ML20207F845
| ML20207F845 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/02/1999 |
| From: | Michael B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-302-98-301, NUDOCS 9903110341 | |
| Download: ML20207F845 (100) | |
See also: IR 05000302/1998301
Text
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March 2,1999
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NOTE TO:
NRC Document Control Desk
Mail Stop 0-5-D-24
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JL-.L W L A
FROM:
Beverly Michael, Licelising @stant, Operator Licensing and Human :
Performance Branch, Division of Reactor Safety, Region ll
SUBJECT:
OPERATOR LICENSING EXAMINATIONS ADMINISTERED JUNE 29 -
JULY 1,1998, AT THE CRYSTAL RIVER NUCLEAR PLANT
l
DOCKET NO. 50-302
During the period June 29 - July 1,1998, operator licensing examinations were
administered at the referenced facility. Attached, you will find the following information for
,
processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:
Item #1 -
a)
Facility submitted outline and initial exam submittal,
,
designated for distribution under RIDS Code A070.
b)
As given operating examination, designated for distribution under
RIDS Code A070.
Item #2 -'
Examination Report with the as given written examination attached,
designated for distribution under RIDS Code IE42.
Attachments: As stated
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9903110341 990302
ADOCK 05000302
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July 31, 1998
Mr. John P. Cowan, Vice President
Nuclear Operations
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Florida Power Corporation
ATTN: Manager Nuclear Licensing (SA2A)
!
Crystal River Energy Complex
15760 West Power Line Street
Crystal River, FL 34428-6708
SUBJECT:
NRC EXAMINATION REPORT NOS. 50-302/98-301
Dear Mr. Cowan:
On June 29 - July 1. 1998, the Nuclear Regulatory Commission (NRC)
administered operating examinations to employees of your company who had
applied for licenses to operate the Crystal River Nuclear Plant.
At the
conclusion of the examination, the examiners discussed the examination
questions and preliminary findings with those members of your staff identified
in the enclosed report.
The written examination was administered by your
staff on June 26. 1998.
The answer sheets and your post-examination comments
were received July 7. 1998.
Based on our evaluation of the examination process. the NRC is concerned with
the number of technical errors in the examination material.
These errors were
found by the NRC during the administration of the walkthrough and
administrative portions of the operating examination.
These errors created
difficulties for the applicants and made it difficult to administer the
examination.
The NRC concludes that this material was not adequately
validated or reviewed by your staff prior to examination administration.
i
A Simulation Facility Report is included in this report as Enclosure 2.
A
copy of the written examination questions and answer key was retained by your
facility following administration and is included as Enclosure 3.
Post-
examination comments are included as Enclosure 4.
The NRC's response to the
comments is included as Enclosure 5.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of
this letter and the enclosures will be placed in the NRC Public Document Room.
4
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DISTRIBUTION CODE
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IE42
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Should you have any questions concerning this examination please contact me
at (404) 562-4600.
Sincere 1 .
Thomas /. Peebl s. Chief
Operator Licen ing and
Human Performar ce Branch -
'
Docket Nos. 50-302
License Nos.
Enclosures: 1.
Report Details
2.
Simulation Facility Report
3.
Written Examination and Answer Key (SRO)
(Document Control Desk Only)
4.
Licensee Post-Examination Comments
,
5.
NRC Response to the Post-Examination Comments
CC w/encis:
1
Charles G. Pardee. Director
Nuclear Plant Operations (NA2C)
4
Florida Power Corporation
,
Crystal River Energy Complex
4
s
15760 West Power Line Street
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Crystal River. FL 34428-6708
4
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Robert E. Grazio. Director
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Nuclear Regulatory Affairs (SA2A)
Florida Power Corporation
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Crystal River Energy Complex
15760 West Power Line Street
Crystal River. FL 34428-6708
i
Gregory H. Halnon. Director
Quality Programs
(SA2C)
Florida Power Corporation
Crystal River Energy Complex
15760 West Power Line Street
Crystal River. FL 34428-6708
R. Alexander Glenn
Corporate Counsel
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Florida Power Corporation
MAC - ASA
>
P. O. Box 14042
St. Petersburg. FL 33733-4042
Attorney General
Department of Legal Affairs
The Capitol
Tallahassee. FL 32304
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(cc w/encls cont'd - See page 3)
)
~ 3
,
(cc w/encls cont'd)
.
William A. Passetti
Bureau of Radiation Control
Department of Health and
Rehabilitative Services
2020 Capital Circle SE. Bin #C21
Tallahassee. FL 3?399-1741
Joe Myers Directo^
Division of Emergency Preparedness
Department of Community Affairs
2740 Centerview Drive
Tallahassee, FL 32399-2100
Chairman-
Board of County Conmissioners
Citrus County
110 N. Apopka Avenue
Inverness FL 36250
Robert B. Borsum
-
Framatome Technologies
1700 Rockville Pike. Suite 525
Rockville, MD 20852-1631
Distribution w/encis:
L. Plisco, RII
R. Schin, RII
L. Wiens, NRR
R. Schin. RII
P. Steiner. RII
PUBLIC
NRC Resident Inspector
U. S. Nuclear Regulatory Commission
6745 N. Tallahassee Road
1
Crystal River. FL 34428
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U. S. NUCLEAR REGULATORY COMMISSION
.
REGION II
Docket Nos.
50-302
License Nos..
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Report Nos.-
50-302/98-301
Licensee:
Florida Power Corporation
Facility:
Crystal River Unit 3
,
Location:
Crystal River. FL
Dates:
June 26 - July 2. 1998
Examiners:
P.
Steiner.
Chief License Examiner
J.
Bartley. Resident Inspector Farley
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Approved by:
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NUles. Chief. Operator Licensing
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and Human Performance Branch
Division of Reactor Safety
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EXECUTIVE SUMMARY
Crystal River Unit 3
NRC Examination Report No. 50-302/98-301
During the period June 26 - July 2. 1998. NRC examiners conducted an announced
operator licensing initial examination in accordance with the Examiner
Standards. NUREG-1021. Interim Revision 8.
This examination implemented the
operator licensing requirements of 10 CFR 555.41, S55.43. and 555.45.
Doerations
Four Senior Pseactor Operator Instant (SRO-I) applicants received written
examinations and operating tests.
The examinations were developed by
the facility training staff and the operating tests were administered by
NRC operator licensing examiners.
The operating tests were administered
June 29 - July 1. 1998.
The written examination was administered by the
facility licensee on June 26. 1998. All four applicants passed the
examination.
Two of the four candidates demonstrated performance
>
weaknesses during the examination.
Acolicant Pass / Fail
R0
Total
Percent
Pass
4
0
4
100%
Fail
0
0
0
0%
The examiners identified a training weakness in the area of piggy-back
operations during a Small Break Loss of Coolant Accident (SBLOCA)
(Section 05.1).
The examiners identified several technical inaccuracies in the
administered operating examinations (Section 05.2).
The examiners identified several procedure deficiencies during
validation and administration of the operating examinations (Section
05.3).
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Rcoort Details
Summary of Plant Status
During the period of the examinations. Unit 3 was at 100 percent power.
I. Doerations
05
Operator Training and Qualifications
05.1
Initial Doerator Licensina Examinations
a.
Examination Scooe
NRC examiners conducted regular. announced. operator licensing, initial
examinations during the period June 26 - July 2.1998.
The facility
licensee developed and NRC examiners administered examinations in
accordance with the Operator Licensing Examiner Standards. NUREG-1021.
Interim Revision 8.
This examination implemented the operator licensing
requirements of 10 CFR S55.41. 555.43. and S55.45.
Four SRO-I
applicants received written examinations and operating tests.
b.
Observations and Findinas
Written Examinatina
The written examination was administered by the facility licensee on
June 26. 1998. All applicants passed the written examination with one
applicant demonstrating a performance weakness with a score of 81.0. The
facility licensee submitted two post-examination comments.
After review
of the post-examination comments. the NRC changed the correct answer on
both questions.
This change resulted in an applicant with a failing
grade of 79.0 being regraded to a passing score of 81.0.
The licensee's post-e.-amination review of the written examination
identified 5 generic knowledge weaknesses. A generic weakness is any
question where less than 50% of the applicants passed.
It should be
noted, that due to the small class size. this method may not be
effective at identifying generic' weaknesses.
A review of lesson plans
and training notes should be conducted to identify any weaknesses in
these topic areas.
Doeratina Test
The operating examination was administered during the period of
June 29 - July 1. 1998.
All applicants passed the operating examination
with one applicant demonstrating performance weaknesses on both the
walkthrough, and scenario portions of the operating examination.
The examiner's post-examination review of the operating tests identified
the following generic knowledge weaknesses:
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1)
Picoy-back Operations Durino SBLOCA
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The examiners identified a training weakness during conduct of the
walkthrough portion of the o>erating examinations.
All candidates
demonstrated performance weacnesses while performing E0P-14
Enclosure 19. "ECCS Suction Transfer". during a SBLOCA.
Step 19.3
requires the operator to:
Adjust Low Pressure Injection (LPI) flowrate for R8 sump
operation.
The procedure details state:
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" Throttle and maintain each LPI train to achieve a
maximum flowrate without exceeding 2200 gpm/ train
using the following:
DHV-5
DHV-6"
Because of the SBLOCA. Reactor Coolant System (RCS) 3ressure was
at - 220 psig.
This was above the shutoff head of t1e LPI pumps.
resulting in 0 gpm LPI flow.
DHV-5 & -6 were full OPEN.
The
required action, given the SBLOCA conditions was to go on to the
next step.
The applicants attempted either to shut DHV-5 & -6 or
to reduce RCS pressure in order to achieve LPI flow.
Reducing RCS
pressure results in a loss of subcooling margin. The applicants
did not demonstrate the ability to complete the procedure step or
to understand its basis.
Follow-up questioning of the applicants
revealed that they had only faced this step during a large break
LOCA when there was a high volume of LPI flow. this condition
required DHV-5 & -6 to be throttled closed.
2)
Observation of Precautions __and Limitations
The examiners identified an applicant performance weakness during
a Walkthrough examination Job Performance Measure (JPM). This JPM
required the applicant to perform procedure SP-356. section 4.1.
ES "A" RBIC. Refueling Manual Actuation Test.
Four components
failed to respond as required by the test procedure, making this
an alternate path JPM.
Step 4.1.3.5 required the operator to
confirm proper system response by verifying the indicating lights
were not lit.
Four lights remained on.
Limits and Precautions
3.5.1 stated. "Do not proceed past a step if any of the status
(Normal. Bypass or Test) indicating lights at the Test light panel
fail to operate as specified by the applicable test step.
To back
out of the procedure. see Enclosure 2.~
Enclosure 2. of section
4.1 directed the applicant to:
" Reset manual actuation by momentarily depressing the ES
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RB ISO MAN TEST RESET" push button.
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Perform Section 4.1.5
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Perform Section 5.0 as applicable."
The applicants correctly identified the four components that did
not properly respond: however, they failed to implement Enclosure
2 as required.
The applicants incorrectly performed the next
sequential ste). 4.1.3.6.
which directed performing step 4.1.5.
As a result t1e manual actuation was not reset and section 5.0.
was not completed.
3)
Identification and Diaanosis of RCP Seal Failures
The examiners identified an applicant knowledge weakness during
walkthrough questioning.
The applicants were asked to evaluate a
table of second and third stage seal pressures for the RCPs. The
applicants were able to identify that the RCP-1C seal pressures
were abnormal. but incorrectly determined which seals had failed.
c.
Conclusions
All of the applicants passed the initial operator licensing examination.
However, there were several individual performance weaknesses. 'Some of
these weaknesses were accentuated by poor procedures and poor
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examination materials.
05.2 Damination Develooment
a.
Insoection Scoce
The NRC examiners worked with the Crystal River training staff for three
months prior to administration of the examination.
The licensee sent a
team of training staff members to the Region II office in Atlanta in May
of 1997 to discuss examination content with the chief examiner.
This
meeting was helpful in identifying and rectifying weak areas in the
proposed examination.
The training staff was cooperative in
incorporating the examiner's feedback into the final examination
product.
b.
Observations and Findinos
Written Examination
The NRC examiners reviewed and validated the written examination over a
period of six weeks.
After an initial review. Operator Licensing
management made the decision to continue with the validation of the
submitted material. based on the overall quality of the original
questions.
However, many changes to the examination were required prior
to receiving final NRC approval on June 25. 1998.
The training staff
was very responsive to the examiner's comments and were prompt in making
the necessary corrections.
The quantity of comprehension / analysis
questions met NRC standards.
The following is a brief summary of
changes:
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100
Questions were submitted to the NRC for review.
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Ouestions were re-written entirely. per NRC request.
20
Questions were modified technically.
6
Questions were administratively changed.
Although the examination was not acceptable for administration as
submitted, the number of modifications required to meet NRC standards
was consistent with other Region II licensees.
Two post examination
comments' were submitted for the written examination.
These comments
were accepted after NP.C review.
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Ooeratina Examination
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The examiners reviewed and validated all portions of the operating test
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during the prep-week of June 5.1998. using the Crystal River simulator
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and walking down in-plant JPMs.
While administering the walkthrough
examination and during the post-examination phase, the examiners
identified several deficiencies in the examination material which
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resulted in questions being deleted, or multiple answers accepted.
Some
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of the questions had information that was confusing to the applicants
and created undue stress.
The following is a compilation of the
{
weaknesses:
1)
JPM NRC10. Question 2. was a graphic of the diamond control panel
with highlighted indications.
The applicants were given plant
conditions and required to determine the basis for the Auto
Inhibit light.
The correct answer was a loss of ICS auto power.
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Although this answer was technically correct. the graphic provided
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to the applicants was not representative of indications an
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operator would expect to see at the given plant conditions.
This
resulted in the applicants providing the examiners with incorrect
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answers and, in general . created confusion.
This question was
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deleted from the examination.
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2)
Administrative JPM Al required the applicants to perform an RCS
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boron change calculation in accordance with OP-304.
Soluble
Poison Concentration Control". Rev. 11.
The applicants were given
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initial and final plant conditions and were required to determine
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the amount of boric acid needed for the change.
Data for initial
and final values for Saxon Xenon Worth were given.
The values
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were -0.013 delta k/k and -0.012 delta k/k respectively.
refers to Saxon Xenon Worth in % delta k/k.
One applicant
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correctly converted the value to a percentage. however the answer
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key did not.
The examiners had to work through the calculation
manually to determine that the applicant's answer was correct.
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The examiners then had to determine whsther or not the other three
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applicants should be downgraded for not providing the correct
answer.
The examiners determined that. for examination grading
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purposes, both answers would be accepted.
All applicants
demonstrated the proper technique for performing a boron change
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calculation.
However. during actual plant operations. a mistake
of this nature could result in a significant reactivity error. and
subsequent undesired reactor plant transient.
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3)
JPM NRC03. question 1. required the applicants to describe the
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control room indications and Technical Specification (TS)
implications following a service water leak into the spent fuel
pool which resulted in a final pool level of 158 ft 9 inches.
The
answer key stated that annunciators for Spent Fuel Pool Level High
and Cask Area Level High would be in alarm and TS 3.9.1 actions
should be taken.
The answer key was inaccurate in that no
annunciators would be in alarm at this level.
The. annunciator.
should not alarm until +0.5 ft on SF-1-LI1/LI2 which corresponds
to 159 ft.
This confused the applicants.
The examiners
eventually realized the error and either modified the question or
asked follow-up questions regarding when the annunciators should
alarm
4)
JPM NRC02. question 2. required the applicants to perform a leak
rate calculation given plant conditions. After the examination,
the examiners questioned why all the applicants came up with the
wrong answer.
The chief examiner immediately requested that the
training staff perform the calculation again.
The next day the
training staff informed the chief examiner that the calculation
was accurate as stated on the answer key.
Five days later, a
post-examination comment was submitted to the NRC describing that
the answer key was in error.
The answer key was changed to
represent the comment,
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5)
Administrative JPM A2 required the applicants to approve a
clearance order which contained four prescripted errors.
One of
the errors was in an item descripticn which read. "FWV-3B" vice
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the required "FWP-3B".
Identification of the error was a critical
task per the JPM standard.
Although this may be a good training
tool
it is a poor evaluation tool.
Failure to perform a critical
,
step results in failure of the JPM.
Because of the high weighting
of the administrative section of the initial examination, three of
the four applicants could have been denied a license based on
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failure to identify a typing error.
The examiners deleted this
task as a critical step.
c.
Conclusions
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The quality of the written examination as an evaluating tool improved
over previous licensee generated examinations.
The simulator scenarios
were modified during prep-week into acceptable evaluating tools.
The
,
areas of greatest weakness wcre the walkthrough and administrative
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examinations.
The number of errors that were discovered during
examination review and administration was high.
Also, several direct
look-up questions were submitted initially.
During prep-week. these
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questions were identified by the examiners. The replacement questions
and tasks were not validated and reviewed to the level of detail
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required for an NRC initial examination final submittal.
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05.3 Licensee Procedures
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a.
