 2 J00/ serp 6 910001 nave 1 - Consposerr --MU-V-1 VALVE
- 3. : s .
d DurITs ' MADIATICII LEVELS aEEE 400 ass /am AT MU-V-1 VALVE
- LIMITING CoeIDITICIIS N GIIs.,15 asR/um META 200 naAD/ER svavET # 165910174 l
UNITS i CourrumTIoet LEVELS asnt 40,000 Drit/100Cas, AT a MU-V-1 VALVE ! LIaEITING COEIDITICEBS N GElf 5,000 Dret/100as svavs? 9 165910174 AIRBORNE IJrVEL5 5 AMPLE 9 0910059 1.0 x 10 -10 UCI/CC 0.1DAC 4 LIMITING CosEDITIcess N 5AaEFLE 9 UCI/CC DAC j RADCoef IIDEtITOR COIrTINUQUS N IEFFERIEITTElff Y DURING KT5TEIE ESEACE N , moTIrr MAD com omII.T PRIOR w erART r, .Go wits w N + ENTER UPDATE F1-RWP PART2 F2-RWT PART3 F4-TEMPLATES FS-DRESSOUT F7-NEXT RNP8 F8-RW REQUEST F9-SPEC PREC P10-RWP CLOSE Fil-PRINT RWP P9 F12- N ATTACHMENTS F13-RWP REV F20-REWRN F21-RADCON MENU F22-MAIN F23-SIGNOFF F24-HELP
> Radiation Levels r Contamination Levels r Airborne Levels .
r Rad Con Monitoring Requirements r Limiting Conditions: If marked "Y" (yes), the maximum or bounding radiological conditions for which the RWP is valid will be listed. W WORD 97NNDOUTCORPCEic18CATil DM pggg e 47 COWRGHT M7 GPU Nuclear ins
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Revision 5 l - 06/19/98 Part 3A of the RWP describes the various worker descriptions. The Automated Access System { will always set the ESRD setpoint as d' efined for Worker Type # worker type differs from the Worker Type #1 setpoint, ask Radiological Controls to change your ESRD alarm setpoint accordingly.
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The RWP Dressout screen describes t$c'following information for each worker type; . . - WMRO11 R11 -; 3pfF DRESSOFF RwPG 910001 Nave 1 REQD AVAIL EIr 300 anung SPBCIAL PRECAUTIONS Y
$1 INSPEC'IVR ( WORKER AT VALVE )
DOSIM WM ~ ' ' RNRD DOSE AIARM 250 asa PROT CLOTE: LOW CONTAMINATION RESPs EECEF710IN: IF SYSTEM LEAKAGE IS DETECTED, IMMEDIATELY CONTACT RAD CON AND RAD con WILL PROVIDE CONTINUOUS COVERAGE DURING INSPECTION. WB TLD PLACEMENT ON THE Ot?rSIDE OF CLOTH COVERALLS. 92 UTILITY WORKER ( OUTSIDE ) DO,I., - asaD Dose u. amer 250 sen F1-RWP PART1 F2-PART2 F3-PART3A FS-MORE F9-BACK F10-QUAL LOOKUP-SSN l F9-SPEC PRECAITTIONS F20-RE'IVRN F21-RWP KENU F24-HELP r' The Minimum Protective Equipment Required
> Dosimetry Requirements
! k ESRD Alarm Setpoints l k Protective Clothing / Respiratory Protection Equipment Required: One of the following standardized protective clothing sets will be specified:
> none 9
l k partials
> low contamination
'r high contamination
'r very high contamination / wet The exact protective clothing items for each standardized set will be listed on a chart posted at the dressout area. Rad Con may grant exceptions to these items in reference to the protective clothing standard for the RWP.
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( 'r Pre Job Briefing / Discussion Requirements The RWP Special Precautions and Instructions screen describes special requirements and information important to the safe radiological conduct of the work evolution for which the RWP is written. Personnel are responsible for complying with the requirements set forth by this section. This section will also include a summary of each RWP revision. WMROO7 R07 RADIATION WORK FERMIT SPECIAL FRECAUTIOWs Jo#/ENPS 910001'
... ****== sr:CIAL FRacAuTIOms AuD InsTRDcTIOss ..........
- 1. ESRD DOSE ALARM SETPODrr IS 250 MR.
- 2. RAD CON WILL PROVIDE MONITORING AT THE START OF THE JOB AND THEN DrrERMITTENTLY THEREAFTER.
