IR 05000327/1998302
| ML20207M716 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/12/1999 |
| From: | Hopper G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207M675 | List: |
| References | |
| 50-327-98-302, 50-328-98-302, NUDOCS 9903190166 | |
| Download: ML20207M716 (137) | |
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J U. S. NUCLEAR REGULATORY COMMISSION
REGION 11
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Docket Nos.: 50-327,50-328 License Nos.: DPR-77, DPR-79 Report No.:
50-327/98-302 and 50-328/98-302
Licensee:
Tennessee Valley Authority
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Facility:
Sequoyah Nuclear Plant Units 1 and 2 i
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Location:
Sequoyah Access Road Hamilton County, TN 37379 i
Dates:
September 28 - October 1,1998 Written Exam October 5,1998 Examiners:
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George T. Hopper #7 Chief License Examiner Jonathan Bartley, License Examiner Approved by:
Thoma(A.'Peebles, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety
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Enclosure 1
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9903190166 990312 PDR ADOCK 05000327 V
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EXECUTIVE SUMMARY NRC Examination Report No. 50-327/98-302 and 50-328/98-302 During the period September 28 through October 1,1998, NRC examiners conducted an announced operator licensing initial examination in accordance with the guidance of Examiner Standards, NUREG-1021, Interim Revision 8. This examination implemented the operator licensing requirements of 10 CFR @55.41,655.43, and S5.45.
Operations Four Senior Reactor Operator (SRO) candidates received written examinations and
operating tests. NRC licensing examiners administered the operating tests during the period September 28-30,1998. The written examination was administered by members of your training staff on October 5,1998.
Candidate Pass / Fail
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i Fail
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0 The NRC concluded that significant improvement was needed by the licensee in the
areas of written examination review and JPM question development and review.
Increased emphasis needs to be placed in ensuring the technical accuracy of examination materials.
The examiners concluded that candidate performance on the written examination was
weak. Overall performance on the operating test was satisfactory with some significant weaknesses noted in the areas of diagnosing events and understanding integrated plant system response.
The facility initially proposed an examination that was inadequate; however, they
corrected the deficiencies and provided a good final product.
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Report Details Summary of Plant Status During the period of the examinations Unit 1 was in an outage and Unit 2 was at 100 percent power.
l. Ooerations
Operator Training and Qualifications 05.1 General Comments NRC examiners conducted regular, announced operator licensing initial examinations during the period September 28 through October 1,1998. The written examination was adrninistered by members of your training staff on October 5,1998. Four SRO upgrade applicants received written examinations and operating tests. NRC examiners administered examinations developed by the licensee's training department, under the
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requirements of an NRC security agreement,in accordar.ce with the guidelines of the Examiner Standards (ES), NUREG-1021, interim Revision 8.
05.2 Pre Examination Activities a.
Scope
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l The NRC reviewed the licensee's examination submittal using the criteria specified for examination development contained in NUREG 1021 Interim Rev 8.
b.
Observations arid Findinas
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The licensee developed the SRO written examination, one Job Performance Measure (JPM) set, and two dynamic simulator scenarios for use during this examination. All materials were submitted to the NRC on time. NRC examiners reviewed, modified, and approved the examination prior to administration. The examiners conducted an on-site l
preparation visit during the week of September 16,1998 to validate examination j
materials and to become familiar with the details required for examination administration.
(1) Written Examination Development l
1, The written examination was submitted on time. The organization of the l
examination with some of the reference material attached expedited the l
examination review process. However, the initial exam submittal did not meet I,
the criteria specified in NUREG 1021 Interim Rev. 8," Operator Licensing Examination Standards for Power Reactors."
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Seventy-five questions were reviewed and edited by the NRC. Substantive L
changes were made to a majority of these questions which required additional
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unplanned expenditure of NRC resources. The most notable deficiencies included four questions that had no correct answer, three questions that had two correct answers, and two questions that had incorrect answers identified.
Several questions had low operational validity and/or low discriminatory value to j
be used as test items. In addition, approximately one third of the questions reviewed contained distractors (choices) that were not plausible and which could be easily eliminated.
Due to the poor quality of the written examination, the NRC chief examiner and the licensee's representatives held a meeting at the Region ll Office to discuss the specific deficiencies of each question and to develop a corrective action plan to correct the examination. The licensee was requested to review and upgrade 25 questions and replace or rewrite six additional questions. The licensee edited these questions and presented them during a second meeting with the chief examiner. The chief examiner found that the licensee's second review of the questions was effective, in that, they correctly identified deficiencies contained within these questions and made appropriate cc.rections. The fundamental guidelines that were found deficient concerned, technical accuracy and adherence to the development guidelines to ensure the examination was a
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valid means of measuring an operator's competence.
The final version of the original written examination met the criteria specified in NUREG 1021, Interim Rev. 8, and was administered without delay. However, a few deficiencies remained. Three post examination comments were submitted following examination administration and are included in this report as enclosure (3). NRC resolution of these comments is included as enclosure (4). The post examination review resulted in changing the answer in two questions and deleti.ng one question because there was no correct answer.
(2) Operating Test Development The NRC reviewed one JPM set and administrative section for the walk-througn examination. The examiners found the JPMs were at the appropriate level of difficulty and the overall quality of the JPMs was deemed satisfactory. However, some minor technical errors were noted such as the incorrect designation of critical steps, and the quality of the JPM questions and administrative questions was poor. Most of the submitted questions lacked significant operational validity and were non-discriminatory in value or werc direct look-up open reference questions where the answer could easily be found in the allowable reference material. Suitable JPM questions are open reference analysis, synthesis, and application level questions that require higher order cognitive
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thought processes and avoid direct look-ups. Increased attention in this area is needed to comply with the guidelines of NUREG 1021, Interim Revision 8. The licensee successfully corrected the deficiencies noted during the NRC review.
The NRC reviewed two simulator scenarios for the examination. Some minor changes and additions were made to the scenarios to provide the examiners i
Fufficient opportunity to observe Candidates perform the required Competencies.
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Overall. the scenarios were found to be challenging and at the right level of difficulty.
c.
Conclusion The NRC concluded that significant improvement was needed by the licensee in the areas of written examination review and JPM question development and review.
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increased emphasis needs to be placed in ensuring the technical accuracy of examination materials. This is necessary to ensure a fair, accurate and valid measurement of a candidate's skills is achieved during the operator licensing process.
Substantial NRC review and comment was required for the submitted examination to meet minimum requirements.
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05.3 Examination Results and Related Findinas. Observations. and Conclusions a.
Scope The examiners reviewed the results of the written examination and evaluated the candidates' compliance with and use of plant arocedures during the simulator scenarios and JPMs. The guidelines of NUREG-1021, Forms ES-303-3 and ES-303-4,
" Competency Grading Worksheets for integrated Plant Operations," were used as a basis for the operating test evaluations.
b.
Observations and Findinas.
The examir ers reviewed the results of the written examination and found that four of four candidates passed. Overall, SRO candidate performance on the written examination was weak with all candidates achievin9 Orades between 80.8 and 82.8 percent. The licensee conducted a post examination item analysis and determined that three test questions were faulted. In addition, the licensee identified nine questions that three or more candidates answered incorrectly. The examination results indicated generic weaknesses in the knowledge level of the candidates and /or the training program. As described NUREG 1021, ES 403.D.3.a,"If it appears that the training program was deficient, (the licensee should] determine the need for remedial training and/or a program upgrade."
Examiners also identified several weaknesses in candidate performance during the operations portion of the examination. Details of the discrepancies are described in each individual's examination report, Form ES-303-1, " Operator Licensing Examination Report," which have been forwarded under separate cover to the Training Manager.
This will enable you to evaluate the weaknesses and provide appropriate remedial
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training for those operators as necessary.
The NRC examiners observed instances during the dynamic simulator portion of the
examination where candidates had difficulties in correctly diagnosing events and understanding integrated plant response. During one scenario mvolving the loss of 120 VAC inverter and 125V Vital Battery Board 1, two candidate's took over 15 minutes to diagnose the loss of the busse?. They originally pursued a belief that the 1 A SDB was deonergized despi e the fact that voltage indication existed for that bus as well as supply t
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breakers indicating closed. Also, the candidates did not notice the annunciators for loss
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l of the 125V DC bus until 20 minutes after the start of the event. During two scenarios j
l involving a loss of condenser vacuum event, none of the candidate's recognized that condenser pressure was increasing over a seven minute period while they were scanning their boards. The candidates only recognized the problem after receiving the condenser high pressure alarm (2.7 psia), at which point a manual reactor trip was required by procedure. No compensatory measures were taken to avoid the plant trip
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because the candidates were not aware of the increasing trend in pressure. Errors such as these are unusual on initial exams, especially when committed by candidates
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who already hold an RO license.
One examiner observed a candidate (currently licensed RO) exhibit poor procedural usage while performing a JPM during the walk-through examination. The JPM required performing section 8.3 of 0-SO-62-7," Boron Concentraton Control,"Rev. 9. The procedure cover sheet identified that the procedure was " Continuous Use" except for section 5.1. Procedure SSP-2.51, " Rules of Procedure Use,"Rev. 7, stated that,"Where sign-offs are required, sign each step as complete before proceeding to the next step."
During the performance of the JPM, the examiner observed the candidate perform several steps (three to five) and then initial the steps as complete. Because the candidate is a licen5ed RO, this was identified to the training staff and resident inspectors for follow-up as a potential weakness in proedural usage by operations personnel, c.
Conclusion
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The examiners concluded that candidate performance on the written examination was weak. Overall performance on the operating test was satisfactory with some significant weaknesses noted in the areas of diagnosing events - nd understanding integrated plant system response.
V. Management Meetinas X1. Exit PAeeting Summary At the conclusion of the site visit, the examiners met with representatives of the plant staff listed on the following page to discuss the results of the examinations and other issues.
None of the material provided to the examiners was identified by the licensee as proprietary.
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PARTIAL LIST OF PERSONS CONTACTED Licensee
- M. Bajestani, Site Vice President
- T. Deasley, Site Quality Manager
- R. Driscoll, Site Training Manager P. Gass, Operator Training
- J. Hamilton Quality Assurance Assessment Supervisor j
- J. Herron, Plant Manager j
"W. Hunt, Operations Training Manager
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l E. Keyser, Operator Training l
- R. Proffitt, Licensing Engineer
- P. Salas, Licensing Manager
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- T. Smith, Licensing Supervisor NRC M. Shannon, Senior Resident inspector i
D. Starkey, Resident inspector
ITEMS OPENED, CLOSED, AND DISCUSSED Closed
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Enclosure (2)
SIMULATION FACILITY REPORT Facility Licensee: Tennessee Valley Authority - Sequoyah Nuclear Plant Facility Docket Nos.: 50-327 and 50-328 Operating Tests Administered on: September 28-30,1998
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This form is to be used only to report 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 certification or approval of the simulation facility other than to provide information that may be used in future l.
evaluations. No licensee action is required in response to these observations, While conducting the simulator portion of the operating tests, the following items were observed (if none, so state):
ITEM DESCRIPTION No discrepancies were noted.
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l Tennessee Valley Authority, Post Office Box 2000, Soddy Daisy. Tennessee 37379-2000 l
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Nkscud Bajestani l
9 e %ce Pre s h nt t
Smmyan Mehr Rmt j
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October 9, 1998
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Mr. Luis A. Reyes
U.S. Nuclear Regulatory Commission l
Region II
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61 Forsyth St.,
SW, Suite 23T85 l
Atlanta Federal Center l
Atlanta, Georgia 30323-3415 l
f Attention:
Mr. Tom Peebles
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In the Matter of
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Docket Nos. 50-327
Tennessee Valley Authority
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50-328
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SEQUOYAH NUCLEAR PLANT (SQN) - SENIOR REACTOR OPERATOR (SRO)
LICENSING EXAMINATIONS-
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This letter provides our comments concerning questions on the written' licensing examination administered on October 5, 1998.
Specifically, we are requesting your consideration to delete l
one question and accept a different answer on two other
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questions. 'These comments were discussed with the chief i
examiner, George Hopper, on October 6, 1998.
The enclosure.
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request.
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If you have any questions, please call me at (423) 843-7001 or Pedro Salas'at (423) 843-7170.
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Sincerely, p
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M%LCIO
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>Masoud Bay.stani f
l Enclosure cc:
See page 2
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ENCLOSURE (3)
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Nuclear Regulatory Commission
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Page 2 October 9, 1998 Enclosure cc:
Mr.
R.
W.
Hernan, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North
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11555 Rockville Pike Rockville, Maryland 20852-2739
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NRC Resident Inspector
Sequoyah Nuclear Plant i
2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 l
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ENCLOSURE SEQUOYAH NUCLEAR PLANT
SENIOR REACTOR OPERATOR (SRO) LICENSING EXAMINATIONS COMMENTS ON WRITTEN EXAMINATION ADMINISTERED ON OCTOBER 5, 1998
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HLC9809 NRC EXAM CONTESTED QUESTIONS
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QUESTION SUPPORTING No.
EXAM NUMBER ACTION / JUSTIFICATION DOCUMENTATION
SRO-6 TVA's Comment:
SQN Lesson Plan TAA-13, l
RCS pressure at 750 psig does not indicate a large break LOCA. As shown in the new pg. 22. Description of Loss reference material, lesson plan TAA-13 pg. 22, a break of this size is defined as a'
of Reactor Coolant Accident.
i category 3 LOCA which is a Small Break LOCA and Secondary Heat sink not required.
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TVA's Recommendation:
Accept "B" as the ONLY correct answer.
2.
TVA's Comment:
Steam Tables.
