IR 05000424/1993300
| ML20057B507 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 09/13/1993 |
| From: | Lawyer L, Mcwhorter R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20057B498 | List: |
| References | |
| 50-424-93-300, NUDOCS 9309220151 | |
| Download: ML20057B507 (106) | |
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/ p meo %,t UNITED STATES NUCLEAR REGULATORY COMMISSION g'
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REGION ll
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101 MAR!ETTA STREET, N.W., SUITE 2900
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Report No.:
50-424/93-300 Licensee: Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 Docket No.:
50-424 i
License No.:
NPF-68 Facility Name: Vogtle Electric Generating Plant Examination Conducted: August' 16-20, 1993 Chief Examiner: M$/
M4 9-/J-93
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Richard D. McWhorter, Jr.
'U Date Signed i
Examiners:
L. Sherfey, PNL e
l Approved By:
/b 2#Me d 9-c-93 f
Lawrence L. Lawyer, Chief'
Date Signed
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Operator Licensing Section 1
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Operations Branch
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Division of Reactor Safety SUMMARY
Scope:
NRC examiners conducted regular, announced operator licensi_ng initial examinations during the period August 16-20, 1993.
Examiners administered
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examinations under the' guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7.
Four Senior Reactor Operator (SR0) candidates' received written and operating examinations.
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CK 05000424
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Results:
Candidate Pass / Fail:
SR0 Percent
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Pass
100%
Fail
0%
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Examiners identified an inspector follow-up item regarding a failure to i
provide operators with tools to accomplish emergency procedure 19100-C,
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step 14 (paragraph 3.e.)
No violations or deviations were identified.
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REPORT DETAILS
1. Persons Contacted Licensee Employees
'J. Beasley, General Manager
- R. Brown, Supervisor, Operations Training
- C. Christiansen, Supervisor, Safety Assessment
- R. Doorman, Manager, Training and Emergency Preparedness
- W. Dunn, Unit Superintendent
- W. Gabbard, Specialist, Technical Support
- W. Kitchens, Assistant General Manager R. Magnuson, Instructor, Operations Training
- T. Mozingo, Oglethorpe Power Corporation C. Salter, Instructor, Operations Training J. Williams, Unit 1 Operations Superintendent Other licensee employees contacted included instructors, engineers, technicians, operators, and office personnel.
NRC Personnel
- R. Starkey, Resident Inspector
- P. Steiner, License Examiner
- Attended exit interview 2. Discussion a. Scope i
i NRC examiners conducted regular, announced operator licensing initial
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examinations during the period August 16-20, 1993.
Examiners adminis-tered examinations under the guidelines of the Examiner Standards (ES),
NUREG-1021, Revision 7.
Four Senior Reactor Operator (SRO) candidates received written and operating (simulator and walk-through)
examinations.
b. Candidate Performance Examiners evaluated candidate performance during three examination phases: written, simulator and walk-through. Candidates were tested in their knowledge of the plant and the ability to safely operate the plant in accordance with established plant procedures.
Examiners found that candidate overall knowledge level was good, and that candidates were j
strong in their ability to supervise the use of Emergency Operating Procedures (EOPs.) However, examiners also noted several areas of candidate deficiencies:
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Two candidates did not correctly classify an event using procedure 91001-C, " Emergency Classification and Implementing Instructions,"
The c' ndidates erroneously classified an event with a Revision 12.
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Report Details
faulted and ruptured steam generator as an " Alert." The correct classification was a " Site Area Emergency."
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Several candidates had difficulty correctly answering written examination questions concerning E0P bases or principles for mitigating core damage. This is based on the fact that several written examination questions in these areas were missed by more than one operator.
Although the small number of candidates in this examination limit general conclusions on training department effectiveness, examiners noted that these areas warrant additional attention for future licensed operator training classes.
c. Results Four of four SR0s examined passed for a 100% pass rate. When compared-to similar pressurized water reactor licensees, examiners judged that written and walk-through examination performance was average, and simulator performance was above average.
d. Simulator Facility Examiners observed simulation facility operation throughout the simulator and walk-through portions of the examination.
Examiners found that the simulation facility performed well in support of the examination, and no fidelity problems were identified.
During one examination scenario, the simulation facility generated unplanned nuclear instrument rate alarms for which candidates responded appropriately.
Licensee staff iden*ified the cause as intermittent
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problems in the simulator interface (input / output) system, and stated that this system has been scheduled for upgrading in the near future.
Examiners judged the licensee's planned action to be adequate to avoid similar problems on future examinations.
Additionally, examiners reviewed simulator issues discussed in Examination Report 50-424/92-302.
This report had identified that the simulation facility did not have prepared malfunctions for several important plausible plant failures.
Examiners found that the licensee had taken action to incorporate most these failures into prepared malfunctions.
Examiners concluded that licensee response to this issue was good.
e. Procedures Examiners observed candidate use of procedur'es throughout the simulator I
and walk-through portions of the examination.
Candidates were capable of using procedures to successfully accomplish tasks and mitigate
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emergency events. However, during the course of the examination, three problems were noted by examiners l
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Report Details
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Procedure 19100-C, "ECA-0.0 Loss of All AC Power," Revision 11, step 14, required operators to open the doors of specified electrical bus rooms.
Examiners tested the candidates' abilities to accomplish this task and found that the operators were required to search for
heavy objects (fire extinguishers, ladders, blank flanges, etc.) to prop open the spring-loaded doors. Although all candidates were eventually successful, the difficulty accomplishing this step should have been discovered and corrected during E0P validation. The step would have also been more difficult to accomplish during an actual loss of all AC since lighting conditions in the area would be poor.
Examiners expressed concern that this E0P local action step apparently had not been adequately validated to identify and correct i
this difficulty. The importance of this procedural step is underscored by its identification in station blackout analysis and individual plant examinations as a critical operator action required to avoid possible vital DC bus common mode failures.
Examiners identified this item for follow up as Inspector Follow-up Item (IFI)
50-424/93-300-01, " Failure to provide operators tools to accomplish Emergency Procedure 19100-C, step 14."
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Procedure 19100-C also contained an " Attachment A" which listed DC loads to be stripped in an emergency.
Examiners tested candidates'
abilities to accomplish this task and found that the format of the attachment led to some operator confusion. The attachment listed
. loads by breaker number and load affected, but failed to note that the description of the load affected did not match the nomenclature on the breaker label plate.
This led to unnecessary delays when candidates were surprised to find that the list in the attachment did not match the actual breaker nomenclature.
Examiners expressed concern that this was a second example of an E0P local action step which may not have been properly validated to identify potential difficulties in step accomplishment.
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Examiners tested candidates' abilities to perform procedure 19014-C,
"ES-1.4 Transfer to Hot Leg Recirculation," Revision 7, with one train of hot leg injection inoperable.
During examination preparation for testing operators' use of this procedure, licensee training staff pointed out to examiners that the logic for step 4 of this procedure was defective, in that it could lead operators to shut valve HV-8835 inappropriately when only one train of hot leg injection was operating.
Examiners were informed that this problem had been previously discovered, and a procedere revision had been initiated.
Examiners obtained a copy of the Procedure Revision Suggestion Form (number 92-0369), and verified that action was being taken to correct the step logic. However, it was noted that the facility had been slow to take action to correct the problem, which had been originally identified in March 199 Report Details
4. Exit Interview At the conclusion of the site visit, the examiners met with representatives of the plant staff listed in paragraph 1 to discuss the results of the
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examinations. The licensee did not identify as proprietary any material-provided to, or reviewed by the examiners. The examiners further discussed in detail the inspection finding listed below.
Dissenting comments were not received from the licensee.
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Item Number Description and Reference 50-424/93-300-01 Inspector follow-up item regarding the failure to provide operators tools to accomplish
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Emergency Procedure 19100-C, step 14
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(paragraph 3.e.)
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ENCLOSURE 2
'J SIMULATOR FACILITY REPORT
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Facility Licensee: Vogtle Electric Generating Plant
Facility Docket No.:
50-424 Operating Tests Administered Or.: August 18 & 19, 1993 This form is to be used only to report observations. These observations do
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not constitute audit or inspection findings and are not, without further verification and review,' indicative of noncompliance with 10 CFR 55.45(b).
l These observations do not affect NRC certification or approval of the
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simulation facility other than to provide information that way be used in
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future evaluations. No licensee action is required in response to these j
observations.
While conducting the simulator portion of the operating tests, the following
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items were observed:
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ITEM DESCRIPTION
None
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l 93,3&o U. S. NUCLEAR REGULATORY COMMISSION SITE-SPECIFIC WRITTEN EXAMINATION l
APPLICANT INFORMATION Name:
Region:
II Date:
August 16, 1993 Facility / Unit:
Vogtle 1 & 2 License Level:
SR0 Reactor Type:
PWR-WEC4
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INSTRUCTIONS l-l Use the answer sheets provided to document your answers.
Staple this cover sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires a final grade of at
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least 80 percent.
Examination papers will be picked up 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the examination starts.
All work done on this examination is my own.
I have neither given nor received aid.
Applicant's Signature
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RESULTS Examination Value 100 Points Applicant's Score Points Applicant's Grade Percent
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SEf00R REACTOR OPERATOR Page 2
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ANSWER SHEET
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Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
MULTIPLE CHOICE 023 a
b c
d 001 a
b c
d 024 a
b c
d i
002 a
b c
d 025 a
b c
d 003 a
b c
d 026 a
b c
d
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004 a
b c
d 027 a
b c
d 005 a
b c
d 028 a
b c
d 006 a
b c
d 029 a
b c
d 007 a
b c
d 030 a
b c
d 008 a
b c
d 031 a
b c
d i
009 a
b c
d 032 a
b c
d
_,
010 a
b c
d 033 a
b c
d 011 a
b c
d 034 a
b c
d 012 a
b c
d 035 a
b c
d 013 a
b c
d 036 a
b c
d 014 a
b c
d 037 a
b c
d 015 a
b c
d 038 a
b c
d 016 a
b c
d 039 a
b c
d 017 a
b c
d 040 a
b c
d 018 a
b c
d 041 a
b c
d
019 a
b c
d 042 a
b c
d 020 a
b c
d 043 a
b c
d 021 a
b c
d 044 a
b c
d 022 a
b c
d 045 a
b c
d
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ANSWER SHEET
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Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
046 a
b c
d 069 a
b c
d 047 a
b c
d 070 a
b c
d 048 a
b c
d 071 a
b c
d 049 a
b c
d 072 a
b c
d 050 a
b c
d 073 a
b c
d 051 a
b c
d 074 a
b c
d 052 a
b c
d 075 a
b c
d 053 a
b c
d 076 a
b c
d 054 a
b c
d 077 a
b c
d 055 a
c d
078 a
b c
d
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056 a
b c
d 079 a
b c
d 057 a
b c
d 080 a
b c
d 058 a
b c
d 081 a
b c
d 059 a
b c
d
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082 a
b c
d 060 a
b c
d 083 a
b c
d 061 a
b c
d 084 a
b c
d i
l l
062 a
b c
d 085 a
b c
d l
l 063 a
b c
d 086 a
b c
d 064 a
b c
d 087 a
b c
d 065 a
b c
d 088 a
b c
d q
066 a
b
d 089 a
b c
d 067 a
b c
d 090 a
b c
d 068 a
b c
d 091 a
b c
d
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ANSWER SHEET
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Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank.
092 a
b c
d i
093 a
b c
d 094 a
b c
d 095 a
b c
d 096 a
b c
d 097 a
b c
d 098 a
b c
d 099 a
b c
d 100 a
b c
d (********** END OF EXAMINATION **********)
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS
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During the administration of this examination the following rules apply:
1. Chrating on the examination means an automatic denial of your application and could result in more severe penalties.
2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examination.
3. Restroom trips are to be limited and only one applicant at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
4. Use black ink or dark pencil ONLY to facilitate legible reproductions.
5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.
