IR 05000338/1993300
| ML20059G286 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/20/1993 |
| From: | Aiello R, Lawyer L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20059G269 | List: |
| References | |
| RTR-NUREG-1021 50-338-93-300, 50-339-93-300, NUDOCS 9311080056 | |
| Download: ML20059G286 (186) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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-S REGION 11
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101 MARIETTA STREET, N.W., SUITE 2900
y ATLANTA, GEORGIA 30323-0199
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Report Nos.:
50-338/93-300 and 50-339/93-300 Licensee: Virginia Electric and Power Company
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5000 Dominion Boulevard Glen Allen, VA 23060 Docket Nos.:
50-338 and 50-339 License Nos.: NPF-4 and NPF-7
Facility Name:
North Anna 1 and 2 Examination Conducted: August 25, S tember 7-10, and September 21-24, 1993
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O Chief Examiner:
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Ronald F; ATello Dat'e Sicfned
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Accompanying Personnel:
R. Baldwin, RII R. Pugh, PNL J. Bumgardner, PNL N\\
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Approved by:
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Date Signed t
' Lawrence L. Lawyer, Chief (/
Operator Licensing Section
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Operations Branch Division of Reactor Safety SUMMARY Scope:
NRC examiners conducted regular,. announced operator licensing requalification
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examinations during the periods of August 25, September 7-10, and September 21-24, 1993.
Examiners administered examinations under the guidelines of the Examiner Standards (ES), NUREG-1021, Revision 7.. Six Senior
. Reactor Operators (SR0s) and five Reactor Operators (R0s) receivediritten and operating examinations. Two SR0s and three R0s received written examinations only. These operators took and passed the operating portion of the
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examination in September of 1992 (Report No. 50-338/92-302).
For..the i
simulator portion.of the examination this year, operators comprised two crews.
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9311080056 931021
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PDR ADOCK 05000338 V
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Results:
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Operator Pass / Fail:
SR0 R0 Total percent Crews Percent i
Pass
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100
100 Fail
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Examiners judged that the North Anna Nuclear Station requalification program
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was satisfactory based on the results of the examinations.
Examiners identified weaknesses in crew operations (paragraph 2.e.(3)),
e operator knowledge of transient and accident analysis, (paragraph 2.e.(1)),
and operator knowledge the electrical distribution. system (paragraph 2.e.(1))
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Examiners identified a strength in the development and use of the stand alone classroom simulator, the computerized control room narrative logs and the
,1 Virginia Power Personnel Qualification Standard (VPPQS) (paragraph 2.h).
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Two Inspector Followup Items (IFIs) were identified (paragraph 2.e.(3) and paragraph 2.1).
No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- M. Allen, Supervisor, Operations Training
- H. Christ, Supervisor, Station Procedures
- B. DeLamorton, Supervisor, Nuclear Training (Simulator)
- L. Edmonds, Superintendent, Nuclear Training
- J. Hayes, Superintendent, Operations
- G. Kane, Station Manager
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- P. Kemp, Supervisor, Licensing
- J. O'Hanlon, VP, Nuclear Operations NRC Personnel
- J. York, Senior Resident Inspector
- R. Pugh, PNL
- J. Bumgardner, PNL
- Attended exit 2.
Discussion a.
Scope NRC examiners conducted regular, announced operator licensing.
requalification examinations during the periods of August 25, September 7-10, and_ September 21-24, 1993.
Examiners administered examinations under the guidelines of the Examiner Standards, NUREG-1021, Revision 7.
Six SR0s and five R0s received written.and
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operating examinations. Two SR0s and three R0s received written examinations only. These operators took and passed the operating portion of the examination in September of 1992 (Report No.
50-338/92-301).
For the simulator portion of the examination this year, operators comprised two crews.
b.
Reference Material The examination team reviewed the reference material and determined that the reference material was adequate to support the e'xamination.
The examiners also reviewed the licensee's 1992 Licensed Operator Requalification Program sample plan. The sample plan was compared to NUREG-1021, ES-601,-Attachment 2.
The sample plan met all the r
guidelines for NRC administered examinations.
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Report Details
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Examination Development
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The examination team conducted a comparison of the facility's proposed-i written, walk-through, and dynamic simulator examinations to the guidance of NUREG 1021, ES-602. The NRC substituted new material for-
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about five percent of each section of the examination..The team
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approved the remainder of the examination with the following minor
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changes.
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(1) Written examination The team either changed or replaced several distractors to make:
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question on Part A and four on Part B.
The examination team-
developed one new question for the written Part A examination to investigate-a weakness that was suspected by the examiners during
~the preparation week.
(2) Simulator Scenarios
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The examination team reviewed all simulator scenarios that were i
selected for the examination week. These scenarios were generally well _ constructed, were in keeping with the guidelines
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of the examiner standards, and ran properly on the simulator.
The examiners identified no significant concerns, and made no major substitutions to the facility developed scenarios.
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discrimination of the simulator examination. The team added time criticality to manually tripping-the main turbine in scenario
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SXG-23. The team also made the rod drive failure occur.on a
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trigger of reactor power at 50 percent versus manually inputting the failures in the same scenario..
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(3) Job Performance Measures (JPMs)
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The JPMs were well structured and formatted. The JPM bank-covered many subject areas that were important to plant safety.
~i Critical: steps were properly identified and didinot. require-i modifications. The NRC added turbine trip failure.to. Task Performance Evaluation-(TPE)-185.4B, Immediate Actions for E-0 for memory, thus making ~ the JPM more discriminating.
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Examination Administration J
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The licensee's administration of the examinations was well~ planned and.
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coordinated. The examination administration satisfactorily met'all:
the recommendations delineated in NUREG 1021, ES-601 and ES-604,.
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Paragraph D, respectively. The team determined-that all facility
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evaluators were satisfactory.
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Report Details
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Operator Performance Five operators took and passed a written examination only.
Eleven licensed operators passed a written examination and an operating test consisting of a walk-through examination and a dynamic simulator examination, developed and administered in accordance with NUREG 1021, ES-602, ES-603, and ES-604, respectively.
(1) Written Question 588 was developed by the NRC. Three of the five R0s and two of the six SR0s missed this question. This part
"A" question tested the operators knowledge of the Reactor Coolant System (RCS) response to a loss of one Reactor Coolant Pump (RCP). All five candidates who missed this question chose distractor "c" which stated incorrectly that the indicated flow in the affected loop would drop to zero and stay at zero. This indicated a weakness in the Operators knowledge of transient and accident analysis. Actual indication on the control board would reflect a significant amount of reverse flow through the loop with the secured RCP.
Four of the five R0s missed question 572 which tested the operators knowledge of how the electrical system will respond to a fault on the IB station service bus. All four operators chose distractor "a", which stated incorrectly that the IB and 1H busses would remain energized. This indicated a weakness in the operators knowledge the electrical distribution system and ability to determine what actions should automatically occur during off normal conditions.
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(2) Job Performance Measure (JPM) Performance Operator performance on the JPM part of the examination was satisfactory.
No significant concerns were identified.
(3)
Simulator Crew performance in the simulator part of the examination was satisfactory. However, one crew took approximately 90 minutes-to recover from a Steam Generator Tube Rupture (SGTR). The Westinghouse Owners Group (WOG) background document for E-3 stated, in_ part, that accumulation of water in the secondary side can lead to an overfill condition which can severely aggravate the radiological consequences and increase the likelihood of.
complicating failures. The WOG further stated that timely operator intervention is necessary to limit the radiological releases and prevent SG overfill. - The Final Safety Analysis Report (FSAR) stated under " Analysis of Effects and Consequences" that "the operator identifies the accident type and terminates
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break flow to the faulty SG within 30 minutes of accident initiation." This issue of timeliness is a concern and is identified as IFI 50-338/93-300-01: Operator timely response and.
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compliance with the FSAR regarding a SGTR. Overall, operator performance was satisfactory.
f.
Program Evaluation Based on the examination results, the North Anna Nuclear Station Requalification Program meets the criterion established in NUREG-1021, ES-601.
The examiners judged that the North Anna Nuclear Station requalification program was satisfactory.
g.
Simulator Facility NRC examiners reviewed the licensee's record of simulator usage during calendar year 1992. The inspectors found that the simulator. was.
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available for a total of 6264 hours0.0725 days <br />1.74 hours <br />0.0104 weeks <br />0.00238 months <br /> based on a Monday through Friday usage.
During these hours of availability, the simulator was used~for operator training related functions 69 percent of the time.
It was
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used for non-operator training functions 8.8 percent of the. time, and
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it was idle 22.2 percent of the time. The examiners concluded that
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the simulator availability for operator training usage was satisfactory. The examination team identified no simulator facility problems.
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General Observations The Virginia Power Nuclear Training Center has a stand alone classroom portable simulator. This system provides the instructor the capability to provide dynamic presentations on plant systems in the classroom.
using equipment identical to the plant. Some of its attributes are
that the system is completely portable, _ full scale simulator changes
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can be developed and tested on the Classroom Simulator. prior to installation in the full scale simulator, control boards are identical to control room including color, system controls operate the same with the same look as the actual controls, dynamic system operations can be demonstrated and observed and much more.
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The Utility also has a method of computerized control room narrative logs. This log is similar to a word processor and_ includes a spell checker.
" Late Entries" may be_ added to the log at any time by simply.
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inserting a blank line. Once the log is " Turned Over" and printed, it -
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is locked from any further editing. The Operations Department commenced using computerized narrative logs on August 30, 1993.
Presently the logs are printed at the end of_ each shift and signed by the watchstander. The log is _then placed in the log book prior to transmittal to Station Records. The' plan is to install an NRC
approved electronic storage system (Laser Storage System) sometime in 1994 and then transmit the narrative logs (among other documents) to
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records electronicall Report Details
The Virginia Power Personnel Qualification System (VPPQS) is a micro-computer based software application designed to allow the user to easily and rapidly create watchbills, track training, and track qualifications.
It is an administrative barrier against assigning a non-qualified person to a watch station. Some of its attributes include rapid verification of qualifications for watchstanding and quick generation of watchbills, monitoring of attendance at requalification training and easy determination of classes missed and electronic signatures on qualification cards and TPEs.
The examiners determined the stand alone classroom simulator, the computerized control room narrative logs and the VPPQS to be training and operations department strengths.
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Procedures Title 10 CFR Part 50, Appendix B, requires that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
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The inspector examined the E0Ps listed in the Appendix for quality and usefulness and found them to be satisfactory.
The examiner noted that Step 15 of FR-H.1 required that the operator initiate SI.
One crew decided not to perform this step since SI had already been initiated and reset. The facility staff could not demonstrate that the crew's action was within the basis of this step.
This is identified as IFI - 50-338/93-300-02.
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Exit Interview J
At the conclusion of the site visit, the examiners met with the represent-
atives of the plant staff indicated in paragraph 1 to discuss the results of the examinations. The facility licensee did not identify as proprietary any material provided to or reviewed by the examiners.
Dissenting comments were not received from the licensee. The examiners further described the areas inspected and discussed in detail the inspection findings listed below.
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Report Details
Item Number Description / Para
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50-338/93-300-01 IFI - Operator timely response and compliance with the FSAR regarding a SGTR.(paragraph 2.e.(3))
50-338/93-300-02 IFI - Deviating from the procedure regarding the initiation of SI in Step 15 of FR-H.1, Response -
to Loss of Secondary Heat Sink (paragraph 2.1).
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APPENDIX LIST OF E0Ps REVIEWED Procedure #
Procedure Title Rev. #
l-E-0 Reactor Trip Safety Injections
1-E-1 Loss of Reactor or Secondary Coolant
1-E-2 Faulted Steam Generator Isolation
1-E-3 Steam Generator Tube Rupture
1-ES-0.2A Natural Circulation Cooldown With CRDH Fans
l-ES-0.28 Natural Circulation Cooldown Without CRDH Fans
1-ES-0.4 Natural Circulation Cooldown With Steam Void in Vessel (Without RVLIS)
I l-ES-1.1 SI Termination
1-ES-1.2 Post-LOCA Cooldown and Depressurization
1-ES-3.1 Post-SGTR Cooldown Using Backfill
1-ES-3.2 Post-SGTR Cooldown Using Blowdown
1-ECA-0.0 Loss of All AC Power
1-ECA-1.1 Loss of Emergency Coolant Recirculation
1-ECA-2.1 Uncontrolled Depressurization of All Steam enerators
1-FR-C.1 Response to Inadequate Core Cooling
1-FR-H.1 Response to Loss of Secondary Heat Sink
1-FR-P.1 Response to Imminent Pressurized Thermal Shock Condition
1-FR-S.1 Response to Nuclear Power Generation /-
1-FR-Z.1 Response to High Containment Pressure
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ENCLOSURE 2 SIMULATOR FACILITY REPORT Facility Licensee: North Anna Nuclear Station Facility Docket No.:
50-338 Operating Tests Administered During: The weeks of September 5 and 19, 1993 This form is used only to report observations. These observations do not constitute, in and of themselves, 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 which may be used in future evaluations. No licensee. action is required solely in response to these observations.
The examination team identified no simulator facility problems.
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VIRGINIA' POWER
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NORTIT ANNA NUCLEAR TRAINING
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LICENSED OPERATOR REOUAL PROGRAM
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NRC ADMINISTERED REACTOR OPERATOR EXAMINATION l
PART "A" AEO-10 AUGUST 25.1993
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NAME:
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(PRINT)
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Number of Questions
Possible Points
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80
% OR 12 Points is passing.
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Time limit:
I hour (s).
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PLEDGE I have neither given nor received any aid on this test.
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Signature:
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TEST RESULTS Grade:
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Graded by:
Date:
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Date:
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TEST Sinnatures (Required for master only)
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Submitted by:
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Date:
DC-D -9)
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<fI7fTS Approved byC
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AEO-10-KEY
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i VIRGINIA POWER NORTII ANNA NUCLEAR TRAINING LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED REACTOR OPERATOR EXAMINATION PART "A" AEO-13 AUGUST 25.1993
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NAME:
DATE:
(PRINT)
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INSTRUCTIONS 1.
Number of Questions
Possible Points
80
% OR 12 Points is passing.
2.
Time limit:
hour (s).
PLEDGE I have neither given nor received any aid on this test.
Signature:
TEST RESULTS Grade:
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Graded by:
Date:
Validated by:
Date:
(If Applicable)
TEST Signatures (Required for master only)
Submitted by:/
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Date:
0:1-n-93
/'k Writer" Approved by:
h Date:
Th 7/#E r
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Supen'isor AEO-10-X
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Vircinia Power North Anna Power Station Licensed Operator Requalification Progen
Static Simulator Scenario Loss of Offsite Power with EDG Failure AE010
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c STATIC SIMULATOR SCENARIQ Loss of Offsite Power with EDG Failus Summan';
Unit I is operating at 100% power with a core bumup of 4000 MWD /MTU when a loss of offsite power occurs with the subsequent failure of 'lH' EDG to remain on line after starting.
SIMULATOR SETUP:
Select IC-1 (100% Power)
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Advance charts and ensure they are inking
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Ensure VCT level >42%
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Ensure FCV-1113A is in manual with a controller output of 31.5 %
Enter the following malfunctions on the malfunction processor:
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MEL01, 60 sec. TD MRC1602, I sec. TD, I sec. ramp, 85 % deg, RX23 trigger Switch Override FWP3B_STOP=ON FWP3B_ START =OFF FWP3B_ASTOP=OFF FWP3B_ASTART=OFF Allow simulator to mn 10 minutes to establish trends.
After coine to freere:
RR_ RAD (3) set to zero RMR1731 set to zero XRMPROC=F Go to mn for one second and back to freeze.
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TIME EVENT MALFUNCTIONS OVERRIDES ACTIONS TAKEN T=0 T = 30 sec. on
'H' EDG trips on overspeed ON SIMLOCH:
MELO) timer EDGH_OVSPDJRP=T T=3 min an MELO1 timer Manually open Reactor trip switch T=5 min. on MEIA1 timer Place light switch LCl in instmetor booth in hand VERIFICATIONS:
'B' AFW pump did not start VCT level >5 %
Ensure instruemnt air has recovemd and RM trip valves are open.
Ensure 1-FW-FI-100A is >340 GPM and not pegged high
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TIJRNOVER:
The unit has been running at 100% power for three months when a loss of offsite power
occurred 6 minutes ago for reasons yet unknown. 'H' 4160 V Bus voltage has not been restored. ONLY THE BfMEDIATE OPERATOR ACTIONS OF E-0 HAVE BEEN
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PERFORMED.
Core Bumup is 4000 MWD /MTU.
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EOUIPMENT OUT OF SERVICE:
None.
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RA-0156 What is the concern one would have if there was no red light indication for B charging pump with amps indicated? (Assume light bulb is NOT burned out.)
No positive way of verifying the pump is running.
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b.
Trip Coil is not functional.
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Could cause emergency bus overload due t'o failure to strip from bus on an undervoltage c.
signal.
d.
Breaker closing coil is not functional.
Answer: b Points: 1.0 References: Simulator Configuration; ESK-5AM K/A: 063000.K3.02 RO 3.5/SRO 3.7 Sys. 06 004020.A4.02 RO 3.7/SRO 3.3 Sys. 09 l
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What will be the response of "A" CC pump if "H" 4160V bus is restored? (Assume the Abnormal Procedure was not used to restore "H" 4160V bus.)
a.
Pump will automatically restan after a 20 see time delay.
b.
Pump will automatically restan after manually resetting the stub bus.
Pump will automatically restart after 15 see time delay.
c.
d.
Pump will remain locked out.
Answer: c Points: 1.0 References: ESK-5P K/A: 008000.A3.01 RO 3.2*/SRO 3.0* Sys.12 062000.K3.01 RO 3.5/SRO 3.9 Sys. 26
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RA-0161 Which one of the following must be reset in order to shutdown the "A" AFW pump? (Given that "H" bus is restored with "A" AFW pump mnning.)
a.
Safety injection.
b.
Feedwater isolation. (if one should occur)
c.
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d.
Under voltage on "H" Emergency Bus.
Answer: c Points 1.0 References: ESK-5AA K/A: 061000.K4.06 RO 4.0"/SRO 4.2* Sys. 04
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Which one of the following statements is tme if N-35 (Intermediate Range NI) is substantially l
over compensated.
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N-35 will remain higher than P-6; Source range re-initiation will occur when N-36 falls l
below P-6.
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N-35 will be lower than P-6; Source range re-initiation will not occur as N-36 falls
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below P-6.
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N-35 will remain higher than P-6; Source range re-initiation will not occur when N-36 I
falls below P-6.
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N-35 wi.i be lower than P-6; Source range re-initiation will occur as N-36 falls below i
P-6.
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Answer-d points: 1.0 i
References: 1-AP-4.2; NA-DW-5655D33/Sh. 3
K/A: 000033.EA2.11 RO 3.1/SRO 3.4 Sys. 28-OE
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012000.K6.10 RO 3.3/SRO 3.5 Sys. 57 (
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Select one of the following which best completes the statement. RM-RMS-159 Low Flow alarm is locked in due to:
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Phase "A" isolation signal.
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Hi-Hi mdiation alarm.
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c.
Loss 9 power to the associated sample pump.
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d.
Closure of RM-TV 100D.
