IR 05000440/1997309
| ML20199A354 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/13/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20199A336 | List: |
| References | |
| 50-440-97-309OL, NUDOCS 9711170147 | |
| Download: ML20199A354 (164) | |
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U - S. NUCLEAR REGULATORY' COMMISSION - -!
REGION lil
i Docket No: 50-440- -i License No: NPF-58 Report No: 50-440/97309(OL):
Licensee: Centerior Service Company i
. Facility: Perry Nuclear Power Plant Location: P.- O. Box 97, A200 Perry, OH 44081 i
Dates: August 25 - 30,1997 I
Examiners: D. McNeil, Chief Examiner, Rlli J. Lennartz, Examiner, Rill H. Peterson, Examiner, Rlli T. Jones, Examiner (In training), Rlli Approved by: M. N. Leach, Chief Operator Licensing Branch
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9711170147 971113
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EXECUTIVE SUMMARY Perry Nuclear Power Plant NRC Examination Report 50-440/97309 A licensee developed and NRC approved initial operator licensing examination was administered to seven Senior Reactor Operator license applicants and two Reactor Operator license applicant Results:
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One Reactor Operator license applicant and one Senior Reactor Operator license applicant did nat meet minimum NRC standards for passing the examination and were denied operator licenses. All other applicants passed all portions of their examinations and were issued operstor/ senior operator license Overall operator performance during the examination was determined to be satisfactor Eyamination Summarv:
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The SRO applicants taking this examination were not as well prepared as those at previous examinations. (Section 05.1)
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The written examinations were challenging, discriminating, and written at the right difficulty level. (Section 05.2)
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The administrative Job Performance Measures used to examine the applicants proved to be a good evaluation tool for determining applicant competence. The applicants were well prepared for this portion of the examination. (Section 05.3)
l None of the applicants were able to correctly calculate an Average Power Range Monitor thermal trip setpoint when provided with a specified core flo (Section 05.4)
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Licensee prepared operating Job Performance Measures were considered goo The follow up questions were not sufficiently discriminating. They were memory level or direct look up questions. (Section 05.4)
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Licensee prepared dynamic scermios were considered good. Significant applicant weaknesses were found in the areas of feedwater level control, recire flow control, and execution of PEl-B13, RPV Control (ATWS) when manually inserting control rods. (Section 05.5)
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No simulator deficiencies were noted during the examination. (Section 05.7)
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Reoort Details 1. Operations
- 05 Operator Training and Qualificdion 05.1 General Comme 013 Operator in!tial license examinations were administered at the Perry Nuclear Power Plant (PNPP) to seven Senior Reactor Operator (SRO) applicants and two Reactor Operator (RO) applicants during the week of August 25,1597. Two SRO applicants were previously licensed at the fac!!ity as ROs. One RO applicant failed the written
. examination and one SRO applicant failed the written, Job Performance Measures (JPMs), and dynamic simulator scenario examirations. All other applicants passed all portions of their examinations and were issued operator license . The PNPP training department participated in a pilot process !n which the initial license examination was developed by PNPP training department instructors and approved by the NRC in accordance with guidance prescribed by NUREG 1021," Operator Licensing Examiner Standards," Revision 7, and superseded in part by Interim Pilot Examination Guidance provided in Generic Letter (GL) 95-06, " Changes in the Operator Licensing Program." As part of the pilot program, the NRC administered the operating test and the licensee administered the written tes All materials developed by the licensee for the examination were submitted to the NRC on or ahead of schedule. The submitted material was of high quality and with only a few exceptions, was used as written by the examiners The examiners observed that the SRO applicants taking this examination were not as well prepared as those at previous examinations performed at PNPP. The examiners noted that SRO applicants examined at previous PNPP examinations were all well prepared to take the examination. Some SRO applicants presented for this examination appeared well prepared for their examination but others were weak in their execution of SRO responsibilities. It was the opinion of the NRC Chief Examiner that, with the exception of a few applicants, the SRO command and control skills of the applicants were generally lower than those observed during previous examinations. This intormation is provided for feedback into the facility systematic approach to training (SAT) program, no written response is necessar During the administration of the examination a significant weakness was noted in all applicants in the areas of the feedwater system and the reactor pressure vessel level control system. Applicants missed questions in this area on the written examination, during JPM follow up questions, and during dynamic simulator scenario demonstration Additional training in these areas may be warranted for applicants examined during this examination. This information is provided for feedback into the SAT program, no written response is necessar m
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- 05.2 - Written Examination ' Examination Scone i
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Using NUREG-1021," Operator Licensing Examiner Standards," GL95-06," Changes'in-the Operator Licensing Program," and NUREG/BR-0122, " Examiner's Handbook for Developing Operation Licensing Written Examinations," examiners reviewed each written examination question. Each question was reviewed for comprehension, validity and level of difficulty, Observations and Findinos NRC examiners made minor comments on approximately twenty of the one hundred and twenty five questions submitted by the licensee. This was substantially below the number and depth of comments returned to other facilities involved in the p!!at examination process. This was an indication of the high quality of the submitted examinatio The written examinations were challenging, discriminating, and written at the right -
difficulty level. Suggestions were made by the NRC examiners to change some questions on the RO examination as they involved tasks normally assigned to SRO The examination author was able to provide lesson plan objectives demonstrating the facility's requirement for the ROs to be knowledgaaole on the examined material. An analysis of the examination grading revealed eleven questions m!: sed by both RO applicants and four question missed by more than 50% of the SRO applicant Enclosure 4, Written Examination Analysis, contairis the questions missed by ooth ROs -
and more than 50% of the SRO An NRC examiner was in the examination room when the written examination was administered. The licensee's examination administrator conformed to all requirements for administering the written examination. All examination rules were read to the applicants and enforced by the examination administrator. When the written examination wLs completed one RO applicant approached the examination administrator and indicated that a reference, PEl-B13, RPV Control (Non-ATWS), was
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not in the reference materials at his table. The examination administrator checked the
- reference material packages for each examination table and confirmed that the reference was at the applicant's table and was overlooked by the applican Two applicants failed the written examination with scores of 76 (RO) and 77 (SRO). All other applicants passed the written examination with scores ranging from 81 to 9 Conclusions-The written examination provided by the training depariment proved to be a high quality evaluation tool for determining applicant competence. The training department has the capability of developing a challenging, discriminating $ written examinatio '
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05.3 Administrative Job Performance Measures
' Examination Scope Using NUREG 1021, Operator Licensing Examiner Standards, examiners reviewed each administrative Job Performance Measure (JPM). Each administrative JPM was reviewed for applicability, importance, and safety significance. The aggregate of the administrative JPMs was reviewed to ensure all required areas of the administrative JPM examination were repreNnte QhsgIyations and Findings The office review of the submitted administrative JPMs indicated they met all NRC guidelines. The administrative JPMs were well planned and prepared.- No significant problems were encountered by NRC examiners while validating or administering the examinatio The appliccats were well prepared for the administrative JPM portion of the examination. No significant adverse comments were made by any of the examiners coaceming applicant performance co. the administrative JPM Conclusions The proposed administrative JPMs provided to tne NRC indicated creative thought and good preparation. The administrative JPMs used to examine the applicants proved to be a good evaluation tool for determining applicant competence. The applicants were well prepared for this portion of the examination. The licensee's training department has the capability of writing administrative JPMs that meet NRC examiner requirement .4 Ooerating Job Performance Measures Examination Scoce Using NUREG-1021, Ooerator Licensing Examiner Standards, examiners reviewed the operating JPMs. Each JPM was reviewed for applicability, importance, and safety significance. The aggregate of the JPMs was reviewed to ensure all required areas of the JPM examination were represented. JPM follow up questions were reviewed for applicability and to determine if they were considered direct look up or memory level question Observations and Findings The licensee submitted JPMs that met standards prescribed in NUREG-1021 and GL 95-06 for addressing the various types and numbers of functional safety systems, alternate path and shutdown plant requirements. Critical tasks were correctly identified within the JPM ._ _
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Many JPM questions were considered direct look up questions by Ele NRC examiners, it was determined that facility tralners did not understand the correct definition of ? direct look up! questions. - NRC examiners provided guidance to the facility trainers who
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the status of other direct look up questions to closed reference question A broad spectrum of individual weaknesses was detected during JPM administratio The following syrtopsis of the weaknesses found is provided as input for the training _
depadment's SAY program, no response is required: _
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Severe applicants were unable to correctly identify the loads supVed by the .
RCIC g'and seal compressor. This was a JPM follow up question; no reference allowe Several applicants were unable to correctly identify the RCIC system response to an increasir g suppression pool water level when the RCIC system was sligned in the Condensate Storage Tank to Condensate Storage Tank mode (full flow test). This was a JPM follow up question; no reference allowed. All of the operators missing this question believed the rising water level would have no ,
effect on the RCIC system because.of the test lineup.
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Several applicants were unable to correctly explain tne plant response to a failure of a Reactor Pressure Vessel (RPV) level detector when it was selected for feedwater level control. This was a JPM follow up question; no reference allowe Several applicants were unable to provide the location of the facility hydrogen detectors used for post accident analysis. This was a JPM follow up question; no reference allowe None of the applicants were able to correctly calculate an Average Power Range Monitor (APRM) thermal trip setpoint when provided with a specified core flo This was a JPM follow up question; reference allowed.
- Condusions Facility instructors presented a good package of JPMs that met all NRC requirements for use to examine the applicants except for the JPM follow up questions. Most JPM follow up questions did not meet the criteria for open reference questions until reworked or replaced. Although many individual weaknesses were apparent to the NRC examiners, in general, the operators were well prepared for the operating JPM portion of the examination. The facility's training department has the capability of writing operating JPMs that meet NRC examiner requirement .
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05.5 Dynamic Simulator Scenatios Examination Scoce Using NUREG-1021, Operator Licensing Examiner Standards, examiners reviewed each dynamic simulator scenario. Each scenario was reviewed for content, applicability, and safety significaric Observations and Findings The dynamic scenarios submitted by the licensee had the required number of malfunction events for each applicant. The initial conditions were different for each scenario and included low power scenarios. The submitted scenarios included equipment or instrument failures after the major transient had occurre The office review of the scenariw showed that the submitted scenarios needed only minor modifications to be acceptable for use as an NRC examination. Some events did not provide the opportunity for the NRC examiners to fully evaluate applicant skills, but by inserting minor changes such as changing the ramp rate for malfunction insertion, changing the order of the events, and replacing events, the examiners were able to ensure compliance with NUREG 1021 guidelines. The scenarios used in the examination provided a good opportunity for NRC examiners to evaluate applicant skill Individual candidate weaknesses were detected during simulator scenario administration. The following synopsis of the weaknesses found is provided as input for the training department's SAT program, no response is requ: red:
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Several applicants had difficulty with feedwater control after a manual start of the motor driven feedwater pump. It was apparent they did not understand the relationship between individual pump controis, the startup level controller, and the master water level controlle Some applicants had dif'iculty controlling the recirc flow control system after a runback had occurred. The operators reported to the Unit Supervisor that they had lost control of the recirc flow control system controllers when, if fact, control was still available to the operator. They apparently did not understand the set points and time constants involved within the floT! controller Three of three SRO applicants given an Anticipated Transient Without Scram (ATWS) failed to promptly take action to insert control rods into the core. Each lowered reactor pressure vessel water level to reduce reactor power, but failed to execute the parallelleg of the PEl-B13 procedure (Emergency Operating Procedure) to inser' control rods (PEl-SPI 1.3, Manual Rod Insertion). In two cases the applicants retraced PEl-B13 and noted the failure to complete the parallelleg and ordered the correct actions. In one case the SRO applicant l failed to enter the parallel leg until prompted by another applican .
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Although a few instances of informal or incorrect communications were observed during tha dynamic simulator scenarios, cominunications were considered to be good. The applicants appeared to conform to the facility policy of closed loop, three-way repeal back communication Conclusions The dynamic simulator scenarios submitted by the facility instructors met all requirements of NUREG 1021 but needed some enhancement to improve the opportunity for applicants to demonstrate license skills and for NRC examincts to observe those skills. The facility training department has the capability of writing discriminating dynamic simulator scenario Significant ap,nlicant weaknesses were fcand in the areas of feedwater level control, recirc flow control, and execution of PEl-B13, RPV Control (ATWS), in the area of manually inserting control rod .6 East Examination Activities The licensee provided one post examination comment for the Reactor Operator written examination. There were four clarifications issued by the facility monitor while the written exai.lination was being administere Post Examination Comment Question #16 (RO)
Which one of the following statements is correct concerning an uncoupled control rod during coupling checks? Up to two attempts may be made to recouple the drive, If attempts to recouple the rod sre unsuccessful, the rod should be disarmed at that positio Digital display of control rod position on the Rod Display Module will go blank if the rod is uncouple When the ROD UNCCUPLED indicating light has illuminated, it will not clear until the rod is successfully recouple Answer: Reference: Off-Normal Instruction (ONI)-C11-2, " Uncoupled Control Rod (Unit 1)"
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ONI C112, section 1.3, states " Digital display of control rod position on the Rod ,
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Display Module goes blank." This step provides the reference for distracter ' Distinctor *a.' is also correct because it is not incorrect to attempt to couple a control rod twice. ONl C112, section 4.3.0 states, *lf approved by the Shift Supervisor, repeat the attempt to recouple the control rod." The original intent of this distractor was to imply that a maximum of two attempts could be made; howt;ver, the distractor is not clear in this implication. Therefore, distractor *a" is also correc Facility Recommendatlun:
Both distracter ' a" and "c" are correct; therefore, either should be accepte NRC resolution: The NRC agreed with the comment and amended the RO written examination answer key to allow answers a. and c. to be accepted as correct answer Examination Clarifications:
All examination clarifications were discussed with the NRC examiner present at the examination prior to making the clarification. The examiner agreed with the need to make each of the clarification RO 76/SRO 48 The weeds " reactor scram was not on Level 3, a level 3 signal did ne' occur," were added to the question. This infom1ation was necessary to disqualify distractors a. and RO-83/SRO 55 Prior to the last sentence in the question, the following words were added,"An ATWS resulted in power stabilizing at 30%." In addition to this modification, distractor d. was eliminated. The modification was required to allow candidates to determine that the redundant reactivity control system had been initiated and the distractor was eliminated because it provided a second correct answer tha*. was trivial and not the answer solicited by the questio SRO 80 Candidates were informed that the distractors were listed in the wrong order, i.e. *b-a d-c" We "a-b-c-d." Candidates were told to use the letter that occurs before the choices and not to re-order them. This was a typographical erro SRO 94 The words, *Drywell hydrogen concentration can be maintained
<90%" were added to the end of the question. This was necessary to make the correct answer totally correct in that the Plant Emergency Instruction step that was used to answer this
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Plant Emergency Instruction step that was used to answer this question included this parameter and the candidates asked what the status of this parameter wa > Simulator Fidelity The examiners observed no simulator modeling deficiencies during the examination administration. The lack of simulator deficiencies is docurranted in Enclosure 2, Simulator Fidelity Repor V. Management Meetings t
X1 Exit Meeting Summary The chief examiner presented the examination team's observations and findings to Members of the licensee's management on August 29,1997. The licensee acknowledged the findings presented. No proprietary information was identified during the examination or at the exit meetin .
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PARTIAL LIST OF PERSONS CONTACTED
Licensee i D. Bauguess, initial Training Lead !
D. Johnson, Instructor i W. Kanda, General Manager i J. Messina, Operations Manager !
C. Persson, Operations Training Supervisor !
T. Rausch, Director, POPDD ;
E. Root. Training Manager l L. Zerr, RAS / Compliance l
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D. Kozicff, Senior Resident inspector j ITEMS OPENED, CLOSED, AND DISCUSSED Opened None Cloiad None Dhgussed None
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Enclosure 2 s
SIMULAOR FIDELITY REPORT ,
Facility Licensee: Perry Nuclear Power Plant Facility Licensee Docket No: 50-440 Operati,ng Tests Administered: August 25 29,1997 The following documents observations made by the NRC examination team during the August 1997,initiallicense examination. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of non compliance with
10 CFR 55.45(b). These observations do not affect NRC certification or opproval of the simulation facility other than to provide information which may be used in future evaluation No licensee action is required in response to these observation During the conduct of the simulator portion of the operating tests, the following item was observed: __
ITEM DESCRIPTION
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Enclosure 4 Written Examination Analysis The following is provided for input into the training department's systematic approach to training (SAT) p.vgram. No written response is required. The following questions were missed by both Reactor Operator applicant RO Question #6 ,
During a full flow test of the RCIC system (CST to CST), a problem is encountered and the Control Room operator makes the decision to isolate the RCIC system. He arms and depresses the RCIC Manual isolation pushbutton. How does the RCIC System respond? The RCIC Turbine trip The RCIC System isolates, Division 1 only, The RCIC System isolates, Division 1 and Nothing happens, RCIC System doesn't isolat Answer Both applicants selected distractor b as the correct answe RO Question ff9 While operating at 100% Rx power an l&C Technician asked you to place the RWCU lsolation Bypass switch,1E31 S1 A on H13 P632, to the BYPASS position for the performance of his surveillance test. If you performed this action, what implications willit have on the RWCU system? The low RPV level 2 RWCU isolation will be bypasse l The RWCU isolation due to SLC Initiation will be bypasse The High RWCU differential flow isolation will be bypasse The Non-Regerierative Heat Exchanger high outlet temperature isolation will be bypasse Answer Both applicants selected distractor d as the correct answe i l
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y RO Question #18 s SELECT the statement that describes the expected response of the Containment Vessel and Drywell Purge Supply (M14, System following 6 refueling accident where an exposed ,
luel bundle is dropped and is damaged. The M14 system is currently operating in the Refuel 3 . mode. The M14 system isolation valves: close AND the M14 supply and exhaust fans will trip.
. 4 .rgi b, close which cause the M14 exhaust fans to trip. The M14 supply fans continue to operat close which cause the M14 supply fans to trip. The M14 exhaust fans continue to operat remain open. The M14 supply and exhaust fans trip on a containment ventilation exhaust high radiation signa Answer Both applicants selected distractor c as the correct answe _ _ _ _ _ _ _ _ . _ _ _ _ __
RO Question #20 An ATWS with a MSIV closure has occurred resulting in the following plant conditions:
- Reactor power is 15%.
- Reactor pressure is 890 psi Reactor levelis stabilized at 100" with the MFP on the Startup Level Controlle Rx modo switch is in SHUTDOWN
- Containment pressure is 0.5 psi DW pressure is 2.1 psi Suppression Pool temperature is 112degF.'
- Suppression Pool Level is 18.7 fee RHR Status: NB/C have been terminated and prevente NB are lined up for outside the-shroud injection.
Why should all injection into the RPV be terminated and prevented with the exception of boron and CRD? This terminates or reduces, as much as possible, any continued containment pressure increas This ensures that if Emergency Depressurization is required later, the heat rejected to the containment will be minimized, This ensures that an Emergency Depressurization will not result in exceeding the suppression chamber design temperature, ECCS pumps which take suction on the Suppression Pool could lose adequate NPSH and containment integrity could be lost.
Answer d.
Both applicants selected distractor b as the correct answe . - .
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RO Question #23 Given the following plant conditions:
Rx Power - 100%
RFPT A & B on the Master Level Controlle Narrow Range Channel"A"is selected for input into C3 A loss of bus D1B occurs. if no operator action was taken, what would be the impact on RPV level control? RPV level will rapidly increase due to partial loss of RPV level and feedwater flow signals, RPV level will increase then stabilize near the level setpoint as RFPT "B" speed increase RFPT B speed initially increases then decreases as level error overcomes the flow error signa Signal to LOW FLOW RX LEVEL CONTROL,1C34 R614, falls causing 1N27-F175 to ramp closed if in AUT Answer Both app!! cants selected distractor b as the correct answe RO Question 40 The following plant conditions exist: ECC pump 'A'is running, ECC pump 'B'Is in STANDBY READINESS, ECC Area Cooling Fan 'A'Is RED TAGGED OUT, ECC Area Cooling Fan 'B'
is in STANDBY READINESS Control Complex 'B' Chiller & Chilled Water Pump are operatin As the operator at-the-controls, which one of the following actions should you take? Start ECC pump 'B' from the control roo Have the Nuclear Island rounds taker start ECC Area Cooling Fan 'B'. ECC pump "A" must be secured due to no Area Cooling available, Have Maintenance personnel install temporary ventilation directed at ECC pump 'A'.
Answer One applicant chose distractor a, one applicant chose distractor d as a correct answer to this question.
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RO Question M1 i
There are certain responsibilities that can only be pedormed by the Emergency Coordinator when the Site Emergency Plan is being implemented. Which one of the following cetions is allowed to be performed by someone other than the Emergency Coordinator? Direct the notification of offsite agencies and organization !t Determine the emergency classification including reclassification or t terminatio ! Recommend protective actions for the general public to State and local County Official Coordinate and direct the actions necessary to terminate or mitigate the effects of the emergenc Answer Both applicants selected distractor c as the correct answe RO Question #44 l Due to a xenon transient li1 the core, it becomes necessary to increase Rx power using recire flow to maintain 1250 MWe In accordance with Operations Policy 2 9, Operational -;
Activity Evaluation, how should this evolution be classified by the Shift Supervisor? . As risk significant, As a risk contributo I As an unplanned change in reactivity per ONI-C5 As an Infrequently Performed Test / Evolution (IPTE).
Answer Both applicants selected distractor b as the correct answe ,
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- A reactor startup is in progress with the Rx Mode Switch in "STARTUP/ STANDBY" The !
following is the present status of the APRMs versus LPRM inputs, and indicated power: ,
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APRM A B C D E F G H !
D LevelInputs: 4 5 4 3 4 4 6 6
C Levelinputs: 4_ 3 3 4 6 2 4 4 -
B Levellnputs: 3 4 3 4 4 4 _6 4 t
A Levelinputs: 3 3 3 4 6 4 4 2 l
, indicated Power: 11 % 16 % 12% 11 % 12% 10% 12 % 10%
SELECT the correct f;PS response to the above data: { No trips l Rod Dlock only ! Half scram only j Full scram Answer r i Both applicants selected distractor b as the correct answe ,
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RO Question #S8 The plant is operating at 90% power and stable. The following alarms are suddenly received at H13 P680: -
RPV Level 7 alarm Stm Flow / Feed Flow Mismatch alarm P601 alarm j The plant stabilizes at approximately 92% power, with level at 196". Choose from below the correct response for this transien Secure the HPCS pump, Reset the ADS Initiation logi Manually trip the RCIC turbin : Take the SRV control switch to OF '
Answer Both applicants selected distractor c as the correct answe ,
RO Question #76 A Rx Scram has occurred with both Reactor Feed Pump Turbines (RFPTs) operating. RFPT A's flow controller is in MANUAL and RFPT B's flow controller is in AUTOMATIC. The Startup Level Control Select switch is selected to "MFP", RPV pressure is 825 psig and ,
decreasing at approximately 2 psig per minute. Wh;ch ONE of the following describes the plant response to SLOWLY decreasing the speed on RFPT "A" down to 1100 rpm with the intent of removing it from service? RPV level will... , be held fairly constant at approximately 178" by RFPT "B". decrease as RFPT "A" is removed from service, unless the MFP is started, be held constant for approximately 10 minutes and then increase to level 8, tripping both RFPT be held fairly constant at the tapeset level selected on the Master Level Controller by RFPT "B".
Answer Both applicants selected distractor d as the correct answer, 1 It
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The following questions were missed by more than 50% of the SRO applicants:
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An emergency drywell entry is required, Rx power is 7% MWt with the main generator off- l 4 line. Which of the below statements is correct with respect to walving the ALARA review for the drywell e ntry? ' An ALARA review is not required for a drywell entry under these plant conditions per PAP-0118, ALARA Progra , An ALARA r6 view is required for work activities under these plant conditions !
and may not be waived per PAP-0118, ALARA Progra ! The Unit Supervisor may waive the ALARA review by declaring this an urgent situation per PAP-0512, Radiation Work Permit Progra The Shift Supervisor may waive the ALARA review by declaring this a Priority 1 Emergency per PAP-0902, Work Request Syste Answer '
Five applicants selected distractor b, one applicant selected distractor c, and one applicant selected the correct Onswe .