This question had a calculation errorin it which resulted in *D* being stated as the correct answer. At 587 deg. F, saturation pressure is 1400 psia which equates to 1385 psig NOT 1415 psig as shown in the answer key. The ONLY correct ar
'is *C'
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(1315 psig) based on the 1 decreasing pressure at which core voiding auld.
INITIALLY form.
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Accept *C" as the ONLY correct answer.
.SRO
TVA's Comment:
FHI-3 pg.15 of 100 (
j As given in the stem of the question the offloaded fuel assembly that is in the Rx side OPL271C137
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upender is only the 2"d fuel assembly out of approved storage locations. (The fuel f
assembly in the Spent Fuel Pit is not counted toward the refueling canal.) As stated in r
FHl-3, Ill, b,3 three fuel assemblies are allowed within the refueling canal Therefore i
fuel assembly (B) does NOT have to be removed from the SFP side upender prior to.
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the next fuel assembly being offloaded.
t TVA's Recommendation:
l DELETE question based on no correct answer.
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6.
Given the following plant conditions:
Reactor trip and Si have actuated
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The operating crew enters E-0," Reactor Trip or SI"
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The operators are unable to establish AFWflow, so they enter FR-H.1, " Loss of i
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Secondary Heat Sink" RCS pressure is now 750 psig and S/G pressure is approximately 890 psig
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The operators are directed by FR-H.1 to enter E-1, " Loss of Reactor or Secondary
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Coolant" Which ONE of the following correctly summarizes plant conditions?
a.
Small Break LOCA in progress; secondary heat sink required.
b.
Small Break LOCA in progress; secondary heat sink NOT required.
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c.
Large Break LOCA in progress; secondary heat sink required.
d.
Large Break LOCA in progress; secondary heat sink NOT required.
Answer; D
K/A:
300011A201
[4.2/4.7)
Reference:
EPM-3-FR-H.1, pages 10 Objective:
OPL271C401, B.2 Level:
Memory
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Description of the loss of Reactor Coolant Accident
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In order to describe the various phenomena that can occur during a LOCA i. is necessary to define LOCA categories depending on approximate break size and safeguard equipment status (one-train operation of the ECCS). The selected LOCA categories that will be described in this section are as follows:
1.
Break diameters less than or equal to 3/8 inch.
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Breakdiametersbetween3/8inchand1 inch;RCSpressurestabilIzes
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above S/G pressure.
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Break diameters between 1 and 2 inches; RCS pressure eventually (> 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) stabilizes below S/G pressure, but above accumulator injection'pressurc.
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4.
Break diameters between 2 and 13-1/2 inches (1 square foot); RCS depressurization occurs along with accumulator injection.
5.
Vapor space breaks such as the inadvertent opening of a pressurizer relief or safety valve.
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Break areas between I square foot and double-ended; RCS rapidly depressurizes to values close to the containment atmospheric pressure.
The small-break transient behavior, modes are summarized in table 1.
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Category 1 is an RCS leak, c~ategories 2 through 4 and category 6 are small-
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break LOCAs,- and category 5 is a large-break LOCA. The intent of this section p
is to describe the general phenomena that would be expected to occur during a LOCA. The values and figures presented are from the Sequoyah Unit 1 Final Safety Analysis Report (FSAR), with the exception of the 1-inch break analyses.
This break is not analyzed in the FSAR. The results are for a typical Westinghouse four-loop plant.
It should be noted that the break-size ranges stated above are approximate and.are. highly dependent on the break location and the performance of the ECCS. Therefore, understanding the physical processes
3730s:4 soeTAA-13 6/90
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C 1990 Westinghotse Electric Corp,
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SEQUOYAH
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UNIT
2 Inch Cold Leg Break RCS Pressure Figure 20
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RCS Pressure
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29.
Given the following piant conditions:
A Unit i reactor trip and safety injection has occurred
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Operators have correctly transitioned to E-1, Loss of Reactor or Secondary Coolant
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Core exit T/C are stable at 587'F
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RCS hot leg temperatures are stable at 577'F
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RCS cold leg temperatures are stable at 535'F
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RCS pressure is 1600 psig and decreasing i
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At which ONE of the following RCS pressures will a void initially form in the reactor vessel head?
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890 psig.
I b.
1015 psig.
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1315 psig.
I d.
1415 psig.
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Answer D
K/A-000009K102- [3.5/4.2)
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Reference:
Steam Tables Objective:
OPL271C385, B.1 Level:
Analysis Source:
New question (Developed 7/24/98)
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- Question #29:
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~587 deg. F = 1400 psia or 1385 psig.
At 587 deg. F Boiling is occurring at 1385 psig.
Of the 4 choices for answers, the 1" pressure at which Core Void will fonn is answer "C".
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t 91.
Given the following conditions:
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Unit 1 is in Mode 6 with core offload in progress.
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l One offloaded fuel assembly has been placed in the RCCA change fixture.
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l One offloaded fuel assembly (A) is being moved by the spent fuel pool (SFP) bridge
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toward its assigned storage cell within the SFP racks.
.. One offloaded fuel assembly (B) is horizontal in the reactor side upender awaiting transfer to the SFP side.
.. The refueling machine has been positioned above the reactor cavity over the next fuel assembly to be offloaded.
Within the following chronological sequence of events, when can the next fuel assembly be offloaded and lifted from its seated position in the reactor cavity?
a.
Immediately, providing that the next fuel assembly to be offloaded is being moved ~~
directly to storage in the RCCA change fixture.
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b.
When the offloaded fuel assembly (B) has been transferred to t!)e SFP side upender in the horizontal position.
c.
When the offloaded fuel assembly (B) has been raised to the vertical position in the SFP side upender, and offloaded fuel assembly (A) has been placed in its assigned SFP storage cell.
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d.
When the offloaded fuel assembly (B) has been removed from the SFP side upender and is in transit to its assigned SFP storage cell.
Answer; D
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K/A:
2.2.26
[2.5/3.7]
Reference:
OPL271C137, page 24; FHl-3, pg.15 Objective:
OPL271C137, B.7 Source:-
New question (Developed 9/8/98)
Level:
Comprehension
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MOVEMENT OF FUEL Rsv: 31
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11.
PRECAUTIONS Continued
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D.
MAIN CONTROL ROOM The applicable control room fuel accountability tag boards shall be kept i
current during major fuel-handling operations to display current fuel
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assembly locations / movement within the affected fuel storage and fuel handling areas. The applicable tag boards will be updated continually by t
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operations and/or Reactor Engineering personnel as each fuel movement f
occurs.
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LIM.lTATIONS A.
The hoist " slow" speed shall be used prior to entering a confined area
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(such as a fuel rack or upender) when bottom fuel nozzle of fuel assembly is approximately 10 inches above the confined area and continue in " slow"
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speed until bottom fuel nozzle is inside the confined area. The hoist
" slow" speed shall also be used prior to bottom of fuel assembly reaching
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bottom of the confined area, (when bottom of fuel assembly is
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approximately 10 inches from bottom) until fuel assembly reaches the
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B.
The maximum quantity of nuclear fuel assemblies allowed out of approved storage locations for Sequoyah Nuclear Plant shall be as listed below- [C.5]
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1.
One unirradiated nuclear fuel assembly shall be allowed within the fuel-handling area, outside of metal shipping containers, or the
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t-new fuel storage vault.
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2.
One nuclear fuel assembly shall be allowed within the spent fuel storage pool boundary when not seated in a storage cell. The
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spent fuel storage pool boundary includes the cask loading area and fuel transfer canal excluding the upender.
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3.
Three nuclear fuel assemblies shall be allowed within the refueling canal. The refueling canal includes the fuel transfer tube boundary, the rod cluster control changing fixture and the upender. This allows for two nuclear fuel assemblies to be in the rod cluster control changing fixture while the third nuclear fuel assembly is being transferred through the fuel transfer tube, is in the upender, or is in transit to or from the reactor cavity.
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4.
One fuel assembly shall be allowed within the reactor cavity.
C.
DO NOT operate the RCS and Spent Fuel Pit below 68 de_grees. Less than 68 degrees is outside the bounds of the criticality analysis. [C.9)
D.
Transportation of loads over the Cask Loading Area while spent fuelis being stored in the Cask Loading Pit must be within the guidelines provided in the FSA....
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OPL271C137 i
Revision 10
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Page 24 of 59 X. LESSON PLAN
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INSTRUCTOR NOTES Q.r.
2.
Stress the following limitations:
,
The hoist " slow" speed shall be used prior to entering a.
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a confined area and continued until the bottom of the assembly is in the confined area. " Slow" speed will
,
also be used as the bottom of the assembly
'
approaches the bottom of the confined area.
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N b.
The rnaximum quantity of nuclear fuel assemblies obj. B.7
[
allowad out of approved storage as listed in this see section xn: c.1 section (to preclude uncovering a fuel assembly in New fuel, sFP. & Cntmt areat the event of a reactor cavity seal failure).
1) One (1) in area of refueling floor where new fuel assemblies are stored or handled.
2) One (1) within the bounds of the SFP (including transfer canal).
,
b 3) Three (3) within the refueling canal: Two in the
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RCCA change fixture and one in the transfer system (transfer cart or either upender) or is in
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transit to or from the Rx cavity.
4) One (1) allowed within Rx Cavity.
c.
Due to possible seal failure on the SFP gate; an obj. 0.7 operator is required to be at the Rx side controls at all times when the Fuel Transfer Cart is on the Rx side.
In addition an operator is required at all times on the i
SFP side when the blind flange and wafer valves are i
open. This operator must be prepared to close the
'
wafer valve.
.
I d.
G.
Discuss / Review FHl-3, Section A, " Transfer of Fuel assemblies use the htest copy of FHl.3 within the Auxiliary Building",
dunns class.
.
1.
Purpose: to provide the guidance to move new fuel assemblies to and/or from the new fuel inspection stand, Ercar tack ofintedocks new fuel storage racks, SFP side upender or SFP storage between Elevator. upender. and racks, sFP Crane.
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I
l Enclosure (4)
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NRC RESOLUTION OF FACILITY RECOMMENDATIONS Question 6: Recommendation accepted. The answer key was changed to reflect (b) as the correct answer.
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i Question 29: Recommendation partially accepted. The licensee provided information stating
'
l that the correct answer was 1385 psig. Since 1385 psig was not one of the 'our choices listed i
in the question, the licensee requested accepting the next lowest value of p;nsure listed on the
question as the pressure at which a void will initially form in the reactor vessel head. While this
is the first listed pressure at which a void would be present, it is not the p; essure at which a void willinitially form in the vessel head. Since no clarification questions wme asked by the -
candidates during the examination on this question, the NRC will accept choice (c) as the correct answer. However, future use of this question as a validated NRC test item will require modification of text and/or answer to be acceptable. The answer key was changed to accept i
choice (c) as the correct answer.
i
. Question 91:_ Recommendation accepted. Review of the test item indicated that there was no absolute correct answer, item (a) was partially correct but contained a (false) statement that l
was not a requirement. Therefore, it is not a correct answer. The answer key was changed to
delete the question from the examination.
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U. k NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR REACTOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME:
FACILITY; Sequoyah REACTOR TYPE:
PWR-WEC4
,
DATE ADMINISTERED:
10/05/98
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INSTRUCTIONS TO CANDIDATE:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer j
sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80% Examination papers will De picked up four (4) hours after the examination starts.
CANDIDATE'S TEST VALUE SCORE
%
100 Points
%
TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature t
SRO Exam NRC 10-5-98 L
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination, the following rules apply:
1.
Cheating on the examination means an automatic denial of your application and could result in
!
more severe penalties.
2.
After you complete the examination, you must sign the statement on the cover sheet indicating that the work is your own and you have not given or received assistance in completing the examination. This must be done after you complete the examination.
3.
Use black ink ONLY to facilitate legible reproductions.
4.
Print your name in the blank provided on the examination cover sheet and each an.swer sheet.
5.
Mark your answers on the answer sheet provided. Do not leave any questions blank.
6.
If the intent of a question is unclear, ask questions of the examiner only.
7.
To pass the examination, you must achieve a grade of 80% or greater.
8.
Each question is worth one point.
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9.
There is a time limit of four (4) hours for completion of the examination.
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10.
Restroom trips are to be limited and only one applicant at a time may leave. You must avoid all j
contacts with anyone outside the examination room in order to avoid even the appearance or possibility of cheating.
11.
When you complete the examination, assemble the completed examination with examination questions, examination aids and answer sheets and give it to the proctor. Remember to sign the statement on the examination cover sheet.
12.
When you are done and have tumed in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA.) If you are found in this area while the examination is still
{
in progress, your license may be denie'd or revoked, j
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i SRO Exam NRC 10-5-98
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l SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 1.
Given the following plant conditions:
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Unit 1 is operating steady-state at 86% power
.
Control Bank "D" is at 196 steps
.
Turbine control is in IMP OUT
.
Which ONE of the following describes how the Overtemperature Delta-T trip setpoint will change if a governor valve f ails closed?
)
,
a.
Setpoint will decrease because pressurizer pressure will increase until pressurizer spray initiates.
'
b.
Setpoint will increase because pressurizer pressure will decrease after Sprays
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initiate.
c.
Setpoint will increase because the RCS Tavg willincrease.
d.
Setpoint will decrease because the RCS Tavg will increase.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 2.
Given the following plant condfjons:
Unit 2 was stable at 100% power
.
Control Bank C (CB C) rod K-10 dropped
.
AOP-C.01, Rod Control System Malfunctions, has been entered
.
ROD CONTROL SYSTEM URGENT FAILURE alarm (M-48, A-6) illuminates when
.
rod motion begins during recovery of rod K-10 Which ONE of the following explains why ADDITIONAL alarms associated with rod control occur during the rod recovery?
a.
Errors exist in P/A converter and plant computer points.
.
b. Alarms occur because ALL CB C lift coils have been disconnected.
t c.
