IR 05000354/1997008

Forwards NRC Operator Licensing Exam Rept 50-354/97-08 (Including Completed & Graded Tests) for Tests Administered on 970929-1002

ML20203D235
Person / Time
Site:	Hope Creek
Issue date:	02/19/1998
From:	Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
354-97-08, 354-97-8, NUDOCS 9802250386
Download: ML20203D235 (1)

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O e February 19, 1998

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NOTE T0: NRC Document Control Dest Mati Stop 0 5 D 24 .

FROM: $4 I boafen . Lic OperatingLicensigBranch,R4.ensingAssistant SUBJECT:

OPERATORLICENSINGEXAMINATIONADMINISTEREDON besT. .t v he r.t 19 97 , AT _Ho u Ceu k ,

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DGCKEl #50 di e /

On cob 9- @efot e r97 Operator Licensing Examinations were administered at tt4e referenced fac'111ty. Attached, you will find the following information for processing through NUDOCS and distribution to the NRC (

staff, including the NRC PDR

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Item #1 - a) Facility submitted outline and initial exam submittal, l

designated for distribution under RIDS Code A070.

b) As given operating examination, designated for distribution under RIDS Code A070. --s item #2 - Examination Report with the as given written examination attached, designated for distribution under RIDS Code IE42.

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$ E REGloN I 476 ALLINoALE RoAo KING oF PRUsstA, PENNSYLVANIA 1lW61416

• • • • • October 15, 1997 Mr. Leor, Chief Nuclear Officer and President Nuclear Business Unit Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 SUBJECT: HOPE CREEK SENIOR REACTOR OPL IATOR INITIAL EXAMINATION REPORT 50 354/97 08 (OL)

Dear Mr. Eliason:

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This report transmits the findings of the senior reactor operator (SRO) licensing exams m.-

conducted by the NRC during the waek of September 29 October 2,1997, at the Hope Creek Nuclear Generating Station. Based on the results of the exams, four of the five SRO applicants passed all portions of the exams. One individual failed the written exam and passed the operating test. At the conclusion of the exams, Mr. J. Williams discussed the preliminary findings with Mr. J. Zerbo and Mr. L. Wagner, as well as other members of h

your staff.

/ The exam addressed areas important to public health and safety and was developed and administered under Interim Revision 8 of the Examiner Standards (NUREG 1021). Hope Creek personnel and their contractor developed all segments of the exam, while the NRC provided review, oversight and final approval prior to its administration. Hope Creek training personnel subsequently administered the NRC approved written exam, while the operating tests were administered by the NRC.

In accordance with 10 CFR 2.790 of the NRC's Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

No reply to this letter is required, but should you have any questions regarding this examination, please contact me at 610-337 5211, or by E-m-il at GWM@NRC. GOV.

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Glenn W. Meyer, Chief I Operator Licensing and Human Performance Branch Division of Reactor Safety

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Docket No. 50-354 cnclosure: Initial Examination Report No. 50 354/97-08(OL) w/ Attachments 12 e

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Mr. Leon Distribution w/enci and Attachments 12:

DRS Master Exam File DRS OL Facility File PUBLIC Nuclear Safety Information Center (NSIC)

V. Curley, DRS J. Munro, NRR/DRCH/HOLB G. Siegfried, NRR/DRCH/HOLB Distribution w/ enc!: w/o Attachments 1-2:

Region i Docket Room (with concurrences)

J. Wiggins, DRS J. Williams, Chief Examiner, DRS J. Linville, DRP S. Morris, SRI C. O'Daniell, DRP W. Axelson, DRA J. Linville, DRP  ;

S. Barber, DRP l L. Harrison, DRP

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NRC fesident inspector D f.3 1-> L e Distri'aution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):

W. Dean, OEDO K. Kennedy, OEDO

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R. Frahm, Jr., NRR l

D. Jaffe, Project Manager, NRR R. Ballo, OLB/NRR J. Stolz, PDI 2, DRR DOCDESK e

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g*****,o*[ KING oF PRUsslA. PENNsYLvMIA 19406 1415 October 15, 1997 Mr. Leon Chief Nuclear Officer and President Nuclear Business Unit Public Service Electric & Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038 g SUBJECT: HOPE CREEK SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT 50 354/97-08 (OL)

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Dear Mr. Eliason:

This report transmits the findings of the senior reactor operator (SRO) licensing exams J' %

conducted by the NRC during the week of September 29 October 2,1997, at the Hope Creek Nuclear Generating Station. Based on the resuits of the exams, four of the five SRO applicants passed all portions of the exams. One individual f ailed the written exam and passed the operating test. At the conclusion of the exams, Mr. J. Williams discussed the preliminary findings with Mr. J. Zerbo and Mr. L. Wagner, as well as other members of your staff.

I f The exem addressed areas important to public health and safety and was developed and administered under Interim Revision 8 of the Examiner Standards (NUREG 1021). Hope Creek cersonnel and their contractor developed all segments of the exam, while the NRC provided review, o"ersight and final approval prior to its administration. Hope Creek training personnel subsequently administered the NRC-approved written exam, while the operating tests were administered by the NRC.

l In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

No reply to this letter is required, but should you have any questions regarding this examination, please contact me at 610 337-5211, or by E-mail at GWM@NRC.GC'.

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Glenn W. Meyer, Chief I Operator Licensing and Human Performance Branch Division of Reactor Safety Docket No. 50-354 Enclosun  :' Examination Report No. 50-354/97-08(OL) w/ Attachments 1-2 e

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Mr. Leon ,

cc w/enct; w/o Attachments 12:

L. Storz, Senior Vice President - Nuclear Operations E. C mpson, Senior Vice President - Nuclear Enginaering

E. Salowitz, Director - Nuclear Business Support A. F. Kirby, Ill, External Operations - Nuclear, Delmerva Power & Light Co.

