IR 05000373/1997303
| ML20148H810 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/05/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20148H776 | List: |
| References | |
| 50-373-97-303OL, 50-374-97-303OL, NUDOCS 9706110238 | |
| Download: ML20148H810 (163) | |
Text
{{#Wiki_filter:. . U. S. NUCLEAR RtiGULATORY COMMISSION REGION 111 Docket Nos: 50-373; 50-374 Licenses No: NPF-11; NPF-18 Reports No: 50-373/97303(OL): 50-374/97303(OL) Licensee: Commonwealth Edison Company (Comed) Facility: LaSalle County Nuclear Generating Station, Units 1 & 2 Location: 2601 North 21st Road Marseilles,IL 61341 Dates: April 21 - April 24,1997 Examiners: D. R. McNeil, Chief Examiner J. A. Lennartz, Examiner M. E. Bielby, Examiner D. S. Muller, Examiner in Training J. D. Ellis, Examiner in Training D. R. Roth, Examiner / Resident inspector, Dresden Station Approved by: Melvyn Leach, Chief Operator Licensing Branch 9706110238 970605 I PDR ADOCK 05000373 V PDR u
. _ _ _ . . EXECUTIVE SUMMARY LaSalle County Nuclear Generating Station, Units 1 & 2 NRC Examination Reports 50-373/97303; 50-374/97303 An NRC developed initial operator licensing examination was administered to two Senior Reactor Operator (SRO) applicants and one Reactor Operator (RO) applicant. The examination process incorporated a one week on-site period for review of the examination by facility personnel.
Results: All applicants passed all portions of their respective examinations but were not issued operator licenses. Waivers for the required reactivity manipulations were granted by the NRC to allow the applicants to take the examination. Completion of required reactivity manipulations and enrollment in the st6 tion requalification training program is necessary before the licenses will be issued.
During this examination, three occurrences of poor examination material control were
identified and investigated by licensee and NRC personnel. Training Department instruction (TDI) 211, Examination Security, was reviewed by the examiners and determined to contain significant weaknesses.
Numerous procedural errors were found as NRC examiners developed the examination.
One error dealing with a non-conservative technical specification was considered an inspection follow up item. (Section 03.1) Summan: The following is a summary of performance items: The applicants demonstrated some difficulties answering selected job performance i
measure (JPM) follow-up questions. (Section 05.3.b) When compared with previous examinations, candidate communications practices
were improved. (Section 05.4.b) A lack of attention to detail exists in the examination security area, as
demonstrated by three separate occurrences of examination security breakdown.
' Review of the facility's examination security procedure was considered an inspection follow-up item. (Section 05.5b)
. Reoort Details 1. Ooerations
Operations Procedures and Documentation 03.1 General Comments Numerous typographical errors were discovered in various operating procedures during examination development. Most of these errors were minor in nature and did not affect safety of the plant. These errors were discussed with the Supervisor, Operations Support Group. The supervisor indicated they would make changes to the affected procedures.
Unit 1 Technical Specification 3.10.8 allows plant operations with suppression pool temperature above 110 F provided reactor power is below 60% rated thermal puwer. The technical specification gives no maximum allowed suppression pool temperature, but requires cooling the suppression pool to less than 100 F within 24 hours after exceeding 110 F. The facility emergency operating procedures (LGAs) require operators to scram the reactor if the suppression pool temperature reaches 110 F. Since the LGAs are based on the heat capacity of the suppression pool, it appears that technical specification 3.10.8 is non-conservative. The inspectors considered this an inspection follow up item (IFl 50-373-97303-01).
Unit 2 Technical Specification 3.10.8 was previously deleted.
Operator Training and Qualification 05.1 General Comments The examination was an NRC prepared and administered examination. The examination was prepared using the guidance provided in NUREG 1021, Operator Licensing Examiner Standards, Revision 7, Supplement 1. The examination contained: A written test consisting of 98 multiple-choice questions and 1 matching e question, focusing on a broad spectrum of plant system design and operation.
e A walkthrough section of the examination that focused on administrative topics and operating job performance measures (JPMs) related to control room and inplant systems.
e A performance-based section focused on dynamic, integrated plant conditions using dynamic simulator scenario sets.
. . -. - ._.
- l ,. . 05.2 Written Test a.
Scoce (NUREG-1021) The written test was developed using the guidance contained in NUREG 1021, Revision 7, Supplement 1, and consisted of 98 multiple-choice questions and 1 matching question. The matching question was assigned a value of two points, resulting in a 100 point examination.
b.
Observations and Findinos Facility personnel assigned to the validation team made several suggestions that enhanced the quality of the written examination. Examination scores for the applicants were above 90%. No post examination comments were developed by the facility. NRC review of the results of the written examination revealed one common weakness, displayed by all three applicants. None of the applicants were able to correctly predict the results of a selected failure in the turbine generator electro-hydraulic control (EHC) system.
c.
Conclusions The high scores on the written examination were indicative of a good training program that provided the applicants with the ability to respond to a broad spectrum of knowledge items.
05.3 Walkthrouch Test (JPMs) a.
Scone (NUREG-1021) The applicants were administered ten JPMs to test their knowledge of plant systems and operating procedures. The administrative p)rtion of the examination consisted of five JPMs administered to the SRO applicants and four JPMs administered to the RO applicant. The RO applicant was asked two questions in lieu of a JPM in the Emergency Plan category.
b.
Observations and Findinos The applicants appeared well prepared for the walkthrough portion of the examination involving operator performance, but were judged to be weak in answering JPM follow-up questions. Each candidate missed several JPM follow-up questions.
During simulator JPMs the simulator instructors gave cues to the applicants that caused the applicants to violate proceoores. These cues were normally in the case where a prerequisite had to be met to continue a system lineup. The simulator , instructor recognized that the system lineup would not be necessary in the l simulator and instructed the applicant to disregard the prerequisite by stating, { " don't worry about that,its ok." This was judged by the examiners to be a poor practice as it may lead the applicant to bypass system lineup prerequisites in the ' plant.
, l
. . c.
Conclusions Applicant performance on the walk through examination (administrative and operational) was indicative of a good training program that provided the applicants with the ability to respond to a broad spectrum of operating situations. The JPM follow-up questions were more difficult than those administered on provious NRC examinations, but reflected the depth and complexity of questions the NRC is trying to achieve. Since the NRC is attempting to administer more complex JPM questions, some attention to training applicants on answering open reference examination questions may be warranted. Follow-up discussion with the simulator instructors on how to properly phrase instructions to trainees concerning completed prerequisites is warranted.
05.4 Dynamic Simulator Examination a.
Examination Scoce (NUREG-1021) Using NUREG-1021, Operator Licensing Examiner Standards, Revision 7, Supplement 1, the examiners developed and administered three dynamic simulator scenarios.
b.
Observations and Findinos Applicant use of procedures and annunciators was good. They were able to readily diagnose plant conditions. Control board operations were swift, accurate, and in many cases, double checked by a second operator. The use of shift briefs and communications in general were improved when compared with previous NRC administered examinations. No significant weaknesses were noted in the applicants during the dynamic simulator examination.
c.
Conclusions The applicants appeared to be well prepared for the dynamic simulator scenarios.
When compared with previous examinations, applicants appear to be improving in the area of communications.
05.5 Examination Security a.
Scoce (NUREG-1021) Examination development guidance and security measures were prescribed by NUREG 1021, " Operator Licensing Examiner Standards," Revision 7, Supplement 1.
Facility personnel also used LaSalle County Station Training Department Instruction (TDI) 211, Exarnination Security, to implement the guidelines of NUREG 1021.
b.
Observations and Findinas l During examination validation by facility instructors and operators, three separate occurrences involving examination security took place.
. . (1) An electrical maintenance instructor that had not signed the NRC examination security agreement ignored signs posted on the copy room door indicating authorized personnel only were to enter the room. The instructor believed he was authorized since he was an instructor. The clerk that had posted the signs challenged the instructor, indicating that he should not be in the room while she was copying the examination. The instructor ignored the challenge by the clerk and remained in the room while the examination copying was completed. The clerk stated to examiners that the instructor was never close enough to the copying machine to see the examination material being copied. The incident was investigated by facility personnel and disciplinary action taken against the instructor. After interviewing the clerk the examiners concluded that no compromise of the written examination had occurred.
(2) An instructor assigned to validate the written examination determined that three pages of a copy of the written examination given to him for review were missing. Facility management personnelinterviewed all personnel that ' had access to the examination. The clerk that copied the examination stated that after making the examination copy, she page checked the copy and all pages were in the copy. She stated she had sent additional material through the copying machine to verify there was no examination materialleft in the copying machine. Management personnel were unable to determine how or when the pages were removed from the examination. At the conclusion of their investigation a management representative stated to tho examiners that they believed the copy machine never actually copied the missing pages and that the page check performed by the clerk may have been inaccurate. The representative indicated he believed the clerk may have been distracted during the page check because the previously mentioned electrical , maintenance instructor was in the copy room while the clerk was page checking the examination.
Examiners interviewed the administrative clerk and some facility personnel participating in the validation of the examination. The results of the interviews were inconclusive. Since neither facility personnel nor the examiners were able to determine the status of the three missing pages, the examiners replaced all questions on those pages with new questions.
(3) During the week of April 14,1997, one of the applicants entered the Production Training Center (PTC) library to check out a procedure for review in preparation for taking the examination. The applicant picked up a checkout card used to check out procedures and noted that three facility personnel on the examination validation team had signed the checkout card for a procedure concerning personal radiation monitoring. The applicant chec:.ed out another procedure using the same card as the three validators.
Upon returning the procedure to the PTC library and retrieving the checkout card, the applicant became concerned that the appearance of the three validator's signatures on the checkout card with his signature on the checkout card would give the appearance of examination compromise. The applicant then destroyed the card. Shortly after destroying the card the
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_ _ ._ _. - __ _ _ _ _ __ ___ .__ . . . applicant decided he had not taken the correct action and reported the event to his immediate supervisor.
Upon learning of the event, examiners directed the training manager to collect all of the reference library checkout cards using sorneone on the security agreement. The checkout cards were retrieved and held until the examination was completed. The checkout cards indicated that the <
operating procedures for the reactor recirculation system, the reactor building ventilation system and the turbine generator's electro-hydraulic ' control system had been reviewed by persons on the security agreement during the examination validation time period. Because these system operating procedures were sizable, and questions from these procedures had a high probability of appearing on an NRC examination, examiners concluded l it was not necessary to change any questions on the examination.
Examiners reviewed TDI-211, Examination Security, to determine if adequate examination security measures were being implemented during requalification i examinations. While reviewing the TDI, examiners reviewed Attachment TDl-211 A,
which contained a definition of the phrase, " Involved in the Training," used by the NRC for security agreements. A note indicated the following tasks are not , considered to constitute " Involvement in the Training:" ! (1) Operation of the simulator (Booth Operator) provided that the person did not
select the training content or participate in the critique.
. (2) Acting as an evaluator during performance mode evaluations on the ' i i simulator (including providing performance feedback) provided the person does not select the evaluation content, and are not involved in the ' remediation of candidates.
(3) Preparation and administration of the certification exam (iruuding providing performance feedback) provided the person (s) are not in,olved in the remediation of candidates.
Examiners considered all three of these exceptions to be " involvement in the training" of the applicants. Examiners directed the lead instructor for initial license ' training to inform everyone signed on the security agreement they were to have no professional contact with the applicants. Examiners provided the applicar.ts with a
list of everyone signed on the examination security agreement and instructed the applicants to have no professional contact with anyone on the security agreement.
Facility trainers have indicated they will compare TDI-211 with the upcoming Revision 8 of NUREG 1021 and revise any areas not meeting the guidelines of the NUREG. Review of the changes to TD1-211 is considered an inspection follow-up i item. (IFl 50-373/97303-02) , c.
Conclusions A lack of attention to examination security issues exists, requiring management attention in this area.
_ , . 05.6 Simulator Fidelity a.
Examination Scoce (NUREG-1021) Using NUREG-1021, Operator Licensing Examiner Standards, Revision 7, Supplement 1, the examiners observed the performance of the plant specific simulator during job performance (JPM) walkthrough and dynamic scenario sets.
, This fidelity review was conducted in relation to simulation tasks performed on the plant specific simulator during the examination process, b.
Observations and Findinos During the performance of one dynamic scenario, the examiners observed one problem with the simulated plant process computer which is described in Enclosure 2, Simulation Facility Report.
c.
Conclusions The plant specific simulator was able to accurately mimic plant parameters during a variety of malfunctions and plant conditions.
V. Manaaement Meetinas X1 Exit Meetino Summary Examiners presented the examination team's observations and findings to members of the licensee's management on April 24,1997. The licensee acknowledged the findings presented and the cornmitment to enroll the applicants in the continuing operator license program. No proprietary information was identified during the exarnination or at the exit meeting.
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i PARTIAL LIST OF PERSONS CONTACTED Licensee P. Barnes, Regulatory Assurance Supervisor J. Connon, ILT Lead instructor J. Davis, Training Manager J. Drago, NRC Coordinator, Reg Assurance G. Ford, Operating Staff Supervisor L. Guthrie, Plant Manager, U2 G. Kaegi, Operations Training Supervisor A. Magnafici, operations Manager, U1 S. Smith, Plant Manager, U1 NRC M. Huber, Senior Resident inspector R. Crane, Resident inspector ITEMS OPENED, CLOSED, AND DISCUSSED Onened 373-97303-01 IFl Conflict between LGAs and Tech Spec 3.10.8 373-97303-02 IFl TDI-211 Exam security procedure weakness Closed , NONE i Discussed i NONE
, . . . Enclosure 2 SIMULATION FACILITY REPORT Facility Licensee: LaSalle County Generating Station, Units 1 & 2 Facility Licensee Docket No: 50-373, 50-374 Operating Tests Administered: April 21 - April 23,1997 The following documents observations made by the NRC exemination team during the April 1997,initiallicense examination. These observations do not constitute audit or ' inspection findings and are not, without further verification and review, indicative of non-compliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observations.
During the conduct of the simu!ator portion of the operating tests, the fo!!owing items were observed:
ITEM DESCRIPTION Plant Process A radiation monitor alarm for the reactor building ventilation Computer system was inserted as a result of a malfunction. The simulated plant process computer indicated the fault was with the spent fuel pool radiation monitor.
Switch Check Several switches have flags that are not switch checked during reset. This led to several flags indicating equipment was in a different operating state than the actuallineup indicated by other i simulator indications.
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ES-401 Site-specific Written Examination Form ES-401-1 Cover Sheet U. S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION _ APPLICANT INFORMATION Name: 091on: Ill MASTER EXAMINATION Date: April 24, 1997 Facility / Unit: LaSalle Co. Station
Units 1 & 2 License Level: R0 Reactor Type: GE INSTRUCTIONS Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80 percent.
Examination papers will be picked up 4 hours after the examination starts.
All work done on this examination is my own.
I have neither given nor received aid.
Applicant's Signature RESULTS ~ Examination Value 100.0 Points _ i Applicant's Score Points Applicant's Grade Percent + Examiner Standards 6 of 7 Rev. 7. January 1993
.. . _ -.._-. - . . _ _. -. -- .. i ' l l REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your se!ection in the blank.
MULTIPLE CHOICE O22 a b c d _ 001 a b c d 023 a b c d 002 a b c d 024 a b c d 003 a b c d 025 a b c d 004 a b c d 026 a b c d 005 a b c d 027 a b c d i 006 a b c d 028 a b c d 007 a b c d 029 a b c d 008 a b c d 030 a b c d 009 a b c d 031 a b c d 010 a b c d 032 a b c d 011 a b c d 033 a b c d 012 a b c d 034 a b c d 013 a b c d 035 a b c d 014 a b c d 036 a b c d 015 a b c d 037 a b c d - 016 a b c d 038 a b c d 017 a b c d 039 a b c d 018 a b c d 040 a b c d 019 a b c d 041 a b c d 020 a b c d 042 a b c d
O21 a b c d 043 a b c d ( .
. - - _ - -. .__ _ _ _ -. - _ _ -. _ -.. ... - ._ ! i REACTOR OPERATOR Page 3 ANSWER SHEET Multiple Choice (Circle or X your choice) l 044 a b c d 067 a b c d , 045 a b c d 068 a b c d 046 a b c d 069 a b c d 047 a b c d 070 a b c d 048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d i 052 a b c d 075 a b c d 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d i 058 a b c d 081 a b c d i i 059 a b c d 082 a b c d . 060 a b c d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b c d I 065 a b c d 088 a b c d 066 a b c d 089 a b c d - l l
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! l REACTOR OPERATOR Page 4 l ANSWER SHEET f Multiple Choice (Circle or X your choice) i 090 a b c d _ l l 091 a b c d _ 092 a b c d _ 093 a b c d _ 094 a b c d _ 095 a b c d _ 096 a b c d 097 a b c d _ 098 a b c d _ 099 MATCHING a _ b _ c _ d _ _ (* * * * * ' * * * END OF EXAMINATION * * * * * * * * * *) l l ! l l -
.. . . -.. - - .--... . - I i l l l l I ! , Page 5 l ES-402 Policies and Guidelines Attachment 2 i for Taking NRC Written Examinations
( 1.
Cheating on the examination will result in a denial of your application and could result in more severe penalties.
2.
After you complete the examination, sign the statement on the cover sheet I indicating that the work is your own and you have not received or given assistance in completing the examination.
i 3.
To pass the examination, you must achieve a grade of 80 percent or greater.
4.
The point value for each question is indicated in parentheses after the question number.
5.
There is a time limit of 4 hours for completing the examination.
6.
Use only black ink or dark pencil to ensure legible copies.
7.
Print your name in the blank provided on the examination cover sheet and the answer si'eet.
8.
Mark your answers on the answer sheet provided and do not leave any question l blank.
9.
If the intent of a question is unclear, ask questions of the examiner only.
10.
Restroom trips are permitted, but only one applicant at a time ' sill be allowed to i leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.
I 11.
When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or proctor. Remember to sign the statement on the examination cover sheet.
-
12.
After you have turned in your examination, leave the examination area as defined by ' the examiner.
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.._-_ _ -_- _ - _- - - .. . _ -.-. _... - -. - - -... -.... - - . .- l l REACTOR OPERATOR Page 6 ! . ! OUESTION: 001 (1.00) For the following plant conditions: l ' ' Drywell pressure at 1.0 psig.
- l Drywell Hydrogen Concentration at 1 % - l . Suppression Pool Water temperature at 95 F.
l - l Suppression Pool Water level at + 6 inches.
l - l Reactor Building Radiation level at 6 mr/hr.
- - , RPV water level at 22.5" ! l - l , Rx Bldg d/p.25" we - t l Which of the following LGAs would be entered? l a.
LGA 01 only [ ! b.
LGA 03 only i ! ' ! c.
LGA 01 AND O2
! i j d.
LGA 01,03, AND 04 l QUESTION: 002 (1.00) l There is a leak into the drywell from the recirculation system. The leak does not , ! depressurize the reactor but drywell temperature is rising. Which of the following is the l effect on panel level instruments? I ! a.
Indicated level may read above actual level.
l.
> , l b.
RPV level instrumentation is no affected by drywell temperature.
i i c.
Variable leg flashing may occur causing erratic levelindication.
l d.
Indicated level will be correct unless reactor building temperature exceeds 140*F.
! ,
I , t i ! ! I = , P - -,,, , , , v n -,.-, , - -.
REACTOR OPERATOR Page 7 QUESTION: 003 (1.00) LGA-03, Primary Containment Control, directs the operator to maintain suppression pool temperature below the HCTL and if you cannot, then reduce reactor pressure to stay inside the HCTL. Reducing reactor pressure to stay inside the HCTL is to... ensure there is adequate margin to the ECCS suction piping design a.
temperature in the event of a full reactoi depressurization.
b.
ensure the suppression pool has enough capacity to accept a full reactor depressurization without exceeding the design temperature of the suppression pool, c.
prevent inadequate steam condensation in the event of a full reactor depressurization, resulting in the suppression pool to drywell vacuum breakers opening.
d.
allow the operator to depressurize the reactor to a pciot where RHR and LPCS can inject prior to the suppression pool temperature exceeding the RHR NPSH limit.
QUESTION: 004 (1.00) WHICH of the following tasks are you allowed to perform WITHOUT a written procedure present? a.
Initiating a Temporary Procedure Chenge.
b.
Performing the monthly surveillance on the O D/G.
c.
Performing a normal unit shutdown to meet T.S. 3.0.3.
- d.
Installing jumpers per the alternate rod insertion LGA.
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. _..... _ _ _. _. _ __ .._.___._.m..-__-_- ~._._. _ _. _. _ _ _ _ _ __._ i i l ) . , I REACTOR OPERATOR Page 8 i i i . QUESTION: 005 (1.00)
WHICH of the following control rod manipulations. REQUIRE second verification by a technically qualified individual? { a.
Weekly individual rod exercising for rods at notch 48.
L b.
. Weekly individual rod exercising for intermediate. rods.
. . i l c.
Insertion of CRAM rods on entry into the stability region.
I I d.
Individual control rod movements with the Mode Switch locked in REFUEL l and the one-rod-out interlock is OPERABLE.
[ QUESTION: 006 (1.00) You are in the RCIC room in full protective clothing (no mask) working in a contaminated area. The assembly alarm SOUNDS and a plant announcement directs all personnel to , ' proceed to their assembly area. What action, if any, should be taken prior to proceeding to the assembly area? a.
remove all protective clothing.
b.
no action, exit the RCIC room as you are, c.
remove your outer gloves and rubber shoes.
d.
remove all protective clothing except the cotton liners and plastic boots.
_
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_. - _ _._ _ _ _. _.__.... _ _. _ _ _ -~ . _ _ _ _ -_ - __ - _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _. _ _ _.. _. REACTOR OPERATOR Page 9 QUESTION: 007 (1.00) A control rod drifted in the out direction, was inserted and hydraulically isolated at its HCU.
After hydraulic isolation the rod continued to drift out. From this, it can be concluded that: a.
The rod has a stuck collet.
b.
The rod has blown scram pilot valve fuses.
CRD cooling water flow and pressure are too high.
c.
d.
The rod has a failed withdraw / supply directional control valve.
QUESTION: 008 (1.00) Tech Staff has written a special test procedure which will require placing the RMCS in Manual Self Test and Manual Scan at the Rod Drive Control Cabinet. What effect will these modes have on the NSO's r5ility to move control rods? The RMCS will: a.
permit rod motion in either mode, b.
