IR 05000387/1985025
| ML20198A822 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/21/1985 |
| From: | Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198A813 | List: |
| References | |
| 50-387-85-25, 50-388-85-27, NUDOCS 8511060190 | |
| Download: ML20198A822 (47) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT
I FACILITY DOCKET NOS.: 50-387/50-388 i
FACILITY LICENSE NO.: NPF-14; Construction Permit - CPPR-102
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LICENSEE: Pennsylvania Power & Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 FACILITY: Susquehanna 1 and 2 EXAMINATION DATES: August and 7, 1985
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CHIEF EXAMINER: /I
/p. Lange, Reacter
[ gineer Examiner
/#1/ M Date REVIEWED BY: h//
R. M. Keller, Chi 6f, Projects Section 1C
/ /0/41 [Q{
' Da nlA APPROVED BY:
H. B. Kisie
>Mb hief, Projects Branch NoT 1 h / /}[
/ Da,fe'
SUMMARY: This report transmits the results of the Operator Licensing Examina-tions administered at the Susquehanna Steam Electric Station on August 6-7, 1985. All seven (7) Reactor Operator candidates successfully passed the oral,
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simulator, and written sections of their respective examinations.
b 8511060190 851022 PDR ADOCK 05000387 G PDR l
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REPORT DETAILS TYPE OF EXAMS: Replacement X EXAM RESULTS:
1 RO l l Pas / Fail l l l l l l l Written Exam I 7/0 l 1 I I I I I l Oral Exam I 7/0 l l I l l l l I Simulator Exam l 7/0 1 I l l l l l l Overall l 7/0 l l- 1 I Chief Examiner at Site: D. Lange, USNRC Region I Other Examiners: M. King, USNRC (Contractor) Pacific Northwest Laboratory
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j Summary of generic strengths or deficiencies noted on exams:
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All candidates demonstrated good teamwork and communication during the Simulator exams. An overall strength was noted by both examiners in the l candidates ability to respond to abnormal situations that involved the use l of prescribed procedures.
4 The candidates familiarity with in plant components was also noted as a i strength, i
, Personnel Present at Exit Interview:
i l NRC Personnel
! D. Lange, Chief Examiner, USNRC Region I
! L. Plisco, Resident Inspector, USNRC Region I I
j NRC Contractor Personnel
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M. King, BWR Examiner, EG&G Idaho
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Facility Personnel i
i T. Markowski i J. Seek
H. Palmer
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A. Fitch T. Logsdon
- W. Lowthert I
G. Ward j J. White T. Crimmins M. Peal Attachments: Written Examination and Answer Key (RO) Facility Comments on Written Examinations Made After Exam Review with NRC Resolution
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l}Ti/ICHMENT 1
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II . R. HilCl FAR REGill A10RY COMMISCTON REACTOR OPERATOR LICENGr EXAMINATION FAFIlITY: SilSollFHANNA 1&?
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REACTOR TYPE: BWR-GE4
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DATE ADMINISTERED: 05/08/06
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EXAMCNER: LANGE, _________________________
APPLICANT: _
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INSTRUCTIONS TO APPLICANT:
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Ilse separate paper for the answer Write ansoers nn one side nn) Staple question sheet on top of the answer sheet Points for es cie question are indiested in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers ei)) be pieked up six (6) hours after the examination start % OF CATEGORY % OF APPLICANf'S CATEGORY VALUF TOTAL SCORF V Al ltF CATEGORY
________ ______ ___________ ________ ___________________________________
?5.00 25.00
________ ______ ___________ ________ PRINCIPtFG OF HilClFAR POWER PLANT OPERATION, THERM 00YNAMTC HEAT TRANCFER AND FtUID FLOW 25.00 25.00
________ ______ ___________ ________
?. PLANT DFSIGN INCLUDTNG GAFF 1Y AND EMERGENCY GYSTEMS 25.00 25.00
________ ______ ___________ ________ INGTRUMENTS AND CONTR0tG
_ _1_ __ _ l__ ___________ ________ PROCEDURFC - NORMAL- ADNORM Al ,
EMERGENCY AND RADIOLOGICA CONTROL 100.00 100.00 TOTALS
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FINAL. GRADE _________________%
All work done on this examination is my own. 3 have neither given nor roccived ai ~
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' PRINCIPLFC OF Ni3CLFAR POWFR PLAN 1 OPERA 1 ION, PAGE ?
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OllFOTION 1.01 (?.?5)
Susquehanna's reactor's are operated within three (3) specified Thermal Limits. List each of the three limits and explain the specific heat trans-fer related problem that the limit protects agains (?.?S)
GUESTION 1.02 (2.00)
You are the Reactor Operator in the Control Room. An instrument technician has just completed calibration work on the RCIC pump suction pressure i n s t r uaie n t . Dur2ng a review of your pancis you notice that RCIC pump suction pressure reads 2 psig. This is not a normal reading for a normal standby lineup. Calculate what RCIC pump suction pressure should read given the following data available to you in the Control Room: Verified water level in the CST = 15 fee ?. The difference in elevation between the base of the CSI and RCIC pump suction = 50 fee . The RCIC system is in a normal standby lineu (Note: Show all uork used for your calculations)
4tlFST10N 1.03 (?.00)
How and why does control rod worth vary for the following changes (Note - address both local and core average flux changes in part b)
A. As moderator temperature increase (1.0)
B. As the position of an adjacent rod is change (1.0)
UUESTION 1.04 (7.00)
Using the attached Steam fables, calculate a reactor cooldown rate for a reactor pressure decrease from 800 psis to 700 psis in ONE hour. SHOW ALI. WOR (mmr** CA1FGORY 01 CONTINUFD ON NEX1 PAGF xxrrr)
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~ PRINCIPLES OF NUCLEAR POWER Pl. ANT OPERATIO PAGE 3
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00ESIION 1.05 (2.25)
Describe HOW and WHY eentrifugal pump discharge flow is affected for the following?
A. Suction pressure increas (0.75)
B. Svetion pressure decrease.(Assume a continuous decrease) (0.75) Throttling down on discharge valv (0.75)
GUESTION 1.06 (1.50)
Assume your reactor has been operating at rated power for two weeks and then scramsi a. What reactor period could you expect to see after the initial prompt drop in reactor power? (0.50)
b. Explain your anseer for part (a). (1.00)
00ESTION 1.07 (2.00)
Define condensste depression and briefly explain why EXCESSIVE condensate depression is undesirable in the main condenser .