J. A. Isabella, Manager, Joint Generation Atlantic Electric M. Bezilla, General Manager - Hope Creek Operations J. McMahon, Director - Quality Assurance & Nuclear Safety Review D. Powell, Manager - Licensing and Regulation R. Kankus, Joint Owner Aff airs A. C. Tapert, Program Administrator Jeffrey J. Keenan, Esquire Consumer Advocate, Office of Consumer Advocate William Conklin, Public Safety Consultant, Lower Alloways Creek Township '

State of New Jersey

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State of Delaware cc w/enci and Attachn.ents 12:

J. McMahon, Director - Nuclear Training J. Nichols, Manager, Training Operations O. Havens, Piincipal Training Supervisor

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. Mr. Leon Distribution w/enel and Attachments 12:

DRS Master Exam File DRS OL Facility File PUBLIC Nuclear Safety Information Center (NSIC)

V. Curley, DRS J. Munro, NRR/DRCH/HOLB G. Siegfried, NRR/DRCH/HOLB Distribution w/ encl: w/o Attachments 1-2:

Region I Docket Room (with concurrences)

J. Wiggins, DRS J. Williams, Chief Examiner, DRS J. Linville, DRP S. Morris, SRI C. O'Daniell, DRP W. Axelson, DRA J. Linville, DRP S. Barber, DRP L. Harrison, DRP

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NRC Resident inspector O ( D l-< L M

! Distri')ution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):

l W. Dean, OEDO j K. Kennedy, OEDO l R. Frahm, Jr., NRR I D. Jaffe, Project Manager, NRR R. Ballo, OLB/NRR J. Stolz, PDI 2, DRR DOCDESK

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V. S. NUCLEAR REGULATORY COMMISSION *

• REGION 1 Docket No.: 50-354 Report No.: 97-08 -j License No.: NPF-57 Licensee: Public Service Electric & Gac Company Post Office Box 236 Hancocks Bridge, New Jersey 08038

Facility: - Hope Creek Nuclear Generating Station Location: Hancocks Bridge, New Jersey Datas: . - September 29- October 2,1997

Chief Examiner: J. Williams, Senior Operations Engineer / Examiner, Region i Examiners: L. 3riggs, Senior Operations Engineer / Examiner, Region I B. Maier, Operations Engineer / Examiner, Region I Approved By: Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch Division of Recctor Safety i.. .

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EXECUTIVE SUMMARY Hope Creek Nuclear Generating Station Inspection Report No. 50-354/97 08 Operations Fivs Hope 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.

Overall, candidate performance during the operating tests was determined to be good. The examiners did not determine any generic performance weaknesses.

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Reoort Details ,

l. Ooerations 05 Operatur Training and Qualifications

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05.1 Senior Reactor Ooerator 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 l' censing tests to five senior reactor operator (SRO) instant candidates. The written exam was administered by the facility's training organization, b. Observations and Findingg The results of the SRO exam is summarized below:

SRO Pass / Fail Written 4/1 Operating 5/0

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Overall 4/1

The written exam, job performance measures (JPMs) and simulator scenarios were developed by Hope Creek in accordance with interim Revision 8 of NUREG 1021

" Examiner Standards." The f acility's exam development team was comprised of contractors,and training and operation department representatives. Allindividuals signed onto a security agreement once the development of the exam commenced. ,

The NRC subsequently reviewed and validated all pcrtions 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 the revised answer.

The c,perating tests were conducted from September 30- October 2,1997. The operating tests consisted of three simulator scenarios, ten JPMs and an admini:;tration portion for each candidate. All JPMs were followed up with two system-related questions. The administrative portion consisted of a mix of questions and JPMs.

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. 2 Based on the grading of the written exam, the following question subject areas were missed by more than half of the 9pplicants. This could indicate a weakness in l the general understanding of the subject area.

• Requirements for SROs to be present in the control room. (Q2)

• Knowledge of the basis for SPDS parameters. (Q7)

• Requiremants for working copies of procedures. (Q8)

• Requirements for equipment tagging. (011)-

• Knowledge of CRD HCtj valve positions. (Q19)

• Knowledge of RWM rod blocks. (O21)

• Knowledge of impact of failure of RPV pressure instrumentation. (Q38)

• Knowledge of RCIC operation in pressure control mode. (Q39)

• Knowledge of behavior of levelinstrumentation. (Q49)

• Diesel generator restart af ter losc of electrical bus. (Q63)

• RPV emergency depressurization at hijh temperature. (Q89)

• Knowledge of the maximum pressure for CS injection. (Q93)

• Basis for runback before tripping the recirculation pumps during an ATWS.

(Q99)

The f acility indicated that it has reviewed each of the above subject areas to correct possible weaknesses and willimplement programmatic changes as necessary.

Simulator performance by the candidates was good. The examiners noted that crew briefings were routinely performed by the SROs and were. effective.

In the administrative segment of the operating portion of the exam, candidate

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performance was generally good, however, two candidates had difficulty with emergency classification and recommending appropriate protective actions.

There were four seperate principal Hope Creek contacts during the exam develnrment process. This caused some problems in communications between NRC and the facility. Some NRC comnients were lost initially, with the exchange of Hope Creek contacts. All of the facility contacts worked well with the NRC.

c. Conclusions a

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Four of the five GRO instant candidates passed all portions of the exam. One candidate failed the written exam. While the Hope Creek liaisen with NRC was frequently changed and resulted in some problems, the facility provided adequate support for the exam. '

E8 Review of UFSAR commitments A recent discovery of a licensee operating their facility in a manner contrary to the updated final safety analysis report (UFSAR) description highlighted the need for a special focused review that compares plant practices, procedures and /or parameters to the UFSAR descriptions. While performing the 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. Menaaement Meetinos X1 Exit Meeting Summary On October 3,1997, the examiners 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 l

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. Attachment 1 HOPE CREEK SRO WRITTEN EXAM W/ ANSWER KEY

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U.S. Nuclear Regulatory Commission Site-Specific Written Examination

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I ApplicantInformation Name: Region:l Date: 9/29/97 Facility / Unit Hope Creek License Level: SRO ReactorType: GE StartTime: Finish Time:

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Instructions Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percent. Examination papers will be collected four hours after the examination starts.

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Applicant Cerlification All work done on this examination is my own. I have neither given nor received aid.

Applicant's Signature Results Examination Value Points

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Applicant's Score Points Applicant's Grade Percent

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o 1. What is the first point during a startue when a licensed operator trainee can withdraw control rods WITHOUT the Operations Manager's written authorization?

a. After all IRMs are on or above Range 3.

b., Aner reactor power is above the Point of Adding Heat.

c. After the MODE switch ir placed in RUN.

d. After the Generator is synched to the Grid.

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2. The Nuclear Engineer has requested the Nuclear ShiR Supervisor (NSS) come to the computer room to discuss a P-1.

The NSS is allowed to enter the computer room:

a. after ensuring another SRO is in the control room.

l b. after ensuring that another cn-shift SRO is available in audible ralige of the Reactor l Operator (RO) at the controls or audible range of the control room annuaciators.

c. for up to 15 minutes without any other action.

d. only after conducting a formal turnover to another SRO.

