ML20140C158
| ML20140C158 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 05/30/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20140C084 | List: |
| References | |
| 50-334-97-03OL, 50-334-97-3OL, 50-412-97-03OL, 50-412-97-3OL, NUDOCS 9706090133 | |
| Download: ML20140C158 (215) | |
See also: IR 05000334/1997003
Text
U. S. NUCLEAR REGULATORY COMMISSION
REGION 1
Docket Nos.: 50-334 and 50-412
Report Nos.: 97-03 and 97-03
License No.: DPR-66 and NPF-73
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Licensee: Duquesne Light Company
Facility: Beaver Valley Units 1 and 2 Nuclear Power Plant
Location: Shippingport, Pennsylvania
Dates: March 17 - 21,1997, and
April 28 - May 2,1997
Chief Examiner: P. Bissett, Senior Operations Engineer / Examiner, Region i
Examiners: L. Briggs, Senior Operations Engineer / Examiner, Region i
D. Prawdzik, NRC Contract Examiner, LITCO
Approved By: Glenn W. Meyer, Chief, Operator Licensing and
Human Performance Branch
Division of Reactor Safety
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9706090133 970530 l
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PDR ADOCK 05000334
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EXECUTIVE SUMMARY
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Beaver Valley Units 1 and 2 Nuclear Power Plant
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Inspection Report Nos. 50-334/97-03 and 50-412/97-03
- Operations
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l Three Unit 1 senior reactor operator (SRO) candidates (one instant and two upgrades), and
j six Unit 2 SRO candidates (two instant and four upgrades) were administered initial
l licensing examinations. All candidates passed all portions of the license examination.
- Generic weaknesses were noted during the Unit 1 examination in the area of crew
! communications and control board awareness during the simulator portion of the operating
- examination. These weaknesses were presented to facility representatives at the
conclusion of the Unit 1 examination. Following completion of the Unit 2 simulator
j examination, it was evident that corrective action had taken place in an effort to correct !
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these identified weaknesses. The NRC examiners observed communications to be more
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direct, succinct, and that all crew members were kept well informed. Also, the candidates
'. readily identified various equipment malfunctions purposely incorporated into the scenarios
as control board awareness exercises. The facility also placed an increased ernphasis on
the Unit 2 written examination development after previous comments from the NRC during l
their initial review of the Unit 1 written examination.
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Report Details j
l. Operations
05 Operator Training and Qualifications
05.1 Senior Reactor Operator initial Examinations
a. Scope
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The examinations were prepared by Beaver Valley Power Station (BVPS) personnel
in accordance with the guidelines in Revision 7, Supplement 1, of NUREG-1021,
" Examiner Standards," and Revision 5 of NUREG-BR-0122, " Examiners' Handbook
for Developing Operator Licensing Written Examinations." The examiners
administered initial operating licensing examinations to three Unit 1 senior reactor
operator (SRO) candidates and six Unit 2 SRO candidates. Unit 1 had one instant
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SRO candidate and two SRO upgrade candidates. Unit 2 had two instant SRO i
candidates and four SRO upgrade candidates. The written examinations were j
administered by the facility's training organization. I
b. Observations and Findinas
The results of SRO examinations for Units 1 and 2 are summarized below:
SRO Pass / Fall Total Pass / Fail
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Written 9/0 9/O '
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Operating 9/0 9/0
Overall 9/0 9/0
The written examinations, job performance measures (JPMs) and simulator l
scenarios for both Units 1 and 2 were developed by Beaver Valley Power Station I
(BVPS) representatives in accordance w'th generic letter guidelines, GL 95-06, j
" Changes in the Operator Licensing Frogram." The exam development team was 1
comprised of BVPS training and operation's representatives. Allindividuals signed )
onto a security agreement once the development of the examination commenced. i
The NRC subsequently reviewed and validated, along with BVPS personnel, all l
portions of the proposed examinations. Various changes and/or additions to the
proposed examinations were requested by the NRC following their review. BVPS
personnel subsequently incorporated the NRC's comments and finalized the
examinations.
The written examinations were administered on March 17,1997, for Unit 1 and on ,
April 28,1997, for Unit 2. Both written examinations consisted of 100 multiple i
choice questions for both Unit 1 and Unit 2. There were no comments by either
the NRC or the utility concerning the validity of questions on the Unit 1 written
examination. For the Unit 2 examination, the answer key for one question was in j
error, and there was one queston that was replaced during the administration of the
written examination due to an administrative oversight. In this instance, one
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question of the examination also had the answer to the question included. This j
i question was replaced with a question of similiar K/A value previously ' verified !
- _ during the examination prep week visit. The NRC and the facility had verified an
extra seven questions during the examination preparation week. I
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j. The operating examinations were conducted from March 18 - 21,1997 and i
j April 29 - May 2,1997. For both examinations, the operating examinations
l consisted of three simulator scenarios and ten JPMs for the instant SRO candidates
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and three simulator scenarios and five JPMs for all SRO upgrade candidates. All
JPMs were followed up with two system-related questions. All candidates were i
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also examined using JPMs to evaluate the administrative requirement portion of the
examination for both Beaver Valley Units,
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Due to the small number of applicants on the Unit 1 examination, genenc ,
, weaknesses based on the written examination were not identified. Based on the .l
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grading of the Unit 2 written examination, the following questions were missed by ;
more than half of the applicants, indicating a weakness in the general understanding
of the subject area. l
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- 12 Recent changes to the independent verification program. )
- 38 Thermodynamics associated with a leaking pressurizor PORV. j
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- 73 Prompt operator actions required on a loss of a vital 12Ovac, bus. i
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- 84 Nuclear inst ument control and instrument power functions and arrangement. !
- 90 Electrical'.nerlocks associated with the circulating water system. j
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Simulator performance by the Unit 1 candidates was, for the most part, good,
however communications at various times was lacking, in that, not all crew
members were kept informed of all actions and events that had occurred. Also, in ;
one instance, it was evident that not all crew members were cognizant of control !
panel conditions of a system malfunction. '
Simulator performance by the Unit 2 candidates was very good. The examiners !
noted that crew briefings were routinely performed by the SROs. One SRO upgrade l
candidate experienced difficulty regaining control of steam generator levels during a ;
low power scenario when the automatic level control system allowed level to :
experience a significant swing of steam generator levels. Communications, in !
general, were very good. The training staff and the candidates appeared to have !
corrected the weaknesses previously identified during the Unit 1 examinations.
In the administrative segrnent of the operating portion of the examination,
administrative job performance measures (JPMs) were used in lieu of administrative
topic questions. The examiners determined that the candidate performance was
good. Two of the Unit 2 candidates, both SRO epgrades, missed an administrative .
JPM concerning the review of a primary system leak rate calculation. Both i
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candidates failed to identify that a correction factor for a change in average coolant
temperature (Tave) had not been used, although a minor change in temperature had l
l occurred. All other administrative JPMs were performed acceptably by all l
candidates.
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JPM followup questions were improved upon from one exam to the next. A review
of the Unit 1 followup questions indicated that several were memory and/or direct i
lookup questions. Tnis was brought to the attention of the training department I
during the Unit 1 preparation week, and corrections and changes were made as
necessary to raise some questions more to the level of cognitive thinking, it was I
noted, during the review of the Unit 2 examination, that JPM followup questions '
were more thought provoking, thus indicating that the facility had placed an
increased emphasis in this area following the identification of the NRC's concern.
c. Cgnclusions
The candidates performed well on both the written and operating examinations, and )
thus were issued licenses. The candidates appeared to be well prepared for the j
examinations. Crew communications and control board awareness improved from
one exam to the next, indicating that an increased emphasis was placed in these
areas by the training department. The BVPS training department did an excellent
job in adhering to the examiner standards and in developing the examination
materials needed to administer the examinations.
05.2 Steam Generator Tube Leakane Procedure Deficiencv .
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a. Scoce of Inspection
The examiners reviewed a portion of procedure 2OM-53C.4.2.6.4(ISS1 A),
Abnormal Operating Procedure (AOP) 2.6.4, Steam Generator Tube Leakage,
Revision 8, (hereafter AOP 2.6.4) during the Unit 2 examination preparation week
(April 14-18,1997). The procedure was reviewed during the scenario verification
process.
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b. Obpjvations and Findinas
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The procedure's wording addressed a steam generator tube leak of 150 gallons per
day (GPD) with a rate of increase of leakage of less than 60 GPD per hour and a SG
tube leakrate that is increasing at more than 60 GPD per hour, The procedure did
not clearly state whether it pertained to a pre-existing leak or a new leak and was
confusing. The procedure wording was discussed with the training staff to resolve
the procedure's intent. The facility agreed that the procedure was not clear in its
intent and that it would be revised to clarify it after the examination. The procedure I
had been recently issued (April 9,1997), and the candidates had not used the new
revision during training. The procedure had been revised to address EPRI document
TR-104788, Primary to Secondary Leak Guidelines.
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During the two performances of scenario 97-3, a tube leak of 700 gallons per day
(GPD) was initiated and remained steady at that leak rate. In both scenarios
! procedure AOP 2.6.4, was referenced by the candidates as expected. However,
l one candidate initiated a normal shutdown and one candidate initiated an
! emergency shutdown. Both actions had been previously discussed and determined
to be acceptable. Facility management had stated during the examinatien
preparation week that the coirset interpretation was that an existing tube leak was
degrading and the type of shutdown (normal or emergency) was dependent on the
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tube leakrate and/or the rate of the tube leakage increase.
l The procedure had two statements concerning primary to secondary tube leakage
l for one steam generator (SG), as follows:
The first statement was, "SG leakrate rate of rise equal to or greater than 60
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GPD/HR." Then, " Perform an emergency shutdown in accordance with AOP
2.51.1, Emergency Shutdown, and be in mode 3 as quickly as possible."
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The second statement was, "SG 1eakrate equal to or greater than 150 GPD AND !
rate of rise less than 60 GPD/HR." Then, " Shutdown plant and be in Mode 3 within !
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />." l
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The scenario initiated a tube leak of 700 gallons per day (GPD) that remained
stable. One candidate interpreted the leak as 700 GPD and steady and initiated a
normal shutdown. The other candidate interpreted it as a leak that initiated from
zero and had increased at a rate of 700 GPD in one hour, which exceeded the 60
GPD/HR rate and initiated an emergency shutdown,
c. Conclusions
The examiners determined during the preparation week, and as demonstrated during
the examination, that Abnormal Operating Procedure 2.6.4, Steam Generator Tube i
Leakage, did not clearly specify the operator actions pertaining to a SG tube leak. {
The facility agreed to revise the procedure wording to clarify the procedure's intent i
subsequent to the examination. The actions to delay procedure revision until after
the examination were considered acceptable to prevent a possible security
compromise of the examination content.
E8 Review of UFSAR commitments
A recent discovery of a licensee operating their facility in a manner contrary to the
updated final safety analysis report (UFSAR) description highlighted the need for a
special focused review that compares plant practices, procedures and /or
parameters to the UFSAR descriptions. While performing the preexamination
! activities discussed in this report, the inspectors reviewed applicable portions of the
l UFSAR that related to the selected examination questions or icpic areas. No
l discrepancies were identified as a result of this review.
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V. Manaaement Meetinas
X1 Exit Meeting Summary
On March 21,1997, and May 2,1997, the examiners discussed their observations from
the examinations with Beaver Valley Unit 1 and Unit 2 operations and training management
representatives. The examiners discussed generic candidate performance, detailed in
paragraph 5.1.b above, concerning communications, control board awareness, JPM
followup questions, and written examination questions developed at the SRO level of
testing. The examiners also stated that significant improvements had occurred frorn one
examination to the next, indicating that increased emphasis and attention had occurred
between the two examinations. The examiners also expressed their appreciation for the
cooperation and acsistance that was provided during both the preparation and examination
week by licensed operator training personnel and operations personnel. Beaver Valley Unit
1 and Unit 2 personnel present at the exit meeting included the following partiallisting in
alphabetical order:
BEAVER VALLEY
Jesus Arias, Director, Licensing
Thomas Burns, Director, Operator Training
James Cross, President, Generation Group
Sushil Jain, Vice President, Nuclear Services
Ronald LeGrand, Vice President, Nuclear Operations / Plant Manager
George Storolis, Techical Assistant General Manager Nuclear Operations
Brian Tuite, General Manager, Nuclear Operations
NRC
Paul Bissett, Senior Operations Engineer, Chief Examiner
Larry Briggs, Senior Operations Engineer
David Kern, Senior Resident inspector, Beaver Valley Units 1 & 2
Attachments:
1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key
2. Beaver Valley Unit 2 SRO Written Examination w/ Answer Key
3. NRC Resolution of BV2 Written Examination Comments
4. Simulation Facility Report
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Attachment 1
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BV-1 SRO WRITTEN EXAM W/ ANSWER KEY
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, RTL #A5.620.H DUQUESNE LIGHT COMPANY Volume 3
' Nuclear Power Division Procedurr 5-5 '
Training Administration Mnual Figure 5-5.1
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i Revision 10
Page1ofI
WRITTEN EXAMINATION COVER SHEET
PROGRAM: Initial Licensed OperatorTraining
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CLASS NUMBER: 2-LOT-1
SUBJECT: Senior Reactor Operator, April 1997 - NRC Initial Licensed Operator Exam.
By this signature,I state that all
of the work done on this examination
is my own. I have neither given nor
received aid.
SIGNATURE DATE
NAME ANSWER KEY DLC EMP #
(Please Print)
COMPANY
(if other than DLC)
POSSIBLE POINTS 100 SCORE
Instructor
Initials
TRAINING DIRECTOR / SUPERYJSOR
PREPARED BY David C.Sibson/ Rich Brooks AP-ROV
SIGNATURE NA V /[V y-r: 1 $$O8
~ [' y ' ueb A-7L ]
" Date
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) ES-402 Policies and Guidelines
for Taking NRC Written Examinations
Attachment 2
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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 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.
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Examiner Standards 5 of 6 Rev. 7. January 1993
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Question 2-971
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.What action is necessary, if any, to defeat interlocks on " Equipment important
to Nuclear Safety" with no permanently installed defeat mechanism?
A. An OSC reviewed and approved procedure.
I B. Written approval from the NSS.
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C. Written approval from the plant manager / designee.
D. These interlocks are NEVER to be defeated.
ANSWER: A. Source: LOT - 0123
REFERENCES: 1/20M-48.3.B pg. B.1 Issue 3 Rev. B
1/2LP-SOS-40.1 OBJECTIVE: 12
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K/A I: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9
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Question 2-97 2
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Which of the following is the LOWEST Emergency Action Level classification that
requires the Operations Support Center to be activated?
A. Unusual Event
B. Alert
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C. Site Area Emergency
D. General Emergency
ANSWER: D. Source: LOT - 0126
REEERENCES: EPP/I-3 pg. 7 Rev. 11
LP-EPP-57.81 OBJECTIVE: 1 j
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K/A #: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 j
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BVPS . Rnr. 2
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Question 2-97-3
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Which of the following identifies the correct sequence for racking out a 4160V '
breaker?
1. Open the 125 VDC control power breaker.
2. Check that the breaker is open. L
3. Turn wrench until breaker is in the disconnect position. '
4. Engage racking wrench, release lock, and turn counter clockwise.
5. Discharge the operating springs.
A. 5, 2, 4, 1, 3.
B. 2, 5, 1, 4, 3.
C. 1, 2, 5, 4, 3.
D. 2, 1, 4, 3, 5.
ANSWER: D. Source: LOT - 0100
REFERENCES : 1/20M-36.4A pg. A1&A2 Issue 3 Rev. 4 l
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2LP-SQS-36.1 OBJECTIVE: 9rr
K/A f: 194001.Kl.07 K/A IMPORTANCE: 3.6/3.7
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DVPS Rev.2
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Question 2-97-4 l
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An operator is assigned to perform a non-energency evolution in a High Radiation !
Area measuring 4 REM / hour. His accumulated dose for the current calendar year I
is 1000 mrem TEDE. Which of the following is the MAXIMUM time that he can !
remain in the area without exceeding the Beaver Valley administrative TEDE Dose
Guide?
A. 15 minutes i
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B. 30 minutes
C, 60 ininutes
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D. 90 minutes
ANSWER: D. Source: M-LOT - 0313
REFERENCES: BVNPD Directive 1.3.1 pg. 2 Rev. 2
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2LP-SOS-GERT OBJECTIVE: 3-5
K/A #: 194001.K1.03 K/A IMPORTANCE: 2.8/3.4 l
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DVPS Rev.2
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Question 2-97 5
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Per Technical Specifications, which of the following describes the process of
making a qualitative assessment of an instrument channel's behavior during
operation, by visually comparing the indication to independent instrument
channels measuring the same parameter?
A. Charmel verification.
B. Channel functional test.
C. Channel check.
D. Channel calibration.
ANSWER: C. Source: LOT - 0316
REFERENCES: TS Definitions
2LP-SQS-TS OBJECTIVE: 7
K/A #: 194001.A1.13 K/A IMPORTANCE: 4.3/4.1
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BVPS - Rev. 2
Question 2-97-6
Which of the following statements describes the MINIMUM required Control Room
staffing when the Unit is in Mode 67
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A. One qualified SRO and one qualified RO must be in the Controls Area.
B. One qualified SRO must be in the Controls Area.
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C. One qualified EO must be in the Control Room Area.
D. One qualified SRO must be in the Control Room Area.
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- ANSWER
- C. Source: LOT - 0323
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REFERENCE 0: 1/20M-48.1 Fage S A.1 and Issue 3 Rev. 4
T.S. 6.2
- 1/2LP-SOS-48.1 OBJECTIVE
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, K/A I: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9 l
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Question 2. 97-7
Which of the following statements is applicable to Temporary Operating
Procedures (TOPS)?
A. TOPS should be used for OM revisions that are required to prevent a plant
trip or a temporary reduction in power.
- B. TOPS must be reviewed by 2 members of the plant staff, one of whom must
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hold an SRO license on the affected unit.
C. TOPS should be used in lieu of OMCNs when the OM changes are extensive,
known in advance, temporary or may be in use longer than 90 days.
D. TOPS must be reviewed by the OSC and approved by the GM - Nuclear
Operations within 14 days of implementation.
ANSWER: C. Source: NEW
RE FERENCES: 1/20M-48.2B Pg. 6 item B.3 Issue 4 Rev. 9
1/2LP-SOS-48.1 OBJECTIVE: 11
l K/A 8: 194001.A1.03 K/A IMPORTANCE: 2.5/3.4
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Page 7
BVPS - Rev. 2
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Question 2-97-8
Under which of the following situations could the performance of an independent I
verification of a valve on a valve list be waived? i
A. Independent verification would result in a personnel exposure greater
than 5 mR.
B. The valve can be verified during the performance of a Temporary Operating
Procedure (TOP) .
C. The valve is located such that scaffolding must be erected to perform the i
independent verification.
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D. The valve can be verified by a functional test using qualified ;
instrumentation, j
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ANSWER: D. Source: New l
REFERENCES: 1/20M-48.3.D.VI.A.6.f Issue 3 Rev. 17
1/2LP-SQS-48.1 OBJECTIVE: 30
K/A f: 194001.Kl.01 K/A IMPORTANCE: 3.6/3.7 )
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DVPS - Rev. 2
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Question 2-97 9 -
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An MOV that is required to be CLOSED on an SI signal and is subject to large i
thermal stresses, is manually placed on its backseat to allow for maintenance. !
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Assuming the backseating was the only action / work performed on the valve, prior {'
l to returning the valve to OPERABLE status, the valve must be
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electrically stroked closed and open, one time.
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l B. electric =11y stroked closed and open, two times. '
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D. nranually stroked closed and open and then electrically stroked closed and '
open.
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ANSWER: B. Source: LOT - 0388 )
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REFERENCES: 1/20M-48.3.D.VI.A.10.e pg. Issue 3 Rev. 16
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1/2LP-SOS-48.1 OBJECTIVE: 23
K/A i: 194001.Kl.01 K/A IMPORTANCE: 3.6/3.7
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BVPS . Rev. 3
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Question 2-9710
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In which of the following areas could one expect to receive a MAXIMUM dose of I
150 mrem in any one hour period?
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1. Radiation Area
2. High Badiation Area
A. 1 Only.
B. 2 Only.
C. 3 Only. i
D. 2 and 3.
ANSWER: B. Source: M-LO* . n482
REFEPINCES: 10CFR 20.1003 pg. 286
2LP-SQS-GERT OBJECTIVE: 4-8
K/A 8: 194001.K1.03 K/A IMPORTANCE: 2.8/3.4
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DVPS Rev.2 '
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Question 2-9711
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The LOWEST Emergency Classification that REQUIRES the implementation of Site !
Accountability is a(n) (1) . This Accountability must be completed within '
(2) .
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,r,, A. Alerts 30 minutes. '
B. Site Area Emergencys 30 minutes.
C. Alerts 60 minutes. j
D. Site Area Emergency; 60 minutes.
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ANSWER: B. Source: LOT - 0723 i
REFERENCES: EPP/I-4 pg. 2 NOTE Rev. 11
LP-EPP-57.81 , OBJECTIVE: 19
K/A f: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 !
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Question 2-97-12
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Which of the following statements is applicable to independent verifications?
1. The independent verification SHALL be performed using the * hands on"
method.
2. Accurate remote or reliable local indication MAY be used.
3. Independent verification for valves on a clearance switching order
CANNOT be substituted by a verification performed by a valve list or
procedure.
4. The original verification and independent verification of a component
MAY be performed simultaneously under certain circumstances.
A. 1, 2 and 3.
B. 1, 2 and 4.
C. 1, 3 and 4.
D. 2, 3 and 4.
ANSWER: A. Source: LOT - 0804
REFERENCES: 1/20M-48.3.D.VI.A.6.c Issue 3 Rev. 17
1/2LP-SOS-48.1 OBJECTIVE: 30
K/A 9: 194001.K1.01 K/A IMPORTANCE: 3.6/3.7
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BVPS - Rev. 2
Question 2-97-13
When taking actions IAN 10CER50.54 (x) that intentionally deviate from Technical
Specifications, NRC notification is REQUIRED taking the actions that
deviated from the Technical Specifications.
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'A. prior to:
B. within 30 minutes after
C. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.after
D. within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after
ANSWER: C. Source: New
REFERENCES: 1/20M-4 6.1. F.VI . A. 3. a Issue 3 Rev. 9
1/2LP-SOS-48.1 OBJECTIVE: 10
K/A 8: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9
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DVPS . Rev. 2
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Question 2-97-14
The VOND marking 'VDM" next to a motor-operated valve (lM OV) indicates that the
MOV.
A. may drift open if the handwheel is engaged.
B. may drift open if the motor is engaged.
C. SHALL NOT be used as a clearance point.
D. can ONLY be used as a clearance point in a LOW pressure (<100 psig)
system.
ANSWER: A. Source: LOT - 0830
REFERENCES: NPDAM 3.4 pg. 12 Rev. 6
1/2LP-SOS-40.1 OBJECTIVE: 25
K/A f: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1
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DVPS . Rev. 2
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Question 2-97-15
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Which of the following describes the Operations Standard with regard to
reactivity control during Mode 1 operations?
j A. All operator initiated reactivity manipulations are to be announced to
the control room crew and approved by an SRO.
B. SRO approval need NOT be received for normal reactivity manipulations but
must be received to initiate a Rx Trip.
1 C. SRO approval for reactivity manipulations is only required for turbine
load adjustments normal rod and boron concentration adjustments may be
made without SRO approval.
( D. SRO approval for reactivity manipulations is only required for rod and
boron concentration adjustments; turbine load adjustments may be made
,
without SRO approval.
ANSWER: A. Source: H-LOT - 0994
REFERENCES: Operations Standards A.14 Rev. 14
1/2LP-SQS-48.1 OBJECTIVE: 39
K/A 8: 194001.A1.03 K/A IMPORTANCE: 2.5/3.4
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BVPS Rev.2
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l Question 2-97-16
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Use of a CAUTION tatt is PROHIBITED for which of the following conditions?
,
A. Special additional manual actions would be required to manipulata a
j component.
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B. Operation of a component will be affected because a portion of the system
! is NOT in NSA.
C. As a temporary replacement for a component label that has fallen off.
D. As a warning that operation of a component will cause erratic indication.
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ANSWER: C. Source: 1-97 Audit
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REE'ERENCES: 1/20M-48.3.L,IV.A.II.C Issue 4 Rev. 2
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! 1/2LP-SQS-48.1 OBJECTIVE: 15
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K/A f: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1
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BVPS - Rev. 2
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Question 2-97-17
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During a maintenance outage, the boundaries on a clearance permit need to be l
moved to support a partial system restoration. The electrical department Work
Party Leader signed on the clearance is offsite and CANNOT be reached. At a j
minimum, whose authorization is necessary to move the boundaries of this
clearance?
i
A. The electrician that assisted the Work Party Leader in performing the
work under the clearance.
I
B. The Work Party Leader's direct supervisor.
C. The Outage Manager.
D. The Site Safety Engineer.
I
ANSWER: B. Source: New
REFERENCES: NPDAP 3.4.IV.A.1.1 Rev. 6
1/2LP-SOS-48.1 OBJECTIVE: 33
K/A 6: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1
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Page 17
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Question 2 97-18
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which of the following st.atements describes the control actions that occ2r when ;
ANY rod control system Rod Stop (C-1 thru C-5) is active? !
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With the Rod Control Selector Switch in
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A. AUTOMATIC, outward rod motion is always prevented, but inward rod
!
action is possible. ;
B. AUTOMATIC, BOTH outward and inward rod action is prevented. !
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C. MANUAL, outward rod motion is always prevented, but inward rod motion '
is possible. .
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, D. MANUAL, BOTH outward and inward rod motion is prevented.
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ANSWER: A. Source: Braidwood - 56 l
REFERENCES: 20M-1.2.B, Page 8 Issue 4 Rev.'3
2LP-SOS-1.3 OBJECTIVE: 10
<
K/A 6: 3.01.001.050.K4.01 K/A IMPORTANCE: 3.4/3.8
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BVPS Rev.2
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Question 2-97-19 ;
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Given the followings
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- The Unit is operating at 2006 power with all systems in their at-power, '
NSA configurations.
,
l * Annunciator (A4-3C] TAVG DEV FROM TREF is received.
- Computer point indicates TAVG is LOW.
You are directed to adjust RCS temperature using the Boration/ Dilution controls. i
l Deterinine the expected valve lineup to accomplish this task.
!
. A. [2CHS*FCV113A) Boric acid flow control valve - OPEN
l [2CES*FCV114A] Primary water flow control valve - CLOSED
[2CHS*FCV113B] Makeup stop valve to the charging pump suction - CLOSED
[2CHS*FCV114B) Makeup stop valve to the VCT - OPEN
Y
B. [2CHS*FCV113A) Boric acid flow control valve - OPEN
[2CHS*FCV114A] Primary water flow control valve - CLOSED
[2CHS*FCV113B) Makeup stop valve to the charging pump suction - OPEN
(2CHS*FCV1148) Makeup stop valve to the VCT - CLOSED
C. [2CHS*FCV113A) Boric acid flow control valve - CLOSED
(2CHS*FCV114A] Primary water flow control valve - OPEN
l [2CHS*FCV113B) Makeup stop valve to the charging pump suction - OPEN.
[2CHS*FCV114B) Makeup stop valve to the VCT - CLOSED
D. (2CHS*FCV113A] Boric acid flow control valve - CLOSED
12CHS*FCV114A] Primary water flow control valve - OPEN
[2CHS*FCV113B] Makeup stop valve to the charging pump suction - CLOSED
(2CES*FCV114B] Makeup stop valve to the VCT - OPEN
ANSWER: D. Source: Modified 1-97-004
REFERENCES: 20M-7.1.D Issue 4 Rev. 3
- 2LP-SQS-7.1 OBJECTIVE: 7
- K/A 9
- 3.01.004.010.A4.03 K/A IMPORTANCE: 3.9/3.7
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BVPS - Rev. 2
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Question. ?-97-20 l
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The Onit is operating at 1004 power with all systems in their at power, NSA I
configurations with the exception of RCS Letdown. The Excess Letoown MX is in 1
service with flow to the VCT, with Normal Letdown secured. A Reactor trip and l'
Safety Injection are manually actuated. PRIOR to resetting the Safety Injection
Signal, Excess Letdown Flow will be_
A. STOPPED, due to the CIA signal CLOSING the Excess Letdown flow control
valve (2CHS*HCV137]. t
B. STOPPED,'due to the CIA signal CLOSING the containment seal return
isolation valves (2CHS*HOV378 and 381).
