ML20203F362
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| ML20203F362 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 02/11/1999 |
| From: | Michael B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-413-97-300, 50-414-97-300, NUDOCS 9902180130 | |
| Download: ML20203F362 (1) | |
See also: IR 05000413/1997300
Text
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February 11,1999
NOTE TO: NRC Document Control Desk
Mail Stop 0-5-D-24
FROM: Beverly Michael, Licensing
&]Ab
istant,' Operator Licensing and Human
Performance Branch, Divi n of Reacto: Safety, Region 11
' SUBJECT: OPERATOR LICENSING EXAMINATIONS ADMINISTERED AT CATAWBA
NUCLEAR STATION - DECEMBER 1 - 5 AND ;5 - 19,1997
DOCKET NOS. 50-413 AND 50-414
During the period December 1 - 5 and 15 - 19,1997, Operator Licensing Examinations
were administered at the referenced facility. Attached, you will find the following
infonnation for processing through NUDOCS and distribution to the NRC staff, including
the NRC PDR:
W)item #1 -
@b)
A
a Facility submitted outline and initial exam submittal,
designated for distribution under RIDS Code A070.
As given op,erating examination, designated for distribution under
4.
RIDS Code A070.
Item #2 - Examination Report with the as given written examination attac,hed, ,
designated for distribution under RIDS Code IE42.
Attachments: As stated
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9902190130 990211
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PDR ADOCK 05000413
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February 2. 1998 l
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Duke Energy Corporation
ATTN: Mr. G. R. Peterson
Site Vice President
Catawba Site
4800 Concord Road
York. SC 29745-9635
SUBJECT: NRC SPECIAL INSPECTION AND EXAMINATION REPORT NO. 50-413/97-300
AND 50-414/97-300 AND NOTICE OF VIOLATION
Dear Mr. Peterson:
On December 1-19, 1997, the Nuclear Regulatory Commission (NRC) inspected and
administered examinations to employees of your company who had applied for
licenses to operate the Catawba Nuclear Station Units 1 and 2. Preparation
for the examination was conducted November 17-21. 1997. At the conclusion of
the inspection and examination, the examiners discussed the preliminary
findings and the examination questions and with those members of your staff
identified in the enclosed report.
A Simulation Facility Report is included in this report as Enclosure 3. A
copy of the written examination questions and answer key, as noted in
Enclosure 4 was provided to the members of your training staff at the
conclusion of the examination. Four post-examination comments to the written
examinations were submitted by your letter dated December 18, 1997. This
submittal was later amended by your letter dated January 13. 1998. Both sets
of comments are included in this report as Enclosure 5. NRC resolution of
your post-examination comments is provided as Enclosure 6. Additionally, the
NRC identified one other question with two correct answers and changed the
answer keys accordingly.
Thirteen of fourteen candidates tested passed the examination. However five
of the twelve candidates who passed were identified as having some performi.nce
weaknesses. The individual examination reports should be reviewed to
determine if adjustments to the training program, as well as individual
remediation, are needed. Additionally. two violations of NRC requirements
were identified.
The violations are cited in the enclosed Notice of Violation (Notice) and the
circumstances surrounding the issues are described in detail in the enclosed
report. Please note that you are required to respond to this letter and
should follow the instructions specified in the enclosed Notice when preparing
your response. The NRC will use your response, in part. to determine whether
further enforcement action is necessary to ensure compliance with regulatory
requirements.
In accordance with 10 CFR 2.790 of the NRC's ~ Rules of Practice ~ a copy of
this letter. its enclosures, and your response will be placed in the NRC
Public Document Room (PDR).
DISTRIBUTION CODE
A nm /5
p.A cr w .7/Hi v .. f(e L IE42
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DEC 2
l
Should you have any questions concerning this letter. please contact me at I
(404) 562-4600. I
Sincerely. I
!
Original signed by !
Johns P. Jaudon .
,
Johns P. Jaudon. Director l
l Division of Reactor Safety .
Docket Nos. 50-413 and 50-414
License Nos. NPF-35 and NPF-52
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Enclosures:
1. Notice of Violation
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2. Report Details
3. Simulation Facility Report
4. Written Examination and Answer Keys (SR0 & RO) !
(Document Control Desk Only) ,
- 5. Facility Post-examination Comments ;
! 6. NRC Resolution of Post-examination Comments
,
cc w/encls-
M. S. Kitlan :
'
l Regulatory Compliance Manager
Duke Energy Corporation
4800 Concord Road '
York. SC 29745-9635
Paul R. Newton
Duke Energy Corporation
422 South Church Street
'
- Charlotte. NC 28242-0001
Robert P. Gruber
Executive Director
- Public Staff - NCUC
P. 0 Box 29520
Raleigh. NC 27626-0520
J. Michael McGarry. III. Esq.
Winston and Strawn
1400 L Street. NW
Washington D. C. 20005
l l North Carolina MPA-1 l
- Suite 600
P. O. Box 29513
Raleigh. NC 27626-0513
- :
(cc w/encls cont'd - See page 3)
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DEC 3
l
(cc w/encls cont'd) '
l -Max Batavia. Chief .
'
l Bureau of Radiological Health !
S. C. Department of Health :
and Environmental Control '
2600 Bull Street .!
' Columbia. SC 29201 '
,
Richard P. Wilson. Esq. ,
Assistant Attorney General ;
S. C. Attorney General's Office
P. O. Box 11549
'
Columbia. SC 29211
;
Michael Hirsch ,
Federal Emergency Management Agency !
500 C Street. SW. Room 840 i
Washington. D. C. 20472
'
North Carolina Electric
Membership Corporation
P. O. Box 27306
l Raleigh. NC 27611 j ! '
Karen E. Long l
Assistant Attorney General '
N. C. Department of Justice l
P. Box 629 !
l
Raleigh. NC 27602 :
- ,
l Saluda River Electric
Cooperative. Inc. -
P. O. Box 929
Laurens. SC 129360
i i
County Manager of York County York County Courthouse York. SC 29745 Piedmont Municipal Power Agency 121 Village Drive Greer SC 29651 G. A. Copp Licensing - EC050 Duke Energy Corporation P. 0: Box-1006 Charlotte. NC 28201-1006 (cc w/encls cont'd - See page 4)
.
6
7-
DEC 4
(cc w/encls cont'd)
Peter R. Harden IV
Account Manager
Energy Systems Sales
Westinghouse Electric Corporation
P. O. Box 7288
Charlotte. NC 28241-7288
B. G. Addis
Training Manager
Catawba Nuclear Station
'
4850 Concord Road
York. SC 29745
Distribution w/encls:
P. Tam. NRR
R. Carroll RII
C. Payne. RII
PUBLIC
NRC Resident Inspector
U.S. Nuclear Regulator Commission
4830 Concord Road
' York. SC 29745
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OFFICE Ril:DRS g, Rll:DR$ Ril:DRP , Rll:DRS // RI1:DRS
SIGNATURE //g/' y j gg // gp j
NAME CPayne:p( OL ,Py CDgle IPeebles JJa h l
DATE 1/ a: /98 /1/W 1/ g ,q /98 ff p /98 3/ ) /98 1/ /98 1/ /98
- COPY? 'YES) NO YES No YES NO lYq NO ,/YES NO TES NO YES No
"'
DFFICI AL RECORD COPY DOCUMENT NAME: S:\DRSL97300RPT.DLP
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DISTRIBUTION CODE,'
IE42-
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NOTICE OF VIOLATION
Duke Energy Corporation Docket Nos. 50-413. 50-414
Catawba Nuclear Station License Nos. NPF-35. NPF-52
During an NRC inspection and examinaticn conducted on November 17-21, 1997 and
December 1-19. 1997, violations of NRC requirements were identified. In
accordance with the " General Statement of Policy and Procedure for NRC
Enforcement Actions." NUREG-1600. the violations are listed below:
A. 10 CFR 50.120(b). Training and Qualification of Nuclear Power Plant
Personnel Requirements, as implemented by Catawba Operator Training
Management Procedure. " Design and Development (OTMP 3.0), requires *
licensees to establish, implement and maintain a training program
derived from a systems approach to training as defined in 10 CFR 55.4
for non-licensed operators. Element (5) of the systems ap3 roach to
training requires evaluation and revision of the training Jased on the
performance of trained personnel in the job setting.
Contrary to the above, as of December 19. 1997, selected officers from
the site security force had been trained to perform the non-licensed
operator emergency tasks for starting and operating the Safe Shutdown
Facility (SSF) diesel generator and had not been retrained nor had their
performance evaluated since initial training and testing in December.
1996. When tested in December 1997. four of seven security force SSF
Operators failed all or part of their written and walkthrough
examinations.
This is a Severity Level IV violation (Supplement 1).
B. 10 CFR 55.59(a)(1). Requalification Requirements, requires, in part,
that each (operator) licensee successfully complete a requalification 1
l program developed by the facility licensee. It also requires the 1
program be conducted for a continuous period (cycle) not to exceed 24 l
l months in duration. Additionally.10 CFR 55.59(c)(4). Requalification
- !
Eva7uation requires, in part that the requalification program include
comprehensive written examinations and annual operating tests. These
requirements were implemented by Duke Power Catawba Operations Training
Procedure OTMP 3.0 Revision 7.
j Contrary to the above. between January 1991 and December 1996, the ! comprehensive, biennial written and annual operating tests which should , l have been administered during each of the three 24-month requalification 1
cycles were actually conducted between three and eleven months after
completion of the requalification cycle.
This is a Severity Level IV violation (Supplement 1).
Enclosure 1
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Pursuant to the provisions of 10 CFR 2-.201. Duke' Energy Corporation is hereby 1
required to submit a written statement or explanation to the U.S. Nuclear .!
Regulatory Commission.' ATTN: Document Control Desk. Washington. D.C. 20555 !
with a copy to-the Regional: Administrator. ' Region 11. and a copy to the NRC :
Resident Inspector at the Catawba facility, within 30 days of the date of the l
letter transmitting this Notice of Violation-(Notice). This reply should be -i
clearly marked as a " Reply to a Notice of Violation" and should include for ;
each violation: (1) the reason for the violation, or, if contested, the basis
Efor disputing the violation or severity level. (2) the corrective steps that l
have been taken and the results achieved. (3) the corrective steps that will
' be taken to avoid further violations, and (4) the date when full compliance
- will be achieved. - Your response may reference or include previous docketed
l' correspondence. if the correspondence adequately addresses the required.. l
response. If an adequate reply is not received within the time specified in '
this Notice, an order or a Demand for Information may be issued as to why the i
license should not be. modified, sus) ended, or revoked, or why such other
action as may be proper should not ]e taken. Where good cause is shown.
consideration will be given to extending the response time.
If you contest this enforcement action, you should also provide a copy of your
response to the Director. Office of Enforcement. United States Nuclear
Regulatory Commission. Washington, DC 20555-0001.
Because your res)onse will be placed in the NRC Public Document Room.(PDR). to
the extent possi)le. it should not include any personal privacy. 3roprietary,
or safeguards information so that it can be placed in the PDR wit 1out
redaction. If personal privacy or proprietary information is necessary to !
provide an acceptable response, then please provide a bracketed copy of your
response.that identifies the information that should be protected and a-
. redacted copy of your response that deletes such information. If you request
- withholding of such material, you must s)ecifically identify the portions of
your response that' you seek to have with1 eld and provide in detail the bases
for your claim of withholding (e.g., explain why the disclosure of information
will create an unwarranted invasion of personal privacy or provide the
~information required by 10 CFR 2.790(b) to support a request for withholding.
confidential commercial or financial information). If safeguards information
. is necessary to provide an acceptable response please provide the level of
protection described in 10 CFR 73.21..
Dated at Atlanta. Georgia s
this 2nd day of February 1998
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F Enclosure 1 '
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U. S. NUCLEAR REGULATORY COMMISSION
REGION II
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Docket Nos. 50-413. 50-414
License Nos. NPF-35. NPF-52
Report Nos. 50-413/97-300. 50-414/97-300
Licensee: Duke Energy Corporation
Facility: Catawba Nuclear Station. Units 1 & 2
Location: 4800 Concord Road
York. SC 29745
Dates: November 17 through 21. 1997 and December 1
through December 19. 1997
Examiners: '/ NMa-
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D.CharlesPayne.ChipfLicenseExaminer
George T. Hopper. License Examiner
Paul M. Steiner. License Examiner
Approved by: 7 w
Thomas A. Peebles. Chief.
Operator Licensing and Human
Performance Branch
Division of Reactor Safety
Enclosure 2
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EXECUTIVE SUMMARY
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Catawba Nuclear Station.' Units 1 &'2 i
NRC Examination Report Nos. 50-413/97-300,-50-414/97-300 i
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, During the period November 17-21, 1997 and December 1-19, 1997. NRC examiners !
conducted an announced inspection and operator licensing initial examination !'
. in accordance with the guidance of Examiner Standards. NUREG-1021. Interim
Revision 8.- This examination implemented the operator licensing requirements i
of 10 CFR S55.41. S55.43, and 355.45.
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Doerations !
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. . Control room activities were observed during the examination validation i
week'and examination administration week. The operators were found to
be attentive and professional in their duties. (Section 01.1)- :
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.- The examiners identified discrepancies with one Emergency Operating i
Procedure and.one normal operating procedure. (Section 03.1) j
P
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Eleven Senior Reactor Operator candidates and three Reactor Operator ,
candidates received written examinations and operating tests. The NRC ;
administered the operating tests during the weeks of December 2-5. 1997.
and December 15-18.'1997; The licensee administered the written i
examination on December 12. 1997. ( Mction 05.1)
. In general. the examiners found that the as-submitted examination was 1
good. -While extensive effort was expended to optimize the examination, i
the initial licensee submittal tested the proper areas of knowledge and :
was set at the appropriate level of difficulty. The. examiners. ;
identified-the need for higher level written examination questions in !
the area of Radiation Monitoring Systems as requiring staff attention or :
improvement. (Section 05.2) j
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. Thirteen of fourteen candidates passed the ~ examination. Four candidates l
. were identified as. exhibiting performance weaknesses during the i
operating test. One candidate was identified as exhibiting performance -l
weaknesses on the written examination. (Section 05.3)
]
. Candidate Pass / Fail
SR0 R0 Total Percent
Pass 10 3 13 93%
Fail 1 0 -1 6%
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Enclosure 2
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The examiners identified one simulator fidelity discrepancy. The
Emergency Diesel Generators on the plant specific simulator were not
modeled to remain tripped on loss of lubricating oil as plant design
requirements specify. (Section 05.4)
.
One violation was identified regarding the implementation of the
licensee ~s licensed operator requalification program. The facility was
conducting its 10 CFR 55.59 required annual operating and biennial
written requalification examinations three to eleven months after the
end of the 24-month requalification training cycle. (Section 05.5)
.
The examiners identified one unresolved item associated with the Updated
Final Safety Analysis Report. Two examples (Auxiliary Feedwater
condensate storage tank outlet valve. 1(2)CA-6 tagged shut on both
units and emergency operating procedure direction for o]erator defeat of
automatic swapover of Refueling Water Storage Tank to t1e containment
sump during selected emergency conditions) were identified where Updated
Final Safety Analysis Report wording was not consistent with the current
plant operating practices and procedures. Also at issue was whether a
proper evaluation per 10 CFR 50.69 was conducted in each case. (Section
08.1)
Enoineerina
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The examiners identified one unresolved item associated with the four
site Emergency Diesel Generators. Given the error identified in the ,
simulator diesel lubricating oil low pressure trip logic (Section 05.4). '
an engineering evaluation is needed to confirm that the plant's EDG
lubricating oil " low-low pressure" trip logic was properly built to
design specifications. (Section E7.1)
Plant Suocort
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. On two occasions, the examiners noted poor personnel radiological i
l monitoring practices at the Radiological Control Area exit. (Section l l R4.1) i l . The examiners identified that a weakness existed in the facility's l contamination control and monitoring procedures that could allow workers i
to unknowingly remove potentially cross-contaminated hand carried items !
from the Radiological Control Area after being checked in the Small
Article Monitor. (Section R4.2) i
l . The examiners identified that the Alert and Site Area Emergency action ,
levels for Event Category # 4.1.9. " Natural Disasters and Other l
, Hazards.~ of RP/0/A/5000/01. were not clearly and specifically written
to achieve the results expected by licensee management for properly
classifying a hurricane event. (Section P3.1)
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Enclosure 2
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. The examiners identified one violation associated with the licensee's
failure to conduct periodic training and evaluation of security officers
qualified as Safe Shutdown Facility Operators. When tested one year
after qualification, with no requalification training conducted, four of
seven security officer examined failed their tests. The examiners
concluded that the licensee failed to properly im)lement a systems
approach to training for this program to ensure tlat the selected
security officers could adequately and safely perform the function of
Safe Shutdown Facility Operator if called upon. (Section S5.1)
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Enclosure 2
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Reoort Details i
Summary of Plant Status
During the period of the examinations the Unit I was shut down and Unit 2 was
at 100 percent power.
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1. Ooerations
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01 Conduct of Operations t
01.1 Control Room Observation
During validation and administration of the examination the examiners
observed the conduct of operations by currently licensed operators in i
the control room. The Reactor Operators (R0s) were attentive to the '
evolutions in progress. The Senior Reactor Operators (SR0s) limited
personnel access for official business only, which contributed to a
quiet, professional atmosphere.
03 . Operations Procedures and Documentation
03.1 Review of Ooerations Procedures ;
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a. Scope
The examiners revtewed normal. abnormal. and emergency operating
l procedures during the examination development and administration for
clarity, accuracy, and ease of use. Plant operators and other site
! personnel should be able to perform all procedures correctly without
- error or confusion while successfully accomplishing the intent of the 1
l . subject procedure. ;
b. Observations
l The examiners observed operator license candidate performance utilizing l all or part of over 70 licensee procedures. Two areas for improvement ! were identified. 1
(1) Post-LOCA Cooldown and Depressurization. EP/1/A/5000/ES 1.2,
Revision 6
Step 35.d. pages 33-35 of 47. directed the operators to establish
conditions for placing the Residual Heat Removal (RHR) system in the RHR
i mode. Among the plant conditions required to accomplish this step were ! '
reactor coolant system (RCS) T-Hots less than 350 degrees F. RCS
pressure less than 385 psig. and the RHR pump secured for the train to
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be placed in the RHR mode.. When these conditions were met. then step
35..d.3) directed the operator to place one train of the RHR system in
the RHR mode per OP/1/A/6200/04. " Residual Heat Removal System ~
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Enclosure 2 ,
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However, the initial conditions of this procedure (OP-04) state that two
trains of RHR are operating. Consequently. the transition steps were
confusing as to the required plant status and how to proceed. The
transition between ES-1.2 and OP-04 should be consistent with required
plant conditions for this situation and accurate for proper operator
performance.
(2) Containment Purge System. OP/2/A/6460/15. Enclosure 4.1, Retype
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#7
Steps 2.9.1 and 2.9.2. page 3 of 4. directed the operator to start the
Containment Purge System (VP) supply and exhaust fans in one of two
manners depending upon the system flow rate. The procedure referenced
50% system flow as the critical flow rate for this decision point.
However, the instrumentation in the plant gives system flow in "scfm."
Nowhere in the subject procedure was some percentage of system flow
translated into "scfm." Consequently, operators without this knowledge
could not be assured of performing the correct step. Three license
candidates were tested on this procedure and all were confused as to 1
what to do. In all cases, the candidates assumed the upper gage range :
for system flow rate was equivalent to 100% system flow and continued
with the task on that basis. In this situation, the candidates were !
correct. But such decisions should not be left to chance. The l
procedure should be written in a manner consistent with indications the !
operator sees in the plant.
The training department staff opened Problem Investigation Process (PIP) ;
action items for both the above procedure problems following completion j
of all licensing examinations. !
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05 Operator Training and Qualifications !
05.1 General Comments j
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NRC examiners conducted regular, announced operator licensing initial i
examinations during the period December 1-19, 1997. The examiners '
! administered examinations developed by members of the Catawba training
staff and a contractor, under the requirements of an NRC security l
l agreement, in accordance with the guidelines of the Examiner Standards
(ES). NUREG-1021. Interim Revision 8. Eleven SRO and three R0 license
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applicants received written examinations and operating tests. 05.2 Pre-Examination Activities
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This examination represented the licensee's first attempt at developing
written examinations and operating tests for the NRC's operator
licensing process. In general, the examiners found that the as-
submitted examination was good. Licensee response to examiner changes
Enclosure 2
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and comments wa's prompt'and accommodating. The operating tests were
validated during the weekLof November 17-21, 1997. The written
examinations were finalized and approved the week of December 8. 1997.
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.The facility examination developers submitted 125 multiple choice
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cuestions for NRC examiner review. The R0 and SR0 written tests were
cesigned to share 75. questions with 25 additional questions that were :
license level specific. The enminers had no comments on 67 questions. ;
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Of the remaining 58 questions, the. majority of examiner t. aments l
concerned editorial changes to assure clarity in the question stem and !
to enhance the quality of the incorrect answers.
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The examiners requested that 10 questions be replaced. Six of these i
questions were judged to be too simplistic for an operator license test. !
Most~ of these were questions concerning radiation monitoring equipment. :
~The remaining four questions were replaced because they-tested at the l
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wrong level or were too difficult to fix. The examiners determined- ;
another 15 questions to be good test items. but each required
significant effort before receiving final approval. i
The examiners-noted that development of additional comprehension and I
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analysis level questions for various radiation monitoring systems would I
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enhance the quality of future examination submissions. Also, the
! examiners identified six questions with technical inaccuracies during L their review des)ite several levels of licensee review. The NRC
expectation is tlat the as-submitted draft examination will not contain
any technical inaccuracies.
-The facility examination developers submitted three simulator scenarios
and one saare for NRC review. The examiners considered the simulator
tests to )e good. The. examiners modified.one scenario to add an
Emergency _ Operating Procedure (EOP) transition to a Functional Recovery
Procedure (FRP). The examiners also re-ordered the sequence of scenario
l administration to better fit the candidate crew mix. Other minor
changes were made to enhance the test items of each scenario. !
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The walkthrough examinations sets submitted by the facility examination
developers contained Job Performance Measures (JPMs) that met the
l' guidelines of the ES and were of the appropriate level of difficulty.
Minor JPM content additions were made to improve their usability by the
'
examiners. The JPM follow-up questions were good and above average in
quality.
05.3 Examination Results and Related Findinas. Observations. and Conclusions
L Thirteen of fourteen candidates )assed the examination. One SR0
- candidate-failed the plant walktarough portion of the operating test. !
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The examiners determined that four other candidates exhibited i
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Enclosure 2
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performance deficiencies on the walkthrough portion of the operating
test. Detailed candidate performance comments have been transmitted
. under. separate cover for management review and to allow appropriate
candidate remediation. !
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L _In a letter dated December 18. 1997, the licensee submitted four post- !
examination comments on the written test for NRC consideration. The !
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examiners
! question #78.questioned
In a letterthe accuracy
dated Januaryof
13.one comment
1998, for R0
the licensee examination
amended
L its' comment for this question. The licensee acknowledged that the
correct answer to R0 question #78 was as originally stated on the answer-
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key. - However, the licensee then contended that the knowledge being :
.. tested by this _ question was beyond the sco3e required for licensed ;
, operators. The examiners dissented with t1e revised comment concerning ! l '
R0 question #78 and accepted the remaining three comments. The ;
licensee's post-examination comments are provided in Enclosure 5 and i
NRC's resolution of these comments is provided in Enclosure 6. Also. i
the NRC identified one other. question (R0 #55. SR0 #42) with two correct ;
answers. The answer keys were modified to reflect these changes. l
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Post-examination grading item analysis of the written examination !
identified six questions where most SRO candidates exhibited knowledge !
deficiencies. A similar analysis of the R0 examination identified l
l. eleven questions where most RO candidates exhibited knowledge .j
deficiencies. .The examiners concluded that no generic knowledge i
weaknesses existed where multiple questions on the same system or topic ;
were missed by a large number of candidates.
The examiners did not note any significant candidate 3erformance I
weaknesses during the simulator examinations except tlat periodic crew j
briefings on the status of plant conditions and progress in mitigating i
the< casualty by the SRO were rare. The few briefings that were made did i
not provide useful.information to the crew.
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During the plant walkthrough examinations five of the eleven SR0
candidates missed 20% or more of the JPMs they were administered.
Examiner post-examination review of these results did not identify any
generic weaknesses in candidate performance in this area.
05.4 . Simulator Fidelity j
During performance of dynamic simulator tests, a scenario was conducted
i that resulted in a loss of all AC power with the IB Emergency Diesel l Generator (EDG) out of service and the 1A EDG failing after start on low
- lubricating oil pressure. The examiners observed that. after tripping.
" the 1A EDG automatically restarted and reloaded. Since the EDG still ,
had: low . lubricating. oil pressure, the EDG tripped again. This cycle of
starting. loading and tripping proceeded for a total of five times until
- 'the_model simulated'using up all the starting air in the receivers.
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Enclosure 2
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L The lubricating oil " low-low pressure" trip was disarmed on initial EDG
start until after a 60 second time delay. Once a loss of lubricating
oil trip was incurred even with a start diesel signal still present.
the EDG should remain tripped and not restart again. Investigation by
the simulator staff determined that the EDG trip circuit software did
not accurately reflect the actual trip circuit design. This item was
added to the simulator deficiency list and will be worked in the future.
The examiners did not note any other simulator fidelity discrepancies.
05.5 Licensed Doerator Recualification Procram
a. Scope
In August 1997 the licensee's Operations Training Supervisor contacted
the Region II staff when he became aware of a potential problem
regarding when the comprehensive biennial written examination and the
annual operating tests were being conducted at Catawba. The examiners
analyzed licensee records during the site visit to determine the facts
of this issue.
b. Observations
While conducting the license examination preparation visit on
November 17-21. 1997, the licensee provided records documenting the
conduct of 10 CFR 55.59 required examinations for the licensed operator
requalification (LOR) program from 1991 through 1997. These records
also included a schedule for the proposed examinations for 1998 and
1999. The examiners noted that Catawba Operator Training Management
Procedure (OTMP) 3.0. " Design and Development". Revision 7. section
21.2. defined the 10 CFR 55.59 required 24-month LOR training cycle as
beginning in January and ending the following year in December. The
licensee stated that this cycle begins again every odd year (i.e. 1991.
1993, etc.). The examiners also noted that section 21.10 of OTMP 3.0
.-stated (1) the " operating exam. shall be administered annually" and
(2) the ~ comprehensive written exam shall be administered-at the
completion of the 24 month cycle and no later that April 30 of the year
following the cycle.~
Based on the records provided, the examiners determined that the annual
operating tests were routinely conducted three to eleven months after
the conduct of training. Specifically, for the years 1991 to 1993 the
operating '.ests were conducted three to five months after the com)letion
of the applicable year's training. For the years 1995 to 1997, tie
subsequent year's training was nearly completed before the previous
year's testing had been performed. In 1994, the NRC granted the
licensee a one time, five month extension for conducting the annual
operating tests due to a plant outage. The NRC was not aware at that
time that the regularly scheduled test dates were three to five months
after training had been completed. Also, the licensee stated that a
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change in supervisors at this time resulted in their misunderstanding I
that this was a one. time only extension. This oversight contributed to *
lwhy all-examinations in years 1995 to present were. ten to eleven months !
after the training year vice only three to five months after, i
l
Similarly. the examiners identified that the comprehensive biennial
.
l
. written examinations were routinely conducted three to eleven months i
following the completion of the applicable LOR training cycle between l
1991 and 1997 (i.e., three to eleven months into the subsequent LOR >
cycle). l
The examiners also noted that the Catawba LOR program was accredited by +
the Institute of Nuclear Power Operations (INPO). Per OTMP 3.0. the LOR ;
. program was based. in part upon the systems approach to training (SAT) 1
process. !
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c. Conclusions '!
The SAT process is defined in 10 CFR 55.4. Element (4) of the SAT
process requires ' evaluation of trainee mastery of the objectives during .
training" and element (5)' of this process requires " evaluation and !
revision of the training based on the performance of trained personnel ;
.in the job setting". Also 10 CFR 55.59(c)(4) requires that the !
l facility LOR program include " comprehensive recualification written i
examinstions and annual operating tests which cetermine areas in which -
retraining is needed to upgrade licensed operator and senior operator
i- knowledge." NRC has interpreted the above to mean that the 10 CFR
- 55.59(c)(4) testing be conducted within the 24-month training cycle'in a-
manner that effectively meets the intent of elements (4) and (5) of the
L SAT process. , ,
The examiners concluded that the Catawba LOR program conducted the
, required annual operating tests and biennial comprehensive written
examinations'. However. the annual operating-tests were not effectively-
.
scheduled to-allow the licensee to timely determine areas where operator
! retraining was needed and. if needed, to feed back necessary program !
improvements or to redirect course content for the next year. As noted
above. from 1990 to 1997 (four years) the subsequent year of training
. was nearly completed without having performed elements (4) and (5) of
1 the SAT process for the previous year. Similarly, the written
examinations were conducted outside of the applicable LOR cycle such
that'.'nearly half of the next 24-month cycle of training was conducted
without the benefit and guidance of lessons learned from the previous
cycle.
-The examiners concluded that the failure to conduct the annual operating
,
tests and biennial comprehensive written examinations within the 24-
4
month requalification cycle was a violation of 10 CFR 55.59(c)(4).
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This was identified as VIO 50-413.414/97-300-01: Failure to Conduct . 1
Annual and' Biennial Requalification Examinations within the 24-Month '
Training Cycle. ,
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08~ Miscellaneous Operations Issues !
08.1 Soecial Review of UFSAR Commitments
a. Scope
l
The discovery of ~ a' licensee operating its. facility in a manner contrary - l
to the Updated Final Safety Analysis Report (UFSAR) description ;
highlighted the need for a special focused review that compares plant ;
practices procedures and/or parameters to the UFSAR descriptions. ;
While performing the examinations discussed in this report, the i
examiners reviewed the applicable portions of the UFSAR that related to j
the area that were being tested. - The examiners intent was to verify ,
that the UFSAR wording was consistent with the observed plant practices, !
. procedures and/or parameters. -
~b. Observations ,
,
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:(1) :1(2)CA-6 Valves Tagged Shut
'
The examiners reviewed the UFSAR sections describing the Auxiliary
Feedwater. System (AFW) The examiners noted on the simulator that Valve :'
'1CA-6.AFW suction from the AFW Condenstate Storage Tank-(CACST). was
' tagged shut. Normally this valve is open to provide an immediate source
of condensate grade water to the AFW system upon system initiation. The
examiners determined that the simulator simply reflected a change that
-had been made on both units since May 1997. The examiners reviewed i
licensee records that documented-recently identified concerns with :
possible air entrainment into the AFW supply piping from this tank
causing equipment damage or loss (Operations Technical Memorandum 1
L#97-01). The examiners also noted that Framatome Technologies had been !
contracted to analyze the problem with a report due to the licensee on i
February 28. 1998 (CA-6 Update to PIP 0-C-97-1579). The examiners- *
identified the following discrepancies in the Catawba UFSAR:
Section 7.4.1.1 - Auxiliary Feedwater System Instrumentation and
Control Description. page 7-67. The second paragraph stated that
during normal operation. the AFW~ system condensate grade supply
i valves and pump discharge valves are aligned to provide AFW flow
without re)ositioning on an automatic start. The tagging of ICA-6
defeated t1is expected alignment of condensate grade water
-sources.
Section 9.2.6.3 - Condensate Storage System Safety Evaluatfon.
page 9-56. The first paragraph stated the preferred source of
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cles water supply'for the AFW pumps is provided by the main !
condenser hotwell and upper surge tanks on each unit and the !
shared AFW condensate storage tank (CACST). This paragraph also ;
stated these were not safety-related since the Nuclear Service :
Water.(NSW) system served that function. However, the CACST was ;
not available with the 1(2)CA-6 valves shut. t
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Section 10' 4.9.2 - Auxiliary Feedwater System. System Description.
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pages 10-49. 50. The last paragraph of page 10-49 stated that'all !
the preferred sources of condensate quality water were normally
aligned to the AFW pump suctions and that the reserve for each
unit was maintained among the sources listed in a table on page
i
10-50. The CACST was the first source listed in this table. The ,
first paragraph of page 10-50 also stated that the AFW pumps i
should normally be aligned to condensate quality water to maintain 1
steam generator chemistry. especially for fast recovery events as :
blackout loss of normal feedwater. or main steam system !
malfunction. All necessary means to prevent inadvertent injection ;
<
of out of chemistry nuclear service water to the steam generators (
must be employed. The station has lost up to 42.500 gallons (15% ;
I of the unit's clean water su] ply) of condensate grade water by , l
shutting the CA-6 valves. T1e CACST was shared by both units. i
Additionally. this paragraph stated that a reliable means of ;
detecting loss of condensate source and automatic transfer of the i
pump suctions to the NSW source was employed. This system was l
designed to handle any of four postulated failures of the
condensate. supplies including: depletion of all condensate sources
and partial or complete loss of source due to air leakage into the
system from ; pipe _ crack, or failure to isolate a depleted source. ]
The licensee's basis for shutting 1(2)CA-6 was concern about the
reliability or effectiveness of this automatic feature, Therefore
this aspect of the issue was also not in accordance with the
UFSAR.
, (2) Automatic Swapover of Emergency Core Cooling System (ECCS) Suction '
Sources to Containment Sump on Refueling Water Storage Tank (RWST)
Low low Level
The examiners reviewed the UFSAR sections describing the Refueling Water
System. The examiners noted during simulator scenario validation that
EP/1/A/5000/ECA-11. " Loss of Emergency Coolant Recirculation". step 4
directed manual operator action to defeat the automatic swap of ECCS
pump suctions from the RWST to the containment sump on RWST low-low
level when an intersystem LOCA (ISLOCA) has occurred. Since the LOCA
would be into another plant system rather into containment, no water
would be available in the emergency sump (i.e.. sump level less than 3.5
feet) to supply these pumps. Manual operator actions which defeat
automatic. emergency design features may be Unreviewed Safety Questions
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(US0) which require NRC review and approval before imalementation. The !
' examiners identified the following discrepancies in t1e Catawba UFSAR: i
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R 'Section 6 3.2.8 -~ Emergency Core Cooling System Manual Actions. l
, pages 6-107.108. The first paragraph stated the only manual l
!
action required by the operator for proper ECCS operation during i
the injection mode was to isolate the Coolant Charging Pump (CCP)
minimum flow bypass line at 1500 psig and to unisolate it again if 1
.
pressure rises above 2000 psig. The second paragraph stated that !
l the. changeover from injection mode to recirculation mode was ;
-initiated automatically and completed manually. No mention was !
made to the quite different system response to and operator !
. ' actions for an ISLOCA. ;
;
Table 6-93 - Sequence of Changeover 0)eration From Injection to '
Recirculation. Appendix 6. The fourtl full paragraah on page 6- ;
.108 referred the reader to Table 6-93 for the switclover sequence, a
This table listed the swapover from the RWST to the containment l
sump as the only two automatic actions in this sequence. All the ;
- rest were manual operator actions. None of the 15 manual operator
l
actions recuired defeating the atomatic swapover by depressing the
train A anc train B " DEFEAT" pushbuttons (C-LEG RECIRC FWST TO
L CONT SUMP SWAP). Also, there was no separate table in Appendix 6
that listed the automatic (if any) and manual operator actions in :
- response to an ISLOCA.
l' Section 7.6.5.1 - Refueling Water System Instrumentation and ! l Switchover From Safety Injection to Recirculation Mode i !
Description. pages 7-94.95. The third paragraph stated that a
- low-low RWST level coincident with a safety injection signal .
I
automatically initiated realigment of the ECCS to the
recirculation mode after a LOCA by opening containment sump .
u isolation valves and isolating the RWST. Further this paragraph l stated that, in the Solid State Protection System (SSPS) output !
cabinet, the slave relay which. picked up and initiated the Safety !
l Injection (SI) signal.. remained latched until manually reset by o the sump valve automatic open circuit reset switch. Therefore. l resetting the SI signal would not inadvertently prevent the ! automatic swapover function but still allow manual o)erator ,
control of the ECCS equipment when needed. The UFSAR also stated
l
that the. function of the reset switch was to permit the operator
- to unlatch the slave relay in the event the corresponding sump
valve must be closed and retained in a closed position following
L '
the LOCA such as for maintenance. No mention was made to
ISLOCAs or to use of the reset switch as part of the operator
actions to override this automatic safety function during
conditions where sump level was inadequate.
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Section 9.2.7.2 - Refueling Water System. System Descriptfon. page
9-57. Paragraph 5.a provided a component description of the RWST
and stated that it was sized to allow sufficient time for the
operators to properly assess the situation and establish the
recirculation mode following a LOCA. No mention was made to the
effect on the recirculation mode if an intersystem LOCA occurred.
Chapter 15 - Accident Analyses. Generally. the consequences of an ,
ISLOCA should be bounded by the large break LOCA (LBLOCA) accident
analysis. However, no mention was made in this chapter to show
that an ISLOCA was considered in the LBLOCA analyses nor was this
accident's effect and impact on plant response discussed.
Licensee review of this chapter in these regards may be
appropriate.
c. Conclusion
The examiners concluded that the UFSAR wording in the sections discussed I'
above was not consistent with the current plant operating practices and
procedures. These discrepancies are collectively identified as examples
of URI 50-413.414/97-300-02. Catawba UFSAR Discrepancies, which will be
combined with other items the resident inspectors are reviewing.
08.2 (Closed) IFI 50-413.414/96-018-01: Verification of Corrective Actions i
for Documentation of Training and Qualification. 1
This issue involved the identification of deficiencies in the Employee !
Training and Qualification System (ETOS) by the Operations Department.
A PIP report was issued to document and track this problem. At the time
the inspector follow-up item (IFI) was generated, licensee corrective
action lad not yet been completed.
The inspectors reviewed the licensee *s corrective actions to the
problems encountered in implementing the ETOS system for new equipment i
and associated tasks. The inspectors found the licensee's level of I
attention and resolution of the problems to be adequate. This IFI is
now closed.
