IR 05000354/1997008
| ML20198L953 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/15/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198L938 | List: |
| References | |
| 50-354-97-08OL, 50-354-97-8OL, NUDOCS 9710270245 | |
| Download: ML20198L953 (49) | |
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U.- S. NUCLEAR REGULATORY COMMISSION,
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REGION 1
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i Docket No.:!
- 50 354:
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Report No.:-
97-08 License No.:
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- Licensee: 2 Public Service Eiectric & Gas Company.
Post Office Box 236 -
Hancocks Bridge, New Jersey 08038 Facility:'-
Hope Creek Nuclear Generating Station L Location:
Hancocks Bridge, New Jersey Dates: -
September 29-October 2,1997 Chief Examiner:
J. Williams, Senior Operation's Engineer / Examiner, Region i I
Examiners: ^
L. Briggs, Senior Operations Engineer / Examiner, Region i B. Maier, Operations Engineer / Examiner, ~ Region 1 Approved By:
Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety
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EXECUTIVE SUMMARY Hope Creek Nuclear Generating Station inspection Report No. 50-354/97-08 Or>erations Five flope Creek senior reactor operator (SRO) instant candidates were administered initial licensing exams.- Four candidates passed all portions of the license exam. One candidate failed the written exam.
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Overallicandidate performance during the operating tests was determined to be good.- The examiners did not determine any generic performance weaknesses.
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Reoort Detal's 1. Ooerations
Operator Training and Qualifications!
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05.1 Senior Reactor Operator Initial Examinations a.
Scope The exam was prepared by Hope Creek in accordance with the guidelines in interim Revision 8, of NUREG-1021, " Examiner Standards." The examiners administered initial operating licensing tests to five senior reactor operator (SRO) instant candidates. The written exam was administered by the facility's training organization.-
i-b.
Observations and Findinas The results of the SRO exam is summarized below:
SRO Pass / Fail Written 4/1 Operating 5/0 I
Overall 4/1 l
The written exam, job performance measures (JPMs) and simulator scenarios were developed by Hope Creek in accordance with Interim Revision 8 of NUREG-1021
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" Examiner Standards." The facility's exam development team was comprised of contractors,and training and operation department representatives. Allindividuals signed onto a security agreement once the development of tne exam commenced.
The NRC subsequently reviewed and validated all portions of the proposed exam.
Various changes and/or additions to the proposed exam were requested by the NRC following their review. Hope Creek personnel subsequently incorporated the NRC's
. comments and finalized the exam.
The written exam was administered on September 29,1997. The written exam consisted of 100 multiple choice questions. There was one post exam comment by the facility concerning a question on the SRO written _ exam. The NRC reviewed the facility's comment and justification for the change and accepted the question with
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the revised answer.-
The operating tests were conducted from September 30- October 2,1997. The operating tests censisted of three' simulator scenarios, ten JPMs and an administration portion for each candidate. All JPMs were followed up with two system-related questions. The administrative portion consisted of a mix of _
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questions and JPMs.-
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Based on the grading of the written exam, the following question subject areas were missed by more than half of the applicants. This could indicate a weakness in the general understanding of the subject area.
- Requirements for SROs to be present in the control room. (02)
Knowledge of the basis for SPDS parameters. (Q7)
Requirements for working copies of procedures. (08)
Requirements for equipment tagging. (011)
Knowledge oi CRD-HCU valve positions. (019)
Knowledge of RWM rod blocks. (Q21)
Knowledge of impact of failure of RPV pressure instrumentation. (038)
Knowledge of RCIC operation in pressure control mode. (Q39)
Knowledge of behavior of levelinstrumentation. (O49)
Diesel generator restart after loss of electrical bus (O63)
RPV emergency depressurization at high temperature. (O89)
Knowledge of the maximum pressure for CS injection. (Q93)
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Basis for runback before tripping the recirculation pumps during an ATWS, (Q99)
The facility indicated that it has reviewed each of the above subject areas to correct
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possible weaknesces and willimplement programmatic changes as necessary.
Simulator performance by the candidates was good. The examiners noted that
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crew briefings were routinely performed by the PROS and were effective.
In the administrative segment of the operating portion of the exam, candidato performance was generally good, however, two candidates had difficulty with emergency classification and' recommending appropriate protective actions.
There were four separate principal Hope Creek contacts during the exam development process. This caused some problems in communications between NRC and the facility. Some NRC comments were lost initially, with the exchange of Hope Creek contacts. All of the facility contacts worked well with the NRC.
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Conclusiong
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Four of the five SRO instant candidates passed all portions of the exam. One candidate failed the written exam While the Hope Creek liaison with NRC was frequently changed and resulted in some problems, the facility provided aiequate support for the exam.
E9 Review of UFSAR commitments A recent discovery of a licensee operating their facility in a manner contrary to the updated final safety analysis report (UFSAR) description highlighted the need for a special focused review that compares plant practices, precedures and /or parameters to the UFSAR descriptions. While performing the preexamination activities discussed in this report, the inspectors reviewed applicable portions of the UFSAR that related to the selected examination questions or topic areas. No discrepancies were identified as a result of this review.
V. Mananement Meetinas
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X1 Exit Meeting Summary On October 3,1997, the exarniners discussed their observations from the exams with Hope Creek operations and training management, led by Joe Zebo, Acting Training Manager, and Larry Wagner, Operations Manager. The examiners expressed their appreciation for the cooperation and assistance that was provided during both the preparation and examination week by licensed operator training personnel and operations personnel.
There were no observed discrepancies between the simulator and the plant, and as such, none were discussed at the exit meeting.
Attachments:
1. SRO Written Examination w/ Answer Key 2. NRC Resolution of Hope Creek Written Exam Comment
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e Attachment 1 HOPE CREEK SRO WRITTEN EXAM W/ ANSWER KEY
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ATTACHMENT 1
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U.S. Nuclear Regulatory Commission Site-Specific Written Examination
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ApplicantInformation Name:
Region:l Date: 9/29/97 Facility / Unit Hope Creek License Level: SRO ReactorType: GE StartTime:
Finish Time:
l Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the l
answer sheets. The passing grade requires a final grade of atleast 80.00 percent. Examination papers l
will be collected four hours after the examination starts.
(d Applicant Certification All work done on this examination is my own. I have neither given nor received aid.
Applicant's Signature Results Examination Value Points Applicant's Score Points Percen5 Applicant's Grade
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What is the first point during a startup when a licensed operator trainee can withdraw control rods WITilOUT the Operations Manager's written authorization?
a.
After all IRMs are on or above Range 3.
b..
After reactor power is above the Point of Adding lleat.
c.
Afler the MODE switch is placed in RUN.
d.
After the Generator is synched to the Grid.
2.
The Nuclear Engineer has requested the Nuclear Shift Supervisor (NSS) come to the computer room to discuss a P-1.
The NSS is allowed to enter the computer room:
after ensuring another SRO is in the control room.
a.
b.
after ensuring that another on shift SRO is available in audible range of the Reactor Operator (RO) at the controls or audible range of the control room annunciators.
for up to 15 minutes without any other action.
c.
d.
only after conducting a format turnover to another SRO.
