ML20197B915
| ML20197B915 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/15/1986 |
| From: | Bill Dean, Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20197B901 | List: |
| References | |
| 50-250-OL-86-03, 50-250-OL-86-3, NUDOCS 8610310030 | |
| Download: ML20197B915 (145) | |
Text
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ENCLOSURE 1 EXAMINATION REPORT 250/0L-86-03 Facility Licensee:
Florida Power and Light Company P. O. Box 14000 Juno Beach, FL 33408 Facility Name:
Turkey Point Units 3 and 4 Facility Docket Nos.:
50-250 and 50-251 Written and oral requalification examinations were administered at Turkey Point Nuclear Plant near flo ida City, Florida, y
M Chief Examiner:
/[
M(U du o
e William M. Dean Date Signed Approved by:
/wbE385 Tohn F. Munrei Acting Section Chief ITate Signed Sumary:
Requalification examinations on September 8-11, 1986 Written requalification examinations were administered to 4 R0s and 9 SR0s. Oral requalification examinations were administered to 2 R0s and 7 SR0s. All of the R0s and 6 of the SR0s passed these examinations.
(Note: Waivers of the oral examination were granted to 2 R0s and 2 SR0s who passed this portion of the requalification examination in April.)
Additionally, a Section 8 written re-examination was administered to an SRO in the Region II office on July 24, 1986. He passed this examination.
(Note: No facility comments regarding this examination were received.)
The performance during this portion of the requalification examinations is an improvement, and though not above the NRC's minimum pass rate for a satisfactory evaluation (80%), it is close enough to demonstrate that corrective actions applied to the Turkey Point requalification training program are successfully correcting past deficiencies. Continued attention is required to ensure success-ful standards are achieved and maintained.
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REPORT DETAILS 1.
Facility Employees Contacted:
- J. Kappas, Acting Plant Manager
- T. Finn, Assistant Operations Superintendent
- W. Miller, Training Supervisor (Nuclear)
- P. Baum, Operations Training Supervisor
- G. Hollinger, Training Staff
- Attended Exit Meeting 2.
Examiners:
- William M. Dean Chuck Casto Lawrence Lawyer
- Chief Examiner 3.
Examination Review Meeting At the conclusion of the written examinations, the examiners provided your training staff with a copy of the written examination and answer key for review.
The comments made by the facility reviewers are included as to this report and the NRC Resolutions to these comments are listed below.
R0 Written Examination Resolutions (SR0 questions in parenthesis):
a.
(1) Question 1.04 (5.03)
NRC Resolution:
Do not concur.
Although an attempt is made to ensure the material covered in the examination is referenced to a facility learning objective, in some cases, material covered by Hot License classes may not be covered by the most recent requalification cycle.
This does not diminish the importance of the material to all candi-dates.
No change in answer key is required.
(2) Question 1.09 (5.09)
NRC Resolution:
Concur that redistribution effects due to power changes are a reasonable answer.
The answer key is revised to accept answers which accurately explain redistribution due to power changes at BOL in accordance with facility reference material.
In addition, the accurate explanation of two of the three reasons will suffice for full credit.
Answer key has been modified.
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(3) Question 2.02 (6.02)
NRC Resolution:
Concur.
Recent change to procedures not held by NRC, modified the Hot leg Recirculation lineup.
Answer key has been changed to "b".
(4) Question 2.06 NRC Resolution:
Concur.
System modification not reflected in facility provided material held by NRC and should be corrected.
Answer key has been changed.
(5) Question 2.08a (6.06a)
NRC Resolution:
Concur.
Answer key will be modified to accgpt indication as another use of the current transformers for full credit, based on new material provided to the NRC.
Question 2.08b NRC Resolution:
Concur.
Answer key has been modified to require alarm in control room as part of the answer, as well as accepting suitable ground detection methods.
(6) Question 2.09a (6.10a)
NRC Resolution:
Concur.
Inconsistencies in facility reference material provided as enclosures for this examination may have led to incomplete 4
answers by the candidates.
Answers which state "CS to start" or
" breaker closed and CS to start" will be considered equivalent answers.
Suggest facility correct reference material.
(7) Question 2.12a (6.12a)
NRC Resolution:
Do not concur.
This fact was clarified by the proctor to all examinees during the exam.
No change to answer key is required.
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3 Question 2.12c NRC Resolution:
Do not concur.
Purpose of question was to determine if the examinee knew what, if any, effects the control circuit for the.
trip throttle valve has on system operations.
No change to answer key is required.
Question 2.12e NRC Resolution:
Concur.
Facility reference material depicts the main oil pump supplying the governor.
The examiner has determined by local observation that this is not the correct flow path.
Therefore, answer key has been changed to accept "will inject" with an accompanying explanation of the effects of loss of lubrication.
Suggest facility correct reference material.
(8) Question 2.14 NRC Resolution:
Concur. Will accept any correct design flow balance based on orifice selected, and the value of CCW flow through the thrust bearing.
(9) Question 3.05b (6.07b)
NRC Resolution:
Concur. Question asked for actions, not a description of a transient.
Answer key has been changed.
Question 3.05c NRC Resolution:
Concur.
Answer key has been changed to add trip.
Facility should correct existing reference material.
(10) Question 3.08a and 3.08b (6.08a and b)
NRC Resolution:
Concur.
Recommended answers are equivalent to answer key.
No change to answer key is required.
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(11) Question 3.09a (6.09a)
NRC Resolution:
Concur.
Setpoint not elicited by question.
Answer key has been changed.
(12) Question 4.17 (7.17)
NRC Resolution:
Concur, in part.
Correct discussion of the target based limit will be acceptable for +0.75 points, as it is discussed in the referenced procedure, but coincidence of NIs is required to gain full credit.
b.
SR0 Written Exam Resolutions:
(1) Question 5.02 NRC Resolution:
Do not concur.
The traces provided in this question are of sufficient quality and are based upon a generic PWR transient.
Without a copy of the analysis of the events which were simulated at Westinghouse, no conclusions can be formulated.
However, based upon Turkey Point's TAVG program, a possibility exists that Tcold may increase slightly depending upon the steam dump pressure setpoint.
Because the steam pressure setpoint was not provided in the question, this question is deleted.
(2) Question 5.06 NRC Resolution:
Concur.
The responses to this question will be graded in accordance with the facility reference material provided.
Any equivalent answers will be evaluated on a case-by-case basis for their adequacy.
No change to answer key is required.
(3) Question 5.07 NRC Resolution:
Concur.
The question is based upon a theoretical core and did not refer to either Turkey Point Unit in particular.
However, it is acceptable to include the effects of cycle X.
In addition, answers which describe two bases for flattening of the flux in the middle of the core are acceptable for full credit due to wording of question possibly leading examinees to this answer.
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(4) Question 8.15 NRC Resolution:
Concur.
Part a changed to accept "4" as a correct response due to SI pump requirements also contained in TS.
Note that this requirement conflicts with requirements of TS referenced in answer key.
c.
It should be noted that a telecon reference is not sufficient documentation to justify changing or modifying the answer key.
Documentation, utilizing controlled information, is required to support facility comments.
d.
Post examination review resulted in the following change:
SR0 5.08 (R0 1.08).
Original answer is actually a combination of redundant responses, which would not be elicited by the wording of the question.
The answer key will be changed as follows:
" Bank overlap is used to provide a more uniform differential control rod worth (+1.0) and a more even flux profile +(1.0)."
Answe,s that address these two areas will be accepted as correct.
4.
Exit Meeting At the conclusion of the site visit, the examiners met with representatives of the plant staff to discuss the examination process and any observations.
There were no generic weaknesses noted during the oral examinations.
It was noted that the operators had a better grasp and understanding of recently implemented emergency and abnormal operating procedures.
The cooperation given to the examiners was also noted and appreciated.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the examiners.
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NUCLEAR PEGULATORY COMMISSION REACTOR OPERATOR REGUALIFICATION EXAMINATION CACILITY:
TURKEY POINT 3&4
~-----
REACTOR TYPE.
PWR-WEC3 DATE ADMINISTERED: S6/0?/08 EXAMINER
- CASTO, C CANDIDATE
- INSTRUCTIONS TO CANDIDATE:
Read the attached instruction P3Se c a r e f u 1 ~_ y.
This enan: nation replaces the current cycle facility administered requclificction 2:: a m i n a t i o n.
Retraining requirements for failure af this examination are the same as
'for failure of a requalificction examinellon prepared and edninistered by your training staff.
Points for each question are indicated in parentheses after the question.
The pensins grade requirec et least 70%
in each category and e final-grade of at least 80%.
Examination papers Oill be picked up four (4) hours cfter the exam:. nation starts.
". O F CATEGORY
- . O F
' CANDIDATE'S CATEGORY
-VALUE TOTAL SCORE VALUE CATEGORY 8.00 25.00 PRINCIPLES OF NUCLEAR COWER PLANT OPERATIONr THERMODYNAMICE.
HEAT TRANSFER AND FLUID FLOW
S'00
2.
- 'LANT DESIGN INCLUDING SAFETY
"" 00
^
AND EMERGENCY SYSTEME 8.00 25.00
_--- 3.
INSTRUMENTS AND CONTROLS to'00 or'"'00
4.
PROCEDURES - NORMAL-A9NORMALr
^~
EMERGENCY AND RADIOLOGICAL CONTROL l
72.00 Totals l
rinal Grade i-s A ll wo r :- done on this examination is ny cun,
! have neither given nor received eid.
i Cendidete's Signature i
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-r----- - -, - - - - - - ~ -. - -. ~ -, - +
N 3 K80%
UNITED STATES k
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NUCLEAR REGULATORY COMMISSION
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REGION il 5
3 101 MARIETT A STREET, N.W.. SUITE 2900 I
ATLANTA GEORGIA 30323 o
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NRC RULES AND GUIDELINES COR LICENSE EXAMINATIONS Durine the administration of this examination the followin.c rules apio l v, :
1.
Cheating on the examination neans an autonatic denial of your application and could result in nore nevere penalties.
2.
Restroom trips are to be limited and only one' candidate at a time may leave.
You must avoid all contact; with anyone cutside the exaninction room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only te facilitate lesible reproductions.
4.
Print your name in the blank previded on the cover sheet of the examination.
5.-
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 #irst page cf each section of the answer sheet.
8.
Consecutively number each answer sheet, wrise 'End of Catescry __' as appropriate-ctart each category on e new pager write only on one side of the paper, and write "Last Page' on the lest answer sheet.
9.
Number each answer as to cate3 cry and nunbeer for exanple, 1.4r 6.3.
- 10. Skip at leact three linet between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your deck or table.
- 12. Use abbreviations only if they are conmonly used in facilitj literature.
- 13. The point value for each question is indicated in parentheses after the question and can be oced as a guide for the depth of answer required.
14 Show all ecleuletions, methodsr cr assumptions used to obtain en answer to mathematical problems whether indicated :n the -;u e s t i o n or not.
- 15. Partial credit may be giver 1 Therefore, ANSWER ALL PARTS Or THE QUESTION AND 00 NOT LEAVE ANY ANSWER 9 LANK.
- 16. If parts of the examinatior a*e not clect
<E te intent, ask quest ons of the_ examiner only.
- 17. You wst sign the statement an the cover sheet that indicates that the work is your eun and you 5 rc e not received or been giver eccictance ir Co%pleting th e e;< a n in e t i c n.
This must be done after the e": a m i n a t i o n has been completed.
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- 10. 'Ahen you : comple te your examination, you shall*
c.
Assemble your examinction as follows:
i (1)
Exam questions on top.
4 j.
(2)
Exam sids - f:gurest tables, etc.
l (3)
Answer pages including figures which are part of the-answer.
b.
Turn in your copy of the examination and all pases used to answer l
the examination questions.
l c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
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1-d.-
Leave the examination areer as defined by the examiner.
If after leavin3r you are found in this area while the exanination is still I
in progressr your license may be denied or revoked.
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
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T QUESTION 1.01 (2.00)'
Indicate how the parameters below respond and the reason why they react that way, on a loss of naturel circulction following a trip from 100%
equilibrium conditions:
a)
RCS wide range Delta T
b)
Difference eetween T(cold) and P(steam)
QUESTION
- 1. 0 2 (1.00)
Which set of parameters below best describes centrifugal pump runout conditions?'
a.
High discharge pressure, Icw flow, high power demand
-b.
High discharge pressure, low flow, low power demand c.
Cow discharge pressure, high flow, high power demand d.-
Low discharge pressure, high-flow, low power demand e.
Law discharge pressure, low flow, high power demand
. QUESTION 1.03 (1.00)
Mhich of the following curves (see attached pege) representing Xenon concentration is correct for the given power history?
QUESTION 1.04 (1.00)
For each of'the following parameters compare a Core Thimble Cell to a Normal Core Cell and indicate which of the.two has the higher value.
(rated conditions, average values) 1.
Total power production 2.
Power density 3,.Enthelphy rise 4.
Mass flow
- 5. DNDR
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. - PRINCIPLES OF NUCLEAR POWER OLANT DPERATION7 PAGE 3
THERMODYNAMICSr HEAT TRANSFEP AND FLUID FLOW GUESTION. 1.05 (1.00)
During a reactor startup, equal increments of reactiv.ty are added and the count rate is allowed to reach equillbt iun ecch time.
Choose the bracketed
( E]) word (s) that describe what is obser'/ed on the Source Rcnge recorder and/or SiJR meter.
a.
The change in equi;ibrium count rate is Clarger] Ethe sane] [cmsller2 each time.
b, The time "equired to coach equilibrix is E l o r, g e r ] Cthe sane] : shot ter]
eech time.
QUESTION 1.06 2.00)
Unit 3 has ;ust rectarted followine a refuelina cutage while Unit 4 is a
a neer EOL.
Answer the fellowing rescrding the dif"etencer in plent esponse between the two unito(enplain your answers);
a)
At a steady power level of 10EE(-8) emps during a st ar tup r equal reactivity additions are made (appron:.mately 100 pcm).
Wh:ch Unit will have the higher steady state startup rete?
b)
At 501 powerr a control cd(100 pcm) drops.
Assuiing NO EUNBACK or OPERATOR ACTION, whien Unit will have the lower steady state T a v g ')
GUESTION 1.07 (2.00)
Refer to figure t 363r discus; the two factors which affect the Axial Power Distribution at EOL cnd result an a relatively flat couer distribution.
GUEST. ION 1.09 (2.00)
EXPLAIN the two bases for control "od bank. ;verlsp.
%Cv '
QUESTION 1.07 (CO )
Eupla.n two reasons why the acmputec moderator-only power d e '^ e c t unde.
ectimates the ectual defe;"..se to redistrib> tion at BOLr 0:suming a calculational value of MTC of
-3 ict/deg.r. : 'r' ". Cb 1200 ;:p c ]
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- CAT 7PO:' 01 CONTINUED ON NEXT RAGE *'* * > * )
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' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE.
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1 TOUESTION 1.10 (2.00) i Referring to Figures S-27A and S-279, answer the followin3 questions
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'concerning a losc of-one RCP transient from 22% power without a reactor j
trip.-
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a.
Why does loop 2 RCS flow increase at point 3?
(0.5) b.
Why does loop 1 RCS flow increase at point 2?
(0.5) i c.
Why does-loop 2 S/G 1evel decrease (Shrink) at point 4?
(0,5) d.
Why does_auctioneered high Tave (operating loop) increase et f
point S?
(0,5)
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t QUESTION 1.11 (1.00) f If reactor power increases from 50% to_100%, the average flux in the
. reactor doubles.
With this average flux increase, why does the rod i
. worth remain essentially constant, i
i GUESTION 1.12 (1.00) 2 What are the two reasons far shifting the SI mode froe cold les recirculation to hot les recirculation approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after e LOCA?
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(**xxx END OF CATEGORY 01
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PLANT DESIGN INCLUDING SAFETY AND' EMERGENCY SYSTEMS CAGE 5
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GUESTION 2.01
($.00)
Which of the following statements concerning the automatic isolatien of component cooling water RCP thermal barrier flow it correct?
a.
High flow sensed in an individual thernal barrier return j
line will automatically close its individual return line j
icolation valve, j
b.
High flow sensed in an Individual thernal barrier return line will automatically close the combined return line isolation valve.
c.
H i;3 h #10w sensed in'the combined thermal barrier return line will autonctically close all three individual return j
line isolation v21ves.
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d.
High flow sensed in the combined thernal barrier return i
line will automatically close the' combined return line isolation valve.
t DUESTION 2.02 (1.00) l I
l Which of the'following statements correctly describes the RHR System lineup l
.when HOT LEG RECIRCULATION is esteblished?
(assume both trains of RHR are available)
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a.
Both trains are used to supply hat les recirculation exclusively.
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b.
One train is used to-supply hot les recirculation and the other
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train is used to continue cold les recirculation.
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c.
E: o t h trains supply both hot and cold les recire simultanecesly.
5 d.
One train is used for hot les recire and the other train is put in ttandby (ie. recirculates from RHR HXer outlet to RHR pump inlet).
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CATEGORY 02 CONTINUED ON MEXT PAGE ** tex)
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6
i 0UESTION 2.03 (1.00)
Which of the following.flowpaths describing new power is normally 3
supplied to o typical vital instrument bus is correct?
a.-
480 VAC from vital bus, rectified to 125 VDC, inverted to 120 VAC, and supplied tc instrument bus, b.
480 VAC from vital bus, transformed to 120 VAC, and supplied te instrument bus, c.
125 VDC from battery, supplied to battery bus, inverted to 120 VACf and supplied to instrument bus.
d.
480 VAC Pron vital bus, rectified to 120 VDC, and supplied i
to instrument bus.
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QUESTION' 2.04' (1.00) f -
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Fill in the blanks in the statement below regarding the Standby Steam Generator Feedwater Pumps (SSGFP):
Besides be:ng used during startup and shutdewn, the SSGFPs are also used as c(n) __________.
These pumps cre located adjacent to the
[
__________ tan. and take e suction on the __________ tank.
Flow-to
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o the S/Gs is controlled by the valve (s).
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OUESTION 2.05 (1.50) t The following questions deal with the Intake Cooling Water 'ysten (ICWS).
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c.
What are the THREE loads supplied by tile ICMS?
Dc NOT :nclude l
redundant loads.
(0.5) b.
What valve in the ICWS is directly effected by a safety injection c:gnalo Include in your answer, how and why it
)
is effected.
(1.0) i.