- 3. NO BREACH OF CLOSED SYSTEM IS PERMITTED. IF LEAKAGE IS DETECTED THEN NOTIFY RAD CON IMMEDIATELY.
l l F1-RWP PART1 F2-PART2 F3-PART3A F8-MORE F9-BACK F10-QUAL LOOKUP-SSN F9-SPEC PRECAUTIONS F20-REWRN F21-RWP MENU F24-HELP 3 By logging into the REM System using any method (auto access, manual, CPAT), the worker is acknowledging that they have read, understand, and will comply with the radiological requirements specified for the work to be performed as outlined within the Radiation Work Permit. No signature is required. Attachment sheets are legal documents and all information entered must be in black ink only. LO: State the worker's responsibility for complying with RWP requirements. Plant managers and supervisors expect full compliance with all RWP requirements. Failure to comply with the requirements established by an RWP may result in a radiological event or problem. This could result in unnecessary exposure to personnel, the plant being fined or other regulatory action, and/or possible disciplinary action against the worker. WW N4#CWW WEMWhW PaQO49 TuE/*?!
-- - .- - ~ - - - - - - _ _e__ _2 . -A Revision 5
' 06/19/98
( LO: Given a completed survey map, be able to state the following information: l k survey numtier > ' maximum radiation level and location , k date/ time > work area radiation level
- k. location / reason > . strength and location of hot spots (if applicable)
T " air sample 'data- , 1 >" method of recording contact dose rates, " ;
~ ' ^
> low dose waiting area ' .
general area dose rates, smear locations, and k maximum contamination level beta and neutron radiation levels
> work area contamination level 4
)
The survey map generally shows a drawing of the area covered by an RWP and includes information on dose rates, contamination levels, and other radiological concerns. While survey j maps can vary between Three Mile Island and Oyster Creek, they contain the same basic information. - Survey information instrunent Data Asr Sample Survey Nun 6er: 961234 Contananation Suryoy Redantson Survey Date Tinn Smeerable Contansnauon L*"*"*" W AcGyl y
. W RMM M Locati n 6K ] Commente Air Sample No 10 gg g gy gag g,y
(' CetDue_g11MCatDue b TimM Col Due k1M Cal Due Notee: Rediatione in n@h 3 W 90M Tech Boe Smem EN G EN MM done rates are general eres. Skg 30 CPM l pp g, 8ka fech Tom Mvers and weitansnat6on reautto in Reviewed St M Jones Tech Harev Thomoson Notes: dpnv100 cac unnees Ceaa asemaes Cir=w eihen e.n.ed. C e Smear Locat6on l l 4R 1.5R 2 6008 ; WTP-A 9R RWT-A C C RWT-B nome,se. 2.5R j 180 I ( Located on the survey map is, for example, the RWP number, date of the survey, area of the plant surveyed, and the purpose of the survey. Also on the survey map is a drawing of the area w womo.wcouTammmeC4Tii D* Page 50 Tut *,s'a,%
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~{ surveyed as well as radiological information such as dose rates, contamination levels, and radiological postings.
Symbols and/or designators are used to. identify the different types ofinforpution contain the survey map. ) A smear location is shown by a numbered square. For example, the figure at the right would indicate on the area drawing where smear number I was taken. i 1 The activity of the smear is entered under location 1 of the smearable contamination column. The symbol "K" is used to represent thousand(s). For example,20K represents 20,000 DPM/100cm2,
> Dose rates are indicated by the following:
r A number by itself indicates a general area gamma dose rate in units of mR/hr. For example, '10' would indicate a general area gamma dose rate of 10 mR/hr. 4 ' r A circled number indicates a contact gamma dose rate in units of mR/hr. For example, the figure at the right would indicate a contact gamma dose 6
- rate of 6 mR/hr.
> A number followed by either an "R" or an "R/hr" indicates dose rates in the R/hr range.
r Dose rates of radiation other than gamma radiation are noted by a number followed by a B, for beta radiation. or a N, for neutron radiation. For example, '4 B' indicates a beta dose rate of 4 mrad /hr while '15 N' indicates a neutron dose rate of 15 mrem /hr. Unless otherwise noted, beta radiation dose rates are in units i of mrad /hr and neutron radiation dose rates are in units of mrem /hr. l LO: State the required action (s) to be taken if the work scope or radiological conditions change such that they are no longer within the scope of the applicable R W P. Radiological conditions can change with reactor power level, equipment status changes, movement of shielding, and other reasons. If radiological conditions are different than expected, or if the conditions change unexpectedly, inform others that may be in the area, exit the area, and contact Rad Con. Or if, while working on a task in an RCA, the job scope changes, such as having to completely disassemble a component instead ofjust part ofit, contact Rad Con before
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proceeding. LO: State the procedures for entering and exiting an RWP, including: r Automated Access ( 'r Manual Access r' Control Point Admission Ticket W WN97NNDOUTCORPGETC18CATil DM pggg $j COP Md 4, w,; , ; - n+s
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,, w, C For most RWPs, workers use the Automated lAcceis (Auto Access) system to log themselves onto the REM-On-Line System thMugh the'folloWingpfoe'sdure:
c, .m. w+y : E An ESRD is picksd up' fro'm the'sforage area at the Control Point and placed on the PDR-1 reader according to the diagram on the reader.. The PDR-1 reader will ; check the ESRD for battery condition and calibration data.' If the reader reports a problem,with the ESRD, notify Rad Con and selec't another ESRD.