ScpD2
The examiners reviewed licensee procedures during the validation of
examination materials, and after the administration of the operating
examinations,
b.
Observations and Findinos
During the pre-week, and while administering'the o3erating examination,
the examiner:; came across six procedure problems tlat could present a
challenge to operators.
One of the procedure problems resulted in all
applicants demonstrating a performance weakness. This resulted in the
failure of a critical, task (See section 05.1.b.2).
The examiners
discussed the task and procedure problem with NRC management.
It was
determined that both the applicants and the procedure were deficient.
However. the applicants would not receive a failing grade for the JPM.
Licensee Precursor Card number 3-C98-3185 has been generated to track
the procedure weaknesses.
c.
Conclusions
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The examiners were concerned with the number of procedure deficiencies
encountered during the initial examination.
The number of deficiencies
discovered is high relative to other Region II initial examinations.
This was compounded by the fact that, because of the small class size,
there was a relatively small amount of examination material generated.
Also. this was a repeat experience from the previous examination in June
,
of 1997.
The examiners were concerned that the training staff is
teaching the applicants how to interpret the procedures and work around
problems, instead of revising then.
The procedure deficiency problem
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was particularly significant when applicants became confused or failed
tasks because of poorly written procedures. The examiners have opened
an inspector follow-up item to track the progress of the procedure
modi fications .
The facility is tracking these modifications under
Licensee Precursor Card number 3-C98-3185 (IFI-302/98-301-1).
V. Management Meetings
X1.
Exit Meeting Summary
At the conclusion of the site visit. the examiners met with
representatives of the plant staff listed below to discuss the results
of the examinations and other issues.
The topics of poor applicant
performance on the walkthrough questions, piggyback operations during a
SBLOCA, and the large number of procedure deficiencies were discussed.
None of the material provided to the examiners was identified by the
licensee as proprietary
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$
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PARTIAL LIST OF PERSONS CONTACTED
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Licensee
J.
Baumstark. Director. Engineering
S.
Bernhoft, Licensing
T.
Catchpole, Licensing
J.
Cowan Vice President
R.
Davis. 0]erations Director
B.
Hickle, Director. Training
J.
Holden, Site Director
M.
Kelly, NSS
T.
Klyder, Regulatory Compliance
J.
Lind. Manager, Operations Training
C.
Pardee. Plant Manager
W.
Pike, Regulatory Compliance
J.
Smith, Supervisor Operations Training
J.
Springer, Supervisor, Nuclear Simulator Training
I.
Wilson Manager Nuclear Plant Operations
MC
S.
Cahill. Senior Resident Inspector
J.
Jaudon. Director. Division of Reactor Safety
S.
Sanchez, Resident Inspector
ITEMS OPENED. CLOSED, AND DISCUSSED
OPENED
302/98-301-01
IFI
Various procedure deficiencies
outlined in PC-C98-3185 (Section
05.3)
CLOSED
NONE
DISCUSSED
None
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LIST OF ACRONYMS USED
A0P
Abnormal Operating Procedure
Annunciator Response Procedure
CFR
Code of Federal Regulations
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COLR Core Operating Limits Report
E0P
Emergency Operating Procedure
Emergency Feedwater
Diesel Generator
IFI
Inspector Followup Item
Knowledge and Ability
LOCA Loss of Coolant Accident
Low Pressure Injection
NI
Nuclear Instrument
NRC
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
OTSG Once Through Steam Generator
Reactor Coolant Pump
R0
Reactor Operator
RIL
Rod Insertion Limits
Small Break
Safety Injection
,
Senior Reactor Operator
SRO-I Senior Reactor Operator Instant
TS
Technical Specification
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SIMULATION FACILITY REPORT
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Facility Licensee:
Florida Power and Light - Crystal River Unit 3
Facility Docket Nos.:
50-302
Operating Tests Administered on:
June 29 - July 1. 1998
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This form is to be used only to report observations.
These observations do
not constitute audit or inspection findings and are not. without further
verification and review. indicative of noncompliance with 10 SCFR 55.45(b).
These observations do not affect NRC certification or a] proval of the
simulation Yacility other than to provide information t1at may be used in
future evaluations.
No licensee action is required in response to these.
observations.
While conducting the simulator portion of the operating tests, the following
items were observed (if none, so state):
IIEtl
DESCRIPTION
1)
The simulator was not equipped with the licensee's PSAM software.
This
restricted evaluating the applicant's ability to use the PRA software to
the Main Control Room space.
2)
The RCP vibration monitoring equipment in the simulator does not match
the installed equipment in the plant.
This created difficulty when
evaluating an applicant's ability to respond to a high vibration
annunciator after starting a RCP.
The annunciator response procedure
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required the applicant to monitor indications that were not installed on
the simulator.
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Enclosure 2
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WRITTEN EXAMINATION AND ANSWER XEY
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Enclosure 3
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Power
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July 6,1998
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Mr. Paul Steiner
U.S. Nuclear Regulatory Commission
Atlanta Federal Center, Region II
61 Forsyth Street SW, Suite 23T85
Atlanta, GA 30303
Subject:
Crystal River Unit 3
June 1998 SRO Examination
Dear Mr. Steiner:
i
Attached you will find the post examination comments for the Senior Reactor Operator (SRO)
examination given at Crystal River Unit 3 (CR-3) the week of June 29, 1998.
These
comments are being submitted in accordance with NUREG 1021 ES 501 Section C.I.a.
Florida Power Corporation (FPC) recognizes that this shows short falls in our review process.
We will perform a root cause analysis, prior to August 7,1998, to identify corrective actions
to improve our process.
Attachment A contains the requested answer key corrections. Attachment B contains the
writrn examination grading quality assurance checklist.
Attachment C contains the
post-examination comment documentation. Attachment D contains the commitment made in
this letter.
If you have any questions regarding this letter, please contact Mr. Johnie G. Smith at
(352) 795-0504, extension 6107.
Sincerely
I /{L f
chn Paul CoWan
Vice President
Nuclear Operations
J PC/jg.;
Attachments
ENCLOSURE 4
CRYSTAL RIVER ENERGY COMPLEX: 15750 W. Powe Line Street * Crystal River, Florida 34428-6708
(352)795 4486
A Florida Progress Company
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' ATTACIDENT A
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' Requested Answer Key Corrections
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Justification for Florida Power Corporation's (FPC) requested answer key corrections to
two (2) written exam questions and one (1) Job Performance Manual walkthrough question.
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1.
Question #54, Correct answer should be "C".
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The question originally submitted for NRC approval contained RCS pressure values in
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each of the four choices. Detail 1 of Step 3.32 of Emergency Operating Procedure
(EOP) 6 requires the operator to maintain minimum adequate SCM.
For the RCS
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pressure values originally supplied, distractors "C" and "D" would have been incorrect
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due to creating a condition ofless than adequate subcooling margin. Answer "A" would
nave been incorrect due to no rea: tor coolant pumps available resulting in no available
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pressurizer spray. The correct answer would have been "B" under these circumstances.
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By removing the RCS pressure values from the choices, the correct answer, per Detail 5
of Step 3.32, is choice "C".
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The answer key should have been changed to reflect this prior to administering the exam.
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RECOMMENDATION:
Accept choice "C" as correct answer.
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2.
Question #74, Correct answer should be "C".
When determining how far Rod 7-3 was below the group average, API PI panel
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indication was used instead of API computer indication. This resulted in a value of
6.625% below group average instead of the correct value of 6.386%.
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While a rough calculation can be made from PI panel inc'ications, the panel lights (PI and
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Diamond) are actuated from the direct voltage- signal supplied to a group average
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amplifier and an H/L limiter. This signal is displayed on the plant computer and is more
accurate than PI panel indication.
API computer values should have been used for this calculation.
RECOMMENDATION:
Accept choice "C" as correct answer.
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U.S. Nuclear Regulatory Commission
Attachment A
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Page 2 of 2
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3.
Question #NRC02Q2, Correct answer should be >> 395 gpm.
Leak rate estimate should consist of all water being injected into the RCS minus all water
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being letdown (or removed in a controllable manner) from the RCS.
This question has letdown isolated so all the water leaving the RCS is due to the OTSG
tube leak. This calculation should consist only of the following:
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a.
MUV-31 flow
130 gpm
b.
Scal injection flow
41 gprn
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c.
MUV-24 flow
200 gpm
d.
Level change in the PZR
2 x 12.2 = 24.4 gpm
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TOTAL leakage of
395 gpm
Contrary to this, *be answer key included the change in MUT level as another source of
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water injection into the RCS. This was in error as the MUT and/or the BWST serve
only as a suction source for the Makeup Pumps.
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Correct leak rate estimate value should be 395 gpm.
RECOMMENDATION:
Accept 395 gpm as correct answer.
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Page 1 of 1
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SR0 54)
Licensee Recomendation: The correct answer is C.
NRC Response:
The NRC agrees with the licensee's discussion
and accepts answer C as the only correct answer.
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SR0 74)
Licensee Recomendation: The correct answer is C.
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NRC Response:
The NRC agrees with the licensee's discussion
and accepts answer C as the only correct answer.
JPM NRC02
Licensee Recomendation: The answer is 395 gpm.
Question 2
NRC Response:
The answer is 395 gpt.
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Enclosure 5
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U. S. Nuclear Regulatory Commission
Site-Specific
Written Examination
Applicant information
Name:
Region:
II
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Date:
Facility / Unit:
License Level: ~SRO
Reactor Type:
BW
Start Time:
Finish Time:
Instructions
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Use the answer sheets provided to document your answers. Staple this cover
sheet on top of the answer sheets. The passing grade requires a final grade of at
least 80.00 percent. Examination papers will be collected four hours after the
examination starts.
Applicant Certification
All work done on this examination is my own. I have neither given nor received
aid.
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Applicant's signature
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Results
Examination Value
_ Points
Applicant's Score
Points
Applicant's Grade
Percent
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Name:
1. ROT 456 001/ F8///2.1.7/ 3.7/4.4/ 33/ LOCA
The following plant conditions exist:
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- Plant is at 100% power.
- Pressurizer levelis at 220 inches and stable.
Letdown flow is at a constant 80 gpm.
There is no increase in the reactor or auxiliary building sumps.
- The Nuclear Services Closed Cycle Cooling (SW) surge tank (SWT-1)levelis
slowly increasing.
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- The liquid radiation monitor for the SW system, RM L3, is in alarm.
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What is the cause of these indications?
A.
A CRD stator cooler is leaking.
vB.
The primary sample cooler is leaking.
C.
A reactor coolant pump Seal Return cooler is leaking.
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D.
The Reactor Coolant Drain Tank cooler is leaking.
Reasons:
A. The CRD stator cooler leaking would not elevate RM-L3 or increase SWT 1
level.
b
C. & D.
The Seal Return cooler and RCDT is at a lower pressure than the SW
system pressure. This would decrease SWT-l's level, not increase.
ROT 4-58 Section 2.4 & 2.5; ROT-4 5G Section 1.5; 2/3-4
NRCCP97
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2. ROT-4-11 001/ B5/ / / 017K5.03/ 3.7/ 4.1/ 33/ INCORE
The following plant conditions exist:
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- RCS pressure is 386 psig.
-Tincore (CET)is 520* F.
Which of the following describes the CET inputs to subcooling margin indications
and the status of core cooling?
A.
Tb average of the 12 selected core exit thermocouples input to the
SPDS indicates that subcooling margin has been lost but the core is
being adequately cooled.
B.
The average of the 6 selected core exit thermocouples input to each
Subcooling Margin Monitor indicates that subcooling margin has been
lost but the core is being adequately cooled.
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C.
The highest one of the 12 selected core exit thermocouples input to the
SPDS indicates that an inadequate core cooling event is in Progress.
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vD.
The highest one of the 6 selected core exit thermocouples input to each
Subcooling Margin Monitor indicates that an inadequate core cooling
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event is in progress.
Reasons:
A.
The incore display to SPDS is an average of the 5 highest core exit
thermocouples of twelve. For this pressure and temperature the core is not
adequately being cooled.
B.
The SMM uses the highest of the 6 selected core exit thermocouples for
indication. For this pressure and temperature the core is not adequately
being cooled.
C.
The incore display to SPDS is an average of the .5 highest core exit
thermocouples of twelve.
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2. ROT-4-11 001/ B5/// 017K5.03/ 3.7/ 4.1/ 33/ INCORE
ROT-4-11 Section 3.5.3; 2/3 - 27
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3. ROT-4-54 001/ B13/ / / 2.4.9/ 3.3/ 3.9/ 33/ OP-404
The following plant conditions exist:
- The reactor is shutdown.
- Core cooling is being provided by Decay Heat Pump 1A (DHP-1A).
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- DHP-1B is in standby.
- DHP-1A discharge flow is oscillating between 500 gpm and 2500 gpm.
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Based on the above conditions which of the following action (s) should be taken to
maintain core cooling?
A.
Throttle closed on DHV-3 (reactor coolant outlet) to limit flow to
DHP-1A.
B.
Start DHP-1B to increase core cooling.Gow.
C.
Start DHP 1B and trip DHP-1A.
vD.
Trip DHP-1A and start DHP-1B after resolution of the problem.
Reasons:
A.
DHP-1A is cavitating and must be tripped.
B.
DHP 1A is cavitating and must be tripped. DHP-1B should not be started
until the reason for cavitation of DHP-1A is resolved.
C. DHP-1B should not be started until the reason for cavitation of DHP-1A is
resolved.
OP-301 Step 3.2.G; OP-404 Step 3.2.2; 2/3-14
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4. ROT-4-13 004/B6///022K4.03/3.6/4.0/33/ESAS
- The following plant' conditions exist:
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- A steam line rupture has occurred in the RB.
- RB pressur9 is 7.8 psig.
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- An SW surge tank Low Level alarm has actuated.
- ES status lights are as indicated.
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Component
Light
Component
Light
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AHV-1B
GREEN
AHV-1A
GREEN
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AHV-1C
GREEN
AHVID
GREEN
BSV-3
GREEN
BSV-4
GREEN
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BSV-12
GREEN
BSV-11
GREEN
BSV 17
GREEN
BSV-16
GREEN
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CAV-1
GREEN
CAV-2
GREEN
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CAV3
GREEN
CAV-6
GREEN
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CAV-4
GREEN
CAV-7
GREEN
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CAV 5
GREEN
SWV-79
AMBER
CAV-126
GREEN
SWV-80
AMBER
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SWV 81
AMBER
SWV-82
AMBER
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SWV-83
AMBER
SWV-84
AMBER
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SWV 85
AMBER
SWV86
AMBER
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Based on the above conditionc which of the following describes the status of ES
components?
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VA.
The SW valves are indicating open and should be closed; all o'ther
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components are in their expected position.
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B.
The SW valves are indicating closed and should be open; all other
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components are in their expected position.
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C.
The SW valves are indicating open and should be closed; BSV-3 and
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BSV-4 are indicating open and they should be closed,
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D.
The SW valves are indicating closed and should be open; BSV-3 and
BSV-4 are indicating closed and they should be open.
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4. ROT-4-13 004/ 86// / 022K4.03/ 3.6/ 4.0/ 33/ ESAS
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Reasons:
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B.
The SW valves are indicating open and should be closed due to the RBIC
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and coincident SW surge tank low level.
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C.
' BSV-8 and BSV-4 are indicating open and should be open.
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D.
The SW valves are indicating open and should be closed due to the RBIC
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and coincident SW surge tank low level. BSVJ3 and BSV-4 are indicating
open and should be open.
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ROT 413 Table 6; 2/3 28
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5. ROT 5-95 001//// E08EK3.2/ 3.0/3.6/33/ EOP-08
The following plant conditions exist:
- EOP-8, LOCA Cooldown, performance is in progress.
RCS pressure is 1600 psig.
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- RCS temperature is 450' F.
- RB pressure is 4.2 psig.
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Which of the following describes the status of the RB main fans (AHF-1A,1B and
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IC)?
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A.
Both ES selected R.B main fans are in slow speed and being cooled by
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Nuclear Services Closed Cycle Cooling.
B.
Both ES s. elected RB main fans are in slow speed and being cooled by
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Industrial Cooling.
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vC.
One ES selected RB main fan is in slow speed and being cooled by
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Nuclear Services Closed Cycle Cooling.
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D.
One ES selected RB main fan is in slow speed and being cooled by
Industrial Cooling.
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Reasons:
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A.
RBIC has actuated which cascades a signal for HPI actuation. HPI starts
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ONE main fan in slow speed. Recent MAR.
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B.
RBIC has actuated which cascades a signal for HPI actuation. HPI starts
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ONE main fan in slow speed. The RBIC signal will swap cooling to SW.
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D.
RBIC has actuated which cascades a signal for HPI actuation. The RBIC
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signal will swap cooling to SW.
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ROT 4-63 Section 2.2; EOP 8 Step 3.10; 2/3-86
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6. ROT 5-50 002/F2///034A1.01//3.2/88/FH -
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Which of the following describes the Sensotec Load Cell overload setpoint and its
basis for FHCR-3, Spent Fuel Handling Bridge?
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A.
