ML20236M847
| ML20236M847 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/27/1987 |
| From: | Moorman J, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236M550 | List: |
| References | |
| 50-348-OL-87-03, 50-348-OL-87-3, NUDOCS 8711130309 | |
| Download: ML20236M847 (147) | |
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3 ENCLOSURE 1
. EXAMINATION REPORT 348/0L-87-03 Facility Licensee:
Alabama Power Company 3
600 North 18th Street i
Birmin0 ham, AL 35291-0400 i
Facility Name:
J. M. Farley Nuclear Plant
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l Facility Docket No.:
50-348 and 50-364 1
Written examinations and operating tests were administered at J. M. Farley 1
Nuclear Plant near Ashford, Alabama.
j Chief Examiner: r/79vf/hNcntw #
/0 /d/s 7 Ja e H. Moorman, III Date' Signed
/d/c27[f7.'-
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l Approved by:
John F. Munro,~Section Chief
~ Tdte Signed l
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SUMMARY
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Examinations were administered on August 24-27, 1987.
Written and operating j'
l (oral and simulator) examinations were given to five Reactor Operators (RO) and nine Senior Reactor Operators (SR0).
Three R0s passed the written examination and five R0s passed the operating examination.
Nine SR0s passed the written l
and operating examinations.
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Based on the results described above, three of five R0s and nine of nine SR0s i
passed the overall examination.
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8711130309 871003
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PDR ADOCK 05000348 i
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REPORT DETAILS j
1.
Facility Employees Contacted:
- R. Hill,' Operations' Manager j
- T. Horne, Instructor 1
- C, McLean, Instructor
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- 0. - Morey,i Assistant: General Manager-Operations
- W. Shipman, Assistant-General-Manager-Support
- R. Wiggins,: Operations Training Supervisor
- L. Williams, Training Manager j
2.
Examiners:
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- R. Baldwin,. Region'II l
- S. Bitter, Region II H
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- J. Moorman,-Region II I
- C, Rapp, Region II j
F. Victor, Sonalysts, Inc.
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Examination Review Meeting
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- W. Bradford, SRI I
- W. Miller, RI
- Chief' Examiner
- Attended Exit Meeting
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,j At the conclusion of the written _ examinations, the examiners provided youri 1
training staff with a copy of the-written' examination and answer key for y
review.
The comments made by the facility reviewer are incluaed as j
, to this report.
The NRC Resolutions to facility comments are u
listed below.
Questions marked with an asterisk (*) weie commented on due Lto being -
prepared with inadequate or insufficient reference t material.provided.
by the facility.
Four of twenty-five (16%) changes were'made to the.
o answer key as a result of this inadequate l training material.
n Several questions required additional examiner: effort toTresolve sincelthe.
l facility provided insufficient reference material to prove the accuracy ofJ q
their comment.
This additional ~ effort caused some. delay in resolution of a
the examination comments and subsequent processing ~ of the: examinations.
The facility. is reminded that ' definitive and: substantive ~ referenceE q
material must be provided to support examination question comments.
a.
NRC Resolution to Written Examination Comments Question 1.04/5.01:
Facility comment: accepted.- The answer key.will be. change'd'to: accept 1 either (s) or (c).for full credit.
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7 Question 1.15(b)/5.16(b):
Facility comment accepted.
The answer key will be changed to require INCREASES for full credit.
Question 1.15(d):
Facility comment noted.
The additional facility documentation supplied to support this comment indicates the correct response ir dependent upon the assumptions made.
Due to its highly subjective nature, this part of the question will be deleted from the examination and the point value adjusted.
Question 2.14(c):
Facility comment accepted.
The additional facility documentation supplied to support this comment indicates that the seal ring.is not located as stated.
The answer key will be changed to require FALSE for full credit.
- Question 2.19(d):
Facility comment noted.
Additional documentation supplied by the facility to support this comment indicates that local operation is required to complete the line-up.
However, the additional material does not support this line-up as a " design feature" as solicited by the question and defined in the facility reference material.
to the answer key required.
. No change Question 2.20:
Facility comment accepted.
The answer key will be changed to require the respon e recommended by the facility for full credit.
Question 3.10(a):
Facility comment noted.
Since this part of the question was not sufficiently specific to focus attention on the Motor Driven AFW pumps, it will be deleted from the examination and the point value adjusted.
Question 3.12(a):
Facility comment accepted.
The answer key will be changed to require
</= 402.5 psig for full credit.
The answer key and the facility suggested response are functionally equivalent.
Question 3.18(b):
Facility comment noted.
Interlock conditions are not required for full credit.
No. change to the answer key required.
8 Question 3.22(a):
Facility comment accepted.
The answer key will be changed to accept answers related to dropping of control rods.
Question 3.24:
Facility comment accepted.
The question will be deleted from the examination as recommended by the facility.
Question 4.05(c):
Facility comment accepted.
The answer key will be changed to require COULD NOT for full credit.
Question 4.09/7.14:
Facility comment accepted.
The answer key will be changed to require the response recommended by the facility for full credit and the point value adjusted.
Questior. 4.10/7.15:
Facility comment accepted.
The answer key will be changed to accept OTdT Rod Stop alarm as an additional possible answer.
Question 4.14(b)/8.09(b):
Facility comment accepted.
The answer key will be changed to require the response recommended by the facility for full credit.
Additionally, the answer key for part (a) of this question will be changed to maintain consistency.
The point value will be adjusted to allow for the additional required responses.
Question 4.19:
Facility comment not accepted.
The supporting documentation provided by the facility specifically restates the required response on the answer key.
However, due to potential confusion with the ' term INCREASED and incorrect proctorial guidance during the examination, the question will be deleted from the examination.
The scope of this question is considered to be a valid testing area.
- Question 4.22:
Facility comment accepted.
The answer key will be changed to delete response 5 and to accept removal of the control power fuses as an alternative for response 1.
The point value will be adjusted accordingly.
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Question 5.09:
3 comment accepted.
The answer key will be changed to Facility (b) for full credit.
require Question 5.19:
Facility comment accepted.
The answer key will be changed to accept l
the facility recommended responses for full credit.
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Question 5.20:
Facility comment accepted.
It is reasonable to assume that the reactor could achieve criticality because a single steam dump valve accounts for 5% of rated steam flow. - However, consideration will be i
given to candidates who explicitly state that the extra reactivity inserted is insufficient to cause criticality.
Furthermore, the answer key will be changed to add the effect of MTC in turning' and leveling power.
Question 5.21:
Facility comment not accepted.
Although MTC would be slightly affected by rod movement, the FTC will be the first in reacting to a sudden power change.
No change to the answer key warranted.
- Question 5.22:
Facility comment noted.
Part (e) of the question will be deleted because it is too theoretical in concept for an operator and an operator may not be able to discern response time from control room indications. The facility is reminded that lack of training material or learning objectives does not preclude the use of questions in areas with high importance values listed in NUREG 1122.
Question 6.03:
Facility comment accepted. The onswer key will be changed to require TRUE for part (b) and FALSE for part (c) for full credit.
Question 6.08:
Facility comment accepted. The answer key will be changed to require 475 F as the value for the RHR/ pressurizer vapor space temperature interlock.
10
- Question 6.09:
Facility comment accepted.
The response recommended by the facility will be added to the answer key as.an additional correct answer.
Question 6.13:
Facility comment accepted.
The answer key will be changed to' accept the response recommended by the facility for full credit only if the candidate specifically states that the turbine is latched.
Question 6.16:
Facility comment accepted.
The answer key will be changed:to' accept the response recommended by the facility for full credit.
Question 7.12:
Facility comment accepted.
The. typographical error in the answer key will be corrected.
Question 8.06:
Facility comment accepted.
The typographical error in the answer key will be corrected.
b.
Other Examination Changes The answer key for question 4.05 was changed to allow a more comprehensive list of responses as elicited by the question.
The answer key for question 4.10/7.15 was. changed to allow for responses that a cognizant operator would realistically expect in 1
diagnosing an inadvertent dilution.
Question 7.03 was deleted since there were two possible answers.
Question 5.17 and 7.18 were deleted since the answers were included -
on the students copy of the exam.
The answer to Question 8.11 has been amended to cover all possible situations posed by the question.
4.
Exit Meeting:
At the conclusion of the site visit, the' examiners met with: members of your staff to discuss items pertinent to the operating. exams and the examination process.
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a W'eaknesses ~ in the. construction of facilit9 -
discussed, as well as a ' generic ' weakness i. learning objectives were again 4
1 n the candidates' ab.ility to' identify sources of, measure and state' hazards associated with' activation-d products.
q The candidates' general knowledge of administrative, requirements - as ;
j demonstrated during. the walk-through portion of the. operating.' exam was-
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noted'as a generic strength.
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The cooperation given to the' examiners was noted and appreciated.
The licensee did not identify any of the material"provided to or reviewed:
by the examiners as' proprietary, 5.
Generic Written Examination Comments
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o Weakness in. the candidates' knowledge of ~ Steam Dump _ System operation was -
noted during-grading of the written examinations and during the simul' tor I
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examinations.
Additionally, the' candidates did not' fully understand their~
responsibilities as stated in FNP-0-AP-16.
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NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
FARLEY 1 & 2 REACTOR TYPE:
PWR-WEC3 l
DATE ADMINISTERED:
87/08/24 EXAMINER:
RAPP. CW CANDIDATE NA((k INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 29.S0
_25.71 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 30.00 26.14 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 28.00 24.40 3.
INSTRUMENTS AND CONTROLS 27.25 23.75 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 114.75 Totals Final Grade All work done on this examination is my own.
I have neither given nor received aid.
Candidate's Signature
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1 NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate 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.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
S.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Prin t your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in I
completing the examination.
This must be done after the examination has l
been completed.
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- 18. When you complete your examination, you shall a.
Assemble your examination as follows:
(1)
Exam questions on top.
l (2)
Exam aids - figures, tables, etc.
I (3)
Answer pages including figures which are part of the. answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
I c.
Turn in all scrap paper and the balance of the paper that you did j
not use for answering the questions.
d.
Leave the examination area, as defined by the examiner.
If after
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1eaving, you are found in this area while the examination is still e
in progress, your license may be denied or revoked.
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li PRINCIPLES OF NUCl.E,tk["JWER PLANT ~ OPERATION.'
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THERMODYNAMICS, HEAT TNANSFER AND FLUID FLOW l
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GUESTION 1.01 (1.00) 1)<
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Select the one shatement below that correct 1r completes the following
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During a reacth+ startup, just[ 'hbI0ka to reaching criticality, the SUR indication willir'espond to a given'ab.ount of rod. withdrawal by s
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. l rising 9(1owly f alling spwly' to then zero.
a.)
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.plI19 therp f alling s }9 'owly, to,ztycb b b.) rising y d
5 th[enfallingrapidzyrtogero.
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c.) rising s1W ly N
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d.) rising rap \\dly ther falling rapic'f y. to zero,
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1 QUESTION 1.02 (1.00) l l
Using STEAM TAEd.ES ONLY, select the one correct answer from the following l
for the given situation l below.
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l A system containing saturated steam at 585 pnfg is. leaking throuqh a j
j crack in a pipe wall.
The pressure downstreas of,the crack is 335 psig.
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The condition c1 the flui)d downstream of the track ist j
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wet vapor wdth, ontality near 100%.
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we t va po r v'.t t h a gua l i ty n e.'1e r 60%?
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superheatey # apt;.e with less t h an *[.E dhgrues superheat.
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I superheatet /aporb(ith greator than 2 degrees superheats d.)
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1.
PR5NCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page S
THERMODYNAMICS. KEAT TRANSFER AND FLUID FLOW OUESTION 1.03 (1.00)
Which one of the following statements, concerning the power defect, is correct?
.is a.)
The power def ect causes control rods to be withdrawn as reactor power is decreased.
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b.)
The power' defect increases the rod height requirements necessary-to maintain the desired shutdown margin following a reactor trip.
c.)
Because of the higher baron concentration, the power defect is more negative at beginning of core life.
d.)
The power defect necessitates the use of a ramped Tavg program
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to maintain an adequate Reactor Coolant System subcooling margin.
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QUESTION 1.04 (1.00)
A reactor is shutdown with a baron concentration of 1100 ppm and a SDM of.2 percent delta K per K.
The source range channel indicates 10 cps.
The baron concentration is then reduced by 100 ppm.
Assuming a DBW of -10 pcm/ ppm, choose the one answer that is CLOSEST to the final Keff.
h.) Keff
.980
=
.985 b.) Keff
=
c.) Keff =.990
.995 d.) Keff
=
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1.
' PRINCIPLES OF NUCLEAR POWER PL, ANT OPERATION.
.Page 6
THERMODYNAMICS, HEAT TRANSFER-AND FLUID FLOW QUESTION 1.05 (1.00)
The unit startup procedure, FNP-1-UDP-1.2, requires that criticality data be taken while at 10E-8 amps in the intermediate range.
If, during a xenon-f ree reac tor startup at MOL, the operator " overshot" 10E-9 amps.and instead leveled off at 10E-7 amps, which one of the following statements is correc t?
a.) At 10E-7 amps, tnere-are few or no effects from nuclear h e a t '.
- However, because the reactor power is a decade. higher, the criticality data. rod-position will"be higher than it would be at'10E-8 amps, b.) At 10E-7 amps, the effects of. nuclear heat are partially offset by the effects of being at a higher reactor power; therefore, the criticality.
data rod position will be lower than it would be at.10E-8 amps.
c.) At 10E-7 amps, there are few or no effects from nuclear heat; therefore, the criticality data rod position will be the same as.it would be at 10E-8 amps.
d) At 10E-7 amps, there are' substantial effects from nuclear heat; therefore, the criticality data rod position.will be higher than it would be at 10E-8 amps.
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QUESTION 1.06 (1.00) 1 Which one of the following BEST describes the indications -that would be observed if a centrifugal pump were. started and operated with its discharge]
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valve shut as compared to with its discharge valve open. (Assume no recirculation flow.)
a) digher starting current and lower running amperage b)
Lower starting current and lower running amperage i
c)
Higher starting; current and higher running amperage d)
Lower starting current,and higher running amperage l
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3 1.
PRI.NCIPLES OF' NUCLEAR-POWER PLANT OPERATION.
'Page
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THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
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QUESTION 1.07 (1.00)
Which one of the following will cause the Axial Flux Difference to b'ecome l
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more positive (less negative)?
l a.) A power increase with power def ec t compensated for by dilution only.
b.) A power increase with power defect compensated for by. rod withdrawal,
- only, c.) The buildup of xenon in the top portion of the core.
d.) The burnup of xenon in the bottom portion of the core.
QUESTION 1.08 (1.50) l l
Two identical reactors are taken critical.
Reactor A'has a rod speed of 40J i
steps per minute.
Reactor B has a rod speed of 30 steps per. minute.
Assuming a continuous rod withdrawal in each case, answer A, B,
or THE SAME; to each of the f'ollowing questions.
.j a) Which reactor will achieve criticality first?
b) Which reactor will have the highest critical rod height?
I c) Which reactor will have the highest source range count at criticality?
QUESTION 1.09 (0.50)
State whether the decay heat load would be HIGHER,-LOWER, or REMAIN THE SAM the situation below.
A reactor has been operating at 100% ' rated thermal power for 12 months when a reactor trip occurs.
The decay heat load after the. reactor ~ trip has bran calculated to be 6% rated thermal power.
If the total rod worth inser.ed on the reactor trip had been 1000 pcm LESS, the. decay heat load -!
after the reactor trip would l
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 8 Tavg by 6 degrees F.
f-b.)
Turbine Tripped, Mode Control in Tavg Mode, Loop A cold leg RTD failedi HIGH, and the Steam Dump Control Reset switch is MAINTAINED in the
" Bypass" position.
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c.)
00% power, 7.5% ramp decrease in turbine load for 3 minutes, Tavg> Tref j by 7 F, Mode Control in Tavg Mode f
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CAT 000RY 3 CONTINUED ON NEXT'PAGE *****)
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3.
INSTRUMENTS AND CONTROLS Page :24.
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I QUESTION 3.07 (1.50) l Indicate whether the Overtemperature Delta-T trip setpoint will INCREASE,.
DECREASE, or REMAIN THE SAME for each'of the following situations.
a.)
A Power Range upper detector fails LOW
-j b.)
A hot leg RTD fails HIGM l
l c.)
A pressurizer level' transmitter fails LOW q
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l QUESTION 3.08 (1.50) l l
Answer TRUE or FALSE to each of the following statements concerning ESFAS. H i
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a.)
In order to generate a "P"
signal, 2 out of.3 HI containment pressure I OR 2 out of 3 HI-H1' containment-pressure. signals are required.
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b.)
The S/G delta-P "S" signal is. generated when 2 out of 3 S/G pressure I
transmitters for one S/G are 100 psi GREATER than_the other twos /G's.]
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c.)
A reactor trip in coincidence with a Low Tavg will result in Main Feed]
l Water control valves closure, but the Main Feed Pumps will continue to l run if running at the time of the trip.
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1 QUESTION 3.09 (1.00)
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l Answer TRUE or FALSE to each of the following:
'{
I a.)
With the Rod Control System in MANUAL and no rod motion in progress, the rod speed meter on the control board will indicate a rod speed of. j 0 spm.
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b.)
The BOU will ONLY count steps if the rod control bank selector switch is in AUTO or MANUAL.
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INS.TRUMENTS AND CONTROLS Pege 25.
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.1 QUESTION 3.10 (1.00)
I Answer TRUE or FALSE to each of the f ollowing statemen ts' concerning the AFWj system.
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a.)
The discharge valves may be completely shut after an autostart ONLY j
after placing BOTH-NORM / RESET handswitches in the RESET.' position.
(THIS PART DELETED)
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b.)
The air operated isolation valves that admit steam to the'TDAFWP.must q
l be LOCALLY operated if instrument air is lost.
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c.)
When'the CST. level decreases below the LO-LO level setpoint, the I
L Service Water Intake valves will automatically open to provide a.
l continuous water supply to the AFW pumps.
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1 QUESTION 3.11 (1.00)
Answer TRUE.or FALSE to each'of the f dllowing statements concerning the
-]
- Emergency Diesel Generators and Auxil.iaries.
4 a.)
The main starting air valves for the F-Ms can'be MANUALLY overridden.
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b.)
The standby circulating pump for the F-Ms continuously circulates oil' to prelube the engine.
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OUESTION 3.12 (2.00)
Complete the following statement regarding the RHR System.
k (NOTE:
Appropriate units should be included as part of the answer as j
necessary.)
a.)
The Loop
'C' Inlet isolation valve can be opened.when'RCS' pressure is. j (1) as sensed by pressure transmitter 403 and will autoclose if
- i RCS pressure increases to (2) as sensed'by pressure transmitter I
(3)
The Loop
'A' Inlet isolation valve can be opened when RCS
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pressure is
-(4) as sensed by pressure transmitter 403 and will j
autoclose if RCS pressure increases to (S) as sensed by pressure 3
transmitter (6)
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6.)
What is the purpose of this arrangement?
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3.
INSTRUMENTS AND' CONTROLS Page 26 1
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GUESTION 3.13 (1.25) q List five (5) valves that SHUT on a Phase
'B' Containment Isolation signal.l l
i GUESTION 3,14 (0.75) l State the three (3) protective features the operator can MANUALLY b1cck l
when nuclear power has been increased above P-10.
1 QUESTION 3.15
'(1.00) 1 With Unit 2 in Mode 1 and the pressurizer level control selec tor switch in the III/II position, an instrument failure causes the following sequence of1 events:
.j
.1.)
Charoing flow reduces to minimum 2.)
Pressurizer level decreases j
3.)
Le tdown secures and ALL pressurizer heaters trip off l
4.)
Pressurizer level increases until a high level trip occurs Assuming no operator action was taken,' state the instrument that failed AND the direction in which it failed.
