IR 05000443/1996009
| ML20129H347 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 10/25/1996 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20129H301 | List: |
| References | |
| 50-443-96-09OL, 50-443-96-9OL, NUDOCS 9610310246 | |
| Download: ML20129H347 (70) | |
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I U.S. NUCLEAR REGULATORY COMMISSION 1 REGION l
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Licensee: North Atlantic Energy Service Corporation Facility: Seabrook Station ,
Dates: September 30 - October 4,1996 Examiners: Joseph D' Antonio, Operations Engineer l Steve Barr, Operations Engineer Approved by: Gierin W. Meyer, Chief Operator Licensing and Human Performance Branch l Division of Reactor Safety
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9610310246 961025 PDR ADOCK 05000443 V PDR _
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EXECUTIVE SUMMARY
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Seabrook Power Station NRC Inspection Report 50-443/96-09 Operations
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From September 30 - October 4,1996, two examiners administered initial licensing
, examinations to four senior reactor operator upgrade (SROU) candidates and one senior reactor operator instant (SROI) candidate at Seabrook. This examination had been developed by the facility training department in accordance with NRC guidance, it was then approved, administered, and graded by NRC.
I All candidates passed their examinatio The facility did a good job in developing the examinatio !
In the simulator, communications were generally excellent, STAR practices were evident,
- and the SROs assigned tasks effectivel One weakness was noted in both the simulator and walkthroughs in that candidates had I i
difficulty deciding that " local manual" action was appropriate when " local powered" actions were unsuccessful. Also, the lack of a procedure for a leak in a secondary system )
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within containment appeared to delay effective mitigating action ,
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Report Details
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j l. Operations
- 05 Operator Training and Qualifications (:
05.1 Operator initial Examinations j Egggg l The facility staff developed the written and operating examinations and submitted l these proposed examinations for NRC review and approval. The NRC reviewed the i proposed examinations, provided comments, and approved the examination for
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administration. The examinations administered reflected incorporation of the NRC 4 comments.
i j Observations and Findinas
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j The fecility did a good job of exam development. The NRC had minor comments on i approxmately ten percent of the written exam. One proposed JPM (job
! performance measure) was not suitable as a system JPM, but was used in the i administrative exam instead. Two JPM questions were replaced, and one event
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j Summarv of Results i l'
SRO i Pass / Fail -
i 4 Written 5/O
Simulator 5/O j Walk-through 5/0 l
Overall 5/0 Ooeratina Examinations
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Command and control and overall communications were strong. The crews did a s generally excellent job of ensuring that communications were clearly delivered and l acknowledged. When on the board, good STAR (stop, think, act, review) practices
{ by the candidates were evident. When in the SRO position, the candidates did a j good job of assigning particular tasks to a specific board individual to move activity j along without confusio I
One area of procedural confusion was noted in a station blackout scenario in which i the output breaker for an operating diesel was not able to be closed locally from the diesel control panel. The crew spent several minutes deciding to continue on in the
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procedure, and never did order the breaker locally MANUALLY closed. The examiners expected that this would be attempted; however, the facility makes a distinction between " local" and " local manual" operations. The procedura directs a local, but not local manual breaker closure. This same confusion was apparent when one candidate took an excessive amount of time to complete a JPM requiring local operation of a diesel output breaker. From the time the candidate was directed to locally close the breaker, it took 10 minutes for the candidate to decide that local MANUAL closure was require Also, the examiners noted that in response to a feed leak in the containment, the facility had no procedure for a small secondary leak, in the scenario as run, the crew realized they had a secondary leak, but entered the procedure for a reactor coolant leak since this was the only procedure available with entry conditions matching plant symptoms. After executing this entire procedure with no useful results, the SRO decided to perform the expected action of shutting down the plan The above difficulties, resulting from lack of a procedure or lack of sufficiently explicit guidance, were discussed with the operations manager and training department following the scenarios, who agreed to further evaluate these difficultie ,
i Overall, walkthrough performance was good. The candidates performed the JPM l tasks in a careful manner, checking and verifying the appropriate procedure section and components to be manipuiste Written Examination The following items were missed by three or more of the candidates tested and represented areas of generally weak understanding:
Question # Tooic 2 Conditions for placing RHR in servic Rod control system component operation When to suspend core alteration Response to a dropped ro Automatic actions for radiation monitor alar i 66 EDG response to SI and loss of offsite powe Technical Specification immediate action Radiation Work Permit Containment purge requirements.
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Review of UFSAR Commitments A recent discovery of a licensee operating their facility b a manner contrary to the updated final safety analysis report (UFSAR) descripti' .i highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the activities discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to selected examination topic The NRC reviewed facility 18-month equipment response time surveillance procedures to verify that ECCS equipment loading times were consistent with the times listed in section 6 of the UFSAR. The inspectors verified that the loading sequence and times were consistent with the UFSAR statement Conclusions The candidates were well prepared and generally performed well. All five candidates passed the examination ars were issued license V. Mansaament Meetinas XI Exit Meeting on May 10,1994
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The NRC expressed its appreciation for the facility examination development, validatien efforts and facility accommodation of the needs of the examination process. Generic l strengths and weaknesses observed in the operating examinations were discusse l The following key facility personnel attended the exit meetin l l
J. M. Grillo Operations Manager l Tom Grew Training Manager D.Roy Training Supervisor L. Carlsen Training Supervisor T. Cassidy Sr. Instructor Attachments:
1. Written Examination and Answer Keys 2. Simulation Facility Report i
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e ATTACHMENT 1 WRITTEN EXAMINATIONS AND ANSWER KEYS
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NRC Official Use Only
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-i Nuclear Regulatory Commission Operator Licensing Examination ;
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f-This document is removed from Official Use Only category on date of examination l i
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NRC Official Use Only
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U.S. NUCLEAR REGUIATORY CObe(ISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1 CANDIDATE'S NAME:
FACILITY: SEABROOK REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 9/30/96 INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE %
100 POINTS % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received ai .
Candidate's Signature i
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- NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS 1 J
During the administration of this examination the following rules
- PP lY: l d Cheating on the examination means an automatic denial of your
- application and could result in more severe penaltie ,
, After the examination has been coupleted, you must sign the statement on the cover sheet indicating that the work is your own
- ] and you have not received or given assistance in completing the examination. This must be done after you complete the examinatio . Restroom trips are to be limited and only one applicant at a time
- may leave. You must avoid all contacts with anyone outside the ;
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- examination room to avoid even the appearance or possibility of cheating.
! Use black ink or dark pencil ONLY to facilitate legible
- reproduction . Print your name in the blank provided in the upper right-hand
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corner of the examination cover sheet and each answer sheet, a
a Mark your answers on the answer sheet provided. USE ONLY THE
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PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
s 1 Before you turn in your examination, consecutively number each
! answer sheet, including any additional pages inserted when writing your answers on the examination question page.
i If the intent of a question is unclear, ask questions of the examiner ONLY.
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j When turning in your examination, assemble the completed
- examination with examination questions, examination aids, and answer sheets. In addition, turn in all scrap pape ]
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! 10. Ensure all information you wish to have evaluated as part of your j answer is on your answer sheet. Scrap paper will be disposed of ;
- immediately following the examination.
j 11. To pass the examination, you must achieve a grade of 80% or s greater.
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12. There is a time limit of four (4) hours for completion of the j examination.
4 13. When you are done and have turned in your examination, leave the
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examination area (EXAMINER WILL DEFINE THE AREA). If you are i found in this area while the examination is still in progress,
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your license may be denied or revoke i . _ _, _ - _ _ _ _ - - _ _ - _ . _ . . - - . _ . . _ _ _ _
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QUESTION: 001 OS1202.05, " Reactor Coolant System High Activity", directs the operator to maxmuze letdown j flow to reduce activity levels during a high RCS activity conditio Under which of the followmg conditions would increasing letdown flow be most effective? High RCS activity is caused by fission products from a leaking fuel assembl b. High RCS activity is due to non-soluble radioactive gasse High,RCS activity is caused by high tritium level d. High RCS activity is caused by high levels of non-ionic contammant :
QUESTION: 002 I
. l PLANT CONDITIONS: l l
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= MODE 4 e 'Ihe crew is preparing to place RHR in service '
. RCS Tavg is 349'F e RCS pressure indication on PT-405 is 375 psig l
. RCS pressure indication on PT-403 is 360 psi l
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Which of the following statements describes the ability to place RHR in service for cooldown? Both trams may be lined up for RHR cooldow b. RHR train 'A' may be lined up for cooldow RHR train 'B' may be lined up for cooldow d. Neither train may be lined up for cooldow l l
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QUESTION: 003 fLA_NT CONDITIONS:
- Operating at 90% power
- 'Ihe operator receives a DRPI rod position deviation alarm for ROD H * 'Ihe crew determines that rod H8 has become W%M 14 steps below the rest of CB The crew is attempting to realign rod H8 per OS1210.06," Misaligned Rod" Which of the following actions will freeze the Bank Overlap Unit at its present count? Placing the Rod Control Selector Switch to the CB D positio Placing the Pulse to Analog converter AUf0-MAN switch to MANUA I Placing the unaffected CB D rods in the ' DISCONNECT' positio Depressing the Rod Control alarm RESET Pushbutto l l
l QUESTION: 004 Reactor power is increasing. Which of the following describes the operation of Control Interlock C-27 Energizes when 1 of 2 Intermediate Range detectors indicates a current equivalent to 20%
power. Inhibits Rod withdrawal in Manual or Auto b. Energizes when 1 of 4 Power Range detectors indicates 103%. Prevents Rod withdrawal in Man or Auto Energizes when 1 of 2 Intermediate Range detectors indicates 20%. Prevents Rod withdrawl in AUTO onl d. Energizes when 1 of 4 Power Range detectors indicates 103%. Prevents Rod withdrawl in AUTO onl Page 2 of 54
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QUESTION: 005'
Reactor power is 6% during a shutdown when intermediate range channel N-36 fails HIGH.