C. DIVERTED to the PrLaary Drains Transfer Tank (2DGS-TK21] due to the
CIA signal positioning the Excess Letdown flow directing valve
(2CHS*HCV389] to the (2DGS-TK21] position.
D. DIVERTED to the PRT via the seal water return line relief valve
(2CHS*RV382A] due to the closure of the containment isolation valves
(2CHS*MOV378 and 381).
ANSWER: D. Source: Byron - 22
REFERINCES: 20M-7.1.C Issue 4 Rev. 3
2LP-SOS-7.1 OBJECTIVE: 4
K/A #: 3.01.004.020.A1.08 K/A IMPORTANCE: 3.0/3.0
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DVPS . Rev. 2
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Question 2-97-21
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The Unit is operating at 886 power with all systems in th3ir at-power, NSA
configurations. The following indications are observed: ,
!
- Reactor power is RISING. !
- T. is greater than T,.r.
- PZR PORV [2RCS*PCV455C] is OPEN.
- PZR level is RISING.
Which of the following would cause the above listed conditions to occur?
A. An OT/AT Turbine Runback.
B. An Uncontrolled Rod Withdrawal.
C. A Failed OPEN S/G Safety Valve.
D. Power Range channel N-44 failed HIGH.
ANSWER: B. Source: Draidwood - 89
REFERENCES : 20M-53C.4.2.1.3.B Issue 1A Rev. 2
2LP-SOS-53C.1 OBJECTIVE: 2
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K/A 4: 3.01.000.001.EKl.06 K/A IMPORTANCE: 4.0/4.2
Page 21
BVPS - Rev. 2
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Question 2-97-22
Which of the following is the correct sequence of steps that the operator needs
to perform to recover a Dropped Control Rod in Control Bank D (CBD), IAW AOP
20M-53C.4.2.1.5, Dropped RCCA? *
1. Withdraw the Dropped Rod to the height of the remainder of the CBD
rods.
2. Place the Rod Group Selector Switch in MANUAL.
3. Place the Rod Group Selector Switch in the CBD position.
4. Reset the ROD CONTROL SYSTEM URGENT ALARM.
5. Align the CBD Lift Coil Disconnect Switches and Reset the CBD Group
Step Counter and the P/A Converter to Zero.
A. 2, 5, 1, 4.
B . 2, 5, 4, 1.
C. 3, 5, 1, 4.
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D. 3, 5, 4, 1.
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ANSWER: C. Source: New ;
REFERENCES : 20M-53C.4.2.1.5 Issue 1A Rev. 4
.
2LP-SOS-53C.1 OBJECTIVE: 5 "
K/A 8: 3.01.000.003.EA1.02 K/A IMPORTANCE: 3.6/3.4
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Page 22
DVPS . Rev. 2
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Question 2-97-23
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The Unit is operating at 40% power with all systems in their at power, NSA
configurations. An event occurs that causes the following indications:
e Annunciator [A4-8G) ROD POSITION DEVIATION is lit.
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- Annunciator (A4-9F] ROD AT BOTTOM is lit.
- RCS T. dropped from 559'F to 54 9'F.
- PZR Pressure dropped from 2235 psig to 2180 psig.
- Control Bank D Group Step Counters are at 145 steps.
Based on these conditions, which of the following statements describes the
Technical Specification required actions? ,
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A. Restore T.,. to >$51*F within 15 minutes or be in HOT STANDBY ]
within the next 15 minutes.
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B. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; Restore the dropped rod to OPERABLE status within
i 12 steps of its group step demand counter, or adjust the
remainder of the CBD rods to within i 12 steps of the dropped rod.
C. Determine the position of the rod with the failed Digital Rod )
Position Indicator by the movable incore detectors at least once i
per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
D. Restore PZR Press to >2206 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be < 5% of RATED
THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
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Provide the following references: TS 3.1.3.1 Movable Control Assemblies,
3.1.3.2 Position Indication Systems - Operating, 3.2.5 DNB Parameters.
1
ANSWER: D. Source: Braidwood - 116
REFERENCES: U2 TS 3.2.5 Amendment No. 51
2LP-SQS-TS OBJECTIVE: 1 i
K/A 6: 3.01.000.003.G03 K/A IMPORTANCE: 3.3/3.8
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BVPS . Rev. 2
Question 2-97-24
The Unit is operating at 65%. power with all systems in their at-power, NSA
configurations.- Power is being raised with Rod Control in AUTO with Control-
Bank D Group Step Counter at 170 steps, when ONLY the following Annunciators are
received:
'= [A4-8G) ROD POSITION DEVIATION
(A4-5H] NIS POWER RANGE HIGH/ LOW SP FLUX DEVIATION / AUTO DEFEAT
e
[A4-4F] NIS POWER RANGE COMPARATOR DEVIATION
Which of the following events would cause the plant conditions listed above?
A. Rod Control Urgent Failure on the Control Bank D Group 1 Power
Cabinet.
B. One control- rod is misaligned from its group step counter by
greater than 12 steps.
C. A single Digital Rod Position Indicator in Control Bank D has
failed at 157 steps.
D. A single control rod in close proximity to a Power Range Neutron
Detector has dropped.
ANSWER: B. $ource: Zion - 120
REFERENCES: 20M-2.4.AAD/AAK Issue 1/1 Rev. 2/2
2LP-SOS-1.3 OBJECTIVE: 15.1
K/A i: 3.01.000.005.EA2.01 K/A IMPORTANCE: 3.3/4.1
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Page 24
BVPS . Rev. 2
Question 2-97-25
Given the following:
An Automatic Reactor Trip has just occurred.
- The Main Turbine did NOT trip automatically.
The Main Generator Output Breakers (PCB-352 and 362] are still CLOSED.
Which of the following describes the procedural action and bases required for
this situation?
A. Open the Main Generator Output Breakers (PCB-352 and 362) to prevent
motoring the Main Generater.
B. Open the Main Generator Output Breakers (PCB-352 and 3621 to actuate
an additional Main Turbine Trip.
C. Manually Trip the Main Turbine to prevent a Loss of Heat Sink.
D. Manually Trip the Main Turbine to prevent an uncontrolled RCS
cooldown.
ANSWER: D. Source: Braidwood - 97
REFERENCES : 20M-53D.4.E-0 Issue IB Rev. 3 j
2LP-SOS-53A.1 OBJECTIVE: 3
K/A 5: 3.01.000.007.EKl.03 K/A IMPORTANCE: 3.7/4.0
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Page 25
BVPS . Rev. 2
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Questi:n 2-97-26
.
Given the following:
The Unit was operating at 754 power with all systems in their at-
power, NSA configurations when the 'B' Main Feed Pump (2FWS-P21B),
tripped. i
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The operators are responding to the transient IAW Annunciator (A6-10A] 1
STM GEN FEEDPUNP 21A/B AUTO-STOP, ARP 20M-24.4. AAE and the EMERGENCY
SHUTDOWN procedtce, 20M-53C.4.2.51.1.
- The Main Turbine is being run back at St/ minute.
- Rod Control is in AUTOMATIC.
The following additional it.dications are observed: i
e All S/G Levels are approximately 35% and RISING. j
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Annunciator (A6-6A) MOIST SEP DRAINS RCVR TANK / PUMP TROUBLE is LIT. ;
- Annunciator (A4-8Dj ROD CONTROL BANK D LOW is LIT. i
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Annunciator (A4-9D] ROD CONTROL BANK D LOW LOW is LIT. !
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Based on the above indications, which of the following should be done FIRST? l
A. Stop the EMERGENCY SHUTDOWN procedure, and commence a Normal
Boration per 20M-7.4.K, Blender Boration Operation.
B. Continue with the EMERGENCY SHUTDOWN procedure, and
concurrently enter 20M-23B.4.C, HEATER DRAIN SYSTEM SHUTDOWN to ,
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remove the Heater Drain Pumps from service.
C. Continue with the EMERGENCY SHUTDOWN procedure, and
concurrently commence a Normal Boration per 20M-7.4.K, Blender
Boration Operation.
D. Continue with the EMERGENCY SHUTDOWN procedure, and f
concurrently commence an Emergency Boration per 20M-7.4.Q,
Emergency Boration.
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ANSWER: D. Source: Braidwood - 78
REFERENCES: 20M-1.4.AAM Issue 4- Rev. 0
2LP-SQS-1.3 OBJECTIV.C: 17.f
K/A f: 3.01.000.024.EK3.01 K/A IKCRTANCE: 4.1/4.4
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Question 2-97-27
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l The Unit is operating at 100% power with all systems in their at-power, NSA
l- configurations with the following exceptions:
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! * Testing is in progress on Train 'B' of the Solid State Protection
System (SSPS).
- The 'B' Rx Trip Breaker (RTB) is OPEN.
For the above listed conditions, which of the following will result in a Rx
Trip?
-
1. Racking IN but NOT CLOSING the 'A' Train Rx Trip BYPASS breaker.
2. Racking IN and CLOSING the 'A' Train Rx Trip BYPASS breaker.
,
3. CLOSING the 'A' Train Rx Trip BYPASS breaker in the Racked OUT position.
A. 1 and 2. j
B. 2 and 3.
C. 2 ONLY.
D. 1, 2, and 3.
Provide the following reference: FSAR Logic Diagram - Sheet 2, Figure 7.3-7,
Rev. 7.
ANSWER: C. Source: New
REFERENCES: UFSAR Figure 7.3-7 Rev. 7
2LP-SOS-1.2 OBJECTIVE: 0
K/A f: 3.01.000.029.EK2.06 K/A IMPORTANCE: 2.9/3.1
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WE W g $wld h 2- 47- 2-7 8 , a- f
AW Y P O S O- VA WY
M d m p A fa4 a 2-nz7,
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BVPS - Rev. 2
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Given the following conditions: J
e The Unit is operating in Mode 5 with all systems in their required l
lineups for the existing mode. '
s
Annunciator (A11-10D] AUXILIARY BU.LDING SUMP LEVEL HIGH went into
alarm 5 minutes ago followed shortly thereafter by Annunciator )
[A6-1H] PRI COMP COOLING WATER SYSTEM TROUBLE.
l
e The NCO notices that CCP surge tank level is slowly DROPPING.
Which of the following could be the location of the CCP leak?
1
C. Excess L/D HX
D. RCP Motor Cooler.
ANSWER: B. Source: North Anna - 65 i
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REFERENCES: 20M Figure 15-1 Rev. 7
2LP-SOS-15.1 OBJECTIVE: 2 and 13
K/A #: 3.10.000.026.EA2.02 K/A IMPORTANCE: 2.9/3.6
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Question 2-97-28 .
Step 8 of FR-S.1, " Response to Nuclear Power Generation - ATWS", requires that
RCS pressure be verified less than 2335 PSIG and if not, reduce RCS pressure to
less than 2135 psig. The reason RCS pressure is reduced at this point in the
procedure is to
A. maximize boration flow from the high head charging pumps.
B. minimite the possibilii.y of a 77It PORV failing open or leaking
excessively.
,
C. prevent opening the PZR Code Safety Valves.
D. prevent exceeding the RCS subcooling limit assumed in the UFSAR
accident analysis.
ANSWER: A. Source: Byron - 67
REFERENCES: 20M-5 3 B . 4 . FR-S .1 Issue IB Rev. 4
2LP-SQS-53A.1 OBJECTIVE: 3
K/A I: 3.01.000.029.EK3.12 K/A IMPORTANCE: 4.4/4.7
Page 28
DVPS - Rev. 2
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Question 2-97-29 .
Which of the following describes the method used to position the group step
counters and digital rod position indicators (DRPI) to zero steps following the
initial startup of the Rod Drive MG Sets and the DRPI system?
Depress the (1) Pushbutton to reset the group step counters to zero steps
and (2) if any DRPI does not indicate .zero steps.
A. (1) ALARM RESET; (2) depressing the ALARM RESET pushbutton will reset
DRPI to zero.
B. (1) STEP COUNTER RESET; (2) depress the ALARM RESET pushbutton_
C. (1) AIARM RESET; (2) Cycle the ACCURACY MODE SELECTOR switch to the A
ONLY position then back to the A+B position.
D. (1) STEP COUNTERS RESET; (2) request I&C to calibrate DRPI_
ANSWER: D. Source: New
REFERENCES: 20M-50.4.D pg. 10 Issue 1 Rev. 25
2LP-SOS-1.2 OBJECTIVE: 2.d.8
K/A #: 3.01.014.000.A4.04 K/A IMPORTANCE: 2.7/2.7
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Question 2-97-30
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? During a reactor startup, the " ROD AT BOTTOM" annunciator will be LIT when
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1. CBA Rod Bottom Lights.are OFF and CBB, CBC and CDD Rod Bottom Lights are
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2. CBA, CBB and CBC Rod Bottom Lights are OFF and CBD Rod Bottom Lights are j
e LIT. i
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j 3. CBA, CBC and CBD Rod Bottom Lights are OFF and ONLY 1 CBB Rod Bottom .;
j. Light is LIT. ;
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A. 1 and 2 l
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B. 1 and-3 l
C. 2 and 3 f
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D. 3 ONLY
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ANSWER: D. Source: Zion - 4
REFERENCES: 20M-1.4.AAA Issue 4 Rev. 1
2LP-SCS-1.4 OBJECTIVE: 0 3
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-K/A 8: 3.01.014.000.Kl.01 K/A IMPORTANCE: 3.2/3.4
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Question 2-97-31
,
Which of the following components'is credited in the UFSAR accident analysis as
being DESIGNED to mitigate the pressure rise in the RCS following a Turbine Trip.
WITHOUT an accompanying Reactor Trip?
A. PER Code Safety Valves.
B. PER Spray Velves.
C. S/G Atmospheric Dump Valves.
D. PZR Level Control System / Surge Volume.
ANSWER: A. Source Zion - 63
REFERENCES: UFSAR 15.2-6&7 Rev. 0
2LP-SQS-6.5 OBJECTIVE: 3.c
K/A 8: 3.02.002.000.K4.10 K/A IMPORTANCE: 4.2/4.4
Page 3I
BVPS - Rev. 2
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QtGstion 2-97-32
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During a normal RCS cooldown the following data was recorded for the loop with
the LOWEST temperatures.
Time Taor (*F) Teot.o (*F) Tavr (*F)
0430 372 368 370
1 0500 362 358 360
0530 352 348 350
1
0600 342 338 340
0630 332 328 330
0700 322 318 320
Which of the following is the LATEST time frame that the required number of
charging pumps would have been placed in the PULL TO LOCK position to comply
with Technical Specifications?
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A. Between 0500 and 0530.
,
B. Between 0530 and 0600.
C. Between 0630 and 0700.
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D. Between 0900 and 0930.
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ANSWER: C. Source: New
REFERENCES: TS 3.5.3 Ar.endment No. - Original
2LP-SOS-ll.1 OBJECTIVE: 10
K/A #: 3.02.006.000.G05 K/A IMPORTANCE: 3.5/4.2
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Question 2-97-33
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Given the following:
i
- The Unit is operating at 100% power with all systems in their at- i
power, NSA configurations. !
e The pressurizer level selector switch is in position 1, 459/460. I
e [2RCS*LT459] PZR level transadtter, fails at $4%.
Assuming NO operator action is taken, which of the following describes the i
system response when plant load is REDUCED to 50%? j
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Charging flow- I
A. AND actual PZR level will RISE. The Reactor will trip on high PZR level.
1
3. will rise AND actual PZR level will be maintained at 54%. The D/U l
heaters will energize, and the Reactor will NOT trip. l
C. AND actual PZR level will DROP. At 14% actual PZR level, letdown will I
isolate and all PZR heaters will de-energize. I
D. remains constant but actual PZR level DROPS. No control or protective
actions will occur.
ANSWER: C. Source: Zion - 112
REFERENCES: UFSAR Figure 7.3-16 Rev. 3
2LP-SQS-6.4 OBJECTIVE: 14 l
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K/A 1: 3.02.000.028.EA2.02 K/A IMPORTANCE: 3.4/3.8
Page 33
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Question 2-97-34 ;
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'Given the following:
i e The Unit is operating at 100% power with all ayutems in their at-
! power, HSA configurations. t
- The PZR level selector switch is in position 1 459/460.
- [2RCS*LT459] PZR level transmitter reference leg develops a slow leak ;
^
- at the transmitter connection. ;
Which of the following describes the plant response to this reference leg leak?
1
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PIR Level PZR Level VCT level >
l Indicator Indicator Indicator
'
2RCS*LI459 2RCS*LI460 2 CMS *LIll2 ,
A. RAISE DROP RAISE
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B. DROP RAISE RAISE
C. PAISE DROP DROP ,
!
D. DROP RAISE CROP
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- ANSWER
- A. Source: Byron - 99
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REFERENCES: UFSAR Figure 7.3-16 Rev. 3
2LP-SQS-6.4 OBJECTIVE: 14
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K/A 8: 3.02.000.028.EKl.01 K/A IMPORTANCE: 2.8/3.1 1
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Question 2-97 35
2 -
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An RCS cooldown is in progress IAW 20M-53A.4.ES-0.3, * Natural Circulation
Cooldown With Steam Void In Vessel (With RVLIS)." All plant parameters are
, currently STABLE.
'!
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Adjusting charging flow GREATER THAN letdown flow will (1) the size of the
, Rx vessel head void, and cause PZR level to (2) .
,
,> (1) (2)
A. INCREASE RISE
- B. REDUCE RISE
y
( C. INCREASE DROP
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3 D. REDUCE DROP
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ANSWER: D. Source: North Anna - 41
REFERENCES: 20M-53B.4.ES-0.3 pg. 19 Issue IB Rev. 3
2LP-SOS-53A.1 OBJECTIVE: 6 1
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K/A 8: 3.02.011.000.K5.15 K/A IMPORTANCE: 3.6/4.0
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Question 2-97-36
.
-Given the following:
c' An automatic Safety Injection has occurred.
- The operators are about to RESET the Safety Injection signal.
!
Which of the following describes why manual action may be required to restart
safeguards equipment if offsite power is lost after the SI signal is reset?
When the SI signal is RESET.
A. the SI signal to the EDG sequencer is removed, disabling the EDG
sequencer from loading ANY Safeguards Equipment on a subsequent Loss
of Offsite Power.
B. the start signal to SOME of the Safeguards Equipment is removed, even
if a valid SI demand signal still exists.
C. the EDG sequencer is completely disabled until the Rx Trip Breakers
are CLOSED4 to remove the P-4 lockout.
D. the ONLY way to restart the Safeguards Equipment following a
subsequent loss of offsite power is to nanually initiate SI when the
EDG sequencer has finished loading the blackout loads.
ANSWER: B. Source: New
REFERENCES: 20M-53B.4.E-3 pg. 52 Issue 1Bb Rev. 5
a
2LP-SQS-53A.1 OBJECTIVE: 4
K/A 6: 3.02.013.000.A4.02 K/A IMPORTANCE: 4.3/4.4
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Question 2-97 37 l
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When performing a plant heatup and RCS pressure rises above the P-11 setpoint on
(1) channels, the (2) Steam Line Pressure, Steam Line Isolation
will be automatically (3) .
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(1) (2) (3)
A. 1/3 Low unblocked
B. 2/3 High Negative Rate blocked i
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C. 2/3 Low blocked
D. 1/3 High Negative Rate unblocked
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ANSWER: B. Sourcet New j
Rev. 7 i
REFERENCES: FSAR Logic 7.3-12
2LP-SOS-1.1 OBJECTIVE: 5
K/A #: 3.02.013.000.K4.03 K/A iMPORTANCE: 3.9/4.2 !
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Question 2-97-38
.
The throttling process through a leaking Pressuriter PORV is a constant
,
A, enthalpy process which always results in saturation conditions at the
PORV. outlet, dependent only upon PRT pressure.
B. enthalpy process which could result in saturation or superheated >
conditions at the PORV outlet, dependent upon the enthalpy of the ;
steam at the PORV inlet and PRT pressure.
C. entropy process which could result in saturation or superheated *
conditions at the PORV outlet dependent upon the entropy of the steam !
at the PORV inlet and the PRT pressure. ,
D. entropy process which always results in saturation conditions at the
PORV outlet dependent only upon the PRT pressure. j
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ANSWER: B. Source: New (
REFERENCES: Mollier Diagram Issue Rev.
1/2LP-NOMCD-1.1 OBJECTIVE: 8 !
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K/A 8: 3.03.000.008.EK3.02 K/A IMPORTANCE: 3.6/4.1
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Question 2 97-39
.
Given the following:
The Rx was at full power with all systems in their at-power, NSA
configurations when a Rx Trip and Safety Injection occurred.
- All systems functioned as designed.
- S/G Pressures are: 'A'=1005 psig, 'B'=1000 psig, 'C'=1000 psig.
- S/G Narrow Range levels are: 'A'=40%, 'B'=42%, 'C'=35%.
- All RCP's are running.
e
HHSI Flow is indicated on (2 SIS *FI943).
i
e RCS pressure is 1700 psig and slowly RISING.
e RCS subcooling based on CET's is 63*F.
e PZR level is 52% and RISING.
e Containment pressure is 12 psia,
e The operators have made the following EOP transitions:
- E-0, Rx Trip or Safety Injection, to
e
E-1, Loss of Reactor or Secondary Coolant, and are currently at
Step 8 " Check If SI Flow Can Be Terminated."
Based upon these conditions, the operators should._
A. trip ALL RCPs and transition to ES-1.1, SI Termination.
B. maintain all RCPs in service and transition to ES-1.1, SI
Termination.
C. maintain all RCPs in service and continue on in E-1 until SI
tencination criteria are reached.
D. M7NUALLY initiate Containment Spray and transition to FR-Z.1,
Response to High Containment Pressure.
ANSWER: B. Source: Byron - 86 (Modified)
REFERENCES: 2OM-53A.1.E-1 pg. 7 Issue IB Rev. 4
2LP-SOS-53A.1 OBJECTIVE: 6 ,
K/A #: 3.03.000.009.EA2.34 K/A IMPORTANCE: 3.6/4.2
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BVPS . Rev. 2
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Question 2-97 40
The small break LOCA recovery procedure ES-1.2, " Post-LOCA Cooldown and
Depressurization" performs an evaluation to ' Check If An RCP Should Be Started.'
This group of steps includes a check that ensures PER level is greater than 144
prior to starting an RCP.
The bases for this pressurizer level check ensures that when an RCP is started
and the Rx vessel head void is collapsed.
A. sufficient RCS inventory exists to prevent PZR level from dropping off l
scale low. '
B. SI re-initiation criteria for RCS subcooling will NOT be met. i
I
C. RCS pressure does NOT drop below 215 psig, thus maintaining RCP seal
integrity.
D. - adequate PZR steam space is availablo, to limit the RCS pressure rise.
ANSWER: A. Source: Byron - 86, Modified.
REFERENCES: 20M-53B.4.ES-1.2 Step 17 Issue IB Rev. 4
2LP-SOS-53A.1 OBJECTIVE: 3
K/A 9: 3.03.000.009.EK3.21 K/A IMPORTANCE: 4.2/4.5
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Qu3stiln 2 7-41
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j Given the following: *
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l * The Unit was operating at 100% power with all systeme in their at-
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power, NSA configurations when a Loss of Rx Coolant Accident occurred. ,
>
- A Safety Injection and CIB have automatically actuated. !
- All equipment functioned as designed EXCEPT [2RSS*P21C & D) the 'C' '
and 'D' Recirculation Spray Pumps, failed to start and CANNOT be ,
restarted.
- . The operators are in the Loss of Emergency Coolant Recirculation
procedure, ECA-1.1.
90 minutes have elapsed since the Rx Trip and Safety Injection, j
l For the above listed conditions, the Emergency Core Cooling System flowpath and i
flow rate should be one Charging /HHSI pump providing flow through the
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A. INJECTION flowpath, at MAXIMUM rate.
, B. INJECTION flowpath and one LMSI pump through its injection
! flowpath, both at MAXIMUM rate.
I
i C. CHARGING flowpath, throttled to -140 gpm.
!
D. INJECTION flowpath, throttled to -200 gpm.
Provide the followin'g reference: 20M-53A.l.A-4.7 Issue 1B, Rev. 1, Minimum SI
Flow After Trip.
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ANSWER: D. Source: New
l REFERENCES: 20M-53A.l.ECA-1.1 Step 16 Issue 1B Rev. 4
and 21.
.
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2LP-SOS-53A.1 OBJECTIVE: 6
K/A #: 3.03.000.Oll.EA1.ll K/A IMPORTANCE: 4.2/4.2
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BVPS Rev.2 '
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' Qu3stion 2-97-42 !
The Unit has experienced a large-break LOCA. The operators have just i
transitioned to ES-1.3, " Transfer to Cold Lag Recircult$ ion" due to the low '
water level in the RWST. The following conditions are reported by the STA after'
e. A Core Cooling ORANGE path exista due to Core Exit Thermocouples
reading 845'F.
e
An Integrity RED Path exista due to RCS conditions to the LEET of
Limit A.
- LHSI Pump [2 SIS *P21B) has TRIPPED on overcurrent.
All other systems are functioning as designed.
What course of action should the operators take?
A. Immediately transition to FR-P.1, Response to Imminent Pressurized j
Thermal Shock, Step 1. i
B. Complete the alignment to Cold Leg Recirculation (steps 1-6 of ES-
1.3), and then transition to FR-C.2, Response to Degraded Core '
Cooling, Step 1.
C. Immediately transition to ECA-1.1, Loss of Emergency Coolant
Recirculation, Step 1.
D. Complete the alignment to Cold Leg Recirculation (steps 1-6 of ES- ,
1.3), and then transition to FR-P.1, Response to Imminent
Pressurized Thernal Shock, Step 1. l
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ANSWER: D. Source: Braidwood - 93
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REFERENCES: 1/20M-53B.2 pg. 8 Issue IB Rev. 2 l
2LP-SQS-53A.1 OBJECTIVE: 6 , -
)
K/A #: 3.03.000.Oll.G12 K/A IMPORTANCE: 4.0/4.1
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Qu:stion 2-97-43
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The Unit is operating at 83% power with all systems in their at-power, NSA. ,
configurations. Pressurizer Pressure Protection Channel (2RCS*PT455) was '
,
declared inoperable seven (7) hours ago with the associated bistables tripped ,
pe r OM-6. 4. IF, " Instrument Failure Procedure." The on-coming RO notes a concern !
during turnover that the P-11 bistable is NOT tripped because the bistable light l
on BB-B is NOT lit.
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The off-going RO:should
.
A. ct ck the P-11 B/S light bulbs. If it is determined NOT to be a bulb
problem, report the B/S trfpping error to the NSS/ANSS.
-
B. check the P-11 B/S light bulbs. If it is determined NOT to be bulb .
problem, have the other on-duty NCO trip the P-11 B/S. !
!
C. inform the on-coming NCO that the P-11 bistable light is NOT fed from ;
Pressurizer Pressure Protection Channel (2RCS*PT455]. !
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D. inform the on-coming NCO that the P-11 bistable light is NOT required to i
be LIT for the current plant conditions. l
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ANSWER: D. Source: Braidwood - 84
REFERENCES: 20M-6.4.IF pg. I.1 Issue 4 Rev. 4 i
2LP-SQS-1.1 OBJECTIVE: 5 * i
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K/A #: 3.03.000.027.EA2.16 K/A IMPORTANCE: 3. 6/3. 9 - i
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Question 2 97-44
Given the following:
The Unit is operating at 100% power with all systems in their at-
power, NSA configurations.
All procedural and Technical Specification actions for the failed 'A'
Rx coolant loop NR T-COLD RTD are completed.
PZR Pretsure Protection Char.nel II (2RCS*PT456) subsequently fails
HIGH.
,
which of the following describes the actions required to be taken fer the
subsequent PER Pressure instrument failure?