6. Mark your answers on the answer sheet provided. USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
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7. Before you turn in your examination, consecutively number each answer sheet,
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including any additional pages inserted when writing your answers on the examination question page.
8. Use abbreviations only if they are commonly used in facility literature.
Avoid using symbols such as < or > signs to avoid a simple transposition
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error resulting in an incorrect answer. Write it out.
9. The point value for each question is indicated in parentheses after the question.
10. Show all calculations, methods, or assumptions used to obtain an answer to
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l any short answer questions, 11. Partial credit may be given except on multiple choice questions.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
12. Proportional grading will be applied. Any additional wrong information that is provided may count against you.
For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points, and you give five responses, each of your responses will be worth 0.20 points.
If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answers.
13. If the intent of a question is unclear, ask questions of the examiner onl )
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14. When turning in your examination, assemble the completed examination with
examination questions, examination aids and answer sheets.
In addition,
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turn in all scrap paper.
15. Ensure all information you wish to have evaluated as part of your answer is on your answer sheet.
Scrap paper will be disposed of immediately following the examination.
t 16. To pass the examinution, you must achieve a grade of 80% or greater.
- 17. There is a time limit of four (4) hours for completion of the examination.
18. When you are done and have turned 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 denied or revoked.
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QUESTION: 001 (1.00)
Which one of the following steady state plant conditions will result in the generation of an ATWT Mitigation System (AMSAC)
trip signal?
All Four Turbine Feedwater Impulse Pressures Flows (Each)
5%
38%
43%
b.
15%
42%
44%
c.
25%
44%
47%
d.
35%
49%
51%
QUESTION: 002 (1.00)
Given the following Unit 1 parameters:
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Reactor power is 100%
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Pressurizer pressure control is in AUTO.
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Proportional pressurizer heaters are at MINIMUM output.
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Normal pressurizer spray valves are FULL OPEN.
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Pressurizer PORVs are CLOSED.
Which one of the following is the LOWEST RCS pressure appropriate for the above conditions?
a.
2280 psig b.
2295 psig c.
2315 psig d.
2340 psig
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QUESTION: 003 (1.00)
Given the following Unit 1 parameters:
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A steam generator tube rupture has occurred.
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The RCS is at 520 degF.
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Containment pressure and temperature are normal.
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The ruptured steam generator has level indication in the control room of:
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99% narrow range
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85% wide range Which one of the following correctly explains the reason for the variance in SG level indications?
a.
Due to rapid pressure fluctuations experienced in the ruptured steam generator, the water has been drawn out of the narrow range reference leg.
b.
Temperature stratification in the steam generator is exposing the transmitters to different density water, causing inaccurate level indication, c.
The wide range reference leg condensing pot has been overfilled due to the high level condition resulting in an erroneously low indication.
d.
The wide range is cold calibrated and indicates low due to the density differences between the SG water and the reference le SEN,IOR REACTOR OPERATOR Page 9
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QUESTION: 004 (1.00)
Given the following:
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Unit 1 is in MODE I at 45 percent power.
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Condenser vacuum reads 20.5 inches Hg.
Which one of the following should have occurred as a direct result of condenser vacuum?
a.
Reactor trip INITIATING a turbine trip b.
Turbine trip INITIATING a reactor trip c.
Turbine run back d.
Turbine trip ONLY i
l QUESTION: 005 (1.00)
Procedure 19100-C, "ECA-0.0, Loss of All AC Power," cautions operators not to allow steam generator pressure to drop below 200 psig after cooldown has been initiated.
Which one of the following is the basis for this caution?
a.
Assure that RCS temperature does not drop to the point where the plant returns to a critical condition.
b.
Assure the RCS and Secondary conditions are maintained in a range where auxiliary feedwater can remove decay heat.
c.
Maintain adequate pressure on the RCP seals to avoid a LOCA due to seal cocking.
d.
Prevent the introduction of non-condensible gasses into the RC SENJ0R REACTOR OPERATOR Page 10
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QUESTION: 006 (1.00)
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Which one of the following will cause RV-18, the radioactive waste discharge isolation valve, to attomatically close?
a.
Fuel Handling Building sump p' imp autostarts, b.
Waste monitor tank #1 level is 4%.
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c.
Total discharge flow to the river is 4750 gpm.
d.
RE-018, waste liquid effluent monitor, is in ALERT.
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QUESTION: 007 (1.00)
Which one of the following radiat.an levels would be expected to be posted as a HOT SPOT?
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Contact General Area Radiation level Radiation level a.
100 mrem /hr 10 mrem /hr
b.
200 mrem /hr 25 mrem /hr
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c.
300 mrem /hr 50 mrem /hr d.
400 mrem /hr 100 mrem /hr
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QUESTION: 008 (1.00)
Given the following:
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A procedure is being performed inside a contaminated area.
The work copy at the job site is contaminated.
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A separate, clean copy is located outside of the contaminated area, and out of a direct line of site of the individual performing the task.
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The procedure is projected to take 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, with about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for each of the two sections of the procedure.
Which one of the following correctly describes the MINIMUM requirement which must be followed by the individual performing the task to ensure steps on the clean copy accurately reflect the status of the activity?
a.
Ensure the clean copy status is correct at the end of each step.
b.
Ensure the clean copy status is correct at the end of each shift.
I c.
Ensure the clean copy status is correct at each change of the job foreman.
d.
Ensure the clean copy status is correct on completion of each procedure section.
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QUESTION: 009 (1.00)
Which one of the following correctly states the basis for the immediate action step to " Verify Turbine Trip" in procedure 19211-C, "FR-S.1, Response to Nuclear Power Generation /ATWT?"
a.
Limits the mass lost from the RCS during the transient.
b.
Limits the mass lost from the steam generators during the transient.
c.
Limits the pressure excursion in the RCS during the transient.
d.
Limits the pressure differential across the steam generator U-tubes during the transient.
QUESTION: 010 (1.00)
Given the following:
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Unit 2 is in MODE 3.
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The RCS is at 2250 psig and gradually increasing.
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RCPs 2 and 3 are operating.
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Pressurizer level and pressure control are in AUTO.
Which one of the following is the LOWEST RCS pressure that will result in an automatic opening of PORV 456?
a.
2262 psig
b.
2295 psig c.
2338 psig
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d.
2360 psig
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QUESTION: 011 (1.00)
Given the following:
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Unit I has tripped from 100% power.
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SI has actuated and has been subsequently reset.
Which one of the following will re-enable automatic safety injection actuations?
a.
Manually blocking SI from the control board.
b.
Placing both SSPS trains in test.
c.
Shutting the reactor trip breakers.
d.
Automatic or manual reset of ESF sequencer e Q
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QUESTION: 012 (1.00)
Given the following:
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A feedwater isolation occurred due to a HI-HI steam generator (SG) level.
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RCS average temperature is 570 F.
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The reactor trip breakers are open.
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SG levels are all reduced below the HI-HI setpoint, and are steady.
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Safety Injection is not actuated.
Which one of the following actions will reset the feedwater isolation, AND ensure another feedwater isolation does NOT occur during a subsequent cooldown to MODE 4?
a.
Taking SSPS Train A to test.
b.
Closing and reopening both reactor trip breakers.
c.
Operating the MCB FWI Reset switches (train A and B)
momentarily to the RESET position.
d.
Closing and leaving closed both reactor trip breaker.
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QUESTION: 013 (1.00)
Given the following:
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Unit I has tripped from 100% power.
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Operators are performing 19000-C, "E-0, Reactor Trip or Safety Injection," step 3.
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It is determined that only one 4160 lE bus is energized.
Which one of the following correctly describes the required action?
a.
Continue on with 19000-C, with no further action to restore power required, b.
Continue on with 19000-C while attempting to restore power to the de-energized bus.
c.
Enter 19100-C, "ECA-0.0, loss of All AC Power," and perform required actions until both 4160 lE busses are energized, d.
Remain at 19000-C step 3 until all 4160 lE busses are re-energized using the RNO column for guidance.
QUESTION: 014 (1.00)
t!hich one of the following is the MINIMUM number of RCPs that must trip to directly cause a reactor trip at 38% reactor power?
(Assume all four pumps running initially.)
a.
b.
c.
d.
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QUESTION: 015 (1.00)
Given the following:
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Unit 1 is operating at full power when a transient occurs.
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NI power range channels indicate:
Channel 1 104%
Channel 2105%
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Channel 3107%
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Channel 4105%
Assuming the plant is at normal full load values, with no trip
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setpoint penalties or credits, which one of the following should have occurred as a result of these conditions?
j a.
NI overpower rod stop
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b.
NI overpower reactor trip c.
OP Delta T runback d.
OP Delta T reactor trip
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QUESTION: 016 (1.00)
Given the following:
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Reactor power is 99%.
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Pressurizer level is 60%.
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Letdown is in service at 120 gpm.
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A charging line check valve (located immediately before RCS loop penetration) disc hinge pin fails, allowing the disc to fall and totally block charging flow.
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Seal injection increases to 45 gpm.
Assuming N0 operator action is taken, which one of the following statements correctly describes the FINAL pressurizer response?
a.
Stable lower pressurizer level.
b.
Stable higher pressurizer level.
c.
Decreasing pressurizer level to the low pressure trip setpoint, d.
Increasing pressurizer level to the high level trip setpoint.
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QUESTION: 017 (1.00)
Given the following:
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Unit 2 is in MODE 2.
-
Reactor power is VJE-8 amps.
-
Critical data is being taken.
-
N-35 intermediate range control power fuses blow due to an internal fault.
Which one of the following is the appropriate action?
a.
Enter 19000-C, "E-0, Reactor Trip or Safety Injection,"
at Step 1.
b.
Hold power at 10E-08 amps until repairs are made.
c.
Insert rods to lower neutron flux level until both source range NIs energize.
d.
Place the N-35 Level Trip switch in the bypass position and continue with startup.
.
QUESTION: 018 (1.00)
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Which one of the following could result due to OVERCOMPENSATION of both Intermediate Range Nuclear Instrumentation Detectors?
a.
Source range nuclear instrument reactor trip during a reactor startup.
b.
Permissive P-6 will not automatically unblock on a-I reactor shutdown.
c.
Start up rate indication will be lower than expected when intermediate range indication is near 10E-8 amps, d.
Intermediate range detectors will eventually be damaged due to excessive current flo '
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QUESTION: 019 (1.00)
Given the following:
-
A reactor startup is in progress.
-
Source' Range channel N31 fails LOW.
-
The Intermediate Range Nuclear Instrument channel
'
indications are:
-
N35 - 2E-Il and increasing, SUR positive i
-
N36 - 1.5E-Il and increasing, SUR positive
'
Assuming that N31 repairs will take 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to complete, which one of the following actions must be performed in accordance with Technical Specifications?
l a.
Commence logging N32 readings at least every 15 minutes.
b.
Place the affected bistables in trip.
c.
Restore N31 to OPERABLE status prior to exceeding P-10.
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d.
Stop all positive reactivity changes.
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QUESTION: 020 (1.00)
Given the following:
-
The RCS has a stuck open Pressurizer safety valve.
-
The reactor tripped and safety injection initiated.
-
The RCS rapidly depressurized to saturation conditions.
-
Pressurizer level initially dropped and then began to rise rapidly.
Which one of the following correctly characterizes the relationship between pressurizer level and RCS inventory under these conditions?
a.
Level is an accurate indication of inventory, because voiding would occur first in the pressurizer due to the high temperature of the pressurizer walls.
b.
Level is an accurate indication of inventory, because hydraulic pressure would force any voids to the pressurizer steam space and out the safety.
c.
Level is NOT an accurate indication of inventory, because RCS voiding may result in a rapidly increasing pressurizer level.
d.
Level is NOT an accurate indication of inventory, because at higher temperatures the cold calibrated pressurizer level channels cause erroneously high indication..
SEN,10R REACTOR-0PERATOR Page 21
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1 QUESTION: 021 (1.00)
j I-Which one of the following is the reason that A0P-18038,
" Operations from Remote Shutdown Panel," step 1, " Trip the
,
reactor," has no RNO action?
a.
