LIC-14-0076, 2014-05 Post Exam Analysis

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2014-05 Post Exam Analysis
ML14170A040
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 06/06/2014
From: Cortopassi L
Omaha Public Power District
To: Garchow S
NRC Region 4
laura hurley
References
LIC-14-0076, NUREG-1021, Rev 9
Download: ML14170A040 (54)
v • d • e


Text

Omaha Publ^Power District444South 16"'Street Mall Omaha, NE 68102-2247 10CFR55 LIC-14-0076 June 6, 2014 Mr.Steve GarchowChiefExaminer, OperationsBranch,Region IVU.S.Nuclear Regulatory Commission 1600 East Lamar BoulevardArlington,TX 76011-4511 Fort Calhoun Station,UnitNo.1 Renewed Facility Operating License No.DPR-40 NRC Docket No.50-285

Reference:

NUREG 1021Revision9,"Operator Licensing Examination Standards for Power Reactors"

SUBJECT:

NRC Licensed Operator Written Exam Reviews In accordancewithNUREG1021, the Fort Calhoun Station Training Department has completedareviewof the NRC licensed operator written exam that was conductedonMay27, 2014,atthe Fort Calhoun Station.During the exam review, the applicants identified issueswithfiveof the exam questions.

In three cases it was determined that the answer in the key was incorrect, in one case a question hadtworight answers, and it was determined that one question hadnoright answer.Therefore, Fort Calhoun Station is submitting recommended changes to theoriginalwritten exam key based on the applicant feedback.Twelve (12)questions on the reactor operator (RO)section of the exam were missed by 50%or more of the candidates.Eight(8)questions on the senior reactor operator (SRO)exam were missed by 50%or more of the SRO candidates.

These questions were reviewed for quality issues.All of the questions that were missed by 50%or more of the exam candidates were discussed with the candidatesonMay28, 2014.For each question that was missed by 50%or more of the candidates, the training material and training provided were reviewed.Ifyou require additional information, please contact Terrence W.Simpkin, Manager, Site Regulatory Assurance, at (402)533 6263.No commitments are made in this letter.The following items were providedtoyou electronicallyonMay27, 2014: 1.Copies of ES-401-7 and ES-401-8 cover sheets and each applicant's answer (bubble)sheetsinpdf format2.A copy of the exam analysis spreadsheet 3.Seating chart Employment with Equal Opportunity U.S.NuclearRegulatoryCommission LIC-14-0076 Page 2 The following enclosures are includedwiththis transmittal:

1.Grading quality checklist (ES-403-1).2.Theapplicant questionsandproctor answersprovidedduringthe written exam administeredonMay27, 2014.3.The original ES-401-7 and ES-401-8 cover sheetsandeachapplicant'sanswer (bubble)sheets.4.Copiesofthe originalapplicantsanswer sheetsgradedtothe originalanswerkey.5.Copyofthe SRO and RO answer keys.6.Requested changestoFCS exam answer key.7.Two copiesofthe exam analysis spreadsheet, one with applicants' names and one with the applicants' names redacted.8.Original seating chart.9.ES-201-3Examination Security Agreement.

Respectfully, Loiitfp^Cortopassi Site Vice President and CNO LPC/epm Enclosuresc:M.L.Dapas, NRC Regional Administrator, Region IV, w/o enclosures J.M.Sebrosky, NRC Senior Project Manager, w/o enclosures J.C.Kirkland,NRC Senior Resident Inspector, w/o enclosures Document Control Desk, w/o enclosures Requested Changes to FCS Exam Answer Key Summary.The following issues wereidentifiedduring the 2014 NRC Examreviewwith the applicants.

Each of the following requested changes have been reviewed and approved for submittal as changes to the exam by theFCSTrainingDirector,Shift Operations Superintendent,andSiteVice President.Inaddition,a peerreviewof each change requested to the exam was conducted by the Exelon CorporateTraining-Operations Training Program Specialist.

Question 46-Missed bv all Applicants Plant References provide contradictory informationonhowlong the batteries will last afterminimizingDC loads.The USAR providesinformationto supportboth4 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.However, the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> statedfora station blackout(SBO)in the USAR refers to the commitment to meetafour hour duration during a SBO and does not mean the batteries will only last that long.The bounding"maximum" time is actuallyforaDBAin accordance with calculation FC-05960.A noteinMVA-24, Minimizing DC Loads supports 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for bothaDBA and SBO.FCS respectfully requests that choice D be accepted as the only correct answer.Condition Report 2014-0664 has beenwrittento document this condition.(See attachment One)Question 47-Missed bv 5 of 6 Applicants FCS requests choiceDis accepted as the only correct answer.ChoiceDis supported by surveillance testEM-ST-EE-0006,BatteryNo.2 (EE-8B)Capacity Discharge Test data, provided as attachment Two.Question 58-Missed bv all Applicants Based on the CET temperature stated in question 58, this question has only one correct answer.If the temperature indicated by the failed CET is significantly different from the temperatures of the other CETs, the failed CET will be rejected and will not be used in the calculation of"Representative CET Temperature."Thisis based on the Chauvenet Criterion using standard deviationofvalid samples.If the applicant assumed normal operating conditions,whichisa standard assumption if conditions are not stated, then a CET temperature of 123.4°F would be outside one standard deviation and not be used in calculations by the QSPDS.If the temperature indicated by the failed CETisnot significantly different from the temperature of the other CETs, the failed CET will be used in the calculation of"Representative CET Temperature." Therefore, choiceCis correct based on the value provided for the CET indication in alarm.FCS respectively requests that choice C be accepted as the only correct answer.(See Attachment

3)

Requested Changes to FCS Exam Answer Key Summary.Question 83-Missed bv both SRO Applicants FCS believes this question has two correct answers.Choices A and B both describe actions that are takeninAOP-08(Attachment4).Althoughthe event occurredinthe containmentandnotthe Auxiliary Building, step8wouldeventuallybeperformedfor either location.The stem states radiation monitors arenotin alarm;therefore VIAS willnothaveoccurred.

If VIAS hasnotoccurred,thecontingencyactionfor step8requires manual VIAS actuation making choice A correct.ChoiceBis also an action taken in AOP-8 as askedbythequestion"whatisanactiontaken."FCSrespectfully requests that choices A and B both be accepted as correct.Question 98-Missed bv 1 of 2 SRO ApplicantsQuestion98 asks"Whatisthe maximum dosethatan individualatthesiteboundary could experience in thefirsttwo hours following the accident?" The question's choices are based on the regulatory limits and not the actual USAR Chapter 14.14 Steam Generator Tube Rupture Accident analysis.The tableinUSAR section 14.14.4 indicates the actual dose would be 1.0REMfor either a Pre-Accident Spikeora Concurrent Spike condition.

The limits are listed as 25 REM and 2.5REMfor each condition respectively.(Attachment 5)Therefore, FCS believes this question has no correct answer and requests it be deleted.

Attachment 1 Question 46;Station Battery Capacity Following Minimizing DC Loads.

Original Question questions report for 2014 Draft NRC WRITTEN EXAM (46)QUEST10N NUMBER: 046 The plant has experienced a station blackout (SBO).What is the maximum time the station batteries are designed to maintain design voltagetoDC Bus 1 and2in this condition?A.4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />swithno operator action.B?'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if non-essential DC loads are removed.C.8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />swithno operator action.D.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> if non-essential DC loads are removed.Plant References provide contradictory information on how long the batteries will last after minimizing DC loads.The USAR provides information to support both 4 and 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.However, the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> statedfora station blackout (SBO)in the USAR refers to the commitment to meet a four hour duration during a SBO and does not mean the batteries will only last that long.The bounding"maximum" time is actuallyforaDBAin accordance with calculation FC-05960.A noteinMVA-24,MinimizingDC Loads supports 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for bothaDBA and SBO.FCS respectfully requests that choice D be accepted as the only correct answer.Condition Report 2014-0664 has been written to document this condition.

Thursday, April 24,2014 3:59:27 PM 89 EPG Step: Safety Function 1 Deviation:

1)EOP criteria does not discuss the option of boration to compensate for CEAs that arenotfully inserted.10CFR50.63 and REG GUIDE 1.155 require only that Station Blackout procedures address Station Blackout events with the only further complication being RCS inventory loss (rates of approximately 100 gpm at hot standby RCS pressures, 25 gpm leak rate from each RCP, and normallimitsof on-line leakage 10 gpm).Hence, existing minimum NRC requirements do not require station blackouts with failures of passive reactivity control systems (CEA failures)to be addressed.