08.3 (Discussed) IFI 50413.414/96-300-01: FSAR vs. ARP on D/G Loss of Cooling '
Time to Failure
This issue involved the amount time it would take for the EDGs to fail 1
I
if EDG cooling water were not available immediately. The UFSAR stated a
very short time frame to failure (1-2 minutes) while the loss of EDG
cooling water annunciator pr cedure stated a relatively longer time ;
frame to failure (10-15 minutes). This information defined how critical 1
it would be for the operators to identify such a failure quickly.
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The examiners reviewed special engineering calculations intended to i
demonstrate that the EDGs have a significant period of time, even fully [
loaded before a loss of EDG cooling water would have an effect. The ;
examiners could make 'no conclusions as to the adecuacy or accuracy of. i
these calculations. Also, the licensee had not acdressed the issue of !
the EDG cooling water low flow alarm being cut out until the EDG cooling *
water inlet valve-is fully open. This IFI is still open pending
additional engineering review into the issue.
III. En.aineerina i
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E7 Quality Assurance in Engineering Activities
E7.1 Emeraency Diesel Generator Trio Circuitry i
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As discussed in Section 05.4 above, the' examiners identified a problem !
with the loss of lubricating oil EDG trip logic on the simulator. The
UFSAR describes a lubricating oil low-low pressure trip that will
shutdown and protect the EDG for all EDG operating modes. The examiners !
raised a concern that the trip logic installed on the plant's EDGs may l
not meet the intent of the UFSAR. Should the lubricating oil low-low '
pressure trip on the site's EDGs function as it'did in the simulator,
not'only would the EDG be damaged by repeated starts without lubricating
oil. but all safety-related loads that are started by the sequencer l
would exceed their allowable start cycles. The examiners asked the
. licensee to review the as-built design and installation of this
circuitry on the 1A.1B. 2A and 2B EDGs in the plant. As of the exit .
meeting. the licensee indicated that actual alant installation of this
circuitry was in accordance with the design ) asis but that investigation i
was continuing. The issue will be tracked as IFI 50-413.414/97-300-03, i
~ Verification of Units'1.and 2 Emergency Diesel Generator Low-Low '
Lubricating 011 Trip Circuitry with Design.
IV. Plant Succort
l R4 Staff Knowledge and Performance in RP&C
R4.1 Poor Personnel'Radioloaical Monitorina Practices at the Radioloaical
Control Area (RCA) Exit
'a . Scope
During.the course of the examination, the examiners observed members of
the plant staff. including Radiation Protection (RP) technicians
: perform whole body frisking of persons who had received contamination
alarms after using the automated personnel counters (PCM-1).
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b. Observations
On two occasions, the examiners noted that frisking was performed by
placing the instrument probe directly in contact with the affected ,
person's skin or clothing. This method of frisking was contrary to the ,
site management's RP expectations and also risked transferring !
contamination to the instrument probe had the person being frisked
actually been contaminated.
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c. Conclusions-
The examiners concluded that a weakness existed in the facility's
contamination control and monitoring procedures that could allow
contamination of the personnel radiation monitoring equipment by RP
technicians as well as by less knowledgeable plant personnel. ;
R4.2 Poor Contamination Control Practices Durino Staff Use of the Small
Article Monitor at the RCA Exit
a. S.c.op_g
During the course of the examination, the examiners observed members of *
'
the plant staff during their use of the Small Article Monitor (SAM).
- The SAM is used to monitor small hand carried items (e.g.. books,
papers, tools, etc.), that had been used in the RCA and thus were -
l potentially contaminated, prior to their removal from the RCA. There ,
were three SAMs at the RCA exit of which one was out of service. .
l Typically a worker exiting the RCA.would place his small items in the l counter, begin the count sequence and then enter a PCM-1 to check his , l body and clothing for contamination. The SAM count cycle would take i
about 15 seconds to complete whereas the body counter would take one or
l two minutes to complete. Consequently, several minutes would pass after
the SAM finished its count cycle before the owner could reclaim his
items from inside the monitor.
,
l b. Observations
t
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The examiners observed that every time other workers (who had not yet
monitored themselves for contamination) walked up to use the SAM for
their own items, they would remove the just counted items and replace
them with their own. Consequently, the just counted items were
potentially cross-contaminated by these newly arrived workers. Once the
l owner of the just counted items finally cleared the body counter, the l person would retrieve the now potentially contaminated items and exit ;
- the RCA. On two occasions. the examiners noted that the worker who had
removed another's items from the SAM. subsequently set off the body
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counter alarm. Since the owner did not know who had removed his items ;
from the SAM. he or she would not be aware that the items could now .
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actually be. contaminated and exited the RCA. In both cases, the second
worker turned out to be uncontaminated after being frisked by an RP l;
technician. i
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c. . Conclusions
'
The~ examiners' concluded that a weakness existed in the facility's !
contamination control and monitoring procedures that'could allow workers !
to unknowingly remove potentially cross-contaminated hand carried items i'
from the RCA after being checked in the SAM.
P3 EP Procedures and Documentation ,
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P3.1 Emeroency Action Level (EAL) Guideline for Hurricanes
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a. Scope I
i
Several SR0 license candidates were tested on their ability to classify l
an event per the EAL guidance of RP/0/A/5000/01 for a set of conditions :
related to an approaching hurricane. j
b. Observations
f
The candidates were presented plant conditions of an approaching
hurricane with National Weather Service (NWS) reported winds of 105 mph. l
The hurricane was about 150 miles from the site and was traveling at
about 25 mph over land. The plant was in the aredicted path of the l
.
Storm. but it was still about six hours from tie site. Wind speeds were i
not-expected to diminish much before. reaching the site and it was i
anticisated it would take about 60 minutes.for the hurricane to pass !
over tie site. Based on the above conditions the candidates were asked !
what. ._if any, notifications would be made? i
The examiners-expected the candidates to determine the criteria to
declare a Site Area Emergency (SAE) was met (notify State and county -
officials within 15 minutes and the NRC within I hour). However. !
licensee management expectation was the declaration should be based on ;
onsite damaae as a result of the approaching hurricane. Since there was
no report of damage yet. no declaration of any kind should be made (not
u even a Notice of Unusual Event [NOUE]). I
;
l One of the three candidates tested with this question declared an SAE. ' <
He stated he would notify State and county officials within 15 minutes
! -and the NRC within I hour. The examiner investigated the candidate's
basis for this determination, and found that the NOUE EAL for Event :
Category.# 4.1.9. " Natural Disasters and Other Hazards." required the ;
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severe weather (i.e.. hurricane) to be ~onsite." In contrast, the EALs
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for Alert and SAE did not specify that the severe weather be onsite
before declaration. Licensee emergency preparedness personnel later
confirmed the expectation for the Emergency Coordinator to declare an
Alert or an SAE only if the EAL conditions are experienced on the site.
c. Conclusions
The examiners concluded that the Alert and SAE EALs for Event Category
# 4.1.9. " Natural Disasters and Other Hazards." of RP/0/A/5000/01. were
not clearly and specifically written to achieve the results expected by
licensee management for properly classifying a hurricane event. '
55 Security and Safeguards Staff Training and Qualification l
55.1 Security Guard Trainina and Oualifications for Ooeration of the Safe !
Shutdown Facility (SSF) Diesel Generator
a. Scop.e
During validation of plant JPMs at the SSF. the examiners were told that !
selected officers from the site Security Force had been trained and 3
tested in December 1995 to perform the emergency duties and l
responsibilities of the non-licensed operator (NLO) for operating the l
SSF diesel generator. The examiners reviewed the training and testing
these officers received against the requirements of 10 CFR 50.120 for i
'
NL0s.
b. Observations
The examiners reviewed records and documents provided by the training
! department about this specialized part of the NLO job functions. i Training attendance records identified that 16 Security Officers
participated in the initial training program the week of December 9.
1996. This program was developed per the guidelines of OTMP 3.0 and the
licensee stated their intention had been to follow the SAT process in
,
implementing this program. The examiners inspected the materials used
l in this training. These materials were essentially NLO lesson plans and l JPMs that provided the technical and administrative knowledge, skills l and abilities (KSAs) a successful officer would need to adequately
perform the SSF operator emergency functions.
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The initial outcome of this training program was good. The officers I
were required to ) ass a written and JPM walkthrough examination. The
licensee stated t1e intention was to re-evaluate the officer's KSAs in .
December 1997 and then redirect the program, if needed. On December 10. !
1997, a requalification examination was administered to seven officers
qualified as SSF Operator. Only three of seven officers (43%) passed
the test. Based on these results the licensee elected to conduct
refresher training for all SSF Operator qualified officers prior to any
further testing. On December 17. 1997. the day after refresher training
was conducted all thirteen SSF Operator qualified security officers
were administered and passed a written examination. The JPM walkthrough
tests were scheduled to be conducted on December 19, 1997.
Enclosure 2
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Based on the above results'..the licensee determined that some periodic
refresher training would be appropriate and plans were made~to determine
the content.and frequency of SSF Operator. retraining in early 1998. The-
examiners noted that.'the licensee made a poor decision in extrapolating
future ~ security officer performance based on past NLO experience. The
. examiners also noted that NL0s benefit from operating plant equipment on ,
a daily basis as.well as receiving periodic requalification training ;
that generically hones administrative and operating skills. The i
examiners determined that the licensee placed too much reliance on the ;
site security officers retaining critical SSF O j
:alone'without some periodic refresher training.perator Given the.KSAa by memory
uniqueness of ;
this program and the licensee's lack of truly comparable ex>erience, the !
examiners believe more conservative decision-making should lave been l
used'to establish the need for and scope of periodic security officer ;
retraining on SSF Operator knowledges. skills. and abilities, !
c. Conclusions -
;
.The examiners identified that selected officers from the site security i
force had been trained to perform the non-licensed operator emergency i
tasks for starting and operating the SSF diesel generator and determined 1
:that these security officers had not been retrained, nor had their ,
performance been evaluated. since initial training and testing in
December,1996. 'When tested in December,1997, four of seven security
force SSF Operators failed all or part of their written and walkthrough-
examinations. The examiners concluded that the licensee failed to
properly implement a systems approach to training for this program to
ensure that the selected security officers could adequately and safely-
perform the function of SSF Operator if called upon. Consequently. a
violation of 10 CFR 50.120 occurred. This item is identified as VIO 50-
413.414/97-300-04. Failure to Conduct Periodic Training and Evaluation
of Security Officers. Qualified as SSF Operators.
V. Manaaement Meetinos
X1 Exit Meeting Summary
At the conclusion of the site visit, the examiners met with
representatives of the plant staff listed on the following page'to
discuss the results of the examinations. The licensee's management
representatives dissented with the examiners that tne PRI and one IFI
identified were regulatory issues. The representatives stated that
Nuclear Service Water System was the safety-related water supply for AFW
and thus temporary isolation of the CACST was not an issue. They also
stated that the Catawba E0Ps followed the Westinghouse Emergency
- Response Guidelines (ERGS) and therefore manual operator action to
i defeat'the automatic swapover from the RWST to the containment sum) was ! not an issue; Finally. the licensee representatives stated that t1ey l believed the problem identified with the EDG low-low lubricating oil '
,
trip logic was isolated to the plant reference simulator and at the
,
time of the exit meeting, did not believe the plant EDGs were affected. ;
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The representatives also stated that regardless of a possible design or
construction error, the plant was designed to handle a single component
failure and again there should be no regulatory issue. The examiners
acknowledged the above dissenting comments and stated they would be
brought to Region Il management attention during review of the
associated issues. No proprietary information was identified.
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PARTIAL LIST OF PERSONS CONTACTED
Licensee-
T. Beadle. Nuclear Instructor
D. Bradley Shift Operations Manager
S. Bradshaw. Operations Training Manager
M. Glover. Operations Superintendent
R. Jones. Station Manager i
S. Miller, Site Training Manager
K. Nicholson. Compliance Specialist
G. Peterson. Site Vice President ;
M. Purser. Senior Engineer
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J. Suptela. Operations Training
NRC
D. Roberts. Senior Resident Inspector :
R. Franovich. Resident Inspector
ITEMS OPENED CLOSED. AND DISCUSSED
Ooened
50-413.414/97-300-01 NOV Failure to Conduct Annual and Biennial
Requalification Examinations within the
24-Month Training Cycle (Section 05.5) !
50-413.414/97-300-02 URI Catawba UFSAR Discrepancies (Section 08.1)
50-413.414/97-300-03 IFI Verification of Units 1 and 2 Emergency
Diesel Generator Lubricating Oil Low-Low
Pressure Trip Circuitry with Design ,
(Section E7.1) l
50-413.414/97-300-04 NOV Failure to Conduct Periodic Training and ,
Evaluation of Security Officers Qualified i
as SSF Operators (Section 55.1)
Closed
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50-413.414/96-018-01 IFI Verification of Corrective Actions for
Documentation of Training and
Qualification (section 08.2)
Discussed
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50-413.414/96-300-01 IFI FSAR vs. ARP On D/G Loss of Cooling Time
to Failure (Section 08.3) i
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Enclosure 2
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LIST OF ACRONYMS USED
CA Auxiliary Feedwater
CACST -Auxiliary Feedwater Condensate Storage Tank
CCP Coolant Charging Pump
CFR Code of Federal Regulations
EAL Emergency Action Level
ECCS Emergency Core Cooling System
EDG Emergency Diesel Generator-
E0P Emergency Operating Procedure
EP: Emergency Plan Procedure
ERG Emergency Response Guidelines
ETOS Employee Training and Qualification System
ES Excminer Standards (NUREG-1021)
FRP Functional Recovery Procedure
FWST Fresh Water Storage Tank (a.k.a. RWST)
IFI . Inspector Follow-up~ Item
INP0 Institute for Nuclear Power Operations
ISLOCA Intersystem LOCA
JPM - Job Performance Measure
KSA- Knowledge. Skill and Ability
LBLOCA Large Break LOCA
LOCA Loss of Coolant Accident
LOR Licensed 03erator Requalification I
m)h Miles per lour
N .0 - Non-Licensed Operator
'NOUE Notification of Unusual Event i
NOV Notice of Violation '
NRC Nuclear Regulatory Commission
NSW Nuclear Service Water
NWS National Weather Service.
0AC Operator Aid Computer. i
OTMP' Operator Training Management Procedure
PIP Problem Investigation Process
psig Pounds per square inch
RCA Radiological Control Area
RCS~ Reactor Coolant System
RHR Residual Heat Removal
R0 Reactor Operator
RP Radiction Protection
RP&C Radiological-Protection and Chemistry Controls
RWST. Refueling Water Storage Tank
-SAE. Site Area Emergency-
SAM Small Article Monitor
SAT Systems Approach to Training
scfm. Standard cubic feet per minute
S/G Steam Generator
SGTR Steam Generator Tube Rupture
SI- Safety Injection
SRO Senior Reactor Operator
SSF Safe Shutdown Facility
SSPS Solid State Protection System
i Enclosure 2
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STA Shift Technical Advisor
TS Technical Specifications
UFSAR Updated Final Safety Analysis Report
URI Unresolved Item
USQ Unreviewed Safety Question
.VP Containment Purge System
URI ' Unresolved Item
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Enclosure 2
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SIMULATION FACILITY REPORT
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; Facility Licensee: Catawba Nuclear Station !
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Facility Docket Nos.: 50-413 and 50-414 l
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. Operating Tests' Administered on:
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December 2-5. 1997 and December 15-18. 1997
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L This form is to be used only to report observations. These observations do- j
liot. constitute audit or inspection findings and are not, without further ,
verification and review, indicative of noncompliance with 10 CFR 55.45(b). l
These observations do not affect NRC certification or a] proval of the l
l ' simulation facility other than to provide information tlat may be used in
future evaluations. No licensee action is required in response to these
ooservations.
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, While conducting the simulator portion of the opt sting tests. the following ! ! -items were observed (if none, so state). ; ,
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JEM DESCRIPTION i
EDG Lo-Lo Lube A scenario was conducted that resulted in a loss of all
011 Trip AC power with the IB EDG out of service and the 1A EDG' l
Circuitry 1failing after start on low lubricating oil pressure. Once a
i loss of lubricating oil trip was incurred. even with a start i diesel signal still present, the EDG should have. remained
tripped and not restart again. The examiners observed that,
after tripping. the 1A EDG automatically restarted and 4
reloaded a total of five times. Investigation by the l
simulator staff determined that simulator modeling of-the
i EDG lubricating oil low-low pressure trip circuit did not l properly reflect the actual trip circuit design. This-item
was added to the simulator deficiency list and will be
worked in the future.
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Enclosure 3
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. - - - - . -- -- -- - - - - - - - - - - --. - . - - - . , . - - - - - . . - - - - , - .,..,------,.,------------.------.-m- - - - - - --. ---- - - - , . - - , - -- - - - , . . - - - - - - - - - . - - , - - - - - , , , . - - - - - - . - - - -
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FACILITY POST-EXAMINATION COMMENTS
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( Encir',re 5 s
Duke ibwer Compa:y N] bl (g y p y ggg
Catawba Training Center
4850 ConcordRoad
York. SC29745
) DUKEPOWER
December 18,1997
Mr. Thomas Peebles, Chief
Operations Branch Division of Reactor Safety
U. S. Nuclear Regulatory Commission
61 Forsythe Street, S. W.
Suite 23T85
Atlanta,GA 30323
SUBJECT: Catawba Nuclear Station Senior Operator and Reactor Operator Exams
(12/12/97) Post Examination Comments
The following comments are offered conceming the written examinations
administered at Catawba Nuclear Station on December 12,1997. Our pre-exam
review failed to identify two instances where the examination answer key had the
wrong answer selected. We also failed to identify two questions that have no correct
answer. These questions as well as references to justify our position are attached.
We apologize for any inconvenience this may cause.
Sincerely,
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Gary . Peterson
Site Vice President
Catawba Nuclear Station
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cc: Mr. Charles Payne
Mr. William Miller
Mr. Scott Bradshaw
Attachments: Questions and Justifications
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N*d"108 mcyc8sc ptv
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RO Exam Question 10 (BANK #249)
SRO Exam Question 8
The correct answer is A. He flux level stated in the stem (8x10-11 amps)is less than 10% power (P-10).
The P-10 interlock allows the Intermediate Range instmment output to be blocked to preclude a reactor trip
on % Intermediate Range instruments >25% amp equivalent reactor power. Because the Intermediate
Range instruments can not be blocked in this condition, any Intermediate Range bi-stable in the trip
condition will result in a reactor trip. When control power or instrument power fuses blow, the associated
instrument bi-stables fail to the safe state. In this case, the result is a reactor trip.
Answer B is not correct because there is no action to reduce pewer (insert rods). Also the answer implies
that you are not in the source range, however, the flux level specified in the stem is within th e source range. '
Answer C is not correct because the reactor will trip in this condition. He plant would be in Mode 3 at this
time and the Tech Spec for Intermediate Range instruments does not apply. This distractor is based on
NOT getting a reactor trip AND being less than the P-6 setpoint (stem supports being < P-6).
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Answer D is not correct because the reactor will trip in this condition. He plant would be in Mode 3 at this
time and the Tech Spec for Intermediate Range instruments does not apply. This distractor is based on
NOT getting a reactor trip AND being greater than the P-6 setpoint (stem supports being < P-6).
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References attached: I
* Technical Specification 3.3.1
* Operations Training Lesson Plan OP-CN-IC-ENB (Excore Nuclear Instrumentation)
Recommendation:
Change the correct answer from C to A.
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[ QUEST 10N #10 CATAWBMI NUCLEARSTATION RO EXAM i
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1Pt(s) Unit 2 is conducting a reactor startup in mode 2 with the following conditions: I
. Reactor power is at 8x10'" amps and increasing.
.
Operators are evaluating overlap between intermediate and source range.
.
The instrument power fuses blow for the N-35 intermediate range channel.
Which one of the following actions are required by procedures?
A. Enter E-0 (Reactor Trip or Safety injection) at step 1.
B. Insert rods to lower neutron flux level to a value within the source
range. Within one hour, repair N-35 or shutdown to mode 3.
C. Do not exceed 1.0x10 amps until N-35 has been restored. Verify
that P6 permissive status light is not lit within one hour.
D. Continue startup procedure but do not exceed 10% power until
N-35 has been restored to operable status.
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I i L Ques 249a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97 L ,
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3/4.3 INSTRUMENTATION
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3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION
LIMITING CONDITION FOR OPERATION 1
3.3.1
As a minimum, the Reactor Trip System instrumentation channels and
interlocks
Table 3.3-2. of Table 3.3-1 shall be OPERABLE with RESPON
;
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS
4.3.1.1 !
automatic trip logic shall be demonstrated OPERABLE b >
Reactor
Table Trip System Instrumentation Surveillance Requirements specified in
4.3-1. t
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4.3.1.2
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shall be demonstrated to be within its limit at least once pe
Each
s test per
once shall include
36 months andatoneleast one
channel per train
function such thatallboth
such that trains
channels are are teste
;
;
tested at least once every N times 18 months where N is the total number of I
redundant channels in a specific reactor trip function as shown in the " Total
No. of Channels" column of Table 3.3-1. The response time of RTDs associated
with the Reactor
least once Trip System shall be demonstrated to be within their limits at
per 18 months.
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CATAWBA - UNIT 1 3/4 3-1 Amendment No. 148
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TABLE 3.3-1
REACTOR TRIP SYSTEM INSTRUMENTATION
MINIMUM
TOTAL NO. CHANNELS CHANNELS APPLICABLE
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABIE MODES ACTION
1. Manual Reactor Trip 2 1 2 1, 2 1
2 1 2 3*, 4*, 5* 10
2. Power Range, Neutron Flux
a. High Setpoint 4 2 3 1, 2 2
b. Low Setpoint 4 2 3 1###, 2 2
3. Power Range, Neutron Flux 4 2 3 1, 2 2
High Positive Rate
+4. Intermediate Range, Neutron Flux 2 (1' 2 (1#ff, 28 f3'
5. Source Range, Neutron Flux
a. Startup 2 1 2 2ff 4
b. Shutdown 2 1 2 3*, 4*, 5* 10
6. Overtemperature AT
Four Loop Operation 4 2 3 1, 2 6
7. Overpower AT
Four Loop Operation 4 2 3 1, 2 6
8. Pressurizer Pressure-Low 4 2 3 1 6
CATAWBA - UNIT 1 3/4 3-2 Amendment No. 148
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TABLE 3.3-1 (Continued)
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TABLE NOTATIONS
*0nly if the Reactor Trip System breakers happen to be in the closed I
position and the Control Rod Drive System is capable of rod withdrawal. '
ffBelow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. j
fiffBelow the1P-10-(Low Setpoint Power-Range Neutron Flux Interlock)-Setpoint? I
fiffAbove the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint. l
ACTION STATEMENTS l
ACTION 1 - With the number of OPERABLE channels one less than the Minimum
Channels OPERABLE requirement, restore the inoperable channel ,
to OPERABLE status within 48 hours or be in at least HOT j
STANDBY within the next 6 hours. '
ACTION 2 - With the number of OPEPABLE channels one less than the Total- l
Number of Channels, STARTUP and/or POWER OPERATION may proceed !
provided the following conditions are satisfied: l
a. The inoperable channel is placed in the tripped condition
within 6 heu.s,
b. The Minimum Channels OPERABLE requirement is met; however,
the inoperable channel may be bypassed for up to 4 hours
for surveillance testing of other channels per
Specification 4.3.1.1, and
c. Either, THERMAL POWER is restricted to less than or equal I
to 75% of RATED THERMAL POWER and the Power Range Neutron '
Flux trip setpoint is reduced to less than or equal to 85%
of RATED THERMAL POWER within 4 hours; or, the QUADRANT ,
POWER TILT RATIO is monitored at least once per 12 hours o!
per Specification 4.2.4.2. !
ACTION 3 - With the number of channels OPERABLE one less than the Minimum l
Channels OPERABLE requirement and with the THERMAL POWER level:
a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE L,
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status prior to increasing THERMAL POWER above the P-6
Setpoint; or
b. Above the P-6'(Intermediate Rarige Neutron Flux Interlock) !
Setpoint but below 10% of RATED THERMAL POWER, restore the
inoperable channel to OPERABLE status prior to increasing
TilERMAL POWER above 10% of RATED THERMAL POWER.
ACTION 4 - With the number of OPERABLE channels one less than the Minimum l
Channels OPERABLE requirement, suspend all operations involving
positive reactivity changes.
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CATAWBA - UNIT 1 3/4 3-5 Amendment No. 148
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TABLE 3.3-1 (Continued)
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ACTION STATEMENTS (Continued)
, ACTION 5 - Delete ! . ACTION 6 - With the number of OPERABLE channels one less than the Total i Number of Channels, STARTUP and/or POWER OPERATION may proceed
provided the following conditions are satisfied:
I a. The inoperable channel is placed in the tripped condition
within 6 hours, and '
b. The Minimum Channels OPERABLE requirement is met; however,
the inoperable channel may be bypassed for up to 4 hours
for surveillance testing of other channels per
Specification 4.3.1.1.
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ACTION 7 - With the number of OPERABLE channels one less than the minimum
channels OPERABLE requirement, restore the inoperable channel
to OPERABLE status within 6' hours or be in at least HOT STANDBY
within the next 6 hours; however, one channel may be bypassed !4
for up to 4 hours for surveillance testing per Specification ll
4.3.1.1, provided the other channel is OPERABLE.
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ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within
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I hour detennine by observation of the associated permissive
status light (s) that the. interlock is in its required state for
l the existing plant condition, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERABLE cha Nels one less than the Minimum ,
i Channels OPERABLE requirement, oe in at least HOT STANDBY I I
within 6 hours; however, one channel may be bypassed for up to
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2 hours for surveillance testing per Specification 4.3.1.1, ,
provided the other channel is OPERABLE. !
ACTION 10 - With the number of OPERABLE channels one less than the Minimum
Channels OPERABLE requirement, restore the inoperable channel
to OPERABLE status within 48 hours or open the Reactor trip
breakers within the next hour.
ACTION 11 - With the number of OPERABLE channels less than the Total Number j
of Channels, operation may continue provided the inoperable i
channels are placed in the tripped condition within 6 hours.
ACTION 12 - With one of the diverse trip features (Undervoltage or shunt
trip attachment) inoperable, restore it to OPERABLE status
within 48 hours or declare the breaker inoperable and apply l
ACTION 9. The breaker shall not be bypassed while one of the
diverse trip features is inoperable except for the time
required for performing maintenance to restore the breaker to
OPERABLE status. With the breaker b passed, apply ACTION 9.
L
ACTION 13 - With any reactor trip bypass breaker inoperable, restore the
bypass breaker to OPERABLE status prior to placing it in
, service. .
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l CATAWBA - UNIT 1 3/4 3-6 Amendment No. 148
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DUKEFOWER -
CATAWBA ERATIONS TRAINING ;
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- 2. Bistable Relav Drivers
a) P-6 (1/2 lR greater'than 10 * amps) !
b) Low Power Rod Stop )
1) Current equivalent to 20% full power (1/2 channels). !'
2) Rod stop in manual or automatic.
3) Blockable at P-10 (2/4 PR > 10% power)
-c) ReactorTrip
1) aCurrent equivalent to 25% full power (1/2 channels)
L'
m 2) : Blockable at P-10
3. Isolation Amplifier
a) Isolates IR channel from remote equipment.
b) Provides output for following: )
1) SUR Circuit l
(a) Converts rate of change of power level to SUR in DPM.
l
(b) Reads out on C/B.
l 2) C/B Indication - Meter calibrated in amps (10-11 to 10-3),
3) C/B Recorder
(a) 1NR-45 two pen recorder.
l (b) Records IR level in amps when selected. !
l. IR Drawer Panel (LPRO/LPSO #6; PTRQ #4)
1. Amoere Neutron Level Meter
a) Indicates current output of detector l
b) Indicates in amps - Eight decades (10-11 to 10-3 amps)
2. Instrument Power "ON" Lamo - AC instrument power applied to drawer. l
3. Control Power "ON" Lamo - AC control power applied to driver assembly
control circuits.
l 4. Channel On Test Lamo - Indicates OPERATION SELECTOR switch is i in a position other than " NORMAL". i 5. Level Trio Bvoass Lamo - Indicate LEVEL TRIP switch in " BYPASS" 4 '
position.
6. Hiah Level Trio Lamo - ON when neutron flux in IR exceeds current
equivalent to 25% full power. (Approximately 2 x 10-5 amp).
.
OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
Page 16 of 38
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DUKE POWER CATAWBAL.-ERATIONS TRAINING
C. Intermediate Range fails * Low"
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1. Indication indicates " Low" for the failed channel. This can be due to a
blown instrument power fuse (loss of instrument power), a failed
detector or a circuit problem. l
2. Operator actions:
)
a) Enter AP/16, Case lil, intermediate Range Malfunction. !
b) Symptoms for entering this case are: l
Indication lost or erratic "l/R Hi Voltage Failure" annunciator
c) There are no immediate operator actions for this case. The
operator should follow the prescribed procedural steps as given in I
the procedure.
D. Intermediate Range fails "High"
1. Indication indicates "high" for the failed channel. This can be due to a
detector or circuit problem.
2. Opsiator actions:
a) Enter AP/16, Case Ill, Intermediate Range Malfunction,
b) Symptems for entering this case are:
Indication lost or erratic "l/R Hi Voltage Failure" annunciator l
"l/R Compensating Voltage Failure" annunciator
"l/R Hi Flux Level Rod Stop" annunciator l
l c) There are no immediate operator actions for this case. The l operator should follow the prescribed procedural steps as given in
the procedure.
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E. Power Range Channel fails " Low"
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1. Indication indicates lower than actual power level as compared to other
l power range channels. This can be due to any one of the following:
a) Instrument power failure for that channel.
b) Upper or lower detector failing low.
j c) Circuit malfunction.
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OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
Page 25 of 38
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RO Exam Question 76(BANK #316)
The correct answer is A. With conditions as stated in the stem, PORV NC-32B will actuate.
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Answer B is not correct because PORY NC-34A is associated with "B" loop W/R_ pressure which is out of ,
service based on conditions given in the stem of the question.- '
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Answer C is not coTect because the lift setpoint of the PORV NC-32B is 400 psi while the ND Pump _
suction line relief valve will not relieve until 450 psi. !
' Answer D is not correct because the lift setpoint of the PORY NC-32B is 400 psi while the NV letdown line l
relief valve will not relieve until 600 psi. i
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References attached:
l
e Operations Training Lesson Plan OP-CN-PS-ND (Residual Heat Removal System)
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Operations Training Lesson Plan OP-CN-PS-IPE (Pressurizer Pressure Control System) . 1
e Operations Training Lesson Plan OP-CN-PS -NV (Chemical and Volume Control System)
1
Recommendation: i
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Change the correct answer from C to A.
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QUESTION #76
O CATAWBA NUCLEARSTATIONO ;
'
RO EXAM
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1 Pt(s) Unit 1 is in mode 5 preparing to startup after a refueling outage. The fill and r
vent ofthe NC system isin progress. ,
. NC temperature is 120 F
.
Letdown is in service via the ND system through NV-135
. Charging is in service
. The "B" loop wide range pressure instrument is out of service for i
calibration '
. The key switch for PORV-32B is selected to " Low Pressure"
. The key switch for PORV-34A is selected to " Normal"
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Which one of the following statements describes the first NC system response
to an over pressurization event?
A. PORV-32B will open to relieve pressure ;
B. PORV-34A will open to relieve pressure
C. The relief valve in the suction line of the ND pump will open to
relieve pressure
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D. The relief valve in the letdown line will open to relieve pressure I
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Ques 316a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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DUKE POWER CATAWB ERATIONS TRAINING
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7 c) If not " ENABLED" during an S sand the level of the FWST has
decreased to the Auto Swap over setpoint, Manual action is
required and as interlocks for opening NI-185A,184A must be met. ;
d) This auto swap over can be blocked by pressing the " Defeat"
pushbutton (i.e.: LOCA outside CNT, prevents ND pumps from
losing suction source)
3. Interlocks to OPEN N1-185A (184B) (LPRO/LPSO #4; PTRQ #1)
a) ND1B or 2A (36B or 37A) closed
b) FW-27A (55B) closed
E. ND Pump. Suction Reliefs (ND-3,38)
.
1. Provide overpressure protection of the ND System suction piping.
2. While the ND System is open to the NC System, they assist the
Pressurizer Power Operated Relief Valves (PORVs) in providing low
temperature overpressure protection (LTOP).
3.;. Relieve to the PRT at 450 psig -
4. Capacity of each valve is a minimum of 900 gpm. This is in excess of
the combined flowrate of the NV pumps at that pressure.
F. ND Pumps (ISS/NLO/LPRO/LPSO #3)
1. Vertical, centrifugal pumps.
l ! 2. KC cools Mechanical Seals and Motor. (Essential Header.) l (ISS/NLO/LPRO/LPSO #2.2) l 3. Design flow of 3,000 gpm at 161 psi
4. Powered from: A-ETA, B-ETB (LPRO/LPSO #3.1)
5. Mini-flow (ND-25A, 598)
a) May be manually operated from MC-11
b) Automatically operated based on flowrate downstream of ND
pumps. (NDFT 5040,5050)
1) OPEN when dow is < 500 gpm
j 2) CLOSE when flow is > 1000 gpm or on pump shutdown ! c) Protects the pumps from potential low flow induced overheating and '
vibration.
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6. ND pumps are operated from: (LPRO/LPSO #4.2; PTRQ #2)
- a) MC-11
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b) ASP's: A 'A' ND Pump; B 'B' ND Pump
OP-CN-PS-ND FOR TRAINING PURPOSES ONLY REV. 21
Page 12 of 23
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DUKEPOWER CATAWB ERATIONS TRAINING
1 l . L ! l 5. Low Temperature Overpressure
a) Enabling Prerequisites !
! 1) NC temperature < 3000F i
2). Low press mode selected 1
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3) PORV must be in ' Auto' i
b) . Purpose - Protect NC and ND systems while at low temperatures. i
L c); System Operation i l !
With NC temp <300*F and LTOP selected PORV's 34A + 32B will
! .be aligned to B/U N2 from CLA A&B and will operate to maintain
the NC System <400#. The temp, press and N2 B/U inputs are
.
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PORV specific and are as follows: '
.PORV,34A Associated with +
- B Loop W/R press
- D Loop W/R TH -
- A CLA N2 B/U isol opens
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PORV 32B Associated with .
- C Loop W/R press
- C Loop W/R Tc
[ - B CLA N2 B/U isol opens l The LTOP System functions as follows for each PORV l - When directed by Procedure to swap to LTOP (<300'F and <360 # l in NC Sys) operator will swap to "Lo Press" on key switches for i
both PORV's (32B and 34A) .
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- When LTOP selected the associated B/U N2 supply from the CLA
will open
"
-if associated W/R Press '(32B = C Loop W/R Press and 34A'= B '
l Loop W/R Press) >400 # then PORV will actuate * l
- If associated W/R Temp (32B = C Loop W/R Tc and 34A = D Loop
j. T/H) >300 F LTOP is disarmed and the PORV will not actuate.
- ,
! I f.
OP-CN-PS-IPE FOR TRAINING PURPOSES ONLY REV.13
p Page 15 of 21 l
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- DUKE POWER CATAWBA ERATIONS TRAINING
m -
d) First temperature reduction for demineralizer protection.
e) Letdown Temperature
1) Indicates temperature exiting the regenerative Hx
2) Indication and high alarm on MCB
3. Orifice Isolation Stop Valves NV11 A,10A,13A
a) Air operated globe valves, fail closed, operated from MCB or Aux
S.D. panel (NV-11 A & NV-13A only from ASP-A)
b) Cannot be opened unless UD ISOL valves are open. ..
1
c) Auto close signals
1) Low PZR level (17%)
2) Containment ISOL. (ST)
3) Closure of NV1 A or NV2A (cannot open NV10A,11 A or 13A
unless NV1 or 2 are open)
4) All 3 charging pumps tripped
4. UD Orifices (2) Isolation Valves NV-11 & 13
...~ a) Reduce coolant press - 1900 psig at design flow rate
b) One 75 gpm orifice (normally used) (NV-13 Block Valve)
c) One 45 gpm orifice (NV-11 Block Valve).
Used to obtain greater UD flow in conjunction with 75 gpm orifice.
Max UD flow of 120 gpm
5. UD Manual Flow Control Valve (NV849) (NV10 Block Valve)
a) Used to warm up downstream piping
b) Flow rate of 5 to 110 gpm (when NCd pressure is 5 385 psig)
c) Controlled from MCB via manual loader. .
6. (Relief Valve 1NV14
a) Overpressure protection for low press. piping and tube side of UD
heat exchanger
b) - Relief setpoint 600 psig
c) Relieves to Pressurizer Relief Tank (PRT)
d) Capacity -max flow rate through all orifices.
l l !
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OP CN-PS NV FOR TRAINING PURPOSES ONLY REV.17 l
Page 15 of 43
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' Encl re 5 ;
F
!
RO Exam Question 78 (BANK #318) !
;
There is no completely correct answer. .
1
Answer A is the most correct answer because a reactor trip will occur due to bi-stables failing to the safe F
position combined with Source Range instruments NOT being blocked, however, the statement that the l
Source Range instrument N31will fail high is incorrect because the instrument will fail low on a loss of j
power. 'Ihe stem of the question does not specify whether the Source Range instruments output is blocked. i
The stem provides information that I of 2 Intermediate Range instruments is >P-6 (which allows Source
Range instruments to be blocked), however, there is no information in the stem about whether the Source -
!
Ranges are actually blocked. Blocking is accomplished manually. !
!
Answer P is not correct because even though I of 2 Intermediate Range instruments is > P-6, the Source
!
Range instruments are not blocked automatically (blocking is a manual action). i
.!
Answer C is not correct because neither Source Range instrument is greater than the hi flux trip setpoint and I
only 1 of 2 Source Range instruments greater than the hi flux trip setpoint is required to trip the reactor.
Answer D is not correct because even though the output of the Source Range channel fails low, the ;
bi-stables fail to the safe position for the failed channel and the reactor trips on 1 of 2 logic. ;
.i,
.
