ML20205P486
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| ML20205P486 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/15/1999 |
| From: | Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-354-98-03OL, 50-354-98-3OL, NUDOCS 9904200194 | |
| Download: ML20205P486 (1) | |
See also: IR 05000354/1998003
Text
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NOTE TO: -NRC Document Control Desk-
Mail Stop 0-5-D-24
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FROM: EC#I b o a le o .L
, Oper#;1ng Licens17 Branch. Rpsing Assistant
SUBJECT: OPERATOR LICENSING EXAMINATION ADMINJSTERED ON
2123/92*- 3 /4 /92' . AT Haar Cece k .
00CKET #50 agy /
On ANff ,3/4/P/ Operator Licensing Examinations were administered
at t'he referenced facility. Attached, you will find the following
information for processing through NUDOCS and distribution to the NRC
staff, including the NRC PDR:
Item #1 - a) Facility submitted outline and initial exam submittal,
designated for distribution under RIDS Code A070,
b) As given operating examination. designated for
distribution under RIDS Code A070.
Item #2 - Examination Report with the as given written examination
attached. designated for distribution'under RIDS Code IE42.
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***** April 28, 1998 I
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Mr. Harold W. Keiser ]
Executive Vice President
Nuclear Business Unit
Public Service Electric and Gas Company
Post Office Box 236
Hancocks Bridge, NJ 08038
SUBJECT: HOPE CREEK EXAMINATION REPORT 50-354/98-03(OL)
Dear Mr. Keiser:
During the period of February 23,1998 through March 4,1998, the NRC administered
initial exams to eleven employees of your company who had applied for licenses to operate
the Hope Creek Generating Station. We also conducted reviews of related information
from March 4 through March 12,1998. At the conclusion of the onsite operating portion
p of the examination, the preliminary findings were discussed with Mr. McMahon and
members of your staff.
The results of the exam are still under review and will be finalized at a later date. We may
issue a revised report when the final results are available.
Your staff developed the initial exams and submitted the exams to NRC for approval.
However, we concluded that your submittal was inadequate to discriminate between safe
and unsafe operators. After postponement of the exam for two weeks and significant
interaction between the NRC staff and your staff, an exam of sufficient adequacy was
developed and subsequently administered to the eleven applicants. In this regard, we
would like to meet with you for your perspective and underlying causea along with
corsective actions.
i
Based on review of information during this exam, the NRC staff had determined that the
I required five significant control manipulations were not acceptable for certain applicants
and the manipulations had to be redone before administration of the examination. We
would like to meet with you on this problem also. /
We wili contact you shortly for a meeting in Region I on these issues. [
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Mr. Harold W. Keiser
In accordance with 10 CFR 2.700 of the NRC's " Rules of Practice," a copy of this letter
and its enclosures will be placed in the NRC Public Document Room (PDR). Should you
have any questi ons regarding this information, please contact me at
(610)337-5211.
I Sincerely,
/d
Richard Conte, Chief
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Operator Licensing
and Human Performance Branch
'
Division of Reactor Safety
,
Docket No.: 50-354
5
Enclosures: Examination Report 50-354/98-03 (OL)
cc w/ encl: w/ Attachments 1-5:
J. McMahon, Director - Nuclear Training
ec w/ encl: w/o Attachments 1-5:
L. Storz, Senior Vice President - Nuclear Operations
,
E. Simpson, Senior Vice President - Nuclear Engineering
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E. Salowitz, Director - Nuclear Business Support
A. F. Kirby, Ill, External Operations - Nuclear, Delmarva Power & Light Co.
C. Bakken, General Manager - Salem Operations
D. Powell, Director, Licensing, Regulation and Fuels
R. Kankus, Joint Owner Affairs
A. Tapert, Program Administrator
J. J. Keenan, Esquire
j M. J. Wetterhahn. Esquire
J. A. Isabella, Manager, Joint Generation
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Atlantic Electric
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Consumer Advocate, Office of Consumer Advocate
William Conklin, Public Safety Consultant, Lower Alloways Creek Township
Public Service Commission of Maryland
.} State of New Jersey
., j - State of Delaware
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
- Docket No: *50-354
License' Nos: NPF-57
Report No. 50-354/98-03 (OL)
Licensee: Public Service Electric and Gas Company.
'
Facility: Hope Creek Generating Station
Location: P.O. Box 236
Hancocks Bridge, New Jersey' 08038
Examination Period: February 23,1998 - March 4,1998 (onsite)
March 4 - March 12,1998 (inoffice)'
Chief Examiner: D. Florek, Senior Operations Engineer
Examiners: J. Caruso, Operations Engineer
T. Fish, Operations Engineer
Approved by: R. Conte, Chief, Operator Licensing
and Human Performance Branch
Division of Reactor Safety
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EXECUTIVE SUMMARY
Examination Report 50-354/98-03(OL) -
initial exams were administered to six senior teactor operator (SRO) instant applicants and -
five reactor operator (RO) applicants during the period of February 23 - March 2,1998, at
the Hope Creek Generating Station.
OPERATIONS
PSE&G staff submit initially an inadequate examination to administer to applicants for an
operator's license. A good majority of the test items of each portion of the examination
required replacement or significant modifications. Significant interactions between the
NRC and PSE&G and an exam postponement for two weeks were required to develop an
exam that was consistent with the NRC Examince Standards.
Also, there was insufficient controls, criteria, or data recorded in the controlling documents
as evidence that the required control manipulations were significant and were properly
credited. Because of this, not all of the applicants performed five significant control
manipulations which had to be redone. This area is unresolved item pending further
enforcement review by NRC staff with respect to meeting 10 CFR 55.31(a)(5)(50-354/98-
03-01).
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'"' i :. 1
Report Details 05 Operator Training and Qualifications
' 05.1 ~ Operator Initial Exams
a. Scope
The NRC examiners administered initial exams to five RO and six SRO instant
applicants in accordance with NUREG-1021, " Examiner Stendards," Interim -
Revision 8. The exams were prepared by PSE&G staff and were approved by the
NRC. PSE&G staff administered and graded the written exam. The NRC
administered the operating exam.
b. Observations and Findinas
The Hope Creek exam was initially scheduled for the week of February 9,1998, but
due to the inadequate submittal by PSE&G, the exam was delayed and rescheduled
for the week of February 23,- 1998. The PSE&G staff involved with the
development of these exams signed socority agreements to ensure the integrity of
the initial exam process.
The PSE&G staff submitted their proposed sample plan on December 9,1997, .
which was later than requested in the NRC letter dated November 19,1967. The
sample plan was generally acceptable. Because of the reduced time for review, the
NRC Chief examiner made some general comments regarding low power JPMs and
the apparent lack of technical specification assessment on the written exam. The
Chief Examiner also informed PSE&G that because of the reduced time for review..
some comment ~ % result from the review of the initial proposed exams and
these, in the fin. i: . W out to be minor in nature.
The PSE&G proposec 7#Ne ims were submitted for NRC approval on
January 5,1998. The PL.L .nu . smitted exam was not adequate with
respect to discriminating between safe and unsafe license candidates. The exam
required significant modification to meet NRC Examiner Standards.
PSE&G submitted a revised exam over the period January 20-22,1998. A NRC
review of this submittalidentified similar difficulties with the exam, but to a slightly -
lesser degree. Following this submittal, Region I staff discussed in detail each of .
the specific items of the exam at the Hope Creak training center on
January 26-27,1998. The NRC subsequently issued a letter, dated
February 2,1998 officially. delaying the exam and offering PSE&G an additional
opportunity to have the NRC administer the exam if PSE&G could submit a adequate
exam by February 9,1998.
PSE&G submitted their third version of the exam on February 9,1998. .The NRC
concluded that the quality, ;while not at the totally acceptable level, was sufficient
- to proceed with the on-site preparation activities. The PSE&G staff was able to .
revise the exam materials during this NRC on-site preparation visit to a level that
allowed the exam to be administered.
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While the written question topic areas were generally acceptable, the difficu!ty with'
the specific writt.a question generally related to the discrimination validity of the -
question. The following summarizes the problems noted with the PSE&G written
exam submittals (Some examples from the initial submittal are identified):
- Poorly written question distractors which were easily eliminated. (38,43,.
65,67)
- Questions with multiple correct answers. (15,55,76).
- Questions with no correct answer as written. (50,76,104,110)
- Questions that did not correlate with the assigned K/A. (31,98,116)'
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Awkwardly worded questions. (6, 96,102)
- Questions stems that did not solicit the answer in the answer key. (52,59,'
90)
- Questions not appropriate for the license level. (56,58)
The following summarizes the problems noted with the walkthrough portion of the
exam submittals:
- Insufficient JPM coverage against the safety function specification.
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Insufficient JPMs to assess low power conditions.
-- Inadequate standards in the JPMs.
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JPM and administrative questions written as simple memory or direct look up
rather than "open reference" use.
The simulator scenarios were deficient because they lacked sufficient depth to
properly assess applicant performance against the required competencies, as well as
details regarding the actions expected of the applicants. Contributing to this was
insufficient description of the scenario objectives, insufficient description of the
specific malfunction effects, insufficient critical task specification, and improper
completion of the forms in NUREG 1021 to assess the simulator exam.
The NRC ' examiners administered 'the operating exams in the period of
February 23-27,1998. PSE&G administered the written exam on March 2,1998.
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By letter, dated March 6,1998, PSE&G staff Identified answer key comments on
eleven questions. A copy of the PSE&G letter is contained in Attachment 3. The
NRC resolution of the PSE&G comments on the written exam is described in
Attachment 4. PSE&G also graded the wr;tten exam based on answer key revisions
consistent with their comments. The NRC regraded the written exam based on the
NRC resolution of the facility comments.
During the administration of the walkthrough portion of the operating test, several
items were identified that demonstrated a poor quality product in the exam. JPM
initiation cues and JPM questions contained typos in significant data that confused
the applicant and required the examiner to revise on the spot. One JPM and one
admin question had incorrect answers in the answer key. The admin JPMs did not
contain sufficient cues to provide to the applicant and did not contain all the
required attachment material to determine whether the applicant's action was
correct. These rec aired considerable post exam interact ion between the NRC
examiners and the USE&G staff to resolve .
c. Conclusions
PSE&G staff submitted initially an inadequate exam to administer to applicants for
an operator's license. A majority of the test items of each portion of the
examination required replacement or significant modifications. Significant
interactions between the NRC and PSE&G, and an exam postponement for two
weeks, were required to develop an exam that was consistent with the NRC
Examiner Standards.
05.2 Sianificant Control Manipulations
a. Scope
The examiner reviewed in detail the evidence of significant control manipulations
performed by the applicants. These manipulations were required per 10 CFR
55.31(a)(5). Guidance contained in information notice IN 97-67," Failure to Satisfy
Requirements for Significant Manipulations of the Controls for Power Reactor
Operator Licensing" was also used.
b. Observations and Findinas
PSE&G criteria and supporting documentation were not sufficient to assure that
applicants performed five significant control manipulations as required by 10 CFR
55.31(alI5). The criteria of "at least one notch for a minimum of eight rods" did not
assure that a manipulation was significant. This could be a very significant
manipulation with clearly observable power changes or not significant with no
power changes depending on the rods selected and its location and position in the
core, in addition PSE&G did not record supporting data ( initisi power level, time
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start, final power level, time end ) to demonstrate that the actual manipulation in
Mode 1, whether it was by recirculation flow or control rods, was significant and
that multiple credit was not provided for the rnanipulation.
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The PSE&G control for documenting significant control manipulations was the " Reactivity Manipulations Documentation Guide /' dated January 31,1997. The guide documented each manipulation with a signature and date with no additional specific detail provided as to what the applicant specifically performed. All the applicants that took this exam, performed significant control manipulations while the plant was in Mode 1. The PSE&G method and criteria for these manipulations were: - Core Flow in Mode 1 - a change in reactor power, as indicated by the
APRMs, of at least 5%.
- Individual Control Rod Manipulation in Mode 1 - at least one notch for a
minimum of eight rods.
All applicants had at least five significant control manipulations documented. Many of the applicants had several of the significant control manipulations performed on the same day. The data in the summary were not sufficient to determine if an applicant took multiple credit for an extended continuous power change, an issue identified in Information Notice 97-67. PSE&G was requested to provide additional data as to what was the extent of each of the significant control manipulations. The initial PSE&G response provided on February 4,1998 provided some data (control room logs and control rod pull sheets) on some of the manipulations, but the data was not sufficient to determine if all the control manipulations were significant. Additional discussions with the PSE&G staff on February 13,1998 provided no new additional data. As a result, on February 18,1998, the NRC informed PSE&G that many of these manipulations were not acceptable because PSE&G could not provide supporting data an the extent of many of the manipulations and provide information that these manipulations were significant. On February 19,1998 PSE&G staff met in the Regional office and were able to provide data using reactor engineering logs, additional control rod pull sheets, which were not provided in earlier discussions, which allowed many of the significant control manipulations to be accepted. The reactor engineering logs provided data on power history when many of the manipulations were performed. Some of these significant control manipulations performed on the same day were acceptable and some were not. In the final analysis, five applicants from the February 1998 exam did not have the required five significant control manipulations and one applicant had seven
acceptable significant control manipulations, but one of the submitted control
manipulation did not meet PSE&G criteria. The problems with the applications
were:
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No supporting documentation was available to conclude that the
manipulation resulted in an observable affect on power or that the
manipulation was not part or a continuous power change.
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The supporting documentation indicated thc. the manipulation was part of a
continuous power change and multiple credit was taken when only one
manipulation should have been credited.
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PSE&G credited partial withdrawal of control rods following a single rod
scram test. This was not considered significant since this type of
manipulation provided little, if any, integrated response and training value.
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Credit was taken for movement of the same four rods twice when the
PSE&G criteria was to move eight rods.
The following summarizes these applicant's significant control manipulations. The
details are contained as Attachment 5.
Docket No. Credited Acceptable Additional Required
55-62176 7 2 3
55-62178 5 2 3
55-62183 5 2 3
55-62187 5 2 3
55-62175 5 4 1
55-62174 9 7 0 (1 not reviewed)
55-60813 6 4 1
Based on the concerns and findinge of the NRC, the five applicants and the one
operator performed the required additional significant control manipulations on Hope
Creek on February 21,1998 by lowering or raising reactor power by at least 5% by
adjusting recirculation flow.
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c. Conclusion
There was insufficient controls, criteria or data recorded in the controlling document
to assure that the control manipulations were significant and were properly credited.
Because of this, not all of the applicants performed five significant control
manipulations which had to be redone. This area is unresolved item pending further
enforcement review by NRC staff with respect to meeting 10 CFR 55.31(a)(5)(50-
354/98-03-01).
E.8 Review of UFSAR Commitments
A recent discovery of a licensee operating their facility in a manner contrary 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 exam activities discussed in this report, the examiner reviewed portions of the UFSAR that related to a control rod withdrawal accident exam question. The selected exam question reviewed was consistent with the UFSAR.
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V. Maneaement Meetinos
.X1 Exit Meeting Summary
On March 4,1998, the examiners discussed their observations'of the exam process with ~
members of PSE&G management. The examiners noted that no simulator fidelity concerns
had been observed or identified. PSE&G management acknowledged the examiner
observations.
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
NUMBER TYPE DESCRIPTION
50-354/98-03-01 URI Significant control manipulations is unresolved item pending
further enforcement review by NRC staff with respect to
meeting 10 CFR 55.31(a)(5).
PARTIAL LIST OF PERSONS CONTACTED
Licensee
P. Doran, Operations Training
H. Hanson Jr., Operations Superintendent
K. Krueger, Assistant Operations Manager
J. McMahon, Director Trair.ing, OA and EP
M. Swartz, Simulator Supervisor
B. Thomas, Licensing
Attachments:
1. SRO Exam and Answer Key
2. RO Exam and Answer Key
3. PSE&G Comments on the Written Exam
4. NRC Resolution of PSE&G Comments on the Written Exam
5. Significant Control Manipulation Details
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ATTACHMENT 1
. SRO EXAM AND ANSWER KEY
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U.S. Nuclear Regulatory Commission ~
L S.ite-Specific .
Written Examination.
Applicant Information
Name: Region: 1
Date:. Date: 2/23/98
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Facility: Hope Creek -
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License L'evel: SRO Reactor. Type: GE - ,
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Start Time: Finish Tinie:~
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instructions i
Use the answer sheets provided to document your answers. Staple this cover sheet
on top of the answer sheets. The passing grade requires a final grade of at least
80.00 percent. Examination papers will be collected four hours after the examination
starts.
Applicant Certification _
All work done on this examination is my own. I have neither given nor received aid.
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Applicant's Signature
Results
Examination Value Points
Applicant's Score Points
Applicant's Grade Percent
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S:nior Rsector Opsretor An:w:r Shorts
Circle the correct answer if,an answer is changed write it in the blank. .
1. a b c.d 26. a b .c d
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23. a b c d 48. a b c d _
24. a b c d 49. a b c d
25. a b c d 50. a b c d
Page1
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S nior Recctor Op:rctor An:wsr Sheets
Cirde the correct answer. If an answer is changed write it in the blank.
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60. a b 'c d . 85. a 'b c d
61. a .b fc d 86. alb c d
62. a b c d 87. a b c d
63. a b c d 88. a b c d
64 a ti c d '89. a b c d
65. a b c d 90. a b c d
66. a b c d 91. a b c d
67. a b c d 92. a b c d
68. a b c d 93. a b c d $!!
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70. a b c d 95. a b c d
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74. a b c d 99. a b c d
75. a b c d 00. a b c d
Page 2
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Senlar R;cctor Operator Extmination
1. Which of the following evolutions is NOT allowed to be performed by the Reactor Building
Equipment Operator?
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c. Transferring an RPS bus to its alternate power supply with the tcMor at ;;/uwer.
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b. T'est scrammingYcontrol rod froth the indikidust tsst switcheion tis h9draulici ontrbl"
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c. Operating the Standby Liquid Control system in'the Test Tark to Test Tank' mode.
d. n%cing hydraulic control unit nitrogen pressure to the normal band with the associated
. m -101 rod withdrawn. ,
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2. 'Given theiollowing conditions- .
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. Afully qualified Nuclear Control Operator (NCO) with an actis license has justl .
* returned from 10 days vacation -
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.On the first day back on shift, this NCO worked a normal 12 hour shift and then
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accepted and worked 4 heua of overtime ',
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Which of the following is the. maximum number of hours this NCO may work on the second .
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d::y back on shift? (Assume rn i.dditional authorizations have'been made.)
c. 8 hours
b.12 hours
c. 14 hours
d. 16 hours
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3. Which of the following conditions require the Operations Superintendent to perform a formal
turnover when delegating his Control Room Command Authority to another individual?
a. Command Authority is being delegated to the current on-shift Nuclear Control Operator
(RO) and the plant is in Op Con 4. ;
b. Command Authority is being delegated to the current on-shift Control Room Supervisor.
c. Command Authority is being delegated to a current on-shift Nuclear Control Operator
(RO) and the plant is in Op Con 3..
d. Command Authority is being delegated to a Senior Reactor Operator with an active
license who is not a member of the current on-shift crew.
Page 1 of 46
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S:nior Rocctor Operctor Excmination
4. A tagging request with switching order has been received from the System Operator. The
- Sw;tching Order has been confirmed and the tags prepared. The System Operator has
contacted Hope Creek and directed the performa.nce of the tagging request and switching.
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Which of the following personnel are required to be present in the'500KV swit& yard '
blockhouse for' completion of the tagging request and switching order? -
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a. A Nuclear Equipment Operator and a Nuclear Control Operator.
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b.' Two Nuclear. Equipment Operators. .
c. A Nuclear Equipment Operator and a Control Room Supervisor.
. d. A Nuclear Equipment Operator and a member of the Sy.ste.ms Operatipn D.epartment.. .
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5. Following shift turnover the Nuclear Control Operator (RO) notes that data enter the
narrative log by the previous shift.is incorrect. - . .
The RO draws a Hogle lirie through the incorrect entry,'makes the ect entry and initials
and dates the change. Which of the.following describes how
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RO should highlight and
explain the change? ..
a. The correct entry shoulr' be circled in red wi n explanation placed in the comments
section.
b. The correct entry should be cire in red with an explanation made next to the corrected
entry,
c. The incorrect entry uld be circled in red with an explanation placed in the comments
section.
d. The in ect entry should be circled in red with an explanation made next to the
cted entry.
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Page 2 of 46
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Senior R=ctor Operator Excmination ]
6. During a valid high reactor pressure condition, the Rscirculation Pumps did NOT .
automatically trip as designed. -
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Which of the following actions must be taken by the Control Room to open the Recirculation " '
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* Pump Trid"(RPT) Brsak' e rs.' ' -
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- a. Man 0 ally initiate both channels of the Redundant Reactivity' Control System (RR.CS). .,'
b. Verify the RPT Breakers trip when the Recirculation Pump MG Set Drive Motor Breakers.
are. opened.
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' c. Direct-the local tripping of the RPT Breakers. < .* -
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.d. Depress the RPT Breaker '. Trip" pushbuttons.
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7. ~Which of the following are the MINIMUM guidelines for Operations Superintendent (OS)
*
- review of critical plant parameters (reactor power, level, ' pressure and turbine load) during . . normal . steady . state plant operations? -
. .
,
..
'The OS stiould: .
.
'
.a. receive.a verbal report from the. Control Room Supervisor (CRS) every hour. .
b. review the current operating bgs, review CRIDS, or parforni a panel walkdown at least
twice during the 12-hour shift.
- c. view current plant conditions on the Control noom information Display System (CRIDS)
every hour,
d. walk-down the control room panels at least four times during the 12-hour shift.
.
f .
. . Page 3 of 46
. _ . . . _ . . .
Senior R: actor Operator Excmination
8i Given the following conditions:
.. A plant shutdown with control rod insertions occurring is in progress . ,
* Reactor power is 22% with generator output at 2.42 MWe " "' ~ ' '
" * * '
~-
J'.' The 5ecorid NCO (PO) begins deinerting the'drywell' *** -
- 1. The CRS is reviewing procedures at the CRS desk
.
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* - ' * -
:
. No other personnel are in the Control Room
Which of the following additipnal requirements, if met, would allow a License Class instant
'.
~
. SRO trainee under direction of.the Nuclear Control Operator (RO) to continue rod. motion for. '
the given conditions? - . .
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.e. Operations Manager written permission.to allow a. License Class trainee to insert control --
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ragg, . .
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. . T 'b. Anothes technically qualified member of the unit technical"sta(to observe rod movement. ~
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c. Verification that the Rod Worth Minimizer is operating prope..ly before reducing power -
- -
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. below 20%.
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d' A Reactor Engineefs presence to satisfy Technical Specifm' ation requirements..
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9. Given the following conditions:
The plant is shutdown for a maintenance outage
. A Red Blocking Tag (RST) is hung on 4160 VAC breaker
. The breaker is tagged in the " Test Disconnect" position >
Later in the outage, the breaker is being removed from its cubicle for maintenance
Whkh of the following describes the required tagging actions for the given conditions?
a. The RBT shall remain on the breaker, the breaker removed from the cubicle and an -
additional RBT installed on the rope! tape placed across the opening.
b. The RBT shall be removed from the breaker but kept active and maintained in the
physical possession of Operations while the breaker is out of the cubicle.
c. The RBT shall be removed from the breaker, the breaker removed from the cubicle and
the same RBT installed on the safety rope / tape placed across the cubicle opening.
d. The RBT shall remain on the breaker, the breaker removed from the cubicle and a White
Caution Tag installed on the safety rope / tape placed across the cubicle opening.
' - Page'4 of 46.'
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S:ninr Rrctor Operat r Excminrtion
10. Which of the following describes how the Operations and Chemistry Departments coordinate
installing Red Blocking Tag.s on the Hydrogen injection System?.
a. - Operations positions all system components
. Chemistry. monitors the system component positioning . .
- Ope, rations installs the tags
- Chemistry performs the independent verification
l
b. - Chemistry positions all system components
- Operations monitors the system component positioning
- Chemistry installs the tags . .
.
- Operations performs the independent verification
c. - Operations positions all system components
'
- Chemistry' monitors the system component positidning . .
-- Chemistry installs the tags- .