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3. Following sevcn 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 WITHOUT approval from the department mc.ager?

a. No overtime can be worked.

b. Only on the first day.

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 Shift Supervisor (SNSS)

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wants w call out the EHC system engineer, in order to call out the EHC system engineer, the SNSS:

a. can directly call out the EHC system engineer, but MUST notify the Technical Manager as

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soon as possible.

b. can directly call out the EHC system engineer.

c. must obtain permission from the Operations Manager to call out the EHC system engineer.

d. must contact the Technical Manager to call out the EHC system engineer.

5. Which of the ivilowing is an acceptable power excursion?

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

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6. A Limiting Condition for Operation (LCO), other than LCO 3.0.3, is entered at 1000 on October 12,

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1997. This LCO has a requirement to be in HOT 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 :en hours.

When, on October 12,1997, is the shutdown required to be commenced?

a. I100 b. 1400 c. 1600

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d. 2000

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7. Following a LOCA, the SPDS Cooling System injection Status display has Core Spray labeled as

"INJ" j

This indication should:

a. be used by the operators as an indication that both core spray subsystems are injecting at

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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 other indications because its based on system flow and the test flow valve being closed.

d. only be used with other indication because it uses indications that are NOT environmentally qualified.

8. A surveillance procedure 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. 7 days.

c. 30 days.

d. indefinitely unless revised, deleted or superseded.

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9. Reactor temperature is 185 * F. The MODE switch has been placed in STARTUP/ HOT STANDBY for RPS testing.

The 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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10. Identify how to perform an independent verification on a throttle valve that is required to be open three turns.

a. Verify that the valve is open. .

b. Observe the valve being opened three turns. Count the valve turns.

c Close the valve one turn, then reopen the valve one turn.

d. Close the valve fully, then reopen the valve three turns.

I 1. A breaker is Red Blocking Tagged in the TD position. It is presently in the DISCONNECT position. In order to :ack the breaker to the TEST position: 9 y gg a. the Red Blocking Tag (RBT) must be released and a new RBT issued for the breaker in the TEST position.

b, the operator must verify RBT required position is TD.

c. a temporary tagging release for that breaker is issued and the RBT is rehung in the TEST position.

d. a Circuit Breaker Position Change Request is completed by thejob supervisor and approved by the Work Control Center SNSS/NSS.

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12. Identify the individual (s) who are responsible for controlling access to the refuel bridge during vessel refueling.

a. Always the Refuel SRO b. The Refueling Coordinator c. The Refueling Bridge Operator d. The Refueling SRO, wher. on the bridge, otherwise it is the Refueling Bridge Operator ,

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.

Identify the activity that is NOT subject to this restriction.

a. Venting, for pressure control, via any monitored flow path.

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b. Venting, for pressure control, via the 2 inch bypass lines.

c. Venting, in order to inert the drywell, via any monitored flow path.

4. Venting, in order to inert the drywell, via the 2 inch bypass lines.

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14. A Site Area Emergency has been declared. A person called out for a TSC position indicates he 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:

a.. report to the site for a breathalyzer test before entering the protected area.

b. come in after the five hours since drinking the beer had expired.

c. standby to be called out if no other individuals qualified for that TSC position could be contacted.

d. report to the site for a behavioral observation before entering the protected area. ,

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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A' JHC malfunction caused pressure oscillations.

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The crew inserted a manual so.am because of the oscillations.

Review of the alarm printout indicated that RPS failed to scram the reactor on high pressure prior to the manual scram.

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No Emergency Declarations or NRC Notifications were made.

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Identify the required actions.

a. Declare an Alert. With the notification of the Alert also cancel the Emergency Condition.

b. Declare a Site Area Emergency. 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 /> report.

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16. An Alert has just been declared. Select when accountability will be performed.

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a. Always on the Alert.

b. At the SNSS discretion during the Alert, but ifin a Site Area Emerge icy, accountability will be req ~uired.

c. During the Alert only if fuel damage hv occurred or high radiation levels are identified.

d. During the Alert only on a loss of one or more fuel barriers.

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17. 'An Unusual Event hasjust been upgraded to an Alert.

The Operations Support Center (OSC):

a. may have been activated at the UE, but is required to be activated for the Alert.

b. , may have been activated at the UE and activation remains optional for the Alert.

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 requked 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. 1 This indicates:

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a. the Withdraw Drive Valve failed to close when required, b. both the Withdraw Exhaust Valve and Withdraw Drive Valve failed to open when required.

c. the Withdraw Exhaust Va' /e failed to close when required.

d. the Withdraw Exhaust Valve failed to open when required.

i 19. Which of the following combinations ofvalve positions can damage a control rod drive if a scram were to occur 7 a .- 1-BF-V102, Withdraw Riser Valve - Closed 1-BF-V101, insert Riser Valve - Open b. 1-BF-V101, insert Riser Valve - Closed 1-BF-V102, Withdraw Riser Valve - Open c. 1-BF-V103, Drive Water Riser Valve - Open 1-BF-VI12, Scram Discharge Riser Valve - Open d. 1-BF-V101, insert Riser Valve - Closed 1-BF-VI 12, 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.

Tne rod has failed reed switches at positions 18 and 20.

Which of the following describes the actions, in RSCS, required to withdraw the control rod?

a. The rod will not have to be bypassed to withdraw to position 20, but will have to be bypassed to withdraw to position 22.

b. The rod will have to be bypassed to withdraw to position 20.

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 t 21. Given the following conditions:

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Reactor is being shutdown with reactor power at 5%.

Rods in RWM group 63 are being insened.

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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 insen error and the rod can be moved.

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22. Given the following conditions:

Reactor power is 20%.

An EHC 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.

'a . 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 zero percent feedwater flow signal to be sensed by the recirculation flow control system. The instmmentation l

technician recognizes the en or 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 I

l a. Initially recirculation will runback to 45%. When the signal is removed the recirculation

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pumps will immediately begin increasing in speed.

b. Initially recirculation will mnback to 45%. When the signal is removed the recirculation pumps will remain at 45% until the runback is reset.

c. Initially recirculation will runback to 30%. When the signal is removed the recirculation pumps will immediately begin increasing in speed.

d. Initially recirculation will runback to 30%. When the signal is remov:d the recirculation pumps will remain at 30% until the runback is reset.