(7.00)
GUESTION 1.08 (2.00)
list five (5) heat inputs and three (3) heat outputs that vould be used to calculate a heat balance at your plan (?.00)
(*xxxx CATEGORY 01 CON INUCD ON NEXT PAGE xxxxx)
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~ PRINCIPL ES OF NUCL EAR POWER PLANT OPERATION, PAGF 4
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OllFCIION 1.09 (?.00)
Your reactor has just scrammed from extended full power operatio Ten (10) hours later cooldown is complete, and the SDh is determined to be 1% dk/k, since all rods did not inser FXPLAIN the changes to the SOM AND any possible adverse consequences for the next 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> OllFCIION 1.10 (?.00)
For the following events listed below , identify which react 2vity coef-ficient would first cause power to change. Indicate in your answer tho direction of change, (inc. nr der.), for each coefficien . Reactor feed pump tri ?. A single safety relief valve lift . Turbine stop valves closo , (no scram).
4. Control rod drop acciden (0.50 for each correct ans.)
OUESTION 1.11 (2.00)
Concerning control rod vorth during a reactor startup with 100% peak Xenon versus a star tup with Xenon free conditions, WHICH STATEMENT IS CORRECT?
lHGT]FY YOUR CHOIC PERIPHERAL control rnd worth will be 10WFR during the 100% peak Xenon startup than during the Xenon free startu CENTRAL control rnd porth will be HIGHER during the 100% peak Xenon startup than during the Xenon free startu PFRIPHFAI control rod worth will be H]GHER during the 100 %
PEAK XENON startup than during the XENON FREE startu ~ E10TH CFNTRAL and PERIPHERAL control rod worths will be the SAME regar dless of core xenon concentratio (***** CATFCORY 01 CONTINUED ON NEXT PAGF *****)
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GUESTION 1.12 (2.00)
The plant is operating at 90% power. Fxtraction steam to the HP feed water heaters is shut off. A visitor, observing that turbinn load has increased by ?O MWF after extraction steam was shut off, concludes that this action has improved the plant's thermodynamic officienc *
Do you agree or disagree? Fnplain using relevant plant indications to support your positio (2.00)
1 OllFGTION 1.13 (1.00)
Indicated reactor vater level at 100% power differs from the actual water level above the core (that which is present in the steam separstors or within the dryer skirt). Which level (actual or indicated) is higher and by how many inches? (0.50) Explain why tho above difference occur (0.50)
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OUESTION 2.01 (2.00)
How is the integrity of FCCO piping inside the reactor vessel verified during normal operation (include sensing points, specific system (s) who's piping is verified, why its verifiede and response of the instrumentation to a loss of integrity in your answer)? ( 2.00)
DilFGTION ?.0? (3.00)
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Concerning the Reactor Water Cleanup system i
- a. What two conditions will cause the Drain Flow Regulator control valve,(F033). to shut and in each case why does it shot?
(1.00)
b. What is the purpose of the restricting orifice in the blowdoun line and when may it be bypassed? (1.00)
c. List four (4) leak detection signals that will close F001 A F004. (1.00)
4tlFC170N ?.03 (3.00)
With regard to the Control Rod Drive Hydraulic System efollnwing a Reactor Scrami s. What causes, sbe specific), the on-line Flow Control Valve (FCV)
to shut and why is this response desirable? (1.50)
b. What would be the consequences of an HCll scram inlet valve sticking shut with the scram outlet valve open? (Consider reactor pressure (1) high, ' 400 psis, and (?) low, < 400 psig, in your answer.) (1.50)
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' Pl. ANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7
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GUESTION 2.04 (2.50)
Regarding the Standby Gas Treatment (COGT) System ; ) Other than a lack of power available, what are 3 of the 6 l remaining interlocks which will cause the selected ' auto' SDGT train to fail to starte or trip after starting, following an asito initiation signal ? (1.50)
B. What system interrelationship does the SCGT system have with the following two systems i 1. HPCI sy ?. Main Steam sy (1.00)
GUESTION 2.05 (2.75)
A. Will the fuel pool cooling mode of RHR be able to maintain the fuel pool temperature beloo !?5 deg during emergency heat load (EHl.) conditions without the Fuel Pool Cooling Syste (0.50)
O. Fuel Pool Cooling and Cleanup is desi gned t o maintain clar i ty
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and quality in what four (4) systems / areas? (1.00)
C. What five emergency water makeup snurces are availabic should the fuel pool require makeup due to evaporation or leaks ? (1.?5)
fiUFGTION ?.06 (P.50)
A. LIST FIVE (5) plant systems that may be cross-connected between Unit 1 and Unit (2.00)
B. WHO may AU1HORI7F the cr oss -connecting of systemsi (0.5)
GUESTION 2.07 (3.00) LIGT FIVF (5) systems or major parts of DIFFERENT systems that may be operated from the Unit i Romote Shutdown Panal (1C201).
OF SPECIFI (P.5) HOW would the Control Room Operator know if a transfer switch on 1CP01 was placed in FMERGENCY? (0.5)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
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GllFST 3 0N 7.00 (3.00)
A jet pump failure occurs on your shift. For the t'ollowing plant parameters indicate whether an INCREAGE or DECREASE 9ill occur and briefly explain WH . Recirculation flo . Actual core flo . Thermal Powe . Generator electrical outpu . Indicated core flo . Core delta (0.50 for each correct)
QUESTION 2.09 (2.00) Which of the following too transients would be more severe to the reactor core ? (briefly explain). (1.00)
1. A 65% DBA. 10CA 2. A 100% 00A 1.0CA B. Which of the following failures vould be more severe during a DDA LOCA ?
(briefly c:: Plain).
