ML20210V420
| ML20210V420 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 10/02/1986 |
| From: | Burdick T, Graves D, Lanksbury R, Anthony Mendiola, Miller L, Munro J, Wiens L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20210V405 | List: |
| References | |
| 50-341-OL-86-02, NUDOCS 8610100692 | |
| Download: ML20210V420 (108) | |
Text
. _...... _
.U.S. NUCLEAR REGULATORY COMMISSION i
REGION III.
- Report No. 50-341/0L-86-02(DRS)
Docket No'. 50-341 License No. NPF-33 Licensee: The Detroit Edison Comp ny 2000 Second Avenue Detroit, MI 48224 Facility Name: Enrico Fermi Nuclear Power Plant, Unit'2 Examination Administered At: Enrico Fermi Nuclear Power Plant, Monroe, MI Examination Conducted: July
-31 and August 27-28, 1986
/4/2[0 Examiners:
ns Date A. J.
endiola
_/0/2/[g f bi
/d2/56 L. R. Miller Date' 8 [ 2 /98 J. M. Munro Wbe7' ek Ya~te h?h(dJ t
/#[2[8 L. Wiens A)$Rfg f
Approved By:
T. M. Burdick Section Chief e'
~
geA22882!8ta8ljgi lg ir
g Examination Summary Recualification examination administered on July 29-31 and August 27-28, 19E6 (Report No.50-3417dE~-BB~-W(DTGT)
IfegualTfiFa~tTon examinations were administered to four reactor operators and eight senior reactor operators.
In addition, four. additional senior reactor operators and one additional reactor operator were given a simulator only examination.
Results: All reactor operators and senior reactor operators passed the written, operating, and simulator examinations (as applicable)..
2
m REPORT DETAILS 1.
Examiners
- R. D. Lanksbury _
A. J. Mendiola L. R. Miller
+**J. M. Munro
+D. N. Graves
+L. Wiens
-* Chief Examiner for initial exams (July 29-31,1986)
- Chief Examiner for second set of simulator exams (August 27-28,1986)
+ Denotes the examiners for the second set of simulator examinations 2.
Ex_ amination Review Meeting At the conclusion of the requalification written examination, the questions and answers were given to the facility training staff for review and comment.
Subsequent to this on August 1, -1986, the licensee provided their consnents to the Reactor Operator and Senior Reactor Operator requalification written examinations. The facility consents and the examiner's resolutions to each are on the enclosed attachment.
3.
Simulator Examina_ti_on_s_
Simulator examinations were initially administered to four Reactor Operators (RO) and eight Senior Reactor Operators (SRO) on July 30-31, 1986. The-initial results of these examinations indicated that the four RO's had passed and that five of the eight SR0's had failed. Further assessment of the SR0 examinations revealed that a number of irregularities existed (eg inconsis-tencies between written and simulator examination results; lack of agreement between NRC personnel on the grading the simulator examination results; and a two-week delay by the examiners in reaching a conclusion regarding failure of the simulator examination after informing the chief examiner, superiors and the licensee that there were no significant problems identified). Based on this assessment it was concluded that the results were indeterminate and the decision was made to administer a second set of simulator examinations to the five SR0,s. On August 19, 1986, aConfirmatoryActionLetter(CAL)
(RIII-CAL No.86-005) detailing the above and stating the NRC's understanding that the five SR0's had been removed from any licensed duties pending completion of the second set of examinations was issued.
The second set of simulator examinations was administered on August 27-28, 1986.
In addition to four of the five suspect SR0's, simulator examinations were administered to one additional R0 and four additional SR0's who had not been previously examined. One of the suspect SR0's was not available to take the reexamination and will continue to be removed from licensed duties.
All eight of the SR0's and the one R0 who participated in the simulator reexamination passed. Based on their successful completion of the simulator re-examination the four suspect SR0's were allowed to return to licensed 3
r activities on August 29, 1986. The above, as well as the successful completion of all actions and conmitments in the August 19, 1986, CAL was documented in a September 10, 1986, letter (J. G. Keppler to B. R. Sylvia) to the licensee.
'4.
Exit Meeting On August 1, 1986, an exit meeting was held. The following personnel were present at this meeting:
Licensee:
B. R. Sylvia, Group Vice President C. E. Aldepson, Advisor to the Vice President, Nuclear Operations F. E. Agosti. Vice President, Nuclear Operations J. T. Coleman, Supervisor, Nuclear Operations Training G. R. Overbeck, Superintendent of Operations J. E. Conen, Licensing Engineer S. V. Heard, Operations Coordinator T.E.Lang, Contractor (FRGCorp.)
NRC:
R. D. Lanksbury, Chief Examiner, RIII L. R. Miller, Examiner, NRC Headquarters A. J. Mendiola, Examiner, NRC Headquarters The following general weakness's with regard to the the SR0's were presented:
a.
Some SR0's exhibited difficulty with procedural compliance.
b.
Some SR0's exhibited conmunication difficulties (use of terms such as "the level is high" without elaborating how high the level was or how close to a setpoint, etc.).
On August 29, 1986, an exit meeting for the simulator reexaminations was held. The following personnel were present at this meeting:
Licensee: B. R. Sylvia, Group Vice President R. S. Lenart, Plant Manager G. R. Overbeck, Superintendent of Operat.ons L. C. Lessor, Advisor to Plant Manager S. J. Latone, Director Nuclear Training J. T. Coleman, Supervisor, Nuclear Operations Training G. Reece, Operations Training Coordinator G. E. Abramson, Asst. Operations Engineer, Simulator J. E. Conen, Licensing Engineer NRC:
E. G. Greenman, Deputy Dir., Division of Reactor Projects, RIII C. W. Hehl, Operations Branch Chief, RIII T. M. Burdick, Operator Licensing Section Chief, RIII J. M. Munro, Acting Operator Licensing Section Chief, RII L. Wiens, Examiner, NRC Headquarters D. N. Graves, Examiner, RIII R. DeFayette, Project Engineer 4
m The following general weaknesses with regard to the SR0's were presented:
a.
Procedure implementation by shift supervisors, b.
Connunications by shift supervisors.
In addition, the satisfactory results of all nine SR0's was made known and the licensee was informed that the CAL hold on the four suspect SR0's, preventing them from participating in licensed activities, was terminated.
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ATTACHMENT FACILITY EXAMINATION COMMENTS AND EXAMINER RESOLUTION FOR FERMI II EXAMINATION ADMINISTERED JULY 29, 1986 1.
Question 1.01 Facility Co_nment:
Due to the broad nature of the question, there are numerous answers to this question.
i.e. burnable poison load, rod speed, fuel enrichment etc.
Resolution:
Commented accepted - Answers such as those provided above were accepted.
2.
Question 1.06b. and d.
Facility Coment:_
b.
Depending on local / average flux value and location of the control rods in group 5 vs group 7, other answers may be acceptable with justification, d.
Typically, the operators are prohibited from changing the size of the control rods.
Resolution:
b.
Commented accepted - other answers were considered with adequate justification.
d.
Connent understood, however question was a theory question and ment to test the operators kncwledge of rod worth.
3.
Question 2.02 Facility Comment:
We do not require setpoint memorization of unimportant facts. The purpose of the valves is important, not a value which the operator can neither nionitor nor control.
Resolution:
Coment accepted - Question and answer key have been changed to not require the setpoint of the equalizing valves.
4.
Question 2.03
[ac,11i_ty_ Comment:
There is no air heater or air flow in standby. A comparable answer would be to say that we built a building around the diesels.
Resolution:
Coment accepted - The answer key answer is the flow path as stated in the reference material provided. The answer key does not state that there is air flow. However, for clarity the portion of the answer about the air coolant system is not required for full credit and the answer key has been changed to reflect this. Point values were also redistributed.
5.
Question 2.04.6 Facility Coment:
We normally discuss the negative pressure as leading to failure of the drywell, not giving excess loading to the floor.
Resolution:
Coment accepted - This answer is considered equavalent to that in the answer key.
6.
Question 2.05.a Facil_ity Comment:
An acceptable answer should be "to allow more time between the first and second lift to allow the tailpipe to drain."
Resolution:
Comment accepted - The answer key has been changed to include the above i
answer.
7.
Question 2.06 Facility Comment:
Again, we do not require that operators memorize the less significant setpoints. Also, a review of tech specs shows that:
1.
A HPCI isolation occurs at an exhaust diaphram pressure of 20 psi.
2.
There is an interlock between the Exhaust valve and the ability of HPCI to initiate. See HPCI Student handout.
_Resolu tion:
Comment accepted - The answer key was changed to include the above two answers and to eliminate the requirement to know the rupture disk setpoint.
8.
Question 3.0.2.a. b, and c Facility Comment:
a.
The RFP's will decrease in speed. This is the initial response, all else subsequent or attendant response.
b.
Decrease to minimum speed, all else is subsequent.
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c.
Depending on the assumption of the examinee, either an increase or decrease in speed is possible.
If the pumps are assumed to be operating above 78% speed, they will of course decrease in speed.
If, on the other hand the pumps are running at less than 78% speed, the loss of the suction flow from the heater drains will result in an increase in speed. The limit of 78% will still apply.
Resolution:
- a. and b. Comment accepted - Answer key changed, but answer still requires a proper explanation of the control system response.
c.
Conment understood - Will accept either answer, but answer still requires a proper explanation of the control system response.
9.
Question 5.03 Facility Connent:
This question could be answered correctly as either A or B.
B is correct since if I select A-3/4 with A-1/2 rods at some intermediate position the selected groups of rods will backlight. Also these indications are clearly specified in our Procedure 23.623. We do not require that the operator memorize all of the possible combinations of things that can be done to get a particular light configuration. We concentrate on recognizing the proper lighting display for the proper operation.
Resolution:
Conment rejected - Question states that a normal startup is in progress.
In this case, all A-12 rods must be withdrawn to establish the
" checkerboard pattern" before any A-34 rods can be moved.
- 10. Question 3.04 a and b Facility Conment a.
We feel that the error limiting network is of limited importance in this situation. We feel that 0.75 points should be assigned to the 45% limit and 0.25 points assigned to the Error Limiting network.
b.
The Reactor recirc pump will trip if the valve is not greater than 95% open after the start sequence is complete.
See the same student handout used as reference in the question.
Resolution:
a and b comments accepted - The answer key has been changed to reflect the above.
- 11. Question 3.06.a facility _Connent:
This question is sonewhat misleading.
You are asking for actions where non exist.
3
1 Resolution:
Coment accepted - Part "a" has been deleted.
- 12. Question 3.08 Facility Coment:
We have trained the operators to exercise caution anytime two MSIV's will be closed by emphasizing that this is a situation which may cause a 1/2 scram. We have not emphasized the possible combinations which will actually cause the 1/2 scram.
Resolution:
Coment rejectea - Question is intended to determine if operator knows RPS logic for closure of the MSIV's.
- 13. Question 4.06
)
facijity Coment:
This question is SR0 level. The R0's are not typically involved beyond the preparation of the PCR.
Resolution:
Comment rejected - Answer key gives the steps leading up to, and including writing of, a PCR. Procedure 12.000.07 states that it is the authors responsibility to get the required interim signatures. Therefore you have to know who can sign the PCR.
- 14. Question 4.07.c Facili_tX Coment:
Either the SRM's or IRM's may be used.
Resolution:
Connent accepted - The answer key has been changed to reflect the 4
comment.
- 15. Question 5.02 facili_ty Coment:
As worded, this question may be interpreted as True/ False.
a.
T c.
F b.
F d.
F Concent understood, will accept any correct answer method.
1
- 16. Question 5.04 Facility Consnent:
4 This reference is unknown to us. Please provide us with a copy. Also, the B part of the question should be answered as DECREASE.
Resolution:
The correct reference is the " Decrease in core coolant flow. events study guide," NT/R334/6.0 from the 0C & P course. Comment for Part B accepted 7
and answer key changed.
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- 17. Question 6.01 Facility Coninent:
I This would be true 'if we have a loss of power or as a general statement.
Normally, the battery chargers are operating on a float above the battery voltage so the logic could be considered to be powered from the chargers.
Either true or false should be acceptable in this case.
NRC Resolution:
f Accepted, answer key changed.
2
- 18. Question 6.01 Facility Comen_t:
. According to drawing 2095-7, the signal is sent to the LOGIC but does not actuate it. Again, either true or false appears to be correct.
NRC Resolution:
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Comment rejected.
On Page 14 of PIS E21, it states that two Core Spray pump 3
l running signals are required to send a permissive signal to the ADS logic.
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- 19. Question 6.03.b i
I Facility Comment:
r l
The second part of the answer is inappropriate. The question does not ask why the time delay is set at 117 seconds, it asks why there is a time delay.
NRC Resolution:
{
Coment accepted. Answer changed.
l
- 20. Question 6.03.c l
Facility Comeg:
Should be turned off pumps.
5 i
,ms-.--
.~~~,.-,-,-,,,~.._,__._._.,.,-.._,._,.,.,.e.,~,_.,,--.._,,___---...,.
,_.m.,.,,,....m,,,,...,
NRC Resolution:
Comment accepted. Any action which will prevent the low pressure pumps from starting will be accepted.
- 21. Question 6.04.a Facility Comment:
The RFPs will decrease in speed. This is the initial response, all else is subsequent or attendant response.
NRC Resolution:
Comment accepted. Answer Key changed, but answer still requires a proper explanation of the control system response.
- 22. Question 6.04.b Facility Comment:
Decrease to minimum speed, all else is subsequent.
NRC Response:
Comment accepted. Answer key changed, but answer still requires a proper explanation of the control system response.
- 23. Question 6.04.c Facility Comment:
Depending on the assumption of the examinee, either an increase or decrease in speed is possible.
If the pumps are assumed to be operating above 78% speed, they will of course decrease in speed.
If, on the other hand the pumps are running at less that 78% speed, the loss of the suction flow from the heater drains will result in an increase in speed.
The limit of 78% will still apply.
NRC Resolution:
Comment understood. Will accept rather answer, but answer still requires a proper explanation of the control system response.
- 24. Question 6.06 b Facility _ Comment:
Which local panel? The diesel can be tripped from the local panel adjacent to the diesel.
6
NRC Resolution:
Connent rejected. Understanding that there are two diesel local panels,
- during orals when the operators were questioned about the diesel local control panel, they led the examiners to the control panel in the EDG switchgear room. From this panel the EDG cannot be tripped when it is running in the emergency mode.
- 25. Question 6.06.c Facility Connent:-
"EACH" is misleading.
It implies TEN (10) starts which is false.
NRC Resolution:
This part of the question is deleted. The student handout is specific in wording (that it implies that the air receivers have enough air for 10 starts page B of student handout).
75% of the SR0s answered this question TRUE and some of them stated that the air receivers had enough air for 10 starts during the plant walk-throughs.
