IR 05000400/1993301
| ML18011A155 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/01/1993 |
| From: | Baldwin R, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18011A154 | List: |
| References | |
| 50-400-93-301, NUDOCS 9309130190 | |
| Download: ML18011A155 (87) | |
Text
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Report Nos.:
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W., SUITE 2900 ATLANTA,GEORGIA 303234199 50-400/93-301 Licensee:
Carolina Power and Light Company P. 0.
Box 1551 Raleigh, NC 27602 Docket Nos.:
50-400 Facility Name:
Shearon Harris Nuclear Project Examination Conducted:
Week of August 2, 1993 License Nos.:
NPF-63 Chief Examiner:
Richard S. Baldwin
~/il~z Date Signed Examiners:
B. Hemming, INEL Approved By:
Qg~~)(.u..
Lawrence L. Lawyer, Chief Operator Licensing Section Operations Branch Division of Reactor Safety Date Signed SUMMARY Scope:
NRC examiners conducted regular, announced operator licensing requalification examinations during the period of August 3-5, 1993.
Examiners administered examinations under the guidelines of the Examiner Standards (ES),
NUREG-1021, Revision 7.
Ten Senior Reactor Operators (SROs)
and two Reactor Operators (ROs) received written and operating examinations.
For the simulator portion of the examination, operators comprised three crews.
The examiners also reviewed the record of simulator usage during January through June of 1993.
Results:
Operator Pass/Fail:
SRO RO Total Percent Crews Percent Pass Fail
12 100
%
0%
100
%
0/
9309130190 930901 PDR ADOCK 05000400 V
Examiners judged that the Shearon Harris Nuclear Plant requalification program was satisfactory based on the results of the examinations.
Examiners identified weaknesses regarding licensee evaluator's analytical ability (paragraph 2.d), the development of Part A (static simulator)
examination written questions (paragraph 2.c(l)), Annunciator Panel Procedure usage (paragraph 2.f(l)(a)), and Shift Supervisor's involvement during simulator scenario events (paragraph 2.f(l)(b)).
Examiners identified strengths regarding crew communications (paragraph 2.f(2)(a)),, EOP usage (paragraph 2.f(2)(b)),
and the STA's problem solving ability (paragraph 2.f(2)(c)).
Examiners identified an inspector follow-up item (IFI) regarding adequacy of procedure FRP-H.1,
"Response to Loss of Secondary Heat Sink" (paragraph 2.d(l)).
No violations or deviations were identifie REPORT DETAILS 1.
Persons Contacted Licensee Employees
- R. Bright, Simulator Support
- J. Collins, Manager, Training
- W. Robinson, General Manager
- J. Pierce, Manager, Initial Training
"A. Powell, Manager, Operations Training
- D. Tibbits, Manager, Operations
"A. Williams, Manager, Shift Operations Other licensee employees contacted included instructors, engineers, technicians, operators, and office personnel.
NRC Personnel T. Peebles, NRC Region II
- J. Tedrow, SRI Shearon Harris
- D. Roberts, RI Shearon Harris
"Attended exit interview 2.
Discussion a.
Scope NRC examiners conducted regular, announced operator licensing requalification examinations during the period of August 3-5, 1993.
Examiners administered examinations under the guidelines of the Examiner Standards (ES),
NUREG-1021, Revision 7.
Ten Senior Reactor Operators (SROs)
and two Reactor Operators (ROs) received written and operating examinations.
For the simulator portion of the examination, operators comprised three crews.
The examiners also reviewed the record of simulator usage during January through June of 1993.
b.
Reference Material The examination team reviewed the reference material and determined that it was adequate to support the examination.
The examiners also reviewed the licensee's 1991-1992 Licensed Requalification Program sample plan.
The sample plan was compared to NUREG-1021, ES-601, Attachment 2.
The sample plan met all those guidelines for NRC administered requalifi-cation examinations.
However, using the sample plan to manually verify the actual test was a very cumbersome and tedious process.
The document provided to the NRC was limited in that it did not contain all subtopic information available on the computerized database located at the facility.
The examiners attempted to verify that the examination questions were a valid sample of the learning objectives, but found that this can only be done utilizing the computerized database.
Therefore, NRC examination validation can only be done on sit Report Details c.
Examination Development The examination team conducted a review of the licensee's proposed written, walk-through, and dynamic simulator examinations in accordance with the guidance provided in NUREG 1021, ES-601,.ES-602, ES-603 and ES-604.
The analysis of this review follows:
Written Examination The licensee submitted an examination that adhered to the require-ments of the sample plan.
However, the Part A portion of the examination did not meet the guidelines of NUREG-1021, ES-602, Attachment 2, in that it did not effectively discriminate between a
competent licensee and one who is not.
Each examination was comprised of a common Part A, an RO Part B,
and an SRO Part B.
The combined written examination exceeded the total number of ques'tions described in NUREG-1021.
This was due to short question validation time.
The examination team reviewed the questions for level of knowledge, discriminatory value, technical content, and time validation.
The Part 8 examination distractors required enhancements.
Some of the questions'istractors were not plausible and were rewritten.
One question concerning an immediate operator action from an EOP had a
test time of four minutes.
The examination team considered this excessive.
This question's allotted time was decreased.
Two questions were added to the RO examination in order to ensure the total examination time validation met ES-602 guidelines.
Thirty percent of the Part B examination required minor changes.
The Part A portion of the written examination required the most work in order to meet the requirements of ES-602.
The examiners had to modify or replace approximately 45 percent of the proposed questions.
The examination team identified many questions on this portion as lists of true/false statements that could be determined incorrect simply from the static simulator setup.
The majority of the questions required significant revisions to the stem or distractors.
(2)
Job Performance Measures (JPHs)
The examination team considered the proposed JPHs to be satisfactory.
Minor changes were made to ensure designated critical steps met the guidelines of ES-603.
In some instances, initially assigned critical steps were not critical and non-critical steps were found to be critical.
The examination team changed these accordingly.
The licensee proposed five JPHs for the examination.
The NRC added two additional JPHs in order to provide a larger sample for observation.
Each operator performed five JPMs during the plant walk-throug Report Details (3)
Simulator Scenarios The examination. team reviewed the two scenarios proposed by the licensee.
The NRC examiners proposed an additional simulator scenario outside of the sample plan.
The examination team used one proposed simulator scenario and the NRC substituted scenario for the two scenarios used during the examination.
The additional licensee proposed scenario was maintained as a spare scenario.
The examination team reviewed the simulator scenario critical tasks in accordance with ES-604, Attachment 1, Critical Task Methodology.
The team determined that critical task grading criteria was not well defined.
Proposed critical tasks needed to be customized or tailored to the conditions or events in the scenario in order for the NRC and licensee examiners to grade consistently.
The examination team added additional initial condition information that was common between the two scenarios used for the examination.
Having the same initial conditions relieved stress on the operators.
d.
Examination Administration The NRC administered the examination in accordance with the guidelines of NUREG-1021, Revision 7.
The licensee's aggressive schedule of three days to accomplish the entire examination became delayed during the administration of the JPM portion of the examination.
This required approximately a 14-hour day to complete the JPMs.
The licensee took great care to minimize the stress levels of the operators; however, when the JPMs were administered, one group of operators had to wait for a number of hours for their examination.
One JPM, CR0-015,
"Perform Heat Balance Calculation," required altering cues during the administration of the JPM.
The cues were not explicit, in that it was not evident what method was to be used to calculate the heat balance based upon the initial conditions.
Sequestering of the operator s was well planned and executed by the licensee.
During the simulator examination, the licensee's evaluators were thorough and asked pertinent follow-up questions.
However, when crews performed the same steps under similar conditions differently, the evaluators did not display the analysis ability necessary to determine what these differences represented (e.g.,
the need for procedural change or additional operator training, etc.).
Because the crew passed the critical task, the utility evaluators neither noted this nor questioned the'operators.
Examples of this lack of questioning attitude were:
(1)
Step number 19 of FRP-H.1,
"Response to Loss of Secondary Heat Sink,"
REV 6, requires the operators to "Actuate SI."
Two crews decided not to perform this step.
Manual initiation of SI would have required them to re-establish Instrument Air and Feedwater systems.
Prior to this step, the crews had already recovered from the first Safety Injection (SI) initiation.
Not following
Report Details procedural steps during the course of an event requires operator decision.
This could lead to errors in determining whether some aspect of the SI actuation had failed to occur.
It is not the intent of the Emergency Operating Procedures to require operators to determine whether a step should or should not be performed during an event.
Additional guidance is necessary in this matter in order to ensure all operators perform steps as procedurally directed.
This item was identified as Inspector Follow-up Item, IFI 400-301-01.
(2)
One crew decided to initiate the feed and bleed procedural steps prior to meeting the fold out page criteria of FRP-H-I.
This decision was based upon all three Pressurizer PORVs (Power Operated Relief Valves) cycling from full open to full closed with the pressurizer water solid.
Additional review of this procedure by the licensee was required in order to determine whether supplemental foldout page criteria was warranted.
e.
Procedures The examination team noted one procedural problem during the course of the examination.
AOP-004, Remote Shutdown, REV. 6, step 2b, describes the "B" train SW/NB switch as SW/NA.
f.
Operator Performance The examination team noted overall performance of the operators was considered satisfactory.
The NRC team considered the administered examination to be safety significant and discriminating.
All operators passed the examination.
(I)
The examination team noted two generic weaknesses during examination administration which are discussed below.
(a)
During the simulator examinations, very few operators used the Annunciator Panel Procedures (APPs) for guidance.
However, no adverse effects were noted due to the lack of use of the APPs.
