IR 05000334/1997003
| ML20149E258 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/14/1997 |
| From: | Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-334-97-03OL, 50-334-97-3OL, 50-412-97-03OL, 50-412-97-3OL, NUDOCS 9707180145 | |
| Download: ML20149E258 (2) | |
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July 14, 1997 NOTE T0: NRC Document Control Desk Mail Stop 0-5-D-24 FROM: Neos/ Oon /e v , Licensing Assistant Opethting Licen ng Branch, R F SUBJECT: OPERAT08LICENSINGEXAMINATIONADMINISJEREDON Maech 171711 . AT &ae a vnth~ d DOCKET #50-d3//-- g kNe tdeeFor J//7/97 Operator Licensing Examinations were administered attherefdreficed' facility. Attached, you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
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Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070.
b) As given operating examination.cdesignated for distribution under RIDS Code A070.
4 tem-#2- DannnattoH Re)orNtiems,Jiven WrMtPin=inat4M~
stt9ed, designated for-estr-itdir.n . h oin W e IE42, 4----- &c ~-> !
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9707180145 970714 PDR ADOCK 05000334 V_ PDR
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i i NOTE.TO: NRC Document Control Desk i Mail Stop 0-5-D-24 )
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FROM: botI o n le n . L' nsing Assistant ,
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Opera';1ngLicensirp" Branch, R
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i SUBJECT: OPERAT08LICENSINGEXAMINATJ,0NADMINISTEREDON l Rom / M 1997 . AT eenvete Mllu, f. i
! DOCKET #50-V/2. f Oh toy.k c[ f/97 Operator Licensing Examinations were administered atthereferenpedfacility. Attached, you will find the following ,
information for processir.g through NUDOCS and distribution to the NRC *
staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and initial exam submittal, f designated for distribution under RID 5 Ccde A070 b) As given operating examination, designated for distribution under. RIDS Code A070.
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DP W kb MAY 30, 1997
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. Mr. J. President Generation Group l Duquesne Light Company l Post Office Box 4'
l Shippingport, Pennsylvania 15077 I
i SUBJECT: BEAVER VALLEY UNIT 1 AND UNIT 2 SENIOR REACTOR OPERATOR l INITIAL EXAMINATION REPORT NOs. 50-334/97-03 (OL) AND i
l 50-412/97-03 (OL)
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Dear Mr. Cross:
This report transmits the findings of the senior reactor operator (SRO) licensing l examinations conducted by NRC examiners during the week of March 17 - 21,1997, and April 28 - May 2,1997, at the Beaver Valley Unit 1 and Unit 2 Nuclear Power Plants.
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Based on the results of the examinations for both units, all SRO applicants passed all i l portions of the examinations. At the conclusion of each examination, Mr. P. Bissett i j discussed the preliminary findings with you and other members of your staff.
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L These examinations addressed areas important to public health and safety and were !
developed and administered under Revision 7 to the Examiner Standards (NUREG-1021), ;
and the guidance of Generic Letter 95-06, " Changes in the Operator Licensing Program,"
which provided the opportunity for facilities to develop and, in part, conduct initial l. licensing examinations with NRC oversight. For these examinations, Beaver Valley Power l Station (BVPS) personnel developed all segments of the examinations, while the NRC provided oversight and final approval prior to the administration of the examinations. BVPS training personnel subsequently administered the NRC-approved written examinations, while the operating examinations were administered by the NRC.
l In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.
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! Mr. J. No reply to this letter is required, but should you have any questions regarding this examination, please contact me at 610-337-5211, or by E-mail at GWM@NRC. GOV.
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Sincerely, l
I Glenn W. Meyer, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety i
Docket Nos. 50-334; 50-412
Enclosure:
Initia! Examination Report No. 50-334/97-03 (OL) AND 50-412/97-03 (OL)
w/ Attachments 1 through 4 '
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
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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 l l D. Prawdzik, NRC Contract Examiner, LITCO Approved By: Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety I
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EXECUTIVE SUMMARY ,
Beaver Valley Units 1 and 2 Nuclear Power Plant inspection Report Nos. 50-334/97-03 and 50-412/97-03 Operations Three Unit 1 senior reactor operator (SRO) candidates (one instant and two upgrades), and six Unit 2 SRO candidates (two instant and four upgrades) were administered initial 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 conclu; ion of the Unit 1 examination. Following completion of the Unit 2 simulator examination, it was evident that corrective action had taken place in an effort to correct these identified weaknesses. The NRC examiners observed communications to be more direct, succinct, and that all crew members were kept wellinformed. Also, the candidates readily identified various equipment malfunctions purposely incorporated into the scenarios as control board awareness exercises. The f acility also placed an increased emphasis on the Unit 2 written examination development after previous comments from the NRC during their initial review of the Unit 1 written examination.
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Report Details I
1. Operations l 05 Operator Training and Qualifications l 05.1 Senior Reactor Operator Initial Examinations I
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a. Scope
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The examinations were prepared by Beaver Valley Power Station (BVPS) personnel l in accordance with the guidelines in Revision 7, Supplement 1, of NUREG-1021, i
" Examiner Standards," and Revision 5 of NUREG-BR-0122, " Examiners' Handbook i
for Developing Operator Licensing Written Examinations." The examiners l administered initial operating licensing examinations to three Unit 1 senior reactor i I
operator (SRO) candidates and six Unit 2 SRO candidates. Unit 1 had one instant SRO candidate and two SRO upgrade candidates. Unit 2 had two instant SRO candidates and four SRO upgrade candidates. The written examinations were I 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 / Fail Total Pass / Fail Written 9/0 9/0 Operating 9/0 9/O Overall 9/0 9/0 The written examinations, job performance measures (JPMs) and simulator scenarios for both Units 1 and 2 were developed by Beaver Valley Power Station (BVPS) representatives in accordance with generic letter guidelines, GL 95-06,
" Changes in the Operator Licensing Program." The exam development team was comprised of BVPS training and operation's representatives. Allindividuals signed onto a security agreement once the development of the examination commenced.
The NRC subsequently reviewed and validated, along with BVPS personnel, all 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 choice questions for both Unit 1 and Unit 2. There were no comments by either l the NRC or the utility concerning the validity of questions on the Unit 1 written I examination. For the Unit 2 examination, the answer key for one question was in error, and there was one question that was replaced during the administration of the written examination due to an administrative oversight. In this instance, one i
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question of the examination also had the answer to the question included. This 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.
The operating examinations were conducted from March 18 - 21,1997 and April 29 - May 2,1997. For both examinations, the operating examinations consisted of three simulator scenarios and ten JPMs for the instant SRO candidates 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 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 1 weaknesses based on the written examination were not identified. Based on the grading of the Unit 2 written examination, the following questions were missed by 1 more than half of the applicants, indicating a weakness in the general understanding of the subject area.
- 12 Recent changes to the independent verification program.
- 38 Thermodynamics associated with a leaking pressurizer PORV.
- 73 Prompt operator actions required on a loss of a vital 120vac bus.
- 84 Nuclear instrument control and instrument power functions and arrangement.
- 90 Electricalinterlocks associated with the circulating water system.
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 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 segment 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 upgrades, missed an administrative JPM concerning the review of a primary system leak rate calculation. Both
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l 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 occurred. All other administrative JPMs were performed acceptably by all candidates.
l JPM followup questions were improved upon from one exam to the next. A review 1
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of the Unit 1 followup questions indicated that several were memory and/or direct )
i lookup questions. This was brought to the attention of the training department !
i during the Unit 1 preparation week, and corrections and changes were made as I necessary to raise some questions more to the level of cognitive thinking. It was
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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 i l increased emphasis in this area following the identification of the NRC's concern.
c. Conclusions 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 examinations. Crew communications and control board awareness improved from l
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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 l job in adhering to the examiner standards and in developing the examination l materials needed to administer the examinations.
I 05.2 Steam Generator Tube Leakaae Procedure Deficiency a. Scope of Inspection The examiners reviewed a portion of procedure 20M-53C.4.2.6.4(ISS1 A),
Abnormal Operating Procedure (AOP) 2.6.4, Steam Generator Tube Leakage, i
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Revision 8, (hereafter AOP 2.6.4) during the Unit 2 examination preparation week l l (April 14-18,1997). The procedure was reviewed during the scenario verification l
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b. Observations and Findinas l 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 l tube leakrate that is increasing at more than 60 GPD per hour. The procedure did i 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 l the procedure's intent. The facility agreed that the procedure was not clear in its l
intent and that it would be revised to clarify it after the examination. The procedure l had been recently issued (April 9,1997), and the candidates had not used the new
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revision during training. The procedure had been revised to address EPRI document TR-104788, Primary to Secondary Leak Guidelines.
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i During the two performances of scenario 97-3, a tube leak of 700 gallons per day
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procedure AOP 2.6.4, was referenced by the candidates as expected. Hawever, one candidate initiated a normal shutdown and one candidate inhed an emergency shutdown. Both actions had been previously discussed and determined to be acceptable. Facility management had stated during tha examination I ( preparation week that the correct interpretation was that an existing tube leak was degrading and the type of shutdown (normal or emergency) was dependent on the i tube leakrate and/or the rate of the tube leakage increase.
The procedure had two statements concerning primary to secondary tube leakage i for one steam generator (SG), as follows: l The first statement was, "SG leakrate rate of rise equal to or greater than 60 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."
The second statement was, "SG leakrate 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 1 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."
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 l normal shutdown. The other candidato interpreted it as a leak that initiated from l 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 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 i the examination were considered acceptable to prevent a possible security compromise of the examination content.
i 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 UFSAR that related to the selected examination questions or topic areas. No 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 IV,y 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 i followup questions, and written examination questions developed at the SRO level of testing. The examiners also stated that significant improvements had occurred from 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 assistance 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 l 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 j David Kern, Senior Resident inspector, Beaver Valley Units 1 & 2 i 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 i
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BV-1 SRO WRITTEN EXAM W/ ANSWER KEY l
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RTL #A5.620.H . DUQUESNE LIGHT COMPANY Volume 3 i
Nuclear Power Division Pmcedure 5-5 Training Administration Manual Figure 5-5.1 Revision 10 Page 1 of 1 WRITTEN EXAMINATION COVER SHEET l PROGRAM: Initial Licensed OperatorTraining l
' CLASS NUMBER: 2-LOT-1 l
SUBJECT: Senior Reactor Operator, April 1997 - NRC Initial Licensed Operator Exam.
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By this signature,I state that all l of the work done on this examination l l is my own. I have neither given nor received aid. -
SIGNATURE 'DATE NAME ANSWER KEY DLC EMP #
(Please Pr'nt)
COMPANY (if other than DLC)
l POSSIBLE POINTS 100 SCORE
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Instructor Initials l
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i TRAINING DIRECTOR / SUPER OR PREPARED BY Dtvid C.Sibson/ Rich Bmoks AP OV
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SIGNATURE NV /[3 y-sr f JCd8 bek. A 71-3]
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Policies and Guidelines for Taking NRC Written Examinations Attachment 2 i
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.
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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
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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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l 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.
B. Written approval from the NSS. I C. Written approval from the plant manager / designee.
D. These interlocks are NEVER to be defeated. I ANSWER: A. Source: LOT - 0123 REFERENCES: 1/20M-48.3.B pg. B.1 Issue 3 Rev. 8 1/2LP-SQS-48.1 OBJECTIVE: 12 K/A 9: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9 1 I
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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 l
I C. Site Area Emergency
D. Genecal Emergency ANSWER: B. Source: LOT - 0126 :
a REFERENCESi EPP/I-3 pg. 7 Rev. 11 LP-EPP-57.81 OBJECTIVE: 1 !
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K/A #: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 l
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i (Nestion 2-97-3 i
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Which of the following iden*.ifies the correct sequence for racking out a 4160V breaker?
1. Open the 125 VDC control power breaker.
2. Check that the breaker is.open.
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.
I A. 5, 2, 4, 1, 3.
B. 2, 5, 1, 4, 3.
( C. 1, 2, 5, 4, 3.
D. ' 2, 1, 4, 3, S.
I ANSWER: D. Source: LOT - 0180 REFERENCES: 1/20M-36.4A pg. A1&A2 Issue 3 Rev. 4 2LP-SOS-36.1 OBJECTIVE: 9rr K/A 8: 194001.K1.07 K/A IMPORTANCE: 3.6/3.7 l
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BVPS Rev.2
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Questi:n 2 97-4
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An operator is assigned to perform a non-emergency e<olution in a High Radiation Area measuring 4 REM / hour. His accumulated dose for the current calendar year 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 B. 30 minutes C. 60 minutes D. 90 minutes ANSWER: B. Source: M-LOT - 0313 REFERENCES: BVNPD Directive 1.3.1 pg. 2 Rev. 2 2LP-SQS-GERT OBJECTIVE: 3-5 K/A #: 194001.K1.03 K/A IMPORTANCE: 2.8/3.4 i
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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. Channel verification.
B. Channel functional test.
C. Channel check. )
'D. Channel calibration. j ANSWER: C. Source: LOT - 0316 REFERENCES: TS Definitions t 2LP-SOS-TS OBJECTIVE: 7 j K/A #: 194001.A1.13 K/A IMPORTANCE: 4.3/4.1
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. Question 2-97-6
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Which o. the following statements describes the MINIMUM required Control Room staffing when the Unit is in Mode 67
1 A. One qualified SRO and one qualified RO must be in the Controls Area. ,
B. One qualified SRO must be in the Controls Area.
C. One qualified RO must be in the Control Room Area.'
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D. One qualified SRO must be in the Control Room Area.
! ANSWER: C. Source: LOT - 0323 REfIRENCES: 1/20M-48.1 Page 5 A.1 and Issue 3 Rev. 4 l T.S. 6.2 l
j. 1/2LP-SOS-48.1 OBJECTIVE: 2 K/A #: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9 l
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Question 2-97-7
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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 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 i
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Operations within 14 days of implementation.
ANSWER: C. Source: NEW REFERENCES: 1/20M-48.2B Pg. 6 item B.3 Issue 4 Rev. 9 1/2LP-SOS-48,1 OBJECTIVE: 11 K/A 8: 194001.Al.03 K/A IMPORTANCE: 2.5/3.4 l
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Question 2-97 8
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Under which of the following situations could the performance of an independent verification of a valve on a valve list be waived?
A. Independent verification would result in 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 independent verification.
D. The valve can be verified by a functional test using qualified instrumentation.
ANSWER: D. Source: New REFERENCES: 1/20M-48.3.D.VI.A.6 f 1ssue 3 Rev. 17 1/2LP-SQS-48.1 OBJECTIVE: 30 i K/A i: 194001.K1.01 K/A IMPORTANCE: 3.6/3.7 l
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r 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 thermal stresses, is manually placed on its backseat to allow for maintenance.
Assuming the backseating was the only action / work performed on the valve, prior i to returning the valve to OPERABLE status, the valve must be A. electrically stroked closed and open, one time.
B. electrically stroked closed and open, two times.
C. manually stroked closed and then electrically stroked open. l D. manually stroked closed and open and then electrically stroked closed and ^!
open. )
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j ANSWER: B. Source: LOT - 0308 l
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REFERENCES: 1/20M-48.3.D.VI.A.10.e pg. Issue 3 Rev. 16 D6 1/2LP-SOS-48.1 OBJECTIVE: 23 K/A 8: 194001.Kl.01 K/A IMPORTANCE: 3.6/3.7 i
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Ch:estion 2-97-10 i
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i In which of the following areas could one expect to receive a MAXIMUM dose of I 150 mrem in any one hour period? I l
1. Radiation Area '
2. High Radiation Area j 3. Locked High Radiation Area A. 1 Only.
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B. 2 Only.
C. 3 Only.
D. 2 and 3.
ANSWER: B. Source: M-LOT - 0482 REFERENCES : 10CFR 20.1003 pg. 286 I 2LP-SQS-GERT OBJECTIVE: 4-8 K/A f: 194001.K1.03 K/A IMPORTANCE: 2.8/3.4
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Page 10 BVPS . Rev. 2
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Question 2-97-11
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The LOWEST Emergency Classification that REQUIRES the implementation of Site t
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Accountability is a(n) (1) . This Accountability must be completed within (2) .
A. Alert; 30 minutes. -
B. Site Area Emergency; 30 minutes.
C. Alert; 60 minutes.
D. Site Area Emergency; 60 minutes.
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ANSWER: B. Source: LOT - 0723 l
REFERENCES: EPP/I-4 pg. 2 NOTE Rev. 11 l
.LP-EPP-57.81 , OBJECTIVE: 19
K/A 4: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 ;
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l Question 2 97-12 , .
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Which of the following statements is applicable to independent verifications?
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! -- l '. 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 i MAY be performed simultaneously under certain circumstances. >
l-l A. 1, 2 and 3.
l B. 1, 2 cnd 4.
C. 1, 3 and 4. !
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ANSWER: A. Source: LOT - 0804 ( REFERENCES: 1/20M-48.3.D.VI.A.6.c Issue 3 Rev. 17 l
I 1/2LP-SOS-48.1- OBJECTIVE: 30 K/A 8: 194001.K1.01 K/A IMPORTANCE: 3.6/3.7 l
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Questi:n 2-97-13
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When taking actions IAW 10Cm50.54 (x) that intentionally deviate from Technical taking the actions that Specifications, NRC notification is REQUIRED deviated from the Technical Specifications.
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 i 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 Issue 3 Rev. 9 REFERENCES: 1/ 20M-4 0.1. F.VI . A. 3. a
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1/2LP-SQS-48.1 OBJECTIVE: 10 K/A IMPORTANCE: 4.1/3.9 K/A 4: 194001.A1.02
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- Question 2 9714
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I The VOND marking "VDM" next to a motor-operated valve (MOV) indicates that the l
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MOV I
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A. may drif t open if the handwheel is engaged.
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B. may drift open if the motor is engaged.
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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 ,
. I REFERENCES: NPDAM 3.4 pg. 12 Rev. 6 ,
1/2LP-SOS-48.1 OBJECTIVE: 25 l
K/A i: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1 l
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- Question 2-97-15
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Which of the following describes the Operations Standard with regard to
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reactivity control during Mode 1 operations?
l l A. All operator initiated reactivity manipulations are to be announced to thn 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.
C. SRO approval for reactivity manipulations is only required for turbine
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load adjustments; normal rod and boron concentration adjustments may be l made without SRO approval.
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D. SRO approval for reactivity manipulations is only required for rod and l boron concentration adjustments; turbine load adjustments may be made
! without SRO approval.
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ANSWER: A. Source: M-LOT - 0894 RE FERENCES : Operations Standards A.14 Rev. 14 1/2LP-SQS-48.1 OBJECTIVE: 39 l
K/A #: 194001.A1.03 K/A IMPORTANCE: 2.5/3.4 i
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i- Question 2-97-16
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Use of a CAUTION tag is PROHIBITED for which of the following conditions?
A. Special' additional manual actions would be required to manipulate a
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component.
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B. Operation of a component will be affected because a portion of the system f f is NOT in NSA. .,
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C. As a temporary replacement for a component label that has fallen off. l 1 '
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D. As a warning that operation of a component will cause erratic indication.
ANSWER: C. Source: 1-97 Audit l
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REFERENCES: 1/20M-48.3.L.IV.A.II.C Issue 4 Rev. 2 1/2LP-SQS-48.1 OBJECTIVE: 15 K/A #: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1 i
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Page 16 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 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 minimum, whose authorization is necessary to move the boundaries of this clearance?
A. The electrician that assisted the Work Party Leader in performing the work under the clearance.
B. The Work Party Leader's direct supervisor.
C. The Outage Manager.
D. The Site Safety Engineer.
ANSWER: B. Source: New REFERENCES : NPDAP 3 4.IV.A.l.1 Rev. 6 1/2LP-SQS-48.1 OBJECTIVE: 33 K/A 8: 194001 Kl.02 K/A IMPORTANCE: 3.7/4.1 Page 17 BVPS Rev.2
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~ Question 2-9718
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Which of the following statements describes the control actions that occur 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 A. AUTOMATIC, outward rod motion is always prevented, but inward rod motion is possible.
B. AUTOMATIC, BOTH outward and inward rod motion is prevented.
C. MANUAL, outward rod motion is always prevented, but inward rod motion is possible.
D. MANUAL, BOTH outward and inward rod motion is prevented.
ANSWER: A. Source: Braidwood - 56 REFERENCES: 20M '.2.B, Page 8 Issue 4 Rev. 3
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2LP-SQS-1.3 OBJECTIVE: 10 K/A 9: 3.01.001.050.K4.01 K/A IMPORTANCE: 3.4/3.8
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l Qu stion 2-97-19
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Given the following: I i
e The Unit is operating at 1004 power with all systems in their at-power, l NSA configurations. !
- Annunciator (A4-3C] TAVG DEV FROM TREF is received.
- Computer point indicates TAVG is LOW. j You are directed to adjust RCS temperature using the Boration/ Dilution controls. l Determine the expected valve lineop to accomplish this task. ]
A. (2CHS*ECV113A] Boric acid flow control valve - OPEN (2CHS*FCV114A] Primary water flow control valve - CLOSED (. [2CHS*FCV113B] Makeup stop valve to the charging pump suction - CLOSED 1
[2CHS*FCV114B] Makeup stop valve to the VCT - OPEN .
