IR 05000382/1993013

Insp Rept 50-382/93-13 on 930301-05.Major Areas Inspected: Qualifications of Applicants for Operator Licenses at Facility,Including Administration of Comprehensive Written & Operating Exams

ML20035F294
Person / Time
Site:	Waterford
Issue date:	04/13/1993
From:	Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20035F290	List: IR 05000382/1993013 ML20035F289
References
50-382-93-13, NUDOCS 9304210121
Download: ML20035F294 (92)

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t APPENDIX A

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U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report:

50-382/93-13 Operating License:

NPF-38 Licensee:

Entergy Operations, Inc.

l P.O. Box B i

l Killona, Louisiana 70066 Facility Name: Waterford 3 Steam Electric Station (SES)

Inspection At: Taft, Louisiana l

l Inspection Conducted: March 1-5, 1993 l

Inspectors:

J. Keeton, Reactor Inspector, Operations Section Division of Reactor Safety l

T. McKernon, Reactor Inspector, Operations Section l

Division of Reactor Safety

Accompanying i

Personnel:

Nancy Maguire-Moffitt Pacific Northwest Laboratory (PNL)

s Ray Pugh (PNL)

Jim Nickolaus (PNL)

l Larry Sherfey (PNL)

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Approved:

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J. L. Pellet, C ef, Operations Secti#n Date

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Division of R ctor Safety j

Inspection Summary

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Areas Inspected: Announced inspection of the qualifications of applicants for operator licenses at the Waterford 3 SES facility, which included an eligibility determination and administration of comprehensive written and operating examinations. The examination team also observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant 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.

Concurrent with the written examination for the SR0 applicants, a

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requalification examination written retake was administered to one licensed operator who had failed the previous NRC administered requalification examination. The examination was administered in accordance the licensee's requalification failure remediation requirements with review and concurrence I

l 9304210121 930414 l

PDR ADOCK 05000382

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l of the NRC as stated in the package transmitted July 9, 1991, subject, " Pilot Requalification Examination at Waterford-3."

Results:

Twelve applicants for senior reactor operator licenses satisfied the

requirements of 10 CFR 55.33(a)(2) (Section 1).

The reference material provided by the training department for

examination development was acceptable (Section 1.1).

Applicants performed well on the writtea examination, with scores

ranging from a low of 85 percent to a high of 94 percent with an average j

of 90 percent overall (Section 1.2).

All applicants performed well on the operating examinations.

No generic

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performance issues were identified (Section 1.3).

Simulator fidelity appeared acceptable with the exception of two

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discrepancies that impacted operator performance by altering event mitigation strategy (Section 1.4).

Summary of Inspection Findings:

There were no findings that were assigned a tracking number identified

during the course of this inspection.

Attachments:

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Attachment 1 - Persons Contacted and Exit Meeting

Attachment 2 - Simulation Facility Report

Attachment 3 - Written Examination Keys

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f I-3-DETAILS 1 LICENSED OPERATOR APPLICANT QUALIFICATION EVALUATION (NUREG-1021)

During the inspection, the examiners evaluated the qualifications of 12

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license applicants, 8 upgrades to senior reactor operator (SRO) currently licensed as reactor operators (R0s), and 4 SR0s not currently licensed. The inspection assessed the eligibility and administrative and technical competencies of the applicants to be issued licenses to operate and direct the operation of the reactivity controls of a commercial nuclear power facility in accordance with 10 CFR 55 and NUREG-1021, " Operator License Examiner Standards," Revision 6, Sections 200 (series), 300 (series), and 400 (series).

Further, the inspection included evaluations of facility materials, i

procedures, and simulation capability used to support development and l

administration of the examinations. These areas were evaluated using the guidance provided in the areas of NUREG-1021 cited above.

• 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

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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, lesson plans, question banks, simulator scenarios, and job performance measures (JPM). The procedures and lesson plans were acceptable.

The facility bank of JPMs, questions, and scenarios were useful for examination development. Although, outdated material was identified in all three areas, there was evidence of a dynamic process of examination material review and update.

There is no regulatory requirement for a facility to develop and maintain a bank of valid test items (questions, JPMs, and scenarios) for NRC use to develop examinations. However, due to the significant savings in development time, the NRC has expressed willingness to use such material if it is i

available and meets the standards of NUREG-1021.

i 1.2 Written Examinations The chief examiner developed a comprehensive written SRO examination in accordance with the guidelines of NUREG-1021, Revision 6, Section 401. The examination consisted of 100 multiple choice questions.

During the week of February 15, 1993, members of the facility training department, under the provisions of NUREG-1021, which require execution of a non-disclosure security agreement, reviewed the examination in the Region IV office. The NRC

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-4-considers the pre-administration 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 March 1, 1993.

The chief examiner provided the facility training staff with a copy of the "as l

administered" written examination and key along with the pre-administration

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review comments on March 1,1993, immediately following the completion of the i

examination by the applicants. The facility took that opportunity to further review the written examination and no comments were made as a result of that review.

The chief examiner reviewed applicant performance on individual questions and

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observed that only six questions were missed by 50 percent or more of the

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applicants responding to the question. Those questions were numbers 2, 53, 70, 83, 97, and 99.

Refer to Attachment 3 for the complete question and answer.

The chief examiner concluded that no specific area of significant knowledge

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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.

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Overall, applicants performed well on the written examinations.

Scores ranged

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from a low of 85 percent to a high of 94 percent with an average of 90 percent

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overall.

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1.3 Operating Examinations The examiners developed comprehensive operating examinations in accordance

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with the guidelines of NUREG-1021, Revision 6, Section 301. The operating

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examinations consisted of two parts, a dynamic simulator scenario portion and

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a control room / plant walkthrough portion. The chief examiner previewed and F

validated the various portions of the operating examinations in the Region IV

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office and on site with the assistance of facility training personnel under security agreement during the week of March 1, 1993. The examination team administered the operating examinations during tFa period of March 1-5, 1993.

j 1.3.1 Dynamic Simulator Scenarios

The examination team evaluated four crews (each consisting of three SRO

applicants) on three scenarios each using the Waterford 3 SES plant-specific simulation facility.

The examiners compared applicants' actual performance

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during the scenarios with expected performance in accordance with the

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requirements of NUREG-1021, Revision 6, Section 303, to evaluate applicants'

competencies on this portion of the operating examinations.

l The examination team noted that communications among crew members was a

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strength in all four crews.

Additionally, the examinatio., team concluded that

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the crews displayed effective command and control attributes, specifically, prioritization of responses to concurrent events competing for limited resources.

The examination team observed no generic weaknesses during this portion of the operating examinations. All applicants passed this portion of the operating i

examination.

Individuals and crews generally performed very well during the scenarios.

i 1.3.2 Walkthrough Examinations The examination team evaluated each of the four instant SR0 applicants using ten JPMs relating to tasks within the scope of potential duties of a licensed SRO (which include non-licensed operator tasks outside the control room). The examination team evaluated the eight upgrade SR0 applicants on five tasks each. The applicants performed some of.the tasks in the simulation facility in the dynamic mode. They simulated (through discussions) the remainder of the tasks in the plant integrated control room and 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 task. The questions solicited "short-answer" responses and permitted the applicants to use operationally controlled references to aid in their responses unless specifically annotated to require response from memory. 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.

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Overall, the applicants performed very well. All applicants passed this

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portion of the operating examination with satisfactory performance on all j

systems and tasks.

1.4 Simulator Fidelity

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During the preparation and conduct of the operating examinations, the

examination team observed two discrepancies in simulator fidelity that

impacted operator performance by altering the event mitigation strategy during the examination. The facility staff noted the discrepancies for further investigation. Attachment 2 contains the specifics of the discrepancies.

i 1.5 Conclusions

The examination team concluded that the performance of the twelve applicants

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for senior operator licenses satisfied the requirements of 10 CFR 55.33(a)(2)

and recommended that licenses be issued.

In general, the examination team concluded that:

Individual applicants and crews performed well.

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-6-Communication and command and control were strong.

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The applicants demonstrated no generic weaknesses.

• Facility material submitted was acceptable for examination development.

• The simulator fidelity discrepancies observed during dynamic operation l

impacted the planned scenario mitigation strategy, but did not effect

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the overall evaluations.

2 LICENSED OPERATOR REQUALIFICATION EVALUATION (NUREG-1021, PILOT PROGRAM)

A requalification examination written retake was administered to one licensed operator who had failed the previous NRC administered requalification examination. This action was necessary to satisfy the requirement for renewal of the operators license. The examination was administered in accordance the licensee's requalification failure remediation requirements with review and concurrence of the NRC as stated in the package transmitted July 9, 1991, subject, " Pilot Requalification Examination at Waterford-3."