GUESTION 3.16 (1.50)
State whether each of the following would cause a Rod Control System URGENT..
or NONURGENT alarm, a.)
Loss of Main 100 VDC Power in the Logic Cabinet b.)
Slave Cycler fails to start counting upon receiving "GO" pulses c.)
Loose circuit card in the Power Cabinet GUESTION 3.17 (1.00)
List the four (4) systems that receive input from the Auctioneered High l
l Tavg control circuit excluding indications and alarms.
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INSTRUMENTS AND CONTROLS Page 27 i
QUESTION 3.18 (1.25) l f
Assuming the appropriate control switch is in AUTO, list ALL the signals that automatically reposition / shut each of the following valves associated with the CVCS.
a.)
Orifice Isolation valves b.)
VCT Outlet Isolation valves OUESTION 3.19 (1.00)
The reactor has just been taken critical at 2500 cps.
Control rods are withdrawn 20 steps to increase reactor power.
No further operator action is taken and a reactor trip eventually occurs.
a.)
What caused the reactor trip?
b.)
If N-36 was UNDERCOMPENSATED when the control rod withdrawal occurred, would the reactor trip have occurred SOONER, LATER, or AT THE SAME TIME?
QUESTION 3.20 (2.00)
State all the AUTOMATIC actions, if any, associated with the following Process Radiation Monitoring System channels, a.)
Steam Generator Blowdown Discharge Monitor R-23D b.)
CCW Monitor R-17A c.)
Containment Purge Exhaust Monitor R-24A d.)
Plant Vent Gas Monitor R-14 OUESTION 3.21 (1.00)
During core alterations, the high power trip function on two (2) Power Range NI's are inadvertently being tested concurrently.
State any impact this testing would have on the remainder of the NI system.
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CATEGORY 3 CONTINUED ON NEXT PAGE *****)
1 3.,
INSTRUMENTS AND CONTROLS Page 28
.1 i
1 QUESTION 3.22 (1.25)
J a.)
State the consequence (s) that could be expected in the Rod Control i
System DC Hold Cabinet if two (2) or more GROUPS of rod drive d
mechanisms, except for control bank
'D',
were placed on HOLD.
1
'I b.)
State the two (2) different'DC voltages used in the DC Hold Cabinet j
and the usage of each.
1 1
QUESTION 3.23 (1.50)
{
i The plant is operating at 75% power with all primary. plant controls in j
AUTOMATIC when pressurizer pressure transmitter 444 fails HIGH.
Assuming l
no operator action is taken, describe the' response of'the pressurizer j
pressure control system until pressure is stable or a reactor trip'has 1
l occurred.
(INCLUDE ALL APPLICABLE SETPOINTS.)
l 1
1 i
1 l
OUESTION 3.24 (0.00)
I l
Explain why a Tcold failing HIGH would result in a greater rod speed signalj L
than a That failing HIGH. (THIS QUESTION DELETED) i l
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(*****
END OF. CATEGORY 3 *****)
.__a
4.
PROCEDUREG -~ NORMAL. ADNORMAL. EMERGENCY Pcge.29 ANb-RADIOLOGICAL CONTROL i
1 l
a GUESTION 4.01 (1.00)
Which one of the following correctly dese.ribes the basis for allowing,RCP
)
restart in FNP-1-FRP-C.1, Response to Inadequate Core Cooling.
j l
a.)
To mix the SI flow to protect the reactor vessel.from. cold water.-
)
i b.)
Once subcooling is established, restarting the RCPs helps to i
collapse voids that may have formed in the reactor vessel head.
c.)'
Allows for restoration of pressurizer pressure control using norcal spray valves.
j d.)
Provides for cooling of the core when secondary depressurization I
does not alleviate inadequate core. cooling.
I l
j i
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I QUESTION 4.02 (1.00) 1 l
Which one of the following would cause the GREATEST biological damage to an l individual?
]
l l
a.)
0.1 Rad of Fast neutron
(
l l
b.)
1 REM of Gamma f
j c.)
10 REM of Beta j
d.)
0.05 RAD of Alpha l
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Cf.7EGORY 4 CONTINUED ON NEXT PAGE *****)
v.
40 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page 30
' ANd RADIOLOGICAL CONTROL tl I
i l
QUESTION 4.03 (1.00)
Which one of the following statements is CORRECT 7
~
l a.) A radiological exclusion area is a high radiation area but a high f
radiation area is NOT necessarily a radiological; exclusion area.
{
b.) A contaminated area will be a radiation area and a. radiation area will be a contaminated area.
l c.). Radiological restricted areas are required to be surveyed routinely.
d.) All areas in the RCA are either' radiation areas or high radiation f
-f j
areas.
1
(
l
(
QUESTION 4.04 (1.00) f Pressurizer PORV leakage falls under which one of~th'e following~ Technical' l
Specification leakage classifications?
a.) IDENTIFIED LEAKAGE j
i b.) PRESSURE BOUNDARY LEAKAGE c.) CONTROLLED LEAKAGE i
l 1
d.) UNISOLABLE LEAKAGE l
1 1
i
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
Vt 4.
PROCEDURES - NORMAL, ADNpHMAL, EMEHGECGy.'
Page 31
\\
AND RADIOLOGICAL COM,TRCL l
l QUESTION 4.05
/t.50)
,1 Answer each of the following situations if te.nporary revisions to operating procedures SHOULD or CHOULD NOT be issued.utp* ding to'FNP-0-AP-16, Conductj of Operations - Operations Group-a.)
While pe: forming a valve line-up, the procedure requires that the pump be running prior fa opening the assoc.iated suction ibolation valve.
b.)
During a reac tor s t'Je tup, it is necessary for the accumualtors to q
remain isolated until;RCS pressure ex eeds 2100 ping to allow for-the completion of pos'-maintenance testing.
c.)
Prior to beginning a plant cocidown using RHR, it fs necessary to cross-connect both CCW trains with a temporary connection in order to perform of surveillance testing on the CCW. pumps.
j l
QUESTION 4.06 (1.50) j Answer TRUE or FALSE to each of the following precautions assoc.iated with FNP-1-SOP-21.0, Condensate and Feedwater System.
a.)
The AUTO start feature of the condensate pumps can only be used after three pumps are running.
b.)
One or more adjacent heaters may be removed from service with no operating restrictions.
j c.)
Resetting the first out panel of A SGFP may cause an inadvertent trip of A SGFP.
l l-l t
l l
i' l
l
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l I
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l.
j' 1
V 4.
PRQCEDURES - NORMAL.fhBN QMAL, EMEFfbidNCY Pcge 32 !
AND RADIOLOGICAL CONTRO g' p
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d
)
)
l OUESTION 4.07 (2.,001 4
d toFNP-UOP-1.2[,$tartupofUnitfromHotStandby actions t the power level at which they arej Match the followitg procedural to be per f ormed e.:ci:rding jj
)
to Minimum Load.
y f,M i
e j
ACTION PGWER LEVEL
(
i sy a.) P1 ace steam dumpy in Tavg /jode
/k.)f'%
feedwatbr,to synch speed
.Oc %
turbine b.) Roll main
.i to SGFP'r 3.. ) 15 %
c.) Transfer 4160 V hsses to Uf 4.).petween 20 % and 25 %
d.) T rar.s f er
/
5 3) 30 %
p
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,P t GUESTION 4.08 (2.00) T 1,
j l
s State the reason f or c at of the foil Og precautions given in FNP-1-UOP-1.1, Stattur T om Cold Shutd Wn to Hot Stjandby.
9 4
f
\\
d, 1
a.)
The reactor coolInt dumps must
['be operchqd wher? the number 1 wP l 3 fir VCT pressure is lessl j
differential pressure in iess tt/M i 200 'dg /,
.t than 18 psig.
- 'j U,
reachar nukejg water b.)
With no RCP's.if operation, pump al} be one racked out.
V' a
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ei I h/
che or more ob,the
)
9 c.)
A reactor coolant (ph p shall not be starteo RCS cold leg tem pe rei tu r e s less than or eqt( )to 10Funless,+hej pressurizer watert plume is less than 24'4 wi :e rapge cold pg*gssurtzer indicationgr'chesecondarywaterhemperatu'(eofeach%.ceam
)
level precautih( y.
generator is l ess, tr4 an 50 F above each 6,f)the RCS/ cold leg temperatures.
(There dre two (2) reast % for this i.
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b,'
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QUESTION 4.09 (1.00)
,j
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List six (6) indications, other than radiation Nr
.dring, that are
(
symptoms of excessi d RCS leakage according to hW 4-AOP-1.0, Excessive RCS Leakage.
(Per new rhvision of FNP-1-AOP-1.0, Excessive RCS Leakage, only' 3
two (2) indications are rSqudtd.)
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7 4 PROCEOURES'- NORMAL. ABNORMAL.~ EMERGENCY Pagm 33, AND5 RADIOLOGICAL CONTROL
.N.)
QUESTION 4.10 (1,20)
List FIVE (5) possible Main Control Board alarms that would: indicate'an inadvertent'digution while operating at 1 0 0'/. p o w e r.
(Setpoints are NOT q
required) 10 l
~ 3,,
QUESTION 4.11 (1.50) j
+-
FNP-1-ECP-0.0, Loss of.All AC Power, requires that the RCS-be checked"for isolation.
How is this accomplished per FNP-1-ECP-0.0?'
I if l
QUESTION 4.12 (4 25) a e
t Step S.3 of'FNP-1-UOP-1.2, Startup of Unit from Hot' Standby to Minimum:
I Load, is preceded by a NOTE stating that RCS Tavg shall be greaterithan:or-equal to 541 F prior to criticality.
State the five (5) bases for this.
te pera ture limitation.
'i i
f,, d QUESTION 4.13 (1.00)
/
l The Operator-At-The-Controls (OATC) is responsible for operation of'all.
]
- q
- g g j equipment in the At-The-Controls (ATC) area.
According.to FNP-0-AP-16,-
j Conduct of Operation - Operations Group, what is meant by the-term
'd
- m\\
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operation"?
lj f
i I
GUESTION 4.14 I (1.50) i State the definition of the following-terms per FNP-0-AP-14, Safety Clearance and Tagging:
a.). Nagging Official b.)
Designated Operator y <- ai
'd c.)
Clearance i1 7
?
r
(*****
CATEGORY 4 CONTINUED ON NEXT PAGE *****).
('
P i
i
g.
K 4.; PPO.CEDURES - NORMAL. ABNORMAL.: EMERGENCY Pagal34 AND RADIOLOGICAL CONTROL
(?'
4 i
'l 9
l
-GUESTION
'4/13 (1.00) s Unit 2 is in t1 ode 3 with a normal' reactor start-up is in progress when ALL:
s power to the DRPI. display is lost.
State what actions. must: be : taken according to Technical Specifications.
3 QUESTION 4.16 (1.00)
State the four'(4)-Emergency Classification categories per FNP-0-EIP-9L,-
Radiation Exposure Estimation ~and Classification of' Emergencies.
1 l
.1 QUESTION 4.17 (1.75) l Step 7 of FNP-1-EEP-1, Loss of Secondary or Reactor Coolant, has the I
operator check.if SI can be terminated.
State the SI termination 1 criteria assuming normal containment' conditions.
)
OUESTION 4.18 (1.72)
When commencing NCS cooldown per FNP-1-AOP-28.1, Fire in Cable Spreading l
l Room, tne operator is cautioned to use extreme' caution'when adjusting the l
S/G atmospher'ic relief valves when the S/G's are isolated.
Excepting.
thermal' considerations'(i.e.:
cooldown rate), what is the bases for this cautica.
l GUESTION 4.19 (0.00)
FNP-b-EEF -3 r Steam Generatet Tube Rupture, has the operator verify that the.
atmospheric relief valves are set at 1035 psig and in AUTO.
Explain why i
the atmospheric relief valves 1.ift.setpoint.is INCREASED to 1035 psig.
(THIS QUESTION DELETED)
QUESTION 4.20 (1.00)
According to FNP-0-A Conduct of Operation - Operations Group, when'and' under what condit1on,P-16, s cen an UNLICENSED individua11 manipulate the controls?:
7,
[ ('. i l r, l
1
(*****' CATEGORY 4 CONTINUED'ON NEXT'PAGE *****)
l l
2 l'
'i
4.
PROCE?URES' NORMAL'. ABNORMAL. EMERGENCY
.Page 3S ANO-RADIOLOGICAL CONTROL-1 i
' QUESTION 4.21 (1.00)
L During a reactor startup per FNP-UOP-1.2, Startup of Unit from Hot.St'andby
to Minimum Load, criticality was achieved at 110 steps on Bank C..
State what actions, if_any, must be taken and why'.
1 1
QUESTION 4.22 (1.00) l.
In GENERAL terms, describe the operator. actions that-are taken to remove
' failed power range channel N-44'from service'per FNP-1-AOP-10.0', Nuc' lear l
Instrumentation Malfunction, including any features that are bypassed or' l
defeated.
j l
QUESTION 4.23 (1.00)
State the two (2) Technical Specification limits on steam generator tube leakage.
i QUESTION 4.24' (1.00) 1 Step 2.1 of FNP-1-EEP-2, Faulted Steam Generator Isolation, requires that'a check be made if any S/G pressure is dropping in an uncontrolled manner.
Explain what is meant by the term " uncontrolled".
e
(*****
END OF CATEGORY 4 *****)
(**********
END OF EXAMINATION **********)
,i
1-a PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
-Page'36 THERMODYNAMICS. HEAT-TRANSFER AND FLUID FLOW
.f i
ANSWER 1.01 (1.00) b.)
REFERENCE l
FNP NUCLEAR ENERGY TRAINING MOD'ULE 3 - REACTOR OPERATIONS Unit 12.4,-Learning Objective #1 015000K506 192008K105 192008K194 192008K103-
..(KA's) i ANSWER 1.02 (1.00)
']
a.)
1 REFERENCE l
1 FNP Nuclear Energy Training Modul'e 4, Plant Performance j
193004K125 193003K125
..(KA's) 1 ANSWER 1-.03 (1.00) 1 b.)
REFERENCE Westinghouse Reactor Physics, pp. I-5.26 & 27 l
SHNPP RT-LP-1.10, p 13-15 License Retraining Learning Objective #17 192004K108
..(KA's)
ANSWER 1.04 (1.00) c.) or a.)
(1.0) f REFERENCE NUS, Nuclear Energy Training, Module 3, Reactor Operation
. Unit 12.1 Learning Obj ective #1 192003K102 192002K111 192002K114 192002K110
..(KA's)
(*****
CATEGORY 1 CONTINUED ON NEXT PAGE-*****)
1, PRINCIPLES OF' NUCLEAR POWER PLANT' OPERATION.
, Page3'7aj THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW l.)
l 1
jl
. ANSWER 1.05 (1 00) c.)-(1.0) j 1
REFERENCE N
t
-l NUS, Nuclear Energy Training, Module 3 - Resctor: Operation j
Unit 13.5, Learning Objective #2 1
19200BK115 192008K114 192008K113
..(KA's) l i
ANSWER 1.06 (1.00)
I b) (1.0)
J i
REFERENCE NUS, Nuclear Energy Training, Module 4, Plant Perf ormance j
License Retraining Learning Objective #17 2
191005K105 191005K104
..(KA's) j AflSWER 1.07 (1.00) b.) (1.0)
REFERENCE 1
)
NUS, Nuclear Energy Training,. Modules 2 and 3, Reactor Operation 001000K506 192005K110
..(KA's) 1 ANSWER 1.08 (1.50) a) A (0.5) b) THE SAME (0.5) c) 9 (0.5)
REFERENCE NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation 192008K104 192008K103 192003K101
..(KA's)
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CATEGORY 1 CONTINUED.ON NEXT PAGE *****)-
m_
____.mmaf'm.a a.m_im.__ _ _.. _ _ _
m
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- 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 30-!
-TH5RMODYNAMICS. HEAT TRANSFER AND FLUID FLOW-I
-ANSWER l'.09 (0.50)
REMAIN THE SAME (0.50)
{
l REFERENCE
.l FNP Nuclear. Energy' Training Module 3 - REACTOR OPERATIONS Unit 14.3, Learning Objective #1
-{
19200SK127
..(KA's)
ANSWER 1.10 (2.00)
'l
]g a.)
LESS THAN b.)
LESS THAN c.)
GREATER THAN I
d.)
THE SAME AS REFERENCE
~
FNP Nuclear Energy Training Module 3 - REACTOR. OPERATIONS 0010104207 001010KS26 192006K110 192006K107
..(KA's) i
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1 ANSWER 1.11 (1.50)
.I a) INCREASES (0.5) l b) INCREASES (0,5) j c) INCREASES (O.5) j REFERENCE
)
NUS, Nuclear Energy Training, Module 4, Plant Performance Unit 6.5, Learning Objective #1,2 004000K604 191004K106 191004K115
..(KA's) 1 ANSWER 1.12 (2.00) a) DECREASE (0.5) o) INCREASE (0.5) c) INCREASE (0.S) d) INCREASE (0.S) i l.
i
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CATEGORY 1 CONTINUED ON NdXT PAGE *****)'
l I
-PRINCIPLES OF NUCLEAR POWER. PLANT OPERATION.
.Pago 391' 1.'
- TH$RMODYNAMICS, HEAT TRANSFER'AND FLUID FLOW REFERENCE NUS, ~ Nuclear Energy Training,-Module 3 - Reactorf0peration
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..(KA's)
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ANSWER' 1.13-(2.50) a) INCREASES (0.S) b) DECREASES (0.5) c) INCREASES ( 0. S ) -
d)~ INCREASES (0,5) e) DECREASES (0.S) j i
REFERENCE l
NUS, Nuclear Energy Training, Module 4 - Reactor Operation Unit 8.2, Learning Objective #1-193000K105
..(KA's) l ANSWER 1.14 (1.00) l i
a) MORE NEGATIVE (0.5)
{
l b) MORE NEGATIVE (0.S) 1 REFERENCE j
l l
NUS, Nuclear Energy Training, Module 3 - Reactor Operation Unit 9.3, Learning Objective #1 Unit 9.2, Learning Objective #1 l
4 192004K106 192004K103
..(KA's)
ANSWER 1.15 (1.50) a) INCREASES (0.5) i b) INCREASES (0,5) c) DECREASES (0.5) d) REMAINS THE SAME (0.00)
(THIS PART DELETED) 1 i
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1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 40
'. THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i
1 REFERENCE NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation
- 4
.Licanse Retraining Learning ObjectiveL#8 002020KS08
..(KA's) 4 i
ANSWER
'1.16 (2.00)
]
.I a.
FALSE (0.S) b.
TRUE (0 S)
- c. FALSE (0.S)
.j d.
FALSE (0.S)
)
REFERENCE.
l 1
i l
NUS, Nuclear. Energy Training, Module 3, Reactor Operation
.j l
Unit 13.6 Learning Objective.#1 j
l Unit 10.2 Learning Objective #1,2
]
l
-192006K113 192006K107 192006K106 192006K104 192006K103 l
..(KA's)
I ANSWER 1.17 (1.00)
)
a.)
TRUE (0.S0)
]
b.)
FALSE (0.50)
REFERENCE NUS, Nuclear Energy Training, Modules 2 and'3, Reactor Operation FNP TS Bases 3/4.2.2 REFERENCE l
193008K103 193009K105
..(KA's)
ANSWER 1.18 (1.50) a) FALSE (0.S) b) FALSE (0,5)
.c) FALSE (0.S)
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i 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
Page 41.i
~
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW REFERENCE I
i NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation j
l 193007K108 193007K106 015000A101 015000K504
..(KA's)-
q i
i ANSWER 1.19 (1.25) 1
)
a.)
