IR 05000305/2005301

Er 05000305/2005301(DRS); 12/14/2005-12/18/2005; Kewaunee Nuclear Power Plant; Initial License Examination Report

ML060120466
Person / Time
Site:	Kewaunee
Issue date:	01/10/2006
From:	Hironori Peterson NRC/RGN-III/DRS/OLB
To:	Christian D A Dominion Energy Kewaunee
References
50-305/05-301 50-305/05-301
Download: ML060120466 (28)

Text

January 10, 2006

Mr. David A. ChristianSenior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT: KEWAUNEE NUCLEAR POWER PLANTNRC INITIAL LICENSE EXAMINATION REPORT NO. 05000305/2005301(DRS)

Dear Mr. Christian:

On November 18, 2005, the NRC completed initial operator licensing examinations at yourKewaunee Nuclear Power Plant. The enclosed report documents the results of the examination which were discussed on November 18 and December 8, 2005, with Mr. M. Gaffney and Mr. S. Johnson, respectively, and with other members of your staff.NRC examiners administered the operating test during the week of November 14, 2005. Members of the Kewaunee Nuclear Power Plant Training Department staff administered the written examination on November 18, 2005. One Reactor Operator (RO) and six SeniorReactor Operator (SRO) applicants were administered license examinations. The results of the examinations were finalized on December 22, 2005. Five applicants passed all sections of their examinations, four of these applicants were issued respective operator or senior operator licenses. Two SRO applicants failed the written examination and will not be issued licenses. One applicant scored less than 82 percent on the written examination; and, in accordance withthe guidelines of NUREG 1021, "Operator Licensing Examination Standards for PowerReactors," ES-501.D.3.c, his license will be withheld until any appeal rights of the failedapplicants are exhausted. In accordance with 10 CFR Part 2.390 of the NRC's "Rules of Practice," a copy of thisletter and its enclosure will be available electronically for public inspection in the NRCPublic Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site athttp://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

D. Christian-2-We will gladly discuss any questions you have concerning this examination.

Sincerely,/RA/Hironori Peterson, ChiefOperations Branch Division of Reactor SafetyDocket No. 50-305License No. DPR-43

Enclosures:

1.Operator Licensing Examination Report 050000255/2005301(DRS)2.Simulation Facility Report3.Post Examination Comments and Resolutions4.Written Examinations and Answer Keys (RO & SRO)cc w/encls 1 & 2: M. Gaffney, Site Vice PresidentC. Funderburk, Director, Nuclear Licensing and Operations Support T. Breene, Manager, Nuclear Licensing L. Cuoco, Esq., Senior Counsel D. Zellner, Chairman, Town of Carlton J. Kitsembel, Public Service Commission of Wisconsincc w/encls 1, 2, 3, and 4: G. Winks, Training Manager D. Christian-2-We will gladly discuss any questions you have concerning this examination.

Sincerely,/RA/Hironori Peterson, ChiefOperations Branch Division of Reactor SafetyDocket No. 50-305License No. DPR-43

Enclosures:

1.Operator Licensing Examination Report 050000255/2005301(DRS)2.Simulation Facility Report3.Post Examination Comments and Resolutions4.Written Examinations and Answer Keys (RO & SRO)cc w/encls 1 & 2: M. Gaffney, Site Vice PresidentC. Funderburk, Director, Nuclear Licensing and Operations Support T. Breene, Manager, Nuclear Licensing L. Cuoco, Esq., Senior Counsel D. Zellner, Chairman, Town of Carlton J. Kitsembel, Public Service Commission of Wisconsincc w/encls 1, 2, 3, and 4: G. Winks, Training ManagerDOCUMENT NAME: E:\Filenet\ML060120466.wpd G Publicly Available G Non-Publicly Available G Sensitive G Non-SensitiveTo receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copyOFFICERIIIRIIINAMEBPalagi:coHPetersonDATE01/10/0601/10/06OFFICIAL RECORD COPY D. Christian-3-ADAMS Distribution

HKN JFS2 RidsNrrDirsIrib

GEG KGO SXB3 CAA1 C. Pederson, DRS (hard copy - IR's only)

DRPIII DRSIII PLB1 JRK1 ROPreports@nrc.gov MAB1 U. S. NUCLEAR REGULATORY COMMISSIONREGION IIIDocket No.:50-305 License No.:DPR-43 Report No.:05000305/2005301(DRS)

Licensee:Dominion Energy Kewaunee, Inc.