'
A. Trip the histables associated with (2RCS*PT456) I AW 20M-6. 4. IF,
Instrument Failure Procedure within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
B. Carry out the Immediate Actions for E-0, " Reactor Trip and SI."
C. Restore at least one of the failed instruments to operable status
within I hour or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Restore at least one of the failed instruments to operable status
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ANSWER: C. Source: Braidwood - 80
REFERENCES: U2 TS 3.3.1.1 & 3.0.3 Amendment No. 4
2LP-SQS-TS OBJECTIVE: 1
K/A 4: 3.03.000.027.G03 K/A IMPORTANCE: 3.1/3.6
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Question 3-97-45
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The
plantoperators
parameters: are evaluating a Steam Generator tube leak with the following
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e
[2CHS-FI150) Letdown flow indicator reads 105 gpm. '
(2CHS*P21A] The 'A' charging pump is the only running charging pump.
-* PRER level is STABLE. ;
!
e Seal injection and leakoff flows are NORMAL. l
e Charging flow is 110 gpm.
e
Identified RCS leakage is 0.9 gpm. -
e 12CHS*FCV122) Charging riow Control Valve is in AUTO. I
Which
leakage? of the following is the approximate amount of primary to secondary 1
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A. -4 gpm. l
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i C. 19 gpm.
D. 29 gpm.
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ANSWER: C. Source: Byron - 91 !
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REFERENCES: 20M Figure 7-1A Rev. 6
2LP-SQS-7.1 OBJECTIVE: 4
K/A 8: 3.03.000.037.EA2.12 K/A IMPORTANCE: 3.3/4.2
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Question 2-97-46
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( With normal letdown in service, which of the following methods should be
utilized to control PRZR level and the ruptured S/G water level when performing
a Post S/G Tube Rupture Cooldown IAW ES-3.1, " POST-SGTR COOLDOWN USING
BACKFILL"?
'
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l 1. Raise charging flow to raise PRZR level to 76%, isolate charging flow
! to allow PRZR level to drop to no less than 14%, and continue to
l repeat this process.
,
2. Adjust charging flow as necessary to maintain PRER level between 14%
and 76%, but maintain greater than 30 gpm charging flow.
3. Feed the ruptured S/G to 70%, minimize feed flow and allow the
j ruptured S/G water level to drain to no less than 23%, and continue to
l repeat this process.
4. Maintain a relatively constant feed rate to the ruptured S/G such that
l the ruptured S/G water level is maintained at approximately 33%.
l
A. I and 3.
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l B. 2 and 3.
C. 1 and 4.
D. 2 and 4.
ANSWER: B. Source: New
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l REFERENCES: 2OM-53.B.4.ES-3.1 Steps 6-8 Issue IB Rev. 3
'
l 2LP-SOS-53A.1 OBJECTIVE: 6 I
K/A #: 3.03.000.030.EA1.39 K/A IMPORTANCE: 3.6/3.7
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, Question 2-97-47 :
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When carrying out the actions of E-3, " Steam Generator Tube Rupture,*' an RCS I
cooldown is conducted prior to the initial RCS depressurization.
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The temperature at which this _ initial RCS cooldown is terminated is based on the
S/G pressure (s) and ensures
A. INTACTs the Technical Specification cooldown rate limit is NOT
exceeded.
!
B. RUPTURED; the Technical Specification cooldown rate limit in NOT
exceeded.
C. RUPTURED that when the RCS is subsequently depressurized, adequate
RCS subcooling is maintained.
l D. INTACTJ that when the RCS is subsequently depressurized, adequate RCS
j' subcooling is maintained.
ANSWER: C. Source: New
REFERENCES: 20M-53B.4.E-3 Step 15 Issue IB Rev. 5
2LP-SQS-53A.1 OBJECTIVE: 3
K/A #: 3.03.000.038.EK3.06 K/A IMPORTANCE: 4.2/(.5
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Question 2 9748
{
Given the followings ;
l
e The Unit was operating at 100% power with all systems in their at- !
power, HSA configurations. ;
e
Pressurizer Pressure Protection Channel I (2RCS*PT455) was declared -
inoperable and taken out of service with the appropriate bistables '
placed in the tripped condition.
.The controlling pressurizer pressure channel (2RCS*PT444] subsequently
fails HICH. .
Assuming no operator action, which of the following describes the plant response
to this subsequent instrument failure?
A. BOTH PZR spray valves and ALL three PZR PORVs will OPEN resulting in-
a LOW PZR Pressure Rx Trip and SI.
B. The reactor will trip immediately on HIGH PZR. Pressure and a
subsequent SI will actuate on LOW PZR Pressure due to BOTH of the PZR
spray valves OPENING.
C. ALL PZR heaters will turn OFF and BOTH PZR spray valves and CNE PZR I
PORV will OPEN resulting in a LOW PZR Pressure Rx Trip and SL. '
D. ALL PZR heaters will turn OFF and BOTH PZR spray valves and hLL PZR
PORVs will remain CLOSED.
ANSWER: C. Source R-LOT - 0172
l- REFERENCES: 20M-6.4.IF pg. 23 Issue 4 Rev. 4
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2LP-SOS-6.4 OBJECTIVE: 12
K/A 8: 3.03.010.000.K6.01 K/A IMPJRTANCE: 2.7/3.1
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Question 2-97-49
i
'If the Unit was operating at 2006 power and Grid frequency dropped to 59.5
Hertz, the Rx core Critical Heat Flux will and DNBR will .
.
l ' A. drop,- drop !
B. drop, raise
.
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C. raise, drop
l. D. raise, raise
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ANSWER: A. Source: North Anna - 21
REFERENCES: LP-TMO-7 pg. 30 Rev. 6
2LP-TMO-7 OBJECTIVE: 11 & 12- l
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K/A 4: 3.04.003.000.KS.01 K/A IMPORTANCE; 3.3/3.9 l
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Question 2-97-50
Given the following:
!
- .There is a Small Break LOCA inside Containment. !
- All systems responded as designed.
- All S/G Pressures are -1000 psig.
- ~ RCS Pressure is 1223 psig and DROPPING SLOWLY.
Containment Pressure is 8.7 psig.
!
Which of the following describes the reason the RCP's must be' tripped?
A. To prevent excessive depletion of RCS inventory whidh could
lead to severe core uncovery.
B. To prevent RCP motor bearing damage due to the loss of cooling.
C. To prevent an RCP seal failure due to the loss of the seal
water return flowpath.
D. To prevent RCP motor damage due to the high temperature, high ;
humidity operating environment of the containment. j
ANSWER: B. Source: Braidwood - 52
REFERENCES: 20M-53A.1.E-0 Left Hand Pg. Issue IB Rev. 3
2LP-SOS-53A.1 OBJECTIVE: 3
K/A 4: 3.04.003.000.K6.04 K/A IMPORTANCE: 2.8/3.1
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Page 50
BVPS . Rev. 2
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Question 2-97-51
Given the following: l
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- The Unit is operating in Mode 5 with the PZR water solid. l
'
- All letdown orifice isolation valves are OPEN. I
i
Letdown from RHR via RHR Letdown Flow Control Valve (2CHS*HCV142) is I
in service.
l
- The 'A' Train Charging Pump (2CHS*P21A] is in service. )
.
- Letdown Pressure Control Valve (2CHS-PCV145) is inadvertently CLOSED. l
Which of the following describes the plant response as a result of the closure
of (2CHS-PCV145]?
A. The 'A' Train RHR flow will DROP due to the valve closure, RCS
pressure will RISE due to the resulting heatup.
B. The RCS pro =mure will DROP due to total letdown flow being
gra: Lor than cherging flow.
C. RCS pressure will RISE due to coatinued charging flow until the
Cold Overpressure Protection System (OPPS) actuates.
D. There will be NO effect on the RCS because auto control of
charging flow will maintain balanced conditions between letdown
and charging.
ANSWER: C. Source: Byron - 37
REFERENCES: OM Figure 7-1A Issue Rev. 7
2LP-SQS-7.1 OBJECTIVE: 4
K/A 9: 3.04.005.000.K5.05 K/A IMPORTANCE: 2.7/3.1
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DVPS Rev.2
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Question 2-97-52
.
During plant operation with reactor power at 854, the following events occurs
e Annunciator,.(A2-4F) " REACTOR COOLANT PUMP OIL TROUBLE" actuates, for
l the 21A RCP due to a low bearing oil reservoir level.
Annunciator (A2-4E) "RCP MTR BRG TEMP-TR448B HIGH" actuates due to a
i HIGH 21A RCP motor bearing temperature.
I
e All CCP and seal injection flows to the RCPs are normal.
- 21A RCP motor bearing temperature is 201*F and RISING at a rate of
l 5'F/ min.
Which of the following describes the required operator action?
A. Reduce Rx power to <30%, then STOP 21A RCP.
B. Reduce Rx power to <10%, then STOP.21A RCP.
C. Immediately STOP 21A RCP, then trip the Rx.
D. Immediately Trip the Rx, then STOP 21A RCP.
.
Provide the following references: ARPs for (A2-4E) 20M-6.4.AAB & ( A2-4 F] 20M-
6.4.AAC
ANSWER: D. Source: M-LOT - 0802
REFERENCES: 20M-6.4.AAB pg. 8 Issue 4 Rev. 1
2LP+SQS-6.3 OBJECTIVE: 12
K/A 9: 3.04.000.015.EA2.10 K/A INPORTANCE: 3.7/3.7
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DVPS . Rev. 2
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Question 2 97-53 i
.
1
i The Unit is operating at 100% power with all systems in their at-power, NSA l
, configurations when the following annunciators are received:
,
- [A2-4D] REACTOR COOLANT PUMPS SEAL TROUBLE;
j leakoff flow.
e
'
[A2-5D] REACTOR COOLANT PUMP SEAL VENT POT LEVEL HIGH/ LOWS
e Computer point reveals the 'A' Loop RCP (2RCS*P21A) to have a HIGH
j seal vent pot level.
!
! The NCO reports the following additional information on the 'A' Loop RCP
,
t2RCS*P21A):
!
' * No. 1 seal leakoff flow is 0.4 gpm.
I e Seal water outlet temperature is 140*F and STABLE.
- Bearing outlet temperature is 145*F and STABLE.
0 .
. .
j Based on the above information, which of the following events has occurred to
- the 'A' Loop RCP (2RCS* P21A)?
!
A. 42 Seal has failed open.
B. 12 Seal has failed closed.
C. Il Seal has failed open.
D. Il Seal has failed closed.
ANSWER: A. Source: Zion - 123 l
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REFERENCES: 20M-6.4.AAE.B/7.4.AAH.5 Issue 4/1 Rev. 7/16 l
2LP-SQS-6.3 OBJECTIVE: 4.e
K/A #: 3.04.000.015.EK2.07 K/A IMPORTANCE: 2.9/2.9
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Page 53 i
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Question 2-97-54
~
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1
Given the following:
- The Unit was at full power with all systems in their at-power, NSA
configurations.
- An Automatic Safety Injection occurred.
e The operators are .'n E-0, Rx Trip or Safety Injection Response, at
Step 11, " Verify SI status." ~]
e RCS pressure is 105 psig. 1
e The 'following indications are observed on the 'A' Train LHSI Pump
[2 SIS *P21A):
- The Pump Control Switch is in Auto.
- Motor Amps are fluctuating between 18 and 24 amps.
- LHSI Injection Ficw Meter is fluctuating between 1800 and 3000
gPC.
The A Train LHSI Pump.[2 SIS *P21A] is
A operating at runout conditions and the pump discharge valve
(2 SIS *MOV8880A] should be throttled CLOSED.
B. operating as designed, the pump miniflow valve [2 SIS *HOV8890A).is
cycling due to the RCS being at saturation conditions.
C. . cavitating and the pump sucticn valve [2 SIS *MOV8809A) should be
verified OPEN.
D. cavitating and the Recirc Pump Discharge to Safety Injection
Header valve [2 SIS *MOV8811A] should be verified OPEN. l
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ANSWER: C. Source: New i
REFERENCES: 20M Figure 11.1 Rev. 6
2LP-SOS-11.1 OBJECT J* t: 8
K/A 6: 3.04.000.025.EA1.09 K/A IMPORTANCE: 3.2/3.1
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Page 54
DVP$ . Rev. 3
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! Question 2-97-55
4
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Given the following:
j e The Unit was'at 25% power with all systems in their at-power, NSA
.; configurations.
i e A loss of ALL Offsite AC Power has occurred.
- T. is $30'F.
- Teen is at 535'F.
j- * Tm s is at- 541'F.
- The average of the 10 hottest CET's. is $48'F. .
e- PZR pressure is 2160 psig.
f
2
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Which of the following is the current RCS subcooling?
1
1
j A. 93*F.
i
< B. 100*F.
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C. 107'F.
{ D. 110*F.
p
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ANSWER: B. Source: Zion - 97
7-
', . REFERENCES: Steam Tables
'
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1/2LP-NOMCD-1.1 OBJECTIVE: 8
! K/A 8: 3.04.000.074.EKl.01 K/A IMPORTANCE: 4.3/4.7
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Page 55
BVPS . Rev. 2
Question 2 97-56
Given the following:
- The unit is operating at 60% power with ALL systems in their at-power,
NSA configurations.
- Control bank 'D' rods are in AUTO at 190 steps.
The following indications are then observed:
e All S/G Steam Flows, Feed Flows and Water Levels RISE, and then return
to their original values.
- T.. DROPS and then returns to its original value.
- Rx power as indicated on Recorder [2NME-NR45] RISES and then returns
to its original value.
- Control rods move OUT rapidly and then slowly step in to approximately
their original positions.
i
Which of the following failures has occurred? l
A. The Main Steam Residual Heat Release Valve (2SVS*HCV104] has
failed OPEN. -
l
B. The #4 Main Turbine Governor Valve [2TMS-GV4] has failed OPEN. j
i
C. The HP Turbine Impulse Pressure Transmitter [2 MSS *PT446] has l
failed HIGH. l
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D. Power Range Nuclear Instrument channel (2NMP-NI44B] has failed l
LOW. I
ANSWER: B. Source: Braidwood - 73
REFERENCES : LP-ATA-3.1 Rev. O I
2LP-SOS-ATA-3.1 OBJECTIVE: 1&2
K/A 9: 3.04.035.010.K5.01 K/A IMPORTANCE: 3.4/3.9
Page 56
DVPS . Rev. 2
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Question 2-97-57
With the Unit operating at 100% power and all systems in their at-power, HSA
configurations, which of the following statements describes the response of a
main feedwater break as compared to a steamline break?
A. A feedline break will cause the affected S/G to depressurite BEFORE the
Rx Trip; for a steamline break the affected S/G will depressurite AFTER .
the Rx Trip.
)
B. A feedline break will blowdown ALL S/Gs until the FW isolation occurs; a
steamline break will only blow down one S/G.
C. A feedline break initial primary response is a RISE in T.,.; for a
steamline break T. continaously DROPS.
D. A feedline break may result in indicated SGWL RISING or DROPPING,
depending on the location of the breaks for a steamline break SCWL will
always initially DROP.
>
ANSWER: C. Source: Byron - 69
REFERENCES: ATA LP-4.1 & 4.3 Rev. 0
2LP-SOS-ATA-4.1 & 4.3 OBJECTIVE: 1 and 2
K/A 8: 3.05.000.040.EA2.01 K/A IMPORTANCE: 4.2/4.7
.
Page 57
DVP3 - Rev. 2
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Question 2-97-58
l
Given the following:
i
! ~e The Unit was operating at 100% power with all systems in their at-
power, NSA configurations, when a Rx Trip and Safety Injection
occurred.
e 'A' S/G pressure is DROPPING rapidly,
e 'AS/G steam flow is 2.5E6 lbm/hr.
- RCS cold leg temperatures are 238'F and DROPPING.
I e Containment Pressure is 8.5 psig and RISING.
- All MSIVs and Bypass Valves are CLOSED.
- Total AW flow is 395 gpm.
l
e Highest reading Power Range instrument is 1.5% and DROPPING.
l Assuming all ESF Equipment functioned as designed and NO operator action, which
l Critical Safety Function is of the MOST concern?
!
A. Subcriticality
B. Heat Sink
C. RCS Integrity
l D. Containment
j ANSWER: C. Source: Zion - 82
!
l REFERENCES: 20M-53A.1.E-0 Left Hand Issue IB Rev. 3
Pg.
2LP-SQS-53A.1 OBJECTIVE: 6
1
.
K/A # 3.05.000.040.EKl.06 K/A IMPORTANCE: 3.7/3.8
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Page 58 I
BVPS . Rev. 2
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Question 2-97-59
) The Unit has experienced a Reactor Trip and Safety Injection due to'a S/G Tube
- Rupture on the 'A' S/G. The operators are about to commence the initial RCS
-
,'
cooldown at the maximum rate IAW E-3, " Response to a S/G Tube Rupture." The
. following conditions exist:
i
, e 'A' S/G Water Level is 65% Narrow Range and RISING.
tt * RCS T. is 540*F and stable.
j. e Main condenser vacuum is 15" Hg absolute and stable.
e ?B' and 'C' ' Cooling Tower Pumps [2CWS-P21B and C) have tripped.
- e *A' and 'D' Cooling Tower Pumps [2CWS-P21A and D) are running.
$ Which of the following actions are necessary'to conduct the RCS cooldown IAW E-3
"S/G Tube Rupture"?
A. Take the Steam Dumps to the Steam Pressure Mode and manually OPEN
the Steam Dumps.
B. Take the-Steam Dumps to the Steam Pressure Mode, take BCTTH Steam
Dump Bypass Selector Switches momentarily to the DEFEAT TAVG
position, and then manually OPEN the Steam Dumps.
C. Manually OPEN ALL S/G Atmospheric Steam Dump Valves
[2SVS*PCV101A,B&C).
D. Manually OPEN the 'B' and 'C' S/G Atmospheric Steam Dump Valves
(2SVS*PCV101B6C].
ANSWER: D. Source: Zion - 25
REFERENCES: 20M-53A.1.E-3.14.c RNO Issue IB Rev. 5
2LP-SQS-53A.1 OBJECTIVE: 6
K/A ft 3.05.000.051.EK3.01 K/A IMPORTANCE: 2.8/3.1
Page 59
DVPS . Rev. 2
Question 2-9'/-60
Given the followings
e The Unit is operating at 90% power with all systems in their at-power,
NSA configurations.
e RCS T.,, is 574*F and slowly rising on all 3 loops.
- RCS pressure is stable at 2235 psig.
- Stearn Flow on each S/G is 3.78 E6 lbm/hr.
e 'C' S/G feed flow is off scale HIGH.
e 'C' S/G Main Feed Reg Valve (2FWS*FCV498] is full OPEN.
- 'C' S/G level is DROPPING.
- Containment pressure and humidity are RISING.
Which of the following events is in' progress?
A. 'C' S/G Main Feed Reg Valve [2fvS*FCV498) has failed OPEN.
B. 'C' S/G Feed Flow Indicator has failed HIGH.
C. 'C' S/G Feed Line Break INSIDE Containment.
D. 'C' S/G Feed Line Break OUTSIDE Contaitunent.
ANSWER: C. Source: Braidwood - 100
REFERENCES: 20M-53.B.4.E-2 pg. 3 Issue IB Rev. 2
2LP-SOS-24.1 OBJECTIVE: 5
K/A #: 3.05.000.054.EKl.01 K/A IMPORTANCE: 4.1/4.3
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BVPS Rev.2
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i Question 2-97-61
,
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I The Unit is operating at 804 power with all systems in their at-power, NSA
l configurations. One condenser TCV steam dump valve feils full OPEN. Assuming ]
that NO operator action or automatic runback occurs, what will be the resulting j
l Rx power level't -i
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-
A. Cl I
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B. 806
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C .' 05%
D. 904
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ANSWER: C. Sources M-LOT - 0590
l PE f'ERENCES : 20H-21.1.C pg. 5 Issue 4 Rev. 4 l
<
2LP-50S-21.1 OBJECTIVE: 5' ]
l
K/A 8: 3.05.039.000.A2.05 K/A IMPORTANCE: 3.3/3.6
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BVPS - Rev. 2 ;
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i
. Question 2-97-62.
-
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During the performance of FR-S.1, ' Response to Nuclear Power Generation - ATWS"
,
,
launediate action step 4, 'Close Condenser Steam Dump Valves,' the operator
places ONLY the Train 'A' Steam Dump Bypass Selector Switch to- the 'OFF'
,
l- l
l . position (Train 'B' switch was left in the 'ON' position). Which of the i
j following describes the Steam Dump system response't i
i
l
A. The first two banks of valves will CLOSE, but the last two banks are
still ARMED.
B. The last two banks of valves will CLOSE, but the first two banks are
l still ARMED. l
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C. ALL banks of valves CLOSE and are BLOCKED from actuating.
D. NO banks of valves CLOSE and ALL are ARMED for operation.
l
! ANSWER: C. Source: LOT - 0023
l
- REFERENCES
- 20M-21.5.A.12 Issue 4 Rev. O
i
!. 2LP-SQS-21.1 OBJECTIVE: 3
K/A I: 3.05.041.020.A4.08 K/A 1MPORTANCE: 3.0/3.1
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Page 62
DVPS . Rev.2
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e
. Question 2-9743
.
The Unit is operating at 100% power with all systems in their at-power, NSA
- configurations when an EHC equipment failure caused a rapid load rejection to
,
~40% power.
4
3 In response to this rapid load rejection, the Condensate Feedwater Heater Bypass
Valve (2CNM-AOV100) will automatically OPEN
A. if a low main feed pump suction pressure is detected and will
, automatically CLOSE once normal pressure has been restored.
, B. if a low main feed pump suction pressure is detected and may be
s manually CLOSED after a four minute time delay.
!
5 C. on a C7B signal and will automatically CLOSE once normal pressure has
i been restored.
>
.. D. on a C7B signal and will automatically CLOSE after four minutes has
] elapsed following the load rejection.
ANSWER: D. Source: M-LOT - 0411
REFERENCES: OM Figure 22A-12 & 23B-ll Rev. */
2LP-SOS-22A.1 OBJECTIVE: 4
K/A #: 3.05.056.000.Kl.03 K/A IMPORTANCE: 2.6/2.6
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Page 63
BVPs.Rev.2
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Question 2 97-64
.
The Unit is operating.at 954 power with all systems in t!eir at-power, NSA
configurations. The selected steam pressure input to the 'B' S/G water level
control (SGMLC) systea [2 MSS *PT485), fails LOW. Which of the following
' describes the plant response, if any, to this instrument failure?,
A. There will be No effect to the SGWLC system due to the median select
design feature.
B. 'B' S/G Main Feed Reg Valve [2WS'FCV488] throttles CLOSED to maintain
33% level. I
C. 'B' S/G Main Feed Reg Valve [2 NS*FCV488] initially throttles OPEN and
then CLOSED to maintain 44% level.
D. 'B' S/G Main Feed Reg Valve.[2FWS*FCV488]. initially throttles CLOSED
and then OPEN to maintain 44% level.
ANSWER: ' D. Source: M-LOT - 0195-
REFERENCES: 20M-24.1.0 pg. 5 Issue 4 Rev. 2
2LP-SQS-24.1 OBJECTIVE: 5
K/A 9: 3.05.059.000.A2.ll K/A IMPORTANCE: 3.0/3.3
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BVPS - Rev. 2
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Question 2-97-65
.
,
- l
l
Given the following: I
l
- The Unit was operating at 100% power with all systeins in their at- l
l power, NSA configurations when Grid instabilities caused a Unit Trip.
! e All systems functioned as designed EXCEPT that an 'A' S/G AN flow
control valve (2 NE*HCV100E) is mechanically bound in the full OPEN
position and CANNOT be SHUT.
e 'A' S/G Wide Range Water Level is now 78% and RISING.
!
!
Which of the following should be done to prevent overfilling the 'A' S/G yet
maintain an adequate Heat Sink?
i
i 1. Throttle CLOSED the other 'A' S/G AFW flow control valve
[2 NE*HCV100F].
2. STOP the 'A' Train Motor Driven AW Pump [2 NE*P23A). l
,
3. STOP the 'D' Train Motor Driven AW Pump (2FWE*P23B) . ,
l 4 STOP the Steam Driven A W Pump (2NE*P22).
A. 1 ONLY. ,
{
B. 1, 2 and 4. t
C. 2 and 3.
D. 3 and 4
'
ANSWER: D. Source: New
REFERENCES: OM Figure 24-3 Rev. 7 !
2LP-SQS-24.1 OBJECTIVE: 7
K/A 4: 3.05.061.000.A2.07 K/A IMPORTANCE: 3.4/3.5
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DVPS . Rev. 2
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Question 2-97-66
~
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I
Given the followings
- The Unit is operating at 100% power with all systems in their at-
power, NSA configurations EXCEPT that the 'A' Train Service Water Pump ;
(2SWS*P21A] is being put on Clearance for motor replacement. '
[2SWS*P21A) 4KV breaker has been racked out and tagged OPEN.
- The 'B' and 'C' Service Water Pumps (2SWS*P21B] and [2SWS*P21C) are
running.
- Upon review of the Clearance paperwork it was noted that when
performing the Clearance, the operator inadvertently racked the Swing
Service Water Pump (2SWS*P21C) onto the 2DF 4KV bus instead of the 2AE
4KV bus,
j
What are the Technical Specification implications, if any, for the Swing Service
Water Pump [2SWS*P21C) being powered from the wrong 4KV bus?
A. There are NO Technical Specification implications provided at i
least one Standby Service Water Subsystem is OPERABLE. 1
I
B. Both trains of Service Water may be considered OPERABLE based I
on the provisions of Technical Specification 3.0.5. ;
1
C. Establish the 2AE 4KV bus as the power supply to [2SWS*P21C) l
within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Establish the 2AE 4KV bus as the power supply to [2SWS*P21C)
,
within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
l
j
Provide the following references: TSs 3.0.3, 3.0.5, 3.7.4.1 and 3.7.13.1. l
l
ANSWER: D. Source: New
REFERENCES: TS 3.7.4.1 Amendment No. - Original
i 2LP-SOS-TS OBJECTIVE: 1
l
K/A 8: 3.05.076.000.G05 K/A IMPORTANCE: 2.8/3.2
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BVPS . Rev. 2
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, Question 2-97-67
1
!
Under which of the following conditions would the Containment Integrity
1. Technical Specification be satisfied?
!
4
I
NOTE: Assume there are NO blank flanges or pipe caps installed on equi.pment
that is disassembled or removed.
A. e
21A S/G Blowdown isolation valve [2BDG*AOV101A1) is
REMOVED,
'
21A S/G Blow own isolation valves [2BDG*AOV101A2) and
[2BDG*465) are CLOSED and
- . Drain valve [2BDG*547) is OPEN with the pipe cap removed.
'
B. e Containment Equipment Hatch installed with three closure
bolts, and
i e BOTH doors on the Emergency Personnel Access Hatch are
CLOSED, and
e All containment purge dampers are CLOSED.
.
C. e CCP return header relief valve [2CCP*RV105) is removed,
o
CCP return header isolation valve [2CCP*MOV157-2) is
CLOSED, and [2CCP*MOV157-1) is OPEN and
e Drain valve [2CCP*926) is OPEN with the pipe cap removed.
D. * The CVCS charging line is disconnected at the Regen Heat
Exchanger inlet,
e Charging Header vent valve [2CHS-730) is locked OPEN,
e Charging Header Isolation valve [2CHS*MOV289) is OPEN, and "
l
- Charging Header Manual Isolation valves [2CHS*30] and !
[2CHS*477) are CLOSED. l
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Provide the following references: VOND Figures; 7-1A Rev. 6, 15-2 Rev. 8 and
25-1 Rev. 9.
1
ANSWER: A. Source: M-1-97-057
REFERENCES: 20M Figures 7-1A, 15-3 and Rev, 6, 2 and 9
25-1.
2LP-SOS-TS OBJECTIVE: 1
%
.J
K/A 8: 3.06.000.069.G08 K/A IMPORTANCE: 3.4/4.1
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Question 2-97-68 I
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A Large Break Loss of Coolant Accident has occurred inside containment. The l
2AE 4KV tie breaker to the 2N 480v bus [ACB-2 Ell) has tripped and CANNOT be !
closed.
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The Transfer to Recirculation signal is now present. Which of the following i
describes how the recirculation components will or will not be affected?
A. No affect, ALL components will function as designed due to the auto
transfer of the 2N bus feed from the 2AE bus to the 2P bus.
B. Transfer to Recirculation will NOT occur because Recirculation Spray
pumps [2RSS*P21AEC) are deenergized.
C. Transfer to Recirculation will occur on 'B' Train components ONLY
because the 'A' Train motor operated valves are deenergized.