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l 24. The reactor is in cold shutdown with loop "A" of RHR in shutdown cooling. A break results in a

. loss of reactor coolant inventory.

j Water level has lowered to -140 inches.

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, a. RHR pump (AP202) will be tripped.

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RHR Injecti on Valve (HV-F017A) will be full open.

RHR Suppression Pool Suction Velve (HVp04 A) will be full open.

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b. RHR pump (AP202) will be tripped.

RHR Injection Valve (HV-F017A) will be full open.

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RHR Suppression Pool Suction Valve (HVp04A) will be full closed, c. RHR pump (AP202) will be running RHR Injection Valve (HV-F017A) will be full closed.

RHR Suppression Pool Suction Valve (HV-004A) will be full open.

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d. RHR pump (AP202) will be running RHR Injection Valve (HV-F017A) will be full closed.

RHR Suppression Pool Suction Valve (HVp04A) will be full closed.

25. Which of the following components must be tagged in the closed position in order to prevent inadvertent draining of the reactor to the Suppression Pool during shutdown cooling loop A

'

operation?

a. HV-F004A, RHR Pump Suppression Pool Suction Valve.

b. HV-F007A, RHR Pump Minimum Flow Valve.

'

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.

What will be the effect on HPCI if an initiation signal is received?

,

,

a. The HPCI Turbine Stop valve (HV-4880) will be tripped.

I b. The turbine will startup, then trip on high exhaust pressure.

c. The turbine will operate at a lower speed due to the exhaust pressure.

i d. The HPCI Turbine Steam Supply Isolation valve (HV-F001) will not open.

11

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27. The Core Spray system is operable except, Instrumentation Technicians have just reported that 1-BE-PISH-N651A, Cere Spray Loop A Header Pressure is failed.

Select the action that will satisfy Technical Specific ation actions statements. .

a. Restore to operable status within seven dayc or verify core spray pressure less than 475 psig

. every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> fe. 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 inoperable within seven days.

28. The CORE SPRAY LINE BREAK annunciator has alarmed. This indicates that a break has occurred anywhere:

l a. between the vessel and outside the shroud.

b. inside the shroud.

c. downstream of the loop injection valve (HV-F005).

d. dot nstream of the te. . table check valve.

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29. When will a SLC/RRCS INITIATION FAILURE alarm occur if both SLC pumps are inoperable when a failure to scram occurs?. Assume RPV level stabilize; at -50 inches and reactor power remains at 8%.

a. When the RRCS POTENTI'.L ATWS alarm occurs.

b. 30 seconds afler the RRCS POTENTIAL ATWS alarm occurs.

c. When the RRCS CONFIRMED ATWS alarm occurs, d. 30 seconds afler 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 vessel level = -30 inches.

c. One RRCS manual initiation permissive and one manual initiation pushbutton in the same channel depressed. -

d. Suppression pool temperature = 110 * F.

l 31. Following a reactor scram the CRD SCRAM DISCH VOL WTR LVL HI annunciator is illuminated. The Mode Switch is in Shutdown.

SbV He'$h tel b u ypM .

I Of the power levels listed, which is the maximum power that will allow the scram to be reset without bypassing RPS?

a. 100/125ths on IRM Range 8 b. .

c. 12 %

d. 25'

32. Given the following conditions:

Recirculation flow unit A - 50%.

-

Recirculation flow unit C - 55%.

-

Control rod 22 .'7 is withdrawn.

The ALARM REF ZT H1 is displayed on 10C651C.

The Rod Block Monitor "A" will block rod withdrawal at:

a. 68%.

b. 66%.

c. 73%.

d. 77%.

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33. A reactor startup is in progress with IRMs indicating 40/125ths on Range 2. IRM "A" detector drive is selected when the SRM detectors are withdrawn.

What action will occur?

a. IRM "A" detector will withdraw with no r.! arms until its indication lowers to 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 after its indication lowers to 5/125ths of scale.

34. Given the following conditions:

-

A reactor stanup is in progress.

-

Power is increasing on a stable period.

-

SRM detectors are withdrawn except for SRM "A" which fails to withdraw.

-

The SRM UPSCALE OR INOPERATIVE alarm has been received.

-

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 IIGis are on Range 8 or higher.

c. Associated IRMs are on Range 9 or higher.

d. The Mode switch is p; aced to RUN.

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35. Given the following conditions:

a Reactor power is 73%.

.

No rod blot a or scram signals are present.

-

Flow unit A output fails downscale.

Select the status after placing the Division I Flow Unit bypassjoystick 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 red 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 "B" The APRM inputs ta the RBMs w111 be:

a. RBM "A" - APRM "E", RBM "B" - APRM "F" b. RBM "A" - APRM "E", RBM "B" - APRM "D"

'

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c. RBM "A" APRM "A", RBM 'B" - APRM "D" d. RBM "A" - APRM "A", RBM "B" - APPai "B" 37.- Given the following conditions:

-

A Loss of Coolant Accident occurred.

-

A cooldowm 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. increased from -50 inches.

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,

_s

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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 falhng to close.

39. A loss of reactor feed has resulted in a reactor scram and automatic initiation of HPCI and RCIC.

With RPV level at +25 inches, which of the following conditions would prevent using RCIC in the pressure control mode?

(Assume bypasses ofinterlocks allowed by EOPs are performed)

a. RCIC tripped on high RPV water level and then lowered to 425 inches b. Low CST level.

c. Drywell pressure is 2.1 psig.

d. The RCIC Initiation Logic " Reset" pushbutton has not been depressed.

i

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I 40. Given the following:

I t=0 see LOCA occursAcue t=2 see High Drywell signal is generated and all equipment responds as required.

. t=20 see ADS CH B and D INITIATION PENDING (RPV Level 1) annunciators alarm.

t=48 see ADS CH B and D INITIATION PENDING (RPV Level 1) annunciators clear.

+

t=60 sec ADS CH B and D INITIATION PENDING (RPV Level 1) annunciators alarm.

When will ADS initiate?

a. t=107 sec b. t=125 sec c. t=137 sec d. t=165 sec

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41. The reactor building to suppression chamber vacuum breakers fail to operate when required.

This may result in a failure of the:

a. suppression chamber caused by external pressure, b., suppression chamber caused by internal pressure.

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.

