IR 05000298/1993020
| ML20045E754 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/25/1993 |
| From: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20045E747 | List: |
| References | |
| 50-298-93-20, NUDOCS 9307060014 | |
| Preceding documents: |
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| Download: ML20045E754 (75) | |
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-APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION
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REGION IV
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Inspection Report:
50-298/93-20
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Operating License: DRP-46
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Licensee: Nebraska Public Power District P.O. Box 499 Columbus, Nebraska 68602-0499 Facility Name: Cooper Nuclear Station (CNS)
Inspection At: CNS, Brownville, Nebraska Inspection Conducted: May 24-28, 1993
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Inspectors:
T. O. McKernon, Reactor Inspector, Operations Section Division'of Reactor Safety John L. Pellet, Chief, Operations Section Division of Reactor Safety Accompanying personnel: Mark P. Morgan, Contractor
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Arthur T. H ll, Dejidty D'irector Date Division'o Reactor Safety Inspection Summary Areas Inspected: Announced inspection of the qualification of applicants for operator licenses at the Cooper Nuclear Station, which included eligibility determination and administration of comprehensive written and operating examinations. The examination team also observed the performance of onshift operators and plant conditions incident to the conduct of the applicant
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evaluations.- The examiners used the guidance provided in NUREG-1021,
" Operator Licensing Examiner Standards," Revision'6, Sections 201, 202, 203, 301, 302, 303, 401, 402, and 403 issued June 1, 1990.
Further, the-inspection included the review of inspection findings from previous inspections.
Results:
All six of the applicants for reactor operator licenses satisfied the
requirements of_10 CFR Part 55.33(a)(2) and have been issued the appropriate license (Sections 1.2, 1.3).
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9307060014 930629 PDR ADOCK 05000298 G
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The reference material provided by the training. department for
examination development wa's generally satisfactory. However, the facility developed test items (job performance measures and related questions) were not in all cases current (Section 1.1).
Applicants performed well on the written examination, with scores
ranging from a low of 84 percent to a high of 94.9 percent with an
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average of 88 percent (Section 1.2).
No generic weaknesses were observed during the conduct of the operating-
examinations (Section 1.3).
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Simulator fidelity appeared acceptable with no notable discrepancies
observed (Section 1.4).
During the dynamic simulator scenarios, crew communications appeared
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strong. However, some instances of operator inattention to detail were observed (e.g., allowing the reactor core isolation cooling turbine to overspeed without manually tripping it) (Section 1.3.1).
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During a review of procedures used to support the job performance
measures, it was observed that an emergency plan implementation.
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procedure contained erroneous computation instructions for determining the estimated drywell noble gas curie content. This finding was considered a violation for failure to maintain appropriate plant procedures (Section 1.3.2).
Summary of Inspection Findings:
(0 pen) Violation 298/9320-01:
Failure to maintain appropriate plant
procedures (Section 1.3.2).
(Closed) Violation 298/9211-01:
Failu're to maintain procedures for
responding to emergency conditions (Section 2.2).
(Closed) Inspection Followup Item 298/9006-02: -NUREG 0737, II.B.2,
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" Design Review of Plant Shutdown and Environmental Qualification of Equipment for Spaces / Systems Which May be used in Post Accident Operations" (Section 2.1).
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Attachments:
Attachment 1 - Persons Contacted and Exit Meeting
Attachment 2 - Written Examination Keys
Attachment 3 - Facility Post-Examination Review Comments
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-3-DETAILS 1 LICENSED OPERATOR APPLICANT QUALIFICATION EVALUATION (NUREG-1021)
During the inspection, the examiners evaluated the qualifications of-six license applicants, all reactor operators (RO). The inspection assessed the eligibility and administrative and technical competencie of the applicants to -
be issued licenses to operate the reactivity controls of a commercial nuclear power facility in accordance with 10 CFR Part 55 and NUREG-1021, " Operator License Examiner Standards," Revision 6, Sections 200 (series), 300 (series),
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and 400 (series).
Further, the inspection included evaluation of facility i
materials, procedures, and simulation capability used to support development and administration of the examinations. These areas were evaluated using the guidance provided in the areas of NUREG-1021 cited above.
In addition, the inspection included the review of previously identified inspection findings.
Performance results for individual applicants are not included in this report because inspection reports are placed in the NRC Public Document Room as a matter of course.
Individual performance results are not subject to public disclosure.
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1.1 Facility Materials Submitted for Examination Development The chief examiner reviewed the licensee's materials provided for development of the examination, which included station administrative and operating procedures, question banks, simulator scenarios, lessons plans, and job performance measures (JPMs). The procedures and question banks, as well as the lesson plans, appeared current and adequate to support the examination
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development.
The JPMs, while relevant, were not in all cases current with the la'.est revision of the procedure.
For example, JPM SKLO34-20-04 used Revisun 7? of the procedure; yet Revision 72 graphs (not current) were used in'the simulator. None of the problems identified during validation of the_JPMs required major rewrites of the JPMs.
The facility maintained a good bank of questions, JPMs, and scenarios.for
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evaluating initial license applicants. The chief examiner was able to develop scenarios of varied scope and complexity and to select JPMs which challenged the applicants both inside the control _ room and in the plant.
The licensee's question bank, JPMs, training material, and scenarios proved helpful in the examination development.
1.2 Written Examinations
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The chief examiner developed a comprehensive written R0 examination in accordance with the guidelines of NUREG-1021, Revision 6, Section 401.
The examination consisted of 98 multiple choice questions.
During the week of May 10, 1993, members of the facility training department, under the
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provisions of NUREG-1021, which require execution of a non-disclosure security agreement, reviewed the examinations. The NRC considers the preadministration
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review of the examination by the facility as part of the examination development process.
Therefore, the specific comments resulting'from that-review are not reported or'otherwise retained. The chief examiner-incorporated the facility review comments and administered-the. examinations to the license applicants on May 25, 1993.
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The chief examiner provided the facility training staff with a copy.of the "as administered" written examination and key along with the pre-administration'
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review comments on May 25, 1993, immediately following the completion of the examination by the applicants. The facility took that opportunity to further review and comment on the written examination'. The facility's post-examination review are contained in Attachment 3.
After careful consideration of the post-examination review comments the chief examiner took the following actions:
The recommendation to delete Question No. 012 was accepted on the basis
of the justification provided by the facility.
The recommendation to allow credit for either Answer (c) or (d) to
Question No. 034 was accepted on the basis of the justification provided
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by the facility.
The recommendation to delete Question No. 055 was accepted on the basis
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of the justification provided by the facility.
The chief examiner reviewed applicant performance on individual questions 'and
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observed that the following questions were missed by more than 50 percent of the applicants.. The questions are referenced here only by. question number..
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Refer to -Attachment 2 for the complete question and answer. Questions:
56,
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57, 64, 72, and 87.
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The chief examiner concluded that no specific area of significant knowledge weakness was apparent in the responses to the above. questions. -Therefore, the information is provided to the facility training staff for consideration as feedback into future training needs.
Overall, the applicants performed well on the written examination. Scores ranged from a low of 84 percent to a high of 94 percent with an average of.
88 percent.
q 1.3 Operatina Examinations The examiners developed comprehensive operating examinations in accordance with the guidelines of NUREG-1021, Revision 6, Section 301. The operating examinations consisted of two parts, a dynamic simulator scenario portion and a control room / plant walkthrough portion. The examiners previewed and validated the various portions of the operating examinations onsite with the
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-5-assistance of facility training personnel under security agreement during the week of May 24, 1993. The examination team administered the operating examinations during the period of May 25-27, 1993.
1.3.1 Dynamic Simulator Scenarios The examination team evaluated three crews (each consisting of one not examined, qualified surrogate filling the control room supervisor position and two R0 applicants) on two scenarios using the Cooper Nuclear Station plant-specific simulation facility. The. examiners compared applicants' performance during the scenarios with expected performance in accordance with the requirements of NUREG-1021, Revision 6, Section 303, to evaluate applicants'
competencies on this portion of the operating examinations.
The examination team noted that communications among crew members was generally good.
The applicants responded well to alarms and annunciations.
However, in a few instances some of the applicants missed certain actions or overlooked component parameters.
Examples of these included not isolating recirculation loop after securing the recirculation pump and not manually tripping the reactor core isolation cooling turbine after the turbine exhibited overspeed conditions.
The examination team observed no generic weaknesses during this portion of the operating examination. All applicants, without exception, passed this portion of the operating examination. The applicants generally performed well during~
the scenarios.
1.3.2 Walkthrough Examinations The examination team evaluated each of the six R0 applicants using one of two.
sets of ten JPMs relating to tasks within the scope of potential duties of a licensed R0 (which included non-licensed operator tasks outside the control
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room). The applicants performed most of the tasks in the simulation facility in the dynamic mode. They simulated (through discussions) the remainder of tasks either in the plant control room or at local operating stations throughout the plant.
Immediately following the performance of each task, the examiners asked pre-scripted questions relating to the system involved in the
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task.
The questions solicited "short answer" responses and permitted the applicants to use controlled references unless instructed otherwise.
The examiners combined the applicants' task performance and question responses in accordance with the guidelines of NUREG-1021, Revision 6, Section 303, to evaluate performance on this portion of the operating examination.
Overall, the applicants performed well. All applicants passed this portion of the operating examination with satisfactory performance on all systems and tasks with the exception of two applicants failing the same'JPM task on the simulator.
During the validation of JPM SKLO34-20-54, " Release Rate Determination," the examiners determined that Attachment 4 of EPIP 5.7.16, " Release Rate l
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-6-Determination," Revision 16, was in error. The procedural instruction (Step 8.4.5) required the user to calculate the estimated noble ' gas curie -
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content (Ci) by dividing Column (2) (R/hr) by Columns (3) (R/hr) and (4)-(C1)
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rather than multiplying by Column (4).
The resultant, Column (5), should have been in units of (Ci). However, the resultant obtained by the erroneous
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instruction would have the inverse units, and be in error by roughly 10" in the nonconservative direction. A further review of the prior procedure revision indicated that same error was not present. 'A similar error existed with Procedure Step 8.3.5 and Attachment 3.
The inspectors concluded that the
errors were incorporated during the change to Revision 16. As such, the-
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inspectors considered that the errors represented a failure to maintain appropriate plant procedures for responding to emergency conditions as-
required by Technical Specification 6.3.2 (Violation 298/9320-01). This is
similar to previous violation 298/9211-01 for failure to maintain procedures
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for responding to emergency conditions, which is discussed in section.2.2
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below.
The inspectors concluded that this was not a repetitive violation because emergency operating procedures and emergency plan implementing procedures are handled by separate groups and processes by the licensee.
1.4 Simulator Fidelity During the preparation and conduct of the operating examinations, the
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examination team did not observe any discrepancies in simulator fidelity.
2 FOLLOWUP The following items were reviewed to ascertain whether the licensee had taken sufficient action.
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2.1 Followup (92701)
IClosed)
Inspection Followup Item 298/9006-02:
NUREG 0737. Item II.B.2.