D. Recirculation Spray pump (2RSS*P21A] will NOT supply the LHSI Header
due to the Recirculation Spray pump discharge isolation valve
[2RSS*MOV156A] being deenergized. !
!
ANSWER: C. Source: New
REFERENCES: 2OM-37.5.B.7 Table 37.7 Issue 4 Rev. 7
pg.38-132 & Fig. 10080-RE-lC
2LP-SOS-ll.1 OBJECTIVE: 6 and 11
K/A #: 3.06.026.000.K2.02 K/A IMPORTANCE: 2.7/2.9
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Question 2-97-69
A Large Break Loss of Coolant Accident has occurred inside containment.
Thel,'B' r Train Hydrogen Analyzer (2HCU*HA100B)
A, will receive an automatic start signal from the Train 'B' SIS signal.
.
B. will automatically start after a time delay if (2HCS*HA100A) fails to
achieve adequate sample flow.
C. must' be manually started from the Control Room when directed in the
EOP network.
D. must be manually started from its local Control Panel when directed in .,
the EOP network.
'
ANSWER: A. Source: New
REFERENCES: 2OH-46.1 pg. 1 Issue 1 Rev, 5 .i
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2LP-SQS-46.1 OBJECTIVE: 5
K/A ft 3.06.028.000.A4.03 K/A IMPORTANCE: 3.1/3.3
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Question 2-97-70
Given'the following conditions:
e The Unit is operating at 1004 power with all systems in their at-
power, NSA configurations.
-e A large, audible leak is reported at the equipment hatch seal into
. containment..
Which of the following describes the action to be taken?
A. Quantify the leakrat's to ensure the maximum allowable equipment
hatch leakrste is NOT exceeded.
B. Reduce power to less than 50% until the leak is repaired.
C. Restore containme'nt' integrity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. Restore containment integrity within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in HOT
STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
>
ANSWER: D. Source: North Anna - 53
REFERENCES: TS 3.6.1.1 Amendment No. 80
'2LP-SQS-TS OBJECTIVE: 1
K/A 6: 3.06.103.000.K3.02 K/A IMPORTANCE: 3.8/4,2
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- Question 2-97-71
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The Unit was operating at 1004 power with all systems in their at-power, NSA !
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2
configurations when a complete Loss of Offsite Power occurred. The EDGs start '
and energize the emergency AC buses. A Natural Circulation cooldown is being '
performed IAW ES-O.2 " Natural Circulation Cooldown.' The following major action
steps have been accomplished:
(
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e Cold shutdown boron concentration has been verified, l
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- Two CRDN fans are running, )
e RCS cooldown to cold shutdown has been initiated.
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At this point, RCS hot leg temperatures are checked to determine if they are
j- less than 550'F. l
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What is the purpose of this RCS hot leg temperature check at this point in the f
l_ procedure?
2
( A. To determine if the RCS cooldown has resulted in steam void formation !
! in the upper head of the reactor vessel. ;
i
B. To verify that natural circulation flow still exists between the core !
and the S/Gs.
I
C. To verify that the RCS cooldown has not resulted in a challenge to the j
RCS integrity. critical safety function.
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D. To ensure that at least 50*F RCS subcooling will be maintained during l
the subsequent RCS depressurization.
!
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ANSWER: D. Source: New !
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REFERENCES: 20M-53B.4.ES-0.2 Step 7 Issue IB Rev. 3 l
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2LP-SOS-53A.1 OBJECTIVE: 3
i
K/A 4: 3.07.000.055.EA1.01 K/A IMPORTANCE: 3.7/3.9
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Question 2-97-72
During the performance of ECA-0.0 * Loss of All AC Power", a rapid S/G
depressurization to 300 psig is performed to reduce RCS tem.c::rature and
pressure.
The bases for STOPPING the S/G depressurization at 305 psig is to ensure that
A. the maximum Technical Specification cooldown rate is NOT exceeded.
B. a steam void will NOT be created in the Rx vessel head.
C. the challenge to the RCS Integrity Critical Safety Function is limited
to that assumed in the Accident Analysis.
D. RCS pressure is maintained above the minimum pressure to preclude
injection of Safety Injection Accumulator N2 into the RCS.
ANSWEPt D. *
Source: New
REFERENCES: 20M-53.B.4.ECA-0.0 pg. 116 Issue IB Rev. 3
2LP-SOS-53A.1 OBJECTIVE: 3
K/A 8: 3.07.000.055.EK3.02 K/A IMPORTANCE: 4.3/4.6
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Question 2-97-73
.
The Unit was operating at 100% power with all systems in their at-power, NSA
configurations when a loss of 120 VAC Vital Bus III occurred. Which of the
following will require expeditious manual operator control action to prevent a
Rx Trip?
A. Pressuriter Level.
B. Pressurizer Pressure.
C. S/G Feed Flow.
D. Main Turbine Load.
ANSWER: C. Source: Byron 75
- REFERENCES: 20M-38.4.V pg. 1 Caution Issue 1 Rev. 5
.
2LP-SQS-38.1 OBJECTIVE: 8.h
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K/A f: 3.07.000.057.EA1.06 K/A IMPORTANCE: 3.5/3.5
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BVPS Rev.2
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., Question 2-97 74
.
1
The reactor is critical at 1.0 E. CPS in the source range. The 120 VAC Vital f
'
Bus II inverter _ output breaker tripped OPEN. This causes Vital Bus II to be ;
doenergized. I
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A reactor trip will occur due to the loss of power to [
!
A. SSPS Train 'B' Logic Cabinet. ,
B. SR channel N-32. ,
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C. IR channel N-35. 7
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D. Rod Control low voltage power supply. 1
.
!- ANSWER: B. - Source: Braidwood - 91 I
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REFERENCES: 20M-2.3.C pg. 4 Issue 4 Rev. 2 $
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2LP-SOS-2.1 OBJECTIVE: 6 j
.K/A is 3.07.000.057.EA2.19 K/A IMPORTANCE: 4.0/4.3
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Question 2-97-75
.
Given the followings
\
- The Unit was operating at 100% power with all systems in their at-
power, HSA configurations EXCEPT;
- The 'A' Charging Pump [2CHS*P21A] is NOT running but its
breaker IS racked onto the 2AE bus.
- The 'C' Charging Pump [2CHS*P21C] is running on the 2AE bus.
- A loss of DC control power to the 2AE bus has occurred.
- While stabilizing the unit, a spurious S1 occurred.
Which of the fo13owing charging pump combinations will exist as a result of
these failures?
[2CHS*P21A1 (2CHS*P21B] (2CHS*P21C)
h. Stopped Running Stopped
B. Stopped Running Running
C. Stopped Running Stopped
D. Stopped Stopped Running
.
ANSWER: B. Source: North Anna - 95
REFERENCES: [2CES*P21C) 12241-E-SDM Rev. 11
2LP-SQS-7.1 OBJECTIVE: 7
K/A 8: 3.07.000.058.EA2.03 K/A IMPORTANCE: 3.5/3.9
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Questi:n 2-97-76
.
The Unit is operating at 1004 power with all systems in their at-power, NSA
configurations. Under which of the following conditions would the System
Station Service Transformer (SSST) Normal 41N Breaker [ACB-42A) CLOSE?
5
A. The Live Bus Transfer Switch is placed in the ON position with the
SSST Normal 4KV Breaker (ACB-42A) set up for auto transfer.
B. The Unit SST Normal 4KV Breaker (ACB-42C] trips OPEN on an
OVERCURRENT fault.
C. The Unit SST Normal 4KV Breaker (ACB-42C] is Manually OPENED from BB-
C.
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D. The Live Bus Transfer Switch is placed in the ON position and the 1
control switch for [ACB-42A].is then placed in the CLOSE position.
ANSWER: D. Source: New
REFERENCES: 20M-36.1.E pg. 30 Issue 4 Rev. 4 j
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!' 2LP-SOS-36.1 OBJECTIVE: 6
l
! K/A 6: 3.07.062.000.K4.03 K/A IMPORTANCE: 2.8/3.1
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Question 2-97-77
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i 'Giventhefollowibg: i
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The Unit is operating at 1004 power with all systems in their at-
, power, NSA configurations.
- - *
EDG No. 2-1 is running unloaded to cooldown following the monthly OST.
i
A Loss of.the DC SWBD 2-1 has occurred.
8
' Based on the above information, which of the following actions will STOP the 2-1
.EDG7
_
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} 1. Simultaneously depressing BOTH of the EDG STOP pushbuttons on BB-C.
,
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2. Depressing the EDG STOP pushbutton on the Local EDG control panel.
e
3. Placing the mechanical governor lever on the EDG fuel racks to the STOP
- - position.
'
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A. - 1 and 3.
!
- B. 2 and 3.
O
i C. 1, 2 and 3.
1
- - D. 3 ONLY.
2
ANSWER: D. Source: Zion - 3, Hodified
REFERENCES: OM Figure 36-3 & 12241-E- Revs. 9,7,8,8,2.
12H,J,K&L
2LP-SOS-36.1 OBJECTIVE: 6
K/A 8: 3.07.063.000.K3.01 K/A IMPORTANCE: 3.7/4.1
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~ DVPS . Rev. 2
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Question 2-97 78
.
Given the following
e
The Unit is operating at 100% power with all systems 'in their at-
power, NSA configurations when a complete Loss of Offsite Power
occurs, at 10:00:00.
A spurious safety injection signal is generated at the same time.
The operators are responding to the transient IAW E-0 " Reactor Trip or
Safety Injection."
e
At 10:02:10 the AFW pump status is checked; the turbine driven AFW
pump (2FWE*P22) is running, but the motor-driven AFW pumps i
[2fWE*P23A&B] are NOT running.
i
The Motor Driven AFW Pumps (2FWE*P23A&B) should. ;
A. . NOT be running, the EDG sequencers will NOT start the motor-driven AFW
pumps for another 10 seconds.
D. NOT be running, the motor-driven AFW pumps will NOT start
automatically unless the turbine-driven AFW pump fails to start. j
C. be running, the motor-driven AFW pumps should have started immediately
upon the trip of the second main feedwater pump.
D. be running, the EDG sequencers should have already started the motor-
driven AEW pumps.
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ANSWER: D. Source: M-LOT - 0232 !
REFERENCES: 20M-36.1.C pg. 14-15, 20M- Issue 4/4 Rev. 2/2
24.1.D pg. 16 and TS 3.3.2.1 Table 3.3.5
2LP-SOS-24.1 OBJECTIVE: 10
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K/A #: 3.07.064.000.K4.11 K/A IMPORTANCE: 3.5/4.0
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Question 2-97-79
-
.
Under which of the following situations should the station air to containment
instrument air cross tie valve (2IAC-MOV131] be directed to be OPENED?
A. Only ONE Containment Instrument Air Compressor is operational.
.B. The normal containment instrument air containment isolation valve
[2IAC*MOV130] is failed CLOSED.
C. - During the recovery from a Large Main Steam Line break inside
containment;
D. Following a Loss of Offsite Power without an SI and the Black Diesel
Generator fails to start.
'* ANSWER: C. Source: New
REFERE:CES: 20M-53.B.4.ES-1.1 Step 8 Issue 1B Rev. 5
2LP-SQS-53A.1 OBJECTIVE: 6
K/A 0: 3.08.000.065.EK3.04 K/A IMPORTANCE: 3.0/3.2
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Page 79
BVPS - Rev. 2
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3 Question 2-97-80
.
!-
i Which of.the following indications are available on the Emergency Shutdown Panel
(SDP) ?
,
!
' A. Auxiliary Feedwater Flow, Containment Pressure, Charging Flow. '
I. B. RHS HX Outlet Temperaturei S/G Wide Range Level, RCS Wide Range
Temperature.
1
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C. Pressurizer. Level, Rx Trip Breaker Position, Steam Generator Pressure.
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D. Letdown Flow,. Intermediate Range SUR, Charging Header Pressure,
j
~ ANSWER: B. Source: Braidwood - 92
3- REFERENCES: 20ST-45.2[ Issue 1 Rev. 8
2LP-SQS-53C.1 OBJECTIVE: 6
K/A I: 3.08.000.068.EK2.01 K/A IMPORTANCE: 3.9/4.0
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DVPS Rev.2
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Question 2-97 81
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l The Unit is operating at 100% power with all systems in their at-power, NSA
l. configurations.
l
If the Source Range High Voltage (HV) Manual control switch for N-31 is placed
in the "ON" position, N-31 High Voltage will (1) , the high flux reactor
l
. trip status light will (2) , and a Rx Trip (3) occur.
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! (1) (2) (3)
A. remain OFF remain OFF will NOT
l B. turn ON remain OFF will NOT
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C. turn ON turn ON will NOT
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D. turn ON turn ON will
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ANSWER: C. Source: R-LOT - 0763
REFERENCES: 20M-2.1.B pg. 7 & 20M2.2.5 Issue 4/1 Rev. 12
figure 2-2
2LP-SQS-2.1 OBJECTIVE: 5
K/A 8: 3.09.000.032.EA1.01 K/A IMPORTANCE: 3.1/3.4
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Question 2-97-82
A Technical Specification Action Statement entry would be required if the Unit
is and I&C reports that_
A. at 84 power; the N-35 Hi Flux Trip Bistable setpoint is the current
equivalent of 35%.
B. conducting a Rx startup with IR level at 1.0 E amps; ALL RCP
underfrequency trip relays were calibrated with ~ a frequency meter that
was out of calibration in the NON-conservative direction.
C. in Mode 1; BOTH Source Range instruments should be declared
inoperable due to the failure of the detector cables.
D. in Mode 3; the Turbine Impulse Pressure Transmitter that feeds P-13
should be declared inoperable due to a leaking capacitance bellows
assembly.
ANSWER: A. Source: New
REFERENCES: TS 3.3.1.1 Amendment No. 10
2LP-SOS-TS OBJECTIVE: 1
1
K/A #: 3.09.000.033.G08 K/A IMPORTANCE: 2.8/3.4
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DVPS . Rev. 2
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Question 2-97-83
The Unit is operating at 100% power with all systems in their at-power, NSA l
configurations. What are the affects on the Solid State Protection System i
(SSPS) if the 120vac Vital Instrument Bus III is de-energized? I
A. ALL of the Train B, Output Bay Slave Relays will NOT function on a
Safety Injection signal.
D. ONLY the #2 Emergency Diesel Generator Load Sequencer will load the l
required components on a Safety Injection signal, l
C. ALL of the Train B, Input Bay Relays will de-energize resulting in a
Rx Trip and Safety Injection.
D. Train B will function as designed due to the auctioneered power
supplies to the Logic Bay.
ANSWER: D. Source: M-1-97-69
REFERENCES: OM Figure 1-41 Issue 1 Rev. 5 l
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2LP-SOS-1.2 OBJECTIVE: 7 i
K/A #: 3.09.012.000.Kl.01 K/A IMPORTANCE: 3.4/3.7
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Question 2-97-84 ' '
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A reactor startup is in progress with IR power at 3.0 E*** amps. The source range
High Flux Trip has NOT been blocked. Which of the following describes the . o
Reactor Protection System response if a CONTROL POWER fuse blows on the N-31 1
Source Range instrument with the Level Trip Bypass Switch in the positions ;
indicated? l
i
Level Trip Bypass Switch Position
1;ORMAL BYPASS
A. NO Trip NO. Trip
i
B. NO Trip Rx Trip )
C. Rx Trip NO Trip
D. Rx Trip Rx Trip
!
ANSWER: D. Source: Braidwood - 17 I
REFERENCES: OM Figure 2-8 Issue 1 Rev. 3
2LP-SQS-2.1 OBJECTIVE: 3
K/A 8: 3.09.015.000.K6.04 K/A IMPORTANCE: 3.1/3.2 i
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BVPS . Rev 2
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Question 2-97-85
Given the following:
- The Unit is operating at 100% power with all systems in their at-
power, NSA configurations.
- RCS Tan control channels are indicating as follows:
e 'A' loop - 575'F
- '.B ' loop - 577'F
e 'C' loop - 576*F
e 'B' loop Tcas instrument begins to flafLY fail LOW.
Which of the following describes the response of the Tau Control System to this
failure?
As _ Ja 'B' loop Tan drops, the selected Tan will swap from 'C' loop to
'A' loop, in *- to 'B',, then finally back to' 'C'
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B. 3' loop, chen to 'A', then finally back to 'C' I
C. 'B' loop, then to 'A' and remains there. f
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D. 'A loop and remains there.
)
le.SWER: C. Source: M-North Anna - 44 '
P'FERENCES: 20M-6.1.D pg. 18 Issue 4 Rev. 0 l
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lLP-SQS-6.5 OBJECTIVE: 5.a .
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K/A 8: 3.09.016.000.A3.01 K/A IMPORTANCE: 2.9/2.9 !
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Chiestion 2 97-86
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Given the following
- The Unit was operating at 1006 power with all systems in their at-
power. NSA configurations when a Rx trip and Safety Injection
occurred. !
- Plant status'is as follows: i
e All RCPs are stb:? PED.
- All Tmn. are 567'F and slowly RISING. i
e All Tema. are 510*r and slowly DROPPING. I
e The 5 highest CETs are 572'T and slowly RISING. j
a Pressurizer Level is 01.
- Pressurizer Pressure is 1180 psig. !
e All S/G Pressures are 1005 psig and STABLE. t
Based on the above indications, what is the status of natural circulation flow
and decay heat removal?
i
A. Single phase natural circulation flow is occurring and is
adequately removing decay heat. j
B. Two phase natural circulation flow is occurring and is adequately ;
removing decay heat. !
C. Single phase natural circulation flow is occurring but adequate l
decay heat removal is NOT occurring, i
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D. Natural circulation flow has stopped and adequate decay heat ,
removal is NOT occurring. {
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l' Get Objective /LP E
ANSWER: D. Source: New
REFERENCES: 20M-53B.5.GI-4 Issue IB Rev. 0
2LP-SOS-53.2 OBJECTIVE: 12
l K/A.f: 3.09.017.020.A3.01 K/A IMPORTANCE: 3.6/3.8
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Question 2-97-87
.
Unit 1 and Unit 2 are both at 100% power when a HIGH alarm is received on
[2RMC*RQ201] and [2RMC*RQ202), Control Room area radiation monitors. Which of
the following is the expected automatic system response to these HIGH alarms?
1. Unit 2 CR ACU Air Intake and Exhaust Dampers (2HVC* MOD 201A,B,C,D]
receive a CLOSE signal.
2. Unit 1 CR Air Intake and Exhaust Dampers {1VS-D-40-1A,1B,1C,1D]
receive a CLOSE signal.
3. The Control Room Emergency Bottled Air Pressurization System is
actuated.
4. The Unit 1 and Unit 2 Emergency Ventilation fans start after a 60
minute time delay.
>
A. 1, 2, and 3
B. 1, 2, and 4
C. 1, 3 ONLY.
D. 2, 3, and 4
1
ANSWER: A. Source: M-LOT - 0509
RE FERENCES: 20M-43.5.D.3 pg. 1 Issue 4 Rev. 0
2LP-SOS-43.1 OBJECTIVE: 5
K/A #: 3.09.072.000.K1.04 K/A IMPORTANCE: 3.3/3.5
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-Question 2-97-88
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An event has occurred that resulted in an Automatic Safety Injection. During
l the performance of.the EOPs it is noted that the entry conditions for FR-Z.3,
.. Response to High Containment Radiation Level, have been met. Containment
j. radiation levels are 80 mr/hr. What type of accident could have caused this
+
level of radiation to be in the containment?
i
1. A Loss of Rx Coolant Accident inside containment, with No fuel damage.
2. A Steam Line Break inside containment, with the allowable Technical-
i Specification S/G tube leakage.
1
i 3. Fuel cladding damage with NO indication of fuel melting.
!
! 4. Fuel cladding damage WITH indications of fuel melting.
1
A. 1 Only.
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B.
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1 and 2.
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C. I and 3.
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D. 1 and 4. I
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- ANSWER: B. Source: New
- -REFERENCES
- 2OM-53B.4.FR-Z.3 pg. 1 Issue 1B. Rev. 1
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! 2LP-SOS-53.A.1 OBJECTIVE: 5
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K/A 6: 3.09.073.000.A1.01 K/A IMPORTANCE: 3.2/3.5
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Question 2-97-89 -
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With the Unit operating at 100% power with all systems in their at-power, NSA
, annunciator (A2-5F] REACTOR COOLANT PUMP COOLING WATER TROUBLE if the 2B RCP
thermal barrier CCP isolation valve (2CCP*AOV107B] CLOSED on a high flow signal?
i
I A. Within 30 minutes,' reduce power to <30% then trip the affected RCP.
a
~ B. Verify adequate RCP seal injection flow and continue power operations.
C. Immediately Trip the Rx and then trip the affected RCP.
J
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D. Declare (2CCP*AOV107B] INOPERABLE per T.S. 3.6.3.1 " Containment Isolation
, Valves."
s
ANSWER: B. -
Source: Braidwood - 5
REFERENCES: 20M-6.4.AAG pg. 5 Issue 4 Rev. 1
2LP-SQS-6.3 OBJECTIVE: 12
K/A f: 3.10.000.000.K3.01 K/A IMPORTANCE: 3.4/3.5
.
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Question 2-97-90
Which of the following interlocks are provided to ensure a flowpath for service !
water return to t.he Circulating Water System? ,
i
A. ALL cooling tower pump suction valves [2CWS-MOV101A-D] CANNOT be
CLOSED simultaneously.
B. ALL cooling tower pump discharge valves (2CWS-MOV110A-D) CANNOT be
CLOSED simultaneously. ,
C. On the whole condenser, no more than ONE out of FOUR condenser ;
waterbox inlet valves (2CWS-MOV106A-D] can be CLOSED at any given I
time.
D. On one condenser half, no more than ONE out of TWO condenser waterbox
outlet valves (2CWS-MOV100A/B(C/D)] can be CLOSED at any given time. ,
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A"SWER: D. Source: New j
REFERENCES: 20M-31.2 P&Ll18 Issue 4 Rev. 0
2LP-SQS-31.1 OBJECTIVE: 5
K/A 1: 3.10.075.000.K4.01 K/A IMPORTANCE: 2.5/2.8
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Question 2-97-9I
Given the following:
- The Unit is in a refueling outage.
- Core on-load is almost complete.
- The refueling cred in the fuel handling building (FHB) is moving a
fuel element through the weir gate that separates the spent fuel pool
and the fuel transfer canal.
- The refueling SRO notices that cavity level is slowly DROPPING.
Based on the above information, the suspended fuel element should be_
A. lowered into the fuel transfer cart and left on the FHB side.
B. placed in the reactor vessel.
C. returned to the SFP.
D. placed in the RCCA change fixture.
ANSWER: C. Source: North Anna - 99
REFERENCES: 20M-20.4.AAD Issue 4 Rev. 0
2LP-SOS-20.1 OBJECTIVE: 9g
K/A f: 3.ll.000.036.EAl.04 K/A IMPORTANCE: 3.1/3.7
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Question 2-97-92 !
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l Due to component leakage, an inadvertent radiological liquid release has just l
- occurred. This release has resulted in the following radiation monitor j
readings.
1
- [2SWS-RQ-101] Component Cooling HX Rad Monitor at 5.73 E*' uCi/cc. l
- '[28WS-RQ-102] Component Cooling Service Water Rad Monitor at 6.03 E"' !
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uCi/cc. I
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I hours -
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A. NO Emergency Plan Classification will be necessary.
[
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B. an Unususi Event should be declared IMMEDIATELY. l
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l C. an Unosual Event should be declared if an E-Plan Assessment CANNOT f
I be performed within'60 minutes. *
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D. an Alert should be declared if an E-Plan Assessment CANNOT be i l
l perfonned within 15 minutes.
}
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Provide the following references: EPP classification tabs.
[
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ANSWER: C. Source: New
REFERENCES: EPP/I-lb Att. 1 Table 7-1, Rev. 6
Fif. 7-A. I
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2LP-SOS-53A.1 OBJECTIVE: 6 I
K/A 4: 3.ll.000.059.G02 K/A IMPORTANCE: 2.6/3.9 !
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Question 2-97-93
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i During sampling of the PER vapor space, the outside containment foolation valve ~
)
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[2SSR*AOV112A2) developed a severe packing leak. A HIGH radiation alarm on the
Leak Collection Vent Radiation Monitor [2RMR*RQI301) has resulted.
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This alarm will result in
A. NO automatic actions. The operators must manually CLOSE the Normal
, Leak Collection Dampers [2HVS* MOD 201AEB) to terminate the release.
J
B. Normal Leak Collection Dampers [2HVS*WOD201A&B) CLOSING and Filtered
,
Leak Collection Dampers [2HVS* MOD 202A&B) OPENING.
C. Containment Purge Diverting Dampers [2HVR* MOD 21622] swapping to the '
j- Filtered Release Path.
I D. Normal Exhaust Fans [2HVS*FN263A&B] STOPPING and the associated fan
. discharge dampers CLOSING.
1
ANSWER: B. Source: New
REFERENCES: 20M-43.5.B.3 Issue 4 Rev. 0
2LP-SOS-43.1 OBJECTIVS: 5
K/A 8: 3.11.000.060.EA2.05 K/A IMPORTANCE: 3.7/4.2
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Question 2 97-94 l
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What indication is available to alert personnel that a CO2 discharge is imminent '
inside of a protected zone?
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A. A red revolving light inside the zone.
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B. A pre-discharge horn sounding inside the zone. I
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C. A wintergreen odorizer floods the zone prior to the discharge. I
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D. The announcement from Security over the Page Party System. .!i
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ANSWER: B. Source: .SQS - 1165 '!
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REFEPINCES. 20M.33.1.D Issue 4 Rev. 2 '
2LP-SOS-33.1 OBJECTIVE: 4.d
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K/A ft 3.11.000.067.G05 K/A IMPORTANCE: 3.4/3.8 l
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Page 94
- BVPS . Rev. 2
Question 2-97-95
Technical Specification 3.4.8.1 - RCS Specific Activity, action s*.atement
requires the RCS to be cooled down to <500*F if the specific activity limits of
the reactor coolant are exceeded. What is the bases for reducing Tave to <500*F7
A. .To prevent the release of activity should a S/G tube rupture, since
Tsar of the RCS would be below the lift pressure of the S/G atmospheric
steam relief valves.
B. To ensure additional iodine spiking will NOT occur due to the reduced
thermal energy in the fuel rod gas voltune.
C. To ensure the projected site boundary thyroid dose will be maintained
less than the 10CFR Part 20 limits following a postulated SGTR.
D. To prevent having to make a Protective Action Recommendation (PAR) I
should a SGTR concurrent with a Faulted S/G Outside Containment occur.
ANSWER: A. Source: New
REFERENCES: TS 3.4.0 Bases pg. 4-6 Amendment No. - Original
2LP-SQS-TS OBJECTIVE: 4
K/A 9: 3.11.000.076.G04 K/A IMPORTANCE: 2.1/3.7
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BVPS Rev.2
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Question 2-97-96
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j Given the following:
- The Unit is in mode 6 for a refueling outage.
- Off-load of fuel is 554 complete and ongoing.
<
e Containment Purge and Exhaust is in service. ;
- IEC has just reported that the current HI setpoints for the !
Containment Purge Radiation Monitor (2HVR*RQIl04A] was incorrectly set ;
, two decades HIGH. '
i
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What action should be directed based on this information? .
I
A. Suspend core off-load until containment atmosphere grab samples can be i
obtained a:P1 double verified <MPC.
1
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- B. Suspend core off-load until the containment purge and exhaust valves
, are declared OPERABi2. 3
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, C. Continue core off-load and direct HP to perform continuous air i
monitoring of the containment.
D. Continue core off-load and verify purge exhaust is directed through
the Main Filter Bank.
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- . ANSWER
- B. Source: North Anna - 46 I
REFERENCES: TS 3.9.9 Amendment No. - Original
.
- - 2LP-SOS-TS OBJECTIVE
- 1
.
j K/A #: 3.ll.029.000.G11 K/A IMPORTANCE: 2.8/3.5
.
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DVPS . Rev. 2
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t - Question 2-97-97 !
.
.
. The suct. ton piping of the speat fuel pool cooling pumps (2FNC*P21A&B) has
ruptured and CANNOT be isolated. Which of the following, by design, is the
LOWEST spent fuel pool level that could result?
l
A.110 feet above the spent fuel assemblies. ,
B. 23 feet above the spent fuel assemblies.
C. A' level equal to the top of the spent fuel assemblies.
.D. A level equal to the top of the cask area weir.
