ML20127G689
| ML20127G689 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/01/1985 |
| From: | Lawyer S, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127G645 | List: |
| References | |
| 50-302-OL-84-02, 50-302-OL-84-2, NUDOCS 8505210116 | |
| Download: ML20127G689 (100) | |
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ENCLOSURE 1 EXAMINATION REPORT 302/0L-84-02
-Facility Licensee: . Florida Power Corporation Facility Name: Crystal River Unit 3 Facility Docket No. 50-302-Written examinations were administered at Crystal River Training Center .
near Crystal River, Florida. Oral exam cation were administered at Crystal River Power Plant near stal River, M)orida. s Chief Examiner: 1 /
- Sahdy Lawyer Date Signed
[/
Approved by: . d.k%
BrWe 4'. Wilson, Section Chief E
Date Signed Summary:
Examinations on December 17 - 20, 1984 Examinations were administered to seven SRO candidates, six of whom passed. '
Examinations were administered to four R0 candidates, all of'whom passed.
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4 REPORT DETAILS
- 1. Persons Examined SR0 Candidates: R0 Candidates:
Gallion, Earnest J. Ashworth, Donald C.
Harmon, Steven F. Carr, William G.
Long, Ronald A. Garrett, Jimmy Sr.
Hickle, Bruce J. Rawls, Ricky E.
Kirk, Michael W.
Metcalf, Thomas E.
Welch, Earl E.
Other Facility Emoloyees Contacted:
- J. R. Cuneo, Lic. Op. Training Supervisor
- V. R. Roppel, Plant Engineering and Technical Services Manager
- G. L. Boldt, Plant Operations Manager
- B. E. Crane, Training Manager
- J. L. Bufe, Nuclear Compliance Specialist
- W. P. Ellsberry, Technical Training Supervisor
- J. F. Bel:er, Nuclear Support Supervisor L. Giles, Nuclear Training Instructor C. D. Arouthnot, Nuclear Operator Instructor )
R. C. Zareck, Nuclear Operator Instructor l J. P. Haerle, Nuclear Operator Instructor J. L. Springer, Nuclear Operator Instructor
- Attended Exit Meeting
- 2. Examiners:
Bruce Wilson Mike King Tom Morgan
- Sandy Lawyer
- Chief Examiner t
- 3. Examination Review Meeting At the conclusion of the written examinations, the examiners met with L. Giles, C. D. Arbuthnot, R. C. Zareck, J. P. Haerle, J. L. Springer and J. R. Cuneo to review the written examination and answer key. The following comments were made by the facility reviewers:
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SRO EXAM
- 1. QUESTION 6.04 Facility Comment: The STM provided to the NRC was incorrect. The Emergency Diesel Generator crankcase pressure switch does not shutdown the engine; it provides an alarm function only.
a.
l NRC Resolution: The drawing provided confirmed the facility comment.
However, the question did not ask whether it shutdown the engine or not. The question as stated can only be correctly answered with choice (d). No change to the answer key is required.
- 2. QUESTION 6'.13 Facility Comment: This question is not appropriate. Memorization of remote interlocks on fuel handling equipment should not be required.
NRC Resolution: Choosing the correct answer from among four answers does L not require memorization, but . rather recognition. In
! addition, raemorization of important load limits on the fuel handling bridge is required of SR0 candidates.
- 3. QUESTION 6.14 I Facility Comment: The STM provided to the NRC was incorrect and unclear.
I As a result, all answers are correct.
f- NRC Resolution: Review of R0-100 provided at the exam review confirms the facility comment. Question 6.14 has been deleted.
- 4. QUESTION 6.17 l Facility Comment: We feel it is unreasonable to expect an SRO to know every vital load on every vital bus.
{
NRC Resolution: Choosing the correct answer from among four answers
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given does not requite " memorization of every vital load on every vital bus." Recent LERs have demonstrated the importance of .this subject matter. No change to the answer key is appropriate.
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- 5. QUESTION 8.02, A, B & C Facility Comment: 8.02A - The answer is 3-and 4. -~The "B" HPI Pump would be placed in service prior to removal of "A" from service; hence, two HPI pumps would always be operable.
8.028 - The answer is 4 and 7. - (3.5.1/3.5.4) - For the same reason given above.
8.02C - The answer is:
1 (Shutdown to hotstandby) if the assumption is made that repair is made after initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> but before the 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> time expires. ,
3 (Shutdown to cold shutdown) with the assumption of no repair completed, or with no assumption.
4 (Shutdown not required) if assumption is made that repair is effected in the initial 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. ,
i NRC Resolution: The comments are accepted and the answer key.has been I changed accordingly. l
- 6. QUESTION 8.10C Facility Comment: The phrase (steady state power to steady state power) is used to denote equilibrium conditions at CR-3.
Candidates will probably choose " Decrease" as the change direction for Xenon.
NRC Resolution: The meaning of the phrase " steady state power" was explained to all candidates during the exam. Therefore, no change to the answer key is appropriate.
- 7. QUESTION 8.16 Facility Comment: The material sent to the NRC is not up-to-date. A new letter has been generated changing the distribution.
NRC Resolution: The question has been deleted because of the extensive change to the fire brigade team manning requirements as verified by our review of the new letter. The answer key has been changed accordingly.
. ,; ; a 4'
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4 RO' EXAM
- 5 13 : QUESTION 1.05:
. Facility-Comment: -Answer (d) also appears to be correct. .
lNRC Response: Upon review of reference material, . answer (d) was found-to also be correct. The answer key. was -changed to
- reflect this.
2.1 QUESTION 1.22 Facility Commenti We do not. believe this is appropriate for R0 level,. it-
.is.possibly OK-for SRO.
[NRCResponse: The question ~ was based on six curves- from the plant
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curve book- The intent of the question was to test'.RO's familiarity with and basic. knowledge of Control- Rod Insertion limit curves and how these. limits are affected-by parametric changes such as number .of operating RCPs and core ; burnup. . We believe it is appropriate at'. R0 level since all that is required to be known is that insertion . limits become .more -restrictive under - both circumstances. :No change to question ' or - answer -is warranted.
1 3. [ QUESTION-1;25
- Facility Comment
- ' Answer (b) -is also correct - reference Tech Spec pg. B
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3/4 2-1.
NRC Response: Agree. - Answer was based on T.S. pg :3/4 1-3 which did not include ' choice (b). .Since there is no correct
-answer to the question, it was deleted.
[4. QUESTION 2.06
- Facility Comment: Choices (b) and (d) are both correct.
NRC Response: Reference STM 5-4 shows answer (b) is- correct but
, a ' incomplete. The 4-psi signal also opens the BWST outlet-valves (DHV-34 and -35) and the sodium ~ thiosulfate tank valves, although these valves are tagged out.
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5 Choice (b) was incorrectly worded in that it should have
.said the only valves affected were the NaOH tank outlet and RB spray pump suction valves. Either choice (b) or (d) was accepted.
~5. QUESTION 2.07 Facility Comment: Choice (b) is.also correct. Training material provided as reference to NRC was incorrect. Reference Elementary drawing.
NRC Response: Either choice (b) or (c) was accepted based on new reference material.
- 6. QUESTION 2.09 Facility Comment: Choice (a) is also a correct response to the question.
-Reference material provided to NRC was incorrect. New reference provided was AP-402, "PSA G. Annunicator Response." Number G-5-3 shows IAV-30 closing at an air pressure of < 80 psi.
NRC Response: Either choice (a) or (d) was accepted based on reference supplied.
- 7. . QUESTION 2.17 Facility Comment: All of the choices are incorrect. Reference provid'ed was Elementary Diagram 208-026.
NRC Response: Question was reworded - during exam because reference material provided for developing exam was suspect.
Reference diagram 208-026 showed attempt to save question was unsuccessful. Since there is no correct answer to the question, it was deleted.
- 8. QUESTION 2.19 Facility Comment: Choice (b) and (d) are also proper answers. STM provided to develop exam was wrong. New STM was provided as reference.
NRC Response: New STM provided during exam review shows answers (a),
(b), and (d) to be correct responses. All three choices will be accepted.
- 9. QUESTION 2.20 Facility Comment: Although there is no problem with the concept of what the question is asking for, there is a great deal of
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concern about asking specifics of responses to annuncia-tors. .It is unreasonable to expect operators to know all annunicator responses and this is, in fact, what the annunciator response manual is for.
NRC Response: We acknowledge the comment and agree with the concerns expressed. The question however, was a design question and expected the operator to know some design specifics of the closed ' cycle cooling system rather than the response to a particular annunciator. No change to question or answer is appropriate.
- 10. QUESTION 3.01 Facility Comment: Choice (d) is also a correct response.
NRC' Response: In choice (d), the limit of 60% was originally written as 75%. It was inadvertently changed during final exam proof and typing. Both choices (c) and (d) are accepted.
- 11. QUESTION 3.19 Facility Comment: The answer, choice (c) is incorrect since the valve op is maintained at 80 psid rather than 35. A new draft STM was provided as a reference. The answer as written was obtained from two different chapters of the plant's STM.
NRC Response: Since none of the statements are TRUE, the question will be deleted.
- 4. Exit Meeting
' At the conclusion of the site visit the examiners met with representatives of the' plant staff to discuss the results of the examination. Those individuals who clearly passed the oral examination were identified. ,
-a . ,, . - - - , , - , , . , , . - - , - - - -
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There were no generic weaknesses. noted during the oral examination. The cooperation given to the examiners and the effort to ensure an atmosphere in the control room. conducive to oral examinations was also noted and appreciated.
- 5. The following additional changes were made to the examination and answer key based upon NRC review during the grading process:
- a. Question 5.06.c Answers 3 and 4 are acceptable based on assumption o' beta effective.
No changes to candidates grading were required.
- b. Question 5.08.d.
Answers 1 and 2 were accepted. Depending upon the reference material both concepts are often taught.
- c. Question 5.09.b.
Answers 2, 3 and 4 were accepted. Although answer 2 was the correct choice from facility supplied information, other references show answer 3 is also correct. Also, there is too much overlap between the distractors,
- d. Question 7.13.a Question is not worded sufficiently clearly to define its intent, i.e.,
is a trip required for all seismic events? Procedure AP-961 does require a reactor trip as an immediate action under certain circumstances.
Credit was given for both answers.
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E r\CloSO/ L 5 i U. 5. NUCLEAR REGULATORY COMMISSION
$, SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: _CRYSIAL_RIMER___________
REACTOR TYPE: _ EWE =B1W122______________
TERC Y DATE ADMINISTERED:_B&Z12212________________
EXAMINER: _KIWGa_H.________________
APPLICANT: _________________________
IWSIEUCIIONS_ID_eEELICeWII Use separate paper for the answers. Write answers on one side on1w.
Starle uuvstion sheet on tor of the answer sheets. Points for each cuestion are indicated in parentheses after the uuestion. The passins drade reauires at least 70% in each categorw and a final grade of at loast 80%. Examination papers will be ricked up six (6) hours after the examination starts.
% OF CATEGORY % OF APPLICANT'S CATEGORY
__UALUE. _IDIAL ___ SCORE ___ _UALUE__ ______________CAIEGORY_____________
.25.00__ _25.00 ___________ ________ 5. THEORY OF NUCLEAR POWER PLANT '
OPERATION, FLUIDS, AND THERH0 DYNAMICS
_25.00__ _25.00 ,___________ ________ 6. PLANT SYSTEMS DESIGNe CONTROL, AND INSTRUMENTATION
- 7. PROCEDURES - NORMAL, ABNORMAL,
_25.00__ _25.00 ___________ ________
EMERGENCY AND RADIOLOGICAL CONTROL
- 8. ADMINISTRA11VE PROCEDURES, l _25.00__ _25.00 ___________ ________
i CONDITIONSe AND LIMITATIONS 100.00__ 100.00 ___________ ________ TOTALS FINAL GRADE _________________%
l All work done on thin examination is av own. I have neither diven nor reevived aid.
APPLICANT'S SIGNATURE a
5.. IW3'DB L'U OH5LbtKO- vusts -a ,
IMERMODYW^WICS l 4
h QUESTION 5 01 (3.00)
True or False 7
- a. The differential temperature necessarv to transfer heat is inverselv Proportional to heat flux. (0.5)
- b. Pune runout is the term used to describe a centrifugal pump ,
when it is operating with its discharge valve shut. (0 5) :
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- c. The latent heat of vaporization is another term for the latent heat of condensation. (0.5)
- d. One of the Pump laws for centrifugal pumps states that Power reouired bw the Pump motor is direct 1w Proportional to the sauare of the Pump speed. (0.5)
- e. The faster a centrifugal Pump rotates, the greater the NPSH reuuired to Prevent cavitation. (0.5)
- f. When comparinM & Parallel-flow heUt exchanger to a Counter-flow heat exchanger, the temperature difference between the two fluids along the LENGTH of the heat exchanger tubes is MORE uniform for the Parallel-flow heat exchanger. , (0.5)
GUESTION 5.02 (1.00)
True or False
- a. At high boron concentrations (prm), as the modera' tor temperature increases, the ppm boron decreases resulting in a positive (+)
moderator temperature coefficient. (0.5)
QUESTION 5.03 (1.00)
Which fluid swstem is subcooled bw greater than 30 F7 TEMP. PRESS. (Psia)
- a. 540 1000
- b. '560 1500
- c. 665 2000
- d. 640 2400
34__IhE04Y_DE_WUCLEet_20WER EL4WI_ DEERS 110Wa_GLU88Sa_oWB vo@S a IMERMODP'OCICS A
C QUESTION 5 04 (1.50)
Critical Heat Flux (CHF) is defined au the heat flux at which Departure from Nucleate Boiling (DNB) occurs. For un INCREASE in osch of the Parameters below, tell how the CHF will change. (Consider cach Parameter seperatelv.)
Limit wour answer to INCREASE, DECREASEe or REMAINS UNCHANGED.
- c. Reactor Coolant Flow Rate.
- b. Reactor Coolant Temperature.
- c. Reactor Coolant Pressure. [3 9 0.5 eu3 (1.5)
QUESTION 5.05 (1.50)
Assume that wour Plant has experienced a degraded Power condition and that wou are monitoring the Plant # G CouldOWO cn natural circulation.
