IR 05000387/1986001
| ML20141E665 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/01/1986 |
| From: | Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20141E652 | List: |
| References | |
| 50-387-86, 50-387-86-01, 50-387-86-1, 50-388-86-01, 50-388-86-1, NUDOCS 8604220345 | |
| Download: ML20141E665 (41) | |
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EXAMINATION REPORT Examination Report No /86-01; 50-388/86-05(0L)
Facility Docket No: 50-387/388 Licensee: Pennsylvania Power and Light C North Ninth Street Allentown, Pennsylvania 18101 Facility: Susquehanna 1 and 2 Examination Dates: January 14-16, 1986 Chief Examiner: 0 *
M David Lange, LVad 'Reacto# Engineer 3'IT//P6
' Dafe (Examiner)
Reviewed by: f h fen 0:K Robert Kefler, Chief ff/f{h Dhte Projects Section 1C Approved by: 8[Ng [ 4 572 Harry B.' Kister,4Tiief Date
' Projects Branch No. 1 Summary: Operator licensing examinations were administered to five Senior Reactor Operator Upgrade candidates and one Instructor Certification candidat All candidates passed the examinatio PDR ADOCK 05000387~
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REPORT DETAILS TYPE OF EXAMS: Initial Replacement X Requalification EXAM RESULTS: l SRO Upgrade l Inst.Certl l Pass / Fail l Pass /Faill I I I I I I I IWritten Exam l 5/0 l 1/0 l l l 1 I I I I I 10ral Exam l 5/0 l 1/0 l l l l l l l l l ISimulator Examl 5/0 l 1/0 l l 1 1 I I I I I l Overall l 5/0 l 1/0 l l 1 I I Chief Examiners at Site: David Lange, NRC Other Examiner: Frank Crescenzo, NRC Summary of generic strengths or deficiencies noted on oral exams: Strengths were noted in the following areas:
* Application and interpretation of technical specification * General system knowledg *
Application of the relatively new Emergency Operating Procedure * Responsibilit Weaknesses were noted in the following areas:
* Classification of events and implementation of the emergency pla * Specific auxiliary effects due to electrical malfunctions.
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3 Personnel Present at Exit Interview: NRC Personnel David Lange, Lead. Reactor Engineer (Examiner) Frank Crescenzo,-Reactor Engineer Examiner Rick Jacobs, S. Loren Plisco, Facility Personnel T. M. Crimmins, Plant Superintendent R. J. Thompson, Assistant Plant Superintendent W. G. Ward, Manager, Nuclear Training H. J. Palmer, Operations Supervisor
'T. R. Markowski, Operations 'R. M. Peal, Operations Training A'. Fitch, Operations Training W. G. DiDomenico, Operations Training R. Chin, Operations Training Summary of NRC Comments made at exit interview: * The generic strengths and weaknesses noted in paragraph 3 were discusse * Training and Operations department personnel were cooperative'
throughout the examination proces . Summary of facility comments made at exit interview:
* The facility acknowledged the NRC comments noted in paragraph * The facility felt the written examination was fair and had challenged the objectives of SR0 training at Susquehann . Changes made to written exam during examination review:
All comments to the written examination were resolved during the examina-tion review. The following represent significant changes made to the examination as a result of these resolution Question N Change Reason 6.1 Add condensate storage CST can be crosstied tank as an acceptable to refuel water alternate answe storage pump .: G
Question N Change Reason 7.0 Delete E0-100-105 from Question was misleading answer unless candidate in that "no other assumes a valid re- alarms" was given leas Valid release is verified by Stack
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Monitor Alarm .0 Answer is tru Answer key was incorrec Attachment: Written Examination and Answer Key (SRO)
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frrncb n7e11f l e U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: SUSOUEHANNA 1&2
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REACTOR TYPE: BWR-GE4 ___---_--_--------------- DATE ADMINISTERED: 86/01/14
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APPLICANT: __ d _8_d_ h j_(_________ INSTRUCTIONS TO APPLICANT:
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Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet Points for each ( question are indicated in parentheses after the question. The passing ' Srade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination start _