NOT permit red motion in either mode.
c.
permit rod motion in Manual Self Test, but NOT in Manual Scan mode.
o.
permit rod motion in Manual Scan, but NOT in Manual Self Test mode.
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f
{'.
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_ _ _ _ _ _... _. _ _ _ _ _ _ _ _.. _ _ _ _ _ ._, _.__ _.___ _ __..._.._-..___ _ _ __. _ __.._ _ _ _., l - c ! i ! REACTOR OPERATOR Page 10 i ! , QUESTION: 009 (1.00) i f The Standby Liquid Control System (SBLC) is in its normal, STANDBY lineup. WHICH of ! i l the following describes the correct indications on control room panel H13 P603 for this - i i lineup? The SQUlB continuity lights for valves F004A/B are: l
! a.
LIT, the SBLC pump suction stop valves (F001 A/B) indicate OPEN.
l l b.
LIT, the SBLC pump suction stop valves (F001 A/B) indicate CLOSED.
l t c.
NOT LIT, the SBLC pump suction stop valves (F001 A/B) indicate OPEN.
! . ! l d.- NOT LIT, the SBLC pump suction stop valves (F001 A/B) indicate CLOSED.
! QUESTION: 010 (1.00) , , !
l Which ONE of the following Intermediate Range Monitor (IRM) Rod Blocks is bypassed j l - when the Range Selector is positioned to RANGE 1?
l ' a.
IRM HI , b.
IRMINOP ! i c.
IRM DOWNSCALE ! d.
Detector Not Full in
~ i [ , i , , - .-, e -. -,, , - -. ..-n., , w .- -r - ~.
i ! ! ! REACTOR OPERATOR Page 11 l QUESTION: 011 (1.00) Given that your unit is in Mode 1, which one of the following RCIC valves is NOT in the normal standby line-up position? a.
Turbine governor valve (F361) indicates open.
b.
Pump minirnum flow valve (F019) indicates closed.
c.
Turbine steam supply stop valve (F045) indicates open.
d.
Cooling water to tube oil cooler stop valve (F046) indicates closed.
QUESTION: 012 (1.00) Which of the following describes the effect of losing 125 VDC bus (112Y) power to the Automatic Depressurization System (ADS) with a subsequent valid ADS initiation signal? a.
ADS logic "A" will function. Only 4 ADS valves will open.
b.
ADS solenoid "A" will function. All ADS valves will open.
c.
ADS solenoid "B" will receive backup power from 125 VDC bus 111Y. All ADS valves will open.
d.
ADS logic "B" will receive backup power from the Auxiliary Electric Equipment Room (AEER). Only 4 ADS valves will open.
.. M
l ! i REACTOR OPERATOR Page 12 l QUESTION: 013 (1.00) After an auto initiation signal has occurred, why is one train of the Standby Gas Treatment l System (VG) immediately shutdown? l a.
Prevent excessive stack release rate.
b.
Prevent overhcat of the charcoal adsorbers.
l c.
Ensure the proper operation of the flow control dampers.
, ! d.
Ensure the proper flowrate to the isokinetic probe of the VG WRGM.
l QUESTION: 014 (1.00) ! l Given the following plant conditions: - A reactor startup is in progress Reactor power is in the intermediate range -
! - NO rod blocks or errors exist The next to the last rod in the group, rod 30-43, is notched to position 10, instead of l stopping at the Banked Position Withdrawal Sequence (BPWS) position of 08. The RWM ' will: a.
allow the rod to be moved to position 8.
' b.
allow entering an alternate limit of 10 for rod 30-43.
l c.
allow the entire latched group to be moved to position 10.
d.
lock-up, and not allow any control rod movement without bypassing the rod.
i
2 . l . l
REACTOR OPERATOR Page 13 l QUESTION: 015 (1.00) l The reactor is initially at 100% power, with both Reactor Recirculation Pumps (RRPs) in FAST speed. The following transient occurs: ( - The main turbine trips ' Aux. power successfully transfers to the SAT, NO AC or DC busses are lost.
- Which of the following actions is expected to occur as a DIRECT result of the turbine trip? a.
Bcth RRPs trip OFF.
b.
RRPs shift to SLOW speed c.
Both Flow Control Valves LOCKUP.
l d.
Both Flow Control Valves RUNBACK.
l l t ! QUESTION: 016 (1.00) Unit 1 shutdown from Power Operation to Hot Shutdown is in progress. Warmup of Shutdown Cooling is about to commence. Further RPV depressurization at this point could result in which ONE of the following conditions? a.
Increased RHR conductivity.
b.
Flashing within the RHR System.
i c.
RPV low pressure isolation of RHR.
d.
Excessive Shutdown Cooling warmup rate.
- l ! l
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_ _. _ _ _ .._____-.__m ._ __m . _-_ _ _ _ _.. _ _ _ _ _...._ _ _ _. _ _ _ _. _ _..... _. _., f t REACTOR OPERATOR Page 14 ! i ! OUESTION: 017 (1.00) l A single control rod that is scrammed from position 10 using the Scram test switches, will l ALWAYS result in each of the following EXCEPT: ,
a.
an HCU TROUBLE Annunciator.
i
b.
a Control Rod Block Annunciator.
. c.
a Control Rod Drift Annunciator.
d.
a Blue " SCRAM" light on full core display.
OUESTION: 018 (1.00) Given the following conditions: I i Reactor Power at 60% i - APRM "C" is selected as the reference APRM input for RBM Channel."A" - APRM "C" fails DOWNSCALE - -- APRM "C" has NOT been bypassed - WHICH one of the following describes the response of RBM Channel "A"? a.
RBM Channel "A" is automatically Lypassed b.
automatically shifts to APRM "E" as reference APRM c.
generates an RBM DOWNSCALE FAILURE alarm and rod SELECT block ] d.
generates an RBM DOWNSCALE FAILURE alarm and rod WITHDRAWAL block l ., -.---.m.
- .,, - , . , - -
__. _ ____ _. _ _ _. _ _. ... _ _ _ _ _. _ . __ . _. _ _ _ _. _ _.. _ _ _ _ _ _ _ _. _.
I l ! REACTOR OPERATOR Page 15 i r i . QUESTION: 019 (1.00) ' l The RHR system is in the suppression pool cooling mode. The temperature limits between , suppression pool water and lake water are maintained by-
, f i j a.
Varying the number of running RHR service water pumps.
b.
throttling the RHR Service Water heat exchanger outlet valve (E12-F068A/B).
- l l
c.
throttling the RHR heat exchanger bypass valve (E12-F048A/B) and RHR > heat exchanger inlet valve (E12-F047A/B).
' d.
throttling the RHR heat exchanger bypass valve (E12-F048A/B) and RHR heat exchanger outlet valve (E12-F003A/B).
L QUESTION: 020 (1.00) l , Which of the following describes how a loss of the TSC UPS will affect plant operations? ! I a.
TDRFP speed control on Unit 2 ONLY will auto swap to its respective alternate power supply.
l b.
TDRFP speed control on BOTH units will auto swap to their respective alternate power supply, t c.
Reactor water level will be uncontrollable on BOTH units due to a loss of i TDRFP speed control.
< d.
Reactor water level will be uncontrol!able on Unit 2 ONLY due to a loss of TDRFP speed control.
- l l l ! l . d i -.. , - .-. - _ .--
_ _ _.. _ _ __ i , I REACTOR OPERATOR Page 16 QUESTION: 021 (1.00) l Which of the following conditions will automatically start the control room ventilation ! emergency makeup train? l ' a.
Smoke in the control room b.
Ammonia in the outside air j l c.
High radiation in the outside air d.
Smoke in the auxiliary equipment room ! l QUESTION: 022 (1.00) With the refueling platform near the core,. which of the following, by itself, will prevent further refueling platform movement toward the core? l l a.
All rods NOT full in.
l b.
Mode switch in " REFUEL".
l l c.
Mode switch in "STARTUP".
I d.
Any refueling hoist loaded.
. , - l , l l I ! < d i L
_ ___ _ __ l REACTOR OPERATOR Page 17 < QUESTION: 023 (1.00) With the main turbine generator on line at 40% reactor power, the following parameters were noted: Turbine Lube Oil Cooler Outlet Temperature 115*F - Turbine Bearing Oil Drain Temperature 155*F -
- Turbine Bearing Vibrations 5 to 6 mils - Turbine Exhaust Hood Temperature 230 F l - Turbine Back Pressure 2.0" Hg absolute You should: a.
Reduce turbine load, check turbine lube oil cooler service water outlet temperature, monitor bearing temperatures.
b.
Verify auto actions for a turbine trip, generator trip and reactor scram have l occurred. Verify recirc pump trip to slow speed, bypass valves open and
power transfers to the SAT.
Verify the turbine generator trips, in-house loads transfer, and bypass valve c.
operation. SHUTDOWN the unit in accordance with LGP 2-1, Unit Shutdown -! from Power Operations to Hot Shutdown.
. d.
Start the Emergency Bearing Oil pump, verify oil flow in sightglasses, and verify service water cooling flow, if turbine bearing and hood temperatures does not decrease, then manually depress the turbine trip pushbutton, verify auto actions for turbine trip.
..
BEACTOR OPERATOR Page 18 QUESTION: 024 (1.00) The unit was operating on the 100% flow control line. One recirc pump tripped, followed shortly by a trip of the second recirc pump. No automatic scrams were generated. The resulting combination of power and flow placed the plant in Region I of the Power vs. Flow map. The LPRM noise levels are within 8% peak to peak. APRMs are reading 48% with noise swings between 42% and 53%. You should: immediately place the reactor mode switch in the SHUTDOWN position.
a.
b.
insert in-sequence control rods to leave the region, be in HOT SHUTDOWN within the next six (6) hours.
c.
insert cram arrays to reduce reactor power below 36% of RATED CORE THERMAL POWER, take immediate action to restore one loop to operation.
d.
immediately reduce CORE THERMAL POWER by inserting control rods, observing the indicated APRM and LPRM noise levels and complete power reduction to below 36% of RATED CORE THERMAL POWER within two (2) hours.
QUESTION: 025 (1.00) Which of the following states the effect of losing BOTH divisions of the 24V DC power system while at power? a.
Reactor recirculation flow control valves willlock up.
' b.
Loss of all SRMs administratively prohibits a shutdown.
c.
A reactor scram occurs if the mode switch is in "STARTUP". ~ d.
The main stack cannot be sampled for particulates and lodin.-. ._ - . _ _ _ . __ _ _. _ _ _ _ _. _ _ _. _ _ _ _.. _ _. _. _. _ _.. _. _ _ _. , ! i t REACTOR OPERATOR Page 19 ! QUESTION: 026 (1.00) !
Given the following conditions: !
reactor power at 85% steady-state - l 3-element control in service ' - l recirc system operating in loop manual mode - Which ONE of the following describes the behavior of the reactor if a loss of feedwater heater 16A occurs due to closure of the steam extraction stop check valve? Core power I will... !. J l a.
remain constant at present steady state level.
b.
decrease, then return to the pre-transient power level.
c.
initially decrease, then stabilize at a lower steady state level.
d.
initially increase, then stabilize at a higher steady state level.
! QUESTION: 027 (1.00) The reactor is operating at 100% rated power when the running RBCCW pump and the running CRD pump auto trip. Attempts to restart the pumps and the standby equipment-has failed. Based on this, you should: a.
trip both reactor recirc pumps and scram the reactor.
l b.
downshift both reactor recirc pumps and insert cram array control rods.
c.
reduce core flow to 49 M#/hr and downshift the reactor recirc pumps if seal failure occurs, d.
monitor RR pump temperature and downshifting RR pumps only if motor bearing temperature exceeds 200 F.
!-
"
-. , _. .. - - - - - _ . -
-. r i REACTOR OPERATOR Page 20 QUESTION: 028 (1.00) l l Unit 1 reactor building ventilation exhaust radiation is increasing and reaches 10 mr/hr.
Select the expected automatic action.
l a.
Standby Gas Treatment (VG) would start on UNIT 1 ONLY.
! b.
Reactor building supply and exhaust dampers (VR) close on BOTH units.
c.
Purge Filter Train Suction isolation Dampers (VQ) close on Unit 1 ONLY.
d.
Nitrogen Inerting and Makeup Isolation Valves (VQ) close on BOTH units.
QUESTION: 029 (1.00) i The Reactor was operating at full power when an l&C tech started conducting manual ' scram pushbutton testing. The l&C tech asked the NSO to arm and depress the A1 and A2 manual scram pushbuttons. No channel A solenoid lights extinguished. The l&C tech , then asked the NSO to arm and depress the B1 and B2 manual scram pushbuttons without attempting to reset any scram signals from channel A. No channel B solenoid lights j extinguished. The action to be taken under these conditions is to l a.
enter LGA-10, Failure to Scram, b.
take action to be in HOT SHUTDOWN within 12 hours.
c.
lock the reactor mode switch in the SHUTDOWN position within one hour.
d.
immediately run recirc to minimum, trip the main turbine and place the reactor mode switch in SHUTDOWN.
- < l I .a
.. . _ _.. .- _.. -. - _. _. -.. _ _. _. _ _ _. -. _ _. _ _. _ _ _. _ _. _ > _ _. _ _ _ _. _.... l i REACTOh OPERATOR Page 21
i t QUESTION: 030 (1.00) What is the purpose of the Low-Low Set feature of the Safety Relief Valves (SRV's)? , a.
Minimize SRV fatigue ,
b.
Minimize containment fatigue ! ] ' c.
Prevent exhausting the drywell pneumatic supply (lN) ! d.
Prevent thermal overload of the SRV's pneumatic actuator solenoids ,
' QUESTION: 031 (1.00) i , ! LGA-05, RPV FLOODING states, if you cannot get 5 or more SRVs open or cannot hold RPV pressure 52 psi or more above suppression chamber pressure, then flood the drywell.
. , Why is it necessary to maintain RPV pressure 52 psi ABOVE suppression chamber pressure ' ,, during the vessel flood? a.
To ensure decay heat is adequately removed.
b.
To ensure collapse of steam voids in the reactor.
i c.
To prevent closure of SRVs, resulting in less than 5 open'SRVs.
I d.
To prevent the intrusion of non-condensible gases from the suppression i chamber into the reactor vessel.
. &
REACTOR OPERATOR Page 22 OUESTION: 032 (1.00) During an ATWS, why is the operator directed to delay plant cool down until the reactor is shutdown without boron injection or until the cold shutdown weight of boron has been injected? a.
The cool down will add positive reactivity, taking the reactor critical with no method available to shut down the reactor.
b.
Core flow may become restricted due to the cooler water causing the boron to come out of solution and depositing on core surfaces.
Higher reactor pressure helps mixing of the boron due to the higher steaming c.
. rates and minimizes the time to complete the shutdown.
d.
Lowering reactor pressure would cause the water in the reactor to flash to steam, thus lowering the boiling boundary and increasing the chance to damage fuel.
QUESTION: 033 (1.00) On increasing suppression pool temperature, a manual Scram is required prior to exceeding what temperature? a.
95 F b.
105 F c.
110 F d.
120 F . O
- _., -.. - - _._ _ _. _._.
_..__ l , I
REACTOR OPERATOR Page 23 l l QUESTION: 034 (1.00) The NSO in the control room has just received 1H13-P601-B110, "RB RAD Hl." Upon
investigation it is found that 1D21-K601T, "RB 740 - RB CRD Storage" on control room panel 1D21-P600 indicates 40 mr/hr. Its associated amber "HIGH" light is illuminated.
Which of the following statements is correct? The ARM signal may be erroneous. Contact Rad Protection to verify dose ! a.
readings in the area and troubleshoot and repair the area rad monitor as
required.
b.
The amber light on the ARM is a warning light indiceting that the area in the RB on the 740 elevation is increasing and precautionary action should be taken to restrict access to the area.
c.
The amber light is indicating an airborne high radiation condition. Actions should be taken to evacuate all personnel, isolate ventilation to the area and notify Rad Protection to initiate air samples, d.
The ARM is indicating an area high radiation condition. All personnel in the - area should be evacuated and access restricted. Entry into Secondary Containment Control, LGA-02, should be checked. Notify Rad Protection to survey and sample.
QUESTION: 035 (1.00) WHICH ONE (1) of the following conditions is an entry condition into LGA-02, " Secondary Containment Control?" i a.
Fuel Pool exhaust 8 mr/hr . b.
Reactor building Differential pressure at +0.5" wc c.
Steam pipe tunnel area temp 130 F d.
There is water on the floor in the RCIC room, the surnp is NOT overflowing.
! ! .
_ _ . REACTOR OPERATOR Page 24 QUESTION: 036 (1.00) Which of.following conditions require entry into LGA-02, " Secondary Containment Control"? a.
Reactor building d/p at -0.5 inches of water.
b.
Reactor building floor drain sump overflowing.
) c.
Fuel Pool ventilation radiation level at 3 mr/hr.
, . d.
Reactor water cleanup pump room temperature above normal.
QUESTION: 037 (1.00) , Given that: - The plant was operating at 20% power when a reactor scram with a turbine trip occurred The turbine bypass valves have not yet opened - - All other expected automatic actions have occurred as designed Under these conditions, which of the following will open the turbine bypass valves? Manually decreasing the... a.
Pressure Setpoint b.
Load Limit setpoint.
c.
Steam Throttle Pressure value.
d.
Max Combined Flow Limiter setpoint.
... _ _ _ _ _. - .__ _ _ . _ __._ _ __ _. _. _ -_ _ = _ _ _. _ _. _.. _ _._. ! i i r ! REACTOR OPERATOR Page 25 i i ! ! ! l QUESTION: 038 (1.C0) ! With drywell pressure at 1.8 psig and drywell temperature at 310 F (rising), which of the ., l following must be satisfied to allow drywell spray valves F016A and F017A to be opened ! l simultaneously? j
a.
LPCIinjection valve F042A closed.
! ! b.
Suppression pool spray valve F027A closed.
! I c.
RHR Heat exchanger bypass valve F048A open.
j r d.
Suppression chamber pressure less than 8 psig.
I ! l , QUESTION: 039 (1.00) The reactor was at 100% power when 1PM02J B101, "H2 PANEL TROUBLE" alarms.
! Investigation by the EA at the local panel 1PL19J indicates alarm B101, " MACHINE GAS PRESSURE HIGH/ LOW" is in alarm. The NSO confirms that hydrogen pressure is at 45 psig. What is the concern with hydrogen at this pressure? i a.
Hydrogen seal oil will collect in the Hydrogen Retraining Section and cause a liquid detector alarm.
b.
Hydrogen purity will not be valid due to the lower pressure and decreased flow to the hydrogen purity monitor.
c.
If hydrogen pressure were to fall below service water pressure, service water could begin filling the generator casing.
- d.
Service air, via the service air purge line for carbon dioxide removal, will overcome hydrogen pressure and cause a potentially explosive concentration of hydrogen and air.
i l - - i I ! e , +, - - , . - -,. - ~. - -, - - >~ - - - >
...... -.. ~ - - . . _.
. . - ~ . -.-. -.-... .., -... r b I !
REACTOR OPERATOR Page 26 i i ! QUESTION: 040 (1.00) Unit 1 is at 100% power with the SAT supplying electrical power to bus 143. If an . overcurrent condition occurs on bus 143, what is the expected status of the following I components? ! Breaker 1432 (SAT feed to bus 143) ! - Breaker 1433 (18 DG feed to bus 143) - DG 1B - - a.
1432 trips OPEN,1433 remains OPEN, DG 1B remains SHUTDOWN.
b.
1432 trips OPEN,1433 remains OPEN, DG 18 STARTS.
c.
1432 trips OPEN,1433 CLOSES then trips on overcurrent, DG 1B STARTS.
_
d.
1432 remains CLOSED,1433 remains OPEN, DG 1B remains SHUTDOWN.
? QUESTION: 041 (1.00)
t Which of the following is NOT directly affected by a loss of 24/48 VDC? f a.
Source Range Monitors i b.
Process Radiation Monitors c.
'ntermediate Range Monitors ! d.
Local Power Range Monitors .. QUESTION: 042 (1.00) Which ONE of the following Off-Gas parameters is indicative of a fire upstream of the recombiner? - i a.
rapid increase in recombiner temperature ' b.
rapid decrease in recombiner temperature c.
rapid increase in charcoal bed temperature , d.
rapid decrease in charcoal bed ternperature - . v - ,- - - - -- -- - e- -- -+ -~- ., - - -w o---r -,, - ~ r-
_ _ l ' REACTOR OPERATOR Page 27 QUESTION: 043 (1.00) Unit 1 is operating at 100% power with the "B" offgas post treatment Process Radiation Monitor (PRM) inoperable and its selector switch in STANDBY. If the "A" PRM channel fails downscale, which one of the following is a direct response? a.
Offgas realigns for the TREAT mode b.
Offgas suction valves automatically close c.
Offgas discharge valve automatically closes d.
Offgas causes MSIV closure and Reactor Scram QUESTION: 044 (1.00) If a Carbon Dioxide (CO2) hose reel was lifted from its holder for fighting a fire on the turbine deck, what indication is provided to the fire brigade member to tell him the hose is capable of ejecting Carbon Dioxide? The local CO2: a.
indicating light would turn ON.
b.
indicating light would turn OFF.
c.
audible annunciator would turn ON.
d.
pressure gauge would activate, showing CO2 pressure.
- -
._._ _ - - - ! l l REACTOR OPERATOR .Page 28 - QUESTION: 045 (1.00) , If both trains of the Standby Gas Treatment System are inoperable, which of the following , conditions exist? a.
Primary Containment Integrity is no longer assured, b.
Secondary Containment integrity is no longer assured.
c.
Desired Drywell pressure can no longer be maintained.
d.
Reactor Building negative pressure can no longer be maintained.
i QUESTION: 046 (1.00) A TIP probe was traversing the core when an event occurred causing Reactor Water Level to drop to + 10". Which one of the following statements describes TIP System response to this event? The TIP system... a.
continues its sequence for the present core position.
b.
Shear valve immediately fires, shearing the cable and isolating the TIP System.
c.
Ball valve immediately CLOSES, cutting the cable and isolating the TIP System.
d.
immediately reverses and withdraws the probe from the core into the shield, and the ball valve CLOSES.
!
l
l i ! REACTOR OPERATOR Page 29 , l
' l QUESTION: 047 (1.00) i i l ' ! Unit 1 is in Hot Standby with RCIC turbine in operation. It may be necessary to reduce . RCIC steam flow when warming the Main Steam Lines to prevent: a.
Inadvertent SRV actuation.
b.
Excessive RCIC pump cavitation.
c.
Excessive carryunder in the RPV.
d.
Excessive reactor power increase.