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3.
Followirig severi vacation days, an operator is scheduled to work regular shifts for four continuous days.
Identify the days, if any, the operator can work four hours of overtime and still work all regularly scheduled hours WITilOUT approval from the department manager?
a.
No overtime can be worked.
b.
Only on the first day.
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c.
Only on the first and third day.
d.
Only on the first and last day.
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4.
EHC logic problems are occurring on a weekend and the Senior Nuclear Shin Supervisor (SNSS)
wants to call out the EllC system engineer.
In order to call out the E11C system engineer, the SNSS:
a.
can directly call out the EHC system engineer, but hiUST notify the Technical hianager as
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soon as possible.
b.
can dirvetly call out the EHC system engineer, c.
must obtain permission from the Operations hianager to call out the EliC system engineer.
d.
must contact the Technical hianager to call out the EHC system engineer.
5.
Which of the following is an acceptable power ewirsion?
a.
103 % for 10 continuous minutes b.
102 % for 20 continuous minutes, c.
101 % for 25 continuous minutes.
d.
100.5 % for 75 continuous minutes (~
A Limiting Condition for Operation (LCO), other than LCO 3.0.3, is entered at 1000 on October 12, 6.
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1997. This LCO has a requirement to be in 110T SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Current time estimates for the repair are six to ten hours.
When, on October 12,1997, is the shutdown required to be commenced?
a.
1100 b.
1400 c.
1600 d.
2000 f
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17; Following a LGCA, the SPDS Coollig System Injection Statur display has Core Spray labeled as
"lNJ".
Thisindicstion should; a.
- be used by the operators as an inGcation that both core spray subsystems are injecting at
- thdr design flow rate,
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b.
be used by the operators as an indication that at least one core spray subsystem is injecting at its design flow rate.
c.
only be used with otner indications because its based cu system flow and the test flow valve being closed.
d.
only be used wit' ather indication because it uses indications that are NOT environmentally qualified.
' 5t. A surveillance procedurn is to be used as a working copy by an Equipment Operator (NEO).
The working copy is valid for:
a.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
/daya i. )
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30 days.
d.
-indefinitely unless revised, deleted or superseded.
9.
Reactor temperaturc is 185 * F. The MODE switch has been placed in STARTUP/IiOT STANDBY for RPS testing.
Ti e reactor mode is in:
a.
Cold Shutdown and all requirements for being in Operational Condition 2 must be met, b.
Cold Shutdown and operation in this condition is allowed provided control rods are verified to be fully inserted.
c.
Startup/ Hot Standby and all requirements for being in Operational Condition 2 must be met.
d.
Startup/ Hot Standby and operation in this condition is allowed provided control rods are verified to be fully inserted.
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l10 : Identify how to perform'an independent verification on a throttle valve that is required to be open -
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2 three turns.
a.
.- Verify that the valve is open.-
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b; L Observe the valve being opened three turns.' Count the valve turns.
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.cf _ Close the valve one turn, then reopen the valve one turn.
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d.
' Close the valve fully, then reopen the valve three turns.
11. A breaker is Red Blocking Tagged in the TD position. It is presently.n the DISCONNECT
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. position. In order to rack the breaker to the TEST position:
9ygg a.
the Red Blocking Tag (RBT) must be released and a new RBT issued for the breaker in the
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TEST position.-
b.
the operator must verify RBT required position is TD.
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a temporary tagging release for that breaker is issued and the RBT is rehung in the TEST
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position.-
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1 d.
a Circuit Breaker Position Change Request is completed by the job supervisor and approved
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by the Work Control Center SNSS/NSS.
lD; 12. Identify the individual (s) who are responsible for controlling access to the refuel bridge during vessel refueling.-
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Always the Refuel SRO b.
The Refueling Coordinator
c.
The Refueling Bridge Operator The Refueling SRO, when on the brid e, otherwise it is the Refueling Bridge Operator d;
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13. - Purging or venting the containment is limited to 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> per 365 days.
= ldentify the activity that is NOT subject to this restriction.
- a.
Venting, for pressure control, via any monitored flow path.
b.
. Venting, for pressure control, via the 2 inch bypass lines.
c.
Venting, in order to inert the drywell, via any monitored flow path.
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d
' Venting, in order to inert the drywell, via the 2 inch bypass lines.
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1.4.
A Site Area Emergency has been declared. A person called out for a TSC position indicates he
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consumed one beer four hours before the call. The individual also indicated that he felt that he was -
' Fit for Duty".
The individual should be instructed to:
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report to the site for a breathalyzer test before entering the protected area.
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b.
come in after the five hours since drinking the beer had expired.
standby to be called out if no other individuals qualified for that fSC position could be c.
contacted, d.
report to the site for a behavioral observation before entering the protected area.
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15. A review of a previous scram provided the following information.
+ - The turbine generator had been on the grid approximately 30 minutes.
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. An EHC malftmetion caused pressure oscillations.
- The crew inserted a manual scram because of the oscillations.
- Review of the alarm pintout indicated that RPS failed to scram the reactor on high pressure prior to the' manual scram.-
. No Emergency Declarstions or NRC Notifications were made.
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Identify the required actions.
a.
Declare an Alert. With the notification of the Alert also ca::cel the Emergency Condition.
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Declare a Site Area Ercergency. With the notification of the Site Area Emergency also cancel the Emergency Condition.
c.
Perform a i hour report.
_ d.
Perform a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reportc
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16. An Alen h'as just been declared.: Select when accountability will be performed.
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Always on the Alert.
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At the SNSS discretion during the Alert, but ifin a Site Area Emergency, accountability will-be required.
c; During the Alen only if fuel damage has occurred or high radiation levels are identified.
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d.
- During the Alert only'on a loss of one or more fuel barriers.
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17i - An Unusual Event hasjust been upgraded to an Alert.
- The Operations Support Center (OSC):
may have been activated at the UE, but is required to be activated for the Alert, a.
may have been activated at the UE a' d activation remains optional for the Alert.
b. ',
n c.
shall not be activated for a UE and activation for the Alert is optional, d.
. shall not be activated for a UE but is required to be activated for the' Alert.
18. A control rod has been inserted one notch but the SETTLE light does NOT illuminate indicating the SETTLE function failed.
This indicates:
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a-the Withdraw Drive Valve failed to close when required.
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b.
- both the Withdraw Exhaust Valve and Withdraw Drive Valve failed to open when required.
c.
- the Withdraw Exhaust Valve failed to close when required.
d.
- the Withdraw Exhaust Valve failed to open when required.
6*.) ' 19. Which of the following combinations of valve positions can damage a control rod drive if a scram
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were to occur?
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a.
1-BF-V102, Withdraw Riser Valve Closed 1-BF-V101, insert Riser Valve - Open
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b.
1-BF-V101, insert Riser Valve - Closed 1-BF-V102, Withdraw Riser Valve - Open c.
1 BF-V103, Drive Water Riser Valve - Open
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1-BF-VI 12, Scram Discharge Riser Valve - Open d.
