ML20238A104
| ML20238A104 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/13/1987 |
| From: | Burdick T, Hare E, Lanksbury R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20238A067 | List: |
| References | |
| 50-331-OLS-87, 50-331-OLS-87-0, NUDOCS 8708200371 | |
| Download: ML20238A104 (91) | |
Text
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.i-U.S. NUCLEAR REGULATORY COMMISSION REGION III i
Report No. 50-331/0LS-87-01(DRS)
Docket No. 50-331 Licenses No. DPR-49 Licensee:
Iowa Electric Light and Power Company I.E. Towers P. O. Box 351 Cedar Rapids, IA 52406 i
Facility Name:
Duane Arnold Energy Center Examination Administered At:
Duane Arnold Energy Center Examination Conductp4L July 15-16,1987' 1
i Datelnln Examiners: Q Lhriksbur 03 a
Ob t, /& 7
.. rc Date' Approved By:
dick, Chief 8h/D Operating Licensing Section Date Examination Summary Examination administered on July 15-16, 1987 (Report No. 50-331/0LS-87-01(DRS))
Written and oral examinations were administered to two Senior Reactor Operator (SRO) candidates.
Results:
Both SR0 candidates passed the written and oral examinations.
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- 1 8708200371 870818 PDR ADOCK 05000331 V
p REPORT DETAILS i
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Examiners t
R. D. Lanksbury, Chief Examiner E. A. Hare 2.
Exit Meeting At the conclusion of the examinations, an exit meeting was conducted.
The following personnel attended this exit meeting.
Facility Representatives C.-R. Mick, Operations Supervisor D. Wilson, Training Superintendent M. Meyer, Senior Instructor l
R. J. Bucker, Acting Training Supervisor, Operations NRC Representatives R. D. Lanksbury, Operating. Licensing Examiner E. A. Hare, 0perating Licensing Examiner J. S. Wiebe, Duane Arnold Senior Resident Inspector The following items were requested from the licensee:
a.
A copy of Technical Specifications Interpretations, b.
Ensure complete and up to date material is sent to examiners, especially Technical Specifications.
Reference material should not be sent too early (i.e., significantly before the required 60 days,unlesssorequested).
During the scenario walkthrough, the examiner noted a discrepancy in referenced Operating Instructions (01) procedure numbers in the Emergency Operating Procedure (EOP) and the currently used 01 numbers. The E0Ps referenced the.old two digit procedure numbering system. A new operating instruction numbering' system was implemented which changed the number system to three or four digit codes. The E0P's have not been updated to reference the proper 01 number. The examiner requested the facility L
to review their E0P and correct all reference material.
This item was turned over to the Resident Inspectors for followup, i
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Examination Review l
Specific facility comments concerning written examination questions,
'c followed by the NRC responses, are enumerated in the attachment.
The examiners noted that significant portions of the. facility supplied reference material were out of date at the time of the examination.
This material ranged. from the' Technical Specifications themselves-(which were outdated by several amendments) to the Operating Instructions-(which in
'one case was outdated by.two' revisions).- Investigation.into.the matter l
revealed that the changes had been issued as must as four months prior l
to'the~ examination.
The facility was requested to ensure that in the future reference materials were up to date.
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e ATTACHMENT 5.04 E
t Facility Comment:
We agree with the' answer key, however, we feel the question was misleading since "several hours" was not quantified.
Also, this question required the candidate to make an assumption that is beyond the scope of his classroom 1
training.
The' assumption he'had to make was that the burnout of Xenon at that power level is less tha'n the buildup of-Xenon due-to I-135 decay.,
In operating the plant during a startup, he is trained
^
to correlate rod movement at constant pawers due to Xenon 1
concentration changes.
The Xenon concentration' amounts, i
changes, and when they occur is more a function of the Reactor Engineer.
i The training conducted on Xenon Equilibrium is based on conceptual. knowledge of:
1.
The Xenon production and removal terms.
2.
Operational transient effects of Xenon concentration.
3.
Relative changes in Xenon Reactivities for reactor startup, shutdown, and poweE operation.
We feel the question would have been clearer if "several hours" would have been quantified and a phrase explaining that Xenon burnout is insignificant.
Also, Answer "b" is vague so we feel it could be interpreted as the correct answer for the following reasons.
1.
The phrase " rapidly insert" is meaningless because
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there is only one rod speed.
The operator has no way to increase rod. insertion other than to' reduce the time between rod selection.
2.
The phrase "high rate of Xenon burnout" is vague and he is not required to quantify Xenon burnout with a power level.
He does not Pnow whether it is a high rate of burnout nor..is h6 ad to know when the
.w burnout rate is high.
Al*, v-n the different half lives of Iodine and Xeno<
^ lours for Iodine and i
9.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for Xenon) t.re
't into play it becomes very difficult to evalua*
a phrase as being incorrect.
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ATTACHMENT 2
We request that credit be given for Letter a or b.
NRC Resolution:
Disagree with comment.
The facility reference material did not substantiate that b would also be an appropriate answer.
In addition, " rapidly insert" rods is a generic industry term not related to rod speed, but rather to how l
quickly the operator performs the rod insertions.
Sufficient information was provided to indicate to a knowledgeable operator that Answer 'b' would not be correct (i.e., Rx just critical means low neutron flux levels which implies low burnout of Xenon).
To state in the question.that Xenon burnout was minimal would make 'b' an obvious wrong answer and not worth using.
It should be also noted that the candidates were briefed prior to the start of the examination that if any question was not clear they should ask the examiner for clarification.
The question will be reviewed for any necessary revisions for clarity.
5.06 B Facility Comment:
We agree with the answer key for 5.06 B, but we also l
feel that there are additional acceptable responses as stated in the reference used.
Several of the major sources of hydrogen are long term sources such as:
Radiolysis of Water i
Corrosion of Zinc based paint Aluminum corrosion i
Since the question did not specify the time period, we l
request the answer key be modified to accept any of the major sources of hydrogen listed below:
{
Steam:
Zirconium Steam:
Steel Core:
Concrete Radiolysis of Water Corrosion of Zinc based paint i
Corrosion of Aluminum
{
NRC Resolution:
Disagree with comment.
The question asked for the principal source of hydrogen in containment following
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a LOCA.
A Steam:
Zirconium reaction is the major l
producer of Hydrogen as shown in MCD Page 6-22, 1
Figure 6-F.
The answer key will not be changed.
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ATTACHMENT 3
5.06 C Facility Comment:
Technical Specification Section 3.7 list a limit of.4%
for Oxygen in the Primary Containment.
The bases for Technical Specification 3.7 lists two values, 4% and 5%.
System Description E-12 states that the alarm is set for 4%.
We realize that the Technical Specification bases and the 4
alarm setting are set conservatively low to ensure that the flammability limit stated in Mitigating Core Damage is not exceeded.
However, we believe it is more important for the operator to be trained using the Technical.
Specifications 4% limit since it is a licensed document.
The bases for Technical Specifications also implies there t
is more than one flammability limit, the 5% limit and the AEC (now NRC) recommendation of 4%.
The question also did not state the document that the operator should reference for the Oxygen limit.
Since we believe it is absolutely imperative that the SR0 is familiar with Technical Specifications and complies with Technical Specifications at all times, we feel an Oxygen limit of 4% is also a correct answer.
We request that full credit be given for stating that the Oxygen limit is 4%.
I NRC Resolution:
Disagree with comment.
The candidates were specifically asked for the maximum flammable concentrations of hydrogen and oxygen in the containment following an LOCA.
The question was for LOCA conditions - not normal operations l
as stated in your Technical Specification reference.
The answer key has not be changed.
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5.10 l
Facility Comment:
We believe there is a math error in the answer key.
EOL period calculation is incorrect and should be 45 seconds.
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l This will also change the final answer to 17 seconds.
l 62 - 45 = 17 l
We request the answer key be changed to reflect an E0L
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l period of 45 seconds, and a change of 17 seconds.
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' ATTACHMENT-4 NRC Resolution:
Agree.with comments.
The correct answer will be changed to BOL = 62, E0L = 45, Change = 62 - 45 = 17 seconds.
5 11 A i
Facility Comment:
The question did not ask for the reason or mechanism by which enhanced operation using barrier fuel is obtained.
The question asked for the difference between barrier fuel and other fuels.
l The answer key includes the difference in the first sentence.
The second part of the answer key includes the mechanism by which cracking is inhibited.
We request that the second sentence of the answer for 5.11 A be deleted and full credit be given for stating that a layer of zirconium is bonded to inner surface.
NRC Resolution:
Agree with comment.
The second sentence will not be required for full credit.
The answer key will reflect this change.
5.11 B Facility Comment:
Iowa Electric may or may not eliminate PCIOMR restraints.
It is true that Preconditioning can be eliminated by use of barrier fuel.
However, from an operational standpoint the main significance of barrier fuel is the ability of the fuel to accommodate large power changes without pellet-clad interaction failures.
The operator has v
been informed that PCIOMR may still be followed in an abbreviated manner to provide additional i
conservatism.
The use of the term " target exposures" is more of a
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Reactor Engineer term than an operator term.
The operator is more familiar with the term power level and we feel the two terms are synonymous.
We request full credit for answering in terms of allowing large power changes without fuel damage.
We also request that no points be taken off if the operator does not state that preconditioning can be eliminated.
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' ATTACHMENT-5 NRC Resolution:
Partially agree with comment. The answer key specifically states either answer would be acceptable for full credit.
Therefore, the candidate does not need to state that.
preconditioning can be eliminated to receive full credit.
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. However, the examiner does not agree' that power level and target exposures are synonymous terms.- One aspect of. a.
target exposure'is power level. The other aspect is the amount of time at a given power level. Therefore, the two are not synonymous, but are related, 5.13 Facility Comment:
This question involves a conversion from psig to psia, determination of the saturation temperature i
and a subtraction calculation to determine the number of degrees subcooled or superheated.
Any single mistake in this calculation such as improper or no conversion from psig to psia or a reading error on
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Table 2 could result in an incorrect response for both answers in the answer key.
We request that any single error in the calculation not be carried forward when grading the question.
NRC Resolution:
Agree with comment.
It is not NRC policy to double jeopardize a candidate when a question of this type is asked. A candidate would not lose full credit for a single error in his calculation.
5.15 B Facility Comment:
We agree with the answer key. We also feel there are alternate answers that are just as correct.
As per the attached reference (System Description C1 Page 47) we feel the below' listed responses should also be given full credit:
1.
Pump minimum flow ' valves 2.
Pump minimum flow valves which open on low flow (400 gpm)
We request the answer key be modified and full credit be I
given for the above mentioned items or similar responses.
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ATTACHMENT 6
NRC Resolution:
Agree with co;.1 ment.
Will also accept " Pump minimum.
flow valves" or " Pump minimum flow valves which open on low flow (<400 gpm).
6.01 A Facility Comment:
We agree that Technical Specifications list the five items listed in the answer key, but the list is for the second items and not the actual rod blocks.
See attached Technical Specification Bases.
The attached System Description (I-8) and Figure of refueling rod blocks describe the refueling rod blocks in greater detail.
We feel that a better answer to the question would be to list four conditions that result in a rod block.
We request the answer key be modified to accept any four of the below list:
MODE switch in REFUEL and:
1.
Trolley mounted hoist loaded with platform over or near core.
2.
Frame mounted hoist loaded with platform over or near core.
3.
Fuel grapple loaded with platform over or near core.
4.
Fuel grapple no full up with platform over or near core.
f 5.
Not all rods in and selection of a second.
6.
Service platform hoist loaded.
MODE switch in STARTUP and:
7.
Refueling platform over or near core.
8.
Service platform hoist loaded.
NRC i lution:
Agree with comment.
Will also accept the above mentioned list as answers.
ATTACHMENT 7
6.02 A Facility Comment:
We agree with the answer key for each condition, however, we do not feel it is absolutely necessary for the operator to answer the question using terms such as " level error, level set, steam flow / feed flow error" in order to demonstrate understanding of the feedwater control system.
We feel the format of this question could be improved by separating the questions from the initial condition statement.
It is very difficult for an operator to determine the question he is being asked if the questions are all written as one sentence next to the initial condition.
It is also difficult for the operator to check his answers following a lengthy exam if the questions are not readily identifiable.
The answer key for the last 0.5 points states that the "feedwater control valves will open to match new higher level." This statement is confusing and we do not agree with this answer.
If the "A" level detector fails low, then the feedwater control system will see a large level error signal no matter how far level increases.
Reactor level will increase until the feedwater pumps trip at 211".
(See attached System Description 0-15, Page 11.)
We request full credit be given if the explanation includes a final resolution of the event, i.e., " level increases until feed pump trip" because we feel this shows understanding of feedwater control.
NRC Resolution:
Disagree with comment.
The question asked how the reactor level will initially respond (immediately following the failure).
Therefore, feedwater pump trip at 211" is the final response due to increasing reactor water level.
The answer key will not be changed.
The candidate need not use the exact terminology used in the answer key to receive credit.
The question will be reviewed for any changes necessary to enhance clarity.
6.02 C Facility Comment:
We request full credit be given if the explanation includes final resolution of the event, i.e., "Feedwater l
valves close due to Feed flow / Steam flow error until level drops enough for level error to counteract the l
flow error.
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ATTACHMENT 8
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NRC Resolution:
Agree in part with comment.
The candidate need not have exact wording as the answer key to get full credit.
Each candidate's res.ponse~will be reviewed for its individual content.
6.03 8 Facility Comment:
We do not agree with the answer key (prevent lockout) and feel it.is more correct to answer that the diesel would fail to start.
01 324 states that to prevent an inadvertent lockout, a start signal should not be initiated within one minute i
of resetting an engine trip.
This precaution implies that diesel generator lockout may occur.
It will not occur if a start signal occurs before the one minute time delay.
We feel this precaution is in error and have sent a procedure change request to our Procedure Development Group.
The System Description for the Diesel Generator states that "to ensure there is not a failure to start, a start signal should not be initiated within 60 seconds of resetting a trip."
We feel this is the more correct answer based on review of the DAEC start circuitry and breaker control circuitry.
A short explanation of this circuitry follows:
A diesel generator trip will energize the SDR which seals in and energizes Relay "5" which shuts down the diesel generator fuel racks and prevents a restart.
Once the SDR is reset by clearing the " trip" condition i
and resetting PB4.
The "5" relay is deenergized with a 60 second time delay before it drops out.
Until the "5" relay drops out, the diesel cannot be started because the air start solenoids will remain deenergized.
The only way we feel a lockout could occur would be if the operator attempted to close the diesel generator output breaker without having the diesel running and a failure of the Sync check relay.
(Sync check - 25, i
prevents breaker closure if both sources are not in phase.)
ATTACHMENT 9
Since this entails a procedure violation and equipment failure, we feel this to be unlikely.
We request the answer key be changed to read the following.
The Diesel Generator will not start.
NRC Resolution:
Agree with comment.
Answer key will be changed to read as.follows, "The Diesel Generator will not start."
6.05 Facili ty. comment:
.The lower limit to the Standby Liquid Control Boron Injection Rate is no longer applicable, and no longer in our Technical Specifications.
The ATWS changes 1
(10 CFR 50.62) required both pumps to start on S8LC initiation via one.
Attached is a copy of Technical Specification bases which includes the lower limit.
Our' System Description (C-4 SBLC) which was the reference for the exam question will be revised.
NRC Resolution:
Agree with comment.
The answer key will be changed to reflect the deletion of the lower limit.
Full credit will be given to a candidate that states the lower limit no longer exists.
6.07 8 Facility Comment:
We believe the correct answer for 6.07 B is that the Recirc Drive Motor breaker will trip.
The Drive Motor breaker must be closed in order to start the sequence.
If the field breaker has not closed within 15 seconds of the Drive Motor breaker closing, then the incomplete-sequence timer will trip the Drive Motor breaker.
We request the answer key be changed to state that "The Recirc Drive motor breaker will trip."
NRC Resolution:
Agree with comment.
The answer key will be changed to read as follows, "The recirc drive motor breaker l
will trip."
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-ATTACHMENT 10
.6.08 A Facility Comment:
The valves listed in Question 6.08 A were.not correct
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in that-these valve numbers were for the Head Spray Valves.
The answer key is also confusing for this question.
l because two requirements are included'in one statement (2/3 core coverage and LOCA signal).
As stated in System i
Description C-1, Page 37 and Figure 17 (see attached).
There are four (4) (not three) conditions that must be satisfied in order to open.the containment spray valves after a LOCA.
.i These conditions are.
1.
Drywe11' pressure greater than 2.psig and 2.
Reactor vessel level is above -39 inches (2/3 core coverage) and 3.
LPCI initiation signal present and 4.-
Containment Spray Valve Control Switch to MANUAL
- We request the answer key be modified so that full credit is given for any three (3) of the above listed four (4) items.
We also recommend that M0 numbers not be used in future j
questions unless necessary, in order to avoid confusion.
NRC Resolution:
Agree with comment.
The question will be changed requiring three out of four conditions.
The answer key will be changed to include " Containment Spray Valve Control Switch in Manual." The examiners prefer to include valve numbers in case the candidate does not recognize the valve by name.
6.08 B Facility Comment:
As stated in System Description C-1, Page 37 and 1
Figure 17, the Containment Spray Valve Control Keylock switch bypasses the -2/3 core coverage and the LPCI initiation signal, i
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- ATTACHMENT.