4 Identify whether the following statements are TRUE or FALSE:
- a. A slow downward trend in indieuted Tuve is alwawu a good indication of well-established natural circulation flow. (0.5)
- b. A difference between wide-range T h und wide-range T c of 65"F and slow 1w increasing indieutes that natural circulation flow is developing. (0.5)
- c. Natural circulation flow rate can be increased bv raPidlw increasing steam flow rate bw "5%. (0.5)
- - - g,_, - - - v IEERMODYM&MICS t
t CUESTION 5.06 (3.00)
The reactor is shutdown with a K vf f of 0.9 ' ^^ and the source range indicates 100 ces. Rods are withdrawn and suurev range now indicates 200 ces.
Choose the correct answer for each of the three (3) auestions below
- o. The new K off will be?
- 1. 0. 93 de lla_JC/K
- 2. 0.95 dwli. K/K
- 3. 0.97 celi. X/K 0.99 delt, r/r (g,o) 4.
- b. The amount of reactivitw added was?
- 1. 0.0449 delta K/K
- 2. 0.0526 delta K/K
- 3. 0.0585 delta K/K
- 4. 0.0635 delta K/K (1.0)
- c. If the same amount of reactivit'u were added again the reactor would bei
- 1. Sub-critical
- 2. Critical ,
- 3. Super-critical
- 4. Prompt-critical (1.0)
GUESTION 5.07 (1.00)
Are the following statements about the DuPPlur Coefficient TRUE or FALSE?
a .- Doppler coefficient becomes more negative from 0-100% Power due to the increased overlapping of resonance peaks at higher fuel tWmPeratures. (0.5)
- b. Doppler coefficient becomes more negative over core life due to the buildup of Pu240 and fission Products with large resonances in the writhermal range. (0.5) 2
tb._ vwas e; IWERBODXWeMICS
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T GUESTION. 5.08 (2.00)
Match the following terms with their definitions:
TERMS:
- a. Natural circulation
- b. Saturated Liould
- c. Enthalew
- d. Departure from Nucleate Boiling DEFINITIONS:
- 1. The Point at Which Partial f 21m boiling begins.
- 2. The Point on the boiling curve where heat transfer surface temperature will rise sharelw with little or no increase in heat flux. .
- 3. -The movement of a fluid base and the differences in density of the fluid caused bw a differential temperature.
- 4. The total energw of a substance Per unit mass.
- 5. A liauid that cannot absorb anw more energw without starting to vaporize.
- 6. The internal energw of a svstem that is no longer available to do work. ,
[4 9 0.5 es3 (2.0)
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IMERBODYO^141CS
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s TOUESTION 5 09 (3.00)
Usins the followins fisurese chcose the correct answer for each Cf the three (3) euestion asked below!
100 Peer I
0 Time Al A2 Il2 Y ,
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- o. What is the approximate time from Al to A27
- 1. 10 hours
- 2. 30 hours ,
- 3. 50 hours
- 4. 70 hours (1.0)
- b. What is the aeProximate time from B1 to B27
- 1. 1-3 hours
- 2. 2-6 hours
- 3. 5-7 hours
- 4. 6-9 hours (1.0)
- c. Whw does Xe concentration decrease from A1 to D17
g.. MJMs r ICERNODXMADICS s
OUESTION 5 10 (2 00)
If steam moes throush an ideal throttling process from a high Pressure oteamline to steospherice will the following INCREASEe DECREASE or RENAINS CONSTANT 7 Eno exPlaination reuuired3
- c. entroPw
- b. enthalPw
- c. Pressure
- d. specific volume [4 9 0.5 ea3 (2.0) 0UESTION 5.11 (1.00)
What percentage of delawed neutrons are born at thermal energies?
O. 49%
- b. 25%
- c. 7% ,
- d. 0%
GUESTION 5.12 (2.00)
Answer the following two (2) auestions with regard to Plutonium
- a. Which of the following dveict how Pu239 is formed.
- 1. U23B + oN1 -> U239 -B -> NP239 -B -> Pu239
- 2. U235 + oN1 -> U236 +d2 -> NP238 +P1 -> Pu239
- 3. U238 - oN1 -> U237 -B -> NP237 +d2 -> Pu239
- 4. U235 - oN1 -> U234 +Hv4 -> Pu238 foN1 -> Pu239 (1.0)
- b. Which of.the following describec the effect Pu239 has on reactor operation:
- 1. Pu239' increases fuel temperature coefficient over core life.
- 2. Pu239 increases reactor response time for reactivitu chanses.
- 3. Pu239 increases core life due to being a fuel.
- 4. Pu239 increases the amount of decaw heat after a shutdown. (1.0)
ma _ m c IMERNODYWeMICS s
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GUESTION 5 13 (3.00)
Indicate which of the following are TRUE and which are FALSE INDICATIONS that the Point of addins heat (PDAH) has been reached.
[ Assume normal Plant operation 3
- a. SUR decreases j i
- b. Pressurizer level decreases j i
I
- c. T-hot increases
- d. Turbine bweass controller station (Bailev) output decreuses,
- e. Pressurizer Pressure decreases
- f. T-ave increasing C6 e 0.5 vu] (3.01 0
- 6. _ _ E L e WI . B Y STETSTD E a 1031_LtDWLs _006_3L13maistrdDwn00 vum v l
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OUESTION 6.01 (1.00)
Which of the following describes how a thermoccur ill indicate when it fails open?
- o. It will exhibit maximum resistance across its output leads which corresponds to a high or maximum temperature.
- b. It will exhibit minimum resistance across its output leads which corresponds to a low or minimum to *erature.
- c. It will exhibit maximum rwsistance across its outrut Ivads which corr;,,sponds to a low or minimum temperature.
- d. It will exhibit a different resistance across itu output Ivads ,
which corresponds to the temperature at the Point of the failure.
QUESTION 6.02 (2.00)
Determine if the following statements about the Nuclear Instrument-otion swstem are TRUE or FALSE.
- c. The intermediate range detectors are compensated i o n-c h a m b e r s,.
The lined chamber is sensitive on1w to gamma raus while the unlined chamber is sensitive to both neutron and summa raws. (0.5)
- b. UndercomPensation of the intermediate range detector will cause it to indicate a higher Power level than actual core Power.
(0.5)
- c. The source range detector Power suPP1w has a high voltage cutoff from the intermediate range which turns off the high voltage at 1 X 10 E-10 ames. (0.5)
- d. The PuWer rande difference amplifier output iG the difference in the Power in the bottom and the top of the reactor core.
This difference is called Power imbalance. Power imbalance is (0.5) eaual to power in the botton minus Power in'the top.
, 6A__BLCOM_D8bMLE_L'15G8TfJi L@l3L(lih cDLCEDEmnst%wmIUnsLU3 uv unsts r s
GUESTION 6.03 (2.00)
Choose the correct statement for each of the uuestions asked below about the Emersonew Diesel Ownerators.
- o. The EDG can operate loaded at 3300 kw with a 0.8 Power factor!
- 1. continuous 1w
- 2. for 2000 hrs
- 3. for 600 hrs
- 4. for 30 min (1.0)
- b. The following statement is refering to which one of the below listed controllers?
The CONTROLLER can be set to automatica11w divide and balance the load between engines Parallelled on an vivetrical swstem. As the CONTROLLER reduced toward zero, the unit becomes able to change .
loads without chenging speed. As a general rulee units running alone should have the CONTROLLLER set on zero.
- 1. Load Limit Control
- 2. Speed Droop Control
- 3. Compensation Adjustment Control
- 4. Swnchronizer Control , (1 0)
QUESTION 6.04 (1.00)
-The 2 out of 3 low oil Pressure and the 2 out of 3 high crankeese Pressore shutdowns for the Emergenew Diesel are activated when!
- a. Generator speed is > 710 rpm for longer than 20 see- OR Low Coolant Pressure is reset for longer than 20 see AND Low Oil Pressure is reset for longer than 20 sec.
- b. Generator speed is > 810 ren for longer than 20 see AND Low Coolant Pressure is reset for longer than 20 see AND Low Oil Pressure is resel for longer than 20 sec.
- c. Generator speed is > 710 rem for longer than 20 sec OR Low Coolant Pressure is reset for longer than 20 sec OR Low 011 Pressure is reset for longer than 20 sec.
- d. Generator speed is > 810 rpm for longer than 20 sec OR Lcw Coolant Pressure is reset for longer than 20 see AND Low Gil Pressure is reset for longer than 20 sec.
En5L%EUBSmsna trt39s Gb s
.00ESTION 4.05 (1.00)
Which of the following securatelw dvPicts tb reuuired core flood tank number and the amount of core covera *'s that the core flood owstem will Providet
- c. One core flood tank will cover 3/4 of the core.
- b. Two core flood tanks will cover 3/4 of the core.
- c. One core flood tank will cover the entire core.
- d. Two core flood tanks will cover the entire core.
QUESTION 6.06 (1'.00)
With regard to the Engineered Safeguard Actuation Swstem, determine if the following statements are TRUE or FALSEt
- a. When the HPI swstem is bwPassede during a normal shutdown, HPI is prevented from initiating when RC Pressure reaches 1500 Psis, but can still be activated bw a RC Pressure of 500 Pside , (0.5)
- b. Depressing the two manual actuation Pushbuttons for the LPI Swstem, will Position the LPI valves to their Engineered Safeguards Position and Start the dvCav heat Pumps. (0.5) 4 O
6a _EBLI L M M EG8 L% ME60A.
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. s GUESTION 6.07 (2.00)
In reference to the Control Rod Drive Sustem answer the followins two nuestions.
- o. While in automatic on the Operator Control Panel (Diamond Panel),
which of the following indicating lamese when illuminated, will.
also switch the diamond panel to MANUAL 7
- 1. Seeuence-Inhibit Lamp
- 2. Automatic-Inhibit Lamp
- 3. Aswometric Rods Lamp
- 4. Out-Inhibit Lamp (1.0)
- b. When the Out-Limit Lamps for groups 1-8 illuminate, this indicates .
that at Iwast one rod out of its respective group is at the Out-Limit of 1 1/2 inches Past 100% withdrawn except _____?_____
which the Out-Limit is 91.4% withdrawn.
- 1. group 5
- 2. group 6 ,
- 3. group 7
- 4. group 8 (1.0)
QUESTION 6.08 (1.00)
Which of the following condition (s) will Put the Integrated Control Swstem in to the tracking mode?
- a. Cross Limits
- b. Steam Generator Reactor Demand Hand / Auto Station in 'HANUAL'
- c. A Feedwater Loop Muster Hand / Auto Station in 'HANUAL"
- d. Both the Diamond Control Station in ' MANUAL
- AND the Rusetnr Demand Hand / Auto Station in ' HAND' O. Turbine E.H.C. not in operator I.C.S. mode of control
- f. A generator output breaker tripped
- d. The Reactor tripped
trtasts tra s
GUESTION 6.09 (1.00)
Which of the choices listed below correct 1w dveiets two conditions that cause I.C.S. to runback?
- o. o Loss of 1 RC Pump with 4 runnings I.C.S. runs back to 75%
reactor power at 50%/ min.
o Loss of a feedwater pump or feedwater booster pump, I.C.S.
runs back to 45% unit Loud demand at 50%/ min.
- b. o Aswametric Rode I.C.S. runs back to 60% reuetor power at 20%/ min.
o Loss of 2 RC Pumps with 4 running, I.C.S. runs back to 45%
reactor power at 50%/ min.
- c. o Loss of 1 RC Pump with 4 runninge I.C.S. runs back to 75%
unit load demand at 50%/ min.
o Reactor coolant flow limite I.C.S. runsback to a reactor ,
demand level eaual to 1.1 time flow at 30%/ min.
- d. o Loss of either a feedwater pump or feedwater booster pump, I.C.S. runs back to 55% unit load demand at 50%/ min.
o Reactor coolant flow limite I.C.S. runs back to a unit load demand level voual to 1.1 times flow at 20%/ min.
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- 6. ~_%CCEL18h A- e 4
GUESTION 6.10 (1 00)
Cssumind the I.C.S. is in its normal automatic linvun and Power Which one of the following statements most cutput is at 750 MWE.
cccuratelv describes the response the_1.C.S. Would take if one of the bvPass valves on the 'A' side failed open?
- o. The increased steam flow would start to dvervase loon 'A' Tc.
The delta Te controller would reratio feedwatere cutting back on the 'A' side and increasing 'B' side feed. 12 . H . C . will decrease turbine throttle setting to return header Pressure to setting.
- b. The increased steam flow would start to dvervose loor 'A' Tc.
The delta Tc controller would reratio feedwater reducing 'A' feed and increasing 'B' side feed to balance delta Tc. Reactor demand would Pull rods to recover Tave.
- c. The increased steam flow would start to decrease steam header Pressure which would then cause an error signal between header Pressure and set Pressure. This error sinnal would then be given to the control valves to close to compensate for the increased steam flow.
- d. The increased steam flow would cause a decrease in Tave ,
therebw causing the reactor demand to pull rods to compensate for the decrease. With the correction being greater than 5%e feedwater would be cross limited and inervased bw 2%
to makeup for the increased steam flow.
QUESTION 6.11 (2.00)
Listed below are four Parameters that input into the BTU calculator.
For wache indicate how each would have to change [ INCREASE, DECREASE or REMAIN THE SAME3 in order for,t.he BTU limit to be increased.
- a. Feedwater Temperature
- b. Hot Les Temperature i
- c. 0.T.S.G. Pressure
- d. R.C. Swsten Flow [4 9 0.5 va] (2.0)
Or,- - unats ce; s
GUESTION 6.12 (3.00)
Place the followind Hakeup 8 Purification Swetum components in the Proper ordere startins with the origin of Ivtdown to the suetion of the makeue Pumps.
- o. MU 8 P Domins 1. MUV-49
- b. RCP Seal return J. SuPP1w to RM-L1
- c. Prefilters
- k. LPI Piggw back surplu conn.
- d. Post filters 1. Return TO DHR
- o. Block Orifice m. Connection from FEED Supplies
- f. SuPP1w FROM DHR n. Letdown Coolers
- u. MU Tank u. MUV-64
- h. Connections to Cation Domins
[15 9 0.2 eu3 (3.0)
QUESTION 6.13 (2.00)
The operation of the control rod grapple,_ on the Main Fuel Handling Bridges is limited bw load indication on the Dillon Load Cell.