(b)
The Shift Supervisor (SS),
was not directly involved in the recovery of the plant after he entered the Emergency Action Level (EAL) flow path on some occasions.
The flow path was time consuming and kept the SS from helping the Senior Control Operator (SCO)
when necessary.
Additional assessment by the licensee was necessary to determine whether the SS should be predominately occupied with EAL flow path during the course of plant events.
(2)
The examination team also noted three generic strengths during examination administration as discussed belo Report Details (a)
All three crews exhibited very good verbal communications.
They did a good job in maintaining each other's awareness of plant conditions.
They also did a good job of informing personnel outside of the control room.
(b)
All three crews used the Emergency Operating and Abnormal Operating procedures very well.
The crews referenced the correct procedure and used them with knowledge and confidence.
(c)
The examination team noted the strength of the STA's knowledge.
All STAs exhibited excellent diagnostic abilities.
Their presence during the scenarios proved their ability as part of the operating crew.
(3)
All operators passed the written examinations.
The grades ranged from 85.7 percent to 100 percent.
(4)
Two SROs failed one JPH each.
No individual operator failed the operating examination because of the JPH failures.
g.
Simulator Facility The simulator facility performed well during the course of the examinations.
No hardwa} e or software failures occurred.
3.
Simulator Usage The inspectors reviewed the schedule for a six-month time period; January to June 1993.
The inspectors noted that the simulator was available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day for seven days per week.
This review determined that Licensed Operator Requalification (LOR) simulator training had top priority.
LOR simulator training always occu} red during 0730-1230 hours and 1730-2230 hours, a total of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per day for each 5-day week.
This occurred for 4 days per week, for a total of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week.
The Training Unit used the simulator on Friday for simulator scenario development and validation.
The Training Unit staff used the simulator during 1230-1730 hours for extra training.
The time available to the operators for LOR training was approximately 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />.
The simulator was available approximately 1034 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.93437e-4 months <br /> for training for the period under review.
This provided the LOR classes with approximately 40 percent of the time used for requalification training during either the day or the swing shift. It was determined that all additional tr aining was accomplished on the swing or midnight shifts.
The total hours in the period were 4344, of which 1713 hours0.0198 days <br />0.476 hours <br />0.00283 weeks <br />6.517965e-4 months <br /> represented maintenance hour Report Details 4.
Exit Interview At the conclusion of the site visit, the examiners met with representatives of the plant staff listed in paragraph I to discuss the results of the examinations.
The licensee did not identify as proprietary any material provided to, or reviewed by the examiners.
The examiners further discussed in detail the inspection findings listed below.
Dissenting comments were not received from the licensee.
Item Number Descri tion and Reference IFI 400-301-01 Inspector follow-up item regarding the documentation that establishes the basis for actuating or verifying SI when implementing FRP-H.l (paragraph 2.d(l)).
ENCLOSURE 2 SIMULATION FACILITY REPORT Facility Licensee:
Shearon Harris Nuclear Plant Facility Docket No.:
50-400 Operating Tests Administered on:
Week of August 2-6, 1993 This form is to be used only to report observations.
These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b).
These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations.
No licensee action is required in response to these observations.
While conducting the simulator portion of the operating tests, the following items were observed (if none, so state):
ITEM DESCRIPTION None Note LOR EXAMINATION WEEK 3 WRITTENEXAM EXAMINATIONPACKAGE CONTENTS ga>>( s 93-ZO)
Wad Ye Part A Static Simulator Scenario (1) and Questions (19)
Scenario:
SSS-001, SMALL BREAK LOCA A02-005 A02-052N A02-070 A02-093 A03-056 A03-063 A03-081 A05-010 A05-025N A07-001 A07-033 A09-013 A09-045 A01-018 A01-023 A03-082 A04-020 A05-008 A06-009 Part B Exam Questions RO PART B EXAM (25)
SRO PART B EXAM (23)
BOO-011 BOO-032 BOO-035 B01-001 B01-003 B01-010 B01-023 B01-052 B02-006 B02-013 B03-024 B03-025 B03-028 B03-122 B04-024 B04-030 B05-012 B05-020 B07-002 B07-005 B07-041 B08-001 B08-002 B09-010 B11-006 BOO-005 BOO-011 BOO-035 B01-001 B01-002 B01-010 B01-052 B02-011 B02-047 B03-010 B03-024 B03-028 B03-116 B04-024 B04-030 B04-048 B05-020 B05-046 B07-005 B07-050 B08-001 B11-006 B11-008 Job Performance Measures (5)
JPM-CR-006 JPM-CR-015 JPM-CR-103 Groups 1 and 2:
Group 3:
JPM-IP-099 JPM-IP-050 JPM-IP-112 JPM-IP-118 Dynamic Simulator Scenarios (2)
DSS-017 DSS-002 Exam Package Compiled By:
Date:
Simulator:
JPMs:
Written:
Total:
125
135 313 minutes minutes minutes minutes REEK 3 1993 LICENSED OPERATOR REQUALIF*CATION EXAM Q~ p.p-i'z 93-~ ~J
/ qz/e~
RO and SRO Static Simulator (Part A) Written Examination Composition QUEST TOPIC CONTENT TIME A02-005 SYSTEMS/CFW SI and FRV closure 0.6
- A02-052N SYSTEMS/ESFAS A02-070 SYSTEMS/SIS Evaluate secpxencer-affected components Calculate RWST volume until auto swa over 0.6 1.6
- A02-093 SYSTEMS/RHR RHR pump recirculation flow path when running dead-headed 0.6 A03-056 SYSTEMS EDG Emer enc non-emer enc, trips 1.3
- A03-063 SYSTEMS/PRZ PRZ low-low level heater interlock and reset
- A03-081N MODS/CVC A05-010 SYSTEMS SG
- A05-025N SYSTEMS SDs PMi-.abX
- A07-033 5FSTEM'S Pj~JN,"'":i 5%7&4%iDG SYSTEMS/ESW fiick-,0'ii'i
%~le'o>bio,'.
A09-013 SYSTEMS/NIS A09-045 SYSTEMS/AMSAC A01-018 SYSTEMS CRD A01-023 SYSTEMS/CVC CSIP alt. mini-flow response Determine SG PORV set oint Low-low tern steam dum interlock kvatua8a:-'.Matey-,:.,yead~enewioE,:;:;41!
Im act of SI on ESW pumps Source Ran e NI reenergization Determinin AMSAC status Power cabinet urgent failure Diagnose inadvertant dilution caused by letdown cooling Pii~yjgessure:.';.:con%ai~w~'pic Saturated 0.3
- A04-020
- A05-008
- A06-009 SYSTEMS/RHR SYSTEMS/AFW SYSTEMS/SPRAY Impact of loss of instrument air at half-loop Affect of transfer to ACP on the AFW automatic start signals Spray addition tank isolation
1.3
18.3
- indicates exam measure was changed by exam team
Limits and Controls (Part B) Examination Composition Questions Common to the RO and SRO Examinations QUEST.
TOPIC B00-011 PEPs/PEP-341 B00-035 ~
PEPs/PEP-301
- B01-001/
AOPs/AOP-019
- B01-010 ~
Normal Ops/
GP-004 CONTENT Dose ro)ection met. data Off-site fire assistance RCS low pressure recpxiring reactor trip Actions for inadvertent critical-ity during reactor startup TIME 0.6 0.6 0.6 1.3
- B01-052 /
AOPs/AOP-002 Total boration for stuck rods followin reactor tri 1.3
ECA-3.1 FRP applicability in EPP-010/
ES-1.3 1 3 B04-024 AOPs/AOP-018 Actions for loss of CCW to RCPs B04-030 r B05-020, Normal,Ops/
RHR pump starting duty Criterion to shift AFW to ESW 0.6 B07-005
~
EOPs/EPP-001 Criterion to secure cooldown (SG de ressurization based on SR SUR
1.3 B08-001 AOPs AOP-036 SG level tar et 1.3 B11-006 Rad Control New CP&L annual whole body dose limit
12.9
- indicates exam measure was changed by exam team
Limits and Controls (Part B) Examination Composition Questions Unicpae to the RO Examination QUEST TOPIC CONTENT TIME BOO-032 PEPs PEP-321 Emer enc facility activation 0.6
- B01-003 B01-023 EOPs/FRP-S.1 Normal Ops/
Boration Alternate emergency boration path for ATWS Calculate boric acid addition to reposition control rods
,3 2.5 B02-006 EOPs/Path-1 Protocol if safeguards component fails to start on se encer 0.3
- B02-013 EOPs/EPP-008 Criteria to reestablish BIT in-jection following SI termination in EPP-008 ES-1.1 5~~~'in@ i!ccin'd3Pkensiztl~ia~
1.3 B03-122 'OPs/FRP Rules SOP+P&~OA~
Rule to stay in FRP-J.1 until completion unless higher priority FRP re ired B07-002 AOPs/AOP-015 B07-041 EOPs/EOP-005 Actions for grid underfrequency with turbine tripped CRDM fan availability and requir-ed subcooling margin in EPP-005 (ES-0.2)
1.3
- B08-002 B09-010 AOPs/AOP-017 Refueling/
Tech.
Specs.
Criterion to trip reactor with lowerin Instrument Air pressure Audio count rate monitor requi.re-ments during refueling
13.3 B05-012 substituted for B05-058 due to substituting loss of secondary heat sink (FRP-H.1) simulator exam DSS-002 for the steam break (EPP-014) scenario DSS-009.
- indicates exam measure was changed by exam team
Limits ana Controls (Part B) Examination Composition Questions Unique to the SRO Examination QUEST.