" Design Review of Plant Shutdown and Environmental Qualification of Eauipment
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for Spaces / Systems Which May Be Used in Post Accident Operations" This item involved concerns related to post-accident reactor building reentry to support emergency operating procedures (E0Ps) ' implementation.
Specifically, the concern addressed whether radiation levels in the reactor
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were adequately evaluated during reentry into the reactor building in order to perform E0Ps and whether the existing emergency procedures, such as EPIP 5.8.11, " Reactor Pressure Vessel (RPV) Venting During Primary Containment (PC) Flooding," met the intent of post-THI action item NUREG 0737, II.B.2.
In subsequent review and discussions, the licensee committed to revise the E0Ps including the emergency support procedures (ESPs), which required local
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actions to be performed using an enhanced verification and validation (V&V)
methodology. The V&V included radiological considerations, appropriate cautions, and prioritization of alternate success paths.
Further, the procedures would be revised to ensure a health physicist accompany the
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In addition, the licensee committed to pursue with the BWR Owners Group (BWROG) whether a conflict between NUREG 0737, Item II.B.2, and the BWR0G's position existed with respect to E0P requirements.
During this inspection, the inspector reviewed licensee documentation, which indicated agreement between the BWROG emergency procedures committee and the licensee's position. Draft ESPs and E0Ps were reviewed to verify that the licensee was incorporating the cautions and prioritized local action requirements based upon exposure dangers and use of the enhanced V&V to identify local action steps which might be prohibited due to high radiation levels.
Further, the procedures required a health physicistLto accompany the response team to assist in determining radiation levels.
In addition, the E0P
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flowcharts were being revised to incorporate the appropriate cautions and
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prioritize success paths (local actions). The licensee had also incorporated training on the procedures into their licensed operator training program. At the time of this inspection, the procedures were to be reviewed and approved by the onsite committee prior to the end of the 1993 spring outage.
The inspector concluded that the licensee was in the process of fulfilling their commitment related to revising the E0Ps and ESPs and appeared-to have satisfied those requirements of NUREG 0737, Item II.B.2.
It was further noted that both the BWROG and NRC staff has reviewed this issue and rm. curred as acceptable those items detailed in licensee letter to the M.iC dated
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November 2, 1992.
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2.2 Followuo items of Noncomo11agre (92702)
(Closed) Violation 298/9211-01:
Failure to Maintain Procedures for Respondina to Emeraency Conditions.
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This violation involved the failure to maintain appropriate plant procedures in that ESP-5.8.11, " Reactor Pressure Vessel (RPV) Venting during Primary
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Containment (PC) Flooding," contained procedural steps which would not have-worked under certain accident conditions and the procedure was not revised to correct the discrepancies.
During this inspection, the inspector verified that the appropriate procedures had been revised.
It was further noted that Procedure CNS 0.22, " Emergency Operating Procedure Maintenance Program," had been revised to include provisions for prioritizing E0P comment forms. The procedure provided expedient document change measures for significant items which affected the
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E0P flowcharts, the emergency support procedures, or the E0P development documents.
It was further noted that the licensee was in the process of revising the E0P flowcharts and the ESPs with intended issuance prior to the end of the 1993 spring outage.
Based upon the above review, the inspector concluded that the licensee had taken the appropriate corrective actions.
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3 Conclusions The chief examiner concluded that the performance of the six applicants _ for-operator licenses satisfied the requirements of 10 CFR Part 55.33(a)(2) and recommended that licenses be issued. The appropriate licenses have been-issued.
In general, the examination team concluded that:
i Individual applicants performed well.
- Individual and crew communications were good.
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The applicants demonstrated no generic weaknesses.
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With few noted exceptions, the developmental material (question bank,.
JPMs, and scenario banks) were useful in evaluating the applicants and saved time in developing the examination.
No discrepancies with simulator fidelity were observed.
- s One violation for a failure to maintain appropriate plant procedures was
identified.
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ATTACHMENT l'
PERSONS CONTACTED 1.1 Licensee Personnel
- B. Ackerman, Training Instructor
- B. Black, Operations Supervisor
- J. Boyd, Lead License Instructor
- B. Brungardt, Operations Manager
- R. Creason, Operation Training Supervisor
- M. Dean, Nuclear Licensing & Safety Supervisor
- J. Dutton, Training Manager
- J. Florence, Simulator Supervisor
- R. Gardner, Plant Manager
- E. Jackson, Training Instructor
- S. Jobe, Assistant Nuclear Training Manager
- W. McKinzey, Training Instructor
- J. Meacham, Site Manager 1.2 NRC Personnel
- T. McKernon, Reactor Inspector / Examiner
- J. Pellet, Chief, Operations Section
- W. Walker, Resident Inspector 1.3 Other M. Morgan, Consultant In addition to the personnel listed above, the examiners contacted other personnel during this inspection period.
- Denotes personnel that attended the exit meeting.
EXIT MEETING An exit meeting was conducted on May 28, 1993. During this meeting, the examiners reviewed the scope and generic findings of the inspection.
The-examiners did not disclose preliminary results of individual evaluations since they are subject to change during the final review and approval process.
The licensee did not identify as proprietary any information provided to, or reviewed by, the examiners.
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ATTACHMENT
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' OOOPER NUCLEAR STAtcN 8 O. BOX 98,. BROWNVfLLE. NEDRASKA 68321
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NTM93137 May 28. 1993
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U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington. TX 76011 Attention:
Mr. John Pellet Subject:
Comments - SRC Examinati on Administered May 24 through May 28. 1093.
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Dear Mr. Pellet:
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ittached please find our comments on questions associated with - the subject examination given on May 25. 1993.
Te believe that our comments should.be-considered in the grading of the written examinations.
Please contact us if you would like to pursue our comments further or if additional clarification is desired.
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,wtc.M/ * MW iJihn M. Meacham ate Manager JWD/dj m Attachments
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G.R. Horn w/ attachments R. L. Gardner w/ attachments
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R. 3rungardt w/o attachments J. 'i. Dutton w/ attachments TM File w/ attachments i
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i CHALLENGE QUESTIONS FOR 1993 CNS HOT LICENSE EXAM QUESTION:
012 (1.00)
RPS Motor Generator (MG) Set lA has tripped. Upon. completion of-repairs and testing, what are the correct recovery actions needed to return the components to service?
a.
No immediate actions are required since alternate power is supplied from HCC-T to RPS power panel 1A due to Auto Switching upon loss of MG set lA.
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If interim power is supplied from MCC-T, then transfer back to the
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MG Set lA power source must be accomplished with the RPS PPIA deenergized locally.
c.
Transferring power back to the MG set lA will result-in a temporary 1/2 SCRAM but will not affect the Primary Containment Isolation System (PCIS).
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d.
Transferring power back to the MG set 1A from the critical distribution panel CDPlB can_be accomplished through reset of the 1/2 SCRAM, a power transfer switch in the control room, and resetting the affected electrical protection assemblies from the control room.
ANSWER:
012 (1.00)
b.
i REFERENCE:
CNS System. Manual COR002-01 IV.A l
{3.4/3.9]
212000A201
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Comments
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Answer a is incorrect; there is no Auto Switching.of PRS PPlA upon the.
loss of the MG set.
Answer b (key answer); interim power is not supplied from MCC-T. It is supplied from MCC-TX via CDP 1B (SOP 2.2.22 Section 8 4.2' Attached).
Answer c is incorrect; PCIS is affected on a transfer of RPS Power.
Answer d in incorrect; the EPAs cannot be reset from the control room.
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'Je recommend this question be deleted from the exam since there is no correct
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answer.
The reference given in the exam key seems to be incorrect: it should probably be COR002-21 IV.A.
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QUESTION:
034 (1.00)
y Which most accurately describes RHR system Loop
"A" lineup for containment spray?
a.
RHR-MO 39A Open, RHR pump "A" running, RHR-MO-38A (Torus Spray.
Throttle valve) closed, and RHR-MO-31A Drywell inboard spray valve
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open, and throttle with RHR MO-26A Outboard Valve, b.
RHR pump "A" running, RHR.-MO-39A Open,.RHR-MO-38A Open, RHR-MO-26A-l Drywell Spray Putboard valve Open and throttle the-inboard RHR-MO-31A valve.
c.
RHR-MO-39A (Suppression Pool Cooling) valve Open, RHR pump "C"
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running, RHR-MO-38A Open, RRR-MO-31A, Inboard Drywell Spray Valve Open, and Throttle with RHR MO-26A Outboard valve.
d.
RHR SW system in service, RHR M0 39A closed, RHR pump
"C" running, RHR-MO-38A closed, RHR-MO-31A Open, RHR-MO-66A throttled elosed,-
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and RHR-MO 26A throttled Open.
ANSWER:
034 (l~00)
c.
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REFERENCE:
System Procedure 2.2.69.3, pg 2
[3.1/3.4]
226001A105
.(KA's)
Comments Answer c (key answer) is correct; this lineup will spray;the Torus-and the Drywell but it does not reference any cooling lineup.
Answer d is correct; this lineup will spray the Drywell with cooling lined up.
By CNS EOP convention, Reactor Operators are not given the order to.
lineup for " containment spray", they are told to lineup for either Torus or Drywell sprays, or both.
These specific directions insure E0Ps are not violated.
Per E0P 3A PRIMARY CONTAINMENT CONTROL step DW/T-8, the Drywell will be sprayed for elevated temperature cont *ol making answer d the most correct.
When spraying for containment pres ure control per
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E0P 3A steps PC/P-5 thru PC/P 10 answer c would be the most correct.
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The operator is required to determine which.is more correct, without sufficient containment parameters.
We recommend accepting either answer c or d, because there was no clear direction to the operator on which areas of containment to be sprayed.
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QUESTION:
055 (1.00)
,i With the reactor at power and the "B" RWCU Pump _ out-of-service for maintenance, the reactor operator observes annunciator 9-4-2/B-4, RWCU Pump
"A" Low Flow, al a r,as. After notifying the CRS, which of the following L
descrfbes the reactor operator's required actions?
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Closes RWCU-MO-15, MO-18, opens RWCU M0-74 (Demineralized Water-Suction Bypass) and closes CRD-189 (RWCU Minipurge isolation),
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rejects water to the Radwaste/ Main Condenser.
b.
Closes RWCU-MO-15, MO-18, opens RUCU-MO-74 (Damineralized Water-suction Bypass), and closes CRD-189 (RWCU minipurge Isolation).
e.
Opens RWCU-MO-74, opens RWCU-MO-15 -18; rejects water to g
L Radwaste/ Main Condenser; and calls chemistry to' monitor Reactor'
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Uator Chemistry parameters, d.
Verifies a Group 3-isolation, opens RWCU-MO-70 (Demin. Suct.
Bypass), opens CRD-189; and commences reactor shutdown.
ANSWER:
055 (1.00)
C.
REFERENCE:
COR001-20, pg 15
[3.3/4.3]
294001Alli
.(KA's)
Comments Per AP 2.3. 2.25 annunciator 9-4 2/B-4 RWCU PUMP A LOW FLOW ( ATTACHED),
the automatic actions are:
trip of "A" RUCU pump and the closing of'
RWCU-AO-15A EFFLUENT VALVE. -(No te :
RUCU-AO-15A and RWCU-MO-15A are-two different valves).
