IR 05000423/1997004
| ML20217P749 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/12/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20217P748 | List: |
| References | |
| 50-423-97-04OL, 50-423-97-4OL, NUDOCS 9708280221 | |
| Download: ML20217P749 (255) | |
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U. S. NUCLEAR REGULATORY COMMISSION
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Docket No.:- 50-423 i
Report No.: 97-04- I
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License No.: NPF-49
Licensse: Northeast Nuclear Energy Company
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P. O. Box 128 Vvaterford, CT 06385- .
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. h i Facility: Mlistone Nuclear Power Station, Unit 3
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Location: Waterford, Connecticut Dates: July 7 - 11,1997 ,
J Chief Examiner: L. Briggs, Senior Operations Engineer / Examiner, Region i Examiners: J. D' Antonio, Operations Engineer / Examiner, Region 1-F. Jaggar, NRC Contract Examiner, LITCO Approved By: Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety
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-9708280221-970812 PDR ,ADOCK 05000423 V_ PDR;
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EXECUTIVE SUMMARY Millstone Nuclear Power Station, Unit 3 Inspection Report No. 50-423/97 04 Ooerations Four Millstone Unit 3 senior reactor operator (SRO) upgrade candidates and four reactor operator (RO) candidates were administered initial licensing examinations. All candidates passed all poitions of the license examination.
Overall, candidate performance during the exam was determined to be good, There were no candidate weaknesses or actions that would compromise safe plant operations observed by the examiners. The NRC examiners observed good three point communications among the operating crews, The three point communications were effective and ensured that information and directions were understood by the operating crews but appeared somewhat cumbersome during lengthy information exchanges, ii i
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Report Details I. Operations
. 05 Operatoe Training and Qualifications
- 0 Reactor Operator and Senior Reactor Operator Initial Examinations 4- Scope
- The exam was prepared by Millstone Unit 3 personnel in acwrdance with the i guidelines in interim Revision 8, of NUREG-1021, " Examiner Standards." The examiners administered initial operating licensing exams to four Unit 3 senior reactor
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operator (SRO) upgrade candidates and four Unit 3 reactor operator (RO)
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candidates. The written examinations were administered by the facility's training organization.
' Observations and Findinan
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The results of the SRO and RO exams are summarized below:
4 SRO Pase/ Fail RO Pass / Fall Total Pass / Fail Written 4/O 4/0 8/0
Operating 4/O 4/0 8/0 Overall 4/O 4/0 6/0
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The written examinations, job performance measures (JPMs) and simulator scenarios were developed by Millstone Unit 3 representatives in accordance with
- Interim Revision 8 of NUREG 1021 " Examiner Standards." The exam development
team was comprised of Millstone Unit 3 training and operation's representative Allindividuals signed onto a security agreement once the development of the
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examination commenced. The NRC subsequently reviewed and validated, along with Millstone Unit 3 personnel, all portions of the proposed examinations. Various
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. changes and/or additions to the proposed exams were requested by the NRC
! following their review. Millstone Unit 3 personnel subsequently incorporated the
. -NRC's comments and finalized the examination The written exam was administered on July 7,1997. Both the SRO and RO written examinations consisted of 100 multiple choice questions. There was one comment by the utility concerning one question on the RO written exam that had two correct answers. ' The NRC reviewed the utility's comment and justification and accepted
, the question with two correct answers. There was also a typographical error on the
- SRO answer key for one question which war. identified and corrected during the-
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- grading of the exa .
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The operating exams were conducted from July 7-10,1997. The operating exams consisted of two simulator scenarios conducted twice and ten JPMs for the RO candidates and five JPMs for the SRO upgrade candidates. All JPMs were followed up with two system-related questions. All candidates were also examined using a mix of questions and JPMs to evaluate the adminittrative requirement portion of the exa Based on the grading of the written exam, the following question subject areas were missed by more than half of the RO, SRO, or a combination of both applicants for the common questions. This indicated a weakness in the general understanding of the subject are * Controls at the hot shutdown panel (RO)
- Auxiliary feedwater flow versus steam pressure (RO)
- Shrink and swell effects of a reactor coolant pump trip (RO)
- Loss of Bus 34A and impact on plant operations (RO)
- Auxiliary feedwater flow versus decay heat level (RO & SRO)
- Pressurizer pressure heater operations (SRO)
The facility should review each of the above subject areas (and several additional areas identified during the writtei) exam grading) to correct possible weaknesses and implement programmatic changes if necessary. The facility should pay particular attention to the safety related auxiliary feedwater system for both RO and SRO candidate Simulator performance by the candidates was good. The examiners noted that crew briefings were routinely performed by the SROs. Three point communications, in general, were good; however, lengthy information exchanges appeared somewhat cumbersom Simulator deficiencies that occurred are discussed in detail in Attachment 4 of this report. The facility should place increased emphasis on maintaining simulator fidelity and problem free operation to mitigate the negative training impact on licensed operators, in the administrative segment of the operating portion of the examination, a combination of questions and administrative job performance measures (JPMs) were used to address admin lstrative topics. The examiners determined that the candidate performance was goo , ---------J
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3 Conclusions The candidates performed well on both the written and operating exams. The SRO upgrades were issued licenses. The RO candidates will be issued licenses when NRC and utility program requirements, that cannot be completed while the plant is shutdown, are completed. The candidates appeared to be well prepared for the exams. The training department did a good job in adhering to the examiner standards and in developing the exam materials needed to administer the exam The simulator experienced more problems than normally oberved during an NRC exam.
E8 Review of UFSAR commitments A recent discovery of a licensee operating their f acility in a manner contrary to the updated final safety analysis report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and /or parameters to the UFSAR descriptions. While performing the preexamination activities discussed in this report, the inspectors reviewed applicable portions of the UFSAR that related to the selected examination questions or topic areas. No discrepancies were identified as a result of this revie V. Manaaement Meetinas X1 Exit Meeting Sutnmary On July 11,1997, the examiners discussed their observations from the exam with Millstone Unit 3 operatius and trainir.g management representatives. The examiners discussed candidate perfvmance, detailed in paragraph 5.1.b above, concerning communications, preliminary writ %n exam results, and simulator performance during the exam. The examiners also expressed their appreciation for the cooperation and assistance that was provided during both the preparation and examination week by licensed operator training personnel and operations personnel. Millstone personnel present at the exit meeting included the following partial listing in alphabetical order:
Millstone M. Brothers, Vice President, Millstone Unit 3 B. Carns, Senior Vice President, Millstone D. Lazarony, Supervisor Operator Training, Millstone Unit 3 R. Lueneberg, Senior Instructor, Millstone Unit 3 C. Miller, Operator instructor, Millstone Unit 3 L. Palone, Assistant Operations Manager, Millstone Unit 3 t R. Rowe, Licensed Operator Initial Training Coordinator, Millstone Unit 3 J. Smith, Manager Operator Training J. Stankiewicz, Recovery Training Manager R. Stotts, Assistant Supervisor Operator Training, Millstone Unit 2 J. Thayer, Vice President, Nuclear Engineering and Support
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MBC Larry Briggs, Senior Operations Engineer, Chief Examiner Joseph D' Antonio, Operations Engineer
- Russell Arrighi, Resident inspector, Millstone Unit 3 Attachments:
1. Millstone Unit 3 SRO Written Examination w/ Answer Key
- 2. Millstone Unit 3 RO Written Examination w/ Answer _ Key 3. NRC Resolution of Millstone Unit 3 Written Exam Comment 4. Simulation Facility Report
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Attachment 1 Millstone Unit 3 SRO WRITTEN EXAM W/ ANSWER KEY
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SRO1 PLANT CONDITIONS:
- A loss of all AC power has occurred
- SBO diesel is supplying "A" Train 4160 V buses e Charging and letdown are secured
. Pressurizer levelis 10%
- RCS subcooling is 35'F
. RCS pressure is 1600 psia
. Highest CET is 574*F
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. Containment temperature is 115*F Which of the following actions should be taken by the crew? Initiate RCS cooldown by depressurizing Steam Generator Consult with the DSEO and depressurize the RCS to inject the accumulator Perform ECA-0.2, Loss of All AC Power Recovery With S1 Require Consult with the DSEO and align one charging pump in the injection mod ANSWER:
' Consult with the DSEO and align one charging pump in the injection mod REFERENCE: ECA-0.3 Ceution prior to Step JUSTIFICATION: Distracters "A" and "C" are incorrect because these actions would be performed if the crew could not energize any AC Emergency Bus from the SBO Diesel at Step 7 of ECA- Distracter "B" is incorrect because depressurizing the RCS is a strategy used past Step 7 in ECA-0.0. The operators will transition from Step 7 of ECA-0.0 to ECA-0.3 when the bus is re-energize Per the Caution prior to Step 6 in ECA-0.3 Distracter "D" is correc K/A: 062 K2.01 Loss of power to major loads Exam Item 687
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SRO2.
i The plant is in MODE 6. Core ofiload is in progres The alarra circuitry for one of the Spent Fuel Pool area monitors fails. I & C is investigating. No other operator actions have been take Which of the following describes the required ACTION, if any, to be taken regarding the fuel movement in progress? No ACTION required, all LCOs are satisfied, fuel movement may continu ! Fuel movement may continue for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while adjusting the setpoint to within the limi Fuel movement may continue for up to 30 day Fuel movement must be suspended until an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool are ANSWER: Fuel movement must be suspended until an appropriate portable continuous ,
monitor with the same Alann Setpoint is provided in the fuel storage pool are ,
REFERENCE: Technical Specification 3.3.3.1, ACTION 28 JUSTIFICATION: A incorrect, Tech Spec Minimum channels required is 2, cmrently only B incorrect, the setpoint is not the problem, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit does not appl C incorrect, ACTION required is to provide the backup or suspend fuel movemen D correct, until the backup is provided, no fuel movement can occur, then the monitor must be retumed to OPERABLE status in 30 days or suspend fuel movemen K/A: 072 K3.02 Effects on fuel handling operations OBJECTIVE: RMS05C, RMS07C, RMS08C New question -
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-- SRO 3 Containment purge and exhaust are in operatio Fuel Drop monitor,3RMS'RE41 goes into high alamt Which of the following describes the automatic response of the system to the alarm? All four containment purge supply and exhaust valves close onl All four containment purge supply and exhaust valves close and the running exhaust fan,3HVR-FN4A or 4B stop One supply and one exhaust valve closes onl One supply and one exhaust valve closes and the running exhaust fan. 3HVR-
. FN4A or 4B stop ANSWER:
- C. - . One supply and exhaust valve closes onl REFERENCE
- LSK 22.27B -
JUSTIFICATION: - Purge isolation takes place if EITHER fuel drop monitor goes into-alarm. (One monitor isolates the inside containment valves, the other monitor isolates the outside containment valves.) -The' fans do not get tripped, (C ~ correct only) ,
OBJECTIVES: RMS05C; RMS07C; RMS08C K/A:' ape 061 Al.01, Automatic actuation of ARM 3.1/3.6
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SRO 4 Plant Cor.ditions:
.'e ! An ATWS has occurred due to a fire in the switchgear room ie- The reactor could not be tripped locally because the breakers are fused together
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e- Reactor power is 40% and decreasing-e Tav is increasing
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e' Pressure is being maintained by the PORVs cycling around 2350 psia The PREFERRED sequence to add negative reactivity to shutdown the reactor is: Perform an immediate boration, drive rods in auto or manual, initiate a safety injection
'# Drive rods M auto or manual, shift charging pump suction to the RWST, perform an immediate boratio Perform an immediate boration, maximize charging flow, drive rods in auto or manua Shift charging pump suction to the RWST, drive rods in auto or manual, perform an immediate boration,.
ANSWER:-
' i Per'orm an immediate boration, maximize charging flow, drive rods in auto or manua l REFERENCE: . FR-S.11 JUSTIFICATION: The first 4 steps of FR-S.1 contain the preferred sequence for
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OBJECTIVES: . FS102C
'K/A: 029 EK3.12 Actions in EOP '
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SRO5 The plant is in Mode 1. The following personnel are in the control room The CO who is the " Operator at the Controls", the US and the SM. The other CO i on a plant tour.
When, if at all, may the US be considered the " Operator at the Controls?" F If the " Operator at the Controls" must leave the red carpeted area (Operations Area). If the " Operator at the Controls" is around back of the main control board acknowledging an annunciator. If the Shift Manager is in the surveillance area and a proper turnover occur The US cannot be considered the "Operato: at the Controls".
ANSWER: If the Shift Manager is in the surveillance area and a proper turnover occur REFERENCE: COP-200.1 Section 1.7 & COP-200.1 Attachments 5 & 6 Tech. Specs 6. JUSTIFICATION: "A" incorrect due to CO not restricted to Operations area (COP 200.1, Section 1.6)
"B" incorrect because CO allowed behind boards to acknowledge main control board annunciators and is still the " Operator at the Controls"(COP 200,1, Section 1.6)
"C" correct. With the SM in the surveillance area and with a proper relief the US can become the " Operator at the controls"
"D" incorrect. Section 1.7 of COP 200.1 states I licensed operator in the surveillance area at all times in addition to the SM or the US in the control room. If other personnel are to be considered in the surveillance area, they shall meet the normal relief requirement OBJECTIVES: NAD407 K/A: 2.1.2, Operator responsibilities Major rewrite of Question 2189
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SRO 61 I
. Safety-related plant equipment is known to have been operated in a manner which had the -
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- potential to damage the equipmen Which of the following des .ibes an action which must be taken in accordance with COP 200.1 CONDUCT OF OPERATIONS?
. Shift Manager shall notify the Duty Officer and initiate a C A notification must be made to the NRC within twenty-four hours of the even The Unit Supervisor shall initiate a priority;1 AWO for maintenance to investigat The Operations Manager shall notify the Technical Services Engineering Manager and cognizant system enginee ANSWER:
, Shift Manager shall notify the Duty Officer and initiate a C '
REFERENCE: C OP 20 JUSTIFICATION: 'If plant equipment is' known to have been operated in a' manner which had the potential to damage the equipment, the following actions shall be taken:
Operator shall infomt the US of the proble US shall direct equipment or plant be placed in a safe conditio SM shall describe the problem in the SM log. -
If potentially damaged equipment is safety related, the following _
additional actions shall be taken:
' SM/US- shall check _ applicability. . of Technical Specifwations or Technical Requirement SM shall inform the Duty Office SM shallinitiate an C SM shall initiate appropriate . investigative action to
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determine status of patentially damaged equipmen '
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i OlljECTIVES: NAD413 K/A: 2. .7/ .1.7, Evaluate performance / judgment 3.7/ Question #3093
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SRO7-Which of the following is a difference between a dual verification and independent verification?
- An Independent Verification can be performed by the same individual perfonning the initial verification. Dual Ve'ification requires two individual An Independent Verification is performed prior to the tack being performed. Dual Verification is performed upon completion of the tas Dual Verification can be performed using the " time and distance" method Independent Verification is usually performed concurrently with the initial verifie D, Dual Verification is performed concurrently with the task. Independent
?/erification is performed after the task or evolution has been complete ANSWER: Dual Verification is performed concurrently with the task. Independent Verification is performed after the task or evolution has been complete REFERENCE: WC-6, Attachment 2 JUEnFICATION: DUAL VERIFICATION: Dual verification is performed concurrently with the task. In general, dual verification is ,
performed when an action could result in an immediate threat to safe and reliable plant operation. For dual verification, concurrently means the " performer" and the " verifier" together determine and agree the work location and component are correct for the specified action. And, the task, to the best of their knowledge, will result in 'he desired outcome. The person performing the task and the person performing the verification both must positively identify the component, determine the actual and required position or state, and agree on the method to be tised prior to the action taking place. The verifier essentially completes the verification of the intended action, before the taion is undertaken and then witnesses the task performed. It thould be noted that if a specific task requires a dual verification during its performance, the need for a separate independent verification of the same evolution is not required. [v Comm 3.1]
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INDEPENDENT VERIFICATION: Independent verification is performed after a task or evolution has been completed to ensure it has been performed in the correct location, on the correct equipment, or the desired results have been obtaine ' An independent verification is also performed periodically to ensure systems, equipment, components, etc. are still in the condition they were left following their last manipulation or verificatbn. To satisfy the requirements for an independent verification, when plant conditions or the situation allows, a good rule of thumb to ensure real independence, is to apply the " time end distance" method. This requires the independent verifier to not visually observe the person who initially perform the tas Conducting the verification in this manner will alleviate the possibility that the performer goes to the wrong item or place and y{
the independent verifier, watching the performer, simply goes to b the same (wrong) location and verifies the performer's actions were completed correctly (at the wrong location 1). In lieu of using the time and distance technique, as with all verifications, using attentiveness and attention to detail during their performance will
prevent inadvertent problems, it is the responsibility of the verifier to ensure he or she is in the correct location, checking the required equipment or components, and determining if they meet the specilled acceptance criteri OlUECTIVES: NAD613 K/A: 2.1.29, Conduct / verify valve lineups Question #3268
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~SRO 8 -- _
-When lia second Main Condensate Pump required to be operating? , ? A, - : When pumping the Main Condenser Hotwell, When a Main Feed Pump is operating.-
E C, .When the Mr.in Circulating Water Pumps are operatin . Anytime steam is being released from the Steam Generator ANSWER:
B,i _When a Main Feed Pump is operating, REFERENCE: OP 3319A section 4,1
- JUSTIFICATION': Pumping the hotwell requires only one condensate pump and is
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long recycle (no MFPs operating) - ('A' incorrect)
Operation of the Cire pumps requires that tube sheet seal water be :
applied, This can be supplied from the CNS system during plant shutdown conditions and does not require use of the condensate
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pumps ('C' incorrect)-
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being supplie OP 3319A section 4,1 states "A minimum of two condensate pumps must be running while feeding steam generators with main
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feed pumps." This requires that a 2nd condensate pump be
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operating any time a main feed pump is operating ('B' correct)
OBJECTIVES: CNM06C (a)
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iK/A:- : 2.1.32, System limitations / precautions -
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SRO9 l The following plant conditions exist:
- The plant was operating at 100% power when a large LOCA occur * A Containment Depressurization Actuation Signal (CDA)is generated due to high containment pressur * All safeguards equipment operates as designed except the B EDG fails to auto start and cannot be starte minutes later, while performing E-0, Reactor Trip or Safety injection, all offsite power is los Approximately 2 mir.utes later, the RO checks power availability to the Safeguards equipmen What should be the status of the RSS pumps? All four pumps should be runnin Only RSS pumps "A" and "C" should be runnin Only RSS pump "A" should be runnin None of the pumps should be mnnin ANSWER: Only RSS pumps "A" and "C" should be runnin REFERENCE: LSKs 24-9.4A,24-9.4B,24-9.4Q, and 27-11J JUSTIFICATION: The CDA signal starts the 11 minute timer to start the pump Once started, if an LOP occurs, the pumps are restarted by the sequencer after 60 second A is incorrect, because the "B" EDG is not running C is incorrect Both "A" train pumps will start. Only the "A" pump would start ifin the SI/Recire mod D is incorrect because the "A" train pumps will star .- -_- . _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ .__ _
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OBJECTIVES: CDA06C(1); CDA07C K/A: 103 Kl.08 SIS /CDA including reset Bank item
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SRO 10 ' "
' A Safety injection has occurred and the crew was conducting a brief at the end of Step 14 i of E-0 when a complete loss of Off-Site power occur The Emergency Diesel Generators start, their output breakers close to restore power to
- the vital busses, and the sequencers complete their sequencing on ofload What is the status of Containment Air Recirculation (CAR) fans?
- All the CAR fans will be mnnin The "A" and "B" CAR fans will be runnin Only the "C" CAR fan will be runnin D.- None of the CAR fans will be runnin ANSWER:
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REFERENCES: LSK 22-27.C, LSK 24-9.4a
' JUSTIFICATION: - The crew has stopped the "C" CAR fan at Step 12 of E-0 ("A" and
"C" wrong).
The "A" and "B" fans will start on the SI signal, and when the LOP
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39 seconds. ("B" correct, "C" and "D" wTong.).
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OBJECTIVE: CVS03C (6.4)
K/A: 022 A3.01 Initiation of Safeguards .
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SRO 11 During an uncontrolled rod withdraw from 175 steps on "D" bank, the fmal steady state actual reactor power will , and RCS Tave will . (Assume no operator action /reacter does not trip) Increase, increase Increase, remain the same Remain the same, decrease Remain the same, increase ANSWER: Remain the same, increase REFERENCES:
JUSTIFICATION: The p reactivity addition of the control rods will cause power and Tave to initially increase. The increasing Tav will drive power down to approximately the initial level. The end result approximately at the same power level with an elevated temperature. The temperature increase will feedback negative reactivity to offset the positive rcactivity originally added by rod motio OBJECTIVES: ROD 06C (e)
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K/A: 00lKl.03, Relationship of reactivity and Rx power to rod movement New Question
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SRO 12 Given the following conditions:
. MP3 is holding at 75% power following a refueling outag . Rod control is in automati . Rod height is 220 on Bank Power range channel N-44 fails hig WHICH ONE of the following describes the response on the rod control system? Rods will drive in until a Tav-Tref error develops which will result in the rods driving ou Rods will not mov Rods will continuously drive in unl:ss stopped by operator interventio Rods will drive in until the power mis-match circuit output decays awa ANSWER: Rods will drive in until the power mis-match circuit output decays awa REFERENCES:
JUSTIFICATION: The over power rod stop will prevent outward motion (part a is incorrect).
Rod stops don't prevent inward auto motion (B is incorrect).
Temperature error and power mismatch circuit output decaying away will stop rod motion. (D is correct)
C is incorrect because the rods will eventually stop driving in without operator interventio OBJECTIVES: NIS07C (g)
K/A: 001 Kl.05, Cause/effect between CRDS and NIS and RPS l
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Modified item 2207 SRO 13 During a Reactor Startup, you must ensure the reactor goes critical above the rod inscion limit. The reason that this is safety significant at this time is to ensure: Peak cent;rline fuel melt temperatures are not exceeded if you ejected a rod during startu Suflicient positive reactivity is available by the control rods to offset power defect on the power escalatio No power tilt is introduced in the core because Bank C is partially withdrawm with Bank D still on the botto The reactor will have adequate shutdown margin for a steam line break acciden ANSWER:' The reactor will have adequate shutdown margin for a steam line break acciden REFERENCES:
JUSTIFICATION: Tech Spec Bases. The hot zero power Rod Insertion Limit is to ensure reactor DNB limits are not exceeded on large Steam Line Break and the reactor doesn't return to critical on small steam line break, (D is correct)
A is incorrect because HCP rod ejection do not cause fuel melt concerns, rod ejection is limiting at HF Power defect must be offset by rods and dilution on power escalations. Normal rod height for criticality is about Bank D at 150 steps. (B is incorrect).
QPTR is not a concern less than 50%. QPTR insures FQ (LOCA)
calculations are within limits. This is not a concern at hot zero power. (C is incorrect).
OBJECTIVES: ROD 03C (c); kOD08C (b)
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K/A: 001 K5.08, Rll Setpoint New Question
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SRO14 PLANT CONDITIONS:
- A sinall break LOCA has occurred
- An RCS cooldown has been initiated by dumping steam to the atmospher Which of the following statereents describes the optimum reactor coolant pump configuration, and the basis for this configuration? All RCPs sh r x stopped to minimize RCS inventory loss following break uncover, and , . .vent steam valding in the reactor vessel on subsequent RCS depressurization, i One RCP rhould be run to produce effective heat tramfer and RCS pressure control, yet minimize RCS heat input.
< One RCP chould be run to produce effective heat transfer and RCS pressure control, yet minimize RCS invutory los Two RCPs should be run to ensure symmetric heat transfer to the intact sos, to enhance RCS pressure control, and to prevent steam voiding in the reactor vessel /
head on the subsequent RCS depressurizatio ANSWER:
1 One RCP rhould be run to produce effective heat transfer and RCS pressure control, yet minimize RCS heat inpu REFERENC!! ES 1.2 llackground JUSTlFICA'110N: Forced coolant flow is the preferred mode of operation to allow for nomial RCS cooldown and provide PZR spray. All but onc should be stopped to minimize heat input into the RCS. "A" is incorrect because if all RCPs are stopped voiding snuld occur in the vessel head during the depressurization. "C"is incorrect because the RCS inventory loss is based on the existing differential pressure and not on the forced flow through the RCS. While the reasons in distracter
"D" are correct, the procedure does not address running two (2)
RCP OBJECTIVES: bl203C i
R/A: WEST 03 EAl.3, Desired results .
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SRO 15 Initial Plant Conditions:
e - The plant is at 100% power at EOL e - Bank D rods are at 225 steps
--*- All controllers are in automatic
'.The follow;ing MB annunciators air received:
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TAVEfrREF DEVIATION-
- ROD POSITION DEVIATION
- POWER RNO CIIANNEL DEVIATION
- POWER RNO FLUX RATE HI Based on the above indications, which of the following events has occurred? The contolling first stage perssure transmitter has failed hig An NIS power range upper detector has failed lo C.- A rod position indication channel has failed lo _ A control rod has dropped into the cor ANSWER:' A control rod has dropped into the cor REFERENCE: AOP 3552 i i JUSTIFICATION: The dropped control rod will ger.erate the negative rate signal and the NIS power channel deviation because of the power tilt in the -
core. The dropped rod will cause the plant to cooldown causing the Tave. Tref deviation alann. (D is correct) ,
ODJECTIVES: A5203C; ROD 07C(c)
K/A: . 003A2.03, Dropped rod using in-core /ex core inst., loop tem ~ New Question -
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SRO 16 Which of the following reprewnts the safeguards signal that actuates tle listed CVCS system valves?
RCP Seal Return Charging Flow Control Letdown isolation isolation CIB CIA Si Bi CIA SI St CIA- SI CIA D.- CIB CIA CIB ANSWER: CI S1 CIA-REFERENCES:
JUSTIFICATION: The seal retum MOVs close on CI A and CVCS letdown and charging flow control isolation valves close on an S OBJECTIVES: ECC03C (6.1)
K/A:- 006 K4.09 modified test item 2382
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SRO 17 Which of the following describes how the Emergency Diesel Generators are placed in the
" Speed Droop" mode of operation? Automatically selected when started hianually from hialn Iloard 8 and no safeguards signals are presen I Taking the mode selector switch on hiain 130ard 8 to " Unit". Taking the mode selector switch on hiain 11oard 8 to " Parallel". Automatically selected for all Emergency Diesel Start ANSWER: Taking the mode selector switch on hiain 130ard 8 to " Parallel".