BOO-005 TOPIC PEPs PEP-101 CONTENT Classif the event TIME 2.2 B01-002 EOPs FRP-S.1 B02-011 / ADMIN/
Reporting
- B02-047
~
EOPs/EPP-005 B03-010 ~
EOPs/Bases SSOA'4Y:
SOP'P.:.:0 What if SI actuates in FRP-S.1 Reporting requirement for ESF component actuation CRDM fan availability and requir-ed subcooling margin in EPP-005 (ES-0.2)
Bases for minimum SG level in EPP-017 (ECA-3.1)
~ee,yjsecjfgc i,",'eoetm~~c.
0.6 1.3 1.6 0.6 B04-048 ~
EOPs/CSFSTs Determine critical safety func-tion status 1.3 B07>>050
'~ Tech.
Specs Bll-008 < Refueling Requirement to bar an EDG for testin with other EDG inoperable Reason for last four core loca-tions filled during refueling
1.3 12.5 B05-012 substituted for B05-058 due to substituting loss of secondary heat sink (FRP-H.1) simulator exam DSS-002 for the steam break (EPP-014) scenario DSS-009.
go5 - 4'('-
Qo7 c-ISR~
- indicates exam measure was changed by exam team
JOB PERFORMANCE MEASURES JPM CR-006
OST-1004 EOPs/Path-1 SYSTEMS/ELEC SYSTEMS/ESW/
AOP-022 EVOLUTION Perform a Dro ped Rod Recovery Perform Heat Balance Calculation Verify Reactor Trip and Turbine Tri - Turbine Did Not Tri Deenergize Power to the Se encers Supplying Air Compressors From ESW TIME
20 2.5 1.3 6.3
17. 1 IP-050 lp-118 SYSTEMS/ACP SYSTEMS/CVCS ADDITIONALJPMs Transfer control to the ACP Establish CSIP suction from the RWST
10 3.8 3.2
7.0
- - indicates exam measure was changed by exam team
EXAMS DSS-017 Alternate TOPIC AOPs TECH SPECS AOPs EOPs EOPs EOPs PEPs EVOLUTION
Tech.
Spec.
3.4.6.2 FRV auto failure (AOP-010)
300 gpm SGTR SG 1B Path-1 300 gpm SGTR SG 1B Path-2 SGTR with loss of pressure control EPP-022 Classify the event (SS-N only)
TIME
12
NM
3.2 1.6 2.5 NM 29.3 DSS-009 AOPs TECH SPECS AOPs EOPs EOPs EOPs EOPs EOPs PEPs NI-41 fuse failure (OWP-NI)
Tech.
Spec.
3.3.1 Loss instrument bus
(AOP-024)
ATWS (Manual trip successful)
Reactor tri Path-1 Reactor tri EPP-004 Steam break SG 1A (Path-1)
Steam break SG 1A (EPP-014)
Classify the event (SS-N onl
)
10
NM
3.35 1.0 3.35 0.3 1.7 1.7 4.0 2 '
NM 18.1 SUBSTITUTION FOR DSS-009 AS REQUESTED BY NRC EXAMS DSS-002 TOPIC
,AOPs/AOP-025 TECH SPECS AOPs/AOP-001 EOPs/PATH-1 (E-0)
EOPs/FRP-H. 1 EVOLUTION Loss of power to 6.9 kV emergency bus 1B-SB TS 3.8.1.1 and 3.8.3.1 Loop 1A control T. fails high SG 1A feed line break/reactor trip and safety injection/alternate high head injection Implements bleed and feed and restores one MFW pump TIME
10
20
4.8 1.6 3.2 3.2 6.4 19.2
- indicates exam measure was changed by exam team
1993 LOR E ATION WEEK 3 WRITTEN EXAM STATIC S ATOR (PART A~ EXAM
C5 SHEARON HARRIS NUCLEAR POWEk PLANT LZCENSED OPERATOR REQUALZFICATION EXAMBANK QUESTION NUMBER A02-005 REVZSZON:
SCENARIO:
i7OB CLASS!
POINT VALUE:
ESTIMATED TIME:
QUESTION:
01/02/03'4'4'6 REACTOR OPERATOR QUESTION 1.0
MINUTES What is the current status of the SG 1B main feedwater regulating valve (1FW-249)?
a.
The valve has been shut by a P-14 generated Feedwater Isolation signal.
b.
The valve is 1004 open in response to a SG level error.
c.
The valve is 1004 open in response to a steam flow/feed flow mismatch error.
d.
The valve has been shut by a Safety Injection generated Feedwater Isolation signal.
ANSWER d.
REFERENCE:
EOP-PATH-1 108D831,S13 SD-134 OMM-004 SD-103 PATH-1 CP&L SHNPP FUNCTIONAL DIAGRAMS, FEEDWATER CONTROL AND ISOLATION CONDENSATE AND FEEDWATER POST TRIP/SAFEGUARDS REVIEW REACTOR PROTECTION/ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
45
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER A02-052N REVISION
SCENARIO:
JOB CLASS:
POINT VALUE:
ESTIMATED TIME 01,
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTIONS What component is not in the proper condition for the sequencer actuation that has occurred?
a.
1B1 supply breaker b.
Containment Spray Pump 1B-SB c.
CSIP 1B-SB ANSWER d ~
REFERENCE 108D831,S8 SD-103 EOP-PATH-1 OMM-004 CP&L SHNPP FUNCTIONAL DIAGRAMS, SAFEGUARD ACTUATION SIGNALS REACTOR PROTECTION/ENGINEERED SAFETY FEATURES ACTUATION SYSTEM PATH-1 POST TRIP/SAFEGUARDS REVIEW
29
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
A02-070 REVISION:
SCENARIO:
J'OB CLASS POINT VALUE ESTIMATED TIME 011
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION Approximately how much water will be pumped out of the Refueling Water Storage Tank (RWST) between now and the time that automatic swapover to the containment recirculation sumps occurs?
a.
275,000 gallons b.
300,000 gallons c.
325,000 gallons d.
350,000 gallons ANSWER:
b.
[RWST at 984 (460,000 gal.)
RWST at 23.4~
(152,000 gal.)
= 308,000 gal.]
REFERENCE:
APP-ALB-004 CURVE BOOK SD-110 MAIN CONTROL BOARD OPERATIONS CURVE BOOK SAFETY INJECTION SYSTEM
D-1
SHEARON HARRIS NUCLEAR POWEk PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
A02-093 REVISION:
SCENARIO JOB CLASS POINT VALUE ESTIMATED TIME 01'3i 04i 16i 17'0 REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION What is the current status of Residual Heat Removal Pump A-SA?
a. It is recirculating to the refueling water storage tank.
b. It is recirculating back to its suction.
c. It is injecting into the RCS cold legs.
d. It is supplying the suction of CSIP 1A-SA.
ANSWER b.
REFERENCE SD-111 RESIDUAL HEAT REMOVAL SYSTEM
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER A03-056 REVISION
SCENARIO:
JOB CLASS:
POINT VALUE:
ESTIMATED TIME 01'4'7'0 REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION:
The TRIP HIGH TEMP JACKET WATER annunciator has just actuated on Emergency Diesel Generator (EDG) A-SA with a jacket water temperature of 200'F.
Considering current plant conditions, what is the expected response of EDG A-SA?
a.
The EDG will trip immediately.
b.
The EDG willtrip after a 10 second time delay.
c.
The EDG will trip when safety injection is reset.
d.
The EDG should continue to run after safety injection is reset.
ANSWER:
c ~
REFERENCE:
SD-155.01 APP-ALB-024 EMERGENCY DIESEL GENERATOR SYSTEM MAIN CONTROL BOARD
17
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATION EZAMBANK QUEST10N NUMBER A03-063 REVISION
SCENARIO JOB CLASS POINT VALUE:
ESTIMATED TIME>>
QUESTION
REACTOR OPERATOR QUESTION 1.0
MINUTES What action will cause pressurizer heater group 1C to energize?
a.
Position the control switch to OFF then back to ON.
b.
Restore pressurizer level to greater than 174 and reenergize the electrical bus that supplies power to group 1C.
c.
Restore pressurizer level to greater than 174, control switch operation is not required.
d.
Restore pressurizer level to greater than 174 and cycle the control switch to OFF then to ON.
ANSWER d.
REFERENCE:
SD-100.03 108D831,S11 APP-ALB-009 PRESSURIZER AND CONTROLS CP&L SHNPP FUNCTIONAL DIAGRAMS, PRZ PRESSURE LEVEL CONTROL MAIN CONTROL BOARD
36
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER A03-081 REVISION
SCENARIO:
JOB CLASS POINT VALUE:
ESTIMATED TIME
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION:
With current plant conditions, how is the charging/safety injection pump (CSIP)
1A-SA alternate mini-flow valve (1CS-752)
operating?
a.
The valve has received an open signal from safety injection (SI).
b.
The valve will open if RCS pressure exceeds 2300 psig.
c ~
The has closed in response to containment spray actuation with RCS pressure less than 1750 psig.
The valve will only respond to demands from its control switch.
ANSWER:
b.
REFERENCE:
SHEARON HARR1S NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICAT1ON EXAMBANK QUESTION NUMBER A05-010 REVISION:
SCENARIO:
JOB CLASS:
POINT VALUE ESTIMATED TIME Olg 02'3'5'6'1'3'6'7~
19/
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION:
Which one of the following is the current pressure setpoint for an automatic opening of the "A" SG PORV?
a.
1079 psig b.
1105 psig c.
1131 psig d.
1157 psig ANSWER b.