QUESTION: 048 (1.00) If the output of the selected EHC pressure regulator failed high while operating at 100% reactor power, which one of the following describes the plant response? (Assume no operator action is taken.)
a.
MSIV closure on low reactor pressure i b.
Main turbine trip on overspeed due to TCVs opening c.
Reactor scram on high flux or high reactor pressure i d.
Reactor power and pressure stabilize as the backup regulator take control i M , I i I "
__ _ _ _ _. _
I ' i l i ) REACTOR OPERATOR Page 30 l ' \\ f QUESTION: 049 (1.00) , Unit 2 is at 100% power with both TDRFPs in 3-element control, and the level setpoint at ! 36". If the selected levelinput channel fails downscale, which ONE of the following ! describes the expected plant response? j
1.
RPV water level will decrease, and the reactor will scram on low water level.
) a.
l b.
RPV water level will increase, Main Turbine will trip, and the reactor will ! )- scram due to the turbine trip.
j I The reactor will scram immediately, because the selected Narrow Range 1, c.
channel feeds both RPS channels.
! - i d.
No scram will result because reactor water level will be maintained above the' ! ! scram setpoint due to the combined effects of Recirculation Pump downshift, i } direct setpoint setdown and dynamic compensator.
I
I QUESTION: 050 (1.00) l
i Which one of the following : 1als will auto start the Standby Gas Treatment System (VG)
from it's normal standby linec ?
, . a.
RPV level O inches ! l b.
Drywell pressure 1.0 psig , c.
Fuel pool ventilation exhaust rad levels of 5 mR/hr l , d.
Reactor building ventilation exhaust rad levels of 10 mR/hr _ h .
l REACTOR OPERATOR Page 31 l l l QUESTION: 051 (1.00) The 1 A DG Cooling Water pump trips while the 1 A DG is running under 75% load during a surveillance. Assuming no operator action occurs, which ONE of the following describe the expected impact on continued 1 A DG operation? The 1 A DG will trip on high cooling water temperature.
a.
b.
The pump trip will directly actuate the DG lockout which will trip the 1 A DG off.
The 1 A DG SW cross connect AOV will automatically open, allowing SW to c.
cool the 1 A DG.
d.
The 1 A DG governor will runback the loadset to 10% which is within the DG's cooling capacity.
QUESTION: 052 (1.00) The Unit 1 NSO identified a need to quickly reduce power from 100 to 50%. The NSO selected the Power Reduction Mode on the RWM, and began selecting and driving cram array rods to position 00. One cram array rod would not latch at position 00, but settled to position 02 and latches. Which ONE of the following is the RWM response? a.
INSERT ERROR generated, ROD BLOCK applied.
b.
INSERT ERROR generated, no ROD BLOCK applied.
c.
WITHDRAWAL ERROR generated, ROD BLOCK applied.
d.
WITHDRAWAL ERROR generated, no ROD BLOCK applied.
- . +
, { REACTOR OPERATOR Page 32 , l OUESTION: 053 (1.00) l Which of the following will result in automatic closure of the Reactor Water Cleanup ,' (RWCU) blowdown flow control valve (G33-F033)? a.
Reactor water level at -53 inches.
b.
75 gpm RWCU system differential flow.
c.
142 psig pressure downstream of the flow control valve.
d.
145 F non-regenerative heat exchanger outlet temperature.
QUESTION: 054 (1.00) Which one of the following is used by the reactor recire flow control valve runback circuit to determine the number of operating reactor feed pumps? a.
MDRFP breaker position i b.
MDRFP discharge pressure i c.
TDRFP vacuum d.
TDRFP discharge pressure . l !
l !
j
. . __ _ _. _ _ _ _.. _ _ _ .__ _ _.___.-__._ _.
__._..._- ... _ _ ! "
I
l REACTOR OPERATOR Page 33 } QUESTION: 055 (1.00)
} Given that the unit was initially operating at 80% reactor power with "A" and "B" TDRFPs i
in 3-element AUTO, the following occurred: i - - The 'A' TDRFP tripped - Reactor water level is at 30" and lowering The lead HPU for the FCVs has tripped, the backup HPU has started and is - ] operating normally - Recirculation flow control is in the loop manual mode Which of the following is the CORRECT response of the Recirculation Flow Control , l System? The FCVs will i a.
be LOCKED UP.
, b.
be UNAFFECTED.
, c.
start to CLOSE.
b ' d.
RUNBACK to its minimum position.
i
i
QUESTION: 056 (1.00)
I RHR Loop "A" has automatically initiated in the LPCl mode on a high drywell pressure i j' signal, it is injecting to the core through F042A, RHR Loop "A" injection valve. Under the j present condition, which one of the following statements describes the "A" RHR system
response when its handswitch is taken to the OPEN position? a.
F027A and F042A will both open and stay open.
. a b.
F027A will open after a 10 minute time delay, c.
F027A will not move as long as F042A is open.
d.
F027A will open and F042A will automatically close when the full open limit is reached on F027A.
a
REACTOR OPERATOR Page 34 QUESTION: 057 (1.00) After a scram the NSO prepares to reset the scram by placing the Scram Discharge Volume (SDV) High Water Level bypass switch in the BYPASS position, but fails to depress the SDV Vent and Drain valves Test pushbuttons. Which ONE of the following will result from the NSO's actions? SDV Vent and Drain valves will: a.
OPEN without the scram valves CLOSING.
b.
OPEN without the scram valves OPENING.
c.
CLOSE without the scram valves CLOSING.
d.
CLOSE without the scram valves OPENING.
QUESTION: 058 (1.00) WHICH ONE (1) of the following statements correctly describes the shorting links that are used in Reactor Protection System (RPS)? a.
Installation of the shorting links makes SRM scrams available.
An SRM scram will occur if ANY SRM reaches it's Hi-Hi setpoint.
. b.
Removal of the shorting links makes SRM scrams available.
An SRM scram will occur if ANY SRM reaches it's Hi-Hi setpoint.
c.
Installation of the shorting links makes SRM scrams available.
An SRM scram will occur based on a coincident logic scheme of SRM channels reaching their Hi-Hi setpoints.
~ d.
Removal of the shorting links makes SRM scrams available.
An SRM scram will occur based on a coincident logic scheme of SRM channels reaching their Hi-Hi setpoints.
, ) e
_. -. . _ _ _ _... _.. _. _ _. - _ _ _ _. _.. _ _ _. _ _.... _ _ _. _. .__m.__ _ _.., ! , ! ! ! REACTOR OPERATOR Page 35 ! ! t L l QUESTION: 059 (1.00) l l-A reading of 20 on Range 9 of the intermediate Range Monitors (IRM) recorders would indicate WHICH ONE (1) of the following reactor power levels? l a.
0.8 %
i l b.
2.0% t l
c.
8.0% ' I i d.
20.0 % I QUESTION: 060 (1.00) The plant was operating at 60% reactor power when the following occurred: level began decreasing rapidly and is at -45" and lowering.
- drywell pressure rapidly increased to 1 psig and continues to rise - 'due to a malfunction, both HPCS RPV level 8 switches are locked up and - tripped at + 55,5" - .the RX VESSEL WTR LVL 8 Hi annunciator lit at panel H13-P601 A.
Which ONE of the following would cause HPCS to inject if the NSO manually started the HPCS pump from the control room? - a.
RPV level drops to -50" b.
A High Drywell Pressure signal is received by HPCS c.
The NSO depresses the Lo Level /Hi Drywell Pressure reset pushbutton.
d.
The NSO at panel H13-P601 takes the control switch for the HPCS INJECTION valve (E22-FOO4) to OPEN.
! l l '
_ - - _ _ .. _ _ _ _ _ _. - _ . _ . _.. _.. _ _ - _ - - _ _. _.___.-..._ _ I . t i i i ' I ! + REACTOR OPERATOR Page 36 l l i i s ! QUESTION: 061 (1.00) ' i Given the following plant conditions: i
l A small break LOCA has occurred ' - Dryweli pressure is 1.9 psig and rising i - RCIC is injecting -
HPCS has been secured by operator action I - Emergency RPV Depressurization has been initiated ! - i Which of the following conditions will result in the closure of RCIC Vacuum Breakers (E51-F080 and F086)? l l a.
Reactor Water Level (-129") l b.
Reactor Vessel Pressure (57 psig) c.
RCIC Exhaust Diaphragm Pressure (10 psig) d.
RCIC Equipment Area Differential Temperature (120*F) l .
QUESTION: 062 (1.00) l Which of the following activities does NOT require the use of electrical protection I equipment? l a.
A 480 VAC circuit breaker is being racked out.
b.
Breaker 1423 is being closed by local manual operation.
l c.
A GE 4160 volt circuit breaker is being placed in " Test."
- ! d.
A 250 VDC circuit breaker is closed by local manual operation.
! i i i [ i . .- , -, , , -,
. -. . . -- i REACTOR OPERATOR Page 37 QUESTION: 063 (1.00) You have been assigned to the training department for a two year rotational assignment.
During that time you have allowed your license to go INACTIVE. In accordance with 10CtR55.53, what must you do to restore your license to ACTIVE status? a.
Stand 7 eight-hour shif ts as the NSO.
b.
Pass an annual requalification examination.
c.
Notify the NRC that you 'are resuming watchstanding duties.
( d.
Stand 40 hours of watch under instruction at your watchstation.
QUESTION: 064 (1.00) Which ONE of the following is the NORMAL sequence that should be used for tagging a motor operated valve out of service (OOS)? Assume the valve is in the proper position.
a.
Tag the remote control switch OOS, open the power supply breaker for the valve and tag it OOS, tag the valve OOS.
b.
Open the power supply breaker for the valve and tag it OOS, tag the remote control switch OOS, tag the valve OOS, c.
Tag the valve OOS, open the power supply breaker for the valve and tag it OOS, tag the remote control switch OOS.
d.
Open the power supply breaker for the valve and tag it OOS, position the clutch to the manual position and tag the clutch OOS, tag the valve handwheel OOS.
- i l
m
l l l I REACTOR OPERATOR Page 38 , QUESTION: 065 (1.00) Which of the following work activities may be performed without a Nuclear Work Request (NWR)? Adjusting a protective relay setpoint for a Non-Safety related instrument.
a.
b.
Aligning a loose limit switch on an open MOV to give the proper indications in the control room.
c.
During a surveillance test the Alarm Typer is turned off due to a nuisance alarm saturating the typer.
d.
Tightening live load packing on a pump leaking at 20 drops per minute by an operator during his inplant rounds.
QUESTION: 066 (1.00) Conditions develop for which no action consistent with the Plant Technical Specifications will provide for adequate protection of the public health and safety. (NOTE: All shift personnel are present in the control room.)
Select the appropriate action for the reactor operator, Take only those actions that will comply with approved plant operating a.
procedures.
b.
Obtain approval from the Shift Operations Supervisor, then take actions to protect the public health and safety.
c.
Take actions to protect the public health and safety and subsequently inform the Shift Manager, d.
Obtain approval from the Shift Manager, then take actions to protect the public health and safety.
e , .
.,. .... _. . ... - - - -. - -... ~.- -. -. - - -.-- --..- - - i i ! l REACTOR OPERATOR Page 39 ! i , ' , , . I , l QUESTION: 067 (1.00) . Which one of the following statements is correct in respect to conduct of operations
associated with valve operations? I
i a.
Operators making their inplant rounds may loosen the packing on MOVs if } the packing is too tight.
!.
I b.
Operation of MOVs are limited to 30 starts within 1 minute followed by a 5 i minutes cooling off time.
i l c.
When placing the ECCS minimum flow valve in the shut position for maintenance, the associated ECCS pump is placed in pull-to-lock.
, s d.
When locally closing a motor operated Gate Valve with the Reactor Coolant ' system at normal operating pressure and temperature, mechanical assistance ! (cheater bars) may be used to prevent leakage past valve seats.
' QUESTION: 068 (1.00) ' Which ONE of the following RADIATION SIGNS is posted to indicate an area accessible to personnel in which there exist radiation at such levels that a major portion of the whole body could receive, in any one (1) hour, an exposure in excess of 100 mrem? < l l a.
RADIOACTIVE LSA l i b.
CAUTION-RADIATION AREA c.
DANGER-HIGH RADIATION AREA d.
CAUTION-AIRBORNE RADIOACTIVITY AREA - , i i
,
_ _ _ _ _ . -.. .. _ _ _ _ _ .. _ _ _ _. _ _ _ _. _., _ _ _. _ _. _ _ _ _ _ _ _ _ _ _ _ _. ! l
, I
! REACTOR OPERATOR Page 40 , t i i
QUESTION: 069 (1.00) j i ! You have been designated to give a plant tour for several visitors (not badged). You will , ' have the prime escort duties. What is the maximum number of people you are allowed to l escort into the plant? ! I ! ! l a.
No more than 5 in any area.
j b.
No more than 10 in any Owner-Controlled Area at one time.
, i l c.
Up to 5 in the Vital Areas, and up to 10 in the Protected Areas.
l 'd.
Up to 5 in the Vital and Protected Areas, and no more than 15 in the Owner-Controlled Area.
l ! i QUESTION: 070 (1.00) I
i Which one of the following describes why the off-gas mechanical vacuum pump should not i be used while operating above 5% of rated thermal power?
I a.
the mechanical vacuum pump cannot maintain condenser vacuum less than . the turbine trip setpoint.
i b.
there is a possibility of steam reaching the mechanical vacuum pump and ! causing damage to the blading, , c.
there is the potential formation of detonable concentration of hydrogen and ! oxygen in the untreated mechanical vacuum pump flow path, d.
the flow path through the mechanical vacuum pump and the off gas charcoal
adsorbers provides insufficient holdup time to allow radioactive gasses to ' decay.
, '
t I > a _., , .., -. _ _.- _ _ _ , ..,..
_-. _
i REACTOR OPERATOR Page 41 l l QUESTION: 071 {1.00) l The plant was operating at 95% power when an event occurred resulting in the following plant parameters: Reactor scram - RPV pressure 800 psig trending down - RPV level -70" trending down rapidly - - HPCS injecting into the RPV > The NSO armed and depressed the "A" RHR/LPCS Manual Initiation pushbutton to help control reactor water level. Under these conditions, which of the following LPCS lineups would occur? a.
LPCS Pump OFF. LPCS/RH Water Leg Pump OFF. No system flow.
b.
LPCS Pump OFF. Water Leg Pump running; taking suction from suppression pool and discharging back to the suppression pool.
c.
LPCS Pump running; taking suction from suppression pool and discharging to vessel through LPCS Injection Valve. LPCS Water Leg Pump running.
d.
LPCS Pump running; taking suction from suppression pool and discharging back to suppression pool through Minimum Flow Valve. Water Leg Pump is running.
.,
. _.... ~ m.
.___4__ . _ _. _ _. _. ._ ... _ _ _ _ _ _ _ _.. _ _ _ _ _... _ _ - _ !l i ! i REACTOR OPERATOR Page 42 l
l QUESTION: 072 (1.00) i The High Pressure Core Spray (HPCS) automatically initiated on a valid low reactor water ! level signal. The Unit Supervisor then directed the operator to manually override and close ! the HPCS Injection Valve (E22-F004) to maintain RPV water level at Level 4. What is ' required to reinstate the " automatic opening" feature of the HPCS Injection Valve? I a.
Drywell pressure rises to 3 psig and subsequently decreases below 1.69 { psig.
! I b.
The "Hi WATER LEVEL" reset pushbutton is depressed while reactor water l level is below + 55.5 inches.
l } ( c.
Reactor water level increases to above -50 inches and subsequently l decreases to less than -129 inches.
j d.' The "HI DRslWELL PRESSURE /LO WATER LEVEL'" reset pushbutton is depressed after the HPCS initiation signal clears.
l
l - ! , . . . , a ... ~ ... -... , __ _ .. , -.. - _ __
_ _ . i ,
i REACTOR OPERATOR Page 43 QUESTION: 073 (1.00) A unit startup is in progress with the reactor still suberitical. AFTER each rod withdrawal, neutron level willincrease for a period of time and then stabilize at a higher level. WHICH of the following describes this effect as the reactor gets closer to criticality? a.
The proximity to criticality has NO effect on how long it takes for neutron ! level to stop increasing following a rod withdrawal.
b.
The closer the reactor is to criticality, the LONGER it will take for the neutron level to stop increasing and level out following a rod withdrawal.
c.
The closer the reactor is to criticality, the LESS TIME it will take for the l neutron level to stop increasing and level cut following a rod withdrawal.
d.
The closer the reactor is to criticality, the less likely that neutron level will stabilize following a rod withdrawal. At a given reactivity, reactor power will continue to increase (after rod motion has ceased) until the reactor achieves criticality.
QUESTION: 074 (1.00) During steady state power operations, the APRM Gain Adjustment Factor (AGAF) for APRM-C reads 1.04 on a current OD-3. Under these conditions, the operation of APRM C is because , a.
conservative / APRM reading is LESS THAN actual power b.
conservative / APRM reading is GREATER THAN actual power c.
NON-conservative / actual power is LESS THAN APRM reading d.
NON-conservative / actual power is GREATER THAN APRM reading ! ! ! -
._ . . _ _. _. ~.. .. _.. - . .. _ , i ! l l l t ! ! REACTOR OPERATOR Page 44 - ! QUESTION: 075 (1.00)- The following plant conditions exist: , Automatic Depressurization System - actuated Reactor water level-( 100) inches and trending up . All RHR pumps - running , 105 second timer - timed out ! 7 ADS SRVs - open ' Drywell pressure - 1.5 psig and decreasing if both the DIV I and DIV 11 RPV LOW LEVEL RESET pushbuttons are momentarily depressed, which of the following describes the result on the Automatic Depressurization System? a.
SRVs will remain open.
P b.
' SRVs will close and remain closed.
c.
SRVs will close and then reopen after 105 seconds, d.
SRVs will close and then reopen after 450 seconds.
QUESTION: 076 (1.00) Suppression Chamber to Drywell differential pressure is maintained LESS than 0.5 PSID to prevent: a.
uncovery of the SRV dischargas a b.
loss of containment emergency vent capability - c.
structural failure of the containment downcomers d.
opening the suppression chamber to drywell vacuum breakers i - . .- .- _.
__
.__.-_. _ __... _..... _. _.. -. . _... _.. - - -... .. _ _ _ _, -.. _ -. _ _ _.. _ _ _ - ..! i , - ! ! REACTOR OPERATOR Page 45 ! I i
QUESTION: 077 (1.00)' .! l A Unit 1 shutdown is in progress per LGP 2-1 with the following conditions: t -' The reactor vessel pressure is 940 psig l The mode switch is in STARTUP ! - ! The main condenser low vacuum Main Steam Isolation Valve (MSIV) isolation signal will be automatically bypassed by taking the low condenser bypass switches to BYPASS... a.
OR by closing the turbine stop valves.
b.
OR by closing the turbine bypass valves.
c.
AND shutting the turbine stop valves.
l i d.
AND shutting the turbine bypass valves.
I
QUESTION: 078 (1.00) i i Which of the following would be an indication of an open or leaking safety relief valve? i l a.
Indicated Total Steam Flow INCREASE with a CONSTANT Generator Load.
) b.
Indicated Total Steam Flow CONSTANT with an l'NCREASE in Generator Load.
c.
Indicated Total Steam Flow DECREASE with a CONSTAl4T indicated heactor i Power.
d.
Indicated Total Steam Fbw DECREASE with a DECREASE in indicated Reactor Power.
I
. l -. .-. - -- - . -... -, - -..
, _,, J +--A - l , i
1 REACTOR OPERATOR Page 46 l i QUESTION: 079 (1.00) i The following plant conditions exist: 100 percent reactor power ' - - reactor water level is 36 inches ' MDRFP is OFF and the FRV is CLOSED - Both TDRFPs in 3-element control - - Level Setpoint is 36 inches if the "A" TDRFP flow transmitter rapidly fails UPSCALE, HOW will the plant respond to this event? (Assume no operator action.) Reactor level: a.
increases causing a turbine trip.
b.
stabilizes at a level below normal.
c.
stabilizes at a level above normal.
' d.
decreases causing a reactor scram.
QUESTION: 080 (1.00) Given the following plant conditions.
A Normal reactor shutdown is in progress - - Reactor pressure is 950 psig Reactor power is 10% - Which of the following Reactor Protection System automatic scrams is bypassed when the j mode switch is taken from "RUN" to "STARTUP"? - a.
APRM Thermal Hi-HI b.
Low CRD header pressure c.
Turbine Stop Valve Closure d.
Main Steam Line isolation Valve Closure l -
REACTOR OPERATOR Page 47 , QUESTION: 081 (1.00) If the scram discharge volume vent and drain valves do not go closed when a scram has occurred, which of the following would be the adverse consequence? a.
There will be a primary to reactor building leak.
b.
The CRD discharge path has insufficient back pressure.
c.
The control rod insertion speed will exceed design limits.
d.
The timer allowing scram reset after 10 seconds will not initiate.
QUESTION: 082 (1.00) Given the following plant parameters:
- Turbine Throttle Pressure 965 psig - Pressure Setpoint 935 psig.
Load Set 110 percent - ' - Load Limiter 100 percent - Max Combined Flow Limiter 120 percent - RRFC Individual Manual WHICH of the following describes the expected plant response if the Max Combined Flow Limiter is inadvertently reduced to 80%? a.
TCVs close and BPVs open. Rx power will control at SOE b.
TCVs open and BPVs remain closed. Rx power will control at 80% c.
TCVs open and BPVs open. The reactor will scram on MSIV closure.
d.
TCVs close and BPVs remain closed. The reactor will scram on high flux or ' pressure.
. - . , .. - - - - - l REACTOR OPERATOR Page 48 ! l QUESTION: 083 (1.00) l Given the following plant conditions: ! RPV pressure 450 psig - l Drywell temperature 250 F - ! Drywall pressure 20 psig - ! - Rx building temp 125'F - RPV level is increasing due to HPCS/RCIC injection ! ! Which of the following may be used to determine that water levelis at the Top of Active Fuel (TAF)? a.
Fuel Zone (only) b.
Wide Range (only) c.
Upset and Wide Range d.
Fuel Zone and Wide Range and Upset Range QUESTION: 084 (1.00) During a large break LOCA, vessel levelindication was lost and the SRO directed the initiation of ADS LAW LGA-05 "RPV FLOODING". The NSO reported back to the SRO that the SRVs did NOT open. Plant parameters at time of the event: reactor vessel pressure 800 psig - suppression pool water level-10 feet - suppression pool temperature 220*F - drywell pressure 69 psig - - drywell temperature 220 F WHICH of the following is the reason for the SRVs not opening? a.