1 QUESTION 2.06 (1.50) 1i I
List the 6 essential safety related loads served by the Component Cooling I
Weter System (redundant components like CCW hect exchangerc are one load).
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PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEh5 PAGE 7
GUESTION 2.07 (1.00)
List the 4 sets of ECCS related valves required to mitigate a LOCA which t
have their control power breakers racked out durins critical operations.
t OUESTION 2,09 (1.50) l a.
Refer to figure 1398 Heater Power Distribution.
State the purpose of the Current Transformers (CT's) wrapped on the heater supply lines.
1 b.
How are small electrical grounds (e.g. scounds which do not trip
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breakers) detected / indicated on Pressurizer' Heater strings?
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How is a bus stripping signal to Backup Heater Group B Pressuriner heaters overriden? (include location of controls) i
. 0UESTION 2.09 (1.50) a.
Refer to figures $397 1 and 4397.2 Reactor Coelant Pumps, LIST all interlocks which must be met to start an Oil Lift Pump (OLP).
E0.53 l
b.-RCP 3A start sequence is in progress (i.e.
OLP control switch to neutral j
and OLP breaker closed) with ' 400 psis oil pressure.
Prior to the l
expiration of the 120 see timer the signal from 'and" gate i fails open (no signal
- 0').
Explain the resultant affect on the completion of the RCP start sequence.
El.03 I
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l ODESTION 2.10 (1.50) i l
a.
If EDG
'A' is tied onto the 4KV bus with greater than ninimum leading and the operator places the DG voltage ad ust switch in the 'rcise' e
direction causing DG current to decrease, would tne machine be operatins in the lead or las mode?
E0.53 i
b, By using figures 6404.1 and 404.2 determine any actions a bus "ault on 3A E 39 would have on the EDG including start /stop of EDG and E0.53 i
breaker :nterlocks.(Assume all control switches are in normal positicos) c.
What it the consequences of r failure of the soak back oil pump during EDG speed reductior., from rated speed to shutdown conditions?
E0.52 3
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- CATEGORY O2 CONTINUED ON NEXT PAGE
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2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE S
i OUESTION 2.11 (1.50) a.
Which system supplies the water flow for the Emergency Containment Filter System (ECFS) dovsins?
1 b.
Once air flow has been restored to the ECFS is it possible te shutdown
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the dousing system fron the' control room?
j c.
The drain line for the ECFS has an installed loop seal. What adverse j
conditions would occur during fan operation as a result of the loop seal beins drained?
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OUESTION 2.12 (2.00)
For eacn of the followins abnormal plant conditions state whether or not i
AFW pump 3A would inject into the Stean Generators et rated conditions AND j
for each condition in which the AFWP does inject state all adverse consequences of the given condition (assume e valid start signal occurs).
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Consider each case seperately.
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a.
The discharge flow transmitter on AFWP
'A' has a failed "high* high j
pressure tap. (Prior to start) i I
b.
SG 39. supply to Train 1 (valve 3-1404)-fails closed.
I c.
AFWP 3A Trip a Throttle valve trip solenoid hat an open circuit.
d.
AFWP 3A Trip i Throttle valve connecting rod spring is detached from I
the rod.
e.
The main oil punp shaft fails.
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i GUESTION 2.13 (1.00) i l
Refer to figure 1403 NIS Power Supply. On an operatins unit inverter supply l
to 3P07 has been transferred to INV AS fron, INU 3A.
Prior to the transfer j
j INV AS output vcltage was greater than that of INV 3Ar EXPLAIN why this i
higher voltase would have no effect on the operation of the Power Range j
NI instrument power.
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l GUESTION 2.14 (1.00)
Refer to figure 4309 '8asic CVCS Flow Balance *.
or points 1 - 5 state the j
design values for " low (spn) at each point.
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END OF CATEGORY 02
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INSTRUMENTS AND CONTROLS PAGE 9
OUESTION 3.01 (1.00)
Of the following signals that de-energine the diesel
- Ready To Start' light, which is also interlocked to prevent the diesel from starting?
a.
Low Starting Air Pressure.
i b.
Key Switch in Bypass c.
Low Skid Tank Level l
l d.
Low Prelvbe Oil Pressure e.
Low Prelvbe Oil Temperature b
RUESTION 3.02 (1 00) 5 l'
At 100% pcwer, with the steam dump control system in the Tave moder a 10%
l step loss of load occurs.
Assuming no reactor trip occurs, the condenser l
is available, and the reactor operator manually operates the control tads, which of the following would occur if Bank 1 (CV-2827 and CV-2828).
steam dump valves failed to open?
f a.
Denk 2 (CV-2829 & CV-2830) would open.
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b.
Atmospheric dumps would open.
t c.-
S/G safeties would open.
d.
No other steam valves would open.
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GUESTION 3.03 (1.00) i i
Which one of the following conditions will result in a reactor trip?
j a.
Low flow on 2/3 detectors in 1/3 loops when o-7 and < P-8.
b.
Low flow on 2/3 detectors in 1/3 loops when P-8.
c.
Low flow on 1/3 detectors in 2/3 loops when < P-7 end P-8.
d.
Low flow on 1/3 detectors in 2/3 loops when / P-S.
f 1
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(r****
CATEGORY 03 CONTINL'ED ON NEXT PAGE
- xx*)
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- 3.
INSTRUMENTS AND CONTROLS PAGE 10.
GUESTION
~3 04 (1.50) i Indicate whether the following will cause a Rod Centrol System URGENT or-NON-URGENT FAILURE alata.
a)
Loss of Main 100 VOC' Power in.the Logic Cabinet b). Slave Cycler fails to start counting upon receipt of
'Go*
pulses c)
Loose circuit card in the Power Cabinet 1s ]5'* up tV QUESTION 3.05
( do )
Unit 3 is operating at full powere all control systems in auto, Pressurizer level channel 1.(LT 459) is selected.
Answer the following*
a.
Would a conmon failure on LT d61 and LT 462 causing both to read maximum result in a reactor trip? (yes/no) 6.
The reference leg for LT 459 PRZR level develops a significant packing
- leak causing the reference les to drein.
Stete all the auto actions ~
i affecting the pressuriner as a result of this nalfunction.
c.
List all functions of Pressurizer pts 455, 456 and 457.
-GUESTION 3.06 (2.50) a.
The charging pumps have two ' Low Speed Stop' settings.
Disci ss the modes of operation in which these stops are applicable AND how their values differ, b.
Refer to figure 1425 'VCT/ Holdup Tank Diverting Valve (LCV-115A?'r EXPLAIN the signals which will operate SV 115A-1 and SV 115-2 in both the VCT and Avtc Divert positions.
p
-(**ttr CATEGORY 03 CONTINUED ON HEXT PAGE
- )
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3.
INSTRUMENTS'AND CONTROLS PAGE 11 1
1 i
s.
v j
GUESTION _3.07 (2.50) a.
List all inputs to the Steam Generator Level / flow controller FC 478,
[
in 'autonatic?
b.
E::actly how.does Steam Generator level flow controller FC 478 operate
-the. Main Feedwater Control
'J a l v e ?
Include motive force (s) and si~gnal conditioning. equipment.
f c.
There are three solenoid operated vent valves associated with the MFWCVs l:
' with time delays of Sr5,and 20. sect.
STATE the. interlocks associated with these valves AND whether the valves energine/deenergize;f,o ticip?
j I
i QUESTION 3.08 (2.00) 1 l
a.
For the Unit 3-Rod Position Indication (RPI) power supplyr describe.
i
. theisupply to the DC power sys, ems including both' normal inverter; supply I
and alternate supply.
Include all automatic features associated with'
{
these two sources.
. o l
b.LThe Rod Devistion Monitor generates a ' Rod in Motion' signal.
EXPLAIN j
the basis'for this-signal AND the actions which are initiated by the i
signal.
I t.
i l-
.0UESTION 3.09 (1.00)
Assune the following CCW, pump alignments:
S 8 C pumps are racked in and in AUTO S pump is running A pump is racked out ond open j
a.
If
'B' pump-received an over current lockout what signal (s) would gujo start
'C' pump AND.et what time would it start?
'i, b.
Would the events in
'a' above be different with the local / remote switch for
'C' in ' local'?
i OUESTION 3.10 (1.50) d i
For the CCW temperature C4
'_4 PLAIN how the input for temperature control in AUTO -is derived.
s l
(*****. CATEGORY 03 CONTINUED ON NEXT PAGE
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i3. - : INSTRUMENTS AND CONTROLS
'PAGE 12 4
l GUESTION 3.11-(2.00).
Refer to figure 4420 ' Emergency Containment Cooler Fan" and answer the followins:
a, For.the mattir positions marked 1 - 5 what position should the valve e
be in'for the given control switch position?
i b.
If the fan switch is-in.the STOP position and a sequencer start signal
. is received,-HOW should-the fant outlet valve and inlet / outlet-bypass l.
valves respond?
!.3-l~
b q
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(***o END OF CATEGORY 03 ****x) i E
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4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13 i~
RADIOLOGICAL CONTROL
{
QUESTION
.4.01 (1.00)
OP 0202.1 ' Reactor Startup...'r cautions the operator not to exceed 2000 psi until Cteem Gener ator pressures are > 585 psis.
Explain why this caution is'necess*rf GUESTION 4.02 (1.00)
HOP 0205.1 ' Unit Shutdown...' cautions the op:*ator not to pull control power fuses on the Source Range channel (s) in the event the channels do not-4 reenersize after the P-6 defea.t buttons'are depressed.
Explain the basis for this caution.
c QUESTION 4.03 (1.00)
I What conditions would allow a Tenporary or Shcrt Tern Operator Turnover versus a formal Shift Turnover?
Include as a part of your. answer any time restrictions which may apply and d o.c u m e nt a t i a n which nay be required.
DUESTION 4.04 (1.00) r l
Which of the following statements regardins plant clearances is correct?
a.
When executing a cleetance order, each operation 'nust be conducted in the order in which it was given.
- b. When returning systens to service, the electrical switching is done first.
t I
- c. Writter permission from all clearance holders must be obtained before performing e temporary lift for test purposes.
d.
If conducting a clearance in a contaminated area, the or13inal copy of the clearance is left at the entrance to the area and is initialed appropriately by the operaten upon e: it.
(*****
CATEGORY 04 CONTINUED ON NEXT DAGE
- )
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'4.
-PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE ~ 14
~~~~~---------
~~~~55015E55iE5t 56NTE5L GUESTION 4.05 (1.00)
Which of the following statements describes the correct usage of the-Critical Safety Function (CSF) Status Trees while performing E0P-E ' Reactor Trip or Safety Injection'?
a.
The CSF Status Trees are ONLY monitored when E0P-E-0 directs.
b.
Awareness of Red Path Cr.iteria is required at all tinese but the CSF Status Trees are monitored only after it is determined that SI can NOT be terminated, c.' Monitoring of the CSF Status Trees. commences as soon as the immediate action steps are completed.
d.
CSF Status 1rees are required to be monitored as soon as the proceduce is entered and a valid SI is determined to have occurred.
-0UESTION a.06-(1.00)
A hydrogen bubble formed in the reactor vessel is eliminated by a.
increasing pressuriner temperature above core thermocouple readings.
- b.. injecting oxygen into the reactor coolant system via the c h e nii c a l and volume control system.
c.
maximizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 spmy 3nd placing the Cation I
bed demineralizer in service in parallel with the mixed bed demineralizer.
d.
venting the reactor vessel head.
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14.
PROCEDURES --NORMAL, ABNORMAL, EMERGENCY AND PAGE 15.
!~
RADICLOGICAL CONTROL I-i i
t' GUESTION 4.07 (1.00).
i
't.
Which-of the'followins indications require SI reinitiation following a spurious SI that has been secured?
I a.
RCS pressure at 1950 psig.
i i
b.
RCS.subcooling at 25 destees C.
F c.
Pressuriner level at 17%.
i d.
All steam generator levels at 15%.
i OUESTION 4.08 (1.00)
Indicate whether the following conditions will place you in a CSF RED PATH or NOT:
a)
Containment pressure of 37 psis.
b)
All 3 S/Gt at 25% level and AFW Flow of 300 3pm.-
i
'0UESTION 4.0?
(.50) l
.Will' adverse containment conditions cause affected control room indications j
to indicate higher or lower than actual conditions?
QUESTION 4'.10
(.50)
Fill-in-the-Blanks'
[
In accordance with DNOP-12308 ' Power Range NI Malfunction', the Power Range I
meters everase reading must be within ______%
at greater then or equal to i
90% reactor power and at less than 90% reactor power must be within _____%.
QUESTION 4.11 (1.00)
What are the shutoff head pressures listed ;n E0P-E-0 for both the HMSI and LHSI pumps below which SI flow needs to be verified?
(*****
CATEGORY 04 CONTINUE 9 ON MEXT PAGE
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4.
PROCEDURES - NORMAL, ADNDRMALr EhERGENCY AND PAGE 16
~~~~EIDE5L5EiEst E5sTR5E-------~~~~~~~~~~~~~~-~~
GUESTION 4,12 (1.00)
State how the Normal Containment Coolers are re-started after a safeguard system initiation signal is received.
.(Address sequence of. switch manipuistions)
GUESTION 4.13
(.50)
In accordance with 0-OP-023 ' Emergency Diesel Genertor's how is the start failure relay reset?
00ESTION 4.14 (1.00)
What plant condition is assoned in the valve / breaker ' Normal" position designation for the Operating Frocedure alignment sheets?
- QUESTION 4.15 (1.00)
During implementation of ECA-0.0 " Loss of all A/C Power *, a red path CSF on containment occurs.
Which procedure should the operator perform?
EXPLAIN.
GUESTION 4.16 (1.00)
FRP-P.1 ' Response to Imminent Pressuriced Thermal Shock", has the operator l'
check for SI termination criteria relatively early in the procedure and with less restrictive conditions than in the EPs.
Give TWO bases for securing SI early into this procedure.
QUESTION 4.17 (1.00)
Define *Inside the Target Sand" in accordance with DNCP-1008.6 ' Operations outside the Target Band'.
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4.
PROCEDURES'- NORMALr.ADNDRMALr EMERGENCY AND PAGE 17 RADIOLOGICAL CONTROL j
r i t.
i QUESTION 4.18 (1,50)
'After Natural-Circulation has been established, what 3 indications are t
monitored to determine RCS COOLDOWN, according to ES-0+2,
" Natural I-Circulation Cocidown
t QUESTION 4.19 (1.00)
I j
In accordance with.3-OP-075 'AFW system', outline the flow paths for AFW i
Train i during Dual Unit operation to ensure compliance with the require-ments of Tech Spec 3.8.
Include in your discussion suction, discharger steam-supply flow paths and valves.
i
?
I
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(*xx**
END OF CATEGORY 04
- )
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yo3 FIGURE
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t BASIC CVCS FLOW BALANCE l
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FIGURE
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EMERGENCY CONTAINMENT COOLER FAN EMERGENCY CONTAINMENT INLET OUTLET B P SS COOLER FAN (A, B OR C)
(2903,2904, (2008, 2907, (2810 2812 SWITCH POSITION P
OR 2005)
OR 2008)
OR 2814)
AUTO-FAN NOT RUNNING y7; l
(NORMAL CONDITION) 4..
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p AUTO-FAN STARTS OPEN OPEN OPEN HELD IN ST#3T OPEN OPEN
. Yo STOP CLOSED CLOSED
- r. _ _
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_ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _' 0 W E R - P L A N T O P E R A T I O N,___-___________-___-_
1.
PRINCIPLES OF NUCLEAR C PAGE 18 THERM 00YNAMICSr 9 EAT TRANSFER AND CLUID FLOW ANSWERS -- T'JRKEY POINT 3&4
-86/09/08-CASTOR C ANSWER 1.01 (2.00) a)
Delta T will increase
(+,7) as That goes up due to boiling in the core and Tecid remains fairly conciant
(+.3) 4 b)
Tcold will not follow Pstm
(+.7) as Pstn decreases due to boiling off in the S/G and Te remains fairly constant
(+.3)
REFERENCE Westinghcuse Mitigating Core Damager pp 1.15/16 EPE-017; EK1.01(4.4/4.6)
I i
ANSWER 1.02 (1.00) e REFERENCE NUSr Vol 4r pp G-8 Nuclear T' w r Plant O p e r.3 t o r Trng Prgm, HTFF and Thermo, Sect ZE CNTDr ' T her nal/ Hydr aulic Prineples and Applications, II'r pp 10-43/44 App A: Pumps /Centrtfugal, Operating at Runout Conditions (3.1/3.1) 1 C
JeO ANSWER 1.03 (M
DN
[ (-
i w p -og (0A.i c-l REFERENCE Comprehensive Nuclear Training Operations (CNTO)r pp u-16/27 001/0007 K 5.13 '. 'J 7 / 4. 0 )
ANSWER 1.04 (1.00) 1.
normal 2.
thinble 3.
thimble 4
normel 5.
normal l
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PRINCIPLES OF NUCLEAR POWER PLANT DPERATION, PAGE 19 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW i
ANSWERS -- TURXEY POINT 3&4
-86/09/08-CASTD, C L
l REFERENCE TPT. CNTO Vol II 13-27 i
l ANSWER 1.05 (1.00)
[-
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Larger b.
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ANSWER 1.06 (2.00) a)
Unit 4
(+.5) due to a lower Beta coefficient at EOL
(+.5) l b)
Unit 3
(+.5) due to MTC being less negative, so Tcv3 must i
decrease come to add + r e a c t i v :. t y )
(+,5)
REFERENCE CNTO ' Reactor Core Control'r pp 3-21 1 ' Fundamentals of Nuclear Reactor Physics *r pp 7-31 001/0001 K5.49(2.9/3.4) E K5.10(3.9/4.1) l ANSWER 1.07 (2.00) i
-More Negetive MTC EO.52 fuel depletion at core bottom CO.SJ fuel depletion is due to the BOL fiv: distribution f l u:. peak at uppar core at EOL E0.5]
i The negative HTC produces more negative teattivity in the top of the core at EOL than at EOL.
a p g/ g f () g o.52 REFERENCE i' g.( e TPT CNTO Core Centro 8 - 2 *.
/t Cyc-c.aqL4 7 w, &r l.rt. %i$
~.c*%
l sq eg. gun.LLLy, ~J g.
r i
1 l
i i
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.?
P R.. T N.{t 01_ C_ C p ;- y t f r i C_ a. ?