It is expected the operator will take whatever action is necessary to trip the reactor.
b.
A reactor trip is not a necessary condition for the performance of 18038.
I c.
19000-C, "E-0, Reactor Trip or Safety Injection,"
directs the appropriate RNO action upon entry.
d.
Immediate transition to 19211-C, "FR-S.1, Response to Nuclear Power Generation /ATWT," is required.
QUESTION: 022 (1.00)
Which one of the following individuals is responsible for ensuring a Temporary Change to Procedures does not change the
intent of the original procedure?
a.
Shift Superintendent b.
Originator c.
Cognizant Supervisor
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d.
Department Manager
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SENJ0R. REACTOR OPERATOR Page 22
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- QUESTION: 023 (1.00)
Which one of the following is the MAXIMUM time between last consumption of alcohol and overtime call out that requires an individual to notify their supervisor?
a.
b.
c.
d.
QUESTION: 024 (1.00)
Given the following:
A Reactor Operator must leave the control room for 45
-
minutes during his assigned shift.
Which one of the following is NOT required of the relieving operator PRIOR to taking the watch?
a.
Notify the Support Shift Supervisor, Balance of Plant Operator, and Plant Equipment Operators.
b.
Review the narrative log for the preceding 5 days or since the last shift worked, whichever is less.
c.
Complete the relief checklist and obtain complete information on current plant status, d.
Obtain permission from the Unit Shift Superviso n a
U SENJ0R REACTOR OPERATOR -
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QUESTION: 025 (1.00)
Which one of the following correctly describes the preferred method for independently verifying the position of a normally open manual valve that is required to be positioned slightly off the backseat?
a.
Move the valve slightly in the closed direction, then back to the original position.
b.
Move the va've slightly in the open direction, then back to the original position.
c.
Fully open the valve, then return to the procedurally established position.
d.
Visually observe the stem position or local indicator.
QUESTION: 026 (1.00)
Given the following:
-
Unit 1 is operating at 100%.
-
The normal feeder breaker tripped open for IAA02.
-
The diesel generator started, but the output breaker did NOT close.
Which one of the following actions should be performed FIRST?
a.
Reset the sequencer, then close the diesel generator
)
output breaker.
)
b.
Trip the diesel generator.
l c.
Manually close the alternate feed breaker.
j d.
Manually close the diesel generator output breaker.
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QUESTION: 027 (1.00)
Given the following:
-
An operating procedure is being performed to restore a system to service following system maintenance during an outage.
-
An error is discovered in the sequence of steps in the procedure which, if performed, would result in starting a pump without the required seal water.
Which one of the following actions should be taken?
a.
Continue with the procedure, performing the steps in the correct sequence, since the errors are known to be typographical.
b.
Continue with the procedure performing the steps in correct sequence, and request a procedure change to correct the order of the steps after the fact.
c.
Obtain the Unit Shift Supervisor's permission to perform the steps out of sequence.
d.
Terminate the performance of the procedure at the incorrect step, and request a procedure chang SEN,10RREACTOROPERATORO G
Page 25
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t QUESTION: 028 (l'. 00)
Given the following:
l
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Unit 1 is at 100% power.
'
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All steam generator level and pressure control board indications fail downscale.
-
All main feedwater control valve demand position indications start steadily increasing.
Which one of the following is the MINIMUM level in the organization that can authorize a Reactor Trip in this situation?
a.
Reactor Operator b.
Unit Shift Supervisor c.
Shift Superintendent d.
Assistant General Manager - Plant Operations
QUESTION: 029 (1.00)
Which one of the following individuals, by title, authorizes containment entry while in MODE 27 a.
Security Supervisor b.
Control Point Personnel c.
Health Physics Supervisor d.
Shift Superintendent
.
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SENJ0R REACTOR.0PERATOR Page 26
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QUESTION: 030 (1.00)
' Which one of the 'following conditions is acceptable for using solenoid valves and relays that are energized when in the desired position as a clearance point?
a.
When the solenoid valve or relay has two sources of power (i.e. normal AC and battery backup).
b.
When concurrence is obtained by the cognizant work group supervisor, Shift Superintendent, and Manager Operations, c.
When the solenoid ' valve or relay is mechanically blocked in the desired position.
d.
When the Unit Shift Supervisor concurs that the proposed isolation is adequate, and Maintenance Engineering concurs that adequate reliability exists.
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QUESTION: 031 (1.00)
l Given the following:
l
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A clearance exists with four subclearance holders.
-
The Electrical Plant Supervisor wants a tag changed on an MCC breaker door to allow opening the door for cubicle cleaning.
-
The MCC is de-energized with all feeder breakers HOLD tagged open.
-
The Electrical Plant Supervisor holds one subclearance.
-
All other subclearance holders are working on fluid systems under the clearance.
f'
Which one of the following is the level of approval required to change the tag on the MCC door?
a.
Unit Shift Supervisor ONLY b.
Electrical Plant Supervisor ONLY c.
Unit Shift Supervisor and Electrical Plant Supervisor d.
Unit Shift Supervisor and al' subclearance holders
.
QUESTION: 032 (1.00)
t!hich one of the following deficiencies requires a Work Request Tag?
a.
Tightening of packing on a manual valve.
b.
Tightening of non-safety related fitting leak.
Adjustment of non-safety related pump packing.
c.
d.
Adjustment of packing on a motor-operated valve.
l
SEN.IOR REACTOR OPERATOR Page 28
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QUESTION: 033 (1.00)
Which one of the following positions is responsible for assuring required surveillance tests have been performed prior to each MODE change?
a.
Department Managers b.
Surveillance Tracking Coordinators c.
Support Shift Supervisor d.
Unit Shift Supervisor QUESTION: 034 (1.00)
Given the following:
-
It has just been discovered that a 31 day surveillance on the Component Cooling Water (CCW) system has not been performed for 60 days.
-
A Unit Shift Supervisor review determined that operability of BOTH trains of CCW is impacted.
-
A review of the surveillance procedures indicate that it will take 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to perform the surveillance.
Which one of the following is the corrective action that allows the MAXIMUM time for plant operation in accordance with Technical Specifications?
a.
Ensure the surveillance is completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
Ensure the surveillance is completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate action to place the unit in a MODE that the specification does not apply.
d.
Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate action to place the unit in a MODE in which the specification does not appl,
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3/4 LIMITING CONDITICNS FOR OPERATION AND SURVEILLANCE REOUIREMENTS*
3/A.0 APPLICABILITY l
LIMITING CONDITION FOR OPERATION
-
3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.
If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met, except as provided
'
in the associated ACTION requirements, within I hour action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
a.
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 />, b.
At least 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 />, and c.
At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation, Exceptions to these requirements.are stated in the individual specifications.
,
This specification is not applicable in MO{ 5 or 6.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval.
Entry into an OPERATIONAL MODE or specified condi-tion may be made in accordance with ACTION requirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION e quirements.
Exceptions to these requirements are stated in the individual specifications.
.'
- Where specific instrument numbers are provided in parentheses they are for information only, and apply to each unit unless specifically noted (to assist in identifying associated instrument channels or loops) and are not intended to limit the requirements to the specific instruments associated with the number.
V0GTLE UNITS - 1 & 2 3/4 0-1 Amendment No. 59 (Unit 1)
Amendment No. 38 (Unit 2)
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APPLICABILITY LIMITING CONDITION FOR OPERATION (Continued)
3.0.5 Unless specifically noted, all the information provided in the Limiting Condition for Operation including the associated ACTION requirements shall apply to each unit individually.
In those cases where a specification makes reference to systems or components which are shared by both units, the affected systems or components will be clearly identified in parentheses or footnotes declaring the reference to be " common".
Whenever the Limiting Condition for Operation refers to systems or components which are common, the ACTION require-ments will apply to both units simultaneously.
(This will be indicated in the ACTION section.) Whenever certain portions of a specification refer to sys-tems, components, operating parameters, setpoints, etc., which are different for each unit, this will be identified in parentheses or footnotes or in the APPLICABILITY section as appropriate.
SURVEILLANCE RE0VIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2 shall constitute non-compliance with the OPERABILITY requirements for a Limiting Condition for Operation.
The time limits of the ACTION teguirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.
The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the com-i pletion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Surveillance Requirements do not have to be performed on inoperable equipment.
EntryintoanOPERATIONALMODEorotherspekifiedconditionshallnotbe 4.0.4 made unless the Surveillance Requirement (s) associated with the Limiting Condi-tion for Operation has been performed within the stated surveillance interval or as otherwise specified.
This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.
a
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V0GTLE UNITS - 1 & 2 3/4 0-2 Amendment No.59 (Unit 1)
Amendment No.38 (Unit 2).
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i APPLICABILITY SURVEILLANCE REOUIREMENTS (Continued)
i 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME
'i Code Class 1, 2, and 3 components shall be applicable as follows:
Inservice inspection of ASME Code Class 1, 2, and.3 components and
!
a.
inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the_ASME Boiler
'
and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50, Section 50.55a(g), except where specific written
,
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relief has been granted by the Consission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(1);
,
b.
Surveillance intervals specified in Section XI of the ASME Boiler I
and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and-Pressure Vessel Code and applicable Addenda shall be applicable as
,
follows in these Technical Specifications:
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ASME Boiler and Pressure Vessel Required fragttencies for
Code and applicable Addenda terminology for inservice performing inservice
'
inspection and testino activities inspection and testing i
activities
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Weekly
,
Monthly At least once per 7 days.
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t Quarterly or'every 3 months At least once per 31 days
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Semiannually or every 6 months At least once per 92 days Every 9 months At least once per 184 days
.
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l Yearly or annually At least once per 276 days l
At least once per 366 days
The provisions of Specification 4.0.2 are applicable to the above
.
c.
!
required frequencies for performing inservice inspection and testing j
activities; e
(
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t d.
Performance of the above inservice inspection and testing activities
.
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shall be in addition to other specified Surveillance Requirements; i
and
,
Nothing in the ASME Boiler and Press 0re Ve,ssel Code shall be construed e.
I to supersede the requirements of any Technical Specification.
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PLANT SYSTEMS
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3/4.7.3 COMPONENT COOLING WATER SYSTEP.
LIMITING CONDITION FOR OPERATION 3.7.3 Two independent component cooling water trains shall be OPERABLE with at least two pumps per train.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
With only one component cooling water train OPERABLE, restore the inoperable train 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 at least HOT STAN0BY 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 />.
SURVEILLANCE REQUIREMENTS 4.7.3 At least two component cooling water trains shall be demonstrated OPERABLE:
At least once per 31 days by verifying that each valve that is not a.
locked, sealed, or otherwise secured in position is in its correct position; and b.
At least once per 18 months during shutdown, by verifying that each Component Cooling Water System pump starts automatically on a Safety Injection test signal.
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V0GTLE UNITS - 1 & 2 3/4 7-11
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QUESTION: 035 (1.00)
Which one of the following is the MAXIMUM allowed whole body
,
exposure for planned lifesavino emergencies per 00920-C,
!
" Radiation Exposure Limits and Administrative Guidelines?"
a.
25 Rem b.
75 Rem c.
100 Rem
,
r d.
125 Rem i
QUESTION: 036 (1.00)
Given the following:
-
A 25 year old male raaiation worker and VEGP employee has received the following exposure:
-
300 mrem for the quarter
-
3900 mrem for the year
-
24 Rem lifetime Ubich one of the following is the MAXIMUM additional exposure the individual can receive today, without exceeding ADMINISTRATIVE limits? Assume N0 signed extensions.
a.
O mrem b.
600 mrem c.
700 mrem
,
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d.
1000 mrem
.
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QUESTION: 037 (1.00)
Which one of the following is the allowable time limit (minutes)
for notification of the agencies below following the initial declaration of an emergency?
County and NRC State a.
60 b.
45 c.