References attached: j
i
e' Technical Specification 33.1 . :
. Operations Training Lesson Plan OP-CN-IC-ENB (Excore Nuclear Instrumentation) l
Recommendation:
. Delete this question from the examination due to no correct a .swer.
,
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O CATAWBA NUCLEARSTATIONO
'
'
QUESTTON #78 RO EXAM i
7
! s ,
1Pt(s) Unit 1 is conducting a reactor startup. Given the following conditions:
. Iw Channel N35 indicates 8X10 " amps !
. IR channel N36 indicates 1.1X10 amps !
.
SR channel N31 indicates 7.1X10' CPS
. SR channel N32 indicates 6.8X10' CPS
If the N31 SR detector high voltage power supply fails, which one of the
'
following statements describes the plant response?
l
A. A reactor trip signal is generated and a reactor trip occurs
l
because the N31 output fails high. l
B. A reactor trip signal is generated but no reactor trip occurs
because one IR channel is above P-6 and the SR high level trip is
automatically blocked.
C. No reactor trip signal is generated because the required
coincidence for the SR high level flux trip is 2/2 SR channels.
D. No reactor trip signal is generated because the channel N31
output fails low.
!
!
I i l l . 1 4
.
Ques 318a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
I
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3/4.3 INSTRUMENTATION t
\
3/4.3.1
REACTOR TRIP SYSTEM INSTRUMENTATION
LIMITING CONDITION FOR OPERATION
3.3.1 _
As a minimum, the Reactor Trip System instrumentation channels and
interlocks
Table 3.3-2. of Table 3.3-1 shall be OPERABLE with RESPO
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1. ;
'
SURVEILLANCE RE0VIREMENTS
4.3.1.1
automatic trip logic shall be demonstrated OPERABLE
Reactor
Table Trip System Instrumentation Surveillance Requirements specified in
4.3-1. l.
4.3.1.2
shall be demonstrated to be within its limit at least Each
once pe
_,
once per 36 months and one channel per function such !
!
tested at least once every N times 18 months where N is the total number of
redundant channels in a specific reactor trip function as shown in the " Total l
No. of Channels" column of Table 3.3-1. The response time of RTDs associated y
1
with the Reactor
least once Trip System shall be demonstrated to be within their
per 18 months. limits a
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CATAWBA - UNIT 1 3/4 3-1 Amendment No. 148
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TABLE 3.3-1
REACTOR TRIP SYSTEM INSTRUMENTATION
MINIMUM
TOTAL NO. CHANNELS CHANNELS APPLICABLE
FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
1. Manual Reactor Trip 2 1 2 1, 2 1
2 1 2 3*, 4*, 5* 10
2. Power Range, Neutron Flux
a. High Setpoint
b. Low Setpoint
4
4
2
2
3
3
1, 2
1###, 2
2
2
h
3. Power Range, Neutron Flux 4 2 3 1, 2 2
High Positive Rate
4. Intermediate Range,' Neutron Flux 2 r1 2 . :1###v f2*. p:3 4
5. Source Range, Neutron Flux
a. Startup 2 1 2 2## 4
b. Shutdown 2 1 2 3*, 4*, 5* 10
6. Overteeperature AT
Four Loop Operation 4 2 3 1, 2 6
7. Overpower AT
Four Loop Operation 4 2 3 1, 2 6
8. Pressurizer Pressure-Low 4 2 3 1 6
i
CATAWBA - UNIT 1 3/4 3-2 Amendment No. 148
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TABLE 3.3-1 (Continued)
) 1
TABLE NOTATIONS
l *0nly if the Reactor Trip System breakers happen to be in the closed ! j
position and the Control Rod Drive System is capable of rod withdrawal.
##Below the P-6 (Intennediate Range Neutron Flux Interlock) Setpoint. g
.
'
'W#8e16WithetP40F(t'6WSetpointrPowef4RangetNeutron FluxrinterlockT+Sitp^oint/
####Above the P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint. (
ACTION STATEMENTS
ACTION 1 -
With the number of OPERABLE channels one less than the Minimum
Channels OPERABLE requirement, restore the inoperable channel '
to OPERABLE status within 48 hours or be in at least HOT
STANDBY within the next 6 hours.
ACTION 2 - With the number of OPERABLE channels one less than the Total
.
Number of Channels, STARTUP and/or POWER OPERATION may proceed
provided the following conditions are satisfied;
a. The inoperable channel is placed in the tripped condition
within 6 hours,
#
b. The Minimum Charinels OPERABLE requirement is met; however,
the inoperable channel may be bypassed for up to 4 hours
for surveillance testing of other channels per
Specification 4.3.1.1, and 3
c. Either, THERMAL POWER is restricted to less than or equal
to 75% of RATED THERMAL POWER and the Power Range Neutron
Flux trip setpoint is reduced to less than or equal to 85%
of RATED THERMAL POWER within 4 hours; or, the QUADRANT
POWER TILT RATIO is monitored at least once per 12 hours
per Specification 4.2.4.2.
50T10NU72S/ With the number of channele OPERABLE one less than the Minimum
Channels OPERABLE requirement and with the THERMAL POWER level:
f
a. BelontheiP;6' (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE
status prior to increasing THERMAL POWER above the P-6
Setpoint; or
b. Above'.the P-6'(Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the
inoperable channel to OPERABLE status prior to increasing
THERMAL POWER above 10% of RATED THERMAL POWER.
ACTION 4 - With the number of OPERABLE cnannels one less than the Minimum
Channels OPERABLE requirement, suspend all operations involving
positive reactivity changes.
(
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CATAWBA - UNIT 1 3/4 3-5 Amendment No. 148
I
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O V
TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 5 - Delete
ACTION 6 - With the number of OPERABLE channels one less than the Total
Number of Channels, STARTUP and/or POWER OPERATION may proceed
provided the following conditions are satisfied:
a. The inoperable channel is placed in the tripped condition
withi.. 6 hours, and
b. The Minimum Channels OPERABLE requirement is met; however,
the inoperable channel may be bypassed for up to 4 hours
for surveillance testing of other channels per
Specification 4.3.1.1.
ACTION 7 - With the number of OPERABLE channels one less than the minimum
channels OPERABLE requirement, restore the inoperable channel
to OPERABLE status within 6 hours or be in at least HOT STANDBY
within the next 6 hours; however, one channel may be bypassed Q
for up to 4 hours for surveillance testing per Specification
4.3.1.1, provided the other channel is OPERABLE.
[
ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1
1 hour detennine by observation of the associated permissive '
status light (s) that the interlock is in its required state for i
the existing plant condition, or apply Specification 3.0.3.
ACTION 9 - With the number of OPERABLE channels one less than the Minimum-
Channels OPERABLE requirement, be in at least HOT STANDBY h
within 6 hours; however, one channel may be bypassed for up to
'
2 hours for surveillance testing per Specification 4.3.1.1,
provided the other channel is OPERABLE.
Tl
ACTION 10 - With the number of OPERABLE channels one less than the Minimum
Channels OPERABLE requirement, restore the inoperable channel
')
'
i
to OPERABLE status within 48 hours or open the Reactor trip l
breakers within the next hour. 1
4
ACTION 11 - With the number of OPERABLE channels less than the Total Number 8
of Channels, operation may continue provided the inoperable ,
'
channels are placed in the tripped condition within 6 hours. ,
1
ACTION 12 - With one of the diverse trip features (Undervoltage or shunt >
trip attachment) inoperable, restore it to OPERABLE status I
within 48 hours or declare the breaker inoperable and apply
ACTION 9. The breaker shall not be bypassed while one of the i
diverse trip features is inoperable except for the time
,
4 l
required for performing maintenance to restore the breaker to ,
OPERABLE status. With the breaker bypassed, apply ACTION 9.
ACTION 13 - With any reactor trip bypass breaker inoperable, restore the
bypass breaker to OPERABLE status prior to placing it in i
service. ,
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CATAWBA - UNIT 1 3/4 3-6 Amendment No. 148
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DUKEPOWER V CATAWBAu2 RATIONS TRAINING \
5. Hioh Lsvel Reactor Trio
a)' Setpoint- 1 out of 2 SR channels > 105 cp's
b) ; Blockable greater than P-6 (1 out of 2 Intermediate Ranges greater
than 10-10 Amps.) Blocked by tuming 2 switches on C/B to " Block".
By taking both switches to " Block" the high level reactor trip for both
channels of SR are blocked and power is automatically removed
from both SR detectors causing all SR indication to decrease to
zero. If only one switch is taken to " Block", only that channel's high
level reactor trip is blocked and both detector remain energized.
D. Source Range Drawer Panel (LPRO/LPSO #6; PTRQ #4)
1. Detector Volts Meter - Indication of high voltage supplied to proportional
counter. i
2. Neutron Level Meter- Scale 100 to 106CPS. l
3. Instrument Power "0N" Lamo - 118 volts AC Inst. power applied to ;
drawer.
4. Control Power"ON" Lamo - 118 volts AC control power applied to {
drawer. I
5. Channel "On Test" Lamo - indicates the operation selector switch is in a l
position other than " NORMAL". i
6. Loss of Detector Volt Lamp - indicates high voltage to detector off g
low.
7. Level Trio Lamo - indicates neutron level greater than trip setpoint in
Source Range.(105 CPS)
8. Level Trio Bvoass Lamo - On when Level Trip switch in " Bypass" for test :
or calibration. l
9. Hiah Flux at Shutdown Lamo - Neutron level greater than 1/2 decade
above normal shutdown level.
10. Bistable Trio Spare Lamo - No function !
11. Instrument Power Fuses - Overcurrent protection for power supply 4
circuits. Instrument power supplies the meters, circuit processing
components, high voltage supply and detector power. This is true for
the IR and PR drawers / circuits also.
12. Control Power Fuses - Overcurrent protection for control signal circuit ,
transformers. Control power supplies the lights on the drawer and 118 l
VAC to the bistable relay drivers to the plant relays. (High flux at
shutdown alaim and SR high level trip). This is true for the IR and PR
drawers / circuits a!so.
i l OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
Pag? 12 of 38
DUKE POWER
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CATAWBAGERATIONS TRAINING
13. Level Trio Switch - Two position switch. '
a) Normal- Switch inactive
b) Bypass - Enables operation selector switch for test and calibration. l
Provides AC signal to prevent Rx trip signal during testing.
14. Operation Selector Switch
a) Eight Position Switch i
b) Enabled by Level Trip Switch to ' Bypass'
c) Normal Position
d) Six Test Positions
e) Level Adjust - Channel On Test lamp lites & Level Adjust
Potentiometer switched into test circuitry,
f) Actuates "N/I System Channel on Test" annunciator when taken out
of ' Normal' position.
l
15. Level Adiust Potentiometer - Adjustable test signal into level amp. - ;
Enables adjustment of the trip level of various bistables.
16. Hiah Flux at Shutdown Switch -Two position switch.
a) Normal-allows circuit to provide 'High Flux at Shutdown" and !
" Containment Evacuation" alarm when setpoint is exceeded.
l b) Block-used during startup - Blocks High Flux at Shutdown Alarm ! and Containment Evacuation Alarm.
E. Source Range interlocks
l 1. During reactor startup as power increases into the Intermediate Range '
and increases above 10" amps (1 out of 2 channels) permissive P-6 is
energized automatically. This permissive allows the operator to block
the SR High Flux Reactor Trip.
2'.' When reactor power is reduced to less than P-6 (2 out of 21R channels
less than 10" amps. There is some amount of' dead band', so power
may indicate somewhat less than 10" amps before P-6 clears) the SR
j detectors are automatically energized and the SR High Flux Reactor l Trip is placed back in service for both channels. ! i
OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
Page 13 of 38
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DUKE POWER O CATAWBA (\.sERATIONS TRAININ
.
2.3 Abnormal Operation (AP-16) (LPRO/LPSO #11 & 12)(PTRQ #9) '
~
A. Source Range Channel fails " Low" in Modes 2, 3,4 or 5.'
1. Indication indicates " Low" for the failed channel. This can be due to a
, blown instrument fuse, the detector output failing Low or a circuit -
problem which causes that SR channels output to indicate " Low".
- 2. Operator actions should be as follows
a) Enter AP/1/A/5500/16, Case I, Source Range Malfunction.
b) Symptoms for entering this case are:
Indication lost or erratic "S/R Hi Voltage Failure" annunciator
c) Immediate operator actions:
Suspend any positive reactivity additions.
B. Source Range Channel fails "High"in Modes 2,3,4 or 5.
1. Indication indicates " Higher than Actual Counts". This can be due to the
detector output failing to a higher output or a circuitry problem causing
the channel to indicate high. Should this failure cause channel output to
5
exceed 10 CPSu in Mode 2 or 3 a reactor trip should occur and the
crew should enter EP/E-0, Reactor Trip or Safety injection, and perform
the required procedural steps.
2. Operator actions, (if this failure is not severe enough for counts to
5
increase to > 10 CPS) should be as follows:
a) Enter AP/1/A/5500/16, Case 1, Source Range Malfunction.
b) Symptoms are:
Indication lost or erratic "S/R Hi Flux Level at Shutdown"
annunciator.
c) Immediate operator actions:
Suspend any positive reactivity additions.
OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
Page 24 of 38
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Enclosure 5
..
RO Exam Question 96 (BANK #336)
This question has no correct answer.
1
Answer A is not correct because EMF-36 will close VQ-10 (release), however, EMF-36 will not close !
VQ-13 (air addition). l
1
'
Answer B is not correct because EMF-38 signal will result in an SH signal which in turn closes the VQ
containment islation valves (NOT VQ-10 and VQ-13).
Answer C is not correct because VQ-10 will not close until pressure returns to 0 psig and this distractor says ;
pressure is 0.2 psig and decreasing. l
Answer D is not correct because Containment Air Fan low flow is a result of VQ-10 closing rather than a
cause.
References attached:
* Operations Training Lesson Plan OP-CN-CNT-VQ (Containment Air Addition and Release System)
l
Recommendation- l
Delete this question from the examination due to no correct answer.
I ! 1 . l i
. . . .. . . . . .
,
QUESTION #%
O CATAWB4 NUCLEARSTATIONO RO EXAM
O I
1 h(s) Which one of the following conditions will automatically:
e terminate a containment air relesse by closing VQ-10 or
e terminate a containment air addition by closing VQ-13
during startup if EMF-39 (Containment Monitor) is out of service?
A. EMF-36 (Unit Vent Monitor) trip 2 signal
B. EMF-38 (Containment Monitor) trip 2 signal
C. Containment Pressure = +0.2 psig and decreasing
D. Containment Air Fan low flow (at the dischage)
.%
v i
{ r
Ques 336a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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DUKE POWER
(~h
\) (3
CATAWBAq,ERATIONS TRAINING
2.2 Components
A. Containment Air Release and Addition Fans
1. Two fans per unit (A and B fans)
2. Controlled mcnually from switches located on the HVAC control panel in
the control room (ON/OFF).
3. Fans stop automatically upon low air flow of 80 SCFM. (LPRO/LPSO
Obj. #1.4; PTRQ Obj. # 3)
4. Heater in filter train must have its supply breaker closed to run the
associated fan.
5. A time delay upon fan start bypasses the low flow fan trip until fan is
running.
B. Air Release Filters
1. Two filters per unit, each filter contains:
a) Varicealprefilter
b) Astocelfilter
c) Two carbon trays
2. Differential pressure gauges are provided for the prefilter and astocel
filter to determine when the filters need replacing.
3. A temperature alarm and a deluge connection are provided for fire
control.
4. A three kilowatt heater is provided for air temperature control.
C. VQ Fans Discharge to Unit Vent VQ10
1. Controlled by a manual loader located on the main control room HVAC
panel.
2. Closes automatically at 0 psig containment pressure decreasing and
upon high radiation, as seen by EMF 35,36 or 37.
3. Manual loader must be re-zerced before VO-10 will open.
4. Receives open permissive at 0 psig increasing.
D. Containment air addition inlet valve VQ13.
1. Controlled from the HVAC control panel in the main control room
(OPEN/CLOSE).
2. Closes automatically at 0 psig containment pressure increasing. Will not
open unless containment pressure is < 0 psig decreasing.
OP-CN-CNT-VQ FOR TRAINING PURPOSES ONLY REV.14
Page 10 of 15
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DUKE POWER O CATAWBAGERATIONS TRAINING
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:
E. Containment isolation valves VQ2A and VQ3B~
1. Normally closed valves on the fan intake line from containment. '
Operated from the HVAC control panel in the Main Control Room. i
2.~ Close automatically upon containment ventilation isolation signal (Sa) to ,
isolate containment. '
F. Containment isolation valves VQ15B and VQ16A.
1. Normally closed valves on the fan exhaust line to containment.
Operated from the HVAC control panel in the main control room.
2. Close automatically upon containment ventilation isolation signal (Sg) to j
isolate containment. [
G. Instruments
1. Flow
1
2. Discharge Flow l,
a) Flow totalizer
3. Pressure
a) On HVAC board
b) Annunciator at + .125 psig and -0.00 psig
2.3 Operation
A. Review Limits and Precautions, refer to per OP/1/A/6450/17 (LPRO/LPSO
Obj. #3.3; PTRQ Obj. #7)
B. Containment Air Addition end Release (Objective ISS/NLO Obj. #3; ,
LPRO/LPSO Obj. #2.1,2.2; PTRQ Obj.# 5,6) l
l
1. Air Addition Mode is performed if Containment pressure is decreasing
and is 5-0.03 psig.
l
a) Verify cumulative release and addition time will not exceed 3,000
hours for a calendar year. This is recorded in PT/1/A/4450/16
b) Open valves VQ-13,15B and 16A
c) Air enters containment frrem the Auxiliary Building
d) VQ-13 will open only if containment pressure is < 0 psig and will i
close when pressure returns to 0 psig.
e) Close VO-15B and 16A
,
OP-CN-CNT-VQ FOR TRAINING PURPOSES ONLY REV.14
Page 11 of 15
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DUKE POWER h CATAWBA OPERhNS TRAINING
4. Signals or conditions that terminate a release or addition (LPRO/LPS'O i
Obj. #1.3; PTRO Obj. #2)
a) At any time by manual operator action from the control room.
b) Automatically when containment pressure retums to O psig by
i closing VQ-10 or VQ-13.
- c) The Containment Air Releasa Fans automatically shut off when low .
flow is detected at fan discharge.
. l. d) EMF-35,' 36,' 37 will'aDto'clo's's VQ-10.'
e) Any Sg signal will close the containment isolathn valves.-
, 5. Initiating and terminating a GWR form (OP/1/A/6450/17)
,
a) Initiation
'
1) Must be signed and dated by the Operations Shift Manager or
designee on duty.
2) Verify that controlling EMF is operable and source checked.
3) Set the EMF setpoints (Trip 1 and Trip 2) as required by the
GWR per OP/1/A/6450/17.
4) Make all appropriate log entries for a release. 3.
*
(a) EMF log book
(b) Release log book ,
(c) Release recorder 1
(d) Start section of release record
b) Termination i
1) Make all appropriate entries on the GWR ' termination" section.
2) Must be signed by Operations Shift Manager or designee on
duty.
.
6. Containment Ventilation isolation (Sn) effect on VQ (LPRO/LPSO Obj.
#1.5; PTRO Obj. #4)
a) With a release in progress the containment isolation valves will
close creating a low flow condition tripping the fans.
b) With an addition in progress the containment isolation valves will
close.
OP-CN-CNT-VQ FOR TRAINING PURPOSES ONLY REV.14
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t
Bank Question: 318 Answer:D
1
1Pt(s) Unit 1 is conducting a reactor startup. Given the following conditions:
* IR ChannelN35 indicates 8X10-" amps ;
e IR channelN36 indicates 1.1X10# amps l
. SR channelK31 indicates 7.1X10' CPS )
. SR channelN32 indicates 6.8X10' CPS !
l
d
if the N31 SR detector high voltage power supply fails. which one of the i
following statements describes the plant response?
A. A reacter trip signal is generated and a reactor trip occun
because theN31 output fails high.
B. A ranciar trip signalis genersted but no reactor tdp occurs
because one IR channet is above P.6 and the SR high level trip is
automatically blocked.
C. No reactor trip signalis generated because the required
coincidence for the SR high level flux trip is 2/2 SR chanads.
D. No reactor trip signalis generated because the channel N31
output failslow.
<
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e
Ques 31Ba. doc
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P .. . . .,
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.
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,
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,
,
,
L l i l'
. RO Exam Qaestloa78 (BANK #318) {
i
"Ihis question is beyond the scope of operator knowledge.
Knowledge of the camponents associated whh the sourm range nu: lear instruments is limitad to knowing !
l
.
ihe function of the component as it seinses to indication and control output. Intemel component failums are -
I not specifically addressed during training, since the &agnosis of an internal failuss and any conective j
malaisnance of the insawnent is accomplished by lastruanent and Electroales(I&E) techni:ians. j
l
i Failures of electrical power supplies extemal to the source range instrument drawers is taught and addressed '
by annunciator sospoose W= 'Ihis includes the failure of an instrument fuse and a failure of the vital !
*
instrument electncal power supply (annunciator responsc pmcedure) Thelesson plan states that abe ;
operstor should fonow proemdurn1Suidance far the ~5/R Hi Voltage Fadure annunciator and to enser j
,
AP/N550W16, Malfunction of Nuclear 1nstrumentaaion System. Neitherexed especificallyaddressesa
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power supply failus imernal to ibe source range instrwnent drawer. j
.
This question was ,-eiried frtun the 1996 RO n==tantion.' 'Ihat questien specifically staand that the j
instnanent fuss had blown causing a high vohage failure. Modification of the question changed the sought [
knowledge to that of a failure of the high vo'lage power supply whhin the source range instrument drawer, l
! sa==lhmg beyond that required for liczas=d opcretor use and training. l l f- The question is technically correct as written with a correct answer (D). homever as modified, it is beyond 1
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the scope of expected operator kncwledge.
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References ansched: j
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* Operations Training lesson P!an OP-CN-IC-ENB (Excore Nuclear Instnanentation) *
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e OP/1/B/6100/10C (Annun:iator Response for Panel 1 AD.2), page 36 of 67
. . AP/l/AJ $00/16 (Loss cf Nuclear Instrumer.ta: ion Systern), Case 1 l
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Delete this questien from the examination due to inappropriate knowledge requirement. i
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EXCORE NUCLEAR INSTRUMENTATION
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(ENB)
LESSON PLAN
@ 1995 Duke Power
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Prepared By: M,
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Reviewed By: b ,- - Date f-d' k7
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Ops Review : A//// Date
Approved By: mT Date 7//f//7
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( 1. OVERVIEW
This lesson covers the purpose and the functional parts of the Excore Nuclear
instrumentation System and compares Excore Nuclear instrumentation to calculated
reactor power when using Heat Balance inforrnation.
2. REFERENCES:
2.1 Nuclear Instrumentation System. Technical Manual, CNM-1399.04-96
'
2.2 Source and Intermediate Ranoe Housina, Technical Manual, CNM-1399.04-97
2.3 Power Ranae Assembiv Uncomoensatari lonhation Chamber. CNM-1399.04-98
2.4 Precautions, Limitations, and Setpoints, Reactor Control and Protection System, !
Westinghouse (preliminary)
2.5 Electrical One Line Diagram, CN-1705.01-02
2.6 FSAR. Section 7.2 Reactor Protection System
2.7 OP/1/A/6100/01 (Controlling Procedure for Unit Startup)
2_8 OP/1/A/6100/02 (Controliing Procedure for Unit Shutdown)
( 2.9 OP/1/A/6100/03 (Controlling Procedure for Unit Operation)
2.10 AP/1/A/5500/16 (Malfunction of Nuclear Instnamentation System)
2.11 OP/1/A/6700/01 (Unit 1 Data Book), fig 2.1.4 and 22
2.12 CNS Technical Specifications
2.13 Accident Mitigation Training, Westinghouse
2.14 SER 96-10 l
3. AIDS: !
3.1 Handout (s) as prepared by the instructor
32 Transparencies selected by the instructor i
3.3 Slides, videos and/or other motivational tools at the instructor's discretion
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( LPRO TRAINING OBJECTIVES
1. State the purpose of the ENB System.
2. Describe the principle of operation of each detector used.
3. Describe the overlap provided by each range. .
I
4. Describe the function of each output from each range of nuclear
instrumentation.
.
5. Explain the function of each portion of the individual ranges when given a
block diagram of each range.
6. Explain the function of allindications and controls associated with ENB.
7. Describe the " Gamma Compensations" used by each range.
8. Describe the effects of "over" and "under" compensation in the
intermediate range.
,
[
9. Descdbe the plant response to a given detector or instrument failure. l
10. Describe the Tech Specs associated with the ENB system.
11. List the symptoms as given for each case in AP/1/A/5500/16 (Malfunction ,
of Nuclear Instrumentation)
12. Erom memory state the immediate actions as required by AP/1/A/5500/16
(Malfunction of Nuclear Instrumentation).
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DUKEPOWER CATAWBA PERATIONS TRAINING
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13. Describe the use of Heat Balance information for calibration of the Excore
( Nuclear Instrumentation.
. 1
13.1 List inputs to the Heat Balance Calculation.
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13.2 State the effect on Actual Power ya Indicated Power if incorrect information is
used to determine core power.
13.3 State the purpose in using the Venturi Fouling correction factor.
'
13.4 Describe the reasons why procedures require resetting the Venturi Fouling factor !
following planttransients.
l
13.5 Ust the indications used to verify the accuracy of the Thermal Power Best *
Estimate Calculation. -
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14. Describe the source range instrumentation response for voiding in the
core /downcomer region and for core uncovery.
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Time: 4.0 Hrs
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( LPSO TRAINING OBJECTIVES
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1. State the purpose of the ENB System.
2. Describe the principle of operation of each detector used. ,
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3. Describe the overlap provided by each range. .
-
4. Describe the function of each output from each range of nuclear -
instrumentation.
. .
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, 5. Explain the function of each portion of the individual ranges when given a -
block diagram of each range.
6. Explain the function of allindications and controls associated with ENB. I
7. Describe the " Gamma Compensations" used by each range.
8. Describe the effects of "over" and "under" compensation in the
intermediate range.
g. Describe the plant response to a given detector or instrument failure.
10. Describe the Tech Specs and Bases associated with the ENB system.
11. List the symptoms as given for each case in AP/1/A/5500/16 (Malfunction
of NuclearInstrumentation)
12. From memorv state the immediate actions as required by AP/1/A/5500/16
(Malfunction of Nuclear Instnamentation).
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, 13. Describe the use of Heat Balance information for calibration of the Excor6
[ Nuclear instrumentation.
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13.1 Listinputs to the Heat Balance Calculation.
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13.2 State the effect on Actual Power,yg Indicated Powerifincorrectinformation is
used to determine core power. .
," 13.3 State the purpose in using the Venturi Fouling correction factor.
~
13.4 Des cribe the reasons why procedures require resetting the Venturi Fouling factor
following plant transients.
'
13.5 Ust the indications used to verify the accuracy of the Thermal Power Best
Estimate Calculation. ,
.
14. Describe the source range instrumentation response for voiding in the
cors/downcomer region and for core uncovery.
.(.
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. Time: 3.0 Hrs
OP CN-IC-ENB FoR TRAINING PURPOSES ONLY REV.19
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( L1. INTRODUCTION
1.1
Purpose - To protect the reactor core by monitoring neutron flux ard generating . j
appropriate alarms and trips during all power levels from shutdown to full power l
operation.(LPRO/LPSO #1) '
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2. LESSON PRESENTATION I
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2.1 Functional / Component Description .
A. Source Range Detector (LPRO/LPSO #2) .
:
1. BF3gas filled pisperdonalcounter.
2. Detects thermalleakage neutrons
a) .N' = ,B" -+ ,B" -+ ,Lf*" + 2He*** l
d '
b) Positive charged ,He " and ,lf*** create voltage pulse through
lonization of gas.
B. Source Range Circuitry (LPRO/LPSO #5; PTRQ #3)
1. Pleamolifier located near Rx to minimize noise pickup. Receive pulses
, from detector.
( a) Fumishes high voltage coupling to detector (-2000 VDC).
-
Generated from 118 VAC supplied to drawer and rectified to
produos DC.
b) Generates pulse signal for testing and calibration. (60 pulse per
second and 10* pulse per second oscillators and multivibrator)
2. Pulse-amo/ Discriminator
a) Amplifies signalfrom pre-emp.
. b) Discriminates against noise & Gamma induced pulses. (PTRQ #5;
LPRO/LPSO #7)
c) Provides a signal to audio count rate through an isolation amplifier.
3. Pulse Shaper - Shapes pulses into uniforrn square waves
4 Pulse Driver- Matches impedance between Pulse Shaper and Log
Pulse integrator
5. Loo Pulse intearator- Changes pulse signal to voltage output
proportional to logarithm of pulse rate.
!
6. Level Amplifier- Amplifies signal from Log Pulse integrator by a factor of !
.
40. Sends signalto Bistable Relay Drivers and Isolation Amp. ,
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7. Bistable Relav Driver provides:
a) "High Flux at Shutdown" Alarm (set by lAE on shutdown to half
decade > shutdown countrate) .
b) " Containment Evacuation" Alarm (set at same time as high flux at
shutdown alarm is set. These are set at the same setpoint)
c) "High Level Trip" Signal (setpoint of 10 5cps as seen on 1 out of 2
SR detectors) -
8. Isolation Amp provide noise discrimination be' tween Pulse Amp and
Audio Count rate channel and provides signal from Level AMP to the
following;
a) Computer 1
b) SUR Circuit
c) Control Board Meter
d) Control Board Recorder- NR-45
i
it isolates the SR channel from faults in other circuits or equipment. !
C. Source Range Circuit outputs (LPRO/LPSO #4: PTRQ #2)
1.
Control Board Indication - reads out in Counts Per Second (CPS) with a
(' range of10"to 106 CPS
_ 2. Control Board Recorder- CPS recorded on NR-45 Recorder, when
selected.
3. Startuo Rate Circuit Sicnal
a) Source Range Output (CPS) converted to rate signal. l
b) Display on control board and Comparator and rate drawer in
decades / min.
4. Hioh Flux of Shutdown l
4
a) Seipoint - 1/2 decade above nonnat shutdown counts.
b) Actuates containment evacuation alarm in containment.
c) Sounds annunciator in Control Room.
d) Blockable during startup
e) Not required by Tech Specs unless Boron Dilution Mitigation
System is inoperable (either train).
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OP-CN-lC-ENB FOR TRAINING PURPOSES ONLY REV.19
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7. Bistable Relav Driver provides:
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a) "High Flux at Shutdown" Alarm (set by lAE on shutdown to half -
l decade > shutdown countrate) ,
b) " Containment Evacuation" Alarm (set at same time as high flux at
shutdown alarm is set. These are set at the same setpoint)
c) "High Level Trip" Signal (setpoint of 10' cps as seen on 1 out of 2
SR detectors)
8. Isglation Amp provide noise discrimination be' tween Pulse Amp and
Audio Count rate channel and provides signal from Level AMP to the
following:
a) Computer 1
b) SUR Circu:1 !
c) Control Board Meter
d) Control Board Recorder- NR-45
It isolates the SR channel from faults in other circuits or equipment.
C. Source Range Circuit outputs (LPRO/LPSO #4; PTRQ #2)
1.
Control Board indication - reads out in Counts Per Second (CPS) with a
( range of108 to106CPS
2. Control Board Recorder- CPS recorded on NR-45 Recorder, when
selected.
3. Startuo Rate Circuit Sicnpj
a) Source Range Output (CPS) converted to rate signal.
b) Display on control board and Comparator and rate drawer in
de::ades/ min.
4. Hioh Flux at Shtddown
a) Seipoint 1/2 decade above nonnat shutdown counts.
b) Actuates containment evacuation alarm in containment.
c) Sounds annunciater in Control Room.
d) Blockable during startup
e) Not required by Tech Specs unless Boron Difution Mitigation
System is inoperab!e (either train). l
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DUKE Pot 9ER CATAWBADPERATIONS TRAININ3
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5. Hioh Level ReactorTrio
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( . a) Setpoint- 1 out of 2 SR channels > 105 eps
b) Blockable greater than P-6 (1 out of 2 Intemiediate Ranges greater
'
than 10-10 Amps.) Blocked by tuming 2 switches on C/B to " Block".
. By taking both switches to " Block" the high level reactor trip for both l channels of SR are blocked and power is automatica!y removed ! from both SR detectors causing all SR indication to decrease to
zero. If only one switch is taken to " Block", only that channe!'s high
level reactor trip is blocked and both detector remain energized.
'
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D. Source Range Drawer Panel (LPRO/LPSO #6: PTRQ #4)
1. Detector Volts Meter- Indication of high voltage supplied to proportional
counter.
2. Neutron Level Meter- Scale 100 to 106 CPS.
3. Instrument Power "0N" Lamo - 118 volts AC inst. power applied to
drawer.
4. Control Power"ON" Lamo - 118 volts AC control power applied to
drawer.
5. Channel "On Test" Lamo -indicates the operation selector switch is in a
( position other than * NORMAL".
. 6. Loss of Detector Volt Lamo -indicates high vahage to detector off p_r
low.
7. Level Trio Lamo - indicates neutron level greater than trip setpoint in
Source Range.(105 CPS)
8. Level Trio Bvoass Lamn - On when Level Trip switch in " Bypass" for test ;
or calibration. !
9. Hich Flux at Shutdown Lamo - Neutron level greater than 1/2 decade
above normal shutdown level.
10. Bistable Trio Soare Lamo - No function i
11. Instrument Power Fuses - Overcurrent protection for power supply I
circuits. Instrument power supplies the meters, circuit processing !
components, high voltage supply and detector power. This is true for
the IR and PR drawers / circuits also. i
12. Control Power Fuses - Overcurrent protection for control signal circuit
transformers. Control power supplies the lights on the drawer and 118
VAC to the bistable relay drivers to the plant relays. (High flux at
i shutdown a! arm and SR high level trip). This is true for the IR and PR
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drawers / circuits siso.
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13. Level Trio Switch -Two position switch. i
a) Normal-Switchinactive
b) Bypass - Enables operation selector switch for test and calib' ration. -
Provides AC signal to prevent Rx trip signal during testing. -
i
14. Ooeration Selector Swich
a) Eight Position Switch
b) Enabled by Level Trip Switch to ' Bypass'. !
c) Normal Position
d) Six Test Positions
e) Level Adjust - Channel On Test lamp !!tes & Level Adjust !
Potentiometer switched into test circuitry.
f) Actuates "N/l System Channel on Test" annunciator when taken out
of' Normal' position.
15. Level Adiust Potentiometer- Adjustable test signalinto level amp. -
Enables adjustment of the inp level of various bistables.
'
16. Hioh Rux at Shutdown Switch - Two position switch.
a) Normal -allows circuit to provide 'High Flux at Shutdown * and
,(
, ' !
" Containment Evacuation" alam1 when setpoint is exceeded.
, ' , b) Block-used during startup - Blocks High Flux at Shutdown Alarm
and Containment Evacuation Alarm.
E. Source Range interiocks
l 1. During reactor startup as power increases into the intermediate Range
and increases above 10* amps (1 out of 2 channels) permissive P-6 is
energized automatically. This permissive allows the operator to block
the SR High Flux ReactorTrip.
2. When reactor poweris reduced to less than P-6 (2 out of 2 IR channels
less than 10* amps. Thero is some amount of ' dead band', so power
may indicate somewhat less than 10* amps before P-6 clears) the SR
detectors are automatically energized and the SR High Flux Reactor
Trip is placed back in service for both channels.
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OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
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3. When reador power is increased greater than P-10 (2/4 PR channels >
( 10% reactor power) the SR detedor power is further controlled and
, assured to be de energized. This controlis automatically removed
when reactor power is reduced below the P-10 setpoint and P-10 is de-
energized. Should P-10 not de-energize (either due to 3 out of 4 PR
channel P-10 bistables not clearing or the 2 out of 4 'and' gate not going ..
.
to the 'NOT P-10' state) then the SR detectors can not be re-energized
eltherautomaticallyormanually. '
F. Intermediate Range Detector (LPRO/LPSO #2) -
1. Detector- Compensated ion Chamber- (gamma compensation required
only below 10* amps) -
a) Two Volume Detector
'
1) Outer volume Baron lined, sensitive to neutrons and gammas.
2) Inner volume sensitive only to gamma - not bomn !!ned.
3) Each volume generates current output.
b) Inner volume current due to gamma interacdons with N 2 gas.
c) Outer volume current generated from neutron interactions with
Boron 10 lining and gamma interactions with N2 gas.
( d) A high de voltage of approximately 800 volts is applied to the
. detector to provide for the collection of aff charged particles for each
ionizing event. A low de voltage between 0 and -100 volts is applied
to the gamma compensating electrode. Compensation is necessary
because after sustained fu!I power operation of the reactor, there is
an appreciable amount of residual gamma flux. The compensation
voltage is adjusted to cancel out the signal due to gamma flux
leaving an output from the detector which is proportional to neutron
flux only. Due to the high rate of neutron pulses detected in the
intermediate Range the output frorn the compensated ion chamber
detector is direct current, and is coupled directly to log current
amplifier.
2. Under Compensation (LPRO/LPSO #7; PTRQ #5)
a) Higher power indication than actual power level.
b) Can restit in attaining an extremely high SUR without seeing it on
SUR meters.
c) Can possibly prevent P-6 clearing and prevent automatic re-
energizing of SR detectors following a Reactor Trip,
d) Will energize P-6 earlier than ~ expected during Reactor startup.
OP-CN-IC-ENB FOR TRAINING PURPOSES ONLY REV.19
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- 2.3 Abnormal Operation (AP-16) (LPRO/LPSO #11 & 12)(PTRO #9) -
'
A. Source Range Channel fails " Low"in Modes 2,3,4 or 5.
1. Indication indicates * Low" for the failed channel. This can be due to a '
blown instrument fuse, the detector output failing Low or a circuit
problem which causes that SR channels output to indicate " Low".
2. Operator actions should be as follows:
a) Enter AP/1/A/5500/16, Case 1, Source Range Malfunc:lon.
b) Symptoms for entering this case are:
Indication lost or erratic "S/R HI Voltage Failure * annunciator
, c) Immediate operator actions:
I
Suspend any positive reactivity additions.