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_; f* 3 j y t.tC Q u,[.i. n v y tz g;s t
+ Trq
- ; p 3. T
. RV a l' 1 p = 3; q pi er
~
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~
1 a 'r ANSWE93 -- TUR EY POINT 3Pa
-96/09/02-CASTO., '.,
wepCp 4
no
<z +nn)
...w eu z s bGnk OV9P1]p 15 US9d tO PIOVid" s3 DOP9
'n i #C " F:
diffGPent121 CRW 2nd 3 "009
'irii f 0 P B 3: J1 nUUtPOP f20 dif t r ibUllCT dUring COntPCl P C C.' nE'n9UVOPE [ h eE n o re-U n i f o r m aulcl
- 10:
distPlbu', ton coUld caUse abnormally high power i
ie i E.5 In the COP 9 P t'2 5 0 l + 1 * * ; in fuel dODage.((D2A Unlf0P3 l t f 0 " O "f t 10 ) Worth
'l e
r r
i always ensOPus ' ii J t rod U.otion OPoduc95 a change ;n Te'?ctlfity [7.O]
& pLI p L_ C.E.' p C_ ?l (' C M
(tTh 9 tutu tai MS&vtf tot mJ T. O L~ o n e L, a n, P o. 6 b,
L z
'l
- M
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b ANowE -
1.09 c coi (c 4
The MrC is terperature d ";~ e a d.. r n 2nd '. t does not change 11: a: 'y.
At HFP MTC has c s 10e of O,c c r cea.-
ct th-on t e.nicL (557> :: n.
-1.
.er' dea,;
u at the c a t-OUt.e*
/617' For <' a l c u l a t i o n.31
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-1
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N G f.3 9 " ? t C T t O '*! D O P D b O P C Of OC dug, I 1"
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'0100 IC E h d' 1 1 e 7 t h E. n thO Q P 0 ? V O, e MTC
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'. E ' ' 3 / fj 9 G.a.C.
e O ' [ftt h O r $ 5 P D i i.
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D.
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(dU9 10 ' 15i ' 0 t.R 1 COFF f 1 C l.: )
(0.5)
NO 10nler
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- i O 3 t intn
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w Of U E' t 9 ? ' I. P u PI 1n S[b (0*b) d.
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.1.
PRINCIPLES OF NUCLEAR POWER PLANT' OPERATION, PAGE 21 i
~~~~ 5ER556isd55657~5EST TR5U5fER~5UE~ FLU 56~ELUU
~
~
T ANSWERS -- TURKEY POINT 3R4
-86/09/08-CASTD, C-r ANSWER 1.11 (1.00) i Because rod worth is proportional to the ratio of the f'.ux at the tip f
of the rod to the average flon (0.7).
A change in power does r.ot j
significantly change this ratio (0.3).
(1.0)
~
t i
l REFERENCE j'
' Turkey Point, Reactor Core Control, Chapter 6 1
001/0001 K5.02(2.9/3.4) 4 i'
ANSWER 1.12 (1.00) remove boric acid that is precipitated.on upper core surfaces CO.53 terminate any hoiling or steam formation in upper head region [0.52 REFERENCE TPT SD-21 'ECCS* pp 26 1
I l
i 1
i I
i r
\\
l I
r i
r h
)
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 22 ANSWERS -- TURKEY POINT 384
-86/09/08-CASTOR C ANSWER 2.01 (1.00) d REFERENCE Turkey Point, Requal Lesson Plan, RCPr Fig. 16 003/000; K1.12(3.0/3.3)
ANSWER 2.02 (1.00)
- 1. < s Y'
c V,,
REFERENCE TPT E0P E-1.Er 1.3r 1.4 006/020; A4.02(3.9/3.8)
' ANSWER 2.03 (1.00) a REFERENCE VEGP, Training Text, V o l u n. e 8, p.
16b-3 and Fig. 16b-1 d 6 VCS, GS-2r Safesucrds Power Systemt pp. 27 E 2E TPT DWG 5610-T-E-1592 l
062/000; K4.09(2.4/2.9)
ANSWER 2.04 (1.00) i B a c!< u p t o AFW; Un:t 4 Condensate; Demin Water Storasei CWRV Bypass (&.25ea)
REFERENCE TPT Requal Cycle IV-1985, Oay 3 TFT SD*12 ' Condensate and Main Feedwater'. pp 6/7 05?/000i K1.03(3.1/3.3) l
i i
4
}
2.
PLANT DESIGN INCLUDING SACETY AND EMERGENCY SYSTEMS PAGE 23.
)
ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTOR C
+
ANSWER 2.05 (1.50) i t
a.
1.-
CCW HEXs (3)
-(0.2) 2.
TPCW HEXs (2)
(0.2) 3.
Lube Water System (0.1) b.
1.
CV-2201 (TPCW HEX 00+1et Temp Control Valve)
(0.4)
Clones (0.3) i 3.
Allow all ICW to be directed to CCW HEXs (for cooling of
('
vital safeguard equipment)
(0.3)
I REFERENCE I
T u r '<.e y P o i n t, Requal Lessco Plan, ICWSr pp. a8 10 1
076/000i M1.01(3 4/3.3) cnd EPE-026; EM3.02(3.6/3.9) 1^
4 1
1 ANSWER 2.06
(?.50)
J 1)
RHP H/ers
(+
25 ee) 1 2)
RHR Pump seal cooleru l
3)
SI Pump oil coolers and secls j
a) c~
tm f ass /
- r/ ^ S't$-Y&
5)
CS Pump soci coolers 6)
Emerjency Containment Coolers REFERENCE l
TPT SD40 *CCW'r pp 9 008/000; M1.02(3.3/3.4)
ANSWER 2,07 (1.00) 1)
862 A cod B (RWST to RHF; (4.25 ee for any 4) 2)
864 A and 8 (RWST isolation) 3)
065 A, O and C (Accumul: tor Icalation) 4)
066 A and S (GI
o t Leg Injection) 5)
863 A and D (R4R to HMSI/CS pump suctions)
REFERENCE TPT SD21 "ECCS*r pp 36; "T
T3 3.4.1.a.7 006/000; K4.00(3.6/3.7) l t
4
y-----
. ~. -
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS c' A G E 24 ANSWERS -- TURKEY POINT 3E4
-86/09/08-CAST 07 C
/
69//
'.. /
<g,
ANSWER 2.08 (1.50) h 9-'3 h. CT's
/4 protectit,o.la "t'wlica bll',t '
a are used for overcurrent b.
C- :.& e900
'n'"
r*
_ -Mped QM c
- 3. c; w u c :.. u -. _ u.
ev2. me L _c w
phase selecto p i.
c.
n-h j -
. s. -
- 4..2aii
,2 c a un
.une uree ei-v..#.
- s, a
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q-m st eme.
_1
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' L
.s +.2 bvT
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,s L i sd
/ cm [o ' b(9 -' ' ' r'<
w1i6
- 4'4 4 O E0.5 2 4N A 4 '* # t M
l f O # L 2
(/
ea & OL.<,e wwJ n.1 ; c. f, on [o. 23) c' REFERENCE TPT C0-9 h> Ud ANSWER 2.0?
(1.50)
_ y,, py i.
g r,. / rr i
- a. OLP CG start and breaker closed E 0 + 2 5 e a ) c/t. ( h<
' /
b.
The CCP will not stcrt due t"
stcppins of the CLP (contactor open) the not gate will have a
'O' sig al in and a
'l' out thus tripping open the contactor this will result in c locc on signol(s) to the
'and' late prtor to the 120 see timer. E1.02 REFERENCE TPT SD-0 BCPS ANSWER 2.10 (1.50) 3.
Lead E0.52 b.
EDG would start CO.252 breaker vould be locked out E0.252.
c.
Sufficient oil prescure would not te developed for coact down due to sheft driven pump coastdown, specificallyr possible demase to the turbocharser.
CO 52 REFERENCE TPT UD 107 ANSWER 2.11 (1.50' a.
Yes C,
Air drJun into the filter tySten r' duces the effici:.>ncy for removing radicitatopes.
CO.S ti a. 2
2.
PLANT DESIGN INCLUDING GAFETY AND IMERGENCY SYSTEMS PAGE 25 ANSWERS -- TURKEY POINT 3&4
-86/07/08-CASTO, C REFERENCE TPT ED 27 ANSWEP 2 12 (2.00) 3.
would not inject E0.25]
b.
would inject loss of r e d u n d e ric y for steam supply E0.52 c.
would inject no electrical trippin2 c3Pability CO.5J d.
would inject potsible fcilure to mechnically overtpeed trip and locally e.
.d not injuctF w. Il /s) ^ t sw F. L I'**'< b '
53 I' O REFERENCE TPT SD 117 ANSWER 2.13 (1.00)
The i n s t,r u m e n t power is supplied through a SOLA trans"ormer.
The function of the SOLA transformer it to provide a conttent voltage tupply.
- 7. s t - S c
<f = 4 /
- ANSWEP 2.14 (1.00)
/
)
4<<
/,le S /7 e
F
.g fl~ j*f^^'*
p,, fw 60 spn q, q,.j f,
- a, F
.r
+
2.
69 gpn t
3.
24 3pm a.
8 v,m 5.{p m an 2 5ys'as ff 'v Ynu
' ' ' ' ~ ' ' ~ ' '
5.
REFEFENCE TPT SD-12 1
l i
t i
l t
I l
l
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.. ~....
i 1
t-p i
f I
i i
3 1
I i
3.
INSTRI.
PAGE 26
__ _ _ _ __ ___'M E N T S AN D C ON T ROLS ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTO, C i
1 i
i.
i ANSWER 3.01
- (1.00) l I
j, c
REFERENCE Turkey Point, Drawing 5610-T-L1r Sheets 9A and 98 064/000; K4.05(2.8/3.2) i i
ANSWER 3.02 (1 00) d i
i REFERENCE
[
Turkey Point, SDCS SDr pp. 2&?
t l
041/020; K4.14(2.5/2.S) l ANSWER 3.03 (1.00) 1 b.
i
(
REFERENCE j
23 i
i i.
ANSWER 3.04 (1.50) i I
.a)
Non-Urgent
(+.5 ea) b).
Urgent c)
Ursent f
REFERENCE I
j TPT SD5 ' Rod Contral Systen"r pp 46-48
,. 001/-0107 K6.05(2.9/3.2) i i
[
i k
4.
,i.
I h
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e l
l 4
4 I
(
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- ~
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..,,----.,nnn.
~
1 3.
INSTRUMENTS AND CONTROLS PAGE 27 ANSWERS -- TURKEY POINT 284
-86/09/09-CASTG, C
/
"L.7 5 ANSWER 3.05
( t - t"0 )
3.
No E0.5]
, y r-b.
Cherging ' low decreases until cetual level decreases isolating letdown.
After letdopo isolates oreucur rer level increases unt:1 at 9 2 *;
t<
< 4 r tri'o 9
<.r+,',
..n _ r,. ere r c~
2..u
_e unt: 1 su;;u
-c M.b. m*W i - twptJa,G7 eg -
>/,
c.
OT delte T trio and runbsets
( <"L <e'k n
cs
.pg, 2n00 os4 PORV i n t e r l o c k 's E^"
ea.]
L.f. Gf kl ltr jw1,wrc,.
L f $ 1..
ud dem n e r r o.s,rdL-r hur-TPT SD 009 ANSt'ER 3.06 (2.50) a.
A Low speed stop prevents the pump fror decreasing below the apeed corresponding te (21 gen) E0.25: in the auto mode E0.252. In manuel E0.25] the pump can be lowereo to the speed equal to
'9.5 gpm) a lower value E0.25:
b.
(LT 112 provides a signal to modulate LCV 115a to caintain the normal band), et 86% level LT 115 will generate e backup signal to trip LCV lt5s to the full holdup tank divert E0.52. Modulation of LCV 1150 rod the backup trip to divert will function only if the two pcsition switch is selected to the " a u t. o - d i v e r t ' position.EO.5]IN the VCT position full letdown flow is oesse to the VCT E0.52
- e r. e r r. 5r. %' c Rs As w
TPT SD 13 ANSWER 3.07 (2,50) a.
Channel III S/G 1evel and is+ stcse pressure E0.252
-or-programmed S/G 1evel Feedwater #10w E0.25] cnd Steam flow E0.252 b.
The output from the FC is converted to air precsu'= by an I/F converter E0.25]. This pneunstic signc! is used by the velve positioner, increasin pressure to open and decreasing pressort to close E0.25]
c.
De-energine E0.52 Hi H2 : team generator i e'r e _ (2/2 nar ow nenSe) 5 SeCE
~
SI 5 secs Rt trin and low T e v c.
55a dea. r 20 sect t
- r. n.,a-
. a..
REFERENCE TPT SD 11
3.
INSTRUMENTS AND CONTROLS PAGE 28 ANSWERS -- TURMEY POINT 3&4
-86/09/08-CASTO, C 4
ANSWER 3.08 (2.00) 4 a.
Inverter is supplied from AC (30766) and DC (3001-49) E0.5J and auto i
transfers to alternate EO.25] MCC
'A' Lts panel ELP 317] E0.25].
b.
When control rods are in motion the trip setpoint of the rod deviation bistable is changed [0.52. The change prevents fals? alarms from spurious signal inputs E0.5].
REFERENCE TPT SD 006 l
ANSWER 3.09 (1.00) 1 a - S.-/7_53
No E0.53 REFERENCE TPT SD 40 i
. ANSWER 3.10 (1.50)
A temperature elenent is instcIled on the A and 9 headers [1.0]. The controller selects the higher of the two TE's for control of CU 2202. E0.52 REFERENCE TPT SD 40 i'
ANSWER 3.11 (2.00) a.
open closed open closed closed I
- b. Fan starts, outlet opens, Inlet / Outlet bypass closed, 2 s' e c l
REFERENCE TPT SD - 40 l
[
l l
[
i.,..
...... _ -.., _.. _.. _,. _. _.. _,... _.... _... _ _.... _. _... _ ~... _. _. _ _ _ _ _ _ _. _. _. _
_. ~. - -
'*ee i
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~ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29 i
l RADIOLOGICAL CONTROL' ANSWERS -- TURMEY POINT 3&4
-86/09/08-CAST 0r C r
4 ANSWER 4.01 (1.00) l An SI will automatically be intitiated by'hish steam line - stean generator diff. pressure if this ic exceeded.
3 i
F REFERENCE TPT OP-0202.1 p.
35'
[
t ANSWER 4.02 (1.00)
I Pulling the control power fuses on 1/2 channels will result in a reactor E
l trip. Ecaf' I
I l
REFERENCF l
TPT-OP 0205.1 p.
12 l
l-I l
ANSWER 4.03 (1.00)
[
- 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> E0.252 An informal turnover must be completed prior.to.the operators _ departure E0,252. Note chanseover in operators los prior to' j
departure E0.252 and up)n return E0.252.
O i-i.
REFERENCE TPT AP 0103.2 pp. 15~(8.2)
I ANSWER 4.04 (1.00) a 1
REFERENCC I
TPT ADM-0103.4, pp 13-15 i-k F
ANSWER 4.05 (1.00) ic 5-/3'Y" l
y REFERENCE E0P-E-0 i
I I
1 I
i
.f I
4 i
L h
r 4
i -
t 4.
PROCEDURES - NORMAL, ADNORMAL, EMERGENCY AND PAGE 30 I"
-~~~EA5i5t55iEAL 55sTE5L
-~~~~~~~~---------------
i ANSWERS -
TURKEY POINT 3&4
-86/09/08-CAST 07 C i
l l
l ANSWER.
4.06 (1.00)
I d
REFERENCE MNS EP/2/A/5000/16.3 i
CNS EP/1//A/5000/2F3r p.7.
l NAPS 1-FRP-I.3Ar p.3.
ANSWER 4.07 (1.00) g 4
l b.
REFERENCE NAPS 1-ES-0.2r_ Foldout page.
l TPT Coldout page 1
l ANSWER 4.08 (1.00)
I i
a)
NO
(+.5 ec)
9 ' ' I -W b) w j^/# lro h..e-h < m st an s
REFERENCE 4
j TPT E0P-F-0, CSF Status Trees PWG-10:
E0 Entt y Level C o n <ii t i o n s (4.1/4.5)
)
)
ANSWER 4.09
(.50)
I Higher REFERENCE I
North Anna EP-0. foldout page j
TPT E-0 ANSWER 4.10 C.50)
- 8.
- i
m.
t g
i i
1 iE 4.
PROCEDURES - NORMAL, ASNORMAL, EMERGENCY AND
? AGE 31 I
b 1
RADICLOGICAL CONTROL f
!~
ANSWERS -- TURKEY POINT 384
-86/09/08-CASTO, C i.
-ANSWER.
4 11 (1.00) e
-HHSI 1590 psig (1500 - 1600)
LHSI 225 pois (150 - 250) j REFERENCE TPT E0P-E-0 p.
6 6
k i
ANSWER 4.12' (1.00)-
1 j
The safegaurd initiation signal must be reset E0.52 and then the NCC reset l
pushbutton must be depresced E0,5].
REFERENCE
(.
'TPT 3-OP-057 citep 4.4 i
-l t
ANSWER 4.13
(.50)
[
By depreccing the 'Alstm Recet end Stop' pushbutton, REFERENCE TPT-OP-023 EDG j
ANSWER 4.14 1*
00) je.,
S'/3-N L_ u_ v.
c., n v..J O W n u
i k
~cc _o chc.
w.
.m o_
TPT ADM 201 ANSWER 4.15 (1.00)
The operator should remain in ECA-0.0 since the FRPs are written on the premise that-at lecct one E-but is energized.
i REFERENCE Wes,. background info. for ICA-0.0 4
t i
i e
l 6
i 4
'I L
P i
(
t I
i I
l r
i
=
4.
- c. :e n_ e._ c e; r.,m-e._ a - Nn_ a.ntm, r.e.m.n em.,e _, r u,r w r,r e r_ v.
- v. m.
- c.,s. c e w
u 4
--- Fs67.5.I(ys;F F F595.5 -~~~~~~~~---~~~~~~~~-~~~
ANSWERS -- TURKEf :'OINT 3?.
- D A / 0 7 / 9 0 - C A S T 'J r C s.y e u r n, 4.4. A
- r. 4. n. n ).
s s
m 1.
SI i c w :. 3 _ ciaenific. + c e n t r : ;u t o r-to a n "i ce!d 1m decreece CC.5:
- t e m.a e r n u r e 2.
It can also be e sign.'inant contributo.- to a overornsure concition i
1 <. + p. c..