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B. [2CHS*FCV113A) Boric acid flow control valve - OPEN l (2CHS*FCV114A] Primary water flow control valve - CLOSED 5
[2CHS*FCV113B] Makeup stop valve to the charging pump suction - OPEN l
[2CHS*FCV114B] Makeup stop valve to the VCT - CLOSED C. [2CHS*FCV113A] Boric acid flow control. valve - CLOSED ,
(2CHS*FCVil4A] 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 j l- D. [2CHS*FCV113A] Boric acid flow control valve - CLOSED
[2CHS*FCV114A] Primary water flow corttrol valve - OPEN
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! (2CHS*FCV113B] Makeup stop valve to the charging pump suction - CLOSED '
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[2CHS*FCV114B] Makeup stop valve to the VCT - OPEN l
ANSWER: D. Source: Modified 1-97-004
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REFERENCES: 20M-7.1.D Issue 4 Rev. 3 2LP-SOS-7.1 OBJECTIVE: 7
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K/A 6: 3.01.004.010.A4.03 K/A IMPORTANCE: 3.9/3.7
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l Question 2-97-20
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I The Unit is operating at 100% power with all systems in their at power, NSA configurations with the exception of RCS Letdown. The Excess Letdown HX is in
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service with flow to the VCT, with Normal Letdown secured. A Reactor trip and t Safety Injection are mancally actuated. PRIOR to resetting the Safety Injection l Signal, Excess Letdown Flow will be_ ;
l A. STOPPED, due to the CIA signal CLOSING the Excess Letdown flow control ,
valve [2CHS*HCV137). l l
B. STOPPED, due to the CIA signal CLOSING the containment seal return j isolation valves [2CHS*MOV378 and 381]. ;
i C. DIVERTED to the Primary Drains Transfer Tank [2DGS-TK21] due to the !
CIA signal positioning the Excess Letdown flow directing valve }
[2CHS*HCV389] to the [2DGS-TK21] position. ;
I 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 f
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20M-7.1.C
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REFERENCES: Issue 4 Rev. 3 l 2LP-SQS-7.1 OBJECTIVE: 4 K/A 8: 3.01.004.020.A1.00 K/A IMPORTANCE: 3.0/3.0 i
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i Question 2-97 21 !
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I The Unit is operating at 884 power with all systems in their at-power, NSA t
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configurations. The following indications are observed:
- Reactor power is RISING. ;
- T. is greater than T,.r.
- PZR PORV [2RCS*PCV455C) is OPEN.
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- ?ZR level is RISING.
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Which of the following would cause the above listed conditions to occur? [
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- A. An OT/AT Turbine Runback. i
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B. An Uncontrolled Rod Withdrawal.
i C. A Failed OPEN S/G Safety Valve. 1 D. Power Range channel N-44 failed HIGH. -l l
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I ANSWER: B. Source: Braidwood - 89 <
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REFERENCES: 20M-53C.4.2.1.3.B Issue 1A Rev. 2 ;
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2LP-SOS-53C.1 OBJECTIVE: 2 .i 4.0/4.2 [
K/A #: 3.01.000.001.EKl.06 K/A IMPORTANCE: t I
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Page 21 BVPS . Rev. 2
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- Question 2-97-22
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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
2OH-53C.4.2.3.5, Dropped RCCA? l
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1. Withdraw the Dropped Rod to the height of the remainder of the CBD ,
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2. Place the Rod Group Selector Switch in MANUAL. I C
3. Place the Rod Group Selector Switch in the CBD position.
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4. Reset the ROD CONTROL SYSTEM URGENT ALARM.
S. Align the CBD Lift Coil Disconnect Switches and Reset the CBD Group Step Counter and the P/A Converter to Zero.
i A. 2, 5, 1, 4.
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B. 2, 5, 4, 1. !
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C. 3, 5, 1, 4.
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D. 3, 5, 4, 1.
i ANSWER: C. Source: New e e
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REFERENCES: 20M-53C.4.2.1.5 Issue 1A Rev. 4 f
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2LP-SOS-53C.1 OBJECTIVE: 5 l,
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K/A #: 3.01.000.003.EA1.02 K/A IMPORTANCE: 3.6/3.4 !
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, Question 2-97-23 The Unit is operating at 40n power with all. systems in their at power, HSA configurations. An event occurs that causes the following indications:
- Annunciator (A4-8G) ROD POSITION DEVIATION is lit'.
- Annunciator [A4-9F) ROD AT BOTTOM is lit.
- RCS T. dropped f rom 559'r 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?
A. Restore T.,. to >551*F within 15 minutes or be in HOT STANDBY within the next 15 minutes.
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 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 />.
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.
ANSWER: D. Source: Braidwood - 116 REFERENCES: U2 TS 3.2.5 Amendment No. 51 2LP-SQS-TS OBJECTIVE: 1 K/A i: 3.01.000.003 G03 K/A IMPORTANCE: 3.3/3.8
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Q estion 2 97-24
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The Unit is operating at 65% power with all systems in their at-power, HSA 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 i received:
e (A4-8G) ROD POSITION DEVIATION e
(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. Source- Zion - 120 REFERENCES: 20M-2.4.AAD/AAK Issue 1/1 Rev. 2/2 2LP-SOS-1.3 OBJECTIVE: 15.1 K/A #: 3.01.000.005.EA2.01 K/A IMPORTANCE: 3.3/4.1
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- QuGstion 2 97-25
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l Given the'following:
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- An Automatic Reactor Trip has just occurred. '
e The Main Turbine did NOT trip automatically, e The Main Generator Output Breakers (PCB-352 and 362) are still CLOSED.
Which of the following describes the procedural action and beses required for this situation?
A. Open the Main Generator Output Breakers (PCB-352 and 362) to prevent ;
motoring the Main Generator. '
B. Open the Main Ge:.erator Output Breakers (PCB-352 and 362) to actuate l . an additional Main Turbine Trip.
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l C. Manually Trip the Main Turbine to prevent a Loss of Heat Sink.
I D. Manually Trip the Main Turbine to prevent an uncontrolled RCS
- cooldown.
ANSWER: D. Source: Braidwood - 97-REFERENCES : 2OM-53B.4.E-0 Issue IB Rev. 3 2LP-SOS-53A.1 OBJECTIVE: 3 '
K/A'8: 3.01.000.007.EKl.03 K/A IMPORTANCE: 3.7/4.0 ,
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Question 2-97 26
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Given the following:
- The Unit was operating at 75% power with all systems in their at-power, NSA configurations when the 'B' Main Feed Pump (2FWS-P21B],
tripped.
- The operators are responding to the transient IAW Annunciator (A6-10A]
. STM GEN FEEDPUMP 21A/B AUTO-STOP, ARP 20M-24.4. AAE and the EMERGENCY SHUTDOWN procedure, 2OH-53C.4.2.51.1.
e The Main Turbine is being run back at St/ minute.
e . Rod Control is in AUTOMATIC.
The following additional indications are observed:
e All S/G Levels are approximately 35% and RISING.
= Annunciator (A6-6A] MOIST SEP DRAINS RCVR TANK / PUMP TROUBLE is LIT.
e Annunciator (A4-BD] ROD CONTROL BANK D LOW is LIT.
e Annunciator (A4-9D] ROD CONTROL BANK D LOW LOW is LIT.
Based on the above indications, which of the following should be done FIRST?
A. Stop the EMERGENCY SHUTDOWN procedure, and commence a Normal Boratim 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 remove the Heater Drain Pumps from service.
C. Continue with the EMERGENCY SHUTDOWN procedure, and concurrently commence a tiarmal Boration per 20M-7.4.K, D1 ender Boration Operation.
D. Continue with the EMERGENCY SHUTDOWN procedure, and concurrently commence an Emergency Boration per 20M-7.4.Q, Emergency Boration.
ANSWER: D. Source: Braidwood - 78 REFERENCES: 20M-1.4.AAM Issue 4 Rev. 0
'2LP-SOS-1.3 OBJECTIVE: 17.f K/A i: 3.01.000.024.EK3.01 K/A IMPORTANCE: 4.1/4.4
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Question 2 97-27
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The Unit is operating at 100% power with all systems in their at-power, NSA 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.
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: 8 K/A 1: 3.01.000.029.EK2.06 K/A IMPORTANCE: 2.9/3.1 M d
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z-97-oz7A .
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Given the following conditions:
e The Unit is operating in Mode 5 with all systems in their required lineups for the existing mode.
- Annunciator.(A11-10D]-AUXILIARY BUILDING SUMP LEVEL HIGH went into-alarm 5 minutes ago followed shortly thereafter by Annunciator (A6-1H] PRI COMP COOLING MATER SYSTEM TROUBLE.
e The NCO notices that CCP surge tank level is slowly DROPPING.
Which of the following could be the location of the CCP leak?
A. RER HX B. CCP HX C. Excess L/D HX D. RCP Motor Cooler.
ANSWER: B. Source: North Anna - 65 REFERENCES: 2OH Figure 15-1 Rev. 7 2LP-SQS-15.1 OBJECTIVE: 2 and 13 K/A 8: 3.10.000.026.EA2.02 K/A IMPORTANCE: 2.9/3.6 l
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Qu:stion 2-97-28 .
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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. minimize the possibility of a PER PORV failing open or leaking excessively, i
C. prevent opening the PTR Code Safety Valves.
D. prevent exceeding the RCS subcooling limit assumed in the UFSAR accident analysis.
ANSWER: A. Source: Byron - 67
. REFERENCES: 20M-53B.4.FR-S.1 Issue IB Rev. 4 2LP-SOS-53A.1 OBJECTIVE: 3 K/A #: 3.01.000.029.EK3.12 K/A IMPORTANCE: 4.4/4.7 i
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l Question 2-97-29 ,
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l 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?
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Depress the (1) Pushbutton to reset the group step counters to zero steps !
ar.d (2) if any DRPI does not indicate zero steps. l A. (1) ALARM RESET; (2) depressing the ALARM RESET pushbutton will reset DRPI to zero j B. (1) STEP COUNTER RESET; (2) depress the ALARM RESET pushbutton C. (1) ALARM 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 I4C to calibrate DRPI ANSWER: D. Source: New REFERENCES: 20M-50.4.D pg. 10 Issue 1 Rev. 25 2LP-SQS-1.2 OBJECTIVE: 2.d.0 K/A 8: 3.01.014.000.A4.04 K/A IMPORTANCE: 2. 7 /2. ~1 l
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. Question 2-97 30 ;
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During a reactor startup, the ' ROD T.T BOTTOM" annunciator will be LIT when 1. CBA Rod Bottom Lights are OFF and CBB, CBC and CDD Rod Bottom Lights are i LIT.
2. CBA, CBB and CBC Rod Bottom Lights are OFF and CBD Rod Bottom Lights are ;
LIT.
3. CBA, CBC and CBD Rod Bottom Lights are OFF and ONLY 1 CBB Rod Bottom Light is LIT.
A. 1 and 2 l l B. I and 3 C. 2 and 3 l
D. 3 ONLY ANSWER: D. Source: Zion - 4 REFERENCES: 20M-1.4.AAA Issue 4 Rev. 1 2LP-SQS-1,4 OBJECTIVE: 0
- K/A i
- 3.01.014.000.K1.01 K/A IMPORTANCE: 3.2/3.4 +
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Question 2-97-31
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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 I
WITHOUT an accompanying Reactor. Trip?
i A. PZR Code Safety Valves.
B. PZR Spray Valves, i I
C. S/G Atmospheric Dump Valves. l t
'D. PZR Level Control System / Surge Volume.
ANSWER: A. Source: ' Zion - 63 REFERENCES: UFSAR 15.2-6&7 Rev. O
'2LP-SOS-6.5 OBJECTIVE: 3.c
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K/A #: 3.02.002.000.K4.10 K/A IMPORTANCE: 4.2/4.4 i
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Quistion 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 Tuor (*F) Teoto (*F) Tava ('F)
0430- 372 368 370 l
0500 362 358 360 0530 352 348 350 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?
A. Between 0500 and 0530.
B. Between 0530 and 0600.
- C. Between 0630 and 0700.
D. Between 0900 and 0930.
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ANSWER: C. Source: New REFERENCES: TS 3.5.3 Amendment No. - Original l 2LP-SOS-ll.1 OBJECTIVE: 10 ,
. 1 l K/A 8: 3.02.006.000.G05 K/A IMPORTANCE: 3.5/4.2 i
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Question 2-97-33
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Given the following:
- The Unit is operating at 100% power with all systems in their at-power, NSA configurations.
- The pressurizer level selector switch is in position 1, 459/460.
- (2RCS*LT459] PZR level transmitter, fails at 54%.
Assuming NO operator action is taken, which of the following describes the system response when plant load is REDUCED to 50%?
Charging flow _
A. AND actual PZR level will RISE. The Reactor will trip on high PZR level.
B. will rise AND actual PZR level will be maintained at 54%. The B/U heaters will energize, and the Reactor will NOT trip.
C. AND actual PZR level will DROP. At 14% actual PZP 'uvel, letdown will isolate and all PZR heaters will de-energize. -
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-SOS-6.4 OBJECTIVE: 14 K/A I: 3.02.000.028.EA2.02 K/A IMPORTANCE: 3.4/3.8 t
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' Question 2-97-34
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Given the following:
e The Unit is operating at 100% power with all systems in their at-power, HSA configurations.
- The PZR level selector switch is in position 1 459/460.
e (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?
PZR Level PZR Level VCT level Indicator Indicator Indicator l
2RCS*LI459 2RCS*LI460 2CHS*LIll2 l
A. RAISE DROP RAISE B. DROP RAISE RAISE l
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! C. RAISE DROP DROP D. DROP RAISE DROP
ANSWER: A. Source: Byron - 99 l
l REFERENCES : UFSAR Figure 7.3-16 Rev. 3 l
l 2LP-SOS-6.4 OBJECTIVE: 14 )
2.8/3.1
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K/A #: 3.02.000.028.EK1.01 K/A IMPORTANCE:
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Question 2-97-35
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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) .
t 4 Y (1) (2)
i A. INCREASE RISE B. REDUCE RISE
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C INCREASE DROP 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 CBJECTIVE: 6 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:
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- An automatic Safety Injection has occurred.
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- The operators are about to RESET the Safety Injection signal. i
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I which of the following describes why manual action may be required to restart i 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 l if a valid SI demand signal still exists. 1
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C. the EDG sequencer is completely disabled until the Rx Trip Breakers are CLOSED to remove the P-4 lockout.
D. the ONLY way to restart the Safeguards Equipment following a subsequent loss of offsite power is to manually initiate SI when the EDG sequencer has finished loading the blackout loads. I
ANSWER: B. Source: New REFERENCES: 20M-53B.4.E-3 pg. 52 Issue 1Bb Rev. 5 2LP-SOS-53A.1 OBJECTIVE: 4 K/A 8: 3.02.013.000.A4.02 K/A IMPORTANCE: 4.3/4.4 l
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. ' Question 2-97-37 ' . .
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r when performing a plant heatup and RCS pressure rises above the P-11 setpoint on
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(1) channels, the (2) Steam Line Preasure, Steam Line Isolation will be automatically (3) .
(1) (2) (3) i
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A. 1/3 Low unblocked )
B. 2/3 High Negative Rate . blocked ,
C. 2/3 Low blocked D. 1/3. High Negative Rate unblocked t ANSWER: B. Source: New Rev. 7 REFERENCES: FSAR Logic 7.3-12 !
2LP-SQS-1.1 OBJECTIVE: 5 K/A 4: 3.02.013.000.K4.03 K/A IMPORTANCE: 3.9/4.2 r
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Question 2-97-38
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The throttling process through a leaking Pressurizer 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 l
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steam at the PORV inlet and PRT pressure.
C. entropy process which could result in saturation or'superheated l
conditions at the PORV outlet dependent upon the entropy of the steam at the PORV inlet and the FRT pressure.
D. entropy process which always results in saturation conditions at the PORV outlet dependent only upon the PRT pressure. -
l ANSWER: B. Source: New i
i REFERENCES : Hollier Diagram Issue Rev.
1/2LP-NOMCD-1.1 OBJECTIVE: 8 ,
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K/A I: 3.03.000.008.EK3.02 K/A IMPORTANCE: 3.6/4.1 l
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- Question 2-97-39
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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. !
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- 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' = 4 0 % , 'B'=42%, 'C'=35%.
- All RCP's are running.
- RCS pressure is 1700 psig and slowly RISING.
,
- PZR level is 52% and RISING.
- Containment pressure is 12 psia.
- The operators have made the following EOP transitions:
- E-0, Rx Trip or Safety Injection, to
- E-1, Loss of Reactor or Secondary Coolant, and are currently at
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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 termination criteria are reached.
D. MANUALLY initiate Containment Spray and transition to FR-Z.1, Response to High Containment Pressure.
ANSWER: B. Source: Byron - 86 (Modified)
REFERENCES: 20M-53A.1.E-1 pg. 7 Issue IB Rev. 4
2LP-SOS-53A.1 . OBJECTIVE: 6 K/A 8: 3.03.000.009.EA2.34 K/A IMPORTANCE: '3.6/4.2 l
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Question 2-97-40
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The small break LOCA recovery procedure ES-1.2, " Post-LOCA Cooldown and l Depressurization" performs an evaluation to ' Check If An RCP Should Be Started.'
l This group of steps includes a check that ensures PZR level is greater than 14%
! prior to starting an RCP.
l The bases for this pressurizer level check ensures that when an RCP is started i-and the Rx vessel head void is collapsed A. sufficient RCS inventory exists to prevent PZR level from dropping off scale low.
I B. SI re-initiation criteria for RCS subcooling will NOT be met.
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l C. RCS pressure does NOT drop below 215 psig, thus maintaining RCP seal l integrity.
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l D. adequate PZR steam space is available to limit the RCS pressure rise, l
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ANSEER: Source: Byron - 86, Modified.
REFERENCES: 20M-53B.4.ES-1.2 Step 17 Issue 1B Rev. 4 ,
2LP-SOS-53A.1 OBJECTIVE: 3 l K/A f: 3.03.000.009.EK3.21 K/A IMPORTANCE: 4.2/4.5 l
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Qusstion 2-97-4I
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Given the following:
- The Unit was operating at 100% power with all systems in their at-power, NSA configurations when a Loss of Rx Coolant Accident occurred.
- A Safety Injection and CIB have automatically actuated.
e All equipment functioned as designed EXCEPT (2RSS*P21C & D) the 'C'
- and 'D' Recirculation Spray Pumps, failed to start and CANNOT be ,
restarted. j e The operators are in the Loss of Emergency Coolant Recirculation l procedure, ECA-1.1. '
e 90 minutes have elapsed since the Rx Trip and Safety Injection. 1 l
For the above listed conditions, the Emergency Core Cooling System flowpath and flow rate should be one Charging /HHSI pump providing flow through the I
A. INJECTION flowpath, at MAXIMUM rate.
B. INJECTION flowpath and one LHSI pump through its injection ;
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flowpath, both at MAXIMUM rate.
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C. CHARGING flowpath, throttled to -140 gpm.
D. INJECTION flowpath, throttled to ~200 gpm.
Provide the following reference: 2OM-53A.1.A-4.7 Issue 1B, Rev. 1, Minimum SI Flow After Trip.
ANSWER: D. Source: New REFERENCES: 20M-53A.1.ECA-1.1 Step 16 Issue 1B Rev. 4 and 21,
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, Question 2-97-42 i i
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l The Unit has experienced a large-break LOCA. The operators have just I transitioned to ES-1.3, " Transfer to Cold Leg Recirculation" due to the low l water level in the RWST. The following conditionis are reported by the STA af ter SI is RESET in ES-1.3: ]
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e A Core Cooling ORANGE path exists du<a to Core Exit Thermocouples l reading 84 5'F.
i e An Integrity RED Path exists due to RCS conditions to the LEFT of l
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j e LHSI Pump [2 SIS *P21B] has TRIPPE9 on overcurrent.
- All other systems are functioning as designed. i i
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What course of action should the operators take?
l l A. Immediately transition to FR-P.1, Response to Imminent Pressurized l l Thermal Shock, Step 1.
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i l B. Complete the alignment to Cold Leg Recirculation (steps 1-6 of ES- l
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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 Thermal Shock, Step 1.
J ANSWER: D. Source: Braidwood - 93 RE FERENCES : 1/20M-53B.2 pg. 8 Issue IB Rev. 2 2LP-SOS-53A.1 OBJECTIVE: 6 K/A i: 3.03.000.Oll.G12 K/A IMPORTANCE: 4.0/4.1 i
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Question 2-97-43 .
The Unit is operating at 834 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 per OM-6.4.IF, " Instrument Failure Procedure." The on-coming RO notes a concern during turnover that the P-ll bistable is NOT tripped because the bistable light on BB-B is NOT lit. ;
The off-going RO should.
A. check the P-11 B/S light bulbs. -If it is determined NOT to be a bulb problem, report the B/S tripping 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) .
D. inform the on-coming NCO that the F-11 bistable light is NOT required to be LIT for the current plar.t conditions.
ANSWER: D. Source: -Braidwood - 84 REFERENCES : 20M-6.4. IF pg. 21 Issue 4 Rev. 4 2LP-SCS-1.1 OBJECTIVE: 5 *
K/A 4: 3.03.000.027.EA2.16 K/A IMPORTANCE: 3.6/3.9
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Question 2 97 44
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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.
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PZR Pressure Protection Channel II (2RCS*PT456] subsequently fails [
HIGH.
Which of the following describes the actions required to be taken for the-subsequent PZR Pressure instrument failure?