~ 2800/2700 pounds (heavy / lite); to provide electrical overload protection
for the fuel hoist motor.
.
vB.-
= 2800/2700 pounds (heavy / lite); to limit the amount of withdrawal force
on the fuel assembly.
C.
= 3200/3100 pounds (heavy / lite); to provide electrical overload protection
for the fuel hoist motor.
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D.
= 3200/3100 pounds (heavy / lite); to limit the amount of withdrawal force
on the fuel assembly.
Reasons:
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A., C., & D. The highest load limit is 2900 lbs. with the concern being damage to
the fuel assemblies or fuel racks.
FP 6010 Step 3.4.5.1; ROT 5-50 page 20; 1
BANK; rot 4 26 #30; NRC 96; ROTS M - T5A; ROTS M - 5B
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7. ROT-4-54 002/B13///005K6.01/2.4/2.6/33/DH
Which of the following describes the minimum continuous DHP flow and
maximum time period to prevent accelerated pump wear?
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VA.
100 - 1200 gpm for 30 days.
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B.
300 1500 gpm for 30 days.
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C.
100 - 1200 gpm for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
D.
300 - 1500 gpm for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
Reasons:
B., C. & D.
Per OP-404, Decay Heat Removal System,100 to 1200 gpm for 30
days is the limit to minimize wear of the DH rotating assembly.
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2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> is an EDG time limit.
OP-404 Step 3.2.3; 1-54
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8. ROT-5-61 001/ A1/ / / 076K3.07/ 3.7/ 3.9/ 33/ AP.330
A step in AP-330, Loss of Nuclear Service Cooling, directs the operator to stop all
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the SWPs and RWP-1, RWP-2A and RWP-2B. Why are the SW Raw Water
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Pumps stopped?
A.
To aid in the diagnosis of SWHE tube leakage.
.B .
To minimize the thermal stresses on the SWHE as the SW system
temperature increases.
C.
To prevent salt water from entering the SW system following SWP
shutdown.
vD.
To ensure the SW Raw Water Pumps are not run without cooling water.
Reasons:
A.
This could be where the leak is however the concern at this point is
running the RWPs with no cooling.
B. & C.
The overriding concern is damage to the RWP motors due to lack of
cooling.
AP-330 Step 3.25; 1-57
BANK; ROT 5 61 #10; ROTS J - Final 96; ROTS K - T2
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9. ROT-4-64 003/ G6/// 063K302/ 3.5/ A7/ 77/ DC
A plant startup is in progress .vith all four reactor coolant pumps running. The
DC distribution sys tem power is lost to 6900 VAC Reactor Auxiliary Bus "3A"
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AC power to the same bus is then lost. Which of the following describes the
response of the reactor coolant pumps if AC power is restored?
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vA.
RCPs will start.
B.
The RCP'a breakers tripped when AC pwer was lost and the pumps will
not restart.
C.
AC power cannot be restored to a bus with RCP breakers closed. The
RCP brea:<ers must be opened prior to energizing the bus.
D.
The RCP's breakers tripped when DC power was lost and the pumps will
not restart.
Reasons:
B.
These breakers would have tripped on a loss of AC power if DC power were
available.
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C.
There is no interlock associated with load breaker position.
D.
These breakers require a trip signal and available DC power to open.
ROT 4-04 Section 3.2.2: OP-705 Step 4.5.1; 1-34
BANK; ROT 4-64 #17; ROTS J - Final 96
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10. ROT-5-48 002/ A2/// 059AA2.05/ 3.6/3.9/44/ WASTE
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An evaporator condensate storage tank (ECST) is being released. The release is
terminated by a high alarm on RM-L2. What action (s) should be performed?
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VA.
Verification that release isolation valves WDV-891 and 892 are closed
and contact chemistry for RM-L2 evaluation.
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B.
Verification that release isolation valve SDV-90 is closed and contact
chemistry for RM L2 evaluation.
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C.
Verification that release isolation valves WDV 891 and 892 are closed
and contact chemistry for RM-L2 flush sample.
D.
Verification that release isolation valve SDV-90 is closed and contact
chemistry for RM-L2 flush sample.
Reasons:
B.
SDV-90 is not in the release path for ECST releases nor is it closed by
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an RM-L2 actuation. SDV-90 is used for releases from the secondary
side, SDT-1.
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C.
Chemistry is directed to check the operability of the radiation monitor
not to take samples from it.
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D.
SDV-90 is not in the release path for ECST releases nor is it closed by
an RM-L2 actuation. Chemistry is directed to check the operability of
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the radiation monitor not to take samples from it.
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AR-403 EP 1788; ROT 4-59 Section 4.3; 1-70
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BANK; ROT 5-48 #20; NRC 6-97; ROTS M - TGB
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11. ROT 429 001/F5///33K1.02/2.5/2.7/33/SF
To prevent pump dan' age to SFP-1A or 1B which of the following must be adhered
to?
A.
Maintain the maximum flow through each SF pump < 1800 gpm.
vB.
Maintain a minimum sustained flow through each SF pump of > 1000
gpm.
C.
If flow instrumentation is not available, pump differential pressure
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should be maintained at < 15 psid.
D.
Ensure the maximum SF pump inlet temperature is < 150* F.
Reasons:
A., C. & D. Limit and precaution in OP-406.
OP-400 Step 3.2.3;1-45
BANK; ROT 4-29 #4; ROTS J - T9; ROT N - T2 & T2A
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12. ROT-4-69 003/ B7/// 051G2.2.22/ 3.4/4.1/ 33/ VACUUM
The following plant conditions exist:
- The plant is at 50% power.
- Condenser vacuum is 25 in-Hg and steady.
- Low pressure turbine exhaust temperature is 258* F.
Based on these conditions which of the following action (s) should be taken?
A.
Restore vacuum to > 26.5 in-Hg within five minutes or trip the turbine.
B.
Immediately reduce power to < 30% and trip the main turbine within
five minutes.
vC.
Trip the main turbine immediately.
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D.
Initiate hood spray.
Reasons:
A.
If reactor power was less than 30% then this would be the correct action
to take, based on vacuum only.
B.
If vacuum was < 24.5 in-Hg then this would be the correct action to take,
based on vacuum only.
D.
Vacuum is adequate for this power level however with low pressure
turbine exhaust temperature > 250 F OP-607 requires the main turbine
to be tripped immediately.
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OP-607 Section 4.5;2/3-66
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13. ROT-4-25 003/ B3/// 072 <1.01/ 3.1/ 3.5/ 33/ RMS
The plant is currently at 100% power. AHF-3B (Reactor Building Operating Floor
Fan "B") is declared inoperable, which also renders RM-A6 inoperable. Other
than repairing AHF 3B which of the following describes the action (s) that can be
taken to restore RM-A6 to operable status?
A.
Start the back-up sample pump on RM A6.
B.
Start AHF-3A (Reactor Building Operating Floor Fan "A").
vC.
Open the alternate sample line isolation valve and close the normal
sample isolation valve.
D.
Have Chemistry analyze grab samples of the Reactor Building
atmosphcre every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Reasons:
A.
RM-AG normal sample point is from the ductwork of AHF-3B. Starting the
backup sample pump will not alter the sampling point.
B.
Starting AHF-3A will not help because the sample point is still aligned to
the ductwork of AHF-3B.
D.
This is a required action per TS due to the loss of RM-AG however it does
not restore RM-AG to operable status.
ROT 4-25 page 20; TS 3.4.14; OP-505 Step 4.3; 1-23
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14. ROT 4 62 001/B3///026A2.04/3.9/4.2/33/8S
The following plant conditions exist:
A large break LOCA is in progress.
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. HPI, LPI, RBIC and BS have actuated.
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- HPI, LPI and RBIC were bypassed following actuation.
. RB pressure is currently 15 psig.
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- BSV-3 failed to automatically control and was taken to manual and closed.'
- The HPI seal-in permit was reset and the "A" BS pump was secured.
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Which of the following methods of BS flow control are available when BSV-3 is
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repaired and returned to service?
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VA.
Manual only.
y
B.
Remote / Auto only.
,
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C.
Local / Auto only.
.
D.
Local / Auto and Manual.
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Reasons:
l
B., C. & D.
With the HPI seal in reset and the BS pump secured all automatic
i
functions have been bypassed. The only way to restore automatic
i
control (Remote / Auto or Local / Auto) would be to re-actuate RBIC.
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ROT 4-62 Section 2.2.3;2/3-29
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15. ROT-4 64 001/ F2///063K1.03/2.9/3.5/88/DC
The following readings were taken on the "A" battery charger following~a
BATTERY A DISCHARGE HIGH alarm in the control room:
- 120 volts
1
- 60 amps
Shortly after these readings were taken the amp meter increases to 360 amps.
Which of the following action (s) will occur following this increase?
A.
The "A" and "C" inverter will trip.
vB.
The "A" battery charger will trip and the battery will supply bus loads.
C.
The "A" and "C" inverter will not trip but will swap to the AC input.
D.
The "A" battery charger will trip and the "C" battery charger will
automatically be placed in service.
Reasons:
A. There may be problems with the inverters but they should not trip.
C. There may be problems with the inverters but they should not trip. The
inverter is normally supplied from its AC source.
'
D. The "A" charger will trip but the "C" charger does not have any automatic
closing feature.
ROT-4-64 Section 2.2 AR-701 EP 169 & 1945; OP-705 Step 3.1.7; 1-33
NRCCPDi
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16. ROT-5-114 001/ A2/// 009EK3.21/4.2/4.5/44/ AP-520
The following plant conditions exist:
The plant is in Mode 3.
RCS pressure is 2155 psig.
RCS temperature is 534 F.
MUV-31 has increased from 65% to 95% open.
- PZR level, after an initial decrease, is being maintained at setpoint.
- A Security Officer has called the control room to report water dripping from the
ventilation ductwork in the Seawater Room.
-
Based on the above which of the following describes the EOP/AP that should be
entered and probable plant conditions?
A.
Entry Conditions for EOP 2, Vital System Status Verification, are met
due to excessive RCS leakage.
B.
Entry Conditions for EOP-5, Excessive Heat Transfer, are met. The PZR
level decrease is due to the decrease in RCS temperature.
C.
Entry Conditions for EOP-8, LOCA Cooldown, are met. The PZR level
decrease is due to an RCS leak into the SW system.
vD.
Entry Conditions for AP-520, Loss of RCS Coolant or Pressure, are met.
The PZR level decrease is due to an RCS leak into the SW system.
Reasons:
A.
AP-520 Entry Conditions are met.
B.
RCS temperature is normal for these plant conditions.
C.
EOP-8 is only entered if directed to by another procedure.
ROT 5114 page 27; AP 520 pages 1,29 & 33; 2/3-84
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17. ROT-4-15 002/ BS/// 040AK3.01/ 4.2/4.5/ 33/ EFIC
The following plant conditions exist:
- A plant startup is in progress.
- A main steam lit.e rupture downstream of MSV 55, MS supply to EFP-2, has
occurred.
- A" OTSG pressure is 540 psig.
"
"B" OTSG pressure is 780 psig.
-
Based on the conditions above which of the following describes current plant
configuration?
A.
MSIVs on the "A" OTSG are open; MSIVs on the "B" OTSG are open;
MFW is controlling the "A" OTSG at Low Level Limits.
B.
MSIVs on the "A" OTSG are closed; MSIVs on the "B" OTSG are open;
EFW is controlling the "A" OTSG at Low Level Limits.
C.
MSIVs on the "A" OTSG are open; MSIVs on the "B" OTSG are open;
MFW is controlling the "B" OTSG at Low Level Limits.
vD.
MSIVs on the "A" OTSG are closed; MSIVs on the "B" OTSG are open;
EFW is controlling the "B" OTSG at Low Level Limits.
Reasons:
A.
MSIVs close at 600# unless bypassed. TS requires that EFIC be in service
at > 750# OTSG pressure. MSLI and MFLI would have actuated and
isolated MFW.
B.
FOGG would isolate EFW to the "A" OTSG.
C.
MSIVs close at 600# unless bypassed. TS requires that EFIC be in service
at > 750# OTSG pressure. MFLI will trip both running MFWPs, securing
MFW to both OTSGs, because FWV-28 would not be closed at this point.
ROT 4-15 Section 2.2.2.2 & 2.2.3.3;2/3-75
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18. ROT 3-03 001/ B2///1517AK3.07/4.1/4.2/33/NAT CIRC
The following plant conditions exist:
- EOP-9, Natural Circulation Cooldown, is in progress.
- Levelin both OTSGs is = 73%.
- The NSS directs you to determine if Natural Circulation has been established.
Which of the following conditions indicates that natural circulation is in progress?
A.
0900
T
= 520" F
ave
0905
T
= 518' F
ave
vB.
0900
Tcold = 515' F
OTSG T
= 522 F
sat
0905
Tcold = 513 F
OTSG Tsat = 520' F
.
C.
0900
T
= 525 F
Tincore = 555' F
hot
0905
T
= 523" F
Tincore = 557 F
hot
D.
0900
T
= 525* F
Tcold = 475 F
hot
0905
T
= 528" F
Tcold = 472 F
hot
Reasons:
A., C. & D.
Natural Circulation indications are as follows:
1. T
tracking T
hot
incore'
cold racking OTSG Tsat-
2. T
t
minus Teold) stable or lowering.
3. AT (Thot
EOP-9 Step 3.3; 164
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19. ROT-5-48 001/ A2///2.3,6/2.1/3.1/ 44/ WASTE
The NSM/NSS must sign all radioactive liquid release permits prior to the
initiation of the release. What is the purpose of the SSOD signature?
VA.
It serves to acknowledge and confirm the approval to complete the
release.
B.
It serves to acknowledge and confirm the appropriate liquid radiation
monitor is operating properly.
C.
It acknowledges and confirms the estimated volume of fluid to be
released.
D.
It acknowledges and confirms the estimated amount of radioactivity to
be released to the environment.
Reasons:
B. This is done after the signature.
C. Volume of the release is determined by the size of the tank.
D. The signature acknowledges only the completion of the chemistry portion of
the permit not the amount of reactivity release.
OP-407A Section 4.3; 1-11
BANK; ROT 5-48 #18; NRC 6'-97; ROTS M - 5B
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20. ROT 5-14 003/ B5// NTS/2.4.16/3.0/4.0/33/ EOP/AP
The following plant conditions exist:
- A reactor trip has occur. ed.
- RM-A3, Auxiliary Builning Exhaust (Waste Gas Area), is in high alarm.
- The NNI-X status light on the Redundant Instrument Panelis extinguished.
- All the lights on the NNI-X power supply monitor are lit.
Which of the following describes the Procedure Director's response to these
indications?
A.
Enter EOP 02 and complete prior to performing AP-250.
B.
Concurrently perform EOP 02 and AP-581 by completing first the
immediate actions in both procedures and then performing the
follow-ups,
vC.
Enter EOP-02 and once the immediate actions are complete concurrently
perform AP-250 and AP-581.
D.
Enter EOP-02 and once the immediate actions are complete concurrently
perform AP-250 and then subsequently perform AP-581.
Reasons:
4
.
A. EOP 02 does not need to be completed prior to entry into AP-250.
B. EOP-02 immediate actions should be completed prior to entering another
procedure.
D. A loss of NNI-X has also occurred. AP 581 should be concurrently performed.
ROT 514 page 10 & 13; 2/3-15
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20. ROT-5-14 003/BS// NTS/2.4.16/3.0/4.0/33/EOP/AP
The following plant conditions exist:
- A reactor trip has occurred.
- RM-A3, Auxiliary Building Exhaust OVaste Gas Area), is in high alarm.
The NNI X status light on the Redundant Instrument Panelis extinguished.
- All the lights on the NNI-X power supply monitor are lit.
Which of the following describes the Procedure Director's response to these
indications?
A.
Enter EOP-02 and complete prior to performing AP-250.
B.
Concurrently perform EOP 02 and AP 581 by completing first the
immediate actions in both procedures and then performing the
follow-ups.
vC.
Enter EOP-02 and once the immediate actions are complete concurrently
D.
Enter EOP-02 and once the immediate actions are complete concurrently
perform AP-250 and then subsequently perform AP-581.
Reasons:
A. EOP-02 does not need to be completed prior to entry into AP-250.
B. EOP 02 immediate actions should be completed prior to entering another
procedure.
D. A loss of NNI-X has also occurred. AP-581 should be concurrently performed.
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ROT-5-14 page 10 & 13; 2/3-15
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21. ROT-2-32 002/ G3/ // G2.3.5/ 2.3/ 2.5/ 77/ RAD
.
Which of the follovring is the required posting for a space where it is estimated
that an individual could receive 200 mr in any hour at 15 centimeters?
A.
Vital Area.
vB.
Radiation Area
C.
D.
Grave Dtnger Radiation Area.
'
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Reasons:
.
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A., C. & D.
At 30 centineters the dose rate should be 50 mr/hr.
This meets the 5 to 100 mr/ hour definition of a radiation area.
,
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22. ROT-5-31003/B3///011K6.05/3.1/3.7/33/AP-990
The following plant conditions exist:
- The plant is in Mode 3 with RCS pressuie at 2150 psig.
