ML20248L608

Forwards Exam Rept W/As Given Written Exam for Test Administered on 971215-18 at Facility

ML20248L608
Person / Time
Site:	Beaver Valley
Issue date:	03/05/1998
From:	Curley V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9803230274
Download: ML20248L608 (1)

Text

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March 5, 1998 NOTE T0: NRC Document Control Desk Mail Stop 0-5-0-24 FROM: SctI on fe n . Li sing Assistant Opera'ingLicensigBranch.R

SUBJECT:

STERED ON OPERATOR

• Deeenfu rt is-ir 1971.

DOCKET #50-aap LICENSING AT (kaee Va g

EXAMINATION ADMIN)lkv On eennbee 15-/t/Pr1 Operator Licensing Examinations were administered at the referenced' facility. Attached, you will find the following information for processing through NUDOCS and distribution to the NRC staff, including the NRC PDR:

Item #1 - a) Facility submitted outline and initial exam submittal, designated for distribution under RIDS Code A070.

b) As given operating examination. ' designated for distribution under RIOS Code A070.

Item #2 - Examination Report with the as given wr.itten examination attached, designated for distribution ~ under RIDS Code IE42.

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UNITED STATES NUCLEAR REGULATORY COMMISSION 5

l REGloN I o

g 475 ALLENDALE ROAD KING oF PRUSSIA, PENNSYLVANIA 19406-1415 g ,o January 9,1998 Mr. J. E. Cross President Generation Group Duquesne Light Company Post Office Box 4 Shippingport, Pennsylvania 15077

SUBJECT:

BEAVER VALLEY UNIT 1 SENIOR REACTOR OPERATOR INITIAL EXAMINATION REPORT NO. 50-334/97-10 (OL)

Dear Mr. Cross:

This report transmits the findings of the senior reactor operator (SRO) licensing ,

examinations conducted by NRC examiners during the week of December 15 - 18,1997 at the Beaver Valley Unit 1 Nuclear Power Plant. Based on the results of the examinations, all SRO applicants passed all portions of the examinations. At the conclusion of the examination, Mr. P. Bissett discussed the preliminary findings with other members of your staff.

These examinations addressed areas important to public health and safety and were developed and administered under interim Revision 8 to the Examiner Standards (NUREG-1021). For this particular examination, Beaver Valley Power Station (BVPS) personnel developed all segments of the examination, while the NRC provided oversight and final approval prior to the administration of the examination. BVPS training personnel 1 subsequently administered the NRC-approved written examination, while the operating portion of the examination was 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.

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(____________--_--

l Mr. J. E. Cross 2 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. GOV.

Sincerely, t c,  ;

Glenn W. Meyer, Chief Operator Licensing and Human Performance Branch Division of Reactor Safety Docket No. 50-334

Enclosure:

Initial Examination Report No. 50 334/97-10(OL) w/ Attachments 1 and 2 cc w/ encl; w/o Attachments 1-2:

Sushil C. Jain, Vice President, Nuclear Services R. LeGrand, Division Vice President, Nuclear Operations Group & Plant Manager W. Kline, Manager, Nuclear Engineering Department  !

8. Tuite, General Manager, Nuclear Operations Unit M. Pergar, Acting Manager, Quality Services Unit J. Arias, Director, Safety & Licensing Department J. MacDonald, Manager, System and Performance Engineering M. Clancy, Mayor Commonwealth of Pennsylvania State of Ohio State of West Virginia cc w/enci and Attachments 1-2:

K.Beatty, General Manager, Nuclear Support T. W. Burns, Director of Training T. Kuhar, Supervisor of Licensed Operator Training

1 Mr. J. E. Cross 3 Distribution w/enci and Attachments 1-2:

DRS Master Exam File PUBLIC Nuclear Safety information Center (NSIC)

V. Curley, DRS Distribution w/ encl: w/o Attachments 1-2:

Region 1 Docket Room (with concurrences) 1 J. Wiggins, DRS P. Bissett, Chief Examiner, DRS P. Eselgroth, DRP ,

D. Haverkamp, DRP l N. Perry, DRP l W. Axelson, DRA l NRC Resident inspector DRS OL Facility File DRS File Distribution w/ encl: w/o Attachments 1-2 (VIA E-MAIL):

I D. Brinkman, NRR R. Correia, NRR S. Bajwa, NRR W. Dean, OEDO (WMD)

R. Frahm, Jr., NRR S. Richards, OLB/NRR ,

J. Stolz, NRR J. Harold, NRR D. Kerns - SRI - Beaver Valley G. Wunder, NRR F. Talbot, NRR DOCDESK Inspection Program Branch, NRR (IPAS) i I

)

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U. S. HUCLEAR REGULATORY COMMISSION REGION 1 Docket No.: 50-334 l

l Report No.: 97-10 l

License No.: DPR-66 Licensee: Duquesne Light Company Facility: Beaver Valley Unit 1 Nuclear Power Plant Location: Shippingport, Pennsylvania Dates: December 15-18,1997 Chief Examiner: P. Bissett, Senior Operations Engineer / Examiner Examiners: T. Kenny, Senior Operations Engineer / Examiner J. Caruso, Operations Engineer / Examiner Approved By: Glenn W. Meyer, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safe;y M CX1 M % _-101tY_ L

i EXECUTIVE

SUMMARY

l l Beaver Valley Unit 1 Nuclear Power Plant inspection Report No. 50 334/97-10 Operations Three Unit 1 senior reactor operator instant (SROl) candidates were administered initial i licensing examinations. All three candidates passed all portions of the license examination.

Generic strengths were noted during the Unit 1 examination in the area of crew communications, control board awareness, and crew briefings during the simulator portion of the operating examination. The NRC examiners observed communications to be direct, succinct, and that all crew members were kept wellinformed. Crew briefings were routinely held during those instances in which time permitted. The written exam was developed at the appropriate SRO knowledge level, as were the job performance measures and follow-up questions. Several JPMs, in lieu of questions, were appropriately developed to test the knowledge level of the applicants in the administrative area of the exam.

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i Report Details I. Ooerations 06 Operator Training and Qualifications 05.1 Senior Reactor Operator Initial Examinations

a. Scope l

The examinations were prepared by Beaver Valley Power Station (BVPS) personnel in accordance with the guidelines in interim Revision 8, of NUREG-1021," Examiner Standards," and Revision 5 of NUREG-BR-0122," Examiners' Handbook for Developing Operator Licensing Written Examinations," and Revision 1 of NUREG-1122, " Knowledge and Abilities Catalog for Nuclear Power Plant Operators:

Pressurized Water Reactors." The examiners administered initial operating licensing examinations to three Unit 1 senior reactor operator instant (SROI) candidates. The written examinations were administered by the facility's training organization.

b. Observations and Findinas The results of SRO examinations for Unit 1 are summarized below:

SRO Pass / Fail Written 3/0 Operating 3/0 Overall 3/0 The written examinations, job performance measures (JPMs) and simulator scenarios were developed by Beaver Valley Power Station (BVPS) representatives in accordance with NUREG-1021. The exam development team was comprised of BVPS training and operation's representatives. All individuals signed a security agreement once the development of the examination commenced. BVPS personnel also validated the exam prior to their submitting it to the NRC. The NRC subsequently reviewed and validated, along with BVPS personnel, all portions of the proposed examinations. Also, various changes and/or additions to the proposed examinations that were requested by the NRC following their review, were subsequently validated and approved. BVPS personnel subsequently incorporated the NRC's comments and finalized the examinations.

The written examination was administered on December 15,1997, and consisted of 100 multiple choico questions. There were no comments by either the NRC or the utility concerning the validity of questions on the written examination, however, the answer key for one question was in error due to a typographical mistake.

Based on the grading of the written examination, one question, #49, was missed by all three candidates, indicating a weakness in the general understanding of the subject area dealing with the effects and indications of a dropped rod at 100%

power.

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l The operating examinations were conducted from December 16 - 17,1997, and l- consisted of three simulator scenarios and ten JPMs. All JPMs were followed up with two system-related questions. All candidates were also examined using JPMs and/or questions to evaluate the administrative requirement portion of the examination.

During the exam preparation week, the examiners noted that the facility routinely developed JPMs involving failed instruments, transmitters, pumps, etc. with each l

failure having already occurred prior to the candidate starting performance of the

JPM. Thus, the candidate would start the JPM with the simulator frozen and the failure already having occurred. The NRC examiners stated that JPMs involving instrument or component f ailures would be greatly enhanced if the individual were to assume the watchstander's position with all systems operating normally and have the failure subsequently occur. The facility stated that failures of this type were routinely performed and evaluated during the simulator scenario portion of the exam. The NRC stated that the JPM setting would allow for a greater number of individuals to be evaluated in one particular area of performance and also would be more realistic for these types of failures. The facility agreed with this approach and stated that JPMs of this type would be developed for future evaluations, Simulator performance by the Unit 1 candidates was, for the most part, very good.

Communications was also good, including the use of repeatbacks. The examiners noted that crew briefings were routinely performed by the SROs. Control board awareness by the operators was evidnt throughout each of the three scenarios.