NIS POWER RANGE HIGH NEUTRON FLUX RATE alarm (M-6A, B1) illuminates if
K-10 is withdrawn in a high flux zone.
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d.
ROD POSITION D.C. AUX POWER ON alarm (M4B D6) illuminates when the first lift coil in affected bank is disconnected.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 3.
Given the following plant conditions:
Unit is at 90% power with Control Bank D at 220 steps
.
RCS boron concentration is 150 ppm
.
Rod H8 in CB D has just dropped to the bottom of the core
.
RCS Tavg is 570 F and stable
.
Which ONE of the following describes the operator action required to stabilize the plant and maintain Tavg equal to Tref?
a.
Reduce turbine load, b.
Withdraw control rods
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c.
Borate RCS.
d.
Dilute RCS.
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SEQUOYAH NUCLEAR PLANT
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NRC LICENSE EXAMINATION 10/05/98
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4.
Given the following plant conditions:
t Unit 2 at 100% power l
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The Reactor Engineer informs you that Unit 2 control bank D (CB D) rod M-4 has
.
been determined to be misaligned by 23 steps
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AOP-C.01, Rod Control System Malfunctions, has been entered j
e Which ONE of the following explains why any attempt to realign rod M-4 should be coordinated
!
with the Reactor Engineer?
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a.
Prevent possible fuel failure and minimize xenon oscillations.
b.
Prevent possible localized power peaking and fuel failure when establishing T-avg at
,
T-ref.
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I c.
Because the operators failed to identify M-4 as a stuck rod, the Reactor Engineer must determine if the nuclear instrumentation is functioning properly, d.
Because the operators failed to identify M-4 as a stuck rod, the Reactor Engineer must determine if OP Delta T and OT Delta T calculator is functioning properly.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 5.
Given the following plant conditions:
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- Unit 2 is at 80% power with all systems in Automatic
.
Control Bank D is at 160 steps on the step counters
-.
Boron is added to the RCS to step Control Bank D out to 190 steps
.
Control Bank D has just started stepping out in response to the boron addition
.
The OATC reports the following individual Control Bank D rod positions:
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D-4160 steps -
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D-12162 steps
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All other rods in Control Bank D are at 176 steps
.
The Reactor Engineering Group has just confirmed control rod D-4 and D-12
.
positions.
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Which ONE of the following describes the action required by AOP-C.01, Rod Control System Malfunctions, for these conditions and the basis for the action?
a.
Dilute the RCS to insert Contrcl Bank D rods to 160 steps to maintain OPTR within limits.
>
b.
Dilute the RCS to insert Control Bank D rods to 172 steps to maintain all Control
,
Bank D rods within 12 steps.
j c.
Remove Unit 2 from service in a controlled manner to comply with Technical
,
Specification 3.1.3, Group Height, i
d.
Decrease turbine load to balance the effects of the boron addition to allow mig to initiate maintenance on control rods D-4 and D-12.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
- 6.
Given the following plant conditions:
Reactor trip and Si have actuated
.
The operating crew enters E-0, " Reactor Trip or SI"
.
The operators are unable to establish AFW flow, so they enter FR-H.1," Loss of
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.
Secondary Heat Sink" RCS pressure is now 750 psig and S/G pressure is approximately 890 psig
.
The operators are directed by FR-H.1 to enter E-1, " Loss of Reactor or Secondary l
.
Coolant" i
Which ONE of the following correctly summarizes plant conditions?
a.
Small Break LOCA in progress; secondary heat sink required.
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b.
Small Break LOCA in progress; secondary heat sink NOT tequired.
,
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c.
Large Break LOCA in progress; secondary heat sink required.
d.
Large Break LOCA in progress; secondary heat sink NOT required.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 7.
Given the following plant conditions:
Unit 2 is responding to a large break LOCA
.
Containment pressure is 12.0 psid
.
Pressurizer level is 0%
.
The operators have just transferred to ES-1.3, " Transfer to RHR Containment
.
Sump", due to low RWST level Sl is reset per ES-1.3
.
The STA reports the following conditions:
.
RVLIS lower range is 70% and Core Exit TCs reading 750 degrees F.
.
RCS PTS conditions to the LEFT of Limit A
_
.
RHR Pump 2B-B tripped on overcurrent
.
Which ONE of the following is the proper action to take?
a.
Immediately go to FR-l.2, " Low Pressurizer Level", Step 1.
b.
Immediately go to FR-Z.1, " Response to High Containment Pressure, Step 1.
c.
Complete the alignment of ECCS for Cold Leg Recirculation, then go to FR-P.1,
" Pressurized Thermal Shock", Step 1.
d.
Go to FR-C.2, " Degraded Core Cooling", Step 1 and concurrently complete the alignment of ECCS for Cold Leg Recirculatio _
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SEQUOYAH NUCLEAR PLANT
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NRC LICENSE EXAMINATION 10/05/98
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8.
Given the following plant conditions:
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Reactor trip and Safety injection occurred on Unit 1 l
=
EG-1,1, "Si TERMINATION" has been entered
!
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I Which ONE of the following combinations describes the'ES-1.1 Si termination method and
!
ECCS flow reduction methods performed in other EOPs? NOTE: This question is not referring l
to ONE or TWO handed operation, j
l Si TERMINATION ECCS FLOW REDUCTION j
a.. Stops pumps one at a time. Exits the Stops pumps one at a time. Exits EOP set to appropriate plant procedure.
The EOP set to appropriate plant
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procedures.
!
b.
Stops pumps two at a time. Exits the Stops pumps one at a time. Remains
EOP set to appropriate plant procedure, within the EOP set.
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c.
Stops pumps one at a time. Remains.
Stops pumps two at a time. Exits the within the EOP set.
EOP set to apprcpriate plant procedure, d.
Stops pumps two at a time. Remains Stops pumps two at a time. Exits the within the EOP set.
EOP set to appropriate plant procedure.
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SEQUOYAH NUCLEAR PLANT i
L NRC LICENSE EXAMINATION 1G/05/98 l
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9.
Given the following plant conditions:
l
. - Unit 1 at 28% power
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. - All control systems are in Auto
.. RCP W 1 trips on overcurrent
!
Which ONE of the following is the expected plant response to the RCP #1 trip?
I a.
S/G #1 level decreases and steam flow on S/G #1 decreases.
i b.
S/G #1 level decreases and steam flow on S/G #1 increases.
l c.
S/G #1 level increases and steam flow on S/G #2 decreases.
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d.
S/G #1 level increases and steam flow on S/G #2 increases.
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SEQUD MH NUCLEAR PLANT l
NRC LICENSE EXAMINATION 10/05/98
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10.
Given the following plant conditions:
i Unit 2 in MODE 3 for maintenance
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Panel 0-XA-55-27B-D Annunciator A-4, MISC EQUlPM SUPPLY HEADER FLOW
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LOW, starts alarming i
Panel 0-XA-55-278-D Annunciator A-6, LETDOWN HX OUTLET FLOW / TEMP
.
ABNORMAL, starts alarming
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Which ONE of the following events could cause both alarms to actuate?
!
a.
f b.
Letdown HX tube rupture.
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d.
Loss of charging flow.
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l SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
4 11.
Which ONE of the following conditions can cause "AMSAC INITIATED" annunciator to i
alarm? (Assurne unit initially operating at 100% power)
a.
2 of 4 steam generator narrow range levels at 10% for 45 seconds, one minute after a reactor trip.
b.
3 of 4 steam generator narrow range levels at 7% for 15 seconds, four minutes
'
after a reactor trip.
c.
3 of 4 steam generator narrow range levels at 5% for 25 seconds, two minutes after a reactor trip, d.
4 of 4 steam generator narrow range levels at 15% for 30 seconds, threr
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minutes after a reactor trip.
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NRC LICENSE EXAMINATION 10/05/98 f
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12.
Given the following plant conditions:
'
r Unit 2 has experienced an ATWS i
.
RCS pressure is 2360 psig
.
FR-S.1," Nuclear Power Generation /ATWS", step 4 directs operators " INITIATE
!
.
emergency boration of the RCS" i
Which ONE of the following is the basis for adding negative reactivity by initiating emergency l
boration instead of actuating Sl?
!
a.
Boration via Si actuation flow isolates main feedwater.
(
)
b.
SI flow rate (boration) is limited by high RCS pressure.
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c. Emergency boration is the most direct available method of adding negative reactivity
,
to the core.
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d.
Emergency boration limits additional RCS pressure increase that could cause
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pressurizer PORV to operate,
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SEQUOYAH NUCLEAR PLANT f
NRC LICENSE EXAMINATION 10/05/98 i
i 13. -
Given de following plant conditions:
,
Unit 1 was at 73% power
.
. A reactor trip / safety inbetion on low steam line pressure occurred 21 minutes ago
.
Averege Core Exit TC temperature is 400*F
..
.
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RCS pressure is 1350 psig e
All S/G pressures are DECREASING slowly
.
- 2 and #3 S/G levels are 5% NR and DECREASING slowly
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- 1 S/G levelis 6% NR, and INCREASING slowly
.
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- 4 S/G levelis FTEADY at 2% NR i
Total feedwater flowis 340 gpm
.
- PZR levelis 10% and INCREASING
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RCS T-cold temperature is 325'F and DECREASING slowly
.
Containment pressure is 5 psid and INCREASING slowly
.
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At this point, which ONE of the following Critical Safety Functions is the MOST degraded?
]
a.
Heat Sink i
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b.
Core Cooling I
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Containment
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d.
Pressurized Thermal Shock i
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NRC LICENSE EXAMINATION 10/05/98 14.
Given the following plant conditions:
A large main steam line break has occurred in containment
.
All MSIVs failed to close
.
. RCS has cooled down to M5'F within 15 minutes following initiation of the event
'
Narrow range level in all S/Gs is -15% and decreasing slowly j
.
l Which ONE of the following describes the Auxiliary Feedwater Flow requirements under these conditions (per ECA-2.1, Uncontrolled De pressurization of All Steam Generators) and the basis j
for the flow requirements?
a. Reduce flow to 25 gpm to each S/G to prevent thermal shocis to the steam generator j
shell.
-
b.
Reduce flow to 440 gpm total to minimize RCS cooldown, and ensure adequate heat sink.
,
i c.
Reduce flow to 25 gpm to each S/G to minimize the RCS cooldown and the total
)
amount of steam released to containment.
d.
Reduce flow to 440 gpm total to minimize the RCS cooldown and to prevent Pressurized Thermal Shock (PTS).
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r SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
,
15.
Given the following plant conditions:
.' Unit 2 is at 81% power Rods are in manual
.
The following Panel XA-55-5C annunciators have just Illuminated:
.
A-6 TS-68-2M/N RC LOOPS T AVG /AUCT T AVG DEVN HIGH-LOW
.
B-6 TS 68-2/A/B R'" ACTOR COOLANT LOOPS AT DEVN HIGH-LOW l
.
C-6 TS-68-2P/Q REAC COOL LOOPS T REF T AUCT HIGH-LOW j
.
Which ONE of the following plant transients would cause these three annunciators to begin alarming at the same time?
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a.
Steam line break.
],
b. - Dropped control rod.
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c.
RCS Dilution, d.
Steam generator tube leak.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 16.
Given the following plant conditions:
Unit load is at 28%
.
A condenser vacuum leak has occurred
.
Operators are decreasing turbine load in an attempt to maintain vacuum
.
Condenser pressure is 2.1 psia
.
Which ONE of the following should occur first if condenser pressure continues to increase?
a.
Manual turbine trip.
c.
Automatic turbine trip on low vacuum.
d.
Loss of steam dumps due to C-9 interlock.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 17.
Given the following plant conditions:
!
Unit 1 stable at 97% power
.
A plant trip and safety injection occurred due to a LOCA i
.
A loss of all AC power occurred while the crew was implementing E-0
.
,
The crew transitio ed from E-0 to ECA-0.0 following the loss of all AC power
.
Power has just been restored to shutdown board 1 A-A l
.
The crew transitions to ECA-0.2, " Recovery From Loss of All AC Power With Sl i
.
Required"
[
The following conditions exist upon entry into ECA-0.2:
)
.
i TDAFW pump tripped and will not reset
.
.
All S/G levels are 10% - 20% Narrow Range
e Core Exit T/Cs average 760*F
.
.
Containment pressure is 3.1 psid and increasing slowly
.
Which ONE of the following actions is required at the time of entry to ECA-0.2?
a. Transition to FR-C.1.
i b.
Transition to FR-H.1.
c.
Continue with ECA-0.2 and monitor Status Trees for information only, d.
Continue with ECA-0.2 and perform FR-H.1 and FR-Z.1 actions concurrently, as conditions allow.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 i
!
18.
Given the following plant conditions:
Unit 1 is at 100% power
!
.
Rod control is in automatic
.
.
'
Pressurizer level is selected to 1-LT-68-339 for auto control v
Pressurizer pressure is selected to 1-PT-68 323 for auto control l
.
' The unit has just experienced a loss of 120 VAC Vital Instrument Power Board 1 11.
.
Which ONE of the following describes the effect on the plant, assuming NO operator action is taken?
I a. NIS channel N-41 fails low.
j
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b. Tave will initially decrease.
!
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Automatic rod control is lost but manual rod control is available.
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Pressurizer spray valves close ar :te backup heaters energize.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 19.
Given the follow.ing plant conditions:
Reactor trip and safety injection have occurred on Unit 1
.
The ERCW supply header 1 A to the Auxiliary Building has ruptured
.
Which ONE of the following aescribes the operational effect frorn the ERCW supply header 1 A rupture?
a.
CCS 1 A pump bearing failure occurs in 10 minutes.
b.
SI 1 A pump bearing failure occurs in 10 minutes.
c.
Turbine-driven AFW pump bearing failure occurs in 15 minutes.
