ML20210N863
| ML20210N863 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/12/1997 |
| From: | Moody B UNION ELECTRIC CO. |
| To: | Bundy H NRC |
| Shared Package | |
| ML20210N765 | List: |
| References | |
| NUDOCS 9708260150 | |
| Download: ML20210N863 (15) | |
Text
. _ _ _ _ _ _.
liruce Moody i
Union Electric Callaway Plant P.O. Ilox 620 Fulton, Mo. 65251 June 12,1997 l
Mr. Iloward 11 undy U.S. Nuclear Regulatory Commisslor 611 Ryan Plaza Drive, Suite 400 Arlington, Tex. 76011 8064
Dear floward:
1his letter accompanies material discussed via telcon on June 12 concerning the Callaway RO Retake Exam scheduled for June 27,1997. The following material is enclosed with this letter per your request:
j 1)
Revised reference sheets for questions 1,$,6,10,16,17,20,32,66 and 93 to reflect changes as i
discussed.
2)
Revised examination key.
3)
Lesson plan objectives to support questions 6 and 11, 4)
Revised Outline to reflect replacement question 5, 5)
A copy of the exam as it will be administered.
If you have any ques' ions conceming this material, or if additional information is needed to support your exam review, pleste call me at 573-676 8194, Sincerely, MT llruce Moody 9708260150 970020 PDR ADOCK 05000403 V
QUESTION #1 A Safety injection has occurred due to a Steam Generator Tube Rupture. The crew has just completed verifying that both NB01 and NB02 are energized per Step 3 of E-0, Reactor Trip or Safety injection. Which one of the following describes the HVAC flowpath for the Fuel Building at this time?
A.
Fuel Building supply and normal exhaust stops; emergency exhaust dampers align to the l
Aux Building.
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B.
Fuel Building supply and normal exhaust stops; emergency exhaust dampers align to Fuel and Aux Building.
C.
Fuel Building supply keeps running or starts; normal and emergency Fuel Building exhaust isolates.
D.
Fuel Building supply and exhaust keeps running or starts; emergency exhaust dampers align to the Fuel Building.
ANSWER:
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A. Fuel Building supply and normal exhaust stops; emergency exhaust dampers align to the Aux Building.
VJA #: 013K113 2.8/3.1 KA DESCRIPTION: FB Ventilation on SIS OBJECTIVE #:
0110390D
REFERENCES:
T61.0110.6 LP 39, Page 40 AUTHOR: RBM SOURCE: BANK Modified Y - L DISTRACTER EXPLANATION:
Response B is incorrect because the emergency exhaust does not align to the Fuel Bldg. Response C is incorrect because the Fuel Bldg supply does not keep runnin0. Response D is incorrect because the Fuel Bldg supply and exhaust does not keep running and the dampers do not align to the Fuel Bldg.
RO Outline #32
QUESTION #5 i
1 The following plant conditions exist:
- Reactor power is 1 x 10E-8 amps
- A reactor startup is in progress 1
- N-35 Intermediate Range channel fails LOW Which ONE of the following actions is correct per Technical Specification 3.3.1, " Reactor Trip System Instrumentation *?
A.
Restore inoperable channel prior to exceeding P 6.
B.
Restore inoperable channel prior to exceeding 5% power.
C.
Restore inoperable channel prior to exceeding 10% power.
D.
Bypass the inoperabie channel and continue with reactor startup.
l ANSWER:
B. Restore inoperable channel prior to exceeding 5% power.
K/A #: 015A201 3.5/3.9 KA DESCRIPTION: NIS Faliure OBJECTIVE #:
003B500C
REFERENCES:
T61.0038.6 LP-#50 AUTHOR: RBM SOURCE: BANK DISTRACTER EXPLANATION:
A.
Alpha is not correct since power is already above P-6.
C.-
Would be correct if power was above SE D.
Delta is incorrect because both intermediate ranges are required to perform a reactor startup.
RO Outline #51
i QUESTION #6 i
The plant is operating at 100% power with all equipment in its normal lineup. A spurious Si occurs during some I&C testing. All equipment functions as designed.
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Which ONE of the following procedures will the CRS transition to upon completing ES-1,1, Si Termination?
