Text

Exam Rept 50-321/93-300 on 930218-25.Exam Results:Seven Senior Reactor Operators Passed Initial Exam & One Senior Reactor Operator Passed Requalification Retake Exam.No Violations or Deviations Noted
	ML20056C167
Person / Time
Site:	Hatch
Issue date:	03/19/1993
From:	Bartley J, Lawyer L; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20056C163	List: IR 05000321/1993300; ML20056C162;
References
	50-321-93-300, NUDOCS 9303300175
	Download: ML20056C167 (114)
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.,o(c#"%g UWTED STATES l

NUCLEAR REGULATORY COMMISslON j

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101 MARIETTA STREET.N.W.

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ATt.ANTA GEORGIA 30323

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ENCLOSURE _1 y

' Report No.: 50-321/93-300

Licensee: Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 j

Docket Nos.: 50-321 and 50-366

License Nos.: DPR-57 and NPF-5

Facility Name:

E. I. Hatch Nuclear Plant

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Inspection Conducted:

February 18 - 25, 1993

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rn Chief Examiner:

A7Jfrodkw ~

3/8/93 Jona 3 H. Bart1 Q Date signed Examiners:

M. Daniels, Sonalysts, Inc.

l G. Harris, NRC

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Approved By:

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Lawrence L. Lawyer,' Chief Date Signed OperatorLicensingSectid(n/

Operations Branch

Division of Reactor Safety

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SUMMARY q

Scope:

Regular, announced initial examinations were conducted during the weeks of February 15 and 22,1993. The examinations were

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administered under the guidelines of the Examiner Standards (ES),

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NUREG-1021, Revision 6.

Written and operating examinations were adndnistered to seven Senior Reactor Operator (SRO) applicants. A requalification retake examination was administered on February-17, 1993, to one SRO who failed the requalification examinations given in September 199?.

Results: Seven of seven SR0s passed the initial examination for a 100 percent f

pass rate. One of one SR0 passed the requalification retake

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examination.

A' strength was noted in the area of crew communications (paragraph.

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2. e)..

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No violations or deviations were identified.

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9303300175 930322 PDR ADOCK 05000321

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REPORT DETAILS

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Persons Contacted Licensee Employees R. Belcher, Instructor, Operations Training i

J. Betsill, Operations Unit Superintendent

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S. Crosby, Operations Training Supervisor (Classroom)

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0. Fraser, Site Supervisor, SAER

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G. Goode, Manager, Engineering Support

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S. Grantham, Supervisor, Operations Training

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M. Gunn, Plant Instructor, Operations Training

J. Lewis, Operations Manager

D.

Read, AGM-PS

L. Sumner, General Manager

S. Tipps, Manager, Nuclear Safety and Compliance Other licensee employees contacted included instructors, operators, engineers, and office personnel.

NRC Personnel

E. Christnot, Resident Inspector i

L. Lawyer, Section Chief, Region II

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T. McKernon, Examiner, Region IV i

Attended exit interview 2.

Discussion a.

Results

Seven SR0 candidates passed the initial examination. One SRO passed

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a requalification retake examination.

b.

Reference Material

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The licensee submitted reference material as required by NUREG-1021.

The materials were well organized and easy to use. The training staff promptly informed the examiners when updates were made to procedures during examination development. This helped alleviate i

last minute changes during the preparation and examination weeks.

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Examination Development Representatives from the licensee training staff reviewed the written examination in the regional office on January 25 - 27, 1993.

The review was thorough and provided valuable input to improve the accuracy of the information contained in the written examination.

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Report Details

The facility provided the NRC with a proposed walkthrough l

examination for the requalification retake examination. The l

examiners reviewed the proposed examination for content and

structure per the guidance of NUREG-1021. The examiners determined the examination to be adequate and administered it without any I

modifications. Additionally, the examiners reviewed the operator's l

remedial program and determined it to be adequate.

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d.

Examination Administration The examinations were administered without major problems. The simulator primary computer failed twice during the evaluation of one i

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crew which caused a 1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months delay in the administration of the

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simulator examinations. Licensee support in running the simulator

and providing other assistance was good. Cooperation received from

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shift personnel was excellent.

The licensee provided an evaluator to assist in the administration j

of the requalification retake examination. The evaluator did a good l

job and no problems were encountered.

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Candidate Performance

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l Candidate performance during the examinations was excellent.

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Written examination scores ranged from 90 percent to 97 percent, no i

Job Performance Measures (JPMs) were failed, and overall performance during the simulator examinations was excellent.

During the simulator portion of the examination, the candidates'

communication skills were excellent. No instances of missed i

i communications or mis-communications occurred. The SR0s gave crew briefings consistently and asked the crew for feedback at the end of

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the briefings. Candidates in the R0 positions ensured the SR0 l

understood and acknowledged their reports.

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Simulator Facility

The simulator worked well with the exception of downtime caused by I

the simulator computer, a failed switch, and the inability to

override automatic actuation signals.

(1) The simulator primary computer failed twice during the I

evaluation of one crew.

This caused a 1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months delay in the i

I administration of the simulator examinations. The second l

failure of the primary computer required switching to the backup computer which accounted for the majority of the

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downtime.

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(2) The switch for the CS/RHR Diagonal Pump Room Cooler, 2T41-B002B, failed mechanically during the examinations and

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was replaced.

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Report Details

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(3) The simulator did not have the capability to override automatic actuation signals while still allowing manual actuation. This prevented the examiners from evaluating the candidates'

response to plausible events where systems fail to actuate automatically and require manual operation. The simulator

operator could override the system or component switch to

off/close and remove the override to allow manual actuation.

However, it was difficult for the simulator operator to monitor the candidate's actions so he knew when to remove the override to allow manual actuation. As aa example, the examiners wrote i

a JPM for a manual initiation of Core Spray with a failure of

the minimum flow valva to shut. To accomplish this, the simulator operator overrode the minimum flow valve switch to i

open. The simulator operator removed the override when the

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candidate positioned the minimum flow valve switch to close.

On back panel systems, such as the Standby Gas Treatment System, this could not be done at all because the simulator

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operator could not monitor the candidate's actions.

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Procedures

The examiners noted two procedural problems during the course of the examination.

(1) 31E0-E0P-106-2S Revision 1, " Restoration of RPV Water Level Following RPV Flooding," step 3.9 stated:

"0 PEN Torus Spray or Test Viv 2 Ell-F025A(B) by placing jumpers

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as follows:"

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This left the control switch in the close position which resulted in the valve repeatedly cycling full stroke. The facility training staff submitted a procedure change request to correct this.

t (2) 34AR-601-406-2S Revision 4, "0/G AVG ANNUAL REL LIMIT WILL BE j

EXCEEDED," step 5.2, directed the operator to check radiation levels on monitor 2011-K601 or 2D11-K602 at panel 2Hil-P600.

These monitors, 2D11-K601 and 2011-K602, are NUMAC radiation

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monitors which are located on panel 2Hil-P604.

The monitors on 2Hil-P600, 2011-R601 and 2Dll-R602, are pen recorders that monitor the same points as 2Dll-K601 and 2Dil-K602. This

discrepancy caused confusion and resulted in the delay of l

reporting radiation levels to the SRO.

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h.

Plant Material Condition

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On Tuesday of the exam week, while administering JPMs, an examiner noted numerous lights for the LPRM four rod display in the back of panel P603 were out. On Unit 1, there were 8 of 16 lights out, and

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Report Details

on Unit 2, there were 3 of 16 lights out. Thursday, the examiner determined the light bulbs provide the light for the illuminated pointer on the 16 LPRM modules and the meter is not functional if the light is out. The examiner went to the Control Room to

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follow-up with Control Room personnel and noted the lights had not i

been replaced. The Control Board Operators (CB0) did not know the

lights were out until questioned. When asked if it was a problem, the CB0s stated they would only use them at power if there was a problem with power oscillations. The examiner informed the Shift Operations Supervisor who took prompt action to replace the bulbs.

The label plate for the Containment Spray Viv Control switch did not include the alphanumeric designation, 2E11-S17A(B). This was inconsistent with the rest of the labeling in the Control Room.

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Training Program Review f

The examiners reviewed shift manning practices and compared them to both Technical Specification requirements and requalification

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training crew size. The maximum Technical Specifications manning levels requires two SR0s, three R0s and a Shift Technical Advisor

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(STA) to be on shift. The requalification crews on the simulator are composed of two SR0s, three R0s and an STA for one unit. The facility assigns four SR0s, seven R0s and an STA to each shift to cover both units. The facility makes it a policy to maintain a

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minimum of four SR0s, five R0s, and an STA on shift. Although this

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policy was not documented or committed to by facility management, a review of the shift manning records for the previous two months showed that the facility always exceeded these minimum manning

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levels on shift.

The examiners determined the facility shift manning significantly exceeds the levels required by Technical Specifications. The examiners also determined, based on shift manning records, the crew size used during requalification training could be manned on both units by shift personnel during a dual unit casualty.

The examiners also reviewed simulator usage and availability for the period January 1992 to December 1992. The examiners determined the simulator availability was approximately 99 percent. During the time available, the simulator was used for operator training related functions approximately 32 percent of the time and was' secured for i

maintenance approximately 7 percent of the time.

The facility did not maintain records to document simulator use of non-operator training functions. However, based on discussions with the Training Department staff, the examiners determined outside organizations were not affecting the simulator availability for the Training Department.

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No violations or deviations were identified.

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Report Details

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3.

Action on Previous Inspection Findings

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(Closed) IFI 92-301-01, review of facility implementation of procedures i

and operator training to assure that potential level indication errors will not result in improper operator actions. This item concerned the facility implementation of training and procedures in response to issues identified in NRC Generic Letter 92-04, " Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F)." During this examination, the examiners reviewed the licensee's response to Generic Letter 92-04, incorporation of the concerns into initial license and requalification training, and procedure guidance for level instrumentation problems.

Examiners found the issues of the Generic Letter were incorporated into lesson plans and the operators had adequate procedural guidance.

Examiners considered the licensee's corrective action to be adequate and this Inspector Follow-up

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Item is closed.

4.

Exit Interview At the conclusion of the site visit, the examiners met with representatives of the plant staff listed in paragraph 1 to discuss the J

results of the examinations.

The licensee did not identify as

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proprietary any material provided to, or reviewed by the examiners.

l Dissenting comments were not received from the licensee.