- 26. Question 6.06.d Facility Conment:
Should be false. The burn rate for a loaded diesel is 180 to 200 GPH.
The capacity is 210 gallons if at minimum.
NRC Resolution:
Comment rejected. Page 11 of the student handout specific 611y states two hours.
- 27. Question 6.07.a Facility Connent:
Should also accept ERR 0NE0US rod withdrawal in areas of high power density during high power operation. This is stated in the student handout as part of the system.
NRC Resolution:
Connent accepted. Answer changed to reflect the student handout explanation of the purpose of the RBM.
- 28. Question 6.07.c Facility _Connent:
Nulling sequence also causes a bypass of the RBM.
NRC_ Resolution:
Convent accepted. Answer key chan ed.
- 29. Question 6.08 Facility Coment:
This question could be answered correctly as either A or B.
B is correct since if I select A-3/4 with A-1/2 rods at some intermediate position the selected groups of rods will backlight. Also these indications are clearly specified in procedure 23.623. We do not require the operator memorized all the possible combinations of things that can be done to bet a particular light configuration. We concentrate on recognizing the proper lighting display for the proper operation.
NRC Resolution:
Coment rejected. Question states that a " normal" startup is in progress.
In this case, all A-12 rods must be withdrawn to establish the
" checkerboard pattern" before any A-34 rods can be moved.
- 30. Question 7.03.a Facility Coment:
TSC-0BA DSC-RTC (3RD FL TB/AB)
E0F-NOC NRC Resolution:
Concent accepted.
- 31. Question 7.03.b Facility _ Comment:
Plant Manager, Superintendent of Operations, Operations Engineer.
NRC Resolution:
Comment accepted. Answer key changed.
- 32. Question 7.04 Facility Connent:
Also during radiography. Same reference as question.
NR_C Re_ solution:
Comment accepted. Answer key changed.
8
-33.
Question 7.06 d Facility Comen_t:
Steam Line Pressure Low -- Alarms in two locations. APRMs alarm in six places.
NRC Resolution:
Coment rejected. Procedure details what specific alarms must clear.
- 34. Question 8.04 Facility Coment:
There are several other conditions external to the three specified on that particular page of 12.000.07.- For instance, the procedure must be approved by DRSO within 14 days, can't change the acceptance criteria (setpoints).
NRC Resolut_i_o_n:
Coment partially accepted. Question tests SRO knowledge of items he must consider before approving any general change. Answers will be reviewed to determine if adequate knowledge of the procedure is displayed.
- 35. Question 8.06 Facility Comment:
This method is normally used in the " transient" condition.
It is acceptable but not the desired method.
NRC Resolution:
Coment accepted. No changes made.
- 36. Question 8.07.c Facility Coment:
There is no wrong answer.
NRC Resolution:
Coment accepted. This part of the question is deleted.
- 37. Question 8.07.d Facility Coment:
False - Generally the ones that are locked are locked with a keylock switch. The question can be misread quite easily.
q 9
NRC Resolution:
Comment accepted. This'part of the question is deleted.
- 38. Question-8.09.b Facility Coment:
True or False.
If fire occurs onsite inside the protected area the.
answer is True.
If fire occurs onsite inside the controlled area but outside the protected area, the answer is false.
NRC Resolution:
Comment accepted. Answer key changed.
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I U.
S.
NUCLEAR REGULATDRY COMMISSION SENIOR REACTOR OPERATOR REQUALIFICATION EXAMINATION FACILITY:
_EE8M1_2___ _
i REACTOR TYPE:
_BWB-GE3_________________
f DATE ADMINISTERED: _@bl9Zl22__ - -==__________
4
_MENQ1gLA _A ____________
EXAMINER:
3 z
j CANDIDATE:
__h_h_3 _k ___
JNSI60CIlgNS_IQ_Q8NDID91Ej.
Read the attached instruction page carefully.
This examination replaces the current cycle facility administered requalification examination.
Rotraining requirements f or failure of this examination are the same as l
for failure of a requalification examination prepared and administered by j
your training staff.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70%
in each category and a final grade of at least 80%.
Examination papers will be picked up four (4) hours after the examination startr..
% OF CATEGORY
% OF CANDIDATE'S CATEGORY
__YBLUE_ _IQIGL
___SCQBE___
_y66UE__ ______________GGIEGQBy_____________
_1Dz99__ _20the
.._______ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_19z59__ _29122
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, 1
AND INSTRUMENTATION
_19199__ _20xb3
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL I
_19199__ _22123
________ B.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS Totals
_30159__
Final Grade
?
lt All work done on this examination is my own.
I have nei ther given 1
nor receivad aid.
Candidate's Signature
'i 1
D 1
1
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of thir examination the following rules apply:
$g i-1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
i fj 2.
Restroom trips are to be limited and only one candidate at a time may l
1 eave.
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 gely to facilitate legible reproductions.
IJ 4.
Print your name in the blank provided on the cover sheet of the examination.
4 5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
f, 7.
Print your name in the upper right-hand corner of the first page of gach i
section of the answer sheet.
B.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a gew page, write gnly gn gne side of the paper, and write "Last Page" on the last answer sheet.
9.
Number each answer ac to category and number, for example, 1.4, 6.3.
- 10. Skip at least thtee lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face j
down on your desk or table.
i
- 12. Use abbreviations only if they are commonly used in facility litgtatute.
- 13. The point value for each question is indicated in parentheses after the 4
J question and can be used as a guide for the depth of answer required.
assumptions used to obtain yl an annwer
- 14. Show all calculations, methods, or to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not cicar as to intent, ask questions of the gxeminet only.
i
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination.
This must be done after the examination has been completed.
A,4 N.
.d
- 10. When you complete your_ examination, you shall:
Assemble your examination as follows:
a.
(1). Exam questions on top.
)
(2)
Exam aids - figures, tables, etc.
the answer.
(3)
Answer pages including figures which are part of lI b.
Turn.in your copy of the examination and all pages used to answer the examination questions.
c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
o d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
y' s
.1 14 L '(
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PAGE.
2 Dz__IME9By_QE_NUC6E88_EQWEB_E68NI_QEEBBIlgN _ELUlp@2_8ND 3
ISEBdQQyNOMICS QUESTION 5.01 (2.00)
For each of the following conditions describe the effect (INCREASE, DECREASE, or REMAINS THE SAME) on Critical Power assuming all other variables remain unchanged?
(0.50 EA) l ASSUME NORMAL FULL-POWER OPERATING CONDITIONS.
?
1.
Inlet subcooling is DECREASED.
2.
Reactor Pressure is INCREASED.
3.
A::i al Power Peak is RAISED.
4.
Coolant flow rate is INCREASED.
QUESTION 5.02 (1.50)
The reactor trips from full power, equilibrium XENON conditions.
Five hours later the reactor is brought critical-and power level will be maintained on range 5 of the IRMs for several hours to allow test ing.
Which of the following statements is CORRECT concerning control rod motion (1.50)
IMMEDIATELY after startup?
a.
Rods will have to be withdrawn due to Xenon build-in.
reactor b.
Rods will have to be rapidly inserted since the critical will cause a high rate of Xenon burnout.
c.
Rods will have to be inserted since Xenon will closely f ollow its normal decay rate, d.
Rods will approximately remain as is as Xenon establishes its equilibrium value f or this power level.
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'l CATEGORY 05 CONTINUED ON NEXT PAGE
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Dz__IBE98Y_9E_NWGLE98_E9 WEB _EL801_9EEB8II9N _ELUlpg3_@ND 2
ISEBD99XU8dlGE QUESTION 5.03 (1.50) feedwater subcooling.
What is.
[
A_BWR is operating at 985 psig with 217 F of (1.50) the f eedwater temperature?
STEAM TABLES ATTACHED FOR YOUR REFERENCE i
c.
321.5 F b.
323.8 F c.
325.7 F d.
327.5 F QUESTION 5.04 (1.50)
Assume the reactor is operating at 100% power in 3-clement control and one recirculation pump trips.
Indicate how each listed parameter would REMAIN THE SAME).
i INITIALLY respond (INCREASE, DECREASE, or (0.50) a.
Reactor power (0.50) b.
Reactor pressure (0.50)
- c. Feedwater flow i
DUESTION 5.05 (1.50) 1 the VOID COEFFICIENT of REACTIVITY change l
How does the MAGNITUDE of 3
(MORE NEGATIVE, LESS NEGATIVE, OR REMAINS THE SAME) for each of the
!I f ollowing changes in core condition?
1.
(0.50)
I INCREASE in core void fraction a.
(0.50)
- b. DECREASE in fuel temperature (0.50) c.
INCREASE in core age
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t f7k i'I CATEGORY 05 CONTINUED DN NEXT PAGE *****)
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5 __IBEQBy_QE_NyCLg66_EQWEB_EL6Ml_QEEBSIlgN _ELQ1pg2_6ND 2
t IHEBdQQyN9Migg i
QUESTION 5.06 (1.50) j Concerning control rod worth during a reactor startup with 100% peak Xenon versus a startup with a Xenon free condition, which ONE of the following
( 1. 50) etatements is CORRECT 7 PERIPHERAL control rod worth will be HIGHER during the Xenon free i
a.
4-startup than the 100% peak Xenon startup.
b.
CENTRAL control rod worth will be HIGHER during the 100% peak Xenon startup than during the Xenon free startup.
l c.
PERIPHERAL control rod worth will be LOWER during the Xenon free j
startup than during the 100% peak Xenon startup.
1 d.
Both CENTRAL and PERIPHERAL control rod worth will be the same 1
I regardless of the core Xenon concentration.
QUESTION 5.07 (2.00)
Match the Failure Mechanism f rom column
'1' and the Limiting Condition from column
'2" with the associated Power Distribution Limits (a-c) below.
Your (2.00) should have letter-number-number.
answer Linear Heat Generation Rate (LHGR) a.
- b. Average Planer Linear Heat Generation Rate (APLHGR) i
- c. Minimum Critical Power Ratio (MCPR) 1-FAILURE MECHANISM 2-LIMITING CONDITION
- 1. Fuel clad cracking due 1.
1% Plastic Strain to lack of cooling caused by DNB.
l
.l
- 2. Fuel clad cracking due to 2.
Prevent Transition Boiling high stress from pellet expansion
- 3. Gross clad f ailure due to 3.
Limit clad temperature to 2200 F decay heat and stored heat f ollowing a LOCA 4
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b PAGE 5
D___ISE9By_QE_NgC6E88_EQNEB_E68NI_9EEB811gN3_E6UlpS,_6NQ IMEBd9DYN6NICS QUESTION 5.08 (2.00)
State whether each of following statements are TRUE OR FALSE:
- a. As condenser vacuum is INCREASED (pressure decreased), MORE (0.50) energy can be extracted from the steam.
- 6. The use of feedwater heaters places the reactor FARTHER f rom (0.50) transition boiling than NOT using f eed water heaters.
Air ejectors would NOT be needed if the condenser was (0.50) c.
absolutely air tight.
The main condenser uses the conduction mode of heat transfer d.
(0.50) to reject heat to the circulating water system.
OUESTION 5.09 (1.50)
For each of the events li sted below, state which REACTIVITY COEFFICIENT responds first and HOW it adds positive or negative reactivity to the event.
(0.50)
One Safety Relief Val ve opens at 100% power.
(0.50) a.
b.
Control rod drops at 100% power.
(0.50) c.-
I sol ati on of a feedwater heater string at power.
i' I
5 i
I!
i 1
I
'}
i
(***** END OF CATEGORY 05
- )
3 PAGE 6
6 __EL6MI_EYSIEdQ_QEQ1QN2_CQNIBQL3_6NQ_lNSIBUdEN16I1QN t
i DUESTION 6.01 (2.00)
ANSWER THE FOLLOWING OUESTIONS EITHER TRUE i
REGARDING THE CORE SPRAY SYSTEM, (0.50 EA)
OR FALSE.
i 1.
When the Core Spray Pump control switch is placed in the 0FF/ RESET f
position, a valid core spray initiation signal will start the pump.
If a core spray pump is running and its associated electrical power 2.
bus auto transfers to the emergency diesel generator, the core spray pump will trip and will need to be restarted by the operator.
1 3.
The core spray logic for the "A" loop is powered from the Division I 1
battery.
4.
If one of the Core Spray pumps start and is running, a permissive signal is sent to the ADS logic system.
QUESTION 6.02 (2.00)
REGARDING THE REACTOR RECIRCULATION FLOW CONTROL SYSTEM:
The Speed Limiter (Dual-Speed Demand Limiter) gives a high and a low the master controller.
What are these two limits and what a.
limit on (1.00) are their bases during normal critical operation?
b.
The purpose of the #2 speed limiter is to limit recirculation pump to a maximum value of 45% if what speed by limiting the demand signal (1.00) conditions exist?
i DUESTION 6.03 (1.50)
REGARDING ADS SYSTEM LOGIC:
a.
What is the basis for requiring both a Level 1 and a Level 3 trip (0.50) prior to ADS initiation?
i
( less than or equal to 117 i
Why is there a time delay in the logic (0.50) i b.
seconds) before actuation of the ADS?
c.
If the ADS timer has begun to count down, LIST TWO ways the operator (0.50) can manually prevent ADS blowdown.
lm CATEGORY 06 CONTINUED ON NEXT PAGE *****)
(*****
PAGE 7
' 6 __EL6UI_SYSIEUS_pEgigN3_GQNI6963_999_193IBUDENISIlgd t
QUESTION 6.04 (1.50)
Feedwater Level Control System is in 3-element control using reactor l evel detector channel A.
Reactor power is at.85% steady state.
Using the attached Feedwater Level Control System block diagram, state how (0.50 EA) the system will respond initially to the following failures.
.B Feedwater %wngr Flow signal fails HIGH.
a.
b.
LOSS of Control Signal to B Reactor Feed Pump Speed Controller.
Check valves between the #5 heaters and the heater drain flash tanks c.
CLOSE.
l OUESTION 6.05 (1.50)
For each HPCI' system component failure listed below, STATE:
1.
If HPCI will auto inject on a valid injection signal.
2.
ONE adverse effect or consequence of the component f ailure during system operation.
(0.75) a.
The gland seal e>thaust pump f ails to operate
?
the system conYitions b.
The minimum flow valve fails to auto open when (0.75) require it to be open.
r DUESTION 6.06 (1.50)
REGARDING THE EMERGENCY DIESEL GENERATORS, ANSWER THE FOLLOWING TRUE OR (0.50 EA)
FALSE:
a.
If the LOCAL / REMOTE switch (located on the local, control panel) for a diesel is switched to the LOCAL position, the diesel will start on a valid start signal.
b.
If a EDG is running in the emergency mode, it can be tripped at the local control panel.
c.
The day tanks for each EDG have enough fuel to support 120 minutes of full load operations.