1-BF-V101, Insert Riser Valve - Closed
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1-BF-Vil2, Scram Discharge Riser Valve'- Closed -
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20. Given the following conditions:
- Reactor power is 8%.
A control rod is being withdrawn from position 12 to 24.
- The rod has failed reed switches at positions 18 and 20.
Which of the following describes the actions, in RSCS, requir:d to withdraw the control rod?
a.
The rod will not have to be bypassed to withdraw to position 20, but wi" have to be bypassed to withdraw to position 22.
b.
The rod will have to be bypassed to withdraw to position 20.
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c.
A substitute position will not be required to withdraw to position 20, but will require a substitute position to withdraw to position 22.
d.
A substitute position will be required to withdraw to position 20, then a substitute position will also be required to withdraw to position 22 21. Given the following conditions:
Reactor is being shutdown with reactor power at 5%.
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- Rods in RWM group 63 are being inserted.
A rod is selected from group 61.
RWM will generate a select error and:
a.
the rod cannot be moved.
b.
the rod can be moved. Following movement an insert error will be received but no insert block.
c.
the rod can be moved. No insert error will be received when the rod is moved.
d.
an insert error and the rod can be moved.
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22. Given the following conditions:
- Reactor power is 20%.
- An EllC fluid leak has occurred causing complete depressurization of EHC.
During the resulting transient, the recirculation pumps trip.
What caused the recirculation pumps to trip?
a.
Low ETS pressure.
b.
Turbine stop valve closure.
c.
Reactor vessel water level decreasing to -25 inches.
d.
Reactor pressure increasing to 1080 psig.
23. The reactor is operating at 48% power when an instrumentation technician causes a rero percent feedwater flow signal to be sensed by the recirculation flow control system. The instrumentation technician recognizes the error and removes the cause of the low feedwater flow signal.
Select the response of the initial low flow signal and the response when the signal is removed.
Initially recirculatica will mnback to 45%. When the signal is removed the recirculation a.
pumps will immediately begin increasing in speed.
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b.
Initially recirculation will runback to 45% When the rignal is removed the recirculation pumps will remain at 45% until the runback is reset.
Initially recirculation will runback to 30%. When the signal is ren.oved the recirculation c.
pumps will immediately begin increasing in speed.
d.
Initially recirculation will runback to 30%. When the signal is removed the recirculation pumps will remain at 30% until the mnback is reset.
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i 24. The reactor is in cold shutdown with loop "A" of RHR in shutdown cooling. A break results in a l
- loss of reactor coolant inventorys
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- Water level has lowered to.140 inches, j
- Select RHR loop "A" status.-
a[
RHR pump (AP202)will be trippedc
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- RHR Injection Valve (HV-F017A) will be full open. -
RHR Suppression Pool Suction Valve (IW)X)4A) will be full open.
= b..
RHR pump (AP202) will be tripped.
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. RHR Injection Valve (HV-F017A) will be full open.
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RHR Suppression Pool Suction Valve (HVpA) will be full closed.
- c.
,HR pump (AP202) will be nmning.
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RHR Injection Valve (HV-F017 A) will be full closed.
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RP.R Suppression Pool Suction Valve (HVf A) will be full open.
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.RHR pump (AP202) will be running RHR Injection Valve (HV-F017A) will be full closed.-
RHR Suppression Pool Suction Valve (HVf A) will be full closed.
V). 25.- _ Which of the following components must be tagged in the closed position in
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inadvertent draining of the reactor to the Suppression Pool during shutdown cooling loop A operation?
HV-F004A, RHR Pump Suppression Pool Suction Valve.
a.-
b.
HV-F007A, RHR Pump Minim 9m Flow Vaive.
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c.
HV-F024A, RHR Loop A Test Return Valve
d.
HV-F027A, Suppression Chamber Spray Isolation Valve (26. _ During a surveillance, HPCI Turbine Exhaust Isolation valve (HV-F071) breaker trips before the CLOSE light goes out when'the valve is being opened.
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What will be the effect on HPCI if an initiation signal is received?
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The HPCI Turbine Stop valve (HV-4880) will be tripped.-
- b.
The turbine wili startup,- then trip on high exhaust pressure.
c.
l The turbine will operate at a lower speed due to the exhaust pressure.
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d.
The HPCI Turbine Steam Supply isolation valve (HV-F001) will not open.
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.1 27.. Tim Core Spray system is operable except, Instmmentation Technicians have just reported that
.JrBB-PISil N654A, Core Spray Loop A Header Pressure is failed.
Select the action that will satisfy Technical Specification actions statements.
Restore to operable status within seven days or verify core spray pressure less than 475 psig a.
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every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for thirty days.
b.
Restore to operable status within thirty days and verify core spray pressure less than 475 psig every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
Declare core spray inoperable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d.
Declare core spray inoperaHe within seven days.
2. The CORE SPRAY LINE BREAK annunciator has alarmed. This indicates that a break has
occurred anywhere:
a.
between the vessel and outside the shroud.
b.
inside the shroud.
c.
downstream of the loop injection valve (HV-F005).
d.
downstream of the testable check valve.
29. When will a SLC/RRCS INITI ATION FAILURE alarm occur if both SLC pumps are inoperable when a failure to scram occurs? Assume.RPV level stabilizes at -50 inches and reactor power remains at 8%
a.
When the RRCS POTENTIAL ATWS alarm occurs.
b.
30 seconds after the RRCS POTENTIAL ATWS alarm occurs.
c.
When the RRCS CONFIRMED ATWS alarm occurs.
d, 30 seconds after the RRCS CONFIRMED ATWS alarm occurs.
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30. A failure to scram has occurred. All APRMs indicate 25%.
Which additional condition must exist in order to initiate the timer to inject SBLC7 a.
Reactor Pressure = 1080 psig.
b.
Reactor vessellevela 30 inches.
c.
One RRCS manual initiation permissive and one manual initiation ;.ashbutton in the same channel depressed.
d.
Suppression pool temperature = 110 * F.
31. Following a reactor scram the CRD SCR AM DISCll VOL WTR LVL lit annunciator is illuminated. The Mode Switch is in Shutdown.
SbV Ntp 14wl b u y M.
Of the power levels listed, which is the maximum power that will allow the scram to be ter.ct without bypassing RPS?
a.
100/125ths on IRM Range 8 b.
3%
c.
12 %
()
d.
2r4 32. Given the following conditions:
- Recirculation flow unit A - 50%.-
+ Recirculation flow unit C 55%.
- Control rod 22 27 is withdrawn.
- The ALARM REF SIIT111is displayed on 10C651C.
The Rod Illock Monitor "A" will block tod withdrawal at:
a.
68%.
b.
66%.
c.
73%.
d 77%.
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l 33. A scactor startup is in progress with IRMs hdicating 40/125ths on Range 2. IRM "A" detector
!
drive is selected when the SRM do,ectors are withdrawn What action will occur?
a.
IRM "A" detector will withdraw with no alarms until its indication lowers io 5/125ths of
.
scale.
b.
IRM "A" will generate an INOP scram signal.
c.