REFERENCE: LSK 24 9.3C JUSTIFICATIGN: Speed droop is selected when th: mode selector switch is in the
" Parallel" position, providing no safeguard signals are presen Speed Droop allows the diesel speed to change allowing load sharing. When the diesel is the sole source of power,(i.e. in Unit)
no speed droop is allowed to ensure equipment is running at full spee OBJECTIVES: EDG02C(v); EDG06C(f)
K/A: 064 A4.06, manual start, stop of EDO New question
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SRO 18 A large break LOCA occurred on unit 3. All systems responded as designed. When transitioning from E-0, the US directs you to perform an evaluation of the CSFs. You identify an orange path on integrity, a red path on containment and a yellow path on core cooling, a yellow path on heat sink, and a yellow path on inventor Based on this information, the operating crew should: Transition to E Transition to FR Transition to FR Transition to FR- ANSWER: Transition to FR REFERENCES:
JUSTIFICATION: First transition to the highest priority critical safety function procedure which is FR Z 1 - the only red path (correct), other answer are incorrect in accordance with EOP rules of usag OBJECTIVES: E0004C R/A: 2.4.1, EOP entry control /immediate actions
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Modified Question 533
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SRO 19 The following plant conditions exist:
- Plant is in Mode 5 with allloops FULL
- Train "A" of RiiR is in service providing shutdown cooling
- Train "11" electrical outage is in progress and is expected to last another 6 hr * Due to a plane crash in the switchyard, all ofTsite power is lost
- The "A" EDO starts but fails to load onto Bus 34C
- nc Unit Supervisor enters the appropriate procedure for loss of shutdown cooling
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llow should the crew re-establish shutdown cooling? Transition to EOP 3501, Loss Of All AC Power (Mode 5,6 and Zero), and testore power to be able to resttui coolin Satrt and align the Silo Diesel onto 13usses 34A/C and then perform Attachment B, Loss of Shutdown Coolong And/Or RCS Inventory Mode 5, of EOP 350 Initiate decay heat removal as the RCS heats up by duaping steam from at least one available S Inject avai!able St Accumulators into the RCS, ANSWER: Transition to EOP 3501, Loss Of All AC Power (Mode 5,6 and Zero), and restore power to be able to restart coolin IEFERENCE: EOP 3505 Step 1 JUSTIFICATION: The appropriate procedure for loss of shutdown cooling in Mode 5 is EOP 3505
"A" is conrect. Step 1 RNO will direct a transition to EOP 3501 if neither Bus 34C or 34D is energize "B" is incorrect because the SBO diesel should only be started using the guidance of EOP 3501. EOP 3501 directs the operators to
, attempt manual loading of the operating EDO before directing action to start / load the SBO diese . .-
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"C" and "D" are inconcet because they are strategies used in EOP 3501 maer all attempts at energizing operable arxt degraded busses
- have been unsuccessfu OBJECTIVES: E05805C(3)
K/A: 2.4.9, Low power operations in EOP Exam Bank item 2935
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SRO 20 1he plant is operating normally at 100% powe Which of the following conditions require a reactor trip followed by a trip of ALL RCPs? VCT temperature increases to 140' . Isolation of a single train of RPCCW to containmen Seal injection flow to each pump decreases to 5.5 OP The "A" RCP bearing oil temperature increases to 200* ANSWER: VCT temperature increases to 140* REFERENCE: AOP 3561 Foldout Page JUSTIFICATION: (A correct) If VCT temperature is greater than 135'F AND RCS temperature is above 400*F, then all RCPs must be stopped (RCPs are also stopped at any time if VCT is above 150*F)
(B incorrect)Both trains of RPCCW to containment must be lost to trip RCPs (C incorrect) A reactor and pamp trip is required if seal injection flow is less than 6 GPM AND thermal barrier cooling is also lost, (D incorrect )lligh oil temperature only requires a trip of the all'ected pump OlljEC11VE: A61761D (1)
K/A: 2.4.11 Knowledge of abnormal operating procedures Question # 275
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SRO21 The plant tripped from 100% power A large break LOCA has ocentred. The control room has carried out E 0 and is now entered E-1. A computer failure has occurred and the US is performing a manual status tac chec The following conditions exist:
SR energized with a negative SUR Core Exit TC's 640 degrees Subcooling 28 degrees RVLMS Plenum 100%
S/O NR's 4%
Total AFW flow 500 GPM Cold leg temp decrease in last hour 20 degrees RCS temperature 480 degrees Containment Pressure 60 psia Pzr levelis at 0%
RVLMS upper head is at 100%
What answer best describes the sequence for dealing with the Critical Safety Functions Ileat Sink Containment IL Containment llent Sink Ileat Sink Core Cooling Containment Containment lleat Sink Core Cooling ANSWER: Ileat Sink Containment REFERENCES: CSF status Trees. OP3272 EOP Users Guide Section 1.6
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JUSTIFICATION:
SR energized with a negative SUR Green Path Subtriticality Core Exit TC's 640 degrees Subcooling 28 degrees RVLMS Plenum 100%- Yellow Path Core Cooling S/O NR's 4%
Total AFW flow 500 GPM Hed Path liest Sink Cold leg temp decrease last hour 20 degrees RCS temperatuie 480 degreesGreen Path Integrity Contailunent Pressure 60 pslaRed Path Containment Pzr levelis at 0%
RVLMS upper head is at 100% Green Path inventory ( A correct) Red paths are first. lical sink is before Containment (11 incorrect) Containment is aller lleat sink (C is incorrect) Core cooling is a yellow path and is before Containment which is a Red pat ( D incorrect) lleat Sink wn! Containment are out oforder of priority OllJECTIVES: E0004C K/A: 2.4.21, Knowledge of status trecs/ logic's for CSF New question e
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SHO22 PLANT CONDITIONS:
Plant is in MODE 3 Tave is $57 F Main steam pressure is 1092 psig, controlled via turbine bypass valves Atmospheric steam dump valves (ASDVs),3 MSS *PV20A, B, C and D are in Atno, set at 1150 psig What is the proper method for changing ASDV set points? Place turbine bypass valve controller (MSS PK507)in M ANUAL, Place all ASDV controllers in MANUAL. Adjust each ASDV set point to the desired value, then place the ASDV in Atno. Return MSS PK507 to AUT Place one ASDV controller in MANUAL. Adjust the set point to the desired value, then place the controller in AUTO. Repeat this process for each remaining ASD C, Slowly dird the ASDV thumhwhe:Is to the desired value, ensuring the set point tracks properly and the valves do not ope Place MSS PK507 in MANUAL. Ensure all ASDVs are closed, then place cach ASDV in MANUAL, one at a time, and adjust set points to the desired valu ANSWER: Place one ASDV controller in MANUAL. Adjust the set point to the desired value, ..
then place the controller in Aino. Repeat this process for each remaining ASD REFERENCES: OP 3203 JUSTIFICATION: 4.26 Place the Atmospheric Steam Dump controllers in ,
MANUAL one valve at a time prior to making any setpoint changes, then, Return the controller.to AUTO. This prevents the Atmospheric Steam Dumps from opening rapidly causing steam pressure transients OBJECTIVES: NAD106; NAD108 K/A: 2.2.2, Local / manual operations of controllers Question: 2131
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l SHO 23 While perfonning a new Surveillance on the Safety injection pumps, the CO perfonns a step as written and notices that the Safety injection pump does not have adequate recirculation flo What is the fut action (s) the CO should take? Detennine the cause of the proble . Continue with the surveillance, but consult with the Unit Superviso Stop the surveillance and place the Safety injection pump in a stable or safe conditio Initiate a procedure modification in accordance with DC 1, Administration of hiillstone Procedures and Fomi ANSWER Stop the surveillance and place the Safety injection pump in a stable or safe conditio REFERENCE: DC-4, Procedural Compliance, JUSTIFICATION: Inadequate or Unexpected Results 1.9.1 IF procedure appears to be inadequate, OR yields unexpected results while executing work activity, PERFORhi the following: STOP work activit IP applicable, PLACE equipment or system in stable or safe condition. (C correct) CONSULT First Line Supervisor for direction. (D incorrect) DETERhilNE cause of problem. (A !ncorrect) IF necessary, kefer To DC 1, " Administration of hiillstone Procedures and Fonus" and INITIATE modification to procedure to rectify problem (Stephen E. Scace memo " Procedure Compliance - hianagement Expectations" to hiillstone Station Personnel, number h1P 91 801, dated October 10,1991, states: "If you think you can't follow the written procedure, consult First Line Supervision to determine what actions ( procedure changes) are necessary before proceeding. If a procedure cannot be followed as written: a. Stop the task and place the equipment or system in a safe condition, b. Change the procedure using the -
procedure change process. c. Proceed with the task.")
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OllJECTIVES: 11AD544 K/A: 2.2.12, Knowledge of Surveillance Procedures Question : 3260 modified
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o SHO24 When preparing a clearance, which of the following system / equipment conditions should be isolated from the work area by two closed valves in series?
- A fluid system which operates at 170* A gas system which operates at 50 psi Caustic or acid systems at any temperature or pressure.
<. Syrtems from confined work space ANSWER: Systems from confined work space REFERENCE: WC2
.iUSTIFICATION: If practical, isolate fluid or gas systems that operate at greater than 200'F or 100 psig from the work area with two closed valves in series. Isolate systems from confined work space with two closed valves in serie OBJECTIVES: NAD318 K/A: - Generic 2.2.13 Tagging / Clearances 97 LOIT Remediation Exam i
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SRO 25 Giwn the following conditions:
- MP3 was operating at 100% power when a spurious Si occur * All systems respond as designed with the exception of the "B" reactor trip breaker, which did not open and remains close * The crew responds I AW E0Ps, eventually transitioning to ES 1.1, "Si TERMINATION".
- The crew takes both Si reset switches to RESE WillCil ONE of the following describes the status of Sl? Both trains of Si are reset and automatic initiation blocke Neither train of Si is re. set, nor is automatic initiation blocke Both trains of Si are reset, only the "A" train automatic initiation is blocke Both trains of Si are reset, only the "B" train automatic initiation is blocke ANSWER: Both trains of Si are reset, only the "A" train automatic initiation is blocke REFERENCES:
JUSTIFICATION: Both trains of Si can be reset but only Train "A" is blocked because P-4 is not enabled because B reactor trip breaker is close ("C" is conect,)
Si cannot be blocked in Train "B" because P-4 is not present. ("A" and "D" are incorrect.)
Si can be reset in both trains. ("B" is incorrect.)
OBJECTIVES: RPS012C R/A: 013 K4.01, SIS Reset New Question
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l SRO 26 Following a Loss of Coolant Accident, adverse containment conditions exist. The following values have been recorded by the ST ISE CONTAINMENT TEMP CONTAINMENT RADIAYION LEVELS 0800 185"F 5 x 10' R/lIR
- 5 0815 190 F 2 x 101UllR
0830 180"F 2 x 10 IUllR
0845 175 F 1 x 101UllR d
0900 170"F 9x 10 lUllR When, if ever, may the crew suspend the use of adverse containment values? At 0845 because containment temperature has decreased below it's adverse setpoin . At 0900 because both containment temperature and radiation levels are below their adverse value Adverse values can never be relaxed once they are entered if containment temperature limits were exceede Adverse values can never be relaxed once they are entered if containment radiation limits were exceede ANSWER: Adverse values can never be relaxed once they are entered if containment radiation limits were exceede IEFERENCE: OP 3272 JUSTIFICATION: "A" is incorrect because containment radiation levels are still adverse and adverse values will always appl "11" is inconect because 10 1UllR limits are/were exceede . .
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"C" is incorrect because of containment radiation levels haven't exceeded 10' R/IIR containment temperature regarding adverse values can be relaxed when temperature drops < 180* Ol3JECTIVES: ' E0003C R/A: W 16 EK 1.3,111 Rad Alanus and Actions Modified Exam item 3213
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SRO 27 FR C 1 is entered when core exit therinocouples are greater than 1200 F. Implementation of FR C.1 is safety significant at this time because additional operator action is required to: Prevent core uncove . Provide core cooling to stop the hydrogen generation due to zircaloy water reaction.
. Limit containment pressure to less than design pressur Provide wre cooling to prevent exceeding peak clad temperature limit ANSWER: Provide core cooling to prevent exceeding peak clad temperature limit REFERENCES:
JUSTIFICATION: When FR-C.1 is entered the core is already uncovered. ("A" is incorrect.)
113 generation from zircaloy water reaction starts ~ 1800 - 2200F, ("11" is incorrect)
FR-C.1 is written for a small break LOCA with no high head injection. Car fans and spray will limit pressure. FR established injection flow to cool core. ("C" is incorrect.)
OlljECTIVES: MC1004 K/A: 017 A2.02, Mitigating Core Damage New Question
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SR028 Givra the following conditiorts:
e The unit is at 4$% power with all loops in operatio . Control Rods M12 and D4 in Bank "D" , group 2, are stuck and misaligned and wG rot mw . Tb '4inted yMa ait trippabl * The Bank D sitch rods indiente 156 and 162 Steps respeev 'ely on DRPI while the other Bank D rods and step cc.unters indicate 180 step Assmning the rods cannot be repaired within the next week, which one of the following conectly describes the actions required by Technical Specifications for the misaligned rods? Verify shutdown trargin requirements are met within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
- Verify QPTR within I hour and apply LCO 3. Align the remaining Bank "D" rods to i 12 steps of the inoperable rod Be in llOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ANSWER: Be in llOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> REFERENCE: Technical Specification 3.1.3.1, AOP 3552 " Malfunction of the Rod Drive System" JUSTIFICATION: Technical specification 3.1.3.1 outlines the actions to be taken for a
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single stuck but trippable control ro A is incorrect because shutdown margin must be satisfied in one hour not B is incorrect because QPTR requirements are not applicable when less than 50% powe C is incorrect because multiple rods in the same group are misaligned by greater than 12 steps. Alignment is not allowed and the unit must shutdown and be in hot standby because of the multiple misaligned rods in the same group (D is correct).
K/A: 005 K3.06, Actions in EOp
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OllJi?CTIVil: ROD 08C Modified NRC IIxam item 95 LOIT INam ,
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SRO 29 INITIAL CONDITIONS:
- MODE 5 with the RCS solid, e Temperature being maintained 140*F to 150'F by the "A" RiiR train.
- The plant is currently 4 days into a scheduled 8 day "B" train electrical outage, with 34B and 34D deen:rgiwd. "B" train load centers are NOI cross tied to the
"A" train.
Assuming NO operator actian, which of the below statements describes the plant response to a loss of the 'A' instrument Air Compressor? RCP Thermal Barrier cooling flow from RPCCP will decrease. Letdown flow will increase resulting in the RCS depressurizin RCS temperature will increase due to increased RPCCP flow through the RllR 11 eat Exchange RCS temperature will decrease due to increased RilR flow through the RilR lleat Exchange ANSWER: RCS tempersture will decrease due to increased R11R flow through the R11R lleat Exchange REFERENCES: P&lD 104A,121 A,112A,121B JUSTIFICATION: 'B' is incorrect because a loss ofI AS will cause 11CV128 to fail closed which will result in a loss ofletdown and a resultant increase in RCS pressur 'C' is incorrect because a loss ofIAS will cause FV66A to fail AS IS resulting in NO change in RCS temperature from CCP flow. 'D'
is correct because FCV 618 fail closed and llCV 606 fails open on a loss of1AS which will result in maximum flow through the RllR ll 'A' is incorrect because the CCP return valves from the thermal barriers have a LOCK UP feature to prevent them from being I
affected by a loss ifl AS. The CCP CTMT isolation valves are MOVs and therefore are not affected by a loss ofI A OBJECTIVES: PAS 07C; RilR07C K/A: 078 K3.02, Pneumauc control valves 96 AOP Exam
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SRO 30 Given the following:
. A loss of offsite power has occurre * Tave is S$2'F e " Turbine Bypass Tav Interlock Bypassed" is illuminate * Steam dumps are in steam pressure mode of contro * Steam dumps demand is manually INCREASED to begin a cooldow * The steam dumps failed to ope Which ONE (1) of the following explains why the steam dumps will NOT open? P-12, LO LO Tave, has disarmed the steam dump P-4, Reactor trip, has locked out the steam header pressure signa C-9, Condenser available, interlock is not me The plant trip controller has not rese ANSWER: C-9, Condenser available, interlock is not me REFERENCES:
JUSTIFICATION: "C" is correct because power is not available to cire pumps on loss of offsite power. Blocking signals override arming signal "A" is incorrect because P 12 has been bypasse "B" is incorrect because P-4 locks out the load rejection controller in the Tave mode of contro "D" is incorrect because plant trip controller was reset when you shifted to pressure mod I
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Olul!CTIVES: SDS06C K/A: 051 K3.01, Steam Dump Operation Loss of Vacuum Exam Item: 2422
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SRO 31 Which of the following conditions occurring concurrently with a large 1.OCA will required entry into ECA-l.1 - Loss of Emergency Coolant Recirculation?
- Off site power is lost and the "B" EDO did not stan. The power lockout relays fall to operate and all the white power lockout indicator lights are dim. The A charging pump tripped on overcurrent and the B SI pump is tagged out for maintenance. The "B" & "D" Recirculation spray pumps are damaged and cannot be starte ANSWER: The power lockout relays fail to operate and all the white power lod ov' indicator lights are di REFERENCE: ES 1.3 step 2, notes prior to step 2 & JUSTIFICATION: A is incorrect. The A EDO is still available to supply A train components for cold leg recir C is incorrect - One charging pump and one Si pump are still available for cold leg recirculatio D is incorrect because the A train of RSS is still operabl B is correct because without the power lockout operating power not available to operate some of the recirculation valves, ES- directs the operator to ECA- OBJECTIVES: A1101C K/A: W/E 11 K2.1, Control, function, system Modified 2853
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SRO32
- Which of the following plant conditions will cause the TD AFW pump to auto start? /4 SO level detectors at low - low level in two so Safety injection Signa Main Feedwater Isolation Signa Loss of Battery Bus 5 ANSWER: /4 SO level detectors at low low level in two so REFERENCES:
JUSTIFICATION: 2/4 low low signals in at least 2 SO will start TD AFW pum Only the motor driven AFW pumps start on SI (D is incorrect).
Main feedwater isolation will only isolate MFW but does not start any AFW pump TD AFW pump will start on loss of batt Bus 1/2 not Bus 5 (D is inconrect).
- OBJECTIVES: FWA04C -
K/A: 059 A4.ll, Auto start AFW 4.2 -
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SRO 33 Assume that prior to a stanup, work on the Intermediate Range nuclear instrumentation resulted in HOTil channels being OVER COMPENSATE Which of the following describes the expected system response to this condition? During stanup the Intermediate Range indication will be less than actual, and during shutdown the Source Range may not be automatically reinstate . During startup the Intermediate Range indication will be greater than actual, and .
during shutdown the Source Range tr.ay be reinstated prematurely causing an unwanted reactor tri During stanup the Intermediate Range indication will be less than actual, and dming shutdown the Source Range rnay be reinstated prematurely causing an unwanted reactor tri During stanup the intermediate range indication will be greater than actual, and during shutdown the source range may not be automatically reinstated due to the P 10 pennissive being activ ANSWER: During startup the Intennediate Range indication will be less than actual, and during shutdown the Source Range may be reinstated prematurely causing an unwanted reactor tri REFERENCE: Funct. Diag. Sht. 3 & 4 JUSTIFICATION: A is incorrect because on shutdown the Intermediate range detectors will read lower than actual and be automatically reinstated prematurel B & D are incorrect because readings on stanup will be lower than actual not higher,
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OBJECTIVES: NIS06C(a); NIS05C l
K/A: 032 A2.04 - SIUIR overlap Dank item 2217
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SRO 34 -
PLANT CONDITIONS:
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i * 100^4 power
- All systems in AUTOMATIC
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LT-459 se!ected for control of Pressurizer level control selected to LT-459 a
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PT-456 selected for control of Pressurizer Pressure
- Instrument for "A" and "C" steam generators selected to Channel 1
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i * Instruments fbr "B" and "D" steam generators selected to Channel 11
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E Aloss of VIAC-2 occur Which of the following lists controllers which should be taken to MANUAL as a result of
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the VIAC-2 failure? Rod control Pressurizer Pressure Pressurizer Level E
E - Rod Control 7- Pressurizer Pressure s Master main feed pump controller
! Pressurizer Pressure
_ Pressurizer Level *
i . Feed Regulating Valves fo "A" & "C" SGs E
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Pressurizer Pressure Master main feed pump controller Feed Regulating Valves for "B" & "D" SGs
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ANSWER:
h Pressurizer Pressure Master main feed pump controller Feed Regulating Valves for "B" & "D" SGs
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REFERENCE: AOP 3564, Process sheets 10 and 11,25 JUSTIFICATION:
Pressurizer level controller will be affected because the bachup
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, 'y Pressurizer pressure is affected because its controlling channel is channel 11 Main feed pump speed control is affected'due to loss of two steam-flow channels, lo ' Rod Controlis not afTected The Feed Regulating Valves on only the "B" and "D" SGs will be'
affecte Only D correc .
OBJECTIVES: 12005C
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K/A: ape 057 Al'.06, Manual control of components New question i
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i SRO 35 -
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. The plant is operating at 100% power. The steam dumps are in the Tave mode of control
- and rods are in AUTO.-
[ 3 MSS *PT507, Main Steam Header Fressure, fails high. This will:-
?A.- .Open the steam dump cooldown valves, and the TDFWPs speed will decrease,
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- B.- Block the load rejection controller from arming the steam dumps, and the
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TDFWPs speed willincreas C.- Arm the steam dumps but they won't open, and the TDFWPs speed will decreas '
- . Have no effect on steam dumps, and the TDFWPs speed willincreas ANSWER: ., Have no effect on steam dumps, and the TDFWPs speed will increas REFERENCES:
- JUSTIFICATION: PT 506, not PT 507, feeds the load rejection controller and arms )
the steam dumps (B & C are incorrect).
The PT 507 effect on the cooldown valves function is only available in the pressure mode of control (A is incorrect).
. PT_507 is only in effect when in the pressure mode of control (D is correct)
PT 507 feeds the TDFWP speed control circuitry. An increase in -
- pressure will cause the pumps speed to increas OBJECTIVES: SDS07C K/A: 039 A2.04 Modified 2426
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- SRO 36 INITIAL CONDITIONS:
.- The Unit is operating at 48% power with the "NIS POWER RANGE P-9 '
PERMISSIVE" Blue Light NOT Lit.-
- - _ Due to an instrument' failure, actual level in the "B" S/O level has increased to
- 80% resulting in an automatic FW1 actuatio .- _ All equipment operated as designe . Assuming no actions are taken in the instrument rack room, which of the following must -
occur to allow resetting the FW1 signal from the main boards?
A. - Clear P 14 and Reset P-4 B.' Clear P-14 and P-9
' Clear P-14 only
' I). Reset P-4 only ANSWER: Clear P-14 and Reset P-4
- REFERENCE: Functional Sheet 13 '
- JUSTIFICATION: - Permissive is lit below P-9, iflevel reaches the turbine' trip setpoint, the reactor will trip, therefore to reset the FWI, both P-4 and P-14 will have to clea ;
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. OBJECTIVES: .NISO4C (b.4) -
- K/A: 059 A 4.11 Permissives; Question 152 ,
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SRO 37-Initial Plant Conditions
- Plant is at 60% power
- - Rod controlis in manual
- Steam Dumps are in Tave mode of control
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' A turbine trip occurs. The turbine trip fails to cause a reactor trip and actuate the steam dumps on the turbine trip controlle .The next autoniatic reactor trip signal to be generated for this transient would be:
A. - High Pressurizer Level Trip - _ OTAT-
, High Pressurizer Pressure Trip
- OPAT ANSWER:-
- High Pressurizer Pressure Trip REFERENCES:
JUSTIFICATION: The power mismatch will cause a rapid increase in pressurizer pressure causing the reactor to trip (C is correct).
Pressurizer level will backup the high pressure trip (A is incorrect)
B is' incorrect. The temperature increase will drive power'down
' and pressure up. Both of these factors are benefits with regard to OTAT.'
e D is incorrect. Power will decrease during the trans!ent. The -
margin to the OPAT trip will be increasin OBJECTIVES: A5002C; MC0302 K/A: 045 A1.05 RCS following turbine trip
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New Question
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SRO 38 Plant History:-
- - A loss of off-site power occurred 10 minutes ag . The crew has stabilized the plant and just completed ES- * RCS temperature has stabilizxt at 557'F .
e Steam Generator levels are between 25 - 30% in the narrow rang . Total AF_W flow to the steam generators is 535 GP The Operations Manager has directed the plant be maintained at the current plant-conditions for RCS temperature and steam generator levels.
O . One day from now the total AFW flow should be approximately:7 sg
' The sam ' B. __ 110 - 140 GP ,
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. ANSWER: GP . REFERENCE:
'
JUSTIFICATION: One minute after a reactor trip the decay heat level is -
approximately 3-4%. After one hour it is approximately 1.5-2% .
.
- and after one day.it is 0.7-l%. Consequently after one day the existing decay heat is approximately'one quarter ofits value -
following the trip. Therefore since the AFW system is maintaining-SG level, the flow will reduce by one quarter to approximately 134
- GPM. .
.' 4
- OBJECTIVES: S0103C K/A: W/E 9 EK Relationship betWeen emergency feedwater flow to S/G and decay heat removal for facility heat removal following a tri New Question
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SRO 39
- PLANT CONDITIONS:
=- - Reactor is operating at 100% rated thermal power c
- ' Annunciator 5-3 on MB4C "PR UP DET lil FLUX DEV/ AUTO DEFEAT" has alarmed
.- All control rods are positioned within 12 steps of their group demand counters
. - Maximum QPTR based on plant computer program 3R5 is 1.04 .
Assuming QPTR is not reduced, within two hours reactor power must be reduced to
, and NIS overpower trips reduced to within the next four hour %,59% - . 50*$,55% %,97 % %,103 %
ANSWER:
C, 88 %,97 %
REFERENCE: Tech. Spec. 3.2.4 Action Statement c.2; OP 3273
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JUSTIFICATION: If QPTR is greater than 1.02 but less than 1.09 then within 2 hrs reduce thermal power 3% of rated power for every 1% greater than l.0 and similarly reduce the overpower trip setpoints within the 74?%{
'"';!At;
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next four hour . OBJECTIVES:
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NIS08C (b)
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K/A: 015 Al.04,NIS/QPTR Modified 1058 l
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- SRO 40)
^ Given the following conditions:
e' The unitis critical
'e
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- The crew is holding power at 1.0 x 10-8 amps.-
'* : SO levels are being controlled on the bypasses, in automati N.36 control power fuse blow WHICH ONE of the following describes the plant's response?. An IR high flux rod stop will be received and the reactor will remain critica L B, The reactor will remain critical with no rod stop J' U The reactor will trip on SR high flux when they automatically energiz ~ D.-- 1The reactor will trip on IR high flu ANSWER:
D.' The reactor will trip on IR high flu : REFERENCES:
iJUSTlFICATION: Loss of control power fuses causes a trip signal to be sent to RPS through the Reactor Protection System and will also generate a rod stop and reactor trip on 1/2 coincidence (d is correct)
' A is incorrect because the trip will cause the reactor to go -
suberitical. (This also makes b incorrect).
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C is incorrect as source ranges will not automatically energiz nOBJECTIVES: NIS07C K/A: 015 K2.01, NIS channels, power supplies Modified item 2.269
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_ SRO 41_
The following conditions exist:
- Crew is in EOP 3503, Shutdown Outside The Control Room
- Control room is filling with dense smoke e Control room is ordered evacuated
. - The reactor is tripped from 100% power-
- The turbine is tripped Si occurs after trip due to the steam dumps malfunctioning Which of the following describes the procedural flow path under these conditions: Com}Jete EOP 3503 and then enter E- B. - Exit EOP 3503 and enter E- Perform E-0 in parallel with EOP 350 D.- Complete EOP 3503, then perfonn cooldown in accordance with EOP 350 ANSWER:
D.- Complete EOP 3503, then perform cooldown in accordance with EOP 350 REFERENCES:-
JUSTIFICATION: EOP rules of usage - if Si or Rx Trip occurs in EOP 3503, you should remain in EOP 350 OBJECTIVES: EOU (1733)
"T a K/A: 067 K3.04 Actions in EOPs "
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SRO 42 Given the following conditions:
. Unit is at 100% power
. Pressurizer level control is selected to 459/461
. VCT makeup controlis in Automatic A reference leg leak occurs in pressurizer level transmitter LT-45 Assume no other operator action, which of the following occurs: Letdown isolation occurs 3Cl-I FCV-121 ramps open Auto makeup occurs Unit will eventually trip on high pressurizer leve Cil-FCV-121 ramps closed VCT diverts Letdown isolation occurs Pressurizer level will begin to increase Unit will eventually trip on high pressurizer leve , Letdown isolation occurs
>
3CII-FCV-121 ramp open Auto make-up occurs VCT swaps over to RWST Plant cools down Unit trips on low pressurizer pressure because heaters are de-energize Cil FCV-121 ramps closed VCT diverts Letdown isolation occurs Pressurizer level will continue to decrease due to seal leakof Unit trips on low pressurizer pressure ANSWER: CH-FCV-121 ramps closed VCT diverts Letdown isolation occurs Pressurizer level will begin to increase Unit will eventually trip on high pressurizer leve ,
u
REFERENCES:
JUSTlFICATION: Reference leg failure will cause indicated level to fall high this
- causes FCV-121 to ramp close. Thus A & C are incorrect).
Act pressurizer level will decrease and the renaming channels will cause letdown isolation. Seal injection will fill pressurizer and cause eventually a high level trip (B is correct).
D is incorrect because seal injection still occurs even if the FCV-121 is closed and level will begin to increas .
OBJECTIVES: A5503C; PPLO6C; PPLO7C K/A:~ 011 K3.01, Loss of PZR Level effect on CVCS,
- Modified 375 i
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SRO 43 Given the following conditions:
- The unit is at t power.
- Plant startup is in progress
Pn level instrument LT-459 has failed LOW.
- All actions of AOP 3571 "Instnunent Failure" Attachment C are complete.
Which of the followmg describes the course of action the crew should take if a subsequent failure of Pzr level instnament LT-460111011? Verify reactor trip. Stop the startup, and restore one of the failed channels of pressurizer level to OPERABLE status piior to increasing power above 10%. Stop the startup, and restore both of the failed channels of pressurizer level to OPERABLE status prior to increasing power above 10%. Within one hour initiate ACTION ta be in at least ilOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ANSWER: Stop the startup, and restore ONE of the failed channels of pressurizer level to OPERABLE status prior to increasing power above 10%.
REFERENCE: AOP 3571 " Instrument Failure" Attachment C, Pzr Level and Pressure Control Lesson Plan, Technical specification 3.3.1 and ftmetional sheet 11
.IUSTIFICATION: With all actions of the AOP complete, the bistable associated with the high Pzr level Rx. trip has been placed in a tripped condition When the second channel fails high, the coincidence for a high pressurizer level reactor trip is met, however, the trip is blocked less than 10%. (A incorrect)
.
Technical specifications require 2 channels to be OPERABLE, however, this is required below P-7 (10%), and to increase above 10%, the bistable, must be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, B correct, D incorrec _ _ _ _ _ _ _ _
._ _ _ _ _ _ _ _ - - _ _ _ - _ - _ _ - _ _ - _ _ _ _ - _ _ _ - _ - - - _ - _ - _ _ _ - - - . - - _ _ - _ - _ . - - - . - . - - - - _ - - - - _ _ - _
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it is not required to have both channels OPERABLE to increase - '-
l above 10%,(C incorrect)--
OBJECTIVES: -
'PPLO7C- -+
!. - K/A: 028 M.01, Pressurizer level histables -
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. modified from 1995 MP3 NRC exam -
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SRO 44 Given the following conditions:
- The reactor is trippe * A Loss of Offsite Power has occurred
- Safety injection is actuated from a small LOCA.
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= All ECCS equipment is operating as expecte * Pressurizer level is 4E% and increasing on all channel * RCS pressure is 1700 psia and decreasing slowly on all channel Which one cf the following describes a leak location that is consistent with the indications given?
.A . A leaking pressurizer safety valv The letdown line relief valve lifting A reference leg break on pressurizer level instrumentatio * A failed open spray valv ANSWER: A leaking pressurizer safety valv REFERENCES:
JUSTIFICATION: A PORV or safety valve failing open will cause pressurizer level to increase and pressure to decrease on all channels. ("A" is correct.)
"B" is incorrect because this break location will be isolated by the SI/ CIA signa "C" is incorrect. On the effected reference leg the indicated level channel would increase and pressure would decrease. However, on the non-affected channel level will decrease as well as pressurizer pressur . _ _ _ _ _ _ _ _ - _
(.,,,,,,., , , , , , , , . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
"D" is incorrect a failed open spray valve will only cause
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- pressurizer pressure to decreas OBJECTIVES: -
A5503C
- K/A: 008 A1.01, Operation Monitoring Instrumentation from for PORV, -
sprays
' New question
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SRO 45 During performance of the shift control room rounds, the Control Operator discovers that 311VQ-RE49, ESF Building Normal Ventilation Monito indicates OFF-LINE for both Data-A and Data B at the RMS Consol Which of the below statements describes the operating status of the radiation monitor? The radiation monitor may be considered operational once its data and operation is verified at the Local Indicating Control panel (LIC). The radiation monitor must be considered inoperabl The radiation monitor continues to indicate properly at the RMS Console but all radiation monitor control functions must be performed manually at the RMS Consol The radiation monitor may still be considered operational because it will still 4 notify control room stafTof high radiation conditions by actuating the
"RADI ATION ALERT" and " RAD 111" annunciators on MB ANSWER: The radiation monitor may be considered operational once its data and operation is verified at the Local Indicating Control panel (LIC).
REFERENCES: P&lD 152A, RMS073T and RMS073C Handouts, PIR 391-043, MP3 Memo MP-3-0-385 dated 3/25/91, Kaman Instrumentation Operation - Maintenance Manual volumes I thru 3 and RMS Console iIelp displa JUSTIFICATION: 'B' is incorrect because each local unit is completely self contained, requiring the computer room computer ONLY for transmitting data to the control roo 'C' is incorrect because if off-line from both data-A and data-B, all communications between control and the RMU is terminate 'D' is incorrect because the alarms at MB2 are a function of the computer. If the RMU computer is not communicating with the computer room computer, it can not cause the MB2 alarms to actuat 'A' is correct because although not communicating with the control room, each RMU is completely a stand alone unit and is designed to function without the control room computer. Once the data and operation has been verified correct for the RMU at its LIC, the unit may be considered operational per the SS (Memo MP-3-0-385 and PIR 391-043 and RMS073 handouts)
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011JECTIVES: RMS08C K/A: 073 A4.02, RMS Control Panels / indications Exam Item 2407
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SRO 46
- Oiven the following:;
- e: - A twenty five (25) year old Maintenance Contractor with complete exposure
- records has the followilig exposure record for the current calendar year:
. Shallow Dose Equivalent -. 2.55 REM e Committed Dose Equivalent - 0.75 REM e Deep Dose Equivalent - 2.13 REM -
- ' Lens Dose Equivalent - 3.08 REM a Committed Effective Dose Equivalent _- - 1.95 REM WHICH ONE (1) of the following is this individuals Total Effective Dose Equivalent (TEDE) for the current calendar year?