- a. The graPele cannot be LOWERED if the Dillon reads less than? ,
- 1. 1800 44 --
- 2. 2100 4
- 3. 600 4 (1,0)
- 4. 900 4
- b. The graPPlv cannot be RAISEDs if selveted for orifice rod testing, if the Dillon Load Cell reading is greater than?
- 1. 2750 4
- 2. 2650 4 - wmma.S
- 3. 24 50 4 - o n( m c**M ,
- 4. 2350 4 - M % ^ed M g gg,o) i
,-_d.,r - s ---- -- -, , , , -
"M M *
- Ca .
s A
GUESTION 4.14 (1.00) $)
e e - oovuv = = am - w omuaur- - - - - - - - - - -
BADIOLOGICAL_CDWIROL QUESTION 7 14 (2.00) I
' f I
You (as NSS) have diven permission to the I & C technician to perform the calibration check on the emergenew feed water DT8G A low low level trie.
- c. The I 1 C Technician has how long to perform this surveillance item ?
- 1. 30 minutes
- 2. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- 3. 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
- 4. 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> i
- b. At the end of the alloted time, the I & C technician reports f
that he will have to rvrair the bistable in order to finish the calibration check. Which of the following would be reuuired ?
l l 1. Declare the channel inoperable and allow I 1 C 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to rePaire ,
l' before reauiring a shutdown.
- 2. Dwelare the channel inoperative and be in hot stundbv in six hours.
- 3. Declare the channel inoperative and. commence the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> timins before going to hot standbv.
- d. Dwelare the channel inoperative and place the channel in a tripped Condition within i hour.
l l
GUESTION 7.17 (1.00) l l
Selvet the CORRECT statement from the following:
- a. Visitors at Crwstal River saw not reevive a dose in exevss of 1.25 ren/atr even if a NRC form 4 is cumpleted and furnished.
- b. Emergenew exposure (25 rem during or accidient or 75 run for life saving) must be included in the individual radiation exposure historw.
- c. The Nuclear Plant Manager has the authoritw to allow an individual to receive up to 3000 mrum in one week for special work assignments,
- d. The maximum week 1w exposure is 300 mrum. The Shift Superivor saw autorizer exposures to 600 mrum bw use of form 912801.
Bo__6D510lSIROMlWE_E80CEDWBLEA CDhD1110Hhc_6ub_LAbA16AAuub FAbt ea ,
l
.~
GUESTION 8.01 (1.00)
Which of the following would revuire action Per the Technical Specifications durins Power operations 7
- a. Condensate storse tank Ivvel of 140,000 mal.
- b. Hot well level of 145,000 sal.
- c. Hotwell level of 8 ft. 5 in. i l
- d. Condensate storage tank Ivvel of 33 ft.
QUESTION 8.02 (3.00)
After assumins the Nuclear Shift Supervisor watch and reviewing .
the logs wou are aware of the following Plant conditions.
- 1. HPI A is tassed out for maintenance (in Progress for 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />)
- a. For which of the above Plant Conditions do Technical Specifications reouire entrw of an action statement?
- b. Which of the below Technical Specification action statements APP 1w 7 l
- 1. 2 2.1
- 2. 3.1.3.7
- 3. 3.3.1.1
- 4. 3.5.1
- 5. 3.5.2
- 6. 3.5.3
- 7. 3 5.4
- c. In addition to the above Plant conditions 1 thru 5 wou are informed that the 3B Emergenew Diesel Generator (EDG) i has a large air leak in the air start motor and the swstem has isolated. Which of the below is the correct Technical Specification action ?
- 1. Shutdown to Hot Standbv ,
l
- 2. Shutdown to hot Shutdown
- 3. Shutdown to Cold shutdown
- 4. Shutdown not reuuired l
I 1
1 l
_. .- n s
s GUESTION 8.03 (1.00)
On wour shif t m . month 1w -surveillance item is discovered overdue.
Remuired due date was 25th of the monthe assume tudsw is the 31st,
'and the Performance of the SP has begun. All Previous surveillances were completed on time as scheduled. Which of the statements below is correct about the surveillance (SP)?
- a. The SP has been missed and the swstem must be dwelared inoperable until the SP is completed satisfactorw.
- c. The swstem is operable because the technical seveification allows a time extension and the extension has not been exceeded.
- d. The swstes is inoperable because the 3.25 time interval for 3 consecutive SP was not met.
QUESTION 8.04 (2.00)
Due to an ' event" wou are the Acting Emergenew Coordinator (AEC) per the Radiological Emergenew Responaw Plan. Indicate whether the following statements are TRUE or False.
- a. The AEC is responsible for activities in the control complexe the designated wavrsenew coordinator is reseunsible for the balance of Plante
- b. During an voorgenew the AEC is responsible for the direction of activities at all Crwstal River Plants (CR1 thru CR5)
- c. The AEC must obtain WPProval from the Nuelvar Plant Manager Prior to ordering an site evacuation.
1
L _ _4D B 1 W 1000488 WL5_LtLYM5MLifL1a _ w e
CUESTION 3.05 (1.50)
TRUE or FALSE ,
- o. The Radiation Emersonew Response Plan is designed to Provide procedures for the reponse to incidents / accidents other than ,
the FSAR postulated accidents.
1
- b. An on site fatalitw reuuires at least an alert clasification I to obtain reeuired offsite medical assistance. -l j
regardless of the emersonew classification.
GUESTION 8.06 (1.00) , j i
If reeuired to notifw the State Warning Pointe Tu11uhausev (SWPT) l the order of telephone preferenew is ?
- c. COMME NAWAS, SHRDTS (NOTE: COMM= COMMERCIAL NAWAS= NATIONAL WARNING SYSTEM
- b. NAWAS, COMME SHRDTS SHRDTS= STATE HOT RING DOWN TEp. SYS.)
- c. SHRDTS, COMM, NAWAS
- d. SHRDTSe NAWAS, COMM QUESTION 8.07 (1.00)
If a general emergenew has been declared what is the minimum Protective action recomendation 7
- a. Evacuate all People within a i mile radius and shelter all PeoPlv in a 2 mile radius. (affected sectors)
- b. Evacuate all people within a 2 mile radius and shelter all Poorlv in a 5 mile radius. (affveted svetors)
- c. No protective action is to be recomended univss radiation exposure to the Public will exceed 5 mr.
- d. Station additional securitw at site entrance to prevent unethorized Personnel from entering a radiation area.
aus-1sueastxusens tseuuensueueu-vuur:mswmsumsetzt, e -
o e
QUESTION 8.08 (1.00)
Telephone notification of an emergenew is comelvtud when which of the below is established 7
- a. Voice contact with responsible rweresentative.
- b. Voice contact with adenew's telephone operator.
- c. Voice contact with agenew's sverstarw.
-d. Emersonew information Placed on a recording device GUESTION 8.09 (1.00)
TRUE or FALSE. To remain at Power after a DC swstem is declared ,
inoperable wou must verifw the operabilitw of the redundant DC swstem within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
QUESTION 8.10 (3.00)
Indicate the direction the following Plant Parameters will move (INCREASES DECREASE, NO CHANGE) as a result of a Power reduction
! from 75% to 25% (steadv state eewer to steadv state nowe+)
- a. Th
- b. Tc
- c. xenon
- d. Pzr level
- e. OTSG 1evel
- f. OTSG superheat GUESTION B.11 (2.00)
Indicatee TRUE or FALSE, whether each of the following work schedules are in compliance with the OSIM guidelines.
< a. Workins from 8 A.M. to 10 P.M.
- b. Working daw shift Mondaw (0800-1600) and midnights Tuesdaw (0000-0800).
- c. Working midnight to noon (0000-1200)
- d. Workins Mondaw thru Sundawe 0800-2000
,, , - - -,-,,--n-n. . .. . - . , ..v,- .----,..---.-.--,_r., , - , , , . - - - - -- , - . - . , , , ,,
Ba__6ENI e i
4 (1.00) ;
QUESTION 8.12
.The operations studw book is used to disseminute information and is considered to be Part of!
- a. The SS loss
- b. Training material
- c. SOTA's responsibilities
- d. Shift relief O
l
Lk- 03RDS trwL% r
- a. o QUESTION 8.13 (2.00)
Durins a reactor starture with Power starting to indiente on the Power renser a intermediate rense channel failu low.
- a. Which of the following is the correct action to be taken in accordance with technical seveifications ?
- 1. With the number of operable channels one less than the total number of channels, startuP and Power operation muu proceed Provided the inoperable channel is placed in the tripped condition in one hour.
- 2. With the number of operable channels one leuu than reuuired bw the minimum channel operable ruuuirements and with thermal Power Ivvel less than or eeusi to 5% of rated, ruutore the inoperable channel to operable status Prior to increasins thermal power above 5% of rated thermal power. .
operation saw continue until the next reuuired channel functional test provided the inuPerable channel is placed in the tripped condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- b. While trwing to satisfw the action statement in part se the reactor Power increased and mode 1 was entered.
Which of the following is the corrvet action to be taken ?
- 1. Reenter mode 2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and complete the reuuired action Previous 19 initiated.
- 2. Complete the reuuired action and then continue the Power accession.
- 3. Within one hour commence a reactor shutdown and be in hot standbw within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- 4. Continue the Power accession and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notifw the NRC of the tvehnical specification violation.
uy__u9JUUUuuu_n5cus_wutxxsu umwe -
o s'
00ESTION 8.14 (1.00)
The nuclear instrumentation rouuired to be cPerable in the control room Curins core alterations Per FP-203 is/are :
- o. 2 SR
- b. 1 SR
- c. 2 SR and 2 IR
- d. 1 SR and 1 IR QUESTION 8.15 (1.00)
Chich statement below describes the corrvet operator uuv of the ..
PORV to Prevent an over Pressure trip of the reactor 7 a.-The PORV is not to be used anwtime bw the oPerstor if the reactor is critical unless RCS Pressure exceeds the PORV setPoint.
other responsibilities).
, QUESTION 8.16 (2.50) e ---
}. . . s. . '
Fill in the blanks with the reuuired number of PerGonnel.
The Operations department will suPPlw __a__ Personnel for the fire brigade team from each shift. The Chemistrv and radiation control Dept. shall Provide __b__ fire brigade team member (s) on duu shift and __e__ team member (s) on the back and aid shift, __d__ daw (s) a week.
The maintanence Duet. shall Provide __e__ team member __g__
(s) on daw shift and __f__ team member (s) on the back and mid shift daw (s) a week.
I l
_ _ _ - . . _ ~ __ -. . . _ . . _ _ . . . . . _ _ . , _ _
$ EQUATION SHEET o f = ma v = s/t Cycle efficiency = (Net work cut)/(Energy in) l 2
w = mg s = Vg t + 1/2 at 2 , l E = mc KE = 1/2 my A = IN A = Ag e'*
a = (Vf - Yo )/t PE = mgn w = e/t 1=
Vf = V, + at tn2/t1/2 = 0.693/t1/2 t
, p
- 2 1/2*If " E5tU)(t)3 b A= 4 [(t1/2) + (tb I) t.E = 931 an ** c av Ao I,Ie Q = mCpat I
- I c
e~"
6 = UAa T .
Pwe = Wfah I=I g 10-x/TVL TVL = 1.3/u P.= P 10 sur(t) HVL = -0.693/u t
P = Po e /T SUR = 26.06/T SCR = S/(1 - K,ff)
CR x = S/(1 - K,ffx)
SUR = 26o/t* + (a - o)T CR)(1 - Kdf1) = CR2 (I ~ keff2)
T = ( t*/o ) + [(a - o ]/ Io ] M = 1/(1 - Kdf) = CR)/CR, T = 1/(o - 8) M = (1 - K ,ffa)/(1 - Keffl)
T = (3 - o)/(Io) SUM = ( -Kgf)/K,ff
/K eff t= 10 seconds a = (Kgf-1)/K ,ff = t.Keff I = 0.1 seconds-I o = [(t=/(T K,ff)] + (a,ff /(1 + IT)]
I)d) = I d P = (tov)/(3 x 1010) I d) 2 ,2gd 2 22 2 I = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (feet) ,
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbm. I curie = 3.7 x 1010 dps 1 gal. = 3.78 liters i kg = 2.21 lem 1 ft3 = 7.48 gal. I np = 2.54 x 103 Stu/hr Density = 62.4 lbm/ft3 1 mw = 3.41 x 10o Stu/hr Density = 1 gm/cn.3 lin = 2.54 cm Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32 Heat of fusion = la4 Stu/lbm 'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf i ft. H O 2
= 0.4335 lbf/in.
A
- :. , , *NCO 1. 0C.WWFL Feu.EDW[ILLGLCON2 cI@3 ms
' ' ' 'Ik'EFMI Gyscb1CS '
-[' CNS lr.RS -
CRYGTAL RIVER -84/12/17-KING, k.
MASTERCOPY C.NSWER 5.01 (3.00)
- o. Falso
- b. False
- c. True
- d. False
- o. True C6 0 0.5 va3 (3.0)
- f. False REFERENCE Westinshouse Thermal Science, Chapters 3e5 1 10.
ANSWER 5.02 (1.00)
A. False
[2 E 0.5 eu3 (1.0)
B. False REFERENCE B&W swstem descriptions ANSWER 5.03 (1.00) b.
REFERie:CE Thermodunamics, Fluid f)uwe and Heat Transfer for Nuclear Power Plants ANSWER 5.04 (1.50)
- a. Increase.
- b. Live re s s e .
Increase. [3 9 0.5 eu3 (1.5) c.
REFERENCE Westinghouse Thermal Seierawl Ch. 13e FP 33-52.
5 __ISEORX DE WUCLL6E EDWLE t'Lmmi utthalAuse_cuuluso chu tsut sa
, TWEREDDYC_^.2ICS
- - CESTERS -- CRYSTCL RIVER -84/12/17-KING, M.
CNSWER 5.05 (1.50)
- o. False
- b. False
- c. True [3 0 0.5 eu3 (1.5)
REFERENCE CNO 1 AOP 1203 13 PM i of 5e E0 1203.01 pg 71 of 146 and STM-1-69 PW 12 ANSWER 5.06 (3.00)
O. 42
- b. 43
- c. 3 [ answer for Part c will be gruded indveendantlw of Part b3 C3 9 1.0 vu3 (3.0) hNd, mm 6 f. coq
+H REFERENCE '
Crwstal River Question Bank Categorw 115 Question 417 ANSWER 5.07 (1.00)
- a. False
- b. True [2 0 0'.5 vu3 (1.0)
REFERENCE Crwstal River Question Bank Catesura 1 1 5 Question 426 ANSWER 5.08 (2.00) a-3 b-5 e-4 (2.0) d -{f) gja, C4 0 0.5 eu3 g) {*
_ pu3 sn- &Am 1-3 REFERENCE &M 164 ( A)8C Mu NI FF g i Crvstal River Question Bank Catsuurw 1 1 5 Question 45
5... ICE DR Y .L%
, 2ES20DXWODICS
- -84/12/17-KING, M.