A. Trip the bistables associated with (2RCS*PT456) IAW 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 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. 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 />.
l ANSWER: C. Source: Braidwood - 80 )
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l REFERENCES: U2 TS 3.3.1.1 & 3.0.3 Amendment No. 4 2LP-SQS-TS OBJECTIVE: 1 K/A f: 3.03.000.027.G03 K/A IMPORTANCE: 3.1/3.6 l
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QuGstion 2-97-45 i
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' The operators are evaluating a Steam Generator tube leak with the following plant parameters: -
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e (2CHS-FIl50] Letdown flow indicator reads 105 gpm. l
- e (2CHS*P21A] The 'A' charging pump is the only running charging pump. J e PRER level is STABLE.
- Seal injection and leakoff flows are NORMAL.
- Charging flow is 110 gpm.
- Identified RCS leakage is 0.9 gpe.
e (2CHS*FCV122] Charging Flow Control Valve is in AUTO.
Which of the following is the approximate amount of primary to secondary l leakage? '
)
A. 4 gpm. i B. 5 gpm.
C. 19 gpm.
D. 29 gpm.
ANSWER: C. Source: Byron - 91 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: 3.3/4.1
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l Question 2-97-46
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With normal letdown in service, which of the following methods should be
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I 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 764, isolate charging flow to allow PRZR level to drop to no less than 144, and continue to repeat this process.
2. Adjust charging flow as necessary to maintain PRZR level between 144 and 764, but maintain greater than 30 gpm charging flow.
3. Feed the ruptured S/G to 70%, minimize feed flow and' allow the ruptured S/G water level to drain to no less than 234, and-continus to repeat this process.
4. Maintain a relatively constant feed rate to the ruptured S/G such that the ruptured S/G water level is maintained at approximately 334.
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A. 1 and 3.
B. 2 and 3. j i
C. 1 and 4. j
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D. 2 and 4. I
l ANSWER: B. Source: New REFERENCES: 20M-53.B.4.ES-3.1 Steps 6-8 Issue IB Rev. 3 2LP-SQS-53A.1 OBJECTIVE: 6 K/A -#: 3.03.000.038.EA1.39 K/A IMPORTANCE: 3.6/3.7 i
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ChEstion 2-97-47 '
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When carrying out the actions of E-3, " Steam Generator Tube Rupture," an RCS cooldown is conducted prior to the initial RCS depressurization. i The temperature at which this initial RCS cooldown is terminated is based on the i i S/G pressure (s) and ensures.
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A. INTACT; the Technical Specification cooldown rate limit is NOT
); exceeded. g B. RUPTURED; the Technical Specification cooldown rate limit is NOT
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exceeded.
C. RU PTURED; that when the RCS is subsequently depressurized, adequate -
,
RCS subcooling is maintained.
D. INTACT; that when the RCS is subsequently depressurized, adequate RCS i subcooling is maintained. .
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I ANSWER: C. Source: New t
RE FERENCES : 20M-538.4.E-3 Step 15 Issue IB Rev. 5
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Question 2-97-48
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Given the following:
- The Unit was operating at 100% power with all systems in their at-power, NSA configurations.
- 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 HIGH.
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 ONE PZR PORV will OPEN resulting in a LOW PZR Pressure Rx Trip and SI.
D. ALL PZR heaters will turn OFF and BOTH P7R spray valves and ALL PZR PORVs will remain CLOSED.
ANSWER: C. Source: R-LOT - 0172 REFERENCES: 20M-6.4.IF pg. 23 Issue 4 Rev. 4 2LP-SQS-6.4 OBJECTIVE: 12 K/A #: 3.03.010.000.K6.01 K/A IMPORTANCE: 2.7/3.1 l
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Question 2 97-49 ;
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If.the Unit was operating at 1004 power and Grid frequency dropped to 59.5 .
Hertz, the Rx core Critical Heat Flux will and DNBR will . 1
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A. drop, drop -
B. drop, raise C. raise, dro.p ..
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D. raise, raise ANSWER: A. Source: North Anna - 21 REFERENCES: LP-THO-7 pg. 30 Rev. 6 l
2LP-TMO-7 OBJECTIVE: 11 & 12 K/A 8: 3.04.003.000.KS.01 K/A IMPORTANCE: 3.3/3.9 e
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i Question 2-97-50
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Given the following e There is a Small Break LOCA inside Containment.
- All systems responded as designed.
- All S/G Pressures are -1000 psig.
- RCS Pressure is 1220 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 whic'h 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.
ANSWER: B. Source: Braidwood - 52 REFERENCES: 20M-53A.I.E-0 Left Hand Pg. Issue 1B Rev. 3 2LP-SQS-53A.1 OBJECTIVE: 3 K/A #: 3.04.003.000.K6.04 K/A IMPORTANCE: 2.8/3.1 Page 50 BVPS - Rev. 2
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Question 2-97-51 .
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Given the following:
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e The Unit is operating in Mode 5 with the PZR water solid, e The 'A' Train of RHR is in service maintaining RCS temperature. i e All letdown orifice isolation valves are OPEN.
- The 'A' Train Charging Pump (2CHS*P21A1 is in service.
e Letdown Pressure Control Valve (2CHS-PCV145] is inadvertently CLOSED.
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
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pressure will RISE due to the resulting heatup.
B. The RCS Pressure will DROP due to total letdown flow being greater than charging flow.
C. RCS pressure will RISE due to continued charging flow until the
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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. ,
f ANSWER: C. Source: Byron - 37 REFERENCES: OM Figure 7-1A Issue Rev. 7 ,
2LP-SQS-7.1 OBJECTIVE: 4 '
K/A #: 3.04.005.000.K5.05 K/A IMPORTANCE: 2.7/3.1 t
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Question 2 97-52
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t l During plant operation with. reactor power at 854, the following events occur: [
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- Annunciator (A2-4r) " REACTOR COOLANT PUMP OIL TROUBLE" actuates, for the 21A RCP due to a low bearing oil reservoir level.
- Annunciator [A2-4E] "RCP MTR BRG TEMP-TR448B HIGH" actuates due to a l HIGH 21A RCP motor bearing temperature. ,
( e All CCP and seal injection flows to the RCPs are normal.
j e 21A RCP motor bearing temperature is 201*F and RISING at a rate of l 5'F/ min.
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Which of the following describes the required operator action?
A. Reduce Rx power to <30%, then STOP 21A RCP.
B. Reduce Rx power to <101, then STOP 21A RCP.
l C. Immediately STOP 21A RCP, then trip the Rx.
! D. Immediately Trip the Rx, thea STOP 21A RCP.
, Provide the following references: ARPs for (A2-4E) 20M-6.4.AAB & [A2-4F) 20M-l 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
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Question 2-97-53 ,
The Unit is operating at 100% power with all systems in their at-power, NSA
. configurations when the following annunciators are received:
e (A2-4D) REACTOR COOLANT. PUMPS SEAL TROUBLE; e Computer point reveals the 'A' Loop RCP (2RCS*P21A] to have a LOW seal 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 seal vent pot level.
The NCO reports the following additional information on the 'A' Loop RCP (2RCS*P21A]:
- No. I seal leakoff flow is 0.4 gpm.
- Seal water outlet temperature is 140*F and STABLE.
- Bearing outlet temperature is 145*F and STABLE.
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Based on the above information, which of the following evente has occurred to the 'A' Loop RCP (2RCS*P21A)?
A. 52 Seal has failed open.
B. #2 Seal has failed elosed.
C. Il Seal has failed open.
D. 51 Seal has failed closed.
ANSWER: A. Source: Zion - 123 REFERENCES: 20M-6.4.AAE.B/7.4.AAH.5 Issue 4/1 Rev. 7/16 2LP-SQS-6.3 OBJECTIVE: 4.e K/A 8: 3.04.000.015.EK2.07 K/A IMPORTANCE: 2.9/2.9 Page 53 BVPS - Rev. 2
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! Question 2-97 54 -
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Given the following:
- The Unit was at full power with all systems in their at-power, HSA configurations.
- An Autouatic Safety Injection occurred.
- The operators are in E-0, Rx Trip or Safety Injection Response, at ,
, Step 11, " Verify SI Status."
l * RCS pressure is 105 psig.
l e The following indications are observed on the 'A' Train LHSI Pump l [2 SIS *P21A]:
e The Pump Control Switch is in Auto. ;
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- Motor Amps are fluctuating between 18 and 24 amps.
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e LHSI Injection Flow Meter is fluctuating between 1800 and 3000 gPm.
The A Train LHSI Pump (2 SIS *P21A] is.
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l A. operating at runout conditions and the pump discharge valve (2 SIS *MOV8888A] should be throttled CLOSED.
B. operating as designed, the pump miniflow valve [2 SIS *MOV8890A] is I cycling due to the RCS being at saturation conditions. l
C. cavitating and the pump suction valve [2 SIS *MOV8809A] should be
]
verified OPEN.
]
D, cavitating and the Recire Pump Discharge to Safety Injection l Header valve (2 SIS *MOV8811A] should be verified OPEN.
ANSWER: C. Source: New REFERENCES: 20M Figure 11.1 Rev. 6 2LP-SOS-11.1 OBJECTIVE: 8 .
K/A 8: 3.04.000.025.EA1.09 K/A IMPORTANCE: 3.2/3.1 i
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' Question 2-97 55 i
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Given the following: }
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e The Unit was at 254 power with all systems in their at-power, NSA configurations.
e A loss of ALL Offsite AC Power has occurred.
- T. is $38*F.
- 7. 14 is at 535'F. '
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- - Tn.s is at 541*F.
- The average of the 10 hottest CET's is 548'F.
- PZR pressure is 2160 psig, i i
Which of the following is the current RCS subcooling?
A. 93'F. ;
B. 100*F. ,
C. 107'F. I i
D. 110'F.
t ANSWER: B. Source: Zion - 97 !
REFERENCES: Steam Tables i
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1/2LP-NOMCD-1.1 OBJECTIVE: 8 K/A #: 3.04.000.074.EK1.01- K/A IMPORTANCE: 4.3/4.7
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, . Question 2-97-56 Given the followings e The unit is operating at 60% power with ALL systema in their at-power, NSA configurations.
e Control bank 'D' rods are in AUTO at 190 steps.
The following indications are then observed:
- All S/G Steam Flows, Feed Flows and Water Levels RISE, and then return to their original values, e T. DROPS and then returns to its original value.
e- Rx power as indicated on Recorder (2NME-NR45) RISES and then returns to its original value.
e Control rods move OUT rapidly and then slowly step in to approximately their original positions.
Which of the following failures has occurred?
A. The Main Steam Pesidual Heat Release Valve (2SVS*HCV104] has failed OPEN.
B. The 64 Main Turbine Governor Valve (2TMS-GV4] has failed OPEN.
C. The HP Turbine Impulse Pressure Transmitter (2 MSS *PT446] has failed HIGH.
D. Power Range Nuclear Instrument channel (2NMP-NI44B) has failed LOW.
ANSWER: B. Source: Braidwood - 73 REFERENCES: LP-ATA-3.1 Rev. 0 2LP-SQS-ATA-3.1 OBJECTIVE: 'l& 2 K/A 0: 3.04.035.010.KS.01 K/A IMPORTANCE: 3.4/3.9 Page56 BVPS . Rev. 2
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. Question 2-97-57
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With the Unit operating at 100% power and all systems in their at-power, NSA ;
configurations, which of the following statements describes the response of a i main feedwater break as compared to a steamline break't-A. A feedline break will cause the affected S/G to depressurize BEEVRE the Rx Trips for a steamline break the affected S/G will depressurize AFTER the Rx Trip.
B. A feedline break will blowdown ALL S/Gs until the IW 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. continuously DROPS.
'D. A feedline break may result in indicated SGWL RISING or DROPPING, depending on the location of the break; for a stear.line break SGWL will always initially DROP.
ANSWER: C. Source: Byron - 69 L i
REFERENCES: ATA LP-4.1 s 4.3 Rev. O r
2LP-SOS-ATA-4.1 & 4.3 OBJECTIVE: 1 and 2 K/A #: 3.05.000.040.EA2.01 K/A IMPORTANCE: 4.2/4.7 I
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Questi:n 2-97-58
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Given the following:
- The Unit was operating at 100% power with all systems in their at-power, NSA configurations, when a Rx Trip and Safety Injection occurred.
- 'A' S/G pressure is DROPPING rapidly. -
- 'A' .S/G steam flow is 2.5E6 lbm/hr.
- RCS cold leg temperatures are 238'F and DROPPING.
- Containment Pressure is 8.5 psig and RISING.
- All MSIVs and Bypass Valves are CLOSED.
- Total AFW flow is 395 gpm.
- Highest reading Power Range instrument is 1.51 and DROPPING.
! Assuming all EST 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 D. Containment i.
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ANSWER: C. Source: Zion - 82 REFERENCES: 2OM-53A.l.E-0 Left Hand Issue 1B Rev. 3 Pg.
2LP-SOS-53A.1 OBJECTIVE: 6 K/A #: 3.05.000.040.EKl.06 K/A IMPORTANCE: 3.7/3.8 1 l
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Questi:n 2-97 59
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The Unit has experienced a Reactor Trip and Safety Injection due to a S/G Tube j 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:
e 'A' S/G Water Level is 65% Narrow Range and RISING.
- RCS T.,. is 540*F and stable. '
- 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.
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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 Fressure Mode and manually OPEN i the Steam Dumps.
B. Take the Steam Dumps to the Steam Pressure Mode, take BOTH 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].
Manually'OPEN the and S/G Atmospheric Steam Dump Valves I D. 'B' 'C'
[2SVS*PCV101B&C].
ANSWER: D. Source: Zion - 25 REFERENCES: 20M-53A.1.E-3.14.c RNO Issue 1B Rev. 5 ,
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2LP-SOS-53A.1 OBJECTIVE: 6 K/A 8: 3.05.000.051.EK3.01 K/A IMPORTANCE: 2.8/3.1
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Questi:n 2-97-60
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Given the following:
- The Unit is operating at 904 power with all systems in their at-power, NSA configurations.
l *- RCS T.., is 574*F and slowly rising on all 3 loops.
- RCS pressure is stable at 2235 psig, i
e Steam Flow on each S/G is 3.78 E6 lbm/hr.
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- 'C'. S/G feed flow is off scale HIGH.
l * 'C' S/G Main Feed Reg Valve (2fwS*FCV498] is full OPCN.
! * 'C' S/G level is DROPPING.
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- Containment. pressure and humidity are RISING.
Which of the following events is in progress? l l
A. 'C' S/G Main Feed Reg Valve (2FWS*FCV498} has failed OPEN.
l 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 Containment.
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ANSWER: C. Source: Braidwood - 108 i Rev. 2 l REFERENCES: 20M-53.B.4.E-2 pg. 3 Issue 1B I 2LP-SQS-24.1 OBJECTIVE: 5 1 I
I K/A #: 3.05.000.054.EKl.01 K/A IMPORTANCE: 4.1/4.3
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. . Question 2 9741 .
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The Unit is operating at 80% power with all systems in their at-power, NSA configurations. One condenser TCV steam dump valve fails full OPEN. Assuming a that NO operator action or automatic runback occurs, what will be the resulting Rx power level? ;
A. On l B. 80% !
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D. 900
,
ANSWER: C. Source: M-LOT - 0590
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REFERENCES: 20M-21.1.C pg. 5 Issue 4 Rev. 4
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2LP-SOS-21.1 OBJECTIVE: 5'
K/A #: 3.05.039.000.A2.05 K/A IMPORTANCE: 3.3/3.6 r
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Question 2-97-62 .
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During the performance of FR-S.1, " Response to Nuclear Power Generation - ATWS" immediate action step 4, 'Close Condenser Steam Dump Valves,' the operator places ONLY the Train 'A' Steam Dump Bypass Selector Switch to ihe 'OFF'
position (Train 'B' switch was left in the 'ON' position). Which of the following describes the Steam Dump system response?
l A. The first two banks of valves will CLOSE, but the *ast t+;o banks are
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still ARMED. l B. The last two banks of valves will CLOSE, but the first two banks are still ARMED.
C. ALL banks of valves CLOSE and are BLOCKED from actuating.
D. NO banks of valves CLOSE and ALL are ARMED for operation.
I I ANSWER: C. Source: LOT - 0023
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REFERENCES: 20M-21.5.A.12 Issue 4 Rev. O I 2LP-SQS-21.1 OBJECTIVE: 3 K/A #: 3.05.041.020.A4.08 K/A IMPORTANCE: 3.0/3.1 )
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Question 2-97-63 .
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The Unit is operating at 100% power with all systems in their at-power, NSA l configurations when an EHC equipment failure caused a rapid load rejection to-40% power. .
In response to this rapid load rejection, the Condensate Tcedwater 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 manually CLOSED after a four minute time delay.
C. on a C7B signal and will automatically CLOSE once normal pressure has ,
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-11 Rev. 7 2LP-SQS-22A.1 OBJECTIVE: 4 i
K/A 8: 3.05.056.000.Kl.03 K/A IMPORTANCE: 2.6/2.6 Page 63 BVPS - Rev. 2 i
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Question 2-97-64 .
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The Unit is operating at 954 power with all systems in their at-power, NSA t~
configurations. The selected steam pressure input to the 'B' S/G water level i control (SGWLC) system [2 MSS *PT485], fails LOW. Which of the following j 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
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design feature. -
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B. 'B' S/G Main Feed Reg Valve [2 WS*FCV488] throttles CLOSED to maintain 33% level. i
l C. 'B' S/G Main Feed Reg Valve [2WS * FCV488 ] initially throttles OPEN and l
then CLOSED to maintain 44% level.
D. 'B' S/G Main Feed Reg Valve [2FWS*FCV48d] initially throttles CLOSED and then OPEN to maintain 44% level.
ANSWER: D. Source: M-LOT - 0195 REFERENCES: 20M-24.1.D pg. 5 Issue 4 Rev. 2 f
2LP-SQS-24.1 OBJECTIVE: 5 K/A 4: 3.05.059.000.A2.ll K/A IMPORTANCE: 3.0/3.3 l
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k Question 2 97-65
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Given the following:
e The Unit was operating at 100% power with all systems in their at-power, NSA configurations when Grid instabilities caused a Unit Trip.
e All systems functioned as designed EXCEPT that an 'A' S/G AW flow control valve (2 WE*HCV100E] is mechanically bound in the full OPEN position and CANNOT be SHUT.
e 'A' S/G Wide Range Water Level is now 76% and RISING.
Which of the following should be done to prevent overfilling the 'A' S/G yet maintain an adequate Heat Sink?
1. Throttle CLOSED the other 'A' S/G A W flow control valve
[2 WE*HCV100F] .
2. STOP the 'A' Train Motor Driven AW Pump (2WE*P23A).
3. STOP the 'B'. Train Motor Driven AW Pump [2 WE*P23B). ,
4. STOP the Steam Driven AW Pump [2WE*P22) . ~
A. 1 ONLY.
B. 1, 2 and 4.
C. 2 and 3.
D. 3 and 4.
i ANSWER: B. Source: New REFERENCES: OM Figure 24-3 Rev. 7 2LP-SQS-24.1 OBJECTIVE: 7 K/A i: 3.05.0Gl.000.A2.07 K/A IMPORTANCE: 3.4/3.5 -
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Question 2-97-66
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Given the following: ]
- The' Unit is operating at 1004 power with all systems in their at- !
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power, NSA configurations EXCEPT that the 'A' Train Service Water Pump l
(2SWS*P21A] is being put on Clearance for motor replacement. ,
l e '[2SWS*P21A] 4KV breaker has been racked out and tagged OPEN.
The 'B' and 'C', Service Water Pumps (2SWS*P21B] and (2SWS*P21C] are
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e l' 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. ,
What are the Technical Specification implications, if any, for the Swing Service l
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Water Pump (2SWS*P21C) being powered from the wrong 4KV bus?
A. There are NO Technical Specification implications provided at least one Standby Service Water Subsystem is OPERABLE. i r
I B. Both trains of Service Water may be considered OPERABLE based l on the provisions of Technical Specification 3.0.5. j 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 />. j
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l D. Establish the 2AE 4KV bus as the power supply to (2SWS*P21C)
l. 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 />. .
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i Provide the following references: TSs 3.0.3, 3.0.5, 3.7.4.1 and 3.7.13.1.
l ANSWER: D. Source: New <
REFERENCES : TS 3.7.4.1 Amendment No. - Original 2LP-SQS-TS OBJECTIVE: 1 i
K/A 8: 3.05.076.000.G05 K/A IMPORTANCE: 2.8/3.2 l
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s Question 2-97-67
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Under which of the following conditions would the Containment Integrity Technical Specification be satisfied?
NOTE: ' Assume there are NO blank flanges or pipe caps installed on equipment that is disassembled or removed.
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A. e 21A S/G Blowdown irolation valve [2BDG*AOV101A1] is ,
REMOVED, . !
o 21A S/G Blowdown isolation valves (2BDG*AOV101A2] and ,
(2BDG'465] are CLOSED and e Drain valve (2BDG'547] is OPEN with the pipe cap removed.
B. e Containment Equipment Hatch installed with three closure bolts, and e BOTH doors on the Emergency Personnel Access Hatch are CLOSED, and e All containment purge dampers are CLOSED.