' .88 REM .08 REM .21 REM - .43 REM ANSWER:- < 4.08 REM REFERENCE: IRPM I. Get RAD Worker Training--
- JUSTIFICATION:
-
TEDE = CEDE + DDE = 1.95 REM + 2.13 REM = 4.08'
OBJECTIVES: GET Radworker training
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K/A: 2.3.1,.10CFR20 Radiation Limits e 1 New Question
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SRO 47-l The rad waste PEG is dispatched to change LWS-FLT3. This PEO has not performed
= this task before. The lip technician informs the PEO that the dose rate on the outside of the filter housing is 1 R/h Which one of the following is nel an example of ALARA techniques for reducing exposure for filter replacemen 'A Long handled tools to remove the old filter -
B Place the filter in a shielded drum upon removal to reduce exposure llave the rad waste PEC oe assisted bv the Turbine Building PEO who has done the task several times befor Di llave the new PEO perform the filter replacement on a mockup first, i l
ANSWER: Have the rad waste PEO be assisted by the Turbine Building PEO who has done the task several times befor REFERENCE: RPM 5.2.4 Section 1.2,1.3, JUSTIFICATION: RPM 5.4.2 list 3 main areas to reduce Radiation exposure. Time Distance and Shielding A is Distance B is Shielding
-
D is Reduced Time by practice on a mockup prior to the jo RPM 5.2.3 Section 1.1 states that individual exposures within a work group are balanced consistent with experienc C will not balance the exposure if the experienced person shvays does thejo ' OBJECTIVES: NAD721; NAD722; NAD723 K/A: 2.3.10, ALARA - procedures to reduce radiation exposure New Question
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SRO 48 --
L Given the following conditions: MP3 is at 35% powe * "RCP B STANDPIPE HI LEVEL" has li .- : "B'.' seal injection flow is 8.2 GPM.-
t* "B" seal leak-off flow is 0.2 GP * Seal retum temperature is 150*F and rising steadil *: Pump radial bearing - rising slowly @ 145'F
- Based on the above indications, the operating crew should:
=1 Trip the unit, recure "B RCP and close its No. I seal leekoff valve within 2
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minute _ B. _ Trip the "B" RCP and close its No I seal leakoff valve after the pump has been stopped for five minute Close the "B" RCP's No. I seal leakofTvalve within 5 minutes and shutdown the
, unit within the next 30 minutes then secure the "B" RC Trip the "B" RCP and close its No 1 seal leakoff aner the pump has been tripped
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for two minute .
-ANSWER:-
- Trip the "B" RCP and close its No 1 seal leakoff aAer the pump has been tripped - j for two minute REFERENCE: OP 3554 -
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' JUSTIFICATION: - Since power less than P-8, the RCP can be stopped without tripping the unit (a is incorrect).
OP 3554 requires tripping RCP and closing the seal leakoff valve within two minutes (D is correct) The RCP must be removed from service within 5 minutes of failure (not within 5 minutes of closing .
_
seal leakoff valve);(B and C are incorrect)
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' OBJECTIVES: 'AS403C" K/A:- 015 A2.01, Cause of RCP failure
. Modified question 1104 -
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SRO 49 '
WHICH ONE of the following interlocks must be satisfied to start an RCP7__ RCP _#1 seal AP must be greater than 200 psid - The overcurrent trip selector switches must be in the cold ' positio Cold leg and hot leg isolation valves must be ope D.- Cold leg isolation valve must be open and the loop bypass and hot leg isolation
- valve must be closed.
~ ANS4TR:
C. - Cold leg and hot leg isolation valves must be ope REFERENCES: RCP text; OP 3301B JUSTIFICATION: ' A and B are incorrect because _they are procedure administrative requirements but are not part of the interlock circuitry'
C is correc D is incorrect. The cold leg stop valve must be closed with the bypass fully open to satisfy the RCP interloc ~ OBJECTIVES: RCS04C K/A:
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003 K6.14, RCP starting requirements Modified question 2159
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SRO 50 :
c A small break Loss of Coolant Accident has occurre The current plant conditions exist at the completion of E-0 step 14:
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- l - S1 has occurre *: . All SI equipment starte *; Containment Temperature is 185' *~ RCS pressure is 1800 psia and stabl * -- CET's are 520' . _
-* _
Pressurizer level is 50% and slowly increasin Assuming conditions do not significantly change, you would expect to stop one charging -l
- pump in:
-A.: E-0 Reactor Trip on Safety Injection
~ B.- ES 1.1 - Si termination.
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' ES-1.2 - Post LOCA Cooldown and Depressurization
' DJ ES - 1.3 - Transfer to Cold Leg Recirculation
' ANSWER:
!
i C. - ES-1.2 - Post LOCA Cooldown and Depressurization-
~ REFERENCES:
- JUSTIFICATION: LA is incorrect because adequate subcooling doesn't exist and RCS pressure is less than 1950 psia (adverse containment) to stop a'
- charging pump in E- B is incorrect because adequate subcooling .doesn't exist to make the transition to ES- ' C is correct because a 100"F/hr cooldown will be started which -
. will increase subcooling major to allow stopping a charging pump y in ES- D is incorrect because charging pump will be stopped la ES-1.2, ,
and cooldown will place plant on RHR, ES-1.3 will not be entered
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for a small break LOC __- -__
OBJECTIVES: S1203C K/A: E02 EK2.1, SI Termination Modified Exam item 1533 l
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SRO 51?
- Given the following conditions:
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E * -- The Unit was operating at 75% powe * : A small break LOCA occurred in coincidence with a loss of off-site powe .
RVLIS indicates that a void exists in the reactor vessel hea *
The cooldown was stopped and RCS pressure raised to regain subcooling margin, increasing RCS pressure will the size of the void and the leakage from the RC Increase; increase
-: Decrease; increase Increase; decrease . Decrease; decrease ANSWER: " Decrease; increase REFERENCE:
- JUSTIFICATION: Raising the pressure will decrease the size of the void but increase the leakage for the RCS. ("B" is correct)
OBJECTIVES: MC0703 (c)
K/A: 009 K3.06, Inventory Balance During Small Break Los New Question -
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SRO52 PLANT CONDITIONS:
.- Plant is in Mode 6
- No fuel movements are in progress
= "A" Train Electrical outage is in progress
- Computer is available
. "B" Spent Fuel Pool cooling pump caught fire and tripped l_ hour ago e Spent Fuel temperature - 115'F and slowly increasing
. Spent Fuel Pool level - 37% and decreasing slowly ,
e Reactor Cavity seal-intact
.- Fuel Building Monitoring Group 11istogram - NORMAL e- RWST level - 1,000,000 gallons ;
For the existing plant conditions, which one of the following corrective actions should be -
taken? Align RWST to gravity feed the spent fuel poo Establish emergency makeup using the fire water syste Supply makeup to the Spent Fuel Pool from the Primary Grade Water System Establish emergency makeup to the Spent Fuel Pool from Service Wate ANSWER:
At Align RWST to gravity feed the spent fuel poo REFERENCE: EOP 3505A, Att. A, Step 3 JUSTIFICATION: Gravity feed is the preferred method to the spent fuel pool. (A is correct)
Emergency makeup using the fire water system is only used if the additional attempt of emergency makeup from the RWST is attempted after the gravity feed method does not work (B incorrect).
C and D are least preferred and are only done if RWST is not
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available (C & D are incorrect).
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s OBJECTIVES: E0503C; E05A3C K/A: 033 A2,02, Loss of Spent Fuel Cooling EOP actions Exam item 1295 c
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SRO53 The following events have occurred:
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1 - A SGTR has occurred subsequent to a steam break inside containment.-
- The crew has transitioned to E-3, "SGTR", from E-2," Faulted Steam Generator =
-Isolation."
- - ; They have identified and isolated the ruptured Steam Generator, which is not
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faulted, and are preparing to initiate RCS cooldown. -
Current Plant Conditions:
v_4
-+ Containment temperature is 185*F
- RCS pressure is 1420 psia
- Ruptured Steam Generator pressure is 895 psig ..b
- +' Intact Steam Generator pressures are 850 psig
- - Faulted Steam Generator is 600 psig e Core exit temperature is 485*F Using the provided reference, determine the required core exit temperature to be achieved by the RCS cooldown,if necessar A cooldown is not necessary, core exit temperature is already less than the required temperatur *F *F-
- *F
-ANSWER:-- F-
. PROVIDE ATTACHMENT TO STUDENTS -
REFERENCE: -- E-3 Step 14a graph (Adverse CTMT parameters to be used) -
. JUSTIFICATIOR CTMT 185*F Adverse parameters. Step states not to interpolate, ,
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therefore,885 psig on graph to be used. A incorrect since RCS is above the required temperatur F = non adverse number for 850 psig (D incorrect).
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OBJECTIVES: E3003C K/A: 038 A1.34, Cooldown to specific temperature
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97 EOP Exam 681
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SRO 54 PLANT CONDITIONS The unit is nutning down in power from 100% to take the unit of f line.
Loop 3 Tave falls unobserved to a constant output of 572'F.
Which one of the following describes where pressurizer level will stabilize under these plant conditions? Assume no operator action taken.
Pressurizer level will stabilize at: %
1 % % % '
ANSWER: %
REFERENCES:
JUSTIFICATION: Pressurizer level is progratamed with auctioneered hig'a Ta Pressurizer level will deercase, : hen control at program level for 572*F which is 45%. (C is correct).
13 is inconect because pressurizer level will not be decreased no load valu A is incorrect because pr ssurizer level will not decrease to cause let down isolatio D is incorrect as pressurizer level will not increase above program i valu OBJECTIVES: PPLO7C R/A: 0094 A1.02, CVCS/Tav/Pressurizerlevel
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SRO55 Which of the following signals will cause 3LWS IIV77. Waste to Discharge Tunnel Stop valve to CLOSE7 liigh radiation, only, it liigh radiation and Low flow from the Cire Water syste liigh radiation and liigh flow through the monito liigh radiation and Low flow through the monito ANSWER: liigh radiation, onl REFERENCE: P & ID 106A; LWS 068T Text JUSTIFICATION: The stop valve closes if radiation reaches 2x the setpoint, onl OBJECTIVES: WF1305C K/A: ape 059 K3.01, Tc..nination of release
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modified from exam bank item 3429
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SROS6 PLANT CONDITIONS:
- Plant startup in progress
+ Reactor power is 42%
+ Turbine load is 425 MWe For the current plant conditions which of the following would immediately start the Motor Driven AFW ptunps?
. * Safety injection .
- Low-Low level in one steam generator 1 * Safety injection
- AMSAC * Loss of DC Ilus I
+AMSAC * Loss of DC Bus 1
+ Low Low levelin one steam generator ANSWER: * Safetyinjection
. Low-Low level in one steam generator REFERENC Functtonal Sheet 15 JUSTIFICATION: The MDAFW pamps start on Si or Low -Low level in i of 4 SGs Based on IM given conditions, turbine load less than 40%,
AMSAC will not be armed. Therefore B and C are not correct, l..oss of DC Bus I will result in the TDAFW pump starting but not (ce MDAFW pump OBJECTIVES: , FWA04C K/A: 061 K4.02, auto start of Aux. feed
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PM4 AllI4 I El lliii I ll modified from 2815
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SRO57 With the plant at 100% power, the "B" RPCCW pump trip The crew enters AOP 3561 LOSS OF REACTOR PLANT COMPONENT COOLING WATE The crew immediately OPENS the RPCCW CTMT header cross-connect valves 3CCP' AOV179A, D, C, and D; and CLOSES the RPCCW CTMT Supply and Retum header isolation valves for the "13" trai If no further operator action is taken, how long can the plant remain at full power and satisfy Technical Specifications? Ihour 13, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> hours Indermitely ANSWER: - ' hours REFERENCE: Tecimical Specification 3. JUSTIFICATION: As long as only the "13" pump is runalng, only one loop is OPERA 13LE, both are required, the crew has 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore to two OPERABLE loops. (C correct)
OBJECTIVES: CCP08C R/A:- 008 A2.01, Loss of RPCCW pump MODIFIED FROM 3437
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SR O 58 The crew is attempting to determine the location of a small leak, approximately 10 GPM, in the charging system.- All systems are in automati mlnnian after the leak begins, the PEO reports an indication ofleakage from 3CilS*MV8438C, Charging Pump A/C Discharge isolation valv What would be a oositive Indicailon the operator would have on the main board that 3CllS*MV8438C is the.nource of the leak?
A. VCT level decreasing with letdown flow constan B. No change in indicated charging flow from its initial valu C. VCT level decreasing with increased t. cal injection flo D. Increased charging flow indicated and pressurizer level decreasin ANSWER:
B. No change in indicated charging flow from its initial valu REFERENCE:-- AOP 3555 P&lD 104A CVCS Text JUSTIFICATION: A leak downstream of CilS*FCV121 would INITIALLY result in a decrease in pressurirer level. FCV 121 will open to restore level and would remain at this amount to maintain leve A leak upstream of FCV121 would have the same initial decrease in pressurizer level, however, the increase in charging flow to restore pressurizer level would only be indicated as the original value because the leak is not "seen" by the flow indicator FT-121, (B correct)
C incorrect because a leak at 3CllS*MV8438C would not cause sealinjection flow to increase D incorrect, charging flow will not be increased and pressurizer level will be constant.
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pumi A is incorrect. VCT level decreasing with letdown flow constant only confinns that the leak is not in the letdown system. It does not pmvide any information relativa to the location of the leak in the charging syste Ol3JECTIVES: A5501C K/A: 004 A3.11, Auto ops. charging / letdown 3.4/ modified from 2581
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l SRO 59 The plant is operating at 100% power, the following leak rate data exists:
- Total RCS leakage,10.0 GPM
+ Known leakage is as follows:
- Secondary leakage:
- A steam generator,0.25 GPM
- B and C, none detected
. D steam generator,0.33 GPM
+ Leakage into the PRT,4.0 GPM
+ Leakage into the Primary Drain Transfer Tank,4.5 GPM
+ Leakage into the Containment Drain Transfer Tank,0.5 GPM Which of the following RCS Leakage Technical Specifications, if any, hcVe been exceeded? Identifie . Unidentifie Reactor to Secondar None, all leakage is within Technical Specification ANSWER: None, all leakage is within Technical Specification REFERENCES: Tech Spec 3.4.6.2 and T.S. definitions for leakag JUSTIFICATION: There is 0.58 GPM leakage total to the SGs, and less than 500 gpd thru any 1 SG (about 0.35 GPM), C incorrect, Identified leakage is 9.73 GPM (4 + 4.5 + 0.5 + 0.58), TS requires greater than 10, A incorrect. Total leakage is 10.0 therefore Unidentified is 0.42, less than Spec of 1.0 B incorrect, D correc GBJECTIVES: RCS09C(b); A5502C K/A: 2.1.33, Recognize entry into LCO modified from 652 l
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SRO 60 The following plant conditions exist:
+
1he plant was operating at 100% power when a large LOCA occurs.
+ 1he crew is preparing to swap over to cold leg recirculation
- The "A" and "B" RSS pumps failed to start and are currently considered out of senice.
+
The crew has elected to use the "C" RSS pump for cold leg recirculation.
Which of the following actions minimizes the potential for overheating the "C" RSS pump? The flowpath from the "C" RSS pump to the containment spray header will not be isolated until a flowpath to the Charging /SI pumps la established, 1 The "A" Train Sequencer will be placed in " Test 2" mode, with .he "C" RSS pump " Test / Inhibit" switch in " Inhibit". The "C" RSS pump is equipped with a motor operated recirculation valve that will return a minimum amount of flow back to the containment sum The "11" train charging and Si pumps are aligned in the injection Mod ANSWER: The flowpath from the "C" RSS pump to the containment spray header will not be isolated until a flowpath to the Charging /S1 pumps is establishe REFERENCE: ES-1.3," Transfer to Cold Leg Recirculation": step 2.f and Att. C P&lD EM 112C JUSTIFICATION: Step 2.f RNO directs the operator to align the "C" RSS pump using Att. C of ES 1,3 only if the "A" and "II" RSS pumps are unavailable. This means the "B" train flowpath will not be realigned for Cold Leg Recirculation, llowever, distracter D is incorrect because the "C" RSS pump does not provide flow to the
"B" train flowpath until after flow is established to train "A". (See ES 1,3, Att.C steps 3-9). This means the pump would heat up unless flow through the pump is assured as soon as flow is established in train "A".
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C is incorrect because the "C" RSS pump does not have a min flow recirculation valve /lin is incorrect because the sequencer lineup has nothing to do with protecting the pump from overheating. The steps ensure the pump restants on an LOP The lineup the operators follow specifically ensures that the flowpath to the charging /S1 pumps is established prior to isolating flow to the containment spray header. (A is correct)
OlljECTIVES: CDA05C K/A: 026 A2.04, Failure of spray pump modified from 2473
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SR0 61 Which of the following action (s)is necessary to transfer control of the steam generator atmospheric dump valves to the Auxiliary Shutdown Panel and enable valve position indication? Select " LOCAL" on the controller selector switch to both transfer control and enable valve position indicatio . Select " LOCAL" on the push button below each controller to transfer control and select " LOCAL" on the controller selector switch to enable valve position indicatio Select " LOCAL" on the push button below each controller to transfer control and enable valve position indicatio Select " LOCAL" on the controller selector switch to transfer control and select
" LOCAL" on the push button below each controller to enable valve position indicatio ANSWER: Select " LOCAL" on the controller selector switch to transfer control and select
" LOCAL" on the push button below each controller to enable valve position indicatio REFERENCE: EOP 3503 JUSTIFICATION: Step informs the operator to perform 2 steps, select LOCAL on the controller selector switch to transfer control and select LOCAL on the push button below the controller to enable valve position indicatio OBJECTIVES: ASP 07 K/A: APE 068 EK2.03, Relationship controller /positioners
New question
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SHO 62 The control switch for PORY 3RCS*PCV456 at MB4 is taken to CLOSE and the LOCAL / REMOTE switch at the TSP is placed in LOCA Which of the following describes the response of the PORY if RCS pressure increased to 2350 psla?- The PORV will not open until the switch at MB4 is taken back to AUT The PORV will open but will not close until the switch at the ASP is placed in CLOS The PORV will open and close normall : The PORV will not cpen until the REMOTE / ISOLATE switch at the FTSP is also taken to LOCA ANSWER: The PORV will open and close normall REFERENCE: LSK 251.2B JUSTIFICATION: In LOCAL or REMOTE, the PORV will open as long as the switch at the ASP (when in LOCAL) or MB4 (REMOTE) is in AUTO. (A incorrect, C correct)_
B incorrect, going to close will close the PORV however, this action is not required, the PORV will close once RCS pressure is below the 111111 setpoin D incorrect, there is not a FTSP switch for this POR OBJECTIVE: ASP 08 K/A: 068 EA1.21. Transfer control panel
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SRO 63 The plant is at ful' : owe The nonnal power supoly to inverter 6 trips ope minutes later, the operators locally aligns to the alternate supply to IAC-6 using the
" ALTERNATE SOURCE TO LOAD" push betto minutes afler transferring to the attemate supply, an LOP occur Which of the following describes the status ofinverier 67 Invener 6 is de-energize . Inverter 6 is energized but will de energize in 5 minute Inverter 6 is energized but will de-energize in 20 minute Inverter 6 is energized but will de energize in 30 minute ANSWER: Inverter 6 is de energize REFERENCE: EE lllA,063T JUSTIFICATION: When the normal power sup sly is lost, a timer in the breaker to the inveder begins to time out, afler 30 minutes the breaker trips open, isolating the DC supply (and the nonnal AC supply) from the inverter. Aligning the alternate supply does not change or reset the timer, in this case, when the LOP occurs the 30 minute has timed out and therefore the inverter will be de energized. (A correct)
Ol3JECTIVES: 12507C K/A: 063 K4.02, auto swap modified from exam b:mk item 2490
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SRO 64
'lhe RWST is at the Technical Specification minimum when a design basis LOCA occurs. Only one train of.LGefrtomponents actuat G9PrGMRas Approximately how long will it take to reach the automatic RilR pump trip setpoint? minutes 11, 35 - 45 minutes minute D.- 85 95 minute ANSWER:
C.- 50 60 minute REFERENCE: ECC006T JUSTIFICATION: Runout Flow from:
1 CilS pump 550 GPM at 650 psia 1 SIS pump 650 GPM at 650 psia
! QSS pump 5000 GPM at 120 psia 1 RllR pump 5000 GPM at 120 psia total: 12,000 GPM TS minimum level: 1,166,000 gallons Switchover setpoint: 520,000 gallons amount to be injected: 646,000 gallons time to inject 646/12 = 54 minutes
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OBJECTIVES: ESF13C K/A: 006 KS.06 ECCS llow/ pressure modified from exam bank item 2643
l SRO 65
'llr "A" EGLS is in Test I to perfonn testing of the RilR pump. The "A" RllR pump toggle switch aligned to the " TEST" position, all other components are in "lNiilillT".
A loss of offsite power occur Which of the following describes the response of the "A" EGLS7 The EGLS sees both the LOP and the Test signal, resets and starts all equipment in the required mode and the RilR pump is OF I The EGLS sees both the LOP and the Test signal, but only starts the RilR pump.
t The EGLS comes out of Test, resets and starts all equipment in the normal LOP mode, the RilR pump is OF The EGLS comes out of Test and resets, however, none of tl.. .:quipment in
" Inhibit" is sequenced in the normal LOP mode and the RllR pump is runnin ANSWER: The EGLS comes out of Test, resets and starts all equipment in the normal LOP nmde, the R1IR is OF REFERENCE: LSK 24 9.411 & 27 7K JUSTIFICATION: If an actual accident occurs during a Test I sequence, the EGLS will automatically reset to normal and carry out its automatic actions. Test /lrihibit toggle switch position will have no effect on equipment operation. (A,13, and D incorrect)
C is correct, the RiiR pump is not started for the LOP and will not
, be started by the sequencer. All other equipment, such as the service water pumps will start in their normal sequenc Test 2 functions identically to Test I with two major exception The first is that in addition to responding to the local Test pushbuttons it will also respond to external input signals, whether generated by l&C or by actual plant conditions. The second is that an actual external input signal will not reset the sequencer to nonnal; therefore, the Test / Inhibit switches remain in effect and will determine if each piece of sequenced equipment will actually start (test position) or not (inhibit position).
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OllJliCTIVES: EDS06C K/A: 064 K4.11, Autoload sequencing modified from exam bank iteia 258 i
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I SRO 66 The crew has entered FR.S.1, " Response to Nuclear Power Generation /ATWS" and is aligning for immediate boration of the RCS through one gravity feed boration valv Which of the following would be an acceptable flowrate? Any flow between 5 and $0 gp . Any flow between 25 and 75 gp Any flow between 40 and 90 gp Any flow greater than 100 pp answer: Any flow between 40 and 90 pp REFERENCE: FR 3.1, step 4 JUSTIFICATION: Step 4b. RNO states the limit when using at least one gravity feed boration valve is less than 100 gpm. Step 4c has the operator verify flow equal to or greater than 33 gp A and 11 incorrect, minimum flow is less than require C correct, both values are acceptabl '
D incorrect, flow is greater than allowe K/A: APE 024 AK2.01, Relationship boration flow / valves OBJECTIVE: FS103C New question l
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r SRO 67 Elant conditions e 100% power
- lt is discovered that the crew has violated the Reactor Core Safety Limi What ACTION is trattired? Take action to be within the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, be in llOT STANDBY within the next hour, and comply with the requirements of Specification 6. . Be in llOT STANDBY withm I hour and notify the NRC within the following hou lie in llOT STANDilY and notify the NRC within I hou Ile in llOT STANDilY and notify the NRC, the Senior Vice President and CN Millstone within 1 hou ANSWER: lie in llOT STANDilY and notify the NRC within I hou REl'ERENCE: Technical Specifi.:ations 2.1.1,2.1.2, the 2.1 bases, and section 6. JUSTIFICATION: If Specincation 2.1.1 (Reactor Core) is violated the plant must be in llOT STANDBY within I hour and comply with Specification 6.7.1. Specification 6.7.1 requires the following:
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Unit in llOT STANDBY within I hour
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NRC Operations Center notilled as soon as poesib!c and in all cases within I hour, Senior VP Millstone and the Chairperson of NSAB notified within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Safety Limit Report prepared
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Report submitted to the NRC within 14 days
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Operations not to resume until authoorized by the NRC OBJECTIVES: SAD 821 R/A: 2.2.25, Knowledge of Safety Limits
7pubuu WMii New question
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SRO 68 The SIGM A refueling machine operator hasjust removed a fuel assembly from the core and started moving toward the upender.
A failure of the refueling cavity seal occurs. The fuel transfer cart is inside containment and the water level in the refueling cavity is dropping quickly.
What action should the SIGMA nefueling machine operator take? Return to the core and place the es'sembly in a core location.
1 Place the fuel assembly in the upender, and return to horizontal position. Place the fuel assembly in the north saddle area of the refueling cavity and unlatch the assembly. Place the fuel assembly in the north saddle area of the refueling cavity and lower the assembly until the cable is slack.
ANSWER: Retum to the core and place the assembly in a core location.
REFERENCES: EOP 3572 JUSTIFICATION: Desired location for an assembly is back in the core if the SIGMA is near the core, if the operator hasjust started moving, then this would be the appropriate location. (A correct)
B inconect, this is the location for an assembly when the upender is in containment and the refueling machine is away from the cor C is incorrect, it is never desired to unlatch the assembl D incorrect, this location is appropriate if the refueling machine is away from the core and the transfer carriage is not in containment OBJECTIVES: FilS06C K/A: ape 036 A1.04, refuel:ng operations / fuel handling accident 2.2.31 modified from exam bank i'em 209
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SR0 69 The plant is at full powe ESF testing results in inadvertent 1112 actuation. All equipment operates as designe Whici of the following reactor trip signals will be generated? Safety injectio . liigh Steam Pressve Rat Low-Low Steam Genera 1or Leve Low Pressurizer Pressur ANSWER: Iow-Low Steam Generator leve REFERENCE: Westinghouse functional sheet 8 JUSTIFICATION: A, B, and D incorrect,1112 will not actuate St. The !!igh Steam Pressure Rate will only cause a MSLI. Pressurizer pressure will not go low in this situatio C correct, MSiv..will shut, causing steam generators to shrink off scale, causing a reactor trip. Another possibility would be high pressurizer pressur OBJECTIVES: RPSO4C (g)
K/A: 013 Kl.01, ESF initiation signals modified from 2657
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SRO 70 A loss of refueling cavity seal has occurred and level in the refueling cavity is decreasing slowly.
An operator on the spent fuel side begins to close the transfer tube gate valve prior to a spent fuel assembly being transferred from containment to the spent fuel building. When the valve is moved back to the open position, the operator cannot return it to the full open position.
Which of the following interlocks, if any, would have to be bypassed to allow the assembly to be transferred to the spent fuel building? None.
1 The " valve interlock" only. The " traverse interlock" only. Iloth the " valve interlock" and the " traverse interlock".
ANSWER:
1 The " valve interlock" only.
REFERENCE: OP3303C JUSTIFICATION: The_ bypass interlock key for " Valve Interlock." allows the cart to bc_ traversed with the transfer tube valve not full onen. (A incorrect,il correct)
The bypass interlock key for " Traverse Interlock," allows the cart to be traversed with no power to containment side control panel and/or the " Traverse Control," switch in the "OFF" position on the containment side control panel (C and D incorrect)
OllJECTIVES: FSI105C K/A: ape 036 K2.01, Fuel handling equipment new (based on 1582)
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SRO 71 PLANT CONDIT:0NS:
- The unit is in MODE 5
- RWST levelis 450,900 gallons lhe crew is evaluating conditions for filling the refueling cavity from the RWS Should the crew make the decision to fill the cavity? Yes, there is adequate level in the RWST to fill the cavity, there is no minimum
, RWST level required in MODE 5, and the water in the cavity can be used to satisfy the shutdown risk minimum inventory requirement No, to fill the refueling cavity requires more than 450,000 gallon Yes, there is adequate volume in the RWST to fill the cavity and still be above the required Technical Specification leve No, filling the cavity would result in an RWST level less than the minimum amount required by Technical Specifications and shutdown risk minimitm inventory requirement ANSWER: No, filling the cavity would result in an RWST level less than the minimum amount required by Technical Specifications and shutdown risk minimum inventory requirement ,
REFERENCE: TS 3.1.2.5 & SFC033T, OP3305, OP3260A-4 JUSTIFICATION: The cavity holds about 260,000 gallons (B incorrect) and is filled with water from the RWST in preparation for refueling. TS
< requires 250,000 gallons. filling the cavity will leave 190,000 gallons,(C incorrect).
If the Boric Acid storage system is OPERABLE,'.he RWST could be drained without entry into Technical specification 3.1.2.5, however, OP3260 requires the RWST be maintained greater than 250,000 gallons. (A inco Tect, D correct)
OBJECTIVE ,2523 and 2041
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K/A 034 A1.02, Levelin refueling canal
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SRO 72 PLANT COND1110NS:
CETs are 1250*F
+
A loss ofoffsite power has occurred
+
"A" EDO did not stan
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' "B" Charging and "ll" SI pumps have tripped and cannot be restaded
+
No AFW pumps are running, narrow range levels in all steam generators is off scale low
+
Wide range levels in all four SGs are approximately 45% and trending down slowly Which of the following is the PREFERRED method for cooling the core for the existing plant conditions? Depressurire the secondary system at a rate not to exceed a cooldown rate of 100'F/ hour, 1 Dump steam to the condenser at the maximum mt Restart reactor coolant pumps, one at a time until CETs are less than 1200* Open all pressurizer PORVs and reactor head vent ANSWER: Open all pressurizer PORVs and reactor head vent REFERENCE: FR- JUSTIFICATION: No offsite power, RCPs unavailabl No AFW pumps and narrow range less than 6%, cannot depressurize the secondary No liigh head pumps, cannot restart high head injectio only available option is opening PORVs and reactor head vents (D correct)
OBJECTIVES: FC102C K/A: 074 K1.03, Process of removing heat modified from exam bank item 2520
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SRO 73 in accordance with Millstone 3 procedures, when is the appropriate time for tmnsferring to hot leg recirculation? hours afler the LOCA occurre hours aRet Emergency Core Cooling has been placed in Cold Leg Recirculatio hours after the LOCA has occurred hours after the Emergency Core Cooling Systems have been placed in Cold Leg Recirculatio ANSWER: hours after the LOCA has occurred REFERENCE: Millstone 3 Basis Document for E 1. Step 22 JUSTIFICATION: 'Ihe requirement in the Mp3 procedures is to shift to llot Leg Recirculation 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after the initiation of the event (LOCAXD is-incorrect). The procedures do require the operators to prepare for llot Leg Recirculation 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> aRer the event started (A and B are incorrect)
OBJECTIVES: ESFilC K/A:- 011 A.l.11. Long term core cooling _
modified from exam bank item 866
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SRO 74 The crew is responding to a LOCA in the Auxiliary Building using ECA 1.2,"LOCA Outside Contaloment."