Answer: c i
Points: 1.0
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References: Load List; Loop Diagram RM-159 K/A: 073000.A2.01 RO 2.5/SRO 2.9* Sys. 54 073000. A4.02 RO 3.7/SRO 3.7 Sys. 54 I
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RA-0167 The following action will result in which one of the below responses:
The Shift Supervisor, desiring to have two operable charging pumps running on the "J" 4160V bus, orders the control switch for 15H7 placed in P-T-L and directs an operator to rack out 15H7 and rack in 15J7 under current plant conditions.
a.
Auto stan of "A" charging pump (breaker closure).
b.
All thme charging pump breakers closed.
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c.
No charging pumps running.
d.
"C" charging pump ranning on its alternate breaker.
Answer: c Points: 1.0 References: ESK-5AL; ESK-5AM; ESK-5AN K/A: 004000.K2.02b RO 3.3/SRO 3.5 Sys. 09
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RA-0169 -
THIS-QUESTION DOES NOT PERTAIN TO THE
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SCENARIO.
i Which one of the following statements is true concerning system response if breaker 15A2 ("A" l
Station Service (SS) normal feeder breaker) was inadvertently opened by the operator. (Given that the event occured while the unit was at 100% power condition.)
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"A" 4160V SS would swap to "A" Reserve Station Service (RSS) feeder brehker,_"B" a.
SS and "C" SS busses would not swap to their respective RSS bus feeder breakers.
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"A", "B" and "C" 4160V SS feeder breakers would swap to their respective RSS bus
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feeder breakers.
c.
"A" 4160V SS feeder breaker would not swap to "A" RSS feeder breaker;
"B" SS and
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"C" SS busses would swap to their respective RSS bus feeder breakers.
d.
"A", "B" and "C" 4160V SS feeder breakers would not transfer to their respective RSS feeder breakers.
Answer: d Points: 1.0 i
References: FE-21G e
i K/A: 062000.K4.03 RO 2.8*/SRO 3.1 Sys. 06
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RA-0172 l
t Note that a "'B" loop AT/Tavg has failed. Such a failure with the UNIT AT 90% POWER
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would result in which one of the following?
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OT AT setpoint increasing.
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b.
OP AT setpoint decreasing.
c.
P-12 interlock met prematurely, i
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Rod insertion limit decreasing.
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Answer; b Points: 1.0 References: Simulator; TS 3.1.3.6; Curve Book l
K/A: 002000.A1.09 RO 3.7/SRO 3.8 Sys. 56 001050.A1.01 RO 4.0?/SRO 4.27 Sys. 61
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TIUS QUESTION DOES NOT PERTAIN TO THE i
SCENARIO.
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Complete the following statement concerning reactor coolant loop flow indication. (Given that
all three reactor coolant pumps (RCPs) are initially running.)
If "C" RCP trips then indicated flow in the "C" loop will
, while indicated flow in the "A"
and "B" loops will
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a.
decrease to zem then increase slightly; increase b.
decrease to zem then increase slightly; decrease c.
decrease to zem and stay at zero; increase i
d.
decrease to zero and stay at zero; decrease
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Answer; a
Points: 1.0 e
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References: HT-FF
K/A: 002000.A1.05 RO 3.4/SRO 3.7 Sys. 56 l
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Which one of the following pieces of equipment is in-a condition that WOULD NOT be
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expected for current plant conditions?
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FW-P-3A ("A" Auxiliary Feedwater Pump),
b.
FW-P-3B ("B" Auxiliary Feedwater Pamp).
c.
CH-P-1 A ("A" Charging Pump).
!
.
'
d,.
FW-P-1C1/1C2 ("C" Main Feed Pump).
Answer: b Points: 1.0 References: Simulator K/A: 061000.K4.06 RO 4.0*/SRO 4.2* Sys. 04
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RA-0180 Choose the statement below which correctly reflects CC flow rate through the "A" RCP Thermal Barrier.
Cannot be determined with event in progress.
a.
b.
No flow due to loss of the IH emergency, bus.
c.
Approximately 21 GPM.
d.
No flow due to a thermal barrier heat exchanger leak.
Answer: b Points: 1.0 References: Simulator; FM-79B/Sh. 2; Loop Diagriun CC-64; l-AP-15 K/A: 008000. A3.01 RO 3.2*/SRO 3.0* Sys.12 003000.A4.08 RO 3.2/SRO 2.9 Sys. 56
,
-
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.
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With the current "B" Loop temperature indication failure, is the position of FCV-1122 correct?
' No,100% full open should be indicated due to loss of power to 1-EI-CB-56.
!
a.
!
b.
Yes, Tave has shifted to Lo-Select.
c.
Yes, the median selector circuitry has compensated for its failure.
!
,
d.
No,100% full open should be indicated.
-
Answer: c
'
' Points: 1.0 References: Imop Diagrams CH-1, RC-68
-
K/A: 002000.A1.09 RO 3.7/SRO 3.8 Sys. 56 002000.Kl.06 RO 3.7/SRO 4.0 Sys. 56 l
.
.
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. RA-0483 THIS -QUESTION DOES NOT PERTAIN TO THE.
_
SCENARIO.
j Select the one correct statement'concerning operation of the Steam Generator Water Level Control (SGWLC) system. (Assuming the unit is stable at 50% power.)
,
If Channel 2 Ixvel Transmitter on "B" S/G fails low, feed flow to "B" S/G will initially -
a.
decrease.
b.
If Channel 3 Ixvel Transmitter on "C" S/G fails high, actual level in "C" S/G will
decrease.
'
,
If the selected channel of Feed Flow for "A" S/G fails low, "A" S/G level will decrease.-
'
c.
d.
If the Steam Pressure input to "B" Steam Generator Steam Flow fails "as is" and then
[
Unit power is increased from 50% to 100 %, indicated steam flow will be less than actual steam flow.
,
Answer: b
.
Points: 1.0 References: NA-DW-5655D33/Sh.13
i K/A: 059000.A2.11 RO 3.0*/SRO 3.3*
Sys.39
-
039000.A1.06 RO 3.0/SRO 3.1 Sys.42
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Which one of the following is correct concerning restoration of the UNIT 1 Radiation Monitors?
a.
Requires the "H" bus be restored.
b.
May be restored by swapping power to the Unit 2 "B" 4160V SS bus.
'
c.
May be restored by swapping power to the Unit 1 "J" Bus, d.
Must await the return of the "A" Station Service Bus'.
Answer: a Points: 1.0 References: Loop Diagrams; Load List K/A: 072000.A4.01 RO 3.0*/SRO 3.3 Sys. 54 062000.A2.04 RO 3.1/SRO 3.4* Sys. 06
.
.
-
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-
Which statement below best describes the response of the "B" SG PORV, if the reactor operator the controller for "B" Steam Generator power operated relief valve to a pot setting of 6.07 (
'dering current plant conditions)
A setting of 6.0 corresponds to a setpoint of 1080 psig. The valve would open further.
a.
b.
A setting of 6.0 corresponds to a setpoint of 1080 psig. The valve would shut.
A setting of 6.0 corresponds to a setpoint of 840 psig. The valve would open further.
c.
d.
A setting of 6.0 corresponds to a setpoint of 840 psig. The valve would shut.
Answer: b Points: 1.0 References: Loop Diagrams, MS-P.12 K/A: 041020.A4.06 RO 2.9*/SRO 3.1 Sys. 42 039000.Kl.02 RO 3.3/SRO 3.3 Sys. 42
.
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Od a.5 /r s. '?) Jn i , . ~. It)o,,//t /7n n ~ VIRGINIA 10WER g6pvfyL - ~ NORTII ANNA NUCLEAR. TRAINING-LICENSED OPERATOR REOUAL PROGRAM ! NRC ADMINISTERED REACTOR OPERATOR EXAMINATION OPEN REFERENCE PART "B" i AUGUST 25.1993 l . NAME: ! DATE: ' (PRINT) ~
, INSTRUCTIONS i 1.
Number of Questions
Possible Points
80 % OR 16 Points is passing.
- 2.
Time limit:
hour (s).
- PLEDGE I have neither given nor received any aid on this test.
Signature: TEST RESULTS Gmde: / % Graded by: Date: Validated by: Date: - (If Applicable) TEST Sirnatures (Required for master only) Submitted by: l' - >\\ Date: 88 /7@, /' Writer p .. Approved by:' Date: I//7[93 ' Supervisor PART B-KEY
'. _. .
. . VIRGINIA POWER. NORTil ANNA NUCLEAR TRAINING LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED REACTOR OPERATOR EXAMINATIOE OPEN REFERENCE PART "B" AUGUST 25.1993 . NAME: ! DATE: (PRINT) INSTRUCTIONS 1.
Number of Questions
Possible Points
j i
c4 OR 16 Points is passing.
2.
Time limit:
hour (s).
PLEDGE l , . I have neither given nor received any aid on this test.
' . Signature: . TEST RESULTS ' , Grade: / % Graded by: Date: Validated by: Date: (If Applicable) TEST Sinnatures (Required for master only) , Submitted by: 'N / Date: 7B~/7'S3 /W' Writer () Approved by: O D C Date: T[/'7[9~L Supervisor ' '
PART B-X ' ., _.. _ -, . - - .
. - - J %.) ., , Policies and Guidelines for Takinn NRC Written Examinations . 1.
Cheating on the examination will result in an automatic denial of your application and could result in more severe penalties.
2.
After you complete the examination, you'must sign the statement on the cover sheet indicating that the work is your own and tliat you have not received or given assistance in completing the examination.
3.
To pass the examination, you must achieve a grade of 80 percent or greater.
' 4.
The point value for each question is indicated in parentheses after the question number.
5.
You have a total of 3 hours to complete both sections of the examination.
6.
Use only black ink or dark pencil to ensure legible copies.
7.
Print your name in the blank provided in the upper right-hand corner of the examination
cover sheet and of each answer sheet.
8.
Mark your answers on the answer sheets provided.
USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGES.
9.
Use abbreviations only if they are commonly used in facility literature. Avoid using symbols such " < " (less than) or " > " (greater than) to avoid a simple transposition error resulting in an incorrect answer. Write it out.
10.
Show all calculations, methods, or assumptions used to obtain an answer to any short-answer questions.
11.
You may receive partial credit on questions that are not multiple choice format.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
. 12.
If the intent of a question is unclear, ask questions of the examiner only.
' l 13.
Limit restroom trips, if possible; only one applicant at a time will be allowed to leave.
Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheating.
14.
After you complete your examination, assemble the examination questions, the examination aids, the answer sheets, and all scrap paper into a package and give it to the examiner or proctor. Remember to sign the statement on the examinadon cover sheet.
r . , Y - . 15.
Ensure all information you wish to have evaluated as pan of your answer is on your answer sheet. Scrap paper will be disposed of immediately after the examination.
16.
After you have turned in your examination, leave the examination are as defined by the examiner. If you are found in this are while the examination is still in progress, you license may be denied or revoked.
, i a
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". RB-0480
Which one of the following actions must the crew take if while performing E-3 " Steam Genemtor Tube Rupture", (due to a tube mptum in "C" S/G), it is found that."C" MSTV and ' NRV cannot be closed? a.
close all mmaining MSTVs, NRVs and bypasses and cooldown with intact S/Gs
atmospheric relief valves.
b.
close all remaining MSTVs and cooldow'n with the ruptured S/G atmospheric relief valves.
~ . cooldown by dumping steam fmm "A" and "B" S/Gs through the AFW pump terry c.
turbine.
e d.
transition to ECA-3.1 "SGTR. with Loss of Reactor Coolant - Subcooled Recovery".
,
Answer: a
Points: 1.0 References: 1-E-3 K/A: 000038.EA2.08 RO 3.8/SRO 4.4 System Code: 42-OD 000038.G12 RO 3.8*/SRO 4.0* System Code: 42-OD ', !
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. RB-0518 How does cycling Pzr level enhance Upper Head Cooling in ES-0.4 Natural Circulation Cooldown with Steam Voids in Vessel (Without RVLIS)? The operator is reminded by a note preceding step 9 that, "In order to continue overall system depressurization, cycling PZR level (cycling pressure) may be necessary to enhance Upper Head cooling, As the pressurizer level is cycled, the heat transfer fmm the pressurizer metal is greater, a.
thus causing the pressurizer upper head to, cool faster.
b.
As pressurizer pressure is increased, the water in the PZR will be forced out, collapsing the steam bubble in the reactor upper head, thus cooling upper internals.
As pressure is increased, the amount of RCS subcooling also increases, so that heat c.
transfer rate from the upper head to the RCS will increase, d.
As pressurizer pressure lowers the water in the reactor vessel will flash to steam, removing the latent heat of condensation from the water in the upper head region, resulting in better upper head cooling.
Answer: b Points: 1.0 References: ES-0.4 Background Document; l-ES-0.4 K/A: 000009.EK3.10 RO 3.4/SRO 3.6 System Code: 62-OD 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-UD 000009.EKl.01 RO 4.2/SRO 4.7 System Code: 62-OD l
i
_ _ - ., .. - (N (j .qf
RB-0467 Which one of the following is the basis for closing the charging pump recirculation valves when . RCS pressure is below 1275 psig and the RCPs are tripped in accordance with the Continuous Action Summary of E-0? protect the Charging Pumps from runout operation.
a.
' b.
avoid Charging Pump cavitation.
I c.
maximize SI flow.
- d.
prevent damage to the VCT Relief Valve due to overfilling the VCT.
Answer: c Points: 1.0 , References: E-0 Background K/A: 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-OD 006000.K5.06 RO 3.5/SRO 3.9 System Code: 62 , > f ,
f
. ,m - . RB-0459 Why is there NO requirement on pressurizer level for starting an RCP, while in 1-ECA-1.1 - less of Coolant Recirculation, with RCS subcooling greater than 25'F {75*F}? Adequate subcooling ensures that pressure is high enough to prevent DNB from a.
occurring as the RCP is started.
b.
Pressurizer level is assured to be great enough to support RCP operation at this point in 1 -ECA-1.1.
Only adequate subcooling is required since it is likely pressurizer level will be offscale c.
low for this tnmsient.
d.
Pressurizer level requirements for starting an RCP only apply when inadequate RCS subcooling exists.
Answer: c Points: 1.0 References: 1-ECA-1.1 Background Rev.1B EOP Obj D K/A: 000011.EK3.12 RO 4.4/SRO 4.6 System Code: 62-OD
- f) '. .() V- . RB-0543 ' Which one of the following is correct if the crew has been unsuccessful in reducing core exit.
r thermocouples to <1200*F and FR-C.1 directs the operators to stan RCP(s) in an attempt to - lower core exit temperatures? (Note: RCP suppon condition / systems cannot be established.)
The first RCP may be staned without suppon conditions, but suppon conditions must be
a.
available prior to staning additional RCPs b.
RCPs should not be started without suppdn conditions. If RCPs were staned without suppon conditions then severe RCP damage would result requiring extensive RCP outage time to repair.
, RCPs may be started without suppon conditions being available provided permission is c.
obtained from the Station Emergency Manager (SEM).
d.
RCPs should be started even if suppon conditions cannot be established. RCPs may provide temporary cooling of the core even under highly voided conditions.
. Answer: d Points: 1.0 ! References: 1-FR-C.1; FR-C.1 Background Document K/A: 000074.EK3.07 RO 4.0/SRO 4.4 System Code: 62-OD ' i a +
- O.
b, . RB-0572 Why should the operator continue with RCS depressurization until pressurizer level is greater than 36% [50%] in accordance with step 29 of 1-AP-17 (Shutdown LOCA), if they notice RCS subcooling has lowered to 0*F7 After pressurizer level is restored, the depressurization will be stopped and RCS pn:ssure a.
will then begin to rise, which willIpstore the required subcooling.
b.
Subcooling will be restored by the cooldown after the depressurization is stopped.
Subcooling is not a concern because boiling is not a concern during shutdown conditions.
c.
d.
By increasing the pressurizer level, the head of water above the reactor will restore the RCS subcooling.
Answer: b Points: 1.0 References: ARG 2; l-AP-17 K/A: 000009.EK3.26 RO 4.4/SRO 4.5 System Code: 62-OD i .
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RB-0302 How close is the pressurizer to being solid if: RCS temperature is 150*P and pressure is 320 psig. PZR temperature is 350*P. PZR level is indicating 100% on Protection Channels I, II, and III and 85% on the Stanup level indicator. Assume Unit 1 is being cooled down and depressurized for a refueling outage. The PZR is being filled to a solid condition in preparation l for hydrogen peroxide addition.
' a.
The PZR is solid.
. b.
The PZR is 4% below a solid condition.
. t c.
The PZR is 13 % below solid condition.
d.
The PZR is 15% below solid condition.
Answer: b ' Points: 1.0 References: 1-OP-3.4; 1-SC-23.1 K/A: 002000.Al.02 R0 3.6/SRO 3.9 System Code: 56 , I & f
1
! _ _ _
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RB-0098 Which one of the following is required to be performed if it is desimd to continue to cooldown given that the following has just occured? While performing a Unit I cooldown in accordance with 1-OP-3.2, the operator perfonns the following as Tavg goes below 543*F: Blocks both trains of High Steam Flow SI'
Arms the two cooldown steam dumps by going to ' bypass interlock'
Then Tavg is allowed to increase to 546 F and then is lowered again to 543*F.
a.
Neither action is required to be performed again.
b.
Block the High Steam Flow SI again, the cooldown steam dumps am still available.
Use " Bypass Interlock" to mstore the cooldown steam dumps, the high steam flow SI c.
remains blocked.
d.
Both actions are required to be performed again.
Answer: d Points: 1.0 References: NA-DW-108D014 Sh. 2 K/A: 013000.K4.12 hO 3.7/SRO 3.9 System Code: 62 013000.K4.03 RO 3.9/SRO 4.4 System Code: 62 041020.K4.09 RO 3.0/SRO 3.3 System Code: 42 , f
O O - ' r +. . ' RB-0018 + Why is it not acceptable for the OATC to allow a ' Step 6' operator to perform a Rx stanup as part of his Rod control system check out? The operator has not completed the step program.
a.
b.
The operator has never attended a reactor, operator license class.
, The operator has not obtained the shift supervisor's permission.
c.
d.
The operator is not enrolled in a reactor operators license class.
Answer: d Points: 1.0 References: 10-CFR-55-3, VPAP-1404 t K/A: 194001.G29 RO 3.1/SRO 4.7 194001.G23 RO 2.8/SRO 3.5 194001.G31 RO 3.1/SRO 3.1
001000.G01 RO 3.7/SRO 3.8 System Code: 61 i . i I -
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RB-0173 .When may another RCP stan attempt be made given the following? At 1500 the "A" RCP ' tripped due to a faulty relay. - The relay was replaced and at 1615 the operator started the "A" RCP, but then inadvenently stopped it as it was reaching full speed. Upon the shift supervisors request the opentor again attempted to start "A" RCP at 1645 which also proved unsuccessful.
a.
1715 ' , b.
1745 I - ! c.
1815 , d.
I845 Answer: a Points: 1.0 References: 1-OP-5.2 K/A: 003000.G10 RO 3.3/SRO 3.6 System Code: 56 ' 003000.K6.14 RO 2.6/SRO 2.9 System Code: 56 ,
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- -
. - RB-0380 ,
.Which one of the following requires activation of both the "!.C and OSC? a.
any eme.rgency plant classification.
b.
any emergency condition classified higher than notificadon of unusual event.- only emergency condition classified as either a site area emergency or general c.
emergency.