'

a. Neither HV-F001 or HV-F004 close b. Only HV-F001 closes.

c. Only HV-F004 closes.

d. Both HV-F001 and HV-F/004 close.

i 43. Which of the following describes ALL shutdown cooling isolation protection available from the Remote Shutdown Panel.

a. SDC will be prevented fiom being placed in senice when pressure is too high.

b. SDC will be prevented from being placed in senice when pressure is too high and if pressure increases while in SDC it will isolate.

c. SDC will isolate if pressure increases while it is in senice.

- d. SDC will not be provided any protection from the Remote Shutdown Panel.

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44. The refuel bridge is over the core. There are no loads on refuel bridge cranes. All control rods are fully inserted.

Which of the following lists all Reactor Mode switch positions that will result in a rod block? ,

a. SHUTDOWN and STARTUP b.' SHUTDOWN and REFUEL c. SHUTDOWN, 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 Steam Isolation Valve while its control switch is placed in CLOSE. The pushbutton sticks in the depressed position.

Select the response of the Main Steam Isolation Valve.

a. When the valve reaches 90% it will reopen.

b. When the valve reaches 90% it will stop closing and remain at 90%.

c. The valve will fast close once the 90% closure point is passed.

d. The valve will continue to slow close unless the control switch is placed in the OPEN position.

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.

.

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47. Given the following conditions:

• A LOCA has occurred causing the RPV to fully depressurize.

One inboard MSIV failed to fully close.  ;

All required conditions were met to place MSIV Scaling System in senice. l

+ A high nitrogen supply flow condition for the inboard sealing system has occurred.

Select the item which lists the action (s) taken by the MSIV Scaling system.

_

a. Reduces seal system flow rate until flow is within the allowable range.

b. Isolates inboard seal system flow only to the Main Steam Line with the leaking MSIV.

c. Isolates inboard seal system flow to all main steam lines.-

j d. Isolates inboard and outboard seal system flow to the Main Steam Line with the leaking MSIV.

48. Reactor power is 97%. After decreasing pressure set on EHC, the setpoint continues to lower.

With r.o operator action, what will be the effect of the setpoint decreasing.

a. EHC will shift to the other pressure controller and pressure will stabilize at a lov.er value.

t. i b. Bypass valves will open until the maximum 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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49. Given the following: -

-

Reactor power is 85%

• Narrow Range "A" (PDT N004 A) = 36 inches.

• Narrow Range "B" (PDT N004B) = 35 inches.

• Narrow Range *C" (PDT N004C) = 34 inches.

If Narrow Range "A" (PDT-N004A) drifts 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" iedicates bad quality, then it will return to 35 inches.  ;

50. Given the following:

• Reactor power is 98%.

.

All three feedpumps are in service.

• The operator observes RFP A speed inc~ asing then stabilizing.

The RFP TURBINE AUTO XFR TO MANUAL annunciator has ala ned. .

Select how RFP A speed can be controlled.

a The lockup signal must be reset, then the controller can be controlled in manual or automatic.

b. The speed can be varied using the INC SPEED and DEC SPEED buttons.

c, The speed can only be varied locally.

d. The controller must be transferred to single element then controlled manually.

51. Following a steam break inside the drywell, an operator observes that FRYS has initiated and RBVS has isolated. HPCI has initiated but RCIC dicI * ot.

Idesty the cause of the FRVS initiation.

a. Low RPV level b. liigh Drywell pressure -

c. Reactor Building ventilation high radiation d, A signal from the LOCA sequencer.

,, . -. --- _a - - - -----

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$2 The plant is in a normal, full power lineup. The operator depresses TRIP pushbutton on control room panel 10C0SIE for breaker 40201, " Normal Feed Breaker for 10A402."

What will be the status of bus 10A402 and EDG "B*7 a. Bus 10A402 will be energized with breaker 40208, Alternate Feed Breaker, closed. EDG

.

• B" will NOT be running.

b. Bus 10A402 will be energized with breaker 40208, Alternate Feed Breaker, closed. EDG

"I4" will be running.

c. Bus 10A4C.' will be energized with breakst 40207, Diesel Generator Output Breaker, closed. EDG "B" will be running.

d. Bus 10A402 will be deenergized. EDG *B' will not be running.

53. The static inverier section of a IE 20 KVA Uninterruptable Power Supply (UPS) output has failed to zero.

Output power to the dhtribution 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.

'

i c. normal AC power because the static inverter section is only used by the battery as a backup to the AC source.

d. normal AC power because the static inverter section is only used by the battery as the primary power source.

54. 125 VDC bus 1CD417 is deenergized and a diesel start signal is received.

Which of the following describes the effect on Diesel Generator ICG400?

l a The diesel will automatically start but the output breaker can only be shut manually. '

b. The diesel generator will not automatically start.

c. The diesel generator will automatically start but in droop mode.

d. The diesel generator will start but the automatic trips will be disabled

,

e

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_ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ . _

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$5. The diercl generator is running due to a loss of offsite power. A spurious loss of field voltage signal i has tripped the Test Lockout relay.

!

i Select the expected status of the diesel generator.

a Running supplying the 4160 V bus.

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b Running with the output breaker open due to a lockout on the 4160 V bus.

c. Running with the output breaker open due a lockout on the EDG Supply Breaker,

, d. Tripped with the 4160 V bus deenergized.

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56. Following a loss of ofTsite power (lop), where the diesel generators stan and supply their l respective busses, the drywell cooling fans will:

a. restart, aller a time delay, in the speed they were in prior to the LOP, b, restart, after a time delay, in High Speed.

c restan, afler a time delay, in Low Speed.

d. not restan until the operator restores power to htCCs 10B252 and 108262.

,

57. During a simultaneous Loss of Offsite Power and Loss of Coolant Accident, Control Area Battery Exhaust fan (AV410) will automatically stan 60 seconds aller 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 resulted in operation in the " Exit" region of the power to flow map.

Which of the foilowing lists two indications which are both acceptable for monitoring for power oscillations?

a. Al Rh1 Chart Recorder and SPDS computer b. - APRh1 hieters and CRIDs c. APRh1 Chart Recorders and period meters d. CRIDs and LPRh1 meters

,

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ____ _ ___

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59. Given the following conditions:

• Reactor 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 powc: oscillations to ur.

b. inaccurate indicated recirc loop flows  ;

c. idle loop cooldown rates to exceed allowable limits.

d. excessive cooldown of the idle loop.