Pnc.i c 1 r. 4. a n_ t,
r r>. 9_ '_
m-s Q U t r_ R, C_ t *' c_
L West. b a c !> g r o u n d info. #or r;P-P.1 4h s
/
\\
/
Y (*.00) 13 ANSWER 4.17 channo'm\\cT C' a a n d' ti 2/c
'"41-N44) inuirate incide the defined tarzet "or the respective c h :: n n e l.
nt.r n ym. r e r-c.,, c_
T T' T 0 H n t'
'. n. ".> 0,
s
.m AN9WER 4.18 (1.50)
Core Enit
/C
(+.5 ea) r.
t! u, +s GCS Subcooling
&-. c c. r e r b; r, c
->s-w SGN ES-0.3, p,o 6
T, o r.
- c. e _., enP
., cJ
=
J i
.a s
I C' E () 7,d,
- wf. e C '
- n..O.
[ 9, ~ h /
9 **
us r
s r i i :*
-c r-
, < r P1 M L ;ri t ?
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, a,
e n' f G
'*'r.
1 r,
s? ~,, g,,e {, r,.
e't s
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eo'l a _.1
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L
.sa n _,
m
'E a, c,,
, 3 p.
r-.
a,a.,6 y
,v
.. i a m
c.m SuCtlCn f'er the CST [0.252 2; n ' disch?"OO
- ,0 3/C 1 hr O'!3)
ICV C V - 2 C l o r 23177 and 2318 C O. :' 5 2,
'ne 31, 2 r s o;.p / for thu AFW Pumps whai: be f r o r.
MDV-3-1405 :: n. 1404 CO,252 fS? tr c u w ;l c!:o.nclude c1_ -' i,c i n q cr<_
c E
s' b
$3 b 5 4 Chb.rb be O Y 'b a
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S. NUCLEAR REGULATORY l'OMMISSION SENIOR REACTOR OPERATOR REOUALIFICATION EXAMINATION CACILITY:
TURKEY POINT 3a4 REACTOR TYPE:
PWR-4EC3 OATE ADMINISTERED: 86/09/00 EXAMINER:
CASTO, C CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Read the attacned instruction page carefully.
This exanination replaces the current cycle facility administered requalification examination.
Retraining requirements for failure of this examination are the same as for failure of a requalification exanination prepared and administered by your training staff.
Points for each question are indicated in
- parentheses after the question.
The passins grade requires at least '70%
in each category and a final grade of at least 80%.
Examination papers will be picked up four (4) hours after the examination starts.
% OF CATEGORY
% OF CANDIDATE'S -CATEGORY l
VALUE TOTAL SCORE VALUE CATEGORY f
___1_00____'_"1_0
________ 5.
THEORY OF NUCLEAR POWER PLANT 18
[0_
CPERATION, FLUIDS, AND THERMODYNAMICS 5
1 1
- _I 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 18.00
________ _['"i_00
________ 7 PROCEDUPES - NORMAL, ABNORMAL.
[
i EMERGENCY AND RADIDLOGICAL I
CONTROL
- 0.00 1
"5.00 o
l
________ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 72.00 Totalt j
- - i n a l Grade i-All work done on this examination is my own.
I have neit!'er siven not received aid.
Candidate's Signature i
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination tne following rules apply:
1.
Cheatirf g on the examination means an automatic denial of your application and could; result in more-severe pencities.
2.
Restroom trips are to be limited and only one candidate at a time nay 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 lesible reproductions.
t l
4.
Print your.name in the blank provided on the cover sheet of the j
examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
. Print your name in the upper right-hand corner of the first page of each 7.
section of the answer sheet.
8.
Consecutively number each answer sheetr write "End of Category __
as appropricter start each cete3ory on e new pager write only on one side of the paper-and write 'Last Page' on the last answer sheet.
o.
Number each answer as to category and numbert for exampler 1.4r 6.3.
4 l
- 10. Skip at lecst three lines between each answer, i
- 11. Separate answer sheets from pad and place #inished answer sheets face i
down on your desk or table.
~12. Use abbreviations only if they are commonly used in #acility literature.
j.
- 10. 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.
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- 14. Shou cll c a l c u l a i l o r)s, methodsr or assumptiers used to obtain an answer to mathematical prcblems whether indicated in the question or not.
- 15. ParticI credit may be given.
Thereforer ANSWER ALL PARTS OF THE OUESTION ANO DO NOT LEAVE ANY ANSWER ELANM.
i
- 16. If pcrts of the examination are net c!est as to intent. ask questions of j-l the examiner only.
l i7. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been siven assistance in completing the exanination.
This must be done after the examination nas been completed.
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- 18. When you complete your e:: a m i n a t i o n, you shall:
a.
Assemble your examinction as follows:
(1)
Exam questions on top.
(2)
Exam aids - figurest tables, etc.
(3)
Answer pages includins fisures which are part of the answer.
b.
Turn in your copy of the e:: amination and cil-pages used to answer the examination questions.
L c.
Turn in c11. scrap paper and the balance of the paper that you did not use for answering the questions.
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 proeress, your license may be denied or revoked.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 2
~~~~TUEE5665 NAM5C5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 5 '01 (1.00)
For each of the following, select from the choices in the parentheses, the most restrictive condition for Shutdown Margin requirenents, cssuning Mode l or 2 conditions exist a)
Time in core life (90Lt MOL er EOL) t I
b)
Accident (Rod Ejection, Steamline 9 teak cr-Inadvertant Dilution) i QUESTION 5.02 (1.00) i
!=
A reactor has been operating at full power for three months when a manual reactor trip occurs.
All systems are operatichal and the stean dunps are innediately-placed in the steam pressure control node.
Ten minutes after the reactor trip, all RCPs are tripped. Twenty minutes after the reactor tript Loop i RCP is jogged momentarily. Which set of traces (a - d)'on i
figure 4 174 most closely represents the previously described events?
l i!'
OUESTION 5.03 (1.00)
I i'
For each of the following parameters compare a Core Thimble Cell to a l
Normal Core Cell and indicate which of the two has the higher'value, j
(rated conditions, average values) 1.
Total pouer production 2.
Power density 3.
Enthelphy rise 4.
Mass flow 5.
(*****
CATEGORY 05 CONTINUED ON NEXT PACE kvr**)
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LTHEORY Or NUCLEAR POWER PLANT OPERATIONr FLUIDS, AND PACE 3
t i
t DUESTION 5.04 (1.50) l i
Unit 4 is at 50% power with control rods in MANUAL when~the turbine is ramped up to 60%.
' Indicate whether the parameters below will increaser decrease or remain the same during both the initial response (first 30 seconds of the transient) and after turbine power has stabililted relative l
to the initial conditions. (Assume the following: No changes to baron /nenon Loop transport t'.me is 10 seconds l
j.
No operator actions) l NOTE: No answer required where it is already filled in below.
l Initial Response Steady State l
1 i
i 3)
S/G pressure NO ANSWER RORD b)
Reactor Power NO ANSWER RORD i
c)
Teold' l
d)
Tave QUESTION 5.05
(.50) t TRUE or FALSE.During cold plant conditions, you would expect the COLD
[
calibrated PZR level instrument to indicate HIGHER than the HOT i
calibrated level instrument.
(0,5) t i
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QUESTION 5.06 (1.00) 1 l
Restrictions on F(G)(T) limit include penalties due to uncover and reficod rates post-LOCA, tup tin one other basis #ct core height restrictions I
on F(G)(T).
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L QUESTION 5.07 (2,00) i t
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Refer to figure 1 363, discuss the two factors which sffect the Anial Power Di s tr ibuti or. a t EOL and result in a relatively flat power distribution.
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GUESTION 5.00
-(2.00) r EXPLAIN the two bases for control rod bank everlap.
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(*****
CATEGORY 05 CONTINUED ON NEXT DAGE
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5.
THEORY OF' NUCLEAR POWER PLANT OPERATION, CLUIDSr AND PAGE 4
OUESTION 5.07 (2.00)
Explain two reasons why the computed' moderator-only power defect under-estimates the actual defeetr due to redistribution et SQL, assuming a calculational value of MTC of
-4 pcm/ des.F. EHFPr Cb = 1200 ppm 2 GUEST 10N 5.10 (1.00) a.-Enplain why decreasing-the H20/U ratio (moderator / fuel) results in a reduction of the resonance esecpe probablity.
b ~. Explain why decreasing the H20/U ratio increases the value of
'f' in the six-factor formule.
QUESTION 5.11 (2.00)
A rod drops and sticks at the core c:id-position from full power conditions with all rods out.
A Reactor Trip does not occur.
If this condition were t o. p e r s'i s t for an entended. period of time (well be/ond T/S-limits), what will be the effect on the Excore Axial Offset of the Power Rense NI for the quadrant in which the dropped / stuck rod occurs <
Include a discussion of xenon effects and a definition of anici offset.
QUESTION 5.12 (1.00)
TG 3.2.2-requires power be reduced to ' 75% if a misaligned control rod w
cannot be aligned within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
What is the basis for reducing power in this situation?
QUESTION 5.12 (2.00)
Indicate how-the parameters below respond and the reason uhy they react that wayr on a loss cf natural circulation followins a trip fece 100%
equilibrium conditions; c)
RCS wide ranse Delta T i..
b)
Difference between T(cold) and P(steam) 1 (xxxxx END OF CATEGORY 05 r****)
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PLANT SYSTEMS DESIGN 7 CONTROL, AND INSTRUMENTATION PAGE 5
4 i
i GUESTION 6.01 (1.00)
Which of the followins statenents concerning the autenatic isolation i
of component cooling ucter RCP thermal barrier flow is correct?
l a.
High flow sensed in an individual. thermal barrier return l.
j line will autonatically close its individual return line l
l isolation valve.
i i-b.
High flow sensed in an individual thermal barrier return j.
line will 'automaticelly close the combined return line
{
isolation vc1ve.
l c.
High flow sensed in the combined thermal barrier return
[
line will automatically close all three individual return l
line isolation valves.
l l'
l d.
High flow sensed in the combined thermal berrier return j
line will automatically close the conbined return line isolation valve.
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i 00ESTION 6.02 (1.00)
Which of the following statements correctly describes the RHR System lineup l
l uhen HOT LEG RECIRCULATION is established?
Cassume both trains of RHR sre available) a.
Both trains are used to supply hot les recirculation exclusively.
b.
One train.is used-to supply hot les recirculation and the other t r a f. n is used to' continue cold les recirculation.
c.
Both treins supply both hot and cold les recirc simultaneously.
d.
One train is used for het les recirc and the other train :s put in standby (ie. ree:rculates from RHR HXer outlet te RHR pump inlet).
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(*****
C A T EG O R'f 06 CONTINUED ON NEXT PAGE ***'**)
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P_ ANT SYS'TEMS DESIGN, CONTROLr AND INSTRUMENTATION PAGE 6
i 6. --
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.0UESTION 6.03 (2.50)
Indicate whether the following will cause a Rod-Control System URGENT or NON-URGENT FAILURE alsem.
]-
1 a)
Loss of Main 100 VDC Power in the Logic Cabinet U
l b)
Slave Cycler fails to start counting upon receipt of'"Ge' pulses i
c)
Loose circuit card in the Power Cabinet GUESTION 6.04 (1.50)
Answer TRUE or FALSE to the following.
s a.
With no rod motion in progress and the rods in Manucir the rod speed j
meter on the' control board indicates a speed of 0 spm.
l b.'The bank overlap unit will ONLY count steps if the rod control bcnh
~
1 elector switch it in AUTO cr MANUAL.
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c.
Only one of the three (movable, stationaryr or lift) coilt of the CRDM
+s energized at any one time.
QJESTION 6.06 (2.00)-
f hatch the interlock descriptions in Colon.n A with the appropriate logic required to ecuse rod withdrawal to be blocked in Calumn B.
(column S items may ce used more than once) f COLUMN A COLUMN 9 l
E a) Power Ranse High clux 0 103% Power 1.
1/2 m,
,a,,a b) Overtemperature Delte T rod stop 3.
1/3 4,
2/3 c) Intermediate Range High Flur 5.
1/4 6.
2/4 d) Power Renge. Rod Drop 7.
3/4 (xuvs CATECORY 06 CONTINUE 9 ON NEXT PAGE *****)
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PLANT SiSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7
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i QUESTION 6.06 (1.50) t j
a.
Refer to figure 4398 Heater Pewer Distribution.
State the purpose of the Current Transformers (CT's) wrapped on the heater supply lines.
b.
How are snall electrical grounds (e.s. grounds which do not trip i
breakers) detected / indicated on Pressuriner Heater strinsc0 1
c.
How is a bus stripping signal to E:ackup Heater-Group B Pressuriner heaters overriden? (include location of controls) 1 1
J QUESTION 6.07 (2.00) i Unit 3.is operating at full powers all control systems in autor Pressurizer level channel 1 (LT 459) is' selected.
Answer the following; a.
Would a connon failure on LT 461 and LT 462 causing both to read maximum result in a reactor trip? (yes/no) b.
The reference _les for LT 459 PRZR level develops a significant packing leak causing the reference les to decin.
State all the aute actions j
affecting the pressuriner as a result of this nalfunction.
j c.
List cll-functions of Pressurizer pts 455, 456 and 457.
1 1
- DUESTION-6.08 (2,00) 3.
For the Unit 3 Pod Position Indication (RPI) power suppl /> describe I
the supply to the DC power systene including both nernal inverter supply r-and alternate supply.
Include'all autonatic features assceiated with these two sources, j
b.
The Rod Deviation Monitor generates a ' Rod in Motion' signal.
EXFLAIN j~
the basit for this signal AND the actions which-are initiated by the
- signal, f
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(T*x**
CATECORY 06 CONTINUED ON NEXT PAGE
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-PLANT SYSTEMS DESIGN, CONTROL, AND INS PAGE 8
__________________________________________'RUMENTATION____________
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I-t 6.-09 (1.00).
00ESTION Assume the following CCW pump alignments; i
R & C pumps are racked in and in AUTC B pump is running A pump is racked out and open a.
If
'C' pump received an over current lockout what sigt.al(s) uculd auto start
'C' pump ANO at what time would it start?
j
' b.
Would the events in
'a above be different with the 1ccel/renote switch i
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for
'C' in ' local'?
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QUESTION 6.10 (1.50) smrG j-a.
Refer to figures $397.1 sad $397.2 Reactor Coolant Pumpar E:,'
all interlocks which must te met to start an/ 011 Lift Pump (OLP).
CO.5]
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b.
RCP 3A start sequence is in progress (i.e.
OLP control switch to neutral l
and OLP breaker closec) with > 400 psis oil pressure.
Prior to the i
expiration of the 120 see timer the signal fron 'and' sate 1 fails open l
(no signal
'O').
Explain the resultant affect on the completion of the l
RCP start sequence.
El.02 GUESTION 6.11 (1.00)
Refer to figure 4403 NIS Power Supply.
An operating unit ;nverter si_ipply r i or to the transfer
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to 3P07 has been transferred to INV AS from INV 3A.
D IN'> A S output voltage was greater than that or INV 3A, EXPLAIN why this higher voltese would have no effect on the operation of the Power Range NI. instrument power.
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(****"*
CATEGORY 06 CONTINUED ON NEXT PACE
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6.
PLANT. SYSTEMS DESIGNr CONTROLr AND INSTRUMENTATION PAGE 9
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QUESTION 6.12 (2.00) l l
For the following abnormal.P ant conditions state whether or rot AFW l
punp 3A would inject into the Steam Generators at rated conditions AND i
for each condition irphich the AFWP does inject state all advc.se l
consequences (assume a valid start signal occurs).
Consider each case seperately..
~
a.
The discharge flow transmitter on AFWP
'A' has a failed "high' high pressure. tap.
b.
SG 39 supply to Train 1 (valve 3-1404) fails closed.
c.
AFWP-3A Trip & Throttle valve trip solenoid hos en open circuit.
d.
AFWP 3A Trip & Throttle valve connecting rod spring is detached fron the rod, e.
The :t a i n oil pump shaft fails.
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(*****
END OF CATEGORY 06
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7.
PROCEDURES - NORMALi 'ABNORMALr EMERGENCY AND PAGE 10-
--- ss518c851Est E8 ave 8c------------------------
i GUESTION 7.01 (1.00)
OP 0202.1
- Reactor Startup...'. cautions the operator. not to exceed 2000 psi until Steam Generator pressures are > 585 psis.
Enplain why this caution is necess --
i QUESTION 7.02 (1.00)
OP 0205.1 ' Unit Shutdown...' cautions the operator not to pull control power fuses an the Source Range channel (s) in the event.the channels do not reenergine after the P-6 def at buttons are depressed.
Explain the basis i
f o r t ri:us caution.
QUESTION 7.03
'(1.00) l A hydrogen bubble formed in.the reactor vessel is' eliminated by a.
increasing pressuriner temperature above core thermocouple readings.
b.'injteting onysen into the reactor coolant system via.the chemical and volute control system.
l c.
manimizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 spm, and piccing the catien bed demineralizer in service in parallel with the mined bed demineralizer.
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- d. venting the reactor vessel head.
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PROCEDURES
.NORMALr ABNORMAL, EMERGENCY AND DAGE 11
~~~~E5656 LUG 5CdL C6UTEOL
~
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QUESTION 7.04 (1.00) i l
Which of the following reasons correctly describes the basis for allowing t
- Response to Inadequate Core Cooling'.
s.
. Helps to mi:: the SI flow to protect reactor vessel from cold water.
b.
Once subcooling is established, restarting the RCPs helps to collapse voids that may have formed in the reactor vessel head.
c.
Allows restoration of PZR pressure control using nort:al sprays.
i d.
Provides for. cooling of the core when secondary depressurinction does not elleviate inadequate core cooling.
QUESTION 7.05 (1.00) l below requires initiation of Emergency Soration?
{
i Which of the sitvations t
t a.
Following a R: Trip, the rod position indicators show TWO rods i
l which are NOT fully inserted.
i b.
Rod Bank D Low Limit Alarm is actuated.
i c.
An uncontrolled RCS Heatup following a reactor trip occurs.
i l
d.
An unexplained decrease in reactor power occurs while at
<0% rated power.
QUESTION 7.06 (1.50)
Indicate whether the 'ol]owine conditions will place you in a CSF RED PATH or NOT*
3)
Containment pressure of 3/ psis.
b)
All 3 S/Cs at 25%-level and.AFW Flow cf 300 gpm.
t c)
FCS Fubcooling is 10 degrees and PZR level is 12%.
