IR 05000277/1990024
| ML20028H887 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 01/18/1991 |
| From: | Conte R, Pullani S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028H878 | List: |
| References | |
| 50-277-90-24OL, 50-278-90-24OL, NUDOCS 9102010010 | |
| Download: ML20028H887 (53) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT Examination Report Nos.: 50-277/90-24 and (OL) and 50-278/90-24 (OL)
Facility Docket Nos.
50-277 and 50-278 Facility License No..
Philadelphia Electric Company Nuclear Group Headquarters Correspondence Control Desk P. O. Box 195 Wayne, Pennsylvania 19087-0195 Facility:
Peach Bottom Atomic Power Station, Unit 2 and 3 Examination Dates:
December 18 - 20, 1990 Examiners:
R. Miller, Examiner, Sonalysts, Inc.
S. Pullani, Senior Operation Engineer J. Williams, Senior Operations Engineer l
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Chief Examiner:
_.. _ lani, Sr. Operations Engineer Date'
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Approved By:
[k kichardJ.Conteghief
/ Date BWR Section, Operations Branch Division of Reactor Safety
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Examination Summary: Written examinations and operating tests were administered to nine Senior Reactor Operator candidates limited to fuel handling (LSR0).
All the nine candidates were previously licensed at Limerick Units 1 and 2 as LSR0s and received waivers that the written and operating portions of the examination will involve " differences" in procedure and design between Peach Bottom and Limerick units.
The written examination was prepared by the facility as previously agreed to by NRC, but was reviewed, administered, and graded by the facility and NRC in parallel.
Both grades were identical. The operating test was prepared, administered, and graded by NRC.
All nine candidates passed the examinations.
Within this report, individual strengths and deficiencies were listed as feedback to the licensed operator training program.
9102010010 910125 PDR ADOCK 0D000277 V
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DETAILS
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1.0 Introduction and Overview The NRC examiners administered written examinations and operating tests i
to nine LSRO candidates.
The examinations were administered in accord-
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ance with NUREG 1021, Examiner Standards, Revision 6.
The draft standard l
ES-701, Administration of Senior Reactor Operators Limited to Fuel Handl-
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ing Fuel Handling (LSRO) Examinations, Revision 6, was used as guidance to the extent possible.
Further, as approved.by the Of fice of Nuclear Reactor Regulation, Operator Licensing Branch, the NRC staff administered a " differences" examination since the candidates were already licensed as fuel handlers at Limerick, Because of similar procedures and designs in fuel handling
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equipment between the two facilities, the examination primarily-tested'
differences between the facilities in the subject areas.
The differences examination also involved the licensee proposing the written portion, subject to approval by NRC staff.
The results of the examinations are summarized below.
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2.0 Persons Contacted l
2.1 Nuclear Regulatory Commission (NRC)
L. Bettenhausen, Chief Operations-Branch (1)
R. Conte, Chief, BWR Section (1) (Part-Time)
T. Easlick, Operations Engineer (1) (part-Time)
R. Miller, Examiner, Sonalysts, Inc. (3)
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S. Pu11ani, Senior Operations Engineer (1), (2), (3)
J. Williams, Senior Operations Engineer (1), (2), (3)
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2 Philadelphia Electric Company (PECo)
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R. Andrews, Supervisor-0perations Traising (1), (3)
D. Filson Instructor-LSR0 (4)
R. Helt, NTS Supervisor (1), (2), (5), (4)
J. House, PECO Operations Training (1)
R. Klemm, Manager, Nuclear Trainirg (1)
R. Krich, PECO Branch Head, Licessing (1)
J. Lytc, Senior Instructor-taltial (3)
S, Mannix, Operations-Siitt Manager (3)
R. Nunez, Supervisor-Operations Training, Limerick (1)
D, McClellan, Senior Instructor-Licensed Operator Requel. (3), (4)
2.2 Philadelphia Electric Company (PECo) (Cont'd.)
K. Patek, Senior Instructor (3)
C. Schwanz, Shif t Manager (4)
G. Stewart, Licensing (1)
E. Till, Superintendent-Training (3), (4)
J. Volz, Health Physics Supervisor (4)
Notes (1) Denotes those present during the meeting in Region I on November 2, 1990, to discuss the scope of the examination (2) Denotes those present during the pre-examination review of the written examination on December 6, 1990 (3) Denotes those present during the entrance meeting on December 18, 1990 (4) Denotes those present dur.ing the exit meeting on December 20, 1990.
l l-3.0 Pre-Examination Activities
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i 3.1 Examination Scope Meeting l
A meeting between NRC and PECO was held'at the NRC Region I of fice at King I
of Prussia, Pennsylvania, on November 2, 1990, to discuss the scope of the
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examination.
During this meeting, it was agreed that the' draft standard ES-701, Administration of Senior Reactor Operator Limited to Fuel Handling (LSRO) Examinations, be used as guidance for the' conduct of the-examination.
3.2 License Application Review The license applications were reviewed in accordance with NUREG 1021, Examiner Standards, Revision 6.
The applications contained sufficient
information to determine the eligibility of the applicants to appear for the examination or to grant the waivers as requested.
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l 3.3 Examination Preparation The written examination and operating test were prepared in accordance with NUREG 1021, Examiner Standards, Revision 6. As noted in Section 1, the written examination was prepared by the facility and the operating test was prepared by NRC.
The reference material provided by the licensee was found adequate for the preparation, i
3.4 Pre-Examination Review Prior to edministration of the written exaniination, on December 6,1990, the NRC reviewed the examination at the Regional Of fice with the facility representative who prepared the examination. All NRC con.ments were discussed and resolved during the review session.
The examination was revised-to incorporate those changes.
The facility individual involved with preparation and the review of the examination materials signed-security
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agreement to ensure that there was no compromise of the examination.
3.5 Entrance Meeting An entrance meeting with the licensee was held on December 18, 1990, at its training facility. The purpose of the meeting was to discuss the plan and schedule for the examination.
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4.0 Examination - Related Findings and Conciusions The following is a summary of the strengths and deficiencies noted during the administration of the written examination and operating tests.
This information is being provided to aid the licensee-in upgrading the LSR0 training programs.
No licensee written response is required.