REFERENCE:
GP-005 SD-126.01 POWER OPERATION (MODE 2 TO MODE 1)
MAIN STEAM AND STEAM DUMP SYSTEM
SHEARON HARRZS NUCLEAR PORER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTZON NUMBER A05-025N REVZSZON:
SCENARIO:
JOB CLASS:
POINT VALUE:
ESTIMATED TIME:
QUESTION:
01 g 02'4
~
REACTOR OPERATOR QUESTION 1.0
MINUTES With current plant conditions, how are the Group 1 steam dump valves operating?
a.
They are blocked from operation by the low-low T-avg interlock.
b.
They are blocked from operation due to loss of condenser vacuum.
c.
They are armed and are being controlled by the turbine trip controller.
d.
They are armed and are being controlled by the load rejection controller.
ANSWER a ~
REFERENCE:
108D803,S18 SD-126.01 EOP-PATH-1 PROCESS CONTROL BLOCK DIAGRAM, STEAM DUMP MAIN STEAM AND STEAM DUMP SYSTEM PATH-1 13-15
sHEARON HARRZS NUCLEAR POWE+ PLANT LICENSED OPERATOR REQUALIFZCATZON EXAMBANK QUESTION NUMBER A07-001 REVZSZON:
SCENARIO
JOB CLASS ~
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME QUESTZON0 1.0
MINUTES How would Emergency Diesel Generator lA-SA respond if the SAFETY INJECTION TRAIN A RESET switch was positioned to RESET and then the DIESEL GENERATOR A-SA STOP/START switch was positioned to STOP at this time?
a. It would stop and could not be restarted.
b. It would stop and then immediately restart.
c. It would stop and could be re-started from the Main Control Panel.
d. It would continue to run.
ANSWER d.
REFERENCE SD-155.01 SD-103 EMERGENCY DIESEL GENERATOR SYSTEM REACTOR PROTECTION/ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATION EXAMBANK QUESTION NUMBER:
A07-033 REVISION:
SCENARIO:
JOB CLASSi POINT VALUE ESTIMATED TIME:
QUESTION:
01'2'4'6 REACTOR OPERATOR QUESTION 1.0
MINUTES What is primarily responsible for the current status of the Emergency Service Water pumps?
a.
Trip of Normal Service Water Pump 1B.
b.
Startup of the Emergency Diesel Generators.
c.
Safety Injection actuation.
d.
Low pressure in the "B" emergency service water header.
ANSWER c ~
REFERENCE:
SD-139 108D831,S8 OMM-004 SERVICE WATER SYSTEM CP&L SHNPP FUNCTIONAL DIAGRAMS, SAFEGUARD ACTUATION SIGNALS POST TRIP/SAFEGUARDS REVIEW
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER A09-013 REVZSZON:
SCENARIO JOB CLASS POINT VALUE ESTIMATED TIME:
QUESTION
REACTOR OPERATOR QUESTION 1.0
MINUTES If the scenario were allowed to continue, which one of the following describes how the Source Range channels would respond?
(Assume no operator action)
a.
Neither Source Range channel will automatically reenergize b.
Both Source Range channels will automatically reenergize c.
Only NI-32 will automatically reenergize d.
Only NI-31 will automatically reenergize ANSWER a 0 REFERENCE:
108D831i S4 SD-105 EOP-PATH-1 CP&L SHNPP FUNCTIONAL DIAGRAMSg NI PERMISSIVES AND BLOCKS EXCORE NUCLEAR INSTRUMENTATION PATH-1
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER:
A09-045 REVZSION:
SCENARIO JOB CLASS POINT VALUE:
ESTIMATED TIME QUESTION:
01, 04,
REACTOR OPERATOR QUESTION 1.0
MINUTES What is the current status of the AMSAC System?
a.
Neither ARMED or TRIPPED b.
ARMED and TRIPPED c.
TRIPPED but not ARMED d.
ARMED but not TRIPPED ANSWER d.
REFERENCE:
SD-103.01 ANTICIPATED TRANSCIENT WITHOUT SCRAM MITIGATING SYSTEM ACTUATION CIRCUITRY AMSAC
THE FOLLOYVI1VGQUESTIONS DO NOT REFER TO PRESENT PLANT CONDITIONS
SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER:
AOl-018 REVISION
SCENARIO:
JOB CLASS POINT VALUE:
ESTIMATED TIME:
QUESTION~
REACTOR OPERATOR QUESTION 1.0
MINUTES THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS The following conditions exist:
The plant is at 804 power A dropped rod in group 2 of control bank "D" has occurred A recovery of the dropped rod has begun and the
"ROD CONTROL URGENT ALARM" annunciator has been received Select the power cabinet that is the source of the
"ROD CONTROL URGENT ALARM" annunciator.
a 0 1BD b.
2BD c
1AC d.
2AC ANSWER:
a 0 REFERENCE:
SD-104 ROD CONTROL SYSTEM 24-25
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIF1CAT1ON EXAMBANK QUESTION NUMBER A01-023 REVISION:
SCENARIO:
JOB CLASS POINT VALUE.
ESTIMATED TIME REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS
'I The plant is operating at 1004 power with rod control in manual when the TAVG TREF DEVIATION and HIGH LOOP TAVG annunciators actuate.
A slow increase in Tavg is observed on the chart recorder, but there have been no blender operations or changes in turbine load since the beginning of shift 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ago.
Which one of the following conditions could be causing the increasing Tavg?
a.
TCV-144 (CCW from letdown heat exchanger flow control valve) is found to be failed full open.
b.
LCV-115A (letdown to VCT/RHT divert valve) is found to be fully diverting to the recycle holdup tank.
c.
A main steam line safety valve on SG lA is found to be leaking significantly.
d.
LCV-115D (RWST to charging pump suction) is found to be leaking.
ANSWER.
a ~
REFERENCE:
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER A03-082 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE.
ESTIMATED TIME 1.0
MINUTES QUESTION THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS Approximately 20 minutes ago the plant experienced a reactor trip and safety injection from 1004 power.
The following conditions exist:
Pressurizer level RCS pressure RCS temperature SPDS subcooing margin Containment pressure SG pressures 04 (off-scale low)
1015 psig (slowly decreasing)
548'F (slowly decreasing)
0 F (stable)
9 psig 1010 psig (stable)
With these conditions, which of the following actions will MOST EFFECTIVELY reduce RCS pressure?
a.
Initiate pressurizer auxiliary spray.
b.
Open a pressurizer PORV.
c.
Dump steam from the steam generators.
d.
Reduce safety injection flow.
ANSWER:
c ~
REFERENCE:
EOP-EPP-009 POST LOCA COOLDOWN AND DEPRESSURIZATION
iiEh SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATION EXAMBANK QUESTION NUMBER:
A04-020 SCENARIO REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE EST1MATED TIME QUESTION:
1.0
MINUTES THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS Which one of the following describes how the RHR System would be affected during a loss of instrument air while RHR is operating in mid-loop conditions?
a.
RHR flow will remain the same.
b.
RHR flow would increase to greater than the maximum limit.
c.
RHR flow would decrease, but remain above the value required to prevent boiling in the core d.
RHR flow would decrease to the point where the pump is running at shut-off head ANSWER:
b.
REFERENCE:
AOP-017 LOSS OF INSTRUMENT AIR
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFICATION EXAMBANK QUESTION NUMBER A05-008 SCENARIO~
REVZSZON:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION:
THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS After a transfer to the Auxiliary Control Panel (ACP) following a fire in the Main Control Room, which one of the following describes the effect on the Auxiliary Feedwater (AFW) system?
a.
The automatic start signals for the turbine-driven (TD)
and motor-driven (MD) AFW pumps are blocked.
b.
Both the TD AFW and MD AFW pumps retain their auto start signals.
c.
Only the auto start for the MD AFW pump is blocked.
d.
Only the auto start for the TD AFW pump is blocked.
ANSWER:
d.
REFERENCE:
AOP-004 SAFE SHUTDOWN IN CASE OF FIRE OR CONTROL
ROOM INACCESSIBILITY
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER:
A06-009 REVISION:
SCENARIO:
iYOB CLASS POINT VALUE:
ESTIMATED TIME REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION:
THIS QUESTION DOES NOT REFER TO PRESENT PLANT CONDITIONS.
Assume a large break LOCA has occurred and containment spray has actuated automatically.
How will the containment spray additive tank isolation valves (1CT-11 and 1CT-12)
be shut during this accident?
a.
They must be manually shut by the operators b.
They will shut automatically when Containment Spray is reset c.
They will shut automatically on receipt of the SPRAY ADDITIVE TANK EMPTY alarm d.
They will shut automatically when the Spray Addition Timer has timed out ANSWER:
c ~
REFERENCE.
SD-112 APP-ALB-001 CONTAINMENT SPRAY SYSTEM MAIN CONTROL BOARD
27
RO PART B LIMITSAND CONTROI,S WRITTEN EXAM
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EZAMBANK QUESTION NUMBER:
BOO-011 REVISION
JOB CLASS.
POINT VALUE ESTIMATED TIME~
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION While performing a manual dose calculation, which of the following should be used for wind speed and direction?
(Assume these are the only values available)
a.
ERFIS indicates 18 mph from 1854 b.
National Weather Service at the Raleigh-Durham airport reports 20 mph from 1804 c.
Corporate Weather Center reports 15 mph from 175 d.
Wind sock indicates 9 mph from 1804 ANSWER a 0 REFERENCE:
PEP-341 MANUAL DOSE CALCULATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER:
BOO-032 REVISION:
JOB CLASS!