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ENCLOSURE 2

i SIMULATOR FACILITY REPORT

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t Facility Licensee: Georgia Power Company f

Facility Docket Nos.: 50-321 and 50-366

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Operating Tests Administered On: February 22 - 25, 1993 This form is to be used only to report observations. These observations, in

and of themselves, do not constitute audit or inspection findings and are not, without further verification and review, indicative of non-compliance with 10

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CFR 55.45(b). These observations do not affect NRC certification or approval

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of the simulation facility other than to provide information which may be used j

in future evaluations. No licensee action is required solely in response to these observations.

Item Comment l

Unable to override The simulator does not have the capability to l

automatic actuation override automatic actuation signals while still allowing manual actuation.

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Simulator Primary Primary computer for the simulator failed twice.

computer failed The second failure was a hard failure that required shifting to the backup computer.

Pump Room Cooler The switch for the CS/RHR Diagonal Pump Room Cooler, switch failed 2T41-B002B, failed mechanically during the examinations and was replaced.

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NRC Official _Use Only 7 J 3 CD lll3 - / 9/7.5 Nuclear Regulatory Commission Operator Licensing Examination i

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

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NRC Official Use Only

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U. S. NUCLEAR REGULATORY COMMISSION

SITE SPECIFIC EXAMINATION l

SENIOR OPERATOR LICENSE

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REGION

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CANDIDATE'S NAME:

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FACILITY:

E.

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Hatch 1 & 2 j

REACTOR TYPE:

BWR-GE4'

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DATE ADMINISTERED: -93/02/19 i

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INSTRUCTIONS TO CANDIDATE:

i Use the answer sheets provided to document your answers. ' Staple this cover.

j sheet on top of the answer sheets.

Points for each question are indicated in-

.i parentheses after the question.

The passing grade requires a final grade of

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at least 80%.

Examination papers wil] be picked up four (4) hours after the-i examination starts.

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CANDIDATE'S I

TEST VALUE SCORE

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I 100.00

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TOTALS-i FINAL GRADE

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All work'done on this examination is my own.

I have neither given nor received aid.

Candidate's Signature

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O SENIOR REACTOR OPERATOR Page

~2 ANSWER SHEET Multiple Choice (Circle or X your choice)

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

MULTIPLE CHOICE 018 a

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MULTIPLE CHOICE 030 a

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d 015 a

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d 038 a

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d 017 a

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ANSWER SHEET

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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041 a

b c-d 064 a.

b c

d

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i 042 a

b c

d 065 a

b c~

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043 a

b c

d 066 a

b c

d 044 a

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d 067 MATCHING 045 a

b c

d a

046 a

b c

d b

l 047 a

b c

d c

l 048 a

b c

d d

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049 a

b c

d MULTIPLE CHOICE t

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d 068 a

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051 a

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d 069 a

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052 a

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d 07v a

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d 071 MATCHING l

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d a

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055 a

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d b

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b c

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057 a

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d MULTIPLE CHOICE 059 a

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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.

077 a

b c

d 094 MATCHING 078 a

b c

d a

079 a

b c

d b

080 a

b c

d c

081 a

b c

d d

082 a

b c

d MULTIPLE CHOICE 083 a

b c

d 095 a

b c

d 084 a

b c

d 085 a

b c

d 086 a

b c

d 087 a

b c

d 088 a

b c

d 089 a

b c

d 090 a

b c

d 091 MATCHING a

b c

d MULTIPLE CHOICE 092 a

b c

d 093 a

b c

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

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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:

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Cheating on the exan.* nation means an automatic denial of your application and could result in mure severe penalties.

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l 2. After the examination has been completed.

'cn2 must sign the statement on

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the cover sheet indicating that the wo-

.e your own and you have not

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received or given assistance in complet.ng the examination.

This must be

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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 l

leave.

You must avoid all contacts with anyone outside the examination j

room to avoid even the appearance or possibility of cheating.

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

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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

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

i 7. Before you turn in your examination, consecutively number each answer sheet,.

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I 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 c 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.

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

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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.

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

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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 g

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0.20 points.

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

deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.

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

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i 14. When turning in your examination, assemble the completed examination with

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examination questions, examination aids and answer sheets.

In addition,

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turn in all scrap paper.

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en your answer sheet.

Scrap paper will be disposed of immediately following

the examination.

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I 16. To pass the examination, you must achieve a grade of 80% or greater.

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i 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 l

l area (EXAMINER WILL DEFINE THE AREA).

If you are found in this area while j

the examination is still in progress, your license may be denied or revoked.

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QUESTION: 001 (1.00)

A high radiation of 20 mr/hr in the Unit 1 Reactor Building Exhaust Ventilation has resulted in initiation of the Standby Gas Treatment Systems.

Unit 1 and Unit 2 Standby Gas Treatment System auto start.

Which ONE of the following describes the suction flow path for the two units?

a.

Unit 1 SBGT takes suction from only the Unit 1 Refueling floor and Unit 2 SBGT takes suction from the Unit 2 Refueling Floor and the Reactor Building.

b. Unit 1 SBGT takes suction from only the Unit 1 Reactor Building and Unit 2 SBGT takes suction from only the Unit 2 Reactor Building.

c.

Unit 1 SBGT takes suction from the Unit 1 Reactor Building and Refueling Floor and Unit 2 SBGT takes suction from only the Unit 2 Refueling Floor, d.

Unit 1 SBGT takes suction from the Unit 1 Reactor Building and Refueling Floor and Unit 2 SBGT takes suction from only the Unit 2 Reactor Building.

QUESTION: 002 (1.00)

DCR 91-171 was generated to install pushbuttons on Panels 1H11-P654 and P657 to reset automatic ESF actuation.

Which ONE of the following describes ALL the additional initiating signals which can be reset by these pushbuttons?

a.

Unit 2 Reactor Building and Refueling Floor Ventilation.

b.

Unit 2 Reactor Building and Units 1 and 2 Refueling Floor Ventilation, c.

Unit 1 Reactor Building and Refueling Floor Ventilation.

d.

Unit 1 Reactor Building and Units 1 and 2 Refueling Floor Ventilatio.-

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QUESTION: 003 (1.00)

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Unit 2 is operating at rated power.

Hydrogen Recombiner

"A" has been j

tagged out of service for the last 10 days due to a failed heater j

section.

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Maintenance has just determined that the RHR water supply valve to

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Hydrogen Recombiner

"B" is in the closed position and can NOT be j

operated.

I Which ONE of the following is the required action?

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a. Restore at least one of the Hydrogen Recombiners to operable j

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status within 30 days or be in Hot Shutdown within the next 12 l

hours.

I b. Restore both of the Hydrogen Recombiners to operable status

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within 20 days or be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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Be in Cold Shutdown within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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d. Be in Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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QUESTION: 004 (1.00)

Unit 2 is operating at 78% power.- The outer Primary Containment Air Lock Door is found to be partially open and cannot be closed.

The

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inside Air Lock Door is shut and locked.

l Which ONE of the following is the required action to be taken?

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a. Be in Hot Shutdown in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

j b. Repair the door in the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in Hot Shutdown in

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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c.

Operation may continue until the next overall airlock leakage j

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test.

d. Repair the door within 30 days or be in Hot Shutdown in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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QUESTION: 005 (1.00)

Unit 2 is operating at 60% power when an event occurs resulting in the i

following plant conditions:

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"A" Feedwater Pump flow 16%

"B" Feedwater Pump flow 28%

RPV Water Level 30 inches i

Which ONE of the following describes the response of the Reactor Recirculation Pumps?

a.

Pumps run back to 22%, and reset automatically when "A" feedwater flow increases above 20%.

b.

Pumps run back to 44 %, and reset automatically when

"A" feedwater flow increases above 20%.

c.

Pumps run back to 22%, but must be manually reset when "A"

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feedwater flow increases above 20% and RPV level increases

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above 32 inches.

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d.

Pumps run back to 44%, but must be manually reset when "A" feedwater flow increases above 20% and RPV level increases above 32 inches.

QUESTION: 006 (1.00)

Which ONE of the following describes the speed changes which are enforced by the Unit 2 recirculation speed controller?

a. Manual speed decreases using the master controller are limited to a MAXIMUM of 2.5% per second.

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b. Manual speed increases using the master controller are limited to 2.5% per second.

c. Manual speed increases using the individual M/A stations are limited to 1.0% per second.

)

d. Manual speed decreases with the individual M/A stations are limited to 2.5% per secon.,_.

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QUESTION: 007 (2.00)

j i

With regard to the RHR System, MATCH each of the valves in Column A with l

its associated setpoint OR interlock from Column B.

(Items in Column B j

may be used once, more than once, or not at all.

Only one answer may j

'

occupy a space in Column A)

(4 required at 0.5 each)

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COLUIG A COLUMN B

!

(Valves)

(Setpoints/ Interlocks)

a.

RHR HX bypass valve 1.

105 psig i

2.

12.5 inches b.

LPCI Outboard Injection Valve 3.

2/3 core coverage l

(F017) auto open (Unit 1)

permissive l

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4.

138 psig l

c. LPCI Inboard Injection Valve 5.

Open for 3 minutes I

(F015) auto open (unit 2)

following a LOCA signal.

f 6.

325.psig

.

i d. Shutdown Cooling Valves close 7.

425 psig with LOCA j

signal present.

l Interlocked open for

.,

duration of LOCA i

.

B.

449 psig with LOCA

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signal Present.

l Interlocked open for

.

10 minutes.

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a QUESTION: 008 (1.00)

l

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A Technical Specification Quarterly Surveillance for a Unit 2 LPCI system was due July 1st.

Because of an LCO on the system, the

,

surveillance was not completed until July 15th.

{

Which ONE of the following is the date when the next quarterly i

surveillance becomes due?

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a.

October 1 b. October 16

i

f c. October 23 r

d. November 8

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SENIOR REACTOR OPERATOR Page 11 QUESTION: 009 (1.00)

Unit 2 RCIC is being used for level control following a level transient which resulted in a reactor scram.

RCIC turbine speed increases to 5150 RPM.

Which ONE of the following describes the response of the RCIC turbine to the increased speed?

a. Continues to operate with increased flow.

b. Isolates but can be realigned when turbine speed decreases below 4950 RPM.

c. Trips on mechanical overspeed and will have to be reset locally, d. Trips on electrical overspeed, but not mechanical overspeed.

QUESTION: 010 (1.00)

Unit 2 RCIC is injecting to control reactor vessel level.