4.1 Written Examination Strengths The knowledge of the following topics-(the related question number in parathesis) was noted as strengths:
Refueling interlocks (Question 3)
Staff working hour restrictions (Question 4)
Diesel Generator automatic start signals (Question 11)
a Technical Specification requirements for -SRM (Question 12)
LSR0 overlap requirement during shif t turnover (Question 14)
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Limit on maximum number of fuel bundles that can be transferred per
day during initial core off-load or during a fuel reshuffle (Question 17)
The types of radiation that can be measured by DRD and TLD (Question
18)
Suction source for the Shutdown Cooling System (Question 19)
Response of Electrical Distribution System for a LOOP signal
(Question 22)
Locations of CCTAS and the official sign off copy (Question 27)
Methods to control contamination (Question 28)
Isolation signals for Reactor Building HVAC (Qrestion 30)
Access restrictions to Drywell during irradiated fuel movements
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(Question 35)
Emergency exposure limits (Question 36)
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SLC system controls and interlocks (Question 37)
i Normal cooling medium for the FPC system (Question 40)
Technical Specifications for inoperable LPRM (Question 46)
Reason for shaking the grapple while the fuel assembly is going into
the core (Question 49)
Technical Specification Safety Limit - bases concerning the 75 psig
RHR shutdown cooling mode isolation (Question 50)
Definition of TS Surveillance Frequency and its margin (Question 51)
Conditions for terminating fuel handling operations (Questicn 52)
Bases for Technical Specification requirement for minimum water
level in spent fuel pool (Question 53)
Deficiencies l
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The knowledge of the following topics was noted as deficiencies:
HPSW pumps response to high drywell pressure (Question 1)
Purpose and operation of Core Spray Leak Detection system (Question 8)
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Reasons to dis:ontinue the mcVement of fuel or other equipment
(Question 23)
Plant Housekeeping Zones end methods of cleanliness controls
(Question 39)
Effect of loss of instrument air on SLC storage tank level
indication (Question 42)
Service Water pump trip signals (Question 44)
4.2. Operating Test Strengths With respect to the applic'nts' training limited to the differences
between Limerick and Pear' Bottom, the applicants demonstrated good knowledge in administra' ive topics, systems, and procedures irnportant to fuel handling oper-lons at Peach Bottom, except those noted as deficiencies.
During the fuel handling portion of the operating test using dummy
fuel bundles, the applicants demonstrated their ability to transfer fuel from one location to another in the Spent Fuel Pool.
The applicants, in general, demonstrated their knowledge to interface
with maintenance crew and the control room operating crew for proper conduct of the interfacing activities during refueling operatiors.
Deficiencies During the performance of' Job Performance Measures (JPMs), the appli-a ants generally exhibited difficulty in locating certain system valves or instrumentation.
The systems were:
Reactor Building HVAC, Reactor Building Closed Cooling Water, and Fuel Pool Cooling. The root cause of this deficiency appears to be inadequate training in performing JPMs associated with systems important to fuel handling.
The licensee indicated that LSR0s are not normally responsible for these types of job functions.
However, the license plans '.o provide the required remedial training to conform to NRC expectations in the Examiner Standard (ES-701), prior to January 10, 1990.
5.0 Exit Meeting An exit meeting was conducted on December 20, 1990, following the admi-nistration of the examination.
The licensee representatives that attended the meeting are listed in Section 2. of this report.
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The generic strengths and deficiencies noted on the operating examinations were presented (see'Section 4.2 of this report).
General observations by the examiners while administering the operating tests were discussed.
The chief examiner stated that the results of the examinations would not be presented at the exit meeting, but would be contained in the examination report and that every effort would be made'to send the results in approxi-mately 30 working days.
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Attachment:
LSR0 Written Examination and Answer Key E
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Attachment i
LSRO Written Exaaination and Answer Key-
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PHILADELPHIA ELCCTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION DELTA, PENHSYLVANIA 17314
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DOCTYPE #:
346 DOCUMENT TITLE: OPERATIONS OUIZ - PBAPS Peach Bottom Atomic Power StationJTm No-LSRO Dif ferences ExnminnH nn (LECTURE NO). _ m _
(List separately)
QUIZ TITLE:
i LIMITED SENIOR REACTOR OPERATOR
_ LSRO CODE:
TYPE:
- /d******************************
PAYROLL #
- NAME HI (Print)
Last First
SOC. SEC NO. _
- SIGNATURE
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I have neither given nor received
All work on this. quiz is my own.
assistanco.
SCORE f
- DATE
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Approved by: / -
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Date:
It;STRUCTIIDN TO CANDIDATE:
l indicated in paratheses after the quest an.
Points for each quest *an a*:
To pass this exami"stion, you must achieve an overall grade of at least 80%.
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Examinati,n pape.s will be picked up three (3) hours af ter the exataination starts.
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l-l HRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
During the' administration of this examination the following rules apply:
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Cheating on the examination means an automatic denial of your application 1. and could result in more severe penalties.
After the examination has been completed, you must sign the statement on 2.
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the cover sheet indicating that the work:is your own and you have not
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received or given assistance in completing the examination.
This must be
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done after you complete the examination.
l-Restroom trips are to be. limited and only one candidate 'at,a' time :may
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3.
f You must avoid all contacts with anyone outside the' examination leave.
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room to avoid even the appearance or possibility ofLcheating.
Use black ink or dark pencil only to facilitate 1cgible reproductions.-
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Print your name in the blank provided in the upper,right-hand corner of S. the examination cover sheet.
Fill in the date on the cover sheet of the examin'ation (if necessary).
6.
You may write your answers on the examination-question page or on a 7.
separate sheet of. paper.- USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON
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THE BACK SIDE OF THE PAGE.
If you write your answers on the examination question :page and -you need 8.
more space to answer'a-specific question, use a separate sheet of-the _
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paper-provided and insert it directly after-the specific question.
DO NOT
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WRITE ON THE BACK SIDE OF THE EXAMINATION QUESTION PAGE.
Print your name in the upper right-hand-corner _of the first page_of answer 9. sheets whether you use the examination question pages or separate: sheets of paper.
Initial cach of-the following answer pages.
10.-Before you turn.in your examination, consecutively number each' answer sheet, -including any a' ditional pages inserted when writing your answers.
d on the examination question page.
If you are using separate sheets, number each an'swer and skip at _least f 3 11. lines between answers to allow space for_ grading.
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12. Write "Last Page" on the last-answer sheet.
13. Use abbreviations only if duey are commonly used in facility-literature.
Avoid using symbols such as < or > signs to avoid a simple transposition t
l error !resulting in an incorrect answer.
Write it out, l
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for each question is indicated in parentheses after the 14. The point valueThe amount of blank space on an examination question page is question.NOT an indication of the depth of answer required.
Show all calculations, methods, or assumptions used to obtain an answer.
15.
16. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION
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NOTE: partial credit will NOT be AND DO NOT LEAVE ANY ANSWER BLANK.
given on multiple choice questions.
Proportional grading will be applied.
Any additional wrong information For example, if a question is 17. that is provided may count against you.
each of which.is worth 0.25 l
worth one point and asks for four responses,each of your responses will be worth points, and you give five responses,If one of your five responses is incorrect, 0.20 will be 0.20 points.
deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.
intent of a question is unclear, ask questions of the examiner 18. If the caly.
19. When turning in your examination, assemble the completed examination with In addition, examination questions, examination aids and answer sheets.
turn in all scrap paper.
To pass the examination,.you must achieve an overall grade of 80% or 20.
greater.
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There is a time limit of (
) hours for completion of the examination.
(or some other time if less than the full examination is taken.)
21.
When you are done and have turned in your examination, leave the examin-ation area as defined by the examiner.
If you are found in this area 22.
while the examination is still in progress, your license may be denied or revoked.
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SENIOR REACTOR OPERATOR Page 1 of 22 l
QUESTION: 001 (1.00)
Assume Drywell pressure increases to 2 psig and liigh Pressure Service Water; (llPSW) pumps are running.