' l d.
only emergency conditions chssified as general emergency.
i Answer: b i Points: 1.0 References: Emergency Plan K/A: 194001.A1.16 RO 3.1/SRO 4.4* System Code: 99-OC
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RB-0261 Select the correct response by the rod control system given that unit 1 is operating at'100% power with control rods in automatic and I. cop "B" cold leg temperature fails high? a.
Contml rods will initially step IN when instrument fails but _will cease movement once instrument is pegged high.
b.
Control rods will initially step IN at 72 steps per minute, i c.
Control rods will step in and stop when Tavg decreases 1.5*F.
d.
No control rod motion will occur at all.
-Answer: d Points: 1.0 References: PLS C.2; NA-DW-108D014 Sheet 1 of 17 K/A: 001050.K5.01 RO 3.3/SRO 3.6 System Code: 61 ,
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RB-0566 r
i What will be the new RCS boron concentration if while placing RHR in service on Unit 1, RHR l boron concentration is 1900 ppm and RCS boron concentration is 2200 ppm? l !
a.
1918 ppm ! b.
1953 ppm , , j c.
2028 ppm i ,
d.
2182 ppm ' i ,i Answer: d l r Points: 1.0 l Refeences: 1-OP-14.1 ' , K/A: 005000.K5.03 RO 2.9*/SRO 3.1* System Code: 60 005000.Kl.09 RO 3.6/SRO 3.9 System Code: 60 i
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. _ _ _.
_ . , ! . RB-0574 l
Which one of the following statements best describes the required actions, if a review of the - ' - most recent test results indicate that the final "as-left" setting for the intermediate range neutron flux trip on Channel N-36 is equivalent to 29% power while the reactor is in Mode 3 with the ' shutdown banks fully withdrawn in prepamtion for a reactor startup? , Correct the level trip setpoint to the current equivalent to 25 % rated thermal power prior ) a.
to increasing thermal power above the P-6 setpoint.
-l b.
Correct the level trip setpoint to the current equivalent to 25 % rated thermal power prior to increasing power above 5 % of rated thermal power.
. Place Channel N-36 in the tripped condition prior to entering Mode 2.
c.
d.
No action is required.
l > Answer: d ' Points: 1.0 References: TS 2.2.1 K/A: 015000.K3.01 RO 3.9/SRO 4.3 System Code: 28 l 012000.A2.03 RO 3.4/SRO 3.7-System Code: 57 012000.G05 RO 3.4/SRO 4.3* System Code: 57 ' 012000.A1.01 RO 2.9*/3.4* System Code: 57 , .' \\ i ! l l .I l
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RB-0474
Which one of the following is correct concerning the basis for establishing redundant HHSI flow j paths during a LOCA (two HHSI pumps are running)? ' ! provides additional protection against passive failures during the recirc phase of the a.
'
LOCA.
b.
enhances the flow indication reliability by' placing the altemate path flow transmitter in i the flow path.
' . provides additional makeup flow to the RCS because of less system 1:ead loss.
c.
d.
enhances reliability of HHSI pumps by reducing the backpressure on discharge line.
, Answer; a t Points: 1.0
References: E-1 Background Document K/A: 006000.K4.18 RO 3.3/SRO 3.8 System Code: 62 l i
1 ) i
.
9
- .
RB-0037 j Which one of the following is the response of source range counts for the given situation.
While a reactor stanup is in progress the RO stops contro' rod motion when the reactor is close j to criticality, but still subcritical? ! Continue to increase but at a slower rate., The stanup rate should stabilize at a lower a.
positive value.
! b.
Continue to increase for a short time and then plateau. The startup rate should gradually . decrease to zero.
l Stop increasing and stabilize at its present value. The stanup rate should'immediately c.
decrease to zero, d.
Begin to slowly decmase. The stanup rate should gradually decrease to zero from a slightly positive value.
., i Answer: b Points: 1.0 i References: 1-OP-1.5-1 K/A: 192008.K1.03 RO 3.9/SRO 4.0 System Code: Rx-Th ' 192008.Kl.05 RO 3.8/SRO 3.9 System Code: Rx-Th
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RB-0330 What will be the control rod position at 100% power given the following: Initial Conditions: Final Conditions: i
- Reator Power = 25%
- Reactor Power = 100% ~
- Equilibrium Xenon = -1570 pcm o Equilibruim Xenon = -2875 pcm
- Boron Concentration = 1300 ppm
- Boron Concentration = 1025 ppm
- Control Rod Position = 'D' bank at 60 steps i (Assume core burnup is 4000 ABVD/MTU and differential boron worth is -6.88 pcm/ ppm.)
h a.
D Bank @ 225 b.
D Bank @ 209 c.
D Bank @ 194 d.
D Bank @ 177 Answer: C - This is Unit 1 Cycle 10 Dependent Points: 1.0 References: Station Curves K/A: 001000.K5.05 RO 3.5/SRO 3.9 System Code: 61 001000.K5.09 RO 3.5/SRO 3.7 System Code: 61 001000.K5.22 RO 2.1/SRO 2.5 System Code: 61 001000.G06 RO 2.9/SRO 3.8 System Code: 61 , l l . - . - - . - -
- ,, $ lj - - ! RB-0049 (RO Only) While the control room operators are performing FR-S.1 (transitioned to from step 1 of E-0), the STA presents the shift supervisor with a red path on heat sink. Upon completion of FR-S.1, the shift supervisor should instruct the procedure reader to: a.
Tmnsition to FR-H.1 b.
Transition to E-0 c.
Perform E-0 and FR-H.1 concurrently d.
Transition to ES-0.0 rediagnoses Answer a Points: 1.0 References: OPAP-0002 K/A: 000029.G12 RO 4.I"/SRO 4.2* System Code: 00-OD . I - .
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' %.s - ! RB-0581 (RO Only) Which one of the following statements is correct conceming defeating annunciators? A piece of equipment covered by Technical Specifications may never have an associated a.
annunciator defeated.
b.
If the LCO for a Technical Specification piece of equipment is not satisfied, then the annunciator associated with it may NOT be defeated, The annunciator may be defeated as long as the equipment it is associated with has an c.
action statement with a time frame of less than 24 hours.
d.
When the power supply is lost to an annunciator panel, the entire panel of annunciators should be defeated.
Answer: b Points: 1.0 References: 1-AP-6; Operations Standards / tab A K/A: 016000.G10 RO 3.8*/SRO 3.9* System Code: OE
. A & - w - 't RB-0226 (RO Only) Which one of the following decisions and subsequent reasons is correct regarding manually initiating SI given the following:
An ATWS has occurred on Unit 1.
Auxiliary operators are unable to locally trip the reactor.
- The RO is manually inserting rods.
, The turbine has tripped.
'
Reactor power is 20%. a.
Yes, the Safety Injection may result in a Reactor Trip.
b.
Yes, a Safety Injection is the fastest way to emergency borate the RCS.
c.
No, the Safety Injection may result in a loss of heat sink.
d.
No, the Safety Injection may result in overfilling the pressurizer, resulting in degraded RCS pressure control.
Answer: c Points: 1.0 References: WOG Executive Volume - User's Guide /1-FR-S.1 K/A: 000029.EK3.12 RO 4.4/SRO 4.7 System Code: 00--OD ,, k
r . ' ' o O /lka.s/en9)Je . . JD c. //t Anna VIRGINIA POWER NORTH ANNA NUCLEAR TRAINING "" LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED REACTOR OPERATOR EXAMINATION OPFN REFERENCE PART "B" SEPTEMBER 21.1993 NAME: DATE:
(PRINT) - INSTRUCTIONS 1.
Number of Questions
Possible Points
80 % OR 16 Points is passing.
2.
Time limit:
hour (s).
PLEDGE I have neither given nor received any aid on this test.
Signature: TEST RESULTS G mde: / % Graded by: Date: Validated by: Date: (If Applicable) TEST Sienatures (Required for master only) Submitted by:
2 / Date: 08 /4-93 / ~ "' Writer () ' Approved by: _ Date: T[/ 9 /9~2, Supervisor PART B-KEY r
f I . n O v . i RB-0188 Which one of the following is the correct action the operating team should take given the following? A SGTR has occurmd on Unit 1, and E-3 is being utilized by the operating team. "A" SG was identified as the ruptured SG and was subsequently isolated. The RCS has been cooled down to 490*F and depressurized to stop primary to secondary leakage and recover PZR level.
.While attempting to establish 25 gpm charging flow, the BOP operator repons that "B" SG , LEVEL is increasing in an uncontrolled manner with all feed /AFW isolated; pmssurizer level is 30% and falling.
a.
Transition to ECA-3.1, step 1.
b.
Transition to E-3, step 1.
Continue with E-3, subsequent actions will provide a sufficient mitigation response.
c.
d.
Re-initiate RCS depressurization to backfill 'B' steam generator to the RCS.
Answer: b
Points: 1.0 References: 1-E-3; WOG Background K/A: 000038.EA2.02 RO 4.5/SRO 4.8 System Code: 42-OD 000038.EA2.01 RO 4.1/SRO 4.7 System Code: 42-OD 000038.EK3.06 RO 4.2/SRO 4.5 System Code: 42-OD l , >
! . . I \\ . . ,
, ,- $ - , RB-0370 Which one of the following describes the response of Natural Circulation to the given infonnation? The plant experiences a loss of all AC power. Efforts to restore AC power prove unsuccessful.
RCS inventory depletion from RCP seal leakage continues, eventually draining the upper head of the reactor vessel and causing steam voids to form in the S/G U-tubes.
Natural Circulation stops and all means of decay heat removal will be lost. Extensive a.
core damage will occur soon after the intermption of natural circulation.
b.
Natural Circulation stops and reflux boiling will remove decay heat until enough inventory is lost to prevent decay heat removal. Then inadequate core cooling may occur.
Natural Circuladon stops but reflux boiling will provide adequate decay heat removal for c.
as long as necessary.
d.
Natural Circulation decreases but continues to provide adequate decay heat removal for as long as necessary.
Answer: b Points: 1.0 References: 1-ECA-0.0; WOG Training Part B Item #000-055-11 K/A: 000055.EA2.02 RO 4.4/SRO 4.6 System Code: 06-OE 000055.EKl.02 RO 4.1/SRO 4.4 System Code: 06-OE .
o O b,) - - - . RB-0190 The preferred method to accomplish a post-SGTR cooldown is: a.
ES-3.1, " Post-SGTR Cooldown using Backfill" because it limits the release of radioactive contaminates and therefore limits the offsite dose.
b.
ES-3.1, " Post-SGTR Cooldown using Backfill" because it allows rapid cooldown and depressurization of the RCS.
c.
ES-3.3, " Post-SGTR Cooldown using Steam Dumps" because it limits the release of ' radioactive contaminates and therefore limits the offsite dose.
d.
ES-3.3, " Post-SGTR Cooldown using Steam Dumps" because it allows a rapid cooldown and depressurization of the RCS.
Answer: a Points: 1.0 References: 1-E-3; ES-3.1 Background Documents; l-ES-3.1 K/A: 000038.EK3.06 RO 4.2/SRO 4.5 System Code: 42-0D . $
w L u_ J ( ' . v . . RB-0515 Should an RCP be started in the following situation? Unit I has experienced a small break LOCA resulting in safety injection and a manual trip of all RCPs due to inadequate subcooling. - All emergency procedures are performed correctly and
all equipment operated as expected. The crew is presently at Step 12 of ES-1.2 Post LOCA Cooldown and Depressurization. All conditions are met to start an RCP but the RCS is solid , with subcooling at 80*F.
, No, the resultant pressure surge will cause the LOCA to worsen, resulting in even lower ' a.
subcooling.
b.
Yes, the leak will act as the steam bubble and limit the pressure insurge.
a No, the increase in pressure will cause the leak rate to be greater than the ECCS makeup c.
rate.
d.
Yes, forced flow heat transfer must be established regardless of consequences, to ensure core damage does not occur.
Answer: b Points: 1.0 References: 1-ES-1.2, ERG Feedback DW.90-051, Response Letter ERG 91-012 , K/A: 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-OD -
I . l ! l . -l
.. ... . .
I . e
- - . RB-0069 Which one of the following statements concerning the power defect is correct? The power defect is the diffemnce between the measured power coefficient and the a.
predicted power coefficient.
b.
The power defect increases the rod worth requirements necessary to maintain the desired shutdown margin following a reactor trip.
c.
Because of the higher boron concentration at the beginning of core life, the power defect is more negative.
d.
The power defect necessitates the use of a ramped Tavg program to maintain an adequate RCS subcooling margin.
Answer: b Points: 1.0 References: Unit 1 Plant Curves K/A: 001000.K5.49 RO 3.4/SRO 3.7 System Code: 61 . k.
. . . n O . RB-0031 At what rate may the operator realign the control rod in the following situation? Unknown to the contml mom crew, the reactor had been operating at 100% power with a 'D' Bank contrcel rod misaligned 9 steps above demand position. When the operator moves the controlling bank for reactor control, he receives a " Computer Alarm Power Range Tilt Rod-Deviation Sequence" annunciator alarm. The RPIs indicate a 13-step deviation between demand _ position and the affected rod's position.
, a.
No restriction apply ~ , b.
2 steps / hour c.
4 steps / hour d.
12 steps / hour Answer: a Points: 1.0 References: 1-AP-1.3; OP-58.5; T.S. 3.1.3 K/A: 000005.EK3.03 RO 3.6/SRO 4.1 System Code: 61-OE 000001.EK3.04 RO 3.4/SRO 4.1 System Code: 61-OE . i b t . e y - .,, -.. ,_ m -,
. O O ' . RB-0176 Which of the following will be the new RCS boron concentration after RHR is placed in service? The Reactor Coolant System (RCS) is being maintained @ 325"F and 350 psig by steam dumps.
The Residual Heat Removal (RHR) System is being placed in service for normal cooldown.
Chemistry has reported that the RCS boron concentation is 2010 ppm and the RHR boron concentration is 1990 ppm.
, > , i a.
2019 ppm b.
2009 ppm
c.
1999 ppm d.
1991 ppm Answer: b Points: 1.0 References: 1-OP-14.1
K/A: 005000.K5.00 RO 3.2/SRO 3.4 System Code: 60 , -
- . . -
. O O . RB-0443 , Which one of the following is cormet? During a normal shutdown, RCS pressure and i temperatum are reduced to below 1950 psig and 543*F respectively, and the Low Pressurizer > Pressure and High Steam Flow SI signals are blocked.
If pressure rises above P-11 setpoint, the Low PZR Pressure SI is automatically [ a unblocked.
- b.
If pressure rises above P-11 setpoint, the 6pentor unblocks the Low PZR Pressure and High Steam Flow SI signals.
If Tavg increases above P-12 setpoint, the Low PZR Pressure arid High Steam Flow SI c.
signals are automatically unblocked.
. d.
If Tavg incmases above P-12 setpoint, the Law PZR Pressure.SI is automatically unblocked.
Answer: a Points: 1.0 References: NA-DW-108D014 P. 3,7; TS 2.3.3 K/A: 010000.Kl.02 RO 3.9/SRO 4.1 System Code: 49 Ol3000.A1.01 RO 4.0/SRO 4.2 System Code: 62 > . r , b t b .-v <-r -- .~,
. A, O_ . RB-0298 What will be the long term effects of maintaining the unit at the following conditions? The RCS is on a VCT float with 20% level in the PZR. An N2 blanket is being maintained on the PZR from the PRT with the PZR PORVs open. RCS wide range pressure is 22 psig and PRT pressure is 8 psig. Charging flow is 25 gpm. Seal injection is 2 gpm per RCP, and letdown is 31 gpm. VCT pressure is 40 psig.
a.
No long term adverse effects. All conditions remain stable.
b.
PZR level decreases due to difference in charging and letdown flow.
Gases come out of solution in Rx head due to high VCT gas over pressure, forming head c.
bubble.
d.
Vortexing in RHR pump impeller due to higher than normal equilibrium N2 content in RCS fluid.
Answer: c Points: 1.0 References: 1-OP-3.4 K/A: 002020.K5.06 RO 3.4/SRO 3.8 System Code: 56 . h
.. \\ O - w , - RB-0332 Which of the following is the appropriate action for this situation? You assume the watch with the unit in Mode 3 at 547*F waiting for RCS activity results from chemistry. The RCS activity prior to the first RCP start was 7.5 x 10r2 micro-curies /gm.
Chemistry informs you that the present RCS activity is 9.5 x 10 micro-curies /gm.
l Present RCS activity is greater than the initial activity, continue RCS purification until a.
initial activity is restored.
' b.
Present RCS acdvity is less than or equal to the maximum allowed RCS activity, withdraw SD banks and commence RCS dilution.
Present RCS activity is greater than the allowed maximum RCS activity, feed and bleed c.
of the RCS is required.
d.
Present RCS activity is greater than allowed by technical specifications. Continue RCS purification until less than the allowed activity limit.
Answer: b Points: 1.0 References: 1-OP-1.5; TS 3/4.8 K/A: 001000.G10 RO 3.3/SRO 3.5 System Code: 61 000076.G07 RO 2.9/SRO 3.4 System Code: OC-OE - . --
' , A ) =w RB-0451 Which one of the following is the reason foi tripping RCPs in E-0 if containment pressure has exceeded 28 psia? a.
RCPs must be secured to prevent electrical grounds due to condensation on the motor windings.
b.
CC-TV-105A, B, and C closed on Phase B isolation resulting in a loss of cooling for the RCP stator windings, CC-TV-101 closed on Phase B isolation resulting in loss of cooling to the RCP thermal c.
barriers.
d.
CC-TV-104A, B, and C dosed on Phase B isolation resultir.g in a loss of cooling for the RCP motors.
Answer: d Points: 1.0 References: 1-E-0; FM-79 Series K/A: 003000.K6.04 RO 2.8/SRO 3.1 System Code: 56 003000.G14 RO 4.0*/SRO 3.9 System Code: 56 _ $
. O
. RB-0221 Given the following data, determine the estimated critical position by performing a 1/M plot on the attached sheet: N-31 Reading Rod Height Bank _ 2.0x10' 228 A 2.2x10' 172 .B 2.4x10' 220 B i 3.1 x10' 144 C 5.7x10'
D a.
130 steps on "D" bank b.
138 steps on "D" bank c.
160 steps on "D" bank d.
168 steps on "D" bank Answer: b Points: 1.0 References: OP-1.5 K/A: 001010.K5.16 RO 2.9/SRO 3.5 System Code: 61 001010.K5.05 RO 3:3*/SRO 3.4 System Code: 61 015000.K5.05 RO 4.1/SRO 4.4 System Code: 28 015000.K5.06 RO 3.4/SRO 3.7 System Code: 28 001010.A2.07 RO 3.6/SRO 4.2 System Code: 61 .
_ _ _ _.
_____ _ _______-_--_-___ _ _ - - _ - - - - - - )SiM 2 a.
o.OO - - $ ro .- >m O< A Criticalay is projected to ocx:ur at the point that an extrapolated line drawn between two 1/M points crosses the 0.0 axis.
1.0
- bbbi
..
- :::::
~--~ ~~~~~ Time N-31 N-32 N-35 N-36 h - 1/M _____ Height Puow _..___ _____
- 1.l:::
____
- _
__ 0.8 _____ ____ ---.--
.