For the administrative segment of the operating portion of the exarriination, administrative job performance measures (JPMs) were used in a number of instances in lieu of administrative topic questions. The examiners determined that candidate performance was good as evaluated in this area,

c. Conclusions The candidates performed well on both the written and operating examinations, and thus were issued licenses. The candidates appeared to be well prepared for the examinations, indicating that the facility thoroughly evaluated the knowledge and ability of each candidate in an effort to determine which individuals were ready to sit for an NRC exam. Crew communications, control board awareness, and crew briefings were very good. As noted in the past, the BVPS training department, again, did an excellent job in adhering to the examiner standards and in developing the examination materials needed to administer the examinations.

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 I special focused review that compares plant practices, procedures and /or i

3 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.

1 V. Manaaement Meetinas X1 Exh Meeting Summary On December 18,1997, the NRC examiners discussed their observations from the examinations with Beaver Valley Unit 1 operations and training management representatives. The examiners discussed generic candidate performance, including communications and briefings, both of which were seen as being very good.-

The examiners also 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. Beaver Valley personnel contacted and/or present at the exit meeting included the following partial listing.

BEAVER VALLEY K. Beatty, General Manager, Nuclear Support R. Brooks, Sr. Nuclear Operations Instructor T. Burns, Director, Operator Training S. C-Jain, Vice President, Nuclear Services R. Hart, Senior Licensing Supervisor T. Kuhar, Licensed Operator Training Supervisor B. Tuite, General Manager, Nuclear Operations NBC P. Bissett, Senior Operations Engineer, Chief Examiner T. Kenny, Senior Operations Engineer D. Kern, Senior Resident inspector, Beaver Valley Units 1 & 2 Attachments:

1. Beaver Valley Unit 1 SRO Written Examination w/ Answer Key

2. Simulation Facility Report

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l Attachment 1 j BV 1 SRO WRITTEN EXAM W/ ANSWER KEY i

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RTL #AS.620.H DUQUESNE LIGHT COMPANY Volume 3 Nuclear Power Division Procedure 5-5 Training Administrative Manual Figure 5-5.1 Revision 12 Page 1 of 1 WRITTEN EXAMINATION COVER SHEET PROGRAM: Licensed Operator Trainino CLASS NUMBER: 1 LOT 3

SUBJECT:

SRO initial NRC Exam By this signature, I state that all of the work done on this examination is my own. I have neither given nor received aid.

SIGNATURE DATE 12/15/97 NAME DLC EMP, #

(Please Print)

COMPANY (if other than DLC)

POSSIBLE POINTS 100 SCORE Instructor initials PREPARED BY A Beckert TRAINING DIRECTOR / SUPERVISOR SIGNATURE [Id<# APPROVAL / Y7 L/ d /z-r.ffy

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-1 RCS cooldown rate and total flow while on RHR is controlled by...

A. manually adjusting RHR flow through the heat exchanger and automatically controlling CCR flow through the heat exchanger.

B. automatically controlling RHR flow through the heat exchanger and manually adjusting CCR flow through the heat exchanger.

C. automatically controlling RHR flow through the heat exchanger and manually adjusting total RHR flow through the system.

D. manually adjusting RHR flow through the heat exchanger and automatically controlling total RHR flow through the system.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12 97-2 i

An Operating Manual precaution in OM 24 states that two (2) Condensate pumps must be running before starting the second Feedwater pump. The reason for this is to...

A. ensure adequate flow capability to prevent a low suction pressure Feed pump trip.

B ensure adequate flow cape.bility to prevent robbing the Steam Generators when the Feed pump recirculation valves open.

C. ensure adequate flow capability to prevent run-out on the Heater Drains pumps.

D. ensure adequate flow to prevent water hammer in the first point heaters.

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. . NRC Exam: 1 LOT 3, Rev 1 Question 12-97-3 l Given the following:

. Reactor trip and Safety injection due to a Small Break LOCA.

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. All ESF equipment operated as designed.

. Various PAB and Safeguards Area Radiation Monitors are in alarm.

The procedure that should be entered based upon the above conditions is:

A. ECA-1.1," Loss of Emergency Coolant Recirculation."

B. FR-Z.2," Response to Containment Flooding." l C. ECA-1.2,"LOCA Outside Containment."

i D. FR-Z.1," Response to High Containment Pressure."

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BVPS-1

' NRC Exam: ILOT3, Rev 1 .

Question 12-97-4 Which of the following events would result in a Containment Particulate and Gas Radiation Monitor [RM-lRM-215 A(B)] alarm?

A. RCS leak at the incore Seal Table.

B. PZR Safety Valve seat leakage.

C. Steam Generator Tube Rupture.

D. RCP #1 seal failure.

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1 BVPS-1 NRC Exam: lLOT3, Rev 1 Question 12-97-5 Given the following plant conditions: l i

• Unit i is in Mode 2.

• Reactor power is at 10 amps in the I.R. (P-6 activated).

• Preparing to block the Source Range detectors per the start-up procedure.

• The detector for N31 Source Range fails low.

Which action is required by procedure?

A. Enter E-0," Reactor Trip or Safety injection,"in response to the automatic Reactor trip.

B. Hold power at 10 amps, until repairs are made.

C. Insert Control Rods until power is less than P-6.

D. Place the N31 Level Trip Switch in the Bypass position.

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BVPS-1 NRC Exam: 1 LOT 3, Rev i -

_ Question 12-97 Given the following conditions:

• A Reactor trip and Safety Injection have occurred due to a SGTR in the 'C' S/G.

• E-3, " Steam Generator Tube Rupture,"is in effect with a cooldown staned at maximum rate.

• The highest S/G Pressure is 900 psig.

. RCS pressure is 1000 psig and dropping.

• High Head Safety injection flow is approximately 300 gpm.

Which of the following describes what should be done with the Reactor Coolant Pumps?

A. RCP's should be tripped because RCP trip criteria is currently met.

B. RCP's should be tripped because RCP trip criteria applies after an operator initiated RCS

' depressurization is commenced.

C. RCP's should not be tripped because RCP trip criteria does not apply once an operator l initiated RCS cooldowm is commenced.

D. RCP's should not be tripped because RCP trip criteria does not apply until the operator initiated RCS cooldown is completed.

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NRC Exam: lLOT3, Rev 1 Question 12-97-7 The following conditions exist:

• A Reactor trip from 100% power has occurred.

• The Turbine has failed to trip both automatically and manually.

Which of the following actions is/are required, by procedure, to be performed next?

A. Runback the Turbine using the " Turbine Manual" and " Fast Down" pushbuttons.

B. Close the Main Steam Trip and Bypass Trip Valves.

C, Close all Reheat Stop and Interceptor Valves.

D. Place the running EHC Pump in Pull-to-Lock.

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NRC Exam: lLOT3, Rev 1 Question 12-97-8 Given the following:

• Reactor power is 90%.

• Tavg is stable at 574 F on all three loops.

• PZR level is 52% and stable.

• RCS pressure is stable at 2235 psig.

• Containment pressure and humidity are rising.

. Net Charging flow is 0 gpm.

• Steam flow on all three loops is 3.78E6 lbm/hr.

• On the "C" S/G:

. Pressure is stable.

. NR level is dropping.

. Feed flow is pegged high.

Which of the following events is in progress?

A. There is a Charging line leak inside Containment.

B. "C" S/G controlling Feed flow channel instrument line is ruptured.

C. There is a Feed line break inside Containment.

D. There is a Small Break RCS LOCA inside Containment.

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BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-9 Given the following conditions:

. The Reactor has been shutdown for 2 days.

• RCS temperature is 150 degrees F.

• RCS pressure is Atmospheric.

• PZR is at normal level for shutdown cooling.

Assume RHR cooling is lost. Which of the following describes the time available until core boiling occurs?

A. Less than 10 minutes.

B.1I to 20 minutes.

C. 21 to 30 minutes.

D. 31 to 40 minutes.

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BVPS-1 NRC Exam: lLOT3, Rev 1 -

Question 12-97-10 The following conditions exist:

• PZR Spray valves [PCV-IRC-455A & B] are closed.

PZR PORV [PCV-1RC-455C] is closed.

• PZR PORV [PCV-1RC-456] is cycling.

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• RCS pressure is approximately 2000 psig.

Which of the following Pressurizer pressure transmitter failures has occurred?

A. [PT-1RC-445] failed low.

B. [PT-1RC-445] failed high.

C. [PT-1RC-444] failed low.

1 D. [PT-1RC-444] failed high.

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BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-11 Given the following:

Unit 1 is operating at 100% power with all systems in their normal system arrangement. .

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. Number i DC Bus voltage is 0 VDC.

. A Reactor trip occurs.

. The Main Feedwater pumps are secured per the Alarm Response Procedure.

Which of the following Auxiliary Feedwater pump (s) will auto start?

A. MDAFW pump 3 A and the TDAFW pump.

B. MDAFW pump 3B and the TDAFW pump. -

C. TDAFW pump only.

D. MDAFW pump 3B only.

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NRC Exam: 1 LOT 3, Rev 1 I Question 12-97-12 The following plant conditions exist:

• Unit I is operating at 100% power.