1.1 PC T failur (1.00)
2. CORE SPRAY failur FSTION ?.10 (1.?5)
SSES conducted a star tup test that entailed a loss nf offsite power to Unit-2. fhe Diesel Generators failed to star . What caused the Diesels failure to start ? (bves)
2. What has been done to assure that this problem does not reoccur ? ,1(
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OllFSTION 3.01 (3.00)
How is total core flov indication obtained ? Your answer should provide the following informationi 1. What is being physically measured and where ? (1.00) What signal conversion and/or conditioning is done to obtain the final core flow indication ? (1.00) What provision is made in t..e circuitry to assure accuratn flow indication eith only one recirculation loop operating and hoe does it perform this function ? (1.00)
I.'UFC130N 3.0? (?.50)
Concerning the Automatic Depressor i: ation System (ADC) logic i A. What is the power supply for the logic? (0.50)
D. Why is the reactor vessel low level contacts in the logic sealed in upon satisfying all conditions for auto A09 intttatio (1.00)
C. How are tho Core Spray and RHR pump running per missivo for AOS sensed (including setpoints)?
(1.00)
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QUESTION 3.03 (2.50)
Assume a Scoop Tube 1orFop on Recire. HG set 'A' orenred on your shift with the plant oper ating at r ated conditionsi List two (?) conditions which could have automatically caused the locku (1.00)
B< Why is it important to reset the Sconp Tube lockup as soon as
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possible? (0.50)
C. What ? indications alert the reactor operator that the conditions which caused the lockup have cleared ?
(1.00)
(xxxxx CAIEGORY 03 CONTINUE 0 ON NEXT PAGE xxxxx)
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OLIFST3DN 3.04 (?.50)
A. What conditions will cause a ROM to be bypassed ? (1.00)
B. Which LPRH detectors in each LPRh string are monitored by RBH channel (0.50)
C. How are the APRM's used by the ROM ? Your answer should include which APRM's and how they are being use (1.00)
OUESTION 3.05 (2.00)
Co.scerning control room narrow range level indscation (0-60') response!
Answer the following either increase, decrease or remain the sam A. The level transmitter bellows ruptures while actual vessel level remains constant? (0.50)
B. The source of power to the control room indicator is lost while actual vessel level is increasing? (0.50)
C. The reactor is cooled down from normal operating temperature to cold conditions while actual vessel level remains constant?
(0.50) The variable leg ruptures?
(0.50)
00ESTION 3.04 (3.00) Unit 1 is in Shutdown Conling en Division 1 RHR vith the
'C' pump running. While conducting a startup test, the lini t ? PCO depressed the o.anual initiation button f or linit ?
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Division 1 RilR . WHAT pumps will start and stop on DOTH UNITS ?
62303 EXPL AI (1.50) To ensure Unit I will have full compliment of ECCS pumps in the event of a iOCA, WHICH pumps on linit ? wonld yon USE for Shutdown Cooling ? (0.75) FXPLAIN the purpost for the LOCA/ False 10CA togi (0.75)
(***** CATEGCRY 03 CONTINilED ON HEXT PAGE *****)
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l GilFSTION 3.07 (3.00)
With regard to Reactor Recirculation Control: WHAT plant cor.ditions/ events vill place the 4? speed limiter
- in control? DE SPECIFI (1.5) With the 'B' recire MG controlled P 75% in Master Manual, the generator tachometer output fails to zero due to an electrical fault. FXPLAIN HOW this fault will effect the 'D' recite M BE SPECIFI (1.5)
Ol!FSTION 3.00 (3.00)
With the linit operating at 75% power, an electrical fanit c a u s o =, t. h e Maximum Cos.bined Flow Setpoint to drop to minimum. HOW W II. I . the fo1109ing parameters RFSpnND after the fanit and WHY? Consider th2cr response for ONE MINUTE following the fault. Assum? NO OPI.RATOR ACTIO Attached FIGilRE 8, EHC logic. is provided for referenc a. Turbine control valvo position (1.0) i b. Dypass valve position (1.0)
t Reactor power a pressure (1.0)
l OllFSTION 3.07 (1.?S)
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What signals / conditions must be satisfied to i ni t i a te Pl ant Ao;<i l l a ry Load Sheddin (1.25)
Ot!FSTION 3.10 (?.75)
Concerning a Stuck Open or Leaking Relief Valvel A. List five (5) operational parameters yno vould check to verify that a relief was leaking, (1.25) How would a stuck open Relief Valve initially effect Feedwater Temperature. (increase or decrease and why). (1.00)
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QUESTION 4.01 (3.00) LIST the entry conditions and setpoints for EO-100-021 ' Level Control'. (1.5) After an ECCS automatic initiation, it may be placed in manual or secured if adequate core cooling is assured OR ____________
______________________________. (fill in blank) (0.5) Adequate core cooling is assured if any one of three conditions are satisfie LIST the THREE (3) condition (1.0)
GUESTION 4.02 (2.00)
Unit 1 is operating at 90% power when a drop in generator MWE is noted, and 'hain Steam SRV Open' is annunciated i A. If Reactor Recirculation flow control is in NASTER MANUAL will reactor power increase , decrease , or remain the same ? (0.50)
8. How could you determine which safety relief is open ? (0.50)
C. What is your major concern during this event and why ? (1.00)
OUEST10N 4.03 (3.00)
As time permits, WHAT TEN (10) actions will be performed prior to Evacuating the Control Room, according to EO-100-009? (3.00)
GUESTION 4.04 , , , , , (2.00)
According to EO-002, Loss of Instrument Air, at what point should you scram the reactor and why ? (2.00)
(***** CATECORY 04 CONTINUED ON NEXT PAGE *****)
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GUESTION 4.05 (1.50)
Concerning a Plant Shutdown from outside the control room, answer the following either true or false. (if false briefly explain why). Transfer of RHR to the remote shutdown panel defeats the automatic initiation of 'BOTH' RHR loops in the LPCI mode ? v"J-#' 'hLg (0.50)
?. The RCIC high level trip at Icvel 8 is bypassed when cont. rolled from the remote shutdown panel ? (0.50) If the red ,pover available , light for the RCIC topar i nve r t.c r is not on at the shutdown panel, RCIC is prevented from being manually started ? (0.50)
GUESTION 4.06 (2.00)
Concerning the RHR SYSTEM i Prior to starting an RHR pump. in the shutdown cooling mode, procedure OP-149-002 cautions you on a flow and time requirement once the pump has started. Furlain the reason for these requirements ? (1.00) When operating thc RHR system in the supression pool cooling modo, on Unit- 1 , why as it desircable to use RHR pumps I P -- ? O ? , C and D ?
(1.00)
QUESTION 4.07 (1.50)
When removing extraction steam from a feedwater heater string, procedure OP-147-001 lists several precautions. Briefly e:: plain these precaution . The highest numbered F.W entraction supply should be removed firs (0.50) Removal of extraction steam supply from more than onn operating heater string is prohibite (0.50) heaters 1 & 2, AOC extraction steam supply and F.W. heater 1-ABC and drain cooler 6-ADC cannot be isolated eithout a turbine trip. (0.50)
(***** CATEGORY 04 CONTINUED ON HEXT PAGE *****)
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- . PROCEDURES - NORMAL, ABNORMAL, FMERGENCY AND PAGF 14
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QUFDTION 4.08 (3.00)
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Concerning procedure FO-030, Unit-? Response To Station Blackouti 1. List the entry conditions . (1.00)
?. What level indications are available to you during this transient ?