-
d.
Motor-driven AFW pump 1 A-A bearing failure occurs in 30 minute.-
SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 20.
Given the following plant conditions:
i The control room has been evacuated due to a fire
.
All controls have been transferred per AOP-C.04
.
MDAFW pumps 1 A-A and 1B-B are injecting into the steam generators
.
The TDAFW LCVs have been placed in PTL
.
Steam generator pressures and levels are decreasing.
.
'
Which ONE of the following describes how ">G water level will be controlled ?
'
a.
The TDAFW pump LCVs will control in auto when transferred in the auxiliary position.
b.
The MDAFW pump level control valves will automatically control steam generator levels at 33%.
!
c.
The MDAFW pump level control valves will have to be manually adjusted using the i
Manual Output Adjust in the L-381 cabinet.
d.
The discharge pressure for the MDAFW pumps will have to be manually adjusted by throttling the manual valves at the LCVs.
.
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-
.
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SEQUOYAH NUCLEAR PLANT
NRC LICENSE EXAMINATION 10/05/98
'
21.
Given the following plant conditions:
'
>
Both units were stable at 100% power when the Control Room was evacuated in
!
.
response to a fire.
.
Plant operations are being conducted in accordance with AOP-C.04, Control Room inaccessibility.
All Unit 1 RCPs are stopped.
.
e Which ONE of the following is the reason the pressurizer auxiliary spray handswitch in the Unit 1 I
Auxiliary Control Room must be shifted from P-AUTO to CLOSE if Unit 1 letdown flow is lost
under these conditions?
'
_ To preclude an uncontrollable RCS pressure decrease due to cold auxiliary spray".
j a.
water being used.
l
b.
To preclude loss of RCS pressure coiltrol due to filling the pressurizer to solid water l
conditions.
c.
To prevent an excessive differenths temperature across the pressunzer spray nozzle.
d.
To prevent thermal shock to the letdown regenerative heat exchanger.
.
I
.
SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 22.
FR-Z.1, "High Containment Pressure," directs that if ECA-1.1, " Loss of RHR Sump Recirculation," is in effect, then operate containment spray pumps using applicable ctops in ECA-1.1, Which ONE of the following describes the basis for operating the containment spray pumps per ECA-1.1 while in FR-Z.17 a.
To conserve the level in the RWST.
b.
To raise level in the containment sump to restore RHR pump operation.
c.
To insure Si rumps have sufficient NPSH from the containment sump.
d.
To prevent automatic swapover of the spray pumps to the containment sump.
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_ _ _ _ _ _
SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 23.
Given the following plant conditions:
FR-C.1," inadequate Core Cooling", has been entered due to a RED path on Core
.
Cooling Core exit temperatures (TCs) are 1250 F and increasing
.
NO Feedwater/ Aux Feedwater is available
.
At step 12, the CRO checks the S/G NR levels and reports all are <10%.
.
As the SRO you should: (Select ONE of the following)
a.
Go to FR-H.1," Loss of Secondary Heat Sink".
b.
Depressurize all intact S/Gs to atmospheric pressure to dump accumulators.
~
c.
Start RCPs one at a time, until core exit TCs are less than 1200 F.
d.
Open pressurizer PORVs and block valves
-
o SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 24.
Given the following plant conditions:
At 0100 on 7/2/98 reactor power was 65%
.
At 0100 on 7/2/98 DOSE EQUIVALENT l-131 was 75 Ci/gm
.
At 0100 on 7/3/98 reactor power is 75%
.
At 0100 on 7/3/98 DOSE EQUIVALENT l 131 is 85 Ci/gm
.
,
Which ONE of the following actions is required for the given plant conditions 7
>
~
a. Be in HOT STANDBY with Tavg less than 500'F not later than 0700 on 7/3/98 to ensure RCS saturation pressure is less than the lift setpoint of the S/G atmospheric
!
relief valves.
'
b.
Be in HOT STANDBY with Tavg less than 500 F not later than 0700 on 7/2/98 to ensure the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the site boundary does not exceed 10 CFR Part 100 limits.
,
c.
Be in HOT STANDBY with Tavg less than 500*F not later than 0700 on 7/4/98 to ensure the 2-hour thyroid dose at the site boundary remains within limits following a
SGTR.
d.
Perform sampling and analysis of the RCS once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the specific activity -
of the primary coolant is restored to within its limits.
I' SEQUOYAH NUCLEAR PLANT ~
WRC LICENSE EXAMINATION 10/05/98 j
i i
!
25.
Given the following plant conditions:
l l
Unit 1 was stable at 100% power
.
A spurious reactor trip has just occurred on Unit 1
.
E-0, Reactor Trip or Safety injection, has been entered
.
j Which ONE of the following reasons explains the basis for closing the MSIVs as an immediate Action?
!
a.
Prevent overcooling if turbine fails to trip.
!
!
b.
Prevent overcooling if steam line break indicated.
I
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t c.
Prevent a release to the environment if a steam generator tube rupture occurs.
l d.
Prevent secondary plant contamination if a steam generator tube rupture occurs.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
'
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26, Given the following Unit 1 plant conditions:
)
. ' A PZR PORV has failed open
!
Reactor trip and Si actuated
.-
PORV block valve was closed from its breaker j
.
Containment pressure is 3.0 psid
.
All steam generator narrow range levels 17 to 19%
j
.
. Which ONE of the following conditions would prohibit Si termination?
'
a.
AFW flow 420 gpm.
~
b.
RCS subcooling 47 F.
c.
RCS pressure stable at 1550 psig.
!
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d.
Pressurizer level 23%.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 27.
G!ven the following plant conditions:
Unit 1 is at 100% power
.
RCS pressure is 2235 psig
.
RCS Tavg is 578*F
.
Safety injection (Cold Leg) pressure isolation check valve 63-553 is leaking through
=
internally at the rate of 0.25 gpm All other systems operating normally
.
Which ONE of the following describes the type of leakage and the action required by Technical Specifications?
(Assurr'e NO other RCS leakage)
~
a.
Identified leakage that requires shutdown.
l b.
Controlled leakage NOT requiring shutdown.
c.
Isolation valve leakage NOT requiring shutdown.
d.
Pressure boundary leakage that requires shutdown.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 28.
Which ONE of the following statements explains how the operator can control small break LOCA break flow rate to prevent core damage?
a.
ECCS flow can be decreased to increase the AP between the RCS and containment.
,
b.
ECCS flow can be increased to decrease the AP between the RCS and containment.
i c.
Steam generator atmospherics (S/G PORVs) can be used to reduce the break flow rate, d.
Steam dumps can be isolated to prevent excessive cooldown and reduce the break flow rate.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 29.
Given the following plant conditions:
A Unit 1 reactor trip and safety injection has occurred
.
Operators have correctly transitioned to E-1, Loss of Reactor or Secondary Coolant
.
Core exit T/C are stable at 587'F
.
RCS hot leg temperatures are stable at 577'F
.
RCS cold leg temperatures are stable at 535'F
=
RCS pressure is 1600 psig and decreasing
.
At which ONE of the following RCS pressures will a void initia!!y form in the reactor vessel head?
a.
890 psig.
,
b.
1015 psig.
c.
1315 psig.
d.
1415 psig.
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SEQUOVAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 30.
Given the following plant conditions:
Reactor power is STABLE at 90%
.
Pressurizer level is DECREASING
.
VCT levelis DECREASING
.
The following annunciators are actuated:
e PRESSURIZER LEVEL HIGH-LOW
.
LTDN HX OUTLET TO DEMIN TEMP HIGH
.
REGENERATIVE HX LETDOWN LINE TEMP HIGH
.
Which ONE of the following events would most likely cause these indications?
,
a.
Letdown header rupture.
b.
Charging header rupture.
c.
Isolation of CVCS letdown.
d.
Pressurizer level control valve fails closed.
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SEQUOYAH NUCLEAR PLANT
)
.
NRC LICENSE EXAMINATION 10/05/98
,
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~
31.
Given the following plant conditions:
Unit i reactor trip and safety injection occurred due to a large-break LOCA
.
Hot leg recirculation was initiated per ES 1.4, Transfer To Hot Leg Recirculation
.
All RHR flow has just been lost
.
Which ONE of the following explains how RCS cooling flow is affected by the loss of RHR flow?
a.
SIP containment sump recirculation flow is lost, b.
Only SIP flow from the containment sump is still in service, I
c.
Only CCP flow from the containment sump is still in service.
-
d.
SIPS and CCPs continue to inject into the RCS.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 32.
Given the following plant conditions:
Unit 2 reactor power is 100%
.
PT-68-340, Pressurizer Pressure Channe;, failed high yesterday
.
PT-68-340 bistables were tripped in accordance with AOP-l.04, Pressurizer
.
instrument Malfunction PT-68-340 was bypassed 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago to perform surveillance testing of the other
.
pressurizer pressure channels PT-68-334, Pressurizer Pressure Channel, can NOT be reset after testing the low
-.
pressure trip setpoint (it failed low)
Which ONE of the following describes the actions required to be taken for this failure?
a.
Commence shutdown withir.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 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 /> and Hot Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
Take PT-68-340 out of bypass 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 />, c. Take PT-38-340 out of bypass within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and trip the bistables on PT-68-334 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 PT-68-334 or PT-68-340 to operation 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 />.
,
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 33.
During a unit 1 startup, Which ONE of the following conditions will prevent the RTBs from being closed?
a.
120 VAC VITAL INSTRUMENT POWER BD 1-1 is de-energized.
b.
120 VAC VITAL INSTRUMENT POWER BD 1-11 is de-energized.
c.
120 VAC VITAL INSTRUMENT POWER BD 1-111 is de-energized.
r d.
120 VAC VITAL INSTRUMENT POWER BD 1-IV is de-energized.
'
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 34.
Given the following plant conditions:
Unit 1 is in Mode 2 durin a plant startup Reactor power at 5 x 10'g% on N-35 and N-36
.
.
SOURCE /INTERMED RANGE CH 1 TROUBLE annunciator illuminates (N-35 non-
.
operate bistable actuated)
Which ONE of the following actions is required by Tech Specs?
a.
Power must remain above 5 x 10% until N-35 is returned to OPERABLE.
b.
Restore N-35 to OPERABLE prior to increasing power above 5% RTP.
_
c.
Power raust be decreased by inserting rods until both source range channel energize.
d.
Restore N-35 to OPERABLE prior to increasing power above the P-6 setpoint.
,
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
I t
!
35.
Given the following plant conditions:
l
!
Unit 2 is operal ng at 29% power in accordance with 0-GO-6, Power Reduction From i
.
30% Reactor Power to Hot Standby Unit 2 will be going to Cold Shutdown for maintenance
.
intermediate Range N-36 has just failed high e
-
Which ONE of the following actions must be performed before reducing reactor power below
,
10%?
!
i a. Remove N-36 control power fuses, b.
Place N-36 Level Trip switch in BYPASS.
!
-
!
c.
Remove N-36 instrument power fuses.
d.'
Manually trip the reactor to prevent an automatic reactor trip.
!
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c SEQUOYAH NUCLEAR PLANT l
'
NRC LICENSE EXAMINATION 10/05/98
,
!
Unit 1 is evaluating a Steam Generator tube leak with the following plant parameters:
l l
Letdown flow is 75 gpm
!
.
One CCP is running i
.
. Presmrizer level is STABLE l
Seal injection flow to each RCP is 8 gpm j
..
Seal leakoff flow from each RCP is 3 gpm
.
Charging flow is 115 gpm i
.
Preexisting leakage is 3 gpm of identified leakage to the RCDT
.
,
Which ONE of the following is the approximate amouni of primary to secondary leakage?
a.
25 gpm.
b.
37 gpm.
c.
40 gpm.
i d.
57 gpm.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
.
37.
After a ruptured steam generator (S/G) has been isolated, E-3, " Steam Generator Tube Rupture," directs a rapid RCS cooldown to a selected core exit thermocouple temperature. Which ONE of the following is M
%. sis for the selected temperature?
a.
Ensures saturation pressure of primary coolant is less than lift pressure of S/G safeties.
b.
Minimizes inleal, age into the ruptured S/G until the subsequent RCS depressurizati'.,n can be initiated.
c.
Ensures 20 degrees of subcooling will be maintained during the subsequent RCS depressurization.
d.
Prevents back leakage from the ruptured S/G until the subsequent RCS depressurization can be initiated.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 38.
Given the following plant conditions:
Unit 1 was stable at 100% power
.
The following Unit 1 annunciators have just illuminated:
.
1-XA-55-3B (A-1) MAIN FEEDWATER PUMP TURBINE 1 A ABNORMAL
.
1-XA-55-3B (A-4) MFPT A OIL PRESSURE LOW
.
1-XA-55-3B (A-2) TRIPPED
.
1-XA-55-3C (A-1) MFP 1 A DISCH FLOW LOW
.
1-XA-55-3C (C-3 & E-3) STM GEN #1 LEVEL LOW
.
1-XA-55-3C (C-4 & E-4) STM GEN #2 LEVEL LOW
.
1-XA-55-3C (C-5 & E-5) STM GEN #3 LEVEL LOW
.
1-XA-55-30 (C-6 & E-6) STM GEN #4 LEVEL LOW
.
Unit 1 turbine runt,ack to 880 MWE is in progress
.
Which ONE of the following actions is required per AOP-S.01, Loss Of Normal Feedwater?
a. Ensure the TDAFW Pump is running.
b.
Trip the reactor and enter E-0 Reactor Trip or Safety injection.
c.
Reduce turbine load to 75% load using the valve position limiter, d.
Close the steam dumps to reduce steam flow from the S/Gs to limit further drop in S/G levels.
,
.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 39.