Li
-4 CATEGORY 06 CONTINUED ON NEXT PAGE *****)
(*****
.{_
w--
PAGE
-8 61__P68NI_gy@IgMS_DESJGN _CQNIB063_8ND_INSIBUDENIBIlON 3
QUESTIDN 6.07 (2.50)
The Rod Block Monitor protects the~ fuel from damage during operation.
(0.50)
Y What EVENT'does the Rod Black Monitor protect.for7 a,
b.
Referring to Technical Specifications, when is the Rod Block Monitor f'
(0.50) required to be in operation?
List THREE RBM bypasses and briefly state when the bypass would be c.-
(1.50) i used.
l
'OUESTION 6.08 (1.00) f A' normal cold plant startup is in progress with the RSCS Sequence Mode i
Selector (SMS) Switch in " Withdraw."
Which of the following depicts the FEWEST number of operator actions that will cause the dim backlights f or the "A-12" sequence rods to EXTINGUISH and the diu backlights for the (1.00)
"A-34" sequence rods to ILLUMINATE 7 Fully withdraw all "A-12" sequence rods; Select "A-34" on the Rod a.
Sequence Selector (RSS) Switch, b.
The position of the "A-12" sequence rods is immaterial; Select "A-34" on the Rod Sequence Selector (RSS) Switch.
c.
The position of the "A-12" sequence rods is immaterial; Select "A-34" on the Rod Sequence Selector (RSS) Switch, and Select any "A-34" Rod.
d.
Fully Withdraw all "A-12" sequence rods; Select _"A-34" on the Rod 2
Sequence Selector (RSS) Switch, and Select any "A-34" rod.
}
QUESTION 6.09 (1.00) j Following a Control Room evacuation, the plant shutdown is being controlled from the Remote Shutdown Panels.
During the shutdown, system isolations i
occur.
For each of the f ollowing i sol ati ons, list j
a)
If the isolation can be bypassed.
b)
If it can be bypassed, on which Remote Shutdown Panel are the bypasses.
i i
1.
HPCI isolates on High Flow 4
RCIC isolates on RCIC room High Ambient Temperature 2.
3.
RHR Shutdown Cooling Isolation on High Pressure W
[
(***** END OF CATEGORY 06 *****)
PAGE 9
Zt__EBgCEQUEEQ_ _NQBd6(3_@ENQBM862_EMEBQENCy_ANQ 88Q196901C86_CQNIBQL QUESTION 7.01 (1.00)
I-As part of the scram procedure (20.000.21), the operator is directed to insert the SRMs and the IRMs.
{
EXPLAIN how these systems could be used to provide a crude indication of water level after-a scram if level cannot,be confirmed by normal (1.00) instrumentation.-
l 1
1 4
QUESTION 7.02 (2.50) i LIST.the FIVE entry conditions with their setpoints for PRIMARY CONTAINMENT I
(0.50 EA) l CONTROL, Procedure 29.000.03.
't OUESTION 7.03 (2.00)
,{
Concerning the Emergency Plan implementation:
STATE the locations of the lechnical Support Center,-the Operations t
a.
(0.50 EA)
Support Center'and the Emergency Operations Facility.
a the NSS assumes the duties of-the fj b.
Upon initiation of an emergency, (0.50)
'1 Emergency Director until relieved by whom?
-l
?
i
~l OUESTION 7.04
-(2.00)
! 1 j
According to procedure 12.000.13, " Radiation Work Permits", list FIVE
-t
.- i conditions when a Specific Radiation Work Permit is required.
(0.40 EA)
?
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et
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(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
y PAGE 10
- 2___EBgCEDUBES_ _NQBd863_@pNQBd862_EDEBggNCy_6NQ 609196ggICe6_CgNIBQ6 1
QUESTION 7.05 (2.00)
Assume an emergency exists which requires _ control room evacuation and room (procedure 20.000.19):
plant shutdown from outside the control a.
Except for necessary announcements, what THREE operator actions should be performed.if possible prior to leaving the control room?
(1.50) from is the procedure recommended method to scram the reactor b.
What
( 0. 50) outside the. control room?
i o
l DUESTION 7.06 (2.50)
THE FOLLOWING OUESTIONS PERTAIN TO A COLD STARTUP TO RATED POWER:
Referring to the Plant Startup Master Checklist:
s.
On the Checklist, who decides if a system lineup is required or not?
(0.50) or his delegate may direct b.
TRUE OR FALSE -- The Operations Engineer the NSS or NASS to pl ace the OE's initials on the Checklist "per (0. 50) telecon".
Referring to the Startup from Cold Shutdown to Rated Power Checklist:
l According to the Checklist, when should the Reactor be declared l
c.
(0.50)
I critical?
i State THREE alarms which must be cleared and verified just prior to d.
(0.33 EA) placing the Reactor Mode Switch in RUN.
1 1
. ]!
QUESTION 7.07 (1.00) i the At what point during a LOSS OF STATION AND/OR CONTROL AIR does the reactor?
(1.00) procedure direct you to manually scram q
}
e
.i
.}-
CATEGORY 07 CONTINUED ON NEXT PAGE *****)
gj
(*****
-l
'PAGE ' 11.
'7-
-PROCEDURES - NORMAL,_ADNQBMSL _gMEBQENgY_AND 3
889196991989_CgNIBg6 QUESTION 7.08 (2.00) l While perf orming the immediate actions of procedure 20.000.02, ABNORMAL I
' RELEASES OF RADIOACTIVE MATERIALS, a Reactor Building Vent Exhaust Radiation Monitor Upscale Trip occurs.
LIST the FIVE' automatic. actions you (0.40 EA)-
must verify.
l 4
)
i I
t i
1 t
i 4
.I
'i i j.
y
.}j b
.. j r
k!
..l
(***** END OF CATEGORY 07 *****)
tjj
PAGE 12 Ez__09M101SIBeIlyE_eggCEQUBES3_CQNQlligN@2_@Np_ LIM 11811gN@
i QUESTION B.01 (1.50)
"C" Tip Ball valve did not auto close.
The TIP Following a TIP Trace, thejogging to close this valve due to a sticking limit machine requires extra l
ewitch.
The ball valve is subsequently closed.
(0.50)
)
STATE whether Primary Containment Integrity is satisfied.
1 a.
1 I
Is additional action required by Tech. Specs?
If so, STATE.the ACTION J
b.
EXPLAIN the TS basis for your decision.
j required.
If not, (1.00)
APPLICABLE TS ARE ENCLOSED FOR REFERENCE NOTE:
d DUESTION 8.02 (3.00)
Plant.
a LIST the FOUR TECHNICAL SPECIFICATION SAFETY LIMITS for the Fermi 1
(0.75 EA) 9 OUESTION 8.03 (1.50) a normal r eact or startup.
At 10%
You are the NSS during a shift performing the Rod Worth Minimizer is and the mode selector switch in STARTUP, INOP.
You station the second licensed operator at the reactor power console to verif y compliance with the required rod sequence check declared e
d control off list.
You then continue the startup.
j.
(0.50) 1 Are you in Technical Speci fication violation?
r answer using as necessary the attached TS excerpts.
(1.00) i DOCUMENT your
-j-i 5-
{
]
Dl!EST ION 8.04 (2.00) is performing a surveillance to the HPCI system and,
..q The reactor operator some procedural steps become impossible to i
due to a system modification, is determined that multiple changes are needed.
q perform.
It
~ l this condition can multiple temporary changes be made to the
)
a.
Under (0.50) fj HPCI surveillance procedure?
b.
In general, what THREE conditions must be f ollowed when making(0.50 EA) temporary changes to the procedure ?
.V si; CATEGORY 08 CONTINUED ON NEXT PAGE
- )
(*****
=
PAGE 13 B z__8DMINi@l8@IlVE _EBQQE DQBE@ t _CQNDillQN S z_6NQ _L lMil@llgN S
'OUESTION B.05 (2.00)
Preparations have been completed for the startup of the reactor, with the exception of last minute repairs to a jet pump instrument circuit that'are The I&C Technician tells'you that his repair work will still in pr ogress.
finished in less than one hour and the Technical Specification LCO be allows operation f or up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with one or more jet pumps inoperable.
Can you proceed with the Reactor Startup?
If so, STATE the TS' actions required.
If not, EXPLAIN the TS basis for your decision.
( 2.' 00)
OUESTION 8.06 (1.50)
ACCORDING TO THE FERMI STANDING ORDERS, ANSWER THE FOLLOWING TRUE'OR FALSE:
(0.50 EA) a.
Whenever a SRM channel is declared inoperable, it shall remained bypassed unless required to be un-bypassed for purposes of testing and troubleshooting or the channel is declared operable.
b.
If a procedure calls for an independent verification, the operator ordered to mal:e this verification may observe the initial operator during the performance of the step.
c.
The NSS can sign a EDP as "NSS COMPLETE" after the PSE has been contacted and veri fied that no verbal ECR's are outstanding against the EDP.
QUESTION B.07 (1.00)
ACCORDING TO FERMI ADMINISTRATIVE OPERATIONS PROCEDURES, ANSWER THE(0.50 EA)
FOLLOWING TRUE OR FALSE:
a.
If a Motor Operated Valve (MOV) fails to operate due to tripped thermal overloads, the thermal overloads may be reset after a 15 minute cooling period.
b.
If the Motor fuses for a MOV blow, they can be replaced and the valve operation attempted one more time.
If they blow again, a PN-21 shall be initiated.
CATEGORY 08 CONTINUED ON NEXT F' AGE *****)
(*****
]
PAGE 14 az__0DMINigIBSIlyE_BBgCE99 beg 3_CQN91119Ng3_8N9_LIMIIBIlgNS QUESTION B.08
(.50) tells you that he wishes to adjust C SRM channel output, A I&C technician it responds with the necessary range and accuracy to as necessary, so that the known Source count.
Technical Specifications defines this as a (0.50) a.
Channel Check b.
Channel Functional Test c.
Channel Calibration i-i d.
Logic System Functional Test DUESTION 8.09 (1.00)
Consider the following situations. Identify ONLY those which are REPORTABLE, WITHIN ONE HOUR, to the NRC.
APRM F f ails downscale at 75% power. All other APRMs are functional.
a.
Large grass fire on site requiring outside fire-fighting assistance.
b.
c.
Car accident on site injures two maintenance men.
d.
Rainfall damages the Technical Support Center with heavy water damage.
I i-i
.. j i
i 5
il
(***** END OF CATEGORY 08 *****)
- )
PAGE 15 Q __IBE981_QE_NQCLE@8_EQWE8_E(@N1_QEg8@l1QN _E(Q1pS _8NQ 1
3 l
IBEBdQQyN@dlC@-
t
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI 2 ANSWER 5.01
-( 2. 00 )
1.
DECREASE
.2.
DECREASE
/
3.
DECREASE
~f 4.
INCREASE REFERENCE NUCLEAR POWER PLANT ~ THERMAL SCIENCES ANSWER 5.02 (1.50)
A -- Rods will have to be withdrawn due to Xenon-build-in.
REFERENCE REACTOR THEORY FUNDAMENTALS ANSWER 5.03 (1.50) d i.
REFERENCE STEAM TABLES
,4 i
ii
!i ANSWER ~
5.04 (1.50) 1 (0.50)-
a.
Decrease (0.50) l b.
Decrease (0.50)-
c.
Decrease REFERENCE Decrease in core coolant flow events study guide NT/R334/6.0 f rom the DC&P' i
,l course 1
i
.e' $
k 9
s lA
V.
PAGE 16
'5.
THEORY OF' NUCLEAR POWER PLANT OPERATION _FLUIQSS_ANQ 1
IBEBMODYNOMICS
' ANSWERS -- FERMI 2
-86/07/29-MENDIOLA, A.
ANSWER 5.05 (1.50) 1 a.
More negative.
b.
.Less negative.
c.
Less negative.
REFERENCE REACTOR THEORY FUNDAMENTALS F
i ANSWER 5.06 (1.50)
C T
REF'ERENCE REACTOR THEORY REVIEW CHAPTER 10 REQUAL QUESTION NUMBER O2-02-OO-01-03-E31 ANSWER 5.07 (2.00)
FAILURE MECHANISM LIMITING CONDITION A.
LHGR 2
1 B.
APLHGR 3
3 C.
MCPR 1
2 (6 ANSWERS AT O.33 EACH)
REFERENCE NUCLEAR POWER PLANT THERMAL SCIENCES i
f-ANSWER 5.08 (2.00)
(0.50) a.
True (0.50) b.
False (0.50) c.
False (0.50) d.
True i
REFERENCE NUCLEAR POWER PLANT THERMAL SCIENCES
PAGE 17 Dz__IMEQBy_gE_NQCLE88_BQWEB_E66NI_9EEB611gN3_E(ylpg3_9ND IUEE.UgpyN6dICS
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI 2 ANSWER 5.09 (1.50) increase in i
Void Coefficient.
The decrease in pressure causes an P
a.
voids adding negative reactivity to the core.
b.
Fuel Element Temperature Coefficient.
Rapid addition of positive causes an increase in power thus reactivity due to the rod removalThis causes negative reactivity to be increasing fuel temperature.
added to the core, Moderator Temperature Coefficient.
The removal of feedwater heating causes a decrease in feedwater temperature adding positive reactivity.
c.
REFERENCE REACTOR THEORY FUNDAMENTALS
- a h
n i
6 3
i i
-I 11 i
I 1, 9
PAGE 18 st__Eb8NI_EYSIEMS_DEE1@N _CQNIBQL1_QND_lNSIBUMENI911QU i
~B6/07/29-MENDIDLA, A.
ANSWERS -- FERMI 2 r
ANSWER 6.01 (2.00) 1.
FALSE 2.
FALSE j
3.
TRUE or FALSE 4.
FALSE' s-REFERENCE CORE SPRAY STUDENT HANDOUT PIS E21 REV 4 i
ANSWER 6.O2 (2.00) a.
The upper limit of 102.5% ensures that thermal limits are not exceeded and fuel channel bowing does not occur.
l The lower limit of 45% ensures than there are no flow instabilities from the master due to controlling both recirculation pumps c ontr ol l er..
b.
- 1) A or B reactor feed pump is not running AND
- 2) Reactor water level is less than level 4 (192.5")
REFERENCE REV 4 RECIRCULATION FLOW CONTROL SYSTEM STUDENT HANDOUT D31 ANSWER 6.03 (1.50)
To confirm that a low vessel level does indeed exist, thus preventing a.
a spurious initiation of'the ADS.
b.
To allow enough time for the HPCI System to recover l evel, (yet not so late that the LPCI and CS systems are not too late to prevent fuel damage).
Reset the Time Delay with the push button c.