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As stated, this design allows the operator to open the l
containment spray valves 'following a LOCA if either:
1.
Level is below -39 inches l
'2.
LPCI' initiation signal is not present.
We request the answer key be changed to accept either of the conditions listed above.
NRC Resolution:
Agree with comment.
The answer' key will be changed
-to state as follows, "It allows opening of containment spray valves by bypassing the requirements for 2/3 core coverage (-39 inches) or LPCI Initiation signal present."
Either answer will be given full credit.
6.10 D Facility Comment:
This part of Question 6.10 is not defined in enough detail to ensure there is only one correct answer.
APRM.B upscale is insufficient because the APRM Upscale alarm is a rod block.
The APRM Upscale Trip is-an RPS trip.
The operator could interpret the APRM upscale to be either the rod block or the scram.
'It sh'ould be noted that upon checking the references used for this' exam question, inconsistencies in terminology for the APRM upscale Trip and Alarm were found.
Change forms have been promulgated to Material Development to correct these deficiencies.
We request the answer key be modified for Question 6.10 0 to accept either " rod block" or "1/2 scram" as correct answers.
NRC Resolution:
Agree with comment.
The answer key has been modified to accept either answer.
The question will be reviewed for necessary changes so that only one answer will be acceptable.
i ATTACHMENT 12 6.12
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Facility Comment:
OI 388 precaution No. 3 states that the battery charger cannot supply emergency power by itself and would trip under this condition.
The question does not state that emergency loads are being supplied by the 250 VDC system when the battery
-is disconnected.
If emergency loads are not being carried and the battery
-is. disconnected, the charger would probably not trip.
System Description G-5 Page 14 also states the charger cannot supply current under all load conditions, meaning that the charger can supply loads under certain -
conditions.
As per Technical Specifications Section 3.8.8.2.C HPCI must be considered inoperable and requirements of 3.5 must be met.
The question did not ask for a procedural response, therefore, we feel the candidates should not be held responsible for mentioning Technical Specifications.
We request the answer key be modified and full credit be given for any of the following response:
1.
HPCI inoperable or 2.
Charger will carry the system as long as Emergency loads are not started.
NRC Resolution:
Agree in part with comment.
The candidate was not asked the status of the HPCI system.
The candidates were not held responsible for meeting DAEC Technical Specifications since that was not the nature of the question.
Therefore, the answer "HPCI inoperable" will not be accepted.
The answer key will be changed to add "or the charger will carry the system as long as Emergency loads are not started." Full credit will be given for this answer.
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-7.01 Facility Comment:
We agree with the answer key, but we feel the bases for HCTL curve as stated in NEDC 30796 (see attached)_ states the same answers as Answer No. 2.
We feel it is also correct to state that the pressure or energy contained i
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ATTACHMENT 13 I
in the reactor is low enough so that a LOCA or SRV blowdown will still be adequately amdensed by water in the torus.
We request that full credit be given if above answer i
is used for Answer No. 2.
j NRC Resolution:
Agree in part with comment. The candidate need not have
.the exact wording as the answer key to get full credit.
Each candidate's response will be reviewed for its indiridual content.
.7.02 Facility Comment:
We feel th'at the question is a good question to ask on an oral or simulator exam, but is inappropriate to ask i
on a written exam for the following reasons:
Iowa Electric's accredited SR0 training program does not require the operator to memorize the E0P's.
The operator must demonstrate the ability to use the E0P's at the simulator.
The question is a complex problem requiring the operator to enter several E0P procedures.
We believe a question of this nature would be appropriate for a written exam only if accompanied by the E0P's.
ES402 states that the operator must have a complete understanding of immediate actions.
We feel the question addresses those actions that are not immediate actions that the operator would take without a procedure.
Recommend Question 7.02 be deleted because it is beyond the scope of the SRO.
If Question 7.02 is not deleted, then recommend the following changes, i
1.
Allow credit in each section based upon the candidate's assumptions and subsequent. actions.
2.
The answer key for Part A should take into account that the initial conditions stated that Boron Injection is not required, but the answer requires
{
Boron Injection to be explained.
This could be confusing to the candidate.
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1 ATTACHMENT 14 3.
The answer key for Part B does not mention that MARFP also takes into account the minimum flow stagnation power level (8%) for no natural circulation.
Maintaining pressure constant ensures that the RPV is being properly flooded.
(Reference NED0-30796 Page 8-149.)
4.
The answer to Part C should also include operational considerations such as if Boron Injection is being used, or the Rx is shutdown. Again due to the nature of the question (i.e., written vs oral) it is difficult to assume that only some things are changing.
5.
The answer to Part D should also include a discussion of the " quality" content of the water, (i.e., Rx water vs river water) and credit given accordingly.
NRC Resolution:
Disagree with comment.
This question does not require memorization of the E0Ps.
It simply describes particular plant conditions and corresponding actions taken in the E0Ps and then asks why these actions are taken.
Since the candidate is actually told what the E0PS require, the E0Ps do not need to accompany the question for it to be answered.
It is not unreasonable to expect an SR0 to understsod why i
he is taking particular actions in an emergency procedure.
10 CFR 55.43(b) prescribes emergency operating procedures as a basis for questions on the written exam.
In addition, the question is supported with knowledge and ability ratings that are extremely high.
This question is appropriate for a written exain in that its objective is not to see if the
' candidate has memorized the E0Ps, but instead to determine if he really understands them.
Requiring an SRO to know the basis of the emergency procedure is consistent with industry practice and guidance provided to NRC examiners.
If as this facility comment implies, the SR0s at DAEC do not understand j
the emergency procedures, then this would certainly cast i
doubt on their ability to understand and control emergency
]
conditions.
Since the facility received a satisfactory j
requalification evaluation in December 1986, it is assumed l
l that SR0s at DEAC do understand emergency procedures and I
l that this comment was inappropriate.
1.
All assumptions indicated by the candidates will be evaluated on a case-by-case basis.
However, all assumptions needed to answer the question correctly were given in the initial conditions.
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ATTACHMENT 15 l
2.
Although in this instance boron injection is not required, the procedure does address this system whether or not it is required at that point.
This question specifically provides further clarification that boron infection must be addressed in the answer to receive full credit.
3.
The answer given in the facility comment for why RPV pressure is kept above MARFP is the same as that given in the answer key, but stated a little more specifically.
The more specific answer given in the l
facility comment is not required for full credit, but would certainly be considered an appropriate answer.
4.
These operational considerations are given in the initial conditions such as to arrive at the correct ansvers.
5.
The quality content of the water is not listed in tha references on the answer key.
This facility comment was not accompanied by any further references to substantiate this.
Therefore, the additional answer given in the facility comment is not acceptable.
In summary, both the question and the answer are considered entirely appropriate and remain in the exam as is.
7.03 A Facility Comment:
The answer key references Page 25 of IP0I-2, but only lists four conditions as stated at the top of the page.
In actuality, the page lists other requirements such as Drywell Deinerted, Oxygen concentration greater than 19.5%
by volume, and IP01-7 (Special Instructions).
The question implies there are only four (4) conditions, but procedurally there are many more conditions that must be met.
We request that full credit be given for any four (4) conditions that are stated per IPOI-7.
NRC Resolution:
Agree with comment.
Full credit will be given for any four (4) conditions required prior to Drywell entry that are stated in IP01-7.
The answer key is changed to include the following, "S.
Drywell is deinerted, 6.
Oxygen concentration is greater than 19.5% by volume, 7.
Ensure requirements of Plant Radiation Protection Manual are followed for Drywell/ Steam Tunnel Inspection.
8.
Verify containment air purge in progress.
9.
Verify at least one CAD 0 analyzer recalibrates for high 0.
10.
Verify 7
7
ATTACHMENT 16 that the TIP detectors are in their shields and the drives deenergized.
11.
Have Health Physics perform necessary surveys."
7.05 C Facility Comment:
Per EPIP 2.5, the OSS functions as the Emergency Coordinator until relieved.
EPIP 1.1 also directs the OSS to make the plant notifications per EPIP 1.2.
We believe that if the overall responsibilities of the OSS are evaluated using the entire Emergency Plan instead of one procedure it is true that the OSS acts as the Emergency Coordinator until relieved.
Iowa Electric's SR0 candidates are trained to function as the Emergency Coordinator until relieved.
If an event occurred that was classified as an alert or higher and the Emergency Coordinator has not relieved the OSS, then the OSS should maintain contact with the NRC.
We request full credit be given for a "true" answer in Part C and we also recommend this question be modified so there is only one correct answer.
hitC Resolution:
Agree with comment.
Since under varied circumstances the answer could be either True or False, this portion of the question has been deleted and the overall point value of the question modified accordingly.
7.06 Facility Comment:
The question asks for specific items as addressed in l
only one part of the EPIP's.
We believe that since the l
candidates are trained on all phases of the EPIP's and log i
entries into the OSS logbook are done in accordance with l
ACP 1410.3, Operating Logs, that credit be given for
]
I additional items that could be part of any log entry j
l made by an OSS when dealing with emergencies.
This would include additional items such as:
L 1.
Time of log entry 2.
Changes in plant operating status l
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ATTACHMENT 17 l
3.
Other entries.
These might include, Deviation Report numbers, 10 CFR 50.72 or 50.73 reports mace.
We request the answer key be modified to include other logged items during an emergency per ACP 1410.3.
NRC Resolution:
Disagree with comment.
The candidates were specifically asked.about the log entries required upon declaring an emergency classification of an event.
The log entries called out in ACP 1410.3 are entries made during each shift.
It is not stated that these entries are required to be logged during an emergency.
The answer key will not be changed.
7.08 8 Facility Comment:
The answer key lists reasons why the minimum speed setting must be correct, yet the question does not ask why the minimum speed setting is correct.
The question is asking for the consequence of improper minimum speed setting, namely that the pump would not start.
If more of an answer was desired, than an " explain why" phrase should have been added to the question.
Given the wording of the question we request full credit be given for Part B, by stating that the pump would not start.
NRC Resolution:
Agree with comment.
The answer key will be changed to state that "the RWCU pump would not start."
7.08 C Facility Comment:
The answer key states that the pump will trip, but no mention is given for the actions that would possibly occur to the rest of the reactor water cleanup system (i.e., DEMIN BEDS going into HOLD).
See attached.
We request credit be given for answers pertaining to other parts of the system such as DEMIN BEDS going into HOLD.
NRC Resolution:
Disagree with comment.
The root cause of the reduced flow to the point where the Demin Beds go into hold is the pump having tripped (as stated in the answer key).
If the candidate states additional items such as stated in the
ATTACHMENT 18 r
comment, they will be accepted, but for full credit the candidate will have to state that the pump. trips. The
' answer key will not be changed.
7.09-Facility Comment:
We do not.see how candidate No. 3 can be accepted even though he exceeds 5(N-18).
Even though 10 CFR 20.101 technically allows exceeding 5(N-18), we do not advocate exceeding of this limit nor do we believe any other agency wouldadvocateexceeding5(N-18).
Both the National Council for Radiation Protection (NCRP) and the International Council for Radiation-Protection (ICRP) reconmend that 5(N-18) not be exceeded.
Duane Arnold's Radiation Protection Training advocates conservatism for personnel exposure.
We request that full credit be given for. rejecting candidate No. 3 based on exceeding 5(N-18).
NRC. Resolution:
Disagree with comment. You are correct in the fact that 10 CFR-20.101 technically allows an individual to exceed 5(N-18). More importantly, the individual is allowed to legally exceed this in the particular question. The j
document you referenced is not a legal document and i
can only recommend guidance. The-10 CFR is a binding I
document.
Since no administrative limits are in place preventing an individual from exceeding 5(N-18) and an individual is legally able to receive up to 1.25 Rem /qtr.
regardless of 5(N-18) the answer key will not be changed.
7.10 Ei Facility Conrnent:
The question as written leaves off the part of trying l'
to first shutdown the reactor by the normal method.
This could be misleading to the candidate as he would be also looking at the operational conditions that would normally be followed prior to shutting down the reactor l-protection systein. We believe credit should be given for answers that utilize those operating conditions.
t
ATTACHMENT 19 We also request that full credit be given for stating that RPS should be reset so that the CRD system can provide adequate cooling to the CRD's (see attached).
NRC Resolution:
Disagree with comment.
The question asked, "What is the reason for this," not "what should be done" as was implied should be correct answer.
No evidence was provided indicating that in the shutdown conditions that normally would exist when the RPS system is shutdown that CRD cooling is a concern.
The only concern that is documented is that listed in the answer key.
The answer key will not be changed.
7.11 A Facilit.y Comment:
We agree with your answer, but we also feel that credit should be given for other operational concerns created by a H/X low level.
A low level could also result in cavitation of the RCIC pumps due to reduced NPSH.
RCIC Turbine Trip - Turbine trips at 15" Hg abs.
suction pressure.
We also feel credit should be given for stating the precaution is based on draining the H/X dry since the thermal shock is due to RHR-SW flow through the tubes of a hot dry H/X.
We request the answer key be modified to give full credit for any of the following, j
1.
Minimize thermal shock to H/X 2.
Prevent cavitation of RCIC pump i
3.
Prevent RCIC Turbine trip l
4.
Prevent draining H/X dry NRC Resolution:
Disagree with comment.
The material that was referenced does not substantiate that low level in the heat exchanger would directly cause a RCIC Turbine Trip.
DAEC System Description C-1 RHR Page 43 states the following.
"The RCIC pump suction is also supplied from the Condensate Storage Tanks so that if the pressure from the RHR heat exchanger drops below the Condensate j
i
y ATTACHMENT 1 20 i
Storage Tank pressure, a check valve opens to supply water from that. source so that the RCIC pump is not in danger of losing. suction pressure."
1 The. facility answers requested to be added to the answer key are not correct.
Prevent Draining the H/X dry was not accompanied by any'further references to substantiate this would in fact happen.
Therefore, the additional-answer given in. the facility comment is. not acceptable.
The answer key will'not be changed.
7.11 B Facility Comment:
The caution in 0I 149 is not based on a loss of inventory to the Torus.
Step (2) of Section 5.3 has the minimum flow valve closed and the breaker tagged open.
Also, the caution r
directs the H/X. Bypass Valve be throttled open instead of the LPCI outboard throttle. valve.
A note in Section 2.0 specifically warns against using the outboard throttle-valve.
Failure to open the Bypass Valve within ten seconds will
. result in the RHR pump operating at shutoff head.
This will result in inadequate minimum flow.
We request the answer key be modified and full credit be given for either of the following:
{
Provide adequate minimum flow
,i Prevent operation at shutoff head NRC Resolution:
Agree with comment.
The answer key will be changed to read as follows, " Provide adequate minimum flow for the.
)
pump or prevent operation at shutoff head."
']
The examiner would like to note that the 10 referenced in the test was 10 149 Rev. O dated July 24, 1986.
The licensee supplied reference was 10 149 Rev. 2 dated April 30, 1987, to support this comment.
The NRC question was correct from the information supplied by the licensee.
7.13 A Facility Comment:
The question as written could be confusing because of the use of the word " basis." This word could imply the Technical Specifications bases and could cause the 1
l' l
1
1 l
1 s
J 1
-ATTACHMENT:
21 i
s candidate to' utilize ti,em'in developing-his answer.
.l We believe credit should be.given as appropriate
^'
'(i.e.,use~" reason").
We also request credit be given for other operational
' concerns that are generic.for' pumps operating at shutoff head such as:
1.
Overheating of Pump 2.
- Seal Damage i
j!
3.
Cavitation NRC Resolution:
Disagree with. comment.
This. facility comment was j
not accompanied by any references to substantiate 1
it.
Therefore, the additional. answers given in the facility comment are not acceptable for full credit.
l 7.14' l
Facility Comment:
The answer key referen:es E0?-6, but the question is l
worded such.that other~means of shutting down the reactor j
can be.used, other than E0P-6.i We request credit be given to: reflect-other operatiot al means, of scramming the reactor to include the below listed and any other n ans identified by the candidate.
s 1
1.
Locally trip the' turbine (power above 30%)
2.
DeenergizeRPSMGSetsbYopening.breakerat 1832 and 1842.
s 3.
NRC Resolution:
Agree with comment.
The answer key will be changed to include the following.
"c, Locally trip turbine (power i
above 30%) at Turbine front standard on operating deck;.
l
- d. Deenergize RPS MG Sets by opening breakers at 1832 and 1842 located in the 1A3 and 1A4 switchgear room." Vent the scram header was not included since it was already j
l discussed in 7.14b of the answer key.
2 8.03 Facility Comment:
The question is extremely long and multipart in nature.
l ES-202-E.12 states that mulitpart questions should be j
broken down into logical sequential parts.
As currently written, we believe that the question is not clear in j
4
(
t o
= ik,-
- - - - - - - - - - - - - - = - - - - - - - - - - - - - - -
7
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4 ATTACHNENT 22 c%e stating the requiremppts expected of the-:andidate.
The-question containD both the long initial conditions section and the actuaP two part requirement of the candidate.
'We agree with the answer!kdy in part,'however, We believe that credit should be given for actions taken that would
- n check for the actual systems ~taken'out of service and the '
operability. requirements since Technical Specifications 3.13C and 3.13E states that only certain.
deluge, sprinkler systems, and hose stations are required to be operable.