Drywell temperature is too high.
b.
Containment pressure is too high.
I c.
Suppression pool water level is too low, d.
Suppression pool temperature is too high.
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QUESTION: 085 (1 00) ! During plant operations the following plant conditions were noted to occur over a 3 minute > period.
I Reactor pressure - decreased to 800 psig, now stable.
- - Reactor Water Level - 34 inches trending to normal.
l Reactor power - decreased 5%, now stable at 50% -
- Generator output - decreased to 250 Mwe from 500 Mwe.
j Reactor Protection Sys - No actuations have occurred.
. l - l '
l You should immediately: i
a.
Trip the main turbine.
, , b.
. Scram, Close MSIV's.
! r c.
Adjust reactor pressure control to increase pressure, d.
Increase Recirculation flow to increase power and pressure, i i i I i l I ! l ' _ _ .
.. - . . _ . - - - ! l REACTOR OPERATOR Page 50 QUESTION: 086 (1.00) The plant is operating at 25% reactor power when condenser vacuum begins to decrease.
! NO operator action occurs and Vacuum decreases to zero over a 30 rninute period. Which one of the following statements CORRECTLY describes the response of the plant to the decreasing vacuum? (Decreasing vacuum means pressure in the condenser is approaching atmospheric pressure.)
a.
The MSIVs will shut causing a reactor scram on MSIV position.
b.
The reactor will trip. There is a direct reactor scram on low condenser vacuum.
c.
The turbine will trip giving an RPS trip signal from low turbine trip oil pressure (TCV fast closure).
d.
The turbine will trip but SRVs will open to control reactor pressure. No reactor scram will occur.
_ l !
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' l I l QUESTION: 087 (1.00' i i L i , ' A loss of offsite power initially occurs on Unit 2 while the unit is shutdown. While j mitigating the Unit 2 problem the following occurred: l' ., - Unit 1 receives an ECCS initiation signal on high drywell pressure ! Unit 1 suffers a loss of offsite power ' - , The 1B diesel generator fails to start - i f l Assuming all other systems operates as expected, which of the following pump ! configurations is available for this event? UNIT 1 UNIT 2 ' r a.
LPCS pump LPCS pump, RHR pumps A,B,C l RHR pumps A,B,C HPCS pump ! , b.
LPCS pump RHR pumps A,B,C RHRB,C HPCS pump l ! l ! - c.
LPCS pump RHR pumps B,C l RHR pumps A,B,C HPCS pump d.
LPCS pump RHR pumps B,C l l HPCS pump HPCS pump.. '
- ,
i i i f .. I
[ ' l I ' !
.
. -.- r w-a w, - , . ., ,- -m , p -
.- ,. . - - --- -_- _ E l
REACTOR OPERATOR Page 52 QUESTION: 088 (1.00) While operating at 70% power, a Feedwater Control System malfunction has caused reactor vessel level to reach Level 8 (55.5 inches), resulting in a reactor scram and main turbine trip. The reactor feed pumps have FAILED to trip automatically and manually. You should: a.
manually CLOSE the MSIVs if vessellevel reaches 73 inches.
'b.
lineup and blowdown with the RWCU System to the main condenser if
vessel level reacnes 73 inches.
c.
lineup and blowdown with the RWCU System to the main condenser if vessel level reaches 125 inches.
l l d.
trip the condensate / condensate booster pumps to cause and additional . TDRFP trip signal to prevent overfilling the RPV.
' OUESTION: 089 (1.00) 'I Which of the following will trip the 1B Primary Containment Ventilation Supply Fan (VP)? a.
Undervoltage on Unit 1 Division 2 ESF bus and Drywell Pressure of 1.69 psig on Unit 1.
b.
Undervoltage on Unit 2 Division 1 ESF bus and i Drywell Pressure of 1.69 psig on Unit 2.
c.
Undervoltage on Unit 2 Division 2 ESF bus and Reactor Water Level of -129 inches on Unit 1.
j d.
Undervoltage on Unit 1 Division 1 ESF bus and Reactor Water Level of -129 inches on Unit 1.
.
_ _ .. - _ - _ _. - l r l REACTOR OPERATOR Page 53 . QUESTION: 090 (1.00) Which one of the following interlocks or trip functions is still operable from the Remote Shutdown Panel after placing all remote transfer switches in their " emergency" Position? a.
LLS for SRVs K and P if manually actuated b.
RHR "B" SDC Suction / Full Flow Test valve interlock c.
RCIC High Exhaust Diaphragm Pressure Isolation / Trip d.
RHR "A" SDC Suction / Suppression Pool suction valve interlock QUESTION: 091 (1.00) While operating at 100 percent reactor power, the following alarms are received: RBCCW EXPANSION TANK LVL Hl/LO (A103) - - RBCCW PMP AUTO TRIP (A101) In addition to the RBCCW System, WHICH of the following components require increased monitoring during this event? a.
Drywell Chillers b.
Reactor Recirculation Pumps c.
Fuel Pool Cooling Heat Exchangers d.
Auxiliary Building HVAC Heat Exchangers _ Y
._-_.._ _._._ _.-._ _.._ _ ._ _ _ _ _ _ _ _ _. _ _. _ _. _ _ _ _. _ _. _ - -. _ .. _. _ i' f REACTOR OPERATOR Page 54 , i QUESTION: 092 (1.00) While operating at 100% power, a complete loss of instrument Air (IA) occurs. WHICH of the following describes the correct failure positions for the listed valves as a result of the loss of IA? a.
MDRFP Feed Reg Valve - FAILS AS IS TDRFP Min Flow Valves - Fall OPEN b.
MDRFP Feed Reg Valve - FAILS OPEN TDRFP Min Flow Valves - Fall OPEN c.
MDRFP Feed Reg Valve - FAILS AS IS TDRFP Min Flow Valves - FAIL AS IS d.
MDRFP Feed Reg Valve - FAILS OPEN TDRFP Min Flow Valves - Fall CLOSED i ! QUESTION: 093 (1.00) The following conditions exist: l - A reactor startup is in progress The mode switch is STARTUP - The main turbine is tripped - A valid Group i isolation has occurred - No Auto Scram Signalis present- - Of those given, select the only signal that could have generated the Group iisolation: a.
Low reactor water level - b.
Low main steam line pressure c.
High main steam line radiation
d.
High main steam line tunnel differential temperature i ! ' l I i
._ _ _ _ ,,s.._.__
_ __ ! REACTOR OPERATOR Page 55 l l ' OUESTION: 094 (1.00) Which one of the following statements describes why you are required to take the control switches for both CRD pumps to start simultaneously when starting a second CRD pump? This is done to bypass thermal overload protection since the second pump a.
would NOT be started except for an emergency.
b.
This is done to delay the low suction pressure trip logic and prevent the operating pump from tripping on low suction pressure.
This is done to enable the CRD flow control valves gain change logic which c.
is needed to compensate for the flow characteristics of two pumps in simultaneous operation.
d.
This is done to enable the low suction pressure trip, which is normally ! bypassed, since damage from low suction pressure CANNOT occur if only
one pump is running.
QUESTION: 095 (1.00) The RCIC turbine was manually started by an NSO to perform an operability surveillance.
The "B" man assigned to locally monitor the RCIC room reported steam leakage near the RCIC turbine governor valve. The NSO immediately depressed the RCIC isolation pushbutton. This will: a.
not cause a RCIC system isolation.
b.
cause an isolation by closing the steam shutoff (F045) only.
i c.
cause an isolation by closing the outboard isolation valve (F008) only.
d.
cause an isolation by closing the inboard (F064) and outboard (F008) isolation valves.
.
, . _ _ _. _ _. _. _ _. .. _.. .. _..... _ _ - _ _ _ __ l \\ r > i REACTOR OPERATOR Page 56 - l \\ l QUESTION: 096 (1.00) . Which ONE of the following statements is correct concerning the RHR Service Water
System Process Radiation Monitor (PRM)? j J l a.
Only required to be in service when in the Shutdown Cooling Mode of l l Operation, i i-l ,
b.
Provides indication and alarm functions only, i.e. NO automatic isolations.
' i ! c.
Measures the cumulative activity release to the environment from the RHR ! ! Service Water System.
' , d.
Measures the activity of the fluid which passes through the shell side of the ! RHR heat exchangers.
I . i OUESTION: 097 (1.00) l i . Given the following plant conditions: ! assume no operator actions - reactor operation in condition # 4 - RHR loop B in shutdown cooling mode { - reactor moderator temperature at 140'F and holding j -
Which ONE of the following ' describes the expected behavior of the reactor water level and moderator temperature upon receiving a trip of the operating RHR service water pump (s)? l a.
reactor level decreases as moderator temperature increases
b.
reactor level increases as moderator temperature increases - c.
reactor level decreases as moderator temperature decreases ] d.
reactor levelincreases as moderator temperature decreases ! i
l ~ ! l l l ._ , . . - . -. -
. _. _ _ ,,. . _ __ _.. _ _ _ - _ -. ._. .. _ _. _ r I t l I i ' REACTOR OPERATOR Page 57 ! i QUESTION: 098 (1.00)
Which of the following situations is ALLOWED, concerning the manipulations of controls? { NOTE: Assume that the NSO has full knowledge and consents with the manipulations, and i all manipulations are per an authorized procedure.
t l a.
A license candidate pulls control rods to criticality under direct supervision of
' the NSO.
! ! , i b.
A license candidate repositions the reactor mode switch while the NSO is I checking Secondary Containment parameters.
l ! c.
With Unit 1 at 25% power, an l&C technician conducting a surveillance test, l adjusts recirculation flow while under the constant direction of the NSO.
! d.
With Unit 1 in mode 3, a system engineer conducting a post maintenance i test, depresses the HPCS manual initiation pushbutton while under the ' constant direction of the NSO.
- l
! l l QUESTION: 099 (2.00) I ! MATCH the LGA procedure (s) in Column B which are entered as a result of the conditions
described in Column A.
, Column A Column B a. Reactor Scram 1. LGA-01 l b. Suppression pool temp 105'F 2. LGA-01 and LGA-03 , - c. Drywell pressure greater 3. LGA-03 than 1.69 psig 4. LG A-02 d. Reactor Bldg differential
pressure equal to 0 inches i of water i l ( * * * * * * * * * * END OF EXAMINATION * * * * * * * * * * ) ' $
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-
!- t
ANSWER: 001 (1.00) ANSWER: 006 (1.00) ! l b c.
l REFERENCE: REFERENCE: ' LGA 01,02,03 & 04 , LRP-1000-1.
i l 295029G011 ..(KA's) i ANSWER: 002 (1.00) 294001K103 ..(KA's) l l a.
ANSWER: 007 (1.00) REFERENCE: a.
l REFERENCE: i RVI Ch 3 system description, pg 28, sect ' E LOA-RD-06 j RVI SYSTEM EKO 12.d.1 l 295028A203 ..(KA's) 201001G013 ..(KA's) l ANSWER: 003 (1.00) ANSWER: 008 (1.00) i b.
b.
j REFERENCE: REFERENCE: l ! L.P. LGA-03 Primary Containment LSD #17, pg.22 ! Control rev.11 page 52 l t 201002K301 .. (KA's) l 295026K301 ..(KA's) ' ANSWER: 004 (1.00) ANSWER: 009 (1.00) j
a.
b.
REFERENCE: REFERENCE- , LAP-100-40 LOP-SC-02, Rev 7, Section F.1.a, p.2.
294001A102 ..(KA's) 211000 GOO 9 ..(KA's) ANSWER: 010 (1.00) - ANSWER: 005 (1.00) c.
b.
REFERENCE: REFERENCE: LSD #12, pp.17,18; LOP NR-02 LAP-1600-3 REV 8 215003K401 ..(KA's) 294001A106 ..(KA's) ' l I ! - -,.. ., _ , ..,, ,,. _ _ _
- I f REACTOR OPERATOR Page 59 ANSWER: 011 (1.00) ANSWER: 016 (1.00) f c.
b.
REFERENCE: REFERENCE: LSD #41, Fig. 41-01b and text; LOP-LOP RH-07, p. 3: RI-05 205000G010 ..(KA's) 217000A403 ..(KA's) ! ANSWER: 012 (1.00) ANSWER: 017 (1.00) b.
b.
REFERENCE: REFERENCE: l LDS #37, pp. 4, 6, 22 LSD #17, pg.35; LSD #8, pp.18,20; LOA-RD-06 l 218000K606 ..(KA's) l 214000A202 ..(KA's) ! ANSWER: 018 (1.00) ANSWER: 013 (1.00) a l d.
REFERENCE: REFERENCE: LSD #15 pp.14,15; LOP NR-05, pp.1, LSD #51 pg. 22; LOP VG-01
, f l 261000A301 ..(KA's) 215002A203 ..(KA's) ANSWER: 014 (1.00) ANSWER: 019 (1.00) a.
d.
REFERENCE: REFERENCE: l LOP-RW-01; LOP RW-02; LSD #18, pp.
LOP-RH-13, pg. 3; LSD #39, pg.18 7,A2.
-- l 219000A101 ..(KA's) t 201006K512 ..(KA's) ANSWER: 020 (1.00) { ANSWER: 015 (1.00) b.
b.
REFERENCE: ! REFERENCE: ' LSD-29B, TDRFP Speed Control LSD #5, pg.14A; LOP RR-05, pg. 4 262002G009 ..(KA's) f 202001K413 ..(KA's) .
. ..... _ _ _. _ -. _ _. _ _. _ _. _ _.. _ . . _ _ _ _ _ . _ l l REACTOR OPERATOR Page 60 - l ANSWER: 021 (1.00) ANSWER: 025 (1.00) c.
c.
REFERENCE: REFERENCE: -
LaSalle Licensed Operator System LOA DC-03; Tech Spec 3.3.7.6; LSD Description Manual, Chapter 59, Control
- 11, pp.17, 24, 27; LSD #12, pp.10, Room Ventilation, page 38, Section V.
11,'18; LSD #43 App. B, pp. 48, 49 290003K501 ..(KA's) 295004K303 ..(KA's) ANSWER: 026 (1.00) ANSWER: 022 (1.00) d.
c.
REFERENCE: REFERENCE: Updated SAR 15.1.1, Loss of Feedwater LSD #67, pg.10 and Fig. 67-06 Heater (p.15.1-3) LOA-FW-01, Rev 12, March 22,1988, Loss of a feedwater heater (s) 234000A302 ..(KA's) LOA-1(2)PM03J-B401, Rev 4, Oct,1989, ANSWER: 023 (1.00) HP HTR16(26) EXTR STEAM CHECK VLV b.
NOT OPEN ! REFERENCE: 295014A107 ..(KA's) LOA TG-101, LOSS OF TURBINE ANSWER: 027 (1.00) GENERATOR, LOAD GREATER THAN a r 25 % REFERENCE: 295005G011 ..(KA's) LOA WR-101 ANSWER: 024 (1.00) a REFERENCE: 295018K202 ..(KA's) ANSWER: 028 (1.00) T.S. 3.4.1.5, Thermal Hydraulic Stability b - REFERENCE: 295001A201 ..(KA's) LSD #49, Appendix B, pg. 54; LOA VR-101,201; LOA PC-101 288000K402 ..(KA's) . ,
! I l l REACTOR OPERATOR Page 61 i ANSWER: 029 (1.00) ANSWER: 034 (1.00) b.
d REFERENCE: REFERENCE: T.S. 3.3.1, ACTION 1 LOA AR-101,1H13-P601 B110 295015K204 ..(KA's) 295023K203 ..(KA's) ANSWER: 030 (1.00) ANSWER: 035 (1.00) b b.
REFERENCE: REFERENCE: LSD #21, pg.10 LG A-02.
295025K309 ..(KA's) 295035G011 ..(KA's) ANSWER: 031 (1.00) ANSWER: 036 (1.00) a b.
REFERENCE: REFERENCE: EOP Lesson Plan #7 pp. 43,44 LGA-02 Entry Conditions 295025A205 ..(KA's) 295036G011 ..(KA's) ANSWER: 032 (1.00) ANSWER: 037 (1.00) a.
a.
REFERENCE: REFERENCE: LSD #26, EHC Logic Diagram, pg. 38; L.P. LGA-10 Failure to Scram Rev. 3, LGP 2-1 page 26 295037K305 ..(KA's) 295005K307 .. ( KA's)- ANSWER: 033 (1.00) ANSWER: 038 (1.00) c.
a.
REFERENCE: REFERENCE: LGA-03, Pool Temperature LSD #39, pg.26 295013G011 ..(KA's) 226001A401 ..(KA's) .
. _. ~ _ _ _ . _ _ _ _ .. _ _ _ _ _ -. _ _ _ _ _ _ _ _.
. . l I < t ! t ' REACTOR OPERATOR Page 62 l ANSWER: 039 (1.00) ANSWER: 044 (1.00) ! c.
b ! REFERENCE: l REFERENCE:
LOA HY-101, pg. 4; LOP HY-07,
l Attachment A LSD #70, pg.18; LOP CO-01, pg. 3 j ! i
245000K507 ..(KA's) 286000A304 ..(KA's) ANSWER: 040 (1.00) ANSWER: 045 (1.00) 5.
b.
REFERENCE: q REFERENCE: l Lasa!!e Unit 1 Tech Spec Definition LSD #42, pp. 20, 22, 59 1.39, Page 1-7 LSD #51, pg.18 262001K602 ..(KA's) ANSWER: 041 (1.00) 290001K603 ..(KA's) d.
ANSWER: 046 (1.00) REFERENCE: d.
REFERENCE: LOA DC-03; LSD #43, Appendix B, pg.
-49 LSD #16, pg.15; LSD #49, Appendix C, LOP PC-03 263000K201 ..(KA's) ANSWER: 042 (1.00) 215001K401 ..(KA's) b.
ANSWER: 047 (1.00) REFERENCE: d.
REFERENCE: LSD #32, pg. 50 and Fig. 32-1 LGP 1-3, REV 27, p.14; - ! 271000A102 ..(KA's) ' ANSWER: 043 (1.00) 290002K101 ..(KA's) c.
ANSWER: 048 (1.00) j REFERENCE: a.
LSD #72, pg. 44; LOP PR-03, pg.1: LOA REFERENCE: PR-101 pg. 9 LSD #26 pg. 44, EHC logic diagram.
' 272000K305 ..(KA's) 241000A107 ..(KA's) - .. - . _ -.
REACTOR OPERATOR Page 63 ANSWER: 049 (1.00) ANSWER: 054 (1.00) b.
a.
REFERENCE: REFERENCE: LOA-FW-01: LSD #31, pg. 32 LOP-RR-07, Ch 6, p12 259001K108 ..(KA's) ANSWER: 050 (1.00) 202002K111 ..(KA's) d.
ANSWER: 055 (1.00) REFERENCE: d.
REFERENCE: LSD #51 pg.12; LOP VG-01 LSD #6, pp.16,32; LOP RR-11,pg.4; LOR H13-P602-A101 261000K401 ..(KA's) ANSWER: 051 (1.00) a.
202002A108 ..(KA's) REFERENCE: ANSWER: 056 (1.00) LOR-1DG03J-2-1; LSD #47 pg.29; LOP c.
DG-02 REFERENCE: LSD #39, RHR 264000K607 ..(KA's) 203000A308 . (KA's) ANSWER: 052 (1.00) ANSWER: 057 (1.00) d.
a.
REFERENCE: REFERENCE: LOP-RW-01, pg.17; LSD #18, pg.18 LGP 3-2, REV 36; LSD #20, pg.18 201006K101 ..(KA's) 212000K106 ..(KA's) ANSWER: 053 (1.00) ANSWER: 058 (1.00) c.
b.
REFERENCE: REFERENCE: LOP-RT-09, page 2, section E.1 LSD #20, pg. 23; LOP NR-01 i 204000A304 .. (K A's) 212000K502 ..(KA's) i -
._..___._m.
_ _. _ _. _ _. _ _. _...._.._.__- ._.. _.__.
- _....... .. _... _.. _ _ -. - ,
i
REACTOR OPERATOR Page 64 ! ANSWER: 059 (1.00) ANSWER: 064 (1.00) c.
a.
REFERENCE: REFERENCE: t
LSD #12 Lap-900-48, Rev 4, p.9, Attachment A.
t l 215003A401 ..(KA's) 294001K102 ..(KA's) i ! ANSWER: 065 (1.00) ! c i ANSWER: 060 (1.00) REFERENCE: I a.
REFERENCE: LAP 1300-7, LAP 100-14, LAP 1600-2, LAP 300-31 ! LSD CH, 38, pg. 24 294001A102 ..(KA's) 216000K304 ..(KA's) ANSWER: 066 (1.00) l ANSWER: 061 (1.00) d.
! b REFERENCE: ! l REFERENCE: 10CFR50.54(x) and 10CFR50.54(y) l LSD CH. 41, pg. 26; LSD CH. 49, LAP-1600-2 pages 6-7.
l Attachment B, pg. 57 294001A109 ..(KA's) 217000A203 ..(KA's) ANSWER: 067 (1.00) ' ANSWER: 062 (1.00) c d.
REFERENCE: REFERENCE:
LOP-AP-05, LOP-AP-12.
LAP 1600-2, LAP AA-04.
- 294001K107 ..(KA's) ANSWER: 063 (1.00) 294001K101 ..(KA's) d.
ANSWER: 068 (1.00) l REFERENCE: c.
t REFERENCE: j 10 CFR 55.53.
10CFR2O p 294001A103 ..(KA's) l 294001K103 ..(KA's) ! - ~, - -,, - - -,... , ,. -, .,. ,
.., -.. - . _. -. _ - -. - = _ - - - - ---.- -.--.. -. - .- . _ -.. -. _ I.
' ! l ' REACTO9 OPERATOR Page 65 l-ANSWER: 069'(1.00) ANSWER: 074 (1.00) c d.
REFERENCE: REFERENCE: LAP 1100-3 LOS-AA-S1, Attachment B; LSD #14, pg.
294001K105 ..(KA's)
ANSWER: 070 (1.00) l c 215005A208 ..(KA's) l REFERENCE: ANSWER: 075 (1.00) l b.
' Chapter 32, Off Gas System, LOP l OG-01.
REFERENCE: { l LSD #37, pp. 6,8; Figs. 37-02, 37-04, ' l 294001K115 ..(KA's) ADS logic j ANSWER: 071-(1.00) d.
218000A206 ..(KA's) l REFERENCE: ANSWER: 076 (1.00) d.
LSD #38, pg.28 REFERENCE:
LOP-VO-07, Rev 4, Section F.4.b, p.2.