(1.00)
3. How is vessel level controlled durin3 a station blackou (0.50)
4. At what point, and what must be done, 'o restore power to the full core displa (0.50)
GUESTION 4.09 (2.00)
Complete the following chart of Operational condition MODE SWITCH AVE. REACTOR CONDITION POSITION COOLANT IfM _________ __________ _____________
1. Power Operation Run Any Temperature 2. Startup (a.) (b.)
3. Hot Shutdown (c.) (d.)
4. Cold Shutdown (e.) (f.)
5. Refueling (3.) (h.)
(0.25 for each correct ans.)
GUESTION 4.10 (1.00)
Conterning Procedure GO-100-003, (Pover Operation) f A. Why is Jet Pump operability a concern prior to execeding 25 % of rated power ? (0.50)
D. Why is the Turbine Load Set limited to 100 MWe above actual Inad until actual load is reached ? (0.50)
(***** CATEGORY 04 CONTINUED ON NEXf PAGE xxzza)
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QUESTION 4.11 (2.00)
According to procedure AD-GA-309, Primary Containment Access a Controli What criteria must be satisfied , concerning containment penetrations required to be closed during accident conditions, to be assured of Primary containment integrity ?
OUESTION 4.12 (2.00)
Using the attached suppression pool heat capacity figures from procedure EO-100-023, Containment Control, determine the minimum suppression pool water level allowable given the following plant condition . RPV pressure = 500 psig 2. Suppresion pool temperature = 175 deg. i40TE: shcw all calculations
(***** END OF CATEGORY 04 *****)
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EQUATION SHEET
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f = :na v = s/: Cycle efficiency = (Nec work out)/(Energy in) ,
o j , = mg s = Vg t + 1/2 at-i = mc- -
,
KE = 1/2 mv
3 , (yf ,.j 9)ft 4 , x3 ; , g o,-At '
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PE = mgn Vf=V - at 4 = e/t x = tn2/t1/2 = 0.693/t1/2
- W = v :.P
~
n0 2 t
1/2 8 *bd A= 4 [(c/2I l * (*0I3 i .E = 931 sn -
""V n Ao -T.x l ,
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Q = mCpat
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6 = UAa7 I = I ce~" '
Pwr = Wfah I = I, 10-*/II'
T/L = 1.3/u l P = P 10 5U"It) o HVL = -0.693/u ,
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P = Po e*/'
- SUR = 26.06/T SCR = S/(1 - K,ff)
l CR x = S/(1 - K,ffx)
SUR = 2So/t* + (s - o)T CR j (1 - K,ffj) = CR2 (I ~ keff2)
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T = ( t*/c ) + [(s - o '/ To] M = 1/(1 - K,ff) = CR /CR j 3 l 7 = 1/(o - a) M = (1 - K,ffo)/(1 - K,ff j)
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T = (5 - o)/(Io) SDM = ( -K,ff)/K,ff
! o = (Keff-1)/K ,ff = aK,ff/K eff t* = 10 seconds
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I = 0.1 seconds-l o = [(t*/(T K,ff)] + [i,ff /(1 + IT)]
Id jj=Id P = (t*V)/(3 x 1010) Id jj 2 =2Id 22
2
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I = cN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (feet) ,
Water Parameters Miscellaneous Conversions 1 gal. = 8.345 10 I curie = 3.7 x 1010 eps i 1 ga: . = 3.78 liters 1 kg = 2.21 lem J l ft* = 7.48 ga I np = 2.54 x 103 Stu/nr
! Density = 62.4 lbqi/f t3 1 mw = 3.41 x 106 Stu/hr
! Density = 1 gm/cv lin = 2.54 cm I Heat of vaporization = 970 Stu/lom 'F = 9/5'C + 32 i Heat of fusion = 144 Stu/lbm 'C = 5/9 ('F-32)
i 1 Atm = 14.7 psi = 29.9 in. H BTU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/in.
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- PRINCIPLES OF NUCLEAR POWER PLANT OPERAfIO PAGF 16
---~isEss55isisiCs, sEAi iRAsSFER Aso FtUi6 FD15
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ANSWERS -- SUS 00EHANNA 182 -05/08/06-LANGE, ANOWFR 1.01 (?.75)
1. MCPR protects against the onset of transition boilin ( .75 )
2. LHCR protects against exceeding 1% plastic strain on the clad due to excessive heat generation i the fue ( .75 )
3.MAPLHGR- ensures that peak fuel clad temperature will not exceed 2?00 degrees F dur ing a DDA-L OC ( .75 )
REFERENCE GGFS, Reactor Theory , Thermal limits., SC073-3 ANSWER 1.02 (2.00)
.
Total head of 9ater at PCIC pump suction!
15' + 50' = 65'
From data shect! I ft. i4 a t e r = .43351bf/in-in (65 ft)(.4335 lb f / t re-in-f t ) = 70.10 psia 28.10 - M7 = 1Y5 psiq rg m er/ (3.0)
REFFRENCE 5*# '# ' /
SESS- Thermodynamics and Fluid Flov .
_
.
-
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 17
-
~~~~YUERUU5YUAbiUS ~UEAi~ FRAU 5fER UU5 FLUi5 FG 4
____________________________________________
ANSWERS -- SUS 00EHANNA 187 -05/08 /06-L A!!GE , ANSWFR 1.03 (?.00)
Rod Worth = (i f 4 Ls) x (Orod/0 avg) squared (Equation not required for full credit)
A. As temperature increases, thermal diffusion length (.33) and slowing down length increases (.33) (as moderator density decreases)
thus rod worth increases (.33) (1.0)
B. As an adjacent rod moves it will change both the local flun around the red and the core average flux (.5) If the change Produces a local flun rise over the average flux, rod worth increases.(.5) (1.0)
REFERENCE SSFS, Units of Instruction, SC073-A6, Specific learning objectives 17&18 .