ANSWER: A. Source: R-SQS - 1103
i .
! REFERENCES: 20M-20.1.B pg. 3 Issue 4 Rev. 1
2LP-SOS-20.1 OBJECTIVE: 1 i
K/A 8: 3.ll.033.000.K4.01 K/A IMPORTANCE: 2.9/3.2 [
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Question 2-97 98
.
Which of the following manipulator crane features helps to prevent lifting a
fuel assembly with excessive force?
A. Dillon load cell circuit.
B. Gripper interlock circuit.
C. Bridge - trolley interlock.
D. Slack cable limiting circuit.
ANSWER: A. Source: R-LOT - 0206
REFERENCES: Refueling Manual Issue Rev.
2LP-SQS-6.12 CBJECTIVE: 5.c j
K/A 8: 3.11.034.000.K6.01 K/A IMPORTANCE: 2.1/3.0
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DVPS . Rev. 2
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Question 2 97 99 !
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During a Liquid Waste Discharge, flow control valve (2SGC-HCV200] is in manual
and controlling flow at 30 gym. A High Radiation Alarm'is received on the f
Liquid Waste Process Effluer.t Radiation Monitor (2SGC-R0100). Which of the i
following explains the effece that this Radiation Alarm will have on the Liquid ;
Waste Discharge't j
!
A. The discharge will be terminated immediately due to the automatic !
CLOSURE of (2SGC-HCV100). !
,
B. The discharge will continue, (2SGC-HCV100) will NOT automatically !
CLOSE while in manual. )
C. The discharge will be terminated by diverting the discharge flowpath
from Unit I cooling tower to the Steam Generator Blowdown Hold Tanks.
D. The discharge will continue for 30 seconds, if the High Radiation ;
Alarm is still present, (2SGC-HCV100) will automatically CLOSE.
ANSWER: B. Source: M-SOS - 0608
REFERENCES: 20M-25.1.D Pg. 10 Issue 4 Rev. O i
2LP-SOS-17.1 OBJECTIVE: 5
-K/A 8: 3.ll.068.000.A3.02 K/A IMPORTANCE: 3.6/3.6 f
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BVPS . Rev. 2
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(hiestion 2-97-100 l
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Which of the following describes the gaseous waste disposal system response if
the gaseous waste surge tank [2GWS-TK21] rupture disc [2GWS-PSE126] were to
<
rupture?
A. The Auxiliary Building Supply and Exhaust Fans will automatically
STOP.
B. A flammable mixture of radioactive gases will form in the Unit 2
Auxiliary Building.
,
C. The release of gas to the environment will be terminated by trip valve 1
[2GWS-A0V105).
.
D. The surge t'ank will relieve via relief valve [2GWS-RV101] to the Unit i
1 waste gas relief header.
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ANSWER: D. Source: R-LOT - 0201
REFERENCES: 20M-19.1.B pg. 2 Issue 4 Rev. 0
2LP-SOS-19.1 OBJECTIVE: 2
K/A 9: 3.11.071.000.K4.01 K/A IMPORTANCE: 2.6/3.0
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Attachment 2
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BV-2 SRO WRITTEN EXAM W/ ANSWER KEY j
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___ _. . __ . . .. . _ _ _ . _ _ _ _ . - - . - _ _ __ _ _ _ _ = __ _ _ ._.__._ __. ._ _ _ _ _ _ . _
RTL #A5.620.H DUQUESNE LIGHT COMPANY Volume 3
Nuclear Power Division Pmcedure 5-5
Training Administration Manual ' Figure 5-5.1
Revision 10
Page1of1
WRITTEN EXAMINATION COVER SHEET
,' PROGRAM: Initial Licensed Operator Training -
,
) CLASS NUMBER: 1-LOT-2
SUBJECT: Senior Reactor Operator, March 1997 - NRC Initial Licensed Operator Exam.
4
4
l
By this signature, I state that all
of the work done on this examination
I is my own. I have neither given nor
received aid.
l
SIGNATURE DATE March 17,1997
NAME ANSWER KEY DLC EMP #
(Please Print)
COMPANY
(if other than DLC)
POSSIBLE POINTS 100 SCORE
Instructor
Initials
1
.
TRAINING DJRECTOR/ SUPERVISOR
PREPARED BY David C. Gibh APPROVAL i
SIGNATURE
~
/3 M '
3//3/97
/ '
^
{/ ' Date
'
( 3 /3-97
._ _ __ . . _ _ . _ _ _ _ __ _. . _
!
l( ES-402 Policies and Guidelines
for Taking NRC Written Examinations
Attachment 2
1. Cheating on the examination will result in a denial of your application
- and could result in more severe penalties.
j 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
j 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
i question number.
5. There is a time limit of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 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.
,
k Examiner Standards 5 of 6 Rev. 7, January 1993
Question Number 1
i
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Select the statement that describes the safety equipment REQUIRED to be worn
when handling sodium hydroxide (NaOH).
l
A. Safety glasses and appropriate gloves ONLY.
B. Safety glasses, appropriate gloves and a paper surgical mask, i
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C. Goggles /faceshield and impervious clothing ONLY. '
D. Goggles /faceshield, appropriate gloves and impervious clothing.
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POINTS: 1.00
ANSWER: D.
REFERENCES : AOP 1/2 53.C.4A.75.7, Att. 1, page 12 - Issue 1A, Rev. 6.
!
ILP-SQS-53C.1 OBJECTIVE: 4 NUMBER: 1-97-093, R-0117
JTA #:
K/A #: 194001.K1.10 K/A IMPORTANCE: 3.0/3.3
Final Revision, Rev. 3 A
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QuestionNumber 2 l
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An MOV designated as "VDM", (valve drifts manually), has been closed manually
for a clearance.
Which of the following actions is required to prevent this valve from drifting
due to system pressure effects?
i
A. Ensure that the MOV manual operating lever remains in the fully )
" ENGAGED" position following manual closure. 1
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B. Manually return the MOV manual operating lever to the " DISENGAGED" j
position following manual closure.
)
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C. Restore power to the MOV following manual closure and leave the ;
control in the * NEUTRAL" position. j
D. Restore power to the MOV following manual closure and take the control
switch to the " CLOSED" position.
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POINTS: 1.00
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ANSWER: D,
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REFERENCES: HPDAP 3.4, Rev. 6, Page 12, item o.
ILP-SQS-48.1 OBJECTIVE: 25 NUMBER: 1-9*1-094, M-0121
JTA #: '
K/A 0: 194001.Kl.01 K/A IMPORTANCE: 3. 6 /3. ~1
'
Final Revision, Rev,3A
_ . - - _
QuestionNumber 3
i
Given the following: I
- RCS pressure is 225 psig.
- RCS temperature is at 210*F.
Chemistry has just called in the most recent RCS sample results.
Which of the following would exceed the Technical Specification limit for the
s RCS transient chemistry specifications?
A. Fluoride = 1.0 ppm.
, B. Chloride = 1.6 ppm.
C. Dose Equivalent Iodine - 131 = 0.5 uci/gm.
D. Dissolved Oxygen = 1.1 ppm.
,
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POINTS: 1.00
ANSWER: B.
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Provide the following references: Unit 1 Technical Specification 3.4.7,
Chemistry, and 3.4.8 Specific Activity,
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REFERENCES: Unit 1 Technical Specification 3.4.7, Table 3.4-1, Original.
ILP-SOS-CHEM,-19 OBJECTIVE: 4 NUMBER: 1-97-095, M-0125
JTA 8:
K/A ft 194001.A1.14 K/A IMPORTANCE: 2.5/2.9
final Revision. Rev. 3A
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QusstionNumber 4 l
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An operator is required to be continuously stationed at a valve in a confined l
- area for 30 minutes. Radioactive material on/in this valve is exposing the !
j-
operator to 200 mR/hr. Three feet behind the operator is another valve that
emits 100 mR/hr at 25 cm. Which of the following is applicable to this -
situation?
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- A. - The valve behind the operator must be labeled a " Hot Spot" and the *
area posted as a Radiation Area.
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B._ The operator will exceed their 10 CFR 20 ' dose limits. !
C. The area must be posted as an High Radiation Area and the operator
1 should have an integrating dose rate meter. l
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D. An HP technician should be present to monitor the radiation in the
4
room with a portable neutron meter while the operator is stationed at
j the valve.
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POINTS: 1.00 ;
ANSWER: C.
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REFERENCES: 10CFR20.003 and Unit 1 Tech Specs 6.12 Amendment No. 188.
1/2LP-RC-02 Rev. 17 OBJECTIVE: 4-9 NUMBER: 1-97-096, M-OLES
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K/A #: 194001.Kl.03 K/A IMPORTANCE: 2.8/3.4
Final Revision, Rev.3 A
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Question Number 5
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Which of the following are required for a confined space entry?
1. A minimum of two qualified individuals shall enter the confined space,
one of which will act as only a safety man.
2. A method of communication shall be established to maintain contact with
personnel within the confined space.
.
3. An SCBA for emergency use shall be located near the entrance of a
confined space when certain tasks are being performed within.
4. For conditions where an SCBA used by rescuers may be impracticable, the
ventilation flow can be increased as an additional precaution.
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A. 1, 2, and 3.
B. 1, 2, and 4.
C. 2, 3, and 4.
D. 2 and 3 ONLY.
l POINTS: 1.00
ANSWER: C.
REFERENCES: NGAM 3.7
1/2LP-GM-6040 OBJECTIVE: 7 NUMBER: 1-97-097, M-0173
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JTA i:
K/A 4: 194001.K1.13 K/A IMPORTANCE: 3.3/3.6
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i I" mal Revision. Rev. 3A
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QuestionNumber 6
Listed below in random order are steps regarding the preparation of the
Valve / Switching Procedure Form.
1. The NCO checking the Form shall check the clearance point (s) (using
control room prints) to ensure the procedure is proper.
2. The ANSS or NSS approving the Form shall ensure the equipment is being
cleared properly and that required equipment is not made inoperable.
3. The operator performing the switching wi11 present the Form to NCO.
4. The operator completing the Form shall fill in the pertinent
information (i.e., Clearance 9, Clearance Point, Tag Type, etc.).
Which of the following groups is in the proper order?
A. 4, 3, 1, 2
B. 2, 4, 3, 1
C. 4, 2, 1, 3,
D. 1, 2, 3, 4,
POINTS: 1.00
ANSWER: A.
REFERENCES: SAP 42, pages 18 and 19, Rev. 6.
ILP-SQS-48.1 OBJECTIVE: 30 NUMBER: 1-97-098, M-0181
JTA #:
K/A ft 194001.K1.02 K/A IMPORTANCE: 3.7/4.1
Final Revision. Rev. 3A
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Question Number 7
During the performance of an Operating Surveillance Test (OST) that takes
several days to finish, which of the following are required?
A. Complete the OST within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to ensure validity of all data taken.
B. Restart the OST if not completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
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C. Re-perform all steps of the OST that are more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> old.
D. Re-verify the working copy of the OST against the Controlled copy
every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
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POINTS: 1.00
ANSWER: D.
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REFERENCES: 1/20M 48.2.C, issue 3, Rev. 17 page CB of 19 item 14c
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1/2LP-SQS-48.1 OBJECTIVE: 10 NUMBER: 1-97-110, LOT - 0726
JTA #:
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K/A #: 194001.A1.01 K/A IMPORTANCE: 3.1/3.2
Final Revision, Rev. 3 A
Questi1n Number 8
Who must approve an on the spot change (OMCN) to an operating procedure?
A. The respective, department st;pervisor of the individual requesting the
change ONLY.
B. TWO members of the plant management staff, ONE of whom must hold an
SRO license for the affected Unit.
C. ONE member of the plant management staff who holds an SRO license for
the affected Unit.
D. The respective department supervisor and BOTH of the on-shift licensed
SROs for the affected Unit.
POINTS: 1.00
ANSWER: B.
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REFERENCES: 1/20M48.2.B - Issue 3, Rev. 13.
ILP-SOS-48.1 OBJECTIVE: 8 NUMBER: 1-97-100, M-0385
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JTA i:
K/A f: 194001.A1.01 K/A IMPORTANCE: 3.3/3.4
Final Revision. Rev. 3A
Qusstion Number 9
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hhen performing a station startup IAW OM Chapter 50 " STATION STARTUP," steps j
marked by a filled diamond sign indicate that the step :
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A. may be skipped at the discretion of the NSS.
B. may be omicted by the NSS provided the UOM initials the omitted step.
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C. cannot be omitted but may be started out of sequence.
D. cannot be omitted and must be performed in the specified sequence.
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REFERENCES: 10M48.2.C - Issue 3, Rev. 13.
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ILP-SOS-48.1 OBJECTIVE: 10 NUMBER: 1-97-101, M-0386
JTA #:
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K/A #: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9
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Final Revision, Rev.3A
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Question Number 10
which of the following is the reason why a nitrogen blanket is maintained on the
Pressurizer Relief Tank (PRT) ?
A. To provide an additjsnal volume of gas to ensure that the PRT Rupture
Disc relieves at th( design pressure.
B. To provide a driving force for the sampling system.
C. To reduce the potential for an explosive mixture of hydrogen and
D. To ensure the water volume in the PRT remains subcooled when a PZR
PORV or Safety Valve lifts.
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POINTS: 1.00
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ANSWEF: C. '
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REFERENCb5: 10M1.5.1.C - Issue 4, Rev. 1.
ILP-SOS-6.4 OBJECTIVE: 7 NUMBER: 1-97-102, M-0661
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, JTA #:
K/A i: 194001.K1.15 K/A IMPORTANCE: 3.4/3.8
Final Revision, Rev. 3A
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QuestionNumber 11
Which of the following conditions MUST be met to allow a non-licensed person to
manipulate the control rods?
1. Under the direct supervision of a licensed operator or SRO.
2. Enrolled in a training program to acquire a license.
3. In the presence of a licensed operator or SRO.
4. Member of the Training Department enrolled in an SRO Certification
program.
A. 1,2, and 3.
B. 1,2, and 4.
C. 1,3, and 4.
D. 2,3, and 4.
POINTS: 1.00
ANSWER: A.
REFERENCES: 1/20M 48.1.B issue 3, Rev. 17 page B7 of 9.
1/2LP-SOS-48.1 OBJECTIVE: 39 NUMBER: 1-97-111, LOT - 0728
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JTA #:
K/A #: 194001.A1.09 K/A IMPORTANCE: 2.7/3.9
Final Revision. Rev. 3A
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QuestionNumber 12
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Which of the following criterion would require the Technical Support Center
(TSC) to be activated?
A. A pressurizer PORV fails to close following a valid open signal with
NO Safety Injection actuation required and the associated PORV Block
Valve operable.
B. The Rx fails to trip when an automatic trip signal is generated but
trips when activated manually at the benchboard.
C. A simultaneous loss of ALL annunciators, sequence of events recorders,
and SPDS for >l5 minutes in Mode 5.
D. A report by plant personnel of a chlorine gas release'within the site
perimeter that renders the chlorine building inaccessible.
POINTS: 1.00
ANSWER: B.
Provide the following references: EPP/IP Tab 2.2.
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REFERENCES: EPP Tab 2.2.
1/2LP-EPP-57.81 OBJECTIVE: 1& 11 NUMBER: 1-97-104, M-6280
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JTA #:
K/A #: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4
final Revision, Rev. 3A
QuestionNumber 13
Which of the following approved portable fire fighting equipment should be used
to combat a flammable liquid fire?
1. Dry Chemical.
2. Water.
3. COz .
4. Foam.
A. 1, 2 or 3.
B. 1, 3 or 4.
C. 1, 2 or 4.
D. 2, 3 or 4.
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POINTS: 1.00
ANSWER: B.
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REFERENCES: 1/2.56A.4.H - Issue 3, Rev. 1, and LP 9339. I
- ILP-SOS-9339 OBJECTIVE
- 7 NUMBER: 1-97-105, New
JTA 4:
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K/A #: 194001.Kl.16 K/A IMPORTANCE: 3.5/4.2
Final Revision, Rev. 3 A i
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.QuestionNumber 14 j
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The plant is operating at 50% load with minimum shift compliment. At the
beginning of.the shift, the Reactor Operator is seriously injured and is sent to 1
the hospital. Select the required actions with less than minimum shift
compliment.
A. If a replacement cannot be contacted within 15 minutes, a Unit
shutdown must be commenced.
- - -B. Operations with one less than the minimum compliment may continue
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until shift change, provided that immediate action is taken to bring
the compliment up to the minimum.
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C. Operations with one less than the minimum compliment may continue for -}
up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, provided that immediate action is taken to bring
compliment up to the minimum.
D. Operations with one less than the minimum compliment may continue
l, indefinitely, provided that within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. action is taken to bring
l compliment up to the minimum.
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' POINTS: 1.00
ANSWER: C.
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REFERENCES: OM-1/2.48.4.B Inst. A Issue 3, Rev. 17
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ILF-SQS-48.1 OBJECTIVE: 5 NUMBER: l-97-ll2, LOT - 0808
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JTA W:
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} K/A i: 194001.A1.03 K/A IMPORTANCE: 2.5/3.4 i
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QuestionNumber 15
Technical Specification Surveillance 4.8.1.1.2 requires in part, that each
Emergency Diesel Generator (EDG) be started from ambient conditions at least
once per 31 days, on a STAGGERED TEST BASIS. Which of the following would be in
compliance with this STAGGERED TEST BASIS requirement?
Start il EDG on the (1) day of each 31 day period and start #2 EDG on
the (2) day of the (3) 31 day period.
(1) (2) (3)
A. 1** 8'" same
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B. 1** 15*" same
C. 1** l'* next
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D. 8'" 31 same
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POINTS: 1.00
ANSWER: B.
REFERENCES: Technical Specification Definitions Amendment No. 192.
ILP-SQS-TS OBJECTIVE: 3 NUMBER: 1-97-107, New
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JTA i:
K/A f: 194001.Kl.17 K/A IMPORTANCE: 2.1/2.5
Final Revision, Rev. 3A
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Question Number 16
Given the following:
- Reactor power has just been reduced from 100% to 60%.
- Control Bank D rods (CBD) are at 122 steps.
- T. . - Tr.: error is 0*F.
- Boron concentration is 375 ppm.
e Reactor Engineering reports that over the next hour, Xenon will add a
negative 115 pcm.
- Core burn-up is 11,500 MWD /MTU.
Over the next hour, how much boric acid or primary grade water must be added to
the RCS in order to withdraw CBD to the fully withdrawn position AND keep Rx
power and T. constant?
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A. 299.0 gallons of Boric Acid.
B. 361.5 gallons of Boric Acid.
C. 5,866 gallons of Primary Water.
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D. 7,190 gallons of Primary Water. l
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POINTS: 1.00
ANSWER: B.
- CB10B - Rod motion will add +550 pcm.
e Xenon will add -115 pcm.
e Boron must add -435pem.
e CB28 - Boron worth is -0.4pcm/ ppm.
Baron conc. must be increased by 52 ppm for a total of 427 ppm.
e T.v. = 565*F at 60% power.
- WAG table 565*F, iteration method:
- From 375 ppm to 425 ppm = 347.57 gal of BA.
e From 425 ppm to 427 ppm = 13.96 gal of BA.
- Total of 361.5 gallons of boric acid must be added.
Provide the following references: Unit 1 Curve Book, Cycle 12, Issue 12 Rev. 1,
and the WAG Tables.
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l REFERENCES : Unit 1 Curve book, CB-10B, CB-28, WAG Tables 560*F.
ILP-SQS-RT-6 OBJECTIVE: 15 NUMBER: 1-97-100, M-LRT 2.3.4
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JTA 6:
K/A #: 194001.A1.08 K/A IMPORTANCE: 2.6/3.1
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Question Numtxt 17 j
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Following a Safety Injection (SI) signal, which of the following describes the
control logic necessary to OPEN the feedwater regulating BYPASS valves-[FCV-lFW-
479, 489, & 499)?
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A. ONLY the SI signal needs to be Reset.
B. ONLY the Feedwater Interlock signal needs to be Reset.
C. DOTH the SI signal AND the Feedwater Interlock signals need to be
Reset.
D. The feedwater regulating BYPASS valves CANNOT be OPENED with a
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standing SI signal even if it is Reset.
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POINTS: 1.00
ANSWER: D.
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Provide the following references: UFSAR Figure 7.2-1, Instrumentation and
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Control System Logic Diagram, Sheet 1 and 13.
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RE FERENCES : UFSAR Figure 7.2-1, Sheet 13, Rev. 10.
ILP-SOS-24.1 OBJECTIVE: 4 NUMBER: 1-97-109, New
JTA #:
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K/A #: 194001.A1.07 K/A IMPORTANCE: 2.5/3.2
Final Revision, Rev. 3A
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Chus6on Number 18 I
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The plant is at 75% power with the rod bank selector switch in manual and all !
other systems in automatic. What would be the effect on the Rod Insertion I
Limits (RIL) and the Shutdown Margin (SDM) if the main generator electrical i
output was raised by 34, and the only other operator action was to restore T.
to program using the Boration/ Dilution controls?. {
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RIL will (1) , and SDM will (2) .
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(1) (2) I
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A. raise, lower. 'l'
B. Iower, raise. '
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C. Iower, lower. l
D. raise, raise. !
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POINTS: 1 00 }
ANSWER: A.
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REFERENCES: Tech. Spec. Definitions
ILP-SOS-1.4 OBJECTIVE: 6 NUMBER: 1-97-001, M-0395 '
JTA #1 0010110104
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K/A #: 3.01.001.000.K5.08 (001K5.04) K/A IMPORTANCE: 3.9/4.4
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Final Revision. Rev.3A
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Question Number 19
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Given the following:
- Control Rod D12, a control bank "C" Group 1 rod, has fallen into the
core due to an equipment failure.
The equipment failure has been corrected and all retests are completed I
satisfactory.
The dropped rod recovery is in progress per AOP-1.1.5, " Dropped RCCA."
e All applicable switches are in their correct position for the rod l
recovery.
- AOP step 1.1.5.17.a directs the operator to " Anticipate rod control !
system urgent failure alarm." '
The ROD CONTROL SYSTEM URGENT ALARM is caused by a_
A. Logic cabinet failure and ALL rod motion will be inhibited. ;
B. Logic cabinet failure and ONLY those rods aligned to power cabinet 1AC
will move.
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C. Power cabinet 1AC failure and ONLY those rods aligned to power cabinet I
2AC will move.
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D. Power cabinet 2AC f ailure and ONLY those rods aligned to power cabinet
1AC will move,
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POINTS: 1.00
ANSWER: D.
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REFERENCES : 10M-1.1.D - Issue 4 Rev. 1, AOP 1.1.5 - Issue 3A Rev. 5
ILP-SOS-1.3 , OBJECTIVE: 10 11 UMBER: 1-97-002, M-0055
JTA 8:
K/A 9: 3.01.001.050.A2.01 (001A2.14) K/A IMPORTANCE: 3.7/3.9
final Revision, Rev. 3 A
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Question Number 20 !
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Given the following:
- The plant has been at 1004 power for 20 days.
- All systems are.in their at-power, NSA configurations.
- RCS T. is stable at 577'F.
e There have been NO sump pump runs for the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- Pressurizer level and pressure are stable at their program values.
- Charging flow is 89 gpm.
I e Letdown flow is 104 gpm.
- RCP seal injection flow is 24 gpm.
e Total RCP seal return flow 9 gpm.
- There have been three auto make-ups to.the VCT in the past 30 minutes.
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A. Letdown flow control valves (LCV-1CH-112 & 115A), are partially
diverting letdown flow to the CRT's.
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I B. Letdown line relief valve [RV-1CH-203), has lifted and failed to
- resent causing a portion of letdown flow to be diverted to the PRT.
i C. RCP seal injection control valve [HCV-1CH-186], has failed open
! causing excessive seal injection flow and VCT depletion.
D. Letdown isolation valve [LCV-1CH-460A), has developed a severe packing
leak causing a small loss of reactor coolant condition.
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POINTS: 1.00
ANSWER: A.
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REFERENCES: lOM-?.1.D - Issue 4 Rev. 2, IOM Fig. No. 7-1 - Issue 9 Rev. 9. l
ILP-SQS-7.1 , OBJECTIVE: 6 NUMBER: 1-97-003, New
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K/A #: 3.01.004.020.K1.02 K/A IMPORTANCE: 2.7/2.8
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final Revision.Rev.3A
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Question Number 21
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( Given the following ~!
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1 * The plant is at 100% power. 3
l e All systems are in their at-power, NSA configurations.
l e Annunciator [A4-51] "RCS h . HIGH", is lit.
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l Determine the mode of makeup control required and the expected corresponding i
i valve lineup. ,
A. Borate; [
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Boric acid flow control valve (FCV-1CH-113A] - OPEN l'
l Primary water flow control valve (FCV-1CH-114A] - CLOSED
Makeup stop valve to the charging pump suction [FCV-1CH-113B] - CLOSED !
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Makeup stop valve to the VCT (FCV-1CH-114B] - OPEN !
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B. Borate; I
Boric acid flow control valve (FCV-1CH-113A] - OPEN
Primary water flow control valve (FCV-1CH-114A] - CLOSED
Makeup stop valve to the charging pump suction [FCV-1CH-113B] - OPEN I
l Makeup stop valve to the VCT [FCV-1CH-1148] - CLOSED
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C. Dilute;
Boric acid flow control valve [FCV-1CH-113A] - OPEN
Primary water flow control valve [FCV-1CH-114A] - OPEN i
Makeup stop valve to the charging pump suction (FCV-1CH-113B] - OPEN.
Makeup stop valve to the VCT [FCV-1CH-114B] - CLOSED
D. Dilute;
Boric acid flow control valve [FCV-1CH-113Al - CLOSED
Primary water flow control valve [FCV-1CH-114A] --OPEN
Makeup stop valve to the charging pump suction [FCV-1CH-113B] - CLOSED
- Makeup stop valve'to the VCT [FCV-1CH-114B] - OPEN
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POINTS: 1.00 ,
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' ANSWER: B.
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REFERENCES: '10M-7.1.D - Issue 4, Rev. 2.
! ILP-SOS-7.1 OBJECTIVE: 5 NUMBER: 1-97-004, New
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Final Revision, Rev.3A
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QuestionNumber 22
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Given the following:
e A plant startep is in progress with Rx power at 17%.
Steam dumps are in the Main Stema Pressure Control Mode maintaining
Main Steam Pressure at 1005 psig.
e The Main Generator Output Breakers have just been closed.
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e Rod control is in manual. l
e CBD is at 100 steps with normal rod sequencing.
- The Rx core is Xenon free.
A spurious equipment failure caused an intermittent continuous rod withdrawal of
CBD that stopped with CBD at 110 steps. Assuming that no Rx trip occurs and T.v.
remains on program, detenmine the level at which Rx power will stabilize.
A. 14%
B. 17%
C. 18%
D. 20%
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ANSWER: D.
CB-13: 705 ppm = 825 MWD /MTU. - Use MOL curves.
CB-21: Power Defect is 20 pcm/% power.
CB-11E: 100 steps on CBD = 839 pcm.
110 steps on CBD = 780 pcm.
Delta - 59 pcm.
59 pcm - 20 pcm/% power = 3% change. 17% + 3% = 20%.
Provide the following references: BVPS Unit 1, Curve Book, Cycle 12, Issue 12
Rev. 1, .
l REFERENCES: Unit 1 Curve Book, Cycle 12, Issue 12 Rev. 1, curves CB-11E,13,21
! and 24B.
ILP-SOS-LP-P)T-6 OBJECTIVE: 9 HUMBER: 1-97-005, New
JTA #: ,
K/A #: 3.01.000.001.EKl.03 K/A IMPORTANCE: 3.9/4.0
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Question Number 23
Due to an equipment malfunction in the automatic rod control system, a
continuous rod withdrawal was initiated from an initial power level of 726. -
'After 5 seconds of rod withdrawal, the reactor operator selected " MANUAL" on the !
Rod Control Selector Switch which terminated the rod movement. All other
systems are aligned in their at-power, NSA lineups. Assume that no reactor trip
occurs, and NO other operator action is taken. Which of the following
-parameters will return to essentially the same value that it was before the
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A. RCS Tave.
. B. Pressurizer level. l
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D. Reactor power.
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! ANSWER: D.