2.1 Written Examination The chief examiner reviewed the written examination prepared by the facility for the retake examination. Only minor revisions were requested by the chief examiner prior to concurring with the facility on the content of the final examination. The examination met the standards of NUREG-1021 and the approved Pilot Program.

The chief examiner also reviewed the facility grading of the written I

examination and was in complete agreement with their results. The operator passed the requalification retake examination. This satisfies the requirement for license renewal.

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ATTACHMENT 1

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1 PERSONS CONTACTED I

1.1 Licensee Personnel

• M. Ferri, Training Manager

• J. O'Hern, Operations Training Supervisor

• T. Brown, Operations Supervisor

• A.

Vest, Senior Instructor B. Litzke, Senior Instructor In addition to the personnel listed above, the examiners contacted other personnel during this inspection period.

• Denotes personnel that attended the exit meeting.

2 EXIT MEETING

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A working exit meeting was conducted on March 5, 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 2 SIMULATION FACILITY REPORT

Facility Licensee:

Entergy Operations, Inc.

Facility Docket:

50-382 Operating Tests Administered at: Waterford 3 SES Operating Tests Administered on: March 1-5, 1993 These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations.

No licensee action is required in response to these observations.

During the conduct of simulator scenario portion of the operating tests, the following items were observed:

ITEM DESCRIPTION 1.

During a simulated fuel failure event, the model actuated a series of area radiation monitors rather than the letdown radiation monitor which was the one expected. This changed the mitigation strategy of the operators who responded correctly for the existing indications.

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However, the fuel failure was never diagnosed.

2.

When containment spray was actuated during one scenario, the B train l

pump flow indicated 1500 gpm while discharge pressure indicated 0 psig.

l This caused some confusion among the operators, but the A train

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indication was satisfactory and only one train was required.

These discrepancies distracted and confused the panel operators; however, they did not significantly affect evaluations.

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ATTACHMENT 3

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WRITTEN EXAMINATION KEYS

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MRC Official Use Only

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MASTER Col'Y

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Nuclear Regulatory Commission Operator Licensing

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Examination

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This document is removed from Official Use Only category on date of examination.

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I NRC Official Use Only o -

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NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION

CANDIDATE'S NAME:

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FACILITY:

WaterFord 3 REACTOR TYPE:

PWR-CE80 DATE ADMINISTERED:

93/03/01 INSTRUCTIONS TO CANDIDATE:

l Use the answer sheets provided to document your answers.

Staple this cover l sheet on top of the answer sheets.

Points for each question are indicated in l parentheses after the question.

The passing grade requires a final grade of l at least 80%.

Examination papers will be picked up four (4) hours after the l examination starts.

CANDIDATE'S TEST VALUE SCORE

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l 100.00

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TOTALS

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FINAL GRADE l

All work done on this examination is my own.

I have neither given nor

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received aid.

Candidate's Signature

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SENIOR REACTOR OPERATOR Page

• ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 023 a

b c

d

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001 a

b c

d 024 a

b c

d 002 a

b c

d 025 a

b c

d

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003 a

b c

d 026 a

b c

d l

004 a

b c

d 027 a

b c

d 005 a

b c

d 028 a

b c

d (

006 a

b c

d 029 a

b c

d

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007 a

b c

d 030 a

b c

d l

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008 a

b c

d 031 a

b c

d

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009 a

b c

d 032 a

b c

d 010 a

b c

d 033 a

b c

d 011 a

b c

d 034 a

b c

d 012 a

b c

d 035 a

b c

d

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013 a

b c

d 036 a

b c

d i

014 a

b c

d 037 a

b c

d

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015 a

b c

d 038 a

b c

d

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016 a

b c

d 039 a

b c

d 017 a

b c

d 040 a

b c

d

018 a

b c

d 041 a

b c

d

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i 019 a

b c

d 042 a

b c

d 020 a

b c

d 043 a

b c

d 021 a

b c

d 044 a

b c

d 022 a

b c

d 045 a

b c

d

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SENIOR REACTOR OPERATOR Page

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ANSWER SHEET j

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Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

046 a

b c

.d 069 a

b c

d 047 a

b c

d 070 a

b c

d

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048 a

b c

d 071 a

b c

d

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049 a

b c

d 072 a

b c

d 050 a

b c

d 073 a

b c

d l

051 a

b c

d 074 a

b c

d

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L 052 a

b c

d 075 a

b c

d 053 a

b c

d 076 a

b c

d j

i 054 a

b c

d 077 a

b c

d j

055 a

b c

d 078 a

b c

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056 a

b c

d 079 a

b c

d 057 a

b c

d 080 a

b c

d 058 a

b c

d 081 a

b c

d i

059 a

b c

d 082 a

b c

d i

060 a

b c

d 083 a

b c

d 061 a

b c

d 084 a

b c

d

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t 062 a

b c

d 085 a

b c

d

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063 a

b c

d 086 a

b c

d j

064 a

b c

d 087 a

b c

d 065 a

b~

d 088 a

b c

d c

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066 a

b c

d 089 a

b c

d 067 a

b c

d 090 a

b c

d 068 a

b c

d 091 a

b c

d

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ANSWER SHEET

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Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

092 a

b c

d 093 a

b c

d

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'094 a

b c

d

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095 a

b c

d

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096 a

b c

d 097 a

b c

d 098 a

b c

d 099 a

b c

d

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100 a

b c

d

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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 following rules apply:

1.

Cheating on the examination mea-an automatic denial of your application j

and could result in more severe penalties.

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'2. After the examination has been completed, you must sign the statement on j

the cover sheet indicating that the work is your own and you have not l

received or given assistance in completing the examination.

This must be done after you complete the examination.

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3. Restroom trips are to be limited and only one applicant at a time may

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I leave.

You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.

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4. Use black ink or dark pencil ONLY to facilitate legible reproductions.

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Print your name in the blank provided in the upper right-hand corner of l

the examination cover sheet and each answer sheet, t

6. Mark your answers on the answer sheet provided.

USE ONLY THE PAPER PROVIDED

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AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.

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7.

Before you turn in your examination, consecutively number each answer sheet,

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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.

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Avoid using symbols such as < or > signs to avoid a simple transposition

error resulting in an incorrect answer.

Write it out.

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9. The point value for each question is indicated in parentheses after the

question.

i 10. Show all calculations, methods, or assumptions used to obtain an answer to i

any short answer questions.

l 11. Partial credit may be given except on multiple choice questions.

Therefore,

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ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.

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t 12. Proportional grading will be applied.

Any additional wrong information

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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

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points, and you give five responses, each of your responses will be worth i

0.20 points.

If one of your five responses is incorrect, 0.20 will be

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deducted and your total credit for that question will be 0.80 instead of l

1.00 even though you got the four correct answers.

i 13. If the intent of a question is unclear, ask questions of the examiner only.

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14. Whe~n turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets.

In addition, turn in all scrap paper.

15. Ensure all infonnation you wish to have evaluated as part of your answer is l

on your answer sheet.

Scrap paper will be disposed of immediately following j

the examination.

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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.

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18. When you are done and have turned in your examination, leave the examination

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area (EXAMINER WILL DEFINE THE AREA).

If you are found in this area while the examination is still in progress, your license may be denied or revoked.

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SENIOR REACTOR OPERATOR Page

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QUESTION: 001 (1.00)

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t The reason that a motor operated valve (MOV) should NOT be manually opened and backseated (unless power to the motor operator has been

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lost) is because:

i the operator may not know how many turns of the handwheel

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are required to open the valve fully.

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b.

electrical power to the motor operator may be suddenly l

returned, causing possible operator injury.

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manual operation increases the likelihood of subsequent r

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t leakage past the valve seat.

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d.

The motor operator may be unable to close the valve

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during subsequent electrical operation.

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f QUESTION: 002 (1.00)

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When performing a valve lineup verification, the position of a locked throttle valve should be verified by:

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t unlocking the valve and closing the valve fully while

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counting the number of turns required.

The valve should i

then be reopened the same number of turns and relocked.

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b.

unlocking the valve and opening the valve fully while l

counting the number of turns required.

The valve should

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then be reclosed the same number of turns and relocked.

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observing the valve stem, position indication, or flow instrumentation.