ANSWER 1.04 (2.00)
- Convert pressure to psis'
800 + 14.7 = 814.7 psia
?OO + 14.7 = P14.7 psia (.75)
n Obtain corresponding temps from steam tables by interpolation:
814.7 psia = 5?O F P14.7 psia = 307 F (.75)
n Dotormine temperature changet 5?O - 307 = 133 F/HR = 2.22 F/ MIN (0.5)
REFERENCE
!; team Tables ANSWER 1.05 (2.25) Flow increases due to pump work requirements decreas (0.75)
8. Pump flow oscillates and NPSH approaches toro and as the pump cavitates flow will approach zer (0.75)
C. Flow decreases as the system resistance incroses, et (0.75)
.
. .
' PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGL 10
--- iniEs55isEAIEE- AFEi iERLEFFE Es5 FEDi5 FEBE
____________________________________________
ANSWERS -- SilSOUEHANNA 1&2 -85/08/06-LANGE, REFERENCE GSFS , Fluid Mechanics, Pumps. Unit objectiver i 3, ANSWER 1.04 (1.50) second period. (0.50) The -80 second period is due to long lived precersor with 5 second half life (0.5).
Using 1.44 times doubling or halving time T=-1.44 x 55.6 = -80 second period (0.5).
REFERENCE SSES, Units of Instruction SC073- ANSWER 1.07 (2.00)
Condensate depression is the temperature difference between the saturation temperature for the existing condenser vacuum and the temperature of the condensate. Fxcessive condensate depression decreases the operating efficiency of the plant since the subcooled condensate m u s t. be reheated in the reactor. ( for definition) (1.0 for reason).
REFERFNCE SSES, Thermodynamic; Steam Plant Cycles, Specific Objective # ANSWER 1.00 (2.00)
2.6 Heat inputs: feedwater Rv vater cleanup water CRD hydraulic water.Recire pump Core thermal ppwor Heat out: Steam,R:, water cleanup vatorj Fixed losses REFERENCE GSFS, Steam cycle components, Specific Objective 4 ._
.
. .
- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGF 19
--- isiss56isEsiEs- sEEi isAssFEE Es5 FEUi6 FE50
____________________________________________
ANSWERS -- SUSQUEHANNA 1&2 -05/08/06-LANGE, ANSWER 1.09 (2.00)
Since the reactor was shut down by 1% dk/k as determined at the time of peak Xenon, then the SOM will decrease as Xenon decays.[1.01 Since Xenon (pcak) is greater than the 1% dk/k a reactor restart would occur.01.01 alt. answer:
The SDM does not change.01.0] The T/S definition lists the SDM as being Xenon free.[1.03 REFERENCE SSES Rx Theory Section 6 pg. 6-7 & Section 8 pg. 6 ANSWER 1.10 (2.00)
1. Moderator / Void. (decrease) Void (decrease) Void (increase) Doppler (decrease) (0.50 for each correct ansect) ,
REFERENCE GSFD. Reactor Theory course i SCO23A-7, pg. 1-1?.
ANSWER 1.11 (2.00)
C is the correct answer (0.50), The highest concentrations will be in the c e r.tc r af the core (0.50), the high flux region from the previous oper-ating period (0.50). This will increase the flux 1cvels in the area of peripheral control rods, thus increasing their worth (0.50),
i:JFERENCE SSES, Units of Instruction,SCO?3A-8, specific objective # _ _ _ ._ - *
lo PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGF 20
~~~~iUEkUU65U555C5~~U5d5~Tk5UUFEE~dU6~ FLU 56~FEUU
____________________________________________
ANSWERS -- SUSQUEHANNA 1&? -05/08/04-LANGE, ANSWER 1.12 (7.00)
Disagre (0.5)
Thermodynamic efficiency compares energy in to versus energy out. The inercase in generator output resulted from decreasing the amount of extraction steam diverted to the HP FW heaters (0.5)
This condition requires additional energy output from the reactor to raise feedwater temperature to the same scturation temp as before. (0.5)
This is evidenced by a decrease in feedwater temperature and an increase in reactor power or MW (0.5)
REFFRENCF Susquehanna SCO23 D-10,11 specific objectivns D-10-7, D-11-1 ANSWER 1.13 (1.00) Indicated is highe (.?S)
7-10 inches higher (.25) Indicated icvel is sensed outside the dryer skir (.?5)
Steam flow throu3h the steam separator / dryer at 100% causes a backpressure of 7-10 inches H? (.?5)
REFERENCE Gusquehanna SYO17 J-7 pg.71 l
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~ PLANT DESIGN INCLUDING SAFETY AND FMERGENCY SYSTEMS PAGF ?!
_______________________________________________________
ANSWERS -- SUSQUFHANNA 1&2 -85/00/06-LANGF, ANSWER 2.01 (2.00)
A differential pressure sensor is used to confirm the integrity of the CORE SPRAY piping within the reactor vessel ( between the inside of the vessel and the core shroud).
To continuously monitor the integrity of the core spray piping, a Delta P switch measures the pressure difference between the two loops, which is effectively the inside of each Core Spray sparager pipe, just outside of the Rx vesse If the core spray sparager is intact, this pressure difference will be rer If integrity is lost, this pressure differential will include the pressure drop across the steam seperator. Alarms at.5 psid in the control room (?.00)
REFERENCE OSFS, Core Spray Lesson Plan,SYO17-C7, objective t ANSWER 2.02 (3.00) . Lou Pressure (less than or equal to 5 psig.) upstream of F030. (0.25)
To prevent draining the system. (0.25) High pressure (140 psig.) downstream of F033 . (0.25)
To prevent overpressurization of downstream piping. (0.25) . Restricts blowdown flow to prevent overloading the nonrege heat exchange (0.50)
7. During low reactor pressure operation. (0.75)
C. 1. High differential temp.(between the inlet and outlet equip. room ducts. ) (0.?5)
7. High ambient equipment room temperature. (0.25)
, 3. High system differential flow (between system inlet and combined outlet and drain flows, 60 spm, 45 see. time delay.) (0.25)
4. High system flow.( measured by the common line from the vessel to the containment penetration.) (0.25)
REFERENCE DCFS, SY017 ,L-1, pg. 12 & 1 .-. , .- _. . _
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?. PLANT DESIGN INCLUDING SAFETY AND FMERGENCY SYSTEMS PAGE ??
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ANSWERS -- SUSQUEHANNA 1&2 -85/00/06-LANGE, ANSWER 2.03 (3.00) . The controller sees a high flow from the t'l o w element which is sensing charging flow to the scram accumulators. (0.75) This action diverts most of the system floe to the accumulators to prepare them for subsequent scrams as soon as possible. (0.75) . With reactor pressure sufficiently high (> 400 psis), the rod would scram but at a slower than normal rat (0.75) With reactor pressure low (< 400 psis), the rod would not scra (0.75)
REFERENCE SSES, Lesson Plan CRD HYD. SYO17-K2, Specific 1carning Objective 4 ' ANSWER 2.04 (2.50)
A (1). Overload trip. (?). Upstream HEPA filter differential pressur]e[and duct heater differential temperature)does not fall within limits within a short time after fan starts. (3).