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SRO Question #29 The following plant conditions exist:
Rx power 100 %
Busses L11 & L12 are being powered from the Unit 1 Aux. Trar.sformer Bus L10 is being supplied by the Unit 1 Startup Transformer Div 1 & Div 3 DGs are in standby readiness Div 2 DG is in " LOCAL" running fully loaded for a WO retest Preferred Source Breakers for EH11 EH12, and EH13 are all closed, With these plant conditions, a bus L10 lockout occurs, a scram signalis generated due to the trip of both Rx Recirc pumps and level 2 ts reached. After auto initiating, the HPCS pump motor catches fire and causes a lockout to occur on bus EH 13. The lowest RPV level achieved during this transient is 100". Which of the following statements correctly describes the steady-state conditions you would expect to see following this transient? Div.'1 DG is the only DG running AND supplying its respective bus, Only Div.1 & 3 DGs are running, supplying their respective busse Div. 2 DG will auto start, but not load when its local / remote switch is taken out of LOCA All three DGs received an auto start on a LOOP signal and are supplying their respective busse Answer Two applicants selected the correct answer, two applicants selected distractor b, one applicant selected distractor c, and two applicants selected distractor ,
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Plant Conditions:
Containment sprays have been running (RHR A and RHR B) for 15 minute .
Containment pressure is 1.5 psi l Drywell pressure is 2.0 psig, ;
Reactor pressure is 50 psig, j Which of the following plant responses describes what you would expect to occur if an !
operator depresses the CNTMT SPRAY A(B) SEAL IN RESET pushbuttons? Nothing, containment sprays continu Containment sprays secure AND the RHR pumps secure, Containment sprays secure AND the LPCI Mode of RHR A and RHR B irAtiat The RHR pumps continue to run, supplying flow to the containment spray header AND the LPCI injection flowpat Answer Four applicants selected distractor d, one applicant selected distractor c, and two selected distractor d as the correct answe SRO Question #48 (RO Question #76)
A Rx Scram has occurred with both Reactor Feed Pump Turbines (RFPTs) operating. RFPT A's ll?w controller is in MANUAL and RFPT B's flow controller is in AUTOMATIC. The Startuo I wvel Control Select switch is selected to "MFP", RPV pressure is 825 psig and decrear.ng at approximately 2 psig per minute. Which ONE of the following describes the plant response to SLOWLY decreasing the speed on RFPT "A" down to 1100 rpm with the intent of removing it from service? RPV level will... be held fairly constant at approximately 178" by RFPT "B".
b, decrease as RFPT "A" is removed from service, unless the MFP is starte be held constant for approximately 10 minutes and then increase to Level 8, tripping both RFPT be held fairly constant at the tapesel level selected on the Master Level Controller by RFPT "B".
Answer Six applicants selected distractor d as the correct answer, one selected c as the correct answe . . - . . - . - - - .- _ _ . .- - - .- .-
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ES-401 Site specific Written Examination Form ES 401 1 [
Cover Sheet i U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC WRITTEN EXAMINATION APPLICANT INFORMATION Name: HASTER EXAMINATION Region: lli Date: August 30,1997 Facility / Unit: Perry Nuclear Power Plant
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License Level: RO Heactor Type: GE
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INSTRUCTIONS ,
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percen Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination start All work done on this examination is my own. I have neither given nor received aid, r
Applicant's Signature RESULTS Examination Value 100.0 Points Applicant's Score Points Applicant's Grade Percent
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- _ - - -. .- . . -- ES 402 Policies and Guidehnes Attachment 2 for Taking NRC Written Examinations Cheating on the examination will result in a denial of your application and could result in more severe penaltie . Af ter you complete the examinati 1, sign the statement on the cover sheet indicating that the work is your own and you . ave not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80 percent or greate . The point value for each question is indicated in parentheses after the question numbe . There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet ano the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan . If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leav Avoid all contact with anyone outside the examination room to eliminato even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or procto Remember to sign the statement on the examination cover shee . Af ter you have turned in your examination, leave the examination area as defined by the examine . -, . - - - - - - . - . - .
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QUESTION: 001 (1.00)
Which one of the below responses correctly describes the purpose behind the Rod Control &
'information System enforcing notch limits above the Low Power Setpointt , Prevent exceeding MCPR or LHGR during a postulated control rod drop accident, Prevent exceeding MCPR or LHGR during a postulated contres rod continuous withdrawal accident. - Restrict control rod patterns to those analyzed resulting in acceptable' fuel enthalples during a Rx startu Restrict control rod patterns during a power increase to prevent exceeding the Reactivity Anomalies technical specificatio ,
QUESTION: 002 (1.00)
'6 following initial conditions on Reactor Recire Hydraulic Power Unit "A":
. ..oop 1: Operational, LEAD, READY selectod Subloop 2: READY selecte Select the condition that would cause Subloop 1 and Subloop 2 to shif t to the MAINTENANCE .
mod Oil reservoir level decreases to 60 gallons, Oil reservoir temperature increases to 145' e Hydraulic pump discharge pressure increases to 2050 psi , Fullers Earth Filter differential pressure increases to 10 psi .
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REACTOR OPERATOR Page 5 OUESTION: 003 (1.00)
Emergency Se.vice Water pump *B" has just been declared INOPERABLE due to a shaft seizure. It has been determined that RHR Loop "B" needs to be lined up in its Containment Flooding inode. Which of the following sources of water are still available to perform this operadore? Diesel Fire Pum The suppression pool, Emergency Service Water Loop "A". Emergency Closed Cooling Loop "B".
QUESTION: 004 (1.00)
RHR "A" pump was running in Suppression Pool Cooling when the Unit Supervisor requested you start ths Low Pressure Core Spray System and place it on Minimum Flow. The System Engineer will be coming up to the control room shortly to assist you in a LPCS Work Crder retest. This would not be done because it would result in: a loss of flow to the LPCS pump, a loss of flow to the RHR "A" pump, excessive load on the Div.1 Diesel Generator, check valve cycling on the common SP return lin __
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REACTOR OPERATOR Page 6 GUESTION: 005 (1.00)
The HPCS system has automatically initiated due to BOTH a high drywell pressure AND a low reactor water level. Which one of the following statements is correct rbout resetting the initiation? The initiation signal can be reset with both initiating sig.ials still present, Goth the high drywell and Icw reactor water level signals raust clear in order to tese Only the low reactor water level s!gnal must clear as the high drywell pressure c.lgnol can be lucked out by the rese Only the high drywell nressure signal must clear as the low reactor water level signal can be locked out by the riuet.
QUESTION: 006 (1.00)
During a full flow test of the RCIC system (CST to CST), a problem is encountered and the Control Room operator makes the decision to isolate the RCIC system. He arms and depresses the RCIC ' .inual isolation pushbutton. How does the RCIC System respond? The RCIC Turbine trips, The RCIC System isolates, Division 1 only, The RC: ; System isolates, Division 1 and Nothing happens, RCIC System doesn't isulat . - . - - . - . - . - . - - - .
REACTOR OPERATOR Page 7 GUESTION: 00'/ (1.00)
The unit is operating at 100% power with both RFPTs in service. Narrow Range level instrument N004A has FAILED HIGH. Feedwater level control is selected to "B" level instrument. Narrow range levelinstrument N004C has just FAILED HIGH. Wh;ch ONE of the following correctly describes the Feedpumps response? Both feedpumps will tri The feedpumps willlock in the last called for speed positio HFPT A ONLY will trip because Narrow Range level instrument "B" 15 still indicating properl The feedpumps would not be affected since levelinstrument .NOO4B is used for ir:put into Feedwater Level Contro QUESTION: 008 (1.00)
Following a small steam line break in containment, the differential pressure between the Containment and Drywell builds up to a positive .5 psid. Which of the following is the expected response of the D.ywell Vacuum Relief System? The Vacuum Relief MOV iso Valves (F010A/B) will open: but can be closed and will remain closed by taking the control switch to CLOS but will automatically close and remain closed if a BOP LOCA condition is presen but can be closed and will remain closed only if the Drywell vacuum condition clears, i only if a BOP LOCA condition is present with the Drywell vacuum condition present, REACTOR OPERATOR Page 8 QUESTION: 009 (1.00)
While operating at 100% Rx power an l&C Technician asked you to place the RWCU lsolation Bypass switch,1E31 S1 A on H13 P632, to the BYPASS position for the performance of his surveillance test. If you performed this action, what implications willit have on the RWCU system? The low RPV level 2 RWCU isolation will be bypasse ' i he RWCU isolation due to SLC initiation will be bypasse ; The High RWCU differential flow isolation will be bypasse The Non-Regenerative Heat Exchanger high outlet temperature isolation will be bypassed.
QUESTION: 010 (1.00)
With EHC control in the STANDBY mode, SELECT the system response to a main turbine OVERSPEED conditio The turbirie will not trip. Overspeed protection NOT is provided in the STANDBY mode, The BOST Unit will trip the main turbine at 105% speed, due to operating in STANDB The main turbine will trip on high MSOP discharge pressure as turbine speed increases, At 100.5% speed, the turbino control valves will start to close and be fully closed at 105.5% speed, if speed continues to increase, the intercept valves will start to close at 105.5% speed and be fully closed at 107.5% speed. If speed continues to increase, the MT will trip at 110% spee _
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OUESTION: 011 (1.00) ;
Which of the below statements is correct ' assuming that Hotwell Storage Tank is i approximately 62" during normal full power operation? , The Normal Hotwell Makeup valve would begin to open at this level, SOI N21, Condensate System, airects securing all Hotwell and Condensate l Booster pumps at this level, i The Emergency Hotwell Makeup valve would open and the Normal Hotwell '
Makeup valve would clos Dissolved oxygen concentration of the condensate in the hotwell storage tank-- .
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would most likely increase. -
OUESTION: 012 (1.00)
SELECT the statement that correctly describes the operation of the Residual Heat Removal (RHR) System, Closure of E12 F105, RHR PUMP C SUPR POOL suction valve, is an automatic pump tri Suppression Pool Test Return Valves (F024A/B) cannot be opened with a LPCI initiation signal presen A Containment Spray initiation signal will cause the RHR Pump A to start if it had been manually overrid< n while in the LPCI mod RHR Pump A and B suction volves (F004A/B) will automatically realign for the LPCI mode when an initiation signal is receive .
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I REACTOR OPERATOR Page 10 QUESTION: 013 (1.00)
While operating at 100% power, alarm window " GENERATOR CORE MONITOR P864" on panel H13+680 energizes. The following indications exist on the Generator Validation Control Panel H13 P864: MACHINE HEATINO LED is ON, and the current reading is 80%. What-IMMEDIATE action is the operator required to perform? Insufficient information is available to determine operator action Notify l&C of equipment malfunctio Within 5 minutes, remove generator from service per 101 14, Fast Unload and Trip of Main Turbine, Monitor generator hydrogen pressure. Maintain generator load within the limits of the Generator Capability Curv ,
QUESTION: 014 (1.00)
While withdrawing control rod 30 31, a " channel disagree" light comes on and a rod withdrawal block is generated, it is determined that, at the current position, there is a faulty position reed switch in channel "A". Can the operator continue to withdraw control rod 30 31, why/why not? No, both reed switches must be operable per Tech. Specs, to continue rod withdrawal.
I Yes, a data substitution must be made for the faulty reed switch using the ,
channel "B" dat ,
. Yes, the control rod is required to be bypassed at the Rod Gang Drive Cabinet,
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but can then be moved.
l~ , No, substitute data is not allowed to replace good data, and since the faulty reed switch is considered real data at that position, then substitute data is not '
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REACTOR OPERATOR Page 11 QUESTION: 015 (1.00)
SELECT the statement that describes the response of the Fire Service Water system pumps to an inadvertant actuation of the deluge system for the main transformer, The motor and diesel fire pumps both auto start when the deluge initiation signal is receive The motor fire pump auto starts at 120 psig, the diesel fire pump auto starts when pressure is less than 105 psig in the fire heade The diesel fire pump auto starts when pressure is less than 120 psig and the motor fire pump auto starts at 105 psig in the fire header, The motor fire pump auto starts by receipt of the deluge initiation signal, the diesel fire pump auto starts when pressure is less than 105 psig in the fire header.
QUESTION: 016 (1.00)
Which one of the following statements is correct concerning an uncoupled control rod during coupling checks? Up to two attempts may be made to recouple the dr!ve, If attempts to recouple the rod are unsuccessful, the rod should be disarmed at that position, Digital display of control rod position on the Rod Display Module will go blank if the rod is uncoupled, When the ROD UNCOUPLED indicating light has illuminated, it will not clear until the rod is successfully recouple _
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REACTOR OPERATOR Page 12 OUESTION: 017 (1.GO)
A Rx start-up and heat up is in progress. RPV water levelis being controlled by use of the RWCU DUMP TO MAIN CNDR STOP VALVE,1G33-F046. In order to lower RPV level at a f aster rate,it is suggested to open RWCU DUMP TO RADWASTE OTOP VALVE,1G33-F035, ,
to provide an additional dump flowpath. Why is this not advisable? ! This could cause possible runout conditions on a running RWCU pum This could cause formation of a void space and excessive water hammer to equipmen The two flowpaths are fed by a common line and thus total dump flow would not be increased, This <:ould cause a loss of condenser vacuum and possible damage to the Waste Collector Tan QUESTION: 018 (1.00)
SELECT the statement that describes the expected response of the Containment Vessel and Drywell Purge Supply (M14) Cystem following a refueling accident where an exposed fuel bundle is dropped sad is damaged. The M14 system is currently operating in the Refuelmod The M14 system isolation valves: close AND the M14 supply and exhaust fans will tri b, close which cause the M14 exhaust f ans to trip. The M14 supply f ans continue to operate, close which cause the M 14 supply f ans to trip. The M14 exhaust f aris continue to operat remain open. The M14 supply and exhaust fans trip on a containment ventilation exhaust high radiation signa __-
i REACTOR OPERATOR Page 13 OUESTION: 019 (1.00)
A plant transient is in progress. The following indications are observed:
- A Reactor Scram signal has been receive Pressure set et 940 psi No con?ol rods have inserted into the cor Reactor power is approximately 100%.
- The MSIVs are open; the main turbine is on lin Recirculation pumps are in FAS Heactor water level is 196 inche N!ch ONE of the following describes tlie consequences of tripping the reactor recirculation pumps? Assume the operators take no other actions, A fevel transient may cause a turbine trip. Bypass valves will open, and will control reactor pressure at pressure se A level transient may cause a turbine trip. Bypass valves will open, and reactor pressure willincrease until an SRV open Tripping the recirculation pumps will result in an immediate power redu: tion, with a subsequent decrease in reactor pressure and leve Tripping the recircolation pumps will rapidly incrense vessel water level, resulting in a large pcwer transient and possible fuel damag ___ . - _ . . . , .
REACTOR OPERATOR Page 14 QUESTION: 020 (1.00)
An ATWS with a MSIV closure has occurred resulting in the following plant conditions:
- Reactor power is 15%.
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Reactor pressure is 890 psi Reactor levelis stabilized at 100" with the MFP on the Startup Level Controlle Rx mode switch is in SHUTDOWN
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Containment pressure is 0.5 psi DW pressure is 2.1 psi Suppression Pool temperature is 112' Suppression Pool Level is 18.7 fee RHR Status: A/B/C have been terminated and prevente A/B are lined up for outside the shroud injection.
Why should allinjection into the RPV be terminated and prevented with the exception of boron and CRD? This terminates or reduces, es much as possible, any continued containment pressure increase, This ensures that if Emergency Depressurization !s required later, the heat rejected to the containment will be minimized, This ensures that an Emergency Dapressurization willnot resultin exceeding the suppression chamber design temperatur ECCS pumps which take suction on the Suppression Pool could lose adequate NPSH and containment integrity could be los . ~ _ _
REACTOR OPERATOR Page 15 QUESTION: 021 (1.00)
The reactor was operating at 100% power with HPCS OOS. A trip of the Feed System resulted in a level 3 scram. The RCIC system auto started but RPV water level decreased to 10" before the Motor F eed Pump was manually started. The Motor Feed Pamp is being used to restore level. With RPV level under control the Unit Supervisor has directed you to maintain reactor pressure 800# 900#. How are you going to maintain this pressure band? Manually cycle Safety Relief Valves in the order listed on H13-P601 to maintain this pressure band, Lower the pressure set setpoint down to a maximum of 900# which will allow for automatic pre:sure control, Manually throttle the output of the Bypass Valve Jack to use Bypass Valves to maintain this pressure ban Leave the Steam Bypass and Pressure Control sys, tem as-is and it will automatically control pressure in this Land.
QUESTION: 022 (1.00)
The plant is operating at 50% power with the electric plant in its normal operating lineup and all divisional and non-divisional batteries being supplied by their normal chtrgers. Bus L11 suddenly experiences a bus lockout. What effect is there on both the divisional and non-divisional DC systems? No effects, tSe normal chargers will continue to supply their DC loads and batterie Both non-divisional DC systems will be supplied by their batteries, all divisional DC systems will be unaffected, The non-divisional 1B DC system will switch to its alternate charger, the non-divisional 1 A DC system will be supplied by its battery, and the divisional DC systems will be unaffected, The non-divisional 1 A DC systern will switch to its alternate charger, the non-divisional 18 DC system will be supplied by its battery, and the divisional DC systems will be unaffecte .~ . ... ,. .. . . -, - - - - - . . - - -
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QUESTION: 023'(1.00)-
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Given the following plant conditions:
Rx Power- -1100% .
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RFPT A & B on the Master Level Controlle Narrow Range Channel _"A" is selected. for input into C3 ' A loss of bus D1B occurs. If no operator action was taken, what would be ihe impact on RPV level control? - ,
.a.- RP_V level will rapidiv increase due to partial loss of RPV level and feedwater
- flow signal ' RPV level will increase then stab!lize near the level setpoint as RFPT "B" speed ' ;
increases, RFPT B speed initially increases then decreases as level error overcomes the flow error signa F Signal to LOW FLOW RX LEVEL CONTROL,1C34- R614, fails causing 1N27-F175 to ramp closed if in AUT QUESTION: 024 (1.00)
Regarding the Remote Shutdown System (C61), all Division 1 Remote Shutdown Transfer-Switches are in their NORMAL position. MSIV pilot solenoid "B" control switch is in the
- NORMAL position. -Which ONE of the following describes the response of the inboard MSlVs if the operator takes MSIV pilot solenoid "A" control switch.to the close position and the reason for this response? Inboard MSIVs close because the "A" pilot solenoids for the MSIVs de-energize.
" Inboard MSIVs remain open because only the "A" pilot solenoids for the MSIVs de energized, c, inboard MSlVs close because this switch energizes one pilot solenoid for each MSIV with 125 VDC.
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d.- Inboard MSIVs remain open because control has not been transferred to the
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REACTOR OPERATOR ' Page 17 OUESTION: 025 (1.00)
PEl T23, Containment Temperature Control, requires the operator to Emergency Depressurize the RPV prior to exceeding 185'F in the Containment. This is performed to: Prevent Primary Containment f ailure due to high temperatur Keep levelindication from fleshing due to reference leg heating, To further minimize heat input into the containment from the reactor vesse To prevent a negative pressure in containment prior to containment spray initiatio QUESTION: 026 (1.00)
Which one of the following s+atements correctly describes a method that would initiate the Suppression Pool Makeup System assuming that Suppression Pool tevelis currently at 1 feet and decreasing and no LOCA signal is present?
. Arm & depress the LPCS / LPCI "A" Manual Initiation pushbutton to insert a LOCA signa Place the SPMU "A" Logic switch in OFF, then arm & depress the SPMU "A" Manual Initiation pushbutto Place the SPMU "A" Full Flow Test Perm. switch in TEST, then arm & depress the SPMU "A" Manual Initiation pushbutto Place the SPMU "A" Logic switch in OFF, open SPMU First Shutoff 1G43-F030A, then open SPMU Second Shutoff 1G43-F040 e ..
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- QUESTIONi'027 (1.00)
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- A LOCA has occurred and the Emergency Coordinator has just declared the site to be in ar)- 3
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persor.nel, The Unit Supervisor must enter PEl-D17, Radioactivity Releaseilf not already in i The Emergency Coordinator should make Protective Action Recommendations per EPI-B8, to the State and County Authorities, The Unit Supervisor should place both trains of Control Room- HVAC in Emergency Recire per ON1 D17, High Rad Levels Within Plan QUESTION: 028 (1.00)
PEl B13, RPV Control (Non-ATWS), and PEl N11, Containment Leakage Control, have both been entered.- SELECT which of this following conditions would require entry into PEl-B13, Emergency Depressurization?
a, the feed system is discharging into the steam tunnel, all the affected area water levels cannot be restored and maintained less than the entry _ condition values, c, with no primary system discharging into the area, the' HPCS pump room is 142*F, and the LPCS pump room is 148" with the feed system discharging into the steam tunnel, the HPCS pump room is 143*F and the steam tunnel is 313 ,- ._ _
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REACTOR OPERATOR Page 19 QUESTION: 029 (1.00)
The plant is in a startup and the RCIC system is being restored to Standby Readiness. Reactor pressure is 100 psig. You are performing the warmup of the RCIC system and reach a step in SOI-E51, Reactor Core Isolation Cooling System, to throttle open the E51-F076, Steam Warmup Valve, and you find the valve already FULL OPEN, You should: proceed with the procedure, since the valve is already open and the intent of the steam line warmup is accomplishe proceed with the procedure af ter notifying your Supervisor who will change the valve position to open on the verificat!on checklist and initiate a non-intent temporary change, stop the procedure and obtain your Supervisor's permission to annotate the step FIC and continue the instructio stop the procedure and obtain your Supervisor's permisolon to appropriately position the valve and document action taken in the Unit log.
QUEST!ON: 030 (1.00)
The 'B' Turbine Building Closed Cooling (TBCC) Pump is RED TAGGED OUT of service for maintenance. Aloss of the 'A'TBCC pump has occurred and a plant operator reports that the
'B' pump looks like it is fully reassembled and no one is working on it. The Maintenance Supervisor, who is the person in charge on the tegout, has left the site for lunch. Assuming all are Person-in-Charge qualified, which of the following can authorize tag removal? The Unit Supervisoi The lead mechanic on the crew The responsible system engineer The operations foreman who walked down the "B" TBCC Pump with the plant operator
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OUESTION: 031 (1.00)
John Smith, age 33,is a radiation worker at Perry. John's current calender quarter radiation exposure is 250 millirem TEDE, all received at Perry Station. John was assigned to Clinton station earlier this year and received 125 millirem TEDE. John was also given two chest-x-rays during July 1997 for an additional O.050 millirem. How much dose is John allowed to receive during this calendar year before he is required to get supervisor's permisslor) to receive additional dose?- .95 millirem millirem .95 millirem millirem QUESTION: 032 (1.00)
An emergency drywell entry is required, Rx power is 7% MWt with the main generator off-line. Which of the below statements is correct with respect to waiving the ALARA review for the drywell entry? An ALARA review is not required for a drywell entry under these plant conditions per PAP 0118, ALARA Program, An ALARA review is required for work activities under these plant conditions and may not be waived per PAP- 0118, ALARA Progra The Unit Supervisor may waive the ALARA review by declaring this an urgent situation per PAP-0512, Radiation Work Permit Program, The Shift Supervisor may waive the ALARA review by declaring this a Priority 1 Emergency per PAP-0902, Work Request Syste _: ~ 4 _m_ .