30 d.
15
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QUESTION: 038 (1.00)
Given the following:
-
Unit I tripped from 100% power.
-
RCS pressure rapidly dropped to below 1000 psig.
-
SI actuated, and started injection into the RCS.
-
Source range nuclear instrumentation energized, and initially showed a decreasing count rate.
-
Count rate then turned and increased, stabilizing at two decades higher than a normal shutdown two hours into the accident.
Which one of the following is the reason for the higher than normal source range reading?
a.
The SI water is cool, increasing the moderating effect and fission count rate over the amount normally expected, b.
As the core void fraction approaches 100%, the fission rate increases due to boron displacement from the core.
c.
Core boiling is occurring, resulting in a decrease in boron concentration in the lower part of the core, increasing fission rate.
d.
Downcomer voiding is occurring, allowing more neutrons to leak out of the cor O b
SENJ0R REACTOR OPERATOR w Page 32 w
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QUESTION: 039 (1.00)
Given the following:
-
Unit I reactor power is 1%.
-
S/G levels indicate:
-
S/G 1 is 68%
-
S/G 2 is 70%
-
S/G 3 is 76%
-
S/G 4 is 88%.
-
RCS Tavg drops to 555 degF.
Which one of the following automatic actions should directly result from the above situation?
Reactor Feedpump Trio Trio a.
Yes Yes b.
Yes No c.
No Yes d.
No No
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QUESTION: 040 (1.00)
Given the following:
-
The A Train Emergency Diesel Generator (EDG)
automatically started due to a Safety Injection ESFAS signal.
-
The EDG tripped after four minutes of operation on low lube oil pressure.
The Emergency Start signal is still present.
-
The condition that caused the low lube oil pressure is still present.
!!hich one of the following describes the status of the EDG following the operator depressing the EMERGENCY STOP RESET push button?
a.
The EDG will IMMEDIATELY restart with the low lube oil pressure trip disabled.
b.
The EDG will IMMEDIATELY restart and subsequently trip again on low lube oil pressure, c.
The EDG will NOT restart until the low lube oil trip condition is cleared and reset.
d.
The EDG will NOT restart until the start signal is cleared and another start signal initiated.
QUESTION: 041 (1.00)
Which one of the following is the MAXIMUM temperature at which a Core Exit Thermocouple output is designed to operate?
a.
708 degF b.
1200 degF c.
2300 degF d.
3200 degF
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QUESTION: 042 (1.00)
Which one of the following is the reason that all the RCP
,
breakers open on underfrequency?
a.
To protect the RCP from damage to the anti-rotation
!
device due to abnormal coastdown.
'
'i b.
To reduce the probability of a stress induced RCP
'
sheared shaft accident.
-
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c.
To avoid water hammer transients in the RCS induced by
rapid RCP speed change.
d.
To preserve the RCP flywheel kinetic energy and to
!
ensure a reactor trip signal occurs before core damage.
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QUESTION: 043 (1.00)
!
Which one of the following Reactor Vessel Level Indication System (RVLIS) ranges can ONLY be used during natural circulation?
!
a.
Upper Range AND Full Range
b.
Full Range AND Dynamic Range c.
Dynamic Range AND Upper Range
'
d.
Lower Range AND Full Range
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QUESTION: 044 (1.00)
,
Which one of the following parameters discriminate between a large secondary loss of coolant accident, and a large primary loss of coolant accident shortly after the accident begins?
a.
Pressurizer level
'
b'.
SG level
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c.
Pressurizer pressure d.
Adverse Containment Conditions
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8 QUESTION: 045 (1.00)
Given the following:
-
Unit 2 has tripped from 100% power.
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-
All systems are operable and in automatic.
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Actual Tavg is 13 degF greater than no-load Tavg.
Which one of the following describes the expected status of the steam dump valves?
Bank 1 Bank 2 Bank 3 Bank 4 a.
Full open Full open Full Open Modulating b.
Full open Full Open Modulating Closed c.
Full open Modulating Closed Closed j
d.
Modulating Closed Closed Closed QUESTION: 046 (1.00)
Which one of the following correctly describes which parameter (s)
are varied to control the RCS cooldown rate when using RHR to perform a normal cooldown of the RCS?
a.
Both CCW flowrate and RHR flowrate passing through the RHR heat exchanger.
b.
CCW flowrate through the RHR heat exchanger, while maintaining RHR flowrate constant.
c.
RHR flowrate passing through the _RHR heat exchanger, while maintaining total-RHR system flowrate constant. -
d.
Total RHR system flowrate varied between 3000 gpm and 5000 gpm.
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'EF Pago 36
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QUESTION: 047 (1.00)
Given the following:
-
Unit 1 is heating up in preparation for startup.
-
RCS temperature is 340 degF.
-
The Cold Overpressure Protection System is armed.
Which one of the following is the approximate MINIMUM pressure that will result in opening of the primary PORVs?
a.
550 psig b.
750 psig c.
950 psig d.
1150 psig QUESTION: 048 (1.00)
Given the following:
-
Unit 1 is at 100% power.
-
AFD has been determined to be outside the COLR limits.
Which one of the following actions will comply with Technical Specification requirements?
a.
Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for every 1% that AFD exceeds the limit, b.
Reduce THERMAL POWER to 46% of RATED THERMAL POWER within 25 minutes.
Restore the indicated AFD to within limits within 30 c.
minutes.
d.
Log AFD at least once per hour until it is within
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limits.
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QUESTION: 049 (1.00)
Given the following:
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The Volume Control Tank (VCT) level is at 40%.
-
Automatic make-up is in progress.
-
A leak develops in the reference leg associated with the automatic level controller (LT-ll2).
-
Assume no operator action.
Which one of the following describes the FIRST system response?
a.
LT-112 will begin modulating letdown flow via LCV-ll2A, b.
Charging pumps will automatically shift suction.o the RWST.
c.
Indicated VCT level will decrease.
d.
Automatic makeup to the VCT will stop.
QUESTION: 050 (1.00)
Given the following:
-
Unit 1 is in at 3% power.
-
Tavg indication is 540 degF.
Which one of the following actions is the MINIMUM required to comply with Technical Specification 3.1.1.4, " Minimum Temperature For Criticality?"
a.
Increase Tavg to 560 degF within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
b.
Increase Tavg to 560 degF within 15 minutes.
c.
Increase Tave to 550 degF within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
d.
Increase Tave to 550 degF within 15 minutes.
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QUESTION: 051 (1.00)
Given the following:
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A main steam line rupture has occurred on Unit 1.
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Containment pressure is 53 psig.
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NR levels in all steam generators are at 0% narrow range.
-
- Total auxiliary feedwater flow to intact steam generators is 450 gpm, which is the maximum available.
-
Core exit thermocouples indicate 320 degF and subcooling is 35 degF.
-
RVLIS indication is 50%.
-
Power range indication is 2%.
Which one of the following Critical Safety Functions has the HIGHEST order of priority under these conditions?
a.
Containment
.
b.
Subcriticality c.
Core Cooling d.
Heat Sink
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QUESTION: 052 (1.00)
Given the following:
-
The RCS is at 120 degF.
-
RCS pressure is 350 psig.
-
RHR is in shutdown cooling mode.
-
The pressurizer is solid.
Which one of the following methods is used to control RCS pressure under these plant conditions?
a.
Regulating the pressure in the VCT.
b.
Controlling letdown rate to CVCS.
c.
Modulating charging pump recirculating flow.
d.
Changing the number of CVCS orifices in service.
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QUESTION: 053 (1.00)
Given the following:
-
Unit 2 has tripped from 100% power.
-
A total loss of AFW and MFW has occurred.
-
Steam generator levels are gradually decreasing.
t!hich one of the following are the conditions that would FIRST require the immediate initiation of RCS bleed and feed cooling during adverse containment conditions?
Wide Range Level in %
48
63 b.
36
52
'
c.
24
48
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d.
18
36 QUESTION: 054 (1.00)
Which one of the following is the reason for promptly closing the seal leakoff isolation valve for an RCP with high leakoff flow from the number 1 seal?
a.
Prevention of damage to the bearing and seal package due to high temperature from excessive leakoff.
b.
Prevention of damage to the thermal barrier due to high fl ow.
c.
Minimize the amount of RCS water that is routed to containment sumps.
d.
Assure a minimum back pressure is maintained on the number 2 sea O,--
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QUESTION: 055 (1.00)
Which one of the following combinations of hydrogen and oxygen in the gaseous radwaste system REQUIRES immediate suspension of addition of waste gas, and reduction of the oxygen concentration in the waste gas holdup system, in accordance with Technical Specifications?
Hydrogen %
0xyaen %
a.
9 b.
7 c.
5 d.
3 QUESTION: 056 (1.00)
Given the following:
-
RE002 and RE003, the containment low range radiation monitors, are aligned for normal operation.
Which one of the following actions will occur if the containment low range radiation monitor, RE002 is in high alarm, and the containment low range radiation monitor, RE003 is NOT in high al arm?
a.
Radiation Monitoring System alarm ONLY b.
Containment Ventilation Isolation c.
Containment Phase A actuation d.
Containment evacuation horn sounds i
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QUESTION: 057 (1.00)
Given the following:
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Unit 1 is at 30% power.
-
Power escalation is in progress.
-
Control rods are in MANUAL.
-
After a control rod withdrawal of several steps, the rods continue outward when the IN-HOLD-0VT lever is returned to the neutral position.
-
The R0 determines a turbine runback is NOT in progress.
Which one of the following actions is required, per 18003-C, " Rod Control System Malfunction?"
a.
Place the Rod Bank Selector Switch in AUTO and check for continued rod motion.
b.
Trip the Unit and enter 19000-C, "E-0, Reactor Trip or Safety Injection."
c.
Check for failed instruments and select or block
,
appropriate instrument inputs.
,
d.
Hold the IN-HOLD-0VT lever to the IN position and check for continued rod motion.
.
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QUESTION: 058 (1.00)
-Which one of the following correctly describes the plant conditions existing AFTER the plant stabilizes.following a turbine runback with control rods in AUT0? (Assume all other plant control systems are operable and in automatic.)
RCS PZR PZR Tavo level Press a.
Decrease On program Normal b.
Decrease Below program Decrease c.
Increase Above Program Increase d.
Increase On Program Decrease QUESTION: 059 (1.00)
Given the following:
-
The fire system is in normal alignment.
-
Fire system pressure is 93 psig.
Which one of the following is the expected status of the fire pumps?
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Jockey Diesel Fire Diesel Fire Pump Pump 1 Pump 2
a.
Off Off Off s
b.
Running Off Off r
c.
Running Running Off d.
Running Running Running
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QUESTION: 060 (1.00)
Given the following:
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Unit 1 is at 100% power.
-
Megawatt output is DECREASING.
-
Window B04, "TURB CNDSR L0 VAC", alarm is illuminated.
-
Condenser vacuum is 23" Hg and is DECREASING (trending towards 0" Hg.)
Which one of the following is the FIRST required action?
a.
Start additional vacuum pumps.
b.
Runback the turbine rapidly, c.
Notify the Technical Staff to trend operating time at increased backpressure.
d.
Trip the reactor and turbine, and enter 19000,
"E-0, Reactor Trip or Safety Injection."
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QUESTION: 061 (1.00)
Given the following:
-
Unit l'has been shutdown for 3 days following 100 days at 100% power.
-
The RCS temperature is 120 degF.
-
The RCS is at midloop.
-
Hot leg dams are installed with no vent path established.
-
A total loss of RHR flow occurs.
-
No other action is taken.
Which one of the following is the approximate MINIMUM time required for the RCS to boil?
a.
1/6 hour b.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> c.
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> d.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />
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Page 46
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QUESTION: 062 (1.00)
Given the following:
i The plant is in MODE 6.
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,
-
Midloop operations are in progress.
i
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SG hot and cold leg manway covers are removed.
-
SG nozzle dams are installed on the hot legs.