Which of the following statements describes how this failure affects the reactor shutdown and subsequent operatum of the Nuclear Instrumastatum Systan? De reactor will trip on high IR flux, and source range NIs will reemrgize when N-35 decreases to the proper setpoin b. He reactor will trip on high IR flux, and source range Nis will have to be manually re-energized. De reactor will not trip, and source range NIs will re-energize when N-35 decreases to the proper setpoint.
d. The reactor will not trip, and source range Nis will have to be manually re-energize I
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QUESTION: 006
'Ihe followmg plant conditions exist:
- e i A rupture in the piping downstream of the Chargmg Cold 14 njection Valves (SI-V138/139)
has w . 4 j = 'Ihe check valves on the piping e=-> - to the RCS have faded causing a LOCA into the ,
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Caa*=:- pnion area of the PAB
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j e 'Ihe Reactor has tripped and Safety Injection Las =ce==*ad on low PZR Pressure
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= 'Ibe Crew has entered E-0 and has completed the I=M *a Actions l
. RCS pressure is 1780 psig and decreasing j = Area a xi airborne high radiation alarms have acin=*ad in the PAB
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- Assunung plant conditions do not signi6cantly change arxl the leak is unisolable, what is the
expected flow path through the EOP Network for this accident?
l E-0, " Reactor Trip or Safety Injection" to E-1, " Loss of Reactor or Secondary Coolant" to j ES-1.2," Post LOCA Cooldown & Depressurization"
b. E-0, " Reactor Trip or Safety Injection" to E-1, " Loss of Reactor or Secondary Coolant" to
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ECA-1.2, "LOCA Outside Contamment" to ECA-1.1, " Loss of Emergency Coolant i j Recirculation". l
I E-0, " Reactor Trip or Safety Inj--tion" to ECA-1.2, "LOCA Outside Contamment" to E-1, ,
I i " Loss of Reactor or Secondary Coolant" d. E-0, " Reactor Trip or Safety Injection" to ECA-1.2, "LOCA Outside Containment" to ECA-j 1.1, " Loss of Emergency Coolant Recirculation".
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j; QUESTION: 007 ;
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3 A plant event has resulted in implementation of ECA-2.1 ' Uncontrolled Depressurization of All j Steam Gmerators".
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j Whde attemptag to control the RCS ra Av. kring this procedure, the operator throttles EFW l
- flow wiuch results in a RED path on the Hon. Sink Critical Safety Function
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- What action should be taken? Raise at least one stram generator narrow range level to greater than 5%, then continue with ,
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j b. Increase EFW flow to 500 gpm to clear the " RED" condition and contmue on in ECA- j i ! Increase EFW flow to 500 gpm to clear the " RED" condition and transition to FR-H.1, i " Response to Loss of knad=y Heat Sink".
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- QUESTION: 008 l
l Due to a Small Break LOCA, a Plant Trip and Safety Injection has occurred.
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- The following conditions exist
- All automatic equipment responds as expected e Containment pressure is 3.2 psig and slowly increasing
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j e Contamment High Range RaAntion Monitors RM-RI-6576-A & B read ~ 3 R/hr j e RCS pressure is 1750 psig and slowly decreasing i
e Subcooling margin is 32 degrees F and slowly decreasing e Pressurizer level is 22% and slowly decreasing l
i Assunung conditions do not signi6cantly change, in which of the followmg procedures would you
- expect to be directed to stop one charging pump?
- In ES-0.2, " Natural Circulation Cooldown".
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b. In E-1, "Ioss of Reactor or Secondary Coolant".
] In ES-1.2. " Post-LOCA Cooldown and Depressurization".
d. In ES-1.1,"S1 Termir.ation".
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QUESTION: 009 PLANT CONDITIONS
- 'Ihe crew is respondmg to a small break LOCA e One charging pump has been stopped
. All other high head ECCS pumps are running
. No RCPs are runmng
- Pressurizerlevelis 30%
Sufficient subcooling exists and the w4ur stops the "A" Safety Injection pum Immadataly aAer stopping the pump, RCS pressure and PZR level begin to decreas What acten should the crew take in respense to the decrease in RCS pressure and PZR level? Immediately restart the "A" Safety Injection pump to restore RCS pressure to its previous valu b. Immediately reinitiate Safety Injectio Monitor RCS subcooling, and pressurizer level to assure these parameters stabilize above their ECCS reinitiation values before continuing with ECCS flow reductio d. Monitor RCS pressure and subcooling. If they stabilize above their SI reinitiation values, place normal Charging and Letdown in service, i l
QUESTION: 010 ,
A small break LOCA has occurred, all RCPs are runmng, and the operating crew is in ES-1.2,
" Post-LOCA Cooldown and Depressurization." An RCS cooldown has been initiated by dumping steam to the <= dance i Which of the followmg statements describes the optunum reactor coolant pump configuration, and the basis for this configuration? All RCPs should be stopped to mimnuze RCS inventory loss followmg break uncovery, and mmmuze heat input to the RC b. Only one RCP should be run to allow for a normal RCS cooldown and provide PZR spray, yet i muunuze RCS heat inpu Only one RCP should be run to produce effective heat transfer and RCS pressure control, yet muunuze RCS inventory los d. Two RCPs should be run to ensure symmetric heat tmnsfer to the intact SGs, to enhance RCS pressure control, and to prevent steam voiding in the reactor vessel head on the subsequent RCS depressurizatio Page 6 of 54
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i QUESTION: 011 Which of the following describes the operation of the Emergency bus fnt kysl undervoltage protection scheme? normally e s d undervoltage relays. When one of 2 relays sense bus voltage less than 70% of nominal for 1.2 senande (RAT available), it initiates a sequence ofload stripping and subsequent bus reenergization by the D ;
b. 2 normally energized undervoltage relays. When bus voltage drops below 25% of nommal, they deenergize, initiating auto closure of the RAT supply breake normally w asundervoltage relays. When both relays sense bus voltage less than 70%
of nonunal for 1.2 seconds (RAT avadable), they imtiate a sequence ofload stripping and subsequent bus ;-sh by the D ! normally energized undervoltage relays. When both relays sense bus voltage less than 95%
of nonunal coincident with an SI existing for greater than 10 seconds, they initiate a sequence .
ofload stripping and subsequent bus reenergization by the D .
QUESTION: 012 PLANT CONDITIONS: ,
e MODE 5, e RHR train 'A' is in service e RHR letdown is in service
. T% is 190*F e Pressurizer is soli * RCS pressure is being controlled at 300 psig by the use of the Letdown backpressure control valve,(CS-PCV-131)in AUTO A bus fault causes a loss of 120 VAC Vital Instrument Panel PP-lE Before any operator action is taken, what is the effect on plant operation? RCS temperature will decrease, and RHR letdown flow will increas ,
b. RCS temperature will increase, and RHR letdown flow will increas RCS temperature will decrease, and RHR letdown flow will decreas d. RCS temperature will increase, and RHR letdown flow will decreas ,
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QUESTION: 013 4 De plant is at 80% power, steady stat Which of the following describes the effect that a loss of 125 VDC bus 11 A has on DG-1 A7 l less of DG-1 A breaker remote open/close capability from the Control Room only.
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i b. loss of all engine protective tripping capabi:ity.
j less of all normal and emergency engine start circuit d. Loss of all engine operating parameter indication.
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j QUESTION: 014 i
l He plant is in MODE 6, Refueling is in progress.
i Under which of the following circumstances would CORE ALTERATIONS be suspended?
l j Both doors of one personnel airlock are opened by a dedicated door operator during personnel j entry into Containmen b. De equipment hatch is in place but not bolted.
l A manual containment isolation valve is open in preparation for Containment leak testing. ne
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responsible Test Engineer is stationed at the valve.
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i d. All CAP valves are open for refueling pge operations.
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QUESTION: 015 he plant is at 100 % power, all control systems are in their normal automatic shgnmen De followmg sequence of events occur:
+ Control Bank D rods begin to step ou * After checking board indicatas, the operator places the Rod Control Selector Switch to MANUA * All Control Bank D rods stop moving with the excepten of one Control Rod, which continues to ste Which of the following actions should be taken?
i l Place the Control Bank Selector Switch in the CBD position and verify no rod movemen b. Manually insert control rods to restore program Tav Check PT-505 Turbine impulse Pressure indication - NORMA d. Trip the reactor and go to E-0, Reactor Trip or Safety Injection, Step QUESTION: 016 While operating at 92% power, a manual boration is initiated due to a malfuncuon in the makeup control system. All control sysem are in AUT While performmg the lineup the operator is momentarily distracted and when he completes the lineup he inadvertently opens the emergency boration valve (CS-V426) instead of the boration flow control valve (CS-FCV-110A).
Which of the following symptoms would indicate that the emergency boration valve (CS-V426)
was inadvertently opened mstead of flow control valve (CS-FCV-110A)?
. T., begins to fall, the boric acid batch integrator advances more rapidly than normal, flow indicator (CS-FI-!83A) reads 70 GPM, and control rods begin to step ou T., remams steady, the boric acid batch integrator does not advance, flow indicator (CS-FI-183A) reads 70 GPM, and control rods do not move, T., begins to fall, the boric acid batch integrator advances more rapidly than normal, flow indicator (CS-FI-183A) reads 70 GPM, and control rods begin to step i T., begins to fall, the bo-ic acid batch integrator does not advance, flow indicator (CS-FI-183A) reads 70 GPM, and control rods begin to step ou l Page 9 of 54
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QUESTION: 017 )
l The plant is at 100% power when the followmg VAS alarm is received:
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-D5734 Vital UPS lEINV PWR FUSE BLOWN
1.ocally at EDE-CP-lE (static transfer switch), the si~..io. observes that the reverse transfer lig'at . !
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i Which of the followmg events is the likely cause of this alarm / indication? j l PP-lE is being supplied from it's DC suppl l i
b. PP-1E has transferred to it's alternate suppl PP-1E maintenance supply breaker has tripped ope PP-1E is de-energize QUESTION: 018 j l
A loss of 123 VAC Mal Instrumentation Panel PP-1 A has occurred. The crew is using ]
OS1247.01, i'.oss of a 120 VAC Vital Instrument Panel, to stabilize the plan i At step 2, the =~=dary operator notices that the Steam Dumps are open and closes the dumps using the steam dump interlock control switch as directed by the procedur Which of the followmg plant conditions would have. to exist for the Steam Dumps to open when PP-1A was lost? The C-7A signal from a prior load rejecuon was not rese b. The Steam Dumps were in the " Steam Pressure" mode of operatio FW-PT-506 failed hig C-16 was actuate Page 10 of 54
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QUESTION: 019
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& plant is at 50% power when a total loss of Main Feedwater occurs. The reactor does not trip, 4 and the crew enters FR- i What W6 if any, will the ATWS Mitigsten System provide under these conditions?
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- a. The ATWS Mitigation System is not armed under these ~=&6e= l
i b. The ATWS Mitigation System will send a start signal to the EFW pumps when 1/4 SG NR levels are less than 5%.