Facility:Kewaunee Nuclear Power PlantLocation:N490 Highway 42Kewaunee, WI 54216Dates:November 14 through 18, 2005 Examiners:B. Palagi, Chief ExaminerC. Phillips, Examiner D. Reeser, ExaminerApproved by:H. Peterson, ChiefOperations Branch Division of Reactor SafetyEnclosure 1 2Enclosure 1 2

SUMMARY OF FINDINGS

ER 05000305/2005301(DRS); 12/14/2005-12/18/2005; Kewaunee Nuclear Power Plant; InitialLicense Examination Report.The announced operator licensing initial examination was conducted by regional examiners inaccordance with the guidance of NUREG-1021, "Operator Licensing Examination Standards forPower Reactors," Revision 9.Examination Summary

• Seven examinations were administered (one Reactor Operator and six Senior ReactorOperator).*Five applicants passed all sections of their examinations, four of these applicants wereissued respective operator or senior operator licenses. Two SRO applicants failed the written examination and were not issued licenses. One applicant scored less than 82 percent on the written examination; and, in accordance with the guidelines ofNUREG 1021, "Operator Licensing Examination Standards for Power Reactors,"ES-501.D.3.c, his license will be withheld until any appeal rights of the failed applicantsare exhausted.

3Enclosure 1 3

REPORT DETAILS

4.OTHER ACTIVITIES (OA)4OA5Other.1Initial Licensing Examinations

a. Examination Scope

The NRC examiners conducted an announced initial operator licensing examinationduring the week of November 14, 2005. The licensee used the guidance established in NUREG-1021, "Operator Licensing Examination Standards for Power Reactors,"Revision 9, to prepare the examination outline and to develop the written examinationand operating test. The NRC examiners administered the operating test November 14through 17, 2005. Members of the Kewaunee Power Station Training Department administered the written examination on November 18, 2005. One Reactor Operator(RO) and six Senior Reactor Operator (SRO) applicants were examined.

b. Findings

Written ExaminationThe licensee developed the written examination. During their internal review, the NRCexaminers determined that the examination, as submitted, was within the range ofacceptability expected for a proposed examination. Written examination commentsdeveloped during review by the NRC staff, and as a result of examination validationwere incorporated into the written examination in accordance with the guidance contained in NUREG-1021.A total of nine post-examination comments (6 RO; 3 SRO comments) were submitted bythe applicants and station training department personnel on November 28, 2005. The results of the NRC's review of the comments are documented in Attachment 3, Post Examination Comments and Resolutions. Operating TestThe NRC examiners determined that the operating test, as originally submitted by thelicensee, was within the range of acceptability for a proposed examination. Theexaminers validated the operating test during the validation week and replaced or modified several items in the proposed operating test. Test changes, agreed upon between the NRC and the licensee, were made in accordance with NUREG-1021guidelines.Examination ResultsFive applicants passed all sections of their examinations, four of these applicants wereissued respective operator or senior operator licenses. Two SRO applicants failed the 4written examination and were not be issued a licenses. One applicant scored less than82 percent on the written examination; and, in accordance with the guidelines ofNUREG 1021, "Operator Licensing Examination Standards for Power Reactors,"ES-501.D.3.c, his licenses will be withheld until any appeal rights of the failed applicantsare exhausted.

.2 Examination Security

a. Inspection Scope

The NRC examiners briefed the facility contact on the NRC's requirements andguidelines related to examination physical security (e.g., access restrictions and simulator considerations). The examiners observed the implementation of examination security and integrity measures (e.g., security agreements) throughout the examination process.

b. Findings

No findings were noted in this area. The licensee staff was observed to be enforcingcorrect examination security procedures.4OA6Meetings.1Exit MeetingThe chief examiner presented the examination team's preliminary observations andfindings on November 18, 2005, to Mr. M. Gaffney and other members of the Operations and Training Department staff. A subsequent exit via teleconference was held on December 8, 2005, with Mr. S. Johnson following review of the site post examination comments. The licensee acknowledged the observations and findings presented. No proprietary information was identified by the station's staff during the exit meeting.ATTACHMENT:

SUPPLEMENTAL INFORMATION

1Attachment

1SUPPLEMENTAL INFORMATIONKEY POINTS OF CONTACTLicensee

M. Gaffney, Site Vice President

K. Hoops, Site Director

F Winks, Manager Nuclear Training

J. Ruttar, Manager Nuclear Operations

D. Fitzwater, Supervisor Nuclear Operations Training

S. Johnson, Operations Training

NRC

S. Burton, Senior Resident InspectorITEMS OPENED, CLOSED, AND DISCUSSEDOpened, Closed, and DiscussedNoneLIST OF ACRONYMS

USEDADAMSAgency-Wide Document Access and Management SystemDRSDivision of Reactor Safety

NRCNuclear Regulatory Commission

PARSPublicly Available Records

ROReactor Operator

SROSenior Reactor Operator

2SIMULATION FACILITY REPORTFacility Licensee:Kewaunee Power StationFacility Docket No.:50-305

Operating Tests Administered:November 14 - 17, 2005

The following documents observations made by the NRC examination team during the initialoperator license examination. These observations do not constitute audit or inspection findings

and are not, without further verification and review, indicative of non-compliance with

CFR 55.45(b). These observations do not affect NRC certification or approval of thesimulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observations.During the conduct of the simulator portion of the operating tests, the following item wasobserved:ITEMDESCRIPTIONEmergencyDieselGeneratorA problem was noted with the use of the Emergency Diesel GeneratorSpeed Control switch. When attempting to adjust frequency from 62 to

Hz, frequency increased uncontrollably resulting in a Diesel

Generator trip. This problem occurred on both the "A" and "B" Diesel

Generators.

3Question Number 32. Given the following:- The plant startup is in progress.- Reactor power is 1.4%.- FW-07A & B, S/G Main Feed valves, AND FW-10A & B, S/G Bypass Feedvalves, indicate closed.- Feedwater Pump A has just been started.- SG B level begins to slowly rise.What action is required?

A. Direct the NAO to locally close FW-9B, Feedwater Main Control Valve 1B Bypass Valve Inlet.

B. Cycle FW-10B, S/G B Bypass, fully open and closed.

C. Close FW-12B, S/G B Feedwater Isolation Valve.

D. Stop Feedwater Pump A.Answer: CApplicant Comment:An applicant commented that answer "D" should also be accepted as correct. Heargued that if level was near the high level trip of 67% stopping the Feedwater Pumpwould be a viable answer to preclude damage to the secondary system and the turbine.Facility Proposed Resolution:The facility management argued that there is only one correct answer for the givenconditions. Based on the given conditions, the guiding procedures are N-O-02, PlantStartup From Hot Shutdown to 35% power, and N-FW-05A, Feedwater System NormalOperation. Per N-O-02, the initial conditions are Steam Generator (SG) level ismaintained between 33% and 50%. Therefore, based on the question stem a normalplant startup level band should be assumed. Furthermore, the stem states that SG Blevel begins to slowly rise. Thus, there is no basis for assuming SG level is near orapproaching the SG Feedwater Isolation setpoint of 67%.

3NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. Based on the plantconditions given in the question stem (normal startup in progress) it is unreasonable topostulate that steam generator level is near the high level trip. A slowly rising steamgenerator level is anticipated by plant procedure under the conditions given thisquestion, and procedural guidance is provided to close the feedwater isolation valveunder these conditions (answer "C"). Therefore, "C" is the only correct answer.

3Question Number 34.Given the following:- The plant is at 100% power.- AFW Pump B is running for a surveillance test in progress.- Annunciator 47061-M, AFW PUMP B LOW OIL PRESS, alarms.What is the expected operator response for this condition?

A. Trip AFW Pump B, and go to N-FW-05B, Auxiliary Feedwater System.

B. Trip AFW Pump B, and go to A-FW-05B, Abnormal Auxiliary Feedwater System Operation.

C. Verify the Auxiliary Lube Oil Pump is running, and go to N-FW-05B, Auxiliary Feedwater System.

D. Verify the Auxiliary Lube Oil Pump is running, and go to, A-FW-05B, Abnormal Auxiliary Feedwater System Operation.Answer: D

Applicant Comment:An applicant commented that answer "B" should also be accepted as correct. Heargued that per the stem of the question, a surveillance test was in progress withAnnunciator 47061-M, AFW PUMP B LOW OIL PRESS, O

N. Since, the stem providesno information as to whether the aux lube oil pump starts and the alarm clears, andknowing the Auxiliary Feedwater (AFW) Pump is being run only for testing, it would becorrect to trip the AFW pump. The applicant sights UG-0 (Rev. F), User's Guide ForEmergency And Abnormal Procedures, section 6.1.2, which states that during abnormalconditions, an operator has the authority to take action that is clearly needed based onobserved condition. With the annunciator 47061-M not clearing, there is a lube oilproblem. Conservative Decision Making, would require the AFW pump be tripped.