IRM "A" will generate a rod block when it is withdrawn.
d.
IRM "A" will generate a rod block only afler its indication lowe:a to 5/125ths of scale.
34. Given the following corditions:
- A reactor t.tartup is in progress.
- Powcr is increasu.g on a stable period.
- The SRM is NOT bypassed.
Select the FIRST point, as power increases when rods will be able to be withdrawn.
a.
Associated IRMs are on Range 3 or higher.
b.
Associated IRMs are on Range 8 or higher.
c.
Associated IRMs are on Range 9 or higher.
d.
The Mode switch is placed to RUN.
l f-
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-
__ __ _ ___ _ __
_ - _ _ _ - _ _ _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
.
.
35. Givea the following conditions:
P
+ Reactor power is 73%.
'
+ No rod blocks or scram signals are present.
Flow unit A output fails downscale.
- Select the status after placing the Division 1 Flow Unit bypass joystick to "A".
a All rod blocks and scram signals will clear.
b.
All rod blocks will clear. Scram signals will still exist.
c.
All rod blocks from the flow units and comparator will clear. Other rod blocks and scram signals will exist.
d.
The od blocks from the comparator will clear but other flow unit rod blocks will still exist.
36. The APRM bypass switches are positioned to channels "C' and *ll".
The APRM inputs to the RBMs will be:
a.
RilM 'A" APRM "E", RBM *B" - APRM F" b.
R11M "A* - APRM "E", RBM *ll" - APRM "D"
,
i' !
c.
RilM "A" - APRM "A", RilM *B" - APRM *D" d.
RilM "A" APRM "A", RBM B" APRM "B" 37. Given the following conditions:
- A Loss of Coolant Accident occurred.
- A cooldown is in progress.
- During the cooldown wide range level indication was constant at -50 inches.
During the cooldown actual water level was:
a.
constant at 50 inches, b.
lowered from -50 inches.
c.
lowered towards -50 inches.
d.
incicased from 50iaches.
15 _______._______.e_______.__
_.___
.
.
.
t 38. Failure down scale of one RPV pressure input to the Low Low Set (LLS) Logic will result in:
a.
neither LLS SRV operating in the LLS mode.
b, only one LLS SRV operating in the LLS mode.
c.
one LLS operating at the correct setpoint but the other will remain open once it has opened,
,
d.
both LLS SRVs opening but failing to close.
39. A loss of reactor feed has resulted in a reactor scram and automatic initiation orilPCI and RCIC.
With RPV level at +25 inches, which of the following conditions would prevent using RCIC in the prensure control mode?
(Assume bypasses ofinterlocks allowed by EOPs are performed)-
ItCIC tripped on high RPV water level and then lowered to +25 inches.
a.
b.
Low CST level.
c.
Drywell pressure is 2.1 psig.
d.
The RCIC Initiation Logic " Reset" pushbutton has not been depressed.
40. Given the following:
t=0 see LOCA occurs. hum
.
t=2 sec liigh Drywell signal is generated and all equipment responds as required.
.
t-20 sec ADS Cil 11 and D INITIATION PENDING (RPV Level 1) annunciators alarm.
.
t=48 see ADS Cll 11 and D INITIATION PENDING (RPV Level 1) annunciators clear.
.
t=60 see ADS Cil 11 and D INITIATION PENDING (RPV Level 1) annunciators alarm.
.
When will ADS initiate?
a.
t=107 sec b. -
t=125 see c.
t=137 sec d.
t=165 sec
'
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'4 41. The reactor building io suppression chamber vacuum breakera fail to operate when required l
l This may esult in a failure of the:
a.
suppiession chamber caused by external pressure.
b.,
suppression chamber caused by internal pressure.
j c.
drywell caused by external pressure.
d.
_ drywell caused by internal pressure.
42. A leak has occurred in the RWCU system.' The operator depresses the "C" and "D" NSSSS manual isolation pushbuttons.
Select the RWCU system isolation valves response, if any, i
a.
Neither liv F001 or ilV F004 close
,
b.
Only liv-F001 rioses.
c.
Only liv F004 closes.
d.
Both IIV-F001 and liv F/004 close.
'
43. Which of the following describes ALL shutdown cooling isolation protection available from the Remote Shutdown Panel.
'
I
'
SDC will be prevented from being placed in service when pressure is too high.
a.
b.
SDC will be prevented from being placed in service when pressure in too high and if pressure increases while in SDC it will isolate.
c.
SDC will isolate if pressure increases while it is in service, d.
SDC will not be provided any protection from the Remote Shutdown Panel.
.
b
&
.
.
l.
I
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44. The refuel bridge is over the core. There are no loads on refuel bridge eranes. All control rods are
,
fully inserted.
.
t i
Which of the following lists all Reactor Mode switch positions that will result in a rod block?
a.
S11UTDOWN and STARTUP
- b.'
SIIUTDOWN and REFUEL c.
SilUTDOWN, REFUEL, and STARTUP d.
REFUEL and STARTUP
[
.
-
-
-
-
,
45. A surveillance is being performed which requires the operator to operate the test pushbutton for the Main Stearn isolation Valve while its control switch is placed in CLOSE. The pushbutton sticks in the depressd position.
t
.
Select the response of the Main Steam twh4. Wve.
a.
- When the valve reaches 90% it will reopen.
b.
When the valve reache 90% it will stop closing and remain at 90%.
c.
The valve will fast close once the 90% closure point is passed.
L.4 d.
The valve will continue to slow close unless the control switch is placed in the OPEN position.
P 46. Loss of RPS A will deenergize one solenoid on:
,
a.
all MSIVs.
b.
only the Inboard MSIVs,
c.
only the Outboard MSIVs.
d, all MSIVs on two Main Steam Lines.
L h
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e 47. Given the following conditions:
.
One inboard MSIV failed to fully close.
- All required conditions were met to place MSIV Sealing System in senice.
-
A high nitrogen supply Cow condition for the inboard sealing system has occurred.
- Select the item which lists the action (s) taken by the MSIV Sealing system.
a.
Reduces seal system flow rate until flow is within the allowable range.
b.
Isolates inboard seal systern flow only to the Main Steam Line with the leaking MSIV.
Isolates inboard seal system flow to all main steam lines.
c.
d.
Isolates inboard and outboard seal system flow to the Main Steam Line with the leaking MSIV.
.
48. Reactor power is 97%. Afler decreasing pressure set on EllC, the setpoint continues to lower.
With no operator action, what will be the effect of the setpoint decreasing.
EllC will shill to the other pressure controller and pressure will stabilize at a lower value.
a.
(
)
b.
11ypass valves will open until the muimum combined flow limit is reached then pressure
,
will stabilize.
c.
A reactor scram and MSIV isolation due to low reactor pressure.
d.
Turbine control valves will open until the maximum combined flow limit is reached then pressure will stabilize.
--
.
-
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,
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.. _ _ - -. _,, _._,.
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,
.-
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.
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.
.
,
49. Given the following:
. Reactor power is 85%
- Narrow Range "A' (PDT N004A) = 36 inches.