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION:
You have just been directed in your capacity as Emergency Communicator to contact SHNPP emergency response personnel for an ALERT.
What emergency response facilities should you expect to be activated for this condition?
a.
TSC only b.
b.
REFERENCE PEP-321 NOTIFICATION OF SHNPP EMERGENCY RESPONSE
PERSONNEL
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER BOO-035 REVISION:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTIONS Security reports a fire in the Met tower shack.
Which one of the following off-site agencies should be requested for assistance?
a.
Holly Springs Fire Department b.
Apex Fire Department c.
Chatham County Fire and Rescue Department d.
Fuquay-Varina Fire Department ANSWER:
b.
REFERENCE:
PEP-301 NOTIFICATION OF NON-CP&L EMERGENCY RESPONSE ORGANIZATIONS
SHEARON HARRZS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER:
B01-001 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION:
While operating at 1004 power a pressurizer PORV spuriously opens and cannot be closed.
The pressure decrease is finally stopped when the RO closes the PORV's block valve.
Two over-temperature delta-T runbacks have occurred and the following conditions currently exist:
RCS T-avg Pressurizer pressure Pressurizer level Reactor power 5864F, stable 1945 psig, stable 56%
964'hat restrictions apply to continued operation?
a.
Power operation may continue for a maximum of 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> per Technical Specification 3.0.3.
b.
Power operation may continue, but pressurizer pressure must be restored to ~ 2185 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
c.
Power operation may continue, but power level may not be increased until pressurizer pressure is restored to normal.
d.
The reactor must be tripped immediately.
ANSWER d.
REFERENCE:
AOP-019 SD-100.03 SD-103 MALFUNCTION OF RCS PRESSURE CONTROL PRESSURIZER AND CONTROLS REACTOR PROTECTION/ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER B01-003 REVISION:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1.0
MINUTES QUESTION:
With the plant initially at 1004 power, Reactor Coolant Pump (RCP)
1A tripped, but the reactor trip breakers have failed to open and the control rods will not insert.
The Control Operator has just reported that neither boric acid transfer pump will start.
How should the reactor be brought subcritical?
a.
Initiate safety injection.
b.
Open CSIP suction from RWST valve LCV-11SB (1CS-291),
close VCT outlet valve LCV-115E (1CS-166),
and verify at least 30 gpm flow indicated on FI-122.
c.
Open CSIP suction from RWST valve LCV-115D (1CS-292),
close VCT outlet valve LCV-115C (1CS-165),
and verify at least 90 gpm flow indicated on FI-122.
d.
Open emergency boric acid addition valve 1CS-278.
ANSWER c ~
REFERENCE EOP-FRP-S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICAT1ON EZAMBANK QUESTION NUMBER B01-010 REV1SION:
JOB CLASS:
POINT VALUE ESTIMATED TZME:
REACTOR OPERATOR QUESTION 1 '
MINUTES QUESTION A reactor startup with MOL conditions (206 EFPD) is in progress.
The ECP predicts criticality at 120 steps on Control Bank D with an RCS boron concentration of 700 ppm.
The Reactor Operator declares the reactor critical at 26 steps on Control Bank D.
What action should the operating shift take?
a.
Perform a normal reactor shutdown using GP-006 b.
Emergency borate per AOP-002.
c.
Continue the startup without a reactivity management analysis since this situation is within acceptable limits.
d.
Continue the startup, but have a reactivity management analysis performed.,
ANSWER:
a 0 REFERENCE GP-004 CURVE BOOK PLP-106 REACTOR STARTUP (MODE 3 TO MODE 2)
OPERATIONS CURVE BOOK
TECHNICAL SPECIFICATION EQUIPMENT LIST
PROGRAM AND CORE OPERATING LIMITS REPORT
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
B01-023 REVISION
JOB CLASS:
POINT VALUE ESTIMATED TIME:
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION The following plant conditions exist:
Core age is 250 EFPD Reactor power has been at 100% power for the past 3 days RCS boron concentration is 700 ppm Bank "D" rods are at 138 steps Boric acid tank boron concentration is 7050 ppm It is desired to borate the Bank "D" rods out to 200 steps.
Which one of the following values is the amount of boric acid required for this maneuver?
a ~
b.
C ~
d.
250 gallons 280 gallons 310 gallons 340 gallons ANSWER:
b.
REFERENCE OP-107 CURVE BOOK CHEMICAL AND VOLUME CONTROL SYSTEM OPERATIONS CURVE BOOK
SHEARON HARRIS NUCLEAR POSER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
B01-052 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE.
ESTIMATED TIME:
1.0
MINUTES QUESTION The following plant conditions exist:
A reactor trip has occurred from full power at 52 EFPD
The SUBCRITICALITYCritcal Safety Function is GREEN
4 control rods have failed to fully insert By procedure, what quantity of boric acid must be added to provide adequate shutdown margin for this condition, assuming the plant will be maintained in HOT STANDBY?
a.
5,800 gallons b.
6,200 gallons c ~
7,240 gallons d.
103,600 gallons ANSWER co REFERENCE:
EOP-EPP-004 AOP-002 REACTOR TRIP RESPONSE EMERGENCY BORATION
SHEARON HARRIS NUCLEAR PO'HER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER:
B02-006 REVZSION
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIMEi QUESTION 1.0
MINUTES The operating shift is responding to a reactor trip and safety injection using Path-1.
When the SRO directs the RO to "VERIFY ALL CSIPs AND RHR PUMPS RUNNING" the RO notices that only CSIP "A" is running.
What should the RO do?
a.
Inform the SRO that "B" CSIP is not running, but do not attempt to start the pump until directed to do so by the SRO.
b.
Do not attempt to start CSIP "B" unless injection flow rate is less than 200 gpm.
c.
Inform the SRO that CSIP "B" did not start and attempt to start the pump any time after load block 1.
d.
Inform the SRO that CSIP "B" did not start and attempt to start the pump any time after load block 9.
ANSWER:
d.
REFERENCE:
EOP-PATH-1 EOP USERS GUIDE PATH-1 EOP USERS GUIDE
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATZON EZAMBANK QUESTION NUMBER B02-013 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1 ~ 0
MINUTES QUESTION:
A reactor trip and safety injection occurred from 1004 power due to RCS leakage in excess of makeup capability.
The operating shift has secured one charging/safety injection pump (CSIP)
and has just completed realigning CSIP discharge from the BIT to the normal charging line per Step 6 EPP-008 (SI Termination).
Before normal letdown can be established, it is observed that pressurizer level is 104 and decreasing with 150 gpm indicated charging flow.
Which of the following should the operating shift do first in accordance with plant procedures?
a.
Start the second CSIP.
b.
Realign CSIP discharge from the normal charging line back to the BIT.
c.
Align the alternate charging path.
d.
Actuate safety injection.
ANSWER:
b.
REFERENCE:
EOP-EPP-008 EOP USERS GUIDE SI TERMINATION EOP USERS GUIDE
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER B03-024 REVISION:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME:
1 '
MINUTES QUESTION~
The operating shift is performing step 18 of EPP-020 (SGTR With Loss of Reactor Coolant:
Subcooled Recovery).
The following conditions exist:
RCP 1B running
.
2 CSIPs running
RHR "A" running
BIT valves are open
RWST level = 454 RCS hot leg temp.
SPDS RCS subcooling Containment pressure Pressurizer level 385oF 40oF 2 '
psig 524 What should the operating shift do?
a.
Leave CSIPs running and go to step 16.
b.
Leave CSIPs running and go to step 29.
c.
Stop one CSIP and go to step 19.
d.
Leave CSIPs running and continue depressurizing the RCS to inject the accumulators.
ANSWER c ~
REFERENCE:
EOP-EPP-020 SGTR WITH LOSS OF REACTOR COOLANT:
SUBCOOLED RECOVERY
hs, SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER B03-025 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION:
FRP-P.1 has just been completed and an RCS temperature soak has just been started.
What action below could be performed by the operating shift with this soak in progress?
a.
Increase AFW flow above minimum to increase steam generator levels.
b.
Re-establish steam generator blowdown.
c.
Stop a residual heat removal pump.
d.
Energize pressurizer heaters to increase subcooling margin above the required minimum.
ANSWER.
ci REFERENCE EOP-FRP-P.1 RESPONSE TO IMMINENT PRESSURIZED THERMAL
SHOCK
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZF1CATZON EZAMBANK QUESTZON NUMBER B03-028
'REVISION
iTOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME QUESTION:
1.0
MINUTES EPP-010 (Transfer to Cold Leg Recirculation)
has just been entered from EPP-009 (Post LOCA Cooldown and Depressurization)
three hours following a small break loss of coolant accident.
The following conditions exist:
SPDS status:
RED on HEAT RWST level RCS pressure RCS temperature SPDS subcooling margin =
SINK 204 400 psig 395 oF 530F Containment pressure
= 5 psig Pressurizer level 55~o SG pressures
=
100 psig CSIPs
"A" running RHR pumps secured What should the operating shift do?
a.
Go to FRP-H.1.
b.
Immediately restart CSIP "B" and open both pressurizer PORVs.
c.
Restart CSIP "B" and return to EPP-009.
d.
Continue EPP-010 until the cold leg recirculation lineup is established.
ANSWER:
d.
REFERENCE EOP-EPP-010 TRANSFER TO COLD LEG RECIRCULATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFICATZON EZAMBANK QUESTION NUMBER:
B03-122 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME QUESTION:
1.0
MINUTES The EOP network has been entered due to a reactor trip and safety injection initiation.
Due to a RED condition on the CONTAINMENT critical safety function, FRP-J.1 (Response to High Containment Pressure)
has been implemented.