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- QUESTION
- 033 (1.00)
You have been assigned to escort several visitors for a plant tour that willinclude the contial
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room. You should: process through the turnstiles ahead of.your visitor when entering and exiting the Protected Are ensure that no more than 10 individuals are escorted in the Protected Areas, ,
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and no more than 5 individuals lo the vital area , use your key card first when traveling through key card controlled doors, then hold the door open while your visitors use their issued key cards.-
d- keep escorted indivicials under con'tinuous observation and report any indications of aberrant behavior following your visitor's exit from the plan '
QUESTION: 034 (1.00)
PAP-0504, Electrical Operating Rules and Practices, requires that an evaluation and pre-job inspection be done whenever working on energized equipment. Many of the PEl-SPIs invol*<e use of lifted leads.and jumpers in energized panels, however an evaluation and pre-job inspection is not done*by operators performing PEl-SPIs. How is this in compliance with PAP-0504? The inspection / evaluation was done once and is applicable for all PEl SPI The inspection / evaluation is not required due to the low voltage in the associated panels, PEl SPIs are done in what is considered a " plant emergency" and therefore the requirement for an inspection / evaluation is waive This requirement is only applicable when 'it is not prudent to use safety equipment, however, safety equipment is used during the performance of PEl SPI r
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QUESTION: 035 (1.00)
PAP-0110, Shift Staffing anc. Overtime, states that the Fire Brigade must be on site per PAP-1910, Fire Protection Program. Which of the following correctly states the requirements for the site Fire Brigade? The Fire Brigade is a five person team consisting a Fire Brigade Leader, the Fire Engineer, and three Fire Brigade members, If less than the minimum number of Fire Brigade members is on-site, the minimum number will be restored within four hour The on-shift Shift Supervisor shall not be assigned to the site Fire Brigade, and the SO at the-controls should not be assigned as the Fire Brigade Leade With the exception of the Fire Brigade Leader, who is a licensed operator, Fire Brigade members will have no duties during a fire except those directly related to manual fire fighting.
. QUESTION: 036 (1.00)
Which ONE of the following statements conforms to the recommended good operating practices listed in PAP-0201, Conduct of Operations? Disconnect switches or breakers opened as a result of sol or ell performance must be DANGER tagged, The Unit Supervisor should have a nuisance alarm disabled if the wc,rk to correct the problem cannot be completed in a reasonable time, defined as one hour, To prevent damage of remotely operated throttle valves when closing them, the control switch should be released as soon as the CLOSED indication is receive Whenever possible, place systems having an automatic actuation feature in a lineup that prevents an automatic- actuation, prior to performing system troubleshooting or repai >-
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OUESTION: 037 (1.00)
Which of the following defines the requirements for the licensed operator At The Controls (ATC Operator) to leave the horseshoe area of the control room without relief? The ATC Operator may leave the horseshoe area of the control room: only with a proper watch relief, any time a Supervising Operator is within the horseshoe area to verify receipt of an annunciator during an emer:!ency that affects safe operation of the plan to retrieve a system description manual when an emergency has occurred affecting safe operation of the plan QUESTION: 038 (1.00)
SELECT the correct action to be taken when taking logs if the Containment Humidity instrument is out of service for preventative maintenance which willcontinue for several day Place an "X" in the data block and explain in the remarks section, Record the out of servico instrument reading and explain in remarks section, Obtain the Unit Supervisor's permission to use alternate instxmentation, asterisk in data block, explain in remarks section, Wait until the preventative maintenance has been completed on the instrument-and then record the instrument reading in the rounds, i
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OUESTION: 039 (1.00)_ _
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Given the following' Rx cooldown data:
TIME TEMPERATURE *F TIME TEMPERATURE'F ;
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0800- -520- 100 '
- 0815 490' 1015 .310 0830 460 1030 275 0845 445 1045 250 0900 430 1100 225- !
0915 400 1115 205 0930 370 +
0945- -340 Choose the correct response from below with regard to the results of the cooldow The cooldown was acceptable and no limits were exceede . The cooldown rate was exceeded during only one one hour perio The cooldown rate was exceeded during 'only two one hour period d The cooldown rate was exceeded during only three one hour periods, d.
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REACTOR OPERATOR Page 25 OUESTION: 040 (1.00)
The following plant conditions exist: ECC pump 'A' is running, ECC pump 'B'is in STANDBY-READINESS, ECC Area Cooling Fan 'A'is RED TAGGED OUT, ECC Area Cooling Fan 'B' is in STANDBY READINESS, Control Complex 'B' Chiller & Chilled Water Pump are operating.
As the operator at-the-controtti, which one of the following sctions should you take? Start ECC pump 'B' from the control roo Have the Nuclear Island roun:'s taker start ECC Area Cooling Fan 'B'. ECC pump "A" must be secured due to no Area Cooling availabl Have Maintenance personnel install temporary ventilation directed at ECC pump 'A'.
QUESTION: 041 (1.00)
There are certain responsibilities that can only be performed by the Emergency Coordinator when the Site Emergency Plan is being implemented. Which one of the following actions is allowed to be performed by someone other than the Emergency Coordinator? Direct the notification of offsite agencies and organization Determine the emergency classification including reclassification or terminatio Recommend protective actions for the general public to State and local County Official Coordinate and direct the actions necessary to terminate or mitigate the effects of the emergenc REACTOR OPERATOR Page 26 OUESTION: 042 (1.00)
The plant is operating at 100% power witn all of the "A" pumps operating when bus EH11 is suddenly de-energized. The Div.1 DG auto starts, but does not tie into bus EH11 Which ONE of the following responses would you expect to see in this situation?
. CRD Mechanism High Temperature alarm on P60 Rx Recirc pump A willlose al! t,aal cooling flo RHR Pump A will auto start on a LOOP sigual, and run on min. flo Div.1 DG will run continuously unloaded due te auto bypass of the DG trips on a LOOP signal.
QUESTION: 043 (1.00)
In which one of the following conditions will the RC&lS allow control rod movement? Rods in Group 5 may each be withdrawn from 00 to 12 prior to selecting the next ro Groups 3 and 4 are fully withdrawn, and a control rod in Group 1 is selecte An attempt is made tc. fully withdraw the rod, Groups 1 through 4 are fully withdrawn with the exception of one rod left at position 44. An attempt to withdraw a rod in Group 6 is made, Groups 1 through 4 are fully withdrawn. Groups 7 and 8 are withdrawn to position 12. Groups 5 and 6 are stillinserted.' An attempt is made to withdraw a rod in Group ,
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Page 27 OUESTION: 044 (1.00)
Due to a xenon transient in the core, it becomes necessary to increase Rx power using recirc flow to maintain 1250 MWe. In accordance with Operations Policy 2-9, Operational Activity Evaluation, how should this evolution be classified by the Shift Supervisor? As risk significan As a risk contributo As an unplanned change in reactivity per ONI-C5 As an Infrequently Performed Test / Evolution (IPTE).
QUESTION: 045 (1,00)
Select from below the conditions that would result in the "CNTMT Spray B Start Signal Received" annunciator, LPCI "A" automatic LOCA start signal sealed-in for 10 minutes, drywell pressure is 3.0 psig, and containment pressure is 9.0 psig, LPCI "A" & "C" automatic LOCA start signal sealed-in for 15 minutes, drywell pressure is 2.5 psig, and containment pressure is 7.0 psig, Cntmt Spray "B" Manual Initiation Pushbutton armed and depressed for 45 seconds, drywell pressure is 1.5 psig, and containment pressure is 10 psi LPCI "B" and "C" automatic LOCA start signal sealed-in for 12 minutes, drywell pressure is 2.5 psig, and the Cntmt Spray "B" Manual Initiation Pushbutton is armed and depressed for 40 second . .- __ _ _ - _ . . . . . _
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. QUESTION: 046 (1.00) -
An operator reports that the "A" Standby. Liquid Control (SLC) squib valve continuity meter
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f indicates 5 mA of current. SELECT the action that should be take I Do nothing, since 5 mA is a normal indicatio . . . i Declare the squib valve inoperable 'since 5 mA indicates the squib valve has
. fire * Declare the squib volve inoperable since 5 'mA indicates current is too low and the squib valve may not fire, Declara the squib valve inoperable since 5 mA indicates current is too high and
- . the ignitor element may have decompose QUESTION
- 047 (1.00)
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During power operation with the Mode Switch in RUN, an event has resulted in the following conditions:
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RPV level decreases to Level RPV pressure peaked at 1113 psi Drywell pressure has reached 1.2 psi Which ONE of the following lists the expected status of the scram, backup scram and ARI
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valve solenoids. No operator actions have been take . SCRAM - BACKUP SCRAM ARI deenergized deenergized deenergized deenergized _' energized deenergized deenergized energized . energized energized . deenergized energized
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QUESTION: 048.(1.00) ;
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- A transient occurs and it becomes apparent that the RPS "B" MG Set has been lost. Which .
bus normally powers this equipment? i F1808 F1C08 l F1C12 F1D12'
QUESTION: 049-(1.00)
A reactor startup is in progress with the Rx Mode ' Switch in "STARTUP/ STANDBY". The - ,
following is tha present status of the APRMs versus LPRM inputs, and indicated power:
APRM A- B C D E F G H D Level inputs: 4 5 4~ 3 4 4 6 6 C Level inputs:- 4 3 3 4 6 2 4 4~
B Level Inputs: 3 4 3 4 4 4 6 4 A Level inputs: 3 3 3 4 6 4 4 2 indicated Power: 11 % 16 % 12% 11 % 12% 10% 12% 10 %
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- SELECT the correct RPS response to the above data: No trips
~ Rod block only
, Half scram only . Full scram -
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- 050 (1.00).
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A discrepancy is noted between RPV levelinstruments where Wide Range Level "B" is reading lower than the others Which ONE of the following conditions could cause this channel to read lower than actual vessel level? > Drywell temperature locally at WR -"B" increase . b .- . Drywell temperature locally et WR "B" decrease c. - WR "C" level transmitter equalizing vsive leak # A steam leak occurs at the condensing pot for WR "B" levelinstrument.' ;
QUESTION: 051 (1.00)
A plant ' transient occurred such that the Automatic Depressurization System (ADS)
automatically initiated. RPV level is steady at 10", all rods are inserted, and RPV pressure is
'25 psig. Which ONE of the following actions would result in ADS logic being reset? With RPV level steady at 10", ADS logic cannot be reset, Increase RPV water level to 20", then place the ADS Logic Inhibit switches in INHIBI Placing the ADS Logic inhibit switches in INHIBIT, then depressing the ADS Logic Seal-in Reset pushbuttons, Depress the ADS Logic Sealin Reset pushbuttons, then take each ADS Valve Control Switch to the OFF position.
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s OUESTION: 052 (1.00)
in order to reduce the number of SRVs that reopen following a reactor isolation event, the Low Low Set function was added to several SRVs. Which ONE of the following statements; correctly describe this function? At 1103 psig, F051D opens, arming LLS and 1821-F051C opens. IF pressure
' increases to 1113 psig, then four LLS valves open' and close at 946 psi FOSIC. cycles between 1103 and 936 psig and F0510 cycles between 1033
. psig and 926 psi > At 1103 psig,' F051D opens, arming LLS at 1103 psig, F051C opens; and at
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1123 psig, the other four LLS valves open, and then close at 946 psig. F051C cycles between 1103 psig and 930 psig, and F051 D cycles between 1033 psig .
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and 926 psig.- At 1103 psig, F051D opens, arming LLS and opening F051C. If pressure increases to 1113 psig, the other four LLS valves open and cycle between 1113 psig and 946 psig. F051C cycles between 1073 psig and 936 psig, and F051 D cycles between 1033 psig and 926 psi At 1103 psig, F051D opens, arming LLS, if pressure increases to 1113 psig, .
four LLS valves will open, arming F051C, which will then open and cycle between 1077 psig and 936 psig The four LLS valves will cycle between
.1113 and 946 psig ~ F051D will cycle between 1033 and 926 psig.
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OUESTION: 053;(1.00):
- Reacto.' power is approximately 40%, recirc pumps have been shifted to fast speed, and a plant startup is in progress. Which of the following statements correctly describes the plant
' resposise following a downscale failure of the in-service Main Steamline pressure regulator? -
a .~ Rx pressure willincrease causing bypass valves to open and Rx pressure will-return to the pre-transient value, A slight pressure transient may occur, however pressure will remain relatively constant throughout this even Pressure will increase, bypass _ valves will open in an attempt to reduce pressure, however the reactor will scram, Pressure willincrease to the point where both the scram and RRCS setpoints for Rx pressure will be reache QUESTION: 054 (1.00)
, Both RFPTs are running, being controlled by the Master Level Controllet. The MFP control switch is in OFF. The Startup Level Controller Select Switch is selected to RFPT "A". What -
would be the resultant pump combination if RFPT "B" tripped, with no operator actions being taken? RFPT "A" would be running alone on the ML RFPT "A" would be running alone on the SUL The MFP would start and be controlled on the MLC with RFPT "A".
- The MFP would start and be controlled on the SULC with RFPT ' A".
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QUESTION: 055-(1.00)- "
The plant is operating at 70% power with RFPT *B* and the MFP on the master lovel r controller, RFPT "A" is danger tagged OOS. - Which of the following statements correctl describes the response of the feedwater level control system to the trip of RFPT "B"?.
e The level demand signal is that signal corresponding to the tapeset of the
- master level controller and will remain so continuousl/. The level demand signal is that signal corresponding to'the tapeset of the startup level cuntroller and will remain so continuously, i i
c .- For 10 seconds the level demand signal is that signal corresponding to the tapeset of the master level controller, it then drops to a level domand signal of approximately 178". For 10 seconds the level demand signal is that signal corresponding to the tapeset of the startup level controller, it then drops to a level demand signal of approximately 178".
QUESTION: 056 (1.00)-
Assuming that the plant is operating at 100% Rx power with AEGTS Fan "A" running, which ONE of the following conditions would provide an automatic start of the "B" Annulus Exhaust Gas Treatment System? Annulus pressure sensed as being -0.50" H2 Low flow sensed on the discharge of the "A" AEGTS fan,
. De energizing the Division 2 RHR LOCA Relay 1E12- K1100, Arm and depress the High Pressure Core Spray Manual initiation pushbutto .
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LOOESTION: 0571(1.00)
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- The following plant conditions exist
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- 8x power - -100 % .
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Busses L11;& L12 are being pc'ered from the__ Unit 1_ Aux. Transformer __
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- Bus L10 is being supplied by the Unit 1_ Startup Transformer -- "
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Div 1 & Div 3 DGs are in standby readiness ' '.;
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Div 2 DG 1s in " LOCAL" running fully loaded for_ a WO retest
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Preferred Source ' Breakers for EH11, EH12, and EH13 are all close '
With these plant conditions, a bus L10 lockout occurs, a scram sigitalis generated due to the trip of both Rx Recirc pumps and level 2 is reachea. After auto initiat'ng, the_ HF CS pump - .
motor catches. fire and causes a lockout to occur on bus EH 13. The lowest RPV level-
.' achieved ouring this transient is 100". Wh;ch of the following statements correctly describes the steady-state conditions you would expect to see following thin transient?
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, _Div.1 DG is the only DG running AND supplying its respective bu Only Div.1 & 3 DGs are running, supplying their respective busse ' _
' Div. 2 DG will auto start, but not load when its local / remote switch is taken out of LOCAL. ; All three DGs received an auto start on a LOOP signal and are supplying their respective busses.
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- 058: (1.00i 1; The plant is r.perating at 90% powcr and stable. The following alarms are suddenly _ received; Let H13 P680:= ~
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1 ~ RPV i.evel 7 alarm'
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Stm Flow / Feed Flow Mismatch alarm - ,
s !P601 alarm
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, -. l The plant stabilizes at approximately 92% power, with level at 196"i Choose from below the correct response for this transient;-
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a,- ' Secure'the HPCS pump... , Reset the ADS Initiation logic. .
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- Manually trip the RCIC turbino, Take the SRV control switch to OF .
QUESTION: 059 (1.00)
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-Assume that the plant is in Mode 5 with the Rx Mode Switch in REFUEL, fuel movement is occurring in the containment, and the shorting links are removed. Which of the following would result in a Rx scram?
. SRM Channel A fails downscals.
. - - - SRM Channel B fails upscale.
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OUESTION: 000 (1.00)
= The reactor is at 72% power and 65% core flow. The mismatch between recirculation loops ;
must not exceed: j l %'of rated recirculation flo q % of rated recirculation flo l c.- 10% of rated recirculation flo . % of rated recirculation flow.
QUESTION: 061 (1.00)
The plant is operating at 100% power and the only equipment that is OOS is SLC pump "A".
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/An -ATWS occurs and you are directed by the Unit Supervisor to initiato SLC "B". Upon initiating SLC "B", which of the fol'owing statements correctly describe what you would expect to occur? RWCU valves will only isolate if both SLC pumps are started, RWCU will continue to run, G33 F001, RWCU Suction Containment inboard Isolation Valve will close, RWCU pumps will tri G33 F004, RWCU Suction Containment Outboard Isolation. Valve will close, RWCU pumps will tri G33-F001 RWCU suction containment inboard isolation valve AND G33-F004, RWCU suction containment outboard isolation valve will close, RWCU pumps will_ trip.
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QUESTION: 062.(1.00)2
-_ The following plant conditions _ exist:
- =- The plant is in MODE Reactor Recirculation (RR) Pump "A" is secured for maintenanc Reactor Recirculation (RR) Pump "B" is in operatio ' Residual Heat Removal (RHR) "A" is in standby readines Residual Heat Removal (RHR) "B" is operating in shutdown coolin Reactor coolant temperature is 120* ;
- RHR "B" pump is required to be shutdown for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for a scheduled motor operated valv ,
stroking surveillance. RHR "B" will be returned.to shutdown cooling following the surveillanc From the following, choose the action required in- accordance with Perry Technical Specification Demonstrate the operability of at least one alternate method of decac heat-removal within one hou Place the "A" loop of RHR in the Shutdown Cooling mode prior to removing the
"B" loop of RHR from servic None, one RHR shutdown cooling subsystem may be inoperable for up to 2
- hours for the performance of a Surveillance, Verify reactor coolant circulating by aii alternate method in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and monitor
- reactor coolant temperature and pressure, f
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REACTOR OPERATOR Page 38 QUESTION: 063 (1.00)
Which ONE of the following statements is true regarding operation of Residual Heat Removal (RHR)in the Suppression Pool Cooling Modo? When 1E12-F024A(B) is open, system flow through 1_E12 F024A(B) shall be maintained between 2000 and 7300 gp The RHR system shall be declared inoperable whenever it is in a secondary mode of operation EXCEPT when in Suppression Pool Coolin RHR Loop "B" shall not be operated in suppression pool cooling when SPCU is in operation through suppression pool return line bypassing RHR "A". A loss of pumping power during any operation with RHR A(B) TEST VALVE TO SUPR. POOL,1E12-F024A(B), open will cause voids to be drawn in the higher elevations of the affected loop.
QUESTION: 064 (1.00)
Plant Conditions:
- Containment sprays have been running (RHR A and RHR B) for 15 minute Containment pressure is 1.5 psi Drywell pressure is 2.0 psi Reactor pressure is 50 psig.
Which of the following plant responses describes what you would expect to occur if an operator depresses the CNTMT SPRAY A(B) SEAL-IN RESET pushbuttons? Nothing, containment sprays continue, Containment sprays secure AND the RHR pumps secur Containment sprays secure AND the LPCI Mode of HHR A and RHR B initiat The RHR pumps continue to run, supplying flow to the containment spray header AND the LPCIinjection flowpat .<-, n ... - -. . - - -. -
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- QUESTION
- 065-(1,00)_ --
j j _ - Given the following plant' conditions:
-:- - A lockout has occuned on LH-1 B _and tripped LH 1- V Breaker H1101, Bus H11. Normal Supply Breaker, is opening.- -
Which ONE of the following sets of conditions must be satisfied in order for breaker H1102, ..
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3 Bus H11 Alternate Supply Breaker, to automatically _ close?- -
a.;- No lockout on bus H1.1 and no lockout on the Auxiliary transforme b. .. No lockout on bus H11 and.no lockout on Interbus Transformer LH 1' C.- - The ATS must be in OFF and normal voltage on the line side of breaker H110 ,
i L The ATS must be in AUTO, no lockout on bus H11 and normal voltage on the line side'of breaker H110 .
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QUESTION: 066 (1.00)
.Tne normal power supply to the plant vita! inverter is bus: _D1 b, D1B . ED1 A_
' ED1B
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QUESTION: 067 (1.00)
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Which of the following signals will cause the Off Gas Discharge Valve (N64-F632) to isolate? Post treat rad monitor A is reading 0, post treat rad monitor B is alarming Hi Hl.- Post treat rad monitor A is alarming Hi-Hi, post treat rad monitor B is alarming ,
Hi-H Post treat rad monitor A is downscale, post treat rad monitor B is alarming - -
, Hi-Hi-H . . w J Post treat rad monitor A is alarming Hi-Hi-Hi, post-treat rad monitor B is alarming Hi-H QUESTION: 068 (1.00)
- While operating at 100% power, is it possible to vedly the HIGH alarm setpoint on the Turbine Building West Area Radiation Monitor without actuating the INal red warning light?. No, depressing the Alarm Trip / Test button will peg the ratemeter and actuate the local light, Yes, placing the Function Selector Switch in ALARM, and depressing the High Alarm button will do this,
, , No, depressing the Fail / Check Source button will bring in the High Alarm and actuate the local light.
- _ Yes, depressing the Horn Silence button while depressing the High Alarm button will do thi !
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REACTOR OPERATOR Page 41 ;
a OUESTION: 069 (1.00)
Perry was operating at 100% power when a Control Room HVAC high radiation condition was sensed in the sup,c!y duct. SELECT the correct response of the system cornponents listed below to this conditio Supply Return Supply Exhaust Return Fan Fan Damper Damper Damper C001 C002 F010 F130 F110 Run Stopped Open Open Closed Run- Stopped Closed Closed Closed Stopped Stopped Closed Open Closed Run Run Open Closed Open QUESTION: 070 '1.00)
While doing a building inspection, a steam leak is discovered by one of the PPOs on feedwater heater #6A. Which of the below statements describes a possible source of this steam? Heater 6A steam drains to the 2A MSR Drain Tan Heater 6A steam drains to the #4 direct contact feedwater heater.
. High pressure turbine 4th stage extraction steam supply to heater 6 High pressure turbine 7th stage extraction steam supply to heater 6 . . . . . . -,. .- - ._ - _ -
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OUESTION: 071' _ (1.00)- 1,
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' A main turbine shutdown is;in progress. . Generator load was approximately 300 MwE when '
the RO at the controls noticed that main condenser vacuum was 5.1"_ HgA.- Which of the
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below statements is true regarding these plant conditions?.
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a.- Operation may continue as long as vacuum does not increase to greater than
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- 5.5" HgA. - [
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b.: Operation under these conditions = is not allowed, perform - a fast reactor shutdow . c. - - Operation under these conditions is not allowed, trip the mainiturbine immediately, . Vacuum should be lowered to less than 5" HgA but main turbine operation may continue as long as' generator load does not fall below 275 Mw '
QUESTION: 07; (1.00)
SELECT the statement that describes the response of the Fuel Pool Cooling and Cleanup (G41)
._ System as a DIRECT result of reactor water level decreasing to below Level The G41 containment isolation valves willisolate (G41- F100, F140, and F145). * The standby G41 pump will automatically start if the control switch is in AUT The G41 filter demin bypass valve (G41-F360) will fail open and fully bypass the tilter demin, . The G41 filter demins willisolate by closing G41-F280, F285, F290, and F295
- (Filter demin isc% tion valves).--
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QUESTION: 073 {1.00)
SELECT the statement that describes the response of the Inboard Main Steam isolation Valve !
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Leakage Controf System (MSIV LCS)to reactor pressure increasing to 25 psig af ter the system has been operating for one hour, The E32.F003 valve for each subsystem receives an open signal returning the i subsystems to the bleed-off mod !
! The heaters for each subsystem trip, the inboard blower trips, and the i E32 F003 valve for each subsystem receives an open signal, , The E32 F001, E32 F002, and E32 F003 valves for each subsystem receive l close signals, the cubsystem heaters trip, and the inboard blower trip The heaters for each subsystem trip and the E32 F001, E32 F002, and .,
E32 F003 valves for. each subsystem receive an open signal returning the subsystems to the bleert off mod ;
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Q-UESTION: 074 (1.00)
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Which of the following is a violation of Technical Specifications safety limits? ; Reactor steam dome pressure is 1300 psi Core flow is 13% AND reactor power lo 205
- Reactor steam dome pressure is 760 psig AND reactor power is 18% Reactor steam dome pressure is 1020 psig AND core flow is 60% AND MCFR is 1.05, i
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REACTOR OPERATOR Page 44 OUESTION: 075 (1.00)
Due to a level control problem on Moisture Separator Reheater 2A, a f ast unload of the main turHoe was in progress esing 10114, Fast Unload and Trip of Main Turbine. At 50% reactor power, a turbine trip signal was generated. Which of the following statements correctly diceribes the plant response to this transient? A reactor scram will occur due to: a pressure increase up to 1065 psi the closure of turbine stop and control volve level 3 following the collapse of voids, high neutron flux following the collapse of volds.