~~ % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY -------- ------ ----------- -------- -----------------------------------
25 24 ___I_00___ ___1_51_ ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS .- ! 24 0 23.53 ___1_0___ ______ ___________ ________ PLANT SYSTEMS DESIGH, CONTROL, AND INSTRUMENTATION
--25I-0-- 0 24 5 --_1-1 ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 28.00 27.45
________ ______ ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS !102.00 100.00 TOTALS
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FINAL GRADE _________________% All work done on this examination is my own. I have neither lgiven nor received ai ~~~~~~~~~~~~~~ 5PPL555UT I5~55GU5TURE l l
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R THEORY OF NUCLEAR POWER FLANT' OPERATION, FLUIDS, AND PAGE 2
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QUESTION 5.01 (2.00) During a reactor startup, criticality is achieved when a positive period is maintained without further positive reactivity addition The definition of critical states Keff equals 1.0 and reactivity equals 0.0 and period would therefore be infinite. WHY then is the reactor declared critical when the period is positive? (FULLY EXPLAIN) DUESTION 5.02 (2.50) The Reactor is on a 85 second perio Moderator temperature is 160 des F. With no operator action, WHAT will the moderator temperature be when the reactor is again on an infinite period? Assume * EOL for time in Reactor lif STATE ANY ASSUMPTIONS YOU HAK SHOW ALL HOR GUESTION 5.03 (3.00) For each of the pairs of conditions listed below, state WHICH condition
- would have the GREATER differential rod worth and briefl), EXPLAIN WH Reactor moderator temperature of 150 des F or 500 des (1.0) For a inserted rod next to a fully withdrawn control rod or next to a fully inserted control rod. [ Assume average (1.0)
core flux is constant] For a rod at position 10 or position 40 of a core operatin3 at 100% powe (1.0)
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. ... . . . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 3 ---- ------------ ------------------------- -----_--------
00ESTION 5.04 (3.00) Following an AUTO INITIATION of HPCI at a reactor pressure of 800 PSIC,
' reactor pressure decreases to.400 PSI HOW are the following parameters affected (INCREASES, DECREASES, REMAINS CONSTANT) by the change in reactor pressure? BRIEFLY EXPLAIN your choic ASSUME the HPCI System is operating in automatic as designe HPCI flow to the reacto (1.0) HPCI pump discharge head (assuming NPSH remains constant). (1.0) HPCI turbine RP (1.0)
GUESTION 5.05 (3.00)
-The curve for NPSH vs Percent Power for the reactor recirculation pump is given below. Explain why NPSH behaves as it does for line segments 1,2, and (3.0)
300Fe--
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NPSMA ! p
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100R-
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, x ^ # ' ,%HR 222T 5N 50 % M)%
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1d0% W o%- J eowen -
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. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4 ---- -------------------------------------- --_-----------
QUESTION 5.06 (2.50) Three (3) minutes following a reactor scram from high power, indicated reactor power is 75 on IRH range 4 and decreasin3 a. WHAT will INDICATED power be one (1) minute later?
(Show-calculations) (1.5)
b. Explain WHY power decreased at this rat (1.0) GUESTION 5.07 (3.00) Concerning THERMAL LIMITS: Since MCPR is not a directly measurable parameter, WHAT are THREE (3) measurable core parameters needed by the process computer to calculate MCPR?~ (1.0) b. With regard to MAPRA (2.0)
' WHAT is the RELATIONSHIP between MAPRAT & MAPLHGR? The process computer prints out a MAPRAT of 1.05. Is this acceptable?
3. WHAT physical consequence could occur if the MAPRAT limit is exceeded? QUESTION 5.08 (2.50) For each of the followin3 events, state which COEFFICIENT of reactivity would act FIRST to change reactivit Control Tod -drcp at power -
(0.5) SRV opening at ynwer (0.5) Loss of shutdown cooling when shutdown (0.5) One recire pump trips while at 50% power (0.5) Loss of one feedwater heater (extraction steam isolated) (0.5)
GUESTION 5.09 (1.50) Could the same reactivity change result in different periods in the saae core? EXPLAI .