If the break is isolated, which procedure will the crew transition to from ECA-1.27 ES 1.2," Post LOCA Cooldown and Depressurization." E 1, " Loss of Reactor or Secondary Coolant."
w ECA l.1," Loss of Emergency Coolant Recirculation." ES-1.1, "Si Tennination."
ANSWER: E 1, " Loss of Reactor or Secondary Coolant."
REFERENCE: ECA-1.2, Step 5 JUSTIFICATION: Step 5 of ECA-l.2 has the operator check if the break has been isolated. If the break is isolated the crew transitions to E-1. If the break is not isolated the crew is transitioned to ECA 1.1 per the step 5 RNO. (B correct, A, C and D incorrect)
"A" is incorrect because there is no entry into ES-l.2 from ECA- "B" is correct because if the break is isolated the crew should go E- "C" incorrect, this procedure is the choice if the break is not isolate "D" is incorrect because the crew must first go to E-1 and subsequently to ES-1,1 OBJECTIVES: Al204C K/A: West E04, EK 1.2, Associated nonnal, abnormal, EOPs New question
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SKJ '/5 Opo:thn of which of the following RPCCW valves,if any, would be DIRECTLY affectvd by Safety injection actuation? No RPCCW valves are directly alTected by Si actuation.
1 RPCCW system cross-connect to chilled water system valves,3CCP'h10V222 thru 229. Containment header cross-connect valves,3CCP'AOV179A&B and 3CCP'AOV180A&B. The A/B RPCCW non safety train headers,3CCP'AOVl97A/B,10A/B,194A/B, 19A/B, ANSWER: Containment header cross-connect valves,3CCP* AOV179A&B and 3CCP'AOVl80A&B.
REFERENCE: P&lD 121 A & B JUSTIFICNIlON: B and D incorrect, valves operate on CIA, not Si C conect, the containment cross-connect valves, will close, if open on S A incorrect, the cross-connect valves are directly affected by the St.
OBJECTIVES: CCP04C K/A: ape 026 EK3.02, Auto actions on ECCS actuation modified from millstone LOIT exam 1995
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SRO 76 FR Z.1, Response To liigh Containment Pressure, places the hydrogen recombiners in service: hours after a large break LOCA When llydrogen concentration reaches 4% Following any reactor vessel head venting
- When Hydrogen concentration reaches 0.5%
ANSWER: When Hydrogen concentra; ion reaches 0.5%
REFERENCE: FR-Z.1, Step 12.c, starts hydrogen recombiners if hydrogen s concentration reaches 0.5%
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_ JUSTIFICATION: The FSAR analysis assumes the hydrogen recombiners are started
.rm 'n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA with an estimated hydrogen m .~ .:-* ration of 1.6%. The FSAR notes that by procedure the wcombiners are started well before this level is reached (A is incorrect)
FR-1.3, Response to Voids In The Vessel, ensures hydrogen
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recombiners are already operating prior to venting the head. It
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determines a venting time to reach a concentration ofless than 3%.
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' 3 and C are incorrect)
I OBJECTIVES: FZ103C K/A: West F14 EKl.2, Loss of containment integrity new question
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- SRO 77l
- PLANT CONDITIONSi-
- .~ plant shutdown in progress C- reactor poweris 7%
Intermediate range channel N35 fails as indicated by "lR 1 Loss of Compensatio Voltage" annunciator alanning at MB4C 5- .
1 The cirw has entered AOP 3571, " Instrument Failure Response",
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Which of the following actions should be taken by the crew?
A,'- Continue with the shutdown, actuate both SR reset switches when reactor power is believed to be in or near the Source Racge, 4 . Reduce power to below 5% and remain there until the channel is restored to OPERABLE statu C.- Stop the shutdown and remain in MODE 1, less than 10%, until the channel is restored to OPERABLE statu Continue with the shutdown, and trip the applicable bistables, ANSWER:
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- A; -- Continue with the shutdown, actuate both SR reset switches when reactor power !
is believed to be in or near the Source Range,'
REFERENCE: AOP 3571 Rev. 3 Attachment E-JUSTIFICATION: AOP 3571, step 2 states that if the IR fails during a shutdown DR
- has symptoms ofimd-comnenution, Actuate both source range reset when reactor power is believed to be in or near the Source -
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Range (A correct) _
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The Technical Specification ACTIONS apply'to increasing power,_
-(B and C incorrect)-
AOP 3571 informs the operator that there are no bistables to be tripped, (D incorrect).
- OBJECTIVES: NIS05C; NIS07C -
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K/A: 033 K3.02, Loss ofintermediate range 1995 LOIT SRO exam
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SRO 78 An LOP occurs and both diesels start and energize buses 34C and 34D.
Subsequently, the operator actuates CDA by pressing one of the 2 sets of 2 push-buttons for CDA on MB2.
Two minutes later, the operator notices that the "A" RPCCW pump is running and the
"B" RPCCW pump is not.
Which of the following describes a probable cause for these indications? The "A" sequencer has malfunctioned and the "B" sequencer has operated correctly because both trains of CDA are actuated by pressing either of the 2 sets of 2 push-buttons. The "A" sequencer has functioned correctly and the "B" sequencer has malfunctioned because all 4 push-buttons must be pushed to actuate both Trains ofCDA. Both sequencers have operated correctly because the manual CDA signal is train specific and only the "B" train must have been actuated by the operator. The "A" sequencer has malfunctioned and "B" sequencer has operated correctly because both trains of CDA have actuated and the RPCCW pumps are locked out from starting on a CDA for 6 minutes.
ANSWER: The "A" sequencer has malfunctioned and the "B" sequencer has operated correctly because both trains of CDA are actuated by pressing either of the 2 rata of 2 push-buttons.
REFERENCE: LSK 24-9.4A,27-18A JUSTIFICATION: For a LOP and LOP /SI, the RPCCW pumps are restarted. For a CDA they are not. Actuating CDA by using either set of 2 push-buttons will actuate both trains of CDA (C incorrect, B incorrect).
The "A" pump is running, therefore, the "A" EGLS has not performed properly, the "B" pump should not be running, therefore, the "B" EGLS has perform correctly. (Only A correct).
There is no lockout on the RPCCW pumps, (only the RSS pumps)
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OI)JECTIVES: CCP06C K/A: ,
EPE 056 A2.47, Proper OPS, diesel sequencer New question
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SRO 79--
What is the basis for isolating the RCPLSeal Supply Isolation valves, 3CHS'MV8109A,B,C and D in ECA 0.0, " Loss of All AC Power"?
- Protect the system from steam formation due to RCP thermal barrier heating.
' Ensure the necessary amount of ECCS flow is available, if required, when an emergency bus is energize Allow starting a charging pump in the normal charging mode as pan of recover witi,out concem for damaging the RCP Prevent a LOCA outside containment of potentially a few hundred gallons per minute should RCP number 1 seal failure occu !
ANSWER: Allow starting a charging pump in the normal charging mode as pan of recovery without concern for damaging the RCP REFERENCE: ECA-0.0 Bkgd Doc, step 8 JUSTIFICATION: Isolating the RCP seal injection lines prepares the plant for recovery while protecting the RCPs from seal and shaft damage that may occur when a charging /SI pump is started as part of the recovery. With the =ani iniection lines isolated. a charging /SI pumn can be started in the normal charging mode without concern -i for cold meal iniection flow thermally shocking the RCPs. Seal injection can subsequently be established to the RCP consistent i with appropriated plant specific procedure ,
OBJECTIVES: A0002C K/A: 022 K3.07, Isolating charging modified from exam bank item 1999
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SRO 80 The plant is in MODE 5 at mid loop in accordance with OP 3270A, " Reduced Inventory Operation Mode 5 (IPTE)" Both trains of RHR are in service at 1000 GPM, RCS level begins to decrease and amps for the RHR pumps begin to oscillat Which of the following actions should be taken by the crew? Maximize charging flow to increase RCS level while maintaining RilR flow at 1000 gp Trip both RHR pumps and immediately perfona the appropriate actions for venting and restarting the pumps in accordance with OP3310A," Residual Heat Removal System". Trip both RHR pumps and Go to EOP 3505, " Loss of Shutdown Cooling and/or RCS Inventory". Trip one RHR pump, allow the system to stabilize, vent and restart the tripped pump, then repeat the sequence for the other RHR pum ANSWER: Trip both RHR pumps and Go to EOP 3505, " Loss of Shutdown Cooling and/or RCS Inventory".
REFERENCE: OP 3270A JUSTIFICATION: 4.2.2 At any time, IF impending loss of both RHR trains is evident or an uncontrolled decrease in RCS level occurs, TRIP the RHR pumps and Go To EOP 3505, " Loss of Shutdown Cooling and/or RCS Inventory "
question has both indications, C is correc OBJECTIVES: E0502C K/A: EPE 025 K3.03, Auto actions on ECCS question modified from 1297, which was probably not used in program (still in 3 distracter format)
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ESRO 81 initial Conditions:
- The unit is at 100% power e The South 345kv bus was lost due to a fault, it will be restored to service in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
. A fault on "B" NSST results in a reactor tri Following the reactor trip, all AFW flow is lost. Attempts to establish Auxiliary _
Feedwater have failed. All steam generator wide range levels are 35% and decreasin Assuming no change in plant conditions or trends, which of the following will be the next recovery action taken by the crew? Establish Main Feedwate Bleed and Feed of the Reactor Coolant Syste Depressurize the secondary by dumping steam to the condenser to establish .
- condensate flow to at least one steam generato Depressurize the secondary by dumping steam to the atmosphere to establish-
, condensate flow to at least one steam generator. -
ANSWER: ! Bleed and Feed of the Reactor Coolant Syste REFERENCE: Potential Core Damaging Event: Loss of Secondary Heat Sink, FR-H.1, EE-1 A, LSK 24-2D,24-3B and 24-2B JUSTIFICATION: Under the step to establish _a secondary heat sink, using main feed '
is preferred over condensate or bleed and feed. Either condensate
- or main feedwater will restore secondary heat sink. ~ However, both are unavailable because the fast transfer of 6.9KV buses is defeated by the South bus being de-energized. If the current trend continues, the crew will be initiating bleed and feed once wide range levels are less than 27%.
OBJECTIVE: MC1104
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K/A: W/E 05 EKl.2,- Procedures for -
modified from exam bank item 3197
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SRO 82 Inverter 2 output breaker to VI AC-2 trips on a spurious signa Which of the following describes how to re-energize the VIAC7 The static switch will automatically transfer to the DC sourc Manually transfer to the altemate AC power source using the manual bypas The static switch will automatically transfer to the attemate AC power sourc Manually transfer to the altemate AC power source using the static switc ANSWER: Manually transfer to the altemate power source using the manual bypas REFERENCE: EE-1BG, OP 3345B JUSTIFICATION: If the output breaker trips, the static switch will not be available, (D, C and A incorrect), the crew will have to manually switch to i the altemate source using the manual transfer switch. (B correct).
OBJECTIVES: 12006C K/A: 057 EA1.01, Manual inverter swapping taken from static sim exam question 26-4 l
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SRO 83
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A plant start up is in progress.- The blue P-10 permissive light hasjust come "0N".
No operator actions have been taken.-
Which of the following will result in an automatic r tactor trip?
- Trip of one RC Power range channel N41 fails HIG = Intermediate Range channel N35B fails HIG The operator at MB-4 places the Block / Reset Switch for NIS Channel 31 in the
" Reset" position and then depresses the butto ANSWER: Intermediate Range channel N35B fails HIG REFERENCES: NIS Text (NIS015T); Data Chapter,Section II.A and Functional Sheet 3 and 4. OP3203, step 5.13-JUSTIFICATION: 5,13.1 Observe the P7 permissive blue light is OFF (MB4D,5-3)
and the P10 permissive blue light is ON (MB4D,4-3) then-5.13. _ BLOCK both channels of Power Range Reactor Trip (low range).
5.13. BLOCK both channels ofIntermediate Range Reactor Tri Above P-10 and below P-8, reactor trip occurs on 2 RCPs trip (A incorrect)
Above P-10, the Source Ranges are automatically blocked from energizing,(D incorrect)
Power range high flux (high or low) is 2 of 4, even ifit assumed that the high failure causes the bypass reg valve to open, high'
steam generator level does not cause reactor trip,(B incorrect)
A failure of an IR will cause a trip because until the trip is blocked by the' operator it will function, (C correct)
OBJECTIVES: NIS04C (a.2)
K/A: 015 K4.08, NIS permissives 3.7/ modified from exam bank questions 988 and 56 o
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' SRO 84 -
- INITIAL CONDIT10NSi
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The plant is operating at 30% power. .
- The 'N TDFW pump is in sersic *
All control systems are operating in Automati The controlling channel of Steam Flow for the 'D' Steam Generator fails lilGl L Assuming NO operator actions, which of the following describes the expected plant =
response?
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' + ; Main Feed pump speed increases a
"D" Main Feed Reg valve opens
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Actual level in "D" steam generator increases
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FWI and turbine trip occurs a Reactor trip on Low-Low SG level
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B .- Main Feed pump speed increases
"D" Main Feed Reg valve opens
Actual level in "D" steam generator increases ;
FWI and turbine trip occurs
. Reactor trip on turbine trip L
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' D" Main Feed Reg valve opens
- Level error turns level prior to FWI
"D" Main Feed Reg valve throttles closed to maintain steam generator
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level
+-
"D" steam generator level returns to program level Dc +
- Main Feed pump speed increases
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"D" Main Feed Reg valve opens >
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- Actual level in "D" steam generator increases
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_ FWI, MSI and turbine trip occurs
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- Reactor trip on Low-Low SG level
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ANSWER: *
Main Feed pump speed increases
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- D" Main Feed Reg valve opens
- - Actual level in "D" steam generator increases
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FWI and turbine trip occurs
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. Reactor trip on Low-Low SG level
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REFERENCE: SGWLC T.M., Functional and process drawings, Simulator JUSTIFICATION: Increase in steam flow on 'D' S/G causes total steam flow to increase which causes programmed MFP DP to increase above actual DP and MFP speed to increase, increase in steam flow causes about 100% flow error (133% steam flow,30% feed flow)
flow error for the 'D' S/G FRV control to increase causing the FRV to open mor Combination ofincreased speed and the FRV being more open causes 'D' S/G level to increase rapidly to the P-14 setpoint resulting in a FW1 and tripping of the MFP's and Main Turbine, At 80% actual level, FWI will occur on P-14. (no reactor trin. below P-9 and P-14 is not a reactor trin signal. "B" incorrect)
FWI causes the main feed lines to isolate to all S/G's and the main feed pumps to trip with MSIV's still open ("D" incorrect); steam dumps will open to remove decay heat and RCP heat, which results in a decrease in level in all S/G's until the AFW pumps start and Auto Rx Trip occurs due to the Low Low level ("A" is correct).
Another main feed pump trip signal will occur after 15 seconds due to the reactor tri OBJECTIVES: FWS07C (9.3)
K/A: 016 K3,14, Effects on SG modified from exam bank item 1519 i
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=. . Which of the following describes the actions, if any, necessary to restore the operation of l the Group A and B pressurizer backup heaters following a loss of power or low pressurizer level cutout? -
A.- For low pressurizer level, heaters will be restored as soon as pressurizer level is restore For a loss of power, the heaters must be tumed "OFF" and then back to " AUTO" B. - For the low pressurizer level, level must be restored and the heaters must be -
turned "OFF" and then back to " AUTO".
For a loss of power, the heaters must be turned "OFF" and then back to " AUTO" C.- For the low pressurizer level,' level must be restored and the heaters must be >
tumed "OFF" and then back to " AUTO".
For a loss of power, the heaters are restored as soon as the manual start block is move > ~ For low pressurizer level, heaters will be restored as soon as pressurizer level is restored.-
For a loss of power, the heaters are restored as soon as the manual start block is remove ANSWER:- For low prescurizer level, heaters will be restored as soon as pressurizer level is restore For a loss of power, the heaters must be turned "OFF" and then back to " AUTO"
- REFERENCE: OP3353.4B,6-3, LSK 25-1.2F JUSTIFICATION: loss of power has locked out backup heater group A or B, CYCLE associated control switch to "OFF" and back to " AUTO" (MB4),
for low pressurizer level, as soon as level is restored, the heaters should be restored, A correc OBJECTIVES: PPLO4C K/A:- 010 K6.03, Effects heaters / spray /PORV New question i
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SRO 86 A failure of a pressurizer pressure channel occurred during which the operator mistakenly adjusted the pressurizer pressure master pressure controller,3RCS PK455A, down by 100 psia. All actions in AUP 3571, Instrument Failure Response, for the failed pressure channel have been completed. The operator still has PK455A in manual and RCS pressure is 2235 psia, trending up slowly. The control room team is preparing to place the ,
inessurizer pressure master pressure controller,3RCS PK455A in Automati r The US has directed the RO to use the applicable procedures and restore pressurizer pressure control to automatic at normal RCS pressure.
Which of the following describes the appropriate sequence of steps necessary to place the controller in automatic? * Set setpoint to desired pressure
- Check output is zero
. Place pressurizer pressure master pressure controller in auto
- Place ALL heaters and spray in auto . Set setpoint to desired pressure
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Adjust pressurizer pressure to setpoint
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Place pressurizer pressure master pressure controller in auto
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Place heaters (except, turn on "C") and spray in autc, =
Adjust pressurizer pressure to desired pressure
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Set setpoint to match pressurizer pressure
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Place heaters (excepi, turn on "C") and spray in auto
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Place pressurizer pressure master pressure controller in auto *
Adjust pressurizer pressure to desired pressure
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Set setpoint to match pressurizer pressure
. Place pressuriz.cr pressure master pressure controller in auto
. Place ALL heaters and spray in auto ANSWER: * Set setpoint to desired pressure
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Adjust pressurizer pressure to setpoint
Place pressurizer pressure master pressure controller in auto
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Place heaters (except, turn on "C") and spray in auto REFERENCE: OP3301G, Section JUSTIFICATION: Procedure OP330lO is a General Use procedure and contains the following steps:-
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4.1.4 SET setpoint controller,3RCS PK453A (MB4), pressurizer pressure master controller, to desired pressure to be maintained (normal operating pressure is 2,250 psia).-
4.1.5 ; With pressurizer heaters and spray valves in manual, ADJUST pressurizer
, pressure to match master controller setpoin .1.6 - To place Pressurizer Pressure Control System in automatic, PERFORM following, PLACE 3RCS-PK455A (MB4), pressurizer pressure master controller, in
" AUTO."
subsequent steps have heaters, A, B, D and E in auto, C "on" and then spray controllers in auto OBJECTIVES: PPLO4C (c)
-K/A: 010 K6.03, Controllers and Politioners New question -
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SRO 87 The unit is in MODE Why does OP3337, Radioactive Gaseous Waste System, inform the operator that if the stack discharge stop valve,311VR*V42 is OPEN, either a process vent fan,3GWS-FNI A or 3GWS FN1B, or a supplementary leak collection and release exhaust fan, 3HVR*FN12A or 3HVR*FN12B, should be in operation? Ensure appropriate dilution air flo , Ensure the cal:ulated radiation monitor alarm setpoints are based on the actual discharge flowiat Provide a positive air flow to the Unit I stack, pieventing backflow from the Unit I stac Provide a recirculation flow path for the GWS syste ANSWER:
v Provide a positive air flow to the Unit I stack, preventing backflow from the Unit I stac REFERENCE: OP3337 JUSTIFICATION: Precaution in OP3337 During periods when the Supplementary Leak Collection and Release System is required to be OPERABLE, and stack discharge stop valve,3HVR*V42 is OPEN, either a process vent fan,3GWS-FN1 A or 3GWS-FNIB, or a supplementary leak collection and release exhaust fan,3HVR*FN12A or 3HVR*FN12B, should be in operation so that positive flow exists to the Unit I stack. IF none of the fans are available, stack discharge stop valve,3HVR*V42, should be close OBJECTIVES: WFB06C K/A: 2.3.3, Radiation control, aux, system outside control room
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SRO 88 PLANT CONDITIONS:
- - Steam break inside containment occurs from full power -
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All systems operate as designed . .
- - Narrow range level for intact steam generators is 15% with AFW flow at 100:
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GPM to each intact Steam generato * The crew is now in FR P.1,_" Response to Imminent Pressurized Thermal Shock .
Condition" RCS temperaure is stable -
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+- RCS pressure is stable with only the control group of pressurizer heaters energized -
The crew has determined a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> soak is require Which of the following evolutions could be performed by the crew in the next houc?-
A.- - Energize additional pressurizer heater B.: Place auxiliary spray in service. - Increase Al W flow to 300GPM per steam'ge, crator to raise steam generator levels to 50%. : Recire RHR to establish boron concentration and place it in servic : ANSWER:
B.- Place auxiliary spray in servic REFERENCE: FR-P.1, step 25
' JUSTIFICATION: Step 25: Determine if RCS temperature soak is required ' Perform ALL of the following steps:
DO NOT cooldown RCS until temperature has been stable -e
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for I hr (increasing SG level would not be allowed, C -
incorrect) -
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DO NOT increase RCS pressure during the I hr-stabilization period (A incorrect)
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Perform the actions of other procedures in effect which do~
NOT decrease RCS temperature or increase RCS pressure-until the I hr stabilization period is comp.lete -
- -
- Maintain RCS pressure and cold leg temperature within the limits shown on Attachment A (B adverse containment)
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- Maintain cooldown rate in RCS cold legs LESS TilAN
- 50'F in any 60 minute period
"A" is incorrect. - Additional heaters would increase pressur " "C" is incorrect. - Additional feedwater would result in additional :-
cooldow "D" incorrect, placing RilR in service aAer recire would result in a cooldow "B" is correct because lowering pressure is allowe OBJECTIVES: FP103C K/A:- W/E 08 EAl.1, Function / safety systems -
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- SRO 89 x
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LINITIAL CONDITIONS:
.- l The unit is at 100% power
".. '- Master pressure controller set to control at its normal pressure -
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_ Pressurizer backup heaters are "OFF"
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Pressurizer spray valves are " MODULATING"
+
' Pressurizer PORV tailpipe temperature is 120*F .
Which of the following RCS p.tessures would cause this response?
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- psi ' psi = psi psi h
. ANSWER:
C, 2280 psi ,
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REFERENCE: Functional sheet 11
.-JUSTIFICATION: Pressurizer backup heaters tum off at 2233 psia. Spray valves y begin to modulate at 2275 psia,'(25 psia above normal setpoint),
, because spray valves are modulating, "A" and "B" incorrec Pressurizer PORVs lin at 2350 psia, (setpMat plus 100 psia),
tailpipe temperatures are normal, therefore, pressure must be less than 2350, "D" incorrec '
' "C" correct, spray valves would be opening, heaters off, PORVs close ~ OBJECTIVES: PPLO4 K/A:- 002 A3.01, Pressure / temperature / flow - . modified from question 12 l
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SRO 90-With the plant operating at 25% power, one of the 4 operating RCPs trip Which of the following describes the status of the steam generator pressures in the OPERATING loops one minute after the RCP trips? Increased due to increasing steam generator temperatur Decreased due to reactor trip on low - low steam generator leve '
4 No change due to a constant steam deman Decreased due to increased steam flow, ANSWER: Decreased due to increased steam flo REFERENCE: Decrease in Reactor Coolant System Flow Rate (MCORE04)
JUSTIFICATION: When the pump stops, the running pumps will reverse flow through the idle loop. The affected steam generator will do veiy little steaming due its pressure dropping. The unaffected loops will increase their steaming, this will cause Pstm in the steam ;
generators to decrease and Tcold in the loops to decrease als OBJECTIVE: MC0402
- K/A: - 003 K5.04, RCP effects on secondary modified 3181-4 .
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SRO 91 A reactor trip has occurred due to a turbine trip from full power. Narrow range steam generator levels are off scale lo Why does ES-0.1, Reactor Trip Response instruct the operator to feed the steam generators at greater than 530 GPM7 To enhance natural circulation, To provide an adequate heat sink for decay heat remova To ensure the steam generator U-tubes remain " wet" preventing hot and dry steam generator To prevent the formation of steam in the steam generator feed rin ANSWER: To provide an adequate heat sink for decay heat remova REFERENCE: E-0, Background JUSTIFICATION: AFW flow is necessarv for secondary heat sink. If SG level is in the narrow range in at least one SG, a heat sink is availabl However, if narrow range level has not been established, feeding at greater than 530 GPM verifies the ensures a heat sink for decay heat removal. If adequate AFW flow for decay heat removal cannot be established, the transition to the FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK, is necessary to establish an altemate source of feed flow or an alternate heat sin (B correct)
"A" incorrect, RCPs could be running, and neither steam generator level or AFW flow is checked to verify natural circulatio "C" incorrect, the " wet" U-tube concept is a concern in FR- aller generators have dried out and flow is to be established to the "D" incorrect, the J-tubes in the feed ring are to prevent steam formation in the feed lin _ _ _ _ _ _ _
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OBJECTIVES: S0103C K/A: 007 EKl.06, Stabilization / relationship to decay heat modified from exam item number 560
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SRO 92
- Which of the following would prevent the operator from OPENING RHR loop suction
- isolation valve RHS*MV870187-
' RHR to CHG/SI valve (3SIL*MV8804A) CLOSED RSS to RHR cross-connect valves (3RSS*MV8837A) CLOSE C, RWST suction isolation valve (3SIL*MV8812A) OPE Di RCS pressure on PT405 380 psi ANSWER:
- RWST suction isolation valve (3SIL*MV8812A) OPE REFERENCES: P/ID 112A, LSK 27-7B JUSTIFICATION: . To open (MV8701B) -
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3SIL*MV8804A RHR PI A to CHG Pump Valve Closed ("A" incorrect) !
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3SIL*MV8812A RWST to RHR PI A Valve Closed ("C" correct)
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3RSS*MV8837A RSS to RHR XCONN Valve Closed ("B" - "
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incorrect)
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3RSS*MV8838A RSS to RHR XCONN Valve Closed
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Wide-range loop 1 hot leg pressure, as read by PT405/405A must be less thari 390 psia ("D" is wrong)
OBJECTIVES: RHR04C (a) -
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K/A: 005 K4.07, System interlocks 3.2/ modified from exam bank item 169
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SRO 93 While operating at 90% power a control Bank "D" Group 1 rod becomes misaligned higher than the rest ofits group.
The crew has entered AOP 3552, Malfunction of the Rod Drive System, and is currently aligning the affected rod to the rest of the bank.
During the rod alignment, the procedure requires the operators to insert the affected rod until the next lower DRPI LEDjust changes state. The operator then resets the affected group step counter to a value of two steps higher than the affected rod's indicated DRPI position.
Which of the following describes the basis for resetting the affected group step counter to a value of two steps higher than the affected rod's indicated DRPI position? This action will reset the logic cabinet master cycler to ensure proper group stepping. Having determined the rods position, this action will reset the P/A converter to the ,
proper value. DRPI necuracy is two steps above the actual position, therefore, when the next DRPI indicator lights, the rod is actually two steps higher. DRPI indicators will light when the rod is actually within *2 steps of actual, setting the group step counters at the high end is conservative.
ANSWER: DRPI accuracy is two steps above the actual position, therefore, when the next DRPI indicator lights, the rod is actually two steps higher.
REFERENCE: AOP 3552 attachment A, Basis document for step 6. Align Rod, DRPI handout JUSTIFICATION: The wils are placed 6 steps apart with the LED positioned 1/2 the distance between the coils. As the rod steps through the coil, the LED above the coil will light. The band is 2 steps above and 3 steps below the indicated position. With the position of the misaligned rod now located, the affected group step counter can be reset to that position and the rod then inserted to the same height as the rest of the rods in the group as recorded initially. (C correct)
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OBJEC11VES: RP104C K/A: 014 A2.04, Misaligned rod impac't modification of 1539,(not used in previous exams)
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SRO 94 Power to VIAC-1 has been lost and cannat be immediately restore Assume that the operators are taking prompt action per AOP 3564 " Loss of One Proteedve System Channel" and that no bistables are in the tripped condition prior to VIAC . deenergizin Which statement describes the immediate consequence, if any, of this event?
' The reactor will trip if power is below 10% The reactor will trip if power is above 10% The reactor will trip, regardless of power leve The reactor will not trip, regardless of power leve ANSWER: The reactor will trip if power is below 10%.
REFERENCE: AOP3564, Functional sheet 4 JUSTIFICATION: Sheet 4 shows the IR are powered from protective channel I and 1 If VIAC-1 deenergizes, IR 35 will de-energize, causing a trip on Iligh Flux if power is less than 10%
OBJECTIVES: RPSO4C (2)
K/A: 012 Kl.01, RPS & 120v vital 3.4/3,7 New question
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SRO 95 Channel 1 is selected as the primary channel or the only input to all controllers on the Main Board . The loop D Thot instrument has failed lilG The contml room team completed all of the actions in AOP 3571, Instrument Failure Response, for this instrument failure and all applicable controls are in automati Subsequently, pressurizer pressure instrument PT-457 fails HIG Which of the following describes the expected plant response, if any? Reactor remains at powe Reactor trip will occur due to OTAT Reactor trip will occur due to OPAT Reactor trip will occur due to high pressurizer pressur ..
~ ANSWER:- Reactor remains at powe . REFERENCES: AOP 3571 -
JUSTIFICATION: The Thot failure corrective action trips the OT and OP AT bistables -
for loop "D".- When the pressure channel fails, the hi'r oressure trip B/S for that channel will actuate, but it will be the only one, requires 2 of 4, "D" incorrec "A" is correct because the plant will not tri "B" is incorrect because the reactor trip will not occur on OTAT-because failing high will increase the setpoint for the "B" loop and will not energize the required second bistable, (one is tripped in AOP 3571).