,
-
SG nozzle dams are NOT installed on the cold legs.
j
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The only RCS vent path _is the SG cold leg manways.
-
RHR cooling flow is lost and not restored.
Which one of the following will occur as a result of this event?
a.
Steam formation in the hot leg will cause an erroneously high RCS temporary level indication.
b.
Steam formation in the reactor vessel head will depress reactor vessel water level and displace water out the cold leg manways, c.
Steam formation in the reactor vessel head will increase RCS pressure and blow out the hot leg nozzle dams.
d.
Steam formation in the hot leg will ultimately collapse, resulting in severe water hammer.
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Page 47
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i QUESTION: 063 (1.00)
Which one of the following is the significance of clearing water from the SG U-tubes (" blowing the inverted loop seal") during a cold leg small break LOCA?
a.
System mass loss will continue prior to the steam vent
path being established from the core to the break, b.
Core cooling is lost after the loop seal is blown.
c.
Pressure control is lost, challenging pressurized
'
thermal shock limits.
d.
The heat sink effect of the water in the intermediate leg is lost when the loop seal is lost, degrading core cooling.
QUESTION: 064 (1.00)
Which one of the following is a reason for the time delay between a turbine trip and a generator trip?
a.
RCP overspeed can occur during a major LOCA resulting in damage to the containment liner.
b.
A loss of RCS flow due to a failure of auto bus transfer would not be as serious because the reactor would have been shutdown for the duration of the time delay.
c.
The thermal stresses imposed on the RCP shafts are far greater if the RCPs are stopped immediately (due to high RCS delta T) after turbine trip.
j d.
The time delay allows circulating transformer currents induced by the trip transient to stabilize following the plant trip, reducing the probability of damage to the RCP motors.
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QUESTION: 065 (1.00)
Which one of the following parameters will best allow an operator to discriminate between a vapor space LOCA and a non-vapor space LOCA?
a.
Pressurizer level b.
Pressurizer pressure c.
RCS temperature d.
RCS subcooling QUESTION: 066 (1.00)
Given the following:
-
Following a LOCA the core exit thermocouples indicate 710 degF with an RCS pressure of 350 psig, Which one of the following conclusions should be drawn from the
'
above information?
a.
Core melting is occurring.
b.
Core uncovery is occurring.
i c.
Core exit thermocouples are no longer valid.
d.
Core exit thermocouples have failed.
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QUESTION: 067 (1.00)
When in MODE 1, Which one of the following is the LOWEST RCS pressure that is a one hour reportable event per Technical Specifications Safety Limits for RCS pressure?
a.
2463 psig b.
2721 psig c.
2885 psig d.
3001 psig QUESTION: 068 (1.00)
Which one of the following is the basis for reducing Tave to less than 500 degF following a shutdown required by a high Dose Equivalent I-131 level of 90 microcuries per gram?
a.
Slows coolant / fuel reaction rate, immediately reducing the source term of the activity.
b.
Reduces the potential release of activity following a steam generator tube rupture.
c.
Minimizes the temperature related degradation of the CVCS demineralizers while the RCS clean-up is in progress.
d.
Minimizes the iodine spiking phenomena which occurs due to the large change in THERMAL POWER level caused by the unit shutdow.SENJ0R REACTOR OPERATOR Page 50
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QUESTION: 069 (1.00)
Given the following:
-
Following a reactor trip, the crew transitioned from 19000-C, "E-0, Reactor Trip or Safety Injection" to 19001-C, "ES-0.1, Reactor Trip Response."
-
A loss of all AC occurs.
-
The Support Shift Supervisor reports the status of the CSFs as follows:
-
Subcriticality - GREEN
-
Core Cooling RED
-
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Heat Sink
- RED
-
Integrity
- GREEN
-
Containment
- GREEN
-
Inventory
- ORANGE Which one of the following procedures should next be used to mitigate the event?
a.
19221-C, "FR-C.1, Response to Inadequate Core Cooling."
b.
19262-C, "FR-I.2, Response to Low Pressurizer Level."
c.
19231-C, "FR-H.1, Loss of Secondary Heat Sink."
d.
19100-C, "ECA-0.0, Loss of All AC Power."
QUESTION: 070 (1.00)
Which one of the following parameters is used to determine the target temperature at which RCS cooldown is terminated when using 19030-C, "E-3, Steam Generator Tube Rupture?"
a.
An RCS temperature limit of 450 degF.
b.
The temperature at which RHR can be placed in service.
c.
Indicated ruptured S/G pressure, d.
Indicated subcooling in'the RC SEN,IOR REACTOR OPERATOR Page 51
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QUESTION: 071 (1.00)
Given the following:
A Reactor trip and SI have occurred.
-
Operators are using 19030-C,"E-3, Steam Generator Tube i
-
Rupture," to mitigate the event.
Which one of the following correctly describes the applicability of the RCP trip criteria on the foldout page while performing steps of 19030-C?
a.
RCPs should be tripped during performance of 19030 ANY TIME the foldout page RCP trip criteria are met.
b.
RCP trip criteria is not applicable in 19030 EXCEPT as an RNO if excessive RCS inventory loss is experienced.
c.
RCPs should be tripped during 19030 ONLY if the RCP trip criteria are met before beginning the cooldown and depressurization.
t d.
RCPs should be tripped during 19030 ONLY if the RCP trip criteria are met at Step 1 of 19030, when the l
operator is specifically required to check the
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criteria.
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QUESTION: 072 (1.00)
Given the following:
-
Technical Specification 3.5.4, " Refueling Water Storage Tank," limits the MINIMUM boron concentration in the Refueling Water Storage Tank to 2400 ppm.
Which one of the following is the reason for this limit?
a.
Ensures the reactor has a negative moderator temperature coefficient during accident conditions.
b.
Ensures that boron precipitation does not occur prior to cold leg recirculation during a LOCA.
c.
Ensures the reactor will remain subcritical during accident conditions, with all control rods inserted except the most reactive assembly.
d.
Ensures that long term cooling is not compromised by having a water / boron solution of such a high viscosity that SI pump overcurrent conditions will result.
QUESTION: 073 (1.00)
Which one of the following is the correct supply path and nominal leakoff flow rate and path for the number 3 reactor coolant pump seal?
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Supply Path Leakoff Flow and Path a.
- 2 seal leakoff 400 cc/hr to the RCDT b.
Standpipe 800 cc/hr to the RCDT c.
- 2 seal leakoff 800 cc/hr to the Containment Sump d.
Standpipe 400 cc/hr to the Containment Sump
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QUESTION: 074 (1.00)
Which one of the following correctly describes both the effect that a high temperature and pressure in containment will have on the steam generator level instruments, and the reason for the effect?
a.
The instruments will read LOWER due to high containment temperature causing an expansion of the water in the reference leg.
b.
The instruments will read LOWER due to high containment temperature causing static shift in the transmitters.
c.
The instruments will read HIGHER due to high containment temperature causing an expansion of the water in the reference leg.
d.
The instruments will read HIGHER due to high containment temperature causing static shift in the transmitter SEN,IOR REACTOR OPERATOR Page 54
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QUESTION: 075 (1.00)
Given the following:
-
In response to a large break LOCA a transition from 19000-C, "E-0, Reactor Trip or SI," to 19010-C, "E-1, Loss of Reactor or Secondary Coolant," was performed.
-
Due to a ORANGE path on the CORE COOLING status tree, a transition to 19222-C, "FR-C.2, Response to Degraded Core Cooling," was subsequently performed.
-
During performance of 19222, you observe that the CORE COOLING status tree has changed from a ORANGE to a YELLOW condition, while you identify a ORANGE path on the CONTAINMENT status tree.
Which one of the following describes the proper procedural transition, and the correct reason for the transition?
a.
Complete 19222, since it was entered due to an orange path, it must be completed unless a higher priority path occurs.
b.
Immediately transition to 19251-C, "FR-Z.1, Response to High CNMT Pressure," since an orange path is a higher priority than a yellow path, c.
Complete 19222, since once a functional recovery procedure is entered, it must be completed regardless of the status of other status trees.
d.
Perform the actions of 19222 and 19251 simultaneously, since functional recovery procedures on different trees
can be executed together.
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QUESTION: 076 (1.00)
Which one of the following will be a SHORT TERM result of starting an RCP during a natural circulation cooldown with a large void in the reactor vessel?
a.
Increase in subcooling margin.
b.
Rapid INCREASE in pressurizer level.
c.
Rapid DECREASE in pressurizer level.
d.
Decrease in the required heat removal from the RCS.
QUESTION: 077 (1.00)
Given the following:
-
A Steam Generator Tube Rupture is in progress in accordance with 19030-C, "E-3, Steam Generator Tube Rupture."
-
The RCS has been cooled down and depressurized, and ECCS flow has been terminated.
,
-
Normal charging and letdown have been established.
-
Ruptured steam generator level is INCREASING.
-
Pressurizer level is 17% and DECREASING.
Which one of the following actions should be taken?
a.
OPEN pressurizer sprays and INCREASE charging flow.
b.
OPEN pressurizer sprays and DECREASE charging flow.
c.
Turn ON pressurizer heaters and INCREASE charging flow.
d.
Turn ON pressurizer heaters and DECREASE charging flow.
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QUESTION: 078 (1.00)
Given the following:
-
An ATWT has occurred and 19211-C, "FR-S.1, Response to Nuclear Power Generation /ATWT," is in progress.
-
While implementing step 6, " Verify the following trips have occurred." an SI occurs and all rods insert.
Bhich one of the following actions should be performed?
a.
Immediately exit 19211-C, and perform 19000-C, "E-0, Reactor Trip or Safety injection."
b.
Immediately exit 19211-C, and perform 19010-C, "E-1, Loss of Reactor or Secondary Coolant "
c.
Perform 19211-C and 19010-C, "E-1, Loss of Reactor or Secondary Coolant," simultaneously, after the red path has cleared.
d.
Remain in 19211-C until completed or directed to transition to another procedure.
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a
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QUESTION: 079 (1.00)
Gi ren that the following occurred in sequence:
-
A small break LOCA occurred which resulted in a reactor trip and SI.
-
During performance of 19010-C, "E-1, Loss of Reactor or Secondary Coolant," the SI signal was reset.
-
A loss of offsite power occurred.
-
The diesel generators started and loaded as designed.
Assuming no operator actions, which one of the following would be the status of the loads on the 4160 lE busses?
a.
All equipment that was operating prior to the loss of power will be restarted.
b.
Only the equipment which would be started during a normal loss of power will restart automatically; ESF equipment must be manually energized.
c.
Manual loading of all loads is required.
d.
All equipment powered from the vital bus with the control board switch in automatic will be energize. _.
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QUESTION: 080 (1.00)
Which one of the following is the correct speed and total time delay for the containment coolers when started froro the sequencer during a loss of offsite power?
Containment Total Time Delay
!
Coolers Speed (seconds)
j a.
Eight fans Low 30.5 b.
Eight fans High 50.5
,
l c.
Four fans low 30.5 Four fans low 50.5 d.
Four fans High 30.5 l
Four fans High 50.5
i QUESTION: 081 (1.00)
Given the following:
-
-
Unit 2 is in MODE 4.
Which one of the following correctly describes conditions that will actuate the interlock shown for the RHR pump loop suction valves (HV-8701A/B and 8702A/B)?
'
Effect of interlock Condition on 1000 suct. valves
,
a.
RCS pressure of 770 psig Causes auto-closure j
b.
Associated train RWST suction Prevents opening valve is shut c.
RCS pressure of 380 psig Prevents opening d.
RCS temperature of 370 degF Causes auto-closure j
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QUESTION: 082 (1.00)
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Which one of the following is the LOWEST PRESSURE SET that will result in an SI actuation?
Containment Pressure Transmitters:
PT-934 PT-935 PT-936 a.
2.2 psic 2.8 psig 3.2 psig t
b.
2.8 psig 3.2 psig 3.8 psig
.
c.
3.2 psig 3.8 psig 4.2 psig d.