A large oil leak develops on the in-service RCIC oil filter which results in decreasing oil pressure.

Which ONE of the following describes the response of the RCIC system as the oil pressure decreases?

a. The Governor Valve will close and turbine speed will decrease _to-zero RPM.

b. The Governor Valve will open and turbine speed will increase possibly resulting in a turbine trip.

c. The Auxiliary Oil Pump will start, the Trip Valve will close and turbine speed will decrease to zero RPM.

d. The turbine will trip but the steam stop valve (F045) will fail open on low oil pressure.

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SENIOR REACTOR OPERATOR Page 12

~ QUESTION: 011 (1.00)

The operating' procedure for the High Pressure Coolant Injection (HPCI)

system cautions against prolonged operation with turbine speed less than 2000 RPM.

Which one of the following is the reason for caution?

a. The potential exists for exhaust check valve _ chatter and reduced oil supply to the turbine governor'and bearings.

b.

The Booster Pump may not provide adequate net positive suction head to prevent cavitation of the Main Pump.

c. The rate of steam flow through the HPCI turbine may not be enough to prevent it from overheating.

d. Cooling water flow from the Booster Pump to the lube oil system heat exchanger is inadequate.

QUESTION: 012 (1.00)

During an ATWS condition on Unit 2, the operator places the SBLC control switch to the

"A" pump operate position and observes that pump "A" starts but neither Squib valve fires. He then places the switch in the

"B" pump operate position and the Squib valves fire.

Which one of the following describes the condition of the Standby Liquid Control system after placing the switch to the

"B" Pump' position?

a. Both pumps running with 82 gpm injecting into the RPV.

b.

"B" Pump running with 41 gpm injecting into the RPV c.

"A" Pump running with 41 gpm injecting into the RPV d. Both pumps tripped with

"0" gpm flow rate indicate O O

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SENIOR REACTOR OPERATOR Page 13 i

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QUESTION: 013 (1.00)

Which ONE of the following statements describes the method used for re-establishing the required SRM minimum count rate during Unit 2 core reload?

a. Spiral reload the core until 3 cps is established.

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b.

Increase the voltage of the SRM high voltage power supply to

!

obtain the required count rate.

i c.

Prior to fuel reload, install neutron sources in the i

source tubes to establish 3 cps.

j i

d. Load up to four fuel bundles into core positions next to each j

SRM until 3 cps is established.

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QUESTION: 014 (1.00)

i A transient at Unit 2 has occurred, resulting in the following plant j

conditions-

'

I Reactor Water level-110 inches

[

Drywell Pressure 1.52 psig I

HPCI Equipment Room Temperature 150 deg F

.j

.

RCIC Tbrbine Exhaust Diaphragm Pressure 5 psig-j RWCU Equipment Room Temperature 125 deg F

!

Which ONE of the following groups of isolations should have occurred?

l

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a.

1, 3,

and 5

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b.

1, 5,

and 6 I

c.

2, 3,

and 4 j

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d.

2, 3,

and 5

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I l

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i QUESTION: 015 (1.00)

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A LOCA has occurred with a simultaneous loss of_125VDC bus 2A, j

"

Which ONE of the following describes the response _of the ADS system j

logic to these conditions?

'j

,

a t

a Both

"A" and

"B" logics will swap to alternate power.

.l

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b.

"A" logic will remain deenergized and the

"B" logic will swap _to i

alternate power.

!

I c.

"B" logic will remain deenergized and the

"A" logic will swap to f

alternate power.

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d. No effect ADS logic is not supplied by 125VDC bus 2A f

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QUESTION: 016 (1.00)

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A LOCA has occurred on Unit 2.

The following plant conditions occurred.

i at the times indicated:

l RPV level 3 (12.3 inches)

at 2048

~

RPV level 1 (-101 inches)

at 2103 Drywell pressure constant at 1.1 psig i

i Core Spray Pump Discharge pressure 163 psig at 2101

'

Which one of the following is the earliest ADS will initiate?

(Assume

,

RPV level and Drywell pressures do not change)

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a. 2050 i

b. 2101

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c. 2105 l

d. 2118 f

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&W SENIOR REACTOR OPERATOR Page 15 QUESTION: 017 (1.00)

Unit 2 is operating at 23% RTP.

The following plant conditions exist.

Drywell pressure 0.6 psig

,

Unit 2 Drywell venting is in progress via valves 2T48-F319 and

'

2T48-F320 During this time the Unit 1 Refuel Floor Vent Exhaust Radiation Monitors, 1D11-D611 A through D reach their isolation setpoints.

Which ONE of the following describes the effect on the Unit 2 Drywell venting?

a. Venting would continue with Unit 2 SBGT taking suction on the Unit 2 Drywell only.

b. Venting would continue with Unit 2 SBGT taking suction on the Unit 2 Drywell and Refuel Floor.

c. Venting would stop due to Unit 2 Drywell Vent and Purge valves

,

'

F319 and F320 closing.

d. Venting would stop due to the Unit 2 filter crains suction dampers realigning to take suction on the Refuel Floor.

.

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l QUESTION: 018 (1.00)

i s

l Unit 2 has scrammed from 100% power due to a Main Turbine trip.

The following plant conditions exist:

All rods in l

The-following valves have amber and green indicators illuminated:

j F013B, F013C, F013D, F013F and F013G i

The Red indicator for F013B is lit l

Plant pressure is 860 psig and slowly decreasing l

Which ONE of the following describes the status of the LLS system and i

the SRVs?

f l

a. LLS is NOT armed, F013B, F013C, F013D, F013F and F013G lifted

and are now closed.

j

b. LLS is NOT armed, F013C, F013D, F013F and F013G lifted and have

{

a closed, F013B lifted and should have reclosed.

!

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c. LLS is armed, F013C, F013D, F013W and F013G lifted, and are'now

,

'

closed and F013B is open.

J.

d..LLS is armed, F013B, F013C, F013D, F013F and F013G are open and

,

j wil.'. close at 851 psig.

.j i

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j i

QUESTION: 019 (1.00)

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Which ONE of the following is a reason for nornally operating with the l

l Reactor Water Level Selector switch selected to "B"?

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j a. If

"A" were selected a failure of the

"B" reference leg would l

result in a turbine trip on high level.

!

i b. With

"B" selected and a rupture of

"A" reference leg, reactor i

water level control would prevent level from decreasing, j

i c.

If

"A" were selected and the

"B" reference leg fails low, l

reactor level would increase until terminated by the operator.

i I

d d. With

"B" selected if

"A" reference leg fails low, the reactor j

.

will scram on low level.

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SENIOR REACTOR OPERATOR Page 17 QUESTION: 020 (1.00)

Given:

Unit 2 in control of D/G

"B" D/G

"B" Mode Switch in TEST (Surveillance being performed)

Electrical distribution NORMAL (Full Power Lineup)

D/G

"B" is at rated speed and voltage, but not synchronized, when power to 4160 volt Bus 2F from SUT 2D is lost.

Which ONE of the following describes the system operation due to the power loss?

a.

Bus 2F can be powered by D/G

"B" when the operator takes the Output Breaker Switch to CLOSE with the SYNC SCOPE activated.

b.

Bus 2F will be powered by D/G

"B" automatically, after 12 seconds; appropriate loads will be picked up sequentially.

c.

Bus 2F can not be powered by D/G

"B" while it is in the TEST mode, given these conditions.

d.

Bus 2F can be powered by D/G

"B" after the operator resets the Lockout Relay, activates the SYNC SCOPE, and takes the Output Breaker to CLOSF.

QUESTION: 021 (1.00)

i Unit 2 is operating at rated power when an event occurs which results in I

the following plant conditions:

CRD Accumulator Low Press or High Level annunciator CRD Hyd High Temp annunciator Charging water header pressure indicating

"0" psig CRD system flow indicating

"0" gpm Khich ONE of the following is the event which caused the conditions?

.

a.

Cooling water pressure control valve failed closed.

,

l b. Operating CRD Pump tripped.

I

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c. CRD flow control valve failed closed.

d.

Flow stabilizing valves failed closed.

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QUESTION: 022 (1.00)

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f

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Which ONE of the following parameters is used by the Rod Worth Minimizer i

'

to determine the point at which the RWM stops enforcing rod blocks while INCREASING power?

a.

APRM indicated power or First Stage Turbine Pressure l

i b.

Feedwater Flow or Steam Flow (

t c.

First Stage Turbine Pressure and APRM indicated power f

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l d.

Feedvater Flow and Steam Flow l

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l OUESTION: 023 (1.00)

j

Which CNE of the following is the. reason that continued plant operation with an inoperable (or failed) jet pump is restricted?

f a.

Invalid APRM Flow Biased SCRAM setpoints due to the change in

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flow through a failed jet pump

?

l b.

Increased blowdown area during a Loss of Coolant Accident

!

l c.

Unbalanced neutron flux across the core due to flow variations f

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I d.

Physical core and fuel cladding damage from a loose piece of I

l the damaged jet pump

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SENIOR REACTOR OPERATOR Page 19 QUESTION: 024 (1.00)

Unit 2 is operating at 800 power when an event results in the following conditions:

Reactor water level-12 inches RWCU Room ambient temperature 128 deg F RWCU Nonregenerative Heat Exchanger Outlet Temp.

145 deg F Which ONE of the following describes the response of the RWCU system?

a. The operating RWCU Pump trips with no isolation.

b.

Inboard Isolation valve F001 closes.

c. Outboard isolation valve F004 closes d. Both inboard and outboard isolation valves close.

QUESTION: 025 (1.00)

Unit 1 is in operation at 80% RTP. Unit 2 is in refueling and preparations are being made to remove the Unit 2 Drywell head.

Which ONE of the following defines the conditions which must be maintained while removing the Unit 2 Drywell head?

a. At least one train of Unit 2 Standby Gas Treatment must be in operation and one train of Unit 1 SBGT must be operable.

b. The Unit 2 Drywell must be maintained " air tight" until the Drywell Bulkhead Manways are closed and sealed.

c. The Unit 2 Drywell bulkhead manways must be closed and sealed until the Unit 2 Drywell equipment hatch is. opened.

d. If the Unit 2 bulkhead manways are_open, the Drywell hatch may_

only be opened if the Unit 2 SBGT system is operatin.. _.

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SENIOR REACTOR OPERATOR Page 20-

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i

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i QUESTION: 026 (1.00)

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Which ONE of the following sets of plant conditions will require the

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operator to manually initiate the Main Steam Isolation Valve Leakage

!