HIGH-HIGH temperature condition in absorber bed (450 des F). (4).
On LOCA starts only, Reactor Building pressure is not negative within a short time after fan starts. (5). Do not have 3000 SCFM or greater air flov at combined discharge header within a short g ,,g et time after fan starts.66). High radiation alarm in section of " Agg,L p(MNT SBGT sy, stem room associated with the train 0.5 ea)
starting]f (3 of 6 at 8,4 .3 D.1. HPCI system- removes contaminents from the HPCI barometric conden-ser when in operation. (0.50)
7. Main Steam system- removes contaminents from the effluent of the MSIV leakage control system when in operatio (0.50)
REFERENCE SSES. Lesson Plan, SYO17-t-3., Specific Unit Objectives t 3 and 8;
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ANSWERS -- SUSOUEHANNA 1&2 -05/08/06-LANGE, ANCHER 2.05 (2.75)
A. yes (0.50)
B. Fuel pool Reactor well Dryer / separator storage pool Shipping cask storage pool (0.25 each)
C.1. Demiralized water storag . Refueling water storag . RHR service wate ~+ A9f 4. Emergency service wate . Fire protection system using fire hose (0.25 each)
REFERENCE SSES, Lesson Plan ,
SYO17, Fuel Pool Cooling & Cleanup. Specific 1carning objectives t 4 E ANSWER 2.06 (?.50) . Service water Instrument air Service air Mechanical vacuum pump CRD system (. . RHR Service vater RBCCW p. "~g & 500 KV Switchyards (5 required P 0.4 each) O ktC W E-- The Shift Supervisor only.[0.53 REFFRENCE SSES Unit 1/2 Differences pg. 70
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ANSWERS -- SUSQUFHANNA 1&2 -85/00/06-LANCE, I
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ANSWER 2.07 (3.00) . C201 systems 1
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- RHR i
- RCIC
- 3 non-ADS SRV's (A, 8 & C)
! - B & D ESW pumps i R RHRSW pump
- Containment Instrument Gas ,
j - R Recire Pump Suction Valve l (5 required at 0.5 each) The loss of indication (s) for effected system and by Control Room Bypass Indication System Annunciator.00.51
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REFERENCE SSES OP-100-001, FO-00-009 & F-105 sheet 10 Bypass Ind. Sy .
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ANSWER 2.08 (3.00) inercase (0.25). Due to less back pressure (0.25). decrease (0.25). Due to reverse flow through the broken jet pump.(0.25)
- deercase (0.25). Due to less core flow. (0.?5)
i decrease (0.25). Due to loss thermal power. (0.25)
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' increase (0.25). Due to less of a change in delta press. (0.25) '
6. decrease (0.25). Due to less flow through the core. (0.25) '
REFERENCE T~ * N #Y SESS. Opor.lic. Systems vol.fi, Unit SYO17, J1 pg. 9 16 and SYO17 L-8,4-11.
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] ANSWER 2.09 (2.00)
covered for a longer period of time. (0.50)
B. LPCI failure (0.50). This is due to LPCI having a larger flow rate and
, no counter current flow limitation effects (at the top of core).(0.50)
REFERENCE
<
DSFS. Transient Analysis. CH.410, Pg. 267-269 and pg. 273-27 :
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ANSWERS -- SUSQUEHANNA 1&7 -05/08/06-LANGE, ANSWER ?.10 (1.25) The D/G start logie had been inadvertenly de-energized due to the wrong D/C knife switch being repositione (0.50) The D/C knife switches have been labled accordingly and the knife switch for the D/G auto start logics has been painted re (0.75)
REFERENCE SSES, Loss of Off Site Power Test problems, and ressedial training required for operator .
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- INSTRUMENlS AND CONTROL S PAGE ?6
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ANSWERS -- SUSQUEHANNA 1&2 -05/08/06-LANGE, ANSWER 3.01 (3.00) The (1) 20 non-calibrated jet pump differential pressure signals, obtained by comparing each (2) jet pump throat pressure (l.p. tap) to the (3) SBLC injection line (H.P. tap), are (4)
converted to flow signals by square root extractor (1.00) The (1) 10 jet pump flou signals from each bank are then (2)
summed to obtain two (3) recire loop total jet pump finu signal The (4) two recire loop total jet pump flows are then summed to provide total core flo (1.00) The two recire loop total floe signals are actually sent tn (1)
two summer One summer calculates the (2) algebraic som of the tuo signal The other summer calculates the (3) algebraic differences between the two signal The (4) first summer supplies the total core flow indication eith both pumps operatin However, (5) when a recire pump trips the total core flew indication is automatically transferred to the latter summer to provide (6) accurate indication with the reverse flow signals from the 10 jet pumps in the idle loo (1.00)
REFERENCE Iacensed operator systems, Reactor Recire Sys. SYO17-L-0., specific learning objectives 4 38 ANSWER 3.0? (?.50)
A . 125 VDC LO.50) To (1) prevent the logic from cycling the (2) ADS valves open then shut as the (3) reactor level instruments respond to the (4)
swell, level increase, when the ADS valves ope (1.00)
C(1) Core Spray pump discharge pressure >145 psig.(?) 9HR pump discharge pressure 2125 psi (1.00)
e REFERENCE SSES, SYO17,C-4, ADS, specific objectives 4 ?, 3, 7.
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- INSTRUMENTS AND CONTROLS PAGE 27
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ANSWERS -- SUSQUEHANNA 1&2 -05/08/06-LANGE, ANCHER 3.03 (2.50)
A. Speed control signal failure or loss of power to the scoop tube positione (1.00) + ou -i 6 4 B. There are no runback capaiblities (speed limiters) with a Scoop Tube Locku (0.50)
C. (1) Control Signal Failure Alarm is cleared (0.50) and the (2) white power available light above the Scoop Tube 'A' LOCK / RESET Pushbutton is illuminate (0.50)
REF: PP&L Reactor Recire Lesson Plan and SSES OP-64-001, Section ANSWER 3.04 (2.50) Manual switch on panel 1C651. (0.33)
Edge rod selected. (0.33)
Reference APRM < 30 %. (0.33)
B. B&D l e v e l AER+i ' s . (0.50)
tre -s 1[ Bl[*
C. Two APRM outputs are used as reference core power signals for each RBM channel. One normal ar.J one backu (0.50)
RBM channel A- APRM C is the normal and E is the backup. (0.25)
RBM channel B- APRM D is the normal and F is the backup. (0.25)
REFERENCE SSES, SYO17-K5, specific objective f ANSWER 3.05 (2.00) Increases b. Decreases c. Increases d. Decreases (0.50 each)
REFERENCE SSES, Reactor Vessel Instrumentation, SYO17-J2, specific objective 4 9.