The valve locking device should not be

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d.

attempting to open and close the valve slightly La ensure movement is not possible.

The valve locking device should not be removed.

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SENIOR REACTOR OPERATOR Page

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QUESTION: 003 (1.00)

Following a design basis excess steam demand accident, the major-concern associated with RCS cooldown caused by the steam generator blowdown phase is

the major concern during the RCS refill phase is

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a.

reactor restart; excessive subcooling b.

void formation; reactor restart c.

excessive subcooling; boron plateout d.

excessive core delta-T; pressurized thermal shock QUESTION: 004 (1.00)

Operation of reactor coolant pumps with a void in the upper plenum region of the vessel will result in RVLMS indicating:

a.

a higher than actual level.

b.

a lower than actual level.

c.

excessive delta-T.

d.

excessive heater power.

QUESTION: 005 (1.00)

RAS is designed to delay initiation for at least 20 minutes after a large LOCA occurs to ensure that:

a.

the RCS will have completed blowdown.

b.

the LPSI pumps will have adequate NPSH.

c.

containment pressure will be reduced by half.

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flow from one HPSI pump will prevent core heatup.

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I QUESTION: 006 (1.00)

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CET temperatures exceeding hot leg temperatures by more than 10 degrees F is an indication of

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open thermocouples.

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thermocouples undergoing thermionic emission.

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loss of single phase natural circulation.

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c.

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d.

supersaturated conditions at core exit.

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QUESTION: 007 (1.00)

Main feedwater is completely lost for 30 minutes.

Only one motor

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driven EFW pump is operating.

OP-902-006, " Loss of Main Feedwater (

i Recovery Procedure," directs you to stop ALL reactor coolant pumps

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to reduce:

a.

RCS inventory loss.

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heat load on the EFW system.

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electrical load.

d.

possibility of lifting SG safeties.

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QUESTION: 008 (1.00)

I Following an uncomplicated reactor trip with the feedwater control

system operating properly, the startup feedwater control valves (FW-166 A/B) will be open:

a.

4%.

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b.

14%.

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24%.

d.

34%.

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SENIOR REACTOR OPERATOR Page 10

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QUESTION: 009 (1.00)

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An EFAS signal has started the AB Emergency Feedwater Pump.

The

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NAO accidently trips the overspeed trip mechanism while conducting l

a visual check of the pump.

If the NAO resets the mechanical trip l

linkage locally and the NPO reopens the turbine stop valve (MS-416)

frcm CP-8 with no other operator action, the pump will respond by:

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NOT starting because the electrical overspeed trip must

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also be reset.

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b.

starting and operating normally.

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c.

starting but will trip on actual overspeed.

l l

d.

starting and remain at idle speed.

l

!

j QUESTION: 010 (1.00)-

l What are two sources for obtaining meteorological data if the plant

,

computer is out of service?

a.

QSPDS and National Weather Service.

j t

i b.

QSPDS and Met tower recorders.

c.

SPDS and National Weather Service.

d.

Met tower recgrders and National Weather Service.

>

QUESTION: 011 (1.00)

The LOCA Recovery Procedure states, "IF RAS occurs, THEN verify Containment Spray actuated."

The basis for this step is to:-

a.

ensure mixing of the trisodium phosphate.

b.

entrain Iodine into the SI sump.

i c.

prevent Boron hideout.

I d.

provide cooling for injection water.

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SENIOR REACTOR OPERATOR Page 11 a

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l

'

,

'

t QUESTION: 012 (1.00)

l l

If all components respond properly to a Recirculation Actuation l

Signal during a Large Break LOCA, which action must be performed by j

an operator to complete the system line up?

'

a.

Open the SI sump outlet valves.

l t

l b.

Close the RWSP outlet valves.

I f

c.

Stop the LPSI pumps.

i d.

Start the HPSI pumps.

l

T QUESTION: 013 (1.00)

!

Which factor is NOT considered when Xenon-free shutdown margin

l boron concentrations are calculated:

l

.

a.

reactor fuel burnup (EFPD).

'

b.

RCS temperature (Tave).

c.

equilibrium Samarium concentration.

d.

pressurizer Boron concentration.

[

t QUESTION: 014 (1.00)

An excess steam demand and a steam generator tube rupture have occurred.

You have entered OP-902-008, " Safety Function Recovery i

Procedure," and identified Safety Functions.IV, V,

and-VI as being

!

in jeopardy.

You have implemented subprocedure IV-2, but are l

unable to satisfy the Success Path Criteria.

You should now go to:

-!

i

a.

Success Path IV-3.

i b.

Success Path IV-1.

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c.

the jeopardized Safety Function V.

f d.

OP-902-000, " Emergency Entry Procedure," Diagnostics.

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SENIOR REACTOR OPERATOR Page 12

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-

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i QUESTION: 015 (1.00)

,

While implementing OP-902-008, " Safety Function Recovery

Procedure," Safety Functions III, IV, and V were in jeopardy.

SF's l

III and IV now meet their success path criteria, but V does not.

The Vital Auxiliaries function no longer meets the Success Path I-1 criteria.

As CRS, you should now:

i

!

i a.

address Safety Function V.

b.

address the Vital Auxiliaries Safety Function.

v l

c.

attempt to upgrade all Safety Functions in jeopardy to higher priority Success Paths.

d.

return to OP-902-000, " Emergency Entry Procedure,"

,

Diagnostics.

!

!

QUESTION: 016 (1.00)

l With the plant operating steadily at full power, the primary

,

operator notes that since his last meter scan the megawatt output t

'

meter has decreased by about 25 MW, while reactor power, Tref, and t

!

Tavg have remained constant.

What is the most likely cause of the

l change in generator output?

'

a.

A load reduction on the grid.

f b.

A feed excursion causing increased SG 1evels.

c.

A leaking SG code safety valve.

d.

An air leak-in the main condenser.

,

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i SENIOR REACTOR OPERATOR Page 13 i

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.

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r

-

t QUESTION: 017 (1.00)

{

If RCS specific activity exceeds the limits of Technical i

Specification 3.4.7, why are you required to go to hot standby and

!

reduce Tavg below 500 degrees F?

j

a.

This temperature has a saturation pressure that is below

'

i the lift pressure of the main steam safeties.

b.

This temperature will contract small fuel defects and

stop the release of fission products to the RCS.

{

!

c.

Reducing RCS temperature will increase the efficiency of l

the CVCS letdown ion exchangers.

l

!

i d.

Reducing RCS temperature will increase the adherency of l

the RCS corrosion film and minimize potential crud

{

'

.

bursts.

t

i i

i QUESTION: 018 (1.00)

A small feed line break inside containment increases containment

!

temperature to 160 degrees F while RCS and containment pressure l

remain constant.

Pressurizer level indicated by LIC-110X/Y will

,

be-i

highey't$1n actual level because the reference leg fluid I

a.

density will decrease.

b.

lower than actual level because the reference leg fluid will expand and spill back into the pressurizer.

c.

higher than actual level because the elevated containment

,

temperature will cause flashing in the reference leg.

i d.

lower than actual level because of the high temperature

effects on the differential pressure detector.

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SENIOR REACTOR OPERATOR Page 14 l

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!

QUESTION: 019 (1.00)

l A manual reactor trip followed by an automatic SIAS actuation was initiated because of an uncontrollable pressure decrease.

Current

indications are as follows:

- RCS pressure 1600 psia and decreasing,

.

i

- Tavg 537 degrees F, I

- pressurizer level (LIC-110X) 23% and decreasing,

- pressurizer level (LIC-110Y) 100% and stable,

- containment pressure 15.7 psia, j

- containment sump high level alarm,

-

- quench tank level 80% and stable,

!

- quench tank temperature 100 degrees F and stable.

!

t Which failure caused the above events and indications?

h

!

a.

A pressurizer safety valve has failed open.

I

!

b.

A pressurizer spray valve has failed open.

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c.

A pressurizer reference leg has ruptured.

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d.

A pressurizer heater well has ruptured.

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SENIOR REACTOR OPERATOR Page 15 i

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'

l QUESTION: 020 (1.00)

Following a LOCA, you are performing OP-902-002, "LOCA Recovery Procedure," with the following conditions:

I

- all RCPs are OFF,

- RCS pressure 1200 psia and decreasing,

!

- highest T-hot 540 degrees F and stable,

!

- highest core thermocouple 550 degrees F and stable, i

- RVLMS 100 % (plenum and head),

i

- pressurizer level 31% and increasing, I

- #1 S/G WR level 53% and increasing,

- #2 S/G WR level 50% and increasing.