209001K201 ..(KA's) l ANSWER: 072 (1.00) d.
223001K501 ..(KA's) REFERENCE: ANSWER: 077 (1.00)
c.
LSD #36, pg. 23; LOP HP-04 REFERENCE: l LSD #49, Appendix B, pg. 53 209002A403 ..(KA's) - ANSWER: 073 (1.00) b.
223OO2A403 ..(KA's) REFERENCE: ANSWER: 078 (1.00) ' l c.
BWR Reactor Theory Manual, Chapter 7 REFERENCE: " Operational Physics", Section ll.B, p.12 of 49.
LOA-SRV-01 , l i 215004A102 ..(KA's) 239002A101 ..(KA's)
i - , -- _- . _. ,_,
..., _ -. _.
. _. _ _ _. _ _ _ _ _ _ _.... _. - m.
_._ _. _. _.. _.._.__ _ _ . _ _ _ _.. _ _ _ _ _ _. ,
I ' REACTOR OPERATOR Page 66 l ANSWER: 079 (1.00) ANSWER: 084 (1.00) d.
b.
REFERENCE: i REFERENCE: LGA 03; LGA 05; EOP lesson plan #5, LSD #31 pg. 32 pg.34 259002K301 ..(KA's) 295024K208 ..(KA's) ANSWER: 080 (1.00) ANSWER: 085 (1.00) d.
b.
REFERENCE: REFERENCE: ) LSD #20, pg. 22; LOP AA-03, pg. 6 LOA RP-102, pg.3; LSD #21, pg. 39, LSD #20 , 239001A301 ..(KA's) ANSWER: 081 (1.00) 295037G010 ..(K A's) a.
ANSWER: 086 (1.00) REFERENCE: a.
REFERENCE: CRD(H) system description, pgs 14 & 15, sect K LaSalle Technical Specifications, Tables LGP-3-2, pg 4, sect 8.a.
2.2.1-1 and 3.3.2-2; CRD(H) EKO 6.d.9.
LSD #20, pg. 52; LSD #21, pg. 24; LSD #23, pg. 34; LSD 295006K203 ..(KA's)
- 26,pg.20 ANSWER: 082 (1.00)
d.
REFERENCE: 295002K201 ..(KA's) LSD #26A, EHC Logic Diagram ANSWER: 087 (1.00) - c 295007K201 ..(KA's) REFERENCE: ANSWER: 083 (1.00) a.
LSD #42, pp.14-22 and Fig. 42-02; LSD
- 47,pp.33,34 REFERENCE:
Detail LGA-D1 of LGA-01; LSD #3, pg.
295003A102 ..(KA's)
295009A201 ..(KA's) . ,4 ,,- - -.In-
! REACTOR OPERATOR Page 67 l ANSWER: 088 (1.00) ANSWER: 093 (1.00) a.
d.
REFERENCE: REFERENCE: LOA-1 H13-P603-A310 Tech Specs Tables 2.2.1 -1, 3.3.2-1, 3.3.2-2; LSD #20, pp. 20, 22; 295008A103 ..(KA's) LSD #21, pp. 22, 23 ANSWER: 089 (1.00) a.
REFERENCE: 295020A206 ..(KA's) ANSWER: 094 (1.00) LOP VP-10, pg. 4; LSD #52, pp.14, 20, b 26,32; LSD #42, pg. 64 REFERENCE: 295012A102 ..(KA's) LOP-RD-03, pp. 2, 3; LSD #8, pg. 30 ANSWER: 090 (1.00) b.
REFERENCE: 295022G007 ..(KA's) ANSWER: 095 (1.00) RHR SDM, Ch 74 a.
REFERENCE: 295016K201 ..(KA's) LSD #41 ANSWER: 091 (1.00) 295034A201 ..(KA's) b.
ANSWER: 096 (1.00) REFERENCE: b.
REFERENCE: LOA-WR-101, pp. 8, 9; LOA 1 H13-P602-A205, A304, B205,8304 LaSalle License System Description, Chapter 72, " Process Radiation Monitoring System", Appendix 295018K101 ..(KA's) 8, Section XI.H., p.70.
, ANSWER: 092 (1.00) a.
REFERENCE: 295038K206 ..(KA's) LOA-IA-101, Attachment B, pg.14
295019K203 ..(KA's) l .
. . _ -.. _. - _ < <. _.... .. _. _ _ . _..... _ _. _ _ _ _. _ _ _ _ _ _. _.. _. _. _ _ _... . i ! REACTOR OPERATOR - Page 68 i . I ANSWER: 097 (1.00) i b.
! REFERENCE: LaSalle Licensed System Description
- 39, Residual Heat Removal
. LOA-RH-03, Rev 4, Jan 21,1989, Loss of RHR Service water i 295021K101 ..(KA's) - l i i ANSWER: 098 (1.00) l a.
REFERENCE: ' i F LAP-1600-2, Ops Memo 2 i
294001A109 ..(KA's) ' f ANSWER: 099 (2.00) i a.1 - ! i b. 3 c. 2 + d. 4 REFERENCE:
1.
LGA-01 THRU 03.
j i l . 295010G011 ..(KA's) . - t '! . ! ,
i i i iy
't l l I ( * * * * * * * * * *. EN D O F EX AMIN ATIO N * * * * * * * * * * ) ! -
l I ! i . i ' __ _ _ - - - - , - _ _ _ _. - .- .,
_ _ _ .. - REACTOR OPERATOR Page 69 ANSWER KEY MULTIPLE CHOICE 023 b ' 001 b 024 a i ' 002 a 025 c 003 b 026 d 004 a 027 a 005 b 028 b
006 c 029 b 007 a 030 b 008 b 031 a 009 b 032 a 010 c 033 c 011 c 034 d 012 b 035 b 013 d 036 b 014 a 037 a 015 b 038 a ~ 016 b 039 c 017 b 040 b 018 a 041 d 019 d 042 b 020 b 043 c l ! 021 c 044 b , 022 c 045 b , l l
.- - - - . _-.- - - . _ . - _ _ =. .. _. REACTOR OPERATOR Page 70 ANSWER KEY l
i MULTIPLE CHOICE 068 c 046 d 069 c 047 d 070 c 048 a 071 d 049 b 072 d 050 d 073 b l 051 a 074 d 052 d 075 b 053 c 076 d 054 a 077 c 055 d 078 c 056 c 079 d ' 057 a 080 d 058 b 081 a < I 059 c 082 d i 060 a 083 a ._ 061 b 084 b 062 d 085 b l 063 - d 086 a I ' 064 a 087 c l 065 c 088 a 066 d 089 a 067 c 090 b i , .
. _ _, _ .. - ._ - - i ! i REACTOR OPERATOR Page 71 !
> ! ANSWER KEY MULTIPLE CHOICE l 091 b
092 a 093 d 094 b 095 a 096 b 097 6 098 a 099 MATCHING a 1 b 3 c 2 d 4 _ (* * * * * * * * * * END OF EXAMINATION * * * * * * * * * *) l l
-
. _..~
-
_ - - _ - .... . - - -. , ,
l l l ! - ' ES-401 Site-specific Written Examination Form ES-401-1 l Cover Sheet i i U. S. NUCLEAR REGULATORY COMMISSION
SITE-SPECIFIC L WRITTEN EXAMINATION ! l APPLICANT INFORMATION i Name: MASTER EXAMINATION Region: 111
Date: April 24,1997 Facility / Unit: LaSalle Co. Station , Units 1 & 2 i License Level: SR0 Reactor Type: GE l k i INSTRUCTIONS l l Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated ' in parentheses after the question.
The passing grade requires a final ) grade of at least 80 percent.
Examination papers will be picked up 4 hours ' after the examination starts.
All work done on this examination is my own.
I have neither given nor l received aid.
> Applicant's Signature RESULTS ~ Examination Value 100.0 Points Applicant's Score Points ! Applicant's Grade Percent ! [ Examiner Standards 6 of 7 Rev. 7. January 1993 ... .. .- - . .
_ . . . i l l SENIOR REACTOR OPERATOR Page 2 ANSWER SHEET l Multiple Choice (Circle or X your choice) i If you change your answer, write your selection in the blank.
l MULTIPLE CHOICE O23 a b c d _ 001 a b c d 024 a b c d l 002 a b c d 025 a b c , _ 003 a b c d 026 a b c d,_ I _ 004 a b c d 027 a b c d
005 a b c d 028 a b c d 006 a b c d 029 a b c d 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d
011 a b c d 034 a b c d j 012 a b c d 035 a b c d j 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d i .. 016 a b c d 039 a b c d 017 a b c d 040 a b c d ,
018 a b c d 041 a b c d j i 019 a b c d 042 a b c d ' 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d - .
. . .-. . . _. ._ - = _ -. . _. _ l i SENIOR REACTOR OPERATOR Page 3 ANSWER SHEET
Multiple Choice (Circle or X your choice) ' l If you change your answer, write your selection in the blank, ! MULTIPLE CHOICE 068 a b c d - ( l 046 a b c d 069 a b c d , 047 a b c d 070 a b c d 048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d _ 074 MATCHING 052 a b c d a 053 a b c d b 054 a b c d c 055 a b c d d 056 a b c d _ MULTIPLE CHOICE 057 a b c d 075 a b c d 058 a b c d 076 a b c d 059 a b c d 077 a b c d 060 a b c d 078 a b c d . 061 a b c d 079 a b c d 062 a b c d 080 a b c d 063 a b c d 081 a b c d 064 a b c d 082 a b c d l 065 .a b c d 083 a b c d l 066 a b c d 084 a b c d l 067 a b c d 085 a b c d -
.. . .,-4M SENIOR REACTOR OPERATOR Page 4 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blank.
MULTIPLE CHOICE 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 a b c d _ 092 a b c d _ 093 a b c d _ 094 a b c d _ 095 a b c d _ 096 a b c d _ 097 a b c d _ 098 a b c d _ 099 a b c d _ .. l I l ( * * * * * * * * * * END OF EXAMINATION * * * * * * * * * *) i l -
_.
. - ~ . - - - ! l i ! i SENIUM HEAGIUH OPEHAIGH Fage b ES-402 Policies and Guidelines Attachment 2 for Taking NRC Written Examinations
i l 1.
Cheating on the examination will result in a denial of your application and could result in more severe penalties.
< 2.
After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination.
3.
To pass the examination, you must achieve a grade of 80 percent or greater.
4.
The point value for each question is indicated in parentheses after the question number.
5.
There is a time limit of 4 hours for completing the examination.
6.
Use only black ink or dark pencil to ensure legible copies, 7.
Print your name in the blank provided on the examination cover sheet and the answer sheet.
8.
Mark your answers on the answer sheet provided and do not leave any question blank.
9.
If the intent of a question is unclear, ask questions of the examiner only.
10.
Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate I even the appearance or possibility of cheating.
11.
When you complete the examination, assemble a package including the examination questions, examination aids, and answer sheets and give it to the examiner or proctor. Remember to sign the statement on the examination cover sheet.
12.
After you have turned in your examination, leave the examination area as defined by the examiner.
i ! -
. _ _ _ . . - .. _ -, . , l \\ l ! ' l SENIOR REACTOR OPERATOR Page 6 l l . l QUESTION: 001 (1.00) Given the following conditions: - reactor power at 85% steady-state ! 3-element cor,,rol in service - ! recirc system operating in loop manual mode -
i Which ONE of the following describes the behavior of the reactor if a loss of feedwater i heater 16A occurs due to closure of the steam extraction stop check valve? Core power will... , l l a.
remain constant at present steady state level.
b.
decrease, then return to the pre-transient power level.
c.
initially decrease, then stabilize at a lower steady state level.
d.
initially increase, then stabilize at a higher steady state level.
) l l QUESTION: 002 (1.00) \\ The reactor is operating at 100% rated power when the running RBCCW pump and the running CRD pump auto trip. Attempts to restart the pumps and the standby equipment i has failed. Based on this, you should: ) a.
trip both reactor recirc pumps and scram the reactor.
b.
downshift both reactor recirc pumps and insert cram array control rods.
c.
reduce core flow to 49 M#/hr and downshift the reactor recirc pumps if seal failure occurs.
d.
monitor RR pump temperature and downshifting RR pumps only if motor j bearing temperature exceeds 200 F.
i I l l' l - i
SENIOR REACTOR OPERATOR Page 7 QUESTlON: 003 (1.00) Unit 1 reactor building ventilation exhaust radiation is increasing and reaches 10 mr/hr.
Select the expected automatic action, a.
Standby Gas Treatment (VG) would start on UNIT 1 ONLY.
b.
Reactor building supply and exhaust dampers (VR) close on BOTH units, c.
Purge Filter Train Suction isolation Dampers (VQ) close on Unit 1 ONLY.
d.
Nitrogen inerting and Makeup Isolation Valves (VQ) close on BOTH units.
QUESTION: 004 (1.00) The Reactor was operating at full power when an I&C tech started conducting manual scram pushbutton testing. The l&C tech asked the NSO to arm and depress the A1 and A2 manual scram pushbuttons. No channel A solenoid lights extinguished. The I&C tech then asked the NSO to arm and depress the B1 and B2 manual scram pushbuttons without attempting to reset any scram signals from channel A. No channel B solenoid lights extinguished. The action to be taken under these conditions is to: a.
enter LGA-10, Failure to Scram.
b.
take action to be in HOT SHUTDOWN within 12 hours.
c.
lock the reactor mode switch in the SHUTDOWN position within one hour, d.
immediately run recirc to minimum, trip the main turbine and place the i reactor mode switch in SHUTDOWN.
..
i SENIOR REACTOR OPERATOR Page 8 l QUESTION: 005 (1.00) What is the purpose of the Low-Low Set feature of the Safety Relief Valves (SRVs)? a.
Minimize SRV fatigue b.
Minimize containment fatigue c.
Prevent exhausting the drywell pneumatic cupply (IN) d.
Prevent thermal overload of the SRV's pneumatic actuator solenoids QUESTION: 006 (1.00) LGA-05, RPV FLOODING states, if you cannot get 5 or more SRVs open or cannot hold RPV pressure 52 psi or more above suppression chamber pressure, then flood the drywell.
Why is it necessary to maintain RPV pressure 52 psi ABOVE suppression chamber pressure during the vessel flood? a.
To ensure decay heat is adequately removed.
b.
To ensure collapse of steam voids in the reactor.
c.
To prevent closure of SRVs, resulting in less than 5 open SRVs.
d.
To prevent the intrusion of non-condensible gases from the suppression I chamber into the reactor vessel.
. ,
-. . . _.. - .. - _. -.. -. -... -. - -.. . . _. .... -., - -,.. .._. ;
!
a ! l I ' Page 9 f SENIOR REACTOR OPERATOR i- ! t
QUESTION: 007 (1.00) l
. l During an ATWS, why is the operatcr directed to delay plant cool down until the reactor is
shutdown without boron injection or until the cold shutdown weight of boron has been ' l injected? , i a.
The cool down will add positive reactivity, taking the reactor critical with no i method available to shut down the reactor.
b.
Core flow may become restricted due to the cooler water causing the boron ' to come out of solution and depositing on core surfaces.
Higher reactor pressure helps mixing of the boron due to the higher steaming [ c.
rates and minimizes the time to complete the shutdown.
' d.
Lowering reactor pressure would cause the water in the reactor to flash to steam, thus lowering the boiling boundary and increasing the chance to , . damage fuel.
' QUESTION: 008 (1.00) ! \\ On increasing suppression pool temperature, a manual Scram is required prior to exceeding what temperature? i a.
95*F b.
105*F c.
110*F d.
120 F - . m. - a -. - e . - - - -
_. - _ . _ _. _.. _ _ _. _ _ _ _. _ _.. _ __.. _. - -. ! i
! SENIOR REACTOR OPERATOR Page 10 ! QUESTION: 009 (1.00) The NSO in the control room has just received 1H13-P601-B110, "RB RAD Hl." Upon ' investigation it is found that 1D21-K601T, "RB 740 - RB CRD Storage" on control room panel 1D21-P600 indicates 40 mr/hr. Its associated amber "HlGH" light is illuminated.
. Which of the following statements is correct?
The ARM signal may be erroneous. Contact Rad Protection to verify dose j a.
readings in the area and troubleshoot and repair the area rad monitor as required.
j b.
The amber light on the ARM is a warning light indicating that the area in the , RB on the 740 elevation is increasing and precautionary action should be taken to restrict access to the area.
c.
The amber light is indicating an airborne high radiation condition. Actions , should be taken to evacuate all personnel, isolate ventilation to the ares and notify Rad Protection to initiate air samples.
d.
The ARM is indicating an area high radiation condition. All personnel in the ! area should be evacuated and access restricted. Entry into Secondary Containment Control, LGA-02, should be checked. Notify Rad Protection to survey and sample.
QUESTION: 010 (1.00) WHICH ONE (1) of the following conditions is an entry condition into LGA-02, " Secondary Containment Control?" a.
Fuel Pool exhaust 8 mr/hr ' b.
Reactor building Differential pressure at +0.5" we c.
Steam pipe tunnel area temp 130* Fahrenheit d.
There is water on the floor in the RCIC room, the sump is NOT overflowing.
m9
- ~. . -.. -. -.. ........ -. -. _.. . ~ - - -....._ ._.--. .-- .. ..-_.- ! i ! SENIOR REACTOR OPERATOR Page 11 .! , QUESTION: 011 (1.00) i
Which of following conditions require entry into LGA-02, Secondary Containment j Control"?
I i l i a.
Reactor building d/p at -0.5 inches of water.
i b.
Reactor building floor drain sump overflowing.
! c.
Fuel Pool ventilation radiation level at 3 mr/hr,
! I i d.
Reactor water cleanup pump room temperature above normal.
) ! QUESTION: 012 (1.00) Given that: The plant was operating at 20% power when a reactor scram with a turbine - trip occurred , The turbine bypass valves have not yet opened - > - All other expected automatic actions have occurred as designed i Under these conditions, which of the following will open the turbine bypass valves? Manually decreasing the... a.
Pressure Setpoint b.
Load Limit setpoint.
c.
Steam Throttle Pressure value.
_ d.
Max Combined Flow Limiter setpoint.
, i l l ., . .. -. -. -. . .-
. -.- . .. _. _ _ _ _ _ -. _ - _ .. _ _. _ _ _ _ _ _... _. - _ _ _ _ _. _ . _ _ - _ ! ! i i t SENIOR REACTOR OPERATOR Page 12 ,
!
QUESTION: 013 (1.00) f With drywell pressure at 1.8 psig and drywell temperature at 310aF (rising), which of the l following must be satisfied to allow drywell spray valves F016A and F017A to be opened , simultaneously? ' a.
LPCIinjection valve F042A closed.
, Suppression pool spray valve F027A closed.
I b.
i i c.
RHR Heat exchanger bypass valve F048A open.
! i d.
Suppression chamber pressure less than 8 psig.
QUESTION: 014 (1.00)
The reactor was at 100% power when 1PM02J B101, "H2 PANEL TROUBLE" alarms.
Investigation by the EA at the local panel 1PL19J indicates alarm B101, " MACHINE GAS , PRESSURE HIGH/ LOW" is in alarm. The NSO confirms that hydrogen pressure is at 45 psig. What is the concern with hydrogen at this pressure? l a.
Hydrogen seal oil will collect in the Hydrogen Retraining Section arid cause a liquid detector alarm.
b.
Hydrogen purity will not be valid due to the lower pressure and decreased flow to the hydrogen purity monitor, c.
If hydrogen pressure were to fall below service water pressure, service water j could begin filling the generator casing.
d.
Service air, via the service air purge line for carbon dioxide removal, will overcome hydrogen pressure and cause a potentially explosive concentration of hydrogen and air.
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.. .._ .... _. _. _. _..... ~. _ _ _ _. _ -. _ . _ _. _... _ _.. _. _. _ _. _ ! ! ! I L l SENIOR REACTOR OPERATOR Page 13 l <l l ! . l QUESTION: 015-(1.00) ! ! .. , Unit 1 is at 100% power with the SAT supplying electrical power to bus 143. If an ! i l overcurrent condition occurs on bus 143, what is the expected status of the following l components?. ' l Breaker 1432 (SAT feed to bus 143) - ! Breaker 1433 (1B DG feed to bus 143) - . !, DG 1B - j- .a.
1432 trips OPEN,1433 remains OPEN, DG 1B remains SHUTDOWN.
l b.
1432 trips OPEN,1433 remains OPEN, DG 1B STARTS.
! i c.
1432 trips OPEN,1433 CLOSES then trips on overcurrent, DG 1B STARTS.
,d.
1432 remains CLOSED,1433 remains OPEN, DG 18 remains SHUTDOWN.
!
QUESTION: 016 (1.00) f ! Which of the following is NOT directly affected by a loss of 24/48 VDC? l ! a.
Source Range Monitors ) i b.
Process Radiation Monitors ! I c.
Intermediate Range Monitors d.
Local Power Range Monitors ! I ._ QUESTION: 017-{1.00)
l Which ONE of the following Off-Gas parameters is indicative of a fire upstream of the i ! recombiner? l i ( . a.
rapid increase in recombiner temperature i-b.
rapid decrease in recombiner temperature i ' i
c.
rapid increase in charcoal bed temperature -[ i- .d.
rapid decrease in charcoal bed temperature ! " i t ,,, . - - .. . - _ _, . _.
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__ . -___._.____.. _ _ _. _. _ _.._ _ -. _ .---__ _ _ _ _ _ _ _ _ - _ -.. _ _. _ _ _ l l l SENIOR REACTOR OPERATOR Page 14 i . ! QUESTION: 018 (1.00) i i Unit 1 is operating at 100% power with the "B" offgas post treatment Process Radiation , Monitor (PRM) inoperable and its selector switch in STANDBY. If the "A" PRM channel I fails downscale, which one of the following is a direct response?
a.
Offgas realigns for the TREAT mode j i b.
Offgas suction valves automatically close + r c.
Offgas discharge valve autornatically closes ! d.
Offgas causes MSIV closure and Reactor Scram l !
! QUESTION: 019 (1.00) i If a Carbon Dioxide (CO2) hose reel was lifted from its holder for fighting a fire on the j turbine deck, what indication is provided to the fire brigr,de member to tell him the hose is i capable of ejecting Carbon Dioxide? The local CO2: l ' a.
indicating light would turn ON.
b.
indicating light would turn OFF.
j c.
audible annunciator would turn ON.
l f d.
pressure gauge would activate, showing CO2 pressure.
l \\ j ' l l l
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_ _ SENIOR REACTOR OPERATOR Page 15 ! QUESTION: 020 (1.00) If both trains of the Standby Gas Treatment System are inoperable, which of the following conditions exist? a.