C. * CCP return header relief valve (2CCP*RV10S] is removed, e
CCP return header isolation valve (2CCP*HOV157-2] is CLOSED, and [2CCP'MOV157-11 is OPEN and ;
e Drain valve (2CCP*926] is OPEN with the pipe cap removed.
D. e The CVCS charging line is disconnected at the Regen Heat Exchanger inlet, o Charging Header vent valve (2CES-730] is locked OPEN, o Charging Header Isolation valve (2CHS*MOV289] is OPEN, and ,
e Charging Header Manual Isolation valves '[2CHS*30] and (2CHS*477] are CLOSED. ,
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Provide the following references: VOND Figures; 7-1A Rev. 6, 15-2 Rev. 8 and 25-1 Rev. 9.
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ANSWER: A. Source: M-1-97-057 REFERENCES: 20M Figures 7-1A, 15-3 and Rev. 6, 2 and 9 25-1.
i 2LP-SOS-TS OBJECTIVE: 1 l K/A f: 3.06.000.069.G08 K/A IMPORTANCE: 3.4/4.1 l
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, Question 2 97-68 J
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.
The Transfer to Recirculation signal is .aow present. Which of the following 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 Recircuir, tion will NOT occur because Recirculation Spray pumps (2RSS*P21A&C) are deenergized.
C. Transfer to Recirculation will occur on 'B' Train components ONLY J'
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 I (2RSS*MOV156A) being deenergized.
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ANSWER: C. Source: New REFERENCES: 20M-37.5.B.7 Table 37.7 Issue 4 Rev. 7
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pg.38-132 & Fig. 10080-RE-lC 2LP-SOS-11.1 OBJECTIVE: 6 and 11~
K/A 8: 3.06.026.000.K2.02 K/A IMPORTANCE: 2.7/2.9 j l
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Question 2-97-69 i -
A Large Break Loss of Coolant Accident :tas occurred inside containment.
The 'B' Train Hydrogen Analyzer (2HCS*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. l
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D. must be manually started from its local Control Panel when directed in the EOP network.
ANSWER: A. Source: New REFERENCES: 20M-46.1 pg. 1 Issue 1 Rev. 5 l
2LP-SQS-46.1 OBJECTIVE: 5 K/A 8: 3.06.020.000.A4.03 K/A IMPORTANCE: 3.1/3.3 l
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I Question 2 97-70
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Given the following conditions:
e The Unit is operating at 100% power with all systems in their at-power, NSA configurations, e A large, audible leak is reported at the equipment hatch seal into l containment.
Which of the following describes the action to be taken?
A. Quantify the leakrate to ensure the ma:timum allowable equipment hatch leakrate is NOT exceeded.
B. Reduce power to less than 50% until the leak is repaired.
C. Restore containment 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 i STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 1 I
( ANSWER: D. Source: North Anna - 53 REFERENCES: TS 3.6.1.1 Amendment No. 80 2LP-SQS-TS OBJECTIVE: 1 K/A 4: 3.06.103.000.K3.02 K/A IMPORTANCE: 3.8/4.2 l
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Page 70 DVPS . Rev. 2
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Question 2 97 71
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The Unit was operating at 100% power with all systems in their at-power, NSA 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-0.2 " Natural Circulation Cooldown." The following major action steps have been accomplished:
e Cold shutdown boron concentration has been verified,
At this point, RCS hot leg temperatures are checked to determine if they are less than 550*F.
What is the purpose of this RCS hot leg temperature check at this point in the procedure?
A. To determine if the RCS cooldown has resulted in steam void formation in the upper head of the reactor vessel.
B. To verify that natural circulation flow still exists between the core and the S/Gs.
C. To verify that the RCS cooldown has not resulted in a challenge to the RCS integrity critical safety function.
D. To ensure that at least 50*F RCS subcooling will be maintained during the subsequent RCS depressurization.
ANSWER: D. Source: New REFERENCES : 20M-53B.4.ES-0.2 Step 7 Issue IB Rev. 3 2LP-SOS-53A.1 OBJECTIVE: 3 K/A 8: 3.07.000.055.EA1.01 K/A IMPORTANCE: 3.7/3.9
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, Question 2-97-72 ,
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During the performance of ECA-0.0 * Loss of All AC Power", a rapid S/G l depressurization to 300 psig is performed to reduce RCS temperature and pressure.
l The bases for STOPPING the S/G depressurization at 300 psig is to ensure that I
- A. the maximum Technical Specification cooldown rate is NOT exceeded.
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l B. a steam void will NOT be created in the Rx vessel head.
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C. the challenge to the RCS Integrity Critical Safety' Function is limited i to that assumed in the Accident Analysis.
D. RCS pressure is maintained above the minimum pressure to preclude l injection of Safety Injection Accumulator N2 into the RCS.
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ANSWER: D. Cource: New i
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 1 I
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_ Qu:stion 2-97-73
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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
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Rx Trip?
A. Pressurizer Level.
B. Pressuriser Pressure.
J S/G Feed Flow.
C.
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D. Main Turbine Load.
ANSWER: C. Source: Byron - 75
REFERENCES: 2OH-38.4.V.pg. 1 Caution Issue 1 Rev. 5 i
2LP-SCS-30.1 OBJECTIVE: 8.h i i
K/A #: 3.07.0004057.EA1.06 K/A IMPORTANCE: 3.5/3.5 .
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,- Question 2-97-74
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The reactor is critical at 1.0 E" CPS in the source range. The 120 VAC Vital Bus II inverter output breaker tripped OPEN. This causes Vital Bus II to be doenergized.
A reactor trip will occur due to the loss of power to A. SSPS Train 'B' Logic Cabinet.
B. SR channel N-32.
C. IR channel N-35.
D. Rod Control low voltage power supply.
ANSWER: B. Source: E.>'... wood - 91 REFERENCES 20M-2.3.C pg. 4 Issue 4 Rev. 2 2LP-SOS-2.1 OBJECTIVE: 6 K/A 4: 3.07.000.057.EA2.19 K/A IMPORTANCE: 4.0/4.3 I
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Qtustion 2-97 75
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Given the following:
s The Unit was operating at 100t power with all systems in their at-power, NSA configurations EXCEPT; ;
- The 'A' Charging Pump (2CHS*P21A] is NOT running but its '
breaker IS racked onto the 2AE bus, j
e The 'C' Charging Pump (2CHS*P21C) is running on the 2AE bus. ;
e A loss of DC control power to the 2AE bus has occurred. i e While stabilizing the unit, a spurious S1 occurred. j Which of the following charging pump combinations will exist as a result of these failures?
[2CHS*P21A1 (2CHS*P21B1 (2CHS*P21C)
A. Stopped Running Stopped B. Stopped Running Running I C. Stopped Running Stopped '
D. Stopped Stopped Running
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ANSWER: B. Source: North Anna - 95 .
I RE FERENCES: (2CHS*P21C) 12241-E-5DM Rev. 11 j 2LP-SQS-7.1 OBJECTIVE: 7 K/A 1: 3.07.000.058.EA2.03 K/A IMPORTANCE: 3.5/3.9 l
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Question 2-97-76 ,
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1 The Unit is operating at 100% power with all systems in their at-power, NSA '
configurations. Under which of the following conditions would the System ]
i Station Service Transformer (SSST) Normal 4KV Breaker (ACB-42A] CLOSE? {
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l A. The Live Bus Transfer Switch is placed in the ON position with the l SSST Normal 4KV Breaker (ACB-42A) set up for auto transfer. l l
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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-l C.
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l D. The Live Bus Transfer Switch is placed in the ON position and the
'l control switch for [ACB-42A) is then placed in the CLOSE position.
ANSWER: D. Source: New
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REFERENCES: 2ON-36.1.E pg. 30 Issue 4 Rev. 4 2LP-SOS-36.1 OBJECTIVE: 6 4 I I K/A f: 3.0~1.062.000.K4.03 K/A IMPORTANCE: 2.8/3.1 l l
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QuIstion 2 97-77
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Given the following: ,
e Tho' 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. '
- A Loss of the DC SWBD 2-1 has occurred.
Based on the above information, which of the following actions will STOP the 2-1 i EDG7 1. Simultaneously depressing BOTH of the EDG STOP pushbuttons on BB-C.
l j 2. Depressing the EDG STOP pushbutton on the Local EDG control panel.
3. Placing the mechanical governor lever on the EDG fuel racks to the STOP position.
A. 1 and 3.
B. 2 and 3.
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C. 1, 2 and 3.
D. 3 ONLY.
ANSWER: D. Source: Zion - 3, Modified REFERENCES: OM Figure 36-3 & 12241-E- Revs. 9,7,0,0,2.
12H,J,K&L 2LP-SOS-36.1 OBJECTIVE: 6 K/A 0: 3.07.063.000.K3.01 K/A IMPORTANCE: 3.7/4.1 ,
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. Question 2-97-78
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l l Given the following:
, * The Unit is operating at 100% power with all systems in their at- i ( power, NSA configurations when a complete Loss of Offsite Power (, occurs, at 10:00:00.
e A spurious safety injection signal is generated at the same time.
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e The operators are responding to the transient IAW E-0 " Reactor Trip or Safety Injection."
l e At 10:02:10 the AFW pump status is checked; the turbine driven Afv pump [2fwE*P22) is running, but the motor-driven AEW pumps
[2fvE*P23A&B] are NOT running.
l The Motor Driven AfW Pumps (2fWE*P23A&B] should A. NOT be running, the EDG sequencers will NOT start the motor-driven AFW ,
l pumps for another 10 seconds.
B. NOT be running, the motor-driven AFW pumps will NOT start l automatically unless the turbine-driven AEW pump fails to start. I C. be running, the motor-driven AfW pumps should have started immediately upon the trip of the second main feedwater pump. ,
'l D. be running, the EDG sequencers should have already started the motor-driven AfW pumps.
ANSWER: D. Source: M-LOT - 0232 i
REFERENCES : 2OM-36.1.C pg. 14-15, 20M- Issue 4/4 Rev. 2/2 l 24.1.D pg. 16 and TS 3.3.2.1 Table 3.3.5 i 2LP-SQS-24.1 OBJECTIVE: 10 K/A 9: 3.07.064.000.K4.11 K/A IMPORTANCE: 3.5/4.0
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Question 2-97-79
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Under which of the-following situations should the station air to containment f instrument air cross tie valve [2IAC-MOV131) be directed to be OPENED? '
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' A. Only ONE Containment Instrument Air Compressor is operational.
I B. The norinal containment instrument air containment isolation valve (2IAC*MOV130) 1., failed CLOSED.
C. During the recovery from a Large Main Steam Line break inside ,
containment. I I
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 IB Rev. 5 2LP-SQS-53A.1 OBJECTIVE: 6 K/A 8: 3.08.000.065.EK3.04 K/A IMPORTANCE: '3.0/3.2 i
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Question 2-97-80
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Which of the following indications are available on the Emergency Shutdown Panel 6
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(SDP)?
i A. Auxiliary Feedwater Flow, Containment Pressure, Charging Flow.
B. RMS RX Outlet Temperature, S/G Wide Range Level, RCS Wide Range Temperature.
C. Pressurizer Level, Rx Trip Breaker Position, Steam Generator Pressure.
D. Letdown Flow, Intermediate Range SUR, Charging Header Pressure.
e ANSWER: B. Source: Braidwood - 92 REFERENCES : 20ST-45.2 Issue 1 Rev. 8 2LP-SOS-53C.1 OBJECTIVE: 6 j K/A 8: 3.08.000.068.EK2.01 K/A IMPORTANCE: 3.9/4.0
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Qu stion 2-97-81
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The Unit is operating at 1004 power with all systems in their at-power, NSA 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. !
(1) (2) (3)
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A. remain OFF remain OFF will NOT B. turn ON remain OFF will NOT C. turn ON turn ON will NOT D. turn ON turn ON will r
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-SOS-2.1 OBJECTIVE: 5 l
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K/A #: 3.09.000.032.EAl.01 K/A IMPORTANCE: 3.1/3.4 l
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Question 2-97-82
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A Technical Specification Action Statement entry would be required if the Unit is and Isc reports that A. at 84 powers the N-35 Hi Flux Trip Bistable setpoint is the current
equivalent of 351. j B. conducting a Rx startup with IR level at 1.0 E*' amps: ALL RCP '
underfrequency trip relaya were calibrated with a frequency meter that was out of calibration in the NON-conservative direction. I l
C. in Mode 1; BOTH Source Range instruments should be declared 1 inoperable due to the failure of the detector cables.
l in Mode 3; the Turbine Impulse Pressure Transmitter that feeds P-13 I D.
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 K/A #: 3.09.000.033.G08 K/A IMPORTANCE: 2.8/3.4 l
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I Question 2 97 83 .
The Unit is operating at 1006 power with all systems in their at-power, NSA configurations. What are the affects on the Solid State Protection System (SSPS) if the 120vac Vital Instrument Bus III is de-energized?
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A. ALL of the Train B, Output Bay Slave Relays will NOT function on a Safety Injection signal.
B. ONLY the #2 Emergency Diesel Generator Load Sequencer will load the !
required components on a Safety Injection signal. j
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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 2LP-SQS-1.2 OBJECTIVE: 7 K/A #: 3.09.012.000.Kl.01 K/A IMPORTANCE: 3.4/3.7 I
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Question 2-97 84 i
A reactor startup is in progress with IR power at 3.0 E-18 amps. The source range High Flux Trip has NOT been blocked. Which of the following describes the Reactor Protection System response if a CONTROL POWER fuse blows on the N-31
Source Range instrument with the Level Trip Bypass Switch in the positions indicated?
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Level Trip Bypass Switch Position
,1 i NORMAL BYPASS a i
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A. NO Trip HO Trip I
B. NO Trip Rx Trip I C. Rx Trip NO Trip D. Rx Trip Rx Trip ANSWER: D. Source: Braidwood - 17 REFERENCES: OM Figure 2-8 Issue 1 Rev. 3 2LP-SQS-2.1 OBJECTIVE: 3 K/A 5: 3.09.015.000.K6.04 K/A IMPORTANCE: 3.1/3.2 I
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e Qu:,stion 2-97-85
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Given the following:
e The Unit is operating at 1004 power with all systems in their at-L power, NSA configurations.
- RCS Tave control channels are indicating as follows:
e 'A' loop.- 575*F e 'B' loop - 577'F e 'C' loop - 576* F e 'B' loop Teoto instrument begins to SLOWLY fail LOW.
Which of the following describes the response of the Tave control System to this failure?
As the 'B' loop Tave drops, the selected Tavs will swap from 'C' loop to A. 'A' loop, then to 'B' , then finally back to 'C'
B. 'B' loop, then to 'A', then finally back to 'C'
C. 'B' loop, then to 'A' and remains there.
D. 'A' loop and remains there.
ANSWER: C. Source: M-North Anna - 44 REFERENCES: 20M-6.1.D pg. 18 Issue 4 Rev. 0 2LP-SOS-6.5 OBJECTIVE: 5.a ,
K/A #: 3.09.016.000.A3.01 K/A IMPORTANCE: 2.9/2.9
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l , Question 2-97-86
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l l Given the following:
- The Unit was operating at 100% power with all systems in their at-power, HSA configurations when a Rx trip and Safety Injection occurred, o Plant status is as follows:
- All RCPs are STOPPED.
= All Tex. are 567'T and slowly RISING.
- All Tcow, are 510'F and slowly DROPPING.
- The 5 highest CETs are 572*F and slowly RISING. ;
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- Pressurizer Level is 0%.
e Pressurizer Pressure is 1180 psig.
= All S/G Pressures are 1005 psig and STABLE.
Based on the above indications, what is the status of natural circulation flow and decay heat removal?
A. Single phase natural circulation flow is occurring and is adequately removing decay heat.
B. Two phase natural circulation flow is occurring and is adequately removing decay heat.
C. Single phase natural circulation flow is occurring but adequate decay heat removal is NOT occurring.
D. Natural circulation flow has stopped and adequate decay heat removal is NOT occurring.
Get Objective /LP I ANSWER: D. Source: New REFERENCES : 20M-53B.5.GI-4 Issue IB Rev. 0 2LP-SQS-53.2 OBJECTIVE: 12 K/A #: 3.09.017.020.A3.01 K/A IMPORTANCE: 3.6/3.8 i
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Question 2-97-87 .
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Unit 1 and Unit 2 are both at 1004 power when a HIGH alarm is received on
[2RMC*RQ201) and (2RMC*RQ2021, 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 CI4SE 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. ,
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A. 1, 2, and 3
B. 1, 2, and 4 ,
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C. 1, 3 ONLY.
D. 2, 3, and 4 !
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ANSWER: A. Source: M-LOT - 0509 REFERENCES: 20M-43.5.B.3 pg. 1 Issue 4 Rev. 0
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t I-Question 2-97-88
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l 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,
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Response to High Containment Radiation Level, have been met. Containment radiation levels are 80 mr/hr. What type of accident could.have caused this level of radiation to be in the containment?
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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 Specification S/G tube leakage. >
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! 3. Fuel cladding damage with NO indication of fuel melting.
4. Fuel cladding damage WITH indications of fuel melting. .
A. 1 Only.
B. 1 and 2.
C. 1 and 3.
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D. 1 and 4.
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l ANSWER B. Source: New l
REFERENCES: 20M-53B.4.FR-Z.3 pg. 1 Issue IB Rev. 1 2LP-SQS-53.A.1 OBJECTIVE: 5 l
K/A l': 3.09.073.000.A1.01 K/A IMPORTANCE: 3.2/3.5 l
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e Question 2-97 89 . t
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With the Unit operating at 100% power with all systems in their at-power, NSA configurations,'which of the following actions must be taken IAW the ARP for annunciator [A2-5F) REACTOR COOLANT PUMP COOLING WATER TROUBLE if the 2B RCP i'
thermal barrier CCP isolation valve (2CCP*AOV107B] CLOSED on a high flow signal?
A. Within 30 minutes, reduce power to <30% then trip the affected RCP.
B. Verify adequate RCP seal injection flow and continue power operations.
C. Inunediately Trip the Rx and then trip the affected RCP.
D. Declare [2CCP*AOV107B) INOPERABLE per T.S. 3.6.3.1 " Containment Isolation valves." i ANSWER: B. Source: Braidwood - 5 .
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REFERENCES: 20M-6.4.AAG pg. 5 Issue 4 Rev. 1 ,
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2LP-SQS-6.3 OBJECTIVE: 12 K/A #: 3.10.000.000.K3.01 K/A IMPORTANCE: 3.4/3.5 i
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Question 2-97-90
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Which of the following interlocks are provided to ensure a flowpath for service water return to the Circulating Water System?
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 time.
D. On one condenser half, no more than ONE out of TWO condenser waterbox outlet valves (2CWS-MOV100A/B(C/D)l can be CLOSED at any given time.
A::SWER: D. Source: New RE FERENCES: 20M-31.2 P&Lil8 Issue 4 Rev. 0 2LP-SOS-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 91
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Given the following: l l
- The Unit is in a refueling outago. ,
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o Core on-load is almost complete.
- The refueling crew 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. l l
i Based on the above information, the suspended fuel element should be-
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A. lowered into the fuel transfer cart and left on the FHB side. l 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 8: 3.ll.000.036.EA1.04 K/A IMPORTANCE: 3.1/3.7 i i
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, Question 2-97-92 i
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' Due to component leakage, an inam . % A.liological liquid release has just occurred. This release has resulted in the following radiation monitor l- readings:
- [2SWS-RQ-101] Component Cooling HX Rad Monitor at 5.73 E"' uCi/cc.
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- (2SWS-RQ-102) Component Cooling Service Water Rad Monitor at 6.03 E**
, uCi/cc.
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Assuming that these radiation monitor readings will be constant for the next two.
hours i
A. NO Emergency Plan Classification will be necessary.
- B. an Unv.sual Event should be declared Ite4EDIATELY.
C. an Unusual Event should be declared if an E-Plan Assessment CANNOT l
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be performed within 60 minutes.
i D. an Alert should be declared if an E-Plan Assessment CANNOT be.
performed within 15 minutes.
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Provide the following references: EPP classiliastion tabs.
ANSWER: C. Source: New i
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REFERENCES: EPP/I-lb Att. 1 Table ~1-1, P::; . 6
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K/A 4: 3.11.000.059.G02 K/A IMPORTANCE: 2.6/3.9
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Question 2 97-93 .
During sampling of the PZR vapor space, the outside containment isolation valve (2SSR*A0V112A2] developed a severe packing leak. A HIGH radiation alarm On the !
Leak Collection Vent Radiation Monitor [2RMR*RQI301] has resulted.
This alarm will result in A. NO automatic actions. The operators must manually CLOSE the Normal Leak Collection Dampers (2HVS* MOD 201A&B] to terminate the release. *
B. Normal Leak Collection Dampers (2HVS* MOD 201A&B] CLOSING and Filtered Leak Collection Dampers (2HVS*MQD202A&B] OPENING. ,
C. Containment Purge Diverting Dampers (2HVR* MOD 21&22] swapping to the Filtered Release Path.
D. Normal Exhaust Fans (2HVS*FN263A&B] STOPPING and the associated fan discharge dampers CLOSING.
ANSWER: B. Source: New REFERENCES: 00:4-43.5.B.3 Issue 4 Rev. 0 2LP-SQS-43.1 OBJECTIVE: 5 .j K/A 8: 3.11.000.060.EA2.05 K/A IMPORTANCE: 3 . ~1/ 4 . 2 l
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- Qutstion 2 97 94 l
What indication is available to alert personnel that a CO2 discharge is imminent !
inside of a protected zone ~t A. A red revolving light inside the zone.