Facility Proposed Resolution:The facility management agreed with the applicant that, without information about thecurrent oil pressure value and trend, an operator could not make a valid decision as towhether the Auxiliary Feedwater (AFW) Pump should be tripped or not. The questionwould need to be revised to make only one of the answers correct. It wasrecommended that two answers be accepted, "B" and "D."

3NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept two answers as correct. Because, the stem of thequestion states that the auxiliary feedwater pump low oil pressure alarm annunciates itwas expected that the applicants would select answer "D" which contained the actionscalled out in the annunciator procedure (verify the auxiliary oil pump is running). However, neither the question stem nor answer "D" provided information as to whetheroil pressure recovered. Therefore, it is equally reasonable to assume oil pressure hasnot recovered. Knowing that there has been a low oil pressure alarm, believing oilpressure to still be low, and from the stem knowing there is no emergency conditionbecause it is stated that the auxiliary feedwater pump is being run for a surveillancetest, it would also be correct to trip the auxiliary feedwater pump (answer "B"). Therefore both answers "B" and "D" were accepted as correct.

3Question Number 44.Which of the following situations requires entry into a Technical Specification LCOAction?A. Pressurizer Pressure Transmitter P-429 fails low while at 38% power.

B. Pressurizer level decreases to less than 17% with a reactor startup in progress.C. Pressurizer Backup Heaters energize after a 8% load reduction.

D. Pressurizer Pressure Transmitter P-449 fails high while on RHR Cooling.

Answer: A

Applicant Comment:An applicant commented that answer "D" should also be accepted as correct. Heargued that answer "D" requires entry into a Technical Specification LCO Action toMAINTAIN HOT SHUTDOW

N. Facility Proposed Resolution:The facility management disagreed with the applicant. They argued that with the planton RHR Cooling, as stated in answer "D," the reactor would be shutdown with ReactorCoolant System (RCS) temperature is less than 400ºF and RCS pressure less than425 psig. In these conditions the plant is in INTERMEDIATE SHUTDOWN, andtherefore, the LCO for PT-449 does not apply. NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. The applicantcontends that answer "D" should be considered correct because, although no action isrequired, the failure would place a Mode change restriction on the plant. A review ofTechnical Specification 3.5, Table 3.5-2, and Table 3.5-3 confirmed the facility positionthat entry into a Technical Specification LCO Action was not required for the conditionslisted in distractor "D." The fact that Transmitter P-449 has a Technical Specificationfunction in another plant Mode of operation is what makes it a good distractor. Only theconditions given in answer "A" would require Technical Specification directed actionsand therefore "A" was retained as the only correct answer.

3Question Number 62.Given the following:- Plant is at 90% power.- Power Range Instrument N-42 fails and is removed from service per A-MI-87,Bistable Tripping for Failed Reactor Protection or Safeguards Inst.- The surveillance for SP-47-011A, Reactor Coolant Temperature andPressurizer Pressure Instrument Channel I (Red) Calibration, comes due.- The decision is to bypass the failed NIS channel inputs to allow the calibration(Technical Specification 3.5.d).What is the coincidence for the OTT reactor trip while the surveillance is in progress?A. 2 out of 2B. 2 out of 3C. 1 out of 4

D. 1 out of 3

Answer: B

Applicant Comment:An applicant commented that answer "D" should also be accepted as correct. Heargued that, in a normal lineup, the OTDT trip coincidence is 2 out of 4 (4 channelsprovide input, 2 required to actuate). With N-42 input to Channel II - OTDT tripbypassed, only 3 channels provide input, 2 are still required to actuate to trip thereactor, therefore coincidence is 2 out of 3. During SP-47-011A, after initial switchlineups and verifications, the bistables associated with the RED Channel Pressure /Temperature trips are placed in TES