___-. __ ,, -_ _-_ ~_:_. : -_-_ v g 0.5 _-__- g29 ._____ do9 0.4 gm-4
---._-, _,___.. [@ _ <
_-__.
_._.
- -
, ,,
.
epuum.-umqu- -+==-peunurem m.mmums,,a-p -====us-W .--- - .. ..- ---. .-- - ---. . -- ..-- --_ - -
.-- .. ... - _ .-,-- ._-- . _.
- - --._- - _----
-----
_
g ___ __ _ ,, P "B" Bank 120 140 160 180 200 220 228 [<;; I iiii.iiiiiiiiiiiiiiiiiiiiiii;i is i ei.i i s i i i,i si i g "C" Bank 0
40
80 100 120 140 160 180 200 220 228 l 4 Steps / Increment l i6iiIiiiil1348 i I3i iiiI 6iiIiiiii Iii1 1iIi iiii iiii I MO "D" Bank 0
40
80 100 120 140 160 180 200 220 228 A
anr a. w. arms s - A =is O MOZ - .
. a a w - . RB-0175 In accordance with 1-GOP-13.1, determine time required to reach 212*F in the primary given the situation.
Sixty hours ago Unit 1 experienced a SGTR. The RCS is now at 140*F, drained to 10" above centerline. The "A" RHR pump is running and "B" RHR pump is in pull-to-lock because of a damaged mechanical seal. The "A" RHR pump trips on overturrent rendering actions to cool the RCS unsucessful.
a.
5.5 minutes b.
7 minutes c.
8.5 minutes d.
10 minutes Answer; a Points: 1.0 References: 1-GOP-13.1 K/A: 000025.G02 RO 2.7/SRO 3.7 System Code: 60-OE 000025.Gl1 RO 3.6/SRO 3.9 System Code: 60-OE 005000. A1.06 RO 2.7/SRO 3.1* System Code: 60 . , l '
. . .. bd O v , RB-0002 What is the possible consequence of not establishing Redundant Charging /SI Flows in E-1 " Loss of Reactor or Secondary Coolant," step 25 " Establish Charging /SI pump Redundant Flow Paths," following a LOCA on Unit I? A single passive failure could result in loss of charging /SI flow to the core.
a.
' b.
Possible core overheating due to boron plateout on the fuel rods at the core exit.
Limits HHSI additional makeup flow to the RCS because of more system head loss.
c.
d.
Possible HHSI pump damage caused by insufficient cooling due to operating pumps in parallel.
Answer; a ' , Points: 1.0 ? References: 1-E-1; WOG Background E-1/FM 96A Sheet 3 (SI); ESK-6EP, ESK-6EQ, ESQ-6ET K/A: 006000.K6.01 RO 3.4/SRO 3.9 System Code: 62 006000.K6.01 RO 3.4/SRO 3.9 System Code: 62 , -, f .l ' . S ? E 'l . l s . -.. -.. - - - - - - --
. A.
& - . P&O344 Without the CRDM fans in operation, which one of the following is the greatest concern given the following situation? A natural circulation cooldown is in progress in accordance with ES-0.2A, ' Natural Circulation Cooldown with CRDM Fans.' The RCS is at 510 F and 1950 psig. All CRDM fans have just tripped and cannot be restarted.
a.
Damage may occur to the IRPI coils because of overheating.
b.
Nil Ductility Temperature requirements are more likely to be exceeded for the reactor vessel head flange welds.
c.
Damage may occur to the Excore Nuclear Instrumentation because of overheating.
d.
Form 4ies of a reactor vessel head steam bubble is more likely.
Answer: d Points: 1.0 References: 1-ES-0.2 Basis K/A: 015000. A2.02 RO 3.1/SRO 3.5* System Code: 28 001000.A2.01 RO 3.1/SRO 3.7 System Code: 61 002000.K5.15 RO 4.2/SRO 4.6 System Code: 56 _ '! I .
'. e & w - ' . RB-0150 Calculate the present core exit temperature, assuming RCS pressure is 1932 psig and the ICCM indicates 42 F subcooling.
a.
586*F b.
590$F c.
670*F d.
674*F Answer: b Points: 1.0 References: Steam Tables K/A: 001000.K5.56 RO 4.2/SRO 4.6 System Code: 61 000074.EKl.04 RO 3.7/SRO 4.1 System Code: 21 - O
.. g - - . RB-0084 Given the following information, the crew should: A Main Steam Line break has occurred on Unit One. The operating crew has completed E-0 thru Step 15 when the STA informs them that an Orange Path: had existed on containment but pressure has now decreased below the Orange Path entry limit.
Transition to FR-Z.1 immediately.
a.
b.
Transition to FR-Z.1 at step 21 of E-0.
c.
Remain in E-0.
d.
Transition to FR-Z.1 at step 27 of E-0.
Ans wer: C Points: 1.0 References: OPAP-0002 K/A: 194001.A1.02 RO 4.1*/SRO 3.9 System Code: 99-OD t a . . .
. . . .. -. O O
- l
> RB-0236 (RO Only) _ l Regarding procedureal direction, should the opemtors remain in 1-ECA-2.1 or transition to ! another guideline, given the following information? l . 1-FCA-2.1, " Uncontrolled Depressurization of All Steam Generators", is in effect, SG levels ! are all between 30% and 50% NR. After the perfonnance of Attachment 5, SG Isolation, 'A'- , SG pressure starts to increase on a steady ramp from 650 to 850 psig.
.
Continue with ECA-2.1 until SI terminatiob is completed, then transition to E-2.
l a.
' b.
Tmnsition to E-2 because "A" SG is no longer faulted.
c.
Continue with ECA-2.1 until unit is in Mode 5.
! d.
Transition to E-3 because "A" SG has indications of a SGTR.
, Answer: b ' + Points: 1.0 l .i , References: 1-ECA-2.1 ! ,.; K/A: 039000.G15 RO 3.1/SRO 3.2 System Code: 42 l t i '
, . . . - . ? f-i
!
- )
. ., -. . -
s [) ("I v - . RB-0378 (RO Only) Which one of the following gives the required completion time of Protective Action Recommendations to the State given that at 1215 a Site Area Emergency was declared and the initial report to State and Local government was completed at 1227, then at 1230 the Emergency Classification was upgraded to General Emergency? a.
1245 b.
1257 c.
1300 d.
1327 Answer: a Points: 1.0 References: EPIP-1.05, EPIP-1.06 K/A: 194001.Al.16 RO 3.1/SRO 4.4* System Code: 99 194001. A1.05 RO 3.6/SRO 3.8 System Code: 99 _ w i i l ' i
s & A w w
, P"-0468 (RO Only) Which one of the following accident's severity will be increased if the RCPs are not manually tripped, in a;cordance with the continuous action page for E-0, when RCS subcooling is less then 20 F [65"F] and one charging pump running (flowing to the RCS) but then trips later in the event? a.
Small Break LOCA.
b.
Main Stream Line Rupture.
c.
RCP Locked Rotor.
d.
Large Break LOCA.
Answer: a Points: 1.0 References: E-0 Background K/A: 000009.EK3.23 RO 4.2/SRO 4.3 System Code: 62-OD - .
. -] T.. _g b ' - . . .. l0 C.s le ~ 3 3-3 00 I . 7-VIRGINI A' POWER'
- 4A"*'
' NORTH ANNA NUCLEAR TRAINING "" LICENSED OPERATOR REOUAL PROGRAM ~ ! ! NRC ADMINISTERED REACTOR OPERATOR EXAMINATION PART "A" AEO-32
f SEPTEMBER 21.1993 , , NAME: DATE: (PRINT) ~ f i INSTRUCTIONS 1.
Number of Questions
Possible Points
i >
% OR 12 - Points is passing.
2.
Time limit: I heur(s).
' , PLEDGE l ' I have neither given nor received any aid on this test.
Signature: ! I
TEST RESULTS l l
Grade:
% ! Graded by: Date: ! I Validated by: Date: (if Applicable) ! TEST Sienatures (Required for master only) ^ ' Submitted by: /Y Date: Ob-/9"l3 V Writer ()
I[I' D Approved by: Date: Supenisor i l AEO-32-KEY.
, Y ,, , - ,, - , - - - <
(ax g . , .- . . to Vireinia Power North Anna Power Station Licensed Operator Requalification Program Static Simulator Scenario Loss of RHR - Shut Down LOCA AE032
O O ^ - ' t - . _ ' STATIC SIMULATOR SCENARIO Loss of RHR - Shut Down LOCA . P Summarn The Unit was shut down 150 hours ago, and is cooling down in Mode 4 with RCS temperature approximately 205 F. The "B" charging pump and the "B" RHR pump are running.
, The "C" charging pump is tagged out for electrical maintenance. The 1H Emergency Diesel Generator is tagged out for starting air system maintenance, ' the tag for the EDG output breaker (15H2) has been lifted and the breaker is 1, racked to Test for Control Operations breaker testing.
- A loss of the IH Emergency Bus occurs as a result of the 1H normal feeder breaker 15H11 tripping open. It is reported that the breaker has been ! damaged and cannot be returned to service in a rapid manner.
' The "B" RHR pump tripped and could not be re-started, resulting in no
RHR pumps available.
The H Emergency Bus is returned to service via the alternate feed from the "B" Station Service bus.
Power range channel N-41 fails high.
1-AP-11, Loss of RHR, is used to start the "A" RHR pump. A loss of i Coolant Accident occurs due to thermal and hydraulic stresses upon restoration of flow. 1-AP-17, Sh'itdown LOCA, has been entered and is completed through step ' 12.
. SIMULATOR SETUP: [ ,
Advance charts and ensure they are inking
Ensure the correct PORV setpoint plackard is on the control board.
- Select IC-54 (AEO32_PRELOAD)
J Place SYSTEM TAGGED STICKERS the following control switches: )
15H7 "C" charging pump (normal) i 1537 "C" charging pump (alternate) 15H2 - 1H EDG output breaker .
Place a " BREAKER IN TEST" magnet on 15H2.
Run simulator 5 minutes to establish trends.
Run malfunction timer, and start the "A" RHR pump before the LOCA occurs.
) i ,
a_ O -
. Allow PRZR level to drop approx.15%, then perform the following steps of 1-e.
AP-17.
1: Increase make-up flow 2: Isolate RCS drain paths 9: Align the BIT 10: Actuate Phase A isolation , Let simulator run for approx. 2 minutes, then freeze simulator and acknowledge e the board. Do NOT allow any RCS temperature to drop below 200 degrees.
(Mode 5) ' OR Select IC-88 (Mode 5,195 degrees) e Enter the following malfunctions on the malfunction processor:
MNIO801,5 sec. TD, -100% deg., (N35 undercompensated) MMS 0102,5 sec. TD, -100% deg., (CH IV A SG steam flow xmitter failure) MRH0502,10 sec. TD, ("B" RHR pump trip) MNIO101,10 sec. TD,100% deg., (N-41 fails high) MRC04,30 sec. TD,5 sec. ramp,100% deg. (RCS pressure boundary leak) On SIMLOCH: HBUS_UVRESET=T e e Place 1H EDG mode selector switch in MAN-LOCAL Place a " BREAKER IN TEST' magret on 15H2, and PTL breaker.
- Place SYSTEM TAGGED STICKERS the following control switches:
15H7 "C" charging pump (normal) 1537 "C" charging pump (alternate) 15H2 - 1H EDG output breaker Run simulator and malfunction timer. When "B" RHR pump trips, then stop the e malfunction timer.
Manually open 15H11, then try to re-close (leave in AUTO-AFTER-START with
breaker disagreement) Restore the 1H bus from the IB bus IAW 1-MOP-6.70. (Close 15H1 and 15B11) e Allow simulator to run approximately 10 minutes, until all wide range RCS e temperatures are greater than 215 degrees.
Run malfunction timer, and start the "A" RHR pump before the LOCA occurs.
- A
& . - - ' Allow PRZR level to drop approx.15c7, then perform the following steps of 1-
c AP-17.
1: Increase make-up flow 2: Isolate RCS drain paths 9: Align the BIT 10: Actuate Phase A isolation '
- ~
Let simulator run for approx. 2 minutes, then freeze simulator and acknowledge the board. Do NOT allow any RCS temperature to drop below 200 degrees.
(Mode 5) ~ ,
. . . (G - ) . A Ilhg EVENT M ALFUNCTIONS - OVERRIDES -) ACTIONS TAKEN T=0:00 Run simulator ~i . T= 5:00 Run malfunction timer i , Start "A" RHR pump '
T=5:10 LOCA ~ IAW-1-AP-17: . Open CH-FCV-1122 I Isolate RCS drains - Align BIT Phase A isol.
i T= 7:00 Freeze the simulator and acknowledge the board VEILIFICATIONS: All RCS temperatures indicate >200 degrees.
Level in at least one SG is < 75 %. t
1-RC-MOV-1535 is closed.
1-MS-TV-109A/B indicates open.
-
' . O A( / ) v
. TURNOVER: ., The Unit was shut down for refueling 150 hours ago, and is cooling down in
Mode 4 with RCS temperature approximately 205*F.
The IH EDG is tagged out for starting air system maintenance. the Tag for
the EDG output breaker,15H2, has been lifted and the breaker racked to TEST for Control Opemtions Department testing.
A loss of the IH emergency bus odcurred when 15H11 tripped open. The
breaker is damaged and cannot be returned to service.
The "B" RHR pump tripped and could not be re-started, resulting in no RHR
pumps available.
The IH bus was reenergized from the IB station service bus.
- The "A" RHR pump was started in accordance with 1-AP-11, Loss or RHR.
, A loss of coolant accident occurred due to thermal and hydraulic stresses upon restoration of flow.
1-AP-17, Shutdown LOCA, has been completed through step 12. (See attached
procedure.)
, ' EOUIPMENT GUT OF SERVICE: IH Emergency Diesel Generator (air system maint.)
"A" charging pump in PTL,15H6 in CONNECT
"C" charging pump tagged out (electrical maint.)
"A" SG channel IV steam flow in TRIP , ,
A A . w -- Approval and Revisions Record Licensed Operator Reaualification Procram North Anna Power Station Scenario Number AEO32 Revision number 0 Scenario Title less of RIIR - 9mrDown LOCA Written by: Date submitted: ' , + v _!'k Date approved: ?/"U Approved by:f ,. / . ( . Scenario Number Revision number Scenario Title Reason for change i Submitted by: Date submitted: Approved oy: Date approved: , I Scenario Number Revision number Scenario Title Reason for change Submitted by: Date submitted: i Approved by: Date approved:
l Scenario AEO32 Page 7 Revision 0
- - - - q .
o
. .- ' 't RA-568 When RWST level drops to , then the UISI pump suction valves to the Containment sump. (Given that the accident in progress is allowed to continue.)
) 20%, will automatically swap a.
, b.
20%, should be manually aligned , , c.
23%, will automatically swap ! -
d.
23%, should be manually aligned answer: d t points: 1.0 .; e Reference: 1-AP-17/ rev. 2
, K/A: 006020.A1.09 RO 3.3/SRO 3.9 Sys. 62 006020. A4.02 RO 3.9/SRO 3.8 Sys. 62
l !
! !
7
, r '! f > a t a ' + - t 'P ' , - -- ,. . '
) . - RA-569 Tills QUESTION DOES NOT PERTAIN TO THE SCENARIO.
Which one of the following statements is correct concerning the operation of the Stub bus breaker (15H12) during an under-voltage condition? The breaker will open on UV ONLY if the associated RHR pump is running.
a.
I b.
The breaker has NQ auto-close feature.
- The breaker will open 15 seconds after a UV condition occurs, QHLY if the RHR c.
pump breaker is still closed.
d.
The breaker will automatically close 15 seconds after a UV clears ONLY if the associated RHR pump breaker is open.
Points: 1.0 answer: d Reference: ESK 5AQ K/A: 062000.A2.04 RO 3.1/SRO 3.4* Sys. 06 062000. A2.05 RO 2.9/SRO 3.3* Sys. 06 005000.K2.01 RO 3.0/SRO 3.2 Sys. 60 l
, -I l ! ' i
A S - - - . RA-570 Which one of the following alternate core cooling methods is preferred? (Given the current conditions.)
a.
Natural circulation b.
Reflux boiling c.
Forced feed and spill d.
Gravity feed and spill answer: a points: 1.0 Reference: 1-GOP-13.0/Rev.1 K/A: 000025.EKl.01 RO 3.9/ SRO 4.3 Sys. 60
O O - - . RA-574 Which one of the following statements would be correct conceming the operation of the charging pumps, if an operator were to rack breaker 1537 for the "C" charging pump to CONNECT! "B" charging pump continues to mn, "A" and "C" pumps could be manually staned, a.
' b.
"B" charging pump trips, only "A" pump could be manually staned.
"B" charging pump continues to mn, "A" pump could be manually staned.
c.
d.
"B" charging pump trips, NO charging pumps could be manually staned.
answer: b points: 1.0 reference: ESK 5AL,5MN,5AN; LER 90-11 K/A: 006000.K4.05 RO 4.3/SRO 4.4 Sys. 62 , . -. -
, .. - ~ O O -
. a RA-575 J l Which one of the following statements is true conceming the actions required to re-start the spent fuel pit cooling pump,1-FC-P-1A? (Given that 1-FC-P-1A was running at the start of this event.)
None, the pump should have automatically started 30 seconds after IH bus voltage - 'j a.
was restored.
, b.
The pump will start if the contml switch is placed in START.
, The control switch must be placed in stop to reset the UV signal, then placed in c.
START.
d.
The UV signal must be reset by locally cycling the breaker, then place the contml ] switch in START.
answer: b points: 1.0 references: ESK 6GH K/A 033000.K3.03 RO 3.0/SRO 3.3 Sys. 67 033000.A2.02 RO 2.7/SRO 3.0 Sys. 67 i . . n & J b
'. ()
v - v . , RA-577 What was the cause for the steam generator blowdown radiation monitor alarms? - (Note: These alanns were noticed while the operator was checking the status of the Radiation. Monitoring System in accordance with 0-AP-10.)
a.
Containment phase A isolation t , b.
Steam generator tube leakage i c.
Loss of electrical power d.
Improper calibration ANSWER: C reference: Loop diagram /RM-16 K/A 073000.G08 RO 3.3/SRO 3.3 Sys. 54
, .: ,
>
a
- - RA-578 Annunciator C-E4, LOW TC INITIATE NDT PROTECT, is in alaiTn. What is the cause of this alarm? Equipment failure; alarm should NOT be locked in, a.
b.
RCS pressure is high and the PORVs have failed to open.
c.
The PRZR PORV control switch should lie in OPEN.
- d.
PRZR PORV block valve 1-RC-MOV-1535 is closed.
ANSWER: d reference: AR C-E4 K/A: 000008.EA2.05 RO 3.9/SRO 3.9 Sys. 56 0100000.G08 RO 3.5/SRO 3.5 Sys. 49 0100000.G12 RO 3.6/SRO 3.6 Sys. 49
,. O.
l,_) . " RA-581 Which one of the following statements would be correct, if this scenario were allowed to continue with NO operator actions? The RHR pump suction relief valve will open.
a.
b.
The RHR pump suction valves will close.
ONLY one PRZR PORV will open.
I c.
d.