B. A pre-discharge horn sounding inside the zone. ,
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C. A wintergreen odorizer floods the zone prior to the discharge.
.D. The announcement from Security over the Page Party System.
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t ANSWER: B. Source: SOS - 1165 REFERENCES: 20M.33.1.B Issue 4 Rev. 2 2LP-SQS-33.1 OBJECTIVE: 4.d K/A #: 3.11.000.067.G05 K/A IMPORTANCE: 3.4/3.8 l
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Qutstion 2-97-95
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Technical Specification 3.4.8.1 - RCS Specific Activity, action statement 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'F?
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.
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B. To ensure additional iodine spiking will EOT occur due to the reduced thermal energy in the fuel rod gas volume.
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 Roccamendation (PAR)
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 l 2LP-SOS-TS OBJECTIVE: 4 K/A f: 3.11.000.076.G04 K/A IMPORTANCE: 2.1/3.7 l
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- , Question 2-97-96 i
Given the following: ,
e The Unit is in mode 6 for a refueling outage, e Off-load of fuel is SSL complete and ongoing.
l * Containment Purge and Exhaust is in service.
! e IEC has just reported that the current HI setpoints for the l Containment Purge Radiation Monitor (2HVR*RQIl04A1 was incorrectly set two decades HIGH.
What action should be directed based on this information?
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A. Suspend core off-load until containment atmosphere grab samples can be
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obtained and double verified <MPC.
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are declared OPERABLE. r i 1 l- C. Continue core off-load and direct HP to perform continuous air )
monitoring of the containment.
I D. Continue core off-load and verify purge exhaust is directed through l
!- the Main Filter Bank.
ANSWER: B. Source: North Anna - 46 REFERENCES: TS 3.9.9 Amendment No. - Original ;
l 2LP-SOS-TS OBJECTIVE: 1 !
l K/A i: 3.11.029.000.G11 K/A IMPORTANCE: 2.8/3.5
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Question 2-97-97 , ,
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' The suction piping of the spent fuel pool cooling pumps (2fWC*P21AEB] has ruptured and CANNOT be isolated. Which of the following, by design, is the LOWEST spent fuel pool level that could result?
A. 10 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 :
REFERENCES: 20M-20.1.B pg. 3 Issue 4 Rev. 1 2LP-SQS-20.1 OBJECTIVE: 1 K/A 6: 3.11.033.000.K4.01 K/A IMPORTANCE: 2.9/3.2 i
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. Question 2-97-98 I
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Which of the following manipulator crane features helps to prevent lif ting a l fuel assembly with excessive force? !
l A. Dillon load cell circuit.
B. Gripper interlock circuit.
C. Bridge - trolley interlock.
D. Slack cable limiting circuit.
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ANSWER: A. Sources R-LOT - 0206 l
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! REFERENCES: Refueling Manual Issue Rev.
2LP-SOS-6.12 OBJECTIVE: 5.c K/A f: 3.11.034.000.K6.01 K/A IMPORTANCE: 2.1/3.0 l
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t Question 2-97-99
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During a Liquid Waste Discharge, flow control valve (2SGC-HCV100) is in manual and controlling flow at 30 gym. A High Radiation Alarm is received on the Liquid Waste Process Effluent Radiation Monitor (2SGC-RQ100). Which of the following explains the effect that this Radiation Alarm will have on the Liquid Waste Discharge?
A. The discharge will be terminated imediately 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 1 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.
l ANSWER: B. Source: M-SQS - 0608 )
REFERENCES: 20M-25.1.D Pg. 10 Issue 4 Rev. 0 2LP-SOS-17.1 OBJECTIVE: 5 K/A 8: 3.11.068.000.A3.02 K/A IMPORTANCE: 3.6/3.6 I
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Question 2-97-100
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A. The Auxiliary Building Supply and Exhaust Fans will automatically
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. B. A flammable stixture of radioactive gases will form in the Unit 2 Auxiliary Building.
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C. The release of gas to the environment will be terminated by trip valve ,
3 (2GWS-AOV105).
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1 D. The surge tank will relieve via relief valve (2GWS-RV101] to the Unit ,
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1 waste gas relief header. 1
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. ANSWER: D. Source: R-LOT - 0201
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a REFERENCES: 20M-19.1.B pg. 2 Issue 4 Rev. 0
~ 2LP-SQS-19.1 * OBJECTIVE: 2 K/A 9: 3.11.071.000.K4.01 K/A IMPORTANCE: 2.6/3.0
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l Attachment 2 -
l BV-2 SRO WRITTEN EXAM W/ ANSWER KEY l
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, RTL #A5.620.H DUQUESNE LIGHT COMPANY Volume 3
! Nut. lear Power Division Procedure 5-5 i Training Administration Manual Figust 5-5.1 i l
Revision 10 Page1 of1 WRTTTEN EXAMINATION COVER SHEET l
l PROGRAM: Initial Licensed Operator Training CLASS NUMBER: 1-LOT-2 SUBJECT: Senior Reactor Operator, March 1997 - NRC Initial Licensed Operator Exam.
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By this signature, I state that all of the work done on this examination l
is my own. I have neither given nor received aid.
l SIGNATURE DATE March 17,1997 l
NAME ANSWER KEY DLC EMP #
(Please Print)
COMPANY (if other than DLC)
POSSIBLE POINTS 100 SCORE Instructor Initials
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TRAINING DIRECTOR / SUPERVISOR PREPARED BY David C. Gih APPROVAL SIGNATURE /3 M 3//3/77 l
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Date
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I ( ES-402 Policies and Guidelines Attachment 2 for Taking NRC Written Examinations ,
1. Cheating on the examination will result in a denial of your application and could result in more severe penalties.
2. After you complete the examination, sign the statement on the cover sheet 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. l 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 I the answer sheet. !
8. Mark your answers on the answer sheet provided and do not leave any !
question blank. 1 (' 9. If the intent of a question is unclear, ask questions of the examiner only.
i 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 l 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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I k. Examiner Standards 5 of 6 Rev. 7, January 1993
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Qu:stionNumber I
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Select the statement that describes the safety equipment REQUIRED to be worn when handling sodium hydroxide (NaOH) .
A. Safety glasses and appropriate gloves ONLY.
B. Safety glasses, appropriate gloves and a paper surgical mask.
C. Goggles /faceshield and impervious clothing ONLY.
D. Goggles /faceshield, appropriate gloves and impervious clothing.
POINTS: 1.00 j
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ANSWER: D.
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REFERENCES: AOP 1/2 53.C.4A.75.7, Att. 1, page 12 - Issue 1A, Rev. 6.
.i 2LP-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. 3A
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Qu:stionNumber 2 An MOV designated as "VDM", (valve drifts manually), has been closed manually for a clearance, .
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Which of the following actions is required'to prevent this valve from drifting due to system pressure effects?
A. Ensure that the MOV manual' operating lever remains in the fully
" ENGAGED" position following manual closure. .
B. Manually return the MOV manual operating lever to the " DISENGAGED" position following manual closure.
C. Restore power to the MOV following manual closure and leave the control in the " NEUTRAL" position.
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: MP ,J a. , nev. 6,'.aga 12, item o.
ILP-SOS-48.1 OBJECTIVE: 23 NUMBER: 1- 97-094, M-0121 JTA i: -
K/A #: 194001.K1.01 K/A IMPORTANCE: 3.6/3.7 i
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i Question Number 3
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l- Given the followings i
l * RCS pressure is 225 psig.
j e RCS temperature is at 210*F.
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- Chemistry has just called in the most recent RCS sample results.
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Which of the following would exceed the Technical Specification limit for the
RCS transient chemistry specifications?
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h s J .A. Fluoride = 1.0 ppm. ?
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B. Chloride = 1.6 ppm.
j -: -i i C. Dose Equivalent Iodine - 131 - 0.5 uC1/gm.
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! D. Dissolved Oxygen = 1.1 ppm.
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ANSWER: B.
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Provide the following references: Unit 1 Technical Specification 3.4.7, Chemistry, and 3.4.8 Specific Activity.
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 4:
K/A 4: 194001.A1.14 K/A IMPORTANCE: 2.5/2.9
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- QuestionNumber 4 ,
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, An' operator is required to be continuously stationed at a valve in a confined i area for 30 minutes. Radioactive material on/in this valve is exposing the 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. i
The operator.will exceed their 10 CFR 20 dose' limits.
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C. The area must be posted as an High Radiation Area.and'the' operator should have an integrating dose rate meter. -
D. An HP technician should be present to monitor the radiation in the i room with a portable neutron meter while the operator is stationed at-
-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. '
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1/2LP-RC-02 Rev. 17 OBJECTIVE: 4-9 NUMBER: 1-97-096, M-0165 JTA #: '
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. K/A.ft 194001 K1.03 K/A IMPORTANCE: 2.8/3.4 >
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Qurstion Number 5
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Which of the following are required for a confined. space entry?
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1. A minimum of.two qualified individuals shall enter the confined space,
- one of which will act as only a safety man.
i 2. A method of connunication 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 i
ventilation flow can be increased as an additional precaution.
A. 1, 2, and 3. I B. 1, 2, and 4.
C. 2, 3,.and 4 i
D. 2 and 3 ONLY. j
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REFERENCES: NGAM 3.7 1/2LP-GM-6040 OBJECTIVE: 7 NUMBER: 1-97-097, M-0173
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JTA 6:
K/A f: 194001.Kl.13- K/A IMPORTANCE: 3.3/3.6 Final Revision, Rev. 3A
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Question Number 6 r
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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 will 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, '
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ANSWER: A. ;
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REFERENCES : SAP 41, pages 18 and 19, Rev. 6.
ILP-SQS-48.1 OBJECTIVE: 30 NUMBER: 1-97-098, M-0181 JTA #:
K/A f: 194001.K1.02 K/A IMPORTANCE: 3.7/4.1 l Final Revision Rev.3A
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- Qu
- stionNumber 7 l
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- During the performance of an Operating Surveillance Test (OST) that takes several days to finish, which of the following are required? l l
l 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 />.
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.
4 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 .
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REFERENCES: 1/20M 48.2.C, issue 3, Rev. 17 page CB of 19 item 14c 1/2LP-SOS-48,1 OBJECTIVE: 10 NUMBER: 1-97-110, LOT - 0726 JTA 8:
K/A #: 194001.A1.01 K/A IMPORTANCE: 3.1/3.2 Final Revision, Rev. 3A
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QuestisnNumber 8
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Who must approve an on the spot change (OMCN) to an operating procedure?
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A. The respective department supervisor of the individual requesting the ,
change ONLY.
B. TWO members of the plant management staff,'ONE of whom must hold an i 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 POTH of the on-shift licensed -
SROs for the affected Unit.
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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 JTA #:
K/A #2 194001.A1.01 K/A IMPORTANCE: 3.3/3.4 l
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. Final Revision. Rev. 3A i
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Question Number 9 l' When performing a station startup IAN ON Chapter 50 " STATION STARTUP," steps 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 omitted by the NSS provided the UOM initials the omitted step.
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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- POINTS: 1.00 ANSWER: C.
REFERENCES: 10M48.2.C - Issue 3, Rev. 13.
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ILP-SOS-48.1 OBJECTIVE: 10 NUMBER: 1-97-101, M-0386
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- JTA i
l i K/A i: 194001.A1.02 K/A IMPORTANCE: 4.1/3.9 l
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QuestionNumber 10
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Which of the following.is the reason why a nitrogen blanket is maintained on the Pressurizer Relief Tank (PRT)?
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A. To provide an additional volume of gas to ensure that the PRT Rupture
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Disc relieves at the design pressure.
B. To previde a driving force for the sampling system. !
C. To reduce the potential for an explosive mixture of hydrogen and 3 oxygen.
D. To ensure the water volume in the PRT remains subcooled when a PZR ,
PORV or Safety Valve lifts.
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I POINTS: 1.00 ANSWER: C.
REFERENCES: 10M1.6.1.C - Issue 4, Rev. 1.
ILP-SOS-6.4 OBJECTIVE: 7 NUMBER: 1-97-102, M-0661
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JTA 4:
K/A 4: 194001.Kl.15 K/A IMPORTANCE: 3.4/3.8 Final Revision. Rev.3A
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- Question Number 11 l
Which of the following conditions MUST be met to allow a non-licensed person to i manipulate the control rods?
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1. Under the direct supervision of a licensed operator or SRO.
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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 .
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A. 1,2, and 3. (
B. 1,2, and 4. ,
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C. 1,3, and 4.
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POINTS: 1.00 ANSWER: A.
1 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 l
, i JTA #: i K/A 8: '194001.A1.09 K/A IMPORTANCE: 2.7/3.9 !
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(Nestion Numtxt 12 ;
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.
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' C. A simultaneous loss of ALL annunciators, sequence of events recorders, and SPDS for >15 minutes in Mode 5. l D. A' report by plant personnel of a chlorine gas release within the site perimeter that render the chlorine building inaccessible.
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POINTS: 1.00 !
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ANSWER: B. .
Provide the following references: EPP/IP Tab 2.2.
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REFERENCES : EPP Tab 2.2.
i 1/2LP-EPP-57.81 OBJECTIVE: 1& 11 NUMBER: 1-97-104, M-6200 '
JTA #:
K/A #: 194001.A1.16 K/A IMPORTANCE: 3.1/4.4 '
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Final Revision, Rev.3A
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Question Number 13 ;
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Which of the following approved portable fire fighting equipment should be used !
to combat a flammable liquid fire?
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1. Dry Chemical.
2. Water.
a 3. CO2 4. Foam.
A. 1, 2 or 3. l
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B. 1, 3 or 4.
C. 1, 2 or 4.
s D. 2, 3 or 4.
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POINTS: 1.00 ANSWER: B.
REFERENCES: 1/2.56A.4.H - Issue 3, Rev. 1, and LP 9339.
1LP-SQS-9339 OBJECTIVE: 7 NUMBER: 1-97-105, New
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JTA #:
K/A #: 194001.Kl.16 K/A IMPORTANCE: 3.5/4.2 Final Revision. Rev. 3A
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Questi:nNumber 14
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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 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 until shift change, provided that immediate action is taken to bring the compliment up to the minimum.
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 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 compliment up to the minimum.
POINTS: 1.00 ANSWER: C.
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REFERENCES: OM-1/2.48.4.B Inst. A Issue 3, Rev. 17.
l ILP-SQS-48,1 OBJECTIVE: 5 NUMBER: 1-97-112, LOT - 0808 l
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K/A #: 194001.A1.03 K/A IMPORTANCE: 2.5/3.4
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- QuestionNumber 15 t
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i Technical Specification Surveillance 4.8.1.1.2 requires in part, that each !
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Emergency Diesel Generator (EDG) be started from ambient conditions at least ,
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n once per 31 days, on a STAGGERED TEST BASIS. Wnich 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. l'" 8 "" same B. l 15 same i l
C. 1" l'* next
D. 8'" 31 same.
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i POINTS: 1.00 ANSWER: B.
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REFERENCES: Technical Specification Definitions Amendment No. 192.
ILP-SQS-TS OBJECTIVE: 3 NUMBER: 1-97-107, New
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JTA #2 K/A #: 194001.K1.17 K/A IMPORTANCE: 2.1/2.5 Final Revision, Rev.3A
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Question Number 16
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Given the following:
e Reactor power has just been reduced from 100% to 60%.
- Control Bank D rods (CBD) are at 122 steps.
e T.,. - Tr.: error is 0*F.
e 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?
A. 299.0 gallons of Boric Acid.
B. 361.5 gallons of Boric Acid.
C. 5,866 gallons of Primary Water.
D. 7,190 gallons of Primary Water.
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POINTS: 1.00 i
ANSWER: B. l l
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e CB10B - Rod motion will add +550 pcm. l e Xenon will add -115 pcm. j e Boron must add -435pcm. l
- CB28 - Boron worth is -8.4pcm/ ppm.
- Boron conc. must be increased by 52 ppm for a total of 427 ppm.
- T... - 565'F at 60% power.
e WAG table 565'F, iteration method:
- From 375 ppm to 425 ppm - 347.57 gal of BA.
- 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.
REFERENCES : Unit 1 Curve book, CB-10B, CB-28, WAG Tables 560*F. '
ILP-SOS-RT-6 OBJECTIVE: 15 NUMBER: 1-97-108, M-LRT 2.3.4 JTA #:
K/A #: 194001.A1.08 K/A IMPORTANCE: 2.6/3.1 Final Revision, Rev. 3A
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Question Number 17
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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-1FW-479, 489, s 499]?
l A. ONLY the SI signal needs to be Reset. I
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B. ONLY the Feedwater Interlock signal needs to be Reset.
C. ' BOTH the SI signal AND the Feedwater Interlock signals need to be Reset.
l D. The feedwater regulating BYPASS valves CANNOT be OPENED with a l standing SI signal even if it is Reset. ,
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POINTS: 1.00 ANSWER: B.
Provide the following references: UFSAR Figure 7.2-1, Instrumentation and Control System Logic Diagram, Sheet 1 and 13. ]
REFERENCES: UFSAR Figure 7.2-1, Sheet 13, Rev. 10.
1LP-SQS-24.1 OBJECTIVE: 4 NUMBER: 1-97-109, New
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JTA #:
( K/A #: 194 J ')1. A1. 07 K/A IMPORTANCE: 2.5/3.2 Final Revision, Rev.3A
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QuestionNumber 18
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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 ,
l - Limits (RIL) and the Shutdown Margin (SDM) if the main generator electrical i l output was raised by 3%, and the only other operator action was to restore T.,.
to program using the Boration/ Dilution controls?
RIL will (1) , and SDM will (2) .
(1) (2)
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A. raise, lower.
B. Iower, raise.
4 C. Iower, lower.
? D. raise, raise.
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POINTS: 1.00 ANSWER: A.
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i REFERENCES: Tech. Spec. Definitions j. ILP-SOS-1.4 OBJECTIVE: 6 NUMBER: 1-97-001, M-0395 i' .
JTA #: 0010110104 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:
o Control Rod D12, a control bank "C" Group 1 rod, has fallen into the core due to an equipment failure.
e The equipment failure has been corrected and all retests are completed t
satisfactory.
e 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 recovery.
e 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.
C.
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Power cabinet 1AC failure and ONLY those rods aligned to power cabinet 2AC will move.
D. Power cabinet 2AC failure and ONLY those rods aligned to power cabinet 1AC will move. l l
l POINTS: 1.00 ANSWER: D.
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RE FERENCES : 10M-1.1.D - Issue 4 Rev. 1, AOP 1.1.5 - Issue 3A Rev. 5 ILP-SQS-1.3 OBJECTIVE: 10 NUMBER: 1-97-002, M-0055
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j K/A #: 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 e 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. ,
- 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.
- Letdown flow is 104 gpm.
- RCP seal injection flow is 24 gpm.
- Total RCP seal return flow 9 gpm.
- -There have been three auto make-ups to the VCT in the past 30 minutes.
Which of the following is the malfunction?
A. Letdown flow control valves [LCV-lCH-ll2 & 115A), are partially diverting letdown flow to the CRT's.
B. Letdown line relief valve [RV-1CH-203), has lifted and failed to reseat causing a portion of letdown flow to be diverted to the PRT.
C. RCP seal injection control valve [HCV-1CH-186], has failed open causing excessive seal injection flow and VCT depletion. l
D. Letdown isolation valve [LCV-lCH-460A), has developed a severe packing j leak causing a small loss of reactor coolant condition.
POINTS: 1.00 ANSWER: A.
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REFERENCES : lOM-7.1.D - Issue 4 Rev. 2, lOM Fig. No. 7-1 - Issue 9 Rev. 9.
lLP-SOS-7.1 , OBJECTIVE: 6 NUMBER: 1-97-003, New JTA i:
K/A #: 3.01.004.020 Kl.02 K/A IMPORTANCE: 2.7/2.8 Final Revision, Rev. 3A
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Question Number 21 3
- j Given the following:
e The plant is at 1004 power.
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- All systems are in their at-power, NSA configurations.
l e Annunciator [A4-51] 'RCS T.,. HIGH", is lit. l i
You are directed to adjust RCS temperature using the Boration/ Dilution controls.
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Determine the mode of makeup control required and the expected corresponding valve lineup.
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I A. Borates Boric acid flow control valve [FCV-1CH-113A] - OPEN Primary water flow control valve [FCV-1CH-114A] - CLOSED Makeup stop valve t> the charging pump suction [FCV-1CH-ll3B] - CLOSED j Makeup stop valve to the VCT [FCV-lCH-ll4B] - OPEN B. Borates' ;
Boric acid flow control valve [FCV-1CH-113A] - OPEN !
Primary water flow control valve [FCV-1CH-114A] - CLOSED l
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Makeup stop valve to the charging ptr ) suction [FCV-1CH-113B] - OPEN Makeup stop valve to the VCT [FCV-1CH-114B] - CLOSED l C. Dilute;
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Boric acid flow control valve (FCV-1CH-113A] - OPEN Primary water flow control valve [FCV-lCH-ll4A] - OPEN Makeup stop valve to the charging pump suction [FCV-1CH-113B] - OPEN.
Makeup stop valve to the VCT [FCV-lCH-114B] - CLOSED D. Dilute; Boric acid flow control valve [FCV-1CH-ll3A] - CLOSED Primary water flow control valve [FCV-lCH-114A] - OPEN Makeup stop valve to the charging pump s action [FCV-lCH-113B] - CLOSED l Makeup stop valve to the VCT [FCV-1CH-114B] - OPEN l
POINTS: 1.00 ANSWER: B.