Operator actions are: verify the' pump tripped, and determine the cause of system low flow.
The probable causes listed for this annunciator are:
pump or filter demineralizer valve malfunction.
Given the conditions stated, no actions will be required for RWCU-MO-15 or 18, the valves are open and would remain open. With 15 end.18 open,
'RWCU-MO-74 does not need to Le opened. -Rejection of reactor water is not required.
Therefore, answer e is also incorrect, We request that this question be deleted from the exam due to the above reasons.
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lNRChOffic'ial~Use'Only
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ATTACHMENT 2
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Nuclear Regulatory Commission
Operator Licensing-
-l Examination
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l This document is removed from:
Official Us.e Only category:on date of examination.
- .-NRC Official:Use Only.
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NUCLEAR REGULATORY-COMMISSION
SITE SPECIFIC EXAMINATION.
REACTOR OPERATOR LICENSE REGION
f CANDIDATE'S NAME:
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FACILITY:
Cooner Station
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REACTOR TY1-E:
BWR-GE4 DATE ADMINISTERED:
.93/05/24 INSTRUCTIONS TO CANDIDATE:
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.Use-the answer sheets provided to document your-answers.
Staple this~ cover
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sheet dn top of the' answer. sheets.
Points for each question areLindicated'in
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parentheses after the question.
The passing grade requires:a final.' grade of.
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-at11 east 80%.
Examination papers will be-picked up four-(4) hours after the
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CANDIDATE'S
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TEST VALUE SCORE i
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- TOTALS FINAL GRADE
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All work done on this examination is my own.
I have neither given nor
received aid.
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Candidate's Signature
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-. REACTOR: OPERATOR ~
.Page-2 A N S W'E R
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Multiple Choice (circle or X your choice)
If you change your answer, write your selection in the blank.
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MULTIPLE CHOICE 023 a
b
.c d:
.I 001 a
b.
c d
024 a
b c
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002 a
b c
d 025 a
b c'
d 003 a
b c
d-026 a
b c
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- 004.
a b
c d
027 a
b c
d
'005 a
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. REACTOR-OPERATOR
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A NlS~W'E RL S~H E:E-T i
Multiple Choice (Circle or-X your. choice)
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If you change your answer, write your selection in-the blank.
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. REACTOR OPERATOR Page-4 A N'S W E R~
S H'E E T
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. Multiple Choice-(Circle or-X your choice)
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If.you change'your answer,. write your selection in the blank.
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END OF EXAMINATION **********)
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NRC RULES AND GUIDELINES FOR LICENSE ~ EXAMINATIONS During the administration of this~ examination the followiny rules apply:
1.. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on
';
the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination..This must be done after you complete the examination.
3.
Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4.
Use black ink or dark pencil ONLY to facilitate legible reproductions.
5.
Print your name in the blank provided in the upper right-hand corner-of the examination cover sheet and each answer sheet.
'
6. Mark your answers on the answer sheet provided.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
7.
Before you turn in your examination, consecutively number each answer sheet,-
including any additional pages inserted when writing your answers on the
!
examination question page.
'
.8.
Use abbreviations-only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition.
',
error resulting in an incorrect answer.
Write it out.
9. The point value for each question is indicated in parentheses after the question.
. :U). Show all calculations, methods, or assumptions used to obtain an answer'to any short answer questions.
i
,
-11.
Partial credit may be given except on multiple choice questions.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
-
<
12. Proportional grading will be applied.
Any additional wrong information that is provided may count against you.
For example; if a question is-worth one point and asks-for four. responses, each of which-is worth 0.25 points, and you give five responses, each of your responses will be worth
.20 points.
If one of your five responses is incorrect, 0'.20 will be w> ducted and your. total credit for that quastion will be 0.A0 instead of 1.00 even though you got the four correct answers.
13. If the intent of a question is unclear, ask questions of the examiner only.
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"4'. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets.
In addition, turn in all scrap paper.
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'
15. Ensure all information you wish to have evaluated as part of your answer is.
on your answer sheet.
Scrap paper will be disposed of immediately following
'
the examination.
16. To pass the examination, you must achieve a grade of 80% or greater.
17. There is a time limit of four (4) hours for completion of the examination.
,,
18.'When you are done and have turned in your examination, leave the examination
.
area (EXAMINER WILL DEFINE THE AREA).
If you areffound-in'this area while, _
!
the examination is still in progress, _ your license may be denied or revoked.
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. REACTOR' OPERATOR Page
i QUE.ON: 001 (1.00)
.Which one of the following is the proper method of verifying that!an INFORMATION ONLY copy of'a procedure to fill and vent HPCI is the correct
. procedure revision,. including all changes, to-use it for the task?
,
a.
Refer to a hard copy.of the CNS Procedure Index~.
i b.
Insure the procedure is.the same'as the one.in the control room.
d All copies are maintained current so verification isLnot required.
c.
d.
A controlled copy must be used for this task.
QUESTION: 002 (1.00)
.,
Which one of the individuals below is NOT permitted to operate reactor-l
'
controls under the instruction or supervision of a licensed operator?
Auxiliary operator enrolled in current license training course to a.
!
obtain an operator' license.
b.
A licensed reactor operator who recently failed a NRC administered i
Senior Reactor Operator Upgrade examination.
,
c.
A. licensed reactor. operator whose license has become inactive per the requirements of 10.CFR 55.
d.
Individual enrolled in a current license training course to obtain
=
an-instructor certification.
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. QUESTION: 003 (1.00)
A data item on an SPDS display has. changed color to magenta.
This signifiess a.
An EOP entry condition setpoint is being approached.
,
'
b.
An.EOP entry condition setpoint has been. exceeded.
c.
The data is an operator substituted value.
,-
d.
The data is suspect or bad.
QUESTION: 004 (1.00)
-
,
'What is the function of the SCRAM Inlet valves during a SCRAM?
'
a.
Position to pressurize the CRDH overpistons' areas.
'
b.
Causes the HCU Direct Control Valves'to operate.
c.
Position to pressurize the CRDH underpistons' areas.
d.
Position to vent the SCRAM air header.
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.lREACTOR OPERATOR Page
'9 P
P QUESTION: 005 (1.00)
l i
What will happen if the inservice CRD pump's suction strainer becomes
,
plugged?
'
,
a.
The CRD pump will continue to operate on mini-flow' recirculation to the CST at 20 gpm.
'
b.
The CRD pump will trip on undervoltage and annunciate alarm'on-
~5 panel 9-5 in the control room, c.
The running CRD pump will trip on low suction pressure, annunciate:
,
in the control room, and cause an auto start of the backup CRD i
pump.
- d.
The running CRD' pump will trip on low suction pressure, annunciate
'
in the control room, but will require'a manual start of the backup.
CRD pump.
QUESTION: 006 (1.00)
The south scram discharge volume level in 50 inches,.what. automatic actions have occurred?
a.
SDV not drained alarm (10 inches) and reactor scram (46Tinches)
,
b.
SDV not drained alarm (11-1/2 inches) and reactor scram-(46 inches)
c.
Rod Block (11-1/2 inches) and reactor scram (46 inches)
'!
d.
SDV not drained alarrn (11-1/2 inches) and rod block ~ (46 inches)
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REACTOR OPERATOR-PageT10
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,
QUESTION: 007 (1.00)
6 i f
Which of the following conditions will'NOT'cause a scoop' tube lockout?
,
a.
High fluid drive oil temperature b.
Low voltage on the MG_ set motor supply bus c.
High fluid drive oil pressure
,
d.
Actuation of scoop tube positioner limit switch J
QUESTION: 008 (1.00)
What actions must be taken to reset a scoop tube lockout condition caused'by a loss of power to the MG set drive motor after acknowledging the~ alarms on.~
,
. panel 9-4?
a.
Clear. condition on upstream power supply, locally'
_
t manually match the speed demand with the MG speed at the scoop tube positioner motor, and: depress the scoop tube.eset push button on panel 9-4.
l b.
Reestablish the speed control signal to the scoop tube' positioner,--'
take the M/A transfer _ speed controller 1to manual in.the control.
,
room and match the speed demand 'to _- actual MG ' speed,. and depressthe ;
'
scoop tube reset push button'on panel 9-4.
r c.
Lower MG set lube oil. level such.that operating pressure is consistently < 28 psig, match the speed demand with the actual.MGJ
!
i speed, and depress the scoop tube reset push button on panel 9-4.
d.
Reestablish the speed control signal toJthe scoop tube. positioner,
and depress the scoop tube reset: push button.on Panel-9-4.
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. REACTOR OPERATOR Page 11
.
QUESTION: 009 (1. 00 )
Select the statement that describes the purpose of the Reactor Equipment Cooling system, Supplies cooling water to critical and non-critical components in.
a.
the Reactor, Radwaste, Control, and Turbine buildings.
.
!
b.
Provides cooling water to critical and non-critical, contaminated or potentially contaminated components in the Reactor, Radwaste, Augmented Radwaste, and control buildings.
,
Provides cooling water for the removal of excessive heat from the c.
drywell atmosphere, d.
Acts as'a backup heat sink for those components normally cooled by
plant service water.
QUESTION: 010 (1.00)
The SGT system has received a Drywell high pressure initiation signal.
What.
is the SGT system automatic response from a normal standby lineup?
Only the SGT train in-AUTO starts and draws air from the reactor
'
a.
,
building exhaust plenum and the primary containment exhaust'line.
b.
Both SGT trains auto start and draw air from the reactor building i
exhaust plenum and the primary containment exhaust line.
!
!
c.
Both SGT trains auto start and draw air from'the primary containment exhaust line only.
d.
Both SGT trains auto start and draw' air from the reactor building exhaust plenum only.
,
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L
- QUESTION
- 011 (1.00)
,
While operating at 100% power, the Core Spray System Loop "B".line break detection indication is reading approximately -3.5 psid.
With all other plant parameters normal, what is the status of the Core Spray-system?
a.
A core spray line
"B" break has occurred between the inboard
injection valve and the testable check valve.
b.
The standby liquid " outer" pipe has a leak, core spray flow will go
, ;
directly to the area just below the core. plate.
c.
A core spray line
"B" break may have occurred.inside the vessel but-outside the shroud. A leak detection alarm should have occurred.
d.
A core spray line
"B" break may have occurred inside the shroud but'
core' spray
"A" loop will provide full' spray coverage.
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. REACTOR _ OPERATOR Page 13
QUESTION : 41-&
(1.00)
RPS Motor Ge ator (MG) Set 1A has tripped.
Upon completion of repairs and
!
testing, what a the correct recovery actions needed to return the component
',
to service?
a e required'since alternate power is supplied a.
No immediate a i
'
from MCC-T to RP-oweugpane 1A due to Auto Switching upon loss of
MG set 1A.
'N
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'
s, t7 interim power is suppl'h4 ',ac o
CC-T,
.en transfer back to the b.
If
,
1A power source must Bts
d with the RPS PP1A de-MG Set
. energized locally.
s
c.
Transferring power back to the MG set will result in a temporary-1/2 SCRAM but will not affect the Prima ontainment Isolation System (PCIS).
d.