3.8 psig 4.2 psig 4.8 psig QUESTION: 083 (1.00)
Given the following:
-
Unit 1 is in MODE 3 at normal operating temperature and pressure.
-
AY2A is de-energized due to an inverter failure.
-
A loss of all offsite power occurs.
Which one of the following pumps will be started by the sequencer in this situation?
!
a.
Train A ACCW Pump
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b.
Train A RHR Pump i
c.
Train B ACCW Pump I
d.
Train B RHR Purap
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QUESTION: 084 (1.00)
Given the following:
-
The plant is operating at 100% power when a pressurizer safety valve inadvertently lifts.
-
Which one of the following should be the approximate tail pipe temperature of the safety valve?
a.
185 degrees b.
225 degrees c.
550 degrees d.
620 degrees QUESTION: 085 (1.00)
Bhich one of the following plant conditions will result in a greater chance of a restart accident following a steam line rupture?
Core Life Reactor Power a.
EOL FULL Power b.
BOL FULL Power c.
EOL ZERO Power d.
BOL ZERO Power
O b
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QUESTION: 086 (1.00)
Which one of the following parameters is NOT used to verify natural circulation in accordance with 19001-C, "ES-0.1, Reactor Trip Response?"
a.
RCS subcooling b.
Core exit thermocouple temperature c.
RCS pressure d.
SG pressure QUESTION: 087 (1.00)
Given the following:
-
A natural circulation cooldown and RCS depressurization are in progress in accordance with 19002-C, "ES-0.2, Natural Circulation Cooldown."
-
One CRDM fan is out of service.
Bhich one of the following plant conditions is a MINIMUM acceptable RCS subcooling and cooldown rata?
Subcoolina Cooldown Rate a.
50 F 15 F/hr b.
75 F 25 F/hr c.
100 F 35 F/hr d.
125 F 45 F/hr
)
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QUESTION: 088 (1.00)
Given the following:
-
A steam generator tube leak has occurred on unit 1.
-
18009-C, " Steam Generator Tube Leak," step 1 RNO states that if loss of PRZR level or pressure is imminent to trip the reactor and enter 19000-C.
Which one of the following is the HIGHEST pressurizer level at 100% power normal full load Tave, that will result in less than 9% pressurizer level at normal no-load Tave?
a.
50%
b.
40%
c.
30%
d.
20%
QUESTION: 089 (1.00)
Given the following:
-
Unit 1 is at 100% power.
-
I & C is performing a monthly ACOT on pressurizer level channel LT 461 (trip channel III).
-
Power is lost from 120 VAC instrument bus lAYlA.
Which one of the following is a consequence of this loss of power?
a.
Loss of power to the safety sequencer b.
All power is lost to main control board annunciator systems d.
Actuation of a safety injection signal
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QUESTION: 090 (1.00)
Which one of the following is the principal concern during an uncontrolled rod withdrawal ATWT?
a.
RCS overpressure b.
DNBR reduction c.
Containment overpressure d.
RCS overtemperature QUESTION: 091 (1.00)
Given the following:
-
A loss of all AC power has occurred on Unit 1.
-
19100-C, "ECA-0.0, Loss of All AC Power," is in progress.
-
Step 3, " Check if RCS is isolated:" reveals that the pressurizer PORVs are OPEN.
-
RCS pressure is 2340 psig.
Which one of the following actions complies with 19100-C?
a.
Closing the PORVs when RCS pressure is 2330 psig, b.
Closing the PORVs when RCS pressure is 2290 psig.
c.
Immediately closing the PORVs.
d.
Immediately closing the PORV block valves.
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QUESTION: 092 (1.00)
Given the following:
-
An ATWT has occurred on Unit 1.
-
19211-C, "FR.S-1, Response to Nuclear Power Generation /ATWT," is in progress.
-
At step 2, it is determined that the turbine did NOT trip, and will NOT manually trip.
Which one of the following is the next action that should be taken per 192117 a.
Trip the generator.
b.
Trip tt.a turbine locally.
c.
Run the turbine back.
d.
Shut the MSIVs.
QUESTION: 093 (1.00)
Given the following:
-
The B Train Essential Chill Water coil in the cubicle ccoler for the CCW pump room was locally isolated after it developed a leak.
Bhich one of the following is the operability impact of this failure?
a.
B Train ESF room cooler is INOPERABLE.
b.
B Train RHR pump is IN0PERABLE.
c.
No impact as long as the normal chilled water cooling coil is in service, d.
No impact as long as CCW pump room temperature is below 105 O.-
O,--
SENIOR REACTOR OPERATOR Page 65
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QUESTION: 094 (1.00)
Which one of the followir.g is the reason why pressurizer level is programmed with Tave?
a.
To maintain a constant mass of coolant in the RCS.
b.
To maintain a constant volume of coolant in the RCS.
c.
To minimize the effect of Tave change on RCS pressure.
d.
To minimize the effect of reactor power change on RCS pressure.
QUESTION: 095 (1.00)
Which one of the following plant conditions will clear the interlock to allow the CCP-1A to RWST recirculation valve (HV-8508A) to be MANUALLY opened?
RHR to CCP/ SIP Suction Valves VCT Outlet Valves 8804A 8804B 1128 112C
,
a.
Open Open Closed Open b.
Closed Open Closed Closed c.
Open Closed Open Open d.
Closed Closed Open Closed j
I
.
... SEMIOR REACTOR OPERATOR Page 66
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QUESTION: 096 (1.00)
Which one of the following correctly describes a cause for a R0D DEVIATION / RADIAL TILT annunciator?
a.
Rod sequence error b.
Quadrant power tilt ratio greater than 1.02 i
c.
DRPI general warning alarm
,
d.
12 step difference between any two control banks demand position
!
QUESTION: 097 (1.00)
If a contingency action in the Response Not Obtained (RNO) column of an E0P cannot be performed or is not successful, and further contingency instructions are NOT provided, which one of the following demonstrates the actions required to be taken?
a.
Remain at the RNO step until the actions can be successfully performed.
-
b.
Transition to the appropriate FRG procedure for the function desired to be restored.
I c.
Transition to 19005-C "ES-0.0, Rediagnosis."
d.
Continue with the next step or substep in the left hand
column.
,
i
SERIOR REACTOR OPERATOR I )~
e Page 67
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QUESTION: 098 (1.00)
Which one of the following signals will directly cause the steam generator blowdown sample isolation valves to automatically close?
a.
MDAFW auto start signal.
b.
Feedwater isolation signal.
c.
Containment isolation phase A signal.
d.
RE-0021 steam generator blowdown high radiation signal.
QUESTION: 099 (1.00)
In MODE 6 with refueling cavity level less than 23 feet above the reactor vessel flange, two trains of RHR are required to be operable by Technical Specifications.
Which one of the following correctly describes the basis for this requirement?
a.
To ensure the effect of a dilution accident is minimized b.
To ensure a source of makeup water is available to the refueling cavity.
c.
To ensure a single failure will not result in a complete loss of residual heat removal capability d.
To ensure both NSCW and CCW trains will be operating with a heat load
. -
.
e
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SEN.IOR REACTOR OPERATOR -
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QUESTION: 100 (1.00)
Which one of the following is the correct reason that RCPs are required to be tripped at 1375 psig during a small break LOCA?
a.
To eliminate RCP heat input into the RCS.
b.
To limit hydrogen buildup which could preclude core cooling.
c.
To avoid the higher clad temperature consequences of RCP trip later during the accident.
d.
To prevent RCP seal damage during subsequent RCS depressurization.
(********** END OF EXAMINATION **********)
-
O O
SEN.IOR REACTOR OPERATOR Page 69 w
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ANSWER:
001 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28, E.2, C.I.j VEGP LO-LP-28103-14-C, pg. 14 and 36 Training Text 8C, Figure 8c-6 KA 061000A301 (4.2/4.2) 061000G014 (3.9/4.0)
061000K402 (4.5/4.6) 000054A203 (4.1/4.2) 000054G010 (3.4/3.3)
061000A301
..(KA's)
ANSWER:
002 (1.00)
c.
(+1.0)
REFERENCE:
Heaters at min.=2250, Sprays full open = 2310, PORVs open = 2335 Operator System Master Plan Cluster 16 Pressuriver pressure control and protection, A.3 LO-LP-16303-13-C P.9, 10, 11 and 12 KA 010000A302 (3.6/3.5)
010000K403 (3.8/4.1)
010000A302
..(KA's)
ANSWER:
003 (1.00)
d.
(41.0)
REFERENCE:
Operator System Master Plan Cluster 18, F.8
Text Chapter 13a, P.52 LO-LP-18501-10-C P.14
^
Al KA's)
ANSWER:
004 (1.00)
d.
(+1.0)
,
"'
&
/
SEN.IOR REACTOR OPERATOR V Page 70 w
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REFERENCE:
VEGP 17019-1, rev. 11, pp. 11, and 19.
Operator System Master Plan Cluster 30, D.9.a LO-LP-30201, p. 15 KA 000051A202 (3.9/4.1)
000051A202
..(KA's)
ANSWER:
005 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37, H.5.b & H.9 LO-LP-37031-10-C P.10 KA 000055K302 (4.3/4.6)
000055K302
..(KA's)
ANSWER:
006 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 47, Liquid Waste Proc., 2 & 5 VEGP Text 17b, P. 20, and 24.
KA 068000A404 (3.8/3.7)
068000A404
..(KA's)
ANSWER:
007 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63 Radiation and Contamination Control, 2.f LO-LP-63930-05, pg. 23 KA 19400lK103 (2.8/3.4)
19400lK103
..(KA's)
-
-
-
'i g
(J SEN,IOR REACTOR OPERATOR,
Page 71
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ANSWER:
008 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63, Rules for Performing Procedures, 1.a LO-LP-63054-07-C, pg. 11 KA 194001A102 (4.1/3.9)
194001A102
..(KA's)
ANSWER:
009 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37, A.6 & 7 LO-LP-37041-07-C P.11 19211-C KA 000029K312 (4.4/4.7) 000029G010 (4.5/4.5)
000029K312
..(KA's)
ANSWER:
010 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 16, H.4 LO-LP-16303-13-C P. 12 KA 010000K403 (3.8/4.1)
010000K403
..(KA's)
ANSWER:
011 (1.00)
c.
(+1.0)
a
&
SEN.IOR REACTOR OPERATOR Page 72
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-
,
.
.
REFERENCE:
Operator System Master Plan Cluster 28 RPS/ESFAS Signals, A.8 LO-LP-28103-14C P.23 and 50 KA 013000K412 (3.7/3.9)
013000K412
..(KA's)
ANSWER:
012 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28, RPS/ESFAS signals, C.7 LO-LP-28103-14C, P.25, 26, and 49 Logic 7243D07 sheet 13 Feedwater control and isolation KA 059000A411 (3.1/3.3)
059000A411
..(KA's)
ANSWER:
013 (1.00)
b.
(+1.0)
REFERENCE:
,
Operator System Master Plan Cluster 37, Respond to Rx Trip, D.6 LO-LP-37011-08-C P.10 and 21 19000-C, Step 3 RNO Column.
KA 000056G010 (3.7/3.9)
000056G010
..(KA's)
ANSWER:
014 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28, RPS/ESFAS signals A.3 LO-LP28103-14-C P.8, 18, 19, and 39 KA 003000K304 (3.9/4.2)
003000K304
..(KA's)
&
A SENJOR REACTOR OPERATOR iur 4WF Page 73
,
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ANSWER:
015 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28, RPS/ESFAS Signals A.3 LO-LP-28103-14C P.16, 17, 11, 39 KA 000007G011 (4.1/4.3)
000007G011
..(KA's)
ANSWER:
016 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 9, CVCS Overview, A.7 LO-LP-09001-06 P.14, 23, 24 LO-LP28103-14-C P.18, 39 LO-LP-16301-16-C P.10 KA 011000A102 (3.3/3.5)
011000A102
..(KA's)
'
ANSWER:
017 (1.00)
a.