Do plant conditions meet the criteria-for throttling HPSI flow in l

accordance with OP-902-002?

j

!

a.

Yes, but only if at least one S/G has >150 gpm EFW flow.

'

b.

Yes, but only if pressurizer level continues to increase.

l

c.

No, because pressurizer level criteria is not satisfied.

i i.

I d.

No, because subcooling criteria is not satisfied.

t

?

i l

QUESTION: 021 (1.00)

l

Where are the keys located for opening the fuse drawers for the 6.9 l

KV and 4.16 KV ground metering pts?

I t

a.

In the applicable switchgear cubicle.

!

b.

On Auxiliary Operator key rings.

I c.

In the key locker in the shift supervisor's office.

d.

In the RAB +7 key control office.

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SENIOR REACTOR OPERATOR Page 16

.

.

QUESTION: 022 (1.00)

During a loss of coolant accident with SIAS actuation, pressurizer level momentarily went to 25% and has been restored to 32%.

What are the MINIMUM actions necessary to restore the A train pressurizer proportional heaters to operation?

a.

Leave proportional heater bank. 1 (PH-1) control switch in NORM.

b.

After 205 seconds, take PH-1 control switch to ON.

c.

Leave PH-1 control switch in NORM and close A32 feeder breaker after 205. seconds.

d.

After 205 seconds, close A32 feeder breaker and take PH-1 control switch to ON.

QUESTION: 023 (1.00)

You are the CRS monitoring'the safety function status checklist of OP-902-001, " Uncomplicated Reactor Trip Recovery Procedure."

If the criteria for a sa#ety function cannot be maintained, you-should:

a.

restore the safety function using the contingency l

actions.

i f

b.

return to the diagnostics of OP-902-000, " Emergency Entry

!

Procedure."

t c.

go to OP-902-008, " Safety Function Recovery. Procedure."

d.

refer to the resource assessment tree for that safety

function.

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SENIOR REACTOR OPERATOR Page 17

,

.

.

QUESTION: 024 (1.00)

'

The Control Room Supervisor can exit from OP-902-008, " Safety Function Recovery Procedure," when:

a.

all Safety Function Status Checklist Criteria are being l

met, b.

less than two Safety Functions are in jeopardy.

c.

the criteria for any success path are being met for each

!

Safety Function.

l d.

all success paths have been implemented for all Safety i

'

Functions.

l

l QUESTION: 025 (1.00)

l l

Which documentation is required to be completed before plugging a l

[

set of floor drains to monitor inleakage?

i

!

a.

An equipment out of service log entry.

!

l I

b.

A caution tag log entry.

l c.

A temporary alteration request.

l l

l d.

A nonconformance condition identification (CI).

j i

i i

l QUESTION: 026 (1.00)

l

!

i The LOWEST management level allowed to approve a deviation from the l

l plant's working hours policy for nuclear safety-related work by an j

i individual (up to 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in a 7-day week) is the:

!

a.

Operations Superintendent.

j

b.

Shift Supervisor.

I c.

Operations and Maintenance Manager.

I d.

General Manager Plant Operations.

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SENIOR REACTOR OPERATOR Page 18

.

.

QUESTION: 027 (1.00)

]

In accordance with UNT-005-002, " Condition Identification (CI),"

f the SS/CRS is responsible for approving:

l a.

all CI's initiated on the back shifts.

l i

b.

all controlled maintenance CI's.

l c.

only uncontrolled maintenance CI's.

!

d.

only nonconformance CI's.

j

l QUESTION: 028 (1.00)

l Installed quality related equipment found to have an indeterminate

!

qualification status will be documented with a:

a.

Quality notice.

l.

b.

Reject CI Disposition report.

j f

c.

Nonconformance CI.

f i

l d.

Controlled Maintenance CI.

[

l

!

l

&

I

i l

QUESTION: 029 (1.00)

j When performing the diagnostics of OP-902-000, " Emergency Entry f

Procedure," as CRS, you are unable to determine an optinal recovery

!

procedure.

You should:

I

!,

l a.

continue with the diagnostics of OP-902-000 until an

!

l optimal recovery procedure can be selected.

!

i b.

perform the immediate actions of OP-902-000 again, i

!

c.

go to OP-902-001, " Uncomplicated Reactor Trip Procedure."

d.

go to OP-902-008, " Safety Function Recovery Procedure."

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!l SENIOR REACTOR OPERATOR Page 19

.

]

l QUESTION.: 030 (1.00)

j f

During a valve lineup verification, an open valve should be l

verified open by:

!

observing the valve stem travel indicator / follower to

a.

ensure it is fully extended.

!

b.

verifying firm tightness exists when trying to move the

valve toward the open direction.

j i

c.

closing the valve slightly to ensure movement in the shut i

direction; reverse the valve movement until firm tightness is obtained.

j i

d.

closing the valve fully while counting the number of I

handwheel turns required; reverse the valve-movement the

,

same number of handwheel turns.

l

,

i QUESTION: 031 (1.00)

According to UNT-005-002, " Condition Identification (CI)," who is

!

responsible for initiating a CI for a Non-conformance?

'

a.

Individual identifying a non-conformance.

-

,

b.

Lead maintenance planner.

t c.

Operations Coordinator.

l d.

SS or CRS if designated.

.

!

,

i QUESTION: 032 (1.00)

i

!

According to UNT-005-004, " Temporary Alteration Control (TAR), " if

{

any remote or local control switches are affected by a TAR, a(an)

l will be issued to ensure operators are aware of the l

l effects.

r

a.

temporary procedure change

'

b.

work authorization J

c.

Operator Aid d.

Caution Tag l

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SENIOR REACTOR OPERATOR Page 20

.

,

QUESTION: 033 (1.00)

When performing operations in the Emergency Operating Procedures, which Control Room Panel is the primary responsibility of the

Secondary Nuclear Plant Operator - (SNPO) ?

l l

a.

CP-2 i

,

b.

CP-4

c.

CP-7

-

d.

CP-8 j

l

QUESTION: 034 (1.00)

e One responsibility of the Nuclear Plant Operator (NPO) during shift l

turnover is to:

i a.

review and sign the Auxiliary Operator Watch Station j

Turnover Checklists.

.,

b.

ensure the Operations Security Keys Turnover Sheet and Checklist is completed.

'

c.

review the status of annunciators and the Annunciator and Alarm Status Log book.

d.

ensure a shift meeting is held and that all appropriate l

shift personnel attend.

,

i QUESTION: 035 (1.00)

t According to OP-100-009, " Control Of Valves And Breakers," it is I

prohibited (NOT ALLOWED) to:

a.

lock a breaker closed.

!

b.

lock a breaker open.

!

!

l c.

lock a valve open.

d.

lock a valve closed.

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SENIOR REACTOR OPERATOR Page 21

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l i

QUESTION: 036 (1.00)

!

'

!

According to 01-002-000, " Annunciator, Alanm, and Control Room l

Instrumentation Status Control," which action is required if an I

annunciator on CP-18 must be disabled?

f f

a.

Make an Annunciator Status Log entry.

]

b.

A Temporary Alteration (TA) must be generated.

l I

c.

A Radiant Yellow dot must be placed on the annunciator j

window.

l i

d.

A Condition Identification (CI) must be placed on the

!

annunciator window.

j

!

-:

QUESTION: 037 (1.00)

f

,

During an emergency, the state and local emergency organizations are normally notified via the-l I

a.

Emergency Notification System.

b.

Operational Hot Line.

!

i i

c.

Industrial Hot Line.

i d.

LP&L Emergency Dial.

i i

i

!

t QUESTION: 038 (1.00)

l

!

l According to HP-001-110, " Radiation Work Permits," RWPs for i

repetitive tasks may be approved for:

a.

the duration of the task only.

I b.

one calendar year.

!

c.

one calendar quarter.

d.

one month.

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SENIOR REACTOR OPERATOR Page 22 l

.

,

i i

'

,

QUESTION: 039 (1.00)

i Which RCS chemistry specification is a Technical Specification

!

limit ONLY when Tavg is greater than 250 degrees F?

!

a.

Dissolved oxygen.

b.

Chloride.

c.

Fluoride.

d.

Specific Activity.

.

f

QUESTION: 040 (1.00)

The CEDMCS " Timer Failure" alarm has annunciated.

What happens if the " Hold Bus HV" pushbutton is pressed?

!

i a.

Voltage is applied to the upper and lower gripper coils i

of the-subgroup placed on the hold' bus.