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- THEORY'0F NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE ..--------------------------------- ----_----__---
GUESTION 5.10 (2 00) ,rT A. What is' SHUTDOWN MARGIN? ( T.S befinin'e u)
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.: . ,-. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 6 .------------------------------------------------------
QUESTION 6.01 (3.25) Give the five (5) signals required for the Auto Depressurization System (ADS) Intiation Logic Division I to automatically open the valves AND the PURPOSE for each of these signal ESetpoints are not required] QUESTION 6.02 (1.50) The "B' recire MG is controlling at 75% in master manual when the Senerator tachometer output fails to zero. EXPLAIN how this fault will effect the 'B' recire M BE SPECIFI GUESTION 6.03 (3.00) When a LOCA initiation signal is present and the Diesel Generators are supplying the ESS bus power, seven (7) loads will be sequenced o WHAT are five (5) of these loads? ETimes are not required, BUT proper sequencin3 order is required.] , QUESTION 6.04 (3.00) When the reactor water level reaches Level #3 (+13'), it initiates a reactor scra WHAT are si:: (6) additional automatic actions caused or contributed to by this level condition? DUESTION 6.05 (3.00) LIST FIVE (5) SPECIFIC systems or major parts of DIFFERENT systems that may be operated from the Unit 1 Remote Shutdown Panel (1C201).
BE SPECIFI (2.5) HOW would the Control Room Operator know if a transfer switch on 1C201 was placed in EMERGENCY? (0.5)
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. .. ' . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 7 ------------------------------------------------------
QUESTION 6.06 (2.50) For the followin3 conditions WILL the PCIC system INJECT if the condition was present at the time that an automatic initiation signal was received? (JUSTIFY YOUR ANSWER.)
1. Turbine steam e::h a u s t valve (F059) is NOT FULLY OPE (0.75) Loss of the RCIC Topar inverter outpu (0.75) List two RCIC turbine trips NOT bypassed by placing RCIC into 1C201 contro (1.0) OUESTION 6.07 (2.25) Unit 1 is in Shutdown Cooling on Division 1 RHR with the
'C' pump running. While conducting a startup test, the Unit 2 PCD depressed the manual initiation button for Unit 2 Division 1 RHR. WHAT pumps will start and stop on BOTH UNITS ?
EXPLAI (1.5) To ensure Unit 1 will have full compliment of ECCS pumps in the event of a LOCA, WHICH pumps on Unit 2 would you USE for Shutdown Cooling ? (0.75) DUESTION 6.08 (3.00) With r e3ard to Process Radiation Monitors * A trip of the Liquid Radwaste Discharge RMS will close isolation valves HV06432A1 & A2 stopping the discharge. LIST 3 conditions that may have initiated this tri (1.5) l The Process Ventilation RMS monitoring the Refuel Floor Wall ) ' Exhaust, increased to its trip point. LIST 3 automatic actions this trip will generat (1.5) 00ESTION 6.07 (1.00) Concerning the Reactor Water Cleanup (RWCU) System: What tuo innlation signals will only close the outbeard 1_olatiu. velveo (1.0)
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l PLANT SYSTEMS DESIGNr CONTROLr AND INSTRUMENTATION PAGE 8
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OUESTION 6.10 (1.50) A. Will the fuel pool cooling mode of RHR be able to maintain the fuel Pool temperature below 125 des during emergency heat load (EHL) conditions without the Fuel Pool Cooling Syste (0.50) List four emergency water makeup sources that are available should the fuel pool require makeup due to evaporation or leaks ? (1.00) f (*** * END OF CATEGORY 06 *****) _ -
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____________________ 00E?TTON 7.01 (3.00) Answer the following questions with regard to The Loss of Startup Ous 10 Of f Nor mal Procedure ON-003-001, when startup Bus 10 was Ic.st due to a fault from Startup Transformer T-1 Undervultase load shedding occurs on WHICH 4.16 kv busses? (1.0) Tie breaker OA 10502 closes to feed startup bus if WHAT tPree (3) conditions exist? (1.0) Which 4.16 kv EGS bus breakers close if thier control switchs are in the Normal After Trip position? (1.0) l t DUESTION 7.02 (1.50) Ycu ci e the Shift Supervisor and the Reactor has been at 99% rated power for several hours. You notice that a Bypass Valve is particlly opened. Reactor Power is now at 102% rated powe ' .b u ru e no other thermal limits have been exceeded) a. What is the first action per SSES procedures you would have ! the reactor operator take ? i (0.75) Mhtt is the n.a n i m u m time interval you may be at this
, sower level?