Which of the following describes how the llPSW pumps respond to the high drywell pressure.
a.
The llPSW pumps continue to run.
The Emergency Bus load shed is initiated on low reactor water level (minus 160"),
b.
The ilPSW pumps continue to run.
The llPSW pump trip on Emergency Bus low voltage, c.
The llPSV pumps trip. The Emergency Bus-load shed is initiated on liigh Drywell Pressure (2 psig),
d.
The llPSW pumps trip and will auto restart when the Diesel starts.
t QUESTION:
002 (1.00)
Which of the following statements best describes Double verification in accordance with A 42 " Control of Temporary Plant Alterations" (TPA) ?
a.
Includes reinstallation of removed or disturbed items and removal of temporary installations to arrive at the correct configuration and includes inspection of the locations (s) to assure that there was no resulting damage, b.
The process by which two individuals simultaneously verify the affected component or circuit is the correct one specified by-the TPAF, prior to disconnecting any leads or installing any jumpers.
c.
The process by which the TPAF is reviewed by PORC and verified by the TPA Installer prior to disconnecting any leads or installing any j umpers,
d.
An act of ensuring that a condition conforms to the specified requi.rements, accomplished by an individual (s) working. independently of the person (s) performing the activity.
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SENIOR REACTOR OPERATOR Page 2 of 22 f
QUESTION:
003 (1.66; Concerning ST 12.1 "Retuoling '.oterlock Functional Test", when the main fuel-
. grapple is not in use, whou pssition mustLit be in?-
a.
full down b.
full up c.
full down and deenergized d.
locked in place QUESTION:
004 (1.00)-
Concerning site staffing working. hour restrictions, complete the following:
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"An individual shall not be permitted to work more than hours in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than.
hours in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period nor_more than hours in any 7 day period, all excluding shift turnover time, s
a.
12, 24, 70 b.
16, 24, 70 c.
16, 24, 72 d.
12, 30, 72 QUESTION: 005 (1.00)
When the fuel bundle is loaded into the core, and is properly positioned,-what reactor vessel component provides lateral alignment for the bottom end of the fuel assembly?
a.
Control Rod Drive Housing Support b.
Baffle Plate c.
Control Rod Guide Tube (Upper End)
d.
Orificed Fuel Support-
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SENIOR REACTOR OPERATOR Page 3 of 22 l
QUESTION:
006 (1.00)
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ALL of the following_are applicable to temporary changes to refueling procedures-EXCEPT:
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The intent of the original procedure is not altered.
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The change is approved by at_least two members of the plant management
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staff, who hold a SRO license.
The change is reviewed by PORC prior to implementation.
c.
d.
The' change is reviewed and approved by the-Plant Manager within 14 days of implementation.
QUESTION:
007 (1.00)
Refueling operations are in progress. The Refueling Floor SRO is informed that one train of Standby Gas Treatment has become inoperable. Which of the following is the correct evaluation of the Secondary Containment system?
a.
Secondary Containment integrity'has been lost, and suspension of the handling of irradiated fuel, core alterations and operations with the potential for draining the-reactor vessel is required, b.
Secondary Containment has not been lost, all refueling operations may continue, since only one train of Standby Gas' Treatment is required L
during refueling operations.
Secondary Containment has been lost, but handling of irradiated fuel, c.
core alterations and operations with potential for draining the reactor vessel may continue for seven (7) days, d.
Secondary ( mtainment has not been lost, handling oi irradiated fuel, core alterations and operations with-the potential for draining the
reactor vessel may continue for seven (7) days.
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SENIOR REACTOR OPERATOR Page 4 of 22 QUESTION: 008 (1,00)
Concerning the Core Spray Leak Detection System, ALL of the following statements are correct EXCEPT:
The purpose is to detect a Core Spray Piping break between the inside a.
vessel wall and outside shroud wall.
b.
Differential pressure indication will be zero during cold shutdown and will become negative at rated power.
The purpone is to. detect a core spray piping break between the inside c.
shroud wall and core spray sparger, d.
Control Room indication is by alarm only.
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QUESTION: 009-(1.00)
Which of the following defines a " Level I LHRA" (Locked High Radiation Area)?
a.
Area less than 10 R/hr and no potential for extremely high exposure rates b.
Area less than 1000 mr/hr and no potential for extrenely high exposure rates Area greater than 10 R/hr where there is a potential for extremely high
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exposure rates d.
Area less than 1000 mr/hr where there is a potential for extremely high exposure rates l
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QUESTION:
010 (1.00)
According to SE-12. " Injury Response When Site Is Not In A Classified Emergency", which of the following statements is nat an immediate ' operator action?
Chief Operator make Public t.ddress and radio announcement for " Medical a.
Response Team" to report to requested location and repeat several times.
b.
Establish First Aid / Search and Rescue Group.
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Obtain additional information about injured from reporting source.
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tiotify Shift Management.
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SENIOR REACTOR OPERATOR Page 5 of 22 l
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011 (1.00)
Concerning the 4 KV system and the Emergency Diesel Generators, ALL of the following statements are correct EXCEPT:
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Diesel Generators automatically start on -160"Rx-water level, a.
b.
Diesel Generators automatically start on 2 psig Drywell pressure.
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Diesel Generators automatically start on a Dead' bus, and Feeder Breakers.
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Diesel Generators automatically start 0" Rx water level and 1.68 psig Drywell pressure.
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QUESTION:
012 (1.00)
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As the LSRO during a PBAPS Unit 2 refueling outage, your shif t has completed the core off-load to the spent fuel pool. Twenty (20) days later, you are assigned to the first shif t of core on-load.
The on-shift ~ Reactor Operator informs you that the SRM System channel readings are as follows:
Channel A - 2.0 CPS Channel B - 1.3 CPS Channel C -
.9 CPS f
Channel D - 1.1 CPS As a result of this situation, WHICH ONE of the-following best describes the
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Technical Specification requirements?
Core on load cannot commence until " dunking" chambers are installed with a.
at least two of the appropriate SRM channels are reading at least 3 CPS.
b.
SRM operability requirements do not apply, since all control rods are inserted and the shutdown margin (SDM) specification is uredicted to be'
met, SRM operability requirements do not have to be met until a sufficient c.
number of fuel assemblies are loaded into the core to produce 3 CPM on at least two of the appropriate SRM channels, d.
SRM operability requirements do not have to be met until a maximum of sixteen fuel assemblies have been loaded into their original core'off-load position around the associated SRM detecto.....
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SENIOR REACTOR OPERATOR-Page 6 of 22 QUESTION: 013 (1.00)
According to A-64 Fuel and Special Nuclear Material Accounting and Safeguards -
Directive, all of the following are Material Balance Areas for fuel EXCEPT:
i a.
Unit 2, Fuel Pool b.
Unit 2, Reactor c.
Unit 2, New Fuel Vault d.
