ML20234D994
| ML20234D994 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/22/1987 |
| From: | Elin J, Johnston G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20234D949 | List: |
| References | |
| 50-344-OL-87-01, 50-344-OL-87-1, NUDOCS 8707070416 | |
| Download: ML20234D994 (151) | |
Text
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U. S. NUCLEAR REGULATORY COMMISSION l
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i REGION V 4
Examination Report No. 50-344/0L-87-01 Docket Nos. 50-344 Licensee: Portland General Electric Co. ,
i 121 S. W. Salmon Street Portland, Oregon 97204 Facility Name: Trojan Nuclear Power Plant, Prescott, Oregon q Examination Conducted: May -31, Chief Examiner: ff&hx 722 Gar (I ohnstph~ ~
'~
gategigned l l Ope ,t Li ehsing xaminer Approved by:
21 Ofate Sidned .
/Jhn0.Elin hief, Operations Section !
Summary:
1 Examinations on May 28-31, 1987 (Report No. 50-344/0L-87-01) .
Written and oral examinations were administered to two Senior Reactor Operator (SRO) and five Reactor Operator (RO) candidates. Two SRO and two ,l R0 candidates passed both examinations. Two R0 candidates failed the written '
examination. One R0 candidate failed the oral examination.
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I 8707070416 870622 l PDR ADOCK 05000344 )
V PDR l
a REPORT DETAILS
- 1. Examiners:
G. Johnston, RV (Chief Examiner)
T. Guilfoil, Sonalyst (Contractor) ;
- 2. Persons Attending-the Exit Meeting:
G. Johnston, RV S. Nichols, Training Manager G. Ellis, Training Specialist M. Peterson, Training Specialist i J. Bauer, Regulation Engineer i l
- 3. Written Examination and Facility Review:
The written examinations were administered to two Senior Reactor Operator (SRO) candidates, and five Reactor Operator (RO) candidates on l May 28, 1987. 1 t
At the conclusion of the written exam, the facility staff reviewed the- ;
The comments made by the staff at the conclusion of the review exams.
are included in the attachment (1). These comments were discussed with !
the staff and appropriate changes were made to the exams prior to the grading of the exams. ;
i One generic concern associated with the SRO examination concerned the l candidates' ability to determine the correct surveillance intervals. )
The candidates did not demonstrate a clear understanding of time j intervals associated with determining when surveillance tests would i exceed the appropriate intervals.
An item of concern about technical reviews of vendor supplied i information being incorporated into procedures was identified during the I review of the information supplied by the facility reviewers. There is a caution on page 5 of Operating Instruction 0I-3-4, " Reactor Coolant System Normal Operation," which appears only to. address the high leakrate potential for Seal No. 2 of the Reactor Coolant Pumps during a loss of seal injection, when the concern is clearly the limited capacity of the thermal barrier heat exchanger as evidenced in the Westinghouse technical bulletin supplied by the reviewers. This, clearly, is an indication of a failure to adequately review the technical bulletin's content prior to incorporating the caution in the procedure. The facility is advised to review this matter which raises a concern about adequate technical reviews of vendor supplied information.
- 4. Operating Examinations One RO candidate failed the oral portion of the examination. The other four R0 candidates and the two SR0 candidates passed the oral portion. 1
2 I
A generic concern associated with Nuclear Instrumentation (NI) was identified during the control room portion of the oral examinations <
The candidates did not evidence an understanding of the effects of malfunctions affecting NI.and the consequent symptoms they could observe.
The candidates (in particular the R0 candidates) could not-provide the examiner with a clear sketch of the' functional layout of the various NI l channels. This concerns the examiners because of the importance of this instrumentation to the operation of the facility.
- 5. Reference Material l
The examiners experienced some problems with the facility provided reference material. Specifically, there were no Technical -4 Specifications or General Operating Instructions provided as requested -
with the reference. material that was received. The Chief Examiner discussed this with the facility personnel during the exit interview.
The examiner did concede that there were substantial revisions to the 1 reference material over the previous submissions.. However, the 90-day letter does provide a. specific list of the information required. The ,
examiners were able to get the portions they needed prior to the administration of the examinations.
- 6. Exit Meeting lihe Chief Examiner met with the facility representative denoted.in Paragraph 2 on May 31, 1987. The examiner discussed the findings to that point and the examination process.
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I ATTACHMENT 1 1
Resolution of Facility Comments l
Reactor Operators Examination i Facility Comment: Question 1.15 The reviewers indicated that part a should include " fluctuates" as an 1 answer. j Resolution:
The examiner agrees that the closure of the suction valve could lead to possible cavitation which would cause fluctuation of current, i 1
i Facility Comment: Question 2.01 ,
i The reviewers provided a summary of feedv'ater pump trips that indicate that there are 12 definable trips for the feedwater pumps. Including in lieu of Overspeed a Mechanical Overspeed and Electrical Overspeed.
Thrust Bearing Wear and High Discharge Pressure were also identified.
Resolution: i The examiner agrees and will change the key per the reference.
l Facility Comment: Question 2.02 The reviewers made the comment that it was not clear what was meant by
" sources of makeup water" until the key was read.
l Resolution:
The examiner disagrees and finds no basis to assume that the candidates l cannot discern that the response sought means from where the RWST or Accumulators receives the water they require for makeup.
Facility Comment: Question 2.05 i "We have added a drawing in case the student describes it and doesn't I use the word ' scoop'."
Resolution:
The examiner would in any case evaluate the response of the candidate to determine if he has adequately described how the Bypass line is provided with a driving head. No change will be made to the key.
4 2
Facility Comment: Question 2.06 The reviewers indicated that they would like to see the addition of the system acronyms in parentheses.
Resolution:
The examiner will include these on the basis no change to the actual key is intended.
J Facility Comment: Question 2.11 The reviewers indicated for part (b.) " requirement no longer exists."
They provided a page from EI-1 confirming this. For the answer to part 3 (c.) the reviewers indicated that the candidates may not respond that the Service water would be supplied through a Service Water Booster
' pump, but only that Service Water is provided to the Jacket Cooling Water Heat Exchanger.
Resolution:
For part (b.) the examiner will delete the question from the exam. Part
- (c.) will be changed to place the words "via a Service Water booster l pump" in parentheses to indicate that portion is not required for the l response.
Facility Comment: Question 2.12 1
The reviewers indicated that for part (d.) (provided by reference l material supplied during review) that no Containment Isolation Signal is l present and Pressurizer level greater than 17% are also needed l interlocks that must be satisfied.
I Resolution:
The examiner reviewed the reference material and agrees that the two should be added with the points distributed at 0.25 points apiece.
l Facility Comment: Question 2.13 The reviewers indicated that the candidates may not provide the flow-path as described. In particular that the manual valves mentioned may not be included.
Resolution:
The examiner agrees that the question does not provide clear instruction l
to include valves in the response. Therefore, the portion "via manual valves" will be replaced with "from".
l
3 Facility Comment: Question 3.01 l The reviewers indicated that PCV-455A would receive an open signal but will not open because no arming -signal was present. (Reference material was provided.)
Resolution:
In reviewing the reference material, the examiner feels the term " arming signal" was inappropriate. The valve will receive an open signal but will not open unless an interlock provided from channel 458 is satisfied. j The examiner will change the key to read " Power operated relief valve l
PCV-455 will receive a signal to open." This will imply that the valve may or may not open.
Facility Comment: Question 3.03 The reviewers indicated they felt the question for part (a.) was misleading as far as the permissives listed " arming" the steam dumps.
They expressed that "no P-12, Lo-Lo Tavg" and "no Lo-Lo S/G 1evel for >
5 min. (TCVs only)" should be included as answers. They also indicated that "C-7 and C-8 both are not required. Either can arm dumps, while the permissives are required."
The reviewers indicated that the question stated the mode selector switch position of "AUT0" is not on the switch. The position marked corresponding to this was "Tavg".
Resolution:
The examiner feels that adding "no P-12" and "no Lo-Lo S/G 1evel for > 5 min." is not acceptable from the context of the question. They represent the absence of permissives and that is not what the question asked. It is true that either C-7 or C-8 will allow arming of the dumps, however, either one would be required for the action, therefore, they would be required to answer the call of the question. The examiner :
will not make a change to the key. 1 The examiner notes the comment about the mode selector switch and will change the question prior to uploading the question onto the examination bank. The examiner points out that Tavg and AUTO are analogous.
Facility Comment: Question 3.06 The reviewers indicated that the automatic function of the charging
! pumps should include the flow control valve for the centrifugal pumps.
They also indicated the setpoint has changed for the high pressurizer level trip (88% from 92%).
L_____-- _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _
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Resolution:
The examiner recognizes the flow control valve needs to be included. 1 The primary mode of operation of the charging system uses the j centrifugal pumps, the examiner will accept the flow control valve i opening or positive displacement pump speed increases. The key will be changed to include the 88% level for part (b.).
l Facility Comment: Question 3.10 The reviewers indicated that the Main Feed Pump nomenclature is 1 not 1A. They also indicated that the stator cooling water trip uses l generator current not impulse pressure.
Resolution:
The examiner will change the key prior to uploading the exam to the Exam Question Bank. :
Facility Comment: Question 3.11 The reviewers indicated that the P-6 permissive energizes the source range instruments and not the P-10 permissive.
Resolution:
1 The examiner agrees and will accept (c.) as a viable answer. )
i Facility Comment: Question 4.05 The reviewers indicated that answer 2 showed that secondary side temperature is < 50 F above RCS Tcold.
Resolution:
The examiner agrees and will include this in the key.
Facility Comment: Question 4.06 The reviewers indicated that Inventory does not exceed a Yellow path in any of their charts.
Resolution:
The examiner agrees, but points out that in any case the candidates should be able to place in order of priority the Critical Safety Functions based on the relative status of the color. No change will be made to the key for grading, however, the examiner will change it prior to uploading the question to the exam bank.
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. 5 l Facility Comment: Question 4.07 The reviewers stated that the question for part (b.) does not solicit the values of the parameters, but only the parameters.
1 Resolution: 1 The examiner agrees and will place the values in parentheses.
Facility Comment: Question 4.09 The reviewers indicated that part 4 is 35,000 mrem not 3,500 mrem.
Resolution:
The examiner agrees and will change key. (Correct answer is in parentheses as 35 Rem.)
l Facility Comment: Question 4.14 "We have a G0I (General Operating Instruction) so they don't have to j memorize all of this." '
Resolution:
The examiner quotes the Examiners Standard ES-202 "Tne canaidate is not expected to have normal procedures committed to memory but should be able to explain reasons, cautions, and limitations of normal operating '
l- procedures." Further, " Technical Specification questions for reactor
) operators should be conceptual in nature (e.g. , recognition of limiting conditions for operation and Technical Specifications that exist for a given are (a.))." The examiner points out the responses for these ,
questions appear to fall within these guidelines.
1 Facility Comment: Question 6.07 The reviewers commented that part (c.) of the question was not correct in their opinion. They believe that the level will return to 44% and not stay at 54%. They also indicated that part (d.) was a Turbine and not a Reactor trip.
Resolution:
The examiner points out that Lesson Plan B-11-LP was the basis for the Steam Flow Channel Failure. The reference clearly states the level will l be maintained at 54%, therefore, no change will be made to the key. As for part (d.), the examiner told the reviewers that the topic came up during the examination and he told the candidates that the trip was a Turbine trip and not a Reactor trip as stated in the exam.
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i Facility Comment: Question 6.11 I The reviewers stated that the candidates "should not have to memorize I which Flow Transmitters come off each buss." They added a comment to part (f.) that the RHR suction valve will close if power is not removed. j Resolution:
The examiner feels that the case for memorizing the appropriate buss for each Flow Transmitter is a task that would be made very hard if only from the sheer volume of transmitters, and therefore, will delete part (a.). The examiner will add to'part (f.) that the valve will receive a
.close signal. )
I Facility Comment: Question 6.12 I
The reviewers made an informational comment about part (b.). They I expressed that the signal will decay away rather than drop off l immediately as the answer in the key would seem to imply.
Resolution:
The examiner expressed that the comment would be taken into account but noted that it would not matter in grading the question. The answer could apply in either case and will not be changed.
l Facility Comment: Question 7.03 The reviewers indicated they wanted the answer to part (a.) changed to
- read
- " Thermal barrier heat exchanger is not sized to cool the excess l
I flow through the bypass line." The reviewers prctided a Westinghouse technical bulletin that indicated the seal bypass valve b uld only be opened if normal conditions existed in the RCS and with the puaips.
Resolution:
It is apparent that the chief reason is indeed the sizing of.the thermal barrier heat exchanger, the examiner will accept the comment. It is also apparent that the procedure does not appear to incorporate the technical bulletin's warning about the operation of the seal bypass valve during a loss of injection water casualty.
The caution on page 5 of OI-3-4 appears only to address the high leakrate potential for seal No. 1, when the major concern is the limited capacity of the heat exchanger. This is clearly an indication of a failure to adequately review the bulletin's technical content prior to incorporating the caution in the procedure. The facility is advised to review this matter which raises a concern about adequate technical reviews of vendor supplied information.
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Facility Comment: Question 7.04 l The reviewers supplied a new (very recent revision) of the procedure j ONI-7, " Reactor Control Malfunction." The revision deletes the action
~
to proceed to EI-5.
Resolution: '
The examiner will accept the comment and delete the requirement to go to EI-5 if the rods fail to insert.
Facility Comment: Question 7.05 The reviewers requested that part (b.) be deleted on the basis that the I area was not a High Radiation area but a High Radiation Exclusion area. l They identified that a High Radiation exclusion area is an area in excess j of 25 Rem and not 1000 mrem.
1 1
Resolution:
l The examiner will not delete part (b.). The reference provided by the reviewers indicates that a High Radiation Exclusion area is subject to !
the same requirements of a High Radiation area. It merely adds the requirement that additional approvals must be met to allow access.
Therefore, part (b.) is still applicable. The examiner will change the answer for part (a.) to 25 Rem.
Facility Comment: Question 7.06 The reviewers provided a reference that indicates that the entrance to the controlled area is posted indicating that radiation areas of 2 to 5 mrem an hour may be encountered. The posting at the door would not be made unless the radiation levels exceeded 5 mrem an hour.
Resolution:
, The question clearly asks what the posting would be at the door to the l room. The examiner, therefore, will delete Radiation Area as one of the posting requirements for part (b.). The examiner feels the weighting of the points warrants changing the point values from 0.33 to 0.5 and will do so, as this will be consistent with part (a.).
l Facility Comment: Question 7.08 The reviewers indicated that the answer to part (a.) should read: "When directed to by EI-0, or when transitioning out of EI-0."
1
8 Resolution:
The examiner feels this is not inconsistent with his reading of the procedure. In fact, the procedure specifically directs the operators to monitor the CSFSTs when the immediate actions are complete. The examiner will add in parentheses the comment by the reviewers "When directed to by EI-0." The second part of the comment will not be accepted because the question refers specifically to monitoring during the conduct of EI-0.
Facility Comment: Question 7.11 The reviewers indicated that Trojan does not have a low steamline pressure SI signal.