(0.75)
l CUESTIOM 7.03 (2.00) Ccncer ning the operation of the Recirculation System, briefly discuss ahy
m the following precautions are given in OP-64-001, Operation of .
Recircul.; tion System: Consult with the Reactor Ensineer prior to restarting an idle iecirculation pump at powe (0.75) The ret iiculation isolation valves should not be closed for oor e than app oximately 5 minutes while coolin3 dow (0.75) Stop ' ur.t ila t ing the recirculation MG set within 30 minutes
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with both recirculation pumps shutdow (0.5)
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. .. . .' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 10 - ~~~~R d65UL55iEkt E6NTRUL'~~~~~~~~~~~~~~~~~~~~~~~ --------------------
GUESTION 7.04 (3.00) UNIT ONE has Leen operating at 95% sustained operatio Suddenly you receive 'MN STM LINE HIGH RAD'N' and "0FF-GAS HI RADIATION' alarm No other plant alarms have been received. A ' quick check of the MSL Radiation Monitors show that Channel A&B read 4 X Normal and C & D read 3 X Normal. STILL no other alarms have been receive What two emergency plant conditions exist that require immediate actions? (1.0) What THREE SSES Emergency Operating Procedures (EO-100-XXX) would you enter? Include why each procedure is applicabl (2.0)
(Procedure name or number is acceptable)
GUESTION 7.05 (3.00) Concerning the procedure for Containment Control (E0-100-103): Why should injection into the RPV from sources EXTERNAL to the primary containment be stopped if primary contain-cent water level reaches 120 feet? (TWO reasons)
(1.5) In accordance with EO-100-102, what defines ' adequate core cooling *? (NOTE: Give two ways to define it) (1.5)
OUESTION 7.06 (2.00) Plant conditions require RCIC to control RPV water leve ' RCIC turbine speed is currently 2000 rpm. RPV level is constant. Based on these conditions how should RCIC and RPV level be controlled per EO-eet-100 ? Include why these actions are required in your answe ioo < bw hy O &" Cd(4 k" *s i
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RA65ULUUIUIE~EUUTRUL~~~~~~~~~~~~~~~~~~~~~~~~ ____________________ QUESTION .7.07 (2.50) The SSES E0P's covers manual operation of the ADS-SRVs and SRVs in ALPHA orde a. Explain under what conditions the ADS-SRV's are manually . _ operated and why they are preferred over the SRV's (NON-ADS). (1.25) b. Explain under what conditions the NON-ADS SRV's are manually operated and why they are preferred over the ADS-SRV' (1.25) GUESTION 7.08 (2.00) l The SSES CDP EO-200-101, Scram 101, has as the first action
* Place the roode switch to shutdown". What are the two major Soals of this action ?
QUESTION 7.09 (2.00) During a normal reactor startup, after criticality is achieved, the Plant Startup and Heatup Procedure GO-100-002 directs the operator to check the SRM/IRM overla a. What is considered an adequate overlap (in decades) and what action must be taken if proper overlap is not observed? (1.50) ! Answer TRUE or FALSE : If an IRM channel increases from 25 on range 2 to 25 on range 3, the indication has increased by one decad (0.50)
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____________________ QUESTION 7.10 (2.00) A. When removins extraction steam from a feedwater heater strins, procedure OP-147-001 lists several precautions. Briefly explai . The hishest numbered F.W. heater extraction steam supply should be removed first. WHY? (0.50) 2. Removal of extraction steam supply from more than one operatin3 F.W. heater strins is prohibited. WHY? (0.50) 3. F.W. heaters 1 & 2, ABC extraction steam supply and F.W. heater 1-ABC and drain cooler 6-ABC cannot be isolated without a turbine trip. WHY? (0.50) B. When testins of feedwater heater bleeder trip valves , procedure OP-147-001 lists the precaution "Each feedwater heater string shall be tested individually'. Briefly explain WHY this precaution exist (0.50)
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OUESTION 7.11 (1.00) Concernins Procedure GO-100-003, (Power Operation) i A. Why is Jet Pump operability a concern prior to exceeding 25 % of rated power ? (0.50) B. Why is the Turbine Load Set limited to 100 MWe above actual load until actual load is reached ? (0.50) OUESTION 7.12 (1.00) Concerning the RHR SYSTEM i Prior to ctartins 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. Explain the reason for these requirements ? (1.00)