Both PRZR PORVs will open.
answer: d - reference: AR C-F4 K/A: 005000.A2.02 RO 3.5/SRO 3.7 Sys. 60 010000.Kl.02 RO 3.9/SRO 4.1 Sys. 49 ) -
c.
' C's O \\_j - . RA-582 Which one of the following statements is correct concerning CC Dow to the RCPs? (Annunciator C-C4, RCP 1ABC THERMAL BARRIER CC HI/LO FLOW, has alarmed.)
a.
CC flow still exists, ONLY indication of flow has been lost.
b.
CC flow was lost when the H bus deenergized, CC Dow was lost when containment phasd A isolation was actuated.
c.
d.
CC flow was isolated due to a thermal barrier heat exchanger leak.
. answer: b reference: ESK 6MC; load list K/A: 003000.A4.08 RO 3.2/SRO 2.9 Sys. 56 003000.G12 RO 3.4/SRO 3.5 Sys. 56 000026.G09 RO 3.3/SRO 3.4 Sys.12
, ' O O w w . RA-583 What was the cause of annunciator J-H4, RES SS XFMR OVERLOAD /CKT TROUBLE OR LS IN DEFEAT alanning? The "H" emergency diesel generator is tagged out.
a.
b.
One of the RSS transformers is overloaded, c.
' Die "H" bus alternate feeder breakers are closed.
i d.
Imad shed is i. '4' FEAT.
answer: d reference: AR J-H4; 0-OP-26.7/rev. 2 K/A: 062000.G12 RO 3.1/SRO 3.1 Sys. 06 062000.G08 RO 3.1/SRO 3.1 Sys. 06
r.
&,w O
. - . ~ RA-0588 THIS QUESTION DOES NOT PERTAIN TO THE SCENARIO.
Complete the following statement concerning reactor coolant loop flow indication. ~ (Given that all three reactor coolant pumps (RCPs) are initially running.)
If "C" RCP trips then indicated flow in the "C" 160p will , while indicated flow in the "A" and "B". loops will - . decrease to zero then increase slightly; increase a.
b.
decrease to zero then increase slightly; decrease c.
decrease to zero and stay at zero; increase d.
decrease to zero and stay at zero; decrease Answer; a Points: 1.0 References: HT-FF K/A: 002000.A1.05 RO 3.4/SRO 3.7 Sys. 56
c _ .
- a
e- .o o - . ., ' RA-144 (RO ONLY)
Which one of the following actions is REQUIRED in order to OPEN MOV-1380 and MOV-1381 (seal water return isolation valves)? (Given current plant conditions.)
a.
Reset SI, Depress OPEN pushbuttons, b.
Reset Phase "A", Depress OPEN pushbuttons.
< c.
Reset SI, Reset Phase "A", Depmss OPEN pushbuttons.
d.
No action required except depressing OPEN pushbuttons.
. ANSWER: b POINTS: 1.0 ' REFERENCES: NA-DW-5655d33/sh. 8 KA 003000.A2.01 RO 3.5/SRO 3.9 Sys. 56 l
, I E
E-O O - ?,
f
.RA-388'(RO ONLY) Which one of the following correctly describes the Hi Steam Flow SI actuation signal if the " Reset - Train A" switch was taken to " RESET"? (Given current plant conditions and ' ! assuming SSPS fuses are installed.)_ i a.
Would be enabled - SI would NOT occur.
b.
Would NOT be enabled since Tavg is <$43*F.
. i , c.
Would be enabled - SI would occur.
, d.
Would NOT be enabled since BOTH reset switches must simultaneously be taken to i " RESET".
, ANSWER: a
. POINTS.: 1.0 REFER 11NCES: NA-DW-5655D33/sh. 7 , KA 012000.K6.04 RO 3.3/ SRO 3.6 Sys. 57
,a i ,
- - - - - _ _ -__ _ __ ____ - _ 4- _ , & / ,4-O v... 6 /q.
o.
&o IMAGE EVALUATION // A th . 9 A (7
//g i / ~ ' '$f' j {d e{g
f' <! TEST TARGET (MT-3) v #3 4 / e N %. s.43, ,# \\ &r \\ S,h#' k, $[D'(f \\ VW 1.0 d" " _" m.... - g L l,l k w, 0lll I.8 mm I.25 ] =m.4 l l e m.6
i s , - ..... > 150mm 4--- p 4--__-_-___.
__ s~ +,;x , A*s: s e w,
- g/,/
, - &gs,DI Ny a //x
Z J.
.s ,x o, o - - , ' t.# / Ih y l e , ' , 7, Oy,j' qy , , - ji i - -_ ,,
___-____ ' g e .% t +}. $' -g <.49 I ,e '4 q . 7 e.
- T l* 'a IMAGE EVAL.UATION
- //k'^~[g,;? " f
Nff ' 4 If TEST TARGET (MT-3) R 4lff N ///j -:' //
& y Y %h' R4'k% 4'
- ' 7 8 l25 1.0
- - m n, Kna- ,y $" l ',.
i.i . _=_ Ill I.8 IH= L l.25 ! l ____ l l.4 !r i.6! eme = " 150mi.
- - - - - - - - 4______--.___ ._.-_..-----------6" -
4____ y x,, 4> +//s~ ., mp /e /F.
- - 4$,4 s , q~ - 76\\ 'es .: c, /.f en - ,7. _._ s .- g,t,7 \\pf,';,, "<;; , ' U ' g//j/, / qq e j S ._ ._
%;; I.
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a
' i' TEST TARGET (MT-3) T6 <f Qf {g IMAGE EVALUATION /f s \\;g/ 4, q N## ft 4, I h % f ,
E kt l $ \\0\\\\Eu l - - l.I e e p=a
- c
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> = = A 1.25 1.4 1 1.6 _ __ i _ l t I l ! {
150mm >
I
- - 6" >
h i kb'%, 4?4 a g g p //// y,n s,pa., -- , _ - , . g_.
, . ,- ~ '#,
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- j
): m
" (RA-572 '(RO' ONLY) . i P Which one of the following statements cormctly describes the response of the Electrical j Distribution system if the normal _ feeder breaker to the IB station semce bus (15B2) were to ' trip open due to a fault on the IB station service bus? _ , 15B1 would automatically close;~the IB and IH buses would remain energized.
q - a.
15B11 and 15H1 would open; 15B1 woul' automatically close; the IB bus would be b.
d [ re-energized; the IH bus would be re-energized by the EDG.
! , c.
15B11 and 15H1 would open; the IB and 1H buses would be dead.
- 'd 15B11 and 15H1 would open; 15B1 would automatically close; the IB bus would be ' ' . re-energized; the IH bus would remain de:,d.
- ! - answer: c !
- points: 1.0 reference: FE-21G -! K/A: 062000.K4.03 RO 2.8*/SRO 3.1 Sys. 06 062000.A2.04 RO 3.1/SRO 3.4 Sys. 06 ! 062000.K1.04 RO 3.7/SRO 4.2 Sys. 06 i q >
, , ! E
l f I _f'. V , i
-.
2
- -
i ', C')
o - ' RA-580 (RO ONLY) N-31 and N-32 did NOT automatically re-energize because . (Given that the operating crew has not attempted to manually reenergize the source range nuclear instruments.)
intermediate range NI, N-35, is undercompensated.
a.
b.
power rangc N1, N-41, has failed high.
c.
the steam flow channel is in trip.
d.
Intennediate range NI, N-36, is overcompensated.
answer: a reference: NA-DW-5655D33/ sh. 3 K/A: 000032.EA2.02 RO 3.3/ SRO 3.6 Sys. 28 000032.EA2.05 RO 2.9*/ SRO 3.2* Sys. 28 000032.EA2.08 RO 3.3/ SRO 3.4 Sys. 28 l .
' .a.
O O w w Nash n 93-3 n VIRGINIA POMTR NORTII ANNA NUCLEAR TRAINING "" M 0 " " A QL LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATION PART "A" AEO-10 AUGUST 25.1993 NAME: DATE: (PRXf) INSTRUCTIONS 1.
Number of Questions
Possible Points
80 % OR 12 Points is passing.
2.
Time limit:
hour (s).
PLEDGE I have neither given nor received any aid on this test, Signature: TEST RESULTS Grade: / % Graded by: Date: Validated by: Date: (If Applicable) TEST Sienatures (Required for master only) / f ) Submitted by: f.'2 Nr b ~ Date: ofh / 7 43 ' / Writer () Approved by: - k@x Date: Y[I7['D Supervisor AEO-10-KEY
. _ ~ O O ' _ . . VIRGINIA POWER NORTII ANNA NUCLEAR TRAINING LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATION PART "A" AEO-10 AUGUST 25.1993 , NAME: DATE: (PRINT) -
INSTRUCTIONS 1.
Number of Questions
Possible Points
,
% OR 12 Points is passing.
2.
Time limit: I hour (s).
PLEDGE
I have neither given nor received any aid on this test.
Signature: TEST RESULTS Grade: / % Graded by: Date: ' Validated by: Date: (If Applicable) TEST Siznatures equired for master only) Submitted by: // Date: of-n 93 /* Writer V Approved by: C h N/7b3 Date: Supervisor AEO-10-X . . - .-.--
. - ,. O O + Vircinia Power North Anna Power Station Licensed Operator Requalification Progen [
. Static Simulator Scenario Loss of Offsite Power with EDG Failure , AE010 >
- l , I l
i >
.. A e w w - STATIC SIMULATOR SCENARIO Loss _of Offsite Power with EDG Failure Summarv: Unit 1 is operating at 100% power with a core bumup of 4000 MWD /MTU when a loss of offsite power occurs with the subsequent failure of 'lH' EDG to remain on line after starting.
SIMULATOR SETUP: Select IC-1 (100% Power) - - Advance chans and ensure they are inking Ensure VCT level >42% - Ensure FCV-Ill3A is in manual with a controller output of 31.5 % - Enter the following malfunctions on the malfunction processor: - MELO), 60 sec. TD MRC1602, I sec. TD, I sec. ramp, 85 % deg, RX23 trigger Switch Override FWP3B_STOP=ON FWP3B_ START =OFF FWP3B_ASTOP=OFF FWP3B_ASTART=OFF - Allow simulator to run 10 minutes to establish trends.
After coine to freeze: RR_ RAD (3) set to zero RMR173I set to zero XRMPROC=F Go to mn for one second and back to freez n e a w w . TIME EVENT MALFUNCTIONS OVERRIDES ACTIONS TAKEN T=0 T = 30 sec. on 'H' EDG trips on overspeed ON SIMLOCH; MEL01 timer EDGH_OVSPD_'IRP=T T=3 min an MEIl)1 timer Manually open Reactor trip switch T=5 min. on MELOl timer Place light switch , LCl in instnictor booth in hand VERIFICATIONS: 'B' AFW pump did not start VCT level > 5 % Ensure instruemnt air has recovered and RM trip valves are open.
Ensure 1-FW-F1-100A is >340 GPM and not pegged high
a s % i " . TLTRNOVER: The unit has been running at 100% power for three months when a loss of offsite power occurred 6 minutes ago for reasons yet unknown. 'H' 4160 V Bus voltage has not been restored. ONLY THE BihiEDIATE OPERATOR ACTIONS OF E-0 HAVE BEEN PERFORMED.
Core Burnup is 4000 MWD /MTU.
EOUIPMENT OUT OF SERVICE: None.
i
L,7
RA-0156 !
What is the concem one would have if there was no red light indication for B charging pump ! with amps indicated? (Assume light bulb is NOT burned out.)
. No positive way of verifying the pump is mnning.
a.
i b.
Trip Coil is not functional.
, Could cause emergency bus overload due t'o failure to strip from bus on an undervoltage c.
signal.
, d.
Breaker closing coil is not functional.
Answer: b ] Points: 1.0 References: Simulator Configuration; ESK-5AM l K/A: 063000.K3.02 RO 3.5/SRO 3.7 Sys. 06 004020.A4.02 RO 3.7/SRO 3.3 Sys. 09
$ i . ' i
) I
i l .
, O.
d . RA-0159 What will be the msponse of "A" CC pump if "H" 4160V bus is restored? (Assume the Abnonnal Procedum was not used to restom "H" 4160V bus.)
Pump will automatically restan after a 20 see time delay.
a.
b.
Pump will automatically restan after manually resetting the stub bus.
t Pump will automatically restan after 15 see time delay.
c.
d.
Pump will remain locked out.
Answer: c Points: 1.0 References: ESK-5P K/A: 008000.A3.01 RO 3.2*/SRO 3.0" Sys.12 062000.K3.01 RO 3.5/SRO 3.9 Sys. 26 , l l
.. a O ^ . ~ . I ~ - i RA-0161 l r Which one of the following must be reset in order to shutdown the "A" AFW pump? (Given > that "H" bus is restored with "A" AFW pump running.)
a.
Safety injection.
, , b.
Feedwater isolation. (if one should occur), i c.
AMSAC.
i ' d.
Under voltage on "H" Emergency Bus.
Answer: c . ! ' Points 1.0
References: ESK-5AA , t K/A: 061000.K4.06 RO 4.0*/SRO 4.2* Sys. 04 ' ! ! I l s l .
I i l
l i l l
. .. - - - - . _ .. .. .. _.
. -. _ _ - _ _ _. v
l .- l RA-0162 Which one of the following statements is tme if N-35 (Intermediate Range NI) is ~substantially I over compensated.
N-35 will remain higher than P-6; Source range re-initiation will occur when N-36 falls a.
below P-6.
. b.
N-35 will be lower than P-6; Source range re-initiation will not occur as N-36 falls - below P-6.
. N-35 will remain higher than P-6; Source range re-initiation will not occur when N-36 { c.
falls below P-6.
' .; d.
N-35 will be lower than P-6; Source range re-initiation will occur as N-36 falls below
P-6.
Answer: d points: 1.0 , References: 1-AP-4.2; NA-DW-5655D33/Sh. 3 i K/A: 000033.EA2.ll RO 3.1/SRO 3.4 Sys. 28-OE 012000.K6.10 RO 3.3/SRO 3.5 Sys. 57
i ! h
. I h . ! i , L ! '! i r i i f e I i ... . - - .. .. .
. . . -. , . . . .
RA-0166 Select one of the following which best completes the statement. RM-RMS-159 Low Flow alarm is locked in due to: a.
Phase "A" isolation signal.
b.
Hi-Hi radiation alann.
, i c.
loss of power to the associated sample pump.
. d.
Closure of RM-TV 100D.
Answer: c , ! Points: 1.0 .! References: Load List; Loop Diagram RM-159- ' ! K/A: 073000.A2.01 RO 2.5/SRO 2.9* Sys. 54 ' 073000.A4.02 RO 3.7/SRO 3.7 Sys. 54 l ,
.
L F -> t
.
O v w . RA-0167 The following action will msult in which one of the below respcmses: The Shift Supervisor, desiring to have two operable charging pumps running on the "J" 4160V bus, crders the control switch for 15H7 placed in P-T-L and directs an opentor to rack out 15H7 and mck in 15J7 under current plant conditions, Auto start of "A" charging pump (bmaker closure).
a.
b.
All three charging pump bmakers closed.
c.
No charging pumps running.
d.
"C" charging pump mnning on its alternate breaker.
Answer: c Points: 1.0 References: ESK-5AL; ESK-5AM; ESK-5AN K/A: 004000.K2.02b RO 3.3/SRO 3.5 Sys. O'1
' ~ ' ~' ' . ? , - i RA-0169 TIIIS QUESTION ' DOES NOT PERTAIN TO THE SCENARIO.
- Which one of the following statements is true concerning system response if breaker 15A2 ("A" Station Service (SS) nonnal feeder bmaker) was inadvertently opened by the operator. (Given that the event occured while the unit was at 100% power condition.)
, i "A" 4160V SS would swap to "A" Reserve Station Service (RSS) feeder breaker, "B" a.
SS and "C" SS busses would not swap to their mspective RSS bus feeder breakers.
i b.
"A", "B" and "C" 4160V SS feeder breakers would swap to their respective RSS bus , feeder breakers.
"A" 4160V SS feeder breaker would not swap to "A" RSS feeder breaker; "B" SS and ' c.
"C" SS busses would swap to their respective RSS bus feeder breakers.
' d.
"A", "B" and "C" 4160V SS feeder breakers would not transfer to their respective RSS feeder breakers.
Answer: d . Points: 1.0 . References: FE-21G j K/A: 062000.K4.03 RO 2.8*/SRO 3.1 Sys. 06 i l i l l -- -- - ,
.- ... . - . .. O Os -
. ' RA-0172 ,
i Note that a "B" Loop AT/Tavg has failed. Such a failure with the UNIT'AT 90% POWER would result in which one of the following? a.
OT AT setpoint increasing.
b.
OP AT setpoint decreasing.
. , ! P-12 interlock met prematurely.
l c.
d.
Rod insertion limit decreasing.
Answer-b Points: 1.0 .
References: Simulator; TS 3.1.3.6; Curve Book K/A: 002000.A1.09 RO 3.7/SRO 3.8 Sys. 56 001050.A1.01 RO 4.07/SRO 4.27 Sys. 61 ' o j , I ' __ . - . . - -
.. _ . . RA-0588
THIS QUESTION DOES NOT PERTAIN TO THE
SCENARIO.- l Complete the following statement concerning reactor coolant loop flow indication. (Given that
all three reactor coolant pumps (RCPs) are initially mnning.)- i .
If "C" RCP trips then indicated flow in the "C" loop will , while indicated flow in the "A" , and "B" loops will - . decrease to zero then increase slightly; increase a.
i b.
decrease to zero then inemase slightly; decrease
c.
decrease to zem and stay at zero; ir ease i
d.
decmase to zero and stay at zero; decrease p Answer; a
Points 1.0 Refemnces: HT-FF K/A: 002000.A1.05 RO 3.4/SRO 3.7 Sys. 56 ) ) ! l I _
" -s v - . RA-0178 Which one of the following pieces of equipment is in a condition that WOULD NOT be expected for current plant conditions? FW-P-3A ("A" Auxiliary Feedwater Pump).
a.
b.
F\\WP-3B ("B" Auxiliary Feedwater Pump).
CH-P-I A ("A" Charging Pump).
c.
d.
FW-P-lC1/lC2 ("C" Main Feed Pump).
Answer: b Points: 1.0 References: Simulator K/A: 061000.K4.06 RO 4.0*/SRO 4.2* Sys. 04
r-a S - - . RA-0180 Choose the statement below which correctly reflects CC flow rate through the "A" RCP Thermal Barrier.
a.
Cannot be determined with event in progress, b.
No flow due to loss of the IH emergency, bus.
c.
Approximately 21 GPM.
d.
No flow due to a thennal barrier heat exchanger leak.
Answer: b Points: 1.0 References: Simulator; FM-79B/Sh. 2; loop Diagram CC-64; l-AP-15 K/A: 008000.A3.01 RO 3.2*/SRO 3.0* Sys.12 003000.A4.08 RO 3.2/SRO 2.9 Sys. 56 -
- a a
w w . RA-0181 With the current "B" Loop temperature indication failure, is the position of FCV-1122 correct? No,100% full open should be indicated due to loss of power to 1-El-CB-56.
a.
b.
Yes, Tave has shifted to Io-Select.
- Yes, the median selector circuitry has compensated for its failure.
c.
d.
No,100% full open should be indicated.