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ANSWERS -- SUS 00EHANNA 1&2 -35/08/06-LANGE, ANSWFR 3.06 (3.00) Unit 1 - The 'C' RHR pump will stop.CO.51 Unit 2 - All four RHR pumps start.CO.53 This is due to the LOCA/No LOCA Logic.CO.51 Any pump may now be used for S/D Cooling.CO.753 To prevent ESS buses and or the DGs overloading due to post L OCA equiptment loading.CO.75]
REFERENCE SSES,12/12/83 exam & RHR lesson plan, SYO17,C-1, specific objective 4 ANSWER 3.07 (3.00) Speed Limiting whenever*
1. Any condensato pump disch, pressure <100 psig CO.31 OR Individual FW flow (20% CO.33 OR Any FW Heater 1 or 2 Hi-Hi E0.31 AND Reactor water level is below the low 1cvel alarm point of 30'00.33 2. A protective trip of any 1 of 4 CW pumps CO.31 This fault will produce two effects. The Speed Control Loops error limiting network attempts to increase generator speed to maximum.CO.53 The Voltage Control Lcop sees a loss of its demand signal (speed) and reduces generator voltage.EO.51 Resulting in a Generator Lockout.EO.53 REFERENCE SSES Rx Recire Control p , 10-11 & 14-15
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- INSTRUMENTS AND CONTR0 PAGE 29
ANSWERS -- SUSOUEHANNA 1&2 -05/08/06-LANGE, ANSWER 3.08 (3.00) The TCV's will close to 50% flou position 00.53 The 1CV LVG passes a MCF signal of 50% rather than the signal fron the Pressure contro11er.00.53 The BPV's will remain closed through the transient.CO.53 The MCF summer will send a cro signal to the BPV tVG.[0.53 Reactor power & pressure will rapidly increase following the fault.EO.53 The reactor will scram on High Flux A/OR High Pressure. Reactor pressure will be controlled by the TCV's.CO.51 REFERENCE SSES FHC Pressure Control & Logic and Figure C ANSWER 3.09 (1.25)
LOCA signal sensed by Core Spray logic, (-129' level or 1.72 D/W Press.),
WITH a Main Generator Primary or Backup lockou (1.25)
REFERENCE SSES, Lesson Plan, SYO17-G-5, specific objective 4 ANSWER 3.10 (2.25)
A. High bdV tailpire tem SRV opening solenoid illuminated Accoustic monitor light li Decrease in generator load with Rn. power constant Indicated FW flow greater than indicated Steam flo Toros temp. increas DW/ Torus pressure increasin ( 5 required at 0.25 each)
B. Feedwater temp will decrease (0.50). , since Feedwater flow is higher than Turbine Steam floe (0.50).
REFERENCE p o ssg g Le 4- S toleves r Co ca,' -
SSES, Simulator Malfunctions # 80, and Relief Valve and ADS lesson pla .
40 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30 E5NTR5L-------~~---------------
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ANSWERS -- SUSQUEHANNA 1&2 -85/08/06-LANGE, ANSWER 4.01 (3.00) Entry Conditions: RPV water level <+13'.E0.53 Drywell pressure >1.72 psig.CO.5] An isolation condition exists which requires or initiates Rx scram.EO.53 By at least two independent indications, misoperation in automatic is confirmed.CO.53 Adequate core cooling: The active fuel is covered with liquid or a two phase mixture.E.34] ECCS flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembly.E.333 Steam flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembly.E.333 REFERENCE SSES EO-100-021 Level Control pg. 2& 5 ANSWER 4.02 (2.00)
I F- GrPLMoIva ~
A. Remain the same. (0.50) =) o R S LE$ l ily l,n w er D. Check discharge line temperatures , suppression pool for oscillations and acoustic monitor (0.50)
C. The major concern is the resulting HEATING of the Suppression Pool to a point where m a v .' n u m condensation following a LOCA is no longer available. (1.00)
REFERENCE SSES ON-183-001 os. 2 & 3 l
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ANSWERS -- SUSOUEHANNA 1&2 -05/08/06-LANGE, ANSWER 4.03 (3.00)
1. Scram the reactor Place the Mode Switch to Shutdown Trip the Main Turbine Verify all control rods inserted Insert SRM's and IRM's Close the MSIV's and MSI. drains (HV-1F016 and HV-1F019)
7. Trip all operating Reactor Feedwater Pumps 8. Close all RFP discharge isolation valves (Hv-1060-3A,3D&3C)
9. Place FW low load demand signal indication in Auto set for 10'.
(10 required at 0.3 each )
lb.Crea H Pc t REFERENCE yg, y , , ,, p o g, p p (p g. g r- c o l)
Susquehanna, Plant Shutdown from Outside the Control Room, E0-100-009, Rev 0, 3/24/84, Sec 3.1 pg ?
ANSWER 4.04 (2.00)
If it is determined that systene air pressure CANN01 be restored and is de-creasing to 65 psis. In any case, PRIOR to the scram inlet and outlet valves opening. (1.00).
This loss of air pressure could Icad to a significant Scram Discharge Volume inleakage to an extent that the scram function could be adversly affected or prevented. (1,.00)
REFERENCE FO -2 00-002, loss of Instrument Ai .o t ANSWE{,,4.05 (1.50) a
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'F #~ # " ' *~ * * Lent. (0.50) . True (0.50) Falser The system can be manually started but vill trip on overspee The inverter supply's control power to the speed control loo (0.50)
REFERENCE FO-100-009, pg 1- . __- _. - . - . -. . .- - - -. . . - - - - _ - - - .