Resolution:
The examiner will accept the indicated changes the reviewers provided.
Facility Comment: Question 7.14 The reviewers questioned why the examiner expected the candidates to recall from memory the maximum time the Diesel Generators could be run at 4920 kw.
Resolution:
The examiner quotes the Examiners Standard ES-402: "This category contains questions on the procedures for the operation of the reactor and auxiliary systems.. 0perating restrictions and limitations in the facility license, including Technical Specifications, may be included to the extert they are directly applicable to a senior reactor operator."
From this, the examiner recognizes that this may fall outside the required knowledge and will drop the question from the exam.
i Facility Comment: Question 7.14 j The reviewers indicated that the answer did not appear to be an immediate operator action. 3 Resolution: )
1 The examiner strongly disagrees, the requirement to conduct a shutdown ]
margin calculation within one hour comes from the Technical Specifications. The candidate is expected to know those actions required by the Technical Specifications to be done within one hour. Procedure ,
ONI-16 merely ensures that this be done to meet the requirement.
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i Facility Comment: Question 7.15 l
The reviewers commented that "We already know these things: why check j them?
Resolution:
The examiner was merely providing the parameters, and indicating the condition that would be necessary to confirm. There would appear to be confusion from the context of the question, therefore, the examiner will delete this question.
Facility Comment: Question 8.05 The reviewers indicated that the computer would not necessarily be out 1 of commission and therefore the evaluation would be for an Alert fer I part (a.).
Resolution: ;
I The examiner agrees and will change the answer for part (a.) to Alert.
1 Facility Comment: question 8.09 i The reviewers indicated that the Fuel Handling Procedure FHP-5-1 indicates that two operators and not two R0 are required. They also indicated that there are other people needed for the evolution.
Resolution:
The examiner reviewed the reference material provided and will change the answer to two operators from two P.Os. The other personnel are not required and are part of the normal shift complement. The examiner points out the question asked for those personnel that must be present during fuel movement.
i Facility Comment: Question 8.12 The reviewers indicated that the question in part (b.) asks about changing a procedure. They contend that a change is not a deviation of l
a procedure.
Resolution:
The examiner regrets the apparent possibility of confusion. The question however still appears valid and will not be dropped. The examiner indicates that the term change does not conflict with the intent of the question or the context of the Administrative Order.
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y U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION =
- N Facility: Trojan -
v Reactor Type: westinghouse
/ Date Administered: April 28, 1987 gg/ , Examiner: carv w. Johnston 1 Candidate:
l INSTRUCTIONS TO CANDIDATE: l Use separate paper for the answers. Write answers on one side only. Staple ;
question sheet on top of the answer sheets. Points for each question are indi-
. cated in parentheses after the question. The passing grade requires at least ,
70% in each category and a final grade of at least 80L Examination papers '
will be picked up six (6) hours after the examination starts.
% of Category % of Candidate's Category Value Total Score _ Value Category
. 25.0 ~
25 5. Theory of Nuclear -
.cwor
- Plant Operation,
-t:1 aids, and Thermo-
. dynamics l
25.0 25 6. ' Plant Systeras Design, Control, and I Instrumentation 25.0 25 7. Procedures - Normal, ,
Abnormal. Emergency,
{
and Radiological Control 25.0 25
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- 8. Administrative Pro-cedures, Conditions, and Limitations 100.0 Totals Final Grade All work done on this examination is my own, I have neither given nor received l aid.
Candidate's Signature s.
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D NRC RULES AND CUIDELINES FOR LICENSE EXAMINATIONS During the administration of this' examination the following rules apply:
- 1. Cheating on the . examination means, an automatic denial of your application and could result in more severe penolties. .
- 2. Restroom trips are to h limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or popibility of cheating.
- 3. Use black ink or dark pencil 3,y, 1 to facilitate legible repteductions.
- 4. Print your name in the blank provided on the cover sheet of the assafaation. i
- s. .
- 5. Fill in the date on the cover simet of the examination (if necessary).
S. Use only the paper provided for answers. !
- 7. Print your name in the typer right-hand corner of the first page of each I section of the answer sheet. l S. Consecutively number each answer sheet, write "End of Category
- as appr*spriate, start each categor the paper, and write "Last page{ on a new on the"Tast answerpage, sheet. write only gne Egle of S. Number nach answer as to categosy and number, for example,1.4, 5.3.
- 10. Skip at least three lines between each answer.
al. Separate answer sheets from pad and place finished answer sheets face down en your dent, or table.
- 12. Use abbrevidtfons only if they are commonly used in facility literature.
- 13. The point value'for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer requir,ed.
- 14. Show all calculations, methods, er assumptions used to obtain an answer
. to anthematical problems whether indicated in the question er met.
- 15. partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAYE ANY ANSWER ~4LAMIL . .
- 35. If parts of the examination are not clear as to intent, ask iguestions of the examiner only. ,
- 17. You must sign the statement on the cover sheet ta't a indicates that the work is your own and you have not received or been given assittance in completing the examination. This must be dote after the examination has been completed. .
Examiner standards .-
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- 13. When you complete your examination, you shall: ,
- a. Assemble your examination as follows: l
. (1) Exas questions on top. .
j
. (2) Exam aids - figures, tables, etc. -
j (3) Answer pages including figures which are a part of 16e answer.
i
- b. Turn'in your copy of the examination and all pages used to answer l the examination questions. , .
- c. Turn in all scrap paper and the balance of the paper that you did -
not use for answering the questions. !
s- . .
- d. Leave the examination ersa, as defined by the examiner. If after ..
leaving, you are found in this area while the examination is still in progress, your license may be denied or revokad. - . .
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4
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1 o
O 0
9 1
9 8
.i D g e
Examiner standards
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REQUIREMENTS FOR ADMINISTRATION OF WRITTEN EXAMINATIONS .
- 1. A single room shall be provided for completing the written examina- . 1 tion. The location of this room and supportin restroom fact 11 ties shall be such as to prevent contact with all ok,her facility and/or contractor personnel during the duration of W uritten examination.
If necessary, the facility should make arrangeseds for the use of 4 a suitable room at a local school antel, or other building. 0b-tainingthisroomistheresponsi$ilityofthelicensee. -
- 2. Niriisus spacing is required to ensure examination integrity as deterstnen, by the chief examiner. Minisua spacing should be one candidat < per table with a 3-ft space between tables. Ito wall charts, models, and/,or other training antarials shall be present in .
the examination room. .
- 3. Suitable arrangements shall be made by the facility if the candi-dates are to have lunch coffee or other refreshments. These arrangements shall comply with Item 1 above These estrangements shall be reviewed by the examiner and/or proctor.
- 4. The facility staff shall be provided a copy of the written examination
- and answer key after the last candidate has completed and handed in bis written examination. The facility staff shall then have five working days to provide formal written comments with supporting documentation on the examination and answer key to the chief examiner er to the regional office section chief. .
- 5. . The licensee 'shall provide pads of 8-1/2 by 11 in. Ifned paper in unopened packages for each candidate's use in completing the exam-instion. The examiner shall distribute these pads to the candidates. ;
All reference material needed to complete the examination shall be i furnished by the examiner. Candidates can bring pens pencils calculators, c? slide rules into the examination toss, and no other equipment er reference material shall be allowed.
8.
Only black ink or dark penctis should be used for writ,ing answers to questions.
.*e Exaniner Standards .
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EDUATION SHEET fAV
$} = $ 2= Constant ,
= +
E = a5 a = obv -
TM P
3.10 X 1020 I=A7 0 = v(aP) + A..Y + II.oEl P = P.10(SUR t) 2ge ge )
P-Pe*N I .
o . v(ap) + a,L,2 + M + hL g 2gc Sc Keff = y , , {
Keff - 1 y 0 = vCAP) + g2 + g,Ltd + ht , hp g
- ~ L*
T'
+ Ar1L.
1 ,17 ht a f ,L E
_ D 2ge T-
'*"j," htkt 72 28c l
- ~ ~
ht.ksi,2 ,,
n g - MS, 2ge eg, Wp = m (ah) = m (au + v(aP)) = mv (aP)
ICRR CR g PHP = h p , ideal = ivCAP) '
M- U"5 CR.
ggp,h,real=
p v(6P)
- l 1
M~ 1 Keff
. "P "p = E BHP CR 3 (1 - Keff s ) - CR (1 - Keffs ) ,
J C, h= BHP-FHP KE - 1/2 mS2 2g,J At = " O ~1) cp3 ("P
N a 0 (Flow)
N2 a AP (Head)
N3 o SHP (Power) hSAT - vPSAT a
hr > hSAT CONSTANTS & CONVERSIONS l -
NPSHA - 7h (Suction) hsat A- 6.024 X 10:s atoms / mole 2 1 watt - 3.1 X 1088 fissions /second CPSHA - vP 2 + -hSAT
&c 1 Curie - 3.7 X 101' disintegrations per second !
(
1 amu - 931 Mev {
NPSHA = vP1 + b -hL -hSAT Ec 5 - 32 ft/sec8 hEL " hh+hPE 88 - 32 t-lba/ sect-lbf fr - Ibf J - 778 hp<hEL I'N 1
'T - (9/5'C) + 32
'C - (*F-32) 5/9 TDH = 3AI (Fcold * # hot) 'R = 'F + 460 ;
Se i
- K 'C + 273 .
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1
$ a AT /2 $ /3 c p - 1.0 BTU /lba *F G STP .
3 H 0 - 62.4 lbm/ft8-1 5m/cm8 9 STP
. I = I.e*#* - 1.10*I* M )
I nd: = Inds
.I d -Indj
/a D.
R/hr , CE (meters)
R/hr - 6 CE/d8 (feet) '
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i SECTION 5 Theory of Nuclear Power Plant Operation, Fluids, and Thermodynamics l'
QUESTION 5.01 l
l If the speed of a Positive Displacement Pump is varis3d the flow rate of the pump will' vary proportionately. Refering to Figure 5.01:
How could the head of the pump be maintained constant if the speed of the pump is lowered? (1.0)
- ANSWER By throttling closed the discharge valve. (1.0)
.
- REFERENCE O2-H-12-HO, FF-26 I
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1 4
1 1
1 4
u !
' 3' &
8 s +
o U i m E +"
o s !
- I
( 0
.spm h pump _ _ _ _ _ _ curse YdG6 .
Volumetric Flow Rate (V) !
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TYPICAL POSITIVE DISPLACEMENT PUMP CHAR ACTERISTIC CURVE i
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j
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=
4 QUESTION 5.02 After a Reactor Trip from a Loss of Offsite Power. event'the-Reactor Operators determine that natural-circulation is occuring.
If the Steam Generator pressure was lowered by using the Atmospheric Dump valves, approximately how soon would the effect be seen on the Core Exit Thermocouple? . (1.0)
(No calculation necessary.)
- ANSWER Transport time f or natural circulation is on the order of 5 to 10 minutes. (1.0) 4
- REFERENCE O2-H-12-LP page 41, 02-H-12-HD part FF-29.
1 4
a GUEST 10N 5.03 l 1
~
After taking critical data at 5x10 amps in the intermediate-range, a reactor operator establishes a stable startup rate of 0.15 DPM at 0.1% of full power. The reactor is at no-load Tave,.
and the Boron concentration is 800 ppm. The plant was below the point of adding heat.
- a. What would the Reactor power be after 2 minutes? (1.25)
- b. What defines the point of adding heat (PDAH)? (0.5)
- c. At what power level does the point of adding heat' begin? (0.5)
- ANSWER j
' p = p to(SUR) (t) (0.75) o 0.15)(2)
= (0.1)10 = 0.199526 % (0.5)
Will accept round off to 0.2% l
- b. The point at which heat energy addition to the coolant due to the fission rate exceeds the heat losses to ambient.
(0.5) l' A -b
- c. About 2h10 amps or 1% to 2% of full power. (0.5) l
- REFERENCE l H-07-LP pages 15,23.
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QUESTION 5.04 If the Reactor Coolant System (RCS) is operated at satur'ation ,
conditions, a f orced two phase fluid flow will result. Under !
these conditions the indicated RCS flow rate will be higher than l
- 1 the actual flow rate.
- a. What physical property of the two phase liquid causes the actual flow rate to be less than indicated flow 4 rate? (0.5) ,
- b. For the same mass flow rate, why would system head !
losses be greater with forced two-phase flow than l forced single phase (liquid) flow? (0.5)
- ANSWER
- a. Lower density of two-phase fluid. (0.5) J
- b. Higher fluid velocity. (0.5)
- REFERENCE O2-H-12-LP page 39.
1 1
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I OUEST10N 5.05 .
i
- During a ref ueling of the Reactor a 1/M plot is being maintained in the Control Room. Looking at Figure 5.05_which shows three 1/M plots and three detector positions around the core (X, Y, and Z):
- a. Which detector in the drawing of detector layout best represents the corresponding output in the 1/M plot for curve (a)? (0.5)
- b. Which detector in the drawing of' detector layout best represents the corresponding output in the 1/M plot f or curve (b) ? (0.5) l
- c. Why.would the curve (c) 1/M plot be considered less conservative than the curve (b) 1/M-plot? (1.0)
- ANSWER
- a. Detector Y. (0.5)
- b. Detector X. (0.5)
- c. The slope of curve (b) would indicate criticality with fewer
. fuel elements than actually needed, while curve (c) will l initially indicate more f uel elements than actually a l requireded for criticality. (1.0) )
- REFERENCE H-04-LP pages 6 to 19, figures 4 to 9.
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QUESTION 5.06 ;
The Reactor Coolant System is being heated by the use of the Reactor Coolant Pumps f or F10DE changes f rom MODE 5 through MODE I
- 3. .
- a. Why will the current drawn by a Reactor Coolant Pump ciecrease as the system is heated f rom 200 F to 564 F7 (1.0)
- b. What will happen to the motor current of the Reactor Coolant Pumps if cavitation occurs'? (1.0)
- ANSWER
- a. The density decrease means less work for the pump. (1.0)
- b. There should be fluctuations in the current. (1.0)
- REFERENCE H-12-LP, H-12-HD 4
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I QUESTION 5.07 i As fuel temperature incr eases, the reasonance absorption ~ peaks for U-238 become lower in height and the' bands broaden, but the )
area under the curve remains theoretically constant. .
Why then does a heatup of the f uel result in a negative <
reactivity i nserti on? '(1. 0) ;
)
- ANSMER l l
Due to decrease in self shielding of the fuel pellet as the range j of neutron energies in the absorption band increases. (1.0) l
- REFERENCE 1 l
H-06-LP, pages 28-29 i 1
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1
QUESTION 5.08 After operating for several weeks at 100% power, a react'or at EOL is shutdown for maintenance. .An Estimated' Critical. Position (ECP) is calculated for a startup to be conducted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the. shutdown. Why is the Actual Critical Position (ACP) LOWER than the ECP if the following occur:
- a. The Steam Dump Pressure Controller is changed from 1080 psig to 900 psig af ter the ECP calculation. (1.0)
- b. The startup is delayed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (1.0)
(Consider both cases independently and assume temperature coefficients of reactivity to be negative.)