"C" is incorrect because pressure does not impact the setpoint.--
- OBJECTIVES: RPS04C -
-K/A: 012 K6.03, Trip logic circuits 3.03 New question ,
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. SHO 96 A loss of All AC power has occurred. The crew is attempting to locally start the "A"
,rm,
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ne operator at the diesel reports that the Service Water valve. 3SWP' AOV39A is close According to ECA 0.0, what actions,if any, are required? Restore instrument air and then open the valv . Vent air from the valve operator
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r Vent air using 3SWi"IIV39A in the EDG enclosur No action required, the valve should open when the diesel is starte ANSWER: Vent air using 3SWP*llV39A in the EDG enclosur REFERENCE: ECA 0.0, attachment E, step 1. P & ID 133D JUSTIFICATION: Attachment 1, step 1 has the operator open the service water valve, ifit is not open by venting air by using AOV39A., (C correct)
"D" incorrect, the LOP should have already opened the valve ( by the stem of the question indicating the valve is closed, this failed to work).
- B" inconect, this is the action required for 3SWP* AOV398
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"A" incorrect, the valve fails open on lostof ai OllJECTIVES: A0003C K/A: 055 EA2.03, Actions to restore power 3.7/ New question
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SRO 97 With the plant operating at 100%, a total loss of ofTsite power occurs, ne crew has entered ECA 0.0," Loss of All AC Power" ne SB0 diesel has been started.
Which of the following will allow the operator to energize emergency bus 34C from the S00 diesel After 60 seconds have clapsed, the operator resets the station LOP signal at the sequencer, 11 After 6 minutes have clapsed, the operator resets the station LOP signal at the sequencer. Afler 60 seconds have clapsed, the operator resets the LOP signal at MB2 and the station LOP signal at the sequencer. After 6 minutes have clapsed, the operator resets the LOP signal at MB2.
ANSWER: After 6 minutes have clapsed, the operator resets the sequencer LOP signal at MB REFERENCE: LSK 24 3K JUSTIFICATION: Once the LOP occurs,if the RSST does not energize the bus after 1.8 seconds, the bus is locked out for 6 minutes to provide sufricient time to restore from the emergency diesel, if the diesel is not able to be placed on the bus, aller 6 minutes, the LOP lockout can be reset and either oft site or the SB0 could be placed on the bus. ("D" correct)
"A" inconect, both signals do not need to be reset and 6 minutes must have elapse "B" inconect, the station LOP signal is the wrong signal to rese "C" inconect, the station LOP is incorrect signal and don't need to do bot OBJECTIVES:
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K/A: 055 EA2.06, Lockouts to restore bus 3.7/ modified from exam item number 2077
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f SRO 98 A plant shutdown due to a large steam generator tube leak is being perfomied. 'the crew believes the tube leak is in the *C" steam generato During the shutdown, the plant trips and Safety injection actuate What is the earliest the operator can isolate steam flow from the *C" steam generator? ' When narrow range level in the "C" steam generator is greater than 6%. Aner completing the inunediate actions of E- Once the transition is made from E-0 to E When directed to do so in E- ANSWER:
D, When directed to do so in E-3.-
REFERENCE: OP 3272 EOP User's Guide Attachment 3 "Special Considerations" JUSTIFICATION: An operator may not isolate the steam line to a ruptured steam generator until directed by the procedure. This is not an identified safe condition unless specifically directed by the SGTR procedure OBJECTIVES: E3003C K/A: 037 EA2.ll, Isolate one/more SG modified from 2277
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SRO 99 A leak develops in the reference leg for the wide range level indication for one of the steam generators.
Which of the following describes the indication the operator would see at the main boards for the affected steam generator? Wide range level would INCREASE, narrow range levels would not change. Wide range level would DECREASE, narrow range levels would not change. Wide range level and one narrow range level would INCREASE. Wide range level and one narrow range level would DECREASE.
ANSWER: Wide range level and one narrow range level would INCIEASE.
REFERENCE: Non Nuclear Instrumentation Text (NN1016T), P&lD 130C and D JUSTIFICATION: AP = Pi u:r - PVARIAllli = 0 at 100% indicated SG Level Parr decreases . As AP approaches 0 tbc levelindicated is lilGIIER than actua The Wide Range instruments share a reference leg with one narrow range steam generator level.(A and il incorrect)
'Iherefore, the level in both the wide range and one narrow range level will cach be INCREASING. (C correct, D incorrect)
OBJECTIVES: MC0202 K/A: 035 K6.03, Loss SG level detectors 2.6/ taken from exam bank items 2469,2168 and 1519
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SRO 100 PLANT CONDITIONS:
- Reactor trip from full power and ESF actuation 5 minutes ago
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RCS Pressure: 1500 psia and increasin +
RCS Tav: 470*F and decreasing
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PZR Level: 10% and increasin +
SO Narrow Range LcHr N! offscale lo * SO Pressure: At; 600 psig
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AFW Flow: 300 GPM per steam generator
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Containment Pressure: normal
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ECCS Equipment: All operating at nonnal flow Which of the following describes the cause of the transient? Loss of Coolant Acciden Steam Generator Tube ruptur Steam break inside containmen Steam break outside containmen ,
ANSWER: Steam break outside containmen REFERENCE: Increase in lleat Removal by the Secondary System (MCORE02)
JUSTIFICATION: Containment pressure is normal, therefore, a steam break inside containment or a LOCA has not occurred, also RCS pressure is increasing. (A and C incorrect). '
If a tube rupture had occurred, RCS pressure and pressurizer level would be not increasing and Temperature would not have dropped to 470 degrees. (B incorrect)
A steam break outside containment would result in lowered steam generator pressure until the MSIVs closed, then RCS pressure and pressurizer level would begin to increase. RCS temperature would decrease due to AFW flow to all steam generator .
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OBJECTIVES: MCO202 K/A: 040 EA2,03, Difference between steam line break vs LOCA New ques.tlon
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Senior Hemeter Operator Answer Key 1. d 2 b 76. d 2. d 2 $2. a 77,a 3. c 28. d 53. c 78,a 4. c 2 . c 79 c 5. c 30. c 55. a 80 g C W PoG f.MM L 6. a 31. b 56, a 81. b enstad. - sem.s . d 82, b
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Attachment 2 Millstone Unit 3 RO WRITTEN EXAM W/ ANSWER KEY
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Which of the following conditions would require the crew to initiate immediate boration of the RCS7 In MODE 5, with dilution paths not isolated and cold calibrated pressurtzer level at 50%
core bumup is 6,000 MWD /MTU and RCS boron concentration is 2050 PPM. (See attached curves.) With the plant at 100% power, an unexplained event is causing Tave and reactor power to increase, and control rods to inser After a reactor trip, with the crew performing ES-0.1, * Reactor Trip Response", the RCS cools down uncontrollably to 540' While performing a, rapid downpower IAW OP 3275, the ROD CONTROL BANKS LIMIT LO Alarm is receive ANSWER: With the plant at 100% power, an unexplained event is causing Tave and reactor power to increase, and control rods to inser JUSTIFICATION: "B"is correct, since this is indication of an unexplained positive reactivity additio CM emed?" "
- A"is ensens since there is adequaR boron per the " loops filled" curve, hM and PZR level is > 40%.
"C" is wrong since the setpoint is 530 "D"is wrong since Rod LO-LO is the immediate Borate setpoin OBJECTIVE: CHS08C (a)
REFERENCE: AOP 3566 Entry Conditions K/A: 024.EK3.01 Emergency Boration Requirements Exam item: 2373 ta.w Geda 4 4 L p (w- M lee % a. wA la,
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R32 The following plant conditions exist:
. The plant has just completed a 428 day full power ru . The plant is cooling down in MODE 3
- RCS temperature is 490*F
. RCS pressure is 1800 psia
. Excess letdown is in service A leak develops in train ' B' of the RPCCW system causing the train 'B' surge tank level to INCREAS WHICH of the following could be the potential source of leakage into train *B*? Excess letdown heat exchanger
- *B* RHR heat exchanger Letdown heat exchanger Seal Water heat exchanger ANSWER: Excess letdown heat exchanger REFERENCES: P&lD 14C-82020A, CC AOP 3555, RCS Leak, Attachment B JUSTIFICATION: Only the letdown heat exchanger, the excess letdown heat exchanger and the seal water heat exchanger have reactor coolant moving through 15em in MODE 3 for the stated cond,tions. The letdown heat exchanger is on Train *A*, The seal water return pressure is 30 - 50 psig, or less than RPCCW pressure. A leak into the *B* RPCCW surge tank can only come from excess letdown heat exchanger. The 'B' heat exchanger is isolated from the RC OBJECTIVE: CCP07C (e), CCP06C (e)
K/A: APE 026 A2.01, Exam 8
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RO3 Which of the following signals will cause the RPCCW to CDS cross-ties in containment (MOV226 229)to close? High RPCCW flow rate Containmentisolation Phase A Containmentisolation Phase B Low RPCCW surge tanklevel !
ANSWER:
D, Low RPCCW surge tank level REFERENCE: LSK g-10 JUSTIFICATION:
OBJECTIVE: CCP03C (b), CCP07C (e)
K/A: 026 EA 1.05 RPCCW Surge Tank, include level control, alarms in RM EXAM BANK ITEM: 2177
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R34 The following conditions exist:
. The plant is currently in operational mode * Reactor power is 4%.
. Preparations to increase power to 100% are in progres * The Pressurizer Pressure Channel Selector Switch is in the PT-455/PT-456 positio * The annunciator for HIGH PRESSURE ALARM, actuate * PORV PCV 450 open * Actual pressurizer pressure is 2200 psia and decreasing rapidl Which one of the following describes the instrument malfunction which caused this transient and the response of the PORV (PCV 450) PT-455 failed high and PORV will remain open in AUT PT-456 failed high and PORV will close in AUTO when actual pressure decreases below 2200 psia PT-456 failed high and PORV will remain open in AUTO PT 455 failed high and PORV will close in AUTO when aciual pressure decreases below 2200 psia ANSWER: PT 456 failed high and PORV will close in AUTO when actual pressure decreases below 2200 psia REFERENCES: OP 3353.MB4A(3-4); AOP 3571; PPL system text JUSTIFICATION: In the CHl/CHil position PORV456 is controlled by PT456. On high pressurizer pressure on channel 456 the PORV will open and will close when 2/4 pressurizer channel decrease to less than 2200 psia if the PORV is in AUTO. (B is correct)
A is incorrect because PT 456 failed and the PORV will close in AUT C is incorrect because PORV will close if in AUT D is incorrect because PT456 failed not PT45 OBJECTIVES: PPLO3D (a)
K/A: 027 A1.01 Ability to operate / monitor pressurizer heaters, spray, PORV Modified question 2118
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RO6 Given the following situation:
- A safety injection actuation has occurred due to a steam line break upstream of the
. MSI e The operators have completed the actions in E 0 and Isolated the affected SG in accordance with E 2 * Faulted CG loolation".
Which one of the following correctly describes the plant response, during the next hour, once the affected steam generator is empty? (Assume no further operator actions.) Pressurizer pressure, level, and RCS hot leg temperature in the affected loop decreas Pressurizer pressure, level, and RCS hot leg temperatures continue to decreas Pressurizer pressure and levelincrease, and RCS hot leg temperatures continue to decreas Pressurizer pressure, level, and RCS hot leg temperatures increas ANSWER: Pressurizer pressure, level, and RCS hot leg temperatures increas REFERENCE: MITCOR Text; FSAR Chapter 14, Increase in Heat Removal JUSTlFICATION: Once the faulted SG has blown dry, ECCS injection flow will cause an increase in pressurizer pressure and level. Decay heat will cause RCS hot leg temperatures to in::resse. As a result, *D" is the correct answe OBJECTIVE: E0203C K/A: 040K1.03, RCS Shrink /depressurization 3.8/ LOIT NRC Exam
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RO6 While performing E 3, " Steam Generator Tube Rupture", prior to initiating the RCS COOLDOWN, the Unit Supervisor directs you to determine if the ruptured steam generator pressure is greater than 420 psi If pressure is less than 420 psig, E 3 directs you to transition to ECA 3.1,"SGTR with Loss of Reactor Coolant Subcooled Recovery Desired".
. Which of the following describes a reason for ensuring the ruptured SG pressure is greater than 420 psig prior to performing the cooldown of the RCS? To ensure subsequent cooldown does not cause a PTS concer To ensure the ruptured SG pressure is greater than the intact SG pressure To ensure that a low steam line pressure Safety injection does not occu To ensure subsequent RCS cooldown does not cause a red path on Suberiticality due to a loss of shutdown margi ANSWER: To ensure subsequent cooldown does not cause a PTS concer REFERENCE: E-3, Step 13 Basis, WOG Background Document JUSTIFICATION: Although it is important that the ruptured SG pressure be greater than intact SG pressure, this verification is made at step 13 of E-3, and appropriate action taken if these conditions do not exist ("B"is incorrect).
Since intact SG pressure must be less than ruptured SG pressure to maintain RCS subcooling, the RCS cooldown that would be required to maintain the necessary pressure differential between the ruptured and intact SG's would present an PTS concern in the RCS ("A"is correct).
Distracter "C" cannot be correct because automatic Si actuation at this point in the recovery would have been defeated since Si is reset in Step 8 of E 3. "D"is incorrect because a Suberiticality red path only comes in for reactor power greater than 5% which could not occur based on the conditions provide OBJECTIVE: G300*C K/A: WEST E08 EK2.2, PTS relationship to Heat Removal Used 97 exam 8 (modified distracter)
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RO7 Given the following conditions:
. Reactor trip due to station blackout has occurre . Natural Circulation cooldown has been establishe . The crew is in ECA 0.0 dumping steam at the maximum rat The RO states that he is unsure if natural circulation is providing adequate heat remova Which one (1) of the following indications would indicate a potential problem with core heat removal by natural circulation. (Consider each condition independently). SG pressures have decreased from 700 psia to 600 psia and are now starting to decrease at a slower rat Pressurizer levelis 15% and decreasing while subcooling is 80'F and increasin T>or is 520'F and Tcoto is 460*F and the AT between the two is increasin Troy is 4g0'F and decreasing and CET's are decreasing slowly.
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ANSWER: Tsor is 520*F and Tcoto is 460*F and the AT between the two is increasin REFERENCE: ECA-0.0; MITCOR Text JUSTIFICATION: For adequate natural circulation the following should be occurring:
SG pressure stable or decreasing 4 - This condition is satisfied. It is expected that the depressurization rate slows as pressure gradient decreases. (A is not a problem).
The pressurizer is expected to empty under these conditions. The RCS is adequately subcooled. The contraction cannot be compensated for due to the loss of power. (B is not a problem),
The core AT is greater than full power AT which is indicative of degrading natural circulation and cooldown. AT should be stable or decreasing during the cooldown, not increasin Tnot and CET must be decreasing if a cooldown is in progres OBJECTIVE: A0003C K/A: 055 AK1.02 Natural Circulation Cooling New Question / Modified 2467
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RO8 The plant is at 100% power when the following alarm is received on Main Board 8A:
e Inverter 1 Trouble 1.ocally at Inverter 1, the' operator observes that ority the reverse transfer light is li Which of the following events is the likely cause of these alarms / indications? Inverter 1 is de-energize Inverter 1 DC feeder breaker has opene Inverter 1 is being supplied from it's DC suppl Inverter 1 has transferred to it's alternate suppl ANSWER: Inverter 1 has transferred to it's alternate suppl REFERENCE: OP 3353.MB8A,15 JUSTlFICATION: "A" is incorrect, if the reverse power is lit, the UPS must have transferred to its alternate suppl "C" is incorrect because the reverse transfer light would not light if Inverter 1 is being supplied by its backup DC power supply. Additionally, the backup DC would have to go through the same UPS as the normal AC suppl "B"is incorrect because if the cause was only the DC feeder breaker opening the UPS would still be supplied by its normal AC power supply,
"D" is correct because the local indication is that the inverter 1 is still powered otherwise there would be more alarms than just the reverse transfe OBJECTIVE: A64764D (1)
K/A Rating: APE 057 A2,06, inst. Bus Alarms for Alt Power 97 Exam 8
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R39 t Given the following conditions:
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The control room hao been evacuated due to a fir * All remote shutdown system line ups have been complete * RCP #2 is runnin . The RCS is borated to cold shutdown conditions e A plant cooldown is in progress Which of the following describes the PREFERRF.D sequence to depressurize the RCS to 1950 psia in accordance with EOP 3504 over the next 4 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> De-energize all pressurizer heaters, use one PORV, use aux spra Use normal spray, use aux spray, use one POR De-enorgize all pressurizer heaters, use aux spray, use one POR Use aux spray, use normal spray, de-energize all pressurizer heater ANSWER: De energize all pressurizor heaters, use aux spray, use one POR REFERENCE: EOP 3504 JUSTIFICATION: Normal spray is not available from the auxiliary shutdown panel, thus b and d are incorrec The preferred way to depressurize the RCS is to deenergize heaters and allow the RCS to depressurize slowly due to ambient losse If need to depressurize further, procedure guides them to place letdown in service and use aux spray. If letdown not available, then use one PORV. Thus c is correc A is incorrect sequence is reversed. Tested caution note in procedur OBJECTIVE: ASP 03 K/A: 068 A1.12 Aux Shutdown Panels / controls New Question
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RO 10
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Plant conditions are as follows:
- Reactor Trip & Sl have occurred
- No high head SI pumps or charging pumps are running e RCS pressure is 800 psia e Contalnment p,.asure is 20 psia .
. RVLMS 0% (plenum)
- CET's are 723'F and increasing The crew has properly transitioned to the appropriate functional recovery procedur Why is the operator directed to check if one RCP should be stopped? To minimize RCS inventory pumped or lost out of the brea To reduce the heat input into the RCS and help maintain RCS inventor C, To reserve one RCP for future use and aid in the plant recover To reduce the inventory lost when the seal packages fail in the running RCPs ANSWER: To reserve one RCP for future use and aid in the plant recover REFERENCE: EOP Basis Document JUSTlFICATION: EOP 35 F-02 Background document. One RCP is preserved to be able to be restarted in FR-C.1 if plant conditions degrade to an inadequate cooling situatio OBJECTIVE: FC203C K/A: 074K3.04 Tripping RCPs New Question l
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RO 11 Plant Conditions e The operators are performing ES 1.3, ' Transfer to Cold Leg Recirculatior..'
. The RO is aligning the RHR and RSS Systems for cold leg recirculation using the recir an'sy.
The STA informs the SM that a red path just came in on the Integrity CSF Status Tree.
Which of the following is the correct course of action? Transition to FR-P.1 immediately because vessel integrity is in jeopardy. The SM and BOP will perform FR P.1 concurrent with the US and RO completing E .3 Continue in ES 1.3 until directed to return to the procedure in effect, and then go to FR-P.1. Continue in ES 1.3 until the cold leg recirculation alignment is completed, and then transition to FR-P.1 ANSWER: Continue in ES 1.3 until the cold leg recirculation alignment is completed, and then transition to FR P.1 REFERENCE: ES-1.3, NOTE prior to Step 1.
JUSTlFICATION: Must complete steps 13 to ensure long term cooling by cold leg recirculation is established prior to transitioning to other FRPs.
OBJECTIVE: EOU (1733)
K/A: 011 K3.12, Actions Contained in EOPs Modified Exam Bank 2577
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RO 12 The LOCA outside containment procedure (ECA-1,2) is entered based on: Inadequata inventory of the containment sump to allow shiftover to cold leg recirculatio Si actuation with Indications that the RCS is intact, SG tubes are intact, and SG boundaries are intac Abnormal radiation level in the auxiliary building or ESF building with Si actuatio Abnormal radiation levels in the spent fuel building with Si actuatio ANSWER: Abnormal radiation level in the auxiliary building or ESF building with Si actuatio REFERENCES:
- JUSTIFICATION: Symptom which requires entry into ECA 1,2 from E-1/E "A"is incorrect there are symptoms of a Loss of Emergency Coolant Recirculatio *B" is incorrect they are possible indications for an inadvertent SI signal or extremely small brea "D"is incorrect because no safeguards release paths or components are located in the spent fuel buildin OBJECTIVES: A1201C K/A: W/E 04 EK 1.1 - 3.5 Components, Functions, and capacitie Modified question 2579
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RO 13 The following conditions exist:
. A steam generator tube rupture has occurre * The operators are presently at step 4 of E 3 * Steam Generator Tube Rupture * verifying ruptured SG level greater than 6% (42% adverse) leve Which one of the following describes the basis for maintaining the ruptured SG level greater than 6% (42% adverse) level after the SG is isolated? Provides filtering of the element iodine present in the ruptured steam generato Maintain a thermal stratification layer over the ruptured SG U tubes, allowing for depressurization of the RCS to the ruptured SG pressur Maintain an adequate heat sink available in the ruptured S Minimize streases on the SG U tube bends to prevent subsequent failures when the RCS is cooled down and depressurize ANSWER: Maintains a thermal stratification layer over the ruptured SG U tubes, allowing for depressurization of the RCS to the ruptured SG pressur REFERENCE:
JUSTIFICATION: WOG ERG background document for E 3 * Steam Generator Tube Rupture * states the reason for a minimum level in the SG is to maintain a thermal stratification layer over the top of the SG U-tubes. This will insulate the affected SG steam bubble and prevent the affec5d SG frcm depressurizing during the RCS cooldown. This makes matar.tlon 'B'
correc K/A: 037K3.07 Actions in EOPa OBJECTIVE: E03030 COMMENTS: SRO #77,95 NRC Exam (modified distracter D)
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RO 14 The plant is at 100% powe A loss of DC bus 5 occur Which of the following describes the expected plant response? Turbine Stop valves close due to low ETS pressure, the plant trips on Low-Low Steam Generator level or OTA ETS pressure switches in the Turbine Trip / Reactor Trip circuit deenergize causing a reactor tri C, ETS oil pressure is dumped causing the turbine to trip, which causes a reactor tri Turbine stop valves close due to low ETS pressure, causing a turbine trip on reverse
power, which causes a reactor tri ANSWER:
B, ETS pressure switches in the Turbine Trip / Reactor Trip circuit deenergize causing a reactor tri REFERENCE: AOP 3563, Attachment E JUSTIFICATION: ETS pressure la powered from 301C-1AO, from battery bus 5 OBJECTIVES: A6302C K/A: 058 A2.03 Impact on Indications 97 Exam 7(modified)
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R815 A loss of off site power has occurred.
While the Emergency Generator Loading Sequencer (EGLS)is in the process of completing the stepping sequence, an automatic safety injection occurs.
Which of the following describes what occurs upon initiation of the Safety injection? Sequencer resets to step 0, Emergency Diesels remain running, Emergency Diesel output breakers remain closed, all previously running loads remain running unless not required for the Si and the sequencers sequence the Si load , Emergency Diesel output breakers open to strip the bus, sequencer reset to step 0, 'Sl'
signal causes the Emergency Diesel output breakers to close and the sequencers to restart all the loads in the SI/ LOP stepping sequence. Sequencers stop at the step in progress, Emergency Diesels remain running, Emergency Diesel output breakers remain closed, the EGLS' start any Sl loads that should be running and then resumes sequencing at the steps in progress in the SI/ LOP mode. Emergency Diesel output breakers open to strip the bus, Emergency Diesels remain running but the breaker will not close for at least 6.8 seconds. The sequencers reset to step 0 in the SI/ LOP mode and will restart the entire sequence when 6.8 seconds have elapsed and the 'second* SI/ LOP is sensed by the sequencers.
ANSWER: Sequencer resets to step 0, Emergency Diesels remain running, Emergency Diesel output breakers remain closed, all previously running loads remain running unless not required for the Si and the sequencers sequence the Si loads.
REFERENCES: LSK 24 9A JUSTIFICATION: 'B' & 'D' are incorrect because the diesel output breaker does not ope 'C' is incorrect because the sequencer resets to time 0. "A" is correct, the sequencer resets, any loads not needed for the m ode, such as RPCCW pumps for CDA would be tripped, any Si loads would be starte OBJECTIVES: EDSOSD (f)
K/A: APE 056 A2.49 Component, Capacity, Functions 97 Exam 8
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RO 16 The control room is responding to an overpressure condition in *B' SG. SG pressure is currently 1250 psig and no safety valves or atrnospheric relief valves are open, in order to depressurize the SG as fast as possible, in accordance with FR H.2, the operators should: Maximize blowdown flow from *B" S Feed *B" SG to cooldown the SG and lower the pressur Dump steam from the 'B' SG to lower the pressur Dump steam from the other SG's to cool down the RCS to depressurize the 'B' S ANSWER: Dump stearn from the "B' SG to lower the pressur REFERENCE:
JUSTIFICATION: The quickest way to remove heat from the SG is to dump steam. (C is correct)
A cannot remove much heat. Increasing blowdown is a possible action for a high SG level not pressur Feed will cool the SG but procedure caution against feeding SG until steam heat removal path is available. (B is incorrect).
D cannot remove as much heat as C. D is the second alternative in accordance with FR-H.2 to depressurize the S OBJECTIVE:
K/A: W/E 13 K Components, capacity and function New Question i
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l RO 17 Given the following conditions:
- MP3 is operating at 80% powe .
- Rod controlis in automatic, e Rods are at 225 on Bank * The controlling Iraf signal fails to a volun of $75' Control rod speed will indicate a speed of steps per minute, and the indication will be li , out : 72. out , in , in
- ANSWER: , in REFERENCE:
JUSTIFICATION: Tav (80%) = 557 + 24 = 581 Tref = 575 Tav-Tref = 6'F
. Max rod speed at 72 spm (d is correct)
A is incorrect - wrong speed / direction -
B is incorrect -wrong direction C is incorrect wrong speed OBJECTIVE: A5202C; ROD 02C (f.6)
K/A:- 001 K6.02 Sensor feed to RCS Modified 382
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RO 18 The analysis for Reactor Coolant Pump locked rotor loss of flow accident predicts the initial trend over the first few seconds is: Both DNBR and Pressurizer pressure increase. Both DNBR and Pressurizer prossure decrease. DNBR increases, Pressurizer pressure decreases. DNBR decreases, Pressurizer pressure increases.
ANSWER: DNBR decreases, Pressurizer pressure increase REFERENCE: FSAR Chapter 14, Decrease in Reactor Coolant System Flowrate (MCORE04)
JUSTIFICATION: Loss of flow causes the DNBR to DECREASE (closer to DNB), the loss of flow also causes the RCS to heat up as less heat is removed, and pressure will increase (D" is correct).
OBJECTIVE: RCS08C; MCORE04 K/A: 003 K 5.01 - RCS Flow / Core & RCS pressure 97 EOP Exam Question 2609 l
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Which one of the following is the reason that RCP #1 seallenkoff is isolated at RCS pressures below 125 psla? Leskoff flow decreases at low pressures so #1 sealleskoff is isolated to force more flow through #2 sea Backflow from the VCT through the seal leakoff line could flush contaminants into the seal Controlled leakage limits may be exceeded due to excessive seal in}ection flow at low pressure Leakoff flow instruments are not accurate at low pressures and excessive leakoff could go undetected ANSWER: Backflow from the VCT through the seal leakoff line could flush contaminants into the seal REFERENCE: Reactor Coolant Pump Text JUSTIFICATION: The VCT may be at highor pressure than the RCS and backflow through the seal leskoff line may result. This flow is not subject to filtration (no seal injection filters on this line) so VCT contaminants may be introduced into the seals and result in mechanical damage when the pump is ret, tarte K/A: 003 K1.03, RCP Seals 3.3/ OBJECTIVE: RCP06C g5 NRC exam
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RO 20 What is the purpose of the interlock requiring the Letdown isolations (3CHS*LCV459/460) to be open prior to opening the orifice isolations (3CHS*AV8149A-C)? To prevent flashing downstream of the Letdown pressure control valve (3CHS-PCV131).
- To prevent damage to the Letdown Regenerative Heat Exchanger shell due to r%sure
- spike To prevent excessive Letdown Heat Exchanger temperatures overloading the RPCCW syste To prevent the letdown relief valve from lifting due to the heatup of the water between 3CHS*LCV459/460 and 3CHS*AV8149A- ANSWElk: To prevent damage to the Leidown Regenerative Heat Exchanger shell due to pressure .
spike REFERENCE: NSSS TEXT, VOL 1; JPM28 JUSTIFICATION: The interlock which prevents opening or closing the letdown isolation valves unless the orifice isolation _ valves are closed prevents damage to the shell side of the regenerative heat exchanger due to pressure spike If LCV459 and LCV460 were opened with the orifice isolation valves
- open, the hot letdown flo v would depressurize to VCT pressure as LCV4e and LCV460 left their closed seats and acted like throttle valves, and the water would flash to steam, This would cause a pressure spik If LCV459 and LCV460 closed with the orifice isolations open, the hot letdown water between the letdown isolation valves and the regenerative heat exchanger would again depressurize and flas OBJECTIVES: CHS04C (a1.a2)
j K/A: 004K4.15 - Intertocks with letdown orifice isolation valves
,; Modified Bank 818
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Which of the following is a reason for placing the control switches for the. GN ; ng pumps in pull-to-lock during a Loss of all AC Power? To prevent the injection of cold seal injection water into the RCP seal packages when an emergency bus is restored, possibly damaging the seal packages or bowing the RCP pump shaf To prevent PTS by thermal shocking the reactor vessel downcomer with cold, high pressure Safety injection water when an emergency bus is restored after the Gea are d3 pressurize To prevent thermal stressing the cl'arping loop penetrations by injecting cold, safety injection water into the RCS possib'/ t ueating an unisolable LOC To prevent a pressurizer overfill situation when power is restored, since FCV-121 will be full open and the pressurizer will be empty due to the cooldown when the SGs were depressurize ANSWER: To prevent the injection of cold seal injection water into the RCP seal packages when an emergency bus is restored, possibly damaging the seal packages or bowing the RCP pump shaft, r
REFERENCE: ECA-0.0 Step 6 basis JUSTIFICATION: Defeating automatic loading of Charging /SI pumps functions to protect RCP from d/ mage when AC power is restored. This action prevents the automatic delivery of relatively cold seal injection flow into the RCP number 1 seal chamber and shaft are which has the potential for thermal shock and subsequent damage to the RCP seals and shaf OBJECTIVES: CHS06C (c.1); A0003C K/A: 004 K2.02, Chargirsg Pump Power Supplies Test item 690 modified i
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RO 22 The plant is opersting normally at 100% power.
Which of the following would be a result of the Letdown Pressure Transmitter (3CHS*PT131)
falling LOW? The Regenerative Heat Exchanger Outlet Temperature increases above its original value. The Letdown Heat Exchanger Outlet temperature increases above its original value. The CDTT levelincrease The PRT levelincreases.
ANSWER: The PRT levelincrease REFERENCE: P&lD 104A,102F; CVCS Text JUSTIFICATION: Indicated pressure decrease = PCV131 close = actual pressure increases = relief lifts to the PR A is incorrect. When PT 131 falls low, letdown pressure control valve PCV*131 will close. Regenerative heat exchanger outlet temperature will decrease because more heat is being removed by the charging flo When the letdown relief opens, the temperature will increase slightly, but it will not increase above the original valu B is incorrect. Letdown flow will decrease. The initial heat exchanger outlet temperature will decrease and the RPCCW valve will clos C is incorrect. The letdown relief lifts to the PRT r.;f the CDT OBJECTIVE: CHSO4C (a.6, a.7)
K/A: 004 A4.05, Letdown Pressure / Temp. Control Valves NSSS Exam ID 2213 l
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RO 23 PLANT CONDITIONS:
. . Plant is at 100% power e Pressuriier pressure channel, PT-455, failed high three (3) hours ago.