60 The generator was just placed in service when a loss of vacuum occurs. With no operator action, the reactor will e, ram at:

a 7 P llGA on a Turbine Stop Valve Closure signal.

b. 10.0" llGA on an RPV Low Water Level.

c. 22.9" 11GA on a liigh Reactor Pressure Signal.

d. 21.5" llGA on a Main Steam Line Closure Signal.

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.

,

E

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62. Following a Loss-c f Offsite Power (LOP) RACS will REALIGN to supply cooling to specific chill water loads to enst te cooling to prevent:

a. excessive temperatures in the control room.

.

b. an automatic containment isolation.

c: exceeding design environmental conditions for the 1E switchgears.

d. a loss ofincrument air from also occurring.

63. While a diesel generator was running for a surveillance the " Emergency Stop" pushbuttons were b Ne depressed as part of the surveillance. One minute later a loss ofits respectivcA160 v bus occurs LoF 5)p\ .

Select how the diesel generator can be started or vill automatically start, a. Following a time delay the engine will n. start without operator action b. Following a time delay, the operator rnust reset the engine and the generator. lockout, then the diesel will restart-c. The operator must reset the engine and generator lockout then, afler a time delay, the diesel engine will restart.

d.

Ah The operator must reset the engine and generator lockout, then the* generator will automatically restart.

64. Shutdown cooling is in operation on RilR loop "B" with RCS temperature at 265 * F. RPS "A" is deenergued Select the status of RllR a. The suction dowpath would be isolated.

b. The return flow path on ILp "B' would be isolated but loop "A" can be aligned for SDC.

c. The return Dow path for loop *B" would be isolated and loop "A" cannot be aligned SDC.

d. The suction Dowpath and return Dow path for loop "B" would be isolated.

24 l l

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65. Reactor power is 92%.

Which of the followinst conditions describes the effect on the turbine controls following a loss of 125 V DC7 i a. The turbine will trip on loss of 125 V DC.

t. On a trip signal an operator must trip the turbine from the front standard.

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. l 66. Given the following conditions:

Power ascension is :n progress and control rods are being withdrawn.

. TCV FAST CLOSURE /MSV TRIP BYP is NOT in alarm.

1 . A turbine trip occurs and reactor power stabilizes at 33%.

Select the required operator action.

a. Drive control rods to reduce powcr to within the capacity of the bypass valves.

b. linter ilC.OP-EO.ZZ-0101, RPV Control.

c. Perform actions for the loss of feedwater heating that occurred on the turbine tnp.

d. Enter liC.OP EO.ZZ-0100, Reactor Scram.

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67. Given the following:

• A MSIV isolation occurred due to a loss of PCIO.

• The reactor failed to scram and reector power is 10%.

• The operator observes the following during a revie w of the pressure trend:

• 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?

a. LLS is properly controlling pressure with both LLS valves.

b. LLS is properly controlling pressure with SRV *I1" c. LLS is properly controlling pressure with SRV "P",

d. LLS is NOT properly centrolling pressure, l

68. A high drywell pressure condition has occurred due to a small break LOCA causing liPCI to initiate. l

'

IIPCI is the only source ofinjection and the leak is less than the capacity ofIIPCI.

In this condition llPCI will:

a. restert as soon as the high level signal clears following a high level trip.

b. not allow the operator to manually control the li'Cl flow rate.

c. following a high level trip, require the reset of the initiation logic in order to restan liPCI before level lowers to level 2.

d. maintain level between level 2 and level 8.

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69 The following events occurred in the sequence given:

• Hoth "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 val,e 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:

l * A LOCA has occurred.

• Drywell temperature is 240 F.

l l. +

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 nq tive design pressure of the primary containment, b. causing a stress failure of downcomer and vent header junction due to cyclic condensation.

c. excessive accumulation of non-condensibles in the suppression chamber.

d. diywell depressurization that exceeds the capacity of tha suppression chamber to drywell vacuum breakers.

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71. Given the following conditions:

• IIPCI 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 llPCI surveillance is completed.

c. if Suppression Pool temperature is greater than 95 * F one hour aner the test is terminated.

'

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 feedwi.ter heater st.ing 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.

c. Operations hianager as a verbal approval.

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. hiaintain 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 recirculation flow. Insert control rods per the stufT sheet to prevent an APRh1 flow biased high flut scram.

d. Immediately reduce power to at least 75% using control rods per the stuff sheet. hiaintain power below the 100% rod line using recirculation flow.

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74. Given the following conditions:

• Reactor was operating at 94%.

{

• A htSIV isolation occurred due to improper testing.

. All rods remained withdrawn following the scram.

• CRD SCRAM DISC 11 VOL NOT DRAINED is not in alarm.

• APRMs are indicating between 7% and 11%.

• No Blue lights are lit on 10C650, Of the listed methods, which would most efTectively correct the cause of the failure te scram.

a. Isc, late and vent the scram air header using IIC.0P EO.ZZ-0306.

b. Manually insert control rods, defeat RSCS and RWM interlocks using IIC.OP EO.ZZ 0307 and ilC.OP-SO.SF-0003 l c. Reset the scram and initiate a manual scram. Defeat RPS interlocks using IIC.0P-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

• llPCl is injecting.

-

RCIC is shutdown

• Reactor pressure is 820 psig

.

RPV level is 440 inches and increasing To prevent RPV overfill, llc.OP lO.ZZ-0008, " Shutdown Outside the Control Room" requires IIPCl:

a. be tripped at the RSP prior to c;meeding liigh L evel S.

b. discharge valve be shut from the RSP prior to exceeding liigh 1.evel 8.

c. liPCI discharge valve be 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.

.

b

)

.

.

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 eeactor pressure will be controlled by:

a.' turbine bypase valves.

b. IIPCI and RCIC operating in CST to CST mode.

c. SRVs operating in LLS 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 "B"is running

.

S ACS pump "D" is in standby,

+ Due to improper testing, a low SACS pump "A" differential pressure signalis generated.

Following all automatic actions, identify the running SACS pumps and cooang to TACS loads.

a. SACS pumps *B", "C", and "D" are runnirg and SACS loop "B" is supplying TACS loads.

b. S ACS pumps "B", "C", ar.d "D" are running and both S ACS loops are supplying TACS loads.

c. SACS pump' "A", "B", "C", and "D" are running and SACS loop "B" is supplying TACS loads.

d. SACS pumps "B", "C", and "D are running ad SACS loop "A" is supplying TACS loads.

78. A main stum 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 Building to Suppression Chamber Vacuum Breakers

- d. Drywell Equipment and Drain Sump Coolers

)

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79. A leak has occurred in the instrument air header.