• Pressurizer Spray valve (PCV-1RC-455B] is stuck OPEN.

• All effons to close the Spray valve have failed.

• RCS pressure is dropping rapidly.

Which of the following actions should be taken?

l A. Trip the 'A' Reactor Coolant Pump, then trip the Reactor.

B. Trip the 'C' Reactor Coolant Pump, then trip the Reactor.

C. Trip the Reactor, then trip the 'A' Reactor Coolant Pump.

D. Trip the Reactor, then trip the 'C' Reactor Coolant Pump.

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BVPS.1 NRC Exam: 1 LOT 3 Rev i Question 12-97-13 Which of the following would indicate that the PRT rupture disk had blown following a Pressurizer PORV failing open?

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PRT Pressure PRT Temperature PORV Tailpipe Temperature A. Drops Remains the same Drops B. Drops Drops Remains the same C. Drops Drops Drops D. Remains the same Drops Drops I

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Question 12-97-14 Given the following:

. A Turbine / Generator trip has caused a Reactor trip.

. The operators are in ES-0.1," Reactor Trip Response," at step 4, " Check RCS Temperature Stable at or Trending to 547F."

. RCS pressure is 1810 psig and slowly dropping.

. Pressurizer level is 22% and stable.

. Core exit T/Cs are 575*F and slowly rising.

. Containment pressure is 19 psia and slowly rising

. All S/G NR levels are 20% and slowly rising.

Which of the following actions should be taken?

A. Dump steam to the Condenser and proceed to step 5 of ES-0.1.

B. Initiate Si and go to E-0," Reactor Trip or Safety injection," step 1.

C. Transition to FR-H.1," Response to Loss of Secondary Heat Sink."

D. Transition to FR-Z.1," Response to High Containment Pressure."

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BVPS-1

' NRC Exam: ILOT3, Rev 1

' Question 12-97-15 Given the following:

j- . The Unit is in Mode 1 at 50% power proceeding to full power following a refueling outage.

. All control systems are in automatic.

l . All Pressurizer Heaters are turned on to allow for boron mixing as power is raised.

. ' A loss of Containment Instrument Air occurs.

t I-Assuming no Operator actions are taken, which of the following will occur?

1. Reactor will trip on high Pressurizer level.

I 2. Letdown willisolate. ,

f 3. Pressurizer pressure will rise.

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4. Charging will isolate.

A.1,2, and 3 only.

B. I and 2 only.

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L l D. 1,2,3, and 4.

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BVPS-1 NRC Exam: ! LOT 3, Rev 1

( Question 12 97-16 The plant is in Mode I with all systems in their normal system arrangement. Containment pressure transmitter [PT-LM-101C] failed low and actions were completed per lOM-1.4.lF.

" Instrument Failure." Subsequently, Containment pressure transmitter [PT-LM-101B] fails high.

Which of the following will occur?

1. SIS i

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! 3. CIA 4.CIB 1

A. 1,2,3, and 4.

1 B. 1,2, and 3 only.

C. I and 3 only.

D. I and 2 only.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-17 The plant is preparing for a start-up. The Shutdown Bank rods are fully withdrawn, and all Control Banks are fully inserted. In preparation for the start-up the Control Room operators are performing an Operations Surveillance Test (OST) on Source Range detector N31. Which of the following is correct regarding the performance of this test?

A. The Shutdown Banks must be inserted and the Reactor trip breakers opened. The OST will generate a Reactor trip signal.

B. The Shutdown Banks must be insened and the Reactor trip breakers opened. This will provide a lower baseline Source Range count to allow all setpoints to be tested.

C. The Shutdown Banks can be left withdrawn. No Reactor trip signal is generated during the perfonnance of the OST.

D. The Shutdown Banks can be left withdrawn. Placing the Level Trip switch to Bypass will prevent the OST from causing a Reactor trip.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 -

Question 12-97 Given the following:

• A spurious Reactor trip and Safety injection have occurred during a Reactor start-up.

• Tavg is stable at 547'F.

• PZR pressure is stable at 2235 psig.

• All Steam Generator water levels and pressures remained stable at 33% and 1005 psig respectively.

Before resetting Si, which of the following pumps can provide feedwater to the S/Gs?

1. Main Feedwater pumps

2. Auxiliary Feedwater pumps

3. Dedicated AFW pump 1

A. I and 2.

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l C. 2 and 3.

D. 1, 2, and 3.

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• NRC Exam: ILOT3, Rev 1 -

Question 12 97-19 When determining subcooling requirements while progressing through the EOPs, the value for subcooling that is read on the inadequate Core Cooling Monitor (ICCM) is the difference

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between the saturation temperature for the...

A. Pressurizer pressure and the highest loop Hot Leg temperature.

B. wide range RCS pressure and the average loop Hot Leg temperature.

C. Pressurizer pressure and the average of the five hottest Core exit T/Cs.

D, wide range RCS pressure and the average of the five hottest Core exit T/Cs.

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BVPS-1 L NRC Exam

lLOT3, Rev 1 .

l Question 12-97-20

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l The plant has entered Mode 5 for a refueling outage. Containment vacuum has been broken.

l. Shortly after placing Containment Purge and Exhaust into service, Containment Purge Exhaust Radiation Monitor [RM-IVS-104A] fails high. This will result in...

1. Containment Purge and Exhaust dampers closing.

2. Auxiliary Building exhaust fans tripping.

3. Main Filter Bank inlet dampers opening.

4. Containment Purge and Exhaust fans tripping.

A. I and 2 only.

B. 1,2, and 3.

C. 1,3, and 4.

D. 2 and 3 only.

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BVPS-1 NRC Exam: ILOT3, Rev i Question 12-97 21 The plant has entered a mini-outage to repair Condenser tube leaks. The Reactor is stable in i Mode 3. A decision is made to completely isolate the Condenser Circulating Water system and repair all four Condenser sections simultaneously. This evolution should NOT be allowed to occur because this will result in a loss of...

A. a flowpath for the Reactor Plant River Water system for Appendix R requirements.

B. normal Feedwater and would require Auxiliary Feedwater to maintain Steam Generator levels which is a violation of Technical Specifications.

C. the ultimate heat sink requirement for Technical Specifications.

D. Condenser Steam Dumps and there will be no way to get the plant to Mode 5 if needed. 1 21 ,

NRC Exam: 1 LOT 3, Rev 1 Question 12-97-22 When fighting an electrical fire using foam, which of the following are precautions to be exercised by the Fire Brigade members?

1. Anticipate and avoid the run off from the electrical equipment being sprayed.

2. Always wear rubber boots for electrical insulation.

3. Always use an MSA 401 SCBA when using foam.

4. Maintain a minimum distance of 15 feet from the electrical equipment being sprayed.

A. I and 3 only.

B. 2 and 4 only.

C. I and 4 only.

D. 2,3, and 4 only.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-23 Which of the following actions are directed in AOP-1.6.5," Shutdown LOCA"?

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1. Manually initiate Safety injection (both pushbuttons) to recover Pressurizer level.

i 2. Align HHS1 pumps through the Hot Leg Injection flowpath.

3. Manually depressurize and cooldown the RCS to place RHR into servics.

4. Manually start HHS1 pumps to recover Pressurizer level.

A. 1,2, and 3.

B. 2 and 4 only.

C. 3, and 4 only.

D. 2,3, and 4.

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BVPS-1 NRC Exam: ILOT3, Rev 1 .

Question 12-97-24 What is the expected procedure flow path for a Small Break LOCA that is too small for two HHSI pumps, but too large for one HHSI pump, i.e., a " smart break"?

1. lOM-11.4.M," Recovery from Safety injection."

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2. lOM-10.4.A, " Residual Heat Removal System Startup."

3. E-0, " Reactor Trip or Safety Injection."

- 4. ES-1.2, " Post LOCA Cooldown and Depressurization."

5. ES-1.1, "Si Termination."

6. E-1, " Loss of Reactor or Secondary Coolant."

A. 6,5, then 2.

B. 3, 6, 5, then 4.

C. 3, 6, 5, 6, then 1.

D. 3,6, then 1.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1

. Question 12-97-25 Which of the following actions does ECA-1.1," Loss of Emergency Coolant Recirculation" direct?

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l. Provide guidance on aligning the HHSI pump suction directly to the Containment sump.

2. Terminate Cold Leg Recirculation and restore Charging and Letdown.

3. Cooldown and depressurize the RCS to allow RHR to be put into service.

.4. Provide methods to make-up to the RWST.

A. 1,2, and 3.

B. 2, and 4 only.

C. 3, and 4 only. .

D. 2,3, and 4.

i 25 l

1

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 -

Question 12-97-26 The unit is at 100% power with all systems in their NS A configurations for the current power level. The RO inadvertently changes the Pressurizer Pressure Master Controller setpoint to 2185 psig. Assume a step change in the setpoint and assume the controller remains in automatic.

Which of the following is the immediate automatic response of the system?