.A SEERS -- CRYSTCL RIVER .
C%SWER 5 09 (3.00) b 2 -- Du$m.ch w% h M A M3 QM- j C3 e 1.0 va3 (3.0) e-3 CEFERENCE Crwstal River Guestion Bank Catwsurv 185 Question 412 ANSWER 5.10 (2 00)
- o. increase
- b. remain constant
- c. decrease E4 8 0 5 wu3 (2 0)
- d. increase REFERENCE Crwstal River Question Bank Categurw 155 Ouvstion 422 ,
CNSWER 5.11 (1.00) d.
REFERENCE Reactor Fundamentals ANSWER 5.12 (2.00)
- c. - 1 (2.0)
- b. -3 C2 8 1 0 vu3 REFERENCE Crwstal River Question Bank Catusurw 185 Ouvstion 441 i
l i
itcysts 83 5 a . ..I W E O E Y .0 E _ W U C L E e B _ E Q W E E _f] L COM L' ELM 366Bl'O a -lie @JEE a CrJ6
, ICERMODYCOCICS
~
C% SEERS -- CRYSTCL RIVER -84/12/17-KINGe M.
CMSWER 5.13 (3.00)
- c. True
- b. False
- c. True
- d. False
- o. Falso
- f. True C6 9 0.5 eu3 (3.0)
REFERENCE BSW Operation fundamentals t
h
6.._ELeWI 51583t19_L9196LOz- v gngts _e
-84/12/17-KING, M.
b>: SEERS--CRYSTAL _ RIVER ANSWER 6.01 (1.00) -
c.
REFERENCE Crwstal River Swster Training Manual See 7 Non-NucIvar Instrumentation es STM-7-26 1 27 CNSWER 6.02 (2.00)
- o. False
- b. True
- c. False
- d. False C4 6 0.5 eu3 (2.0).
REFERENCE Crwstal River Swstes Trainind Manual Sec 6 Nuclear Instrumentation Sws Ps STM-6-11, 15, 17 1 25 o
CNSWER 6.03 (2.00)
- o. - A
-2 E2 6 1.0 vu3 (2.0) b.
REFERENCE Crwstal River Swstem Training Manual Sec 10 Emerdeneu Diesel Generators es STM-10-1, 37 1 38 CNSWER 6.04 (1.00)
- d. CCAF3 REFERENCE Crwstal River Swstem Training Manual See 10 Emergenew Diesel Generators es STM-10-45, 46 1 47 ANSWER 6.05 (1.00) b.
REFERENCE Crwstal River Swstem Training Manual See 11 Engineered Safeguard Actuation es STH-11-14
(tsg gr en. tLeWI_arsIEns_DEsloWa_CDWIBDLa CWS 8C680KBD8688@0 CNS ERS -- CRYSTCL RIVER -84/12/17-KINGe M.
-CNSWER 6.06 (1.00)
- o. True
- b. False [2 9 0.5 vu3 (1.0)
REFERENCE Crwstal River Swstem Training Manual See 11 Enginevred Sufvuusrds Acuation Ps STM-11, 12 1 13 ANSWER 6.07 (2.00)
O. - 1.
- b. - 3. [2 9 1.0 eu3 (2.0).
~
REFERENCE Crwstal River Swstem Training Manual See 12 Control Rod Drive em STM-12-13e 14 1 15 s
ANSWER 6 08 (1.00) se be-de e 1g C5 9 0.2 wu3 (1.0)
REFERENCE' Crvstal River Swstem Training Manual See 13 Integrated Control Sustem Ps STM-13-19 ANSWER 6.09 (1.00) d.
REFERENCE Crwstal River Swr, ten Training Manual See 13 Integrated Control Swstem PW STM-13-21 ANSWER 6.10 (1.00) c.
REFERENCE Crvstal River Swstem Training Manual See 13 Integrated Control Susten Ps STM-13-6 thru 38
, - -- _ ,, , _ . - , _ - , . _ - __m . . , , - - - , , , . . _ , , -
hMS"ERS -- CRYSTAL RIVER -84/12/17-KING, M.
CNSWER 6.11 (2.00) ,
- s. Increase
- b. Increase
- c. Decrease C4 0 0.5 eu3 (2.0)
- d. Increase REFERENCE Crwstal River Question Basnk Cat 3 1 6 Question #16 ANSWER 6.12 (3.00) ne le we fe Je ce he se me de le be de ce k E15 5 0.2 vu3 (3.01 REFERENCE Crwstal River Question Basnk Cat 2-6 Question 435 Swstem Training Manual Sec 17 MU S P Pd STM-17-2 Simplified Flow Diagram o
ANSWER 6.13 (2.00)
- a. -
1.
- 4. [29 1.0 ea3 (2.0) b.
REFERENCE Crvstal River Swstem Training Manual See 21 Fuel Handling, es STM-21-47 6.14 (1.00) 7' ANSWER 1
- c. ECAF3 REFERENCE Crwstal River Swstem Training Manual See 27 Fevdwater pg STM-27-56e 59, 60, 61 8 67 ANSWER 6.15 (1.00) d.
' 6 a...
vums ev-q bMSWERS'-- CRYSTCL RIVER -84/12/17-KIN 0s M. )
REFERENCE Crvstal River Fvedwater Swstem operatins Procedure OP-605 Rev 28 Ps 2 ANSWER 6.16 (2.00)
- o. - Deluse
- b. - Frvon FE-1301
- c. - Delude
- d. - Wet Pipe Sprinkler C4 5 0.5 va3 (2.0)
REFERENCE Crwstal River Swstem Trainins Manual Sec 38 Plant Fire Protection Swstem em STM-38-3 1 5 T .
CNSWER 6.17 (1.00) c.
REFERENCE Crwstal River Swstem Trainins Manual See 16 Communications Sustem Ps STM-16-3, 5, 10, 13 1 15 ,
I t .
,,,,..,__,,,,m.
_ . _ _ . . , .,_.[___,,, ._ , , _ _ . _ _ _ _ , _ _ . _
victr w
- 2. ..ELT'EfAIR193_ = _
&&DIOLDOIceL_CDMIRDL AN3:ERS -- CRYSTAL RIVER -84/12/17-KING, M.
CNSWER 7 01 (1.00)
Deced,3e (4 8 0.25 es.)
REFERENCE CR Gues. Cat 4 1 7e 427 CNSWER 7.02 (1.00) b REFERENCE OP-203, em 2 CNSWER 7.03 (1.00)
- d. ,
REFERENCE Radiation Protection Manual RP-101 em 5 86 ANSWER 7.04 (3.00)
- o. 3
- b. 1
- c. 3 REFERENCE CR Dues Cat 4 1 7,439 (OP-204,AP-521)
ANSWER 7.05 (2.00)
- o. F
- b. F
- c. F
- d. T (4 8 0.5 ea.)
REFERENCE OP-502 rev 11, em 3 (sec 4.0)
',- Z o...t B O C E DUR E S _ :: _ UL'11XL a _hsTe --'- m c1-I
, _R&DIOLOGICAL_COWIROL
, ." -84/12/17-KYMG, M.
. CNSWERS -- CRYSTAL RIVER 1
ANSWER 7 06 (1.00) b REFERENCE OP-502e PW 14 ANSWER 7.07 (1.00) c REFERENCE OP-504 rev 7, Ps 5 ,
ANSWER 7 08 (1.00) c o
REFERENCE
'OP-603 rev 21, em 4
-ANSWER 7.09 (1.00) b REFERENCE OP-605 rev 28, Ps 5 CNSWER 7.10 (2.00)
- c. True
- b. False
.c. False
- d. True REFERENCE EM-202 rev 23, em 5
UA - N-
' j d&DIOLOGICCL_CD'IROL
. C;;s'.ERS -- CRYSTCL RIVER
' -84/12/17-KING, M.
CN8WER 7.11 (1.00)
- c. F
- b. F REFERENCE OP-607 rev 9e Ps 3
' ANSWER 7.12 (2.00)
- o. Falso
- b. True
- c. False
- d. True REFERENCE CR Ques CAT 4 1 7 630 (AP-390)
ANSWER 7.13 (2 00) 74 k'/ ,
[,, g a & M-%i N *l."$r
- b. True
- c. True
- d. True REFERENCE AP-961 Ps 2/AP-580 rev 2 PW 1/AP-543 rev 1e pg 2/CR Tech Spec. 3/4 1-18 ,
I ANSWER 7.14 (1.00) l 1
REFERENCE FP-203 rev 12, pg 12e sec 5.18 j 1
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- -84/12/17-KING, M.
CNSWERS -- CRYSTAL RIVER CNSWER 7.15 (2.00)
- o. suidance
- b. guidance
- c. suidance
- d. step-bw-ster REFERENCE AI-500 Ps 8 89 ANSWER 7.16 (2.00)
- a. 2
- b. 2 REFERENCE '
CR Tech Spec. Table 3.3-3 action ice pg 3/4 3-14 ANSWER 7.17 (1.00) b.
REFERENCE Radiation Protection Manual RP-101 eg 9, 11 L 12 und IN bulletin 84-40
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-84/12/17-KING, M.
. 'CKS'_'ERS -- CRYSTCL RIVER ANSWER 8 01 (1.00) -
O REFERENCE OP-603 rev 21e Ps 2 e ;
M'L_**a ANSWER 8.02 (3.00) ,
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- b. 4e 87.(3.5.is 3 2, 8 3.5.4) i-
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c 3 s- A. s REFERENCE CR Tech Specs ANSWER 8.04 (2.00)
- o. F
- b. T
- c. F
- d. F REFERENCE Radiological Emersenew Response Plan Pg 6-Se sec 6.3.1 C.NSWER 8.05 (1 50)
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- a. True
- b. False
- c. True REFERENCE Radiological Emergenew Response Plane PW 8-5/8-7/D-1
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d REFERENCE Radiolusical Emergence Response Plan es 9-6 CNSWER 8.07 (1.00) b REFERENCE EM-202 rev 23, Ps4 ANSWER 8.08 (1.00)
O REFERENCE ,
EM-202 rev 23, Ps 7 l
CNSWER 8.09 (1.00)
False REFERENCE OP-705 rev 3e es 5
-CNSWER 8.10 (3.00)
- c. decreasse
- b. increase
- c. . increase
- d. no change
- o. decrease
- f. decrease *'-
REFERENCE Various CR ICS and RX theurv lessons
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'CNS:.'ERS -- CRYSTCL RIVER -84/12/17-KINGe M.
ANSWER 8.11 (2.00) -
- a. Falso
- b. False
- c. True
- d. False REFERENCE OSIM Ps III-5 CNSWER 8.12 (1.00) ,
b REFERENCE OSIM Ps III-11 CNSWER 8.13 (2.00)
- a. 2
- b. 3 REFERENCE CR Tech Spec 3/4.3 table 3.3-1 69 Ps 3/4 3-2,3e8 4 3.0.3 Ps 3/4 0-1 ANSWER 8.14 (1.00) a REFERENCE FP-203 rev 12e Ps 1 ANSWER B.15 (1.00) c REFERENCE OSIM see Ve Ps V-22e Interoffiew Corrww. OP83-204
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, IWSCERS--CRYSTALRIVER CNSWER 3.14 (2.50) -
- o. 3
- b. 1
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b
- d. 7 ' i
- o. 1 /
- f. 2
- 0. 7 REFERENCE OSIM see V, en V-21
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I NUCLEAR REGULATORY mEGloN H ON COMMISSI/ :
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ATLANMlUEOnG A 30323 6ndoso(E 3 REACTOR OPERATOR LICEN ON ION Facility:
Reactor Type:h 1 River PWR B&W INSTRUCTIONS 10 CANDIDATE:
Da te AdministerDecenib'er'T7 Examiner:
Candidate: 5i Use separate paper for sheet on top of answers.
parenthesis cate after the questionthe
. Points answers foron one side sheets.
only. Write answers six 6) (goryhours and a final after the grade examinatio .
of at least 80*4The passing grad n n starts. Examination papers will be ri keast 70 Category
% of c ed up Value % of Total Candida te 's -
25 Score Category 25 Value 1.
Principles of Nuclear 25 25 Thermodynamics, HeatPo ,
Transfer Flow and Fluid 2.
25 25 Safety and ErrergencyPla Systems 25 3.
25 Instruments and Jontrols 4
100 Procedures - Normal Abnorma l, Emergency,,
and Radiological Control Final Grade TOTALS All work done aid. %
1 on this examination is my own .
I have neither given nor e ved rec i
+ 4
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.- l CATEGORY 1: Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow (25.0) 1.01 Select the statement that best describes how the three heat transfer regions in the OTSG change as power increases.
Nucleate Boiling Film Boiling Superheat
- a. remains constant increases decreases
- b. increases remains constant decreases
- c. decreases remains constant increases
- d. increases decreases remains constant 1.02 Which of the following statements is most nearly accurate regarding -
control rod worth?
(a) It is proportional to reactor power.
(b) It is proportional to rod speed.
(c) It is higher in regions of higher relative neutron flux. ,
(d) It is not dependent upon rod position.
1.03 Which of the following is NOT a characteristic of subcritical multiplication?
l (a) If the reactor is shutdown long enough, the source range instruments will lose their ability to. determine the subtritical multiplication level even though the core may still be at MOL.
(b) Doubling the indicated count rate by reactivity additions will reduce the margin to critical by approximately one half.
(c) For equal reactivity addi ti or.s , it takes longer for the equilibrium subcritical multiplication level to be reached as Keff approaches unity.
l (d) If ten inches of rod withdrawal increases the subcritical multiplication level by 10 cps, then twenty inches of rod withdrawal will increase the subcritical multiplication level by-approximately 20 cps.
l l
^
- 2 1.04 Which of the following demonstrates the effects of the delayed neutron fraction changing over core life?
(a) A lower boron concentration.
(b) A higher rod bite.