A.
OTG-ZZ-0001 A, Shutdown Bank Withdrawal 1
B.
OTG ZZ-00002, Reactor Startup C.
OTG-ZZ-00005, Plant Shutdown 20% Power to Hot Standby.
D.
OTG-ZZ-00006, Plant Cooldown Hot Standby to Cold Shutdown ANSWER:
- C. OTG-ZZ-00005, Plant Shutdown 20% Power to Hot Standby K/A #: E02EA2.2 3.5/4.0 KA DESCRIPTION: Procedural Guidance Following SI Termination O'3JECTIVE #:
003D090G
REFERENCES:
ES-1.1 AUTHOR: PJM SOURCE: NEW L i
DISTRACTER EXPLANATION:
Upon a Reactor Trip and Safety injection, E 0 will be entered. ES-1,1 will be transitioned to from there.
Upon completion of ES 1,1, the guidance is to perform OTG-ZZ-00005 or 8.
A. Incorrect B. Incorrect C. Correct.
D. Incorrect RO Outline #87
i QUESTION #10 The crew is responding to a plant transient and are currently in procedure ECA 1.2, 'LOCA Outside Containment".
Why should operators wait some amount of time during each valve repositioning per this procedure?
A.
Prevents overcurrent trips on va!ve motor breakers.
B.
Allows system pressure to respond to repositioning.
C.
Prevent valve motor overheating due to excessive operation.
D.
To allow check on indications of leak in auxillary building.
ANSWER:
B. Allows system pressure to respond to repositioning.
l K/A #: E04EK1.2 3.7/4.0 KA DESCRIPTION: Precaution During Valve Strokes in ECA 1.2
~
OBJECTIVE #:
003D140D
REFERENCES:
T61.003D.6 LP 14 1
AUTHOR: FXB SOURCE: BANK - Modified N - L -
DISTRACTER EXPl.ANATION:
A.
Breakers overcurrent trips are jumpered C.
Valve motor overheating is not a concem.
D.
No remote indication required, but note on page 2 has personnel searching RO Outline #86
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4 QUESTION #16 The plant is at 8% power preparing to synchronize the main generator to the grid when the I
running main feed pump trips. As the Balance of Plant operator you observe the following steam generator narrow range levels:
A indicates 16%
B indicates 15%
C indicates 14%
D indicates 17%
All other plant parameters are normal.
i j
Which one of the following correctly describes the status of the Auxiliary Feodwater (AFW)
System?
)
A.
No AFW pumps are running l
B.
Only the turbine driven AFW pump is running O.
Only the motor driven AFW pumps are running D.
All the AFW pumps are running ANSWER:
C.- Only the motor driven AFW pumps are running K/A#: 054AA2.03 4.1/4.2 KA DESCRIPTION: AFW PP Start Signals OBJECTIVE #:
01102SOH
REFERENCES:
T61.0110.6 LP 25 OTO-SA 00001 AUTHOR: RBM SOURCE: NEW-HO DISTRACTER EXPLANATION:
All answers _are plausible, depending on whether a MDAFAS or TDAFAS signal has been actuated.
With the given conditions, only a MDAFAS signal would be generated, therefore C is the correct answer.
RO Outline #84
QUESTION #17 The following conditions exist:
. PRZR Relief Tank Level Hi/Lo -- ALARMING on HIGH LEVEL j
. PRZR Rollef Tank Pressure
- ALARMING on HIGH PRESSURE Which ONE of the below cornbinations contain sources, ALL of which should be monitored for leakage into the PRT7 A.
RHR Pump Suction Reliefs (EJ8708A/B), RCP SealI.eakoff Relief (BG8121), and CVCS Letdown Rollef (BG8117).
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B.
ECCS Accumulator Rollefs (8855A-D), RHR Pump Suction Rollef (EJB708A/B), and CVCS Leidown Relief (BG8117).
C.
RCP Seal Leo'.off Relief (BG8121), CVCS Letdown Rollef (BG8117) and RHR Dischorge Reliefs (EJ8856A/B).
D.
Safety injection Pump Suction Rollefs (EM8858A), RHR Pump Suction Rollefs (EJ8708A/B), and RCP Seal Leakoff Rollef(BGB121).