. B. Source Range Channel fails *High"in Modes 2,3,4 or 5. '
1. Indication Indicates * Higher than Actual Counts". This can be due to the .
, detector output failing to a higher output or a circuitry problem causing
the channel to indicate high. Should this failure cause channel output to
exceed 105 CPSu in Mode 2 or 3 a reactor trip should occur and the
crew should enter EP/E-0, Reactor Trip or Safety injection, and perform
the required procedural steps.
2. Operator actions, (if this failum is not severe enough for coun+s to
. -
increase to > 10' CPS) should be as follows:
l a) Enter AP/1/A/5500/16, Case 1, Source Range Malfunction.
b)' Symptoms are:
Indication lost or erratic *S/R Hi Flux Level at Shutdown"
annunciator.
c) Immediate operator actions;
Suspend any positive reactivity additions.
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OP-CN-LC-ENB FOR TRAINING PURPOSES ONLY REV.19
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-
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CIRCUIT
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BISTADC$ RELAY DRIVER
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OP/1/B/6100/010C -
PANEL: 1AD-2 Page 36 of 67 i
i
S/R HI VOLTAGE FAILURE D/1. . !
!
SEITOINT: 100 volts belowhivoltage setting
NOTE: ' Ibis alarm is disabled when above P-10 (10% reactor power). L
,
ORIGIN: Bistable hivoltage to detector
PROBABLE 1. Source range drawer instrumant power fuse blown.
CAUSE: 2. Power supply failure.
3. S/R Hi Hux trip blocked and detectors de z._M on startup at
P-6
AUIOMATIC Possible reactortrip.
ACI10NS:
NOTE: If due to power supply failure at ERPA, I/R channe135 and P/R ehanaal 41 A & B control
and instnanent power is lost. Ifpoweris lost at ERPB, I/R channel 36 and P/R
rhannal 42 A & B control and instnanent power is lost.
IMMEDIATE Refer to AP/1/A/5500/016 (MalFaMon ofN6cicar Instrumentation
ACTIONS: System) ifant due to trip being blocked.
SUPPLEMENTARY 1. Restore power from ERPA or ERPB ifnecessary.
. ACTIONS: 2. Refer to OP/1/A/6350/008 (125VDC/120VAC Vital Instrument and
ControlPowerSystem)as 7
3. Refer to Tech Spec 3.3.1 (Reactor Trip System I ..=ntation).
. . 4 Ifin Mode 6, then refer to Tech Spec 3.9.2 (Ra&he Operations.
*
lastnurmatation). .
REFERENCES: 1. CNM-1399.04-96 -
2. CN-1705-01.02-01
3. Tech Specs 3.3.1 and 3.9.2
l
. ! 't
-
- . -
. _ _
-- -
t} e -
Q l
meson l
:
Duke Power Company (1)lD No.AP/1/A/5500/10 I
PROCEDURE PROCESS RECORD Rew.sonsig
i
PREPARATION * '
(2) station CATAWBA NUCLEAR STATION
I
(3) Procedure Title MALFUNCT10N OF NUCLEAR INSTRUMENTATION SYSTEM
.
(4) Prepared By OJ lls Me Date /o-/S -97 l
(5) Requires 100FRSO,59 evaluation?
g' Yes jNew procedure pr revision with major changes)
No :Beymon with nunor changes)
No i,To encorpora:e approved changes)
(6) Reviewed By Auu m (QR) Date /O-/[-77
Cross-Desc# nary Review By haihbEA (QR) NA Date 10[P9h7
Reactrvity Mgmt. Review Dy
(QR) NA b __ _Date
(7) Additional Reviews
Reviewed By vm 4%_ =-- (b/(.7%) Date _- 10-/6-9; 7
Reviewed By h Date
(0) Temporary Approval p/nocessary)
By_---.
(SPO/QR) Date
- By__ .
.
(OR) Date
(9 Approved By M, *r Aff ) b> >[A/ Date 10/2;t/97
PERFORMANCE (Compare with control
every 14 cale dar days while work is being performed.)
(10) Compared with Control Copy Date
Compared with Control Copy
Date
l
Compared with ControlCopy Date
(11) Date(s) Perfonned
Work Order Number (WO4)
COMPLETION
(12) Procedure Completion Veri'ication
i
O Yes D N/A Check lists and/or blanks propertyinitialed, signed, dated, or filled in N/A or N/R, as appropriate?
O Yes D N/A Usted enclosures attached?
O Yes D N/A Data sheets attached, completed, dated, and signed?
O Yes D N/A Charts, graphs, etc. attached and property dated, identified, and marked?
O Yes D N/A Procedure requirements met?
Verified By _
Date
(13) Procedure Completion Approved--
Date
(14) Remarks (attach additionalpages, if necessary)
-r
INFORMAr0N ORY
_ _ _ _ . _ _ . _ .
, , . _ __.
L. O~ O i
CNS
MALFUNCTION OF NUCLEAR INSTRUMENTATION SYSTEM PAGE NO.
AP/1/A/5500/16
1 of 13
Revision 16
! l l A. Purpose
*
To verify the proper response in the event of a nuclear instrumentation malfunction.
.
B. Symotoms
Case 1. Source Range Malfunction
* Indication lost or erratic
* 1 AD-2, D/1 "S/R Hi VOLTAGE FAILURE"- LIT
* 1 AD-2, D/3 "S/R Hi FLUX LEVEL AT SHUTDOWN" - LIT -
* 1 AD-2, D/4 "S/R HI FLUX LEVEL AT SHUTDOWN" - LIT.
I
Case 11. Audiu Count Rate Malfunction
* Audible count rate lost.
>
Case 111. Intermediate Range Malfunction
= Indication lost or stratic
* 1 AD-2; C/1 "l/R HI VOLTAGE FAILURE" - LIT
* 1 AD-2, C/2 "l/R COMPENSATING VOLTAGE FAILURE" - LIT
l * 1 AD-2, C/3 "l/R HI FLUX LEVEL ROD STOP" - LIT
* S/R failure to re-energize during shutdown.
Case IV.' Power Range Malfunction
!
* Indication lost or erratic
l * 1 AD-2, All "P/R HI NEUTRON FLUX RATE ALERT"- LIT
a
i 1 AD-2, A/2 *P/R HI NEUTRON FLUX LO SETPolNT ALERT" - LIT i *
1 AD-2, A/3 "P/R HI NEUTRON FLUX Hi SETPOINT ALERT" - LIT
i ' * 1 AD-2, B/3 "COMPARATOR P/R CHANNEL DEVIATION" - LIT
* 1 AD-2, B/5 *P/R HI VOLTAGE FAILURE" - LIT
i * 1 AD-2, E-B "OVER POWER ROD STOP"- LIT.
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CNS MALFUNCTION OF NUCLEAR INSTRUMENTATION SYSTEM
AP/1/A/5500/15 PAGE NO.
c ,, j 2 of 13
Source Range Malfunctions Revision 16
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED -
C. Operator Actions
i Stop any positive reactivity additions in ,
progress.
2. Identify the affected S/R channel:
__ = N31
OR
.
_ = N32.
3. Verify the following alarms - DARK: Perform the following:
! ~
e 1 AD-2, D/3 "S/R Hi FLUX LEVEL AT a. Depress "OFF" for the containrnent
~
SHUTDOWN" evacuation alarm.
1
_ = 1 AD-2, D/4 "S/R Hi FLUX LEVEL AT _ b. Notify plant personnel to disregard
SHUTDOWN". the containment evacuation alarm.
I
4. At the affected S/R drawer, perform the
following:
_ a. Place the " LEVEL TRIP switch for i
the affected channelin " BYPASS". i
_ b. Verify the LEVEL TRIP BYPASS"
light on the affected S/R drewer -
LIT. .
'
_
5. Verify the effected S/R channel trip
bypass status light (151-19) - LIT.
I
i l
-
6. Verify 1 AD-2, C/4 "N/l SYS S/R & 1/R
, . TRIP BYPASS" - LIT. .
i
e
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,
_ _ . _ . _ _ . _ . .
_ _ . _ . . _ _ _ _ _ _ . . _ , . , _ _ _
. O O
CNS
AP/1/A/5500/16 MALFUNCTION OF NUCLEAR INSTRUMENTATION SYSTEM PAGE NO.
Case 1 3 of 13
Source Range Malfunctions .
Revision 16
ACTION / EXPECTED RESPONSE
RESPONSE NOT OBTAINED -
7. hek the high flux at shutdown
function as follows:
_ u. At the affec'ted S/R drawer, place , i
the "HIGH FLUX AT SHUTDOWN"
switch for the affected channel in
j " BLOCK".
l
~
b. Verify 1 AD-2, D/2 "S/R H1
SHUTDOWN FLUX ALM BLOCKED"
-LIT.
l l
8. Ensure the following components -
ALIGNED TO THE OPERABLE S/R
CHANNEL:
- . " CHANNEL SELECTOR" switch on
the Audio Count Rate Channel panel I
_ * *NIS RECORDER".
- 9. Determine and correct cause of SIR I
malfunction. I
i
10. Ensure compliance with appropriate
Tech Specs: ,
I
_ * 3.3.1 (Reactor Trip System
Instrumentation)
l
_ * 3.9.2.1 (Refueling Operations
Instrumentation)
_ * 3.3.3.11 (Boron Dilutior) Mitigation
System). i
t
I i 11. Determine required notifications:
_ * REfEB T_Q RP/0/A/5000/001
(Classification Of Emergency) '
a
_ REFER TQ RP/0/B/5000/013 (NRC
Notification Requirements).
1
, . . _ . - _ . _ _ .____.----- __--.__ -
-_ _ . . . - _ . . _ _ _ __
. _ _ . _ . , . . _ _
_
- . .
- - -
.
CNS MALFUNCTION OF NUCLEAR INSTRUMENTATION SYSTEM PAGE NO.
AP/1/A/5500/16 4 of 13
Source Range Malfunctions Revision 16
.
ACTION / EXPECTED RESPONSE RESPONSE NOT OBTAINED -
--
__
_ 12. Notify Reactor Group Engineer of
occurrence.
'
_ 13. WHEN the affected S/R channel is
repaired, THEN ensure IAE returns the
channel to service.
l
~ 14. Determine long term plant status.
HETURN IQ procedure in affect.
ENQ.
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Source Range Response to Loss of High Voltage.
By: Jeny White
Sr. Tech Spec
CNS CEN-E Engineering
01/12/97
' The followingis CNS Engineering's response, to a question from CNS Operator Training staff.
The question concerned the effect oflosing the High Voltage Power Supply in the NIS Source
Range drawen to the plant.
h7S EQUIPMENTRESPONSE
The effect oflosing the Nigh Voltage Power Supply for the Source Range NIS Syo.e.ut is as follows:
1. If the High Voltage Power Supply failed totally and abmptly, (output went from setpoint voltags
to zero volts), then:
a) The LEVEL TRIP bistable would NOT trip.
b) HIGH FLUX AT SHUTDOWN would NOT trip or would reset for the effected chan.._1.
c) Loss of High Voltage Bistable would illammate.
d) All indications wuald read Zero Counts.
c
2. If the High Voltage Power Supply did not failed totally and abruptly (output spiked high befote
going to zero volts), then:
a) The LEVEL TRIP bistable may or niay not trip, depending on whether or not the count
- rate was close to setpoint when the spike occuned. The spike probably would not have :
an effect; however, there is a possibility. After total supply failure, the bistable would
then go to the non- tripped state.
b) HIGH FLUX AT SHIJIDOWN may or may not uip, depending on whether or not the
count rate was close to serpoint when the spike occurred. This bissable would most likely
trip, if not already tripped. After total supply failure. the bistable would then go to the
non- tripped state.
c) Loss of High Voltage Bistable would illuminate.
d) All indications would read Zero Counts after total supply failure.
PLANT RESPONSE
The effects to the plant, during the above evolution, follow: (Note: This review assutned a loss of ONE
drawer et a tirne. Multiple failures were not reviewed.)
1. Source Range
a) If poweris belowP6.
i) Total and abmpt loss of the High Vohage power supply.
a) With LEVEL TRIP Normal, will NOT cause a reactor trip.
b) With LEVELTRIP Bypassed, will NOT cause a reactor trip.
c) With HIGH FLUX AT SHUTDOWN Normal will NOT cause an alann.
Page1
--.
O O
d) With HIGH FLUX AT SHUTDOWN Blocked, will NOT cause an
alarm.
e) Power supply failure alarms.
ii) Output <=&ine hinh during High Voltage power supply failure.
a) With LEVEL TRIP Normal, may cause a reactor trip.
b) With LLVEL TRIP Bypassed, will NOT cause a reactor trip.
c) With HIGH FLUX AT SHUTDOWN Normal may cause an alarm.
d) With HIGH FLUX AT SHUTDOWN Blocked, will NOT cause an
alarm.
e) Power supply failure alarms.
b) If power is above P6 /G2 Source Range Trip is Manually BLOCKED by the operator,
or poweris above PIO.
i) ne Soura: Range High Voliage power supply is turned off when the channelis
blocked, No effect on plant systems.
>
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Page 2
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NRC RESOLUTION OF POST-EXAMINATION COMMENTS
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RO Question #10 . j
SR0 Question #8 j
)
l
Recommendation accepted. The additional information provided by and :
l instrument drawer bench testing conducted by the licensee, as well as !
-information independently obtained by the examiner, demonstrated that the i
Intermediate Range instrument reactor protection system bistables will fail to !
<
the safe condition (trip >ed) if the instrument power fuses blow. Since a j
reactor trip signal.is t1us' generated, the operators should enter E0P E-0 at i
step 1 which makes answer A the correct response to this question. The. j
correct answer is changed from C to A. ;
R0 Question #76
Recommendation accepted. The correct answer to this question had always been :
the response given in answer A. However, during the course of question i
review, modification, distractor rearrangement and finalization, the answer i
key was not updated properly.to reflect the location of this answer. The !
' correct answer is changed from C to A. l
!
R0 Question #78 l
Recommendation not acce)ted. The licensee originally recommended deletion of ]
this question because t1ey inferred that the loss of the Source Range High
Voltage (SRHV) power supply was due to loss of instrument power. If this were:
the case, then the reactor protection system bistables would fail to the safe !
condition (tripped) and a reactor would tri). Since none of the possible I
answers properly desc.ribed this response, tie licensee recommended deletion of.
the question. However, the question specifically stated that the SRHV Jower
supply had failed. There was no reference made in the question as to t1e !
status of instrument power elsewhere in the drawer. When no reference is made
. to a problem with a particular system or component, the candidates are
' instructed to assume that the system is in its normal condition. Thus for
- this question, instrument power should be assumed to be energized and
operating properly except where the loss of SRHV impacts its function.
Upon.further questioning and clarification by the examiner, the licensee
reviewed plant drawings and instrument technical information. Based on this
additional review.' the licensee concluded that a reactor trip signal would not
generated under the conditions stated in the question. Consequently, the
licensee amended its original post-examination review comments to reflect new
understanding of the question and agreed the answer was correct as originally
approved;
However, the licensee also stated the question required knowledge that was
j. outs.ide the scope of expected operator knowledge and was inappropriate for use
- on the examination. The NRC disagrees. While the stem of the question asked
l-
Enclosure 6
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- _ _ _ _ _ _ _
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- - - . - - . - . - . - - - . , - - . . . . - - - - . - - _
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L the plant response to a specific failure of one component within the Source i Range instrument drawer, the question actually measured the candidate's l l understanding'of the instrument logic (SR tria coincidence - 1/2 vice_2/2. ;
1/21R' channels'above P6), interfaces with otler equipment (IR instrument and i
reactor protection). general instrument design (detectors fail low on loss of :
aower).- setpoints (SR hi flux trip setpoint. P6 set)oint) and permissives-(SR i
alocking) . All the above were within the expected (nowledge level of an :
l . operator and were within the system lesson plan.
.
:
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Additionally the control room has a specific annunciator (AD-2. D/1) that j
l warns operators of a failure of SRHV. The arinunciator arocedure states a ) '
reactor trip may occur (depending on the cause of the S1HV failure). For ;
example, a loss of instrument power would cause a loss of SRHV and its ;
associated annunciator: and if a plant startup were in arogress with the SR hi i
flux trip unblocked. a reactor trip should occur. If t1e reactor did n t t
trip, a failure of the RPS occurred and the operators should manually trip the l
reactor;
The NRC' believes it important that an operator know when how and why a valid
-reactor trip signal should occur. . Conversely. an operator should also know ;
when, how and why an instrument failure should not cause a reactor trip. This -
understanding is vital to the operators correctly responding to the '
malfunction while avoiding unnecessary challenges to the plant's safety !
equipment.
)
The question is not' deleted from tae examination and the correct answer
remains D. ,
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RO Question #96 ;
,
Recommendation accepted. This was a new question written by the licensee's ;
contractor for this examination. Several attempts were made to clarify and !
'
enhance the question during the examination review. However. the question
-developer, the licensee reviewers and the NRC examiner reviewers all
misinterpreted the VQ system lesson plan and thus, did not recognize that no
single condition would cause automatic closure of both VQ-10 and VQ-13 during
startup with EMF-39 out'of service. Consequently none of the possible
answers provided for this question were correct. This question is deleted
from the examination.
RO Question #55
SRO Question #42
During a post-examination question writing workshop the NRC identified that
-
the stem of this question was not specific enough to elicit the desired
response. Despite the initial conditions and supporting information this
question only asked for the initial average cooldown rate to remain within
procedural limits. _ No time frame was specified. As long as an operator did
not cooldown longer than 12 minutes at an initial average rate of 50 F/hr. a
plant cooldown of 10 F would occur and the operator would still be within
-
limits.. . (Such a course of action would also require no further cooldown for
l
Enclosure 6
,.-7+g. ,-_-.v-----n ---.n - - - -
- .. .- - . . . - . . - - . . - _ . . . - - . . - _ - - . _ - -
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the ensuing 18 minutes.) Consequently, the Technical Specification cooldown
limit of 50 /hr (answer B) is also an acceptable response for this question.
The answer key was changed _to accept either A or B as correct responses.
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Enclosure 6
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; eArmazu,ria
5-'N1 'll Yl97-3c
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NRC Official Use Only /2['-/f[97
(No Yorema
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Nuclear Regulatory Commission
Reactor Operator Licensing
- Examination .
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This document is removed from
Official Use Only category on
date of examination
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NRC Official Use Only l
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ES-401 Site-specific Written Examination Form ES-401-1 l -
Cover Sheet !
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U. S. NUCLEAR REGULATORY COMMISSION
SITE-SPECIFIC ,
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WRITTEN EXAMINATION
APPUCANT INFORMATION
Name: Region: I /g/111/ IV / V
Date: Facility / Unit: CATAWBA i
4
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License Level: @] / SRO Reactor Type: [_W2/ CE / BW / GE
I
INSTRUCTIONS
Use the answer sheets provided to document your answers. Staple this cover
sheet on top of the answer sheets. Points for each question are indicated in
parentheses after the question. The passing grade requires a final grade of at
least 80 percent. Examination papers will be picked up 4 hours after the
examination starts.
l
All work done on this examination is my own. I have neither given nor received
aid. ;
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Applicant's Signature
RESULTS ,
1
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Examination Value 100 Points
Applicant's Score Poirits
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Applicant's Grade Percent
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L 402 Policies and Guidelines Attachment 2
for Taking NRC Written Examinations
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1. Cheating on the examination will result in a denial of your application and could result in I
more severe penalties.
]
1
2. After you complete the examination, sign the statement on the cover sheet indicating that l
thd work is your own and you have not received or given assistance in completing the ! examination.
3. To pass the examination, you must achieve a grade of 80 percent or greater. 4. The point value for each question is indicated in parentheses after the question number. 5. There is a time limit of 4 hours for completing the examination. 6. Use only black ink or dark pencil to ensure legible copies. 7. Print your name in the blank provided on the examination cover sheet and the answer sheet. 8. Mark your answers on the answer sheet provided and do not leave any question blank. i
i
.. If the intent of a question is unclear, as questions of the examiner only. ; 10. Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination room to e!iminate even the appearance or possibility of cheating.
11. When you complete the examination, assemble a package including the examination
questions, examination aids, and answer sheets and give it to the examiner or proctor. 1
Remember to sign the statement on the examination cover sheet.
.
12. After you have turned in your examination, leave the examination area as defined by the ;
examiner. '
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REACTOR OPERATOR A N S W E R -S H E ET Page 1 - ! M' 1PLE CHOICE -(Circle or X your choice) If ha change your answer, write your selection in the blank. .,
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001 a'b c @ D 035 @ b ~c d A 069 a b @ d C, i
002 a b'c @ O 036 a b @ d C. . 070 a b c @ l ;
003 @ b c d 1 ' 037 a b @ d C, 071 a @ c d 1 :
004 a b c @ l - 038 a b c @_D., 072 a b @ d L !
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005 a b @ d 1 - 039'a b c @ A 073 a b @d L ;
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006 @ b_c d A 040 a,b @ d 1 074abc@_Q,
007'a b'@ d,_C., 041 a b @ d 1 ' 075 a @ c d 1 l
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008 @ b c'd A ~ 042 a ' b @ d 1 076 a b @ d L ,
009 a b c @ l 043 a @ c d 1 077 a b c @ ;
010.a b @ d L .044 @ b c d 1 078 a b c @ A !
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011 @ b c d i 045 a b @ d L 079 @ b c d A -
012'a b c @ A 046' a b c @_D_ 080a@cdR j
:013 a @ c_d 1 -
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047 a b c @ A 081 @ b c d A ;
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-014.a @ c d 1 048 @ b c d 1 082 @ b c d A
015a@cdj_ 049 a @ c d 1 083 @ b c d A
016.a b @ d 1 050a@'cdH 084 a b c @ l
017 a @ c d 1 -051 a @ c d 1 085 a @ c d 1
018abc@,,Q, 052 a b c @ A 086 @ b c d ' A
.019.a b c @ f 053 a b @d _Q 087 @ b c d A
020 a:b @ d 1 054 @ b c d A- 088 a @ c d B
021 a b c @ A ~
055 @ b c d A 089 a b c @ j_,
022 a @ c'd B 056 a b c Q 090 @ b c d A
023 @ b'c_d A 057a@cdJ_' 091 a b @ d _(,_
024 a b @ d_C_ 058 @ b c d A 092 a b c @ A
025 @ b c-d A' 059 @ b c d A 093 a b @ d L .]
026 a b @ d L _ 060 @ b c d A1, 094 a @ c d G_
027 a @ c d 1 061 @ b c _ d A 095 a @ c d 1 _
028 @ b :c d - A 062_ a b c @_Q,- 096.a@cd.1 )
' 029 @ b c d . A' 063 a.b @ d_C_ 097 a b @ d L
030'a'b c @_D. 064 a b @ d C, 098 a b @ d L i
031'a @ c'd & . 065-a b c @ A 099 a b c @ n- .
032 a b:c @ D 0661a'b@d_C,_, 100 a @ c d 1
033 a'b @ d_L_ 067 - a @ c d A ,
034-.a b'c @ A 068ab'c@A_ !
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(' " " " " END OF EXAMINATION * * * " " " *)
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Question #1 CATAWBA NUCLEAR STATION RO EXAM i
Bank Question: 240 Answer: D i
1 Pt(s) Unit I was operating at 100% power when a multiplexing thyristor for bank C
in the rod control system failed causing bank C to be energized at the same
time as bank D. If rod control is in automatic, which one of the following r
sequence of events describes the response of the rod control system following
this failure?
,
A. Banks C and D will move at the same time in response to rod
withdrawal or insertion signals.
B. Banks C and D will move at the same time in response to rod i
insertion signals only.
C. Banks C and D will only move together for manual rod insertion
signals. Automatic signals will have no effect on md movement.
l D. Banks C and D will not move for rod insertion or withdrawal j
signals. i
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For Official Use Only Ques 240a NRC EXAM 12/12/97
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QUESEON #2 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 241 Answer: O
1Pt(s) Unit 1 is operating at 100% power when the supply breaker from LTC to #1
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Control Rod Drive MG se opens. Which one ofthe following sequence of *
events will occur to the reactor trip breakers (RTB) and the reactor 3 trip b pass
breakers (RYB)?
,,
A. RTB "A" and RYB ""B" will open
B. RTB *B" and RYB "A" will open
C. RTB "A" and "B" will open
,, ._ D. No RTB or RYB bruken willopen I
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Ques 241a. NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Quesbon #3 CATAWBA NUCLEAR STATION RO EXAM !
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._ Bank Question: 242 Answer: A
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1 Pt(s) Unit 1 is operating at 100% power. Given the following conditions on the 1 A l
NCP: i
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Time 0200 0210 0220 0230 :
Motor bearing temp (F ): 181 184 193 196 i
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Shaft vibration (mils): 2 3 4 5
Pump seal pressure AP (psig): 196 201 223. ?.35 i
Pump seal outlet temp (F ): 208 212 221 228 I
When are the operators required to trip NCP-1 A?
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A. 0200
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B. 0210 -
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C. 0220
D. 0230
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Question #4 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 243 Answer: D
1Pt(s) Unit 1 is conducting a plant stanup in mode 1. The operators have reached
12% power when a momentary electrical transient occurs resulting in the
following conditions:
Bus ITA ITB !TC ITD
Frequency (Hz) 55 60 55 60
Voltage (VAC) 6410 6900 6410 6900
Which one of the following sequences would occur?
A. A reactor trip does NOT occur and NCPs I A and IC trip on
under-frequency while NCPs IB and ID continue running.
B. A reactor trip occurs and NCPs I A and IC trip on under-voltage
while NCPs 1B and 1D continue running. )
C. A reactor trip does NOT occur but NCPs I A and IC do trip on
under-voltage while NCPs 1B and ID continue running.
D. A reactor trip occurs and all four NCPs trip on under-frequency
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Question #5 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 244 Answer: C
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.1Pt(s) Which one of the following precautions are required by procedure when
,
initiating auxiliary spray to prevent thermal shock? ;
-A. Auxiliary spray shall NOT be initiated when pressurizer vapor ,
space temperatureis LESS than 320 F
!
B. Auxiliary spray may be initiated after warming up the spray line
if the differential temperature is GRFATER than 320 *F between
pressurizer and NC system. l
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C. Auxiliary spray shall NOT be initiated when differential
temperature between the pressurizer and the NV system is ,
GREATER than 320 F. )
D. Auxiliary spray shall NOT be initiated when the differential ,
temperature between the pressurizer and the NV system is LESS
than 320 *F.
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Question #6 CATAWBA NUCLEAR STATION RO EXAM
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_' Bank Question: 245 Answer: A I
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1Pt(s) Given the following plant conditions on Unit 2:
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e Reactor poweris at 100%
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* All control systems and components are in automatic
* VCT level is at 40% with auto makeup m progress !
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If the channel i VCT level transmitter (NVLT-5761) fails high as indicated on
'
the Operator Aid Computer and NO operator action is taken, which one of the
following describes the response of ACTUAL VCT level?
A. Decrease to 0%, where charging pump cavitation causes loss of
charging
B. Decrease to 4.3%, where swap-over to the FWST suction occun .
C. Increase to 49.3%, and then cycle between 32.7% and 49.3%
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D. Increase to 85.3%, and then cycle between 70.1% and 85.3%
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Quesbon #7 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 246 Answer: C ,
1Pt(s) . Which one of the following conditions will automatically UNBLOCK both the
Unit 1 pressurizer low pressure and steam line low pressure safety injection
actuation signals?
A. T increases to 564 *F
B. Steam perssure increases to 614 psig ;
.
C. NC system pressure increases to 2015 psig j
D. Main steam line differential pressure increases to 100 psid
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For Official Use Only Ques 246a NRC EXAM 12/12/97
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Question #8 CATAWBA NUCLEAR STATION RO EXAM
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. Bank Question: 247 Answer: A
1 Pt(s) - Which one of the following Unit 2 safety injection actuation signals can be i
blocked below P-1I? l
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A. Pressuriurlow pressure ;
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B. Steam line low pressure J
C. High containment pressure j
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D. Pressurizer low pressure AND steam line low pressure
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For Official Use Only Ques 247a NRC EXAM 12/12/97
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Question #9 CATAWBA NUCLEAR STATION RO EXAM
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} Bank Question: 248 Answer: D
1Pt(s) The following plant conditions exist on Unit 1:
* A plant power increase is in progress at 85%.
. Channel I pressurizer pressure detector has previously failed LOW and has ,
been removed from service, including tripping all effected bistables. j
e Power range channel N-43 upper detector has just failed HIGH.~ l
Which one of the following trip function bistables associated with the N-43
Power Range channel failure will require a plant shutdown to Mode 3 to-
comply with Technical Specifications under these conditions?
A. Loss of Coolant or Steam Break Protection !
B. Steam Generator Lo-La Level
C. NC Loop Over-power Delta-T (OPDT)
.
D. NC Loop Over-temperaturr Delta-T (OTDT)
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For Official Use Only Ques 248a NRC EXAM 12/12/97
Question #10 CATAWBA NUCLEAR STATION RO EXAM
] Bank Question: 249 Answer: C
1Pt(s) Unit 2 is conducting a reactor startup in mode 2 with the following conditions:
* Reactor power is at 8x10-" amps and increasing.
. Operators are evaluating overlap between intermediate and source range.
* The instrument power fuses blow for the N-35 intermediate range channel.
Which one of the following actions are required by procedures?
A. Enter E-0 (Reactor Trip or Safety Injection) at step 1.
;
B. Insert rods to lower neutron flux level to a value within the source
range. Within one hour, repair N-35 or shutdown to mode 3.
C. Do not exceed 1.0x10- amps until N-35 has been restored. Verify
that P6 permissive status light is not fit within one hour.
D. Continue startup procedure but do not exceed 10% power until i
N-35 has been restored to operable status.
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For Official Use Only Ques 249a NRC EXAM 12/12/97
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l Question #11 CATAWBA NUCLEAR STATION RO EXAM l l .
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1 Bank Question: 250. Answer: A l
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1Pt(s) Which one of the following NC system penetrations is sensed to provide
RVLIS indication?
A. Reactor Vessel head vent
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B. loop B cold leg
C. Reactor vessel post accident system sample line (at the seal table) l
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D. Reactor vessel flange leakoffline :
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For Official Use Only Ques 250a NRC EXAM 12/12/97
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Question #12 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 251- Answer: D I
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1Pt(s) Unit 1 is responding to a loss ofoffsite power. Given the following events and :
conditions: -!
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. Both EDGs load nonnally on the 4160 VAC essential busses. '
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The "YV/RN Cool Water Ctrl" is in the " CTRL RM" position
;
.Which one of the following statements describes the containment cooling water
alignment under these conditions?
A. Cooling water is isolated until manual operator action is taken to :
reston the RN system.
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B. Cooling water auto-swaps from thc RN system to the YV system
C. Cooling water is4solated until manual operator action is taken to
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restom the YV system. 1
D. Cooling water auto-swaps from the YV system to the RN system
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Question #13 . CATAWBA NUCLEAR S /ATION RO EXAM t
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.,- Bank Question: 252 Answer:B
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iPt(s) Unit 1 is operating in mode 1 Which one of the following conditions are
- outside of administrative limits but within tech spec limits for containment '
pressure and temperature? i
P
Containment Aux. Building Upper CNT '
Pressure (osic) Pressure (osic) Temoerature ( F) l
A. -0.01 -0.05 95 ,
B. -0.09 -0.02 86
C. -0.10 -0.03 72 i
D. - -0.34 -0.25 85
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. Question #14 CATAWBA NUCLEAR STAT'9N RO EXAM i
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} Bank Question: 253 Answer: B
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i Pt(s) Given the following outputs / indications from the excore nuclear instmment
channels with power stable during a Unit I stanup: !
* N-41: 20 % ,
1
. N-42: 12 % (Upper detector failed LOW, power mismatch defeat 1
switch positioned to defeat the detector) )
. N-43 22 % !
. N-44: 24 % i
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Which one of the following is the power range value that is used to establish
the arbitrator signal for the steam flow circuit in the steam generator level and i
feedwater pump speed control system?
)
A. 18 % :
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B. 20 %
C. 22 %
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; D. 24 % )
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Question #15 CATAWBA NUCLEAR STATION RO EXAM
1 Bank Question: 254 Answer: B
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1 Pt(s) Which one of the following statements describes the operation of the ice
condenser lower inlet doors during a LOCA into containment?
A. The doors open when differential pressure exceeds I psid across
the doors.
B. The doon open when differential pressure exceeds I lb/ft' across
the doors.
C. The doors open following a 9 minute time delay from when the
Containment Air Return Fans (VX) start
D. The doon open by the weight of the cold air after the mechanical !
latch on the doors is released by a phase B isolation signal. ;
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For Official Use Only Ques 254a NRC EXAM 12/12/97
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QUESTION #16- CATAWBA NUCLEAR STATION RO EXAM
Bank Quesuon: 256 Answer: C
1 Pt(s) Unit I was operating at 100% power when en inadvertent reactor tr:p and
safetyinjection occurred. Given the following esents and conditions:
-
Phase A imlation occurred
.
The NF system glycol that was trapper: between the ireide and outside
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containment retum isolation va)ves expanded due to heating.
&
Which one of the foDowing statements describes the system response to the
glycci expansion.
:
A. The containment isolation valve disks an designed to rdieve
.
.
trapped glycolinto the hT containment header. - - : . ;
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B. A smallline with a check valve is installed on the auxiliary !
, building side of the penetration to relieve trapped glycolinto the '
NF supply header. I
C.
A smallline with a check valve is installed on the containment side -
of the penetration to relieve trapped glycol into the hT ;
containment header.
D. A relief valve is installed on the auxiliary bui! ding side of the
penetntion to relieve trapped glycolinto the NT dmin header,
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Question #17 CATAWBA NUCLEAR STAlION RO EXAM !
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.i ' Bank Question: 257' Answer: B
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1Pr(s) Unit 2 is operating at 100% power when a feedwater control valve for the B
S/G fhils full open. Assuming no operator action, what time would the turbine
trip on hi-hi S/G level?
- Time Channel 1 Channel 2 Channel 3 Channel 4 !
A. 0200 76 % 78 % 76 % 75 %
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B. 0205 76 % 80% . ,78% 76 %
C. 0210 79 % 84 % 82 % 80 %
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D. -0215 .82 % 86 % 84 % 82 % :
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For Official Use Only Ques 257a NRC EXAM 12/12/97
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Question #18 CATAWBA NUCLEAR STATION RO EXAM
a Bank Question: 258 Answer: D r
.
1Pt(s) Unit I was operating at 100% power with plant control systems in automatic.
Which one of the following conditions would cause the condensate load ;
rejection bypass valve (CM-83) to open automatically?
A. Feedwater pump suction pressure is reduced by 100 psig.
B. Turbine impulse pressure is reduced by 50 psig.
. C. Feedwater pump recirculation flow decreases to 200 gpm.
I
D. Condensate booster pump suction pressure decreases to 100 psig. !
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: For Official Use Only Ques 258a NRC EXAM 12/12/97
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Question #19 CATAWBA NUCLEAR STATION RO EXAM -
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l Bank Question: 259 Answer: D
1Pt(s) The operators are conducting a cooldown in Mode 3 on Unit 1. Given the
following plant conditions:
i
e Tave = 510 F :
* NC system pressure = 1900 psig >
e- I A Feedwater pump is maintaining S/G levels
. 1B Feedwater pump is tripped / shutdown '
. Pzrlow pressure Si has been blocked i
. Steamline low pressure SI has been blocked
i
e Both auxiliary feed (CA) train AUTO-START-DEFEAT buttons have !
'
been depressed
Which one of the following events will cause the motor-drivei. A pumps to
start automatically?
A. I A Feedwater pumps trips causing all S/G levels to decstase below l
, the Lo-Lo level trip setpoint. 1
B. NC loop "B" spray valve, NC-29, fails open causing NC pressure l
'
l to decrease to 1500 psig !
. . .
C. A steamline empture causes the "C" S/G pressure to decrease to
700 psig at a rate of 200 psig/second.
l D. A feedline sipture causes the "D" S/G level to decrease below the l Lo-Lo level trip setpoint and containment pressure to increase to l
2 psig.
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For Official Use Only Ques 2S9a NRC EXAM 12/12/97
4
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Question #20 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 260 Answer: C
1Pt(s) Unit 2 is responding to a loss of main feedwater event from 100% power.
Given the following conditions:
. The reactor has tripped
. The 2A and 28 Motor-driven CA pumps started in auto
. The Turbine-driven CA pump (CAPT) started in auto
e' Train A of"CA SYS VLV CTRL" has been reset
. The train B reset has failed to actuate
. The CA pumps are aligned to the hotwell
. CA suction pressure drops to 5 psig as hotwelllevel decreases
Which one of the following system responses (if any) will occur without
operator action.
A. 2A CA pump suction shifts to the RN system
2 CAPT suction remains aligned to the hotwell
2B CA pump suction shifts to the RN system
. B. 2A CA pump suction shifts to the RN system
) 2 CAPT suction remains aligned to the hotw ell
2B CA pump trips
C. 2A CA pump trips
2 CAPT suction shifts to the RN system
2B CA pump suction shifts to the RN system
D. 2A CA pump trips
2 CAPT suction shifts to the RN system
2B CA pump trips j
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For Official Use Only Ques 260a NRC EXAM 12/12/97
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Question #21 CATAWBA NUCLEAR STATION RO EXAM
*
1
Bank Question: 261 Answer:D
1Pt(s) Unit 1 is operating at 100% power. A liquid waste release has been approved
for the shift. Given the following conditions:
. lEMF-49 has been set in accordance with the LWR form.
. 1 RN and 2 RL pumps are required to be operating per the LWR form
. 1 RN and 3 RL pumps are currently operating
Which one of the following situations requires securing the liquid waste release
(and not immediately restarting it) after beginning the release?
A. The EMF-49 low sampie flow alarm actuates immediately upon
commencement of the discharge and clears within 20 seconds.
B. The EMF-49 "Hi Rad" spikes into alarm 10 minutes after the
release started.