!

.

b.

Voltage is increased momentarily to the lower gripper coils.of the subgroup placed on the hold bus.

c.

Voltage is increased momentarily to the upper gripper

.;

coils of the subgroup placed on the hold bus.

d.

Voltage is increased and stays that'way until the CEAs in i

the subgroup are disconnected from the main bus.

i

,

QUESTION: 041 (1.00)

r During an approach to criticality, the NPO starts pulling the

'

,

regulating group of CEAs in the manual sequential mode.

When should criticality be anticipated?

!

a.

When CEAs reach the ECC height.

j

b.

At 5x10E-4% on the Log Safety channel, j

!

At 5-7 doublings on the Countrate Doubling Data Sheet.

c.

l

'

d.

At any time during CEA withdrawal.

l

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I l

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._

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.

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.. _. _ _

.

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.

i SENIOR REACTOR OPERATOR Page 23

.

.

QUESTION: 042 (1.00)

i While at 100% power operation, an Auxiliary Operator inadvertently opens breakers TCB-1, TCB-3, and TCB-7.

This action will directly cause:

a.

all CEAs to insert into the core, j

i b.

the regulating group of CEAs to insert into the core.

'

c.

half of the CEAs to insert into the core.

[

$

d.

three of four CEDMCS Bus undervoltage relays to actuate.

l l

QUESTION: 043 (1.00)

.

The RCP speed sensor devices are used to provide an input for:

a.

backup overspeed protection.

!

b.

the COLSS calculation of RCS flow.

l

'

c.

the compensation for temperature shadowing.

d.

input to the shape annealing factor.

QUESTION: 044 (1.00)

With RCS pressure at 2250 psia and the following seal pressure i

indications:

,

!

vapor seal = 100 psia,

'

upper seal = 1125 psia, middle seal = 1125 psia, which seal has failed?

a.

Lower seal.

i b.

Middle seal.

c.

Upper seal.

d.

Vapor sea t-

-

i

,

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-

SENIOR REACTOR OPERATOR Page 24

[

i f

E

i

!

QUESTION: 045 (1.00)

i Which systers must be manipulated to secure the A chiller and place

!

I-the A/B Essential chiller in service?

!

!

a.

CHW, PMU, CMU l

b.

PMU, CCW, CHW

{

i

'

c.

CCW, ACCW, CHW l

d.

CHW, CMU, CCW

!

t

!

I.

!

QUESTION: 046 (1.00)

!

Following a SIAS the essential chilled water system will:

a.

continue to run in the same configuration as it was in

,

prior to the receipt of.the SIAS.-

b.

split the A loop from the B loop and isolate the non-safety loop.

c.

split the A loop from the B loop, isolate the non-safety

!

loop, then start the A/B chiller to supply the non-safety loop.

r d.

split the A loop from the B loop, isolate the non-safety-loop, then start the A/B chiller as a backup for the

'

safety loop.

f QUESTION: 047 (1.00)

Outside air is completely isolated from the control room envelope l

during a:

.

,

l i

a.

containment isolation actuation.

b.

safety injection actuation.

c.

toxic gas actuation.

d.

high radiation actuation.

i

!

.

,

a

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..

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,

SENIOR REACTOR OPERATOR Page 25 i

i

i QUESTION: 048 (1.00)

l

Which signals are provided by the reed switch position transmitter I

(RSPT) to control CEA withdrawal or insertion?

!

t i

a.

Upper sequential permissive and lower sequential j

i permissive.

,

!

b.

Upper control limit and lower control limit.

l

!

i c.

Upper group stop and lower group stop.

d.

Upper electrical limit and lower electrical limit.

l

?

l

!

<

QUESTION: 049 (1.00)

!

!

The reason for maintaining a specified minimum. level in the l

containment spray header riser is to:

l a.

minimize pressure surges in the spray header.

f

b.

provide a water seal for the spray valves.

l

!

c.

provide back pressure to seat the check valve.

,

i d.

minimize the time for spray to reach containment.

,

,

QUESTION: 050 (1.00)

If a CSAS has occurred, which sequence of manipulations is required i

to close containment spray header isolation valve CS-125A?

-

!

a.

The CSAS must be reset, the valve control switch taken to

!

OPEN, and then to CLOSE.

i b.

The CSAS must be reset, the valve will then stroke closed

'

automatically.

c.

The valve control switch must be taken to OPEN and then to CLOSE.

d.

The CS pump must be tripped and then the control switch taken to CLOSE.

,

i l

.

SENIOR REACTOR OPERATOR Page.26

.

t QUESTION: 051 (1.00)

The letdown stop valve, CVC-101, automatically closes on an actuation signal from either:

a.

SIAS or CIAS.

b.

SIAS or high letdown temperature.

c.

SIAS or RAS.

d.

CIAS or high letdown temperature.

QUESTION: 052 (1.00)

Valves that receive an actuation signal from SIAS include:

a.

letdown inside containment isolation valve, CVC 103, letdown outside containment isolation valve, CVC 109, and letdown stop valve, CVC 101.

b.

letdown inside containment isolation valve, CVC 103, RWSP suction to charging pumps, CVC 507, and letdown stop valve, CVC 101.

c.

VCT discharge valve, CVC 183, letdown outside containment isolation valve, CVC 109, and letdown stop valve, CVC 101.

d.

letdown inside containment isolation valve, CVC 103, VCT makeup isolation valve, CVC.510, and VCT discharge valve, CVC 183.

.

-

.

-

,

l

-

SENIOR REACTOR OPERATOR Page 27 j

i

-

!

,

,

QUESTION: 053 (1.00)

,

I If a fresh mixed bed demineralizer in the CVCS is placed in service

]

prior to the resin saturation treatment, how would the RCS f'

chemistry respond?

a.

The pH will decrease and boron concentration will

!

decrease.

{

b.

The pH will decrease and boron concentration will increase.

!

!

c.

The pH will increase and boron concentration will

,

decrease.

!

!

d.

The pH will increase and boron concentration will_

l increase.

l t

I

,

QUESTION: 054 (1.00)

l The auxiliary feed water system was designed to:

f a.

supply feed water to the steam generators during'all l

normal and accident conditions.

,

b.

supplement the main feed water system at high steam l

generator pressures.

,

c.

ensure adequate core cooling-in the event of a high

pressure LOCA (small line break).

,

d.

deliver water to the steam generators when the

-

condensate / feed water system is not available.

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SENIOR REACTOR OPERATOR Page 28

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QUESTION: 055 (1.00)

'

Which additional interlocks must be satisfied to start the

auxiliary feed water pump with recirc valve, AFW-113, OPEN?

i i

a.

Pressure control valve, AFW-124, OPEN, and suction pressure > 20 ft water.

b.

Pressure control valve, AFW-124, CLOSED, and discharge valve, AFW-144, CLOSED.

c.

Pressure control valve, AFW-124, CLOSED, i

and suction pressure > 20 ft water.

j a.

Pressure control valve, AFW-124, OPEN,

'

and discharge valve, AFW-144, OPEN.

QUESTION: 056 (1.00)

{

i The reactor is at 50% power and the feed water control system

,

(FWCS) is in automatic.

If the feed flow signal to the FWCS fails low, the flow error developed will cause:

i i

a.

an increase in actual feed flow, offset by level error,

{

resulting in level stabilizing on-program.

!

b.

an increase in actual feed flow, offset by level error,

[

resulting in level-stabilizing above program.

l

.

'

c.

a decrease in actual feed flow, offset by level error, resulting in level stabilizing below program.

!

d.

a decrease in actual feed flow, offset by level error, resulting in level stabilizing on program.

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SENIOR REACTOR OPERATOR Page 29 l

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QUESTION: 057 (1.00)

_

Following an emergency feed water actuation signal with NO operator action, the final steam generator level will automatically be

!

controlled in the range of

,

a.

68-71% narrow range.

i

b.

68-71% wide range.

I c.

55-71% narrow range.

l

d.

55-71% wide range.

.

i QUESTION: 058 (1.00)

'

s Given the following plant status:

'

s

'

- RCS pressure 1700 psia

- DRTS enabled

,

- DEFAS enabled i

- EFW valve control in automatic

- SG 1 pressure 505 psia, decreasing

- SG 2 pressure 707 psia, decreasing

-

- SG 1 level 28% wide range, dropping

- SG 2 level 39.5% wide range, dropping

,

These parameters indicate the occurrence of:

a.

DEFAS and DRTS automatic actuation..

l

b.

DEFAS and EFAS automatic actuation.