Answer: c Points: 1.0 References: Loop Diagrams CH-1, RC-68 K/A: 002000.A1.09 RO 3.7/SRO 3.8 Sys. 56 0020C0.Kl.06 RO 3.7/SRO 4.0 Sys. 56 i
.e- ~ O O . -
,- i RA-0483 THIS QUESTION DOES NOT PERTAIN TO THE SCENA.RIO.
Select the one correct statement concerning operation of the Steam Generator Water Level ! ~ Control (SGWLC) system. (Assuming the unit is stable at 50% power.)
I ,
If Channel 2 Level Transmitter on "B" S/G fails low, feed flow to "B" S/G will initially a.
i decrease.
' i b.
If Channel 3 Level Tmnsmitter on "C" S/G fails high, actual level in "C" S/G will decrease.
t If the selected channel of Feed Flow for "A" S/G fails low, "A" S/G level will decrease.
i c , d.
If the Steam Pressure input to "B" Steam Generator Steam Flow fails "as is" and then I Unit power is increased from 50% to 100%, indicated steam flow will be less than actual i steam flow.
ju ! Answer: b i Points: 1.0 ' i References: NA-DW-5655D33/Sh.13 , K/A: 059000.A2.11 RO 3.0*/SRO 3.3* Sys.39 i 039000. A1.06 RO 3.0/SRO 3.1 Sys.4' t ! i i ! ! ' , ! ! E ! }
e
i
t . -. -. . .. -- .. '
e O O w - . SA-0519 (SRO ONLY) ' Classify the event in accordance with EPIP-1.01. (Assume that the present plant conditions have existed for 20 minutes.)
a.
Notification of Unusual Event.
b.
Alert.
t c.
Site Area Emergency.
I d.
General Emergency.
Answer: a Points: 1.0 ' References: EPIP-1.01 K/A: 000055.G02 RO 2.9/SRO 4.l* Sys. 06-OE
a ,.- e u .. SA-0520 (SRO ONLY) What actions must be taken given the following? Utilize current plant conditions e The "B" Boric Acid Transfer pump can not be placed in service e a.
Place the unit in HOT SHUTDOWN within the next 6 hours and at least COLD SHUTDOWN within the following 24 hours.
b.
Restore a flowpath from BAST within 72 hours or be in at least HOT STANDBY and borated to a SDM of 1.77% delta K per K at 200 F within the next 6 hours.
c.
Restore Dowpatn from RWST within one hour or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
d.
Suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one bomted soume is restored to operable status.
Answer: b Points: 1.0 References: TS 3.1.2.2 K/A: 003000.G03 RO 2.6/SRO 3.9" Sys. 56 . i l . - - - - - - -
_ ' o a ' /Vl o.c le n 9J-Jet VIRGINIA POWEll ' " ' NORTII ANNA NUCLEAR TRAINING ' A E Gu r? L LICENSED OPERATOR REOUAL PROGRAM , NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATION OPEN REFERENCE PART "B" ' AUGUST 25.1993
! l ' ' NAME: DATE: (PRINT) ! INSTRUCTIONS ! 1.
Number of Questions
Possible Points
80 % OR 16 Points is passing.
2.
Time limit:
hour (s).
l . PLEDGE t I have neither given nor received any aid on this test.
. b Signature:
TEST RESULTS Grade: / % Graded by: Date: l Validated by: Date: (If Applicable)
TEST Sienature (Required for master only) Submitted by: / / N __ Date: @ /7-63
- ~
Wnter Q 'd[/'7[G Approved by: '9 Date: ~ Supervisor PART B-13Y i i
e e - - . VIRGINIA POWER NORTH ANNA NUCLEAR TRAINING i LICENSED OPERATOR REOUAL PROGRAM NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATION OPEN REFERENCE PART "B" ' AUGUST 25.1993
NAME: DATE: (PRINT) , INSTRUCTIONS 1.
Number of Questions
Possible Points
80 % OR 16 Points is passing.
2.
Time limit:
hour (s).
PLEDGE I have neither given nor received any aid on this test.
Signature: TEST RESULTS Grade: / _% G mded by: Date: Validated by: Date: (If Applicable) TEST Sinnatures (Required for master only) [1' <>\\,- ' Submitted by:, t Date: CB - / 713 / Writer () Approved byb_ k Date: Y [7 M Sujiervisor ' PART B-X !
a & w - . RB-0480 Which one of the following actions must the crew take if while performing E-3 " Steam Generator Tube Rupture", (due to a tube rupture in "C" S/G), it is found that "C" MSTV and NRV cannot be closed? a.
close all remaining MSTVs, NRVs and bypasses and cooldown with intact S/Gs atmospheric relief valves.
b.
close all remaining MSTVs and cooldown with the ruptured S/G atmospheric relief valves.
cooldown by dumping steam from "A" and "B" S/Gs through the AIAV pump terry c.
turbine.
d.
tmnsition to ECA-3.1 "SGTR with loss of Reactor Coolant - Subcooled Recovery" Answer: a Points: 1.0 References: 1-E-3 K/A: 000038.EA2.08 RO 3.8/SRO 4.4 System Code: 42-OD 000038.G12 RO 3.8"/SRO 4.0-System Code: 42-OD i
,
. , q .. RB-0518 l How does cycling Pzr level enFance Upper Head. Cooling in ES-0.4 Natum! Circulation ,, Cooldown with Steam Voids in Ve,sel (Without RVLIS)? The operator is reminded by a note preceding step 9 that, "In order *.s continue overall system depressurization, cycling PZR level (cycling pressure) may be necessary to enhance Upper Head cooling.
I As the pressurizer level is cycled, the heat transfer from the pressurizer metal is greater, a.
thus causing the pressurizer upper head to cool faster.
a
b.
As pressurizer pressure is increased, the water in the PZR will be forced out, collapsing / i the steam bubble in the reactor upper head, thus cooling upper internals.
As pressure is increased, the amount of RCS subcooling also increases, so that heat c.
transfer rate from the upper head to the RCS willinemase.
> d.
As pressurizer pressure lowers the water in the reactor vessel will flash to steam, , removing the latent heat of condensation from the water in the upper head region, j resulting in better upper head cooling.
Answer: b + Points: 1.0 References: ES-0.4 Background Document; 1-ES-0.4 K/A: 000009.EK3.10 RO 3.4/SRO 3.6 System Code: 62-OD 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-O D 000009.EKl.01 RO 4.2/SRO 4.7 System Code: 62-O D
O O w - . RB-0467 Which one of the following is the basis for closing the charging pump recirculation valves when RCS pmssure is below 1275 psig and the RCPs are tripped in accordance with the Continuous Action Summary of E-0? protect the Charging Pumps from runout operation.
a.
' b.
avoid Charging Pump cavitation.
. c.
maximize SI flow.
d.
prevent damage to the VCT Relief Valve due to overfilling the VCT.
Answer: c Points: 1.0 References: E-0 Background K/A: 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-OD 006000.K5.06 RO 3.5/SRO 3.9 System Code: 62
. A A - -
RB-0459 Why is there NO requirement on pressurizer level for staning an RCP, while in I-ECA-1.1 - Loss of Coolant Recirculation, with RCS subcooling greater than 25 F {75'F}? A&quate subcooling ensures that pressure is high enough to pre,cnt DNB from a.
occurring as the RCP is started.
b.
Pressurizer level is assured to be great enough to support RCP operation at this point in 1 -EC A-1.1.
Only adequate subcooling is required since it is likely pressurizer level will be offscale c.
low for this transient.
d.
Pressurizer level requirements for staning an RCP only apply when inadequate RCS subcooling exists.
Answer: c Points: 1.0 References: 1-ECA-1.1 Background Rev. IB EOP Obj D K/A: 000011.EK3.12 RO 4.4/SRO 4.6 System Code: 62-OD I I
1
e
- -
RB-0543 Which one of the following is correct if the crew has been unsuccessful in reducing core exit thermocouples to < 1200*F and FR-C.1 directs the operators to stan RCP(s) in an attempt to lower core exit temperatures? (Note: RCP support condition / systems cannot be established.)
The first RCP may be started without support conditions, but support conditions must be a.
available prior to staning additional RCPs b.
RCPs should not be staned without suppoit conditions. If RCPs were started without support conditions then severe RCP damage would result requiring extensive RCP outage time to repair.
RCPs may be started without suppon conditions being available provided permission is c.
obtain.<l from the Station Emergency Manager (SEM).
d.
RCPs should be staned even if support conditions cannot be established. RCPs may provide temporary cooling of the core even under highl voided conditions.
Answer: d Points: 1.0 References: 1-FR-C.1; FR-C.1 Background Document K/A: 000074.EK3.07 RO 4.0/SRO 4.4 System Code: 62-OD
. _ _ _ _.
. O O - ' RB-0572 i i Why should the operator continue with RCS depressurization un 'il pressurizer level is greater - ! than 36% [50%) in accordance with step 29 of 1-AP-17 (Shutdown LOCA), if they notice RCS i subcooling has lowered to 0*F7 l After pressurizer level is restored, the depmssurization will be stopped and RCS pressure a.
will then begin to rise, which will restore the required subcooling.
( i
b.
Subcooling will be restored by the cooldown after the depressurization is stopped, c.
Subcooling is not a concern because boiling is not a concern during shutdown conditions.
' d.
By increasing the pressurizer level, the head of water above the reactor will restom the , RCS subcooling.
' Answer: b l Points: 1.0 - , References: ARG 2; l-AP-17 i , ! K/A: 000009.EK3.26 RO 4.4/SRO 4.5 System Code: 62-O D l t E ,
!
i
i ) % . ! . 3,
y ,,., ,, w '+-w--- ' ~ - m-
= v
'
I v -
RB-0302 Ilow close is the pressurizer to being solid if: RCS ternperature is 150 F and pressure is 320 psig. PZR temperature is 350*F. PZR level is indicating 100% on Protection Channels I, II, and III and 85% on the Startup level indicator. Assume Unit 1 is being cooled down and depressurized for a refueling outage. The PZR is being filled to a solid condition in preparation for hydrogen peroxide addition.
a.
The PZR is solid.
. b.
The PZR is 4% below a solid condition.
c.
The PZR is 13% below solid condition.
d.
The PZR is 15% below solid condition.
Answer: b Points: 1.0 References: 1-OP-3.4; l-SC-23.1 K/A: 002000.A1.02 R0 3.6/SRO 3.9 System Code: 56
O O - - . RB-0098 Which one of the following is required to be performed if it is desired to continue to cooldown given that the following has just occured? While performing a Unit 1 cooldown in accordance with 1-OP-3.2, the operator performs the following as Tavg goes below 543*F: Blocks both trains of High Steam Flow SI
Arms the two cooldown steam dumps by going to ' bypass interlock'
Then Tavg is allowed to increase to 546*F :md then is lowemd again to 543 F.
Neither action is required to be performed again.
a.
b.
Block the High Steam Flow SI again, the cooldown steam dumps are still available.
Use " Bypass Interlock" to restore the cooldown steam dumps, the high steam flow SI c.
remains blocked.
d.
Both actions are required to be perforned again.
Answer: d Points: 1.0 References: NA-DW-108D014 Sh. 2 K/A: 013000.K4.12 RO 3.7/SRO 3.9 System Code: 62 Ol3000.K4.03 RO 3.9/SRO 4.4 System Code: 62 041020.K4.09 RO 3.0/SRO 3.3 System Code: 42
,. a
- -
RB-0018 Why is it not acceptable for the OATC to allow a ' Step 6' opentor to perform a Rx stanup as part of his Rod contml system check out? a.
The operator has not completed the step pmgram.
b.
The operator has never attended a reactor, operator license class, c.
The operator has not obtained the shift supervisor's permission.
d.
The operator is not enrolled in a reactor operators license class.
Answer: d i
Points: 1.0 References: 10-CFR-55-3, VPAP-1404 K/A: 194001.G29 RO 3.1/SRO 4.7 194001.G23 RO 2.8/SRO 3.5 194001.G31 RO 3.1/SRO 3.1 001000.G01 RO 3.7/SRO 3.8 System Code: 61 i ) J - -.
. O a_ i RB-0173 When may another RCP stan attempt be made given the following? At 1500 the "A" RCP tripped due to a faulty relay. The relay was replaced and at 1615 the operator staned the "A" RCP, but then inadvenently stopped it as it was reaching full speed. Upon the shift supervisors request the operator again attempted to stan "A" RCP at 1645 which also proved unsuccessful.
a.
1715 b.
1745 c.
1815 d.
I845 Answer: a Points: 1.0 References: 1-OP-5.2 K/A: 003000.G10 RO 3.3/SRO 3.6 System Code: 56 003000.K6.14 RO 2.6/SRO 2.9 System Code: 56 ) . i l
. n O m -
RB-0380 Which one of the following requires activation of both the TSC and OSC? a.
any emergency plant classification, b.
any emergency condition classified higher than notification of unusual event.
only emergency condition classined as either a site area emergency or general c.
I emergency.
d.
only emergency conditions classiGed as general emergency.
Answer: b Points: 1.0 , References: Emergency Plan K/A: 194001.A1.16 RO 3.1/SRO 4.4* System Code: 99-OC l
l l l
. A e w -
RB-0261 Select the correct response by the md control system given that unit 1 is operating at 100% power with control rods in automatic and loop "B" cold leg temperature fails high? a.
Control rods will initially step IN when instmment fails but will cease movement once instrument is pegged high.
b.
Control rods will initially step IN at 72 steps per minute.
c.
Control rods will step in and stop when Tavg decreases 1.5 F.
d.
No control rod motion will occur at all.
Answer: d Points: 1.0 References: PLS C.2; NA-DW-108D014 Sheet 1 of 17 K/A: 001050.K5.01 RO 3.3/SRO 3.6 System Code: 61
a a w -
RB-0566 What will be the new RCS bomn concentration if while placing RHR in service on Unit 1, RIIR boron concentration is 1900 ppm and RCS boron concentration is 2200 ppm? a.
1918 ppm b.
1953 ppm c.
2028 ppm < d.
2182 ppm Answer: d Points: 1.0 References: 1-OP-14.1 K/A: 005000.K5.03 RO 2.9*/SRO 3.1* System Code: 60 005000.Kl.09 RO 3.6/SRO 3.9 System Code: 60 .
. . . . . i ' O O
i i i RB-0574 f Which one of the following statements best describes the required actions, if a review of the most recent test results indicate that the final "as-left" setting for the intermediate range neutron ' flux trip on Channel N-36 is equivalent to 29% power while the mactor is in Mode 3 with the , shutdown banks fully withdrawn in preparation for a reactor startup? ' ' , Correct the level trip setpoint to the curent equivalent to 25 % rated thermal power prior i a.
to increasing thermal power above the P-6 setpoint.
i ! .: b.
Correct the level trip setpoint to the current equivalent to 25 % rated thermal power prior to increasing power above 5 % of rated thennal power.
! Place Channel N-36 in the tripped condition prior to entering Mode 2.
c.
d.
No action is required.
Answer: d Points: 1.0 , References: TS 2.2.1
. K/A: 015000.K3.01 RO 3.9/SRO 4.3 System Code: 28 [ 012000.A2.03 RO 3.4/SRO 3.7 System Code: 57 012000.G05 RO 3.4/SRO 4.3* System Code: 57 , 012000.A1.01 RO 2.9*/3.4* System Code: 57 ] i l l ! > k i i
I l > . I i + !
! i .! -, . - . ..
O Of w v i RB-0474 Which one of the following is correct concerning the basis for establishing redundant HHSI flow paths during a LOCA (two HHSI pumps are mnning)? < provides additional protection against passive failures during the meirc phase of the a.
LOCA.
b.
enhances the flow indication reliability by' placing the alternate path flow transmitter in the flow path.
provides additional makeup flow to the RCS because of less system head loss.
c.
d.
enhances reliability of HHSI pumps by reducing the backpressure on discharge line.
Answer: a Points: 1.0 References: E-1 Background Document K/A: 006000.K4.18 RO 3.3/SRO 3.8 System Code: 62
l i . . .. - - - - .
.
(^') O O - I RB-0037 Which one of the following is the response of source nmge counts for the given situation.
While a reactor stanup is in progress the RO stops control rod motion when the reactor is close to criticality, but still suberitical? Continue to increase but at a slower rate. The stanup rate should stabilize at a lower a.
positive value.
' b.
Continue to increase for a shon time and then plateau. The startup rate should gmdually decrease to zero.
Stop increasing and stabilize at its present value. The stanup rate should immediately c.
decrease to zero.
d.
Begin to slowly decrease. The stanup rate should gradually decrease to zero from a slightly positive value.
Answer: b Points: 1.0 References: 1-OP-1.5 K/A: 192008.Kl.03 RO 3.9/SRO 4.0 System Code: Rx-Th 192008.Kl.05 RO 3.8/SRO 3.9 System Code: Rx-Th j i j i l I
I
e_ e_ s RB-0330 What will be the control rod position at 100% power given the following: Initial Conditions: Final Conditions:
- Reator Power = 25%
- Reactor Power = 100%
- Equilibrium Xenon = -1570 pcm
- Equilibmim Xenon = -2875 pcm o Baron Concentration = 1300 ppm
- Boron Concentration = 1025 ppm
- Control Rod Position = 'D' bank at 60 steps
~ (Assume core burnup is 4000 MWD /MTU and differential boron worth is -6.88 pcm/ ppm.)
a.
D Bank @ 225 c.
D Bank @ 209 c.
D Bank @ 194 d.
D Bank @ 177 Answer: C - This is Unit 1 Cycle 10 Dependent Points: 1.0 . References: Station Curves K/A: 001000.K5.05 RO 3.5/SRO 3.9 System Code: 61 001000.K5.09 RO 3.5/SRO 3.7 System Code: 61 001000.K5.22 RO 2.1/SRO 2.5 System Code: 61 001000.G06 RO 2.9/SRO 3.8 System Code: 61 .
. .
a a w -
RB-0187 (SRO Only) Which ONE of the following best describes your course of action to be taken if during normal full-power operation of unit 1, radiation monitors for Steam Generator Blowdown and Condenser Air Ejector Discharge were observed to be increasing? Chemistry then samples the RCS and all S/Gs and identifies a primary to secondary leak on "C" S/G of.06 gpm. No other primary or secondary parameters exhibit any changes.
Reduce leakage within 4 hours or be in at least HOT STANDBY within the next 6 hours, a.
and in COLD SHUTDOWN within the following 30 hours.
b.
Trip the reactor and go to 1-E-0.
Reduce power to <50% with in 90 minutes, no other power re.fuction is required.
c.
d.
Reduce power to <50% within 90 minutes and below Mode I within 2 hours of detecting excessive leakage.
Answer: d Points: 1.0 References: 1-AP-24; TS 3.4.6.3 K/A: 000037.EA2.13 RO 4.1/SRO 4.3 System Code: 42-OE 000037.EA1.13 RO 3.9/SRO 4.0 System Code: 42-OE , ' 000037.G11 RO 3.9/SRO 4.1 System Code: 42-OE 000037.EA2.01 RO 3.0/SRO 3.4 System Code: 42-OE i I
,. . , .. . -: O O , . RB-0364 (SRO Only)
Determine the operability of the diesel driven fire pump if an operator on rounds notes the diesel . driven fire pump fuel oil tank (FP-TK-4) indicates 17 inches in the sight glass? Pump is operable, fuel oil level is satisfactory.
a.
b.