A. Spray valves open; Variable heaters go to minimum output.

B. PORV 455C opens; Spray valves open; Variable heater go to maximum output.

C. Spray valves open; Variable heaters go to maximum output.

D. Spray valves close; Variable heaters go to minimum.

26

BVPS-1 NRC Exam: 1 LOT 3, Rev 1

[ Question 12-97-27 Given the following: )

i i e Control Rod D12, a control bank "C" Group 1 rod, has fallen into the core due to an equipment failure.

l e The equipment failure has been corrected and all retests are completed satisfactorily.

l e The dropped rod recovery is in progress per AOP-1.1.5," Dropped RCCA." ,

l J

! e All applicable switches are in their correct position for the rod recovery. l I

i e AOP-1.1.5, step 17.a directs the operator to " Anticipate a Rod Control System Urgent l 1

l Failure Alarm."

The Rod Control System Urgent Alarm is caused by a:

1 l

l A. Logic Cabinet failure and all rod motion will be inhibited.

1 B. Logic Cabinet failure and only those rods aligned to Power Cabinet 1 AC will move.

C. Power Cabinet l AC failure and only those rods aligned to Power Cabinet 2AC will move.

D. Power Cabinet 2AC failure and only those rods aligned to Power Cabinet l AC will move.

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27

l 1

BVPS-1 -

NRC Exam: ILOT3, Rev 1 l Question 12-97-28 l Technical Specifications restrict the quantity of radioactive liquids in [lLW-TK-7B] to less than 10 curies (excluding tritium and dissolved or entrained Noble gases). The basis for this limit is to...

l A. prevent over-exposure to personnel who must work near or pass by the tank (Transient Pathway Radiation Levels).

1 B. prevent exceeding 10CFR20 Appendix B limits at the nearest surface water supply in the event of an accidental release.

C. maintain activity low enough so that the tank may be discharged with minimum design dilution flow.

D. prevent exceeding 10CFR100 limits for child thyroid dose at the site boundary in the event of an accidental release.

l l

l 28

. l BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-29

-Which of the following does the NSS sign for when discharging a Gaseous Waste Decay Tank?

1. Verification of the proper tank and approval for the discharge.

2. The appropriate Rad Monitor alarms have been adjusted.

3. The opposite units NSS has been informed of the discharge.

4. Only one batch discharge is being done at a time.

A.1, and 2 only.

( B. 1,2, and 4.

C. 3 and 4 only. )

D. 2,3, and 4.

1 i

29

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 .

Question 12-97-30 Which of the following must be met in order to terminate Safety injection following a Small Break LOCA?

1. Adequate subcooling.

2. At least one Reactor Coolant Pump in operation.

3. Adequate secondary heat sink.

4. RCS pressure stable or rising.

5. RCS Hot Leg temperatures stable or rising.

6. Adequate Pressurizer level.

7. Charging and Letdown available.

A.1, 2,4, and 6.

B. 2,3, 5, and 7.

C.1,3, 5, and 7.

D.1, 3,4, and 6.

30

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97 Given the following:

A Reactor trip from 100% power has occurred.

The 'B' RTB has failed to open.

i I

The Condenser Steam Dump (1) controller will maintain Tavg at l

(2) .

(1) (2)

A. Load Rejection 549'F B. Load Rejection 552 F 1 ~

C. Reactor Trip 547'F D. Reactor Trip 549*F l

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31

r- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

BVPS-1 NRC Exam: ILOT3, Rev 1 ~

, Question 12-97-32 The plant is in Mode I with Reactor Plant River Water Pumps [1WR-P-1 A] and [1WR-P-1B] in service. In order to replace [1WR-P-1 A] with [1WR-P-lC], which sequence of operations must be performed?

1. Rack on [1WR-P-lC).

2. Rack off[ LWR-P-1 A).

3. Start [ LWR-P-lC).

4. Stop [1WR-P-1 A].

l A. 2, 3, 4,1.

B. 2, 4, 3,1.

C. 4,1, 3, 2.

l D. 4, 2,1, 3.

l l

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r 32

. BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-33 Which of the following conditions will prevent the Emergency Diesel Generator output breaker from closing to re-energize the 1 AE bus following a transient?

A. l AE Emergency Bus feeder breaker, [ACB-1 A10] trips on overcurrent.

B. l AE Emergency Bus reverse phase PT has a blown fuse.

C. 1 A Normal 4KV Bus feeder breaker, [ACB-41C) trips on overcurrent.

D. l A Normal 4KV Bus feeder breaker, [ACB-41C) trips due to the Main Generator tripping on overcurrent.

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33

BVPS-1 )

NRC Exam: ILOT3, Rev 1 . j Question 12-97-34 f

In ECA-0.0, " Loss of Emergency AC Power," prior to restoring power to an emergency bus, all major loads are placed into Pull-to-Lock with the exception of the Reactor Plant River Water

( Pump. This pump is left in automatic to...

A. provide a load for the Diesel Generator to prevent it from tripping on overspeed when started in the emergency mode.

B. provide cooling for the Diesel Generator to prevent overheating and possible failure of the

! Diesel.

C. provide cooling for the Charging pumps oil coolers so that make-up to the RCS can begin immediately to replace the RCS lost through the RCP seals.

j D. provide cooling to the Control Room emergency back-up cooling coils to maintain Control Room habitability.

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34 l

BVPS-1 NRC Exam: ILOT3, Rev 1

, . Question 12-97-35 The unit is at 100% power when the following Radiation Monitors go into alarm:

e

[RM-1MS-102A] N-16 SG Leak Monitor.

~

[RM-1BD-100] S/G Blowdown Effluent Monitor.

i

• [RM-ISV-100] Condenser Air Ejector Vent.

Within 15 minutes, analysis reveals a 0.12 gpm tube leak with 0 GPD/HR rate of rise on the 'A' S/G. Which of the following actions should be taken?

A. Enter TS 3.4.6.2 action for RCS leakage and restore the leak to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in Hot Shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

B. Enter AOP-1.6.4, "S/G Tube Leakage," and commence an emergency shutdown to be in Mode 3 as soon as possible.  !

l C. Enter AOP-1.6.4,"S/G Tube Leakage," and place the plant in Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

]

D. Manually trip the Reactor and initiate Safety injection and enter E-0.

1 35

BVPS-1 NRC Exam: ILOT3, Rev 1 -

Question 12-97-36 With the unit at 100% power, a HIGH-HIGH alarm is received on Condenser Air Ejector Vent Radiation Monitor [RM-ISV-100]. This HIGH-HIGH alarm will result in the Condenser Air Ejector exhaust...

A. being isolated.

B. diverting through the Main Filter Banks.

C. diverting to Containment.

D. diverting to the Gaseous Waste Surge Tank.

36

I

< s VPS-1 \

\

NRCExam: .

Question12-97-37ILOT3, Rev 1 There has been injection"has en bea Reactor trip caused b a o done t step 6, entered.

be yaloss A transition foff-sitep ow the step ewher th cause bothtoemergencECA-0.0, "L er. E-0, " Reactor T i y4KV o

1. 2 Diesel ey are to determine busses w oss fEmergency4K V rp or Safety which following is theGeneratoris bus s startedl e ACPow houldbeerede-energized er "w Th as appropriate response?ocally us is selectedas and the.the IDF crewhas reachedb A. Continue automaticallyloadedcross-tie e bus A . Which ofthe on in toEC B. Skip ahead th -0.0 until b \

C. Transition to E 0e step which determioth emergency bus es D. Concurrently nesorthe

, " Reactor Trip a

Safppropriate rare restored.

ecov p

perform E-0 andetyinjection."

ECA ery rocedure.

-0.0 until the eis cross ti

\

completed for Bus I AE.

6

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.8 (O

e6

BVPS-1

• NRC Exam: 1 LOT 3, Rev 1 Question 12-97-38 Which of the following statements is correct if the Power Range channels have been adjusted based on a calculated calorimetric?

A. If the Blowdown flow is ignored in calculating the calorimetric, then actual Reactor power would be lower than indicated Reactor power.

B. If the Blowdown flow is ignored in calculating the calorimetric, then actual Reactor power would be higher than indicated Reactor power.

C. If the Feedwater temperature used in calculating the calorimetric had been 10 degrees lower than actual Feedwater temperature, then actual Reactor power would be higher than indicated Reactor power.

D. If the Feedwater temperature used in calculating the calorimetric had been 10 degrees higher than actual Feedwater temperature, then actual Reactor power would be lower than indicated Reactor power.

b 38

l BVPS-1 NRC Exam: ILOT3, Rev i Question 12-97-39 Given the following: .

. The plant was in Mode 1 when a loss of offsite power occurred.

. Both Emergency Diesel Generators and the ERF D/G have started and loaded their respective busses.

Control of air-operated valves outside of Containment,is.'..

A. not possible until offsite power is restored.

B. not possible until the Diesel Air Compressor is started.

C. Possible since the Station Air Compressors will auto load onto the Emergency Diesel Generators.

D. Possible since the Station Air Compressors will auto load onto the ERF Diesel Generator.

l l

e 39 L

BVPS-1 NRC Exam: 1 LOT 3, Rev i '

Question 12-97-40 l

I Which of the following is an indication that natural circulation exists in the RCS?"