. The operating crew carried out the actions of AOP 3571, instrument Failure Response and removed the channel from service.
. Channel PT457 is now the controlling channel.
. All systems have been retumed to automatic control A loss of 120 VAC vital instrument panel VIAC 3 has just occurred.
Which of the following describes the impact on the plant of the loss of VIAC 3? The plant will remain at 100% power. However the automatic actuation of the PORVs has been lost. A safety injection will have occurred due to low pressurizer pressure logic coincidence being met. One PORV will open due to a high pressure signal and this will eventually lead to a safety inbetion on low pressurizer pressure. The master pressure controller will cause the pressurizer control heaters to go to minimum output and close the spray valves.
ANSWER: A safety injection will have occurred due to low pressurizer pressurtz logic coincidence being met.
REFERENCES: AOP 3571, Functional sheets 12 and 12a JUSTIFICATION: The operators will have tripped the bistables associated with the failure of the pressurizer pressure channel. This will include the low passsure Si bistable. When the vital AC bus is lost a second low pressurizer pressure bistable will be received and an SI will occur. The OBJECTIVES: RPS07C K/A: 013 A2.04 Loss of instrument bus 97 Exam 8
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RO 24 A Safety injection Signal occurs while VIAC 2 is deenergize How will the Train 'B' ESF equipment respond? Diesel Generator *B' starts, Bus 34D is stripped and Train B ESF loads are immediately loaded onto the bu Diesel Generator *B" starts, Bus 34D is not stripped, and the Train B ESF loads do not star Diesel Generator 'B' does not start, Bus 34D is not stripped, and the Train B ESF loads do not star Diesel Generator *B' does not start, Bus 34D is stripped, and the Train B ESF loads are immediately loaded onto the bu ANSWER: Diesel Generator "B" does not start, Bus 34D is not stripped, and the Train B ESF loads do not star REFERENCE: AOP 3564, Caution prior to Step 6 JUSTIFICATION: The caution prior to Step 6 of AOP 3564 states that:
" Loss of VIAC 1(2) results in diesel generator A(B) being de-energized, if an ESF actuation takes place during this condition, the following items will not occur automaticall . The A(B) diesel will not start (except LOP)
. Loads will not be stripped from the emergency busses
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A(B) train loads will not start" Distracter "C" is therefore correct. Distracters *,Y & *B" are incorrect because they state the "B" EDG will start even though the questien does not assume an LOP is present. Distracter "D" is incorrect because loads will not be stripped from Bus 34 OBJECTIVES: A64764D (1); A64764D (3)
K/A: 013 A3.02 Actuation of safeguards equipment 97 Mitcore Exam 2370
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l RO 25
- Given the following conditions:
. The crew !s performing a reactor startu * The RO has just pulled the control rods several steps and is waiting for source range counts to stabiliz Assuming the reactor is very close, but not yet critical, source range counts should: Stop increasing and stabilize immediately, with SUR decreasing to zer Increase at a constant rate with a constant, positive SU Continue to increase, but at a slower rate, with SUR stabilizing at a lower positive valu Continue to increase for a short period of time, then plateau, with SUR decreasing to zer ANSWER: Continue to increase for a short pedod of time, then plateau, with SUR decreasing to zer REFERENCE: Reactor startup procedure (op 3202) and fundamentals +
JUSTlFICATION: A is incorrect. Doesn1 consider the effects of suberitical multiplication and its impact as criticality is approache B is incorrect. The reactor is still subcritica C is incorrect because SUR will decrease to zero, and count will stabilize
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at higher volum D is correc OBJECTIVE: 60202C K/A: 015 K5.05, Criticality and its indications New Question -
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(L RO 26 The plant is operating at 95% power, The Balance of Plant (BOP) operator inadvertently
.: partially opens the Low Pressure Foodwater Hester Bypass Valve (CNM-MOV88).
How will opening the valve affect plant cycle efficiency?
A, Decrease since acre energy from the reactor is required due to feedwater entering the -
steam generator at a lower temperature.= Increase due to less work required to pump the water through the low pressure
- feedwater heater Decrease because mass flowrote entering the steam generators has increase D; Ind ase due to an increase in feedwater temperature resulting in more efficient high-pressure feedwater heater performanc ' ANSWER: Decrease since more energy from the reactor is required due to feedwater entering the steam generator at a lower temperatur REFERENCE: Westinghouse heat transfer fundamentals chapter 7, Mollief diagra JUSTIFICATION:: Shutting the bypass valve will result in an increase in feedwater-temperature due to mom flow thru the feedwater heaters to be preheated -
an no " cold" feedwater bypassing the heater strings. As_ a result, .
feedwater will enter the steam generators at a higher temperature. The -
enthalpy change of the working fluid across the steam generator is smaller therefore, the heat transferred into the system (Q.) is smaller, A smaller heat addition results in increased cycle efficienc OBJECTIVES: FWH06C; FWH07C K/A: ' 056 A212, Heater String Bypass SOURCE: Modified exam item 127'
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I RO 27-
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0900 A turtine trip / reactor trip occurs 0905 A loss of offsite power occurs A loss of all main feedwater occurs Aux %ry feedwater initiates with approximately 1200 GPM total flow 0915 Narrow range level in the 'C' SG is 6% and increasing 0918 Auxiliary feedwater total flow continues at approximately 1200 GPM -
What are the consequences of not throttling back on AFW flow to all of the SGs at this time?. SG levels will increase until SG overfill protection isolates AFW flow to all SG Full AFW flow will cause the RCS to cooldown and depressurize. Letdown will isolate and an SI will occu Full AFW flow will cause a plant cooldown unul low-low Tav is reached and the feedwater isolation signal stops AFW flow to all SG The SGs will cooldown and depressurize. This will result in a safety injection on the high steam pressure rat ANSWER: Full AFW flow will cause the RCS to cooldown and depressurize. Letdown will isolate and an SI will occu I-JUSTIFICATION: A is incorrect because the high-high signal trips the turf 'ne, which ids already tripped and provides no isolation of AF C is incorrect because a feedwater isolation signal has no affect on AFW flo D is incorrect because the high steam pressure rate will result in a main steamline isolation, not aa C OBJECTIVE: EOOO3C K/A: 059 K3.04, Effects on RCS Temperature New Question
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Given the following conditions:
. _ MP3 was operating at 100% powerc
.- 1The station experienced a totalloss of all AC power. .
e; The crew is currently depressurizing the SGs to 260 psig in ECA- As SG pressures are lowered, available AFW flow will'
A; Decrease to zero (0) gallons total flow Drop to 100 GPM per steam generator C.- Remain above a total of 530 GPM L Remain near the original flow of approximately 1150 GPM
' ANSWER:
- C.- Remain above 530 GPM s
REFERENCE:
JUSTIFICATION: - The Terry Driven AFP is sized to remove the decay heat and cooldown j the plant to go on RHR at 250 psig in RCS (SG Press -100 psig) l The Terry Turbine will provide adequate heat sink for reactor and be above red path value. (C is correct)
A & B are incorrect. B is the value of flow due to the flow restricting venturi for feedline break protection (incorrect). A is from number for SG pressure 150 when RHR is put in service, (A is inconect).-
D is incorrect because flow decreases when pressure drops especially when SG pressure drops < 600 psig (FSAR).
OBJECTIVES: FWA03D (6)
- K/A: . 059 A1.02 Steam generator pressure effects on terry turbine 3.6 ~
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RO 29 Given the following conditions:
.: Unit is operating at 100% powe . - A Hi alarm is received on the containment gaseous radiation monito * Charging flow is 135 GP * Seal retum flow is 3.2,2.8,3.3,3.4 GPM (A, B, C, D)
. Seal injection flow is 9.8, 9.0, 9.2, 9.0 GPM (A, B, C, D)
. Letdown flow is 75 GPM,
. - RCS Tave is on progra * Containment sump level is increasin .- PRZR levelis on progra . VCT levelis 30% and decreasin . PDTT levelis stable at 25%.
- Based on the above information, the current leak rate is GP A.- 4 .0-C, 8 .0 ANSWER:
C, 8 '
REFERENCE:
JUSTIFICATION: Leak Rate = Charging + Seal Injection -(Letdown + Seal Retum)
Leak Rate = 135 + 37 - (75 + 12.7)
Leak Rate = 172 - 8 Leak Rate = 8 . OBJECTIVES: .A5502 K/A: - 002 A2.01 RCS inventory balance / loss of inventory Modified Exam item 659
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_ RO 30 INITIAL CONDITIONS;
. Reactor power is 60%
. - Loop 1 Delta-T indicates LOW e Loop 1 Tave indicates HIGH Which of the following Loop 1 RTD failures could cause these indications? T-hot failed HIG B.- T-cold failed LO , T-cold failed HIG T-hot failed LOW.-
' ANSWER: T-cold failed HIG REFERENCE:
JUSTIFICATION: Tave = TH Tc AT = TH - IC If Tc fails high, Tav indicates high AT indicates low, (c is correct)
A is incorrect because AT would indicate hig D is incorrect because AT would indicate hig D is incorrect Tav could indicate lo OBJECTIVES: RCS02C (1)
K/A:- 002K 5.12 Relation of temperature indications Modified item 146 . .. .
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- RO 31l Given the following conditions:
A small break LOCA has occurre Fuel failure occuned during rapid downpower prior to trip. Crew is currently in ES-1.2 " Post-
- LOCA Cooloown and Depressurization.'
- RCS pressure is 100 psia and decreasing Wide range Tc's are 550'F and decreasing Wde range Th's are 560'F and decreasing- --
- CET's are 565'F and appear stabl <
Containment pressure is 20 psia and decreasin Containment temperature is 185'F and decreasing.-
' SG levels are being maintained at 35%.;-
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' Pressurizer levelis 25% and stable i Baeed upon the above indications, Si flow: ; Can be reduced because CET's are stable.~ Cannot be reduced since subcooling is inadequat Can be reduced because the SG's are an adequate heat sin Cannot be reduced since containment pressure is still greater than the Si actuation setpoin ANSWER:
- Cannot be reduced since subcooling is inadequate.-
JUSTIFICATION: B is correct. Subcooling is less than 115'F. Thus Si flow cannot be reduce A is incorrect because Si reduction criteria is a function of subcooling and pressurizer level _ not CET's alon C is incorrect. SG's levels aren't greater than those required for heat removal under adverse containment condition D is incorrect.- Si flow can be reduced under circumstances when containment pressure has not been restored to less than the actuation -
setpoint (18 psla)
1 OBJECTIVES: S1202C; S1203C K/A: 006 A1.19' Effects of subcooling New Question
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RO 32 The plant is in MODE 5 with Tave approximately 185'F. The crew is in the process of drawing a bubble in the pressurizer. Pressurizer level has just started to come on scal A complete loss of instrument air occurs.
.Which of the following describes plant response with NO operator action? khe plant slowly depressurize RCS pressure willincrease to the COPS setpoin RCS pressure will increase to safety valve setpoin The plant will maintain pressure until pressurizer heaters tri ANSWER: RCS pressure will increase to the COPS setpoin REFERENCE: OP 3201 Attachment 1 JUSTIFICATION: Letdown isolates, spray valves fail closed. COPS is armed. Setpoint for COPS at the current temperature is = 400/450 psia. Pressure will increase until COPS open OBJECTIVES: RCS08C K/A: 010 K4.03 Overpressure Control 97 Exam 9 2184
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RO 33 t Given the following conditions:
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. MP3 is holding at 9% to complete turbine shell warmin . PZR pressure channel ll bistables are in trip because channel previously drifted hig . SG level is being controlled with the FRV bypass valves, which are in automati . FRVs are isolate . RCS temperature is being controlled by the steam dump WHICH ONE cf the following events would, if no corrective actions were taken, result in a reactor trip?
A Running charging pump trip A second PZR pressure channel fails to 1900 psi Loss of power to the FRV bypass solenoid EHC piping ruptur ANSWER: Loss of power to the FRV bypass solenoid REFERENCE: Tech Specs, Functional sheets 6 & 7 JUSTIFICATION: The loss of power to the FRV bypass solenoids will result in them failing closed. This will result in a reactor trip on SG LO-LO level, a trip that is not bypassed by P- Power < P-7 (10%) blocks or bypasses the following trip:
Turbine trip - Rx trip bypassed by P-9 (D is incorrect)
Pressurizer low press Rx trip (B is incorrect)
High Pressurizer level Running charging pump trip will cause auto start of standby pump and will not cause a trip. RPS trip is off high pressurizer level (A is incorrect)
An Si signal, causing a reactor trip, will not be generated because the setpoint is 1892 psi OBJECTIVES: FWSO4C (p)
K/A: 012 'R1 Perr6.ssive circuits New Question
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RO 34 Pressurizer pressure instrument PT-455 has failed high. The crew has tripped the following bistables:
liigh pressurizer pressure reactor trip
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Low Pressurizer Pressure
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Pressurizer pressure low pressure SI
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Low pressurizer pressure PORV logic
\ht additional bistables, if any, must still be tripped to complete AOP 3571, Instrument Failure
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Response? No additional bistables must be tripped, all the bistables listed in AOP 3571 for the failed channel have been trippe OTAT C-3 OTAT OPAT OPAT C-3 ANSWER: OTAT C-3 REFERENCE: AOP3571, attachment B JUSTIFICATION: There are 7 bistables listed in the attachement, the crew has tripped The remaining 2 are OTAT and C-3, (Low Pressure Reliefinterlock)
K/A: 012 A3.05, RPS Channel Trips OBJECTIVE: PPLO8C
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RO 35 Which of the following situations could cause a General Waming on the DRPl display? One central control card differs from the other tw One dropped rod at 100% power with normal rod alignment prior to the rod droppin Either Data "A" or Data *B" card inputs are invali Two rods within the same bank deviate from one another by more than 12 step ANSWER: Either Data "A" or Data "B" card inputs are invali REFERENCES: DRPI Technical Manual; AOP 3552; ARP MB4C JUSTIFICATION: A is incorrect. A central control card failure will cause the control card failure alarm on the DRPI display and an RPI Non Urgent Alarm B is incorrect. A dropped rod will result in "ONE ROD BOTTOM *
annunciator, " ROD DEVIATION" alarm. and a rod bottom light on DRP D is incorrect. For this situation a " ROD DEVIATION * alarm and " ROD POSITION DEVIATION" alarm would be generate C is correct. The general waming DRPI display LEDs are illuminated for Data "A" or Data *B' input being rejected or any cause for an DRPI urgent failure alarm which are:
- Loss of both A and B data or errorin BOTH; or
- > 1 bit difference in A and B gray codes: or
- Combined data sum is > 38 OBJECTIVES: RPl07C K/A: 014 A1.02, Control Board Indications Exam item: 48
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I RO 36 What will be the effect on Feed Regulating Valve 3FWS*FCV510 due to the controlling channel for 'A' Steam Generator level instantaneously failing high? The feed regosting valve will start to close after a predetermined time due to the built in lag circuit ofihe controller, The feed regulating valve will immediately start to close and then close at a speed determined by the magnitude of the error signal and the time the error signal is presen The feed regulating valve will immediately start to close, then change its closing speed based on the time the error signal is present onl The feed regulating valve will immediately start closing at a constant rate for a predetermined time, then begin opening at a rate determined by the actual level erro ANSWER: The feed regulating valve will immediately start to close and then close at a speed determined by the magnitude of the error signal and the time the error signal is presen REFERENCE: Functional drawing 108D685, instrumentational and Operational Analysis, Nuclear Energy TRNG. Module 7 JUSTIFICATION: The valve will close immediately without any lag. (A is incorrect)
The valve will close immediately at a rapid rate then close at rate dependent on the integral time constant. (C is incorrect).
Because the proportional and integral signals are additive, the rate is never constant and the valve is always closing due to the system being level dominant. (D is incorrect).
OBJECTIVES: FWS07C (a1)
K/A: 016 A3.01 Auto select / signals to control currentl _ _ _ _ _ _ _ _ - - _
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RO 3 Following a CDA signal, what is the purpose of the lime delay associated with the start of the Recirculation Spray System (RSS) Pumps?.
- To allow the operators time to override the starting of the RSS pumps, on art inadvertent -
CDA signal, preventing an undesired spray down of Containmen To allow time for the RHR pumps to trip off on low-low RWST level and prevent exceeding the heat removal capabilities of the service water syste To be consistent with the sequencer time delays for a CDA/ LOP situatio D. . To allow time for proper NPSH to the RSS pumps to be establishe ANSWER:
D. - To allow time for proper NPSH to the RSS pumps to be establishe REFERENCE: FSAR Section 6.2. JUSTIFICATION: The 11 minutes allows CTMT sump level to increase with QSS water, and condensed RCS or MSS steam, preventing RSS pump starts prior to achieving and adequate not positive suction sourc OBJECTIVES: CDA06C (g 4); CDA03C (e)
-K/A: 026 K1.01 Relationship with ECCS Exam item 2473 i
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L RO 38
- Given the following conditions:
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- L The Unit has just synched on line and is ramping past 15% powe . "B" reactor coolant pump trip Assuming no operator actions, "B" steam flow will and "B" SG level will
' (consider only the immediate effects). Increase, increase B, increase, decrease -- Decrease, increase Decrease, decrease ANSWER: Decrease, decrease REFERENCE:
JUSTIFICATION: Steam flow will decrease as pressure drops and level will decrease, due
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to colder RCS when flow reverde OBJECTIVES: RCS08C-K/A: 035 K5.03 Shrink and Swell Modified Exam item 1818
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RO 39 The plant is at full power in a normal electric plant lineu The NSST breaker for bus 34A open Which of the following would result in a slow transfer rather than a fast transfer? The 34A 34C bus tie breaker opens at the same time as the NSST breake The NSST *early actuation * signalis presen An RSST output voltage of 3600 volt Bus 3dA normal supply lockout relay failed to actuat ANSWElt The 34A-34C bus tie breaker opens at the same time as the NSST breake REFERENCE: LSK 24-3D JUSTlFICATION: A is correct. The tie breaker being closed is a condition for fast transfer and the breaker being open is a condition for slow transfe B is incorrect. The early actuation signal is a requirement for fast transfe O is incorrect because low voltage on the RSST would prevent either transfe D is incorrect because the supply lockout relay will impact both transfer scheme '
OBJECTI~. " ' 4KV07C; 4KV06C K/A: # K4.03 Interlocks, auto transfer Modified bank item itm9
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l- RO 40 t -
In accordance with the Millstone 3 FSAR, which one of the followin'g describes the expected
= battery capacities for a loss of DC on battery bus (301 A-1/301B-1)? - hou . hours Cc 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hours ANSWER: hours REFERENCES: Technical Specification 3/4.8.2 and FSAR JUSTlFICATION: The battery capacity is assumed to be 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. (TS 3/4.8.2) "B"is correc '
OBJECTIVES: 12501C; 12502C K/Ai 063 K4.03, Battery Charger Capacities New Question
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RO 41-The crew is in E-0, " Reactor Trip or Safety injection", due to Si actuation Prior to resetting the SI signal the operator depresses both Emergency STOP push-buttons on MB8 for the *B' Emergency Diesel Generato Which of the following describes the response to the "B" Emergency diesel Generator?
. (No further operator action occurs). The diesel continues to ru The diesel would stop and remain shutdow The diesel would stop but automatically restart after 140 seconds, and restart EGLS in the mode SI/ LO The diesel would stop and then immediately restart and restart EGLS in the mode
- Sl/ LO ANSWER: The diesel would stop and remain shutdow REFERENCE: LSK 24-09.3A,B and J JUSTIFICATION: The emergency stop push-buttons energizes the shutdown relay which energizes the solenoid to cut fuel to the diesel, preventing a start. The -
' diesel will remain shutdown. . (B correct)
- OBJECTIVES: EDG06C; EDG04C K/A: 064 K3.01, impact of Auto Loading Modified from 443
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RO 42-The plant is in:
Mode 3
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- RCS is at 150*F and 340 psia e SG temperatures are 200*F after natural circulation Which of the following will cause an RCS overpressure transient in the RCS? Loop A Ts falls lo Bl- RCP B is restarted, C. ' Letdown Pressure Control valve fails open.- Starting an RHR pum ANSWER: RCP B is restarted, REFERENCE: 4 JUSTIFICATION: ' "A" is incorrect because COPS valves are armed and won't open unless RCS pressure exceed setpoint. This failure will not cause any transient in RC "B" is correct - starting "B" RCP is solid plant with COPS armed and a 50AT will ceuse a pressure increase and the COPS valves will ope "C" is incorrect - letdown pressure control valve failing open will cause
- RCS pressure to decreas "D"is incorrect - starting the RHR pump should cause the RCS pressure to decrease slightly due to increase flow and hoat removal in the RHR heat exchanger. PCV 113 will reposition in auto to maintain RCS -
pressur ' OBJECTIVES: RHR05C; RHR08C 1 K/A: 005 A2.02, Cold Overpressere Protection
.New Question
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RO 43 Maximum allowed RHR flow and the bases for this maximum during reduced inventory conditions are: RHR flow limited to 3000 GPM to prevent exceeding the heat removal capabilities of the RPCCW syste RHR flow limited to 1000 GPM to prevent loss of flow due to vortex entrainment of air at pump suctio RHR flow limited to 1000 GPM to minimize potential inventory loss in the event of a leak RHR flow limited to 3000 GPM to prevent overloading the pump motor from operating near runout condition ANSWER: RHR flow limitad to 1000 GPM to prevent loss of flow due to vortex entrainment of air at pump suctio REFERENCES:
JUSTIFICATION: "B" is correct. EOP 3505 limits flow to 1000 GPM to minimize effects of vortexin "D" is incorrect. RHR flow is released from 3000 GPM to 1000 GPM as reduced inventory procedures are implemente "B" is incorr60t because pressure limits flow from leaks not RHR flowrat "A"is incorrect because flow is limited to 1000 GPM and the RPCCW system is cabable of removing heat at higher then 3000 GPM flow rate OBJECTIVES: RHR05C .
KIA: - 005 A1.02 RHR flow rate New Question
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RO 44 The plant has tripped and experienced a safety injection and ' CIA'.
Which of the following describes how the reactor plant chilled water system is affected? Chilled water containment isolation valves close, the RPCCW to CDS cross-ties open, and all other loads outside containment isolate. All loads outside containment isolate, the chilled water loads in containment are mairstained unless a "CIB' actuates. Chilled water containment isolation valves isolate, the RPCCW to CDS cross-ties open, flow to all other loads is maintained. All loads outside containment isolate, the chilled water loads in containment are maintained unless a "CDA' actuates.
ANSWER: Chilled water containment isolation valves isolate, RPCCW to CDS cross-ties open, flow to all other loads is maintained.
REFERENCE: P&lD 122B and 121B JUSTIFICATION: CIA closes the containment valves and opens the RPCCW cross ties (C Correct)
CDA will not directly cause any valves to reposition. (D incorrect)
CIB will cause the RPCCW supplies to CDS to close. (B Incorrect)
CIA will not cause a loss of all loads outside containment. (A Incorrect)
OBJECTIVES: CDS03C; CDS06C K/A: 008 A3.05, Control / auto isolation valves Bank item 1504
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~RO 45.-
- The following plant conditions exist:
- A plant cooldown from 557'F has been starte ei _ The steam dumps are being utilized in the steam pressure mod *- ' At 553'F all the steam dumps clos Select the action which will allow continuation of the cooldown on steam dump A. - Take the " mode selector" switch to " reset".
. Take bypass interlock selector switches to " bypass". Take bypass interiock selector switches to "off/ reset" positio Take the steam pressure controller to manual to open the steam dump valve ANSWER: Take bypass interlock selector switches to " bypass".
-JUSTIFICATION: To bypass low low Tav interlock you must take both bypass interlock -
switches to bypass. ("B"is correct.)
- Taking the mode switch to reset - resets the load rejection arming memory (C-7)'only. ("A"is incorrect.)
Taking the bypass interlock select switches to off/ reset position will block steam. dump operation. ("C" is incorrect.) - .
Shifting steam pressure controller to manual doesn't ovemde/ bypass the -
block signal ("D"is incorrect.)-
OBJECTIVES: SDS05C (c); SDS02C (g)
K/A: 041 A4.02, Use of cooldown valves -
Modified from 353 (Confidence Weighted Question)
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. Plant Conditions:.
- A & B SW pumps selected as lead pumps C & D SW pumps selected as' follow pumps The following plant conditions exist
. Reactor trip has occurred due to a loss of off-site powe . - The "A" EDG starts and immediately closes in to supply bus 34C
. = The "B' EDG starts but its output breaker does not close automatically. The operator closes the output breaker after the diesel has been running for 30 sec On a board walkdown the expected following service water pump combinations should be
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A.- All 4 service water pumps should be runnin The A and D SW pumps should be runnin . C, - The C and D SW pumps should be runnin ! ' The A and B pumps should be running off their respective emergency diesel _
e ' ANSWER: . The A and B pumps should be running off their respective emergency diesel REFERENCES:
JUSTIFICATION: The lead pumps should be running. The follow pumps only start if a lead pump fails to start. (A, B, & C are incorrect.)
OBJECTIVES: SWPO4C(a); SVlMC(d) ,
a K/A' 076 K4.02, Aato Stari Features
- Modified Exam item 1997
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RO 47 Initial Conditions:
. Unit is at 100% power
. All circulating water pum[.s are running
. It is August 20th and Long Island Sound average temperature is 71'F
= All heater dra3n pumps are running
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. A & C TPCCW pumps are running
. B Stator liquid cooling pump is running A * BUS 34A BUS DIFF' alarm occurs on main Board 8. The operators should: Go to E-0 because the unit will trip on OTAT due to the MSIVs closing due to low air pressur Reduce load in an attempt to maintain vacuum in the condense Trip the reactor and go to E-0 because the main feedwater pumps tripped on low suction pressur Stabihe the unit after the turbine runback due to high stator water cooling temperature ANSWER: Reduce load in an attempt to maintain vacuum in the condense JUSTIFICATION: A is incorrect. ihe MSIVs would close on loss of DC power to the solencio%cr tha loss of instrument air. DC power is not affected by the given AC power loss. The "B" TPCCW pump will auto start on low TPCCW pressure. Even if TPCCW was lost the air compressors would shift to the dumestic water system as a backu C is incorrect because even though heater drain flow will be reduced and main feed pump suction will start to decrease, the condensate pumps will make up for the lost fkTw, As feed flow decreases, the feed regulating valves open, causing feed pressure to decrease which in tum causes main feed pump speed b increase. The unit will not trip on loss of the main feed pumps or loss of suction pressur D is incorrect because the stator water cooling pump has not been lost (B pump not powered by 24A). Additionally the unit can operate for up to one hour without statot water tooling or cooling to the stator cooling heat exchangers by TPCCW. The "B" TPCCW pump will start in auto on low TPCCW pressur B is correct. With only one circulating water pump running for each bay of the condenset, vacuum will start to decrease if the plant remains at 100% powe OBJECTIVE KV06C; 4KV07C K/A: 062 K2.01 Loss of power to major loads New Question
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RC <8 The plant is in MODE 6. Core ollload is in progres .
The alann circuitry for one of the Spent Fuel Pool area monitors fails.1 & C is investigatin No other operator actions have been take Whic 3 of the fo'. lowing describes the required ACTION,if any, to be taken regarding the fuel movement in progress? No ACTION required, all LCOs are satisfied, fuel movement may continu . Fuel movement may continue for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while adjusting the setpoint to within the limi Fuel movement may continue for up to 30 day Nel movement must be suspended until an appropriate portable continuous monito "ith ic same Alarm Setpoint is provided in the fuel storage pool are ANSWER: Fuel movement must be suspended until an appropriate portable continuous monitor with the same Alann Setpoint is provided in the fuel storage pool are REFERENCE: Technical Specification 3.3.3.1, ACTION 28 JUSTIFICATION: A incorrect Tech Spec Minimum charmels regt. ired is 2, currently only B incorrect, the setpoint is not the problem, the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit does not appl C incorrect, ACTION required is to provide the backup or suspend fuel movemen D correct, until the backup is provided, no fuel movement can occur, then the monitor must be returw to OPERABLE status in 30 days or suspend fuel movemen K/A: 072 K3.02 Effects on fuel handling operations OBJECTIVE: RMS05C, RMS07C, RMSOSC New Question
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RO 49 Containment purge and exhaust are in operation.
Fuel Drop monitor,3RMS'RE41goes into high alarm.
Which of the following describes the automatic response of the system to the alarm? All four containment purge supply and exhaust valves close only. All four containment purge supply and exhaust valves close and the running exhaust fan,3HVR FNdA or 48 stops. One supply and one exhaust valve closes only.
D. - One supply and one exhaust valve closes and the running exhaust fan,3HVR FN4A or 4B stops.
ANSWER: One supply ana exhaust valve closes only.
REFERENCE: LSK 22.27B JUSTIFICATION: Purge isolation takes place if EITHER fue'. drop monitor goes into alar (One monitor isolates the inside containment valves, the other monitor isolates the outside containment valves.) Th6 fans do not get trippe (C correct only)
OBJECTIVES: RMS05C; RMS07C; RMS08C K/A: APE 061 A1.01, Automatic Actuation of ARM 3.6/3.6 modified from 398 and 2117
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- RO 60 l
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Plant Conditions:
- An ATWS has occurred due to a fire in the switchgear room
- The reactor could not be tripped locally because the breakers are fused together o Reactor power is 40% and decreasing
- Tav is increasing
- Pressure is being maintained by the PORVs cycling around 2350 psia The PREFERRED sequence to add negative reactivity to shutdown the reactor is: Perform an immediate boration, drive rods in auto or manual, initiate a safety injection Drive rods in auto or manual, shift char 0i ng pump suction to the RWST, perform on immediate boratio Perform an immediate boration, maximize charging flow, drive rods in auto or manua Shift charging pump suction to the RWST, drive rods in auto or manual, perform an immediate boratio ANSWER: Perform an immediate boration, maximize charging flow, drive rods in auto or manua REFERENCE: FR JUSTIFICATION: The first 4 steps of FR-S.1 contain the preferred sequence for adding negative reactivity in an ATWS situatio OBJECTIVES: FS102C K/A: 029 EK3.12 Actions in EOP New question
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RO 51 The plant is la Mode 1. The following personnel are in the control room The CO who is the " Operator at the Controls", the US and the SM. The other CO is on a plant tour.