ALL of the following at (0.25) each 1.) FAILS HIGH 2.) FAILS LOW 3.) FAILS LOW 4.) DOES NOT CHANGE b.) T/C (0.25)
'i REFERENCE l
J Oconee OP-OC-IC-RCI pg 16 l
191002K114 191002K113
..(KA's) l l
'l
~
ANSWER 1.20 (1.75)
Thera is a heated and an unheated junction'(0.25) thatgenerates.adelta-Tfj signal (to the RVLIS) (0.25).
When water surrounds (0.25) the HJTC's, the i delta-T between the heated and unheated junctions is small (0.25).
When i
the HJTC is uncovered (0.25), the heated junction temperature begins to l
rise (0.25) due to poor (lower) heat transfer by steam (0.25).
i
-i REFERENCE l
l FNP ICCMS OPS-52202E pg 5,6 License Retraining Learning Objective #1 l
002000K107
..(KA's) i l
ANSWER 1.21 (1.50)
At EOL Beta-effective is approximately equal to 0.0057. (0.50)
By the period equation,. addition of 0.0057 dk/k reactivity causes the i
delayed neutron contribution to become 0 (0.50) leaving only the prompt l
neutron function controlling reactor period. (0.50) l l
l l
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4
j' 1
' PRINCIPLES OF~ NUCLEAR' POWER PLANT OPERATION.-
Pege-42.'!
l TH5RMODYNAMICS. HEAT TRANSFER AND FLUID FLOW 4,
REFERENCE Basic Reactor Theory 192003K108
..(KA's)
ANSWER 1.22 (1.00) b) (1.0)
I REFERENCE j
[
NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation-1 002000A105 002000KS10
..(KA's) a i
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i l
f' 1
i a
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END OF CATEGORY 1 *****)
f
1 2,
PLANT' DESIGN INCLUDING SAFETY AND EMERGENCY:
Page-43' SYSTEMS i
.b j
ANSWER 2.01 (1.00) i b.)
1 REFERENCE l
010000K403
..(KA's)
,jj ANSWER 2.02 (1.00) k a.)
l REFERENCE l
FNP ECCS PG S Learning Objective #S 006000K302
..(KA's) q l
ANSWER 2.03 (1.00) l l
c.)
i REFERENCE l
..(KA's) l t
l ANSWER 2.04 (1.00) l b.)
l l
REFERENCE 1
l FNP Main and Reheat Steam pg 8 1
035010K602
..(KA's) l l
l l
I i
ANSWER 2.0S (1.00) l l
a.)
l l
l
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l h___.
i
~ 2, PLANT DESIGN INCLUDING SAFETY AND' EMERGENCY
'Page 44
~
SY$TEMS-
. REFERENCE FNP.SGWLCS pg 18 Condensate and Feedwater License Retraining Learning Objec tive. #6 -
I OS9000K419
..(KA's)
)
l 1
l ANSWER 2.06 (1.00) b.)
i REFERENCE l
FNP CVCS License Retraining Learning Objective #1.
004000K602'
..(KA's) l l
l I
ANSWER 2.07 (1.00) c.)
l l
REFERENCE l
FNP CS and Containment Cooling pg 9 026000K404 026000K402 026000K401'
..(KA's) 1 1
l I
l ANSWER 2.08 (1.00) c.)
REFERENCE l
l 061000K602
..(KA's) l l
,i ANSWER 2.09 (1.00) j u
a.)
')
REFERENCE
-)
..(KA's)
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- )
___O
20 PLANT DE' SIGN INCLUDING 1 SAFETY AND-EMERGENCY Page 45 SYSTEMS ANSWER 2.10
. ( 1. 00 ) -
H b.)
1 l
]
REFERENCE i
FNP ECCS License Retraining Learning Objective #3 006000K506
..(KA's)
)
ANSWER 2.11 (1.00)
]
l 1
l d.)
]
A REFERENCE j
l l
)
l 064000K302 064000K301
..(KA's) 1 l
1 l
)
l ANSWER 2.12 (1.00) l l
l b.)
q 1
REFERENCE FNP RHR II L 3G 17 1
003000K410
..(KA's)
.-j i
l ANSWER
.2.13 (t.S0) a.
DOTH UNITS (0.50) b.) UNIT 2 ONLY (0.50) l c.) UNIT 1 ONLY (0.50)
.j
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REFERENCE l
FNP SGWLC X B PG 7, 13 FNP UI AND UII DIFFERENCES OPS-52108H l
035010K401
..(KA's) j
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2 PLANT-DESIGN INCLUDING SAFETY AND EMERGENCY Page 46.
SY$TEMS l
i 1
ANSWER 2.14 (1.00) j a.) FALSE
( 0. 5, > )
b.) FALSE (0.5L,
. REFERENCE FNP RCS-1 003000K602
..(KA's) i I
l ANSWER 2.15 (1.50) a.) FALSE (0.'50) b.) FALSE (0.50) c.) FALSE (0.50) i REFERENCE FNP REACTOR MAKE-UP II H Learning Objective #7 004000K404
..(KA's) l l
l ANSWER 2.16 (1.00) a.)
FALSE (0.50) b.)
FALSE (0.50)
_ REFERENCE FNP GFFD PG 19 Learning Objective #5 ANSWER 2.17 (1.50) a.) FALSE (0.50) b.) FALSE (0.50)-
c.) TRUE (0.50)
REFERENCE FNP TURBINE AND AUX VI B PG 28-37 l
045010K423
..(K4's)
.i
(*****
CATEGORY 2-CONTINUED-ON:NEXT PAGE *****)
2 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 47-SY$TEMS ANSWER 2.10 (1.50) a.) 3 (0.50) h
]
b.).1 (0.50) c.) 2 (0.50)
REFERENCE j
l FNP CVCS II G PG 8,9 004000K401
..(KA's)
I i
ANSWER 2.19 (1.00)
I ALL of the following at (0.25) eacht a.)
Auto closure.cf LTDN (and orifice) isolation valves, j
b.)
LTDN. delay tanks (pipes) c.)
Spraying of LTDN flow into VCT vapor space.
d.)
Excess LTDN divert valve REFERENCE I
FNP CVCS TI-G Learning Objective #8 004000K407 004000K404
'004000K401
..(KA's) l 4
ANSWER 2.20 (1.75)
Any SD bank (0.10) rod < 211 steps (0.05)
Any CBA (0.10) rod < 6 steps (0.05)
Any CBB (0.10) rod < 6 steps (0.05) and (0.033) any CBB (0.10) rod >/= 12 steps (0.05) or CBC (0.10)' rod >/= 6 steps (0.05) or-CBD (0.10) rod >/= 6 steps (0.05)
Any CBC (0.10) rod < 6 steps (0.05) and (0.033) any Ci m
')
rod >/= 12 steps (0.05) or CBD (0.10) rod >/= 6 steps (0.05)
Any CBD (0.10) rod < 6 steps (0.05) and (0.033) any CBD (0.20) rod >/= 12 steps (0.05) l
(*****
CATEGORY 2 CONTINUED ON NEXT PAGE *****)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY Page 48 SY$TEMS 1
d I
REFERENCE i
i FNP DRPI XF PG 13 Learning Objective #5 l
014000K406
..(KA's)
'i 1
ANSWER
'2.21 (i.00) 1
]
ALL'of the following:
1.)
Startup of an idle RCP (0.25)'with secondary water-temperature of l
the associated S/G less than or equal to 50 F above RCS cold leg j
temperatures (0.25).
2.)
The start of three (3) charging pumps (0.25)'and their injection into a water' solid RCS (0.25).
.]
l REFERENCE l
l l
FNP RHR II L pg 0,9,23 005000K407 005000K401
..(KA's) l l
ANSWER 2.22 (1.00) l (0.50)-
I L
All non-AFW system equipment (that use the CST as a water source)
I have their respective suctions above a certain elevation (standpipes)(0.50;l (that ensures a minimum water volume available to the AFW system).
I REFERENCE l
FNP AUX FEEDWATER III I PG 8 l
061000K401
..(KA's)
ANSWER 2.23 (0.75)
The gamma flux contribution (0.25) in the power range is insignificant l
compared to the neutron' flux contribution'(0.25) and is also proportional to power (0.25)
REFERENCE FNP EXCORE NUCLEAR INSTRUMENT XC PG 7 Learning Objective' #4 015000K502
..(KA's)
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CATEGORY 2 CONTINUED ON NEXT PAGE *****)
- 12. ' PLANT DESIGN INCLUDING SAFETY AND-EMERGENCY Pagn 49 )
SY$TEMS
]
l ANSWER
'2.24 (2.50) a.) ALL four (4) of the following at--(0.25) each:.
j 1.)
Open.or shorted coil or cable in either Data A or Data B.
j 2.)
Loose or removed detector / encoder card in Data-A or Data B.
I 3.)
Loss of one power supply
(+/-
15-VDC.or 6 VAC)'in a-Data A or.
i Data B cabinet.
4.)
Loss of one power supply in the display console.
b.) With the DRPI system'in' half-accuracy (0.50), the. rods;will be no more than +/- 10 steps from indicated position (0.50).
The, Technical Specification requires that tim indicated position to be no more than
+/- 12 steps from demanded..(0.50)
I REFERENCE FNP DRPI pg 13,14 Learning Objective #6 l
014000K406 014000A202 014000A102
..(KA's) l ANSWER 2.25 (1.00) l Countrate would rapidly decrease (0.50) (due to phase separation) then
.i would increase (0.50) (as the downcomer would empty).
REFERENCE FNP Mitigating Core Damage pg 17-24 015020A202
..(KA's)
I ANSWER 2.26 (1.00) l Ensures flow to intact S/G's (0.50)
Prevents (limits) containment overpressure OR limits flow to a' faulted'S/G' i
(0.50) 1 REFERENCE FNP AFW III I pg 5 l
l l
(*****
END OF CATEGORY 2 *****)
1
-' ll 3.
INSTRUMENTS AND CONTROLS Pdga SO i i
l i
li s
ANSWER 3.01
.(1.00)
)
l c.)
I REFERENCE
-l FNF RPS X I 012000K610
..(KA's) i 1
ANSWER 3.02 (1.00) l c.)
l o
. j REFERENCE 1
FNP EXCORE NI X C License Retraiteing Learning Objective #5,7 l
015000K401
..-(KA's) l l
s l
ANSWER 3.03 (1.00) l j
d.)
REFERENCE FNP Tavg, Delta-T, and Pimp X G 001000A103 001000A101
..(KA's)
ANSWER 3.04 (1.00) c.)
REFERENCE FNP RPS X I 012000K402 012000A306
..(KA's)
ANSWER 3.05 (1.00) b.)
i
(*****
CATEGORY 3 CONTINUED ON NEXT PAGE *****)
l l-
4 3.
INSTRUMENTS AND CONTROLS
'Pagn 51 l
REFERENCE l
FNP ROD CONTROL SYSTEM X E Learning Objective #10
- j 001000K408
..(KA's)
]
1 I
ANSWER 3.06 (1.S0) 1 1
a.)
ARM ONLY (0.50) b.)
ARM AND ACTUATE (0.50) c.)
HAVE NO EFFECT (0.50)
REFERENCE 1
FNP Steam Dump pg 23-20 Learning Objective #6 l
041020K418
'041020K414
-..(KA's) 1 ANSWER 3.07 (1.50) a.)
DECREASE (0.50) b.)
DECREASE (0.50)
I c.)
REMAIN THE SAME (0.50)
REFERENCE FNP RPS X I 012000K402
..(AA's) i i
ANSWER 3.00 (1.50) j a.) FALSE (0.50)
I b.) FALSE (0.50) c.) TRUE (0.50)
REFERENCE FNP ECCS III C FNP RPS X I License Retraining Learning Objec tive #7 006000K102
..(KA's) l ANSWER 3.09 (1.00)
.1 a.) FALSE (0.50) 6.) TRUE (0.50) l
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CATEGORY 3 CONTINUED ON NEXT PAGE'*****)
~
-!l 3,
INSTRUMENTS'AND CONTROLS
.Pago 52
-REFERENCE
- l FNP Rod Control System'X E Learning Ob.fective #4, #6
.f 001SWG04
..(KA's)
I 1
ANSWER 3.10 (1.00)
])
a.) FALSE (THIS PART DELETED)-
b.) FALSE
.l c.) FALSE
{
REFERENCE FNP AFW PG 23,12,05 061000K404 061000A202
..(KA's)
Il ANSWER 3.11 (1.00)
{
i a.)
FALSE (0.50) b.)
FALSE (0.50)
REFERENCE FNP EDG and Au.t 064000A101 064000K105
..(KA's)
ANSWER 3.12 (2.00) a.)
ALL of the following:
1.) </=402.5 psig (0.20); (0.05) j 2.) 700 psig (0.20); (0.05)
-]
3.) 402 (0.25) 4.) </= 402.5 psig (0.20); (0.0S) 1 5.) 700 psig (0.20); (0.05) j 6.) 403 (0.25) b.)
Prevents complete loss of overpressure protection.(0.25).due to instrument failure. (0.25) i REFERENCE-FNP RHR II L Learning Objective #7 1
005000K407
..(KA's)
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Li_____.__._.
i
3, INSTRUMENTS AND CONTROLS Pege 53 i ANSWER 3.13 (1.25)
Any five (S) of the following at (0.25) each 1.)
CCW from RCP Thermal Barrier Heat exchanger High Pressure. Isolation q
valve (3184) i 2.)
CCW from RCP Thermal Barrier Heat exchanger High Flow Isolation valve (3045)
)
3.)
RCP Oil Cooler CCW Return Isolation Valve (3046) 4.)
RCP Oil Cooler CCW Return Isolation Valve (3182) i S.)
RCP CCW Supply Valve (3052) 6.)
Instrument air to containment Isolation valve (3611) i REFERENCE
..(KA's)
ANSWER 3.14 (0.75)
ALL three (3) of the following at (0.25) each:
1.)
IRM High Flux Reactor Trip 2.)
IRM High Flux Rod Stop j
3.)
PRM High Flux Reactor Trip - Low Setpoint REFERENCE i
..(KA's)
ANSWER 3.15 (1.00)
Pressurizer Level channel III (0.50)
High (0.50)
REFERENCE FNP Pressuri.ter Level and Pressure Control X H pg 22,23 Learning Objec tive #6 011000A210
..(KA's)
(*****
CATEGORY 3 CONTINUED ON NEXT PAGE *****)
~
R
.{
I 3_,
INSTRUMENTS AND CONTROLS PagucS4 1
~
l j
l 1
ANSWER 3.16 (1.50) 1y a.)
NONURGENT (G.50) b.)
URGENT (0.S0) 4 c.)
URGENT (0.50)
REFERENCE j
]1 FNP Rod Control System X E pg 27-31 Learning Obj ec tive #15,16 001010K605
..(KA's) 1 i
' d 1
i ANSWER 3.17 (1.00) j l
l ALL of the following at (0.25) each:
l 1.)
Rod control 2.)
Pressurizer l evel ' cor:tro l j
l 3.)
Steam dump control q
j 4.)
RIL computer 4
I i
l REFERENCE l
4 i
j FNP TAVG, DELTA-T, AND Pimp X G
]
l 016000K101
..(KA's) i l
I l
ANSWER 3.18 (1.25) 1 i
l ALL of the following at (0.25) each:
l l
a.)
Letdown Line Containment Isolation valves not fully open 1
LO Pressurizer-level
(<
15%)
l Containment Isolation Phase
'A' signal present
('T' signal) b.)
LO-LO VCT level
(<S%)
Safety Injection signal
('S' signal)
REFERENCE FNP CVCS II G Learning Objective #8 004010K403 004000K403
..(KA's)
ANSWER 3.19 (1.00) a.)
SR High Flux.
(0.50) b.)
AT THE SAME TIME (0.50) i
(*****
CATEGORY 3 CONTINUED ON NEXT'PAGE,*v***)
1 l
-_-__.---.w_-~.a
3.,
INSTRUMENTS AND CONTROLS Page SS REFERENCE FNP EXCORE NI X C License Retraining Learning Objec tive #3,5 015000K405 015000K101 001000K105
..(KA's)
ANSWER 3.20 (2.00) a.)
Isolates discharge valve to river (if opened)
(0.50) b.)
Isolates CCW surge tank vent (0.50) c.)
Isolates the Containment purge and exhaust lines (0.50) d.)
Isolates waste gas release valve (0.50)
REFERENCE FNP Radiation Monitoring System OPS-52106D (OLT)
Learning Objective #4 073000K401
..(KA's)
ANSWER 3.21 (1.00)
Would cause both Source Range (0.50) detectors (N-31 and N-32) to deenergize (0.50).
REFERENCE FNP Excore NI RPS License Retraining Learning Objective #9 01SSWG15 01SSWGOS 015000K401
..(KA's)
ANSWER 3.22 (1.25) a.)
overload (overheat) the cabinet. (0.25)
(also accept dropping of control rods) b.)
125 VDC (0.25) - Latching rods (0.25) 70 VDC (0.25) - Holding rods (0.25)
REFERENCE FNP Rod Control System XE 001050G01
..(KA's) i l
1
(*****
CATEGORY 3 CONTINUED ON NEXT PAGE *****)
)
l
y
.q 3.,, INSTRUMENTS AND' CONTROLS'
'!Page-56 q l
H L.
l ANSWER 3.23 (1.50) ld 1
l Spray valves (444C and 444D) open (0.25)-
PORV 444D opens (0.25)!
pressure decreases (0.25)'
p PORV.444D shuts at 2000 psig.
( 0. 2 5 ) ',- ( 0. 1 2 5 ) -
l reactor trip occurs at'1865 psig (0.25), (0.125)-
I REFERENCE j
l l
FNP.PRESSURI ZER PRESSURE AND LEVEL CON 'ROL X'M l
~
l 010000A202 010000K603 010000K402 010000K101
..(KA's) l l
i l
b I
l ANSWER 3.24 (0.00) j 1
'I A Tcold failed high would produce an Auctioneered High Tavg equal.to Thot. j l
(0.50)
A That failed high would.cause only a small increase'in
'i Auctioneered High_Tavg.(0.50)
The temperature error (generated by the Rod I l
Control System) would be larger for a Teold failure high than for a Thot i
i failure high. (0.S0)
(THIS QUESTION DELETED)
REFERENCE f
FNP ROD CONTROL SYSTEM X E Learning Objective #10' 00iOOOK403
..(KA's)
^
'[
1 i
i
]
a q
i.i l
l
)
i
(*****
END OF CATEGORY 3 *****)
~
)
1
i 4.
PROCEDURES - NORMAL. ABNORMAL,,,,gMERGENCY Page S7-l
'AND RADIOLOGICAL CONTROL I
ANSWER 4.01 (l'.00) 4 d.)
REFERENCE Westinghouse background information RCP Trip / Restart Generic Issue 000074K307
..(KA's) l ANSWER 4.02 (1.00) c.)
REFERENCE l
10CFR20.5 194001K103
..(KA's)
ANSWER 4.03 (1.00) a.)
REFERENCE FNP-0-M-001 pg 10 194001K103
..(KA's)
ANSWER 4.04 (1.00) a.)
REFERENCE FNP TS DEFINITION 1.14 002000K40S
..(KA's) 1 I
l
\\
l
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
l l.
L __
p f
'4-PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page.50 i
'AND RADIOLOGICAL-CONTROL j
i l
ANSWER 4.05 (1.50) a.)
COULD j
b.)
COULD NOT c.)
COULD NOT j
REFERENCE
.l 1
FNP-0-AP-16 pg 47
{
194001A101
..(KA's) i 4
4 ANSWER 4.06 (1.50)
]
)
a.)
FALSE (0.50) b.)
FALSE (0.50) l c.)
TRUE (0.50) i l
REFERENCE FNP-1-SOP-21.0 I
056020G01
..(KA's) i ANSWER 4.07 (1.00) a.)
3.
(0.25) b.)
2.
(0.25) c.)
1.
(0.25) d.)
4.