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i The A1WS Mitigation System will send a start signal to the EFW pumps when 2/4 detectors
on 1/4 SGs are less than 14%.
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d. The ATWS Mitigation System will send a start signal to the EFW pumps when 3/4 SG NR j levels are less than 5%.
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- QUESTION
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i E-3, " Steam Generator Tube Rupture", directs the operators to cooldown the RCS using the intact
SG's and then to depressurize the RCS below ruptured SG pressure.
l Why doesnt the ruptured SG depressurize as a result of the RCS cocidown?
, The tube rupture provides an energy input to the RCS, from both the high temperature water i and mechanical compression of the steam space.
i j b. The ruptured SG is isolated and ruptured SG water level is maintained above the U-tubes so
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- that a layer of hot water insulates the steam space from the U-tubes as the RCS cools.
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' The rupturcxl SG is isolated preventing energy removal.
l The RCP in the mptured loop is stopped reducing the heat transfer rate in that steam j generator, i
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- QUESTION: 021 :
j Step #12 of ECA-3.1, "SGTR With Loss of Reactor Coolant - Subcooled Recovery Desired",
l l provides a check to deternune if a subcooled recovery is appropriate. Why is a transition to ECA-3.2, "SGTR With loss of Reactor Coolant - Saturated Recovery Desired" directed if RWST level
- is less than 290,000 gallons and C-aiament sump level is not within the " expected region" of
- Figure ECA-3.1-l? Maia*=iaias saturated RCS conditions facilitates PZR pressure control and allows RCP
- operation without concern for pump damag b. Reducing pressure to saturate the RCS and thereby reduce RCS leakage is appropriate if significant leakage is occurring outside Contamment as RCS makeup water supply may be madequate.
, It is advantageous to maintain saturated RCS conditions in order to provide an indication of
, coolant inventory trends and ensure significant margin to core uncovery.
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d. With subcooled conditions in the RCS no indication of RCS inventory trends exists until the
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core uncovers.
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QUESTION: 022 While performing E-3, " Steam Generator Tube Rupture", prior to initiating the RCS cooldown, the USS directs you to determme if the ruptured steam generator pressure is greater than 225 psig.
Reactor Coolant - Subcooled Recovery Desired".
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Which of the following describes the reason for transitioning to ECA-3.l? l l
l j ECA-3.1 is more appropriate because if ruptured SG pressure is this low it may be an
- indication of a ruptured-faulted SG.
i b. ECA-3.1 is more appropriate because it ensures that the ruptured SG pressure is greater than the intact SG pressures for any subsequent cooldow :
' ECA-3.1 is more appropriate because it contains steps that ensure that the subsequent cooldown does not cause a low steam line pressure Safety Injection actuatio d. ECA-3.1 is more appropriate because it ensures that the subsequent cooldown does not j J decrease RCS pressure below accumulator injection pressur l Page 12 of 54
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sm QUESTION: 023
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"Ihe plant is in MODE 1,100% powe i A loss of Off site Power occurs, and "A" DG fails to star Which of the following describes the expected electrical power flowpath to vital instrument power panels PP-1 A, PP-1B, and PP-lE? Battery B-1 A -+ DC bus 1I A -+ UPS-I-1 A -+ PP-1 A j, Bus E61 * MCC E612 * UPS-I-1B -+ PP-1B Battery B-1 A * DC bus 11A * UPS-I-lE -+ PP-1E b. Battery Charger BC-1A -+ DC bus llA -+ UPS-I-1A -+ PP-1A
, Bus E61 -+ MCC E612 -+ UPS-I-1B -+ PP-1B Bus E61 -+ MCC E612 * UPS-I-lE * PP-1E !
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! Battery B-1 A -+ DC bus 11 A -+ UPS-I-1 A -+ PP-1 A
- Battery Charger BC-1B -+ DC bus 1IB 4 UPS-I-1B * PP-1B l Battery B-1 A -+ DC bus 11 A -+ UPS-I-lE * PP-1E I
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d. Battery Charger BC-1 A * DC bus 11 A -+ UPS-I-1 A -+ PP-1 A I Bus E61 -+ MCC E612 * UPS-I-1B -+ PP-1B l
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Bus E53 -+ MCC E531 -+ 480/120vhansformer via Static Transfer switch -+ PP-1E
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QUESTION: 024
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The plant is at 100% power at EOL with control bank D at 225 steps. A bank "D" rod drops into l the core. No operator actions are take ;
Which of the following represents the expected plant response to the dropped rod? Initially there will be a prompt drop in reactor power and then reactor power will slowly increase to a value equal to the initial powe l
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b. Reactor power will decrease to a new value that represents the negative reactivity worth of the ;
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dropped ro ' Initially there will be a prompt drop in reactor power and then reactor power will slowly
- increase to a value less than the initial power that represents the negative reactivity worth of the dropped ro ;
d. Reactor power will increase slightly and then decrease to a value equal to the initial powe .
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QUESTION: 025 PLANT CONDITIONS:
. Cc '== Vacuum is 22.5 inches Hg and slowly decreasing e lead Reduction is in progress
- Turbineload is 350 MWE Which of the following actions should be taken by the operating crew? Immmu*1y trip the turbine and verify all stop valves closed and the generator breaker opens b. Continue the load decrease to increase har vacuum to > 25 inches H l Immediately trip the reactor and go to E- Continue the load decrease and if vacuum remains greater than 22.4 inches Hg remove the turbine generator from service IAW OS1000.06, Power Decreas ~
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l QUESTION: 026 )
l Reactor trip and Safety Injection signals have actuated safeguards equipment. The Reactor Trip breakers failed to open and the crew has transitioned from E-0 to FR- At the step requiring the crew to initiate Emergency Boration, what is the appropriate flowpath? One boric acid pump runmng, chargmg flow maintained greater than 50 GPM, and letdown flow adjusted to maintain VCT leve b. CS-Y426 open, at least one boric acid pump runmng, charging flow control in Manual and at manmum, suction aligned to RWST with VCT isolate Charging flow maintamed greater than or equal to 110 GPM, suction aligned to RWST, VCT ,
suction isolated and letdown flow adjusted to maintain VCT leve !
d. CS-V426 open, at least one boric acid pump running, charging flow control in auto and set at 120 GPM, suction aligned to RWST and VCT suction isolate I
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QUESTION: 027 PLANT CONDITIONS:
. The unit has just returned to 100% power folkming a refueling outag . All Shutdown Banks are fully withdraw . Control Banks are wi+1' drawn with Control Bank D at 180 step . Rod Controlis operatiri in AUT . A high stator cooling water temperature causes the turbine to begin runrung bac During the runback D7762 "CNTL BK D INSERTION LIhC 0 LO-LO"is receive What is the significance of this alarm? De Relaxed Axial Offset Limits have been exceeded b. He MODE I Shutdown Margin Limit may have been exceede The runback rate has exceeded the capabilities of the Control Rod Drive Syste The level of turbine runback has ber.n excessiv l
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QUESTION: 028 4 PLANT CONDITIONS:
. The plant is operating at 100% steady state
. All Shutdown and Control Bank rods are at 229 steps
- De RO is performmg rod exercises in accordance with the applicable surveillance procedur ,
. As directed i,y the procedure, Control Bank "D" is exercised using " Bank Select" for CB The praedure directs the operator to return the rods to their initial position but the operator accidentally leaves CB D rods at 224 steps following completion of surveillance testin l Waat is the consequence ofleaving CB D rods at 224 steps as opposed to retuming them to their l i JGal position of 229 steps?
' The Rod Insertion Limit (RE) computer will calculate a RIL that is lower than it should be for the given power leve I
- b. The Digital Rod Position Indication (DRPI) for the bank will read lower than the Bank d
Demand Positio The Control Bank Overlap Unit will sense that CB D rods are still at 229 steps and rwerlap control banks based on this input.
j d. A Logic Cabinet Non-urgent Failure will be generated the next time CB D rods are moved in !
AUTO or MANUAL.
i QUESTION: 029 PLANT CONDITIONS
- 22% powe * Pressurizer Pressure is stable at 2235 psig Which of:he following plant conditions would require the affected Reactor Coolant Pump to be tripped in accordance with OS1201.01, "RCP Malfunction"? Seal Water Inlet Temperature is 200" Upper Radial Bearing Temperature is 165' RCP Frame Vibration is 2.5 mils and increasing at 0.2 mils per hou d. RCP Seal Leakoff Total Flow is greater than 9 gp Page 16 of 54
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QUESTION: 030 i
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Which of the following conditions will permit operung of the Regenerative Heat Exchanger Outlet l
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Isolation Valve (CS-V-145)?
- Pressurizerlevel of 20*/.
I and Letdown Flow Isolation Valves (RC-LCV-460 and 459) - BOTH OPEN
b. Pressurizer Level of 20% l d
letdown Flow Isolation Valves (RC-LCV-460 and 459) - EITHER ONE OPEN ! Pressurizer leve! of 20%
and letdown line Isolation Valve (RC-V81) - OPEN
d. Pressurizer level of 20%
and i
letdown Flow Isolation Valves (RC-LCV-450 and 459) - BOTH CLOSED r
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- f QUESTION: 031
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- Given the followmg
- A Large Break LOCA has occurred e All Safeguard Systems respond as designed
e The operstag crew enters E-0, Reactor Trip or Safety Injection
' e Based on observed plant conditions the crew transitions to E-1, Loss of Reactor or Secondary l Coolant.
4 e In E-1 the crew resets the SI signal and shuts down the EDGs which have been runnmg l uloaded i e When RWST Iow-Iow level is reached at 125,000 gal. the Crew transitions to ES-1.3,
! Transfer To Cold leg Recirculation.
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{ How will the process of transfemng to cold leg recirculation be affected with the SI signal having i
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i Automatic opening of the Containment Sump Recirculation Valves (V-8 and V-14) will not occur. These valves will need to be manually opened in accordance with ES-1.3.
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b. The 'S' signal must be reset to allow manual closure of the RWST suction valves (V-2 and V-j 5).
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' Automatic opening of the Contamment Sump Recirculation Valves (V-8 and V-14) aEl
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automatic closure of the RWST Suction Valves (V-2 and V-5) will not occur. These vah es
- will need to be manually operated in accordanz with ES-1.3.