- AP-990, Shutdown from Outside the Control Room, has been entered and
transfer to the Remote Shutdown Panelis complete.
- MUV 31 has failed. C.Ja.5 t.o
- The NSS directs that PZR level be maintained at an indicated = 100 inches.
Which of the following actions should be taken and what would be the
approximate actual PZR level for these conditions?
.
A.
Open MUV-27 and direct the PPO to open MUV-30, bypass around
,
MUV-31; = 160 inches.
<
vB.
Use an available HPI valve; = 160 inches.
C.
Open MUV-27 and direct the PPO to open MUV-30, bypass around
MUV-31; = 40 inches.
D.
Use an available HPI valve; = 40 inches.
,
Reasons:
,
A.
Per step 3.48 an available HPI valve should be used.
C.
Per step 3.48 an available HPI valve should be used. Actuallevel for
these conditions should be = 160 inches.
D.
Actuallevel for these conditions should be = 160 inches.
1
ROT 4-09 Section 2.2.5; AP-990 Step 3.48; ROT 4-16 page 9; 2/3-40
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23. ROT-5-34 002/ A1// / 2.4.44/ 2.1/4.0/ 55/ E-PLAN
'
Thirty (30) minutes after an event the plant conditions are as follows:
RCS pressure is 600 psig and slowly decreasing.
- RCS temperature (incores)is 850 F and slowly increasing.
'
- RM-G29 is 27,000 R/hr.
RM G30 is 28,000 R/hr.
,
- MUP 1A and MUP-1B are running.
'
- Both BSPs are running.
- A release is underway.
Determine the correct Protective Action Recommendations for the conditions
listed above.
A.
No Protective Action Recommendations are required.
B.
0-2 miles, Evacuate 360 ; 2-5 miles, Evacuate downwind sectors and
shelter remaining sectors; 510 miles, Shelter downwind sectors.
C.
0-2 miles, Evacuate 360 ; 2-5 miles, Evacuate 360*; 5-10 miles, Shelter
360*
vD.
0-2 miles, Evacuate 360*; 2-5 miles, Evacuate 360*; 5-10 miles, Evacuate
'
360*.
Reasons:
.
'
A., B. & C.
Pressure and temperature conditions indicate Region 3 however
RM-G29/30 readings indicate Region 4. With a release underway
evacuation 010 miles is required.
EM 202 Enclosure 8; 2/3-17
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24. ROT-4-14 004/B6///A01AA22/3.5/3.8/33/lCS
The following plant conditions exist:
- The reactor is producing 1842 MWthermal.
Three reactor coolant pumps are operating.
- Control rod group 7 is 60% withdrawn.
Which of the following describes the ICS response, and required additional
operator action (s), if any, if control rod 7-3 dropped fully into the core?
A.
ICS will automatically run back to 60% ULD demand. No additional
operator actions are required.
B.
ICS will automatically run back to 60% reactor power. No additional
operator actions are required.
vC.
ICS will automatically run back to 60% ULD demand. The operator
must ensure that reactor power is maintained less than 45%.
D.
ICS will automatically run back to 60% reactor power. The operator
must ensure that reactor power is maintained less than 45%.
Reasons:
A.
With only three RCPs in operation the operator must reduce and maintain
reactor power less than 45%.
B.
ICS runback is based on ULD demand and with only three RCPs in
operation the operator must reduce and maintain reactor power less than
45%.
D.
ICS runback is based on ULD demand.
ROT-4-14 Section 3.1.4; AP-545 Step 3.21; 2/3-82
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25. ROT-4-60 002/ 818/// 062AA2.06/2.8/ 3.1/ 33/ RC
The following plant conditions exist:
MUV-16, seal injection flow control valve, failed closed while in automatic.
- MUV-16 manual control has been selected and sealinjection flow is being
restored.
- Sealinjection flows to each Reactor Coolant Pump (RCP) are:
RCP-1A 8 gpm.
RCP-1B 9 gpm.
RCP-1C 0 gpm.
RCP-1D 8 gpm.
What action (s) should be taken with respect to the RCPs if SWflow to the RCPs is
lost?
A.
RCP-10 must be tripped within five minutes. All other RCPs can
continue to operate.
vB.
RCP 1C must be tripped within two minutes. All other RCPs must be
tripped within five minutes.
1
C.
RCP-1C must be tripped immediately. All other RCPs must be tripped
within five minutes.
D.
All RCPs must be tripped immediately.
Reasons:
A. & C. RCP-1C must be tr; ped within two minutes due to loss of both SI and
SW.
D.
RCP-1A,1B & ID may operate a maximum of five minutes with a loss of
SW.
OP-302 Step 3.2.4 & 3.2.5; 2/3-71
NRCCP97 - Modified for new time limit.
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26. ROT-4-09 004/ B1// / 027AA1.01/4.0/ 3.9/ 33/ NNI
)
The following plant conditions exist:
- RC3A-PT1, RCS narrow range pressure transmitter, was selected for control
1
when the pressure transmitter fails high.
- SASS has failed to transfer to the alternate instrument.
Based on these failures which of the following describes the resulting plant
response?
A.
ES Channel I will be inoperable.
B.
Channel A of the Diverse Scram System (DSS) will actuate.
C.
RCS pressure will decrease only because the PZR heaters are receiving a
false high pressure signal.
vD.
RCS pressure will decrease rapidly because RCV-14, RCS Spray valve,
and RCV-10, PORV, are receiving a false high pressure signal.
Reasons:
A.
Wide range pressure transmitters feed the ES system. This failure has no
effect on ES.
B.
This transmitter does not feed DSS so no actuation will occur.
C.
The PZR heaters are receiving a false high pressure signal however RCS
pressure will decrease rapidly due to the Spray valve and PORV opening.
ROT 4-09 Section 2.4 and Figure 6; 2/3-90
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27. ROT-4-90 001/Bi///062K1.04/3.7/4.2/33/ ELECT
i
'Which of the following power sources could feed the "A" ES'4160V bus if a plant
i
startup was in progress with the turbine at 1700 rpm?
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A.
Unit 3 Startup transformer, Offsite Power transformer or Backup ES
transformar.
i
vB.
Offsite Power transformer, Backup ES transformer or "A" EDG.
C.
Unit Auxiliary transformer, Offsite Power transformer or Backup ES
transformer.
l
D.
Unit 3 Startup transformer, Offsite Power transformer or "A" EDG.
Reasons:
,.
A. & D.
The Startup transformer is no longer able to feed the ES buses.
C.
With the turbine at 1700 rpm and the output breakers open there is no
power input to the Unit Auxiliary transformer.
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ROT 4-90 Section 2.1; 1-50
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28. ROT-3 25 001/B4///011EK1.01/4.1/4.4/33/ICC
The following plant conditions exist:
- A LOCA has occurred.
j
- RCS pressure is 485 psig.
- Incore thermocouples indicate 427' F.
Which of the following is the required OTSG level?
.
A.
> 20" EFIC Lo Range.
B.
> 65% EFIC Hi Range.
,
C.
> 70% EFIC Hi Range.
vD.
> 90% EFIC Hi Range.
Reasons:
A., B. & C.
40 F subcooling me.rgin exists. For this pressure 50 F
subcooling margin is required. > 90% OTSG levelis required
to enhance reflux boiling.
EOP-13, Rule 1 and Rule 3; 2/3 61
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29. ROT-5-38 001/G4///2.1.3/3.0/3.4/77/01-04.
The oncoming Auxiliary Building operator (PPO) must complete the following
prior to the " Shift Crew Briefing" as part of being properly relieved.
l
i
A.
Receive s.ssigned keys and review noteboard, STIs, COCs, and OOS Lag,
vB.
Receive assigned keys and review PPO log and noteboard.
C.
Review of noteboard, OOS log, OSBs and walkdown of auxiliary
building.
D.
Review of STIs, COCs, OOS Log and walkdown of auxiliary building.
!
Reasons:
A., C. & D.
The PPO must receive assigned keys and review PPO log and
noteboard prior to the " Shift Crew Briefing".
0104 page 30; 1-2
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30. ROT-5-01002/ A3///2.1.5/2.3/3.4/44/TS
The following plant conditions exist:
4
The plant is in Mode 3.
- One of the two available'PPOs assigned to the Auxiliary Building slips and
severely sprains his ankle while performing a walkdown of the Reactor
Building.
The PPO is contaminated and is escorted to the hospital by both available
Health Physics technicians.
,
Which of the following describes the appropriate response, relating to shift
staffing, for this situation?
A.
No action is required. Minimum staffing levels 'are still met.
B.
Ifit is two hours or less until shift turnover is scheduled to occur no
action is required.
C.
Another PPO should be called in immediately and should arrive within
two hours.
vD.
Another HP technician should be called in immediately and should
arrive within two hours.
Reasons:
A.
A minimum of one HP technician is required when fuelis in the reactor.
B.
Efforts must be made immediately to replace the HP technician within two
hours.
C.
Only one PPO is required to meet staffing levels.
TS 5.2.2; AI-500 Section 4.6; 1-3
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31. ROT-4 51001/B2///074EA1.01/4.2/4.4/33/RCITS
Which of the following conditions will allow the reactor vessellevel indication
portion of the RCITS to provide the most accurate information?
A.
The Reactor Vessel head must be removed.
B.
The RCS pressure must be > 40 psig.
vC.
There must be no flow in the reactor vessel.
D.
There must be at least one RCP operating.
Reasons:
A.
If the RV head is removed the piping associated with RCITS is
disconnected.
B.
Overpressure has no effect on RCITS operation.
D.
RCITS is not accurate with RCS flow through the system.
i
ROT 4-51 Section 2.1.3; 1-78
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BANK; ROT 4 51 #1G; ROTS K - Final 97
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32. ROT-5-38 002/ A9// / G2.2.20/2.2/ 3.3/ 55/ 01-07
The following plant conditions exist:
- The plant is at 100% power.
- RC-3B-PT3 has failed low.
- ES Channel 3 (RC3 & RCG) has actuated.
1
- MP 531. Troubleshooting Plant Equipment, is about to start.
Based on the above conditions determine the Risk Level, if NSM approvalis
,
required and if the Maintenance Supervisor is required to be present in the field?
'
.
A.
Risk Level 1; NSM approvalis required; Supervisor must be present.
B.
Risk Level 1: NSM approvalis required; Supervisor not required in the
field.
vC.
Risk Level 2; NSM approvalis required; Supervisor must be present.
D.
Risk Level 2; NSM approval is not required; Supervisor not required in
the field.
Reasons:
A., B. & D.
These conditions meet Risk Level 2 criteria because the channel
is already actuated and working on this pressure transmitter will
not cause actuation of a safety system. NSM approval and a
Maintenance Supervisor in the field is required per NOD-22.
OI-07 Step 1.13; NOD-22 Section II; MP-531 Step 3.1.12; 2/3-8
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33. ROT-4-64 002/F2///058G2.1.28/3.2/3.3/88/DC
Which of the following ueecribes the purpose / function of the DC Test Jacks
installed on the mrdn Cla.a IE battery disconnect switches?
A.
To allow an alternate method for testing the " Class 1E Battery
Disconnect Open" alarm.
B.
To allow an alternate method to verify the batteries resistance to ground
(ground detection).
vC.
To allow nn alternate method to verify the disconnect is closed.
D.
To allow an alternate method to verify battery charger voltage.
Reasons:
A.
" Class 1E Battery Disconnect Open" alarm does not exist.
B.
The test jacks were installed to verify the disconnect closed, not as a way to
test for grounds. Recent MAR.
D.
With the disconnect open the test jacks will only read battery voltage, not
battery charger voltage.
ROT 4-64 Section 2.9 & Figure 7; 1-96
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34.' ROT-5-30 001/ F2///008K2.02/ 3.0/ 3.2/ 88/ RW
The following pl, ant conditions exist:
Backup ES Transformer is supplying the "A" ES bus.
-
- Offsite Power Transformer is supplying the "B" ES bus.'
-- A sudden pressure fault on the Startup transformer has just occurred.
'
Based on the above conditions which of the following Raw Water Pumps (RWPs),
'
if any, would be in operation?
A.
"A" Emergency Duty Nuclear Services RWP, RWP-2A.
vB.
"B" Emergency Duty Nuclear Services RWP, RWP-2B.
C.
D.
No RWPs would be operating.
Reasons:
A. There is no low pressure start for RWP-2A and no ES actuation has occurred.
C. RWP-1 is powered from the "A" Unit 4160V bus which is lost due to the
Startup transformer failure.
D. RWP-2B will auto-start on low pressure.
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If the Shutdown Margin is -4.6% Ak/k from a calculation using 3.62% Ak/k
reactivity from xenon, which of the following statements will apply?
)
.
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A-
Section 6 of Enclosure 1 in SP-421 must be completed. Sufficient
,
shutdown margin will be preserved even when xenon decays to zero.
'
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.
A
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vB.
Section 6 of Enclosure 1 in SP-421 must be completed. Insufficient
shutdown margin will result when xenon decays to zero. Boron must be-
l
increased to maintain adequate shutdown margin.
'
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C.
Section 6 of Enclosure 1 in SP-421 need not be completed. Sufficient
l-
shutdown margin will be preserved even when xenon decays to zero.
+
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D.
Section 6 of Enclosure 1 in SP-421 need not be completed. Insufficient
'
i
shutdown margin will result when xenon decays. Boron must be
l
increased to maintain adequate shutdown margin.
I
i
Reasons:
i
.
b
A.
Shutdown margin will be less than the minimum 1% A k/k after xenon
i
decays to zero.
1
L
,
!
C.
Section 6 must be completed because xenon was used in Section 4.
Shutdown margin will be less than the minimum 1% A k/k after xenon
!.
decays to zero.
!,
.
}
D.
Section 6 must be completed because xenon was used in Section 4.
1
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SP-421 Step 4.1 and Enclosure 1:1-7
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BANK; ROT 5-10 #4
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36. ROT-3-25 002/B4///074G2.4.6/3.1/4.0/55/ICC
The following plant conditions exist:
RCS pressure is 2420 psig.
- Incore temperature is 800* F.
- Levelin both OTSGs is 5 inches.
- Pressure in both OTSGs is 800 psig.
Which of the following actions should be taken to aid in the establishment of core
cooling?
.
A.
Decrease RCS pressure to = 700 psig.
B.
Decrease RCS pressure to = 900 psig.
C.
Decrease RCS pressure to = 1400 psig.
vD.
Decrease RCS pressure to = 1600 psig.
Reasons:
A., B. & C.
For these plant conditions RCS pressure should be reduced to the
higher of the following:
1.100 psig above OTSG PRESS or,
2.100 psig above next ICC region curve.
EOP 7 Step 3.7; EOP-7 Figure 1; 2/3-77
BANK; ROT 3-25 #7
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37. ROT-4-90 002/ B3/ / / A05 AK3.1/ 3.2/ 3.4/ 44/ ELECT
The following light indications are present
on the main contrcl board,
c# sic * cmc
O
<
Startup
3203
3204
BLDCK CLOSING
BEST
3205-
3200
'
$7$gos
O
Unit Aux
3207
3208
3209
3210
asm .
ACTUATED 3207
OPT
3211
3212
BLDCK CLO5(NO
g
,
ACTUA Tc0 321l
A plant startup is in progre.ss with reactor power at 10%. Which of the following
sets of conditions will cause the 9bove indication? Assume sufficient time for
automatic actions to have occurred.
i
A.
The offsite power transformer is OOS (tagged out) for oilleak repair.
'
SP 354A is in progress with Breaker 3209 closed.
A spurious 'A' train ES actuation has just occurred.
B.
SP-354A is in progress with Breaker 3209 closed.
The Startup Transformer is OOS (tagged out) for oilleak repair.
A loss of Off-Site power has just occurred, Bkr 3211 has failed to open.
vC.
SP 354B is in progress with Breaker 3210 closed.
The Startup Transformer is OOS (tagged out) for oil leak repair.
A loss of Off-Site power has just occurred, Bkr 3212 has failed to open.
j
i
D.
The offsite power transformer is OOS (tagged out) for oil leak repair.
SP-354B is in progress with Breaker 3210 closed.
A spurious 'B' train ES actuation has just occurred.
Reasons:
A.
These conditions have 3205,3200 and 3209 closed. 3210 will be blocked.
B.
These conditions have 3209,3211 and 3212 closed. 3210 will be blocked.
D.
These conditions have 3205,3206 and 3210 closed. 3209 will be blocked.
'
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37. ROT-4-90 002/ B3// / A05AK3.1/ 3.2/ 3.4/ 44/ ELECT
ROT 4-90 Section 2.5; AR-702 EP 1183; 2/3-100
NEW
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38. ROT-4-60 004/ B20/// 015AK2.07/2.9/ 2.9/ 55/ RCS
The plant is operating at 100% full power with the following RCP seal data:
RCP SEAL STAGE PRESSURE (psig)
2nd
3rd
2nd
3rd
2nd
3rd
2nd
3rd
Time
Stage
Stage Stage - Stage
Stage Stage
Stage
Stage
RCP-1A
RCP-1B
RCP-1C
RCP-1D
0900
1300
700
1400
800
1550
900
1425
725
0910
1325
725
1375
825
1575
925
1425
775
0920
1300
700
1400
800
1550
950
1400
775
0930
1825
725
1400
800
1575
1035
1450
800
0940
1350
725
1400
800
1575
1125
1450
800
Dumpster
3
2
3
2.