(0.25) i REFERENCE l
FNP-UDP-1.2 194001A102
..(KA's)
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3 l
4e PROCEDURES - NORMAL. ABNORMAL,' EMERGENCY Pege 59 "f
'AND RADIOLOGICAL CONTROL l
j i
1 ANSWER 4.00 (2.00) 3 a.)
Ensure adequate seal flow (0.50) 6.)
Prevent an inadvertent' dilution While on RHR (0.50)
J I
c.)
Ensure sufficient surge volume for any RCS thermal expansion f
(RCS pressure increase)
(0.50)
Ensure that any RCS thermal expansion (RCS pressure increase) is
'{
within RHR relief capacity.
(0.50) 1 l
REFERENCE FNP-1-UOP-1.1 RCP License Retraining Learning Objective #5
- ]
003000G13
..(KA's)
]
l
^
?
l l
ANSWER 4.09 (1.00) l l
i l
ANY six (6) of the following at (0.25) each:
q 1.)
Increase in charging flow requirements 2.)
Decreasing VCT level
'l 1
REFERENCE
-l FNP-1-AOP-1.0 000028A106
..(KA's)
I i
ANSWER 4.10 (1.25)
Any FIVE (5) at (0.25) each I
1.)
Tavg/ Tref deviation alarm 2.)
Overpower Rod Stop g
3.)
RCS High Delta-T l
4.)
OPdT Rod Stop l
S.)
RCS High/ Low Trivg 6.)'
RIL low limit alarm 7.)
RIL lo-lo limit alarm 8.)
OTdT Rod Stop Other alarms that a cognizant operator would realistically expect in diagnosing this condition j
I
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CATEGORY 4 CONTINUED ON NEXT PAGE *****)
1 j
a
4.
PROCEDURES - NORMAL 'ABN RMAL. EMERGENCY r
40
?
q,Y h h
( e[!
'AND RADIOLOGICAL COffTROL p
3
\\'
i <.
, )
/,
'rk' 5
(
REFERENCE l
s/ 1
\\
,f' i
');
'\\>
\\
f,
,(. H FNP-1-AOP-27.0 n
I
),,
l FNP-1-ARP-1.6
/
004020G08
..(KA's) j
[
s,g
-(
q i
T a
t,r
(
'g g s; s
i
\\
/pc'
,l, ANSWER 1.11 (1.50) j
,\\
\\h ALL'of the following at (0.50) each:
I&
GJ.S pressure d.' 2[>15, p<41g l
1.)
Pressurizer PORV's closed when vakes s.h6t"\\ 3 /
2.)
Letdown'line OR orifice isolation q
shutI
'/
3.)
Excess letdown line isolation valves q
j REFERENCE s
j
\\
FNP-1-ECP-0.0 pg 3/28 License Retraini.g'L ning Objectivl 49
/
000056G11
..(KA's)
/
-O l
t
?
\\\\
I 1
ANSWER 4.12 (1.25)
,l
.]
\\
i ALL of the following at (0.25) each:
j
'1.) 'The MTC is within its analyzed tempera'j are range-(\\
)' '
2.) The protective instrumentation is within its normal operating rar41e 3.) The P-12 interlock is above its setpoint
/ N
'h f
4.) The pressurizer is capable of being in gr', OPERADLE s ta t.lswitt 'a steam J bubble
)
(
5.) The reactor vessel is above its minimum RT-NDT temperature l',
j d
REFERENCE
\\
J4 g\\
q FNP TS Bases for 3/4.1.1.4 pg B 3/4 1-2
{
1,1vj License Retraining Learning Objective #9 N
)
001050G05
..(KA's)
[
g
]
l Lhs j
t i
ANS'4ER 4.13
( 1. 23 0 )
7{
Ns
'l l
y
(,
ALL of the following at (0.25) eacht
(
r
)
1.)
Actual controls manipulation.
(
2.)
Monitoring equipment and systems parametersd
/ k, >' ;,
k,g\\ '
evolutions and significant eyapt.s relhter) j to tho3e
\\ -
3.)
Documentation of systems end/or componen ts controlled from thr"ATy, i
4.)
Initiating or performance of operator actions required by pla'n't
[I Emergency Operating Procedures, Abnormal Operating Procedures, and Annunica tor Response Procedures applicable to the ATC p,
'j 7
s o;
(*****
CATEGORY 4 CONTINUED ON NEXT PAGE *****)
5 y(?
l
.1s l
1
'/,
N r.
a.
11
n g
I 4.
' PROCEDURES - NORMAL,-ABNORMAL, EMERGENCY PageL61 '
4
'AND RADIOLOGICAL CONTROL
')
)
lI REFERENCE j
FNP-0-AP-16 pg 20-001050G01
..(KA's) f L.
I ANSWER 4.'14 (1.50) t taggingprocedureano{bh!thePlant or any individual authorized l
- a. )?
SS,.SF, PO, (0.05 each) i Manager (0.05) to implement the with whom the final responsibility of insuring the adequacy of Tagging. Operations
(
(
Orders rests.
(0.30),,
\\
3',
'.s u
i.
b.)$_ P0, 50,. ELECTRICIAN, CHEM /HP TECel, ASSIST C&HP TECH,,I&C TECH '(0.05 i
i 0.05) to each), or any. individual authoriead by the ' Plant Mar.ager, (\\
l '
execute Tagging Operations Orders.
(0.15)
,,s
- i. t c. 'i The formal ( theorization (0.25) to an individual or a classification
\\
or pers?ns worki g under-his supervision which provi the individual portion erf a systp{m that he is working on.l assurance th t the system or has been isolated to prevent personrei ' injury or equipmen t damage -
(0.25).
l.
i i
RL{goENCE
.x j,
FRP-0-AP-14 pg 2,3 Learning Objective f44 1
IA09001K102
..(KA's)
/
i s
/
ANSWER
'4.15 (1.00)
/
, )%(3.j (s
Coen'RTB's immediately(
c
, (
~
'gi i
REFERE'FCE g
)
/
ni.
- gl4000K301
..(KA's)
(
{,
Ai s
s ANSWER
/#.16 (1.'00)
/s s,
l
,ALL of the following at (3.25) each:
)
g.p 1.-)
No ti f ica tion ' o fy Un usua l Event (or NOUE)
- i l
,f' 2.b Alert 1
3.)
Site Area Emergenct 4,)
General Emergency E.
t
\\.
I I
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CATEGORY 4 CENTINUED ON NEXT PAGE *****)
t i
6 4
'I O
l j.
m
7, 77 ' (f
'it a
';v 4.
PRQjEDURSS - NORMAL'. ABNORMAL, EMERGENCY Page 62 I l,
'ANhfRADIO'OGICAL CONTROL w
(ptr L
i j
4e U
REFERENCE a
t l
I FNP-0-ETF-9 Learning Objective #16 i
194001A116
..(KH's) l OV ANSWER 4.17 (1.75) l y
l Subcooling (0.25) > 28 F ( B'.10 )
Total Aux feed flow (0.25) > 377 gpm (0.10) or Any one S/G narrow range.
level (0.25) > 6% (0.10)
RCS pressure (0.25) stable or increasing (0.10)
Pressurizer level (0.25) > 7% (0.10)
REFERENCE' l
Ff 4P-1-EEP-1 194001A102
..(KA's)
ANSWER 4.18 (1.00)
The steam pressura differential SI cannot be blocked. (0.50)
Unequal (excessive) steaming on one S/G could cause an unnecessary SI. (0.50)
REFERENCE FNP-1-AOP-28.1 000017G07
..(KA's)
ANSWER 4.19 (0.00) l l
Less than the lift wetpoint ter the first S/G safety (0.S0) to minimize challenges to the S/G safties (0.50).
l Greater that no-load pressure (E'.50) to minimize. rad release from the i
faulted S/G (0.50).
l (THIS QUC? TION DELETED) l l
REFERENCE 1
l Westinghouse L'S Bac'eground for SGTR:
l 000038G12
..:NA'r) 1 1
1 t
.(*****
CATEGORY 4 CONTINUED ON NEXT PAGE *****)
(b ' PROCEDURES NORMAL. ABNORMAL. EMERGENCY
-Pcge 63 l
'ANb RADIOLOGICAL CONTROL ANSWER 4.20 (1.00)
When is is part of training to qualify for an operators license (0.50)
~Only under the direction and in the presence of a licensed RO or SRO (0.50)
REFERENCE FNP-0-AP-16 pg 38 001050G01
.. ( Ki4 ' s )
ANSWER 4.21 (1.00)
Emergency borate and fully. insert all control-bank rods. [0.50]
Criticality was achieved below the O power RIL. [0.50]
REFERENCE FNP-UOP-1.2 pg 8 Learning Objec tive #5, #10 000023K301
..(KA's)
ANSWER 4.22 (1.00)
ALL of the following at [0.25] each;-
1.) Trip all bistables associated with the failed channel--(or remove control power fuses) 2.) Bypass the rod stop for the failed channel 3.) Defeat the upper and lower comparator input-4.) Defeat the comparator input 5.) Defeat the input to the rod control system (THIS PART DELETED PER NEW REVISION)
REFERENCE FNP-1-AOP-18.0 015000K604
..(KA's)
ANSWER 4.23 (1.00) 1.) 1 gpm total tube leakage for all generators
[0.50].
2.) 500 gpd per S/G
[0.503
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f l
3 4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page'64
- 'AND RADIOLOGICAL CONTROL l
l
'f
' REFERENCE J
FNP TS Bases 3/4.4.7.2 pg B 3/4 4-4 000037G03
..(KA's) i 1
1 1
.I ANSWER 4.24 (1.00) j J
Not under the control of the operator (0.25) AND (0.25) incapable of being )
controlled by the operator (0.25) using available equipment (0.25).
l REFERENCE l
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. Westinghouse Background Document for E-2
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194001A102
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END OF EXAMINATION-**********)'
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l TEST CROSS' REFERENCE Page 1 l 1
OUESTION VALUE REFERENCE 1.01 1.00 ZZZ0000001 1.02 1.00 ZZZ0000003 1.03 1.00 ZZZ0000005 1.04 1.00 ZZZ0000006
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1.05 1.00 ZZZ0000010 l
1.06 1.00 ZZZ0000013
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1.07 1.00 ZZZ0000014
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1.08 1.50 ZZZ0000019 1
1.09 0.50 ZZZ0000002 1
1.10 2.00 ZZZ0000004 1.11 1.50 ZZZ0000008 i
1.12 2.00 ZZZOO00009 1.13 2.50 ZZZ0000012 1.14 1.00 ZZZ0000013 l
1.15 1.50 ZZZ0000017 j
1.16 2.00 ZZZ0000007 i
1.17 1.00 ZZZ0000015 l
1.18 1.50 ZZZ0000016 1.19 1.25 ZZZ0000020 1.20 1.75 ZZZ0000021 1.21 1.50 ZZZ0000022 1.22 1.00 ZZZ0000018 l
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e TEST CROSS REFERENCE.
.Page 2
QUESTION VALUE REFERENCE 3.02 1.00 ZZZ0000057 3.03 1.00 ZZZ0000058 3.04 1.00 ZZZ0000059 3.05 1.00 22Z0000064 3.06 1.50 Z2Z0000053 3.07 1.50 ZZZ0000069 3.08 1.50 ZZZ0000052 3.09 1.00 ZZZ0000060 3.10 1.00 ZZZ0000070 3.11 1.00 ZZZ0000071 3.12 2.00 ZZZ0000056 3.13 1.25 ZZZ0000049 3.14 0.75 ZZZ0000050 3.15 1.00 ZZZ0000051 3.16 1.50 ZZZ0000054 3.17 1.00 ZZZ0000063 3.18 1.25 ZZZ0000065 3.19 1.00 ZZZ0000066 I
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'U.'S.
NUCLEAR REGULATORY COMMISSION' cSENIOR REACTOR OPERATOR LICENSE EXAMINATION-FACILITY:
FARLEY '
.]
REACTOR TYPE:
PWR-WEC3 DATE ADMINISTERED:'
87/08I24-J 1
EXAMINER.
BITTER.S' CANDIDATE j
l INSTRUCTIONS TO CANDIDATE:
Sj Use. separate. paper.for the answers. -Write answers on one side.only..
StapleLquestion sheet; on top'of the answer sheets..
Points.for each l
' question:are indicated in parentheses after the question.
.The passing tj grade requires at least 70% in each category and~a. final ' grade of,at.
L least 80%.
Examination papers will'be picked up six1(6)- ' hours after-the examination starts.
L
% OF CATEGORY
% OF CANDIDATE *S
' CATEGORY VALUE TOTAL SCORE-VALUE
-CATEGORY-a l
30.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS,AND THERMODYNAMICS 30.00 25.00 6.
PLANT SYSTEMS DESIGN,' CONTROL, AND. INSTRUMENTATION' 30.00-25.00 7.
PROCEDURES NORMAL, ABNORMAL,
~
EMERGENCY.AND RADIOLOGICAL CONTROL:
30.00 25.00 8.
ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND-LIMITATIONS 120.0 Totals Final Grade All work'done on this examination is my,own.
I have-neither given-nor received aid.
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Candidate's. Signature.
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l
During the administration of this examination the following rules apply:
- 1. - Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to.be limited and only one candidate 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.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
l 4.
Print your name in the blank provided on the cover sheet of the I
examination.
1 S.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
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8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10 Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
13..The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE l
OUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
'17.
You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
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18e When you complete your examination, you shall:
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a.
4ssemble your exar,ination as follows:
(1)
Exam questions on top.
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(2)
Exam aids - figures, tables, etc.
'l (3)
Answer pages including figures which are part of the answer.
I b.
Turn in your copy of the examination and all pages used to answer j
the examination questions.
I c.
Turn in all scrap paper and the balance of the paper that you did j
not use for answering the questions.
j d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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THEORY OF-NUCL' EAR POWER PLANT OPERATION.
Page. 4L l
FLUIDS.AND THERMODYNAMICS i
t' l
'OUESTION 5.01
.(1.00)
I l
A reactor is shutdown' with a boron concentration of 1100 ppm,
'and a SDM of 12 percent delta'K.per K.
Th'e source range.
~
j channel indicates 10 cps.
The boron concentration 11s:then j
reduced'by 100 ppm.. Assume that DBW is -10 pcm/ ppm.
Choose.
the answer that is closest to the final Keff.
1 l
.a) Keff =.980 b)'Keff ='.985 1
c) Keff
=.990 d) Keff =.995 l..
l' GUESTION-5.02 (1.00)
Answer EACH'of the following statements concerning differential boron-worth-(DBW): TRUE or FALSE.
-a)
DBW increases as Tavg> increases, b)
DBW increases as boron concentration increases.
l' QUESTION 5.03
. (1.00)
A system containing saturated vapor at'585 psig is. leaking via a crack 11n a L
pipe wall.
The pressure downstream of the crack is 335 psig.'
Which ONE of the following represents the condition of the fluid' downstream of the~
crack?
NOTE
-The steam tables should be.used.in lieu of1the Mollier Diagram a)
Wet vapor-with a very high quality (quality is'near:100 percent)~'..
b) - Wet vapor \\with a quality of approximately,60' percent.
c)
Superheated vapor with the amount of superheat'..'less than 2 1
degrees F.
a) ' Superheated -vapor with the amount of E superheat grea ter f than 121
- degrees F.
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THEORY OF-NUCLEAR POWER PLANT OPERATION, Page S-FLUIDS.AND THERMODYNAMICS QUESTION S;04.
(1.00)
All~three RCPs are operating.
Each has a flow rate " m " ~. 'The combined flow:
rate (core. flow)'is "M".
From the four possible answers below,1 choose the
.l ONE CORRECT ANSWER that describes the RCS response if one RCP is: secured..
a) The core flow (M) will increase.
b) The core flow (M) and the. flow of each of the operating RCPsf(m) will.
-l increase.
c) The' core flow (M) will decrease and the flow of each of the operating.
j RCPs (m) will increase.
d) The core flow-(M) will not changa because the remaining RCPs'will increase their. individual flows (m).
1 l
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OUESTION 5.05 (1.00)
The unit startup procedure, UO'-1.2, requires that criticality data be taken while at 10E-8 amps in the. intermediate' range.
If, during a j
xenon-free reactor startup at MOL, the operator " overshot" 10E-8. amps and instead leveled off at 10E-7 amps, which.ONE of'the1 following statements is correct?
i a) At 10E-7 amps, there are few or no effects'from nuclear' heat. 'However, because the reactor power is a decade' higher, the. criticality data rod position.ill be higher than it would be at 10E-8' amps, b) At 10E-7 amps, the effects of nuclear heat are-partially. offset by the effects of being at a higher reactor power; therefore,.the criticality data rod position will be lower than.it would be at'10E-8 amps.
c) At 10E-7 amps, there are few or no effects from nuclear-heat;.therefore, the criticality data rod position will be the same as it would'be.at 10E-8 amps.
d) At 10E-7 amps, there are' substantial effects from nuclear heat; therefore, the criticality data rod position will be higher-than it would be at 10E-8 amps.
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THEORY OF NUCLEAR POWER PLANT OPERATION, Page 6
FLUIDS.AND THERMODYNAMICS OUESTION 5.06 (1.00)
Assume that the reactor is critical (Keff 1.0000) at 10E-8 amps.
If +5
=
pcm is added to the core, choose the ONE statement from those below that BEST describes reactor behavior, a)
Keff will increase.
Neutron population and core power will continue to increase indefinitely until the operator inserts -5 pcm by using control rods or boron.
b)
Keff will increase.
Neutron population and core power will increase by an amount equal to il.005 percent of their original values, c)
Keff will increase.
Neutron population and core power will increase.
This will add negative reactivity due to power defect.
The power defect will cancel the +5 pcm.
The core power will level out by itself at a slightly higher power.
d)
Keff will increase.
Neutron populatir' and core power will increase.
This will add negative reactivity due to power defect.
The power defect will cancel the +5 pcm.
The core power will rotarn to the original power.
QUESTION 5.07 (1.00)
During a reactor startup, just prior to reaching criticality, the SUR meter indication will respond to a given amount of rod withdrawal by:
(choose the one CORRECT answer) a)
Rising slowly, then slowly falling off to zero b)
Rising rapidly, then slowly falling off to zero c)
Rising slowly, then rapidly falling off to zero d)
Rising rapidly, then rapidly falling off to zero
(*****
CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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.THEURY OF NUCLEAR POWER PLANT' OPERATION.
Page 7
j FLUIDS.AND THERMODYNAMICS.
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' -QUESTION' S.08 (1.00) l
~
Which ONE of the following statements. correctly describesLthe indications
.that would be observed if a centrifugal pump were started and operated with-1 its discharge valve shut.' Assume no recirculation flow.-
l a)
High starting current and minimum-running amperage L
.)
b)
Minimum starting current and minimum running amperage c)
High starting current and high running amperage-Minimum starting. current'and high running. amperage j
d)
QUESTION S.09 (1.00)
Which one of the following will cause the Axial Flux Difference.to become more positive (less negative)?
a) A power increase with power defect compensatedEfor by. dilution only.
b) A power increase with power defect compensated'forfby' rod. withdrawal only.
c) i'.e buildup of xenon in the top portion of the core.
d) The burnup of xenon in the bottom portion of the core.
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1 50 THEORY OF NUCLEAR POWER PLANT OPERATION, Page 8
FLUIDS.AND THERMODYNAMICS l
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1 QUESTION 5.10 (1.50) l Two identical reactors are taken critical.
Reactor A has a rod speed of 40 steps per minute.
Reacter B has a rod speed of 30 steps per minute.
Assuming a continuous rod withdrawal in each case, answer A, B,
or THE SAME to each of the following questions:
1 a) Which reactor will achieve criticality first?
b) Which reactor will have the highest critical rod height?
1 c) Which reactor will have the highest source range count at criticality?
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QUESTION 5.11 (1.50)
Indicate whether each of the following INCREASES, DECREASES, or DOES NOT l
AFFECT the AVAILABLE NPSH of a centrifugal pump; a) The pump discharge valve is throttled shut.
b) The temperature of the pump suction fluid decreases.
c) The nitrogen blanket pressure on the suction side supply tank increases.