In the extremely unlikely case that all regulating and shutdown CEAs arenotfully inserted, as long as electrical power is not available to charging and HPSI pumps boration success paths are not available anyway.If more than one CEA was not fully inserted during Station Blackout the following sequence would occur: The fact that all CEAs had not been inserted would be identified during performance of Standard Post Trip Actions.Ifall rods were not insertedorcold shutdown boron concentration achieved by the time that an initial event diagnosis was made operators would be directed to the Functional Recovery Procedure EOP-20.As soon as power was restored to bus1A3or1A4, EOP-20 would require that boration to the RCS be implemented using charging or safety injection systems.This justifies deviation 1.For all emergency events, the reactor must be shutdown.Reactor power lowering in conjunction with negative startup rateisa positive indication that reactivity control has been established.

The criterion that no more than one regulating or shutdownCEAisnot insertedisin compliance with Technical Specification requirements.

10CFR50.63 andREGGUIDE 1.155 require only that Station Blackout procedures address Station Blackout events.Hence SBO procedures are not required to address Station Blackout events that are complicated by failures of multiple CEAs tofallinto the core, causing an event where a reactortripis required 2.Maintenance of Vital Auxiliaries ACCEPTANCE CRITERIA a.125V DC Bus 1 and 2 energized.b.DC loads have been minimized per Attachment MVA-24, Minimizing DC Loads.TBD-EOP-07 Page 50 of 63 R17 TBD-EOP-07 Page 51 of 63 ERG Step: Safety Function 2Deviation:1)TheEOP requiresbothDC busestobe energized,theEPGonly requires one train.2)The EOP confirms that DC loads have been minimized.

3)The EPG requires onetrainof 120 VAC instrument bus power be energized.

The 125V DC system is designed as one of the basic sources of power for plant control and instrumentation.Duringa StationBlackout,bothDC buses should remain energized.

One DC bus is required asaminimumtoprovidemonitoring andlimitedcontrolof other safety functions.

The loss of one DC bus limits theabilityto take an optimal approachtoplant recovery.Forthis reason, both DC buses must be energized to remain in the Station Blackout procedure.

The DC buses provide powerforDC control power and thefour(4)vitalAC instrument bus inverters.TwovitalAC inverters receive powerfromDCBus one (1), the othertwofromDC bustwo(2).By requiring both DC buses to be energized, power will be availableforallfourAC instrument bus inverters andDCcontrolpower.This justifies deviation 1.Studies at FCS indicate that USAR requirementsfor8 hours of batterylife,DC loads must beminimized.This condition is applicable during any DBA, not just Station Blackout.This safety function status check item implements the requirements of FC05960, and helps ensure that batterylifeis extended.This justifies deviation 2.Directions to recover a deenergized DC bus during a Station Blackout are in the Functional Recovery Procedures.IfaDC bus is deenergized the Safety Function Status Check for Station Blackout vital auxiliaries cannot be met and the Functional Recovery Procedures will be implemented.

The EOP does not address 120 VAC instrument power.The 120 VAC instrument buses are powered via the vital inverters fed from the DC buses.Thus,ifavital bus is energized along withaDC bus, the 120 VAC bus will remain powered.This justifies deviation 3.

USAR-8.4 Information Use Page9of19 Emergency Power Sources Rev.18 Loading of each diesel-generatorunitis functionally diagramed in Figure 8.4-2;the control circuits and automatic load sequencer operation are discussed in Section 7^.The arrangement of the motor control centers supplying Engineered Safeguards and essential loads and the load distribution are shown in USAR Figure 8.1-2.Periodical maintenance and inspection of the Emergency Diesel Generators is performed and controlled by the plant Preventive Maintenance Program.8.4.1.3 Design Analysis The capacity of each Diesel-Generator is adequate to support the operation of required Engineered Safeguards under the most restrictive design basis accidentfrominitiation through long term post accident cooling.8.4.2 Station Batteries 8.4.2.1 Design Bases Station batteries are an emergency source of d-c and a-c power for instrumentation and control, and are elements of the d-c systems generally described in Section 8.3.4.The battery installation is designed to survive without interruption of output or impairment of function of the environmental design bases cited in Section 8.1.1 for Class 1 seismic.The capacity of the storage batteries in the two, separate d-c systems is adequateforupto8 hours operation of control and instrumentation devices required in the eventofaDBA, or for reactor shutdown and standby, without battery charger operation.

To ensure 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of battery capacity manual actions mu^be taken to minimize DC system loads.The battery worst case loading schedule assumed during these ernergenc^

conditions accounts for operation of the following:

DC Bus#1 1.Emergency Bearing Oil Pumps (including starting transient and subsequent shutdown)2.EmergencyLighting(a portion is load shed per EOPs)3.Breaker transient loads and DG start attempts 4.Continuous instrument and control loads.

USAR-8.4 Information Use Page 10of19 Emergency Power Sources Rev.18 5.Shutdown of nonessential loads.DC Bus#2 1.Emergency Turbine Seal Oil Pump (including starting transient and subsequent shutdown)2.EmergencyLighting(aportionis load shed in minutes)3.Breaker transient loads and DG start attempts 4.Continuous instrument and control loads.5.Shutdown of nonessential loads.Analysis has demonstrated that the installed station batteries have adequate capacity to meet the present eight hour load demand.The capacity of the batteries therefore meets the required load demand criteria.The battery capacity calculation, FC05690, Battery Load Profile and Voltage Drop Calculation, verifies that the class 1E batteries have sufficient capacify to meet station blackout (SBO)loads for four hours assumirrg certain loads hot needed to cope with an SBO are removed;These loads are identified in the battery capacity calculation FC05690 and are Incorporated into the applicable emergency/abnormal operating procedure forminimizingDC loads.(Reference 8.7.2)8.4.2.2 Description and Operation Two storage batteries are provided.Each battery has sufficient capacity to meet the power demands as described in Section 8.4.2.1.Each battery is installedina separate room for physical segregation and protection; the rooms are separately ventilated.

Battery racks are designedtohold the battery cells in position in the event of the maximum hypothetical earthquake.

The arrangement of the battery rooms and their ventilation is as shown in Figure 8.4-3.During normal operation, the batteries share in meeting control power demand only during peaks when the battery charger rating is exceeded;otherwise the battery chargers meet system demand and simultaneously float-charge the batteries to maintain full charge.

Attachment 2 Question 47: Battery Discharge Capacity Verses Time Original Question questions report for 2014 Draft NRC WRITTEN EXAM (47)QUESTI0N NUMBER: 047 Battery Charger 1, EE-8C,isfaultedand hasbeenremoved fromservice.Battery 1,EE-8A,is supplying Battery Bus AI-41A.AssumingtheDC loads remain constant, thebatteryvoltage will.A.Lower at a linear rateuntilfully discharged.

B/Remain stable, and dropsrapidlyat end.C.Lower exponentiallyuntilfully discharged.D.Initial short rapid drop off and then lower at a linear rate.FCS requests choiceDis accepted astheonly correct answer.ChoiceDis supported by surveillance test EM-ST-EE-0006, BatteryNo.2(EE-SB)

Capacity Discharge Test data.Thursday, April 24,2014 3:59:27 PM 91 Voltage ChangesDuringDischarge.At theendofacharge,andbefore opening the charging circuit, the volt^e of ea;hceilis about2.5to2.7 voitaAssoon as thechargingcircuitis opened, the ceii voltage drops rapidly to about2.1volts, within threeorfour minuteaThisis due to the formationofathinlayeroflead sulphate on the surface of the negative plate and b^een the lead peroxide and themetalof the positive plate.Rg.21 shows how the volt^changes during thelasteight minutes of charge, andhowit drops rapidly as soon ssthe charging circuit isopened.Thefmal value of the voltage after the charging circuit isopened isabout 2.15-2.18 voitaThisis more fully explained in Chapter6.Ifa current is drawn from thebatteryat the instant the charge is stopped, thisdropis morerapid.At the beginning of the discharge the voltage has already hadarapid drop from the final voltage on charge, due to the formation of sulphate as explained above.When a current is being drawn from the battery, the sudden drop is due to the internal resistance of the cell, the formation of more sulphate, and the abstracting of theacidfrom the electrolyte which filIs the pores of theplate.Thedensityof thisacidishighjust before the dischargeisbegun.Itis dilutedrapidlyatfirst,buta balanced condition is reached between the density of theacidin the platesandin themainbodyof the electrolyte, the acid supply in the plates being maintainedata lowered densitybyfresh acid flowing into them from themainbodyof electrolyte.