,
clicks per
minute at
0940.
Based on the above data which of the following describes the proper course of
action?
l
A.
Immediately trip RCP-1C and allow the ICS to run back the plant.
l
B.
Reduce power to < 72% per AP-510, Rapid Power Reduction, and trip
RCP-10.
'
.
!
VC.
Reduce power to < 72% per OP-204, Power Operations, and trip RCP-10.
D.
RCPs are all within expected leakage for the operating condition. Total
RCS leakage should be verified by performance of SP 317.
-
Reasons:
A., B. & D.
RCP-1C total sealleakage is 2.55 gpm and increasing gradually.
Per OP-302, RCP Operation, the correct action is to reduce power
to < 72% per OP-204, Power Operations, and trip RCP-1C.
d
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38. ROT-4-60 004/ B20/ / / 015AK2.07/ 2.9/ 2.9/ 55/ RCS
OP 302 Section 4.7.2 and Enclosures 3 & 4; 2/3-63
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39. ROT-4-14 007/B1///00'K1.05/4.5/4.4/33/NNI/ICS
The following plant conditions exist:
- NI-5 indicates 73% reactor power.
NI-6 indicates 75% reactor power.
- NI-7 indicates 76% reactor power.
- NI-8 indicates 74% reactor power.
- NI 5/6 selected for control.
Which of the following describes the expected plant response if NI-6 failed low?
.
i
A.
The neutron power signal from RPS to ICS would be 76% power; SASS
'
would transfer and select NI-7/8 for control; CRD system would initially
insert control rods.
B.
The neutron power signal from RPS to ICS would be 76% power; SASS
would not transfer; CRD system would initially withdraw control rods.
C.
The neutron power signal from RPS to ICS would be 73% power; SASS
would transfer and select NI-7/8 for control; CRD system would initially
4
insert control rods.
,
4
vD.
The neutron power signal from RPS to ICS would be 73% power; SASS
would not transfer; CRD system would initially withdraw control rods.
Reasons:
4
A.
The power signal would be the highest of NI-5/6 (73%). SASS would not
transfer because the error seen is only 3% (3.75% error required for
,
transfer). CRD system would initially withdraw control rods.
B.
The power signal would be the highest of NI-5/6 (73%).
C.
SASS would not transfer because the error seen is only 3% (3.75% error
required for transfer). CRD system would initially withdraw control rods.
.
ROT 4-09 Section 2.1.3, 2.1.4 & Figure 27;2/3 18
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40. ROT-5-100 002//// 055EK1.02/4.1/4.4/ 33/ EOP 12
The following plant conditions exist:
- A Station Blackout has occurred.
RCS pressure is 2000 psig.
- Subcooling margin is 38' F.
Which of the following describes an available method of core cooling?
A.
FWP-7 with flow control valves automatically maintaining level and the
ADVs automatically maintaining OTSG pressure.
vB.
EFP-2 with flow control valves automatically maintaining level and the
ADVs automatically maintaining OTSG pressure.
C.
FWP-7 with flow control valves automatically maintaining level and the
TBVs automatically maintaining OTSG pressure.
D.
EFP-2 with flow control valves automatically maintaining level and the
TBVs automatically maintaining OTSG pressure.
Reasons:
A.
FWP-7's flow control valves cannot automatically maintain level.
C.
FWP-7's flow control valves cannot automatically maintain level and
the TBVs are not available.
D.
TBVs are not available.
i
EOP 12; 1-68
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41. ROT-2-32 001/G7///2.3.10/2.9/3.3/77/ RAD'
!
A pre-planning meeting for the performance of OP-407S, Makeup and Spent Fuel
j
Filter Changes, is in progress. The following post-filter dose rates are listed:
l
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'
Filter housing on contact - 25 mR/hr.
'
Filter housing general area - 10 mR/hr.
- Room above the filter room (slots in floor) general area - 2 mR/hr.
.
I
In order to keep personnel exposure at a minimum which of the following
i
combinations of workers should be used for this filter change?
!
.
i
vA.
Two people in the upper room for 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using long handled tools.
t
I
B.
Two people, one in the upper room for 0.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> using long handled
i
tools and the other in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
!,
{
C.
Three people, one in the upper rnom for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> using long handled
i
tools, one in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and the last one in
]
contact with the filter housing for 0.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
i
D.
Three people, two in the filter housing area for 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> each and the
last one in contact with the filter housing for 0.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />.
-
!
l
Reasons:
i
B. The total dose received is 6.5 mR which is greater than 6 mR.
I
-
C. The total dose received is 12.25 mR which is greater than 6 mR.
4
!
I
l
D. The total dose received is 16.25 mR which is greater than 6 mR.
i
.
.
,
,
,
ROT-2 32 pages 19 through 24; 2/3-12
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42. ROT-3 22 001/ B4/// E04EA2.1/3.2/4.4/33/ EOP/AP
l
The following plant conditions exist:
(assume allImmediate Actions are complete for EOP-2, Vital System Status
Verification)
.
i
.
- A reactor trip has occurred.
- The "A" steam generator levelis 87% and increasing.
.
- The "B" steam generator levelis 43% and increasing,
j
- All Reactor Coolant Pumps (RCP) are operating.
- Both Emergency Feedwater Pumps (EFW) are' operating.
!
- Reactor coolant temperature is 533*F.
,
- Reactor coolant pressure is 1850 psig.
'
.
Based on the above conditions which of the following describes the appropriate
Emergency Operating Procedure and associated rule for this situation?
1
i
A.
EOP-2, Vital System Status Verification; EOP-13 Rule 4, Pressurized
'
Thermal Shock.
l
B.
EOP-2, Vital System Status Verification; EOP.-13 Rule 3, EFW Control.
,
,
1
!
C.
EOP 5, Excessive Heat Transfer; EOP-13 Rule 4, Pressurized Thermal
i
Shock.
'
vD.
EOP-5, Excessive Heat Transfer; EOP-13 Rule 3, EFW Control.
Reasons:
A. & B. Due to RCS temperature the conditions are met for entering EOP 5.
C. RCS temperature is not low enough for entry into Rule 4 (< 380 * F.)
ROT 5 96 B4; EOP-13 pages 7 and 9; EOP-5 page 1 ROT 3-22 page 1; ROT 5-14
pages 20 & 21; 2/3-95
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43. ROT-4-12 002/ B2///012K3 01/3.9/4.0/33/RPS
'
The following sequence of events are in progress:
{
Unit is at 100% power with the "A" RPS channel in bypass for testing.
-
]f
- A feedwater flow problem causes RCS pressure to exceed the RPS high pressure
trip setpoint.
- RPS channels "B" and "D" actuate as designed.
-
- RPS channel"C' does not actuate due to a failed RCS pressure bistable.
,
1-
l
Which of the following describes the expected response of the 'RPS/CRD system?
,
1
j
A.
No CRD breakers will open.
!
VB.
All CRD breakers will open.
,
C.
The "B" & "D" breakers and the "F" electronic trip will open. This will
3
l
not result in a reactor trip because the "A" and "C" CRD breakers and
j
the "E" electronic trip do not open.
!
l
D.
The "B" & "D" breakers and the "F" electronic trip will open. This will
I
result in a reactor trip even though the "A" and "C" CRD breakers and
"E" electronic trip do not open.
!
\\
Reasons:
4
A., C. & D.
With one RPS channelin bypass the trip logic is modified to a 2 out
'
of 3 scheme. RPS channels "B" and "D" actuated as designed and
'
tripped all CRD breakers.
!
!
.
4
1
!
ROT-4-12 Section 4.4.2;1-41
BANK; ROT-412 #32
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44. ROT-4-69 002/ B8/// 055K3.01/ 2.5/ 2.7/ 33/ AR
The following plant conditions exist:
- The "A" Air Removal Pump (ARP-1A) has tripped.
ARP-1B auto-starts.
Condenser vacuum is 26 inches Hg and decreasing.
4
- Unit load is at 85% full power.
i
What will be the mode of operation of ARP 1B and the status of the plant if
j
condenser vacuum decreases to 25 inches Hg?
j
A.
ARP 1B will be in the holding mode and the plant will trip.
,
B.
ARP-1B will be in the hogging mode and the plant will trip.
!
VC.
ARP-1B will be in the holding mode and the plant will be at 85%.
D.
ARP-1B will be in the hogging mode and the plant will be at 85%.
j
Reasons:
'
A. Condenser vacuum did not get low enough for the plant to trip.
B. At 25 inches Hg ARP-1B should be in the holding mode. Condenser vacuum
did not get low enough for the plant to trip.
D. At 25 inches Hg ARP-1B should be in the holding mode.
4
ROT-4-69 Section 4.2.2.2; OP-607 Step 3.1; AR-602 EP 0025; 2/3-49
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45. ROT-5-99 001/ B2/// 025AK1.01/ 3.9/ 4.3/44/ AP-404
A plant cooldown is in progress. The "A" decay heat system has just been put in
service. If both trains of decay heat are lost at this point, what alternate method
of core heat removal should be used?
A.
Establish cooling with the spent fuel system.
4
vB.
Establish OTSG cooling.
!
C.
Establish BWST gravity drain cooling.
D.
Establish LPI cooling.
,
Reasons:
1
A.
Because of seismic concerns spent fuel cooling is not used for decay heat
removal.
C.
RCS is not vented to atmosphere.
D.
With DH just put into service LPI would not inject water into the RCS
due to the high pressure.
.
AP-404 Steps 1,3.2, 3.5-3.11, 3.15; 1-88
BANK; ROT 5-99 #13; ROTS J - Final 96; ROTS K - T2
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46. ROT 5-01011/ A1///103K3.02/ 3.8/ 4.2/ 55/TS -
The following plant conditions exist:
'
The plant is at 40% power.
< SP-181, Containment Air Lock Test, is in progress on the_ personnel hatch.
- The ISI test engineer informs you that the Leak Rate Monitor reads off-scale
high.
- There are indications ofleakage around the shaft of the handwheel for
operating the outer hatch.
The inner hatch is operable.
-
Which of the following describes the required actions?
A.
Check inner hatch for leakage. Ifinner hatch is acceptable, initiate
work request for leakage around the shaft of the handwheel for the outer
hatch. Repair must be complete within 31 days.
vB.
Verify the operable door is closed in the affected air lock within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />;
lock the operable door closed in the affected air lock within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and
verify an operable door is locked closed in the affected air lock one per 31
days.
1
C.
Initiate action to evaluate overall containment leakage rate; verify a
'
door is closed in the affected air lock in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; and restore air lock to
operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and Mode 5 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Reasons:
,
A.
TS 3.6.2 must be entered if either air lock door is inoperable.
C.
This action is not required when at least one air lock door is operable.
l
D.
This action is only required if both air lock doors are inoperable and the
.
actions of Condition C are not met.
1
i
TS 3.6.2; SP-181 Step 3.6 and 5.2; 2/3 53
.
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47. ROT-4-06 001/ 89/// 064A2.03/ 3.1/ 3.1/ 33/ EDG
SP 354A, Monthly Functional Test of the Emergency Diesel Generator, is in
progress with the EDG output breaker closed and an electricalload of three
megawatts. A grid disturbance occurs and grid frequency decreases.
Which of the following describes the effect this will have on the operating EDG?
A.
The EDG output breaker will automatically open due to a Volts / Hertz
lockout rclay actuation.
B.
The EDG output breaker will automatically open due to a reverse power
relay actuation.
C.
There should be minimal effect to the EDG due to the Unit / Parallel
switch being selected to Parallel.
vD.
There should be minimal effect to the EDG due to the Speed Droop
being set at 60.
1
Reasons:
,
A.
A Voltz/ Hertz lockout relay actuation will only open the main generator
j
output breakers.
1
i
B.
If grid frequency decreases the could only attempt to increase power cut.
C.
The Unit / Parallel switch being selected to Parallel only affects voltage
droop. The speed droop selection determines how the EDG responds to
changes in frequency.
.
ROT 4-06 Section 4.G; ROT 2-16 Sections 7.16 & 7.18;2/3-52
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48. 30T-5-101 002/ A2/ / / 2.3.11/ 2.7/ 3.2/ 55/ EOP-6
The following plant conditions exist:
- A tube rupture has occurred in the "A" OTSG.
Prior to the tube leak dose equivalent I-131 activity was 0.5 Ci/cc.
-
- As the reactor is being shutdown a Startup transformer protective relay
,
actuates.
- Chemistry samples indicate a transient peak dose equivalent I-131 activity of 97
.
Ci/cc.
Based on the above information which of the following describes the requirements
for operation of the OTSGs?
VA.
The "A" OTSG must be isolated as soon as possible. Steam the "B"
OTSG to atmosphere.
B.
The "A" OTSG must be isolated as soon as possible. Steam the "B"
OTSG to the condenser.
C.
Neither OTSG needs to be isolated. Continue the cooldown steaming
both OTSGs to the condenser.
D.
Neither OTSG needs to be isolated. Continue the cooldown steaming
both OTSGs to atmosphere.
Reasons:
B. The condenser is not available.
C. The "A" OTSG's peak dose equivalent I-131 is too high and must be isolated as
soon as possible. The condenser is not available.
D. The "A" OTSG's peak dose equivalent I-131 is too high and must be isolated as
soon as possible.
ROT-5-101 pages 10 & 11; EOP-6 Step 3.36 and Figures 1 & 2;2/3-13
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49. ROT-4-63 003/B1///029K4.03/3.2/3.5/33/ PURGE
A reactor building (RB) purge is in progress when the following occurs:
- AHF-7A, the "A" purge exhaust fan trips (AHF-7B is still operating).
- RM.Al gas channel has reached its warning setpoint.
What is the condition of the purge supply fans (AHF-6A/6B) and AHV-1C (supply
valve inside RB)?
A.
Both fans are operating; AHV-1C is open.
B.
Neither fan is operating; AHV-10 is open.
vC.
Neither fan is operating; AHV 1C is closed.
D.
One fan is operating; AHV-1C is closed.
Reasons:
A. Both pu ge exhaust fans must be running for a supply fan to remain running.
AHV-1C automatically closes with the loss of both supply fans.
B. AHV-10 automatically closes with the loss of both supply fans.
D. Both purge exhaust fans must be running for a supply fan to remain running.
ROT-4-63 Section 2.8;2/3-44
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50. ROT-4-91001/F2///057AK3.01/4.1/4.4/33/VITALPOWER
With the plant at 100% power a catastrophic failure of VBIT-1C rendered itself
inoperable and caused both of the VBXSs that it feeds to fail as is and not transfer
to their alternate power supply.
Which of the following describes the EOP/AP action (s), if any, that should be
taken?
.
A.
AP 581, Loss of NNI-X, should be entered.
'
B.
AP 582, Loss of NNI-Y, should be entered.
vC.
AP-430, Loss of Control Room Alarms, should be entered.
D.
Trip both MFW pumps and the reactor due to the loss ofICS power.
EOP-2, Vital System Status Verification, and Rule 3, EFW Control,
should be entered.
Reasons:
A.
The ABT for NNI-X should transfer to VBDP-1 on a loss of VBDP-5.
B.
Neither VBDP-5 or 9 feed NNI-Y therefor no loss of power should occur.
D.
This is the correct response for a loss ofICS power however neither
VBDP-5 or 9 feed ICS therefor no loss of power should occur.
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ROT 4-91 Figure 1: AP-430 Step 3.5: 2/3-69
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51, ROT-5-31002/A2///06E AA2.09/4.1/4.3/44/RSP
Step 3.34 of AP-9E'0, Shutdown from Outside the Control Room, requires the
performance of Enclosure 2, RSD Panel Log Readings. Natural Circulation is in
progress with EFIC controlling OTSG level. The following data is recorded:
OTSG 'A' Operate Level
91%
-
OTSG 'B' Operate Level
92%
-
T
545'
-
cold
-
T
572
hot
-
T
590
incores
- RCS Wide Ran ge Pressure
1600 psig
Based on the above readings which of the following describes the condition of the
vA.
Adequate Subcooling Margin does not exist. EFIC is controlling at the
required level.
B.
Adequate Subcooling Margin does not exist. EFIC should be controlling
level at the Natural Circulation setpoint due to RCPs being secured.
C.
Adequate Subcooling Margin does exist. EFIC is controlling at the
required level.
D.
Adequate Subcooling Margin does exist. EFIC should be controlling
level at the Natural Circulation setpoint due to RCPs being secured.
Reasons:
B.
EFIC should control at the ISCM setpoint, not the Nat Cire setpoint.
C.
Adequate Subcooling Margin does not exist.
D.
Adequate Subcooling Margin does not exist and EFIC should control at the
ISCM setpoint, not the Nat Circ setpoint.