RCS RCS indicated S/G RCS Cold CNMT  !

Pressure Subcooling Pressure Leg Temp. Pressure

' i on ICCM A. 600 psig 40 F 400 psig 400 F 10 psia l

B. I100 psig 30 F 500 psig 470 F 7 psig C. 1500 psig 35 F 600 psig 490 F 10 psia l D. 2000 psig 40 F 700 psig 550 F 7 psig l

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40 L________ _ _ _ _ _ _ _ . _ _ _ _ _ _ - - _ - . . -

BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-41 The RCP Thermal Barrier Component Cooling Outlet Trip Valve, [TV-lCC-107B],

l fails closed on the loss of (1) Instrument Air and will auto close on high CCR (2)  ?

l l

(1) (2) l A. Containment flow

! B. Containment pressure C. Station flow D. Station pressure l

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-l 41

q BVPS-1 -

)

NRC Exam: 1 LOT 3, Rev i Question 12-97-42' 1 The location of the 18 inch escape manway associated with the CNMT Personnel Airlock Doors is...

A. on the inner and outer doors with no associated interlocks.

B. only on the inner door with no associated interlocks.

C. on the inner and outer doors, interlocked to prevent any inner and outer door from being opened simultaneously.

D. on the inner door, interlocked to prevent any inner door and the outer door from being.

opened simultaneously.

l 42

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BVPS-1 NRC Exam: lLOT3, Rev 1 Question 12-97-43 Given the following conditions:

e Reactor power level is 97% and dropping.

. Control Rods are in automatic and stepping in.

• Tavg is 574 F and dropping.

• Pressurizer pressure is 2225 psig and dropping.

• Pressurizer level is 53% and dropping.

. MW Recorder is 820 MW and stable.

Which of the following actions should be :aken first by procedure?

A. Check the In-Hold-Out lever is in the Hold position.

B. Enter AOP-1.51.1," Emergency Shutdown."

C. Place the Control Rod selector switch in Manual.

D. Reduce Turbine load to stabilize primary plant parameters.

1 l

d 43 i

BVPS 1 NRC Exam: ILOT3, Rev 1 .

Question 12-97-44 The following conditions exist:

. Loops 1 and 3 Tavg indicates 576 F.

. Loop 2 Tavg indicates offscale high.

• Loops 1 and 3 Delta T indicates 100%.

• Loop 2 Delta T indicates 0%.

Which of the following is the cause of these indications?

A. Loop 2 Tcold failed low.

B. Loop 2 Tcold failed high.

C. Loop 2 Thot failed low.

D. Loop 2 Thot failed high.

l 1

1 44

._ _____ ___-___.-______ _ f

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-45 The following events have occurred:

. The unit is at 100% power with all systems in their NSA configurations for the current power level when a Reactor trip and Safety injection occur.

1

. 1 A S!G pressure is dropping rapidly.

e l A steam line indicates 2.5E6 lbm/hr steam flow.

. Containment pressure is 8.5 psig and rising.

• All ESF actuations occur as designed.

Assuming no operator actions taken, a possible consequence of this accident is..

A. an acidic Containment spray solution since the steam will dilute the spray ring water.

B. insufficient Containment sump levels to support Recirculation Spray pump operation.

C. a postulated flaw in the Reactor Vessel wall propagating if RCS pressure rises.

D. a loss of the steam driven AFW pump due to all three steam lines depressurizing.

45

l BVPS-1 L NRC Exam: 1 LOT 3, Rev 1 Question 12-97-46 Given the following:

. The unit is at 100% power with all systems in their NSA configurations for the current power level when a Reactor Trip and Sl occur.

. The operators transition to FR-H.1," Response to Loss of Secondary Heat Sink

(: due to low S/G NR levels and low AFW flow.

. RCS pressure is less than S/G pressure and FR-H.1 directs a transition to E-1,

" Loss of Reactor or Secondary Coolant.

Based on this information, select the statement that correctly summarizes plant conditions:

A. Large Break LOCA in progress; secondary heat sink required.

i B. Large Break LOCA in progress; secondary heat sink not required.

C. Small Break LOCA in progress; secondary heat sink required.

D. Small Break LOCA in progress; secondary heat sink not required.

46 i

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 I

i l

Question 12-97-47 l E-1," Loss of Reactor or Secondary Coolant," step 24 directs the operators to isolate the Accumulators if at least two RCS Hot Leg temperatures are less than 390 F.

The basis for this RCS temperature ensures...

A. Adequate core cooling is established prior to isolating the Accumulators as a water l injection source.

B. Sufficient Accumulator water and nitrogen volumes are injected into the RCS prior to isolating the Accumulators.

C. Saturation pressure of the RCS is less than the Accumulator nitrogen pressure when the Accumulator water volume is fully discharged.

D. That the injected Accumulator nitrogen volume has expanded sufficiently to maintain RCS saturation temperature less than the UFSAR design basis.

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47 l l

BVPS 1 NRC Exam: ILOT3, Rev 1 .

Question 12-97-48 Given the following conditions:

• The 'A' train of RHR is in service. i e All systems are in their NSA configurations for the current mode of operation.

. RCS pressure is 200 psig.

. Which of the following would indicate a tube leak in the "A" RHR heat exclianger?

A. CCR Surge Tank level rising.

B. VCT level rising.

C. RHR return flow to the RCS dropping.

D. Pressurizer level rising.

48

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l BVPS-1 1.

NRC Exam: lLOT3, Rev 1 Question 12-97-49 Which of the following is indicative of, or the result of, a dropped rod with an initial power level  !

l of 100%?

1. Rod Insenion Limit drops. ]

2. "NIS Power Range Comparator Deviation" alarm.

3. Tavg drops rapidly.

4. Axial Flux Difference becomes more negative.

! A.1,2, and 3.

B.1,2, and 4.

C. I and 3 only. '

l D. 2 and 3 only.

I

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49

BVPS-1 -

NRC Exam: 1 LOT 3, Rev 1 Question 12-97-50 When responding to a Degraded Core Cooling condition in FR-C.2, the operator is directed to l

l " Verify SI Valve Alignment" with the SI system in the Cold Leg injection mode. Which of the following valves should be closed for the current plant conditions?

A. RWST Discharge to Charging Pumps Suction Valve [MOV-lCH-115B1 B. Regen HX/Chg Header Inlet CNMT Isolation Valve [MOV-lCH-289].

C. AFW Turbine Steam Supply B Train Trip Valve (TV-lMS-105B].

D. BIT Outlet isolation Valve [MOV-ISI-867D].

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1 50

BVPS-1 NRC Exam: lLOT3, Rev 1 Question 12-97-51 Given the following conditions:

. The unit was at 100% power with all systems in their NSA configurations for the current power level.

3 . A Main Steam Line break has occurred inside Containment.

. The crew is in FR-P.1, " Response to Imminent Pressurized Thermal Shock," which directs a " soak of the RCS."

)

Which of the following evolutions can be performed during this soak?

A. Warm up the RHR system and commence a cooldown of the RCS.

B. Raise the faulted S/G water level to 50% and secure the TDAFW pump.

C. Place PZR Auxiliary Spray in service to control RCS pressure.

D. Energize the PZR heaters to raise the saturation temperature of the PZR.

l 51

BVPS-1 .

NRC Exam: ILOT3. Rev 1 Question 12-97-52 Given the following conditions:

• The unit is at 92% power with all systems in their NSA configurations for the current power level.

• The Reactor did not trip, and the crew is responding with FR-S.1, " Response to Nuclear Power Generation /ATWS."

• After completing the immediate Manual Actions, the Reactor trip breakers are opened locally and the Reactor is shutdown.

• No other malfunctions or actuations have occurred.

Based on the above information, the crew should...

A. perform the first 15 steps of E-0 while continuing in FR-S.I.

B. transition to E-0, perform the first 15 steps, and then return to FR-S.I.

C. return to procedure and step in effect.

D. continue with procedure and step in effect.

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1 52 L___--__________

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12 97-53 The following conditions exist:

e Unit 1 is in Hot Standby following an unscheduled maintenance outage.

• The Plant Operator is performing valve stroke OST 1.47.3B.

. [TV-ISS-105A2), RCL Hot Leg Samples Outside Cnmt Isol valve's closing time is found to be 3.2 seconds.

• The acceptable range listed in the OST is s; 2.0 seconds, the ASME limiting stroke time is 2.0 seconds, and the Technical Specification limiting time is 21.0 seconds.

Based on this information, the OST should be marked:

A. unsat, and penetration declared inoperable.

B. unsat, and the valve declared inoperable.

C. sat, because the average stroke time was less than Technical Specification limit, but frequency of testing must be doubled.

D. sat, because the second stroke time was within 25% of the ASME limiting stroke time which is acceptable for valves with stroke times less than 1 minute.

53 l

1 BVPS-1 '

NRC Exam: ILOT3, Rev 1 Question 12-97-54 i The Technical Specification limit for RCS activity ensures that the dose at the site boundary will not exceed a small fraction of the Part 100 limits in the event that a occurs.