Select the response of the air supplies to lowering instrument air header pressure.

a. If pressure has lowered to 90 psig, both service air compressors will be supplying the instrument air header, b.' If pressure has lowered to _80 psig, only one service air compressor and the emergency instrument air comptessor will be supplying the instrument air header, c. If pressure has lowered to 90 psig, only one sersice air compressor will be supplying the instrument air header, d. If pressure has lowered to 80 psig, instrument air will be isolated from service air and the l

emergency instrument air compressor will be supplying the instrument air header, u 80. An unrecoverable loss of primary containment instrument gas has occurred. Operation below 45 i psig is prohibited because:

a. ADS will become inoperable below 45 psig.

b. multiple rod drifts may occur while the reactor continues to operate.

c. incomplete closure of the MSIVs may occur due to low nitrogen pressure.

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d. the inboard MSIVs will close causing a larger pressure and power transient if the reactor is operating when the MSIVs close, t 81. -A loss of shutdown cooling has occurred and MSIVs are open. RPV water level is at +95 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.

c, acceptable because it is below the level of the Main Steam Lines.

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:

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RPV Temperature is 275 * F.

• Shutdown cooling is in service.

• RPV level decreases to -5.nches before being recovered to + 40 inches.

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Isolations initiated by the level transient cannot be reset due to equipment failure.

Which of the following lists both systems that can be used as alternate methods of decay i at removal?

a. RWCU and RilR llead Spray b Condensate Transfer system and RWCU c. RIIR llead Spray and Condensate Transfer system d. Fuel Pool Cooling and RIIR llead Spray 83. Due to a loss of drywell cooling, drywell pressure has increased to 2.0 peig.

Which one of the following lists ALL actions that s ,uld have to be performed to restart a CRD pump?

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a. Depress the CRD pump STOP pushbutton, then depress the CRD pump START

.

pushbutton.

b. Depress the LOCA OVERRIDE pushbutton, the depress the CRD pump START pushbutton.

c. Depress the 1.OCA OVERRIDE pushbutton, then depress the CLOSE for the IE breaker 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. left grappled in the fuel transfer chute.

b. returned to its original location in the vessel.

c. placed in the fuel pool.

d placed in the first avaliable location in the vessel

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j 85. Dr>well sprays are required to be secured when pressure is reduced below a .pecific pressure to

prevent

a. exceeding the containment negative design pressure.

i b. reducing the supprersion chamber pressure below the RHR vortex limits.

cf exceeding suppression chamber to drywell vacuum breaker capacity.

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d. exceeding design cooldown rates for the drywell structures i

86. Given the following:

I

• A LOCA has occurred.

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i + SRVs have not been opened.

i

+ Suppression chamber pressure hasjust exceeded the Pressure Suppression

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Pressure Curve.

Exceeding the Pressure Suppression Prescure 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 PClG is lost while controlling RPV pressure with SRVs, the SRVs are required to be placed to CLOSE or AUTO.

With normal drywell pressure, the accumulators are sized to provide a minimum of-a. 2 cycles of the SRVs.

,

b. 3 cycles of the SRVs

! c. 5 cycles of the SRVs.

!

d. 7 cycles of the SRVs.

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a

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T-7 '-P!@ =1r>w -, w- -^ r?% o m ---r v de w y- -iewe -------u-eM--w-s-- =*+-7-y c ,7- -7

. _ . .

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88. When the combination of Suppression Pool temperature and IU'V pressure cannot be maintained below the licat Capacity Temperature Limit, 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.

c. exceeding suppression chamber design temperature during emergency depressurization.

d. exceeding the range of the suppression pool temperature instrumentation.

89. A steam break has occurred in the drywell coincident with a failure to scram. Emergency depressurization was not initiated at the required drywel' a iperature.

Select the effect of this condition.

a. Drywell spray, ifinitiated, will rapidly vaporize causing a rapid pressure increase.

b. The ability to emergency depressurize cannot be assured c. The ability to iaonitor drywell temperature is lost.

d. Design temperature of containment seals has been exceeded.

90. Du.ing high primary containment water level conditions, suppression pool water level indications cannot be used.

Operation of which system will invalidate the abernate method used for determining primary containment water level?

a. RCIC b. Core Spray c. RhR d. IIPCI

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91. Suppression pool level has just increased from 120 inches to 130 inches.

Select ti.e consequences of this level increase.

a. If Drywell sprays are initiated the negative design pressure of the containment may be 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.

l 91 Which one of the following components becomes uncovered when suppression pool level drops I below 38.5 inches?

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a. HPCI 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 procedure, 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 c. 350 psig d. 300 psig t

b i

Ik2

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94. Given the following:

• A LOCA has occurred.

• RPV Pressure is 100 psig. ,

. RPV level is .130 inches.

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 7 a. A2266 and A2283 b. A2283 and A2287 c. A2280 and A2284 d. A2274 and A2281 95. Actions of11C OP.EO.71-0201, 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 ernergency depressurizing, what is the maximuni clad temperature that will be reached?

a. 1500 * F.

b. 1800*F.

c. 20M * 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 exhau:t ventilation radiation is increasing.

Select the alarm setpoint when Reactor Iluilding Control, llc.OP E02Z-0103, entry is required and when RIIVS isolates and FRVS initiates.

a. Reactor Iluilding Control entry is required before Alert alarm level. RDVS isolation and FRVS initiation occurs at the Alert alarm level. ,

b. Reactor Iluilding Control entry is required at the Alert alarm level. RDVS isolation and FRVS initiation occurs at the Alert alarm level c. Reactor Iluilding Control entry is required at the Alert alarm level. RDVS isolation and FRVS initiation occurs at the liigh alarm level d. Reactor Building Control entry is required at the liigh alarm level. RIlVS isolation and FRYS initiation 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-EO ZZ 0207, I,evel/ Power Control, has directed terminating and preventing injection.

Which of the following conditions, by itself, will allow RPV injection to be reestablished?

a. APRM downscale lights are cycling on and off 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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98. IIC.OP EO ZZ-0207, Level Power Control, is being implemented.

Which ONE of the following sets of conditions indicates that adequate core cooling does NOT exist?

a. 1 SRV open RPV pressure is 500 psig b. 3 SRVs open RPV pressure is 500 psig c. 4 SRVs open RPV pressure is 300 psig d. 5 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 line.