Unit 2, Fuel Floor QUESTION:
014 (1.00)
The LSRO shall be responsible as the Fuel Handling Supe'rvisor for an overlap during a turnover for at least move (s), or until confident of the situation, prior to the relieved individual leaving, n.
b.
c.
d.
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015 (1.00)
According. to GP 20, Temporary Defeat of Core Spray and RHR Pump and Diesel Generator Auto Starts During.CRD Exchange and _ core reload, all of the
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following statement (s) is/are correct concerning this1 procedure EXCEPT:
allows an RilR Pump to be run in Shutdown Cooling mode a.
b.
allows an RHR Pump to be run in TORUS cooling mode c.
allows the RilR Pump to be bumped for rotation-
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d.
allows stroking RilR suction valves for maintenance l
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SENIOR REACTOR OPERATOR-Page 7. of 22 i
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QUESTION: 016 (1,00)
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Refuel floor, Hose stations, Carbon hexafloride extinguishe.s, dry chemical n.
extinguishers.
b.
Hose stations, foam, dry chemical extinguisherc.
Dry chemical extinguishers, hose stations, halon c.
Dry chemical extinguishers, hose stations, carbon dioxide extinguishers d.
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QUESTION: 017 (1.00)
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The LSRO shall be responsible as the Fuel Handling Supervisor for maintaining a maximum limit of bundles per day being transferred to the fuel pool during the initial core off-loadaor during a fuel shuffle.
a.
150 b.
160 l
c.
170 d.
180 OUESTION: 018 (1.00)
,
The pocket dosemeter (DRD) and TLD badges worn by personnel in the power block
-
are designed to measure radiation. What types of radiation does each measure?
f.
a.
TLD - Gamma and alpha DRD Gamma and beta b.
TID - Gamma and beta DRD Gamma only c.
TLD - Gamma and beta DRD - Beta only d.
TLD - Gamma and fast neutrons
'DRD - Beta only f
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SENIOR REACTOR OPERATOR Page 8 of 22 i
QUESTION:
019 (1.00)
Where is the suction for Shutdown Cooling from the recirculation system?
"B" recirculation loop upstream of-the recirculation pump suction valve, a.
b.
Either recirculation loop via the LPCI injection lines.
t
"A" recirculation loop upstream of the recirculation pump suction valve,
,
c.
d.
"B" recirculation loop downstream of'the recirculation pump suction l
valve.
QUESTION:
020 (1.00)
Regarding Standby Gas Treatment System, all of the foll'owing are functions of the SBCTS EXCEPT:
Used for backup to maintain (negative)
.25 inches water gauge in the a.
Reactor Building
l b.
Provides a means for venting Primary Containment-Assists in purging /inerting operations of Primary Containment f
c.
d, Used for leak rate testing of Primary Containment QUESTION:
021 (1.00)
Uhich of the following statements does-not describe the.' term Constant. coverage-as it applies to li.P. practices?
The ll.P.' Technician accompanies'and-is in close proximity to the' worker l
a.
i b.
The ll.P. Technician provides coverage such that he is aware of the-
-workers activities The '11.P. Technician maintains line of sight and audio communications c.
with the workers
'
d.
Surveillance is provided using a TV camera and dedicated communication link.
-
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.
.
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.
.
.
.
.
SENIOR REACTOR OPERATOR Page 9 of 22 l
l l
QUESTION:
022 (1.00)
ALL OFF-SITE Power is Lost. The (4Kv) Emergency Bus voltage has decreased to less than 25% voltage.
Which of the following is the correct 4Kv system response?
a.
The alternate feeder breaker auto closes, the diesel starts and the diesel generator output breaker closes on the-emergencj bus,
'
b.
The alternate feeder breaker will not close, the diesel starts, and diesel generator output breaker closes on the emergency bus.
c.
Tbc normal feeder breaker will trip, the diesel starts, and the alternate feeder breaker closes on the emergency bus.
d.
The normal feeder breaker will trip, the diesel starts and the normal feeder breaker closes when power is restored.
,
QUESTION:
023 (1.00)
Which of the following, in accordance with FH-35 " Control of Material Movement in the Fuel Pool" is not a reason to discontinue the movement of fuel or other equipment?
a.
Failure or inoperability of any equipment or components of the refueling platform, b.
A drop in fuel pool water level below Elevation 232'.
c,
. Loss of Secondary Containment, Containment Isolation System or Main Control Room Normal Ventilation System.
d.
The dropping or damaging of a spent fuel assembly during the handling operations.
.
l
.,
_ - _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
_
,
.
I I'
l SENIOR REACTOR OPERATOR Page 10 of 22 l
[
QUESTION:
024 (1.00)
'
When tne reactor is suberitical and the reactor water temperature is less than 212 degrees F, in accordance with Technical Specifications which (one) 1 of the'following trip functions is not required to be operable?
,
a.
Mode switch to " Shutdown" b.
Reactor low water level.
l c.
High flux IRM d.
Scram Discharge Instrument volume high leval QUESTION:
025 (1.00)
PBAPS Technical Specification Surveillance Requirements 4.3. A.1 (Reactivity Margin - core Loading) states:
Sufficient control rods shall be withdrawn following' a refueling outage when core alterations were performed to demonstrate a margin of 0.38%
l delta K/K that the core can be made suberitical.
l The Technical Specificatien bases (Reactivity Limitation) state that the,
!
reactor is suberitical by at least (E 0.38% delta K/K)
l l
Which of the following define the value of E?
a.
The difference between the calculated value of maximum core reactivity during the operating cycle and the calculated beginning-of-life core reactivity.
b.
The difference between the actual value of maximum core reactivity during the operating cycle and the calculated end-of-life core reactivity.
c.
The difference between the actual value of maximum core reactivity during the previous operating cycle and the calculated'end-of-life core i
reactivity.
d.
The difference between the calculated value of maximum core reactivity
l during' the refueling cycle and the calculated beginning-of-life core reactivity.
I
,
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l SENIOR REACTOR OPERATOR Page 11 of 22
)
!
QUESTION:
026 (1.00)
ALL of the following are correct concerning the Intermediate Range monitoring system EXCEPT:
Detect and indicate neutron flux in a range between the SRM and power a.
l range detection capability, i
b.
Generate trip signals to prevent fuel damage from' single operator error or single equipment malfunction, i
Detect between 10 8 and 10 3% of rated power, c.
d.
Provide a backup to-the APRM SCRAM 1SSS for fuel cladding-integrity.
.
QUESTION:
027 (1.00)
FH-6C (Fuel Movement and Core Alteration Procedure During a Fuel Handling Outage) prerequisites 5.17 states:
copics of the applicabic procedures FH-6C Core Component transfer Authorization Sheet (CCTAS) Appendix 1 through 7 are provided at the following locations
2
,
,
with the
copy as the official signoff copy.
a.
1) Control Room, 2) Refueling Bridge, 3) Shif t Supervisors Desk: 4)
j Control Room l
b.
1) Control Room, 2) Refueling Floor, 3) Shift Supervisors Desk-4)
Shift Supervisors Desk c.
1) Refueling Bridge, 2) Refueling Floor, 3) Control Room:
4) Refueling Bridge d.