Control System (MSIV-LCS) with a suspected fuel element failure and a

!

LOCA signal present?

j

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a.

MSIV leakage is suspected, Main Steam Line High High Radiation-l

'

is present, Reactor pressure is less than 31.5 psig and Main

!

Steam Line pressure is less than 35 psig.

l t

d.

10 minutes have elapsed since the Loss of Coolant Accident l

(;OCA), Reactor Pressure is less than 31.5 psig and, Main Steam i

L Line Pressure has bled down to 0 psig.

!

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c.

MSIV leakage is suspected, 10 minutes have elapsed since the

!

Loss of Coolant Accident (LOCA), and both Main Steam line j

pressure and reactor pressure are less than 31.5 psig.

j q

b.

No steam line High or High High Radiation, Reactor pressure has

}

been less than 31.5 psig for a minimum of 10 minutes and, Main

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j Steam Line pressure is less than 35 psig.

)

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j QUESTION: 027 (1.00)

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l t

Which ONE of the following sets of parameters will cause a main turbine

)

i trip following a turbine runback due to a loss of stator water cooling?

j (ASSUME a 18095 amp load prior to failure of stator water cooling)

f

r i

TIME SINCE LOAD i

FAILURE l

i a.

4.8 minutes 5290 amps j

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b.

3.2 minutes 4675 amps j

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c.

2.6 minutes 15230 amps j

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d.

2.3 minutes 16340 amps i

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SENIOR REACTOR OPERATOR Page 21-

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QUESTION: 028 (1.00)

{

i Unit 2 is operating at 90% RTP when the following alarns are received:

}

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"APRM Upscale"

{

" Heater Trouble" i

"4th Stage HTR B007A Level High" i

Which ONE of the following is the IMMEDIATE action which must be taken?

.

l (Assume no reactor scram).

!

i a.

Reduce thermal power to 72% by decreasing recirculation flow.

!

I b. Reduce thermal power to 80% by decreasing recirculation flow.

{

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c. Reduce thermal power to 45% with recirc flow or until the heater

!

trouble alarm clears.

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d. Manually scram the reactor if the APRM Upscale alarm has not l

cleared within two (2) minutes.

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QUESTION: 029 (1.00)

I f

p Which ONE of the following sets of plant parameters would result'in a j

,

trip of the

"B" Reactor Feed Pump Turbine on Unit 2?

(Assume all time

delays have timed out.)

'

a. Reactor Water Level

+9 inches t

)

Bearing Oil Pressure 3 psig i

Pump Suction press 175 psig

-;

Condenser Vacuum 24' inches Hg j

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l b. Reactor Water Level

+38 inches l

Bearing Oil Pressure 5 psig Pump Suction press'

220 psig Condenser Vacuum 27 inches Hg l

!

c. Reactor Water Level

'47 inches

+

Bearing Oil Pressure 5 psig I

Pump Suction press 230 psig

,

Condenser Vacuum 23 inches Hg

!-

d. Reactor. Water Level-3 inches Bearing Oil Pressure 6 psig i

Pump Suction press 235 psig cl Condenser Vacuum-28 inches Hg l

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QUESTION: 030 (1.00)

Unit 2 is in HOT SHUTDOWN with a reactor pressure of 805 psig and MSIVs

,

are closed. Operability Surveillances are performed on all of the MSL Radiation Monitoring System Channels 2D11-K603A through D.

Channels A and D test UNSAT, while Channels B and C test SAT.

Maintenance has no estinate of repair time and will not be able to commence troubleshooting and repairs for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

Which ONE of the following actions describe the allowances and/or limitations imposed by the Technicr.1 Specifications in this instance?

a. No action required; function not required to be OPERABLE.

b.

Place one Trip System in the tripped condition within one hour; no additional action is required.

'

c.

De in Cold Shutdown within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d.

Place one Trip System in the tripped condition within one hour; be in Cold Shutdown within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

QUESTION: 031 (1.00)

From the drawing provided as Figure 1,

which ONE of the following sets of contacts are closed under NORMAL conditions for the Vital AC Static Invertor?

a.

B1, B2, and B3 b.

B2 c. B2 and B3 d.

B3 and B4

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QUESTION: 032 (1.00)

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i A complete loss of Unit 1 Service Air has occurred.

l 1-

'Which one of the following describes the effect this loss will have on

[

the Puel Pool Transfer Canal inflatable seals?

c

,.

"

Pneumatic pressure will be immediately lost to'only the inner

'

a.

gate seals.

i

b. Pneumatic pressure will be immediately lost to only the outer gate seals.

.

c.

Pressure to neither seal will be immediately lost due to the air receivers that are available to supply air pressure to the

seals.

-

I d.

Pressure to the seals will not be lost due to a backup nitrogen

-,

a bottle automatically supplying pressure.

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QUESTION: 033 (1.00)

Which ONE of the following REOUIRED actions should be taken if feedwater

conductivity cannot be maintained less than 0.2 umhos/cm?

i

a.

Operation may continue for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while efforts are

being made to return the conductivity to normal.

,

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b.

Perform a normal shutdown within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if conductivity cannot

'

be restored to less than 0.1 umhos/cm.

i i

c. Scram the Reactor and terminate feedwater injection if HPCI or

,

RCIC are available.

,

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d. Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and COLD SHUTDOWN l

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within the next 24' hours.

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i SENIOR REACTOR OPERATOR Page 24'

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l QUESTION: 034 (1.00)

l

The Control Rod Reed Switches provide input to the Rod Position f

Indicating System for use in the rod control systems.

!

Which ONE of the following DOES NOT receive input from the Reed l

Switches?

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a. Rod Worth Minimizer

!

b. Reactor Manual Control System i

c. Process Computer d. Rod Block Monitor System

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f QUESTION: 035 (1.00)

i Unit 2 is at rated power when the following indications occur:

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Fluid Drive

"A" Scoop Tube Lock annunciator alarns Speed Control

"A" Signal Failure annunciator alarms j

>

Feedwater flow decreases PSW Div I pressure indication is zero f

Inverter 2R44-S002 Trouble annunciator alarms

Which ONE of the following conditions would result in the listed i

indications?

-

i a. A loss of Vital AC.

.

b. A trip of the

"A" Recirc Pump.

f

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c. A recirc system runback to the #2 speed limit.

l

d. A loss of Instrument Bus

"A" i

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SENIOR REACTOR OPERATOR Page 25

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l

QUESTION: 036 (1.00)

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WHICH ONE of the following conditions would be a Technical Specification Safety Limit violation?

a. Unit 2 is operating at 100% power.

Reactor steam dome pressure increases to a maximum of 1100 psig without a scram.

The operator manually scrams and restores pressure to 920

psig.

I i

b. Unit 1 is in Cold Shutdown with the RHR system operating in

Shutdown Cooling.

The reactcr steam dome pressure increases

to 155 psig without occurrence of an isolation.

,

c. Unit 1 is operating at 22% pcwer end the EHC pressure

'

regulator fails.

The reactor preosure drops to 820 psig before the reactor scrams on M31V closure.

d. Unit 2 is in Refueling.

RPV level is lost due to an

unisolable leak.

Level drops to -167 inches before it is l

restored to -15 inches with emergency systems.

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QUESTION: 037 (1.00)

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The EOPs direct emergency depressurization of the RPV prior to entering i

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the UNSAFE region of the HCLL due to low Suppression Pool level.

j

Which ONE of the following is the reason for this action?

i

'

a. Ensures that the RPV is depressurized prior to evacuating 99% of

-

the non-condensibles to the Suppression Pool.

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b. Ensures the downcomers will not become uncovered causing loss of l

pressure suppression on a LOCA.

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c. Ensures that unstable steam condensation (chugging).does not occur in the downcomer during a LOCA.

!

d. Ensures that excessive hydraulic stress is not-applied to the f

SRV tailpipes and supports on a LOCA.

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&w SENIOR REACTOR OPERATOR Page 26 QUESTION: 038 (1.00)

Which ONE of the following is the bases for maintaining the MINIMUM water level in the Spent Fuel Storage Pool per Unit 2 Technical Specifications?

a.

To keep the fuel below a Keff of 0.95 delta K/K.

b.

To provide adequate shielding to personnel working in the area.

c.

To remove iodine gap activity f ollowing a fuel rupture accident.

d.

To insure adequate heat removal from irradiated fuel.

QUESTION: 039 (1.00)

Which ONE of the following is the advantage of venting the primary containment through the suppression pool?

a.

Condenses any steam from the primary containment.

b.

Prevents excessive drywell-to-torus differential pressure.

c. Minimizes amount of radioactivity released.

d. Dilutes the explosive hydrogen concentrations.

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f QUESTION: 040 (1.00)

f

During testing of the Unit 2 Standby Gas Treatment' System (SBGT)

!

Train

"A",

the following plant conditions exist:

!

~f Reactor power 80%

!

SBGT

"A" control switch In STANDBY l

SBGT

"A" Discharge _ Isolation Valve 2T48-F002A OPEN j

"

SBGT

"A" Suction Isolation Valve F076 CLOSED l

Train

"A" CR1 heater outlet temperature 305 deg F-

!

Which ONE of the following is the response of the SBGT Train "A"

!

following a high drywell pressure initiation?

I

,

a. Will not start due to the control switch being in STANDBY.

{

!

b. Will not start due to high heater outlet temperature.

,

j c. Will start after the Discharge Isolation Valve realigns.

,

q d. Will start and the Suction Isolation Valve will realign.

!

I i

I

"

i QUESTION: 041 (1.00)

'

,

During an ATWS, reactor power is 35%.

Which ONE of the following actions would be taken to MINIMIZE the heat addition to the suppression pool?

.

l

.

a.

Boron injection is commenced if temperature approaches 110 F

}

to ensure'that the Technical Specifications limits are not

!

exceeded.

b.

If power incres.ses to greater than 45% then immediate tripping j

of the recircu3ation pumps from their present speed is

>

necessary to reduce heat added to the suppression pool.

c.

If, due to a failure of the SLC system, injection of boron is

.

delayed, then lowering of= level should be delayed until boron l

is injected regardless of heat addition to the pool.

i

!

d.

Level should be reduced until EITHER level is at the top of i

active fuel (TAF) OR power is less than 3% with suppression l

. pool _ temperature less than 110-F, AND all SRV's closed.