Transferring power back to the MG set 1A from th ritical'
distribution panel CDP 1B can be accomplished throug eset of_the
_
1/2 SCRAM, a power transfer switch in the control room, and resetting the affected electrical protection assemblies -
m the control room.
l
,
QUESTION: 013 (1.00)
.
A purpose of the APRM Subsystem is to:
a.
Initiate SCRAM trips when reactor power is < 15% and the mode switch is not in RUN.
b.
Provide a reference power signal to the rod worth minimizer.
c.
Initiate rod insert blocks when reactor power < 2.5%.and mode switch in RUN.
d.
Prevent fuel damage during local neutron flux level excursions.
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.' REACTOR OPERATOR Page 14
,
,
QUESTION: 014 ( l'. 0 0 )
I During refueling operations, which of the below. conditions must. exist in.
order for the reactor protection system to receive a-scram signal from the-
,
Source Range Monitors?
a.
At least two IRM channels above the upscale trip _setpoint.
b.
One SRM channel > 1 + E5 cps.
E5 cps and the RPS shorting links ' switches'
c.
One SRM channel > 5 +
-
"open".
'
d.
Refueling bridge near or over the core,.and one SRM channel > 100 cps.
-
QUESTION: 015 (1.00)
During Startup operations the RO notices constant neutron flux indications at Panel.9-5 while attempting to move the SRMs.
After notifying the SS, which of the following describes the appropriate subsequent RO actions.
a.
Bypass the detector channel and continue power' increase.
b.
Monitor reactor power on the other channels, check power supplies and detector drive fuses, and attempt to drive the detector in both-
- t directions.
-
'
c.
Bypass the channel, issue a clearance order and continue startup.
d.
Cease all startup operations.
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' Page.15-
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QUESTION: _ 016 (1.00)
.
.During power operations (Mode Switch in RUN) and APRM calibration,-power is-lost-to APRM-A.
What resultant conditions should the operator. expect?
a.
Full SCRAM when any of the APRMs in Trip _ system "B" are' bypassed.
b.
Full' SCRAM when reactor power reaches 15%.
c.
1/2. SCRAM (Trip System "A")
-
,
d.
Rod Block due to Power to Flow restrictions.
l QUESTION: 017 (1.00)
Which statement correctly describes the-relationship between the Reactor-Recirculation system Jet Pumps and the core flow measuring system?
~
a.
All 20 Jet Pump flow indicators are calibrated identically to the.
same standard.
b.
Reverse flow from a single Jet Pump.is ignored in_the measuring process for total flow.
,
c.
Jet Pump flow indicator signals. (NBI-FI-92A & B) are subtracted'and used as a substitute value for PMIS during single.: loop operation.
- d.
All 20 Jet Pump d/p measuring devices utilize an identical. upper'
diffuser' tap.
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. REACTOR OPERATOR
.Page 16 i
,
QUESTION: 018 (1.00)
.
The Reactor Core Isolation Cooling (RCIC) system has' initiated on-low reactor-t water level.(-37 inches).
Which one of the following is the EXPECTED
"
sequence of events once reactor vessel level recovers to +58.5 inches?
-
,
(Assume the RCIC system is the only system injecting.)
a.
The RCIC turbine trips allowing' level to decrease to.-37 inches.
.
.where it (high level signal)'must be manually reset to allowJtha turbine to restart to raise level.
,
b.
The Steam Supply Blocking Valve (MO-131) will close' allowing level to decrease to -37 inches where it-(high level signal) must be manually reset to allow MO-131 to reopen to restart the turbine to raise level.
l t
'
c.
The turbine trips and the Steam Supply Blocking Valve (MO-131)L closes automatically. resetting the trip-throttle and allowing level'
to decrease to -37' inches where MO-131 reopens to' restart the turbine to raise level.
d.
The RCIC turbine governor valve goes shut allowing level.to-decrease to -37 inches where the governor valve is automaticallyi i
reset restarting the turbine to raise level.
.
!
l QUESTION: 019 (1.00)
'
During RCIC operation a steam line break occurs between RCIC-MO-16 and RCIC-MO-131 causing an increase in area temperature.
.Which statement correctly describes the subsequent events?-
,
J a.
Actuation.of. switches (DPIS-83,84), a Group V isolation,cand a plant announcement warning personnel of the steam line break.-
b.
Closure of RCIC-MO-16, while RCIC-MO-151 remains:open and the RCIC-'
l turbine' trips on overspeed.
.
c.
RCIC steam line high D/P Alarms, RCIC-MO-15 remains open; and the RCIC turbine coasts down.
'
d.
RCIC-MO-131 remains open and RCIC trips on sensed back pressure.
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. REACTOR OPERATOR Page'17.
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,
.
QUESTION: 020 (1.00)
.,
What.is the purpose of the LOW LOW SET logic of the SRVs?
a.
Minimize.the cyclic stresses applied to the reactor pressure
'
vessel.
'
~
b.
Minimize the reactivity effects on the-core due to SRV cycling.
->
c.
Reduce high frequency loadings on the containment.
'
d.
Prevent SRV actuation if reactor water' level is below TAF.
.
.s I'
Q~UESTION: 021 (1.00)
i
- t What.is the purpose of the containment atmosphere monitoring' system?-
"
a.
Monitors Oxygen concentrations in the secondary containment.
.
b.
Monitors Hydrogen and Oxygen concentrations-in the drywell and.
s suppression pool air spaces.
c.
Provides input signals to'the PCIS for automatic. isolation.
d.
Supports the SGT in processing effluent from the Primary
'
Containment.
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QUESTION: 022 (1.00)
What' indications.are available to'the reactor _ operator in thefcontrol roomL for a leaking' nuclear pressure relief system (NPR) relief valve condition?
'
a.
The valve's amber indication light on' control' panel 9-3fwill
illuminate.
'
b.
The leaking relief' valve would cause the vacuum breaker to operate.
and indicate in the control room.
c.
Discharge piping temperature iscrecorded in the control room 300 deg F on panel 9-3.
and will annunciate:at >
,
d.
The discharge pipe acoustic monitor will-sense the leakage and annunciate on panel 9-3.
-
QUESTION: 023 (1.00)
During a turbine trip condition, one of the four control' valves fails to
'close.
What additional protection exists to prevent the turbine from overspeeding?
a.
Turbine Bypass Valves close b.
68A Pressure Regulating Valve closes c.
Interceptor Valves close d.
Main Steam Stop Valves close l
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.DREACTOR OPERATOR Page 19~
.
. QUESTION: 024 (1.00)
,
What is the purpose of a turbine trip on high reactor waterJ1evel +58.5'
j inches?
I
~
a.
To prevent a generator motoring condition..
b.
To prevent a turbine overspeed condition.
' I
'
c.
To preclude moisture carryover to the turbine.
d.
To preclude a sudden loss of condenser Vacuum.
,
,
'
r QUESTION: 025 (1.001
'_Why is it not recommended to place the RFPT speed control Pneumatici.AssemblyL
- -in,the manual regulation mode?
The manual mode defeats any feedback control 1(speed and fl'ow)..
a.
b.
Taking manual control will cause the LP and HP~stop. valves-to full-open resulting in a-turbine overspeed trip.
q c.
Taking manual control allows control of the turbine speed with'the-
!
high pressure stop valves at higher power levels.
,
d.
Taking manual control will defeat the turbine hydraulic overspeed d
trip.
-j
QUESTION:'026'
(1.00)
. What will cause the RFPT standby AC lube oil' pump to start?'
a.
Depressing the Emergency Oil Pump Test switch, 4QTE1A(B).
~
i b.
A loss of TEC to the oil coolers.
j
.c.
. Low control and bearing-oil pressure.
d.
Depressing the Thrust Bearing Wear test push button 4TBT-1A(B).
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. REACTOR OPERATOR Page.20 b
,
.
'
QUESTION: 027 (1.00)-
~For the Emergency Diesel Generator which of the following will'cause an-f actuation of the 86 lockout relay?
'
,
a.
A reverse power condition or a l'oss of field
.
b.
A trip of the overspeed governor at 500 rpm.
I c.
High vibration of the diesel engine.or generator.
'
d.
Low Lube Oil pressure or High jacket water temperatures.
'
-
t
-
QUESTION: 028 ( 1.' 0 0 )-
'
- What are the consequences.of the Diesel GeneratorJBreaker mode switch 1(Bd-C)
being in the REMOTE position when an automatic start signallof 1F(1G). Bus' -
.Undervoltage~is received?
.
a.
The DG will start..The DG Breaker will-~ automatically!close.
b.
The DG will NOT' respond to a Bus Undervoltage Automatic' Start
.
Signal.
-
,
c.
The DG will start.
The DG Breaker will.NOT automatically close.
d.
The DG will start.
The DG Breaker must be CLOSED manually-from the,
local DG Control Panel.
'
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J. REACTOR OPERATOR lPage 21-
&
P
-QUESTION: 029 (1.00)
-i Why should an idle loop Reactor Recirc pump'NOT be started.when'the idle loop i
-
temperature is > 50 deg F of the operating loop's temperature and >50.deg F
of the core inlet temperature?
a.
Starting the pump could suddenly increase reactor power.
l b.
It could cause damage to the pump and induce abnormal stresses on l
the vessel bottom head.
!
c.
It could cause bottom head fluid temperatures of 145'deg.F >
i Saturation temperature of core fluid temperatures.
e
'
d.
It could cause damage to the jet pump riser bracing.
QUESTION: 030 (1.00)
e Which one of the following will NOT cause,the reactor recirculation MG set
.
drive motor breaker-to trip?
,
'
a.
Motor Generator set motor or generator air temperature > 260 deg F.
,
!
b.
Startup transformer or Normal transformer undervoltage.
,
c.
Generator Lockout.
d.
Suction valve greater than 90% open (MO-43A or B).
'
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t/ REACTOR' OPERATOR Pagei22
i
,
!
'
' QUESTION: 031 (1.00)
l t
Select the statement that describes the RHR Shutdown Cooling (SDC) mode of
,
operation.
i a.
SDC provides a means of reducing the primary
.;
containment temperature.
i b.
SDC removes decay. heat from the reactor after shutdown and-maintains cold shutdown condition.
Lt c.
SDC removes decay heat by condensing reactor' steam and returning the condensate with the RCIC system.
d.
SDC removes stored and decay heat from the reactor under' accident conditions.
.;
-QUESTION: 032 (1.00)
The fuel storage pool has 1/4 core discharged.from the previous refueling.
outage and the entire core off loaded from the past operating cycle.
This.
'
heat load has required supplemental cooling with the RHR system.
A loss of-
,
the RHR system could result in.which of the following?
j a.
An increase in Fuel PoolLtemperature.
y b.
A potential drop ~in fuel pool water level.
j
.
c.
Causes an' auto lineup of the Fueling Pool Cooling system to the
-
Service Water system.
i d.
Causes an auto lineup between.the skimmer. surge tanks andLthe-condensate transfer system.
.!
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oREACTOR OPERATOR-Page-23
,
h
- QUESTION:1033 (1.00)
Select the condition which will actuate the Rod : Position and Information System (RPIS) Rod Overtravel annunciator.
Scram signal is present with control rods'in past 00 position.