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51. ROT-5-31002/A2///068AA2.09/4.1/4.3/44/RSP
AP 990 Step 3.32 and Enclosure 2; EOP-13 Rule 1 and 3; 2/3-74
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52. ROT-4-00 002/ B4/// 064A4.01/4.0/4.3/ 33/ EDG
The following plant conditions exist:
- A Main Feedwater leak has occurred inside the RB.
- RCS pressure is 1750 psig.
- RCS temperature is 510' F.
- RB pressure has increased to 4.8 psig.
Based on the above conditions which of the following describes the status of the
EDG?
A.
The EDG is not running because RCS pressure has not decreased below
the HPI bistable trip setpoint.
B.
The EDG is not runding because an undervoltage condition has not
occurred.
C.
The EDG is running with the output breaker closed.
vD.
The EDG is running with the output breaker open.
Reasons:
A. & B.
The EDG is running due to RBIC cascading up and actuating HPI.
C.
The output breaker is not closed because no undervoltage condition
exists.
ROT 4-06 Section 1.2.4;2/3-51
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53. ROT-5-01006/ A1///015A1.04/3.5/3.7/55/TS/COLR
The plant was operating at 100% power when control rod 7-2 drops fully into the
core. AP 545, Plant Runback, was entered and the plant is stabilized at 58%
power. A QPT calculation is required per AP-545. Power Range NIs indicate as
follows:
- NI 5
58% power
- NI-6
57% power
- NI-7
58% power
- NI-8
42% power
Which of the following actions would be acceptable for these conditions?
A.
Enter TS 3.2.4 and comply with Condition A.
vB.
Enter TS 3.2.4 and comply with Condition B.
C.
Enter TS 3.2.4 and comply with Condition D.
D.
Enter TS 3.2.4 and comply with Condition F.
Reasons-
A., C. & D.
QPT ratio is 7.9 based on NIs. This is greater than the transient
limit but less than the maximum limit. TS requires actions
associated with Condition B.
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54. ~ ROT-5101 001/ A2/// 033EA2.15// 4.4/ 33/ EOP-06
1
The following plant conditions exist:
~
- An OTSG tube rupture has occurred.
Reactor coolant pressure is 1705 psig.
- Incore temperature is 570*F.
- A LOOP has occurred.
EOP 6, Steam Generator Tube Rupture, Step 3.32 states "Begin depressurization
of RCS". As proced'.ure director which of the following would describe your
recommendatic,n fcr a means of depressurization? -
.
A.
Pressurizer spray.
h.
High pres.sure a
iliary} spray.
,lI $ 9 V
Ter%4e.>o..Comm.)s
/C.
Open the PORV.
D.
De-energize the pressurizer heaters.
.
Reasons:
A.
PZR spray is unavailable.
C. & D.
1535 psig is the minimum pressure allowable without losing
subcooling margin.
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'55. ROT-5-102 001/ A1/// E04EK3.1/ 3.5/ 3.7/ 33/ EOP-04
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EOP-04, Inadequate Heat Transfer, contains the following step:
'
IE at any time adequate subcooling margin is lost,
THEN trip all RCPs within 2 minutes ofloss of adequate SCM.
Why is this action required?
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A.
It insures that the RCPs are stopped in order to reduce break flow.
i
vB.
It insures the reactor coolant pumps are stopped before the RCS void
fraction is too high.
-
C.
It insures that the reactor coolant pumps are available for use should
i
the transient degrade'to an inadequate core cooling situation.
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D.
It insures the reactor coolant pumps are stopped before running for an
!
extended period of time with inadequate suction pressure. This could
i
lead to seal failure due to increased vibration.
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Reasons:
I
A., C. & D.
The two minute concern is based on the possibility of creating a
j
70% void fraction which could lead to core uncovery if RCPs are
secured after this event.
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56. ROT-4-14 008/B1///G2.2.2/4.0/3.5/33/ICS
,
The following plant conditions exist:
- The plant is at 83% power.
- '
The Reactor Demand control station has been selected to Manual.
- All other ICS stations are in their procedurally directed conditions.
In preparation to return the Reactor Demand station to Auto, you select
" MEAS VAR" and note that the pointer is indicating 54%. Which of the following
<
is the method to be used to place the Reactor Demand station back into Auto?
.
.
.
A.
You may place the station into Auto provided that T
is 2 F from
ave
setpoint.
,
B.
The error should be adjusted to zero before selecting Auto by placing the
feedwater masters and the SG/Rx master into ma nual and adjusting the
,
SG/Rx master.
vC.
The error should be adjusted to zero before selecting Auto by using the
t
'I' ave setpoint knob on the T
module.
ave
D.
The error should be adjusted to zero before selecting Auto by using the
" Raise / Lower" toggle switch on the Reactor Demand station.
,
Reasons:
,
'
A.
Procedurally T
error must be at 0 prior to placing the station in Auto
ave
'
to prevent unwanted rod motion.
B.
When selected to "MeasNar" on the Reactor Demand station the error
,
represented is T
error. Manipulating the SG/Rx station will not
ave
'
correct for T
error.
ave
.
.
.
'
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D.
Using the " Raise / Lower" toggle switch on the. Reactor Demand station
,
will directly cause control rods to insert or withdraw.
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56. ROT-414 008/B1///G2.2.2/4.0/3.5/33/ICS
OP-504 Section 4.10; 1-6
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~ 57. ROT 5-31001/A2///067AA2.13/3.3/4.4/44/AP-990
The following plant conditions exist:
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A " Control Complex Fire Alert" alarm has been received.
-
~ A Halon bank has actuated in the Cable Spreading Room.
-
Multiple plant components / equipment are cycling erratically causing a loss of
-
plant control.
Based on the abova conditions which of the following describes the action (s) that
,
should be initiated?
<
.A.
Enter AF-880, Fire Protection, and complete Immediate Actions.
Concurrently perform AP-510, Rapid Power Reduction.
B.
Enter AF-880, Fire Protection, only and perform required actions.
C.
Enter AP-990, Shutdown from Outside the Control Room, only and
perform required actions.
v'D.
Enter AP-990, Shutdown from Outside the Control Room, and AP-880,
Fire Protection, and perform required actions.
Reasons:
A.
AP-990 should be entered due to the loss of plant control. There are no
Immediate Actions for AP 880 and the plant should be tripped.
.
!
B.
AP-990 should,also be entered due to the loss of plant control.
<
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C.
AP 880 should also be entered.
.
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AP-990 Step 1; AP-880 Step 1; 1-72
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58. ROT-4 26 001/F4///036AA2.02/3.4/3.9/88/FH
.;
Refueling operations are underway. As a fuel assembly is removed from the core
RM-G16, Fuel Handling Bridge Radiation Monitor, goes into high alarm. What is
!
the appropriate response to this situation?
1
i
A.
Direct HP to start a survey of the area to validate alarm and ensure
- ]
containment purge is secure.
!
B.
Insert tl e fuel assembly in either the deep end of the fuel transfer canal
or the spent fuel pool, whichever is closer, then evacuate the fuel
handling area.
C.
Insert the fuel assembly back into the core and then evacuate the fuel
handling area.
'
vD.
Evacuate the fuel handling area immediately.
Reasons:
A., B. & C.
Per the limit and precaution in FP-203 personnel are to evacuate
the affected area immediately even if there is a fuel assembly in
the mast.
1
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FP-203 Step 3.2.14; 1-98
BANK; ROT 4-26 #23; ROTS J - Final 96
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59. ROT-5-106 001/B1///2.1.32/3.4/3.8/33/OP-209
A Limit and Precaution in OP-209, Plant Cooldown, states:
The Main Feedwater flow rate through the OTSG high nozzles (EFW nozzles)
6
must not exceed .7 x 10 lbm/hr.
What is the basis for this Limit and Precaution?
A.
To prevent possible damage to EFW nozzles.
vB.
To prevent exceeding OTSG cross flow limitations.
,
C.
To prevent exceeding analyzed stresses on the EFW piping.
D.
To limit the possibility of an overcooling event.
Reasons:
- ,
A., C. & D.
Since the EFW nozzles spray directly on the OTSG tubes too high a
flow may cause damage.
OP-209 Step 3.2.8; 1-5
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60. ROT-5-102 002/ A1/// E04EK1.2/4.0/4.2/44/ EOP-4
The following plant conditions exist:
- An Inadequate Heat Transfer event is in progress due to a loss of main and
emergency feedwater.
- HPI/PORV :ooling has been established in accordance with EOP-4, Inadequate
Heat Transfer.
One RCP in each loop is operating.
OTSG integrity does not exist.
- Adequate subcooling margin has just been lost.
.
Based on these conditions which of the following describes the actions that should
be taken?
A.
Transition to EOP-3, Inadequate Subcooling Margin, due to the higher
priority symptom.
sB.
Stay in EOP-4 and follow Rule 1, Loss of SCM.
C.
When AFW is aligned ani eady for operation then transition to EOP-8,
LOCA Cooldown, beginniag with Step 3.1.
D.
When AFW is aligned and ready for operation then stay in EOP-4 and
maintain > 300 gpm flow to each OTSG.
Reasons:
A.
If adequate subcooling margin is lost due to HPI/PORV cooling EOP-4
requires you to stay in this EOP.
C.
The transition to EOP-8 is only required if OTSG heat transfer
restoration is not expected.
D.
Per Table 2 of EOP-14 the required flow rate is < 300 gpm to each
OTSG.
EOP-4 Step 3.10; 2/3-79
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61. ROT-4-14 006/B1///041 A1.02/3.1/3.2/33/ICS
During a unit start-up the following conditions exist:
.,
!
-- Turbine header pressure is 880 psig and increasing slowly.
'
- Turbine is in operator auto with one generator output breaker closed.
- Megawatt output is 130 megawatts.
- Reactor power in 17%.
- All turbine bypass valves are closed.
f
If the turbine header pressure continues to increase the turbine bypass valves
'
should begin to open when header pressure exceeds:
,
1
A.
885 psig
vB.
935 psig
C.
1010 psig
D.
1025 psig
Reasons:
A., B. & C.
With the turbine and reactor not tripped and all TBVs closed with
less than a 10#~ header pressure error the bias applied is 50#.
1
ROT 414 Section 3.2.5; 2/3-5G
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62. ROT-5-100 001/ A1/// 0051 AK3.01// 3.1/44/ EOP-12
The following plant conditions exist:
- A loss of off-site power (LOOP) has occurred.
- The "A" Emergency Diesel Generator (EDG-1A) has failed to start.
- The "B" Emergency Diesel Generator (EDG-1B)is running and supplying the
"B" 4160V ES Bus.
- All "B" side emergency safeguards components are operating properly.
How is the reactor core being cooled?
A.
The Electric Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Atmospheric Dump Valves.
B.
The Electric Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Turbine Bypass Valves.
vC.
The Steam Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Atmospheric Dump Valves.
D.
The Steam Driven Emergency Feedwater Pump is feeding the OTSGs.
Steaming is through the Turbine Bypass Valves.
Reasons:
A. EDG-1A is not operating therefore the Electric Driven Emergency Feedwater
Pump is not operating.
B. EDG-1A is not operating therefore the Electric Driven Emergency Feedwater
Pump is not operating. If all off-site power is lost then there is no vacuum and
the TBVs cannot be used.
D. If all off-site power is lost then there is no vacuum and the TBVs cannot be
used.
ROT 5-100 page 9; 2/3-67
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63. ROT-4-10 001/ B2/// 0032AK1.01/2.5/ 3.1/ 33/ NI
A plant startup is in progress when the detector power supply for NI-1 fails. The
following detectors indicate as follows:
5
- NI 2 indicates 2 x 10 counts per second
- NI-3 indicates 2 x 1010 amps
- NI 4 indicates 2 x 1010 amps
Based on the above conditions which of the following describes the effect this
failure will have on the control rod withdrawalinhibit bistable?
.
A.
No effect. NI-2 will trip the rod withdrawalinhibit bistable if SUR
exceeds 1.0 dpm.
vB.
No effect. NI-2 will trip the rod withdrawalinhibit bistable if SUR
exceeds 2.0 dpm.
C.
This failure will defeat the rod withdrawalinhibit bistable trip signal
from the SR detectors. IR detectors will trip this bistable if SUR exceeds
2.0 dpm.
D.
This failure will defeat the rod withdrawalinhibit bistable trip signal
from the SR detectors. IR detectors will trip this bistable if SUR exceeds
3.0 dpm.
Reasons:
.
A.
The rod withdrawalinhibit bistable trip setpoint is 2.0 dpm. Reset is 1.0
dpm.
C.
NI-2 is still operational. The rod withdrawalinhibit bistable trip setpoint
for the IR detectors is 3.0 dpm. Reset is 2.0 dpm.
D.
NI-2 is still operational.
.
ROT-4-10 Section 2.0; 2/3-91
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64. ROT-4-60 006/ B20//0020101018////33/ RC
A limit and precaution of OP 302, RCP Operation states:
Seal Injection Flow Control Valve MUV-16 should be closed if seal injection is lost
and then gradually restored to 10 gpm/ pump.
Why is seal injection restored gradually?
vA.
Seal temperatures may be high and the cool sealinjection could cause
damage due to thermal stress if flow is increased rapidly
B.
Rapid increase in seal injection flow will alter the seal staging which
could cause damage to the seals.
C.
Seal lubrication is being provided by water from the reactor coolant
system and this must be changed gradually to prevent vapor forming in
the SW system
D.
To prevent damage to the sealinjection filters.
Reasons:
.
B., C. & D. The overriding reason for slowly restoring sealinjection is the
concern with the thermal stresses it creates on the seal packago.
.
OP-302 Step 4.9.1; OP-402 Step 3.2.10; 1-87
BANK; ROT 4-60 #11; ROTS J - T6; ROTS K - T2
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65. ROT-4-52 003/ B1T// A02AK1.3/ 3.8/ 3.8/ 33/ MU
The plant has experienced a loss of NNI-X power. This causes the letdown high
temperature interlack to isolate letdown. The NSS directs you to reduce seal
injection flow to decrease the rate that pressurizer level is increasing due to a
possible delay with opening MUV-49. How would you accomplish this?
A.
Direct the PPO to manually reduce flow by throttling MUV-18,~ seal
.
j
injection isolation valve.
B.
Reduce the setpoint for MUV-16, seal injection control valve.
-
1
'
vC.
Operate MUV-16, seal injection control valve, hand / auto station in
manual.
D.
Direct the PPO to isolate MUV-16, sealinjection control valve, and use
the manual bypass valves to allow continued sealinjection.
6
Reasons:
A. & D.
There is no procedural guidance for this action.
1
B.
MUV-10 will not work in automatic with a loss of NNI-X.
AP-581 Step 3.14; ROT 4-9 Section 3.3; ROT 4-52 Section 1.4.24;1-81
BANK; ROT 4-52 #2; ROTS J - T5
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66. ROT-4-15 003/ B7/// 039K4.05/ 3.7/ 3.7/ 33/ EFIC
Which of the following conditions will cause both MSIVs on the "A" OTSG to close?
A.
Plant shutdown is in progress. All EFIC channel bypass pushbuttons
are depressed when both OTSGs reach 690 psig. "A" OTSG is now 590
psig and "B" OTSG is 610 psig.
B.
Plant shutdown is in progress; All EFIC channel bypass pushbuttons
have not been depressed when both OTSGs reach 760 psig. The MFLI
pushbuttons for the "A" OTSG are depressed.
C.
Plant startup is in progress. All EFIC channels had been bypassed.
When both OTSGs reach 690 psig the "A" OTSG develops a steam leak
and pressure decreases to 590 psig. "B" OTSG remains at 690 psig.
vD.
Plant startup is in progress. All EFIC channels had been bypassed.
When both OTSGs reach 700 psig the "A" OTSG develops a steam leak
and pressure decreases to 590 psig. "B" OTSG remains at 760 psig.
. Reasons:
A.
EFIC is bypassed in this condition and will not actuate at the 600 psig
"
setpoint.
B.
EFIC is not bypassed at this point however a manual MFLI will not
close the MSIVs. Only an automatic MFLI.
,
C.
Since both OTSGs were below 732 psig the EFIC system was still
bypassed. The bypass is automatically removed when the OTSGs
increase above 732 psig.
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67. ROT-5-116 002/ A1/// 011 EA2.11/ 3.9/4.3/ 33/ EOP
Under which of the following conditions is HPI throttling criteria met?
A.
HPI may be throttled if:
"A" OTSG was isolated due to a tube rupture.
RCS pressure is 985 psig.
RCS T
is 500* F.
incore
B.
HPI may be throttled if:
"A" OTSG automatically isolated due to a steam leak.
RCS pressure is 900 psig.
RCS T
is 500* F.
incore
vC.
HPI must be throttled if:
A LOOP has occurred.
RCS pressure is 900 psig.
RCS T
is 350 F.
incore
D.
HPI must be throttled if:
A large break LOCA has occurred.
RCS pressure is 500 psig.
RCS T
is 400* F.
incore
,
Reasons:
A.
RCS pressure due to a tube rupture meets part of the throttling
,
criteria however since adequate SCM does not exist HPI cannot be
throttled.