QUESTION 5.12 (2.00)
An estimated critical position has been calculated for reactor startup that is to be performed 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after a trip from a 100 day full power run.
For each of the following events / conditions, state whether ACTUAL critical rod position is HIGHER THAN, LOWER THAN, or DOES NOT DIFFER FROM the ESTIMATED critical rod position. Consider each event / condition separately, a) Steam dump pressure setpoint is decreased by 100 psig.
b) Startup is delayed for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
c) The present baron concentration is 40 ppm lower than that used in the ECP calculation.
d) Condenser vacuum decreases by 2 inches Hg.
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CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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9.
THEORY OF NUCLEAR POWER PLANT OPERATION.
Page AP FLUIDS.AND THERMODYNAMICS
.OUESTION S.13 (2.00)
For'each of the following, indicate.whether the differential. rod'worthtof' an individual _ control. rod will' INCREASE, DECREASE, or IS NOT AFFECTED.
Consider each case separately.
a) An adjacent rod is inserted to the same. height.
b) Moderator temperature is increased.
(
c) Boron concentration is decreased.
l.
d) An adjacent burnable poison rod depletes.
f GUESTION S.14
( 1. 50 ).
Indicate how each of the following parameters changes.would' affect DNBR.
Use the term's INCREASES, DECREASES, or HAS NO EFFECT.
No. explanation is required.
1 a)
Tavg decreases b)
Reactor coolant flow increases c)
Pressurizer pressure increases GUESTION 5.15 (1.00)
Indi ate whether each o the following will-mak the mo rator temperature coe ficient LESS NEGAT VE, M E NEGATIVE, or H S NO EFFE T.-
a) ncreasing modera or temperature j
b)
Jecreasing baron concentra1. ion p
ANEWER a)
MORE NEGATIV
.(0.5)
I b)
ORE NEGATI (0.5)
(*****, CATEGORY S CONTINUED ON NEXTDPAGE *****)
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THEORY'OF NUCLEAR POWER PLANT OPERATION, Page.10 i
FLUIDS.AND THERMODYNAMICS QUESTION 5.16 (1.50) i How defeach of the following parameters: change (INCREASE, DECREASE, or REMAINS THE SAME) if.one main steam isolation. valve closes with-the-plant at 50 percent load.
Assume that control rods are in: manual and that-no trip occurs'.
a).Affected' loop' cold leg temperature s'
)
b) Affected loop's steam generator pressure c) Unaffected loop cold leg temperature OLESTION 5.17 (2.00)
Answer EACH of the following statements.concerning xenon-135.
production / removal: TRUE or FALSE.
a.
At full power, with equilibrium conditions,'approximately one half of the xenon is produced by iodine decay and the other half is produced directly from fission.
b.
When the reactor is shutdown, xenon burnup' effectively stops,'while.the L
decay of iodine continues. Therefore, the xenon concentration starts to-l increase.
c.
The production of xenon from iodine decay continues for up.to approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after shutdown.
Therefore, because all the production processes of. xenon have ceased, thefxenon reaches'its' minimum leval in the core at this time.
d.
Xenon. production.and. removal increases linearly as. power-level increases; i.e.,
the value of 100 percent equilibrium-xenon 1is twice the value of150 percent equilibrium xenon.
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THEORY OF NUCLEAR POWER PLANT OPERATION.
Page 11' FLUIDS.AND THERMODYNAMICS i
OUESTION 5.18 (1.00)
Assume.that the power range channels have been-adjusted based-on a calculated calorimetric.
Answer.each.of the following statements,: TRUE or FALSE::
i a) Iffthe-feedwater-temperature used in calculating the calorimetric had.
been 10 degrees lower than actual feedwater temperature,1then actual 1
reactor power would be higher than -indicated. reac tor power.
b)-If a main' steam atmospheric' relief. valve-had been' leaking during the-data-taking portion,ofsthe' calorimetric,-then' actual. reactor power: would be higher than indicated reactor power.
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L OUESTION 5.19 (1.00)
Name TWO effects of operating a motor-driven centrifugal pump at.or.beyond a runout condition.
I OUESTION 5.20 (2.50) j l
The reactor is subcritical at 2500 cps.When one crf'the steam dump' valves l
suddenly fails open.-, Assume that the reactor is UNDERMODERATED, at BOL, no operator actions are allowed, and that no trip occurs. Explain HOW and'WHY Tave'AND reactor. power change.
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GUESTION 5.21 (1.00) 01,the coefficients that comprise the power defect:
a)
Which'one contributes most to the change o'f power defect over-core life?
]
b)
Which one reacts first to a. sudden power change due:to rod movement 7 t-1
(*****
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.Sc ~ THEORY OF NUCLEAR POWER PLANT" OPERATION;
.Page 12 f
l' FLUIDS.AND THERMODYNAMICS-I
.i OUESTION 5.22 (1.50)
Answer each of'the following:
1 a) If'an_RTD element fails OPEN, what happens to indicated temperature?
I b) IfLan RTD fails'due to a SHORT CIRCUIT, what'happens to' indicated-1 temperature?
c) If a thermocouple' fails OPEN, what~happens;to indicated temperature?
I j
d): If a. thermocouple SHORT CIRCUITS 'in ' its instrument-w' ll,fwhat' happer.s to e
indicated temperature?
e).Which of-the two temperature detection elements,' RTD orL t he rmoc ou p l e,.'-
1 I
has the faster responseLtime?
f) Which of the two temperature detection elements, RTD orithermoccuple, j
~
develops its own electrical output?
j
't OUESTION 5.23 (1.00)
A reactor is producing 100 percent. rated thermal power at a core delta T of
~
42 degrees and a core mass flow rate of 100' percent.when a. blackout occurs.
Natural circulation is established and core delta T goes to 28 degrees.
If.
l decay heat is 2 percent, what is the percent core massLflow rate?.
j a) 1.5 percent b) 3 percent-c) 5 percent d) 7.5 percent l
(*****
END OF CATEGORY.
55*****)
d 6.
PLANT SYSTEMS DESIGN, CONTROL, AND' INSTRUMENTATION-
'Pageci3
.GUESTION
' 6 ; 01.
,(.i.00)
Which'.ONE of the following actions does th'e. operator perform to reduce'the RCS temperature'when the RHR system is in service for a normal plant.
cooldown?
a).He throttles open CCW from the RHR heat exchanger' outlet isolation valve.
b ): 'He throttles"open the RHR. heat exchanger outlet isolation valve.
l c) He' throttles open'.the RHR heat exchanger bypass valve, d) He. throttles open the-RHR miniflow recirculat' ion valve.
l
.OUESTION 6.02 (1.00)
For a design basis-(large break) LOCA,-which one of the following j
statements indicates the correct order in which ECCS components will produce flow?
H a.)
High Head, Containment Spray, Cold Leg Accumulators, Low Head j
b.)
Cold Leg Accumulators, High Head,. Low Head, Containment Spray c.)
High Head, Low Head, Cold Leg. Accumulators,-Containment Spray l
d.)-
. Cold Leg Accumulators, Low Head, High' Head, Containment Spray' l
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.6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION.
Page 14' OUESTION 6.03 (2.00)
Answer each.of.the-following: statements, concerning.the'AFW system. TRUE: or FALSE:
a.)'
TDAFP low speed coincident with an auto start signal will-NOT result-1 71 a common alarm on the MCB.
b.)
The TDAFP.and the MDAFPs in Unit One have their miniflow. flow control valves failed open.
c.)
A loss of instrument air will result in the fail-safe opening.of.TDAFP; steam supply valves 3235A'and 3235B.
d.)
Individual AFP discharge flow'CAN NOT be observed at the MCB when'all 3 AFPs are running.'
)
GUESTION 6.04 (1.00)
Incore instrumentation can be used in post-accident situations to perform three functions.
State TWO of the functions of the Incore Instrumentation system during post-accident situations.
' QUESTION 6.05 (1.00)
Hot Shutdown Panel (HSP) A contains-indications for'the HSPs.
.Two examples of indications found on HSP A'are SG waterg level and SG pressure? List-FOUR OTHER indications found on HSP A..
QUESTION.
6.06 (2.00)
State the five' automatic actions that occur due to.any mainigeneratoritrip..
.('*****-CATEGORY 6 CONTINUED ON'NEXTLPAGE.*****)~
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'6.
PLANT SYSTEMS DESIGN.' CONTROL, AND' INSTRUMENTATION'
.Page IS'
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'GUESTION 6.'07
' ( 1. 50 ) -
~
THREE conditions / interlocks must be met'in order to open letdown orifice isolation valve 8149A.
State the THREE conditions.
OUESTION 6.08 (2.00) l The train A RHR suction valve, 8701A, cannot'be opened unless'FOUR interlocks are met.
State the FOUR interlocks. Include setpoints;as applicable.
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OUESTION 6.09 (2.00)
Containment Cooler Fan A has.been running in fast speed for ten minu'tes.
Containment Cooler Fan B is not running.
The control switch'for each fan l
is in the AUTO position and the LOCAL / REMOTE selector switch:for each fan is in the REMOTE position.
Answer EACH of the fol'1owing1 questions independently.
l a)' State the TWO fast speed auto-start signals for Fan B.
b) State the TWO fast speed auto-trip signals for Fan A.
' QUESTION 6.10 (2.00)
Diesel engine shutdowns are divided into two (2) categor'iest Essential,and Non-essential.
a.)
' List the four ESSENTIAL engine protection shutdowns,.
(Setpointscare NOT. required.)
b.)'
List.FOUR NON-ESSENTIAL engine protection. shutdowns.
(Setpoints are NOT required.)
(***** CATEGORY 6 CONTINUED ON.NEXT:.PAGEv*****)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page.16 l
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-QUCSTION 6.11 (1.50) l Assume that a power cabinet URGENT failure has occurred in the Rod Control System.
State whether.each of the 3 types of CRDM coilst the lift coil, i
the stationary coil, and the movable coil, is" ENERGIZED or DEENERGIZED.
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QUESTION' 6.12-(1.00) 1 Indicate if.each of the following statements applies'to. UNIT 1 ONLY, UNIT 2
~
ONLY, or BOTH UNITS.
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a.)
Programmed S/G level varies from 33% at 0%.RTP'to 44% at 20% RTP and
(
is constant at 44% from 20% RTP to 100% RTP.
r b.)
The Delta-P program for feed pump speed l control varies linearly 4 rom
~
20% (43 psi) to 100% (215 psi) power.
1 QUESTION 6.13 (2,00) a.)
Indicate whether each of the.following situations: ARMS ONLY, ARMS AND ACTUATES, or HAS NO.EFFECT on the steam dump system, i
1.)
50% power, 18% step load INCREASE, Tavg is 6 F LESS than Tref, i
steam dumps are in the Tavg mode of' operation.
2.)
80% power, 7.5%/ min ramp DECREASE in turbine loadIfor 10' minutes, Tavg is 7 F GREATER than Tref, steam dumps are in the Tavg mode of operation.
3.)
Hot Zero Power, Tavg = 549 F,. steam dumps are in the.Tavg mode operation with 985 psig set into the steam. pressure controller.
4.)
Reactor trip,.Tavg = 549 F, steam dumps are in the Tavg mode'of operation.
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-6, PLANT SYSTEMS DESIGN. CONTROL, (ND INSTRUMENTATION Page.17-i i
i, OUESTION 16.14 (1.00) 1 State the 2 reasons.for the thirty second, time delay between'the. turbine trip and the main generatorftrip.
i L
QUESTION 6'15 (2.00)'
For EACH of'the following. cases, state whether the reactor willitr.ip in-response to a simultaneous failure low of both Intermediate RangeLchannels.
Justify yourl answers.
a) A reactor startup is'in progress;and. power.is'b percent..
b) The reactor-is at 20. percent steady-state power.
i
.)
OUESTION 6.16 (2.00)
R The Post-LOCA Atmospheric Control System consists of'four. (4)-subsystems.
l One of these subsystems is the Post-LOCA Air Mixing System...
L l
a.)
List the remaining three (3)-subsystems.
1 I
b.)
How is the automatic start signal.for,the Post-LOCA Air Mixing., fans reset?
l OUESTION 6.17 (2.00)
L l
4.16 kV bus IF has six (6) UV protection relays associated with it --Lthree (3) that provide degraded grid protection and three (3)'that provide LOSP protection.
a')
State.the automatic. action that occurs.when 2 ofm3 degraded. grid' voltage relays sense 90 percent nominal 1F bus, voltage.
b.)
State the 31 automatic actions'that occur as a result of the FIRS'T signal that"is generated when 2 of 3 LOSP undervoltage rel'ays. sense 70% nominal 1F bus vol tage.-
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page 18
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t i-QUESTION' 6.18 (1.00)L Explain the: design purpose of the J-tubes that are' installed'on the ShG' feedring.
1
.a QUESTION'.
6,19
-(2.00):
I
-The plant is operating lat 200.MWe and all ' primary plant controls' are'in
]
automatic.
The controlling pressurizer level channel. f ails ? low.. '.' Assuming '
1 l
no operator action, EXPLAINithe sequence of actions'that-will occur'and f
specif,y what' reactor trip,.if any,'will occur. DO NOT' include q
-annunciators / alarms in your explanation.
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PROCEDURES'- NORMAL, ABNORMAL, EMERGENCY-Page.19
)
AND RADIOLOGICAL CONTROL U
- QUESTION 7.01 (1.00) i l
The SUBCRITICALITY Status Tree indicates an ORANGE. path. 'From the j
statements below,; choose the one correct action to be taken..
a)
Go immediately to the referenced procedure'.
b)
Go.to the' referenced' procedure as. time permits c)
Monitor.other status: trees.-
If no higher'< priority. condition exists,,
then go torthe referenced procedure.
d)
Monitor other status trees.
If no' higher. priority condition exists, s
then go:to referenced procedure _as time-permits.
R
[
QUESTION 7.02 (1.00) 1 Which one of the following reasons CORRECTLY describes /the' basis for allowing RCP restart in-FNP-1-FRP-C.1, Response.to. Inadequate Core Cooling.
,p 1
a.)
Restarting the RCPs helps to mix the SI flow to protect the reactor R
vessel.from cold water.
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b.)
Once subcooling is established,-restarting the.RCPs helps-to collapse voids that may have formed in the' reactor vessel head.
c.)
Restarting the RCPs allows for restoration.of-pressurizer pressure control using. normal spray.
d.)
Restarting the RCPs provides for cooling of the core when' secondary depressurization does not alleviate inadequate core cooling.
QUESTION 7.03 (1.00)
Which o e of the following si iro s below DOESfNOT'.re uir.e 1 itiation of emergen y boration?'
- 1 a.)
T o or more control-ods fail.to insert on a r actor.tr; p
I'b.)
C trol Bank
'D' LO limitialar is' actuated.
' c.')
An uncontrolled c ntrol rod with rawal occur An uhe Vxplained d.)
eactor power incr aseioccu s-while a tH 4 5'/. power.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page.20 AND RADIOLOGICAL CONTROL 1
l OUESTION 7.04 (1.00)
L Which ONE of the following statements is CORRECT?
a.) A radiological exclusion area is a high radiation area but'a h'igh
')
radiation area is'NOT'necessarily a radiological exclusion area.
_j 1
b.) A contaminated area will be.a radiation area and a radiation. area will be a contaminated area.
~
c.) Radiological restricted. areas are : required to be. surveyed routinely.
I d.) All areas in the RCA'are either radiation. areas or hi g h. r ad i a't ion.
(
areas.
L QUESTION 7.05 (1.00) i State the two IMMEDIATE operator actions for loss of main feedwater. per FNP-1-ADP-13.0, Loss of Main Feedwater.
QUESTION 7.06 (1.00)
-t List the emergency exposure limits and explain when each is;used..
OUESTION 7.07 (2.00)
State the four reactor vessel vent termination' criteria per.FNP-1-FRP-I.3, Response'to Voids'in Reactor Vessel.
QUESTION 7.08 (1.50)
L List the six Critical Safety Functions in order of. priority.
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-(***** CATEGORY 7 CONTINUED ON NEXT.PAGE
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7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY Page :21 AND~ RADIOLOGICAL CONTROL QUESTION 7.09 (2.00)
FNP-1-ESP-0.2, Natural. Circulation-Cooldown'To Prevent. Reactor. Vessel.' Head Steam Voiding, has a step that directs the operator to' check for proper natural circulation.
State the five criteria per, FNP-1-ESP-0.2,- that indicate proper natural. circulation is taking place.
QUESTION 7.10 (1.00)
Assume the reactor is operating at 100% power when ONE-Intermediate Range channel fails h.gh, followed immediately by a reactorLtrip (from.other causes).
State the action that will have to be.taken to-ensure proper operation of the Nuclear Instrumentation System and~ explain.the reason'ing a
~
for taking: this action.
OUESTION 7.11 (1.50)
FNP-1-SOP-1.1, Reactor Coolant System, contains the RCP starting duties.
Answer each of the following questions concerning RCP starting duties.
a.)
How many RCPs may be started at any one time?
b.)
After any running period OR after any attempted start, how long.must
~
the operator wait before attempting a restart?
c.)
What is the maximum number of starts or attempted starts allowed in a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period?
OUESTION 7.12 (1.00)
FNP-1-ESP-0.2, Natural Circulation-Cooldown To Prevent Reactor Vessel Head-Steam Voiding, contains a. statement on.the foldout page.that calls'for the operator to monitor " adverse con tainment criteria. ", State the criteria that constitute ADVERSE ' containment conditions. -List setpoints..
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7.
- PROCEDURES - NORMAL,' ABNORMAL,. EMERGENCY Page 22; AND RADIOLOGICAL CONTROL t
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GUESTION 7.13 (2.00)
/
((;
. State the reason for each of the-following precautions given in:
s "
9 FNP-1-UOP-1.1, Startup from Cold Shutdown to Hot' Standby.
I a.)
Thereactorcoolantpumpsmustnotbeoperated.when.thWwMicr'.-1. seal differential pressure is less than 200-psid or VCT pressure is'.less l_
than 18!psig.
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reactor makeup water pump'snfil be b.)
With no RCPs in operation, one 9
'f I
racked out.
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)
A reactor coolant pump shall not be started with one or{.7dtjzd.e
- /
i c.)
of the(
RCS cold' leg temperatures less than or equal'to 310 F unlesswj 24% wide range cold presku(n' ke-j pressurizer water volume is less than i
each sf a, level' indication or the secondary. water temperature of h
- i.
'l '
kff-]
generator is~less than'50 F above each of theoRCS cold leg g
temperatures.-(There are 2 reasons 1or this procedure) y Jt
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IE '
OUESTION 7.14 (1.50) k, Lis t si+46-)-indic a t'ioris r--no-t-i*velvmg raa nit icn mon 4-torintyr t ha t 'a re -
symptoms of excessive RCS leakage according to FNP-1-AOP-1.0, Excessive RCS Leakage.
QUESTION 7.15 (1.50) i List FIVE (5) possible alarms on the Main Control Board that would indicate that an inadvertent dilution is in progress while operating at' power.
(Setpoints are NOT required)
GUESTION 7.16 (1.00)-
A CAUTION statement in FNP-1-FHP-5.13, Manipulator. Crane.. states'that' during'the insertion of fuel into.the: core, the-Weight Indicator must be
- watched continuously to ensure that indicated weight does not drop belowt 2250 pounds. Explain the reason for being concerned with.the weight-dropping below 2250 pounds.
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t-s) 4 7.}hND RADIOLOGICAL CONTROL
- ROCE.DURES - NORMAL,-ABNORMAL, EMERGENCY
'Page.23-k, j
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.( 1. 9 0 )~ 4, 5 QUESTION 7.17,
-(
One of the Emergenc'y 11 n\\mplementingPro u es is FNP-0-EIP-14,.Re-Entry
~ Procedures.