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REFERENCES: LP-RT-6 Rev. 2
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- ILP-SQS-RT-6 OBJECTIVE
- 15 NUMBER: 1-97-006, M-0497.
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! K/A #: 3.01.000.001.EA2.04 (001EA2.04) K/A IMPORTANCE: 4.2/4.3 !
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QuestionNumber 24
Given the following:
e A reactor startup is in progress with power at IE-8 amps in the IR. i
e All systems are aligned in their normal lineups for the current power
level.
- Main feed pump [FW-P-1B] is in service with SGWL control in manual
using the bypass feedwater regulating valves (FCV-FW-479,489, and
499).
- An electrical fault occurs that causes a sustained loss of the 1A 4KV
bus.
- The 1AE bus is re-energized from the No. 1 Emergency Diesel Generator.
Determine the expected configuration of the Reactor and the Rod Drive MG Sets
following the loss of the 1A 4KV bus.
A. The Reactor will trip on low flow due to the loss of the 1A RCP.
Neither rod drive MG sets are affected due to the automatic bus
transfer of the 1A 480V bus feed to the IC 4KV bus via the 480V. bus
tie breaker.
B. The Reactor will not trip. ,
Rod drive MG set Rod-MG-1 will be lost due to the loss of the 1A 4KV
bus but Rod-MG-2 has sufficient capacity to maintain power to all
C. The Reactor will trip due to loss of Rod Drive power.
Rod drive MG set Rod-MG-1 will be lost due to the loss of the 1A 4KV
bus and Rod-MG-2 does not have sufficient capacity to maintain power
to all control rods.
D. The Reactor will not trip.
Neither rod drive MG sets are affected due to the automatic bus
transfer of the 1A 480V bus feed to the 1C 4KV bus via the 480V bus
tie breaker.
POINTS: 1.00
ANSWER: D.
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REFERENCES : 10M-37.1.D - Issue 4 Rev. O, 10M-1.5.B.4 - Issue 2 Rev. 10.
1LP-SQS-37.1' OBJECTIVE: 2 NUMBER: 1-97-007, New I
JTA #:
K/A #: 3.01.000.003.EK2.05 K/A IMPORTANCE: 2.5/2.8
Final Revision, Rev. 3A
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Question Number 25 ,
Why is Reactor Coolant System (RCS) pressure NOT used in the diagnostic steps in
E-0 Reactor Trip or Safety Injection?
Because RCS pressure will.
A. NOT be affected by a loss of secondary coolant accident.
B. ONLY be affected by a loss of primary coolant accident.
C. be affected by ONLY a loss of primary coolant accident and a loss of
secondary coolant accident.
D. be affected by the loss of primary and secondary coolant accidents, and
steam generator tube rupture accidents.
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POINTS: 1.00 !
ANSWER: D. ,
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REFERENCES: NOMCD Lesson Plan
ILP-SOS-53.3 OBJECTIVE: 3 NUMBER: 1-97-009, R-0360 '
JTA 8: 3010010601
)
K/A 9: 3.01.000.007.EK3.01 K/A IMPORTANCE: 4.0/4.3
Final Revision, Rev. 3A
QuestionNumber 26
An equipment malfunction has caused a demand for a Rx trip. The automatic Rx
trip signal did not open the Rx trip breakers (RTB's) . Which of the following
describes the locations from which the RTB's and Rod Drive MG Set Supply
breakers can be opened?
1. Locally on the front of the respective breaker.
2. Lift Coil Disconnect Switch Panel.
3. Rod drive MG Control Panel.
Rod Drive MG Set
RTB's Supply Breakers
A. 1 ONLY. 1 and 3 ONLY.
B. 1 and 2 ONLY. I and 3 ONLY.
C. 1 ONLY. 3 ONLY.
.
D. 1 and 2 ONLY. 3 ONLY.
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POINTS: 1.00 l
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ANSWER: A.
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REFERENCES: 10M-1.3.C - Issue 4, Rev. 1.
ILP-SOS-53.3 OBJECTIVE: 6 NUMBER: 1-97-011, New
JTA #2
K/A i: 3.01.000.029.EA1.12 K/A IMPORTANCE: 4.1/4.0
i
Final Revision, Rev. 3A
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Question Number 27
With the Unit at 100% power with all systems in their at-power, HSA
configurations; * MALFUNCTION" is displayed on the RVLIS Train A plasma display
Dynamic Head indicator. Which of the following caused this indication't
A. A high volume sensor bellows is leaking causing a hydraulic isolator
e
limit switch to actuate.
B. Pressurizer pressure protection transmitter channel I (PT-1RC-455),
has failed HIGH.
C. This indication is normal, the RCP breaker auxiliary contacts disable
the Dynamic Head indications when ANY RCP is running.
D. A single T-hot narrow range channel I RTD (TRB-RC-412B1], has failed
OPEN.
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POINTS: 1.00
ANSWER: A.
REFERENCES : 10M-6.1.D - Issue 4 Rev. 1
ILP-SQS-6.5 OBJECTIVE: Sh NUMBER: 1-97-012, New
,
JTA #:
K/A #: 3.02.000.K6.03 K/A IMPORTANCE: 3.1/3.6
Final Revision. Rev. 3A
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I
Question Number 28
r
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A large break LOCA and Safety Injection is in progress in the cold leg injection
i phase. What automatic actions will occur within 2.5 minutes of receiving
Annunciator (Al-25] "2/4 RWST LO LEVEL & SI AUTO XFR SI INJ TO RECIRC"?
a
A. CNMT sump to LHSI pump suction valves [MOV-1SI-860A&B] will CLOSE.
B. HHSI to RCL Cold Leg isolation valve [MOV-1SI-836] will OPEN.
C. RWST to HHSI valves [MOV-1CH-115B&D] will OPEN.
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D. LHSI to HHSI cross connect valves (MOV-1SI-863A&B] will OPEN. '
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POINTS: 1.00 !
ANSWER: D.
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REFERENCES: 10M-53.A.1 Att. 1-G, Issue IB, Rev. 1
ILP-SOS-11.1 OBJECTIVE: Sc NUMBER: 1-97-013, M-0421
,
JTA 8:
K/A f: 3.02.006.020.K4.03 (006A3.08) K/A IMPORTANCE: 3.2/3.6
i
Final Revision Rev.3A
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Question Number 29
Given the following:
e The plant is operating at 100% power.
e All systems are aligned in their at-power, NSA configurations.
e The controlling pressurizer level channel LT459, fails high.
Assuming NO operator actions are taken, what will be the First Out Annunciator
for the ensuing Rx Trip? '
A. Low Pressurizer Pressure.
B. Low Pressurize." Level.
C. High Pressurizer Level.
D. High Pressurizer Pressure,
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POIWTS: 1.00
ANSWER: C.
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REFERENCES : 10M-6.1.D - Issue 4, Rev. 1
ILP-SOS-6.4 OBJECTIVE: 14 NUMBER: 1-97-014, M-0658
JTA f:
K/A 4: 3.02.011.000.A2.10 (011K3.01) K/A IMPORTANCE: 3.4/3.6
Final Rension, Rev. 3A
.
Questi:n Number 30
Given the Following:
- A normal plant cooldown and depressurization is in progress using the
condenser steam dumps.
- All S/D rod banks are withdrawn.
- RCS T.,. is 501*F and PZR pressure is 1925 psig,
e The cooldown and depressurization is temporarily terminated to conduct
shift turnover.
- During the turnover T.,. has drifted up to 514*F and PZR pressure has
risen to 2025 psig.
When the cooldown and depressurization is recommenced, which of the following
must be performed?
A. Immediately depressurize the RCS to less than 1945 psig to prevent
exceeding the S/G tube differential pressure limit.
B. Reset the condenser steam dump cooldown valve interlock to restore
manual operator control of the cooldown.
C. Verify the PZR low-pressure reactor trip is bypassed prior to 1945
psig to keep the shutdown banks withdrawn.
D. Re-block the PZR low-pressure SI signal when pressure is reduced below
1980 psig.
POINTS: 1.00
ANSWER: D.
REFERENCES: 10M-51.4.C, Issue 4, Rev. 9 - Coution.
,
I LP-SOS-51.1, OBJECTIVE: 3/9 NUMBER: 1-97-015, M-0255
JTA 4:
K/A 0: 3.02.013.000.K4.12 (010Kl.02) K/A IMPORTANCE: 3.7/3.9
Final Revishn, Rev. 3 A
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QuestionNumber 31
The plant is in Mode 3 making preparations for a Rx startup, when a small break
loss of coolant accident occurs. All systems function as designed and no other
transient exists. Which of the following actions will automatically occur if
all four containment pressure instruments reach a maximum of 5.0 psig?
A. MSLI actuation, Feedwater Isolation actuation and ALL Auxiliary
Feedwater Pumps auto start.
B. Containment Spray Actuation, Emergency Diesel Generators auto start,
and the Motor Driven Auxiliary Feedwater Pumps auto start.
C. MSLI actuation, Emergency Diesel Generators auto start and load, and
the Turbine Driven Auxiliary Feedwater Pump auto starts.
D. CREBAPS actuation, Feedwater Isolation actuation, and the Turbine
Driven Auxiliary Feedwater Pump auto starts.
POINTS: 1.00
ANSWER: A.
REFERENCES: 10M-1,11,12, & 13.2.B - Issue 4, Rev. 4,2,1 & 3 Respectively.
1LP-SOS-11.1/13.1 OBJECTIVE: 5/5 NUMBER: 1-97-016, New
JTA #:
( .
, K/A #: 3.02.013.000.A4.03 K/A IMPORTANCE: 4.5/4.7
l
Final Revision. Rev. 3 A
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i QuestionNumber 32.
1
! - Given the followings
e The Unit is operating at 1004 power with all systems in their at-
power, NSA configurations.
- The following Annunciators are. actuated
e (A4-4), " PRESSURIZER CONTROL LOW LEVEL DEVIATION,"
e [A3-115), " REGEN HX LETDOWN OUTLET TEMPERATURE HIGH,"'
e [Al-39), * CONTAINMENT SUMP LEVEL HIGH."
e PZR level is 42% and dropping slowly.
e Demineralizer Bypass valve [TCV-1CH-143), is in the VCT position.
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- Charging flow control valve [FCV-1CH-122), is full open.
- Charging flow meter (FI-1CH-122), is pegged high.
- Letdown flow meter [FI-1CH-150), indicates 105 gpm .
! *
All RCP seal injection flow meters indicate between 8.0-8.5 spm.
All RCP seal return flow recorders indicate between 2.8-3.2 gpm.
l The location of the leak is on the.
A. in service RCP seal injection filter inlet line.
i
B. charging line between the containment penetration and the Regen
HX.
C. letdown line between the Regen HX and the Letdown Orifices,
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, D. PZR AUX spray control valve-[MOV-1CH-311), inlet line.
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POINTS: 1.00
ANSWER: B. i
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[ REFERENCES: OM Figure 7-1 Rev. 9
- 1LP-SOS-7.1- OBJECTIVE: 2 NUMBER: 1-97-017, New
S' JTA #1 '
]
K/A #: 3.02.000.022.EA2.01 K/A IMPORTANCE: 3.2/3.8
i-
Final Revision, Rev.3A
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l Question Number 33
i Given the following: l
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- The plant is at 47% power. '
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l e All systems are in their at-power, NSA configurations.
! e The pressurizer level selector switch is in position 1, 459/460.
- The reference leg for pressurizer level transmitter (LT-1RC-460)
develops a leak.
Assuming no operator actions are taken, which of the following will occur? l
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A. Pressurizer level will stabilize at the full load setpoint of 59%. l
B. The Rx will eventually trip on low presscrizer pressure.
I
C. Annunciator [A4-1), " PRESSURIZER CONTROL LEVEL HIGH" will actuate.
1
D. All letdown orifice isolation valves will immediately trip closed. '
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POINTS: 1.00
ANSWER: C.
RE FERENCES : 10M-6. 4IF Issue 4, Rev. 5
ILP-SQS-6.5, OBJECTIVE: Sh NUMBER: 1-97-018, M-0116
JTA 8:
K/A I; 3.02.000.028.EKl.01 (028 AKl . 01) K/A IMPORTANCE: 2.8/3.1
Final Revision, Rev.3 A
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QuestionNumber 34
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The pressurizer level controller utilizes an integral control function. Which
of the following describes the action of this integral control function?
The pressurizer level controller integral control action will (1) the
[ demand signal to the Pressurizer Level Control Valve, [FCV-1CH-122) (2) .
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l A. (1) raise
(2) for as long as actual 1,n is below program level.
B. (1) raise
l (2) only as long as actual (pu is dropping.
l
! C. (1) lower
(2) only as long as program 1,u is dropping.
I D. (1) provide
(2) in proportion to the difference between actual and program L,u.
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POINTS: 1.00
ANSWER: A.
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REFERENCES: 1/2LP-ICS-1.4
i ILP-ICS-1.4 OBJECTIVE: 2 NUMBER: 1-97-019, New
-
, JTA f:
K/A #: 3.02.000.028.EK2.03 K/A IMPORTANCE: 2.6/2.9
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Final Revision, Rev. 3A
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QuestionNumber 35
The Unit is in Mode 5 with the RCS water solid. PZR level is >100% and the RCS
heating up. . What valve should be adjusted to control RCS pressure, and in which
direction should this valve be moved in order to maintain RCS pressure stable?
A. RH Letdown to Non Regen HX Inlet Flow Control Valve [MOV-lCH-142),
should be CLOSED.
B. PZR Spray Valve (PCV-lRC-455A), should be OPENED.
C. CCR HX Bypass Temperature Control Valve [TCV-CC-100), should be
OPENED.
D. LP Letdown Back Pressure Regulating Valve (PCV-lCH-145), should be
OPENED.
POINTS: 1.00
ANSWER: D.
REFERENCES: lOM-6. 4. F - Issue 4, Rev. 6
ILP-SOS-6.5 OBJECTIVE: 4 NUMBER: 1-97-020, New
,
JTA 9:
K/A #: 3.03.010.000.A1.04 K/A IMPORTANCE: 3.6/3.8
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Find Futision,:Rn.3A i
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Qu:stionNumber 36
Given the following:
A plant heatup is in progress with the RCS at 366'F/1900 psig.
- Due to an equipment failure, a PZR PORV failed OPEN.
- The PRT Rupture Disc has ruptured.
.
The associated PORV Block Valve has been CLOSED.
AFTER CLOSING the PORV Block Valve the PZR is saturated at 1600 psig.
- Containment pressure is 10 psia.
What will be the PORV Tailpipe Temperature if the Block Valve is NOT fully
closed?
A. 193*F
B. 212*F
C. 233*F
D. 24 6*F
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POINTS: 1.00
ANSWER: C.
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i REFERENCES: OM Fig. 6-2 Rev. 9
ILP-SQS-6.4 , OBJECTIVE: 4 NUMBER: 1-97-021, New
JTA 1:
K/A #: 3.03.000.008.EA1.01 K/A IMPORTANCE: 4.2/4.0
!
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Final Revision, Rev. 3A
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QuestionNumber 37
When trying to establish RCS flow during a loss of ALL 4KV power, which of the
following would cause Natural Circulation flow to RISE 7
A. Lowering the RCS cooldown rate using the Condenser Steam Dumps.
B. Raising the setpoint on the S/G Atmospheric relief valves.
C. Lowering RCS pressure using Aux Spray.
D. Raising all SGWL's using Aux Feed.
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POINTS: 1.00
ANSWER: D.
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PEFERENCES: lOM-53B.5.GI-4 - Issue IB, Rev. 1
i
1LP-SOS-53.2 OBJECTIVE: 11 NUMBER: 1-97-022, New
JTA #:
K/A 4: 3.07.000.055.EA2.02 K/A IMPORTANCE: 4.4/4.6
Final Revision, Rev.3A
.
. . . . .. - .. . - -
Question Numtwr 38
What is the bases for the RCP Trip Criteria Setpoint of RCS/111ghest SG D/P = 150
paid (450 psid Adverse]?
A. Provides for timely RCP trips for small break LOCA events but reduces
the probability of RCP trips for SGTR's and non-LOCA events.
B. Allows the RCPs to remain in service for core cooling until cavitation
damage potential reaches FSAR limits.
C. Ensures RCS fluid level never drops below the elevation of the break
during a small break LOCA thus ensuring re-pressurization of the RCS.
D. Ensures the RCPs are removed from service during a Loss of Secondary
Coolant event which limits the RCS cooldown rate. !
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POINTS: 1.00
ANSWER: A.
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REFERENCES: 10M-53.B.S.GI-6, Issue 1B, Rev. 1
l
ILP-SOS-53.2 OBJECTIVE: 1 NUMBER: 1-97-023, New
. '
JTA #:
K/A #: 3.03.000.011.EK3.14 K/A IMPORTANCE: 4.1/4.2
Final Revision, Rev.3A
. __ - ,-. -. . -_ - _ . _ - _ . .. -.
QuestionNumber 39
.
I A large break LOCA has occurred. The crew has made the following procedure
( transitionss E-1, Loss of Reactor or Secondary Coolant, to ES-1.3, Transfer to
l Cold Leg Recirculation, completed the cold log recirculation line up and then
returned to the procedure and step in effect, E-1 Loss of Reactor or Secondary
l Coolant, step 24. Step 24 of E-1 directs actions to isolate the SI Accumulators
l if at least two RCS hot leg temperatures are less than 390'F.
I The bases for this RCS temperature ensures.
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A. adequate core cooling is established prior to isolating the Accumulators as
a water injection source.
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B. sufficient Accumulator water and nitrogen volumes are injected into the RCS
prior to isolating the accumulators.
C. saturation pressure of the RCS is less than Accumulator nitrogen pressure
when the accumulator water volume is fully discharged.
l D. that the injected Accumulator nitrogen has expanded sufficiently to maintain
RCS saturation temperature less than the UFSAR design bases.
POINTS: 1.00
ANSWER: C.
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! REFERENCES: lOM-53B.4.E-1 - Issue 1B, Rev. 3
l
l lLP-SOS-53.3 OBJECTIVE: 3 NUMBER: 1-97-024, New
JTA #: -
K/A 8: 3.03.000.0ll.EK3.12 K/A IMPORTANCE: 4.4/4.6
.
.
Final Revision, Rev. 3 A
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Question Number 40
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A major plant transient is in progress with the Unit at full power. The current
plant parameters are:
1
(Assume parameters are stable unless otherwise stated.) j
e Highest Core Exit Thermocouple = 623*F
= All RCS T-hot's are between 608-610*F j
e All RCS T-cold's are between 54 4-54 6*F i
e RCS Press = 2112 psig and dropping, j
e A S/G: Press = 800 psig, Level = 44%, Feed Flow = 3.6 E5 lbm/hr.
- B S/G: Press - 805 psig, Level = 43%, Feed Flow - 3.7 E5 lbm/hr.
- C S/G: Press - 790 psig, Level = 44%, Feed Flow = 3.5 E5 lbm/hr.
- Containment:
o Press = 12 psia,
e Temp = 110*F,
o Particulate Rad Monitor [1RM-RM-215A] = 1.3 E5 cpm with its HI i
alarm LIT. I
e Gaseous Rad Monitor [1RM-RM-215B) = 1.9 E5 cpm with its HI-HI
alarm LIT.
In-Core Transfer Device Rad Monitor [1RM-RM-204) = 1000 mn/hr
with its HI-HI alarm lit.
Which of the following events is occurring?
A. A Loss of Secondary Coolant.
B. A Steam Generator Tube Rupture.
C. A Loss of Reactor Coolant.
l D. A single S/G is Ruptured and Faulted inside ccatainment. .
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ANSWER: C. I
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REFERENCES: 10M-53A.1.E-0 - Issue 1B, Rev. 4
1LP-SQS-53.3 OBJECTIVE: 6 NUMBER: 1-97-025, New
,
JTA f:
K/A i: 3.03.000.011.EA2.13 K/A IMPORTANCE: 3.7/3.7
Final Revision. Rev. 3A
i
- CNestion Nundwr 41 f
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'Given the following: I
e .The. Unit is in Mode 4 with a plant heat-up in progress. I
e RCS Temperature is 220*F, being maintained by RHR. !
,
e There is a bubble in the pressurizer with PZR level at 22%. ;
,
- RCS Pressure is 340 psig. ;
.
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.e 1A RCP is in service. l
- Pressurizer overpressure protection system (OPPS) is in service. l
e IEC is performing maintenance on the pressurizer pressure umster and ;
slave controllers, and requests the controllers to be placed in the '
following line-ups i
e PZR Spray Valve controller, (PCV-1RC-455A] in AUTO.
- PZR Spray Valve controller, [PCV-lRC-455B) in MANUAL and shut. .
- PZR Group A Heater control switch in AUTO.
!
e PZR Group B,C,D, and E Heater control switches in OFF. t
e Master Pressure Controller in AUTO.
What effect will this line-up have on PZR pressure with no further operator ,
action?
l
A. No effect. The master pressure controller output is bypassed
with the OPPS keyswitches in AUTOMATIC.
i
B. PZR pressure will be maintained at 340 psig with [PCV-lRC-455A]
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and Group A Heaters operating in AUTO.
+
C. PZR pressure will rise to 410 psig and cause ONLY OPPS Relief l
valve [PCV-lRC-455D) to open. [PCV-lRC-455C) will not open. t
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D. PZR' pressure will rise to 410 psig and cause BOTH OPPS Relief ;
valves [PCV-lRC-455C) and [PCV-lRC-455D) to open. '
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POINTS: 1.00
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ANSWER: D. *
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lOM-6.1.D Issue 4, Rev. 1
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' REFERENCES: i
'1LP-SQS-6.4 OBJECTIVE: 5 NUMBER: 1-97-026, New
,
JTA 6:
K/A #2 3.03.000.027.EA2.03 K/A IMPORTANCE: 3.3/3.4
Final Revision, Rev. 3A -
!
' Qu:stion Number 42
i
Given the followings
e The Unit is in Mode 1 with all systems in their at-power, NSA
configurations.
- The 2B and 2E Pressurizer heaters are in MANUAL and ON to equalize the
PER and RCS boron concentrations.
- The Auto / Man station for PZR spray valve (PCV-RC-455A), has failed,
causing the valve to go full OPEN.
In responding to the transient, the operator placed both PZR sprey valves (PCV-
1RC-455A and B) in NANUAL,'and was able to CLOSE BOTH valves. PZR pressure at
that point was 2156 psig.
If no further operator actions are performed, PZR pressure will-
A. DROP, resulting in an OT/AT Reactor Trip.
B. DROP, resulting in a Low Prescurizer Pressure Reactor Trip.
C. RISE, and cause a PZR Safety valve [RV-1RC-551A] to OPEN.
D. RISE, and cause PZR PORV [PCV-lRC-455C] to OPEN at < 2335 psig
POINTS: 1.00
ANSWER: D.
REFERENCES : lOM-6.4.ABU Step 5 Caution, Issue 3, Rev. O
ILP-SQS-6.4 , OBJECTIVE: 11 NUMBER: 1-97-027, New
JTA #:
K/A 4: 3.03.000.027.G05 K/A IMPORTANCE: 3.3/3.3
- Final Revision, Rev. 3 A
- - - - - - - -. -.~- .. -. _. . _ . - _ - . - _ _
Question Number 43 l
Which of the following is an entry condition for AOP 1.6.4 - Steam Generator
Tube Leakage? l
1. PZR Level Dropping.
2. S/G Blowdown Rad Monitor Rising.
3. SGWL Rising. I
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A. 1 ONLY. ;
!
B. 2 ONLY.
C. 3 ONLY.
D. 1 OR 3.
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POINTS: 1.00
ANSWER: B.
REFERENCES: AOP 1.6.4 Issue 3A, Rev. 7
ILP-SQS-53C.1 OBJECTIVE: 2 NUMBER: 1-97-028, New
,
JTA #2
K/A 8: 3.03.000.037.G11 K/A IMPORTANCE: 3.9/4.1
Final Revision. Rev. 3 A I
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Question Number 44
A continuous action step in E-3, " Steam Generator Tube Rupture", directs the
operator to maintain feed flow to the ruptured S/G until narrow range level is
greater than St. What is the basis for establishing this minimum level in the
ruptured S/G?
A. To compress the ruptured S/G steam bubble and raise its pressure, thus
minimizing break flow.
B. To minimize the ruptured S/G depressurization during the subsequent
RCS cooldown.
C. To dilute RCS water with SG water in anticipation of an uncontrolled
radiological release to the environment.
D. To prevent thermal stratification in the ruptured S/G which would
extend the time required to stop break flow.
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POINTS: 1.00
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ANSWER: B. '
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REFERENCES: 10M-53B.4.E-3 - Issue 1B, Rev. 3
ILP-SOS-53.3 OBJECTIVE: 3 NUMBER: 1-97-029, M-0536
, .
JTA #:
K/A #: 3.03.000.038.EK3.06 (03BEK3.06) K/A IMPORTANCE: 4.2/4.5
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Final Revision. Rev.3A
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Qu:stion Number 45
Given the following:
- The Unit was operating at 100% power with all systems in their at-
power, NSA configurations.
occurred.
e
The operators have transitioned from E-0, Reactor Trip or Safety
Injection, to E-3, Steam Generator Tube Rupture, and have completed
,
isolating 1A S/G.
S/G parameters are:
e 1A S/G; Level = 18%, Pressure = 1010 psig
e IB S/G; Level = 23%, Pressure = 995 psig
e IC S/G; Level = 15%, Pressure = 1000 psig
At this point, a circuit malfunction causes ALL of the steam dumps to trip
CLOSED. Which of the following will occur FIRST in response to the RCS heatup?
A. 1A S/G Atmospheric Relief Valve (PCV-1MS-101A), will modulate open at
1035 psig.
B. 1B S/G Atmospheric Relief Valve (PCV-1MS-101B], will trip open at 1060
psig.
C. IC S/G Atmospheric Relief Valve [PCV-1MS-101C), will trip open at 1060
psig.
D. 1C S/G Atmospheric Relief Valve [PCV-1MS-101C), will modulate open at
1035 psig.
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POINTS: 1.00 l
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ANSWER: C.
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l REFERENCES: 10M-21.3.A - Issue 4, Rev. 4.
ILP-SOS-21.1, OBJECTIVE: 3 NUMBER: 1-97-030, New
'
JTA 9:
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K/A 9: 3.03.000.038.EA1.16 K/A IMPORTANCE: 4.4/4.3
.
Final Revision, Rev. 3A
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Question Number 45
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Which of the 'following describes the response of 1A S/G NR water level and 1A
S/G steam flow when lA RCP is stopped with Rx power initially at 20%?
A. SGWL will initially DROP and steam flow will DROP due to reverse flow
in the loop.
B. SGWL will initially DROP and steam flow will DROP due to the reduced
load on the main generator.
C. SGWL will initially RISE and steam flow will RISE due to reverse flow
in the loop.
,
D. SGWL will initially RISE and steam flow will DROP due to reverse flow
1 in the loop.
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POINTS: 1.00
ANSWER: A.
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REFERENCES: 1LP-SOS-6.3 i
1LP-SQS-6.3 OBJECTIVE: 11 NUMBER: 1 07-031, New
JTA f:
K/A #: 3.04.035.010.Kl.09 K/A IMPORTANCE: 3.8/4.0
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Final Revision, Rev. 3A l
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Question Number 47
With the Unit at full power, the number one seal on 1B RCP has failed. After
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tripping the Rx and stopping the 1B RCP, which of the following is performed to .
limit the temperature rise of the IB RCP lower radial bearing */ j
A. OPEN the RCP Seal bypass valve [MOV-1CH-307).
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B. OPEN.the 1B RCP Thermal barrier flow throttle valve [1CCR-307).
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C. CLOSE the Common seal return isolation valve [MOV-1CH-378 or 381). ;
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D. CLOSE the IB RCP seal leakoff isolation valve [MOV-1CH-303B]. !
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POINTS: 1.00
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ANSWER: D. r
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REFEPENCES: 10M-7.4.ABE - Issue 3, Rev. 3
f ILP-SQS-6.3 OBJECTIVE: 12 NUMBER: 1-97-032, New
l JTA la
+
K/A 4: 3.04.003.000.A2.01 K/A IMPORTANCE: 3.5/3.9
,
Final Revision, Rev. 3A
_ . . . - _ - _ _ , . . . . . _ . . , , _ . - . . _
. . _ _ ._. _ _ . .