-- Chemistry performs the independent verification ' - -
d. - Chemistry positions all system components
'
- Operations monitors the system component positio'ning ~ ~
- Operations installs the tags
- Operations performs the independent verification
,
11. Given the following ec;iditions:
Power is 89%
~
At 1200 on 2/16/98 is discovered that, due to a recent procedure change, part of a TS
required surveillance was not performed.
The last complete satisfactory surveillance was completed at 1200 on 1/15/98
The incomplete surveillance was performed on 2/13/98
The surveillance is required to be performed at least once per 31 days
The action statement requires that inoperable equipment must be restored within 72 hrs,
or be in Hot Shutdown within 12 hrs and in Cold Shutdown within next 24 hours.
If the surveillance is not satisfactorily performed, which of the following identifies the date
when the unit must be in Hot Shutdown?
c. 2/18/98
b. 2/19/98
c. 2/23/98
d. 2/26/98
Page 5 of 46
Senior R:cctor Operator Excmination
12. Given the following ccnditions:
A General Emergency has been declared .
. All Emergency Response Orgarpization facilities ha've been' activated
. ' ~
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Planned emergency axposures are'necessary to evccuate injured plant personnel
= The Radiation Protection Supervisor - Exposure Contrors ALARA Analysis shows
' '
expected rescue team individual exposures of 6500 mrem-
The Operations Support Center Coordinator, Operations Superintendent and
Radiological Assessment Coordinator have determined that emergency exposure
'
.mu'st be'receiv'ed-
.
Which of the following individuals must authorize the emergency exposure for t;ie given
conditions? v ,
c. Ernergency Duty Office'r , ,
b. . Emergency Coordinator
c ' Radiological Assessment Coordinator .
d. Operations Support Center Coordinator , .
.
.
13. The estimated time to independently verify a valve position is 15 minutes.
Of the listed dose rates, which is the MINIMUM dose rate that would allow waiving the " Hands
On" independent verification requirement for the corditions given?
c.10 mrem /hr ]'
b. 30 mrem /hr
c. 45 mrem /hr
d. 60 mrem /hr
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a
c
Page 6 of 46
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S:ni r Rrctor Operater Excmin::. tion i
i
14. An emergency has occurred immediately requiring reasonable actions to be taken that depart
from Technical Specifications. No actions consistent with Technical Specifications that can
provide adequfe equivalent protection are immediately apparent.- i
'
Which'of the fcilowing' identifies who is required to approve the action and under what' *
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'
*
conditions the action can be performed?
'
'
O. The Control Room Supervisor approves actions to be taken to protect the health and
safety of facility personnel.
b.' The Control Room Supervisor approves actions to be.taken to protect the health and
safety of the public. .
c. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to be
'
taken to protect the health and safety of facility personnel.
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.
di The Emergency, Coordinator, in the Emergency Ope, rating Facility, approves actions to be ',
taken to protect the health and safety of the public.
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' 15. ' Niich of ths'fdllowing is the first no'ification
t requirement and when must that notificati~or: be j
made when a plant. event requires declaration of.an Alert?
c. To the N'RC - within 15' min' utes of the event occurring.
b. To the State and Local agencies - within 15 minutes of the event occurring.
c. To the NRC - within 15 minutes of the Alert declaration.
d. To the State end Local agencies - within 15 minutes of the Alert declaration.
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Page 7 of 46 .
Seni: r Rrctor. Operator Examination -
16. Given the following conditions:
. . . .
. .' .
A major plant transient has occurred
*
The plant is now in a stable condition .
'
<- - '--
' ' ' ' . Posf transient reviewindicates an Alert should habe beeh' declared 30'~minOtes*
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ango but the. conditions do not currently exist ,
. .
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Which of the following describes the requirements for event declaration and notification by the . -)
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Operations. Supervisor (OS)? .
'
'c. Ihe OS should declare the Alert, make the appropriate State, Local and NRC J *'
n6tifications and immediately downgrade or terminate the ' classification as appropriate for -
,
current plant concLtions. .
.
s. ... .. .. . . .. .. .
. .
' b. The OS need not. declare the Alert but should make a'non-emergency one hour report to "
,' the NRC Operations Centet. ,
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!- c. The OS should declare the Alert, make the State, Local and NRC notifications and hold
- at this classification until the Emergency Duty Officer (EDO) terminates the. event. . . ' . l
d.' The OS need not declare the Alert but should make a non-emergency four. hour report to
'
the NRC Operations Center.
'
.
,
17. Given the following conditions:
The plant is performing a shutdown in accordance with 10 0004, " Shutdown
From Rated Power To Cold Shutdown"
At 20% power the shutdown is completed by placing the Reactor Mode Switch
to " Shutdown"
All plant systems responded as designed during the scram
Post Scram Actuation Review is in progress in accordance with HC.OP-AP.ZZ-0101,
Post Reactor Ocram/ECCS Actuation Review and Approval Requirements
Which of the following should be the FIRST reactor scram signal identified when reviewing
the Sequence Of Events printout?
a. Reactor Mode Switch in " Shutdown"
b. IRM Neutron Flux - High
c. Scram Discharge Volume Water Level - High
d. APRM Neutron Flux - Upscale, Setdown
Page 8 of 46
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Senior RIctor Operator Excmin: tion
18. Given the following conditions:
* The plant is at normal operating pressure and temperatures .,
b Allplant syste'ms are operating as designed
*'
*
' i The "A" add "B" scrarh toggle' switches at'the hydraulic co~ntrol unit for * -
-
control rod 42-03 have been placed in ' Test" .-
:
Which of the following is the expected response of the Scram Pilot Valves for Control Rod 42-
03 and the Scram Dump Valves for the given conditions?
a. - The Scram Pilot Valves reposition to vent the Scram Inlet and Outlet Valves- '
- The Scram Dump Valves remain in their initial positions ,
.b. - The Scram Pilot Valves remain in their initial po~sitions. -
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'
- The Scram Dump Valves remain in their initial positions ,
- . . ,, .. . .
. . .
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c . '- Th' Scram
e Pilot Valves reposition to vent the Scram inlet and Outlet. Valves
- The Scram Dump Valves reposition to vent the Scram Discharge Vent and Drain
'
Valves . . . .
d. - The Scram Pilot Valvas remain in their initial positions. , , .
- The. Scram Dump Valves reposition to vent the Scram Diccharga Vent and Drain
'
Valves ,
, ._
19. Given the following conditions:
The plant is performing the control rod exercise surveillance
The Nuclear Control Operator (RO) selects control rod 34-19 on the rod select module
Only one half of the selected rod pushbutton illuminates
Which of the fcl lowing describes what has failed and how that affeds the ability to move
control rods?
a. The selected control rod activity control card is in the scan mode and rod motion is
allowed.
b. The selected control rod activity control card is in the scan mode and rod motion is not
allowed.
c. Only one of the two RMCS transmitter cards has successfully selected the control rod
and rod motion is not allowed.
d. Only one of the two RMCS transmitter cards has successfully selected the control rod ,
I
and rod motion is allowed.
Page 9 of 46
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S::nior R cctor Operator Exfmination -
20. G ven the following conditions:
The plant is operating at 25% power peiforming a startup ,
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"
. Control rod 10-23 has been determined to be stuck
" ' -
While attempting 'to withdraw the control Nxi, indicated ~ drive water flow is~readire ' 'c' -
"0" gpm ,
. -
,
. . . . .
Which of the following is the cause of this indication?
c. Hydraulic Control Unit Directional Control Valve (122) has failed to reposition. .
b. The 2.gpm Stabilizing Valve has failed to reposition. , ,
. ,
, c. Both Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed
'
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** *- ' " '
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open. . .
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.d. The Drive Water Header Pressure ,Cbntrol Valve has failed closed.
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21. Given the following co'nditions: ,
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1
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Control rod insertions are in progress for a plant shutdown .
* The last control rod in Group 35 was inserted to Notch "02" y .
-
* The fitst three control rods in Group 34 were then fully inserted
e insart and withdraw limits for these two Groups are Notch ~"00" and Notch "12"
respectively
Which of the following describes what the Rod Worth Minimizer (RWM) will be displaying for
the given conditions?
a. The RWM will be displaying normal operation parameters with no alarms or errcrs in
effect.
b. The RWM will be displaying a select error with no other alarms or errors in effect.
c. The RWM will be displaying a select error with the Group 35 control rod at Notch 92" in {
the withdraw error box. A rod withdrawal block is in effect. ]
d. The RWM will be displaying a select error and three insert errors. A rod insert block is in
effect. l
, Page 10 o' f 46
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Senior Reactor Operator Excmination -
a
22. Given the following conditions:
.~ A reactor startup is in progress .
.
* Reactor power is,30%
'
,
. Thd total steam flow signal' output from thG Feedw6tl'svol Contt6l SpsteM faits tothe ' ~
'
equivalent of 18% power. ,
,
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Which of the following describes how the Rod Worth Minimizer will enforce control rod
movement for the given condi.tions? ;
. ,.
'
The Rod. Worth Minimizer wi!' allow continued control rod niovement but only in single :
'
c ~ ' ' '
notchincrements. .
'
. b. .The Rod Worth Minimizer will allow all normal control rod motion until actual reactor
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,
power is I,ess than the Low Power Setpoint; , ,
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c. The Rod Worth Minimizer will immediately prevent all controt rod insertions nnd
'
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withdrawals. - . .
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Id. The Rod Worth IUlinimiser~will brevent cohtrol rod'withdrtwals if any control rM is-
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withdrawn past its withdraw limit. .
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23. Given the following conditions:
The plant is operatirig at 75% power
* Confirmed seal failures have occurred on the "B" Recirculation Pump
The pump has just been tripped
Which of the following describes the preferred order for isolation of the "B" Recirculation
Pump and the reason for that order?
a. Close the Suction Valve', isolate seal purge and close the Discharge valve - This order
ensures further damage is not done to the seil package from overpressure.
b. Close the Discharge Valve, isolate seal purge and close the Suction Valve - This order
ensures the Discharge Valve is stroked against a minimal differential pressure.
c. Close the Suction Valve, isolate seal purge and close the discharge valve - This order '
.
ensures the Suction Valve is stroked against a minimal d"'erential pressure.
d. Close the Discharge Valve, isolate seal purge and close the Suction Valve -This order
ensures further damage is not done to the seal package from overpressure.
4
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Page 11' of 46'
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Senior Rozctor Operator Excmination
22. Given the following conditions:
.
: A reactor startup is in progress ,
.
' '
Reactor poweris 30% . ~
Th4 t'otal steam flow signal output from the Feedw' a terl'6 vel Contr'ol Systern fails tothe *~
- '
-
e4Jivalsnt of 16% power. <. .
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,-
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. . ,
.
Which of the following describes how the Rod Worth Minimizar will enforce' control rod .
. movement for the given conditions? .
- - -
. .
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.. .
* ' a.- The' Rod Worth Minimizer will allow continued control rod' movement but only in single
1,0tch increments.
'
. b. .The Rod Worth. Minimizer will allow all normal control rod motion until actual reactor '
l.
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power is less than the Low Power Setpoint , , ,
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c. The Rod Worth Minimizer will irhmediately prevent all control rod insertions and * - 1
-
withdrawals. - .
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' ' '
Yd.' The Rod Worth slinimiser'will brevent coFitrol ro'd withdrawals if any contx1 rod is-
withdrawn past its withdraw limit. .
9
. . . s . .
.,
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23. Given the following conditions: i
The plant is operating at 75% power
Confirmed seal failures have occurred on the "B" Recirculation Pump .
The pump has just been tripped
Which of the following describes the preferred order for isolation of the "B" Recirculation
Pump and the reason for that order?
a. Close the Suction Valve', isolate seal purge and close the Discharge valve -This order
ensures further damage is not done to the seal package from overpressure,
b. Close the Discharge Valve, isolate seal purge and close the Suction Valve - This order
ensures the Discharge Valve is stroked agaiast a minimal differential pressure.-
c. Close the Suction Valve, isolate seal purge and close the discharge valve - This order
ensures the Suction Valve is stroked against a minimal differential pressure.
d. Close the Discharge Valve,-isolate seal purge and close the Suction Valve - This order
ensures further damage is not done to the seal package from overpressure.
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: Page 11 of 46 '
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Senior R=ctor Operat:r Ex: vdnation
24. Given the following conditions:
Prepatations are complete to trait the "A" Recirculation Pump '
. The P. ump Discharge Valve (F031 A)is closed ' ..
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Which of the following describes how the "A" Recirculation Pump trip on tho disdarge valve -
-
-
cloere is bypassed to allow the pump to.be started? .
;
c. This trip is bypassed until the pump start sequence is complete within prescribed time
~
. limits. - - .
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. .
,
b. This trip is bypassed untii th's discharge valve'has reached' the 10d% 'open' position.
c. This trip is bypassed until the pump has been running for 9 seconds. .
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*
d.' Tiils trip is bypassed until'the dischafge. valve jog.(open) dircuit has timed out. '
.
- - -
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'* - - . . . - .
.
. .
*
25. Given the following conditions: ,
.
. . . . .
The plant is operating at 60% power with the "B" Recircuistion' Pump scoop tube locked
'
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The operator is preparing to reset the scoop tube
-
Speed demand on.the "B". Recirculation Pump is slightly 1.ES'S then indicated speod ,
Which of the'following actions is the operator directed to perform if pump speed begins to >
,
'
slowly rise at approximately 2% per minute after the Scoop Tube Trip / Reset pushbutton is
prerad? ,
a. Immediately press the "A" Recirc MG Set Drive Motor Breaker Trip pushbutton.
b. Attempt to control speed with the increase / Decrease arrows on the Pump Speed Control
Station for the "B" Recirc pump,
c. Press the Scoop Tube Trip pushbutton for the "B" Recirc pump.
d. Press the Recirc Pump Trip (RPT) Breaker Trip pushbutton for the "B" Recirc pump. .
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LPage 12 of 46 -
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- S nior R::ctor Operttor Examinr.ti:n
26. Given the following conditions:
. The plant is operating at 75% power , , .. ,
. Valve, stroke time testing is in progress on the "A" RHR PumpTorus Suction
, , - ' -
- - ' *
Valve I,F004A) ' l . ' ~ 'l - -
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* The valve is currently closed T -
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. All other RHR systemT.sTipon de are in their normal standby lineup -
-
.
_ A' steam break causes drywell pressure to reach 2.0 psig.
Which of the followird describes.the' response'of the F004A vafve and the"A" R'HR. pump?
' d.' The F004A valve automatically' opens and the "A" RHR Pump automatically starts after
F004A is fully open. , ,,
I
' b. 'The F004A valve must be manually opened and the "A" RHR Pump sutonsaticall'y starts. '
.. afte'r F004Ais;. fully,6 pen.
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' c. The F004A valve automatically. opens but the"A" RHR Pump must be started. by the,
' '
-
operator afterF004Ais fully open. '
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,
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d.. The F004A valve must be manually opened and the "A".RHR Pump manuaily started ,
after F004A is fully open. *
,
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27. Given the following conditions:
The plantis operating at 90% power .
The Reactor Water Cleanup (RWCU) Inboard Isolation Valve (F001) has just
stroked closed
No othei RWCU valve repositioned
RWCU responded as designed
Which of the following initiated the RWCU isolation?
a. RWCU system differential flow is excessive.
b. The RWCU Filter /Demineralizer inlet temperatures are excessive.
c. The "A" Reactor Protection System MG set tripped.
d. The "A" and "D" NSSSS Manual isolation pushbuttons have been armed and depressed
simultaneously.
Page 13 of 46
&
S;nior R:cctor Operctar Excmination
28. Which of the following describes the reason for having the capability to bypass the Residual
Heat Removal (RHR) Pump suction path interlocks?
c. Allows operation'of the RHR Pumps for shutdown cooling from the Remote Shutdown
o
.s Panel. .
- -
: - . .
b. Allows the "C" and "D " RHR Pumps to be utilized for an alternate path of suppressicn
pool heat removal. .
c. Allows operation of the RHR Pumps for cooling the containment Hydrogen Recombiners
post-LOCA.
d. Allows the "C" and "D" RHR Pumps to be utilized for an attemate path' of core decay heat
removal.
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29. The plant is 'in Mode 4 with Shutdown Cooling in service on 'the "A" Residual' Heat Removal"
(RHR) loop.with the "A" RHR Pump running.
.
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Which of the following describes how a loss of the "B" Reactor Protection System (RPS) bus
will affect the inboard and Outboard Sh'utdown Cooling isolation Valves (F008 & F009)?
. c. The.F008' arid F009 valves' b'oth.close.
b. The F008 valve closes and the F009 valve remains open. ~
c. The F008 and F009 valves both remain opeti. I
d. The F008 valve remains open and the F009 valve closes. ,
30. Given the following conditions:
)
The plant is shutdown
The reactor head is removed but no fuel has been removed from the vessel
Shutdown Cooling is in service on the "B" Residual Heat Removal loop
Reactor coolant temperature decreases to 65 'F
Which of the following would be the expected result of the low reactor coolant temperature?
a. The reactor vessel flange thermal stress limits will be exceeded.
b. The Technical Specification reactor coolant chemistry conductivity limit will be exceeded.
c. The reactor teroperatures can no longer be monitored.
d ;he calculated shutdown margin wpuid be invalid.
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Page 14 of 46
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S:ntur R:: ctor Oper tor Extminition
31. During a loss of feedwater, a manual start of the High Pressure Coolant injection (HPCI)
system was done at a water level of -20 inches by operator manipulation of the system
. components.
~
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Mich of the fdlloWing describes'thiiHPCI system response as reactor water level' con'tinues
to change? .
,
.
c. It will automatically trip at +54 inches and will automatically restart at -38 inches.
b. It requires operator action to secure injection when level is greater than +54 inches and
automatically restarts at -38 inches.
c. It requires operator actions to secure injection when level is greater than +54 inches and
to restart when level is less than -38 inches.
d. It wili automa'tically trip at +54 in'ches aiid.Will require operator' action to restaitOhen ievel
is less than '-38 inches. ,
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32. Given the following coriditions:
.
. Aloss of coo! ant'accider.! has occurred ,
Reactor water level reached -140 inches and is currently -50 inches and rising
Drywell pressure is 6 psig
. All plant systems responded as designed
For the given conditions, which of the following describes the system isolation capabilities for
the Core Spray System (CSS) Downstream Loop Injection Valve (F0058) and the CSS
l
Upstream Loop Injection Valve (F0048), should Core Spray Loop "B" isolation be required?
l
a. Only r005B valve may be closed. i
b. Neither the F0048 or F0058 valves may be closed.
c. Only the F004B valve may be closed.
4
d. Both the F004B and F0058 valves may be closed. ;
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Page 15 of 46
S;nior Rrctor Operatcr Extmin; tion
33. Given the following conditions: ;
!
* A failure-to-scram with Main Steam isolation Vaive (MSIV) closure has occurred -
The pressure spike.on the MSIV closure wa.s 1120 p,sig '
>
'. Rdac' tor poweris 16% and Water Isvel is -25 inches as the 3.9 minute' timer times out
Only Division 11 of the Redundant Reactivity, Control System automatica.lly initiates
. No operator actior's are' taken J
Which of the following it the expected plant response for the given conditions.
a. Both SLC Pumps start, both' Squib Valves fire and the RWCU isolation Valv'es (Inboard -
^
F00'1 & Outboard - F004) close.
>
b.: The "B" SLC Pump . starts,.the "B" Squib. Valve, fires and orily the RWCU inboard Isolation
Valve (F001), closes.
,
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c. Both SLC Pdmps start,' both Squib Valves fire and only the RWCU ' inboard Isolation
Valve (F001) closes... .-
- : >
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d. The "B" SLC Pump starts, the "B" Sq0ib Valve fires and only the'RWCU Outboard :
isolation Valve (F004) closes.
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.
. .
34. Given the following conditions:
The plant i.s in a failure-to-scram condition
. Standby Liquid Control (SLC) has been initiated by the operator
Approximately 13 minutes later the operator noted SLC Storage Tank level analog
indication on Panel 10C651 is "0" gallons
No additional SLC system abnormalities were noted
Which of the following describes how boron injection would be continued for the given
conditions?
a. Boron injection would continue with two SLC Pumps running.
b. Boron injection would continue with the "A" SLC Pump running.
c. Boronin; . tion would continue with the "B" SLC Pump running.
d. Boron injection would have to be transferred to RWCU as directed by EOP-0304.
,
Page 16 of 46
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" Senior Rrctor Operat:r Examinatirn i
I
35. Given the following conditions:
!
* The reactor scrammad and HPCI and RCIC initiated on low reactor wat'er level l
following a loss of feedwater * *
.
*
* Water'le' vel has bee'n restored to'the' normal b' and ' ~
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_
+ All required operator actions were taken on the scram , .
,
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,All Scram Roset switches have been placed in RESET and released
e - Which of the following would prevent the scram air header from repressurizing for the
!
conditions given? , ,,
c. The Scram Discharge Volume High Level Scram Bypass Switch was not retumed to
i . NORMAL. ~
" ~ '
b. The RPS trip Iogic channels B1 and 82 fail to reenergize when RPS 'is reset.
.
. ;
. .
-
c. 125 VDC power is lost to one Backup Scram valve. .
d. The Redundant Reactivi,ty Control System Alternate Rod Insertion logic is not reset.
.
36. Given the following conditions: ,
. The plant was performing a sikrtup following a refueling outage when a reactor
scram occurred (all rods inserted)
The sequence of events printout shows that just prior to the scram, Average
Power Range Monitoring (APRM) chaanels "B" and "D"were upscalo Hl-HI
Which of the following additional conditions, by i self, could have caused the full reactor
scram signal?
c. Rod Block Monitor Channel "A" has failed.
b. RPS Bus "B" has deenergized.
c. SRM Channels "A" and "C" are reading 1.5 E5 counts per secor.d.
d. The Reactor Protection System shorting links are removed.
4
Page 17 of 46
s-
S:ninr R cctor Operator Excmin ti:n 1
37. Given the following conditions:
1
* The plant is operating at 100% power
' '
* APRM Channel,"D"is bypassed with the joystkk '
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. . Control ^ rod 30-31 is selected' - -
e' All other plant. systems are operating as designed - .
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, ,
Which of the following occurs if APRM Channel "F" fails full "downscale" for the given
conditions?
a. Rod Block Monitor dharinel "B" automatically shifts'to the "B" APRM as'its reference.
. b. Rod Block Monitor Channel "B" generates a rod withdrawal block on a failure to null.
c. ~ Rod Block Monitor Channel"B"is. indicating 0%. -
.
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d./ Rod Block Monitor Channel"B"is bypassed on the reference.APRM dosvnscale. ,
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. . :.
38. Given the tollowing conditions:
The plant is . performing control rod withdrawals for a reactor startup
The reactor is si Witical- ~
R'eactor power is 75 counts per second (CPS) in the source range
The Nuclear Control Operator (RO) selects the '"B" Source Range Monitoring (SRM)
detector then holds its " Drive Out" pushbutton in the depressed position
:
Which of the following describes the plant response?
a. The "B" SRM detector will not withdraw due to the current power level,
b. The Retract Permit light will extinguish and the SRM Det Removal Not Permitted alarm
will be received.
c. The "B" SRM detector will retract untii source range indicates less than 3 cps.
d. A Control Rod Withdrswal Block will be generated.
i
Page l18 of 46
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Senirr Rrcctor Operc. tar Ex minrtisn
39. Given the following conditions:
.
The plant is operating at 55% power
Average Power Range Monitoring
(APRM). Channel"C" currently has 14 " good"
'
' ' '
i' '
- ~ ' * - ~
LPRM input signals ,
Which of the following will result in receipt of the APRM Sys A Upscale' Trip /inop alarm (C4 on
Section C3)?
c. APRM "C" meter function switch is placed in !' Flow".
b. .One of the " good" LPRMs mode switch is placed in "C"(Calibrate).
c. APRM "C" meter function switch is placed in " Average".
~ '
~ ^
. d.' One of the " goo.d".LPRMs fails "down$cale". { l .
40. ;Which of the following describes the. difference in actuaireactor water level versus indicated
wide range reactor water Ievel and the expected change in that difference during a power
- '
reduction from 100% to 65%?
O. Actual water level is. lower than indicated level and the difference will get larg'er du' ring
' '
the power reduction.
b. Actual water level is higher than indicated level and the difference will get larger during
the power reduction.
c. Actual water level is lower than indicated level and the difference will get smaller:during
the power reduction.
d. Actual water level is higher than indicated level and the difference will get smaller during
the power reduction.