1) Refueling Bridge, 2) Refueling Floor, 3) Control Room:
4) Refueling Floor
l l
-
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____
.
.
.
SENIOR REACTOR OPERATOR Page 12 of 22 QUESTION:
028 (1.00)
Concerning Fuel Handling Operations, if a fuel bundle must remain in the fuel l
prep machine unattended, two requirements must be met.
These two requirements
,
are:
The fuel prep machine must be in the fully lowered position; the air a.
supply must be disconnected.
b.
The fuel prep machine must be in the fully raised position; the air supply must be disconnected, The fuel prep machine must be'in the fully lowered position; the power c.
supply must be disconnected, d.
The fuel prep machine must be in.the fully raised position; the power supply must be disconnected.
.
QUESTION:
029 (1.00)
rocedure FH 74 Actions in Response to an Unexpected Loss of Fuel Fool, Reactivity Cavity, or Equipment Storage Pool Water Inventory has four (4)
one (1) cf the four (4) initiating events is also an
,
(
entry condition to T-103 Secondary Containment Control.
Select the Initiation Event that requires entry to T-103 Secondary Containment control.
a, Reactor water level decrease of one (1) foot or more for no known reason, with the RPV head removed.
b.
Fuel Pool 1cvel decrease of one '(1) foot or more for no known reason, An unexpected or unexplained Fuel Floor Area Radiation Monitor (ARM)
c.
alarming.
d.
An unexpected or unexplained vent stack high-high radiation alarm.
!
l-
.
..
.
-
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-
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.
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.
'
SENIOR * REACTOR ~0PERATOR-Page 13 of 22 l
QUESTION: 030 (1.00)
Concerning the Reactor Building Exhaust Radiation monitors, ALL of the'
l following will cause a Reaccor Building Ventilation Isolation EXCEPT:
Exhaust duct high radiation - 16 mr/hr a.
b.
Low reactor water level
"0" inches High Drywell pressure - 2 psig
'
c.
d.
Exhaust duct radiation monitor inoperable l
QUESTION:
031 (1.00)
.o
-
Concerning the Refuel Floor Ventilation System, an operator is directed place the refuel floor ventilation system in-service according to s
S0 40B.1.A-2.
After starting two refuel floor ventilation supply fans, he leaves the area to answer the page. Assume the refueling floor hatch is installed and no operator action. ALL of the following is/are correct 'EXCEPT:
Refuel floor internal pressure would increase and become positive a.
b.
Supply fans should trip at +.5 inches water gauge c.
Exhaust fans should auto start at +.5 inches water gaugo (
l d.
Refuel floor blowout pancis blowout l
-
QUESTION:
032 (1.00)
All of the following are entry conditions for T-103 Secondary Containment Control EXCEPT:
Unexplained Reactor Building Ventilation Exhaust High Radiation ALARM a.
l b.
Uaexplained Secondary Containment ARM ALARM Unexplained Refuel Floor Ventilation Exhaust lligh Radiation ALARM c.
d.
Unexplained Vent Stack High Radiation A1 ARM
,-
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.
SENIOR REACTOR OPERATOR-Page 14 of 22 QUESTION: 033 (1.00)
In regards to e Design Basis Refueling Accident, which (one) 1 of the
'ollowing statements is correct?
Design basis accident is irradiated fuel uncover a.
b.
Reactor Building Ventilation should isolate No offsite release is anticipated c.
d.
Standby gas treatment will remove most of the activity from the noble gases.
QUESTION: 034 (1.00)
HP-321, " Health Physics Requirements for Irradiated Fuel Transfers", defines the radiological controls for the fuel floor and drywell during irradiated fuel navement.
Which (one) 1 of the following is not a p.erequisite for HP 3217 The Area Radiation Monitor on the fuci bridge shall be in operation.
a.
'
b.
With the Fuel Shuttle Shield removed no access shall be allowed to elevations above the 163'. elevation of the Drywell.
Radiological monitoring and communications capabilities shall be c.
utilized, d.
The Drywell Control point,.lealth Physics Technician, shall post all ladders with signs prohibiting access =to Drywell clevations.above the 163' elevation.
I i
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.
.
i SENIOR REACTOR OPERATOR-Page 15 of 12
i Qt'ESTION: 035 (1.00)
,
The Emergency Exposure limits for Protection of Health and Safety of the public are:
.
Whole Body Thyroid j
Dose Dose a.
b.
25 Rom 125 Rem c.
5 Rem 25 Re. i
.
QUESTION: 036 (1.00)
The RO is directed to place the Standby Liquid Control System Switch to the pump "B" position.
ALL of the following are correct EXCEPT:
a.
SLC pump "B" starts b.
SLC System Squib primers "A" and "B" in explosive valves fire c.
RWCU System isolates d.
Only "B" Squib primers in
"B" explosive fire QUESTION:
037 (1.00)
Procedure A 30, "Plar.: llousekeeping Controls", designates.the refuel' platform as Housekeeping Zono If refuel-platform was determined to be
.
a Hot Particle Zone.
The area would be posted with "
Zone"-
inserts, n.
I, Red b.
II, Yellow c.
I, Yellow d.
II, Red
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SENIOR REACTOR OPERATOR Page 16 of 22 l
QUESTION:
038 (1.00)
.
!
j Which of the following is the back-up Cooling Medium for the Fuel Fool Cooling i
System?
Emergency Service Water (f"O
!
n.
,
I b.
Turbine Building Closed Cooling Water (TBCCW)
c.
'
d.
Reactor Building Closed Cooling Vater (RBCCW)
l l
QUESTION:
039 (1.00)
-
What is the effect of a loss of instrument air on the Standby Liquid Storage Tank level indication?
a.
Indicated level vill increase b.
Indicated IcVel will not change
!
c.
Indicated level will be erratic i
d, Indicated level will decrease QUESTION:
040 (1.00)
Any accessibic area in which a major portion of the whole body would receive greater than 100 mrem in any one hour is the definition of:
_.
.a.
Radiation Arca-b.
liigh Radiation Area e,
controlled Area d.
Locked liigh Radiation Area -
l
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SENIOR REACTOR OPERATOR Page 17 of 22 l
QUESTION:
041 (1.00)
,
i Assume the Emergtecy Cooling Water System must be placed in service and the sluice gates control switches places in " lower" position. Which of the
.
i following statements is/are correct?
'
Service Water pumps receive a trip sig"al a.
b.
Bay level controllers will begin controlling level a
ESW wash to pond valve will close c.
i d.
All screen wash pumps receive a start signal
.
QUESTION:
042 (1.00)
With the PBAPS Unit 2 core approximately half on-loaded, the Refuel Floor Supervisor informs you that one LPRM string has been determined to be inoperable and must be replaced. The following conditions exist:
RHR B, C and D are in various stages of maintenance and are unavailable.
,
l
- Core spray A and C are currently inoperable due to motor rebuilds.
The spent fuel pool gates are removed and reactor,well/ spent fuel pool water levels are in specification.
In accordance with Technical specifications, WHICH ONE of the following most accurately describes what is allowed?