- ANSWER
- a. A decreased pressure setting lowers Tave.. Assuming a negative MTC, this adds positive reactivity so the ACP is LOWER. (1.O)
. b. A 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> delay adds positive reca.tivity since Xenon is decaying beyond its ECP value. The ACP is LOWER. (1.0)
- REFERENCE H-06-LP, pages 18-22, 45-47
QUESTION 5.09 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after a reactor trip you are directed to take'the plant to 300 F to do some maintenance on a supporting sysgem. The )
reactor is shut down by 3.6% at Hot.Zero Power (557 F),,and the reactor coolant system is at 900 ppm boron. The reactor has ,
negative temperature coefficients of reactivity (average [557 to 3003 MTC = -7 pcm/ F) and the Differential Boron worth i s -12 pcm/ ppm.
- a. How much (in pcm) is the reactor shut down at Hot Zero Power? (0.5) l
- b. How much reactivity (in - pcm) is inserted into the j reactor by cooling down to 300 F7 (0.5) ;
- c. How much (in pcm) must the reactor be shut down to be at 5% shutdown? (0.5) ;
- d. What baron concentration must the plant have to be 5%
shutdown at 300 F7 (1.0)
- (Ref erence material on following pages.)
- ANSWER O.964 - 1
- a. rh = --------- = - 3734 pcm at HIP (0.5)
HZP 0.964 1799
- b. NEW SD = -3734-+ (300 - 557)(-7) = -4235- p c m (0.5) 1 0.95 - 1
- c. rho (5% SD) = -------- = ~5263 pcm (0.5) 0.95
- d. baron pcm to adjust = -5263 -(-1935) = -3328 pcm (0.5)
Differential boron worth BOL (900 ppm HZP) = -12 pcm/ ppm
-3328 ;
boron conc. = 900 + ----- = 900 +277 = 1177 ppm (0,5) j
-12 (Examiner will allow errors to be carried without penalty, and will allow 3 digit accuracy.)
- REFERENCE H-07-LP pages 20 to 23. I
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QUESTION 5.10 The following conditions exist in the Reactor That = 565 F Tavg ='549 F Tcold =:533 F Pressurizer pressure = 2185 psig 1
Steam Generator pressure = 850 psig
- a. What is the subcooling margin of the plant with the above conditions? (1.0) -
l b. If power is raised f rom 50% to 100%, why will the .
subcooling margin decrease? (1.0)
- ANSWER J
- a. Tsat for 2200 psia (2185 psig) (0.5)
= 649.45 F SCM = Tsat - That = 649.45 - 565 = B4.45 F (0.5)
. b .- SCM will decrease because of the increase of Thot. (1.0)
- REFERENCE Steam Tables, H-12-LP, H-12-HD 1
l
QUESTION 5.11 i
When reactor power. is increased f rom 50% to 100% the f ol' lowing
-changes occur to DIFFERENTIAL rod worth from the conditions stated. .
Considering each case separately, what causes the DIFFERENTIAL l rod worth to change in the direction described?
I
- a. Differential rod worth decreases with temperature
' decrease, with rod position and boron concentration held constant. (1.0)
- b. Differential rod worth decreases as.the. rods are withdrawn from 150 steps to maintain temperature constant, with boron held constant. (1.0)
.q
- c. Dif f erential rod worth increases as boron concentration-is diluted to maintain temperature constant, with rod j
position held constant. (1.0)
- ANSWER
- a. Lower temperature increases the density of the coolant'.and increase the boron concentration.per. unit volume, thus reducing the number of neutrons available for' capture by the rods. (1.0) l b. There is less flux density available f or interaction as the l rods move towards the top of the core. (1.0)
- c. Decreased boron concentration increases the number of.
neutrons available for interaction with the rods. ( 1. 0) .
- REFERENCE H-06-LP l
_ _ _ _ = _ - - _ _ _ _ _ _ _ _ - _ _ - _ _ _ - - - _
r QUESTION 5.12 The reactor is inititally operating at.75% power, beginning of. j l
life, with bank D rods at 150 steps. Baron concentration is 900 ppm. Using the attached curves (Figures 1.1, 1.6, and 1.20) answer the following questions considering each question separately.
- a. With no rod motion, what change in boron concentration is necessary to increase power to 100%7 (1.0)
- b. What will be the rod insertion of Bank D immediately after a power reduction to 25% without any boron concentration change? (1.25)'
- ANSWER
- a. at 75% BOL power defect is +1146 pcm (+/- 20 pcm) at 100% it is +1490 pcm (+/- 20 pcm) (0.5)
Change in reactivity required = 344 pcm Boron worth = 9.15 pcm/ ppm (+/- O.1 pcm/ ppm) (0.25) 344 pcm/ (9.15 pcm/ ppm) = 37.6 ppm (+/- 2 ppm) (0.25)
- b. at 75% BOL power defect is +1140 pcm (+/- 20 pcm) at 25% it is 438 pcm (+/- 20); delta rho = 702 pcm . (0. 5 )
Bank D at 150 steps IRW = 350 pcm (+/-20) 350 + 702 = 1052 pcm (0.5) therefore Bank D = 55 steps (+/- 5) (0.25)
- REFERENCE CROCTRM l
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Positive worths shown for use TrueBoronWorth=hxh in COI-8 ECP .
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- FIGURE 1.1 !
DIFFERENTIAL BORON WORTH VS. BORON CONCENTRATION, BOL AND EOL. HZP j Revision 60 j Hay 30, 1986 ,
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_A '8ANK C 100 228 BANK B 100 223 DIFFERENTIAL AND INTEGRAL ROD WORTH VS. STEPS WITHDRAWN, BANKS D, '
C, AND D MOVING WITH 100 STEP OVERLAP, AT BOL, HFP, EQUILIBRIUM XENON Figure 1.6 Revisaon 60 May 30, 1986
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BOL 1900 PPM +1178 +912 6848 +358 0 SQL 1200 PPM 61359 41049 6733 4405 0 SQL 800 PPM 91448 11148 1803 +438 0 NOL 800 PPM 41820 91240 1870 sett 0 EOL 300 PPM e1764 etaas s963 +548 0 -
EOL 0 PPM 41924 t1485 t1040 e588 0 FOWER DEFECT VS. POWER LEVEL AT 80L (150 MWD /MTU), MOL (5000 MWD /MTU)
- AND EOL (10050 MWD /MTU)
Figure 1.20 Revision 60
.= . ..
QUESTION 5.13 i
Considering the diagram on the following page. Assume Trojan is at.100% power. -
. j I
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- a. Draw the temperature profile from the centerline of the fuel to the centerline of the coolant channel. (2.0) i
- b. Draw the heat production profile from'the centerline of ]
the fuel to the centerline of.the coolant channel? l (1.0)
- ANSWER Key of graph on next page. ;
I
- REFERENCE j General Physics Academic Program for Nuclear Power Plant Personnel l
END OF SECTION 5 GO ON TO SECTION 6 l
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SECTION 6 Plant Systems Design, Cortrol and Instrumentation
~
QUESTION 6.01 1
Regarding the Rod Control System rod stop signals: I l
- a. What are the two rod stops that are active only when 1 the Rod Control System is in automatic? (1.0) i
- b. What rod stops also generate a Turbine runback? (1.0) {
i
- ANSWER l l
- a. Low power interlock (C-5). (0.5)
(0.5) I Bank "D" withdrawal limit (C-11).
- b. OTdT (C-3) (0,5) J
{
OPdT (C-4) (0.5) l I
- REFERENCE 4 B-09-LP page 19.
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QUESTION 6.02-MATCHING. Match each of the actions with the appropriate pressurizer pressure control or protection setpoint. - Actions may be used more than once or not at all. Place an(2.0,appropriate 0.33 each) number or numbers after letter on your paper.
SETPOINT (psi g) ACTION
- a. 1865
- b. 1915 1. Normal Pressure
- c. 2220 2. Reactor Trip
- d. 2260 3. Variable Heaters off
- e. 2335 4. Saf ety valves open-
- f. 2385 5. PORV Interlock
- 6. PORV Opens
- 7. Safety Injection Block
- 8. Safety Injection Actuation
- 9. Variable Heaters Full On
- 10. Spray Initiated
-
- ANSWER
- a. ---
2
- b. ---
7
- c. ---
9
- d. ---
10
- e. ---
5 and 6
- f. ---
2 E6 answers e 0.33 ea.3 (2.0)
- REFERENCE B-04-LP pages 12 and 13.
l
l QUESTION 6.03 What two (2) sources of f l ui d , other than Seal Return f!'ow, may be cooled by the Seal Water Heat Exchanger? (1.0)
- ANSWER
~ \
l Excess Letdown. (0,5) ;
CCP miniflow. (0.5)
- REFERENCE A-06-LP page 18,
(
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l DUESTION 6.04 The interlock between the Letdown Isolation Valve and the Letdown Orifice Isolation valves ensures that RCS pressure is kept on the
_ regenerative heat exchanger to prevent flashing of steam and HX damage.
How does the interlock accomplish this f unction? (1,0)
- ANSWER Orifice Isolation Valves cannot be opened unless Letdown Isolation Valves are open, OR; Letdown Isolation Valves nay ba closed only if all Orifice Isolation Valves are closed. (1.0)
(Orifice Isolation valves are always first to close and last to l open.)
- REFERENCE A-06-LP page 6.
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QUESTION 6.05 The Reactor Makeup System is set to makeup to thel Reactor Coolant System automatically.
i a.- Other than "OFF", what three (3) posit _ ions of' the
" Reactor Coolant Makeup Mode Selector Switch" will NOT result in the delivery of a BLENDED makeup flow? (1.5)
- b. Other than to the Reactor Coolant System, to what two (2) other locations can the Reactor Makeup System supply blended output f or makeup? (1.0) l 1 l
- ANEWER
- a. BORATE, DILUTE, and ALTERNATE DILUTE. (1.5)
- b. CVCS Holdup tanks. (0.5)
Refueling Water Storage tank. (0,5)
- REFERENCE l
. D-05-LP i
i I
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j
R
. QUESTION 6.06 The Residual Heat Removal System provides cooling of'the Reactor Coolant System during cooldown from MODE 4 to MODE 5.,
- a. How is the cooldown of the Reactor Coolant System regulated if,the RHR system'is running at a constant flow rate of'5000'gpm?-(1.0)
- b. Why is the RHR system required to provide a flow rate-of 3000 gpm or more_in MODE 67 (1.0)
- ANSWER
- b. To ensure that boron p .9 '-- - '_ _q; stratificati on do' not occur. (3.0)
- REFERENCE
. 02-A-08-LP 1
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QUESTION 6.07 l Steam Flow Channel 512 detector fails HIGH (maximum 133% of normal full flow) during normal full power operation, but a maximum dif f erential pressure at the detector did not exist when the failure occured. ]
- a. What parameters of the Steam Generator Level Control System will INITIALLY cause the Feedwater Regulating valve to open? (1.0)
- b. What parameter of the Feedwater Pump Speed Control ]
System will cause the Main Feed pump to increase in speed as the Feedwater Regulating valve opens? (0.5)
- c. What will cause the Steam Generator level to rise above i the program level of 44% to 54%7 (0.5)
- d. What is the High Level <e--reve trip setpoint? (0.5)
"Tudk )
I
-
- ANSWER j
- a. Mismatch between f eedwater flow and erroneous steam flow.
(1.0) l
- b. Steamline/feedwater D/P (below 195 psid).
1
- c. The steam flow. (0.5) i l
- d. 75% (0.5) l
- REFERENCE B-11-LP page 20.
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QUESTION 6.08 The Over-Temperature delta Temperature (OTdT) trip provides protection for Departure from Nucleate' Boiling. -
- d. What is the justification that explains the reason for NOT using RCS flow as one of the parameters for calculating the OTdT trip setpoint when RCS flow has a direct relationship to DNB7 (1.0) 1
- b. Three parameters are utilized to generate the OTdT l trip, what are those three (3) parameters? (1.5)
- ANSWER
- a. (The flow is assumed to be constant) and several trips exist to provide protection for loss of flow. (1.0)
- b. 1. RCS pressure. (0.5)
- 2. Tavg (0.5)
- 3. Delta I (Axial Flux distribution) (0.5)
-
- REFERENCE B-03-LP pages 33 to 38.
.------._n--__
QUESTION 6.09 Regarding the Pcwer Range Nuclear Instrumentation.
- a. What three blocks are permitted by Permissive,P-10 when power is above 10% of full power? (1.0)
- b. What are the two (2) design events that would cause a trip signal from (+ or -) 5% of reactor power in 2 seconds with a coincidence of 2 out of 47 (1.0)
- ANSWER
- a. Power Range low power (25%). (0.33)
Intermediate Range (25%) (0.33)
IR rod stop (20%). (0.33)
- b. Rod Ejection. (0.5)
Multiple rod drop. (0.5)
- REFERENCE l
B-07-LP page 16.
QUEST. ION 6.10 Containment Spray . system design requirements dictate a rninimum and maximum pH limit during system operation and accident recovery. ,
- a. What is the reason for the minimum limit?'(0.5)
- b. What is the reason for the. maximum limit? (0.5).
- ANSWER
- a. Minimum:- Maintain ' Iodine in solution. (0.5)
(0.5)
- REFERENCE Containment Spray Lesson Plan 4
l i
i QUESTION 6.11 120 Volt AC Pref erred Instrument Bus Y-11 loses power. What: l effect will this failure have on the following instruments or j components?
- a. "C" 'is- . . . J m . i. m T1 414. '^ 254 drog/
- b. Power Range channel N-43. (0.25)
- c. Train 'A'1ESF Sequencer. (0.25)
- d. CVCS Makeup control system. (0.25)'
- e. Steam Generator 1 3evel indication. (0.25)-
1
- f. RHR Suction Isolation Valve MD-8701. (0.25) I 1
- ANSWER
- a. Fails low.
Nothing (Y-13).
(0.25)
(0.25) i o-b.
- c. Will not sequence. (0.25) ]
- d. Automatic makeup lost. (0.25)
- e. Goes to no load level., (0.25)
- f. Closes y N( gvuss 6-.P (0.25) i re8vd
- g l
- REFERENCE {'
C-06-LP
+
i 4
1 I
l a
QUESTION 6.12 The " Power Mismatch Channel" portion of the rod control system utilizes inputs from Turbine Impulse Pressure and Auctioneered High Nuclear Power : ,
- a. Why is Turbine Impulse Pressure used in the Power R I
Mismatch Channel? (0.5) l
- b. What would happen to the Rod Control System if the l Power Mismatch. Channel recieved a HIGH spike (lasting i j second) from the power Range NI's? (1.0) i i
- ANSWER
- a. To represent turbine load on primary. (0.5) j
- b. Will cause a high rate insertion signal only while the spike is occurring as mismatch is a rate error signal. (1.0) {
- REFERENCE-B-09-LP page 16.