C. One RL pump is secured by an NLO due to excessive bearing
noise.
, D. Local area flooding prompts the OSM to initiate a " Site
#
Assembly".
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For Official Use Only Ques 261a NRC EXAM 12/12/97
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Question #16 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 262 Answer: B
1 Pt(s) Which one of the following events will automatically terminate a waste gas
reJense?
A. EMF-53 (Containment Post LOCA Monitor) alarm
B. EMF-50 (WG Discharge Monitor) alarm
C. WG Compressor A tdp
D. WG system tdp with compressor lockout
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For Offical Use Only Ques 262a NRC EXAM 12/12/97
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Question #23
CATAWBA NUCLEAR STATION RO EXAM !
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Bank Question: 263 Answer: A ,
1 Pt(s)
Unit I is shutdown refueling with fuel movement in progress. Given the
following events and conditions:
e i
The new fuel elevator fails to operate in the up direction i
Which one ofthe following statements describes the cause of this proble!
A.
1 EMF-15 (Spent Fuel Building Refueling Bridge Monitor) has
failed high)
,
B. !
IEMF-20 (New Fuel Vault Moniton) has failed high !
C.
The load in the new fuel elevator weighs 1200 lbs
D.
The spent fuel bridge crane is NOT indexed over the new fuel
elevator
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For Official Use Only
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Ques 263a NRC EXAM 12/12/97
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Question #24 CATAWBA NUCLEAR STATION RO EXAM l
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Bank Question: 264 ' Answer: C
1Pt(s) Which one of the following conditions require the operators to reduce reactor
coolant system pressure within 5 minutes to comply with safety limits.
1
NCS Pressure NCS Temperature Mods i
~A. 2742 psig 605 'F 1
B. 2714 psig 550 'F 2
C. 2746 psig 415 'F 3
D. 2716 psig 345 'F 4
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For Official Use Only Ques 264a NRC EXAM 12/12/97
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Question #23 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 265 Answer: A
1Pt(s) Which one of the following conditions is likely to occur while shifting to cold
leg recirculation following a large break LOCA on Unit 2?
A. Radioactive water will be pumped to the FWST if the NI pumps
flow recirculation valves are not shut.
l
B. Radioactive water will be pumped to the FWST if the ND pump
minimum flow recirculation valves are not shut.
C. If both NI pumps are not running when ND is aligned to the NI
pump suction, the suction relief for the non-running N1 pump
may actuate and direct radioactive water to the FWST.
D. Radioactive water will backflow from the containment sump to
the FWST if the NV pump suction valve from the FWST is not
shut.
1
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For Official Use Only Ques 265a NRC EXAM 12/12/97
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Question #26 CATAWBA NUCLEAR STATION RO EXAM
'
Bank Question: 266 Answer: C I
j
iPt(s) Unit 1 is operating at 100% power, steady state when a PZR SURGE LINE
LO TEMP alarm annunciates. No other abnonnal annunciators are alarming.
Which one of the following statements is the most likely exp!anation for this ;
alarm?
A. Pressurizer temperature has slowly decreased due to pressurizer l
heaters being off.
,
B. Small insurge/outsury,e cycles are occurring due to corr xenon ,
oscillations. ;
J
C. Spray valve bypass now has stopped due to orifice fouling ,
problems. l
D. L has deceased due to inadvertent boration flow being ;
undetected. l
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For Official Use Only Ques 266a NRC EXAM 12/12/97
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QUESTION #27 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 267 Answer: B
1 Pt(s) Unit I was operating at 100% power. Given the following events and
conditions:
,
e Complete loss ofc5 site power occurred
.
Only one D/G was available and powering the essential bus
.
.
A rapid decrease in steam line pressure was caused by the addition of cold
CA water
.
A safety injection irdtiation caused the pressurizer to go water solid.'
'
Which one ofthe following statements describes the required plant pressure
control method when the pressurizer is solid following the terrrdnation ofsafety
. s -
. .
injection? -:.
.
.j ,. . - -
A. Pre:sure is increased by manually energizing PZR heaters and
a decreased by opening a letdown line. '
B. Pressure is increased by anowing decay heat to increase
temperature and decreased by allowing NC PORVs to cycle
automaticauy. !
C. Pressure is increased by manually enesti zing PZR heaters and
decreased by opening reactor vessel head vents.
, D. Pressure is controHed by immediately drawing a bubble in the
pressurizer and reestablishing nonnal pressure control.
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Ques 267a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #28 CATAWBA NUCLEAR STATION RO EXAM
;
Bank Question: 268 Answer: A l
l
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1Pt(s) Unit I was operating at 100% power when a LOCA into containment
occurred. Given the following conditions:
* Pressurizer pressure is stable at 1350 psig
. Containment temperature is 155 'F l
. Actual pressurizer levelis 25%
1
Select the combination of answers below that correctly fill in the blanks ,
concerning the effects of these conditions on the pressurizer level indicated on i
channel 3. I
i
'
The environment in the pressurizer causes indicated pressurizer level to
- read (X) than actuallevel. The environment in containment
makes indicated prusurizer level read (Y) than actuallevel.
(X) (Y)
A. higher higher
,
, B. higher lower
-
1
C. lower higher ]
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D. ' lower lower
,
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l l l For Official Use Only Ques 268a NRC EXAM 12/12/97 i
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Question'#29 CATAWBA NUCLEAR STATION RO EXAM l
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Bank Question: 269 Answer: A
\
\
lPt(s) Unit 2 is operating at 10G% power. I&E is conducting a surveillance test on
the A reactor trip breaker. Given the following conditions: .
!
. SSPS Train"A"is in test. l
e l
Reactor Trip Breaker (RTA)is open i
e Bypass Breaker (RYA)is shut
If a subsequent loss of power occurs to SSPS Train "A", which one of the
following statements describes the effect that this will have on the reactor
protective system?
i
A. There will be no change, the General Warning light will remain lit
on SSPS Train "A".
B. A General Warning light will only be lit on SSPS Train "A" after
poweris lost.
C. RYA will open but the reactor will not trip when power is lost.
t
*
D. A reactor trip will occur when power is lost.
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i for Official Use Only Ques 269a NRC EXAM 12/12/97 l l ' . l ! !
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Question #30 CATAWBA NUCLEAR STATION RO EXAM
'
Bank Question: 270 Answer: D
1Pt(s) Unit 1 is operating at 100% power. Given the following indications on the
Digital Rod Position Indication system:
. General warning light for rod D-4 is flashing
. RPI URGENT annunciatoris alarming
. Urgent alarms 1,2 and 3 are flashing
. Rod bottom LED for rod D-4 is lit
Which one of the following describes the condition of rod D-4?
A. Rod D-4 DRPIindication is at half accuracy
B. Rod D-4 DRPI indication is at full accuracy
C. Rod D-4 DRPI indication is valid and the rod is fully inserted
1
D. Rod D-4 position cannot be determined by DRPI
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!' For Official Use Only Ques 270a NRC EXAM 1'J12/97
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l Quesbon #31- CATAWBA NUCLEAR STA110N RO EXAM
1. Bank Question: 271 Answer:B
1 Pt(s) Which one of the following NC system instruments are averaged together to
obtain a single output?
A. Cold leg resistance temperature detecton (RTDs)
B. Hot leg RTDs
l
C. Loop pressure detecton l
D. Loop flow detector high pressure taps j
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For Official Use Only Ques 271a NRC EXAM 12/12/97
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Quesbon #32 CATAWBA NUCLEAR STATION RO EXAM
l
1^
Bank Question: 272 Answer: D
1Pt(s) Unit 2 is responding to a LOCA from a trip at full power. Given the following
conditions:
* A safetyinjection has occurred
* Train B Sp signal failed to actuate
* The 2A NS pump started automatically but the 2B NS pumps was staned i
manually by an operator l
* The Ss signal and sequencer have been reset . l
+ The train A Sp signal has not teen reset !
* Both pumps were stopped for ihifting suctions to the containment sump
If contamment pressure is 0.25 psig, which one of the following statements
describes the operation of the NS pumps upon completion of the swapover? )
A.- Both NS pumps will restart automatically if containment pressure ;
increases above 3.0 psig. l
B. Both NS pumps will restart automatically when their respective )
sump suction valve (NS-ISA, NS-1 B) reaches full open.
.~} 1
C. When the sump suction valves (NS18A, NS-1B) reach full open, I
NS pump 2A will restart automatically and the operator can start
MS pump 2B manually.
'
D. When the sump suction valves (NSl8A, NS-1B) reach full open,
the NS pumps will not restart automatically or manually. l
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For Official Use Only Ques 272a NRC EXAM 12/12/97 .
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Quesbon #33 CATAWBA NUCLEAR STATION RO EXAM l
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;
u __ l Bank Question: 273 Answer: C
1
i
1Pt(s) Unit 1 is shutdown in mode 5. Given the following plant conditions:
l
* The VP systemisin operation
* Both trains ofSSPS arein TEST l
Which one of the following signals will shutdown the VP system? I
A. S signal
l
B. EMF-38 trip 2 signal
C. EMF-39 trip 2 signal
D. EMF-40 trip 2 signal
.
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- For Official Use Only Ques 273a NRC EXAM 12/12/97
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Question #34 CATAWBA NUCLEAR STATION RO EXAM
4
;
,
.I Bank Question: 274 Answer: D
1Pt(s) What feature in the KF pump discharge line into the spent fuel pool prevents
draining down the water and uncovering fuel assemblies in the event of a break
in the dischargeline?
,
A. The discharge line penetration of the spent fuel pool is just below j
the surface of the water and would uncover in the event of a
rupture.
B. The discharge line has a reverse flow check valve that will prevent
!
spent fuel pool water from back-flowing through the line.
l C. The discherve line is sized to ensure that manual isolation could ,
be accomplished prior to draining enough water to uncover the
'
tops of the fuel assemblies
j D. The discharge line has holes below the water line to prevent
draining the water out of the spent fuel pool.
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For Official Use Only Ques 274a NRC EXAM 12/12/97
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Question #35 CATAWBA NUCLEAR STAEON RO EXAM
l
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j Bank Question: 275 Answer: A
!
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1Pt(s)- Which one of the following describes the flow path of the feedwater into the j
Unit 1 Steam Generator main nozzles at 5% power? . '
A. 100% Sow directed into the auxiliary feed nozzle with a small ;
'
amount of How being drawn out of the main feed ring and
returned to the main condenser to maintain CF containment
penetration temperature.
B. 100% How directed into the main feed ring with a small amount of
tempering How into the auxiliary feed nozzle to cool the CA nozzle.
.
' i
.C. 100% Gow directed through the main feed ring with Row directed j
l ' downwarsi then back upward through the preheater counter-How
l section.
j. D. The feedwater flow is split with 15% of the flow is directed into i
the auxiliary feed nozzle while 85% of the flow is directed into the !
main feed ring.
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For Official Use Only Ques 275a NRC EXAM 12/12/97
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1
Question #36 CATAWBA NUCLEAR STATION RO EXAM !
;
Bank Question: 276 Answer: C
i
1Pt(s) Unit 1 is responding to a faulted steam generator inside containment. Given
the following conditions:
* Contamment pressure = 4 psig
* NC system pressure = 1742 psig
*
MSIVs closed on a low steam liae pressure signal
Which one of the fo'Jowing actions will allow main steam isolation valves to
reopen.
1
l
A. Reset both trains of main steam line isolation. I
B. Reset Phase B signal and both trains of main steam line isolation.
C. Block low steam line pressure signal and reset both trains of main
steam line isolation.
D. Reset safety injection and reset both trains of main steam line
isolation. l
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L For Official Use Only Ques 276a NRC EXAM 12/12/97 l
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Question #37 CATAWBA NUCLEAR STATION RO EXAM
. ,.
* Bank Question: 277 Answer: C
1Pt(s) Which one of the following fhilure conditions will cause a lock out relay
actuation for the incoming 6.9KV breaker from IT2A to 1TA?
,
A. Under-frequency
B. Under-vokage
C. Ground fault .
D. Loss of control power ,
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For Omcial Use Only
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Ques 277a NRC EXAM 12/12/97 .
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Question #38 CATAWBA NUCLEAR STATION RO EXAM
!
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Bank Question: 278 Answer: D l
l
1Pt(s)- Unit 1 is operating at 100% power when the following conditions occurred.
. The l A EDG battery was disconnected and removed from service
Which one of the following statements correctly describes the status of the 1 A
EDG?
A. IA EDG is operable because IEDE is energized through 1EADA ;
(VitalI&C)
l
B. I A EDG is operable because the battery charger is designed to
supply the bus without a battery.
C. I A EDG will be inoperable in modes 1-4 but will be operable and
available in modes 5 and 6
D. The 1 A EDG will be inoperable.
,
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For Official Use Only , Ques 278a NRC EXAM 12/12/97
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QUESTION #39 CATAWBA NUCLEAR STATION RO EXAM
i
Bank Question: 279 Answer: D
1Pt(s) )
Unit 1 is operating in mode 3 shuttir.g down for refueling. A surveiDance is in
!
progress on the 1 A EDG Given the following plant conditions-
=
1 A EDG is operating in parallel with off-site power
.
A safety injection signal is receiwxi
__
Which one of the following events will occur?
j
A. The diesel generator breaker will :unain cicsed, unless an under- I
j
voltage condition occun, and non-LOCA loads on IETA will be
tripped.
-
, , -
-
,
B. The diesel generator breaker will remain closed ind load shed of
IETA will occur. The LOCA sequencer will sequenceloads on.
o C. The diesel generator breaker will trip open and a load shed of
IETA will occur. The LOCA sequencer will sequence loads on.
!
D. The diesel generator breaker will trip open and non-LOCA loads
on IETA will be tripped. The LOCA sequencer will . ' quence
!
loads on.
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Ques 279a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #40 CATAWBA NUCLEAR STATION RO EXAM
L
, ,;) Bank Question: 280 Answer: C
.1Pt(s) Which one of the following conditions was EMF-48 (NC monitor) designed to
detect?
A. High N16 gamma activity
B. High coolant gaseous activity
C. Fuel cladding failure
'
D. Crud burst
1
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for Official Use Only Ques 280a NRC EXAM 12/12/97
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' Question #41 CATAWBA NUCLEAR STATION RO EXAM -
l
- . .
} Bank Question: 281 Answer: C
1Pt(s) Which one of the following supplies power to the Circulating Cooling Water
(RC) Pumps?
'A. 600 VAC Unit Power System
B. 4160 VAC Essential Power System
C. 6.9 KVAC Unit Auxiliary Power
D, 13.8 KVAC Normal Auxiliary Power
:
.
:
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l l l For Official Use Only Ques 281a NRC EXAM 12/12/97 l
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Question #42 CATAWBA NUCLEAR STATION RO EXAM l
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Bank Question: 282 Answer: C ;
1Pt(s) Which one of the following statements describes the VI system response to a
loss ofheader pressure?
i
A. ' The Backup Temporary / Diesel Air Compressor auto starts I
80 psig- VS-78 (VS supply to VI) opens )
76 psig- VI-500 (VI supply to VS) closes
B. The Standby Air Compressor auto starts ,
80 psig- VS-78 (VS supply to VI) opens
l
C. The Standby Air Compressor auto starts )
80 psig - VI-500 (VI supply to VS) closes )
76 psig- VS-78 (VS supply to VI) opens j
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D. 80 psig- VS-78 (VS supply to VI) opens
'
76 psig - VI-500 (VI supply to VS) closes
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For Official Use Only Ques 282a NRC EXAM 12/12/97
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Question #43 CATAWBA NUCLEAR STATION RO EXAM
1
Bank Question: 283 Answer: B
1 Pt(s) A fire has occurred in the plant. Fire header pressure drops in response to
demands for water, Given the following conditions:
* The auto-start pressure switch to the "A" fire pump located in the Sersice
Building malfunctions and fails to actuate. '
. The "B" fire pump is tagged out of service for maintenance
l * . The 4160 blackout switchgear 2FTA was deenergized by a ground fault
. Pressure continues to drop in the fire header
l Which one of the following statements describes the correct starting sequence
for a fire pump to pressurize the fire header? !
!
! A. The "A" fire pump auto-starts at 92 psig
i
B. The "A" fire pump auto-starts at 70 psig
C. The "C" fire pump auto-starts at 92 psig
D. The "A" fire pump must be manually started by an operator
.
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For Official Use Only Ques 283a NRC EXAM 12/12/97
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Question #44 CATAWBA NUCLEAR STATION RO EXAM
} Bank Question: 284 Answer: A
1Pt(s) Unit 1 is shutdown in mode 5. Given the following conditions:
. Both trains ofND are operable
. ND train"A" is in operation
. All four NC loops are filled
.
Steam Generator Levels are as follows:
!
S/G A S/G B S/G C S/G D
0% 15 % 15 % 10 %
Mechanical maintenance requests permission to deenergize the IETB 4160V
, bus for a work order. What action (s)(if any) are required to allow I maintenance to proceed? ,
A. Maintenance may proceed without further changes in plant
I
conditions.
B. Raise the ID S/G level to 15%
! l C. Start IB or IC NC pumps
"
/
D. Maintenance cannot deenergize 1ETB under these conditions.
! l l l l , !
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for Official Use Only Ques 284a NRC EXAM 12/12/97
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Question #45 CATAWBA NUCLEAR STATION RO EXAM
l
*
1 Bank Question: 285 Answer: C
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1Pt(s) Unit 1 is responding to a LOCA inside containment. Given the following
conditions:
. Reactor trip and safety injection actuated
;
e Containment pressure peaked at 2 psig !
* FWST level dropped to 36%
. EMF 46 A trip 2 alarmed
. Low-low level alarm on KC surge tank A
. The operators remain in the control room (control not transferred to the
ASP)
Which one of the following KC system loads will still have KC flow under
these circumstances?
A. NC pump thermal barrier heat exchangers
l
B. NCDT Heat Exchanger
!
C. ND heat exchanger i
D. Letdown heat exchanger
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- For Official Use Only Ques 285a NRC EXAM 12/12/97
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Question #46 CATAWBA NUCLEAR STATION RO EXAM
l
' Bank Question:286 Answer: D
,
1Pt(s) Unit 2 is in mode 2 conducting a reactor startup. Given the following
conditions-
;
*
Source range level = 3X10' CPS
*
Shutdown banks are withdrawn
*
The operators have started pulling the control banks
Which one of the following source range levels should be set for the high flux
at shutdown alarm setpoint?
A. 6.1X10' CPS
B. 8.0X10' CPS
C. 9.3X10' CPS
D. Alarm should be blocked
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For Official Use Only Ques 286a NRC EXAM 12/12/97
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Question #47 CATAWBA NUCLEAR STATION RO EXAM
l Bank Question: 287 Answer: D
1 Pt(s) Unit I was operating at 80% power when a reactor trip occurred. Given the i
following conditions: ,
. Reactor trip breaker "A" will NOT open
. Turbine impulse channel II has failed at 80% (as is)
. The Steam Dump Mode Select Switch is in the Tave position
Which one of the following combinations states the response of the Steam
Dump Control System dump valves to these events?
Atmospheric Condenser
Dumps Dumps
A. Opened Opened
B. NOT Opened Opened
C. Opened NOT Opened
'I D. NOT Opened NOT Opened !
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For Official Use Only Ques 287a NRC EXAM 12/12/97
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Question #48 CATAWBA NUCLEAR STATION RO EXAM
J Bank Question: 288 Answer: A
1Pt(s) Unit 1 is operating at 9% power and preparing to increase load after a startup.
Which one of the following conditions or signals will cause a main turbine trip?
A. lA S/G level = 84%
B. Pressurizer perssurr = 1940 psig
C. Pressurizer level = 95%
D. Loss of the l A and IC-NCPs.
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For Official Use Only Ques 288a NRC EXAM 12/12/97
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j Question #49 CATAWBA NUCLEAR STATION RO EXAM j '
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.. l^ Bank Question: 289 Answer: B
1Pt(s) Which one of the following conditions will cause the RN system crossover
l
isolation valves (1RN54A and 1RN53B) to automatically close? j
A. Low RN flow at the outlet of the KC Heat Exchanger (1,895 gpm) i
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B. 14w level at the RN Pump Intake Pit (552 ft) )
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C. Hi differential pressure on the RN Screen (15" water)- l
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D. Low RN Essential Header Pressure (40 psig)
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For Official Use Only Ques 289a NRC EXAM 12/12/97
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Question #50 CATAWBA NUCLEAR STATION RO EXAM
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,1 Bank Question: 290 Answer: B
1Pt(s) Electrical Maintenance requests permission to deenergize Shared Load Center
2SLXC for a breaker repair work order. Which one of the following sets of air
compressors willlose power?
A. The "D" VI Compressor and the "B" VB Compressor
B. The "D" Vi Compressor and the "B" VS Compressor
C. The "E" VI Compressor and the ""B" VB Compressor
D. The "E" VI Compressor and the "B" VS Compressor
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For Official Use Only Ques 290a NRC EXAM 12/12/97 !
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Question #51 CATAWBA NUCLEAR STATION RO EXAM
1
Bank Question: 291 Answer: B
1Pt(s) Unit 1 is responding to a faulted steam generator inside containment. Given
the following plant conditions:
. Containment pressure = 3.5 psig
. EMF-38L trip 2 alarm
. Main steamline pressure = 605 psig
Which one of the following safeguards actuation signals can be reset
immediately after initiation under these conditions?
A. Safety injection (S.)
B. Phase B isolation (S,,)
C. Containment Ventilation Isolation (Sn)
D. Main Steam Line Isolation
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For Official Use Only Ques 291a NRC EXAM 12/12/97
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Question #52 CATAWBA NUCLEAR STATION RO EXAM
i l !
l Bank Question: 292 Answer: D
1Pt(s) Which one of the following events will most likely cause the operators to
implement FR-P.1 (Response to Imminent Pressurized Thermal Shock
Condition) to mitigate an actual PTS challenge during the first 10 minutes of
the event?
A. Excessive CA flow while shutdown in mode 3
i
)
B.. Steam generator tube rupture in mode 1
C. Design basis lage break LOCA in mode i
D. Main Steam line rupture in mode I
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For Official Use Only Ques 292a NRC EXAM 12/12/97
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Question #53 CATAWBA NUCLEAR STATION RO EXAM
; Bank Question: 293 Answer: C
1Pt(s) Unit 1 is conducting a reactor startup when the following events occurred:
. Reactor was subcritical with power stable at 1.2X10' CPS
. While withdrawing Control Bank B, one rod dropped to the bottom of the
core
Which one of the following statements describes the correct actions?
A. Secure pulling control bank B. Ensure that the reactor remains
subcritical while recovering the dropped rod.
B. P.fanually insert control bank B to the bottom of the core and
recover the dropped rod. Ensure that the reactor remains
subcritical while recovering the dropped rod.
C. Manually insert all control rods.
D. Manually trip the reactor.
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For Official Use Only Ques 293a NRC EXAM 12/12/97
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Question #54 CATAWBA NUCLEAR STATION RO EXAM
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,' Bank Question: 294 Answer: A
1Pt(s) Unit I was operating at 100% power when a LOCA occurred into
]
containment. Given the following events: 1
. Orange Path on Containment Integrity )
. The Operators have implemented FR-Z.1 (Response to High Containment
Pressure)
Which one of the following components will continue to receive KC system
cooling?
A. NV Pump Motor Coolen
B. NCDT Heat Exchanger
C. Excess Letdown Heat Exchanger i
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D. Reactor Coolant Pump Motor Oil Coolen -
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For Official Use Only Ques 294a NRC EXAM 12/12/97
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- Question #55 CATAWBA NUCLEAR STABON RO EXAM
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1 Bank Question: 295 Answer: A
1Pt(s)- Unit I was in the process of starting up at 10% power when a loss ofoffsite
power occurred.. Given the following conditions:
* The reactor tripped at 0200
. Dudng the accident, a pressurizer PORV stuck open.
*. An operator closed the PORV block valve
. A safety injection occurred on low pressurizer pressure
* ' The operators implemented E-0 at 0200
* The operators transitioned to ES-0.2 at 0215
. The cooldown was started at 0230 in ES-0.2 with NC system temperature
indicating 517 F.
. The STA reports that there is no void in the reactor vessel head
What is the initial average cooldown rate that the operators must maintain to
remain within procedurallimits? l
A. 20 F/hr
B. 50 F/hr ,
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C. 60 F/hr
D. 100 F/hr
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For Omcial Use Only Ques 295a NRC EXAM 12/12/97
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Question #56 CATAWBA NUCLEAR STATION RO EXAM
i Bank Question: 296 Answer: D
1Pt(s) Unit 1 is operating at 100% power when a large rupture of the KC Essential
Header 1 A occurs. Given the following conditions:
. KC Pumps l Al and 1 A2 are running
. The KC system trains are cross-connected
. All KC supply and retum isolation valves are open
l
Assuming no operating action is taken, which one of the following sequences
will occur automatically?
I
A. The ND heat exchanger inlet valve closes if the train-related KC
surge tank levels decrease to 34%
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B. The KC essential headers will isolate if their train-related KC
surge tank levels decrease to 37%.
C. The KC non-essential headers will isolate if FWST level decreases
to 37%.
j D. The KC non-essential headers will isolate if their tiain-related KC
'~'
surge tank levels decrease to 34%.
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For Official Use Only Ques 296a NRC EXAM 12/12/97
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Question #57 CATAWBA NUCLEAR STATION RO EXAM l
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..; Bank Question: 297 Answer: B
i
1 Pt(s) Unit 2 emergency boration is manually initiated in the plant locally from which !
one of the following locations? l
;
A. 2B NI Pump Room (Rm #244)
B. Unit 2 543' Mechanical Penetration Room (Rm #227)
C. 2B NV Pump Room (Rm #241) j
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D. Unit 2 577' Mechanical Penetration Room (Rm #427)
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For Official Use Only Ques 297a NRC EXAM 12/12/97
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Question #58 CATAWBA NUCLEAR STATION RO EXAM
O
Bank Question: 298 Answer: A
IPt(s) Unit I was operating at 100% power. Given the following conditions:
*
Pressurizer pressure controller is selected to "2&3"
.
Pressurizer pressure controls are in AUTO
+
Pressurizer pressure channel III detector fhils LOW
Which one of the following describes the plant response with no operator
action?
A. High pressurizer pressure reactor trip will occur.
B.
PORY NC-34A will maintain NC system pressure 80 to 100 psig
above normal.
C. No effect on NC system pressure but PORVs NC-32B and
NC-36B will be blocked.
D.
PORY NC-34A will maintain NCS pressure from 100 psig above
__ nonnal to 50 psig below normal.
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For Official Use Only Ques 298a NRC EXAM 12/12/97
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QUESTION #59 CATAWBA NUCLEAR STAHON RO EXAM
Bank Question: 299 Answer: A
1 Pt(s) Unit 1 is operating at 50% power. Given the following conditions:
7:me Main Condenser Vacuum
0200 25in Hg
0205 24in Hg
,,
0210 22in Hg
0215 21 in Hg
What action (s) (if any) are the operators directed to take under these ;
conditions? I
A. Manually trip the main turbine at 0210.
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B.
Wait until 0215 when condenservacuum has decreased below the
, automatic turbine trip setpoint before performing a manual
turbine trip.
C. Wait until 0215 when condenser vacuum has decreased below the I
automatic turbine trip setpoint before performing a manual
reactor trip and verifying turbine trip.
D. Do NOT manually trip the turbine at this powerlevel unless
turbine exhaust hood pressure has exceeded 225 'F.
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Ques 299a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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QUESTION #60 CATAWBA NUCLEAR STATION RO EXAM
Bank Question:300 Answer: A
1 Pt(s) Unit I was responding to a steamline break inside containment en the IC S/G
per E-2 (Faulted Steam Generator).
Which one of the following actions statements correctly describes the expected
method for isolating steam to the CAPT assuming no component failures.
;
-. 1
A. Manually close the maintenance isolation valve (ISA-4) in the
doghouse. !
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B. Manually close the stop-check valve (ISA-6)in the mechanical
penetration room.
_, ._ _. ;_
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C. Manually close the IC MSIV and bypass valve. = ~- ~
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c D. Manually actuate SM isolation.
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- '1 Ques 3003 NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #61 CATAWBA NUCLEAR STATION RO EXAM l
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) Bank Question: 301 Answer: A '
1Pt(s) During step 23 of ECA-0.0. the operators are directed to depressurize intact
S/Gs to 165 psig. What prevents exceeding the FR-P.1 criteria for pressurized
thermal shock (PTS) during this cooldown?
A. The S/G depressurization is stopped at a point where NC system
temperature will be maintained above the minimum temperature ,
for PTS. !
B. The natural circulation cooldown rate is limited by the S/G l
depressurization rate to a rate which is not a PTS conceni.
C. The NC System will be depressurized below the minimum
pressure where PTS is no longer of concern.
D. The natural circulation cooldown rate is much lower than forced
cirrulation cooldown rates however, if the criteria for FR-P.1 is
exceeded, this procedure will be implemented to provide direction
to mitigate the impact of PTS.
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For Official Use Only Ques 301a NRC EXAM 12/12/97 ;
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Question #62 CATAWBA NUCLEAR STATION RO EXAM
'
Bank Question: 302 Answer: D
1Pt(s) Unit 2 is operating at 100% powe Given the following conditions:
* Power is lost to 120 VAC panel board 2ERPB.
Which one of the following events would occur?
A. Reactor trip on power range high flux
B. "B" steam generatorlevel decreases to the Lo-Lo trip point
C. N-42 fails high causing rods to step in initially
D. Over power rod stop would prevent outward rod motion if i
attempted.
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For Official Use Only Ques 302a NRC EXAM 12/12/97
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Quesiion #63 CATAWBA NUCLEAR STATION RO EXAM
; Bank Question: 303 Answer: C
1Pt(s) Unit 1 is operating at 100% steady state power. Given the following i
conditions.
j
1
. KC pumps 1 Al and 1 A2 are mnning
. KC pumps IBl and IB2 arein standby -
. EMF-46A (KC Surge Tank Rad Monitor) is in alarm
. KC surge tank levels indicate 76% and trending up slowly
Which of the following components is the most likely source of the leak? !
i
A. Seal water heat exchanger j
B. NC het leg sample heat exchanger
C. NV letdown heat exchanger
D. NCP motor cooler
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For Official Use Only Ques 303a NRC EXAM 12/12/97
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Question #64 CATAWBA NUCLEAR STATION RO EXAM
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-1 Bank Question: 304 Answer: C
1Pt(s) Unit 1 is shutdown in a refueling outage. A small 6 rein the SSF causes the
activation of the automatic fire suppression system in the vicinity of the fire.
This action immediately extinguishes the fire. Which one of the following
statements describes the FIRST fire protection alarm response to this event.
A. The control room will first receive a fire protection alarm aP.er the
main fire pump starts.
B. The control room will first receive a fire protection alarm when
the RFY pressurizing tank pressure decreases to 128 psig
admitting Nitrogen to the tank. !
C. The control room will fint receive a fire protection alarm as soon
as the fusible link melts in the sprinkler and flow begins. I
D. The control room will fint receive a fire protection alarm when
the SSF CARDOX system actuates.
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For Official Use Only Ques 304a NRC EXAM 12/12/97
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' Quesbon #65 CATAWBA NUCLEAR STATION RO EXAM
.
) Bank Question: 305 Answer: D
1Pt(s) How is assured makeup for the NW System provided during a LOCA?
A. YM is automatically provided on a low-low suge tank level
coincident with a phase A isolation.
B. YM is automatically provided on a low-low suge tank level or
low-low suge tank pressure coincident with a phase A isolation.
C. RL is automatically provided on a low-low suge tank level ;
coincident with a phase A isolation.
D. RN is automatically provided on a low-low suge tank level or low- ;
low surge tank pressure coincident with a phase A isolation.
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For Official Use Only. I
Ques 305a NRC EXAM 12/12/97
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Question #66 CATAWBA NUCLEAR STATION RO EXAM
:
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Bank Question: 306 Answer: C
I Pt(s) Unit 1 is responding to a Station Blackout that lasted for one hour. The TSC i
is not yet operational. Given the following conditions: l
. Both plasma display control panels for the inadequate core cooling monitor
are out of se vice and core exit thermocouple data is not available l
. There is evidence of a LOCA into the containment through NCP seals l
'
. Power has been restored to one 4160 VAC safety bus
. The operators have transitioned to ECA-0.1 and are conducting a
cooldown in natural circulation (due to loss of NCP seals).
. The STA is reviewing critical safety function status trees in F-0
. The STA determines that NC system subcooling margin is -2 *F
(superheated) based on a comparison of:
. The highest b indication
. Saturation temperature for wide range NC system pressure -
determined by using steam tables
. Source range NIs indicate an increasing trend - with some fluctuations
. Based on this determination ofloss of subcooling and indications of
possible loss of reactor vessel level, the STA recommends transitioning to i
FR-C. I .
..-
,-
)
i
Which one of the following actions is correct for this situation?
REFERENCES ATTACHED
A. Immediately transition to FR-C.I. Although core exit ,
thermocouple data is not available, the indications ofloss of l
subcooling with loss of core level is sufficient.
B. Do not transition to FR-C.I. Without having indication of reactor
vessel level or core exit thennocouples, no transition can be made. !
Instead, reference actions in FR-C.3 while continuing on in ECA-
0.1.
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C. Do not transition to FR-C.I. Recalculate the subcooling using the
data book curves-steam tables cannot be used for this j
calculation. I
D. Do not transition to FR-C.I. Recalculate the subcooling margin
by using the average of all four L instruments. This is more
-
representative of actual core conditions. J l
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' For Official Use Only. Ques 306a NRC EXAM 12/12/97
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Question #67 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 307 Answer: B '
i
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1Pt(s) Unit I was operating at 100% power. Given the following events and i
conditions:
]I
e EMF-48 (NC monitor) trip 2 alarm j
e Activity level of 2 X 10 pCi/ gram (Dose Equivalent 1"') in the reactor <
coolant
l
Which one of the following actions are required to correct a high fission !
product activity? l
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A. Purge the VCT with nitrogen l
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B. Place mixed bed demineralizers in service
C. Reduce reactor power below 50%
D. Add hydrogen to the reactor coolant
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For Official Use Only Ques 307a NRC EXAM 12/12/97 l
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Question #68 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 308 Answer: D
<
1Pt(s) Unit 1 is operating at 100% power. Given the following conditions:
* Rod controlis in manual
* Control Bank D is at 200 steps
If the rods in control bank D stan stepping out at 8 steps per minute, what one
of the following actions is required at this time?
!
A. Select Control Bank D on the rod selector switch and manuallv !
insert Control Bank D
'
B. Select " AUTO" on the Bank Select Switch and see if rud motion
stops ;
C. Commence emergency boration
D. Trip the reactor
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For Official Use Only Ques 308a NRC EXAM 12/12/97
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Queston #69 CATAWBA NUCLEAR STATION RO EXAM l
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.,,? Bank Question: 309 Answer: C
1Pt(s) Unit 1 is operating at 100% power. Given the following conditions:
* Rod controlisin automatic
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. Control Bank- D is at 200 steps
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* All other banks are fully withdrawn j
. 'Tm and T,. rare matched *
Which one of the following situations would likely cause a dropped rod?
A.- Repeatedly cycling the in-hold-out switch without pausing 2
seconds between switch movements.
B. Loss of power fmm one DC power supply
C. A malfunction that tdggen a logic error
D. A malfunction that tdggen a multiplexer error
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l For Official Use Only Ques 309a NRC EXAM 12/12/97 i
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Question #70 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 310 Answer: D
1
1Pt(s) A reactor trip has occurred and the crew is in ES-0.1, Reactor Trip Response.
Given the following conditions:
.
A RED path on the CSF for Heat Sink has been recognized on the OAC
and validated by the STA
. The OSM directs the implementation of FR-H.1 (Response to Loss of
Secondary Heat Sink)
. The Heat Sink CSF NOW TURNS GREEN before the procedure has been !
removed from the case. l
Which one of the following actions is correct for these circumstances?
{
)
A. The crew should immediately transition to FR-H.1 and consult I
with the TSC or EOF to evaluate applicable actions and transition
back to ES-0.1.
B. The crew should immediately transition to FR-H.1 because once a
RED path has been validated, the associated functional response
'
procedure must be implemented to completion.
?
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C.
'
The crew should continue in ES-0.1 because guidance to
implement the status trees has not yet been reached in ES-0.1. !
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D. The crew should continue in ES-0.1 because the CSF procedure is
not considered implemented until the first step is read.
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For Official Use Only Ques 310a NRC EXAM 12/12/97
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Quesbon #71 CATAWBA NUCLEAR STATION RO EXAM
i
Bank Question: 311 Answer: B
1Pt(s) Unit I is operating at 50% power. Given the following conditions:
. Pressurizer pressureis 2235 psig
* Pressurizer ReliefTank (PRT) pressure is 25 psig
. PRT temperature is 115 'F
e PRTlevelis 81%
* The PRT is being cooled by spraying from the RMWST
e A pressurizer code safety valve is suspected ofleaking by its' seat
What temperature would be indicated on the associated safety valve discharge
RTD if the code safety was leaking by?
REFERENCES PROVIDED
A. 283 *F
B. 267 F
'
C. 239 F
,
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D. 195 F
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- For Official Use Only Ques 311a NRC EXAM 12/12/97 -
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QUESTION #72 CATAWBA NUCLEAR STATION RO EXAM
i
Bank Question: 312 Answer: C '
-
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1Pt(s) Unit 1 is responding to a small break LOCA inside comainment. Given the
following conditions:
* NI pump 1A isin service
. NI pump 1B has failed to start '
.. .
Both plasma displays on the inadequate core cooling monitor have failed
.
The operators have implemented E-1
NC Loon A NC Loco B NC Looo C NC Looo D
Tw (*F) 579 575 575 579 '
T u (*F) 574 569 565
-$ 569 l
y , , WR Press (psig) 1575 1615 f._:-
Wriich one ofthe following statements correctly describes the NCPs?
o
REFERENCES PROVIDED
i
A. NCPs should be tripped in order to prevent adding unwanted
heat to the coolant and causing core uncovery due to voiding in
the reactor vessel head.