I c.

EFAS automatic actuation with greater flow to SG 1.

d.

EFAS automatic actuation.

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SENIOR REACTOR OPERATOR Page 30

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QUESTION: 059 (1.00)

The quarterly surveillance on all station batteries was. completed as follows during the past year:

- today

- 109 days ago

'

- 228 days ago

- 353 days ago

,

During this time, how many total days were the batteries inoperable?

a.

O b.

,

c.

e d.

!

r

QUESTION: 060 (1.00)

While performing a natural circulation cooldown following a SGTR, l

a sudden change in pressurizer level inconsistent with the cooldown

,

'

rate has occurred.

The most probable cause of this indication is:

,

a.

voiding in the reactor vessel head area, b.

automatic operation-of the charging pumps.

c.

contraction of the RCS during the cooldown.

d.

pressurizer level instrument reference leg flashing.

,

,

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SENIOR REACTOR OPERATOR Page 31

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t QUESTION: 061 (1.00)

J A SIAS affects the containment cooling system by causing the:

a.

containment fan cooler dampers to close.

j b.

shield building ventilation inlet valve to close.

c.

containment fan coolers to start and shift to fast speed.

'

d.

containment fan coolers CCW temperature control valves to fully open.

!

!

!

QUESTION: 062 (1.00)

{

!

l The motor operators for the ESF pumps' suction isolation valves

,

from the SI sump, SI-602A and B, are powered from:

i

'

a.

MCCs 311A-S and 311B-S i

b.

MCCs 312A-S and 312B-S

'

l l

c.

MCCs 313A-S and 313B-S l

,

d.

SUPS A and SUPS B i

!

,

QUESTION: 063 (1.00)

The function of the interlocks associated with the RWSP outlet

valves (SI-106) and the SIS sump outlet valves (SI-602) is as follows:

,

l a.

SI-106A and B receive an open signal on SIAS; and SI 602 l

A and B receive a closed signal on SIAS.

b.

SI-106A and B receive a closed signal on SIAS; and SI 602

l A and B receive an open signal on RAS.

!

c.

SI-106 A and B receive a closed signal on RAS; and SI 602 A and B receive an open signal on RAS.

d.

SI-106 A and B receive an open signal on SIAS; and SI 602 A and B receive a closed signal on RAS.

]

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SENIOR REACTOR OPERATOR Page 32

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QUESTION: 064 (1.00)

!

Which requirement is intended to ensure inventory control when shut down with the RCS level below 18 ft.?

One LPSI pump must be operable and in operation.

a.

b.

Two HPSI pumps must be aligned and available.

i t-All charging pumps must be aligned and available.

c.

d.

One LPSI pump must be operable and available.

)

i r

"

QUESTION: 065 (1.00)

i The reactor was at 70% power when the charging pump tripped.

None

}

of the charging pumps can be started.

At which pressurizer level

!

are you required to manually trip the reactor?

l

!

t a.

48%

t t

b.

35%

l c.

30%

g

d.

25%

I C

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QUESTION: 066 (1.00)

Which parameter is used to derive the pressurizer' level setpoint t

program in the RRS mode?

i

!

a.

Tave, j

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b.

Tref.

l c.

First stage turbine pressure.

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d.

Ex-core power.

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l SENIOR REACTOR OPERATOR Page 33 l

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QUESTION: 067 (1.00)

i The steam generator blowdown system parameter used continuously by

)

COLSS to calculate BSCAL is a default constant of blowdown:

l r

a.

flow inserted by the operator.

b.

pressure inserted by the operator.

t

-

l flow inserted by the operator only if the active point c.

!

goes bad.

d.

pressure inserted by the operator only if the active point goes bad.

'

QUESTION: 068 (1.00)

!

!

The large, rapid changes in the excore nuclear instrumentation

!

count rate during the Three Mile Island accident were caused by:

i

!

a.

containment pressure spikes.

j b.

changes in down comer level, c.

void formation in the reactor head.

d.

voiding in the reactor core.

i i

i QUESTION: 069 (1.00)

l l

Technical Specifications requires that at least 23 feet of water t

must be maintained above the top of irradiated fuel assemblies

!

seated in the spent fuel storage racks to ensure that:

i a.

almost all iodine gap activity released from a ruptured

'

spent fuel assembly will be removed by the water.

>

b.

adequate flow through the skimmer exists to remove the

'

,

fission products released by spent fuel.

• c.

the radiation level at the surface of the pool will not

exceed 10 mrem /hr.

d.

a spent fuel assembly raised to the upper limit on the refuel hoist will not expose the bridge operator to more than 10 mrem /hr.

!

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SENIOR REACTOR OPERATOR Page_34 f

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I

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. QUESTION: 070 (1.00)

After completing a partial stroke test of SG1 MSIV,_the NAO was i

directed to open the A dump valve outlet solenoid valve to i

depressurize the exercise accumulator.

What feature ensures that i

the dump valve inlet has fully closed before the accumulators vent

!

to the hydraulic reservoir?

i a.

The local key switch is spring returned to the normal

position which deenergizes the inlet solenoid.

j b.

A 10 second timer in the test circuit ~ automatically

!

deenergizes the inlet solenoid.

,

"

c.

The MSIV's 90% open limit switch deenergizes the inlet solenoid.

{

!

d.

A 65 second time delay relay-prevents energizing the

'

outlet solenoid to allow time for the inlet valve to

,

close.

I

!

!

!

l QUESTION: 071 (1.00)

l

Which condition requires a cold start of the MSRs?

[

l a.

LP turbine metal temperature > 300 degrees F.

i b.

MSR A tube bundle vented to atmosphere.

c.-

Turbine load at 10% for > 10 minutes.

!

!

d.

Scavenging steam lined up to condenser.

,

,

!

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SENIOR REACTOR OPERATOR Page 35_

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l QUESTION: 072 (1.00)

{

!

If a main steam isolation occurs (all MSIVs close) the steam driven

emergency feed water pump is:

a.

unaffected, because its steam supply is upstream of the j

MSIVs.

'

i b.

not used, because the electric pumps are the only i

available source.

i i

i c.

partially isolated from its steam supply and may need realignment.

.

!

d.

isolated from its steam supply, because its use is

!

undesirable during steam line break.

l

!

!

QUESTION: 073 (1.00)

i

!

During normal power operation, if the 3B safety bus is lost, you

j are required to

,

e

!

a.

immediately restore off site power to the bus.

b.

verify A side components are operab'le.

I i

c.

trip the reactor and trip RCPs 1B and 2B.

f i

f j

d.

commence an immediate shutdown to hot standby.

!

!

l I

QUESTION: 074 (1.00)

!

Which system (s) is(are) likely to be the first to threaten continued power operation if it is lost and can't'be restored?

i i

a.

Control room ventilation.

j i

b.

Containment fan coolers.

l

!

c.

Battery exhaust fans.

i d.

CEDM fans.

I i

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,

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SENIOR REACTOR OPERATOR Page 36 l

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l

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t QUESTION: 075 (1.00)

!

Which radiation monitor (s) provide (s) only a high. radiation alarm I

with NO automatic actuations?

l a.

Chemical and volume control.

I t

b.

Turbine building industrial waste sump.

(

c.

Fuel handling building area monitors, d.

Circulating water discharge.

f

t t

QUESTION: 076 (1.00)

i Which signal initiates an automatic CEA withdrawal prohibit (AWP)?

!

a.

High log power trip, b.

High startup rate.

l c.

Low pressurizer pressure pretrip.

l l

.I l

d.

SBDS demand signal.

!

l l

l r

QUESTION: 077 (1.00)

>

,

When aligning a shutdown cooling train for operation, the LPSI pump miniflow recirculating valve is closed.

How is recirculating flow

provided?

]

a.

A recirculating flow path is not necessary for this l

condition.

!

b.

The shutdown cooling train flow control valve, SI-129A(B), is opened.

c.

The shutdown cooling train warmup valve, SI-135A(B), is opened.

d.

The shutdown cooling train temperature control valve, SI-415A(B), is opened.

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SENIOR REACTOR OPERATOR Page 37

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(

i

I QUESTION: 078 (1.00)

f The maximum number of flow paths allowed while draining.down the RCS is:

.

a.

one.

i b.

two.

t c.

three.

!

i d.

four.

r

!

,

QUESTION: 079 (1.00)

f When SIAS actuates, CCW will be lost to:

a.

boric acid concentrator B.