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> At this point, the worker must have already read (and fully understand) the RWP.
Workers must fully comply with all RWP requirements. 1 k Following the direNs nNeompuler screen, the worker will be asked to input their Social Security Number (SSN), RWP Number and any respirator information, if required. This information may also be scanned in using the bar f code reader. By entering their SSN, the worker acknowledges that they are the V individual logging into the REM System and that the ESRD was issued to them. Upon entering the SSN, the worker's qualifications will be verified with the REM System. The worker's information is retrieved and the worker's name is displayed on the computer screen. Any warning or error message that applies will appear in the appropriate message box. The worker's SSN is stored in the ESRD memory for worker accountability and tracking { Purposes. Upon entering the RWP Number, the system checks for specific RWP requirements such as the Required Available Exposure, current WBC date, and current training date.
> The RWP revision number, work description, and respiratory protection status is displayed. At this point the worker is prompted to determine if the displayed information is correct, then enter "Y" or "N" as appropriate.
By answering "Y", the worker states that the information entered above (name, RWP, respiratory protection status) is correct. A third option at this point allows the worker to correct the respirator information (as required) by entering an "R". If an "R" is entered, the worker will be prompted to enter the Respirator Code (e.g.,01 for Full-Face Negative Pressure Purifying) and the serial number on the respirator facepiece. If a respirator is required by the RWP, the worker should have this device at hand prior to using Auto Access so that the serial number is readily available. A respiratory protection device cannot be worn unless GPU Nuclear's qualification requirements are met for the specific device and fit type. , ,
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4 ( , e if respirator information was corrected, the worker will be prompted again to verify that all information is correct ("Y" or "N"). Respirator qualifications will be checked by the REM system, including
- medical exam, fit test, and the completion of training requiremerits'for the
.N~ ' 1 selected respirator; T -
n , ' k if all worker qualification requirements are met for the selected RWP',!the system
> vill allow entrance to the RWP area and prompt the worker to remove the ESRD from the reader, If Auto Access does not allow entrance, notify Rad Con. Before picking up the ESRD, workers should review the ESRD alarm setpoints and personal dose information displayed by the system.
When exiting an RWP using the Auto Access procedure, workers must log themselves out of the REM System through the following procedure; wp
'r The worker places the ESRD onto the PDR-1 reader according to the provided diagram. Auto Access will then read the name stored in the ESRD and prompt the worker to confirm ("Y" or "N") that the correct name is displayed. If an incorrect name is displayed, notify Rad Con immediately.
'r if the name is confirmed to be correct, Auto Access will read the ESRD to obtain
(_, dose information to update the REM System. k The worker is then prompted to remove the ESRD. Before picking up the ESRD, workers should review the ESRD dose displayed by the system. The ESRD must be returned to the designated storage location. E If Auto Access is not available or(for selected RWPs) Auto Access is not permitted, workers will enter and exit RWPs using the Manual Access Control System. r To sign onto an RWP, workers report to a Control Point where a Rad Con Technician or Control Point Technician will process their RWP entry.
. Workers are required to state their name, SSN, and RWP Number.
The Rad Con or Control Point tech will verify worker qualifications (including respiratory protection if required) using the REM system. The Rad Con or Control Point tech will then issue an ESRD (and/or other dosimetry as required). I w wmomwoourcmmerescaru om Page 53
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e To exit an RWP using the Manual Access Control System, workers'must report to
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a Rad Con or~ Control Point tech at a Control Point. i m
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i ,.. Workers are required to state their SSN and the dose reading on their , > ~ - aN. j ESRD'or SRD.' f ' '. 1
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7., e . ew(( ~. m.. a my w!.3;G,% %g;, M M k. > The Rad Con or Control Point tech will enter the dose reading into the '
' REM system and prepare the ESRD for re-use.-
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j If the REM System is not available to record RWP entry and exit information, workers can be processed by the rnanual entry method using a Control Point Admission Ticket (CPAT). , ! r ~ For entry, the worker's name, SSN, RWP#, date and time of entry, and initial. je l ESRD/SRD dose reading are recorded on the CPAT (along with respirator 1 jg@hl i training code and serial number as required).. 3, /?fC T The Rad Con or Control Point tech will then issue an ESRD (and/or other
- dosimetry as required).
r For exit, the worker must report to the same Control Point used for entry. e The worker's ESRD/SRD dose reading is recorded, along with the date (',: and time of exit.
. The ESRD is returned to the storage location.
- The data from the CPAT is entered into the REM system when it becomes available.
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