(c) A higher startup rate for equal reactivity additions.
(d) A larger (more negative) moderator temperature coefficient.
1.05 An estimated critical position has been calculated for a reactor startup that is to be performed 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> af ter a trip following a 60-day full power run. Which of the following actions will contribute to a higher actual rod position than the calculated ECP?
~
(a) Controlling the OTSG Levels above the low level limit.
l (b) Delaying the startup six hours longer than anticipated.
(c) Increasing the APSR position for criticality from 0 to 20*4.
(d) Using the 579.F reactivity vs Boron curve instead of the 532 F .
reactivity vs Boron curve.
1.06 Which of the following statements describes the behavior of Xenon and
, Samarium?
l (a) After a reactor trip occurs, Xenon concentration initially l increases and Samarium initially decreases. ,
l (b) Af ter a reactor trip occurs, Xenon will eventually decay to a Xenon free condition but a Samarium free condition will not occur until after the next refueling outage.
(c) The Xenon and Samarium peak concentration following a trip occurs at a time independent of the previous power level.
(d) Xenon concentrations may increase or decrease when taking the plant from Mode 3 to full power but Samarium will always decrease during this transient af ter the core's equilibrium Samarium has been reached.
1.07 Which of the following radioactive isotopes found in the reactor coolant would NOT indicate a leak through the fuel cladding?
(a) I-131
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f 1.08 Which of the following is a true statement concerning radioactive decay? Remember the atomic number is the number of protons and the mass number is the number of neutrons plus protons.
(a) When an element decays by beta emission, the new element will have increased in atomic number by one and the mass number will remain the same as the original element. .
(b) When an element decays by alpha emission, the new element will have decreased in atomic number and mass number by two, from the
. original element.
(c) When an element decays by neutron emission, the new element will have increased in atomic number by one and decreased in mass number by one, form the original element.
(d) When an element decays by gamma emission, the new element will .
have increased in. atomic number by one and the mass number will remain the same as the original element.
1.09 During power operation of 'a nuclear reactor the design DNBR will be affected by changes in some of the operating parameters. The four operating parameters of maximum interest to the operator are:
(a) Reactor power, pressure, coolant flow, and average temperature.
, (b) Reactor pressure, coolant flow, average temperature, and boron I
concentration.
(c) Reactor coolant flow, average temperature, boron concentration, and power.
I (d) Reactor average temperature, boron concentration, power, and pressure.
t.
1.10 With the main steam temperature and pressure at 600 F and 900 psia respectively, a main steam relief valve seat begins to leak to atmospheric pressure. The temperature of the steam three feet out of the relief valve is approximately:
(a) 600 F (b) 532 F (c) 444 F (d) 212 F l s-. . . .
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1.11 The quality of steam exiting the HP turbine refers to (a) the ratio of the 1.iquid mass to the vapor mass.
.(b) the ratio of the vapor mass to the liquid mass.
(c) the ratio of the liquid mass to the sum of the liquid and vapor masses. *
(d) the ratio of the vapor mass to the sum of the liquid and vapor masses.
1.12 The reactor coolant system is subcooled by approximately during Mode 3 when Tave is 400 F and the pressurizer pressure is 1000 psia.
(a) 145 F .
(b) 125 F (c) 100 F (d) 75 F 1.13 The following signals are used to derive the BTU limit in the ICS.
Indicate which of these would decrease to increase the BTU limit.
(a) Th (b) S/G pressure (c) RC flow (d) FW temperature 1.14 The amount of aspirating steam (lbm/hr) used in the OTSG .
(a) Increases as power increases from 10 to 100%
(b) Increases as the temperature of the feedwater increases (c) Increases as the feedwater flow decreases (d) Decreases as the temperature of the feedwater decreases l
l 1
4 g
. 5 a-1.15 If a centrifugal pump is operating at 1800 rpm to give 400 gpm at a discharge head of 20 psi, what would be the discharge head if the speed is increased in order to deliver 1600 gpm?
(a) 40 psi (b) 80 psi ,
(c) 160 psi (d) 320 psi 1.16 Which one of the following is NOT one of the four contributors or factors that establish equilibrium Xenon?
(a) Direct production from fission (b) Decay of Iodine (c) Decay of Xenon to Cs (d) Decay of Xenon to Sm 1.17 Which one of the following is TRUE concerning the change e in differential boron worth (% Ak/k) with RCS boron concentration (range of 0 to 1800 ppm) and Tave (range of 532*F of to 579 F)
(a) It increases as Tave and RCS boron concentration increase.
(b) It decreases as RCS boron concentration increases but is constant as Tave increases, (c) It decreases as Tave and RCS boron concentration increase.
(d) It increases as Tave increases but is constant as RCS boron concentration increases.
1.18 Figure 1.18 is a representation of how the resonance peaks of U-238
" flatten cut" or Doppler broaden as fuel temperature increases. Which of the following are the correct labels for the X and Y axes?
(a) X is neutron flux, Y is interaction rate.
(b) X is neutron energy, Y is microscopic capture cross section.
(c) X is atom density of U-238, Y is neutron flux.
(d) X is interaction rate, Y is neutron density.
II ,.
- 6
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1.19. The ratio of Pu-239 and Pu-240 atoms to U-235 atoms changes over core life. Which one of the pairs of parameters below are most affected by this change?
(a) Moderator temperature coefficient and doppler coefficient.
(b) Doppler coefficient and beta ,
(c) Beta and moderator temperature coefficient (d) Moderator temperature coefficient and neutron generation time.
1.20 A moderator is necessary to slow neutrons down to thermal energies.
Which of the following is the correct reason for operating with thermal instead of fast neutrons?
(a) Increased neutron efficiency since thermal neutrons ' are less likely to leak out of the core than fast neutrons.
(b) Reactors operating primarily on fast neutrons are inherently unstable and have a higher risk of going prompt critical.
(c) The fission cross section of the fuel is much higher for thermal energy neutrons than fast neutrons. '
(d) Doppler and moderator temperature coefficients become positive as neutron energy increases.
1.21 Which one of the following factors will help, rather than hinder, natural circulation?
(a) Lowering OTSG level (b) Lowering RCS pressure (c) Increasing RCS temperata e (d) Lowering turbine bypass valve setpoint !
1
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.1.22 Regulating rod group insertion limits change as a function of. core life (EFPD) and number of operating RCPs. Which one of the following is true concerning the change in insertion limits? (An increase in area of acceptable operation means less restrictive rod insertion limits.)
(a) The area of acceptable operation decreases as you go from 4 RCP to 3 RCP operation. .
(b) The area of acceptable operation increases as you go from 4 RCP to
'3 RCP operation.
(c) The area of acceptable operation increases with core life.
(d) The area of acceptable operation is the same as you go from 3 RCP to 2 RCP operation.
1.23 Following a trip from full power with the reactor shutdown and 4 RCPs operating, the 125 psi bias is suddenly removed from the turbine bypass valves. Which one of the following statements best describes plant response?
(a) OTSG pressure drops and levels rise. The increased OTSG levels cause an overcooling of the RCS.
(b) The OTSG saturation temperature drops causing a decrease in RCS T c and a rapid drop in pressurizer level.
(c) Since OTSG pressures drop 125 psi, BTU limit alarms will be received on both generators and feedwater will cut back.
(d) The resulting cooldown of the RCS will probably' decrease the shutdown margin to less than Tech Spec limits.
i
- b. ,
. 8 1.24 The reactor core Safety Limit Curve (Figure 2.1-1 in the Tech Specs) is prevented from being exceeded by a combination of four RPS trips.
Select the combination.of RPS trips below that define the reactor trip envelope.
(a) RCS Outlet Temp.-High Nuclear. overpower Low RCS Pressure Overpower based on RCS Flow and Axial Imbalance (b) Low RCS Pressure High RCS Pressure RCS Outlet Temp.-High Variable Pressure - Temperature Trip (c) High RCS Pressure RCS Outlet Temp.-High .
Overpower based on RCS Flow and Axial Imbalance RCP Power Monitors (d) Low RCS Pressure Variable Pressure - Temperature Trip Nuclear Overpower Overpower based on RCS Flow and Axial Imbalance ,
d Which one of the following is NOT one of the three bases for the g control rod insertion limits?
Spd (a) Ensures acceptable power distribution limits are maintained (b) Ensures maximum fuel clad temperature will . not exceed 2200 F following a LOCA (c) Ensures that the minimum shutdown margin is maintained (d) Limits the potential effects of a rod ejection accident.
END OF SECTION 1
(
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2.0 Plant Design, Including Safety and Emergency Systems 2.01 With regard to Plant Fire.. Protection Systems, the deluge water system protects all but one of the following areas. Which one is NOT protected by the deluge water system?
- a. Charcoal plenums in Auxiliary Building and Control Complex
- b. Startup and unit auxiliary transformers
- c. Diesel generator. rooms
- d. Cable spreading room 2.02 Select the CORRECT statement about the Makeup and Purification System.
- a. .The block orifice has two bypasses, (MUV-51 and MUV-48) both of which
- are remotely operated from the control room,
- b. A temperature element (TE-5) on the letdown line alarms at 130 'F and closes the letdown cooler outlet valves (MUV-40 and MUV-41) at 135 'F.
- c. The letdown line connections to the Decay Heat- Removal System are 2h-inch lines prior to the prefilters and after the makeup filters.
- d. The deborating demineralizer may be operated in parallel or series with the makeup demineralizers.
2.03 Select the CORRECT statement concerning the Makeup Pump Lube Oil System.
- a. If the main gear oil pump control switch is in Auto, the pump will start and run for three minutes after the makeup pump starts,
- b. The backup gear oil pump will start (if in Auto) when oil pressure reaches 7 psig and will automatically stop when oil pressure reaches 20 psig.
- c. If the main lube oil pump control switch is in Auto, the pump will start and run for three minutes after the makeup pump starts,
- d. The backup lube oil pump has no auto start provisions and can be used as a back up for the gear oil system.
.2.04 With regard to the Plant Ventilation System, which one of the following Ventilation Systems are required for emergency operation?
- a. Reactor Cavity Cooling Fans
- b. Reactor Building Purge Supply System
- c. Decay Heat Closed Cycle Cooling Pumps Air Handling Units
- d. Reactor Building Operating Floor Fans
. 2
~~
2.05 Select the CORRECT statement concerning the Nuclear Services Booster Pumps and CRD Cooling System.
- a. One pump is normally operated with the other serving as backup. A drop in line flow (<100 gpm) will start the idle pump.
- b. On an ES signal, the supply and return valves will close and the booster pumps will have to be manually secured. *
- d. Maximum allowable temperature of the cooling water is 120 F; there are no limits on minimum temperature.
2.06 With regard to the Reactor Building Spray System, which of the following statements is TRUE?
- a. The two RB spray pumps are cooled by Nuclear Services Closed Cycle Cooling System.
2.07 A diesel generator has been started as a result of an ESS Actuati , Signal.
Which of the following will NOT cause the diesel to shutdown?
- a. Low oil pressure
- b. High crankcase pressure
- c. Remote STOP pushbutton
- d. Emergency STOP pushbutton
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2.08 Which one of the following correctly describes the trip system of the main turbine?
- a. When the auto-stop (turbine control) oil pressure decreases, the interface trip valve will open allowing the EHC Control Oil to dump to drain. ,
- b. When the EHC Control Oil pressure decreases, the interf ate trip valve will open, allowing the auto-stop (turbine control) oil to dump to drain.
- c. The interface trip valve is solenoid actuated and when open, will dump both auto-stop (turbine control) oil and EHC control oil to drain.
- d. A full turbine trip requires the servo valves for all four sets of turbine valves (throttle, governor, reheat and interceptor) to open. ,
2.09 Which of the following statements about the Instrument Air System is INCORRECT?
- a. In an emergency, the Instrument Air System is automatically cross connected with the House Service Air System through IAV-30.
- b. .The Instrument Air System serves as a backup supply to the Condensate Polishing Air Compressors through manual valve IAV-25.
- c. The normal cooling water supply is Secondary Services Closed Cycle Cooling but in extended outages, may be shifted to Nuclear Services Closed Cycle Cooling.
- d. In an emergency, when cross connected with the House Service Air System, the filters and dryers are bypassed.
2.10 Which statement is TRUE concerning the Core Flood (CF) System?
- a. Isolation valves CFV-5 and 6 receive an open signal following ES actuation even though they are required to be open with their breakers in the " Locked Reset" position.
- b. When the breakers for CFV-5 and 6 are in the " Locked Reset" position, they lose position indication in the control room.
- d. During plant operation, high CF Tank pressure may be relieved by venting to the Reactor Building.
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2.11 Which of the following statements about the Decay Heat Removal (DHR) System
-is INCORRECT?
- a. DHV-3 and DHV-4, RC system isolation valves to DH interlocks, are set at 284 psig RC system pressure.
c.' Maximum pressurizer level is 220 inches when DHV-3 or 4 is open and one makeup pump is in operation. (This precaution limits potential for overpressurization of the DH system.)
- d. The high and low flow alarms (3400 gpm and 2800 gpm respectively) are the same for both the Injection and Recirculation Phase.
2.12 During Long-term Post-Accident cooling, which one of the following flow paths is most desirable?
- b. Condition "B"; Open auxiliary sprey line to pressurizer
- c. Condition "C"; Combination of Conditions "A" and "B"
- d. Condition "D"; Backflush with Reactor Coolant Pump.
2.13 Which of the following statements is TRUE concerning the OTSG?
- a. The minimum level in the generator (low level setpoint) thermodynamically provides the reference for no-load Tave.
- b. The startup range instruments will provide indication of flooding of the aspirating ports.
- c. The auxiliary feedwater header penetrates near the top of the OTSG shell and sprays the feedwater on the upper cylindrical baffle.
- d. Orifice plates, located in the lower downcomer section may be adjusted to balance out the internal circulation system.
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2 2.14 Which one of the following statements concerning the Reactor Coolant Pumps I l
is INCORRECT 7
- a. An RCP motor may be s' tarted three times successively from ambient temperature, or twice from rated motor temperature.
- b. Sufficient lift oil pressure, Upper and Lower oil reservoir levels, and seal injection water flow are four of the starting in'terlocks.
- c. Power must b less than 20*. and Tave greater than 500*F prior to starting the fourth RCP.
'd. The pump is designed for continued operation on either loss of. cooling water or loss of injection fluid, but not both.