The information provided in the question is not complete, and would require the OSS to check on d P&IO exactly what systems were taken out of service.
i NRC Resolution:
Disagree with comment.
The Region does not consider this to be an example of a multipart question as stated in the
' comment. An example of a multipart question would be i
. Question 8.04 with two distinct points.
This question simply listed a situation and then asked what action was q
required and how scon.
Other similar questions, such as 8.11, were not commented on in this context.
The question as stated says that a Technical Specification problem exists.
9 Therefore the candidate does not need to know which specific j4 portion of the deluge and sprinkler system or which hose 1
station is inoperable since the required actions are the same and the'LC0 foR that system is entered as stated in the answer.
As far ab;what was requested of the candidate, i
it is clearly stated in the final retence of the paragraph.
l The answer key will not be changed.
4 3
8.0hj 4
it Faciljty,, Comment:
The question does not follow the Examiner Standards 202
'4
.L in the following ways:
g
.s u
1.
ES-202 2 12 states that multipart questions shou 16 be broken down into logical sequential parts, t
i J
3 2.
ES-202 E.16 states that "if a specific number of s responses are required, the question should clearly.
4 state that expectation...."
3.
ES-202 E.20 states that diagrams or sketches should be used as attachments to written examinations.
It w.
further states that the use of these attachment <,' j's
'I
\\
s b
b s
ATTACHMENT 23 1
provide an effective and easily interpretable way i
for the candidate to demonstrate his knowledge of the topic / concept.
As written, the question could be understood by the candidate as one question with three listed conditions together and answered accordingly.
The answer key, as written, assumes that credit should be given to the candidate based upon his ability to correctly analyze l
the situation using all three of the conditions together.
Furthermore, since a core map was not provided as an attachment, credit should be given to the candidate for i
stating the requirements that would be followed if this operational situation were to arise.
Finally, the answer key for condition No. I states that
}
this situation would have no effect on plant operations.
3 However, the two rods that are inoperable are in separate RSCS groups.
(10-07 in Group A2 and 10-27 in Group B3),
l Since the original scenario has the candidate making preparations to start up the reactor, having these two control rods inoperable and in separate RSCS groups would place a very large operation restraint upon the plant.
I Technical Specification 3.3.A.2.f.(2) clearly states that all rods within a notch group containing an inoperable 4
control rod shall be positioned within one notch of the inoperable control rod whenever the rod sequence control
)
system is required to operable. We believe that credit 4
should be given to the candidate for stating the l
p0ssibility of this constraint, seeing as how a j
core map was not provided, and we do not require i
the candidates to memorize RSCS groups.
NRC Resolution:
Partially agree with comment. This question, as written, is consistent with other Region III examinations.
Each question undergoes an independent QA review by a person other than the author to ensure it meets with the Examiner l
Standards.
The number of responses depends on how the candidate interprets DAEC Technical Specification (also see l
response to Comment 8.03 and 8.08).
The candidate should j
be familiar enough with a full core display / control rod l
positions without needing a diagram to assist him. As far as how credit will be given, each candidate will be l
individually graded for his ability to assess the situation.
i The question was not intended to require the candidates to have the RSCS groups memorized. The examiner will review all assumptions made at arriving at his answer.
The answer for the first condition given will be changed l
to reflect the operational constraint imposed on the plant I
as noted in the comment, j
t
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l
l l
ATTACHMENT 24 8.08 Facility Comment:
The question is stated, does not follow the Examiner Standards.
ES-202 E.16 states that,."If a specific number of responses are required, the question should clearly state that expectation so the candidate will know when the answer is complete."
-The answer key for Question 8.08, list five overtime guidelines that are violated, but there are seven total violations. We believe full credit should be given to the candidate for stating the general guidelines that are violated,.rather than being specific and listing them individually.
NRC Resolution:
Agree in part with ccmment.
The question is not "open-ended" or " vague." The intent of this question is to see how well the candidate can evaluate a given situation and apply his' understanding / knowledge of the overtime guidelines.
Therefore, the number of responses are determined by his interpretation of AP 1410.1.
Since the question by implication asked for all. violations,.this in effect.is the same as asking "How many criteria have been violated." Credit will be given'for the candidate stating the general guidelines violated as well (the-specific answers in. parenthesis were for the graders use only and are not intended to be required).
8.09 B Facility Comment:
The answer key for Question 8.09 B is directly derived from Section 6.3.2 of ACP 1406.3, revision of Procedures and Instructions.
However, there are other places'in ACP 1406.3 that describe situations that would utilize temporary procedure revisions.
In addition, Figure 2 of ACP 1406.3, shows the temporary document change form, which contains a " temporary change information reviews" block that describes reasons for temporary changes. We believe full credit should be given for answers based upon these situations.
NRC Resolution:
Agree with comment.
After a review of the reference material, the following list will be added to the answer l'
key.
"5. When it's impractical to accomplish permanent revision to approved procedure.
- 6. Editorial.
- 7. Improvement.
- 8. New Procedure.
- 9. Inactivation or Reactivation.
- 10. Change of Safety Intent or l
Function.
I i
1 l
1 Q L _-_--_ -_- _--__
s
1
, ATTACHMENT' 25 0
I 8.11 I
Facility Comment:
We agree with the answer key as written, however, we l
believe that full credit should be given to the candidate for describing the concerns about providing an isolable volume (i.e., 2/3 core coverage), as described in the System Description A-2, Reacto'r Recirculation System, 1
when dealing with Jet Pump Operability.
NRC Resolutiom Agree with comment.
Will also accept "provides an isolable volume capable of being flooded after a recirculation line break (2/3 core coverage)"
as alternative answer.
8.12 A j
Facility' Comment:
The question is based upon a new Technical Specification change that is currently having a letter of. interpretation written to clearly define the facilities stand on the issue.
We believe that the question as written does not provide
{
enough information for the candidate to fully answer the question.
The candidate must spend time in developing an assumption of the scheduled due date of the surveillance-to ensure the correctness of his answer.
We believe that credit should be given for answers'that are correct when based upon the candidates assumptions and utilizing Technical Specifications 1.0, Definition No 26, Surveillance Frequency.
Also, per ACP 1407.5, Surveillance Program, it is the surveillance performance coordinator's responsibility in setting up.the due dates on the surveillance test procedures and not the OSS.
Using Definition No. 26, Surveillance Frequency, as stated in Technical Specifications the following information could be used in developing the candidates answer.
Technical Specifications quarterly (92 days) i 25% =
23 days prior to the scheduled due date, or 23 days j
after the scheduled due date.
Combined time interval for any three consecutive surveillance intervals shall noc exceed 3.25.
92 days x 3.25 = 299 days.
j
. _ _ _ _ _ _ _ = _ - _ _ _ _ _ _ _ _ _ _ _ _ _. - _ _ - _ _ - - _ _ _ _ _ _ - _ _ _ - - - _ _ _ _ _ _ _ _ -
p
.}1
[;i.O
' ATTACHMENT.
26 Assuming January 1,1987, is the earliest the 0SS could run a Schedule Surveillance, places the due date at 24-January 1987.
Using.that due date the following applies.
RUN "A" 3 October 1986 (90 days prior to 1 January 1987)
RUN "8" 1 January 1987 (Given)
RUN "C" 16 April 1987 (Given)
Combined time. interval A, B, C = 197 days
(
RUN NO. 1 1 January 1987 (23 days prior to 24 January 1987 due date)
RUN NO. 2 16 April 1987 (10 days prior to 26 April 1987 j
due date)
RUN NO. 3 19 August 1987 (Absolutely last possible date to i
RUN with 27 July 1986 due date) i l
Combined time interval 1, 2, 3 = 230 days NRC Resolution:
Disagree with comment.
It is obvious from this comment that the facility does not understand the new Tech. Spec.
regarding surveillance frequency.
The facility proposed.
answer of August 19, 1987, is.in violation of Tech. Specs.
in two respects.- The maximum allowable extension cannot l
exceed 25% of a surveillance interval.
The time interval between April 16, and August 19, 1987 is greater than 25%
(or 115 days maximum).
Therefore, the date of August 19, i
1987, is not correct.
The combined interval between three consecutive surveillance cannot exceed 3.25 of the l
specified: interval, 92 days in this case, or 299 days for a quarterly surveillance.
The combined intervals for the surveillance from October 3, 1986 to August 19, 1987 is 321 days or 22 days longer than allowed by Tech. Specs.
j Therefore, the date of August 19, 1987, is again not correct.
The facility is apparently keying in on dates 4
and due dates for surveillance rather than intervals as specified in the Tech. Specs.
This is obvious from their calculation of 230 days for the three surveillance
)
conducted on January 1, April 16 and August 19, 1987 to i
show that they met the 3.25 requirement.
This constitutes only two intervals vice the three required by Tech. Specs.,
but did comprise three dates.
The answer key will not be changed.
l
_--___________a
4
%;,.A.. STERGQM3 U.
9.
UUCLFAR REGULATGFY COMMISSION RENINR R E t. C T O R OPEPATOR LICENSE EXAMINATION F AC)! ]I Y :
DUAM ARNOLD REACTOR 1YPr :
_DWR_.GE_4___________
DATE ADMINISTERED +
- 0. 7 _/ 0 7_ _/. _1 0 _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _
~
EXAhTNER:
H_ARE,._F_._A CANDIDATF.
I n s _i rm_ C.T _J _n u e T o c A N ra f: ATF!
une Separate p a ra r for the -nruers.
Write anwers on one side only.
3tep je c;u e x t i c n
- s. h e e t on top of the ensoer theets.
Point-i-for each
..g ie.t i o n ar" i m.j ! c : l e d in parenthe n s f' t e r the o.uestion.
The c, n c <. i n c
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leent 70: in each c a t e a., o r v,
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E n m.i n a t i ci n pepers wi)) be picted up si (6) hout ofter I
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- Er iAiEn0P1 fit P.APPIDf,Tr'c C f TI7 C O R Y
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THEORY DF HUCLFAR P O L! E R Fi M nPERAT3DN-FLUIDF AND T H E R MO D Y N A'il C G t
_. y1_*
6.
PLAN 1 9YSTEMS DLOJGNr CENTHOLr p.. _ _.
AND JN91RUnENTATION
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Nr:C RULES AND CUTDELINES FOR LICENSE EXAMINATIONS-
- During the adkinar,tration of this exsnination the following rules apply:
1.
Cheating on'the ewaniination n.eens an automatic denial of your application and cavid result in
- n. o r e severe penalties.
-l 2.,'Restroom trips are to be lin,ited and only one candidate at.a time.may leeve.. You must avoid all contacts with enyone outside the examinataan room to woid even-the appearance.o r possibility of cheating.
13.
th e blaciJirk or dark penc'il 901'; to facilitste legible reproductions.
4.-
Print 'your name'in the blant provided on the cover sheet of the e;;aminaLion.
6 5.
Fill in the date on the cover sheet of the exenination (if necessary).
-6..
Use.only the peper provided for answers.
.7.-
Print your nan,e in the upper right-hand corner of the first page of e g(n -
section of the ensuer sheet.
C,-
Conercutfively number each answer cheet, utite 'End of Cate. gory _' en appropriece, start each c e teeot y on e nee prae, write.oniv on en* J i r' o or the peper. ond ur Le "Last Page' on the Irst a n ::v e r sheet.
9 Ph u ur each sosuer
- t. n to cctegory tod nun ber. for cuamplee 1.4r 6,3.
' 10. Osip (t least t h r o i-line, between ench ancuct.
, 11, 9epar :Le enroer sheets fror, pad and place finiched answer sheets frec down on your deel or tcble.
l it. Us -hbreviations only if they cre commonly used in facility 111gE512LE.
- 13. The point value for each quection is indicated in parentheses cfte-the question and can be used as a evide for the depth of ensuer required-Show
'l c a l c u l t '. i o n t.
- c. e b h n d c - or resunplioni used to obtcin en Incuer
.i-
- . O rp g t F. p n a l ( e f } },0 0 ', e la u;ielh>1 1ridi(e. led ife t h s,.;
quention o r' not.
15 Psrijal cred.L-niay be given.
Theref or e.
ANSLER ALL P/RTP OF THE D.rCTION AND FC NOT LEA'E WY ANSWFP DLAM.
- 16. If p a r t e.
of
'he e eninction ere not ej
=r as to intent., ask quv.t;on' of the o $ s. in n e i _i ! ],> 4 174 Y Of f th u % i 01 j f-D ie i I. f' I 6 ' o f. n I
) tIID (? O V e f
% Isee$ t lll. b ind1Cateb t'3 E k e not recoaved or bear-given c,sistmae ir work it you. out vid you hi r b n;-- l e ! : n g
o n P o lim t i or.
H nm L
'm.:
dont efter the examin; on hc i
h h e r.
e n ly F l 3 i I (? ij,
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1
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as, When you complet e your examination, you shsil:
a, A Aerable.your examination es follows:
(1)
E n a n, questions or, top.
(2.
Eacn rida - figure -
t.t b l e s, etc.
'7)
Antur psges including figures which are par t of the cnsurr.
Turn in your copy of the examination cod ell pages used to answer th. e;eminction questions.
-c l ui n in all serap paper t= n d the balence of the paper ths+ you dit' not use for snitJering the questions, d.
Lenve the e m m :; n e t i o n creo. as defined by the e n a n. i n e r.
If efter l
1eaving, you tre found in thir, ere-while the entmination ic s: t i l l i T.
pr090e55 YOU! liCen%C (b E y be. den $ed Or r9VOhed.
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g 6_ _. _ _ T H R.O R. Y _. O F _ _N U C L F _A _'i P O W E R_ _ P.. L_ _A N_ _T _ O P E R A. T T 0_ tJ _,_ F L U I_ D_ S_.,
A_ N_ D_
PAGE 2
T H_ _F P. M O D_ Y N _A r.i. T C_ S..
I 1
OUEST3ON G.01 (1.00's Uhi ch on*>
of the following s t. a t e rn e n t s in true r og a r di rig i ntrinsic neiit t on courcen in e shutdowr, recrtor?
c.
Practice 11y 11 tha noorce neutrons f r o rn spontaneous fiscion c a ra e f r o r.,
u.
The alphr-ne utr on 'ource c oie s iton the a l p h r, d e c :n y of U-238, U-239, snd Flutonium which interact with 0-10 in the moderzter to yaeld neon ond i neutron.
c.
The n,; o r corsentt9 tion of source neutrons comes from spontaneous fission of U-238.
d.
The phot o-inutr on tout ce is lec,s significant at BOL than EUL.
DUESTIori D,02 (2.00i TP'ir or FALSE ror a constent reactor pe: iod, il taken the RAME AMOUNT OF llME T D CHMM reactor power f r ort li to t'
et it doer to chmnge it from 10'4 to 505 FXPL All' YOUP At!SWE R F ULi.Y.
MHOW THE CALCULA1 TON YOU L' 0 U L D U S 'i in V F P1 r's YOUR M'9WEP.
D U U' T I D N 5,03 (1.00)
Which one of
.he falloutng s t a t e ni e n t s, it NOT trur iesarding the LHUP
()inver heet g r n.f r e t i o n r,te) thetu-] l i n. I t ?
13.l; LL'ft for both BnB and P0xBG f-l.
Thn
_MG' rH,1 g n l a tn i t
=
7700 b.
The 1 i ro i i is beren on maintaining peak c l w'i is t e m p r r : L u r e
=
deg er 7
C, II e LHGP spo i. I' : t F i
~t ori
'ir e that i'e L. H O P.i I i a n ',> 1 c t' i
jet thf de ' g r-V'lun even if fue3 pel 3 i 1 lnnti facc! ion 1L p o ':: 4.. ' ' ] R t P i
.r.ejedy ii o s i l i.-
T-p., t.i l t in f :iO ] e16d c r i c. I i i.,,, i' i.
d.
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't. ? ) 4 1.1 p i t i.
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NUC.IEAP.PnU.FR-..- - - l Mi t OPFRATION. FLU 1DG-MJD PAGE 3
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THERnnoyNon)en OUES110H
"'i. 0 0 (1.00)
The reacter tript fron, fn11 power, estnlibriun. xenon conditions.
Four ( c 't hour. Jcter the reactor it brought er itical and power level is n, c i n i a ) r a d on rene.> D of th. IRM9 for ;evorc] hours.
Which of the rollowing s t a t e m e n t s, it COPfECT concernine contro) rod motion during this period?
L.
Pcris will h: se to ti e wit heir a un due to venon build-i..
b.
Rods will have T. o b e rapidly inser ted ';i n c e the critical r-nc t or wi;l cause a high rata of eenan burnout.
c.
Pnds vill have to be Inserted since Menon wall closely follou
. ta, n o r nia l decay rate.
c'.
R o W.,
ul)1 appr oxin.atel y r emai n as ic as the xenon establisher itt equilibr ion value for this power 3evel.
40EPTTOM
.05
(*.no) i The ' c: -'or is critaccl at 106 eps WI!ICH of t, h e following best deter iber the behavio-of neutron power following a p r o n.p t intertion of nu p tiv-ie ctivity.
(Assume that not enough negative resetivity is anded l
to shuldm:
the reactor) l Noutron powee tirops imn:ediately to ' bets" (delayed neutron fraction) times the neutron pouer prior to the pr ompt insertion of negative reat iivity.
l h.