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QUESTION- 8.01 (2.00) AccordinS to 10 CFR 50, the reporting requirement of DNE HOUR for the followins' conditions is applicable. ETRUE or FALSEJ The plant is in a condition not covered by operatins and emersency procedure (0.5)
.B. The loss of the offsite notification syste (0.5)
C. A valid automatic initiation of the Reactor Protection System (RPS). (0.5) D. A shutdown was commenced because the plant was in violation of the Technical Specification (0.5) GUESTION lB.02 (2.50) What are five (5) radiological conditions that require a Radiation Work Permit? QUESTION 8.03 (3.00) The Instrument Maintenance Supervisor informs you that while checking relay logic on a 4KV safety bus he found one of the undervoltage relays inoperativer the exact cause is unknow The loss of this undervoltage relay will disable the fast transfer to the other off-site circuit and will disable the auto start feature of the a:sociated diesel senerato In accordance with the Technical Specifications, CAN THE PLANT OPER-ATE UNDER THIS CONDITION? If not WHY? If sor what ACTIONS and/or LIMITATIONS are applicable? (3.0) xxxxxxxxxxxxxxxxxxxxxxurwxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx x NOTE' USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS * x TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE x x SECTIONS OF THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWE x xxxxxxxxxwrxwarwxxxwwrzwruxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx (***** CATEGORY 08 CONTINUED ON NEXT PAGE xxxx*)
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GUESTION 8.04 (3.00) In accordance with the Technical Specifications, the reactor was scrammed,dve to Suppression Chamber water temperature being greater than if0 degrees F. The reactor is now in HOT SHUTDOWN, Suppression Pool Cooling is ONr and Suppression Chamber water temperature is 92 degrees F. CAN YOU STARTUP THE REACTOR AND ENTER OPERATIONAL CONDITION 2? EXPLAIN YOUR ANSWER FULL (3.0) GUESTION 8.05 (3.00) When you assume the midnight to eight A.M. shift the plant is at 85% power and all conditions are normal with the following exceptionst a. APRM channel 'A' is bypassed for maintainence b. APRM channel 'B' is failed low, and bypasse Two hours into the shift APRM channel 'C' fails downscal In accordance with the Technical Specifications, WHAT ACTIONS MUST YOU TAKE IN THIS SITUATION? (3.0) ummxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS *
* TO ANSWER THIS QUESTION. FULLY REFERENCE ALL APPLICABLE x m SECTIONS OF THE T.S. THAT YOU USE TO DEVELOP YOUR ANSWE * ==nxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx QUESTION 8.06 (3.00)
You have just received word from the I & C Supervisor that REACTOR VESSEL STEAM DOME PRESSURE INSTRUMENT PS-1821-N021A is inoperable. The Supervisor estimates that it will take 12 hours to replace and .ecalibrate the instrument. IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS, WHAT ACTIONS MUST BE TAKEN DUE TO THIS INSTRUMENT FAILURE ? (3.0)
(NOTE: A actor Vessel Pressure Instrumentation List is attached to this Examination.)
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NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS TO x n ANSWER THIS QUEyTION. FULLY REFERENCE ALL APPLICABLE SECTIONS OF * n THE T.S. TWAT YOU USE TO DEVELOP YOUR ANSWE * xxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxxx (xxxxx CATEGORY 00 CONTINUED ON NEXT PAGE *****)
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l !: .~ ADMINISTRATIVE PROCEDURES, CONDITIONS, LIMITATIONS PAGE 15 , ___________________________________________AND _______________ ' OUESTION 3.07 (1.50) Procedure AD-DA-421, Significant Operating Occurrence Reports, list'three responribilitiet for the Shift Supervisor. What art they ? OUESTION 8.0S (2.00) l- List t w o ct r.3'ili o n s pe r Procedure AD-DA-306 that an Equipment
! Relesse ro pm ic required with a Work Authorization.
l , QUESTION 8.0? (2.50) l As a shift supervisor on a back shift you receive a call that the l next shift will be short one licensed operator and that overtime j must be scheduled. WHAT are the guidelines soverning overtime l selection when regarding the number of maximum work hours? !