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REFERENCES: 10M-7.1.D - Issue 4, Rev. 2.
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ILP-SOS-7.1 OBJECTIVE: 5 NUMBER: 1-97-004, New JTA f:
K/A #: 3.01.004.010 A4.03 K/A IMPORTANCE: 3.9/3.7
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QuestionNumber 22
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Given the followings e A plant startup is in progress with Rx power at 174.
. Steam dumps are in'the Main Steam Pressure Control Mode maintaining i Main Steam Pressure at 1005 psig. ,
e The Main Generator Output Breakers have just been closed.
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- Rod control is in manual.
e CBD is at 100 steps with normal rod sequencing.
e. RCS boron concentration is 705 ppm.
e The Rx core is Xenon free.
A spurious equipment failure caused an intermittent continuous rod withdrawal of l CBD that stopped with CBD at 110 steps. . Assuming that no Rx trip occurs and T.v. ,
remains on program, determine the level at which Rx power will stabilize.
A. 14%
B. 17%
l C. 18% ?
D. 20%
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ANSWER: D.
l 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. ]
l Delte - 59 pem, i 59 pcm + 20 pcm/% power = 3% change. 17% + 3% = 20%. l l
Provide the following references BVPS Unit 1, Curve Book, Cycle 12, Issue 12 Rev. 1, .
REFERENCES : Unit 1 Curve Book, Cycle 12, Issue 12 Rev. 1, curves CB-11E,13,21 and 24B.
1LP-SOS-LP-RT-6 OBJECTIVE: 9 NUMBER: 1-97-005, New JTA #: ,
K/A 8: 3.01.000.001.EK1.03 K/A IMPORTANCE: 3.9/4.0 Final Revision, Rey,3A
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Qu:stion Number 23
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Due to an equipment malfunction in the automatic rod control system, a ;
continuous rod withdrawal was initiated from an initial power level of 724.
After 5 seconds of rod withdrawal, the reactor operator selected " MANUAL" on the -
Rod Control Selector Switch which terminated the rod movement. All other l 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 transient?
A. RCS Tave.
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B. Pressurizer level.
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l C. Delta I.
D. Reactor power.
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REFERENCES: LP-RT-6 Rev. 2
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i lLP-SOS-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 I
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Questi:n Number 24
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.Given tho'following:
i e A reactor startup is in progress with power at IE-8 amps in the IR.
e All systems are aligned in their normal lineups for the current power level.
e 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).
e An electrical fault occurs that causes a sustained loss of the 1A 4KV bus, e 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 e following the loss of the 1A 4KV bus. '
i A. The Reactor will trip on low flow due to the loss of the 1A RCP. I 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. '
B. The Reactor will not trip, i
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 control rods.
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 l bus and Rod-MG-2 does not have sufficient capacity to maintain power 1 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 IC 4KV bus via the 480V bus tie breaker.
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POINTS: 1.00 ANSWER: D. I
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REFERENCES : 10M-37.1.D - Issue 4 Rev. O, IOM-1.5.B.4 - Issue 2 Rev. 10.
ILP-SOS-37.1' OBJECTIVE: 2 NUMBER: 1-97-007, New JTA #:
K/A 8: 3.01.000.003.EK2.05 K/A IMPORTANCE: 2.5/2.8 Final Revision.Rev 3A
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QuestionNumber 25
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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.
l C. be affected by ONLY a loss of primary coolant accident and a loss of secondary coolant accident.
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D. be affected by the loss of primary and secondary coolant accidents, and steam generator tube rupture accidents.
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I-l POINTS: 1.00 ANSWER: D.
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REFERENCES: NOMCD Lesson Plan.
1LP-SOS-53.3 OBJECTIVE: 3 NUMBER: 1-97-009, R-0360 JTA 8: 3010010601 K/A 4: 3.01.000.007.EK3.01 K/A IMPORTANCE: 4.0/4.3 Final Revision. Rev.3A
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._ Question Number 26
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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. i
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breakers can be opened?
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1. Locally on the' front of the respective breaker.
2. Lift Coil Disconnect' Switch Panel.
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3. Rod drive MG Control' Panel.
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Rod Drive MG Set RTB's Supply Breakers
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A. 1 ONLY. I and 3 ONLY.
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B. 1 and 2 ONLY. 1 and 3 ONLY.
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C. 1 ONLY. 3 ONLY.
D. I and 2 ONLY. 3 ONLY.
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. POINTS: '1.00 ANSWER: A.
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t REFERENCES: 10M-1.3.C - Issue 4, Rev. 1.
ILP-SOS-53.3 OBJECTIVE: 6 NUMBER: 1-97-011, New JTA #1 K/A #: 3.01.000.029.EA1.12 K/A IMPORTANCE: 4.1/4.0 Final Revision, Rev. 3A >
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QuestionNumber 27 j
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1 With the Unit at 1004 power with all systems in their at-power, NSA t configurations; *MALIVNCTION" is displayed on the RVLIS Train A plasma display
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Dynamic Head indicator. Which of the following caused this indication? !
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A. A high volume sensor bellows is leaking causing a hydraulic isolator il - limit switch to actuate.
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B. Pressurizer pressure protection transmitter channel I [PT-lRC-455), -
has failed HIGH.
C. This indication is normal, the RCP breaker auxiliary contacts disable the Dynamic Head indications when ANY RCP is running. 1 3;!
i 'D.~ A single T-hot narrow range channel I RTD [TRB-RC-412Bl], has failed
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OPEN.
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l POINTS: 1.00-ANSWER: A.
I REFERENCES: lOM-6.1.D - Issue 4 Rev. 1 ILP-SOS-6.5. OBJECTIVE: Sh NUMBER: 1-9*1-012, New
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JTA 8:
-K/A #: 3.02.000.K6.03 K/A IMPORTANCE: 3.1/3.6 Final Revision, Rev. 3 A -
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Question Number 28
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7 A large break LOCA and Safety Injection is in progress in the cold leg injection )
'- 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"'t A. CNMT sump to LHSI pump suction valves (MOV-1SI-860A&B) will CLOSE. ;
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B. HHSI to RCL Cold Leg isolation valve (MOV-1SI-836) will OPEN. l C. RWST to HHSI valves (MOV-1CH-115B&D] will OPEN.
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D. LHSI to RHSI 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 1B, Rev. 1 [
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ILP-SOS-11.1 OBJECTIVE: Sc NUMBER: 1-97-013, M-0421 :
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JTA #:
K/A #: 3.02.006.020.K4.03 (006A3.08) K/A IMPORTANCE: 3.2/3.6
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Final Revision, Rev. 3A
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Question Number 29
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Given the following:
, e The plant is operating at 100% power.
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s 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 Pressurizer Level. ]
C. High Pressurizer Level.
D. High Pressurizer Pressure.
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l POINTS: 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 i
l JTA #:
i K/A #: 3.02.011.000.A2.10 (OllK3.01) K/A IMPORTANCE: 3.4/3.6
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Final Revision. Rev.3 A
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QuestionNumber 30
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Given the Following:
- A normal plant cooldown and depressurization is in progress using the condenser steam dumps.
- All S/D rod banks are withdrawn, e ' RCS T.= is 501*F and PZR pressure is 1925 psig.
e The cooldown and depressurization.is temporarily terminated to conduct shift turnover. .
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e During the turnover T.n 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 ;
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must be performed?
A. Immediately depressurize the RCS to less than 1945 psig to' prevent ,
i 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.
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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. i
POINTS: 1.00 ANSWER: D.
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REFERENCES: 10M-51.4.C, Issue 4, Rev. 9 - Caution.
1LP-SQS-51.1, OBJECTIVE: 3/9 NUMBER: 1-97-015, M-0255 JTA f:
K/A i: 3.02.013.000.K4.12 (010K1.02) K/A IMPORTANCE: 3.7/3.9
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Find Revisiort Rn. 3A {
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QuistionNumber 31
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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, j and the Motor Driven Auxiliary Feedwater Pumps auto start. l l
C. MSLI actuation, Emergency Diesel Generators auto start and load, and I the Turbine Driven Auxiliary Feedwater Pump auto starts.
D. CREBAPS actuation, Feedwater Isolation actuation, and the Turbine Driven Auxiliary Feedwater Pump auto starts. j l
l POINTS: 1.00 ANSWER: A.
REFERENCES : 10M-1,11,12, & 13.2.B - Issue 4, Rev. 4,2,1 & 3 Respectively.
ILP-SQS-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 Final Revision. Rev.3A
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QuGstion Number 32
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Given the following:
s The Unit is operating at 100% power with all systems in their at-power, NSA configurations. ,
o The following Annunciators are actuated:
- [A4-4), " PRESSURIZER CONTROL LOW LEVEL DEVIATION,"
e (A3-115), " REGEN HX LETDOWN OUTLET TEMPERATURE HIGH,"
- [Al-39), " CONTAINMENT SUMP LEVEL HIGH."
- P2R level is 424 and dropping slowly.
- Domineralizer Bypass valve (TCV-1CH-143), is in the VCT position, e Charging flow control valve [FCV-1CH-122], is full open.
- Charging flow meter [FI-lCH-122), is pegged high.
- Letdown flow meter [FI-1CH-150], indicates 105 gpm.
- All RCP seal injection flow meters indicate between 8.0-0.5 gpm.
- All RCP seal return flow recordero indicate between 2.0-3.2 gpm.
i 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.
D. PZR AUX spray control valve [MOV-1CH-311], inlet line.
POINTS: 1.00 ANSWER: B.
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REFERENCES: OM Figure 7-1 Rev. 9 I l l 1LP-SQS-7.1 , OBJECTIVE: 2 NUMBER: 1-97-017, New '
JTA #:
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! K/A #: 3.02.000.022.EA2.01 K/A IMPORTANCE: 3.2/3.8 l
l Final Revision. Rev. 3A
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Qusstion Number 33
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Given the following:
] e 'The plant is at 47% power. '
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- All systems are in their at-power, NSA configurations. -
e The pressurizer level selector switch is in position 1, 459/460. i e The reference leg for pressurizer level transmitter (LT-1RC-460]
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develops a leak. ,
f_ Assuming no operator actions are taken, which of the following will occur?
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i l A. Pressurizer level will stabilize at the full load setpoint of 59%.
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B. The Rx will eventually trip on low pressurizer pressure.
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- C. Annunciator [A4-1), " PRESSURIZER CONTROL LEVEL HIGH" will actuate. l
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i D. All letdown orifice isolation valves will immediately trip closed.
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i POINTS: 1.00 >
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ANSWER: C.
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REFERENCES: 10M-6.4IF Issue 4, Rev. 5 ILP-SOS-6.5 , OBJECTIVE: Sh NUMBER: 1-97-018, M-0116 JTA I K/A ft 3.02.000.028.EK1.01 (028AK1.01) K/A IMPORTANCE: 2.8/3.1 Final Revision. Rev. 3A
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Qusstion Number 34 '
The pressurizer level controller utilizes an integral control function. Which of the following describes the action of this integral control function? f i
i The pressurizer level controller integral control action will (1) the !
demand signal to the Pressurizer Level Control Valve, [FCV-lCH-122) (2) .
A. (1) raise !
(2) for as long as actual Lys, is below program level. ;
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B. (1) raise '
(2) only as long as actual Lys, is dropping.
C. (1) lower (2) only as long as program Lpar is dropping.
D. (1) provide'
(2) in proportion to the difference between actual and program Lys,.
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POINTS: 1.00 i I
ANSWER: A. '
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I REFERENCES: : 1/2LP-ICS-1.4 ILP-ICS-1.4 OBJECTIVE: 2 NUMBER: l'97-019, New j
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JTA #: !
K/A fi 3.02.000.028.EK2.03 K/A IMPORTANCE: 2.6/2.9 i
Final Revision, Rev.3A i
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QusstionNumber 35
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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?
l l A. RH Letdown to Non Regen HX Inlet Flow Control Valve (MOV-lCH-142), !
should be CLOSED.
i l B. PZR Spray' Valve (PCV-lRC-455A), should be OPENED.
l C. CCR HX Bypass Temperature Control Valve (TCV-CC-100), should be
!. OPENED.
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!. D. LP Letdown Back Pressure Regulating Valve (PCV-lCH-145), should be OPENED.
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POINTS: 1.00
ANSWER: D.
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I REFERENCES: lOM-6.4.F - Issue 4, Rev. 6 l
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ILP-SQS-6.5 OBJECTIVE: 4 NUMBER: 1-97-020, New i -
i l JTA f:
l l K/A 1: 3.03.010.000.A1.04 K/A IMPORTANCE: 3.6/3.8 i
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l l Final Revision, Rev. 3A -
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QuestionNumber 36
- Given the following
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e 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.
e' The associated PORV Block Valve has been CLOSED.
- AFTER CLOSING the PORV Block Valve the PZR is saturated at 1600 psig,
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e- 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. 2 4 6*F
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t POINTS: 1.00 *
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ANSWER: C.
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' REFERENCES: OM Fig. 6-2 Rev. 9 ILP-SOS-6.4 , OBJECTIVE: 4 NUMBER: 1-97-021, New JTA f:
K/A f: 3.03.000.008.EA1.01 K/A IMPORTANCE: 4.2/4.0 Final Revision, Rev.3A
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QuGstion Number 37
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When trying to establish RCS flow during a loss of ALL 4KV power, which of the l following would cause Natural Circulation flow to RISE? l A. Lowering the RCS cooldown rate using the Condenser Steam Dumps.
B. Raising the setpoint on the S/G Atmospheric relief valves. I C. Lowering RCS pressure using Aux Spray.
D. Raising all SGWL's using Aux Feed.
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POINTS: 1.00
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ANSWER: D.
REFERENCES : 10M-53B.5.GI-4 - Issue 1B, Rev. 1 ILP-SQS-53.2 OBJECTIVE: 11 NUMBER: 1-97-022, New JTA #:
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K/A f: 3.07.000.055.EA2.02 K/A IMPORTANCE: 4.4/4.6 Final Revision, Rev. 3A
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QuestionNumber 38 l
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What is the bases for the RCP Trip Criteria'Setpoint of RCS/ Highest SG D/P = 150 i paid [450 paid 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.
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B. Allows the RCPs to remain in service for core cooling until cavitatien 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 f
ANSWER: A.
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REFERENCES: 10M-53.B.5.GI-6, Issue 1B, Rev. I 1LP-SOS-53.2 OBJECTIVE: 1 NUMBER: 1-97-023, New
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JTA #:
K/A 6: 3.03.000.011.EK3.14 K/A IMPORTANCE: 4.1/4.2 l
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Final Revision, Rev. 3A e
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Question Number 39
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A large break LOCA has occurred. The crew has made the following procedure l transitions; E-1, Loss of Reactor or Secondary Coolant, to ES-1.3, Transfer to Cold Leg Recirculation, completed the cold leg recirculation line up and then returned to the procedure and step in effect, E-1 Loss of Reactor or Secondary Coolant, step 24. Step 24 of E-1 directs actions to isolate the SI Accumulators if at least two RCS hot leg temperatures are less than 390*F. i I
The bases for this RCS temperature ensures. !
A. adequate core cooling is established prior to isolating the Accumulators as a water injection source.
B. sufficient *ccumulator water and nitrogen volumes are injected into the RCS
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prior to isolating the accumulators. l C. saturation pressure of the RCS is less than Accumulator nitrogen pressure when the accumulator water volume is fully discharged.
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 ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-024, New JTA 8: -
K/A i: 3.03.000.Oll.EK3.12 K/A IMPORTANCE: 4.4/4.6 Final Revision, Rev. 3A
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QuIstionNumber 40
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A major plant transient is in progress with the Unit at full power. The current plant parameters are
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(Assume parameters are stable unless otherwise stated.) !
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e Highest Core Exit Thermocouple =. 623*F e All RCS T-hot's are between 608-610*F e All RCS T-cold's are between 54 4-54 6*F e RCS Press = 2112 psig and dropping.
- A S/G: Press'= 800 psig, Level = 444, Feed Flow - 3.6 E5 lbm/hr. !
e B'S/G: Press = 805 psig, Level =.434, Feed Flow - 3.7 E5 lbm/hr.
- C S/G: Press - 790 psig, Level = 446, Feed Flow = 3.5 E5 lbm/hr. ,
o Containment: '
e Press = 12 psia, -
- Temp = 110*F, i e Particulate Rad Monitor [1RM-RM-215A1 - 1.3 E5 cpm with its HI.
alarm LIT.
- Gaseous Rad Monitor (1RM-RM-215B] = 1.9 E5 cpm with its HI-HI alarm LIT. ,
e In-Core Transfer Device Rad Monitor [1RM-RM-204) = 1000 mR/hr '
with its HI-HI alarm lit.
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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.
D. A single S/G is Ruptured and Faulted inside containment.
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POINTS: 1.00 ANSWER: C.
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P REFERENCES: 10M-53A.I.E-0 - Issue 1B, Rev. 4 1LP-SOS-53.3 OBJECTIVE: 6 NUMBER: 1-97-025, New JTA 8:
K/A #: 3.03.000.011.EA2.13 K/A IMPORTANCE: 3.7/3.7 ,
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Final Revision, Rev.3A
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e QuestionNumber 41
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Given the following:
- The Unit is in Mode 4 with a plant heat-up in progress.
- There is a bubble in the pressurizer with PZR level at 22%.
- RCS Pressure is 340 psig.
- 1A RCP is in service, o Pressurizer overpressure protection system (OPPS) is in service.
-e. IEC is performing maintenance on the pressurizer pressure master and slave controllers, and requests the controllers to be placed in the following line-ups e PZR Spray Valve controller, [PCV-1RC-455A] in AUTO.
- PZR Spray Valve controller, [PCV-1RC-455B] in MANUAL and shut.
- PZR Group A Heater control switch in AUTO.
- PZR Group B,C,D, and E Heater control switches in OFF.
e Master Pressure Controller in AUTO.
What effect will this line-up have on PZR pressure with no further operator action?
A. No effect. The master pressure controller output is bypassed with the OPPS keyswitches in AUTOMATIC.
B. PZR pressure will be maintained at 340 psig with [PCV-1RC-455A]
and Group A Heaters operating in AUTO.
C. PZR pressure will rise to 410 psig and cause ONLY OPPS Relief valve [PCV-1RC-455D) to open. [PCV-1RC-455C) will not open.
D. PZR pressure will rise to 410 psig and cause BOTH OPPS Relief valves [PCV-lRC-455C) and [PCV-lRC-455D) to open.
POINTS: 1.00 ANSWER: D.
REFERENCES : lOM-6.1.D Issue 4, Rev. 1 ILP-SQS-6.4 OBJECTIVE: 5 NUMBER: 1-97-026, New JTA #:
K/A 8: 3.03.000.027.EA2.03 K/A IMPORTANCE: 3.3/3.4 Final Revhion, Rev. 3A l
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Question Number 42 '
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, Given the following: '
! * The Unit is in Mode 1 with all systems in their at-power, NSA ,
configurations. ,
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- The 2B and 2E Pressurizer heaters are in MANUAL and ON to equalize the PZR and RCS boron concentrations.
- The Auto /Ma.: 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 spray valves (PCV-
1RC-455A and B) in MANUAL, 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, resultAng in an CT/AT Reactor Trip.
'B. DROP, resulting in a Low Pressurizer Pressure Reactor Trip.
C. RISE, and cause a PZR Safety valve (RV-lRC-551A] to OPEN.
D. RISE, and cause PZR PORV (PCV-lRC-455C) to OPEN at < 2335 psig.
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b POINTS: 1.00 ANSWER: D.
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l REFERENCES: lOM-6.4.ABU Step 5 Caution, Issue 3, Rev. O lLP-SQS-6.4 OBJECTIVE: 11 NUMBER: 1-97-027, New I
JTA #:
K/A # 3.03.000.027.G05 K/A IMPORTANCE: 3.3/3.3
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Final Revision. Rev.3A
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Question Number 43 l
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Which of the following is an entry condition for AOP 1.6.4 ** aam Generator
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Tube Leakage?
1. PZR Level Dropping.
2. S/G Blowdown Rad Monitor Rising.
3. SGWL Rising.
A. 1 ONLY.
B. 2 ONLY.
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C. 3 ONLY.
D. 1 OR 3.
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l POINTS: 1.00 ANSWER: B.
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REFERENCES: AOP 1.6.4 Issue 3A, Rev. 7 1LP-SQS-53C.1 OBJECTIVE: 2 NUMBER: 1-97-028, New
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JTA 4:
K/A 4: 3.03.000.037.G11 K/A IMPORTANCE: 3.9/4.1
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Final Revision, Rev. 3A I l
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t QuestionNurnber 44 :,
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4 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 54. What is the basis for establishing this minimum level in the i ruptured S/G7 A. To compress the ruptured S/G steam bubble and raise its pressure, thus 4 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 anticiphtion of an uncontrolled radiological release to the environment.
D. ' To prevent thermal stratification in the ruptured 'S/G .vhich would l extend the time required to stop break flow. '
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ANSWER: B.