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d. All valves reposition as d%aned; however, both Contamment Building Spray pumps will trip when the RWST Low-law level setpoint is reached. 'Ihe pumps will need to be manually started in accordance with ES- J
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QUESTION: 032 The followmg plant conditions exist:
- The plant was operatmg at 100% power when a large LOCA occure * A loss of offsite power has occurred, the EPS sequece is complete, and both E.w.wy Buses are being powered from their respective diesel generator l e An 'A' train Containment Spray Actuation Signal (CSAS) was generated due to high contammet pressure and 'A' train equipment is operstmg as required. A 'B' train CSAS failed to actuate and required Phase B status panel lights are not lit for Train 'B'.
- The +4-g crew is performmg the RNO actmas of step #14 of E-0 and have manually ara =W both 'B' train CBS/P/CVI manual actuation switches but the 'B' train CBS pump has not started and automatic valves failed to reposition as req' aire . The operator has attempted to manually start the 'B' train CBS pump but it will not start (yellow control switch light ilk m'). The operator scanually repositions other w..ycceats in accordance with the status pane >
What action must the crew take to start CBS-P-9B7 l The CBS pump cannot be started until off-site power is restored. The UAT or RAT breaker must be close b. Since a 'B' train CSAS signal was not generated the CBS pump was not sequenced by the EPS. RMO must be reset before the pymp can be starte l The crew must manually actuate stepping relay SR3 or relay HR8 at the EPS cabinet in the
'B' Train Essential Switchgear Roo d. The RMO bypass switch must be held in the bypass position while the CBS pump breaker is lot. ally closed onto Bus E QUESTION: 033 Why is a Safety Injection signal generated by SSPS in response to a steamline rupture rather than just a reactor trip signal?
! To ensure that EFW is available for subsequent plant cooldow b. To prevent the affected steam generator from dryou To ensure feedwater to the affected steam generator is isolate d. To prevent a return to criticality due to positive reactivity addition from the cooldow Page 19 of 54
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QUESTION: 034 i PLANT CONDITIONS:
. The plant is in MODE 5 e h pressurizer is soli . Normalletdown is in service
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Which of the following explains how RCS pressure is controlled in this situatio Increase charging to increase pressure; increase letdown flow to lower pressur b. Energize pressurizer heaters to raise pressure; deenergize heaters or use aux. spray to lower pressur Throttle PCV-131 closed to raise pressure; throttle PCV-131 open to lower pressur d. Throttle CS-HCV-128 open to raise pressure; Throttle CS-HCV-128 closed to lower pressur QUESTION: 035 The plant is at 100%. The Backup Pressurizer Level Channel (RC-LT-460) rapidly fails low to 0%. Eg operator action is take Which of the following describes the status for the given components / parameters 2 ndnttes a&r the level channel failure occurred?
BACKUP CONTROL LCV-459 LCV-460 ACTUAL PZR HEATERS HEATERS LEVEL OFF ON OPEN CLOSED DECREASING OFF OFF CLOSED CLOSED DECREASING OFF OFF OPEN CLOSED INCREASING OFF ON CLOSED CLOSED INCREASING Page 20 of 54
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QUESTION: 036 The time is 0300 and the plant is in MODE 3 at 549' F with EOL core condition l The followmg events occur:
. The PSO reports that Tavg has begun decreasing at a rate of approxuna:ely l'F/ mi !
L e 'Ihe BOP reports that the low setpomt safety valve on the "C" SG has apparently failed Partally open
. An NSO is dispatched and visually confirms one of the SG Safety Valves on the "C" SG is l passing steam to the atmosphere l . The PSO confirms an RCS cooldown rate of 5'F in the last 5 minutes t
l Assuming no operator action is taken in response to the stuck open safety valve, which of the
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following conditions will exist by time 04007 i 1. Main Steamline isolatio . The Tech. Spec. for RCS Cooldown Rate will be exceeded 3. 'Ihe actual amount by which the reactor is shutdown will decreas . A PTS challenge to reactor vessel integrity will occu l I and 3 l J l
b. 2 Only Only d. 2 and 4
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l QUESTION: 037 What would be the initial effect on a SG's indicated pressure and level if its Main Steam Line Isolation Valve were to inadvertently close with the plant at full power?
l l Both pressure and level would initially mereas l
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b. Pressure would initially increase. Level would initially decreas ! Pressure wculd initially decrease. Level would initially increase.
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QUESTION: 038 :
i The followmg plant conditions exist:
= A Reactor Startup followmg a refuchng outage has just been completed.
l l * The Moderator T ..m e-re Coefficient is + 2 pcmf'F j e Reactor power is currently stable below the Point Of Addmg Heat (POAH) at IX10* nmps . !
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An Atmospheric Steam Dump Valve fails ope l
How will RCS Tavg and reactor power be affected by this failure?
i Both T-avg and reactor power will increase.
l b. T-avg will increase. Reactor power will decreas T-avg will decrease. Reactor power will increas d. Both T-avg and reactor power will decrease.
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i QUESTION: 039 A fire in the Train 'A' Electrical Penetration Area has been confirmed by the Fire Brigade, and the control room crew has entered the appropriate fire response procedure. Subsequently a reactor trip occur The procedure flow path the operating crew will follow is:
> E-0, " Reactor Trip or Safety Injection", OS1200.00," Response to Fire or Fire Alarm j Actuation", OS1200.00A, App. A to Fire Hazards Analysis for Affected Area / Zone, j OS1200.01, " Safe Shutdown and Cooldown From the Main Control Room".
b. OS1200.00," Response to Fire or Fire Alarm Actuation", OS1200.00A, App. A to Fire Hazards Analysis for Affected Area / Zone, OS1200.01," Safe Shutdown and Cooldown From ,
the Main Control Room". E-0, " Reactor Trip and/or Safety Injection", ES-0.1, " Reactor Trip Response" and simultaneously OS1200.00," Response to Fire or Fire Actuation".
l d. OS1200.00," Response to Fire or Fire Alarm Actuation", when directed by OS1200.00 enter E-0, " Reactor Trip or Safety Injection" and retum to OS1200.00 after the immediate action
steps, OS1200.00A, App. A to Fire Hazards Analysis for Affected Area / Zone.
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QUESTION: 040 A reactor startup is in progress and the reactor has just achieved criticalit A Rod Control System Urgent Failure Alann is received and rod B6 irops into the core causing the reactor to go subcritica I Which of the following describes the course of action to be taken? Match Tavg-Tref by adjusting turbine loa b. Trip the reactor and go to E-0, Reactor Trip or Safety Injectio L Refer to Technical Speci6 cations and conduct QPTR and Shutdown Margin Calculation I Conduct a plant shutdown using OS1000.03, Plant Shutdown From Minimum Load to Hot Standby.
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QUESTION: 041 Given the following:
- ONE control rod is misaligned from its group by more than twelve (12) steps and deternuned to be INOPERABL . The Technical Specification ACTION statement limits reactor power to 75% Rated Thermal Powe Which of the following is the reason for this power limit? Reduces the Rod Insertion Limit below the misaligned rod positio '
b. Allows the plant to be operated without performing a re-evaluation of the safety analysis affected by a misaligned ro Relieves the operators of having to calculate Shutdown Margin every 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Provides assurance of fuel rod integrity during continued operation Page 23 of 54
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QUESTION: 042 Which of the following describes the effect ofone UNDERCOMPENSATED Inteanrdiate Range Channel following a reactor trip? Channel indicates HIGH preventing P-6 from automatically energizing the source rang b. Channel indicates LOW prematurely energizing the source rang Channel indicates HIGH, the source range will be energized by P-6 from the other IR channe d. Channel indicates LOW, the source range will NOT be energized until P-6 issupplied from the other IR channe l l
QUESTION: 043 Which of the following describes how the Thermal Barrier Cooling Water System (CC) would be isolated from a tube rupture in an RCP thermal barrier? Manual isolation valves on the supply and retum lines to the RCP b. Check valves in the supply lines and motor operated isolation valves on the return lines from the RCP Mer operated isolation valves in the supply lines and check valves on the retum lines from the RCP d. Motor operated isolation valves on the supply lines and motor operated isolation valves on the return lines from the RCP I
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QUESTION: 044 Which of the followmg is a reason for isolating all feedwater to a faulted Steam Generator (SG)? To reduce the probability of a SG tube rupture in the faulted S ,
b. To muunuze the RCS cooldownand mass and energy release followmg a *=mli- brea To prevent all feedwater flow from entering the faulted SG and filling the SG, causing the ASDV to lift.
d. To allow quicicer identification of a SG tube rupture and thus limit the release to the environment below 10CFR100 limits during a design basis even I I
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I QUESTION: 045 .
'Ihe following plant conditions exist:
. The plant is at 100% power
. The controlling SG water level channel to the 'C' SG has failed low
. The operator is unable to shift the affected feed water regulating valve to manual.
Which of the following methods is used to maintain steam generator water level? Use the main feedwater pump master speed controller together with manual contrel of the unaffected SG feed water regulating bypass valves.
b. Take manual control of the unaffected SG feedwater regulating valves and use the feedwater regulating bypass valve for the affected SG. Use the main feedwater pump master speed controller and locally control the affected SG feedwater regulating valv l d. Immediately trip the applicable bistables so a shift to an unaffected channel can be made and control the affected SG level in aut Page 25 of 54
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QUESTION: 046
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- Given the following
- The control room has been evacuated due to a fir * All runote safe shutdown system lineups have been complete . The local / remote switch on Bus E-5 for RHR pump RH-P-8A is in local
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e A valid SI signal has just been received.
Which of the followmg describes the RHR pump response?
- a. The pump will start and remams runnmg until the "S" signal is reset, at which time the pump will stop.
! b. The pump will start and remams runmng until its associated breaker is opened locally.
i j The pump will not automatically start, but the operator can start /stop the pump using the local
. control switch at the switchgea i
j d. The pump will not automatically start, but the operator can start /stop the pump from the Train
A remote safe shutdown pane l
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QUESTION: 047 Given the following:
- A reactor startup is in progress with the reactorjust critica ]
- The operator has just stopped moving rods l
- Power slowly increases to above the P-6 setpoin One source range (SR) channel fails LOW. The remaining power indications stabiliz l l
WHICH of the following actions is required? j Block the source range since it is not required above P- b. Trip the reactor and enter E-0, Reactor Trip and/or Safety injectio Suspend all operations involving positive reactivity changes until both SR channels are restored to operabilit d. Conduct a reactor shutdown and restore both SR channels to operability prior to the next startu Page 26 of 54
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j QUESTION: 048 The operating crew is establishing Fe-i and Bleed per FR-H.1, " Response to Loss of Secondary Heat Sink".