+ Narrow Range "B"(PDT N004D) = 35 inches.
- Narrow Range "C" (PDT-N004C) = 34 inches.
,
.
If Narrow Range "A" (PDT-N004 A) drills from its present setpoint to zero, actual RPV level will:
a.
remain constant at 35 inches.
b.
increase to 36 inches.
c.
lower to 30 inches then return to 35 inches.
d.
lower until Narrow Range "A" indicates bad quality, then it wi'l return to 35 inches.
50. Given the following:
- Reactor power is 98%.
. All three feedpumps ate in service.
. The operator observes RFP A speed increasing then stabilizing.
- The RFP TURUINE AUTO XFR TO MANUAL annunciator has alarmed.
bI Select how RFP A speed can be controlled, The lockup signal must be reset, then the controller can be controlled ir manual or a.
automatic.
b.
The speed can be varied using the INC SPEED and DEC SPEED buttons.
c.
The speed can onl> be varied locally, d.
The controller must be transferred to single element then controlled manually.
St. Following a steam break inside the drywell, an operator observes that FRVS has initiated and RBVS has isolated. IIPCI has initiated but RCIC did not.
Identify the cause of the FRVS initiation.
a.
Low RPV level b.
1ligh Drywell pressure c.
Reactor Building ventilation high radiation d.
A signal from the LOCA sequencer.
I
l
.
_.
__
__
-. _. _ _ _ _. _ _ _ _. _ _ _. _ _ _ _. _. _ _. _. _
. _ __. _ _ _ _ __ _ _ _
.
.
,
.
,
.
52. - The plant is in a normal, full power lineup. The operator depresses TRIP pushbutton on control room panel 10C651E for breaker 40201, Normal Feed Breaker for 10A402."
'
What will be the status of bus 10A402 and EDG *B"?
Bus 10A402 will be energized with breaker 40208, Altemate Feed Breaker, closed. EDG
!
a.
"B" will NOT be running.
-
b.
Bus 10A402 will be energized with breaker 40208, Alternate Feed Breaker, closed. EDG
- B" will be running.
c.
Bus 10A402 will be energized with breaker 40207, Diesel Generator Output Breaker, closed. EDG 'B" will be running, d.
Bus 10A402 will be deenergized. EIX3 "B' will not be running.
.
$3. The static invener section of a lE 20 KVA Uninterruptable Power Supply (UPS) output has failed to zero.
Output power to the distribution panels will come from the:
a.
125 VDC because it is the backup when the static inverter fails.
_
b.
alternate AC power due to static switch operation.
(, ')
normal AC power because the static inverter section is only used by the battery as a backup
~
c.
to the AC source.
d.
normal AC power because the static invener section is only used by the battery as the primary power source.
54. 125 VDC bus ICD 417 is deenergized and a diesel start signal is received.
Which of the following describes the effect on Diesel Generator ICG4007 a.
The diesel will automatically start but the output breaker can only be shut manually.
b.
The diesel generator will not automatically start.
-
ci The diesel generator will automatically start but in droop mode.
d.
The diesel generator will start but the automatic trips will be disabled.
.
..
.
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-
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.
. - -
___-___ _ _ _ _
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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.
.
55. The diesel generator is running due to a loss of offsite power. A spurious loss of field voltage signal
)
has tripped the Test Lockout relay l
Select the expected status of the diesel generator.
'
a.
Running supplying the 4160 V bus.
)
b."
Running with the output breaker open due to s. lockout on the 4160 V bus.
Running with the output breaker open due a lockout on the EDG Supply Breaker.
c.
.
d.
Tripped with the 4160 V bus deenergized.
56. Following a loss of offsite power (LOP), where the diesel generators start and supply their respective busses, the drywell cooling fans will:
I a.
restart, after a time delay, in the speed they were in prior to the LOP.
b.
restart, after a time delay, in liigh Speed.
,
c.
restart, after a time delay, in Low Speed.
d.
not restart until the operator restores power to MCCs 10B252 and 10B262.
(') 57. During a simultaneous Loss of 0ffsite Power and Loss of Coolant Accident, Control Area Battery
'
Exhaust fan (AV410) will automatically start 60 seconds after the event by the:
a.
LOCA sequencer, ifits control switch is in RUN.
'
b.
LOCA sequencer, ifits control switch is in AUTO.
c.
LOP sequencer, ifits control switch is in RUN.
d.
LOP sequencer, ifits control switch is in AUTO.
58. A trip of a recirculation pump has resultui in operation in the Exit" region of the power to flow map.
Which of the following lists two indications which are both acceptable for monitoring for power
,
oscillations?
a.
APRM Chart Recorder and SPDS computer b.
APRM Meters and CRIDs c,-
APRM Chart Recorders and period meters d.
CRIDs and LPRM r, cts
.
.-
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- --- -.
-
-. -
- - - -.
.
.
.
.
i 59. Given the following conditions:
Reacter power is 65%.
!
Core flow is 100%.
Following a single recirculation pump trip, reducing indicated total core flow less than 40% can
cause:
a.
- excessive power oscillations to occur,
<
b.
inaccurate indicated recirc loop flows I
c.
idle loop cooldown rates to exceed allowable limits.
d.
excessive cooldown of the idle loop.
,
The generator wasjust placed in service when a loss of vacuum occurs. With no operator action,
,
the reactor will scram at:
a.
7.5" IIGA on a Turbine Stop Valve Closure signal.
b.
10.0" IIGA on an RPV Low Water Level, c.
22,9" IIGA on a liigh Reactor Pressure Signal.
d.
21.5" IIGA on a Main Steam Line Closure Signal, l
'
,
61. A loss of all off site power has occurred. All diesel generator have loaded to their respective buses.
When starting Non-Class lE loads diesel generator load is limited to the:
a.
4340 kW.
,
b.
4430 kW.
c.
4774 kW d.-
4873 kW,
,
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62. Following a Loss-of Offsite Power (LOP) RACS will REALIGN to supply cooling to specific chill water loads to ensure cooling to prevent:
l a.
excessi' c temperatuies in the control room.
~
b.
an automatic containment isolation.
.
,
c.'
exceeding design environmental conditions for the IE switchgears.
d.
a loss ofinstrument air from also occurring.
While a diesel generator was running for a surveillance the ' Emergency Stop" pushbuttons were M N 63.
deprened as part of the surveillance. One minute later a loss ofits respective 4160 v bus occurs. t-of Slgw\\.
Select how the diesel generator can be started or will automatically start
'
a.
Following a time deley the engine will restart without operator action.
b.
Following a time delay, the operator must reset the engine and the generator.lackout, then the diesel will restart The operator must reset the engine and generator lockout then, after a time delay, the diesel c.
engine will restart
.
d.
The operator must reset the eagine and generator lockout, then the nerator will
,
automat cally restart.
64. Shutdown cooling is in operation on RilR loop "I3" with RCS temperature at 265 * F. RPS "A" is deenergized.
Sc!cet the status of RilR.
a.
The suction flowpath would be isolated.
,
b.
The return flow path on loop "I3" would be isolated but loop "A" can be aligned for SDC.
c.