After theinitialdrop, the voltage decreases more slowly, therateof decrease depending on the amount of current drawn from the battery.The entire process is showninRg.22.VOI-TS 2.0O 1.9 1.8 1.7 1.6 L vl01:ha RQ<<-4-S HOURS Fig.22.Voltage Changes During Discharge Leai sulphateisbeingformedon the surfa:eaandin the body of the platea Thissulphate hasa higher resistance than theleadorlead peroxide, and the internal resistance of thecellrises, and contributes to thedropin voltsge.As this sulphate fomis in the body of the platea theacidisusedup.Atfirst thisacidiseasilyreplacedfrom themainbodyof the electrolytebydiffusion.Theacidin themainbodyof the electrolyteisatfirst comparativelystrong,or concentrated, causingafreshsupplyof aiidtoflow into the plates as fast asitisusedupin the platea This results in the acid in the electrolyte growing weaker, and thiSk in turn, leadstoa constant decrease in the rate at which the fresh a;id flowa or diffuses into the platea Furthermore, the sulphate, which is more bulky than the lead or lead peroxide fillsthe pores in theplate,makingit more and moredifficultforacidto reach theinteriorof the plate.This increases the rate at which the voltage dropa The sulphate has anothereffect.It fonns a cover over the active material which has not been acted upon, and makes it practically uselesa since theacidis almost unable to penetrate the coating of sulphate.We thus have quantities of active material which are entirely enclosed in sulphate, therebycuttingdown the amountofenergywhichcanbetakenfrom thebattery.Thus theformationof FORT CALHOUN STATION FORM 10 i SURVEILLANCE TEST COVER SHEET FC-1115A R6 Surveillance Test Number: EM-ST-EE-0006 Work Order Packaqe No: 343985 pi Title: BATTERY NO.2 (EE-8B\CAPACITY DISCHARGE TEST Due Dale: Nov 1.2009 Drop Date: Frequency:

103 jPMlD: 00000051 EEQ:NScaffold:NlTrending:

Jbi Special Conditions:

2009 RFO 1.Reason for Test (if other than scheduled):

2.Postponement explanation (if required):

Expected Completion:

Supervisor or System Engineer: Date: 3.Test results acceptable?

tXYes I]No[]Yes Date: I]No T\me\c?73T Tech Specs satisified?(Not required for acceptable test)Supervisor or System Engineer\]lS^xAgy Additional Code analysis requiredwithin72 hours?[3Yes I I INo[]N/A STA Initials: Responsible System Engineer: JQHNSON.IVIATTHEW-R 4.ST Coordinator initial test review WO required?Frequency Increased?

CR required?Retest required?System Engineer test review: X 1ST Coordinator test review:£1 ST Coordinator final test closeout review: (SO-G-23)Initials:[lYes[]Yes I lYes[lYes (.Initiair/Ofe Initials: Initials: i(L=Date: Datef^-3-t^

l/jNo lYl No No]No ate:V^^/'Date:

Fort Calhoun Station Unit No.1 EM-ST-EE-0006

\SURVEILLANCE TEST PAGE1 OF 66CAPACITYDISCHARGE TESTFORSTATIONBATTERYNO.2(EE-8B)

Change No.EC 47417, EC 45495 Reason for Change Added figuresforload bank.Added criteria for termination of test.Changed Independent Verification to Concunrent Verification.

Changed multimeter model from 8060A to 189, Identified Tech Spec numberswitha TS.Requestor Mike Swan, Vern Smolinski Preparer i Ron Shirley Issue Date'10-29-09 3:00 pm R21 FORT CALHOUN STATION SURVEILLANCE TEST CAPACITY DISCH CQMTIMUQUS IISF ARGE TESTFORSTATION BATTERYNO.2(EE-8B)

SAFETY RELATED EM-ST-EE-0006 PAGE 2 OF 66 1.PURPOSE 1.1 This procedureprovidesinstructionsforperforminga dischargeonthebatteryand calculating the battery capacity.1.2 This testisperformedatafrequencyofthefirstrefuelingshutdownafterinstallation, and everythirdrefueling shutdown thereafter.

1.3 This test satisfies,inpart,the requirements of Technical Specifications, TS 3.7(2)b.2.REFERENCES/COMMITMENT DOCUMENTSIons,TS2.7(1)g.

and TS 3.7(2)b.ce Test Program 2.1 Technical Specifica 2.2 SO-G-23, Surveillar 2.3 Ongoing Commitments

  • AR 12818, LIC-92-0192 2.4 TM C173.0010,TechnicalManualfor C&D(CharterPower Systems)Batteries File Description 2.5 Drawings P&IDFig 8.1-1 11405-E-8 2C6289 3.DEFINITIONS 12234 One Line Diagram Plant Electrical 12244 125VDCMisc.Power Diagram 44129 DC Dist.Panel EE-8G Schematic*CICV*Conrected individualcellvoltage*ELC-Electrolyte level correction*ETC-Electrolyte temperature correction
  • ICV-Individualcell voltage*MCM-Thousands of circular mils*SG-Specific gravity*VCT-Voltage correction for temperature R21 FORT CALHOUN STATION SURVEILLANCE TEST CQNITiNUQUS USE Attachment 9.8-Data Sheet 2A Step 7.2 Battery No.2 Isolated Date Time_M£Step 7.7 Electrolyte temperature correction factor Ki 1*0//Step 7.13.3 Battery Terminal Voltage Prior to Test/VDC EM-ST-EE-0006 PAGE 43 OF 66 f Test Time Date/Time Dischar ge Current Battery Terminal Voltage 0 27'7,2.//s-.i 15^J30 272, h 27-^.V/JS',¥//s,o(>30 27Z.Z Z7S.45 2ZoO 272.7 27S.L l/S,l 100 ZZ)S 272.Z 27S,Z//S-J mi?115 Z230 277,Z 27S.ir 130 22.^5-;?7S-,3 S-//H.n//H.2I 145 2ZOO 0.10.1 a 9-7 Z.HH-S'hh.13 200 ii/r llS.11^,95 215 1 2320 17Z-^27S,V//^/7/3.PS-230 J zsyf 27Zr 2?S-,6//s.'f//3;SS 245/^i^oooo Z72.S-20S-Z 7/S,Sr//S,Si 300\00/J" 275,i m,s J/3,'0 315 0030 272^%295^,6/'3,d , JJZ.S^330 7']2.^Z7S,sr//3,o//Z,6S'345 0/00 292.Z Z?S,2./J2,dr'JiZ.i'f 400 0//f 272-Y 2?S,/nz.s-HZ.tk 415 0/SO a->a.^1(1,77 430 aia.^3.1i: t ,ia-o mil 445 0200 c III.T in.34-500 OZ?S'Z7S'.o ll),S m,/z 515 V Oz3o 27S.6//Az I 770.so Cofn PU^R21 FORT CALHOUN STATION SURVEILLANCE TEST COMTINUQUS USE Attachment 9.8-Data Sheet 2A EM-ST-EE-0006 PAGE 44 OF 66 7^Test Time Date/Time Discharge Current Battery Terminal Voltage 530 aeys-2?Z,'?27S.V 545%/a^0300 2??,9 2?S,6//0,5~600"A'/JnOs/s ZIS,^}/0,Z 615"Mol<3330 27Z.^Z7S,Z ms-yo%yo 630 osys-27Z-Z-29^,7.J07>S'06 645"AdfOVco Z7Z,Sr 17 S.6 yo%j\^oa.7s 700"Adi 0 SOZ,Z 217,6 709.6/ce,z3 715 a^2.r tor.1 730 oyys-27Z,/Z7S.Z 7o7,i/07, Z/745 Yty/ofOsoa ZOZ.')ZT^L 7oZo/0^r6/800 2?f,?Z'JS.s/06.S-^06,/z, 815 11Z,S?7^://OS, 9 830 Z7Z,7 Z7s.y/0%S 845 272/9-275,6 703,0 joz, 900"A'/h 06/^peeeO a t^4 Cf rtithi f*X^915 930 945 1000 1 Step 7.13.14A Final Battery Tenninal Voltage Step7.13.14C Final Discharge Current Amps^^^^Step 7.13.140Endof Discharge Date Time%r\\w^R21 Battery OV&Load Graph Ft.Calhoun Unit 1 Battery OV Battery Load 117.5 115.0-i--f-1--1.-1-t" 1 ,ir iml TiU.ai: LJ iT T1 n%L^l jP mŽ>112.5 O S (0 OQ 110.0 107.5 105.0 102.5 00:47:50 01:35:50 02:23:50 03:11:50 03:59:50 04:47:50 05:35:50 06:23:50 07:11:50 07:59:50 Test Time 11/24/2009 Battery: battery 2 eni-st-ee-0006 Test Type: Performance Tested: 11/23/09 I (D*5 I 273.0 272.5 272.0 Page 1 ofI Attachment 3 Question 58;Suspect GET Input to QSPDS Calculations Original Question questions reportfor 2014DraftNRC WRITTEN EXAM (58)QUESTI0N NUMBER: 058 The"A" QSPDS is displaying a suspect alarm for one of the Gore Exit Thermocouples (GET).The alarm has NOT been acknowledged.1)Howis the UN-AGKNOWLEDGED suspect GET alarm displayed on the QSPDS plasma screen, and2)Will this suspect GET input be used in QSPDS calculations:

1)How is the susoect GET disolaved 2)used in calculations A.Normal Modewitha question mark in front of the value 7123.4 Not used B.Normal Modewitha question mark in front of the value 7123.4 Still used G.Inverse Modewitha question mark in front of the value BESi!Not used Dy Inverse Modewitha question mark in front of the value Still used Based on the GET temperature stated in question58,this question has only one correct answer.If the temperature indicated by thefailedGETis significantly different from the temperatures of the other GETS, thefailedGET will be rejected and will not be used in the calculation of"Representative GET Temperature."Thisis based on the GhauvenetGriterionusing standard deviationofvalid samples.If the applicant assumed normal operating conditions,whichisa standard assumption if conditions are not stated, then a GET temperature of 123.4°F would be outside one standard deviation and not be used in calculations by the QSPDS.If the temperature indicated by thefailedGETisnot significantly different from the temperature of the other GETs, thefailedGET will be used in the calculation of"Representative GET Temperature." Therefore, choiceGis correct based on the value provided for theGETindicationin alarm.FGS respectively requests that choice G be accepted as the only correct answer.Thursday, April 24,2014 5:54:05 PM 113 PAGE 1 OF 21 Fort Calhoun Station Unit No.1 OI-QSP-1 OPERATING INSTRUCTION QUALIFIED SAFETY PARAMETER DISPLAY SYSTEM OPERATION Change No.EC 49234 Reason for Change Converted from WordPerfect to Word.Requestor N/A Preparer J.Briggs Issue Date 06-17-10 3:00 pm R8 FORT CALHOUN STATION OI-QSP-1 OPERATING INSTRUCTION PAGE 2 OF 21 QUALIFIED SAFETY PARAMETER DISPLAY SYSTEM OPERATION ATT PURPOSE PAGE Attachment1-Display Hierarchy and Access 4 Attachment2-Parameter Alarms 5 Attachment3-System Errors 7 Attachment4-Operator Error 8 Attachment5-Saturation Margin Calculations 9 Attachment6-Reactor Vessel Level Above the Core Calculation 11 Attachment7-Core Exit Thermocouple (CET)Temperature Calculation 13 Attachment8-System Description 15 PRECAUTIONS 1.When energizing QSPDS the associated inverter power supply may transfer to the bypass transformer.

The Shift Manager shall specify whether the inverter should be manually placed in bypass per OI-EE-4 (requires inverter inoperability), when energizing the QSPDS and/or HJTC heaters.2.Dampers HCV-724A/B and HCV-725A/B have been permanentlyfailedin their accident positions.

Indication for these dampersisfor information only.Abnormal indication needs to be verified by ARP-ERFCS before action is taken.REFERENCES/COMMITMENT DOCUMENTS 1.Technical Specification:

  • 2.21, Post-Accident Monitoring Instrumentation 2.USAR:*Sections 7.5.5 and 7.5.6 R8 FORT CALHOUN STATION OPERATING INSTRUCTION OI-QSP-1 PAGE 5 OF 21 Information Use Attachment2-Parameter Alarms PREREQUISITES 1.Procedure Revision Verification Revision Number Date:2.At least one Qualified Safety Parameter Display System channel is operational.

PROCEDURE NOTES 1.A summary of display behavior for Parameter Alarms is showninFigure2.

2.There are three types of parameter alarms:*Setpoint alarms occur when a parameter value exceedsitshighorlow setpoint (not all parameters have a setpoint).

  • Out-of-Range/Bad Data alarms occur for an out-of-range sensor, a failed sensor, an electrical failure, or a software error.*Suspect alamis occur when aninputtoamulti-input calculated value Is out-of-range/bad data but the value can still be calculated with the remaining Inputs, or if the core exit temperatures are notwithinavalid range.3.In the INVERSE MODE, theNORMALMODE character and background colors are reversed.1.WHENa Parameter Alarm occurs for either a Setpoint or an Out-of-Range/Bad Data Alarm, THEN the relevant Top-Level Page enters the alarm state AND the corresponding System Alarm Indicator on the current Display Page is shown in BLINK MODE.a.IF the parameter is shownona lower level sector page, THEN higher level pages will show the relevant sector numbersinBLINK MODE.b.In the caseofa Setpoint Alarm, the parameter value is shown in INVERSEMODEwith an adjacent asterisk.(£1 INITIALS R8 FORT CALHOUN STATION OI-QSP-1 OPERATING INSTRUCTION l1 PAGE 6 OF 21 Information Use Attachment2-Parameter Alarms PROCEDURES (continued) i£i INITIALS 1c.In the case of an Out-of-Range/Bad Data Alarm, the parameter valuefieldisfilledwith question marks in INVERSE MODE.d.In the caseofa SuspectAlarm,a single question mark is displayed by the affected calculated value OR core exit temperature in INVERSE MODE.2.WHEN the Operator acknowledges an alarm, THEN:a.In the caseofa Parameter alarm, the relevant System Alarm Indicators and sector numbers remain in INVERSE MODE.b.In the caseofa SetpointAlarm,a second asterisk is shown next to the parameter value field.c.In the case of an Out-of-Range/Bad Data Alarm when the parameter hasahigh or low setpoint, the system and sector alarms either stay in the INVERSE or NORMAL MODE but change the question marks to theNORMALMODE.IF the parameter does not have any setpoints, then the system and sector alarms along with the question marks return to the NORMAL MODE.d.In the caseofa Suspect Alarm, the question mark and all alarms return to the NORMAL MODE.R8 FORT CALHOUN STATION OI-QSP-1 OPERATING INSTRUCTION lHI PAGE 13 OF 21 Information Use Attachment7-Core Exit Therrnocouple (GET)Temperature Calculation PREREQUISITES

££]INITIALS 1.Procedure Revision Verification Revision Number Date;2.At least one Qualified Safety Parameter Display System channel is operational.

PROCEDURE 1.The QSPDS displays individual Core Exit Temperatureswitha Core Map, the highest and next highest Core Exit Temperature in each quadrant, and a representative Core Exit Temperature.

2.The Representative Core Exit Temperature is calculated based on statistical analysis with practical checks from other inputs.a.Using the Representative CET Temperature the Pressure AND Temperature Saturation Margins are calculated and displayed.

However, the CET calculated variables will be flaggedwitha questionmarkduring the time that the QSPDS is Identifying Out-of-Range and Suspect CET inpute.b.WHEN the QSPDS completes one (1)complete cycle of the CET algorithm without flagging or removingaflagfrom any CET input, THEN the inputs are considered stable AND accurate and the question mark is removed from the calculated variables.

c.The CET calculated variables will also be flaggedwithall question marks if the numberofvalid CET inputs is less than nine (9).3.Inputs considered invalid areeitherfailedor deviated from themeanof the CET inputsbya specific amount.a.Failed inputs are displayed as all question marks.b.Deviated inputs are displayed as a question markinfrontof the displayed value indicating it is suspect.R8 FORT CALHOUN STATION OI-QSP-1 OPERATING INSTRUCTION 1llPAGE 14 OF 21 Information Use Attachment7-Core Exit Thermocouple (CET)Temperature Calculation PROCEDURE (continued)

(^)INITIALS4.WHENa Saturation Alarm occurs (either RCS or Upper Head),THENall previous valid (before the Saturation Alarm)CET inputs will be used in calculating CET variables.

a.Previously Suspect CET inputs will be used in the calculation if their temperature valuefallswithina specified band of the mean temperature.

5.The Core Exit Temperature alarms if the temperature is greater thanahigh setpoint.However, Suspect inputs will NOT be alarmed.R8 FORT CALHOUN STATION OPERATING INSTRUCTION Figure2-Alarm Behavior OI-QSP-1 PAGE 19 OF 21 PAGE ELEMENTS ALARM STATUS ACK STATUS MECHANISMS EXAMPLES Normal or Return to Normal-Normal Mode CORE PPP 2 System Alarm Indicators and Sector Numbers Parameter Alarm on Associated Page UNACK Blink Mode (CNMT PPP)(2)ACK Inverse Mode (CNMT PPP)(2)ACK Normal Mode Inverse Mode CNMT PPP (2)Normal or Return to Normal-Normal Mode 123.4 Setpoint Alarm UNACK Inverse Mode 1 asterisk (123.4*)ACK Inverse Mode 2 asterisks (*123.4*)Parameter Value Field Out-of-Range

/UNACK Inverse Mode Question Marks (??????)Bad Data Alarm ACK Normal Mode Question Marks Suspect Alarm UNACK Inverse Mode Question Mark (?123.4)ACK Normal Mode Question Mark?123.4 Blink Mode: Alternation of the Normal and Inverse Modes.R8 SDBD-COMP-300 Information Use Page 43 of 112 ERF Computer and QSPDS ComputerRev.21 5.1.2B (continued) 7.QSPDS Computer Functional Design a.The QSPDS computerisa dual-channel seismlcally-qualified display of safety parameters, designed to provide a backup function to the primary SPDS (Ref.10.3.8, and Ref.10.9.3).Refer to Attachment 10 of this SDBD for further design details.b.The QSPDS computer is designed to monitor, process and display information relevant to the core cooling conditions.