2.15 Which of the following statements concerning the Control Rod Drive System is INCORRECT?
- b. Four ball check valves are installed at the base of the thermal barrier to permit in-flow to the CRD mechanism during a reactor trip.
- c. The APSRs are prevented from tripping by physical- restraints on the segment arms; this prevents the arms from pivoting outward.
- d. The stator coils are sequentially energized in a repetitive 2-3-2-3 manner. When rod motion ceases, three coils remain energized.
2.16 Select the combination of breakers and programmers that will NOT result in a Trip Confirm Lamp - Reactor-Trip.
, . 4 6
. Select the CORRECT statement regarding the automatic initiation of the Motor Driven Emergency Feedwater Pump, EFP-1.
jd
'a. In the event of a station blackout, EFP-1 will automatically start five seconds after ES block 1 is loaded onto the 'A' Train D-G.
- c. Thirty minutes after the alarm is received for exceeding the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating.of 3000 kw, EFP-1 will be automatically tripped.
- d. EFP-1 will not automatically start if EDG-3A is running and closed on its respective bus.
2.18 Select the CORRECT statement regarding the 480V to 120V AC distribution -
system.
2.19_With regard to the Main Steam Rupture Matrix, select the INC0" RECT statement.
- a. If the Actuation Test Switch is in " Test M.S. Normal" , hen a steam rupture c curs, the Main Steam Isolation Valves will not : lose on the affected OTSG.
- b. Both*8he 725 psig switches and the 600 psig switches must operate to cause the matrix to actuate.
- c. The emergency block valves (FWV-34 and 35) and the emergency feedwater reg. valves (FWV-161 and 162) are included as valves that auto close when the matrix actuates.
- d. During plant cooldown, at 725 psig the rupture matrix can be bypassed by depressing all four actuation PB's.
7 e-2.20 Which one of the following is NOT an indicated condition of annunciator C-3-12, "NS CCC SYST RB LEAK!'.
- a. Differential flow to AH Units.
- b. Differential flow to letdown coolers,
- c. Differential flow to RC Drain Tank Cooler *
- d. Differential flow to RCP's 2.21 Many important pumps have annunciators which indicate when the pump is out-of-service, for example: E5 Annunciator D-3-3 is labeled "0H Pump 'B' OUT OF SERV". Which one of the following is an indicated condition for this type of annunciator,
- a. No breaker DC control power -
- b. Breaker control switch in normal after start, breaker open, breaker racked in.
- c. Overload relay actuated.
- d. Excessive motor amps. ,
2.22 The secondary cycle system is sampled for pH, Hydrazine, conductivity, oxygen, sodium and silica. Which of the following will generate a computer alarm and lead you to initiate an Abnormal procedure for Secondary Chemistry Control?
- a. pH
- b. Conductivity
- c. Oxygen
- d. Sodium 2.23 Select the INCORRECT statement regarding the Condensate Injection System.
- a. Condensate injection is used for main turbine head sprays in the high pressure turbine, pump seals in the feedwater system, and valve steam sealing to prevent in leakage of air to the condensers.
- b. When condensate pressure is above 220 psig the condensate pumps are supplying seal and spray water.
- c. If the discharge pressure dec'reases to ~200 psig, the G.W.P. that has been selected will start automatically.
- d. The condensate injection system supplies water to the desuperheaters in the auxiliary steam system, steam supplies to both evaporators, and the gland steam system.
7 9 + h 8
2.24 Which of the following chemicals is NOT used in the Makeup and Purification System?
- a. Hydrazine
- b. Lithium Hydroxide
- c. Sodium Thiosulfate *
- d. Sodium Hydroxide 2.25 Which of the following statements is INCORRECT regarding the Emergency Core Cooling-System?
- a. The purpose of the 600 psig CF Tank pressure is to mitigate the sudden Temperature increase following the " blow out" phenomenon.
" Dead-heading" by a recirculation line back to the BWST.
END OF SECTION 2
w- .
3.0 INSTRUMENTS AND CONTROLS 3.01 Which one of the following l'oad limiting conditions and corresponding load limit is CORRECT 7
. a. Loss of 1 RC pump with 4 running - 30%/ min to maximum limit of 75% *
- b. Loss of 2 RC Pumps with 4 running - 30%/ min to maximum limit of 45%
- c. Loss of feedwater booster pump - 50%/ min to maximum limit of 55%
- d. Asymmetric Rod - 30%/ min to maximum limit of 60%. .
3.02 If the Diamond or Reactor Demand Stations are in HAND, the feedwater system will accept responsibility for control of Tave only-if certain conditions are met. Of the following conditions that will prevent feedwater from controlling Tave, which one is stated CORRECTLY?
- a. Either steam generators high level limited '
- b. Either steam generator low level limited
- c. Either steam generators BTU limited
- d. Either loop A or B hand / auto stations in manual.
3.03 Which one of the following statements concerning the Control Rod Drive Position Indication System is TRUE?
- a. The 0% switch is located 1.5 inches below the in-limit switch,
- b. The 100% switch is located 1.5 inches below the out-limit switch.
- c. The first rod in any group to reach the 100% switch will.stop further travel of all rods in that group.
- d. On group 7, the out-limit is at 91.4% withdrawn but this can be bypassed with a key switch in the control room. l l
- =.
2 s -
3.D4 The following statements concern the Control Rod Drive System.
Select the CORRECT statement.
- a. If the speed selector s' witch is left in J0G with the automatic control mode selected, the control rods will still move at RUN speed.
- b. For sequence operation of groups 5 through 7, the group
- select switch may be in any position, including 8, during manual operation in the sequence mode.
- c. Once automatic.is selected; a trip, programmer lamp fault, or sequence inhibit condition will revert control to manual.
- d. The Safety Rods Out Bypass Light will come on if a rod in one of the Safety Groups drops. -
3.05 In the turbine bypass valve controller, +50 or +125 psi biases are sometimes applied. Select the CORRECT statement with regard to these biases.
- a. Before the turbine is synchronized, a +50 psi bias is applied when all bypass valves are closed and real header pressure e is less than 10 psi.
- b. If the real header pressure error is 10 psi or greater, then the U.L.D. must be greater than 10% to have the +50 psi bias applied.
- c. A + 125 psi bias is applied to the turbine bypass. valves whenever the Turbine is Tripped.
- d. The +125 psi bias is removed by pressing TRIP RESET on the Diamond panel.
3.06 Select the CORRECT statement with regard to speed control (Governor) of the Emergency Diesel Generators.
- a. As a general rule, D-G units running alone should have the SPEED DROOP control set on 0 (zero).
- b. The synchronizer motor, mounted on top of the governor, allows the operator to match the voltage of the D-G with running voltage before synchronizing to the system.
- c. The LOAD LIMIT control may be used for shutting down the diesel by turning the LOAD LIMIT control to zero.
- d. The SYNC INDICATOR, located directly below the SYNCHRONIZER control indicates if the D-G is in phase with the system.
4 4
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3.07 Which one of the following statements about temperature detectors is TRUE? .
- a. The thermocouple is connected to one leg of a bridge circuit and as the temperature changes the output voltage across the bridge changes.
- b. When a thermocouple fails open it will respond in the saMe manner as an RTD and will indicate a full scale reading on the meter.
- c. When a thermocouple becomes shcrted, a new thermocouple will exist at the point of the short and the meter will respond to the ambient temperature at the point of the short.
to the temperature difference between the open ends of the wires.
3.08 Which one of the following statements about the ex-core Nuclear Instrumentation is TRUE?
- a. The source range signals originate in BF3 proportional ,
counters and are amplified by pre-amps located in the reactor building.
- b. The Intermediate Range detectors are compensated ion chambers. The detectors consist of two chambers; one is boron lined and is sensitive to neutrons, while the other is unlined and is sensitive to neutrons and gammas.
- c. As the compensating voltage on the IR detectors is increased, the overlap with the source range and power range channels is decreased.
- d. Because of gas multiplication in the BF3 detectors, gamma produced pulses are bigger than the neutron produced pulses and are therefore easy to discriminate out in the circuitry.
4
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3.09 Which of the following is TRUE concerning the source range channel high voltage cutoff?
- a. Either IR channel at 10 ' amps will turn off the high voltage.
- b. If one IR channel fails low while at power, the source range high voltage will be re-energized. *
- c. Power range NI-5 or NI-6 and NI-7 or NI-8 will turn off high voltage at 10% power.
- d. The high voltage is turned on from the intermediate range when the power level decreases to 10 amps.
3.10 When the RPS is in Shutdown Bypass, which one of the following is TRUE?
- a. A high pressure trip of 1720 psig is administratively imposed and an overpower trip of 5% automatically imposed.
- b. The high pressure trip at 2355 psig is bypassed.
- c. The four trips bypassed are high temperature, low pressure, -
variable low pressure and flux / delta flux / flow.
- d. The RCP Power Monitor trip is bypassed.
3.11 With regard to overspeed protection on the main turbine, select the one CORRECT statement.
- a. There is a mechanical overspeed trip at 103% and a backup electrical overspeed trip at 111%.
- b. With the Overspeed Protection Control (0.P.C.) switch in the " Test" position, the electrical overspeed trip is bypassed.
- c. At approximately 103% shaft speed only the governor and interceptor valves will close, while at 111% speed, all four sets of valves will close.
- d. In the "Overspeed Test" position on the 0.P.C., only the Reheat and interceptor valves close.
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3.12 Which one of the following statements concerning the main turbine EHC Indication Panel is INCORRECT?
- a. Speed control: .This light will tue lit before the unit is latched and the generator output breakers are open.
- b. Load control: -This light will be lit when one of the generator output breakers is closed. *
- c. Speed channel: This light will be lit if there is a speed differential between the main speed channel and the auxiliary speed indicator,
- d. Emergency Power Supply: This light will normally be lit on a startup due to the loss of the Permanent Magnet Generator (PMG).
3.13 Cross-Tie Blocking Interlocks are provided to prevent paralleling of both D-G and to prevent paralleling of both 4160v ES buses. Refer to Figure 10.15 and select the CORRECT statement.
- a. If breakers 3209, 3210 and 3205 are all closed, the amber lamp (Block Closing Actuated 3206) will be lit, thus ,
permitting breaker 3206 to be closed.
- b. If breakers 3209 and (1) 3205 and 3206, or (ii) 3207 and 3208, or (iii) 3211 and 3212 are closed, the amber lamp (DG Parallel Block Act) will be lit and breaker 3210 can not be closed.
- c. If the amber lamp (Block Closing Actuated 3208) is lit, it means breaker 3208 can not be closed because the 3B bus is already being fed from the 3A bus (through 3207) and no Diesels are running,
- d. If both Diesels are feeding their respective buses (3209 closed and 3210 closed) all Block Closing Actuated Lamps will be lit.
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3.14 The HPI actuation sequence is divided into four blocks for loading the various equipment on its electrical buses. Which one of the following examples of equipment loaded in each block is CORRECT.
Block I Block 2 Block 3 Block 4
- a. Make up Pump AHF-Run Decay heat RB Spray in slow speed pump pump
- b. Decay heat Makeup Pump NS Closed DH Closed pump Cycle Pump Cycle Pump
- c. Makeup Pump Emerg. NS Decay Heat RB Spray Seawater Pump Seawater Pump Pump
~
- d. AHF-Run in Decay heat Makeup Pump DH Closed Slow speed Pump Cycle Pump -
3.15 When conducting a plant shutdown, several operations are required to prevent inadvertent ES actuation. Which of the following statements is TRUE during a plant cooldown?
- a. When RC pressure is reduced to 1800 psi, the HPI white bypass permit lights will come on. .
- b. If HPI was properly bypassed, the 1500 psi bistable tripped lights will not come on when pressure is reduced below this value.
4
- d. When each channel was bypassed, its respective amber channel bypassed light would have come on, and the green channel function enabled lights and the green bypass / reset lights would have gone out.
3.16 Alarms from various area radiation moritors result in automatic actions that must be immediately verified by the operator. Which
'one of the following is INCORRECT?
- a. RM-A1; Ensure closea AHV-1A, IB, 1C, ID
- b. RM-A2; Ensure stopped AHF-6A, 6B, 7A, 78
- c. RM-A3; Ensure stopped AHF-11A, llB
'd. RM-A4; Ensure AHF-10 stopped
. d
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3.17 Select the INCORRECT statement concerning the Pressurizer Heater controls.
- a. Heater bank A, B and C"use modulating control (SCRs) while banks D and E are strictly on/off control,
- b. If pressurizer level decreases to less than 30 inches, all heater banks will be de-energized. *
- c. Bank C has four groups of heaters which are sequenced on to prevent two groups in the same bank from coming on simultaneously.
- d. Banks A and B contain only one group of heaters, have no staggered turn on, and are both fully on at 2135 psig.
3.18 Which one of the following is consistent with having the out- -
inhibit lamp lit on the Diamond Rod Control Panel?
- a. Power >60% and a 9-inch asymmetric fault exists while in auto
- b. ICS auto power is not available
- c. One or more programmers has out motion with an in programmers command.
- d. One group 7 rod is at the out limit.
Which one of the following statements concerning the main feedwater valve interlocks is TRUE?
i
- a. ICS provides an auto open signal to the lo load valve whan the startup valve is 50% open, and an auto close signal when-the startup valve is 80% closed.
- c. If FWV-28 is closed, and Main Block closed, @ sid is maintained across flow control valves by associated loop AP.
- d. Upon actuation, the main steam rupturs matrix cloces all main feedwater valves and prevents reopening these valves in auto mode but not in manual mode.
i 8
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s 3.20 To start any circulating water pump the four start permissives must be satisfied. Which of the following is NOT one of the permissives? ,
- a. Condenser vacuum of at least 9.5" Hg established
- b. Lube water flow is normal (>19gpm to upper bearing)
- c. Water box has been primed to 2115 feet
- d. Pump trip permit: thirty seconds have elapsed since pump last ran.
3.21 Which one of the following plant fire protection systems is NOT automatically actuated by temperature detectors?
- a. Deluge water spray + -
- b. Wet pipe sprinkler
- c. Carbon dioxide flooding
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3.23 Which one of the following is TRUE concerning the " Air Fail Reset" '
pushbuttons for MUV-16, 31 and 51.
~
- a. The pushbutton only indicates loss of air to the associated valve E/P controller.