The LPRM detector string may be replaced but the core off-load must be a.
suspended.
b.
Core on load and the LPRM defector string replacement may-be accomplished simultaneously. No restrictions apply, Core on-load may c ntinue but the LPRM defector string replacement may c.
not be accomplished until core spray A and C are restored to operable, d.
Under the present conditions, a complete core off load must be accomplished to replace the LPRM detector string.
l
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SENIOR REACTOR OPERATOR Page 18 of 22 QUESTION:
043 (1.00)
ALL of the following are purposes of the Reactor guilding Ventilation system EXCEPT:
f Pretect personnel and equipment from contamination, a.
b.
1 solation of the Secondary Containment to prevent the gross release of airborne radioactivity.
l Protect personnel and equipment from extremes in temperature c.
d.
Isolation of the Standby Cas Treatment System to_ prevent uncontrolled radioactivity releases under accident conditions.
.
QUESTION:
044 (1.00)
During the initial phases of PBAPS Unit 2 core off load, the control room SRO relays the following information:
The reactor mode switch is locked in the REFUEL position.
-
4% *lt' K/".
D Core shutdown margin is calculated to be
-
.
One control rod (in a core quadrant where core alterations are not
-
l taking place) is withdrawn to position 12.
The Standby Liquid Control System operability surveillance has been
-
failed due to insufficient sodium pentaborate solution.
'
All SRM channels are operable.
-
Because of this situation, VHICH of the following statements best applies?
- a.
Core off-load may continue, in compliance with Technical Specifications, b,
Core off-load must be suspended and all control rods inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Core off-load must be suspended the control rod should poi be vithdrawn c.
to position 12.
d.
Core off-load may continue, Standby Liquid Control must be operable
?
within 7 days.
,
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SENIOR REACTOR OPI.RATOR Page 19 of 22
!
!
!
QUESTION: 045 (1.00)
f The refuel platform operator max shake the grapple when the fuel assembly is going into the core:
this will help prevent the fuel assembly from hanging up on the or
.
a.
Core plate; spring clip b.
Upper grid; blade guide
,
<
Core plate; blade guide c.
d.
Upper grid; spring clip
3 Qb6STION: 046 (1.00)
According to Technical Specification Safety Limit bases' concerning the 75 psig I
RHR shutdown cooling mode isolation,.the reason for this limit is:
To prevent a loss of-coolant accident because of a failure of the RHR a.
- pump seals.
b.
To prevent the rupture of the reactor coolant system boundary while operating in the shutdown cooling mode,
/To preclude further RPV cooldown using the RHR system when sufficient c.
steam pressure / flow is available.
.
d.
To prevent excessive radioactive release from the RilR heat exchangers
'
during periods of high suction / discharge pressures.
I
'
QUESTION:
047 (1.00)
Concerning the Technical Specification definition of the term " Surveillance Frequency", complete tha following:
"The-specified surveillance intervals may be exceeded by
, and surveillance tests are not required on systems or l
parts of systems that are
"
n.
25%, in operation b,
25%, tripped c.
3.25%, in operation d.
3.251, tripped l
l-l
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SENIOR REACTOR OPERATOR PaSe 20 of 22
i i
f QUESTION:
048 (1.00)
l In accordance with FH 6C " Fuel Movement and Core Alterations Procedure Duri.ng l
a ruel Handling Dutage", step 9.6 ' Limitation to continued Hovement":
Which of the following DOES EDI require the terminations of fuel handling operations.
!
" Rod Block Interlock #1" light is lit (on) with the refueling platform
!
a.
over the core and the main grapple not full up,
b.
The loss of communications between the fuel floor and the control room.
A single fuel rod is dropped during bundle reconstitution in the fuel
'
c.
pool.
d.
Secondary containment is inoperabic.
.
QUESTION:
049 (1.00)
i The Technical Specification requiring that a minimum uater level must exist
over the top of irradiated fuel seated in the spent fuel storage pool racks is
!
based on all of the following EXCEPT:
Provides adequate water volume required for High Density fuel storage a.
racks.
b.
Provides adequate radiation shiciding for refuel floor personnel, Provides adequate cooling for. decay heat from irradiated fuel.
j c.
d.
Provides adequate iodine level reduction in the event of a. fuel handling accident.
(
,
l l
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l i
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.
SENIOR REACTOR OPERATOR Page 21 of 22
i QUESTION:
050 (1.00)
During refueling cf Unit 2 a fuel assembly is dropped in the spent fuel pool.
Whose permission (by title) must be obtained to recommence fuel handling?
a.
Plant Manager
b.
Superintendent of Operations c.
Vice-President Nuclear d.
Shift Manager
QUESTION:
051 (1.00)
Procedure SO 18.7. A 3 " Transferring Fuel From the Fuel'i'ool to the Reactor",
Step 4.20 visually verify fuel bundle proper orientation prior to insertion.
Which one of the following is an indication of improper fuel orientation?
Spring clip is away from the center of the control rod.
a.
b.
Verify tab on the bail handle points to the control rod, c,
Verify the 2 buttons on the channel face the control rod.
d-.
Verify the serial number on the bail is right side up as viewed from the control rod.
QUESTION: 052 (1.00)
As a Fuel Handling Supervisor, you must be aware of operations which are considered a " Core Alteration". Which of the following is EQI considered a core alteration, n.
Movement of control rods using the CRDH System with the RPV head removed.
b.
Removal of a fuel support piece with fuel in the vessel, c.
Removal of a LPRM detector string with fuel in the vessel.
!
d.
Removal of a control rod blade with the refueling platform with fuel in the vessel.
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SENIOR REACTOR OPERATOR Page 22 of 22
'
QUESTION: 053 (1.00)
According to 1P-16, Breaching and Establishing Secondary Containment, all of the following statement (s) is/are the purpose EXCEPT:
Provides a checklist to facilitate establishing Secondary Containment a.
after it has been breached.
b.
Following a breach authorization, provides a log to track all work when could prevent-establishing second containment.
Following breach authorization provides Suidelines to facilitate c.
irradiated fuel movements.
,
d.
Provide a checklist to ensure compliance with Technical Specifications prior to establishing Secondary Containment.
.
QUESTION: 054 (1.00)
'
Which of the following is not a type of Radiation Work Permit (RWP)?
a.
Special RWP
!
b.
Specific RWP i
'
c.
General RWP i
d.
Standing RWP l
QUESTION: 055 (1.00)
The Refueling Bridge Area Radiation Monitor (ARM) utilizes.a detector for radiation detection.
'
a.
-Proportional b.
Ionization c.
Scintillation d.
Geiger Mueller l
l
'
l END OF EXAMINATION I
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SENIOR REACTOR OPERATOR Page 1 of. 14
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,
ANSVER REFERENCE ANSWER: 001 (1.00)
c
!