The QSPDS uses inputs from the heated junction thermocouples (HJTCs), core exit thermocouples (CETs), and loop temperature and reactor pressure instruments (Attachment 6, Items 7, 8, 9,10, 29 and 30;and Attachment 7, Items 96 through 113)to provide display of ICCI variables and indication of reactor vessel liquid inventory, reactor coolant temperature, and saturation/superheat margin (Ref.10.4.14, Ref.10.4.15, Ref.10.4.35, Ref.10.4.36, Ref.10.4.37, and Ref.10.9.3).In addition to the display of the ICC input variables and of the calculated variables, the QSPDS computer performs the alarming functions of impending ICC, interfacing to the ERF computer for data transmission and display on the SPDS, flagging of suspect calculations and of hardware or software malfunctions, and control of power to HJTC heaters (Ref.10.4.5, Ref.10.4.29, Ref.10.7.1, and Ref.10.9.3).c.Data used by the QSPDS computer is checked for an out-of-range condition, or when the configured setpoint is reached.Thermocouple inputs are checked for an open thermocouple condition.

An alarm is initiated to warn of a bad data orofa setpoint that is reached.The setpoint alarm occurs when an input parameter exceedsitshighorlow value, around the setpoint.The bad data alarm occurs when an input parameter value exceeds its allowable range.This may result from an out-of-range sensor, a failed sensor, an electrical failure or a software error.

SDBD-COMP-300 Information Use Page44of 112ERFComputerandQSPDSComputer

^Rev.21 5.1.2B7 (continued)d.A suspect temperature readingwithina set of CETs is identifi^from the application of the Chauvenet Criterion using standard deviation of the valid samples.Valueswhichfail the validity check are flagged before transmission to the ERF computer and are displayed on the ERF computer operator interface displayswithablinking white question mark (?).These values are displayed on the QSPDS computer PDU, precededbya question mark(?)for suspect data, or are replacedbya series of question marks (???)for out-of-range data (Ref.10.3.8, Ref.10.4.5, Ref.10.4.6, and Ref.10.9.3).e.The QSPDS computer monitors, processes and transmits to the ERF computer for display, selected Regulatory Guide 1.97 variables (Ref.10.4.25 and Ref.10.4.27).Refer to Attachment 12 of this SDBDforalistingof these variables.

f.The QSPDS computer is designed for an operational availability of 99 percent (Ref.10.9.3).C.Monitoring InstrumentationandControl Design1.Monitoring Instrumentation and Control Design a.The SPDS software resident in the ERF computer host system uses the information transmittedtoitby theDASin order to provide a continuous indication of plant parameters or derived variables representative of Fort Calhoun status (Ref.10.4.4, Ref.10.4.5, Ref.10.4.6, Ref.10.4.31, Ref.10.4.42.

Ref.10.4.43, 10.9.6 and Ref.10.9.15).The SPDS software uses three level hierarchical paging system: TopLevel(Level One), Midlevel bargraphs(LevelTwo) and supporting P&ID displays (Level Three)to provide the operator with information about:*Reactivity Control (Ref.10.9.13 and Ref.10.9.14)*Vital Auxiliaries (Ref.10.9.13 and Ref.10.9.14)*Reactor Coolant System Inventory Control (Ref.10.9.13 and Ref.10.9.14)*Reactor Coolant System Pressure Control (Ref.10.9.13 and Ref.10.9.14.*Core Heat Removal (Ref.10.9.13 and Ref.10.9.14)*Reactor Coolant System Heat Removal (Ref.10.9.13 and Ref.10.9.14)*Containment Integrity (Ref.10.9.13 and Ref.10.9.14)

Statistical Rejection of"Bad" Data-Chauvenet's Criterion Occasionally,whena sample of N measurements of a variable is obtained,theremaybeoneormore that appear to differ markedly from the others.If some extraneous influence or mistake in experimental techniquecanbe identified, these"bad data" or"wild points" can simply be discarded.

More difficultisthe common situationinwhichno explanation is readily available.Insuch situations, the experimentermaybe tempted to discard the valuesonthebasis that somethingmustsurelyhavegonewrong.

However, this temptationmustberesisted,since such datamaybe significant either in terms of the phenomena being studiedorin detecting flawsintheexperimentaltechnique.Ontheotherhand,onedoesnotwantanerroneousvaluetobiastheresults.Inthiscase,a statistical criterionmustbeusedtoidentify pointsthatcanbeconsideredforrejection.Thereisnoother justifiablemethodto"throwaway"datapoints.Onemethod that has gained wide acceptance is Chauvenet'scriterion',thisteclmiquedefinesanacceptablescatter,inastatisticalsense,aroundthemeanvaluefromagivensample of N measurements.

The criterion statesthatall data points should be retained that fall within a band aroundthemeanthat correspondstoa probability of 1-1/(2N).In otherwords,datapoints can be considered for rejection only if the probability of obtaining their deviationfromthe meanisless than 1/(2N).This is illustrated below.Frequency distribution Prob=1-1/(2N)Reject data y w Reject data m The probability 1-1/(2N)for retention of data distributed about the mean can be related to a maximum deviation d^axawayfromthemeanbyusingthe Gaussian probabilities in AppendixA.Forthegiven probability, the nondimensional maximum deviation Xmaxcanbe determined from the table where (Xi-X)d max_"max Sx and Sxisthe precision index ofthesample.

Therefore, all measurements that deviatefromthe mean by more than XmaxSxcanbe rejected.A new mean valueanda new precision index can then be calculatedfromthe remaining measurements.

No fiirther application of the criterion to thesampleisallowed; Chauvenet'scriterionmaybeappliedonlyoncetoagivensample of readings.Thetablebelowgivesthemaximumacceptabledeviationsforvarioussamplesizes.

Values of dmax/SxforothersamplesizescaneasilybedeterminedusingtheGaussianprobabilitytablein Appendix A.Chauvenet's Criterion for Rejecting a Reading Number of Readings£N}Ratio of Maximum Acceptable Deviation to Precision Index (dmav/Sv')

3 1.38 4 1.54 5 1.65 6 1.73 7 1.80 8 1.87 9 1.91 10 1.96 15 2.13 20 2.24 25 2.33 50 2.57 100 2.81 300 3.14 500 3.29 1,000 3.48 Example 4 For the 10 temperature measurements in Example 2, determine if any should be rejected by Chauvenet's criterion.

SOLUTION Inspectingthedatafrom Example2,thefifth reading of T=98.5°F appears to deviate substantiallyfromthe othersandis therefore a candidateforrejection.Fromthetableabovefor N=10, the maximumdimensionlessdeviation is dn,ax/Sx=l-96.

Since Sx<<0.49°F in this case, dmax=1.96Sx=(l-96)(0.49°F)=0.96°F.

The deviation for the fifth reading is 1.125°Fsoit canindeedbe rejected, although no others can.Eliminating this point and recalculatingthemean and precision index results inX=97.25°F Sx=0.3rF Comparing these values with those calculated in Example2,Xis decreasedbyonly about 0.13%while Sx is decreased by over 37%.

/m USERS GUIDE FOR THE QUALIFIED SAFETY PARAMETER DISPLAY SYSTEM TD C490.0350 CAUTION DRAWINGS ARE FOR INFORMATION ONLY.CHECK THE DRAWING CONTROL PROGRAM FOR THE LATEST REVISION.CHECK WITH VENDOR PRIOR TO ORDERING ANY PARTS TO VERIFY PART NUMBERS.

rCE-008a(8004)/1r 25 5.3 CORE EXIT THERMOCOUPLE TEMPERATURE The QSPDS displays individual core exit temperatures with a core map, the highest and next highest core exit temperature in each quadrant, and a representative core exit temperature.

The representative core exit temperature is calculated based on averaging all valid CET inputs with practical checks from other inputs.Using the Representative CET Temperature (average temperature), and pressurizer Pressures the Temperature and Pressure Saturation Margins are calculated and displayed.