Prevent Low Pressure Injection pump starts t
REFERENCE
[
ADS STUDENT HANDDUT PIS B21-04 REV 4 i,
i I
E g
-i i
y PAGE 19 f
-st__P68NI_@y@lEdg_DE@lGN,_CgNIBg63_8ND_1NSIBUdENIGIlgN
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI 2 ANSWER 6.04 (1.50)
(Reactor l evel will decrease) due to the level control system having a feed flow the result is a.
steam flow / feed flow error with steam flow <
a decrease in,the speed of the reactor feed pump turbines.
rj b.
Reactor level should drop as the B'RFT will run down to minimum speed, (1660 RPM).
The Control System will see a flow mismatch and attempt to recover it with the A RFT.
Limits feed Pump speed to 78%. (Level will decrease until feed flow c.
equals steam flow.)
REFERENCE REV 4 REACTOR VESSEL LEVEL CONTROL SYSTEM STUDENT HANDOUT PIS N21 t
ANSWER 6.05 (1.50) a.
Will Inject Turbine seal leakage could result in airborne activity in HPCI room.
b.
Will Inject Possibility of pump overheating REFERENCE HPCI STUDENT HANDOUT PIS E41 1
ANSWER 6.06 (1.50) a.
TRUE b.
FALSE c.
TRUE REFERENCE EDG and AUX SYSTEMS STUDENT HANDOUT R 30 REV 3
[!
t l.s
7 P
l PAGE. 20
'6s.__PL6NI_SYSIEd@_DE@lGU3_CONIBg63_8ND_IN@lBUDENI8IlgN
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI 2 E
ANSWER 6.07 (2.50) j Inadvertent rod withdrawal or Erroneous rod withdrawal in areas a.
RBM -
of high power density during high power operation.
b.
RBM is required greater _than to equal to 30% power Joystick bypasses one channel if INOP or for maintenance c.
Less than 30 % power RBM is not required selected the RBM is unnecessary because edge rod Edge rod - if
- i limit.
withdrawals do not produce conditions which approach a thermal T.
Nulling Sequence
.c (Three answers required, 0.50 each)
REFERENCE FERMI REQUAL QUESTION O2-02-OO-01-01-E25 REV 1 FERMI REQUAL-OUESTION 02-02-OO-01-01-E35 REV O ANSWER 6.08 (1.00) a.
REFERENCE ROD SEQUENCE CONTROL SYSTEM STUDENT HANDOUT C11-09 REV 4 i
Li]
ANSWER 6.09 (1.00)
'l 1.
Cannot be bypassed 2.
Can be bypassed with switches on Division I panel 3.
Cannot be bypassed REFERENCE REMOTE SHUTDOWN SYSTEM STUDENT HANDOUT C35-OO AND C36-OO REV 4
'l RHR SYSTEM STUDENT HANDOUT Eli REV 4 i
>=;
'i 5
a '
PAGE 21 Zt__ESQQEQQBES_ _NQBd661_6ESQBd6bi_EMEEGENQY_6NQ
.BG9196991GG6_G9 NIB 96
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI 2 t
(
ANSWER 7.01 (1.00)
By: observing the neutron level while moving.the nuclear instrumentation, b
the operator can determine where the water level is.
A significantly HIGHER count rate would be seen for the UNVOIDED areas of the core as ( 1. 00) opposed to the VOIDED areas of the. core.
1 REFERENCE I
PROCEDURE 20.000.21 AND MITIGATING CORE DAMAGE ANSWER 7.02 (2.50) a.
Torus water level above +2 (124,220 ft )
3 b.
Torus' water level below -2 " (121,080 ft )
Torus water-average temperature above 95 F c.
d.
Drywell atmosphere average temperature above 135 F e.
Drywell pressure above 1.88 psig REFERENCE Procedure 29.000.03 Rev 4 i
ANSWER 7.03 (2.00)
a.
Plant Manager, Superintendent of Operations, or the Operations Engineer i,g REFERENCE EP-110
/
i i
.I
m PAGE_ 22 Z___E89GEQUBEE_ _NQBMOLx_8@NQBM663_EMEBGEUQY_@NQ 86919L901G66_GQNIBQL
-86/07/29-MENDIOLA, A.
ANSWERS -- FERMI'2
\\
ANSWER 7.04 (2.00)
Contaminated Area ~> 10000 dpm/100 cm O.
r b.
Airborne Radioactivity areas i
Neutron Radiation area exposure c.
d.
High Radiation area exposure Unknown conditions in an area te be entered e.
instrumentation'which contain
~
f.
Maintenance of equipment, controls, or radioactive material g.
Radiography REFERENCE 12.000.13 ANSWER 7.05 (2.00)
Place the reactor mode switch to SHUTDOWN (0.5/ea.)
a.
Depress both scram pushbuttons Arm and depress main turbine trip pushbutton from the relay room by taking one operable APRMs l
b.
Scram the Reactor Mode Switch out of operate in Division I (A,C,E) and one in Division II (B,D,F).
a I
REFERENCE Procedure 20.000.19 i
n j.
ANSWER 7.06 (2.50)
Operations Engineer or his designee 1
a.
f b.
TRUE When the-neutron count rate increases at a constant period with no
.]
c.
Control Rod movement.
d.
APRM DDWNSCALE NSSS MN STM LINE LOW PRESS CH A/C TRIP NSSS MN STM LINE LOW PRESS CH B/D TRIP
'k
=l
5 i
PAGE 23 Zz__EBgCgDuggS_:_NQ8d@LS_@@NQEd@L3_EMEB@ENCY_@ND BOQ196991CBL_CQN1BQL
-86/07/29-MENDIDLA, A.
ANSWERS -- FERMI 2 REFERENCE PROCEDURES 22.000.03 AND 22.000.01 ANSWER 7.07-(1.00)
Rod Drift indication Scram valve pilot / air-header High/ Low and Control I'
Ifis received on more than one control rod.
REFERENCE Procedure 20.129.01 i
~ ANSWER 7.08 (2.00)
RB Ventilation System Tripped 1.
RB Div I and II Supply and. Exhaust Isolation Valves close 2.
Primary Containment Purge and Vent Valves Close 3.
4.
SBGT Auto Starts 5.-
CC-HVAC System aligns to the Recirculation Mode.
REFERENCE PROCEDURE 20.000.02 j i-
.?.i i
s e
a
'9 I
PAGE 24
.ai__8DMIN1EIB611YE_EBOGEQUBEEi_QQNpillgN@i_6ND_Lidll611QUE
-86/07/29-M5NDIOLA, A.
, ANSWERS -- FERMI 2 A
d ANSWER 8.01 (1.50)
J.
Primary Containment Integrity is NOT Satisfied.
The Ball valve must be deactivated in its isolated position.to satisf y the TS requirement.
.4 REFERENCE TS 3.6.1.1 and T S 4. 6.1.1 (note) 4
^
ANSWER 8.02 (3.00) a.
Thermal power shall not exceed 25% of rated thermal power with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, b.
The minimum critical power ratio shall not be less than 1.06 with the reactor steam dome pressure greater than 785 psig and the core flow greater than 10% of rated flow.
c.
The reactor coolant system pressure as measured in the reactor vessel.
steam dome, shall not er:ceed 1325 psig.
d.
The reacter vessel water level shall be above the top of the active irradiated fuel.
REFERENCE Tech.
Specs.
2.1.1, 2.1.2, 2.1.3, and 2.1.4 ANSWER.
8.07 (1.50) o NO TS violation TS 3.1.4.1 action statement provides that TS 3.0.4 is not appl i c abl e.
Thus we can continue the startup and enter Condition 1 after we have stationed g
the second operator.
a REFERENCE TS 3.1.4.1 AND 3.0.4 a,
t i l, e
if
- i
- - ~ '
u
'PAGE 25 Et__8DUIU1E18811ME_BBQQEQUBESx_CQUQlllQNQ3_QNQ_blMlIGIlgNQ
-86/07/29-MENDIOLA, A.
ANSWERS'-- FERMI'2 ANSWER 8.04 (2.00) a.-
YES b.
Changes do not. change the ~ intent of the procedure
[
Change:is approved by two knowledgeable members of the plant staff, one of whom is a SRO license holder Upon completion of the performance of-the procedure, the changes will be submitted as a' temporary PCR.
(Other combinations of answers are possible as long as adequate knowledge is displayed)
. REFERENCE POM 12.000.07 ANSWER 8.05 (2.00)
NO into an operational condi tion shall not be made TS 3.0.4 states that entrythe LCO are met without rel2ance on the provision unless the conditions for contained in the action requirements.
REFERENCE TS 3.0.4 I
ANSWER 8.06 (1.50) a.
TRUE b.
.TRUE c.
TRUE 3
REFERENCE y
FERMI STANDING ORDERS 86-12,86-2,06-1 i
ANSWER 8.07 (1.00)
A.
TRUE B.
FALSE
..y 1
.' j J
PAGE 26 Hi__0Dd1NI@IBQllyg_EBgCggdBE@i_CQNpillgN@3_6Np_LidlI@IlgNS-i
-ANSWERS -- FERMI 2'
-86/07/29-MENDIOLA, A.
~
REFERENCE 21.000.01 AND 21.000.14 ANSWER 8.08
(.50)
C c
REFERENCE TS' DEFINITIONS t
t ANSWER B.09 (1.00)
B AND D, or D (alone)
REFERENCE 10CFR50.72 FERMI LEARNING OBJECTIVE OC&P COURSE 6
I C
t i
J 3
I
$ a 4 t I
=-
.PAGE IL TEST CROSS REFERENCE QUESTION VALUE REFERENCE
'05.01
'2.00' AJMOOO2535 05.~O2 1.50 AJMOOO2536
~05.03 1.50 AJMOOO2537 05.04 1.50 AJMOOO2538
~ 05. 05.
1.50 AJMOOO2539-03.06 1.50 AJMOOO2540 05.07
'2.00 AJMOOO2542 05.08
~2.00 AJMOOO2544 05.09 1.50 AJMOOO2545 15.00 06.01 2.00 AJMOOO2546 06.02
'2.00 AJMOOO2547 06.03 1.50 AJMOOO2548 06'.04 1.50 AJMOOO2549
-06.05 1.50 AJMOOO2550 06.06 1.50 AJMOOO2551 06.07 2.50 AJMOOO2552 06.08
.1.00 AJMOOO2553 06.00 1.00 AJMOOO2554 14.50 07.01 1.00 AJMOOO2534 07.02 2.50 AJMOOO2556
'07.03 2.00 AJMOOO2557 07.04 2.00 AJMOOO2558 07.05 2.00 AJMOOO2559 07.06 2.50 AJMOOO2561 07.07 1.00 AJMOOO2562 07.08 2.00 AJMOOO2563 15.00 08.01 1.50-AJMOOO2564 08.02 3.00 AJMOOO2565 08.03 1.50 AJMOOO2566 08.04 2.00 AJMOOO2567 08.05 2.00 AJMOOO2568 08.06 1.50 AJMOOO2569
-~. 0 8. 0 7 1.00 AJMOOO2570 08.08
.50 AJMOOO2571
'08.09 1.00 AJMOOO2572 14.00 I
i 58.50 6
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lo.
IstC RULES Am GUIDELINES FOR LICNISE EXANINATIONS During the administration of this examination the following rules apply:
on the examination means an automatic dental of your application Cheat 11u d result in more severe penalties.
1.
and co 2.
Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of c. eating.
h i
3.
Use black ink or dark pencil only to facilitate legible reproductions.
~
4.
Print your name in the blank provided on the cover sheet of the examination.
Fill in the date on the cover sheet of the examination (if necessary).
h.
5.
~
5.
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.
- as Consecutively number each answer sheet, write "End of Category 8.
appropriate, start each category on a new page, write only one side of the paper, and write "Last Page" on the Tast answer sheet.
Number each answer as to category and number, for example,1.4, 6.3.
9.
- 10. Skip 'at least three lines between each answer.
Separate answer sheets from pad and place finished answer sheets face 11.
i down on your desk or table.
Use abbreviations only if they are connonly used in facility literature.
12.
The point value for each question is indicated in parenthe'sts after the 13.
question and can be used as a guide for the depth of answer required.
i Show all calculations, methods, or assumptions used to obtain an answer 14.
to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTIO AND 00 NOT LEAVE ANY ANSWER BLANK.
If parts of the examination are not clear as to intent, ask questions of 16.
the examiner only.
You must sign the statement on the cover sheet that indicates that the L.
work is your own and you have not received or been given assistance in 17.
This must be done after the examination has completing the examination.
been coupleted.
p.
- 18. When you complete your examination, you shall:
a.
Assemble your examination as follows:
(1)
Exam questions on top.
(2)
Exam aids - figures, tables, etc.
(3) Answer pages including figures which are a part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions.
c.
Turn in all scrap paper and the balance of 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 progress, your license may be denied or revoked.
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- E(1 ) I% Il 5
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MVL = -0.693h P = Po10 l4 P = P et/T o
SCR = 5/(1 - K,gf)
SUR = 26.06/T
~
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CR (1 - K,ffg) = CR (1 - keff2) 3 2
SUR = 26p/a* + ( s-p )T M = 1/(1 - Keff) = CRt/CRo T = (a*/p) + [(s - p)/Ip]
M = (1 - Keffo)/(1 - Kefft)
' T = s/(p - s)
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. T = (s - p)/(Ap)
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Il t t
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Miscellaneous Conversions Water Parameters 10 1 curie = 3.7 x 10 g s 1 gel. = 8.345 Iha.
1 kg = 2.21 the 1 gal. = 3.78 liters 1 hp = 2.54 x 1 Stu/hr 1 ft3 = 7.48 gett -
1 au = 3.41 x 1 Stu/hr Density = 62.4 lbs/ft 1 in = 2.54 ca Density = 1 ga/cm3
- F = 9/5'C + 32 Heat of vaportration = 970 Stu/1ba
- C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lba 1 BTU = 778 ft-1bf 1 Ata = 14.7 Psi = 29.9 in. Hg.
1 ft H O = 0.433 lbf/in2 i
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l 1
- ~ ' '
- l 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION Compliance with the Limiting Conditions for Operation conta ITIONS er other succeeding Specifications is required during the OPERATIONAL CO 3.8.1 i
conditions specified therein; except that upon failure to meet th t
l Conditions for Operation, the associated ACTION requirements sha l
f the Limiting Condition for Operation and associ i
ts are 3.0.2 If the Limiting Condition for not met within the specified time intervals.
. Operation is restored prior to expiration of the specified time J
completion of the Action requirements is not required.
j; ided When a Limiting Condition for Operation is not met, except as prov iti-to the associated ACTION requirements, within one hour action 1
3.0.3 ated to place the snit in an OPERATIONAL CONDITION in wh does not apply by placing it, as applicable, in:
At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.At least HDT S 1.
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 />.
2.
3.
d the ACTION Where corrective measures are completed that permit operation un er ified time requirements, the ACTION may be taken in accordance with the spec
(
i for limits as measured from the time of failure to meet the Limiting C Exceptions to these requirements are stated in the individua Operation.