The recirculation pumps are required to 1,e runback to minimum speed before tripning the pumps to:

a. prevent additional heat loading of the torus if power remains above the bypass valve capacity.

b. prevent power instabiiities due to operating at high power without adequate core flow.

>

c. maintain the largest margin to the hiCPR limit.

d. ensure llPCI will operate when required.

100. A release is in progress following a transient that damaged fuel. A sample analyris of the release has not been performed.

Which of the following would be used to determine if entry into liC.OP-EO.ZZ-0104, Radioactive Release, was required?

a. Performance of a dose assessment.

b Field Measured Dose Rates.

c. Only liigh alarms frem Plant Effluent RMS Channels.

d. liigh alarm from Plant Efiluent RMS Channel and Total Plant Release Rate.

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REFERENCE MATERIAL FOR HOPE CREEK NRC EXAMINATION

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9/29/97 1. Event Classification Guide Sections 1.0 - 11.0 ONLY

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2. J-0651-1 Rev.16 APRM/RBM/ Flow Unit Controls ONLY 3. EHC Control Logic Diagrarn 4. M-13-0 Rev.19 5. M-13-1 Rev. 31 6. M-52-1 Rev. '6 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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Hope Creek NRC Exam 9/39/97 -

Answer Key

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1. d 26. d 2. b 27. a 3. a 28. d 4. b 29. d

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5. c 30. a 6 c 31. c 7. c 32, c 8 d 33. c 9. b 34. b t

l i

10. b 35. c I 1.

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. 36, b 12. a 37. b

! 13. b 38. b 14. c 39. c 15 c 40. d Ifi b 41, a

17, a 42. c 18 d 43. a l

t

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19. a 44, a 20. a 45. d 21. b 46. a i

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22. d 47. c 23. d 48. c 24 b 49. b ( 25. d 50. b l

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Pagei

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Hope Creek NRC Exam 9/39/97 Answer Key

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51. b 76. c

$2. d 77. a 53. b 78. c

$4. b 79. a 55. a 80. d

,,js}gg E w'ene & teJN I

$6. ,#b W C##' 81. b rpeinu4*%.b e '

57. a See e d ed b g82.

$8 c C oppt,<-bg SW 83. d 59. d 84. c 60. b B5. t 61. b 86. c 62. b 87. c 63. b 88. c 64. a 89. b 65. c 90. d 66. b 91. a

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67. d 92. d 68 d 93. c 69. d 94. b 70. b 95. b 71. b 96. c 72. d 97. c 73. c 98. a 74. a 99? a 75 d 100. d e

Page 2

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Applicant: . i

l Circle your answer, if you change >ont answer write it in the blank, ' '

i 1. a b c d .,__ 26. a b c d ,,,,,_

2. a b c d _

27. a b c d ,_

3. a b c d 28. a b c d 4. a

,

b c d , _ _ 29. a b c d ,__,

5. a b c d ,_,_

30. a b c d _

6. a b c d 31. a b c d ,,,_,,

7. a b c d ,_,,,,

32, a b c d ,,,,,,,,

8. a b c d 33, a b c d ,,,,,,

l 9. a b c d ,__,

34. a b c d ,,,,._ I i

l 10. a b c d _ 35. a b c d ,,,,,,

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l 11. a b c d _

3' a b c d ,_,

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I 2. a b c d 37, a b c d ,

13. a b c d 38. a b c d _

14. a b c d ,,,,,,,,

39, a b c d __,,,

15. a b c d _

40. a b c d 16. a b c d ,_._

41. a b c d _

17. a b c d 42. a b c d _.,,_

18. a b c d __

43. a b c d _

,

19. a b c d ,,.__

44, a b c d _

20. a b c d _,_,

45, a b c d __,,

21. a b c d ,,,._,

46, a b c d _

22. a b c d _ , ,

47, a b c d , , , ,

23. a b c d ,_

48. a b c d _

24. a b c d _

49. a b c d 25. a b c d _,

50. a b c d _

l

,_ _

______________ - --- - - - - .

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O Applicant:

, Circle sour answer, if you change your answer write it in the blank.

51, a 'o c d ,, 76. a b c d 52. a b c d _ 77. a b c d ,___

$3. a b c d _ 73. a b c d ,,,,,,,

54. a b c d ,_ 79. a b c d

_

55. a b c d _ 80. a b c d ,.

56. a b c d ,,,,,,,,

81, a b c d ,,_

$7. a b c d , , , _ 82. a b c d ,,_

58 a b c d . , _ , , 83. a b c d __

59. a b c d ,, 84. a b c d ,

l 60 a b c d , , , , 85. a b c d ,,,,,,,,

61. a b c d ,,,, 86. a b c d ,,_

62. a b c d ,_, 87. a b c d ,,,,,,,

63. a b c d ,,,,,,,, 88. a b c d _

t. 64, a b c d ,,,, 89. a b c d ,,,,,,,,,

65. a b c d , 90. a b c d _,,,

66. a b c d ,_,,,

91, a b c d ,_,

67, a b c d ,_,_, 92. a b c d _

68, a b c d , 93. a b c d ,,._,

69. a b c d , , , , 94. a b c d ,,_

70. a b c d 95. a b c d _

71. a b c d 96. a b c d 72. a b c d ___, 97. a b c d ,,_

73. a b c d 98. a b c d ,_,

74. a b c d ,,_

99, a b c d ,_

75. a b c d __ 100 a b c d __

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Attachment 2 ,

HOPE CREEK WRITTEN EXAM COMMENT AND NRC RESOLUTION Question #56 Following a !9ss of offsite power (LOP), where the diesel generatois start and supply

_

their respective busses, the drywell cooling fans will:

a. restart, after a time delay, in the speed they were in prior to the LOP b. restart, af ter a time delay , in high rpeed c. restart, after a time delay, in low speed d not restart until the operator restores power to MCCs 10B252 ar d 10B262 Answer: (a)

Facility Comment: Change answer from (a) to (b).

Basis: The question stem does not mention any special testing occurring t

during the LOP. The unit cooler fans are provided with two speed motors (Iow and high speeds). Thirteen seconds after the sensed LOP, the EDGs will supply electrical power to the MCCs and ultimately the drywell coolers. The crywell 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 fans 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 l sequencer start input to the drywell cooling fans high speed start l circuit. This input is not effected by the position of the control l

switch. There is no input from the LOP sequencer to the low speed start circuit.

NRC Resolution: Comment is accepted.

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