Which ONE of the following is the basis for stopping the RCPs upon entering FR-H.1,
" Loss of Secondary Heat Sink"?
a.
increases ambient heat loss of the RCS and improves RCS cooldown rate.
b.
Allows for a more controlled cooldown via natural circulation when Aux or Main Feedwater is restored.
c.
Reduces heat addition to the RCS and extends the mass inventory in the steam generators.
d.
Reduces RCS pressure allowing a higher flow rate from the CCPs thus increasing the RCS cooldown rate.
-
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 I
!
l 40.,
Given the following plant conditions:
Unit 1 is at 30% power j
.
The Train 8" Aux Control Air header depressurizes to O psig
!
.
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. Which ONE of the following is a component response that occurs as a result of this condition i
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_
a.
1-FCV-62-77, CVCS Letdown Header isolatinn Valve, fails closed.
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b.
1-LCV-3-172. TDAFW LCV to S/G #3, fails closed.
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c. -1-LCV-3-171 A, MDAFW LCV Bypass to S/G #4, fails closed.
d.
1-FCV-1-18 TDAFW Steam Supply isolation, fails closed.
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GEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 41.
Given the following plant conditions-Unit 1 core offload is in progress
.
One fuel assembly is in the RCCA change fixture
.
One fuel assembly is being removed from the upender on the SFP bridge hoist
.
A failure of the Reactor Cavity Seal occurs, cavity level is dropping about 1 inch per
.
minute Which ONE of the following actions is required for these conditions?
a.
Initiate RHR letdown and establish makeup from the RWST using the CCP.
b.
Initiate RHR letdown and establish makeup from the RWST using the RHR pump'
c.
Remove the fuel assembly from the RCCA change fixture and insert the assembly into any available core location.
d.
Remove the fuel assembly from the RCCA change fixture and insert the assembly into the cavity side upender.
,
_
,
,
SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAM: NATION 10/05/98 42.
Given the following plant conditions:
Unit 2 was steady-state at 95% power
.
A reactor trip just occurred due to the grid problems
.
- 1 and #3 RCPs tripped simultaneously with the reactor trip
.
- 2 and #4 RCPs are still running
.
Grid frequency and voltage increased significantly when the trip occurred
.
Which ONE of the following trip circuits caused the reactor trip?
a.
RCP Undervoltage, b.
RCS Loop Low Flow.
~
c.
RCP Underfrequency.
d.
Overtemperature Delta-T i
i
SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 43, Given the following plant conditions:
Unit 1 is in Hot Standby
.
Unit 2 is in Hot Shutdown
.
A loss of offsite power occurs on both units
.
AOP-P.01, Loss of Offsite Power, has been entered
.
Which ONE of the following explains why each diesel generator should be loaded to greater than or equal to 1.6 MW?
a.
Ensures that the diesel generators will be warmed up snd capable of supplying Blackout sequencer loads in the event of a subsequent Blackout.
b.
Ensures that a diesel generator will NOT trip on overspeed if a large load were to trip.
c.
Minimizes the possibility of combustibles accumulating in the diesel generator exhaust piping and turbochargers.
d.
Ensures that minimum required loads (shutdown boards) have been energized.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 44.
Which ONE of the following explains the effect on core operating limits if auctioneered-high AT (reactor power reference) is in error by 5*F high?
a.
Hot channel factor to prevent fuel melt (heat flux) will be more conservative.
b.
Hot channel factor to prevent DNB (enthalpy rise) will be less conservative, Rod insertion limits will be higher than required increasing actual shutdown margin.
c.
d.
Rod insertion limits will be lower than required decreasing actual shutdown margin.
_
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 45.
Given the following plant conditions:
Unit 1 is shutdown with a test in progress on RCP #1.
.
At 0700 RCP #1 was stopped.
.
At 0705 RCP #1 was started
.
At 0710 RCP #1 was stopped
.
At 0750 RCP #1 was started
.
At 0810 RCP #1 was stopped
.
," 0850 RCP # 1was started
.
At 0855 RCP #1 was stopped
.
- Which ONE of the following describes the earliest time when RCP # 1 can be started again?,
a.
0920.
bi 0925.
c.
0950, d.
0955.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
'
46.
Given the following plant conditions:
Unit 1 in Mode 1, following a reactor startup
.
Reactor is at 13% power
.
Tavg is 551*F on all fourloops
.
. - Turbine roll started
' A Loop 1 narrow range Tcold RTD fails high
.
Which ONE of the following describes the INITIAL plant response to this failure? (Assume NO operator action.)-
a.
Pressurizer level low Annuciator illuminates.
,
'b.
Charging flow remains the same.
c.
Charging flow decreases.
d.
Control rods insert.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 r
i 47, 4 Given the following conditions:
' LOCA on Unit 1
.
Containment pressure at 3.0 psig
.
Which ONE of the following describes the MINIMUM action (s) necessary to allow 1B-B i
Containment Spray Pump to be stopped and prevent it from automatically re-starting?
'
a.
Depress the train-B Phase B reset pushbutton on Control Panel M-6.
b. - Depress the train-B Containment Opray reset pushbutton on Control Panel M-6.
c.
Depress the train-B Phase B reset pushbutton and train-B Containment Spray reset pushbuttons on Control Panel M-6.
d.
Reset train-A and train-B Phase A Containment Isolation Signals, depress the train-B Containment Spray reset pushbutton on Control Panel M-6.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 48.
Which ONE of the following conditions will occur if actual rod position differs from bank demand position by 15 steps?
a.
Rod control system urgent failure alarm will illuminate.
b.
Rod control system non-urgent failure alarm will illuminate.
c.
Quadrant power and axial flux difference limits will be exceeded, d.
Control rod insertion limits may NOT ensure adequate shutdown margin.
-
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SEQUOYAH NUCLEAR PLANT
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NRC LICENSE EXAMINATION 10/05/98
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49.
Given the following plant conditions:
Reactor power is at 20% during a unit shutdown
.
Intermediate Range N-36 failed high
.
All operator actions have been performed
.
Which ONE of the following describes the effect of this failure and action during the remainder of l-the shutdown?
a.
The reactor will automatically trip when the Power Range channels decrease below the P-10 setpoint.
b.
Entry from Mode 1 to Mode 2 is prohibited with an inoperable Intermediate Rang 6 channel, so the unit must be manually tripped prior to Mode 2 entry.
c.
Both Source Range channels, N-31 and N-32, must be manually energized when the operable Intermediate Range channel (N-35) decreases below the P-6 setpoint.
d.
Source Ran0e channel N-32 must be manually energized when the operable intermediate Range channel (N-35) decreases below ;he P-6 setpoint; Source Range channel N-31 will automatically energize.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 50.
Given the following plant conditions:
Large Break LOCA is in progress
.
RCS pressure is 550 psig
.
Exosensor indicates 25'F superheat
.
No RCPs are operating
.
Which ONE of the following indications would the operator use along with RCS pressure to accurately substantiate core cooling?
a.
The hottest T/C in each quadrant and one T/C near the core center.
b.
Average value of all core exit thermocouples.
~
c.
Hottest Reactor Coolant wide range Thot value.
d.
Average value of five hottest core exit thermocouples.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 l-51, Given the following plant conditions:
Unit 1 is operating steady-state at 90%
.
l Containment Ventilation is aligned normal for this power level
.
- ERCW Supply Header 1 A is isolated to the Reactor Building due to a leak,
.
Which ONE of the following describes the plant response?
a.
The standby Lower Compartment Cooling Unit will automatically start when lower containment temperature reaches 110'F.
b.
Motor winding temperatures on Reactor Coolant Pumps 1 & 3 will increase above the maximum limit.
~
c.
Containment temperature and pressure willincrease resulting in a Phase A containment isolation.
d.
CRDM suction dampers will automatically realign from the reactc,i vessel shroud area to lower containment when lower containment temperature reaches 100*F.
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P 52.
Given the following plant conditions:
Mechanical maintenance is adjusting the ice condenser inlet doors spring tension
.
The maintenance technician adjusts the zero load spring tension such that the doors
.
l are fully shut and compress their gaskets.
Which ONE of the following describes how the ice condenser operation is affected by the maintenance action?
'
,
a.
Inlet doors will NOT open for large accidents.
i b. This is the normal setting and door position for non-accident conditions.
c.
Inlet flow maldis'ribution occurs for large accidents.
d.
Inlet flow maldistribution occurs for very small accidents,
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 i
53.
Given the following plant conditions:
6.9kV SD boards are aligned as follows:
.
1 A-A from the normal feed
.
1B-B from the attemate feed
.
2A A from the altemate feed
.
2B-B from the normal feed
.
Maintenance is in progress on 1B-B and 2A-A D/Gs
.
A fire is reported in 6.9kvUnit Board 1D
.
6,9kv Unit Board 1D has just been de-energized
.
_
Which ONE of the following is now UNAVAILABLE?
b.
Containment spray pump 1 A-A.
'
c.
Residual heat removal pump 1B-B.
j d.
Component Cooling Water Pump 1 A-A.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 54.
Which ONE of the following describes a required condition from E-1, " Loss of Reactor or
Secondary Coolant," for placing one train of RHR in service for containment spray?
'
a.
At least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> has elapsed since the beginning of the accident and both CCPs are
running.
!
b.
Containment pressure greater than 9.5 psid and at least one CCP pump running c.
RHR discharge aligned to CS pump suction and at least one SI pump running.
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i d.
At least 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> has elapsed since the beginning of the accident and at least one CCP pump 's running.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 55.
Which ONE of the following statements explains why a delay (lag circuit) is applied to
the level signal in the steam generator level controller?
a. Allows SF/FF signal to dominate throughout the level transients.
b.
Allows steam flow / feed flow mismatch signal to dominate at the start of transients.
c.
Allows level signal proportional action to dominate if actual level continues to deviate from the setpoint.
d.
Allows steam flow / feed flow mismatch signal integration to dominate if actual level continues to deviate from the setpoint.
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i SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
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. Given the following plant conditions:
Unit 2 was operating steady-state at 50% power
.
improper testing generated a Safety injection (SI) signal e
Tavg is at 551*F
.
Which ONE of the following describes conditions / actions required to gain control of the Feedwater Isolation Valves?
a. Depress the Train A/B Feedwater isolation Reset pushbuttons.
b.
Reset the Si signal, cycle the Train A/B Reactor Trip Breakers, and depress the F/W isolation reset buttons.
-
c.
Reset the Si signal and depress the Train A/B Feedwater Isolation Reset pushbuttons.
d.
Cycle the Train A/B Reactor Trip Breakers and depress the Train A/B Feedwater Isolation Reset pushbuttons.
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l 57.
Given the following plant conditions:
Train "A" Auxiliary Control Air header has de-pressurized due a significant leak j
.
The reactor has tripped
.
Which ONE of the following describes the operator response required to control AFW flow?
a.
Control the MDAFW LCVs in the MANUAL BYPASS position and close the TDAFW l
LCVs to prevent S/G overfill.
l.
b.
Lower the TDAFW pump to minimum speed and control the MDAFW LCVs in the l
MANUAL position to prevent S/G overfill.
c.
Pull-to-lock 1 A-A MDAFW pump, control the 1B-B MDAFW LCVs in the MANUAL or l-MANUAL BYPASS position, and lower the TDAFW pump to minimum speed to prevent S/G overfill, d.
Pull-to-lock 1 A-A MDAFW pump, control the 1B-B MDAFW LCVs in the MANUAL
,
'
position, and trip the TDAFW pump using the electronic overspeed to prevent S/G overfill.
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SEQUOYAH NUCLEAR PLANT
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NRC LICENSE EXAMINATION 10/05/98
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58.
Given the following plant conditions:
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i Unit 1 is heating up after a refueling outage i
. RCS temperature is 365'F
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.
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Steam Generator pressures are 150 psig
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e Steam supply valves (FCV 1-15 and FCV 1-16) from S/G 1 and S/G 4 have been
.
shut to isolate a crack on the body of FCV 1-18, steam supply valve to the TDAFW j
pump l
Repairs to FCV 1-18 were unsuccessful i
I Which ONE of the following describes the proper action to be taken?
-
i a. Be in HOT SHUTDOWN 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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b.
Reopen FCV 1-15 and FCV 1-18 and continue the heatup.
c.
Apply Technical Specification 3.0.3 and be in COLD SHUTDOWN within the next 36 l
hours.
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d.
Return Terry Turbine to operable status 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 SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 59.
Given the following plant conditions:
Unit 1 is at 100% power
.
l 125V DC Vital Battery Charger I is out of service and disassembled for maintenance
.
(Estimated retum to service 36 - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />)
125V DC Vital Battery Charger 1-S (spare) has been aligned to Battery Board 1
.
A review of 125V DC Vital Battery surveillance records determines the following:
.
125V DC Vital Battery I pilot cell specific gravity is 1.150
.
125V DC Vital Battery 11 pilot cell specific grav!ty is 1.100
.
125V DC Vital Battery lll pilot cell specific gravity is 1.300
.
125V DC Vital Battery IV pilot cell specific gravity is 1.165
.
,
Which ONE of the following act ons is required for the listed conditions?
a.
Restore 125V DC' il Battery Charger 1 to operable status within seven days.
b.
Restore specific gras.,.., all 125V DC Vital Batteries to greater than 1.200 within 48
.
hours.
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c.
Be in at least HOT STANDBY within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
d.
Be in at least HOT STANDBY within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 60.
A release of the Monitor Tank is in progress through the Liquid Radwaste System.
Which ONE of the following conditions will result in an immediate termination of the Monitor Tank release?
a.
Loss of power to RCV-77-43.
b.
Cooling tower blowdown flow 15,500 gpm.
RM-90-122, Liquid Radwaste Release Monitor, flow is 3.0 gpm.
c.
d.