QUESTION: 076 (1.00)
A Rx Scram has occurred with both Reactor Feed Pump Turbines (RFPTs) operating. RFPT A's flow controller is in MANUAL and RFPT B's flow controller is in AUTOMATIC. The Startup t.evel Control Select switch is selected to "MFP". RPV pressure is 825 psig and decreasing at approximately 2 psig per minute. Which ONE of the following describes the plant response to SLOWLY decreasing the speed on RFPT "A" dawn to 1100 rpm with the intent of removing it from service? RPV level will... be held fairly nonstant at approximately 178" by RFPT "B". decrease as RFPT "A" is removed from service. unless the MFP is starte be held constant for approximately 10 minutes and then increase to Level 8, tripping both RFPT be held fairly constant at the tapeset level selected on the Master Level Controller by RFPT "B".
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OUESTION:.. 077 (1.00)
The reactor is operating at BS% power when the following Indications occur: j
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. APRM A/E UPSCALE annunciator energizes [
. ROD BLOCK APRM UPSCALE annunciator energizes ,
- RPV pressure increasing l
- Mwe decreasing .
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= Which ONE of the following transients could result in the above? l l
a.- Maximum combined ficw limit f ails hig j I
~ "A" pressure reguictor falls open, then one minute later, the 'B" pressure regulator slowly falls r;lose !
' "A" pressure regulator fails closed, then one minute later, the "B" pressure
regulator slowly falls ope i During the power increase from 65% to 05% Rx power, the turbine Load Set ,
setpoint was never increase !
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REACTOR OPERATOR Page 46 ;
i OUESTION: 078 (1.00) t
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Following a Rx scram on RPV level 3, the P680 operator is directed by the Unit Supervisor to restor _e RPV water level between 185" 215" using the feed system. While restoring level, the P680 operator notes level at 140" and notices that Rx Recirc pump A is still running in f ast ,
speed. . What is the affect on the reactor coolant system of the failure of the pump to '
downshift?
' Tripping one recirc pump will add sufficient negative reactivity to overcome the -
positive reactivity that would have been added had the main turbine tripped at ,
full power at the end of core life, t Insufficient waterleveland/or subcooling of the waterinthe downcomer region may have caused cavitation to occur somewhere in the recirc "A" syste , With level below the height required for natural circulation there is insufficient
- mixing due to forced circulation such that significant thermal gradients would form, It can be assumed that Rx Recirc pump "A" would have tripped while at power had the delta temperature between the steam dome and the bottom head drain ,
exceeded 100* i QUESTION: 079 (1.00)
A main turbine trip occurred at 45% Rx power, but due to a malfunction of RPS, a Rx Scram failed to occur Rx pressure increased to 1083 psig. SELECT from below the ONE statement that correctly describes the Redundant Reactivity Control System response to this transient 25 seconds after reaching 1083 psig, assuming Rx power remains constant at 45% Alternate rod insertion will actuat The reactor recirculation pumps will trip off, Standby Liquid Control will automatically initiat Feedwater flow is reduced to zero, unless the feedwater pump controllers are in manua *
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REACTOR OPERATOR Page 47 OUESTION: 080 (1.00)
A reactor startup was in progress. Power was approximately 2500 cps on all four source range monitors when a malfunction in RC&lS occurred, resulting in a continuous withdrawal of a four rod gang, and causing a Rx period of 30 seconds. Which of the following statements correctly describes the implications of this event? This is an indication of excess positive reactivity and the operator should insert a manual Rx scram, Rx period will eventually return to between 60150 seconds and therefore no operator action is necessar This is the design basis event behind the Rod Pattern Controller's banked position withdraw sequence (BPWS). This event is more significant if power is initially at the point of adding heat than if power is initially in the source range.
QUESTION: 081 (1.00)
Given the following plant conditions:
- An ATWS has occurre Rx power is steady at 23E
- The Main Turbine has trippe Several control rods are stuck out at various position The Unit Supervisor directs you to insert control rods using PEl SPI, Section 1.
Why is it necessary to bypass the Low Power Setpoint? To bypass the two notch limit, allowing continuous insertion of control rods, To bypass the four notch limit, allowing continuous insertion of control rods, To bypass the bank limits that are in effect because power is being sensed l below the LPS Though not actually needed now, power willeventually decrease to LPSP during !
rod insettlun, requiring the bypas :
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i QUESTION: 082 (1.00)'
Sometime af ter a smailloss of coolant accident, the f ollowing stable plant conditions are found - :
to exist *
- Reactor Power 0%
-- - Reactor Level 180"
- Reactor Pressure 1000 psig !
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- Containment Pressure 4.5 psig
- Drywell Pressure 4.3 psig i
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- Suppression Pool Level 21 f Suppression Pool Temperature 115 ' Which one of the following is being exceeded? !
i f Heat Capacity Limit ! SRV Tailpipe Level Limit
, Pressure Suppression Pressure No PEl SPl curves are being exceede T QUESTION: 083 (1.00)
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The Rx is operating at 100% power when a spurious Main Turbine trip results in a Rx pressure excursion up to 1100 psig. What is the impact on the Feedwater Level Control System? Operation of SULC in MANUAL is available after 30 second i b.- Operation of all controllers is automatically restored in 12.5 minutes, Operation of RFP Controllers is available in MANUAL for the first 30 seconds,
' The Master Level Controller will stay at its tapeset demand signal for 10 seconds, i
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e 1 e s -rN- n-, -a- e- er - m u--- er-e -,cr.- , -4n,w ,vm p.o.- rn-- ,-s-v----.Nreve.------,r-- wn-- -~.wan.w, ,,-g,wn ,,e o>w -r---c- ,
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REACTOR OPERATOR Page 49 OUESilON: 084 (1.00) l l sl B13, RPV Control, has been entered on low reactor water level. Current water level is !
+ 15 inches and slowly decreasirig. RPV pressure is 120 psig. Which of the following l ALTERNATE INJECTION SYSTEMIS)/ SUBSYSTEM (S) can be used for RPV injection for the present plant conditions? RHR Loop D Flood, SPCU alternate injection, Hotwell Dump Line alternato injectio Condensate transfer alternate injection.
QUESTION: 085 (1.00)
A plant startup is in progress with power at 10% with two Circ Water Pumps operating. A problem with the Pteam Jet Air Ejectors causes Main Condenser Vacuum to decrease to 7" HgA. The problem was corrected and vacuum returned to normal. Assuming no operator action, which of the following would be correct after the transient? Reactor power less than 19% due to the Recirc Flow Control Valve runback, Reactor power would be approximately 19% with the Main Turbine still operating, The Main Turbine tripped on low vacuum with the Dypass Valves controlling l Reactor Pressure, The Motor Feed Pump controlling reactor level due to the trip of a Reactor Feed Pump Turbine on low vacuu l
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REACTOR OPERATOR Page 50 QUESTION: 086 (1.00)
High Pressure Core Spray (HPCS) and the Motor Feedwater Pump (MFP) were manually started during a loss of feedwater flow transient to recover level. SELECT the statement that describes the plant response as RPV water levelincreases to the Level 8 setpoin The HPCS pump and MFP will trip, The HPCS injection valve will close and the MFP will tri The HPCS pump will trip, the MFP will operate until manually secure The HPCS Injection valve will close, the MFP will operate until manually secured.
QUESTION: 087 (1.00)
PEl T23, " Containment Temperature Control", requires termination of containment sprays if containment pressure is below 1.75 psig. SELECT which statement below describes the basis for this ste Terminating containment spray at this pressure avoids containment f ailure due to negative pressure, Cooling the containment with the containment sprays will drive the RPV saturation temperature below the curve irto the unsafe 'egion and make level instrumentation inaccurat Continuous operation of the containment sprays will decrease the margin to HCL. Subsequent Emergency Depressurization may be required to stay in the safe region of the HC Since isolations in the secondary containment may have occurred, a Technical Specification limit of -0.1 to 1.0 psid between the Containment and Auxiliary Building cannot be assured, jeopardizing the Auxiliary Buildin _ y -. - - - - - - - . - . . - - - - - _ _ . .
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QUESTION: 088 (1.00)-
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- Given the following plant conditiv.s: !
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- A LOCA has occurre >
- RHR A, B, & C were manually initiated 'n the LPCI mode one minute ago per- ,
PEl SPI 5.2, injection Preventio ;
- MFP is maintaining RPV level 50100".
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- SP Temperature is 100*F. and irnreasing, o
When is the earliest time that you can fully realign RHR "A" to Suppression Pool Cooling?- [
Due to the initiation of RHR in the LPCl mode, RHR "A* cannot be lined up in ! the SP Cooling Mode for nine minute ! ' PEl T23 will not allow RHR to be realigned to SP Cooling at this time because LPCI mode is required until adequate core cooling is establishe ,
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- PEl B13 and PEl T23 state that RHR can be realigned from the LPCl mode to SP Cooling Mode only as required to stay below the Heat Capacity Limit (HCL).-
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d.- The RHR HX BYPASS VALVE 1E12-F048 will not fully open for 110 seconds following a LPCI signal, so 50 seconds is the earliest that SP Cooling can be
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OUESTION: 089 (1.00)
Af ter arming and depressing the RPS manual scram pushbuttons and placing the Reactor Mode Switch in SHUTDOWN, what additional actions are to be taken in accordance with ONI C61, Evacuation Of The Control Room, in the event a Control Room evacuation is required? ;
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a, trip the Main Turbine, initiate HPCS, and place Division 2 DG Control Transfer f Switch in LOCA . verify all rods inserted, initiate RCIC, and place Division _1 DG Control Transfer ;
Switch in LOCAL, verify all rods inserted, initiate RCIC, and place Division 3 DG Control Transfer Switch in LOCAL,
] verify all rods inserted, trip the Main Turbine, and place Division 3 DG Control ,
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- Transfer Switch in LOCA i
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REACTOR OPERATOR Page 52 OUESTION: 090 (1.00)
Which ONE of the following statements is correct concerning a complete loss of Nuclear Closed Cooling water? The Reactor is shutdown in anticipation of... the CRD pump tripping on a high lube oil temperature of 195' the Drywell pressure increasing to 1.68 psig due to the loss of Drywell coolin RWCU Filter Domins, isolating on high NRHX Inlet temperature and-the associated loss of Reactor chemistry, d, tripping the Reactor Recirculation Pumps on high bearing temperature and the loss of Instrument and Service Air compressor QUESTION: 091 (1.00)
SELECT the statement that describes a condition requiring a fast reactor shutdown following a loss of instrument air while operating at 45% power, per ONI-PS2, Loss of Service and/or Instrument Alt, The INST VOL NOT DRAINED annunciator is receive The SCRAM VLV AIR HEADER PRESSURE LOW annunciator is receive The outboard main steam line isolation valve (MSIV) 821 F028A drifts close The ROD DRIFT annunciator (for rod 2211) and CRD MECHANISM TEMP HIGH annunciator (for rod 46 39) is received.
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REACTOR OPERATOR Page 53
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QUESTION: 092 (1.00)
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The plant is in Mode 2 performing a Rx Startup when the running CRDH purop trips. Which of the following describes the plant conditions thtst would require placing the Rx Mode Switch !
in SHUTDOWN immediately, per ONI-C11 1, Inability To Move Control Rods? Rx Pressure is 500 psig. Accumulator f ault on rod 30 31 at 0" comes in. CRD Charging Water Pressure is 1570 psi Rx Pressure is 920 psig. Accumulator faults on rod 30 31 at 0" and rod 20 27 at 24" has been in for 20 minutes. CRD Charging Water Pressure is 1850 psi Rx Pressure is 900 psig. Accumulator f ault on rod 20 27 at 24" CRD Charging Water Pressure is 1500 psig, Rx Pressure is 550 psig. Accumulator fault on rod 20 27 at 24". CRD <
Charging Water Pressure is 1575 psi .
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REACTOR OPERATOR Page 54 OUESTION: 093 (1.00)
Given tL' .,owing initial plant conditions:
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Current deactor Power: 05 %
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Current rod line: 105%
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Current total core flow: 60 Mlbm/hr
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Reactor Recirculation Flow Control in LOOP manual The "A" Recirculation loop flow controller then f allo, causing tha "A" flow control valve to close, decreasing core flow to 41 Mlbm/hr. SELECT the appropriate IMMEDIATE action to be taken per ONI C51, Unplanned Change in Reactor Power or Reactivity, as a result of these plant conditions, insert Cram rods per FTI 802, Control Rod Movements, to exit the IMMEDIATE EXIT region, Immediatelj scram the reactor by armir.g and depressing the RPS MANUAL SCRAM CH A, B, C, and D pushbuttons, Exit the MANUAL SCRAM REQUIRED Region of the Power to-Flow map by increasing flaw in the *B" Recirculation loop, Establish a load line less than or equal to the 100% load line, then shutdown one Reactor Recirculation Pump per sol 83 _
REACTOR OPERATOR Page 55 OUESTION: 094 (1.00)
The following plant conditions exist:
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A small break LOCA and ATWS have occurred
- Reactor power is 8% and slowly decreasing
- Both SLC pumps are running
- 32 control rods f ailed to fully insert
- RPV level / pressure is 90"/900 psig
- Suppression pool temp. / level is 130*F /18.6 feet
- Drywell temp. / pressure is 150'F / 0.3 ptig
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Cntmt. temp. / pressure is 100aF / 0.2 psig What is the required operator action per the Plant Emergency Instructions? Restore and maintain Suppression Pool level between 17.8 and 18.5 feet, Bypass the NCC solation using PEl SPl 2.1 to restore NCC to the Drywell coolers, Enter PEl B13, RPV Control (ATWS) at "X" to enter PEl- 013, Emergency Depressurizatio Manually control reactor pressure using SRVs as necessary to maintain RPV pressure below HC ,
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1 f-r f OUESTION: 095 11.00)
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Containment Flooding has been entered from RPV Flooding. Current plant conditions are as .
follows: !
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Reactor level UNKNOWN
- . Reactor pressure O psig
- Suppression pool water level 62 feet ,
. Containment pressur psig l
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- All Control Rods inserted "00"
- The hotwell pumps are injecting 4 Mlbm/hr into the reacto What actions are required based on these plant conditions? . Terminats injection to the RPV from the Condensate System. . F.xit Containment Flooding and go to RPV Level Control (START). ~ Decrease Condensate flow to lower Suppression Pool Water level to 61 fee Maintain Suppression Poollevel above 61 feet with Condensate System taking >
a suction external to the containmen !
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REACTOR OPERATOR Page 57 OUESTION: 096 (1.00)
Following a LOCA, the following parameters are noted:
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RPV Pressure 100 psig
- Drywell Temperature 250* Containment Temperature 125' ,
RPV LEVELS:
1) Narrow Range 180" 2) Shutdown Range 190" 31 Upset Range 195" Which of these levelinstruments may be used to determine level? All of those instruments can be use None of these instruments can be use Only the narrow range is providing valid level indication, Only the shutdown range is providing valid levelindicatio QUESTION: 097 (1,00)
The plant is shutdown in Mode 4 with RHR "B" in Shutdown Cooling and RHR "A" is OO RHR Pump '8" shaft seizes. Which ONE Of the following is an approved alternate means of decay heat removal per ONI E12-2, Loss of Decay Heat Removal? Operate High Pressure Core Spray (HPCS) to circulate coolant between the suppression pool and the reactor vessel via the head ven If Rx coolant temperature is > 190'F., align the RPV head vents by opening RX HEAD TO DW FIRST/SECOND VENT VALVES,1821 F002/FOO l Operate the condensate system to circulate water between the main condenser j and the reactor vessel to the hotwell pumps via the Main Steam isolation Valves (MSIVs). !
' l Operate Low Pressure Core Spray (LPCS) to circulate the coolant between the l suppression pool and the reactor vessel with the LPCS via two Safety Relief Valves (SRVs).
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OUESTION: 098-(1.00)
During refueling activities, a decrease in upper containment pool level requires suspension of '
- core alterations after placing all irradiated fuel and core components in a cafe conditio Which of the following is a SAFE CONDITION for a fuel handle in the IFTS carriage? ,
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' Bundle properly seated in the IFTS carriage at the Bottom Out position with the I upender vertica l t Bundle properly seated in the IFTS carriage at the Helse Lower Limit pos!Jon !
with the upender incline I Bundle properly seated in the IFTS carriage at the Raise Lower Limit position '!
with the opender vertica i A L Jndle cannot be in a SAFE CONDITION in the IFTS carriage and must be ;
plaosd in a pool storage locatio '
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'OUESTlON: 099 (1.00)
The Maximum Safe Operadng Condit!on Values for Area Temperatures found in PEl N11, ,
_ Containment Leakagt: Control, ore based on: ( Equipment necessnry for safe shutdown of the plant will fall above this -
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temperatur Bar,ed on the high room temperature clarm setpoints for the surrounding
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containment area !
- The availability of alternate instrumentation use in the case of f ailed Installed instrumentatio .j d.- Personnel acem necessary for the safe operation of the plant will be precluded ;
above this tempo. ur ;
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QUESTION: 100 (1.00) ,
Due to a 1.OCA and a steam leak in the RCIC pump room, PEl B13 RPV Control (Non ATWS),
and PEl N11, Containment Leakage Control, heve been entered. The RCIC pump room is i
- inaccessible and steam makes observation of the room impossible. However, the following indications are available to the operators: RCIC Pump Room Area Temperature is 29S'F Thn RCIC Pump Room Sump hi level alarm has been received RWCU Pump Room Area Temperature is 265' Which ONE of the following conditions would require the operators to enter PEl-B13, RPV Control (Non ATWS) at "Z" with the intentions of performing Emergency Depressurization? RCIC Area HVAC Differential Temperature alarming and reading 95'F Aux. Building Ventitation Exhaust Gas radiation monitor reading 34,050 cprn
, Aux. Building 574' elevation Hallway Area Radiation monitor reading 4,200 mR/h Aux. Building 574' elevation Hallway water level observed at 20" above the floo ,
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ANSWER: OR1 (1.00) ANSWER: 008 (1.00)- !
REFERENCE: _
REFERENCE: !
SDMi C 1 1 (' R C & 1 S i LP: SDM: M16 LP: l OT 3036 C11(RCalS), Obj. G OT.3036 M16, Obj. C, E ,
201005G004 . .( K A's) 223001K501 ..(KA's) ;
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ANSWER: 002 (1.00) ANSWER: 000 (1.00) REFERENCE: REFERENCE:
SDM: B33 LP: OT 3036 833, Obj. E f.DM: G33 LP: OT 3036 G33 E31
ANSWER: 003-(l.00)
a, .
ANSWER: 010 (1.00)
REFERENCE: ;
PEl SPl 4.2 LP: OT 340216, Obj. D REFERENCE: i 203000K610 ..(KA's) SDM: N32/C85 LP: OT 3036 N32/C85, Obj. G, D !
241000K403 ..(K A's) .
ANSWER: 004 (1.00) REFERENCE: ANSWER: 011 (1.00)
sol E21 LP: OT 3036 E21, Obj. H G010 ..(KA's) REFERENCE:
SDM: N21 LP: OT 3036 N21, Obj. 8, D ;
2560CJA105 ..(KA's)
ANSWER: 005 (1.00) REFERENCE: ANSWER: 012 (1.00)
SDM: E22 LP: OT-3036 E22, Obj. E K407 ..(KA'n) REFERENCE:
SDM: E12 LP: OT 3036-E12, Obj. F ANSWER: 006 (1.00) 226001K409 ..(KA's)
d- '
REFERENC SDM: E51 LP: OT 3036 E51, Obj. D ANGWER: 013 (1.00) ,
217000A404 ..(KA's) - REFERENCE:
ANSWER:- 007 (1.00) ARl H13 P680 9 LP: OT 3036 N35, Ob e . -- H-REFERENCE: 245000G012 ..(K A's)
-SDM:- C34 LP: OT 3036-C34, Obj. C 259002K503 .. (K A's)
a c _ . _ _ --- . _ . - - - _ . _ - . _ _ ,._-.----_;_;_.-,_ - - - , .
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ANSWER: 014 (1.00) ANSWER: 020 (1.00) ' i REFERENCE: REFERENCE- !
SDM: C11(RCIS) LP: OT-3036 C11(RCIS), PEI Bases LP: OT-340211, Obj. D Ob K303 ..(KA's) r 201003K103 ..iKA's) [
ANSWER: .021 (1.00) [
ANSWEF.: 015 (1.00) a.-
' REFF.RENCE:
REFERENCE: . rEl B13 LP: OT 3402 02, Obj. F ;
SDM: P54 (WTR) LP: 295006K207 ..(K A's)
OT 3036 P54(WTR), Obj. D ,
286000A301 ..(KA's) l ANSWER: 022 (1.00) >
3 ANSWER: 016 (1.00) REFERENCE: SDM: R42 & R10 LP: OT-3036-R42, Obj. B REFERENCE: 295003A203 ..(K A's) ,
ONI-C11-2 LP:- OT 3036-C11(RChlS),
Obj. H _
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201003K402 ..(KA's) ANSWER: 023 (1.00) ;
REFERENCE: *
ANSWER: 017 (1.00) ONl R42 5, Sect.1.3 LP: OT 3036 R42, Ob ;
REFERENCE: 295024A204 ..(K A's) ;
SOI G33 LP: OT 3036 G33, Obj. C -
204000G010 ..(KA's)
ANSWER: 024 (1.00) , '
ANSWER: 018 (1.00) REFERENCE: SDM-C61 LP: OT 3036 C61, Obj. E REFERENCE: 295016A107 ..(KA's) .
SDM: M14 LP: OT 3036 M14, Obj. D 288000A204 ..(KA's)
ANSWER: OM (1.00)
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ANSWER: 019 (1.00) REFERENCE: PEl T23 LP: OT-3402 07 . Obj. C PEl REFERENCE: Bases -
PEl Bases LP: .OT 3402-03, Obj. D 295027K301 - ..(KA's) 1 295037K301 ..(K A's)
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ANSWER: 026 (1.00) ANSWER: 033 (1.00) !
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SDM: G43 LP: OT-3036-G43, Obj. D PAP-0219. Sect. 6.8.2 LP: GEN 1001
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295030A104 ..(K A 's) - 294001K105 ..(KA's)
i ANSWER: 027 (1.00) ANSWER: 034 (1.00) '
REFERENCE: REFERENCE:
FEl D17 LP: OT 340215, Obj. B PAP-0504 LP: OT 3039 00, Obj. B 295038K205 ..(KA's) 294001K107 ..(KA's) :
ANSWER: 028 (1.00) ANSWER: 035 (1.00) - REFERENCE: REFERENCE:
-PEl N11, Containment Leakage Control PAP 1910, Sect. 6.4 LP: OT-3039 00, Ob OT 340217, Objective D 'B 295033G012 - ..(K A's) 294001K116 ..(KA's) l
ANSWER: 029 (1,00) ANSWER: 036 (1.00) ' <
REFERENCE: REFERENCE: :
PAP-0201,6.8.6 LP: OT 3039 01, Obj. A PAP-0201 LP: OT 3039-01, Obj. A !