State thespdepose of this trocs,tdure'according to FNP-0-EIP-14.
t _,
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!p B.' QUEST. ION.'7.18
('2.00)
~
}
ope ator verify' tdatIthe F P-1-EEP-3, Steam.Geherator Tube' Rupture, has.t a mospheric relief.v es re set at 1035 psig nd in JTO.
/.e
.)
' Explain why.t e atmosph ric.)f>alief._ valve 1it.t se : points are I
e, d $' T 'N
^
INCREASED to 035'psig. j7 4
J bi r, Explain Ihy he'atmosphe/ic relief' valves remain in AUTO inst ad ~ of j
'being placed in MANUAL, j
s 6NSWER OE a.)
The setpoin s,are'increesed to 1035 sig so that the new set oints remain'. abo e'no-load' pressure (0.5)-
ut-are still elow the etpoint of th9 1ir t S/G safety (O.5)
Maintaining he setpoi t.above.
'f l
'y(
no-load p essure minligi.'
valve es radiati nl release f rom'. t e fault d S/G !!0.5
. 9 4
\\
b.)
Keepiry)thn[vabes in' auto (enoures that they are capablefof operating;
's t
atLcally. (0.5) j l
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GUESTION 7.19 (1.50)
During a LOCA, there are requirements. to trip the RCPs if certain criteria (RCP trip criteria) are met.' State the reason for-'having RCP trip requirements.
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GUESTION 7.20 (1.00) 1 In performing a cooldown WI1HOUT CRDM fans per ESP-0.2, the subcooling
'I margin is maintained at 223F - 226F until RCS, pressure is below 18001 psig. State the reasons for.-the upper and lower limits.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page :24 AND RADIOLOGICAL CONTROL l
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QUESTION 7.21 (1.50) i During a reactor startup per FNP-UOP-1.2, Startup of Unit from Hot Standby to Minimum Load, criticality was achieved at 110 steps on Bank C.
State i
what ac tions, if any, must be taken and wny.
I l
OUESTION 7.22 (2.00) l a) 9 tate TWO procedure-related conditions that dictate when monitoring of
)
.he Critical Safety Function Status Trees (CSFSTs) should begin.
1 b) State the plant condition during which the CSFSTs will be monitored for information only.
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(***** END OF CATEGORY. 7 *****)
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- 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS,
.PageM25l AND LIMITATIONS GUESTION
- 8.01 (1.00)
~
Pressurizer.PORV leakage falls under which one of the following Technical
. Specification leakage classifications?
a.)' IDENTIFIED LEAKAGE b.). PRESSURE BOUNDARY. LEAKAGE c.)' CONTROLLED LEAKAGE L
.d.)
UNISOLABLE LEAKAGE H
OUESTION 8.02:
(1.00)
Which ONE of'the following statements is NOT a basis for the,RWST, minimum' volume'and boron concentration limits.
a)
Sufficient water is available~to permit recirculation coo' ling' flow to the core.following a LOCA..
b)
RCS temperature will;be less.than_350 degrees'F before the RWST empties-following a LOCA.
c)
The reactor will remain subcritical f ollowing : mixing: of RWST 'and: RCS water volumes following:a LOCA.
I d)
'The solution recirculated within containment following a LOCA will have a pH value between 8.5'and'10.5.
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~ ADMINISTRATIVE PROCEDURES 2 CONDITIONS..
Page 26-AND-LIMITATIONS
- QUESTION 8.03 (1.00)
Answer EACH of the following statements concerning plant-menning: TRUE or' m
FALSE.
a) There will.be a break'of-at least.8Jhours (not including, shift-turnover time) between work periods.
b) An individual will not work more,than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> infany 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor moreethan172 hours 1
in any 7-day period (all including shift turnover' time).
l GUESTION 8.04 (2.00)
Answer-EACH of the following' statements concerning radiation-protection:.TRUE or FALSE.
a)
An entry into the reactor containment? building. requires a~special' radiation work permit.
I '
b)
In an area-(accessible to personnel) with a radiation intensity of-3.0 mr/ hour, one would expect to find'" Radiation Area" signs posted, c)
In general, Radiological Exclusion Areas are excluded from routine surveys.
l d)
The HP technican's approval is needed to LARELY exceed the 100 mr/ week l
administrative limit for whole body 1 cumulative exposure, ii I
.(***** CATEGORY 8, CONTINUED ON NEXi:PAGEs*****)"
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8.
ADMINISTRATIVE PROCEDURES. CONDITIONS,_
Page 27 AND LIMITATIONS QUESTION 8.05 (2.00)
Answer each of the following questions concerning EOP usage:
a.)
Describe the indication that is used in the procedures to denote substeps for which performance in sequence is unimportant.
b.)
Explain the operator action that is required if a " Response Not Obtained" contingency step is required but CANNOT be met and no other contingency steps exist, c.)
Explain the difference between a CAUTION and a NOTE statement used in the procedures.
QUESTION 8.06 (1.50)
List the FIVE individuals, by title, who report to the Technical Support Center (TSC) following full activation of the TSC.
I OUESTION 8.07 (1.50)
Step 5.3 of FNP-1-UOP-1.2, Startup of Unit from Hot Standby to Minimum Load, is preceded by a NOTE stating that RCS Tavr shall be greater than or equal to 541 F prior to criticality.
State the beses for the requirement that Tavg be at least 542 F.
QUESTION 8.08 (1.00)
The Operator-At-The-Controls (OATC) is responsible for operation of all equipment in the At-The-Controls (ATC) area.
According to FNP-0-AP-16, Conduct of Operation - Operations Group, what is meant by the term j
"operetion"?
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4 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 28 AND LIMITATIONS QUESTION 8.09 (2.00)
State the definition per FNP-0-AP-14, Safety Clearance and Tagging, of the following terms:
a.)
Tagging Official-b.)
Designated Operator c.)
Clearance d.)
Hold tags GUEETION 8.10 (1.00)
State the Technical Specification basis for requiring the reactor t be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> before moving irradiated fuel.
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1 GUESTION G.11 (1.00)
The plant has been at 100% power for 5 days following an 18 month fuel cycle refueling.
The A-train RHR pump breaker has been racked out.due to maintenance for 2 days.
A review of the operations surveillance log shows that the B-train SI auto start test th.it was REQUIRED during the refueling outage was not performed.
State what actions must be taken according to Technical Specifications.
See attached Tech Spec 3.5.2/4.5.2.
CUESTION 8.12 (1.00)
Unit 1 is in Mode 3 with a reactor start-up in. progress when all power to the RPIS display is lost.
State what action (s) must be taken according to Technical Specifications.
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CATEGORY 8 CONTINUED'GN NEXT'PAGE *****)
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ADMINISTRATIVE PROCEDURES. CONDITIONS.-
Pagr 29 AND LIMITATIONS l]
.OUESTION 8.13 (1~.00)
.i State the FOUR Emergency Classification categories:per FNP-0-EIP-9, Radiation Exposure Estimation and Classification of Emergencies.
I GUESTION-8.14 (1.00)
How can latching the Main Turbine affect Main Feedwater Pump (MFP)' speed,:
j as described in LER 86-005 for FNP Unit 2,'(Reactor Trip While Performing j
Trip Testing on the Main. Turbine)?
Include in-your' answer'whether MFP
)
speed will INCREASE, DECREASE, or REMAIN THE SAME during Main Turbine latching, and STATE the reason for this effect.
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GUESTION 8.15 (1.00) 1 State the minimum recommendation requirements to off-site agencies I
following a General Emergency Plan.
QUESTION 8.16 (1.00)
According to FNP-0-AP-16, Conduct of Operation - Operations Group, when AND l
under what circumstances is an UNLICENSED individual allowed to. manipulate' the controls?
I 1
l QUESTION 8.17 (1.00)
?
State the Technical Specification basis for restricting loads traveling over fuel assemblies in the spent fuel pit to < 3000 pounds.
l.
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- 8. CONTINUED ON NEXT PAGE *****)
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ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 30.
AND LIMITATIONS j
-QUESTION 8.18 (2.00)
There are TWO Technical Specifications l'imits on s' team generator tube leakage.
State EACH: limit and its corresponding basis.
QUESTION 8.19 (2.50) a) Define SHUTDOWN 1 MARGIN (SDM).
b)' State the minimum SDM allowed by Technical' Specifications for three loop operation while in Mode'1.
c) State the actions required by' Technical' Specifications,-if while in Mode.
3, the SDM is determined to be'lessithan the: minimum allowed value.
QUESTION 8.20 (1.50)
State the basis for the Technical Specification limitation on secondary coolant chemistry specific activity.
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QUESTION 8.21 (1.00)
According to FNP-AP-16, Conduct of Operations-Operations Group:.
a)
Who can authorize deviation from plant procedures?
b)
What extenuating circumstances must a-dev.iation meet?.
D QUESTION.
8.22 (1.00)
State 2 plapes where can Operations. Group AdministrativeLPoliciess,that are.
not identified inLFNP-AP-16 can be lochted.
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ADMINISTRATIVE-PROCEDURES. CONDITIONS.
Page 31 AND LIMITATIONS _
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QUESTION.-
8.23 (1.00)-
Definer'High Radiation Area
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1
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5
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- 4 k a tiA '
- Y ='s/t Cycle Efficiency = (Net Wor'k out)/IEnergy in)-
.c 2
~ 8 a Y t + 1'/2 (et )
W = Mg/ge -
ef 7,)/t 9eernot = T.4-T.g-4 zu E '= Mc2 e = (Y Y
2 KE = MY /2 Yr = Y + At T
o 4
PE ='Mgh v = e/t 2
P = Ogh/ge A = (30 )f4 S urbine = hj-h2r = vk,_2r' i
t
. Wk = P AY
.M=YA gy
' AE = 931( AM) g=1/v h-h vk Q = AAh y
21 1 -2f 2
Q = hC AT gZj + Yt
+ P jv1 + u t + q 1 - 2 = g22+Y2
+PY22+U2 p
4 wk -2 t
Q = UA AT Q
20eJ J
. Q '
29eJ J
J I
I uj + q1-2 = U2 + vk -2 I "l ' P
~
t o
Cy = Au/AT J
-x/TYL Cp = Ah/AT U j + Q -2 = U2 + Wk -2 1
t Wk = h -h J
TYL = 1.3/F g
h = h +%h f
fg HYL = -0.693/ F M = M(h -h )
1 2
1 d; = 1 d22 3
Wk -2 =
P dv h = u + Pv/J 1
J
=J l d), = 1 d22 g
PYg3=PV22 R/hr = (0.55CIE)/d2 (meters)
J 2
R/hr = 6CiE/d (feet) sur(t)
A = AN P = Po10 A = A e-At p '= p,,t/T g
A = 0.693/t SUR = 26.06/T SUR = 26( Ak/k)/i'+(h fr-( Ak/k)) T 1/2 t
eff = ((t 1/2)-(t ))/((t1/2) + (t ))
1/2 b
b E=6N A u1/E
-T = (1*/AK/K) + ((3,77 - ( AK/K))/((kff-( AK/K)h (d( AK/K)/dt))
T = 1/A
( AK/K) = (K,7f - 1)/ Keff A = c/T
( AK/K) = (1*/ T) + (5,ff (1 +X,ff T))-
/
N = 6.02 x 1023 'O l' = 10~ 4 A
0.1 seconds ~ I up power)
X eff = 0.05 seconds"1((I (
RRD = El = N 0 g down pover) -
- 1. 5M.
=
= 0.0125 seconds ~
screm) 1 P = (Elv)/(3 x 10 0)
= 0.08 seconds ~ l'(critical)
N a S - ( 1 -K,ff ) / ( 1 - K,77)
H = No-(K,fr)"
M = 1/(1-K,77) = CR; /CR,
.CRx = S/(1-X,frx)
M = (1-Kerro)/( 1 -K,ff j )
CR (1-K,77j ) = CR (I-Keff2)
. SDM = (1-K,77 )/K rr j
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f.1
Water Peremeters Miscellaneous Conversions 4,_
1 gel. = 8.345 lb I wett = 3.12 x 10 fissions /second m
10 1 gel. = 3.78 liters 1 curie = 3.7 x 10 dps 1 ft = 7.48 gel, 1 kg = 2.21 lbm Density = 62.4 lbm/ft I hp = 2.54 x 10 Btu /hr 3
6 Density = 1 grem/cm 1MW = 3.413 x 10 Blu/hr Heat of Vaporization = 970 Stu/lb I inch = 2.54 cm m
Heat of Fusion = 144 Blu/lb
- F = 9/5*C + 32 m
1 f t. H O = 0.4335 lb / inch
- C = 5/9(*F - 32) 2 7
Cp = 1 Btu /lb
- F
- R = *F + 460 m
Physics.Thermo, and Heat Transfer 1 Btu = 778 f t-lb 7
-8 d = 0.172 x 10 Btu /hr-f t
- R e = 2.718
~9
~
m = 1.67264 x 10 gram 1 ev = 1.60219 x 10 Coulomb-Volt p
~l
~
n = 1.67495 x 10 gram 1 Coulomb-Volt = 1.60219 x 10 Joule m
-8 m = 9.10956 x 10 gram e
0 c = 2.99 x 10 cm/sec
~I9 O = 1.6 x 10 Coulombs p
=,.,,,
~
h = 6.626 x 10 Joules-sec
~
0, =(-) 1.6 x 10 Coulombs 4
G = 6.67 x 10~ I ' N-m /kg 2
2 K = 9.0 x 10 N-m /C
Thermo. Formula Sheet.
'7?0ff-L
_Vt2 Pa-g
_V 2 P2 gc 21
+ 2ge + p ge Z + 2ge
+p
=
V12 Ptyt g2 2_
V2 Pavn scJ + 2scJ + J + ut e qt_
= scJ + 2gcJ + J + ua + Wkt_
p=
1 PE = MZg v
ge y_
KE = 1/2 MV
~
(A/X[
h,. h24 Pu h=u+
J J,2.
P = phg gi_:
To:
=
ge f,g wkt_
Pdu
=
Q1.2 = mCp (AT)
[a = pAV Pu = RT NPSH = (Pdy,- P ) EC s
PV = MRT peg-
- " A' Q _ 0IA! ~ Bev7-LA>Ksiet Q = m(Ah) h/Al h//
Q = UA(AT)
Mc, =
^# ~ T#"I na Water Parameters Conversions Cp = 1 BTU lba of 1 hp = 2.54 x 108 BTU he p = 62.4 lbm 1 MW = 3.41 x 10* BTU ft" he 1 gal. - 8.3 lba 1 inch = 2.54 cm.
1 gal. = 3.78 1 1 BTU ='778 ft -lbf 1 ft* = 7.48 gal 0F = 9/50C + 32 OR = F + 460 OX = oc + 273 l
0273o I
t i
1 EMERGtNCY CORE COOLING SYSTEMS l
i,(
3/4.5.2 ECCS SUBSYSTEMS-- T
> 350*F avg LIMITING CONDITION FOR OPERATION
.}
3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystemsLsha'11 be l
OPERABLE with.each subsystem comprised of:
One OPERABLE centrifugal charging pump, a.
b.
One OPERABLE residual heat removal-heat exchanger, One OPERABLE residual heat removal pump, and c4 d.
An OPERABLE flow path capable of taking suction from the refueling; water storage tank on a safety injection signal and transferring suction to the containment sump during the recirculation phase of operation.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
1 With one ECCS subsystem inoperable, restore the inoperable' subsystem a.
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 STAN08Y 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 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 />.
1 b.
In the event the ECCS is actuated and injects water into the Reactor j
Coolant System, a'Special Report shall be prepared'and submitted to the Commission pursuant to Specification 6.9.2 within 90 days
'l describing the circumstances of the actuation and the total accumu-lated actuation cycles to date.
The current value of the usage factor for each affected safety infection-nozzle shall.be provided-in this Special Report whenever-its value exceeds'0.70.
ep 1 l
l 9
4 FARLEY-UNIT 1 3/4 5-3
- AMENDMENT NO. 26'
i s '.
4 EMERGENCY CORE COOLING SYSTEMS _
SURVEILLANCE REQUIREMENTS
.~
4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that the following valves s.
are in the indicated positions with the disconnect device to the valve operators locked open:
Valve Function Valve Position Valve Number a.
8884, 8886 Charging Pump Closed to RCS Hot Lag b.
8U 2A, 8U 2R Charging Pump Open*
eischarge isolation c.
8889 RHR to RCS Hot Closed Lag In,jection l
At least' once per 31 days by vecifying that each valve (manual,
- b.. power operated or automatic) in the flow path that is not locked,
. sealed, or otherwise secured in position, is in its correct
- position,
>j By a visual inspection which verifies that no loose debris (rags, c.
trash, clothing, etc.) is present in the containment w..fch could be i
transported to the containment sump and cause' restriction of the This visual inspection shall pump suctions during LOCA conditions.
be performed:
1.
For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and Of the areas affected within containment at the completion of
- 2..
each containment entry when CONTAINMENT INTEGRITY is established.
d.
At lesst once per 18 months by:
1.
Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is between 700 psig and 750 psig.
A visual inspection of the containment sump and verifying that 2.
the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, inner cages) are properly installed and show no evidence of structural distress or corrosion.
>j'
'Siill be verified if charging pump 1A is declared' inoperable.
l
a o...
.g.
' =
i' ENERGENCY CORE' COOLING SYSTEMS.
L (0
q l
' SURVEILLANCE REQUIREMENTS (Continued) j e.'
-By verifying the correct position of each mechanical position stop.
for the-following ECCS throttle valves:
1.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of. each valve 'stro' king ;
d
- operation' or maintenance. on the valve when the ECCS subsystems-are required to be-OPERABLE.
l
- 2. -
At leni; once per 18. months.
Valve Number.
g
'CVC-V-8991 A/B/C i
CVC-V S989 A/B/C j
CVC-V-8996 A/B/C CVC-V-8994 A/B/C
.}
l f.
At least once per 18. months, during shutdo9n,= by:
1.
' Verifying that each automatic valve in"the flow path actuates.
1 to its correct position'on a safety injection test signal.
.)
2.
Verifying that each of the following pumpsistart automatically
]
i.
upon receipt of.a safety injection test signal:
4 l
a)
Centrifugal charging puesp l
b)
Residual heat' removal pump-
-l l'
g.
By verifying that each of the following pumps develops'the; indicated.
differential pressure on recirculation flow when tested pursuant to :
Specification 4.0.5:
q 1.
Centrifugal charging pump-
-1 2458'psig 2.
Residual heat removal pump _-
1 136 psig-p h.
Prior 'to entry inio Mo'de 3 from" Mode '4, verify' that: the' mectianical.
~
stops on low head' safety injection valves RHR-HV 603 A/B are intact.
q r
.(
1
]
l 6
=.
1 FARLEY-UNIT l' l3/4 5-5,
AMENDMENT.NC.'26 d
L
_ _- _=
. _ =
o.,
,c EMERGENCY-dOREC00LfNGSYSTEMS I
f-l
' SURVEILLANCE REQUIREMENTS (Continued) _
1 i.
' By performing a flow balance test,:duri.ng shutdown, following l
- completion of modifications to the Ecc5 subsystems that altar.the '
subsystem flow characteristics and verifying the following flow rates:
HPSI-System Single Pumo LPSI System - Single Pump 1193 gpa (each injection leg).
. 1-3981 gpa (total injection)'
l.
l l
I 1
I
)
+
.a l
./
l e
e i
d 6
4 t
(
FARLEY-UNIT 1
. 3/4'.5-6 AMENOMENT NO.'26
- c. -
g.
--- q x..
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- }-
Tg4
! 5.'
" THEORY - OF ' tJUCLEAR POWER PLANT OPERATION,;
Page.32:
j<,
FLUIDSJAND' THERMODYNAMICS i
n a-L L ANSWER' 3
-5.01
( 1. 00 ).