41. The Reactor Core Isolation Cooing (RCIC) system flow controller has failed full downscale
dernanding a "0" gpm flowrate. The controller is in " Automatic".
Which of the following is the expected RCIC turbine response upon receipt of a valid initiation
signal for the given conditions?
a. RCIC will start, accelerate and trip on mechanical overspeed,
b. RCIC will start, accelerate then slow to a stop.
c. RCIC will start, accelerate then will slow to and run at a low speed.
d. RCIC will stan, accelerate to and run continuously at approximately 4000 rpm.
.
. Page 19 of 46 -
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Senlar R::ctor Oper; tor Excminztirn i
~ 42. Given the following conditions:
* A loss of all AC power has occurred '
f
a No Diesel Generators are running '
"
*
* The Reactor Core Isolatio' nC6oling (RCIC) systeih has initiate ~d arid is injecting
~
~.
* A valid ,RCIC steam line high flow signal,is recalved .
Which of the following describes the RCIC inboard and Outboard Steam Supply isolation
Valves (F007 & F008) isolation capabilities and the response of the RCIC turbine for the
' ~
given conditions?. * . .
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c. The F007 and F008 valves remain open but can be closed from the Control Room.
,
b. .The F007.and F008 valves remain open and. cannot be closed.from.the bon. trol Room.
c. Only the.F007. valve. closes. .. ..
' '
~d. Only the F008 volve closes.
. -
.. .. . . . .
43'.' Gihen the following conditions:
* The Automatic Depressurization Systern (ADS) Manual Initiatiori Charinel"B"-
'
and "F" pushbuttons (S6B and S6F) have been armed and depressed
* There is no Safety Relief Valve response
Which of the following "B" Division electrical bus failures caused this system response?
c. A loss of 120 VAC Bus 1BJ481
b. A loss of 250 VDC Bus 10D261
c. A loss of 125 VDC Bus 1BD417
d. A loss of 480 VAC Bus 108420
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Page 20 of 46 {
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S:nier R :ctor Operator Excminttion
44. Which of the following is the MINIMUM number of Safety Relief Valves (SRV) that niust Se
opened during an iimorgency Depressurization and the reason for that minimum number?
c. 4 SRVs provide the minimum depressurization rate required to ensure the low pressure
ECCS systems iriect.soon enough to rpinimize the amount of time water level is below
the top of active fuel. ,
ti. S SRVs provide the minimum depressurization rate required to ensure the low pressure
ECCS systems irject soon enough to minimize the amount of time water level is below
the top of active fuel.
c. 4 SRVs provide the minimum steam' flow through the core required to assure adequate
core cooling.
.
,
d. 5 SRVs p.rovide the minimum steam flow through the . core required to as.sure adequate
' ~
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core coolingc
.
45. Given the following conditions: , .
.
The plant has been operating'at 100% power for several weeks
All systems are operating as designed
Which of the following is the reason why periodic nitrogen makeup to the drywell is required
for the given conditions?
a. Due to leaks from drywell air operated equipment.
b. Due to PClG normal system leakage.
c. Due to normal periodic cycling of the Turus - Drywell Vacuum Breakers.
d. Due to normal drywell air inteakage.
> .
Page 21 of 46
..
Senior R: ctor Operctor Ex min tion
46. Given the following conditions:
.
The plant had been operating at 75%. power .
. .
A loss.of main condenser vacuum caused a complete Main Steam isolation '
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'
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'" -
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'Velve'(MSIV)' closure '
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.
. . The Main Condenser Vacuum Breakers h?ve been opened
-
The main turbino did NOT trip and was NOT manually tripped on'the scram -
<
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The MSlV switches have been placed in "Close"
. Which of the following conditions are required to allow resetting the'NSSSS MSIV isolation
'
logic for the given conditions? . .
c. The Mai.n Con. denser Low Vacuum Bypass switches must be in " Bypass" and the Turbine
.
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*
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Control Valves must be' closed. -
'
' b. The Reactor Mode Switch must be out of "Run" and the Turbine Control Valves must be
'
closed.
c. The Main Condenser Low Vacuum Bypass switches rr ust be in " Bypass".and.the Turbine; s
Stop Valves.must be closed to less than 90% open. ,' ~
d. The Reactor. Mode Switch must be out. of "Run" and the Turbine Stop Valves must be
closed to less than 90% open. -
.
47. Which of the following conditions would prevent opening the RHR "B" Loop inboard and
Outboard Drywell Spray Valves (F021B and F0168) following a LOCA?
a. The LPCI Injection Valve (F0178) is not fully closed.
b. Less than 5 minutes have elapsed since the "B" RHR initiation occurred. .
c. The RHR Full Flow Test Valve (F024B) is not fully closed.
d. Reactor water level is above -129 iriches.
Page 22 of 46
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-S:nier R0cctor Operater Exrmin tion
48. Given the following conditions:
.
* The Fuel Pool Cooling system is operating with one pump and heat exchanger
in service .
' '
' - - - - !
The Fuel Pool Gates areInstalled- - -
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.. No makeup water sources are available , ; .
,
.
.
Which of the following is the expected effect on Spent Fuel Pool water level and cooling
capability if a leak develops on the common FPCC Pump Suction?
~
a. Cooling capability'and sater level will be unchanged.-
b. Cooling capability will be lost and wamr level will lower slightly and stabilize.
c. " Cooling capability will be' unchanged and water level will lower slightly'and stabilize. .
, d. Cooling capability will .be lost and water level will continuously lower.
. ,
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. .. . . ..
49. Which of the following describes how the main , steam line flow restrictors assist in maintaining -
' cdequat' ecore cooling'for steam line break between the flow restrictors and the Main Steam j
isolation Valves? :
~
a. They ensure the to'tal inventory loss from the react'or vessel maintains level above the top
'
of active fuel until one division of low pressure ECCS is injecting.
b. They limit the total inventory loss from the reactor vessel to maintain water level above
the top of active fuel for a minimum of 5 seconds. ,
c. They ensure the total energy release rate to the Primary Containrpent does not result in
_
exceeding suppression chamber design pressure.
d. They limit the total inventory loss from '.ne reactor vessel to maintain level above the top
of active fuel until HPCI is at rated flow.
50. Which of tt's following describes the expected indicated steam flow response with an open
Safety Relief Valve (SRV) and the reason for that response? '
o. Indicated steam flow goes up, because SRV steam flow is seen as additional steam flow
over what is going to the main turbine.
b. Indicated steam flow goes down, because the SRV steam flow is not monitored by the
main steam system flow detectors.
c. Indicated steam flow remains constant, because the Turbine Control Valves and Intercept
Valves throttle open to meintain a steady MWe output.
d. Indicated steam flow remains constant, because the Terbine Control Valves throttle
closed to maintain constant reactor pressure.
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Page 23 of 46
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f g> n ,
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4 150mm *
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k.
._-
Senire' Rrctor Operator Examination
51. Given the fo: lowing conditions:
* The plant is operating at 70% power
* The "B" EHC Pressure Regulator is tagged .out of service '
'
' a' Unknown to the operator, the "A" EHC Pr' essure Reg'ulator output signal is
failed "as is"
Which of the following would be the expected response of the Turbine Control Valves and
Turbine Bypass Valves as the Nuclear Control Operator (RO) begins to raise power using
recirculation flow for the given conditions? (Eigure. attached)
a. - The Turbine Control Valves will close
.
- The Ttsbine Bypass Valves will open *
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. .. '
. b. - The. Turbine Control Valves will close .
'
, Tho Turbine Bypass Valves will not rhove .
c.. - The Turbine Control Valves will not move
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.The Turbine Bypass.valys will not' move ,
-
d. - The Turbine Control Valves will not move.-
, - The Turbine Bypass Valves wiil open
/
52. Given the following conditions:
A loss of off-site power (LOP) has occurred from 75% power j
'
Within 10 seconds a loss of coolant accident (LOCA) occurs
Which of the following is the expected response of the LOP and LOCA sequencers?
c. As soon as power is restored to the buses, the LOCA sequencer will control the l
restoration of allloads. l
b. The LOCA sequencer will begin to sequence until the diesel generator output breakers
close, then the LOP sequencer will complete load restoration. ]
c. As soon as power is restored the buses, the LOP sequencer will control the restoration of
all loads,
d. The LOP sequencer will begin to sequence until the diesel generator output breakers
close, then the LOCA sequencer will compic*e load restoration.
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eage 24 e 4e
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Senior R:cctor Operator Excmination
53. Given the following conditions:
. The "B" Emergency Diesel Ger'erator (EDG) had started following a valid
.LOCA signal .
'
o Some time later the EDG was shutdown'using'the Iocal Emergency Top pushbuttons .
due to Ductuating oil pressure . ,
Concurrent with stopping the EDG, the 10A402 bus lost power
Which of the following describes the actions, if any, regarding resetting the Engine Shutdown
Relay (ESR) and the (86R) Lockout Relay to restart tha "B" EDG and reenergize the 10A402 -
bus?
.
a. ESR must be reset
'
'
' (86R) Lockout Relay reset is not required
b. ESR must be reset '
(86R) Lockout Relay murt be reset
c. ESR reset is not required . .
(86R) Lockout Relay reset is not required
d. ESR reset is not required
(86R) Lockout Relay must be reset
54. Which of the following parameter changes indicate the moisture content of charcoal adsorber ,
bed of the Gaseous Radwaste System (GRW) is rising?
a. GRW post-treatment radiation level due to Krypton is rising.
b. GRW charcoal adsorber bed temperature is lowering. j
c. GRW post-treatment radiatior, level due to lodine is rising.
d. GRW charcoal adsorber bed hydrogen concentration is lowering.
)
[
Page 25 of 46
Senisr R actor Operator Excmination
55. Given the following conditions:
* "he plant has been operating at 100% power for several weeks
* Main Steam, Line (MSL) radiation levels have been averaging *
80 mrem t,ut are now
~ '
' '
* -
slowly trending upwards -
* Chemistry has verified the higher radiation readings are due to failed fuel
What are the immediate Operator Actions required for the given conditions?
o. Place additional Condensate Demineralizers in service if possible.
b. Scram the reactor and close the Main Steam isolation Valves when MSL levels are 4
greater than 120 mrem.
c.' Direct Reactor Water Cleanup flow to'the main condenser to reduce coolant activity. .
d . Reduce rea,ctor power to maintain MSL radiation levels less than 120 mrem.
*
. .
56. Given the following conditions:
* The plant is operating at 50% power j
'
* All systems are orerating normally
.
* One Reactor Building Ver.tilation System (RBVS) Exhaust Fan discharge damper
has failed to the full "open" position with the fan running
* No other RBVS components have changed {
l
Which of the following describes how this will affect the initiation of the Emergency Core i
I
Cooling Systems (ECCS) and the reason for this?
c. ECCS will initiate after L is required because the failed damper lowers Reactor Building )
I
prest.ure resulting in a lower indicated drywell pressure.
b. ECCS will initiate before it is required because the failed damper raises Reactor Building
pressure resulting in a higher indicated drywell pressure, i
c. ECCS will initiate after it is required because th7 failed damper raises Reactor Bui! ding
pressure resulting in a lower indicated drywell pressure.
~ d. ECCS willinitiate before it is required because the failed damper lowers Reactor Building
pressure resulting in a higher indicated drywell pressure.
Page 26 of 46
Senior R3cctor Operator Examination
57. Given the following conditions:
. The plant is operating at 40% power
.
* The Jet Pump operability surveillance indicates that one jet pump has failed
~ '
. Technical Specificat;ons require the plant to' be in hot shutdown within 12 hours -
~
Which of the following describes why such a severe restriction placed on continued operation
for the given conditirAs?
c. A jet pump failure at this low power level will significantly affect the core flows and result
in unacceptable thermal limits (MCPR).
b. A jet pump failure may lim!; esactor water level restoration capability during the reflood
portion of a Loss Of Coolant Accident.
~
c. A jet pump failure combined with the flow restricting orifices may adversely affect core
flow to the higher power fuel bundles.
~d. ' A jet pump failure results in'less conservatiye protective ' action setpoints for ,
' instrumentation using recirculation loop flow as an input signal.
58. Given the followhg conditions:
. The "A" Recirculation Pump has tripped
* The "A" Recirculation Pump discharge valve is open
* RECIRC LOOP A JET PUMP FLOW (TOTAL) indicates 2 mlbm/hr
* RECIRC LOOP B JET PUMP FLOW (TOTAL) indicates 35 mlbm/hr
= RECIRC PMP B FLOW indicates 24,000 gpm
. Recirc pump "B" speed is 49%
Which of the following would be expected values for total JET PUMP FLOW (the flow
I
recorder) and actual core flow?
c. Flow recorder - 33 mlbm/hr, Actual core flow - 33 mlbm/hr
b. Flow recorder - 33 mlbm/hr, Actual core flow - 37 mlbm/hr
c. Flow recorder - 37 mlbm/hr, Actual core flow - 33 mlbm/hr
d. Flow recorder - 37 mlbm/hr, Actual core flow - 37 mlbm/hr
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Page 27 of 46
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Senior Racctor Operator Excmination
;
59. Given the following conditions:
The plant is operating at 90% power
' = All main turbine ~ sealing steam nonpal and backup supplies have been lost
'
'
'
Triere is-no tirrie estimate f6r repair /restorhtion -
-
'
Which of the following are the immediate operator act' ions' for the given conditions?
c. Reduce power as necessary to maintain condenser pressure less than 5.0" HGA.
b. Reduce recirculation flow to minimum, unload'and trip the main turbine. .
c.' Reduce power as necessary to maintain adequate self-sealing steam to the main turbine
seals.
^ '
d. Neduce recirculation flow t'o rnaintain power less tha'n 25% (BypsssValve capacity).
.
.
. . .
60. . During a loss'of off-site power the operator is cautioned not to. acknowledge the flashing ,
,
" Trip" pushbuttons for the 4.16 KV Vital 1 E Bus infeed breakers.
t
Which of the following will occur if these pushbuttons are pressed?
e. 'That bus' feeder breaker will attempt to close until the anti-pump feature causes it to trip
open and remain open.
b. The Diesel Generator associated with that bus, if running, will trip and its output breaker
will open.
c. That bus' alternate feeder breaker will trip open and then immediately reclose when the
pushbutton is released
d. The Diesel Generator associated with that bus will not load.
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Page 28 of 46
Senior Rzctor Operator Excmirction
61. Given the followng conditions:
. . * The plant is at 45% with power ascension to 100% in progress. -
. . . One of the Electrical Protection Asr;My (EPA) breakers on the "B" Reactor
. '
' ' ' ' ' ' ~
' ' "' Protection'Sys'teni(RPS) InG' set N er e kippied *
* Breakerinvestigation shows a trg o Vvaivoltage" .
-
,
. .
.
-
.
Which of the following describes the response of the Racirculation Pumps if a main turbine
trip occurs before the "B" RPS Bus is reenergized for the given conditions?.
. .
.
. .
'
e.' Both Recirculation Pumps runback to " minimum" speed. ,
'
b. The "A" Recirculation Pump trips, the "B" Recirculation Pump runs back to " minimum"
speed. . . .
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. .
,
, . .
.
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c. ~ Both Recirculation Pumps trip. . . .. , .
*
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d. 'The "B" Recirculation ' Pump' trips, the "d" Recirculation Pumh runs tiack'to " minimum" '
.
. , speed.; ,
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62. Given the following conditions: .
. . . ,
* A plant st' artup is in progress with the Reactor Mode Switch in "Run"
The Main Turbine is reset and is at 950 rpm accelerating to 1800 rpm
* A loss of 125 VDC power from distribution panel 1CD318 to the EHC control
logic occurs
.
Which of the following is the expected plant response?
a. Main turbine trips.
b. Main turbine startup would continue at the selected acceleration rate.
~
c. Me'n turbine speed will remain constant at 950 rpm.
d. Mal. turbine control valves throttle closed due to a loss of the speed reference signal.
1
Page 29 of 46
Senior R: actor' Operator Excmination
23. Given the following conditions:
* The plant is operating at 20% power ,, .
t A main generator load reject has just occurred ' ~'
,
d unexpectedly'during the lond reject
-
. The power / load, unbalance' circ 6it triphe'
,
.
. -
- - -
.
.- .
. . .
*
' Which of the following is the expected response of the Turbine Control Valves and the .
Reactor Protection System (RPS) for the given conditions?
c. - The Turbine Control Valves throttle closed -
'
*
- RPS"does not trip ,' *
,, ,
','
'
b. - The Turbine Control Valves fast close ' *
,
.RPS trips . .
. . ', -
. .- .
,.
* *
c. - The Turbine Cisntrol Valves throttle closed:
*
.
-
-
- RPS trips
' '
- - -
-
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d. - The Turtsine Control Valves fast close~
'
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.
-
- RPS ' does not trip -
*
.. . .
64. Which of the following describes when the Main Turbine is required to ba tripped'following a -
rcccior scram?
c. At 50 MWe lowering
b. At 25 MWe lowering
c. At 0 MWe
d. At 50 MWe rising (reverse power)
i
@5, During a failure-to-scram condition, which of the following is the criteria used to determine if - i
HC.OP-EO.ZZ-0100(O), " Reactor Scram", should be exited and HC.OP-EO.ZZ-0101(Q), ,
"Rcactor/ Pressure Vessel (RPV) Control", entered? I
a. Reactor period on SRM Period meters is stable at -80 seconds
b. All APRM "downscale" lights are not illuminated
c. All four RPS logic channels are deenergized
. d. All control rods are inserted to or beyond Notch "02"
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Page 30 of 46
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S:nisr R:rctor Oper; tor Exrmin:.ti:n
66. Following a reactor scram and Main Steam isolation Valve closure, reactor pressure reaches
1050 psig causing the "H" and "P" Safety Relief Valves (SRV) to open.
,, Which of the.following lists the operating,setpoints..for subsequent openings of the ",P" SRV7
.
a. SRV "P" opens at 1047 psig and closes at 935 psig.
b. SRV "P" opens at 1047 psig anci closes at 905 psig. '
c. SRV "P" opens at 1017 psig and closes at 935 psig.
d. SRV "P" opens at 1017 psig and closes at 905 psig.
67. With the plant at 100% power a severe overfeeding transient is' occurring.i Water level is +50 :
inches and rising rapidly.
.
Which of the following reactor water levels require termination of all feed to the reactor,
closing'the MSlVs and a reactor scram assuming none-of these actions have occurred? ,
c. +54 inches
b. +65 inches
c. +90 inch'es '
d. +118 inches
68. Given the following conditions:
)
. The plant is operating at 80% power
. All three Feedwater Pumps are in service
. Feedwater Level Control is in " Automatic - Three Element" control
. Narrow Range level"A"is reading 34 inches
* Narrow Range level"B"is reading 36.5 inches
* Narrow Range level"C"is reading 35.0 inches
Which of the following would be the expected response of the Feed Water Level Control
System and reactor water level if Narrow Range level"B" failed to the low end of the range?
a. It would transfer to Single Element Control and level would remain unchanged.
b. It would remain in Three Element Control and level would remain unchanged.
c. It would transfer to Single Element Control and would raise level by approximately 1.5 ,
inches.
d. It would remain in Three Element Control and would raise level by approximately 1.0
inches.
Page 31 of 46 1
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3:ni;r R::ctor Operctor Excmin tion
69. Which of the following is the basis of the 65 psig Suppression Chamber Pressure limit?
c. 65 psig is the primary containment maximum expected port-1.OCA pressure.
b. Above 65 psig, th'e system lineup required for containment venting may not be able to be
completed.
c. Above 65 psig, the Safety Relief Valves'may not be available when required for an-
Emergency Depressurization.
d. 65 psig is the operational limit of the Torus to Drywell vacuum breakers.
70. Given the following conditions:
The plant is operating at.95% power
* All Drywell Cooling Chilled Water pumps have tripped .
[ * Drywell pressure is rising l . HC.OP-AB.7_Z-0201, ."Drywell High Pressure / Loss Oi 9rywell Cooling", has been .
entered and the Reactor Auxiliary Cooling System (RACS) is being aligned to supply
backup cooling to the Chilled Water System
Which of the following describes the effect of failing to close the Chilled Water isolat' ion
Supply and Return Valves (HV9532-2 and HV9532-1) before the transfer to RACS7
c. The RACS Pump automatic start permissives will be bypassed until the valves are closed.
b. The RACS valves will not automatically sequence open to supply Chilled Water should a
t
loss of off-site power occur.
c. Chilled Water system flow will divert back into the RACS system overflowing the RACS
l head tarik. j
d. RACS system flow will divert back into the Chilled Water system overflowing the Chilled
Water head tank. 1
l
f !
I
I
i
l
Page 32 of 46
_ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _
Senior R cctor Operator Examinttion '
-
71. During a loss of coolant accident the following conditions exist - 3-
-
,
L- - .
- * Reactor pressureis 125 psig .
f-
* Drywell temperature is 325 *F -- - ,
.
Q p /. . .
.
.
. .. .
~
. Which of the following desent;es the andirkding capabilities of wide reactor .
water level indication for the given conditions?
- , -
.
c. They are not providing accurate ~ reactor water level or level trend information.
4
.b. .They are providing acc6 ate reactor water level but level trend is not reliable. .'
-
c. They are providing accurate reactor water level and level trend information.
' -
,
,
.d. They.arenot providing accurate reactor water level but level trend is reliable.
, , ,.
. . .
. . . .
.
, .
.
. . . . .
, .
_
72. Given the following conditions:
. .. . . .
.. , . .
.
.s.
*
4 The plant is operating at 95% power' ' '
+
*' Suppression pool temperature is 92 'F
.
* At 0915, Safety Relief Valve (SRV)"G" opened . .
* After several cycles of the SRV Open and Close pushbuttons, the ' operator notes .
that tailpipe temperature for the SRV is stable at 305 'F and NO other plant parameters
have changed
Which of the following describes the limitations on continued reactor operation for the given
conditions?
a. Reactor operation may continue until pressure set is reduced to less than 850 psig.
b. Reactor operation may continue until suppression pool temperature reaches 120 'F.
c. Reactor operation may continue indefinitely.
d. Reactor operation may continue until 0917.
. Page 33 of 46
,
1
Seni r R:cctor Operator Ex mination'.
73. Given the following conditions: '
. . !
* Reactor power is 82% g *
* HPCI is in operation for a surveillance -
' '
. The "B" loop of RHR is iri Suppres'sloh Pool Co'olin'g -# "
* Suppression pool temperature is 103 'F when the running pump tripped - - -
' '
* HPCIwas secured -
. Subsequently, suppression pool temperature incre to 106 *F
'
Which of the following lists the suppression poo mperatures requiring entry irito HC.OP-
'
EO.ZZ-0102, Primary Containment Control entry into the LCO actions for Tech Spec
3.6.2.17
'
' ' '
c. EOb102 - 96 *F sl '
-
TS 3.6.2.1 - 95 * N ' .
. . .
b. .EO-0102 - 95 * -
TS 3.6.2.1 -
F . .
,
.. .
'
' c. EO-0.102 - 105 *F '
,
- 95 'F
'
TS- 1
d. : -0102 - 105 'F.
- 105 'F
~
'
TS 3.6.2.1 '
'
Q.g.u7: .w a , 7m a .% "!k5 \ '
.
,
WWMY- I.lCc. kO'$. t,(T Ci f tt '. (l V'bE OI f i
74. Given the following conditions: (phtk Was btMb -
/
* The plant is operating at 100% power
. A feedwater heater trip has resulted in a feedwater temperature of 385 *F
. No operator actions have been taken
Which of the following is the operational concern for the given conditions?
c. Entry into the Exit Region of the Power-To-Flow Map.
b. Violation of the Hope Creek Operating License.
c. Immediate thermal hydraulic instabilities.
d. Recirculation Pump damage.
Page 34 of 46
.
. .
. .
..
S:nier Rocctor Operator Ex mination
75. Whi-h of the following describes how the operators would know the Hy ater
Chemistry injection (HWCl) system had NOT been removed from whiie performing a
shutdown in accordance with HC,OP-lO.ZZ-0004(O), "S, som Rated Power To Cold
. Shutdown"?
* . .- .
, .
f
c. Hydrogen explosions in the Mechanica - ump while operating to maintain
condenser va'cuum.
b. Post-shutdown (2 hours ine Building radiation levels would be much higher.
c. Alarms and in ions resulting from a control rod drop accident would not be available
to the o ors as quickly.
d e Primary and Secondary Condensate Pumps will cavitate. ,
.