Neut.ron ;s o u e r decreasec linestly with time siter the initan] pint.pt H
ri r o f.
c, A f *. t r lhe initial pron pt drop. netett on poWCr deC r E RSU 4 On a C o n s.t E n '
negcl1Ve perin#I the it.6 g n i t u d e of the period determined by the c o. n u n i cr neptin r ei n c t i v ; t y inserted.
t l -
d.
Cercose only delpyed neutrons ar e left immed]ately Efter a negr ti u reactivity.inwttion, neutron powor decreases on an 80-second p o:r i o d 1 e g i: r d 3 e c. ta of tho s1:e of the noe.
jve reactivit insertion.
( 'J 3 x:.w 's CATICD?Y O '.i C 0 r1 T I N' t r l+
ON HFXT P'iGF *ttuti
p E _. _ _ T H_ E_ _n R Y _ O F _ _N U_ C I F A R _ P O_ U F_ R _ P t_ A N T _ O_ _P F R_ _A T I. O N.,. _ F L. U_ T_ D_ f_4_,AND PAGE a
l 1 H. F R_ _M O.D..Yr.1 6 _M T C_ P QUESTION 5.06 (3.50)
The Ovene ArnoJd containment atmospheres are inerted with nitrogen to limit the post LOCA oxygen content, e.
What is the princi Pal source o f' oxygen in containment follouins a LOCA9 ti,
Phet ie the pri_ncipal source of hydrogen i rl conLctnment following LOCA?
1 c.
What are the m a n i nc u m permissible concentrations ( 1 1 ni i t of f l a m mi; b l e i
regioni of hydrogen and oxygen (in volume percent) in containmere following a LOCA9 i
I I
OUESTION D.07 (2.50) i 1
Assume that t h e-eactor i t, being started up from Cold Shutdown and rtd drop aceident occurr early in the stattup.
Of the void, doppler, rnd l
t e n p e r a t t.' r e coefficients, uhich vill act first, second, and third to limit the irpid power rise?
EXPLAIN YOUP ANSWER.
QUESTION 5.00 (2.00) 2i The reactivity worth of a s i n g:l e control rod will (For e :h s t a t e rn e n t below, indiccte INCREASE or DECREASE.)
f a.
If the voad content around the rod INCREASES.
b<
If the moder etor temper nture DECREACES.
r, If cn 3 r: j a c e n t contr ol rod is WITHDRAWN 4 1
__f d.
If xe-135 concenttstion :>round the rod DECRE/MF;.
l l
i j
DilESTION 5.V
(:. io )
]
H O 'd dr+
- o r i.? i conrr n Lr - ti en afI'FCI p 'r> r i p h e r a 3 rod Warth following :
<;r-f t on hi h poemr and WHY don thi occur o s
i l
l l
l l
l i
1 l
(***t3 AT'~CORY % CnNTINUED O! N C ):T PAGE
- ?
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{
j
)
n
(
M_ _. _ _1_ I_t _t r: p y_ _ _ _.n. F _' N ll_C L. C_' A R. _ P O _W E R_ _ P L A.N.. T_ _ O_ _P E_.P A T_ T O_ N_ _,_ F_ L_0_3_ D ci _. _ A_ N D_
PAGE.~
S TH_r_-P..M0_pv_N_A__*1_C8_
JUES73 0R 5.10 (1.50)
- Suppose 'Beff over core life decreases from 0.0072 to 0.0055.
With:equc1'
-insertiont-of 0.001 dK/K of, positive rcettivity
.n.
Pilchlate the cbange in reactor period over core 'lifb.
(1.0)
'b.
' 'W h a t -i s the cause for this change.in Deff?
'(0.5)
~
hlJEST10N 5 11-(1.50) a.
Explairi how barrier fuel di f f ers f ron,' the pr evious types of fuel loedvd'into.the core.
(.75)
Lv.-
How uil] thist affect of. change power operation.at your'fecility with
.s furj core lorded wi th: tvr r i er fue17
(.75) i l
l l
O t.!p B T 10 N 5. M'
(2.50)
H
'Re9c7 d.i ng MCPR (Mininam Cnitical Power Retio);
l a.,
What PHENOMENON 00llLD.snist in a fuel bundle if it were operat+d ht an MCPR LES5 THAN ONE (<1 0) and WHAT is the concern if this phennwr on' uas to occur?
- ( 1. 0 )
b.,
. W h y. nn.i n t t h e T e c h n i c a l Specification MCPR lin.it'anclude a M facter when' core #10w i, LESR THAN RATED?
(1.0) j l
.e.
HOW d o r.
CP.TTIC POWEP -chenge (INCREASES, DECPEASES, OR REMAT;K 7HE GAME.) u52n ici subcoeli ng inct ear ei (core inlet water getr cDOlerl' sO.5 Dtl5 N T) OM 5'.15 f1.00) r.
&nd < B '.; pttg.
J1 it mbennled or siip m h' i t. c:
j A # 1 m i c' La at 400 degreen Frsd b y. h 0 0 0Phy'dOgrPD 7 i
l l
t
/
- E F r '6 C A T f G O it'Y Off O O N T 3. N : ! F " ON NEYl PAC-r>**vT
8
~ THEORY O.t Hitrt F AR L,C.
LUIDS.---------AND PAGE 6
OPERATION, F r--------
POWER Pt. ANT
.ldESU991dfl.01EU l
l QUFSTION S.14 (1.50)
The att6ched F:gure 2 shoun a T-S di ag r an, for 6 stean. cycle.
3.
Whst phenomenon is taking place between points 1 and 2?
b.
How does.'this procetas benefii plant operetion'
.e.
What i <s. the divedvsnteae a.tocitted with this process?
00ESTIDh-5.1D (3,00) a,.-
What'is " shut-off head' for a pump and why is it undesirable for 6
. pomp to run in this condition?
(2.0) b.
How doen the'RHR system protect its pumps from operating at n
' shut-off head' condition?
'(1.0)
'l 1
I l
l
(*Wo* Ds0 OF C/4TECDFY 05 A u t"
i
& __5LOUI.li!2IE00_DESI00:_GOUIBOL1_800_IUSIEUdEUI61 IOU
' PACE-7 I
QUESi10ff 6.01; (2.00) i
~
List FOUR ~ (4)' rod biceks anot,i ated with ref ueling equipnient..
(2.0) i
.00ESTION 6 02 (4'.00Y Acwrm the f e e. d w e t. e r '. l e v e l control systr+ is being operated in.3-element e n n't r o ] using. reactor level detector channel
'A'.
Reactor. power it.Et
.EM r4,t+sdy state.
T;or each of the instrument or control signal failures listed below, state how reactor. level. wil] initially' respond (increcse, decreace, or v e nic i n constant) and briefly explain why, in tarnis of what it
.heppening in the L evel Control Systen..and ' Feedwater Systen.. immediciely following the fcilure.
j i
fiOTE' A block disgram of'the feedwater level control svstem is attrehed..
(figure 21 l
a.-
-Channel
'A' - r eactor' flevel detector signal fails low.
b.
I ocn of signal to
- P' feedwater control valve-M/A tran:fer stction.
I c..
"E" foortwater ] ins flow cignal fnils high.
OUFSTION 6.03 (2.00)
The follauing question: concern the starting of the Emergency Diesel
- l Generators.
j a.
Explain the concerns on why a diesel generator should not be run i
u n l o a d e rt.
(1.0) b.
1he r e s.c t or operator just reset c gener ator trip.
If an a Lienpt wee nade to restsrt the diesel before the ono n,inute tine lap::e.
whet would he the strtuc of the d i e s e l '?
(1.0) l l
1 1
(**rr*
ColEGORY 06 C 0 t! T I t W E D O!! t! EXT PAGr rwtrr3 i
6:
6: _ _ E L 69 I _ S I S1E d 9_ D E S I O h _009 I S O L.t _680_ I t! E I E u d E U I 61100 PAGE B
QUESTTON' 6.04 (2 50) e, Driefly enpinin WHY it is i n.p e r t a nt to closely monitor ~RWCU s'/ sten.
f3ourste/tepiperature when operating in-the blowdown mode.~(NegiecL of f eet on vessel level.)
..(1 0) l t.<.
State t h e.
purpose of the RWCU Dlowdown mode.end state when this
- n. e G e ' i t ' u s e d.
(0.5) c, Fo t s t. e. t h e T 'r' O L ( 2 ) datometic. closure signals for RWCU CV 2729
(;
-ithe Clean Up.-Drain Header Contral Uslve)-while operating in the blowdown mode add explain the bssis.of each.
(1.0) 20UEDTION 6.05~
(2.00)
L The Staridliy Liqui d Contr ol System (SDLC) injects sodium pentaborate L
solo' tion into the reactor coo 1Ent.
The punip discharge was desisned to l i n. i t the ba r on_ i n j ec t i on '. r ate.
Why (i. err whet is the basis) are-there lin.itt at.which the solutien mu:,.t be injiated-(both upper end lower limite.cnost be discussed for fu)) credit)?
(2.0) l r
(rr*11 CATEGnRY 05 CONT.TNUED ON NEXT " A C ir i n rx )
l 4
i
' 6 __EL6EI DYSIEtjE_QEDIGU2 COMIS961_6UD_IUEIBUt!EU161190 PAGE 9
i l
l QUES 110rl 6.06 (2.00)
The followins relant c o n d i t i o r.s eulst:
= 3 psid I
Jet Pumps 1 - 53 Differential Preawt e ' Meter) o - 16 Differential Pressur e (Meter)= 15 pnid Jet Pumpt Pecirt Syctem A On]p In Seivite (Annunciator)
Off
=
I a.
Th; TOTAL CORE FLOW Recotder would calculate enre flow by which of the followind methods';
(1 0)
(;)
Loop A Jet Punip Flow + Loop B Jet P i i n. p Flow j
'I (2)
Loop A Jet Pump Flou - Loop B Jet Piimp Flow (3)
Loop E Jet P u m :.. Flov Only
'd)
Loop A Jet Pump flow Only i
b.
After 5 r,rinutes, the operrtor o p e n ts the Dischst se W1ve of Recirc l
Dump D to maintain th+ E: Loop temperature.
The T01AL CORE FLOW i
Retorder would ce;culete core f1ow by which one of the fo3louine wethodt?
(1.01 (1)
Loop A Jet Pump F]ov + Loop D Jet Purap F low l
l (2)
Loop A Jet Pump Flow - Loop D Jet Pump Flow (3)
Loop D Jet P u r,. p F l o w Only i
f4)
Lcop A Jet Piine Flow Only i
O D E b l $.O r!
6.07 (1.50) a.
Wh:t wil) cause RECIPC PYSTEM A ( D '> Sl ARluP SEQUENCE IUC0tU'.f TE
- v. n o u n r i s t o r w h t-n etsiting E recircu]ction pump?
(1 01
'b.
Ul< a l
" t L ori OLCur!
'I
'n InZOmp)Ote start G e qu e ric ** 15 Ci c t e c t e ri.
(0,ns t
(wwwn CATECDPY 06 Corn ItT' D ON NFMT PrGi r m s) t-
r___
l I
f.,
PLANT SYST E MR DESIGN. C0fJTR01.. AND INSTPUMFNTATTON PM;E 10 i
(1.50) actc/LA OUEDTION 6.08 e
\\
7
'c.
List Mm THREE.(3)0 conditions that n,u s t be met in order t'o open the, Conteinnient Spray Uc1ves (MO : ' C,-1H+)
after a LOCA.
( 1. 0 ).
j v
1 j
- h..
. What is the purpose'of the 'Centainn.ent Spriy Vrlve Control Reylock Switch"?
(0.5) 9UEElION 6.0?
(1,00)
The main turbine is at 1800 r p ra in preparation for synchronizing the ncin generator to the grid.
What will heppen if the "all valves closed' pushbutton is depressed?
(Choose the correct answer from the following.)
Nothing.will happen since the synchronous speed select signal is-j sealed in.
b.
The turbine control valves and mein stop valves vill close, but the intercept VElves will t e ni c a n open.
c.
A]1 of the control valves (TCVs and IVs) and u.ain stop valve.; (h0Vs) uil) e]ote.
d.
T h r-contrel valves (TCVs and IVs) will close, but the main stop volves will ren sin open.
i i
1 l
GUE0 TION 6 10 (2.50)
For EACP of the follouing conditions, state whether a scran, half-scrcm, rod block, or no action it generated. -For c o n d i t i o n s, thet p r oduc e nior e than one n e t :i o n. 5 !.a t a the mote severe action (i.e. half-scram is st r i e sever e ther, a red block).
Consider logic only.
o.
Turbine t.t i p et 20% pouer c.
thin stean lines D and D 1solate, Mode suitch in RUN d.
APRM B ups. calc Mode % itch in R,UN
)
e.
S c r a ni discharge volume level is at 60 g a l '. e n c, Mede
- r. witch in i
STARTUP I
1 i
( n t:0 CATECORY 06 C O N T IN L' E D O t! NEXT FACE antx) j i
)
[:
y
-su_ EL6til SISIEDS_DESIGth_G0t!In0Li_6t!9_It!SIBUbEUIBIIgU PACE 11 i
1
' 6.1 1.
(3.00).
QUESTION I
'An automatic RCIC initiation,has occurred.
Subsequently, RCIC-injection was aulcrastical terminated'due to high reactor water level.
i e.'
What con:pon nt in the RCIC. sys t,em f unctioned as
- a. result of the 1
L systen. logic'to autonistically terminste the injection?
(0,5) b.'
Assumins no operator action. how will RCIC respond to a-subsequent continuously decrec;s.ing water level?
(1.0) cc
-If a.RCIC 'Turbfne Test.' had been in progress wheri the initial L
seotomal.ic initiation pignal had been received,-how would the system have - responded?
(1.0) d.
If, following the initietione the RCIC.tur bine had tripped on
'I overspeed. (125% of rated). could RCIC injection be restarted fron, the Control Roon?
EXPLAIN.-
(0 5) 1 l
l QUE51 TOM 6 12 (1 00).
l W h o.l-w i l l happen to the 250 V battery system if the 200 volt b a t t e r ', -i s disconnected and only the charger is sup,tlying bus voltage?
i
e' i
r i _ _t E D G E D U B fi S _ : _ U D S t! 6 L t. 6000 B b 66 : _ E t! E E E E U C 1_600 PAGE 12 l
S69106901G66_G9UIS96 i
QUESTION 7.01 (1.00)'
'Enplein why,.froni the attached Heat Ccpacity Temperature Limit (llCTL)
Curve (figure 31' fro'm EOP-2, Primary Containment Control, the shaded etee does not include' values.less thcn 135 psis.
-(TWO reasons required)
( 1. 0 ) -
o OUESTION 7k07
'(3.b0)-
A transientDhas.oecurred wlith 'a resulting reactor s c r a n,.
Howevery-several j
rods-tencin stuck out-beyond posiLion 02.
(Doron irijection 'is not
- required.)
At the same tinie,.al] Reactor Pressure. Vessel (RPV). water 1ebel-indication'is lost and thus RC/F 'RPV Flooding" (Contingency 461 is l
entered.a RC/P-0, ' Enier genc y Depr e ssur iza tion' (Contingency 42).is be:ng l.
-performed concurrently, i
- a '.
'Euplain why, 'under these circumstances, the flooding procedure first requ2res.thet all injection into the RPV be. terminated and prevented iscept.that :from Doron ;1njection and CRD.
(Full exp.lanation require $
rees:oij uhy injection termineted and why Boron Injection and CRD Jem eaceptions.)
(3.0)
' b'.
'Fuplein why, efter injection is fin; cortr.enced under these circumstances, the floodin>; procer equires that RPV pressure p
'be'ncintained e bt v e the Minin um native RPV Flooding Pressure L
(MARFP) but as low as practical.
Anust explain both for full l'
tredit.)
(1.0)
I l.
[
c.
Euplain what a suddenly decreasing RPV pressure would indicate i
following comnientement of injection'and establishnient of the
- MARFP, (0,5) d.
E::pl ai n why the prefer red systems fo* injection i r, this P r o c e d u r -- ute u
Condenuste/Feedwater, RHR, and CPD.
(TWO renons required.)
(1.0) l
'90EDTION 7.03 (P.SO)
I c'
I P 01 - 74 n'm tup, requires that :e n.nspection of the drywell
- ,c sten
}
tunnel be per f or med.-
List the FOUR (4) conditions that must be net pi ior t r.
drywe)) entt y.
('24 01 4
b, I'
an emergency entry beyond these restrictions is required, who nu-'
i j,
eppt'OVe Ihis type of enlty?
(0,6) t-I 4
1
-(n*n PE Er0RY 0:7 C(!NTINUrD ON NEXT PAGE n**x)
)
l-3 L
)
L __EEDGt:0UBEE __UOBb6L1_6SUDEb6L1_EUEEEEUGl_6UD PAGE 13 E6010LO91066_00dIBOL Q UE S110 tJ 7.04 (1.50)
L13t the THREE (3) reasons why an oFerator would went cooldown rater, sbove 100 deerees F/hr durins Emersency OperaLing Procedur e usege?
(1.5) r DUf.STION 7.05
( 1.j50 )
l I r' eccordance with EFIP 1.2, Notification, the following are the responsibilities of t' h e Operctions Shift Supervisor (answer each as either TRilF or FALSE):
a, Advice the Security and Support Supervisor upon completion of all j
te.:;uired notifications and appraise hini of problems encountered.
b.