- i 1
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QUESTION 6.13 The.High Pressurizer Level reactor trip is' automatically blocked j during shutdown. This allows control rod testing or withdrawal at <
low temperature. ,
What' conditions (i . e. logic and coincidences) are.necessary for this automatic block to occur to prevent the trip during shutdown? (1.0)
- ANSWER
.3/4 Power range channels.less than 10% AND 2/2 Turbine impulse power less than 10%. (1.0)
- REFERENCE , q Reactor Protection System Lesson Plan
. I 1
1 2
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~ QUESTION 6/14 , , j Regarding the Pressurizer Pressure Control System: ,
- a. Why will . the Pressurizer Spray valves NOT automatically open at,the same pressure even though.the setpoint is' not, changed? (1.0)
- b. How would an inadvertent RCS depressurization be prevented.if the pressure signal controlling a Power >
l Operated Relief Walve (PORV), should f ail high7.-;(1.5L ..IJ
+
- ANSWER a.. .The' actual operating.setpoint is modified by duration as,'
I well.as' rate of change. . ( 1. 0 ) 1
.I An interlock signal prevents automatic operation . below -
b.
(2335) psig. The interlock signal is provideo by another: .
channel (IN other words both channels must be' > 2335. ) . l (1.5)
.
- REFERENCE B-02-LP page 35.
END OF SECTION 6 GO ON TO SECTION 7 I
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SECTION 7
-I
. Procedures, Normal, Abnormal ~, and Emergency
~ '
OUESTION 7.011
~
I fin accords.nce with procedure ES-0.2 " Natural Circulation - lj Cooldown"rL 1 a.: What'two (2)-parameters'must be observed'in a downward #)
trend toLverify that the RCG is cooling down? (1'. 0) . , 1
)
.- b.. What' parameter must be observed ' to. be iricreasing to ']
verify'the RCS-isicooling'down? (0.5)
- c. What is the' purpose, in step 16'of procedure:ES-0.2, of having the operators verify'that; pressurizer level response is NORMAL? (1.0) ,
1 l
l
.
- ANSWER
- a. 1. Core Exit Thermocouple (0.5) i
- 2. RCS Thot wide range (0.5) !
- b. RCS subcooling (0.5) !
Large variation in Pressurizer level are indicative of Lvoid c.
formations in the vessel head. (1.0) l
- REFERENCE ES-0.2 pages 7 and 10.
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QUESTION 7.02 In accordance with procedure ONI-10 " Emergency Boration"t am What are the five (5) symptoms that require'the
~
immediate initiation of Emergency Beration by opening valve MOV-81047 (2.0)
- b. What alternate method of Emergency Boration is availiable7 (1.0)
- ANSWER ,
- a. 1. An uncontrolled cooldown of the reactor coolant system following a reactor trip. (0.4)
- 2. An unexplained or uncontrolled reactivity increase.
(0.4)
- 3. Insufficient Shutdown Margin. (0.4)
- 4. Excessive control bank insertion. (0.4)
- 5. One or more rod poroition indicators fail to indicate rods f ully inserted f ollowing plant shutdown. (0.4)
- b. From RWST to Charging pump suction (via MO-112D and
- MO-112E). \ t ,0)
- REFERENCE ONI-10 pages 1 and 2. j i
l s
k
QUESTION 7.03 Answer the f ollowing questions regarding RCP operation: )
~
1
- a. Why should the RCP number i seal bypass valve not be j opened during a loss of seal in.jection- casualty while operating at power? (0.5) 4
)
- b. Under what general plant condition and for what purpose .!
is the RCP number i seal bypass valve opened? (1.0)
- c. Why should seal injection flow to the RCP's be maintained greater than 6.6 gpm to each pump? (0.5)
- ANSWER
- a. Da=n4ng the valve may al;; ::u;; ; "igh 10 ' ratz ::ndition for um is u . 2 .;d . (0.5)
- 6. For the. purpose of providing sufficient cooling' flow to the pump radial bearing if the No. 1 seal leakoff temperature approches it's alarm. (1.0)
O
- c. To avoid high temperature the water reaching the No. 1 seal.
(0.5)
- REFERENCE 01-3-4 pages 5, 6, and 7.
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QUESTION 7.04 During operation at 80% power the Control Operator notic'es-that the Bank D rods are stepping out rapidly with no apparent reason.
What are four (4) of the five immediate actions that must be executed according to ONI-7 " Reactor Control Malfunction"? (2.0)
- ANSWER (0.5 each for any 4)
- 1. Shift rods to manual.
- 2. If they don't. respond maintain plant in stable condition with changes in turbine load and boron.
- 3. If not stable trip reactor.
- 4. If :d: fail iv inseri wm tw I-3.
- 5. Consult EPIPs.
- REFERENCE ONI-7 page 2. -l
)
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- _ _ - _ _ _ _ ~ _ __
QUESTION 7.05 The Radiation Protection Manual describes the variola controlled areas in the plant.
- a. What radiation level determines whether an arsa is a High Radiation Exclusion Area? (0.5) h
- b. fIf an area is going to be a High Radiation Exclusion - /<2#' - -
<F Area f or less than 30 days, what other action, other Ithan providing a lockable entry should.may be taken to C L control access? (0.5) M J hmaia, aa \/C .
7vvvvv v ( A c6 y , -
J I
- ANSWER 25 W em/hr (0.5) {
a.
/
- b. f'mntinn'= ="rvei11mnea- (0.5) p, Ctrn %c>v5 suvvefk c_ [p,$r)
- REFERENCE Radiation Protection Manual pages 2-25 to 2-28.
4 1
- - _ _ . _ _ -_-.__-___ ___ _ m
'
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QUESTION 7.06 For.each of the following survey conditions what would be the '
posting REQUIRED at the entrance to the room? -
l A portion of 10CFR2O is provided for use.
- a. East RHR rooms (1.0)
\
1 General Area Radiation: 125 mrem /hr Airborne Activity: 6 X10E-6 j
l Limiting Nuclide:- Krypton 88
- b. North Charging. Pump rooms (1.0)- (
1 G'eneral Area Radiation: 4 mrem /hr-Airborne Activity: 2 X10E-7 {
Limiting Nuclide: Iodine.131 i Pump Tray and floor 10,000 dpm/100cm 1 1
- ANSWER
- a. Airborne Radioactivity Area (0.5) 1 l
High-Radiation Area (0.5)
- b. 9:dicticr. Arca 'O.03)
Airborne Radioactivity Area Contamination Area (TT 37) g
- REFERENCE 4 Radiation Protection Manual pages 2-25 to 2-20.
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QUESTION 7.07 Regarding Procedure 01-6-1 " Liquid Radwaste":
- a. How many liquid volumes must be recirculated priorito a liquid radwaste release from the Treated Wast & Monitor Tank? (0.5)
- b. What action must be taken, if Plant Effluent Radiation Montitor 9 (PERM-9) is out of service, in order to continue a discharge _in progess? (1.0)
- ANSWER
- a. At least two volumes. (0.5)
- b. May continue, as'long as two independent samples are analyzed, and two independent plant personnel check the discharge valve lineup. (1.0)
-
- REFERENCE 01-6-1 page 12.
I 1
v n QUESTION 7.08 Procedure EI-O " Reactor Trip, Safety Injection, and Diag ~nosis is used on the initiation of a Safety Injection or Reactor. Trip.
- a. WHEN are the operators required to start monitoring of.
the Critical Safety Function Status Trees (CSFSTs)- during the conduct of'EI-07 (1.0)
- 6. What is the order of priority of.the following listed-Critical Safety Function Status Trees from highest.to-lowest? (1.5)
- 1. Containment - Red
- 2. Core Cooling - Red
- 3. Heat Sink - Yellow
- 4. Integrity. - Orange
- 5. Inventory - Green
- 6. Subtriticality - Orange
,* rer.,he 6[ b j F1 te w Ym 5[ [wh N b *
- a. fAfterthecompletionoftheimmediateactionstepsofEI-O}
(1.0)
- b. 2. Core Cooling - Red (0.25 each)
- 1. Containment - Red
- 6. Subtriticality - Orange
- 4. Integrity - Orange
- 3. Heat Sink - Yellow
- 5. Inventory - Green
- REFERENCE EI-O page 10, TAP-141 pages 44, 45, 12, 29.
QUESTION 7.09' What immediate actions must be taken per ONI-23 " Loss of One Feedwater Pump" if the Turbine fails to runback to 60%7 .(1.0)
~
- ANSWER The operator is required to runback the Turbine manually keeping S/G 1evels > 25%. (1.0)
- REFERENCE ONI-23 page 2.
l l
QUESTION 7.10
.During MODE 5 operations with the Reactor Coolant System Drained down to mid loop for installation of the Steam Generator. nozzle dams the plant experiences an inadvertent gas binding.of the-
~~
running RHR pump. How would the operators cycle MD-8812 to an-intermediate position to supply RWST water to the RCS as directed by ONI-13 <
"Malf unction of Residual Heat Removal Loop"? (1.0)
- ANSWER The operator must take the switch to Lockout momentarily, this l will stop'the valve in an intermediate position. (1.0)
- REFERENCE- !
ONI-13 page 3.- l l l t 1 I
QUESTION 7.11 After a Loss of All Feedwater event, using Procedure FR-H.1
" Response to a loss of Secondary Heat Sink", an attempt to restore Auxiliary Feedwater was unsuccessful and the operators ;
have tripped the Reactor Coolant Pumps. The operators are now 1 proceding to establish feed to the Steam Generators with the Condensate pumps.
- a. Why must the Reactor Coolant Pumps be stopped after the j unsuccessful attempt to restore Auxiliary Feedwater? {
(1.0) j
- b. Why must the operators depressurize the Reactor Coolant system to < 1865 psig? (1.0) )
- ANSWER
- a. The RCPs add heat to the RCS which is being removed by the j inventory in the Steam Generators. (Tripping the RCPs I ensures the inventory in the S/Gs is being lost much ]
slower.) (1.0)
- b. This will allow the blocking of the PZR low pressure SI, and 3 low steamline pressure SI to prevent an inadvertent SI i actuation which will cause steamline and f eedline isolation. ]
(1.0)
- REFERENCE FR-H.1, and WOG Emergency Response Guidelines l
l i I Il l
QUESTION 7.12 , i Regarding Operating Instruction DI-5-1'" Diesel Generators and J Fuel" provides precautions and limitations f or the operation: of .{ the Diesel Generators. .
~
i
- a. What is the ma>:imum amount of time the Diesel Generators may be run at rated speed and voltage with -
- ND load? ( 0. 5 ) -
I g Hm ' 1 eng-may-the-Di-esel-GerterwttFiin5Crun-when the 1. cad ] Q4q d is 4?20 kW7 {
- ANSWER
- a. 4 hours (0.5)
-t. _ 200 h J.5) De.)cfe c
- REFERENCE 0I-5-1 page 2.
- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ . . _ _ . _ ]
;)
i
, 1 i
QUESTION 7.13 Procedure 01-5-2 " Safety Injection" cautions against.fil' ling the Accumulators when the Reactor Coolant System is solid.
- a. Why is filling of the Accumulators prohibited when the RCS is solid? (0. 5) . ~
i
- b. Why is it prohibited to make. level increases of more ')
than 3 % to more than one Accumulator at a time when in. MODE 17 (1.0) l
- ANSWER
- a. Potential overpressurization of the RCS. (0.5)
- b. T.S. require that solution concentration of baron be checked when level is increased > or = 1% of volume, the tank is not considered operable until that is done. Two tanks out would place plant in an un-analyzed condition. (1.0)
- REFERENCE
. 0I-5-2 page 6, T. S. 3/4.5.1
l
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QUESTION 7.14
. J During a channel calibration of one Source Range drawer during /
MODE 5, it becomes apparent that both Channels of Source. Range { instrumentation are inoperable. Per DNI-16 "Excore Nucipar - Instrumentation Malfunction": ~
- ;i What action must be taken to provide assurance that an in-advertant criticality will not occur? (1.0) 1
- ANSWER.
Within i hour (per T.S. 3.3.1) determine Shutdown Margin, (and every 12 hours therafter until at least one channel is restored.)- 3 (1.0)
- REFERENCE f j DNI-16 page 3.
l L i a 4 j l I _ _ _ _ _ _ _ _ _ _ _ _ _ __ _j
f QUESTION 7.15 /# th ,
. 67T O, W '
While increasing power f rom 25% to 30% an annuciator on panel l C-17 alarms; " Hydrogen and Stator Cooling Trouble". No Turbine j
- runback occurs, and stator cooling flow is 400 gpm and f alling. j f
What immediate actions must the operators take to verify a i runback is necessary? (1.0) j
- ANSWER Verif y a runback is called f or by checking.
Load > 23% and:
- a. Stator inlet pressure is low. (0.33) j or '
- b. Stator outlet temperature is high. (0.33) or
- c. Stator flow is low. (0.33)
- REFERENCE 01-29 page 3.
END OF SECTION 7 i GO ON TO SECTION 8
i
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SECTION B Administrative Procedures, ! Conditions, and Limitations .
~
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QUESTION 8.01 In accordance with the Trojan Technical Specifications
- a. What is the Minimum Temperature for Criticality? (0.5) l l b. What action must be taken in Mode 1 if the temperature is less than the Minimum Temperature for Criticality?
(1.0)
- ANSWER
- a. 551 F (NO tolerance allowed) (0.5)
. I
- b. Restore to within limit in 15 minutes or be in HOT STANDBY in the next 15 minutes. (1.0) 1
- REFERENCE TS para 3.1.1.5 i
i i l l l
I
-e. 4 QUESTION G.02 Regarding Refueling Operations:
- a. How should the RHR system be operated to ensure that a single f ailure will not cause a complete loss of RHR l
when the water level above the Reactor vessel flange is less than 23Lfeet? (1.0).
- b. What is the basis f or not allowing the movement of f uel in the Reector vessel.unti1~the Reactor has been subtritical for at least 100 hours? (1.0)
- ANSWER
- a. Both trains of RHR must be in operation. (1.0)
- b. This ensures sufficient time has elapsed'to decay off the, short lived fission products. (1.0)
- REFERENCE
- TS B-3.9.3 and B-3.9.8
c___ d a l QUESTION B.03 l l On April 28 during day shift at 0800, your crew oiscovered monthly surveillance due last Monday at 0200, April 20 mid shift (2300 - 0700), for the "A" train Containment Spray System was not performed. This surveillance has been perf ormed on time f or the past six months. The Containment Spray System has been declared inoperable. l
- a. When was the LATEST time this surveillance should have !
l been performed? (Show your work.) (1.0) i
- b. If the surveillance is perf ormed now, 0800 April 28, when will the next surveillance be due at the LATEST?
(Show your work.) (2.0) (A Calendar is provided on the next page.) ( l
- ANSWER
- a. Surveillance intervals f or a monthly surveillance are not to exceed 7.75 days (25*/. of 31 days). (0.5)
- Therefore Monday April 27 at 2000 (7. days 18 hours from j April 20 at 0200). (0.5) i (Therefore 12 hours overdue.) !
l l b. Assuming the prior 2 were on time the next would be dues l 3.25 X 31 = 100.75 days or; 38.75 days (38 days 18 hours) from 0200 April 20 (1.0) l April 20 at O200 + 38 days 18 hours = May 5 at 2000 (1.0) l (Interpretation of surveillance requirement of 3.25 times interval precludes later time or calcualtion f rom April 28 at 0800.)