Question Number 48
When conducting a core off load in a recent refueling outage, the manipulator
crane operator experienced the following:
- Minor vibrations of a suspended fuel assembly in one quadrant of the
Rx vessel near the outer edge of the vessel,
e visual observation of flow turbulence in the same general location in
the Rx vessel.
With all systems in their normal line up for MODE 6, and an RCS sample in
progress, what is the cause of these indications?
A. High charging flow from the HHSI/ Charging pumps via the 'B' Loop
Hot Leg.
B. High flow through the CVCS letdown orifices via the 'A' Loop Cold -
Leg.
C. High letdown flow to the Residual Heat Removal System via the 'A'
Loop Hot Leg.
D. High RCS sample purge flow to the Primary Sample System via the
'B' Loop Hot Leg.
POINTS: 1.00
ANSWER: C.
REFERENCES: lOM Figure 6-1 - Rev. 7
ILP-SOS-10.1 OBJECTIVE: 1 NUMBER: 1-97-033, New
JTA #:
K/A 8: 3.04.005.000.Kl.09 K/A IMPORTANCE: 3.6/3.9
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[ Final Rnision, Rev. 3 A
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Qttstion Numtber 49 i
The Unit is at 374 Rx power conducting a power ascension to full power. All
systems are aligned in their normal lineups for the current power level with the
following exceptions Turbine EHC control is in MANUAL - IMP OUT control due to i
a problem with the EHC first stage pressure transmitter, which is de-energized. !
The operator depresses the GVt pushbutton for 2 seconds to continue the load j
ascension. What is the response of the main feedwater regulating valves to this
action?
The Main Feedwater Regulating Valves will initially throttle _
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A. CLOSED due to the shrink of the SGWL, and then throttle OPEN when I
level drops below 44%.
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B. CLOSED due to the steam flow - feed flow mismatch, and then throttle
OPEN when level drops below 33%.
C. OPEN due to the swell of the SGWL, and then regulate to control level
at 44%.
D. OPEN due to the steam flow - feed flow mismatch, and then regulate to
control level at 44%.
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POINTS: 1.00
ANSWER: D.
REFERENCES: 10M-24.1.D - Issue 4, Rev. 1
lLP-SOS-24.1 OBJECTIVE: 4 NUMBER: 1-97-034, New
JTA #:
K/A #: 3.04.035.010.A3.01 K/A IMPORTANCE: 4.0/3.9
Final Revision, Rev. 3A
QuestionNumber 50
Which of the following would electrically prevent the 1C RCP breaker from
closing?
A. 1C RCP fl seal differential pressure <200 psid.
B. 1C RCP #1 seal leakoff flow <0.2 gpm.
C. 1C RCP Oil Lift pump is de-energized.
D. RCP Lower Bearing Lube Oil Cooling Water Flow Low.
POINTS: 1.00
ANSWER: C.
REFERENCES : OM Figure 6-21 Rev. 10 and 6-22 Rev. 8
ILP-SQS-6.3 OBJECTIVE: 6 NUMBER: 1-97-035, New
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JTA f:
K/A f: 3.04.000.015.EK2.10 K/A IMPORTANCE: 2.8/2.8
Final lkvislom Rey,3A i
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QurstionNumber 51
While operating in Reduced Inventory /Midloop conditions, a loss of RCS inventory
has occurred, resulting in a loss of RHR. Which of the following will be
performed in accordance with AOP 10.2, Loss of RHR While Operating at Reduced
Inventory, to minimize the possibility of. gas binding the RHR pumps when they
are restarted?
1. ~ Raise RCS Level using a Charging /HHSI pump.
2. Shut the RHR temperature and flow control valves [MOV-lRH-758 and
605).
3. Raise RCS Pressure to greater than 150 psig.
4. Establish level in at least 2 S/G's >15% narrow range.
A. 1 & 3.
B. 1&2. l
C. 1, 2 & 3.
D. 2, 3 & 4.
POINTS: 1.00
ANSWER: B.
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REFERENCES: 10M-53C.4.1.10.1 - Issue 3A, Rev. 5
- 1LP-SQS-10.1 OBJECTIVE: 9 NUMBER: 1-97-036, New
< -
l JTA #:
- ,
i K/A 8: 3.04.000.025.EK1.01 K/A IMPORTANCE: 3.9/4.3
i-.
Final Revision, Rev. 3A
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QuestionNundx:r 52 i
Punctional Restoration Procedure, FR-C.1, " Response to Inadequake Ocre Cooling" i
is designed to reduce core exit thermocouple temperatures and recover Rx vessel '
4 level. Which of the following describes the processes used to accomplish this ;
.- objective in the order in which they are performed?
'
1. Run at least one HHSI pump and OPEN at least two PZR PORVs to ;
establish RCS Bleed and Feed. !
2. Perform a rapid secondary depressurization to depressurize the RCS and
inject the SI Accumulators. }
3. Bastart the RCP's to provide two phase flow through the core. '
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4. Restore High Pressure Sefety Injection flow.
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A. 2, 3, 4.
B. 2, 1, 3. {
C. 4, 2, 1.
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D. 4, 2, 3. ,
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POINTS: 1.00
ANSWER: D.
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REFERENCES: 10M-53I- .FR-C.1 - Issue 1B, Rev. 3
ILP-SOS-53.3 OBJECTIVE: 3 NUMBER: 1-97-037, New j
JTA 8: l
K/A i: 3.04.000.074.EK1.03 K/A IMPORTANCE: 4.5/4.9
Find Revision. fky,3A
. - . . - . - . . . - - . . _ - _ _ . . .
QuestionNumber 53
l When responding to a Degraded Core Cooling condition in FR-C.2, the operator is
directed to " Verify SI Valve Alignment" with the SI system in the Cold Leg
Injection Mode. Which of the following valves should be CLOSED for the current
plant conditions?
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l A. RWST Discharge to Charging Pumps Suction Valve [MOV-1CH-115B).
B. Regen HX/Chg Header Inlet CNMT Isolation Valve [MOV-1CH-289).
C. AFW Turbine Steam Supply B Train Trip Valve [TV-1MS-105B).
D. BIT Outlet Isolation Valve [MOV-1SI-8 67D) .
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POINTS: 1.00
ANSWER: D.
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REFERENCES: 10M-53A.1.1-A - Issue 1B, Rev. 1
1LP-SOS-11.1 OBJECTIVE: 3 NUMBER: 1-97-038, New
. ,
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JTA #:
K/A 8: 3.04.000.074.EA1.27 K/A IMPORTANCE: 4.2/4.2
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Final Revision Rev.3A
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QuestionNumber 54
( The 'C' Loop Main Steam Isolation Valve (TV-1MS-101C), inadvertently trips ,
!
closed with the Unit operating at 40% power with all systems in their at-power, ',
NSA configurations. The Rx tripped on low SGWL in the 'C' S/G and NO Safety
Injection occurred. l
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For the above listed conditions, the steam dumps will maintain _
A. ALL S/G pressures at ~1005 psig by maintaining T. . at 449'T (2*F dead
band from no-load T. ) using steam from only 'A' & 'B' S/G's.
l B. 'A' & 'B' S/G pressures at -1005 psig and the 'C' S/G atmospheric
l steam dump valve (PCV-1MS-101C), will trip open if 'C' S/G pressure
reaches 1060 psig.
C. 'A' & 'B' S/G pressures at -1005 psig and the 'C' S/G atmospheric
steam dump valve [PCV-1MS-101C), will modulate open if 'C' S/G
pressure reaches 1035 psig.
! D. 'A' & 'B' S/G pressures at -1005 psig and the 'C' S/G safety valves
,
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will open in succession if 'C' S/G pressure reaches 1075 psig.
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POINTS: 1.00
ANSWER: B.
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REFERENCES: 10M-21.2.B - Issue 4, Rev. 1, 10M-21.3.A - Issue 4, Rev. 4.
ILP-SQS-21.1 OBJECTIVE: 3 NUMBER: 1-97-039, New
l ' JTA #: j
, K/A #: 3.05.039.000.K1.02 K/A IMPORTANCE: 3.3/3.3
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Fina! Revision, Rev. 3A
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QuestionNumber 55
Given the following:
- The Unit is in MODE 1 with Rx power at Bl.
- Boron concentration is 665 ppm. )
- Tave is 547'F. 1
e PZR Pressure is 2235 psig.
- PZR Level is 22%.
- Steam dumps are in Auto in the Main Steam Pressure Control Mode.
o S/G atmospheric relief valve controllers are in Manual.
e All systems are aligned in their normal lineup for the existing power I
level. '
- An inadvertent MSLI occurs.
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Assuming no Rx Trip occurs and no operator action, determine the response of the j
following parameters one minute after the MSLI.
Rx Power 'A' S/G PZR Level
Pressure
A. DROP OTAP RISE
B. RISE RISC DROP
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C. DROP RISE RISE :
D. RISE DROP RISE
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POINTS: 1.00
ANSWER: C.
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REFERENCES: LP-RT-6, Objective 15
ILP-SOS-21.1 OBJECTIVE: 2 NUMBER: 1-97-040, New
JTA #
K/A #: 3.02.002.000.KS.11 K/A IMPORTANCE: 4.0/4.2 !
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Final Revision, Rev. 3 A
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l QuestionNumber 56
The EHC auxiliary governor speed sensor circuits have failed to 104% of rated
speed, causing the 20-1-OPC and 20-2-OPC solenoids to energize. Which of the
following describes the response of the EHC system to this malfunction?
'
A. The Control Valve Emergency Trip Header will be continuously dumped,
causing the Governor, Interceptor and Extraction Steam Ncn-Return
valves to close rapidly and remain closed.
!
B. The Control valve Emergency Trip Header will be dumped for 1.5
seconds, causing the Governor, Interceptor and Extraction Steam Non-
Return Valves to close rapidly and then re-open.
,
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C. The Trip Valve Emergency Trip Header will be continuously dumped,
causing the Reheat and Turbine Stop valves to close rapidly and remain
closed.
D. The Trip Valve Emergency Trip Header will be dumped for 1.5 seconds
i out of every 31.5 seconds, causing the Reheat and Turbine Stop valves
to reduce load at 200%/ min.
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l POINTS: 1.00
ANSWER: A.
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REFERENGS: 10M-26.1.B - Issue 4, Rev. 4
- ILP-SOS.26.3 OBJECTIVE
- 9d NUMBER: 1-97-041, New
JTA #:
j K/A 8: 3.05.045.050.A2.10 K/A IMPORTANCE: 2.7/2.9
Fina! Revision. Rev. 3A
Question Number 57
1
Given the following:
- The Unit is stable at 85%.
- All systems are in their at-power, NSA configurations. j
- The operator notices that the 1C S/G feed reg valve [ FCV-1 FW-4 98 ] I
demand signal is 15% lower than the 1A and 1B S/G feed reg valves .
[FCV-1FW-478 and 488). l
- The S/G 1evel chart recorders show that all three S/G levels have been )
stable at 44% and that all three S/G steam flows and feed flows have '
been stable at 85% for an extended period of time.
Which of the following statements explains the observed plant conditions?
i
A. 1C S/G controlling feed flow transmitter has failed LOW.
B. 1C S/G feed reg valve bypass valve [ FCV-1 FW-4 9 9 ) has failed
OPEN.
C. There is a feed water leak upstream of the 1C S/G feed reg i
valve [FCV-lFW-498). l
D. 1C S/G feed reg valve [FCV-lFW-498) valve stem has become l
uncoupled from its actuator.
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l POINTS: 1.00
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ANSWER: B.
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REFERENCES : 10M-24.1.D - Issue 4, Rev. 1
ILP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-042, M-0153
JTA 1:
K/A 4: 3.05.059.000.Kl.04 (059A2.ll) K/A IMPORTANCE: 3.4/3.4
Final Revision, Rev. 3A
- - . . . . . . - _.
QuestionNumber 58
Which of the following is the power supply to 1EW-P-3A, Motor Driven Aux Feed
Pump?
A. AE 4KV Bus.
C. 8N 480V Bus.
D. 9P 480V Bus.
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POINTS: 1.00 i
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ANSWER: A.
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REFERENCES: 2OM-24.3.C - Issue 4, Rev. 6
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1LP-SOS-24.1 OBJECTIVE: 2 NUMBER: 1-97-043, New
JTA 6:
K/A 8: 3.05.061.000.K2.02 K/A IMPORTANCE: 3.7/3.7
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! Final Revision. Rev. 3A
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QuestionNumber 59
Which of the following would indicate a reduction of total AEW flow capability,
following an automatic Low PZR Precsure Rx Trip from 1004 power?
A. AFW Turbine Steam Supply Trip Valve [TV-1MS-105A] is OPEN.
B. Annunciator [A7-7), "STM UNAVAILABLE TURB DRIVEN FEED PP" is NOT in
alarm.
C. 1B S/G AW Flow Control Valve [MOV-1FW-151C) is OPEN.
D. Motor Driven AfW Pump (1FW-P-3A) Recirc Valve [FCV-1FW-103A] is OPEN.
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POINTS: 1.00 ;
ANSWER: D.
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REFERENCES: 10M-24.2.B - Issue 4, Rev. 0
ILP-SOS-24.1 OBJECTIVE: 4 NUMBER: 1-97-044, New
JTA #:
K/A 8: 3.05.061.000.A3.01 K/A IMPORTANCE: 4.2/4.2
Final Revision, Rev. 3 A
, . . _ _ _ _ . _ _ _ _ ._._..__m.___.______._,,_..___..__.-._.__._.___. .____._.._.m._..
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l QuestionNumber 60
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I which of the following is the initial method of restoring Rx Plant River Water
l Header pressure in the event of a COMPLETE loss of the Normal River Water Intake
l structure?
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A. . Aligning Unit 2 'B' Service Water Header to supply Unit 1 'A' Rx Plant
River Water Header.
l . B. Aligning the Diesel Driven Fire Pump (1FP-P-2] to supply Unit 1 'A' Rx
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Plant River Water Header.
- C. Aligning Unit 1 Aux River Water Pumps to supply that pumps respective
' Unit 1 Rx Plant River Water Header.
D. Aligning Unit 1 Turbine Plant River Water Pumps to supply that pumps
! respective Unit 1 Rx Plant River Water Header,
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- POINTS: 1.00
- ANSWER: C.
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REFERENCES: 10M-30.4.AAC
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ILP-SQS-30.2 OBJECTIVE: 3 NUMBER: 1-97-045, New
-
JTA f:
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K/A f: 3.05.076.000.Kl.21 K/A IMPORTANCE: 2.7/2.9
(. I
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Final Revision.Rev.3A
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QuestionNumber 61
,
Which of the following design features protect plant personnel and Systems,
Structures and Components (SSC's) outside of containment from the effects of a
High Energy Line Break?
A. Automatic S/G blowdown isolation on High Cable Vault Pipe Tunnel Area I
temperature.
B. Programming.of SGWL to reduce the total mass in the S/G's at high '
power levels.
{
C. The Main Steam Line Isolation on a High-2 containment pressure signal. !
D. ' Automatic opening of the Main Steam Valve Room Louvers on high room
temperature, j
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POINTS: 1.0L
ANSWER: A.
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REFERENCES: BVPS UFSAR Appendix. D; Pg. D.1-3 - Rev. 1, and Pg. D.1 -4 - Rev. 4.
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ILP-SQS-25.1 OBJECTIVE: 5 NUMBER: 1-97-046, New
JTA #: i
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K/A #: 3.05.000.040.EKl.06 K/A IMPORTANCE: 3.7/3.8
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Final Revision. Rev.3A
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QuestionNumber 62 l
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The Unit.is at 1004' power with all systems in their at-power, NSA
configurations. At'a MINIMUM, which of the following control switches, needs to l
.be placed in the CLOSE position to prevent 1A S/G from feeding the Main Steam .I
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' Manifold? r
A. TV-1MS101A, 1A SG MAIN STEAM A TRN TRIP VLV.
B. TV-1MS101A, 1A SG MAIN STEAM A TRN TRIP VLV,
MOV-1MS-101A, 1A SG MAIN STEAM BYPASS TRIP VLV.
C. TV-1MS101A, 1A SG MAIN STEAM A TRN TRIP VLV,
Tv-1MS101A, 1A SG MAIN STEAM B TRN TRIP VLV,
MOV-1MS-101A, IA SG MAIN STEAM BYPASS TRIP VLV.
D. TV-1MS101A, 1A SG MAIN STEAM B TRN TRIP VLV,
MOV-1MS-101A, 1A SG MAIN STEAM BYPASS TRIP VLV,
TV-1MS-111A, 1A MAIN STEAM LINE PRE-NRTRN DRAIN ISOL VLV.
POINTS: 1.00 I
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ANSWER: A. I
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REFERENCES: 10M Figures 21-1 Rev. 10 and 21-06 Rev. 4 I
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ILP-SQS-21.1 OBJECTIVE: 3 NUMBER: 1-97-047, New
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l K/A # 3.05.000.040.EA1.04 K/A IMPORTANCE: 4.3/4.3
4
FinalRevision.Rev 3A
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Question Number 63
Assuming all turbine vibration readings are normal, which of the following
sustained conditions would require a manual Turbine and/or Rx Trip ~l l
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Main Generator Condenser
Output (MW) Backpressure
(In. Hg Abs.)
1. 230 4.0
2. 700 4.5
3. 800 6.5
4. 210 2.5
A. 1, 2 and 4 ONLY.
B. 1 and 3 ONLY. j
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C. 3 ONLY.
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D. 2 and 4 ONLY. j
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Provide the following references: 10M-26.4.AAS * Condenser Vacuum Low" alarm
response procedi:re.
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REFERENCES: 10M-26.4.AAS - Issue 3, Rev. 3
- i
ILP-SOS-26.2 OBJECTIVE: 9d NUMBER: 1-97-048, New
JTA 8: i
K/A 8: 3.05.000.051.EA2.02 K/A IMPORTANCE: 3.9/4.1 .
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Final Revision, Rev. 3 A
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QuestionNumber 64
Control for the 1C S/G 'A' Train Auxiliary Feedwater Regulating Valve [MOV-1FW-
j 151B], has been transferred to the Emergency Shutdown Panel (SDP) . What actions
- are nr.cessary to transfer control of [MOV-1FW-151B] back to the Main Control
H
Room Aenchboard (BB-C)?
A. Operate the control rswitch on BB-C for (MOV-1 W-151B] out of the
NORMAL position.
B. Depress the Transfer Pushbutton for [MOV-1 W-151B] on the SDP.
C. Reset the Master Reset Transfer Relay on the SDP.
D. Reset the SDP Transfer Relay for (MOV-1FW-151B] at the respective Aux
Relay Panel.
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ANSWER: D. l
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REFERENCES: 10M-24.1.D - Issue 4 Rev. 1
ILP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-049, New
JTA #:
K/A #: 3.05.000.054.EA1.01 K/A IMPORTANCE: 4.5/4.4
FinalRevisior Rev.3A
_ _ _ _ _ - - ..
, QuestionNumber 65
q With the Unit in Mode 5, which of the following automatic actions will occur
when the Containment Purge Exhaust Monitor [RM-IVS-104A) reaches a High-High
'
alarm condition?
,
A. A CIA signal will be generated isolating all Phase-A flowpaths from
the containment.
B. Containment Purge Supply and Exhaust Fans will trip, and the Purge j
Supply and Exhaust Dampers will close. i
C. The Supplemental Leak Collection and Release System (SLCRS) will be i
f aligned to bypass the Main Filter Bank and provide an elevated release l
flowpath.
D. The Containment Purge Exhaust will be aligned to the SLCRS and then ,
filtered through the Main Filter Banks. !
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POINTS: 1.00
ANSWER: B.
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REFERENCES : lOM-44C.l.B - Issue 4, Rev. O
ILP-SQS-44C (2382) OBJECTIVE: 3 NUMBER: 1-97-050, New
JTA #:
K/A #: 3.06.022.000.K4.03 K/A IMPORTANCE: 3.6/4.0
Final Revision. Rev. 3 A
QuestionNumber 66
Actual containment air partial. pressure is 9.8 psia. If river water temperature
were to rise from 72*F to 77'F, the required Maximum Allowable Operating Air
Partial Pressure would be (1) psia and the actual containment air
partial pressure (2) meet the Technical Specification requirement.
(1) (2)
A. 9.7 would NOT
B. 9.7 would
C. 9.9 would NOT
D. 9.9 would
POINTS: 1.00
ANSWER: A.
Provide the following references: Unit 1 Technical Specification figure 3.6-1..
REFERENCES: BVPS TS, Figure 3.6-1, Amendment No. 174
ILP-SQS-12.1 OBJECTIVE: 10 NUMBER: 1-97-051, New
JTA i:
K/A #: 3.06.022.000.A1.04 K/A IMPORTANCE: 3.2/3.3
Final Revision, Rev. 3A
. . - _. -- ._. -- - - - _ .. . . _ - - . .
Question Number 67
The bases for the OPERABILITY of the Containment Quench and Recirculation Spray
Systems is to ensure containment depressurization and subsequent return to;
A. subatmospheric pressure in the event of a Main Steam Line Break inside
containment.
B. atmospheric pressure in the event of a Main Steam Line Break inside
containment.
C. subatmospheric pressure in the event of a Loss of Rx Coolant Accident.
D. atmospheric pressure in the event of a Loss of Rx Coolant Accident.
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POINTS: 1.00
ANSWER: C.
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REFERENCES: BVPS Unit 1 TS, 3/4.6.2 Bases, Amendment No. 200
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ILP-SQS-13.1 OBJECTIVE: 11 NUMBER: 1-97-052, New
JTA #:
1
K/A #: 3.06.026.000.G06 K/A IMPORTANCE: 2.5/3.8 j
i Finat Pe+br Pay,3A
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' Question Number' 68
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The zirc-water reaction of-the fuel clad is one of the major sources of Hydrogen. !
(H2) generation in the containment following the design bases Loss of Rx Coolant !
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' Accident (DBA LOCA). Which of the following is another major source of Hydrogen ;
generation in the containment following the DBA LOCA? -
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l ' A. H2 gas that has accumulated in the Pzr gas space. l
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(. B. H2 released from the assumed lot failed fuel pins. .
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I .D. H2 produced from the corrosion of the RCS Inconell clad materials.
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ANSWER: A.
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l . REFERENCES: UFSAR Section 14.3, page 14.3-39, Rev.12.
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, ILP-SOS-ATA4.2 OBJECTIVE: 1 NUMBER: 1-97-054, M-0006
. ,
l, JTA #:
l K/A #: 3.06.028.000.K5.03 (028K5.03) K/A IMPORTANCE: 2.9/3.6
!
. Final Revision, Rev.3A
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Question' Number 69
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Given the following: ,
l e
The 60 gpa Ltdn Orifice Cnat Isol Viv [TV-1CH-200B], is to be stro:ced
! closed and timed by de-energizing (SOV-1CH-200) as part of an OST.. I
e ..one 60 gpm, and the 45 gpa Letdown Orifices are in service. [
e The LP Ltdn Back Press Reg Viv (PCV-1CH-145), is in MANUAL. l
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l .In order to maintain letdown system stability when (TV-1CH-200B] is going l
l- CLOSED, (PCV-1CH-145) should be throttled (1) , to (2) letdown i
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pressure as indicated on the BB-A Letdown Pressure Indicator (PI-1CH-145).
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A. CLOSED, RAISE f
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C. OPEN, RAISE
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D. OPEN, LOWER
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POINTS: 1 00
ANSWER: A.
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, ' REFERENCES: 10ST-47.3A - Issue 4, Rev. 15. !
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l ILP-SOS-7.1 OBJECTIVE: 10 NUMBER: 1-97-055, New j
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JTA #2
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K/A #: 3.06.103.000.A4.01 K/A IMPORTANCE: 3.2/3.3
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QuestionNumber 70
Which of the following Containment Airlocks are interlocked to prevent the inner
and outer doors from being opened at the same time?
1. Normal Personnel Air Lock 04-Inch Full Size Doors.
2. Normal Personnel Air Lock 18-Inch Escape Hanway Doors.
3. Equipment flatch Emergency Air Lock Doors.
A. 1 and 2.
B. 1 and 3.
C. 2 and 3.
D. 1, 2 and 3.
POINTS: 1.00
ANSWER: B.
REFERENCES: 10M-47.4.B - Issue 4, Rev. 3, 10M-47.4.C - Issue 4, Rev. 1 and
10M-47.4.D - Issue 4, Rev. 1.
-1LP-SQS-47.1 OBJECTIVE: 4 NUMBER: 1-97-056, New
JTA #:
K/A 8: 3.06.000.069.EK2.03 K/A IMPORTANCE: 2.8/2.9
Final Revhion. Rev. 3 A
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QuestionNumber 71
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Under which of the following conditions would Technical Specifications be
satisfied and allow CORE ALTERATIONS to commence?
NOTE: Assume there are NO blank flanges or pipe caps installed on equipment
that is disassembled or removed.
A. e 1A S/G secondary side manway is removed,
o The 1A S/G MS Trip and Bypass Valves [TV-1MS-101A] and
[MOV-1MS-101A] are CLOSED, and
e All 1A S/G pressure transmitter sensing lines are removed
from the steam lines to install new pressure taps.
B. * CRDH Shroud Cooling Coil relief valve [RV-1CC-113A] is
removed,
o CRDM Shroud Cooling outlet isolation valve (ICCR-188] is
CLOSED, and
e CRDM Shroud Cooling Coil Inlet isolation valve [MOV-1CC-
111A] is CLOSED.
C. e The CVCS charging line is removed from the Regen Heat
Exchanger,
e The charging line vent valve [lCH-390] is locked OPEN, and
e Charging Line Containment Isolation Valve [MOV-1CH-209] is
CLOSED but inoperable due to its motor operator leads being
lifted.
D. * Containment Equipment Hatch installed with three closure
bolts, and
a both doors on the Emergency Personnel Access Hatch CLOSED,
and
- All containment purge dampers are CLOSED.
POINTS: 1.00
ANSWER: B.
Provide the following references: VOND Figures; 7-1 Rev. 9, 15-3 Rev. 2, 21-1
Rev. 7 and Section 3.9 of the Unit 1 Technical Specifications.
RE FERENCES : 10M Figures; 7-1 Rev. 9, 15-3 Rev. 2, and 21-1 Rev. 7, and BVPS
Unit 1 TS 3.9.4 Amendment No. 185.
ILP-SOS-6.13 OBJECTIVE: 12 NUMBER: 1-97-057, New
JTA #:
l K/A #: 3.06.000.069.EA2.02 K/A IMPORTANCE: 3.9/4.4
Final Revision. Rev,3A
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QuestionNumber 72
All of the following Emergency Diesel Generator (EDG) conditions will prevent
the EDG output breaker from closing during a loss of all offsite power EXCEPT
a(n).
A. Electripal Engine Overspeed Trip Signal.
B. Generator Overcurrent Signal.
C. Generator Output Voltage at 2.1KV.
D. Engine Oil Low Pressure Trip Signal.
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POINTS: 1.00
ANSWER: D.
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REFERENCES: 10M-36.1.E - Issue 4, Rev. 1.
ILP-SQS-36.2 OBJECTIVE: 6 NUMBER: 1-97-058, New
JTA #:
K/A 8: 3.07.062.000.K3.02 K/A IMPORTANCE: 4.1/4.4
Final Revision, Rev. 3A
_.___m .. . _ _ ,_ _ _ . . . . _ . . . . _ - _ _ _ _ . . _ . _ . . _ _ _ _ . . . _ . . _ . _ . _ . _ _ _ _ . _ . . _ _ _
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', QuestionNumber 73
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Which of the following describes the relationship between the stations 125VDC
batteries and their respective battery chargers?
In the NSA configuration, the Battery Charger Output Breaker is normally
A. OPEN, and CLOSES automatically to charge the battery when battery.
voltage drops below a preset value.
B. CLOSED, allowing the charger to supply the normal DC loads, and
remains CLOSED on a loss of AC input power to the charger.
C. CLOSED, allowing the charger to maintain a continuous equalizing
charge on the_ battery, and OPENS on a loss of AC-input power to the
charger.
D. CLOSED, allowing the charger to maintain a float charge on the
battery, and OPENS on a loss of AC input power to the charger.
POINTS 1.00
ANSWER: B.