Specifications.
.l This Specification is not applicable in OPERATIONAL CONDITIO l
l Entry into an OPERATIONAL CONDITION or other specif
~
Operation are not be made unless the conditions for the Limiting Condition for f
i nts. This 3.0.4 met without reliance on provisions contained in the ACTION requ reme DITIONS as
}
provision shall not prevent passage through or to OPERATIONA i
required to comply with ACTION requirements.
are stated in the individual Specifications.
i 1
h t
3/4 0-1 FERNI - UNIT 2
l APPLICA81LITY SuitvEILLANCE REQUIREMENTS t
Surveillance Requirements shall be met during the OPERATIONAL C02 1TION i
l or other conditions specified for individual Limiting Conditions for Operation 4.0.1 unless otherwise stated in an individual Surveillance Requirement.
)
l Each Surveillance Requirement shall be performed within the specified b
4.0.2 time interval with:
A maximum allowable extension not to exceed 255 of the surveillanc,e a.
interval, but The combined time interval for any 3 consecutive surveillance inter-vais shall not exceed 3.25 times the specified surveillance interval.
b.
Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirem 4.0.3 Limiting Condition for Operation.
Surveillance requirements do not have to be in the tndividual Specifications.
l performed on inoperable equipment.
Entry into an 0PERATIONAL CONDITION or other specified appilcable co tion shall not be made unless the Surveillance Requirement (s) associated with I
4.0.4 the Limitinq Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.
Surveillance Requirements for inservice inspection and testing of ASME 4.0.5 Code Class 1, 2, & 3 components shall be applicable as follows:
-y Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves I
a.
shall be performed in accordance with Section XI of the ASME Boiler
/
~
and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific l
r (6) (1).
Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice b.
I inspection and testing activities required by the ASME Boiler and L
Pressure Vessel Code and applicable Addenda shall be applicable as ll il follows in these Technical Specifications:
il Required frequencies ASE Boiler and Pressure Vessel for performing inservice Code and applicable Addenda inspection and testing terminology for inservice activities inspection and testing activities At least once per 7 days Weekly At least once per 31 days Monthly At least once per 92 days i
Quarterly or every 3 months At least once per 184 days Semiannually or every 6 months At least once per 276 days Every 9 months At least once per 366 days Yearly or annually l
3/4 0-2 FERMI - UNIT 2
\\
SEACTIVITY CONTROL SYSTEMS CC+ TROL R0D ORIVE ISUSING $UPPORT s
_ LIMITING CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place.
l OPERATIONAL CONDITIONS 1, 2 and 3.
f APPLICABILITY:
ACTION:
With the control rod drive housing support not in pla I
v i'
ih SURVEILLANCE REQUIREMENTS The control rod drive housing support shall be verified to be in place by a visual inspection prior to startup any time it has been disassemble j
(
4.1.3.8 t
when maintenance has been performed in the control rod drive housing suppor I
I-area.
I I
k 3/4 1-15 FERMI - UNIT 2
l l
REACTIVITY CONTADL SYSTEMS 3/4.1.4 CONTROL M D PROGRAM CONTROLS ADO WORTH MINIMIZER LIMITING COMITION FOR OPERATION 3.1.4.1 The rod worth minimizer (RtM) shall be OPERA 8LE.
OPERATIONAL CONDITIONS 1 and 2*, when THERMAL POWER is less than 1
or equal to 25 of RATED THERMAL' POWER, the minimum allowable pres APPLICABILITY:
ACTION:
With the RWM inoperable, verify control rod movement and compliance
- i with the prescribed control rod pattern by a second licensed operator a
a.
or other technically qualified member of the unit technical staff Otherwise, control who is present at the reactor control console. rod movement J
p
[',
the reactor mode switch in the Shutdown position.
The provisions of Specification 3.0.4 are not applicable.
b.
SURVEILLANCE REQUIREMENTS q
The Rhm shall be demonstrated OPERABLE:
4.1.4.1 In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods for the purpose of making the reactor critical, and in a.
OPERATIONAL CONDITION 1 within I hour after RWM automa tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.
- l In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of I
b.
control rods for the purpose of making the reactor critical, by veri-fying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod (after selectton of first control red).
In OPERATIONAL CONDITION 1 within I hour after RWM autom initiation when reducing THERMAL POWER, by demonstrating the c.
withdraw block and insert block functions.
By demonstrating that the control rod patterns and s d.
program into the computer.
- Entry inte OPERATIONAL CONDITION 2 and withdrawal of selected contr permitted for the purpose of determining the OPERABILITY of the RnM prior withdrawal of control rods for the purpose of bringing the reactor to 4
l,J criticality.
I 3/4 1-16 FERMI - INGT 2
y INSTRUMENTATION TRAVER$1M IN-CORE PROSE SYSTEM LIMITIM CONDITION FOR OPERATION The traversing in-core probe system shall be OPERABLE with:
3.3.7.7 Five movable detectors, drives and readout equipment to map the core, a.
and Indexing equipment to allow all five detectors to be calibrated in a 4
b.
common location.
APPLICABILITY: When the traversing in-core probe is used for:
Recalibration of the LPRM detectors, and a.
'b.
ACTION:
With the traversing in-core probe system inoperable, suspend use of the system t
for the-above applicable monitoring of calibration functions. The provisions of Specifica61ons 3.0.3 and 3.0.4 are not applicable.
l-SURVEILLANCE REQUIREMENTS The traversing in-core probe system shall be demonstrated OPERABLE by 4.3.7.7 normalizing each of the above required detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to use when required for the LPRM calibration function.
l
'Only the detector (s) in the required measurement location (s) are required to be OPERABLE.
e l
ll i
FERMI - UNIT 2 3/4 3-65
E l
1 INSTRUENTATION CHLORINE DETECTION SYSTEM f
l LIMITING CONDITION FOR OPERATION I
3.3.7.8 Two independent chlorine detectors shall b equal to 5 ppe.
l APPLICABILITY: All 0PERATIONAL CONDITIONS.
ACTION:
With one chlorine detector inoperable, restore the inoperable detector to OPERABLE status within 7 days or, within the next 1
a.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain isolation of all control room emer-O gency intakes by placing the NVAC system in the chlorine mode of operation.
p With both chlorine detectors inoperable, within I hour initiate and maintain isolation of all control room emergency intakes by b.
placing the HVAC system in the chlorine mode of operation.
The provisions of Specification 3.0.4 are not applicable.
f c.
)
SURVEILLANCE REQUIREMENTS i
Each of the above required chlorine detectors shall be demonstrated I
4.3.7.8 OPERABLE by perfomance of a:
CHAMEL FUNCTIONAL TEST at least once per 31 days, and a.
CHANNEL CALIBRATION at least once per 18' months.
b.
3/4 3-66 FEMI - tstIT 2 E
l 3/4.4 REACTOg enaLANT SYSTEM 3/4.4.1 RECI M 14 TION SYSTEM RECI S A4 TION LOOPS LIMITING ii-31 TION FOR OPERATION Two reactor coolant system recirculation loops shall be in operatio ER When total core flow is less than 45% of rated core flow, then THE 3.4.1.1 oust be less than or equal to the limit specified in Figure 3.4.1.1-1.
-l l
OPERATIONAL CONDITIONS 1* and 2*.
APPLICABILITY:
With one reactor coolant system recirculation loop not in operatio ACTION:
l immediately initiate action to reduce THERMAL POWER to less tha l
a.
j equal to the limit specified in Figure 3.4.1.1-1
?
l 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With no reactor coolant system recirculation loops in operation,'
immediately initiate action to reduce THERMAL POWER to less than b.
equal to the limit specified in Figure 3.4.1.1-1 l'
s and in NOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
With two reactor coolant system recirculation loops in operation total core flow less than 45% of rated core flow and
~~
c.
greater than the limit specified in Figure ?.4.1.1-1:
- i (-
Monitor the APRM and LPRM** noise levels (S Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of entry into this condition 2.
a)
Within 30 minutes after the completion of a THERMAL POW increase of at least 5% of RATED THERMAL POW b) by control rod movement.
With the APRM or LPRM** neutron flux noise levels greater th l
three times their established baseline nois 2.
the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow to OWER greater than 45% of rated core finw or by reducing THE 3.4.1.1-1.
to less than or equal to the limit specified in Figure
- See Special Test Exception 3.10.4.
l detectors A
- Detector levels A and C of one LPRM string per itored when operating with a nonsymmetric control rod pattern.
detectors A and core LPRM string detectors A and C and two ot t
o lI l
3/4 4-1 FERMI - UNIT 2
\\
$WriEILLANCE REQUIREMENTS Each pump discharge valve shall be demonstrated OPERAB i
h each valve through at least one complete cycle of full travel dur ng e 4.4.1.1.1 STARTUP* prior to THERMAL POWER exceeding 255 of RATED l
demonstrated OPERABLE with overspeed se 4.4.1.1.2 102.55, respectively, of rated core flow, at least once per 18 months 4
Establish a baseline APRM and LPRM** neutro 11 ACTION c) the regions for which monitoring is required (Specification 3.4..,
4.4.1.1.3 i d unless within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which mon f ling outage.
i l '
l '
i
~
I i
- If not performed within the previous 31 days.
A
3/4 4-2 FERMI - UNIT 2
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i (031VW %) W3 mod 'imW3H13W03 3/4 4-3 FEltM1 - UNIT 2
REACTOR C0OLANT SYSTEM JET PUMP _5
~
~
l LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
l OPERATIONAL CONDITIONS 1 and 2.
APPLICABILITY:
ACTION:
With one or more jet pumps inoperable, be in at least HOT SHUTO e
I 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
a 21 SURVEILLANCE REQUIREMENTS Each of the above required jet pumps shall be demonstrated OP prior to THERMAL POWER exceeding 25% of RATED THERMA 4.4.1.2 per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- by determining recirculation loop flow, total core diffuser-to-lower plenum differential pressure fo operating at the same speed:
.)
The indicated recirculation loop flow differs by more than 10% from the established pump speed-loop flow characteristics.
I a.
The indicated total core flow differs by more tha q
b.
flow measurements.
l The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from the mean of all jet pump diffe l
c.
j' pressures in the same loop by more than 20% deviation l
deviation.
I l
l l
- 0uring the startup test program, data shall be recorded for the para listed to provide a basis for establishing the specified relationships.
l!
Comparisons of the actual data in accordance with the criteria commence upon the conclusion of the startup test program.
l 3/4 4-4 ll FERMI - UNIT 2 f
. ~.
l
RECIRCULATION MRF5 i
LIMITING CONDITION FOR OPERATION Recirculation pump speed shall be maintained within:
3.4.1.3 5% of each other with core flow greater than or equal to 70K of a.
rated core flow.
10E of each other with core flow less than 70K of rate b.
OPERATIONAL CONDITIONS 1* and 2*.
APPLICABILITY:
i ACTION:
With the recirculation pump speeds different by more than the specif limits, either:
Restore the recirculation pump speeds to within the specified lim i
l a.
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or Declare the recirculation loop of the pump with the slower spe o
in operation and take the ACTION required by Specification 3 b.
(
i
'i l
ll SURVEILLANCE REQUIREMENTS 1,
Recirculation pump speed shall be verified to be within the lim l
4.4.1.3 at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
t l
55ee 5pecial Test Exception 3.10.4.
l l
i !-a
'I 3/4 4-5 FERMI - UNIT 2
\\
\\
REACTOR COOLANT 5YSTEN i
IDLE RECIRCULATION LOOP STARTUP
)
LIMITING CONDITION FOR OPERATION An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolan 3.4.1.4 l
bottom head drain line coolant is less than or equal to 145'F, and:
When both loops have been idle, unless the temperature different l
between the reactor coolant within the idle loop f
a.
to 50*F, or When only one loop has been idle, unless the temperature diffe between the reactor coolant within the idle and operating recircula b.
tion loops is less than or equal to 50*F and the operating loop flow rate is less than or equal to 50% of rated loop flow.
0 OPERATIONAL CONDITIONS 1, 2, 3 and 4.
APPLICA81LITY:
.)
ACTION:
With temperature differences and/or flow rates exceeding the above l suspend startup of any idle recirculation loop.
'4
?
SURVEILLANCE REQUIREMENTS within the limits within 15 minutes prior to start l
i 4.4.1.4 F
loop.
3/4 4-6 FERMI - UNIT 2
- *TAINMENT SYSTEMS 3/4.6 **TAIISENT SYSTEMS
}/4.6.1 PRIMRRY CONTAleMENT PRIMARY e=TA!anENT INTEGRITY LIMITING C0seITION FOR OPERATION PRIfWtY CONTAll8ENT INTEGRITY shall be maintained.
3.6.1.1 j
_ APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.
L e
L ACTION:
Without PRIMARY CONTAllmENT INTEGRITY, restore h
COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS l
PRIMARY CONTAlletENT INTEGRITY shall be demonstrated:
4.6.1.1 I
After each closing of each penetration subject to Type B testing, f
~~
except the primary containment air locks, if opened following Type A a.
or 8 test, by leak rate testing the seals with gas at P,, 56.5 psig,
(
and verifying that when the measured leakage rate for these seals is i
added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.b for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L,.
i At least once per 31 days by verifying that all primary containment penetrations ** not capable of being closed by OP b.
conditions are closed by locked closed valves, blank flanges, or deactivated automatic valves secured in position, except as provided in Table 3.6.3-1 of Specification 3.6.3.
e By verifying each primary containment air lock is in compliance wit the requirements of Specification 3.6.1.3.
l c.
By verifying the suppression chamber is in compliance with the r d.
i ments of Specification 3.6.2.1.
i I
'See Special Test Exception 3.10.1.
- Except valves, flanges, and deactivated automatic valves which are lo inside the containment, and are locked, sealed or other lj SHUTDOWN except such ve'rification need not be closed position.
l' lt than once per 92 days.
3/4 6-1 l
FERMI - WIT 2
I-CONTAINNENT SYSTDt$
4 PRIMARY CONTA!99g:NT LEAKAGE 1,
f LIMITING CONDITION FOR OPERATION Primary containment leakage rates shall be limited to:
3.6.1.2 An everall integrated leakage rate of less than or equal to: L,,
a.
containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P,,
l O.5 percent by weight of the 56.5 psig.
A combined leakage rate of less than or equal to 0.60 L, for all b.
penetrations and all valves If sted in Table 3.6.3-1, except for main steam line isolation valves
- and valves which are hydrostatically tested per Table 3.6.3-1, subject to Type 8 and C tests when pressur-ized to P,, 56.5 psig.
- Less than or equal to 100 scf per hour for all four main steam Ifnes c.
when tested at 25.0 psig.