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CATEGORY 07 CONTINUED ON NEXT PAGE
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PROCEDURES - NORMALr ABNORMAL, EMERGENCY AND PAGE 12 3
~~~~E5656L665656~66UTRUL I
~~~~~~~~~~~~~~~~~~~~~~~~
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i GUESTION 7.07
(.50) i 1
'Will adverse contain! tent conditions cause affected control room indications to indiccte higher or lower than actual conditions?
J QUESTION 7.08
(.30)
Fill-in-the-Blanks:
In accordance with ONOP 12308 ' Power Range NI Malfunction *, the Power' Range meters average reading'must be within ______% at greater than or equal to
-90% reactor power and at less than 90% reactor power must be within _____%.
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DUESTION 7.09' (1.00)
State how the Hornal Containnent Coolers are re-started after a safeguard system initiation signal is received.
]
(Address sequence of switch manipulations)
QUESTION
.7.10
(.50) i In accordance with 0-OP-023 ' Emergency Diesel Genettor's how is the start failure relay reset?
OUESTION 7.11 (1.00) a List the 4 conditions that must be net in order to perform a startup following a Reactor Trip without completing on ECC.-
l l-1 GUESTION 7.12 (1.00) us; g Th u k F l ~
How is the RCS cooleddur ing r efueling opera tions with the refueling 7
i cavity full, if 00TH RHR pumps fail Lc operate' j.
(D b<<:oc % gee r. e,:a h V %,, o h )
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(***** CATEGORY 07 CONTINUED ON HEXT PACE
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 13
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RADIOLOGICAL CONTROL GUESTION 7.13 (1.50)
- Besides required. notifications, what are the innediate operator actions if-you are on chift in the control roon and the' refueling supervicar in the i
containment reports they dropped a spent fuel element in containment?
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GUESTION
/,14 (1.00)
S $ criticality and Core Cooling are the two highest priority CSF Status Trees.to nonitor durins cccident conditions.
List the remainins 4 CSF.
Status Trees in DECREASING order of priority.
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GUESTION 7 15 (1.00) 1 i
What plant condition is assumed in the valve / breaker ' Normal
- position-j designetion for the Operatins Procedure alignment sheets?
i.
GUESTION 7.16 (1.00)
During implenentation of ECA-0.0 ' Loss of all A/C-Power'r a red path CSF on containment occurs.
Which procedure should the operator perforn?
l EXPLAIN.
1 i
QUESTION 7.17 (1.00) l Define 'Irside the Target Band" in accordance with DNOP '008.6 ' Operations i
outside the Terset Band'.
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GUESTION 7.18 (1.50)
-After Natural Circulation has been establishedv what 3 indications are monitored to determine RCS COOLDOWNr ectording to ES-0.2r ' Natural Circulation Cooldown'?
I
(*****
END OF CATEGORY 07
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ADMINISTRATIVE PROCEDURESr CONDITIONSr AND LIMITATIONS CAGE 14 GUESTION 8.01 (1.00)
Which of the following require activation of the TSC?
a.
Either an unusual evente alert, site area emergency or seneral emersency.
b.
Only an alerti site area emergency, or seneral emergency, c.
Only a site crea emersency or general emergency, d.
Only a general emergency.
GUESTION 8.02 (1 00)
Using the attached Tech Specs, which action below would be correct for the following~ situation?
SITUATION: Unit 3 is in startup with Tavs 300 deg.
F.
'C' AFW pump is taken out-of-service for a surveillance test (all other AFW equipment it operable), Unit 4 is at 10% power, a.
Unit 4 must be cooled down to 350 de3 F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- b. Both Onit 3 and Unit 4 must be Shutdown /Cooldown to : 350 des. F within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c.
Both Unit 3 and Unit 4 must be Shutdown /Cooldown to 350 deg.
F.
within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if
'C' AFW pump cannot be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, d.
No actions are required.
QUESTION 8.03 (1.00)
D.' C
'A' is inoperable, LPSIP 39 1s also inoperative.
The Tech Specs for S1 syster.s and Electrical systems are attached.
Which statement below as cost correct concerning Mode 3 operat. ions' a.
Mode 3 must be m e i n t e i n e ti.
b.
Restone D/C to aperable status within 7 days or notify the NRC.
c.
Apply LCO 3.0.1.
d.
Startup activities may continue; Mode 2 tr.ay be entered.
(rxtyx C A T E G O R'l 08 CONTINUED ON NEXT PACE *rxx.*:
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 15 I
E GUESTION 8.04 (1.00)
While performin3 fuel shuffles in accordance with OP-16002.5, ' Refueling
[
Core Shuffle", a change to the procedure is necessary and tine does not i
allow normal processing of a chanse.
The chanse should be made in i
accordance with which one of the followins procedures?
'a.
AP 0109.3 DTSC f
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- b. OP-16002.5 Refuelins Core Shuffle
~
u' c.
OP-16002.7 Refueling Preshuffle d.
AP-0109.6 Temporary Procedures l
GUESTION-8 05 (1.50) l For EACH condition listed below state whether er not a TSA would be l
i needed to complete the work in accordance with ADM-503
'TSAs'.
l a,
Testing _of plant systems requiring disconnection of hydraulic lines i
which is not specifically controlled by a Test Procedure.
1 b.
A tank drain (to 3 floor drain) needs to be bypassed by a temporary j
connection to allow for repair work.
c.
Electrical connections to electronic equipment placed under a
[
surveillance procedure, j
GUESTION G.06
'(1.00)
Given the following alignments determine the operability of the OMS channel l
while in Cold Shutdown. If Inoperative indicate why.
OMS Pel. nary Ali3ned OMS Dsckup Alisned r1.
Aligned" lisht on 1.
' OMS B/U cligned' 1:sht on 1.
' OMS o 2.
MOV-*-535 open 2.
MOV-A-536 open l
3.
PCV-*-456 switch in *ciose' 3.
PCV-r-455C switc? in " auto' 4.
N2 chargea to 2090 psis 4
N2 charged to ?000 psig 5.
Setpoint pesition selected for 1456 5.
Low prese. celected for 455C I
L t
-(*****
CATEGORY 03 CONTINUED ON NEXT PAGE
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 16
[
3 i
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l 1
OUESTION B.07
(.50)
TRUE/ FALSE:
A Temporary.Information Tag is NOT considered to be an Operator Aide for administrative control.
QUESTION G.08 (1.50) i Match the description of responsibility for administration of the.tassins/
clearence program in Column A with the appropriate individual in Column B.
COLUMN A COLUMN B
[
t r
I a)
D e n o t, e s if Independent Verification is
- 1) RCO l
required.
- 2) PS-f'
- 3) Maintenance SPVSR b) 'Normally signs the Clearance Release part
- 4) NWE l
of.the Equipment Clearance Order form.
- 5) Maintenance Worker c)
Enters / Deletes entries for Tech Spec related equipment in the Equipment l
Out of Service book, i
OUESTION 8.09 (1.00) l l
Refer to Figure 4 547, "TSA Administrative Process Flow Chart".
In j
accordance with 0-ADM-503 whet person (by position) would complete the review required in the block marked
'A'?
i l
GUESTION 8.10
(.50)
Fill-in the SLANK for the following; In accordence with 10 CFR 55 'if a licensee has not been actively' perform-ing the function of an Operator or Senior Operator for a period of _______
monthst or looser, he chally prior to resuming activitlec licensed l
' pursuant to this part, demonstrate to the Commission that his knowledge and understanding of facility operations and administration are satisfactory."
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(***** CATECORY 08 CONTINUED ON NEXT PAGE xxx+3)
}
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___.__ _ - _. _ _ =. -. _ _
s F
2 8.
ADMINISTRATIVE PROCEDURES. CONDITIONSr AND LIMITATIONS PAGE
.17 QUESTION 8,11 (1.00)
Provide the minimum number of individuals required by tech specs for the following positions to operate both units at full power.
f a)
______ Plant Supervisor-Nuclear (PS-N) i b)-
______ SR0(s).(NOT including the plant supervisor (s))
i i.
c)
.,_____ R0(s) l d)
A0(s) j.
e)'
______ STA(s)
GUESTION 8.12 (1.00)
In accordance with AP 0103.4, 'Inplant Equipment Clearance Orders' for j
elearances on MOVs used as an isolation boundary, tass will be placed at two locations.
STATE these two locations for the valve.
QUESTION B.13 (1.00)
I During Unit 3 startup with the reactor about 2% power, you find that the 1
i PORV block valve is stuck open and incapable of closins.
Which of the I
following is a correct action per TS?
(see attached LCOs) a.-Continued operation is allowed provided the PORV is operable and power removed forn the block valve.
b.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ensure the 00RV is operable and continue c:cuer operations.
c.
Continue power operations since one other PORV is operable, d,
Be-in a condition with Keff ' O.99 within the n e >: t 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in i
Cold Shutdcun within the follcwins 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
[
GUESTION 8.14 (1,50)
List three responsibilities that the Emergency Coordinator CANNOT delesate.
i I
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(**xt* CATEGORY 08 CONTINUED ON NEXT PAGE **xxx) i i
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i 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18 OUESTION 9.15 (1.50)
List the MINIMUM requirements for each of the necessary electrical systems listed below in order to conduct a reactor startup on UNIT 3,
assumins Unit 4 is shutdown.
a)
Unit 3 and 4 4160 KV Duses b)
Unit G 480 VAC Load Centers j
c)
Battery Chargers i
OUESTION 8.16 (1.00)-
r Unit 3 is operating in Mode 6 with the following conditions present' Water level above the top of the reactor vessel flange <23 feet.
D/G A is inoperative.
RHR Train 9 (the inservice loop) is found to be inoperative due to design flow inadequacies.
Specifically which LCO/ ACTION statement (s) would be entered and carried out?
(LCOs attached) 0UESTION 8.17 (1.00) l During Mode 1 operations of Unit 3 it is found that 2 of 3 channels for Pressurizer Pressure High Reactor Trip are inoperative due to a generic material deficiency (repair tine 14 days).
Using TS LCOs attachedr
' determine what actions nust be taken as a result of this failure.
State specific LCO/ ACTION steps that apply.
l t
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(xtvrx END OF CATEGORY 00
- )
(************* END OF EXAMINATION xxxxxxxx****wwx)
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3.4
- c:::r.nts SATETY FIATURES y nlicabili:y:
Applies to the operati.g sea:us of the Engineered Safety Features.
~~~~ ~ ~~^ ~ ~ ~~' ~
gioctive:
To define those limiting :onditions for operation that are necessary:
(1) to ra=ove decay heat f rom the core in emergency or normal shutdown situations, (2) to re-move heat from containeant is nor=al operating and emergency situations, and (3) to remore airborne iodine from the containment atsesphere in the event of a Maximum Hypothetical Accident.
Seccificatie::
1.
SAFETT INJECYION A';D ?25:3"A:. MEAT RI'*0 VAL SYSTE!S The reactor shall not be made critical, except for a.
low power physics tests, unless the following conditions are
_.e :
5 1.
The refueling ta:ar J_
shall contain not less t-p an 320,000 gal. o,
J yichaboron{ con-7 s.
I g
15 entration ef at 1.za pp=.
~
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2.
The boron 1.je:cisa :ank shall contain not less than 900 gal. of a 20,000 to 22,500 ppa boron solution.
he selution in the tank, and in isolated portions of the inlet andd outlet
' piping, shall be =aintained at a temperature of at least 145F. TlO channela of beac tracing shall be operable for the flow path.*
h 3.
Each accumulat:r shall be pressurized to at least 600 psig and contain 475-891 f t of water with a beren c:ncancration of at lasse 1950 ppm, a.d shall set be isolated.
4 FOUR safety injection pumps shall be operable.
l
~ 7 Sco reference (11) on Page 53.4-2 Q
1 3.4-1 Amendments 78 & 72 e
3.6 CMZMICAL AND VOLU!E COSTRCL ST5TZM A9911cabilitv:
Applias to the operational status of the Chemical a=d volume Control systen.
~
Obdective:
To define thosa conditions of the Chemical and Volume Control System notassary to ensura safe teactor operatEn.
Snecification: a. When fuel is in the reactor there shan be at least one flow path to the, core'for boron injection.
- b. A reactor shall not be =ada critical unless the following Chenical and Vol=== control sys tem conditions are met:
1.
T'4 associated charging pumps shan be operable.
1.
TWO lioric acid tra=sfer pumpi shan be ' operable.
The boric acid canks in servica shall contai= a estal 3.
l l
'of attigast 3,080 ganons of a to 22,500 ;;z baron at a temperaturaf ast j i:
145 T.
a.
System piping, interlocks and valves shan be operable' '
to the exten: of establishing gne flow path fro = de boric acid
.a=ks, and one flow path from the refueling water storage.ank, to the taactor Coolant System.
?
5.
TWO eh====1= of heat tracing shan be operahla, far the l
flow path fr=a the beric acid tanks.
~
r 6.
Th' pr! ary wa:ar storage ta=k conta' ins not less dan s
30,000 gallons of watar.
- c. The second react:r shall not be =ade critical unless te following cenditions are me::
1 i
3.6-1 Mendient Mos. 73 & 57 en,,_------.,_p.,-
a m
3.8 STEA4 AND POWER CONVERSON SYSTEM 5 acolicability:
Applies to the operating status of the steam and power conversion
__-.--.-.,y
- - - - - - - ~ - - ~ ~ ~~ ~ ~ ~ ~
Obiective:
To define conditions of the steam-relieving capacity and auxiliary feedwater system.
Soecification:
1.
When the reactor coolant of a nuclear unit is heated above 3500F, the following conditions must be met:
- a. TWELVE (12) of its steam generator safety valves shall be operable (except for testing).
- b. Its condensate storage tank shall contain a minimu n of 185,000 gallons of water.
- c. Its main stea n stop valves shall be operable and capa5!e of closing in 5 seconds or less.
.! Xl'.
I-L
- d. System fiping, Imerlocks
- .Mes (irectly associated wie. the 1
related componer.u in T5 3.3.1 a, b, e shall be operable.
2.
The lodine-131 activity on the secondary side of a steam gene ator shall not exceed 0.67 uC1/gm.
3.
With the reactor coolant system above 3300F, if an.y of above specifications cannot be met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be shutdown and the reactor coolant temperature reduced below 3300F.
Spec 15 cation 3.0.1 applies.
4.
The following number of independent steam generator =+11ery feedwater trains and their associated flow paths (steam and water) shall be operable wW. the reactor coolant is heated above 3500F:
3.3-1 Amendment Nos. }0
O
- a. Single Nuclear Unit Operation Two independent auxiliary feedwater trains ca
- --- powered from an operable steam supply.
-- pable of being -
~
- b. Dual Nuclear Unit ODeration Two independent auxiliary feedwater trains and a third pump capable of being powered from, and supplying water to either train.
- c. If in accordance with T5 4.10.1, testing is required during start-up of either unit TS 3.8.4.a. or b., as applicable, shall apply for an auxiliary feedwater pump, numps, or associated flow oaths (steam and water) found to be Inoperable.
5.
During power operation, if any of the conditions of 3.8.4 cannot be met, the reactor shall be shutdown and the reactor coolant temperature reduced below 3500F, unless one of the following conditions can be met:
- a. For single mit operation with one of the two required independent auxiliary feedwater trains Inoserable, restore the inoperable train to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the reactor shall be I
shutdown and the reactor coolant temperature reduced below 3500F within the next 12 hows.
..p
- b. For dual mit operation, one auxillary ter pump and its
= Mated piping, valvps, and Inter be Inoperable f
operable for time period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the inoperable j
i, provided two indeperident auxiliary f er trains remain pump cannot be made operable within 72 hows, one reactor shall be shutdown and its reactor coolant temperature reduced below 3500F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. For dual unit operation, with one independent auxillary feedwater train inoperable in one reactor, the affected reactor shall be SHUTDOWN and.its reactor coolant temperature reduced below 350*F within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. T5 3.8.5.a applies for the single unit still in operation.
- d. For dual mit operation, with one independent auxiliary feedwater train inoperable in both units, one reactor shall be 5HUTDOWN and its reactor coolant temperature reduced below 3500F within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. TS 3.8.5.a applies for the single unit still in operation.
3.4
- C;;;T. IRED SATETY FEATURES goticabili:y:
Applies to the operati.g sta:us of the Engineered Safety Features.
__~_--
- --~~-~~~
~
~ ~ ~ ~
Oy ective:
To define those limiting :onditions for operation that are necessary:
(1) to ra=ove decay heat from the core in emergency or normal shutdown situations, (2) to re-move heat from containcent is nor=al operating and emergency situations, and (3) to remove airborne iodine from the containment atscsphere in the event of a Maximum Hypothetical Accident.
Scocificati:=:
1.
SAFETY INJECTION A';3 ?.IS D'Ja'. MEAT RI OVAL SYSTC45 a.
The reactor shall not be made critical, except for l
low power physics tests, unless the following conditions are =et:
1.
The refueling ister tank shall conte got less
- gp 5 ban 320,000 gal. cf water with a p I!
~ f_! ntr ti g
d a on et at least 1950 pp=.
2.
The baron inja:tice tank shall contain not less i
than 900 gel. of a 20,000 to 22,500 ppe boron solution.
~he selution in the tank, and in isolated portians of the inlet an'd outlet piping, shall be =aistained se a temperature of at least 115F.
TWO channels of beac tracing shall be operable for the flow path.*
3.
Each accumulater shall be pressurized to ac least 600 psig and contain 875-891 f t of water with a beren c:ncantration of at lasst 1950 ppm, a.d shall not be isolated.
4 FOUR safety infection pumps shall be operable.
fn See~ reference (11) on Page 53.4-2 3.4-1 Amendments 78 & 72
- 3. TWO residual heat removal pumps shall be operable.
- 6. TWO residual heat exchangers shal: be operable.
~
~
_]._ All_. valves,. interlocks and piping associated with the above --
components and required for post accident operation, shall be operable except valves that are positioned and locked. Yalves 862-A and B; 863-A and B; 864-A and B; 863-A B and Ct and
[
866-A and B shall have power removed from their motor operators by locking open the circuit breakers at the Motor Control Centers. The air supply to valve 738 shall 'n shut off to the valve operator.
- b. During power operation, the requirements of 3.s.la may be modified to allow one of the following components to be inoperable (including associated valves and piping) at any one time except for the cases stated in 3.s.l.b.2.