^
a.
b.
Rod coupling check shows control rod is uncoupled.
c.
Control rod inserted past 00 position using Emergency Rod In.
!
$
d.
Collet finger damage allow rod to drift full out of the~ core.
QUESTION: 034 (1.00)
.Which most accurately describes RHR system Loop'"A" lineup for containment spray?
a.
RHR-MO-39A Open, RHR pump
"A" running,'RHR-MO-38A (Torus: Spray.
.
Throttle valve) closed, and RHR-MO-31A Drywell1 inboard; spray valve
open, and throttle with RHR-MO-26A Outboard Valve.
l b.
RHR pump
"A" running, RHR-MO-39A Open, RHR-MO-38A Open,' RHR-MO-26A Drywell Spray Outboard valve Open and throttle the inboard RHR-MO-
,
i 31A valve.
c.
RHR-MO-39A (Suppression Pool Cooling) valve Open,-RHR pump "C" running, RHR-MO-38A Open, RHR-MO-31A, Inboard Drywell Spray Valve Open, and Throttle with RHR-MO-26A Outboard valve.
d.
RHR SW system in service, RHR-MO-39A closed, RHR-pump "C" running,.
'
,
RHR-MO-38A closed, RHR-MO-31A Open, RHR-MO-66A throttled closed; and RHR-MO-26A throttled Open.
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Page 24 REACTOR' OPERATOR c.
.
QUESTION- 035 (1.00)
Which of the'following^RPV level instrumentation provides an action / interlock-signal for the RHR containment spray control?-
'a.
Fuel zone instrument b.
Wide Range instrument c.
Narrow Range GEMAC d.
Narrow Range Bartons
= QUESTION: 036 (1.00)
.i What is the purpose of the pressure maintenance function of the RHR systera (Containment Spray System Mode)?
Prevents overpressurizationLof RHR system piping)from the a.-
Minimizes potential; damage to the RHR heat exchanger by' decreasing 1
~
~
~
the dif ferential. pressure between the-RHR ~ and RHR Service Water systems.
c.
Maintains piping: full (Drywell & Torus Spray. Lines)-and pressurized below the blowdown permissive pressure; d.
Assists in proper flow to the Fuel-Pool Cooling-system.
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Page 25 QUESTION: 037 (1.00)
!
.
~
Which of the-following statements describes the reason for placing the nain steam line isolation valve's (MSIV) control' switches to the CLOSE position following a Group 1' isolation?
a.
The MSIVs are interlocked to prevent resetting theLisolation signal.
with the control switches in AUTO /OPEN.
,
b.
.The steam line drain valves cannot be opened to equalize pressure with the MSIV control switches in AUTO /OPEN
'
"
The MSIVs will automatically re-open when the Group',1 isolation is c.
reset.
d.
Resetting. the i' solation will re-open the outboard MSIVs creating
'
a large differential pressure across the' inboard MSIVs.
e l
QUESTION: 038 ( 1. 0 0 ).
.
Select the sources of emergency oil pressure'available to supply the_ Main Generator air side gland seals in the event the normal air' side seal oil pump becomes inoperable.
-
1.
Turbine oil transfer pump j
2.
Main lube oil pump 3.
Turning gear oil pump
'
4.
Hydrogen side seal oil pump-5.
High pressure seal oil backup pump
'
Choose from the following:
a.
1, 3 ', 5
,
,
b.
2, 3,
c.
1, 4,
'd.
2, 3,
.
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OLREACTOR OPERATOR Page.26'
,
.
QUESTION: 039-(1.00)
Select the conditions that must be established to manuallyLclose the Main Generator output breakers from the control room.
'
,
1.
Main turbine generator is at rated speed.
,
2.
Breaker's synchronizing-switch is in AUTO.
3.
- Breaker's synchronizing switch is in Manual.
4.
Breaker's selector switch in switchyard in REMOTE.
Choose from the following:
a.
1 and 2 b.
2 and 3 j
'
c.
3 and 4 d.
2 and 4
!
-QUESTION: 040: (1.00)
,
In addition to.the Main Generator output breakers tripping, what automatic actions are initiated by the Main Generator Relays tripping?
a.
Energizes the solenoid trip on protective block.
b.
Running phase bus duct cooling fan trips.
,,
,
c.
Seal oil backup pump starts, d.
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'
.
QUESTION: 041 (1.00)
Which of the following conditions will'cause a trip of-~the condensate pumps?'
a.
Trip of a condensate booster pump.
b.
An overcurrent/ ground condition, c.
When one main condenser hotwell level controller becomes.
inoperative.
d.
Depressing the Condensate system and Condensate: Booster Pump; Minimum Flow Valve reset.
f QUESTION: 042
'(1.00)
Which of the following accurately describes the operation'of the-Feedwater Heaters level control?
a.
The Level Control ~ Valves (LCVs) close on an air pressure signalL>30 psi and sends an Open signal to-the extraction. steam dump. valves on Heater Hi-Hi Level Trip.
b.
The LCVs fully close on an air pressure signal of 6 psi Land; provide -
level control between the #2,3,4,5 Heaters and the main condenser, c.
The Level Control Transmitter output signal serves as a LCV feedback signal and causes the LCVs to' fully close;on lossiof; operating air pressure.
d.
A Hi-Hi Meater Water Level will cause the non-return valve safety pilot va.ves to trip and the extraction steam dump valves _to close.
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/Page 28'
QUESTION: 043-(1. 0 0 ) '
- The' unit is online at 75% power and OCB 1604 -(Auto. Transformer' connect) L and OCB 1606 (Auburn 161kV connect) are open.
What sources of, power are available to--the 4160V AC Distribution syst'em?
1.
Startup Power Transformer 2.
Normal. Station Service Transformer 3.
Emergency. Power Transformer o
!
4.
Standby AC Power-System
.I a.
1-2-3
b.
2-3-4-
,
i c.
1-3-4 i
d.
1-2-4 QUESTION: 044 (1.00)
What condition will'NOT cause an auto transfer of the No-Break Power Supply,
'(NBPP) #1 from its Normal power supply to its emergency backup:AC source?
Overvo1tage or undervoltagelof L+.10% at the inverter
~
a.
b.
An inverter frequency mismatch of + 2 cycles c.
Overcurrent condition at the inverter d.
Failure of the Bypass transformer / inverter selector' switch
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._ REACTOR OPERATOR'
Page'297
.
i
..;
' QUESTION: 045 (1.00)
,
Select the Radiation Monitoring system-instrument that'will' initiate an automatic closure of.the radwaste discharge valvesito the canal'.(AO-231.and;
-
I
'AO-230) when the radiation level setpoint is_ exceeded.
a.
Floor drain sample tank activity monitor i
b.
REC radiation monitor c.
Service Water _ radiation monitor
d.
Radwaste effluent radiation' monitor
!
k QUESTION: 046 (1.00)
-i The computer room halon_ control panel PRE-DISCHARGE' indicating light is
)
illuminated and there is no fire in the computer room.
What action (s):is/are
'
'
"
required to prevent an automatic halon-system discharge?
. Depress-and release Lae Yellow Abort pu'sh button.
>
a.
b.
Depress and hold the Yellow Abort push' button for 20 seconds'
'
then release.
c.
Depress and hold the Yellow Abort push button until the ARM / Disable switch is positioned to Disable.
r d.
Depress and hold the Yellow Abort push button until the Discharge indicating light illuminates then release.
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Page 30
,
QUESTION: 047 (1.00)
What is the purpose of the diesel driven fire pump?
.,
I a.
To supply water to the fire main header automatically upon a header pressure of 65 psig.
b.
To supply water from the FPWST-"B" to the Fire Main Header.
,
c.
To act as a tertiary fire protection backup should the 1E electric fire pump become. inoperable.
d.
To transfer river water to-the fire main header'if necessary.-
!
QUESTION: 048 (1.00).
.
Which of following would NOT be an appropriate reason for the Process computer to require inputs from the TIP system?
,
a.
Determine. substitute data for inoperable LPRMs.
b.
Calculate core thermal limits.
,
c.
Calculate adjustments for LPRMs.
T d.
Determine the location of failed fuel.
.
P QUESTION: 049 (1.00)
'
Which interlocks prevent an inadvertent criticality during refueling.
operations?
a.
Control Rod blocks and Refueling Platform interlocks.
,
b.
Administrative refueling procedures and Technical-Specifications requirements.
c.
SRMs used to monitor reactivity in the core.
d.
SRMs and Control Rod Blocks.
.
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- Page-31'
,
,
!
QUESTION: 050-(1.00)
~
Select from the-following the plant conditions that.will. initiate ' closure oft
- I Ethe mainisteam line isolation valves (Group I isolation).
l
- 1.
Reactor water level'4.5 inches 2.
. Steam-line-. flow 150% of rated flow
.t
~
i 3.
Drywell pressure 2 Psig
.
4.
Main condenser' vacuum 7 inches Hg vacuum Choose from the following:
~
a.
1,2 b.
, 2,3 c.
2,4 a
d.
1,4
.'?
r
- QUESTION: 051 (1.00)
.
. What vessel internals and instrumentation ensure 2/3 core 1coveragefduring-a f
DBA?
a.
-Jet pump diffusers and'LT59 A;& B (Wide Range. Level Instruments)'.
.
b.
Core Shroud and LITS-73-A & B (Fuel-Zone. Level Instruments)..-
c.
' Core Plate and LT-52.A,B,C (Narrow Range Level Instruments).
d.
Core Shroud and LT 61 (Shutdown Range Level Instrument).
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.
_ QUESTION: 052 (1.00)
'Which.one of the following is the correct sequence for returning a' pump to-service before removing the clearance order tag on the-control switch?-
a.
Rack in the breaker,'close the vent or drain valve, open the suction isolation valve, then open the discharge isolation valve.
b.
Close the. vent or drain valve, open the suction valve, open the discharge valve, rack in the breaker.
~
'
c, Rack i.1 the breaker, open the suction isolation valve, then open'
the discharge isolation valve, close the vent or drain valve.
,
d.
Close the vent or drain valve, open the discharge valve, open the suction valve, rack in the breaker.
i
!
QUESTION: 053 (1.00)
i During the performance of a surveillance procedure (SP),.an' indicator which j'
should be lite is not due to maintenance on its power supply.." Verifying the-indicator-lite" is not required to satisfy the SP acceptance criteria.
Which
,
one of the following actions should be taken in accordance with'CNS 0.26,.
- ;
" Surveillance Program" 'to continue the SP.
a.
Request a " Temporary Procedure Change Notice" and. continue when it f
is issued.
-
b.
Record the discrepancy on the data sheet with appropriate comments (
and continue.
i
,
c.
Complete a "Non Conformance Report (NCR), " attach it to the SP and
>
continue.
d.
Complete the SP then fill out a " Component Operability Checklist" and attach it to the SP.
.
.
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.Pagef33
]
-;
-QUESTION: 054:
( 1. 0 0 ) -
During reactor operation, you lose indication on the four-rod' display.. ToJ 1which electrical. distribution panel mustlyou send the: station operator'to-check the RPIS power supply?
{
a.
Critical power-panel.
>
'b.