294001K101 ..(KA's) 294001A102 ..(KA's)
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ANSWER: 030 (1.00) _ ANSWER: 037 (1.00) l REFERENCE: REFERENCE: '
PAP 1401 LP: OT 3039 01, Ob PAP 0110 LP: OT 3039 01, Obj. A 294001K102 ..(KA's) 294001A103 ..(KA's) *
i ANSWER: 031 (1.00) ANSWER: 038 (1.00) ' REFERENCE: REFERENCE:
PAP 0514 LP: OT 3039 00, Obj. B OAl 1702 LP: OT 3039 00, Obj. B 294001K103 ..(KA's) 294001A106 ..(K A's)
l ANSWER: 032 (1.00) ANSWER: 039 (1.00) l
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REFERENCE: REFERENCE:
PAP-0118i Sect. 6. Tech Spec. 3.4.11 LP: OT-3037 08, Obj. D i 294001K104 ..(KA's) 294001A108 ..(KA's)
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REACTOR OPERATOR Page 63
i ANSWER: 040 (1.00) ANSWER: 046 (1.00) { REFERENCE: REFERENCE SDM: M28 LP: OT-3036-M28, Obj. C SO Rounds LP: OT 3036 C41, Obj. E !
294001A113 .. (KA's) 211000A403 ..(K A's) j
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ANSWER: 041 (1.00) ANSWER: 047 (1.00) [ !
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HEFERENCE: REFERENCE:
Emergency Planinstruction EPl A7, section SDM: C71 LP: OT-3036, C71, Obj. D f 4.1 LP: EPL-0823 C22 OT 3036 C22, Obj. D l 294001A116- ..(KA's) 212000A108 ..(K A's) >
i ANSWER: 042 (1.00) ANSWER: -048 (1.00) '
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REFERENCE: _
REFERENCE: ;
SDM: C11(CRDH) LP: OT 3036 C11 SDM: C71 LP: OT 3036 C71, Obj. D t (CRDH), Obj. C 212000K201 ,,(KA's) -
201001K201 ..(KA's)
l ANSWER: 049 (1.00)
ANSWER: 043 (1.00) REFERENCE: !
REFEhENCE: SDM: C 51 ( P R M ) LP:
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SOM: C 11 ( R C & 1 S ) LP: OT 3036 C51(PRM), Obj. D,1 OT 3036 C111RC&lS), Obj. G 215005A104 ..(KA's)
201005K40:s ..(KA's) ,
t ANSWER: 050 (1.00; l ANSWER: 044 (1,00) b, : RE6 ERENCE:
REFERENCE: CDM: B21(NBPI) LP: OT 3036-821(INST), ,
c Ops. Policy 2 9 OT 3039-00, Obj. B PlF Ob i 9C3400 216000K607 ..(KA's)
202001G001 ..(K A's)
ANSWER: DS1 (1.00)
ANSWER: . 045 (1.00) REFERENCE:
REFERENCE: SDM: B21C LP: OT 3036-821C,6bj. E SDM: E12 LP: OT-3036-E12. Obj. F 218000A403 . . (K A's) '
203000KG10 ..(K A's)
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.,mm y - - n e-- c g e -+-te m . m- -s- -+re ve +r-r-et--we<-,- .-%.e v-'-.re-.
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.-rr ,.-i.-v:-.c--,--s --...n--r- -- e .+...=m == -.m- -
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. REACTOR OPERATOR Page 64 i ANSWER: 052 (1.00) ANSWER: 058 (1.00) f c.- !
REFERENCE: REFERENCE: l SDM: B21/N11 ONI E121 LP: OT 3036 E22, Obj. K !
LP: OT-3036 821/N11, Obj. E 209002A201 . . (K A's) .
-
239002A300 ..(KA's) '
ANSWER: 059 (1.00)
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ANSWER: 053 (1.00) ; REFERENCE: <
REFERENCE: - SDM: C51(SRM) !
SDM: N32/C85 LP: OT 3036-N32/C85, LP: OT 3036 C51(SRM), Obj. D !
Ob K402 . . (K A's)
'
'
241000K302 ..(K A's)
ANSWER: 060 (1.00)
ANSWER: 054 (1.00) REFERENCE:
REFERENCE: T.S. 3.4.1 LP: OT 3036 B33, Obj. J SDM: C34 LP: OT 3036 C34, Obj. C 202001G005 ..(KA's)
259001A201 ..(KA's)
ANSWER: 061 (1.00)
ANSWER: 055 (1.00) REFERENCE: ,
'
REFERENCE: SDM: C41 LP: OT 3036 C41, Obj. D SOM: C34 LP: OT 3036-C34, Obj. C 204000K108 . . (KA's)
259002K409 . . ( K A's.)
ANSWER: 062 (1.00)-
ANSWER: 056 (1,00)
, REFERENCE:
REFERENCE: T.S. 3.4.10 LP: OT 3037 08, Obj. C SDM: MIS LP: OT 3036 M15, Obj. E 205000G005 ..(K A's)
261000K401 ..(KA'wl ,
ANSWER: 063 (1.00)
ANSWER: 057 (1,00) REFERENCE:
REFERENCE: sol E12 LP: OT 3036 E12, J SDM: E228 & R43 LP: OT 3036 E22B, 219000G010 ..(K A's) -
Obj. E OT 3036 R43, Obj. D 264000A209 ..(KA's)
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REACTOR OPERATOR Page 65
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ANSWER: 064 (1.00) ANSWER: - 070 (1.00) l l REFERENCE: REFERENCE:- !
SDM: E12 LP: OT-3036 E12, Obj. E, F SDM: N36 LP: OT 3036 N36/25/26, Obj. B t 226001A407 .. (K A's) _239001K110 ..(KA's)
ANSWER: 065-(1.00). ANSWER: 071 (1.00)
b, REFERENCE: REFERENCE: I SDM: R10 LP: OT 3036 R10, Obj. D SOI.N32 LP: OT 3036 N32/C85. Obj. K 262001A302 ..(KA's) ONIN62 l 245000K502 ..(KA's) -
ANSWER: - 066 (1.00)
a,
'AMSWER: 072 (1.00) i REFERENCE: SDM: R14/16 LP: OT 3036 R14/15, Ob REFERENCE:
B R42 OT 3036 R42, Obj. B SDM: G41 LP: OT 3036 G41, Obj. D-263000K201 ..(KA's) 2330n0K408 ..(K A's)
ANSWER: 067 (1.00) ANSWER: 073 (1.00) REFERENCE: REFERENCE:
SDM: N64 LP: OT 3036 N64, Obj. E SDM: E32 LP: OT 3036-E32 Obj. E 271000K408 ..(KA's) 239003A211 ..(KA's)
ANSWER: 068 (1.00) ANSWER: 074 (1.00)
b, d. -
REFERENCE: REFERENCE:
SDM: D21 LP: OT-3036-D21, Obj. B T.S. 2.2 LP: OT 3037 03, Obj. H 272000A101 ..(KA's) - 290002G005 ..(KA's)
ANSWER: 069 (1.00) ANSWER: 075 (1.00) REFERENCE: REFERENCE:
SDM: M25/26 LP: OT 3036-M25/26, SDM: C71 LP: OT 3036-C71, Obj. F Ob K201 ..(KA's)
290003K101 ..(K A's)
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REACTOR OPERATOR Page 66 - l l
ANSWER: 076 (1.00) ANSWER: 082 (1.00# [
REFERENCE: REFERENCE:
SDM: C34 LP: OT 3036 C34, Obj. D, G - PEl-T23 LP: OT 3402 09, Obj. C PEl 295006A102 ..(KA's) Bases t 295024A209 ..tKA's) !
ANSWER: 077 (1.00) r ANSWER: 083 (1.00) i REFERENCE: #
ONI C851 REFERENCE: L LP: _ OT 3036 N32/C85, Obj. N SDM: C34 LP: OT 3036 C34, Obj. D .
295007A105 ..(KA's) C22 OT 3036-C22, Obj. D l 295025K203 ..(K A's)
ANSWER: 078 (1.00) ' ANSWER: 084 (1.00)
REFERENCE: - i SDM: 833 LP: OT 3036 B33, Obj. E REFERENCE:
295009K102 ..(KA's) PEl B13 LF: OT-3402-02, Obj. F - :
295031A108 ..(KA's)
ANSWER: 079 (1.00) ANSWER: 085 (1.00)
REFERENCE: SDM: C22 LP: OT-3036 C22, Obj. D REFERENCE:
295025A107 ..(KA's) ONl N62 LP: OT 3036 N62, Obj. I 295002A201 ..(KA's)
ANSWER: 080 (1.00) i ANSWER: 086 (1.00)
REFERENCE: SDM: C11(RCIS) LP: OT 3039-01, Obj. A REFERENCE:
1011 & 2_ PAP-0201 SDM: E22A LP: OT 3036 E22A, Obj. E 295014K101 ..(KA's) C34- OT 3036-C34, Obj. C ;
295008A101 ..(KA's)
ANSWER: 081 (1.00) ANSWER: 087 (1.00) ,
REFERENCE: . SDM: C11(RCIS)8.P:OT 3036 C11(RCIS),. REFERENCE:
. Obj.= E, G PEl T23 LP: OT-3402-07, Obj. . C PEI
'295015A104 ..(K A's) Bases 295011K101 ..(K A'st
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REACTOR OPERATOR Page 67 ANSWER: 088 (1.001 ANSWER: 094 (1.00) REFERENCE: REFERENCE:
SDM: E12 LP: OT 3036 E12, Obj. E PEl T23 LP: OT-3402 06, Obj. C 295013A101 ..(K A's) 295026G012 ..(KA's)
ANSWER: 089 (1.00) ANSWER: 095 (1.00) REFERENCE: REFERENCE:
SDM: C61 LP: OT 3036 C61, Obj. C PEl T23 LP: OT 3402-14, Obj. C PEl ONIC01 Bascs 295016G01A ..(K A's) 295029G012 ..(KA's)
ANSWER: 090 (1.00) ANSWER: 096 (1.00) REFERENCE: REFERENCE:
ONI P43 LP: OT 3036 P43, Obj. H PEl B13 LP: OT 3402 01, Obj. D 295018K302 ..(KA's) 295028K203 ..(KA's)
ANSV'ER: 091 (1.00) ANSWER: 097 (1.00) REFERENCE: REFERENCE:
ONI PS2 LP: OT 3030-P51/52, Obj. G ONI Ei2 2 LP: OT-303511, Obj. A 295019G010 ..(KA's) 295021 A104 ..(KA's)
ANSWER: 092 (1.00) ANSWER: 098 (1.00) REFERENCE: REFERENCE:
ONI-Cil-1 LP: OT 3036 C11(CRDH), Ob ONi E12-2 LP: OT 3 *36-G41, Obj. E G 295023A103 ..(KA's)
295022A201 ..(KA's)
ANSWER: 099 (1.00)
ANSWER: 093 (1.00) REFERENCE:
REFERENCE: PEl Bases page 341 LP: OT 340217, Ob SDM: 833 LP: OT 3036-B33, Obj. I B PEl N11 1-ONI-C51 295032G012 ..(KA's)
295001A201 ..(KA's)
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REACTOR OPERATOR Page 68 ANSWER: 100 (1.00) REFERENCE:
PEl N11 LP: 340217, Obj. D 295030G012 ..(KA's)
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( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * * )
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LREACTOR OPERATOR . Page 1 ;
, ANSWER KEV !
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MULTIPLE CHOICE 023 a- l t
001 - b 024 b l
002 a 025 c l
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003 a- 026 c-
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- 004 a 027 b I t
005 c 028 d~
f-t 006 d 029 c !
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030 b
- 008 - c 031 d 4
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009 c 032 d !
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010 ~ b - 033 b ,
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011 d 034 a
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. 012 c 035 c ,
013 - c 036 d l 014 b 037 c i 015 b 038 c ,
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016 c kA 039 d 017d 040 b 0.18 La - 041 d .
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019 b : 042 . a - ,
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- 020 - d- 043 .a _
021 a 044 a
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022~a 045 r
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,J .na- ,c.,st~-erw.aw..-, rm e. w s, w v --r+ , < e- nare- :w . -or - e v e += .e e,ym., .w-- *.ev.-.en . -w , ,e-, +- y a.e-**- - mgwe .e g % r g- y em t +y-y- -
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REACTOR OPERATOR Page 2 ANSWER KEY MULTIPLF CHOICE 068 b 046 a 069 b 047 c 070 c 048 c 071 c 049 d 072 d 050 b 073 c ,
051 c 074 d 052 c 075 b 053 b 076 c 054 a 077 b 055 b 078 b
! 056 b 079 b 057 a 080 a 058 a 081 c l
059 b 082 c 060 c 083 a 061 b 084 d 002 c 085 b 003 d 086 b ,
064 d 087 a 065 b 088 a-066 a 089 d 067 c 090 d
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REACTOR OP.ERATOR Page-3 ANSWER KEY M')LTIPLE CHOICE 091 a 092 d 093 b 094.c 095 a 096 c 097 d 098 b 099 a 100 d (' * ' " ' ' ' END OF EXAMINATION " ' * * ' * * )
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l ES-401 Site specific Written Examination Form ES-401 1 l Cover Sheet !
U. S. NUCLEA ? REGULATORY COMMISSION SITE-SPECIFir, WRITTEN EXAMIA VIOii
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APPLICANT INFORMATI')N Name: MASTER EXAMINATION Region: lil Date: August 30,1997 Facility / Unit: Per4y Nuclear Power Plant License Level: SRO Reactor Tyne: GE INSTRUCTIONS Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade re, quires a final grade of at least 80 percent, Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination start All work done on thic examination is my own. I have neither given nor received ai Applicant's Signature RESULTS Examination Value 100.0 Points Applicant's Score Points Applicant's Grade Percent
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ES-402 : Policies and Guidelines' Attachment 2 foi Taking NRC Written Examinations i Cheating on the examination will result in a denial of your application and could result in more severe penaltie l After you complete the examination, sign the statenent on the cover sheet indicating that the work is your own and you have not received or given assistance in completing =
the examinatio . To pass the examination, you must achieve a grade of 80 percent or greate . The point value for each question is indicated in parentheses after the question namber.
- There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for completing the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan . If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted, but only one applicant at a time will be allowed to leav Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, asremble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or procto Remember to sign the statement on the examination cover shee . After you have turned in your examination, leave the examination uea as defined by 4 the examine . . -- .. - . - . - - - . -
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SENIOR REACTOR OPERATOR Page 4
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. QUESTION: 001 (1.00)
The plant is in a startup and the RCIC system is being restored to Standby Readiness. Reactor pressure is 100 psig. You are performing the warmup of the RCIC system and reach a step in SOI E51, Reactor. Core Isolation Cooling System, to throttle open the E51-F076, Steam Warmup Valve, and you find the valve already FULL OPEN. - You should: ;
'
. a, proceed with the procedure, since the valve is already open and the intent of-the steam line warmup is accomplished, proceed with the procedure af ter notifying your Supervisor who wili change the ,
valve position to open on the verification checklist and initiate a non-intent temporary change.
' stop the procedure and obtain your Supervisor's permission to annotate the step f FIC and continue the instruction, stop the procedure and obtain your Supervisor's permission to appropriately position the valve and document action taken in the Unit lo QUESTION: 002 (1.00)
The 'B' Turbine Building Closed Cooling (TBCC) Pump is RED TAGGED OUT of service for maintenance. A loss of the 'A' TBCC pump has occurred and a plant operator reports that the
'B' pump looks like it is fully reassembled and no one is working on it. The Maintenance Supervisor, who is the person in charge on the tagout, has lef t the site for lunch. Assuming all are Person-In-Charge qualified, which of the following can authorize tag removal?- The Unit Supervisor The lead mechanic on the crew The responsible system engineer The operations foreman who walked down the "B" TBCC Pump with the plant operator
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SENIOR REACTOR OPERATOR Par * 5
' OUESTION: 003 (1.00)-
John Smith, age 33,is, a radiation worker at Perry, John's current calender quarter radiation exposure is 250 millirem TEDE, all received at Perry Station. John was assigned to Clinton station earlier this year and received 125 millirem TEDE. John was also given two chest x-rays during July 1997 for an additional 0.050 millirem. How much dose is John ' allowed to receive during this calendar year before he is required to get supervisor's permission to receive adoitional dose? .95 millitem millirern .95 millirem millirem QUESTION: 004 (1.00)
An emergency drywell entry is required, Rx power is 7% MWt with the main generator off-line. Which of the below statements is correct with respect to waiving the ALARA review for the drywell entry? Ar, ALARA review is not required for a drywell entry under these plant conditions per PAP-0118, ALARA Program, An ALARA review is required for work activities under these plant conditions and may not be waived per PAP- 0118, ALARA Progra The Unit Supervisor may waive the ALARA review by declaring this an urgent situation per PAP-0512, Radiation Work Permit Progra The Shift Suporvisor may waive the ALARA review by declaring this a Priority ,
1 Emergency per PAP-0902, Work Request Syste <
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- SENIOR REACTOR OPERATOR Page 6 OUESTION: 005 (1.00)
You have been assigned to escort several visitors for a plant tour that willinclude the control room. You should: process through the turnstiles ahead of your visitor when entering and exiting the Protected Are ensure that no more than 10 individuals are escorted in the Protected Areas, and no more than 5 individuals in the vital area use your key card first when traveling through key card controlled doors, then hold the door open while your visitors use their issued key card keep escorted individuals under continuous observation and report any indications of aberrant behavior following your visitor's exit from the plan QUESTION: 006 (1.00)
PAP-0504, Electrical Operating Rules and Practices, requires that an evaluation and pre-job inspection be done whenever working on energized equipment. Many of the PEl-SPIsinvolve use of lifted leads arid jumpers in energized panels, however an evaluation and pre-job inspection is not done by operators performing PEl SPIs. How is this in compliance with PAP-0504? The inspection / evaluation was done once and is applicable for all PEl SPI The inspection / evaluation is not required dua to the low voltage in the associated panel PEl-SPIs are done in what is considered a " plant emergency" and therefore the requirement for an inspection / evaluation is waive This requirement is only applicable when it is not prudent to use safety equipment, however, safety equipment is used during the performance of PEl-SPI . - _ _
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SENIOR REACTOR OPERATOR Page 7 QUESTION: 007 (1.00)
PAP-0110, Shift Staffing and Overtime, states that the Fire Brigade must be on site per PAP 1910, Fire Protection Program. Which of the following correctly states the requirements for the site Fire Brigade? The Fire Brigade is a five person team consisting a Fire Brigade Leader, the Fire Engineer, and three Fire Brigade member If less than the minirnum number of Fire Brigade members is on-site, the minimum number will be restored within four hour The on-shift Shif t Supervisor shall not be assigned to the site Fire Brigade, and the SO at the-controls should not be assigned as the Fire Brigade Leader, With the exception of the Fire Brigade Leader, who is a licensed operator, Fire Brigado members will have no duties during a fire except those directly related to manual fire fighting.
QUESTION: 008 (1.00)
Which ONE of the following statements conforms to the recommended good operating practicos listed in PAP-0201, Conduct of Operations? Disconnect switches or breakers opened as a result of sol or ELI performance must be DANGER tagge The Unit Supervisor should have a nuisance alarm disabled if the work to correct the problem cannot be completed in a reasonable time, defined as one hou To prevent damage of remotely operated throttle valves when closing them, the control switch should be released as soon as the CLOSED indication is received, Whenever possible, place systems having an automatic actuation feature in a lineup that prevents an automatic actuation, prior to performing system troubleshooting or repai . . -. .. -. . . - . .-
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iSENIOR REACTOR OPERATOR- Page 8
- OUESTION: 009 -- (1.00) -
_Which of the following defines the requirements' for the licensed operator At The Controls
- (ATC_ Operator) to_ leave the horseshoe area of the control room without relief?-- The ATC -
Operator may_ leave the horseshoe area of the control room: only'with a proper watch relie ' b, any time a Supervising Operator is within the horseshoe area.- to verify receipt of.an annunciator during an emergency that affects safe-operation of the plant, to retrieve a system description manual when an emergency has occurred affecting safe operation of the plan QUESTION: 010 (1.00)
SELECT the correct action to be taken when taking logs if the Containment Humidity instrument is out of service for preventative maintenance which will continue for several day Place an _"X" in the data block and explain in the remarks section, Record the out-of service instrument reading and explain in remarks sectio ) Obtain the Unit Supervisor's permission to use alternate instrumentation, i asterisk in data block, explain in remarks sectio Wait until the preventative maintenance has been completed on the instrument -
and then record the instrument reading in the round i l
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- OUESTION: 011' (I'00).-
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' Given the following Rx cooldown data:
TIME - TEMPERATURE TIME ' -TEMPERATURE 0800- 520 1000- 325 081 . 310-0830 460 1030 -275 0845' :445- -1045-- ~250 0900 430 1100 225 0915 '400 1115 205 0930 370 0945 340
Choose the correct response from below with regard to the results of the cooldow The cooldown was acceptable and no limits were exceede The cooldown rate was exceeded during only one one-hour perio The cooldown rate was exceeded during only two one-hour period The cooldown rate was exceeded during only three one-hour period )
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SENIOR REACTOR OPERATOR Page 10 L QUESTION: 012-(1.00)-
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READINESS, ECC Area Cooling Fan 'A'is RED TAGGED OUT, ECC Area Cooling Fan 'B'is in STANDBY READINESS, Control Complex 'B' Chiller & Chilled Water Pump are operatin As the operator at the controls, which one of the following actions should you take? Start ECC pump 'B' from the control room, b, Have the Nuclear Island rounds taker start ECC Area Cooling Fan 'B'. ECC pump "A" must be securad due to no Area Cooling available, Have Maintenance personnel install temporary _venidation directed at ECC pump . ' A'.
QUESTION: 013 (1.00)
There are certain responsibilities that can only be performed by the Emergency Coordinator when the Site Emergency Plan is being implemented. Which one of the following actions is allowed to be performed by someone other than the Emergency Coordinator? Direct the notification of offsite agencies and organization Determine the emergency classification including reclassification or termination, Recommend protective actions for the general public to State and local County Official Coordinate and direct the actions necessary to terminate or mitigate the ef fects of the emergenc l
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- -SENIOR REACTOR OPERATOR - Page 11
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- QUESTION: 014 k(1.00) - -
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- The plant is operating at 100% power with all of the _"A" pumps operating.when bus EH114 is suddenly de energized. The Div.1 DG auto starts, but does not tie into bus EH11L Which
.ONE of the following' responses would you' expect to see in this situation?-
c a. - CRD Mechanism High Temperature alarm on P601, Rx Recirc pump A willlose all seal cooling flo ~ c. - ;RHR Pump A will auto start on a LOOP signal, and run on . min, flo ' d. ~ Div.1 DG will run continuously unloaded due to auto bypass of the DG trips on-
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L a LOOP _ signal..
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OUESTION: 015 -- (1.00) H in which one of the following conditions will the RC&lS' allow control rod movement?-
k Rods in Group 5 may each be withdrawn from 00 to 12 prior to selecting the next ro Groups 3 and 4 are fully withdrawn, and a control rod in Group 1 is selecte An attempt is mada to fully withdraw the ro Groups 1 througn 4 are fully withdrawn with the exception of one rod left at position 44. An attempt to withdraw a rod in Group 5 is mad = Groups 1 through 4 are fully withdrawn Groups 7 and 8 are withdrawn to !
-position 12. Groups 5 and 6 are stillinserted. An attempt is made to withdraw ,
a rod in Group 9.
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- QUESTIONIO16 (1.00)
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Due to's xenon transient in the core l h become's necessary to increase Rx power using recbc L :
, flow to maintain 1250 MWe'. In accordance with Operations Policy 2 9, Operational Activit .
'
- Evaluation, how should this evolution be classified by the Shift Supervisor?
.. As risk'significan l
-. i . As a risk contributo .;
' As an unplanned change in reactivity per ONI C5 ' As an infrequently Performed Test / Evolution (IPTE).
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QUESTION: 017-(1.00) !
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- Select from below th's conditions that would result in the "CNTMT Spray B Start Signal
, - Received" annunciato ,
a, LPCI "A" automatic LOCA. start signal sealed in for -10 minutes, drywell
- pressure is 3.0 psig, and containment pressure is 9.0 psi ..I ; LPCI "A" & "C" automatic LOCA start signal sealed-in for 15 minutes, drywell '
pressure is 2.5 psig, and containmant pressure is 7.0 psi .Cntmt Spray "B" Manual Initiation Pushbutton armed and depressed for 45 ,
seconds, drywell pressure is 1.5 psig, and containment pressure is 10 psig.
' LPCI "B" and "C" automatic LOCA start signal sealed-in foi 12 minutes, drywell pressure is 2.5 psig, and the Cntmt Spray "B" Manual Initiation Pusnbutton is armed and depressed for 40 seconds.