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REFERENCES: 10M-53B.4.E-3 - Issue IB, Rev. 3 ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-029, M-0536
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JTA #:
P K/A f: 3.03.000.038.EK3.06 (03BEK3.06) K/A IMPORTANCE: 4.2/4.5 ,
Final Revision, Rev.3A
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Question Number 45
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f Given the following:
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e The Unit was operating at 100% power with all systems in their at-
- power, NSA configurations. <
i e- A SGTR on the 1A S/G, accompanied by a loss of the 1B 4KV bus has i
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- occurred.
j~ e The operators have transitioned from E-0, Reactor Trip or Safety i Injection, to E-3, Steam Generator Tube Rupture, and have completed j isolating 1A S/G.
S/G parameters are:
l l. * 1A S/G; Level = 186, Pressure = 1010.psig
's 1B S/G; Level - 23%, Pressure - 995 psig
- e IC S/G Level = 15%, Pressure = 1000 psig l
- .. 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?
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A. 1A S/G Atmospheric Relief Valve (PCV-1MS-101A), will modulate open at !
1035 psig. !
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g B. 1B S/G Atmospheric Relief Valve [PCV-1MS-101B), will trip open at 1060 i i psig, i-C. IC S/G Atmospheric Relief Valve [PCV-1MS-101C), will trip open at 1060
.i psig.
! D. IC S/G Atmospheric Relief Valve (PCV-1MS-101C), will modulate open at l i 1035 psig.
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POINTS: 1.00 i
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REFERENCES : 10M-21.3.A - Issue 4, Rev. 4.
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ILP-SOS-21.1 OBJECTIVE: 3 NUMBER: 1-97-030, New JTA #:
K/A #: 3.03.000.038.EA1.16 K/A IMPORTANCE: 4.4/4.3 l
Final Revision, Rev.3A
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'QuestionNumber 46
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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 in the loop.
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a POINTS: 1.00 ANSWER: A.
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REFERENCES: ILP-SOS-6.3 ILP-SQS-6.3 OBJECTIVE: 11 NUMBER: 1-97-031, New l JTA #2 K/A # 3.04.035.010.Kl.09 K/A IMPORTANCE: 3.8/4.0 Final Revision. Rev.3A l
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4- .o QuestionNumber 47
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With the Unit at full power, the number one seal on IB RCP has failed. After
' tripping the Rx and stopping the IB RCP, which of the following is performed to limit the temperature rise of the 1B RCP lower radial bearing?
a A. - OPEN the RCP Seal bypass valve (HOV-1CH-307) .
j- B. OPEN the IB RCP Thermal barrier flow throttle valve (ICCR-307)
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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J POINTS: 1.00 ANSWER: D.
REFEPINCES: 10M-7.4.ABE - Issue 3,'Rev. 3 ILP-SQS-6.3 OBJECTIVE: 12 NUMBER: 1-97-032, New JTA #:
-K/A #: 3.04.003.000.A2.01 K/A IMPORTANCE: 3.5/3.9 Tinal Revision, Rev. 3A i
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Question Number 48 ;
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- When conducting a core off load in a recent refueling outage, the manipulator
! . crane operator experienced the following:
e: Minor vibrations of a suspended fuel assembly in one quadrant of the ,
Rx vessel near the outer edge of the vessel.
- ' Visual observation of flow turbulence in the same general location in ;
the Rx vessel. l With all systems'in their normal line up for HODE 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'
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Loop'
Hot Leg.
B. High flow through the CVCS letdown orifices via the 'A' Loop Cold Leg.
l C. High letdown flow to the Residual Heat Removal System via the 'A' j Loop Hot Leg.
D. High RCS sample purge flow to the Primary Sample System via the
'B' Loop Hot Leg.
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POINTS: 1.00 ;
ANSWER: C. l
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REFERENCES: lOM Figure 6-1 - Rev. 7 ILP-SOS-10.1 OBJECTIVE: 1 NUMBER: 1-97-033, New JTA #:
K/A #: 3.04.005.000.Kl.Os K/A IMPORTANCE: 3.6/3.9 Final Revision, Rev.3A
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O Question Number 49
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The Unit is at 37% 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 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 ascension. What is the response of the main feedwater regulating valves to this action?
The Main Feedwater Regulating Valves will initially throttle _
A. CLOSED due to the shrink of the SGWL, and then throttle OPEN when '
level drops below 44%.
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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ANSWER: D.
REFERENCES : lOM-24.1.D - Issue 4, Rev. 1
- ILP-SOS-24.1 OBJECTIVE
- 4 NUMBER: 1-97-034, New JTA 9:
K/A 9: 3.04.035.010.A3.01
K/A IMPORTANCE: 4.0/3.9
,
i Final Revision. Rev.3A
. Question Number 50 -
Which of the following would electrically prevent'the 1C RCP breaker from closing?
A. 1C RCP #1 seal differential pressure <200 paid.
B. 1C RCP G1 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.
RE FERENCES : OM Figure 6-21 Rev.10 and 6-22 Rev. 8 ILP-SOS-6.3 OBJECTIVE: 6 NUMBER: 1-97-035, New
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JTA i:
K/A #: 3.04.000.015.EK2.10 K/A IMPORTANCE: 2.8/2.0 Final Revision, Rev.3A
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QuestionNumber 51 I
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While operating in Reduced Inventory /Midloop conditions, a loss of RCS inventory 1
,
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 i Inventory,' to minimize the possibility of gas binding the RHR pumps when they j are restarted? j 1. Raise RCS Level using a Charging /HHSI pump, 2. Shut tha RHR temperature and flow control valves (MOV-1RH-758 and 605).
I 3. Raise RCS Pressure to greater than 150 psig. '
j' 4. Establish level in at least 2 S/G's >15% narrow range.
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A. 1& 3.
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B. 1 & 2. ,
j C. 1, 2 & 3.
D. 2, 3& 4.
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POINTS: 1.00 ANSWER: D.
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REFERENCES: 10M-53C.4.1.10.1 - Issue 3A, Rev. 5 ILP-SOS-10.1 OBJECTIVE: 9 NUMBER: 1-97-036, New 1 .
I JTA 4:
i j K/A 4: 3.04.000.025.EK1.01 K/A IMPORTANCE: 3.9/4.3 i
s Final Revision, Rev.3A l
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' QuGstion Number 52
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Functional Restoration Procedure, FR-C.1, " Response to Inadequate Core Cooling" is designed to reduce core exit thermocouple temperatures and recover Rx vessel level. .Which of the following describes the processes used to accomplish this objective in the order in which they are perfo.w d?
1. Run at least one HMSI 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 I inject the SI Accumulators.
3. Restart the RCP's to provide two phase flow through the core.
4. Restore High Pressure Safety Injection flow.
,
A. 2, 3, 4. i B. 2, 1, 3.
C. 4, 2, 1.
D. 4, 2, 3. '
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I POINTS: 1.00 ANSWER: D.
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t REFERENCES: 10M-53B. 4. FR-C.1 - Issue IB, Rev. 3 >
ILP-SQS-53.3 OBJECTIVE: 3 NUMBER: 1-97-037, tiew JTA 9:
i K/A f: 3.04.000.074.EK1.03 K/A IMPORTANCE: 4.5/4.9 l
l I Final Revision. Rev. 3A
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l QuestionNumber 53 r
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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?
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].
t D. BIT Outlet Isolation Valve [MOV-1SI-867D] . ,
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POINTS: 1.00 ANSWER: B.
REFERENCES: 10M-53A.1.1-A - Issue 1B, Rev. 1 ILP-SOS-11.1 OBJECTIVE: 3 NUMBER: 1-97-038, New JTA #:
l K/A #: 3.04.000.074.EA1.27 K/A IMPORTANCE: 4.2/4.2 t
Final Revision. Rev. 3A
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- Questisn Number 54
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The 'C' ' Loop Main Steam Isolation Valve (TV-1MS-101C), inadvertently trips closed with.the Unit operating at 404 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 For the above listed conditions,-the steam dumps will maintain-,
l A. ALL S/G pressures at -1005 psig by maintaining T... at 449'F (2'F dead l band from no-load T.,.)-using steam from only 'A' & 'B' S/G's.
B. 'A' & 'B' S/G pressures at -1005 psig and the 'C' S/G atmospheric i steam dump valve IPCV-1MS-101C), will trip open if 'C' S/G pressure l 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.
l D. 'A' & 'iB ' S/G pressures at -1005 psig and the 'C' S/G safety valves will open in succession if 'C' S/G pressure reaches 1075 psig.
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POINTS: 1.00 ANSWER: B.
REFERENCES : 10M-21.2.B - Issue 4, Rev. 1, 10M-21.3.A - Issue 4, Rev. 4.
t 1LP-SQS-21.1 OBJECTIVE: 3 NUMBER: 1-97-039, New l
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JTA 8:
K/A #: 3.05.039.000.K1.02 K/A IMPORTANCE: 3.3/3.3 I
Final Revision, Rev. 3A
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. Question Number 55 e
Given the following:
- The Unit is in MODE 1 with Rx power at 84.
- Boron concentration is 665 ppm.
- Tave is 547'F.
- PZR Pressure is 2235 psig.
- PZR Level is 224.
- Steam dumps are in Auto in the Main Steam Pressure Control Mode.
- S/G atmospheric relief valve controllers are in Manual.
- All systems are aligned in.their normal lineup for the existing power level.
- . An inadvertent MSLI occurs.
Assuming no Rx Trip occurs and no operator action, determine the response of the following parameters one minute after the MSLI.
Rx Power 'A' S/G PZR Level l Pressure
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A. DROP DROP RISE B. RISE RISE DROP i
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- C. DROP RISE RISE l
D. RISE DROP RISE f
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ANSWER: C.
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REFERENCES: LP-RT-6, Objective 15
! ILP-SOS-21.1 OBJECTIVE: 2 NUMBER: 1-97-040, New l
JTA #:
K/A #: 3.02.002.000.KS.11 K/A IMPORTANCE: 4.0/4.2
[
Final Revision. Rev. 3A i-L
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QuestionNumber 56
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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 Non-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.
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 out of every 31.5 seconds, causing the Reheat and Turbine Stop valves to reduce load at 200%/ min.
.
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ANSWER: A.
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II REFERENCES: 10M-26.1.B - Issue 4, Rev. 4 1LP-SQS-26.3 OBJECTIVE: 9d NUMBER: 1-97-041, New JTA #:
K/A f: 3.05.045.050.A7.10 K/A IMPORTANCE: 2.7/2.9 Final Revision, Rev.3A
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QuestionNumber 57 l
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Given the following
- * The Unit is stable at 85%.
!
- All systems are in their at-power, HSA configurations.
!
- The operator notices that the IC S/G feed reg valve [ FCV-1 FW-4 98 ]
l demand signal is 15% lower than the 1A and IB S/G' feed reg valves l [FCV-1FW-478 and 488).
'
e 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.
l Which of the following statements explains the observed plant conditions?
!
l A. 1C S/G controlling feed flow transmitter has failed LOW.
i B. 1C S/G feed reg valve bypass valve [FCV-1FW-499) has failed l OPEN.
l C. There is a feed water leak upstream of the IC S/G feed reg l valve [FCV-1FW-498).
D. 1C S/G feed reg valve [FCV-1FM-498) valve stem has become uncoupled from its actuator.
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l RE FERENCES: lOM-24.1.D - Issue 4, Rev. 1 j
ILP-SOS-24.1 OBJECTIVE: 4 NUMBER: 1-97-042, M-0153 JTA #:
K/A #: 3.05.059.000.Kl.04 (059A2.11) K/A IMPORTANCE: 3.4/3.4 l
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Final Revision, Rev. 3A l
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Question Number 58
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Which of the following is the power supply to 1EW-P-3A, Motor Driven Aux Feed Pump?
A. AE 4KV Bus.
C. BN 480V Bus.
D. 9P 480V Bus.
POINTS: 1.00 ANSWER: A.
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REFERENCES: 10M-24.3.C - Issue 4, Rev. 6 I ILP-SOS-24.1 OBJECTIVE: 2 NUMBER: 1-97-043, New JTA f: I K/A f: 3.05.061.000.K2.02 K/A IMPORTANCE: 3.7/3.7 Final Revision. Rev. 3A I
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, QuestionNumber 59 j i >
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Which of the following would indicate a reduction of total AW flow capability, i following an automatic Low PZR Pressure Rx Trip from 100% power?
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A. AFW Turbine Steam Supply Trip Valve-[TV-1MS-105A] is OPEN.
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B. Annunciator [A7-7), "STM UNAVAILABLE TURB DRIVEN FEED PP" is NOT in l, alarm.
l C. 1B S/G A m Flow Control Valve (MOV-1 W-151C) is OPEN.
I D. Motor Driven A W Pump [1 W-P-3A] Recirc Valve [FCV-1 W-103A) is OPEN. l
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POINTS: 1.00 i ANSWER:- D.
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l REFERENCES: 10M-24.2.B - Issue 4, Rev. O i ILP-SOS-24.1 OBJECTIVE: 4 NUMBER: 1-97-044, New
]
! JTA #:
- K/A #
- 3.05.061.000.A3.01 K/A IMPORTANCE: 4.2/4.2 i
Final Revision, Rev. 3A
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Question Number' 60
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Which of the following is the initial method of restoring Rx Plant River Water Header pressure in the event of a COMPLETE loss of the Normal River Water Intake structure?
A. . Aligning Unit 2 'B' ' Service Water Header to supply Unit 1 'A'
.
Rx Plant-River Water Header.
B. Aligning the Diesel Driven Fire Pump (15'-P-2) to supply Unit 1 ' A' Rx 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 Pamps to supply that pumps respective Unit 1 Rx Plant River Water Header.
POINTS: 1.00 ANSWER: C.
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I REFERENCES: lOM-30.4.AAC ILP-SQS-30.2 OBJECTIVE: 3 NUMBER: 1-97-045, New JTA 9:
K/A #: 3.05.076.000.Kl.21 K/A IMPORTANCE: 2.7/2.9 Final Revision. Rev.3A
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Question Number 61
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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 l temperature.
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B. Progranning of SGWL to reduce the total mass in the S/G's at high power levels.
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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.
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POINTS: 1.00 ANSWER: A.
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i REFERENCES: BVPS UFSAR Appendix. D; Pg. D.1-3 - Rev. 1, and Pg. D.1-4 - Rev. 4.
ILP-SOS-25.1 OBJECTIVE: 5 NUMBER: 1-97-046, New 4
, 1 JTA 4:
$ K/A 6: 3.05.000.040.EK1.06 K/A IMPORTANCE: 3.7/3.8
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Final Revision, Rev.3A r
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QuestionNmnber 62
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, The Unit is at 100% power with all systems in their_at-power, NSA configurations. At a MINIMUM, which of the following control switches, needs to be placed in the CLOSE position to prevent 1A S/G from feeding the Main Steam '
Manifold? -
A. TV-1MS101A, 1A SG MAIN STEAM A TRN TRIP VLV.
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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, lA MAIN STEAM LINE PRE-NRTRN DRAIN ISOL VLV.
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ANSWER: A.
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l REFERENCES: 10M Figures 21-1 Rev. 10 and 21-06 Rev. 4 ILP-SQS-21.1 OBJECTIVE: 3 NUMBER: 1-97-047, New JTA #
K/A #: 3.05.000.040.EA1.04 K/A IMPORTANCE: 4.3/4.3 Final Revision, Rev. 3A
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' Quation Number 63
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- Assuming all turbine vibration readings are normal, which of the following-sustained conditions would require a manual u tbine and/or Rx Trip?
l Main Generator Condenser ,
Output (MW) Backpressure (In. Hg Abs.)
1. 230 4.0
. 2. 700 4.5
l-3. 800 6.5 4. 210 2.5 A. 1, 2 and 4 ONLY.
B. 1 and 3 ONLY.
C. 3 ONLY. 1
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D. 2 and 4 ONLY.
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POINTS: 1.00 ANSWER: C.
Provide the following references: 10M-26.4.AAS " Condenser Vacuum Low" alarm response procedure.
REFERENCES: 10M-26.4.AAS - Issue 3, Rev. 3 ILP-SQS-26.2 OBJECTIVE: 9d NUMBER: 1-97-048, New l- JTA f:
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l K/A #: 3.05.000.051.EA2.02 K/A IMPORTANCE: 3.9/4.1 Final Revision, Rev.3A
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Question Number 64 -
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Control for the IC S/G 'A' Train Auxiliary Feedwater Regulating Valve [MOV-1W-151B],- has- been transferred to the Emergency Shutdown Panel (CDP) . What actions are necessary to transfer control of (MOV-1 W-151B] back to the Main Control ,
Room Benchboard.(BB-C)?
A. Operate the control switch on BB-C for (MOV-1W-151B] out of the NORMAL positit,n.
~
'B. Depress the Transfer Pushbutton for [MOV-1 N-151B] on the SDP.
C. Reset the Master Reset Transfer Relay on the SDP.
D. Reset the SDP Transfer Relay for [MOV-1 W-151B] at the respective Aux-Relay Panel.
.
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l REFERENCES: 10M-24.1.D - Issue 4 Rev. 1 ILP-SGS-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 l
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I Final Reviskm, Rev.3A l
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Question Number 65 i
' 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.
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B. Containment Purge Supply and Exhaust Fans will trip, and the Purge Supply and Exhaust Dampers will close.
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! C. The Supplemental Leak Collection and Release System (SLCRS) will be aligned to bypass the Main Filter Bank and provide an elevated ~ release flowpath.
D. The Containment Purge Exhaust will be aligned to the SLCRS and then-filtered through the Main Filter Banks.
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REFERENCES: 10M-44C.1.B - Issue 4, Rev. 0 ILP-SOS-44C (2382) OBJECTIVE: 3 NUMBER: 1-97-050, New
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JTA #:
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K/A ft 3.06.022.000.K4.03 K/A IMPORTANCE: 3.6/4.0 Final Revision, Rev. 3A (
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Questi nNumber 66 I
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I 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 *
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' Partial Pressure would be (1) psia and the actual containment air
"
partial pressure (2) meet the Technical Specification requirement. ,
,
i (1) (2) .
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' A. 9.7 would NOT
, B. 9.7 would j
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C. 9.9 'would NOT t
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D. 9.9 would t
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POINTS: 1.00-ANSWER: A. 'f
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Provide the following references: 'Jnit 1 Technical Specification figure 3.6-1 .
REFERENCES: BVPS TS, Figure 3.6-1, Amendment No. 174 ILP-SOS-12.1 OBJECTIVE: 10 NUMBER: 1-97-051, New JTA #:
K/A is 3.06.022.000.A1.04 K/A IMPORTANCE: 3.2/3.3 Final Revision. Rev.3A
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QuestionNumber 67 i
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.The bases for the OPERABILITY of the Containment Quench and Recirculation Spray Systems is to ensure containment depressurization and subsequent return to;
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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. j 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. l l
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POINTS: 1.00 '
ANSWER: C.
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i REFERENCES: BVPS Unit 1 TS, 3/4.6.2 Bases, Amendment No. 200
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ILP-SOS-13.1 OBJECTIVE: 11 NUMBER: 1-97-052, New JTA 9:
K/A # 3.06.026.000.G06 K/A IMPORTANCE: 2.5/3.8 Final Revision. Rev.3A
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- QuestionNumber 68
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i 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 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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A. H2 gas that has accumulated in the Pzr gas space.
B. H2 released from the assumed 10% failed fuel pins.
C. H2 gas that has come out of solution from the RWST injection water. '
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D. H2 produced from the corrosion of the RCS Inconell clad materials.
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POINTS: 1.00 ANSWEP: A.
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I REFERENCES: UFSAR Section 14.3, page 14.3-39, Rev. 12.
ILP-SQS-ATA4.2 OBJECTIVE: 1 NUMBER: 1-97-054, M-0006 JTA 8:
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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e Question Number 69
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Given the followings e The 60 gpm Ltdn Orifice Cnmt Isol Viv [TV-1CH-200B), is to be stroked closed and timed by de-energizing [SOV-1CH-200] as part of an OST.
e One 60 gpm, and the 45 gpm Letdown Orifices are in service.
e The LP Ltdn Back Press Reg Viv [PCV-1CH-145), is in MANUAL.
In order to maintain letdown system stability when [TV-1CH-200B] is going CLOSED, [PCV-1CH-145] should be throttled (1) , to (2) letdown pressure as indicated on the BB-A Letdown Pressure Indicator [PI-1CH-145).
(1) (2)
A. CLOSED, RAISE B. CLOSED, LOWER C. OPEN, RAISE D. OPEN, LOWER POINTS: 1.00 ANSWER: A.
REFERENCES: 10ST-47.3A - Issue 4, Rev. 15.
ILP-SQS-7.1 OBJECTIVE: 10 NUMBER: 1-97-055, New JTA #:
K/A i: 3.06.103.000.A4.01 K/A IMPORTANCE: 3.2/3.3 Final Revision, Rev. 3 A
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QuestionNumber 70
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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 Manway Doors.
3. Equipment Hatch Emergency Air Lock Doors.
i. A. I and 2.
B. I and 3.
C. 2 and 3. .f D. 1, 2 and 3.
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POINTS: 1.00 ANSWER: B.
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l 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 1:
i K/A ft 3.06.000.069.EK2.03 K/A IMPORTANCE: 2.8/2.9 )
i Final Revision. Rev,3A
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- 1 Question Nmnber 71 l
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! Under which of the following conditions would Technical Specifications be j 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.
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A. e 1A S/G secondary side manway is removed, o The 1A S/G MS Trip and Bypass Valves [TV-lMS-101A) and (MOV-1MS-101A] are CLOSED, and l e All 1A S/G pressure transmitter sensing lines are removed I from the steam lines to inctall new pressure taps.