ANSWER:
A. RHR Pump Suction Reliefs (EJ8708A/B), RCP Seal Leakoff Relief (BG8121), and CVCS Letdown Relief (BG8117).
K/A #; 007A205 3.2/3.6 KA DESCRIPTION: Impact of Pressure i on PRT OBJECTIVE #:
01100901
REFERENCES:
M22BB02 AUTHOR: FXB SOURCE: NEW-HO DISTRACTER EXPLANATION:
B-ECCS Reliefs go to atmosphere C-RHR Discharge Reliefs go to RHUT D-Safety injection Suction Relief to RHUT RO Outline #38
l QUESTION #20 Callaway Plant is in Mode 4. "B" RHR is in service. A plant cooldown is in progress. The Reactor Operator is directed to stop the cooldown. EGHV102, 'B' CCW to *B' RHR heat exchanger is CLOSED.
Which ONE of the following events occur?
A.
B.
- B' ESW flashes in the "B" CCW heat exchanger causing water hammer in the "B" ESW.
C.
D.
- B* ESW flashes in the "B" CCW heat exchanger causing the heat exchanger tube side relief valve to lift.
ANSWER:
A. *B' CCW flashes in the *B' RHR heat exchanger causing the "B" CCW surge tank level to increase.
K/A #: 005A103 2.5/2.6 KA DESCRIPTION: Flashing in RHR HX l
OBJECTIVE #:
0110100E
REFERENCES:
SOS 86-0054 AUTHOR: PJM SOURCE: BANK Modified Y - HO DISTRACTER EXPLANATION:
A. When CCW flow was stopped in the plant to the RHR heat exchanger, CCW flashed to steam in the RHR heat exchanger upon heating up.
B. ESW cools CCW. ESW is at a higher pressure so shce!d not flash.
C. B RHR is pressurized so should not flash.
D. ESW cools CCW. ESW is at a higher pressure so should not flash.
RO Outline #27
~
1 I
I l
' QUESTION #32 Which one of the following posted areas must be deposted before plant personnel may be permitted to enter?
A.
CHRA, Caution High Radiation Area B.
-DHRA, Danger High Radiation Area C.
DREA, Danger High Radiation Area Radiological Exclusion Area D.
VHRA, Very High Radiation Area ANSWER:
D. VHRA, Very High Radiation Area K/A #: 2.3.1 2.5/2.9 KA DESCRIPTION: Radiological Posting Requirements OBJECTIVE #:
003A310D 003A310F AUTHOR: RBM SOURCE: NEW L DISTRACTER EXPLANATION:
All of the areas listed have special entry requirements. The entry requirement for a VHRA is that it must be deposted before entry can be authorized.
RO Outline #96
4 QUESTION #66 Which one of the following conditions would require action to be taken within 15 minutes to avoid violating the plant Technical Specifications?
A.
The plant is at 40% power with AFD outside its specified target band.
B.
The plant is at 2% power with Tavg at 550'.
C.
The plant is at 40% power when 120-volt AC bus NN01 loses power.
D.
The plant is at 2% power when SR channel N31 fails.
ANSWER:
B. The plant is at 2% power with Tavg at 550'.
K/A #: 2.1.11 3.0/3.8 KA DESCRIPTION: TS 3.1.1.4 - Minimum Temp for Criticality OBJECTIVE #:
003A03OC
REFERENCES:
TS 3.1.1.4 AUTHOR: RBM SOURCE: NEW-HO DISTRACTER EXPLANATION:
A would be a 15 min response if above 50% power.
B is a 15 min response.
C requires action within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
D would require immediate action if below P-6.
RO Outline #88
QUESTION #93 A plant startup is in progress. Power level is 1E-7 amps. The Reactor is tripped when PA02 is doenergized by Relay Test.
Which ONE of the following is the approximate length of time before the Source Range NI's automatically energize ?
A.
2 minutes 1
B.
5 minutes 1
C.