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32
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RADIOLOGICAL CONTROL
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r ANSWERS -- SUSQUEHANNA 1&2 -85/08/06-LANGE, I i f i
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i ANSWER 4.06 (2.00)
i A. Flow must be increased to > 3000 spm as quickly as possibic to prevent !
j the minimum flow valve from openin3, (10 sec'. time delay), and pumpinq l vessel water to the suppression poo (1.00)
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pumps 1P202,A & B availabic for LPCI initiation on Unit- (1.00)
) REFERENCE
HP-149-002.pg. 17 and OP-149-005 pg.5.
1 ANSWER 4.07 (1.50)
, 1. To limit errosion and/or vibration damnage to the remaining operating j heaters. (0.50) The system design will not accomodate the increased flows and feedwater j temperature dro (0.50) No isolation valves exist in the piping. (0.50) i
. REFERENCE IIP- 14 7 - 001, pg. 9-1 .
ANSWER 4.08 (3.00)
l j' 1. All offsite power supplying in-house loads for both Units is lost. (.5) !
i All four diesel generators fail to star (.5) j i 2. B and Ce narrow range 1cvel indicators on 653 (SIP). .
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gj uPOET . e r. 3 r 1; vel ir.di star e r. 650 ' C IP L. g4es4. d %* / ) -
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j Wide range on 2000 ( 0.G6- f o r each correct answer)
j HPCI/ RCI (.5)
] Wnen inst. AC is restored. By depressing the Full Core Disp.rcset. (.5)
}
! REFERENCE
!.SFS , Procedure EO-200-030.
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ANSWERS -- SUSOUEHANNA 1&2 -85/08/06-LANGE, ANSWER 4.09 (2.00)
a. Startup/ Hot standby b. Any Temperature c. Shutdown d. >200 degrees F Shutdown < or EQ 200 degrees F S. Shutdown er Refuel < or EG 140 degrees F (0.25 each)
REFERENCE SSES-Procedure, AD-0A-309, Rev.6, pg.6 of 2 ANSWER 4.10 (1.00)
A. To assure 2/3 Core Coverage following a LOCA . (0.50)
B. To provide Turbine Overspeed Protectio (0.50)
REFERENCE SSES, Procedure GO-100-003, p .
ANSWER 4.11 (2.00) Capable of beins closed by an OPERABLE primary containment automatic isolation system, or Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed positio (2.00)
REFERENCE SSES, Procedure AD-0A-309, pg.5.
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- ANSWERS -- SUSQUEllANNA la2 -85/08/06-LANGE, D .,
i, 4 ANSWER 4.12 (2.00)
'-
From att. A, the maximum suppression pool temperature for RPV pressure
- of 500 psis is approximately 187 degrees F. Subtracting 175 from i 187 gives a delta temp. of 12 degrees F. From att. E (delta The), minimum l suppression po water level corresponding to 12 degrees F is approximately N!N4 fee Q.60)
b
! REFERENCE N N*t NY* f l Susquehanna EO-100-023 Rev.0
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'T E S T CROSS REFERENCE PAGE 1 GUESTION VALUE REFERENCE
._______ ______ __________
01.01 2.25 DJL0000213 01.02 2.00 DJL0000348 01.03 2.00 DJL0000349 01.04 2.00 DJLOOOO350 01.05 2.25 DJL0000351 01.06 1.50 DJL0000353 01.07 2.00 DJL0000354 01.08 2.00 DJL0000356 01.09 2.00 DJL0000374 01.10 2.00 DJL0000385 01.11 2.00 DJLOOOO374 01.12 2.00 DJL0000399 01.13 1.00 DJL0000400
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25.00 02.01 2.00 DJLOOOO127 02.02 3.00 DJLOOOO359 07.03 3.00 DJL0000360 02.04 2.50 DJLOOOO341 07.05 2.75 DJLOOOO367 02.06 2.50 DJL0000375 0?.07 3.00 DJLOOOO376 02.08 3.00 DJL00003:13 02.09 7.00 DJL0000384 02.10 1.25 DJL00003??
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25.00 03.01 3.00 DJLOOOO357 03.02 2.50 DJLOOOO343 03.03 ?.50 DJL0000370 03.04 2.50 DJLOOOO372 03.05 2.00 DJL0000373 03.06 3.00 DJLOOOO377 03.07 3.00 DJL0000378 03.08 3.00 DJL0000377 03.09 1.25 DJL0000395 03.10 2.25 DJLOOOO396
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25.00 04.01 3.00 DJL0000380 04.02 4 00 DJLOOOO3 tit 04.03 3.00 DJL0000302 04.04 2.00 DJL0000387 04.05 1.50 DJL0000388 04.06 2.00 DJL0000389 04.07 1.50 DJLOOOO390 04.08 3.00 DJLOOOO371
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TEST CROSS REFERENCE PAGE 7 OUESTION VALUE REFERENCE
-_______ ______ __________
04.09 2.00 DJL0000393 04.10 1.00 DJL0000397 04.11 4 00 DJL000039G 04.12 R.00 DJLOOOO401
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?5.00
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100.00
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Facility Comments And NRC Resolution To Susquehanna Units 1 & 2 (RO) Written Exam Conducted On August 6,1985 All comments were resolved during the two hour Post Exam review conducted
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immediately following the (RO) Written Exam. No edditional written comments
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submitted by the License i l
Category #1 l 1.02 - Facility Comment 28.18 psig is the correct answer due to the
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Condensate storage tank being vented and control room instrumentation calibrated Apr
- psi !
NRC Resolution 28.18 is the correct answer if all calcu-l lations shown are correc ,
Category #2
j 2.04 - Facility Comment The Hi Radiation alarm in section of !
S.B.G.T. sys has been removed during plant
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modification.
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i NRC Resolution Answer remains as is, with (5) possible j answers. Three correct answers required j for full credit.
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Category 43
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i 3.10(b) Facility Comment Feedwater temperature could increase if
! candidate explains operation of F.W. 3
element control sys.
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i NRC Resolution Answer is correct as 1s. Will consider
! candidates explanation of 3 element control i for F. W.
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l Category #4 4.02(A) Facility Comment Rx. Power could slightly decrease.
! -NRC Resolution If candidate says a slight decrease, must
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l Explai (1) Facility Comment Correct answer is False, due to Unit 1 & 2 l separatio *
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i 0FFICIAL RECORD COPY OL COMMENTS SUS - 0001. /:
09/09/85 / !
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NRC Resolution Comment accepted, with documentation provide .08(2) Facility Comment Delete, Upset range Level indicator on 653 (SIP) due to procedure chang NRC Resolution Accepted, with documentation provided.
l OFFICIAL RECORD COPY OL COMMENTS SUS - 0002. /09/85
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