High radiation signal on 0-RM-90-212 Cooling Tower Blowdown Rad Monitor.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 61.
Given the following plant conditions:
Unit 1 and Unit 2 are at 100% power
.
"A" Waste Gas Compressor is in service
.
Pressure in the "In-Service" WGDT is 60 psig
.
Waste gas release is NOT in progress
.
Activity level on 1-RM-90-400, Unit 1 Shield Building Monitor, is increasing
.
Which ONE of the following would cause the increasing activity level?
a.
Radiation control valve 0-RCV-77-119 is open.
b.
The relief valve on the "in-service" WGDT is leaking.
~
c.
The ABGTS fan stopped running because 0-RCV-77-119 is open.
d.
0-FIC-77-119 controller on panel 0-L-2A is NOT adjusted to the zero setpoint.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 62.
Which ONE of the following " Common Radiation Monitor 0-XA-55-128* annunciators in alarm is actuated by a Radiation Monitor that only senses gamma type radiation?
a.
(A-5) 0-RA-90-132A SERVICE BLDG VENT MON HIGH RAD.
b.
(B-1) 0-RA 90-101 A AUX BLDG VENT MONITOR HI RAD.
c.
(B-3) 0-RA-90-102A FUEL POOL RAD MONITOR HIGH RAD d.
(C-7) 0-RA-90-125A MAIN CNTRL RM INTAKE MON HIGH RAD
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NRC LICENSE EXAMINATION 10/05/98 I
l 63.
Which ONE of the following corresponds to a RVLIS lower range level of 40% when all l
RCPs are out of service?
a.
Top of the core.
b.
3.5 feet below the top of the core.
c.
3.5 feet above the bottom of the core.
d.
Bottom of the core.
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l 64, Given the following plant conditions:
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'All ESF equipment is operating as required e
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l Swapover to Cold Leg Recirculaticn per ES-1.3, Transfer to RHR Containment
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.
l Sump, is in progress j
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CCP & SIP pump suction valves from RHR, FCV-63-6 & 7, have just been opened
.
Prior to opening FCV-63-8 & 11, RHR Discharge to CCP and SIP, the CRO l
.
announces the RWST levelis 6%
Containment pressure is @ 3.8 psid.
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.
Which ONE of the following actions is required for these conditions?
,
a.
Stop and Pull-to-Lock the Containment Spray pumps, then continue with the transfer
'
,
l b.
Stop and Pull-to-Lock the CCPs, SIPS, and Containment Spray pumps, then l
l continue with the transfer to RHR Containment Sump.
Stop and Pull-to-Lock the CCPs, SIPS, RHR pumps, and Containment Spray pumps, c.
then continue with the transfer to RHR Containment Sump.
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d.
Confirm sufficient RWST level for transfer then continue with transfer to RHR containment sump.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 65.
Given the following plant conditions:
RCS temperature is 250*F
.
RCS pressure is 415 psig
.
All systems are aligned normal for plant conditions
.
Loop 3 wide range Thot i.istrument has failed LOW
.
Which ONE of the following describes the annunciaters which willilluminate 7 a.
PORV/ SAFETY OPEN annunciator illuminates.
b.
COLD OVERPRESSURE APPROACHING LIMIT annunciator illuminates.
c.
RCS TEMP LOW, ARM COPS annunciator and COLD OVERPRESSURE
APPROACHING LIMIT annunciator will illuminate.
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d.
COLD OVERPRESSURE APPROACHING LIMIT Liinunciator and COLD OVERPRESSURE PROTECTION annunciator willilluminate.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 j
l 66.
Given the following plant conditions:
,
Reactor power is 75%
.
Pressurizer pressure is 2235 psig
.
All controls are in AUTOMATIC
.
The backup heaters just ENERGlZED
.
Which ONE of the following is the approximate pressurizer level?
a.
41%.
b. 46%.
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c.
51%.
'
d.
56%.
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NRC LICENSE EXAMINATION 10/05/98
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67.
Given the following plant conditions:
i Reactor power is 40%
.
All control systems are in AUTOMATIC i
.
Bypass breaker BYA is racked in and closed; MEG is working on RTA j
.
t Which ONE of the following will result in an automatic trip signal to the Reactor Protectior l
System?
,
a. Train A" General Waming".
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Bypass breaker BYB racked in.
j c. Train B RPS logic A switch in "ON" position.
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d. Train B RPS memory switch in "OFF" position.
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SEQUOYAH NUCLEAR PLANT
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NRC LICENSE EXAMINATION 10/05/98
'
68.
Given the following plant conditions:
Unit 2 is holding at 15% power for S/G chemistry
.
l S/G water level control is in automatic on S/G #2
,
.
i 2-LT-3-52, S/G #2, has drifted upscale to 50% mig has completed the procedure to e
remove the channel from service.
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The CRO reports that 2-LT-3-55, S/G #2, has just failed upscale
.
l Which ONE of the following describes the plant response to these conditions (Assume no operator action or MIG action is taken)?
l a.
Reactor will trip on S/G Lo Lo level.
b.
Feedwater flow to S/G #2 will increase resulting in a Turbine trip.
c.
S/G #2 will be maintained at approximately 41% by the S/G Level Control Systern.
d.
S/G #2 will be maintained at approximately 44% by the S/G Level Control System.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 69.
Which ONE of the following describes how iodine is removed from the containmsnt atmosphere during a large break loss of coolant accident?
a.
Containment sump pH is maintained at 7.0 - 7.5 with sodium tetraborate from the ice condenser.
b.
Containment sump pH is maintained at 8.0 - 8.5 with sodium tetraborate from the ice condenser, c.
Containment spray helps scrub lodine from the upper containment atmosphere, d.
Ice condensers remove elemental iodine from only the lower containment volume and containment spray removes lodine from the upper containment atmosphere.-
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NRC LICENSE EXAMINATION 10/05/98 l
70.
Which ONE of the following is the upper limit of hydrogen concentration inside containment, above which the EOPs require the operators to obtain direction from the Technical Support Center (TSC) before placing the hydrogen recombiners in service?
a.
2% in dry air, b.
4% in dry air.
c.
6% in dry air.
d.
8% in dry air.
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i SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
,
71.
Given the following plant conditions:
Unit 1 in Mode 3
.
. Unit 2 in Mode 6 with a containment purge in progress
.
Unit 2 wafer valve is open for refueling operations
.
An auxiliary building isolation (ABI) occurs
.
Which ONE of the following actions must be performed on Unit 2?
a. The blast doors must be shut.
b. The wafer valve must be closed.
c.
Fuel handling exhaust fans must be stopped.
d.
The containment purge must be stopped, i
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SEQUOYAH NUCLEAR PLANT i
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NRC LICENSE EXAMINATION 10/05/98 l
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72.
Given the following plant conditions-
.. Unit 1 is at 50% reactor power
. - RCS Tavg is 564*F.
S/G water level control is in AUTOMATIC for all S/Gs
)
.
S/G #1 controlling steam pressure transmitter 1-PT-12A fails high (>1000 psig) over
.
a 1 minute period Which ONE of the following describes the effect on S/G #1 operation? (Assume NO operator action).
a. Feedwater flow will increase resulting in Turbine /Rx trip at 81% level.
b.
Feedwater flow will decrease initially and level will stabilize at program level.
Feedwater flow will increase initially and level will stabilize at the program level, c.
d.
Feedwater flow will decrease resulting in a reactor trip.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 73.-
Technical Specification 3.4.9, Pressure / Temperature Limits, curves for limiting Reactor Coolant System cooldown rates. Which ONE of the following describes the technical specification basis for using the composite curves for limiting reactor vessel cooldown?
,
a. The thermal gradients produced during cooldown produce compressive stresses at the inside of the reactor vessel wall, b.
The thermal gradients produced during cooldown produce tensile stresses at the outside of the reactor vessel wall.
The cooldown procedure is based on measurement of reactor coolant temperature, c.
whereas the limiting pressure is actually dependent on the reactor vessel temperature at the tip of the assumer' flaw.
-
d.
The cooldown procedure is based on measurement of reactor coolant pressure, whereas the limiting temperature is actually dependent on the reactor vessel stress at the tip of the assumed flaw.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
Which ONE of the following loads is supplied by the 480V Unit Boards?
a.
Hotwell pumps.
b.
Heater drain pumps.
c.
Control rod drive M-G sets, d.
Component cooling water pumps, i
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 75.
Given the following plant conditions:
Unit 1 & 2 are operating steady-state at 100%
.
All systems are normally aligned
.
Voltage on 6.9 kV Shutdown Board 1B-B is 6400 volts.
.
Which ONE of the following describes the plant response to ti.ese conditions?
a. After 1.25 seconds, all diesel generators wiii auto start and load shedding will be initiated on 1B-B 6.9 kV Shutdown Board.
b.
After 9.5 seconds,1B-B Diesel Generator will auto start and load shedding will be initiated on 1B-B 6.9 kV Shutdown Board.
-
c.. After 300 seconds all diesel generators will auto start and load shedding will be initiated on 1B-B 6.9 kV Shutdown Board.
d. After 30 seconds all diesel generators will auto start and load shedding will be initiated on all 6.9 kV Shutdown Boards.
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SEQUOYAH NUCLEAR PLANT WRC LICENSE EXAMINATION 'iO/05/98 76.
Diesel Generator 1 A-A has been started by the manual emergency start switch on the M-
1 panel. Which ONE of the following conditions / actions will stop Diesel Generator 1 A-
'
A?
a.
Low lube oil pressure 20 psi.
b.
High Jacket water temperature 205*F.
c.
Actuation of the generator reverse power relay.
d.
Actuation of the generator differential overcurrent relay.
.
.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 77.
Which ONE of the following combinations states the types of radiation detectors used in '
{
PIG (particulate, iodine, and gas) monitors 7 PARTICULATE LODINE GAS a. Beta Scintillation Beta Scintillation Gamma Scintillation l
b.
Beta Scintillation -
Gamma Scintillation Beta Scintillation c.
Gamma Scintillation Gamma Scintillation Beta Scintillation d.
Gamma Scintillation Beta Scintillation Gamma Scintillation
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'i0/05/98
'
78.
.
Given the following plant conditions:
.
Unit 1 is in Mode 6 with fuel shuffle operations in progress
.
The Z-Z axis tape reading is in error such that a fuel assembly is about one half inch
.
.
above the full down position.
!
. Load Cellindicates zero (0)
Which ONE of the following describes the response when the operator releases the gripper?
l a. The fuel assembly will drop and the potential exist to f911 over.
b. ' The console overload and engaged lamps will be illuminated.
c.
The assembly will NOT unlat;h (release).
J.
The fuel assembly will release as designed because the fuel assembly is seated on the lower core plate.
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l SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 79.
Given the following diagram of a YOKOGAWA containment temperature controller:
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TIC-1000 r. o LOOPt ab e
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67.0 E 3 sv:
ss s0.0 53 l
MV:
60.0 >h $
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STC ON E3 Eg PF
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Which ONE of the following responses will occur if the 'PF" pushbutton is depressed?
a. The controller is reset.
b.
The controller switches fror< ihe " Manual" mode to the " Automatic" mode of operation.
c.
The controller switches from the " Automatic" mode to the " Manual" mode of l
operation.
d.
The controller pages forward to the " Tuning Panel Group" or * Engineering Panel G roup".
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SEQUOYAH NUCLEAR PLAN 1 NRC LICENSE EXAMINATION 10/05/98 i
_
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 l
80.~
Given the following plant conditions:
I Unit 1 is in Mode 5
.
l The RCS is being maintained in a water solid condition
.
RCS pressure is 50 psig j
.
1 A A RHR pump is in service for shutdown cooling
.
' Which ONE of the following events would result in an increase in RCS pressure?
a.
Tripout of the 1 A-A RHR pump on an overcurrent condition.
~ b.
1-FCV-74-32 - RHR HTX Bypass valve fails open.
,
c.
Loss of control air to PCV-62-81, Letdown Pressure Control Valve.
d.
Control power failure to FCV-62-77, Letdown Line Isolation Valve.
.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 l-
- 81.
Given the follcwing plant conditions:
Unit 1 isin HOT STANDBY
.
. 480V SD BD 1 A1-A FAILURE OR UNDERVOLTAGE annunciator illuminates Which ONE of the following equipment will now be UNAVAILABLE 7 a. ' Fire Pump 1 A-A.
b.
c.
Component Cooling Pump 1 A-A.
- d. Control Room Air Conditioning Compressor A-A.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 82.
Given the following plant conditions:
Unit 1 is at 100% power
.
All equipment is aligned normal
.
j
. - ERCW strainer B2B-B in the 28 main supply header is blocked-Which ONE of the following describes the impact of this failure?
-
a, ERCW pumps L-B and M-B must be aligned to supply the 28 main supply header.
b.'
ERCW pumps Q-A and R-A must be aligned to supply the 28 main supply header, Component Cooling Water Heat Exchanger OB1 must be aligned to be supplied tty I
c.
the 1 A main supply header, j
d.
Component Cooling Water Heat Exchanger DB2 must be aligned to be supplied by the 18 main supply header.
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i SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 l
83.
Given the following plant conditions:
Unit 1 and 2 are at 100% power
A leak develops on the Control Air System
Control air pressure 74 psig and DECREASING
Which ONE of the following identifies the system response that should occur by the time control air pressure reaches 80 psig?
a.
Auxiliary air isolates from control air, b.
Auxiliary air compressors start and load.
-
i c.
Auxiliary air to containment valves fail closed.
d.
Control and service air compressors start and fully load.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 84.
Given the following plant conditions:
Unit 1 RCS temperature and pressure is 375*F and 1500 psig
.
Unit 2 RCS temperature and pressure is 190 F and 300 psig
.
Which ONE of the following is the MINIMUM Shift Manning requirement for the plant under the conditions stated above per SSP-12.1," Conduct of Operations"?