While performing FRP-J.1, the STA points out that a MAGENTA condition now exists on the RCS INTEGRITY critical safety function.
What action should the operating shift take?
a.
Immediately exit FRP-J.1 and go to FRP-P.1.
b.
Simultaneously implement both FRP-J.1 and FRP-P.1.
c.
Continue with FRP-J.1 until directed to exit.
d.
Suspend FRP-J.1 and return to PATH-1 for guidance.
ANSWER:
C ~
REFERENCE EOP USERS GUIDE EOP-CSFST EOP USERS GUIDE CRITICAL SAFETY FUNCTION STAUS TREE
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER.
B04-024 REVISION:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION With the plant at 1004 power, 1CC-299 (RCP bearing oil coolers returns) is closed by mistake by an Auxiliary Operator performing a
CCW lineup in the RAB and cannot be reopened from the Control Room or by local action.
With BEARING CCW LOW FLOW alarms actuated and RCP motor bearing temperatures increasing, what actions must the operating shift take?
a.
Within 10 minutes of the loss of CCW flow to the RCP bearing oil coolers, trip the reactor then secure the RCPs.
b.
Trip the reactor, then secure the RCPs only if bearing temperature limits are exceeded; there is no time limit to trip the RCPs.
c.
Trip the reactor then secure the RCPs 10 minutes after bearing temperature limits are exceeded.
d.
Commence a plant shutdown at 54/minute and secure RCPs immediately after the reactor is shutdown.
ANSWER a ~
REFERENCE AOP-018 REACTOR COOLANT PUMP ABNORMAL CONDITIONS
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFICATZON EXAMBANK QUESTION NUMBER.
B04-030 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME>>
QUESTION>>
1.0
MINUTES RHR Train "A" was being placed in service for an RCS cooldown.
The following events have occured:
At 18:55, the "A" RHR pump was started for sampling but it immediately tripped At 18:58, the "A" RHR pump was started and a sample was obtained At 19:10, the "A" RHR pump was secured Which one of the following is the earliest time another start of the "A" RHR pump can be attempted?
a>>
b.
c d.
Immediately 19:25 19:40 19:55 ANSWER:
d.
REFERENCE:
OP-111 RESIDUAL HEAT REMOVAL SYSTEM
SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER B05-012 REVISION
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION A reactor trip from 1004 power has just occurred.
As a result of the trip a Main Steam Safety valve on the "A" SG opened and has stuck fully open.
During recovery actions that follow, the feedwater flow to the
"A" SG should be:
a.
Controlled to maintain a minimum level of 10%.
b.
Controlled to maintain a minimum level of 404.
c.
Set and maintained at 12.5 kpph.
d.
Isolated.
ANSWER:
d.
REFERENCE:
EOP-EPP-014 FAULTED STEAM GENERATOR ISOLATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER:
B05-020 REVZSZON:
JOB CLASS:
)
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIMEi 1.0
MINUTES QUESTION The operating shift has just reached Entry Point C of Path-1 following a reactor trip and safety injection from 1004 power.
The following conditions exist:
CSIP status RCP status RCS pressure Power range NIS Core Exist TCs temperature SG A NR level SG B and SG C NR level Containment pressure RWST level CST level both running secured by the operators 1155 psig
564oF 154 54'.5 psig 904 7%'hat action should the operating shift perform next?
a.
Align service water to the AFW pumps.
b.
Implement FRP-H.5.
c.
Implement FRP-C.3.
d.
Align CCW to sample heat exchangers and gross failed.
fuel detector.
ANSWER a ~
REFERENCE:
EOP-PATH-1 PATH-1 PATH-1 USERS GUIDE PATH-1 USERS GUIDE FOLDOUT A
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
B07-002 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIMEi QUESTION 1.0
MINUTES With the plant initially at 504 power, the operating shift has opened the Main Generator output breakers (52-7 and 52-9) to separate from the grid due to an unstable grid frequency condition.
With the plant stabilized at 54 power awaiting directions from the load dispatcher, the Main Turbine trips.
What action should the operating shift take?
a.
Open breakers 105 and 125 to initiate automatic starting and loading of the emergency diesel generators (EDGs)
on the 6.9 kV emergency buses b.
Take no actions the undervoltage condition that will develop on the 6.9 kV emergency buses will automatically cause breakers 105 and 125 to open and the EDGs to automatically start and load.
c.
Manually transfer the 6.9 kV emergency buses to the startup transformers d.
Allow the 6.9 kV emergency buses to automatically transfer to the startup transformers ANSWER:
a ~
REFERENCE:
AOP-015 SECONDARY LOAD REJECTION
SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALIFZCATZON EXAMBANK QUESTION NUMBER:
B07-005 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE 1.0 ESTIMATED TIME
MINUTES QUESTION:
Following a reactor trip from 1004 power, all power is lost to the switchyard and both emergency diesel generators fail to start.
The operating shift has implemented EOP-EPP-001 (Loss of AC Power to 1A-SA and 1B-SB Buses)
and are performing Step 21 to rapidly cooldown and depressurize the RCS to prevent RCP seal damage.
What condition would require termination of this cooldown?
a.
RVLIS full range level less than 394 b.
RCS cooldown rate in excess of 100 F/hour c.
Pressurizer level off-scale low d.
Source Range startup rate of + 0.1 DPM ANSWER+
d.
REFERENCE EOP-EPP-001 LOSS OF AC POWER TO 1A-SA AND 1B-SB BUSSES
SHEARON HARRIS NUCLEAR POSER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER:
B07-041 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME:
QUESTION~
1.0
MINUTES How will a natural circulation cooldown using EPP-005 (Natural Circulation Cooldown)
be affected if the control rod drive mechanism (CRDM) fans are not.available?
a.
There will be no affect because the amount of RCS heat removed by running the CRDM fans is insignificant compared with that removed by steaming the steam generators.
b.
Transition to EPP-006 will be required because the cooldown and depressurization will cause the formation of a steam void in the reactor vessel head.
c.
The cooldown must be stopped for a one hour hold period after each 50 degrees of cooldown.
d.
Greater subcooling must be maintained during the cooldown and depressurization that follow.
ANSWER:
d.
REFERENCE:
EOP-EPP-005 NATURAL CIRCULATION COOLDOWN
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER B08-001 REVISION
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION~
A control room evacuation has occurred and control has been transferred to the Auxiliary Control Panel.
A plant cooldown is being commenced with the following initial conditions:
Primary temperature
= 530'F Pressurizer level = 30%
Primary pressure
= 1900 psig Steam generator pressure
= 1,000 psig What is the minimum steam generator levels that the operators should maintain with these conditions?
a.
104 using narrow range level indication b.
454 using narrow range level indication c.
504 using wide range level indication d.
60% using wide range level indication ANSWER c ~
REFERENCE AOP-036 AOP-004 SAFE SHUTDOWN FOLLOWING A MAJOR FIRE SAFE SHUTDOWN IN CASE OF FIRE OR CONTROL ROOM INACCESSIBILITY
SHEARON HARRZS NUCLEAR POWEk PLANT LICENSED OPERATOR REQUALZFICATION EXAMBANK QUESTZON NUMBER B08-002 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1.0
MINUTES QUESTION~
The plant is operating at 1004 power, when the INSTRUMENT AIR LOW PRESS annunciator is received.
Which one of the following describes a condition that requires a
turbine/reactor trip?
a.
All SG NR levels at 614 and decreasing b.
All SG NR levels at 454 and decreasing c.
IA pressure at 85 psig and decreasing d.
IA pressure at 66 psig and decreasing ANSWER:
b.
REFERENCE:
AOP-017 LOSS OF INSTRUMENT AIR
SHEARON HARRIS NUCLEAR POREk PLANT LZCENSED OPERATOR REQUALIFZCATZON EXAMBANK QUESTION NUMBER:
B09-010 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME:
1.0
MINUTES QUESTION:
During refueling operations (fuel is being moved) the audio count rate in containment stops functioning while the Control Room indication and audio count rate remain operable.
What actions should the operating shift take?
a.
Refueling operation (fuel movement)
may continue provided an audio countrate is maintained available in the Control Room and a dedicated operator is assigned to monitor it.
b.
Verify the boron concentration in the RCS and refueling canal is greater than 2000 ppm once per shift.
c.
Immediately suspend all operations involving core alterations or positive reactivity changes.
d.
Immediately stop movement of irradiated fuel and evacuate containment until the audio countrate is returned to an operable status.
ANSWER:
c ~
REFERENCE TECH.
SPECS.
3.9.2 INSTRUMENTATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
B11-006 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME QUESTION 1.0
MINUTES What is the annual limit for CP&L whole body dose for CP&L employees whose non-CP&L dose for the current year has been determined?
a.
5.0 Rem TEDE CP&L dose ANSWER b.
REFERENCE:
PLP-511 RADIATION CONTROL AND PROTECTION PROGRAM
SRO PART B LIMITSAND CONTROLS TTEN EXAM
SHEARON HARRIS NUCLEAR POWER PLANT L1CENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER:
BOO-005 REVISION
JOB CLASS:
SENIOR REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME:
QUESTION.
1.0
MINUTES The plant had been operating at a steady state power level of 504 with all conditions at equilibrium when Control Bank D began to slowly insert to maintain T-avg at T-ref.
AOP-003 (Malfunction of Reactor Makeup Control)
was implemented and an inadvertent dilution path found and isolated.
Control Bank D is at 156 steps and the total time of the dilution was 23 minutes.
What is the highest Emergecny Action Level (EAL) classification for this event, if any?