Primary Containment Integrity is no longer assured, b.
Secondary Containment Integrity is no longer assured.
c.
Desired Drywell pressure can no longer be maintalaed.
d.
Reactor Building negative pressure can no longer be maintained.
QUESTION: 021 (1.00) A TIP probe was traversing the core when an event occurred causing Reactor Water Level to drop to + 10". Which one of the following statements describes TIP System response to this event? The TIP system... a.
continues its sequence for the present core position.
b.
Shear valve immediately fires, shearing the cable and isolating the TIP System.
c.
Ball valve immediately CLOSES, cutting the cable and isolating the TIP J System.
d.
immediately reverses and withdraws the probe from the core into the shield, and the ball valve CLOSES.
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SENIOR REACTOR OPERATOR Page 16 QUESTION: 022 (1.00) Unit 1 is in Hot Standby with RCIC turbine in operation. It may be necessary to reduce ' RCIC steam flow when warming the Main Steam Lines to prevent: ! a.
Inadvertent SRV actuation.
b.
Excessive RCIC pump cavitation.
c.
Excessive carryunder in the RPV.
d.
Excessive reactor power increase.
QUESTION: 023 (1.00) If the output of the selected EHC pressure regulator failed high while operating at 100% l reactor power, which one of the following describes the plant response? (Assume no operator action is taken.)
a.
MSIV closure on low reactor pressure b.
Main turbine trip on overspeed due to TCVs opening c.
Reactor scram on high flux or high reactor pressure d.
Reactor power and pressure stabilize as the backup regulator take control ._
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SENIOR REACTOR OPERATOR Page 17 . J I QUESTION: 024 (1.00) ' Unit 2 is at 100% power with both TDRFPs in 3-element control, and the level setpoint at 36". If the selected level input channel fails downscale, which ONE of the following describes the expected plant response? a.
RPV water level will decrease, and the reactor will scram on low water level.
l
b.
RPV water level will increase, Main Turbine will trip, and the reactor will I ! scram due to the turbine trip.
The reactor will scram immediately, because the selected Narrow Range
c.
i i channel feeds both RPS channels.
! d.
No scram will result because reactor water level will be maintained above the scram setpoint due to the combined effects of Recirculation Pump downshift, direct setpoint setdown and dynamic compensator.
! QUESTION: 025 (1.00) Which one of the following signals will auto start the Standby Gas Treatment System (VG) from it's normal standby lineup? a.
RPVlevelOinches b.
Drywell pressure 1.0 psig c.
Fuel pool ventilation exhaust rad levels of 5 mR/hr i d.
Reactor building ventilation exhaust rad levels ~of 10 mR/hr - " l , - -- -, .- -, -.
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SENIOR REACTOR OPERATOR Page 18
1' QUESTION: 026 (1.00) The 1 A DG Cooling Water pump trips while the 1 A DG is running under 75% load during a i surveillance. Assuming no operator action occurs, which ONE of the following describe - - the expected impact on continued 1 A DG operation? > i l The 1 A DG will trip on high cooling water temperature.
a.
$ b.
The pump trip will directly actuate the DG lockout which will trip the 1 A DG l off.
The 1 A DG SW cross connect AOV will automatically open, allowing SW to c.
j.
cool the 1 A DG.
d.- The 1 A DG governor will runback the loadset to 10%, which is within the DG's cooling capacity.
L QUESTION: 027 (1.00) a
! The Unit 1 NSO identified a need to quickly reduce power from 100 to 50%. - The NSO selected the Power Reduction Mode on the RWM, and began selecting and driving cram array rods to position 00. One cram array rod would not latch at position 00, but settled to position 02 and latches. Which ONE of the following is the RWM response? a.
INSERT ERROR generated, ROD BLOCK applied.
b.
INSERT ERROR generated, no ROD BLOCK applied.
c.
WITHDRAWAL ERROR generated, ROD BLOCK applied.
d.
WITHDRAWAL ERROR generated, no ROD BLOCK applied.
-
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_ SENIOR REACTOR OPERATOR Page 19 QUESTION: 028 (1.00) Which of the following will result in automatic closure of the Reactor Water Cleanup (RWCU) blowdown flow control valve (G33-F033)? a.
Reactor water level at -53 inches.
b.
75 gpm RWCU system differential flow.
. c.
142 psig pressure downstream of the flow control valve.
d.
145 F non-regenerative heat exchanger outlet temperature.
QUESTION: 029 (1.00) Which one of the following is used by the reactor recirc flow control valve runback circuit to determine the number of operating reactor feed pumps? a.
MDRFP breaker position b.
MDRFP discharge pressure c.
TDRFP vacuum d.
TDRFP discharge pressure _ t ! .
_ _ _.. _ -.. - _ _ _. . _ _ _.. _ _ _ _ _ _. _. _. _, _ _ _. _. _. _ -. _ _.. _ i i ! SENIOR REACTOR OPERATOR Page 20 ?
QUESTION: 030 (1.00) > ! Given that the unit was initially operating at 80% reactor power with "A" and "B" TDRFPs in 3-element AUTO, the following occurred: l . l The 'A' TDRFP tripped - Reactor water level is at 30" and lowering - , l The lead HPU for the FCVs has tripped, the backup HPU has started and is - operating normally Recirculation flow control is in th6 loop manual mode - ! Which of the following is the CORRECT response of the Recirculation Flow Control System? The FCVs will a.
be LOCKED UP.
b.
be UNAFFECTED.
! I c.
start to CLOSE.
!
d.
RUNBACK to its minimum position.
QUESTION: 031 (1.00) RHR Loop "A" has Lutomatically initiated in the LPCI mode on a high drywell pressure signal. It is injecting to the core through F042A, RHR Loop "A" injection valve. Under the present condition, which one of the following statements describes the "A" RHR system response when its handswitch is taken to the OPEN position? a.
F027A and F042A will both open and stay open.
_ b.
F027A will open after a 10 minute time delay, c.
F027A will not move as long as F042A is open.
d.
F027A will open and F042A will automatically close when the full open limit is reached on F027A.
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, I SENIOR REACTOR OPERATOR Page 21 ! QUESTION: 032 (1.00) ! After a scram the NSO prepares to reset the scram by placing the Scram Discharge Volume ! , l (SDV) High Water Level bypass switch in the BYPASS position, but fails to depress the ! - SDV Vent and Drain valves Test pushbuttons. Which ONE of the following may result from the NSO's actions? , i j SDV Vent and Drain valves may: r , t ' l a.
OPEN without the scram valves CLOSING.
, > ! I b.
OPEN without the scram valves OPENING.
c.
CLOSE without the scram valves CLOSING.
i T d.
CLOSE without the scram valves OPENING.
l QUESTION: 033 (1.00) f WHICH ONE (1) of the following statements correctly describes the shorting links that are used in Reactor Protection System (RPS)? a.
Installation of the shorting links makes SRM scrams available. An SRM scram will occur if ANY SRM reaches it's Hi-Hi setpoint.
b.
Removal of the shorting links makes SRM scrams available. An SRM scram will occur if ANY SRM reaches it's Hi-Hi setpoint.
c.
Installation of the shorting links makes SRM scrams available. An SRM i scram will occur based on a coincident logic scheme of SRM channels reaching their Hi-Hi setpoints.
- d.
Removal of the shorting links makes SRM scrams available. An SRM scram will occur based on a coincident logic scheme of SRM channels reaching their Hi-Hi setpoints.
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. SENIOR REACTOR OPERATOR Page 22 QUESTION: 034 (1.00) l A reading of 20 on Range 9 of the Intermediate Range Monitors (IRM) recorders would indicate WHICH ONE (1) of the following reactor power levels? a.
0.8 % b.
2.0% c.
8.0 % d.
20.0 % QUESTION: 035 (1.00) The plant was operating at 60% reactor power when the following occurred: level began decreasing rapidly and is at -45" and lowering.
) - drywell pressure rapidly increased to 1 psig and continues to rise -
- due to a malfunction, both HPCS RPV level 8 switches are locked up and
tripped at + 55.5" the RX VESSEL WTR LVL 8 Hi annunciator lit at panel H13-P601 A.
) - Which ONE of the following would cause HPCS to inject if the NSO manually started the HPCS pump from the control room? a.
RPV level drops to -50" b.
A High Drywell Pressure signal is received by HPCS c.
The NSO depresses the Lo Level /Hi Drywell Pressure reset pushbutton.
d.
The NSO at panel H13-P601 takes the control switch for the HPCS INJECTION valve (E22-FOO4) to OPEN.
i , Y
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I l SENIOR REACTO'3 OPERATOR Page 23 i QUESTION: 036 (1.00) Given the following plant conditions: I - A small break LOCA has occurred Drywell pressure is 1.9 psig and rising - RCIC is injecting - HPCS has been secured by operator action - Emergency RPV Depressurization has been initiated - Which of the following conditions will result in the closure of RCIC Vacuum Breakers (E51-F080 and F086)? a.
Reactor Water Level (-129") b.
Reactor Vessel Pressure (57 psig) c.
RCIC Exhaust Diaphragm Pressure (10 psig) I d.
RCIC Equipment Area Differential Temperature (120 F) QUESTION: 037 (1.00) Which of the following activities does NOT require the use of electrical protection equipment? i a.
A 480 VAC circuit breaker is being racked out, b.
Breaker 1423 is being closed by local manual operation.
c.
A GE 4160 volt circuit breaker is being placed in " Test."
- d.
A 250 VDC circuit breaker is closed by local manual operation.
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l l QUESTION: 038 (1.00) i l You have been assigned to the training department for a two year rotational assignment.
l l During that time you have allowed your license to go INACTIVE. In accordance with l 10CFR55.53, what must'you do to restore your license to ACTIVE status? l
a.
Stand 7 eight-hour shifts as the NSO.
! l l b.
Pass an annual requalification examination.
I
c.
Notify the NRC that you are resuming watchstanding duties, l d.
Stand 40 hours of watch under instruction at your watchstation.
l QUESTION: 039 (1.00) l l Which ONE of the following is the NORMAL sequence that should be used for tagging a
motor operated valve out of service (OOS)? Assume the valve is in the proper position.
a.
Tag the remote control switch OOS, open the power supply breaker for the i l valve and tag it OOS, tag the valve OOS.
, j.
b.
Open the power supply breaker for the valve and tag it OOS, tag the remote l control switch OOS, tag the valve OOS.
c.
Tag the valve OOS, open the power supply breaker for the valve and tag it OOS, tag the remote control switch OOS.
d.
Open the power supply breaker for the valve and tag it OOS, position the l clutch to the manual position and tag the clutch OOS, tag the valve i handwheel OOS.
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SENIOR REACTOR OPERATOR Page 25 l !
! QUESTION: 040 (1.00) ' Which of the following work activities may be performed without a Nuclear Work Request (N W R)? , i l
l a.
Adjusting a protective relay setpoint for a Non-Safety related instrument.
, i b.
Aligning a loose limit switch on an open MOV to give the proper indications I in the control room.
l } c.
During a surveillance test the Alarm Typer is turned off due to a nuisance l alarm saturating the typer.
{ d.
Tightening live load packing on a pump leaking at 20 drops per minute by an ! operator during his inplant rounds.
! QUESTION: 041 (1.00) , Conditions develop for which no action consistent with the Plant Technical Specifications will provide for adequate protection of the public health and safety, i NOTE: All shift personnel are present in the control room.
? ? Select the appropriate action for the reactor operator, a.
Take only those actions that will comply with approved plant operating , procedures.
i b. - Obtain approval from the Shift Operations Supervisor, then take actions to protect the public health and safety.
c.
Take actions to protect the public health and safety and subsequently inform the Shift Manager, d.
Obtain approval from the Shift Manager, then take actions to protect the public health and safety.
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SENIOR REACTOR OPERATOR Page 26 j i ! ! OUESTION: 042 (1.00) ! Wnich one of the following statements is correct in respect to conduct of operations associated with valve csperations? l - Operators making their inplant rounds may loosen the packing on MOVs if ! a.
the packing is too tight.
i l b.
Operation of MOVs are limited to 30 starts within 1 minute followed by a 5 i ! minutes cooling off time.
l
c.
When placing the ECCS minimum flow valve in the shut position for maintenance, the associated ECCS pump is placed in pull-to-lock.
d.
When locally closing a motor operated Gate Valve with the Reactor Coolant system at normal operating pressure and temperature, mechanical assistance (cheater bars) may be used to prevent leakage past valve seats.
! , QUESTION: 043 (1.00) Which ONE of the following RADIATION SIGNS is posted to indicate an area accessible to f personnel in which there exist radiation at such levels that a major portion of the whole ' body could receive, in any one (1) hour, an exposure in excess of 100 mrem? I a.
RADIOACTIVE LSA b.
CAUTION-RADIATION AREA l c.
DANGER-HIGH RADIATION AREA d.
CAUTION-AIRBORNE RADIOACTIVITY AREA - l l l l ! . - r
-. _ -- -_ __ _ r SENIOR REACTOR OPERATOR Page 27 l l QUESTION: 044 (1.00) You have been designated to give a plant tour for several visitors (not badged). You will l have the prime escort duties. What is the maximum number of people you are allowed to escort into the plant? a.
No more than 5 in any area.
b.
No more than 10 in any Owner-Controlled Area at one time.
c.
Up to 5 in the Vital Areas, and up to 10 in the Protected Areas.
d.
Up to 5 in the Vital and Protected Areas, and no more than 15 in the Owner-Controlled Area.
QUESTION: 045 (1.00) Which one of the following describes why the off-gas mechanical vacuum pump should not be used while operating above 5% of rated thermal power? a.
the mechanical vacuum pump cannot maintain condenser vacuum less than the turbine trip setpoint.
b.
there is a possibility of steam reaching the mechanical vacuum pump and causing damage to the blading.
there is the potential formation of detonable concentration of hydrogen and c.
oxygen in the untreated mechanical vacuum pump flow path.
d.
the flow path through the mechanical vacuum pump and the off gas charcoal adsorbers provides insufficient holdup time to allow radioactive gasses to , decay.
l , d I i
- . .. . . -_ ,. . . -. -. .- .- . -. -. - l l l SENIOR REACTOR OPERATOR Page 28 , QUESTION: 046 (1.00) The plant was operating at 95% power when an event occurred resulting in the following plant pgameters: ! Reactor scram - l l - RPV pressure 800 psig trending down
j RPV level-70" trending down rapidly - HPCS injecting into the RPV ' - - The NSO armed and depressed the "A" RHR/LPCS Manual Initiation pushbutton to help
control reactor water level. Under these conditions, which of the following LPCS lineups would occur? i LPCS Pump OFF. LPCS/RH Water Leg Pump OFF. No system flow.
' a.
b.
LPCS Pump OFF. Water Leg Pump running; taking suction from suppression pool and discharging back to the suppression pool, LPCS Pump running; taking suction from suppression pool and discharging to c.
vessel through LPCS Injection Valve. LPCS Water Leg Pump running.
d.
LPCS Pump running; taking suction from suppression pool and discharging back to suppression pool through Minimum Flow Valve. Water Leg Pump is running.
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SENIOR REACTOR OPERATOR Page 29 . ! OUESTION: 047 (1.00) I The High Pressure Core Spray (HPCS) automatically initiated on a valid low reactor water , . level signal. The Unit Supervisor then directed the operator to manually override and close ! the HPCS Injection Valve (E22-FOO4) to maintain RPV water level at Level 4. What is
required to reinstate the " automatic opening" feature of the HPCS Injection Valve? o , l a.
Drywell pressure rises to 3 psig and subsequently decreases below 1.69 l psig.
i l b.
The "HI WATER LEVEL" reset pushbutton is depressed while reactor water
j level is below + 55.5 inches.
. ! l c.
Reactor water level increases to above -50 inches and subsequently !
decreases to less than -129 inches.
'
! d.
The HI DRYWELL PRESSURE /LO WATER LEVEL" reset pushbutton is depressed after the HPCS initiation signal clears.
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SENIOR REACTOR OPERATOR Page 30 QUESTION: 048 (1.00) ' A unit startup is in progress with the reactor still subcritical. AFTER each rod withdrawal, neutron level willincrease for a period of time and then stabilize at a higher level. WHICH of the following describes this effect as the reactor gets closer to criticality? The proximity to criticality has NO effect on how long it takes for neutron a.
level to stop increasing following a rod withdrawal.
b.
The closer the reactor is to criticality, the LONGER it will take for the neutron level to stop increasing and level out following a rod withdrawal.
c.
The closer the reactor is to criticality, the LESS TIME it will take for the neutron level to stop increasing and level out following a rod withdrawal.
d.
The closer the reactor is to criticality, the less likely that neutron level will stabilize following a rod withdrawal. At a given reactivity, reactor power will continue to increase (after rod motion has ceased) until the reactor achieves criticality.
QUESTION: 049 (1.00)
During steady state power operations, the APRM Gain Adjustment Factor (AGAF) for i APRM-C reads 1.04 on a current OD-3. Under these conditions, the operation of APRM C is because , a.
conservative / APRM reading is LESS THAN actual power b.
conservative / APRM reading is GREATER THAN actual power c.
NON-conservative / actual power is LESS THAN APRM reading d.
NON-conservative / actual power is GREATER THAN APRM reading l l ! ! ,
.-- _ - . - i i . SENIOR REACTOR OPERATOR Page 31 . QUESTION: 050 (1.00) The following plant conditions exist: l Automatic Depressurization System actuated Reactor water level - (-100) inches and trending up All RHR pumps - running
105 second timer - timed out t 7 ADS SRVs - open ! Drywell pressure - 1.5 psig and decreasing i If both the DIV I and DIV ll RPV LOW LEVEL RESET pushbuttons are momentarily depressed, which of the following describes the result on the Automatic Depressurization System? . a.
SRVs will remain open.
b.
SRVs will close and remain closed.
l c.
SRVs will close and then reopen after 105 seconds.
d.
SRVs will close and then reopen after 450 seconds.
QUESTION: 051 (1.00) ! Suppression Chamber to Drywell differential pressure is maintained LESS than 0.5 PSID to j prevent: a.
uncovery of the SRV discharges ] b.
loss of containment emergency vent capability - c.
structural failure of the containment downcomers d.
opening the suppression chamber to drywell vacuum breakers -
!
. -. . . . .. ! l l SENIOR REACTOR OPERATOR Page 32 QUESTION: 052 (1.00) A Unit 1 shutdown is in progress per LGP 2-1 with the following conditions: The reactor vessel pressure is 940 psig - - The mode switch is in STARTUP The main condenser low vacuum Main Steam isolation Valve (MSIV) isolation signal will be automatically bypassed by taking the low condenser bypass switches to BYPASS... a.
OR by closing the turbine stop valves.
b.
OR by closing the turbine bypass valves.
c.
AND shutting the turbine stop valves.
d.
AND shutting the turbine bypass valves.
QUESTION: 053 (1.00) l Which of the following would be an indication of an open or leaking safety relief valve? a.
Indicated Total Steam Flow 1NCREASE with a CONSTANT Generator Load.
b.
Indicated Total Steam Flow CONSTANT with an INCREASE in Generator Load, c.
Indicated Total Steam Flow DECREASE with a CONSTANT indicated Reactor Power.
d.
Indicated Total Steam Flow DECREASE with a DECREASE in indicated Reactor Power.
l
SENIOR REACTOR OPERATOR Page 33 OUESTION: 054 (1.00) The following plant conditions exist: - 100 percent reactor power reactor water level is 36 inches - MDRFP is OFF and the FRV is CLOSED - Both TDRFPs in 3-element control - Level Setpoint is 36 inches - if the "A" TDRFP flow transmitter rapidly fails UPSCALE, HOW will the plant respond to this event? (Assume no operator action.) Reactor level: a.
increases causing a turbine trip.
b.
stabilizes at a level below normal.
c.
stabilizes at a level above normal.
d.
decreases causing a reactor scram.
QUESTION: 055 (1.00) Given the following plant conditions: A Normal reactor shutdown is in progress - Reactor pressure is 950 psig - Reactor power is 10% - Which of the following Reactor Protection System automatic scrams is bypassed when the j mode switch is taken from "RUN" to "STARTUP"? -
a.
APRM Thermal Hi-HI t b.
Low CRD header pressure c.
Turbine Stop Valve Closure d.
Main Steam Line isolation Valve Closure ! -
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QUESTION: 056 (1.00) If the scram discharge volume vent and drain valves do not go closed when a scram has occurred, which of the following would be the adverse consequence? a.
There will be a primary to reactor building leak.
. f b.
The CRD discharge path has insufficient back pressure.
c.
The control rod insertion speed will exceed design limits.
d.
The timer allowing scram reset after 10 seconds will not initiate.
QUESTION: 057 (1.00) Given the following plant parameters: Turbine Throttle Pressure 965 psig - - Pressure Setpoint 935 psig.
Load Set 110 percent - - Load Limiter 100 percent - Max Combined Flow Limiter 120 percent RRFC Individual Manual - WHICH of the following describes the expected plant response if the Max Combined Flow Limiter is inadvertently reduced to 80%7 a.
TCVs close and BPVs open. Rx power will control at 80%. b.
TCVs open and BPVs remain closed. Rx power will control at 80%. c.
TCVs open and BPVs open. The reactor will scram on MSIV closure.
d.
TCVs close and BPVs remain closed. The reactor will scram on high flux or pressure.
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, . _. SENIOR REACTOR OPERATOR Page 35 OUESTION: 058 (1.00) Given the following plant conditions: RPV pressure is 450 psig - Drywell temperature is 250*F - - Drywell pressure is 20 psig - RPV level is increasing due to HPCS/RCIC injection - Reactor building temperature is 125 F Which of the following may be used to determine that water level is at the Top of Active Fuel (TAF)? a.
Fuel Zone (only) b.
Wide Range (only) c.
Upset and Wide Range d.
Fuel Zone and Wide Range and Upset Range QUESTION: 069 (1.00) During a large break LOCA, vessel levelindication was lost and the SRO directed the initiation of ADS LAW LGA-05 "RPV FLOODING". The NSO reported back to the SRO that the SRVs did NOT open. Plant parameters at time of the event: - reactor vessel pressure 800 psig suppression pool water level -10 feet - - suppression pool temperature 220 F - drywell pressure 69 psig - - drywell temperature 220 F WHICH of the following is the reason for the SRVs not opening? l a.
Drywell temperature is too high.
b.
Containment pressure is too high.
c.
Suppression pool water level is too low.
d.
Suppression pool temperature is too high.