A. Steam line rupture induced tube leak.

B. Small Break LOCA with a stuck open Atmospheric Steam Dump Valve.

C. Rod Ejection accident.

D. Locked RCP rotor accident.

54

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-55 Given the following: )

. A Steam Generator Tube Rupture has occurred in the IB S/G.

. The Crew is in E-3, " Steam Generator Tube Rupture," preparing to cooldown the RCS.

. 1B narrow range level at 74% and rising.

. PZR pressure is 1900 psig.

. Tavg is 547 F. l

. Main Condenser vacuum is 13" Hg Absolute and stable.

. 1B and 1C Cire Water Pumps are running.

Which of the following actions is/are necessary to commence cooldown in accordance with 1

E-37 A. Take the Steam Dumps to Steam Pressure mode and manually open the dumps to comthence the cooldown.

B. Take the Steam Dumps to Steam Pressure mode, take both Steam Dump Control Selector Switches momentarily to the Defeat Tavg Interlock position, and then manually open the dumps to commence the cooldown.

C. Commence the cooldown using Steam Generator Atmospheric Dumps

[PCV-lMS-101 A, B, & C].

D. Commence the cooldown using Steam Generator Atmospheric Dumps

[PCV-lMS-101 A & C).

l \

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BVPS-1 .

l NRC Exam: lLOT3, Rev 1 j Question 12-97-56' BVPS-1 Steam Generator Atmospheric Steam Dump valves [PVC-1MS-101 A,B,C)

Auto / Manual stations are normally set at 1035 psig. Assuming a malfunction of the Auto / Manual station, at what steam pressure will the Atmospheric Steam Dumps open with no operator action?

l l

A.1050 psig.

l B. 1060 psig.

C. 1070 psig.

D. 1080 psig.

l e

1 56

1 BVPS-1 NRC Exam: ILOT3, Rev 1 l Question 12-97 57 Given the following conditions:

• Reactor tripped due to a loss of off-site power.

! e RCS pressure is 700 psig.

! . Tcold is 370*F.

l

! e Core exit T/Cs are 506*F.

l l

. Pressurizer levelis 68% and rising.

. Operators are performing ES-0.2, " Natural Circulation Cooldown."

Which of the following describes the cause of the abnormally high Pressurizer level?

l l A. Pressurizer level instruments are inaccurate due to the loss of Containment cooling.

B. Letdown has not been placed into service due to the loss of offsite power.

C. RCS pressure has reached the injection point for the Accumulators.

I D. RCS temperature and pressure are at the point where voiding is occuring in the Reactor l Vessel head.

1 I

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BVPS-1 .

NRC Exam: 1 LOT 3, Rev 1

, Question 12-97-58 i:

The following conditions exist:

l

• The Unit is at 92% power with all systems in their NSA configurations for the

. current power level.

• Pressurizer level control channel selector is in the 461/460 position.

• Pressurizer level channels indicate as follows:

Channel 459 is 54%.

Channel 460 is 56%.

Channel 461 is 0%.

Channel 462 is 40%.  !

l Which of the following describes the plant response with no operator intervention?

A. Sackup heaters energize.

B. Charging flow drops to minimum. ,

i C. Reactor trips on high PZR level.

D. Reactor trips on high PZR pressure.

58-i l

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BVPS-1

. NRC Exam: ILOT3, Rev i Question 12-97-59 Rods are being withdrawn in manual during a Reactor start-up, with all systems operable. For the Control Banks, which of the following describes the status of the Rod Bottom Lights at the moment A4-126 " ROD BOTTOM ROD DROP" annunciator clears? -

A. Banks A, B, C, & D - OFF.

B. Banks A, B, C. & D - ON.

C. Banks A, B, & C - OFF; Bank D - ON.

D. Bank A - OFF; Banks B, C, & D - ON.

i i

I 59 l

1 I

BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-60 Administrative controls or interlocks provided with the Charging pumps are designed to 1

accomplish which of the following:

1. If only two pumps are operable, ensure that they are not powered from the same bus.

l

2. Prevent operation of a Charging pump when the primary plant is water solid.

3. Prevent cross-tie of emergency busses through the swing pump breakers.

4. Prevent RCS over-pressurization in Modes 5 & 6 due to excessive flow.

A. I and 3 only.

B. 1,2, and 4.

C. 1,3, and 4.

D, 2 and 3 only.

60 l

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BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-61 Which of the following variables affect the Containment Spray systems capacity to depressurize the Containment in the event of a Design Basis Accident (DBA)?

1. Containment temperature.

2. Containment pressure.

3. RWST temperature.

4. Component Cooling Water temperature.

A. 1,2, and 3.

B. I and 2 only.

C.1 and 3 only.

, D. 2,3, and 4.

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O 61

BVPS-1 .

NRC Exam: 1 LOT 3, Rev 1 Question 12-97-62 Which of the following describes the sources ofinfluent into the Primary Drains Transfer Tank

[1DG-TK-1]?

A. RCP seal leak-off, Sample System drains, and Si valve stem leak-off, j B. CVCS Excess Letdown divert line, RCP seals and VCT drains.

C. Si header drains, PRT, and Sample System drains.

D. . Valve stem leak-off, CVCS Excess Letdown divert line, and Reactor Vessel head 0-ring leak-off.

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BVPS-1 l NRC Exam: 1 LOT 3, Rev 1 ]

Question 12-97-63 1

The following plant conditions exist:

. Unit I has tripped due to a loss of off-site power.

• #1 Diesel Generator has failed to start.

• All Steam Generator narrow range levels have remained above 21%.

Which of the following describes the status of the AFW system 60 seconds after the Reactor trip?

A. No AFW pumps running.

B. Both motor driven pumps are OFF; the steam driven pump is supplying AFW flow.

C. Both motor driven pumps are supplying AFW flow; the steam driven pump is OFF.

D. The ' A' motor driven pump is OFF; the 'B' motor driven and the steam driven pumps are supplying AFW flow, i

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BVPS-1 NRC Exam: lLOT3, Rev 1 -

Question 12-97-64

'Which one of the following is true with regard to Axial Flux Difference?

A. Boration will cause the Axial Flux Difference to become more negative.

B. When Axial Flux Difference is negative, more power is being produced in the top of the core.

C. For power levels greater than 50%, Axial Flux is maintained within + or -7% of target band.

D. When power distribution is distributed equally through the core, AFD is 1.

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l BVPS-1

• NRC Exam: lLOT3, Rev 1 Question 12-97-65 While responding to inadequate core cooling, the operators are unable to establish High Head l Safety injection. Core exit T/C's are 1450'F and RCS pressure is 165 psig. Which of the 1

following states the reason for staning the RCPs under these conditions?

)

A. Flush nitrogen from the S/G tubes so natural circulation flow can be established in subsequent steps. f B. Provide 2-phase forced flow for temporary core cooling to reduce Core exit T/C l

temperatures. l t

C. Assure the core remains shutdown by adding borated water from the loops to the voided I

core.

1 D. Provide forced RCS flow for heat transfer during S/G depressurization, {

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65 i

BVPS-1 .

NRC Exam: ILOT3, Rev 1 Question 12-97-66 Given the following conditions:

. Make-up to the RCS has increased and the following alarms are received:

. Reactor Coolant Pump Seal Leak-off Temp High.

. Reactor Coolant Pump Seal Leak-off Flow High.

. - Reactor Coolant Pump 1 A Seal Vent Pot Level High.

. Reactor Coolant Pump No.1 Seal Differential Pressure Low.

Which of the following has occurred to the l A RCP?  ;

A. #1 seal has failed.

B. #1 and #2 seals have failed.

C. All the seals have failed.

D. Sealinjection has failed.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-67 Given the following:

. - A Large Break LOCA has occcurred.

. SI, CI A, FWl, MSLI, and CIB are actuated. .

In this condition, the Containment Air Recirculation Fans...

l A. must continue to run to ensure Containment pressure returns to sub-atmospheric within Ihour.

B. will continue to run until the Containment sump level reaches the high level trip setpoint.

C. will be tripped due to the CIB.

D. will be tripped due to the SIS.

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BVPS-1 '

NRC Exam: ILOT3, Rev 1 Question 12-97-68 Unit 1 Laundry and Contaminated Shower Drain Tank (ILW-TK-6A] is being discharged to the Unit 1 Cooling Tower Blowdown. A HIGH-HIGH alarm is received in the contrcl room from the discharge Radiation Monitor (RM-lLW-ll6). Identify the automatic actions that occur.

[FCV-lLW-103] Contaminated (TV-lLW-ll6) Contaminated Drains Disposal Flow Control Drains Discharge header Rad Trip Valve A, Closes Closes.

B. Closes Opens.

C. Opens Closes.

D. Opens Opens.

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"- NRC Exam: ILOT3, Rev i Question 12-97-69 Given the fonowing:

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• Unit I has experienced a Large Break LOCA.

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e Containment pressure is 20 psig and dropping.

The Crew has reached the step in the procedure where CIB is to be reset. Which of the following apply in this situation?

CIB cannot be reset until Containment pressure is below the actuation setpoint.

A.