- REFERENCE TS Definitions, Surveillance Interval 1
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QUESTION 8.04 Which of the following events require notification of the NRC within 1 hour 7 (1.5) (Consider one event at a time.) .
- a. During eddy current testing on the "A" Steam Generator. A Health Physics Technician reports that an individual working inside the S/G received a skin exposure of thirty-five (35) Rem.
- b. The Plant Chemistry Supervisor informs the Control Room that both Boric Acid Storage Tanks have a baron concentration of B.6%.
- c. Primary system leakage f or the last f our (4) hours has been verified to be greater than 1.5 gallons per minute. The Shift Supervisor has declared an " Unusual Event." .
- d. An Instrument Control Technician working in the Reactor Protection System racks has accidentally shorted out several terminals. This error resulted in a Reactor
- trip and will keep the f acility shutdown f or at least. 2 weeks.
- e. A worker inside containment has f allen and sustained a life threatening injury. The decision has been made to immediately transport him to St. Johns Hospital in Longview.
- f. The Public Affairs office at corporate headquarters in Portland is planning to release a News Release concerning a discharge of Chlorine into the Columbia River from Trojan's outfall.
- ANSWER
- b. 1 hour (50.72)
- c. 1 hour (accept Immediate Notification) (50.72)
- d. 1 hour (accept Immediate Notification) (20.403 facility shutdown for more than 1 week.)
l l (0.5 each) j
- REFERENCE 10CFR50.72 and 10CFR20.403 l
l l
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~ QUESTION 8.05 l
During operation at 100% power the plant has the East Diesel Generator.out of service after the failure of a voltage ,
, regul ator . After testing the West Diesel Generator and aligning j
it for operation an aircraft crashes through the transmission ;
=
I towers at the top of the hill severing all the transmission lines. The voltage perturbations cause several fuses in the annuciator panels to burn out, the operators call for assitance from Intrument and Controls technicians who appear in 5 minutes. j The operators determine that the trip has aggravated a Steam ', Generator tube leak but the running charging pump is keeping up with the flow rate. l 1
- a. Using the.following figure 1-14, what Emergency Action j Level is applicable to the event described above? (1.0)
- b. If in addition the West Diesel Generator trips after 15 minutes, what Emergency Action Level is applicable? l (1.0)
-
- ANSWER l a.
A le k r Cit; ^ r- rr;eary (1.0) , 1
- b. Site Area Emergency (1.0) j
- REFERENCE EP-1 figure 1-14.
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. 1 QUESTION.B.06 Regarding EP-1 " Emergency Classification and Initial Protective Action Recommendation":
- a. To what position can the authority to reccomend Protective Actions be delegated by the Emergency Coordinator? (0.5)
(
- b. When the rite is being evacuated where should personnel go to be being monitored f or contamination? (0.5)
- c. What are the'two (2) onsite protective actions.that need to be taken.if the Whole Body Dose Rate.is 0.05 i
Rem /HR7 (1.0)
- d. What are the two (2) onsite protective actions that need to.be taken if the Iodine 131 airborne 4 concentration is 25 MPC in the outside.f acade area? -
(1.0)
-
- ANSWER 4
- a. No delegation of this authority allowed. (Can be passed on to subsequent Emergency Coordinators.) (0.5) 1
)
J
- b. Visitors Information Center (0.5) j
- c. Evacuate nonessential personnel. (0. 5) .
Restrict access. (0.5)
- d. Evacuate all personnel not engaged in actions to protect I facility. (0.5)
Consider use of respirtory protection or KI. (0.5) I
- REFERENCE EP-1 pages 1, and 57.
.l
. , l' o .
1 QUESTION 8.07 By definition, Containment Integrity exists when'all valve and ' electrical penetrations required to be tiosed'will be closed or are closed during accident conditions. This includes automatic j isolation valves, maunual: valves, blind flanges, and automatic
. vlaves secured' in their closed positions. -
l What other three conditions must be met to ensure Containment } Integrity? (2.0)
- ANSWER- I
- 1. Equipment hatch is closed. (0.66) j
- 2. Both Airlocks are operable. (0.66)
- 3. The containment leak rates are within the limits of the -
Tech. Specs. (0.66) 4 1
- REFERENCE Tech. Specs. Definitions page 1-2.
l 1
QUESTION 8.08 Technical Specification 3.4.6.1 specifies that three Reactor Coolant System leakage detection systems must be operable. These three include the containment atmosphere particulate j radioactivity monitoring system and the containment sump inventory measuring system. Two other systems.are interchangeable in meeting the requirement that three systems remain operable. l What are the other two interchangeable systems 7 (1.0)
'* ANSWER
- 1. Containment air cooler condcnsate inventory system. (0.5)
- 2. Containment atmosphere gaseous radioactivity monitoring system. (0.5) _
- REFERENCE T. S. 3.4.6.1
1
.1 5 e ,
]
l QUESTION B.09 ! Answer the following questions'rngarding refueling operations: I
- a. What is the Technical Specifications basis _for maintaining a Shutdown I.targin of 0.95 Keff7J (0.5) -
- b. What four (4) personnel, other than the Shift Supervisor, must be present per procedure FHP-5-1
" Refueling Organization" when fuel is being moved in
-the reactor vessel? (1.5)
- ANSWER To assure subcriticality during refueling operations. (0.5) f a.
- b. 1 SRO - Refueling Supervisor 2 - (One on refueling deck with-SRO and one in Fuel 1 Ope S Building.)- -
Engineering Advisor
- REFERENCE
- T. S. 3.9.1 and 3.9.2
QUESTION 8.10 ] In some situations, during a ' radiological emergency as described in EP-3, certain individuals are allowed to receive doses greater than the normal exposure limits. , I What are the whole body exposure limits f or the f ollowing activities 7
- a. Members of survey teams and those assisting in personnel decontamination. (0.5)
- b. Life saving actions.such as the removal of an injured person and provi dir.g first aid, ambulance service, or-medical treatment. (0.5) i
- c. Performing corrective actions to prevent substantial loss of property ora fire fighting. (0.5)
- ANSWER 1
- a. 3 Rem (0.5)
- b. 75 Rem (0.5)
- c. 25 Rem (0.5)
- REFERENCE EP-3 page 6, table 3-1.
l l' l
~ - - - - - _ _ _ _ _ _ _ _ _
QUESTION 8.11 1 Administrative Order AO-3-1 specifies the minimt.m shift l complement f or operations in MODES 1 through 4 (8 to 11 persons depending on qualifications) . Four of the nine specified positions require NRC Operator Licenses. Some positions-require 2 ! persons to f ulfill the complement. f I What are the four NRC .icensed positions that must be j a. filled when in MODE 17 (1.0) I
- b. What are the two positions that require at least twe persons on duty? (1.0) i i
- c. Who can serve as the Fire Brigade Leader in MODE 5?
(0.5) l d. Mcw many persons are required for the Fire Brigade per f l the Technical Specifications? (0.5) l ' 1
- ANSWER 1
Shift Supervisor (0.25) I n. Assistant Shift Supervisor (0.25) Control Operator (0.25) l Assistant Contal Operator (0.25)
- b. Auxiliary Operator (0.5)
Operators trained aa Fire Brigade memebers (0,5) . l l c. Assistant Shift Supervisor or a l icensed operator trained as ! Fire Brigade leader. (0.5) I
- d. 5 (0.5) t
- REFERENCE AO-3-1 page 2.
l I i I i l l
- mq i
i
- I '
- q. ; : , n; i
1
-QUESTION 9.12
~
Administrative Order AD-4-2 .provides - guidance f or deviations from plant procedures. _
'l 1
- a. Who is authorized'to approve deviations to plant procedures 7,(1,0) i
- b. Assuming a proper authorization and a non-emergency.
situation what condition must be met to. allow changing I a procedure? (0.5)
- c. What circumstances would allow a deviation from' plant: ' .
]
procedures when a . license condition would be violated? (1.0) s I ,
- ANSWER ,
l
- a. Two members of ,the pl ant management staf f , at least one.of _
1 whom holds a Senior Reactor Operator 's license. (1.0) l
/
- b. No change is made to the procedures intent. (0.5) l c. If the action was required to protect the health and safety 1 of the public. (1.0) l
- REFERENCE AO-4-2 pages 2 and 5. ,
L h END OF SECTION O ! END OF EXAMINATION l
.l l
t o
r~ . U.S. Nuclear Regulatory Commission Reactor Operator License Examination w Gwa - Facility: Troian Reactor Type: Westinghou're ( Date Administered: April 28, 1987 Examiner: T. Guifo11/ G. Johnston Candidate: INSTRUCTIONS TO CANDIDATE Use separate' paper for the answers. Write answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% of Categocy % of Candidate's Category Value Total Score Value Cateoory 25.0 25 * ~~~~~~ 1. Principles of Nuclear Power Plant Operation, Thermodynamics.
Heat Transfer and Fluid Flow 25.0 25
- 2. Plant Design Including Safety and Emergency Systems -
25.0 25
- 3. Instruments and Controls 25.0 25 4. Procedures - Normal, Abnormal.
Emergency, and Radiological Control 100.0 TOTALS Final Grade %
. All work done on this examination is my okn. I have neither given nor received aid.
Candidate's Signature I 4 t
r , , _ , ,, ok ,e =
. NRC RULES AND CUIDELINES FOR LICENSE EXAMINATIONS During the administration of this' examination the following rules apply:
- 1. Cheatin on the . examination means an automatic denial of your applicatiorY i and cougd result in more severe penalties. ,
- 2. Restroom trips are to be limited and only one candidate at a time may.
leave. You must avoid all contacts with anyone outside the examination , room to avoid even the appearance or popibility of cheating.
- 3. Use black ink or dark pencil Srh to faellitate legible reproductions. I
~
- 4. Print your name in the blank provided on the cover sheet of the examination.
y *
- 5. Fill in the data on the cover sheet of the examination (if necessary).
- 5. ~ Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category
- as appropriate, the paper, and start each write categor{
"Last Page on thTTast answer sheet.on a new page, write;only o S. Number each answer as to category isnd number, for exemple,1.4, 6.3.
- 10. Skip at least three Ifnes between each answer. j
- 11. Separate answer sheets free pad and place finished answer sheets face down on your desk or table. 4
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value'for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer requir,ed.
- 14. Show all calculations, methods, or assumptions used to obtain an answer
, to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK. . ;
- 16. If parts of the examination are not clear as to intent, ask iluestions of )
the examiner only. ,
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assittance in completing the examination. This must be done after the examination has been completed. .
Examiner standards .s
c q l ci e ' . j o l
- 13. tihen you complete your examination, you shall: l ~ '
- a. Assemble your examination as follows:
a
~
. (1) Exam questions on top. .
. l (2) Exam aids - figures, tables, etc. .
(3) Answer pages including figures which are a part of the answer.
- b. Turn'in your copy of the examination and all pages used to answer .
the examination questions. ,
- c. Turn in all scrap paper and the balance of the paper that you did -
not use for answering the questions.
- d. Leave the examination area, as deffn6d by the examiner. If after .. ,
leaving you are found in this area while the examination is still ' in progr,ess, your ifcense may be denied or revokad. -. l Examiner standards
r s n/
- 1
- EOUATION SHEET I i
l N. AW El
- 2 = Constant -
g,g a = OAv P " 3.1 [X 102o
,_ ,,1o(s m c) O=v(AP)+AE2+XI.all 2c8 Ec ,
1
~
**(t/T) '
O=v(AP)+e.f.2+gJ(az,1+ht Keff C C l 1 - p Keff - 1 1 0 = v(aP) + AN + M + hL + hp $ M , 28c Ec Reff -1 l
#~ Keff
, Re = DvD Wie p- U+ T W ]
1 + 5T hL.fLb D 2gc
#* f f ~ "
T-Ip ht.kt i+ l
**C sa - y .
" D ht . ks J _
e " M8o , 2ge CR o Wp = m (ah) = m (au + v(AP)) : av (AP) I ~ CR g PHP = Wp, ideal = mv(aP) H- 1 CR, ggp,{I,real=$v{aP)* p 1 M" ' "P 1 - Keff np . _FHP BHP CR 3 (1 - Keffs ) - CRg (1 - Keffs ) . nm . BHP FE - ME2 3 Q'p = BHP - FHP KE - 1/2 m,v2 . g, 2g,J t = "IAEI ( b -])
*P3 "P
l
.t.: , ,; .
REQUIREMENTS FOR ADMINISTRATION OF WR11 TEN EXAMINATIONS
- 1. A single room shall be provided for completing the written examina- .
tion. The location of this room and supporting restroom facilities 4 l shall be such as to prevent contact with all other facility and/or i contractor personnel during the duration of the written examination. l If necessary, the facility should make arrangements for the use of a suitable room at a local school, motel, or other building. 0b-taining this room is the responsibility of the licensee. -
- 2. Miriisua spacing is reqvfred to ensure examination integrity as detersined by the chief examiner. Mintaus spacing should be one candidate per table, with a 3-ft space between tables. No wall
. charts, models, and/or other training materials shall be present in -
the examination room.
- 3. Sultable arrangements shall be made by the facility if the candi-dates are to have lunch, coffee or other refreshments. These arrangements shall comply with Item 1 above. These arrangements
- shall be reviewed by the examiner and/or proctor.
- 4. The facility staff shall be provided a copy of the written examination and answer key after the last candidate has completed and handed in
- his written examination. The facility staff shall then have five working i days to provide formal written consents with supporting documentation on l the examination and answer key to the chief examiner er to the regional I office section chief. .
1 (
- 5. The Itcensee 'shall provide pads of 8-1/2 by "Il in. lined paper in 1
- enopened packages for each candidate's use in completing the exam-ination. The examiner shall distribute these pads to the candidates.
All reference material needed to complete the examination shall be furnished by the examiner. Candidates can bring pens calculators, or slide rules into the examination room, pencils,
, and no other equipment er reference material shall be allowed.