If tht system is not restored to meet the requirements of 3.4.la within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 3.a.;a are not satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor s.all be placed in the cold shutdown condition. Specification 3.* : applies to 3.4.1.b.
- 1. ONE accumulator may be out of sevice for a period of up to s
- 2. ONE of FOUR safety injection pu-.ps may be out of service for 30 days. A second safety injection pump may be out of 3
service, provided p is restored to operable status
.O within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />si tha FOUR safety injection pumps pf 1
shall be tested' strate ooe ability before initiating maintenance of the inoperable
- ou- ::s.
i
- 3. ONE channel of heat tracing on :ne flow path may be out of service for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.'
4 ONE residual heat removal pu-.o may be out of service.
provided the pump is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In addition the other residua: heat removal pump shall be tested to demonstrate ope-astlity prior to initiatin5 maintenance of the inoperable pump.
- See reference (11) on page B.3.4-2 1kS 1-amJ-ma ad a, fMi..J A t
S. ONE residual heat exchanger may be out of service for a period of 2h hours.
6.
Any valve in the system may be inope able provided repairs _.
--- are - completed within ~ 2C ~ hours P-ior to initiating maintenance, all valves that provide the duplicate function shall be tested to demonstrate operability.
l
- 7. To permit temporary operation of fu valve, e.g.,
for surveillance of valve operability, for tSe purpose of valve maintenance, etc., the valves specified m ').4.1.a.7 may be urdocked and may have supplied air or electric power restored for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i
- c. During power. operation three Reactor Cooiant Loops shall be in operation.
- 1. With less than three Reactor Coolant Laoss in operation, the reactor must be in hot shutdown within one wur.
- d. In hot shutdown at least two Reactor Coc. ant Loops shall be operable and at least one Reactor Coolant Loop shall be in operation.'
- 1. With less than two Reactor Coolant Loces operable, restore the required Coolant Loops 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 reduce Tavg to less ttwn or equal to 353 F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
'l' s
.?
.[ ij2.' With no Reactor Coolant Loop in
' pend all operations involving a reduction in boron ::oncentration of the Reactor Coolant System and immediater..nitiate corrective action to return the required Coolant Looc t: operation.
- e. With average coolant temperature less than 350 F, at least two Coolant Loops shall be operable or immediate corrective action must be taken to return two Coolant Loops t= operable as soon as i
possible. One of these Coolant Loops shall be h operation.*
j
- 1. With no Coolant Loop in operation, susoend all' operations
~
involving a reduction in boron concentration of the Reactor i
Coolant System and immediately initiate carrective action to j
return the required Coolant Loop to opera on.
- All reactor coolant pumps and residual heat removal pumps may be de-energized for up to I hour provided D no operations are permitted that would cause di.=on of the reactor i
coolant system boron concentration, and 2) core outlet temperature.s naintained as last 10 F below saturation tempe-ature.
'l I
~
==
i w
2.
During power ::eration or restarting from hot
~
//e* < / ' ' ' " f i
shutdown the f:llowing components may be inoperable:
i a.
ONE start-up transformer may be out of service provided both diesel generators are operele.
The NRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and be advised of plans to
- restore the transformer ~to~senica.
b.
Power operation may continue 15 Orit diesel generator is out of service ;rovided (;)
the remaining diesel generatar is tested daily and its associated engineered safety features are operable, and start-up transformer is oper(2) either able.
If the diesel outage is to be seven (7) days or more the NRC shall be notified.
ONE battery may be out of service for a c.
period of twenty four hours.
d.
Specification 3.0.1 applies to 3.7.2.
+
.h4i
'y 1
1 G
9 9
i e
e
,---r y,,,eme------
p,-ww-,w.,_
_w-,vm,-r,
---w-e mmm
3.0 LIMITING CONDITIONS POR OPERATION - APPlJCABLITY 3.0.1 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within I hour action shall be initiated to place the unit in a MODE in which the specification does not apply by
. _ _.. _. _ placing it, as applicable, in:._
l a)
At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,'
b)
At least HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and*
c)
At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are 1
stated in the individual specifications.
This specification is not applicable in MODE 5 or 6.
3.0.2 Non-compliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for l
Operation is restored prior to expiration of t5e specified time intervals, completion of the ACTION requirements is not required.
3.0.3 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required duri the OPERATIONAL MODES or C ' '. ether conditions specified thereing t upon failure to meet the miting Conditions for Operation, th9
- d. ACTION requirements shall i
met.
3.0.4 Entry into an OPERATIONAL MODE or other ssecified condition shall not be made smiess the conditions for the Limiting Condition for Operation are met I
without reliance on prowlsions contained in the ACTION require nents. This prowlslon shall not prevent passage through or to OPERATIONAL MODES as requi ed to comply with ACTION requirer.ents.
Exceptions to these l
requirements are stated in the individual specifications.
l l
3.0.5 For purposes of determining if a component is operable for LCO considerations, the. component need not be considered inoperable due to l
Inoperability of its normal or emergency power supply if all of its redundant components are operable with their normal or energency power supplies operable.
i i
- NOTE: Until full conversion to STS, when a LCO actir. statement requires a unit to be placed in HOT SHUTDOWN within 6 hou s. refer to Table 1.1 and place j
the unit on the required status to meet the HOT STANDBY MODE.
i i
i 1
1.0-1 A nendment Nos.114 and 108
b
,,.z,,,,
= =, T :
ae.e:
I Approval Gst a.
c.ADM.503 Control and Use of Temporary System Alterations
- lil4I 4
'ii v
i
_t ENCLOSURE 3
--(Page 1 of 2)
TSA ADMINISTRATIVE PROCESS FLOW CHART I. INITIATION / PERFORMANCE I
REQUf5 TOR Procedure i ection 5 2 1 P STA OR TECM DEPT. ENG.
l ProcedureSection 5 3 I
l P laservice Equipment or Systems. /
Equipment Out of Service / Clearances.
9m Componeng5u M2MComponen Sugtetuton stutions 1 P TECH DEPT.5UPV, l
Procedure Sect.on 5 4 I
541
- POWER PLANT 4
54,3 ENGINEERING
{
S.4.2 Mf h
.j-I
- s..t.s TECH. DEPT. 50PV.
Procedure Sten 5 4 3 4
I REQUESTOR l
Procedure Sten 5 41 I
+
na
-G A n' r
4 1
SHIFT TICHNICAN l
Procedure sarman s.a.1 I
4 I
-.nes..
I
+
I SHIFT TECHNICAN b
II23 Procedure Section 5 6 4 I'
564.4 PNSC q.9u,g h
- /> Sm v &
- c. Pressurizer Safetv Valves,
- 1. ONE valve shall be operable whenever the head is on the
_.. _ _ _. _ _ _ -reactor _yessel except _during hydrostatic. test.__ _ __ ________
- 2. THREE valves shall be operable when the reactor coolant average temperature is above 3500F or the reactor is critical.
- d. Pressurizer The pressurizer shall be operable with a steam bubble, and with at least 125 KT of oressurizer heaters capable of being suoo led by emer5ency oower, when the reactor coolant is heated above 3500F.
- e. Relief Valves
- 1. A power ooe ated relief valve (PORV) and its associated block valve shat! be operable when the reactor coolant is Seated above 3309F.
-y +.
I
'i, i-j 3
- 2. If the average coolant temperature is greater than 35t:F and the conditio s of 3.1.1.e.1 cannot be met because one :- nore PORV(s) is inoperable, within I hour either restre the PORV(s) to operable status or close the associated block valve (s) and emove power from the block valve (s); othe-wise, be in a cond:lon with Keff < 0.99 within the next 6 hors and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
- 3. If the average coolant temperature is greater than 3500F and the conditio-s of 3.1.1.e.1 cannot be met because one x more block valve s' is inoperable, within I hour either restxe the l
block valve (s' to operable status or close the block valve's) and remove power from the block valve (s); otherwise, be in a condition wh. Keff < 0.99 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and L. COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
3.1-la Amendment Nos. C and 104
3.10.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION 3.10.7.1 HICH WATER LEVEL At least one residual heat removal (RHR) loop shall be OPERABLE
" *PE}* *"*
~ ~ ~ - ~ ~
APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet.
ACTION:
With no RHR loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible.
Close all contalnment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
'The RHR loop may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
i 3.10.7.2 LOW WATER LEVEL Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shall be in, d :.. [;
APPLICABILITY: 400E 6, when the water leve of
'1 the reactor vessel flange is less than 23 feet.
ACTION:
a)
With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or ea;2al to 23 feet of water above the reactor vessel flange, as soon as possible.
(
b)
With no RHR looo in operation, suspend all
- operations involving a reduction bi baron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
Close all containment penetrations providing direct access' from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.10.8 BORON CONCENTRATION The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be suf5cient to ensure that the more restrictive of the following reactivity conditions is met; either:
3.10-3
- - - - - Amendment Nos.11LandC
g.
3,5 INSTRUMENTATION
__ Apolicability:
Applies to reactor safety and features and accident monitoring
~ ~ ~ ~ ~ ~ ~ ' ' ~ - - - ~ - ~ ~ - - - - - - - - - - - - - -
Instrumentation systems
.i To delineate the conditions of the instrumentation and safety circuits Oblective:
necessary to ensure reactor safety.
^
Specification:
1.
Tables 3.5-1 through 3.5-5 state the minimum instrumentation operation conditions. ' Specification 3.0.1 applies to Tables 3.5-1 through 3.5-3.
0 j
l
{lf I..,
f i ;
'r
~
l 3.5-1 Amendment Nos.
and
TABLE 3.5-1 INSTRUMENT CPERATI3G CONDITIONS FOR REACTOR TRIP
- - - g -- - -
3 3,
MDI.
MUI.
DEGRgg 0FDATOR AC"ICF OPERABLE OF D CCSDITIC31 C ^ '
30.
FUNCTIONAL UNIT CRANNELS REDUN.
CD4tMN 1 OR 2 DANCY CA5537 BE E' 1.
Manual 1
0 usintain het shotdown 2.
Nuclear Flux Power Range
- 3 2
Maintain hot shutdevn 3.
Overtemperature AT 2
1 Maintais hot abattaen 4.
Overpower AT 2
g uniatain hot shutdown 3.
lov Pressurizer Pressure **
2 1
Mai=tain hot shutdown 4.
Ei, Pressurizer Pressure 2
1 Maintain 14 41' d f
.j : i !f [
7.
Pressurizer-Ri Water Level 2
1 Maintain het shutdown 3.
Iow Flow One Loop > 45% R.P. **
2/ loop 1/ loop Maintain het shutdown inw Flow Tuo Loops > 102 R.P.en 2/ loop 1/ b y Maintain het shutdown 9A. Undervoltage & ET Sus 1/ bus 1
unistain hot shutdown e
Below 102 R.P. operation is per=itted with l'ow trip setting only Reactor Trip nterlocks
- Refer to 2.3 e
e
--.---..---,---,w-emm----m-,,-,,,.,-.
_.-,.m---,i.-.-iew-
C
f s
p' 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 1?
ANSWERS -- TURMEY POINT 3&4
-86/0o/08-CAST 07 C ANSWER 5.01 (1.00) a)
EOL
(+.5 ea) b)
Steamline Break REFERENCE TPT TO B3.2-1 001/050; PWG-5(2.9/4.3)
- ANSWER 5.02 (1.00) gg f, "
$r-j y Y tr df ANSWER 5.03 (1.00) 1.
normal 2.
thimble 3.
thimble 4.
normal 5.
normal REFERENCE TPT CNTO Va: II 13-27 ANSWER 5.04 (1.50) a)
decreasef (ne answer)
(+.25 ea response) b)
(no ans); increase ci decreasei decrease d) dectease; deerease RE.~ERENCE CNTO ' Thermal /9ydraulte rinciples II*r pp 12-39-a5 o
039/000i A2.05(3.3/3.6)
d,'f "Og %,4 NUCLEAR REGULATORY COMMISSION UNITED STATES 3
o REGION il f,
5 101 MARIETTA STREET, N.W., SulTE 2900 i
2 ATLANTA, GEORGIA 30323 os...../
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THEOPY OF HUCLEAS DOWF? oLANT CPrRATION, CLUIDSr AND PAGE 20 THERMODYNAMICS ANSWERS -- TURMEY
'0 INT 3&4
- 06/0c/03-CAST 0r C ANSWER 5.05
(.50) calso I+.5) p~e c e p c. N r r NRC IE Info Notice 84-70
('
Sep 1984)
TPT Lesson Plan for Requel Cycle II-1905 011/000; K4.03(2.6/2.?)
ANSWER 5.06 (1.00)
Coalant temperature increase n it flow urward through th s care E0 52.
Ac c r e t 'f i t. DNBR lt lowest t o w ir e d s, the top of the core E0.51.
REFERENCE TPT CNTO '/ 01 II 12-30 ANSWEP S.07 (2.00)
More N e g.c t i v e MTC ro,5:
fuel depletion 2t core botton
.0.51 ruel d t? ! e t o r, is due to the 90L flu: distr:Dution f l u :, pea? at upper core at 20L E0.32 The negative MTC produces nore negetive tecctivity in the to-of the core at EOL than at 00L.
,]
n,4 e c-f c /.s
/c c<tLO.5:
o
/,iG ~-
perepryge (-Coefli ~ i'w, but etw,4llae c.a f-/3<*
e r<
hT CbTb Core Cai tre 9 '; 1 d;f-('y Q
'""4
'y'/,',[
A U S L' E R
- 1. C' 9
( '_. 0 0 i Cank overlap usea to provido a nor' e :: ? '
": ' e at*
._ C: U in!. n o/. O w'
.a
- nifc,. _ :1 neutreo riu; C.tr ibut1 n
.o; e
-tre' r<
1
. ve*r ; (---
'. flu:
- na ri-'in i f o r n 3. +
Le Dot,i
-e
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..h
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r, uel
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'f.
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w t t ',
- i.
c,ne i
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-; d St,
(>q g fyc j_ ;9
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Rrr mCr rPi rmo ers C;ntre l
r, u e n q v, pr a g r
'_nAo.
c n. W. M=
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- o. c NaT.T o u.,
c' U.T w.
a,N, n c A r., E r.;.
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n n
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T, u,. c tm.. e n v. w,. m, q a
ANSWERS -- TURMcY POINT 3M
-36/09/08-CASTO, C 90 ANSWER 5.09 (d O )
O p i f - Y (-
2 (/3.tryw The M i' C is temoerature d e c e ri d e r. t and 'l does not chanw lineart'. At 4CP Y
i i
w MTC het 2 value of G,oem/deq.c : t the core i nlet (55)
and -11 pcm/de^a.C 3 ! the COTO QUt10t f A I C? ), FCP CalCUlCt10ndl ;; U T' ?' O E -
~4 PCD/ DOS.*
O r
?V1,
'T:O d e r E' t O T t P'P D & r a t U r e D 590 dO..
I i ".
'I L P d,
Thi" YDlUe II C101100 n
e i
hb
. I ' ( f)[ t
-/'a..i e
t I
b h'
y e
I 0 ' The M.o c e r c. t a r tence nture does not reech T,vc.
ct tne core i eI lcne. s1nce s
w
[#CWeI D O 3 !'
l '~~ iD hQ 10 4,.j O r h.j l [ O t' t)) i c o 2.- r T. t 's 1 : : reached Selo-a t r1(2 sk k - & guagg(>(g,uppQujOh' % l$,<s w,
- -.n."Q.bg*
%c CO!.
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r-w-
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pf cG,Gpgl& bsn -
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TPT CNTP Coro Con' n! JC t6ps.i>r v r c.-
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 22 ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTO, C ANSWER-5.12 (1.00)
Ensures design cargins to core limits will be maintained E0.75] under both steady-state and anticipated transient conditions CO.25]
REFERENCE i
TPT TS 3.2.2 i
j ANS4ER 5.13 (2.00) a)
Delta T will increase
(+.7) as That goes up due to boiling in the core and Teold remains fearly
- constant
(+.3) l b)
Teold will not follow Psta
(+.7) an Pstm decreases due to boiling i
off in-the S/G and Tc remains #airly conctant
(*.3)
REFERENCE i
Westinghouse Mitigating Core Damasor pp 1.15/16 1'
EPE-017; EMI.01(4.4/4.6) i i
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PLANT SYSTEMS DESIGNr CONTROLr AND INSTRUMENTATION PAGE 23 ANSWERS -~ TURKEY PGINT 3&4
-06/09/08-CASTO, C ANSWER 6.01 (1.00) d REFERENCE Turkey Point, Requal Lesson Plan, RCP. Fig. 16 003/0007 K1.12(3.0/3.3)
ANSWER 44.02 (1.00) c's#
REFERENCE TPT EOP E-1.2, 1.3, 1.4 006/020; A4.02(3.9/3.8)
ANSWER 6.03 (1.50) a)
Non-Urgent
(+.5 ea) b)
Urgent c)
Urgent REFERENCE TPT S05 ' Red Control System's pp 46-49 001/010i K6.03(2.9/3.2)
ANSWER 6.04 (1.50) c.
Felse b.
True c.
Felse REFERENCE T F T G D ':: a c C o n t r o l op. 16 1 39
6.
DLANT SYSTEMS DESIGNr CCNTROL, AND INSTRUMENTATION c' A C E 24 ANSWERC -- TURKEY 00fNT 3&4
-86/09/03-CASTO, C A fs. cA. : v, v.nc
< m. n t, i s
~'
a) 5
(+
5 ea) b)
4 c) 1 2 3 e
I5r
.)
w p,E r e t, e n n e u-TPT 'jD5 ' Rod Control Sys+ n*,
Tig 12: OWC 5610-LI-1, 9heo+
'7 001/050: F4.01(3.4/2.8)
ANSWE?
6.06 (1.50)
_,3 yv
/ y7 f.,, fd /cm d /
CT's.
ere incd for overcurrcnt 7retection.3, l^d i ( ' bl*
- b. Each 4Sas Load Conter is ; :'2 1 p p e d with a g r o e r;; do*ortion -am
. 1.,m e t e r and p h a t, e telectot s u i t c M U 4^ $ n ^ ( % Ifv/V E C<$~ GI /t< m '**
c.
A key ayerated two position switch on the c r e c e n i-conpa-W.t dcur at the 4 dos lor d c e ri t e r
'O'.
Rotete tho cwiter to the 'Eme.acn:/
.c o s i t i o ri.