DeJare6. .he m , ww' e - - 9 ; s J A - <. W .'- i '? i .
f MN n.v,I,dgthy ier,lr.h,g.
. q,. QST {"A 'i
'T a Y
76. Following a reactor scram all rods are at position "00". except one that is at position "24."
Which of the following describes the capability of the reactor to remain shutdown?
a. Control rods are inserted to beyond the Maximum Suberitical Banked Withdrawal limit,
~
therefore the reactor will remain shutdown under all conditions.
b. Control rods are not inserte'dto beyond the Maximum Subcritical Banked Withdrawal
limit, therefore it cannot be assured the reactor will remain shutdown under all conditions.
c. Design basis shutdown margin is met, therefore the reactor will remain shutdown under
all conditions.
d. Design basis shutdown margin is not met, therefore it cannot be assured that the reactor
will remain shutdown under all conditions.
1
Page 35 of 46
i
'S:nior Roccior Operator Extminati:n
77. Given the following conditions:
. The Control Room has been abandoned in accordance with HC.OP-AB.ZZ-0130(Q), '
" Control Room Evacuation" j
' "
+ Control has been established at the' Remote Shutdown Panelin accorda'nce with'
' -
'
.HC.OP-lO.ZZ 0008(Q), " Shutdown From Outside Control Roorn" -
'
RCIC is operating maintaining reactor water level at +35 inches ,
'
= Safety Relief Valves (SRV) are being used to cooldown
Condensate Storage Tank (CST) level is 135,000 gallons
The Condensate System is not available
Which of the following is correct for the given conditions?
~
'
e. RCIC is operated without overspeed protection.
-
'
b." insufficient CST inveritory is available to rlilow the cooldown.to clear the shutdown
cooling interlocks.
- c. The RCIC Gland Seal Condenser Cx.densate Pump must be manually operated..
. d. SRVs cannot be operated in a rotation that will evenly distribute heat to the Suppression
Chamber.
78. Which of the following describes the effect of failing to restart the Turbine Building Ventilation
System if it trips while operating in HC.OP-EO.ZZ-0104(Q)-FC, " Radioactive Release
Contror"?
c. The Turbine Building will go to a slightly negative pressure.
b. The total off-site release calculations will not be accurate.
c. The Turbine Building releases will be monitored but not treated.
d. The total off-site release will be higher.
79. A loss of Reactor Auxiliary Cooling System (RACS) has occurred.
Which of the following is the MAXIMUM time allowed before a reactor scram is required?
c. An immediate scram is required
b. One (1) minute
c. Ten (10) minutes
d. Twenty (20) minutes
Page 36 of 46
. Sentir Rrctor Operator Excminatian
80. Given the followwig conditions: 1
A loss of coolant accident has occurred .
.T. he Reaclor..Auxiliaries Cpoling System (RACS) ha. s been restored ,
.. .. .
..
. .
.. .
.. ... -. . .
. .
.
Which of the following describes the availability / response of the Emergency Instrument Air .
Compressor (EIAC) for those conditions should instrument air header pressure begin -
lowering?
c. ' The EIAC is.not available until the LOCA signal is cleared, reset and the 1E breaker is'. ~
closed. .
b. The EIAC will automatically start on instrument air header pressure less than 85 psig.
* c. The EIAC is riot Nailable until the Non-1E bre$keris closed and instrument air pressure ,
'
is'less than 85 psig. . . ,,
- . .;. .. .
. .
.. .
.
,
d. The EIAC will not automatically start but may be started manually from the Control Room
or locally. , , ,
,, ,,,',,,, , , ,, , ,
, ; - .
..
8.1. Which of the following describes.the reason control rods insert during a loss of instrument air?
,
~
a. A flowpath is opened to theibottom of the drive mechanism operating piston allowing
reactor pressure to drift the rod in,
b. Drive water flow and pressure both rise enough to drift the control rod in, similar to a
normalinsertion.
c. A flowpath is opened from the top of the drive mechanism operating piston allowing
accumulator pressure to drift the rod in.
d. The normal scram flowpath to and from the drive mechanism operating piston is opened, -
allowing accumulator and reactor pressure to drift the rod in.
82. Following a loss of shutdown cooling, decay heat removal is being transferrou to the
Alternate Shutdown Cooling lineup (Injection to the reactor retuming to the suppression pool
'
via open Safety Relief Valves).
Which of the following indications is utilized to monitor the margin to Op Con 3 entry in this.
lineup?
a. Safety Relief Valve tailpipe temperatures
b. Suppression pool temperatures
c. Reactor vessel skin temperatures
d. Local suction temperatures on the ru m g low pressure ECCS pumps
. Page 37 of 46
-<
-
S niar R cctor Operctor Excminrti:n 1
83. Which of the following describes the conditions requiring the Reactor Mode Switch to be
placed in " Shutdown" on a sustained loss of Control Rod Drive charging water header
pressure (<900 psig) with reactor pressure at 650 psig?
c. Within 20 minutes of determining more than one'CRD accumulator is inoperable and at -
least one of those inoperable accumulator.s is associated with a withdrawn control rod.
'
b. Within 20 minutes'of c'etermining any CRD accumulator is inoperable and the inoperab'le-
accumulator is associated with a withdrawn control rod.
c. Immediately upon determining more than one CRD accumulator is inoperable.and all the )
inoperable accumulators are associated with fully inserted control rods.
d. Immediately upon determining any CRD accumulator is inoperable and the inoperable
accumulator is associated with a withdrawn control rod.
*
4
84. Given the following conditions:
. . . . .
.
The plant is shutdown for refueling
The Reactor Protection System shorting links have been removed
A fuel bundle is being moved from the fuel pool to core.
If SRM "C" fails "downscale", which of the following are the required immediate actions?
4
a. Verify a control rod withdrawal block is received. Terminate fuel movement.
b. Verify a full scram and a control rod withdrawal block is received. Terminate fuel
movement.
c. Verify a control rod withdrawal block is received. Fuel movement is required to be
tr minated ONLY if the fuel bundle is to be placed in the cuadrant monitored by SRM "C."
d. Verify a full scram and control rod withdrawal block is received. Fuel movement is I
required to be terminated ONLY if the fuel bundle is to be placed in the quadrant
monitored by SRM "C."
Page 38 of 46
_
F
Ccni2r R=ctor Operc. tor ExcminaSon
.
85. Given the followng conditions:
A large break loss'of coolant accident has occurred
,
. Drywell pressure reached a maximum of 22 psig
" ' ' ' ~
Suppression chamber sprays twve'NOT been placed in service
'
>
Drywell sprays are in service . .
. -
Drywell pressure is 4 psig and slowly lowering -
Which of the following is the expected positions of the Torus-to-Drywell Vacuum Breakers and
. the Reactor Building 4o-Torus Vacuum Breakers for'the given conditions? -
c. - The Torus-to-Drywell Vacuum Brealiers are open
The Reactor Building-to-Torus Vacuum Breakers are open - .
. . . .. .. .
b.' .- The Torus-to-Drywe Vacuum Breakers are open .
. - The. Reactor Building 4o-Torus Vacuum Breakers.are' closed
-
,
,
,
c. - The Torus-to-Drywell Vacuum Br,eakers are closed
* - The Reactor Building 4o-Torus Vacuum Breakers are closed
.
.. - -
.
d, - The Torus-to-Drywell Vacuum Breakers are closed
_ - The Reactor Building-to, Torus Vacuum Breakers are open ,
,.
,
'
86. Given the following conditions:
The plant has experienced a loss of coolant accident
Suppression chamber sprays were placed in service when required
Drywell sprays were initiated with suppression pool level approximately 145 inches
Which of the following would be the result of these actions?
a. The Residual Heat Removal Pumps'will be operated outside the NPSH Limit Curves.
b. Excessive differential pressures between the suppression chamber and the drywell will
occur.
c. The suppression chamber venting flowpath will be damaged leading to loss of pressure
suppression capability.
d. The suppression chamber spray capacity will be lost.
Page 39 of 46
-
S:ni:r R ct r Operc.tcr Excmin:tian
87. Following a reactor scram with a Main Steam isolation Valve Closure, the plant is being
depressurized using the Safety Relief Valves (SRV).
Which of the following is the reason.why the depressurization should be accomplished with
.
" sustained" SRV opening's'if the pneumatic supply (PCIG and instrument air) is lost to the
SRVs?
c. This prevents exceeding the 100*F/ hour cooldown limit during the depressurization while
conserving the SRV pneumatic supply.
b. This ensures the available SRV pneumatic supply is sufficient to cooldown to less than
the shutdown cooling interlocks.
c. This directs depressurization without regard to the Technical Specification cooldown
.
limits-before the depleted pneumatic supply results in loss of SRV. control. - .
.d. This ensures the SRV accumulat.or pneumatic aupply is available and adequate for later
~
us's if the Ernergency Operating Procedures require Emergency Depressurizatiori.
.
88. The following data was collected following a Group 1 isolation and reactor scram from 100%
. power
.
* The Group 1 isolation was caused by technician error
The reactor scrammed on high reactor pressure
Reactor pressure peaked at 1060 psig
* All control rods fully inserted
. The plant was stabilized in Op Con 3
Which of the following is the basis for a decision not to startup?
a. A safety limit violation has occurred and the requirements of Technical Specification 6.7,
" Safety Limit Violation" must met.
b. The reactor steam dome pressure LCO was violated.
c. The Reactor Protection System did not respond as expected.
d. The Redundant Reactivity Control System did not respond as expected.
Page 40 of 46
i
S ni r Rocctor Operat2r Examin:. tion
89. Which of the following describes the basis for initiating boron injection before exceeding the
Boron injection initiation Temperature (BilT)? -
c. This ensures the reactor will be shutdown and in hot-standby conditions before the
suppression pool reaches the heat capacity level limit.
-
b. This ensures the reactor will be shutdown and in hot-standby conditions before the
suppression pool reaches the heat capacity targnerature limit:
c. This ensures the Primary Containment Pressure Limit will not be exceeded before RPV {
pressure is below the Minimum Alternate Flooding Pressure. I
d. This ensures suppression pool temperature will not exceed 150 *F during an Emergency
Depressurization, if required.
'
90: Given the following condition: -
. The plant is operating in HC.OP-EO.ZZ-0206, " Reactor Flooding"
. Suppression chamber pressure is 22 psig
'
. Reactor pressure is 105 psig -
4 SRVs have been opened and have remained open for 85 minutes
All reactor water levei indicators are off-scale high
Which of the following would be the MINIMUM expected actual reactor water level for the
given conditir 9
c. -209 inches
b. -161 inches
c. +118 inches
d. Filled solid
Page 41 of 46
Senior R :ctor Operator Excmination
91. HPCI and RCIC both started and are injecting in response to a valid low reactor water level.
Current plant conditions are as follows:
Reactor water level is +25 inches, steady .
' '
-
4 Reactor pressure is'845 psig, rising slowly -
'
Drywell pressure is 1.1 psig, steady .
.
. .
RCIC has been aligned to Full Flow Recirc operation (CST to CST) for pressure control
HPCI is injecting to the reactor for level control
After 10 minutes of operation a valid high suppression pool level is received
Which of the following would be the expected response of RCIC if a valid high suppression
pool level is received for the given conditions?
'
' -
o. RCIC will remain in Full Flow Recirculation.
'
l
' '
b. ,RCIC will trip on high turbine exhaust pressure. ,
.
c. RCIC will trip on low suction pressure.
.
'
d. NCIC will' operate on minimum flow.'
.
e.
,
92. During high primary containment water level conditions, suppression pool water level
indications cannot be used.
Operation of which system will invalidate the alternate method used for delpmining primary
containment water level?
a. RCIC
b. Core Spray j
c. RHR i
d. HPCI
)
l
l
Page 42 of 46
a
Senior R actor Operctor Examination
93. Given the following conditens:
,
. A leak has occurred its the suppression pool 1
' '
* The reactor 's shutdown
'
-
~~
-
. A cooldown is being performed using SRVs' -
-
- - *: # -
-
-
. The Heat Capacity Level Limit (HCLL) curve is being monstored - -
,
.
'
..The " Action Required" area of the HCLL curve has been entered for several minutes
Which of the following is a possible effect of initiating an emergency depressunzation with the
'
given conditions? .
.
,- , ,
c. The suppression pool may exceed design temperature.'
,b. Failure of the downcorner vent headerjoints due to " chugging." -
,
-
..
.. .
.
'
c. The SRV Tailpipe Level Limit. curve may be exceeded. .- , .., .
~
,
' ~ ^ ~ '
N. The capacity of the Torus to Drywell vacuum breakers wib be' exceeded.
, . . .. . . . .
.
,
.
'
'94.,' Following the runback of the' Recirculation Pumps on a trip of a Primary Condensate Pump,
the operator may monitor the Source Range Monitoring (SRM) per,iod meters for strong.
deflections above and below " Infinity". *
,
. .
Under which of the following conditions may SRM period indications be considered accurate -
indication of thermal hydraulic instabilities? -
e. Only when the SRM detectors are fully withdrawn f:om.the core.
b. Anytime, regardless of detector position, if the detectors are stationary.
c. Only when the SRM detectors are fully inserted into the core.
d. Anytime the SRM detectors are moving.
Page 43 of 46
S:nior R :ctor Op:rct r Excrnintti:n
95. With the plant at power the Main Steam / Reactor Water Cleanup Area Leak Temperature
High alarm was received and the RWCU system automatically isolated. The leak has been
determined to be in the RWCU Pipe Chase Room 4402.
Whi6h of the following is NOT a required' operator sctionfor the given' conditions'?
~
c. Notify Chemistry to close'the" Manual Sample Line Isolation Valves P-R'C-V9670 ahd 1-
RC-V006.
b. Observing the RWCU inboard and Outboard Isolation Valves (F001 & F004) close.
c. Observing the Recire Sample Line isolation Valves (BB-SV-4310 and 4311) automatically
close.
d. Operate available Reactor Building ventilation fans consistent with plant conditions.
..
96. Given the following conditions:
* The plant was operating 'at rated power when a steam line break occurred in the HPCI
room
+ HPCI isolated due to high room temperatures
* RBVS exhaust radiation levels reached-1.0 E-2 microcuries/ml
Which of the following describes the ventilation system response for the given conditions?
c. RBVS remains in service
b. RBVS isolated,6 FRVS Recirc and 1 FRVS Vent Fans are in sewice
c. RBVS isolated,4 FRVS Recirc and 1 FRVS Vent Fans are in service
d. RBVS isolated,6 FRVS Recirc and 2 FRVS Vent Fans are in service
97. With the Reactor Building Ventilation in a normal lineup, the operator observes that Reactor
Building pressure is .10 inches of vacuum water gauge.
Which of the following is an immediate action to restore Reactor Building pressure to the
required pressure?
a. Place at least two FRVS units in service.
b. Secure a reactor building supply fan.
c. Place an FRVS unit in service and increase FRVS flow rate to maximum.
d. Place the third Reactor Building Exhaust Fan in service.
Page 44 of 46
_-
' l. [ ,
Senior R cctor'Operctor Excmination
98. Given the following conditions:
.
* The reactor hos scrammed from power . .
.
* Pilot, Scram Valve Group 1 solenoids, Logic "A" and Logic "B" did not doenergi,ze
'
' ~ ' '
" - '-
- * The Senkm Discharge Volume isi currently full - -
. .
Which of the following desenbes the difference between inserting control rods in accordance
with HC.OP-EO.ZZ-0303, " Individual Control Rod Scrams" and HC.OP-EO.ZZ-0302, "De-
energization Of Scram Solenoids"?' ,
.
.
a. EO-0302 requires resetting RPS and ARl, EO 0303 does not.
.
b. EO-0303 requires resetting RPS and ARl, EO 0302 does not.
~
c. EO4303 does not isolate the Scram Disdurge Volume, EO-0302 'doesc. - .
,
, d. ,EO 0302 results in a flowpath from the RPV to the Reactor Building Surnp, EO-0303 does'
' '
not.
.
. .. . .
,,, .
.
99: Which 'of the following are the appropriate hydrogen concentration values to complete the
.
following statement following a loss of coolant accident with hydrogen generation occurring?
Rising containment hydrogen concentrations require corrective actions be taken at
and reentry into HC.OP-EO.ZZ-0102, " Primary Cordainment Control", at
.
a. 2.0%,- 0.5%
b. 0.5%, 2.0%
c. 2.0%, 2.0%
d. 0.5%, 0.5%
)
l
Page 45 of 46
g
t ,
Senior R; actor Operator Excmination
J00. Given the follomng conditions:
* A loss of coolant accident has occurred '
* Hydrogen is present in the primary containment
,
' '
- ~~
* The Hydrogen Recombiners have tieen' started. - '
'
.
,
*
1
^
Which of the folloiving is the hydrogen concentration that requires termination of Hydrogen
Recombiner operation and the reason why that value is selected? ,
c. The Hydrogen Recombiners are secured at 4% hydrogen concentration because there is
insufficient oxygen available to support the recombination reaction.
b. The Hydrogen Recombiners are secured at 6% hydrogen concentration because there is
, insufficient. oxygen available to support the recombination reaction.
,
.. .
.
^
c. The Hydrogen Recombiners are secured at f % hydrogen conceritration in order to , , ,
* -
prevent their becoming an ignition source.. -
.
.
~
'
,d. The Hydrogen Recombiners are secured at 6% hydrogen concentration in order to
prevent tiiisir becoming'a'n igriitidn'sou'rcs.
- .
,
.
. .
- -
.
_
)
l
.l
.
1
Page 46 of 46 .
i
t
i
Senior R: actor Operator Answer Key
1
1. b 294001G101 26. d 203000K406
-
2 .. a 294001G102- .. .
27. c.. 204000K115. ..
.
.. .
.
.
.
3. d 294001G104
. .
-
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a y'
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-
.
12. a 294001G304 37. d 215002K604
13. c 294001G310 38. d 215004A104
14. b 294001G412 39. b 215005K104
15. d 294001G440 40. d 216000A301
16. b 294001G441 41. c 217000A210
17. d 294001G448 42. b 217000K201
18. a 201001K405 43, c 218000K201
19. c 201002A405 44. c' 218000K302
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21. a 201006K514 46. c 223002A403
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,
25. b 202002A101 D# '* :l' Ob 50. b 239002A109
Page 1
Senior R: actor Operator Answer Key 51. c 241000K302 76. c 295015A202
4
52, o 262001A304 .
. . 77 . c. . 295016A108 .s . . .
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- 53'. b '264000K603 .. .
-
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-
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59. .a . 295002A1.05 . .
84. a 295023.G232 ..
60. d 295003A101 85. b 295024A116
61. c 295003K204 86. b 295024K101 .
62. a 295004K203 '87. d 295025K102
63. d 295005K201 88, c 295025K201
64. c 295006G449 89. b 295026K304
65. b 295006K103 90. b 295028K302
66 a 295007K304 91. d 295029A104
67, c 295008G123 92. d 295029A201-
68. d 295009K2C2 93. a 295030K103
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3/4.0 AFFLIM mILITY .
% LIMITING CONDITION FOR OPERATION - . . . .
. ... ........................... ....... _
- 3.0.1 Compliance with the Limiting conditions ,for Operation contained.in the- l
succeeding Specifications is required during the'OFERAT10NAL CONDITIOirs or
other conditions specified therein; except that upon failure te meet the -
Limiting conditions for Operation, the associated ACTION requirements, shall be
met.
3.0.2 Noncompliance with a Specification a all exist when the requirements of
the Limiting' Condition for Operation and associated ACTION requirements are
not met within the specified time intervals. If the Limiting condition for
Operation is restored prior to expiration of the specified time intervals,
completion of the Action requirements,is not required.
J'.0.3 Nhen a Limiting cond1+ Nn for Operation is .not met, except as provided
in the associated ACTION r< # resents,' within one hour action shall be -
hich the
, initiated to place the unis in an OPERATIONAL CONDITION in w
,
* speci2Lcation does not apply 'by placing it, ais applicable, in ,
1. At least'STARTUF within-the next 6 hours,
2. At.least NOT SHUTDONN within the following 6 hours, and
3. At least ' COLD SRUTDONN,within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION
- requirements, the ACTION may be taken in accordance with the specified time
limits as measured from the time of failure to eset the Limiting condition for
Operation. Exceptions to these requirements are stated in the individual *
specifications.
This specification is not applicable in OPERATIONAL CONDITIONS 4 or 5.
3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall
not be made when the conditions for the Limiting Condition for Operation are *
not met and the associated ACTION requires a shutdown if they are not met
within a specified time interval. Entry into an OPERATIONAL CONDITION or
other specified condition may be made in accordance with the ACTION
requirements when conformance to them permite continued operation of the
facility for an unlimited period of time. This provision shall not prevent
passage through or to OPERATIONAL CONDITIONS as required to ctmply with ACTION
requirements. Exceptions to these requirements are stated in the individual
specifications.
3.0.5 Equipment removed from service or declared inoperable to comply with
ACTIONS may be returned to service under administrative control solely to j
Perform testing required to demonstrate its CPERARILITY or the OPERABILITY of I
other equipment. This is an exception to Lc0 3.0.2 for the system returned to
service under administrative control to perform the testing required to
demonstrate OPERABILITY.
Amendment No. 63 l
HOPE CREEK 3/4 0-1
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APPLICABILITY
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SURVEILLANCE REQUIREMENTS
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. 4. 0.1 Surv.el11anci Requirements shal'1 be' met during the OPERATIONAL CQNDITIONS ^ ~
cr other conditions specified for individual Limiting Conditions fdr Operation
unless otherwise stated in an individual Surveillance Requirement.
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.4'. 0. 2 Each Surveillance Recutrement shall be nerformed within its specified
surveillance interval with a maximum allowable extension not to exceed
25 percent of the specifled surveillance interval.
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4.0.3 Failure to perform a Surveillance Requirement within the allowed
surveillance interval, defined by Specification 4.0.2, shall constitute a
failure to meet the OPERASILITY requirements for a Limiting Condition .for
-0peration. The time. limits"of the ACTION requirements are applicable at the ,
time it'is identified that a Surveil. lance Requirement has not been performed,
.
The ACTION requirements may be delayed for up to 24 -hours to permit the
'
completion'of the surveillance when the allowable outage time limits of the
ACTION regt. tenants are less than 24 hours. Surveillance roovirements do not.
have to be pe- ormed on' inoperable equi' pment
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4.0.4 EntryintoanOPERATIONALCONDITIONorotherspecifiedapplicable.cond5-
tion shall not be made unless the Surveillance Requirement (s) associated with
the Limiting Condition for Operation have been perfarmed within.the. app 11 cable
surveillance. interval 'or as otherwise specified. 'This provision shall not
prevent passage through or to OPERATIONAL C0feITIONS as required to comply
with ACTION requirements.
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4.0.5 Surveillance Requirements for inservice inspection and testing of ASME
Code Class 1, 2,-& 3 components shall be applicable as follows:
)
a. Inservice inspection of ASME Code Class 1, 2, and 3 components and
inservice testing of ASME Code Class 1, 2, and 3 pueps and valves
shall be performed in accordance with Section XI of the ASME Boiler
and Pressure Vessel Code and applicable Addenda as required by 10
CFR 50, Section 50.55a(g), except where specific written relief has
~
been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)
(6) (1). 4
b. Surveillance intervals specified in Section XI of the ASM! Soiler
and Pressure Vessel Code and applicable Addenda for the inservice
inspection and testing activities required by the ASME Boiler and
.
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HOPE CREEK 3/4 0-2 Amendeent No. 37
APR 3 h190
i
HC.OP-SO CH-0001(Z)
ATTACHMENT 4
(Page1of1)
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MAIN TURBINE CONTROL OIL (EHC) SYSTEM OPERATION - . .
'EHC CONTROL LOGIC DIAGRAM -
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H:pe Creek Page M2 of 84 Rev.19
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ATTACHMENT 2
! i '
RO EXAM AND ANSWER KEY
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U.S. Nuclear Regulatory Commission
'
. Site-Specific .
Written Examination.
Applicant information
Name: Region: 1
Date: Date:. 2/23/98 -
Facility: Hope Creek
License Level: RO Reactor Type: GE
.