Pump is operable, however take actions to restore tank level to > 90% within 8 hours.
j Pump is inoperable, establish and denionstrate the operability of a backup ' fire c.
, suppression system.
- l d.
Pump is inoperable, restore to operable status within 7 days or prepare and submit a - special report to the NRC.
Answer: d , , ' Points: 1.0 References: UFSAR 16.2.1.2; 1-SC-5.27; Log 6E K/A: 086000.G05 RO 3.0/SRO 3.6 System Code: 30 086000.G11 RO 2.7/SRO 3.5 System Code: 30 ' , f i ! , ! L i [ '
- f
! !
) - ! . _, m - -.. , _. _ ,.
O O w w
SB-0522 (SRO Only) Which one of the following is the correct classification for the following event? At 0915 with Unit I at 15% power the condenser air ejector radiation monitor (1-SV-RM-121) Hi-Hi alarm
comes in, reading 3.5x10 cpm, but fails to divert to containment. When the instmment fuses are pulled the valves still fail to swap to containment. Currently (1008) condenser air ejector flow rate is 6.5 scfm and reading 3.5x10 cpm. No other radiation monitors are in alarm and
letdown has been isolated due to pressurizer level lowering, a.
Notification of Unusual Event.
+ b.
Alert.
c.
Site Area Emergency.
d.
General Emergency.
Answer: b Points: 1.0 References: EPIP-1.01, (E.2); l-AP-5 K/A: 000060.EK3.01 RO 2.9/SRO 4.2 System Code: 33 000059.EK3.02 RO 3.2*/SRO 4.5 System Code: 37-OE l
- - - _. _. _ _ _.
.. _.. . _ _
- t<
' . & a s f'e n 33-3@ VIRGINIA POWER NORTH ANNA NUCLEAR TRAINING O#" N- " Rennt LICENSED OPERATOR REOUAL PROGRAM
NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATION . ' PART "A" AEO-32 ' SEPTEMHER 21.1993
' , NAME: I DATE: i (PRINT) ~ l INSTRUCTIONS . -5 1.
Number of Questions
Possible Points - 15
.i
% OR 12 Points is passing.
l 2.
Time limit:
hour (s).
PLEDGE i I have neither given nor received any aid on this test.
, , ' .! - Signature: .i ! ! f TEST RESULTS
Grade: / % l ! Graded by: Date: -;
Validated by: Date: (If Applicable) i .! TEST Sienatures (Required for master only) g - < Submitted by: / 'Date: 0 8-14 "i3 , / Writer ~ [] . b Date: I '33 Approved by: L - Supervisor i ' r L AEO-32-KEY !
> > v.- .* -- .v.. .
- . . L
O ,
I $ l I Vircinia Power
. North Anna Power Station Licensed Operator Requalification
Program ~ '
- ]
, t , Static Simulator Scenario Loss of RHR - Shut Down LOCA , ! AE032 I i i ! l I l < . I o I
. b b , STATIC SIMULATOR SCENARIO Loss of RHR - Shut Down LOCA , Summary: The Unit was shut down 150 hours ago, and is cooling down in Mode 4 with RCS temperature approximately 205'F. The "B" charging pump and the "B" RHR pump are running.
, The "C" charging pump is tagged out for electrical maintenance. The 1H Emergency Diesel Generator is tagged out for starting air system maintenance,- the tag for the EDG output breaker (15H2) has been lifted and the breaker is racked to Test for Control Operations breaker testing.
, A loss of the 1H Emergency Bus occurs as a result of the 1H normal feeder breaker 15H11 tripping open. It is reported that the breaker has been damaged and cannot be returned to service in a rapid manner.
The "B" RHR pump tripped and could not be re-started, resulting in no RHR pumps available.
, The H Emergency Bus is returned to service via the alternate feed from the "B" Station Service bus.
Power range channel N-41 fails high.
1-AP-11, Loss of RHR, is used to start the "A" RHR pump. A loss of Coolant Accident occurs due to thermal and hydraulic stresses upon restoration of ' flow. 1-AP-17, Shutdown LOCA, has been entered and is completed through step 12.
SIMULATOR SETUP: i Advance charts and ensure they are inking e , Ensure the correct PORV setpoint plackard is on the control board.
e Select IC-54 (AEO32_PRELOAD)
Place SYSTEM TAGGED STICKERS the following control switches: e 15H7 "C" charging pump (normal) 1537 "C" charging pump (alternate) 15H2 - 1H EDG output breaker Place a " BREAKER IN TEST" magnet on 15H2.
e
Run simulator 5 minutes to establish trends.
i , Run malfunction timer, and start the "A" RH.R pump before the LOCA occurs.
o
i
. g a w
Allow PRZR level to drop approx.15%, then perform the following steps of 1-
AP-17.
' 1: Increase make-up flow 2: Isolate RCS drain paths 9: Align the BIT 10: Actuate Phase A isolation Let simulator run for approx. 2 minutes, then freeze simulator and acknowledge e' the board. Do NOT allow any RCS temperature to drop below 200 degrees.
(Mode 5) ~ OR Select IC-88 (Mode 5,195 degrees)
Enter the following malfunctions on the malfunction processor: e MNIO801,5 sec. TD, -100% deg., (N35 undercompensated) MMS 0102,5 sec. TD, -100% deg., (CH IV A SG steam flow xmitter failure) MRH0502,10 sec. TD, ("B" RHR pump trip) MNIO101,10 sec. TD,100% deg., (N-41 fails high) MRC04,30 sec. TD,5 sec. ramp,100% deg. (RCS pressure boundary leak) On SIMLOCH: HBUS_UVRESET=T
e Place 1H EDG mode selector switch in MAN-LOCAL Place a " BREAKER IN TEST' magnet on 15H2, and PTL breaker.
- Place SYSTEM TAGGED STICKERS the following control switches:
15H7 "C" charging pump (normal) 1537 "C" charging pump (alternate) 15H2 - 1H EDG output breaker Run simulator and malfunction timer. When "B" RHR pump trips, then stop the
malfunction timer.
Manually open 15H11, then try to re-close (leave in AUTO-AFTER-START with
breaker disagreement) Restore the IH bus from the 1B bus IAW 1-MOP-6.70. (Close 15H1 and 15B11) i e Allow simulator to run approximately 10 minutes, until all wide range RCS e
temperatures are greater than 215 degrees.
' Run malfunction timer, and start the "A" RHR pump before the LOCA occurs.
e
. e
. Allow PRZR level to drop approx.15%, then perform the following steps of 1-
AP-17.
1: Increase make-up flow 2: Isolate RCS drain paths 9: Align the BIT 10: Actuate Phase A isolation Let simulator run for approx. 2 minutes, then freeze simulator and acknowledge e the board. Do NOT allow any RCS temperature to drop below 200 degrees.
(Mode 5) ~
1 e
. S & w -
TIME EVENT MALFUNCTIONS OVERRIDES ACTIONS TAKEN T=0:00 Run simulator T= 5:00 Run malfunction timer Start "A" RHR pump T=5:10 LOCA ~ IAW-1-AP-17: Open CH-FCV-1122 Isolate RCS drains Align BIT Phase A isol.
T= 7:(10 Freeze the simulator and acknowledge the board VERIFICATIONS: All RCS temperatures indicate >200 degrees.
e e Level in at least one SG is <75%. e 1-RC-MOV-1535 is closed.
1-MS-TV-109A/B indicates open.
e
. &
- -
- l TURNOVER:
The Unit was shut down for refueling 150 hours ago, and is cooling down in
Mode 4 with RCS temperature approximately 205"F.
The IH EDG is tagged out for staning air system maintenance. the Tag for
the EDG output breaker,15H2, has been lifted and the breaker racked to l TEST for Control Operations Depanment testing.
A loss of the IH emergency bus odcurred when 15H11 uipped open. The
breaker is dynaged and cannot be returned to sen' ice.
The "B" RHR pump tripped and could not be re-staned, resulting in no RHR
pumps available.
' The 1H bus was reenergized from the IB station service bus.
- The "A" RHR pump was staned in accordance with 1-AP-ll, I.oss of RHR.
- A loss of coolant accident occurred due to thermal and hydraulic stresses upon restoration of flow.
1-AP-17, Shutdown LOCA, has been completed through step 12. (See attached
procedure.)
EOUIPMENT OUT OF SERVICE:
IH Emergency Diesel Generator (air system maint.)
' e "A" charging pump in PTL,15H6 in CONNECT
"C" charging pump tagged out (electrical maint.)
"A" SG channel IV steam flow in TRIP
. /~% ti e - . Approval and Revisions Record Licensed Operator Reaualification Procram North Anna Power Station Scenario Number AEO32 Revision number 0 Scenario Title Imss of RHR - ShtrrDown LOCA Written by: Date submitted: ' , , v !..!d \\, Date approved: 7 - / "d ' Approved by: </ t -. () - Scenario Number Revision number Scenario Title Reason for change Submitted by: Date submitted: Approved by: Date approved: Scenario Number Revision number Scenario Title Reason for change Submitted by: Date submitted: Approved by: Date approved: Scenario AEO32 Page 7 Revision 0
. _ - ... - ._ ' . RA-568
When RWST 1evel drops to , then the LHSI pump suction valves to , the Containment sump. (Given that the accident in progress is allowed to continue.)
J 20%, will automatically swap a.
b.
20%, should be manually aligned
, , c.
23 %, will automatically swap ! I d.
23%, should be manually aligned
^ answer: d points: 1.0 Reference: 1-AP-17/ rev. 2 K/A: 006020.A1.09 RO 3.3/SRO 3.9 Sys. 62 006020.A4.02 RO 3.9/SRO 3.8 Sys. 62 . f .; , ! i
.. t
I ! i , i V , - , >
- o) e - RA-569 THIS QUESTION DOES NOT PERTAIN TO THE SCENARIO.
Which one of the following statements is correct conceming the operation of the Stub bus breaker (15H12) during an under-vcitage condition? The breaker will open on UV ONLY if the associated RHR pump is running.
a.
b.
The breaker has N_Q auto-close feature.
. The breaker will open 15 seconds after a UV condition occurs,.QELY if the RHR c.
pump bmaker is still closed.
d.
The breaker will automatically close 15 seconds after a UV clears ONLY if the associated RHR pump breaker is open.
Points: 1.0 answer: d Reference: ESK 5AQ K/A: 062000.A2.04 RO 3.1/SRO 3.4* Sys. 06 062000.A2.05 RO 2.9/SRO 3.3* Sys. 06 005000.K2.01 RO 3.0/SRO 3.2 Sys. 60 I l l l i
. e a w - . RA-570 Which one of the following alternate core cooling methods is preferred? (Given the current conditions.)
a.
Natural circulation b.
Reflux boiling c.
Forced feed and spill d.
Gravity feed and spill answer: a points: 1.0 Reference: 1-GOP-13.0/Rev.1 K/A: 000025.EK1.01 RO 3.9/ SRO 4.3 Sys. 60 l l
. A S - - t RA-574 Which one of the following statements would be correct concerning the operation of the charging pumps, if an operator were to rack breaker 15J7 for the "C" charging pump to CONNECT? "B" charging pump continues to run, "A" and "C" pumps could be manually started.
a.
b.
"B" charging pump trips, only "A" pump could be manually staned.
"B" charging pump continues to run, "A" pump could be manually staned.
c.
d.
"B" charging pump trips, NO charging pumps could be manually staned.
answer: b points: 1.0 reference: ESK 5AL, SMN, 5AN; LER 90-11 K/A: 006000.K4.05 RO 4.3/SRO 4.4 Sys. 62
. ,~\\ s) _ -
, l RA-575 Which one of the following statements is true concerning the actions required to re-stan the spent fuel pit cooling pump,1-FC-P-1A? (Given that 1-FC-P-1A was running at the start of this event.)
None, the pump should have automatically staned 30 seconds after 1H bus voltage a.
was restored.
I b.
The pump will stan if the control switch is placed in START.
The control switch must be placed in stop to reset the UV signal, then placed in c.
START.
d.
The UV signal must be reset by locally cycling the breaker, then place the control switch in START.
answer: b points: 1.0 references: ESK 6GH K/A 033000.K3.03 RO 3.0/SRO 3.3 Sys. 67 033000.A2.02 RO 2.7/SRO 3.0 Sys. 67
,* -<3 g !_) - t RA-577 What was the cause for the steam generator blowdown radiation monitor alarms? (Note: These alarms were noticed while the operator was checking the status of the Radiation Monitoring System in accordance with 0-AP-10.)
a.
Containment phase A isolation b.
Steam generator tube leakage c.
Loss of electrical power d.
Improper calibration ANSWER: C reference: Loop diagram /RM-16 K/A 073000.G08 RO 3.3/SRO 3.3 Sys. 54
O b , PA-578 Annunciator C-E4, LOW TC INITIATE NDT PROTECT, is in alarm. What is the cause of this alarm? a.
Equipment failure; alarm should NOT be locked in.
, b.
RCS pressure is high and the PORVs have failed to open.
c.
The PRZR PORV control switch should b'e in OPEN.
d.
PRZR PORV block valve 1-RC-MOV-1535 is closed.
ANSWER: d , reference: AR C-E4 K/A: 000008.EA2.05 RO 3.9/SRO 3.9 Sys. 56 0100000.G08 RO 3.5/SRO 3.5 Sys. 49 ' 0100000.G12 RO 3.6/SRO 3.6 Sys. 49 . I
. A G - - . RA-581 Which one of the following statements would be correct, if this scenario were allowed to continue with NO operator actions? The RHR pump suction relief valve will open.
a.
b.
The RHR pump suction valves will close.
ONLY one PRZR PORV will open.
c.
d.
Both PRZR PORVs will open.
answer: d reference: AR C-F4 K/A: 005000.A2.02 RO 3.5/SRO 3.7 Sys. 60 010000.Kl.02 RO 3.9/SRO 4.1 Sys. 49
_ i . h . RA-582
Which one of the following statements is correct concerning CC flow to the RCPs? (Annunciator C-C4, RCP 1ABC THERMAL BARRIER CC HI/LO FLOW, has alarmed.)
a.
CC Dow still exists, ONLY indication of flow has been lost.
b.
CC flow was lost when the H bus deenergized.
! i ! c.
CC flow was lost when containment phasd A isolation was actuated.
d.
CC flow was isolated due to a thermcJ barrier heat exchanger leak.
answer: b i ! ' reference: ESK 6MC; load list K/A: 003000.A4.08 RO 3.2/SR.O 2.9 Sys. 56 003000.G12 RO 3.4/SRO 3.5 Sys. 56
000026.G09 RO 3.3/SRO 3.4 Sys.12 ' . P
I . , ?
.
' ' <~ , u . - RA-583 What was the cause of annunciator J-H4, RES SS XFMR OVERLOAD /CKT TROUBLE OR LS IN DEFEAT alarming? The "H" emergency diesel generator is tagged out.
a.
b.
One of the RSS transformers is overloaded.
c.
The "H" bus alternate feeder breakers are: closed.
d.
Load shed is in DEFEAT.
answer: d reference: AR J-H4; 0-OP-26.7/rev. 2 K/A: 062000.G12 RO 3.1/SRO 3.1 Sys. 06 062000.G08 RO 3.1/SRO 3.1 Sys. 06
. -- .. ~ (3 & v
I v RA-0588 - THIS QUESTION DOES NOT PERTAIN TO THE SCENARIO.
, ' Complete the following statement conceming reactor coolant loop flow indication. (Given that all three reactor coolant pumps (RCPs) are initially mmiing.)
If "C" RCP trips then indicated flow in the "C" loop will , while indicated flow in the "A" and "B" loops will - _ . l a.
decrease to zero then increase slightly; increase j b.
decrease to zem then increase slightly; decrease c.
decrease to zero and stay at zero; increase ,
d.
decrease to zero and stay at zem; decrease Answer; a Points: 1.0 ) References: HT-FF K/A: 002000.A1.05 RO 3.4/SRO 3.7 Sys. 56
. . a l I
- .
. . v . RA-573 (SRO ONLY) Which one of the following statements correctly describes the response of the Electrical distribution system if the Control Room Operator were to mistakenly open the normal feeder breaker to the IB station service bus (15B2)? 15BI would automatically close; the IB and lH buses would remain energized.
a.
b.
15B11 and 15H1 would open; 15BI would automatically close; the IB bus would be re-energized; the IH bus would be re-energized by the EDG.
c.
15B11 and 15H1 would open; the IB and 1H buses would be dead.
d.
15Bil and 15H1 would open: 15B1 would automatically close; the IB bus would be re-energized;'the IH bus would remain dead.
answer: c points: 1.0 reference: FE-21G K/A: 062000.K4.03 RO 2.8*/SRO 3.1 Sys. 06 062000.A2.04 RO 3.1/SRO 3.4 Sys. 06 062000.Kl.04 RO 3.7/SRO 4.2 Sys. 06 i l l
.' C (a'\\ u). .- SA-576 (SRO ONLY) Which one of the following statements correctly describes the actions that must be taken in ' accordance with T.S. 3.8.1.17 (Given that 1-PT-80 is already in progress.)
Perform the operability PT on the IJ EDG within 8 hours; restore ONE of the a.
inoperable sources to operable within 12 hours; restore the other inoperable source within 72 hours.
b.
Perform the operability PT on the IJ EDG within 24 hours; restore the IH EDG to operable within 72 hours.
~ Restore ONE of the inoperable sources to operable within 12 hours; restore the other c.
inoperable source within 72 hours; the operability PT on the IJ EDG is HDT requimd to be performed.
d.
Restore the IH EDG to operable within 72 hours; the operability PT on the IJ EDG ! is NOT required to be performed.
l answer: d points: 1.0 . reference: ' 1-PT-80/Rev.11; TS 3.8.1.1; DR 91-1831; DR 91-1835; LER 91-021 { . ! K/A 064000.G11 RO 3.4/SRO 3.9 Sys. 27 ' 062000.G11 RO 3.1/SRO 3.7 Sys. 06 . .
1 f b l
c.
g.
(,) ,. . _ . RA-579 (SRO ONLY) Which one of the following statements is correct concerning the operation of source range NIs, N-31 and N-32? (Given that the operating crew has not attempted to manually reenergize the source range nuclear instmments.)
The NIs can't be manually energized.
a.
b.
ONLY N-31 can be manually energized. ' BOTH NIs can be manually energized.
c.
d.
ONLY N-32 can be manually energized.
ANSWER: c reference: NA-DW-5655D33/sh. 3 K/A: 000032.EA1.01 RO 3.l*/SRO 3.4* Sys. 28
_ i < .. & S - - .. t . .SA-584 (SRO ONLY) i i
THIS. QUESTION DOES NOT PERTAIN TO THE SCENARIO.
What action would be required in accordance with technical specifications if all Containment air recirc fans were to trip while the unit was in, Mode 1 through 4? ) a.
No action required.
i b.
Immediately secure Containment purge and exhaust, Maintain Containment temperature less than 12.0 degrees using alternate cooling c.
methods.
, i d.
Perform an RCS leak rate calculation every 24 hours.
answer: d - reference: TS 3.4.6.1; LER 93-004 ! K/A: 073000.Kl.01 RO 3.6/SRO 3.9 Sys. 54 073000.G05 RO 3.1/SRO 3.6 Sys. 54 073000.G06 RO 2.5/SRO 3.4 Sys. 54 '
) ,
i l a
.. . .. - - -. . - .. 'l O O } ~ ' , lbfo.r /en 93 Jo* i VIRGINIA POWER i NORTH ANNA NUCLEAR TRAINING M* " /4- !