!
B. NCPs should be tripped to prevent continued mass depletion of !
coolant that is being pumped out the break with potential serious
.
core uncovery if the pumps should later trip.
C. NCPs should NOT be tripped because subcooling does not meet
the foldout page criteria and the pumps are still providing
effective core cooling due to high cort steam flowrate.
D. . NCPs should NOT be tripped because caly one NI pump is in ,
service and without adequate N1 flow,it may not be possible to l
depressurire the plant to the point where accumulators and ND
pusups can ensure core heat removal.
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Ques 312a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #73 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 313 Answer: C l
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1Pt(s) Unit 1 is responding to a large break LOCA into containment. Given the
following conditions: 1
* NC pressure is 40 psig
* PZRlevelis 0%
. All NCPs have been tripped
. Containment levelis 1I feet
.
The operators have reached step 6 in ES-1.3 (Transfer to Cold Leg Recirc) ,
which has them align the S/I system for recirc. j
=
FWST suction valve, FW-27A (ND Pump 1 A Suct From FWST),is open ;
and will not close manually from the control room j
. FW-55B (ND Pump IB Suct From FWST)is closed
. FWST levelis 4%
Which one of the following statements describes the correct action (s) to be
taken?
A. Dispatch an operator to close FW-27A locally and initiate makeup
{
,
to the FWST ;
,
B. Immediately transition to ECA-1.1 (Loss of Emergency Coolant i
Recirculation) !
C. Stop ND pump 1 A 4
D. Immediately stop both ND pumps
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For Official Use Only Ques 313a NRC EXAM 12/12/97
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Question #74 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 314 Answer: D
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1Pt(s) Unit I operators are responding to a LOCA from 100% power. They have i
reached step 4 of ES-1.1 (Safety injection Termination) which states:
4. Establish F7to containmentasfollows:
_a. Ensure IVI-77B (V7 Cont Isol) - OPEN
_b. Venfy VIpre.ssure-GREATER THAN90psig
Given the following conditions:
l
. VI pressure = 85 psig and decreasing slowly
e iVI-77B is open
Which one of the following actions is correct for this condition?
A. Verify that the VS system automatically cross-connects to VI at 76
l psig. VI pressure is only required to be greater than 50 psig for
the PORVs to operate.
B. Verify that ti.e standby air compressor auto starts and raises VI
-
pressure above 90 psig. Do not proceed in ES-1.1 until VI
"
pressure is greater than 90 psig.
C. INI-438A and INI-439B will automatically isolate the VI header -
from the standby Nitrogen supply to allow PORV operation.
D. Follow the RNO and manually align Nitrogen to the PORVs by
opening INI-438A and INI-439B in the control room
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For Official Use Only Ques 314a NRC EXAM 12/12/97
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Question #75 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 315 Answer: B
1Pt(s) Unit 1 is shutdown in mode 3. Given the following plant conditions:
. Train "A" shutdown monitor (SDM) has just been enabled
. Automatic blended makeup to the VCT is in progress from IB Reactor
Makeup Water (RMW) pump and the 1 A Boric acid (BA) pump
.
The operator is preparing to test Train B SDM Monitor before enabling the
monitor
Which one of the following states the annunciators that will alarm if the
operator inadvertently depresses the TEST pushbutton for Train A SDM
Monitor vice Train B SDM Monitor?
A. TRAIN A SHUTDOWN MARGIN ALARM and BA FLOW
DEVIATION
B. TRAIN A SHUTDOWN MARGIN ALARM and TOTAL
MAKEUP FLOW DEVIATION
C. TRAIN A W/R NEUTRON FLUX SYS TROUBLE and BA
-
FLOW DEVIATION
D. TRAIN A W/R NEUTRON FLUX SYS TROUBLE and TOTAL
MAKEUP FLOW DEVIATION
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For Offcial Use Only Ques 315a NRC EXAM 12/12/97
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Quesbon #76 CATAWBA NUCLEAR STATION - RO EXAM
) Bank Question: 316 Answer: C
1Pt(s) Unit 1 is in mode 5 preparing to startup after a refueling outage. The fill and
vent of the NC system is in progress.
. NC temperatureis 120 'F
= Ldown is in senice via the ND system through NV-135
. 9.arging isin senice
. The "B" loop wide range pressure instrument is out of senice for
calibration
. The key switch for PORV-32B is selected to " Low Pressure"
. The key switch for PORV-34.A is selected to " Normal"
Which one of the following statements describes the first NC system response
to an over pressurization event?
, A. PORV-32B will open to relieve pressure
B. PORV-34A will open to relieve pressure
C. The relief valve in the suction line of the ND pump will open to
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relieve perssure
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D. The relief valve in the letdown line will open to relieve pressure
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For Official Use Only Ques 316a NRC EXAM 12/12/97
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- Queshon #77 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question:317 Answer: D e
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.IPt(s) Unit 1 is responding to an ATWS event from 25% power in FR-S.1 (Response
to Nuclear Power Generation /ATWS). Given the following events and
conditions:
:
. Digital Rod Position Indication (DRPI) has been lost for all rods ;
. The reactor was tripped by opening the MG set breakers i
. All immediate actions have been completed in FR-S.1 ,
Which one of the following indications is used to verify reactor trip in FR-S. l?
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A. Power range indication decreasing
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B. Control Rod Drive M-G set breakers open )
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C. Negative intermediate range startup rate '
D. Intermediate range amps decreasing
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For Official Use Only Ques 317a NRC EXAM 12/12/97
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Question #78 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 318 Answer: D
!
!Pt(s) Unit 1 is conducting a reactor startup. Given the following conditions:
i
* IR Channel N35 indicates 8X10'" amps
. 4
IR channel N36 indicates 1.1X10 " amps
* SR channel N31 indicates 7.1X10' CPS
= SR channel N32 indicates 6.8X10' CPS I
!
If the N31 SR detector high voltage power supply fails, which one of the
following statements describes the plant response?
A. A reactor trip signal is generated and a reactor trip occun
because the N31 output fails high.
B. A reactor tdp signal is generated but no reactor trip occurs
because one IR channel is above P-6 and the SR high level trip is
automatically blocked.
C. No reactor trip signal is generated because the required
coincidence for the SR high level flux trip is 2/2 SR channels.
D. No reactor trip signalis generated because the channel N31
output fails low.
1
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! l For Official Use Only Ques 318a NRC EXAM 12/12/97 i !
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Question #79 CATAWBA NUCLEAR STATION RO EXAM
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s Bank Question: 319 Answer: A
1Pt(s) Unit 1 is conducting a reactor startup. Given the following conditions:
. IR Channel N35 indicates 8X10'" amps j
e IR channel N36 indicates 1.1X10 amps
. SR channel N31 indicates 7.1X10' CPS
. SR channel N32 indicates 6.8X10' CPS
The N-36 instrument power fuse blows due to an intemal fault. Which one of ,
the following actions are required? I
:
A. Hold power at present levels until repairs are made.
B. Continue the stadup with no restrictions.
C. Continue the stanup but do not exceed 10% thermal power.
1
D. Insert rods and commence a reactor shutdown.
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For Official Use Only Ques 319a NRC EXAM 12/12/97
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Question #80 CATAWBA NUCLEAR STATION RO EXAM
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Bank Questioni 320 Answer: B
1Pt(s) Unit I was operating at 100% power. Given the following conditions:
* EMF-33 (Condenser Air Ejector Monitor) alarms in trip 2
If all the automatic features operate as designed (without operator i
intervention), which one ofthe following indications will provide the best
]
indication (most sensitive and timely) to identify the leaking S/G and trend the l
magnitude oftheleak?
ll
A. Comparing S/G feed flow to steam flow mismatch
B. Observing EMF-26,27,28 and 29 (steamline monitors)
C. Observing EMF-34 (S/G sample line monitor)
D. Frisking the S/G blowdown demineralizer cation column
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For Official Use Only Ques 320a NRC EXAM 12/12/97
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Question #81 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 321 Answer: A
1Pt(s) In E-3, (Steam Generator Tube Rupture) Enclosure 6 (NC Pressure and
Makeup Control to Minimize Leakage) the operators are directed to energize ,
pressurizer heaters if the mptured S/G level is decreasing and pressurizer level I
is greater than 25% throughout the event. What is the purpose for this action?
A. Maintain pressurizer saturation temperature corresponding to j
niptured S/G pressure to minimize S/G leakage into the NC i
'
system.
B. Maintain pressurizer saturation temperature corresponding to
intact S/G perssure to minimize primary leakage into the S/G.
C. Maintain pressurizer saturation temperature above the !
corresponding niptuitd S/G pressuie to ensure S/G water does j
not flow into the NC system . l
D. Maintain pressurizer saturation temperature corresponding to I
intact S/G pressure to minimize NC pressure transients.
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For Official Use Only Ques 321a NRC EXAM 12/12/97
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Question #82 CATAWBA NUCLEAR STATION RO EXAM
Bank Question: 322 Answer: A
IPt(s) Unit 2 is responding to a loss of main feedwater accident. Given the following
conditions:
. Both main feedwater (CF) pumps tripped to initiate the event
. The reactor tripped on Lo-Lo S/G level
. 2B motor-driven aux feed (MDCA) pump started but was air bound and
no flow or discharge pressure was indicated -
* ' 2A MDCA pump and the turbine-driven aux feed (TDCA) pump are each
discharging approximately 500 gpm flow
Which one of the following sequences describes the condition of the MDCA
isolation valves to the S/Gs if these pump conditions do not change?
i 2CA-58A (CA to B S/G) 2CA-46B (CA to C S/G)
A. remain open remain open
B. close remain open
l C. remain open close
D. close close
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For Official Use Only Ques 322a NRC EXAM 12/12/97
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Question #83 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 323 Answer: A ]
IPt(s) Unit 1 is shutdown, mode 5, draining liquid radioactive waste from the WL
system into an unprotected outdoor storage tank for disposal offsite. Giver tne
i
following tank radiochemistry analysis:
*
Total tank activity = 10.5 Ci with a combined halflife of 50 days
.
Tritium activity = 1.5 Ci with a halflife of 12.6 years
.
Noble gas activity = .4 Ci with a halflife of 48 hours
Which one of the following action (s) (if any) is required for these conditions?
.
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REFERENCES PROVIDED
A. No action is required at this time
B. Immediately stop all additions of radioactive material into the
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tank and wait for the tank contents to decay
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C. !
Immediately reduce the tank contents by transferring radioactive
material to another tank.
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D. Move the tank into the auxiliary building
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>; For Official Use Only Ques 323a NRC EXAM 12/12/97 i
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Question #84 CATAWBA NUCLEAR STATION RO EXAM
i Bank Question: 324
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Answer: D
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1Pt(s) Unit 2 was operating at 100% power when a design basis LOCA into
containment occurred. Given the following conditions:
. 2 EMF-53 A/B (Containment Post LOCA Monitors) are both inoperable
. The area radiation monitors (ARMS) in lower containment are alarming. I
!
Which one of the following indications would most accurately de: ermine the l
area dose rates inside containment for source tenn assessment? !
I
A. ARM indications in lower containment
B. Reactor coolant filter radiation (2 EMF-5 or 2 EMF-6) monitor !
indications !
C. EMF-54 (Unit Vent Monitor) indications.
D. Portable instruments readings taken on the containment wall and
app Spriately scaled for shielding factors. .
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For Official Use Only Ques 324a NRC EXAM 12/12/97
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Question #85 CATAWBA NUCLEAR STATION RO EXAM
f Bank Question: 325 Answer:B
1Pt(s) Unit I was operating at 100% with the pressurizer level controller in the 1-2
position. Given the following conditions and events:
* Charging Dow reduces to minimum
. Backup heaters energize
* Pressurizer level decreases DUT NO operator action is taken
a Letdown isolates
. All pressurizer heaters deenergize
. Pressurizer level tums and increases to the high level reactor trip setpoint.
Which one of the following failures has occurred to cause this plant response?
A. PZRlevel channelI has failed LOW
B. PZR level channelI has failed HIGH
C. Auctioneemi High Tave signal has failed HIGH
, , D. Reference level signal has failed to the NO-LOAD value
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For Official Use Only Ques 325a NRC EXAM 12/12/97
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Question #86 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 326 Answer: A l
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1Pt(s) Unit I was operating at 100% power. Given the following conditions:
)
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. The 1B EDG was out ofservice for maintenance
. A fault in a main power system potential transformer caused a loss of
offsite power
. A reactor trip occurred
. The 1 A EDG started and canied loads on the 4160 VAC busses
Which one of the following events would cause a safety injection actuation i
under these circumstances?
A. Excessive addition of cold CA water into the S/Gs combined with
steam loads that cannot be isolated due to the loss of control
power caused a safety injection on low steamline pressure.
B. Excessive addition of cold CA water into the S/Gs combined with
steam loads that cannot be isolated due to the loss of control
power caused a safety injection on high rate of decrease in
,
steamline pressure. ;
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C. The loss of power caused the CA control valves to close on the 2D
S/G which caused a loss oflevel in the 2D S/G which caused a loss
of pressurizer level due to shrink from the pressure transient.
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D. The loss of offsite power together with one 4160 VAC safety bus
caused the MSIVs to close which caused a safety injection on low
steamline pressure.
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L l l For Official Use Only Ques 326a NRC EXAM 12/12/97 l l
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Question #87 - CATAWBA NUCLEAR STATION RO EXAM !
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I Bank Question: 327 Answer: A
1 Pt(s) Unit 1 is operating at 100%. If an unisolable rupture occurs in the instrument
air (VI) header in the service building, which one of the following conditions
and associated reasons will cause the reactor to trip FIRST?
j A. The main feed regulating valves will close causing a S/G low level l trip. i ! i , B. The letdown isolation valves will close and the pressurizer will fill l ! up causing a pressurizer high level trip i l C. The pressurizer spray valves will fail open causing a pressurizer
low pressure trip
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D. The auxiliary spray valve will fail open causing a pressurizer low !
'
pressure trip.
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i i l , i ~O l l L For Official Use Only Ques 327a NRC EXAM 12/12/97 o
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l Question #88 CATAWBA NUCLEAR STATION RO EXAM !
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) - Bank Question: 328 Answer: B
1Pt(s) Unit 1 is shutdown in mode 5 on the midnight shift.
. The control room operators consist of the following personnel
. The OSM
e The Control Room SRO
+ The Unit I and Unit 2 ROs -who are also the OATCs
. The Unit 1 BOP operator is currently securing a containment air
release. ;
'
e All other on-shift operators are out of the control room at the present
time.
An I&E Tech asks the Unit 1 RO to stand by MC-13 in the control room
during surveillance testing to acknowledge annunciator alarms for EMF-39
ACOTs. Which one of the following statements is correct in regards to the
RO's ability to comply with this request?
A. The RO may pinceed to this area of the control room as it is
within the nonnal surveillance area. i
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B. The RO may only enter this area to acknowledge alarms for a
short period of time as it is part of the limited surveillance area.
C. The RO may enter this area if he/she first temporarily turns over
the OATC position to the Unit 2 RO prior to entering this aita.
D. The RO may enter this area if he/she first must turn over the
OATC position to the Control Room SRO prior to entering this
area.
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For Official Use Only ' Ques 328a NRC EXAM 12/12/97
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Question #89 CATAWBA NUCLEAR STATION RO EXAM
_) Bank Question: 329 Answer: D
1Pt(s) Unit I was operating at 100%. Given the following conditions- i
I
* NC pressure = 2235 psig
* A pressurizer PORV opened
* The operators immediately closed the PORV in manual
What actions are required by Tech Specs within one hour?
A. Maintain the PORY closed in manual
B. Maintain the PORV closed in manual and remove power from the
PORV. i
C. Close the associated PORV block valve.
D. Close the PORY block valve and remove power from the block
valve.
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For Official Use Only Ques 329a NRC EXAM 12/12/97
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Question #90 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 330 Answer: A
1Pt(s) The NCPs are limited to 3 consecutive starts in any 2 hour period with an
additional requirement of a muumum idle period of 30 minutes between
restarts. ht is the reason for this limitation?
!
A. This is an engineering striction to prevent overheating the
motor windings due to high starting currents.
1
B. This is an operational restriction to that assures that the oil j
temperature will decrease to design specifications between restart
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attempts.
C. This is an operational restriction to allow the NCP seals to fully
resent between NCP rotations.
D. This is an administrative restriction that prevents operators from
itstarting without a deliberate approach to ensure that all
precautions and intedocks have been satisfied.
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- For Official Use Only Ques 330a NRC EXAM 12/12/97
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Question #91 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 331 Answer: C
1Pt(s) An operator is conducting a surveillance procedure which requires a series of
approximately 25 sequential steps to be performed while standing in a
contaminated area. He/she is in direct communications with another operator,
the communicator who holds the procedure and reads each step sequentially.
If the performer does not have the procedure in hand as he/she perfomu the
steps. what are the requirements of OMP l-4 regarding the sign off for each
step?
A. Only the performer can sign off the steps upon completion of the
task after leaving the contaminated area along with the time of
each step.
B. The communicator signs ofreach step as the step is completed
using his/her own initials plus the time.
C. The communicator signs off each step as the step is completed
using his/her own initial and the initials of the performer plus the
time.
D. The communicator places a checkmark next to each step as the
step is completed and the performer signs off each step after
-
completion using his/her initials plus the time.
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For Official Use Only Ques 331a NRC EXAM 12/12/97
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Question #92 CATAWBA NUCLEAR STATION ' RO EXAM i
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! Bank Question: 332 Answer: D
1Pt(s) Unit 1 is conducting a plant startup in mode 4. Given the following conditions:
,
l . The pressurizer is full-in a water solid condition
. Numerous, independent failures occurred
. NC system pressure = 2742 psig
Which one of the following statements describes the required action to be
taken in accordance with Tech Specs?
A. Reduce NC system pressure and be in mode 5 within I hour.
l Report this violation to the NRC Operations Center within I
hour.
i
B. Reduce NC system pressure immediately and be in mode 5 within
i 1 hour. Report this violation to the NRC Resident inspector
within 24 houn. i
C. Reduce NC system pressure within 5 minutes and comply with ;
l Tech Spec 3.0.3. Report this violation to the NRC Resident '
Inspector within 24 hours.
-.
D. Reduce NC system pressure within 5 minutes. Report this
violation to the NRC Operations Center within I hour. I
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For Offidal Use Only -
Ques 332a NRC EXAM 12/12/97
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Quesbon #93 CATAWBA NUCLEAR STATION RO EXAM
I
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} Bank Question: 333 Answer: C
;
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, ' 1Pt(s). . Unit 1 is in a refueling outage with fuel movement in progress. What is the j
'
endiest time'when the combinations ofoperable Boron Dilution Mitigation
System (BDMS) and Source Range (SR) Neutron Flux Monitors requires
immediate suspension of all refueling activities?
l L 0200 No train of BDMS + both SR monitors L 0230 One train BDMS + both SR monitors
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'0300 One train ofBDMS + one SR momtor
!
0330 Both trahu ofBDMS + no SR monitors j
A. 0200
l B. 0230
C. 0300
~ l
D. 0330 i
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For Official Use Only Ques 333a NRC EXAM 12/12/97
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Question #94 CATAWBA NUCLEAR STATION RO EXAM !
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Bank Question: 334 Answer: B :
1Pt(s) Dudng a rod swap as part of a Zero Power Physics Test (ZPPT) an operator
was withdrawing shutdown ba .k B, as directed by the Test Coordinator.
Dudng the withdrawal sequence, the Test Engineer became distracted and
failed to propedy observe the indication on the special reactivity computer. A
reactivity excursion occurred when the operator withdrew the shutdown bank
beyond cdticality. The highest start up rate observed was 2.5 DPM when the
operators inserted rods to stop the power increase.
Which one of the following reactor trips would have automatically terminated
the rod withdrawal if no operator action had been taken?
A. Source Range Hi Flux
B. Intermediate Range High Hux
C. Power Range High Hux
D. Intermediate Range Hi Startup Rate
.
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For Official Use Only Ques 334a NRC EXAM 12/12/97
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Question #95 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 335 Answer: B
1Pt(s) Which one of the following is the Emergency Exposure
l- TEDE Dose Limit for the protection ofcritical equipment?
A. 5 rem.
B.10 rem.
l C. 25 rem.
D. 50 rem.
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For Offical Use Only Ques 335a NRC EXAM 12/12/97
r i l Question #% CATAWBA NUCLEAR STATION RO EXAM
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) Bank Question: 336 Answer: B
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1Pt(s) Which one of the following conditions will automatically:
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* terminate a containment air release by closing VQ-10 or
!
. terminate a containment air addition by closing VQ-13
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during startup if EMF-39 (Containment Monitor) is out of service? '
i
A. EMF-36 (Unit Vent Monitor) trip 2 signal
B. EMF-38 (Containment Monitor) trip 2 signal
C. Containment Pressure = +0.2 psig and decreasing
D. Containment Air Fan low flow (at the discharge)
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For Official Use Only Ques 336a NRC EXAM 12/12/97
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Question #97 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 337 Answer: C
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1Pt(s) Unit 1 just completed a shutdown to mode 5 with both ND trains in service
l
prior to core ofiload.
- Which one of the following statements is correct regarding an approved reactor
coolant vent path into containment to mitigate the consequences of a loss of
ND cooling?
!' A. An open reactor head vent will provide an approved vent path in l mode 5. !
B. An open S/G cold leg manway with the hot leg nonle dam
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installed, and the cold leg nozzle dam removed will provide an
approved vent path.
C. An open S/G hot leg manway with the hot leg nonle dam not
installed will provide an appmved vent path.
D. An approved vent path is nw mguire.1 until the reactor head is
removed in mode 6.
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For Offidal Use Only - Ques 337a NRC EXAM 12/12/97
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Quesbon #98 CATAWBA NUCLEAR STATION RO EXAM
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Bank Question: 338 Answer: C
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1Pt(s) Which one of the following statements complies with the requirements of OMP
l-7 regarding the rules of usage for abnormal procedures (APs) when the
EOPs have been implemented?
A. APs may not be implemented when EOPs have been entered.
B. Only one AP at a time may be implemented when EOPs have
been implemented. Concurrent implementation of APs when
EOPs art in use is not allowed.
C. APs may be implemented concurrently with EOPs. However, the
APs were written assuming that SI has not actuated and operators
must be careful when using APs if SI has occurred.
D. APs may be implemented concurrently with EOPs with the
exception of events where SI has actuated. APs were written ,
assuming the SI had not occurred and cannot be used if Si has l
actuated. l
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I i i ! , u ! l. For Official Use Only Ques 338a NRC EXAM 12/12/97 l
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Question #99 ' CATAWBA NUCLEAR STATION RO EXAM
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[ Bank Question: 339 Answer: D
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1Pt(s) E-3, Steam Generator Tube Rupture, step #21.b reads as follows:
"[F A T ANY TIME ruptured S'G(s) pressure is decreasing.... THEN
perform Step 21. "
Which one of the following statements is correct with regards to this step?
A. The step is applicable continuously unless it is determined not to
be pertinent to the recovery effort.
B. The step is applicable while in E-3 and after transition to
subsequent proceduits until alternative guidance is provided.
C. The step is only applicable until another continuous action step is
reached in E-3.
D. The step is only applicable while in E-3.
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For Official Use Only Ques 339a NRC EXAM 12/12/97
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Question #100 CATAWBA NUCLEAR STATION RO EXAM l
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. Bank Question: 340 Answer: B
1
1Pt(s) A LOCA has occurred on Uniti. Given the following conditions: ;
e You are an extra RO on shift but not the OATC.
. You are out in the twtine building performing an independent verification
at the time that the site assembly is announced.
Which one of the following statements con ectly describes the requirements for
site assembly?
A. A site assembly is required for any emergency classifi etion. You
would proceed directly to the Control Room.
B. A site assembly is arquired at an Alert classification. You would
proceed to the Control Room.
C. A site assembly is required at a Site Area Emergency
classification. You would proceed to the TSC.
D. A site assembly is required at a General Emergency classification.
) You would proceed directly to the Evacuation Sites Alpha or
Bravo.
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For Official Use Only Ques 340a NRC EXAM 12/12/97
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NRC Official Use Only N/~' N
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Nuclear Regulatory Commission
Senior Reactor Operator Licensing ,
Examination l
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This document is removed from
l OfIicial Use Only category on
date ofexamination 4 v NRC Official Use Only
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401 Site-specific Written Examination Form ES-401-1
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Cover Sheet
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U. S. NUCLEAR REGULATORY COMMISSION
SITE-SPECIFIC
WRITTEN EXAMINATION
APPLICANT INFORMATION
Name: Region: I /dQ/111/ IV / V
Date: Facility / Unit: CATAWBA
License Level: RO /_I @ ] Reactor Type: M]/ CE / BW / GE
INSTRUCTIONS
Use the answer sheets provided to document your answers. Staple this cover
sheet on top of the answer sheets. Points for each question are indicated in
parentheses after the question. The passing grade requires a final grade of at
least 80 percent. Examination papers will be picked up 4 hours after the
examination starts.
All work done on this examination is my own. I have neither given nor received
aid.
Applicant's Signature
RESULTS "
Examination Value 100 Points
Applicant's Score Points
, v
Applicant's Grade Percent
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m .402 Policies and Guidelines Attachment 2
for Taking NRC Written Examinations
1. Cheating on the examination will result in a denial of your application and could result in
more severe penalties. !
2. After you complete the examination, sign the statement on the cover sheet indicating that l
'the~ work is your own and you have not received or given assistance in completing the
examination.
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3. To pass the examination, you must achieve a grade of 80 percent or greater.
4. The point value for each question is indicated in parentheses after the question number.
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5. There is a time limit of 4 hours for completing the examination. )
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6. Use only black ink or dark pencil to ensure legible copies. !
7. Print your name in the blank provided on the examination cover sheet and the answer sheet.
8. Mark your answers on the answer sheet provided and do not leave any question blank.
.. If the intent of a question is unclear, as questions of the examiner only.
L 10. Restroom trips are permitted, but only one applicant at a time will be allowed to leave.
. Avoid all contact with anyone outside the examination room to eliminate even the
appearance or possibility of cheating. l
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! 11. When you complete the examination, assemble a package including the examination
questions, examination aids, and answer sheets and give it to the examiner or proctor.
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Remember to sign the statement on the examination cover sheet.
L 12. After you have turned in your examination, leave the examination area as defined by the
examiner.
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SENIOR REACTOR OPERATOR ANSWER SHEET Page 1
A TIPLE CHOICE (Circle or X your choice)
If you change your answer, write your selection in the blank.
001 a b c h 035 @ b c d A 069 @ b c d 1
002abc@A 036 a b @d 4_ 070 a @ c d 1
003 @ b c d 1 037 a b c @ l 071 abc@p
004 a b @ d 4_ 038 @ b c d j 072 @ b c d &
005 @ b c d 1 039 a b c g 073 a @ c d 1
006ab@dL 040 a b @ d C. 074 a b @ d L
007 a b c @ A 041 @ b c d _A. 075ab@dC.
008 a b @ d C. 042@bcdA 076 @ b c d A_
009 @ b c d A 043abc@A 077 a b dL
010 a b c @ l 044a@cd3 078ah@cdIR
011 a@cdA 045 @ b c d A 079 a b c @ A
012 a b c @ A 046 @ b c d A 080 a b @ d C
013 a b c @ A 047@bcdA_ 081 abc@H
014 a b @ d L 048 @ b c d A 082 a b @ d C
015 a b c @ A 049 a b @ d L 083 a b c @ R
016 a @ c d 1 050 a b @ d L 084 @ b c d A
017 @ b c d A 051 abc@,D_ 085 a b c @
018 a b @ d C- 052 a b @ d L 086 a b c @ 2
019 a b @d L 053 a@ c d & 087 @ t c d 1
020 a @ c d_B_ 054abc@H 088 a b @ d C.
021 @ b c d A 055ab@dC_,, 089 @ b c d L
022 a b c@)A 056abc@_n_ 090 a @ c d 1
023 a @ c d H 057 a @ c d B_ 091 ab@dC
024 a b c @ l 058 a b @ d L
025 a b @ d C. 059 a b @ d C._ 092
093 a bac b _D_ @@d C
026 a b c @ D 060 a b c @ A ' 094 @ b c d &
027.@ b c d 1 061 a@cdl 095 a b c @H
O28 a b @ d C 062 @ b c d A 096 a b c @ H
029 a b @ d C 063 a @ c d & 097a@cdl
030 a b c @ D 064 @ b c d A 098 @ b c d A
031 ab@dC. 065 @ b c d A 099 a b @ d C
032 a b @ d L 066 @ b c d A 100 a b c @ S
033 a b @ d L 067 a b c @ A
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034a@'cd 068a@cdd
( * * * * * * * * * * END O F EX AMIN ATION * * * * * * * * * * )
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Question #1 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 240 Answer: D
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1 Pt(s) Unit I was operating at 100% power when a multiplexing thyristor for bank C
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in the rod control system failed causing bank C to be energized at the same i
time as bank D. If rod control is in automatic, which one of the following l
sequence ofevents describes the response of the rod control system following i
this failure? i
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A. Banks C and D will move at the same time in response to rod ;
withdrawal or insedion signals.
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B. Banks C and D will move at the same time in response to rod i
insenion signals only.
C. Banks C and D will only move together for manual md insedion j
signals. Automatic signals will have no effect on rod movement. I
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D. Banks C and D will not move for rod insedion or withdrawal {
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signals.
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For Offical Use Only Ques 240a NRC EXAM 12/12/97
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QUEST 1DN #2 CATAWM NUCLEARSTATION 5RO EXAM
Bank Question: 241 Answer: D
1 Pt(s) Unit 1 is operating at 100% power when the supply breaker from LXC to #1
Control Rod Drive MG set opens. Which one of the fo!!owing sequence of
events will occur to the reactor trip breakers (RTB) and the reactor trip bypass
breakers (RYB)?
, , , A. RTB "A" and RYB ""B" will open
B. RTB "B" and RYB "A" will open
C. RTB "A" and "B" will open
., D. No RTB orRYB breakers will open
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Ques 241a. NRC OFFICIAL USE ONLY NRC EXAM 12/12/97 i
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Question #3 CATAWBA NUCLEAR STATION SRO EXAM i
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) Bank Question: 242 Answer: A
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1Pt(s) Unit 1 is operating at 100% power. Given the following conditions on the 1 A
NCP
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Time 0200 0210 0220 0230
. Motor bearing temp (F ): 181 184 193 196
Shaft vibration (mils): 2 3 4 5
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Pump seal pressure AP(psig): 196 201 223 235 i
Pump seal outlet temp (F ): 208 212 221 228
When are the operators required to trip NCP-1 A?
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A. 0200 l
B. 0210
C. 0220
D. 0230
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' For Official Use Only - Ques 242a' NRC EXAM 12/12/97
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Question #4 CATAWBA NUCLEAR STATION SRO EXAM
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,_ ) Bank Question: 244 Answer: C
1Pt(s) Which one of the following precautions are required by procedure when
initiating auxiliary spray to prevent thermal shock?
A. Auxiliary spray shall NOT be initiated when pressurizer vapor
space temperature is LESS than 320 *F
B. Auxiliary spray may be initiated after warming up the spray line
if the differential temperature is GREATER than 320 F between
pressurizer and NC system.
C. Auxiliary spray shall NOT be initiated when differential
temperature between the pressurizer and the NV system is
GREATER than 320 F.
D. Auxiliary spray shall NOT be initiated when the differential
temperature between the pressurizer and the NV system is LESS
than 320 F.
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For Official Use Only Ques 244a NRC EXAM 12/12/97
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Question #5 CATAWBA NUCLEAR STATION SRO EXAM
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} Bank Question: 245 Answer: A l
1Pt(s) Given the following plant conditions on Unit 2:
* Reactor poweris at 100%
* All control systems and components are in automatic
. VCT level is at 40% with auto makeup in progress
If the channel i VCT level transmitter (NVLT-5761) fails high as indicated on l
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the Operator Aid Computer and NO operator action is taken, which one of the I
following describes the response of ACTUAL VCT level? l
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A. Decrease to 0%, where charging pump cavitation causes loss of !
charging '
B. Decrease to 4.3%, where swap-over to the FWST suction occurs i
C. Increase to 49.3%, and then cycle between 32.7% and 49.3%
D. Increase to 85.3%, and then cycle between 70.1% and 85.3%
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Question #6 CATAWBA NUCLEAR STATION SRO EXAM j
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Bank Question: 246 Answer: C :
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1Pt(s) Which one of the following conditions will automatically UNBLOCK both the )
Unit I pressurizer low pressure and steam line low pressure safety injection
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actuation signals? '
A. T.,, increases to 564 *F
B. Steam pressure increases to 614 psig !
C. NC system pressure increases to 2015 psig
D. Main steam line differential pressure increases to 100 psid
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For Official Use Only Ques 246a NRC EXAM 12/12/97
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i l Question #7 CATAWBA NUCLEAR STATION SRO EXAM l ! !
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Bank Question: 248 Answer: D
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1Pt(s) The following plant conditions exist on Unit 1:
. A plant power increase is in progress at 85%
. Channel I pressurizer pressure detector has previously failed LOW and has
been removed from service, including tripping all effected bistables.
. Power range channel N-43 upper detector has just failed HIGH,
Which one of the following trip function bistables associated with the N-43 I
Power Range channel failure will require a plant shutdown to Mode 3 to
comply with Technical Specifications under these conditions? ;
A. Loss of Coolant or Steam Break Protection
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B. Steam Generator Lo-Lo Imel
C. NC Loop Over-power Delta-T (OPDT)
D. NC Loop Over-temperature Delta-T(OTDT)
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For Offical Use Only Ques 248a NRC EXAM 12/12/97
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Quesbon #8 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 249 Answer: C
1Pt(s) Unit 2 is conducting a reactor startup in mode 2 with the following conditions:
* Reactor power is at 8x10'" amps and increasing.
e' Operators are evaluating overlap between intermediate and source range.
. The instrument power fuses blow for the N-35 intermediate range channel.
Which one of the following actions are required by procedures?
A. Enter E-0 (Reactor Trip or Safety Injection) at step 1.
B. Insert rods to lower neutrun flux level to a value within the source
range. Within one hour, repair N-35 or shutdown to mode 3.
C. Do not exceed 1.0x10- amps until N-35 has been restored. Verify
that P6 permissive status light is not lit within one hour.
D. Continue startup procedure but do not exceed 10% power until
N-35 has been restored to opera' ole status.
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For Offical Use Only Ques 249a NRC EXAM 12/12/97
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Question #9 CATAWBA NUCLEAR STATION SRO EXAM j
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Bank Question: 250 Answer: A
1Pt(s) Which one of the fouowing NC system penetrations is sensed to provide
RVLIS indication?
A. Reactor Vessel head vent
B. Loop B cold leg l
C. Reactor vessel post accident system sample line (at the seal table) ,
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D. Reactor vessel flange leakoff hne
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For Official Use Only Ques 250a NRC EXAM 12/12/97 l
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Question #10 CATAWBA NUCLEAR STATION SRO EXAM
\ Bank Question: 251 Answer: D
1Pt(s) Unit 1 is responding to a loss ofoffsite power. Given the blowing events and
conditions:
* Both EDGs load normally on the 4160 VAC essential b, . nes.
. The "YV/RN Cool Water Ctrl" is in the " CTRL RM" pc irion
Which one of the following statements describes the containm e cooling water
alignment under these conditions?
A. Cooling water is isolated until manual operator action is taken to
restore the RN system.
B. Cooling water auto-swaps from the RN system to the YV system
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C. Cooling water is isolated until manual operator action is taken to
restore the YV system.
D. Cooling water auto-swaps from the YV system to the RN system
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For Official Use Only Ques 251a NRC EXAM 12/12/97
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Quesbon #11 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 253 Answer: B
1Pt(s) Given the following outputs / indications from the excore nuclear instrument ,
channels with power stable during a Unit I startup: i
e N-41; 20%
. N-42: 12 % (Upper detector failed LOW, power mismatch defeat -
switch positioned to defeat the detector)
e N-43 22 %
* N-44: 24 %
Which one of the following is the power range value that is used to establish
the arbitrator signal for the steam flow circuit in the steam generator level and
feedwater pump speed control system?
A. 18 %
B. 20 %
C. 22 %
D. 24 %
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For Official Use Only Ques 253a NRC EXAM 12/12/97
I l Question #12 CATAWBA NUCLEAR STATION SRO EXAM ! ! i l
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) Bank Question: 258 Answer: D l
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1 Pt(s) Unit I was operating at 100% power with plant control systems in automatic. I
Which one of the following conditions would cause the condensate load
rejection bypass valve (CM-83) to open automatically?
A. Feedwater pump suction pressure is reduced by 100 psig.
B. Turbine impulse pressure is reduced by 50 psig.
C. Feedwater pump recirculation flow decreases to 200 gpm.
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D. Condensate booster pump suction pressure decreases to 100 psig. !
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For Official Use Only Ques 258a NRC EXAM 12/12/97
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Question #13 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 259 Answer: D
1Pt(s) The operators are conducting a cooldown in Mode 3 on Unit 1. Given the
following plant conditions:
,
. Tave = 510 "F
e NC system pressure = 1900 psig l
. l A Feedwater pump is maintaining S/G levels ;
. 1B Feedwater pump is tripped / shutdown I
. Pzrlow pressure Si has been blocked
* Steamline low pressure SI has been blocked i
e Both auxilia:y feed (CA) train AUTO-START-DEFEAT buttons have !
been depressed j
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- Which one of the following events will cause the motor-driven CA pumps to i
start automatically?
A. lA Feedwater pumps trips causing all S/G levels to decrease below
the Lo-Lo level trip setpoint.
B. NC loop "B" spray valve, NC-29, fails open causing NC pressure
to decrease to 1500 psig
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C. A steamline rupture causes the "C" S/G pressure to decrease to i
700 psig at a rate of 200 psig/second. I
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D. A feedline rupture causes the "D" S/G level to decrease below the
Lo-Lo level trip setpoint and containment pressure to increase to l
2 psig. !