'a ~
b.
c ~
d.
No emergency action level reached Unusual Event Alert Site Emergency ANSWER c ~
REFERENCE PEP-101 EMERGENCY CLASSIFICATION AND INITIAL EMERGENCY ACTIONS AT SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER:
BOO-011 REVISION
JOB CLASS POINT VALUE ESTIMATED TIME QUESTION:
REACTOR OPERATOR QUESTION 1.0
MINUTES While performing a manual dose calculation, which of the following should be used for wind speed and direction?
(Assume these are the only values available)
a.
ERFIS indicates 18 mph from 185 b.
National Weather Service at the Raleigh-Durham airport reports 20 mph from 180'.
Corporate Weather Center reports 15 mph from 175 d.
Wind sock indicates 9 mph from 180 ANSWER:
a ~
REFERENCE PEP-341 MANUAL DOSE CALCULATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER:
BOO-035 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME:
QUESTION:
1.0
MINUTES Security reports a fire in the Met tower shack.
Which one of the following off-site agencies should be requested for assistance?
a.
Holly Springs Fire Department b.
Apex Fire Department c.
Chatham County Fire and Rescue Department d.
Fuquay-Varina Fire Department ANSWER b.
REFERENCE:
PEP-301 NOTIFICATION OF NON-CP&L EMERGENCY RESPONSE ORGANIZATIONS
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFICATZON EXAMBANK QUESTION NUMBER:
B01-001 REVISION
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME
0
MINUTES QUESTION While operating at 1004 power a pressurizer PORV spuriously opens and cannot be closed.
The pressure decrease is finally stopped when the RO closes the PORV's block valve.
Two over-temperature delta-T runbacks have occurred and the following conditions currently exist:
RCS T-avg Pressurizer pressure Pressurizer level Reactor power 586 F, stable 1945 psig, stable 564 964 What restrictions apply to continued operation?
a.
Power operation may continue for a maximum of 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> per Technical Specification 3.0.3.
b.
c ~
Power operation may continue, but pressurizer pressure must be restored to > 2185 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Power operation may continue, but power level may not be increased until pressurizer pressure is restored to normal.
d.
The reactor must be tripped immediately.
ANSWER:
d.
REFERENCE AOP-019 SD-100.03 SD-103 MALFUNCTION OF RCS PRESSURE CONTROL PRESSURIZER AND CONTROLS REACTOR PROTECTION/ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
.c~.
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER B01-002 REVISION:
JOB CLASS:
SENIOR REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIMEi 1.0
MINUTES QUESTION:
The operating shift is performing the step of EOP-FRP-S.1 (Response To Nuclear Power Generation/ATWS) to verify all dilution paths isolated when Safety Injection actuates.
How should the operating shift respond?
a.
Continue actions of FRP-S.1 without delay; do not verify proper operation of safeguards equipment until directed to do so following return to Path-1 b.
Verify proper operation of safeguards equipment while continuing with FRP-ST 1 c.
Go to PATH-1 and verify all automatic actions required for an SI have occurred, then return to the FRP-S.1 step in effect d.
Go to Path-1 and verify all automatic actions required for an SI have occurred, do not return to FRP-S.1 unless a
RED condition still exists on the CRITICALITYCritical Safety Function Status Tree.
ANSWER.
b.
REFERENCE:
EOP-FRP-.S.1 RESPONSE TO NUCLEAR POWER GENERATION/ATWS
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EZAMBANK QUESTION NUMBER B01-010 REVISION:
i7OB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME.
1.0
MINUTES QUESTION A reactor startup with MOL conditions (206 EFPD) is in progress.
The ECP predicts criticality at 120 steps on Control Bank D with an RCS boron concentration of 700 ppm.
The Reactor Operator declares the reactor critical at 26 steps on Control Bank D.
What action should the operating shift take?
a.
Perform a normal reactor shutdown using GP-006 b.
Emergency borate per AOP-002.
c.
Continue the startup without a reactivity management analysis since this situation is within acceptable limits.
d.
Continue the startup, but have a reactivity management analysis'erformed.,
ANSWER:
a ~
REFERENCE:
GP-004 CURVE BOOK PLP-106 REACTOR STARTUP (MODE 3 TO MODE 2)
OPERATIONS CURVE BOOK
TECHNICAL SPECIFICATION EQUIPMENT LIST
PROGRAM AND CORE OPERATING LIMITS REPORT
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATION EXAMBANK QUESTION NUMBER:
B01-,052 REVISION
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE!
ESTIMATED TIME~
1 '
MINUTES QUESTION:
'he following plant conditions exist:
A reactor trip has occurred from full power at 52 EFPD
The SUBCRITICALITY Critcal Safety Function is GREEN
4 control rods have failed to"'ully insert By procedure, what quantity of boric acid must be added to provide adequate shutdown margin for this condition, assuming the plant will be maintained in HOT STANDBY?
a ~
b.
5,800 gallons 6,200 gallons c.
7,240 gallons d.
103,600 gallons
ANSWER:
Co REFERENCE:
EOP-EPP-004 AOP-002 REACTOR TRIP RESPONSE EMERGENCY BORATION
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFICATZON EXAMBANK REVISION
QUESTION NUMBER B02-011 JOB CLASS SENIOR REACTOR OPERATOR QUESTION POINT VALUE 1.0 ESTIMATED TIME
MINUTES QUESTION Due to a failure of an SSPS relay, a residual heat removal pump automatically starts with the plant at 1004 power.
HOW SHOULD THIS EVENT BE REPORTED'
a.
No report required b.
One hour non-emergency report c.
Four hour report d.
Unusual Event ANSWER c.
[ESF actuation per 10CFR50.72(b)(2)(ii)]
REFERENCE:
AP-615 NRC REPORTING REQUIREMENTS
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATZON EXAMBANK QUESTZON NUMBER B02-047 REVISION
JOB CLASSY SENIOR REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1 ~ 0
MINUTES QUESTION:
A natural circulation cooldown is in progress using EOP-EPP-005 (Natural Circulation Cooldown) with the following plant conditions:
Pressurizer pressure Containment pressure
Containment radiation 540oF 1950 psig 0.1 psig normal While performing the RCS depressurization of Step 19, one of the four CRDM cooling fans trips on overload and cannot be restarted.
What action should the operating shift take?
a.
Stop the RCS depressurization and stabilize RCS temperature until the CRDM fan is repaired.
b.
C ~
Go to EPP-006 or EPP-007 based on RVLIS availability.
Stop the RCS depressurization until subcooling can be increased to at least 1104F.
d.
Continue RCS depressurization while maintaining at least 60'F subcooling margin.
ANSWER:
d.
REFERENCE:
EOP-EPP-005 NATURAL CIRCULATION COOLDOWN
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER B03-010 REVISION:
JOB CLASS SENIOR REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1.0
MINUTES QUESTION:
'
In EPP-017 (Post-SGTR Cooldown Using Backfill) why is the ruptured steam generater level controlled by alternately draining to 454 and then refilling to 754?
a.
Decrease RCS dilution.
b.'ore effectively cool the ruptured steam generator while avoiding automatic reinitiation of auxiliary feedwater.
ce d.
Minimize thermal shock to steam generator components.
Minimize the amount of steam generator contaminants and debris from entering the RCS.
ANSWER:
b.
REFERENCE:
EOP USERS GUIDE EOP-EPP-017 EOP USERS GUIDE POST-SGTR COOLDOWN USING BACKFILL
19
s)~
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTZON NUMBER:
B03-024 REVZSZON:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION:
The operating shift is performing step 18 of EPP-020 (SGTR With Loss of Reactor Coolant:
Subcooled Recovery).
The following conditions exist:
RCP 1B running 2 CSIPs running RHR "A" running BIT valves are open RWST level = 454 RCS hot leg temp.
SPDS RCS subcooling Containment pressure Pressurizer level 385oF
F 2 '
psig
= 524 What should the operating shift do?
a.
Leave CSIPs running and go to step 16.
b.
Leave CSIPs running and go to step 29.
c.
Stop one CSIP and go to step 19.
d.
Leave CSIPs running and continue depressurizing the RCS to inject the accumulators.
ANSWER:
C ~
REFERENCE EOP-EPP-020 SGTR WITH LOSS OF REACTOR COOLANT:
SUBCOOLED RECOVERY
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER:
B03-028 REVISION
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1.0
MINUTES QUESTION:
EPP-010 (Transfer to Cold Leg Recirculation)
has just been entered from EPP-009 (Post LOCA Cooldown and Depressurization)
three hours following a small break loss of coolant accident.
lt The following conditions exist:
SPDS status:
RED on HEAT RWST level RCS pressure RCS temperature SPDS subcooling margin =
SINK 20%
400 psig 395 oF 530F Containment pressure
= 5 psig Pressurizer level 55>o SG pressures
=
100 psig CSIPs
"A" running RHR pumps secured What should the operating shift do?
a.
Go to FRP-H.1.
b.
Immediately restart CSIP "B" and open both pressurizer PORVs.
c.
Restart CSIP "B" and return to EPP-009.
d.i Continue EPP-010 until the cold leg recirculation lineup is established.
ANSWER.
d.
REFERENCE:
EOP-EPP-010 TRANSFER TO COLD LEG RECIRCULATION
SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALZFZCATZON EXAMBANK QUESTION NUMBER B03-116 REVISION
JOB CLASS SENIOR REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME:
1.0
MINUTES QUESTION Following a reactor trip and safety injection from full power, EPP-014 was entered from Path-1 due to uncontrolled depressurization of steam generator (SG)
1B.