[0.5 te a. ]
r c c m, r f : r r 1-
---,v-f ?*
,}
,,W 1+4 p r '> r t.1 I
n' (r w '
iJ '
i.; e s,
.v i
3e
'O L' L] + J" 9.
Ki t
t.
b+
C h i', P ;^j l fi j " LOW
'OCri?'I' l'
.,1 1 0 C 10.~l 1 10'/01 d O C P O C 7. O E l t; Cicting lO ldOWri.
1,,+ doun
_r 1
A; pr es wr ; :er level nerea m until 3t 22 '.
r,r*t r
t e _: e t o.- +,,
_ _. _. - m~mwq% a - -p)w ae a g]m p.g
. *r
[
]
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OT delic ! trip end t v W ac' t en,an
- .-..,. o o r;o
a
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._._.,_______.----_,m.m.___.-_
i; 4
1 i
6.
PLANT SYSTEMS DESIGN, CONTROLr AND INSTRUMENTATION PAGE 25 l
ANSWERS ~~ TURMEY POINT 384
-86/09/08-CASTOR-C i
l~
ANSWER' 6.08 (2.00) i.
a.
Inverter is supplied from AC (30766) and DC (3001-49) [0.5] and auto transfers to alternate E0.252 MCC
- A' Lts'penel ELP.3172 E0. 253..
l b.'When control rods are in motion the trip atpoint of the rod deviation bistable is changed-EO.5]+ The change prev.cits false alarms from l
spurious signal inputs CO.S].
l REFERENCE TPT SD 006 i
.f f
ANSWER 6.09 (1.00) UV'4 ath8.5 psi) low
.a.
C would start header pressure CO.25] after 30 sec CO.253
-b.
No [0,53 i
j REFERENCE i
TPT SD 40
?
~ ANSWER 6.10 (1,50) i
-)
l
- a. OLP CS start and breaker closed [0.25 eo) # C5 FM i
b.
The RCP will not start due to stopping of the OLP (contacter open) the not gate will have a
'O' signal in and'a
'1' out thus tripping open the contactor this will reselt in a loss o ri signal (s) to the 'and' gate p r i o r-to the 120 see timer. E1.0J l
l REFERENCE TPT SD-8 RCPS t
1 i
ANSWER 6.11 (1.00) l The-instrument power _is supplied thrcugh a SOLA transformer.
The function of the SOLA trantformer is to provide e conctant volta 3e supply.
REFERENCE TPT SD 4 I
i a
i 1
l l
i C -.~
- m..
e 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATIOW PAGE 26 ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTO, C ANSWER 6.12 (2.00) a.
would not inject E0.252
\\
b.
would inject loss of redundancy for steam supply E0.52 c, would inject no electrical tripping capability E0.52 d.
would inject possible' failure to mechnically overspeed trip cod loca.'.ly e.
ot d not inject A db I
0, 52
<c-c REFERENCE
^"')
TPT SD 117 i
i i
I, 4
e i
i
t i
i-I i.
l 4
1
+
r 7.
PROCEDURES - NORMAL, A E:N O R M AL, EMERGENCY AND PAGE 27 l
~~~~kdb5bEbb5 CAL 65UTEUE~~~~~~~~~~~~~~~~~~~~~~~~
~
l ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTC, C L
ANSWER 7.01 (1.00)
I t
i An SI uill automatically be intitiated by high steam line - stean generator diff, p r e s uv a if thig is onconded.
REFERENCE TPT OP-0202.1 p.
35 ANSWER 7.02 (1.00)
'Pullins the control power-fuses en-1/2 channels will result in a reactor trip. Ceaf]
f r
REFERENCE f
TPT OP 0205.1 p.
12 ANSWER 7.03 (1.00) i d
I REFERENCE I
MNS EP/2/A/5000/16.3 CNS EP/1//A/5000/2F3, p.7.
I NAPS 1-FP.P-1.3Ar p.3.
i TPT EP CRP I ANSWER 7,04 (1.00) d REFERENCE i
Westinghouse background info for TPT E0Ps- "RCP Trip / Restart'- pp 49/50 l
000/074; Et:3. 07 ( 4. 0/4. 4 )
t r
l I
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PROCEDURES - NORMALr ABNORMAL, EMERGENC) AND PAGE 23
~~~~
RE5i5EasiCAE c5sTR5E------~~--~~~---~~~--~~~
i ANSWERS -- TURKEY POINT.3&4
-86/09/08-CAST 0r C
!1 i
ANSWER 7.05 (1.00) i a
REFERENCE i
TFT ONOP 2600.1r pp 1/2 000/024; PWG-10(4.1/4.4) 1 l
ANSWER 7.06 (1.50)
{
a)
NO
(+.5eg);L b)
M ll'O u %^
i-c)
NO REFERENCE TFT E0P-F-Or CSF Status Trees l
PWG-10' EOP E r, t r y Level Conditions (4.1/4.5) i l
ANSWER 7.07
(.50) 1 l
Higher 1
i REFERENCE North Anna EP-0 foldout page.
TPT E-0 ANSWER 7.08
(.50)
\\ ; l' ANSWER 7.09 (1.00)
The safegaurd initiation signal nust be reset E0.52 and then the NCC reset pushbutton must be depressed E0.53.
REFERENCE TPT 3-OP-057 ctep 4.4
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29
~~~~E5056E655U E CU YR6[~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS - ' TURKEY POINT 3&4
-86/09/08-CASTO, C ANSWER 7.10
(.50)
By depressing"the ' Alarm Reset and Stop' pushbutton.
REFERENCE
'TPT-OP-023 EDG ANSWER 7.11 (1.00)
/1 )j CritDqality planned within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />
(+.25 ea)
/L
'2 )
Boron bqncentration < 300 ppm
- ,5 ij 3h Trip was\\from S 40%
1 j
14{
Equilibri - Xenor existed prior to trip
[ REFERENCE N
TPT UP-0202.2, pp 15 001/050; PWG-12(3.7/3.7, A"SWER 7.12 (1.00)
An SI punp (+425) takes a suction on toop C (+.25) via the RHR system and its HXert (+.25) where its cooled by CCW
(+.-25) i OR l
Natural Convect.ag/-gre,ppagigr qgia the refueling cavity being maintained 7
23 ft above the >Fb >I flange.
(+1.0) l l-REFERENCE TPT ONOP-050, pp 5/6; tot TS 9 3.10.7 005/0001 K3.07(3.2/3.6) l l
l
m-7.
PROCEDURES'- NORMALr ABNORMAL, EMERCENCY AND PAGE 30
~~~~kUUEUEUUEUAE~CUUTRUE--------~~~~~----~~~----
ANSWERS -
. TURKEY POINT 3&*
-86/09/08-CAST 0r C ANSWER 7.13 (1.50) 1)
Sound Containment Evacuation Alarm
(+.25 ea) 2)
Stop CNTMT purse supply fan 3)
Stop exhaust 4)
Close the CNTMT Intstrument Air Bleed Uslves 5)
Close CNTM purse supply isolation valves
-6 )
Close exhaust REFERENCE TPT ONOP 16008.2, pp 2/3 034/000; PWG-11(2.0/4.1)
ANSWER 7.14 (1.00) 1)
Heat sink
(+.15 for CSF,
+.1 for correct order) 2)
Integrity 3)
Containment 4)
Inventory REFERENCE E0P-F.0, 'CSF Status Trees' l
i ANSWER 7.15
- ( 1.-60 )
l Cold Shutdown REFERENCE TFT ADM 201 ANSWER 7.16 (1.00)
The-operator should remain in ECA-0.0 since the RPs are written on the premise that at least one E-bus is enersized.
REFERENCE West, background info. for ECA-0.0 t
I a
0 t
i i
1
r 7.
PROCEDURES - 10RMAL, A B N C ';' " Y, EMERGENCY "v; 0 PAGE 31
--~~EE5i5E55 EE-E5 HE5E-~~---~~~~~~~-------~~~~
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Core U"'4 7 '
'+.5 ec.
T-Hot
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r_ e_. n. o,
TPT ES
').27 pp 5 cc.r_n7a; c
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ADMINISTRATIVE PROCEDURES, CONDIT10NSr AND LIMITATIONS PAGE 32-ANSWERS -- TURKEY FOINT 3&4
-86/09/08-CASTO, C ANSWER 8.01 (1.00)
(b)
REFERENCE EPIP 3.02 and 3.03.
TPT EP 20104 pp 6.
PWC-36: E-Plan (2.9/4.7)
. ANSWER 8.02 (l'.00) d.
REFERENCE TPT TS 3.8.1.4 ANSWER 8.03 (1.00) c.
REFERENCE TPT TS 3.7.2.b ANSWER G.04 (1.00) b.
REFERENCE TPT AP 0109.3 OTSC pp.
4.
i-L ANSWER 8.05 (1.50) 1.
would 2.
would not' 3.
would not E0.5'ea.]
REFERENCE TPT ADM 503 see. 4.1 2 l
r u
8..
ADMINISTRATIVE PROCEDURES, CONDITIONS,-AND LIMITATIONS PAGE-33 ANSWERS -- TURKEY POINT 3&4
-86/09/08-CASTO, C I
ANSWER 8.06 (1.00)
Primary Channel misaligned 13 should be in ' auto' Backup Channel aligned properly REFERENCE TPT,AP 0103.32 pp 3 ANSWER 8.07
(.50)
TRUE REFERENCE TPT AP 0103.36 pp 2.
ANSWER 0.08 (1.50) a)
2 cr 4
(+.5 ea) b)
3 c) 1 REFERENCE TPT AP 0103 Sr p 5-9 PWG-14:~Tassing/ Clearance Procedures (3.6/4.0) t ANSWER 0.07 (1.00)
PS-N REFERENCE TPT ADM 503 4
i ANSWER 8.10
(.50) 4 months REFERENCE l
Pwo-23(shirt Starrins/ Activities) (2.e/3.5) 4 e
l e
e 1
M 4
._.,, _. -., -. ~. _ _ _ _ _ _.. _ _ _
r ---
- 0.
ADMINISTRATIVE PROCEDURESr CONDITIONS, ANO LIMITATIONS PAGE 34 i
ANSWERS -- TUEMEY POINT 3&4
-86/09/09-CASTOR C ANSWER 8.11 (1.00) a.
1 b.-
1 c.
3 d.
3 i
e.
1 REFERENCE NA Ul&2 TS table 6.2-1.
TPT TS Table 6.2-1; TPT AP 0103.2r pp 14 PWG-23: Staffing / Activities (2.8/3.5)
ANSWER 8.12 (1.00) i i
Breaker and Handwheel
-E0.5 ee.2 i
REFERENCE t
TPT.AP 0103.4 pp 4.
ANSWER
~ 8.13 (1.00) d.
(
REFERENCE TPT TS 3.1.1.e l
l ANSWER 0.14 (1.50) 1 l
1.
Class fication of event 2.
Decision to notify state and 1ccal authorities 3.
Protective action recommendations i
r REFERENCE
-TPT EP 20101 pp 2 i
f-l
{
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4DMINISTPATIVE *ROCEDURES, CONDITIONSr AND LIMITATIONS PAGE 35 ANSWERS -- TURKEY 00 INT 3&4
-86/00/03-CASTO, C q/q/,wS'5,cg541 ANSWER 8.15 (1.50) 7 7
a) 3A and 39 and either 4A or 48 4160 KV buses
(+
5 ea) b)
3 of a 480 VAC Loed Centers c) a of 6 Cattery Chargers REFERENCE TPT TS 3.7.1 062/000;PWG-5t3.0/4.0) pggd ANSWER 8.16 (1.00) b.fA ed "
titd48
- Y.
- TS 3.10.7.2 action
, p 40MdTN.
- g4 (D/C LCO previously entered)
A M @#
ud kI REFERENCE eeM TP1 TS 3.10.7.2 n
+
- M r5 @
ANSWER 8.17 (1.00) apply LCO 3.0.1/section 3.5 applies REFERENCE TPT LCO 3.0.1 1
4 l
l l
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l
P. o. BOX 10C33. JUN6 8EACH FL 33408 04W 0
ENCLOSURE 3 o
y 1
A 8.
SEPTE.'n ER i 935 L-86-382 Dr. J. Nelson Grace Regional Administrator, Region 11 U. S. Nuclear Regulatory Commission 101 Marietta St., N.W., Suite 2900 Atlanto, GA 30323
Dear Dr. Grace:
Re: Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Operator License Examinations This is to confirm the technical comments presented to your staff at the exit interview following the examinations administered to proposed and current operators at Turkey Point on Septernber 8,1986. These previously discussed technical issues and several additional technical comments may bear significantly on your evaluation of these examinations; accordingly, they are formally transmitted herewith for your consideration. i Also, as your staff is aware, FPL has identifie number of concerns with respect to the examinations which go to their basic validity and which we believe should be reflected in your final action on the above-mentioned examinations. We are prepared to discuss these mqtters with you at your earliest convenience.
As you are also aware, FPL has previously taken steps to improve our operator training program and these programs have now been accredited by the National Nuclear Accrediting Board.
These training improvements and the resources expended in support of these improvements cannot be totally effective if the testing process that our operators are subjected to is not valid.
We look forward to discussing these issues with your staff.
Very truly yours, MJ C.O.W Group e President Nuclear Energy COW /PLP/cvb Attachment cc:
Harold F. Reis, Esquire M M Y %T h 32h-PEOPLE. SERVING PEOPLE
09/11/86 Paga 1 s
s NRC Exam Section 1 QUESTION RO 1.03:(SRO 5.01, SRO Requal 5.02)
A reactor has been operating at full pwer for three months when a manual reactor trip occurs. All systems are toperational and the steam dumps are immediately placed in the steam pressure control mode. Ten minutes after the reactor trip, all RCPs are tripped. Twenty minutes after the reactor trip, Loop 1 RCP is jogged momentarily. Which set of traces (a - c) on figure #174 most closely represents the previously described events?
RESPONSE
We request that answers"a" cr"b" also be accepted or delete question.
Answers "a" and "b" correctly show Te increasing at the beginning of the transient. In addition, the poor quality of the trace reproduction made it difficult tointerpretthe ATbetweenTe and thermocouple readings.
Discussion with Westinghouse confirme'd Eh'at none of the traces given in the question is truly representative of what should actually occur.
REFERENCES:
Telecon with Westinghouse on September 10,1986
- RI-5/cas
09/11/86 l
Page 2 NRC Exam Section 1 QUESTION RO 1.04:(SRO 5.02, SRO Requal 5.03, RO Requal 1.04)
For each of the following parameters compare a Core Thimble Cell to a Normal Core Cell and indicate which of the two;has the higher value. (rated conditions, average values) i
- 1. Total power production l
- 2. Powerdensity
- 3. Enthalphy rise t
- 4. Mass flow l
5.DNBR l
l
RESPONSE
l We request that this question be deleted from Requal Exams.
This topic was not covered in the Accelerated Requal Class. Lesson Plan in l
Accelerated Requal Class did not cover this t'op,ic as shown on reference.
REFERENCES:
Instructor Lesson Plan No. 0017-OL, Appendix A, Page 2, Rev. O,03/31/86
- RI-5/cas l
TESTITEM ANALYSIS SHEET Question # / Point Value:
Question:
A reactor which has been operating at full power tEOla trips due to aj Loss of All Offsite Power.
Natural circulation conditions have been established and steam dump to condenser is operational. Steam Generator levels / pressures are at no-load stable values.
Core decay heat generation how ever. is exceeding the free convection heat transfer rate capability.
EXPLAIN the thermodynamic principles which will cause fuel element temperatures to eventually reduce to lower values.
(assume a relatively constant heat generation rate within the fuel)
Answer:
Since free convection heat transfer rate under subcooled conditions increases as a function of the available temperature difference between wall and bulk Guid (delta T)[0.25l the heat transfer rate to the Rx coolant in the core area will increase
[0.25) causing an increase in core exit temperature (0.251 increasing the available natural circulation driving head [0.25l whicp will cause an increase in mass now [0.251 thereby mereasing the convective heat transfer capability of the RCS.
The higher, mass now will also cause (0.25] a corrective decrease in fuel element temperature.
i....
Reference:
TPT CNTO Vol 1114-26 K/A EK 1.013.7/4.2 Technical Review:
Inaccurate information stated in stem orquestion.
1)
Steam dump to condenser can not be operational during loss of all offsite power. CW pumps will trip, condenser vacuum will decrease, locking out the condenser steam dumps.
2)
Increasing steaming rate will also increase driving head.
3)
Higher mass now resulting solely from increased coolant temp will only stabilize fuel temperatures, n3 cause a
" corrective decrease" Construction Review:
No comments Relevancy Review:
No comments l
i K&A Review:
InsufHeient K/A reference 1
l Recommendation:
Delete question.
Roll lo
- RI-5 d4p 09/13/86
1 09/11/86 Page 3 I
NRC Exam Section 1 QUESTION RO 1.12:(SRO 5.09, SRO Requal 5.06)
Restrictions on F(Q)(T) limit include penalties due to unccver and reflood rates, post-LOCA. Give one other basis for gore height restrictions on F(Q) (T) and explain where and whyitis restrictive. '
RESPONSE
We request the answer key to be expanded to accept answers that concern fuel temperature, clad temperature and DNB during SB LOCA and discussion of K(Z) curve.
F(Q) was taught in Accelerated Requal as a limit on fuel temperature rise to protect the clad.
\\
t
- 6. e,
l l
l l
- R1-5/cas
09/11/86 Pege 4 NRC Exam Section 1 QUESTION RO 1.13c:(SRO 5.10)
Explain why curve "C"is valid at < 15% quality and not at > 15% quality.
c.
I RESPONSE c:
We request that the answer key be expanded to accept an explanation of analysis limitations to generate the curve.
REFERENCES:
Thermodynamic Principles II, page 13-54,55 and 56, Westinghouse copyright 1982.
i b
9 6.*6-l l
- RI-5/cas m-w i
y.,
w.
w
-- =
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1 TESTITEM ANALYSIS SHEET J
Question # / Point Value:
RO 1.13 tSR0 5.10i / t.5 pts.
Refe' to figure e 357. Tavg vs. Power *and answ er the follow mg:
Question:
r a.
On the eraph curve 'A' hounds the maumum operating loop ten $perature at a given system pressure ter a particular reactor p.m. r.
Esplain the basis for graph *A*and w hy it has a negative slope.
b.
Graph 'B' ensures that two adverse conditic.ns. ire not exceeded. State these two conditions.
c.