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for Official Use Only Ques 259a NRC EXAM 12/12/97
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Question #14 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 260 Answer: C
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1Pt(s) Unit 2 is responding to a loss ofmain feedwater event from 100% power.
Given the following conditions:
. The reactor has tripped
. The 2A and 2B Motor-driven CA pumps started in auto
. The Turbine-driven CA pump (CAPT) staned in auto
. Train A of"CA SYS VLV CTRL" has been reset
. The train B reset has failed to actuate
. The CA pumps are aligned to the hotwell
. CA suction pressure drops to 5 psig as hotwell level decreases l
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Which one of the following system responses (if any) will occur without
operator action. i
A. 2A CA pump suction shifts to the RN system l
2 CAPT suction remains aligned to the hotwell
2B CA pump suction shifts to the RN system I
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B. 2A CA pump suction shifts to the RN system
2 CAPT suction remains aligned to the hotwell
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2B CA pump trips
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C. 2A CA pump trips
2 CAPT suction shifts to the RN system 1
2B CA pump suction shifts to the RN system
D. 2A CA pump trips
2 CAPT suction shifts to the RN system
2B CA pump trips
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For Official Use Only Ques 260a NRC EXAM 12/12/97
,- l l Question #15 CATAWBA NUCLEAR STATION SRO EXAM i
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Bank Question: 261 Answer:D
IPt(s) Unit 1 is operating at 100% power. A liquid waste release has been approved
for the shift. Given the following conditions:
+ IEMF-49 has been set in accordance with the LWR fonn. ,
+ 1 RN and 2 RL pumps are required to be operating per the LWR fonn I
+ 1 RN and 3 RL pumps are currently operating-
Which one of the following situations requires secunng the liquid waste release
(and not inunediately restaning it) after beginning the release?
A. The EMF-49 low sample flow alarm actuates immediately upon
commencement of the dischary,e and clears within 20 seconds.
B.- The EMF-49 Hi Rad" spikes into alarm 10 minutes after the
release started.
C. One RL pump is secured by an NLO due to excessive bearing
noise.
D. Local area flooding prompts the OSM to initiate a " Site ;
" Assembly". !
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For Official Use Only Ques 261a NRC EXAM 12/12/97
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Question #16 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 262 Answer: B
1 Pt(s) Which one of the following events will automatically terminate a waste gas
release? ;
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A. EMF-53 (Containment Post LOCA Monitor) alarm j
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B. EMF-50 (% G Dischsaye Monitor) alarm l
C. WG Compressor A trip -
D. WG system trip with compressor lockout
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For Official Use Only Ques 262a NRC EXAM 12/12/97
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Question #17 CATAWBA NUCLEAR STATION SRO EXAM
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i Bank Question: 263 Answer: A ,
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1Pt(s) Unit 1 is shutdown refueling with fuel movement in progress. Given the
following events and conditions:
]
. The new fuel elevator fails to operate in the up direction
Which one of the following statements describes the cause of this problem?
A. 1 EMF-15 (Spent Fuel Building Refueling Bridge Monitor) has
failed high) .
B. IEMF-20 (New Fuel Vault Monitors) has failed high
C. The load in the new fuel elevator weighs 1200 lbs
D. The spent fuel bridge crane is NOT indexed over the new fuel
elevator
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For Official Use Only Ques 263a I" EXAM 12/12/97
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Question #18 CATAWBA NUCLEAR STAT 10N SRO EXAM
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Bank Question: 264 Answer: C
1Pt(s) Which one of the following conditions require the operators to reduce reactor
coolant system pressure within 5 minutes to comply with safety limits. :
NCS Pressure NCS Temnerature Mode
A. 2742 psig 605 F 1
B. 2714 psig 550 F 2 :
C. 2746 psig 415 F 3 !
D. 2716 psig 345 *F 4 i
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For Official Use Only Ques 264a NRC EXAM 12/12/97 l
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Question #19 CA1 AWBA NUCLEAR STATION SRO EXAM i
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1Pt(s) Unit 1 is operating at 100% power, steady state when a PZR SURGE LINE
LO TEMP alarm annunciates. No other abnormal annunciators are alarming.
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Which one of the following statements is the most likely explanation for this l
alarm?
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A. Pressurizer temperature has slowly decreased due to pressurizer
heaters being off. 1
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B. Small insurge/outsurge cycles are occurring due to core xenon
oscillations.
C. Spray valve bypass flow has stopped due to orifice fouling I
pinblems.
D. Tw 'aas deceased due to inadvertent boration flow being
undetected.
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For Official Use Only. Ques 266a NRC EXAM 12/12/97
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QUEST 10N #20 CATAWB4 NUCLEARSTATION SRO EXAM
Bank Question: 267 Answer: B
1Pt(s) Unit I was operating at 100% power. Given the following events and !
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conditions:
. Complete loss of off-site power occurred
. Only one D/G was available and powering the essential bus
_.
. A rapid decrease in steam line pressure was caused by the addition of cold
CA water
. A safety injecdon initiation caused the pressurizer to go water solid.
Which one of the following statements describes the required plant pressure
control method when the pressurizer is solid following the termir.ation ofsafety
1
y, injec: ion? c. . ,
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A. Pressure is increased by manually energizing PZR heaters and
o decreased by opening a letdown line.
B. Pressere is increased by allowing decay heat to increase
temperature and decreased by allowing NC PORVs to cycle
automatically.
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C. Pressure is increased by manually energizing PIR heaters and j
decreased by opening reactor vessel head vents.
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D. Pressure is controlled by immediately drawing a bubble in the
pressurizer and reestablishing norrnal pressure control.
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Ques 267a NRC OFFICIAL. USE ONLY NRC EXAM 12/12/97
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Question #21 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 269 Answer: A
1 Pt(s)' Unit 2 is operating at 100% power. I&E is conducting a surveillance test on
the A reactor trip breaker. Given the following conditions:
. SSPS Train"A"is in test.
. Reactor Trip Breaker (RTA)is open
. Bypass Breaker (RYA)is shut
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If a subsequent loss of power occurs to SSPS Train "A", which one of the
following statements describes the effect that this will have on the reactor
protective system?
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A. There will be no change, the General Warning light will remain lit
on SSPS Train "A".
B. A General Warning light will only be lit on SSPS Train "A" after ;
power is lost. l
C. RYA will open but the reactor will not trip when power is lost.
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D. A reactor trip will occur when power is lost.
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For Official Use Only Ques 269a NRC EXAM 12/12/97
I l Queston #22 CATAWBA NUCLEAR STATION SRO EXAM l !
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Bank Question: 270 Answer: D
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1Pt(s) Unit 1 is operating at 100% power. Given the following indications on the
Digital Rod Position Indication system:
. General warning light for rod D-4 is flashing
. RPI URGENT annunciatoris alarming
. Urgent alarms 1,2 and 3 are flashing
. Rod bottom LED for rod D-4 is lit
Which one of the following describes the condition of rod D-4?
A. Rod D-4 DRPI indication is at half accuracy
B. Rod D-4 DRPI indication is at full accuracy
C. Rod D-4 DRPI indication is valid and the rod is fully inserted
D. Rod D-4 position cannot be determined by DRPI
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For Official Use Only Ques 270a NRC EXAM 12/12/97
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- Question #23 CATAWBA NUCLEAR STAEON SRO EXAM
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._! Bank Question: 271 Answer: B
1Pt(s) Which one of the following NC system instruments are averaged together to
obtain a single output? ,
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A. Cold leg resistance temperature detectors (RTDs) l
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B. Hot leg RTDs
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C.
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Loop pressure detectors
D. Loop flow detector high pressure taps
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For Official Use Only Ques 271a NRC EXAM 12/12/97
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Question #24 CATAWBA NUCLEAR STATION SRO EXAM
'} Bank Question: 272 Answer; D
1Pt(s) Unit 2 is responding to a LOCA from a trip at full power. Given the following
conditions:
. A safetyinjection has occurred
. Train B Sp signal failed to actuate
. The 2A NS pump started automatically but the 2B NS pumps was staned
manually by an operator
. The Ss signal and sequencer have been reset
. The train A Sp signal has not been reset
-. Both pumps were stopped for shifting suctions to the containment sump
Ifcontainment pressure is 0.25 psig, which one of the following statements
describes the operation of the NS pumps upon completion of the swapover?
A. Both NS pumps will restart automatically if containment pressure
increases above 3.0 psig.
B. Both NS pumps will restart automatically when their respective
sump suction valve (NS-18A, NS-1 B) reaches full open.
"
C. When the sump suction valves (NS18A, NS-1B) reach full open,
NS pump 2A will restart automatically and the operator can start
NS pump 2B manually.
D. When the sump suction valves (NS18A, NS-1B) reach full open,
the NS pumps will not restart automatically or manually.
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- . For Official Use Only Ques 272a NRC EXAM 12/12/97 !
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Question #25 CATAWBA NUCLEAR STATION SRO EXAM J
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i Bank Question: 273 Answer: C
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1 Pt(s) Unit 1 is shutdown in mode 5. Given the following plant conditions:
)
e' . The VP system is in operation !
. Both trains ofSSPS are in TEST I
Which one of the following signals will shutdown the VP system?
A. Snsignal
B. EMF-38 trip 2 signal
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C. EMF-39 trip 2 signal
D. EMF-40 trip 2 signal
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For Official Use Only Ques 273a NRC EXAM 12/12/97
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Question #26 CATAWBA NUCLEAR STATION SRO EXAM !
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Bank Question: 274 Answer: D
1 Pt(s) What feature in the KF pump discharge line into the spent fuel pool prevents
draining down the water and uncovering fuel assemblies in the event of a break
in the discharge line?
A. The discharge line penetration of the spent fuel pool is just below
the surface of the water and would uncover in the event of a
rupture. -
B. The discharge line has a reverse flow check valve that will prevent
spent fuel pool water from back-flowing through the line.
C. The discharge line is sized to ensure that manual isolation could
be accomplished prior to draining enough water to uncover the
tops of the fuel assemblies
D. The discharge line has holes below the water line to prevent
draining the water out of the spent fuel pool.
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For Official Use Only Ques 274a NRC EXAM 12/12/97
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Question #27 CATAWBA NUCLEAR STATION SRO EXAM
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.__ Bank Question: 275 Answer: A
1Pt(s) Which one of t'e following describes the flow path of the feedwater into the
Unit 1 Steam Generator main nozzles at 5% power?
A. . 100% flow directed into the auxiliary feed nozzle with a small
amount of flow being drawn out of the main feed ring and
returned to the main condenser to maintain CF containment l
penetration temperatute. l
B. 100% flow directed into the main feed ring with a small amount of
tempering flow into the auxiliary feed nozzle to cool the CA nozzle.
C. 100% flow directed through the main feed ring with flow directed
downwant then back upward through the prehoter counter-flow
section.
D. The feedwater flow is split with 15% of the flow is directed into
the auxiliary feed nozzle while 85% of the flow is Jirected into the
main feed ring.
.
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For Official Use Only Ques 275a NRC EXAM 12/12/97
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Question #28 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 276 Answer: C
1Pt(s) Unit 1 is responding to a faulted steam generator inside containment. Given
the following conditions:
. Containment pressure = 4 psig
. NC system pressure = 1742 psig
* MSIVs closed on a low steam line pressure signal
Which one of the following actions will allow main steam isolation valves to
reopen.
A. Reset both trains of main steam line isolation.
B. Reset Phase B signal and both trains of main steam line isolation.
C. Block low steam line pressure signal and reset both trains of main
steam line isolation.
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D. Reset safety injection and reset both trains of main steam line
'
isolation.
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For Official Use Only Ques 276a NRC EXAM 12/12/97
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Question #29 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 277 Answer: C I
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iPt(s) Which one of the following fhilure conditions will cause a lock out relay
actuation for the incoming 6.9KV breaker from IT2A to ITA? *
: A. Under-frequency !
B. . Under-voltage i
C. Ground fault
D. Loss of control power )
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For Offidal Use Only Ques 277a NRC EXAM 12/12/97
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QUESTTCN #33 CATAWBA NUCLEARSTATION SRO EXAM l
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Bank Question: 279 Answer: D ?
1 Pt(s) Unit 1 is operating in mode 3 shutting down for refueling. A sun eiDance is in
progress on the 1 A EDG. Given the following plant conditions:
=
1 A EDG is operating in parallel with off-site power
.
A safety injection signal is received
1
.
Which one of the following events will occur?
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A. The diesel generator breaker will remain closed, unless an under-
voltage condition occurs, and non-LOCA loads on 1 ETA will be
tripped.
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B. The diesel generator breaker will remain closed anii s load shed of
IETA will occur. The LOCA sequencer will seque' nce loads on.
O C. The diesel generator breaker will trip open and a load shed of
IETA will occur. The LOCA sequencerwill sequence loads on.
D. The diesel generator breaker will trip open and non-LOCA loads
on IETA will be tripped. The LOCA sequencer will sequence
loads on.
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Ques 279a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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,_.- p l Question #31 CATAWBA NUCLEAR STATION SRO EXAM .
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Bank Question: 280 Answer: C
1Pt(s) Which one of the following conditions was EMF-48 (NC monitor) designed to
detect? .
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A. High N16 gamma activity '
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B. . High coolant gaseous activity
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C. Fuel cladding failure
D. Crud burst
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For Official Use Only Ques 280a .NRC EXAM 12/12/97
Question #32 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 281 Answer: C
1Pt(s) Which one of the following supplies power to the Circulating Cooling Water
(RC) Pumps?
A. 600 VAC Unit Power System
B. 4160 VAC Essential Power System
C. 6.9 KVAC Unit Auxiliary Power
D. 13.8 KVAC Nonnal Auxiliary Power
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For Official Use Only Ques 281a NRC EXAM 12/12/97
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1 l l Question #33 CATAWBA NUCLEAR STATION SRO EXAM
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Y Bank Question: 282 Answer: C
1Pt(s) Which one of the following statements describes the VI system response to a
loss ofheader pressure?
A. The Backup Temporary / Diesel Air Compressor auto starts
80 psig- VS-78 (VS supply to VI) opens
,
76 psig - VI-500 (VI supply to VS) closes
B. The Standby Air Compressor auto starts
80 psig- VS-78 (VS supply to VI) opens
C. The Standby Air Compressor auto starts
80 psig - VI-500 (VI supply to VS) closes ,
'
76 psig- VS-78 (VS supply to VI) opens )
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D. 80 psig- VS-78 (VS supply to VI) opens
76 psig- VI-500 (VI supply to VS) closes
s,
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For Official Use Only Ques 282a NRC EXAM 12/12/97
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Question #34 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 283 Answer: B
1Pt(s) A fire has occurred in the plant. Fire header pressure drops in response to
demands for water. Given the following conditions. ;
1
. The auto-start pressure switch to the "A" fire pump located in the Service
Building malfunctions and fails to actuate.
. The "B" fire pump is tagged out of service for maintenance
. The 4160 blackout switchgear 2FTA was deenergized by a ground fault
e Pressure continues to drop in the fire header
Which one of the following statements describes the correct starting sequence
for a fire pump to pressurize the fire header?
A. The "A" fire pump auto-starts at 92 psig
B. The "A" fire pump auto-starts at 70 psig
C. The "C" fire pump auto-starts at 92 psig
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D. The "A" fire pump must be manually started by an operator
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For Official Use Only Quas283a NRC EXAM 12/12/97
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Question #35 CATAWBA NUCLEAR STATION SRO EXAM
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*) Bank Question: 284 Answsr: A l
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1 Pt(s) Unit 1 is shutdown in mode 5. Given the following conditions: I
e Both trains ofND are operable
* ND train "A" is in operation
. All four NC loops are filled
. Steam Generator Levels are as follows:
S/G A S/G B S/G C S/G D
0% 15 % 15 % 10 %
Mechanical maintenance requests permission to deenergize the IETB 4160V
bus for a work order. What action (s)(if any) are required to allow
maintenance to proceed?
A. Maintenance may proceed without further changes in plant
conditions.
B. Raise the ID S/G Icvel to 15%
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C. Start IB or IC NC pumps !
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D. Maintenance cannot deenegize IETB under these conditions. i
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For Official Use Only Ques 284a NRC EXAM 12/12/97
Question #36 CATAWBA NUCLEAR STATION SRO EXAM
,) Bank Question: 285 Answer: C
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1Pt(s) Unit 1 is responding to a LOCA inside containment. Given the following ;
conditions: l
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.* Reactor trip and safety injection actuated
e Containment pressure peaked at 2 psig i
e FWST level dropped to 36% i
e EMF 46 A trip 2 alarmed
. Low-low level alarm on KC surge tank A l
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. The operators remain in the control room (control not transferred to the
ASP)
Which one of the following KC system loads will still have KC flow under
these circumstances?
A. NC pump thermal barrier heat exchangers
B. NCDT Heat Exchanger
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C. ND heat exchanger
l D. Letdown heat exchanger
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For Official Use Only Ques 285a NRC EXAM 12/12/97
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Quesbon #37 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 287 Answer: D
1Pt(s) Unit I was operating at 80% power when a reactor trip occurred. Given the
following conditions:
* Reactor trip breaker "A" will NOT open '
. Turbine impulse channel 11 has failed at 80% (as is) t
. The Steam Dump Mode Select Switch is in the Tave position
Which one of the following combinations states the response of the Steam
l Dump Control System dump valves to these events?
Atmospheric ' Condenser
Dumps Dumps
. .
A. Opened Opened ;
B. NOT Opened Opened
i. C. Opened NOT Opened
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D. NOT Opened NOT Opened I
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For Official Use Only Ques 287a NRC EXAM 12/12/97 !
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Question #38 CATAWBA NUCLEAR STATION SRO EXAM
- 4,
Bank Question: 288 Answer: A
1Pt(s) Unit 1 is operating at 9% power and preparing to increase load after a stanup.
Which one of the following conditions or signals will cause a main turbine trip?
A. IA SK; level-84%
B. Pmsurizer perssure = 1940 psig
C. Pressurizerlevel = 95%
D. Loss of the I A and IC-NCPs.
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For Official Use Only Ques 288a NRC EXAM 12/12/97
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Question #39 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 292 Answer:D
1Pt(s) Which one of the following events will most likely cause the operators to j
implement FR-P.1 (Response to Imminent Pressurized Thermal Shock l
Condition) to mitigate an actual PTS challenge during the first 10 minutes of l
the event? )
A. Excessive CA flow while shutdown in mode 3
B. Steam generator tube rupture in mode 1
C. Design basis lay,e break LOCA in mode 1 i
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D. Main' Steam line rupture in mode 1
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For Official Use Only Ques 292a NRC EXAM 12/12/97
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Question #40 CATAWBA NUCLEAR STATION SRO EXAM
.' Bank Question: 293 Answer: C j
1Pt(s)- Unit 1 is conducting a reactor startup when the following events occurred: ;
i
e Reactor was subcritical with power stable at 1.2X10' CPS
.
While withdrawing Control Bank B, one rod dropped to the bottom ofthe
;
core
- Which one of the following statements describes the correct actions? l
A. Secure pulling control bank B. Ensure that the reactor remains
subcritical while recovering the dropped rod.
B. Manually insert control bank B to the bottom of the cort and
recover the dmpped md. Ensure that the reactor remains
subcritical while recovering the dmpped rod. l
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C. Manually insert all contml rods. ;
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D. Manually trip the reactor. j
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For Official Use Only Ques 293a - NRC EXAM 12/12/97
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Question #41 CATAWBA NUCLEAR STATION SRO EXAM ,
...
) Bank Question: 294 Answer: A
1Pt(s) Unit 1 was operating at 100% power when a LOCA occurred into
containment. Given the following events: :
)
e Orange Path on Containment Integrity
. The Operators have implemented FR-Z.1 (Response to High Containment
Pressure)
Which one of the following components will continae to receive KC system
cooling?
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A. NV Pump Motor Coolen
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B. NCDT Heat Exchanger
C. Excess letdown Heat Exchanger
.D. Reactor Coolant Pump Motor Oil Coolen
.
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For Official Use Only Ques 294a NRC EXAM 12/12/97
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Question #42 CATAWBA NUCLEAR STATION SRO EXAM
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._, Bank Question: 295 Answer: A .
1 Pt(s) . Unit I was in the process ofstarting up at 10% power when a loss of offsite
power occurred. Given the following conditions: ;
* The reactor tripped at 0200
e During the accident, a pressurizer PORV stuck open. l
e An operator closed the PORV block valve
e '
A safety injection occurred on low pressurizer pressure
. The operators implemented E-0 at 0200 j
e The operators transitioned to ES-0.2 at 0215 '
* The cooldown was staned at 0230 in ES-0.2 with NC system temperature
indicating 517 *F.
e ' The STA reports that 1here is no void in the reactor vessel head
What is the initial average cooldown rate.that the operators must maintain to
remain within procedurallimits?
I
A. 20 "F/hr
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B. - 50 *F/hr i
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C. 60 F/hr
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D. 100 F/hr -
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For Official Use Only Ques 295a NRC EXAM 12/12/97 <
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Question #43 CATAWBA NUCLEAR STATION SRO EXAM
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i Bank Question: 296 Answer:D
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1Pt(s) - Unit 1 is operating at 100% power when a large rupture of the KC Essential
Header 1 A occurs. Given the following conditions:
. KC Pumps lAl and 1 A2 are mnning
. The KC system trains are cross-connected j
= All KC supply and return isolation valves are open j
Assuming no operating action is taken, which one of the following sequences
will occur automatically?
A. The ND heat exchanger inlet valve closes if the train-related KC !
surge tanklevels decrease to 34% i
B. The KC essential headers will isolate if their train-related KC
surge tank levels decrease to 37%.
C. The KC non-essential headers will isolate if FWST level decreases
to 37%.
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D. The KC non-essential headers will isolate if their train-related KC
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surge tank levels decrease to 34%.
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For Official Use Only Ques 296a NRC EXAM 12/12/97
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' Question #44 CATAWBA NUCLEAR STATION SRO EXAM
) Bank Question: 297 Answer: B
1Pt(s) Unit 2 emergency boration is manually initiated in the plant locally from which
one of the following locations?
A. 2B NI Pump Room (Rm #244)
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B. Unit 2 543' Mechanical Penetration Room (Rm #227) l
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C. 2B NV Pump Room (Rm #241) l
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D. Unit 2 577' Mechanical Penetration Room (Rm #427) l
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For Offidal Use Only Ques 297a NRC EXAM 12/12/97
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Question #45 CATAWBA NUCLEAR STATION SRO EXAM
') Bank Question: 298 Answer: A
1 Pt(s) Unit I was operating at 100% power. Given the following conditions:
. Pressurizer pressure controller is selected to "2&3"
. _ Pressurizer pressure controls are in AUTO
e Pressurizer pressure channel UI detector fails LOW
Which one of the following describes the plant response with no operator
action?
A. High pressurizer pressure reactor trip will occur.
B. PORY NC-34A will maintain NC system pressure 80 to 100 psig
above normal.
C. No effect on NC system pressure but PORVs NC-32B and
NC-36B will be blocked.
D. PORY NC-34A will maintain NCS pressure from 100 psig above
normal to 50 psig below normal.
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For Official Use Only Ques 298a NRC EXAM 12/12/97
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QUESTION #A6 CATAWBA NUCLEARSTATION SRO EXAM
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Bank Question: 299 Answer: A
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1 Pt(s) Unit 1 is operating at 50% power. Given the fo'Jowing cond'tions:
Ise Main' Condenser Vacuum
0200 j
25 in Hg
-0205 24in Hg
,, 0210 22 in Hg
)
0215 21 in Hg
What action (s) (iiany) r.re the operators directed to take under these
;
, conditions?
.
. .. A. Manually trip the main turbine at 0210. -
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B.
Wait until 0215 w hen condenser vacuum has decreased below the
, automatic turtine trip setpoint before performing a manual
turbine trip.
C. Wait until 0215 when condenser vacuum has decreased below the
automatic turbine trip setpoint before performing a manual
reactor trip and verifying turbine trip.
D. Do NOT manually trip the turbine at this power level unless
t'
turbine exhaust hood pressure has exceeded 225 *F.
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Ques 299a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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^ QUEST 10N #47 CATAWBA NUCLEARSTA170N SRO EXAM
Bank Question: 300 Answer:A
1 Pt(s) Unit I was responding to a steamhne break inside containment on the 1C S/G !
per E-2 (Faulted Steam Gene.stor).
Which one ofthe following actions statements correctly dese:ibes the expected
method for isolating steam to the CAPI assuming no component failures.
. -.
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A. Manually close the maintenance isolation valve (ISA-4)in the
doghouse.
B. Manually close the stop-check valve (ISA-6)in the mechanical <
penetration room.
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C. Manually close the IC MSIV and bypass valve. -~~~'
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, D. Manually actuate SM isolation.
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Ques 300a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #48 CATAWBA NUCLEAR STATION SRO EXAM l
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Bank Question: 301 Answer: A
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1Pt(s)' During step 23 of ECA-0.0, the operators are directed to depressurize intact
S/Gs to 165 psig. What prevents exceeding the FR-P.1 criteria for pressurized
thermal shock (PTS) during this cooldown?
A. The S/G depressurization is stopped at a point where NC system
temperature will be maintained above the minimum tempenture
for PTS.
B. The natural circulation cooldown rate is limited by the S/G
depressurization rate to a rate which is not a PTS concern.
C. The NC System will be depressurized below the minimum
pressure where PTS is no longer of concern.
D. The natural circulation cooldown rate is much lower than forced
cirrulation cooldown rates however, if the criteria for FR-P.1 is
exceeded, this procedure will be implemented to pmvide direction
to mitigate the impact of PTS.
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For Official Use Only- Ques 301a NRC EXAM 12/12/97
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Question #49 - CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 303 Answer: C l
1Pt(s) Unit 1 is operating at 100% steady state power. Given the following
i
conditions:
. KC pumps 1 Al and 1 A2 are mnning
. KC pumps 1B1 and 1B2 arein standby
. EMF-46A (KC Surge Tank Rad Monitor) is in alarm
. KC surge tank levels indicate 76% and trending up slowly
Which of the following components is the most likely source of the leak? I
i
A. Seal water heat exchanger
.
B. NC het leg sample heat exchanger l
C. NV letdown heat exchanger
D. NCP motor cooler
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For Official Use Only . Ques 303a NRC EXAM 12/12/97
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Question #50 CATAWBA NUCLEAR STATION SRO EXAM I
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) Bank Question: 304 Answer: C
{
lPt(s) Unit i is shutdown in a refueling outage. A small fire in the SSF causes the
activation of the automatic fire suppression system in the vicinity of the fire.
This action immediately extinguishes the fire. Which one of the following
statements describes the FIRST fire protection alarm response to this event.
A. The contml room will first receive a fire protection alarm after the
main fire pump starts.
B. The contml room will first receive a fire protection alarm when
the RFY pressurizing tank pressure decreases to 128 psig
admitting Nitrogen to the tank.
C. The control room will first receive a fire protection alarm as soon
as the fusible link melts in the sprinkler and flow begins.
D. The contml mom will first receive a fire protection alann when
the SSF CARDOX system actuates.
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For Official Use Only Ques 304a NRC EXAM 12/12/97
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Question #51 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question:305 Answer: D ,
1Pt(s) How is assured mak.eup for the NW System provided during a LOCA?
A. VM is automatically provided on a low-low surge tank level
coincident with a phase A isolation.
B. YM is automatically pmvided on a low-low surge tank level or
low-low surge tank pressure coincident with a phase A isolation.
C. RL is automatically pmvided on a low-low surge tank level
coincident with a phase A isolation.
D. RN is automatically provided on a low-low surge tank level or low-
.
low surge tank pressure coincident with a phase A isolation.
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For Official Use Only Ques 305a ' NRC EXAM 12/12/97
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Quesbon #52 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 306 Answer: C
1 Pt(s) . Unit.1 is responding to a Station Blackout that lasted for one hour. The TSC
is not yet operational. Given the following conditions:
. Both plasma display control panels for the inadequate core cooling monitor
are out ofservice and core exit thermocouple data is not available
. There is evidence of a LOCA into the containment through NCP seals
e' Power has been restored to one 4160 VAC safety bus
e The operators have transitioned to ECA-0.1 and are conducting a
cooldown in natural circulation (due to loss of NCP seals).
* The STA is reviewing critical safety function status trees in F-0
e The STA determines that NC system subcooling margin is -2 F -
(superheated) based on a comparison of:
* ' ThehighestT6 indication
. . Saturation temperature for wide range NC system pressure -
determined by using steam tables
e Source range NIs indicate an increasing trend - with some fluctuations
e Based on this determination ofloss of subcooling and indications of
possible loss of reactor vessel level, the STA recommends transitioning to
FR-C.1,
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' Which one of the followmg actions is conect for this situation? l
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REFEPENCES ATTACHED
-A. Immediately transition to FR-C.I. Although core exit
thennocouple data is not available, the indications ofloss of
subcooling with loss of core level is sufficient. l
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' B. Do not transition to FR-C.I. Wit!..,ut having indication of reactor i
vessel level or core exit thermocouples, no transition can be ns..%
Instead, reference actions in FR-C.3 while continuing on in ECA-
0.1. ,
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- C. Do not transition to FR-C.I. Recalculate the subcooling using the !
data book curves- steam tables cannot be used for this
.
, calculation.
, , < D. Do not transition to FR-C.I. Recalculate the subcooling margin l - by using the average of all four T,,, instruments. This is more ,
representative of actual core conditions. 1
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For Offiaal Use Only - Ques 306aL NRC EXAM 12/12/97 !
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' Quesbon #53 CATAWBA NUCLEAR STATION SRO EXAM i
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.) Bank Question: 307 Answer:B ,
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1Pt(s) Unit I was operating at 100% power. Given the following events and
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conditions
1
1
. EMF-48 (NC monitor) trip 2 alarm
. . Activity level of 2 X 10 pCi/ gram (Dose Equivalent I"')in the reactor
coolant
Which one of the following actions are required to correct a high fission
product activity?
A. Purge the VCT with nitrogen
B. Place mixed bed demineralizers in service
,
C. Reduce reactor power below 50% l
D. Add hydrogen to the reactor coolant
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For Official Use Only Ques 307a NRC EXAM 12/12/97
Question #54 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 308 Answer: D
1Pt(s) Unit 1 is operating at 100% power. Given the following conditions:
* Rod controlis in manual
* Control Bank D is at 200 steps
if the rods in control bank D stan stepping out at 8 steps per minute, what one
of the followir.g actions is required at this time?
A. Select Control Bank D on the rod selector switch and manually
insert Control Bank D
B. Select " AUTO" on the Bank Select Switch and see if rod motion
stops
C. Commence emery,ency boration
D. Trip the reactor
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For Official Use Only Ques 308a NRC EXAM 12/12/97 1
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Question #55 CATAWBA NUCLEAR STATION SRO EXAM j
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Bank Question: 309 Answer: C l
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1Pt(s) Unit 1 is operating at 100% power. Given the following conditions: j
e Rod controlis in automatic l
* Control Bank D is at 200 steps
. All other banks are fully withdrawn
. Tm and T,,r are matched
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Which one of the following situations would likely cause a dropped rod?
A. Repeatedly cycling the in-hold-out switch without pausing 2 ,
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seconds between switch movements.
B. Loss of power from one DC power supply
C. A malfunction that triggers a logic error
D. A malfunction that triggen a multiplexer errer
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For Official Use Only Ques 309a NRC EXAM 12/12/97
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Question #56 CATAWBA NUCLEAR STATION SRO EXAM
Bank Question: 310 Answer: D
1Pt(s) A reactor trip has occurred and the crew is in ES-0.1, Reactor Trip Response.
Given the following conditions:
* A RED path on the CSF for Heat Sink has been recogmzed on the OAC
and validated by the STA
* The OSM directs the implementation of FR-H.1 (Response to Loss of
Secondary Heat Sink)
= - The Heat Sink CSF NOW TURNS GREEN before the procedure has been
removed from the case.
Which one of the following actions is correct for these circumstances?
A. . The crew should immediately transition to FR-H.1 and consult
with the TSC or EOF to evaluate applicable actions and transition
back to ES-0.1.
B. The crew should immediately transition to FR-H.1 because once a
RED path has been validated, the associated functional response
procedure must be implemented to completion.
C. The crew should continue in ES-0.1 because guidance to
implement the status trees has not yet been reached in ES-0.1.
D. The crew should continue in ES-0.1 because the CSF procedure is
not considered implemented until the first step is read.
,
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For Official Use Only Ques 310a NRC EXAM 12/12/97
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Quesbon #57 CATAWBA NUCLEAR STATION SRO EXAM :
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Bank Question: 311 Answer: B ;
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iPt(s) Unit 1 is operating at 50% power. Given the following conditions:
. Pressurizer pressureis 2235 psig
. i
Pressurizer Relief Tank (PRT) pressure is 25 psig
. PRT temperature is 115 'F ,
. PRTlevelis 81%
. The PRT is being cooled by spraying from the RMWST
. A pressurizer code safety valve is suspected ofleaking by its' seat
'
What temperature would be indicated on the associated safety valve discharge
RTD ifthe code safety was leaking by?
REFERENCES PROVIDED i
A. 283 *F
1
B. 267 'F
C. 239 F i
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D. 195 'F
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For Official Use Only Ques 311a NRC EXAM 12/12/97
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. QUEST 10N #58 CATAWBA NUCLEARSTATION SRD EXAM
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Bank Question: 312 Answer: C
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1Pt(s) Unit 1 is responding to a small break LOCA inside con:eirment. Given the
following conditions:
. NI pump 1Aisin senice
. NI pump IB has failed to start
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Both plasma displays on the inadequate core cooling monitor have failed
. The operators have implemented E-1
NC Loon A NC_ Loon B NC Loon C NC Looo D
Tu (*F) 579 575 575 579
T u (*F) 574 559 565 .569
~
WRPress (psig)..
- -
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1575 1615
_ _ -.-.
Which one ofthe following statements correctly des:ribes the NCPs?
.
REFERENCES PROVIDED
4
A. NCPs should be tripped in order to prevent adding unwanted
heat to the coolant and causing corr uncovery due to voiding in
the reactor vessel head.
B. NCPs should be tripped to prevent conthiued mass depletion of !
coolant that is being pumped out the break with potential serious
1
. . core uncovery if the pumps should later trip.
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C. NCPs should NOT be tripped because subcooling does not meet
the foidout page criteria and the pumps are sti!J providing
effective core cooling due to high core steam flowrate.
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D. NCPs should NOT be tripped because only one NI pump is in l
service and without adequate NI flow, it snay not be possible to - )
depressurire the plant to the point where accumulators and ND l
pumps can ensure core heat removal.
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Ques 312a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Question #59 CATAWBA NUCLEAR STATION SRO EXAM
.) Bank Question: 313 Answer: C
1 Pt(s) Unit 1 is responding to a large break LOCA into containment. Given the
following conditions:
. NC pressureis 40 psig
. PZRlevelis 0%
. All NCPs have been tripped
. Containment levelis 11 feet ;
* The operators have reached step 6 in ES-1.3 (Transfer to Cold Leg Recire)
which has them align the S/l system for recire.
e. FWST suction valve, FW-27A (ND Pump 1 A Suct From FWST), is open
and will not close manually from the control room
l . FW-55B (ND Pump IB Suct Fromt'WST)is closed
. FWSTlevelis 4% -
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Which one of the following statements describes the correct action (s) to be
l taken? !
A. Dispatch an operator to close FW-27A locally and initiate makeup :
,
to the FWST
!
.) B. Immediately transition to ECA-1.1 (Loss of Emergency Coolant
Recirculation)
C. Stop ND pump 1A
D. Immediately stop both ND pumps
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For Official Use Only Ques 313a NRC EXAM 12/12/97
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Questbn #60 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 314 Answer: D l
1Pt(s) Unit 1 operators are responding to a LOCA from 100% power. They have
reached step 4 of ES-1.1 (Safety injection Termination) which states:
4. Establish VIto containmentasfollows:
_a Ensure 117-77B (VI Cont Isol) - OPEN
_b. Venfy VIpressure- GREATER THAN90psig
Given the following conditions:
1
* VI pressure = 85 psig and decreasing slowly j
* 1VI-77B is open
Which one of the following actions is correct for this condition?
A. Verify that the VS system automatically cross-connects to VI at 76
psig. VI pressurt is only required to be greater than 50 psig for
the PORVs to operate.
B. Verify that the standby air compressor auto starts and raises VI
'
perssure above 90 psig. Do not proceed in ES-1.1 until VI ,
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pressure is greater than 90 psig.
C. INI-438A and INI-439B will automatically isolate the VI header
from the standby Nitrogen supply to allow PORV operation.
D. Follow the RNO and manually align Nitrogen to the PORVs by
opening INI-438A and INI-439B in the control room ,
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For Official Use Only Ques 314a NRC EXAM 12/12/97
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Question #61 CATAWBA NUCLEAR STATION SRO EXAM I
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Bank Question: 315 Answer: B
1Pt(s) Unit 1 is shutdown in mode 3. Given the following plant conditions:
. Train "A" shutdown monitor (SDM) has just been enabled
l
. Automatic blended makeup to the VCT is in progress from IB Reactor !
Makeup Water (RMW) pump and the 1 A Boric acid (BA) pump l
.
The operator is preparing to test Train B SDM Monitor before enabling the
monitor
Which one of the following states the annunciators that will alann if the
operator inadvertently depresses the TEST pushbutton for Train A SDM
Monitor vice Train B SDM Monitor?
A. TRAIN A SHUTDOWN MARGIN ALARM and BA FLOW
DEVIATION
B. TRAIN A SHUTDOWN MARGIN ALARM and TOTAL
MAKEUP FLOW DEVLATION
i
! C. TRAIN A W/R NEUTRON FLUX SYS TROUBLE and BA j
' ...' FLOW DEVIATION
,
D. TRAIN A W/R NEUTRON FLUX SYS TROUBLE and TOTAL
MAKEUP FLOW DEVIATION
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l l l For Official Use Only Ques 315a NRC EXAM 12/12/97 l
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Question #62 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 319 Answer: A
1Pt(s) Unit 1 is conducting a reactor startup. Given the following conditions:
. IR Channel N35 indicates 8X10'" amps
. IR channel N36 indicates 1.1X10-*' amps
* SR channel N31 indicates 7.1X10' CPS
* SR channel N32 indicates 6.8X10' CPS
The N-36 instrument power fuse blows due to an internal fault. Which one of
, the following actions are required?