3
b.
4
c.
3
d.
4
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 85.
Which ONE of the following activities will require the use of a full faceshield over safety glasses with sideshields?
a. Acetylene-burning, cutting, or welding.
b.
Replacing fuses on the 250V DC Battery Board.
c.
Racking in or out a breaker on a 480V Unit Board.
d.
Using the turning tool to roll the diesel generator prior to running it.
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SEQUOYAH NUCLEAR PLANT
!
NRC LICENSE EXAMINATION 10/05/98
-'
t i
!
86.
Given the following plant conditions:
i
!
Unit 1 is at 100% power
.
Unit 2 is at 50% power i
.
The Unit 1 CRO is directed to temporarily relieve the Unit 2 OATC l
.
i Which ONE of the following describes the turnover requirements for this relief?
j The Unit 1 CRO shall be briefed by the Unit 2 OATC on all ongoing activities and a.
unit status.
b.
The Unit 1 CRO must be briefed by the Unit 2 OATC and note the transfer of responsibilities in the narrative log,
~
c.
A temporary shift relief is not required for this situation, the Unit 1 CRO need only l
receive authorization from the US/SRO or SM.
d.
The Unit 1 CRO must perform the Pi for Unit 2 OATC and record in the narrative log
!
the time responsibility was assumed.
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SEQUOYAH NUCLEAR PLANT
, *
NRC LICENSE EXAMINATION 10/05/98 t
87.
Which ONE of the following describes how a locked THROTTLED VALVE is verified in
the correct position per SSP-12.6 " Equipment Status Verification and Checking Program"?
The verifier observes the initial valve operator's action (concurrent verification).
l a.
I b.
The verifier checks the mechanical position indicator or stem position to verify that the valve is in the correct position and that the locking device is installed.
With the locking device REMOVED, the valve is opened and tumed in the CLOSE c.
direction the appropriate number of tums by the independent verifier.
!
d.
With the locking device REMOVED, the valve is closed and tumed in the OPEN-l direction the appropriate number of turns by the independent verifier.
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}
NRC LICENSE EXAMINATION 10/05/98 88.
Given the following plant conditions:
Unit 2 operating in accordance with 0-GO-5, Normal Power Operation at 73% with a '
.
power increase to 100% in progress CLA #1 water volume is 7610 gallons
.
Chemistry reports Unit 2 RCS loop 1 accumulator boron concentraticn is 2390 ppm
.
Current time is 0100
.
'
Which ONE of the following acticas must be taken?
a.
Immediately stop the power increase.
b.
Contir.ue the power increase while restoring loop 1 accumulator boron concentration l
to 2400 to 2700 ppm boron within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c.
If loop 1 accumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, be in HOT STANDBY by 0700.
'
i d.
If loop 1 accumulator boron concentration is NOT restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, reduce
pressurizer pressure to 1000 psig or less by 1300.
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SEQUOYAH NUCLEAR PLANT l
NRC LICENSE EXAMINATION 10/05/98
,
,
89.
Given the following plant conditions:
1B-B Thermal Barrier Booster pump is tagged to replace the coupling
.
The coupling has been replaced on the 1B-B pump, but a machinist is still on the
.
hold order The machinist was involved in an auto accident and can't be reached to be removed
.
>
from the clearance 1 A-A Thermal Barrier Booster pump trips due to a motor lockup
.
The Shift Manager determines the 18-B pump needs to be placed back in service.
.
Which ONE of the following descrit ts the process that can be used to release the clearance on the 18-B pump?
-
L a. The US/SRO confirms the work is complete by checking the WR status and releases i
the clearance.
b.
The Ops Superintendent and SM representative ensure that the pump is ready for I
service and the SM releases the clearance.
J c.
THE MMG Foreman confirms the work is complete and releases the clearance.
I d.
The Ops. Supt, and MMG Foreman inspect the pump, confirm it is ready for service i
and the SM Foreman releases the clearance.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 90.
Which ONE of the following identifies the minimum voltage above which a grounding device is required to be installed when working on electrical conductors?
a. 250V.
b.
600V.
c.
1000V.
d.
6900V.
.
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a SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 l
91.
Given the following conditions:
)
l Unit 1 is in Mode 6 with core cffload in progress.
.
One offloaded fuel assembly has been placed in the RCCA change fixture.
.
One offloaded fuel assembly (A) is being moved by the spent fuel pool (SFP) bridge
)
.
toward its assigned storage cell within the SFP racks.
j One offloaded fuel assembly (B) is horizontal in the reactor side upender awaiting
.
' transfer to the SFP side.
{
The refueling machine has been positioned above the reactor cavity over the next '
{
.
fuel assembly to be offloaded.
]
Within the following chronological sequence of events, wh?n can the next fuel assembly be ~
'
offloaded and lifted from its seated position in the reactor cavity?
immediately, providing that the next fuel assembly to be offloaded is bring moved j
a.
directly to storage in the RCCA change fixture.
b.
When the offloaded fuel assembly (B) has been transferred to the SFP side upender in the horizontal position.
When the offloaded fuel assembly (B) has been raised to the v?rtical position in the c.
SFP side opender, and offloaded fuel assembly (A) has been placed in its assigned SFP storage cell.
d.
When the offloaded fuel assembly (B) has been removed fror., the SFP side upender and is in transit to its assigned SFP storage cell.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
.
92.
Given the following plant conditions:
Unit 1 is in Mode 6 with refueling operations in progress.
.
The fuel transfer cart is empty and on the reactor side
.
The blind flange and wafer valves are open
.
Which ONE of the following describes the minimum number of operators that must be staticned at the reactor side controls and/or the spent fuel pool sides?
One operator at the reactor side controls and NO operators on the spent fuel pool a.
j side.
'
b.
One operator at the reactor side controls and one operator on the spent fuel pool'
side, Two operators at the reactor side controls and NO operators on the spent fuel pool c.
side d. Two operators at the reactor side controls and one operator on the spect fue! 9,01 side.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 93.
Given the following plant conditions:
Unit 2 is operating at 100% power
.
The RWP Survey Map for the Letdown Heat Exchanger room on EL. 714 lists the
.
following dose rates for the valve gallery in that room:
1250 mr/hr (General Area)
.
175 mr/hr (General Area)
..
43 mr/hr (General Area)
.
Beta-Gamma Derived Air Concentration (DAC) 3.0 E-8 Cl/cc
.
Beta-Gamma loose e,urface contamination 500 dpm/100 cm
.
Which ONE of the following lists the proper radiological posting for the valve gallery?
~
a.'
CAUTION, AIRBORNE RADIOACTIVITY AREA.
DANGER, HIGH RADIATION AREA.
b.
CAUTION, CONTAMINATED AREA.
DANGER, HIGH RADIATION AREA.
c.
CAUTION, AIRBORNE RADIOACTIVITY AREA.
GRAVE DANGER, VERY HIGH RADIATION AREA.
d.
CAUTION, RADIOACTIVE IWATERIAL AREA.
GRAVE DANGER, VERY HIGH RADIATION AREA.
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l SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 94.
Given the following plant conditions:
A General Emergency has been declared on Unit 1 due to a LOCA in the El. 653
.
pipe chase An offsite release is in progress due to this leak
.
A worker isolating the leak suffered a heart attack e
An emergency responder has volunteered to go in remove the injured worker
.
.. The volunteer has a current year-to-date exposure of 3 Rem
.
Which ONE of the following describes the MAXIMUM dose the emergency responder could be allowed to receive for this activity?
a.
10 Rem.
-
b.
7 Rem.
c.
25 Rem.
d.
22 Rem.
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- SEQUOYAH NUCLEAR PLANT
!
NRC LICENSE EXAMINATl,0N 10/05/98 l
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' 95.
Given the following plant conditions:
The dose rate from a small valve is 4 R/hr at 4 inches away
)
You are working at a distance of 4 feet from the valve e
,
All th, Jose rate in the work area is due to the small valve
'l
What is your dose rate if you work 4 ft. From the valve? (Assume NO shielding in place.)
i a.
22 Mr/Hr.
,
l b.
27 Mr/Hr.
,
c.
250 Mr/Hr.
d.
333 Mr/Hr.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
,
96.
Given the following plant conditions:
The radiation level 5 meters away from a valve is 1,000 R/hr, all due to the valve
.
Iron shielding in tenth-value layer (TVL) and half-value layer (HVL) thicknesses are
.
available to install as temporary shielding Which ONE of the following sets of shielding should be installed to reduce the radiation level to 5 mr/hr?
Number Number of TVL of HVL a.
1
-
b.
3 C.
2 d.
1
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l SEQUOYAH NUCLEAR PLANT i
NRC LICENSE EXAMINATION 10/05/98 97.
EPld 4," User's Guide", describes conditions for Prudent Operator Actions. Which ONE of the following would be considered an approved Prudent Operator Action, per EPM-47 a. The OATC manually initiates Phase B Containment Isolation with containment pressure at 2.7 psig and increasing.
b. The OATC manually starts the Containment Spray pumps 1 minute after a blackout with containment pressure at 8 psig.
The CRO isolates AFW flow to a ruptured steam generator immediately after the c.
OATC ensures the reactor is shutdown.
i d.
The CRO closes the TDAFW LCV to S/G #3 with its level at 75% and increasing-when the crew transitions from E-0 to ES-0.1.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 98.
Which ONE of the following is the LOWEST emergency classification that requires activation of the TSC and OSC?
a. Alert.
b.
Unusual Event.
c.
General Emergency.
d.
Site Area Emergency.
_
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 'iO/05/98 99.
Given the following plant conditions:
A steam line break has occurred inside containment
.
Maximum containment pressure was 6.3 psid
.
The crew has entered E-1 Loss of Reactor or Secondary Coolant
.
Containment pressure has decreased to 2.0 psid
.
Which ONE of the following describes how intact steam generator levels should be maintained?
a.
Maintain intact S/G narrow range levels between 10% and 25%.
b.
Maintain intact S/G narrow range levels between 10% and 50%.
c.
Maintain intact S/G narrow range levels between 25% and 50%.
d.
Sh!ft Manager and Shift Technical Advisor evaluate plant conditions and specify a control band.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98
,
!
i 100.
Which ONE of the following defines the EOP verb VERIFY?
.
a.
Evaluate the status of a parameter to establish whether or not an action should be
!
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performed.
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Check the status of a process parameter repeatedly, at an unspecified interval.
!
c.
Observe that an expected condition 3xists and, if necessary, take action to make the l
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condition occur,
!
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Observe that an expected condition exists, but does not permit action to make the
!
condition occur.
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SEQUOYAH NUCLEAR PLANT NRC LICENSE EXAMINATION 10/05/98 ANSWER KEY 1.
D-26.
A 51.
B 76.
D
2.
A 27.
C 52.
D 77.
B
'
l 3.
A 28.
C 53.
C 78.
A 4.
A 29. gC 54.
B 79.
A
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5.
C 30.
B M
55.
B 80.
A 6.
,EI$j fB 31.
A 56.
B 81.
C 7. C [l'
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D.
57.
C 82.
C 8.
B 33.
B 58.
D 83.
D 9.
'A 34.
D 59.
C 84.
D 10. A 35.
B 60.
A 85.
C 11. C 36.
A 61.
B 86.
A 12. A 37.
C 62.
C 87.
A 13. A 38.
A 63.
C 88.
A 14. C 39.
C 64.
B 89.
D
'
15. A 40.
C 05.
B 90.
B Deleted
- 16. B 41.
C 66.
D
D-48*sgu g fo/4/g 17. C-42.
A 67.
C 92.
B M /*[>NI
- 18. B 43.
C 68.
A 93, A
19. B 44, A
69.
C 94.
C 20. B 45.
D 70.
C 95.
B t
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21. C 46.
A 71.
D.
96.
D 22. A 47.
B 72.
C 97.
D
- 23. D 48.
C 73.
C 98.
A 24. A 49.
C 74.
C 99.
C l
25. A 50.
A 75.
C 100. D
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SEQUOYAH NUCLEAR PLANT
NRC LICENSE EXAMINATION 10/05/98
'
Attachment of Reference Material.
1. Steam Tables 2. Equation Sheet 3. PTS Limits Curve 4. T/S 3.4.8 Specific Activity 5. T/S 3.7.1.2 Auxiliary Feedwater (AFW) System 6. T/S 3.5.1.1 Cold Leg injection Accumulators 7. T/S 3.8.2.3 D.C. Distribution - Operating
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gQUATIONS AND CONVERSIONS HANDOUT SWRT
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'4t EOUATIONS
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.
___________________________________________________________________
P = P,108""t*8 d'= 2hc,aT
.
. P = P e */'8
- o
.
Q = sah -
A = A,e '
Q' = ' UAAT
-
,
CRsfa =. S/ (1 - K,re)
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% c cire CRs (1 - K, ggt) = - CR (1 - K r:2)
"
_ AT = decire 1/M = CR /CRx
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K,tr.= 1/(l'
'p)
DRW " &cip,/Om i
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P=
(K.re - 1)/K,rz y ' = pg
.
SUR = 26.06/t 6 = pap E-P
= baPu I,
pump
'A.fr P E = IR f~
Y
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A * T + 1 + A,gg Eff. = Net Work Out/ Energy In t
u(P -P) + (0 * - D') + 9(Z2 -Z)
=
2
f. = - 1 x 10-4 ~ ' seconds
-
2 g,
-g, A.,, = 0.1 seconds-2
g, = T2.2 lbm-f t/lbf-sec (I f(D )2 (I )(D )2
=
t g
2, CONVERSIONS
_________________________________________________________________ _
3.7 x 10 dps 3.41 x 10 Stu/hr 1 Curie
=
V 1 Mw
=
2.21 lbm
'2.54 x 103 Btu /hr 1 kg
=
-1 hp
=
1 Btu-= ' 778'ft-lbf 1 gal..,=
8.35 lbm
7.48 gal
"C'
.=
(5/9) (*F-- 32)
1 f t.c.,
=
J (9/5) ('C) + 32
- F.