B.
Adequate SCM does not exist. HPI cannot be throttled.
D.
Adequate SCM does exist. HPI may be throttled but is not required.
.
Rule 2, HPI Control; Rule 4, PTS: 2/3 62
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68. ROT-5-01003/A1///001AK3.02/3.2/4.3/55/AP/TS
The following plant conditions exist:
j
- A Continuous Control Rod Motion event was in progress.
AP-525 was entered and control rod motion was stopped.
- Initial reactor power was 30% with a Rod Index of 180.
- Final reactor power is 60% with a Rod Index of 200.
Based on the above conditions which of the following action (s), if any, should be
bitiated?
i
l
A.
No action is required. Rod Index is acceptable for this power level.
N
vB.
Enter TS and verify Fq and F AH are within limits once every two
hours and restore regulating rod groups to within limits in < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l
C.
Enter TS and initiate boration to restore SDM to > 1% Ak/k within 15
minutes and restore regulating rod groups to within restricted operating
region in < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
D.
Enter TS and red.uce thermal power to < the thermal power allowed by
the regulating rod group insertion limits < 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Reasons:
A.
Conditions for entry into TS 3.2.1 are met.
C. & D.
The rods are in the Restricted Region. These actions are only required if
the rods are in the Unacceptable Region.
ROT 5-07 page 1; TS 3.2.1 & SR 3.2.5.1; OP 103D Curve 1; 2/3-58
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69. ROT-S-101003/81/ ROT-449//2.4.1/4.3/4.6/33/ICS/EOP
Assume the following plant conditions:
NI-5 is out of serdce for maintenance
SASS channel for ~ neutron power in bypass with NI-7/8 selected
'
Plant has just had a spurious reactor trip from 100% full power
J
Due to control signalinterference the output voltage signal from the high select
'
module associated with NI-7/8 remains at 100% reactor power.
Select the statement below which describes the operator's most probable actions
with the above conditions.
a
A.
Perform 30P-2, Vital System Status Verification. No adverse conditions
,
were cretted with the above failures.
B.
Perform EOP 2, Vital System Status Verification. Due to the above
j
failures EOP-4, Inadequate Heat Transfer would also be entered.
vC.
Perform EOP-2, Vital System Status Verification. Due to of the above
failures EOP-5, Excessive Heat Transfer would also be entered.
D.
Perform EOP-2, Vital System Status Verification. Due to the above
'
failures EOP-9, Natural Circulation Cooldown would also be entered.
Reasons:
A., B. & D.
Due to this failure RFR will not decrease feedwater and the
reactor cross-limit will also attempt to maintain feedwater flow at
95% demand.
O
ROT 4-14 Section ; ROT 4-09 Section ; 2/3-92
BANK; ROT 414 #113
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Page: 75
70. ROT-4-09 006/B1///027AK2.02//2.6/33/NNI
The following plant conditions exist:
'
- The plant is operating at 100% power.
- RC-2-TE1, PZR temperature, fails low.
i
Which of the following pressurizer levelinstruments, if any, should the operator
select for control?
,
.
A.
LT1
B.
LT2
vC.
LT3
D.
SASS will automatically transfer to the operable level transmitter. No
operator input is required.
Reasons:
A.
RC-2-TE1 is the temperature compensation input to RC-1-LT1.
B.
RC-1 LT2 is an uncompensated levelindication. It does not use a
temperature input and is no longer used for control.
D.
SASS will only transfer to the operable transmitter if the actual LT
were to fail. A failure of the temperature input will not cause a rate
of change at a greet enough magnitude for SASS to sense the
failure.
ROT 4-00 Section 2.2,2.3 & Figure 5; 2/3-89
Modified Bank; ROT 4-09 #26; ROTS N - T3 & 3A
NRCM98.TST Version: 0
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71. ROT-4-25 005/ F2/// 071 A4.26/ 3.1/ 3.9/ 88/ RM
In preparation for a waste gas release a " check source" of the Area Radiation
Monitor (ARM)is required. Which of the following describes the purpose for
performing this action?
A.
To perform a calibration of the ARM meter to a specific value.
B.
To ensure the detector and its circuits respond to an electronic test
signal.
vC.
To ensure the detector and its circuits respond to actual radiation.
D.
To provide the meter reading for adjusting the alarm setpoints.
Reasons:
A., B. & D.
The check source can only verify that the detector and
associated circuits will respond to radiation.
,
ROT 4-25 page 5; 1-3G
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72. ROT 452 002/ B18/// 004A3.09/ 3.3/ 3.2/ 33/ MU
The following plant conditions exist:
- Plant is operating at 100% full power.
"B" MUP is in operation.
-
Channel "B" (LT-2) is selected for MUT indication.
Which of the following describes the response of the makeup system if the
Channel "A" (LT-1), MUT level transmitter fails low?
VA.
Only "A" side suction valve will receive an open signal.
B.
Only "B" side suction valve will receive an open signal.
.
C.
Both "A" and "B" side suction valves will receive an open signal.
D.
Neither BWST suction valve will receive an open signal.
Reasons:
B., C. & D.
These transmitters control their individual BWST suction valves.
ROT 4-52 Section 1.4.20;1-22
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73. ROT-4-13 002/ B6/ / / 006 (4.09/ 3.8/ 4.2/ 44/ ES
The following plant conditions exist:
,
- A large break Loss of Coolant Accident (LOCA) has occurred.
- Reactor Building (RB) pressure is 17 psig.
.
,
- Reactor Coolant (RCS) pressure is 550 psig.
- Borated Water Storage Tank (BWST) levelis 21 feet.
!
Based on the above conditions which of the following describes the suction source
for the High Pressure Injection (HPI) pumps as prescribed by EOP-3, Inadequate
Subcooling Margin?
!
vA.
MUV 58 ind MUV-73, make-up pump supply valves from the BWST.
B.
DHV 42 and DHV-43, decay heat pump supply valves from the RB
l
sump.
l
C.
MUV-64, make-up pump supply valve from the make-up tank.
!
D.
DHV-11 and DHV 12, make-up pump supply valves from the decay heat
system.
Reasons:
'
,
!
B., C. & D.
With the BWST level above 20 feet HPI pump suction will be from
j
the BWST, not the make up tank. Steps are taken at 20 feet to start
!
alignment for piggy-back operation /RB sump suction.
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ROT-4-13 Table 2; EOP 3 Step 3.4 & 3.13; 2/3-39
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74. ROT-4-28 002/B3///014K4.06/3.4/3.7/44/CRD
The following information is available from the CRD PI panel and the computer
for Absolute Position Indication (API) and Relative Position Indication (RPI):
Cont. ol
RPI
API
Rod
(PI Panel)
(Computer) (PI Panel) (Con;puter)
7-1
92
92.023
93
93.789
7-2
93
93.475
92
93.021
7-3
88
89.987
85
85.987
7-4
93
92.895
93
92.245
7-5
94
93.507
94
94.997
70
91
90.842
92
93.042
7-7
91
91.027
92
93.525
7-8
92
92.357
92
91.778
From the above information evaluate the rod position indication with regard to
Asymmetric conditions / faults and determine which of the following is the correct
indication?
.,
A.
Asymmetric Fault PI panel
OFF
Asymmetric Fault Diamond panel
OFF
'
B.
Asymmetric Fault PI panel
OFF
f . cr Pest cs n Conv}sAsymmetric Fault Diamond p/9
anel
ON
S
0A
'll S
'
C
Asymmetric Fault PI panel
ON
Asymmetric Fault Diamond panel
OFF
@.
Asymmetric Fault PI panel
ON
Asymmetric Fault Diamond panel
ON
Reasons:
A- . B. & C.
Rod 7-3 is G.025% below the group average. An Asymmetric
condition does exist and both the Asymmetric Fault lights should be
lit. (PI Panel and Diamond Panel)
NRCM98.TST Version: 0
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74. ROT-4-28 002/ B3/// 014K4.06/ 3.4/ 3.7/44/ CRD
ROT 4-28 pages 15 & 16; 2/3 24
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75. ROT-5-99 002/ B1/// E04EK2.2/ 4.2/ 4.2/ 33/ AP-404
.Which of the following would be an entry condition for AP 404, Loss of Decay Heat
Removal?
'
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vA.
During Mode 5 cperations with DHP-1B in service a Loss of Offsite
Power occurs.
i
B.
During Mode 5 operations with DHP-1A in service RWP-3A trips.
C.
Defueling complete with the core off-loaded to the Spent Fuel Pool. The
,
transfer tubes are closed and a leak develops in the "A" RCP seal
package.
D.
Defueling complete with the core off-loaded to the Spent Fuel Pool. The
transfer tubes are open and a leak develops in the "A" RCP seal package.
!
Reasons:
B., C. & D.
The running decay heat train must be lost to enter AP-404.
i
AP-404 Step 1; 1-80
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76. ROT-4-12 003/B3///003K3.04/3.9/4.2/33/RPS
The plant is operating at 60% power with the "D" reactor coolant pump secured.
'
An opticalisolator for the "B" reactor coolant pump fails. What is the expected
plant response to this situation?
.
A.
None of the RPS channels will trip.
B.
All four LRPS channels will trip resulting in a reactor trip.
j
C.
The RPS channel containing the failed optical isolator will trip and its
associated CRD breaker will open.
.
vD.
The RPS channel containing the failed optical isolator will trip but no
CRD breakers will open.
.,
Reasons:
.,.
j
A.
The RPS channel containing the failed opticalisolator will trip because it
'
sees a loss of two RCPs.
B.
Only the channel containing the failed optical isolator will tiip.
.
1
C.
The CRD breaker requires two RPS channels to actuate.
.
4
ROT 4-12 Section 2.1.4.E; 2/3-20
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BANK; ROT 4-12 #39; ROTS J Final 96
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77. ROT-5-116 003/ A1///013A1.01/4.0/4.2/44/EOP-RULES
The following plant conditions exist:
- A LOCA has occurred.
- EOP-3, Inadequate Subcooling Margin, was performed.
- Current RCS pressure is 1300 psig.
- Current RCS temperature is 400" F.
Based on these conditions which of the following describes the appropriate method
for HPI control?
.
A.
Rule 1, Loss of SCM, is in effect. HPI may not be throttled.
B.
Rule 2, HPI Control, is in effect. HPI may be throttled.
"
Rule 1, Loss of SCM, and Rule 4, PTS, is in effect. HPI must be
throttled.
> :i
Rule 2, HPI Control, and Rule 4, PTS, is in effect. HPI must be
throttled.
'
Reasons:
A. & C.
Adequate SCM exists. Rule 1 is not applicable.
i
B.
HPI must be throttled because a PTS event has occurred and Rule 4,
PTS, is also in effect.
EOP-13, Rules 1,2 & 4; 2/3-19
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78. ROT-5-01010/89///2.2.21/3.1/3.8/55/TS
Given the following plant conditions:
- The plant is in Mode 3.
t
- The diesel fuel storage tank readings are as follows:
EDG "A" - 8'
2"
EDG "B" - 7'
1"
Which of the following describes required action (s) for this situation?
A.
Restore fuel oil to within limits in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
B.
Verify combined stored fuel oil level > 45,834 gallons within I hour.
C.
Both EDGs are inoperable; restore one to operable status in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
vD.
Immediately declare the "B" EDG inoperable and restore to operable
status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Reasons:
A.
Only the second requirement of Condition B is met.
B.
Condition A is not met with the storage tank levels given.
C.
1
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TS 3.8.1 & 3.8.3: OP-103F pages 36 & 37; 2/3-9
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79. TRE-007 001/ 9/// 028A1.01/ 3.4/ 3.8/ 55/ EC
Which of the following describes how post-LOCA hydrogen is controlled?
VA.
If hydrogen concentration is predicted to reach 3.5% within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />
and winds are from 45 then perform intermittent purge for 120
minutes.
B.
If hydrogen concentration is predicted to reach 3.5% within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />
and winds are from 245* then perform intermittent purge for 120
minutes.
C.
If hydrogen concentration is predicted to reach 3.5% within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />
and winds are from 45* then perform a continuous purge.
D.
If hydrogen concentration is predicted to reach 3.5% within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />
and winds are from 245* then perform a continuous purge.
Reasons:
B.
With winds from 245* and concentration < 3.5% intermittent purge
should not be established.
C. & D.
A continuous purge should not be started until hydrogen
concentration is > 3.5%.
EM 225A Steps 3.3, 4.4, 4.5, Enclosure .1 Step 1.18; 2/3-43
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80.' ROT-4-10 002/ B10/// 015K6.01/2.9/3.2/ 33/ N1
The following sequence of events have occurred:
-- A reactor trip has oxurred.
'
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- NI-3 and NI 4 indications have' decreased to = 1 x 10-10 amps when NI-3 fails
j'
mid scale.
Based on the above conditions which of the following describes the impact this
failure will have on the Source Range NI's high voltage power supplies?
4
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.
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.A.
Only NI 1 HV power supply would de-energize.
i
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B.
Only NI 2 HV power supply would de-energize.
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C.
Both NI-1 and NI 2 HV power supplies would de-energize.
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vD.
Neither NI-1 or NI 2 HV power supplies would de-energize.
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Reasons:
!
A., B. & C.
Conditions required to de-energize the HV power supplies:
'
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NI-3 AND 4 > 1 x 10'9 amps
.
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Either NI-5 or G AND NI-7 or 8 > 10% RX power.
i
1
A mid-scale failure of NI 3 is 1 x 10~7
)
amps.
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.
ROT 4-10 Section 2.1.B; 2/3 25
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81. ROT-5-01004/A1///034G2.2.26/2.5/3.7/44/TS
The plant is in Mode 6 with Refueling Operations in progress. The Refueling Shift
Supervisor has been notified that SP-346, Containment Penetrations Weekly
Check During Refueling Operations, has failed due to multiple containment
penetrations not in their required status. Which of the statements below
describes the minimum required actions?
A.
Core alterations may continue but the HP Supervisor must verify that
air flow at the RB Hatch is into the RB.
B.
Core alterations may continue but the Reactor Building Purge System
supply fans must be secured.
C.
Immediately suspend all core alterations however, movement of
irradiated fuelin the fuel transfer canal may continue.
vD.
Immediately suspend all core alterationspf movement ofirradiated fuel
within containment.
d4
Reasons:
A., B. & C.
TS 3.9.3 requires all core alterations or movement ofirradiated fuel
within containment be suspended immediately if one or more
containment penetrations are not in their requned status.
.
.
!
TS 3.9.3; 1-46
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82. ROT-5 96 002/ A1/ ROT-5-14// 026AK3.02// 3.9/44/ Al-505
A step in EOP-02, Vital System Status Verification, states:
lE, at any time, ES systems have,
QE should have actuated,
THEN ensure ES equipment is properly aligned.
.
4
If reactor coolant pressure is 1350 psig which of the following indications and
associated operator responses is in compliance with this step?
A.
The decay heat inlet valve to the reactor coolant system, DHV-5, has a
green ES status light and the control board operator rotates the valve's
switch to open it.
4
'
B.
The "B" Building Spray Pump, BSP 1B, has an amber status light and
the control board operator rotates the pump's control handle to sart it.
C.
High pressure injection valve, MUV-23, has a green status light and the
,
control board operator rotates the valve's switch to open it.
vD.
The "A" Decay Heat C'osed Cycle Cooling Pump, DCP.1A, has an amber
status light and the control board operator rotates the pump's control
'
handle to start it.
1
Reasons:
A. Pressure is too high for an LPI actuation to open DHV-5. A green light
indicates it has opened inappropriately.
B. An HPI actuation will only give a permit for BSP-1B to start. The amber light
was correct and the pump should not be started.
C. A green status light indicates the valve is already open. There is no need to
manipulate the switch.
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ROT-4-13 Table II; 2/3-65
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83. ROT 425 004/B7///068K6.10/2.5/2.9/44/ODCM.
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The following plant conditions exist:
!
- The plant is in Mode 5.
- Chemistry notifies the NSS that RM-L7 has failed its Channel Calibration.
i-
SDT-1 is full and on recirc for release.
.
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What action (s), if any, can be taken to release the contents of SDT-1?
J
!
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A.
The release can continue as planned as long as Chemistry verifles
,
release rate calculations.
,
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B.
The release can continue as planned if a Channel Check, Source Check
'
and Channel Functional Test are performed and meet specified
j
requirements.
I
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C.
The release can continue as planned if the monitor is declared
!
i
inoperable and the applicable TS requirements are met.
,
vD.
The release can continue as planned if the monitor is declared
t
,
[
inoperable and the apphcable ODCM requirements met.
!
i
Reasons:
i
i
A.
Grab samples must be collected and analyzed for gross radioactivity at
least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />,
-
i
B.
Channel Check, Source Check, Channel Calibration and Channel
!
Functional Test are required for operability.
!
!
C.
RM L7 requirements have been removed from TS.
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84.' ROT 5-85 001/ A2///009G2.4.2/ 3.9/4.1/44/ EOP-3
The following plant conditions exist:
,
i -
- -A Loss of Coolant Accident has occurred.