125V DC distribution. panel, c.
No-Break power panel.
,
d.
RPS distribution panel.
.
>
,
QUESTION M W (1.00)
.With thh' reactor at power and the
"B" RWCU Pump.out-of-service for maintenan the reactor operator observes annunciator'9-4-2/B-4, RWCU Pump
"A" Low Flow, - arms..Af(er notifying-the CRS, which of'the following describes the rea or operah 's required actions? -
a.
Closes RWCU-M f M 0-1.,. opens RWCU-MO-74 (Demineralized Water Suction Byp
)
n clos CRD-189 -(RWCU Minipurge t
e t
e Radwaste/ Main Condenser.
isolation), rejects
' CU-MO-74 (Demineralized
~
b.
Closes RWCU-MO-15, MO-18, o Water Suction Bypass), and clo
-
(RWCU minipurge Isolation).
,
c.
Opens RWCU-MO-74, opens RWCU-MO-15,-18)s ejects waterLto l
Radwaste/ Main Condenser; and calls chemis
' to monitor Reactor
,
!
Water Chemistry parameters.
P d.
Verifies a Group 3 isolation, opens RWCU-MO-70 (D in. Suct.
Bypass), opens CRD-189; and commences reactor shutd n.
.
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REACTOR OPERATORL Page.34 s.
.
E QUESTION: 056 (1.00 The plant is shutdown, the vessel head removed,.and RHR shutdown cooling.in-service.
I & C technicians are_ performing'surveillances on the PCIS. system
,
when the reactor operator observes a PCIS Group 2 isolation signal.
Which of i
the following statements describes the operator's required' actions?
l a.
Verifies the operating RHR pump has tripped, RHR-MO-17,- MO-18 valves have closed, calls I&C to secure-from testing, and places
,
the RWCU system in service with maximum' flow.
,
b.
Verifies the-operating RHR pump has tripped, RHR-MO-17, MO-18 valves have closed, calls I&C to secure from testing, aligns RHR to
an available non-running pump and torus-. suction path, resets and.
starts an available pump,.and resets the Group 2'isclation on panel'
l 9-3.
'
c.
Verifies Group 2 isolation, places RWCU.in service with maximum flow, and uses the main condenser.as a heat sink.
]
d.
Verifies a Group 2 isolation has occurred, opens RHR-MO-66A(B),
A Heat Exchanger (HX) Bypass Valve, Close RHR-MO-12A(B) HX Outlet-
,
Valve,'and reestablish RHR flow.
-
'!
>
.,
QUESTION: 057 (1.00)
,
. Select the action (s) that is (are) required to execute a Plant Management-Information System special function f rom the Industrial Data Terminal (IDT)
!
l keyboard.
!
Type'in the TURN ON CODE for desired function / press; return; type a.
NSSS MENU / press return; type line number of desired function / press
~
return.
b.
Type SPECIAL FUNCTION / press return; select desired function using arrow keys / press return.
- c.
Type.SPECIAL FUNCTION, press return; select the desired function.
and press return.
!
L
..
d.
Type NSSS MENU, press return; press down arrow to select desired
'
function; press-return.
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-
Pagef35:
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.
l
- QUESTION
- .058 (1.00)
Given the following conditions:
Today's date is May 20, 1993
'
A fully trained, male radiation. worker is'25-yrs old'today.
.
He has a current Cooper HP-4 (equivalent to NRC' Form 4) ~ on record.
His previous lifetime whole body exposure is 26. Rem.
What is the MAXIMUM whole body radiation exposure for:1993' allowed by Cooper
,
Nuclear Station Health Physics Procedures for this individual WITHOUT-invoking EPIP limits fo. emergency and accidental' exposure?
,
a.
1250 mrem /qucrtet for.the entire. year.
'
'
-
.,
b.
2500 mrem / quarter not to exceed-4000 mrem for the year.
c.
3000 mrem / quarter for 3 quarters in the year.
,
d.
3000 mrem / quarter not to exceed 5 rem total for the year.
, QUESTION: 059 (1.00)
Which one of the following is usually the cause-for Heat Syncope inca'high heat-environment?
a.
Not moving for long periods.
b.
Profuse sweating.
I c..
Rapid area temperature drop.
d.
Drinking large amount of cold liquid.
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-
REACTOR-OPERATOR-Page-36-QUESTION: 06'O (1. 00 ).
i Which of the following statements describes the requirement for working in
~
'
oxygen deficient environments?
l A positive pressure SCBA shall be used when environments have-less; a.
than 30% Oxygen.
b.
SCBAs shall not-be used. in environs of < 18% _ Oxygen.without a-hazardous work permit except in emergency.
,
c.
Users of SCBAs in deficient Oxygen: environs do not require.
additional personnel support as long as the pre.use examination is performed correctly.
d.
SCBA bottles shall be refilled with Grade D or better air and
"
examined bi-weekly.
'
r QUESTION: 061 (1.00)
Which one of the following represent the lower and upper bounds of percent ~
.j Hydrogen in air where the mixture is flammable'or explosive?
,
a.
0 - 8%
,
b.
4 - 18%
c.
18 - 59%
d.
74%
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_ QUESTION: 062 (1.00)
'
-The reactor has experienced a scram but reactor power has remained above:2.5%-
,
power and-attempts-at alternate' rod insertion (ARI) have, failed.
What~
alternative methods for reactivity control procedurally (i.e. EOP Flowcha~rt'
-;
vand EOPs) exist?
a.
Boron injection using the SLC system.
,
,
b.
Boron injection using the SLC system from the SLC tank followed by a
injection of Demineralized. Water using the SLC system.
c.
Boron injection using the SLC system, RWCU or RCIC' systems.
,;
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Boron injection using the SLC system or ' manual injection
~rnm
sodium pentaborate drums into the RWCU system.
,
I-QUESTION: 063 (1.00)
- l I
!
If entry into the EOPs has 7ccurred and a Failure-To-Scram condition exists y
and Emergency RPV Depressur; zation is. required, what,RPV level condition
!
"
-requires the CRS to flood the primary containment?
a.
When RPV level cannot be. maintained ~at TAF or above using Core
,'
Spray and alternate injection subsystems.
,
b.
When RPV level cannot be restored and maintained above.-30 inches.
ji using Core Spray and Alternate injection subsystems.
c.
When RPV level cannot be restored and maintained at +58.5 inches using Core Spray and Alternate injection subsystems.
,
d.
When RPV level cannot be maintained at or above the Low Level using
'
outside shroud injection systems.
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,
. QUESTION: 064 (1.00)
Trip'of a. single reactor recirculation pump occurs at 70% power with no reactor scram.
The procedure requires thatthe discharge valve for the tripped pump be closed for 5 minutes, then reopened for 6-seconds to:
a.
Prevent hydraulic lock on the RR pump discharge valve.
b.
Prevent overpressurization of the RR pump seal assembly.
c.
Minimize valve stem wear.
d.
Minimize-RR loop to reactor vessel differential temperature.
,
QUESTION: 065 (1.00)
Which of the following describes the probable cause for a loss'of main
condenser vacuum?
a.
Failure of a hotwell level controller.
- b.
Closure of RWCU blowdown valves RWCU-MO-56 and'MO-57 simultaneously.
c.
Circ Water pump trip or condenser expansion. joint-leakage.
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d.
Tripping of a condensate booster pump.
.
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QUESTION: 066 (1.00)
'
-The main; generator has just been synchronized to the grid and is carrying.
approximately 100 MWe. - The "TG ROTOR HI VIBRATION" alarm is received: shortly.
'
- thereafter.
A quick check'of the bearing and shaftivibration. recorders show?
,
vibration at 12 mils and trending up rapidly.
'
The1 correct action shouF1 be to:
a.
Immediately unload the turbine and consult the Alarm Procedure.
b.
Trip open the generator output breakers,-3310 & 3312.
l
'
c.
Initiate a reactor scram and trip the reactor.1aALL -
d.
-Immediately trip the turbine manually.
.;
t QUESTION: 067 (1. 0 0 ) '
t
?
During a loss of a 250V DC Bus, the operator is directed by the procedure to-l immediately:
a.
. Ensure that the No-Break' power panel indicates about.240 volts.
l i
"
b.
Switch the Reactor Building starter rack to the unaffected bus.
c.
Advise the Shift Supervisor that the condition may require an' Alert'
to be declared per EPIP 5.7.1.
'
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d.
Switch the 250V DC charger 1C to the affected bus to supply th'e-critical.
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R Page 40
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- QUESTION: -068 (1.00)
l
~ Which of the following statements describes those.setpoints at which-the-operator would manually. trip the main turbine?
Generator motoring (30 psid decreasing on high press turbine).
a.
-
,
- 180 deg F bearing oil temperature.
- or 250.deg1F LP turbine exhaust temperature, b.
- low bearing oil pressure (< 7 psig).
.,
- high bearing metal temperature (225 deg F).
- or condenser low vacuum.< 19 inches Hg.
,
c.
- high bearing oil Temperature (180.deg F).
- high vibration (12 mils).
- or high LP-turbine exhaust temperature (250 deg F).
d.
- high vibration (12 Mils).
- low bearing oil pressure (< 7 psig).
- or high bearing' oil temperature (180 deg F).
.
QUESTION: 069 (1.00)
-
Which statement.below describes the concurrent RPV~ level control actions
'
taken when RPV pressure cannot be stabilized to'below 1045'psig afterLa reactor SCRAM and entry into the EOP?
,
.,
a.
Depressurize the RPV and maintain RPV level above -150 inches using
.
SLC from Demineralized Water.
t b.
Depressurize RPV, maintain RPV level above -150 inches using HPCI
,
with suction from-the ECST.
c.
Depressurize RPV below 80 psig and maintain RPV' level above -40 inches with HPCI (defeat Hi suppression pool water level suction transfer logic maybe necessary).
'
d.
Depressurize RPV and maintain RPV level above -40 inches.using RCIC aligned to the suppression pool.
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+ REACTOR" OPERATOR Page241-
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,
jQUESTION: '070 (1. '00 ) }
.During implementation of emergency operating procedure RPV Pressure Control ~,.
why is.it important,to' maintain' key parameters within the' Heat Capacity'
~
. Temperature Limits (HCTL) of the suppression pool?;
.
a.
To: ensure.RHR pump NPSH limit 11s satisfied.
' '
b.
To. preclude RHR pump vortex conditions.
'c.
To provide a' sufficient accident / transient heat sink.
.
d.
To preclude Core Spray pump vortex conditions.
E
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'
QUESTION: 071 (1.00)
Under what conditions would the RWCU be used to maintain RPV level by means, of the blowdown line to the main condenser?
During reactor startup or refueling with low steam flos;and CRD a.
injection.
'
b.
During reactor shutdown when. RHR is injectiing.
During post' trip conditions when ECCS injection overfillsithe RPV.
c.
,
d.
During hot standby with drain flow rest ricted -to limit: Non-Regenerative Heat Exchanger (NRHX)_ outlet temperatures to less than
,
150 deg F.
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b QUESTION: 072 (1.00)
During primary' containment (PC) venting utilizing drywell vent path, which
.
Hof the following' key parameters should be monitored closely?