The return flow path for loop "13" would be isolated and loop "A" cannot be aligned SDC.
d.
. The suction flowpath and return flow path for loop "I3" would be isolated.
.-
.
.
.
. - -
--
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.
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. _ _ _,, _,
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. _. _. _ _ _ _ _ _. _.__._. _ _.
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65. Reactor power is 92%.
[
Which of the following conditions describes the effect on the turbine controls following a loss of 125 V DC?
!
a.
The turbine will trip on loss of 125 V DC.
b.'
On a trip signal an operator must trip the turbine from the front standard.
e c.
Following a trip reactor pressures will be higher due to slow response of the bypass valves.
d.
Reactor scram from a turbine trip will NOT occur.
66. Given the following conditions:
+ Power ascension is in progress and control rods are being withdrawn.
+ TCV FAST CLOSURE /MSV TRIP BYP is NOT in alarm.
- A turbine trip occurs and reactor power stabilizes a* 33%.
Select the required operator action.
Drive control rods to reduce power to within the capacity of the bypass valves.
a.
b.
Enter llc.OP EO.ZZ-0101, RPV Contrel.
L ' o c.
Perform actions for the loss of feedwater heat:ng that occurred on the turbine trip.
"
,
d.
Enter llc.OP EO.ZZ-0100, Reactor Scram.
,
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.m,
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=
-
,
,...,,..., - - - -...
._.
-
.
_
_ - _ _ _ _
_ _.
__
.
.
.
.
67. Given the following:
- The reactor failed to scrarn and reactor power is 10%.
- The operator observes the following dur ng a review of the pressure trend:
i
- Reactor pressure increases to 1060 psig.
+ Pressure then lowers to 935 psig.
- Pressure then cycles between 1017 and 935 psig.
Which of the following describes how Low Low Set (LLS)is controlling pressure for the given number of LLS valves?
LLS is properly controlling pressure with both LLS valves.
a.
b.
LLS is properly controlling pressure with SRV "11".
c.
LLS is properly controlling pressure with SRV "P".
d.
LLS is NOT properly controlling pressure.
68. A high drywell pressure condition has occurred due to a small break LOCA causing IIPCI to initiate.
I{PCI is the only source ofinjection and the leak is less than the capacity ofilPCI.
in this condition llPCI will:
,
restart as soon as the high level signal clears following a high level trip.
a.
b.
not allow the operator to manually control the llPCI flow rate.
following a high level trip, require the reset of the initiation logic in order to restart liPCI c.
before level lowers to level 2, d.
maintain level between level 2 and level 8.
-
-
--
.
-
-,
. _ _ _ _ _..
_
_
_. _ _ _
_
.
- _ _ _
l
'
.
.
69. The following events occurred in the sequence given:
Both "A" Reactor Recirculation Pump Seals have failed
.
The pump was tripped
.
The pump discharge valve was closed
The pump suction valve close pushbutton was depressed
+
Identify the potential consequence of this event sequence.
a.
Damage to the suction valve.
b.
The suction valve may not be able to be opened after it fully closes.
c.
Increased closure time on the pump discharge valve, d.
An unisolable leak in the RCS.
70. Given the following conditions:
A LOCA has occurred.
- Drywell ternperature is 240 * F.
- Suppression Chamber pressure is 9 psig.
- Suppression chamber sprays are required to be initiated at this pressure instead of drywell sprays to prevent:
a.
exceeding the negative design pressure of the primary containment.
b.
causing a stress failure of downcomer and vent headerjunction due to cyclic condensation.
c.
excessive accumulation of non-condensibles in the suppression chamber.
d.
drywell depressurization that exceeds the capa*y of the suppression chamber to drywell vacuum breakers.
.-
-
- -.
-
-. -
-
.. _ _ _ _
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _
,
.
a i
I
.
71. Given the following conditions:
i
+ liPCI is being operated for a surveillance.
)
+ Suppression pool temperature is 98 ' F.
Technical Specification LCO entry is required:
a.
with the stated conditions.
b.
as soon as the liPCI surveillance is completed.
if Suppression Pool temperature is greater than 95 * F one hour after the test is terminated.
c.
d.
if Suppression Pool temperature is greater than 95 * F 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the test is terminated.-
72. A feedwater heater string is to be removed from service at rated power. Engineering has calculated that RPV feedwater inlet temperature will lower to 390 ' F when the feedwater heater string is removed.
in order to operate with the feedwater heater string removed from service, approval must be received from the:
a.
SORC as a special procedure.
()
b.
Director-Plant Engineering via a memo.
,,
Operations Manager as a verbal approval.
c.
d.
NRC as a license amendment.
73. A significant feedwater heating loss has occurred with reactor power at 95% and core flow at 93%.
Select the required power reduction.
a.
Initially reduce power to at least 75% using recirculation flow. Maintain power less than 75% using recirculation flow, b.
Insert control rods then lower recirculation flow to maintain power less than 75%,
c.
Initially reduce power to at least 75% using reci. ahn flow. Insert control rods per the stuff sheet to prevent an APRM flow biased hig'
.; ram.
d.
Immediately reduce power to at least 75% using control rods per the stuff sheet. Maintain power below the 100% tod line using recirculation flow.
.
.
_ _ _
_,
.-
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..
.
. __. _ _ _ - -
. _ -.__
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74. Given the following conditions:
+ Reactor was operating at 94%.
A MSIV isolation occurred due to improper testing.
.
All rods remained withdrawn following the scram.
CRD SrRAM DISCil VOL NOT DRAINED is not in alarm.
- -APP'
are indicating between 7% and 11%.
- N stue lights are lit on 10C650.
'
Of the listed methods, which would most effectively correct the cause of the failure to scram.
a.
Isolate and vent the scram air header using IIC.OP EO.7A0306.
b.
Manually insert control rods, defeat RSCS and RWM interlocks using IIC.OP-EO ZZ-0307 and llc.OP SO.SF 0003.
c.
Reset the scram and initiate a manual scram. Defeat RPS interlocks using IIC.OP-EO.ZZ-0320.
d.
Vent the control rod overpiston volumes using OP SO.BF 0002.
75. Given the following conditions:
The control room has been abandoned
.
'
+ llPCIisinjecting.
- RCIC is shutdown
- Reactor pressure is 820 psig
+ RPV levelis 440 inches and increasing To prevent RPV overfdl, llc.OP-lO ZZ-0008, " Shutdown Outside the Control Room" requires llPCI:
a.
be tripped at the RSP prior to exceeding liigh Level 8.
b.
discharge valve be shut from the RSP prior to exceeding liigh Level 8.
c.
IIPCI discharge valve b= throttled, by an operator at the valve, to control level, d.
isolation and shutdown to be initiated by opening appropriate circuit breakers at the 125 VDC Distribution Panels.
_ -.
- - -- - - - - _ -
-.-
. _..
., _ - _. - - - -. - - - - - - _. -.-
,
.
76. Given the following conditions:
- A control room evacuation has occurred.
+ Time allowed performance of all control room actions prior to evacuation.