- RCS pressure is 1185 psig.
!
- RCS temperature has been = 530' F for three minutes.
.
.
!
- Reactor Building (RB) pressure is 8,3 psig.
,
1
- One Reactor Coolant Pump per loop is in operation.
1
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Based on the above conditions which of the following actions should be taken?
'
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A.
Trip the operating RCPs due to the loss of SW cooling.
l
B.
Trip the operating RCPs due to the loss of seal return.
l
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C.
Bypass the HPI actuation and restore SW cooling.
i
vD.
Bypass the RBIC actuation and restore seal return.
-
l
Reasons:
A.
RCPs must be kept running due to the loss of adequate SCM for greater
i
l
than two minutes. A loss of SW cooling has not occurred.
B.
RCPs must be kept running due to the loss of adequate SCM for greater
,
than two minutes.
!
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C.
SW cooling has not been lost.
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EOP 3 Step 3.10; 2/3-85
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85. ROT-4-14 009/ B1/// 001 AK1.14/ 3.4/ 3.7/ 33/ AP-525/ICS
Control Rod Group 7 is moving outward. Which of the following is the possible
,
cause of the rod motion?
,
A.
Tcold signal slowly failing high.
i
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vB.
T
signal slowly failing low.
hat
C.
ATeold signal slowly failing high.
D.
Total FWfig, signal slowly failing low.
Reasons:
4
A.
Tcold sign 1 sl wly f iling high will cause rods to insert due to a high
indicated Tave-
C.
ATeold signal slowly failing high will only re-ratio feedwater. Total
l
feedwater flow will remain the same which should not move rods.
D.
Total FW ;g, signal slowly failing low would cause a cross limit to occur
"
7
and control rods would int rt, not withdraw.
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ROT 4-14 Section 3.3.2,3.4.8 & 5.5; 1-60
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Modified Bank: ROT 5-67 #3; NRC 96; ROTS K-Final 97
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86. ROT-5-01005/ 89/// 061G2.1.12/ 2.9/4.0/11/ TS
While performing the surveillance procedure for EFIC Automatic Actuation Logie
(in Mode 1) the Channel A Main Feedwater Isolation (MFWI)is declared
inoperable at 1000,6-15-98. Two hours later the Channel A Main Steam Line
Isolation (MSLI) is ' declared inoperable. All B channel functions are operable.
How much time is allowed for restoration of both functions?
A.
Both functions must be restored by 1000, 6-18-98.
B.
Both functions must be restored by 1600, 6-15-98.
C.
MFWI must be restored by 1600,6 15-98. MSLI must be restored by
1800, 6 15-98.
vD.
MFWI must be restored by 1000, 6-18-98. MSLI must be restored by
1200, 6 18-98.
Reasons:
A., B. & C.
Both channels have 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for completion starting when they
were declared inoperable. Separate condition entry is allowed for
each function.
TS 3.3.13; 131
BANK; ROT 5-01 #46
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The following plant conditions exist:
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A LOOP has occurred.
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- MUP 1B tripped due to a motor overload.
- RCS pressure is 1550 psig.-
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- RCS temperature based on incores is 580* F.
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Based on the above conditions which of the following describes the appropriate
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A.
Enter EOP-3, Inadequate Subcooling Margin, perform Rule 1, Loss of
SCM and manually actuate EFIC.
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- B.
Enter EOP-4, Inadequate Heat Transfer, perform Rule 2, HPI Control,
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and bypass the HPI actuation.
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v'C.
Enter EOP-3, Inadequate Subcooling Margi , perform Rule 1, Loss of
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SCM and manually actuate HPI.
D.
Enter EOP-4, Inadequate Heat Transfer, perform Rule 2, HPI Control,
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and reset the HPI actuation.
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Reasons:
A.
MUP-1B status light will not indicate that the pump is running. Steps
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will have to be taken to ES select and start MUP-1A.
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B. & D.
The conditions are met to enter EOP-3 and an HPI actuation should not
have occurred at this pressure,
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EOP-3; Rule 1; ROT 4-13 Table II; 2/3-99
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88. ROT-4-25 002/F2///0061AK2.01/2.5/2.6/88/RM
The Makeup Pump Area Radiation Monitor, RM.G10, goes into alarm locally and
in the control roor2.
Which of the following describes the actions,if any, associated with this alarm?
A.
Ventilation dampers for the makeup pump area close.
B.
Operatir.g auxiliary building supply fans trip.
C.
Associated waste disposal panel annunciator goes into alarm.
vD.
No other automatic actions are expected.
Reasons:
A., B. & C.
The only automatic actions associated with RM-G10 are the
alarm functions at the local monitor (horn and lights) and in the
control room (annunciator).
ROT-4-25 Section 2.2 & Table 8; 1-97
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- 89. ROT-5-01008/A1///002K3.02/4.2/4.5/55/TS/RCS
' During three RCP operation (RCP-1D secured) the following readings are
recorded:
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- RCS total flow
107 x 10 lbm/hr
- RCS T
"A"1 P
605*F
"B" Loop
605*F
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RCS pressure
"A" loop
2055 psig
"B" Loop l
2090 psig
Based on the above conditions what action (s), if any, are required to be taken?-
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A.
No action required. All parameters are within limits with only three
RCPs in operation.
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B.
A DNBR Safety Limit has been exceed. Be in Mode 3 within one hour.
C.
One DNB parameter is not within limits. Restore the parameter to
within limits in two hours.
vD.
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Two DNB parameters are not within limits. Restore the parameters to
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within limits in two how;s.
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Reasons:
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A. & C.
Two DNB parameters are not within limits.
B.
The DNBR safety limit has not been exceeded per 2.1.1-1.
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TS 3.4.1; TS Figure 2.1.1-1;2/3-37
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90. ROT 5-29 001/ B1/// A04AA2.1/3.3/3.7/33/ AP-660
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The following plant conditions exist:
- The plant is at 30% power.
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Turbine vibration on bearing #7 is 10 mils.
Stator bar. discharge gas temperature is 78'C.
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- Average cold gas temperature is 60*C.
Based on the above conditions which of the following action (s) should be taken?
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A.
Enter AP-510, Rapid Power Reduction, arid perform a plant shutdown.
B.
Enter OP-204, Power Operations, and perform a plant shutdown.
C.
Enter EOP 02, Vital System Status Verification, and trip the reactor.
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vD.
Enter AP-660, Turbine Trip, and trip the turbine.
Reasons:
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.A., B. & C. Average cold gas temperature is above the limit of 55*C per OP-203.
The entry conditions for AP-660 are met.
OP-203 Step 3.2.12; AP-660 Step 1 & 2.1; 1-83
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~91. ROT-4-52 001/B22///004K6.17/4.4/4.6/33/MU
The following plant conditions exist:
- A reactor trip has occurred.
- Control Rods 4 3,6-4 and 6-5 did not fully insert into the core.
- NIs indicate the reactor is shutdown.
If both of the Boric Acid Pumps (CAP 1A and CAP-1B) are incapable of boric acid
injection which of the following would be the appropriate means of " emergency
boration"?
A.
Opening the suction valves (MUV-58 and MUV-73) from the BWST to
the Makeup Pumps and diverting letdown flow to a Reactor Coolant
Bleed Tank. Greater than 10 gpm flow must be insured.
B.
Feed from the highest concentration RCBT through the Batch Feed
Valve (MUV-103) into the Makeup Tank and divert letdown flow to
another RCBT. Greater than 10 gpm flow must be insured.
vC.
Opening the suction valves (MUV 58 and MUV-73) from the BWST to
the Makeup Pumps and diverting letdown flow to a Reactor Coolant
Bleed Tank. Maximum flow rate must be insured.
.
D.
Feed from the highest concentration RCBT through the Batch Feed
Valve (MUV-103)into the Makeup Tank and divert letdown flow to
another RCBT. Maximum flow rate must be insured.
,
Reasons:
A.
This lineup requires > maximum flow through MUV-31.
B. & D.
This is not an approved flow path and most likely wouldn't work due to
the lack of elevation difference.
ROT 4-52 step 4.4.2; EOP-2 Step 3.3; AP-490 Step 3.14; 1-21
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92. ROT-4-09 005/81///002A3.03/4.4/4.6/33/NN!/RCS
During a plant startup a malfunction occurred with the automatic operation of
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MUV-31, RCS makeup control valve, and PZR level decreased to 30". Which of
the following describes how this failure will effect operation of the pressurizer
heaters?
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A.
PZR heater control is not affected by this malfunction.
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vB.
There will be a loss of all automatic and manual P7R heater control.
C.
There will be a loss of all automatic PZR heater control. Manual control
of PZR heater banks "D" and "E" is available.
D.
There will be a loss of automatic PZR heater control of the SCR
controlle 1 heaters. PZR heater banks "D" and "E" will function in
automatic or manual.
Reasons:
A., C. & D.
At less than 40" PZR level all heater functions are locked out.
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ROT 4-9 Section 2.2; 1-38
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93. ROT-5-107 001/B4///005G2.4.4/4.0/4.3/33/EOP-2
SP-333, Control Rod Excercises, is in progress with reactor power at 90%. While
swapping to the auxiliary power supply for Group 4 a malfunction occurs and
control rods 4-3 and 4-4 drop to 70% withdrawn and remain there. Which of the
following action (s) should be taken.
VA.
Trip the reactor and enter EOP-2, Vital System Status Verification.
B.
Reduce reactor power to 60% AP-510, Rapid Power Reduction, and verify
SDM limits are not exceeded.
C.
Reduce reactor power to 60% using AP-545, Plant Runback, and verify
SDM limits are not exceeded.
D.
Reduce reactor power to 60% using OP-204, Power Operations, and
verify SDM limits are not exceeded.
Reasons:
B., C. & D.
Per AI-505 the operator should trip the reactor.
AI-505 Enclosure 1; 2/3-59
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94. ROT-4-09 003/ B1/// 016K4.03/ 2.8/ 2.9/ 33/ NNI
The following plant conditions exist:
- Plant is operating at 100% full power.
- The RC temperature select switch is selected to the "B" position.
- The Auto / Manual T
select switch is selected to the "A LOOP" position.
ave
circuit in the NNI
An I & C technician working on the selected "A" loop Thot
cabinet inadvertently places a test lead in the wrong test jack causing the signal
to begin a slow, steady decrease.
Which statement below is correct concerning Tave?
vA.
Indicated T
n the recorder would not be affected. T
indicated on
ave
ave
the digitalindicator and supplied to the ICS would decrease.
B.
Indicated T
n the recorder would decrease. T
indicated on the
ave
ave
digital indicator and supplied to the ICS would not be affected.
C.
Indicated T
n the recorder and digitalindicator would not be
ave
affected. T
supplied to the ICS would decrease.
ave
D.
Indicated T
n the recorder and digitalindicator would decrease.
ave
T
supplied to the ICS would not be affected.
ave
Reasons:
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B. & D.
The output of the RC temperature select switch will only affect the
recorder. With this switch selected to "B" this output would not
change.
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C.
The output from the Auto / Manual Tave select switch will affect the
digitalindicator. With this switch selected to "A" T
will decrease.
ave
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94. ROT-4-09 003/B1///016K4.03/2.8/2.9/33/NNI
ROT-4-09 Sections 2.10 & 2.11; 1-42
BANK; ROT-4-09 #6 ; NRCCP97; ROTS J T7; ROTS K - T2
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95. ROT-4-15 001/B1///061K2.01/3.2/3.3/33/ EFIC
A maintenance worker inadvertently hits the power supply wiring for EFV-57,
EFP-1 to 'B' OTSG control valve, pulling the wire out of the valve body. The
control room is immediately notifled and a decision is made to isolate this EFW
line and ensure it cannot feed if an EFIC actuation were to occur.
Which of the following action (s), if any, could be taken to ensure this line is
isolated?
A.
No actior. is necessary. EFV-57 and has failed closed due to the loss of
power.
B.
Select mt.nual and closed at EFV-57's control station and select closed
EFV-33, block valve for EFV-57, on the main control board.
C.
Select closed EFV-33, block valve for EFV-57, at the control board and
de-energize its power supply at DPDP-80.
vD.
Select closed EFV-33, block valve for EFV-57, at the control board and
de-energize its power supply at DPDP-8D.
Reasons:
A.
EFV-57 fails open on a loss of power.
B.
EFV-57 fails open on a loss of power. EFV-33 will get an open signalif an
EFIC actu ition occurs.
C.
EFV-33 is powered from DPDP-8D ("B" train power supply).
ROT 4-15 pages 8 and 9; Rule 3; 2/3-32
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96. ROT-5 34 001/ A2///2.4.41/2.3/4,1/55/ EM-202
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The following plant' conditions exist:
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- RMG 29/30 indicate 11,000 R/hr.
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- RCS pressure is 1400 psig.
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- RB Spray has actuated.
RB temperature is 190 F.
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Which of the following is the lowest emergency action level (EAL) that should be
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declared at this time?
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A.
' Unusual Event
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B.
Alert
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Site Area Emergency
D.
General Emergency
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Reasons:
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A.,B.,& D
A Site Area Emergency should be declared if the following
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conditions are met'
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1. > 1000 R/hr and two of the following:
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a. RCS pressure > 1500 psig.
b. Containment pressure > 4 psig.
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c. RB temperature > 180* F.
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EM-202 page 23; 2/3-16
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97.- ROT 5-14 002/B10f//2.1.2/3.0/4.0/33/Al-505
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Which of the following would require the operator to manually trip the reactor?
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A.
One CRD stator temperature is 185* F.
B.
The nuclear services closed cycle cooling surge tank, SWT-1, has a level
of 4 feet.
vC.
Pressurizer level is 95 inches at 100% full power.
D.
One MSIV closes at 80% full power.
Reasons:
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A. Requires at least two stator temperatures to be > 180* F.
B. The manual trip for low SWT-1 levelis 2 feet.
D. Power operation may continue if reduced to 60% full power.
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AI-505 Enclosure 1; 1-1
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98. rot-5-e0 001/ B1// / 069AA1.01/ 3.5/ 3.7/ 33/ AP-250
Which of the following is automatically closed or stopped when the high alarm
setpoint is reached on RMA 1, " Reactor Building Purge Exhaust Duct Radiation
Monitor"?
A.
AHF-1A/1B/1C (RB cooling fans)
B.
AHF 6A/6B (RB purge supply fans)
C.
AHF-7A/7B (RB purge exhaust fans)
vD.
AIW-1A, IB,1C, and ID (containment purge supply / exhaust valves)
Reasons:
A., B. & C.
The automatic actions associated with RMA-1 are to close AHV-1A,
1B,1C, ID and LRV-70,71,72 & 73. Per the AP the purge supply
and exhaust fans must be tripped manually.
AP-250 Steps 3.1,3.3 and Table 1; 1-76
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99. ROM 92 001/ G1///2.1.16/ 2.9/2.8/ 77/ FIRE
Step 3.4 in AP-880, Fire Protection, directs the operator to maintain plant
communications using Party Line 1 and portable radios. Which of the following
describes the reason Party Line 1 is used and how it can be accessed?
VA.
Party Line 1 is the dedicated emergency line and can be accessed from
the UTF 4 Digit Telephone System by dialing 12.
B.
Party Line 1 is the dedicated emergency line and can be accessed from
the UTF 4 Digit Telephone System by dialing 311.
C.
Party Line 1 is the only line that services the entire plant and can be
accessed from the UTF 4 Digit Telephone System by dialing 12.
D.
Party Line 1 is the only line that services the entire plant and can be
accessed from the UTF 4 Digit Telephone System by selecting 311.
Reasons:
B.
Dialing 311 will get the emergency phone in the control room only.
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C.
Party Line 1,2, PAX and the UTF system all service the entire plant.
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D.
Party Line 1,2, PAX and the UTF system all service the entire plant.
Dialing 311 will get the emergency phone in the control room only.
ROT 4 92 pages 4 through 6; OP-704 Step 4.2;173
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100. ROT-5-116 001/ A1/// 059A2.04/ 2.9/ 3.4/ 55/ MFW
The following plant conditions exist:
- The "A" OTSG has blown down to 6" on the EFIC Low Range instrument due to
a failed open MSSV. This valve has now been reseated and gagged closed.
- Main feedwater pumps are the only available source of feedwater.
- Feedwater temperature is 160 F.
As the NSS which of the following actions should be performed?
,
Feed the "A" OTSG via tile EFW nozzles at 450 to 600 gpm.
A.
B.
Feed the "A" OTSG via the MFW nozzles at 450 to 600 gpm.
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C.
Feed the "A" OTSG via the EFW nozzles at < .15 x 10 lbm/hr.
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vD.
Feed the "A" OTSG via the MFW nozzles at < .:5 x 10 lbm/hr.
Reasons:
A.
At temperatures > 150 F the MFW nozzles shcald be used with a flow rate
6
of < .15 x 10 lbm/ min.
6
B.
Flow rate of < .15 x 10 lbm/hr must be maintained.
C.
At temperatures > 150 F the MFW nozzles should be used.
EOP-14 Enclosure 3; EOP-4 Table 3; 2/3-30
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