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SENIOR REACTOR OPERATOR Page- 13
- OUESTION: 018 (1.00)
An operator reports that the "A" Standby Liquid Control (SLC) squib valve continuity meter indicates 5 mA of current. SELECT the action that should be taken, a .' Do nothing, since 5 mA is a normal indication, Declare the squib valve inoperable since 5 mA indicates the squib valve has fired, Declare the squib valve inoperable since 5 mA indicates current is too low and the squib valve may not fire, Declare the squib valve inoperable since 5 mA indicates current is too high and the ignitor element may have decompose QUESTION: 019 (1.00)
During power operation with the Mode Switch in RUN, an event has resulted in the following conditions:
- RPV level decreases to Level RPV pressure peaked at 1113 psig,
- Drywell pressure has reached 1.2 psi Which ONE of the following lists the expected status of the scram, backup scram and ARI valve solenoids. No operator actions have been take SCRAM BACKUP SCRAM ARI deenergized deenergized deenergized b, deenergized energized deenergized deenergized energized energized energized deenergized energized
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f SENIOR REACTOR OPERATOR Page 14
_ QUEST!.ON: 020 (1.00)'
A transient occurs and it becomes apparent that the RPS "B" MG Set has been lost. Which bus normally powers this equipment? -
-a, F1808 F1 C08 F1C12 , FiD12
- QUESTION: 021; (1.00)
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A reactor startup is in progress with the Rx Mode Switch in "STARTUP/ STANDBY" The following is the grasent status of the APRMs versus LPRM inputs, and indicated power:
APRM A B C D- E F G H
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D Level inputs: 4 5 4 3 4 4 6 6 (
C Level inputs: 4 3 3 4 6 2 4 4 B Level inputs: 3 4 3 4 4 4 6 4
- A Level inputs: 3 3 3 4 6 4 4 2 indicated Power: 11 % 16% -12% 11 % 12% 10 % 12% 10 %
SELECT the correct RPS response to the above data:
i No trips
. Rod block only Half scram only i
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d' ; Full scram l
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SENIOR REACTOR OPERATOR Page 15 OUF.STION: 022 (1.00) .
A discrepancy is noted between RPV level instruments where Wide Range f.evel "B"is reading lower than the others. Which ONE of the following conditions could cause this channel to ,
read lower than actual vessel level? Drywell temperature locally at WR *B" increases, Drywell temperature locally at WR "B" decrease WR "B" level transmitter equalizing valve leak A steam leak occurs at the condensing pot for WR "B" level instrument.
QUESTION: 023 (1.00)
A plant transient occurred such that the Automatic Depressurization System (ADS)
automatically initiated. RPV level is steady at 10", all rods are inserted, and RPV pressure is 25 psig. Which ONE of the following actions would result in ADS logic being reset? With RPV level steady at 10", ADS logic cannot be rese Increase RPV water lavel to 20", then place the ADS Logic inhibit switches in INHIBI Placing the ADS Logic inhibit switches in INHIBIT, then depressing the ADS Logic Seal-In Reset pushbuttons, Depress the ADS Logic Seal-in Reset pushbuttons, there take each ADS Valve Control Switch to the OFF positio .
SENIOR REACTOR OPERATOR Page 16 OUESTION: 024 (1.00)
In ordet to reduce the number of SRVs that reopen following a reactor isolation event, the Low Low Set function was added to several SRVs. Which ONE of the following statements correctly describe this function? At 1103 psig, F051D opens, arming LLS and 1821-F051C opens. IF pressure increases to 1113 psig, then four LLS valves open and close at 946 psi F051C cycles between 1103 and 936 psig and F051D cycles between 1033 psig and 926 psig, At 110'., psig, F051D opens, arming LLS at 1103 psig, F051C opens; and at 1 R3 psig, the other four LLS valves open, and then close at 946 psig. F051C cycles between 1103 psig and 936 psig, and F051 D cyc.les between 1033 psig and 926 psig, At 1103 psig, F051D opens, arming LLS and opening F051C. If pressure increases to 1113 psig, the other four LLS valves open and cycle between 1113 psig and 946 psig. F051 C cycles between 1073 psig and 936 psig, and F051 D cycles between 1033 psig and 926 psi At 1103 psig, F051D opens, arming LLS. If pressure increases to 1113 psig, four LLS valves will open, arming F051C, which will then open and cycle between 1077 psig and 936 pai?. The four LLS valves will cycle between 1113 and 946 psig. F051D wM vele between 1033 and 926 psig.
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- SENIOR REACTOR OPERAT_OR - Page 17 GllESTION: 025 (1.00)- -
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Reactor power is' approximately 40%, recirc pumps have been shifted to fast speed, and a plant startup is in progress. Which of the following statements correctly describes the plant .
response following a downscale failure of the in-service Main Steamline pressure regulator? l
'a; Rx pressure willincrease causing bypass valves to open and Rx pressure 3 return to the pre-transient valu '
b.: A slight pressure transient may occur, however pressure will remain relatively constant throughout this even Pressure will increase, bypass valves will open in an attempt to reduce
.' pressure, however the reactor will scram, d.- Pressure willincrease to the point where both the scram and RRCS setpoints
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for Rx pressuru will be reache _
l-QUESTION: 026 (1.00)
Both RFPTs are running,' being controlled by the Master Level Controller. The MFP control
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switch is in OFF. The Startup Level Controller Select Switch is selected to RFPT "A". What would be the resultant pump combination if RFPT "B" tripped, with no operator actions being taken? RFPT "A" would be running alone on the ML RFPT "A" would be running alone on the SUL , The MFP wou!d start and be controlled on the MLC with RFPT "A". The MFP would start and be controlled on the SULC with RFPT "A".
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SENIOR REACTCR OPERATOR Page 18 OUESTION: 027 (1.00)
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The plant is operating at 70% power' with RFPT "B" and tha"MFP on the moster level
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controlles, RFPT "A" is danger tagged OOS. Which of the following strtements correctly; describes the response of the feedwater level control system to r.be trip of RFPT "B"?
, The level demand signal is that signal corresponding to the tapeset of the '
master level controller and will remain so continuousl The level demand signal is that signal corresponding to the tapeset of the ,
startup level controller and will remain so continuously, For 10 seconds the level demand signal is that signal corresponding to the
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- tapeset of the master level controller, it then drops to a level demand signal of approximately 178". f or _10 seconds the level demand signal is that signal corresponding to.the tapeset of the startup level controller, it then drops to a level demand signal of -
approximately 178"..
QUESTION: 028 (1.00)
Assuming thrt the plant is operating at 100% Rx power with AEGTS Fan "A" running, which ONE of the following conditions would provide an automatic start of the "B" Annulus Exhaust Gas Treatment System? Annulus pressure sensed as being -0.50" H: b, Low flow sensed on the discharge of the "A" AEGTS fa De energizing the Division 2 RHR LOCA Relay 1E12 K110 Arm and depress the High Pressure Core Spray Manual Initiation pushbutto . .. .- -. - . - , - - -
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SENIOR REACTOR OPERATOR Page 19 OUESTION: 029 (1.00)
The following plant conditions exist:
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Rx power 100 %
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Busses L11 & L12 are being powered from the Unit 1 Aux. Transformer-
-- Bus L10 is being supplied by the Unit 1 Startup Transformer
- Div 1 & Div 3 DGs are in standby readiness
- Div 2 DG is in " LOCAL" running fully loaded for a WO retest
- Preferred Source Breakers for EH11, EH12, and EH13 are all closed.
With these plant conditions, a bus L10 lockout occurs, a scram signalis generated due to the trip of both Rx Recirc pumps and level 2 is reached. After auto initiating, the HPCS pump motor catches fire and causes a lockout to occur on bus EH-13. The lowest RPV level achieved during this transient is 100". Which of the following statements correctly describes the steady state conditions you would expect to see following this transient? Div.1 DG is the only DG running AND supplying its respective bu Only Div.1 & 3 DGs are running, supplying their respective busses, Div. 2 DG will auto start, but not load when its local / remote switch is taken out of LOCA All three DGs received an auto start on a LOOP signal and are supplying their respective busse .
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I SENIOR REACTOR OPERATOR Page 20 QUESTION: 030 (1.00)
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The plant is operating at 90% power and stable. The iollowing alarms are suddenly received at H13 P680:
- RPV Level 7 alarm
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Mm Flow / Feed Flow h/ismatch alarm I
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P601 alarm The plant stabilizes at approximately 92% power, with level at 196". Choose trom below the correct response for this transien Ser no the HPCS pum deset the ADS Initiation logic, Manually trip the RCIC turbine, Take the SRV control switch to OF QUESTION 031 (1.00)
A:sume that the plant is in Mode 5 with the Rx Mode Switch in REFUEL, fuel movement is occurring in the containment, and the shorting links are removed. Which of the following would result in a Rx scram? SRM Channel A fails downscal SRM Channel B fails upscal SRM Channel C period fails to a reading of zero, SRM Detector D is partially withdrawn.
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OUESTION: 032.(1.00).
The reactor is at 72% power and'65% core flow. The mismatch between recirculation loops-
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must not exceed: ;
a.- 2% of rated recirculation flo : b. . 5% of rated recirculation flow.' i % of rated recirculation flo ,
' % of rated recirculation flo '
- OUESTION: 033 (1.00)
The plant is operating at 100% power and the only equipment that is OOS is SLC pump "A".
An ATWS occurs and you are directed by the Unit Supervisor to initiate SLC "B". Upon ,
initiating SLC "B", which of the following statements co rectly describe what you would expect to occur?
4 RWCU valves will only isolate if both SLC pumps are started, RWCU will continue to ru G33-FOO1, RWCU Suction Containment inboard Isolation Valve will clos RWCU pumps will tri G33 FOO4, RWCU Suction Containment Outboard Isolation Valve will close, RWCU pumps will trip, G33-F001 RWCU suction containment inboard isolation valve AND G33-FOO4, RWCU suction containment outboard isolation valve will close, RWCU pumps will tri ,
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QUESTION:.034 (1.00)-
-The
- f511owing plant conditions exist:"
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- . The plant is in MODE ' Reactor Recirculation (RR) Pump "A".is secured for maintenanc *
-- Reactor Recirculation (RR) Pump "B"_is in' operation. - - ;
(- Residual Heat Removal (RHR) "A" is in standby readines : Residual Heat Removal (RHR):"B".is operating in t,hutdown cooling.-. -J t
--- - Reactor coolant temperature is 120*F.-
- RHR "B" pump is required to be shutdown for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> for a scheduled motor operated valve
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- stroking surveillance.' RHR "B" will be returned to shutdown cooling following the surveillanc j From- theifollowing, choose the - action required in accordance.- with Perry . Technical-Specifications. - .
c.- -. Demonstrate the operability of at least one alternate method of decay heat -f removal lwithin _one hour, b.- Place the _"A" loop of RHR in the Shutdown Cooling mode prior to removing the -
"B" loop of RHR frorn servic None, one RHR shutdov n cooling subsystem may be inoperable for up to 2
' hours for the performance of a Surveillance.
, Verify reactor coolant circulating by an alternate method in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and monitor reactor coolant temperature and pressure.
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- QUESTION: 035'(1.00)- .
Which ONE .,1 the following statements is true regarding operation of Residual Heat Removal-
~(RHR)in the Suppression Pool Cooling Mode? When 1E12 F024A(B) is open, system flow through 1E12 F024A(B) shall be maintained between 2000 and 7300 gp The RHR system shall be declared inoperable whenever it is in a secondary mode of operation EXCEPT when iri Suppression Pool Cooling, RHR Loop "B" shall not be opecated in suppression pool cooling when SPCU is in operation through suppression pool return line bypassing RHR "A". A loss of pumping power during any cperation with RHR A(B) TEST VALVE TO SUPR. POOL,1E12-F024A(B), open will cause voids to be drawn in the higher elevations of the affected loo QUESTION: 036 (1.00)
Plant Conditions:
- Containment sprays have been running (RHR A and RHR B) for 15 minutes,
- Containment pressure is 1.5 psi Drywell pressure is 2.0 psi Reactor pressure is 50 psi Which of the following plant responses describes what you would expect to occur if an operator depresses the CNTMT SPRAY A(B) SEAL-IN RESET pushbuttons? Nothing, containment sprays continu Containment sprays secure AND the RHR pumps secure, Containment sprays secure AND the LPCI Mode of RHR A and RHR B initiate, The RHR pumps continue to run, supplying flow to the containment spray
- header AND the LPCI injection flowpat .
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QUESTION: 037 (1.00j
^ -Given the following plant conditions: .
- A lockout has occurred on LH-1 B and tripped LH-1-B.-
--- Breaker H1101,-Bus'H11 Normal Supply Breaker, is openin Which ONE of the following sets of conditions must be satisfivd in order for breaker H1102, .
Bus H11 Alternate Supply Breaker, to autometically close? No lockout on bus H11 and no lockout on the Auxiliary transforme No lockout on bus H11 end no lockout on Interbus Transformer LH 1- . The ATS must be in OFF and normal voltage on the line side of breaker H110 The ATS must be in AUTO, no lockout on bus H11 and normal voltage on the line side of breaker H110 QUESTION: 038 (1.00)
The normal power supply to the plant vital inverter is bus: DiA D1B ED1A ED1B
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l SENIOR REACTOR OPERATOR Page 25 QUESTION: 039-(1.00)
Which of the following signals will cause the Of f Gas Discharge Valve (N64-F632) to isolate? Post-treat rad monitor A is reading 0, pcst-treat rad monitor B is alarming Hi-H Post-treat rad monitor A is alarming Hi-Hi, post-treat rad monitor B is alarming Hi-H Post treat rad monitor A is downscale, post treat rad monitor B is alarming Hi-Hi-H Post-treat rad monitor A is alarming Hi Hi-Hi, post-treat rad monitor B is alarming H!-Hi.
QUESTION: 040 (1.00)
While operating at 100% power, is it possible to verify the HIGH alarm setpoint on the Turbine Building West Area Radiation Monitor without actuating the local red warning light? No, depressing the Alarm Trip / Test button will peg the ratemeter and actuate the local ligh Yes, placing the Function Selector Switch in ALARM, and depressing the High Alarm button will do thi No, depressing the Fail / Check Source button will bring in the High Alarm and actuate the local ligh Yes, depressing the Horn Silence button while depressing the High Alarm button will do thi . , . - , , _ - _ . . . _, .___ _ .._ . . _ _ - , . _
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-- QUESTION: 041.li1.00) -
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- Perry was operating at 100% power when a Control Room HVAC high radiation condition was - '
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-sensed in the supply duct. : SELECT the correct response of the system components listed -
below to this conditio Supply ' Return = . Supply Exhaust - Return - _
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Fan- Fan' Damper . Damper Damper -- '
. C00 C002 F010- !F130 - F110 !
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Ope Open Closed
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5 b.' . . Run -Stopped- _ Closed ~ Closed' Closed -
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Stopped . Stopped Closed Open - Closed ,
. Run Run -- Open Closed Open QUESTION: 042 (1.00)
While doing a building inspection, a steam leak is discovered by one of the PPOs on feedwater heater #6A.- Which of the below statements describes a possible source of this steam?
' Heater 6A steam drains to the 2A MSR Drain Tank, Heater 6A steam drains to the #4 direct contact feedwater heater.
' h.g pressure turbine 4th stage extraction steam supply to heater 6 ~
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d.- = High pressure turbine 7th stage extraction steam supply to heater 6A.
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- SENIOR REACTOR OPERATOR Page 27 I l
OUESTION: 043 (1.00) ;
- A main turbine shutdown is in progress. Generator load was approximately 300 MwE when the RO at the controls noticed that main condenser vacuum was 5.1" HgA. Which ct the belou statements is true regarding these plant conditions? Operation may continue as long as vacuum does not increase to > 5.5" Hg Operation under these conditions is not allowed, perform a fast reactor chutdow Operation under these conditions is not allowed, trip the main turbine immediatel d, Vacuum should be lowered to <5" HgA but main turbine operation may continue as long as generator load does not fall below 275 Mw QUESTION: 044 (1.00)
SELECT the statement that describes the response of the Fuel Pool Cooling and Cleanup (G41)
System as a DIRECT result of reactor water level decreasing to below Level The G41 containment isolation valves willisolate (G41- F100, F140, and F145). The standby G41 pump will automatically start if the control switch is in AUT The G41 filter demin bypass valve (G41-F360) will fail open and fully bypass the filter demi The G41 filter domins willisolate by closing G41-F280, F285, F290, and F295 (Filter demin isolation valves).
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SENIOR REACTOR OPERATOR Page 28
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OVESTION: 045 (1.00)
SELECT the statement that describes the response of the inboard Main Steam isolation Valve
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Leakage Control System (MSIV LCS) to reactor pressure increasing to 25 psig af ter the system has been operating for one hour, The E32 F003 valve for each subsystem receives an open signal returning the subsystems to the bleed-off mod The heaters for each subsystem trip, the inboard blower trips, and the E32 F003 valve for each subsystem receives an open signa The E32 F001, E32 F002, and E32 F003 valves for each subsyctem receive close signals, the subsystem heaters trip, and the inboard blower trips, The heaters fut each subsystem trip and the E32 F001, E32 F002, and E32-F003 valves for each subsystem receive an open signal returning the subsystems to the bleed-off mod QUESTION: 040 (1.00)
Which of the following is a violation of Tech. Spec safety limits? Reactor steam dome pressure is 1300 psig, Core ilo.' is 13% AND lanctor power is 20%. Reat. tor steam dome pressure is 750 psig AND reactor power is 18%.
' Reactor steam dome pressure is 1020 psig AND core flow is 60% AND MCPR is 1.0 _ _ _ _ _ _ _ _ _ _ . _ . _ ._- - .- .
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SENIOR REACTOR OPERATOR Page 29 OUESTION: 047 (1.00)
Due to a level control problem on Molsture Separator Reheater 2A, a f ast unload nf the main turbine was in progress using 10114, Fast Unioad and Trip of Main Turbine. At 50% reactor power, a turbine trip signal was generated. Which of the following statements correctly describes the plant response to this transient, A reactor scram will occur due to a pressure increase up to 1065 psig, A reactor scram willoccur due to the closure of turbine stop and control valves, A reactor scram will occur due to level 3 following the collapse of voids, A reactor scram will occur due to high neutron flux following the collapse of vold QUESTION: 048 (1.00)
A Rx Scram has occurred with both Reactor Feed Pump Turbines (RFPTs) operating.- RFPT A's flow controller is in MANUAL and RFPT B's flow controller is in AUTOMATIC. The Startup i.evel Control Select switch is selected to "MFP" RPV pressure is 825 psig and decreasing at approximately 2 psig per minute. Which ONE of the following describes the plant response to SLOWLY decreasing the speed on RFPT "A" down to 1100 rpm with the intent of removing it from service? RPV level will be held fairly constant at approximately 178" by RFPT "B". RPV level will decrease as RFPT "A" is removed from service, unless the MFP is started, RPV level will be heid constant for approximatel/ 10 minutes and then increase to Level 8, tripping both RFPT RPV level will be held f airly constant at the tapeset level selected on the Master Level Controller by RFPT "B".
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SENIOR REACTOR OPERATOR Page 30 QUESTION: 049 (1.00)
The reactor is operating at 85% power when the following indications occur:
- APRM A/E UPSCALE annunciator energizes
- ROD DLOCK APRM UPSCALE annunciator energizes
+ RPV pressure increasing
- Mwe decreasing Which ONE of the following transients could result in the above? Maximum combined flow limit fails high, "A" pressure regulator fails open, then one minute later, the "B" pressure regulator slowly fails close "A" pressure regulator f alls closed, then one minute later, the "B" pressure regulator slowly fails ope During the power increase from 65% to 85% Rx power, the turbine Load Set setpoint was never increase .
SENIOR REACTOR OPERATOR Page 31 OUESTION: 050 (1.00)
Following a Rx scram on RPV level 3, the P680 operator is directed by the Unit Supervisor to restore RPV water level between 185" 215" using the feed system. While restoring level, the P680 operator notes level at 140" and notices that Rx Recirc pump A is still running in f ast speed. What is the affect on the reactor coolant system of the failure of the pump to downshift? Tripping one recirc pump will add suf ficient negative reactivity to overcome the positive reactivity that would have been added had the main turbine tripped at full power at the end of core life, Insulficient waterleveland/or subcooling of the waterin the downcomer region may have caused cavitation to occur somewhere in the recirc "A" system, With level below the height required for natural circulation there is insuf ficient mixing due to forced circulation such that significant thermal gradients would for It can be assumed that Rx Recirc pump "A" would have tripped while at power had the delta temperature between the steam dome and the bottom head drain '
exceeded 100' QUESTION: 051 (1.00)
A main turbine trip occurred at 45% Rx power, but c've to a malfunction of RPS, a Rx Scram f ailed to occur. Rx pressure increased to 1083 psig. SELECT from below the ONE statement that correctly describes the Redundant Reactivity Control System response to this transient 25 seconde after reaching 1083 psig, assuming Rx power remains constant at 45%. Alternate rod insertion will actuate, The reactor recirculation pumps will trip off, Standby Liquid Control will automatically initiate, Feedwater flow is reduced to zero, unless the feedwater pump controllers are in manual, i
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QUESTION: 052-(1.00)
A reactor startup was in progress. Power was approximately 2500 cps on all four source ';
range monitors when a malfunction in RC&lS occurred, resulting in a continuous withdrawal
-- of a four rod gang, and causing a Rx period of 30 seconds. Which of the following statements ;
correctly describes the implications of this event? This is an indication of excess positive reactivity and the operator should insert a manual Rx scra i Rx period will eventually return to between 60150 seconds and therefore no !
operator action is necessar , This is the design basis event behind the Rod Pattern Controller's banked -
position withdraw sequence (BPWS). This event is more significant if power is initially at the point of adding heat than if power is initially in the source rang QUESTION: 053 (1.00)
Given the follov4g plant conditions:
- An ATWS has occurre ;
- Rx power is steady at 23%.
- The Main Turbine has trippe Several control rods are stuck out at various position The Unit Supervisor directs you to insert control rods using PEl SPl, Section Why is it necessary to bypass the Low Power Setpoint? To bypass the two notch limit, allowing continuous insertion of control rods,
' To bypass the four notch limit, allowing continuous insertion of control rod To bypass the bank limits that are in effect because power is being sensed below the LPS Though not actually needed now, power will eventually decrease to LPSPduring rod insertion, requiring the bypas ,
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.I QUESTION: 054 (1.00) i Sometime af ter e smailloss of coolant accident, the following stable plant conditions are found ;
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- Reactor Power 0%
- Reactor Level 180"
- Reactor Pressure 1000 psig l'
- Containment Pressure 4.5 psig
- Drywell Pressure 4.3 psig !
- - Suppression Pool Level 21 f :
- - Suppression Pool Temperature 115' l Which one of the following is being exceeded? Heat Capacity IWit l SRV Tailpipe t.ws UM l
' Pressure Suppression Pressure No PEl SPI curves are being exceede ;
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QUESTION: 055 (1.00)
The Rx is operating at 100% power when a spurious Main Turbine trip results in a Rx pressure excursion up to 1100 psig. What is the impact on the Feedwater Level Control System? . Operation of SULC in MANUAL is available after 30 second , Operation of all controllers is automatically restored in 12.5 minutes, Operation of RFP Controllers is available in MANUAL for the first 30 seconds, The Master Level Controller will stay at its tapeset demand signal for 10 seconds, i
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r OUESTION: 056 (1.00) :
PEl B13, RPV Control, has been entered on low reactor water level. Current water level is [
+ 15 inches and slowly decreasing. RPV pressure is 120 psig. Which of the following ALTERNATE INJECTION SYSTEM (S)/ SUBSYSTEM (S) can be used for RPV injection for the present plant conditions? *
~ ' RHR Loop B Floo {
' SPCU alternate injectio Hotwell Dump Line alternate injectio Condensate transfer alternate injectio l QUESTION: 057 (1.00)
A plant startup is in progress with power at 19% with two Circ Water Pumps operating. A problem with the Steam Jet Air Ejectors causes Main Condenser Vacuum to decrease to 7" HgA. The problem was corrected and vacuum returned to normal. Assuming no operator action, which of the following would be correct after the transient? Reactor power less than 19% due to the Recirc Flow Control Valve runbac Reactor power would be approximately 19% with the Main Turbine still operatin The Main Turbine tripped on low vacuum with the Bypass Valves controlling Reactor Pressur " The Motor Feed Pump controlling reactor level due to the trip of a Reactor Feed Pump Turbine on low vacuu .