(+1.0)
!
)
REFERENCE:
Operator System Master Plan Cluster 17, A.5 LO-LP-17201-09-C P.22 and 23 i
KA 000033G0ll (3.2/3.4)
000033G011
..(KA's)
!
!
'i ANSWER:
018 (1.00)
a.
(+1.0)
{
,
a S
SENJ0R REACTOR OPERATOR Page 74 w
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.
REFERENCE:
Operator System Master Plan Cluster 17, A.8 LO-LP-17201-09-C P.24 KA 015000K602 (2.6/2.9)
015000K602
..(KA's)
ANSWER:
013 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 17, A.12 LO-LP-17101-08C P.36 Technical Specifications 3.3.1, 3.0.3 KA 015000G005 (3.3/3.8)
015000G005
..(KA's)
ANSWER:
020 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 36 MCD: Vital Instrumentation, 5 LO-LP-36104-07-C P.7, 17 KA 000008K301 (3.7/4.4)
000008K301
..(KA's)
ANSWER:
021 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60, "]".3 LO-LP-60328-07-C P.38 KA 000068K318 (4.2/4.5)
000068K318
..(KA's)
,
.
.--
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_., lSEN,IOR REACTOR OPERATOR Page 75
.
.
ANSWER:
022 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63, Temporary Changes to Procedures, 2.c LO-LP-63052-05, P.10
'
00052C Rev 7, Page 4, Step 4.2.4 KA 194001A101 (3.3/3.4)
194001A101
..(KA's)
,
ANSWER:
023 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63, Shift Relief, A.I.b LO-LP-63504-09 P.6, B.2.a.3 KA 194001A103 (2.5/3.4)
194001A103
..(KA's)
.
i ANSWER:
024 (1.00)
a.
(+1.0)
,
REFERENCE:
.
Operator System Master Plan Cluster 63, Shift Relief, S.1 & 4
'
10004C P.2 and 3 KA 194001A103 (2.5/3.4)
.'
194001A103
..(KA's)
!
ANSWER:
025 (1.00)
c.
(+1.0)
.
i e
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SENJOR REACTOR OPERATOR,.,
Page 76
,
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..
REFERENCE:
Operator System Master Plan Cluster 63, Independent Verification
Policy, K.2.a & b i
00308-C P.4, 4.2.1.1 KA 19400lK101 (3.6/3.7)
19400lK101
..(KA's)
,
ANSWER:
026 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60 Loss of Class IE Electrical System, X.1 LO-LP-60323-03 P.25 A01-18031, step A.1 KA 000055G010 (4.1/4.3)
000055G010
..(KA's)
ANSWER:
027 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Lluster 63 Rules for performing procedures, D.I.c LO-LP-63054 P.12 00054-C P.3 and 4 KA 194001A102 (4.1/3.9)
194001A102
..(KA's)
ANSWER:
028 (1.00)
a.
(+1.0)
e A
SENJ0R REACTOR OPERATOR ---
Page 77
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,
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REFERENCE:
Operator System Master Plan Cluster 63 Authority to startup and shut down reactors, A.3 LO-LP-63300-05 P.5 and 6 00300-0 P.2 KA 194001Alli (2.8/4.1)
194001A111
..(KA's)
ANSWER:
029 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63 Containment Entry, H.2 LO-LP-63363-03 P.6 and 15 00303-C P.2 KA 194001All2 (3.1/4.1)
194001All2
..(KA's)
ANSWER:
030 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63 Equipment Clearance and Tagging, 1.4.a 00304-C P.12 KA 19400lK102 (3.7/4.1)
,
19400lK102
..(KA's)
ANSWER:
031 (1.00)
d.
(+1.0)
a e
SENJOR REACTOR OPERATOR Page 78
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-
,
.
.
REFERENCE:
Operator System Master Plan Cluster 63 Equipment Clearance and Tagging, 1.6 LO-LP-63304-14 p.38 00304-C p.18 and 21 KA 19400lK102 (3.7/4.1)
194001K102
..(KA's)
ANSWER:
032 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63 Maintenance Program, L.2 LO-LP-63350-09-C P.19 00350-C P.15 KA 194001A112 (3.1/4.1)
194001All2
..(KA's)
ANSWER:
033 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 63, N.4.c LO-LP-63404-09 P.21 00404-C P.6 KA 194001A109 (2.7/3.9)
194001A109
..(KA's)
ANSWER:
034 (1.00)
a.
(+1.0)
_
_
. _.
.
,
h SINIORREACTOROPERATOR Page 79
_,
..
.
REFERENCE:
i Operator System Master Plan Cluster 39 LC0 and Surveillance Applicability, G.4 KA-194001A111 (3.1/4.1)
!
194001Alli
..(KA's)
!
ANSWER:
035 (1.00)
b.
(+1.0)
t i
REFERENCE Operator System Master Plan Cluster 63 Rad. Exp. Limits, "[".6 LO-LP-63920-04-C P.13
,
00920-C P.6 l
KA 19400lK104 (3.3/3.5)
i 19400lK104
..(KA's)
-
,
l ANSWER:
036 (1.00)
b.
(+1.0)
.
REFERENCE:
i Operator System Master Plan Cluster 63 Rad. Exp. Limits, "[".7 LO-LP-63920-04-C P.13
,
'
00920-C P. 3, 4, and 6 KA 194001K104 (3.3/3.5)
>
19400lK104
..(KA's)
,
ANSWER:
037 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 40 EPIP Overview,17
.
LO-LP-40101-21-C P.22 i
KA 194001A116 (3.1/4.4)
194001All6
..(KA's)
l
i i
_
SENJ0R R'EACTOR OPERATOR Page 80
,
.
.
t ANSWER:
038 (1.00)
d.
(+1.0)
REFERENCE:
i Operator System Master Plan Cluster 36 MCD Excore Response to Core Damage, 3 LP-LP-36103-06-C P.10-12, 14-18, KA 000074A101 (4.2/4.4) 015020A202 (3.3/3.8)
000074A101
..(KA's)
,
ANSWER:
039 (1.00)
k c.
(+1.0)
+
REFERENCE:
Operator System Master Plan Cluster 28 RPS/ESFAS Signals, 2.J LO-
.
LP-28103-14-C P.39, 26
'
KA 059000A412 (3.4/3.5)
059000A412
..(KA's)
--
ANSWER:
040 (1.00)
a.
(+1.0)
'
REFERENCE:
Operator System Master Plan Cluster 11 EDG Control and Indication, 16a and b
'
LO-LP-ll201-16-C P.42 "Special Situations" KA 064000A406 (3.9/3.9) 064000K402 (3.9/4.2)
,
064000A406
..(KA's)
'
ANSWER:
041 (1.00)
c.
(+1.0)
)
I
]
O O
SENIOR REACTOR OPERATOR Page 81
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-
,
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.
REFERENCE:
Operator System Master Plan Cluster 17 Incore Instrumentation, 7 LO-LP-17401-12C P.31, 35 (Indication of failed thermocoupie)
KA 017020K403 (3.1/3.3)
017020K403
..(KA's)
ANSWER:
042 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28 RPS and ESFAS Signals, C.4 LO-LP-28103-14-C P.20, 42 LO-LP-16401-18 P.15 Training Text la P.11 (Flywheel)
VA 003000K502 (2.8/3.2)
003000G007 (3.2/3.3)
003000K502
..(KA's)
ANSWER:
043 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 16 Reactor Vessel Level Indication, L.4.
LO-LP-16701-09-C P.5, 6, 15 K/A 002000K603 (3.1\\3.6)
002000K603
..(KA's)
ANSWER:
044 (1.00)
b.
(+1.0)
O O
SENJ0R REACTOR OPEP,ATOR w W
Page 82
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.
.
REFERENCE:
Operator System Master Plan Cluster 37 Operator response to a loss of primary coolant, 5; and 37 Loss of secondary coolant faulted steam generator isolation, 6 LO-LP-37111-09-C P.16 LO-LP-37121-10-C P.16 KA 0000llA213 (3.7/3.7)
0000llA213
..(KA's)
ANSWER:
045 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 21 Steam Dumps, D.10 LO-LP-21201-18 P.14, 27 KA 041020K105 (3.5/3.6)
041020K105
..(KA's)
l ANSWER:
046 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 12 RHR System, 6 LO-LP-12101-31-C P.18 KA 005000K410 (3.1/3.1) 005000K402 (3.2/3.5) 002000A103 (3.7/3.8)
002000A103
..(KA's)
ANSWER:
047 (1.00)
b.
(+1.0)
.
.
~
, SENIOR REACTOR OPERATOR Page 83
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.
.
REFERENCE:
Operator System Master Plan Cluster 16 Cold Overpressure Protection System, 3 LO-LP-16501-06-C P.7 Training Text Id, Figure Id-12 KA 010000K403 (3.8/4.1)
010000K403
..(KA's)
ANSWER:
048 (1.00)
.
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster. 60, Control Rod System Malfunction, 0.21-TS 3.2.1 LO-LP-60303-13-C P.36 KA 000003K305 (3.4/4.1) 000003G003 (3.3/3.8)
000003G003
..(KA's)
ANSWER:
049 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 9 CVCS-L/D flowpath, B.3.j i
Training Text 5A P 29, 33, 34 KA 004010A211 (3.1/3.1)
004010A211
..(KA's)
ANSWER:
050 (1.00)
b.
(+1.0)
A e
SENJ0R REACTOR OPERATOR -
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,
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.
REFERENCE:
Operator System Master Plan Cluster 39 Reactivity Control Systems, A.2 LO-LP-39205-08-C P.15 Technical Specifications 3.1.1.4, 30 minute action, 3/4 1-6 KA 002000G005 (3.6/4.1)
002000G005
..(KA's)
ANSWER:
051 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Inad. core cool., B.7 & 8 LO-LP-37061-08 P.19; LP-LP-37002 p. 17.
19200-C P.4-10 KA 000040G012 (3.8/4.1)
000040G012
..(KA's)
ANSWER:
052 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 09 CVCS Overview, A.S.a LO-LP-09001-06 P.12 Solid Plant Ops KA 004010K101 (3.4/3.9)
004010K101
..(KA's)
ANSWER:
053 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Loss of Heat Sink, J.ll LO-LP-37501-10-C P.21 KA 000054A104 (4.4/4.5)
000054A104
..(KA's)
n S
SENJ0R REACTOR OPERATOR v Page 85
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,
.
.
ANSWER:
054 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 16 Reactor Coolant Pump, I.4 Training Text la P.27 13003-1 RCP Operation P.1 LO-LP-16401-18 P.19, 21 KA 000015K207 (2.9/2.9) 000015Al22 (4.0/4.2)
015000G007 (3.1/3.2)
000015K207
..(KA's)
ANSWER:
055 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 39 TS 3/411 and 12, P.2 i
LO-LP-39215-04 P.5, 10 Technical Specifications pg. 3/411-9, Immediate requirement KA 071000A429 (3.0/3.6) 071000K504 (2.5/3.1)
071000A429
..(KA's)
ANSWER:
056 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 32,12 and 7
'
Logic 7243007 Sht 8 LO-LP-32101-18-C P.11 KA 072000K102 (3.5/3.9)
072000K102
..(KA's)
A e
SENIOR REACTOR OPERATOR Page 86 w
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,
.
.
ANSWER:
057 (1.00)
b.
(+1.0)
REFERENCE:
,
Operator System Master Plan Cluster 60 Rod Control System Malfunction, D.12 18003-C Rod Control System Malfunction, P.8,9 LO-LP-60303-13-C P.33, 34 KA 000001G010 (3.9/4.0)
)
000001G010
.. (rsA's)
<
ANSWER:
058 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60 Control Rod System Malfunction, D.11 LO-LP-60303 P.33 KA 045050K301 (2.9/3.2)
045050K301
..(KA's)
ANSWER:
059 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 43 Fire Protection System, 9 LO-LP-43101-11-C P.32 KA 086000A202 (3.0/3.3) 086000K402 (3.0/3.4)
086000A202
..(KA's)
ANSWER:
060 (1.00)
a.