When, if at all, may the US be considered the " Operator at the Controls?" If the " Operator at the Controls" must leave the red carpeted area (Operations Area). If the " Operator at the Controls" is around back of the main control board acknowledging an annunciato If tl1 Shift Manager is in the surveillance area and a proper turnover occur The US cannot be considered the " Operator at the Controls".
ANSWER: If the Shift Manager is in the surveillance area and a proper tumover occur REFERENCE: COP-200.1 Section 1.7 & COP 200.1 Attachments 5 & 6 Tech. Specs 6. JUSTIFICATION: *A" incorrect due to CO not restricted to Operations area (COP 200.1, Section 1.6)
"B" incorrect because CO allowed behind boards to acknowledge main control board annunciators and is still the " Operator at the Controls"(COP 200.1, Section 1.6)
"C" correct. With the SM in the surveillance area and with a proper relief the US can become the " Operator at the controls"
"D" incorrect. Section 1.7 of COP 200.1 states 1 licensed operator in the surveillance area at all times in addition to the SM or the US
!n the control room. if other personnel are to be considered in the surveillance area, they shall meet the normal relief requirement OBJECTIVES: NAD407 K/A: 2.1.2, Operator Responsibilities Modified question 2189 I
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RO 52 Safety related plant equipment is known to have been operated in a manner which had the potential to damage the equipmen Which of the following describes an action which must be taken in accordance with COP 200.1 CONDUCT OF OPERATIONS? Shift Manager shall notify the Duty Officer and initiate a C A notification must be made to the NRC within twenty four hours of the even The Unit Supervisor shall initiate a priority 1 AWO for maintenance to investigat The Operations Manager shall notify the Technical Services Engineering Manager and cognizant system engineer.
-- ANSWER: Shift Manager she" notify the Duty Officer and initiate a C REFERENCE: C OP 20 JUSTlFICATION: If plant equipment is known to have been operated in a manner which had the potential to damage the equipment, the following actions shall be taken:
Operator shallinform the US of the proble US shall direct equipment or plant be placed in a safe conditio SM shall describe the problem in the SM lo If potentially damaged equipment is safety-related, the following additional actions shall be taken:
SM/US shall check applicability of Technical Specifications or Technical Requirement SM shallinform the Duty Office SM shallinitiate an C SM shall initiate appropriate investigative action to determine status of potentially damaged equipmen OBJECTIVES: NAD413 K/A: 2. ,1.7, Evaluate performance J/udgment 3.7/ Question #3093
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RO 63 l
I Which of the following is a difference between a dual verification and independent verification? An Independent Verification can be performed by the same individual performing the initial verification. Dual Verification requires two individual An Independent Verification is performed prior to the task being performed. Dual Verification is performed upon completion of the tas . Dual Verification can be performed using the " time and distance" method Independent Verification is usually performed concurrently with the initial verifie Dual Verification is performed concurrently with the task. Independent Verification is performed after the task or evolution has been complete ANSWER: Dual Verification is performed concurrently with the task, independent Verification is performed after the task or evolution has been complete REFERENCE: WC-8, Attachment 2 JUSTIFICATION: DUAL VERIFICATION: Dual verification is performed concurrently with the task. In general, dual verification is performed when an action could result in an immediate threat to safe and reliable plant operation. For dual verification, concurrently means the
" performer" and the " verifier" together determine and agree the work location and component are correct for the specified actio And, the task, to the best of their knowledge, will result in the desired outcome. The person performing the task and the person performing the verification both must positively identify the component, determine the actual and required position or state, and agree on the method to be used prior to the action taking place. The verifier essentially completes the verification of the intended action, before the action is undertaken and then witnesses the task performed. it should be noted that if a specific task requires a dual verification during its performance, the need for a separate independent verification of the same evolution is not required. [v Comm 3.1)
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INDEPENDENT VERIFICATION: Independent verificati2n is
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performed after a task or evolution has been completed to I
ensure it has been performed in the correct location, on the correct equipment, or the desired results have been obtaine An independent velification is also performed periodically to ensure l systems, equipment, components, etc. are still in the condition they were left following their last manipulation or verification. To satisfy the requirements for an independent verification, when plant conditions or the situation allows, a good rule of thumb to ensure real independence, is to apply the " time and distance" metho This requires the independent verifier to not visually observe the person who initially perform the task. Conducting the verification in this manner will alleviate the possibility that the performer goes to the wrong item or place and the independent verifier, watching the performer, simply goes to the same (wrong) location and verifies the performers actions were completed correctly (at the wrong locationl). In lieu of using the time and distance technique, as with all verifications, using attentiveness and attention to detail during their performance will prevent inadvertent problems. It is the responsibility of the verifier to ensure he or she is in the correct location, checking the required equipment or components, and determining if they meet the specified acceptance criteri OBJECTIVES: NAD613 K/A: 2.1.29, Conduct / verify valve lineup Question #3268
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RO 54 When is a second Main Condensate Pump required to be operating? When pumping the Main condenser Hotwell. When a Main Feed Pump is operating. When the Main Circulating Water Pumps are operating. Anytime steam is being released from the Steam Generators.
ANSWER: When a Main Feed Pump is operating.
REFERENCE: OP 3319A section 4.1 JUSTIFICATION: Pumping the hotwell requires only one condensate pump and is only performed when the plant is in a shutdown condition and in long recycle (no MFPs operating)-('A' incorrect)
Operation of the Circ pumps requires that tube sheet seal water be applied. This can be supplied from the CNS system during plant shutdown conditions and does not require use of the condensate pumps ('C' incorrect)
Steam may be released from the SIGs to atmosphere with feed being supplie OP 3319A section 4.1 states "A minimum of two condensate pumps must be running while feeding steam generators with main feed pumps." This requires that a 2nd condensate pump be operating any time a main feed pump is operating ('B' correct)
OBJECTIVES: CNM06C (a)
K/A: 2.1.32, System limitations / precautions Question #2964 -
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RO 55
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The following plant conditions exist:
- The plant was operating at 100% power when a large LOCA occur . A Containment Depressurization Actuation Signal (CDA) is generated due to high containment pressur . All safeguards equipment operates as designed except the B EDG fails to auto start and cannot be storie minutes later, while performing E-0, Reactor Trip or Safety injection, all offsite power is los Approximately 2 minutes later, the RO checks power availability to the Safeguards equipmen What should be the status of the RSS pumps? All four pumps should be runnin Only RSS pumps 'A' and 'C'should be runnin Only RSS pump *A" should be runnin None of the pumps should be runnin ANSWER: Only RSS pumps 'A' and 'C'should be runnin REFERENCE: LSKs 24 9.4A,24 9.48,24-9.4Q, and 2711J JUSTIFICATION: The CDA signal starts the 11 minute timer to start the pumps. Once started, if an LOP occurs, the pumps are restarted by the sequencer after 60 second A is incorrect, becauss the 'B' EDG is not running C is incorrect. Both "A' train pumps will start. Only the 'A' pump would start if in the SI/Recirc mod D is incorrect because the "A" train pumps will star OBJECTIVES: CDA06C(1); CDA07C K/A: 103 K1.08 SIS /CDA including reset Bank item
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RO 66
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A Safety injection has occurred and the crew was conducting a brief at the end of Step 14 of E-0 when a complete loss of Off-Site power occur The Emergency Diesel Generators start, their output breakers close to restore power to the vital busses, and the sequencers complete their sequencing on of load What is the status of Containment Air Recirculation (CAR) fans? All the CAR fans will be runnin The "A" and "B" CAR fans will be runnin Only the "C" CAR fan will be runnin None of the CAR fans will be runnin ANSWER: The "A" and "B" CAR fans will be runnin REFERENCES: LSK 22-?7.C, LSK 24 9.4a JUSTIFICATION: The crew has stopped the "C" CAR fan at Step 12 of E-0 ("A" and "C" wTong).
The "A" and "B" fans will start on the Si signal, and when the LOP occurs, the "A" and "B" fans will trip and will sequence back on at 39 second ("B" correct, "C" and "D" wrong.).
OBJECTIVE: CVS03C (b.4)
K/A: 022 A3.01 initiation of Safeguards 97 Exam 8
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RO 57
During an uncontrolled rod withdraw from 175 steps on *D' bank, the final steady state actual reactor power will . and RCS Tave will . (Assume no operator action / reactor does not trip) Increase, increase Increase, remain the same Remain the same, decrease Remain the same, increase ANSWER: Remain the same, increase REFERENCE:
JUSTIFICATION: The p reactivity addition of the control rods will cause power and Tave to initially increase. The increasing Tav will drive power down to approximately the initial level. The end result approximately at the same power level with an elevated temperature. The temperature increase will feedback negative reactivity to offset the positive reactivity originally added by rod motio OBJECTIVES: ROD 06C (e)
K/A: 001 K1.03, Relationship of reactivity and Rx power to rod movement New Question
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RO 58 Given the following conditions:
. MP3 is holding at 75% power following a refueling outag . Rod controlis in automati . Rod height is 220 on Bank Power range channel N-44 fails high.
I WHICH ONE of the following describes the response on the rod control system? Rods will drive in until a Tav Tref error develops which will result in the rods driving ou Rods will not mov Rods will continuously drive in unless stopped by operator interventio Rods will drive in until the power mis match circuit output decays awa ANSWER: Rods will drive in until the power mis-match circuit output decays awa REFERENCES:
JUSTIFICATION: The over power rod stop will prevent outward motion (part a is incorrect).
Rod stops don't prevent inward auto motion (B is incorrect).
Temperature error and power mismatch circuit output decaying away will stop rod motion. (D is correct)
C is incorrect because the rods will eventually stop driving in without operator interventio OBJECTIVES: NIS07C (g)
K/A: 001 K1.05, Cause/effect between CRDS and NIS and RPS Modified item 2207
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l RO 59 During a Reactor Startup, you must ensure the reactor goes critical above the rod insertion limit. The reason that this is safety significant at this time is to ensure: Peak centerline fuel melt temperatures are not exceeded if you ejected a rod during startu Sufficient positive reactivity is available by the control rods to offset power defect on the power escalatio No power tilt is introduced :n the core because Bank C is F artially withdrawn with Bank D still on the botto The reactor will have adequate shutdown margin for a sieam line break acciden ANSWER: The reactor will have adequate shutdown margin for a steam line break acciden REFERENCE:
JUSTIFICATION: Tech Spec Bases. The hot zero power Rod insertion Limit is to ensure reactor DNB limits are not exceeded on large Steam Line Break and the reactor doesn't return to critical on small steam line break. (D is correct)
A is incorrect because HCP rod ejection do not cause fuel melt concerns, rod ejection is limiting at HF Power defect must be offset by rods and dilution on power escalation Normal rod height for criticality is about Bank D at 150 steps. (B is incorrect).
OPTR is not a concem less than 50% OPTR insures FQ (LOCA)
calculations are within limits. This is not a concem at hot zero power. (C is incorrect).
OBJECTIVES: ROD 03C(c); RODOBC(b)
K/A: 001 KS.08, Rll Setpoint New Question
l RO 60 l
PLANT CONDITIONS:
. A small break LOCA has occurred
. The crew is in ES 1.2,' Post LOCA Cooldown and Depressurization*
- An RCS cooldown has been initiated by dumping steam to the atmospher Which of the following statements describes the optimum reactor coolant pump configuration, and the basis for this configuration? All RCPs should be stopped to minimize RCS inventory loss following break uncovery, and pre'ent steam volding in the reactor vessel on subsequent RCS depressurizatio One RCP should be run to produce effective heat transfer and RCS pressure control, yet minimize RCS heat inpu One RCP should be run to produce effective heat transfer and RCS pressure control,
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yet minimize RCS inventory los Two RCPs should be run to ensure symmetric heat transfer to the intact SGs, to enhance RCS pressure control, and to prevent steam voiding in the reactor vessel head on the subsequent RCS depressurizatio ANSWER: One RCP should be run to produce effective heat transfer and RCS pressure control, yet minimize RCS heat inpu REFERENCE: ES 1.2 Background JUSTIFICATION: Forced coolant flow is the preferred mode of operation to allow for normal PCS cooldown and provide PZR spray, All but one should be stopped to minimize heat input into the RCS. 'A' is incorrect because if all RCPs are stopped volding could occur in the vessel head during the depressurization. 'C' is incorrect because the RCS inventory loss is based on the existing differential pressure and not on the forced flow through the RCS. While the reasons in distracter 'D' are correct, the procedure does not address running two (2) RCP OBJECTIVES: S1203C K/A: WEST 03 EA1.3, Desired results Exam 8 l
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l RO 61 Initial Plant Conditions:
e The plant is at 100% power at EOL
- Bank D rods are at 225 steps
- All controllers are in automatic The following MB annunciators are received:
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TAVE/ TREF DEVIATION
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ROD POSITION DEVIATION
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POWER RNG CHANNEL DEVIATION
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POWER RNG FLUX RATE HI Based on the above indications, which of the following events has occurred? The contolling first stage pressure transmitter has f ailed hig An NIS power range upper detector has failed lo A rod position indication channel has failed lo A control rod has dropped into the cor ANSWER: A control rod has dropped into the cor REFERENCE: AOP 3552 JUSTIFICATION: Tne dropped control rod will generate the negative rate signal and the NIS power channel deviation because of the power tilt in the core. The dropped rod will cause the plant to cooldown causing the Tave/ Tref deviation alarm. (D is correct)
OBJECTIVES: A5203C; ROD 07C(c)
KIA: 003A2.03, Dropped rod using in-core /ex-core inst., loop tem New Question
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R3 62 Which of the following represents the safeguards signal that actuates the listed CVCS system valves?
RCP Seal Return Chargina Flow Control LeidgetalSQlAll00
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Isolation CIB CIA Si ' CIA SI SI CIA SI CIA CIB CIA CIB ANSWER: CIA-- SI CIA REFERENCES:
JUSTIFICATION: The seal retum MOVs close on CIA and CVCS letdown and charging flow controlisolation valves close on an S OBJECTIVES: ECCO3C (b.1)
KIA: 006 K4.09, Valve position on an Si modified test item 2382
RO 63 Which of the following describes how the Emergency Diesel Generators are placed in the
" Speed Droop" mode of operation? Automatically selected when started Manually from Main Board 8 and no safeguards signals are present.
> Taking the mode selector switch on Main Board 8 to " Unit". Taking the mode selector switch on Main Board 8 to " Parallel". Automatically selected for all Emergency Diesel Start ANSWER: Taking the mode selector switch on Main Board 8 to " Parallel".
REFERENCE: LSK 24 9.3C JUSTIFICATION: Speed droop is selected when the mode selector switch is in the
" Parallel" position, providing no safeguard signals are present. Speed Droop allows the diesel speed to change allowing load sharing. When the diesel is the sole source of power, (i.e. In Unit) no speed droop is allowed to ensure equipment is running at full spee OBJECTIVES: EDG02C 9v); EDG06C (f)
K/A: 064 A4.06, Manual start /stop of EDG New question
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RO 64 l
A large break LOCA occurred on unit 3. All systems responded as designed. When transitioning from E-0, the US directs you to perform an evaluation of the CSFs. You identify an orange path on integrity, a red path on containment and a yellow path on core cooling, a yellow path on heat sink, and a yellow path on inventor Based on this information, the operating crew should: Transition to E Transition to FR- Transition to FR Transition to FR- ANSWER: Transition to FR REFERENCES:
JUSTIFICATION: First transition to toe highest priority critical safety function procedure which is FR Z.1 ti e only ced path (correct), other answer are incorrect in accordance with EC>P rules of usag OBJECTIVES: E0004C K/A: 2.4.1, Ability to recognize entry conditions for EOPs Modified Question 533
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RO 45 The following plant conditions exist:
Plant is in Mode 5 with allloops FULL
Train 'A' of RHR is in servios providing shutdown cooling
Train *B' electrical outage is in progress and is expected to last another 6 hrs.
Due to a plane crash n the switchyard, til offsite power is lost
The *A* EDG starts I ut fails to load onto Bus 34C
The Unit Supervisc. enters the appropriate procedure for loss of shutdown cooling How should the crew re-establish shutdown cooling? Transition to EOP 3501, Loss Of All AC Power (Mode 5,6 and Zero), and restore power to be able to restart cooling, B, Satti and align the SBO Diesel onto Busses 34A/C and then perform Attachment B Loss of Shutdown Coolong And/Or RCS inventory Mode 5, of EOP 350 Initiate decay heat removal as the RCS heats up by dumping steam from at least one available S Inject available SI Accumulators into the RC ANSWER: Transition to EOP 3501, Loss Of All AC Power (Mode 5,6 and Zero), and restore power to be able to restart cooling, REFERENCE: EOP 3505 Step 1 JUSTIFICATION: The appropriate procedure for loss of shutdown cooling in Mode 5 is EOP 3505
'A' is correct. Step 1 RNO will direct a transition to EOP 3501 if neither Bus 34C or 34D is energize *B' is incorrect because the SBO diesel should only be started using the guidance of EOP 3501, EOP 3501 directs the operators to attempt manual loading of the operating EDG before directing action to start / load the SBO diese 'C' and 'D' are incorrect because they are strategies used in EOP 3501 after all attempts at energizing operable and degraded busses have been unsuccessfu '
OBJECTIVES: E05805C (3)
K/A: 2.4.9, Low power operations in EOP
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Exam Bank hem 2935 RO 66 The plant is operating normally at 100% power.
Which of the following conditions require a reactor trip followed by a trip of ALL RCPs? VCT temperature increases to 1406F. Isolation of a single train of RPCCW to containment. Seal injection flow to each pump decreases to 5.5 GPM. The "A" RCP bearing oil temperature increases to 200'F.
ANSWER: VCT temperature increases to 140*F.
REFERENCE: AOP 3561 Foldout Page JUSTIFICATION: (A correct) If VCT temperature is greater than 135'F AND RCS temperature is above 400'F, then all RCPs must be stopped (RCPs are also stopped at any time if VCT is above 150 F)
(B incorrect)Both trains of RPCCW to containment must be lost to trip RCPs (C incorrect) A reactor and pump trip is required if seal injection flow is less than 6 GPM ANQ thermal barrier cooling is also lost, (D incorrect )High oil temperature only requires a trip of the affected Pump OBJECTIVES: A61761D (1)
K/A:- 2.4.11, Knowledge of abnormal operating procedures Question # 275
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RO 67 The plant tripped from 100% power. A large break LOCA has occurred. The control room has carried out E-O and is now entered E-1. A computer failure has occurred and the US is performing a manual status tree chec The following conditions exist:
SR energized with a negative SUR Core Exit TC's 640 degrees Subcooling 28 degrees RVLMS Plenum 100%
S/G NR's 4%
Wal AFW flow 500 GPM Colo leg temp decrease in last hour 20 degrees RCS temperature 480 degrees Containment Pressure 60 psia Pzr levelis at 0%
RVLMS upper head is at 100%
What answer best describes the sequence for dealing with the Critical Safety Functions Heat Sink Containment B, Containment Heat Sink C.- Heat Sink Core Cooling Containment ,
D.' Containment Heat Sink Core Cooling ANSWER: Heat Sink
- Containment REFERENCES: CSF status Trees. OP3272 EOP Users Guide Section 1,6 JUSTIFICATION:
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.'ed with a' negative SUR- Green Path Subcriticality
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> "s 640 degrees
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RVLMS Plenum 100S Yellow Path Core Cooling S/G NR's 4%
Total AFW fiow 500 GPM Red Path Heat sink Cold leg temp decrease last hour 20 degrees
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RCS temperature 480 degrees Green Path integrity Containment Pressure 60 psia Red Path Containment Pzr levelis at 0%
RVLMS upper head is at 100% Green Path inventory ( A correct) Red paths are first. Heat sink is before Containment (B incorrect) Containment is after Heat sink (C is incorrect) Core cooling is a yellow path and is before Containment which is a Red path ( D incorrect) Heat Sink and Containment are out of order of priority OBJECTIVES: E0004C K/A: 2.4.21, Knowledge of Status Trees / Logics for CSF New question ,
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RO 68 PLANT CONDITIONS:
Plant is in MODE 3 Tave is 557 F Main steam pressure is 1092 psig, controlled via turbine bypass valves Atmospheric steam dump valves (ASDVs), 3 MSS *PV20A, B, C and D are in AUTO, set at 1150 psig What is the proper method for changing ASDV set points? Place turbine bypass valve controller (MSS PK507) in MANUAL, Place all ASDV controllers in MANUAL. Adjust each ASDV set point to the desired value, then place the ASDV in AUTO. Return MSS-PK507 to AUTO. Place one ASDV controller in MANUAL. Adjust the set point to the desired value, then place the controller in AUTO. Repeat this process for each remaining ASDV. Slowly dial the ASDV thumbwheels to the desired value, ensuring the at point tracks properly and the valves do not open. Place MSS-PK507 in MANUAL. Ensure all ASDVs are closed, then place each ASDV in MANUAL, one at a time, and adjust set points to the desired value.
ANSWER: Place one ASDV controller in MANUAL. Adjust the set point to the desired value, then place the centroller in AUTO. Repeat this process for each remaining ASDV.
REFERENCES: OP 3203 JUSTIFICATION: 4.26 Place the Atmospheric Steam Dump controllers in MANUAL one valve at a time prior to making any setpoint changes, then, Return the controller to AUTO. This prevents the Atmospheric Steam Dumps from opening rapidly causing steam pressure transients OBJECTIVES: NAD106; NAD108 KIA: 2.2.2, Local / manual operations of controllers Question: 2131
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RO 69 i
Whlie performing a new Surveillance on the Safety injection pumps, the CO performs a step as written and notices that the Safety injection pump does not have adequate recito flo What is the first action (s) the CO should take? Determine the cause of the proble B.- Continue with the surveillance, but consult with the Unit Superviso Stop the surveillance and place the Safety Injection pump in a stable or safe conditio Initiate a procedure modification in accordance with DC-1, Administration of Millstone Procedures and Form ANSWER Stop the surveillance and place the Safety Injection pump in a stable or safe conditio REFERENCE: DC-4, Procedural Compliance, JUSTlFICATION: Inadequate c. 'Jnaxpected Results 1.9.1 IF procedure appears to be inadequate, OR yields unexpected results while executing work activity, PERFORM the following: STOP work activity, IF applicable, PLACE equipment or system in stable or safe condition. (C correct)
c, CONSULT First Line Supervisor for direction. (B incorrect) DETERr'NE cause of problem. (A incorrect) IF necess y, Refer To DC 1, " Administration of Millstone Procedures and Forms" and INITIATE modification to procedure to rectify problem (Stephen E. Scace memo " Procedure Compliance - Management Expectations" to Millstone Station Personnel, number MP 91-801, dated October 10,1991, states: "If you think you can't follow the written procedure, consult First Line Supervision to determine what actions ( procedure changes) are necessary before proceeding. If a procedure cannot be followed as written: a. Stop the task and place the equipment or system in a safe condition. b. Change the procedure using the procedure change process. c. Proceed with the task.")
OBJECTIVES: RAD 544
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K/A: 2.2.12, Knowledge of Surveillance Procedures Question : 3260 modified
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RO 70
- When preparing a clearance, which of the following system / equipment conditions should be isolated from the work area by two closed valves in series? A fluid system which operates at 170' A gas system which operates at 50 psi Caustic or acid systems at any temperature or pressur D.- Systems from confined work space ANSWER: Systems from confined work space REFERENCE: -WC2 JUSTIFICATION: If practical, Isolate fluid or gas systems that operate at greater than 200'F -
- or 100 psig from the work area with two closed valves in series. Isolate l _
systerns from confiref work space with two closed valves in sef,e '
OBJECTIVES: NAD318 K/A: Generic 2.2.13 Tagging / Clearances
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97 LOIT Remediation Exam I
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i RO.71 Given the following conditions:
- MP3 was operating at 100% power when a spurious Si occur * All systems respond as designed with the exception of the "B" reactor trip breaker, which did not open and remains closed, a The crew responds IAW EOPs, eventually transitioning to ES-1.1, "Si TERMINATION".
- The crow takes both Si reset switches to RESE WHICH ONE of the following describes the status'of St? Both trains of 81 are reset and automatic initiation blocke Neither train of Si is reset, nor is automatic initiation blocke Both trains of Si are reset, only the "A" train automatic iniflation is blocke Both trains of Si are reset, only the "B" train automatic initiation is blocke ,
ANSWER: 3 Both trains of Si are resst, only the "A" train automatic initiation is blocke t REFERENCES:
.,1USTIFICATION: Both trains of Si can be reset but only Train "A" is blocked because P-4 is [
not enabled because B reactor trip breaker is closed. , ("C" is cotrect.) I o Si cannot be blocked in Train '8" because P-4 is not present. ("A" and
"D" are incorrect.)'
Si can be reset in both trains. ("B"is incotrect.) -
OBJECTIVES: RPS012C ,
K/Ai' 013 K4.01, SIS Roset New Question
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RO 72
! Following a Loss of Coolant Accident, adverse containment conditions exist. The following values have been recorded by the ST IlME fcORTAINMENT TEMP CONTAINMENT RADIATION LEVELS 0800 1850F 5 x 10' R/HR
0815 190 F 2 x 10'R/HR
0830 160 F 2 x 10'R/HR 0845 1750F 1 x 10'R/HR 0900 1700F 4 9x 10 R/HR When, if ever, may the crew suspend the use of adverse containment values?
A. - At 0845 because containment temperature has decreased below it's adverse setpoin At 0900 because both containment temperature and radiation levels are below their adverse value Adverse values can never be relaxed once they are entered if containment temperature limits were exceede Adverse values can never be relaxed once they are entered if containment radiation limits were exceede ANSWER: Adverse values can never be relaxed once they are entered if containment radiation
- limits were exceeded.-
REFERENCE: OP 3272
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- JUSTIFICATION: "A"is inconect because containment radiation levels are still adverse and adverse values will always appl "B" is incorrect because 10' RIHR limits are/were exceede "C"is incorrect because of containment radiation levels haven't exceeded
10 R/HR - containment tempurature regarding adverse values can be relaxed when temperature drops < 180 OBJECTIVES: E0003C KIA: W 16 EK 1.3, Hi Rad Alarms and Actions Modified Exam item 3213
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RO 73 FR C-1 is entered when core exit thermocouples are greater than 1200 F. Implementation of FR-C.1 is safety significant at this time because additional operator action is required to: Prevent core uncovery. Provide core cooling to stop the hydrogen generation due to zircaloy water reaction.
C, Limit containment pressure to less than design pressure. Provide core cooling to prevent exceeding peak clad temperature limits.
ANSWER: Provide core cooling to prevent exceeding peak clad temperature limits.
JUSTlFICATION: When FR-C,1 is entered the core is already uncovered. ("A"is incorrect.)
H3 generation from zircaloy water reaction starts ~ 1800 - 2200F. ("B"is in:orrect)
FR-C.1 is written for a small break LOCA with no high head injectio Car fans and spray will limit pressure. FR-C.1 established injection flow to cool core. ("C" is incorrect.)
OBJECTIVES: MC1004 K/A: 017 A2.02, Mitigating Core Damage New Question
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RO 74
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Given the following coaditions:
. The unit is at 45% power with all loops in operatio . Control Rods M12 and D4 in Bank 'D' , group 2, are stuck and misaligned and will not mov . The affected rods are trippabl . The Bank D stuck rods indicate 156 and 162 steps respectively on DRPI while the other Bank D rods and step counters indicate 180 step Assuming the rods cannot be repaired within the next week, which one of the following correctly describes the actions required by Technical Specifications for the misaligned rods? Verify shutdown margin requireme nts are met within 2 hou Verify QPTR within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and app'y LCO 3. Align the remaining Bank *D' rods to i 12 steps of the inoperable rod Be in HOT STANDBY within the nex 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ANSWER: Be in HOT STANDBY within the next 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
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REFEREi4CE: Technical Specification 3.1.3.1, AOP 3552 ' Malfunction of the Rod Drive System'
JUSTIFICATION: Technical specification 31.3.1 outlines the actions to be taken for a single stuck but trippable control ro A is incorrect because sh Jtdown margin must be satisfied in one hour not B is incorrect because OPTR requirements are not applicable when less than 50% powe C is incorrect because multiple rods in the same group are misaligned by greater than 12 steps. Alignment is not allowed and the unit must shutdown and be in hot sta1dby because of the multiple misaligned rods in the same group (D is conect).
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K/A: 005 K3.06, Actions in EOP OBJECTIVE: ROD 08C Modified NRC Exam item 95 LOIT Exam
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'RO 75:
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! INITIAL CONDITIONS:
.- MODE 5 with the RCS soli .' Temperature being maintained 140*F to 150*F by the 'A' RHR trai .. The plant is currently 4 days into a scheduled 8 day *B' train electrical outage, with 34B and 34D doenergized. *B' train load centers are NOI cross-tied to the 'A' trai AssWng NO operator action, which of the below statements describes the plant response to a -
- loss of the 'A' instrument Air Compressor? ' RCP Thermal Barrier cooling flow from.RPCCP will decrease, B,. - Letdown flow will increase resulting in the RCS depressurizin C .- .RCS temperature will increase due to increased RPCCP flow through the RHR Heat Exchange ' RCS temperature will decrease due to increased RHR flow through the RHR Heat Exchange ANSWER:
D.' RCS temperature will decrease due to increased RHR flow through the RHR Heat
' Exchange REFERENCES: P&lD 104A,121 A,112A,121B JUSTIFICATION: 'B' is incorrect because a loss of IAS will cause HCV128 to fail closed which will result in a loss of letdown and a resultant increase in RCS pressur 'C' is incorrect because a loss of iAS will cause FV66A to fail AS IS resulting in NO change in RCS temperature from CCP flow. 'D' is correct because FCV 618 fail closed and HCV 606 fails open on a loss of IAS
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which will result in maximum flow through the RHR HX<
3 'A' is incorrect because the CCP retum valves from the thermal barriers
. have a LOCK UP feature to prevent them from being affected by a loss if IAS. The CCP CTMT isolation valves are MOVs and therefore are not -
affected by a loss of IAS; '
OBJECTIVES: PAS 07C;- RHR07C K/A: .- 078 K3.02, Pneumatic control valves 96 AOP Exam --
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RO 76 Given the following:
. A loss of offsite power has occurrod.
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. Tave is 552'F
. " Turbine Bypass Tav Interlock Bypassed" is illuminate . Steam dumps are in steam pressure mode of contro * Steam dumps demand is manually INCREASED to begin a cooldow . The steam dumps failed to ope Which ONE (1) of the following explains why the steam dumps will NOT open? P-12, LO-LO Tave, has disarmed the steam dump P-4, Reactor trip, has locked out the steam header pressure signa s C-9, Condenser available, interlock is not me The plant trip controller has not rese ANSWER: C-9, Condenser available, interlock is not me JUSTIFICATION: "C" is correct because power is not available to cire pumps on loss of offsite power Blocking signals override arming signal "A"is incorrect because P-12 has been bypasse "B" is incorrect because P-4 locks out the load rejection controller in the Tave mode of contro "D" is incorrect because plant trip controller was reset when you shifted to pressure mod OBJECTIVES: SDS06C K/A: 051 K3.01 Steam Dump Operation - Loss of Vacuum Exam item: 2422 I
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RO 77 Which of the following conditions occurring concurrently with a large LOCA will required entry into ECA-1.1 - Loss of Emergency Coolant Recirculation? Off site power is lost and the 'B' EDG did not start.