When the control room actions are completed reactor pressure will be controlled by:
a.
b.
lil'Cl and RCIC operating in CST to CST mode.
c.
SRVs operating in L1,5 mode, d.
SRVs opening on high RPV pressure.
77. Given the following conditions:
SACS pumps "A" and "C" are running, supplying TACS loads.
.
SACS pump "11" is running
SACS pump "D" is in standby.
.
Due to improper testing, a low SACS pump "A" differential pressure signal is generated.
.
Following all automatic actions, identify the nmning SACS pumps and cooling to TACS loads.
i
' a.
SACS pumps "II", *C", and "D" are sunning and SACS loop "I3" is supplying TACS loads.
'
b.
SACS purnps "11", "C", and "D" are nmning and both SACS loops are supplying TACS loads.
c.
SACS pumps "A", "11", "C", and "D" are running and SACS loop "II" is supplying TACS loads.
d.
SACS pumps "II", "C", and "D" are running and SACS loop "A" is supplying TACS loads.
78. A main steam line leak has occurred in the turbine building. RPV level has lowered to -48 inches.
PCIG has been isolated to which of the following components?
a.
Safety Relief Valves b-Drywell Unit Cooler Chilled Water valves c.
Reactor Iluilding to Suppression Chamber Vacuum Ilreakers d.
Drywell Equipment and Drain Sump Coolers l
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i 79. - A leak has occurred in the instrutnent air header.
-;
Select the response of the air supplie to lowering instrument air header pressure.
,
If pressure has lowered to 90 psig, both service air compressors will be supplying the
n.
instrument a!r header, b.'
If pressure has lowered to 80 psig, only one service air compressor and the emergency i
instrument air compressor will be supplying the instrument air header.
.
If pressure has lowered to 90 psig, only one service air compressor will be supplying the c.
instrument air header.
If pressure has lowered to ' O psig, instrument air will be isolated from service air and the d.
J
emergency instrument air compressor will be supplying the instrument air header.
-.
80. An unrecoverable loss of primary containment instrument ga. hss occurred Operation below 45 psig is prohibited because:
,
a ADS will become inoperable below 45 psig.
,
b.
multiple rod drifts may occur while the reactor continues to operate.
incomplete closure of the MSIVs may occur due to low nitrogen pressure.
c.
bI d.
the inboard MSIVs will close causing a larger pressure and power transient if the reactor is operating when the MSIVs close.
t
A loss of shutdown cooling has occurred and MSIVs are open. RPV water level is at 495 inches.
RPV level is:
a acceptable because natural circulation has been established, b.
not acceptable because RPV level is above the point at which the MSIVs should have been closed.
acceptable because it is below the level of the Main Steam Lines.
c.
d.
not acceptable because RPV level is below the point at which natural circulation would be
,
established.
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82. Given the following conditions:
- RPV Temperature is 275 * F.
- Shutdown cooling is in r,ervice.
- RPV level decreases to 5 inches before being recovered to + 40 in 'tes.
+ lsolations initiated by the level transient cannot be reset due to equipiaent failure.
Which of the following lists both systems that can be used as alternate methods of decay heat removal?
'
a.
RWCU and RilR llead Spray b.
Condensate Transfer system and RWCU c.
RilR licad Spray and Condensate Transfer system d.
Fuel Pool Cooling and RilR licad Spray 83. Due to a loss of drywell cooling, drywell pressure has increased to 2.0 psig.
Which one of the following lists ALL actions that would have to be puformed to restart a CRD pump?
Depress the CRD pump STOP pushbutton, then depress the CRD pump START a.
')
pushbutton.
b.
Depress the LOCA OVERRIDE pushbutton, the depress the CRD pump START pushbutton.
Depress the LOCA OVERRIDE pushbutton, then depress the CLOSE for the IE breaker c.
on 10C650E.
d, Depress the LOCA OVERRIDE pushbutton, depress the CLOSE pushbutton for the IE breaker on 10C650E, then depress the CRD pump START pushbutton.
84. An irradiated bundle is being moved from the vessel to the fuel pool through the fuel transfer chute when it is damaged. Area radiation levels are increasing.
The bundle should be:
a.
lefl grappled in the fuel transfer chute.
b.
returned to its original location in the vessel.
l c.
placed in the fuel pool.
l I
d.
placed in the first available location in the vessel.
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85. Drywell sprays are required to be secured when pressure is reduced below a specific pressure to prevent:
exceeding the containment negative design pressure.
a.
b.
f educing the suppression chamber pressure below the RllR vortex limits.
c.~
exceeding suppressic 'hamber to drywell vacuum breaker capacity.
d.
exceeding design cooldown rates for the drywell structures.
86. Given the following:
- A LOCA has occurred.
- SRVs have not been opened.
. Suppression chamber pressure hasjust exceeded the Pressure Suppression Pressure Curve.
Exceeding the Pressure Suppression Pressure curve indicates that:
a.
all non condensibles have been removed from the suppression chamber.
b.
Primary Containment Pressure limit has been exceeded.
( )
c.
steam exists in the suppression chamber air space.
d.
suppression pool design load has been exceeded.
87. If PClO is lost while controlling RPV pressure with SRVs, the SRVs are required to be placed to CLOSlior AUTO.
With normal drywell pressure, the accumulators are sired to provide a minimum of:
a.
2 cycles of the SRVs.
b.
3 cycles of ti;e SRVs.
c.
5 cycles of the SRVs.
d.
7 cycles of the SRVs.
I
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88. When the combination of Suppression Pool temperature and RPV pressure cannot be maintained below the lleat Capacity Temperature 1.imit, emergency depressurization is required.
Emergency depressurizing at this point prevents:
a.
suppression pool boiling.
b.'
ECCS pumps, which are taking a suction from the suppression pool, from cavitating.
exceeding suppression chamber design temperature during emergency depressurization.
c, d.
exceeding the range of the suppression pool temperature instrumentation.
89. A steam break has occuned in the drywell coincident with a failure to scram. Emergency depressurization was not initiated at the required drywell temperature.
Select the effect of this condition.
Drywell spray, ifinitiated, will rapidly vaporize causing a rapid pressure increase.
a.
b.
The ability to emergency depressurize cannot be assured.
c.
The ability to monitor drywell temperature is lost.
d.
Design tempe ature of containment seals has been exceeded.
,
90. During high prhnary containment water level conditions, suppression pool water level indications cannot be used.
Operation of which system willinvalidate the alternate method used for determining primary containment water level?
a.
ItCIC b.
Core Spray c.
Rllit d.
IIPCI
-
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91. Suppression pool level has just increased from 120 inches to 130 inches.
Select the consequences of this level increar,e.
If Drywell sprays are initiated the negative design pressure of the containment may be a.
exceeded.
b.'
If multiple SRV openings occur dynamic loading of the suppression pool may be exceeded.
c.
The containment can only be vented via the drywell.
d.
If Suppression Chamber Sprays are initiated they will be ineffective.
92. Which one of the following compon :nts becomes uncovered when suppression pool level drops below 38.5 inches?
a.
IIPCI exhaust b.