T. After that step, 3 channels still provide input if aBLUE or a YELLOW channel actuates the reactor will trip, therefore he ar

gued that thecoincidence becomes 1/3. Therefore he believes the correct answer depends on whichactions of SP-47-011A have been completed. Because, it was not stated in thequestion stem which actions of SP-47-011A have been completed, he contends answer"D" should also be accepted as correct. Facility Proposed Resolution:The facility management disagreed with the applicant. They argued that thecoincidence for the OTT reactor trip will always remain 2 out of 3 for the conditionsgiven in the question stem. While it is true that the number of bistables that mustactuate to cause a reactor trip varies between one and two at various times during the

3performance of SP-47-011A, the overall coincidence for the trip is still 2 out of 3. Therewill always be 3 channels, at times during the performance of SP-47-011A one of thethree bistables is actuated and only one additional bistable needs to actuate tocomplete the 2 out of 3 logic.It was also stated that an applicant requested clarification of the status of performingthe surveillance. It was answered that the surveillance had not yet been entered. Thisinformation was provided to all applicants in the room at that time. The individualproviding this feedback was not in the room, as he had completed his examination.NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. The applicantcommented that answer "D" should be considered correct because, at times in theconditions provided in the question stem, the trip of one more instrument would cause areactor trip. While it is true that at times during the completion of SP-47-011A the trip ofone more instrument would cause a reactor trip, that was not the question asked. Thequestion asked what was the reactor trip "coincidence," which in this case was 2 out of3 at all times, answer "B." The concept of what trip coincidence the plant is in is animportant concept for Technical Specification compliance. Because only answer "B"provided the correct reactor trip coincidence for the conditions given in the stem it wasretained as the only correct answer.The Facility noted that a clarification to this question was provided after the applicantproviding this comment had left the room having completed the examination. However, the clarification did not provide information necessary to select the correct answer. Therefore, again it was determined "B" was the only correct answer.

3Question Number 63Given the following:- Power level is 99.9% (1771.5 MWt).- The current UFMD and RTO OPERATING LIMITs are 1772 MWt.- SP-87-125, Shift Instrument Channel Checks - Operating, calorimetric wascompleted 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ago.- The signal is lost to PPCS from Feedwater Flow channel FT-476.- The RTO 1-minute average, PPCS point R5110G, drops to 1200 MWt.- The Computer Group reports the signal will be restored within 10 minutes.What is the affect on the UFMD and RTO Operating Limits, and what action is takenwhen the flow channel input is restored and R5110G reads normal?

A. Both the UFMD and RTO Operating limits will read 1749 Mwt. When the signal is restored, the operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction Factors screen to return both limits to 1772 Mwt.

B. The UFMD Operating Limit will read 1749 MWt and RTO Operating Limit will read 1772 Mwt. Power will be reduced to less than 1749 Mwt, and when the signal is restored, the operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction Factors screen to return the UFMD limit to 1772 Mwt.

C. Both the UFMD and RTO Operating limits will read 1769 Mwt. When the signal is restored, the operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction Factors screen to return both limits to 1772 Mwt.

D. The UFMD Operating Limit will read 1769 MWt and RTO Operating Limit will read 1772 Mwt. Power will be reduced to less than 1769 Mwt, and when the signal is restored, the operator will click the APPLY UFMD LIMIT button on the PPCS UFMD Correction Factors screen to return the UFMD limit to 1772 Mwt.Answer: A

Applicant Comment:An applicant commented that there was no correct answer to this question. He arguedthat when the flow channel is returned, the UFMD limit will go to 1772. Then, when the"APPLY UFMD LIMIT" button is clicked only the RTO Limit changes, not "both" the RTOand the UFMD Limit.

3Facility Proposed Resolution:The facility management maintained that answer "A" is correct. They agree that whenthe flow channel is returned, the UFMD limit will go to 1772. However, to have both thethe RTO and the UFMD Limit read 1772 the "APPLY UFMD LIMIT" button must beclicked. Therefore, answer "A" remains correct.NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to retain the question. This question had a two part answer.First, for the stem conditions what would the UFMD and RTO Operating Limit read, andthen how to reset the limits? Only answer "A" supplied the correct answer for the firstpart of the question. The applicant commented that the second part of answer "A" wasincorrect because the UFMD limit would automatically reset. While the second part ofthe answer could have been worded differently, it is correct. Both the UFMD and RTOOperating Limits would not read 1772 MWt until the APPLY UFMD LIMIT button waspushed. Because "A" is a correct answer the question was retained.