- 5. Only black ink er dark panells should be used for writ,ing answers to questions. l Examiner Standards
r e,' -. h N a 0 (Flow) N2 a AP (Head) N3 a BHP (Power) - t hSAT - vPSAT a hr > hSAT CONSTANTS & CONVERSIONS NPSHA - h 7 (Suction)-hsgT A 6.024 X 102: atoms / mole 2 1 watt - 3.1 X 1018 fissions /second NPSHA - vP2 + ".l -h SAT 2&c 1 Curie - 3.7 X 1028 disintegrations per second L 1 amu - 931 Mev ' NPSHA - vP1 + b-hL -hSAT
&c g - 32 ft/sec2 hgt - hpv + hyg Ec -
32 ft-lbm/sec2 -lbf J - 778 f" ~ Ihf hp<hEL BTU
'F - (9/5'C) + 32 4
*C - (*F-32) 5/9 TDH - 3AI (# cold - phot) 'R 'F + 460 Ec {
'K = *C + 273 . i b a AT !2 a Q !3 c - 1.0 BTU /lbm *F @ STP p -
s *
#H,0 - 62.4 lbm/ft8-1 Em/ cms @ STP
. I - I,e*#* - I,10'I*M)
Ind - I,d,
.ld - In dj
/a R/hr = 0. 2 (meters)
R/hr - 6 CE/d2 (feet) '
~
o
SECTION 1 Principles of Nuclear Power Plant Operation, Thermodynamics, Heat Transfer, and Fluid Flow. QUESTION 1.01
- How will each of the f ollowing af f ect (INCREASE, DECREASE, or have NO EFFECT) on secondary plant efficiency?
a.; Circulating water temperature increases. (0.5)
- b. RCS Tave program is raised by 5 F. (0.5)
- ANSWER
- a. DECREASE (O.5) b.- INCREASE ~(0.5)
- REFERENCE H-10-LP, THERMODYNAMICS. FUNDAMENTALS, P. 59 AND 60 P
1
- {
l l QUESTION 1.02
' The equilibrium count rate in.a subcritical reactor increases by a f actor of ~ FOUR f rom 50 cps to 2OO cps due to a withdrawal of .f control rods, how will the margin to.criti'cality change? (1.0)
- ANSWER Margin to criticality decreases by.'3/4 (1.0).
(Doubling. counts = half way,.50 x 2 = 100, 100 x 2 =l200
.therefore 1x 1/2 = 1/2, 1/2 x11/2 = 1/4, have gone 3/4 of way.to j criticality.)
- REFERENCE H-04-LP, NEUTRON SOURCES AND.SUBCRITICAL MULTIPLICATION, P. 18 W '
)
,1 l
'j l
_____________=_ _______w
e QUESTION 1.03 . SELECT THE BEST ANSWER: The Reactor Protection System would become" unreliable f or DNB protection from the OT delta-T trip if excessive voids were allowed to f orm in the Reactor Coolant System because: (1.0)
- a. The heat transfer coefficient of the cladding is ,
J reduced significantly.
- b. Reactor power is no longer proportional to delta T as -
measured by the loop RTDs. q 1
- c. Thermocouple respond much slower in voids as compared q to a subcooled fluid. l l
i
- d. Voids in the reactor vessel head region will cause the I NI detectors to indicate in error.
- ANSWER b (1.0) 1
- REFERENCE H-08-LP, ACCIDENT-ANALYSIS, P. 6 l
l. e I 4
- - - _ - - -____-__a
QUESTION 1'.04 Af ter a secondary calorimetric and adjustment of the power range instruments, it is discovered that the recorded values' of main f eed flow were higher than actual . . WH Y is the new indicated power higher than the actual power? (1.0)
- ANSWER This is because the higher value of f eed flow used in the calorimetric calculation yielded a higher than actual heat transfer rate across the steam generator U-tubes. . ( 1' . 0)
(Therefore, since NIs.were adjusted to indicate this higher power level, they are indicating too high.)
- j. '* REFERENCE Steam Tables and Mollier chart e
QUESTION 1.05 How does INTEGRAL ROD WORTH vary from:
- a. BOL to EOL (0.5) .
- b. HZP to HFP (0,5)
- ANSWER
- 1. IRW increases (0.5)
- 2. IRW increases (0.5)
- REFERENCE H-06-LP, PWR PHYSICS, P. 12 and 14 i
l e
o GUESTION 1.06 For what two (2) reasons are the control. rods withdrawn 'and inserted.in an overlapping program? (1.0) .
- ANSWER
{ Maintain a more unif orm DRW by ensuring rod withdrawal in the more active region exists at.all times. (0.5) i Maintain a more unif orm axial neutron flux distribution during j
. control rod maneuvers. (0.5) ]
)
- REFERENCE H-06-LP, PWR PHYSICS, P. 12 and 14 j i
. I l I l e l I 1 i
QUESTION 1.07
- Assume that you have established a 0.8 DPM SUR on the reactor while the source range meters indicate 1000 cpm on both channels.
' After 30 seconds, what will be the indicated source range neutron
~
level? SHOW ALL WORK! (1.5)
- ANSWER P (f ) = P(i)
- 10expSUR (t ) (0.75)
[ P (f )- = 1000
- 10exp (0.8) (.5) ' (0. 5 )
P(f) = 1000
- 10exp.4 l P (f ). = 1000
- 2.511886 P(f) = 2512 epm (2500 - 2520 is acceptable) (0.25)
- REFERENCE H-05-LP, REACTOR KINETICS, P. 14 P
QUESTION 1.08 The reactor is operating at 807. power. Consider each case INDEPENDENTLY and assume no other parameters vary. Assume the reactor does not trip. WHY.will the Moderate- Temperature Coefficient vary, as l I described, f or the f e llowing conditions? ' I a. RCS Tave INCREASES by 20 F making MTC.more negative. (1.0)
- b. Control rcas are INSERTED 50 steps making MTC more neagative. (1.0)
- c. .RCS boron concentration is INCREASES by 50 ppm making MTC less negative. (1.0)
- ANSWER
- a. At higher Tave moderator density changes a larger amount per degree than at lower Tave. (1.0)
.O
- b. With rods INSERTED further into the core, the effective' core size decreases and there will be more leakage from the core.
(1.0)
- c. As RCS boron concentration INCREASES, thermal utilization DECREASES which causes MTC to decrease (become less negative). (1.0)
- REFERENCE
- H-06-LP,PWR PHYSICS, P. 21 and 22 f
1 QUESTION 1.09 A CALCULATED Estimated Critical Rod Position (ECP) is performed for a startup to be commenced 6 hours after a trip from 100% power. . Why will the Actual Critical Position (ACP), for each case below, change from the Estimated Critical Position (ECP)? (Consider each INDEPENDENTLY.)
- a. The startup is delayed until 12 hours after the trip causing the ACP to be lower than the ECP. (0.5)
- b. The condenser steam dump pressure controller setpoint is increased to just below the SG atmospheric steam dump controller setpoint causing the ACP to be higher-than the ECP. (0.5)
- ANSWER
- a. -(Xenon will peak at B hours after trip and then decrease.)
At 12 hours after the trip the decaying Xenon will add reactivity-(0.5) O
- b. The corresponding temperature increase (and associated negative reactivity) must be compensated for by a higher critical rod position. (0.5)
- REFERENCE GOI-8, ESTIMATED CRITICAL POSITION, APP. A AND TABLE 1
QUESTION 1.10 The reactor is in MODE 3 f ollowing a reactor trip from 100% power after being at full power for a month. The single most reactive control rod did not trip into the core.and is f ully out. What
~
sources of negative reactivity are to be Considered in determining the Shutdown Margin? (1.5) l
- ANSWER Boron. (0.5)
The rods that did trip in. (0.5) renon. (0.5)
- REFERENCE DI-11-8, SHUTDOWN MARGIN, P. 1 and 2 l
I 1 i l l l I i
QUESTION 1.11 The.retctor will exhibit different responses th'oughout r core life due to burnup. Comparing the End of Life core with the Beginning of Life core: ,
- a. Why will the reactor have a higher steady state Startup Rate at End of Life versus Beginning of' Life? (1.0)
- b. If at 50 percent power, a control'rodLworth,-150 pcm-drops into the core (assuming no operator actions, no runback occurs, and the reactor does NOT trip), why will the EOL core have a hLgher steady. state.Tave than 3 a BOL corewhen the plant steadies out at 50%7 (1.0)
- ANSWER l
- a. Below the point of adding heat the SUR is affected most by the delayed neutron fraction, Beff is lower at EOL due to the buildup of Pu237, and the burnup of U235. (1.0) l
- b. At EOL a more negative MTC prevails, therefore it takes a l
- less appreciable drop in RCS temperature to add the reactivity to balance the dropped rod.' (1.0)
- REFERENCE H-05-LP, REACTOR KINETICS, P. 7 and B H-06-LP,PWR PHYSICS, P.-21 l
1
~
1 { i
L QUESTION 1.12 The reactor is producing 100% rated thermal power at-a core j delta-T of 60 degrees and a RCS mass. flow rate of 100% when a q _ station blackout occurs. Na' ur al circulation is established and core delta-T goes to 28 F. If decay heat is 2% rated thermal power,-what.is the core mass flow rate in percent? (SHOW ALL WORK.) (2.0) I
- ANSWER To determine flow in RCS: f Q = m cp delta-T => 100 = 100
- cp
- 60 (1.0)
Q1=b1cpdelta-T=> 2=bi*cp*28 100
- cp
- 60 j 100
__m______ . _________________ (o,5) 2 mi
- cp
- 28 ;
)
- 214.3 50 = --~~~~--- mi i 1 mi = 4.29 percent (4.0 - 4.5 acceptable) (0,5)
- REFERENCE l H-07-LP, TRANSIENT ANALYSIS, P. 13 l l
e e i 4
QUESTION 1.13 The plant is initially operating at 85 percent power with all control systems in AUTOMATIC (unless stated otherwise) and Bank D q control rods at 150~ steps. Consider steady state power. operation unless stated otherwise. Consider each change INDEPENDENTLY and assume the reactor.does NOT trip. Will the Departure from Nucleate Boiling Ratio (DNBR) INCREASE, - DECREASE, or REMAIN THE SAME f or the f ollowing changes in plant conditions?
- a. All pressurizer heaters are energized with spray valves in manual. (0.5) J
- b. RCP bus frequency increases to 60.5 Hz. (0.5)
- c. Turbine load is increased to 100 percent using RCS boron adjustment to maintain a constant rod height.
(0.5)
- d. The Pressurizer Spray valve opens inadvertently. (0.5)
- ANSWER
- a. INCREASE (0.5 each)
- b. INCREASE
- c. DECREASE
- d. DECREASE
- REFERENCE H-13-LP, DNB & HCF, Section 2.7.2 l
~
)
i l
QUESTION 1.14 The plant is in Hot Standby with pressurizer pressure at'905 psig. .A pressurizer PORV begins leaking to the pressurizer l relief tank which is at 3 psig. , Using the attached Mollier diagram,
- a. What is the downstream tail pipe' temperature? (0.5)
- b. What is the enthalpy of the fluid entering the PRT7 (0.5)
- c. The fluid entering the PRT is as (SELECT-ONE) j' (0.5)
- 1. Superheated vapor
- 2. Wet vapor
- 3. Saturated vapor
- 4. Subcooled liquid
- ANSWER q O
- a. 300 F (290-310 F acceptable) (0,5)
- b. 1192 BTU /LBM (1190-1195 BTU /LBM acceptable) (0.5)
- c. 1. (0.5)
- REFERENCE )
STEAM TABLES
)
1 1 1 l (
QUESTION 1.15 What would happen to the motor current.of the motor driven-auxiliary feedwater pump if the following occur?
- a. The suction valve is closed. (0.5)
- b. The discharge valve.is shut. (0.5)
- c. The pump experiences cavitation. (0.5)
L
- ANSWER l a. DECREASE ( 0. 5 ) 4-~ O V- k(uc
- 0- #5.
- b. DECREASE (0.5)
- c. FLUCTUATES-(0.5)
I
- REFERENCE H-12-LP o.
1 i 1 1 i i i 4! 1
1 i
- f QUESTION 1.16 One purpose of the Reactor Coolant System chemistry control is to i retard the corrosion of the cladding of the fuel.
- a. How is the Oxygen concentration in the RCS controlled during MODE 57 (0.5)
- b. How is the RCS pH controlled when the plant is in MODE 17 (0.5) l
- c. What is the reason f or limiting the amount of fuel cladding corrosion to a uniform 17% of the thickness of the cladding? (1,0)
- ANSWER i I
- a. By the use of Hydrazine (H 22 0 ). (0.5) 1 I b. By the use of Lithium Hydroxide (Li OH) . (0.5) {
l
- c. This precludes a loss of strength and ductility during j accident conditions. (Thermal Shock / brittle fracture). (1.0) {
+ REFERENCE I
H-06-LP,PWR PHYSICS, P. 36-38 I l l l 1
QUESTION 1.17 Why would a f aster positive reactivity insertion result in f ewer source counts at criticality than a slower insertion of positive reactivity? (1.0) (Same amount of reactivity addition 10 both cases.) _ 1 l
- ANSWER 1 The faster the rate, the lower the source range counts at criticality due to the reduced time for delayed neutron subcritical multiplication. ' (1.0)
- REFERENCE H-04-LP, NEUTRON SOURCES AND SUBCRITICAL MULTIPLICATION, P. 14,15 and 16 End of Section 1 Go on to Section 2 P
1 l l l
SECTION 2 Plant Design ~ Including Saf ety and Emergency Systems
~
QUESTION 2.01 p What cre SIX (6) of the conditions (other than manual or emergency trip) which will cause a Steam Generator Feed Pump to l trip? (1.5) l l j
- ANSWER l
- 1. High exhaust temperature (6 at 0.25 each) 1l
- 2. Dircharge valve off full open seat i
- 3. Pump suction pressure f
- 4. Lube oil pressure 1
- 5. h :p =nd-
- 6. Low vacuum turbine exhaust l
- 7. Saf ety Injection
- 8. S/G Hi-Hi level FWIS
- 9. Respective condensate pump trip
- REFERENCE E-12-LP, FEED AND CONDENSATE, P. 15 AND 16 l
f 11c u l a k a { o v a s p e= d
- ,0 mcd a al o ver y ed
/ /. Th v wbh L% weer
- a. s i
//M t D.' da8e1% n !
l
l l l 4
.l
- 1 l
QUESTION 2.02 ) i What are the normal source (s) of makeup water for the following? ,l l
- a. SI accumulators (0.50) .
i
- b. RWST (0.50) j i
- ANSWER l
- a. RWST (0.50)
- b. PWST (CVCS) (0.50)
- REFERENCE' A-07-LP, EMERGENCY CORE COOLING SYSTEMS, P.16 OI-5-2, SAFETY INJECTION, P. 5 I
l. l l 5 l l
i 1 1 i QUESTION 2.03 ) f Where is the cover gas vented to, when venting the SI J accurnulators to reduce pressure? (0.5) {
' l '
- ANSWER Containment Atmosphere (0,5) 1 i
- REFERENCE A-07-LP, EMERGENCY CORE COOLING SYSTEMS, P.16 OI-5 -2, SAFETY INJECTION, P. 5 l
l .; O .j l e-i l l A l l l
l I l l OUESTION 2.04 f 1 1 Regarding the Waste Gas Decay Tanks:
- a. Other than the inservice tank, what is the normal ;
lineup for the other three Waste Gas Decay Tanks? ! (1.0)
- b. What two automatic actions' occur when th.e inservice {
Waste Gas Decay Tank reaches 100 pnig? (0.5) { I
- ANSWER J
- a. 1. One tank in standby to receive gases. ]
i
- 2. One tank supplying cover gas to CVCS HUTS.
- 3. One tank isolated to allow decay of it's contents.