- b. The pushbutton indicates loss of air to E/P controllers.for MUV-16 and 51 and also loss of air to the valve positione'r for MVV-31.
- c. On loss of air supply, the solenoid valve supplying air to the air lock valve will de-energize, causing the affected valve (16, 31 or 51) to close.
- d. When air pressure has increased, depressing the air fail reset pushbutton will unlock MUV-16, 31 or 51.
3.24 Select the CORRECT statement concerning the Nuclear Services Cooling Water System (Seawater and Closed Cycle Cooling)?
- a. The normal duty seawater pump, RWP-1, is backed up by emergency pumps RWP-2A and 28, which start sequentially on low header pressure.
- b. An ES signal will start both emergency pumps, RWP-2A and 28, and will trip RWP-1 after a 15 second time delay.
- d. If any one of the three closed cycle pumps are normally running, an ES signal will automatically start the other.two.
3.25 Which of the following is TRUE concerning the OTSG level instruments?
- a. The startup range (0-250") and the Operate Range (0-100%)
share the same upper and lower level instrument taps
- b. If a startup level transmitter fails low while at power, there will be no noticeable effect on the ICS (all subsystems in auto)
- c. Interlocks from the Operate Range to the ICS include 50%
level on loss of 4 RCPs and Hi level limit.
- d. The startup range has a low level input to the ICS and is temperature compensated.
END OF SECTION 3
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4.0 Procedures - Normal, Abnormal, Emergency and Radiological Control 4.01 Which one of the following 1.imits and precautions of OP-504 " Integrated Control System", is INCORRRECT.
- a. If a feedwater cross-limit occurs while controlling the reactor from either the reactor demand control station or the diamond, reduce reactor power to be
- compatible with total feedwater flow.
- b. If a BTU limit occurs while in manual control of feedwater demand, place the diamond in manual and reduce reactor power until the BTU limiting condi tion just clears and investigate the cause.
- c. Prior to placing the diamond control station in " Auto",
verify that the power range channel is reading at least
- 10% power.
- d. Steam generator load ratio, ATc, cannot be placed in
" Auto" unless either feedwater demand station "A" or "B" is in " Auto".
4.02 Which of the following is NOT correct conccrning the '
condition or probable cause of the respective area radia-tion alarms?
- a. RM-A1; High activity in purge duct exhaust or possible RC leak
- b. RM-A2; Possible makeup / letdown leak or possible steam generator tube rupture
- c. RM-A3; Possible waste gas tank or piping leak
- d. RM-A4; High activity in control complex ventilation return air.
4.03 Which of the following conditions requires the affected Reactor Coolant Pump to be shutdown immediately?
- a. High seal stage pressure drop (less than two-thirds RCS pressure)
- b. High controlled bleed off temperature of 2170 F
- c. Standpipe high level alarm
- d. Total seal outflow exceeds 2.5 gpm and is gradually increasing.
. 2 O. .
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4.04 Which one of the following is a Remedial Action instead of an Immediate Action in AP-580, " Reactor Protective System Actuation"? ,
- s. Ensure main block valves closed
- b. Ensure low load block valves closed
- c. Trip both MFPs.
- d. Close MOV-51, Letdown Block Orifice Bypass 4.05 Emergency Reactivity Control Procedure, EP-140 should be used for which one of the following conditions?
- a. As a remedial action to any of the runback procedures (immediate action is to ensure control rods inserting.) .
- b. If there is excessive positive reactivity in the core or boration from BWST is ineffective.
- c. As a remedial action to AP-580, "RPS Actuation" (immediate action is to ensure GRP 1-7 rods inserted).
- d. Prior to evacuating the control room while conducting the immediate actions of AP-990," Shutdown From Outside Control Room."
4.06 In performing the follow-up actions of EP-290, " Inadequate Core Cooling", the RC pressure has reached 2300 psig and the PORV has been manually opened by the operator. When should the PORV be closed?
- a. When RC pressure falls to 90-100 psig above OTSG pressure.
- b. When subcooling decreases to less than 20 F.
- c. Wnen indicated pressurizer level goes off-scale high.
- d. It should not be reclosed as long as full HPI is injecting.
4.07 Which one of the following reactivity control parameters is NOT part of the verifications performed during VP-540, " Runback Verificaticr Procedure."
- a. Imbalance I i
- b. Quadrant Power Tilt
- c. Heat Balance
- d. Control rod index l
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4.08 Which.one of the following is an INCORRECT method of verifying natural circulation as specified in VP-580,
" Plant Safety Verification Procedure?"
- a. Tc > Tsat of OTSG.
- b. AT develops and stabilizes.
- c. Avg of 5 highest incore thermocouples follow T h within 10*F.
- d. When OTSG pressure is lowered, then Th, Tc and incore thermocouples lower.
4.09 Which one of the following is an INCORRECT immediate action of AP-450, " Emergency Feedwater Actuation?" Ensure closed: ,
.a. MBVs
- b. MSIVs
- c. LLBVs
- d. SUBVs '
4.10 OP-402, " Makeup and Purification System, says, " Loss of flow through makeup pump will destroy the pump within approximately .
- a. 15 seconds
- b. I minute
- c. 3 minutes
- d. 10 minutes 4.11 Which of the following is a CORRECT definition as contained in the Radiation Protection Manual, RP-101?
- a. " Clean Area" - Removable contamination < 30 dpm beta gamma or 300 dpm alpha, as determined by smear surveys representing approximately 100 cm2 of surface area.
- b. " Radiation Area" - Any accessible area where a major portion of the body could exceed a dose rate of 5 mrem /hr or in any seven consecutive days a dose in excess of 100 mrem.
- 4 l
- c. "High Radiation Area" - Any accessible area where a major portion of the body could receive a dose rate in excess of 100 mrem /hr o.r, in any seven consecutive days a dose in excess of 1000 mrem.
C. " Clean Area" - Fixed surface contamination measuring
< 0.25 mrem /hr beta gamma and < 300 dpm alpha as measured by appropriate survey instrument probes.
4.12 Select the CORRECT statement from the following:
a '. Visitors at Crystal River may not receive a dose in excess of 1.25 Rem /Qtr. even if an NRC Form 4 is completed and furnished.
exposure history. '
- c. The Nuclear Plant Manaoer has the authority to allow individuals to receive up to 3000 mrem in one week for special work assignments.
- d. The maximum weekly exposure is 300 mrem. The Shift ,
Supervisor may authorize exposures 600 mrem by use of Form 912801.
4.13 Which one of the;following Crystal River emergency procedures would list these entry " symptoms" and " condition":
Symptoms Condition
- 1. Neutron Flux raising Excessive reactivity exist in core without automatic protection features
- 2. RC pressure raising'.
or Pzr level raising
- a. EP-120, " Inadequate Shutdown Value"
- b. EP-140, " Emergency Reactivity Control"
- c. EP-220, " Pressurized Thermal Shock"
- d. EP-260, " Inadequate Decay Heat Removal" i l
6
- 5.
4.14 Which one of the following procedures has no Immediate Action?
- a. EP-120, " Inadequate Shutdown Value"
- b. EP-140, " Emergency Reactivity Control"
- c. EP-220, " Pressurized Thermal Shock"
- d. EP-260, " Inadequate Decay Heat Removal" 4.15 Crystal River emergency procedure Ep-220, " Pressurized Thermal Shock" lists as a remedial action, " Reduce Subcooling to minimum;" Which is one of the following RC pressure and subcooling margins are appropriate for this remedial action?
- a. s 500 psig - 20 F
- b. s 500 psig - 50 F
- c. 2 1500 psig - 20 F ci . 2 1500 psig - 50 F 4.16 Which one of the following is-true of emergency procedure ,
EP-390, " Steam Generator Tube Rupture"?
- a. Entry Symptom: Subcooling Margin Monitor < 0 F, Incore thermocouples selected.
- b. Entry Symptom: OTSGs are at 95% or greater on the operating range and rising.
- c. Immediate action: Start load reduction at 5%/ min.
- d. Immediate action: Lower and maintain.0TSG Tsat.
L 90-110 F less than Tsat. for the existing RC pressure.
4.17 According to 10 CFR 20, which one of the following is equal to one rem?
- a. A dose of I rad due to X, gamma or beta radiation.
I
- b. A dose of I rad due to X or gamma radiation, or 0.1 rad due to beta radiation. j
- c. A dose of 0.1 rad due to alpha particles or fast neutrons.
- d. A dose of 0.I' rad due to alpha particles or 0.2 rad due to beta radiation or 0.33 rad due to thermal neutrons.
L
. 4
. 6 4.18 There are two classifications of operating procedures (according to AI-500), step-by-step and guidance. Identify the one procedure below that is,to be followed step-by-step.
-a. OP-502, " Transfer Group from DC Hold to Auxiliary Bus."
- b. OP-503, " Initialization of Plant Computer"
- d. OP-605, " Emergency Feed Pump Operation" 4.19 Which one of the following conditions requires the operator to manually trip the reactor?
- a. Loss of RCP (3 running)
- c. Pzt. level of 295 inches
- d. Channel A Th of 602 F 4.20 Which one of the following' conditions requires the operator ,
to manually trip the reactor?
- b. Loss of 1 main feed pump
- c. Loss of Pzr heaters
- d. Condenser vacuum of 26" Hg.
4.21 Which one of the following conditions requires the operator to manually trip the reactor?
- a. Dropped control rod
- b. -BTU limit (ICS)
- c. Two MSIVs close
- d. 0.25 gpm OTSG tube leak
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4.22 Select the one statement below which is correct for t' plant at 90% power, steady state operation.
- a. a control rod suspected of being stuck must-be moved (attempted) in " jog" speed.
- b. at the conclusion of rod exercising with no auto inhibit conditions, an out inhibit indication means a safety control rod not fully withdrawn.
- c. Rod exercising should be performed only on down power transients, not on up powerJtransients.
- d. Motor tube extension operating temperature must not exceed 450 F and stator temperature must not exceed 180*F.
4.23 Which one of the following statements about the deaerator tank level control is correct?
- a. High level in the deaerator tank (~14.5 f t.) will start the second CD pump.-
- b. High level interlock (~14 ft.) will open HDV-83 and dump 7500 ,
gallons to the condensate storage tank.
t
.c. High level interlock (~14 ft.) closes HDV-50 and 54, drains for intermediate pressure feedwater heaters.
- d. The deaertor tank normal operaing level is between 7 feet and ,
14.0 feet.
4.24 Which of the following would be proper action to be taken for a plant-upset.with feedwater controllers in manual?
- a. Place LTc controller in Auto.
ensure adequate heat removal capabilities. l
- c. Feed flow should be adjusted according to RCS pressure. For )
RCS pressure less than 2155 decrease feedflow; for RCS pressure o greater than 2155 increase feedflow.
4.25 Cooling water flow to the CRDs must be:
- a. at least 130 GpM if RCS temperature is above 180 F
- b. 10*F above the dew point at the reactor vessel head.
- c. at least 125"F.
- d. a maximum of 180*F END OF EXAM
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- 7.0 1 % 85 0 OlW9 53.H 144.% 992.5 1136 9 0.2581 1.5586 1.8167 80 182.86 0.01653 47.34 150.79 988.5 1139.3 0.2674 1.5383 1.8057 90 186.26 0 016 % 42 40 1%.22 985.2 1841.4 0.2759 1.5203 1.7962 10 193.21 0.01659 - 3842 161.17 982I 1843.3 0.2835 1.5041 8.7876 14.656 212.00 0.01672 26.80 180.01 970.3 l 150.4 0.31 2 IA446 I.7%6 15 213 03 0.01672 26.29 Ikl.it 9691 1830.8 0.3135 lA415 1.7549 20 227.96 0.01683 20.089 196.16 960.1 11 % .3 0.33 % 1.3962 1.7319 25 240 01 0 01692 16.30) 20h 42 952.1 1860.6 0.3533 1.3006 I.7139 '
30 250.33 0.01701 13.146 218.82 945.3 IIM.I 0.3600 I.3313 IA993 35 259.28 0.0ll'Oli 11.898 227.91 939.2 1867.1 0.3807 f.3063 IA870 40 267.25 0.01715 10.498 2%.03 933.7 1869.7 0.3919 1.2H4 IA%3 '
45 274.44 0.01721 9.401 243.36 928.6 1872.0 0 4019 I.3650 IA669 50 281.01 0 21127 8.515 230.09 924.0 1874.1 0 Allo 1.2414 14585-55 28727 0.01732 7.787 2 %.30 919.6 1175.9 0.4193 1.2316 14509 60 29211 0.01738 7.175 262 09 915.5 1177.6 0.4270 1.2168 14436 65 297.97 001743 6.655 267.50 911.6 1879.1 0.4342 1.2032 14374 10 302 92 0.01748 6.26 272.61 901.9 1880.6 0 4409 1.1906 14315 75 307A0 0 01753 5.816 277.43 904.5 l181.9 0.4472 1.1787 14259 30 312.03 0.01757 5.472 282.02 901.1 1883.1 0.4531 1.16% 14307 85 31625 0.01 % I 5.168 286.39 897.8 1184.2 0 4587 1.1571 14158 90 320.27 0.0146 4.896 290.56 894.7 1185.3 0.4641 1.1471 14112 95 324.12 0.01770 4.652 294.56 891.7 1886.2 0.4692 1.13 % IA069 100 327.81 0.01774 4.432 298.40 888.8 1887.2 0 4740 1.1286 IA026 110 334 77 0.01782 4.049 305.66 883.2 1188.9 0 4832 1.1111 1.5945 i
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TABLE D.lb Properties of Dri Saiurated Sicam termtmurde lemperaturc u, Erecer sotwm k mhet kwer*
7emp
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liquid verce liq uid II
- spor ligvid Esar ..por
- r 'o *, h, 4 6, s, s. s, 350 IM 63 0 01799 3.M2 321.63 870 7 1192.3 0.5029 1.0154 1.5783 360 153 04 0 01881 2 951 332.18 852.2 l194.4 0.5158 1.0519 1.5677 370 173.37 0 0182) 2.625 342 79 853.5 1196.3 0.5286 1.0261 1.5573 380 195.17 0 018 % 2.335 353 45 644 6 119s 1 0.5413 1.0059 l.5471 390 22Q.37 0 01850 2 0636 M4.17 8354 1899.6 0.5539 0 9832 1.5371 400 247.31 0.01864 1.8633 374 97 826 0 1201.0 0.5664 0 9608 1.5272 410 276.15 0.01878 8.6700 385 8) 816.3 1202I 0.5788 0.9386 1.5174 * "
420 308 83 0 01894 1.500r' 396 77 806 3 1203.1 0.5912 0.9l% 1.5078 -
430 343 72 0.01910 1.34L 407.79 796.0 1203.8 0 6035 0.8947 1.4982 44 381.59 0.01926 I.2171 419 90 785 4 1204.3 0.6154 0 8730 1.4887 450 422 6 0.0194 l.0993 430.1 174.5 1204 6 0.6280 0.8513 1.4793 460 466 9 0 01 % 0 9944 441.4 763.2 8204 6 0.6402 08298 1.4700 470 $14 7 0 0198 0.9009 452.8 151.5 1204 3 0.6523 0 8053 1.4606 460 566 1 0 0200 0 8172 464 4 739 4 12037 0.M45 0 7668 1.4513 490 6214 0.0202 0.1423 476 0 726.8 1202.8 0 67M 0.7653 1.4419 ,
500 6806 0.0204 0.6149 487.8 713.9 1201.7 0 6887 0.1434 1.4325 520 812.4 0.0209 0.5594 511.9 686 4 1898.2 0 1130 0.1006 1.4136 540 962.5 0.0215 0.4649 5%.6 6% 6 1193.2 0.7374 0.6%8 1.3942 560 l 133.1 0 0221 0.3M8 562.2 624.2 1886 4 0.1621 0.6121 1.3742 ,
580 1325.8 0.0228 0.3217 588.9 588 4 1I77.3 0.7872 0.5659 f.3532 600 1542.9 0.02 % 0.2668 610.0 548.5 1165.5 0.8131 0.5176 1.3307 620 1786 6 0.0247 0.2201 6467 503.6 1150.3 0.8396 0.4664 1.3062 643 20557 0 0260 0.1798 676.6 452.0 1130.5 0.8679 0 4110 1.2769 MO 2M54 0 0274 0.1442 114.2 390.2 1104 4 0.8981 0.3485 1.2472 680 2706 1 0.0305 0.1115 757.3 309.9 1061.2 0.9351 0.2119 1.2071 700 3093 7 0OM9 0.0761 823.3 172.1 995 4 0.9905 0.1484 1.1389 705 4 3206.2 0 0503 0 0503 902.7 0 902.7 1.0540 0 1.0580
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1 ABLL D.lb Properties ot Dr3 5atursted $leam tamiinects Temperatute !