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SENIOR REACTOR OPERATOR Page 28 I

l QUESTION: 042 (1.00)

!

'

i Select the Emergency Classification that completes the following l

statement.

i

!

When Events are in progress or have occurred which involve actual OR l

imminent substantial core degradation OR melting with the potential for

loss of containment integrity, the Emergency Director would declare i

a(an)

l

.

a.

Unusual Event

!

l b.

Alert f

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c.

Site Area Emergency

!

l d.

General Emergency l

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a l

QUESTION: 043 (1.00)

j i

!

In preparation for plant startup, the operator lining-up the Standby _

l Diesel Service Water (PSW) system informs the Shift Supervisor that a

valve is in the correct position but is listed incorrectly on the_ valve i

checkoff sheet.

]

'

Which ONE of the following describes the action to be taken to make a

'

temporary change to the procedure?

j i

a.

Only the Shift Supervisor needs to authorize this change, since it does not change the intent of the procedure.

,

I

"

b. Two members of management nmst approve the change, since this is

i a safety related procedure.

!

'

c. The change must be reviewed by PRB (can be done by phone) and approved by the applicable manager prior to implementation.

d. The valve should be repositioned with an * beside the desired

'

position and a note sent to Document Control to initiate a i

i procedure change.

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SENIOR REACTOR OPERATOR Page 29

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-

QUESTION: 044 (1.00)

!

Which ONE of the following. Plant Hatch personnel makes the determination

'

that Urgent Maintenance is required to maintain a schedule or plant j

safety?

l

'

!

a. Operation Supervisor on Shift.

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,

b.

Plant Support Manager.

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c.

Plant Manager.

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d. Shift Supervisor on Duty-f i

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SENIOR REACTOR OPERATOR Page 30 QUESTION: 045 (1.00)

During an entry into a HIGH RADIATION AREA, an operator will receive an estimated 70 mrem whole body dose.

The following is the radiation exposure information available on the operators who are available for the task.

Time constraints will not permit authorization of an increase in Plant Hatch administrative limits.

NRC Form 4s are on file unless otherwise indicated.

Operator 1 Operator 2 Operator 3 Operator 4 Sex Male Male Female Male Age

30

20 Wk/ Exposure 200 mrem 0 mrem 5 mrem 98 mrem Otr/ Exposure 1190 mrem 1960 mrem 435 mrem 420 mrem Lif e/ Exposure

-

55370 mrem 2735 mrem 8970 mrem Remarks History NONE 3 months NONE Unavailable Pregnant-Signed Prenatal Document on File Which ONE of the following personnel should be selected for the job?

a.

Operator 1 b.

Operator 2 c.

Operator 3 d.

Operator 4

.

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SENIOR REACTOR OPERATOR Page 31 i

.

QUESTION: 046 (1.00)

l Which ONE of the following procedure changes WOULD alter the intent of

the procedure?

l a. The procedure is changed to indicate an authorized temporary

test gauge will be in place for three weeks while the normal pressure instrument is in the shop for repairs.

'

i t

b. The procedure is changed so that when the APRMs are

!

recalibrated, the neutron trip setpoint would be lower than that i

required by Technical Specifications.

j c

The procedure is changed so that shielding is removed in the f

vicinity of the RWCU Cleanup Filter /Demins which will result in i

a higher radiation dose to the PEO stationed in the area.

d. The procedure is changed such that a trip setpoint listed in the

!

l procedure is altered to agree with the vendor's technical

manual.

j i

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QUESTION: 047 (1.00)

Which ONE of the following IS NOT a valid reason for exiting an EOP flow chart?

I j _

a. The EOP flow charts have directed exiting to a normal operating

[

procedure.

'

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b. All entry conditions for that flow chart have cleared and an

'

emergency no longer exists.

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'

i c. An entry condition to another EOP chart with a higher priority has occurred.

d. An entry condition exists, but-the Shift Supervisor has determined that an emergency no longer exists.

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SENIOR REACTOR OPERATOR Page 32.

QUESTION: 048 (1.00)

,

FILL'IN THE BLANKS The Fire Brigade must consist of a minimum of qualified

,

persons. A minimum of of these persons one must have competent

!

knowledge of safety related systems and components.

a.

5,

b.

6,

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c.

5,

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d.

6,

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QUESTION: 049 (1.00)

!

Following a core damage accident, gaseous effluent is released at the

.

plant. At the site boundary the whole body radiation dose is measured at l

620 mR/hr for 45 seconds and then drops rapidly to 45 mR/hr.

By the end I

of the following 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months the dose rate drops to background levels.

i

'

Which ONE of the following is the lowest level Emergency Classification for this event?

(73EP-EIP-001-OS is enclosed as attachment 1 for i

reference)

!

a.

Unusual Event i

b.

Alert t

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c.

Site Area Emergency i

!

d.

General Emergency I

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t SENIOR REACTOR OPERATOR Page 33

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!

i QUESTION: 050 (1.00)

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!

A reactor startup is being performed on Unit 2 with the Reactor Mode

-!

i

"

Switch in START / HOT STANDBY and Reactor Pressure at 700 psig.

.

A review of the ADS LSFT 42SV-B21-003-2S surveillance data package i

reveals the

"E" ADS valve is NOT acceptable.

HPCI CSS and LPCI j

surveillances are current and not required.

,

Which ONE of the following actions should be taken by the Shift

Supervisor?

a. Enter a 14 day LCO and direct the Unit 2 Startup to be stopped

!

prior to entering the RUN mode.

,

!

b.

Enter a 14 day LCO and continue the Unit 2 Startup to rated

power.

j

'

-c. Direct the Unit 2 Startup to be continued because NO LCO needs to be entered for present plant conditions.

d.

Initiate a reactor shutdown to HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce reactor steam dome pressure to less than or equal to 150 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j i

i

_ QUESTION: 051 (1.00)

,

A valve has been tagged closed for-a clearance on a safety related f,

. system.

Independent verification of the valve position is required.

j

,

!

Health Physics has determined that there is a high radiation level in-

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the area of the valve.

Which ONE of the following describes the disposition of the independent

,

' verification of the valve?

i-i a. Verification can be waived by the Shift Supervisor, but.

i

,

alternate means of verification should be considered.

!

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l b. Verification can be waived by the Shift Supervisor, but must be i

visually checked prior to commencing work.

I c. Verification can be waived only if approved by.the Operations l

Department Manager when visual exposure results in excessive j

exposure.

l d. Verification can be waived only by the Operations Department l

Manager if'an. alternate means of verification is available.

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SENIOR REACTOR OPERATOR Page 34 QUESTION: 052 (1.00)

An event has occurred on unit 2 resulting in an entry into PC-1, Primary Containment Control. The following plant conditions exist:

,

Drywell temperature 122 Deg F Suppression Pool Water Level 148 inches Suppression Pool temperature 105 Deg F Drywell pressure 1.1 psig Which ONE of the following is the required IMMEDIATE action if the Suppression Pool water level increases to 151 inches?

a.

Return to that key parameter, readdress, and continue from there.

b.

Ignore the change since subsequent actions have insured plant safety.

,

c.

Return to top of flow chart and begin to follow all flow paths again.

d.

Go to End Path Manual and follow procedure for that parameter.

QUESTION: 053 (1.00)

A reduction in recirculation pump speed has occurred while operating at 100% power causing the Region of Potential Instabilities to be entered.

Which ONE of the following actions would be taken based on the given nuclear instrumentation respor.se?

(Consider each case independently.)

a.

IF three LPRMs in one quadrant are oscillating at an 8%

bandwidth, rods should be inserted in that quadrant to suppress the oscillations.

b.

If one APRM is oscillating at a 12% bandwidth with the others oscillating at an 8% bandwidth, recirculation flow should be

!

increased to suppress oscillations.

j c.

If two LPRMs in opposite quadrants are oscillating at an 8%

bandwidth, the reactor should be scrammed.

l d.

If all observed LPRMS are oscillating at a 12% bandwidth, the reactor should be scrammed.

-- -.- -

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w SENIOR REACTOR OPERATOR Page 35 QUESTION: 054 (1.00)

An ATWS has occurred on Unit 2 with a steam line isolation. To reopen the MSIVs, EOP 31EOP-111-2S, Emergency Opening of MSIVs, directs the operator to override the high flow isolation on 2P70-F004 and 2P70-F005.

Which ONE of the following is the purpose f or this override?

a.

Restores operating pressure to the outboard MSIVs to allow opening the valves.

b.

Permits resetting the main steam line flow isolation to allow opening the MSIVs.

c.

Restores operating pressure to the inboard MSIVs to allow opening the valves.

d.

Permits resetting the RPV low water level isolation to allow opening the MSIVs.

QUESTION: 055 (1.00)

Unit 2 is at power when a complete loss of the CRD system occurs. The operator has been unable to restore charging water pressure.

Which ONE of the following describes the conditions and actions which should be taken in accordance with 34AB-C11-001-2S, Loss of CRD System?

a.

Commence a FAST REACTOR SHUTDOWN if RPV pressure is LESS than 800 psig and more than four accumulator trouble lights are illuminated.

b.

Commence a FAST REACTOR SHUTDOWN if RPV pressure is GREATER than 800 psig and four accumulator trouble lights are illuminated.

c. Manually SCRAM the reactor if RPV pressure is LESS than 800 psig and four accumulator trouble lights are illuminated.

d. Manually SCRAM the reactor if RPV pressure is GREATER than 800 psig and more than four accumulator trouble lights are illuminate &

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w SENIOR REACTOR OPERATOR Page 36 QUESTION: 056 (I.00)

Which ONE of the following describes the significance of a Primary Containment level of 83.9 feet during Primary Containment Flooding?

a.

The top of active fuel in the RPV.

b.

The structural load limit of the containment walls.

c.

The highest containment vent capable of rejecting all core decay heat.

d.

The bottom of the main steam lines.

QUESTION: 057 (1.00)

After a transient Unit 2 RPV level is being maintained at -150 inches with RCIC as the only high pressure injection system.

Which ONE of the following statements describes RCIC use for pressure control?

a.

RCIC MAY NOT be used because defeating the suction swap is not permitted.

b.

RCIC MAY NOT be used because there is an initiation signal present.

c.

RCIC MAY be used for pressure control after realigning the suction to the CST.

d. RCIC MAY be used for pressure control after overriding the CST test isolation valve interlock _

. _,

.