10 minutes l
1 D.
15 minutes ANSWER:
C.10 minutes
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K/A #: 007EK1.05 3.3/3.8 KA DESCRIPTION: How Long for Source Ranges to Energize on Rx Trip l OBJECTIVE #:
003D060E l
AUTHOR: PJM SOURCE: NEW - LO DISTRACTER EXPLANATION:
SUR following a Reactor trip is -1/3 DPM.
Source Range NI's automatically energize at 6E 11.
1E 7 to 6E-11 is 3.4 Decades 3.4 Decades at.33 DPM = 10.3 minutet, RO Outline #76
CONTROL BOARD CERTIFICATION MOD B,
i B 18 OTO BB 00005 REACTOR COOLANT SYSTEM HIGH ACTIVITY
/
i Objectives A. State the chemistry analytical sampling and trending URO SBR 04ll' requirements for response to reactor coolant system high activity.
Procedure Cover latest.evision of procedure with students.
Review Group IIDt!!sfrom EIP ZZ 00101.
Using OTA SP RM011 (attachment l
11, review plant raduation monnotong locations and Jianctions.
I Review TS 3.4.b' Summary OTO BB-00005 provides guidance for responding /
to RCS high activity, Directions are provided to have Chemistry
'; stan sampling and take actions as directed by Tech Specs. I, References A. OTO BD-00005 B, EIP ZZ-00101 C OTA SP RM0ll T61,00386 1 930914
CONTROL BOARD CERTIFICATION MOD D LP 9, ES 1.1, S. l. TERMINATION ) OBJECTIVES NOTE: OlWECTIVES INDICATED WITil AN "*" ARE APPLICABLE TO T61.0520.6. i l Uno-CRK 022 A. STATE Ti1E PURPOSE OF ES 1.1 0000090.12 4.1/4.3 i URo-CRK 02AA 11. STATE Tile TRANSITIONS FROM ES L1 1 000009G.lt DACK TO M. Tile PARAMETERS USED, AND T11EIR DAS!S. URo-CRK 02AB C. STATE Tile PLANT PARAMETERS WillCIi 000009G.lt AIG EVALUATED TO TRANSITION FROM ES 1,1 TO M. INCLUDE TRANSIT.'ON CRITEIUA AND BASIS. URo CRK-02Ac D. GIVEN T11E ColuGCT ORDER OF 0/m0Al.07
- '##'7 PRESSURIZER PRESSUIG CONTROL PER ES 1.1, EXPLAIN T11E DASIS FOR TlilS 3
OIU)ER. URo-CRK-02AD E. DESCRIBE Tile INDICATIONS USED TO 0000/ 7A1.21 4.4/4.3 VEldFY NATURAL CIRCULATION PER ES-1.1. URo-CRK 02AE F. PREDICT WilEN SOURCE RANGE m m EA1.05
- 0'# #
DETECTORS SliOULD BE ENERG17ED FOLLOWING A IGACTOR TRIP. URo-CRK 02AF G. STATE TiiE PROCEDURE TI-IAT SliOULD BE 0000llEA2.03 TRANSITIONED TO ON COMPLETION OF ES-1.1. URo-CRK-02AG 11. DESCRIBE IN CORIGCT ORDER T11E OllMA2 01 3.2n.1 MET 110DS OF ESTABLISli!NG EXCESS LETDOWN PER ES 1,1. URo-cRK-02AH 1. EXPLAIN Tile CAUTION AND BASIS FOR 013000Kl.02 3.23.6 RESETTING S1 PRIOR TO STARTING RCP "C" 9g,9 URo CRK 02Al J. DESCRIDE llOW ESW IS SUPPLIED TO TliE 079000Kl.02 } 2.2/2.2 AIR COMPRESSORS AFTER A LOSS OF INSTRUMENT AIR. T61.003D.6 1 930901 MOD 961107 \\
ES-401 Ca!!away RO June 1997 RC> Examination Outline Form ES-401-4 ~ Plant Systems -Tier 2 / Group 1 K K K ^ ^ ^ ^ E/ APE # / Name / Safety Function K/A Topic (s) Imp. O# 3 5 6 2 3 4 001 Control Rod Drive X K565 Power Mismatch Effect on Rod Control 32/3.6 7 X A102 Plant Response to Pimp Failure 3.1/3.4 13 X KSO4 Rod Insertion Limit w/P-A Cony. 4.3/4.7 49 Maffunct 003 Reactor Coolant Pump X A404 Monitor #1 Seal DP on RCP 3.1/3.0 48 X K112 RCS Leak Thermal Barrier Isolation 3.0/3.3 53 004 Chemical and