A costined leakage rate of less than or equal to 5 gpa for all con-tainment isolation valves in hydrostatically tested lines which d.
penetrate the primary containment, when tested at 1.10 P,, 62.2 psig Less than or equal to 1 gpm times the number of valves per penetration not to exceed 3 gpa per penetration for any line penetrating contain-l e.
ment and hydrostatically tested at 1.10 P,, 62.2 psig.
When PRIMARY CONTAINMENT INTEGRITY is required per l
APPLICABILITY:
Spect 1 cation 3.6.1.1.
.]
ACTION:
,I With:
The measured overall integrated primary containment leakage rate s.
exceeding 0.75 L,, or The measured combined leakage rate for all penetrations and all
(,
~
valves listed in Table 3.6.3-1, except for main ste b.
subject to Type 8 and C tests exceeding 0.60 L,, or The measured leakage rate exceeding 100 scf per hour for all four l,
c.
l*
main steam lines, or The measured combined leakage rate for all containment isolation v i
in hydrostatically tested lines which penetrate the primary contain d.
ment exceeding 5 gpa, or ifne penetrating primary
,The leakage rate of any hydrostatically testedcontainm l
e.
containment isolation valves per penetration or greater than 3 gpa per penetration, prior to facreasing reactor coolant system temperature above 200*F ff The overall integrated leakage rate (s) to less than or equal to 0.75 ll L,, and a.
j,
- Exemption to Appendix J of 10 CFR Part 50.
3/4 6-2
!i FERMI - WIIT 2
N TAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) i ACTION:
(Continued)
The combined leakage rate for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves
- and b
valves which are hydrostatically tested per Table 3.6.3-1, subject l
to Type 8 and C tests to less than or equal to 0.60 L,, and The leakage rate to less than or equal to 100 scf per hour for all c.
four main steam lines, and l
The combined leakage rate for all containment isolation valves in hydrostatically tested lines which pentrate the primary containment d.
to less than or equal to 5 gps, and The leakage rate of any hydrostatically tested line penetrating primary containment to less than 1 gpa per isolation valve times the number of i
e.
containment isolation valves per penetration or less than 3 gpa per penetration.
e i
~
i SURVEILLANCE REQUIREMENTS
}
y i
The primary containment leakage rates shall be demonstrated at the following test schedule and shall be determined using the methods and pro-4.6.1.2 l-visions described herein:
Integrated Primary Containment Leakage Rate - Type A Test a.
Integrated leak rate tests shall be performed at the testContai 1.
pressure (P,) of 56.5 psig.
permitted to decrease more than 1 psi below P,.
l r
Type A tests should be completed prior to Type 2.
the Type A test results shall have added to it the difference between the "as found" vs. "as left" leakages for all penetra-Type B and C leakages not accounted for in the Type A tions.
test shall be added to the upper confidence limit (UCL) to However, when estimate the overall integrated leakage rate.
adding the leakage rate measured during a Type C test to the I
j:
results of a Type A test, the lower leakage rate of the two isolation valves in a line shall be used.
f
.h l'
3/4 6-3 FERMI - UNIT 2 l
t
,. - --- - -~ ~ ~,,,-. -.
.. - - - - ~. - -.
CaMTAINMENT SYSTEMS 9
$URVEILLANCE REQUIREMENTS (Continued)
~
I If the leakage rate exceeds the acceptance criterion, corrective action shall be required.
If, during the performance of a Type 3.
A test, excessive leakage occurs through locally testable penetrations or isolation valves to the extent that it would interfere with the satisfactory completion of the test, these leakage paths may be isolated and the Type A test continued A local leakage test shall be performed until completion.
The before and after the repair of each isolated leakage path.
sum of the post repaired local leakage rates and the UCL shall be less than 75% of the maximum allowable leakage rate, L,.
Closure of the containment isolation valves for the purpose of the test shall be accomplished by the means provided for normal 4.
operation of the valves.
A Type A test shall last a minimum of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after stabiliza-For a test less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the test procedures and 5.
l-the acceptance criteria provided in BN-TOP-1, " Testing Criteria tion.
for Integrated Leakage Rate Testing of Primary Containment struc-1, 11/1/72 shall be tures for Nuclear Power Plants," Revision met in order to consider a Type A test satisfactorily completed.
6.
Test and Analysis Method J
(a) The absolute test method shall be used.
The mass plot (mass point) analysis technique, as described in ANSI /ANS-56.8-1981, in addition to the total-time method (b) l described in BN-TOP-1, shall be used to compute the con-tainment leakage rate (only the mass plot analysis tech-nique need be performed for test durations greater than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).
For the mass plot method, an upper one sided 95% confidence (c) limit for the leakage rate shall be determined based upon e
For the total-time method, the normal regression theory.
precedures for calculating the leakage rate are described in ANSI M45.4-1972.
7Property "ANSI code" (as page type) with input value "ANSI M45.4-1972.</br></br>7" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
Instrumentation During Type A testing, measurements shall be i
j (a) absolute pressure.
Dry bulb temperature sensors shall have an acf (b) the test 120'F and a repeatability of at least 0.1*F.
3/4 6-4 FERMI - UNIT 2 C-
CNTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
Dewpoint temperature sensors shall have an accurac'y or better over the dewpoint temperature range expected (c) during the test $20*F and a repeatability of at least 20.5*F.
Pressure sensors should have a range such that p, is (d)
Accuracy shall be at between 25 and 75% of full scale.1 east 0.0155 of f l
of 0.001% of full scale.
The number and location of temperature and dewpoint sensers shall be determined prior to each Type A test based on a (e)
I l
temperature survey of the containment.
A sufficient number of dry bulb tamperatu (f) contributes more than 105 to the calculated temperature.
i At least two-thirds of the dewpoint temperature sensorsHowev shall be functioning properly during the test.
(g) data recorded over the last 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ind order to cause error in leak rate calculations, then mal-function of any or all but three of the dewpoint sensors shall not require aborting the test.
(
l At least one precision pressure gauge shall be functio (b) l properly during the test.
Prior to each Type A test and following the failure of any l
sensor, an instrument error analysis shall be performed (i) using the Instrument Selection Eulde (ISG) form l'
ANSI /ANS-56.8-1981 the end of a test except as noted in (g) above.
Three Type A Overall Integrated Conta t
8.
The third P,, 56.5 psig, during each 10 year service period.
for the test of each set shall be conducted during the shutdown 10 year plant inservice inspection.
If any periodic Type A test fails to meet 0.75 L,, the te d
schedule for subsequent Type A tests shall be review 9.
approved by the Commission.f ail to meet 0.75 L,, a t 0.75 every 18 months until two consecutiv 3
3/4 6-5 FERM] - UNIT 2 i
touTAlte4ENT SYSTEMS SURVi!LLANCE REQUIREMENTS (Continuedi The accuracy of each Type A test shall be verified by e supple-10.
j mental test which:
Confirms the accuracy of the test by verifying that the 1
(a) difference between the supplemental data and the Type A test data is within 0.25 L,.
Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental (b) test.
Requires that the rate of gas injected into the containment t
(c) or bled from the containment during the supplemental test to be equivalent to at least 75% but not more than 125% of L,at P,, 56.5 psig, Type B and C tests shall be conducted with gas at P,, 56.5 psig*,
b.
at intervals no greater than 24 months except for tests involving:
1.
Air locks, 2.
Main steam line isolation valves, Penetrations using continuous leakage monitoring systems, 3.
Valves pressurized with fluid from a seal system, j
4.
ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the primary containment, and 5.
Purge supply and exhaust isolation valves with resilient materia 6.
j seals.
~
j Air locks sha)1 be tested and demonstrated OPERABLE per Specifica-c.
tion 4.6.1.3.
Main steam line isolation valves shall be leak tested at least o e
d.
per 18 months.
Type 8 tests for penetrations employing a continuous leakage system shall be conducted at P,, 56.5 psig, at intervals no greater e.
than once per 3 years.
Leakage from isolation valves that are sealed with fluid from a s system may be excluded, subject to the provis f.,
the seal system and valves are pressurized to at least 1.10 P,,
i 62.2 psig, and the seal system capacity is adequate to maintain system pressure for at least 30 days..
- Unless a hydrostatic test is required per Table 3.6.3-1.
3/4 6-6 FERMI - UNIT 2 I
l
CONTAINMENT SYSTEMS
$URVE1LLANCE REQUIREMENTS (Continued)
~
ECC5 and RCIC containment isolation valves i g.
least once per 18 months.
Purge supply and exhaust isolation valves with resilient materia seals shall be tested and demonstrated OPERABLE per Specifica-h.
tion 4.6.1.8.2.
The provisions of Specification 4.0.2 are not applicable to specif l
cations 4.6.1.2a.
4.6.1.2b. and 4.6.1.2c.
1.
l i
e.
O em
.l l
[
P P
i l
l I,
3/4 6-7 FERMI - UNIT 2
...,,,___--_____ _ _._, _ - _ _4
.-.-_,__~_,m.__.
CONTAINMENT SYSTEMS PRIMARY CONTAINNENT AIR LOCKS LIMITING CONDITION FOR OPERATION 1
j Each primary containment air lock shall be OPERA 8LE with:
4 3.5.1.3 Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one a.
air lock door shall be closed, and An overall air lock leakage rate of less than or equal to 0.05 L, at J
b.
l P,, 56.5 psig.
i APPLICABILITY:_ OPERATIONAL CONDITIONS 1, 2*, and 3.
1 ACTION:
With one primary containment air lock door inoperable:
a.
Maintain at least the OPERA 8LE air lock door closed an restore the inoperable air lock door to OPERABLE status within 1.
~
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or lock the OPERA 8LE air lock door closed.
Operation may then continue until performan 2.
air lock door is verified to be locked closed at least once per
}
31 days.
Otherwise, be in at least NOT SHUTDOWN within the next 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
}
and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
The provisions of Specification 3.0.4 are not applicable.
4.
With the primary containment air lock inoperable, except as a res q
of an inoperable air lock door, maintain at least one air lock door b.
closed; restore the inoperable air lock to OPE in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I
- See Special Test Exception 3.10.1.
o I
I 1
3/4 6-8 FERMI - UNIT 2
EMTAIMENT SYSTEMS siNiWILLANCE REQUIREMENTS 8
OPERABLE:
4.6.1.3 Each primary containment air lock shall be demonstrate s'
k is Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following each closing, ex i
hour when the gap between the door seals is pres a.
56.5 psig.
i Prior to establishing PRIMARY CONTAIMENT INTEGRITY lock has been opened during periods when containment in The demonstration shall verify a seal leakage rate b.
less than or equal to 5 scf per hour when the gap between l
not required.
is tasted seals is pressurized to P, 56.5 psig, unless the air lock pursuant to Specification,4.6.1.3.c.2.
d By conducting an overall air lock leakage test at P, 5 i its by verifying that the overall air lock leakage rate is with c.
limit:
prior to initial fuel loading and at 6-month
- intervals 1.
thereafter, i
prior to establishing PRIMARY CONTAINNENT INTEG lock has been opened during periods when containm 2.
h leak was not required, if maintenance which could affect t e last tight integrity of the doors has been performed since t I
successful test pursuant to Specification 4.6.1.3.c.1.
in each At least once per 6 months by verifying that only one door air lock can be opened at a time.**
d.
I
- The provisions of Specification 4.0.2 are not applicable. inte
- Except that the inner door need not be opened to verifyided t when the primary containment is inerted, provlock i t has been l
deinerted.
3/4 6-9 FERMI - UNIT 2 f
1 CONTAINMENT SYSTEMS MSIV LEAKAGE CONTROL SYSTEM LIMITING CONDITION FOR OPERATION Two independent MSIV leakage control system (LCS) subsystems shall be 3.6.1.4 OPERABLE with each subsystem comprised of a flow path from the associated control air division to the main steam lines.
APPLICA81LITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
l With one MSIV leakage control system subsystem inoperable, restore the inoperable subsystem to OPERA 8LE status within 30 days or be in at least HOT SHUTOO the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS Each MSIV leakage control system subsystem shall be demonstrated 4.6.1.4 OPERA 8LE:
At least once per 31 days by cycling each testable valve except the motor-operated MSIVs through at least one complete cycle of full a.
8 travel.
During each COLD SHUTDOWN, if not performed within the previous 31 days, by cycling each valve including the motor-operated MSIVs b.
3-i f.
not testable during operation through at least one complete cycle j
of full travel.
~
l At least once per 18 months by performance of a functional test of the subsystem throughout its operating sequence, and verifying that c.
each interlock operates as designed and each automatic valve actuates to its correct position.
4 By verifying the pressure control (pressure and Ap) instrumentation to d.
be OPERABLE by performance of a:
3 1.
CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, CHANNEL FUNCTIONAL TEST at least once per 92 days, and 2.
CHANNEL CALIBRATION at least once per 18 months.
3.
l 3/4 6-10 l
FERMI - letIT 2 e-
-r----v,,,---r-+,..
,,,-c
--,,-m_,,,,,,.v-----aw,_nw an-,
en
5 CONTAINMENT SYSTEMS PRIMARY CONTA1 M NT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION The structural integrity of the primary containment shall be saintained at a level consistent with the acceptance criteria in Specification 3.6.1.5 4.5.1.5.1 APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
_With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least NOT SHUTDOWN within the nex COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS s
The structural integrity of the exposed accessible interior and exterior surfaces of the primary containment shall be d 4.6.1.5.1 This inspection shall be performed prior to the Type A of those surfaces.
containment leakage rate test to verify no apparent changes in appearance or
(
other abnormal degradation.
Any abnormal degradation of the primary containment struc-ture detected during the above required inspections s 4.6.1.5.2 Reports This report shall include a description of the conditio
(
r i
3/4 6-11 FERMI - UNIT 2
1 CnNTAINNENT SYSTEMS DirkTLL AND SUPPRESSION CHAN8ER INTERNAL PRES $URE LINITING CONDITION FOR OPERATION Orywell and suppression chamber internal pressure shall be maint 3.6.1.6 between -0.10 and +2.00 psig.
OPERATIONAL CONDITIONS 1, 2, and 3.
APPLICA81LITY:
ACTION:
With the drywell and/or suppression chamber internal pressure outside of specified limits, restore the internal pressure to within the limit within I hour or be in at least HDT SHUTDOWN within the next SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIRENENTS The drywell and suppression chamber internal pressure shall be
)
determined to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.6.1.6 e
I h
3/4 6-12 FERMI - UNIT 2 i
1
. t --
~-ww,
^~^
Sta Ott7 Sch DP IRIT U.S. NUCLEAR REGULATORT C000GSSION REACTOR OPERATOR REQUALIFICATICII EIANIRATION Facility:
Fermi 2 Reactor Type BWR-GE4 Date Administered:
86/07/29 Examiner:
Lanksbury. R.