B. CIB cannot be reset until Containment pressure is subatmopheric.

C. CIB can be reset, but will re-actuate as soon as the reset switches are released.

D. CIB can be reset regardless of Containment pressure.

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BVPS .

I NRC Exam: 1 LOT 3, Rev 1 Question 12-97-70 In the event of a gross fuel element failure, which one of the following monitors would be the first to alarm?

l A. CVCS Letdown Radiation Monitor.

B. Containment Area Radiation Monitor.

C. Vent stack Radiation Monitor.

D. SPING Radiation Monitor.

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i BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-71 A plant heat-up is in progress. The following RCS temperatures were recorded at the given {

l times: ,

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TIME TEMP 1000 362F .!

l 1030 383F 1100 412F 1130 440F {

1200 459F Which of the following statements is correct?

A. No Administrative or Tech Spec limits were exceeded.

B. The Administrative limit was exceeded but the Tech Spec limit was not.

C. Both the Administrative and the Tech Spec limits were exceeded.

D Not enough data has been gathered to determine if any limits were exceeded.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 .

Question 12-97-72 r

l The unit is at 100% power with all systems in their NSA configurations for the current power l l

level. {

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Which of the following will occur if Vital Bus IV is de-energized?

A. An OTAT Turbine runback will occur. i

B. All Condenser Steam Dumps will open.

C. 'C' S/G Feed Reg Valve [FCV-lFW-498) shifts to manual.

D. The controlling PZR level channel fails low.

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BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-73 The following conditions exist:

. Unit 1 is operating at 100% power.

. Tavg is 576 F.

)

• Pressurizer pressure is 2235 psig.

• Delta 'l' is +8.

• Delta T is 60 F.

Which of the following plant parameter changes would cause the OTAT setpoint to LOWER?

A. Delta 'l' lowers to -2.

B. Tavg rises to 578 F.

C. Delta T drops to 55 F with Tavg remaining constant.

D. Pressurizer pressure rises to 2260 psig.

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! BVPS-1 NRC Exam: lLOT3, Rev 1 -

Question 12-97-74 l Prior to an outage, the Steam Generators were producing steam at 780 psia with an RCS Tavg of I' 575*F. During the outage 12% of the Steam Generator tubes were plugged.

What will be the new full power steam pressure if the RCS Tavg is to remain constant at the same power level as before the tubes were plugged?

A. 700 psia 10 psia.

B. 725 psia 10 psia.

I C. 780 psia i 10 psia.

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! D. 825 psia 10 psia.

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- NRC Exam: lLOT3, Rev 1 Question 12-97-75 What condition (s) would prevent the Reactor cavity upender from being operated?

! 1. The transfer car is at the Fuel Pool end ofits travel.

~

2. The transfer tube valve is closed.

3. The manipulator crane is over the upender with the mast in the full DOWN position.

A. 1 only.

B. 2 and 3 only.

C. I and 3 only.

D. 1,2, and 3 4

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BVPS-1 l NRC Exam: ILOT3, Rev 1 .

Question 12-97-76 Which of the following describes the condition associated with the starting air system for the i Diesel Generator that would cause the Control Room Diesel Generator Not Available annunciator?

l A. One of the two starting air header pressures less than 165 psig.

l B. Both of the starting air header pressures less than 165 psig.

C. One of the two starting air header pressures less than 200 psig.

l l D. Both of the starting air header pressures less than 200 psig.

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BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97 77 Which of the following indications would lead you to suspect a leak in the Fuel Pool liner?

A. Increased auto-make-up from the blender to the pool.

B. Fuel Building sump level high alarm.

C. Fuel Pool tell-tale' drains receiver level high.

D. Fuel Poal Cooling heat exchanger outlet temperature high.

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A problem in the Instrument Air system caused air pressure to momentarily (30 seconds) drop to 65 psig, then return to 105 psig. What should be the response of the Station Air Header Trip l Valve [TV-lSA-105]?

A. It closes, then re-opens.

B. It closes and remains closed.

C. It opens, then re-closes.

D. It remains open.

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BVPS-1

- NRC Exam: ILOT3, Rev 1 Question 12-97-79 Given the following:

l e Unit I was operating at 75% power

e A Small Break LOCA has occurred .

• RVLIS indicates that a void exists in the vessel head increasing RCS pressure will (1) the size of the void, and (2)

Pressurizer level.

(1) (2)

A. increase increase B. decrease increase C. increase decrease D. decrease decrease O

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E BVPS-1 NRC Exam: 1 LOT 3, Rev 1 Qitestion 12-97-80 l

A leak in which of the following components would result in a rising level in the Component l

Cooling Water Surge Tank?

A. Excess Letdown Heat Exchanger.

B. RCP Upper Bearing Cooling Water Cooler.

C. Regenerative Heat Exchanger.

D. Component Cooling Water Heat Exchanger.

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0 BVPS-1 '

NRC Exam: ILOT3, Rev 1 Question 12-97-81 Which of the following represents the condition of the steam entering the PRT from a leaking PORV if Pressurizer pressure is 1400 psig and the PRT is 15 psig. Assume an ideal thermodynamic process.

A. Superheated steam at 265 F.

B. Superheated steam at 280 F.

C. Saturated steam at 265'F.

D. Saturated steam at 280 F.

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BVPS-1

• NRC Exam: 1 LOT 3, Rev i Question 12-97-82 Given the following:

. Unit 1 in Mode 3 preparing for a normal plant cooldowm.

. Condenser Steam Dump system is in the Steam Pressure Mode controlling Tavg at 547F in automatic.

. Main Steam Header pressure transmitter [PT-lMS-464] fails low.

What manual actions will be required by the operator to continue the cooldown?

A. Manually close steam dumps by switching to the Tavg mode. Cooldown manually in the Tavg mode. j B. Manually close steam dumps by placing the controller to manual and reducing demand.

Cooldowm manually in the Steam Pressure Mode.

C. Manually open ti.e steam dumps by placing the controller to manual and raising demand.

Cooldown manually in the Steam Pressure Mode.

D. Steam Dump control is inoperable. Cooldown manually with the Atmospheric Steam Dump Valves or the Residual Heat Release Valve.

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BVPS-1 NRC Exam: lLOT3, Rev 1 Question 12-97-83 Given the following:

. Both units are operating at 100% power with all systems in normal system anangement.

~

. Operations manning is at minimum shift compliment at each unit.

. An oil fire breaks out in the Unit 1 Turbine building basement.

. The NSS determines a Unit 1 plant shutdown is required.

Which of the following is correct? i I

A. Off-site fire fighting personnel must be called in to combat the fire. Not enough personnel i are available on-site to fight the fire and shutdown simultaneously. .

l B. Off site fire fighting personnel must be called in to combat the fire. NPDAP 3.5," Fire Protection," requires the use of off-site personnel for any oil or chemical fires.

C. The opposite units fire brigade will fight the fire and the affected unit will perform the required shutdown duties.

D. Minimum shift compliment provides enough personnel to safely shutdown the plant and perform the required fire fighting duties.

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NRC Exam: ILOT3, Rev 1 l l Question 12-97-84 l Given the following: ,

l

! . A Control Room fire caused evacuation to the Shutdown Panel (SDP).

. The Reactor was tripped 10 minutes ago.

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. Plant control is established at the SDP with all equipment transferred.

With equipment controlled from the SDP, RCS temperature is controlled using (1) , and RCS pressure is controlled using (2)

I (1) (2)

A. Condenser Steam Dumps PZR heater group C.

B. Atmospheric Steam Dump valves PZR heater groups A and B.

C. Condenser Steam Dumps Auxiliary spray.

D. Atmospheric Steam Dump valves, PZR heater group C.

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i. BVPS-1 1 NRC Exam: ILOT3, Rev 1 Question 12-97-85 l

Given the following:

. A Reactor trip and Safety injection have occurred, along with the loss of all Feedwater.

. The steam driven AFW pump hasjust been returned to service.

. The crew goes to step 28 of FR-H.1, " Response to Loss of Secondary Heat Sink."

(attached)

. The following conditions are noted:

. RCS Hot Leg temperatures are all greater than 520F.

.. All S/G wide range levels are less than 10%.

. Core exit TCs are stable.

j The action that should be taken is...

A. establish AFW Cow ofless than 100 gpm to one S/G until narrow range level is greater than 6%. .

B. establish AFW flow of less than 100 gpm to all S/Gs until narrow range level is greater than l

6%.

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C. establish maximum AFW flow to one intact S/G.

D. establish maximum AFW flow to allintact S/Gs.

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BVPS-1 NRC Exam: ILOT3, Rev I o Question 12-97-86 Which of the following designates an Immediate Manual Action statement in Emergency Operating Procedures?.

A. The high level action statement is underlined.

B. The step number has a triangle around it.

C. The step number has a circle around it.

D. The high level action statement is bold.

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l BVPS1 NRC Exam: 1 LOT 3, Rev 1 Question 12-97-87 Which of the following correctly describes the monitoring of the Critical Safety Function Status Trees?

A. If an orange terminus is encountered, the STA is expected to monitor the remaining trees. If a red terminus is encountered on a lower tree, the crew will continue the orange terminus with the higher priority.