Start Time: Finis'.a Time:
Instructions
Use the answer sheets provided to document your answers. Staple this cover sheet
on top of the answer sheets. The passing grade requires a final grade of at least
80.00 percent. Examination papers will be collected four hours after the examination
starts.
Applicant Certification
All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature
l
Results
Examination Value Points
Applicant's Score Points
Applicant's Grede Percent
_a
R: actor Operator Answer Sheets
Circle the corred answer. If an answer is changed wrne R in the blank.
1. a b c d 26. a b c d:
' 2. a b c d - -
27.. a bac d .
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28. a b c d
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3. a b c d
4. a b c d 29. a b c d
5.abcd 30. a b c d
6. a b c d 31. a b c d
32.' a b'c d
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7. a b c d <
8. a b c d- 33, a b c d - -
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9. a b c d 34. a..b c d . .- , -
10. a b c d 35. a b c d
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11. a b c d 36. a b c d .
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12. a b c d 37. a b,c d
13. a b c d 38, a b c d
14. a b c d 39. a b e d
15. a b c d 40. a b c d
16. a b c d 41. a b c d
17. a b c d 42. a b c d
18. a b c d 43. a b c d
19. a b c d 44. a b c d
20. a b c d 45. a b c d
21. a b c d 46. a b c d
22. a b c d 47. a b c d
23. a b c d 48. a b c d
24. a b c d 49.- a b c d
25. a b c d 50. a b c d
Page1
Reactor Operator Answer Sheets
Circle the correct answer.,if an answer is changed write it in the blank.
51. a b c d 76. a b c d
52. a. b c *d - '
. 77. a b c d .
. ..
53. a b c d 78. a b c d
54. a b c d 79. a b c d
55. a b c d 80. a b c d
56. a b c d 81. a b c d
'57 abcd
*
'82. a b c d
58. a b c.d 83. a b c d
' 59, a b c d , , .
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. 84. a b c d ,
60. a b c d- ,
85. a b c d
61. a b c d 86. a b c d -
62. a b c d 87. a b c d
63. a b c d 88. a b c d
64 abcd 89. a b c d
SS,abcd 90. a b c d
66. a b c d 91. a b c d
67. a b c d 92. a b c d
68. a b c d 93. a b c d
69. a b c d 94. a b c d
70. a b c d 95. a b c d ,
71. a b c d 96. a b c d
72. a b c d 97. a b c d '
73. a b c d 98. a b c d
74. a b c d 99. a b c d ;
75 abcd 00. a b c d
Page 2
Rrctor Operator Excminatian
1' Which of the following evolutions is NOT allowed to be performed by the Reactor Building
'
Equipment Operator?
a. Transferring an RPS bus to its alternate power supply with the reactor at power.
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b. ' Test scramming a control rod'from thelndividual test switches on the hydraulic control
unit.
c. Operating the Standby Liquid Control system in the Test Tank to Test Tank mode.
d. Reducing hydraulic control unit nitrogen pressure to the normal band with the
associated control rod withdrawn.
2. Given the following conditionsf
A fully qualified Nuclear Control Operator (NCO) with an active license has just
returned from 1.0 days vacation
5 On the first day back on shift, this NCO worked a normal 12-hour shift arid then -
accepted and worked.4 hours of overtime
Which of the following is the maximum number of hours this NCO may work on the second
day back on shift? (Assume no additiorial authorizations have been made.)
a. 8 hours
.
b. 12 hours
c. 14 hours -
d. 16 hours
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3. A tagging request with switching order has been received from the System Operator. The
Switching Order hac been confirmed and the tags prepared. The System Operator has
contacted Hope Creek and directed the performance of the tagging request and switching
order.
Which of the following personnel are required to be present in the 500KV switchyard
blockhouse for completion of the tagging request and switching order?
a. A Nuclear Equipment Operator and a Nuclear Control Operator.
b. Two Nuclear Equipment Operators.
c. A Nuclear Equipment Operator and a Control Room Supervisor.
d. A Nuclear Equipment Operator and a member of the Systems Operation Department.
Page 1 of 45
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Rocctor Operator Excminati n
4. Following shift turnover the Nuclear Control Operator (RO) notes that data entered in t
narrative log by the previous shift is incorrect.
The RO draws a single line.through the incorrect entry, make? the corr entry and initials
and dates the change. Which of the following describes how the should highlight and
cxplain the change?
a. The correct entry should be circled in red wit explanation placed in the comments
section.
b. The correct entry should be cir in red with an explanation made next to the j
co acted entry.
c. The incorrect ent ould be circled in red with an explanation placed in the comments
section.
d. The rrect entry should be circled in red with an explanation made next to the
crected entry.
Deleted See cm ~!rds artwo ff4C 3-fAV ,
5. Which o'f the following will identify when Op Co'n 2 is entered'during a reactor startup and
heatup?
a. When the reactor is declared critical.
b. When the first control rod is withdrawn.
c. When the MODE switch is placed in Startup/ Hot Standby.
d. When enough control rods are withdrawn to increase keff to greater than or equal to .99.
6. During a valid high reactor pressure condition, the Recirculation Pumps did NOT
cutomatically trip as designed.
Which of the following actions must be taken by the Control Room to open the Recirculation
Pump Trip (RPT) Breakers.
a. Manually initiate both channels of the Redundant Reactivity Control System (RRCS).
b. Verify the RPT Breakers trip when the Recirculation Pump MG Set Drive Motor Breakers
are opened. ,
c. Direct the local tripping of the RPT Breakers.
t d. Depress the RPT Breaker" Trip" pushbuttons. ,
Page 2 of 45 f
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R::ctor Operator Excminati:n
7. Which of the following are the minimum requirements for the ' Board" Nuclear Control
Operator (RO) to review critical plant parameters (reactor power, level, pressure and turbine
load) and walk down the control boards during normal, steady-state plant operations?
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The RO should:
a. continuously monitor critical plant parameters and perform a complete control board
walk down every hour.
b. monitor critical plant parameters every five (5) minutes and perform a complete control
board walk down every two (2) hours.
c. continuously monitor critical plant parameters and perform a complete control board 8
walk down every two (2) hours.
d. monitor critical plant parameters every five (5) minutes and pe. form a complete control i
board walk down every hour.
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8. Given the following conditions:
A plant shutdown with control rod insertions occurring is in progress
Reactor power is 22% with generator output at 242 MWe
The second NCO (PO) begins deinerting the drywell
The CRS is reviewing procedures at the CRS desk
No other personnel are in the Control Room ,
!
Which of the following additional requirements, if met, would allow a License Class instant
SRO trainee under direction of the Nuclear Control Operator (RO) to continue rod motion for
the given conditions?
a. Operations Manager written permission to allow a License Class trainee to insert control
rods.
b. Another technically qualified member of the unit technical staff to observe rod movement.
c. Verification that the Rod Worth Minimizer is operating properly before reducing power
below 20%.
d. A Reactor Engineer's presence to satisfy Technical Specification requirements.
Page 3 of 45
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R: actor. Operator Examinatinn ,
C Given the following conditions:
. The plant is shutdown for a maintenance outage
.- 'e A Red Blocking Tag (RBT) is hung on 4160. VAC breaker . '
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* Th's breaker is' tagged in the " Test Disconnect" position
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. Later in the outage, the breaker is being removed from its cubicle for maintenance
,
Which of the following describes the required tagging actions fo- the given conditions?
a The RBT shall remain on the breaker, the breaker removed from the cubicle and an
additional RBT installed on the ropettape placed across the opening.-
b. The RBT shall be removed from the breaker but kept active and maintained in the
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physical possession of Operations while the breaker is out of the cubicle. .
c. The RBT shall be removed fr am the breaker, the breaker removed from the cubicle and
the sarde R'BT installed on the' safe'yt rope / tape placed across the cubicle opening.
.d.. The RBT.shall remain on the breaker, the breaker removed from the cubicle md a.
White Caution Tag installed on the safety rope / tape pla'ced across the' cubicle opening.
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10. Given the following conditions:
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+ A Hope Creek radiation worker is fully qualified with current lifetime exposure
records on file
* This individual's current yearly exposure (TEDE) is 355 mrem
. A Site Area Emergency hasjust been declared
Which of the following is the MAXIMUM additional exposure that can be received by this
individual without exceeding eny administrative or procedurally based limits? (Assume no
additional approvals have been received.) .
a. 1645 mrem
b. 4145 mrem
c. 4395 mrem
d. 4645 mrem
Page 4 of 45
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Rccctor Operator Excmination 11. The ostimated time to independently verify a valve position is 15 minutes.
<
Of the listed dose rates, which is the MINIMUM dose rate that would allow waiving the ' Hands
On" independent verification requirement for the~ conditions givsn? ,- , ,
. 'a. 10 mrom/hr
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' b. 30 mrom/hr -
c. 45 mrem /hr
d. 60 mrem /hr
12. An emergency tias occurred immediately requiring reasonable actions to be taken that ' depart
from Technical Specifications. No actions consistent with Technical Specifications that can .
provide adequate equivalent' protection are immediately apparent. -
. Which of the following ident'ifies who is required to approve the action and under what
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_conditioris the action can be performed? ,
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a. The Control Room Supervisor approves actions to be taken to protect the health and
saf,ety of facility personnel. -
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b. The Control Room Supervisor approves actions to be taken to protect the health and
safeiy of the public.
c. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to I
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be taken to protect the health and safety of facility personnel.
d. The Emergency Coordinator, in the Emergency Operating Facility, approves actions to
be taken to protect the health and safety of the public.
Page 5 of 45
R: actor Operator Extminati:n
13.. Given the following condstens:
.
. The plant is performing a shutdown in accordance with 60-0004, " Shutdown
. From. Rated Power To Cold Shutdown" . .
" '.
At 20% power the shutdown is complisted by placing the Reacto'r Mode Switch' -
to " Shutdown" -
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* All plant systems responded as designed during the scram -
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. Post Scram Actuation Review is in progress in accordance with HC.OP-AP.ZZ-0101,
Post Reactor Scram /ECCS Actuation Review and Approval Requirements.
Which of the following should be the FIRST reactor scram signal identified when reviewing .
the Sequence.Of Events printout?
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" a. Reactor Mode ~ Switch in shutdown"
*
b; IRM Neutron Flux-High
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c. . Scram Discharge Volume Water Level - High
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.'d. APRNI N$utron Flux - Upscale. Setdown
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1'4 . Given the following coMitions: .
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The plant is operating at 55% power
. All systems are operating normally in automatic
Which of the following is the. expected response of the Scram Discharge Volume (SDV) vent
End drain system if APRM Channel "A" fails full " upscale"?
a. One Scram Dump Valve repositions, all SDV Vent and Drain Valves close.
b. One Scram Dump Valv.e repositions, all SDV Vent and Drain Valves remain open.
c. The Scram Dump Valves do not change. position, all SDV Vent and Drain Valves remain
open.
d. One Scram Dump Valve repositions, one set of SDV Vent and Drain Valves close,
Page 6 of 45
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R actor Operator Extminatian
15. Given the followng conditions:
* The plant is at normal operating pressure and temperatures -
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. * All plant systems are ope $ating as designed , ,, ,. '
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+ The "A" and "B" scram toggle switches at the hydraulic conti'ol urit 'for [
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control rod 42-03 have been placed in " Test"
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Which of the following is the expected response of the Scram Pilot Valves for Control Rod 42-
03 and the Scram Dump Valves for the given conditions 7
a. - The Scram Pilot Valves reposition to vent the Scram inlet and Outlet Valves
- The Scram Dump Valves remain in their initial positions
b. The Scram' Pilot Valves remain ~in their initial positions .
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. The Scram Dump Va.lves re. main in their initial positions .
c. - The Scram Pilot Valves reposition to vent the Scram inlet and O'utlet Vaives
- The Scram Dump Valves reposition to vent the Scram Discharge Vent and Drain
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Valves
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-d. - The Scram Pilot Valves remain in their initial positions
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- The Scram Dump Valves reposition to vent the Scram Discharge Vent and Drain '
Valves. . . ,
16. Given the following conditions:
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* The plant is performing the control rod' exercise s'urveillance . .
* The Nuclear Control Operator (RO) selects control rod 34-19 on the rod select module
* Only one half of the selected rod pushbutton illuminates
Which of the following describes what has failed and how that affects the ability to move
control rods?
a. The selected control rod activity control card is in the. scan mode and rod motion is
allowed.
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b. The selected control rod activity control card is in the scan mode and rod motion is not
allowed.
c. Only one of the two RMCS transmitter cards has successfully selected the control rod
and rod motion is not allowed.
d. Only one of the two RMCS transmitter cards has successfully selected the control rod
and rod motion is allowed.
H Page 7 of 45
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R cctor Operator Ex::minatirn -
17. Given the following conditions: 1
* The plant is opwating at 25% power performing a startup '
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. Control rod 18-23 has been determined to be stu::k '
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. V tiile attempting to'w'ithdraw the control rod, indicated drive water flow is reading'
'U' gpm.
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Which of the following is the cause of this indication?
a. Hydraulic Control Unit Directional Control Valve (122) has failed to reposition.
b. The 2 gpm Stabilizing Valve has failed to reposition.
c. Bot.h Cooling Water Header to Exhaust Header Pressure Equalizing Valves have failed
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open.
. d. 4 The Drive Vater Header Pressure Control Valve has failed closed.
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18. The current Rod Worth Minimizer (RWM) group has insert and withdraw limits of No' h 24
cnd Notch 36 respectively.
Which of the following are the control rod altemate limits allowed by the RWM for this group?
a. Notch 22 and Notch 34
b. Notch 22 and Notch 38
c. Notch 26 and Notch 34
d. Notch 26 and Notch 38
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Page 8 of 45
R: actor Operator Extmination
19. Given the following' conditions:
The is og ;ating at 75% power . .
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* Confirmed,
, failures,have occurred on the "B",Recircul.ation Pump
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'*The pump has ' tripped .
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Whim of the following de
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' the order for "B" Recirculation Pump valve manipulation that .
must be followed in order to ens pump will be completely isolated?
a. Cicae the Discharge Valve, isolate al purge, isolate RWCU flow from the loop and
close the Suction Valve.
b. Isolate the seal purge, close the Suction Val isolate RWCU flow from the loop and
close the Discharge Valve.
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c. Close the Suction Valve, close the Discharge Valve, i te seal purge, and isolate. ,
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RWCU flow from the loop.
,d. .lsolate the. seal, purge, close.the ,Di,schar Vajve isojate RW ow frope , loop and.
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close the Suction Vbive,
gp
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20: Given the following conditions:' ,
Preparations are complete to start the "A" Recirculation Pump
The Pump Discharge Valve (F031 A) is closed
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Which of the following describes how the. "A" Recirculation Pump trip ~on the discharge valve
closure is bypassed to allow the pump to be started?
a. This trip is bypassed until the pump start sequence is complete within prescribed time
limits,
b. This trip is bypassed until the discharge valve has reached the 100% open position.
c. This trip is bypassed until the pump has been running for.9 seconds,
d. This trip is bypassed until the discharge valve jog (open) circuit has timed out.
21. With the plant at 100% power, which of the following would cause a drop in reactor power and
a rise in the "A" Recirculation Loop drive flow? ,
a. A jet pump has failed in the "B" Recirculation loop.
b. The "B" Recirculation Pump speed has risen.
c. A jet pump has failed in the "A" recirculation loop.
d. The "A" Recirculation. Pump speed has risen.
' Page 9 of 45
R:cctor Operct r Ex min: tion
22. Given the following conditions:
. The plant is operating at 60% power with the "B" Recirculation Pump scoop tube locked
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. 'The operator is preparing to reset the scoop tube
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. Speed demand on the "B" Recirculation Purnp is slightly LESS than indicated speed
Which of the following actions is the operator directed 'o
t perform if pemp speed begins to
slowly rise at approximately 2% per minute after the Scoop Tube Trip / Reset pushbutton is
pressed?
a. Immediately press the "A" Recirc MG Set Drive Motor Breaker Trip pushbutton.
b. Attempt to control speed with the increase / Decrease arrows on the Pump Speed Control
Station - the "B" Recire pump.
c. P.ress u,e Scoop Tube Trip pushbutton for the "B" Recirc pump.
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d. Press the Recirc Pump Trip (RPT) Breaker Trip pushbutton for th'e "B" Recirc pump.
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23" Which of the following is the MAXIMUM speed at which the' Recirculation Pumps can operate !
..with NO Reactor Feedwater Pumps operating?
a. 20%
b. 30% j
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c. 45%
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d. 50%
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Page 10 of 45
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Rrctor Operator Excmin;tian
. 24. Given the following conditions:
The plant is operating at 75% power .
. Valve stroke time testing is in progress on the "A".RHR Pump Torus Suction
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Valve (F004A), ,
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The valve is currently closed _ '. .
t All other RHR' system components are iritheir normal standbi dirmup
- A steam break causes drywell pressure to reach 2.0 psig.
Which of the folicwir.g describes the response of the F004A valve and the "A" RHR pump?
a. The F004A valvo automatically opens and the "A" RHR Pump automatically starts after
, F004A is fully open.
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,
.b. The F004A valve must be manually opened and the "A" RHR Pump automatically starts
.
J
after F004A is fully open. .
.
.
c. The F004A valve automatically opens but the "A" RHR' Pump must be, started by the
'
'-
~ operator after F004A l~s fully open '
'
*
d. The F004A valve must be manu' ally opened and the "A" RHR Pump manually started
after F004A is fully open.
.
25. Given the following conditions:
. The plant is operating at 90% power
The Reactor Water Cleanup (RWCU) Inboard Isolation Valve (F001) has just
I
stroked closed
No other RWCU valve repositioned l
RWCU responded as designed
Which of the following initiated the RWCU isolation? a
a. RWCU system differential flow is excessive.
b. The RWCU Filter /Demineralizer inlet temperatures are excessive,
c. The "A" Reactor Protection System MG set tripped.
.d. The "A" and "D" NSSSS Manual isolation pushbuttons have been armed and depressed
- simultaneously.
l
Page 11 of 45
<
_ .
Rscct2r Operator Extminntian
26. Which of the following describes the reason for having the capability to bypass the Residual
Heat Removal (RHR) Pump suction path interlocks?
a. Allows operation of the RHR Pumps for shutdown cooling from the Remote Shutdown
'
- - -
- - - -
Panel. -
-
-
b. Allows the "C" and "D " RHR Pumps to be utilized for an altemate path of suppression -
' -
pool heat removal.
c. Allows operation of the RHR Pumps for cooling the containment Hydrogen Recombiners
post-LOCA.
'
d. Allows the "C" and "D" RHR Pumps to be utilized for an alternate path of core decay
heat removal.
S
'
27. The plant is in Mode 4 with Shutdowa Cooling in servics on the "A" Residual Heat Removal
(RHR) loop with the "A" RHR Pump running. -
. . ,
,
Which of the follopag describes how a loss of the "B," Reactor Protection System (RPS) bus
will affect the inboard and Ouiboard Shutdown Cooling isolation Valves (F008 & F009)7
a. The F008 and F009 valves'both close. . .
b. The F008 valve closes and the F009 valve remains open.
c. The F008 and F009 valves both remain open.
d. The F008 valve remains open and the F009 valve closes.
28. During a loss of feedwater, a manual start of the High Pressure Coolant injection (HPCI)
system was done at a water leyel of -20 inches by operator manipulation of the system
components.
Which of the following describes the HPCI system response as reactor water level continues
to change?
a. It will automatically trip at +54 inches and will automatically restart at -38 inches.
b. It requires operator action to secure injection when level is greater than +54 inches and
automatically restarts at -38 inches.
c. It requires operator actions to secure injection when level is greater than +54 inches and
to restart when level is less than -38 inches.
d. It will automatically trip at +54 inches and will require operator action to restart when
level is less than -38 inches.
Page 12 of 45
I
Rrctor Operator Examination
29. Given the following conditions:
. j
The plant is operating at 70% power .
An inadvertent initiation of HPCI has occurred , ' ,
,
, .
HPCI injection to the vessel is' occurring -
Which of the following is the required IMMEDIATE action for the given~ conditions?
a. Close the HPCI Main Pump Discharge Valve (F007) and depress the Turbine Trip
pushbutton.
b. Depress the Turbine Trip pushbutton and stop the Auxiliary Oil Pump.
c. Control. reactor water level manually to maintain level between . Level 4 and Level 7.
d. Reduce reactor power as necessary by running back Recirculation flow and inserting
control rods. .
,
4
. . . - .
30. Given the following conditions;
A loss of coolant accident has occurred
Reactor water level'is -110 inches and lowering ~
Reactor pressure is 290 psig and lowering
Which of the following is the minimum combination of the CSS Manual Initiation pushbuttons
that must be armed and depressed to place four Core Spray Pumps in service and injecting?
(Assume the manual initiation pushbuttons are operable.)
a. "A" and "B"
b. "A" and "C"
c. "C" and "D"
d. "A", "B", "C" and "D" 4
Page 13 of 45
R; actor Operator Examihation-
31. Given the following conditions:
.
LA loss of coolant accident has occurred .
'
.* Reactor. water level reached -140 inches and is currently -50 inches and rising :
, , ,
'
'
* Drywell pressure is 6 psig ,
. -
All plant systemsid;+M+1 as designed .
.
,
-
,
For the given conditions, which of the following describes the system isolation capabilities for -
. the Core Spray System (CSS) Downstream Loop injection Valve (F0058) and the CSS
Upstream Loop injection Valve (F004B), should Core Spray Loop "B" isolation be required?
a. Only F005B valve may be closed.
'
. b. Neither the F0048 or F0058. valves may be closed. ,
'
.. c. Only the F004.B valve may be closed.
.
. .
.
.
d. Both the F004B and F0058 valves may be closed.-
, .
.
. . . . . . . . .
,
~
' '
32. Given the following conditions:
* A failure-to-scram with Main Steam isolation Valve (MSIV) closure has occurred
~
.
* The pressure spike on the MSIV closure was 1120 psig
* Reactor powsr is 16% and water level is -25 inches as the 3.9 minute timer times out
Only Division 11 of the Redundant Reactivity Control System automatically initiates
* No operator actions are taken
Which of the following is the expected plant response for the given conditions.
a. Both SLC Pumps start, both Squib Valves fire and the RWCU isolation Valves (Inboard -
F001 & Outboard - F004) close.
b. The "B" SLC Pump starts, the "B" Squib Valve fires and only the RWCU Inboard
Isolation Valve (F001) closes.
c. Both SLC Pumps start, both Squib Valves fire and only the RWCU inboatd Isolation
Valve (F001) closes.
d. The "B" SLC Pump starts, the "B" Squib Valve fires and only the RWCU Outboard
Isolation Valve (F004) closes.
Page 14 of 45
R; actor Operator Examinctinn
33. Given the following conditions:
* The plant is in a failure 4o-scram conddion
. '. Standby Liquid Control (S,LC) has been initiated l>y the operator; . ,
*
* Approximately 13 minutes later the operator noted SLC S',orage' Tank level analog
.
. Indication on. Panel 10C851 is 'tr' gallons . .
* No adddional SLC system abnormalities were noted
Which of the following describes how boron in; action would be continued for the given
conditions?
a. Boron injection would continue with two SLC Pumps running.
b. Boron injection would continue with the "A" SLC Pump running.
c. Boron injection would continue with the "B" SLC Punp running. ,
'
d. Boron injection would have to be transferred to RWCU as directed by EOP-0304.'
.. . . . .
.; : .. . .. . .- - . .,
"34. Which of the following is the reason why the Reactor Protection Sys' tem (RPS) power supplies
contain Electrical Protection Assembly (EPA) breakers for specific protection against
undervoltage, overvoltage and underfrequency conditions? . -
'.
a. To maintain bus parameters during short duration power interruptions (less than 2
seconds).
b. To provide a highly reliable,' stable power supply to the RPS supplied loads, specifically
instrumentation.
c. To maintain a close tolerance power supply for the Scram Pilot Valve solenoids
proventing spurious deenergization.
d. To provide a highly reliable, stable power supply to ensure the Scram Pilot Valve
solenoids will reposition during a reactor scram.
Page 15 of 45
R:cctor Operator Examination
35. Given the following conditions:
The plant was pe', forming a startup following a refueling outage when a reactor .
. .
,., . scram occt '(all rods inserted). , , , , .