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SENIOR REACTOR OPERATOR Page 35 QUESTION: 058 (1.00)
High Pressure Core Spray (HPCS) and the Motor Feedwater Pump (MFP) weh .1anually started -
during a loss of feedwater flow transient to recover level. SELECT the statement that describes the plant response as RPV water levelincreases to the Level 8 setpoint, The HPCS pump and MFP will tri The HPCS injection valve will close and the MFP will tri ,
. The HPCS pump will trip, the MFP will operate until manually secure The HPCS injection valve will close, the MFP will operate until manually ,
secure QUESTION: 059 (1.00)
PEl T23, " Containment Temperature Control", requires termination of containment sprays if containment pressure is below 1.75 psig. SELECT which statement below describes the basis for this step, Terminating containment spray at this pressure avoids containment f ailure due to negative pressure, Cooling the containment with the containment aprays will drive the RPV i saturation temperature below the curve into the unsafe region and make level instrumentation inaccurat Continuous operation of the containment sprays will decrease the margin to HCL, Subsequent Emergency Depressurization may be required to stayin the safe region of the HC Since isolations in the secondary containment may have occurred, a Technical Specification limit r>f 0.1 to 1.0 psid between the Containment and Auxillary Building.cannot be assured, Jeopardizing the Auxiliary Building,
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Given the following plant conditions: J
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- A LOCA Ess occurre j
- RHR A, B, & C were manually initiated in the LPCI mode one minute ago per PEl SPI 5.2, injection Preventio MFP is maintaining RPV level 50100".
- SP Temperature is 100*F. and increasing.-
When is the earliest time that you can fully realign RHR "A" to Suppression Pool Celing? Due to the initiation of RHR in the LPCI rnode, RHR "A" cannot be lined up in the SP Cooling Mode for nine minutes, PEl-T23 will not at:ow RHR to be realigned to SP Cooling at this time because l LPCI mode is required until adequate core cooling is established, PEl B13 and PEl T23 state that RHR can be realigned from the LPCI mode to SP -!
Cooling Mode only as required to stay below the Heat Capacity Limit (HCL). The RHR HX BYPASS VALVE 1E12 F048 will not fully open for 110 seconds following a LPCI signal, so 50 seconds is the earliest that SP Cooling can be s establishe >
i OUESTION: 061 (1.00) l Af ter arming and depressing the RPS manual scram pushbuttons and placing the Reactor Mode Switch in SHUTDOWN, what additional actions are to be taken in accordance with ONI C61, l Evacuation Of The Control Room, in the event a Control Room evacuation is required? trip the Main Turbine, initiate HPCS, and place Division 2 DG Control Transfer ;
Switch in LOCA :
b.- verify all rods inserted, initiate RCIC, and place Division 1 DG Control Transfer !
Switch in LOCA , verify all rods inserted, initiate RCIC, and place Division 3 DG Control Transfer Switch in LOCA d.- _ verify all rods inserted, trip the Main Turbine, and place Division 3 DG Control Transfer Switch in LOCA ;
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I SENIOR REACTOR OPERATOR Page 37 !
OUESTION: 062 (1.00)
Which ONE of the following statements is correct concerning a complete loss of Nuclear
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Closed Cooling water? The Reactor is shutdown in anticipation of...
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a, the CRD pump tripping on a high lube oil temperature of 195a I the Drywell pressure increasing to 1.68 psig due to the loss of Drywell coolin : RWCU Filter Demins. Isolating on high NRHX Inlet temperature and the i associated loss of Reactor chemistry, i
' tripping the Reactor Recirculation Pumps on high bearing temperature and the -
loss of instrument and Service Air compressor I QUESTION: 063 (1.00)
SELECT the statement that describes a condition requiring a f ast reactor shutdown following !
a loss of instrument air while operating at 45% power, per ONI PS2, Loss of Servict and/or instrument Ai , The INST VOL NOT DRAINED annunciator is receive The SCRAM VLV AIR HEADER PRESSURE LOW annunciator is receive The outboard main steam line isolation valve (MSIV) 821- F028A drifts close , The ROD DRIFT annunciator (for rod 2211) and CRD MECHANISM TEMP HIGH
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annunciator (for rod 46 39) is receive .
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SENIOR REACTOR OPERATOR Page 38 j t
OUESTION: 064 (1.00)
The plant is in Mode 2 performing a Rx Startup when the running CRDH pump trips. Which l of the following describes the plant conditions that would require placing the Rx Mode Switch !
in SHUTDOWN immediately, per ONI C11 1, Inability To Move Control Rods? i
, Rx Pressure is 500 psig. Accumulator fault on rod 30-31 at 0" comes in. CRD Charging Water Pressure is 1570 psig, t Rx Pressure is 920 psig. Accumulator f aults on rod 30 31 at 0" and rod 20-27 at 24" has been in for 20 minutes. CRD Charging Water Pressure is 1850 psi c.- Rx Pressure is 900 psig. Accumulator f ault on rod 20 27 at 24". CRD Charging l Water Pressure is 1500 psi ; Rx Pressure is 550 psig. Accumulator f ault on rod 20 27 at 24". CRD Charging -
Water Pressure is 1575 psi ;
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SENIOR REACTOH OPEHATOR Page 39 OUESTION: 065 (1.00)
Given the following initial plant conditions:
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Current Reactor Power: 65 %
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Current rod line: 105 %
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Current total core flow: 60 Mlbm/hr
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Reactor Recirculation Flow Control in LOOP manual The "A" Recirculation loop flow controller then f ails, causing the "A" flow control valve to close, decreasing core flow to 41 Mlbm/hr. SELECT the appropriate IMMEDIATE action to be taken per ONI C51, Unplanned Change in Reactor Power or Reactivity, as a result of these piant condition insert Cram rods per FTI 802, Control Rod Movements, to exit the IMMEDIATE EXIT regio Immediately scram the reactor by arming and depressing the RPS MANUAL SCRAM CH A, B, C, and D pushbuttons, Exit the MANUAL SCRAM REQUIRED Region of the Power to Flow map by increasing flow in the "B" Recirculation loo Establish a load line less than or equal to the 100% load line, then shutdown one Reactor Recirculation Pump per S0183 SENIOR REACTOR OPERATOR Page 40 QUESTION: 066 (1.00)
The following plant conditions exist:
- A small break LOCA and ATWS have occurred
- Reactor power is 8% and slowly decreasing
- Doth SLC pumps are running
- 32 control rods f ailed to fully insert
- RPV level / pressure is 90"/900 psig
- Suppression pool temp. / level is 130'F /18.6 feet
- Drywell temp. / pressure is 150*F / 0.3 psig
- Cntmt. temp. / pressure is 100'F / 0.2 psig What is the required operator action per the Plant Emergency Instructions? Restore and maintain Suppression Pool level between 17.8 and 18.5 fee Bypass the NCC lsolation using PEl SPl 2.1 to restore NCC to the Drywell cooler Enter PEl 813, RPV Coritrol (ATWS) at "X" to enter PEl B13, Emergency Depressurizatio Manual!/ control reactor pressure using SHVs as necessary to maintain RPV pressure below HCL.
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OUESTION: 067 (1.00) !
i Containment Flooding has been entered from RPV Flooding. Current plant conditions are as j follows: !
- Reactor level UNKNOWN i
- Reactor pressure O psig *
- Suppression pool water level 62 feet .['
-- Containment pressure 38 psig
- All Control Rods inserted "00"
- The hotwell purnps are injecting 4 Mibm/hr into the reacto !
What actions are required based on these plant conditions?
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" Terminate injection to the RPV from the Condensate Syste Exit Containment Flooding aad go to RPV Level Control (START). Decrease Condensate flow to lower Suppression Pool Water level to 61 fee d .' -
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Maintain Suppression Poollevel above 61 feet with Condensate System taking l a suction external to the containment.
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SENIOR REACTOR OPERATOR - Page 42 QUESTION: 068 (1.00)
Following a LOCA, the following parameters are noted: .
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RPV Pressure 100 psig
- Drywell Temperature 250' Containment Temperature 125' RPV LEVELS:
1) Narrow Range 180" 2) Shutdown Range 190" 3) Upset Range 195" Which of these level instruments may be used to determine level? All of these instruments can be use ' None of these instruments can be use . Only the narrow range is providing valid level indicatio Only the shutdown range is providing valid levelindicatio QUESTION: 069 (1.00)
The plant is shutdown in Mode 4 with RHR "B" in Shutdown Cooling and RHR "A" is OO RHR Pump "B" shaft seizes. Which ONE Of the following is an approved alternate means of decay heat removal per ONI E12 2, Loss of Decay Heat Removal?
- Operate High Pressure Core Spray (HPCS) to circulate coolant between the suppression pool and the reactor vessel via the head ven If Rx coolant temperature is > 190'F., align the RPV head vents by opening RX HEAD TO DW FIRST/SECOND VENT VALVES,1B21 FOO2/F00 Operate the condensate system to circulate water between the main condenser
. and the reactor vessel to the hotwell pumps via the Main Steam isolation Valves (MSIVs), Operate Low Pressure Core Spray (LPCS) to circulate the coolant between the suppression pool and the reactor vessel with the LPCS via two Safety Relief Valves (SRVs).
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. 1 SENIOR REACTOR OPERATO Page 43 )
l OUESTION: 070 (1.00) .
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I During refueling activities, a decrease in upper containment pool level requires suspension of ,
core alterations after placing all irradiated fuel and core components in a safe condition, {
Which of the following is a SAFE CONDITION for a fuel bundle in the IFTS carriage? l t Bundle properly seated in the IFTS carriage at the Bottom Out position with the upender vertica !
l b. . Bundle properly seated in the IFTS carriage at the Raise Lower Limit position l with the opender incline ,
I Bundle properly seated in the IFTS carnage at the Raise Lower Limit position with the opender vertica : A bundle cannot be in a SAFE CONDITION in the IFTS carriage and must be i placed in a pool storage locatio l
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OUESTION: 071 (1.00)
- The Maximum Safe Operating Condition Values for Area Temperatures found in PEl N11, Containment Leakage Control, are based on: Equipment necessary for safe shutdown of the plant will fail above this temperatur < Based on the high room temperature alarm setpoints for the surrounding containment areas, The availability of alternate instrumentation use in the case of failed installed l Instrumentatio !
' ' Personnel access nocessary for the safe operation of the plant willbe precluded :
above this temperatur l
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SENIOR REACTOR OPERATOR Page 44 !
h OUESTION: 072 (1.00)
Due to a LOCA and a steam leak in the RCIC pump room, PEl-B13 RPV Control (Non ATWS), l and PEl N11, Containment Leakage Control, have been entered. The RCIC pump room is l-inaccessible and steam makes observation of the room impossible. However, the following )
indications are available to the operators: RCIC Pump Room Area Temperature is 295'F The ;
RCIC Pump Room Sump bl level alarm has been received RWCU Pump Room Area Temperature is 265' Which ONE of the following conditions would require the or < ators to enter PEl B13, RPV j
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Control (Non ATWS) at "Z" with the intentions of performing Emergency D3 pressurization?
i a.- RCIC Area HVAC Diffecential Temperature alarming and reading 95'F j r Aux Building Ventilation Exhaust Gas radiation monitor reading 34,050 cpm ; Aux. Building 574' elevation Hallway Area Radiation rnonitor reading 4,200 mR/h ;
d_ Aux. Building 574' elevation Hallway water level observed at 20" above the floor !
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,l SENIOR REACTOR OPERATOR Page 45 )
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OUESTION: 073 (1.00)
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As the Unit Supervisor, you have been brought the following examples of tagouts that require your approval. Select the example below that violates the tagging rules defined in the PAP 1401, Safety Taggin !
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ai Tagout #1 has red danger tags and yellow grounding tags being hung on the j same componen ;
i Tagout #2 has a red danger tag attached to a solenoid valve that is to be
' removed, bench tested, then reinstalled into the syste .
- Tagout #3 has e single danger tag and a portable ground isolating a component j
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from its 500 volt AC power supply because only one break is available,
! Tagout #4 has a danger tag being used to provide separation between Unit 1
and Unit 2, and requires a 10CFR50.59 Applicability Check prior to your authorizatio j
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QUESTION: 074 (1.00)
Assume that the plant is in a Site Area Emergency and you are the Shif t Supervisor acting as the Emergency Coordinator. It becomes necessary to send an individual into an area that will ,
exceed his 10CFR20 dose limits in order to save another individual's life. Per PAP-0514, Perry Plant Personnel Radiation Dose Control Program, the dose should be limited to the following when lower doses are not practical; rem TEDE , rem TEDE l c.- 15 rem TEDE rem TEDE
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t SENIOR REACTOR OPERATOR Page 46 - l I
i OUESTION: 075 (1.00) ,
in accordance with PAP 0201, Conduct of Operations, which of the following individuals is :
. allowed to make changes to the reactor recire flow control valves positions at full power i without direct supervision? . the recirc system engineer to check valve response, providing the consent of ,
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the licensed operator at the controls and the unit supervisor is obtaine , . an unlicensed trainee e' illed in a licensed operator training program providing [
the concent of the lic6nseJ operator at the controls and the unit supervisor i '
obtaine c,- the oncoming unit supervisor, providing .the knowledge.and consent of the q licensed operator at the controls is obtaine ._
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i the oncoming SO at the controls, providing the knowledge and consent of the licensed operator at the controls and the unit supervisor is obtaine ,
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I OUESTION: 078 ('1.00)
Upon accessing an ERIS screen in the control room, you notice a white asterisk in the value ,
box for a parameter. Which ONE of the following statements correctly describes how this _
should be interpreted?. The data that goes in that box has not been validated The parameter that normally provides input to that box has been lost .i The data is good but there is not enough room in the box to display the data
, The data that goes in that box is normally compensated data but is not currently being compensated
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l SENIOR REACTOR OPERATOR Page 47 1
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OVESTION: 077.(1.00)
The Reactor Recirculation pumps have just been shif ted to f ast speed, immediately after the pump shift, the "LPCS OUT OF SERVICE" alarm annunciatesi and upon investigation, you determine that this alarm is due to a LPCS line break. What caused this alarm to come in? ;
, Excessive differential pressure between RHR "A" LPCI nozzle and the LPCS .
Spray Sparger.- -
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b.- Excessive differential pressure between RHR "B" LPCl nozzle and the LPCS Spray Sparger. - Excessive differential pressure between the vessel downcomer regio a and the LPCS Spray Sparge Excessive difforential pressure between the LPCS Spray Sparger and the above !
core plate pressure ta _
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QUESTION: 078 (1.00)
During a plant startup at 5% reactor power, the on shift Chemistry Technican reports the following 'esults with regards to the Standby Liquid Control Storage Tank per SVI C41 T1026:
. NET TANK VOLUME - 4475 gallons % WElGHT SOLUTION CONCENTRATION 13% .
What ACTION (S) are you required to take with regards to the SLC System? No action is required; the SLC system is OPERABL Restore the SLC System to OPERABLE status within 7 day Have the plant shutdown to at least MODE 3 within the next 12 b ur d. . Restore at least one subsystem to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at
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least MODE 3 within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ,
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j SENIOR llEACTOR OPERATOR Page 48 i
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OUESTION: 079 (1.00)
i The plant has experienced a LOCA. Per PEl M51/56, Hydrogen Control, the H2 Analyzers and the H2 Ignitors l' 'e been started. An electrical transient causes power to be lost to Di +
and Div. 21gnitor. 9naware that the Igniters have do-energized, the operators restore power i to the busses approximately two hours af ter it was lost. What effect,if any, could this have ,
on the Containment and/or Drywell? ;
! Restoring the igniter power supply will have no impact on the containment because the ignitors will not start unless restarted by the operators, If hydrogen concentration has continued to increase while power was lost, ;
restoring the igniter power supply could result in excesr!ve Cntmt. pressure t threatening Cntmt. Integrit t If hydrogen concentration has continued to increase while power was lost, restoring the igniter power supply will result in localized combustion resulting in little impact on Cntmt. pressure, Restoring the igniter power supply will have no impact on the containment ,
because it is expected that during a LOCA the containment will be in a steam i inert conditio ,
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SENIOR REACTOR OPERATOR Page 49 OUESTION: 080 (1.00)
Given the following initial plant conditions:
- RHR loop "B" is in the Shutdown Cooling (SDC) Mode
- Reactor coolant temperature is 300* RPV pressure is 65 psig.
SELECT the statement that describes the effect on the Shutdown Cooling (SDC) Suction isolation Inboard and Outboard Valves (1E12 F009 and IE12 F008) if 4.10 KV Bus EH12 de-energizes due to a bus lockout conditio E12 F008 and IE12 F009 will both isolate when pressure reaches 135 psig, E12 FOO8 and 1E12 F000 will not isolate at 135 psig due to a loss of powe E12-F008 willisolate when pressure reaches 135 psig,1E12 F009 will not isolate due to a loss of powe E12 F009 willisolate when pressure reaches 135 psig,1E12 F008 will not isolate due to a loss of power.
QUESTION: 081 (1.00)
The plant is at 100% power. Due to a problem with the Reactor Core Isolation Cooling System, the Unit Supervisor declares the RCIC System inoperable at 0930 on June 1.
Because the High Pressure Core Spray System was also Inoperable, the Unit Supervisor also entered Action 3.0.3 at 0930. A Main Condenser vacuum problem occurred, causing the operators to rapidly unload the turbine at 1000 and eventually place the Reactor mode switch in Startup at 1040. Based on these plant conditions, at the latest, (date and time), when must the plant be in Hot Shutdown (assume neither RCIC nor HPCS can be repaired)? June 1,1630, June 1,213 June 1,223 June 2,103 . -
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SENIOR REACTOR OPERATOR Page 50 t
i OUESTION: 082 (1.00)
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RACS channel 1 is selected for display. IRM B which is assigned to RACS channel 2 fails upscale. No other IRMs are alarming. Which one of the following correctly describes the
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indications the operator will see? [
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a, No rod block present, Withdraw Block indicator lighte No rod block present, Withdraw Block indicator unlighte ,
! Rod block present, Withdraw Block indicator lighte l Rod block present, Withdraw Block indicator unli0 hte ,
QUESTION: 083 (1.00) .
While operating at 100% power you receive the Hot Surge Tank Level Hi annunciator on 1H13 P680. Upon silencing the annunciator, you observe Hot Surge Tank Level go offscale high on recorder 1N21 R323, HOT SURGE LEVEL & CNDS TO HTR 4 FLOW. Assuming that you are unable to stop Condensate flow to the Hot Surge Tank, which of the following statements correctly decribes the correct action to take per Sol- N21? ' Manually isolate Feedwater Heater #4 prior to its auto isolating on High HST Leve Manually trip all Hotwell and Condensate Booster Pumps prior to 120 psig HST pressur Manually trip all Feedwater Booster Pumps and Rx Feed Pumps prior to 120 psig HST pressure, Decrease condensate flow to the HST by securing one of the Condensate Booster Pumps and manually throttling 1N21-F23 ,
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QUESTION:. 084_(1.00)- ,
The plant is in' a condition of Loss of Offsite Power (LOOP). Which ONE of the following l correctly describes a condition whic'1 would cause the plant to degrade into a Station Blackout : !
(SBO)?- l . Allromainirig Divisional DieselGenerators fall, causing a de energization of their ;
respective buse i b'. ' ~ Division 3 Diesel Generator' falls, resulting in a de energization of Bus EH1 ;
_I' Only Div.1 Diesel Generator remains running, resulting in only Bus.EH11
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eno63l red, d.: .Tlee remaining Div.1 OR 2 Diesel Generator fails, resulting in only bus EH13 i energize ..
QUESTION: 085 (1.00) .
A scram has occurred; however, during the transient, it is discovered that station loads have :
f ailed to shif t. Which ONE of the following statements correctly describes the actions you are ;
required to take per ONI C71 1, Rx Scram? ; Close START UP SUPPLY BRKRS, L1006 and L1009, enter ONI R22 2, Loss of Non-Essential-13.8KV/4.16 KV bu ;
I Open NORMAL SUPPLY BRKRS, L1102 and L1202, enter ONI R22 2, Loss of
- Non-Essential 13.8KV/4.16 KV bus, ; Verify AUXlLIARY TRANSFER SWITCH is in the AUTO position, then verify the
- START-UP SUPPLY BRKRS close and NORMAL SUPPLY BRKRS ope I I Restore power to at least one 13.8 KV bus by opening its associated NORMAL i
SUPPLY BRAR and clostrm its associated START UP SUPPLY BRK '
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- SENIOR REACTOR OPERATOR Page 52 :
i OUESTION: 086 (1.00)
l During refueling operations, a fuel assembly was dropped while suspended in the Upper containment pool. The fuel handling crew reported that a large quantity o; bubbles were observed coming to the water surface but no area high radiation alarms were present. Whet are the required immediate operator actions per ONI-J112, Fuel Bundle Rupture During Fuel ,
Handling?
l Suspend all Core Alterations and movement of fuel, and evacuate the Containment upper pool area, Suspend all Core Alterations and movement of fuel, and evacuate the containraent if a high radiation alarm occur t c.- Suspend all Core Alterations and movement of fuel, sound the containment ,
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evacuation alarm, and start containment purg Suspend all Core Alterations and movement of fuel, evacuate the containment of all unnecessary personnel, and increase fuel pool coolin .
QUESTION: 087 (1.00)
An unisolablo steam line break has occurred in the Steam Tunnel and has resulted in a potential off site release. The Emergency Coordinator has declared a Site Area Emergenc Which ONE of the below statements describes a correct action required to be taken for this situation? Within 15 minutes, notify the VP Nuclear, Within 16 minutes, notify the State of Ohio and provide protective action recommendations, Within one hour, notify the NRC Within one hour, notify the Ashtabula County officials
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SENIOR REACTOR OPERATOR Page 53 OUESTION. 088 (1.00)
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With the reactor shutdown, which of the following sets of riant conditions ensures the core is adequately cooled by STEAM COOLING? Reactor pressure is 125 psig, reactor levelis 15 inches, and the HPCS Pump '
is running and injectin Reactor pressure is 135 psig and reactor level is +16.5 inches and the Condensate System is injecting into the reacto Reactor pressure is 850 psig, reactor level is + 10 inches, and no injection subsystems or alternate injection subsystems are running'. Reactor pressure is 450 psig, reactor irvel is -25 inches and no injection subsystems or alternate injection subsystems are runnin QUESTIGN: 089 (1.00)
A LOCA has occurred and PEl T23, Containment Temperature Control, directs the operator to enter PEl-B13, RPV Control (Non ATWS),if Containment Temperature cannot be maintained i < 185'F. Which of the below statements correctly describes the reason for this action? This WB1 minimize containment heating and provide a margin to the contalnment desig8) p*gssur This assures that, if possible, the reactor is scrammed before emergency depressurization is initiated, Further release of energy from the RPV to the containment is minimized by rapidly depressurizing the RP This will terminate containme.nt temperature increase and thereby maintain equipment operability for as long as possibl .
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i OUESTION: 090 (1.00)
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SELECT the ONE statement that describes the actions required to be taken if the followir,g plant conditions occur while testing safety relief valves. PLANT CONDITIC NS:
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+ . Reactor mode switch position: "STARTUP"
- Reactor coolant temperature: 540* ~
Reactor power: 2%.