Candidate:
NihsT E L INSTRUCTIONS TO CANDIDATE:
Read the attached instruction page carefully. This examination replaces the current cycle facility administered requalification examination. Retraining requirements for failure of this examination are the same as for failure of a requalification examination prepared and administered by your training s'aff.
Points for each question are indicated in parentheses after the question, passing 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 value Total Score Value Category 15.50 25.41 1.
Principles of Nuclear Power Plant Operation. Thermodynamics, Heat Transfer and Fluid Flow 15.00 24.59 2.
Plant Design Including Safety and Emergency Systems
.00 24.59 3.
Instruments and controls 15.50,_
25.41 4.
Procedures - Normal. Abnormal.
Emergency, and Radiological Contrcl 00 100.00 Totals Final Grade All work done on this examination is my own, I have neither given nor received aid.
O Candidate's Signature 0
6
.I
s PAGE 2
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSn x AND FLUID FLQH QUESTION 1.01 (2.00)
List four (4) reactor conditions or characteristics which influence the(2.0) point of criticality and the rate at which it is approached.
I QUESTION 1.02 (3.00)
Concerning Thermal Limits:
a.
WHAT are the three (3) peaking factors which make up the Total (1.0; Peaking Factor?
b.
WHAT safety condition does limiting FLCPR to less than one (1) (1.0.
prevent?
WHAT is the definition of MAPRAT?
(1.0 c.
QUESTION 1.03 (2.00) g_
During your Shift, an SRV inadvertantly opens from 100% power and 1000
[;(()1 psia.
Use a Hollier Diagram or the Steam Tables to answer the following (ASSUME a saturated system and instantaneous heat transfer):
WHAT is the SRV tailpipe temperature, assuming atmospheric pressure a.
in the Suppression Chamber and No Reactor Depressurization?
(0.5) b.
If the Suppression Chamber Pressure were to increase, would the
~ R2jkh.L (0.51 tailpipe temperature INCREASE, DECREASE, or REMAIN THE SAME?
l If the reactor is depressurized when the SRV is opened, wil the g
l c.
Tailpipe Temperature AniTIALLi inunsant, vauntASE, or REMAIN THE pg gg d 4 g, 2h (0.5)
SAME?
I j
d.
At WHAT Reactor Pressure will the Tailpipe Temperature be at its l}
MAXIMUM value (during the depressurization)?
(0.5)
U
.i ll 1
l l'
l.
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1.
PRINCIPLES _OF NUCLEAR POWER PLANT OPERATION.
PAGE 3
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (1.00)
During the last refueling outage a center bundle was inadvertently positioned in a peripheral bundle location.
The reactor is then brought to 100% power.
Fill-in-the-blanks with (LARGER, SMALLER, or THE SAME) to indicate the correct response for the thermal hydraulic conditons in the eisplaced bundle.
The flow rate in the misplaced bundle will be
... than the a.
adjacent peripheral bundles.
(0.5) b.
The power level in the misplaced bundle will be than if the bundle had been properly placed in its central location.(0.5)
QUESTION 1.05 (2.50)
State whether the following statements concerning fission poisons are TRUE or FALSE:
The largest production contribution of xenon is the radioactive a.
decay of Iodine.
(0.5) b.
A 25% power reduction from 100% power would have a larger xenon peak than a 25% power reduction from 50% power.
(0.5)
A rapid reactor shutdown from 100% power would have the same c.
resultant xenon peak as a reactor scram from 100% power.
(0.5) d.
Since the production and removal of samarium is a direct function i
of thermal flux, a reactor that has operated at ONLY 50% capacity I
will have the same equilibrium samarium concentration as one that has operated at 100% capacity.
(0.5)
Upon restarting the reactor following a 6-month outage, the samarium e.
concentration will decrease to its 100% full power concentration.
(0.5) l l
l l
_ (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
_,/
PAGE 4
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSnx AND FLUID FLQ){
J QUESTION 1.06 (3.00)
STATE WHY the following situations would or would not change differential control rod worth.
(Also indicate whether rod worth would. INCREASE, DECREASE, or NOT CHANGE.)
A rod is withdrawn from notch 06 to notch 10.
(0.75 g' a.
b.
Pulling the first rod in Rod Group 7 versus pulling the first rod in Rod Group 5?
Assume in either case that all previous rods have(0.75) p been pulled.
Localized voiding of a region not previously voided.
(0.75) c.
d.
A change of the size of the control rod reducing the surface area while maintaining the same boron volume.
(0.75's QUESTION 1.07 (2.00)
Five (5) minutes following a reactor scram from 100% power, reactor power is 15 on IRM Range 4 and decreasing.
WHAT is the minimum IRM Range that you could go to two (2) minutes later without violating any operational limits?
SHOW calculation and EXPLAIN any assumptions made.
(2.0) l 4
1.
l l
l l
L
(***** END OF CATEGORY 01 *****)
PAGE 5
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.
QUESTION 2.01 (2.00)
What are four (4) functions of the SBLC " Pipe within a Pipe"?
(2.0) i 6UESTION 2.02 (1.50)
EXPLAIN the purpose of the pressure equalizing valves in the CRD Hydraulic Include
- tpoir.tc.
(1.5)
System and HOW they accomplish their purpose.
t QUESTION 2.03 (1.50)
( 1. 5 '
HOW is the 11 Diesel Generator kept warm in standby condition?
QUESTION 2.04 (2.00)
ANSWER the following questions concerning the vacuun relief lines between the drywell and the suppression chamber:
(0.5)
WHICH direction does the flow go?
a.
WHAT safety feature do they provide for or prevent?
(0.75) b.
WHAT would be the consequence if a vacuum breaker line had failed (0.75) c.
in the open position during a LOCA?
KLI-Am QUESTION 2.05 (1.00) a ag h gg c-e sn u t LL Concerning the relief valve low-low set (LLS) function:
ayt ru g,
(0.5)
WHAT is the purpose?
a.
DOES the LLS actuate on manual operation of the relief valves, b.
automatic operation of the relief valves, or on either one?
(0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
PAGE 6
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.
QUESTION 2.06 (3.50)
HOW is the HPCI turbine exhaust line protected against overpressure?
(1.5) c.
(Identify two means includihi"eeuvolu^ -).1 yllut.k r
Identify which of the following are direct HPCI turbine trips and(2.0) b.
which are HPCI system isolations:
F LJ N f - / /!/ /
1.
HPCI steam line low pressure P
2.
Reactor high water level
/h A TAIP, 3.
HPCI room high temperature (gg,Mge.,
4.
Pump suction low pressure QUESTION 2.07 (2.25)
Following an auto initiation of RCIC at a reactor pressure of 800 psig, HOW are each of the following reactor pressure decreases to 400 psig.
parameters affected (INCREASE, DECREASE, or REMAINS CONSTANT) by the change in reactor pressure?
BRIEFLY EXPLAIN EACH CHOICE.
ASSUME the RCIC System is operating as designed.
(0 75) a.
RCIC flow to the reactor.
(0.75; RCIC pump discharge head (assuming NPSH remains constant).
b.
(0.75)
j c.
.i 1
ll
't QUESTION 2.08 (1.25)
Listed below are a number of components of the Reactor Water Cleanup Systen l:
Place the components in order of normal FLOW PATH beginning with the (One component MUST components downstream of the inlet isolation valves.
(1.25) be used twice for full credit).
Filter demineralizer l
a.
b.
Pumps Regenative heat exchanger c.
d.
Non-regenative heat exchanger I
l
(***** END OF CATEGORY 02 *****)
PAGE 7
3.
INSTRUMENTS AND CONTROLS QUESTION 3.01 (2.00)
REGARDING THE CORE SPRAY SYSTEM, ANSWER THE FOLLOWING QUESTIONS EITHER TRUE (2.00)
OR FALSE:
When the Core Spray Pump control switch is placed in the OFF/ RESET l~
position, a valid core spray initiation signal will start the pump.
1.
If a core spray pump is running and its associated electrical 2.
the core power bus auto transfers to the emergency diesel generator, spray pump will trip and will need to be restarted by the operator.
The core spray logic for the "A" loop is powered from the Division I 3.
battery.
4.
If one of the Core Spray pumps start and is running, a permissive signal is sent to the ADS logic system.
QUESTION 3.02 (1.50)
Feedwater Level Control System is in 3-element control using reactor level detector channel A.
Reactor power is at 85% steady state.
state how Using the attached Feedwater Level Control System block diagrac, (1.50) the system will respond initially to the following failures:
a.
B Fe D tar Elnw signal fails HIGH.
b.
LOSS of Control Signal to B Reactor Feed Pump Speed Controller.
Check valves between the #5 heaters and the heater dra' n flash tanks i
c.
CLOSE.
Com - Ca;L4 is ow
.Cto 6. 4 -
FN 16. (tom tett ksa 6 A E. Q,
3
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
i
1 i
PAGE 8
1 3.
INSTRUMENTS AND CONTROLS l
QUESTION 3.03 (1.00)
A normal cold plant startup is in progress with the RSCS Sequence Mode Salector (SMS) Switch in " Withdraw."
Which of the following depicts the FEWEST number of operator actions that will cause the dim backlights for tho "A-12" sequence rods to extinguish and the dim backlights for the "A-34" sequence rods to illuminate?
(1.00) a Fully withdraw all "A-12" sequence rods; Select "A-34" on the Rod c.
Sequence Selector (RSS) Switch.
b.
The position of the "A-12" sequence rods is immaterial; Select "A-34" on the Rod Sequence Selector (RSS) Switch.
The position of the "A-12" sequence rods is immaterial; Select "A-34" c.
on the Rod Sequence Selector (RSS) Switch, and Select any 'A-34" Rod.
d.
Fully Withdraw all "A-12" sequence rods; Select "A-34" on the Rod Sequence Selector (RSS) Switch, and Select any "A-34' rod.
QUESTION 3.04 TfT The reactor is operating at power.
What will be the effect on/of the recirculation flow control system due to the following conditions:
EI actor 30% power when an o erator in
. rte y
ibk e "A r
1rc pu ran er s tio o
0.
(1..
S b.
The reactor is at 95% power with recirculation flow control in master manual when full open INDICATION on recirc pump "B" discharge valve (F031B) is lost.
(1.0) o j
gu-A O M AM M-QUESTION 3.05 (2 00)
On the SRM detector drive control a green light indicates a SRM detector may be withdrawn.
List four (4) conditions when the light will be on.(2.0)
N 1
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
.i
PAGE 9
3.
INSTRUMENTS AND CONTROLS QUESTION 3.06
(.00)
Main steam pressure is being controlled by a normal lineup with the Gsvenor Pressure Control (GPC) in control at 65% power.
Recirc control is in master manual.
Explain the actions of the GPC system for the following:
W (1.0) b.
One turbine bypass valve fails open.
QUESTION 3.07 (1.00)
Which of the following are TRUE concerning the Standby Liquid Control (1.0; System:
(CHOOSE ONE)
In the event a remote (outside control room) reactor shutdown is a.
required, SBLC injection can be actuated by the local pump START switch.
b.
The pumps may be operated simultaneously if necessary to shutdown the reactor in an ATWS.
If injected, the SBLC system will provide at least a 0.28% SDM c.
and 1000 ppm boron concentration in the reactor vessel.
l
(:
d.
Nitrogen-charged accumulators assure adequate suction pressure-for the pumps.
L QUESTION 3.08 (3.50)
State what specific RPS action (if any) will occur DIRECTLY from MSIV position if the following main steam lines shut in the RUN mode:
(1.5) a.
Lines B, C, and D.
b.
Lines B and C.
c.
Lines A and B.
4 l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
PAGE 10 3.
INSTRUMENTS AND CONTROLS QUESTION 3.09 (2.00)
What are FOUR (4) automatic actions which occur at the Reactor Water (2.0)
L2 vel 8 (217.1") trip point?
1 k
l i
1 l
(***** END OF CATEGORY 03 *****)
t-
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11 B&DIOLOGICAL CONTROL QUESTION 4.01 (2.50)
LIST the FIVE entry conditions with their setpoints for PRIMARY CONTAINMENT CONTROL, Procedure 29.000.03.
(2.50)
QUESTION 4.02 (2.00)
According to procedure 12.000.13
" Radiation Work Permits", list FIVE conditions when a Specific Radiation Work Permit is required.
(2.00)
QUESTION 4.03 (2.00)
Assume an emergency exists which requires control room evacuatien and plant shutdown f rom outside the control room (procedure 20.000.19):
c.
Except for necessary announcements, what THREE operator actions should be performed if possible prior to leaving the control roon.?
(1.50; b.
What is the procedurly recommended method to scram the reactor from outside the control room?
(0.50)
QUESTION 4.04 (1.00)
At what point during a LOSS OF STATION AND/OR CONTROL AIR does the 1,
procedure direct you to manually scram the reactor?
(1.00; i
~
QUESTION 4.05 (2.00)
While performing the immediate actions of procedure 20.000.02, ABNORMAL RELEASES OF RADIOACTIVE MATERIALS, a Reactor Building Vent Exhaust R2diation Monitor Upscale Trip occurs.
LIST the FIVE automatic actions you (2.00) must verify.
h C
't
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
PACE 12 4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 4.06 (2.00)
Tha reactor operator is performing a surveillance to the HPCI system and, dua to a system modification, some procedural steps become impossible to perform.
It is determined that multiple changes are needed.
b~/ Y Under this condition can multiple temporary changes be made to the(0.50)d*4#'
o.
procedure?
What THREE conditions must be followed when making multiple temporary b.
(1.50) changes to a procedure ?
QUESTION 4.07 (2.50)
In accordance with the approach to criticality steps in the cold startup procedure, 22.000.03, answer the following:
(0.5)
When is the reactor considered critical?
a.
b.
What FOUR (4) items (some items may consist of more ther.1 piece of information) are recorded in the Control Room Log Book when (1.0) criticality is established?
How is reactor period determined other than from reading the c.
(1.0) period meter?
GL.
QUESTION 4.08 (1.50) 00tu
(
Answer the following:
Vh A new employee who is 18 years old is assigned to work for youWhat is the individu a.
and is to be a radiation worker.
accumulated occupational dose (lifetime) to the whole body and V
y allowable whole body quarterly limit per 10CFR20?
'1.0) b.
BRIEFLY explain the difference between a " rad" and a "rer.".
(0.5)
(***** END OF CATEGORY 04 *****)
r*****w******* END OF EXAMIN ATION ***************)
i PAGE 13 1.
PRINCIPLES OF NUCLEAR POWER PLANT OEERATION, THERMODYNAMICS. HEAI TRANSFER AND FLUID FLOW ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 1.01 (2.00) 1.
Xenon concentration 2.
Moderator Temperature 3.
Control rod position 4.
Order of withdrawal 5.
Core exposure (4 @ 0.5 each)
REFERENCE Standard Nuclear Reactor Theory ANSWER 1.02 (3.00) a.
1.
Axial 2.
Local 3.