,
'
The sequerm u: vents printout shows that just prior to the scram; Average -
-
Power Rary tr.aring (APRM) channels "B" and "D" were upscale Hi-Hl .
-
.
Which of the following additional conditions, by itself, could have caused the full reactor
scram signal?
a. Rod Block Monitor Channel %" has failed.
. b. RPS Bus "B" has deenergized.
,
c. SRM Channels "A" and "C" are reading 1.5 E5 counts per second.
d. The Reactor Protection System shorting links,are removed. . .
.
.
,
36. The Nuclear Control Operator (PO) is performing backpanel checks and reports the following
. indications on the Traversing incore Protie (TIP) "A" and "B" subsystem panel (Refer to f
attached figure): ,
Squib Monitor lights - both illuminated
Shear Valve Monitor lights - both extinguished
Ball Valve "Open" lights - both extinguished
Ball Valve " Closed" light's - both illuminated
Which of 'he following is the status of the "A" and "B" TIP shear valves and primary
containment integrity?
a. The TIP Shear Valves are operable and primary containment integrity is met.
b. The TIP Shear Valves are inoperable and primary containment integrity is met.
c. The TIP Shear Valves are inoperable and primary containment integrity is not met.
d. The TIP Shear Valves are operable and primary containment integrity is not met.
Page 16 of 45
Rrctor Operator Excmination
37. Given the following conddions:
The plant is operating at 100% power . _
. .i APRM Channel V is, bypassed with the joystick '
,
, , ,,
-
Control rod 30-31 is selected *
All other plant systems are ' operating 'as designed .
-
'
Which of the following occurs if APRM Channel "F" fails full "downscale" for the 'given
conditions?
a. Rod Block Monitor Channel "B" automatically shifts to the "B" APRM as its reference,
b. Rod Block Monitor Channel "B" generates a rod withdrawal block on a failure to null.
c. Rod Block Monitor Channel"B"is indicating 0%. .-
d. Rod Block Monitor Channel"B"is bypassed on the reference APRM downscale.
.
. .
.
. . .
.. ,
. . .
.
,
38. Given the following conditions.:
Control rod insertions are in progress for scheduled plant shutdown
' Corrent reactor' power is 17%'
Intermediate Range Monitoring (IRM) Channel "A" has failed full " upscale" and
has NOT been bypassed with the joystick
Which of the following describes what will occur as the pov.er reduction continues in
accordance with .HC.OP-iO.ZZ-0004(Q), " Shutdown From Rated Power To Cold Shutdown"
and when it will occur?
a. A half scram will occur when the IRM detectors are fully inserted,
b. A control rod b, lock will occur when IRM "A" is ranged down from Range 8 to Range 7.
c. A half scram will occur when the Mode Switch is placed in Startup.
d. A control rod block will occur when the IRM detectors are fully inserted.
Page 17 of 45
s
Rrctor Operator Ex^mination
39. Given the following conditions:
The plard is performing control rod withdrawals for a reactor startup
'
.The reactor is.suberitical .,
~
Reactor power is 75 counts per second (CPS) Inthe so'u rce range .
,
The Nuclear Control Operator (RO) selects the "B" Source Range Monitoring (SRM)
detector then holds its " Drive Out" pushbutton in the depressed position
Which of the following describes the plant response?
a. The "B" SRM detector will not withdraw due to the current power level.
b. The Retract Permit light will extinguish and the SRM Det Removal Not Permitted alarm
will be received.
c. The "B" SRM detector will retract until source range indicates less than 3 cps.
d. A Control Rod Withdrawal Block will be generated. l
~.- -
~ '
40. Given the following conditions:
'
The plant is operating at 55% power .
Average Power Range Monitoring (APRM) Channel "C" currently has 14 " good *
LPRM input signals
Which of the following will result in receipt of the APRM Sys A Upscale Trip /Inop alarm (C4 on
Section C3)?
a. APRM "C" meter function switch is placed in " Flow".
b. One of the " good" LPRMs mode switch is placed in "C"(Calibrate).
c. APRM'"C" meter function switch is placed in " Average".
~d. ' One of the " good" LPRMs fails "downscale".
Page 18 of 45
. _ - _ -
Rrctor Operator Excmination -
41. With the plant operating at 85% power, steady state conditions, a narrow range water level is
readog 35".
'
Which of the following will be the indicated " level." from this instrument if the differential
' ' '
pressure acros's the det'ector fails to 'V psid for the'given conditions 1 '
..
'
-
a. O inches
b. 30 inches
c. 35 inches
d. 60 inches
42. Which of the following describes the difference in actual reactor water level versus indicated
wide range reactor water level and the expected change in that difference during a power
reduction from 100% to 65%7
' a. Actual water leDel is lower tilan indidated level and the difference will get larger during
the power reduction.
b. Actual water level is higher than indicated level and the difference will get larger during
the power reduction. .
.
c. Actual water level is lower than indicated level and the difference will get smaller during
the power reduction.
d. Actual water level is higher than indicated level and the difference will get smaller during
the power reduction.
1
-
4
I
Page 19 of 45-
o
R: actor Operator Examination
43. Given the following conditions:
The Reactor Core isolation Cooling (RCIC) is operating in Full Flow Recirc
. .. The RCIC flowcontroller is in."Automahc" . "
. , , . ,
^ ~
* RCIC turbine speed is 2450 rpm - -
,
.
.
.
Which of the following desenbes the expected response of RCIC turbine speed and system
flow if the operator throttles the RCIC Test Bypass To CST lsolation Valve (F022) in the
"open" direction for the given conditions?
(Compare the conditions after they stabilize to before the valve was throttled.)
a. - RCIC turbine speed lowers
- System flow remains unchanged
b - RCIC turbine speed lowers
- System flow goes down .
'
**
'
cc - RCIC'tuabine speed raises- -
- System flow remains unchanged
d. - RCIC turbine speed raises
- System flow goes up. .
44. Given the following conditions:
A loss of all AC power has occurred
No Diesel Generators are running
The Reactor Core Isolation Cooling (RCIC) system has initiated and is injecting
A valid RCIC steam line high flow signal is received
Which of the following describes the RCIC inboard and Outboard Steam Supply isolation - l
Valves (F007 & F008) isolation capabilities and the response of the RCIC turbine for'the
given conditions?
a. The F007 and F008 valves remain open but can be closed from the Control Room.
b. The F007 and F008 valves remain open and cannot be closed from the Control Room.
c. Only the F007 valve closes.
d. Only the F008 valve closes.
Page 20-of 45
9
__
R:: actor Operator Examination
45. Given the following conditions:
* The Automatic Depressurization System (ADS) Manual Initiation Channel"B"
and "E" pushbuttons (S6B and S6F) have been armed.and depr,essed
, ,. .,
'
There is no Safety Relief Valve respc.5e
'
Which of the following "B" Division electrical bus failures caused this system response?
a. A loss of 120 VAC Bus 1BJ481
b. A loss of 250 VDC Bus 10D261
c. A loss of 125 VDC Bus 1BD417
d. A loss of 480 VAC Bus 10B420
46. Given the following conditions:
.. . ..
-. . . . . . .
The plant has been operating at 100% power for several weeks i
All systems are operating 'as designed
Which of the following is the reason why periodic nitrogen makeup to the drywell is required
for the given conditions?
a. Due to leaks from drywell air operated equipment.
b. Due to PCIG normal system leakage,
c. Due to normal periodic cycling of the Torus - Drywell Vacuum Breakers. l
t
d. Due to normal drywell air inleakage.
.
l
l
i
Page 21 of 45
I
R:actnr Operator Excmintti n I
47 Given the following conditions:
* The plant had been operating at 75% power .
.
A loss.of main condenser vacuum caused a complete Main, Steam isolation
'
,
Valve (MSIV) closure .
.
Vacuum has been reestablished and is currently.15" Hg absolute - -
-
,
Which of the following conditions is REQUIRED in order to reset the NSSSS MSIV isolation -
logic?
a. The Reactor Mode Switch must be in " Shutdown". 1
b. The Main Condenser Low Vacuum Bypass Switches must be in " Bypass".
c. The MStV control switches must be in "Close".
d. The Turbine Stop Valves must be closed.
-
- -
.
.. . , . ,
48.' Wh.ich of the following conditions would preven.t. opening the RHR "B" Loop. Inboard and
.
Outboard Drywell Spray Valves (F0218 and F0168) following a LOCA?
a. The LPCl injection Valve (F.0178) is not fully close'd. .
]
b. Less than 5 minutes have elapsed since the "B" RHR initiation occurred.
c. The RHR Full Flow Test Valve (F024B) is not fully closed.
d. Reactor water level is above -129 inches.
49. Given the following conditions:
The Fuel Pool Cooling system is operating with one pump' and heat exchanger
in service
The Fuel Pool Gates are installed i
. No makeup water sources are available
Which of the following is the expected effect on Spent Fuel Pool water level and cooling
capability if a leak develops on the common FPCC Pump Suction?
1
a. Cooling capability and water level will be unchanged.
b. Cooling capability will be lost and water level will lower slightly and stabilize.
c. Cooling capability will be unchanged and water level will lower slightly and stabilice.
'I
d. Cooling capability will be lost and water level will continuously lower.
.
I
' Page 22 of 45 -l
b -
_____
l
Rocct:r Op2ratar Examination
50. Which of the following describes how the main steam line flow restrictors assist in maintaining
adequate core cooling for steam line break between the flow restrictors and the Main Steam
Isolation Valves?
a. They ensure the total ~ inventory loss from the reactor. vessel maintains level-above.the
top of active fuel until one division of low pressure ECCS is injecting. ,
b. They limit the total inventory loss from the reactor vessel to maintain water level above
the top of active fuel for a minimum of 5 seconds.
c. They ensure the total energy release rate to the Primary Containment does not result in
exceeding suppression chamber design pressure.
d. They limit the total inventory loss from the reactor vessel to maintain level above the top
of active fuel until HPCI is at rated flow.
51. Given the following conditions:
A reactor scram and Main Steam Isolation Valve (MSIV) closure from 90% power
has occurred
The Safety Relief Valves (SRVs) are cycling to control pressure
Which of the following primary containment parameters indicates that one of the SRV tailpipe j
vacuum breakers has failed open?
a. Suppression chamber pressure will go up each time the SRV cycles.
b. Suppression pool water temperatures will show rapid localized rises from the SRV
discharge flow bypassing the T-quenchers.
c. Drywell pressure will go up each time the SRV cycles.
d. The Torus to Drywell differential pressure will rise each time the SRV opens.
l
52. Which of the following plant systems must be in operation to support the Main Steam
isolation Valve (MSIV) Seal System.
a. Primary Containment Instrument Gas (PClG)
b.125 VDC Electrical Distribution
c. NUMAC Leak Detection System
d. Process Radiation Monitoring System
Page 23 of 45
/
R: actor Operator Examinati:n.-
53. Given the following conditions:
i
* The plant is operating at 70% power.
.'
. The "B" EHC. Pressure Regulator is tagged out of service' . . '
'
'
'
Unknown to thei operator, the "A" EHC Pressure Reg 61ator output signal is .
failed "as is"
Which of the following would be the expected response of the Turbine Control Valves and
Turbine Bypass Valves as the Nuclear Control Operator (RO) begins to raise power using -
recirculation flow for the given conditions? (Figure attached)
a. - The Turbine Control Valves will close
- The Turbine Bypass Valves will open
b. - The Turbine Control Valves will close
. - The Turbine Bypass Valves will not move
c. - The Turbine Control Valves will not move
-
- The Turbine Bypass valve will not move -
~ d. - The Turbine Control Valves will not move .
- The Turbine Bypass Valves will open
54. Due to a main turbine vibration problem with a generator load of 110 MWe, a successful
manual turbine trip is performed.
Which of the following describes when the operator is REQUIRED to open the generator
Output Breakers for the given conditions? (Assume they have not already tripped on reverse
power.)
a. Immediately
~b. Within 15 seconds of the turbine trip
c. Within 60 seconds of the turbine trip
d. Within 90 seconds of the turbine trip
,
i
Page 24'of 45
i
_ _ .
. - - - . - _ _ _ _ _ _ - _ _ _ _ _ -
- _--- - _ - _ - . _.
R: actor Operator Exeminati::n.
r I
55. Given the following initial conditions:
+ The plant is operating at 25% power performing a plant startup
* All plant systerns are oper@ as designed _
._ .
. The "A" Reactor Feedwater Pump is in service in auto at approximately 3850 rpm
. .
Following a plant transient the following conditions exist: 4
The reactor failed to scram when required
Reactor power is 14% and reactor pressure is 1105 psig
The Nuclear Control Operator (RO) notes that the "A" RFP speed has slowed
to less than 1000 rpm
. The RFP TURBINE AUTO XFR TO MANUAL (B3-F3) annunciator is in alarm
Which of the following describes the recson for the "A" RFP speed reduction?
,
a. The "A" RFP is responding properly to a Redundant Reactivity Control . System runback.
i
b.- The "A" RFP is responding to the S~etpoint Setdown feature of Digital Feedwater Contrcl
l calling for a lower level.
c. The "A" RFP is responding to a Control Signal Failure
d. The "A" RFP is responding to a loss of one Primary Condensate Pump and one
Secondary Condensate Pump.
I
56. Given the following conditions:
. A loss of off-site power (LOP) has occurred from 75% power
. Within 10 seconds a loss of coolant accident (LOCA) occurs
Which of the following is the expected response of the LOF and LOCA sequencers? 3
a. As soon as power is restored to the buses, the LOCA sequencer will control the
restoration of allloads.
b. The LOCA sequencer will begin to sequence until the diesel generator output breakers
close, then the LOP sequencer will complete load restoration.
c. As soon as power is restored the buses, the LOP sequencer will control the restoration l
of allloads.
d. The LOP sequencer will begin io sequence until the diesel generator output breakers
close, then the LOCA sequences will completa load restoration. i
i
Page 25 of 45 ]
_ _ _ _ _ _ _ _ _ _ - _ _ -
_
R: actor Operator Excminati:n
57. Given the following conditions:
* The "B" Emergency Diesel Generator (EDG) had started following a valid
LQCA signal '
Some tirhe later the' EDG was shutdown using the locdl Emergency Stop pushbuttons
due to fluctuating oil pressure
Concurrent with stopping the EDG, the 10A402 bus lost power
Which of the following describes the actions, if any, regardi1g resetting the Engine Shutdown
Relay (ESR) and the (86R) Lockout Relay to restart the "B" EDG and reenergize the 10A402
bus?
a. ESR must be reset
(86R) Lockout Relay reset is not required
b. ESR must be reset
(86R) Lockout Relay must be reset
c. ESR reset is not required' -
-
(86R) Lockout Relay reset is not required
d. ESR reset is not required
.(86R) Lockout Relay must be reset
58. Which of the following parameter changes indicate the moisture content of charcoal adsorber
bed of the Gaseous Radwaste System (GRW)is rising?
a. GRW post-treatment radiation level due to Krypton is rising.
'
b. GRW charcoal adsorber bed temperature is lowering.
I
c. GRW post-treatment radiation level due to lodine is rising.
d. GRW charcoal adsorber bed hydrogen concentration is lowering.
l
l
1
:
Page 26 of 45
I
l
- - - - -
- -
-
-
- . . _
_______ ______j
R::ctor Operator Excminatinn
50. Given the following conditions:
The plant has been operating at 100% power for several weeks
Main Steam Line (MSL) radiation levels have been averaging 80 mrem but are now ,
-
slowly trending upwards
Chemistry has verified the higher radiation readings are due to failed fuel
What are the immediate Operator Actions required for the given conditions?
a. Place additional Condensate Demineralizers in service if possible.
b. Scram the reactor and close the Main Steam isolation Valves when MSL levels are
greater than 120 mrem.
c. Direct Reactor Water Cleanup flow to the main condenser to reduce coolant activity.
d. Reduce reactor power to maintain MSL radiation levels less than 120 mrem.
. . ,
60. Which of the following is the basis for raising the Main Steam Line (MSL) radiation monitor
setpoints when the Hydrogen Water Chemistry injection (HWCl) system is placed in service?
a. The setpoint adjustment ensures the higher (approximately two times) background
radiation does not mask a true fuel element failure.
b. The setpoint adjustment prevents unnecessary alarms / protective actions from the higher
(approximately two times) background radiation.
'
c. 's ne setpoint adjustment ensures the higher (approximately ten times) background .
radiation does not mask a true fuel element failure.
d. The setpo;# adjustment prevents unnecessary alarms / protective actions from the higher
(approximately ten times) background radiation.
l
l
'
,
!
I
Page 27 of 45
!
-
Rocct2r Operator Examinatian
61. Given the following conditions:
A valid EDG room high temperature condition has just occurred
The Diesel Generator Room Carbon Dioxide Fire protection system is aligned
~
~ fo'r* automatic operation
Which of the following describes how the Diesel Generator Room Carbon Dioxide Fire
protection system responds?
a. A discharge alarm occurs, CO2 with a wintergreen scent is discharged into the room
immediately.
b. A pre-discharge alarm is activated and a wintergreen scent is discharged into the room.
After a time delay, CO2 is discharged into the room.
c. A pre-discharge alarm is activated. No CO2 is discharged into the room until a valid
smoke detector alarm is received.
', d. A pre-discharge alarm is activated. After a time delay CO2 with a wintergreen scent is
' ~
discharged into the room.
62. Given the following conditions:
The plant is operating at 50% power
All systems are operating normally
One Reactor Building Ventilation System (RBVS) Exhaust Fan discharge damper
has failed to the full "open" position with the fan running
No other RBVS components have changed
-
Which of the following describes how this will affect the initiation of the Emergency Core
Cooling Systems (ECCS) and the reason for this?
a. ECCS will initiate after it is required because the failed damper lowers Reactor Building
pressure resulting in a lower indicated drywell pressure,
~
b. ECCS will initiate before it is required because the Siled damper raises Reactor
~ Building pressure resulting in a higher indicated c.ywell pressure.
c. ECCS will initiate after it is required because ti.s failed damper raises Reactor Building
pressure resulting in a lower indicated drywell pressure.
d. ECCS will initiate before it is required because the failed damn lowers Reactor
Building pressure resulting in a higher indicated drywell pressure.
' Page 28 of 45
_
RIcctor Operator Ex;minati:n
63. Given the followng conditions:
The plant is operating at 40% power
;The Jet Pump operability surveillance indicates that one jet pump has failed .
Technical Specifications require the plant to be in hot shutdown within 12 hours
Wnich of the following describes why such a severe restriction placed on continued operation -
for the given conditions?
a. A jet pump failure at this low power level will significantly affect the core flows and result
in unacceptable therrn<l limits (MCPR).
b. A jet pump failure may limit reactor water level restoration capability during the reflood
portion of a Loss Of Coolant Accident.
c. A jet pump failure combined with the flow restricting orifices may adversely affect core
flow to the higher power fuel bundles.
d. A jet pump failure results in less conservative protective action setpoints for
'
instrumentation using recirculation loop flow as an input sign'al.' '
64. Which of the following is the expected status of the Control Area Ventilation after a valid high
radiation condition at the Control Area Ventilation air intake occurs?
The Control Room Emergency Filtration (CREF) units are processing:
a. air enteing the control room as well as recirculated air and are maintaining a slight
negative pressure.
b. air entering the control room as well as recirculated air and are maintaining a slight
positive pressure.
c. only the current control room atmosphere and are maintaining a slight negative pressure.
d. only the current control room atmosphere and are maintaining a slight positive pressure.
Page 29 of 45
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Reactor Operator Examinatinn
65. :.Given the following conditions:
The "A" Reciraalation Pump has tripped
.
,
.
The "A" Recirculation Pump discharge valve is open - .
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* RECIRC LOOP A JET PUMP FLOW (TOTAL) iridicates 2 mlbm/hr
RECIRC LOOP B JET PUMP FLOW (TOTAL) indicates 35 mlbm/hr
RECIRC PMP B FLOW indicates 24,000 gpm
Recirc pump "B" speed is 49%
Which of the following would be expected values for total JET PUMP FLOW (the flow
recorder) and actual core flow?
a. Flow recorder - 33 mlbm/hr, Actual core flow - 33 mlbm/hr
b. Flow recorder - 33 mlbm/hr, Actual core flow - 37 mlbm/hr -
c. Flow recorder .37 mlbm/hr, Actual core flow - 33 mlbm/hr
- d. Flow recorder - 37 mlbm/hr, Actual pore flow - 37 mlbm/hr .
66. Given the following conditions:
* The plant is operating at 90% power
All main turbine sealing steam normal and backup supplies have beer lost
There is no time estimate for repair / restoration
Which of the following are the immediate operator actions for the given conditions?
a. Reduce power as necessary to maintain condenser pressure less than 5.0". HGA.
b. Reduce recirculation flow'to minimum, unload and trip the main turbine,
c. Reduce power as necessary to maintain adequate self-sealing steam to the main'tutt>ine !
seals.
d. Reduce recirculation flow to maintain power less than 25% (Bypass Valve capacity).
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Page 30 of 45
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Reactor Operator Examinr_ tion
67. During a loss of off-site power the operator is cautioned not to acknowledge the flashing
' Trip" pushbuttons for the 4.16 KV Vital 1E Bus infeed breakers. I
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.Which of the following will occur if these pushbuttons are pressed?
a. That bus' feeder breaker will attempt to close until the anti-pump feature causes it to trip
open and remain open.
b. The Diesel Generator associated with that bus, if running, will trip and its output breaker
will open.
c. That bus' alternate feeder breaker will trip open and then immediately reclose when the
pushbutton is released
d. The Diesel Generator associated with that bus will not load.
68. Given the following conditions:
A plant startup is in progress with the Reactor Mode Switch in "Run"
The Main Turbine is reset and is at 950 rpm accelerating to 1800 rpm
A loss of 125 VDC power from distribution panel 1CD318 to the EHC control
logic occurs
Which of the following is the expected plant response?
a. Main turbine trips.
b. Main turbine startup would continue at the selected acceleration rate.
,
c. Main turbine speed will remain constant at 950 rpm.
d. Main turbine control valves throttle closed due to a loss of the speed reference signal.
Page 31 of 45
Re:ctor Operator Examinati:n
69. Given the following conditions:
* The plant is operating at 20% power
* A main generator load reject has just occurred . . .
The power / load ui1 balance circuit tripped unexpectedly during the load reject
Which of the following is the expected response of the Turbine Control Valves and the
Reactor Protection System (RPS) for the given conditions?
a. - The Turbine Control Valves throttle closed
- RPS does not trip
b. - The Turbine Control Valves fast close
- RPS trips
c. - The Turbine Control Valves throttle closed
- RPS trips
d. - The Turbine Control Valves fast close .
- RPS does not trip
70. Which of the following describes when the Main Turbine is required to be tripped following a
reactor scram?
a. At 50 MWe lowering
b.- At 25 MWe lowering
c. At 0 MWe
d. At 50 MWe rising (reverse power)
71. During a failure-to-scram condition, which of the following is the criteria used to determine if
HC.OP-EO.ZZ-0100(O), " Reactor Scram", should be exited and HC.OP-EO.ZZ-0101(Q),
" Reactor / Pressure Vessel (RPV) Control", entered?
a. Reactor period on SRM Period meters is stable at -80 seconds
b. All APRM "downscale" lights are not illuminated
c. All four RPS logic channels are deenergized
d. All control rods are inserted to or beyond Notch "02"
Page 32 of 45
_ _ _
Rcactor Operat r Examinati::n
72. Following a reactor scram and Main Steam isolation Valve closure, reactor pressure reaches
1050 psig causing the "H" and "P" Safety Relief Valves (SRV) to open.
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Which of the following lists the operating setpoints for subsequent openings of the "P." SRV? !
a. SRV "P" opens at 1047 psig and closes at 935 psig. I
b. SRV "P" opens at 1047 psig and closes at 905 psig.
c. SRV "P" opens at 1017 psig and closes at 935 psig.
d. SRV "P" opens at 1017 psig and closes at 905 psig.
73. With the plant at 100% power a severe overfeeding transient is occurring. Water level is +50
inches and rising rapidly.