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REPERENCES: 10M-39.1 - Issue 4, Rev. 0 '
ILP-SQS-39.1 OBJECTIVE: 1 NUMBER: 1-97-059, New
JTA # ]
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K/A #: 3.07.063.000.K1.03 K/A IMPORTANCE: 2.9/3.5 'I
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Final Revision, Rev.3A
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Qu3stion Number 74 ,
,
With the Emergency Diesel Generator (EDG) LOCAL / REMOTE keylock switch in the
LOCAL position, the EDG will
A. START on an SI or UNDERVOLTAGE signal, but the output breaker WILL NOT
automatically close.
,
B. START on an SI or UNDERVOLTAGE signal, and the output breaker WILL
automatically close.
C. NOT START on an SI signal, but WILL START on an UNDERVOLTAGE signal,
and the output breaker WILL automatically close. ,
D. NOT START on an SI or UNDERVOLTAGE signal, therefore the output
breaker WILL NOT automatically close.
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POINTS: 1.00
ANSWER: D.
REFERENCES: 10M-36.1.E - Issue 4, Rev. 1
ILP-SOS-36.2 OBJECTIVE: 7 NUMBER: 1-97-060, M-0664
JTA #1
g K/A #: 3.07.064,000.A4.01 (064A4.01) K/A IMPORTANCE: 4.0/4.3
4
Final Revision. Rev. 3A
. . _ .
Questionthunber 75
Given the following:
break inside containment has just occurred. '
There are indications of a steam void in the Rx Vessel head. l
- All SG narrow range water levels are still <5%. I
e Containment pressure is 8.5 psig. l
- The 1A Emergency Diesel Generator has just been started locally. '
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Which of the following pumps should be started / verified running FIRST, and the l
bases for this action? I
A. Charging /HHSI pump, to collapse the Rx Vessel head void.
B. Motor Driven Aux. Feedwater pump, to establish a Heat Sink.
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C. River Water pump, to provide cooling to the EDG's. )
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D. Quench Spray pump, to reduce containment pressure. )
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POINTS: 1.00
ANSWER: C.
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RE FERENCES : 10M-53B.4.ECA-0.0 - Issue IB, Rev. 3.
ILP-SQS-53.3 CBJECTIVE: 4 NUMBER: 1-97-061, M-0561
JTA #:
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K/A #: 3.07.000.055.EA1.06 (064K1.02) K/A IMPORTANCE: 4.1/4.5
,
final Revision, Rev. 3A
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QuestionNumber 76 )
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ECA-0.0 " Loss of All Emergency 4KV AC Power" directs the operator to locally
close the seal Water Return Containment Isolation Valve [MOV-lCH-381). Which of
the following is the bases for performing this step?
A. To prevent over pressurizing and possibly rupturing the VCT.
B. To minimize the potential for a radioactive release within the Aux.
Building. <
C. To minimize the chance of RCP seal damage when seal injection and CCR
flow is isolated.
D. To prevent steam formation on the CCR side of the Seal Water Heat
Exchanger. i
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POINTS: 1.00
ANSWER: B.
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REFERENCES: 10M-53B.4.ECA-0.0 - Issue 1B, Rev. 3 l
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ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-063, New '
JTA #:
.
K/A #: 3.11.000.059.EK3.04 K/A IMPORTANCE: 3.8/4.3
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Final Revision, Rev. 3 A
, _ _.___. _ __ . _ . _ _ . . . . . ...m_.__ _.._____m._______ _ _ _ _ _ . . _ _ _ _ _ .
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j QuIstionNumber 77
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The operators are responding to Annunciator (Al-11) " VITAL BUS II TROUBLE" IAW
the associated ARP. They are directed to place the 120vac Vital Bus II inverter
' Man Bypass _ Switch' to the 'STfW-ISOL' position.
, ..
l This switch' position transfers the Vital Bus power source from the *
'
(1) to the (2) and when the normal power source is t
.
restored, allows the (3) _ _ _ transfer back to the normal power source.
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(1) (2) (3)
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i A. AC Static Line Inverter Output Automatic
l Regulator
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I B. AC Static Line Inverter Output Manual
l Regulator ,
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C. Inverter Output AC Static Line Automatic
Regulator ,
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D. Inverter Output AC Static Line Manual
l Regulator
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ANSWIR: D.
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REFERENCES: 10M-38.4.AAC - Issue 4, Rev. 1, and Figure 38-3 - Issue 3, Rev. 6.
ILP-SOS-38.1 OBJECTIVE: 2 NUMBER: 1-97-064, New j
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K/A 6: 3.07.000.057.EK3.01 K/A IMPORTANCE: 4.1/4.4
Final Revision, Rev.3A I
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Questi:nNumber 78
All of the following will RAISE the 125VDC battery capacity EXCEPT-
A. disconnecting from the battery, individual loads that have indications
of being grounded.
B. periodically performing a deep cycle and battery equalizing charge.
C. ensuring the individual cell electrolyte levels fall below the minimum
level before refilling.
D. maintaining the batteries on a continuous float charge.
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POINTS: 1.00
ANSWER: C.
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l REFERENCES: 10M-39.1.D - Issue 4, Rev. 1
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ILP-SQS-39.1 OBJECTIVE: 2 NUMBER: 1-97-065, New
JTA #:
K/A f: 3.07.000.058.EK1.01 K/A IMPORTANCE: 2.8/3.1
Final Revision. Rev. 3A
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QuestionNumber 79
The Diesel Air Compressor has a low engine oil pressure trip for engine
protection. What allows the Diesel Air Compressor to start with a standing low
oil pressure signal before engine oil pressure has a chance to build up?-
The Engine Low Oil Pressure Trip is-
A. automatically bypassed for the first 15 seconds of operation to allow
oil pressure to build up.
B. cleared when the operator starts the auxiliary oil pump prior to
diesel engine start up.
C. manually bypassed by the operator when the ' Diesel Air Compressor
Start Pushbutton' is depressed.
D. manually bypassed by the operator by maintaining the ' Diesel Air '
Compressor Selector Switch' in the ' HOLD' position.
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POINTS: 1.00 1
ANSWER: D.
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. Provide the following references: 10M-34.4.L * Placing Diesel Air Compressor
Into Service".
REFERENCES: 10M-34.4.L - Issue 4, Rev. 1
ILP-SQS-34.1 OBJECTIVE: 5 NUMBER: 1-97-067, New
JTA I:
K/A 4: 3.08.000.065.EA1.04 K/A IMPORTANCE: 3.5/3.4 I
Mnal ikvision, Rev. 3A ;
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Questi:nNumber 80
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What are the affects on the Solid State Protection System (SSPS) if the 120vac
Vital Instrument Bus II is de-energized? l
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Assume initial conditions of 100% power, with all systems in their at-power NSA !
configurations. !
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A. ALL of the Train D, Output Bay Slave Relays will NOT function on a ;
Safety Injection signal.
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B. ONLY the il Emergency Diesel Generator Load Sequencer will load the
required components on a Safety Injection signal.
C. ALL of the Train B, Input Bay P.alays will de-energize resulting in a
Rx Trip and Safety Injection.
D. Train B will function as designed due to the auctioneered power
supplies to the Logic and Output Bays.
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POINTS: 1.00
ANSWER: A.
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REFERENCES: lOM Figure 1-43 - Rev. B l
ILP-SOS-1.2 OBJECTIVE: 2 NUMBER: 1-97-069, New
JTA #:
K/A #: 3.09.032.000.K2.01 K/A IMPORTANCE: 3.3/3.7
Final Rnision, Rev. 3 A
QuestionNumber 81
Given the following:
- The Unit is in Mode 6, conducting a core off-load.
- N-43 power range instrument is on clearance for paintenance.
- N31 and N32 Source Range Instruments HV Manual On/Off switches are in
NORMAL.
- All other systems are in their normal line-ups for the current Mode of
operation.
- The 120vac Vital Bus IV is inadvertently de-energized.
Following the loss of the 120vac Vital Bus IV, Core Alterations may_ 4
A. continue, the 120vac Vital Buses are NOT required by the
refueling Technical Specifications.
B. continue, provided boron concentration of the RCS is determined
at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
C. NOT continue, due to the loss of both source range nuclear
instruments.
D. NOT continue, due to the inability to actuate a complete
containment phase A isolation (CIA).
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POINTS: 1.00
ANSWER: C.
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l - Refueling Operations.
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REFERENCES: lOM-1.5.B.1, Table 1-2 - Issue 2, Rev. 8 & TS 3.9.2. Amend No. 175.
ILP-SOS-2,1 OBJECTIVE: 2 nut!BER: 1-97-070, New
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JTA 8: ,
K/A f: 3.09.015.000.K2.01 K/A IMPORTANCE: 3.3/3.7
Final Revision, Rev. 3A
QuestionNumber 82
Given the following:
- The Unit is in Mode 3 with the shutdown banks fully withdrawn,
preparing to enter Mode 2.
- Source Range counts are:
e N31, 120 cps
e N32, 130 cps
- A review of the most recent N36 nuclear intermediate range (IR)
channel functional test (OST 1.2.2) indicates the "as-left" setting
for the IR high neutron flux trip is equivalent to 33% power. (3.5 x ;
10E-4 amps). .
Which of the following describes the Technical Specifications required actions j
for this condition? '
A. Adjust the N36 high flux trip setpoint to the current equivalent of 25%
rated thermal power prior to raising power above P-6.
B. Adjust the N36 high flux trip setpoint to the current equivalent of 25% i
rated thermal power prior to raising power abcve 5% rated thermal power. )
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C. Place N36 in the tripped condition when greater than P-10 and the low power
trip setpoints have been blocked.
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D. Place the N36 ' Level Trip Bypass' switch in the 'BYPA3S' position prior to
exceeding P-6.
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ANSWER: A. '
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Provide the following references: Unit 1 Technical Specification 3.3.1.1, Table
! 3.3-1.
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REFERENCES: Unit 1 TS 3. 3.1.1, Table 3. 3-1 - Amendment No. 195
ILP-SOS-TS OBJECTIVE: 1 ' NUMBER: 1-97-071, M-0270
JTA 6:
K/A #: 3.09.015.000.G05 (2.2.22) K/A IMPORTANCE: 3.3/3.8
l'inal Revision, Rev. 3 A
Question Number 83
Following a spurious Rx trip from 1006 powers
The turbine trip was delayed for five seconds, causing T... to drop to
54 0*F.
- Two cooling tower pumps have tripped.
1A RCS loop hot leg temperature instrument has failed high.
Assuming no operator action, which one of the following describes how RCS T.v.
will be maintained?
A. At 547'T by the steam dumps with ONLY the cooldown valves (PCV-1MS-
106A, B, & B1] amed.
B. At 547*F by the steam dumps with the first 2 banks of the steam
dump valves armed.
C. By ONLY the atmospheric steam dump valves [PCV-1MS-101 A, B, & C]
cycling at their trip open setpoint.
D. At 543'F by the condenser steam dumps cycling open due to the
failed high Thet instrument and closed by the Lo-Lo T.. interlock. ,
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POINTS: 1.00
ANSWER: B.
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REFERENCES: 10M Figure 21.1.D - Issue 4, Rev. 1.
1LP-SQS-21.1 OBJECTIVE: 3, 4 NUMBER: 1-97-072, M-0188
JTA 9:
K/A 9: 3.09.016.000.K3.03 (041K4.09) K/A IMPORTANCE: 3.0/3.1
Fmal Revision. Rev. 3A
. . . - .. . . - _ .
QuestionNumber 84
Which of the following do the Emergency Operating Procedures use as an
indication of Natural Circulation flow occurring within the RCS?
1. RCS T.,. RISING.
2. S/G Pressures DROPPING.
3. Core Exit Thermal Couples STABLE.
4. RCS T-Hot's AT T-SAT FOR S/G PRESSURE.
A. 1 and 3.
B. 1, 2, and 4.
C. 2 and 3.
D. 2, 3 and 4.
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POINTS: 1.00
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ANSWER: C.
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REFERENCES: 10M-53.A.1 Attachment 2-G - Issue IB, Rev. 1.
1LP-SOS-53.2 OBJECTIVE: 11 NUMBER: 1-97-073, M-0171
JTA i: .
K/A 6: 3.09.017.020.A3.01 (E09EA1.20) K/A IMPORTANCE: 3.6/3.8
l'inal Revision. Rev. 3A
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Which of the following automatic actions should occur if the Fuel Pool Bridge
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-Crane Radiation Monitor [RM-1RM-207), were to fail high while raising a spent !
fuel assembly out of the fuel transfer cart? '
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activity alarm (3) actuated. '
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(1) (2) (3)
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A. UPWARD Automatically IS
Stopped
, B. UPWARD and Automatically 'S
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C. UPWARD and Unaffected IS NOT. {
DOWNWARD 'i
D. UPWARD and Unaffected IS i
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DOWNWARD
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POINTS: 1.00
ANSWER: D.
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-REFERENCES: 10M-43.1.E . Issue 4, Rev. 4.
, ILP-SOS-43.1 OBJECTIVE: 6 NUMBER: 1-97-074, New j
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K/A 4: 3.09.072.000.K3.02 K/A IMPORTANCE: 3.1/3.5
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QuestionNumber 86
Which of the following lists of switch positions is the 100% power NSA switch I
lineup for the N31 Source Range Nuclear Instrument Drawer?
HV MANUAL LEVEL HIGH FLUX l
ON/OFF TRIP AT SHUTDOWN
A. NORMAL NORMAL BLOCK
B. HV OFF NORMAL BLOCK
C. HV OFF BYPASS BLOCK
D. NORMAL NORMAL NORMAL
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POINTS: 1.00
ANSWER: A.
REFERENCES: 10M-2.3.C - Issue 4, Rev. 2.
ILP-SOS-2.1 OBJECTIVE: 3 NUMBER: 1-97-075, New
JTA 9:
K/A 4: 3.09.000.032.EK2.01 K/A IMPORTANCE: 2.7/3.1
Final Revision. Rev. 3 A
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Question Number 87
N35, Intermediate Range Nuclear Instrument Channel I, compensating voltage is
set excessively HIGH. Following a Rx trip, N35 will indicate
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A. HIGH, preventing the source range from automatically energizing. j
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B. HIGH, and the source range will be energized when N36 is <P-6. ,
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C. LOW, and the source range will be energized when N35 is <P-6.
D. LOW, and the source range will be energized when N36 is <P-6. l
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POINTS: 1.00
ANSWER: D.
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REFERENCES: 10M-2.1.C - Issue 4, Rev. 1.
1LP-SOS-2.1 OBJECT [VE: 465 NUMBER: 1-97-076, M-0030
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K/A #: 3.09.000.033.EA2.11 (015K4.07) K/A IMPORTANCE: 3.1/3.4
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final Revision. Rev. 3A 4
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Question Number 88
Which of the following instrument failures would require entry into a Technical
Specification Action Statement?
A. The Auxiliary Feedwater Pump Turbine Exhaust Radiation Monitor fails
low during a plant cooldown with RCS temperature at 290*F.
B. The Waste Gas Decay Tank Hydrogen Monitor fails low during an RCS
degas operation.
C. The Pressurizer Surge Line Temperature instrument fails low during
steady-state operations at 100% power.
D. Demineralized Water Storage Tank (WT-1TK-26), level transmitter fails I
low during a steam plant heatup.
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POINTS: 1.00
ANSWER: A.
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REFERENCES : BVPS UNIT 1 TS 3.3.3.1, Table 3.3-6 2.c.v - Amendment No. 59 I
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ILP-SQS-43.1 OBJECTIVE: 7 NUMBER: 1-97-077, New
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JTA #:
K/A #: 3.09.000.061.G00 K/A IMPORTANCE: 2.6/3.3
Final Revision. Rev. 3 A
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Question Number 89
, With the Unit at 100% power and all systems in their at-power, NSA
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configuration, a tube leak has developed in the Non-Regenerative Heat Exchanger.
With no operator action, this will result in a RISE in CCR Surge Tank Level, and
the CCR Surge Tank
A. overflowing to the Auxiliary Building Sump. I
B. relief valve lifting and relieving to the Gaseous Waste Surge Tank.
C. level control valve [LCV-1CC-100A], dumping water back to the Primary
- Water Storage Tanks [BR-1TK-6A/B].
D. going off scale HIGH and closing Thermal Barrier Isolation Valves [TV-
ICC-107A,B &C). -
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ANSWER: A.
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, REFERENCES: 10M-15.1.C - Issue 4, Rev. 1
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ILP-SQS-15.1 OBJECTIVE: 3&9 NUMBER: 1-97-078, New
4
JfA 1:
K/A #: 3.10.000.000.A1.04 K/A IMPORTANCE: 3.1/3.2
Final Revision. Rev. 3 A
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QuestionNumber 90
Under which of the following conditions do the Emergency Operating Procedures
, allow an operating RCP to REMAIN running, when ALL CCR flow to that RCP is lost?
A. At ALL times, provided adequate HHSI flow AND seal injection flow can
be maintained. '
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B. During a SGTR, AETER the RCS depressurization has conunenced, if the
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RCS/ Highest SG D/P drops to <l50 psid.
C. During a response to Inadequate Core Cooling, if high pressure I
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injection flow AND an adequate heat sink CANNOT be established. ,
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D. When responding to a Loss of Emergency Coolant Recirculation, and ALL
safety injection flow is lost due to the depletion of the RWST.
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ANSWER: C. -
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- REFERENCES
- lOM-53A.l.2-C - Issue 1B, Rev. 2 (Note)
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, ILP-SOS-53.3 OBJECTIVE: 4 NUMBER: 1-97-060, New j
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JTA #:
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K/A #: 3.10.000.026.EK3.03 K/A IMPORTANCE: 4.0/4.2
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Final Revision, Rev. 3A
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Question Number 91
Given the following:
- The Unit is operating at 100% power with all systems in their at-
power, NSA configurations.
- Annunciators (Al-73), ' NEUTRON SHIELD EXPANSION TANK LEVEL LOW" and
[Al-50), *INCORE INSTRUMENT RM SUMP LEVEL HIGH" are received.
- Gross leakage is indicated by a rapidly dropping indication on the
Neutron Shield Expansion Tank Level Indicator (LIS-1NS-101).
What are the required actions?
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A. Perform a plant shutdown at a rate determined by the NSS/ANSS in
accordance with 10M-51.4, " Station Shutdown".
B. Perform a plant shutdown at a rato determined by the NSS/ANSS in
accordance with 10M-53C.4, AOP 1.51.1, " Emergency Shutdown".
C. Restore level to the normal range by operating [TV-1NS-101), Neutron
Shield Tank Makeup Valve.
D. Manually trip the reactor and proceed to 10M-53A.1, E-0 ' Reactor Trip
or Safety Injection."
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POINTS: 1.00
ANSWER: D.
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Provide the following references: Window (Al-50), 10M-9.4.AAB "Incore
Instrument Room Sump Level High" and Window [Al-73), 10M-15.4.AAD * Neutron
Shield Expansion Tank Level Low" ARP's.
REFERENCES : 10M-15. 4. AAD - Issue 4, Rev. 1.
ILP-SQS-15.1 OBJECTIVE: 10 NUMBER: 1-97-081, New
JTA i: s
K/A i: 3.10,000.026.G05 K/A IMPORTANCE: 3.3/3.4
. Final Revision, Rev. 3A
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QuestionNumber 92
Which of the following design features ensures K-eff remains less than or equal
to 0.95 in the spent fuel pool with irradiated fuel in the pool?
1. A minimum of 2000 ppm boron concentration in the pool.
2. The 'Boral' fuel rack installed neutron poison.
3. A minimum center-to-center distance between fuel assemblies.
A. I and 2.
B. 1 and 3.
C. 2 and 3.
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D. 1, 2 and 3.
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POINTS: 1.00
ANSWER: C.
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REFERENCES: lOM-20.1.B - Issue 4, Rev. 3.
lLP-SQS-20.1 OBJECTIVE: 167 NUMBER: 1-97-083, New
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JTA 4: -
K/A 6: 3.ll.033.000.K4.05 K/A IMPORTANCE: 3.1/3.3
-
Final Revision, Rev. 3A
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QuestionNumber 93
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Which of the following is a function of the Refueling Manipulator Crane?
A. Transfer fuel assemblies from the Reactor to the Fuel Transfer System
Upender.
B. Transfer fuel assemblies from the Reactor Containment to the Fuel
Handling Building.
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C. Provide the motive force to raise and lower the Burnable Poison Rod i
Assembly Handling Tool. '
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D. Provide the motive force to raise and lower the Reactor Upper ,
Internals.
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ANSWER: A. I
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REFERENCES: 1RP-10R-3.3 - Issue 0, Rev. O.
ILP-SOS-6.13 OBJECTIVE: 9 NUMBER: 1-97-084, New
JTA #:
K/A 9: 3.ll.034.000.G07 K/A IMPORTANCE: 2.5/3.0
Final Revision, Rev. 3A
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QuestionNumber 94
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Which of the following are sources of hydrogen that could accumulate in the
Waste Gas Disposal System?
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1. Cover gas on the Volume Control Tank.
2. Hydrogen gas in the Main Generator.
3. Gaseous Waste Disposal Blower Effluent.
4. Degasifier Gaseous Waste Charcoal Bed Effluent,
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A. I and 3.
B. 1 and 4.
C. 1, 3 and 4.
D. 2, 3 and 4.
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ANSWER: B.
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REFERENCES: 10M-19.1.C - Issue 4, Rev. O.
ILP-SOS-19.1 OBJECTIVE: 2 NUMBER: 1-97-006, New
JTA f:
K/A 8: 3.11.071.000.K5.03 K/A IMPORTANCE: 2.3/2.9
Final Revision. Rev. 3A
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- Question Number 95
When combating an electrical fire using foam, which of the following are
precautions to be exercised by the Fire Brigade members?
1. Anticipate and avoid the run off facm the electrical equipment being
sprayed.
2. Maintain a minimum distance of 15 feet from the electrical equipment
being sprayed.
3. Always wear rubber boots for electrical insulation.
4. Always use a MSA 401 SCBA when using foam.
A. 1 and 2 ONLY.
B. 2 and 3 ONLY.
C. 1, 3 and 4 ONLY.
D. 1, 2, 3, and 4.
POINTS: 1.00
ANSWER: A.
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REFERENCES : 10M-56.B.2 - Issue 4, Rev. O.
ILP-SQS-33.1 OBJECTIVE: 3 NUMBER: 1-97-087, New
JTA #: l
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K/A #: 3.11.086.000.K5.04 K/A IMPORTANCE: 2.9/3.5
Final Revision. Rev.3A
Question Number 96
What is the bases for the Technical Specification requirement that at least 23 l;
feet of water be maintained above the reactor pressure vessel flange during
refueling operations? !
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A. In the event of a fuel element rupture, the limits of 10 CFR 100 are
maintained.
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B. In the event one train of RER is lost, an adequate heat sink is :
available.
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C. The refueling operators can perform a full core off-load without
exceeding their 10 CFR 20 exposure limits.
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D. In the event of a fuel element rupture, 99% of the assumed 10% iodine ,
gap activity released is removed. l
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ANSWER: D.
REFERENCES: UnitkTechnicalSpecification3.9.10 Bases,AmendmentNo.175.
1LP-SQS-6.13 OBJECTIVE: 12 NUMBER: 1-97-088, New
JTA #: ,
K/A #: 3.11.000.036.004 K/A IMPORTANCE: 2.6/3.8
Final Revision. Rev. 3A
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QurstionNumber 97
Which of the following events is required to be reported to the NRC within 1
hour?
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A. A contract mechanic is found to be intoxicated inside the protected 1
area.
B. A planned radiological liquid release was performed and 2 days later
found to be in excess of 10CFR20 limits.
C. A Unit shutdown is commenced due to the INOPERABILITY of BOTH I
Emergency Diesel Generators. !
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D. A fire has occurred in the Site Engineering Building that took 13
minutes to extinguisa.
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ANSWER: C.
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Provide the following references: NPDAP 5.1, REPORT BEQUIREMENTS.
REFERENCES: NPDAP 5.1 Rev. 5.
1LP-SQS-40.1 OBJECTIVE: 19 NUMBER: 1-97-089, New
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JTA #: 1
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K/A 8: 3.11.000.059.G02 K/A IMPORTANCE: 2.6/3.9
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QuestionNumber 98
- Which of the following radiation monitor automatic actions are designed to
protect the health and safety of the general public if a highly radioactive
Waste Gas Decay Tank were to rupture?
A. Waste Gas Decay Tank Radiation Monitor [RM-IVS-106), will OPEN the
Main Filter Bank inlet damper and CLOSE the Main Filter Bank Bypass
B. Waste Gas Decay Tank Radiation Monitor [RM-IVS-106), will TRIP the
Leak Collection Area Exhaust Fans.
C. Aux Bldg Vent Gaseous Radiation Monitor [RM-IVS-102A), will TRIP the
Leak Collection Area AND Aux Bldg Exhaust fans.
D. Aux Bldg Vent Gaseous Radiation Monitor [RM-IVS-102A), will CLOSE the
Main Filter Bank inlet AND Bypass dampers.
POINTS: 1.00
ANSWER: A.
REFERENCES: 10M-43.5.B.2 - Issue 4, Rev. 1.
lLP-SOS-43.1 OBJECTIVE: 6 NUMBER: 1-97-090, New ;
JTA i:
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K/A is 3.11.000.060.EK2.02 K/A IMPORTANCE: 2.7/3.1
Find Revisiog Rnr 3A I
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Question Number 99
Which of the following is an advantage of using a straight hose stream over a
fog spray when fighting a large building fire?
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A. gets the water to the base of the fire before it can vaporize.
B. has a better heat absorption / cooling effect.
C. provides a better thermal shield to the firefighters.
D. uses less water thus minimizes the potential for a re-flash.
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ANSWER: A. l
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RE FERENCES : 10M-56B56.B.2 - Issue 4, Rev. O and LP 9339.
1LP-SOS-9339 OBJECTIVE: 20 NUMBER: 1-97-091, M-0154
JTA #:
K/A f: 3.11.000.067.EK1.02 (2.4.25) K/A IMPORTANCE: 2.9/3.9
Final Revision, Rev. 3A
QuestionNumber 100
Given the following:
A rapid (5%/ min) power reduction from 100% to 75% was performed due to
Grid instabilities.
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Power has been stable at 75% for seven hours.
The results from the RCS chemistry samples taken four hours after
power was stabilized at 75%, reveal the followings
e Dose Equivalent I-131 (DEI) is 97 uCi/gm.
e Gross coolant activity is 16 uCi/gm.
e The 100/E-Bar limit is 250 uCi/gm.
(E-Bar itself is 0.4 uCi/gm)
What is the status of the RCS activity levels and what actions are required to
comply with Technical Specifications?
A. RCS activity levels are within the LCO limits and NO actions
are necessary to comply with the Technical Specifications.
B. The DEI limit has been exceeded, power operation may continue
provided an isotopic analysis for iodine is performed every
four hours.
C. The DEI limit has been exceeded, be in Hot Standby with T,y.
<500*r within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D. The 100/E-Bar limit has been excteded, be in Hot Standby with
T... <500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
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ANSWER: C.
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Provide the following references: Technical Specification 3.4.8, Specific
Activity,
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REFERENCES : Unit 1 Technical Specification 3.4.8, Amendment No. 102. I
ILP-SOS-6.5 OBJECTIVE: 8 NUMBER: 1-97-092, New l
JTA 4:
K/A'#: 3.11.000.076.G07 K/A IMPORTANCE: 2.9/3.4
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Final Revisio:1. Rev. 3A
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Attachment 3
BV-2 WRITTEN EXAM COMMENTS AND NRC RESOULTION
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Question #9 !
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Facility Comment: The answer key to question #9 is incorrect as a result of a
typographical error. The correct answer should be "B" versus "A" as ,
indicated on the answer key. l
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NRC Resolution: Agree with facility comment in that the correct answer should be "B". l
Answer key will be changed accordingly. ;
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Attachment 4
SIMULATION FACILITY REPORT
Facility Licensee: Beaver Vallev Unit 1 and Unit 2 ;
Facility Docket Nos: 50-334 and 50-412 l
Operating Tests Administered from: March 17-21 and April 28-May 2.1997
This form is used only to report simulator observations. These observations do not
constitute audit or inspection findings and are not, without further verification and review, ;
indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC t
certification or approval of the simulation facility other than to provide information that !
may be used in future evaluations. No licensee action is required in response to these !
observations. i
No simulator deficiencies, that affected the scenario examinations, were identified for ;
either simulator during the conduct of the examinations. j
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