- Suppression pool temperature: 97'F. (has exceeded 95'F for1 hour)
- Suppression poollevel: 17 feet 8 inche Residual Heat Removal (RHR) system loop "A" is in suppression pool cooling mod l Stop all testing which adds heat to the suppression pool, Reduce suppression pool water temperature to less than 95'F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in MODE 4 within the following 24. hours,
' Verify suppression pool water temperature is s110*F once per hour and restore suppression pool water temperature to 595'F within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Restore suppression poal level to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least MODE 3 with' . ine next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> QUESTION: 091 (1.00)
During a LOCA, indication of level has been lost, but SRV tallpipe temperature indicates =
340*F. on recorder 1B21 R614 located on H13 P614. This may be indicative of: Inadequate core cooling, Feedwater should be cut back, SRV has been open for too lon Increased efficiency of heat transfe Y
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SENIOR REACTOR OPERATO Page 55 f
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QUESTION: 092 (1.00) :
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Concerning the use of the HPCS System during an ATWS event, SELECT the correct statement.
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' HPCS injection is not allowed if reactor power is above 4%. At no time is HPCS injection allowed during an ATWS conditio . c, HPCS injection into RPV is allowed if RPV-level cannot be restored and i
maintained above 25 inches.~ HPCS injection is allowed if Feedwater, RCIC, and CRD cannot maintain RPV level between and 0 and 100 inche QUESTION: 093 (1.00) ,
Given the following plant conditions:
- PEl B13 (ATWS) has been entere The reactor f ailed to scra SLC is injectin '
-
SELECT the statement that describes when PEl-B13 (ATWS) can be exited and PEl B13 (Non-ATWS) can be entere Power is on IRM range 4 and decreasing, all SRVs are close The SLC storage tank level is 190 gal, and the SLC pumps have trippe All control rods are fully inserted, power is on IRM range 2 and decreasin i The reactor engineer confirms the reactor will remain shutdown with the present rod configurstion, as long as the boren concentration remains 2:500 ppm in the RP i
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SENIOR REACTOR OPERATOR Page 56 OUESTION: 094 (1.00)
While venting the Containment per PEl M51/56 to restore hydrogen concentration to below HDOL, which of the following conditions requires venting to be secured? Containment pressure is 8 psig and Containment hydrogen is 8%. Radiation levels at the site boundary increase to 1 mr/hr above normal backgroun Containment pressure is less than or equal to 1.75 psig and Containtnent hydrogen cannot be determine Adequate Core Cooling cannot be assured and the radiation level at the site boundary is 0.5 mr/hr above normal background.
QUESTION: 095 (1.00)
During a LOCA condition, an automatic initiation of RHR A & B in the containment spray mode is received. SELECT the correct statement concerning operation of the RHR systems in this mode, RHR Pump flo'n should b 6900-7'800 gpm during containment spray operation With a High Drywell Pressure signal present, the containment spray logic cannot be reset irrespective of containment pressur Realignment from the containment spray mode to suppression pool cooling cannot be accomplished without a reset of the containment spray initiation logic, When containment pressure decreased to less than 1.5 psig, containrrent spray would be terminated per the PEI by overriding closed the E12-F028 and the E12 F537 valves (CONTAINMENT SPRAY FIRST AND SECOND SHUTOFFS).
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SENIOR REACTOR OPERATORL ~ .l G
i m
QUESTlON:'096 {1.00)- ;
'
An ATWS has occurred and Rx power is 37%. Several control rods are stuck at various-- l
- positions in the core and neither Standby Liquid Control pump will start. Which ONE of the
- below actions would ' serve to decrease Rx power? - Maintain Rx level using the Motor Feed Pump betweO .0- 100". j Commence a controlled depressurization not to exes y :.00'F per hou , Sta t both Rx Recirc pumps in slow speed to increase circulation through the -
. Cor ,
'
id . . If no R'x Recirc pumps are running, raise RPV level to 2240" to provide natural circulatio i i
OUESTION: 097 (1.00).
-
SELECT the system that will NOT UTOMATICALLY function if required following a loss of Bus ED1 High Pressure Core Spray (HPCS)
i Turbine Bearing Emergency Oil Pump c .- Reactor Core isolation Cooling (RCIC) ' Automatic Depressurization System (ADS) ,
.9
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SENIOR REACTOR OPERATOR Page 58 OUESTION: 098 (1.00)
The plant is shutdown to Mode 4 with RPV water temperature being maintained 100'F-140'F by RHR "A" h shutdown cooling. l&C is currently performing a Level 3 SVi on Narrow Range Level Channel "A". The Unit Supervisor is relieved by the Shift Supervisor, who allows l&C Technicians to begin Surveillance Testing on RPV Narrow Range Level Channel "B". During performance of their SVis, both insert a RPV Level 3 signalinto their channels. What could be the possible consequences of doing this? RHR Pump "A" will tri RHR Pump "A" Min. Flow Valve,1E12-F064A will open, RHR Pump "A" Shutdown Cooling Suction Valve,1E12- F006A will clos RHR Pump "A" Test Valve to the SP,1E12-F024A can now be opened by the operator on P601.
QUESTION: 099 (1.00)
The Rx is operating at 100% power when CRDH Pump "A" trips. What action, if any, is required to be taken with respect to the Rx Recirculation System? Shift both Rx Recirc Pumps to slow speed and maintain seal temperatures
< 160' Rx Recirc pumps are not allowed to be operated without seal purge in operation, trip both Rx Recirc Pumps to OF Rx Recirc Pump operation may continue, however seal purge should be shifted to its alternate supply via the Mixed Bed Syste Rx Recire Pumps can be operated indefinitely without seal purge, operation may continue but seal purgo should be restored as soon as possibl .~..-. - . -, . . . - . . . ~
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LSENIOR REACTOR OPERATOR _- : Page 59
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i J OUESTION: 100,(1.00)
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! The plant has experiencH a LOCA with an ATWS. It has been determined that an unisolable j steam leak has occurred in the RCIC pump room. However, RCIC is the only system capable 1!
.of injecting into the RPV.LWhile fighting this casualty, the control room receives the AX 574' :l
- Area Radiation. Monitor alarm on H13-P803 and it is reading 4.5E3 mremihr. From this information, which of the following statements would be correct? . it can be assumed that a primary systern is NOT discharging into the affected area.- ,
> RCIC pump room area must be verified at >4 rad /hr by survey to Emergency Depressuriz c.- . RCIC area HVAC differentia; temperature must be > 103*F to shutdown the:
~
.
RCIC pum It can be' assumed that two area radiation levels are greater than their Ma i Safe Operating Leve >
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(* * * " ' ' * * * END OF EXAMINATION * * * * * " ' ' ')
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(SENIOR REACTOR OPERATOR _ ; Page 60; {
q 1 ANSWER: 001 (1.00) ANSWER: '007- (1.00) .
J REFERENCE: REFERENCE:
PAP-0201, 6.8. PAP 1910, Sect. 6.4-LP: OT-3039 01, Obj. A ---
LP: OT-3039 OO, Obj. B -
294001K101 ._. (K A's) -294001K116 .. (K A's) '
l L
- ANSWER: 002 (1.00) ANSWER: 008 (1.00) , REFERENCE: - REFERENCE:
PAP-1401 PAP-0201 LP: OT-3039 01, Ob LP: OT-3039-01, Obj. A _
'294001K102 ..(KA's) 294001A102 .. (KA's)
ANSWER: 003-(1.00) - ANSWER: 009 (1.00) REFERENCE: REFERENCE:
- PAP-0514 PAP-0110 LP: OT-3039 00, Obj. B LP: OT-3039 01, Obj. A 294001K103 ..(KA's) 294001A103 .. (KA's)
ANSWER: 004-(1.00) ANSWER: 010 (1.00)
- REFERENCE: REFERENCE:
PAP-0118, Sect. 6. OAl 1702 294001K104 ..(KA's) LP: OT-3039-00, Obj. B 294001A106 .. (KA's)
ANSWER: 005 (1.00) ANSWER: 011 (1.00)
REFERENCE: PAP-0219, Sect. 6. REFERENCE:
LP: GEN 1001 Tech Spec. 3.4.'i 1
--
294001K105 ..(KA's) LP: OT-3037-08, Obj. D 294001A108 .. (KA's) .
'
ANSWER: 006 (1.00) ANSWER: 012 (1.00)
' REFERENCE:- PAP-0504 REFERENCE:
LP:' OT-3039 OO, Obj. B SDM: M28 294001K107 ..(K A's) LP: OT 3036-M28, Obj. C 294001A113 .. (KA's) '
' ,
Y s,m - --,s
SENIOR REACTOR OPERATOR Page 61 ANSWER: 013 (1.00) ANSWER: 019 (1.00) REFERENCE: REFERENCE:
Emergency Plan Instruction EPl-A2, section SDM: C71 LP: OT-3036, C71, Obj. D LP: EPL-0823 C22 OT 3036-C22, Obj. D 294001A116 .. (K A's) 212000A108 ..(KA's)
ANSWER: 014 (1.00) ANSWER: 020 (1.00) REFERENCE: REFERENCE:
SDM: C11(CRDH) SDM: C71 LP: OT 3036 C11 (CRDH), Obj. C LP: OT 3036-C71, Obj. D 201001K201 ..(KA's) 2120dOK201 ..(K A's)
ANSWER: 015 (1.00) ANSWER: 021 (1.00) REFERENCE: REFERENCE:
SDM: C11(RC&lS) SDM: C51(PRM)
LP: OT-3036-C11(RC&lS), Obj. G LP: OT-3036-C51(PRM), Obj. D,1 201005K403 ..(K A's) 215005A104 ..(KA's)
ANSWER: 016 (1.00) ANSWER: 022 (1.00)-
a, REFERENCE: REFERENCE:
Ops. Policy 2 9 SDM: 821(NBPI)
OT-3039-00, Obj. 8 LP: OT-3036-821(INST), Obj. B PlF 96-3400 216000K507 ..(KA's)
202001G001 ..(K A's)
ANSWER: 023 (1.00)
ANSWER: 017 (1.00) c, REFERENCE:
REFERENC SDM: B21C SDM: E12 LP: OT-3036-821 C, Obj. E LP: OT-3036-E12, Obj. F 218000A403 ..(KA's)
203000K610 . .(K A's)
ANSWER: 024 (1.00)
ANSWER: 018 (1.00) c. REFERENCE:
REFERENCE: SDM: 821/N11 SO Rounds LP: OT-3036-B21/N11, Obj. E LP: - OT-3036-C41, Obj. E 239002A309 ..(KA's)
211000A403 ..(KA's)
SENIOR REACTOR OPERATOR Page 62 ANSWER: 025 (1.00) ANSWER: 031 (1.00) REFERENCE: REFERENCE:
SDM: N32/C85 SDM: C51(SRM)
LP: OT-3036-N32/C85, Obj. E LP: OT-3036-C51(SRM), Obj. D 241000K302 ..(K A's) 215004K402 ..(KA's)
ANSWER: 026 (1.00) ANSWER: 032 (1.00) REFERENCE: REFERENCE:
SDM: C34 T.S. 3. LP: OT-3036-C34, Obj. C LP: OT-3036-B33, Obj. J 259001A201 ..(KA's) 202001G005 ..(KA's)
ANSWER: 027 (1.00) ANSWER: 033 (1.00) REFERENCE: REFERENCE:
SDM: C34 SDM: C41 LP: OT-3036-C34, Obj. C LP: OT-3036-041, Obj. D 259002K409 ..(KA's) 204000K108 ..(KA's)
ANSWER: 028 (1.00) ANSWER: 034 (1.00) REFERE' ~~:: REFERENCE:
SDM: M16 T.S. 3.4.10 LP: OT-3036-M15, Obj. E LP: OT-3037-08, Obj. C 261000K401 ..(K A's) 205000G005 ..(K A's)
ANSWER: 029 (1.00) ANSWER: 035 (1.00)
REFERENCE: d.
SDM: E228 & R43 REFERENCE:
LP: OT-3036-E228, Obj. E sol-E12 OT-3036-R43, Obj. D LP: OT-3036-E12, J 264000A209 ..(KA's) 219000G010 ..(KA's)
ANSWER: 030 (1.00) ANSWER: 036 (1.00) REFERENCE: REFERENCE:
ONI E12-1 SDM: E12 LP: OT-3036 E22, Obj. K LP: OT-3036-E12, Obj. E, F 209002A201 ..(KA's) 226001A407 .. (K A's)
SENIOR REACTOR OPERATOR Page 63 ANSWER: 037-(1.00) ANSWER: 043 (1.00) REFERENCE: REFERENCE:
SDM: R10 SOLN 32 LP: OT 3036 R10, Obj. D LP: OT-3036-N32/C85, Obj. K 262001A302 ..(KA's) ONI-N62 245000K502 ..(K A's)
ANSWER: 038 (1.00) ANSWER: 044 (1.00)
REFERENCE: SDM: R14/15 REFERENCE:
LP: OT-3036-R14/15, Obj. B SDM: G41 I
R42 OT 3036-R42, Obj. B LP: OT-3036-G41, Obj. D-263000K201 ..(KA's) 233000K408 ..(KA's)
ANSWER: 039-(1.00) ANSWER: 045 (1.00) REFERENCE: REFERENCE:
SDM: N64 SDM: E32 LP: OT 3036-N64, Obj. E LP: OT-3036-E32, Obj. E 271000K408 ..(KA's) 239003A211 .(KA's)
ANSWER: 040 (1.00) ANSWER: 046 (1.00) REFERENCE: REFERENCE:
SDM: D21 T.S. LP: OT-3036-D21, Obj. B LP: OT-3037-03, Obj. H 272000A101 ..(K A's) 290002GOOE ..(KA's)
ANSWER: 041 (1.00) ANSWER: 047 (1.00) REFERENCE: REFERENCE:
SDM: M25/26 SDM: C71 LP: OT 3036-M25/26, Obj. E LP: OT-3036-C71, Obj. F 290003K101 ..(KA's) 295005K201 .. (K A's)
ANSWER: 042 (1.00) ANSWER: 048 (1.00) REFERENCE: REFERENCE:
SDM: N36 SDM: C34 LP: OT 3036-N36/25/26, Obj. B L OT-3036-C34, Obj. D, G 239001K110 ..(KA's) 295006A102 ..(KA's)
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iSENIOR REACTOR OPERATOR:- -
Page 641 ANSWER: T C49 . (1.00) . ANSWER: 055 (1.00)- ' a. - -
- REFERENCE: REFERENCE:
ONI-C85 1- SDM: : C34- .
LP:: ~ OT-3036-N32/C85, Obj. N - . LP: - OT 3036-C34, Obj. D
.295007A106' .. (KA's) OT-3036-C22, Obj. D
'
i - C22-295025K203 ..(KA's)
. ~ ANSWER: LOSO ' (1.00) .. ANSWER: -056 (1.00)
- REFERENCE: ' .SDM: B33 REFERENCE:
LP: OT-3036-B33, Obj. E PEl813.
295009K102 . .,X A's) LP: OT-3402 02, Obj. F
-
295031A108- ..(KA's)
ANSWER: -- 051 (1.00) , ANSWER: 057 (1.00)
!- - REFERENCE: SDM: C22 ._
REFERENCE:
LP: OT 3036-C22, Obj. D ONI-N62 295025A107 i (KA's) -
. - LP: .OT-3036-N6e., A,.1-295002A201 ..(KA's)
ANSWER: 052 (1.00)
- ANSWER: 058 (1.00)
REFERENCE: .
,
, SDM: C11(RCIS)
~
.
REFERENCE:
'
LP: OT-3039-01, Obj. A SDM: E22A 5101 1 & 2 LP: OT-3036-E22A, Obj. E PAP-0201 C34 OT-3036-C34, Obj. C
,
295014K101 .. (KA's) 295008A101 . .(KA's)
, - ANSWER: 053 (1.00) ANSWER: 059 (1.00)
-- a,
'
REFERENCE: REFERENCE:
' SDM: C11(RCIS)
-
PEl-T23 LP: OT-3036-C11(RCIS), Obj. E, G LP: OT-3402-07, Obj. C 295015A104: .. (KA's) PEI Bases
'
295011K101 ..(KA's)
' ANSWER: 054-(1.00)
- ' "
REFERENCE:
PELT 23
-
LP:~ i OT-3402-09, Obj. C
- . PEIBases 295024A209 .. (KA's)
c
,
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. . . . - ..
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SENIOR REACTOR OPERATOR Page 65
' ANSWER: 060 (1.00) ANSWER: 066 (1.00) REFERENCE: REFERENCE:
SDM: E12 - = PELT 23 LP: . OT-3036-E12, Obj. E LP: OT-3402-06, Obj. C 295013A101- ..(KA's) 295026G012 ..(KA's)
ANSWER: 061 (1.00) ANSWER: 067 (1.00) REFERENCE: REFERENCE: 1 SDM: C61 i
PELT 23 LP: OT-3036-C61, Obj. C LP: OT-3402-14, Obj. C ONI-C61 PEIBases 7295016G010 ..(KA's) 295029G012 ..(KA's)
- ANSWER: 062 (1.00) ANSWER: 068 (1.00) REFERENCE: REFERENCE:
ONI-P43 - PEl-813 LP: OT-3036-P43, Obj. H LP: OT-3402-01, Obj. D l 295018K302- ..(K A's) 295028K203 ..(KA's) !
ANSWER: 063 (1.00) ANSWER: 069 (1.00) ;
' REFERENCE: REFERENCE:
ONI-PS2 ONI-E12-2 LP: OT-3036-P51/52, Obj. G LP: OT-3035-11, Obj. A 295019G010 ..(KA's) 295021A104 ..(KA's) '
' ANSWER: 064 (1.00) ANSWER: 070 (1.00) REFERENCE: REFERENCE:
ONI-C11-1 ONI-E12-2 LP: OT-3036-C11(CRDH), Obj. G LP: OT-3036-G41, Obj. E 295022A201 ..(KA's) 295023A103 ..(KA's)
. ANSWER: 065 (1.00) ANSWER: 071 (1.00)
b- REFERENCE: - REFERENCE:
SDM: B33 PEI Bases page 341 LP: . OT-_3036 '433, Obj. I LP: OT-340217, Obj. 8-ONIC51 PEl-N11 295001A201- ..(KA's) 295032G012 ..(KA's)
L_ . __. -
-. . - . . - . . _ - . . -. .
- SENIOR REACTOR OPERATOR Page 66 ANSWER: 072 (1.00) ANSWER: 077'(1.00) '
FtEFERENCE: . REFERENCE:
PEl N11 :SDM: E21 LP:-340217, Obj. D LP: = OT-3056-E21, Obj. J 295036G012- . .(KA's) 209001K404 ..(KA's)
ANSWER: - 073 : (1.00) - ANSWER: 078 (1.00) REFERENCE:- REFERENCE:
' PAP-1401, Section 6. _
T.S. 3. LP: OT-3039-01, Obj. A LP: OT-3037-02, Obj. E '
294001K102 ..(KA's) 211000G005 ..(KA's)
ANSWER: 074 (1.00) ANSWER: 079 (1.00) REFERENCE: REFERENCE:
PAP-0514, Section 6.4.1 c . SDM:M56 LP: OT-3039-00, Obj. A LP: OT-3036-M56, Obj. C-294001K103 ..(K A's) OT 3402-10, Obj. C T3001K607 ..(KA's)
ANSWER: 075 (1.00)
ANSWER: 080 (1.00) REFERENCE:
REFERENCE: SDM: E12 LP: OT-3036-821(INST), Obj. F PAP-0201 section B21(NS4) OT-3036-E12 , Obj. F LP: OT-3039-OO, Obj. A '223002K607. ..(KA's)
294001A102 ..(KA's)
ANSWER: 081'(1.00)
ANSWER: 076 (1.00) c.- REFERENCE:
REFERENCE: T.S. 3. SDM: C95 LP: OT-3037-01, Obj. F
'
. LP: OT-3036-C95, Obj.' C 217000G005 ..(r. s)
294001A115 ..(KA's)
.
SENIOR REACTOR OPERATOR Page 67 ANSWER: 088 (1.00)
ANSWER: 082 (1.00) REFERENCE:
REFERENCE: PEl-B13 SDM: C11(RCIS) LP: OT-3402-02, Obj. F LP: OT-3036 C11(RCIS), Obj. D 295031K304 ..(KA's)
215003K103 ..(KA's)
ANSWER: 089 (1.00)
ANSWER: 083 (1.00) REFERENCE:
REFERENCE: PEI Bases, pg. 254 SOI-N21, Sect. 2. LP: OT-3402-07, Obj. C LP: OT-3036-N21, Obj. G 295027K303 ..(KA's)
256000A401 .. (K A's)
ANSWER: 090 (1.00)
ANSWER: 084 (1.00) REFERENCE:
REFERENCE: T.S. 3.6. ONI-R10 LP: OT-3037-05, Obj. 8 LP: OT-3036 R10, Obj. J 295030G008 ..(KA's)
295003K106 ..(KA's)
ANSWER: 091 (1.00)
ANSWER: 085 (1.00) REFERENCE:
REFERENCE: MCD Text ONI-C71-1 LP: OT 3401-06, Obj. F LP: OT-3036-C71, Obj. L 295031K101 ..(K A's)
295006G010 .. (K A's)
ANSWER: 092 (1.00)
ANSWER: 086 (1.00) REFERENCE:
REFERENCE: PEl-B13 ONI-J11-2 LP: OT-3402-11, Obj. O LP: OT-3036-J11, Obj. J 295037G012 .. (K A's)
295023G010 ..(KA's)
ANSWER: 093 (1.00)
ANSWER: 087 (1.00) REFERENCE:
REFERENCE: PEl-B13
' EPI-A2 LP: OT-3402-11, Obj. C LP: EPL 0823 OT-3402-03, Obj. F 295017G002 ..(KA's) 295037A201 ..(KA's)
.- - . . . -. - . . . .
- i
- SENIOR REACTOR OPERATOR Page' 68 ANSWER: 0941(1.00)- ANSWER:' - 099 : (t.00) _;
- REFERENCE
- REFERENCE:
PEl M51/56 SDM:B33 LP: . OT-340210, Obj. C
_
LP: OT-3036-B33, Ob _295038G012_ .' (KA's)- GOI833 295022A103= ..(KA's) ,
ANSWER: 095'(1.00) ANSWER: 100 (1.00)
REFERENCE: - SOI-E12 REFERENCE:
LP: OT-3036-E12, Obj. E PEI Bases page 344 295024A117 ..(KA's) LP: OT 3402-17, Obj. D 295033G012 ..(KA's)
<
'
ANSWER:- 096 (1.00) REFERENCE:
PEl-B13 (ATWS)
'
LP: OT-3402-11, Obj. D 295015A201 ..(KA's)
ANSWER:- 097 (1.00)
-c.
'
- REFERENCE:
SDM: R42 LP: OT-3036-R42,- Obj. B ONI-R42-1 295004K203- ..(KA's)
ANSWER: 098 (1.00) REFERENCE:
SDM: E12 LP: OT-3036-E12, Obj. F 2"5020K209 ..(KA's)
(* * * * * * * * ' END OF EXAMINATION * * " * " * ")
I
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- SENIOR REACTOR OPERATOR Pago 69 ANSWER KEY MULTIPLE CHOICE O23 c 001 c 024 c 002 b 025 b 003 d 026 a 004 d 027 b 005 b 028 b 000 a 029 a .
007 c 030 a 008 d 031 b 009 c 032 c 010 c 033 b 011 d 034 c 012 b 035 d 013 'd 036 d 014 a 037 b 015 a 038 a 016 a 039 c 017 - d 040 b 018 a 041 b 019 c 042 c 020 c 043 c 021 d 044 d 022 b 045 c
_ _ _ _ _ _ .
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SENIOR REACTOR OPERATOR ; _ Page 70 ,
- - A N S W E R -E K E Y
'
' MULTIPLE CHOICE c' '
i 046 ' '~d 069 - d -
- 047 ; b -
. _ 070 : b .
048 c2 071 a-
'
049: b ' 072 d 050: b1 1073 [051 b ._ 074 d:
052 a 075 = d
.053 c 076'c 0 5 4 -- c:/ 077 a-055 a 078- d 1 056. _d 079 a .
- 057 ~ b - '080 d 058:.b 081 c 059 a-- 082 d 0 6 0 -- a 083 b
-
061 d .- 084 d 002 - d 085 :n- '086 a
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LO64I d: 087 c-
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-
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.. Page 7,1 A N S W E R:: K E Y: -
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Li (' * ' * ' * * * END OF EXAMINATION '" * " * ')
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