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Which of the following reactor water levels require termination of all feed to the reactor,
closing the MSIVs and a reactor scram assuming none of these actions have occurred?
a. +54 inches
b. 465 inches
c. +90 inches
d. +118 inches
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74. Given the following conditions: ,
The plant is operating at 80% power
All three Feedwater Pumps are in service
Feedwater Level Control is in " Automatic - Three Element" control
Narrow Range level "A" is reading 34 inches
Narrow Range level "B"is reading 36.5 inches
Narrow Range level "C" is reading 35.0 inches
Which of the following would be the expected response of the Feed Water Level Control i
System and reactor water level if Narrow Range level "B" failed to the low end of the range?
a. It would transfer to Single Element Control and level would remain unchanged.
b. It would remain in Three Element Control and level would remain unchanged. l
c. It would transfer to Single Element Control and would raise level by approximately 1.5
inches.
d. It would remain in Three Element Control and would raise level by approximately 1.0
inches.
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Page 33 of 45
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1
R^ ctor Operator Examinati:n
75. Given the following conditions:
The plant is operating at 95% power
All Drywell Cooling Chilled Water pumps have tripped '
Drywell pressure is rising -
HC.OP-AB.ZZ-0201, "Drywell High Pressure / Loss Of Drywell Cooling", has been
entered and the Reactor Auxiliary Cooling System (RACS) is being aligned to supply
backup cooling to the Chilled Water System
Which of the following describes the effect of failing to close the Chilled Water Isolation
Supply and Return Valves (HV9532-2 and HV9532-1) before the transfer to RACS?
a. The RACS Pump automatic start permissives will be bypassed until tha valves are
closed.
b. The RACS valves will not automatically sequence open to supply Chilled Water should
a loss of off-site power occur,
c. Chilled Water system flow will divert back into the RACS system overflowing the RACS
head tank.
d. RACS system flow will divert back into the Chilled Water system overflowing the Chilled
Water head tank.
76. During a loss of coolant accident the following conditions exist
Reactor pressure is 125 psig
Drywell temperature is 325 'F
Which of the foilowing describes the accuracy and i ding capabilities of wide range reactor
water level indication for the given conditions? l
a. They are not providing accurate re or water level or level trend information.
b. They are providing accurete ctor water level but level trend is not roliable.
c. They are providing a te reactor water level and level trend information.
d. They are not prov' ng accurate reactor water (evel but level trend is reliable.
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Page 34 of 45
( R: actor Operator ExSminati:n
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77. Given the following conditions:
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. The plant is operating at 95% power
. Suppression pool teirpeinture is 92 *F
. At 0915, Safety Relief Valve (SRV) "G" opened
.
'After several cycles of the SRV Open and Close pushbuttons, the operator notes
that tailpipe temperature for the SRV is stable at 305 *F and NO other plant parameters
have changed
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Which of the following describes the limitations on continued reactor operation for the given
conditions?
a. Reactor operation may corninue until pressure set is reduced to less than 850 psig.
b. Reactor operation may continue until suppression pool temperature reaches 120 *F.
c. Reador operation may continue indefinitely. ;
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d. Reactor operation may continue until 0917.
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78. Given the following conditions:
Reactor power is 82%
HPCI is in operation for a surveillance
The "B" loop of RHR is in Suppression Pool Cooling
Suppression pool temperature is 103 *F when the runni R pump tripped
HPCI was secured
* Subsequently, suppression pool temperature in sed to 106 *F
Which of the following lists the suppression temperatures requiring entry into HC.OP-
EO.ZZ 0102, Primary Containment Cont AND entry into the LCO actions for Tech Spec )
3.6.2.17
a. EO4102 - 95 *F
TS 3.6.2.1 - 95 *
5 *F
b. i00 0102 7
13.6.0 P - 105 *F
2 1% * d{VN
ghU
c. t ,'1, f T
~
;'A v >2.1 -Y,
y
d. EO4102. - 105 *F
TS 3.6.2.1 - 105 *F gg
Delet e6 See Ard emm 'YS
Aft 3 -5-1
Page 35 of 45
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Reactor Operatcr Ex0mination
79. Given the following conditions:
The plant is at 75% power
* Control rod 22-27 is being withdrawn one notch to Notch "22"
Which of the following is the required immediate operator action if a control rod drift alarm is
received and the operator notes control rod 22-27 is continuing to move out and power is
rising?
a. Apply a continuous insert signal to control rod 22-27.
b. Place the Rod Select key lock switch to "Off" (de-select the rod).
c. Direct the local operator to perform a single rod scram on control rod 22-27.
d. Runback recirculation flow and insert control rods to reduce power.
80. Given the following conditions:
The plant is operating at 100% power
. A feedwater heater trip has resulted in a feedwater temperature of 385 'F
No operator actions have been taken
Vdiich of the following is the operational concern for the given conditions?
a. Entry into the Exit Region of the Power-To-Flow Map.
b. Violation of the Hope Creek Operating License, j
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c. Immediate thermal hydraulic instabilities.
d. Recirculation Pump damage.
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Page 36 of 45
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R: actor Operatar Excminatirn
81. Following a reactor scram all rods are at position "00" except one that is at position "24."
Which of the following describes the capability of the reactor to remain shutdown?
a. Control rods are inserted to beyond the Maximum Subcritical Banked Withdrawal limit,
therefore the reactor will remain shutdown under all conditions.
b. Control rods are not inserted to beyond the Maximum Suberitical Banked Withdrawat
limit, therefore it cannot be assured the reactor will remain shutdown under all
conditions.
c. Design basis shutdown margin is met, therefore the reactor will remain shutdown under
all conditions.
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d. Design basis shutdown margin is not met, therefore it cannot be assured that the reactor
will remain shutdown under al! conditions.
82. Given the following conditions:
The Control Room has been abandoned in accordance with HC.OP-AB.ZZ-0130(Q),
" Control Room Evacuation"
Control has been established at the Remote Shutdown Panel in accordance with
HC.OP-lO.ZZ-0008(O), " Shutdown From Outside Control Room"
RCIC is operating maintaining reactor water level at +35 inches
Safety Relief Valves (SRV) are being used to cooldown
Condensate Storage Tank (CST) level is 135,000 gallons
The Condensate System is not available
Which of the following is correct for the given conditions?
e RCIC is operated without overspeed protection.
b. Insufficient CST inventory is available '.o allow the cooldown to clear the shutdown
cooling interlocks.
c. The RCIC Gland Seal Condenser Condonsate Pump must be manually operated.
d. SRVs cannot be operated in a rotation that will evenly distribute heat to the 'Juppression
Chamber.
Page 37 of 45
R: actor Operator Ex2 min:ti:n
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83. Which of the following describes the effect of failing to restart the Turbine Building Ventilation
System if it trips while operating in HC.OP-EO.ZZ-0104(Q)-FC, " Radioactive Release
Control"?
a. The Turbine Building will go to a slightly negative pressure.
b. The total off-site release calculations will not be accurate.
c. The Turbine Building releases will be monitored but not treated.
d. The total off-site release will be higher.
84. A loss of Reat., tor Auxiliary Cooling System (RACS) has occurred.
Which of the fol',owing is the MAXIMUM time allowed before a reactor scram is required?
a. An immediate scram is required
5. One (1' minute
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c. Ten (19) minutes
d. Twergy (20) minutes
85. Given the following conditions:
A loss of coolant accident has occurred
The Reactor Auxiliaries Cooling System (RACS) has been restored
Which of the following describes the availability / response of the Emergency Instrument Air
Compressor (EIAC) for these conditions should instrument air header pressure begin
lowering?
a. The EIAC is not available until the LOCA signal is cleared, reset and the 1E breaker is
closed.
b. The EIAC will automatically start on instrument air header pressure less than 85 psig.
c. The EIAC is not available until the Non-1E breaker is closed and instrument air pressure
is less than 85 psig. l
d. The EIAC will not automatically start but may be started manually from the Control
Room or locally. ;
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Page 38 of 45
R: actor Op:: rat r Examinttisn
86. Which of the following describes the reason control rods insert during a loss of instrument air?
a. A flowpath is opened to the bottom of the drive mechanism operating piston allowing
reactor pressure to drift the rod in.
b. Drive water flow and pressure both rise enough to drift the control rod in, similar to a
normal insertion.
c. A flowpath is opened from the top of the drive mechanism operating piston allowing
accumulator pressure to drift the rod in.
d. The normal scram flowpath to and from the drive mechanism operating piston is opened,
allowing accumulator and reactor pressure to drift the rod in.
87. Given the following conditions:
The plant is operating at 20% power following a refueling outage
* An error during a surveillance has resulted in a Group 10 (Drywell Chilled Water)
isolation signal
The isolation goes to completion (all valves are closed)
Drywell pressure is 1.25 psig and rising slowly
Which of the following are the required immediate operator actions for the given conditions?
a. Lineup and commence venting the drywell.
b. Secure drywellinerting.
c. Place the Reactor Mode Switch in " Shutdown".
d. Align RACS to supply cooling to Drywell Chilled Water.
88. Following a loss of shutdown cooling, decay heat removal is being transferred to the
Alternate Shutdown Cooling lineup (Injection to the reactor returning to the suppression pool
via open Safety Relief Valves).
Which of the following indications is utilized to monitor the margin to Op Con 3 entry in this
lineup?
a. Safety Relief Valve tailpipe temperatures
b. Suppression pool temperatures
c. Reactor vessel skin temperatures
d. Local suction temperatures on the running low pressure ECCS pumps
Page 39 of 45
R: actor Operator Excminatinn
89. Which of the following describes the conditions requiring the Reactor Mode Switch to be
placed in " Shutdown" on a sustained loss of Control Rod Drive charging water header
pressure (<900 psig) with reactor pressure at 650 psig?
a. Within 20 minutes of determining more than one CRD accumulator is inoperable and at
least one of those inoperable accumulators is associated with a withdrawn control rod.
b. Within 20 minutes of determining any CRD accumulator is inoperable and the
inoperable accumulator is associated with a withdrawn control rod.
c. Immediately upon determining more than one CRD accumulator is inoperable and all the
inopersle accumulators are associated with fully inserted control rods.
d. Immedir Aly upon determining any CRD accumulator is inoperable and the inoperable
accur.dator is associated with a withdrawn control rod.
90. Given the following conditions:
The plant is shutdown for refueling
* The Reactor Protection System shorting links have been removed
A fuel bundle is being moved from the fuel pool to core.
if SRM "C" fails "downscale", which of the following are the required immediate actions?
(
a. Verify a control rod withdrawal block is received. Terminate fuel movement.
b. Verify a full scram and a control rod withdrawal biock is received. Terminate fuel
movement.
c. Verify a control rod withdrawal block is received. Fuel movement is required to be
terminated ONLY if the fuel bundle is to be placed in the quadrant monitored by SRM
"C ."
d. Verify a full scram and control rod withdrawal block is received. Fuel movement is
required to be terminated ONLY if the fuel bundle is to be placed in the quadrant
monitored by SRM "C."
Page 40 of 45
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R actor Operator Examination
91. Given the following conditions:
* A large break loss of coolant accident has occurred
* Drywell pressure reached a maximum of 22 psig
* Suppression chamber sprays have NOT been placed in service -
* Drywell sprays are in servk:e
* Drywell pressure is 4 psig and slowly lowering
Which of the following is the sqeded positions of the Torus-to-Drywell Vacuum Breakers and
the Reactor Building 4o-Tonas Vacuum Breakers for the given conditions?
a. - The Torus-to-Drywell Vacuum Breakers are open-
- The Reecte Building-to-Torus Vacuum Breakers are open
b. - The Torus-to-Drpil Vacuum Breakers are open
- The Reactor Building-to-Torus Vacuum Breakers are closed
c. - The Torus-to-Drywell Vacuum Breakers are closed
- The Reactor Building-to-Torus Vacuum Breakers are closed
d. - The Torus-to-Drywell Vacuum Breakers are closed
- The Reactor Building-to-Torus Vacuum Breakers are open
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92. Following a reactor scram with a Main Steam isolation Valve Closure, the plant is being
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depressurized using the Safety Relief Valves (SRV).
Which of the following is the reason why the depressurization should be accomplished with
" sustained" SRV openings if the pneumatic supply (PClG and instrument air) is lost to the
SRVs?
a. This prevents exceeding the 100*F/hout cooldown limit during the depressurization
while conserving the SRV pneumatic supply.
b. This ensures the available SRV pneumatic supply is sufficient to cooldown to less than
the shutdown cooling interlocks.
c. This directs depressurizetion without regard to the Technical Specification cooldown
limits before the depleted pneumatic supply results in loss of SRV control.
,
d. This ensures the SRV accumulator pneumatic supply is available and adequate for later
use if the Emergency Operating Procedures require Emergency Depressurization.
Page 41 of 45
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Reactor Operator Examination
93. HPCI and RCIC both started anf are injecting in response to a valid low reactor water level.
Current plant conditions are as Mtws:
Reactor water level is +25 inches, steady
1
Reactor pressure is 845 psig, rising slowly
Drywell pressure is 1.1 psig, steady
RCIC has been aligned to Full Flow Recirc operation (CST to CST) for pressure control
HPCI is injectir. ) to the reactor for level control
After 10 minutes of operation a valid high suppression pool level is received
Which of the following would be the expected response of RCIC if a valid high suppression
pool level is received for the given conditions?
a. RCIC will remain in Full Flow Recirculation.
b. RCIC will trip on high turbine exhaust pressure.
c. RCIC will trip on low suction pressure.
d. RCIC will operate on minimum flow.
94. During high primary containment water level conditions, suppression pool water level
indications cannot be used.
Operation of which system will invalidate the alternate method used for determining primary
containment water level?
a. RCIC
b. Core Spray
c. RHR
d. HPCI
Page 42 of 45
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1
R: actor Operator Examination R
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95. Given the following conditions:
A leak has occurred in the suppression pool '
. The reactor is shutdown
. A cooldown is being performed using SRVs
The Heat Capacity Level Limit (HCLL) curve is being monitored
The " Action Required" area of the HCLL curve has been entered for several minutes -
Which of the following is a possible effect of initiating an emergency depressurization with the
given conditions?
a. The suppression pool may exceed design temperature.
b. Failure of the downcomer vent headerjoints due to " chugging."
c. The SRV Tailpipe Level Limit curve may be exceeded.
d. The capacity of the Torus to Drywell vacuum breakers will be exceeded.
96. Following the runback of the Recirculation Pumps on a trip of a Primary Condensate Pump,
the operator may monitor the Source Range Monitoring (SRM) period meters for strong
deflections above and below " Infinity".
Under which of the following conditions may SRM period indications be considerad accurate
indication of thermal hydraulic instabilities?
a. Only when the SRM detectors are fully withdrawri from the core.
b. Anytime, regardless of detector position, if the detectors are stationary.
c. Only when the SRM detectors are fully inserted into the core.
d. Anytime the SRM detectors are moving.
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Page 43 of 45
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R ccter Opercter Ex mination
97. With the plant at power the Main Steaml Reactor Water Cleanup Area Leak Temperature
High alarm was received and the RWCU system automatically isolated. The leak has been
determined to be in the RWCU Pipe Chase Room 4402.
Which of the following is NOT a required operator action for the given conditions?
a. Notify Chemistry to close the Manual Sample Line Isolation Valves P-RC-V9670 and 1-
RC-V006.
b. Observing the RWCU inboard and Outboard Isolation Valves (F001 & F004) close.
c. Observing the Recirc Sample Line isolation Valves (BB-SV.4310 and 4311)
automatically close,
d. Operate available Reactor Building ventilation fans consistent with plant conditions.
98. Given the following conditions:
. The plant was operating at rated power when a steam line break occurred in the HPCI
room
. HPCI isolated due to high room temperatures
. RBVS exhaust radiation levels reached 1.0 E-2 microcuries/ml
Which of the following describes the ventilation system response for the given conditions?
a. RBVS remains in service
b. RBVS isolated,6 FRVS Recirc and 1 FRVS Vent Fans are in service
c. RBVS isolated,4 FRVS Recirc and 1 FRVS Vent Fans are in service
d. RBVS isolated,6 FRVS Recire and 2 FRVS Vent Fans are in service ,
99. With the Reactor Building Ventilation in a normal lineup, the operator observes that Reactor
Building pressure is .10 inches of vacuum water gauge.
Which of the following is an immediate action to restore Reactor Building pressure to the
required pressure?
a. Place at least two FRVS units in service.
b. Secure a reactor building supply fan. -
)
c. Place an FRVS unit in service and increase FRVS flow rate to maximum. j
d. Place the third Reactor Building Exhaust Fan in service. :
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Page 44 of 45 ;
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), 7
R actor Operator Examinetinn _
100. Given the following conditions:
The reactor has scrammed from power
- Pilot Screm Valve Group 1 solenoids, Logic "A" and Logic "B" dici not doenergize
The Scre.m Discharge Volume is currently full
Which of the following describes the difference between inserting control rods in accordance
with HC.OP-EO.ZZ4303, " Individual Control Rod Scrams" and HC.OP-EO.ZZ4302, "De-
energization Of Scram Solenoids"?
a. EO-0302 requires resetting RPS and ARI, EO-0303 does not.
b. EO-0303 requires resetting RPS and ARI, EO-0302 does not.
c. EO-0303 does not isolate the Scram Discharge Volume, EO4302 does.
d. EO4302 results in a flowpath from the RPV to the Reactor Building Sump, EO4303
does not.
Page 45 of 45
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Reactor Operator Answer Key
1. b ' 2s4=10101 26. d mosoooAio4 2. a. 2samtato2 27. pa. 20soooA2m
ses m u a e ~ l s'.
t s
3. c 204micios 28. a m io2 i k 3* M 'tjftg Q [ g
i. r n ;o;;; 29. d 20enono1
oeseres see Arw-u e n~~i s:s r ne s-s >s
5. c 2s a iot22 M b kW 30. d rocootA302 6. c 2s4001o12e 31. a 20eootA m 7. b 2samtatat 32 a 25toooA20s 8. b 2s4001o202 33. a 2itoooksoe 9. c 2s4ooto213 34. d 2120ooAios
10. b 29400103o4 35. d 2120 m K4tt 11. c 2s40010310 36. b 21smiA4o3 12. b 2s4001c4 2 37. d 21soo2xao4 13. d 2ndootG44e 38. c 21sooskoos 14. c 20tm1A204 39. d 21sm4Ato4 15. a 201m1K405 40. b 21Soo6K1o4 16. c 20im2A435 41. d 2160 m A201 17. a 20ioo3A207 42. d 2teo m Asot 18. a 201oosks14 43. a 2170mA401 19, p 5 =@u,,s 95 m
DW p3[tt it$ 44. b 217000K201
_m--, .. 20. a orJ 202001Aso2 45. c 218moK201
r seenrraa202mixtoe
21. c-
m~., w se (6rMD J-s-es~h 46. b 223ooixtos
22. b 202o02Atoi 47. c 223oo2A403 23. b 202m2xso4 48. a 22ecoix4o3 24. d 2MoooK406 49. b 2330m o o2 25. c 2odom Kit 5 50, b 239001o12s
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ATTACHMENT 3
PSE&G COMMENTS ON WRITTEN EXAM
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Public Sendoe Bectic and Gas Company 244 Chestnut Street Salem, NJ. 08079 Phone 609/935-8560
Nuclear Training Center
March 6,1998
NTC-98-3011
Mr. Don Florek
Chief Examiner
Division of Reactor Safety
US Nuclear Regulatory Commission
475 Allendale Road
King of Prussia, PA. 19406-1415
Dear Mr. Florek
HOPE CREEK SRO/RO EXAM COMMENTS
Attached please find our post-examination analysis comments and related backup information on
the following questions, from our recently conducted Hope Creek RO/SRO examination, Our
comments are on the page with the applicable question, and are broken into three (3) categories:
Exam Answer key corrections
. RO #19
. RO #27 / SRO #29
. RO #98 / SRO #96
Correct Alternate choice answers from oriainal answer kew
. RO #20 / SRO #24
. SRO #69
. RO #76 / SRO #71
. RO #79
. RO #84 / SRO #79
Question Deletions I
. RO # 04/ SRO #05
. RO #78 / SRO #73
*
. SRO #75
If you have any questions, comments or require any additional information, please contact Pete
Doran acting Nuclear Training Supervisor at 609-339-3816 or John Nichols Operations Training
Manager at 609-339-3769.
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Sincerely, I
['/ /
erome F. McMahon
/ Director-OA/NuclearTraining/EP
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INATIONALbo'
11e;merisin yourltuxh. .
95@G9 RLV 4192
EXAM ANSWER KEY CORRECTIONS EXAM QUESTION RO #19 .
Given the following conditions: . The plant is operating at 75% . Confirmed seal failures have occurred on the "B" Recirculation Pump . The pump hasjust been tripped
!
Which of the following describes the order for the "B" Recirculation Pump valve manipulations that j must be followed in order to ensure the pump will be completely isolated. j a. Close the Discharge valve, isolate seal purge, isolate RWCU flow from the loop and close the
suction valve.
b. Isolate seal purge, close the suction valve, isolate the RWCU flow from the loop and close the
discharDe valve
c. Close the suction valve, close the discharge valve, isolate seal purge, isolate RWCU flow from the
loop. ,
d. Isolate seal purge, close the discharge valve, isolate the RWCU flow from the loop and close the j
suction valve.
Ans: C Ref HC.OP-AB.ZZ-0112, " Recirculation pump Trip", rev.13 LP - 0302-000.00H400114-rev. 5 Obj. 3 1. Based on pre-examination discussions and referenced procedures, the critical step sequence is
based on the dim au*m item 5.7 of HC.OP-AB.ZZ-0112, "Recirculat';n pump Trip *(attached) and
.
precautions and limitations 3.1.2 of HC.OP-SO.BB-0002 * Recirculation System Operation"
(attached)
2. The suction valve must be closed before the discharge valve, and the seal purge must be !'
closed prior to pump isolation. This makes 'b' the only correct answer.
Recommendation: Change answer key to choice "b" as correct answer
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EXAM ANSWERKEYCORRECTIONS
EXAM QUESDON RO #27/ SRO #29 - The plant is in Mode 4 with Shutdown Cooling in service on the "A" Residual Heat Removal (RHR) loop with the "A" RHR Pump running. Which of the following desenbes how a loss of the "B" Reactor Protection System (RPS) bus will affect the inboard and the Outboard Shutdown Cooling isolation Valves (F008 & F009)? c. The F008 and F009 valves both close. b. The F008 valve closes and the F009 valve remains open, c. The F008 and F009 both remain open. d. The F008 valve remains open and the F009 valve closes. Ans.B Ref HC.OP-SO.SMM1(O), rev 5, page 3, section 3.1.3 LP 0302-000.00H-000045, rev 12 Obj. R3.b & R4
1. The answer key per the stated reference is incorrect. The correct answer per the stated reference
is "a".
RECOMMEDATION:
Change answer key to choice "a" as the correct answer.
3
EXAM ANSWER KEY CORRECTIONS
EXAM QUEST.ON RO #98 / SRO #96 Given the following plant conditions: . The plant was operating at rated power when a steam line break occurred in the HPCI room. . HPCI isolated due to high room tempsratures . RBVS exhaust radiation levels reached 1.0 E-2 microcuries/mi Which of the following describes the ventilation system response for the given conditions? c. RBVS remairs running. b. RBVS ir. dated,6 FRVS Recirc and 1 FRVS Vent fans are in service. c. RBVS isolated,4 FRVS Recirc and 1 FRVS Vent fans are in service. d. RBVS isolated,6 FRVS Recirc and 2 FRVS Vent fans are in service Ans. A Ref. HC.OP-EO.ZZ-0103, rev.10 LP 0302-000.00H-000127, rev 10, page 8 , Obj. 2 & R6 j 1. The answer key was incorrectly typed, the correct answer should be "b" 2. RBVS exhaust radiation levels reached (1.0 E-2 microcuries/ml) is > 1.0 E-3 which is the isolation )
signal for RBVS and an initiation signal for FRVS see HC.OP-SO.GU-0001 " Filtration,
Recirculation and Ventilation System Operation"
)
3. This is also an entry condition for HC.OP-EO.ZZ-0103, the lesson plan page listed lists the action
of HC.OP-EO.ZZ-0103 for the retention override that
if
. Reactor Bldg. exhaust Rad level exceeds 1 x10 4 l
95
.
Refuel Floor HVAC Exhaust Rad Level exceeds 1 x 10"
l
Then
. Verifyisolation of RBVS
And
. Initiation of FRVS
Recommendation - Change answer key to choice "b" as the correct answer
!
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______________.__________j
. _ _ _
____ _ _ _ - .
CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL A.* 3WER KEY .