I (0.33 each)
- b. 1. Inlet valve will close on inservice WGDT (0.25)
- 2. Inlet valve on standby tank will open (0.25)
- REFERENCE G-04-LP, GASEOUS RADIOACTIVE WASTE SYSTEM, P. 10 AND 13
l
-QUESTION 2.05 Regarding the Reactor Coolant Systems
- a. What three (3) piping connections to:the Reactor Coolant System have thermal sleeves? (1.5) -
- b. How does the Hot Leg RTD bypass line connection provide an increased forced flow through.the line? (0.5)
- ANSWER
- a. Charging lines. (0.5)
Pressurizer surce line (0.5) l Pressurizer Spray line (0,5) 1 l b. By a scoop in the hot leg. (0.5)
- REFERENCE A-01-LP
~
- b. ) hw~' e & we' o'n o$ CokMt* Ek --y
- 1 4 /. g 1J,., a ce c.4,e--. q- t6 4,L.rhesa L /M/g-l
QUESTION 2.06 What is'the order in which the ECCS subsystems will injeet into the RCS during a continued RCS depressurization of 100 psig/ minute caused by a LOCA. (Assume each active component was started by a Safety Injection Signal.) (2.0) i
- ANSWER High head injection [(C fth (0.25 f or each in correct order)
Intermediate head injection ($ rig) Accumulators . , Low head injection (Agg)
- REFERENCE A-07-LP, EMERGENCY CORE COOLING SYSTEMS, P. 9-18
QUESTION 2.07 Regarding Pressurizer Spray:
- a. What are TWO reasons for maintaining a minimum.
pressurizer spray line flow during nor<nal "at' power" operations? (1.0)
- b. What are the two (2) annunciators available to alert the operator that minimum spray flow is not being maintained?- (1.0)
- ANSWER
- a. 1) To reduce thermal shock to the spray line and spray nozzle. (0.5)
- 2) To promote mixing f or chemistry control . .(0.5) .
- b. 1) Pressurizer Spray Line La Temperature alarm (0.5)
- 2) Pressurizer Surge Line Lo Temperature alarm (0.5)
- REFERENCE A-03-LP, PRESSURIZER AND PRESSURE RELIEF _ TANK, P. 3,4,AND 6 ;
1 I l i
l QUESTION 2.08 Regarding the Residual Heat Removal Systems
- a. What is the purpose of the interlock that ensures MD-8812 (RWST to RHRS) is closed before MD-8701 and j MD-8702 (RHR suction header isolation) are opened?
(0.5)
- b. What is the. basis for the' sizing of'the RHR suction relief valve PSV-87087 (1.0) ]
- c. What flow rate in the RHR discharge line will cause FCV-610, in the miniflow recirculation line, to open-when RHR flow is being reduced? (0.5)
- ANSWER
- a. Prevents leakage past check valve (CV-8958) to RWST. (0. 5)
- 6. To pass the flow from all three chaging pumps to prevent overpressurization of RHR (and RCS). (1.0)
- c. 500 gpm (0.5)
- REFERENCE A-08-LP page 22 n
l. i
u------,_.-.--_- - - - - - - - _ _ _,.,_,s _ _ _ , _
)
1 i QUESTION 2.09 Regarding the Reactor Coolant Pump Thermal Barriers
- a. What are TWO (2) indications or symptoms whicb will be observed if a tube leak occurs in a RCP Thermal Barrier.
heat' exchanger. (Assume NO alarm.setpoints are reached.) (1.0)
- b. If the tube leak continues to increase in severity,.
WHAT AUTOMATIC action will occur? (0.5)
- ANSWER
- a. 1. Rising CCW surge tank level
- 2. Increasing CCW system radioactivity
- 3. . Increasing CCW heat exchanger inlet temperature (any two at 0.5 each)
- 6. 'The CCW return isolation valve on the affected thermal barrier heat exchanger will shut (0.5).
- REFERENCE A-11-LP,CCW SYSTEM, P. 10,15,16 and 20 l
l d I _--------___-______i,_.______,___,._____
QUESTION 2.10 For each of the f ollowing electrical loads, What 4160 volt bus provides the normal source of electrical power 7
~
- a. Service Water Pump P-108B (0.5)
- b. Safety Injection Pump P-203A (0.5)
- c. Pressurizer Heater L C Bio (0.5)
- d. Residual Heat Removal Pump P-202A (0.5)
- ANSWER
- a. A2
- b. Ai
- c. A6
- d. A1 (0,5 each for a total of 2.0) i
- REFERENCE
. C-04-LP,4.16KV ELECTRICAL DISTRIBUTION,P. 13-19 i
l
1 l 1 I l QUESTION 2.11
- 1 l
Regarding the Auxiliary Feedwater Pumps:
~
- a. What are the Normal and Emergency sources of water to
~
the Auxiliary Feedwater Pumps. (1.0)
- b. What is'the reason for. requiring that only one dr*(fC1 Auxiliary Feedwater Pump be run when the Condensate Storage Tank is between 60% and 9% of level? (1.0) c.- How is the engine for the Diesel Driven Auxiliary Feedwater pump provided cooling. water? (1.0) f i
- ANSWER
- a. 1. Normal -
CSTs (0.50)
- 2. Emergency - Service Water System (0.50) 1
- b. - The potential exists to exceed the NPSH requirements for the pumps. (1.0)
- c. FromtheServiceWatersystemviaa(ServiceWaterbooster pump)totheJacketCoolingwaterheatexchanger. (1.0)
- REFERENCE A-12-LP, AUXILIARY FEEDWATER SYSTEM, P. 5 and 17 EI-O REFERENCE PAGE l
e , 1 l
.3 1 1 l 1 I
l
QUESTION 2.12 i The f ollowing concern the CVCS. .
- a. What are two (2) of the three functions (purposes) of
- the Letdown Pressure Control Valve (PCV-131) While the plant is in MODES 4 and 57 (1.0)
- b. What is the maximum allowed letdown flow leaving the letdown heat exchanger? (0.5)
~
- c. What is the meximum allowed letdown temperature leaving the heat exchanger? (0.5)
- d. What are the INTERLOCKS that must be satisfied in' order to open an orifice isolation valve? (1.0)
- ANSWER-
- a. - Control letdown flow during plant _ heatup while drawing a-bubble in the pressurizer.
- Maintain RCS pressure control when solid.
- Prevent two-phase flow in the letdown heat ex changer ~ and -
. its upstream piping to the orfices.
(any two at 0.50 each)
- b. 127 GPM (0.5) [wdc.tccf ID Yo 00 y
- c. 137 degrees F (0.5) (tviIk AJC#f f l30tol'/O
- d. - Letdown isolation valves must be open. (0.25)
- At least one charging pump must be running. (0.25)
- REFERENCE A-06-LP, CHEMICAL AND VOLUME CONTROL SYSTEM, P. 6,7 and 8 OI-3-5, CHARGING, LETDOWN AND RCP SEAL WATER, P. 2 dA
-nocc.s (0cd
- vz.p. a.uf > /7% (o.s
__.__-__._.._._m._______-m_______-__
( QUESTION 2.13 The Reactor Coolant System is at 200 F and 355 psig..' Cooling . to remove decay heat is being provided. by the. Residual Heat Removal' System.' What is.the normal flow path to the demineralizers'in the CVCS for this mode of opearation? (1.5) i
- ANSWER Letdown "4-(#
-- :/;i cr! c : dowstream of the RHR HXs, to the CVCS letdown HX, (through PCV-131),.to the demineralizers. (1.5)
- REFERENCE A-08-LP l 6
t
> 1 l
f l
~
/
t 4 1 1 I QUESTION 2.14 h The Fire Main System provides a -Fire Loop that f ully encircles j the plant. j
. i
- a. What three fire protection subsystems are supplied from 1 l
l the Fire Loop? (1.0)
- b. What function does the Jockey Pump provide to the Fire Loop? (0.5)
- ANSWER
- a. 1. blater Spray Systems (Deluge) (0.33)
- 2. Gprinkler Systems -(0.33) .
- 3. Fire Hose stations (0.33)
- b. Keeps Fire Loop pressurized. (0.5)
I
- REFERENCE I-04-LP-
- End of Section 2 Go on to Section 3 1
l 1 L l' 1 i 1 l I j
l l SECTION 3 Instruments and Controls QUESTION 3.01 The plant is operating at 80 percent power with all systems in automatic. The CHANNEL 455 (controlling) pressurizer pressure channel FAILS HIGH. What three (3) initial AUTOMATIC actions occur because of the failure? (Alarms are not required, assume no operator action.) { (1.5) l l l
- ANSWER
- 1. Both spray valves open. (0,5)
~b
- 2. All pressurizer heaters turn off. (0.5) yec,idf " ' l
/
. 3. Power operated relief valve PCV-455A will open. (0.5)
- REFERENCE i l
ARG APP. A, P. 2 and 7
)
i l
l 4 QUESTION 3.02 i Which one of the 'f ollowing will result'in the main feedwater . control valves closing immediately? (1.0) l
- a. ' High Main Steam flow i- b. Reactor trip with.Tave <. 564 F l ' c. Containment isolation signal
- d. -Low-low steam generator water' level
-
- ANSWER :
- b. t
- REFERENCE CROCTRM P. 3.9-2
)
I V I
. I
'l j
i
\
QUESTION 3.03 fl gqq The SDS has been rmed as with the Steam Dump Mode Selec' tor switch in due to prior testing of related instrumentation. The plant is operating at 50 percent power with all control ~ systems in automatic.
- a. What are the three (3) permissives required to arm the Steam Dump System (SDS) with the Steam Dump Mode Selector Switch in AUTO? (1.5)
- b. How will the SDS now respond to a LOOP D Tave instrument failure to 583 F7 Assume no reactor trip occurs. (1.0)
- ANSWER
- a. C-7 (5 percent load decrease in < 120 seconds,10 percent step load decrease) (0.5) g M
C-9 (Condenser vacuum > g Hg, At least one circulating water pump Pc !::p cic :d) (0.5) , cu au s % C-8 (4/4 turbine stop valves closed, 2/3 low emergency trip oil pressures) (0.5)
- b. Two groups (6) of SDS valves will trip open (0.5). The other two groups (6) will modulate open as necessary (0.5).
- REFERENCE CROCTRM P. 3.9-5
~
- a. a>:ll cu,g F-a (1-1. L ) ,, ,, 4
,%"gAc,2:e , 6+ . 3 6 y c_.9 4 o wid c ccy k 1.C c-9 i 3
.n,
& neei<c.pq ea cn n
- f. pd4ni,,ae.
p".) p(ped 6.tL &c-q . f
l .
~
l l' QUESTION 3.04 - 1 The plant is operating;at 100 % steady state power with Containment Pressure Channel III failed'high. A technician
- troubleshooting the _ f ailed channel inadvertently de-energizes the instrument power for Containment Pressure' Channel II.
Why will a Containment Spray Signal NOT be actuated? (1.0)
- ANSWER Channel II must energize to actuate' f or an unsaf e condition (to avoid inadvertent spray actuation). (1.0)
- REFERENCE M1T(13)-8, FUNCTIONAL DIAGRAM SAFEGUARDS ACTUATION SIGNALS A-09-LP, CONTAINMENT SPRAY SYSTEM, P. 20 e
l l i
- j
QUESTION 3.05 Regarding the Intermediate Range nuclear instrumentation:
- a. If an IR instrument is undercompensated during a reactor shutdown so that the indication " froze" at 10E-9 amps, what effect would this have ? (1.0)
- b. WHAT operator action (s) is required to continue a reactor shutdown if one IR channel has failed high?
(0.5)
- ANSWER
- a. At > 10E-10 amps will prevent the SR detectors from automatically energizing. (1.0)
- b. The operator must manually energize the SR detectors with the Source Range Manual Reset switches. (0.5)
- REFERENCE i B-07-LP,EXCORE INSTRUMENTATION SYSTEM, P. 11 l - ONI-16,EXCORE NUCLEAR INSTRUMENTATION MALFUNCTION, P. 5 l
l e - l l 1 l I I I l l l l l
)
QUESTIDN 3.06 The plant is operating at 100% power. The normally selected channel (CH-459) to the pressurizer level control system fails low (O%). , 1
- a. What are the four (4) automatic - actions that will be initiated by this channel failure. (2.0)
- b. If no operator action is taken, what will eventually cause a reactor trip? (0.5)
- ANSWER !
- a. letdown isolation valve closes (0.5)
All letdown orifice isolation valves close (0.5) All pressurizer heater groups are turned off (0.5) Posttive displacement charging pump speed increases (0. 5) . or % casAnt v slue o per . g-
- b. Reactor will trip on high pressurizer level (at . /. ) . (0.50)
- REFERENCE B-04-LP, PRESSURIZER PRESSURE AND LEVEL, P. 16 and 17
=
l l 1 l f
- l
QUESTION 3.07 In the Steam Generator Level Control System, why is Steam Pressure used as a compensating signal for modifying the Steam Flow signal? (1.0) ,
- ANSWER Steam pressure is used to compensate the steam flow signal for density variations in the steam (as steam pressure varies). (1.0)
- REFERENCE B-05-LP,FEEDWATER CONTROL, P. 15 l
l 4 i
" l l
I
i i
~
1 QUESTION 3.08 The plant is operating at 50% power with all control systems in automatic. Bank D rods are at 150 steps. > What causes the rods to move IN when the following events occur? )
- a. B steam generator MSIV inadvertently closes (turbine i load constant). (1.0) ]
l
- b. Loop A narrow-range That instrument fails high. (1.0) )
c.. Turbine load is ramped to 20% at 5%/ minute. (1.0) { i (Assume no operator action (unless' stated) and the reactor does j NOT trip. Consider each case separately.) l i
- ANSWER
- a. Loop B Tcold increases to Thot causing loop B Tave to increase above Tref. (1.0) l l
- b. Loop A Tave will become the auctioneered high Tave which is sensed to be higher than Tref. (1.0)
- c. Power mismatch circuit senses turbine power decreasing at a faster rate than nuclear power. (1.0)
- REFERENCE B-02-LP, CONTROL AND PROTECTION LOGIC DIAGRAMS, P.24 and 25 l
l 1 l
)
1 l l i i l I
+ 1 i 1 QUESTION 3.09 The reactor is critical at 5% rated thermal power during a. normal reactor startup. What are six '(6)' of the eight reactor trips
-which are DISABLED in this condition? (3.0) ,
- ANSWER
- 1. Source range reactor trip
- 2. RCP bus undervoltage
- 3. RCP bus underfrequency 4.- RCP. breakers open trip
- 5. , Pressurizer. low pressure 6.- Pressurizer-high level-
- 7. Loss of flow B. Turbine. trip (6 at 0.5 each)
- REFERENCE B-02-LP, CONTROL AND PROTECTION LOGIC DIAGRAMS, P. 6 - 13 i
1 1 1 1 l I i I l l i 1 l 3 f U-_-______-__-___------__ .__. - - - . - _ _ . _ - _
l 1 QUESTION'3.10 ) Which one of the f ollowing WILL result . in a turbine runback. I (1.0)
- a. One OP delta T channel within 2% of op delta i reactor trip setpoint. 6
- b. Impulse pressure equals 65% and ld j MFP trips. (two ,
i running initially)
- c. Stator cooling water pressure at O psig for 10 seconds' -
- pr ;.e.
with 20% Lag;I 4 g..e JAo,es;vr t
- d. Two OT delta T channels within 4% of the OT delta T reactor trip setpoint.