However the CET calculated variables will be flagged with a question mark during the time that the QSPDS is identifying out of range and suspicious CET inputs.When the QSPDS completes one (1)complete cycle of the CET algorithm without flagging or removing a flagfromany CET input then the inputs are considered stable, and accurate and the question mark (?)is removed from the calculated variables.

Also the CET calculated variables will be flagged with all question marks, if the number of valid CET inputs is less than 2 CET's per Quadrant per channel.Those inputs considered invalid are either failed or deviated from the mean of the CET inputsbya specific amount.The failed inputs are displayed as all question marks.While the deviated inputs have a question mark in front of the displayed value indicating it is suspicious.Whenevera saturation alarm occurs (either RCS orUpperHead) all previous valid (before the saturation alarm)CET inputs will be used in calculating CET variables.

Previously suspected CET inputs will be used in the calculation if their temperature value fall within a specified band of the mean temperature.

The core exit temperature alarm if the temperature is greater than a high setpoint.However, suspected inputs will not be alarmed.Description No.15381-ICE-3227 Revision 04Page25 of 55 Attachment 4 Question 83: Actionsfora Dropped Fuel Assembly Original Question questions reportfor 2014DraftNRC WRITTEN EXAM (83)QUESTI0N NUMBER: 083Anewfuelbundle has dropped off the FH-1RefuelingMachine grapple onto therefuelingcavityfloorand appearstobe damaged.The followingconditionsexist:

The inner PAL door is being held open by an RP tech for trash removal.*The outer PAL door is open and inoperable.

The Equipment Hatch is installed.

All Area and Process Monitors indicate normal background radiation.

1)Which AGP does the CRS enter, and 2)What is an action the operators are directed to perform?Procedure Action A.AOP-08 Fuel Handling Incident Trip VIAS using both CRHS test switches By AOP-08 Fuel Handling Incident Direct theEONAto close the inner PAL door within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C.AOP-12 Loss of ContainmentIntegrityTripVIASusingboth CRHS test switches D.AOP-12 Loss of ContainmentIntegrityDirect theEONAto close the inner PAL door within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> FCS believes this question has two correct answers.Choices A and B both describe actions that are taken in AOP-08.Although the event occurred in the containment and not theAuxiliaryBuilding, step 8 would eventually be performed for either location.The stem states radiation monitors are not in alarm;thereforeVIASwill not have occurred.IfVIAS has not occurred, the contingency action for step 8 requires manual VIAS actuation making choice A correct.ChoiceBis also an action taken in AOP-8 as asked by the question"what is an action taken." FCS respectfully requests that choices A and B both be accepted as correct.Thursday, April 24,2014 6:19:18 PM 15 AOP-08 Page2of 15 1.0 PURPOSE This procedure provides guidanceinthe eventanirradiatedFuelAssemblyisdroppedor otherwise damaged.2.0 ENTRY CONDITIONSAfuel assembly has been damagedwhichmaybeindicatedbyanyofthe following:

A.Area radiation monitors increase.B."RM-050CNTMTPARTICULATEHIGH RADIATION" alarm(AI-33C:A33C).

C."RM-OSI CNTMT NOBLE GAS HIGH RADIATION"alarm(AI-33C:A33C).D."RM-052STACK/CNTMTNOBLEGASHIGH RADIATION" alarm(AI-33C;A33C).

E."RM-062AUXBLDGVENTSTACKHIGH RADIATION" alarm(AI-33C;A33C).

F.Containment air particulate high radiation indication upscale.G.Ventilation Isolation Actuation Signal (VIAS).H.While handling fuel, the Hoist Load Indicator showslowHoist weight.I.Possible damage to Fuel Assembly is observed.R10 AOP-08 Page4of 15 4.0 INSTRUCTIONS/CONTINGENCY ACTIONS INSTRUCTIONS CONTINGENCY ACTIONS 1.Announce and repeat the following over the plant communications system: "Attention all personnel.

Attention all personnel.Afuel handling accident has occurred in the (location).

All non-essential personnel should immediately evacuate the area." 2.Direct the Radiation Protection Technician to survey the affected area.3.IMPLEMENT the Emergency Plan.4.IF the incident y\ras Jn the Auxiliary Building, THEN GO TO Step 8.R10a INSTRUCTIONS CONTINGENCY ACTIONS AOP-08 Page5of 15 NOTEItIs required to close thefollowingwithin1 hour of event initiation:

Room 66 Roll-up Doors or the Equipment Hatch, all containment penetrations open to the outside atmosphere and one door in the PAL.5.Direct the Shift Outage Manager to close the Equipment Hatchwithin1 hour.6.Direct Shift Outage Manager to close all containment penetrations open to the outside atmosphere within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.7.Direct the EONA to close at least one PAL door within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.5.1 IF the Equipment Hatch can not be closed, THEN perform the following:

a.Direct Shift Outage Manager to close the construction access.b.Direct the Security Shift Supervisor to close Room 66 Doors 1009-1, 1013-1 and 1013-4 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.RlOa AOP-08 Page6of 15 INSTRUCTIONS CONTINGENCY ACTIONS8.VerifyVIAS actuation PER8.1IFVIASdidNOT actuate, Attachment A.VIAS Actuation.

THEN ensure VIAS actuates bv performing the following; a.Obtain a kev from oanei AI-30A orAI-30B.b.TripVIAS using both CRMS test switches.*86A/CRHS Test Switch (AI-30A)*86B/CRHS Test Switch (AI-30B)c.Verify VIAS actuation PER Attachment A, VIAS Actuation.

RlOa Attachment 5 Question 98: SGTR Dose at Site Boundary Original Question questions report for 2014 Draft NRC WRITTEN EXAM (98)QUESTI0N NUMBER: 098 The plant was at 100%powerwithall TS LCDs met when a Steam Generator Tube Rupture occurred.*Off-Site Power has been lost.*Natural Circulation is established.

  • All Safeguards equipment functioned properly.What is the maximum dose thatanindividual at the site boundary could experience in thefirsttwo hours following the accident?A.25 mRem TEDE B.75 mRem TEDE 0.5 REM TEDE D/25 REM TEDE Question 98 asks"What is the maximum dose that anindividualat the site boundary could experience in thefirsttwo hours following the accident?" The question's choices are based on the regulatory limits and not the actual USAR Chapter 14.14 Steam Generator Tube Rupture Accident analysis.The table in USAR section 14.14.4 indicates the actual dose would be 1.0REMfor either a Pre-Accident Spike or a Concurrent Spike condition.

The limits are listed as25REM and 2.5REMfor each condition respectively.

Therefore, FCS believes this question has no correct answer and requests it be deleted.Thursday, April 24, 2014 6:19:19 PM 45 Page1of12 USAR-14.14 Safety Analysis Steam Generator Tube Rupture Accident Rev 15 Safety Classification:

Safety Change No.: Reason for Change: Preparer: Issued: Usage Level: Information EC 51265 The USAR title pages and sections are being reformatted into the current format showing the safety classification and usage level.Additional editorial changes (e.g., font, capitalization, title corrections, etc.)are being made for consistency.

M.Edwards 07-12-11 3:00 pm Fort Calhoun Station USAR-14.14 Information UsePage2of12SteamGeneratorTube Rupture Accident Rev.15 Table of Contents 14.14 License Renewal Supplement 4 14.14.1 General 4 14.14.2 Method of Analysis 6 14.14.3 Radiological Consequences for Steam Generator Tube Rupture Accident 6 14.14.4 Results 9 14.14.5 Specific References 10 USAR-14.14 Information UsePage3of12 Steam Generator Tube Rupture Accident Rev.15 List of Tables Table14.14-1-Radiological Analysis Assumptions and Key Parameter Values-SGTR 11 Table 14.14-2-Integrated Break RowfromPrimary Coolant to Faulted Steam Generator-SGTR 12 Table 14.14-3-Cumulative Steam Releases from Faulted Steam Generator-SGTR 12 Table 14.14-4-Cumulative Steam Releases from Intact Steam Generator-SGTR 12 USAR-14.14 Information UsePage4of12SteamGeneratorTube Rupture Accident Rev.15 14.14 License Renewal Supplement Since the performanceoftheFSAR analysis, the steam generator tube rupture accident has been evaluated using recent computer codes (Reference 14.14-6)which also addresses a release pathway through the steamdrivenauxiliary feedwater pump (FW-10).14.14.1 General The steam generator tube rupture accidentisa penetration of the barrier between the reactor coolant system and the main steam system.Theintegrityofthis barrier is significant from the radiological safety standpoint, as a leaking steam generator tube would allow transport of reactor coolant into the main steam system.Radioactivity contained in the reactor coolant wouldmixwith shell side water in the affected steam generator.