Verify that the NRC is contacted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of initial doelaration.
-e.
-4.-a.11..c v en tw e l - % i r i ad--we r n L F P T - tir-tagrirr, v er rf y--t ha.t.
Cont 3 Duct'" 'c O m m o r t i c h< t i o n" h?Vr bron cst ?bli 5he d-tei th the -NRC. '
f 1
QUE"il.ON 7;06
/1.50)
.In necordenc e with EPIP 1.1, Determination of the Emergency Action Level, what THREE (0) i t e n, s are required to be logged in the Shift Supervisor's Lo3 upon declar ing the emergency classifiestion of an eventr
{
l'
'GUELTION 7.07 (1.50) luo e6119 of the 250 volt DC batter y sis ta, are found to have e fin ~s volt gie 2 e xt than ihe required 2.13 volts.
Ones this rffeet the overisbi1ity of this mysten, and other plant synteme?
Enplnin your answer.
I (nvra CATFGORY 07 CONTINtG NN NTYT PACE * * *u )
i I
.q 1
17'..
PROCEDURER - NOPH AL, ARNORMAL. FMERGENCY AND PAGE 14 S6010LDEIGOL_G9UIRgl 1
(
j 1
N 7.00 (2.50F
-00ESTION LWith. respect to Opereting Instruction No. 261, Resctor Water Cleanup l
S y t, t e nr ;
.a' State.the raininoan'sp eed at : which the RWCU pump speed
.euntro)lers shoold be set during pump startuP.
(0.51 b.
Wh'at would heppen if.they'were set at a lower setting?
(1.0) c.
What'would happen'if the RWCU pump is not secured prior to-q transferring between local and r emote operations?
(1.0)
QUESTION 7.09 (1.50)-
A.co.ndition arises that requires the entry of an operator into a HIGH RADIATION AREA.
The operator will receive an estimated whole body dose.of 50 mren.c You.have tho following data svailcble about the three (3) candidates for~the-job:
Candidate 1
2 3'
Age 2-30 24 Exposure:
Wee!
15 mrem 35 meen 5-rtirem Ducrter 2954 m r e n.
?G7 n r e r.,
16 ner e m j
Life 10000 m r e n.
45720 mrem 29995 mrer.,
Rem #rkr!
NRC Forni 4 on File Y E. C YES NO
, Authorized 5000 mrem /yr 1000 mrem /qtr 150 mrem / day F,sch candidcte is techniccily competent and physically cepsble of p e r f c, r m i r.g t h e t asl.
Emer gency limitt do not apply and time conttreints do not permit authorinction f o r-an administrative exponore limit jnerease.
EXPLAIN the reasons l' o r ececpting or ocjecting each condidate.
l a
GUESTION 7.10 (1.00)
Oporeting Instruction No. 350, Peactor Protection-System, notes that in
{
the event that both RPS Dusse rw s t be shut down, a]1 errort shov:d be nd.
to restore at l e a s t, one PPS Dos to oper st r on and recet the s c r a i-n 80 p D F, b i b l e.
Whet is the ietson for thle?
I
( n y :n CATECDPY 07 CON 1: Nuth DN NLXT PAGF n tx v )
1
L l
~E1__C89CE998EE'_~_U98d6LI_bCUO80bk1_EUE89EUC1.6UD PAGE 15 P_ A_ D I G_ l. n G_7_ C_ A_ ! _ C O tJ T R_ O_ L_
l l
lL l
l OUESTI0tt 7.11 (2.001 With resti d to'the RHR Systen, and Opercting Instruction No. 1497 Reuidual l
' Hect Renion1 Systeni; c.-
DI 149 eautions the operator to slowly lower HX shell side water level during initiating of the Steen, Condensing Mode.
WHY?
b.
DJ 149 cautions the operstor t, h a t the LPCI Dutboard Throttle velve should be throttled open within 10 seconds after starting _en RHR p u nip for
- t. h e S h u t d o w r's Cooling Mode.
What is the concern if this is not done?
QUESTION 7.12 (1.50)
The ECCS Opercting Instruction Procedures caution you NOT to secure OR i
override en ECCS Systen: initittion unless certain criteria are ni e t.
What rt ? O iheLe e i' i i, e r i O ?
QU E F) 10 t!
7.13 (2.50)
With veetrd to Operating Instruction No. 264, Reactor Recirculation Systen.'
i n.
01 264 cautions the operator to open the discharge va.!ve as a rm n t-
.possible after etcrting recire. punp.
Whrt is the basis of this pt ecau ti on?
(1.0) h.
01 T6C a ] s-n caut.Jons the oper ator that the recirc. punip shoit: d not be opereted with the p u ru p dischstge valve closed, e ::c e p t when st cling hull llte dOWh N ptim;.
Other than your answer to psrt E.-
lip of t
a b r: v e, wha
- is the beuis for this precaution?
FULLY
- e. plain your i
cn,yer, A c;,q. e i,yi e (tid ! tiig p e n t,
[5 f ri R U fJ Mode and that the other r.~ e t t e, p u u.p is runnin. n o t ni a l '1 y.
(1.5) c o t't P I n n c.1c (2.00)
Durins apietions at 100L,suwer, en exp nsion pnd mubsequent tii e in tho j
c o ri t r e '. r o o m less required immediate evacuction prior to any po:sible O p e l C t 01-LC jpp.
N A M E t w c, (g) D l t e r r. c f., e mcthodi to achieve a r e ts c t u i t
seren ent DIVE their locations.
(Panel tJ c.. not r+ouired)
'T.0) i i
( * * :r a r E" >'l i.:.
'.] F, E l i.:C. O N. Y 07 * 't * " t l
r h - _6 C U I U I SI B o II V E _ E B 90 E 0 !J E E h _ G 0001II O U h _ o u 0_ L I U II o II O U S PAGE 1.6 l
i OUEST10tf 0.01 (2.00)
F o r. +.h e following refueling occurrences,' choose the correct action.
I Recurrences
- 1..
Dr opped f ue 3 e s sen bly wit.hi n -- the rezctor vessel.
E' D anaged 'f uel assenibly with evidence of fission ges releece (buhliles)..
3.-
' Loss of Water' Level Situation.
1 A..
Unexpected Suberitical Multiplication.
Actionc i
l
-a.
Stoppsge of core component movement only.
1 b.
Evacuation of refueling floor, i
i c.
Evecuetion of refueling ricor and drywea.1.
d.
Evaevetion of refueling floor, drywell, torus, and torus ares.
r l
.1 I
GUESTION C'02 (1 50) l TRUE or FALSE?
l a.
A Juroper and Lifted Lead Clearance Form shall only be completed ptior
.j to. r enioving Jumper or Lifted Lead Tags.
b.-
For Lifted Lead r e nt o v a l, return of the Lifted Lead Tag is sufficient
.i verification that they have been reconnected.
I c.
Durin3 not m31 surveillance testing, Jumper end Relay Block ins,tallation verification is not required; however, final Jumper and l
R e l e y R ?. o r-k r e m o v a l v e r i f i c:a t i o n by qualified personnel is r e gin i e d.
'l i
l l
e I
i (wrtrr CATEGORY 08 C0tMINUr0 DN t'E K1 PAGE rrrtr)
h._ _ 6 D D I U l E I E 6 1 1 E _ E B O C E D U E E h _ C O U D I I I O U h _ 6 U O _ L I d I I 6 1 I O U D PAGE 17 outsT1oN a.03 (2 00) 1he reactor is at 100% power and the Fire Protection System Ring Hesder Fluth it in progreu when you take the shift is the Shift Engineer.
j Duririg your revine of the w o r t.
In progrecs. you note that, per procedore.
l 2 vi ' vet have been closed to allow flushing of the cociing tower looE.
M l
the time t, h i s occutred, 2 other post indicating valves on the main heedet enre airerdy closed to allow installation of 2 eiew fare hydrants.
This ituation r ewl tod in isolaLing a section of the deluge, c., p r 1 rs 1 e r, hace.
e red f.i t " ru e. i n n g s t e n.t for the las.t five hour.
F r o ta c Technice]
Opeetf' cal 1on sLandpotnt, thit ereates a problt+.
For the above t
)tuntion. d e s c r i t, e dhat actions should have been taken et the time the 2 valvec were closed to iso] ate the cooling tower loop and in what time frame (si per your technical spoeifications.
GUEE110N C.04 (2.00)
Concerning reactor coolant c h e ra i c t r y :
DAFC Technical Specificttione r e q u i. r e that the iodire c o n c e r.t r : t i c n in the t v. s c t o r cool mit shall not exceed uCi/gm of doce equisalent
'J - 131 far the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> rollowing power transient, If this l j rn. t in e<ceeded, what TWO (2) act:on a the Technical Specifications iequire the operator io t al ei b
uhen in the shutdown or refueling node and with steaning rate: 1ere t h r. r.
100,000 lbc /hr. conductivity shall rio t e :eeed 5 umho/cm a r. d chlor i<le not exceed 0.1 ppn, per the Technical Specifiestions.
With eterming r4Les greater thEn or e e;o a l to 100,000 } t, /hr, conductisil.y bhn31 not PAceed 10 umho/en, said chlorides not (mceed 0.5 ppm.
Why ate the l i ra i ta. for conductivity sod chloride al] owed to p u ;-., hen I
s t e s i. i n g ratee ere above 100,000 lbs/hr?
00"9T10N 0.05 (1.50)
'T h e lechnical Opecificcilori ntato ih follow:n3 "The ten diesel g
- f n e f' a l o r :
Lh,li be operab]r prir4
't h e r e :ha]1 he m j t i ; p t i il of 10e000 g 513.,o - er ri i o. e ' fool an th-die
..]
t iie ] oil tent.".
I umt a.
uw Teenr w spec;c m : tion ru c c f o, thu mount?
( :41 ?" *4 7 r'AT[CDOY O f) C I N I I [ : [ if+
n2' N[yl P A f' f 11WYWl
4
?
Y.
n i'
Si_ 6DdIf.! ISIS $112E_ES9GE9UBEE1_COUDIIIDUSt_6BR_LIdII61100g Pact is 1
i t
3 i
' QUESTION 8'.06 (2.50)
Yod ace making preparatiorfs to start up the reactor the reactor ;is ir; the Stertup. Mode.
The f ollouiing cenditions ar e reported
- l 1.
Control rods;10-07 and 1.0-27 have inoperable accumulators.
i (2.
The scram time,to rod position 46 from full out for control rod N-23 was 0.370 sec.; for control rod 26-19, the time was 0.372 sec.;-
for contr ol. rod 22-19. the time was 0.371 sec.1 and for control rod 22-23,<the time was 0.368 sec.
l J
i 3 '.
Control rod 26-07 cannot be roved with control rod drive pressure.
q
\\
In accordance with DAEC Technical Specifications, what actions must.be i
I taken,in this situation and how does this effect plant operctions?
s
. NOTE:
Use the attach'd sections of Tech Specs to answer this ques 11on'.
e FULLY reference all applicable sections of Tech. Specs.that you use to develop your answer.
QUESTION C.07 (1.50)
While playind softb511 on Sunday afternoon, you severely sprain your E n F. ] e.
Crutches ar e needed to repor t to worl< on Monday as the Shift Supervisof.
a.
Can you ' legally assume the shift?
(0.5)
Why or Why not?
(0,5) i I
b.
Mutt a report be sent to the NRC?
(0,5)
)
. Qllr.CTION 0,06 (2.00)
D.ue
,o vacations and r equir ed tr aining, you have been esked to unrF the
+
in) lowing schedule neNt week.
Review the schedule end identify the overtine guldelines o*
DAEC Adminis!.rt+ive Procedure 1410.1, Shift Di wanin s wn
- Oper,. tion fnd Tur nove r, that ynn would be violeting.
CN01F; Mi tiae e;cludo 9h.ift turnover t : n, e. )
Assune that the pl so' it npc1eting at 100?. pouer, j
Sunday 0700 - 1900 Mondq 0709 - 1800 Tuesdsy 0700 - 2400 Wednesdby 0700 - 1600 Thuisday 0733 - 2000 Friday 0700 - 2000 Saturday 0700 - 1900 l
1 (rr m C A T E r:n n Y CO CONTINUED Oi' ITXT PACF m ir)
I
(
_M _. _ _.A D M _I N 1 S.T R _A _T T _u F _ P R O_ _C _E _D U R E_ S _. _ C _O N_ D_ _T T I O _N G_ _,_ A_ _N D_ _ L_ I M I T A _T I_ O_ N. S__
PAGE 17 i
l OUES110N B.07 (2.50) l i
e.
In accordance with DAEC Ado:inistrative-Procedure 1406.3, Revision of
-Procedures and Iristructions, who shall cpprove'all temportry changera.
(1,0) b.
Give THDEE (3) REASONS (not enamples) why a temporary procedure revision would be issued.
(1 5)
GUESTION B.0 (4,00)
With.r egar d to DAEC Administrative Control P'rocedure 1410,5, Tescut Procedure:
a.
What is c " SAFETY SYSTEM?'
(1.0) b.
What are the required actions if a necessery Hold Card is found to be missing?
(1.5)
L.
la the event that the person whose name is on a Hold Card is not onsite and cannot be r esched, WHO can authorire a relearse of the 4 a l ci '; E r d and WHM r e e,o i r e m e n t '.
't.u s t be n.et be f 'ai e gr ant in 3 the relene?
(1.5)
I GL'EDTIOd 8.11 (1.50) 1 r c hn i c i:1 Specifications require that all jet punips mus t be opertble or be ir et least hoi shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
LIST two (2) chect-for (portbilitv of the jei p u n. p s and CIVE the reason WHY jet p u r. p o p e r, $ 1 ity i t-E centern.
(1,5)
CIII Cl 3 OI C,37 (2,n A Techn c11 Spcc.ftratior, Docr t ei 'y 9^
days) survy 11ance un<
pr r f t ted 4
on the L:ay, Pre _,1v e Cut e Injectior s y d e n.
oci mmri
, 1907, or,o to J cnuct y -
1907, was c c,.. p i e t #
A,, r 1 l 16, * "' C ?
Tir u a / c. 1 a r. m pr:
r.i n e t y (o:
hc before J; nuar y 3r 1 P 7 i
_rtvoly surveillance p': AT thr
, si MONTM and DAY the nc "
e sa r - lit perfarted ;n' h,
m,-!
S e n; -s n qi m n c c of ,: i l i n.; tr 7erfarn m'r c ill ence t II(.: p e C 1 30d i 1 11. 0
'l n t E5 T V 01
i f' Cs ' C i t.: l i d. (
a t.t C C lie d,
l ot u u m m :w n a oc n e
. x i s vrv rrn ett FND Or EKAM]NAT70p rrrx t rm ni nix '
]
EQUATION SHEET f d ma v a s/t Cycle efficiency o (Network out)/(Energyin)
. w = mg s = Vot + 5 atg E = mc2 KE = 5 mv2 a = (Vf - Vo)/t A = AN A = A e-At g
l PE = mgh Vf = Vo + at w = e/t A = in2/tg = 0.693/tg W-6 tgeff a [(tg) (t )]
b
[(tg) + (t )3 f
b 1
AE = 931 Am y, j e-zx Q = mCpat 6 = UAat I = lo ~"
- e Pwr = Wrah I = Io 10- M TVL = 1.3/p 10 ur(t)
HVL = -0.693/p s
P = Po P = P et/T o
SUR = 26.06/T SCR = S/(1 - Keff) f CRx = S/(1 - Keffx) l SUR = 26p/t* + ( e-p )T CR) (1 - Kef f j ) = CR (1 - keff2) 2 T = (t*/p) + [(a - p) 'Ip]
M = 1/(1 - Keff) = CR /CRo 1
T = 1/(p
's)
M = (1 - Keffo)/(1 - Keff1)
. T = (B - p)/(Ap)
SDM = (1 - Keff)/Keff t* = 10-5 seconds p = (Keff-1)/Keff = aKeff/Keff i = 0.1 seconds-1 p = [(1 /(T Keff)] + [Seff/(1 + IT))
'Ijdj = 1 d22 2
P = (rev)/(3 x 1010) lidt 2=1d22 2
R/hr = (0.5 CE)/d (meters) z=N R/hr = 6 CE/d2 (feet)
Miscellaneous Conversions Water Parameters 1 curie = 3.7 x 1010dps l
1 gal. = 8.345 lbm.
l 1 gal. = 3.78 liters 1 kg = 2.21 lbm 3 Btu /hr l
1 ft3 = 7.48 gali 1 hp = 2.54 x 10 1 mw = 3.41 x 106 Btu /hr Density = 62.4 lbm/ft l'
Density = 1 gm/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1bm
- F = 9/5*C + 32 Heat of fusion = 144 Btu /1bm
- C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-1bf 1 ft H O = 0.433 Ibf/in2 2
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DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l
3.3 REACTIVITY CONTROL 4.3 REACTIVITY CONTROL Applicability:
Applicability:
Applies to the operational Applies to the surveillance status of the control rod requirements of the control rod system.
system.
Objective:
Objective:
j To assure the ability of the To verify the ability of the control rod system to control control rod system to control reactivity.
reactivity.
Specification:
Specification:
1 A.
Reactivity Limitations A.
Reactivity Limitations 1.
Reactivity margin - core 1.