1 i
- 1. 0 )'-
lt c ) A (Y lV8 3l2>l67.
- REFERENCE-d
.-NUS,'. Nuclear. Energy Trai.ning,, Module 3, Reactor' Operation,.: Unit $12.1' '
Objective.1=
'192003K102
.192002K111 le/2002K114 192002K110
...(KA's) i
- ANSWER
'5.02 (1.00.).
a). FALSE '(0.5) b.) FALSE (0.5) i 1
REFERENCE
.:q NUS,~ Nuclear Energy Training, Module.3, Reactor Operation, Module-3, Unit 9.3, Objective 1; Unit 9.2, Objective-'1 l
'i 192004K110 192004K109
..(kA's) l i
l i
l l
ANSWER-5.03 (1.00) l
(
a.) (1.0)
REFERENCE l
..q NUS,xNuclear Energy.Traininge Module 4, Plant Performance l,
p 193004K125
~193003K125
..(KA's)
I Lh
.y_
- )
l
'I' a
...j - _
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4 3
I
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-5.
THEORY OF NUCLEAR POWER PLANT OPERATION.
Page 33 FLUIDS,AND T, THERMODYNAMICS I
l a
' ANSWER 5.04 (1.00' l'
c) (1.0) l REFERENCE NUS, Nuclear Training, Module 4, Plant Performance 1
1 191004K114 19100CK109 003000K505
..(KA's)
ANSWER 5.05 (1.00)
[
c) (1.0) 1 l
l REFERENCE NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation, Unit 13.5, Objective 2 192008K115 192008K114 192008K113
..(KA's)
ANSWER 5.06 (1.00) c) (1.0)
REFERENCE NUS, Nuclear Energy Training, Module 3, Reactor Operation, Unit 13.5, Obj ec tives 1,2 192008K11 192008K110
..(KA's) l l
(*****
CATEGORY 5 CONTINUED ~ON.NEXT PAGE *****)
+
- a,
2' i
J
. 5'-
JTHEORY':OF NUCLEAR POWER' PLANT O'ERATION, Page 34 FLUIDS,AND THERMODYNAMICS O
2 i
(1.00).
ANSWER 5.07 Eb)' (1.0)'
,- J i
i s
REFERENCE i
-NUS, Nuclear Energy Training, Module:3, Reactor Operation, Unit l12.4, Objective-1
?
1
-l
-'i 015000K506 192008K105 192008K194'
~192008K103-
..(kA's)
ANSWER 5.08 (1.00)
- l9 E
i b) (1.0)
/.
REFERENCE NUS, Nuclear Energy Training, Module:4, Plant Performance 1
l l
191005K105 191005K104
..(KA's)
ANSWER 5.09 (1.00) b k) (1.0) sto9/2J[87 1
REFERENCE NUS, Nuclear' Energy Training, Modules 2.and 3 Reactor Operation; L'i c en s ed.
Operator Retraining Cycle 4 Review Material 001000K506
.192005K110L
..(KA's)
- (*****-CATEGORY' 5 CONTINUEDf0NJNEXTfPAGE-*****)
r a.
._i_.
.mu 0-
- j
s
' ', W. r,.
j i
j t
i
.5.
. THEORY'OF NUCLEAR POWER-PLANT-OPERATION.
Page135 FLUIDS.AND' THERMODYNAMICS-JANSWER 5.10-(1.50)
.a).A (0.5):.
i b)_THE'SAME. (0.5)
'c).B (0.5) f l
. REFERENCE NUS, Nuclear Energy Training, Modules 2'and' 3',, Reactor. Operation-192008K104 192008K103 192003K101-(KA's)-
' 1 ANSWER 5.11 (1.50) c) INCREASES (0.5) b) INCREASES (0.5).
c) INCREASES (0.5)
REFERENCE l
I NUS, Nuclear Energy Training,-Module 4,LPlant(Performance, UnitU6.'5,-
[
Objectives 1,2 l
l 004000K604 191004K106
'191004K115 (KA's)
I ANSWER
5.12 (2.00) a)-LOWER THAN (0,5) b). LOWER THAN (0,5) e l~
Lc ) LOWER.THAN (0.5) i
-d)'DOES NOT DIFFER FROM-(0.5)
.0 f.-
R l
i
(***** CATEGORY. 5 CONTINUEDLON NEXT..PAGEl*****)f
.__.A "J
[
2-
$.__m a_
A m_m.,,
i
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i *f
<,=.
~
s 5F.
THEORY OF NUCLEAR' POWER PLANT-OPERATION.'
Page"36 FLUIDS.AND. THERMODYNAMICS
.j
' REFERENCE' i
NUS, Nuclear ~ Energy Training,-Module 3,LReactor Operation i
y 1
001010A207 001010K526 192006K110'
'192006'K107;
.. ( K A ' s' )-
ANSWER!
5.13
. ( 2. 00 )L 1
a) DECREASE (0.5)
'b)' INCREASE (0.5) c) INCREASE (0.5) d)' INCREASE'
- ( 0. 5 ):
REFERENCE NUS, Nuclear Energy Training, Modules 12 and 3, Reactor. Operation, Module.3',
Unit'9.4, Objectives 1,3 l
191002K114 191002K113
..(KA's)
ANSWER 5.14
-(1.50) a) INCREASES (0.5)
I b') INCREASES (0.5) c) INCREASES (0.5) l1 L
REFERENCE NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation,l Module 4,3
-Unit 8.2, Objective 1
'193008K105
..(KA's)'
a
\\'
(*****;CATEGOR'Y-5 CONTINUEDf0N NEXTiPAGE *****)
~
E 2
\\
F "i
U ij f.r;.
-f t..y.;
E
\\;
- 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION.
lPa'ge'37' l
. FLUIDS.AND THERMODYNAMICS,
.j REFERENCE' I
NUS, Nuc' lear' Energy Training,-. Modules 2.and 3,
- l
\\'
. Reactor Operation, Module.3, Unit 19.3, ' Obj ec tive 1 ; Uni t 9.2,- Obj ec tive '.1.
.192004K106
-192004K103
..-(KA's) l i
I (1.50).
-l l
ANSWER 5.'16
'l INCREASES [$e[INCAERSff.
d a) jof (0.5) b) RE"^.INC '9E E^"E
"(0.5) c.)' DECREASES (O.5)
REFERENCE NUS,. Nuclear Energy Training,: Modules 2 and'3,--React'or Operation 002020K508
..(KA's)
ANSWER 5.17
-(2.00) a.
FALSE (0.5) b.
TRUE (0.5) c.
FALSE (0.5)
.d.
FALSE (0.5)
REFERENCE
~
NUS, Nuclear Energy Training, Module 3,' Reactor Operation, Unit-13.6,-
Objective 1; Unit 10.2, Objectives ~1,2 i
192006K113 192006K107 192006K106-
.192006K104 192006K103 l
.;(KA's) c
~"
- --(*****nCATEGORY.z 5: CONTINUED GN NEXTfPAGE:*****)'_
o
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+
+
)
j 3
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l
~
- 5.'
THEORY OF NUCLEAR-POWER PLANT OPERATION.-
Page.38 4
j
- FLUIDS.AND THERMODYNAMICS-ii
-: I 1
ANSWER 5.18~
(1.00).,
i o
- a) FALSE (0.5) 1 b) FALSE (0.5)-
l REFERENCE
'I Tl NUS, Nuclear Energy Training',' Modules 2 and.3, ReactorLOperation.;
q 1
193007K108
'193007K10'6 015000A101 015000K504.
..'('KA*s);
' ANSWER 5.19 (1.00) 4 Any two of the following (0.5 each) gctre i
1)' Overload' 9 Mkr Ninbh $UdV' Md t}
- 2). Breaker. trips 9ggg g
- 3) Pump overheats
- /#, M//Nul74/#1
- 4) Pump cavitates gf2.}
- 5) Pump efficiency decreases O) klAX /thit&ttl b([ns9JMi & :
M -
1
$ Increnek amferugyp Bearby overhedug 7
. REFERENCE j
l NUS, Nuclear Energy Training, Module 4, Plant Performance
')
1 n
s 191004K112
..(KA's)
[.
i; ANSWER 5.20 (2.50) g7 2
' The extra gtga. flow causes Tave 'to decrease
'"25)'
This~ inserts'. positive reactivity.
Thus, power. increases.(0. 5)j..When power, reaches the-j POAH, the increasing fuel temperatq/v ' ' 0, _2
)
to;insertinegative;,
'd S; b ef) i n st r
the FTC.(0.25) "Thid
//i reactivity via graduallyf(0.25).until'.the:
.]
power level stabilizes' higher than..the POAHL(0;25)lf Tave willibeLl'ower.than 1
its original value,(0.25): by.an amountLsuf.ficientLto.-add,enough positive:
a reac tivity -( 0.25). to ' coun ter. - the. ef f ec ts, of. the.~ FT (O.'25) i Q TQY??b~ ^'
i 2t!!b?$dNgak()Jc]&M7628'&M44e~&4
- (*****-CATEGORY l.5 CONTINUED
- ON/NEXTsPAGE *****)c
,G kl b5 l '. >
ecc !
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.{
-Se ' THEORY-OF NUCLEAR POWER PLANT ' OPERATION.
Page 39.
1 FLUIDS.AND, THERMODYNAMICS.
8 REFERENCE L
NUS, Nuclear' Energy-Training, Module-3, Reactor. Operation-i s
192008K115 192008K114
.192008K113' (KA's)-
j
- ANSWER 5.21 (1.00)~
a) MTC (0.5) b) Fuel Temperature Coefficient (credit will be awarded <for: FTC, Doppler).
1 REFERENCE NUS, Nuclear Energy Training',. Modules'2 and 3, Reactor Operation,. Module;3,:
j Unit 11.4,. Objective 2; Unit 8, Objective 1 j
'l 1
192004K103 192004K106' 192004K108
('KA*s) 1.
1 ANSWER 5.22 (1.50) h
.,3 o a) indicated T fails high (0.e5) b) indicated T fails low
( 0. 25$ 2 +
g9 D 671 c) indicated T. fails low (0. M )to d) no change
( 0.2fr) Ao
-c; T/C
( 0. 2 51
>(-
.f ) T/C (0.2&)3o j
1
' REFERENCE d
OCONEE, OP-OC-IC-RCI 1
191002K114 191002K113
...(KA's).
l
-)
'f Y
4 i
i l
/
- ( * * * *
- CATEGORY: 5 CONTINUED:ON.NEXT: PAGE.*****).
J!
s
+
.3.
._n___________._-._____
.-__----_-_m--__-=--
- ~ - - -
l S.-
THEORY OF NUCLEAR POWER PLANT OPERATION, Page 40 FLUIDS.AND THERMODYNAMICS il
-l ANSWER 5.23 (1.00) b) (1.0)
]
REFERENCE NUS, Nuclear Energy Training, Modules 2 and 3, Reactor Operation 002000A105
..(KA's) l l
l
_(*****
END OF-CATEGORY S;*****)
A 6
'. ~ ?. 0; l
}..(
y '
,l 4
^
W hi!hilANTSYSTEMSDESIGN.'CONTROU,AND' INSTRUMENTATION Page 41-d I
. ANSWER..
6.01 (l'.00)'
]
1
' - b );..(1.00
' REFERENCE FNPS, Residual. Heat' Removal System, IIL,.pages 14 and 17'
-'005000K402
..(KA's) p
-j ANSWER 6.02 (1.00) l.
p p
b.)
1 REFERENCE' FNPS, Emergency Core Cooling System,-III-C.-
' 006000K603 006000K602' 006000K103 006000K506
.(KA's)
I s-
-ANSWER.
6.03'
.(2.00) a.')
FALSE.
(0.50)
N TRf,/6( 0. 50) gfg y 23 gf l
b.)
c.).
MR-FRW( 0. 50 )
d.)
TRUE (0.50)
REFERENCE r
FNPS, Aux Feedwater System, III I,
pages.10, 11, 12, 14,;and 15 061000G007 061000G008
..(KA's) r 1
i
(*****
CATEGORY 6 CONTINUED'ON NEXT1.PAGE'*****).
j
7
(..
'iN1 s :.
6; ' PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Pageo42'
(
.'Ii e
LANS'WER 6 '. 0 4 -
(1.00)
.v l
Any 12 of'the.following 3 (0.5.each)
- 1) It can determine.the effectiveness of core cooling-determining reactor vessel -level (or L2) It' can provide (an. alternate) method of
'.3)'ItlcanJdetermine.various~levelscof core degradation h cIe]Crni$ 1 66$
Vdid'ing)'
\\
i I
REFERENCE-
.j v.
'1 7
.FNPS,. Mitigating, Core Damage', OPS-532, page 6' FNPS, Incore Nuclear Instrument. System, XD, Objective 13 017020K502 017020K501 "017020A202 015000G004
..(NA's) 1 l
. ANSWER
'6.05 (1.00) l 1
Any 4 of the following 6: (0.25 each) 1
- 1) charging flow a
- '2)-pressurizer pressure L
3)-' pressurizer level L
f4) loop A~ wide range T. hot l-S)(loop A wide range T cold I
- 6) condensate storage tank' level REFERENCE
'FNPS, Hot Shutdown Panels, XR, Objective 2 000b68G006 000068K201 000068A112
..(KA's)
(*****1 CATEGORY 6 CONTINUED ON NEXT PAGE *****)
u
i 6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Pace 43 l
ANSWER 6.06 (2.00)
All b of the following: (0.4 each)
- 1) It trips the switchyard high side breakers for the generator.
- 2) It trips the breakers on 4160 KV buses A, F, C,
D, and E from the UATs.
- 3) It initiates a fast dead bus transfer of 4160 KV buses A,B.
and C.
l
- 4) It trips the 41 breaker.
- 5) It trips the turbine.
q l
REFERENCE FNPS, Main Generator and Auxiliaries, OPS-40202B (SOT) and S2105C (OLT),
Objective 6 045000G014 045010K423
..(KA'c)
ANSWER 6.07 (1.50) 1.
Both letdown isolation valves (LCV-459,460) must be open. (0.5)
]
2.
Pressurizer level must be greater that 15 percent.
(0.5)
R 3.
No phase A CI signal can be present.
(0.5)
REFERENCE FNP3, Chemical and Volume Contro1 System, IIG, page 15 FNPS, Chemical and Volume Control System, OPS-52101F Objective 4 004020K403
..(KA's)
ANSWER 6.00 (2.00) 1.
Train A RWST suction valve (8809A) is shut. (0.5) 2.
Train A charging pump suction valve (8706A) is shut. (0.5) 3.
RCS pressure (0.20) is less than 402.5 psig (0.10) as sensed by the transmitter on loop C (PY-402) (0.20).
f 4.
Pressurizer vapor space temperature (0.40) is less than 47y degrees F (0.10).
$jf fll3 0 ?
(*****
CATEGORY 6 CONTINUED ON NEXT PAGE *****)
a 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
.Page 44 REFERENCE FNPS, Residual Heat Removal System, IIL, page 8 FNPS, Residual Heat Removal System, 1986 Cycle 2, License Retraining, Objective 3 005000K407
..(KA's)
@!dM n-ANSWER 6.01 (2.00) 9 1
v H
j a) overload on Fan A (0.h) low d/p on Fan A (0.5) b) bus UV ( 0.% )'l q 4
SI signa / (O.
I OU@hocLf ((),y REFERENCE FPNS, Containment Spray and Cooling System, IIID, page 21 FNPS, Containment Spray and Cooling System, OPS-52102C, 1987 Cycle 3, License Retraining, Objective 5 4
022000G007
..(KA's)
ANSWER 6.10 (2.00) a.)
ALL of the following at (0.25) each:
1.)
Generator phase differential 2.)
Engine fail to start 3.)
Engine overspeed 4.)
Low lube oil pressure b.)
Any four of the following at (0.25) each:
i 1.)
Local stop pushbutton 2.)
Remote stop pushbutton 3.)
Lube oil high temperature 4.)
High jacket water tempeature 5.)
Low Jacket water pressure 6.)
High crankcase pressure 7.)
Barring device engaged (PC-2's ONLY)
(*****
CATEGORY 6 CONTINUED ON NEXT PAGE *****)
h d
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page 45 REFERENCE i
l FNPS, Diesel Generator and Auxiliaries, IV D, pages 28 and 29 064000K402
..(KA's) i 1
1 ANSWER 6.11 (1.50)
{
i 1.)
Lift coil - deenergized (0.50) l 2.)
Stationary coil - energized (0.50) j 3.)
Movable coil - energized (0.50) j i
REFERENCE I
l FNPS, Rod Control System, XE, objective 18 1
l 001050K401
..(KA's) b 1
ANSWER 6.12 (1.00) a.)
BOTH (0.50) b.)
UNIT 2 ONLY (0.50)
REFERENCE FNPS, SGWLCS, X B, pages 7, 13, 16, 17, and 21 FNPS, VI and VII-Differences, OPS-52108H 035010K401
,,(KA's)
ANSWER 6.13 (2.00) a.)
1.)
HAS NO EFFECT (0.50) 2.)
ARMS AND ACTUATES (0.50) 3.)
ARMSANDACTUATES$f((0.50) t f> ^
4.)
ARMS AND ACTUATES 0.50) b Mi dhEU
-gcakh g $a g: HAS NO 5'W aswf "b 9(b*
U gj 9lulb7
- s
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l
~.
I 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page 46
]
i e
REFERENCE
)
n FNPS, Steam Dump System, X J, objectives 2, 5,
6, 8
1 1
041020K414 014020K409 041020K417 014020K411 041020K105 i
..(KA's)
ANSWER 6.14 (1.00) i i
Both of the following answers:
1
- 1) It ensures that the RCPs continue to run for 30 more seconds prior to l
deenergizing the 4160 KV buses. (0.5) j
- 2) It reduces the chances of the main turbine overspending. (0.5) j a
i REFERENCE FNPS, Main Generator and Auxiliaries, OPS-4020B (SOT) and 52105C (OLT),
Objective 7 j
\\
)
s f
045000G014 045000K413
..(KA's) i ANSWER 6.15 (2.00) a) The reactor will trip (0.25) because with both IR channels failing low, P-6 drops out, the SR instruments reenergize, and a SR level trip occurs. (0.75)
{
b) The reactor will not trip (0.25) because P-10 is in effect.
This interlock will prevent the SR instruments from reenergizing. (0.75) 1 1
1 1
f l
'l i
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I i
i m__________
J
l l
6.
PLANT SYSTEMS DESIGN. CONTROL, AND INSTRllMENTATION Page 47 REFERENCE R
FNPS, Excore Nuclear Instrument System, XC, pages 16, 17, and 20 FNPS, Excore Nuclear Instrument System, OPS-52201D, 1987 License Retraining Objectives, Cycle 1, Obj ec tives 5 and 7 015000K401 015000K104 015000K407 015000K301 015000A202
..(KA's) j l
ANSWER 6.16 (2.00) a.)
ALL of the following at (0.50) each:
1.)
Hydrogen Recombiner System 2.)
Post-Accident Containment Venting System g
3.)
Post-Accident Sampling System (PASS) ([]y h y n A N/t1 # fE M )
6.)
Momentarily (0.10) placing each (0.10) fan's switch (0.10) to the STOP position (0.20). (ng Jy&Q g gypf a
- ,f&
REFERENCE Y
2d FNPS, Post-LOCA Atmospheric Control System, III E,
pages 1 and 15 028000G007 028000G004 029000K103
..(KA's)
ANSWER 6.17 (2.00) l a.)
The signal trips the 1F bus normal feeder breakers (DF01, DF15) (0.50) b.)
The first signal:
l 1.)
Starts the diesel (0.50).
2.)
Sheds loads on the 1F bus (0.50).
3.)
Sends a signal to trip the IF bus normal feeder breakers (DF01, DF15) (0.50).