Radial
( 3 @ 0.33 each) b.
Avoiding boiling transition, c.
APLHGR/MAPLHGR limit REFERENCE Nuclear Power Plant Thermal Sciences.
l ANSWER 1.03 (2.00) a.
295 degrees F (+/- 5 degress F) b.
Increase c.
Increase l
d.
450 psia (+/- 25 psia)
REFERENCE Steam Tables /Mollier Diagram, and Nuclear Power Plant Thermal Sciences.
PAGE 14 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 1.04 (1.00) orifice fixed)
Smaller (2 phase flow larger due to higher power, c.
Smaller (2 phase flow larger due to smaller orifice) b.
REFERENCE Nuclear Power Plant Thermal Sciences.
4 ANSWER 1.05 (2.50) a.
True b.
True c.
False d.
True o.
True REFERENCE Rsactor Theory Fundamentals.
ANSWER 1.06 (3.00)
Increase [0.25] due to higher flux [0.5].
c.
[0.25] due to higher flux [0.5].
b.
Increase [0.25] due to decrease in thermal neutrons [0.5].
c.
Decrease [0.25] due to reduced surface area seen by flux [0.5].
d.
Decrease REFERENCE Roactor Theory Fundamentals.
l s
1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE 15 THERMODYNAMICS. HEAT TRANSn;x AND FLUID FLOW ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 1.07 (2.00)
Using P = Po e**(t/T)
[0.25]
= 15 e**(-120/80)
= 3.35 on Range 4
[0.25]
= 33.5 on Range 1 Therefore Range 1 is the lowest Range [0.5].
Assumptions:
On a down power transient, with large negative reactivity insertions, the stable decay period is determined by the longest lived half-life [0.5].
For this example, it is assumed to be -80 seconds [0.5].
I REFERENCE Reactor Theory Fundamentals; and Fermi Student Handout C51-11.
4 l
l l
1 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 16 ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 2.01 (2.00) 1.
Sodium pentaborate injection 2.
CS line break detection 3.
Core plate delta P measurement 4.
Jet pump delta P measurement 5.
CRD drive water and cooling water delta P tap (4 @ 0.5 each)
REFERENCE Fermi Student Handout PIS C41.
ANSWER 2.02 (1.50)
(Following a scram, the exhaust header will equilize to the scram diccharge volume pressure across directional control valve 121.)
Once the outlet valve is shut (when scram reset) the exhaust water header would remain at some low pressure causing a high differential pressure across the control rod mechar ism if rod motion were attempted [0 75].
The equalizing valves will open(,at approxiemntely 50 psid between exhaust [0.25] and cooling water header [0.25} to facilitate a rapid repressurization of the exhaust hander to el)iminate this problem [075].
REFERENCE Formi Student Handout PIS C-11-50.
ANSWER 2.03 (1.50) pump is used to circulate water f rom the jacket coolant system U>d7[
1.
ndtheaircoolantsystem}[p-5]throughaheaterandbackto hese systems M (O7G 2.
Lube oil, being circulated by a pump, passes through a standby heater.
)
REFERENCE Formi Student Handout R30.
2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 17 ANSWERS --' FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 2.04 (2.00)
From suppression chamber to drywell.
c.
b.
Negative pressure in drywell to limit upward loading of the drywell floor.
The flow would bypass the suppression chamber suppression feature c.
following a LOCA.
REFERENCE Fermi Student Handout T23-01/02.
l ANSWER 2.05 (1.00)
To prevent exessive loading on the TORUS.
(Also accept:
reduces c.
the number of challanges to the SRV's s)g ado o M$d#* k kIIa'dly ta.s ble ti. E t 4se a o-This will function if valves are opene'c! mania'D>
a to a b.
REFERENCE Formi Student Handout PIS B21.
ANSWER 2,06 (3.50)
N/,., E'"- a A r %
o.
1.
Rupture disks op hy eggust I geg, g u v c-rrv
% M one trip.[ m.
Turbi 2.
y"e m gt;,,(
p.m...
AM d to 3 ;4 ubit up % pu. M 1.
s
. M. Heu c6..t ia&h u teu a tue-. camt u.6,9-M b.
1.
Isolation 9
s
)
2.
Turbine trip p g,gg) h 3.
Isolation l:
4.
Turbine trip (4 @ 0.5 each) j REFERENCE Formi Student Handout PIS E41 and FSAR Sec. 6.3.
L l
L l
i I
PAGE 18 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 2.07 (2.25)
Remains constant [.25).
Flow is controlled by the RCIC flow controller c.
which will attempt to maintain a constant output flow regardless of reactor pressure [0.5).
b.
Decreases [0.25).
The flow controller functions to maintain a constant flow, thus pump discharge pressure is decreased along with the decreasing reactor pressure to maintain a constant flow.
OR Since the as reactor flow controller maintains a constant flow to the reacter, preasure decreases, the pump discharge head must decrease to maintain a constant flow (constant NPSH) [0.5),
Decreases [0.25).
Since pump discharge head is decreasing to maintain c.
a constant flow, turbine RPM must also decrease [0.5).
REFERENCE Fermi Student Handout E51.
ANSWER 2.08 (1.25) b,c,d,a,c REFERENCE Formi Student Handout G-33.
3.
INSTRUMENTS AND CONTROLS PAGE 19 ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 3.01 (2.00) 1.
FALSE 2.
FALSE 3.
TRUE 4.
FALSE REFERENCE Formi Student Handout Pls E21.
ANSWER 3.02 (1.50) 4 (Reactor level will decrease) L4cF&r due to the level control system 3o c.
havinE a steam flow / feed flow error with steam flow < feed flow [0 '+65 resulting in a decrease in the speed of the reactor feed pump turbiner
[0.20].
M Reactor level shouldArgppes the B RFT will run down to minimum speed, b.
speed (1660 RPM) [0. RJ.
The Control System will see a flow mismatch and attempt to recover it with the A RFT [0.263 c.
Limits feed Pump speed to 78%
0./5). (Levelwilldecreaseuntilfeed flow equals steam flow [0.251
. REFERENCE Formi Student Handout PIS N21.
L ANSWER 3.03 (1.00)
G.
REFERENCE i
Formi Student Handout C11-09.
I is i4 1
l
I PAGE 20 3.
INSTRUMENTS AND CONTROLS ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 3.04 (2.00)
The control system will see a step increase to 45% demand signal c.
from the dual limiter and will attempt to increase pump speed [075].
However, the error limiting network will see the resulting spike and will automatically limit the rate of change of the pump speed [025]
Loss of open indication for the discharge valve r N c#$ N S ks N b.
l sigr. 1 mrcund the #1 rreed limiter and it outputs a 30% cignal [0.5].
The "0"
emap will then s i v,- Lv Lim demend speed [0.5]. L t : p s.'%
495% e 4 A M L g.
Q,q t
4-weio bli culi3
{
Fermi Student Handout B31.
ANSWER 3.05 (2.00) 1.
SRM > 100 cps.
2.
SRM channel in bypass.
3.
Reactor mode switch in RUN.
4.
All companion IRM's are above range 2.
(4 @ 0.5 each)
REFERENCE Formi Student Handout PIS CSI.
4 i
ANSWER 3.06 (2.00) i fro P
b.
Reactor pressure would decrease due to the pwr/stm flow mismatch.
This would decrease the error signal from the Regulator amp summer (error amp).
The reduced signal would close the turbine control valves, reducing steam flow through the turbine.
Reactor pressure i
would return to the original value with pwr/stm flow (total) equal.
i REFERENCE
+
Formi Student Handout PIS N30-12A.
t 0
PAGE 21 3.
INSTRUMENTS AND CONTROLS ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 3.07 (1.00)
.C.
REFERENCE Formi Student Handout PIS C41; Tech Spec 3.1.1.
ANSWER 3.08 (1.50) o.
Full scram b.
None c.
Half scram REFERENCE Formi Student Handout PIS C71.
ANSWER 3.09 (2.00) 1.
Main turbine trip 2.
Reactor feedpump turbine trip 3.
HPCI turbine trip 4.
RCIC turbine trip 5.
S/B feedwater system valve isolation (4 @ 0.5 each)
REFERENCE Formi Student Handout PIS B21.
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 22
)
RADIOLOGICAL CONTROL ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
1 ANSWER 4.01 (2.50)
Torus water level above +2 " (124,220 ft )
c.
3 b.
Torus water level below -2 " (121,080 ft )
Torus water average temperature above 95 F c.
d.
Drywell atmosphere average temperature above 135 F Drywell pressure above 1.88 psig e.
(5 @ 0.5 each)
REFERENCE Procedure 29.000.03 Rev 4 ANSWER 4.02 (2.00) a.
Contaminated Area > 10000 dpm/100 cm' b.
Airborne Radioactivity areas c.
Neutron Radiation area exposure d.
High Radiation area exposure Unknown conditions in an area to be entered o.
f.
Maintenance of equipment, controls, or instrumentation which contain radioactive material (5@0.qfeach)
REFERENCE 12.000.13 l
l l
i
(
4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 23 RADIOLOGICAL CONTROL ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 4.03 (2.00) e.
Place the reactor mode switch to SHUTDOWN Depress both scram pushbuttons Arm and depress main turbine trip pushbutton (3 @ 0.5 each) b.
Scram the Reactor from the relay room by taking one operable APRMs Mode Switch out of operate in Division I (A,C,E) and one in Division II (B,D,F).
REFERENCE Procedure 20.000.19 ANSWER 4.04 (1.00)
If Scram valve pilot / air header High/ Low and Control Rod Drift indication is received on more than one control rod.
REFERENCE Procedure 20.12c.01 ANSWER 4.05 (2.00) 1.
RB Ventilation System Tripped 2.
RB Div. I and II Supply and Exhaust Isolation Valves close 3.
Primary Containment Purge and Vent Valves Close 4.
SBGT Auto Starts 5.
CC-HVAC System aligns to the Recirculation Mode l
REFERENC Procedure 20.000.02 I
l l
PAGE 24 PRQCEDURES
_ NORMAL. ABNORMAL. EMERGENCY AND 4.
BADIOLOGICAL CONTROL ANSWERS -- FERMI 2
-86/07/29-LANKSBURY, R.
ANSWER 4.06 (2.00)
O.
Yes
- b. (Changes do not change the intent of the procedure.
LChange is approved by two knowledgeable members of the plant staff, one of whom is a SRO license holder.
the changes will jrf$Upon completion of the performance of the procedure, be submitted as a temporary PCR.
(3 @ 0.5 each) 46f1,L64 h ASM QA po x ( 7,,0 0 0, o )
ANSWER 4.07 (2.50) a.
Neutron count rate rises [0.15] with a constant period [0.20] without additional control rod withdrawal [0.15].
s__.
b.
1.
time 2.
rod sequence, rod group, rod, and rod position 3.
period 4.
reactor coolant temperature (4 @ 0.25 each) gunn Period = Time for power to double, from SRM/ recorder, x 1.443.
c.
REFERENCE Fermi General Operating Procedure 22.000.03.
ANSWER 4.08 (1.50)
Accumulated whole body limit = 5(N-18) = 0 rem.
a.
Quarterly whole body limit = 1.25 rem.
The rad is a measure of the energy released in tissue by ionizing b.
radiation whereas the rem is a measure of biological effect caused by that energy.
REFERENCE 10CFR20.101
H/L T 201.5,
k 4192.5 CONM R M
A 2MA)
RRS RUN9ACM ELE MEN L*
1 [ MODE l,
1 I JC atanu V
3
/j> LEVEL -
unit ELEMENT SF. LECT h
h ggL,gy LEVEL / FLOW svnTCH
+
g, mg, A
l SELECT SWITCH COMPARATOR LEVEL I gI f LEVEL-MASTER COMPARATOR RWM LPAP 35*/e CONTROLLER RWM LPSP 30*/.
L STEAM FLOW ALARM
.f A
UNIT DUAL LIMITER g h
2900 RPM A 0
-> FiOW 1r o
STEAM FLOW COMPARATOR gfg STATION TOTAL MfA E
STEAM STATION M CV EG FLOW P
FUNCTION STEAM FLOW __i FUNCTION GEN.
C GEN.
A U
' SIGNAL q
I~~"~~7 FAILURE STM/ FEED p
RFP iSRFP I
FLOW N.
i STEAM FLOW RECORDB l TURBINE 1 TUR8K l D
l lEHC l EHC g
l
^
L _ _ _ _i L _ _ _ _I
,tO, FEED FLOW INTEG.
A TOTAL RRS RUNBACK 20*/.
[
FLOW UNIT FEEO FLOW REF. DECO DWG.
8 61721-2126-1 REV. E 03-15-46 A l FIGURE I m
EQUATION SHEET f = ma y a s/t Cycle Gfficiency o (Netw3rk out)/(Energy in) w = og s = Vot + 5 at2 E = mc2 KE = 5 av2 a = (Vf - Vo)/t A = AN A = Age-At PE = agh Vf = Vo + at w = e/t A = sn2/tg = 0.693/t1 W = -6 tgeff = [(tg) (t )3 b
[(tg) + (t )3 b
AE = 931 am y, yo,-zx 6 = iiCpat 6 = UAat I = Ic -"*
e Pwr = Wrah I = Io 10- M TVL = 1.3/p P = Po10 "
HVL = -0.693/p P = P et/T o
SUR = 26.06/T SCR = S/(1 - Keff)
CRx = S/(1 - Keffx)
SUR = 2Gp/t* + ( s-p )T CRj (1 - Keffj) = CR (1 - keff2) 2 T = (t*/p) + [(8 - p)/Ip]
M = 1/(1 - Keff) = CR1/CRo T = 1/(p - s)
M = (1 - Keffo)/(1 - Keffj)
. T = (6 - p)/(Ap)
SDM = (1 - Keff)/Keff f* = 10-5 seconds p = (Keff-1)/Keff = AKeff/Keff I = 0.1 seconds-1 p = [(1 /(T Keff)] + [s ff/(1 + IT))
e
'Ijdj = 1 d22 2
P = (IeV)/(3 x 1010) 11d1 2=1d22 2
R/hr = (0.5 CE)/d (meters)
I=N R/hr = 6 CE/d2 (feet)
Miscellaneous Conversions Water Parameters 1 curie = 3.7 x 1010dps 1 gal. = 8.345 lbm.
1 kg = 2.21 lbm 1 gal. = 3.78 liters 3 Btu /hr 1 ft3 = 7.48 gal.
I hp = 2.54 x 10 1 mw = 3.41 x 106 Btu /hr Density = 62.4 lbm/ft Density = 1 gm/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1bm
- F = 9/5*C + 32 Heat of fusion = 144 Btu /1bm
- C = 5/9 (*F-32) 1 BTU = 778 ft-1bf 1 Atm = 14.7 psi = 29.9 in. Hg.
1 ft H O = 0.433 lbf/in2 2
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