'

b.

the CEDM coolers.

f c.

shutdown cooling heat exchanger B.

l t

d.

the RCP seal coolers.

[

t

!

!

!

!

QUESTION: 080 (1.00)

j i

If instrument air pressure is decreasing, a reactor trip is

required when pressure gets to:

i a.

60 psig.

!

b.

65 psig.

!

c.

70 psig.

t i

d.

75 psig.

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SENIOR REACTOR OPERATOR Page 38 l

f QUESTION: 081 (1.00)

Which automatic actions, besides the selected subgroups dropping, occur when the reactor power cutback circuitry actuates?

a.

Bypass valves open, turbine setback, and turbine runback.

b.

Bypass valves open, turbine setback, and turbine power i

increase inhibit.

c.

Turbine runback, turbine setback, and turbine power

'

increase inhibit.

,

i d.

Turbine setback, turbine runback, and automatic motion l

inhibit.

}

,

a QUESTION: 082 (1.00)

I The minimum number of CEAs not fully inserted following a reactor trip that requires the operator to emergency borate is:.

I a.

one.

b.

two.

l c.

three.

,

d.

four.

l

QUESTION: 083 (1.00)

,

e Before shutdown cooling can be initiated following a LOCA,

-

pressurizer pressure must be greater than 205 psia to:

a.

ensure sufficient makeup capacity is maintained to the

RCS.

b.

ensure adequate NPSH to the LPSI pump.

c.

prevent flashing in the SDC heat exchanger.

l

'

d.

prevent CCW from diluting the RCS if the SDC heat exchanger leaks.

.

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SENIOR REACTOR OPERATOR Page 39

>

QUESTION: 084 (1.00)

i

,

If emergency boration is required, in which situation must you open f

the gravity feed valves and close the VCT discharge valve?

l f

a.

Charging pumps fail to start.

b.

BAM tank level is less than 15%.

c.

BAM pumps recircs are closed.

!

I d.

BAM pumps fail to start.

l

.

l QUESTION: 085 (1.00)

i

!

If an ATWS occurs, a contingency action directs the operator to open the feeder breakers to 3A32 and 3B32 for 5 seconds then

,

!

reclose them.

Why should they be reclosed?

!

l

a.

To restore power to the CEDMCS to verify that all CEAs l

'

l have inserted fully.

i l

b.

To reenergize the 32 busses before the undervoltage i

j relays time out and strip the individual loads from the j

I busses.

i c.

To restore power to the proportional pressurizer heaters, l

only.

j i

d.

To restore power to all pressurizer heaters.

[

i l

QUESTION: 086 (1.00)

f

'

!

How is a diverse reactor trip (DRTS) designed to interrupt 240 VAC

to the CEDM coils?

l t

l a.

By energizing the shunt trip coils of the reactor trip

circuit breakers.

b.

By deenergizing the undervoltage trip coils of the

,

reactor trip circuit breakers.

!

c.

By opening the MG sets' output load contactors.

'

d.

By opening the MG sets' feeder breakers on the 32A and 32B busses.

i i

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SENIOR REACTOR OPERATOR Page 40'

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QUESTION: 087 (1.00)

A station blackout has occurred that resulted in a reactor trip and

i EXTENDED loss of all offsite and onsite power; RCS Tave can be-r controlled.

!

!

a.

automatically by the steam bypass valves.

b.

automatically by the atmospheric dump valves.

j c.

manually by the atmospheric dump valves.

[

d.

manually by the steam bypass valves.

l t

!

QUESTION: 088 (1.00)

i I

The emergency' diesel generators started on loss of all offsite i

power.

The transformer for MCC-315A developed a short to ground

~

when energized by the sequencer 1 second load block (S1 relay),

[

causing undervoltage on 3A3.

Which actions are caused by the-

!

undervoltage override or the sequencer lockout?

?

a.

The sequencer immediately stops the auto loading due to i

l the undervoltage override.

!

l i

b.

The sequencer immediately stops the auto' loading due to

.

the sequencer lockout.

{

c.

The sequencer stops the auto loading.when-the 7 second load block (S3 relay) is reached.

'

d.

The sequencer stops the auto loading when the'17 second load block (S4 relay) is reached.

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SENIOR REACTOR OPERATOR Page 41

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l

,

QUESTION: 089 (1.00)

For which radiation monitor will the stated action occur when a high radiation alarm is received.

a.

Industrial waste sump; sump pumps stop, j

b.

Dry cooling tower sump; discharge valve closes.

c.

Fuel handling building (FHB) WRGM; FHB emergency exhaust fans start.

d.

Plant stack PIG; containment purge system isolates.

!

!

l

!

l QUESTION: 090 (1.00)

.

i

<

i

?

l If the switchgear HVAC system is placed in purge it will cause the:

l l

I a.

switchgear area AHU (AH-25) OAI dampers to open.

b.

switchgear area AHU (AH-25) recirc dampers to open.

f c.

switchgear area smoke exhaust fan (E-48) to start.

l

<

~

d.

penetration area smoke exhaust fan (E-50) to start.

,

i

!

i QUESTION: 091 (1.00)

!

!

An action that CANNOT be performed at LCP-43 is:

a.

manual control of pressurizer level.

r t

,

b.

manual control of atmospheric dump valves.

!

,

c.

operation of all emergency feed water pumps.

j d.

operation of all main steam isolation valves.

,

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SENIOR REACTOR OPERATOR Page 42

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QUESTION: 092 (1.00)

l

A loss of all offsite power occurred about an hour ago that

!

resulted in a reactor trip.

Offsite power has not been recovered.

l An indication that inadequate core cooling exists is:

l a.

a core delta-T of 65 degrees F.

I b.

subcooling 30 degrees F.

i c.

T-hot and T-cold temperatures decreasing.

I d.

CETs 545 degrees F with T-hot 539 degrees F.

i QUESTION: 093 (1.00)

Indications of an air bound LPSI pump will be differ from i

indications of cavitation by:

a.

amps fluctuating slowly.

t b.

amps indicating low and steady.

l c.

flow fluctuating slowly.

d.

flow indicating high and steady.

i i

QUESTION: 094 (1.00)

,

During a loss of shutdown cooling that resulted because of LPSI

[

pump cavitation, the primary reason for initiating hot and cold leg

!

injection to the SDC train to be started is to a.

prevent boron precipitation.

+

l b.

ensure that all injection does not go out a cold leg j

break.

l c

!

t c.

prevent over pressurization of cold leg nozzle dams.

J d.

provide cool water to the suction of the LPSI pump.

l i

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I SENIOR REACTOR OPERATOR Page 43.

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QUESTION: 095 (1.00)

l l

The purpose of providing emergency power to the pressurizer heaters

!

is to:

i a.

prevent losing the function of the pressurizer safeties..

!

!

b.

insure pressure control if a spray valve fails open.

c.

establish and maintain natural circulation.

l

!

.

provide pressure control for restart of RCPs upon power d.

f i

restoration.

!

QUESTION: 096 (1.00)

The battery chargers for the train A, train B, and train A/B will shutdown on high voltage:

,

a.

instantaneously if charger output voltage exceeds 144

[

VDC.

i

,

b.

when charger output voltage exceeds 144 VDC for more than

{

'

10 seconds.

[

t I

c.

instantaneously if charger output voltage falls below battery voltage.

!

d.

when battery voltage exceeds charger output for more than i

I 10 seconds.

t

!

QUESTION: 097 (1.00)

!-

If instrument air to the turbine closed cooling water (TCCW)

components is isolated, how is minimum pump flow maintained?

,

a.

All TCCW component temperature control valves fail full l

open.

b.

The TCCW system temperature control valv ails to a preset position.

c.

The TCCW system pressure control valve fails full open.

d.

The TCCW system pressure control valve fails to a preset position.

.

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SENIOR REACTOR OPERATOR Page 44

!

.

,

t QUESTION: 098 (1.00)

i The standby instrument air compressor will:

l

immediately start automatically upon SIAS and loss of j

a.

offsite power.

,.

I b.

start automatically on loss of offsite power after the i

I EDG loads are sequenced on.

i start automatically after loss of offsite power and SIAS c

when a low receiver pressure is reached.

,

!

!

'

d.

require manual start on loss of offsite power and SIAS.

.

.

QUESTION: 099 (1.00)

'

Which process instrument qualifies as accident monitoring l

I instrumentation on QSPDS?

.

I a.

Subcool margin.

b.

EFW flow.

I

!

l c.

Pressurizer level.

I d.

Containment pressure.