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350.4 67.97 1037.2 1805.2 0.1295 100 0.9492 0.01613 -
1.9577 0.01617 265 4 77.94 1031.6 1109.5 0.1417 I.8106
- l10 1.2748 1025.8 1113.7 0.lH5 1. % 94 1.93M 14924 0.01620 203.27 87.92 120 0.01625 151.34 97.90 1020 0 till.9 0.1816 1.7296 1.9112 2.2223 1.8894 130 123.01 107.89 1014.1 1822.0 0.1984 14910 140 2.8886 0.01629 l.6537 1.8685 97.01 117.89 10D8.2 1126.l 0.2149 ISO 3.718 OD1634 1A174 1.8485 4.741 0.01639 77.29 127.89 10D2.3 1830.2 0.2311 160 996.3 1134.2 0.2412 l.5822 1.8293 ODlHS 62.06 137.90 5.992 1.5480 1.8109 110 7.510 0 44451 30.23 147.92 990.2 llM.I 0.2630 1.1932 llo ODl657 40.96 157.95 9e4.1 1842.0 0.2785 1.5141 190 9.3M 977.9 1145.9 0.29M l.4824 1.7M2 200 Il.526 OD1663 3344 167.99 1.7598 OD1610 27.82 178.05 973 6 1149.7 0.3000 1.4508 210 14.123 97J 1 1 1850.4 0.3120 1.4446 1.7546 ODI672 26.80 180D7 14 696 1.7440 212 -
0.01677 23.15 188.13 965.2 1153 4 0.3239 1.4201 220 17.186 958.8 1857.0 0.3M1 1.M01 l.7288 20.780 0 01684 19.382 198.23 1.1140 2 30 0D1692 16.323 208.34 952.2 1860.$ 0.3531 l.3609 240 24 969 1.3323 1.4998 0 01700 13.821 .216.48 945.5 1164 0 0 3675 250 29.825 1167.3 0.Mll 1.3043 14860 260 35 429 0.01709 11.%3 228 44 9 38.7 1A727 270 41.'858 0.01781 10.061 238 84 931.8 1170 4 0.3958 1.2%9 IA597 280 49.203 ODI726 8 445 249.06 924.7 1873.8 0.4096 1.2501 14472 57.556 0D1735 7.461 259.31 917.5 11%.8 0 4234 1.2238 290 300 67.013 0.01145 6 466 269.59 910 1 1179.7 0 4369 1.1980 1.6350 310 7748 0.01755 5426 279.92 902A 1882.5 0 4504 1.1727 14231 0 01765 4 914 290.28 894.9 1885.2 0.4637 1.1478 14115 320 89 66 l 330 103.06 0.01776 4.307 300 48 887.0 1187 1 0 4769 1.1233 1.6002 340 118.01 0.01787 3.788 311.13 879.0 1190.1 04900 1D992 1.5891 1
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' CRYSTAL' RIVER: CATEGORY 1 ANSWERS l'.'01 (b)
Ref: STM 2-41 (fig.14) 1.02: (c)
Ref: Duke Power Company, FNRE 1.03' -(d) _
Ref: NUS, NETRO 1.04 (c) ,
Ref: NUS, NETRO
~
1.05 (c) M )
Ref: Duke Power Company, FNRE, Pg. 237 1.06 (d)
Ref: Duke Power Company, FNRE ,
1.07 (c)
Ref: North Anna FSAR, Table 15.1-5 1.08 (a)
Ref: Nuclear Reactor Analysis, Dunderstadt & Hamilton, Pg.13 1.09 (a) 1 Ref: Duke Power Co., FNRE 1.10 (c)
Ref: Steam Tables or Mollier Diagram 1.11 (d)
Ref: North Anna, Thermo lesson Plans, pg. 4 1.12 -(a)
Ref: Steam Tables Tsat @l000 psia = 544.6 - 400 * = 144.6 1.13 (b)
Ref: Oconee question bank 1.14 (a).
Ref: Oconee question bank - Q #150 1.15 (d)
Ref: Duke Power Co. Thermo, HT & FF, pg 161 Also; Cr 3, Question Bank #4
t 1
~
l 10 i i
l 1.16 (d)
Ref:'CR3; Categ 1, Question 9 1.17 (c)
Ref: CR3, Plant Curve book, curve 3.2 A & B (OP-103) 1.18 (b)
Ref: Doppler. Coefficient, M. G. Woram Pg. 7 1.19 (b)
Ref: Paraphrase CR3 Question Categ 1. No. 13 1.20 (c)
Ref: Duke Power Company, FNRE 1.21 (d)
Ref: CR3 Question, Categ. 1, No. 54. (Paraphrase) 1.22 (a)
Ref: CR3 Curve Book, Curves 4.1 to 4.4 c 1.23 (b) '
Ref: CR3 Curve Book 1.24 (b)
Ref: T.S. Figure 2.1-1
-(b)
<g . t f- T.S. E 3/41-3 Y
CATEGORY 2: ' Answers 2.01' _(d)
'Ref: STM 38-3/5 2.02 (c)
Ref: STM 17-4/5 2.03 (a)
Ref: STM 17-11/12 2.04 (c) .
Ref: STM 22-13/40 2.05 -(b)
Ref: STM 23-7 and OP-502, pg 3 2.06 (d)4(b /
Ref: STM 5-1/4 2.07 -(c) +- (b) /
Ref: STM 10-47
!. 2.08 (a)
Ref: STM 28-5
~$ r_
?
10 i .2.09 (d)F(A\
Ref: STM 36-4 2.10 (c)
-Ref: STM-4 and OP-401, pg 2/10 2.11- (b)
Ref: OP-404, pg 4 & 5 L
'2.12 (a)
Ref: OP-404, pg 32 i
2.13 .(d) -
Ref: STM 2-7/54 2,14 (c)
Ref: STM 2-66/106 2.15 (d) ,
Ref: STM 12-4/7 2,16 (c).
Ref: STM 12-13
_' 17 ~ -- (& kk Ref: Attached description and OP-605, pg 5
(
= ,
11 2.18 -(d) ,
Ref:. Dwg EC-206-017 and.0P-703, pg 9 2.19 .(a)..(b +d Ref: STM : 27-63/68 Nov %W 2.20' (c)
Ref: AP-303, ESD Annunciator Response
~ 2.21 (a)
Ref: AP-304, ESD Annunciator Respon.e 2.22 '
(b)
Ref: AP-1071 and STM-21/25 2.23 (a) l Ref: STM 25-13/14 l-l 2.24 (c).
Ref: STM 18-1 2.25 (c)
Ref: STM 4-6/12 t
b.
7 10 3.0 -INSTRUMENTS AND CONTROLS - ANSWERS 3.01 (c)
REF: STM-l'3-21
-3.02.(c)
REF: STM-13-34 3.03 (b)
REF: STM-12-11/12 3.04 (a)
REF: STM-12-18/20 3.05 (d) -
REF: STM 13-15/16 3.06 (a)
REF: STM 10-36/37
' 3.07 (c)
REF: ~STM 7-2/27 3.08(a)
REF: STM 6-3/17 3.09 (c)
REF: STM 6-11 3.10(d)
REF: STM 9-11/21 also T.S. pg 2-6 3.11(c)
.REF: STM-28-6/8 3.12 (a)-
, REF: STM-28-14 3.13 (b)
REF: STM-10-56/57 3.14 (c) -
REF: STM 11-10 3.15 (d)
REF: STM 11-24 3,16 (b)
REF: AP-241/242/243/244
\
+
11
-3.17 (b)
REF: STM 2-121/122 3.18(a)
REF: B&W CRD description pg 43 11^ 'd r.:r. ;T^;
r k
27 ;3/60 b
3.20 (a) .
I REF: STM 35-7 3.21 (d)
REF: STM 38-3/5-3.22 (d)
REF: STM 17-14 3.23 (d)
REF: STM 17-17/18 3.24 (b)
REF: STM 23-2/4 '
-3.25 (c)
REF: STM 24-7 I
L'
i
- s
- e-8 r-4.0 Procedures .
4.01(b)
REF: OP -504
'+
.4.02 (d)
- REF: AP-241/242/243/245 I
4.03(b)'
REF: OP-302, pg. 4&5
- 4.04 (c)
REF: AP-580 4.05(b)
'REF: EP-140 4.06.(a) .
REF: EP-290, pg. 4.
4.07.(c)
REF: VP-540, pg. 3
- 4.08;(a)
REF: VP-580 4.09(b)
' REF: AP-450, pg. 2 4.10(a)
REF: OP-402, pg. 4 4.11(d)
<, REF: RP-101, pg. 5&6 4.12 (b)
REF: RP-101, and IN-84-40 4.13 (a)
, REF: EP-120 " Inadequate Shutdown Value" pg. 1 of 5 1
9
- 4.14(d)
REF: EP-260 " Inadequate Decay Heat Removal" pg. I of 6 4.15(c)
REF: ~ EP-220 " Pressurized Thermal Shock" pg. 2 of 7 4.16-(c)
REF: EP-390, " Steam Generator Tube Rupture" pg. 1, 2~of 11 l
4.17(a)
REF: 10 CFR 20, 20.4 ,
4.18(d)
~REF: AI-500, pg 8&9 4.19(c)
REF: CR Ques. Cat 4&7, #39 (OP-204, AP-521) '
~
4.20(a)
REF: CR Ques. Cat 4&7, #39 (OP-204, AP-521) 4.21 (c)
REF: CR Ques. Cat 4&7, #39 (OP-204, AP-521) 4.22 (d)
REF: OP-502, Rev. 11, pg. 3 (sect. 4) 4.23(c)
REF: 0P-603,'Rev. 21, pg. 4 4.24 (c)
REF: OP-504, Rev. 7, Pg. 5 4.25(b) i REF: CRNS OP-502, pg.3
).
l L
/ AUTOMATIC INITI ATION OF MOTOR DRIVEN EMERGENCY
- FEEDWAILR PUMP EF P-1 Prior to the present refueling outage at Crystal River Unit 3, an automatic start of the actor-driven emergency feedwater pump (EFP-1) was not possible during a station blackout (lor.s of offsite power). Modifications sunnarized below, have been made to control circuits for EFP-1 that now pennit an automat-ic start with a station blackout. j in the event of a station blackout coincident with either a loss : of both main feedwater pumps or low-low levels in both 1 Once Through Steam Generators (OTSGs), EFP-1 will automatical-ly start five (5) seconds after Engineered Safeguards (ES)
Block 1 is loaded onto the "A" train diesel generator.
If an ES Actuation Signal' occurs concurrently, or anytime af-ter the above-listed conditions occur, EFP-1 will be automati-cally tripped, if already loaded on the diesel generator, and au,tomatically restarted five (5) seconds after ES Block 4 has
- been started. This assures that ES Blocks have priority load-ing on the diesel generators.
Under worse case conditions of ES Actuation with *B" Train diesel generator unavailable, concurrent with a station black-out condition, the automatic starting of EFP-1 will result in a voltage drop to 69.7% of nominal (4000 Volts) voltage. The ,
voltage drop, however, will last only 43% of the load-sequence time interval and will, therefore, preclude problems with op-erating rotational equipment. In addition, equipment such as contactors will remain energized since the design dropout voltage is 55% of nominal.
With the addition of EFP-1 to the loads associated with ES Blocks 1 through 4, the "A" train diesel generator, under worst case conditions, will be operating at 3181 kW, which is 96.4% of the 30-minute rating of 3300 kW. In order not to ex-ceed this rating, a watt transducer that monitors the diesel generator output has been installed. Alanns will be seen in the Control Room when the 2000-hour rating of 3000 kW is exceeded, and again when 25 of the 30 minutes have been expended. At 30 minutes, EFP-1 will be automatically tripped and subsequent breaker closure will be prevented until the
' trip circuit is manually reset. Resetting the trip circuit will be administratively controlled.
FPC considers this design change a short-tenn solution. Long-term solutions are still under consideration and future dr. sign changes will be submitted as these final solutions are determined.