All actions of EPP-014 have been performed, but SG 18 is still depressurizing uncontrollably.
Path-2 has now been entered from EPP-014 based on abnormal radiation levels noted on the SG 1B steam line radiation monitor prior to closure of, its MSIV in EPP-014.
II With SG 1A and SG 1C intact, what action should the operating shift take when SG 1B narrow range level decreases to 10<?
a.
Maintain auxiliary feedwater isolated to SG 1B.
b.
Feed SG 1B to maintain narrow range level between 104 and 154.
c.
Feed SG 1B to maintain narrow range level between 104 and 504.
'
d.
Feed SG 1B at a rate of 12.5 kpph.
ANSWER:
a 0 REFERENCE:
EOP-PATH-2 EOP-EPP-014 PATH-2 FAULTED STEAM GENERATOR ISOLATION
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATZON EZAMBANK QUESTION NUMBER B04-024 REVZSZON:
JOB CLASS.
POINT VALUE ESTIMATED TIME:
QUESTION:
REACTOR OPERATOR QUESTION 1.0
MINUTES With the plant at 1004 power, 1CC-299 (RCP bearing oil coolers returns) is closed by mistake by an Auxiliary Operator performing a
CCW lineup in the RAB and cannot be reopened from the Control Room or by local action.
With BEARING CCW LOW FLOW alarms actuated and RCP motor bearing temperatures increasing, what actions must the operating shift take?
a.
Within 10 minutes of the loss of CCW flow to the RCP bearing oil coolers, trip the reactor then secure the RCPs.
b.
Trip the reactor, then secure the RCPs only if bearing temperature limits are exceeded; there is no time limit to trip the RCPs.
c.
Trip the reactor then secure the RCPs 10 minutes after bearing temperature limits are exceeded.
d.:
Commence a plant shutdown at 54/minute and secure RCPs immediately after the reactor is shutdown.
ANSWER a ~
REFERENCE AOP-018 REACTOR COOLANT PUMP ABNORMAL CONDITIONS
SHEARON HARRIS NUCLEAR POSER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER.
B04-030 REVISION:
'OB CLASS REACTOR OPERATOR QUESTION POINT VALUE~
ESTIMATED TIME:
1.0
MINUTES QUESTION.
RHR Train "A" was being placed in service for an RCS cooldown.
The following events have occured:
At 18:55, the "A" RHR pump was started for sampling but it immediately tripped At 18:58, the "A" RHR pump was started and a sample was obtained At 19:10, the "A" RHR pump was secured Which one of the following is the earliest time another start of the "A" RHR pump can be attempted?
a.
Immediately b.
19:25 c.
19:40 d.
19:55 ANSWER d.
REFERENCE OP-111 RESIDUAL HEAT REMOVAL SYSTEM 7,
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER:
B04-048 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION:
Path-1 is being performed following a reactor trip and safety injection from 1004 power.
No other procedures have been implemented yet.
Step
(VERIFY POWER AVAILABLETO PORV BLOCK VALVES) has just been reached and the following conditions exist:
WIDE RANGE SG levels SG pressures Pressurizer pressure Core exit thermocouples Containment pressure Total AFW flow RCP status RVLIS Full Range Level 354 in all three SGs 300 psig, decreasing uncontrollably 2335 psig 740 F and increasing 5 psig 530 KPPH Tripped by the operators 624 What should the operating shift do?
a ~
b.
cod.
Go to EPP-014 Go to EPP-015 Go to FRP-C.2 Go to FRP-H.1 ANSWER:
c REFERENCE.
CSFST EOP USERS GUIDE CSFST EOP USERS GUIDE
.)~
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALZFZCATION EXAMBANK QUESTZON NUMBER B05-020 REVZSZON
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TZMEe 1.0
MINUTES QUESTION0 The operating shift has just reached Entry Point C of Path-1 following a reactor trip and safety injection from 1004 power.
The following conditions exist:
CSIP status
RCP status
RCS pressure
Power range NIS
Core Exist TCs temperature
- 'G A NR level
Containment pressure
RWST level
CST level both running secured by the operators 1155 psig
564oF 154
3.5 psig 904
What action should the operating shift perform next?
a.
Align service water to the AFW pumps.
b.
Implement FRP-H.5.
c.
Implement FRP-C.3.
d.
Align CCW to sample heat exchangers and gross failed.
fuel detector.
ANSWER a ~
REFERENCE:
EOP-PATH-1 PATH-1 PATH-1 USERS GUIDE PATH-1 USERS GUIDE FOLDOUT A
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATION EXAMBANK QUESTION NUMBER.
B05-046 REVISION
JOB CLASS:
SENIOR REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION:
A plant heatup to 557 F is in progress with RCS temperature currently 330'F and the 1A auxiliary feedwater (AFW) pump cleared for maintenance.
What actions should. the operating shift take with these conditions?
a.
Continue the plant heatup to 557'F but restore the 1A AFW pump to OPERABLE condition prior to taking the reactor critical.
b.
Write a temporary change to GP-002 concerning the AFW pump and then continue the heatup to 557 F.
c.
Do not continue the heatup to ~ 350 F until the 1B AFW pump is proven OPERABLE.
d.
Do not continue the heatup to ~ 350'F until the 1A AFW pump is returned to restored to OPERBLE status.
ANSWER d.
REFERENCE:
TECH.
SPECS.
3.7.1.2AUXILIARY FEEDWATER SYSTEM
SHEARON HARRZS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER:
B07-005 REVISION:
JOB CLASS POINT VALUE ESTIMATED TIME:
REACTOR OPERATOR QUESTION 1.0
MINUTES QUESTION.
Following a reactor trip from 1004 power, all power is lost to the switchyard and both emergency diesel generators fail to start.
The operating shift has.implemented EOP-EPP-001 (Loss of AC Power to 1A-SA and 1B-SB Buses)
and are performing Step 21 to rapidly cooldown and depressurize the RCS to prevent RCP seal damage.
What condition would require termination of this cooldown?
a.
RVLIS full range level less than 394 b.
RCS cooldown rate in excess of 100 F/hour c.
Pressurizer level off-scale low d.
Source Range startup rate of + 0.1 DPM ANSWER d.
REFERENCE EOP-EPP-001 LOSS OF AC POWER TO 1A-SA AND 1B-SB BUSSES
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATZON EXAMBANK QUESTION NUMBER B07-050 REVISION
JOB CLASS!
SENIOR REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME QUESTION:
1.0
MINUTES Emergency Diesel Generator (EDG)
1A-SA fuel oil day tank has developed a leak and level has decreased to 94.
What action should be taken to demonstrate the OPERABILITY of EDG B-SB?
a ~
Test EDG B-SB per the current surveillance test schedulei no special testing is required.
b.
Start, and load EDG B-SB without barring per the surveillance test and operating procedure within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c.
Start and load EDG B-SB without barring per the surveillance test and operating procedure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
Bar EDG B-SB to ensure free rotations the diesel does not need to be started and loaded.
ANSWER c
OP-155 OST-1073 REFERENCE:
TECH.
SPECS.
3.8.1.1A.C.
SOURCES, LIMITING CONDITION FOR OPERATION DIESEL GENERATOR EMERGENCY POWER SYSTEM 1B-SB EDG OPERABILITY TEST MONTHLY INTERVAL
SHEARON HARRIS NUCLEAR POWER PLANT LZCENSED OPERATOR REQUALZFZCATZON.EXAMBANK QUESTION NUMBER B08-001 REVISION:
JOB CLASS REACTOR OPERATOR QUESTION POINT VALUE ESTIMATED TIME 1.0
MINUTES QUESTION:
A control room evacuation has occurred and control has been transferred to the Auxiliary Control Panel.
A plant cooldown is being commenced with the following initial conditions:
Primary temperature
= 5304F Pressurizer level = 304 Primary pressure
= 1900 psig Steam generator pressure
= 1,000 psig What is the minimum steam generator levels that the operators should maintain with these conditions?
a.
104 using narrow range level indication b.
454 using narrow range level indication c.
504 using wide range level indication d.
604 using wide range level indication ANSWER c ~
REFERENCE:
AOP-036 AOP-004 SAFE SHUTDOWN FOLLOWING A MAJOR FIRE SAFE SHUTDOWN IN CASE OF FIRE OR CONTROL ROOM INACCESSIBILITY
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFZCATION EXAMBANK QUESTION NUMBER B11-006 REVISION:
JOB CLASS:
REACTOR OPERATOR QUESTION POINT VALUE 1 '
ESTIMATED TIME
MINUTES QUESTION What is the annual limit for CP&L whole body dose for CP&L employees whose non-CP&L dose for the current year has been determined?
a.
5.0 Rem TEDE CP&L dose ANSWER b.
REFERENCE PLP-511 RADIATION CONTROL AND PROTECTION PROGRAM
SHEARON HARRIS NUCLEAR POWER PLANT LICENSED OPERATOR REQUALIFICATZON EXAMBANK QUESTION NUMBER:
B11-008 REVISION
JOB CLASS~
SENIOR REACTOR OPERATOR QUESTION POINT VALUE:
ESTIMATED TIME 1.0
MINUTES QUESTION Which one of the following is the reason core locations R08, R09, P08, and P09 are the last locations filled during refueling?
a.
Prevents causing a non-conservative 1/M plot b.
Increases the time to detect an inadvertant dilution c.
Limits the possibility of attaining a critical configuration if an inadvertant dilution occurred d.
Allows for insertion of fuel latched to the manipulator crane if a refueling cavity seal failure occurred ANSWER d.
REFERENCE FHP-014 FUEL AND INSERT SHUFFLE SEQUENCE