REFERENCE:
LSRO 0400, Obj. 5 (page - 9)
KA 295010 AK2.05 3.8 ANSWER: 002 (1.00)
b
REFERENCES: LSRO-1510, Obj. 2.1. (page 8)
A-42 Rev 15 (page 4)
RA 294001 A1.01 3.4 l
l ANSWER: 003 (1,00)
b REFERENCE: LSRO 004, Obj. 9 (page 8)
ST 12.1 Rev 11 (page 3)
l l
KA 234000 K3.03 3.8 A3.02 3.7 A2.01 3,7 ANSWER: 004 (1.00)
c REFERENCE:
PBAPS Technical Specifications, 6.20.2.2 LSRO-1570, Obj. 2g, (A-40, pages 2, 3)
KA 294001 A1.03 3.7
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SENIOR REACTOR OPERATOR-Page 2 of 14
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ANSWER REFERENCE-i l -
ANSWER:
005 (1.00)
d
- .
REFERENCE: LSRO 0010, obj. 5,13, (page 42)
ANSWER:
006 (1.00)
c
.
REFERENCE:
PBAPS Technical Specifications, 6.8.3 i
KA 294001 A1.01 3.4
-
A1.02 4.2
(
ANSWER:
007 (1.00)-
i d
'
REFERENCE:
PBAPS Technical Specification, 3.7.13.3 (page 175a)
LSRO 1840, Obj 4 KA 290001 SG.5 4-2-
.
290001 Kl.04 3.9 i
l
.
ANSWER:
008 (1.00)
l c
l l
REFERENCE:
LSRO-0350, Obj. 6 (page 10)
KA 290001~K1.13 3.0
.
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SENIOR REACTOR OPERATOR Page 3 of*14 l
.
ANSWER REFERENCE ANSWER: 009 (1.00)
,
i a
REFERENCE:
.
LSRO 1730, Obj.10 (page 6)
MA 294004 K1.04 3.B
.
ANSWER: 010 (1.00)
b
.
REFERENCE:
.
LSRO 1555, Obj 3, (SE 12, page 1)
'
KA 294001 Kl.03 3.8
!
!
ANSWER:
011 (1.00)
,
d
'
i
'
REFERENCE:
LSRO 0660, Obj. 4d (page 13)
l
!
KA 262001 E4.08 3.7
'
.
ANSWER:
012 (1.00)
d REFERENCE:
PBAPS Technical Specifications, (L.C.O. 3;10.B.2)_
LSRO 1840, Obj. 4, SRO KA 215004 SG5.0 _. _.. _ _ _
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SENIOR REACTOR OPERATOR Page 4 of 14 ANSWER REFERENCE ANSWER:
013 (1.00)
e REFERENCE:
LSRO 1570, Obj. 2k, (A.44, pages 5, 8)
KA 294001 Kl.03 3.8 v
ANSWER: 014 (1.00)
a
.
REFERENCE: LSRO 002, Obj. 4, (page 20)
KA 234000 SG.1 3.8 SG.13 3.3 e
I ANSWER:
015 (1.00)
b REFERENCE:
LSRO-1530, Obj. 3, (GP-20, page 1)
KA 294001 A1.02 4.2
!
l l
ANSWER:
016 (1.00)
d REFERENCE:
LSRO 0685, Obj. 4, (PF-57, page 1)
KA 294001 Kl.16 3.8 KA 286000 SG.9 3.8 KA 286000 P.4.01 3.6 FA 286000 K3.01 3.4
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Page 5 of 14 SENIOR REACTOR' OPERATOR ANSk'ER REFERENCE
ANSWER: 017 (1.00)
a i
REFERENCE:
LSRO-002, Obj. 7, (page 19)
FA 234000 SG.1 3.8 K3.03 3.8 ANSWER: 018 (1.00)-
b
.
REFERDJCE:
LSRO 1780, Obj. 3, (pages 7,11)
KA 294001 Kl.03 3.8 ANSWER:
019 (1.00)
C REFERENCE:
LSRO-00' 30, Obj. 3. f, (page' 40)
KA 202001 K3.08 2.9 ANSWER: 020 (1.00)
-- -
d
^
REFERENCE:
LSRO 0210, obj.1, (page 4)
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SENIOR REACTOR OPERATOR Page-6 of 14 ANSWER REFERENCE
.
s ANSWER: 021 (1.00)
b REFERENCE:
LSRO 0007, Obj. 7d, (Page 9)
KA 294001 K1.03 3.8
-
i ANSWER: 022 (1.00)
b
,
REFERENCE:
- ,
LSRO 0660, Obj. 4, (page 13)
KA 262001 A1.03 3.1
.
i i
ANSWER:
023 (1.00)
a REFERENCE:
LSRO 0002, Obj. 8,-(page 24, g)
KA 234000 K4.04 3.3 l
A1.01 3.4 l
.
ANSWER:
024 (1.00)
b REFERENCE:
LSRO-1530, Obj. 4, (T.S. Table 3.1.1)
MA 212000 SG.15 4.7
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SENIOR REACTOR OPERATOR Page 7 of 14
'
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ANSWER REFERENCE i
,
i ANSWER:
025 (1.00)
i
!
a i
REFERENCE: Technical Specification bases (TS 4.4. A.1, Page 106)
ISRO 1820, Obj. 2 FA 290002 K5.05-3.9 i
.01 -3.3 ANSWER:
026 (1.00)
C I
REFERENCE:
LSRO.0250, Obj.1 (page 5)
ANSWER:
027 (1.00)
.
d REFERENCE:
LSRO 0002, Obj. 3.a (page 4, FH 60)
.3 4.5 ANSWER: 028 (1.00)
a REFERENCE:
r LSRO 0002, Obj. 8, (page 24g)
-
KA 234000 SC.1 3.8 l
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SENIOR REACTOR OPERATOR Paga 8 of 14
ANSWER REFERENCE
i ANSWER: 029 (1.00)
e REFERENCE:
LSRO.1560, Obj. 3, (page 14).
LSRO 0002, Obj.12, (Fit 74, page 2)
KA 234000 A1.02 3,8 ANSWER: 030 (1.00)
d
.
!
REFERENCE:.LSRO-0200, Obj. 3 (page 8)
KA 288000 A3.01 3.8
-
ANSWER: 031 (1.00)
c REFERENCE:
LSRO 0200, Obj. 3 (page 9)
MA 290001 R3,04 3.3 t
ANSWER: 032 (1.00)
- d REFERENCE:
13RO 1560, Obj. 2, (page 14)
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SENIOR REACTOR OPERATOR Page 9 of 14 ANSWER REFERENCE
\\
ANSWER: 033 (1.00)
b REFERENCE:
LSRO 1650, obj. 2, (pages 9, 10)
'
KA 295023 AK2.06 3.8
._
ANSWER:
034 (1.00)
'
b
.
REFERENCE:
LSRO 0007, Obj.10 (page 12)
IIP-231 (page 7)
KA 234000 A1.02 3.8 ANSWER: 035 (1.00)
d REFERENCE:
LSRO 1730, Obj 5, (page 11)
KA 294001 K1.03 3.8 ANSWER: 036 (1.00)
d REFERENCE:
LSRO-0310, Obj. 7 (page 5)
KA 211000 A3.08 4.2
..
.
_ - -. _ - _ _ - - _ -. -.