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SENIOR REACTOR OPERATOR Page 37 l

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QUESTION: 058 (1.00)

[

Unit 2 Reactor Building Closed Cooling Water (RBCCW) System is operating t

'

with pumps 2A and 2B in service and pump 2C in standby.

The feeder

,

breaker to 600V bus 2D fails causing the bus to deenergize.

!

Which ONE of the following will be the (RBCCW) system response?

?

,

a.

Pump 2A will trip off, 2C will fail to start and pump 2B will i

continue run.

!

b.

Pump 2B will trip off and pump 2C will start on low header

!

pressure.

l i

.

i

c.

Pump 2A and 2B will trip off and pump 2C will start on low l

header pressure.

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d.

Pump 2A will trip off and pump 2C will be locked out on an l

undervoltage trip.

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.

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l i

,

QUESTION: 059 (1.00)

.

Unit 2 is at rated power when a loss of Instrument and Service air

,

occurs.

I Which ONE of the following indications will require a reactor scram l

after the loss of air?

E l

,

a.

Indication of 1 control rod drifting in coincident with a CRD

!

HYD HIGH TEMP annunciator, i

b.

Indication of 3 control rods drifting in.

i

j c.

Scram pilot air pressure of 46 psig.

,

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d.

SCRAM VLV PILOT AIR HEADER PRES 6URE HIGH/ LOW annunciator is received.

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SENIOR REACTOR OPERATOR Page 38 f

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"

QUESTION: 060 (1.00)

i

!

Which ONE of the following is the definition of " Minimum Alternate RPV j

Flooding Pressure" (MARPVFP)?

j

a.

The lowest pressure at which steam flow through the SRVs will

}

remove the decay heat generated'10 minutes after a scram from full power.

I b. The lowest pressure at which steam generation needed for steam cooling will not be suppressed.

j l

c. The lowest pressure at which steam flow through the SRVs will f

preclude clad temperature from exceeding 2200 deg F.

j

,

d. The lowest pressure at which steam flow through the SRVs will preclude clad temperature from exceeding 1500 deg F.

,

i l

!

QUESTION: 061 (1.00)

i

!

Which ONE of the following level instruments is reliable for the given l

drywell temperature.

!

>

Instrument Reading RTD 2T46-N001K RTD 2T47-N001A

,

L a.

2B21-R610-295" 288 deg F 286 deg F

i

b.

2C32-R606C

+2" 280 deg F 271 deg F

!

c.

2B21-R605'

+37" 255 deg F 260 deg F

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d.

2B21-R604B-144 198 deg F 207 deg F l

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SENIOR REACTOR OPERATOR Page 39 QUESTION: 062 (1.00)

A transient has occurred in which the reactor scrammed at 2237 hours0.0259 days 0.621 hours 0.0037 weeks 8.511785e-4 months and all RPV level indication became erroneous.

RPV Flooding has been initiated.

At 2258 a 50 psid differential was established between RPV pressure and Suppression Chamber pressure with 5 SRVs open.

Which ONE of the following is the earliest time that core coverage can be assured?

a.

2246 hours0.026 days 0.624 hours 0.00371 weeks 8.54603e-4 months b.

2307 hours0.0267 days 0.641 hours 0.00381 weeks 8.778135e-4 months c.

2321 hours0.0269 days 0.645 hours 0.00384 weeks 8.831405e-4 months d.

2342 hours0.0271 days 0.651 hours 0.00387 weeks 8.91131e-4 months

,

QUESTION: 063 (1. 0 0)

While restoring RPV water level following RPV Flooding, jumpers are installed for the RHR Torus Spray Valves (2E11-F02 8A/B).

Which ONE of the following is the reason that the jumpers are installed?

a.

To establish torus spray capability.

b.

To establish suppression pool cooling.

c.

To override the LOCA isolation signal.

d.

To drain the RPV into the suppression pool.

,

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I SENIOR REACTOR OPERATOR Page 40

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QUESTION: 064 (1.00)

l

Which ONE of the following statements explains the reason for inhibiting l

ADS if boron is injecting into the RPV?

!

a. ADS actuation could slow-the rate of boron injection into the

'

core due to the additional two phase flow resistance.

!

t b. ADS actuation results in void formation which reduces control

,

rod worth and lowers shutdown margin.

'

,

c. ADS actuation results in boron dilution by carry over into the i

Torus causing a power excursion.

r L

d. ADS actuation could result in the rapid addition of-cold l

unborated water to the reactor by low pressure injection.

j J

i

I QUESTION: 065 (1.00)

j

Which ONE of the following is a reason that drywell sprays may be initiated IRRESPECTIVE of adequate core cooling per "PC - Primary

[

Containment Control"?

l a.

To prevent excessive drywell-to-torus differential pressure to.

i

,

l prevent f ailure of the drywell floor.

j i

b.

To reduce drywell pressure when above 13 psig and suppression pool water level is above the vacuum breakers.

l I

c.

To prevent tripping tlue RCIC and HPCI turbines due to elevated

'

Suppression Chamber pressure.

'

!

d.

To dilute drywell hydrogen and oxygen concentrations to prevent a containment explosion.

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' SENIOR REACTOR OPERATOR Page 41

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i i

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QUESTION: 066 (1.00)

-An event has occurred on Unit 2 resulting in the following plant

!-

conditions:

!

'

Reactor pressure 1020 psig Main Steam Tunnel temperature 160 deg F

'

Suppression Chamber air temperature 132 deg F i

RWCU Equipment Room temperature 122 deg F j

Drywell Pressure 1.95 psig

!'

Which ONE of the following sets of valves should have isolated?

I i

a.

Reactor Water Sample Line Valves l

i l

b.

Suppression Chamber Vent Valves I

!

c.

HPCI Steam Supply Valves

!

.

f d.

RWCU Suction Supply Valves I

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i SENIOR REACTOR OPERATOR Page 42 I

,

QUESTION: 067 (2.00)

,

'

All Unit 2 Service Air Compressors have tripped and system air pressure

!

is decreasing.

For the actions listed in Column A SELECT the I

,

appropriate air pressure set point from Column B.

(Items in Column B i

.

may be used once, more than once, or not at all.

Only one answer may l

l occupy a space in Column A)

(4 required at

!

0.5 each)

!

!'

CCLUMN A COLUMN B

.,

(Action)

(Setpoint)

,

a. When Turbine Building Instrument Air 1.

90 psig-j pressure drops to the standby t

,

prefilter and afterfilter are 2.

85 psig

!

'

automatically placed in service.

l

,,

3.

80 psig j

b. When Control Air Pressure drops to Startup LCV, 2N21-F111 locks 4.

75 psig

,

up in its existing position.

,

5.

70 psig

!

c. When Non-interruptable air pressure

'

drops to

, Reactor Building 6.

65 psig i

j Instrument Nitrogen to Non-Interruptable l

Air Isolation Valve 2PS2-F565 opens.

7.

60 psig

i i

'

d. When Service Air Pressure drops 8.

55 psig

!

to Service Air Isolation Valve

!

,

'

2P51-F017 closes.

9.

50 psig d

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QUESTION: 068 (1.00)

.;

i

'

l The plant is operating normally at power when Recirculation Pump A Seal

Stage Flow alarms low and the operator notes No. 2 Recirc Pump seal 4 ;'

' pressure decreasing towards zero with No. 1 seal pressure stable at 1000 psig.

Which ONE of the following failures would cause these indications?

(No other alarns are present.

Figure 2 is enclosed for reference.).

.

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a. Failure of No. 1 seal

!

b.

Failure of No. 2 seal

!

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c.

Plugging of the No. 1 internal restricting / breakdown orifice

.

d.

Plugging of the No. 2 internal restricting / breakdown orifice

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t SENIOR REACTOR OPERATOR Page 43

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J

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l

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QUESTION: 069 (1.00)

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,i A Unit 2 reactor startup is in progress with the Plant Operator

withdrawing Control Rods in Rod Sequence Step 9 which has the following j

limits:

I

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>

.

Insert

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)

Withdraw

i i

!

Several Rod Worth Minimizer (;RWM) Group 9 Control Rods are still at l

l position 08 when the operator inadvertently withdraws a Group 9 rod to

i position 18.

Which ONE of the following describes the operational status of the Roc l

Worth Minimizer System and the corrective action to be taken? (No RWM

'

i lights or alarms are present.)

.

l a. The RWM is INOPERABLE and a second licensed operator or I

qualified member of the tech staff should be stationed at the

!

P603 panel to double verify any further rod movement.

-l b.

The RWM is INOPERABLE; all Control Rod movement should be l

,

"

'

terminated and the reactor manually scrammed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months unless the RWM is made operable.

j c. The RWM is INOPERABLE and the RWM Group 9 Control Rod withdrawn

j to position 18 should be immediately inserted back to within its j

withdrawal limit before any other Control Rod movement is

!

i attempted.

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d.

The RWM is INOPERABLE and the RWM Group 9 Control Rods should be l

inserted to their insert limit before any other Control Rod j

movement is attempted.

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SENIOR REACTOR OPERATOR Page 44 QUESTION: 070 (1.00)

DCN 90-066 installed a 3/8 inch bypass line around the EGR actuator for the HPCI turbine sper-d control.

Which CNE of the following is the reason for installing the bypass line?

a.

Reduces the peak speed of the HPCI Turbine on startup to prevent tripping on overspeed.

b.

Keeps the Governor Valve closed until the Stop Valve is fully opened to reduce startup vibration.

,

c.

Removes the Ramp Generator from the speed control during manual operation to allow faster speed increase.

d. Maintains the Governor Valve at approximately 50% open until the stop valve begins opening to prevent Ramp Generator oscillations.

.

_ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _. _ _ _

e A

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w SENIOR REACTOR OPERATOR Page 45 QUESTION: 071 (2.00)

For the components of the Control Rod Drive (CRD) system labeled a through d on Figure 3 make the proper identification in Column A from the list of components in Column B.

(Components listed in Column B will only be used once and only one answer may occupy a space on Column A.)

(4 answers required, 0.5 each)

COLUMN A COLUMN B (Components)

(Component Identification)

a.

1.

Drive Water Pressure Control Valve.

l b.

I 2.

Stabilizing Valve.

3.

Flow Control Valve.

d.

l 4.

Return Line Pressure Control Valve.

5.

Cooling Water Pressure Control Valve j

6.

Suction Filter.

l 7.

Drive Water Filter.

8.

Flow Element.

I i

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' SENIOR REACTOR OPERATOR Page 46'

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I QUESTION: 072 (1.00)

.