. d l l
. l l SENIOR REACTOR OPERATOR Page 36 QUESTION: 060 (1.00) During plant operations the following plant conditions were noted to occur over a 3 minute period.
Reactor pressure - decreased to 800 psig, now stable.
- - Reactor Water Level - 34 inches trending to normal.
- Reactor power - decreased 5%, now stable at 50% - Generator output - decreased to 250 Mwe from 500 Mwe.
Reactor Protection Sys - No actuatior.s have occurred.
- You should immediately: a.
Trip the main turbine.
b.
Scram, Close MSIV's.
c.
Adjust reactor pressure control to increase pressure.
d.
Increase Recirculation flow to increase power and pressure.
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- . . . SENIOR REACTOR OPERATOR Page 37 QUESTION: 061 (1.00) The plant is operating at 25% reactor power when condenser vacuum begins to decrease.
NO operator action occurs and Vacuum decreases to zero over a 30 minute period. Which one of the following statements CORRECTLY describes the response of the plant to the decreasing vacuum? (Decreasing vacuum means pressure in the condenser is approaching { atmospheric pressure.)
' The MSIVs will shut causing a reactor scram on MSIV position.
a.
b.
The reactor will trip. There is a direct reactor scram on low condenser vacuum.
The turbine will trip giving an RPS trip signal from low turbine trip oil c.
pressure (TCV fast closure).
d.
The turbine will trip but SRVs will open to control reactor pressure. No reactor scram will occur.
I wa M
l \\ SENIOR REACTOR OPERATOR Page 38 j OUESTION: 062 (1.00) i A loss of offsite power initially occurs on Unit 2 while the unit is shutdown. While mitigating the Unit 2 problem the following occurred: i
Unit 1 receives an ECCS initiation signal on high drywell pressure - - Unit 1 suffers a loss of offsite power - The 18 diesel generator fails to start Assuming all other systems operates as expected, which of the following pump configurations is available for this event? UNIT 1 UNIT 2 ) a.
LPCS pump LPCS pump, RHR pumps A,B,C RHR pumps A,B,C HPCS pump b.
LPCS pump RHR pumps A,B,C RHR B,C HPCS pump , c.
LPCS pump RHR pumps B,C , RHR pumps A,B,C HPCS pump d.
LPCS pump RHR pumps B,C ' HPCS pump HPCS pump . -
SENIOR REACTOR OPERATOR Page 39 QUESTION: 063 (1.00) While operating at 70% power, a Feedwater Control System malfunction has caused reactor vessel level to reach Level 8 (55.5 inches), resulting in a reactor scram and main turbine trip. The reactor feed pumps have FAILED to trip automatically and manually. You should: i a.
manually CLOSE the MSIVs if vessel level reaches 73 inches.
b.
lineup and blowdown with the RWCU System to the main condenser if vessel level reaches 73 inches.
c.
lineup and blowdown with the RWCU System to the main condenser if vessel ievel reaches 125 inches.
d.
trip the condensate / condensate booster pumps to cause and additional TDRFP trip signal to prevent overfilling the RPV, i QUESTION: 064 (1.00) Which of the following will trip the 18 Primary Containment Ventilation Supply Fan (VP)? , a, Undervoltage on Unit 1 Division 2 ESF bus and Drywell Pressure of 1.69 psig j on Unit 1.
b.
Undervoltage on Unit 2 Division 1 ESF bus and Drywell Pressure of 1.69 psig on Unit 2.
' c.
Undervoltage on Unit 2 Division 2 ESF bus and Reactor Water Level of -129 inches on Unit 1.
. d.
Undervoltage on Unit 1 Division 1 ESF bus and Reactor Water Level of -129 inches on Unit 1.
.
. .. - - -, ~... -. -. - -. . - . .._.- . .... -... -.. -....-...-.. i i ) l SENIOR REACTOR OPERATOR Page 40
i \\ QUESTION: 065 (1.00) . ! Which one of the following interlocks or trip functions is still operable from the Remote l Shutdown Panel after placing all remote transfer switches in their " emergency" Position? ! a.
LLS for SRVs K and P if manually actuated i l b.
RHR "B" SDC Suction / Full Flow Test valve interlock l c.
RCIC High Exhaust Diaphragm Pressure isolation / Trip ' ! d.
RHR "A" SDC Suction / Suppression Pool suction valve interlock I OUESTION: 066 (1.00) . While operating at 100 percent reactor power, the following alarms are received: i RBCCW EXPANSION TANK LVL Hl/LO (A103) ' - - RBCCW PMP AUTO TRIP (A101) l in addition to the RBCCW System, WHICH of the following components require increased monitoring during this event? a.
Drywell Chillers b.
Reactor Recirculation Pumps c.
Fuel Pool Cooling Heat Exchangers d.
Auxiliary Building HVAC Heat Exchangers -
, ! L
I l ! ! . .. - - - -
, SENIOR REACTOR OPERATOR Page 41 QUESTION: 067 (1.00) While operating at 100% power, a complete loss of Instrument Air (IA) occurs. WHICH of the following describes the correct failure positions for the listed valves as a result of the loss of IA? a.
MDRFP Feed Reg Valve - FAILS AS IS TDRFP Min Flow Valves - Fall OPEN b.
MDRFP Feed Reg Valve - FAILS OPEN TDRFP Min Flow Valves - Fall OPEN c.
MDRFP Feed Reg Valve - FAILS AS IS TDRFP Min Flow Valves - Fall AS IS d.
MDRFP Feed Reg Valve - FAILS OPEN TDRFP Min Flow Valves - Fall CLOSED QUESTION: 068 (1.00) The following conditions exist: A reactor stertup is in progress - The mode switch is STARTUP - The main turbine is tripped - - A valid Group iisolation has occurred - No Auto Scram Signal is present Of those given, select the only signal that could have generated the Group I isolation: a.
Low reactor water level - b.
Low main steam line pressure c.
High main steam line radiation d.
High main steam line tunnel differential temperature
) - >
... ~ . - - . . = - _ ~. - - _-._- - - - - - .... -- - . ,
I ! SENIOR REACTOR OPERATOR Page 42 '
l l !
L QUESTION: 069 (1.00) ! l I i Which one of the following statements describen why you are required to take the control ' switches for both CRD pumps to start simultaneously when starting a second CRD pump? l a.
This is done to bypass thermal overload protection since the second pump j would NOT be started except for an emergency.
j i l b.
This is done to delay the low suction pressure trip logic and prevent the
l operating pump from tripping on low suction pressure.
c.
This is done to enable the CRD flow control valves gain change logic which i is needed to compensate for the flow characteristics of two pumps in ! simultaneous operation.
d.
This is done to enable the low suction pressure trip, which is normally
bypassed, since damage from low suction pressure CANNOT occur if only
I one pump is running.
l l
' , I QUESTION: 070 (1.00) l ' The RCIC turbine was manually started by an NSO to perform an operability surveillance.
The "B" man assigned to locally monitor the RCIC room reported steam leakage near the i ' RCIC turbine governor valve. The NSO immediately depressed the RCIC isolation pushbutton. This will: a.
not cause a PCIC system isolation.
b.
cause an isolation by closing the steam shutoff (F045) only, c.
caur. an isolation by closing the outboard isolation valve (FOO8) only.
d.
cause an isolation by closing the inboard (F064) and outboard (FOO8) isolation valves.
i i f i e i
r
! t
_ SENIOR REACTOR OPERATOR Page 43
l QUESTION: 071 (1.00) l Which ONE of the following statements is correct concerning the RHR Service Water ' System Process Radiation Monitor (PRM)? a.
Only required to be in service when in the Shutdown Cooling Mode of Operatior.. b.
Provides indication and alarm functions only, i.e. NO automatic isolations.
c.
Measures the cumulative activity release to the environment from the RHR Service Water System.
d.
Measures the activity of the fluid which passes through the shell side of the RHR heat exchangers.
QUESTION: 072 (1.00) Given the following plant conditions: - assume no operator actions reactor operation in condition # 4 - RHR loop B in shutdown cocling mode - reactor moderator temperature at 140 F and holding - Which ONE of the following describes the expected behavior of the reactor water level and moderator temperature upon receiving a trip of the operating RHR service water pump (s)? a.
reactor level decreases as moderator temperature increases b.
reactor level increases as moderator temperature increases - c.
reactor level decreases as moderator temperature decreases d.
reactor level increases as moderator temperature decreases
SENIOR REACTOR OPERATOR Page 44 QUESTION: 073 (1.00) Which of the following situations is ALLOWED, concerning the manipulations of controls? NOTE: Assume that the NSO has full knowledge and consents with the manipulations, and all manipulations are per an authorized procedure.
' A license candidate pulls control rods to criticality under direct supervision of a.
the NSO.
' b.
A license candidate repositions the reactor mode switch while the NSO is checking Secondcry Containment parameters.
With Unit 1 at 25% power, an I&C technician conducting a surveillance test, c.
adjusts recirculation flow while under the constant direction of the NSO.
d.
With Unit 1 in mode 3, a system engineer conducting a post maintenance test, depresses the HPCS manual initiation pushbutton while under the j constant direction of the NSO.
l l QUESTION: 074 (2.00) MATCH the LGA procedure (s) in Column B which are entered as a result of the conditions i described in Column A.
Column A - Column B a. Reactor Scram 1. LGA-01 b. Suppression pool temp 105 F 2. LGA-01 and LGA-03 . c. Drywell pressure greater 3. LGA-03 than 1.69 psig 4. LG A-02 d. Reactor Bldg differential pressure equal to 0 inches of water -
l l , SENIOR REACTOR OPERATOR Page 45 l QUESTION: 075 (1.00) Given the following conditions: , I - Reactor Power 100 % - Reactor Pressure 1000 psig Drywell Pressure 25 psig - Suppression Chamber Press 25 psig - - Suppression Pool Level + 10 feet - Suppression Pool Temperature 112 F Which of the following may be exceeded if a sustained RPV emergency depressurization occurs a.
SRV tailpipe limit b.
Pressure Suppression Pressure Limit c.
Heat Capacity Temperature Limit d.
Primary Containment Level Limit QUESTION: 076 (1.00) Select the condition under which a Fall OPEN air operated valve may be considered closed for out of service purposes.
a.
The valve is tagged out of service and its gas supply is physically disconnected.
b.
The valve is closed with a gag. The valve and its gagging device are tagged out of service.
c.
The valve is closed with a gag. The valve and its gas supply isolation points are tagged out of service.
d.
The gas supply is physically disconnected. The valve and its gas supply isolation points are tagged out of service.
. r
SENIOR REACTOR OPERATOR Page 46
QUESTION: 077 (1.00) It is estimated that an individual in a 90 mrem /hr radiation area will take 10 minutes to conduct a second verification. As a result, which ONE of the following statements is correct in this case? a.
Second verification cannot be waived.
b.
Second verification can be waived to comply with ALARA.
J c.
Second verification can be waived only if the person is endangered.
d.
Second verification can be waived anytirr.e a radiation area is to be entered.
OUESTION: 078 (1.00) Which ONE of the following Emergency Action Levels is classified as en " Unusual Event"? a.
MCPR confirmed to be 1.06 in plant condition 1.
b.
Sustained winds of 90 MPH while in plant condition 2.
c.
Loss of off-site power and loss of all Diesel Generators for 15 minutes in plant condition 4.
d.
Drywell floor drain isolation valves (RF012 and 013) DO NOT close when RPV level is -50" in plant condition 3.
.
e ..
- . .. .- - _. - SENIOR REACTOR OPERATOR Page 47 QUESTION: 079 (1.00) , Which of the following describes the required emergency notifications when a Site Area Emergency is made as the initial declaration? ' a.
State and local agencies must be notified within 15 minutes after declaration of an emergency and the NRC notified immediately thereafter not to exceed one hour.
b.
State and local agencies must be notified within 15 minutes except for a , Unusual Event. The NRC must be notified within 4 hours after declaration of , the event.
j c.
The NRC must be notified within 15 minutes after declaration of an emergency. State and local agencies must be notified immediately thereafter not to exceed one hour.
d.
State, local agencies, and the NRC must be notified immediately after declaration of an event not to exceed 15 minutes except for an Unusual Event which may not exceed 4 hours.
" l ! _ l l ,
- .- . .. -. _ . .- _ . - - -.. -. ... -. =. -.. -. - .. - , , I ' r SENIOR REACTOR OPERATOR Page 48 ! QUESTION: 080 (1.00) ! .
[ Unit 2 is in op condition 5 with the mode switch locked in the refuel position. Fuel ) l movement is in progress, with one half of the core off loaded. Two (2) RHR loops are i i available for shutdown cooling, with the 'A' RHR loop operating in the shutdown cooling mode. The following event occurs: The 'A' RHR loop heat exchanger suffers a cooling water tube rupture -
The 'A' RHR Heat exchanger is isolated, and the 'B' RHR loop is placed into j - the shutdown cooling mode i As a result of the leak: Reactor Vessel water level is 21'11 1/8" above the top of the vessel flange - Spent fuel pool water level is 23 feet above the top of active fuel i - Whet is the correct action based on the above conditions? l a.
Refueling activities within the RPV may continue.
i b.
Refueling activities in the Spent Fsel Pool Area may continue.
c.
Refueling activities in the Spent Fuel Pool Area and within the RPV may continue.
) d.
Refueling activities in the Spent Fuel Pool Area and within the RPV may continue provided an alternate means of decay heat removal has been i demonstrated within 1 hour and at least once per 24 hours thereafter.
I . - - - - -_ __.
- - _ ~
- . _ _ _ SENIOR REACTOR OPERATOR Page 49
l QUESTION: 081 (1.00) , Which one of the following is the correct reason for Emergency Depressurizing DUE TO High Offsite release? a.
To avoid reaching the General Emergency release rate.
b.
To prevent unnecessary evacuation of on-site personnel.
' c.
To maintain core submergence to minimize any further fuel damage.
d.
To discharge the heat energy and radioactivity to the main condenser vs the environment.
i OUESTION: 082 (1.00) The Drywell Spray Limit curve is to prevent: a.
damage to the drywell spray nozzles b.
implosion of the drywell due to low pressure.
c.
catastrophic failure of the drywell from overpressure.
d.
fatigue failure at the downcomers and drywell floor junction.
~ a
) SENIOR REACTOR OPERATOR Page 50 QUESTION: 083 (1.00) l
A Unit 1 startup in progress per LGP-1-1. For which one of the following sets of conditions would it be permissible to place the reactor mode switch in the startup position? (Assume Unit 2 is in RUN) Set A Set B Set C Set D
- Operable SRMs
3
2 Seismic Monitor OPER INOP INOP OPER RCIC INOP OPER INOP OPER
- SROs
1
1 'NSOs
2
3 , (*Present in the control room) QUESTION: 084 (1.00)
l Which ONE of the following statements is CORRECT regarding Technical Specification (TS) Clarifications? a.
A clarification that is editorial in nature does NOT require a 10 CFR 50.59 Screening.
j ' b.
NRC approval is required prior to any TS clarifications that are NOT editorial in nature c.
If after conducting a 10 CFR 50.59 Screening of the TS Clarification, an Unreviewed Safety Question arises, then a 10 CFR 50.59 Safety Evaluation is required.
d.
If an error is discovered in TS which results in the unit being placed in a very time restrictive action statement, immediate clarification of the TS can be obtained from the NRC Resident inspector.
l i d
SENIOR REACTOR OPERATOR Page 51 QUESTION: 085 (1.00) Given the following plant conditions: Unit 2 is in mode 3, busses 241-Y and 242-Y have had only the Unit 2 SAT { - as a power supply for the past 20 minutes as per part of pre-planned breaker , testing Unit 1 was in mode 1 when the following occur: - RCIC steam line break, all attempts to isolate the break have failed - - The reactor has been scrammed (as a conservative measure), all rods are fullin, the RPV is being depressurized with the main turbine bypass valves - RPV level is being maintained at 12.5" to 55.5" with the feed and condensate system - RCIC Equipment area has reached 155 F What is the correct Emergency Classification? a.
Unusual event ' b.
Alert c.
Site Emergency d.
General Emergency I i i l l , d
.m . . - _ _ .. _ _ _ _. _ _. - _ _ _... _ _ ~. _.. _... _. _.._.
- _ _.... _ _ - .. r , ' i i !
r ' $ j SENIOR REACTOR OPERATOR Page 52 l t QUESTION: 086 (1.00) Which ONE of the following SLC system net tank volume, solution concentration and ! temperature satisfy Unit 1 Technical Specification requirements to make the system operable? , NET TANK VOLUME SOLUTION CONCENTRATION TEMP, F i ' , i a.
4700 gals 12 %
- b.
5000 gals 12 %
c.
5000 gals 13 %
i d.
5000 gals 13 %
OUESTION: 087 (1.00) Which of the following describes a properly oriented fuel bundle? ' a.
The channel spacer buttons face the control rod of the fuel assembly.
b.
The orientation boss on the fuel assembly bail points away from the control rod.
c.
The channel spring clip is located on the outside edge ("away" from the control rod) of the fuel assembly.
d.
Serial number on the bail handle is readable ("right side up") from the outside edge of the fuel assembly.
- l i e i
. r- - y ,, w,- --, -.-
i i f I , I SENIOR REACTOR OPERATOR Page 53 l l l QUESTION: 088 (1.00) The reactor is operating in the Core Thermal Hydraulic Instability region, which ONE of the following statements describes the major safety concern with operation in this region? a.
Core power oscillations may cause APRMs to reach their trip setpoints resulting in a scram.
b.
Core power oscillations may cause pressure surges on the MSLs resulting in ' piping and/or main turbine damage.
c.
Core power oscillations as sensed by the LPRMs may be out of phase and cause erroneous computer readings resulting in RBM trip due to LPRM and APRM reference differences.
d.
Due to the way LPRM inputs are used by the APRMs, core power oscillations sensed by the LPRMs may only be marginally detectable by the APRMs. Core limits could thus be challenged without getting an APRM scram QUESTION: 089 (1.00) When the Turbine Stop Valve closure scram is bypassed, adequate core protection is provided by: a.
MSIV closure scram b.
control valve fast closure scram c.
end of cycle recirculation pump trip d.
reactor pressure or neutron flux scrams - .
_ _ . --. l , SENIOR REACTOR OPERATOR Page 54 l l \\ QUESTION: 090 (1.00) i Unit One is at 100% power. While performing LOS-DC-W1, the weekly surveillance for safety related batteries, the equipment operator reports that one pilot cell float voltage on
the Division i 125 VDC battery is reading 2.06 volts. No other plant equipment is inoperable. Which one of the following statements is correct? l a.
The plant is on a 2 hour time clock to restore Div i 125 VDC to operable status.
b.
The plant is on a 24 hour timeclock to restore the individual cell float voltage above 2.07 volts.
. c.
Div i 125 VDC must be returned to operational status within a 7 day timeclock by restoring the individual cell float voltage above 2.13 volts.
, d.
Plant operation may continue indefinitely as long as the remaining connected cells are above 2.06 VDC with an acceptable total float voltage reading of 128 volts.
i OUESTION: 091 (1.00) Which of the following is the basis for limiting reactor operation following the detection of a failed Jet Pump? a.
Increased blowdown area during a Loss of Coolant Accident.
j b.
Unbalanced neutron flux across the core due to flow variations.
c.
Suspected loose parts from the failed jet pump will cause internal vessel damage.
~ ' d.
Invalid APRM Flow Biased Scram setpoints due to the change in flow through a failed jet pump.
-
. . SENIOR REACTOR OPERATOR Page 55 i QUESTION: 092 (1.00) r Per LZP-1200-1, Classification of GSEP Conditions, which ONE of the following precautions is applicable when determining or verifying total gaseous release from the plant? a.
ADD both the Off-gas Pretreatment monitor release rate AND the Reactor Building Ventilation Exhaust Plenum Radiation Monitor release rate.
b.
ADD both the Standby Gas Treatment Wide Range Gas Monitor release rate AND the Station Vent Stack Wide Range Gas Monitor release rate.
ADD both the Off-gas Post Treatment Monitor release rate AND the Standby c.
Gas Treatment Wide Range Gas Monitor release rate.
l d.
ADD both the Reactor Building Ventilation Exhaust Plenum Radiation Monitor release rate AND the Station Vent Stack Wide Range Gas Monitor release j rate.
QUESTION: 093 (1.00) During refueling, blade guides must be used to provide lateral support to inserted control rod blades if two or more fuel assemblies are removed. Which one of the following describes fuel bundles that should be removed to allow blade guides to be installed? a.
Two diagonally adjacent fuel bundles in different control cells.
b.
Two diagonally adjacent fuel bundles in the same control cell, c.
Two parallel faced adjacent fuel bundles in different control cells.
d.
Two parallel faced adjacent fuel bundles in the same control cell.
.
- _ - _ -.. . -... _ -._..- - .. .. ... -. .. _.. -.. _ . -_- t I e I SENIOR REACTOR OPERATOR Page 56 l .i QUESTION: 094 (1.00) The Automatic Depressurization System (ADS)is inhibited during an ATWS situation to l prevent: a.
a rapid loss of RPV inventory which would result in a loss of adequate core < ' cooling.
l l b.
adding a considerable amount of energy into the suppression pool before it is , necessary or required.
! c.
unusable RPV water levelindication as plant conditions will be driven above the RPV saturation curve (RSL).
d.
a large power excursion due to the injection of relatively cold water after , ! reactor pressure drops below the shutoff head of low pressure ECCS.
1 QUESTION: 095 (1.00)
Should it become necessary to lower reactor level during an ATWS condition, only Group A systems are specified for use to maintain level. Select the reason why only Group A systems are to be used? a.
At this point in the ATWS, reactor pressure precludes use of other systems.
b.
Their point of injection is preferred over other systems available for makeup.
c.
These systems are the easiest for the operator to control and maintain a stable level.
d.
These systems provide the cleanest source of water for injection into the
reactor.
l
, . , , -.. ., ..,. . -.. - ,y ,
I l SENIOR REACTOR OPERATOR Page 57 l l QUESTION: 096 (1.00) i Which of the following describes the response of the Scram Discharge Volume (SDV) valves following a half scram signal? ' a.
One SDV vent and drain solenoid valve repositions. All SDV vent and drain valves close.
b.
One SDV vent and drain solenoid valve repositions. All SDV Vent and drain valves remain open.
c.
One SDV vent and drain solenoid valve repositions. One set of SDV vent and drain valves close.
] d.
SDV vent and drain solenoid valves do not change position. All SDV Vent and drain valves remain open.