Explam why curve 'C' is vstid at <15T ouality and not at.> t54 quality.
Answer:
a.
This curve is based upon preventing hot 'eg temperature f om reaching a saturation temperature as power increases.the difference between Th
& Te increases. (0.251 In order to keep Th below saturation. Tatg must decrease with power (0.251.
b.
To maintain minimum DNBRl0.251 Ensures that the avg. enthalpy at the vessel exit does not exceed the enthalpy of saturated liquid. [0.251 c.
At higher pressure. a higher Tavg is necessary for 15% quality 1.251 at low er pressures DNB occurs at 15%. At higher pressures, the quality of
\\ water may be >20% before DNBoccurs.[0.251
Reference:
cNTo voi n ta.5a.
i i
Technical Review:
Ro i.tsc: The an.wer key does not address the question.
Curve "c'is valid at < 15% quality and not at > 15% quality because the WRB.
1 correlation used to predict DNB only applies to fluid at less than 15% quality.
l Above 15% quality.the WRB.t correlation is not valid.
Construction Review:
Part 'a' is misleading in that it refers to " maximum operstmg loop temperature" but graph labelisT_gyg vs. power.
l l
Part "c' Figure doesn't apply to PTN < ES-201 I apg 5 of 8); Westinghouse told trainees concept has no a pplicability to liN: double jeopardy question iES.202 E 13.t pg 4 cf 6). Thorndike and Hagen t pg 951.
Relevancy Review:
Not relevant for RO candidates. NRC Examiner Standard ES.202. " Scope of Written Examinations Administered to Reactor Operators. Power Reactors".
Rev.1.10/1/84. Part E.1 states that Tech Spec questions for reactor operators should be conceptual in nature, which does not include specific basis l
explanation. (Pa ge 3 cf 6. ES.202).
K&A Review:
No K/A reference Expand Answer Key (per previous submittal) or Recommendation:
Deiete entire question rrom Ro exams.
Delete part "c" from SRO exam.
l
- RI-i d,p 119/ l & S6 Roll.13
1 09/11,86 Page 5 NRC Exam Section 1 QUESTION RO 1.14c:(SRO 5.11)
State three possible initiation signals which may initiate SI upon a steam c.
line break accident.
l
[0.75]
RESPONSE c:
We regttest that the answer key be expanded to accept Hi Steamline Flow with Lo Tave or Lo S/G Pressure
REFERENCES:
Logic diagram 5610-T-L1, Sheet 19, Rev. 5 i
I e
l
- R1-5/cas
1 09/11/86 Peg 26 NRC Exam Section 1 QUESTION RO 1.16:(SRO 5.13, SRO Requal 5.09, RO Requal 1.09)
Explain two reasons why the computed moderator-only power defect underestimates the actual defect, dud to redistribution at BOL, assuming a calculational value of MTC of-4 pcm/deg. F. [HFP, Cb = 1200 ppm]
RESPONSE
We request the answer key be expanded to accept a general discussion of MTC, Power Defect and redistribution due to the large percentage of Operators who had difficulty in understanding the question.
\\
i l
k l
l
- RI-5/cas
09/11/86 Pag 27 NRC Exam Section 1 QUESTION RO 1.19:(SRO 5.16, SRO Requal 5.07)
Refer to figure # 363, discuss the two factors which affect the Axial Power Distribution at EOL and result in a relatively flat power distribution.
RESPONSE
We request the answer key be expanded to accept the affects of redistribution and flux burnout in middle of core since Turkey Point is in Fuel Cycle X and our flux is slightly positive at BOL.
REFERENCES:
Telecon with Turkey Point Reactor Engineering on September 9,1988.
\\
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be #.
a e
l l
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(
1
09/11/86 Page 8 NRC Exam Section 1 QUESTION RO 1.21a:(SRO 5.18) a.Why is this action necessary during Natural Circulation Condition
(.75)
I t
RESPONSE
a.
We request the answer key be expanded to accept responses that address non uniform mixing of the RCS in general, notjust the pressurizer.
s 4
i...
~
a l
l I
N l
l
- RI-5/cas I
I.
09/11/86 Pega9 NRC Exam Section 2 QUESTION RO 2.03:(SRO 6.02, SRO Requal 6.02, RO Requal 2.02)
Which of the following statements correctly describes the RHR System lineup when HOT LEG RECIRCULATION is,stablished? (assume both trains of RHR are available) a.
Both trains are used to supply hot leg recirculation exclusively, b.
One train is used to supply hot leg recirculation and the other train is used to continue cold leg recirculation.
Both trains supply both hot and cold leg recirc simultaneously.
c.
d.
One train is used for hot leg recire and the other train is put in standby (i.e.,
recirculates from RHR HXer outlet to RHR pump inlet).
RESPONSE
We request the answer key be revised to reflect "b" as the correct answer.
\\
REFERENCES:
EOP-ES-1.4, Steps 1 and 2 dated 07/27/86 l
l
- R15/cas
09/11/86 Pagg10 NRC Exam Section 2 QUESTION RO 2.11:(RO Requal2.06)
List the 6 essential safety related loads served by the Component Cooling Water System (redundant components like CCK heat exchangers are one load).
t
RESPONSE
We request the answer key be revised to accept " Pass Heat Exchanger" and delete SFP Heat Exchanger
REFERENCES:
SD 040, Rev.1-7 s
I 9
l l
l l
- RI-5/cas
-,,---__,,---A
-m-
-- - - - -,---,_~ -
n
09/11/86 Pcge 11 NRC Exam Section 2 QUESTION RO 2.12a:(SRO 6.13, SRO Requal 6.10, RO Requal 2.09)
Refer to figures #397.1 and #397.2 Reactor Coolant Pumps.
List all a.
interlocks which must be met to stqrt an Oil Lift Pump (OLP).
[0.5]
RESPONSE
Figure 397.1 requires breaker to be closed, but 397.2 does not. We request a.
that no penalty be levied for a response that does not include " Breaker Closed".
REFERENCES:
Figures 397.1 and 397.2
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09/11/86 P ge 12 NRC Exam Section 2 QUESTION RO 2.13 a and b:(RO Requal 2.08)
Refer to figure #398 Heater Power Distribution. State the purpose of the a.
Current Transformers (CT's) wrapped on the heater supply lines.
b.
How are small electrical grounds ('e.g. grounds which do not trip breakers) detected / indicated on Pressurizer Heater strings?
RESPONSE
We request the answer key to be expanded to accept the response "CT's a.
supply any indication meter."
b.
We request the answer key be expanded to accept the response " annunciator ~
in Control Room for 480 L.C. Ground." In addition, there is no ground detection instrument for pressurizer heater strings.
We request that alternate answers for ground detection be accepted.
\\
REFERENCE:
~
Training BriefNo.64,06/15/85
- RI-5/cas
TEST ITEM ANALYSIS SHEET Question # / Point Value:
R0 2.16/ 0.50 pts.
Question:
Refer to Jigure #401 Charging Pumps. STATE the forcetsi which mdintain seal water flow through the packing of the charging pumps.
Answer:
(thermally induced) natural circulation and head of water form seal tank [0.25 eachl
Reference:
TPT SD 12 Technical Review:
Seal water flow through the oacking of the pump is maintained by the static head of the head tank and has nothing to do with thermally induced natural circulation.
Construction Review:
Answer not relevant to question; answer key does not answer question.
ES-107 C1,(pg.1 of 31, Wood Ipg. 46),Thorndike & Hagen,(pg 79), ES-202 E10,(pg. 4 of 6). I F1L*).
i Relevancy Review:
No comments.
K&A Review:
No K/A reference Recommendation:
Accept seal water static head as answer.
- RI-5/d4p 091& S6 Roll-17
~
09/11/86 Pega13 NRC Exam Section 2 QUESTION RO 2.19 a, c and e:(SRO 6.17, SRO Requal 6.12, RO Requal 2.12)
For each of the following abnormal plqnt conditions state whether or not AFW pump 3A would inject into the Steam G6nerators at rated conditions and for each condition in which the AFWP does inject state all adverse consequences of the given condition (assume a valid start signal occurs).
Consider each case separately.
The discharge flow transmitter on AFWP 'A' has a failed "high" high a.
j pressure tap. (prior to start)
AFWP 3A Trip & Throttle valve trip solenoid has an open circuit.
c.
l e.
The main oil pump shaft fails.
RESPONSE
i We request the answer key to expanded to accept "will inject" with an a.
accompanying explanation on a c,ase,by case basis. The questio'n was confusing to the Operators since there is no discharge flow transmitt'er as such for"A" AFW pump.
We request the answer key be expanded to accept "will not inject" with an l
c.
accompanying explanation. There was insufficient information given 'to determine the affects of an electrical open circuit on the T&T valve trip solenoid.
We request the answer key be expanded to accept "will inject" since the e.
pump will not trip. The pump will run until it seizes and fails.
REFERENCES:
Operating Drawing 5610-T-E-4062
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09/11/86 Pcge 14 NRC Exam Section 2 QUESTION RO 2.22: (RO Requal 2.14)
Refer to figure #409 " Basic CVCS Flow Balance". For points 1 - 5 state the design valu es for flow (gpm) at each point.
RESPONSE
We request the answer key be expanded to accept 45 gpm to 120 gpm for point I due to varied LTDN orifice configurations possible.
We request the answer key be expanded to accept the value of 25 gpm for thermal l
barrier flo.w (point 5) which is the value of CCW flow through the thermai l
barrier. Students asked the proctor if they wanted the CCW flow and the proctor said "yes".
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REFERENCES:
Turkey Point FSAR, Table 9.2-2 l
SD008, Rev.1-19 I
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Page 15 NRC Exam Section 3 QUESTION RO 3.08d:
Match the interlock descriptions in Column A with the appropriate logic required to cause rod withdrawal to be blocked in, Column B. (column B items may be used more than once)
COLUMN A COLUMN B a.
Power Range High Flux @ 103% power
- 1. 1/2 b.
Overtemperature Delta T rod stop
- 2. 2/2 c.
Intermediate Range High Flux 3.1/3 d.
Power Range Rod Drop
- 4. 2/3'
- 5. 1/4
- 6. 2/4
- 7. 3/4 i
RESPONSE
We request the answer key be revised t Seflect"5" as the correct answer.
d.
REFERENCES:
l Operating Diagram 5610-TL-1, Sheet 17 l
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09/11/86 Pag 216 NRC Exam Section 3 QUESTION RO 3.09a:
For each set of plant conditions listed below STATE all functions (of 1,2 and/or 3) which would be generating a rod movegnent signal and STATE the response of Bank D rods (in, out, no movement) to this signal. Bank D at 200 steps.
- 1. Bank Selector Sw.
- 2. In-Out-Hold Lever
- 3. Plant Condition a.
Manual IN Urgent Failure
RESPONSE
a.
We request the answer be expanded to accept " Rods Do Not Move" since this is true for an urgent failure in a logic cabinet.
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REFERENCES:
Control System Diagram 5610-T-D-12A, Sh6et 1 of 1
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NRC Exam Section 3 QUESTION RO 3.14a:(SRO 6.20, SRO Requal 6.09, RO Requal 3.09)
Assume the following CCW pump alignments:
B&C pumps are rackedin andin AUTO B pump is running A pump is racked out and open If'B' pump received an over current lockout what signal (s) would auto start a.
'C' pump AND at what time would it start?
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RESPONSE
We request the Operators not be penalized for not including the low pressure setpoint in the first part of the question. The setpoint appears in the answer key but was not asked forin the question.
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c 09/11/86 Pega 18 NRC Exam Section 3 QUESTION RO 3.15:(SRO 6.21, SRO Requal 6.08, RO Requal 3.08) a.
For the Unit 3 Rod Position Indication (RPI) power supply, describe the supply to the DC power system, including both normal inverter supply and alternate supply. Include all auto'matic features associated with these two sources.
b.
The Rod Deviation Monitor generates a ' Rod in Motion' signal. EXPLAIN the basis for this signal AND the action (s) which are initiated by the signal.
RESPONSE
a.
We request the answer key be expanded to accept a drawing (sketch) of this system. It is our understanding that the Examiner stated that a drawing was acceptable.
b.
We request the answer key be expanded to accept " Alarm at 12 steps Stationary,24 steps moving" s!nce this demonstrates the setpoint change as stated in answer key.
REFERENCE:
SD6, Rev.0-12 l
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09/11/86 Page 19 NRC Exam Section 3 QUESTION RO 3.16b:
b.
Exactly how does Steam Generator level flow controller FC 478 operate the Main Feedwater Control Valve?
Include motive force (s) and signal conditioningequipment.
RESPONSE
b.
We request the answer key be expanded to accept instrument air as a motive force. Question was read as" motive force to move valve"
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F Paga20 NRC Exam Section 3 QUESTION RO 3.18b:(SRO 6.22, SRO Requal 6.07, RO Requal 3.05)
Unit 3 is operating at full power, all control systems in auto, Pressurizer level channel 1 (LT 459)is selected. Answer the following:
t Would a common failure on LT 461 and LT 462 causing both to read a.
maximum result in a reactor trip? (yes/no) b.
The reference leg for LT 459 (PRZR level) develops a significant packing leak causing the reference leg to drain. State all the auto actions affecting the Pressurizer which would occur as a result of the failure.
List all functions of pressurizer pts 455,456 and 457.
c.
RESPONSE
b.
We request the answer key be expanded to accept " backup heaters on." In addition, we request the Oper'ators not be penalized for not including the water solid condition written in the answer key since the question aslied for auto-actions and did not refer to a transient description over time. \\.
We request the answer key be expanded to accept " reactor protection and c.
SI" as stated in SD009 and also accept "Hi and Lo Pressure Trip, Lo Pressure SI and associated alarms."
REFERENCES:
i SD-009, Rev. 0-40 5610-T-D-16A, Sheet 1 of1 j
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QUESTION RO 3.19 Refer to figures 82.1,.2:
What is the purpose of the Droppek Rod Normal / Reset / Bypass or Switch a.
on the Power Range A drawer?
b.
What automatic actions (if any) would occur as a result of a " loss of detector volt alarm"on the Power Range A drawer?
c.
State the purpose of the gain switch on the Power Range B drawer.
RESPONSE
b.
We request the answer key to accept
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1)
Blocks Auto Rod Withdrawal 2)
Initiates a Turbine Runback Signal.
Students were led to believe that the P.R. Detector lost its hi voltage supply, which will cause the detectors output to fail low, therefore tripping the NIS 5%
pwr decrease 5 seconds bistables which generate the above listed signals.
REFERENCES:
TP System Description #004 Pg. #46 & #47 TP Fact Sheet #012 TP Fact Sheet #003 TP Drawing 5610-T-LI Sh. #21 i
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09/11/86 P ge 22 QUESTION RO 4.04 (SRO 7.04)
Which of the following statements describes the correct usage of the Critical Safety Function (CSF) Status Trees while performing EOP-E-0 " Reactor Trip or Safety Injection?
a.
The CSF Status Trees are ONLY monitored when EOP-E-0 directs.
Awareness of Red Path Criteria is required at all times, but the CSF Status Trees are monitored ONLY after it is determined that SI can NOT be terminated.
Monitoring of the CSF Status trees commences as soon as the immediate c.
action steps are completed.
d.
CSF Status Trees are required to be monitored as soon as the procedure is entered and a valid SI is determined to have occurred.
RESPONSE
l We request the answer key be expanded to accept"a" or"b" The question is worded "while performing EOP-E-0" which makes "a" correct also.
REFERENCE:
l
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" Users Guide for T.P. EOP's and Background Documents" 19635:4 Pg. #16 l
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09/11/86 Page 23 NRC Exam Section 4 QUESTION RO 4.16:(SRO Question 7.15)
What plant condition is assumed in the valve / breaker " Normal" position designation for the Operating Procedure! alignment sheets?
RESPONSE
We request the answer key be expanded to accept " Mode 1,"" Mode 2" or " Power Operation." The question did not state as per 0-ADM-201 and the Operators are experienced in the use of alignment sheets for power operation.
REFERENCES:
OP-0202.2, Table 1,5/1/86
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Pag 2 24 NRC Exam Section 4 QUESTION RO 4.19:(SRO 7.20)
List all the valves which reposition as a result of a Phase B ("P") isolation signal.
RESPONSE
We request that the answer key be expanded to include " Containment Spray Pump Discharge Valves"
REFERENCES:
Drawing # 5610-T-E-4510 Sh.1 of 2
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l 09/11/86 Pag 2 25 NRC Exam Section 4 QUESTION RO 4.20:(SRO 7.21)
With rod control in automatic while at 80% power, what are all the immediate operator actions if a single rod drops Kassume no rx trip) as stated in ONOP 1608.1," Full Length RCC Malfunction"?
RESPONSE
We request that the answer key be expanded to include " Verify Turbine Runback" as per ONOP 028.
REFERENCES:
ONOP-028, step 4.5.2,03/28/86 s
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09/11/86 Peg 2 26 NRC Exam Section 4 QUESTION RO 4.24:(SHO 7.25, SRO Requal 4.17, RO Requal 4.17)
Define "Inside the Target Band"in accordance with ONOP-1008.6, ' Operations outside the Target Band'.
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RESPONSE
Expand Answer Key to include"i5% ofTarget"
REFERENCES:
ONOP 1008.6 Pg.1 Sect. 3.2.2,01/29/82
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NRC Eram Section 7 QUESTION SRO 7.19 Besides required notifications, what are the immediate operator actions if you are on t
shift in the control room and the refueling supervisor in the containment reports they dropped a spent fuel element in containment?
RESPONSE
i Request that the Answer Key be expanded to accept the following as an additional answer:
- 1. Sound ContainmentEvacuation Alarm
- 2. Initiate ContainmentVentilationIsolation
REFERENCES:
i E-0, Page 6, Step 12 i
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a 09/11/86 Pcge 28 NRC Exam Section 8 QUESTION SRO 8.22 (SRO Requal 8.15)
List the MINIMUM requirements for each of the necessary electrical systems listed below in order to conduct a reactor startup 4n UNIT 3, assuming Unit 4 is shutdown.
a)
Unit 3 and 4 4160 KV buses b)
Unit 3 480 VAC Load Centers.
c)
Battery Chargers
RESPONSE
We request that the Answer Key be expanded to accept "4 of 4 4160 Volt Buses" as an additional answer to meet the requirements ofTech Spec 3.4.1.A.4
REFERENCES:
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i Tech Spec, Page 3.4-1 4-OP.062, Pages 4 and 24 i
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