Both RFPTs have tripped with a failure of HPCI to start resulting.in a significant reactor water level' transient.

~

The following plant conditions currently exist:

The reactor has scrammed All rods in i

Reactor water level-95 inches decreasing l

Reactor pressure 987 psig decreasing j

Drywell pressure 1.67 psig steady j

Torus water temperature 96 deg F j

An operator observes the Core Spray System is running.

Which ONE of the following describes the correct status and corrective action to be taken for the Core Spray System?

I

.

a. The Core Spray System should have initiated on high-drywell i

pressure and should be allowed to run

-i b. The Core Spray System should have initiated on low reactor water

level and should be allowed to run

'!

'

I c..The Core Spray System should NOT have initiated but can be i

allowed to run until water level is restored.

s

!

'd.

The Core Spray System should NOT have initiated but can be used

!

for pressure control.

>

l l

QUESTION: 073 (1.00)

!

A caution in the procedure for cooling in the steam condensing mode with

the RHR Heat Exchangers states "A pressure of 75 psig on the.RCIC pump

i suction must not be exceeded during the steam condensing mode of

'

l operation.

-!

'Which ONE of the following is the reason for maintaining this pressure.

!

.

below 75 psig?

'

a. Could cause failure of the RCIC suction seals.

l b. Could result in exceeding the RCIC design pressure discharge limit.

~:

i c. Could result in over pressurization of the RCIC Lube Oil Cooling System

.

d. Could result in decreasing the RHR heat exchanger level too-rapidly.

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SENIOR REACTOR OPERATOR Page 47_

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QUESTION: 074 (1.00)

{

i A transient has occurred on Unit 2, resulting in the following

'j conditions.

i Reactor Power All rods in RPV Pressure 950 psig i

Reactor Water Level

+30 inches and stable l

Drywell Pressure 1.2 psig j

'

Drywell Temperature 160 deg F (Bulk temperature)

l Torus Level

+46 inches

!

4160 VAC buses

"A" and

"B" are deenergized.

!

MSIV's are closed.

'

The Primary System is discharging into the reactor building with r

the following radiation levels:

)

,

158' level are N.E.

1100 mR/hr l

130' N.E.

working area 1300 mR/hr l

!

Which ONE of the following methods would the SS direct to Emergency Depressurize the reactor?

i I

!

a. Opens 7 SRV's.

b.

Place HPCI in full flow test.

!

c.

Open the steam.line drains.

,

d.

Equalize around the MSIV's and use the MSIV's with the Bypass l

Jack.

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SENIOR REACTOR OPERATOR Page 48 QUESTION: 075 (1.00)

Unit 2 has experienced a loss of Main Condenser Vacuum resulting in MSIV closure and a reactor scram.

The following plant conditions exist:

Control rods Fully inserted.

HPCI system Inoperable.

RPV water level-10 inches and slowly increasing.

Reactor pressure 905 psig.

While operating RCIC to restore RPV water level, a steam leak develops

'

on the Trip and Throttle Valve, but RCIC continues to operate.

RCIC diagonal ambient 195 deg F temperature Area radiation 55 mr/hr Which ONE of the following is the action the SS would direct?

a.

Operate RCIC area coolers until area temperature or radiation levels exceed Max Safe operating value, then isolate RCIC and use Core Spray to restore RPV water level.

b.

Continue operating RCIC to restore RPV water level until water level is greater than +32 inches.

c.

Immediately isolate RCIC and use CRD to restore RPV water level to between +12.5 and +51.5 inches.

d.

Immediately isolate RCIC and depressurize the plant to use low pressare systems to restore water level per CP-1 Alternate Level Control.

,

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f i

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t QUESTION: 076 (1.00)

]

l CP-3 ATWS Level Control has been implemented to the point of " SLOWLY I

raise injection to restore and maintain RPV water level above - 185 l

inches."

!

Which ONE of the following describes the action (s) to be taken to SLOWLY raise injection?

,

a. Start one low pressure pump at minimum flow and gradually l

increase flow until RPV level begins to increase.

i b. Start one low pressure pump, immediately increase flow to rated

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and start additional pumps sequentially at rated flow until

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level begins to increase.

p n

I c. Start two low pressure pumps at maximum flow until level begins

.;

to increase, then reduce flow to less than 5000 gpm.

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.

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,

d. Start two low pressure pumps at minimum flow, increase flow to

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5000 gpm per pump and start additional-pumps if required.-

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,

QUESTION: 077 (1.00)

l q

t Restarting Turbine Building HVAC during a rad release above an ALERT-

!

-

l Emergency insures that:

'

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-

a. Any release to the Turbine Building is exhausted through SBGT.

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,

b. Accessibility to the Turbine Building is preserved.

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i j

c. The Turbine Building Blow Out Panels do not blow out.

i i

d. Release to the Turbine Building is contained in the Turbine

Building.

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I QUESTION: 078 (1.00)

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The following conditions exist after a Unit 2 scram from 100% power:

Highest Drywell temperature 210 deg F Drywell pressure 1.3 psig

!

Reactor pressure 920 psig

'

-i The following water level instruments read as indicated:

l

,

i Floodup Range, 2B21-R605

+20 inches f

Narrow Range, 2C32-R606A

+6 inches i

Wide Range, 2B21-R604A

+5 inches

[

Fuel Zone, 2B21-R610-20-inches I

l Which ONE of the following Reactor Water Level indicators WOULD NOT be

!

reliable for level trend information in accordance with the EOPs?

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I a. Floodup Range, 2B21-R605

!

I b. Narrow Range, 2C32-R606A j

!

c. Wide Range, 2B21-R604A l

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.

d. Fuel Zone, 2N21-R610

.j

!

I i-i QUESTION: 079 (1.00)

i i

Unit 2 is operating at 75% RTP when SRV 2B21-F013A inadvertently opens.

!

Suppression Pool temperature increases to 115 deg F before the operator

!

,

'

pulls'the fuses for the valve and it closes.

Which ONE of the following is the appropriate IMMEDIATE action to be-

'

taken in accordance with plant procedures and the Unit 2 Technical

!

Specifications?

!

"

!

a. Place the Mode Switch to SHUTDOWN.

!

'

b. Reduce Suppression Pool Temperature to less than 100 deg F.

within the next four hours I

c. Initiate a fast reactor shutdown.

I

!

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d. Place the plant in the HOT SHUTDOWN within 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months and l

COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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i SENIOR REACTOR OPERATOR Page 51

'i OUESTION: 080 (1.00)

!

A loss of the 2C and 2D 4160 VAC switchgear occurs'following a main j

turbine trip.

Which ONE of the following describes the operation of the' Unit 2 Reactor

.

j Recirculation MG set Lube Oil system?

a. The DC oil pump for each Reactor Recirculation MG set will auto-j start, supply oil to the motor and the generator.

,

b. The running AC oil pump for each Reactor Recirculation MG set j

will continue running from its invertor's backup power supply.

c. The A3 and B3 AC Reactor Recirculation MG set oil pumps will f

auto-start supplying oil to the fluid drive bearings.

t

.

d. The DC oil pump for each Reactor Recirculation MG set will auto-l

'

start and supply oil to the fluid drive bearings.

I

i f

.

QUESTION: 081 (1.00)

j

!

'

A pipe break in the Unit 2 Reactor Building resulting in steam tunnel

temperatures greater than Max Normal and an automatic shutdown of the i

Reactor Building HVAC system. The leak has now been isolated.

!

-Which CNE of-the following is a condition'that has to be met prior to

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restarting the HVAC system?

i i

a. Drywell pressure nmst be below 1.85 psig.

l t

b. The reactor building potential contamination area ventilation exhaust radiation level must be below 9.5 mr/hr.

I c. RPV water level must be greater than 0 inches.

'

d.

Not more than one area above the Maximum Nornal Operating

,

.

temperature.

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&w SENIOR REACTOR OPERATOR Page 52 QUESTION: 082 (1.00)

An event has occurred on Unit 2 with the plant operating at 97% RTP.

The following conditions exist:

RE SE Diagonal Area Floor Drain Sump High Alarm RB SE Diagonal Area Instrument Sump High Alarm Southeast Diagonal Area Radiation Level 9 mr/hr Southeast Diagonal Area Ambient Temperature 92 deg F Which ONE of the following describes the action (s) which should be taken by the crew and the reason f or the action (s) ?

a.

Entry into the SC flow chart is NOT required because no entry conditions exist.

b. Entry into the SC flow chart is required because sump levels arc abnormal.

c.

Entry into RC and SC flow charts is required because area radiation levels are above max normal operating level.

d.

Entry into RC and SC flow charts is required because SE Reactor Building diagonal water level is above max normal operating leve,

&

/

W SENIOR REACTOR OPERATOR Page 53 QUESTION: 083 (1.00)

An event has occurred on Unit 2 with the plant operating at 68% RTP.

The following plant conditions exist:

Steam Tunnel Area temperature 145 deg F 130' Northwest Area Radiation 1100 mr/hr 130' Southwest Area Radiation 820 mr/hr It has been reported that steam is leaking from one of the main steam lines.

Which ONE of the following is the action that would be taken based on the plant conditions?

a. The reactor must be shutdown per 34GO-OPS-013-2S, Normal Shutdown.

b. The reactor must be shutdown per 34GO-OPS-14-2S, Fast Reactor Shutdown.

c.

The reactor must be scrammed.

d. The reactor must be scrammed and Emergency Depressurization initiated.

QUESTION: 084 (1.00)

At the completion of an " EMERGENCY IN" rod movement, a " ROD DRIFT" alarm is received.

Which ONE of the following describes the reason for this alarm?

a. The rod is not at an even reed switch and none of the selected relay busses are energized (insert, withdraw, or settle).

b. The rod is not at an odd reed switch and none of the selected relay busses are energized (insert, withdraw, or settle).

c. The sequence timer is bypassed causing an insert and withdraw signal at the same time.

d.

" EMERGENCY IN" bypasses the Rod Position Indication Syste O O

.

!

i SENIOR REACTOR OPERATOR Page-54 QUESTION: 085 (1.00)

>

The " Reactor Recirc Loop B Out of Service" annunciator on P602 failed to reset after the operator completed the startup of Recirc Pump 2B.

The annunciator has been in the alarm condition for the last two (2) shifts.

,

An MWO has been initiated and approved to repair the annunciator.

!