Reactivity margin - core loading loading Sufficient control rods shall be A sufficient number of control withdrawn following a refueling" rods shall be operable so that outage when core alterations were the core could be made performed to demonstrate with a subcritical in the most margin of 0.38 ak/k that the core reactive condition durina the can be made suberitical at any time in the subsequent fuel cycle operating cycle with the with the analytically determined strongest control rod fully withdrawn and all.other strongest operable control rod operable control rods fully fully withdrawn and all other operable rods fully inserted.
- inserted, 2.
Control Rod Exercise 2.
Control Rod Exercise Control rods which cannot be a.
Each partially or fully withdrawn a.
moved with control rod drive operable control rod shril be pressure shall be considered exercised one notch at least once inocerable.
If a partially or each week when operating above 30% power.
This test shall be fully withdrawn control rod performed at least once per 24 drive cannot be moved with drive or scram pressure, the hours in the event power reactor shall be brought to a operation is continuing above 30%
shutdown condition within 48 power with two or more inoperable hours unless investigation control rods or in the event demonstrates that the cause of power operation is continuing the failure is not due to a above 30% power with one fully or failed control rod drive partially withdrawn rod which mechanism collet housing.
cannot be moved and for which control rod drive mechanism damage has not been ruled out.
The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housina failure is not the cause of an immovable control rod.
Amendment No. 19 3.3 1 3/76
,7 LIMITING CONDITION FOR OPERAT10N SURVEILLANCE REQUIREMENT b.
The control rod directional b.
A second licensed operator
~
control valves for inoperable shall verify the confonnance tr.
control rods shall be disanned Specification 3.3. A.2d before a electrically and the control rod raay oe bypassed in the Rod rods shall,,be in such position Sequence Control System, that Specification 3.3. A.1 is met.
c.
Control rods with inoperable c.
Once per week when the plant is accumulators or those whose in operation, check status of position cannot be positively pressure and level alarms for determined shall be considered each CR0 accumulator, inoperable, d.
Control rods with a f ailed d.
Once per quarter verify that:
" Full-in" or " Full-out"
]
position swi.tch may be (1) the Scram Discharge Volune i
bypassed in the Rod Sequence (50V) vent and drain valves Control System and considered close within 30 seconds operable if the actual rod after receipt of a close position is known. These rods signal, and must be moved in sequence to their correct positions (full-(2) after removal of the close in on insertion or full-out on signal, that the 50V vent and drain valves are open.
wi thd rawal),
Once per month verify that the 50V vent and drain valve position indicating i
lights located in the control roan indicate that the valves are open, e.
Control rods with scram times e.
Once per operating cycle verify j
greater than those pennitted that:
by Specification 3.3.C.3 are (1) the SOV vent and drain inoperable, but if they can be inserted with control rod valves close within 30 drive pressure they need not seconds after receipt of a be disarmed electrically.
signal for the control rods to scram, and (2) open when the scram signal is reset, f.
Inoperable control rods shall be positioned such that Specification 3.3. A.1 is met.
1 05/87 Aaendnent No. 143 3.3-2 l
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT In addition, whenever the l
1) reactor is in the startup or run mode elo more than one control rod in any 5 x 5 array may be irfoperable (at least 4 operable control rod', must separate any 2 inoperable ones).
If this Specification cannot be. met, the reactor shall not be started, or if at l
power, the reactor shall be brought to a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2)
All rods within a notch group containing On inoperable rod will be positioned within 1 (one) notch of the inoperable l
rod whenever the Rod Sequence l
Control System is required.
g.
During reactor power operation the number of inoperable control rods shall not exceed-8.
Specification 3.3.A.1 must be met at all times.
B.
Control Rods B.
Control Rods 1.
Each control rod shall be 1.
The coupling integrity shall be coupled to its drive and have verified for each withdrawn rod position. indication control rod as'follows:
available for the " full in" and " full out" position or a.
When a rod is withdrawn the completely inserted and the first time af ter each refuelino control rod directional outage or after maintenance, control valves disarmed observe discernible response of l
the nuclear instrumentation and l
electrically.
This i
requirement does not apply in rod position indication for the
" full in" and " full out" the refuel condition when the reactor is v.ented.
Two position.
However, for initial rods when response is not control rod' drives may be removed as long as discernible, subsequent exercising of these rods af ter Specification 3.3. A.1 is met.
the reactor is above 30% power shall be performed to verify instrumentation response.
{
b.
When the rod is fully withdrawn the first time after each refueling outage or after CR0 maintenance, observe that the drive does not go to the overtravel position.
03/76 Amendment No. 19 3.3-3
\\
(-
OAEC-1 LIMITINGCONDITIdNFOROPERATION SURVEILLANCE REQUIREMENT c.
During each REFUELING OUTAGE l
t observe that any drive which has been uncoupled frain and subsequently recoupled to its control rod does not go to the overtravel position.
l 2.
The control rod' drive housing 2.
The control rod drive housing support system shall be in support system shall be i
place during REACTOR POWER inspected af ter reassembly and OPERATION or when the reactor the results of the inspection coolant system is pressurized recorded.
I above atmospheric pressure with fuel in.the reactor vessel, unless all control rods are fully inserted and Specification 3.3. A.1 is met.
3.a Whenever the reactor is in 3.a. Prior to the start of control the STARTUP,or RUN mode below rod withdrawal towards 1
30% RATED POWER, and the criticality and prior to
, control rod movement is within attaining 30% RATED POWER the group notch mode af ter during rod insertion at 50% of the control rods have shutdown, the capability of the the Rod Rod Sequence Control Systen to been withdrawn,l System (RSCS) properly fulfill its function Sequence Contro shall be OPERABLE.
If the shall be verified by the system is determined to be following check:
inoperable in accordance with checks in Specification Group Notch - Test the six 4.3.3.3, power may be comparator circuits. Co increased above 30% RATED throu-each comparator n!
P' ~R by increasing core flow.
inhit initiate test, ify
'l erre j reset. After b.
aver the reactor is in comp checks initia. test STARTUP or RUN modes and
/e completion of cycle t
aw 30% RATED POWER the indi
. by illumination of Rwa Worth-Minimizer (RWM) test alete light.
shall be OPERABLE or a second 1
Reactor Operator shall verify b.
Pric
] the start of control that the Reactor Operator at rod aithdrawal towards criti-the reactor console is cality and prior to attaining following the control rod 30% RATED POWER during rod
- program, insertion at shutdown, the i
I capability (of the Rod Worth Minimizer RWM) shall be 1
c.
If either Specifications 3.3.B.3.a or.b cannot be met, verified by the following the reactor shall not be checks:
started, or jf the reactor is in the RUN ce STARTUP modes at 1)
The correctness of the Reduced less than 30% RATED POWER.
Notch Worth Procedure sequence control rod moveraent shall not input to the RWM computer shall be permitted, except by a be verified.
Limited control rod l
i movement is permitted for the purpose of determining RSCS or RWM OPERABILITY and shall be verified by a 'second Reactor Operator.
l Amen'dment 142 3.3-4 05/87
DAEC-1 LIMIT!NG CON 0lT10N FOR OPERATf0N SURVEILLANCE REQUIREMENT J
2)
The RWM computer on line diagnostic test shall be successfully performed.
3)
Proper annunciation of the selection ' error of at least one out-of-sequence control rod in
)
each fully-inserted group shall
]
be verified.
)
4)
The _ rod block function of tne RWM shall be verified by withdrawing the first rod as an j
out-of-sequence control rod no more than to the block point.
l 4.
Control rods shall not be 4
Prior to control rod withdrawal withdrawn.in STARTUP or '
in STARTUP or REFUEL modes, REFUEL modes' unless at least verify that at least two Source.
q two Source Range Monitor Range Monitor Channels have an l
Channels have an observed observed count rate of at.least count rate equal to or greater three counts per second.
than three counts per second.
5.
During operation with Limiting 5.
When a Limiting Control Rod
~
j i
Control Rod Patterns, either:
Pattern exists, an Instrunent Functional Test of the RBM a.
Both RBM channels shall be shall be performed prior to l
OPERABLE, or withdrawal of the designated rod (s).
b.-
With one RBM channel inoper-able, control rod' withdrawal
'l Shall be blocked within 24 l
hours, unless OPERABILITY-is restored within this time
. period, or c.
With both RBM channels inoper-able, control rod withdrawal shall be blocked until l
OPERABILITY of at least one channel is restored.
p k
i Amendment 142 3.3-5 05/87 l
'l L
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT l
l C.
Scram Insertion Times C.
Scram Insertion Times l
'1.
The average scram insertion 1.
After each refueling outage all time, base on the de-OPERABLE rods shall be scram
'l energization of the sera time tested from the fully pilot va'Ive at time zero, of withdrawn position to the l
all OPERABLE control rods in drop-out of the reed switch at i
l l
the reactor power operation the rod position required by l
condition shall be no greater Specification '3.3.C.
The nuclear system pressure shall l
than:
be above 950 psig (with l
Average Scran saturation temperature) and the 1
Rod Insertion requirements of Specification
{
Position Times (Sec) 3.3.B.3.a met.
This testing shall be completed prior to l
46 O.35 exceeding 40% power.
Below 30%
38 0.937 power, only rods in those 26 1.86 sequences (A12 and A34 or 812 and 8 4) which are fully 06 3.41 3
withdrawn in the region from 2.
The average scran insertion 100% rod density to 50% rod times for the three fastest density shall be scram time control rods of all groups of tested.
During all scran time four control rods in a 2 x 2 testing below 30% power, the array shall be no greater Rod Worth Minimizer shall be OPERABLE or a second licensed l'
than:
operator shall verify that the Average Scram operator at the reactor console Rod Insertion is following the control rod Position Times (Sec) program.
46 0.37 38 1.01 26 1.97 06 3.62 3.
Maximun scran insertion time to rod position 04 of any 0PERABLE control rod should not exceed 7.00 seconds.
l l
l 1
l l
Anendment 141 3.3-6 03/87 l
l
(
d D:__ISE0El_DE_UUCLEGB_E9EES_CLGUI_0EES6II90:_ELUID01_QU9 PAGE 20 10560221U00103 ANSWERS - DUANE ARNOLD
-07/07/15-HARE, E.
A, t
l
)
'ANSHER
'5.01 (1 00) l d.
REFEREf4CE DAEC RXTH-SH-16, pg. 2-2 292003M110
...(KA*S)
ANSWER 5.02 (2 00)'
True.
= Po e (t/T)
Usins the equstion P solving for time results in the equation:
t-=T,<
In(P/Po)
Fron this it can be seen that cince S/1 yields the same value at 50/10, and since all other factors i r. th( egostion are equc]e the tinie is o c3m l.
(2.0)
REFEREllCE DAEC Foset.or lheory, Period E.quation, Pg. 2
-292003K100
...(MA'S)
ANSWER 5.03 (1.00) b.
REFEREllCE DAEC Heat. 1 r aresi er 14, Miti 3cting Reactor Core D a n.a s e 344 203009K107
...(MA'S)
ANSWER S.0a(1.00) a.
REFEREfICT DAEC RXTH-SH2;-t.
???006K107 4 292006M103
...(KA'S) l
i
~~-- PERATION,O -------- FLUID'in AND PACE 21 Pl. ANT
- ilCl.FAR POWER THFORY OF N 81.
1 l'
1 H F R.t.1 0 D Y N A M'l C S ANSWER? -- OtIANE ARNULD
-87/07/15-HARE, E.
A.
l l
1 l
l ANSWER U.05 (1.00)
.j l
c.
REFERENCE D AEC R):1 H-SH-19e r. pg. 2
?92000K107
...(KA'S)
ANSWER 5.06 (1.50) a.
Padiolytic decomposition of water.
(0.5)
'b.
hotfl (fuel claddin3) to uster r e T,e l i o n or Zr - Water reaction.
(0.5) c.
0 fgon - ';
Sydrogen - 47.
(0.5)
REFERENCE Ouane Arnold Mitigating Reactor Core Damage, pg.
6-5, 6-10, 6-46 220001A204 223001A205 227001M404
...(KA'S)
ANSWEP 5.07 (2.50) s.
Doppler LO.333 is the f'i r s t to add nogstive reactivity.
The increase 1
in pouet 1.w e l causes a rise in fue! t e nip e r a t u r e and a corresponding
=dditico of riog t ti ve pn etivity due to doppler e f' f e c t E 0. 5 "I,
- t..
Mcdelstor T e n.p e r a t u r e Coefficient [0.33] is ne;4t cs there is - tarm
@] ey (fuel '; 1 rw constant) for the heat generated in the fuol pellet to recch the coolant in the channel and cause the t e n> p e r a t u r e it inr:r m e
[0. Z.
c.
Void Cneffi :ent EG.337 is last e 4; the c.ader c t or tempeistore hc >
- t. -
increase to the po3nt of natur mt ion befor e voids are f o c n.e d i r,
- , pr ec intale e,u a n i i t y.
(A1 o aerept if there are no v o.i d s e, c li ii nt;riop' C0.53.
l r
RFFEPrN[I "7,
pyg, 4 and 5 DA((
ov',.. g y.
29200W'.14
...(KA'S)
._ _____________ - _ a
l 5
THrnRY OF N!! CLEAR P O L' E R PtANT nPERATTON. FLI)TDn, AND PAGE 22 1SEEdODit!0blCS UlSWERS -- DUANE ARNOLD
-07/07/15-HARE,'E.
A.
4 1
ANSWER 5.08 (2.00) i a.
DECREASE j
b.
DECREASE c.
INCREASE I
d.
INCPEASE (4 0 0.5 pts each) i REFERENCE i
DACC RXTH-SH-29, pg.
4, SH-27, pas. 4 and 5 292005K104
...(KA'S) l ANSWER S.09 (1 50)
Peripheral rod worth will increase.
(0.5)
High Xe concentration in th_
center of the core (highu:t power tm f o r e r.cr w.)
will depress the t h e r ra a l neutron rive to t h e, t region c :; u s.i n s t h e relative neutron flui:
"t n the 3 er. thus incterce the reletive rod worth in peripher01 bundlet to be hi h
peripheral rud.
(1.0)
PE r E r& N CE DAEC. RXTH-SH-29, pg, 2 292006K107
...(KA'S)
Al'S W E R 5 10 (1.50) 62 seconds (0.33) e.
BOL:
t = F: - P - (.0072 -.001) a I
o P (0.1) (0.001) l' EOL; t
(0.0055 -.001)
= X <;econd-
.'O.331 (0.11 (0.001i Ll5 I
CHANGE 62 37<
'><C, s e c o n d c (0.30) t')
j i
b.
Evildup of f'u-TOO coupitc Lith the burnout of U-235 ccuses detteGsr in t;m ef
- ec i i cr ch 11 'j e d neutr on friction (Deff).
(?.5) l EFFPEt:r p DAIC b'e m e t o r l heer y.
SH-TO. p ; -. B 292003U306
...fMA'B'
1 L __lSE0BX_9E_UUGLESB_E9BEB_ELQUI_0EEB611901_ELUIDS:_600 FAGE 23 10560901060100 A N C U E P'8 -- DUANE ARNOLD
-87/07/15-MARE E.
A.
ANSWER 5,11 (1.50) a.
This design utilinen a layer of circoniun mettl bonded to the inner 5
. u ;., s turface of the c3odding.
-T-;il e m.
iii..
1L.
ui
' i ewh-4+e t w e r n the - ; 11u -
aJm. ;. [(). 7 5 tM b.
W ) t h e-full core load of barrier fuel, it is expected ths.t ope r s.t i o-La tsrgeted espor.vres without pellel-clad interaction failures or operrtionol con: tr eint s on losd swings (preconditioning) can be d C h 3 F. / 0 0,.
f h /.r j 3
FFFERENCF DAEC Systeit. Description A-4, pg, 0-9 TV300?M101 293009K137
...(MA'C)
A N 'fV 5.1?
(:.50) si,
lotnsition boiling n: a y occ ur CO.5J.
Th.s could lead to It 'e 1 d a ll : 3 H EO.5'.
h.
10 c, s t. e the MCPR l i n i t ro a r e conservative to account for the potm;bility of a sudden flou i n e t-e n c e e n c' the resultant pouer incierse.
(1.0) c.
Increase:.
(0,5\\
ram r tE N C D DAF.C - T her m odyn s n i t
d!7 degt ne r
400 degreet
=
r u,
ean1 wnc m F,
- R E F E R E t.'CI:
't h c r m o d y n t sico 10-1 2 ? '1 0 0 3 K ', ? '
...(KA'F)
r___-_-_-__--__--_-
i 1
E. _. _ _ _T H E D _R _Y _ O_ F..PM J _C *L _ A R _ P O W E_ R_ _ P_ L A N _T _ O_ P E_ R A _T.1. 0 N_ _, F L_ U_ T O S. _, _ A N_ D 'PACE 2
E__
T_ H_ F P_ M_ O_ D_ Y N_ A_ h l_ r_. G_
ANSWERS'- oDUANE ARMOLD
-07/07/15-HARE, E.
A.
ANSWER 5 14 11.50) a..
Condensatt depression or subcooling i
l b..
Helps pr ovide NPCM.to e ondero. ate. pumps to prevent cavitation
- e.