This radioactivity would be transported by steam to the turbine and then to the condenser, ordirectlyto the condenser via the steam system dump and bypass valves.Noncondensible radioactive materials would be discharged through the condenser vacuum pumps to the atmosphere.

FW-10 (Auxiliary Feedwater Pump)Was considered as a release path also.Modification MR-FC-91-039 provided Operations with theabilityto override containment isolation actuation signal(CIAS)from steam generator blowdown (SGBD)sampling isolation valves to draw a steam generator blowdown sample.This modification allows for establishment of an acceptable/analyzed blowdown sampling release path should plant conditions warrant the drawingofa blowdown sample.Appropriate Abnormal Operating Procedures (AOPs)/Emergency Operating Procedures (EOPs)instruct the Operator to reroute the SGBD sampling dischargeflowtoa lineup leading to the radioactive waste disposal system.Diagnosis of the accident is facilitated by radiation monitors in the blowdown sample lines from each steam generator and in the condenser vacuum pump dischargeline.Additionally, the CIAS signal for HCV-2506A/B and HCV-2507A/B can be overriddenfora period not to exceed two hours.This allows samples to be obtained from the steam generatorstoaidin the identification of the affected steam generator.

These monitors initiate alarms in the control room and inform the operator of abnormal activity levels and that corrective action is required;the steam generator blowdown is automatically isolated shouldahighactivitylevel be reached.Additionally, steam generator blowdown is automatically isolated upon reactor trip.

USAR-14.14 Information UsePage5of12SteamGeneratorTube Rupture Accident Rev.15Thebehaviorofthe systemsvariesdependinguponthe sizeoftherupture.Forleak ratesuptothe capacity of the charging pumps in thechemicalandvolumecontrol system (CVCS), reactorcoolantinventory canbemaintainedandan automatic reactortripwouldnotoccur.The gaseous fission productswouldbe released to atmosphere from the secondary systematthe condenservacuumpump discharge.

Those fission products not discharged in this way would be retained by the main steam, feedwater and condensate systems.For leaks that exceed the capacity of the charging pumps, the pressurizer water level and pressure decrease and a thermal margin/low pressure (TM/LP)reactor trip results.The turbine then trips and the steam system dump and bypass valves open.The steam generator water level indicatorsaidin detection of these larger leaks since the water inventory in the leaking steam generator may increase more rapidly than that of the intact steam generator following reactor trip.The amount of radioactivity released increases with break size.For this analysis, an area equivalenttoa double-ended break of one steam generator tube is assumed for the rupture size.At normal operating conditions, the leak rate through the double-ended rupture of one tube is greater than the maximum flow available from the charging pumps;the reactor coolant system pressure decreases andalow pressurizer pressure trip occurs.Following the reactor trip, the reactor coolant average temperature is reduced by exhausting steam through the steam dump and bypass system.The radioactivity exhausted through the steam dump and bypass systemflowsto the condenser where the noncondensible gaseous products are released to the atmosphere.

Based on guidance contained in the EOPs, which serve to minimize releases and off-site doses, the operator will use the steam dump and bypass valves to reduce the reactor coolant system hot leg temperature.

Once the dump valves are closed, operation of the bypass valve (PCV-910)allows steam to pass to the condenser to reduce this temperature to 510°F.When the hot leg temperature is below SIO^F, the affected steam generator is isolated to terminate the release source.This hot leg isolation temperature of 510°F bounds the minimumcoldleg temperature requirements for adequate reactor coolant pump (RCP)net positive suction head (NPSH).This steam generatorhotleg isolation temperature value is consistently used in the EOPs and this analysis, regardless of whether RCPs are available (i.e., off-site power is available).

When the reactor coolant temperature is 300°F the operator is assumed to place the shutdown heat removal system into operation and isolate both steam generators.

Although Shutdown Cooling entry can occur at 350°F,alower entry temperature is conservative for this analysis.

USAR-14.14 Information Use Page6of12SteamGeneratorTube Rupture Accident Rev.15 14.14.2 Method of Analysis The analysisofa steam generator tube rupture was performedusingadigital computer simulation of the system.The simulation includes neutron kineticswithfuel and moderator temperature feedback, the effect of the shutdown group of control element assemblies (CEAs)and the reactor coolant and main steam systems including the pressurizer, steam generators, and steam dump and bypass valves.The method of analysis used provides radiological consequences results that bound operator actions that may be taken in accordance with the EOPs (Reference 14.14-7).This results from the analysis assumption that the operator will feed and steam both steam generators for cooldown, for thefirsttwo hours.Cooldown to less than 51CF would occur much sooner than this when using the EOPs.14.14.3 Radiological Consequences for Steam Generator Tube Rupture Accident The Steam Generator Tube Rupture (SGTR)Radiological Consequences assessment followed the guidance provided in Regulatory Guide (RG)1.183 with exceptions noted in Section 14.1.6 and discussed in Reference 14.14-4.Table 14.14-1 lists the assumptions/parameters used to develop the radiological consequences following SGTR and were documented in References 14.14-4 and 14.14-10.Two cases of thermal hydraulic dataofa postulated SGTR were generated for input to the radiological consequences evaluation (Reference 14.14-11).

One case assumed the rupture tube is at the top of the tubesheet on the cold side of the steam generator and is called the cold side break case;the other case assumed the rupture is at the top of the tubesheet on thehotleg side of the steam generator and is called the hot side break case.Due to the assumed simultaneous loss of offsite power with the reactor trip, the reactor is cooled down by releasing steam via the main steam safety valves (MSSVs)/atmospheric dump valve (ADV).The primary coolant with elevated iodine concentrations (pre-accident or concurrent iodine spike)flowsinto the faulted steam generator and the associated activities are released to the environment due to the secondary side steam releases.Before the reactor trip, the activities are released from the main condenser air ejector (AEJ).After the reactor trip the steam releaseisvia theMSSVs/ADV.Aportionofthis steam is released via the turbine exhaust of the turbine driven auxiliary feedwater (AFW)pump.However, the atmospheric dispersion factorofthis releasepointis bounded by that of the MSSVs/ADV (see Section 14.1).Consequently, the dose analyses conservatively assumes thatallof the steam is discharged via the MSSVs/ADV.

The spiking primary coolant activities leaked into the intact steam generator at a conservative 1 gpmprimaryto secondary leakage rate are also released to the environment via secondary steam releases.

USAR-14.14 Information UsePage7of12SteamGeneratorTube Rupture Accident Rev.15 Both the cold side break case and the hot side break case were evaluated.

The doses from the worst case are calculatedindetail and are reported in the results.The most critical parameter that affects the doses is the total break flow that flashes from thetimeof the reactor trip until theisolationoftheaffected steam generatoratt=2hours.Thisdirectly released primary coolant for the hot side break is greater than the cold side break.Additionally,thetotal steam releasefromtheintactSGis essentially the same for the two break cases.Therefore, the hot side break case will produce greater radiological consequences and is evaluated and described in detail.Since thereisno postulated fuel damage associated with this accident, the main sourceofradioactivityis theactivityin the primary coolant system.Two spiking cases were addressed:

a pre-accident iodine spike and a concurrent iodine spike per RG 1.183.a.Pre-Accident spike-the initial primary coolant iodineactivityis assumed to be 60 pCi/gm of DEI-131,whichis the transientTSlimitfor full power operation.

The initial primary coolant noble gasactivityis assumed to be at TS levels.b.Concurrent spike-the initial primary coolant iodineactivityis assumed to be at TSof1 pCi/gm DEI-131 (equilibrium TSlimitforfull power operation).

Immediately following the accident, the iodine appearance rate from thefuelto the primary coolant is assumed to increase to 335 times the equilibrium appearance rate corresponding to the 1ljCi/gm DEI-131 coolant concentration.

The duration of the assumed spike is eight hours.The initial primary coolant noble gasactivityis assumed to be at TS levels.The initial secondary side liquid and steamactivityis relatively small and its contribution to the total dose is negligible compared to that contributed by the rupture flow and is therefore, not considered in this assessment.

USAR-14.14 Information Use Page8of12SteamGeneratorTubeRupture Accident Rev.15 Faulted SG Release Per Reference 14.14-4 a postulated SGTR will resultina large amount of primary coolant being released to the faulted steam generator via the break locationwitha significantportionofit flashed to the steam space.The noble gases in the entire break flow and the iodine in the flashed flow are assumed immediately available for release from the steam generator without retention.

The iodine in the non-flashed portion of the break flow mixesuniformlywith the steam generator liquid mass and is released into the steam space in proportion to the steaming rate and partition factor.Before the reactor trip at 629.5 seconds, the activities in the steam are released to the environment via the main condenser air ejector.All steam noble gases and organic iodine are released directly to the environment.Onlya portion of the elemental iodine carried with the steam is partitioned to the air ejector and released to the environment.