Why are the PORVs manually opened rather than allowing RCS pressure to rise and automatically )
open the valves at their lift setpoints? l 1 RCP damage may occur due to inadequate seal injection at PORV setpoin I b. Waiting for the PORVs to open at their lift setpoints may prevent adequate injection flow and '
lead to inadequate core coolin A solid water RCS at the PORV setpoint has a high potential to challenge the RCS pressure safety limit.
I d. Depressurizing a SG for condensate feed with the RCS at the PORV setpoint will exceed SG
U-tube delta-P limits.
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QUESTION: 049
The plant is operating at 100% power.
i The following events occur:
l . Pressurizer level deviation alarm actuates
. Pressurizer backup heaters energize
- CS-FCV-121 output is slowly decreasing
- Pressurizer level is slowly decreasing on LT-460 and LT-461 l e Pressurizer level reads 100% on LT-459 ,
Which of the following will result if N_Q operator action is taken? Pressurizer level will decrease and be controlled at a lower than normal leve l b. Pressurizer level will decrease, causing Pressurizer pressure to decrease until a reactor trip -
occurs on low pressurizer pressur Pressurizer level will decrease, then increase and be controlled at a higher than normal level.
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QUESTION: 050
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Under degraded core cooling conditions, one RCP is tripped by FR-C.2, " Response to Degraded
Core Cooling" ;
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[ Which of the followmg describes the basis for tripping only one RCP.
i - To reserve one pump from possible damage for future cooling needs yet still maintain sufficient core cooling flo b. To limit heat input to the RCS already in a degraded condition yet still maintain sufficient core cooling flow.
i To conserve SG water inventory yet maintain sufficient boron moung to prevent a reactor restart accident.
l d. To reserve inventory in one SG for future cooling needs yet maintain sufBeient loop flow t ,
j j prevent a stagnant loop (PTS concerns). ;
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QUESTION: 051 The following control room air intake radiation monitors are in "HIGH" alarm:
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East Air Intake Train 'A' RM-6506A East Air Intake Train 'B' RM-6506B Channel Al Channel B1
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West Air Intake Train 'A' RM-6507A West Air Intake Train 'B' RM-6507B Channel Al Channel B2 Which of the following describes the required control room ventilation alignment based upon the above radiation monitor alarms? CBA-FN-27A and CBA-FN-27B makeup air fans - BOTH NOT RUNNIN b. CBA-FN-27A and CBA-FN-27B makeup air fans - BOTH RUNNIN CBA-FN-16A and CBA-FN-16B emergency makeup filter fans - BOTH RUNNIN d. CBA-FN-16A and CBA-FN-16B emergency makeup filter fans - BOTH NOT RUNNIN Page 28 of 54
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The plant is at 100% power,3410.9 MWth.
! 'Ihe crew has determmed that feedwater heaters 25B and 26B must be removed from service to isolate a leakmg 26B heate Which of the followmg actions will have to be taken by the crew?
] Reduce turbine load to prevent exwiae licensed thermal power limits.
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b. 'Ihe 25A and 26A feedwater heaters will also have to be removed to maintain balanced feedwater heatmg through the feedwater strmgs Increase reactor power to ir . ease RCS m d. Increase turbine load to compensate for the load decrease due to the partial loss of feedwater preheatin I QUESTION: 053 in response to a steamline break inside.contamment the crew has transitioned from E-2 to E-1 and the following conditions exist:
. Pressurizer level- 25%
. Intact SG narrow range levels - 52% and slowly increasing
. Total feed flow to intact SGs - 100 gpm
. RCS pressure - 1600 psig and increasing
. RCS temperature-475 F e Contamment pressure - 16 psig Which of the following MUST be increased before the operating crew can terminate Sl?
, Feed flo b. RCS subcoolin Pressurizer leve RCS pressur Page 29 of 54
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QUESTION: 054
'Ihe liquid radwaste test tank discharge radiation monitor (R-6509) has been declared INOPERABL ;
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Which of the followmg describes the Technical Srihion ACTION that will permit contmued l release from the liquid waste system? Liquid waste discharge will not be permitted until the discharge radiation monitor is returned to
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operable statu b. A temporary monitor may be used provided its alarm setpoint is more consenstive than the R-6509 setpoint to allow the oper. tor sufBeier time to manually secure the discharge in the
- event an alarm condition occur Two independent samples of the tank to be discharged must be analyzed, and two technically qualified staff members must independently verify the release rate calculations and the discharge line valve lineup.
- Samples must be taken every 15 minutes while the discharge is in progress, to verify the efUuent is within Technical Specification limits.
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, QUESTION: 055
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Which of the following indications wil; the operator u*ilize to verify natural circulation during a post LOCA cooldown and depressurization? Core exit TCs, RCS hot and cold leg temperatures, RCS pressure and RCS subcooling.
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b. RCS subcooling, RCS hot leg temp., RCS pressure, PZR level and SG pressure Core exit TCs, RCS hot and cold leg temps. RCS subcooling and SG pressure d. RCS subcooling, RCS pressure, SG pressure, PZR level and RCS cold leg temperature.
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QUESTION: 056
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An LOP occur Whde the E..egy Power Sequencer (EPS) is in the process of completing the stepping l sequence, an autonume safety insection occur .
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l Which of the followung describes the operation of the Emergency Power Sequencer upon initiation I of the SafetyInjection? j i Sequencer resets to step 0, diesel remams runnmg, diesel output breaker remams closed, l
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C~ tamment structure cooling fans are tripped, all previously runmng loads renum runmng and the sequencer sequences the Si load b. Diesel output breaker opens to strip the bus, sequencer resets to step 0 in a standby mode, "S" signal causes the diesel output breaker to close and the EPS to restant all the loads in the SI/ LOP stepping sequenc Sequencer stops at the step in progress, daesel remains runnmg, diesel output breaker remams closed, the RA relay actuates to trip the Contamment structure cooling fans and then the EPS completes the stepping sequence in progres d. Diesel output breaker opens to strip the bus, diesel remains running but the "S" signal prevents the EPS from actuating, the sequencer resets to step 0 in a standby mode and will l restart the entire sequence when the operator closes the diesel outptit breake l l
QUESTION: 057 j
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Which of the following will occur on a loss of Vital DC Bus 1 IB? ! Both EFW pumps start and the MFRV and bypass valves fail ope b. & steam driven EFW pump starts, however EFW flow can be throttled only with the "B" train throttle valve W "B" train P-14 solenoids on the MFRV and MFRV bypass valves are deenergized causing these valves to fail close d. The "B" train P-12 solenoids on the steam dump valves are de-energized causing the steam dumps to fail cpe Page 31 of 54
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l QUESTION: 058 j The control room operators are responding to a large-break LOCA inside containment in
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accordance with E-1. Critical Safety Function Status is: l l
Suberiticality-GREEN 1 Core Coohng -YELLOW l
Heat Sink - GREEN Integrity-YELLOW l I 1 The Shift Manager checks contamment conditions to deternune if the Contamment barrier is intac The following conditions exist: j e Contamment pressure: 42 psig and slowly increasing
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e Containment sump level: 3.2 feet and slowly increasing
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. Containment radiation: 8.0 R/hr and stable
. All Containment Phase A & B penetrations are isolated
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Given these conditions the Shift Manager will direct the USS to: Transition to FR-Z.1, " Response to High Containment Pressure" l b. Transition to FR-Z.2, " Response to Containment Flooding" Transition to FR-Z.3, " Response to Containment High Radiation" d. Remain in E-1, Yellow path FRPs are entered upon discretion onl Page 32 of 54
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QUESTION: 059
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The following plant conditions exist: j
- The plant has just completed a 428 day full power ru * 'Ihe plant is cooling down in MODE 3
. RCS temperature is 490*F e RCS pressure is 1800 psig (SI has been previously blocked) i e Excess letdown is in service l A leak develops in loop "B" of the PCCW system causing the loop "B" head tank level to INCREAS WHICH of the followmg could be the potential source ofleakage into Loop "B"?
, Excess letdown heat exchange I
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' "B" RHR heat exchanger, d Letdown heat exchange d. Seal Retum Water heat exchanger.
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l QUESTION: 060
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Which of the following describes plant protection provided for a steam line break accident? . I i If main MemmHan pressure drops to 585 psig on two of four detectors on any steamline after P-
, 11 has been blocked, all the MSIVs will close and a safety injection will occur.
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i b. If main steamline pres:;ure drops to $85 psig on two of three detectors on any steamline, the ,
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MSIV will close on that steamline isolating the brea l If main steamline pressure drops to 585 psig on two of four detectors on any steamline, the MSIVs will close and a safety injection will occur.
' If main steamline pressure drops to 585 psig on two of three detectors on any steamline, the ,
MSIVs will close and a safety injection will occu l i
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QUESTION: 061 Step 10 of E-1, "Ims of Reactor or Secondary Coolant", directs the operator to check all steam generator pressures stable or increasin .
If any steam , w.iaos pressure is decreasing, the operator is directed to retum to step 1 of E- '
Which of the following describes why the operator should not proceed past step 10 with a ,
depressurmag steam generator? l SI ternunation criteria could not initially be met and more restrictive termmation criteria would ,
be -*~ed in ES-1.2, " Post LOCA Cooldown and Depressurization" b. E-1 provides no guidance for faulted steam generator isolation past step 1 The RCS cooldown rate must be under control in order for subsequent E-1 steps to be i effectively implemente i
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d. SI termmaton criteria do not exist in ES-1.2, " Post LOCA Cooldown and Depressurization" A loop back to step #1 ensures that Si termination criteria are met in E- l l
l QUESTION: 062
'Ibe core exit thermocouples (CETCs) are used as the temperature input for subcooling margin calculation. Subcooling margm is one of the criteria for safety injection termmation in all of the Emergency Operatmg Procedures (EOPs). !
t Which of the following explains why the core exit thermocouples are the preferred instrumentation for subcooling detenninatio a. 'Ihe core exit thermocouples are used because any effect due to an isolated steam gmerator i i
will be limited to the core exit thermocouples in the vicinity of the isolated loop reactor vessel inle b. The core exit thermocouples are used because they are automatically compensated for adverse Contamment condition The core exit ti-i.acouples are used because they provide the most direct indication of conditions at the hottest point in the reactor coolant system d. 'Ibe core exit thermocouples are used because they are retractable and can measure localized hot channels at various core height Page 34 of 54
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QUESTION: 063 An inadvertent SI occuned and the operatmg crew is currently in ES-1.1, "SI Terminadon" at step 9, " Check if SI Pumps Should be Stopped".