B. If a red terminus is encountered during the performance of the immediate action steps of E-0,

" Reactor Trip or Safety injection", the crew is expected to immediately perform the Functional Recovery Procedure required by the terminus.

C. A yellow terminus will require continuous monitoring until all conditions are satisified.,

D. If only green or yellow terminus exist, monitoring frequency may be relaxed to once every 10 to 20 minutes, unless a significant change in plant status occurs.

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BVPS-1 -

NRC Exam: lLOT3, Rev 1 Question 12-97-88 Given the following conditions:

• Mode 3.

. RCS Tavg is 547 F.

• RCS boron is 760 ppm.

• Burnup is 8000 MWD /MTU.

• Control Rods are inserted.

. Shutdown Banks are withdrawn.

Using the attached plant curves, determine which of the following statements is correct:

A. Minimum Shutdown Boron requirement is satisfied for the existing conditions.

B. A cooldown to Cold Shutdown can be accomplished at the present boron concentration.

C. A cooldown to Cold Shutdown can be accomplished at the present boron concentration if the Shutdown Banks are inserted.

D. The current boron concentration is adequate to maintain Hot Standby requirements if the Shutdown Banks are inserted.

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l BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-89 Which of the following methods should be used to Independently Verify the position of a CLOSED valve?

A. The person doing the Independent Verification may accompany the person closing the valve and observe him/her close it.

B. Attempt to move the valve in the closed direction.

C. A visual exam is all that is necessary, a " hands-on" verification is not required.

D. The Independent Verification for a procedure step can be substituted with the exact verification from the system valve line-up.

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BVPS-l' ,

NRC Exam: lLOT3, Rev 1 Question 12-97-90 l Given the following plant conditions:

. A Reactor trip and Safety injection have occurred from 100% power.

E-0," Reactor Trip or Safety Injection," has been completed through step 13 (Verify Feedwater Isolation).

e No Feedwater flow to the Steam Generators is indicated.

• AFW pumps cannot be started.

Which of the following actions is required?

A. Transition to FR-H.1 when directed to by E-0.

B. Immediately transition to FR-H.1," Response to Loss of Secondary Heat Sink."

C. Transition to FR-H.1 as soon as a transition out of E-0 occurs.

D. Go to ES-0.0, "Rediagnosis," which will allow a transition to FR-H.l.

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BVPS-1 NRC Exam: ILOT3, Rev 1

' Question 12-97-91 Per NPDAP 2.3, Attachment 5,(attached) which of the following represents a non-intent change to a procedure?

A. A change in initial conditions.

B. A modification to setpoints.

C. Deleting a QC hold point.

D. Addition of steps to return equipment to NSA.

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• NRC Exam: 1 LOT 3, Rev i

! Question 12-97-92 { '

!. Which of the following correctly specifies the maximum time period that a working copy of a procedure made from the Control Room copy may be used for valve manipulations without checking it against the original?

A. Up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the date on the copy.

B. Up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the date on the copy.

C. Up'to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the date on the copy.

D. Must be checked each shift.

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.. - BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-93 Which of the following are considered " extremities" when considering radiation dose limits?

A. Hands and feet only.

B. Hands, elbows, forearms, ankles and feet.

C. Hands, arms, legs, and feet.

D. Hands, elbows, forearms, knees, leg below knees, and feet. i i

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[ BVPS-1 NRC Exam: 1 LOT 3, Rev 1

• l l Question 12-97-94 Operators havejust stabilized the Unit from an unexplained 15% load rejection from full power when another, also unexplained,15% load rejection occurs. Which of the following actions should be taken?

A. Re-stabilize the Unit in accordance with the Load Rejection AOP.

B. Reduce load to <P-9 in accordance with the Emergency Shutdown AOP.

C. Place the Reactor in Hot Standby.

D. Start the second EHC pump, and place the EHC controller in Turbine Manual mode of operation.

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BVPS-1 NRC Exam: ILOT3, Rev 1 Question 12-97-95 During the performance of a Clearance procedure, what job position is responsible for authorizing the ESF checklist if the equipment to be cleared is safety related and required in ' the current mode of operation?

l. NSS.

2. ANSS.

3. STA.

4. RO.

A. I and 3 only.

B. 2,3, and 4.

C. I and 2 only.

D.1,2, and 3.

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BVPS-1 '

NRC Exam: lLOT3, Rev i Question 12-97-96 Which of the following Emergency Shutdown Panel indications are required by Technical Specifications 7

1. Auxiliary Feedwater Flow.

2. RCS Hot LegTemperature.

3. Source Range Flux.

4. Emergency 4kV Bus Voltage.

A. 1, 2, and 3.

B. I and 3 only.

C. 2,3, and 4.

D. 2 and 4 only.

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BVPS-1 NRC Exam: 1 LOT 3, Rev i j Question 12-97-97 An operator is required to be continuously stationed at a valve in a confined area for 30 minutes.

Radioactive material on/in this valve is exposing the operator to 200 mR/hr. Three feet behin'd the operator is another valve that emits 100 mR/hr at 25 cm. Which of the following is t

applicable to this situation?

A. The valve behind the operator must be labeled as a " Hot Spot" and the area posted as a Radiation Area.

B. The operator will exceed his/her 10 CFR 20 dose limits.

C. The area must be posted as a High Radiation Area and the operator must have an integrating dose meter, or Health Physics coverage.

D. A Health Physics technician should be present to monitor the radiation in the room with l

a portable neutron meter while the operator is stationed at the valve.

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l BVPS-1 NRC Exam: 1 LOT 3, Rev i ,

Question 12-97-98 DC Bus #1 has been de-energized due to a fire in the switchboard cabinet. A Reactor trip then occurs. Recovery from this transient is complicated because...

A. control of emergency loads on the I AE bus are not possible fro'm the Control Room.

B. the #1 Diesel Generator will start and load, but not be able to be stopped from the Control Room.

C. Charging and Letdown will not be available due to the loss of the Station Air Compressors.

D. temperature control will be on the Residual Heat Release valve because the Condenser and Atmospheric Steam Dumps will not be available.

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l E BVPS-1" NRC Exam: ILOT3, Rev 1 Question 12-97-99 Given the following:

ES-0.1, " Reactor Trip Response," has been entered from E-0, " Reactor Trip or Safety ,

Injection."

• Total AFW Gow is 330 gpm.

i j e Turbine Driven AFW pump secured.

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• 2 Motor Driven AFW pumps running.

• Tavg is 529 F and lowering slowly.

l- . All narrow range S/G levels are 4%.

A partial Feedwater isolation has occurred, and all Condenser and Atmospheric Steam Dump valves are shut. Assume all trends continue.

The conect operator response to these conditions is to...

A. reduce AFW Cow and initiate Main Steamline Isolation.

B. reduce AFW Gow and commence emergency boration.

C. initiate Main Stj;;mline Isolation and commence emergency boration.

D. Establish SGFW bypass Dow, secure AFW How, and initiate Main Steamline Isolation.

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BVPS-1 NRC Exam: 1 LOT 3, Rev i Question 12-97-100 With the plant operating at 100% power, Intermediate Range detector N35 fails high, followed immediately by N36 failing low. Which of the following is the correct operator response?

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A. Initiate a Reactor Trip and enter E-0," Reactor Trip or Safety injection."

B. Restore at least one Intermediate Range channel in the next hour or be in Hot Shutdown I

within the following six hours. I C. Since the number of channel operable is less than the number required by the Minimum Operability Channel requirement, Technical Specification 3.0.3 applies.

D. Power operation may continue in accordance with Technical Specifications.

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,- g-(* BVPSM NRC Exam: ILOT3, Rev i

! Question 12-97-101

1. D 26. A 51. C 76. A l

2. A 27. D 52. D 77. C l

3. C 28. B 53. B 78. B j l

4. A 29. B 54. A 79. D

5. D 30. D 55. D 80. A 1

6. C 31. A 56. B 81. A

7. B 32. D 57 D 82. C

8. C 33. A 58. C 83. D

9. D 34. B 59. D 84 B

10. B 5. ,B'6 60. C 85. A

11. B 36. C 61. A 86. C

12. D 37. B 62. D 87. D j

13. ,

C 38. A 63. D 88. D

14. B 39. B 64. C- 89. B

15. A 40. C 65. B 90. A

16. B 41. A 66. A 91. D

17. D 42. A 67. C 92. C

18. B 43. C 68. A 93. D

19. D 44. B 69. D 94. C l

20. C 45. C 70. A 95. D i l

21 A 46. B 71. B 96. A -

22. C 47. C 72. C 97. C

23. D 48. A 73. B 98. A

24. B 49. A 74. B 99. C i l 25 D 50. B 75. C 100. D 101

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Attachment 2 SIMULATION FACILITY REPORT Facility Licensee: Beaver Vallev Unit 1 Facility Docket No: LM Operating Tests Administered from: December 16-17.1997 This form is used only to report simulator observations. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observations.

No simulator deficiencies, that affected the scenario examinations or JPMs, were identified during the conduct of the examinations.

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