=
'
l
- _
-.. _ _ _
_
_ - _ _ _ _.
>
i x
i i
p g
y:
-,
a
.
Pressurized Thermal Shock SON
'
F-0.4 FR-0 Curve 3 Page 7 of 11 Rev.11 PTS Limits Cairve
.
3000 iiiisiiii 3000
._
_
__
_
_
.. _
_
_
2560 PSIG
_
.-
---
-
-
-
-
f
.
A
..
-
-.
_
...
..
..
-
..
..
-
..
.
_
..
_.
..
..
-.
-
-.
_
.
. -
-.
,
,
f
..
_
_
_
-
_
_
__
2230 F
/-
--
-
-
-
'
Y
-
_
_
_ _ _'
2050 psig
_
_
__
._
__
".
2000 2000
._
e
}
1500
_
___
U nacceptable
.
Acceptable
__
R egion Rel; ion 1500 mg i
g
_
_
_..
__p
__
__
_
,
.
___
_
._
__
_
__
_
___
y tn
_
-
--
-
-
-
--
'
1000 i
1000.,
.
--
280 F
Limit A m
g I
I 500
!
!
500
.w 8.
191 F
--
- er
-
250 0
.
___
F g3
_
_ S-
-r_
__
.
g s-Y
'
'
o
100 200 300 400 500-c>
.=:2-RCS Temperature 0 F o
3o y
.
SON SQS2-0110-R3.YO3
.
a w
w-
-
_.. _. _ _ _ _ _. _ _ _. - _ _ _ _. _ _ _. _ _. _ _ _ _ _ _ _. _. _ _ _ _ _. _.. _... _ _. _ _
.
-
.
REACTOR'C0OLANT SYSTEM
-
3/4.4. 8 SPECIFIC ACTIVITY i
LIMITING CONDITION FOR OPERATION
3.4.8 The specific activity of the primary coolant shall be limited to:
Less than or equal to 1.0 microcuries/ gram DOSE EQUIVALENT I-131, a.
and
,
b.
Less than or equal to 100 6 microcuries/ gram.
APPLICABILITY: MODES 1, 2, 3, 4 and 5 ACTION:
MODES 1, 2 and 3*
R121 With the specific activity of the primary coolant greater than a.
1.0 microcuries/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANN Y with T"'E less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
..
b.
With the specific activity of the primary coolant greater than 1004 R121 microcuries/ gram, be in at least NOT STANDBY with T"V9 less than.500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
MODES 1, 2', 3, 4 and 3 With the specific activity of the primary coolant greater than a.
1.0 microcuries/ gram DOSE EQUIVALENT I-131 or greater than 1004 microcuries/ gram, perform the sampling and analysis requiree nts of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
.
R121 e
'
.
"With I,yg greater than or equal 500*F.
'
-
SEQUOYAH - UNIT 1 3/4 4-19 Amendment No. 36, 117 June 19, 1989
-.
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---
.-
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...
. -
.
.-
-
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SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.
.
.
.
.
l
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~
.
SEQUOYAH - UNIT 1 3/4 4-20 Amendment No.16 June 19, 1989
..
-
t l
f I
k
.
.
1ABLE 4.4-4
'
$
,
c S
PRIMARY COOLANT. SPECIFIC ACTIVITY SAMPLE E
AND ANALYSIS PROGRAM
'
t TYPE OF MEASUREMENT e$
AND ANALYSIS SAMPLE AND ANALYSIS M00ES IN 1AIICH SAMPLE
'
a FREQUENCY AfG ANALYSIS REQUIRED
-
'
-*
1.
Gross Activity Determination
At least once per 72' hours 1, 2, 3, 4 2.
Isotopic Analysis for DOSE EQUIVA-1 per 14 days
LENT l-131 Concentration 3.
Radiochemical for E Determination 1 per 6 months *
-
4.
Isotopic Analysis for lodine a) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1,2',3,4,5
Including I-131, 1-133, and I-135
w
.
whenever the specific
activity exceeds 1.0
<
pCl/ gram DOSE
>
.
g, EQUIVALENT I-131 i
-
or 100/E pCl/ gram, and
"
l
-.
!
I b} One sample between
2, 8 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following i
.
l a T!!ERMAL POWER
'
'
change exceeding
-
15 percent of the
RATED THERMAL
'
POWER within a one hour period.
i'
- gntil the specific activity of the primary coolant ' system is restored within its il it
' Sample to be taken af ter a minimum of 2 EFPD and 20 days of POWER OPERATIDH have ela
.
m s.
!
reactor was last subtritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
'
e I
l
=
E
_ _ _
.. _. _.... _. - _. _ _ _... _. _ _ _ _.. _ _. _ _ _ _ _ _... _., _. _ _. _ _ _ _. _. _. _ _ _..
I
l
.
i,
.
!
f1&l(LLYSIFA:
'
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.
j ggljJARY FhudATER (AFW SYSTEM i
-
LIMITING CofEITION FOR OPERATION
R210
.
j-3.7.1.2 Three auxiliary feedwater trains shall be OPERABLE.*
h l
APPLICAlqLIU: MODES 1, 2, and 3. MODE 4 when steam generator is relied upon i,
for heat twon).
.
l E:
-
,
i
!
a.
With one AFW train inoperable in MODE 1, 2 or 3 restore the inoperable AFW train to OPERABLE status within 72' hours or be in HDT i
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 /> and in H0T SHUTD0W within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
With two AFW trains inoperable in MODE 1, 2 or 3, be in at least HDT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOW within the following 6.X hours.
c.
With three AFW trains inoperable in MODE 1, 2 or 3, immediately initiate corrective action to restore at least one AFW train to OPERABLE status.**
d.
With the required AFW train inoperable in MODE 4, immediately initiate action to restore the required AFW train to OPERABLE
'
status.
.
SURVEILLANCE RE0UIREMENTS 4.7.1.2.1 At least once per 31 days, verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.
_
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j August 2, 1995 SEQUOYAH UNIT 1 3/4 7-5 Amendment No. 12, 115, 206 a.
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.. _ _ _ _ _ _. _. _ _ _ _ _ _ _..
. _ _ _.. _ _. _.
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PLANT SYSTEMS
.-
SURVEILLANCE REQUIREMENTS (continued)
.
R2LO 4.7.1.2.2 At least once per 92 days, verify the developed head of each AFV pump at the flot test point is greater than or equal to the required developed head.*
4.7.1.2.3 Once every 18 annths, verify each AFW automatic valve that is not lockad, sealed, or otherwise secured in position, actuates to the correct
position on an actual or simulated actuation signal.**
j 4.7.1.2.4 Once every 18 months, verify each AFW pump starts automatically on an actual or simulated actuation signal.***
i
- Not reouired to be completed for the turbine driven AFW pump until 24 hourE after steam supply pressure is greater than or equal to 842 psig.
- Not applicable in MODE 4 when steam generators are relied upon for heat removal.
- Not required to be completed for the turbine driven AFW pump until 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after steam supply pressure is greater than or equal to 842 psig. Not applicable in NODE 4 when steam generator (s) are relied upon for heat
. removal.
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August 2, 1995
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SEQUOYAH UNIT 1 3/4 7-6 Amendiment No. 12, 77, 114, 206
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3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
.
3 /4. 5.1 ACCUMULATORS COLD LEG INJECTION ACCUMULATORS LIMITING CONDITION FOR OPERATION
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3.5.1.1 Each cold leg injection accumulator shall be OPERABLE with:
a.
The isolation valve open, b.
A contained borated water volume of between 7615 and 8094 gallons of RIM
.
borated water,
_
c.
Between 2400 and 2700 ppe of boron, R196 d.
A nitrogen cover-pressure of between 600 and 683 psig, and e.
Power removed from isolation valve when RCS pressure is above
,
2000 psig.
L APPLICABILITY: MODES 1, 2 and 3.*
ACI1QN:
a.
With one cold leg injection accumulator inoperable, except as a result of boron concentration not within limits, restore the inoperable accumulator to OPERABLE status within one hour or be in at
_
least 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 /> and reduce pressurizer pressure to 2000 psig or less within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
,
b.
With one cold leg injection accumulator inoperable due to the boron concentration not within limits, restore boron concentration to within limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 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 /> and reduce pressurizer pressure to 1000 psig or less within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
.
.
"
- Pressurizer pressure above 1000 psig.
SEQUOYAH - UNIT 1 3/4 5-1 Amendment No. 124, 140, 147, 192 De c. ember 27, 1994
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. -.
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.. _
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.
EMERGENCY CORE COOLING SYSTEMS (ECCS)
SURVEILLANCE REOUIREMENTS 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE:
i a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
R196 )
1.
Verifying the containe'd borated water volume and nitrogen cover-pressure in each cold leg injection accumulator, and 2.
Verifying that each cold leg injection accumulator isolation valve is fully open.
b.
At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of greater than or equal to 1% of tank volume, that is not the result of addition from the refueling water storage
.
tank,# by verifying the boron concentration of the cold leg injection accumulatar solution.
c.
At least once per 31 days when the RCS pressure is abova 2000 psig by verifying that power to the isolation valve operator is removed.
,
i
.
.
- 0nly required to be performed for affected accumulators that experienced volume increases.
.
December 27, 1994 SEQUOYAH - UNIT 1 3/4 5-2 Amendment No. 12, 124, 147, 192
,
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_ - _ _ ___. _ _._ - -. _
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ELECTRICAL POWER SYSTEMS D.C. DISTRIBUTION - OPERATING M MITING CONDITION FOR OPERATION 3.8.2.3' The following D.C. vital battery channels shall be energized and OPERABLE:
CHANNEL I Consisting of 125 -volt D.C. board No. I, 125 - volt D.C.
battery bank No. I* and a full capacity charger.
lR41 CHANNEL II Consisting of 125 - volt D.C. board No. II, 125 - volt D.C.
I R41 battery bank No. II*, and a full capacity charger.
-
CHANNEL III Consisting of 125 - volt D.C. board No. III, 125 - volt D.C.
l R41 battery bank No. III*, and a full capacity charger.
CHANNEL IV Consisting of 125 - volt D.C. board No. IV, 125 - volt D. C.
R41 battery bank No. IV*, and a full capacity charger.
APPLICABILITY:
MODES 1, 2, 3 and 4.
ACTION:
a.
With one 125-volt D.C. board inoperable, restore the inoperable
'
board to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least NOT STAND 3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With one 125-volt D.C. battery bank and/or its charger inoperable, restore the inoperable battery bank and/or charger to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least 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 /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
.
.
CD.C. Battery Bank.V may be substituted for any other Battery Bank as needed.
R41
I
.
.
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January 24, 1985 SEQUOYAH - UNIT 1 3/4 8-11 Amendment No. 37 Y
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_______... _ _. _ _. _. _ _ _ _. _ _ _. _.. _.. _ _. _. _ _ _ _ _ _. _ _ _..... _... - _ _ _ _ _.. _ _ _ _ _ _ _ _ _.
.
.
,
ELFCTRICAL POWER SYSTEMS SURVEILLANCE RE0UIREMENTS (Continued)
.
d.
At least once per 18 months by verifying that the battery capacity
is adequate to supply and maintain in OPERABLE status all of the actual j
or simulated emergercy loads for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when the battery is subjected to a battery service test.-
.
a.
At least once per 60 months by verifying that the battery capacity
is at least 825 of the manufacturer's rating when subjected to a R2]l7 performance discharge test. Once per 60 month interval, this
performance discharge test may be performed in lieu of the battery service test.
i j
f.
Annual performance discharge tests of battery capacity shall.be-given to any battery that shows signs of degradation or has reached 855 of the service life ex.>ected for the application. Degradation is indicated when the sattery capacity drops more than 105 of rated
-
capacity from its average on previous performance tests, or is below i
90% of the manufacturer's rating.
,
f
,
R217
1
- Ij
)
.
'
!
!
.
V October 4, 1995 SEQU0YAH - UNIT 1
'
3/4 6-13 Amendment No. 29, 213
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e TABLE 4.8.2
-
BATTERY SURVEILLANCE REQUIREMENTS
.
~
ll)
CATEGORY B(2)
,
CATEGORY A A11owable(3)
Parameter Limits for each Limits for each designated pilot connected cell value for each connected cell cell Elcctrolyte
> Minimum level
> Minimum level Above top of Level indication mark, indication mark, plates, and and 5 k" above and i %" above not overflowing
,
maximum level maximum level
'
indication mark indication mark 12.13 vol'ts(C)
> 2.07 volts
Ficat Voltage
> 2.13 volts Not more than
.
.020 below the
~
u6 average of all connected cells 1 1.195 1*
Average of all Average of all ha a)
t connected cells connectg) cells
> 1.205
> 1.195
)
(a). Corrected for electrolyte temperature and level.
i (b) Or battery charging current is less than 2 amps.
Corrected for average electrolyte temperature.
(c)
For any Category A parameter (s) outside the limit (s) shown, the battery
'
(1)
may be considered OPERABLE provided that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> all the category B j
measurements are taken and found to be within their allowable values, and provided all parameter (s) are restored to within limits within the next 6 days.
For any Category B parameter (s) outside the limit (s) shown, the battery (2)
may be considered OPERABLE provided that they are within their allowable values and provided the parameter (s) are restored to within limits within 7 days.
(3) Any Category B parameter not within'its allowable value indicates an inoperable battery.
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i March 25, 1982 MAR 251982 3/4 8-13a Amendment No. 12
.
SEQUOYAH - UNIT 1