The followmg conditions exist:
- PZRlevelis 35%
- PZR pressure is 2300 psig and increasing e Both SI pumps are running e
ONE chargmg pump is runmng providmg 60 GPM flow through the normal charging hander
. Both RHR pumps are running Subsequently, one PORV opens and fails to reclose. Attempts to close the PORV's associated block valve fai Which of the following conditions will first require the operator to manually start the non-operating charging pump? PZR level drops to less than 17%.
b. RCS Subcooling drops to less than 40 RCS pressure drops to less than 1650 psi Contamment pressure increases to greater than 4 psi QUESTION: 064 The pressure channel input to the master pressurizer pressure controller is slowly failing hig Which of the followmg explains the pressurizer heater response to this failure? h control group of heaters will go to zero output and remain there throughout the failure of the channel. W backup heaters will not energiz b. 'Ibe control group of heaters will go to zero output initially and then return to full output,
, followed by all sets of backup beaters turning o The control group of heaters will go to zero output and then turn back on to full output, followed by the C and D sets of backup heaters turning o d. The control group of heaters will go to zero output initially and then turn back on to full output, subsequently all sets of backup heaters will turn on followed by all heaters (control and backup) deenergizin l
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QUESTION: 065 A total loss ofinstrument air has occurre Which of the following is the failed position for the listed ==panents:
CC-V122 CS-LV-112A CS-VISO CS-HCV-128 PCCW ORC VCT divert IJD HX RHR flow CTMTIsol ORC iso control CLOSED VCT CLOSED CLOSED CLOSED DIVERT OPEN CLOSED OPEN VCT OPEN OPEN OPEN DIVERT CLOSED OPEN QUESTION: 066 The local / remote switch for the 'A' emergency diesel is in the " local" positio An LOP occurs and 30 seconds later a safety injection occur Which of the following describes the status of the "A" emergency diesel? The diesel is runnmg, the output breaker automatically closed and the sequencer is following the LOP /SI load sequenc b. The diesel is running and the output breaker must be closed by an operator before tle sequencer will commence the LOP /SI load sequenc 'Ihe diesel is runmng, the output breaker must be closed by an operator and the LOP /S1 loads must be manually started d. The diesel generator will NOT start until the local / remote switch is returned to the " remote" positio ;
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QUESTION: 067 In accordance with Attachment 'A' to OS1215.07, "Ims of Spent Fuel Pool Cooling or Level",
winch of the following describes the preferred order of EMERGENCY makeup water sources to the spent fuel pool?
l Chemical and Volume Control System Makeup, Demmeralued Water, Gravity Feed from the RWS b. Chemical and Volume Control System Makeup, Gravity Feed from the RWST, Gravity Feed from the CS ~ Gravity Feed from RWST, Gravity Feed from CST, Denunerahzed Wate d. Gravity Feed from RWST, Gravity Feed from CST, Fire Protection Syste l i
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I QUESTION: 068 The plant is at 100% power and a loop 1 Tw mstrument has failed LOW. The operating crew has completed all of the actions in the applicable Abnonnal operating procedure for this instrument i failure, tripped applicable bistables, and all controls are back in automati Subsequently, pressurizer pressure instrument PT-456 fails LO Which of the followmg describes the expected plant response? No effect on the plan b. Reactor trip will occur due to OTAT trip coincidence being me Reactor trip will occur due to OPAT trip coincidence being me i "S" signal on PZR low pressur Page 37 of 54
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l QUESTION: 069 I
j Refueling is in progress when Channel "A" of the fuel manipulator crane radiation monitor (R-j 6535A) spuriously alarms HIGH. ch=M "B" reads normal
- What operator actions are required in acconiance with OS1252.03 " Area High B=Antion"?
t Perform an OPERABILITY surveillance on the RDMS channel to verify that the alarm is l
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i b. Verify a contamment ventilation isolation has occurre Manually close the valves and stop the fans for a contamment ventilation isolatio d. Ensure the contamment purge supply and exhaust valves are ope ,
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l QUESTION: 070 ;
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PLANT CONDITIONS:
= Plant is at 100% power
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- Pressurizer pressure channel, PT-455 failed high
- h operating crew carried out the actions of OS 1201.06, all applicable bistables are tripped e Channel PT-457 is now the controlling channel :
- All systems have been returned to automatic control j A loss of 120 VAC vital instrument panel PP-IC has just occurre Which of the following describes the impact on the plant of the loss of PP-1C?
I & plant will remam at 100% power. PZR pressure control will be in manual and the j automatic actuation of the PORVs has been los ,
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. b. A safety injection will have occurred due to low pressurizer pressure logic coincidence being me The PORVs will open due to a high pressure signal and this will eventually lead to a safety injection on low pressurizer pressur d. He master pressure controller will cause the pressurizer control heaters to go to nummum output and close the spray valve .
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QUESTION: 071 *
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A reactor trip and safety injection have occurre ,
- Wluch of the followag describes the status of RCP seal leakoff 7 i Seal leakoff flow thru the seal return heat + . , to the suction of the chi $g pumps is ;
retained for continued RCP operation t l b. Seal leakoff flow is diverted to the #2 seal when CS-V 167 and CS-V-168 go closed on an
"T" signa i i
, Seal leakoff isolates when CS-V-167 and CS-V-168 close on a 'T' signal; seal leakoff flow is prended via a relief valve to the PR d. Seal leakoff flow isolates when CS-V-167 and CS -V-168 go closed on an 'T' signal and all
, seal flow is directed back to the RCS via a relief valve on the return lin *
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QUESTION: 072 . !
I 'Ihe unit has tripped due to a loss of all circulating water pumps. All other plant equipment has j operated as designed, and no operator actions have been taken.
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! When the plant is stable, what should the average RCS temperature (T, ) be?
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b. 557'F i
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a QUESTION: 073 i
The plant is at 40% powe 'Ihe alanns for 'D' RCP high #1 seal leakoff flow is received. Upon cNMa= applicable parameters the operators dh..iiie that the 'D' RCP should be tripped immediately due to seal
- leakoffproblem .
Which of the following actions should be taken by the operating crew? Trip the reactor and go to E-0. After step 4 of E-0, stop 'D' RCP and close the #1 seal leakoff valve after the pump has stopped.
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stopped, and commence a plant Shutdown j Trip the 'D' RCP, take manual control of 'D' SG feedwater control, and commence a plant 4 shutdow i d. Trip 'D' RCP, close the # 1 seal leakoff valve after the pump has stopped, trip the reactor, and perform the immediate actions of E-0.
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QUESTION: 074
The following alarms have been received:
. Instrument Air Pressure Low UA-54
- e D4976 Inst Nr Hdr Press A Low j e D4981 Inst Air hdr Press B Lcw l When should the reactor be manually tripped per the Abnormal Operating Procedure?
. When the ORC Containment PCCW isolation valves fail closed on loss of air.
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b. When the Main Feed Regulating and Bypass valves fail open on loss of ai I
- When Instrument Air header pressure decreases to 90 psi d. When HCV-182 (RCP seal supply flow control valve) fails closed on loss of ai I Page 40 of 54 i
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QUESTION: 075 ,
'Ihe plant is in MODE 5 on Shutdown Cooling. The following conditions exist:
- RCS pressure: 350 psig
- RCS temp: 160 * F e
"B" RHR Pump is unavailable due to a "B" train electrical outage i
"A" RHR Pump trips on overcurrent due to a seized pump shaf I
'Ihe crew has entered OS1213.01, " Loss of RHR during Shutdown Coohng" If both RHR pumps will be out of service for'an avtaaded period of time and the plant is slowly ,
- heating up, what is the ultunate got ofOS1213.0l?
l Manually control RCS inventory and pressure. Monitor Containment integrity, and use at least 2 SG's for heat remova b. Isolate Containment, increase RCS pressure to increase subcooling, and use Feed and Bleed to control RCS temperatur Manually control RCS inventory and pressure. Isolate Containment and use Feed and Bleed to control RCS temperatur d. Manually control RCS inventory, retain automatic pressure control via letdown, decrease pressure and use any available SG's for heat removal l
QUESTION: 076 From the list below, select the radiation monitor that is both a release path monitor and has an automatic action associated with i IGM 810, Condenser Air Evacuatio ILM805, Turbine Building Sum . IGA 410, Fuel Storage Building Exhaus ING222, Plant Vent to Range Ga Page 41 of 54
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QUESTION: 077 I l
A Safety injection has occurred and the crew is at step 11 of E-0 when a loss of Off Site power I occurs. 'Ihe E.msecy Diesel Generators start and their output breakers close to restore power to j the vital busse What is the status of rantaimnent Structure Cooling (CAH)? 'Ihe CAH fans will initially trip due to low CC flow and then restart at SR3 of the Emergency Power Sequencer on the LOP signa b. The CAH fans will inmally trip due to the "S" signal and then restart at SR3 of the Emergency Power Sequencer on the LOP signa The CAH fans will initially trip due to the "S" signal and will not restart at SR3 of the Emergency Power Sequencer on the LOP signa d. 'Ihe CAH fans will have tripped and will automatically restart when Remote Manual Override (RMO)is rese QUESTION: 078 A steam break in the 'C' SG resulted in a reactor trip and Safety injection. The EFW system l functioned normally. The control room team eventually restored an adequate heat s'mk with the following alignment for the EFW flow control valves:
- CS-4214-A1 Throttled Full closed e CS-4214-B1 Auto-Full Open ,
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e CS-4224-A1 Auto-Full Open
- CS-4224-B1 Throttled Full closed
. CS-4234-A1 Auto-Full closed e CS-4234-B1 Auto-Full closed
. CS-4244-Al 'Ihrottled Full closed e CS-4244-B1 Auto-Full Open Subsequently, power to MCC-615 is los What is the effect on the operators ability to control steam generator level? No effect since the loss of MCC affects only one of the two flow control valves to each steam generato b. 'Ihe crew will be unable to initiate flow to the 'A' and 'D' SG' c. The crew will be unable to initiate flow to the 'B' and 'D' SG' d. The crew will be unable to initiate flow to the 'B' S Page 42 of 54
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QUESTION: 079
While operating at 100% power the following VAS alarm is received:
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D4602 RCP A No I sealleakoffflowlo Wien the operator checks the status of the A RCP on the computer he observes the following
point:
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A4250 RCP A No 2 seal leakoff flow reads 1.2 gpm Which of the followmg is the most likely cause'of these indications? Failure of the "A" RCP thernal bame b. Failure of the "A" RCP #1 seal.