QUESTION: 097 (1.00) A LOCA has occurred, LGA-01 and LGA-03 have been entered. Containrnent spray is NOT available. The following containment parameters exist: l - RPV pressure: 450 psig - Suppression chamber pressure: 28 psig - Suppression pool temperature: 210 F i - Suppression pool level: + 5 feet - Drywell temperature: 300 F Which ONE of the following describes the correct action AND reasoning at this point in the casualty? _ a.
Reduce RPV pressure with the bypass valves to stay inside the HCTL curve.
b.
Vent the containment to stay inside the PCPL curve.
c.
Ensure core cooling is adequate and secure RPV injection (except boron and CRD) due to being unable to stay inside the SRVTPLL curve.
d.
Emergency depressurize due to being unable to stay inside the PSP curve -.
SENIOR REACTOR OPERATOR Page 58 , , l l QUESTION: 098 (1.00) After power operations lasting several weeks, the reactor scrammed on high drywell
pressure at 1635. All rods fully inserted.
' At 1700 Operators discovered that they were not able to determine reactor - water level and entered LGA-05.
- At 1702 ADS was initiated and 7 SRVs opened.
At 1732 RPV pressure was established at 52 psi above suppressior chamber - pressure.
- The operators then waited the required flooding time.
j What is the Maximum Core Uncovery Time Limit under these conditions if a loss of injection occurs? a.
4 minutes ' b.
5 minutes c.
9 minutes d.
10 minutes i l QUESTION: 099 (1.00? In accordance with LGA-05, which of the following statements concerning the RPV Saturation Curve (RSL) is correct? a.
Once below the RSL curve LGA-05 can be exited.
' b.
ANY time the RSL is exceeded, LGA 05 must be entered.
- c.
If above the RSL with Wide Range indication on-scale, entry into LGA-05 is NOT required.
d.
Entry into LGA-05 is permitted only if above the RSL curve AND below Minimum Useable Level curve.
( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * *) ' . .
-~.. ~ _,. .- . _ _.. .- . -- , . -... -. .. - - -. . _.. - l l SENIOR REACTOR OPERATOR Page 59 ANSWER: 001 (1.00) ANSWER: 004 (1.00) l l d.
b.
! l-REFERENCE: REFERENCE: I Updated SAR 15.1.1, Loss of Feedwater Heater (p.15.1-3) T.S. 3.3.1, ACTION 1 LOA-FW-01, Rev 12, March 22,1988, Loss of a feedwater heater (s) LOA-1(2)PM03J-B401, Rev 4, Oct,1989, 295015K204 ..(KA's) HP HTR16(26) EXTR STEAM CHECK VLV NOT OPEN 295014A107 ..(KA's) ANSWER: 005 (1.00) b ANSWER: 002~ (1.00) REFERENCE: a LSD #21, pg.' 10 , i REFERENCE: ' 295025K309 ..(KA's) j i ' 295018K202 ..(KA's) ANSWER: 006 (1.00) a . ANSWER: 003 (1.00) REFERENCE: _ b EOP Lesson Plan #7 pp. 43,44 REFERENCE: 295025A205 ..(KA's) LSD #49, Appendix B, pg. 54; LOA VR-101, 201; LOA PC-101 288000K402 ..(KA's) -
- - ,
._._ _ _ _ _. _.... _. _... . .. _.. _.. _.. _ - _ _ _... ..._.__ .._ _. _ _ _ _ __ _ _ SENIOR REACTOR OPERATOR Page 60 l l l ANSWER: 007 (1.00) ANSWER: 010 (1.00) l- , l a.
b.
! REFERENCE: REFERENCE:
! L.P. LGA-10 Failure to Scram Rev.
LGA-02.
3,page 26 i p 295037K305 ..(KA's) 295035G011 ..(KA's) I ANSWER: 008 (1.00) . ANSWER: 011' (1.00) [ c.
b.
REFERENCE: l REFERENCE: l-LGA-02 Entry Conditions ' LGA-03, Pool Temperature ' 295036G011 ..(KA's) 295013G011 ..(KA's) ANSWER: 012 (1.00) ANSWER: 009 (1.00) a.
l d REFERENCE: REFERENCE: LSD #26, EHC Logic Diagram, pg. 38; LOA AR-101,1H13-P601 B110 LGP 2-1 295023K203- ..(KA's) 295005K307 ..(KA's)
i !;. ~ ! ! .. .
l ! l SENIOR REACTOR OPERATOR Page 61 ANSWER: 013 (1.00) ANSWER: 016 (1.00) -a.- d.
REFERENCE: REFERENCE: LSD #39, pg.26 LOA DC-03; LSD #43, Appendix B, pg.
226001A401 ..(KA's) ' 263000K201 ..(KA's) - ANSWER: 014 (1.00) c.
ANSWER: 017 (1.00) ' REFERENCE: b.
LOA HY-101, pg. 4; LOP HY-07, REFERENCE: Attachment A LSD #32, pg. 50 and Fig. 32-1 245000K507 ..(KA's) 271000A102 ..(KA's) ANSWER: 015 (1.00) ANSWER: 018 (1.00) 6.
c.
REFERENCE: REFERENCE: - LSD #42, pp. 20, 22, 59 LSD #72, pg. 44; LOP PR-03, pg.1: LOA PR-101 pg. 9 262001K602 ..(KA's) 272000K305 ..(KA's) . '
SENIOR REACTOR OPERATOR Page 62 ANSWER: 019 (1.00) ANSWER: 022 (1.00) b d.
REFERENCE: REFERENCE: LSD #70, pg.18; LOP CO-01, pg. 3 LGP 1-3, REV 27, p.14; 286000A304 ..(KA's) 290002K101 ..(KA's) ANSWER: 020 (1.00) ANSWER: 023 (1.00) b.
a.
l REFERENCE: REFERENCE: LSD #26 pg. 44, EHC logic 'Jiagram.
Lasalle Unit 1 Tech Spec Definition 1.39, Page 1-7 LSD #51, pg.18 241000A107 ..(KA's) 290001K603 ..(KA's) ANSWER: 024 (1.00) ANSWER: 021 (1.00) b.
d.
REFERENCE: REFERENCE: LOA-FW-01; LSD #31, pg. 32 LSD #16, pg.15; LSD #49, Appendix C, LOP PC-03 259001K108 ..(KA's) 215001K401 ..(KA's) .
--. _ .. .. _. _ _. > -.. - - _ ... _. _. _. -. _ _ ..... _. _.. _. __._.
.... _ _. .,
1 , !. ! l l l i SENIOR REACTOR OPERATOR Page 63 l ! i l . ANSWER: 025 (1.00) ANSWER: 028 (1.00) i ! ! l d.
c.
l l l REFERENCE: REFERENCE: . LSD #51 pg.12; LOP VG-01 LOP-RT-09, page 2, section E.1.
. 261000K401 ..(KA's) 204000A304 ..(KA's) ANSWER: 026 (1.00) ANSWER: 029 (1.00) a.
a.
REFERENCE: REFERENCE: LOR-1DGO3J-2-1; LSD #47 pg.29; LOP LOP-RR-07, Ch 6, p12 DG-02 ' 202002K111 ..(KA's). 264000K607 ..(KA's) ANSWER: 030 (1.00) ANSWER: 027 (1.00) d.
d.
REFERENCE: REFERENCE: LSD #6, pp.16,32; LOP RR-11,pg 4; LOR H13-P602-A101 LOP-RW-01, pg.17; LSD #18, pg.18 202002A108' ..(KA's) -201006K101 ..(KA's) l- . ..
. . -. -,. -. . -.. -. - _ _ .
.. _ _. __ _... _... _. _. _ _._. . .. _ _. _.... _. _. _ _. _.. _. _.. _. .... _ _. _. _.. _ _.. _., l ' l l ! ' l I s i SENIOR REACTOR OPERATOR Page 64 ANSWER: 031 (1.00) ANSWER: 034 (1.00)- ! c.
c.
! REFERENCE: REFERENCE: , LSD #12
LSD #39, RHR l l j 215003A401 (KA's) .. 203000A308 (KA's) ] .. _ - ANSWER: 035 (1.00) f-- ANSWER: 032 (1.00) a.
a.
i l REFERENCE: REFERENCE: ! LGP 3-2, REV 36; LSD #20, pg.18 LSD CH. 38, pg. 24 . , 212000K106 (KA's) 216000K304 (KA's) ! .. .. ! ANSWER: 033 (1.00) ANSWER: 036 (1.00) b.
b ' REFERENCE: REFERENCE: LSD #20, pg. 23; LOP NR-01 LSD CH. 41, pg. 26; LSD CH. 49, l Attachment B, pg. 57 ' 212000K502 (KA's) - .. 217000A203 (KA's) .. ,
4- , . ,sn- ,, 9- -ww- -
y
. _.
l SENIOR REACTOR OPERATOR Page 65 l
l ANSWER: 037 (1.00) ANSWER: 040 (1.00) i ' d.
c REFERENCE: REFERENCE: LOP-AP-05, LOP-AP-12.
LAP 1300-7, LAP 100-14, LAP 1600-2, LAP 300-31 294001K107 ..(KA's) 294001A102 ..(KA's) ANSWER: 038 (1.00) d.
ANSWER: 041 (1.00) REFERENCE: d.
10 CFR 55.53.
REFERENCE: 294001A103 ..(KA's) 10CFR50.54(x) and 10CFR50.54(y) LAP-1600-2 pages 6-7.
i ANSWER: 039 (1.00) 294001A109 .. (KA's) a.
ANSWER: 042 (1.00) REFEP.ENCE: c Lap-900-4B, Rev 4, p.9, Attachment A.
294001K102 ..(KA's) LAP 1600-2, LAP AA-04.
su 294001K101 ..(KA's) l ! - [
I l SENIOR REACTOR OPERATOR Page 66 ANSWER: 043 (1.00) ANSWER: 046 (1.00) c.
d.
REFERENCE: REFERENCE: LSD #38, pg.28 10CFR20 209001K201 ..(KA's) 294001K103 ..(KA's) ANSWER: 047 (1.00) ANSWER: 044 (1.00) d.
c REFERENCE: REFERENCE: LAP 1100-3 LSD #36, pg. 23; LOP HP-04 294001K105 ..(KA's) 209002A403 ..(KA's) ANSWER: 045 (1.00) l ANSWER: 048 (1.00) c b.
REFERENCE: REFERENCE: Chapter 32, Off Gas System, LOP - OG-01.
BWR Reactor Theory Manual, Chapter 7 " Operational Physics", Section ll.B, p.12 of 49.
294001K115 ..(K A's) 215004A102 ..(KA's) .
. _ _.. _ _. _. _. - .. _. _ _ _... _ _ _ _. _ ____.. _. _ _ _______._._.
m... _. ~ _ _ _ _ .. i.
-!
! ! ! i ! t SENIOR REACTOR OPERATOR Page 67
l . ANSWER: 049 (1.00) ANSWER: 052 (1.00) . t ? d.- c.
i i l REFERENCE: REFERENCE: '; ! LOS-AA-S1, Attachment B; LSD #14, pg.
LSD #49, Appendix B, pg. 53 l !'
. ! 223002A403 ..(KA's) =215005A208 ..(KA's) l , ANSWER: 053 (1.00) ANSWER: 050 (1.00) { c.
b.
{ REFERENCE REFERENCE: ' LOA-SRV-01 LSD #37, pp. 6,8; Figs. 37-02, 37-04, ADS logic 239002A101 ..(KA's) 218000A206 ..(KA's) ANSWER: 054 (1.00) ANSWER: 051 (1.00) d.
d.
REFERENCE: LSD #31 pg. 32 - REFERENCE: LOP-VO-07, Rev 4, Section F.4.b, p.2.
259002K301 ..(KA's) -223001K501 ..(KA's)
! . I _ -. - -- - - . -... - - -. ... i
.-. _ _ _... _ -.. _ _... -. . ... _.. _.. _.. - _ -. _ - _. _. _ __.
... _ _. _. _ ~. _ ~ _ _., K
- -
, r. - ! ! !
I
. . ) SENIOR REACTOR OPERATOR Page 68' l
! ! ! ANSWER: 055~ (1.00) ANSWER: 058 (1.00) '
d.
l a.
I REFERENCE: i REFERENCE: t - - r . Detail LGA-D1 of LGA-01; LSD #3, pg.
l ! LSD #20, pg. 22; LOP AA-03, pg. 6
i i ,
- .
! i . i l 239001A301 .. (KA's) 295009A201 ..(KA's) - i l t .! ANSWER: 056 (1.00) ANSWER: 059 (1.00) !
a.
b.
! REFERENCE: i REFERENCE: l ' CRD(H) system description, pgs 14 & { 15, sect K LGA 03; LGA 05; EOP lesson plan #5, LGP-3-2, pg 4, sect 8.a.
pg.34 l CRD(H) EKO 6.d.9.
295006K203 ..(KA's) 295024K208 ..(KA's) i
l ANSWER: 057 (1.00) ANSWER: 060 (1.00) d.
b.
_ REFERENCE: REFERENCE: ' LSD #26A, EHC Logic Diagram LOA RP-102, pg.3; LSD #21, pg. 39,~ LSD #20 295007K201 ..(KA's) 295037G010 ..(KA's)
!
i - -- -. - . -.. - - ..,. - , .,. -. . _.
- i i i SENIOR REACTOR OPERATOR Page 69 ANSWER: 061 (1.00) ANSWER: 064 (1.00) 8.
a.
REFERENCE: REFERENCE: LaSalle Technical Specifications, Tables LOP VP-10, pg. 4; LSD #52, pp.14, 20, 2.2.1-1 and 3.3.2-2; 26, 32; LSD #42, pg. 64 LSD #20, pg. 52; LSD #21, pg. 24; LSD #23, pg. 34; LSD
- 26,pg.20 295012A102
..(KA's) , l 295002K201 ..(KA's) ANSWER: 065 (1.00) b.
ANSWER: 062 (1.00) REFERENCE: c RHR SDM, Ch 74 REFERENCE: LSD #42, pp.14-22 and Fig. 42-02; LSD 295016K201 ..(KA's)
- 47,pp.33,34 i
295003A102 ..(KA's) ANSWER: 066 (1.00) b.
' REFERENCE: ANSWER: 063 (1.00) LOA-WR-101, pp. 8, 9; LOA 1 H13-a.
P602-A205, A304, B205,8304 REFERENCE: 295018K101 ..(KA's) LOA-1 H 13-P6C3-A310 l l 295008A103 ..(KA's)
i a
i
i i l SENIOR REACTOR OPERATOR Page 70 l i ANSWER: 067 (1.00) ANSWER: 070 (1.00) i a.
a.
REFERENCE: REFERENCE: LOA-lA-101, Attachment 8, pg.14 LSD #41 295019K203 ..(KA's) 295034A201 .. (KA's) ANSWER: 068 (1.00) j ANSWER: 071 (1.00) { d.
b.
' REFERENCE: ! REFERENCE: Tech Specs Tables 2.2.1 -1, 3.3.2-1, - 3.3.2-2; LSD #20, pp. 20, LaSalle License System Description, 22; Chapter 72, " Process LSD #21, pp. 22, 23 Radiation Monitoring System", Appendix B, Section XI.H, p.70.
295020A206 ..(K A's) 295038K206 ..(KA's) ANSWER: 069 (1.00) ANSWER: 072 (1.00) b b.
REFERENCE: REFERENCE: l LOP-RD-03, pp. 2, 3; LSD #8, pg. 30 LaSalle Licensed System Description
- 39, Residual Heat Removal 295022G007
..(KA's) LOA-RH-03, Rev 4, Jan 21,1989, Loss of RHR Service water 295021K101 ..(KA's) .
. ..-.-.-.-.- -._.-.-.- _ . ..-.__..-. . ...-..-... - -... - - l \\ [ ! SENIOR REACTOR OPERATOR Page 71 ! . ANSWER: 073 (1.00) ANSWER: 076 {1.00) a.
c.
' REFERENCE: REFERENCE: LAP-1600-2, Ops Memo 2 LAP 900-4 Attachment.A page 12 i ' 294001A109 ..(KA's) i 294001K102 ..(KA's) l ANSWER: 074 (2.00) ANSWER: 077 (1.00) a.1 b.
g ' b. 3 L - c. 2 [L d. 4 REFERENCE: l REFERENCE: LAP-100-30, REV 14, p. 7; l 1.
LGA-01 THRU 03.
294001A110 ..(KA's) 295010G011 ..(KA's) ANSWER: 078 (1.00) ANSWER: 075 (1.00) d.
- b.
REFERENCE: .. REFERENCE: LZP-12OO-1; Tech Specs 2.1.2 and 6.4; LSD #49, Appendix B LGA 03 Lesson Plan, Basis for PSP i - Curve 294001A116 ..(KA's) l 295010K101 ..(KA's) l I , i i ) , . - _ - .. -..;. . _, - _ _ -_
l l SENIOR REACTOR OPERATOR Page 72 ANSWER: 079 (1.00) ANSWER: 082 (1.00) a.
b.
REFERENCE: REFERENCE: . 10CFR50 Appendix E, Section EOP Lesson Plan, LGA-03, pg. 24 IV.D.3 ' 10CFR50.72(a)(3) LZP-1310-1 section F.4 page 8 295024K301 ..(KA's) 294001A116 ..(KA's) ANSWER: 083 (1.00) . b.
ANSWER: 080 (1.00) REFERENCE: b Tech Specs 3.04 REFERENCE: 294001A113 ..(KA's) Tech Spec 3.9.11,3.9.8,3.9.9 295023A202 ..(KA's) ANSWER: 084 (1.00) c.
ANSWER: 081 (1.00) REFERENCE: B.
LAP-1200-17, LAP-1200-13 . REFERENCE: 294001A102 ..(K A's) LGA-09 Lesson plan 295038K304 ..(KA's)
. _. _ _ _... -, _.. _.-.__-.._,.--._-_._=_...m _.... -. _ _ _.. _ _.. _.. - _. _ , l
! ! t l - SENIOR REACTOR OPERATOR Page 73 l ! ANSWER: :085 (1.00) ANSWER: 088'(1.00) c.
d.
l REFERENCE: . REFERENCE: LOA-RR-101, REV 1, pg.15 l LZP-1200-1, Rev 17 ! l 202001A206 ..(KA's) 294001A109 ..(KA's) l l ANSWER: 089 (1.00) l ANSWER: 086 (1.00) d.
l' d.
I i REFERENCE: REFERENCE: l ' Technical Specifications 4.1.5; Figs LSD #40, pg. 22; Tech Spec Basis 3.1.51,-2 2.2.1.3 , i l l 211000G005 ..(KA's) 295005K201 ..(KA's) ANSWER:. . i 087 (1.00) ANSWER: 090 (1.00) a.
a ! REFERENCE: REFERENCE: - LSD #67, Fig 67-25; LFP 400-1, pg. 4 TS 3/4.8.2.3, Table 4.8.2.3.2-1 234000K505 ..(KA's) ~ 295004G003 ..(KA's)
! l , !
, i l t - > \\ . >. . .. _ _. -,...- .. .- _,. _ _., , , -. ,, . - - -
_ _ __ _ _ _. . I
l SENIOR REACTOR OPERATOR Page 74
i ANSWER: 091 (1.00) ANSWER: 094 (1.00) a d.
REFERENCE: REFERENCE: Tech Specs Bases 3.4.1 EOP Lesson Plan LGA-10, pg.10 295001G004 ..(KA's) 295015K103 ..(KA's) ANSWER: 092 (1.00) ANSWER: 095 (1.00) b.
b.
REFERENCE: REFERENCE: LZP-1200-1, pg.16 EOP Lesson Plan LGA-10, pg. 30 ) 295038K205 ..(KA's) 295037K209 ..(KA's) ANSWER: 093 (1.00) ANSWER: 096 (1.00) b d.
REFERENCE: REFERENCE: LSD #20, pg.18, Fig. 20-1; LSD #8, pp.
LFP 400-2, pg. 2; LSD #67 pg.18 and 14-16 and Fig. 8-7 Fig. 67-17 201001K40E ..(KA's) 234000 GOO 9 ..(KA's)
. - - _ _ _ .________.m __-.. - ... _ _. _ _ -.. -... _. - _ _ _ _. _, _ _ _ _ _. _ _ _.. _ . l~ l I l l
l ! l
SENIOR REACTOR OPERATOR Page 75 l-l l l ANSWER: 097 (1.00) t d.
] , l ) i i REFERENCE: ,
LGA-03 flowchart. j ! 295010K101 ..(KA's) i
!
- ANSWER: 098 (1.00) i b.
REFERENCE: LGA-05
295031A204.
..(KA's) i l { ANSWER: 099 (1.00) ! l b.
l REFERENCE: i.
LGA-05, REV 6-295028A203 ..(KA's) ..
i ! l ! ! (* * "" * * * * END OF EXAMINATION " * " * * * * *) i er , . -, - .- .
l ! l SENIOR REACTOR OPERATOR Page 76 I i ANSWER KEY
MULTIPLE CHOICE O23 a l 001 d 024 b I 002 a 025 d 003 b 026 a
004 b 027 d 005 b 028 c 006 a 029 a 007 a 030 d 008 c 031 c 009 d 032 a 010 b 033 b 011 b 034 c 012 a 035 a ! 013 a 036 b 014 c 037 d 015 b 078 d ' 016 d 039 a 017 b 040 c 018 c 041 d 019 b 042 c l 020 b 043 c 021 d 044 e i 022 d 045 c ! -
.... _... -.... -.. .... -. _. _.. _ - - -... ..___...m.... . _. _ _ _.. -. . _ -. _ _ _ _ _ _ _ _ _.. _. _.. _. -- i -
i SENIOR REACTOR OPERATOR Page 77 l ANSWER KEY 046 d 068 d MULTIPLE CHOICE.
069 b 047-d 070 a i 048 b 071 b ' , 049 d 072 b !
1 050 b 073. a ' 051 d 074 MATCHING
1 052 e a 1 053 c b 3 054 d c 2
055 d d 4 056 a MULTIPLE CHOICE 057 d 075 b 058 a 076 c
059 b 077 b 060 b 078 d .. 061 a 079 a 062 c 080 b 063 a 081 a 064 a 082 b 065 b 083 c 066 b 084 c . 067_ a 085 c ,
- - . __ ., . _ -. __ _, _.
l l ! SENIOR REACTOR OPERATOR Page 78 i ANSWER KEY
MULTIPLE CHOICE 086 d i ' 087 a 088 d 089 d 090 a , 091 a i 092 b
093 b
l 094 d 095 b 096 d 097 d 098 b 099 b , ( * * * * * * * * * * END OF EXAMIN ATION ' ' ' ' * * * * * * ) l
l ! " }}