,
a.
Containment atmosphere. (Hydrogen & Oxygen) and suppression pool level.
b.
ERP effluent radiation monitors and suppression pool temperature if:
c.
Suppression pool-level and RPV pressure.
d.
Suppression pool level and status of torus. exhaust outboard-isolation valve.
- >
,
.
QUESTION: 073 (1.00)
P Why does the Technical Specifications place.a time limit on the drywell and suppression chamber 24 inch exhaust purge and vent system yearlyfoperational
-
period'for when the reactor is critical or when reactor coolant temperature
> 212 deg F?
a.
To decrease the frequency of maintenance on the 24 inch. purge and vent valves.
b.
Reduce the probability'of a LOCA-occurrence.when the.24 Inch l purge and vent valves are opened in series.
c.
So that venting with the inboard exhaust bypass valve and the outboard exhaust valves in-series does not cause damage to the SGT filters during a LOCA.
d.
To ensure primary containment pressure control during PC venting can be maintained utilizing the 1 inch PC vent paths in a post LOCA r
condition.
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QUESTION: 074 (1.00)
-For which of the'f-. lowing suppression pool conditions is it.necessarf to scram the reactor and make the requisite announcements?
,
t a.
During testing.and suppression pool temperatures reach 100 deg F.
b.
During power operation and a leaking relief ~valvo'causes the.
suppression pool temperature to reach 95 deg F.
c.
At any. operating condition and suppression pool temperature reaches 110 deg F.
'
d.
Any mode of operation when suppression pool temperatures reach 100-deg F.
-'
.
QUESTION: 075
.(1.00)
,
Given that:
.-'
-Reactor Pressure is 900'psig,
.
-Suppression pool level is 13 feet, and-
-Suppression pool average' temperature is 140.deg-F
,
Which one of the following is the suppression pool Heat Capacity Level'and-Heat Capacity Temperature. Limits?
(EOP Reference provided)
a.
feet and 200 deg F b.
10 feet and 180 deg F
c.
11 feet and 190 deg F d.
12 feet and 200 deg F
.
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r QUESTION: 076 (1.00)
i During power operations a sudden rise in reactor power _is indicated by neutron monitoring, steam flow and feed flow.
After notify-the SS, which statement describes the operator's required actions if an automatic' reactor scram is not generated?-
a.
Check to verify if cause was due to Recirculation System controls,.
or DEH system controls, and hold power at existing level.
,
b.
Reduce reactor power with Recirculation flow, if 2 or more j
'
LPRMs read >-100, scram the reactor.
-
i c.
If LPRM Upscale alarms and 1 LPRM reads > 100, scram.
'
the reactor.
i d.
LPRM Upscale Alarms, if 2 or more LPRMs read > 100, bypass failed LPRMs and refer to Tech Spec for the LCO.
.t
-QUESTION: 077 (1.00)
'
.
A plant accident has~ occurred and the reactor is-not shutdown.
The turbine ~
building KAMAN reads 2E+8 micro Ci/sec, Whole Body dose rate at the site.
.
>
boundary is 1 Rem /hr.
For a declared. General Emergency what is the MINIMUM
Protective Action Recommendation (PAR)?
_,
Evacuation of the public for a 5 mile radius and 10 miles ^ downwind'
a.
in at least 3 sectors and shelter in_ remainder of.10 mile EPZ.
b.
Evacuation of the public for a 2 mile _ radius and'10 miles' downwind in 3 sectors and shelter in remainder of the.10 mile EPZ.
c.
Evacuation of public for a 2 miles radius, and 5 miles downwind in
.
3 sectors, and shelter remainder of the 10 mile EPZ.
d.
Evacuation of the public for a 5 mile radius and 7 mile downwind in 3 sectors and shelter in remainder of'the 10 mile EPZ.
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-QUESTION: 078
. (1. 0 0)
-During a loss of offsite power and a failure of one diesel to start, which of the following describes how Reactor Equipment Cooling (REC) would be:af fected?
'
a.
The condition would cause the Service Water pump'in the'available~
Loop to start if the pump is in~the Auto Mode.
b.
Causes the SW pump in the available Loop to start 15-seconds after the available-DG starts'and loads, and SW-MO-37'(Loop Crosstie Valve), opens at < 20 psig in
"B" Service Water Loop.
,
c.
Causes the Service. Water pump in the available Loop to' start 15 seconds after the available DG starts and loads; SW-MO-37 Closes ati
< 20 psig in Service Water Loop
"B" and Nbnual' alignment of Service-Water for REC is required due to a loss of the plant air system.
>
d.
Manual loading a Service Water pump'onto the available'4160V bus will_be required; SW-MO-2128 and SW-MO-2129:open to supply Service.
'
Water-to the Service Water pump gland seal system and SW-MO-37 20 psig in SW Loop
"B".
(Loop. Crosstie Valve) closes at <
_
QUESTION: 079 (1.00)
Which one of the following is the MINIMUM reactor-water level necessary to aid in removing decay heat by thermal convection' flow (natural circulation)-
if ALL shutdown cooling is lost?
a.
37.0 inches b.
48.0 inches c.
58.5 inches
,
d.
92.0 inches
.
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-QUESTIONS-080 (1.00)
l
- To' preclude a potential refueling accident-at what.' minimum-level'.'should;the spent fuel pool water-level be maintained?
l
'
'6 1/2'ft. above top of fuel'
a.
-
b.
. 8-1/2 ft, above top of fuel
,
c.
. 10.ft. above top of fuel
.
d.
12 ft. above top of fuel
,,
..,
-QUESTION: 081 (1.00)
o
'
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.!
i During post-accident conditions, at.what suppression pool water level musti
.the Primary Containment vent. paths be changed.from the Torus vent.line to'the
~
Drywell vent'line in order to maximize-suppression pool scrubbing.of fission'
,
yproducts?.
,
.
a.
' < 19.5 feet
-
b.
. > 22 feet q
.c.
< 24 feet
.. i-l d.
> 28=.5 feet
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. QUESTION: 082- (1.00)
'
i For the. conditions listed:
-Indicated' reactor water level on the fuel zone Yarwayfis:-10 inches;
eReactor pressure is 1000 psig
-Drywell Bulk Temperature is 200 deg F Which of the following choices is the difference from actual level'to thetTop.
~
of Active Fuel?
a.
-20 inches
!
b.
+12 inches c.
+18 inches
. ?
.
-
d.
+37 inches
!
-QUESTION: 083 (1.00)
'
t EOP ATTACHMENTS PROVIDED:
' !
For the conditions listed below:
,
-RPV Pressure 1045 psig
.
-PC-TI-505A-D Average Temperature 500 deg F
>
-Drywell Maximum Run Temperature is.525'deg F What indicated RPV water level should be expected on NBI-LI-91A?
a.
-137 inches Fuel Zone
'
b.
20 inches Shutdown Range c.
2.5 inches Narrow Range I
d.
30 inches Steam Nozzle Range r
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- Page-48
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.
. QUESTION: 084 (1.00)
'EOP ATTACHMENTS PROVIDED:
Primary Containment venting is in progress using the torus vent line'and"1 RHR pump is providing suppression pool cooling.
Given the following-Suppression pool water level +15 feet
-Torus pressure 11 psig
,
-RHR pump flow 6000 gpm What temperature limit for suppression pool water exists before RHR. pump Loperation is challenged?
'
.
a.
160 deg F j
b.
190 deg F l
c.
200 deg F d.
230 deg F
,
t i
QUESTION: -085 (1. 00 )'
EOP Flowchart Reference is provided:
During EOP actions, Primary Containment flooding'is required.
Whatjis the
-
Maximum Primary Containment Water Level Limit (MPCWLL) before injection from, external' sources into the primary containment-if_ Torus pressure is 29 psig?'
a.
70 feet b.
80 feet c.
95 feet j
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d.
100 feet
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'Page 49
)a-QUESTION:.086 (l'. 0 0 );
,
During' primary. containment inj ect ion', what is the SRV Tail'. Pipe Level > Limit
-
(SRVTPLL) ? -
q
'
a.
12 feet b.
16 feet
.t c.
18 feet d.
28 feet'
!
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.
QUESTION: 087 (1.00)
,
"'
Secondary Containment reentry following an accident in.which maximum safe -
operating (MSO) radiation levels have been exceeded is required.
What.is the.
-
maximum stay time and what dosimetry would be used?
,
a.
,
.!
b.
I hr, TLD, SRD &. alarming' dosimeter'-
c.
1 hr, dosimetry determined by HP f
d.
2 hrs, dosimetry determined by HP
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!
-QUESTION: 088 (1.00)
.Which of the following statements describes the automatic
' isolation of the control. room ventilation system upon high radiation?
a.
Monitor RMV-RM-1 trips at 4,000 CPM; Booster Fan BF-C-1A starts.if-selector switch is in Auto; and the emergency bypass valve opens.
b.
Monitor RMP-RM-452 trips at 100 mr/hr; booster fan BF-C1A starts; and the emergency bypass valve closes.
Monitor RMP-RM-150-trips at equivalent dose of 25 mrem whole body;
'
c.
Booster. fan BF-C-1A starts if select switch in Auto; and the emergency-bypass valve opens.
4, d.
Monitor RMP-RM-251 trips at 3 X Normal full' power background radiation; booster fan BF-C-1A starts; and. emergency bypass. valve opens.
,
r
!
QUESTION: 089 (1.00)
the three parallel action paths for secondary.
containment _ control-are Secondary Containment temperature, Secondary-
-
Containment Radiation, and Secondary Containment water' level.
Which of the following statements describes actions taken to control Secondary Containment-temperature?
Operating the sump pumps and isolating discharges'into the a.
b.
Operating Reactor Building HVAC if Reactor Building exhaust plenum
!
Radiation levels are below 10 mr/hr.
!
c.
Operating the Standby Gas Treatment. System.
q d.
Operating Reactor Building HVAC as long as Reactor' Building exhaust j
plenum Radiation levels are below-25 mr/hr and high drywell pressure and low RPV water level isolation interlocks have been defeated.
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.Page 511
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.
i QUESTION: 090 I(1.00)
.
.
-Which one of the following THERMAL HYDRAULIC INSTABILITY conditions requires
'an immediate reactor scram?
'
a.
LPRM upscale-to-downscale oscillations.
.
b.
APRM swing of 5% occurring each 2.3 seconds.
,
c.
Increasing oscillations on APRM's with an increasesin thermal power.
'
d.
Increasing' oscillations on'APRM's (10%. peak-to peak) and operation; in the region of instability.
~
~
i
'l
- .
. QUESTION: 091 (1.00)
,
'Which one of the following methods will result.in the. largest.overallL
. differential pressure across the Control Rod Drive piston for inserting-f
'
-
f a.
Open individual scram. test switches',
b.
Maximize CRD cooling water differential pressure..
c.
Drive control rods using maximum drive pressure'.
>
d.
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Pagej52
.-
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'[ QUESTION:.092-(1.00)-
Which one of the.following is.the RPV water level upon which the Hot Shutdown
'
Boron Weight is based?
a.
High level trip setpoint
.
..
b.
Low level scram setpoint c.
Top of active fuel
,
d.