3Question Number 67Given the following:- Waste Gas Decay Tank A is in service.- The relief valve for that tank, WG-14A, lifts and fails to reseat. What is the effect on the plant?

A. The release will be automatically isolated when the Waste Gas Analyzer senses the pressure drop in the Waste Gas Decay Tank. B. The release will be automatically isolated when the Waste Gas Decay Tank pressure reducing control valve WG-201 closes.C. The release will NOT be automatically isolated, but will be monitored by the Aux. Building Vent radiation monitors R-13 and R-14.

D. The release will NOT be automatically isolated, but will be detected by the Charging Pump Room Area Monitor R-4.Answer: C

Applicant Comment:An applicant commented that the stem of the question lists the wrong valve number forthe Waste Gas Decay Tank A relief valve. He argues that if the question is answeredusing the valve with the valve number given in the stem, answer "D" could be a possiblecorrect answer, if a pipe failure was also assumed.Facility Proposed Resolution:The facility management maintained that this question had only a typographical that didnot invalidate the question. They agreed that in the question stem "The relief valve forthat tank, WG-14A, lifts and fails to reseat." should have read "The relief valve for thattank, WG-13A, lifts and fails to reseat." However, they argued that during theexamination no individual questioned the difference between the valve number and thevalve name. Additionally, they argued that even if someone attempted to answer thequestion for valve WG-14A, which is a valve that connects the Waste Gas Decay Tankto a vent header, answer "D" would not be correct.

3NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. This question asksthe effect on the plant of the relief valve on Waste Gas Decay Tank A lifting and failingto reseat. An applicant commented that the wrong valve number was given for therelief valve (which is true), and if one attempted to answer the question using the valvewith the provided valve number (and ignoring the fact that the stem clearly states it isthe relief valve that opens), answer "D" could be right. In fact "D" would not be a correctanswer even if the provided valve number is assumed to be correct. Additionally, noneof the candidates asked a question during the examination about the differencebetween the valve number and the valve description. This was a minor typographical

error that does not appear to have confused any applicant or invalidated the question. Therefore, "C" was retained as the only correct answer.

3Question Number 81Given the following:- The plant is at 100% power.- Surveillance Test SP-42-312B, Diesel Generator B Availability Test, is inprogress with DG B running.- Component Cooling Pump B trips on overcurrent.- BRB-104, ckt 10 supplying DG B tripped open and CANNOT be closed.What is the effect of this condition?

A. DG B is inoperable and must have its fuel supply locally isolated. A plant shutdown must commence within one hour using the Standard Shutdown Sequence (Technical Specification 3.0.c)B. DG B is Degraded but Operable and can be controlled locally. A 24-hour LCO is applicable for restoration of the DC Distribution System and a 72-hour LCO is applicable for restoration of Component Cooling Pump B.C. DG B is inoperable and must have its fuel supply locally isolated. A 72-hour LCO is applicable for restoration of Component Cooling Pump B and a 7-day LCO is applicable for restoration of DG B.D. DG B is Degraded but Operable and can be controlled locally. Only the 72-hour LCO is applicable for restoration of Component Cooling Pump B. Answer: C

Applicant Comment:An applicant commented that answer "A" should also be accepted as correct. Heargued that, this question has two answers dependent on the time frame. He agreesthat answer "C" is correct initially. However, he argues that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of theconditions given in the stem "A" Diesel Generator testing is required, because all "B"train emergency equipment is not OPERABLE this testing can not be performed. Therefore, at that time a plant shutdown must commence within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Facility Proposed Resolution:The facility management maintained that there was no reason, based on the questionstem, to extrapolate what the answer to this question could be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later. Theyargued that based on the conditions given in the question stem "C" is the only correctanswer.

3NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. The applicantcorrectly points out that under the condition of the stem a once per 24-hour test of the"A" Diesel Generator is required and cannot be performed. Therefore he believes that ifthe conditions still existed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later a shutdown must be commenced in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />,making answer "A" correct. While site management contends it is unreasonable toanswer this question by speculating what the answer could be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later, in that the24-hour period until Diesel Generator testing is required would be used to attempt toreturn equipment to service, thereby preventing the need to shutdown. TechnicalSpecifications 3.0 c, 3.3.d.2, 3.7.b.2, and 3.7.c were reviewed and it was found thatthere was no requirement that "a plant shutdown must commence within 1-hour" of theconditions given in the question stem. Therefore, "C" was retained as the only correctanswer.