The rest is partitioned to the condensate, returned to both steam generators and assumed to be available for future steaming release.After the reactor trip, the break flow continues until the primary systemisfully depressurized and in equilibrium with the secondary side of the faulted SG.To maximize the calculated off-site doses, the condenser is assumed unavailable after reactor trip.The steam is released from the MSSVs/ADV.

All activity releases from the faulted steam generator cease whenitis isolated at 120 minutes after the accident.Table 14.14-2 provides the breakflowofprimary coolant into the faulted SG and Table 14.14-3 provides the steam releases from the faulted SG as a function of time (References 14.14-4 and 14.14-11).

Reference 14.14-11 also provides the integrated break flow that flashes in the faulted SG.Intact SG Release The activity release from theintactSGis duetonormalprimaryto secondary leakage and steam release from the secondary side.Aportionof the break flowactivitythatis transferredtotheintactSGvia the condenser before reactor trip will also be released from the intact SG during thecooldown phase.Theprimaryto secondary leak rate is conservatively assumed to be 1 gpm at STP.Steam generator tube uncovery occursatt=679.5 secondsforaperiodof 112 minutes (Reference 14.14-12).

During this tube uncovery period, the halogensinthe flashedportionoftheprimaryto secondary tube leakage are assumed immediately available for release to the environment.

The flash fractions associated with the SG tube leakage during the tube uncovery period are provided in Reference 14.14-11.

USAR-14.14 Information UseSteamGeneratorTube Rupture Accident Page9of12 Rev.15 The halogensinthe non-flashedportionofthe leakage mixes uniformly with theSGliquid mass and is released into the steam space in proportion to the steaming rate and thepartitionfactor.

When the tubes are totally submerged, the halogens in the tube leakage are assumedtomix uniformly with the SG liquid and are released as discussed above for the non-flashedportionof the tube leakage.The noble gases are releasedfreelyto the environment without retention in the SG for the entire duration of the event.The steam releases from the MSSVs/ADV continue for 144.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, at which time shutdowncooling(SDC)isinitiatedvia operation of the SDC system and environmental releases are terminated.

Table 14.14-4 provides the cumulative steam releases from the intact SG (Reference 14.14-4 and 14.14-11).

Per RG 1.183, the 2-hour EAB dose must reflect the"worst case" 2-hour activity release period following the accident.The worst 2-hour EAB dose is expected to occur during thefirsttwo hours of the event because it coincides with the breakflowin the faulted SG and tube uncovery in the intact SG.Also, the noble gas release rate is at its highest at the onset of the event.Reference 14.14-10 shows that the maximum 2-hour EAB dose occurs duringt=0to2 hours for both the pre-accident and concurrent iodine spike cases.Accident Specific Control Room Assumptions The SGTR will resultina safety injection actuation signal (SIAS), 678.5 seconds into the event, which will result in the initiation of the CR emergency ventilation.

An additional delay of 44 seconds was assumed to accountfora coincident LOOP.Due to single failure of the recirculation damper, the emergency recirculation filtration system is assumed to be unavailable for thefirst2 hours after the event.The remaining CR parameters were discussed in Section 14.1.14.14.4 Results The total^fPeetive dose equivalent (TEPE)^(^vafipus jo^tipns are shown below.Pre-Aocident Spike (Rem)R^g yinii Concurrent Spike (Rem)Reg Limit EA-B (2-hour dose)1.0 25 1.0 2.5 LPZ (for duration of event)0.5 25 0.5 2.5 Control Room (30-day dos#1.0 5.0 1.5 5.0 The results remain below the regulatory limits set by 10 CFR 50.67 and RG 1.183.

USAR-14.14 Information UsePage10of12 Steam Generator Tube Rupture Accident Rev.15 Doses were roundedupto the nearest 0.5 Rem.The maximum 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doseperiodfor the EAB dose was in the0to2 hours time period.14.14.5 Specific References 14.14-1 Not Used 14.14-2 Not Used 14.14-3 Not Used 14.14-4Applicationfor Amendment of Operating License, Updated Safety Analysis Report Revision for Radiological Consequences Analysis for Replacement NSSS Components, LIC-05-107, October 26, 2005, OPPD to USNRC Document Control Desk 14.14-5 Not Used 14.14-6 Letter from OPPD (R.L.Andrews)toNRC(J.M.

Taylor), dated April 10,1987 14.14-7 Response to Fort Calhoun Station Condition Report 199700976 ActionItem1 (including 10CFR50.59 evaluation), April 1998 14.14-8 OPPD-NA-8303-P.

Revision 1, Omaha Public Power District Reload Core Analysis Methodology:

Transient and Accident Methods and Verification, January 1993 14.14-9 Letter NRC-93-0301 from NRC (S.D.Bloom)to OPPD (T.L.Patterson), dated November 2, 1993 14.14-10 OPPD Calculation FC06820, Revision 1, Site Boundary and Control Room Dosefollowinga Steam Generator Tube Rupture Accident Using Alternate Source Terms 14.14-11 AREVA Document 32-5051831-01, FCS RSG-Steam Generator Tube Rupture with Cooldown to SDC Entry Conditions, March 11, 2005 14.14-12 AREVA Document 32-5048718-02, FCS RSG: Inputs for Radiological Consequences Analysis-Tube Uncovery, March 30, 2005 USAR-14.14 Information Use Steam GeneratorTubeRuptureAccident Page 11 of 12 Rev.15 Table 14.14-1-Radiological Analysis Assumptions and Key Parameter Values-SGTR Power Level Reactor Coolant Mass Break Flow to Affected Steam Generator Time of Reactor Trip Termination of Release to Affected SG Amount of Break Flow that Flashes Leakrate to intact SG Failed/Melted Fuel Percentage RCS TS Iodine Concentration RCS TS Noble Gas Concentration RCS Equilibrium Iodine Appearance Rates Pre-Accident Iodine Spike Activity Accident Initiated Spike Appearance Rate Duration of Accident Initiated Spike Secondary System Release Parameters Intact SG Liquid Mass (post-accident minimum)Faulted SG Liquid mass (post-accident minimum)Initial Mass in SGs (total)Form of All Iodine Released to the Environment via SGs (total)Time period of Tube Recovery (Intact SG)Iodine Partition Coefficient (unflashed portion)Fraction of Iodine Released (flashed portion)Fraction of Noble Gas Released either SG Partition Factor in Condenser AEJ Cumulative Steam Releases from Faulted SG Cumulative Steam Releases from Intact SG Steam Flowrate to Condenser from Faulted SG Before Trip Steam Flowrate to Condenser from Intact SG Before Trip Termination of Release from SGs Environmental Release Points 1530 MWt 250,000 Ibm Table 14.14-2 629.5 sec 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Reference 14.14-11 1 gpm at STP 0%Table 14.1-6 Table 14.1-6 Table 14.1-7 Table 14.1-7 335 Times Equilibrium 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 45,708 Ibm 70,261 Ibm 157,185 Ibm 97%elemental; 3%organict=679.5 sec tot=7399.5 sec 100 (all tubes submerged) 1.0 (Released to Environ w/o holdup)1.0 (Released to Environ w/o holdup)2000 (elemental iodine), 1 (organic iodine and noble gases)Table 14.14-3 Table 14.14-4 944.8 Ib/s (0-629.5 sees)940.3 Ib/s (0-629.5 sees)144.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Condenser Evac.Discharge (0-629.5 sees)MSSVs/ADV (629.5 sees-144.6 hr)CR Emergency Ventilation:

Initiation Signal/Timing Initiation time (SIAS)678.5 see USAR-14.14 Information Use Page12of12 Steam Generator Tube Rupture Accident Rev.15 Table 14.14-2-Integrated BreakFlowfrom Primary Coolant to Faulted Steam Generator-SGTR Time Integrated Break Flow (sec)(lbs)0 0 629.5 26852.7 722.5 30048.1 3600 119070.9 7200 284570.7 Table 14.14-3-Cumulative Steam Releases from Faulted Steam Generator-SGTR Time MSSV Release (sec)(lbs)0 0 629.5 0.0 722.5 7338.8 3600 106262.6 7200 146230.0 Table 14.14-4-Cumulative Steam Releases from Intact Steam Generator-SGTR Time MSSV Release ADV Release Total Release (hr)(lbs)(lbs)(lbs)0 0.6736 1 2 4 8 24 30 144.6 0 0.0 24686.0 64111.8 104000.3 171279.7 365063.0 425470.0 0 0.0 0.0 0.0 69221.8 186449.0 524165.6 629425.0 3080000.0 0 0.0 24686.0 64111.8 173222.1 357728.7 889228.7 1054895.0 3505470.0

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