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' Failure of the "A" RCP #2 seal.
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, Failure of the "A" RCP #3 seal.
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QUESTION: 080 i l
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Which of the following describes the eventual result of VCT level channel LT-112 failing HIGH?
No operator actions are taken.
- A low VCT level and auto swapover of the charging pump suction to the RWS b. A high VCT level due to continuous makeup from the blender to the VC A low VCT level and loss of NPSH to the charging pump d. A high VCT level and letdown diverted to the PDT
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QUESTION: 081 Refueling operations are in progress with a core offload being conducted in accordance with applicable plant procedures. A fuel assembly has just been placed in the fuel transfer ca A loss of refueling cavity water occurs and the crew commences to carry out the actions of l
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OS1215.05, "Ims of Refuehng Cavity Water". The cr.-w is unsuccessful in moving the fuel
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transfer car to the Fuel Storage Buildin ~
Which of the following is the appropriate action for the crew to take for the fuel assembly timt is ;
out of the cor I Store it in the transfer canal with the refueling machine mast fully evtandeA b. Store it in the RCCA change 6xtur Store it in the upender in a vertical positio d. Store it in the transfer car in a horizontal positio l l
l QUESTION: 082 Due to an operator error during SSPS testing at 100% power, a Safety Injection occur Which of the following procedure steps ternunates SI flow to prevent a pressurizer overfill situation? E-0, " Reactor Trip or Safety Injection", step 15, Verify ECCS flo b. E-0, " Reactor Trip or Safety Injection", step 25, Check if ECCS Flow Should be Reduce ES-1.1, "SI Temunation", step 7, Establish 60 GPM Chargmg Flow, d. ES-1.1,"SI Termmation", step 11, Verify ECCS flow is not require Page 44 of 54
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QUESTION: 083 l,
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Radiation monitor ILM216, STM Gen 'A' Blowdown goes into HIGH alarm ,
Which of the followag describes the expected plant response to this alarm? i
' SG blowdown flash tank d:scharge valve SB-CV4519 goes closed and the blowdown flash tank level control valve (LCV-1909-1) operates to maintain flash tank level at approximately -
25%.
b. SG blowdown flash tank discharge valve SB CV4519 goes closed and flash tank pressure will ,
increase until the blowdown outer C-* ' = isolation valves go closed on high flash tank ,
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' SG blowdown flash tank discharge valve SB-CV4519 goes closed and flash tank level will !
mercase until the blowdown inner Containment isolation valves go closed on high flash tank .
leve ! SG blowdown flash tank discharge valve SB-CV-6519 will remam open until radiatior j monitor ILM215 Blowdown Flash Tank Outlet also goes into high alarm and satisfies the ,
coincidence for closing CV-651 ;
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l QUESTION: 084 l Which of the following sets of conditions would require a plant shutdown as directed by a Seabrook Station Abnormal operatmg procedure? RCP 'A' frame vibration is 2.5 mils and stable, shaft vibration is 15 mils and stable. React power is currently at 30%.
b. Verification that sustained winds in excess of 65 MPH are expected to hit the site within tim next six hour Control Bank D rod H8 cannot be moved in automatic or manual MODE of control. The control room has a power cabinet urgent VAS alarm. The plant is at 100% power.
! d. Chemistry has been monitoring steam generator tube leakage; SG 'A' leakage is 25 gallons per !
l day, SG 'C' leakage is 300 gallons per day. Other SG samples show no detectable actisity j (NDA). Plant power is 75%.
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QUESTION: 085 Which of the following Technical Specification LCO's DOES NOT have Immediate ACTIONS t
associated with it? .7.7, Snubbers (MODES 1,2,3, and 4). ,
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b. 3.4.2.1, Pressurizer Safety Valves (MODES 4 and 5), .1.1.1, Shutdown Margin (MODES 1,2,3, and 4). .9.2, Refueling Operations, Source Range Monitors (MODE 6).
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QUESTION: 086 PLANT CONDITIONS:
. Train 'A' Service Water (SW) was transferred from the ocean to the cooling tower for l quarterly surveillance testin * When transferring Train 'A' .9'll'uack to the ocean, the breaker for cooling tower pump 110A discharge valve, SW-V-54, tripped and the valve was found to be mechanically bound at 60%
ope * Cooling tower pump 110A discharge pressure is 60 psig
- Train 'A'systen flow is 10,000 gp ,
What associated Tech Spec ACTION should the crew enter? (See reference material attached to exam package) T.S. 3.7.4, ACTION a b. T.S. 3.7.4, ACTION T.S. 3.7.4, ACTION d. No T.S. 3.7.4 ACTIONS apply, the crew should enter T.S. 3. l l
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QUESTION: 087 OS 1000.01, "Heatup From Cold Shutdown to Hot Standby", states the following Limitation:
"When changmg RCS boron concentration, pressurizer sprays should be utihzed to maintain the differential borun concetration betwem the pressurizer and reactor coolant loops to less than 50 ppm."
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What is the reason for this Limitation?
l Ensuies proper mixing occurs such that RCS loop boron samples accurately reflect actual boron concentratio b. Prevents boron stratification in the pressurizer spray nozzle Prevents an RCS dilution event from occurring during a pressurizer outsurg Ensures that the boron concentration used in the accident analyses remain in their analyzed !
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QUESTION: 088 l Which of the following statements describe satisfied Technical Specification Limiting Conditions for Operation (LCOs)in MODE 17 ; Accumulator 'D' pressure indicators read 550 psig and 560 psig respectively, b. RWST level is 485,000 gallons and RWST boron concentration is 2750 pp . NG-V-13, Safety Injection Accumulator Nitrogen Outside Containment Isolation valve fails its CLOSE stroke time requirement during surveillance testin d. CBS-V-43, Spray Additive Tank Outlet MOV, fails to OPEN on a HI-? pressure test signal during surveillance testin Page 47 of 54
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I l QUESTION: 089 In accordance with the North Atlantic Procedure Admmistration Manual (NAPA), which of the following is N_QI true regarding the review and approval of non-intent procedure changes? Changes to Staten Procedures, EOPs, and AOPs require a SORC resiew within 14 days of the Effective Dat b. Technical Specifications require that nun-intent changes are reviewed and approved or canceled within 14 days of the effective date.
l If the change is to a department procedure, and the change is needed immediately, then the four l 10CFR50.59 applicability questions must be answered by a SORC member.
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d. If after further review, it is found that the non-intent change should have been an intent change, I then the Operating Experience Manual (SSOE) should be referred to for Adverse Condition
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Report (ACR) applicability.
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QUESTION: 090 l
l Which of the following statements is a tagging requirement of MA 4.2, " Equipment Tagging and Isolation", when applied to 480V Motor Control Center (MCCs) breakers? For load work only - Open the breaker and hang the tag on the breaker operator, b. For load work concurrent with breaker work - Rack the breaker to the connected positio Position the lockout pawls to the locked position. Place the tag on the cubicle door (Maintenance is allowed to remos e the breaker from the cubicle). For breaker work only - Rack the breaker to the disconnect position, position the lockout pawls to the locked position and lock the lower lockout pawl. No Tags are required, l
t I For all work on an MCC breaker - rack the breaker to disconnect, remove breaker from the cubicle, attach proper grounding device, and hang tag on cubicle doo Page 48 of 54 l
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QUESTION: 091 Which of the followmg statements is TRUE regarding Technical Specificaten 2.1, " Safety Limits"?
I 'Ihe basis for the reactor raalant system pressure safety limit is to proted the integrity of the reactor caalant systan piping and componets, which prevents the release of radmouchdes contamed in the RCS to the contamment 4,@ l b. The reactor core safety limit is bounded by a figure using a combinaten of'Ihermal Power, pressurizer pressure, and highest loop auctioneered coolant flow, s If th: safety limit for reactor coolant system pressure is eremiM in MODE 3,4, or 5, then the pressure must be reduced to within it's limit within 30 minutes, and a Safety Lumt Violaten i report must be prepared d. The curves used in the reactor core safety limit are based on a Heat Flux Hot Channel Factor l Fo(Z) at rated thermal power of 2.3 I
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l QUESTION: 092 Which of the following correctly describes a Limitation & Setpoint of OS1000.09, " Refueling Operation"? Allow no more than one irr=AatM assembly in the cavity and canal at any one time. Fuel in the transfer car is not counted as ifit were in the canal until it is latched by the spent fuel handhng too b. The nummum RHR flow in MODE 6 is 2750 gp Maintain at least 20 feet of water above the reactor vessel flange during core alteration d. When there is fuel in the reactor, maintain greater than 120'F in the RCS to ensure that RCS temperature is within its analyzed range for shutdown margin calculation.
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I QUESTION: 093 The fuel handling SRO is performing OS1000.09, " Refueling Operation" Which of the following Technical Specification based procedure Prerequisites is stated correctly?
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I If the contamment purge and exhaust isolata system is needed for CAP or COP, then both j trams of SSPS are in OPERATE.
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b. Both RHR loops should be OPERABLE and in operation (until water level in the reactor
! cavity is greater than or equal to 23 feet above the flange, then both RHR loops should be l
OPERABLE and one RHR loop in operata).
l l The spent fuel pool and fuel transfer area (in the Fuel Storage Building) should be filled to a level at least 23 feet above the bottom of the fuel racks in the spent fuel poo d. At least one source range NI should be operable and in operation including the control room and containment audible indications.
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l I QUESTION: 094 Which of the following statements is correct regarding Radiation Work Permits? Health Physics personnel may authorize deviation from RWP requirements in the field on a
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case-by-case basis provided that the following conditions are met:
1 the deviation is documented, and the deviation permitted is not above the individual's normal approved authorit b. An example of work requiring a Specific Radiation Work Permit would be:
Entry into an area with removable contammation greater than 1,000 dpm/100 cm (beta,
gamma) or 20 dpm/100 cm (alpha) An example of work requiring a Routine Radiation Work Permit would be:
Opening a potentially contammated system
! d. Routine RWPs will normally remain in effect for the duration of the job.