Reduces plant-efficienc'/ --(this.is sdditionzl ha'at energy that
'f n, u s t be ptovided by the r e a c t o r )'.
l
- (3 0 0. 5 e a c h. = 1.51' REFERENCE
-LDAEC Heat.1ransferr-pg. 16-15 293003K116 2730061:110
...(MA'S)
I l
ANSWER 5. '. S.
(3.00) n.
Pump shul-off head it 'the pimp he d ct uhich the m',intainable flow rate is' reduced to nero L.1. 0 3.
It is undesi r able because the pun.p overhsats and nicy eesn'1L 1n rnochanical d;; mage to the pump [1.0:.
'b,
'Provided with e ninimun, flow recirculation valve and return line.
~I (1 0)
L.,.41 w q J. rc 1
l Cr
' b' 3 -
- } ! '
. REFERENCE
',i
' 's ' ' i DAEC.1hermot$4iamicsr Hert Transfer, and Fluid Flow, p3 4-5 LU%A i\\'
i s
DAEC. System Description C-1, pg. 47 293006H1?5 293006l(117
...(KA'S')
l l
l 4
'f
s __ELOUI_01EICb2_0ESIGU _00UIE06:_000_IUSIBudEUI0IIOU PAGE 23 ANSVEPS -- O ll t.N E ARNOLD
-87/07/15-HARE, E.
A.
f.NDUER 6.01 (2.00) 1.
Roroeling platforn, pos:.tioned ne ar or over the core.
'2.
R e f ue ' s rig p i s t f o r n, boists are fuel-loeded (fuel grapple, f r a ca e mounted h o i s ',, monorail courtad hoist).
'3.
Il P O I SPeppIO fiO ! fil 3 up 6-9er"L e platforn$ hoist fuel-loaded.
5.-
0 4e rod withdrawr..
(4 of 5 t u,uired 0 0.5 pts each = 2.0)
REFER TO PAGE 25A REFERENCE DAEC Technical Specifications 3.9.1, pg. 3.9-6
?34000A302 234000K502
...(KA'S)
ANSWER 6.07 (4.00)
Ceuces reactor level ta increcse E0.33] due to the level control syston having a level error, level set indicated leve] [0.53 resulting in the feedwater control valves to open to ni a t c h new higher level Co.53.
q l
b.
Pe+ctor level should remain constant E0.33] because the
'D" M/A transfer e,tetion will lock up E0.5].
The
'A' feedweter control vcive will control leve' E0.5].
c.
Cruso reactor level to decr ease E0.33J due to the level control i
.. y r, t c.. nssing a stear:, flow / feed flow error, stean flow feed fins 1
[ 0. 5 'J resulting in the feedwater control velves to close to netch I
ne : louer level E0.51.
REFEf'ENCE
)
O f 'I t' By-tem Demeription p;.
6, 20-?2, Figure 5, 6
l 250001K108 259001K109 259001K301 259001K302
...(Mo'"'
i PAGE 25A MODE switch in REFUEL and:
1.
Trolley mounted hoist loaded with platform over or near core.
2.
Frame mounted hois't loaded with platform over or near core.
3.
Fuel grapple loaded with platfortn over or near core.
4.
Fuel grapple not full up with platform over or near core.
S.
Not all rods in and selection of a second.
6.
Service platform hoist loaded MODE switch in STARTUP and:
7.
Refueling platform over or near core.
8.
Service platform hoist loaded.
I j
i
u I
6:__EL6dI_SXSIEd3_2ESI901_C9BIE9L _600_IUDIBUUEUI61 ION PACE 26 ANGUER" -- DUAME ARNULD
-87/07/15-HARE.-
E.
A.
ANSWER 6.03 (2.00) e, Operating on unloaded diesel increases the air blower temperatures to the manimum operating value due to decreased air flow. and ruy result in blower damr,ge.
( 1.' 0 )
or l
Corbon-rich cortbustion products could collect in the E D/G exhaust por ts and present a comburtion hazard.
.. f o e, s i.7Jr 4 a r.Tb U t t:cM-d-ewar..
(l.0) b.
M, c c, ' a c~
,>e Ly& 7 REFERENCE 01 324, p g s,. 6 and 7 DAEC System Des c r iption G--2 pg. 33 264000A203 264000M401
...(KA'S)
ANSPER 6 04 (2.50) s.
D e c nu -
t.he regenerat.ve HY h o c.
no flow through the secondary side, bit.edcun flow muct be 3 i rr. i t e d to t h c-capacity of the NRHX in or de-to prevent overheut!n3 of the ite:.ine r v l j r er bed.
(1.0) b.
E:l o u d o u n is used der i r g stc t teap-heatup. or hot standby ope r a t i ore; f0.252 to rednee reactor water inventory E0.253.
c.
In order to p r ev erit draining of the systen. CV 2729 c l o c. o s on lou p r e r.. s u r e sensed up;trean. (5 psig).
This pr events drainir.3 the RWCll s y _ t e r. to nmin condent.er in the event CV 2729 is not fully c.losed.
(0.5)
A3 a rlosec c 146 p:13 u'nsed downstrese, to protect the low pressure piping, prevents pcscable over p r e s s u r i :' e t i o n.
(0.5)
REFERENCE DAE System Description E: - 4.
,r g.
37 13 19 204 000t 10'?
204000M402 204000U407
...(MA'S)
I
- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - I N. S T R t J M F U T A T i. n N PAGE 2-A.
PtANT SYSTEMS DFSIGN, CONTROL, AND
-ANSWERS -- DUANE ARNOLD
-07/07/15-HARE, T.
A.
I 1
1 ANSV?P 6.05 (2.00)
The s y n i e r-diccherge baron injection i. s limited such that the reLe o0 increene in the concentr a tior. of naturel boron in the princry coolant oster is 'ast enough to o v e i r !. d e the rete of reactivit.y insertion caused by cooldown of the reactor following the xenon poison penP. [1.03, e rrr m.,+~gi i;.. r. ' ' e4 h t-- t het9. -i n - ttMimnt crm i > J w baron dcot nct r c c1 r.cul ate - tiu ough ta,e cor e in..en. vs n conc +n t.r et i one. thrt could portib1y.
c a v :> n nuclesi power
- 4. o rite end fn11 cyclice}1y C,1, 0 ]..
):
L mb v(.cno. y ;Vi.;..
- V. t.
-e REcCRENCT DAEC Systeni Description C-4, ps. 9
~h c.
v41
-"k i 7, r-211000A107 211000K405
...(KA'S>
ANSWER 6.06 (2.00) a.
(2)
(1.0) b.
(2)
(1.0)
REFERENC
DAEC System Description A-2 pg. 64 202002A106 20200?A107 202002K603
...(MA'S)
ANSWER 6.07 (1.50) 2.
If the field brealer is not shot within 15 seconds efter the p u m,n cequence is started.
(1.0)
(0.5) b.
The recare MC dr ive n c + or bred er rk - it -krip wn \\ \\
PEr F r7 pr'E DAEC Rytten D, :. e r i p t i e r: A 2 pg. 3?. TsM o 1,
Figure. 16 7 0 "l 0 0 : K112 202001E'r' 202001V6N
...(MA'S)
f-4 l
l P 'd" 20 l
l
.A._ _. _ _ P_..A_ N _' _ _c,Y S _T U _M T;__nrE_I__CH_,_ r_ n N T R O L _,_ A_ N I)_ _ _ T H _':. _' r ' ' I t_F_ N _T A _T.10_ _f!
i A N S tF "
-- DUANE ARNOLD
~B7/07/15-HARE, E.
A.
A N E;U E P 6. O ff
(^.50) a.
r.u p r es s iir e e t ei. t.e r than J p.: i g.
coverese $ nCA cign:1 pree e nt ) - reactor venel leve. i r.
'. ' 3 cor e c o v e X '? 9 ncher Com' : t i n, e ;t Sprey Valve -uitch in ne a riu o l.
i.
1 A,
'R'w.--,... -
- \\ '," T
ya r d, L U.As b a'-
i i
1.0)
( n.., ~ '
t e_
p a_ r_ h :
t b.
' t.
7110ws o;: (ning of cont a i mi en t spray,> e l v e <.. by bypassing the r eq.: iron ent for 2/3 ent e coveregex (0.5>
(
t c;
) p,c (( s@ C '
. 'v i) e-
'~ N
^FFERENCE
'~
D AF C Sys tter. Deser ip tion C -- 1 2 pg. 37 226001K10P 224n01M113
...(KA'S) t4 M c. W E ^
/3. 0 9 r1.00) c.
F: F. F E T E M C E DAFC % tern Descriptier D-11 215000A103
...(HA'S)
A l ' ?. is' E R 6.In (2.501 h c l f - s t r 2 n-b4 O
^ ion s
l'.
h iJ 1 I
- O c i ^ n d4 r c; c' hloch t'
i C r B li (0.5 g 1. 9 i i in )
pgp,
per y
07
', n i i r 5--
5
'l ' r r , s t r n. D r t - i.o t i o!, I -- ;. o 1.
D A E r.
-t
%t:ription 1-3, 4
p;. O '$
s.
^ ? O.l K 1 c '
2000K305
/1500SK101 2 3 cT 011' ' ~
2 2 'i O : e n :
(KA'S)
4.
PtANT SYSTEMS DESIGN,' CONTROL. AND It!G TRUM ENT ATION PAGE 29
)
i ANSWERS -- DUANE ARNOLD
-87/07/15-HAPE, E.
A.
. ANSWER 6.11 (3.C0) a.
The Turbine Stear. Supply Valve (MO-2404) closed.
(0.5) 4
.b.
When level decreases to the initiation level (119.5'), the 2404
)
velve will r.3 0 p e n.
(1.0) c.
The turbine test circuitry would be autonistically bypassed and 1
-the-RCIC system,would revert to f' 10 w control mode.
(1.0) d.
No [0.25].
The mechanical overspeed must be reset locally [0.252.
' REFERENCE DAEC Systen. Descript,ico E-2, pas. 17, 24, 25, 36 38 217000A101 217000A103 217000A301 217000A402
...(KA'S) l AN9WER 6.12 (1.00)
The charger will trip E 0. ". ] (on h.tgh energency load starting currentti and to all associated loads CO.GJ (both n and power would t, hen be lost G [ & <A c. L., a ct f f Qit./.s 1 j f._; 9 f.g,7,?}) &pr emaraency). G f. 3L/s K
j,. _ e.
. rr. _
C..:
a e,. n e,-
_.e.,..
c,.,
v
+-
' 3.X o REFERENCE s
1 DAFC OI 388, pg. U 263000H102 263000K201 263000K302
...(KA'8)
- i -.
l
- - - - R n C F D.t.lP F 9 i'.
P.
- NOPMM. ADNORM
..-----------------..AL.
FMERCFNCY AND PAGE 30 2
R A D I O L.O G I C M--------CONTROL AtiSWERS -- DUANE ARNOLD
-07/07/15-HARE, E.
A.
ANSUER 7.01 (1.00) 1.
Relnu this range. toru ; heat cap; city is less of a concern since the r,hutdown cooling syster,i ru e y b e used for hert r e n> o v a l.
( 0. D '>
7, S t e r n, iloo e ; t.e s duricia 3FV discharge will be suffic1ently low te preclude u n t.t a : o condensation.
(0.5)
PEFERENCE DAEC EO '- 7 NFD0 30796, Danes for En.ergency Operating Procedures; ct DAEC, pa3e '-2 295013!'104 295026A203 2950264102 295030M103
...(KA'Ci ANGWIR 7.02 (3.50)
To avoid rapid i n,j e c t.
cold unboratec' water d u r i rig dept os ufir inatiois
- 2. n d tt the restdting react!-ity encursion.
(0.5)
D o r o r, injection and CRD 11cv ere continued to achieve rNetor shULdown4 (0.5) 6.
Un; i. above MARrP to provide rufficient s t e mi flow thrcuch the o,% n S R th to ensure v4 9uv i e S TE A tt CDOLING tc an uncovered core <
(0.04 low sc c:c'te:1 to W control flooding rate and tho.
Rept p.wsible power
(,:cu r s ions.
(0.5) c.
Wo.Id indicate that water i ruddenly entering the RPV e+
rete m
that is l e r t.
then Uc rete of steam produttion and ther e f er,
7PV w ") t e r ; e' <
15 d e c r e 6 % i II',,;.
(0,5) i
'1 6' L i' t-nEceSLOTy b e C' a v i - the RPV will br c,
I U ! OT -f){ iVP t
it,r s u r i r e d p.- i c.. e to flood,. ;.
(0.5)
C z' a t : m ' injett'ne otit'ide the shToud ero P.'er' o to minitire colt wzter reactivity add:. ton.
( 0. ';)
PIIt R E l.'!' E W "; CDP-;.
p+i.-
51 t hi u Y tt P 3079(, D c4 9 Ni Crergency Operaljng Proceduret et DAEC.
5 1 /*. i. Ii r u C -$ C4 0908 I
w C.
EpPf :( i
't.'
fl C, P r J I. i [. ;
C' r o ( p r.j i i y p p y [, d ' ;L e r, i f. l. 3 C6-i thrU C 'i - 3
,L
]i ;
=
{ C E / '31 C / 1 ^
^ 9 5 0 "t 1 M 3 0 t.
To!.0?'9012
... ( L' - '
[ ], A.", Q l-( ' 7 ^
i
I 7.
P e n r'F D tlP F S - NORMAL. ADNORM A'.. FMERGENCY AND PAGE 31
~~~~EhbiU bb5EEI~Ebi}TRULI~~~~~~~~~~~~~~~~~~~~~~~~
i ANSWERS -- DUANE ARNOLD
-87/07/10-HARE.
E.
A.
1 l
l l
ANSWER 7.03 (2.50) a.
Drywell inspections shall only be performed under the following cor da tions!
l Shift, Supervisors and Operators are noti f'ied o f entry cod e n i t.
l tin es and appropriate los enteies cre mcde.
2.
The reactor' i s soberitical and ' 5% reactor power.
t 3.
RPV pressure is below 400 ps)s, 1.
No evolutions cre performed by operators that would significantly i n c r,e a s,e system pressure.
C. J. - wu, /.
, z s ' c-i (4 E 0.5 pts each = 2.0)3.
. A
.~,v u. /.-
- w. c
,..6,m'
'io
' '. b v.
'r..'
1
, ' '_ s, d b.
E mer gency eritt les beyond these restrictions must be paroved by the, P) ant Superintendent.
(0.5)
.,g,,,,
"/~~' U
~ " '
REFERENCE DAEC IP01-2 pg. 25 h~ i t I
223000C001 223001A106 2230011f116
.. (HA*S) q, Q 4
~
-ANF.WER 7.04 (1.50) l0 1L-(- lt -
- 'i',
1.
To conserve water inventory o.. I I..
L.,
,- i.
i, 1.
To protect c o n t a i nnie n t inventory i'
l '.
i
\\
l 3.
10 ::ni+ 1adacactive release to bc envi r onme n t.
(
.t.
k k$
R i i E '~' E N C E m c m -c, n
r ee c eut. co =. p3 9,
ceou on 1s
~ 2 9 0 0 0 2.c ? n,
2c'OOO2G001
..(MA'E' i
l J
- - - - - - - - - - - - - - -... N n P M A L, A R N 0 0'M A L, EMERCFNCY AND PACE 3;.
<ROCFOURFS Ei2IOLOGIO6L_COUIC0L AW4HERS--- DUANE ARNGLD
-87/07/15-HARE, E!.
A.
l I
e l
ANSLER 7.05-( 1.f 0 )
l s
False j
7 b.
rue 1
1-If e 0.5 pts ench)
294001d116
...(MA'5) l 3
l ANSWER.
7.06 (1.50) 14 The e v e r; t 2.
The t i r<. e declered 3.
Artion(s) t a k ez n REFEREtCE DAEC Eft 1, 3 -
294001M106
...(MA'R) l ANnWEP 7.07 (1.50)
Yes.
CO 53 1he 250 volt DC battery ie coneidered inoperabla
[0.5' i f' nere than one c o l ', i e out of e o r v.i c e ).
Tho F!' Cl B y t t. e rr 'ihall also be coricidered inoperable.
[ 0. r. '
PEFEPENCF D6Er n1 100, pg.
/s,
', 4 1echiict1 9pocifi cet i on ? b.E.?,c 2 /;'16 0 t) fi 0 0 '
2 4 3 O O O f; O A 5 7 / ') O O O l'I' C 1
'?'S) i l
FNCY AND PACC 73 FMERC
D t i p F 9 - n rl R M A L.. A R N O R M A L...
7.
PROCE--- ----------
R A D 'l O L O D T P. A L CONTROL ANGUER3 -- Dl!ANE ARHOLD
-07/07/15-HARE, E.
A.
ANEVEr 7. 0 f' (2.50) v.
20%.
I o. ',1 gj1] not ave-i.) r: e the
(, h r i,.i :; '
- c ( e p.
31 h
1br, i.t n p
- t. a e 1.1 f, {.
Cit r e r' C. r /My 94
- u. q Actsi i rA ~ f..,.'L 7; - i. n it e! ed -- c i ca (1,0.
v,, h ? -
r -. u : t....
.s ho
't, 4 '
s.
., o c.
The unp'will
- i off u 'w n the Mode Swlt,ch is shifte<. b e t u vw n (1.d)p 0o :: tions..
PEFET,ENCE OT 261 R V C l), ons. 41 arid 06 t
204000A401 2040094401
...(RA'S)
ANEWF R
.09 (1.50) t i a l d e *., s.