Volume Control X K107 Plant response to Flux Doub!ing 2.6/2.9 3 X K106 Response to VCT Level Channel 3.1/3.1 6 Failure X A227 RWST Operability in Mode 6 3.5/4 2 25 ~ K113 Fuel B3dg Ventilation on a SIS 2.8/3.1 32 013 Engineered Safety Features Actuation X X K201 Downpower cross-trip tWocks ' 3.6/3.8 33 X A402 Reset cf ESFAS Channels 4.3/4.4 47 015 Nuclear instrumentation X K302 Response to Power Range NI Failure 3.3/3.5 8 X K604 Source Range NI Failure 3.1/32 14 X A201 NIS Failure 3.5/3.9 51 017 in-core Temperature Monitor X K401 Incore Thermocouple Inputs 3.4/3.7 21 022 Containment Cooling X A301 Flow to Containment Coolers on SI 4.1/4.3 30 056 Condensate X K103 FeedwaterTemp Reapera to 2.6/2.6 19 Feedwater Heaterisolation 059 Main Feedwater X A403 FRV/ Feed AP on Powerincrease 2.9/2.9 22 X A212 MFP trips from FRV failure 3.1/3.4 44 069 Auxiliary / Emergency Feedwater X K401 AFAS/LSP Actuation Requirements 3.9/42 12 X K404 Flow Control Valve Operations 3.1/3.4 44 068 Liquid Radwaste X K401 Auto Isolation of Radwaste Discharge 3_4/4.1 28 072 Area Radiation Monitoring X A403 Area Rad MON Source Check 3.1/3.1 46 K/A Category Totals: 5 1 1 4 2 1 1 3 1l4 0 Group Puint Total: 23 Target 23 NUREG-1021 Interim Rev. 8. January 1997
7 ,e 4 CALLAWAY PLANT IPE/PRA Referenced to Callaway Plant Retake RO Exam of 6/27/97 Individual Plant Evaluation RO Written Major Event (% CDF) Question # l TOPIC internalFlooding (30.5%) 40 Monitor NK lineup in the Control Room (Effect of Flooding) 57 Immediate Actions for Loss of NN02 (Effect of Flooding) Station Blackout (30.3%) 54 ECA-0.0 Actions for ESW Pump 58 Plant Equip Response during Loss of All AC 80 Natural Circ Cooldown following Loss of All AC l LOCAs(19.0%) 62 Subcooling requirements following a LOCA 72 ECCS water sources available in ECA 1.1 70 _ Decay heat removal following a LOCA in E 1 83 Basis for blocking low Steamline Pressure in ES 1.2 86 Precaution during valve strokes in ECA 1.2 11 RTD failure effects on plant instrumentation Transients (17.7%) 06 Uncontrolled depressurization of all Steam Generators 71 Loss of steam dumps following a loss of condenser vacuum 78' Loss of Secondary Heat Sink (FR-H.1) 87 Recovery from a spurious Safety injection 90 CSF implementation following a steamline rupture 55 Response to SG Overpressure following a SGTR SGTR (1.4%) 80 Preferred order of depressurizing RCS following a SGTR S Basis for Turbine Trip in Response to ATWS ATWS (0.7%) 81 Operation of the AMSAC System l PRA Risk Sianificant Systems (% Contributing) KJ/NE (30%) 2 HVAC operability for DGs 68 Basis for altomating ECCS trains during recovery actions ~ ESW (23%) 54 ECA-0.0 Actions for ESW Pump AFW (12%) 12 AFAS/LSP Actuation Requirements 44 AFW flow control valves operation 84 AFW pump start signals RHR (7%) 11 RHR valve interlocks - 27 Flashing in RHR Hex 70 Air entrainment during RHR operation CCW(4%) 4 Response to CCW rad monitor alarm 10 Source of CCWleakage 60 Immediate Actions for loss of CCW Pump 94 P&L for CCW supply to Radwaste m}}