A. Hold power at present levels until repairs are made.
l B. Continue the startup with no restrictions. l C. Continue the startup but do not exceed 10% therinal power. l ! D. Insert rods and commence a reactor shutdown. !
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For Official Use Only Ques 319a NRC EXAM 12/12/97
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Question #63 CATAWBA NUCLEAR STATION SRO EXAM i
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Bank Question: 320 Answer: B i
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1Pt(s) Unit I was operating at 100% power. Given the following conditions:
. EMF-33 (Condenser Air Ejector Monitor) alarms in trip 2
If all the automatic features operate as designed (without operator
intervention), which one of the following indications will provide the best
indication (most sensitive and timely) to identify the leaking S/G and trend the
l
magnitude ofthe leak?
A. Comparing S/G feed flow to steam flow mismatch
B. Observing F.MF-26,27,28 and 29 (steamline monitors)
C. Observing EMF-34 (S/G sample line monitor)
D. Frisking the S/G blowdown demineralir.er cation column
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For Official Use Only Ques 320a NRC EXAM 12/12/97
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Question #64 CATAWBA NUCLEAR STATION SRO EXAM
*} Bank Question: 321 Answer: A
1Pt(s) In E-3, (Steam Generator Tube Rupture) Enclosure 6 (NC Pressure and
,
Makeup Control to Minimize Leakage) the operators are directed to energize
pressurizer heaters ifthe ruptured S/G level is decreasing and pressurizer level '
is greater than 25% throughout the event. What is the purpose for this action?
A. Maintain pressurizer saturation temperature corresponding to
ruptured S/G pressure to minimize S/G leakage into the NC
system.
B. Maintain pressurizer saturation temperature corresponding to
intact S/G pressure to minimize primary leakage into the S/G.
C. Maintain pressurizer saturation temperatum above the
corresponding ruptured S/G pressure to ensure S/G water does
not flow into the NC system .
D. Maintain pressurizer saturation temperature corresponding to
intact S/G pressure to minimize NC pressure transients.
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For Official Use Only Ques 321a NRC EXAM 12/12/97
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Question #65 CATAWBA NUCLEAR STATION SRO EXAM l
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)- Bank Question: 322 Answer: A
1 Pt(s) Unit 2 is responding to a loss ofmain feedwater accident. Given the following
conditions: i
. Both main feedwater (CF) pumps tripped to initiate the event
. The reactor tripped on Lo-Lo S/G level
. 2B motor driven aux feed (MDCA) pump staned but was air bound and
no flo or discharge pressure was indicated
. 2A h TA pump and the turbine-driven aux feed (TDCA) pump are each
discha. ing approximately 500 gpm flow
Which one of the following sequences describes the condition of the MDCA l
isolation valves to the S/Gs if these pump conditions do not change?
2CA-58A (CA to B S/G) 2CA-46B (CA to C S/G)
A. remain open remain open
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B. close remain open j
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C. remain open close j
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D. close close
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For Official Use Only . Ques 322a NRC EXAM 12/12/97
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Question #66 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 323 Answer: A
1Pt(s) Unit 1 is shutdown, mode 5, draining liquid radioactive waste from the WL
system into an unprotected outdoor storage tank for disposal offsite. Given the
following tank radiochemistry analysis:
I
. Total tank activity = 10.5 Ci with a combined halflife of 50 days
. Tritium activity = 1.5 Ci with a halflife of 12.6 years
. Noble gas activity = .4 Ci with a halflife of 48 hours '
Which one of the following action (s) (if any) is required for these conditions?
REFERENCES PROVIDED
A. No action is required at this time
B. Immediately stop all additions of radioactive material into the j
tank and wait for the tank contents to decay
C. Immediately reduce the tank contents by transferring radioactive
material to another tank.
D. Move the tank into the auxiliary building
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Queston #67 CATAWBA NUCLEAR STATION SRO EXAM i
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;) Bank Question: 324 Answer: D
1Pt(s) Unit 2 was operating at 100% power when a design basis LOCA into
containment occurred. Given the following conditions: j
i
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. 2 EMF-53 A/B (Containment Post LOCA Monitors) are both inoperable ;
'
l . - The area radiation monitors (ARMS) in lower containmmt are alarmmg.
Which one of the following indications would most accurately determine the i
area dose rates inside containment for source tenn assessment? !
A. ARM indications in lower containment
i
B. Reactor coolant filter radiation (2 EMF-5 or 2 EMF-6) monitor !
indications j
C. EMF-54 (Unit Vent Monitor) indications.
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D. Portable instruments readings taken on the containment wall and
appropriately scaled for shielding facton.
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For Official Use Only Ques 324a NRC EXAM 12/12/97
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Question #68 CATAWBA NUCLEAR STATION SRO EXAM
.: Bank Question: 325 Answer; B
1Pt(s) Unit I was operating at 100% with the pressurizer level controller in the 1-2
position. Given the following conditions and events:
. Charging flow reduces to minimum )
e Backup heaters energize
. Pressurizer level decreases BUT NO operator action is taken
. Letdownisolates
. ' All pressurizer heaters deenergize
. Pressurizer level tums and increases to the high level reactor trip setpoint.
Which one of the following failures has occurred to cause this plant response?
A. PZR level channelI has failed LOW
B. PZR level channelI has failed HIGH
C. Auctioneered High Tave signal has failed HIGH
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_ D. Reference level signal has failed to the NO-LOAD value
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For Official Use Only Ques 325a NRC EXAM 12/12/97
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L Quesbon #69 CATAWBA NUCLEAR STATION SRO EXAM l l
) Bank Question: 327 Answer: A
1 Pt(s) Unit 1 is operating at 100%. If an unisolable mpture occurs in the instrument
air (VI) header in the service building, which one of the following conditions
and associated reasons will cause the reactor to trip FIRST?
A. The main feed regulating valves will close causing a S/G low level
trip.
!
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B. The letdown isolation valves will close and the pressurizer will fill :
up causing a pressurizer high level trip
;
C. The pressurizer spray valves will fail open causing a pressurizer ' j
low pressure trip l
D. The auxiliary spray valve will fail open causing a pressurizer low
pressure trip.
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For Official Use Only Ques 327a NRC EXAM 12/12/97
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Quesbon #70 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 328 Answer: B
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1 Pt(s) Unit 1 is shutdown in mode 5 on the midnight shift. .
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. The control room operators consist of the following personnel
,
. The OSM
!
. The Control Room SRO
. The Unit I and Unit 2 ROs - who are also the OATCs
. The Unit 1 BOP operator is currently securing a containment air j
release. i
. All other on-shift operators are out of the control room at the present
, , f time.
An I&E Tech asks the Unit 1 RO to stand by MC-13 in the control room
! during surveillance testing to acknowledge annunciator alarms for EMF-39
ACOTs. Which one of the following statements is correct in regards to the
RO's ability to comply with this request?
l l A. The RO may proceed to this area of the control room as it is ) ! within the normal surveillance area. !-
B. The RO may only enter this area to acknowledge alarms for a
'~
short period of time as it is part of the limited surveillance area.
C. The RO'may enter this area if he/she first temporarily turns over
the OATC position to the Unit 2 RO prior to entering this area.
D. The RO may enter this aru if he/she first must turn over the
OATC position to the Comrol Room SRO prior to entering this {
area.
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. For Official Use Only Ques 328a - NRC EXAM 12/12/97
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Question #71 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 329 Answer: D
1 Pt(s) Unit I was operating at 100%. Given the following conditions:
* NC pressure = 2235 psig
* A pressurizer PORV opened
* The operators immediately closed the PORV in manual
What actions are required by Tech Specs within one hour?
A. Maintain the PORV closed in manual.
B. Maintain the PORV closed in manual and remove power femm the
PORV.
C. Close the associated PORY block valve.
D. Close the PORV block valve and remove power from the block
valve.
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For Official Use Only Ques 329a NRC EXAM 12/12/97
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, . . i Question #72 '
CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 330 Answer: A
1Pt(s)
The NCPs are limited to 3 consecutive stans in any 2 hour period with an
additional requirement of a minimum idle period of 30 minutes between
restans. What is the reason for this limitation?
A.
This is an engineerino, restriction to prevent overheating the
motor windings due to high starting currents.
,
B.
l
This is an operational restriction to that assures that the oil
!
! temperature will decrease to design specifications between restart
, anemph.
C.
This is an operational restriction to allow the NCP seals to fully
rescat between NCP rotations.
D.
This is an administrative restriction that prevents operaton from
restarting without a deliberate approach to ensure that all
i
precautions and intedocks have been satisfied.
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For Official Use Only Ques 330a NRC EXAM 12/12/97
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Question #73 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 334
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Answer; B
1Pt(s) During a rod swap as part of a Zero Power Physics Test (ZPPT) an operator
was withdrawing shutdown bank B, as directed by the Test Coordinator.
)
'
During the withdrawal sequence, the Test Engineer became distracted and
failed to properly observe the indication on the special reactivity computer. A l
reactivity excursion occurred when the operator withdrew the shutdown bank j
beyond criticality. The highest start up rate observed was 2.5 DPM when the
operators inserted rods to stop the power increase.
Which one of the following reactor trips would have automatically terminated
the rod withdrawal if no operator action had been taken? j
A. Source Range Hi Flux
B. Intermediate Range High Rux
;
C. Power Range High Flux
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D. Intermediate Range Hi Startup Rate ,
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For Official Use Only Ques 334a NRC EXAM 12/12/97
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Question #74 CATAWBA NUCLEAR STATION SRO EXAM :
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Bank Question: 337 Answer: C
1Pt(s) Unit 1 just completed a shutdown to mode 5 with both ND trains in service l
prior to core ofiload.
Which one of the following statements is correct regarding an approved reactor
coolant vent path into containment to mitigate the consequences of a loss of
ND cooling?
l
.
A. An open reactor head vent will provide an approved vent path in
mode 5.
B. An open S/G cold leg manway with the hot leg nozzle dam
installed, and the cold leg nozzle dam removed will pmvide an
approved vent path. !
C. An open S/G hot leg manway with the hot leg nozzle dam not
installed will provide an approved vent path.
D. An approved vent path is not required until the reactor head is
removed in mode 6.
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For Official Use Only Ques 337a NRC EXAM 12/12/97
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Question #75 CATAWBA NUCLEAR STA110N SRO EXAM
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) Bank Question: 338 Answer: C
1 Pt(s) Which one of the following statements complies with the requirements of OMP
l-7 regarding the mles of usage for abnormal procedures (APs) when the
EOPs have been implemented?
A. APs may not be implemented when EOPs have been entered. 5
B. Only one AP at a time may be implemented when EOPs have
been implemented. Concurrent implementation of APs when
EOPs an in use is not allowed.
C. APs may be implemented concurmutly with EOPs. However, the
APs were written assuming that SI has not actuated and operators
must be careful when using APs if SI has occurred.
D. APs may be implemented concurrently with EOPs with the
exception of events where SI has actuated. APs were written
assuming the Si had not occurred and cannot be used if SI has
actuated.
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For Offcial Use Only Ques 338a NRC EXAM 12/12/97
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Question #76 - CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 341 Answer: A ,
,
1Pt(s) The foldout page in E-1, Loss of Reactor or Secondary Coolant, requires
NCPs to be tripped when NC system subcooling margin is less than 0 F. This
step provides protection against a specific scenario. Which one ofthe l
following statements describes this scenario?
A. Operating the NCPs causes excessive mass depletion in the NC
system which will greatly increase the degree and duration of fuel
uncovery i.f the pumps were lost resulting in fuel peak centedine
temperatures (PCTs) exceeding FSAR limits. 1
B. The depressurization of the NC system below shutoff head of the i
ND pumps will cause the NCPs to suddenly pump large quantities
of cold injection water into the hot core and cause excessive
thermal shock leading to clad or fuel damage.
i
C. A loss of NCPs due to steam binding after subcooling margin is i
lost precludes their use later dudng recovery operations when use l
of NCPs may be critical.
D. Continued degradation of NC system conditions due to the
"
heating of the remaining NC system water causes the NCPs to
become a heat source and add to the superheated condition of the
steam in the core causing steam blanketing of the fuel and
excessive peak centedine temperatures (PCTs),
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For Official Use Only Ques 341a NRC EXAM 12/12/97
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Question #77 CATAWBA NUCLEAR STATION SRO EXAM
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_l Bank Question: 342 Answer: C
1Pt(s) Unit 1 is in mode 4 cooling down to mode 5. Given the following conditions:
* The 1 A ND pump is in sersice
.
The 1B ND pumpis operable but secured
. NCPs are secured
Parameter Looo A LoooB LoooC Loop D
S/G level (%) 69 72 68 65
S/G Press (psig) 125 127 125 126
T a ( F) 281 277 280 278
T w ( F) 289 288 290 290
NC Press (psig) 45 psig 44 psig
If the 1 A ND pump stops due to a pump bearing failure, which one ofthe
following statements describes the required Tech Spec action (s) within the first
hour?
REFERENCES PROVIDED
.m
. ,' A. No immediate action is required for one hour during which time
no operations are pennitted that would dilute the NC system.
B. Immediately start one NCP.
C. Immediately initiate corrective action to start the 1B ND pump. :
D. No action required because at least 2 S/G levels are greater than
12 %
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For Official Use Only Ques 342a NRC EXAM 12/12/97
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Question #78 CATAWBA NUCLEAR STATION SRO EXAM
.. Bank Question: 343 Answer: B
1Pt(s) Unit 1 is responding to a LOCA inside containment. Which one of the
following situations describes the case where the operators should use
ES-1.2 (Post LOCA Cooldown and Depressurization) to bring the plant into
cold shutdown?
A. A small break LOCA with a loss of containment sump level due to
a ruptured FWST.
B. A small break LOCA where NC system pressure is above shutoff
head for the ND pumps.
C. A lary,e break LOCA where NC system pressure is below shutoff
head for the ND pumps.
D. A lary,e break LOCA with a loss of both NI pumps.
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, i For Official Use Only Ques 343a NRC EXAM 12/12/97 !
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Quesbon #79 CATAWBA NUCLEAR STATION SRO EXAM
) Bank Question: 344 Answer: D l
1Pt(s) Which one of the following procedures does NOT provide a direct transition
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path into ECA-1. l(Loss of Emergency Coolant Recirculation)
,
!
A. Transition from E-1 (Loss of Reactor or Secondary Coolant) when
recirc sump isolation valves cannot be opened.
B. Transition from ES-1.3 (Transfer to Cold Leg Recirculation) '
when at least one flow path from the sump cannot be established
or maintained.
C. Transition from ECA-1.2 (LOCA Outside of Containment) when
a LOCA outside of containment cannot be isolated.
D. Transition from ES-1.1 (SI Tennination) when NC system
pressure is greater than shutoff head for the ND pumps.
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( ! ! !- For Official Use Only Ques 344a NRC EXAM 12/12/97 l
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Quesbon #80 CATAWBA NUCLEAR STATION SRO EXAM
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'3 Bank Question: 345 Answer: C
1Pt(s) Unit I was operating at 100% power steady state, when a loss of feedwater
occurred. Given the following events and conditions:
.
A condensate header rupture caused a loss of main feedwater pumps
. The reactor tnpped
. Safetyinjection actuated
. The turbine and motor driven CA pumps failed to start
i
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The operators entered FR-H.1 but were unsuccessful in restoring
feedwater to any S/Gs
. Feed and bleed wasinitiated
.
Core exit T/Cs are now reading 355 F and slowly decreasing
i l
Which one of the following emergency event classification is correct for the
conditions described above?
REFERENCES PROVIDED
A. Notification of Unusual Event
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B. Alert I
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C. Site Area Emergency
l= D. General Emergency I , !
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For Official Use Only Ques 345a NRC EXAM 12/12/97
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Queston #81- CATAWBA NUCLEAR STATION SRO EXAM
)- Bank Question: 346 Answer: D
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1 Pt(s) Unit 1 operating at 100% power. Given the following events and conditions: I
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. Engineering has issued an inoperability statement for the following S/G
code safety valves:
!
. SV-2
. SV-8
. SV-10 !
. SV-14
. The valves must be removed for repair.
Management has decided to remain at the highest power level allowable due to l
a power emergency on the grid. What actions are required to continue
operations at power?
!
l REFERENCES PROVIDED 1
A. Reduce power and reset the power range neutrun flux high trip
j setpoint to 87%.
B. Reduce power and reset the power range neutron flux high trip
setpoint to 65%.
l l C. Reduce power and reset the power range neutron flux high trip
setpoint to 58%.
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D. Reduce power and reset the power range neutron flux high trip
,
setpoint to 41%.
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For Official Use Ordy Ques 346a NRC EXAM 12/12/97
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Question #82 CATAWBA NUCLEAR STATION SRO EXAM
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') Bank Question: 347 Answer: C '
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1Pt(s) Unit I was conducting refueling. The refueling team dropped a fuel element )
into the reactor vessel. Given the following events and conditions at 0200: l
. 1 EMF-17 Trip 2 alarm
* Containment evacuation i
+ Containment closure
0210 The OSM declared an alert.
0220 The State and Local Authorities were notified of the emergency.
Assuming that no personnel were injured and no further damage was done, l
what is the latest time that NRC notification must occur by?
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REFERENCES PROVIDED
A. 0225
B. 0300
C. 0310
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D. 0320
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Question #83 CATAWBA NUCLEAR STATION SRO EXAM
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i Bank Question: 348 Answer: D
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1Pt(s) ~ Unit I is operating at 100% power when an engineer informs the OSM that he
just reviewed a Tech Manual change to the Emergency Diesel Generator that ;
requires replacement of the hydraulic oil in the Woodward govemor. The 1
engineer reports that type of oil that is presently installed in both EDG
govemors is not known but may have a tendency to break down over time and
cause the governor to malfunction.
Which one of the following actions is proper for the OSM under the above
circumstances?
A. Immediately declare both EDGs inoperable, enter Tech Spec 3.0.3
and commence a shutdown. Document the decision by entenng j
the LCO in the TSAIL. '
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B. Direct the engineer to investigate the problem further and report
back when he has more specific answen. Document the problem
in the TSAIL and assure that it is logged in the control room
logbook but do not declare the EDGs inoperable.
l C. Immediately call the Station Manager and report the problem. '
Wait for further direction by senior management before taking
any action. Document the problem by entering it in the control
room logbook and in the shift turnover worksheet.
D. Begin the operability evaluation process of the EDGs to perform i
their required FSARfrech Spec function and determine if an
unresolved safety question has been identified or Tech Spec LCO
has been exceeded. Initiate a PIP to document and control the
process.
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For Official Use Only Ques 348a NRC EXAM 12/12/97
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Question #84 CATAWBA NUCLEAR STATION SRO EXAM
a
_) Bank Question: 349 Answer: A
1Pt(s) Which one of the following statements correctly describes the status of
Selected Licensee Commitments (SLCs) and the process by which they may be
changed.
A. SLCs are commitments made to the NRC that require control but
are not appropriate in Tech Specs because they do not involve
design bases accidents. They are part of the FSAR and may be
changed using the 10CFR50.59 process.
B. SLCs are comraitments made to the NRC involving beyond design
bases accidents that are incorporated as a subpart of Tech Specs.
They are changed using the Tech Spec change process which
requires NRC appmval.
C. SLCs are commitmenu made to state and local authorities
regarding emergency planning issues. They may be changed by
modification of the memorandum of understanding between the
cogniz. .it governmental agency and the company.
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D. SLCs are commitments made to other federal agencies besides the
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NRC, such as FEMA, EPA or OSHA. They may be changed with
approval fmm the cognizant federal agency.
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For Official Use Only Ques 349a NRC EXAM 12/12/97
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Queston #85 CATAWBA NUCLEAR STATION SRO EXAM
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) Bank Question: 350 Answer: D ?
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1Pt(s) Unit 1 is in mode 5 two days after a shutdown. Given the following events and !
conditions: )
. NC temperature u 150 F
.
Pressurizer level = 70% (cold cal)
. Both EDGs are operable
. NV pump 1 A is operable ;
= NV pump 1B is tagged per the shutdown procedure
. NI pump 1 A motor is removed for maintenance
. NI pump IB is functional
. Both trains ofND are operable ;
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If IB NI pump breaker is being racked out for maintenance, the IB NV pump
is currently _(1)_and is required to be _(2)_.
A. (1) Functional (2) Operable
B. (1) Functional (2) Functional
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.. C. (1) Available (2) Functional
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D. (1) Available (2) Available
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For Offidal Use Only Ques 350a NRC EXAM 12/12/97
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Question #86 CATAWBA NUCLEAR STATION SRO EXAM
> Bank Question: 351 Answer:D 1
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1 Pt(s) Tech Spec 3.1.1.4 requires NCS temperature to be above 551 F whenever the l
reactor is made critical. What is one of the bases for this requirement?
A. Ensures that reactivity transients associated with cold water
addition accidents are within acceptable limits. l
B. Ensures that the reactor vessel is below its maximum RTsyr
temperature.
i
C. Ensures that moderator temperature coefHeient always remains a
negative value,
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D. Ensures that the tractor trip instrumentation is within its nonnal
operating range.
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For Offidal Use Only Ques 351a NRC EXAM 12/12/97
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Question #87 CATAWBA NUCLEAR STATION SRO EXAM l
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,) Bank Question:352 Answer: A
1Pt(s) Prior to moving fuel assemblies into their final storage positions in the spent
fuel pool, the Staff Engineer responsible for refueling desires to troubleshoot a
problem on the Spent Fuel Pool Manipulator Crane Bridge which requires i
bypassing the" bridge left bypass" l
The statements below have been ranked in order from the lowest approval level .
to the highest approvallevel.
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Which one of the following statements is the first one to correctly describe the
minimum approvals and notifications authorized in Site Directive 3.1.17 for
-
bypassing thisinterlock?
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! A. The Staff Engineer can approve his own request to bypass the !
intedock for testing as long as no fuel assemblies, insert ;
components or dummy fuel assemblics are being handled.
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l B. Verbal approval must be obtained from the Fuel Handling SRO I l
and the Control Room RO must be notified.
C. Verbal appmval must be obtained from the OSM (or designee)
--
and any other exempt employee in the operations group. The
Control Room RO must be notified. l
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D. A written procedure must be prepared to control and document
the performance of this special test. No notification of the Control 1
Room RO is necessary,
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For Official Use Only Ques 352a NRC EXAM 12/12/97
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QUESTION 488 CATAWM NUCLEARSTATION 5RO EXAM
Bank Question: 353 Answer: C
1 Pt(s) A worker needs to repack a valve in an crea which has the following
radiological characteristics:
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The worker's present dose burden is 250 mrem for the year.
. General area dose rate = 32 mrenv'hr
*
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Airborne contamination concentration - 10 DAC
Thejob will take 3 hours with a mechanic weadng a full-face respirator. It vdll
only take 2.5 hours if the mechanic does NOT wear the respirator.
Which ofthe following choices for completing thisjob would maintain the
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workers exposure within the Station ALARA requirements?: E .
, . _ . _ .
,
A. The worter should wear the respirator because station RP
1
a practices require personnel to wear respiraton in areas that have i
measurable airborne contamination levels.
B. The worker should NOT wear the respirator because the dose
received will exceed neither NRC nor site annual penonnel dose ;
limits.
C. The worker should wear the respirator because the total TEDE
dose received will be less than if he/she does not wear one.
~
D. The worker should NOT, wear the respirator because the total
TEDE dose received will be greater than if he/she w can one.
Ques 353a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
QUESTION #89 CATAWBA NUCLEARSTATION SRO EXAM
Bank Question: 354 Answer: A
1 Pt(s) Unit 1 is preparing for a waste gas release. Given the following conditions and
events prior to the release:
.
IEMF-50 (Waste Gas Discharge Monitor) = 5 CPM
e
IEMF-36(Unit Vent Monitor)= 50 CPM
.
,
1 RAD-2 (D/5) IEMF 35/36/37 Unit Vent Loss ofFlow-in alarm
.
The Release Rate Determination Fonn has been co:npleted
Which one of the following actions must be completed by operations prior to
starting the release?
, ._.
A. Reset the IEMF-50 trip setpoints to the value stateil hn the
- - - -
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Release Rate Determination Fonn
,, B. Reset the IEMF-50 trip setpoints to 15 CPht
C. Reset the IDIF-36 trip setpoints to the vrdue stated on the
Release Rate Determination Fonn
D. Reset the 3 DiF-36 trip setpoints to 150 CPM.
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! i I , Ques 354a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97 i l
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Question #90 CATAWBA NUCLEAR STATION SRO EXAM
i Bank Question: 355 Answer:B
1Pt(s) Unit 1 operators are responding to a LOCA in E-1, Loss of Reactor or
Secondary Coolant. Given the following events and cor.ditions:
. The STA reports that the critical safety function for heat sink is in a RED
PATH condition.
.
The operators transition into FR-H. l(Response to Loss of Secondary Heat
Sink)
.
After completing the f.rst several steps, it is determined that a valid red path
never existed in heat sink and that no other red or orange paths exist.
If the TSC is still in the process of being activated, which one of the following
actions is the proper way to recover from this situation? !
A. Evaluate the actions and system alignments perforined in
- FR-H.1 and realign these systems as necessary to restore the plant
to the correct condition for E-1. Stop and return to step 1 of E-1.
B. Evaluate the actions and system alignments performed in l
FR-H.1 and realign these systems as necessary to restore the plant
l to the correct condition for E-1. Stop and return to E-1 at the ' ,
step that was in effect at the time of the transition. '
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C. ES-0.0 will provide guidance concerning which procedure should i
be in effect. Stop and transition to ES-0.0, Rediagnosis, to !
l reanalyze the status of the plant and return all systems to the ; i correct configuration.
D. Restarting in E-0 assures that the critical actions and system
l
alignments are not missed. Stop and return to step 1 of E-0,
l Reactor Trip or Safety Injection. 1 I L 1 i !
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For Official Use Only Ques 355a NRC EXAM 12/12/97
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Question #91
CATAWBA NUCLEAR STATION SRO EXAM
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.l Bank Question: 356 Answer: C
1Pt(s)
l If the SRO detemsnes that emergency actions must be taken that are outside ,
the bounds of the approved procedures, what authority does he/she have to
take this action?
A.
The SRO has the authority of 10CFR50.54x to take reasonable
action outside of the approved procedures in an emergency
prvviding such action does not violate the licensing bases of the
plant. ,
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B. The SRO has the authority under 10CFR50.54x to take
reasonable action outside of approved procedures in an
i
emergency even ifit violates the licensing bases of the plant when
there is a declaration of a state of emergency by the Governor of
the State of South Carolina.
C.
* The SRO has the authority of 10CFR50.54x to take reasonable
l action outside of the approved procedurts in an emergency if such
action is necessary to protect the public heath and safety.
D.
! # The SRO has the authority of 10CFR50.54x to take reasonable
action outside of the approved procedures in an emergency if such
action is necessary to protect the public heath and safety provided
that he/she first obtain approval from the Operations Duty
Manager.
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I, r i i " g For Official Use Only Ques 356a NRC EXAM 12/12/97
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Question #92 CATAWBA NUCLEAR STATION SRO EXAM
)- Bank Question: 357 Answer: C
i
iPt(s) Unit 1 is responding to a LOCA into containment. Given the following events
and conditions:
. The operators completed E-0 (Reactor Trip and Safety Injection) and
transitioned to E-1 (Loss of Reactor or Secondary Coolant)
. A RED PATH on NCS Integrity occurred and the operators transitioned
.
to FR-P.1 (Response to Imminent PTS) at step 16 ofE-1.
* Midway through FR-P.1, the NCS Integrity RED PATH turned GREEN.
. A RED PATH on Heat Sink occurred and the operators transitioned to
FR-H.1 (Response to Loss of Heat Sink) from step 4 of FR-P. I.
. The operators perfonned all required actions in FR-H.1 which placed
feedwater back in service.
. Upon completion of FR-H.1, the STA reports that all CSFs are now
GREEN (including Integrity).
Which one of the following describes the correct procedure flow path?
A. Return to E-1 step 1 and continue.
*
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B. Return to E-1 step 16 and continue.
C. Return to FR-P.1 step 4 and complete the procedure, then return
I to E-1 step 16. j' D. Go to ES-0.0 (Rediagnosis) and rediagnose the situation l l ! j
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For Official Use Only Ques 357a NRC EXAM 12/12/97
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Question #93 CATAWBA NUCLEAR STATION SRO EXAM l
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) Bank Question: 358 Answer: D l
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1 Pt(s) Unit I is operating at 100% power. If the OAC computer input for NCP
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breaker status were to fail in the open condition (input is out of service), what
would the CSF status display on the OAC indicate in the control room?
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A. Core Cooling- Orange !
Heat Sink- Green )
NC Inventory- Yellow l
B. Core Cooling-Orange
Heat Sink-Green l
NC Inventory- Magenta l
C. Core Cooling- Magenta j
Heat Sink-Magenta
NC Inventory-Yellow
D. Core Cooling- Magenta
Heat Sink- Green j
NC inventory- Magenta j
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i l l For Official Use Only Ques 358a NRC EXAM 12/12/97 l l
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Question #94 CATAWBA NUCLEAR STATION SRO EXAM j
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1 Bank Question: 359 Answer: A
1Pt(s) Unit 1 is responding to a small break LOCA in E-1. Given the following
events and conditions. ;
i
. EMF-41 (Aux Bldg. Ventilation) = trip 1 !
* Aux Building area radiation monitors are in alarm
.
EMF-53A and B (Containment Rad Monitors) = 25 R/Hr
. Hydrogen Analyzer = 0.4% in contairanent
* Contamment pressure = 2.8 psig
These symptoms provide entry conditions to which one of the following
procedures?
A. ECA-1.2 (LOCA Outside of Containment)in step 12 of E-1.
B. FR-Z.1 (Response to High Containment Pressure)
C. FR-Z.3 ( Re:panse to High Containment Radiation Level)
D. FR-Z,4 (Response to High Containment Hydrogen
Concmtration)
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For Official Use Only Ques 359a NRC EXAM 12/12/97
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Question #95 CATAWBA NUCLEAR STATION SRO EXAM l
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) Bank Question: 360 Answer:D
1Pt(s) Unit 1 is operating at 100% power when the following events occur: ;
)
. Complete loss of offsite power - the operators diagnose immediately {
e EDG 1 A fails to start !
. The output b'reaker for EDG IB fails to close i
* A reactor trip signal occurs - -
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. Shutdown Banks A and B fail to insen at the reactor tiip
. IR SUR = +.05 DPM
*
* Reactor Power = 15%
l
. . Unit 2 offsite power is aligned to lETA through SATA
Which one ofthe following procedural flow paths is correct? j
A. Initially enter E-0 (Reactor Trip or Safety Injection) !
Immediately transition to ECA-0.0 (Loss of all AC Power)
Immediately transition to FR-S.1 (Response to Nuclear Power - ,
'
Generation /ATWS)
When completed, return to ECA-0.0
,
After one 4160 VAC essential bus is restored, retum to E-0
~
B. Initially enter E-0 l
Immediately transition to FR-S.1
)
When completed, return to E-0
Immediately transition directly to ECA-0.0
When directed, transition to ECA-0.1 (Loss of All AC Recovery
without SI Required)
l C. Initially enter ECA-0.0 l Immediately transition to FR-S.1
When completed, transition to E-0
'
Imnaediately transition ECA-0.0
l After one 4160 VAC essential bus is restored, return to E-0 l ! D. Initially enter ECA-0.0
After one 4160 VAC essential bus is energized and ECA-0.0 is
, completed, transition to ECA-0.1
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When directed, transition to FR-S.1
2
When completed, return to ECA-0.1 i
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For Offical Use Only Ques 360a NRC EXAM 12/12/97
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Question #96 CATAWBA NUCLEAR STATION SRO EXAM
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Bank Question: 361 Answer: D
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1Pt(s) Unit I was operating at 100% power. Which one of the following sequences '
require immediate entry into E-0 and immediate transition to FR-S.1, Response
to Nuclear Power Generation.
)
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l A. An I&E Technician is repairing a pmblem in SSPS when he
inadvertently generates an invalid reactor trip signal without the
trip breaken opening.
B. A reactor trip occurs due to a failed power range nuclear
instrument with another power range channelin test. Four rods
do not dmp to the bottom of the core even after a manualizactor
trip is inserted. Flux is dec easing from a level of 3% at rate of
-0.1 DPM.
C. A loss of power to DRPI causes all rod position LEDs to
deenergize. A subsequent tractor trip on low S/G water level
occun and no rod bottom lights are lit. Flux.is decreasing from
level of 4% at a rate of-0.1 DPM.
"
f D. The operaton perform a manual reactor trip from 50% power
<
when rods in control bank D stop moving during a shutdown. ,
- Four rods do not drop to the bottom of the core. Hux is ]
decreasing from a level of 15% at a rate of-0.1 DPM. l
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For Official Use Only' NRC EXAM 12/12/97
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Ques 361a :
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Question #97 CATAWBA NUCLEAR ST,iTION SRO EXAM
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,) Bank Question: 362 Answer: B
1Pt(s) Unit I is operating at 100% power. Given the following events and conditions:
. Normal lineup for the 120 VAC Vital Instrument Busses
. The supply breaker from lEMXA to Battery Charger 1ECA trips open
. Standby battery charger IECS is in standby alignment
. No operator action is taken
Which statement describes the effect of this failure on the 120 VAC Vital
Instrument Bus, IEDA?
A. The IEDA bus is deenegized
B. The bus remains enegized fmm the battery, IEBA
C. The bus is picked up automatically from auctioneering diode
assemblies, IEADA
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D. The bus is automatically picked up by the standby chager IECS. l
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For Official Use Only Ques 362a NRC EXAM 12/12/97 l
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QUESTION #98 CATAWBA NUCLEARSTATION SRO EXAM
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Bank Question: 363 Answer: A
;
1Pt(s) Unit 1 is in mode 6 conducting refueling operations with fuel movement in j
progress. Given the following conditions and events: i
e N-31 = 15 CPS
. N-32 = 14 CPS (aligned to the audio count rate panel) i
* BDMS train A = 15 CPS
.
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e BDMS train B isinoperable j
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If the instrument power fuses blow for N-32, what action must be taken?
A. Immediately suspend all core alterations or positive reacthity
-< .- - .
- - '
changes. Realign the audio count rate panel to'N-31'and ensure )
that NV-230 is closed. ?-- 1
o B. Restore B train BDMS to operation within 1 bour, or suspend all
core alterations or positive reactivity changes. Ensure that
NV-230 is closed.
C. Realign the audio count rate panel to N-31. No fmther actios is i
required with train A of BDMS in service.
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D. Restore N-32 to operation within I hour, or suspend all core
alterations or positive reactivity changes.
'
Distracter Ana'ysis:
I
B. Incorrect: Tech Specs require an audio count rate channel to be in service any time
that core ahcrations are in progress. :
Plausible: Restoring a second train ofBDMS meets Tech Spec requirements for
l monitoring core suberiticality when core alterations a e NOT in progress.
C. Incorrect: Tech Specs require either BOTH trains ofBDMS to be in operation or both
,
rnurce range NTs to be in operation.
Plausible: The candidate believes that one train ofBDMS can be substituted for one
train of SR NIs. This is acceptable at McGuire but not at Catawba.
l D. Incorrect: Tech Specs requires au core alterations to be stopped immediately.
Plausible: The cr.ndidate may not remember that core alterations must be stopped
immediately and think that there is a 1 hour action statement.
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Ques 363a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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Quesbon #99 CATAWBA NUCLEAR STATION SRO EXAM
'
) Bank Question: 364 Answer: C
1Pt(s) Which one of the following statements is the correct bases for Technical
,
Specification 3.6.5.5, Divider Barrier Personnel Access Doors and Equipment !
Hatches?
A. The doon and hatches must be OPEN and OPERABLE to ensure
that a minimum bypass steam now will occur during a LOCA.
l
B. The doors and hatches must be OPEN and OPERABLE to ensure
that a NCS fluid released during a LOCA wiu be diverted through !
the ice condenser hays for heat removal. !
C. The doors and hatches must be CLOSED and OPERABLE to
ensure that a minimum bypass steam Row will occur during a
LOCA. '
D. The doors and hatches must be CLOSED and OPERABLE to
ensure that exceasive sublimation of the ice bed will not occur '
because of wann air intmsion.
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Distracter Analysis:
A. Incorrect: The door and hatches must be closed not open or else the steam flow will ;
bypass the ice condenser.
Plausible: The second part of the distracter is actually part of the basis for the correct
answer. If the candidate does not understand the flow path in the system but has j
memorized the words in the bases, he/she may select this distracter. I
B. Incorrect: The door and hatches must be closed not open or else the steam flow will
bypass the ice condenser.
Plausible: If the candidate does not understand the flow path through the ice
condenser, he/she may think that these doors and hatches must be open to allow air
flow through the ice condenser.
D. Incorrect: The loss ofice due to sublimation is not the bases for this Tech Spec.
l Plausible: There are other Tech Specs that require minimum amounts ofice to be l resident in the ice condenser bays and places limits on the amount of sublimation that '
can be allowed to occur.
.
a
for Official Use Only Ques 364a NRC EXAM 12/12/97
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QUESTlON #100 CATAWB4 NUCLEARSTATION 5RO EXAM
Bank Question: 36S Answer: D
1 Pt(s) Unit 1 is in mode 4. Given the following conditions
.
l A main bus line is tagged out to the Transmission Department
e All 6.9 kV busses emergized
_.. Unit 2 is in Mode I with a nonnat electrical lineup.
Which one ofthe following alignments to provide power to 1 ETA will enere
Unit I remains fully operable.
REFERENCES PROVIDED .
., ... - --
-- -
A. From ITA via 1ATC ~-
,, B. Fmm 2 ETA via 2 ETA-4
C. From ITC via SATA
D. Fmm 2TC via SATA
.
+
Ques 365a NRC OFFICIAL USE ONLY NRC EXAM 12/12/97
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