- - - _ _ -
....
.
.
_
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.
.
SENIOR REACTOR OPERATOR Page 10 of 14 ANSWER REFERENCE ANSWER:
037 (1.00)
a REFERENCE:
LSRO 1810, Obj. 2, (page 6)
038 (1.00)
d
.
REFERENCE:
LSRO 0750, Obj. 3a, (Page 10)
KA 233000 K6.07 2.8
,
ANSWER:
039 (1.00)
d REFERENCE:
LSRO 0310, Obj. 6, (page 9)
KA 211000 K6.01 2.4 s
ANSWER:
040 (1,00)
b REFERENCE:
LSRO 1730, Obj. la, (page 5)
KA 294001 K1.03 3.8
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b
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SENIOR REACTOR OPERATOR Page 11 of 14 ANSUER REFERENCE-ANSWER: 041 (1.00)
a
.
!
REFERENCE:
!
LSRO 0410, Obj. 3, (page 9)
{
i KA 295018 AK1.01 3.6 j
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,
J l
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'
ANSWER: 042 (1.00)
,
b
!
i
,
REFERENCE:
PBAPS Technical Specifications, (3.5.F.4 b)
'
LSRO 1840, Obj. 4
!
i
KA 215005 SG5.0 4.2 i
l 226001 SGS.O 4.0
234000 SG5.0 4.1 i
!
,
i
,
ANSWER:
043 (1.00)
.
!
d REFERENCE:
LSRO 0200, Obj.1 (page 4)
KA 288000 SC 04 3.4
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l ANSWER:
044 (1.00)
a i
REFERENCE:.PBAPS Technical Specifications, 3.4.A. 3.3.A, Bases 3.4.A, 3.10.A.2 l
.l LSRO-1840, Obj. 4
_.
KA 211000 SC 5-4.4 i
KA 234000 A1.03 3.9 i
e
_ _ _ _ _ _ _ _ _
._
___
SENIOR REACTOR OPERATOR Pago 12 of 14 ANSWER REFERENCE ANSWER: 045 (1.00)
b
__
LSRO-0002, Obj. 5 (Page 22, item 7)
FA 234000 Kl.01 3.7 KA 234000 Kl.02 3.3 ANSWER: 046 (1.00)
b
'
REFERENCE:
PBAPS Technical Specifications, (Bases 1.2, Page 31)
LSRO 1820, Obj. 2 KA 205000 SG.5 3.9 ANSWER: 047 (1.00)'
b REFERENCE:
PBAPS Technical Specifications Definitions, Surveillance Frequency KA 234000 SG.6 2.7 ANSWER: 048 (1.00)
a REFERENCE:
LSRO-0002, Obj. 9 (Page 27 g, Page 28, h, i, j, k. 1)
,
KA 234000 SC.5 4.1
.
.
.
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.
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.
SENIOR REACTOR OPERATOR Page 13 of 14 ANSWER REFERENCE
.
ANSWER:
049 (1.00)
a REFERENCE:
PBAPS Technical. Specifications (Bases 3.10.C)
YA 234000 50.6 3.7
-
233000 SG.6 3.4 ANSWER: 050 (1.00)
b REFERENCE:
.
L3RO-0002, Obj.10 (page 27) (FH 60, Page.13)
s KA 234000 SGK.1 3.8
ANSWER: 051 (1.00)'
a REr'ERENCE:
LSRO-0002, Obj. 15, (S018.7. A-3, Pa;;e 4, Step 4.20)
l
KA 234000 K5.05 3.7 l
SG.13 3.3
.
ANSWER: 052 (1.00)
-
a i
!
-REF2RENCE:
LSRO 0002, Obj.13, (page 40, F.2.a).
KA 234000 GF1 37 i
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SENIOR REACTOR OPERATOR Page 14 of 14
.
ANSWER REFERENCE
!
l ANSWER: 053 (1.00)
e i
REFERENCE:
LSRO-1530, GP-16, Obj 2, (page 1)
KA 290001 SG.13 3.5 ANSWER: 054 (1.00)
a
.
REFERENCE:
LSRO-1760, Obj. 5, (page 4)
KA 294001 K1.03 3.8
ANSWER:
055 (1.00)
l d
REFERENCE:
LSRO 0710, Obj. 1, (pages 4, B1)
MA 294001 K1.05 3.7 c
END OF EXAMINATION
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Date SENIOR REACTOR OPERATOR Page 1 of 3 ANSWER SHEET-
Multiple Choice (Circle or X your choice)
If you change your answer, write your seicetion in the blank.
001 a
b c
d 002 a
b c
d 003 a
b c
d 004 a
b c
d 005 a
b c
d 006 a
b c
d 007 a
b c
d 008 a
b c
d 009 a
b c
d'
010 a
b c
d 011 a
b c
d 012 a
b c
d 013 a
b c
d 014 a
b c
d 015 a
b c
d 016 a
b c
d 017 a
b c
d 018 a
b c
d 019 a
b c
d 020 a
b c
d 021 a
b c
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Name Date SENIOR REACTOR OPERATOR Page 2 of 3 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
022 a
b c
d 023 a
b c
d 024 a
b c
d 025 a
b c
d 026 a
b c
d
,
027 a
b c
d 028 a
b c
d 029 a
b c
d 030 a
b c
d 031 a
b c
d 032 a
b c
d 033 a
b c
d 034 a
b c
d 035 a
b c
d 036 a
b c
d 037 a
b c
d 038 a
b c
d 039 a
b c
d 040 a
b c
d 041 a
b c-d 042 a
b c
d
.
.
.
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.
.
Name
,
e Date SENIOR REACTOR OPERATOR Page 3 of 3
,
ANSWER SHEET Multiple Choice (Circle or X your choice)
If you chango your answer, write your selection in the blank.
043 a
b c
d 044 a
b c
d 045 a
b c
d 046 a
b c
d 047 a
b c
d
,
046 a
b c
d 049 a
b c
d
,
050 a
b c
d 051 a
b c
d-052 a
b c
d
'
053 a
b c
d 054 a
b c
d 055 a
b c
d END OF EXAMINATION i
.
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SENIOR REACTOR OPERATOR Page 1 of 3 ANSVER KEY 001 c
002 b
003 b
004 c
005 d
,
006 e
007 d
008 e
'
,
009 a
010 b
011 d
012 d
013 c
014 a
015 b
016 d
017 a
018-b 019 e
020 d
021 b
022 b
023 a
024 b-
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SENIOR REACTOR GPERATOR Page 2 of 3 i
ANSWER KEY
,
025 a
016 c
027 d
028 a
029 c
.
030 d
031 e
,
032 d
.
033 b
034 b
035 d
l
'
036 d
'
I 037 a
!
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038 d
'
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R 039 d
'
040 b
!
041 a
042 b
043 d
044 a
045 b-046 b
= 04 7 b
048
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SENIOR REACTOR OPERATOR Page 3 of 3 ANSWER KEY 049 a
050 b
051 a
052 a
053 c
054 a
055 d
,
END OF EXAMINATIO'N l
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