EXAM OUEST10N RO #20 / SRO #24 . Given the following conditions: o Preparations are complete to start the "A" Recirculation Pump e o The Pump Discharge Valve (F031 A) is closed Which of the following describes how the "A" Recirculation Pump trip on the discharge valve is bypassed to allow the pump to be started? a. This trip is bypassed until the pump start sequence is complete within prescribed time limits. b. This trip is bypassed until the discharge valve has reached the 100% open position. c. This trip is bypassed until the pump has been running for 9 seconds. ; d. This trip is bypassed until the discharge valve jog (open) circuit has timed out. Ans A R:f 0302-000.00H-000020, rev 15, page 39 & 40, section VI.A.2.e.5).c) l LP 0302-000.00H-000020, rev 15, page 39 & 40, section VI.A.2.e.5).c) Obj RIO 1. The referenced Reciculation Flow Control Lesson Plan does not go into sufficient detail, neither in !
the lesson plan body nor in the leaming objectives, to differentiate between the discharge valve , jog circuit from the pump start sequence as the permissiva for pump start process completion.
2. Upon review of normal Control Room references (attached) it is shown on marked up sheets 8,
14, and 17;
.
That the K51 relay, which is energized during the start sequence, bypasses the 90% open trip
to the drive motor breaker until 85 seconds after the sequence has been initiated. This makes i
choice "a" a correct answer
.
That the K54 relay, which is denergized by the jog circuit timer, bypasses the full closed trip i
signal to the drive motor breaker for the first three seconds of jog circuit operation. This
makes choice "d" a correct answer.
RECOMMENDATION: Accept both a and d choices as correct answers,
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_____ - __ _ ____-__-. . _ - - _ _ _ _ - - _ _ - . _ _ . ._ _ _ _ -
._____ _____- _____-___________ __-
4
CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
EXAM QUESTION SRO #69
.
Which of the following is the basis of the 65 psig Suppression Chamber pressure Limit?
c. 65 psig is the primary containment maximum expected post-LOCA pressure.
b. Above 65 psig, the system lineup required for containment venting may not be able to be
completed.
c. Above 65 psig, the Safety Relief Valves may not be available when required for an Emergency
Depressurization.
d. 65 psig is the operational limit of the Torus to Drywell vacuum breakers.
Ans. C
Ref. 0302-000.00H-001268, " Primary Containment Control -Drywell Pressure" , rev
Obj. R6/R7
1. 0302-000.00H-001268, " Primary Containment Control-Drywell Pressure", rev-11 (attached)
states that 65 psig is the maximum pressure at which SRV's can be opened. This makes "c" the
correct answer
2. 0302-000.00H-00124A, "RPV Water Level Control", rev.10, (attached) states regarding the
Primary Containment Pressure Limit that above this limit
.
The vent valves lit the primary containment vent path above TAF may not open
.
The SRV's rc:ay not be able to be manually opened with PCIG at 90 psig.
3. This obvious discrepancy was discussed with the Operation Department Emergency Operating i
procedure writers, and the Primary Containment Pressure Limit / Maximum Primary Containment
Water Level limit worksheet (PSTG WS-9) identifies both the vent valves opening and SRV
opening as limiting components. This makes "t" also a correct choice
l
, Recommendation: !
Accep1 choices "b" and "c" as correct answers l
.
6
CORRECT ALTERNATE CHOICE ANSWERS FROM OfuGINAL ANSWER KEY EXAM OUEST10N RO #78 / SRO #71
'
During a loss of coolant accident the following conditions exist:
o Reactor pressure is 125 psig
o Drywell temperature is 325 *F
Which of the following describes the accuracy and trendog capabilities of wide range reactor water
level indication for the given condihons?
a. They are not providing accurate reactor water level or level trend information.
b. They are providmg accurate reactor water level but level trend is not rehable.
c. They are providing accurate reactor water level and level trend information.
d. They are not providing accurate reactor water lovel but level trend is reliable.
Ans. C
Ref EOP Caution 1, HC.OP-EO.ZZ-0101 RPV Water Level Control Section,
LP 0302-000.00H-00124A, rev 10
O bj. 7
1. The wide range instruments are calibrated for normal operating pressure and temperature, where
RPV level is significantly below Normal operating range. See attached 0302000.00H-000002
" Nuclear Boiler Instrumentation".
2. At lower than normal operating pressure the wide Gnge indicators read higher than actual leve'
when RPV level is above the mid scale range. See attached temperature compensation curves
from HC.OP-lO.ZZ-0003(O).
3. Since RPV level was not given, the accuracy of the Wide range level instrument is in question,
depending on the assumption of the candidate.
4. The conditions given show that the instrument Reference leg should not be affected by potential
flashing, since we are below the saturation curve, as couk' be determined by steam tables
provided to the candidates, this makes the instrument reliable for trending, as stated in EOP
caution #1
5. Based on the assumption of the candidate, either "c" Accurate level and trend,' or "d"
Inaccurate level but reliable trend would be acceptable answers
RECOMMENDATION:
Accept "c" or "d" as correct answers
,
7
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CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
Exam Question RO #79 Given the following conditions: o The plant is at 75% power o Control rod 22 27 is being withdrawn on notch to Notch "22" Which of the following is the required immediate operator action if a control rod drift alarm is received ana the operator notes control rod 22-27 is continuing to move out and power is rising? c. Apply a continuous insert signal to control rod 22-27. b. Place the Rod Select key lock switch to "off"(de-select the rod). c. Direct the local operator to perform a single rod scram on control rod 22-27. d. Runback recirculation flow and insert control rods to reduce power. Ans. D Ref HC.OP-AB.ZZ-0204 Positive reactivity addition, LP 302H-000.00H-000114, rev 5 Obj. 1 1. Runback recirculation flow and insert control rods to reduce power, is a prescribed method for
power reduction as stated in HC.OP-AB.ZZ-0204 section 3.1 which makes "d' a correct choice.
2. Applying a continuous insert signal to control rod 22-27 is a method of " inserting control rods to
reduce reactor power" and therefore, makes choice "a" a correct answer IAW HC.OP-AB.ZZ- 0204.
3. Additionally, since the question states that "the operator notes that control rod 22-27 is continuing
to move cut and power is rising", the operator could enter abnormal procedure HC.OP-AB.ZZ-
0102 Dropped Control Rod. IAW with this procedure the immediate actions are to:
.
If necessary then Insert control rods, in sequence, to terminate the power increase.
.
If a scram condition is reached, Then ensure the reactor scrams and implement procedure
HC.OP-EO.ZZ-0100(O)
. Ensure that all appropriate automatic actions are complete.
4. Inserting control rod 22-27 would be correct for this abnormal procedure since that would be the
first rod to insert "in sequence".
R' COMMENDATION: c , cept choices "a" and "d" as correct answers
8
CORRECT ALTERNATE CHOICE ANSWERS FROM ORIGINAL ANSWER KEY
EXAM QUESTION RO #84 / SRO #79
A loss of Reactor Auxiliary Cooling System has occurred?
Which of the following is the MAXIMUM time alkwed, before a reactor scram is required?
c. An immediate scram is required
b. One(1) minute
c. Ten (10) minutes
d. Twenty (20) minutes
Ans C
Ref HC.OP-AB.ZZ-0123, rev 5, caution 4.8
LP 0302-OOO.00H-000114, rev 9, page 3
Obj. 3
1. The answer key has "a" as being correct, based on caution (4.8) of HC.OP-AB.ZZ-0123 which
allows 10 minutes to get RACS restored to the recirc pumps or they must be tripped, the operator
is cautioned to place the mode switch in " shutdown" prior to tripping the pumps. This makes "c" a
correct answer.
2. Section 4.9 of the same procedure states if a totalloss of RACS has occurred and cannot be
immediately restored them perform the following:
. Scram the reactor
. Trip both Recirc pumps
. Trip both CRD pumps
. Trip both RWCU pumps
s
3. One SRO candidate aslad the exam proctor if this loss was a * total loss". His response was yes.
Using a totalloss and following that direction, this would make "a" also a correct answer.
Recommendation
Accept choices "a" or "c" as correct answers
9
. _ _ .
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QUESTION DELETIONS
Exam Question RO #04/SRO #05 i'A;&g shift tumover the Nuclear Control Operator (RO) notes that data entered in the narrative log by the previous shift incorrect. The RO draws a single line through the incorrect entry, makes the correct entry and initials and dates the change. Which of the following describes how the RO should highlight and explain the change? c. The correct entry should be circled in red with an explanation placed in the comments section. b. The ;orrect entry should be circled in red with an explanation made next to the corrected entry. c. The incorrect entry should be circled in red with an explanation placed in the comments section. d. The incorrect entry should be circled in red with an explanation made next to the corrected entry. Ans. A Ref HC.OP-AS.ZZ-0002, rev 2, page 20, section - Log Taking LP 0302-000.00H-000113, rev 8 Obj. 125R 1. LP-0302-000.00H-000113, rev 8 objective 125 (attached ), specifically states "Given access to
control room references, distinguish between proper and improper methods of maintaining Operations Department logs IAW HC.OP-AP.ZZ 0110. This procedure was not provided for the candidates to review to determine correct choice.
2. HC.OP-AP.ZZ 0110 (applicable pages attached) defines the use of the Narrative and Comments
section logs. It also describes Data logs and requirements of circling abnormal, unusual, or O.O.S.
data in red ink, additionally it states that any abnormal, unusual, or O.O.S. entnes will be
investigated immediately and recorded on the applicable comments section. HC.OP-AP.ZZ-0110
'
further has a desenption of the Comment Sheets / Sections and states they are the Narrative Log
for operating stations that do not have a formal Narrative Log ledger.
3. HC.OP-AS.ZZ-0002, page 20 (attached) specifically states if an entry is corrected by an individual
other than the person entering the ggts, the correction must be circled in red with an explanation in the 221DDREME.ggg8gn.
4. The NCO Narrative log (attached) is a comments logs in itself and not a data log. Data is taken
on logs such as DL 0002 (attached) which has a comments section. The misapplication of the NCO Narrativr> Log as the Data log vice any DL log supplied with a comments section, prevented the candidates from determining the correct selection.
RECOMMEDATION: Delete question from exam
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QUESTION DELETIONS
~
Exam Question RO #78 / SRO #73 4
.
Given the following conditions:
. Reactor poweris 82%
. HPCI is in operation for a surveillance
. The "B" loop of RHR is in Suppression Pool Cooling
.. Suppression Pool temperature is 103*F when the running RHR Pump tripped
. HPCI was secured . Subsequently, suppression pool temperature reached 106'F Which of the following lists the suppression pool temperatures requiring entry into HC.OP-EO.ZZ-
0102, Primary Containment Control AND entry into the LCO actions for Tech Spec 3.6.27 c. EO-0102 - 95'F
1 S 3.6.2 -
95 F
b. EO-0102 - 95*F
TS 3.6.2 -
105 F
c. EO-0102 - 105'F
TS 3.o.2 -
95*F
d. EO-0102 - 105*F
TS 3.6.2 -
105*F
Ans: D Ref: 0302-000.00H-00125A, " Primary Containment Control - Suppression Pool Temp", rev 10
HC.OP-IS.BJ-0001, "HPCI inservice test", step 5.1.16, rev 29
Obj. 3
1
1. 0302-000.00H-00125A, " Primary Containment Control- Suppression Pool Temp", Rev 10,
objectives do require knowledge of entry conditions to EOP-0102 (attached)
2. HC.OP-IS.BJ OOO1, rev 29, step 5.1.16 states to implement suppression pool average water
temperature monitoring of technical specification 3.6.2.1 prior to and during HPCI operations by performing HC.OP-DLZZ-0026(O) (both attached)
3. No leaming objective in the Hope Creek Operations Traininc program requires commitment to
memory inservice Test cautions and bases behind the cautions.
4. No Leaming Objective for Technical Specification evaluation require determination of Technical
Specification actions without having the applicable section of the procedure availab'a for reference.
5. Testing Technical Specification compliance without the materials available for review is not in the
best interest of the candidate or in compliai.ce with Hope Creek Operations Training Department objectives.
6. Nuclear Business Unit Procedural Compliance requirements, and expectations, for use of a
Catergory I procedure require step by step compliance. The same level of procedural usage should be complied with during examinations, and was not.
Recommendation: Delete quenion .
11
QUESTION DELETl_QHE
EXAM QUESTION SRO #75 Which of the following describes how the operators would know the Hydrogen Water Chemistry injection (HWCl) system had NOT been removed from service while performing a shu;oown in accordance with HC.OP-lO.ZZ-0004(O), " Shutdown from Rated Power to Cold Shutdown"? a. Hydrogen explosions in the Mechanical Vacuum Pump while operating to maintain condenser
vacuum.
5
b. Post shutdown (2 hours) Turbine Building radiation levels would be much higher. c. Alarms and indications resulting from a control rod drop accident would not be available to the
operators as quickly.
d. The Primary and Secondary Condensate Pumps will cavitate. Ans. C Ref HC.OP-AB.ZZ-0102, Dropped Control Rod, rev. 3 LP 0302-000.00H-000225, rev 05 Obj. 6 & 7.1 in order for this situation to occur the operators would be required to violate procedure HC.OP-lO.ZZ- 0004," Shutdown from Rated Power to Cold Shutdown". If the operators failed to have the Chemistry Department remove HWCl from sen.ve at 35% power, they would also have to miss the next step of the procedure which instructs the operatun, to have I&C restore the MSL RMS setpoints. If the ettpoints are restored with HWCl in service then RMS alarms may result which could clue the operators in to the problem with HWCl. This scenario requires multiple procedure violations. There is no power level specified for this question and in order for HWCl to remain in service it would have failed to trip at 30% power (as it is currently designed). Technical Specific.ations require that with reactor power at 20%, the only control rod motion that is d! owed is by a scram if MSL Rad Monitor Setpoints have not been restored. HC.OP-AB.ZZ-0102 " Dropped Control Rod" section 5.3 states "The effects of a rod drop accident above 20% power are minimal; therefore, H2 injection system operation is only permitted above 20% power". There are numerous protections to prevent the conditions specified in this question from occurring. The likelihood of all of those failures and then a rod drop accident are too remote to expect the ctudents to select choice "c" as the correct answer. RECOMMENDATION: Delete question from exam j
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1
HC4P-SO.CH-0001(Z)
ATTACHMENT 4
(Page1of1)
.
MAIN TURBINE CONTROL OIL (EHC) SYSTEM OPERATION
EHC CONTROL LOGIC DIAGRAM .
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ATTACHMENT 4
NRC RESOLUTION OF PSE&G COMMENTS ON THE WRITTEN EXAM
. .
,
RO-4 / SRO-5 The f acility recommended to delete this question from the exam.
Based on a review of the references provided, the NRC staff agreed
with the facility that this question should be deleted from the exam.
There was no clear reference to clearly support a correct answer to
this question.
RO-19 The facility recommended to change the answer key from "c" to "b".
Based on a review of the procedure HC.OP-AB.ZZ-0112, Recirculation
Pump Trip, there was no answer that provided the sequence to isolate
the recirculation pump as required by the procedure. The NRC staff
did not totally agree with the facility recommendation. Since there
were no correct answers to this question, the appropriate action was ~
. to. delete the question from the exam.
RO-20 / SRO-24 The facility recommended to accept both "a" and "d" as correct
answers to 't he question. Based on review oflhe refer' enc ~es, the
NRC st.aff agreed with the facility. The answer key was revised to
accept "a" and "d" as correct answers.
RO-27 / SRO-29 The f acility recommended to change the answer key from "b" to "a".
Based on review of the references and prior exam versions, this was
clearly a typogrephical error, the NRC staff agreed with the facility.
The answer key was revised from "b" to "a".
RO-76 / SRO-71 The facility recommended to accept both "c" and "d" as correct
answers to the question. Caution 1 of the emergency operating
procedures indhted that if drywell temperature and reactor pressure
were below the saturation curve then wide range level indication was j
a reliable instrument. Since the given conditions were below the j
saturation curve and steam tables were available to the applicants, j
they had sufficient information to concluded that the wide range level
instrument was useable for the entire range and thus "c" was could
be a correct answer since no accuracy range was delineated.
The examiner further reviewed the licensee provided curve showing
inaccuracy of the water level instruments over a range of levels and
of reactor coolant system pressures and temperatures. Answer "d" is
also correct in that the level instrument would not be providing 1
accurate water level indication but the trend would be reliable. After i
further review answer "a" is also correct with the "or" condition that I
the instruments "are not providing accurate reactor water level or j
level trend information. !
Accordingly, since the question has three correct answers, it was
deleted from the exam for the reasons noted by the examiner above. !
I
1
I
_
c 3
l
RO-78 / SRO-73 The facility recommended that this question bo deleted from the
exam. The licensee indicated that the question was testing the
applicant's memory of specific technical specification limiting
)
1
, condition for operations (LC) actions or emergency operating.
procedure actions as suggested. The examiner viewed the question
as testing.the applicant's knowledge of the entry conditions into these
documents at the analysis level, which is a more challenging question.
This was a acceptable testing area as identified by the KA assigned to' j
this question and because of the importance of this LC. Since there
was a single correct answer to the question, there was no basis to
delete the question from the exam. An acceptable basis would have
been no correct answer or more than two correct answers. The
facility comment was not accepted.
RO-79 The facility recommended to accept both "a" and "d" as correct
answers to the question. The question required the applicant to
identifV the required.immediate operator actions. Answer "a"'was not .
a required immediate operator action identified in HC.OP-AB.ZZ-0204, -
Positive Reactivity Additiori. The facility recommendation was not l
accepted. ]
~
RO-84 / SRO-79 The facility recommende.d to accept both "a" and "c" as correct
answers since one applicant was told by the proctor, in response to a
question, that this was a totalloss of RACS. The proctor's response
did not alter the question since ten minutes is still the maximum tirne I
allowed before a reactor scram is required and answer "c" is the only
correct answer. There was no change to the answer key.
RO-98 / SRO-96 The facility recommended to change the answer key from "a" to "b".
Based on review of the references and prior exam versions this was
clearly a typographical error and the NRC staff agreed with the
facility. The answer key was revised from "a" to "b".
SRO-69 The facility recommended to accept both "b" and "c" as correct
answers to this finestion. Based on review of the references, the
NRC staff agreed with the facility. The answer key was revised from
to accept both "b" and "c" as correct answers. i
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l
SRO-75 The f acility recommended to delete the question from the exam j
without sufficient supporting justification as to why it should be !
deleted. The question was based on the discussion section of HC.OP-
AB.ZZ-0102, Dropped Control Rod, on why hydrogen injection is :
secured at low power. This was a legitimate testing area as identified j
by the KA assigned to the question. Since the question was valid ;
with the one correct answer to the question, there was no basis to I
delete the question from the exam. There was no change to the
answer key.
1
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ATTACHMENT 5
ElGNIFICANT CONTP.0L MANIPULATION DETAILS
APPLICANT DAT_g TYPE ASSESSMENT - r 55-62176 4/6/97 Flow Acceptable.
4/6/97 Flow Unacceptable - No documentation available to support
that this was not part of a continuous power change.
4/6/97 Flow Unacceptable - No documentation available to support
that this was not part of a continuous power change.
4/6/97 Rods Acceptable.
4/6/97 Rods Unacceptable - Documentation indicated that this was
part of'a continuous power change.
4/6/97 Rods Unacceptable - Documentation indicated that this was
part of a continuous power change.
~
4/6/97 Rods Unacceptable Documentation indicated that this was
part of a continuous power change.
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
APPLICANT DATE IY.EE ASSESSMENT 55-62178 2/1/97 Flow Acceptable
- -
. .. . . . -
Flow Acceptable -
'
3/1/97
.
3/1/97 Rods Unacceptsble - Rod movemew consisted of inserting 5
rods from 16-12 to reduce power and then three
examples of partially withdrawing a control rod
individually scrammed by a licensed operator as part of
individual control rod scram testing, another applicant
also completed the withdrawal. No documentation was
available to support that the control rod movement
resulted in an observable effect on power. Rod
withdrawal to recover from an individual rod scram test
was not considered to be significant.
3/1/97 Rods Unacceptable - Rod movement consisted 6f eig'ht'
'
examples of parti;lly withdrawing a control rod
it.dividUally scrammed by a licensed operator as p' art of
individual control rod scram testing. Another applicant
also completed the withdrawal. Rod withdrawal to
recover from an individual rod scram test was not
considered to be significant.
3/1/97 Rods Unacceptable - Rod movement consisted of three
examples of partially withdrawing a control rod
individually scrammod by a licensed operator as part of
individual control rod scram testing, another applicant
also completed the withdrawal, and withdrawing 5 rods
from 12-16. No documentation was available to
support that the control rod movement resulted in an
observable effect on power. Rod withdrawal to recover
from an individual rod scram test was not considered to
be significant.
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
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APPLICANT DATE TYPE ASSESSMENT
55-62183 - 2/2/97 Flow Acceptable
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3/1/97 Flow Acceptable
2/2/97 Rods Unacceptable - Rod movement consisted of inserting
four rods from 10-06 and then withdrawing the same
four rods from 06-10. This did not meet the PSE&G
acceptance criteria of at lease one notch for a minimum
of eight rods.
3/1/97 Rods Unacceptable - Rod movement consisted of inserting 3
rods from 08-00, four rods from 14-12 and three rods
from 1612. There was no documentation to support
that this resulted in an observable power affect.
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3/1/97- Rods Unacceptable - Rod movement consisted of eight
examples of partially withdrawing a control rod
' individually s' crammed by a licensed operator as' part of
individual control rod scram testing. Another applicant
also completed the withdrawal. Rod withdrawal to
recover from an individual rod scram test was not
considered to be significant.
2/21/98 Flow . Acceptable
2/21/98' Flow Acceptable
2/21/98 Flow Acceptable
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APPLICANT DAIE IyfE ASSESSMENT
55-02187 2/2/97 Flow Acceptable
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3/1/97 Flow Acceptable
3/1/97 Rods Unacceptable - Rod movement consisted of eight
examples of partially withdrawing a control rod
individually scrammed by a licensed operator as part of
individual control rod scram testing. Another applicant
also completed the withdrawal. Rod withdrawal to
recover from an individual rod scram test was not
considered to be significant.
3/1/97 Rods Unacceptable - Rod movement consisted of eight
examples of partially withdrawing a control rod
individually. scrammed by a licensed operator as part, of
individual control rod scram testing. A'nother applicant
also completed the withdrawal. Rod withdrawal to
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recover from an individual rod scram' test wa's not
considered to be significant.
3/1/97 Rods Unacceptable - Rod movement consisted of
- withdrawing 7 rods from 12-16 and one rod from 00-
08. There was no documentation to support that this
resulted in an observable power affect.
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
2/21/98 Flow Acceptable
APPUCANT DATE IYfE ASSESSMENT 55-62175 4/6/97 Rods Acceptable
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4/6/97 Rods Acceptable
4/6/97 Rods Acceptable
4/6/97 Rods Unacceptable - Rod movement consisted of
withdrawing four control rods from notch 00-06 and
then withdrawing the same four rods from notch 06-12.
This did not meet PSE&G acceptance criteria of at least
one notch for a minimum of eight rods.
C/20/97 F!ow Acceptable
2/21,/98 Flow Acceptable .
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- APPUCANT DAH IYE ASSESSMENT 55-62174 4/6/97 Flow Acceptable
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v/3/97 Flow Acceptable
7/10/97 Flow Acceptable
9/4/97 Flow Acceptable
4/6/97 Rods Acceptable
4/6/97 Rods Acceptable
4/6/97 Rods Acceptable
4/6/97 Rods Unacceptable - Rod movement consisted of
witndrawing four control rods from notch 12-14, then -
these same four rods from 14-16, and these same four
rods agairi frcm 16-18. This did not meet the PSE&G
acceptance criteria of at least one notch for a minimum i
of eight rods.
5/9/97 Rods Did not assess since applicant had the required number.
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6PPUCANT DAIE TYPE ASSESSMENT 55-60813
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12/13/97 Rods Acceptable
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12/13/97 Rods Unacceptable - Documentation indicated that this was
part of a continuous power change.
12/13/97 Rods Unacceptable - Documentation indicated that this was -
part of a continuous power change.
12/13/97 Rods Acceptable
12/14/97 Flow Acceptable
12/14/97 Rods Acceptable
2/21/98 Flow Acceptable
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