'2/3 core height i-QUESTION: 093 (1.00)
.
~Which one of the following'is the required action for a sustained reactor
'i period of~30 seconds during reactor startup?-
'a.
Manually scram the reactor.
.
b.
Insert control rods until period is greater than:50 seconds.
n c.
Stop. control rod movement until period: increases. greater than.50 l
seconds.
.
d.
Stop withdrawing control rods and contact-ReactorjEngineering.
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,
JQUESTION: '094 ( 1. 0'0 )
~
During normal condensate system operation,.'theLcondensate booster pump.
bearings are lubricated by:
a.
Condensate booster pump auxiliary oil pumps.
b.
A connection with the feedpump turbine lnbe:. oil.
c.
Self lubrication from the on-board oil 1. reservoir.
d.
A connection with the main turbine lube oil.
,
t Y
. QUESTION: 095 (1.00)
' Select the statement that describes the diesel governor RATED SPEED RESET.
,
a.
Resets the governor afterlan overspeed' trip condition has cleared, b.
Resets the governor for a generator frequency.of'60.2.1
. t c.
Disables the. governor. control switches while manually resetting the governor.
.
d.
Resets the governor for test operation of the~ diesel generator.
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Page.54
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-
QUESTION: 096 (1.00)
'The'RCIC system is operating at 400 gpm with.1120-psig discharge pressure r
-
when the turbine trips.
Which of the statements'below' describes the
'cause of the turbine trip?
a.
RCIC pump suction pressure of 15 psig.
.
e
'
b.
Steam supply pressure of 90 psig.
>
c.
RCIC turbine exhaust pressure of 5 psig.
,
d.
Steam supply line space temperature of 200 deg F.-
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~ QUESTION: 097 (1.00)
-Select the methods used to maintain the Standby Liquid Control poisoniin,
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solution.
a.
Service air mixing and elevated: water temperature.
b.
Heat-traced piping and' elevated water temperature.
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Heat-traced. piping and air bubbler standpipe.
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c.
d.
Elevated water temperature and storage. tank outlet nozzles;on thel side of the tank.
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d-QUESTION: 096 (1.00)
- t Ten minutes after the diesel generators energize'the. critical buses following a loss of all AC power, what is the. power source for the PMIS-UPS
~ distribution panel?
a.
MCC-L b.
MDP-1 c.
MDP-2 d.
.PMIS Battery
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zREACTOR' OPERATOR Page-55
- QUESTION
- 099 (1.00)
-Which syctems comprise the Core Standby Cooling systems?
I a.
c.
ADS, LPCI, HPCI, CS-QUESTION: 100 (1.00)
Which components would isolate on a valid high drywell pressure signal?-
RHR sample valves, Drywell equipment drains, and Recirculation a.
loop sample valves.
b.
RHR sample valves, Drywell equipment drains, and' Steam' supply..to AOG.
c.
RHR sample valves, Recirculation-loop sample valves, and Steam supply to ADG.
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d.
Drywell equipment drains, Recirculation-loop sample. valves, and Steam supply.to AOG.
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END OF EXAMINATION **********)
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ATTACHMENT 2
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- ' REACTOR OPERATOR WRITTEN EXAM 1 NATION Page
ANSWER KEY MULTIPLE CHOICE 023 d
001 d
024 c
002 d
025 a
003 d
026 c
004 c
027 a
005 d
028 c
006 d
029 b
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007 c
030 d
008 a
031 b
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009 b
032 a
010 b
033 b
011 d
034 c or d ei.
.c @4 GN 035 a
013 a
036 c
014-c 037'
a 015 b
038 d
016 c
039 c
017 d
040 a
018 c
041 b
019 a
042 a
020 c
043 b
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021 b
044 d
022 c
045 d
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ANSWER KEY 069 b
046 c
047 b
070 c
048 d
071 a
072 b
049 a
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073 b
050 c
051 b
074 c
,
052 b
075 b
053 b
^76 b
054 c
077 c
?:r 7tItd6 078 c
056 b
079 b
080 b
057 a
058 b
081 d'
059 a
082 d
060 b
083 a
061 d
084 d
062
085 d
063 b
086 b
064 d
087 c
065 c
088 a
066.
d 089 b
067 a
090 d
068 c
091 a
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REACTOR OPERATOR Page
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ANSWER KEY c
092 a
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093 b
094 c
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'095 b
096 d
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097 b
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098 d
099-d s
100 b
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ATTACHMENT-3
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- o BOX 98. BRoWNVILLE. NEBRASKA 68321
- ELEPHoNE f402382S3811 Nebraska Public Power District m==mi
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i STM93137 May 28. 1993 L.S. Nuclear Regulatory Commission
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r 611 Ryan Plaza Drive Suite 1000 Arlington. TX 76011 Attention:
Mr. John Pellet Subject:
Comments - SRC Examination Administered May 24 through May 28, 1993.
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Dear Mr. ?ellet:
Attached please find our comments on questions associated with the subj ec t examination given on May 25. 1993.
'.?e believe that our comments should be considered in the grading of the written examinations.
Please contact us if you.:ould like to pursue our comments further or if
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additional clarification is desired.
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,-WK2dMMW
[JhnM.Meacham
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ite Manager
"JD/dj m
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Attachments pc:
G.R. Horn w/ attachments R.L. Gardner w/ attachments R.
3rungardt w/o attachments
- 'J. Dutton
'/ attachments
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TM File w/ attachments
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WW M W W N
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CRALLENCE QUESTIONS FOR 1993 CNS HOT LICENSE EXAM l
QUESTION:
012 (1.00)
i RPS Motor Generator (MG) Set lA has tripped,.Upon completion of repairs and testing, what are the correct recovery actions needed to return the components'
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to service?
>
a.
No immediate actions are required.since alternate power is
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supplied from MCC-T to RPS power. panel lA due to Auto Switching upon loss of MG set lA.
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b.
If interim power is supplied from MCC-T, then transfer back to the MC Set lA power source must be accomplished with the RPS PPIA deenergized locally.
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c.
Transferring power back to the MG set lA will result in a temporary 1/2 SCRAM but will not affect the Primary Containment Isolation System (PCIS).
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d.
Transferring power back to the MG set lA from the critical
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distribution panel CDPlB can be accomplished through reset of the 1/2 SCRAM, a power transfer switch in the control room, and'
resetting the affected electrical protection assemblies from the control room.
ANSWER:
012 (1.00)
b.
REFERENCE:
CNS System Manual COR002-01 IV.A
[3.4/3.9]
212000A201
,(KA's)
Comments Answer a is incorrect; there is no Auto Switching of PRS PPlA upon the loss of the MG set.
,
Answer b (key answer); interim power is not-supplied'from MCC-T, it is supplied from MCC-TX via CDP-1B (SOP 2.2.22 Section 8.4.2 Attached).
Answer e is incorrect; PCIS is affected on a transfer of RPS Power.
Answer d in incorrect; the EPAs cannot be reset from the control room.
We recommend this question be deleted from the exam since there is no correct answer.
l The reference given in the exam key seems to be incorrect: it should probably be COR002-21 IV.A.
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QUESTION:
034 (1.00)
Which most accurately describes RRR system Loop
"A" lineup for containment spray?
a.
RHR-MO-39A Open, RRR pump
"A" running, RHR-MO-38A (Torus Spray Throttle valve) closed, and RHR-MO-31A Drywell inboard spray valve open, and throttle with RRR-MO-26A Outboard Valve.
b.
RRR pump
"A" running, RHR-MO-39A Open, RHR-MO-38A Open, RHR-MO-26A Drywell Spray Outboard valve Open and throttle the inboard RHR-MO-31A valve, c.
RHR-MO-39A (Suppression Pool Cooling) valve Open, RHR pump
"C" running, RRR-MO-38A Open, RHR-MO-31A, Inboard Drywell Spray Valve Open, and Throttle with RRR-MO-26A Outboard valve.
d.
RHR SW system in service, RHR-MO-39A closed, RHR pump
"C" running, RHR-MO-38A closed, RHR-MO-31A open, RHR-MO-66A throttled closed, and RHR-MO-26A throttled Open.
ANSWER:
034 (1.00)
c.
REFERENCE:
System Procedure 2.2.69,3, pg 2
[3.1/3.4]
226001A105
.(KA's)
Comments Answer c (key answer) is correct; this lineup will spray the Torus
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and the Drywell but it does not reference any cooling lineup.
f Answer d is correct; this lineup will spray the Drywell with cooling lined up.
By CNS EOP convention, Reactor Operators are not given the order to lineup for " containment spray", they are told to lineup for either Torus or Drywell sprays, or both.
These specific directions insure E0Ps are not violated.
Per E0P 3A PRIMARY CONTAINMENT CONTROL step DW/T-8, the Drywell will be sprayed for elevated temperature control making answer d the most correct.
When spraying for containment pressure control per EOP 3A steps PC/P-5 thru PC/P 10 answer c would be the most correct.
The operator is required to determine which is more correct, without sufficient containment parameters.
We recommend accepting either answer e or d, because there was no clear direction to the operator on which areas of containment to be sprayed.
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QUESTION:
055 (1.00)
With the reactor at power and the "B" RWCU Pump out-of-service for
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maintenance, the reactor operator observes annunciator. 9-4-2/B-4, RUCU Pump
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"A" Low Flow, alarms. After notifying the CRS, which of the following describes.the reactor operator's required actions?
a.
Closes RWCU-MO-15, MO-18, opens RWCU-MO-74 (Demineralized Water Suction Bypass) and closes CRD-189 (RWCU Minipurge isolation).
rejects water to the Radwaste/ Main Condenser.
i b.
Closes RWCU-MO-15, MO-18, opens RWCU-MO-74 (Demineralized Water suction Bypass), and closes CRD-189 (RUCU minipurge Isolation).
c.
Opens RWCU-MO-74, opens RUCU-MO-15,-18 ; rej ects water to Radwaste/ Main Condenser; and calls chemistry to monitor Reactor
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Water Chemistry parameters.
d.
Verifies a Group 3 isolation, opens RWCU-MO-70 (Demin. Suct.
Bypass) opens CRD-189: and commences reactor shutdown, ANSWER:
055 (1.00)
C.
,
REFERENCE:
COR001-20, pg 15
[3.3/4.3]
294001Alli
.(KA's)
Comments Per AP 2.3.2.25 annunciator 9-4-2/B-4 RUCU PUMP A LOW FLOW (ATTACHED),
the automatic actions are:
trip of
"A" RUCU pump and the closing of RWCU-AO-15A EFFLUENT VALVE. (Note:
RWCU-AO-15A.and RWCU-MO-15A are two'
different valves).
Operator actions are: verify the pump tripped, and l
de te rmine the cause of system low flow. The probable causes listed for
.
,
this annunciator are:
pump or filter demineralizer valve malfunction.
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Civen the conditions stated, no actions will be required for RWCU-MO-15 ar 18, the valves are open.and would remain open.
With 15 and 18 open, i
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RWCU MO-74 does not need:to be opened.
Rejection of reactor water is.
not required.
Therefore,' answer e is also incorrect.
We request that.
this question be deleted from the exam due to the above reasons.
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