( Co.4s m 4 .: sc (c.) 5 0 c c.. .=.c y'.u t.',. e s . )
( OUESTION G.10 (1.00) Concerning Procedure AD-0A-309 Primary Containment Access Control Why is there a caution to ensure primary containment equipment hatches are not removed when the drywell head is removed and the drywell area ventilation ducts are open? (0.5) Answer TRUE or FALSE. After deinerting the containment and while containment purse is in progress, the installed oxygen analyzers may be used to determine entry condition (0.5) OUESTION 8.11 (1.50) D scribe the conditions necessary to cause the following alarms on the Interlock Status Display Module associated with the refuelins platform.
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(use the attached fi3ure 13 for reference)
, ' Fuel Hoist Interlock" (0.50) I 'Back Up Hoist Limit' (0.50) ' Rod Block Interlock # l' (0.50)
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GUESTION 8.12 (3.00) The plant is operating at 75% power. The IRC supervisor informs you that the EOC-RPT trip systems are inoperable due to all turbine control valve fast closure setpoints being out of spec ( <460 psi). He further informs you ' that the problem is due to the use of faulty test equipment which was used during the last channel calibration. Using the attached tech. specs., list all applicable action statements and state which is most limiting.
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jy[76'/5 /v' C . s E,v'. l 5.- ____ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 17 ______________________________________ ______________ ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, ANSWER 5.01 (2.00) Criticality is achieved low in the power ranse (i.e. source or intermediate ranse). It is difficult and time consumin3 to distinguish between exactly critical (Keff = 1.0) and suberitical , '
.cultiplication and/or the effects of source neutrons. E1.03 If a positive period is maintained without further positive reactivity achieve additions,'it is assured that criticality has been l- The reactor is in fact supercritica E1.03 (2.0)
REFERENCE SSES Reactor Theory SC023 A-4 rev 0, See IX, ps 21 Reactor Theory /Rx Control SC023 A-6 rev 0, SEC I' PS 1 CO-100-002 rev 3, sec 6.24, note,6.25, ps 13 ANSWER 5.02 (2.50) Assumptions: Lambda = 0.1 see -1 (siven) o c g ea cbga 6(oc e Beta Bar = 0.0054 (0.25) g o n el ,7{ J t ' 3 Alpha T = -1 X 10 E-4 (0.25) 4 ~
Period = 85 see (siven) a b su rn h 'a ndC
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(0.5)
T= (B p)/(Ap): solve for p: p=E : Co^5 "" A p = 0.0054 / 1 + 85 X 0.1: p = I-4 delta K/k (1.0) NOTE: The second part will be graded int; pendently of the first par delta T mod = p / alpha T
= 5.7 X 10 E-4 delta K/K / 1 X 10 E-4 delta K/K des F =
5.7 des F
Hoderator Temp = 160 des F + 5.7 des F = 165.7, des F (1.0) REFERENCE SSES Reactor Theory SCO23 A-4 rev 0, ps 3i Reactor Theory / reactor control, SCO A-6 rev 0, ps 4 SCO23 A-7, Rev 0, ps 7 ! !
. . . . . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 18 ---- -------------------------------------- --------------
ANSWERS -- SUSOUEHANNA 1&2 -86/01/14-KING, ANSWER 5.03 (3.00) A. At 500 des [0.253, As moderator temperature increases, neutron leakage out of the fuel bundles is increased, thus the control rod is exposed to higher neutron flux and rod worth increases. [0.753 (1.0) B. The withdrawn rod [0.253, Neutron flux is higher in this area, thus rod worth is greater. [0.753 (1.0) C. At 10 [0.253, The void content of the upper portion of the core is high at operating conditions, the effects of deep control rod with-drawal can be substantial radially. The negative power effect of increased voids above the control rod is not seen because the channel length above the control rod is relatively short. [0.753 (1.0) REFERENCE Susquehanna Rea~ctor- Theory SCO23 A-6 rev 0, ps 9 & 10 and attachment A, pf 1 & ANSWER 5.04 (3.00) Remains constant E0.25 Flow is controlled by the HPCI flow controller which will attempt to maintain a constant output flow regardless of reactor pressure CO.75 (1.0) Decreases [0.25 The flow controller functions to maintain a constant flow, thus pump discharge pressure is decreased along with the decreasin3 reactor pressure to maintain constant flo OR Since the flow controller maintains,a constant flow to the reactor, as reactor pressure decreases, the pump discharge head must decrease to maintain a constant flow (constant NPSH) [0.753. (1.0) Decreases [0.253. To maintain a constant flow, turbine RPM must also decreaseEO.75 (1.0) REFERENCE SSES Fluid Mechanics * pumps, SC023 E-4 HPCI, SYO17 C-6 rev 0 r
. . - -- . . ; - . * . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 19 ---- -------------------------------------- ------------_-
ANSWERS -- SUSOUEHANNA 1&2 -86/01/14-KING, ANSWER 5.05 (3.00) 1-The NPSH available decreases due to the lower density as the water heats up (with constant level in the vessel). There is no feedwater-at this tim (1.0) 2-Recirculation pump speed is constant at 20%. Power is increased by pulling rods. Feedwater flow increases so subcooling increases als (1.0) 3-Power is increased by increasins pump speed. Core flow is increased faster than steam flow increases. More hot water is recirculated so subcoolins decreases causing the NPSH available to decrease als (1.0) l REFERENCE SSES PCO HTX & Fluid Flow, Attachment C-NPSH, ps. 2 '
*
ANSWER 5.06 (2.50) Using P = Po e to the t/T then P = 75 e to 60/-80 P = 75 e to -0.75 = 35 on RanSe 4 [1.53 On down power transients, the rate of power change is limited by the rate of decay of the lonSest lived precursors,thus retardin3 the rate of power decrease.[1.03 (55.6 see half life) REFERENCE SSES Rx Theory, SCO23 A-4, Section 9 ps. 4 , l
. .. * . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 ---- -------------------------------------- -----_----_---
ANSWERS -- SUSOUEHANNA 1&2 -86/01/14-KING, ANSWER 5.07 (3.00) a.- Power (Local power, Flux, or Local flux)
- Flow - Pressure . . - - Inlet subcoolins (3 of 4 req. 0 0.333 each) (1.0) . MAPRAT is the ratio of APLHGR(act) to MAPLHGR(LCO).EO.75] NO E0.53 3. The clad temperature can exceed 2200 des. F during a DBA LOCA.EO.753 (2.0)
REFERENCE l SSES Thermal Limits, SCO23 G-3 l x .-^ l ANSWER 5.08 (2.50) j a. Doppler or fuel temperature b. Void l c. Moderator temperature [ d. Void e. Moderator temperature (0.5 each) (2.5) REFERENCE l SSES Rx Theory, SCO23 A-7, pg. 11-12 .
-ANSWER 5.09 (1.50's YES.EO.53 The period depends upon Beta and reactivit Beta chanses I
over cycle life, so the same reactivity change produces one period at BOC and a shorter period at EOC.E1.03 (1.5) REFERENCE SSES R0 Requal exam #84-3 question ti.08 l
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5. - THEORY OF NUCLEAR POWER PLANT ~0PERATION, FLUIDS, AND PAGE 21
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ANSWERS -- SUS 00EHANNA 182 -86/01/14-KING, ANSWER 5.10 (2.00)- A. Shutdown Marsin shall be the amount of reactivity by which the reactor is soberitical or would be soberitical assumin3 all control rods are fully inserted excePt for the most reactive ( fj rod. (SSES T.S. assume 68 F & Xenon free) ( 1. 0-) c.n , c y
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REFERENCE GA kJtnaar .seggc.. m.msae SSES Reactor Theory, SCO23 A-6, pg. 6&7 o o r p,4 a,, g m ,,,,, g, ,. g m, q {,,g cJ
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. .