1:
Pejectc. would e'ceed the 10 CFP O e,i i a r ' e r 1 y
- 1. i :. i t Cr roli d t, 1T Rejectrde w o e. : : r! e ceed the r ein i ni t t o et i ve c'u c t t e r-2 y iieat C J UI.;
EI Accepted 9 PVen th0 Ugh 5 (U-1@) PhCeeded the 1 i m ;t !.
onlv. L.ooliec 11 ci o a n a to 3000 mrem /ott C a n d :i h t o i
c i
e k0
(
I IP.
r
(
3 j D. ' t. r R '~ I l~ R [ N C F 1 ('. ll t R 20, D M" C He : li, E !, f e, i e n Precedur-3102.1r p ;;. 2 "940011'10?
.(KA'C' AN?,Ur
.10 i;,n'.)
~'
lo la i n i ti. i r e w( si or, the CFD p j <.t ori se
'14 0)
Y.*P',
';, f I e f p 9 O' t 1, I
is L I l-i r_
U(
!% C, I (
rn 4
( A V 6
P,.,_
s s
+, + I,h'y; ' b
')
e, 6
Z.:.__CEOCE00EES_:_UDEt!66 _60UDEd6L:._EUE-SOEUC'f _6UD PAGE 34 E601GLOGIG66_GOUIBOL
+
ANSWERS -- DUAi!F ARNOLD
-87/07/15-HARE, E.
A.
i ANSWER ~
7.11 (2.00) a, lo minimize therms) shot to the HX.
' q ctor b.
The i' l u u b y ;-.
.... im m i
.2
~~
c oo l errt-ninh i., u ; j c->i.
.2
, u s._..
vu,...
<< ;, i n'
!c, o r.j.
p,,3 m
s REFERENCI
, ; pgen c.
,i c,
DrFC OI No. i4?
.205060A110 2050001'102 205000l(502 217000G001 217000M105
...(KA'C)
ANSWER 7.12 (1.50) leert tWo independent indications [0.5] misoycret, ion _in automatic Dy at '
c o n f'i r n e d E 0. 5 3.' D R E d e c;u a t e c o r e cooling is crsured C O, '~ ~.
mode is RF.FERF.NCE DAEC DI Ne 149, 150. 151 end 3 f;"
' 03000H401 206000M407 209003K40F,
...(KA'S) i J
ANSWER 7.13 (2,50) l To ptevent damage to the pumps hydrostatic thrust bear:.ne.
(1.01 l
hi l PCl 1 C40p E C 3 e c t i o rt logi(* Will m15tuhen3y E t f.u m e n O r th D 1 2 pump I
h p e r 7,1,i o ri e h i y i. s even thotigh 1 o f' the purnp di s char ge vs1ves i :;
'I e] c c e ri s;nce it osce p u rn e d e ] t e p to s,e n r e thit a p u r;.p iir runnin3-t (1.5) fEFEREU2F DAEC D) 26/1 2020012.224 202001M11/
..(MA'G1
i
- - - - - - - - - - - - ' - - - - - - - -.i. A.D tJ O R t1 A l..
EMERGENCY AffD PAGE 3D P Tt 0 C E D tlR E S. -- N O R M Al.
..7.
E8010L091G66 00'11806 - ---------------------
ANSWERS -- DU91E ARNDLD
-87/07/15-HARE, E.
A.
. ANCWER -
7,1 '4 (2.00) e.
Turn off circuit breakers (BUS A CUT 02 and BUS D CMT 02) to Power Rcnge' Neutron Monitor ing Systen. -( 0. 5 ) i ri the 1A3 switchgect roon (0.5)
.b.
Claw ai r s uppl y 't o Scr a ri Valve Piloi Air Header located in Rr, Olds.
.757 level (Col. C6) (0.5) and Vent scram air header ct
.PI-iC41/PS-1842'.(0.5)
PCFERENCE
/
. c j.
y-
~
/ '
DAEC EOP-/. pg. 17 )
295016A301 295016G006
..(KA'S) i
[
,1.- 3 L'. l%
[ 0Cr#I k D >. J MTodc$t.v
/
,e
" -5-
.['/
(
g.g
- m. o t
4
(. t, e g
lI
'g.'
g
,f'_
(
.a i'
+r
.j,
w 2.
v
~~
w'
.J n
s m
O
's
,e i
l l
l l
r
)
l i
i l
=...
1
i 1
Sz__60bIUILIB6IIME_EB90EDUEED:_G9E91110BS:_6U2_LIblIGIIQUS PACE 36 AN9WERS ---.DUANE ARNOLD
-87/07/15-HARE, E.
A.
ANSWER B.01
(?.00)
- 1. -
c.
2.
h, 3.
d.
4.
a.
(4
(*
0.5 ptc each = 2.0)
REFERENCE
- I DAEC Fuel and Reactor C o nip o n e n t Handling Procedure fi 5, pgs. 9 and 10 1
DAEC Enabling O b.i e c t i v e. FRCHP tS-3 204000A102 234000A401 234000G014
...(KA'S)
ANSWER 8.02 (1.50) j i
a.
Felne.
li.
- Falco, c.
- True.
i (3 0 0.5 pte each = 1.5) i i
REFERENCE' j
D AFC Adniin i.s t r a L 3 ve Procedur e 1410.6, Fgs.
3, 6,
7 294001l'107
...(K/'S) i ANSWER G.00 (2.00)
Fnc tlentons, if - deluse and Sprinkler system is no r Per letbnde i l 9 J
OpetBDlO. c r.
hour 3 y fire witch patro: vith por tab]e fire extinguishing e qu i pum n t iii sl c rfected area' is r e gi e.i r e d to be est.ablished with.in one i c hose s t r ti vri i rioper t ble, a fire watch p a t t' o l with hout i ', 4 0?
e, rid w i t t por table fire e :11 ngo:i ching equip a r nt shall be established within one hout until en edditiores: hose can be touted from an operable hose striion
'ljo the ef f i p p o t e c (,: d EPe6 E_1. O T.
REFEPrhCE r, A P E 1.. c h n i c,1 9;, cification ^1.13 C End E a ri d LER 86-015-00.
?i4000A105 296000G005 286000KcC1
..(PA*Ci 9
c
& _ _60 dI U I SI B 61 I 2 E _ C E 00 E D U B E 21_ C O U 91II 9 U S 1_689_ L ItJ I I 611000 PAGE 37 AtlSWERS ---DUAHE ARil0LD
-87/07/15-HARE, E.
A.
ANSWER 8.04
.(2.00)
-- a.-
1.'
Chut~down the resctor l
l 2.
Close the tiSIV's l
l-l.
' ( 2 '3 0. 5 o c c h )
l l
b.
Doiling occur.s at higher steaming rates causing dearation of.thb i
!~
recetor'. water, fhus maintaining oxygen concentration at low levels and assuring that the chorlide-okysen content is not such as would tend to induce stress corrosion crocking.
(1,0)
_.REFEREttCE OAEC Technical Specifications 3.6 D.1, 3.6.D.2,' 3.6.B Dases
'29400 A114
...(KA'S)
ANSWER 8.05 (1.50)
The na ni nivn diesel fuel sur. ply of 35,000 gallone w]Il iupp_1y nne diessel generator C0.53_for s ni i n 3 ni u r.: of seven ti t.y s of operation [0,53, REFERENCE
~
DAEC 'iechnica) Specification 3,8. A. 2 ane: Technica1' Specification BAGES 3.C.1.
264000G005 264000G006 264000M105
... ( l( A ' S ).
i 1
l i
i I
_ _. _ ~ _. _.
-- - - --.T N I S T R fi. T T u re------------- CEDilRES. CONDITTONS. AND I3MITATTONS par.E 38 l'
'p.
/, n M PRO- ---------------- --------------- -
l l
ANSWERS - DUANE ARNGLD
-87/07/15-HARE, F.
A.
(: <,
wi.
... :.. (
< t v.-
2.n 6, h a p.
u q...
WL.
e.r s..
e-
~.e me (.,,
s t
AUSWER 8.06 (2.50) 6 '),
i' 1.
Technacal Specification 3.3.A.2.e.
states that control rods uith inoperable o c c u n,u i a t o r i. shs11 be considered i no p e r a b l e 3.--
=
u ' " ertr*--?*'t?-tv et7*r cet elmt otr+cetion.
(0.75)
.a J,. )
- s. <
sed.,.c a... '
r>
. < <.. i. 13 Q.
1 cq<-
oc.
2.
Teci'nicel Specification 3.3.C.2 r.tetes average s c r a n, i ns er ti or, times 2 array shall be no "or t, h u three fastest control rom in a 2 gierter then 0.370 secs. f r o ni rod position 46.
The three fastest rods have an avsrage time of 0.369 secs and
- t. h e r e f o r e, do not affect plant operation.
-(0.75) 3.
Technical Specification 3.3.A.2.a.
staten a rod which cannot be moved by control rod drive pressur e is inopereb]e and since 26-O' le within the 5 5 array of 10-07, lechnical Specification 3.3.A.2.f.1 steter the reactor ihs]1 be brought to cold S/D w i t., h i n 2 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(1.0)
PnFRENCE DMC 'lechnic Specific.tiun 3.3
'?0'0010005
...(MA'S)
AN9WER H.07 (1.50) e.
No.
(0.5)
The bat..le ni e d i c a l qi..ialifications of 10 CFR 55 are nc longer mei.
(0.5)
(Will accept alternate wot ding such as " Unfit for puty".
j b,
Yes.
(0.5) l R E F E RE.N r F.
10 CFr 55.3, 55.41 i
0'1(441A103
... '.lf A ' 9 i
]
l l
l I
l D LIMITAT10N4 PAGE 39 i
CONDITIONS. AN
EDURES.
PROC
.' A D M.I. N I S T R A T I 'M
?
ANSWERS -- DUANE ARNOLD
-07/07/15-HARE, E.
A.
ANSPCR 8.06 (2.00) t 1,
No more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (Tuesday 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />).
, T, n n o
2.
No n, ore than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period (Mondcy - Tuetday/.
-l 3.
At } east 0 hoorn rest between. work periods (Tuesday - Ncdnesday).
- c..
No mote than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in'any 7 day period (86 hrs. Sunday-Saturday).
i 5.
No more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> stoaight (Tuesday).
(5 P.O 4 pts each)
REFERENCF D AEC Arin,i n st r ati ve Procedure 1410.1 204001A103
...(KA'S)
AN5WER 8.0" i' !'. 5 0 )
a.
A men tier of the Plant Management Stsff and an Operations Shift Supervisor.
(1.0)
.I b.
1.
lo provide guidance in unusual situations not within the = cope of procedures.
l 2.
To ensure orderly and uniform operations for short periods when to the plant. e system, or a component of a system is p e r P cir rn i n g l
r m s o r' e i not covered by evist.ing aetailed procedures. nr has i ri b C.a t t modified or PUtef tded in stich 2 m atifie r that por ti Olk7 of 62ther C s do n o l, J: p ;i l y.
0,, ; $_ ti rig p1 O(
P.
Ir dl eet Opet E t i oris durjng tfsling and it e i n t e n s:n c e +
a l
)
4 To correct procedui: ] er rot d i scove r ed dur ing per f or rc = nc e rf the procedurt bS1
~
s'
$ c ~,,.
' k2D'
.,.l<,
n I
.A
,a in (A y 3 i i i'
'v h s - a e c 0,ii l t e d
,s t 0.5 pts each) i.
,j,%
r.
+~ r g g; 7 ;,[jg, f, ;c,.,
PErEntgcr q 7.
DO"I A d ra i n i h t t e, t i V I P ' o cc e d u r e 1406.3 pg4 2.
S; snd 6 T Mis CE /' %
29"SO1A10,
...(p'S)
(b, I,
I e
- j p 4,(
,s.' /, ~ i l'). I O $ 1' V: d,.. > -s.. '
4 6
1 J
132. _ _69 b lU l B I E 6 Il 2 E _ E B O G E Q U E [ h _ C O U D III O U h _6 U O _ L I[il l 6110 S S PAGE 40 LAt1SWEPS -- DUANE ARNOLD
-97/07/15-HARE, E.
A.
1 ANSWER 8.10 (4,00) c.
Mechanical and electrical systems which have direct design featuree involving cecident prevention and/or mitigation.
(1.0) tc. -
Stop sffected work it ^
Fersonnel or safety hazard esists [0 53.
McLe a duplicate lloid Ctrd snd msth it as c duplieste EO.5:.
Note aduplicale" i n ued nent to t.he pertinent entry on the Maintenance Tagging Form E0.5J.
(Work n cy be r esumed when j
proper isolation bcs been verified.)
(1.5) c.
The OSS E0.53, if deemed necessary for safe operation or shutocun of the plant'CO 53 but only a f t, e r investigating and assuring l
iii n /her t.el f that there is no danger to personnel and that othei i
- sersonne] involved have been informed'[0.5].
(1.5)
REFERENCE DAEC Administrative Contr ol Procedure 1410.5 294 0 01 M 10';
..,(l(A'S)
ANSWER 8.11 (1.50)
Operaba2aty:
i a.
Either recirculation loop flow differs (by i 15%) from given I
epeed/ flow cheracteristle.
i b.
Indicated total core flow differs (by 10%) from the vtive me t siir ed f t cn loop flow nie a s u r e r.i e n t s.
c.
O P f '.i :. u. - t o - 1 c L e r p l e m.n.
de] ts P of cny individu 3 jct pump differt f' m norn. (by > 10%).
l (N,y 2 of the above P 0. f.
pts each)
% +.us' s
Eli t.' s fE C r o s' T, 0 C u onF OW !! T p i Or It0WdOWn d t.> r-i n g L O C (~i.
(0.5)
$ f f l h Q k [f L O I' % $ ',
<*'ll.
' $ t.
1 t,
s.
r;t" PEFIRENCE
- A,' A /,.
(
' l-
.('
o s.,
OnEE TechnicT1 Soerifications,. Section?"' 3,6,E.',
C 6 E.1, 9.6.'E.Av
~
- y, s
,.(d// IA
'['
^
~
3nd kt(. h.
L., y h V
.f
- 0eo0 w e:
20:wm 1 c 3 ror em. c u r020n m ia 20 cm e
.. 01M'94 I
1 l
)
i i
f
- t.
ADMIN 1GTRATIVE PROCEDt!RFC, CnfJDITIONS, AND LIMITATInNC PAGE 41 A NS M E R S,
-- DUANE ARNOLD
-87/07/15-HARE, E.
A.
ANGUEi 0.12 (2.00) a July 27, 19C7 (1.0) i.
Utilur e to meet the t i r, e intervc1 for a surveillance c o n s t i i u t e t, a failure to o,e e t t r, e onprobility r equi r r n,en t of the LCO.
(1.0) e TeEFEPENPE DAEC Technical SPecif3 cation 1.0, Definition 26 206000C005
...(Kd'S)
F i
e
==
i i
i
-.e
- TEST ER005 REFERENCE PACE 1
OUCCTIOf!
VALUE REFERENCE 05.01 1.00 EAH0001221 i
05.02 2.00 EAH0001222 0 5. 0 ">
1.00 EAH0001223 05.0c 1.00 EAH0001224 05.05 1.00 EAH0001225 05,06 1.50 EAH0001224
-05.07
?.50 EAH0001227 r
05.03 2.00 EAH0001220 l
05.00 1.50 EAH0001229 i
05,10 1.50 EAH0001200 05.11 1.50 EAH0001231 05.12 2.50 EAH0001232
-05.13 1.00 EAH0dO1233 05.1c 1.50 EAH0001234 05 15 3 00 EAH0001235 24.50 t
I 06.01 2.00 EAH0001236 06 02 4.00 EAH0001237 06.03 2.00 EAH0001230 06.04 2.50 EAH000123o 06.05 2.00 E A H 0 0 012 n 06.06
- 2. ?(
EAH0001241 0 6. V 1.50 EAH0001242 06 08 1.50 EAH0001243 06.0Y 1.00 EAHOOO1244 l
1 06.10
?.50 EAH0001245 I
06 11 3.00 EAH0001246 I
06.12 1.00 FAH0001247 l
25.00 07.01 00 EAH0001243 4
07.02 3.50 EAH0001240 3
0 ?. O '-
' 50 E A H 0 0 01.0 l
07.O" 1. 5 '-
EAH0001251 07.o:
1.. ' 0 EAH000125?
l 07.0d 15' EAH0001253 l
l 07.07 1.50 EriH0001254 I
07.OR
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L 07.10 1 00 EAH0001257 l
l 07.I1
'.00 EAH00017".R k
0~.12 1 50 EAH0001259 07.13
' 50 E.," 2 0 012 6 0 i
07.<
'.On EAH0001261 26.On Or,c-
? oc r( n ; t 0 t ',12 A ;
e' TEST CROSS REFERENCE PAGL 2
OUESTION
A L U E REFERENCE 00 02 1.50 EAH0001263 08.00 2.00 EAH0001264 00.04 I'.00 EAH0001265 00 05 1 50 EAH0001266 00.06 2.50 EAH0001267 08.07 1.50 EAH0001268 00.08 2,00 EAH0001269 08.09
'?.50 EAH0001270 00.10 4.00 EAH0001271 00.11 1.50 EAH0001272 00.12 2.00 EAH0001273 25.00
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