Which ONE of the following describes the status of the annunciator?

a. The annunciator IS CONSIDERED a Problem Annunciator because it has been lit (cannot be reset) for more than one shift, b. The annunciator IS CONSIDERED a Problem Annunciator because an Engineering Evaluation has not been completed.

c. The annunciator IS NOT CONSIDERED a Problem Annunciator because it has not been deactivated.

,

i

'

d. The annunciator IS NOT CONSIDERED a Problem Annunciator because it cannot bend deactivated (reset) and is therefore no distraction to the operator

-

.

!

t QUESTION: 086 (1.00)

Which ONE of the following is the reason for preventing Primary

[

Containment water level from exceeding 103.5 feet during Primary i

Containment flooding?

'

!'

a. This level is equal to the top of active tuel in the reactor vessel.

b. This level exerts pressure equal to the structural design load

,

limit of the containment walls.

!

c. This level is the elevation of the highest containment vent l

capable of rejecting all core decay heat.

'

d. This level is the elevation of SRV pilot valve solenoid i

operators.

4

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Page 55 j

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l i

QUESTION: 087 (1.00)

!

e Fuel movemer is'in progress on Unit 2.with the following plant

!

conditions:

l

!

MODE COOLANT REACTOR i

SWITCH POSITION TtMPERATURE POWER

!

.I Unit 1 RUN 545 deg F 80%

I Unit 2 REFUEL 128 deg F 0%

!

Which ONE of the following is the MINIMUM. shift staffing requ1 red by the

.;

Unit 2 Technical Specifications?

l I

a.

SROs 2 + 1 for Fuel Handling Ros

NLOs

-;

STA

!

i b. SROs 1 + 1 for Fuel Handling

ROs

.

NLOs

.!

.STA

,

c. SROs 1+ 1 for Fuel Handling f

ROs

i'

NLos

STA NONE

'

t d. SROs 2+1 for Fuel Handling (

ROs

4 NL0s

l STA NONE

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v SENIOR REACTOR OPERATOR Page 56 QUESTION: 088 (1.00)

A Temporary Release (TR) to a clearance has been written and pldced in the TR book.

Subsequently a Maintenance Foreman requests an additional subclearance on this clearance from the Shift Supervisor.

Which ONE of the following actions should be taken by the Shift Supervisor?

a. Direct Foreman to obtain the current subclearance holders permission and then proceed with his subclearance, b. Authorize the subclearance after obtaining the approval of all other subclearance holders.

c.

Direct the Foreman to work his job under another supervisors subclearance.

d.

Deny the Foreman a subclearance until the clearance is removed from the TR book.

QUESTION: 089 (1.00)

Which ONE of the following IS NOT in conformance with 30AC-OPS-001-OS, Control of Equipment Clearances and Tags?

a.

When pulling fuses to deenergize power supplies, the DANGER tags will be attached to the blockout fuses.

b.

When danger taggfay-an MOV as a clearance boundary, tagging the breaker and local aperator will satisfy the tagging requirements.

c.

It is permissible to repack a MOV with the valve on its backseat and a DANGER tag on its control switch, breaker, and local operator.

d.

When danger tagging an MOV which requires dependent verification it is permissible for the in(2 pendent verifier to use the valve position indication lights to determine valve positio _ _ _ - -

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SENIOR REACTOR OPERATOR Page 57 l

!

t QUESTION: 090 (1.00)

[

,

Which ONE of the following describes the method for performing a valve l

lineup check of a locked throttle valve?

I a. Unlock the valve, turn the handwheel 1/4 turn in the closed i

direction, then 1/4 turn in the open direction, and relock the (

e valve.

j b. Unlock the valve, turn the handwheel 1/4 turn in the open direction, then 1/4 turn in the closed direction, and relock the

,

valve.

c. Verify the position of the valve by checking the valve position

,

indicator.

d. Verify the operability of the attached throttle valve locking

'

device.

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!

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.

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!

"

QUESTION: 091 (2.00)

t

.i For the components of the OFF GAS SYSTEM labeled "a" through' "d" on Figure 4 l

make the proper identification in Column A from the list of components in l

Column B.

(Components listed in Column B will only be used once and only_one j

answer may occupy a space in Column A) (4 answers required, 0.5'each)

.l t

I COLUMN A COLUMN B (Items)

(Components)

l a.

1.

Charcoal Absorbers j

i l

b.

2.

Off-Gas Condenser c

Intercondenser

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d.

4.

Water Separator i

,

5.

Cooler Condenser

i 6.

Holdup Pipe f

7.

Catalytic Recombiner l

t 8.

Off-Gas Preheater f

i 9.

Moisture Separator j

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!

QUESTION: 092 (1.00)

l

Jul ATWS requiring boron injection has occurred-on Unit 2. An-unisolable steam leak into the RCIC diagonal area has resulted in RCIC isolation

,

and Secondary Containment temperature above Max Safe in more than one

!

area.

The Shift Supervisor has determined that Emergency

-!

Depressurization is required.

!

Which ONE of the following actions should be directed prior to emergency l

depressurization?

a. Terminate and prevent injection from all injection system

,

including CRD.

!

b. Terminate and prevent injection from Core Spray only.

f f

c. Terminate and prevent injection from RHR and Core Spray only.

j d. Terminate and prevent injection from'all injection systems

'

except boron, CRD and RCIC

e

!

QUESTION: 093 (1.00)

l

Which ONE of the following conditions would require Emergency Depressurization of the reactor in accordance with the Emergency Operating Procedure?

!

,

a. Ruptured instrument leg is releasing in excess of 1500 mr/hr to l

'

the Primary Containment and the Primary Containment cannot be isolated.

SGBT is operating.

?

b. Offsite release rate is 1200 mr/hr and a main steam line break

!

in the Tbrbine Building cannot be isolated.

Turbine building

,

ventilation is operating.

.;

!

c. Offsite release rate is 1000 mr/hr and an unisolable RBCCW j

!

break is discharging outside the primary and secondary

'

containment.

d. All the 130 foot elevation radiation levels are in excess of l

500 mr/hr I

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SENIOR REACTOR OPERATOR Page 59

,

QUESTION: 094 (2.00)

I

For the components of the ARI/ SCRAM AIR HEADER system labeled

"a"

'

through

"d" on Figure 5 make the proper identification in Column A from j

the list of components in Column B.

(Components listed in Column B will

,

only be used once and only one answer may occupy a space in Column A) (4 answers required, 0.5 each)

l COLUMN A COLUMN B (Components)

(Component Identification)

I a.

1.

Scram Outlet Valve i

i b.

2.

ARI Valve

"

c.

3.

Scram Inlet Valve

,

i d.

4.

Discharge Volume Drain Valve

.

!

5.

Discharge Volume Vent Valve 6.

Backup Scram Valve j

7.

Scram Air Header Vent Valve

,

QUESTION: 095 (1.00)

Which ONE of the following describes the results of a loss of a

Feedwater Heater Train?

a.

Core inlet subcooling decreases

,

'

Actual bundle power decreases Critical power ratio decreases b.

Core inlet subcooling increases i

Actual bundle power decreases Critical power ratio increases

!

l c.

Core inlet subcooling decreases

Actual bundle power increases

.

Critical power ratio increases r

i d.

Core inlet subcooling increases

!

Actual bundle power increases Critical power ratio decreases

!

(**********

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

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ANSWER:

001 (1.00)

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ti REFERENCE:

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f EIH Student Text LT-ST-03001-01 Page 18 Learning Objective EO-1 r

290001K104 [3.7/3.9)

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l 290001K104

..(KA's)

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ANSWER:

002 (1.00)

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d.

REFERENCE:

l EIH Student Text LT-ST-03001-01 DCN 91-171 page 22

Learning Objective EO-13

,

261000G007 [3.5/3.7)

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..(KA's)

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ANSWER:

003 (1.00)

i d.

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REFERENCE:

EIH: LT-IH-04801, l

Learning Objective E.O. 8.

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EIH: Unit 2 Technical Specification 3.6.6.2 223001G005 [ 3. 3 / 4.1]

223001G005

..(KA's)

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SENIOR REACTOR OPERATOR ~

Page 61

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I REFERENCE:

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EIH Student Text LT-ST-01301-00 Learning Objective EO 27 EIH Technical Specification 3.6.1.3 J-l 223001G005 [ 3. 3 /4.1]

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ANSWER:

005 (1.00)

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REFERENCE:

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EIH: LT-IH-00401-00, Page 27 j

Learning Objective EO 13

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202002K402 [3. 0/3. 0)

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202002K402

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I ANSWER:

006 (1.00)

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EIH Student Notes'LT-ST-00401-00 DCR 90-043 Page 16

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Learning Objective EO 7 i

i 202002A407 [3.3/3.2]

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i 202002A407

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i ANSWER:

007 (2.00)

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i a.: 5 b.

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c. 7

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d. 4

{4 required at 0.5 each)

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i REFERENCE-

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EIH LT-ST-OO701-01 Pages 21-23 l

Learning Objectives EO 11 & _20

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i 203000K410 [3. 9 /4.1]

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ANSWER:

008 (1.00)

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REFERENCE:

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EIH Unit 2 Technical Specifications 3.5.1 f

EIH Student Text LT-ST-30005-00 I

Learning Objective TO 300.010.a

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203000G005 '[3.5/4.4]

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203000G005

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ANSWER:

009 (1.00)

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.
._____
EPE-I Total 25.00 Group II OUESTION VALUE KA 053 1.00 295001G010 075 1.00 295002G012 080 1.00 295005K203 063 1.00 295008A109
'
058 1.00 295018K304 059 1.00 295019G010 067 2.00 295019K303 055 1.00 295022K207 061 1.00 295028A203
,
062 1.00 295028K302
086 1.00 295029A203
,
092 1.00 295032K301 083 1.00 295033G011
..
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TEST CROSS REFERENCE Page
SRO Exam BWR React or Oroani zed bv KA Group EMERGENCY PLANT EVOLUTIONS Group II QUESTION VALUE KA
,
081 1.00 295034K203 082 1.00 295036G011
,
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EPE-II Total 16.00
____..
......
EPE Total 41.00
......
____..
......
Test Total 100.00
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v t e Inspection Report - Hatch - 1993300 (Operator Exams)
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1993 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 13(0) 14(0) 15(0) 16(0) 17(0) 18(0) 19(0) 20(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) Security Operator Exams 300(0) 301(0) Emergency Planning Other

IR 05000321/1993300

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