REFERENCE FNPS, Diesel Generator Sequencers, IV E, page 20 Objectives: Lesson Plans for License Retraining Exam Preparation Reference, Diesel Generator Sequencers, IVE, Objectives 1 and 4 064000K411 064000K410
..(KA's)
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s 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION Page 48 ANSWER 6.18 (1.00)
(Ab 4
The J-tubes maintain water in the feedring so that thermal shock on the feed flow in a S/G at operating initiating /13 6 7 feedring is minimized when W </
temper a tureM (b ALSO ACCEPT THE FOLLOW NG FOR FULL CREDIT:
The J-tubes prevent water hammengdue to the condensation in the feedring that would occur if feed flow [were terminated and then reinstated -tt.e)
(0 ?) 9/23!f 7 W8 YlOlf7 REFERENCE j
FNPS, Steam Generator, II D,
pages 7 and 8 059000K103
..(KA's)
ANSWER 6.19 (2.00)
The following will occur:
1.)
Pressurizer heaters will deenergize (0.50) 2.)
Letdown isolation occurs (0.50) 3.)
Charging flow will increase (RCS inventory increases) (0.50) 4.)
Reactor trip occurs due to high pressurizer level (0.50)
REFERENCE FNPS, Pressurizer Pressure and Level Control, X H, pages 21 through 25 011000K604 011000K401 011000K104 011000K101
..(KA's)
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END OF CATEGORY 6 *****)
7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY
-Page 49 AND RADIOLOGICAL CONTROL ANSWER 7.01 (1.00) c) (1.0)
REFERENCE l
Westinghouse Background Document for EOP Development License Retraining, 1986 Cycle 3, Emergency Response Procedures, Classroom Objectives, Objective 5 194001A102
..(KA's) l ANSWER 7.02 (1.00) d) (1.0) l REFERENCE i
Westinghouse background information RCP Trip / Restart Generic Issue 000074K307
..(KA's)
ANS (1.00) g 'l#
4 1'f b) (1.0) 2 7 REFERENCE FNP-1-AOP-27.0 FNP-1-FRP-S.1 000024PG10
..(KA's)
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 50 l
AND RADIOLOGICAL CONTROL' ANSWER 7.04 (1.00) a) (1.0) l REFERENCE FNP-0-M-001 pg 10 194001K103
..(KA's)
ANSWER 7.05 (1.00)
Both of the following:
- 1) Trip the turbine. (0.5)
- 2) Verify both motor-driven auxiliary feedwater pumps start. (0.5)
REFERENCE FNP-AOP-13.0 FNPS, Condensate and Feedwater System VD, Classroom Objectives. License Training, Obj ec tive 12 059000G014
..(KA's)
ANSWER 7.06 (1.00)
Both of the following:
- 1) 25 Rem (0.25).
This limit applies when emergency onsite action is required to eliminate a source (0.1) or potential source (0.1) that represents a hazard to the general public (0.1' or to prevent a substantialjbf' loss in roperty (0.1).
N( N 7/2 97
- 2) 75 Rem (0.25).
This imit applies to lifesaving operations (0.1).
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4 7,
PROCEDURES - NORMAL, ' ABNORMAL, EMERGENCY Page 51 AND RADIOLOGICAL CONTROL l
REFERENCE FNP-0-EIP-14 FNPS, Cycle 3, Emergency Plan Training, OPS-565, 1987 License Retraining Objectives, Objective 17 194001K103 194001A116
..(KA's) l ANSWER 7.07 (2.00) l l
All of the following at 0.5 each:
- 1) less than 28 degrees subcooling
- 2) minimum pressurizer level is reached
- 3) RCS pressure drops by 200 psi
- 4) maximum allowed venting time is reached REFERENCE FNP-1-FRP-I.3 FNPS, Emergency Response Procedures, OPS-52302, 1987 License Retraining Objectives, Cycle 3, Objective 10 1
002000G014
..(KA's) l l
1
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 52 AND RADIOLOGICAL CONTROL ANSWER 7.08 (1.50)
(0.25 each; 0.15 for each function and 0.10 for each in correct order) 1.
Subtriticality 2.
Core Cooling 3.
Heat Sink 4.
Integrity 5.
Containment i
6.
Inventory REFERENCE l
l Westinghouse Background Document for EOP Development License Retraining, 1986 Cycle 3, Emergency Response Procedures, Classroom l
Objective 4 l
l l
l 000037GG12 194001A102
..(KA's)
)
l ANSWER 7.09 (2.00)
I (All at 0.40 each) 8 1.
SG pressure stable or trending down 2.
Subcooling monitor indicates increasing subcooling 3.
RCS hot leg temperatures stable or trending down 4.
Core exit thermocouple temperatures are stable or trending down l
S.
RCS cold leg temperatures at saturation for SG pressure l
REFERENCE FNP-1-ESP-0.2 l
1 l
002000A402 002020K514 193008K122
..(KA's)
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~
)
l 7.
PROCEDURES - NORMAL, ABNORMAL. EMERGENCY Page 53
)
AND RADIOLOGICAL CONTROL Q
ANSWER 7.10 (1.00)
Manually reset SR trip / block switches for BOTH trains (0.5) to reactivate j
j the SR high voltage and get SR indication (0.5) l l
REFERENCE FNP-1-AOP-18.0 015000K401 015000K407 015000A202
..(KA's) i ANSWER 7.11 (1.50) l l
l a.)
one b.)
thirty minutes j
c.)
three j
l l
REFERENCE 1
1 FNP SOP-1.1 l
003000K614 003000G101
..(KA's)
ANSWER 7.12 (1.00) g y
Containment pressure (0.2%) at or
>4 psig (0.2%)
-GR
( 0. 20 }- 9/h y/H/P 7 r
Containment radiation (0.2Q) at or > TOsE+05 R/hr.
(0.2%)
0 REFERENCE g7 FNP-1-ESP-0.2 103000G015 103000G014 103000G010 103000A101
..(KA's) l l
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1
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1 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 54 AND RADIOLOGICAL CONTROL l
k ANSWER 7.13 (2.00)
}
a.)
To ensure adequate seal flow (0.50) i b.)
To prevent an inadvertent dilution while on RHR (0.50) c.)
1.
To ensure sufficient surge volume for any RCS thermal expansion (RCS pressure increase)
(0.50)
)
2.
To ensure that any RCS thermal expansion (RCS pressure increase) is within RHP relief capacity.
(0.50) l 1
REFERENCE l
}
FNP-1-UOP-1.1
{
Reactor Coolant Pump, OPS-52101D, 1987 License Retraining Objectives, Cycle j
1, Objective 5 003000G010 003000K602 002000G010
..(KA's) i l
ANSWER 7.14 (1.50)
Six of the following at (0.25) each:
1.)
Increase in charging flow requirements (decreasing pressurizer level)
- 2. ) - Pressusei wo ie ty ve1ve t c m peret++re--hi-gk
- - 3. )
Centatement--coal o r dr2in laval high A)
""T level high b-), -Gent-e-i-nment-hunii u i i y ieityh -
- 2. )
rORv-tw porsrure h imn-j g/g7
-7.-)
"R T - -pr-esstyre-iigh
//
A )
RR-T--temps a tu r e hign 2-)
Dec reuw 'iv \\/c7 level l
REFERENCE FNP-1-AOP-1.0 FNPS, 1987 Licensed Retraining, Cycle 3, AOP-1.0 l
000028E106
..(KA's)
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- C**)
b 7.
PROCEDURES - NORMAL, ABNORMAL. EMERGENCY Page 55 AND RADIOLOGICAL CONTROL ANSWER 7.15 (1.50)
Any FIVE (5) of the following at (0.30) each:
1.)
Tavg/ Tref deviation alarm 2.)
Overpower Rod Stop 3.)
RCS High Delta-T 1 g a, g
[y gf//
4.)
OPdT Rod Stop 5.)
RCS High/ Low Tavg Answena r a c c ey*f'r
/rt// /j/
UMM
?tj M'
j i
6.)
RIL low limit alarm gy f}g p,pg,,Q Qpg 7.)
RIL 10-10 limit alarm 8-) OT23~ Kal 94' I c f C#fni an 9
/
REFERENCE v
taqnoS6 dM FNP-1-ARP-1.4,1.6,1.8 e $
I
/4 FNP-1-AOP-27.0 004020PG10
..(KA's)
ANSWER 7.16 (1.00)
Dropping below 2250 pounds means that the fuel assembly is interfering with adjacent assemblies. (1.0)
REFERENCE FNP-1-FHP-5.13 Fuel Storage, Handling, Refueling and Spent Cooling Lesson IX1 034000A203 034000A101 034000K601
..(KA's)
ANSWER 7.17 (1.00)
This procedure provides guidelines for reentry into areas that have been evacuated because of emergency. (1.0)
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY Page 56 AND RADIOLOGICAL CONTROL REFERENCE FNP-0-EIP-14 Excore Nuclear Instrument System, OPS-52201-D, 1987 License Retraining Objectives, Cycle 1, Objective 5 194001A101
..(KA's)
REFERENCE Westinghouse ERG Background for SGTR 000038G012 000038K306 000038K302
..(KA's) l ANSWER 7.19 (1.50) l l
To prevent excessive depletion of RCS inventory (0.50) that would lead to j
uncovering the core (0.5) if the RCPs were to trip (0.5).
ad a bitbof[at{
REFERENCE i
6 fA $ 'll U f ?> 7 j
Westinghouse Background Document RCP Trip / Restart Generic Issue l
I l
l 000009K323
..(KA's) 1 l
ANSWER 7.20 (1.00) i Upper limit to limit the pressure difference across the S/G tubes (0.50]
l Lower limit to ensure no bubble formation in upper head area
[0.50) l REFERENCE I
FNP-0-ESP-0.2 l
Westinghouse Background Document for ERGS 002000K111
..(KA's) t l
i
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b 7.
PROCEDURES - NORMAL. ABNORMAL, EMERGENCY Page 57 AND RADIOLOGICAL CONTROL i
l l
1 l
ANSWER 7.21 (1.50) l Emergency borate (0.5) and fully insert all control bank rods (0.5).
j Criticality was achieved below the O power RIL. (0.5) l l
i REFERENCE i
i FNP-UOP-1.2 pg 8 l
1 000023K301
..(KA's) l r
ANSWER 7.22 (2.00) a)
1.
When directed by EEP-0 (at step 29) (0,5) procedure ("Wkgq 2.
Any time that EEP-0 directs the operator to a different l
(0.5)
/
u SXITny ESP-o b) Loss of All AC Power (1.0) cuiff ye gccgp/hl l
- [ #U Ed REFERENCE 923 27 FNPS, 1986 License Retraining Objec tives, Cycle 3, Em rg ncy Response l
Procedures, item 3 l
194001A102
..(KA's)
)
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i 1
i i
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)
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l r
1 a
F 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS.
Page 58 AND LIMITATIONS ANSWER 8.01 (1.00) a) (1.0)
REFERENCE FNP TS DEFINITION 1.14 002000K405 002000G011 002000G005
..(KA's) l ANSWER 8.02 (1.00) l b) (1.0) l l
REFERENCE FNP TS Bases 3/4.5.5 pg B 3/4 5-2 006000G006
..(KA's)
ANSWER 8.03 (1.00) a) False b) False REFERENCE
- FNP, T.S.,
Administrative Controls 194001A103
..(KA's)
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4 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS.
Page.59 AND LIMITATIONS ANSWER 8.04 (2.00) i a) True (0.5) b) True (0.5) c) True (0.5) d) True (0.5)
REFERENCE 1
FNP-0-M-001, Health Physics Manual 194001K103
..(KA's) 4 ANSWER 8.05 (2.00) a.)
A black dot (0.25) b.)
Return to next step or sub-step on the left side of.the procedure in c.)
A CAUT ON statement contains information which must be followed to j
advoid personal injury, loss of life, or damage to equipment.
(0.25) j A NOTE statement contains additional information necessary to support an instruction or portion of a procedure (0.25), but the information does not meet the criteria for a CAUTION statement.
(0.25)
REFERENCE Westinghouse Background Document for EOP development FNP-0-AP-74 pg 14,15
)
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a.
}j 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 60 i
AND LIMITATIONS ANSWER 8.06 (1.50)
All of the following at 0.3 point each:
Emergency Director Operations Oircctt [fg4/4 pq j
f 23 $}7 1
Maintenance Manager
/
Technical Manager l
Health Physics Manager REFERENCE FNPS, 1987 License Retraining Objectives, Cycle 3, OPS-565, item i 194001A116
..(KA's)
ANSWER 8.07 (1.50) l ALL of the following at (0.30)-each:
1.) The MTC is within its analyzed temperature range 2.) The protective instrumentation is within its normal operating range 3.) The P-12 interlock is above its setpoint 4.) The pressurizer 10 capable of being in an OPERABLE. status with a steam bubble 5.) The reactor vessel is above its minimum RT-NDT temperature REFERENCE FNP TS Bases for 3/4.1.1.4 pg B 3/4 1-2 Classroom Objectives for License Retraining, Cycle 3-Unit Operating Procedures, Objective 9 001000G006
..(KA's) 1 1
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=
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 61-l AND LIMITATIONS ANSWER 8.08 (1.00)
ALL of the following at (0.25) each:
1.)
Actual controls manipulation.
l 2.)
Monitoring equipment and systems parameters.
3.)
Documentation of evolutions and significant events related to those l
systems and/or components controlled from the ATC.
4.)
Initiating or performance of operator actions required by plant Emergency Operating Procedures, Abnormal Operating Procedures, and i
Annunicator Response Procedures applicable to the ATC.
I REFERENCE FNP-0-AP-16 pg 20 l
001000G001
..(KA's)
(6-7/2 ANSWER 8.09 (2.00)
Manager (o.
a.>
SS, 5 6 % any individual-authorized by the Plant v't o implement the tagging procedure and with whom the final responsibility of insuring the adequacy) ofggy g,,,,jgf gg f f(, fe,c &;ciecut $' 3 WC 9/af/g,;
Tagging Oper tions Ord rs reg s if4-b'6c, t% {n,_),m gQ,j;jpjec
)
- b. ) V Any individual authorize by the Plant Ma, nager to execute Tagging
<;,pf l67 0'D O pe r a t i on s O rd e r s _ Jr.GWJ7-( 0. '1.) e q)O c.)
The formal authorization to an individual or a classification which provides the individual or persons working under'his supervision assurance that the system or portion of a system that he is working on has been isolated to prevent personnel injury or equipment damage.
(0,5) d.)
Safety tags which are placed on control points or devices to ensure personnel safety and equipment protection.
(0.5)
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-l 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 62 1
AND LIMITATIONS j
i REFERENCE j
FNP-0-AP-14 pg 2,3 AP-14, 1987 License Retraining Objec tiven, Cycle 1,
Objective 4 1
194001A112 194001A111 194001A110 194001A109
..(KA's)
ANSWER 8.10 (1.00) l 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> allows for the decay of short-lived fission products.
I REFERENCE FNP TS 3.9.3 Bases 034000G006
..(KA's)
ANSWER 8.11 (1.00)
I Apply TS 3.0.3 (initiate action within one hour to place unit in mode in which noncompliance TS does not apply.) (1.0)
REFERENCE FNP TS 3.5.2,3.0.3,4.0.3 005000G005
.. -( K A ' s )
1 ANSWER 8.12 (1.00)
Open RTB's immediately (1.0) l
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- 9 e
e.
8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Page 63 AND LIMITATIONS REFERENCE l
l 001000G014 014000K301
..(KA's) l ANSWER 8.13 (1.00)
ALL of the following at 0.25 each:
- 1) Notification of Unusual Event (NOUE)
- 2) Alert
- 3) Site Area Emergency
- 4) General Emergency I
REFERENCE
)
I i
FNP-0-EIP-9 Emergency Plan Training, OPS-565, 1987 License Retraining Objectives, Cycle 3,
Objective 16 l
l 194001A116
..(KA's) l l
ANSWER 8.14 (1.00) j i
Decrease (0.5), because latching the main turbine can cause significant j
drops in EH fluid pressure (0.5).
)
)
REFERENCE
.i I
l LER 86-005 for FNP Unit 2.
l l
FNP-SOP-28.1 precaution 3.33 l
l l
045000G010
..(KA's) 1 1
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8, ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 64 AND LIMITATIONS ANSWER 8.15 (1.00)
Recommend immediate evacuation (0.2) or shelter (0.2) for all of the general population (0.2) within a two-mile radius of FNP (0.2) and 5 miles downwind of FNP (0.2).
REFERENCE FNPS, 1987 License Retraining Objectives, Cycle 3, OPS-565, item 21 194001A116
..(KA's) l ANSWER 8.16 (1.00) 1 When it is part of training to qualify for an operator's license (0.50) l Only under the direction and in the presence of a licensed RO or SRO (0.50) l REFERENCE FNP-0-AP-16 pg 38 l
l 001000G001
..(KA's)
ANSWER 8.17 (1.00)
In the event the load is dropped, the activity released will be limited and f
any distortion of fuel will not result in a critical array.
REFERENCE FNB TS Bases B 3/4.9.7
)
l I
i i
i' 034000G006
.. -( K A ' s )
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_ _______ _ _ _ _J
9 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS,
.Page 65 AND LIMITATIONS ANSWER 8.18 (2.00) 1.) 1 gpm total tube leakage for all generators (0.50)
Ensures that the dosage contribution from the abe leakage will be limited to a small fraction of Part 100 limits in the event of either a SGTR or MSLB. (0.50) 2.) 500 gpd per S/G (0.50)
Ensures that tube integrity is maintained in the event of a LOCA or MSLB. (0.50)
REFERENCE i
FNP TS Bases 3/4.4.7.2 pg B 3/4 4-4 035000G006 035000G005
..(KA's)
ANSWER 8.19 (2.50) a) SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subtritical or would be subtritical from its present condition assuming all full length RCCAs (control and shutdown) are-fully inserted except for the single RCCA of highest reactivity worth which is assumed to be fully withdrawn. (1.0) b) 1.77 percent delta k per k (0.5) c) Immediately initiate and continue boration (0.25) at greater than or equal to 30 gpm (0.25) of a solution containing greater than or equal to 7000 pppm boron (0.25) or equivalent until the required SDM is restored (0.25).
REFERENCE l
FNP: Technical Specifications 3.1.1.1 Reactor Theory, OPS-313, 1987 License Retraining Objectives, Cycle 4, Objective 19 001000G006
..(KA's)
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e 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 66 AND LIMITATIONS ANSWER 8.20 (1.50)
The resultant offsite radiation dose (0.5) will not exceed a small fraction of 10 CFR Part 100 limits (0,5) in the event of a steam line rupture (0.5).
REFERENCE FNP, Technical Specifications, 3.7.1.4, B 3/4 7.1.4 035000G006
..(KA's)
ANSWER 8.21 (1.00) a) a plant operator (0.25) or licensed supervisory personnel (0.25) b) adherence to the procedure will create an undue hazard to personnel, 1
equipment (0.25) or the health and safety of the general public (0.25) l REFERENCE l
i l
~
l FNP-AP-16 l
FNP, 1987 License Retraining Obj ec tives, Cycle 1,
OPS-52301B, objective 3 194001A103 194001A102 194001A101
..(KA's) l ANSWER 8.22 (1.00)
- 1) Night Orders Book (0.5)
- 2) Operations Standing Policies Book (0.*)
l l
REFERENCE l
FNP, 1987 License Retraining Objectives Cycle 1,
OPS-52301B, item 6 194001A103
..(KA's)
ANSWER 8.23 (1.00) l to personnel (d
Any area accessible Vin which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 100 mrem.-td.0; g'9/D[f7
[ 0>#
lo $)'
w/3['7
<//23
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4 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, Page 67 AND LIMITATIONS REFERENCE FNP-0-M-001, Health Physics Manual 194001K103
..(KA's) l i
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END OF EXAMINATION **********)
i I
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