3Question Number 88What is the reason for the Feedwater Isolation signal generated from High-High SGlevel?A. Preclude excessive SG tilts due to cooler feedwater supplied to ONE S

G.

B. Prevent overfill of the SG that may result in damage to secondary components.C. Ensure containment pressure remains within maximum internal pressure limit with the affected SG faulted inside containment.D. Protect the Feedwater Pumps from operating in runout condition with FW-7A/B, S/G A/B Main Valve, fully open. Answer: B

Applicant Comment:An applicant commented that answer "C" should also be accepted as correct. Heargued that per the Update Safety Analysis Report, one of the 4 major factors thatinfluence release of mass and energy following a steam line break is Steam Generator(SG) water level. Having too much inventory could challenge containment integrity, andthe Feedwater Isolation signal generated from High-High SG level prevents this.Therefore, "C" should also be considered a correct answer.Facility Proposed Resolution

The facility management disagreed with the applicant. They argued that the specificreason for Feedwater Isolation signal on Steam Generator (SG) high level was toprevent damage to secondary components as stated in the Technical SpecificationBasis.NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. The applicantcontended that there are two correct answers given for this question. He believes thatin addition to the reason given in the Technical Specification Basis, the High-Highfeedwater isolation is provided to limit containment pressure in the event of a faultedSteam Generator. This is a misconception on the part of the candidate. The USARanalysis for a faulted Steam Generator is based on normal level (not level at the High-High limit) and a feedwater isolation from the resulting safety injection signal. TheTechnical Specification 3.5 Basis states "Main feedwater isolation actuation occurs as aresult of a Hi-Hi steam generator water level to prevent steam generator overfill

3conditions. Steam generator overfill may result in damage to secondary components;for example, high moisture steam could erode the turbine blades at an acceleratedrate." no other reason for this trip is mentioned. Therefore, "B" was retained as the onlycorrect answer.

3Question Number 90Given the following:- The plant has experienced a fire in the Control Room.- The actions of E-O-06, Fire In Alternate Fire Zone, are being performed.What are the indications of an air dryer filter failure, and what are the proceduralactions performed by the Control Room Supervisor to mitigate the consequences of thisfailure.

A. Instrument Air Drier/Filter 1A/1B differential presure will rise above 5 psid. The CRS will verify SA-121, Air Drier/Filter Bypass CV, opens.

B. Instrument Air Drier/Filter 1c differential pressure will rise above 10 psid. The CRS will open SA-100A, Air Drier 1A Supply, and IA-300, 11/2" Alt IA, and then request the Control Operator A to start Air Compressor

B.C. Instrument air header pressure will drop below 95 psig. The CRS will depress the Air Drier 1C RESET pushbutton to confirm problem, and then align flow through Instrument Air Drier 1

B.D. Instrument air header pressure will drop below 100 psig. The CRS will open SA-70 and SA-71, 11/2" Dedicated Instrument Air Header Isolations, that bypass the Instrument Air Driers. Answer: D

Applicant Comment:An applicant commented that answer "A" should also be accepted as correct. Heargued that high D/P across IA-121 is an early indication of an air dryer filter failure andthat the normal automatic action for that condition is opening of the bypass valve. Therefore, "A" is a correct answer. Facility Proposed Resolution:The facility management disagreed with the applicant. They argued that answer "A"gives the correct actions under normal conditions. However, in the situation given inthe question stem the control room is evacuated and E-O-06 is in progress. Underthese conditions the Air Dryers are bypassed by opening the Dedicated Air Headerisolations, which supply the air directly through a separate in-pipe air dryer, making "D"the only correct answer.

3NRC Resolution

Upon review of the question, the applicant comment, and the facility pr oposedresolution it was decided to accept only the original correct answer. The applicantcontended that there are two correct answers given for this question. He argues thatanswer "A" contains the normal action for air dryer failure and is therefore correct. However, the question stem states that there is a fire in the control room and theactions of procedure E-O-06 are being carried out, these are the action described inanswer "D" not the action described in answer "A." Therefore, "D" was retained as theonly correct answer.

4WRITTEN EXAMINATIONS AND ANSWER KEYS (RO/SRO)RO/SRO Initial Examination ADAMS Accession

1. ML 060060460.