B. * CRDM Shroud Cooling Coil relief valve [RV-1CC-113A) is removed, e CRDM Shroud Cooling outlet isolation valve [1CCR-188) is CLOSED, and l e CRDM Shroud Cooling Coil Inlet isolation valve [MOV-1CC- l 111A) is CLOSED. l C. * The CVCS charging line is removed from the Regen Heat Exchanger, e The charging line vent valve [1CH-390) is locked OPEN, and e Charging Line Containment Isolation Valve IMOV-1CH-289) is l CLOSED but inoperable due to its motor operator leads being lifted.
D. * Containment Equipment Hatch installed with three closure l bolts, and l
- both doors on the Emergency Personnel Access Hatch CLOSED, I and I e All containment purge dampers are CLOSED. l 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.
REFERENCES : lOM 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 #:
! K/A i: 3.06.000.069.EA2.02 K/A IMPORTANCE: 3.9/4.4 l
Final Revision. Rev. 3A
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Question Number 72 t
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All of the following Emergency Diesel Generator (EDG) conditioes will prevent the EDG output breaker from closing during a loss of all offsite power EXCEPT a(n)
A. Electrical Engine Overspeed Trip Signal, t
B. Generator Overcurrent Signal.
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C. Generator Output Voltage at 2.lKV.
D. Engine Oil Low Pressure Trip Signal.
POINTS: 1.00 ANSWER: D.
REFERENCES: 10M-36.1.E - Issue 4, Rev. 1.
ILP-SQS-36.2 OBJECTIVE: 6 NUMBER: 1-97-058, New JTA 6:
K/A i: 3.07.062.000.K3.02 K/A IMPORTANCE: 4.1/4.4 Final Revision. R v. 3A
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Question Number _73 e
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. l POINTS: 1.00 ANSWER: B.
RE FERENCES : lOM-39.1 - Issue 4, Rev. 0 lLP-SOS-39.1 OBJECTIVE: 1 NUMBER: 1-97-059, New JTA #:
K/A #: 3.07.063.000.Kl.03 K/A IMPORTANCE: 2.9/3.5 Final Revision. Rev. 3A
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Question Number 74
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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.
POINTS: 1.00 ANSWER: D.
REFERENCES: 10M-36.1.E - Issue 4, Rev. 1 ILP-SQS-36.2 OBJECTIVE: 7 NUMBER: 1-97-060, M-0664 JTA #:
K/A i: 3.07.064.000.A4.01 (064A4.01) K/A IMPORTANCE: 4.0/4.3 Final Revision, Rev. 3A
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Question Number 75
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Given the followings e A loss of all emergency 4KV AC power in conjunction with a steam line break inside containment has just occurred.
- There are indict.tions of a steam void in the Rx Vessel head.
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- All SG narrow range water levels are still <54.
- Containment pressure is 8.5 psig.-
- The 1A Emergency Diesel Generator has just been started locally.
Which of the following pumps should be started / verified running FIRST, and the bases for this action?
A. Charging /HHSI pump, to collapse the Rx Vessel head void.
B. Motor Driven Aux. Feedwater pump, to establish a Heat Sink.
C. River Water pump, to provide cooling to the EDG's.
D. Quench Spray pump, to reduce containment pressure.
POINTS: 1.00 ANSWER: C.
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REFERENCES: lOM-53B.4.ECA-0.0 - Issue 1B, Rev, 3.
lLP-SOS-53.3 OBJECTIVE: 4 NUMBER: 1-97-061, M-0561 JTA #:
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K/A #: 3.07.000.055.EA1.06 (064Kl.02) K/A IMPORTANCE: 4.1/4.5 l
Final Revision, Rev. 3A !
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QuestionNumber 76 4 f ECA-0.0 " Loss of All Emergency 4KV AC Power" directs the operator to locally close the Seal Water Return Containment Isolation Valve (NOV-1CH-381). Which of f
' the following is the bases for performing this step?
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A. To prevent over pressurizing and possibly rupturing the VCT.
r 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.
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ANSWER: B.
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K/A #: 3.11.000.059.EK3.04 K/A IMPORTANCE: 3.8/4.3
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i QuestionNumber 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 'STBY-ISOL' position.
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This switch position transfers the Vital Bus power source from the
! (1) to the (2) and when the normal power source is l restored, allows the (3) transfer back to the normal power source.
(1) (2) (3)
A. AC Static Line Inverter Output Automatic Regulator B. AC Static'Line Inverter Output Manual Regulator C. Inverter Output AC Static Line Automatic Regulator D. Inverter Output AC Static Line Manual Regulator POINTS: 1.00 M4SWER: D.
REFERENCES: 10M-38.4.AAC - Issue 4, Rev. 1, and Figure 38-3 - Issue 3, Rev. 6.
1LP-SOS-38.1 OBJECTIVE: 2 NUMBER: 1-97-064, New JTA f:
K/A 9: 3.07.000.057.EK3.01 K/A IMPORTANCE: 4.1/4.4
- FinalRevision.Rev.3A
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QuestionNumber 78 1.
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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.
e REFERENCES: 10M-39.1.D - Issue 4, Rev. 1
'1LP-SOS-39.1 OBJECTIVE: 2 NUMBER: 1-97-065, New JTA f:
! K/A 9: 3.07.000.058.EK1.01 K/A IMPORTANCE: 2.8/3.1 Final Revision.Rev.3A
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QuestionNumber 79
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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? i i
i The Engine Low Oil Pressure Trip is A. automatically bypassed for the first 15 seconds of operation to allow ,
oil pressure to build up. ,
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B. ' cleared when the operator starts the auxiliary oil pump prior to
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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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Provide the following references: lOM-34.4.L " Placing Diesel Air Compressor Into Service".
REFERENCES: .OM-34.4.L - Issue 4, Rev. 1 ILP-SQS-34.1 OBJECTIVE: 5 NUMBER: 1-97-067, New
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JTA #:
K/A f: 3.08.000.065.EA1.04 K/A IMPORTANCE: 3.5/3.4 l:
Final Revision. Rev.3A
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~ Question Number 80
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What are the affects on the Solid State Protection System (SSPS) if the 220vac ,
Vital Instrument Bus II is de-energized?
Assume initial conditions of 100% power, with all systems in their at-power NSA configurations.
A. ALL of the Train B, Output Bay Slave Relays will NOT function on a i Safety Injection signal. '
i 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 n: lays will de-energize resulting in a
'Rx Trip and Safety Injection.
D. Train B W function as designed due to the auctioneered power supplies to he Logic and Output Bays.
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REFERENCES: 10M Figure 1-43 - Rev. 8 ILP-SOS-1,2 OBJECTIVE: 2 NUMBER: 1-97-069, New JTA f:
K/A #: 3.09.012.000.K2.01 K/A IMPORTANCE: 3.3/3.7 Final Revision, Rev.3A
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e QuestionNumber 81 l'
l Given the following:
e The Unit is in Mode 6, conducting a core off-load. I t
e N-43 power range instrutaent is on clearance for maintenance. I i e N31 and N32 Source Range Instruments HV Manual On/Off switches are in^
! NORMAL. .
e All other systems are in their normal line-ups for the current Mode of i operation. '
e .The 120vac Vital Bus IV is inadvertently de-energized.
Following the loss of the 120Vac Vital Bus IV, Core Alterations may j l 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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ANSWER: C. l
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Provide the following references: Unit 1 Technical Specification section 3/4 9
- Refueling Operations.
REFERENCES: 10M-1.5.B.1, Table 1-2 - Issue 2, Rev. 8 & TS 3.9.2. Amend No. 175. ,
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ILP-SOS-2.1 OBJECTIVE: 2 NUMBER: 1-97-070, New JTA i:
K/A #: 3.09.015.000.K2.01 K/A IMPORTANCE: 3.3/3.7 Final Revision.Rev. 3A
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e Questi:nNumber 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:
- N31, 120 cps ;
e N32, 130 cps )
e 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 for this condition?
A. Adjust the N36 high flux trip setpoint to the current equivalent of 25% I rated thermal power prior to raising power above P-6. l l
B. Adjust the N36 high flux trip setpoint to the current equivalent of 25% ,
rated thermal power prior to raising power above 5% rated thermal power. I C. Place N36 in the tripped condition when greater than P-10 and the low power trip setpoints have beGa blocked.
D. Place the N36 ' Lev' . Trip Bypass' switch in the ' BYPASS' position prior to I'
exceeding P-6.
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l Provide the following references: Unit 1 Technical Specification 3.3.1.1, Table 3.3-1.
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l REFERENCES: Unit 1 TS 3.3.1.1, Table 3.3-1 - Amendment No.- 195
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1LP-SOS-TS OBJECTIVE: 1 NUMBER: 1-97-071, M-0270 JTA 0 K/A #: 3.09.015.000.G05 (2.2.22) K/A IMPORTANCE: 3.3/3.8
' Final Revision. Rev. 3A
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QuestionNumber 83
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l Following a spurious Rx trip from 100% powers e The turbine trip was delayed for five seconds, causing T. to drop to 5 4 0*F.
- Two cooling tower pumps have tripped.
e 1A RCS loop hot leg temperature instrument has failed high.
Assuming no operator action, which one of the following describes how RCS T.
Will be maintained?
A. At 547*F by the ' steam dumps with ONLY the cooldown valves [PCV-1MS-106A,'B, & B1] armed.
B. At 547'r by the steam dumps with the first 2 banks of the steam dump valves armed. ]
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C. By ONLY the atmospheric steam dump valves (PCV-1MS-101 A, B, & C)
cycling at their trip open setpoint. J D. At 543*F by the condenser steam dumps cycling open due to the i failed high Tues instrument and closed by the Lo-Lo T... interlock. )
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ANSWER: B. I REFERENCES: 10M Figure 21.1.D - Issue 4, Rev. 1.
ILP-SQS-21.1 OBJECTIVE: 3, 4 NUMBER: 1-97-072, M-0188 i
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JTA #:
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-K/A f: 3.09.016.000.K3.03 (041K4.09) K/A IMPORTANCE: 3.0/3.1 l
l Final Revision. Rev. 3A
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.e Qu:stion Number 34
Which of the following do the' Emergency Operating Procedures use as an indication of Natural Circulation flow occurring within the RCS7
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.
l B. 1, 2, and 4. l C. 2 and 3.
D, 2,.3 and 4 ,
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POINTS: 1.00 ANSWER: C.
REFERENCES: 10M-53.A.1 Attachment 2-G - Issue IB, Rev. 1.
1LP-SOS-53.2 OBJECTIVE: 11 NUMBER: 1-97-073, M-01~11 !
JTA 6:
K/A 6: 3.09.017.020.A3.01 (E09EA1.20) K/A IMPORTANCE: 3.6/3.8 Final Revision. Rev.3A
e QuestionNumber 85
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Which of the following automatic actions should occur if the Fuel Pool Bridge Crane Radiation Monitor (RM-1RM-207), were to fail high while raising a spent fuel assembly out of the fuel transfer cart?
All (1) fuel handling crane movement is (2) and a High-High activity alarm (3) actuated.
(1) (2) (3) _
A. UPWARD Automatically IS Stopped B. UPWARD and Automatically IS DOWNWARD Stopped C. UPWARD and Unaffected IS NOT DOWNWARD D. UPWARD and Unaffected IS DOWNWARD l
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POINTS: 1.00 ANSWER: D.
i REFERENCES: 10M-43.1.E - Issue 4, Rev. 4.
ILP-SQS-43.1 OBJECTIVE: 6 NUMBER: 1-97-074, New JTA I:
K/A #: 3.09.072.000.K3.02 K/A IMPORTA' ICE: 3.1/3.5 Final Revision, Rev. 3 A
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QuestionNumber 86
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Which of the following lists of switch positions is the 100% power NSA switch lineup for the N31 Source Range Nuclear Instrument Drawer?
HV MANUAL LEVEL HIGH FLUX 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.
I REFERENCES: 10M-2 3.C - Issue 4, Rev. 2.
ILP-SOS-2.1 OBJECTIVE: 3 NUMBER: 1-97-075, New
1 JTA #:
( K/A #: 3.09.000.032.EK2.01 K/A IMPORTANCE: 2.7/3.1
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Final Revision. Rev. 3A
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e Qu:stionNumber 87 e
N35, Intermediate Range Nuclear Instrument Channel I, compensating voltage is set excessively HIGH. Following a Rx trip, N35 will indicate l
A. HIGH, preventing the source range from automatically energizing, j
B. HIGH, and the source range will be energized when N36 is <P-6. l l
C. LOW, and the source range will be energized when N35 is <P-6. I I
D. LOW, and the source range will be energized when N36 is <P-6.
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POINTS: 1.00 ANSWER: D.
REFERENCES: lOM-2.1.C - Issue 4, Rev. 1.
ILP-SQS-2,1 OBJECTIVE: 4&S NUMBER: 1-97-076, M-0030 JTA 8:
K/A #: 3.09.000.033.EA2.ll (015K4.07) K/A IMPORTANCE: 3.1/3.4 Final Revision. Rev. 3A
e QuestionNumber 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 low during a steam plant heatup.
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REFERENCES: BVPS UNIT 1 TS 3.3.3.1, Table 3.3-6 2.c.v - Amendment No. 59 1LP-SOS-43.1 OBJECTIVE: 7 NUMBER: 1-97-077, New JTA #: >
K/A 8: 3.09.000.061.G08 K/A IMPORTANCE: 2.6/3.3
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Final Revision. Rev. 3A
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l Question Number 89 l s
With the Unit at 100% power and all systems in their at-power, NSA !
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.
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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RE FERENCES : 10M-15.1.C - Issue 4, Rev. 1 ILP-SOS-15.1 OBJECTIVE: 3&9 NUMBER: 1-97-078, New JTA 4:
K/A 4: 3.10.008.000.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 AN') seal injection flow can be maintained.
B. During a SGTR, AFTER the RCS deptessurization has commenced, if the RCS/ Highest SG D/P drops to <l50 paid.
C. During a response to Inadequate Core Cooling, if high pressure injection flow AND an adequate heat sink CANNOT be established.
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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POINTS: 1.00 ANSWER: C.
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RE FERENCES : lOM-53A.l.2-C - Issue IB, Rev. 2 (Note)
ILP-SQS-53.3 OBJECTIVE: 4 NUMBER: 1-97-080, New JTA #:
K/A 0: 3.10.000.026.EK3.03 K/A IMPORTANCE: 4.0/4.2 Final Resision. Rev. 3A
I Question Number 91
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4 6 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.
4 e 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?
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 rate determined by the NSS/ANSS in accordance with 10M-53C.4, AOP 1.51.1, " Emergency shutdown".
C. Restore level to the i rmal range by operating [TV-INS-101), Neutron Shield Tank Makeup Valve.
D. Manually trip the reactor and proceed to 10M-53A.1, E-0 " Reactor Trip or Safety Injection."
POINTS: 1.00 ANSWER: D.
Provide th( following references: Window [Al-50), 10M-9.4.AAB "Incore Instrwment 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-SOS-15.1 OBJECTIVE: 10 NUMBER: 1-97-081, New JTA 8:
K/A I: 3.10.000.026.G05 K/A IMPORTANCE: 3.3/3.4 Find Revision, Rev. 3A j j
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QuestionNumber 92
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e 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. 1 and 2.
B. I and 3.
C. 2 and 3.
D. 1, 2 and 3.
POINTS: 1.00 ANSWER: C.
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REFERENCES: 10H-20.1.B . Issue 4, Rev. 3.
ILP-SOS-20.1 OBJECTIVE: 167 NUMBER: 1-97-083, New JTA #
K/A ft 3.11.033.000.K4.05 K/A IMPORTANCE: 3.1/3.3 Final Revision. Rev,3A
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Question Number 93 :
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I Which of the following is a function of the Refueling Manipulator Crane?
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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. :
C. Provide the motive force to raise and lower the Burnable Poison Rod Assembly Handling Tool.
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D. Provide the motive force to raise and lower the Reactor Upper f Internals.
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POINTS: 1.00 ANSWER: A.
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l REFERENCES: 1RP-10R-3.3 - Issue 0, Rev.'O.
lLP-SQS-6.13 OGJECTIVE: 9 NUMBER: 1-97-084, New JTA #
. K/A #: 3 ll.034.000.G07 K/A IMPORTANCE: 2.5/3.0 t
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Question Number 94 s
Which of the following are sources of hydrogen that could accumulate in the Waste Gas Disposal System?
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.
A. I and 3.
B. 1 and 4.
C. 1, 3 and 4.
D. 2, 3 and 4.
I POINTS: 1.00 ANSWER: B.
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l REFERENCES: 10M-19.1.C - Issue 4, Rev. O.
1LP-SQS-19.1 OBJECTIVE: 2 NUMBER: 1-97-086, New JTA #:
K/A #: 3.11.071.000.K5.03 K/A IMPORTANCE: 2.3/2.9 i
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Question Numter 95
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L 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 from 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.
l C. 1, 3 and 4 ONLY. I D. 1, 2, 3, and 4.
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L REFERENCES: ~10M-56.B.2 - Issue 4, Rev. O.
1LP-SOS-33.1 OBJECTIVE: 3 NUMBER: 1-97-087, New
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JTA f:
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K/A f: 3.11'086.000.K5.04
. K/A IMPORTANCE: 2.9/3.5
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QuestionNumber 96 I i
i What is the bases for the Technical Specification requirement that at least 23 i 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, ,
I B. In the event one train of RHR is lost, an adequate heat sink is available.
C. The refueling operators can perform a full core off-load without i'
exceeding their 10 CFR 20 exposure limits.
D. In the event of a fuel element rupture, 99% of the assumed 10% iodine gap activity released is removed.
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POINTS: 1.00 ANSWER: D.
f REFERENCES: Unit 1 Technical Specification 3.9.10 Bases, Amendment No. 175. l ILP-SOS-6.13 OBJECTIVE: 12 NUMBER: 1-97-088, New JTA #:
K/A f: 3.11.000.036.G04 K/A IMPORTANCE: 2.6/3.8
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Final Revision, Rev. 3A -
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'Wu QuestionNumber 97
't Which of the following events is required to be reported to the NRC within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />?
A. A contract mechanic.is found to be intoxicated inside the protected
. area.
I l B. A planned rac.iological liquid release was performed and 2 days later found to be in excess of 10CFR20 limits.
i C. A Unit shutdown is commenced due to the INOPERABILITY of BOTH l Emergency Diesel Generators.
D. A fire has occurred in the Site Engineering Building that took'13 minutes to extinguish.
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l Provide the following references: NPDAP 5.1, REPORT REQUIREMENTS.
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REFERENCES: NPDAP 5.1 Rev. 5.
1LP-SOS-48.1 OBJECTIVE: 19 NUMBER: 1-97-089, New 5 l
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f K/A 8: 3.11.000.059.G02 K/A IMPORTANCE: 2.6/3.9 i
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. Question Number' 98 '
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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? 5
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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.
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.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 l Main Filter Bank inlet AND Bypass dampers. I
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POINTS: 1.00 ANSWER: A.
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REFERENCES : 10M-43.5.B.2 - Issue 4, Rev. 1.
1LP-SOS-43.1 OBJECTIVE: 6 NUMBER: 1-97-090, New I JTA #1 i
K/A #: 3.11.000.060.EK2.02 K/A IMPORTANCE: 2.7/3.1
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l 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?
The straight hose stream _
A. gets the rater 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. l i
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D. uses less water thus minimizes the potential for a re-flash.
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REFERENCES: 10M-56B56.B.2 - Issue 4, Rev. O and LP 9339.
11.P-SOS- 933 9 OBJECTIVE: 20 NUMBER: 1-97-091, M-0154 JTA #:
K/A #: 3.11.000.067.EKl.02 (2.4.25) K/A IMPORTANCE: 2.9/3.9 Final Revision, Rev. 3 A
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a ~ i_ t QuGstion Number 100 +
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l e A rapid (St/ min) power reduction from 1004 to 75% was performed due to i l
Grid instabilities. ;
Power has been stable at 75% for seven hours. I e The results.from the RCS chemistry samples taken four hours after power was. stabilized at 75%, reveal the followings ;
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e Dose Equivalent I-131 (DEI) is 97 uCi/gm.
- Gross coolant activity is 18 uCi/gm.
- The 100/E-Bar limit is 250 uC1/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?
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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.v.
<500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, t
k D. The 100/E-Bar limit has been exceeded, 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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Provide the following references: Technical Specification 3.4.8, Specific l Activity. )
REFERENCES : Unit 1 Technical Specification 3.4.0, Amendment No. 102. ;
ILP-SOS-6.5 OBJECTIVE: 8 NUMBER: 1-97-092, New [
JTA #; j i
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l- Final Revision. Rev.3A I l
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i Attachment 3
BV-2 WRITTEN EXAM COMMENTS AND NRC RESOULTION Question #9 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.
NRC Resolution: Agree with facility comment in that the correct answer should be "B". ;
Answer key will be changed accordingly, i i
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Attachment 4 SIMULATION FACILITY REPORT
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l Facility Licensee: Beaver Vallev Unit 1 and Unit 2 Facility Docket Nos: 50-334 and 50-412 Operating Tests Administered from: March 17-21 and Aoril 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
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]j 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.
l No simulator deficiencies, that affected the scenario examinations, were identified for either simulator during the conduct of the examinations.
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