'

l QUESTION: 100 (1.00)

,

l During a station blackout you are required to place the containment (

spray pumps to the OFF position.

The purpose of this step is to

'

'

prevent:

a.

overloading the emergency diesel generator, b.

actuating spray when power is regained.

c.

running the pumps deadheaded.

d.

loss of control power to the pumps.

(**********

END OF EXAMINATION **********)

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SENIOR REACTOR OPERATOR Page 45

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!

i ANSWER:

001 (1.00)

!

!

d-

!

.

l REFERENCE:

j

!

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SENIOR REACTOR OPERATOR Page 67

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REFERENCE:

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ZCED-000-00

[2.9/3.1]

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ANSWER KEY

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027 b

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069 a

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071 b

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076 d

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077 c

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i 092 a

i 093 b

l 094 d

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096 b

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098 d

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100 b

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(********** END OF EXAMINATION **********)

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SRO Exam PWR React-or Organized by Ouest ion Number

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QUESTION VALUE REFERENCE

i 001 1.00 20345

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002 1.00 21149

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003 1.00 22426

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004 1.00 22427 l

005 1.00 22428 i

006 1.00 22429

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007 1.00 22430

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008 1.00 22431

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009 1.00 22432

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011 1.00 22439 i

012 1.00 22440 i

013-1.00 22442 t

014 1.00 22443

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015 1.00 22444 l

016 1.00 22445

017 1.00 22446 c

018 1.00 22447 i

019 1.00 22448 l

020 1.00 22449 021 1.00 22456

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022 1.00 22457

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i 023 1.00 22458 024 1.00 22834 i

025 1.00 22836

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026 1.00 22837-j 027 1.00 22840 j

028 1.00 22841

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029 1.00 22842 l

030 1.00 22844 i

031 1.00 27883 (

032 1.00 27885 l

033 1.00 27887 034 1.00 27888

035 1.00 27889

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036 1.00 27890

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l 037 1.00 27891 038 1.00 27894 j

039 1.00 27895

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040 1.00 9000001 i

041

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042 1.00 9000003 j

043 1.00 9000004 i

044 1.00 9000005

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045 1.00 9000006 046 1.00 9000007 047 1.00 9000008 i

048 1.00 9000009

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Orqanized by Ouest ion Number l

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QUESTION VALUE REFERENCE

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050 1.00-9000011 l

051 1.00 9000012

052 1.00 9000013 053 1.00 9000014

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054 1.00 9000015 t

055 1.00 9000016 l

056 1.00 9000017

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057 1.00 9000018 058 1.00 9000019 059 1.00 9000020 060 1.00 9000021 a

061 1.00 9000022

062 1.00 9000023 i

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063 1.00 9000024 064 1.00 9000025 065 1.00 9000026

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066 1.00 9000027 067 1.00 9000028

068 1.00 9000029 069 1.00 9000030 i

070 1.00 900003'.

071 1.00 9000032 072 1.00 9000033 l

073 1.00 9000034

074 1.00 9000035

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075 1.00 9000036

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076 1.00 9000037 077 1.00 9000038 078 1.00 9000039 079 1.00 9000040 080 1.00 9000041 081 1.00 9000062 082 1.00 9000063

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083 1.00 9000064

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084 1.00 9000065

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085 1.00 9000066

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086 1.00 9000067

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087 1.00 9000068 088 1.00 9000069

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089 1.00 9000070 090 1.00 9000071 091 1.00 9000072 092 1.00 9000073 093 1.00 9000074 094 1.00 9000075 095 1.00 9000076 096 1.00 9000077 097 1.00 9000078 098 1.00 9000079

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TEST CROSS REFERENCE Page-3

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I SRO Exam PWR React or

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Organized by Quest ion Number j

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QUESTION VALUE REFERENCE

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SRO Exam PWR React or

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Organized bv KA Group PLANT WIDE GENERICS QUESTION VALUE T4

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024 1.00 194001A102 029 1.00 194001A102

027 1.00 194001A103 026 1.00 194001A103 l

036 1.00 194001A106 l

028 1.00 194001A106

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033 1.00 194001A109 034 1.00 194001A111

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031 1.00 194001A112

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032 1.00 194001A113 039 1.00 194001A114 037 1.00 194001A116 030 1.00 194001K101 i

002 1.00 194001K101

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025 1.00 194001K102 038 1.00 194001K103 035 1.00 194001K107 001 1.00 194001K107

......

PNG Total 18.00 PLANT SYSTEMS

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Group I QUESTION VALUE KA i

041 1.00 001000K518 042 1.00 001000K603 040 1.00 001050A201 043 1.00 003000A304 044 1.00 003000K103 051 1.00 004000A302 052 1.00 004010A205 l

053 1.00 004020A213 l

047 1.00 013000A302 046 1.00 013000A302 j

045 1.00 013000G007 048 1.00 014000G007 l

068 1.00 015020A202 050 1.00 026000A401 049 1.00 026000G001 056 1.00 059000A211 055 1.00 059000G007 054 1.00 059000K102 057 1.00 061000A303 I

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TEST CROSS REFERENCE Page

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SRO Exam PWR React or

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Oroani2 ed by KA Group

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Group 1 l

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QUESTION VALUE KA

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058 1.00 061000K402 l

059 1.00 063000G011 i

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PS-I Total 21.00 Group II t

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060 1.00 002000K514 064 1.00 006000G013 l

061 1.00 006000K102

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062 1.00 006000K204 i

063 1.00 006000K409 i

066 1.00 011000A104 065 1.00 011000K101 067 1.00 012000K608 069 1.00 033000A203

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070 1.00 035010K601

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071 1.00 039000G004

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072-1.00 039000K107

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074 1.00 062000A201 073 1.00 062000A204 075 1.00 073000K101

......

PS-II Total 15.00 Group III QUESTION VALUE KA 078 1.00 005000K301

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077 1.00 005000K411 I

079 1.00 008000A202 076 1.00 041020K401 080 1.00 078000G008

......

PS-III Total 5.00

......

......

PS Total 41.00 EMERGENCY PLANT EVOLUTIONS I

.

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e SRO Exam PWR Reactor

,

I Organi zed by KA Group i

EMERGENCY PLANT EVOLUTIONS l

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Group I

QUESTION VALUE KA 081 1.00 000003K303 082 1.00 000005A203 004 1.00 000011A117 083 1.00 000011A201 012 1.00 000011G012 011 1.00 000011K312 i

005 1.00 000011K315 i

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013 1.00 000024A205 084 1.00 000024K302 086 1.00 000029K206

085 1.00 000029K312 003 1.00 000040K106 i

018 1.00 000040K106 l

016 1.00 000051G011 006 1.00 000055A101 021 1.00 000055A107

088 1.00 000055A206

087 1.00 000055K302 089 1.00 000059A205 l

090 1.00 000067A105

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091 1.00 000068A112

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092 1.00 000074K311 017 1.00 000076K305

......

EPE-I Total 23.00

i Group II l

QUESTION VALUE KA

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I 008 1.00 000007A102 l

023 1.00 000007G012 019 1.00 000008A227

i 020 1.00 000009A234 093 1.00 000025A207 094 1.00 000025G012 022 1.00 000027A101

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095 1.00 000027A104 015 1.00 000038G012

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014 1.00 000038G012

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009 1.00 000054A102 007 1.00 000054K304

096 1.00 00005BA202 l

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010 1.00 000060K104 098 1.00 000065G009

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TEST CROSS REFERENCE Page

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SRO Exam PWR React or l

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Organized by KA Group

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EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA l

097 1.00 000065K307

..____

EPE-II Total 16.00 l

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Group III

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i QUESTION VALUE KA

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t 099 1.00 000028A107

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100 1.00 000056K302

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EPE-III Total 2.00 l

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EPE Total 41.00 i

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Test Total 100.00

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For Operction j

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'OP-901-014 Revision 5

Attachment 6.3 (1 of 1)

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OP-902-008 Revision 7

,

Page II of II

.

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Foldout:

Safety Function Status Checklist (Cont'd)

Safety Function Success Path I-I I-2 I-3 I. Reactivity Control g

II-I II-2 II-3 2. Vital Auxiliaries O

III-I III-2 III-3 3. RCS Inventory and Pressure Control IV-I IV-2 IV-3 IV-4 IV-5 4. RCS and Core Heat Removal O

V-I 5. Containmnent Isolation O

VI-I VI-2 VI-3 6. Containment Temperature and Pressure Control O

VII-1 VII-2 7. Containment Combustible Gas Control O

O l

END

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