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l QUESTION: 095 When performing a contauunent on-line purge to reduce contamment ammonia levels for a planned entry, what is the sequence of events nar===ry to place the COP system in semce aAcr an i approved Gaear=ie EfBuent Release pernut is received? '
5 Set the COP rnantian monitor seapoets to 1 x 10 cpm, unlock and energize the circuits for !
l COP valves 1-4, and place the COP system in semce to mamtain 15.2 to 15.3 psi '
l b. Enter the Tech Specs for the COP radiaten momtors, set the COP radiation monitor setpoints
! to 1 x 10 cpm, unlock and energize the circuits for COP valves 1-4, and place the COP ;
system in service to mamtam 15.2 and 15.3 psia.
l l Enter the Tech Specs for the COP radiation momtors, set the COP rahatiani momtor setpoints
l to 1 x 10 cpm, cater the Tech Spec for the COP valves, unlock and energize the circuits for :
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COP valves 1-4, and place the COP system in semce to maintain 15.2 and 15.3 psi l d. Set the COP radiation monitor setposts to 1 x 10' cpm, enter tle Tech Spec for the COP i valves, unlock and energize the circuits for COP valves 1-4, and place the COP system in j service to maintain 15.2 to 15.3 psi l
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QUESTION: 096 Which of the following describes the flow path for performing an air purge of containment to reduce hydrogen concentration while in the emergency operating procedures? From the service air system into containment via the normal H2 analyzer sample lines. Out of contamment via CGC-V14 and CGC-V28 to the containment enclosure emergency exhaust filters and then out the plant ven .
b. From the service air system via normally locked closed valves into contamment. Out of contamment via CGC-V14 to the inlet of the Train A contamment enclosure emergency exhaust filter and then out to PAH-F-1 From the service air system via normally locked closed valves into containment. Out of containment via CGC-V14 and CGC-V28 to the inlet of the containment erelosure emergency exhaust filters and then out the plant vent, d. From the service air system via normally locked open valves through a containment isolation check valve into containment. Out of containment via CGC-V14 or CGC-V28 to the containment enclosure emergency exhaust filter to PAH-F-16.
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QUESTION: 097 i The plant is at 100% power when the following events occur: 1 l
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Power is lost to MCC-531, hsiilog DRPI indication and causing an indicated Red Path on the Suberiticality CSF Status tree. Soon aAerwards, the Reactor Protection system generates an ,
automatic reactor trip signal resulting in the reactor trip breakers openin Which of the following actions should be performed by the operating cre j Rod bottom lights are lit, the primary side operator should attempt a manual trip using the manual trip switches. Reactor trip is then verified so the crew should proceed on to verify
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turbine tri b. Rod bottom lights are not lit, the primary side operator should attempt a manual trip using the
- manual trip switches. The crew should go to FR-S.1, " Response to Nuclear Power
Generation /ATWS", Step I
- Rod bottom lights are not lit, but reactor trip is verified by reactor trip breaker open indication and neutron flux levels decreasing. The crew should proceed on to verify turbine tri d. Rod bottom lights are not lit, the crew should exit the procedure to FR-S.1, " Response to Nuclear Power Generation /ATWS", Step 1, due to the red path that exists on the subcriticality CSF status tree.
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QUESTION: 098 h operating crew is responding to a LOCA in contamment and have exited E-0 to FR-P.1 due to an ORANGE path on the Integrity CSF status tre Whde perfornung this procedure the crew nobees that there is an ORANGE path indicated on the
? N-w CSF status tree and the Core Cooling CSF status tre l l
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These indications are followed 10 minutes later by a RED path on the Heat Sink CSF status tre Which of the following procedure flowpaths should be used by the operating crew? Complete the act:ons of FR-P.1, and proceed to FR-C.2. Suspend the actions of FR-C.2 when the red path on the heat sink CSF tree is nobeed and proceed to FR-H.l.
b. Complete the actions of FR-P.1, and proceed to FR-Z.l. Suspend the actions of FR-2.1 when the red path on the heat sink CSF tree is noticed and proceed to FR- Suspend the actions of FR-P.1 and proceed for FR-C.2. Complete the actions of FR-C.2 and then proceed to FR- Suspend the actions of FR-P.1 and proceed to FR-C.2. Suspend the action of FR-C.2 when the red path on the Heat Sink CSF tree is noticed and proceed to FR-H.l.
l l QUESTION: 099 For an ATWS event where a loss of normal feedwater has occurred, FR-S.1, " Response to Nuclear Power Generation /ATWS", has the operator verify Turbine trip at step # What is the basis for this requirement? h turbine is tripped to prevent a controlled cooldown,of the RCS due to the steam flow that the turbine would require.
l b. Turbine trip is required (within 30 seconds) to maintam SG inventory.
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! - 'Ihe turbine is tripped to increase primary system pressure and temperature so that feedback mechanisms will add negative reactivity to the core d. The turbine is tripped to decrease the probability of delayed operator action due to misdiagnosis of the A'IWS even Page 53 of 54
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QUESTION: 100 5 He plant has experienced a large LOCA in contamment and the operating crew is performing i actions in E-1, Ioss of Reactor or Secondary Coolant. He CSF status trees have the following indications;
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S - Yellow; C - Orange; Z - Orange; P - Orange, H - Re ,
What Emergency Plan classification should the Shift Manager make?
1 Site Area Emergency, based on P - Orange or S - Yellow.
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b. General Emergency, based on C - Orange or H - Re i
' Site Area Emergency, based on Z - Orange amL S - Yellow d. General Emerg~ency, based on C - Orange and H- Red.
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i PLANT SYSTEMS i -
] 3/4.7.4 SERVICE WATER SYSTEM / ULTIMATE HEAT Sile: '
4 l LINITING t'namITION FOR OPERATION
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3.7.4 The Service Water System shall be OPERABLE with:
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j An OPERABLE service water panphouse and two service water loops with one OPERABLE service water pump in each loop, i
i An OPERABLE mechanical draft coolini tower and two cooling tower j
j service water loops with one OPERAS.E cooling tower service wat.er
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pump in each loop, and -
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l A portable cooling tower makeup system stored in its design operational readiness stat APPLICABILITY: HODES 1, 2, 3, and I i
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i - With one service water loop inoperable, return the loop to OPERABLE
status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at least NOT STAlWBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
j b. With one cooling tower service water loop or one cooli tower cell
inoperable, return the affected loop or cell to status within 7 days, or be in at least NOT STAlWBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in
! COLD SHUTDOWN within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> With two cooling tower service water loops or the mechanical draft cooling tower inoperable, return at least one loop and the mechanical l draft cooling tower to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or be in at i
least HDT STAlWBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the
following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> l With two loops (except as descr.ibed in c) or the service water
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pumphouse inoperable, return at least one of the affected loops and
! the service water to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or be i
in at least HDT within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within
- the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> i With the portable tower makeup pump system not stored in its design
operational readiness state, restore the portable tower makeup pump i system to its required condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or continue i
operation and notify.the IIIC within the following I hour in i accordance with the requirements of 10 CFR 50.72 of actions to ensure i
an adequate supply of makeup water for the service water cooling
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tower for a minimum of 30 days.
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j SEABROOK - UNIT 1 3/4 7-13 Amendment No. 32
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! TzCENICAL CLARIFICATIcet Page 1 of 3
SECTION I - REQUEST "OR CLARIFICATION
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Originator: Gene St. Pierre Date: 6/6/96 Technical Clarification Title: Inoperable Service Water Loops ,
i Technical Clarification No.: TS'-66
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j Type of Clarification:
I Tech' Spec E UFSAR' (Excluding 17.2) O UrSAR 17.2 O Licensing O
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REQUEST FOR CLARIFICATION: (Attempt to state the request as a question.)
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tower loop in the same train are inoperable?
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i i 1 CONCURRENCE: < h Group Manager 7h&
'Date
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SECTION II - INITIATION l RECEIVED BY REGULATORY COMPLIANCE: 7 J- ~
h Regul Cdeplianceppetvisor Date
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NARC FORM 2-4A
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Rev. 47 Page 1 of 2
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j TECHNICAL CLARIFICATION (Continuid)
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- SECTION III - EVALUATION I
l The Seabrook Station Service Water System (SW) consists of 4 loops, Train A and B i SW loops and Train A and B cooling tower SW loop '
When one SW loop and one SW cooling tower loop in the same Train are inoperable, the correct Action to enter is Technical Specification 3.7.4, Action d., which i allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to return at least one of the affected loops to operable statu I Action d. is also applicable for any combination of SW loops and cooling tower SW loops (except 2 cooling tower SW loops) provided the functional equivalent of one full Train i.e. 2 of 4 loops is operabl (Example: Train A SW loop and Train B SW cooling tower loop) l
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Prepared By: . &C!f6 Concurrence: [ 4
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SECTION IV - REVIEW AND APPROVAL (Check Appropriate Boxes) -
D g0engingAMag ~
Y l= RAM //r dgrg (Date)
D (Date)
o % lm 1 ettor abMb-(Datd)
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O O (Date) (Date)
(Date)
SORC MEETING NO.: 477 DATE: B!7!h h'
NARC FORM 2-4A Rev. 47
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i ATTACHMENT 2 l
SIMULATION FACILITY REPORT
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j FACILITY LICENSEE: North Atlantic Energy Service Corporation l FACILITY DOCKET NO: 50-443 Operating Tests Administered: 10/01 -10/02 1996
- This form is to be used only to report observations. These observations do not constitute
audit or inspection findings, and are not, without further verification and review, indicative of noncompliance with 10 Cs Ct 55.45(b). These observations do not affect NRC j
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certification or approval of tc.e simulation facility other than to provide information, which !
I may be used in future evaluations. No licensee action is required in response to these
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observation )
- During the validation and performance of the simulator examination scenarios and Job i j Performance Measures, the following item was observed.
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An invalid position indication was illuminated for a D moisture separator heater steam valve. This was overridden by the simulator operator for the scenario.
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An improvement in modelling capability was noted in a scenario involving FR-C.1.
j in this scenario, core exit thermocouple temperatures of 1650 dog F were achieved, j then brought down with accumulator injection and SI. The simulator response i appeared realistic to the examiners and allowed the running of a scenario resulting in more severe core conditions than had previously been attainable.
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ATTACHMENT 3 1
ANSWER KEY l l
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- NRC LICENSE EXAM - SEABROOK SRO (EXAM DATE
- 9/30/96)
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ANSWER KEY , a 26. b S d 76. -b !
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28. c 5 c 7 d b 29. d 5 c 7 c b 30. a 5 e 8 c i d 31. b 5 a 8 d d 32. b 5 c 8 c
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