- "
(4 L"Qtt A L i LICENSED OPERATOR REOUAL PROGRAM l NRC ADMINISTERED SENIOR REACTOR OPERATOR EXAMINATIQE.
l OPEN REFERENCE PART "B" " SEFFEMBER 21.1993 i NAME: DATE: (PRINT) INSTRUCTIONS
l . f 1.
Number of Questions
Possible Points
!
% OR _16 Points is passing.
,
2.
Time limit:
hour (s).
> I i PLEDGE I have neither given nor received any aid on this test.
t Signature: -) -i TEST RESULTS Grade: / % Graded by: Date: Validated by: Date: (If Applicable) TEST Sinnatures (Required for master only)
f , Submitted by: '/ ///6,/ j/ Date: /7} / 9 - 93 . /~ Writer
Approved by: Date: 6fN/95 C Supervisor , PART B-KEY l
- !
, _.
. - a a w - '. RB-0188 Which one of the following is the correct action the operating team should take given the following? A SGTR has occurred on Unit 1, and E-3 is being utilized by the operating team.
"A" SG was identified as the ruptured SG and was subsequently isolated. The RCS has been cooled down to 490*F and depressurized to stop primary to secondary leakage and recover PZR level.
While attempting to establish 25 gpm charging flow, the BOP operator reports that "B" SG LEVEL is increasing in an uncontrolled manner with all feed /AFW isolated; pressurizer level is 30% and falling.
a.
Transition to ECA-3.1, step 1.
- b.
Transition to E-3, step 1.
Continue with E-3, subsequent actions will provide a sufficient mitigation response.
c.
d.
Re-initiate R.CS depressurization to backfill 'B' steam generator to the RCS.
Answer: b Points: 1.0 References: 1-E-3; WOG Background K/A: 000038.EA2.02 RO 4.5/SRO 4.8 System Code: 42-OD 000038.EA2.01 RO 4.1/SRO 4.7 System Code: 42-OD 000038.EK3.06 RO 4.2/SRO 4.5 System Code: 42-OD .
.
f) b v
RB-0370 ' Which one of the following describes the response of Natural Circulation to the given information? The plant experiences a loss of all AC power. Effons to restore AC power prove unsuccessful.
RCS inventory depletion from RCP seal leakage continues, eventually draining the upper head > of the reactor vessel and causing steam voids to form in the S/G U-tubes.
, t Natural Circulation stops and all means of decay heat removal will be lost. Extensive a.
core damage will occur soon after the intermption of natural circulation.
b.
Natural Circulation stops and reflux boiling will remove decay heat until enough P inventory is lost to prevent decay heat removal. Then inadequate core cooling may occur.
Natural Circulation stops but reflux boiling will provide adequate decay heat removal for c.
, as long as necessary.
d.
Natural Circulation decreases but continues to provide adequate decay heat removal for.
as long as necessary.
> Answer: b Points: 1.0 References: 1-ECA-0.0; WOG Traming Pan B Item #000-055-11 K/A: 000055.EA2.02 RO 4.4/SRO 4.6 System Code: 06-OE
000055.EK1.02 RO 4.1/SRO 4.4 System Code: 06-OE . $
! ? i
I i
. O O '
RB-0190 i The preferred method to accomplish a post-SGTR cooldown is: a.
ES-3.1, " Post-SGTR Cooldown using Backfill" because it limits the release of radioactive contaminates and therefore limits the offsite dose.
b.
ES-3.1, " Post-SGTR Cooldown using Backfill" because it allows rapid cooldown and depressurization of the RCS.
ES-3.3, " Post-SGTR Cooldown using Steam Dumps" because it limits the release of c.
radioactive contaminates and therefore limits the offsite dose.
d.
ES-3.3, " Post-SGTR Cooldown using Steam Dumps" because it allows a rapid cooldown and depressurization of the RCS.
Answer; a Points: 1.0 References: 1-E-3; ES-3.1 Background Documents; l-ES-3.1 ' K/A: 000038.EK3.06 RO 4.2/SRO 4.5 System Code: 42-0D -
I i ,
. A e - -
RB-0515 Should an RCP be staned in the following situation? Unit I has experienced a small break LOCA resulting in safety injection and a manual trip of all RCPs due to inadequate subcooling. All emergency procedures are performed conectly and all equipment operated as expected. The crew is presently at Step 12 of ES-1.2 Post LOCA Cooldown and Depressurization. All conditions are met to start an RCP but the RCS is solid with subcooling at 80*F.
, No, the resultant pressure surge will cause the LOCA to worsen, resulting in even lower a.
subcooling.
b( Yes, the leak will act as the steam bubble and limit the pressure insurge.
No, the increase in pressure will cause the leak rate to be greater than the ECCS makeup c.
rate.
d.
Yes, forced flow heat transfer must be established regardless of consequences, to ensure core damage does not occur.
Answer: b Points: 1.0 References: 1-ES-1.2, ERG Feedback DW-90-051, Response Letter ERG 91-012 K/A: 000009.EK3.21 RO 4.2/SRO 4.5 System Code: 62-OD . I
' S e - - .
RB-0069 Which one of the following statements concerning the power defect is correct? The power defect is the difference between the measured power coefficient and the a.
predicted power coefficient.
b.
The power defect increases the rod worth requirements necessary to maintain the desired shutdown margm following a reactor trip., Because of the higher boron concentration at the beginning of core life, the power defect c.
is more negative.
d.
The power defect necessitates the use of a ramped Tavg program to maintain an adequate RCS subcooling margin.
Answer: b i Points: 1.0 References: Unit 1 Plant Cun'es K/A: 001000.K5.49 RO 3.4/SRO 3.7 System Code: 61 \\ . t
.. , - ' . .
RB-0031 At what rate may the operator realign the control rod in the following situation? ' Unknown to the control room crew, the reactor had been operating at 100% power with a 'D' Bank control rud misaligned 9 steps above demand position. When the operator moves the
controlling bank for reactor control, he receives a " Computer Alarm Power Range Tilt Rod Deviation Sequence" annunciator alarm. The RPIs indicate a 13-step deviation between demand position and the affected rod's position.
a.
No restriction apply ~ ! b.
2 steps / hour
c.
4 steps / hour , d.
12 steps / hour Answer: a ' Points: 1.0 References: 1-AP-1.3; OP-58.5; T.S. 3.1.3 [ K/A: 000005.EK3.03 RO 3.6/SRO 4.1 System Code: 61-OE
000001.EK3.04 RO 3.4/SRO 4.1 System Code: 61-OE , d
1 -
. ' O O w w
P3-0176 Which of the following will be the new RCS boron concentration after RHR is placed in service? The Reactor Coolant System (RCS) is being maintained @ 325 *F and 350 psig by steam dumps.
The Residual Heat Removal (RHR) System is being placed in service for normal cooldown.
Chemistry has reponed that the RCS boron concentration is 2010 ppm and the RHR baron concentration is 1990 ppm.
t a.
2019 ppm b.
2009 ppm c.
1999 ppm d.
1991 ppm Answer: b Points: 1.0 Refemnces: 1-OP-14.1 K/A: 005000.KS.00 RO 3.2/SRO 3.4 System Code: 60
e
RB-0443 Which one of the following is correct? During a normal shutdown, RCS pressure and temperature are reduced to below 1950 psig and 543'F respectively, and the Low Pressurizer Pressure and High Steam Flow SI signals are blocked.
If pressure rises above P-11 setpoint, the Iow PZR Pressure SI is automatically a.
unblocked.
b.
If pressure rises above P-11 setpoint, the operator unblocks the Low PZR Pressure and High Steam Flow SI signals.
If Tavg increases above P-12 setpoint, the Iow PZR Pressure and High Steam Flow SI c.
signals are automatically unblocked.
d.
If Tavg increases above P-12 setpoint, the Ixw PZR Pressure SI is automatically unblocked.
Answer: a Points: 1.0 References: NA-DW-108D014 P. 3, 7; TS 2.3.3 K/A: 010000.K1.02 RO 3.9/SRO 4.1 System Code: 49 013000. A1.01 RO 4.0/SRO 4.2 System Code: 62 - . .
a a w -
RB-0298 What will be the long term effects of maintnining the unit at the following conditions? The RCS is on a VCT float with 20% level in the PZR. An N2 blanket is being maintained on the PZR from the PRT with the PZR PORVs open. RCS wide range pressure is 22 psig and PRT pressure is 8 psig. Charging flow is 25 gpm. Seal injection is 2 gpm per RCP, and letdown is 31 gpm. VCT pressure is 40 psig.
a.
No long tenn adverse effects. All conditions remain stable.
b.
PZR level decreases due to difference in charging and letdown flow.
~ Gases come out of solution in Rx head due to high VCT gas over pmssure, forming head c.
bubble.
d.
Vonexing in RHR pump impeller due to higher than normal equilibrium N2 content in RCS fluid.
Answer: c Points: 1.0 References: 1-OP-3.4 K/A: 002020.K5.06 RO 3.4/SRO 3.8 System Code: 56 . .
' ^ . -* . t - '
i s RB-0332 l , Which of the following is the appropriate action for this situation?^ You assume the watch with the unit in Mode 3 at 547'F waiting for RCS activity results from chemistry. The RCS activity prior to the first RCP stan was 7.5 x 102 micro-curies /gm.
Chemistry infonns you that the pmsent RCS activity is 9.5 x 10-2 micro-curies /gm.
.i
Present RCS activity is greater than the initial activity, continue RCS pu'rification until a.
initial activity is restored.
, '
i b.
Present RCS activity is less than or equal to the maximum allowed RCS activity-i withdraw SD banks and commence RCS dilution.
. Present RCS activity is greater than the allowed maximum RCS activity, feed and bleed c.
of the RCS is required.
d.
Present RCS activity is greater than allowed by technical specifications. Continue RCS purification until less than the allowed activity limit.
! Answer: b , Points: 1.0 , i References: 1-OP-1.5; TS 3/4.8 K/A: 001000.G10 RO 3.3/SRO 3.5 System Code: 61 000076.G07 RO 2.9/SRO 3.4 System Code: OC-OE -
. $ , ! l / \\
i
e t . RB-0451 Which one of the following is the reason for tripping RCPs in E-0 if containment pressure has exceeded 28 psia? RCPs must be secured to prevent electrical gmunds due to condensation on the motor a.
windings.
b.
CC-TV-105A, B, and C closed on Phase B. isolation resulting in a loss of cooling for the RCP stator windings.
CC-TV-101 closed on Phase B isolation resulting in loss of cooling to the RCP thermal c.
barriers.
d.
CC-TV-104A, B, and C closed on Phase B isolation resulting in a loss of cooling for the RCP motors.
Answer: d Points: 1.0 References: 1-E-0; FM-79 Series K/A: 003000.K6.04 RO 2.8/SRO 3.1 System Code: 56 003000.G14 RO 4.0*/SRO 3.9 System Code: 56 .
- _.
, v , . RB-0221 Given the following data, determine the estimated critical position by perfonning a 1/M plot on the attached sheet: N-31 Reading Rod Height Bank 2.0x10' 228 .A 2.2x105 172 .B 2.4x103 220
- B 3.1x103 144 C
5.7x10'
D a.
130 steps on "D" bank b.
138 steps on "D" bank c.
160 steps on "D" bank d.
168 steps on "D" bank Answer: b Points: 1.0 References: OP-1.5 K/A: 001010.K5.16 RO 2.9/SRO 3.5 System Code: 61 001010.K5.05 RO 3.-3*/SRO 3.4 System Code: 61 015000.K5.05 RO 4.1/SRO 4.4 System Code: 28 015000.K5.06 RO 3.4/SRO 3.7 System Code: 28 001010.A2.07 RO 3.6/SRO 4.2 System Code: 61
-- _ - - - _ - _ _ _ _. - - - - _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ - _ _ _ _ _. _ _ )$.$$ i .ya OO O. Q rn ~c , ~ -m , yO ' < i A Critx:ahty is projected to occur at the point that an extrapolated line drawn between two 1/M points crosses the 0.0 axis.
1.0 ._._ .._ ._.. _.__.. .___.
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- :::-: ::::: Z::._____ ZZ
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_ .___ __ __ __ . __.___ _____ ZZ 1:r ::n Z-:: Z:~ ZZ z:_ z . ::z -- : c:- ~ ~ _____ ____._ Z~: :Z: ZZ 'z: ZZ Z::
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"B" Bank 120 140 160 180 200 220 228 b<; l iiii iiii.iiiiliiii;iiiiiiiii.
i s i li ii i is i ii i s ii i e "C" Bank 0
40
80 100 120 140 160 180 200 220 228 f 4 Steps / Increment l 1IiiIIIiiIIiiI IIII.IIIi iii1 IIIi'iiIi iiii i I i I ; I r1TT M O "D" Bank 0
40
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a $ - w
RB-0175 In accordance with 1-GOP-13.1, determine time required to mach 212 F in the pnmary given the situation.
Sixty hours ago Unit 1 experienced a SGTR. The RCS is now at 140"F, drained to 10" above centerline. The "A" RHR pump is running and "B" RHR pump is in pull-to-lock because of a damaged mechanical seal. The "A" RHR pump trips on overcurrent mndering actions to cool the RCS unsucessful, a.
5.5 minutes b.
7 minutes c.
8.5 minutes d.
10 minutes Answer: a Points: 1.0 References: 1-GOP-13.1 K/A: 000025.G02 RO 2.7/SRO 3.7 System Code: 60-OE 000025.G11 RO 3.6/SRO 3.9 System Code: 60-OE 005000. A1.06 RO 2.7/SRO 3.l* System Code: 60 - l ,
" - _ _ . RB-0002 What is the possible consequence of not establishing Redundant Charging /SI Flows in E-1 " Loss of Reactor or Secondary Coolant," step 25 " Establish Charging /SI pump Redundant Flow Paths," following a LOCA on Unit I? A single passive failure could result in loss of chargmg/SI flow to the core.
a.
b.
Possible core overheating due to boron plateout on the fuel rods at the core exit.
Limits HHSI additional makeup flow to the RCS because of more system head loss.
c.
d.
Possible HHSI pump damage caused by insufficient cooling due to operating pumps in parallel.
Answer; a Points: 1.0 References: 1-E-1; WOG Background E-1/FM 96A Sheet 3 (SI); ESK-6EP, ESK-6EQ, ESQ-6ET K/A: 006000.K6.01 RO 3.4/SRO 3.9 System Code: 62 006000.K6.01 RO 3.4/SRO 3.9 System Code: 62 . i l /
o g
RB-0344 Without the CRDM fans in operation, which one of the following is the greatest concern given the following situation? A natural circulation cooldown is in progress in accordance with ES-0.2A, ' Natural Circulation Cooldown with CRDM Fans.' The RCS is at 510*F and 1950 psig. All CRDM fans have just tripped and cannot be restarted , Damage may occur to the IRPI coils because of overheating.
a.
b.
Nil Ductility Temperature requirements are more likely to be exceeded for the reactor vessel head flange welds.
Damage may occur to the Excore Nuclear Instrumentation because of overheating.
c.
d.
Formation of a reactor vessel head steam bubble is more likely.
Answer d Points: 1.9 Rcferences: 1-ES-0.2 Basis K/A: 015000. A2.02 RO 3.1/SRO 3.5" System Code: 28 001000.A2.01 RO 3.1/SRO 3.7 System Code: 61 002000.K5.15 RO 4.2/SRO 4.6 System Code: 56 . . ' l
a
- - . RB-0150 Calculate the present core exit temperature, assuming RCS pressure is 1932 psig and the ICCM indicates 42*F subcooling.
a.
586*F b.
590*F c.
670*F d.
674*F Answer: b Points: 1.0 References: Steam Tables K/A: 001000.K5.56 RO 4.2/SRO 4.6 System Code: 61 000074.EK1.04 RO 3.7/SRO 4.1 System Code: 21 - .
/-. . \\_/ , . RB-0084 Given the following information, the crew should: A Main Steam Line bmak has occurred on Unit One. The operating crew has completed E-0 thm Step 15 when the STA informs them that an "Omnge Path; had existed on containment but pressure has now decreased below the Orange Path entry limit.
a.
Transition to FR-Z.1 immediately.
b.
Transition to FR-Z.1 at st p 21 of E-0.
c.
Remain in E-0.
d.
Transition to FR-Z.1 at step 27 of E-0.
Answer: C Points: 1.0 References: OPAP-0002 K/A: 194001.A1.02 RO 4.1*/SRO 3.9 System Code: 99-OD .
l g , - t . SB-0044 (SRO Only) Which one of the following situations allows you four (4) hours to mport it to the NRC.
a.
Instmment Maintenance personnel inadvenently cause the SPDS to be uut of service for 24 hours.
b.
A deviation from Technical Specifications.per 10CFR50.54(x) has occurred.
2000 gallons of hypochlorite are spilled into the discharge canal. The Environmental c.
Protection Agency has been notified of the spill.
d.
A unit shutdown is performed per the requirements of Technical Specifications LCO 3.0.3.
Answer: c Points: 1.0 References: VPAP-2802; 10 CFR 50.72 2.vi K/A: 000062.G02 RO 2.5/SRO 3.3 _ h , ( .
. .. .. _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ , - w . RB-0374 (SRO Only) Which one of the following statements is correct regarding the follow-up actions required by UFSAR? (Given that a diesel driven fire pump has been out of service for required testing pursuant to UFSAR 16.2.1.2.1.B for 7 days and backup fire suppression has been established.)
a.
Submit a special report by phone within 24 hours in accordance with T.S. 6.9.2.
b.
Submit a special report in writing within 1.4 days following the event outlining actions taken, causes for, and plans for restoratiori.
Submit a special report within the next 30 days outlining plans and redundancy c.
, provisions.
l ! d.
Place the units in HOT SHUTDOWN per Tech Spec 3.0.3.
Answer: c l Points: 1.0 References: UFSAR 16.2.1.2.1 I K/A: 086000.G05 RO 3.0/SRO 3.6 System Code: 30 { OS6000.G03 RO 2.7*/SRO 3.8 System Code: 30 l l l l l . . _ _ _. _ _ _ _ _ _ _ _ _ _ -. _ _ _ _ _ _ _ _ _ _ _ - - _ _. - - - - - - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - --- -- '
m_ &_
SB-0377 (SRO Only) Classify the event, given the following sequence of events: Unit 1 is just commencing a load reduction from 100% pcwer, as a result of a steam generator tube leak, confirmed to be > 50 gpm, when a steamline header rupture occurs resulting in a Rx trip and SI initiation. It is quickly recogmzed that Main Steam Line Isolation has not occurred and the MSTVs are closed manually. The condenser air ejector radiation monitor indicated 4x105 cpm prior to the SI with an air flow of approximately 5 CFM.
a.
Notification of Unusual Event.
b.
Alert.
c.
Site Area Emergency.
d.
General Emergency.
- Answer: b (per Tab G #2)
Points: 1.0 References: EPIP-1.01 K/A: 000040.G02 RO 3.0/SRO 4.0* System Code: 42-OD - 000040.G12 RO 3.8*/SRO 4.l* System Code: 42-OD , . }}