(+1.0)
}
Page'87
, SENIOR REACTOR OPERATOR
.
.
REFERENCE:
.0perator System Master Plan Cluster 30 Turbine Protection, 0.12 17019-1 P.19 LO-LP-30201-10 P.15
KA 000051A202 (3.9/4.1)
000051A202
..(KA's)
-.
t ANSWER:
061 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60 Loss of RHR, 6 18019-C P.10, 24 LO-LP-60315-09-C P.24, 41 Generic Letter 88-17 KA 000025K101 (3.9/4.3)
000025K101
..(KA's)
.
ANSWER:
062 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60, 6 18019-C P.10, 24 LO-LP-60315-09-C P.24, 41 NRC Generic Letter 88-17 Loss of Decay Heat Removal, Enclosure 1, Step 2.1.1 " Pressurization."
KA 002000G010 (3.4/3.9)
002000G010
..(KA's)
,
.
ANSWER:
063 (1.00)
'
a.
(+1.0)
i I
i i
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SENIORREACTOROPERATOR()
(
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Page 88
,
!
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.
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REFERENCE:
Operator System Master Plan Cluster 36 MCD Core Ctoling Mechanics, 12 14 15 and cluster 37 Operator Response to a LOCA, 6 LO-H0-37111-001-04-C P.1-5 (1" to 1 ft sq break)
LO-LP-37111-09-C P.7
,
KA 000009K101 (4.2/4.7)
000009K101
..(KA's)
ANSWER:
064. (1.00)
b.
(+1.0)
REFERENCE:
,
Operator System Master Plan Cluster 31 Main Generator Construction and Operation, A.6 LO-LP-31101-12-C P.30 KA 000007K301 (4.0/4.6)
.
000007K301
..(KA's)
!
ANSWER:
065 (1.00)
a.
(+1.0)
,
'
REFERENCE:
Operator System Master Plan Cluster 37 Response to Inadequate
[
Core Cooling, K.6 i
LO-H0-37061-001-03 P.1-2 & 1-5
,
LO-LP-37061-08 P.15, 16
-
KA 000008A212 (3.4/3.7)
000008A212
..(KA's)
.
I ANSWER:
066 (1.00)
b.
(+1.0)
i
,
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i
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.
.
e
a A
SENIOR REACTOR OPERATOR Page 89
=-
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.
.
REFERENCE:
Operator System Master Plan Cluster 37, K.4 and 6 LO-H0-37061-001-03 P.1-17 (Core Exit Superheat)
KA 000074A207 (4.1/4.7) 017020K503 (3.7/4.1)
017020K503
..(KA's)
ANSWER:
067 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 39 Safety Limits and Limiting Safety System Settings, 3 8 0-LP-39203-06 P.7 Technical Specifications 2.1.2, and 6.6.1 V.A 002000G003 (2.7/4.1)
002000G003
..(KA's)
ANSWER:
068 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 39, I.4 LO-LP-39208-08-C P.18 Technical Specifications Bases 3/4 4-7 KA 000076G004 (2.1/3.7)
000076G004
..(KA's)
ANSWER:
069 (1.00)
d.
(+1.0)
,
I
-
- -- --.--.-----
e a
SENIOR REACTOR OPERATOR Page 90
-
-
,
,
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REFERENCE:
Operator System Master Plan Cluster 37 Loss of All AC Power, 7 and 9 LO-LP-37031-10-C P.7 19100-C P.2 Note KA 000056G012 (3.4/3.6)
000056G012
..(KA's)
ANSWER:
070 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Operator Response to a SGTR, R.5 LO-LP-37311-09-C P.16 KA 000038A136 (4.3/4.5)
000038A136
..(KA's)
ANSWER:
071 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Operator Response to a SGTR, R.10 LO-LP-37311-09-C P. 14 & 30 KA 000038K308 (4.1/4.2)
000038K318
.(KA's)
ANSWER:
072 (1.00)
c.
(+1.0)
....
.
_
_
,
(k
,. SENIOR REACTOR OPERATOR Page 91
'
..
..
REFERENCE:
Operator System Master Plan Cluster 39 TS ECCS, 4
'LO-LP-39209-05-C, p. 10
,
TS Bases P.3/4 5-2 i
KA 006000G006 (2.9/4.0)
006000G006
..(KA's)
ANSWER:
073 (1.00)
-
d.
(+1.0)
REFERENCE:
,
Operator System Master Plan Cluster 16 Reactor Coolant Pump, I.1 Vogtle Text la, figure h-9 LO-LP-16401-18 P.11,12,20,26 KA 003000K602 (2.7/3.1)
003000K602
..(KA's)
,
ANSWER:
074 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 18 SGWLS, F.7 LO-LP-18501-10-C P.25 KA 035010K402 (3.2/3.S)
035010K402
..(KA's)
,
ANSWER:
075 (1.00)
,
a.
(+1.0)
,
REFERENCE:
Operator System Master Plan Cluster 37 Use of E0Ps, B.8.c LO-LP-37002-10-C P.11,17 KA 000011G012 (4.0/4.1)
000011G012
..(KA's)
,
j
,
_
e,--
S SENIOR REACTOR OPERATOR Page 92
--
.
.
.
ANSWER:
076 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Natural Circulation Cooldown, E.10 LO-LP-37012-ll-C P.12,17,18 KA 002000A203 (4.1/4.3)
002000A203
..(KA's)
ANSWER:
077 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Op. Response to SGTR, R.9 19030-C, Step 29 LO-LP-37311-09-C P.20,30 KA 000038K306 (4.2/4.5) 000038A215 (4.2/4.2)
000038A215
..(KA's)
ANSWER:
078 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Use of E0Ps, B.7 LO-LP-37002-10-C P.ll, 13, 17 KA 000029G012 (4.1/4.2)
000029G012
..(KA's)
ANSWER:
079 (1.00)
b.
(+1.0)
a A
SENIOR REACTOR OPERATOR w Page 93
-
,
.
.
REFERENCE:
Operator System Master Plan Cluster 37 Operator Response to a loss of Primary Coolant, M.9 19010-C, Caution before step 9 LO-LP-37111-09-C P.17 KA 064000K411 (3.5/4.0)
064000K411
..(KA's)
ANSWER:
080 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 29 Containment Coolers, A.2 LO-LP-29130-04 P.8; 28201 p. 9 KA 022000K201 (3.0/3.1) 022000A401 (3.6/3.6)
022000K201
..(KA's)
ANSWER:
081 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 12 RHR System, A.8 LO-LP-12101-31-C P.13, 14 and 40 KA 000025K302 (3.3/3.7) 005000K407 (3.2/3.5) 000025A206 (3.2/3.4)
000025A206
..(KA's)
ANSWER:
082 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master olan Cluster 37 SI/ Inadvertent S1 Response, 3 LO-LP-37021-10 P.15 KA 000011G011 (4.3/4.5)
000011G011
..(KA's)
e
SENIOR REACTOR OPERATOR -
Page 94 j
-
,
i
.
e ANSWER:
083 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 28 Sequencer Operation, 3 LO-LP-28201-16-C P.18 KA 000056A206 (3.5/3.6)
000056A206
..(KA's)
ANSWER:
084 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Operator Response to a LOCA, M.2 LO-H0-37061-001-03 P.1-3, 1-4 KA 000008K101 (3.2/3.7)
000008K101
..(KA's)
ANSWER:
085 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 Loss of Sec Coolant / Faulted SG isol, Q.5.b LO-LP-37121-10-C P.8,9; LO-LP-37121-01-C, 1-5 KA 000040K105 (4.1/4.4)
000040K105
..(KA's)
ANSWER:
086 (1.00)
c.
(+1.0)
'
,
SENIOR. REACTOR OPERATOR Page 95
,
t
'
REFERENCE:
Operator System Master Plan Cluster 37 Natural Circulation Cooldown, E.13 LO-LP-37012-11-C P. 6 19001-C, p. 17 KA 000074A106 (3.6/3.9)
000074A106
..(KA's)
ANSWER:
087 (1.00)
d.
(+1.0)
,
REFERENCE:
'
Operator System Master Plan Cluster 37 Natural Circulation, 15 LO-LP-37012-ll-C P.18 19002-C, Caution prior to step 12 KA 000015A121 (4.4/4.5)
000015A121
..(KA's)
ANSWER:
088 (1.00)
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 60 Steam Generator Tube Leak, 5.c and 1 LO-LP-60309-09-C P.20, 12 KA 000037K305 (3.7/4.0)
000037K305
..(KA's)
ANSWER:
089 (1.00)
b.
(+1.0)
j l
i
SENIORREACTOROPERATORO O
Pa9e.96
,
..
.
REFERENCE:
'.0perator System Master Plan Cluster 60 Loss of 125 vac instrument power, Y.1 & 5 LO-LP-60324-05-C P.34, 39 KA 000057A219 (4.0/4.3)
000057A219
..(KA's)
ANSWER:
090 (1.00)
b.
(+1.01 REFERENCE:
Operator System Master Plan Cluster 37, ATWT, I.4 LO-LP-37041-07-C P.8 KA 000029K101 (2.8/3.1)
000029K101
..(KA's)
,
ANSWER:
091 (1.00)
,
b.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37, H.6 LO-LP-37031-10-C P.7 19100-C KA 000055G010 (4.1/4.3)
000055G010
..(KA's)
ANSWER:
092 {1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 37 ATWT, 7 LO-LP-37041-07-C P.16 19211-C P.2 KA 000029G010 (4.5/4.5)
000029G010
..(KA's)
,
.
.
.
...
. - -
,
e A,--
SENIOR REACTOR OPERATOR Page 97
---
,
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.
ANSWER:
093 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 23 ESF Room Coolers, 1 and 4 LO-LP-23203-06 P.15 KA 008000K301 (3.4/3.5)
008000K301
..(KA's)
ANSWER:
094 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 16 Pzr Level control, G.1 LO-LP-16302-09-C P.24 Logic Diagram 7243D07 KA 011000K512 (2.7/3.3)
011000K512
..(KA's)
ANSWER:
095 (1.00)
d.
(+1.0)
REFERENCE:
l Operator System Master Plan Cluster 13 HHSI, B.4 LO-LP-13101-13 P.12,17 KA 006000K409 (3.8/4.1)
006000K409
..(KA's)
ANSWER:
096 (1.00)
a.
(+1.0)
e
, SENIOR REACTOR OPERATOR -
Page 98
-
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,
REFERENCE:
Operator System Master Plan Cluster 27, Rod Pos. Ind. Sys., 7.f LO-LP-27201-09-C P.15 KA 014000G006 (3.4/3.7)
014000G006
..(KA's)
ANSWER:
097 (1.00)
d.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster B.2 LO-LP-37002-02-02, p. 5 & 6 KA 000040G012 (3.8/4.1)
000040G012
..(KA's)
ANSWER:
098 (1.00)
a.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 20,
VEGP LO-LP 24101-15, pg. 12.
KA 061000K414 (3.5/3.7)
061000K101 (4.1/4.2)
061000K414
..(KA's)
ANSWER:
099 (1.00)
c.
(+1.0)
REFERENCE:
Operator System Master Plan Cluster 39 Refueling TS, N.4 VEGP LO-LP-39213-05, p. 10 TS B 3/4 9.8 KA 005000G006 (2.7/3.6)
005000G006
..(KA's)
c f-~)s SENIORREACTOROPERATORk_f)
i Page 99 t.
m
,
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.
ANSWER:
100 (1.00)
c.
(+1.0)
REFERENCE:
LO-LP-37111-09-C, pg. 10 Operator System Master Plan Cluster 37, M.6 KA 000009K323 (4.2/4.3)
000009K323
..(KA's)
hk l
(********** END OF EXAMINATION **********)
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