] The power lockout relays fall to operate and all the white power lockout indicator lights are di The A charging pump tripped on overcurrent and the B SI pump is tsyged out for maintenanc The 'B' & *D' Recirc spray pumps are damaged and cannot be started,
.d.
i ANSWER: Tlie power lockout relays fail to operate and all the white power lockout indicator lights are dim.
M REFERENCE: ES-1,3 step 2, notes prior to step 2 & JUSTIFICATION: A is incorrect. The A EDG is still available to supply A train components for cold leg recir C is incorrect - One charging pump and one SI pump are still available for cold log recirculatio D is incorrect because the A train of RSS is still operabl B is correct because without the power lockout operating power not available to operate some of the recirculation valves, ES-1.3 directs the operator to ECA- OBJECTIVES: A1101C K/A: W/E 11 K2.1. Control, function, system Modified 2853
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RO 78 Which of the following plant conditions will cause the TD AFW pump to auto atart? -
- A. - 2/4 SG level detectors at low - low level in two SGs, Safety injection Signa Main Feedwater Isolation Signa D. - Loss of Battery Bus 5 ANSWER:
. A.- 2/4 SG level detectors at low - low level in two SG REFERENCE:
JUSTIFICATION: 2/4 low low signals in at least 2 SG will start TD AFW pum Only the motor driven AFW pumps start on SI (D is incorrect).
Main feedwater isolation will only isolate MFW but does not start any AFW pump . TD AFW pump will start on loss of batt Bus 1/2 not Bus 5 (D is incorrect).
OBJECTIVES: - FWA04C '
K/A! 059 A4.11, Auto start AFW Modified 349 '
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RO 79 Assume that prior to a startup, work on the Intermediate Range nuclear instrumentation resulted in BOTH channels being OVER COMPENSATE Which of the following describes the expected system response to this condition?
] During startup the Intermediate Range indication will be less than actual, and during shutdown the Source Range may not be automatically reinstate During startup the Intermediate Range indication will be greater than actual, and during shutdown the Source Range may be reinstated prematurely causing an unwanted reactor trip.
] During startup the Intermediate Range indication will be less than actual, and during shutdown the Source Range may be reinstated prematurely causing an unwanted reactor tri During startup the intermediate range indication will be greater than actual, and during shutdown the source range may not be automatically reinstated due to the P-10 permissive being activ ANSWER: During startup the Iniermediate Range indication will be less than actual, and during shutdown the Source Range may be reinstated prematurely causing an unwanted reactor tri REFERENCE: Funct. Diag. Sht. 3 & 4 JUSTIFICATIO A is incorrect because on shutdown the Intermediate range detectors will read lower than actual and be automatically reinstated prematurel B & D are incorrect because readings on startup will be lower than actual not highe OBJECTIVES: NIS06C (a); NIS05C K/A: 032 A2.04, SR/lR overlap Bank Item 2217 l
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R2 80 l
PLANT CONDITIONS:
- 100% power
- All systems in AUTOMATIC a
LT-459 selected for control of Pressurizer level control selected to LT-459
+- PT-456 selected for control of Pressurizer Pressure
- Instrument for "A" and "C" steam generators selected to Channel 1
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Instruments for "B" and "D" steam generators selected to Channel 11 A loss of VIAC-2 occur Which of the following lists controllers which should be taken to MANUAL as a result of the VIAC-2 failure? Rod control Pressurizer Pressure Pressurizer Level Rod Control Pressurizer Pressure Master main feed pump controller Pressurizer Pressure Pressurizer Level Feed Regulating Valves for "A" & "C" SGs Pressurizer Pressure Master main feed pump controller Feed Regulating Valves for "B" & "D" SGs ANSWER: Pressurizer Pressure Master main feed pump controller Feed Regulating Valves for "B" & "D" SGs REFERENCE: AOP 3564, Process sheets 10 and 11,25 JUSTIFICATION:
Pressurizer level controller will be affected because the backup channel will result in a letdown isolation Pressurizer pressure is affected because its controlling channel is channel 11 Main feed pump speed control is affected due to loss of two steam flow channels, lo Rod Control is not affected The Feed Regulating Valves on only the "B" and "D" SGs will be affecte Only D correc ._-_______ \
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OBJECTIVES: 12005C K/A: ape 057 A1.06, Manual control of components New question
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- - _ _ _ _ _ ___ _ _ ___ __- _ _ .RO 81 The plant is operating at 100% power. The steam dumps are in the Tave mode of control and rods are in AUT MSS *PT507, Main Steam Header Pressure, fails high. This will: Open the steam dump cooldown valves, and the TDFWPs speed will decreas Block the load rejection controller from arming the steam dumps, and the TDFWPs speed willincreas Arm the steam dumps but they won't open, and the TDFWPs speed will decreas Have no effect on steam dumps, and the TDFWPs speed willincreas ANSWER: Have no effect on steam dumps, and the TDFWPs speed willincreas REFERENCES:
JUSTIFICATION: PT 506, not PT 507, feeds the load rejection controller and arms the steam dumps (B & C are incorrect),
The PT 507 effect on the cooldown valves function is only available in the pressure mode of control (A is incorrect).
PT 507 is only in effect when in the pressure mode of control (D is correct)
PT 507 feeds the TDFWP speed control circuitry. An increase in pressure will cause the pumps speed to increas OBJECTIVES: SDS07C K/A: 039 A2.04, Manual control of components Modified 2426
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.RO 82 INITIAL CONDITIONS:
. The Unit is operating at 48% power with the "NIS POWER RANGE P-9 PERMISSIVE" Blue Light NOT Li .. Due to an instrument failure, actual level in the "B" S/G level has increased to 80%
resulting in an automatic FWI actuatio . A!I equipment operated as designe Assuming no actions are taken in the instrument rack room, which of the following must occur to allow resetting the FWI signal from the main boards? Clear P-14 and Reset P-4 Clear P-14 and P-9 C.- -Clear P-14 only Reset P-4 only ANSWER:
I Clear P-14 and Reset P-4 REFERENCE: Functional Sheet 13 JUSTIFICATIOh Permissive is lit below P-9, if level reaches the turbine trip setpoint, the reactor will trip, therefore to reset the FWI, both r'-4 and P-14 will have to clea . OBJECTIVES: NISO4C (b.4)
K/A: 059 A 4.11 Permissives Question ID 1520 96 Quiz 6
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RO 83 Initial Plant Conditions
- Plant is at 60% power e Rod controlis in manual
. Steam Dumps are in Tave mode of control l,' A turbine trip occurs. The turbine trip fails to cause a reactor trip and actuate the steam dumps on the turbine trip controlle The next automatic reactor trip signal to be generated for this transient would be: High Pressurizer Level Trip
. OTAT High Pressurizer Pressure Trip OPAT ANSWER: High Pressurizer Pressure Trip REFERENCES:
JUSTIFICATION: The power mismatch will cause a rapid increase in pressurizer pressure causing the reactor to trip (C is correct).
Pressurizer level will backup the high pres:. re trip (A is incorrect)
B is incorrect. The temperature increase will drive power down and pressure up. Both of these factors are benefits with regard to OTA D is incorrect. Power will decrease during the transient. The margin to the OPAT trip will be increasin OBJECTIVES: A5002C; MC0302 K/A: 045 A1.05, RCS following turbine trip New Question i
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._ RO 84
- Plant History:-
.- A loss of off-site power occurred 10 minutes ag . The crew has stabilized the plant and just completed ES- .- RCS temperature has stabilized at 557'F
= Steam Generator levels are between 25 - 30% in the narrow rang . Total AFW flow to the steam generators is 535 GP The Operations Manager has directed the plant be maintained at the current plant conditions for RCS temperature and steam generator level _
One day from now the total AFW flow should be approximately:7 The sam GP GP GP ANSWER: GPM.
- REFERENCE:
JUSTIFICATION: One minute after a reactor trip the decay heat level is approximately 3-4% After one hour it is approximately 1.5-2% and after one day it is 0.7-1%. Consequently after one day the existing decay heat is approximately one quarter of its value following the trip. Therefore since the AFW system is maintaining SG level, the flow will reduce by one quarter to approximately 134 GP OBJECTIVES: S0103C K/A: W/E 9 EK Relationship between emergency feedwater flow to S/G and decay heat removal for facility heat removal following a tri New Question
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RO 85
' PLANT CONDITIONS:
- Reactor is operating at 100% rated thermal power
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Annunciator 5 3 on MB4C *PR UP DET HI FLUX DEV/ AUTO DEFEAT' has alarmed
. All control rods are positioned within 12 steps of their group demand counters
. Maximum QPTR based on plant computer program 3R5 is 1.04 Assuming QPTR is not reduced, within iwo hours reactor power must be reduced to and NIS overpower trips reduced to within the next four hour %, 59 % %, 55 % %, 97 % %,103 %
ANSWER: %,97 %
REFERENCE: Tech. Spec. 3.2.4 Action Statement c.2; OP 3273 Modified Ban JUSTIFICATION: If QPTR is greater than 1.02 but less than 1.09 then within 2 hrs reduce thermal power 3% of rated power for every 1% greater than 1.0 and similarly reduce the overpower trip setpoints within the next four hour OBJECTIVES: NIS08C (b)
K/A: 015 A1.04, NIS/QPTR Modified 1058 -
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RO 8 Given the following conditions:
.. The unit is critical
.! The crew is holding power at 1.0 x 10-8 amp . SG levels are being controlled on the bypasses, in automati N 36 control power fuse blow WHICH ONE of the following describes the plant's response? - An IR high flux rod stop will be received and the reactor will remain critica The reactor win remain critical with no rod stop i The reactor will trip on SR high flux when they automatically energiz The reactor will trip on IR high flu ANSWER: The reactor will trip on IR high flu REFERENCES:
- JUSTIFICATION: Loss of control power fuses causes a trip signal to be sent to RPS through the Reactor Protection System and will also generate a rod stop -
and reactor trip on 1/2 coincidence (d is correct)
A is incorrect because the trip will cause the reactor to go subcritica (This also makes b incorrect).
C is incorrect as source ranges will not automatically energiz OBJECTIVES: NIS07 (c)
K/A: 015 K2.01, NIS channels, power supplies Modified item 2269 l
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RO 87
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The following conditions exist:
. Crew is in EOP 3503, Shutdown outside Tne Control Room
. Control room is filling with dense smoke
- Control room is ordered evacuated
. The reactor is tripped from 100% power
. The turbine is tripped
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Si occurs after trip due to the steam dumps malfunctioning Which of the following describes the procedural flow path under these conditions: Complete EOP 3503 and then enter E- Exit EOP 3503 and enter E- Perform E-0 in parallel with EOP 350 Complete EOP 3503, then perform cooldown in accordance with EOP 350 ANSWER: Complete EOP 3503, then perform cooldown in accordance with EOP 350 REFERENCES:
JUSTIFICATION: EOP rules of usage - if Si or Rx Trip occurs in EOP 3503, you should -
remain in EOP 350 OBJECTIVES: EOU (1733)
K/A' 067 K3.04, Actions in EOPs New Question
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)-- RO 88
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Given the following conditions:
. . Unit is at 100% power
. Pressurizer level control is selected to 459/461
. VCT makeup controlis in Automatic q
A reference leg leak occurs in pressurizer level transmitter LT-45 Assume no other operator action, which of the following occurs: Letdown isolation occurs 3CH-FCV 121 ramps open Auto makeup occurs i Unit will eventually trip on high pressurizer leve B, 3CH-FCV-121 ramps closed
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VCT diverts Letdown isolation occurs Pressurizer level will begin to increase Unit will eventually trip on high pressurizer leve C, Letdown isolation occurs 3CH-FCV-121 ramp open Auto make-up occurs y
Plant cools Awn
_ Unit trips on low pressurizer pressure because heaters are de-energize CH-FCV-121 rampt closed VCT diverts Letdown isolation occurs Pressurizer level will continue to decrease due to seal leakof Unit trips on low pressurizer pressure ANSWER: CH-FCV .121 ramps closed VCT diverts Letdown isolation occurs Pressurizer level will begin to increase Unit will eventually trip on high pressurizer leve REFERENCES:
JUSTIFICATION: Reference leg failure will cause indicated level to fall high this causes !
FCV-121 to ramp close. Thus A & C are incorrect).
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Act pressurizer level will decrease and the renaming channels will cause letdown isolation. Seal injection will fill pressurizer and cause eventually a high level trip (8 is correct).
D is incorrect because seal injection still occurs even if the FCV-121 is closed and level will begin to increase.
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OBJECTIVES: A5503C; PPLO6C; PPLO7C K/A: 011 K3.01, Loss of PZR Level effect on CVCS Modified 375
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RO 89 Given the following conditionsi
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The unit is at 8% powe Plant startup is in progress -
- - Pzr level instrument LT-459 has failed LO *
All actions of AOP 3571 " Instrument Failure * Attachment C are complete.
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Which of the following describes the course of action the crew should take if a subsequent failure of Pzr levelinstrument LT-460 HIGH7 Verify reactor tri !
- Stop the startup, and restore one of the failed channels of pressurizer level to ,
OPERABLE status prior to inc, easing power above _10%. Stop the startup, and restore both of the failed channels of pressurizer level to OPERABLE status prior to increasing power above 10%. Within one hour initiate ACTION to be in at least HOT STANDBY within the next 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ANSWER: Stop the startup, and restore ONE of the failed channels of pressurizer level to OPERABLE status prior to increasing power above 10%.
REFERENCE: AOP 3571 " Instrument Failure' Attachment C, Pzr Level and Pressure Control Lesson Plan, Technical specification 3.3.1 and functional sheet
JUSTIFICATION: With all actions of the AOP compete, the bistable associated with the high Pzr level Rx. trip has been p 'ced in a tripped condition When the second channel fails high, the coincidence for a high pressurizer level reactor trip is met, however, the trip is blocked less than 10%. (Aincorrect)
Technical specifications require 2 channels to be OPERABLE, however, this is required below P-7 (10%), and to increase above 10%, the bistables must be tripped within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, B correct, D incorrec It is not required to have both channels OPERABLE to increase above 10%, (C incorrect)
OBJECTIVES: PPLO7C K/A:- 028 A1.01, Pressurizer level bistables modified from 1995 MP3 NRC exam
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2 RO 90 1Given thu following conditions:
. -The reactoris trippe .- ; A Loss of_Offsite Power has occurred -
e' Safety injection is actuated from a small LOC . . All ECCS equipment is operating as expecte .- Pressurizer level is 48% and increasing on all channel .. -- RCS pressure is 1700 psia and decreasing slowly on all channel : Which one of the following describes a leak location that is consistent with the indications - I given?
- ~ A leaking pressurizer safety valv < The letdown line relief valve lifting
- A reference leg break on pressurizer level instrumentation.-
- A failed open spray valv ANSWER: ;. A leaking pressurizer safety valv REFERENCES:
JUSTIFICATION: L A PORV or safety valve failing open will cause pressurizer level to 1 increase and pressure to decrease on all channels.-- ("A"is correct.) -
"B" is incorrect because this break location will be isolated by the SI/ CIA signal
"C" is incorrect. On the effected reference leg the indicated level channel would increase and pressure would decrease. However, on the non-
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affected channel level will decrease as well as pressurizer pressur "D" is incorrect - a failed open spray valve will only cause pressurizer
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pressure to decreas OBJECTIVES: A5503C-
-.K/A: 008 A1.01, Operation Monitoring Instrumentation trom for PORV, sprays New question
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RO 91 During performance of the shift control room rounds, the Control Operator discovers that 3HVQ-RE49, ESF Building Normal Ventilation Monitor, indicates OFF-LINE for both Data-A and Data-B at the RMS Consol Which of the below statements describes the operating status of the radiation monitor? The radiation monitor may be considered operational once its data and operation is verified et the Local Indicating Control panel (LIC). The radiation monitor must be considered inoperabl The radiation monitor continues to indicate properly at the RMS Console but all radiation monitor control functions must be performed manually at the RMS Consol The radiation monitor may still be considered operational because it will still notify controi room staff of high radiation conditions by actuating the " RADIATION ALERT * and
' RAD Hl* annunciators on MB ANSVER: The radiation monitor may be considered operational once its data and operation is verified at the Local Indicating Control panel (LIC).
REFERENCES: P&lD 152A, RMS073T and RMS073C Handouts, PIR 391-043, MP3 Memo MP-3-0-385 dated 3/25/91, Kaman instiumentation Operation -
Maintenance Manual volumes 1 thru 3 and RMS Console Help displa JUSTIFICATION: 'B' is incorrect because each local unit is comp!etely self contained, requiring the computer room computer ONLY for transmitting data to the control roo 'C' is incorrect because if off-line from both data-A and data-B, all communications between control and the RMU is terminate 'D'is incorrect because the alarms at MB2 are a function of the computer, if the RMU computer is not communicating with the computer room computer, it can not cause the MB2 alarms to actuat 'A' is correct because although not communicating with the control room, each RMU is completely a stand alone unit and is designed to function without the control room computer. Once the data and operation has been verified correct for the RMU at its LIC, the unit may be considered operational per the SS (Memo MP-3-0-385 and PIR 391-043 and RMS073 handouts)
OBJECTIVES: RMS08C K/A: 073 A4.02, RMS Control Panels / indications Exam item 2407 l
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RO 92 Given the following:
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A twenty five (25) year old Maintenance Contractor with complete exposure records has the following exposure record for the current calendar yea . Shal!ow Dose Equivalent - 2.55 REM
-. Committed Dose Equivalent - 0.75 REM
. Deep Dose Equivalent - 2.13 REM
. Lens Dose Equivalent - 3.08 REM
. Committed Effective Dose Equivalent - 1.95 REM WHICH ONE (1) of the following is this individuals Total Effective Dose Equivalent (TEDE) for the current calendar year? .88 REM .08 REM-C- 5.21 REM D.- 5.43 REM ANSWER: .08 REM REFERENCE: RPM 1. Get RAD Worker Training JUSTIFICATION: TEDE = CEDE + DDE = 1.95 REM + 2.13 REM = 4.08 OBJECTIVE: GET Radworker Training K/A: 2.3.1 / 3,0 10CFR20 Radiation Limits New Question
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I RO 93 --
The rad waste PEO is dispatched to change LWS-FLT3. This PEO has not performed this task
- before. The HP technician informs the PEO that the dose rate on the outskie of the filter housing is 1 R/h Wnich one of the following is notan example of ALARA techniques for reducing exposure for filter replacemen A Long handled tools to remove the old filter B Place the filter in a shielded drum upon removal to reduce exposure Have the rad waste PEO be assisted by the Turbine Building PEO who has done the task several times befor Have the new PEO perform the filter replacement on a mockup firs ANSWER: Have the rad waste PEO be assisted by the Turbine Building PEO who has done the task several times befor ,
REFERENCE: RPM 5.2.4 Section 1.2,1.3. JUSTIFICATION: RPM 5.4.2 list 3 main areas to reduce Radiation exposure. Time Distance and Shielding A is Distance i B is Shielding D is Reduced Time by practice on a mockup prior to the jo RPM 5.2.3 Section 1.1 states that individual exposures within a work group are balanced consistent with experienc C will not balance the exposure if the experienced person always does the jo OBJECTIVES: NAD721; NAD722; NAD723 K/A: 2.3.10, ALARA procedures to reduce radiation exposure New question
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RO 94
. Given the following'conddions:
. MP3 is at 35% powe * "RCP B STANDPIPE HI LEVEL" has li *- "B' seal injection flow is 8.2 GP *
.. B" seal leak-off flow is 0.2 GP . Seal return temperature is 150'F and rising steadily. .
. Pump radial bearing - rising slowly @ 145'F Based on the above indications, the operating crew should:
A.' Trip the unit, secure .*B' RCP and close its No.1 seal leakoff valve within 2 minute B, Trip the "B" RCP and close its No.1 seal leskoff valve after the pump has been stopped -
for five minute Close the "B" RCP's No.1 seal leakoff valve within 5 minutes and shutdown the unit within the next 30 minutes then secure the *B' RC Trip the *B' RCP and close its No 1 seal leakoff after the pump has been tripped for two minute ANSWER:
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- Trip the 'B' RCP and close its No 1 seal leakoff after the pump has been tripped for two
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minute REFERENCE: OP 3554
" JUSTIFICATION: Since power less than P-8, the RCP can be stopped without tripping the unit (a is incorrect)
- OP .9554 requires tripping RCP and closing the seal leakoff valve within .
two minutes (D is correct) The RCP must be removed from service within 5 minutes of failure (not within 5 minutes of closing sealleakoff valve). (B and C are incorrect)
. OBJECTIVES: AS403C l K/A: 015 A2.01, Cause of RCP failure-Modified question 1104
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RO 95 WHICH ONE of the following interlocks must be satisfied to start an RCP?
' RCP #1 seal AP must be greater than 200 psi The overcurrent trip selector switches must be in the cold positio Cold leg anctot leg isolation _ valves must be ops Cold leg isolation valve must be open and the loop bypass and hot leg isolation valve must be close ANSWER:
- Cold leg and hot leg isolation valves must be ope REFERENCES: RCP text; OP 3301B JUSTIFICATION: A and B are incorrect because they are procedure administrative requirements but are not part of the interlock circuitry C is correc D is incorrect. The cold leg stop valve must be closed with the bypass fully open to satisfy the RCP interloc .
OBJECTIVES: RCSO4C K/A: 003 K6.14, RCP starting requirements Modified question 2159 s
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RO 96 '
A small break Loss of Coolant Accident has occurred.
The current plant conditions exist at the completion of E-0 step 14:
- Si has occurred, e All Si equipment starte . Containment Temperature is 185' * RCS pressure is 1800 psia and Stabl . CET's are 520* . Pressurizer level is 50% and slowly increasing.
Assuming conditions do not significantly change, you would expect to stop one charging pump in: E-0 Reactor Trip on Safety injection ES-1.1 - Si termination. ES-1.2 - Post LOCA Cooldown and Depressurization ES - 1.3 - Transfer to Cold Leg Recirculation ANSWER: ES-1.2 - Post LOCA Cooldown and Depressurization .
REFERENCES:
JUSTlFICATION: A is incorrect because adequate subcooling doesn't exist and RCS pressure is less than 1950 psia (adverse containment) to stop a charging pump in E- B is incorrect because adequate subcooling doesn't exist to make the transition to ES- C is correct because a 100*F/hr cooidown will be started which will increase subcooling major to allow stopping a charging pump in ES- D is incorrect because charging pump will be stopped in ES-1.2, and cooldown will place plant on RHR, ES-1,3 will not be entered for a small break LOCA.
OSJECTIVES: S1203C K/A: E02 EK2.1, St Termination Modified Exam item 1533
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RO 97
- Gnven the following conditions
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- - 'The Unit was ' operating at 75% power, ,
- A small break LOCA occurred in coincidence with a loss of off. site power -
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e: RVLIS indicates that a void axists in the reactor vessel head;-
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The cookkwn was stopped and RCS pressure raised to regain subcooling margi ; locreasing RCS pressure will
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the size of the void and the leakage from i
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the RCS, u .Increace; increese B; . Decrease; increase Increase; decrease ' Decrease; decrease ANSWER:'
, ; Decrease; increase
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- REFERENCE:
a JUSTIFICATION: Raising the pressure will decrease the size of the void but increase the -
leakage for the RCS, - ("B" is correr:t)
- OBJECTIVES: ;MC0703 (c)
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K/Ah - 009 K3.06, inventory Balance During Small Break Loss.-
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-RD 982 PLANT CONDITIONS:
.- Plant is in Mode 6
. No fuel movements are in progress-
... fA" Train Electrical outage is in progress
. - Computeris available ..
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- . *B' Spent Fuel Pool cooling pump caught fire and tripped i hour ago
.- Spent Fuel temperature - 11S*F and slowly increasing .
. _ Spent Fuel Pool level- 37% and decreasing slowly
. Reactor Cavity seal-intact-a Fuel Building Morioring Group Histogram NORMAL e- RWST level- 1,000,000 gallons For the existing plant conditions, which one of the following corrective actions should be taken?
A. . Align RWST to gravity feed the spent fuel poo #-
B .- L Establish emergency makeup using the fire water syste _ Supply makeup to the Spent Fuel Pool from the Primary Grade Water System
- D.- Establish emergency makeup to the Spent Fuel Pool from Service Wate ANSWER:
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- A. - Align RWST to gravity feed the spent fuel poo REFERENCE: EOP 3505A, Att. A,' Step 3
> JUSTIFICATION: Gravity feed is the preferred ' method to the spent fuel pool. (A is correct)
Emergency makeup using the fire water _ system is only used if the additional attempt of emergency makeup from the RWST is attempted after the gravity feed method does not work (B incorrect).'
C and D are least preferred and are only done if RWST is not available (C & D are incorrect).
OBJECTIVES: 'E0503C; E05A3C K/A:' 033 A2.02, Loss of Spent Fuel Cooling EOP actions Exam item 1295l 1- . . . . . . . . . . _ . . . . . , ;....... -
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RO tt The following events have occurred:
- A SGTR has occurred subsequent to a steam break inskje conta!nmen . The crew has transitioned to E 3, *SGTR', from E.2, *Feulled Steam Generator Isolation.'
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They have identified and Isolated the ruptured Steam Ger'erator, which is not faulted, and are preparing to initiate RCS cooldow Current Plant Conditions:
- Containment temperature is 185'F e RCS pressure is 1420 psia e Ruptured Steam Generator pressure is 895 psig e intact Steam Generator pressures are 850 pclg e Faulted Steam Generator is 600 psig
- Core exit temperature is 485'F Using the provided reference, determine the required core exit temperature to be achieved by the RCS cooldown, if necessar A cooldown is not necessary, core exit temperature is already less than the required I temperatur *F 'F *F ANSWER: *F PROVIDE. ATTACHMENT TO STUDENTS REFERENCE: E-3 Step 14a graph (Adverse CTMT parameters to be used)
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JUSTIFICATION: CTMT 185'F Adverse parameters. Step states not to interpolate, therefore,885 psig on graph to be used. A incorrect since RCS is above the required temperatur *F = non adverse number for 850 psig (D incorrect).
OBJECTIVES: E3003C K/A: 038 A1.34, Cooldown to sperific temperature 97 EOP Exam 681
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-RO 100 PLANT CONDITIONS The unit is running down in power from 100% to take the unit off lin Loop 3 Teve fails unobserved to a constant output of 572' Which one of the following describes where pressurizer le vel will stabilize under these plant conditions? Assume no operator action take Pressurizer level will stabilize at: % - % % %
ANSWER: %
REFERENCES:
JUSTlFICATION: Pressurizer level is programmed with auctioneered high Tav. Pressurizer level wdl decrease, then control at program level for 572*F which is 45%.
(C is correct).
B is incorrect because pressurizer level will not be decreased no load valu A is incorrect because pressurizer level will not decrease to cause let -
down isolatio O is incorrect as pressurizer level will not increase above program valu OBJECTIVES: PPLO7C K/A: 004 A1.02, CVCS/Tav/Pressurizerlevel
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Remetor Examination Answer Key 1.bowda St c Nok! b
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10. e
!M 60, b
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11. d 61. d l 12. c 62. c
- 13, b 63, c 14 b 64, c 15. a 65, a 16, c 66, a 17. d 67. a 1 , b 19. b 69 c 2 . d 2 c 22. d 72. d 23 b 73. d 24. c 74. d 25. d 75. d
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26.a- 7 . b 77. b 28 c 78. a 29. c 7 . c 80. d 31. b 81. d 32. b 82. a 33, c 83. c 34. b 84, b 35. c 85 c 36,b 86. d 37. d 8 .d 88,b 3 b 40,b 90, a 41 b 91, a 4 b 43. b 93. c 44 c 94. d 4 , c 46, d 56, c 47,b 97. b 4 ,a 49. c 99, c 50. c 10 . _ _ _ _ _ _ _ _ _ _ _ - - _ - -
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Attachment 3 Millstone Unit 3 WRITTEN EXAM COMMENT AND NRC RESOLUTION RO Question #1 Facility Comment: "RO Exam Test item - Question #1: The question asks for conditions that would require immediate boration by the Neactor Operator. The facility recommends acceptance of an additional answer (a) that an immediate boration is required in Mode 5 if shutdown margin requirements are not satisfied in accordance with the attached curve From the curves, the required boron under the given plant conditions is 2125 ppm. Additionally nota the entry conditions from the attached copy of AOP 3500 and OP 3200B."
NRC Resolution: Following review of the referenced procedures and curves the examiner agreed with the facility comment. There were two correct answers to RO question No.1. In accordance with Interim Revislen 8 of the Examiner Standard (ES) 403, Paragraph D.1.b, two correct answers were allowed for this question. The answer key was changed accordingly.
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Attachment 4 SIMULATION FACILITY REPORT Facility Licensee: Millstone Unit 3 Facility Docket Nos: 52du Operating Tests Administered from: Julv 7 throunh 10,1997 This form is used only to report simulator observations. These c,bservations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation f acility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observations.
During the performance of simulator scenarios and JPMs there wore several deficiencies that caused confusion or apprehension of the candidates. The following is a listing:
- During the performance of JPM 109, when charging flow was being re established, the flow controller (3CHS*FCV121) went to maximum flow when its outlet isolation valve (3CHS'MV8106) was opened. The controller was previously set to minimum flow by the operator, per procedure. This problem was observed four times during the performance of this JP * During one performance of JPM 109, AOV 197A/194A, the non saiety header isolation valves for the running reactor plant closed cooling water would not open although the safety injection (SI) signal had been reset (SI annunciator was not lit).
The operator reset the Si a second time and the valves functioned properl * During performance of JPM 50, the spray valve controller was supposed to fallin the open position and remain there when shifted to manual to conduct an alternate path JPM. On two occasions when shifted to manual the valve responded normally and allowed the operator to control reactor pressure normally. The incorrect response of the simulator removed the dynamic actions required by this JP Proper operator knowledge on actions for the anticipated failure were verified by subsequent questions by the examiners. The chief examiner also questioned the simulator operator to verify that the simulator had been property r.etup for this JPM and was satisfied with the simulator operator's answer * During several power maneuvers the main turbine was slow to settle at load set causing the operating crew to believe that there was a possible problem with the main turbine control system. This simulator response did not adversely impact the simulator scenarios other than extending the duration of the scenarios and causing operating crew concern.