?
i
- ANSWER
- b. WILL (1.0)
- REFERENCE
. CROCTRM P. 3.9-3 1 e r
QUESTION 3.11-The reactor is operating at 100% power when an inadvertent Safety- l Injection signal is generated. All systems respond as expected except that only the Train B reactor trip breaker opens. (Assume no operator. action.) Which one of f ollowing is f alse? (1.0) 1
- a. The Turbine trips because of P-4 permissive vice the Safety Injection Signal.
- b. Feedwater isolation will occur before Tave decreases below 564 F.
- c. Source range instruments-will automatically energize at P-10.
- d. Containment Ventilation System isolates along with Containment Isolation.
i o
- ANSWER
- a. FALSE.(Directly from SI) (1.0)
- c. Fal s < j 4
- REFERENCE B-02-LP, CONTROL AND PROTECTION LOGIC DIAGRAMS, P. 6,9 and 22 i
e l l l l
~1 l
l
'4 QUESTION 3.12 What are the f our Rod Block Signals that cause both manualu and< -l auto rod withdrawal stops? (Setpoints.are not required.) (2.0)'
- ANSWER
- 1. Intermediate Range Hi (C-1) (0.5)
- 2. Power Range Hi (C-2) . (0. 5) ,
4 L 3. OT delta T (C-3) (0.5) i e (0.5)
- 4. OP delta T (C-4) 1
- REFERENCE CROCTRM P. 3.9-3
- l 1
i
- )
i u
\
l l
-i i
QUESTION 3.13
. What are f our .(4) of.the six' components RESET by the Rod' Control Startup Pushbutton? (1.0) .
- ANSWER
' 1. Bank demand step counter
~
- 2. Master cycler
- 3. Slave cycler
- 4. Bank overlap' unit l 5. Resets urgent f ailure alarms (if cleared) ,
- 6. PA converter counter (any;4 at 0.25 each) (1.0)
- REFERENCE B-09-LP, FULL LENGTH ROD CONTROL, P. ' 3B P
9
i l 1' QUESTION 3.14 , l How will the f ollowing components respond (FAIL OPEN, FAIL I CLOSED, AS IS, REMAIN FUNCTIONAL, DIVERTS TO ..., ETC.) when instrument air pressure is. lost with the plant at 100% power?
- a. Letdown pressure control valve (PCV-131) (0.5)
- b. Main steamline A to TAFP (CV-1451) (0.5)
'c. Letdown bypass to hold up tank (LCV 112A) (0.5) d.' Pressurizer spray valves (PCV 455B and 455C) (0.5)
- ANSWER l a. Fail open l
- b. As is
- c. Open to VCT
- d. Fail closed (0.5 each)
- REFERENCE ]
. ONI-51, LOSS OF INSTRUMENT AIR, P. 4,5,10 and 12 l
L i i l a I
1 I QUESTION'3.15~ The Volume Control Tank (VCT) leveliinstrumentation, in' addition; to control board readouts'and' alarms, initiates THREE. unique automatic actions at .various levels in the VCT., , j What' are these three -(3) AUTOMATIC ACTIONS? (1.0)' H
- ANSWER
- 1. ' Divert to holdup tank (at 93%).- (0.33)
- 2. Automatic makeup control (On at.41% Off at 54%). -(0.33)
- 3. Emergency makeup from RWST (at 1.4%). (0.33)
- REFERENCE A-06-LP, CHEMICAL AND VOLUME CONTROL SYSTEM, P. 12 ..-
LPOO79, P.9 l End of Section 3
- Go on to Section 4 j
- ~
l i 1 A i
--__--y SECTION 4 Procedures - Normal, Abnormal, Emergency and Radiological Controls QUESTION 4.01
.What are the FOUR (4) conditions referenced in the Technical Specifications tha ensure th Hot Channel Dactor limits are being '
maintained within limits during normal power operations?
- ANSWER Control Rods within +/- 12 steps of group demand position. (0.5) l Proper sequencing and overlap of rod groups. (0.5) I Control Rod insertion limits are maintained. (0.5)
AFD is maintained within limits. (0.5)
- REFERENCE TECHNICAL SPECIFICATIONS, P. 3/4 2-4 V
QUESTION 4.02 Following a reactor trip from a loss of all AC power, F.C'A O.0 has the operator depressurize the RCS using Steam Generator Power Operated relief valves'to ecol the plant down. ,
- a. What cooldown rate must be attained during the depressurization? (0.5)
- b. Why must the'RCS be depressurized? (0.5)
- c. Why shouldn't the' RCS be depressurized' below 150 psig? -
(0.5)
- d. Why should the operator continue tha cooldown if
. pressurizer level is lost or vessel head voiding occurs? (0.5) l
*ANSWE)
- a. Maximum attainable. (0.5) i
- b. To minimize RCS inventory loss (via the RCP seals). (0.5)
- c. T;s prevent injection of accumulator nitrogen into the RCS.
(0.5) The inventory loss potential is high and overrides this d.
' concern. (0.5) li
- REFERENCE ECA-O.0, LOSS OF ALL AC POWER,P. 16 ,
K-04-Lty,ECA-O GERIES AND FR-C CORE COOLING,P. 5 and 22
- , j
\
l'
,,J !
QUESTION 4.03 What are the six (6) immediate actions that f ollow the seven f below that have to executed when a Saf ety~ Injection occurs in. l accordance with EI-O " Reactor Trip, Safety Injection, and l
.(
Diagnosis"? (3.0)
- 1. Verify Reactor Trip'
- 2. Verif y Turbiree Trip
- 3. Verify Generator Trip 'j
- 4. Verify power-to AC Emergency Busses .
- 5. Verify SI is actuated 1 R
- 6. Verify Containment Isolation, Phase A
- 7. Ver*if y Feedwater Isolation i
l
- ANSWER !
- 1. Verify AFW flow i 2.' Verify ECCS flow l j
- 3. Check Containment Pressure
- 4. Verify RCS heat removal o 5. Check RCP trip criteria 1
- 6. Check CCP Mini-Flow criteria - (0.5 each)
- REFERENCE i l
EI-0, REACTOR TRIP, SAFETY INJECTION, AND DIAGNOSIS, P. 2-9 [ i i I i l i
)
OUESTION 4.04 Regar-ding Radiation Protection procedures: I.
- a. When must a Self Reading Pocket Ionization Chamber (SRPIC) be re-zerced? (0.5)
- b. What TWO additional provisions (above'those required for a high radiation area) are required for an area. .]
where the intensity of radiation is > 1000 mrem /hr? (1.0) l J
- ANSWER
- a. Prior to reaching 3/4 of full scal ~e. (0.50)-
- b. Locked door shall be provided to prevent un' authorized.
entry and the keys shall be. maintained under the administrative control of the Shift Supervisor on duty. (1.0)
*REFEAENCE RADIATION PROTECTION MANUAL, P. 2-10 and 2-26 I
'1 l-1 I
. -i i
1 s I
~!
h _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ - . _ _ _ _ _ . . _ . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ m_ _ _ _ __ _ _ _ . _ _ . _ _ _ . . -
1 1 J QUESTION 4.05 Technical Specification 3.4.1.4 and GOI-1 specify'that a Reactor Coolant Pump shall not be started with ene or more RCS cold leg
~
temperatures less than or equal to 290 F unless one of three conditions-are met. What are those three (3) conditions? (1.5)
- ANSWER i 4
- 1. Another RCP is running. (0.5) ,, ,g,y .
- 2. The.socondary. side temperature of each S/G <gRCS Tcold, and overpressure protection is operable. (0.5)
- 3. A bubble is present in the Pressurizer. (0.5) ;
i
- REFERENCE T.S. 3.4.1.4 and GOI-1. !
O 1
't
QUESTION 4.06 What is the order of priority of the following? (2.0)
- 1. Inventory - Orange Path ,
- 2. Heat Sink - Red Path' -
- 3. Integrity - Red Path
- 4. Subtriticality - Orange Path
- 5. -Containment - Green Path
- 6. Core Cooling - Yellow Path
- 7. Inventory - Red Path B. Integrity - Orange Path
-* ANSWER 2.
3. 7. 4. 8.
- 1. (0.25 points each) 6.
. 5. (Points awarded on basis.of overall order.)
- REFERENCE FR-0, CRITICAL SAFETY FUNCTION STATUS TREES, P.2
{c**s f U66" %CT i t
- l' W f*
1 l
l QUESTION 4.07 Regarding EI-O " Reactor Trip, Safety Injection, and Diagnosis" and no adverse containment:
~
- a. The reference page for the EI-O series lists THREE conditions that TOGETHER require the operator to trip all RCPs. What are these THREE conditions? (1.0)
- b. According to the reference page for the EO-1 series, two parameters must be monitored to determine if Safety Injection must be reinitiated. What are these TWO parameters? (1.0)
- ANSWER I a. 1. SI is actuated (0.33) l 2.'CCP or SI flow is verified (0.33)
- 3. RCS pressure < 1425 psig (0.33)
- b. 1. RCS subcooling 30 F (0.5) 2.
Pressurizer lev ( 1 (c)annot >
. 5'/)be (0. 5)maintained
- REFERENCE EI-O and 1 SERIES PROCEDURES REFERENCE PAGE l
A
u I I QUESTION 4.08 What is the direction or condition that indicates, for e'ach of the following parameters, that natura1' circulation is occuring in accordance with ES-0.2, Natural Circulation Cooldown? (3.0)
- a. RCS T * ""9" cold ,
I
- b. RCS T Wide Range ]
hat I
- c. Steam Generator Pressure <
i
- d. Core Exit Thermccouples
- e. RCS Subcooling i i
- f. Delta T
- ANSWER
- a. Near saturation temperature of S/G steam pressure. (0.5) ,
- b. Stable or decreasing slowly. (0.5) l
- c. Stable or decreasing slowly. (0.5)
. d. Stable or decreasing slowly. (0.5)
- e. > 30 F. (0.5)
- f. < 66 F and stable. (0.5)
- REFERENCE ES-0.2, NATURAL CIRCULATION COOLDOWN, P. 3 e
1 l
- - _ _ _ _-- _ _ _ _a
O a QUESTION 4.09 What are the f ollowing exposure limits , (non-emergency) for a 25
-year-old male licensed operstor7 (1.0)
- 1. Maximum ADMINISTRATIVE QUARTERLY whole body limit
- 2. Maximum 10 CFR 20 QUARTERLY whole body limit
- 3. Maximum ADMINISTRATIVE YEARLYE whole body limit
- 4. Maximum 10 CFR 20 LIFETIME whole body limit
- ANSWER
- 1. 2500 mrem 1
- 2. 3000 mram (0.25 each for a total of 1.0)' )
- 3. 4500 mrem .
- 4. 35000 mr em (5 x (N - 18) = 5 x 7 = 35 rem)
- REFERENCE j RADIATION PROTECTION MANUAL, P. 2-2 and 2-3 l
l 1 l
l
. j QUESTION 4.10 What INDIVIDUAL may authorize an extension of an ADMINIS'TRATIVE whole body limit f or a licensed operator in each of the following j 1
cases? .
- a. Exposures up to 1000 mrem per quarter. (0.5) j
- b. Exposures in excess of 1.0 Rem in a quarter. (0.5)
- ANSWER
- 1. Radiation Protection Supervisor up to 1000 mrem /qtr. (0.5)
- 2. General Manager in excess of 1.0 rem /qtr. (0.5)
- REFERENCE RADIATION PROTECTION MANUAL, P. 2-2 and 2-3 l
e I
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i l 1 1 OUESTION 4.11 Refer to figure 4-1 f or the f ollowing question. From the indications given below, WHY is the axial flux difference being j maintained within the appropriate Technical Specification limits? (1.0)
'AFD AFD AFD' AFD POWER LEVEL CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL 4 87% -10 Out of service -15 -8
- ANSWER Less than two channels exceed limits (2 out of 3 operable). (1.0)
- REFERENCE Technical Specifications, P. 3/4 2-1 and 2-2 V
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- OPER AT.I.C.IN .
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0 ~~ 30 40 30 20 10 0 10 20 30 40 50 FLUX DIFFERENCE (al) % AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER TROJAN-UNIT 1 3/4 2-4 w ___ _ _ _ _ _. -__ _
e 4 QUESTION 4.12 During a Natural Circulation cooldown using ES-0.2, RCS subcooling is limited to > 50 F. What is the REASON for this limit? (1.0) ,
- ANSWER To prevent voiding in the reactor vessel upper head. (1.0)
- REFERENCE ES-0.2 NATURAL CIRCULATION COOLDOWN, P. O K-04-LP,EI-O SERIES, ES-0.0, ES-0.2, ES-0.3, P. 29 - 31
\ I i i l l I l 4
^
t i 1 i < _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ . _ _ _ . _ _ . _ __ )
i l I OUESTION 4.13 What are.four (4) of the six situations or conditions (unrelated) ; which require the operator to commence EMERGENCY BORATION of the RCS in accordance with ONI-10 " Emergency Boration"? (1.0) ;
- ANSWER l
- 1. One or more rod position indicators fail to indicate the rods fully inserted after a plant shutdown.
- 2. Excessive control bank insertion. '
1
- 3. Unexplained or uncontrolled reactivity increase.
- 4. Insufficient shutdown margin.
- 5. Uncontrolled cooldown following a reactor trip.
- 6. ATWS.
(any four at 0.25 each) j l
- REFERENCE ONI-10, EMERGENCY BORATION,P. 1 and 2 4 1
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= ,
J l QUCSTION 4.14 l During plant shutdown from hot standby to cold shutdown per GOI-4 RCS boron concentration must be changed and only one RCP is in ] operation. .
- a. How must the spray valves be aligned? (1.0)
- b. Prior to reducing RCS cold leg to less than or equal to 290 F what TWO items must be ensured operable to meet Technical Specification limits? (1.0)- ,
- ANSWER
- a. Place the spray valve in the line supplied from the idle loop in NORMAL and CLOSE the valve. (1.0) 4
- b. One SI. pump is operable (0.5) and over pressure protection systems are operable. ( 0. 5 ) '
- REFERENCE GOI-4, P. 3 and 5 TECHNICAL SPECIFICATIONS 3.5.3.2 and 3.4.9.3 .
f I
/
t QUESTION 4.15 [A (76 )'l g.g- > The following questions refer to the current Standing Orders: j
- a. In addition to the normal Control Room personnel who must be in the Control Room during a turbine roll?
(0.33)
- b. When both condensate pumps are lost why must both heater drain pumps be tripped? (0.33) l
- c. If a motor trips unexpectedly what two things must be
'done prior to restarting the motor 7 (0.33)
- ANSWER
- a. Duty Plant Manager or Duty Operations Supervisor (0.33)
- b. To eliminate the possibility of resin back leakage into the hot well (0.33)
- c. Investigate the cause and notify an electrician (0.33) -
- REFERENCE STANDING ORDERS, NO. 6, 9 and 11 _
1 End of Section 4 End of Examination I l l i l}}