RCIC exhaust c.
SRV tail pieces d.
Downcomers 93. Alternate level control is being performed due to a loss of all high pressure feedwater systems. Per the piocedure, Core Spray pumps were started in both Core Spray subsystems.
As reactor pressure lowers during an emergency depressurization, what is the maximum pressure, of those listed below, when core spray will be injecting?
a.
450 psig b,
400 psig i
c.
350 psig d=
300 psig l
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94. Given the following:
f
'
+ A LOCA has occurred.
- RPV Pressure is 100 psig.
+ RPVlevelis 130 inches.
t Exceeding 350 ' F on which TWO Drywell Temperature SPDS points will cause all available RPV level instruments to be considered potentially unreliable due to the temperature near the instrument
runs?
a.
A2266 and A2283.
'
b.
A2283 and A2287
'
c.
A2280 and A2284
?
d.
A2274 and A2281 95. Actions ofilC.OP EO.ZZ-0201, Alternate Level Control, for no injection or alternate injection systems available are being performed.
. If RPV level is allowed to lower to below -200 inches, before emergency depressurizing, what is the maximum clad temperature that will be reached?
'
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a.
1500'F.
b.
1800'F.
c.
2000 ' F.
d. -
2500*F.
.
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96. A fuel bundle is damaged while it is being moved from the vessel to the fuel pool. Refuel floor exhaust ventilation radiation is incteasing.
i Select the alarm setpoint when Reactor 13uilding Control, ilC.OP.E0110103, entry is required
,
and when RIIVS isolates and FRVS initiates.
Reactor fluilding Control entry is sequired before Alert alarm level. RUVS isolation and a-FRVS initiation occurs at the Alert alarm level.
b.
Reactor Building Control entry is required at the Alert alarm level. RDVS isolation and FRVS initiation occurs at the Alert alarm level.
c.
Reactor Building Control entry is required at the Alert alarm level. RBVS isolation and FRVS initiation occurs at the liigh alarm level.
d.
Reactor Building Control entry is required at the liigh alarm level. RBVS isolation and FRVS jnitiation occurs at the liigh alarm level.
97. A failure to scram has occurred resulting in a high suppression pool temperature and drywell pressure of 2.1 psig. IIC.OP.E01&O207, Level / Power Control, has directed tenninating and preventing injection.
Which of the following conditions, by itself, will allow RPV injection to be reestablished?
APRM downscale lights are cycling on and off.
a.
b.
Suppression pool temperature is reduced to below 110 * F.
c.
RPV level is reduced to below the top of active fuel.
d.
All SRVs are closed.
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l 98, llc.0P.EO.ZZ-0207, Level Power Control, is being implemented.
)
I Which ONE of the following sets of conditions indicates that adequate core cooling does NOT i
exist?
j a.
1 SRV open
_
l l
RPV pressure is 500 psig b.
-3 SRVs open i
RPV pressure is $00 psig j
I 4 SRV's open c.
P.PV pressure is 300 psig
. d.
_-$ SRVs open RPV pressure is 300 psig 99. A failure to scram has occurred and reactor power is approximately 65%. The main turbine is on j
line.
l The recirculation pumps afe required to be runhack to minimum speed before tripping the pumps to:
prevent r.3ditional heat loading of the torus if power remains above the bypass valve a
,.
.
capacity.
l C
b, prevent power instabilities due to operating at high power without adequate core flow.
maintain the largest margin to the MCPR limit.
c.
!
d.
ensure llPCI will operate when required.
,
100. A release is in progress followisp treMent that damaged fuel. A sample analysis of the sclease has
,
not been performed.
Which of the following would be used to determine if entry into llc.OP.EO.ZZ-0104, Radioactive Release, was required?
'
i Performance of a dose assessment.
a.
b.
Field Measured Dose Rates.
c,
- Only liigh alarms from Plant Emuent RMS Channels.
d.-
liigh alarm from Plant Emuent RMS Channel and Total Plant Release Rate.
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.
+
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REFERENCE MATERIAL
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FOR HOPE CREEK
~
NRC EXAMINATION 9/29/97 1.
Event Classification Guide Sections 1.0 - 11.0 ONLY 2.
J-0651-1 Rev.16 APRM/RDM/ Flow Unit Controls ONLY 3.
EHC Control Logic Diagram 4.
M-13-0 Rev.19 5.
M-13-1 Rev.31 6.
M-52-1 Rev. 26 7.
Emergency Operating Procedure Flowcharts - Minus the entry conditions 8.
Hope Creek Technical Specifications 3 / 4.0 - 3 / 4.12.3 ONLY
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Ilope Cicek NRC Exam 9/29/97
'
Arner Key
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1.
d 26. d 2.
b 27. a 3.
a 28. d 4,
b 29. d S.
c 30. a 6.
C 31. c 7,
c 32. c 8.
d 33. c 9.
b 34. b 1,
b 35. c II. d 36. b 12. a 37. b 13. b 38. b-i 14. c 39. c 15. c 40. d 16. b 41. a 17, a 42. c 18. d 43. a 19.. a 44. a 20. a 45. d 21. b 46, a 22. d 47. c 23. d 48. c 24. b 49. b 25. d
$0. b Pagei
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Hope Creek NRC Exam 9/29/97
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Answer Key
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$1. b 76. c
$2. d 77. a 53, b 73, c 5,4.
b 79, a 55. a 80. d ph)11 G w s'sns. % !e 6 56.,Vb M CDem b,,
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See M 82. b 57. a 58. c g3, d 59. d 84. c 60. b 85, a 61. b 86, c 62, b 87. c 63, b 88. c 64. a 89, b 65. c 90 d 66. b 91, a 67. d 92. d 68. d 93, c 69. d 94. b 70. b 95. b s
71. b 96. c 72. d 97, c 73. c 98. a 74.
a 99. a 75. d 100. d f* age 2
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Attachment 2
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HOPE CREEK WRITTEN EXAM COMMENT AND NRC RESOLUTION Question #56 Following a loss of offsite power (LOP), where the diesel generators start and supply their respective busses, the drywell cooling f ans will:
a. restart, after a time delay, in the speed they were in prior to the LOP b. restart, after a time delay, in high speed c. restart, after a time delay, in low speed d. not restart until the operator restores power to MCCs 10B252 and 108262 Answer: (a)
- Facility Comment:
Change answer from (a) to (b).
Basis:
The question stem does not mention any special testing occurring during the LOP. The unit cooler fans are provided with two-speed motors (low and high speeds), Thirteen seconds after the sensed LOP, the EDGs will supply electrical power to the MCCs and ultimately the drywell coolers. The drywell cooling fans are started by the LOP sequencers at time = 0. That is, as soon as the EDG restores power to the bus, the drywell cooling f ans are started.
Except during integrated leak rate testing, the fans are run in high speed. The Hope Creek print (H 86 0, sheet 2) shows the LOP sequencer start input to the drywell cooling f ans high speed start circuit. This input is not effected by the position of the control switch, There is no input from the LOP sequencer to the low speed start circuit. -
NRC Resolution:
Comment is accepted.
.
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