' . PLANT SYSTEMS' DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22 ------------------------------------------------------
ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, ANSWER 6.01 (3.25) High drywell. pressure (1.72 psis) (0.25) 2. Low reactor water level (+13 in.) (0.25) 3. Low reactor water level (-129 in.) (0.25) 4. Completion of a timer (102 sec.) (0.25) RHR A or C or Core Spray A and C running (125/145 psis)(RHR/CS) (0.25) 1&3 An increasin3 drywell pressure and a decreasing reactor water level are indicative of a breach in the reactor coolant pressure boundar (0.5) 2. Is a confirmatory level to prevent spurious ADS initiation if any one of the level 1 detectors fail downscal (0.5) 4. The timer allows sufficient time for the HPCI System to initiate and recover reactor water level before allowins ADS to initiat (0.5) 5. RHR and CS are to insure at least one of the RHR or the CS pump are runnin3 and is capable of delivering water to the reactor to make up for the loss of coolant inventory created by the openin3 of the ADS valve (0.5) REFERENCE Susquehanna, ADS, SYO17 C-4, Rev 0, 12/16/82 ps 7 & 11 ANSWER 6.02 (1.50) This fault will produce two effects * 1. The speed control loop attempts to increase senerator speed to maximum.EO.53 2. The voltage control loop sees a loss of its demand si3nal and reduces senerator voltase.CO.53 TC 5 ,t @ ' This results in a Senerator lockout.CO.5] (1.5) L_.__3 c, r eve cvi<cOL ',np
-
I REFERENCE Susquehanna, Recire Control, SYO17-L9, Rev 0, 11/17/83, ps 4, 5, 6, 7 and fisure 2 (specific objectives) _ - _ _ _ _ _ _ _ _ _ _ _
, . _ - . - - .-. -. .. * . - . . , PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 23 ------------------------------------------------------ c ANSWERS -- SUSQUEHANNA 182 -86/01/14-KING, M.
i ANSWER 6.03 (3.00) 1. RHR pumps
'2. CS pumps f3OID 1
"kv T I r ',
l,." # #k on bO 4: LL M I ".' ' 3. ESH pum est+tc- .4 s s Te rc ~><; eo nle . 4. Emersency Switchsear Load Center Cooler's and ESSW pump House RHR SW pump fan (if previously runnins) 5. Control Room Emersency Outside Air Supply System (CRE0 ASS) train Control Structure Chilled Water Loop Pump 'A' Control Structure Chiller B (5 0 0.6 ea) (3.0) i ! REFERENCE l Susquehanna, AC Dist, SYO17-C5, Rev 0, 08/23/82, ps 34 l l ANSWER 6.04 (3.00) Contributes to ADS initiation-confirmatory
- Closes RHR Hx Sample Valves Isolates RHR discharse to main condenser Closes RHR shutdown cooling valves, RPV suction and head spray Contributes to closure of RHR RPV injection valves (when in shutdown coolins mode) Reduces Recire MG set speed to 30%
i Isolates DW floor and equipment drain valves Star.ts Primary Containment pressure recorders Shifts Post Accident Monitoring System recorders from slow to fast spee . Tip withdrawal and isolation.
- 11. Suppression pool clean-up pump isolation i
12. 18' setpoint applied-to FHCS (6 9 0.5 ea) (3.0) REFERENCE i Susquehannar Rx Vs Inst, SYO17-J2, Rev 0, 12/17/82, Table 2, ps 32 l l I
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6.- PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 24 ______________________________________________________ ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, ANSWER 6.05 (3.00)
. C201 systems' - RHR-LPCI loop B - RCIC - 3 non-ADS SRV's, A, B&C - B.& D ESW pumps - D RHRSW pump - Containnient Instrument Gas - B Recire-Pump Suction Valve (5 required at 0.5 each) The loss of indication (s) for effected system and by Control Roon. Bypass Indication System Annunciator.CO.53 . REFERENCE SSES-0P-100-001, EO-100-009 & E-185 sheet 10 Bypass Ind. Sy ANSWER 6.06 (2.50)
a. 1. . No 00.25] The interlock between F059 and F045 will Freuent F045 from openin3 unless F059 is full open.CO.53 2. No CO.25] The light signifies no power to the speed control loop, RCIC will have no speed control & will O/S on startup.CO.53 (1.5) 6.~1. overspeed elec. (110%) 2. overspeed mech. (125%) 3. n.anual (2 0 0.5 ea) REFERENCE GSES RCIC SYO L-5, p , 4, 8-9 & EO-100-009 pg. 5 ANSWER 6.07 (2.25) Unit 1 - The 'C' RHR pump will.stop.CO.53 Unit 2 - All four RHR pumps start.CO.53 This is due to the LOCA/No LOCA Logic.CO.53 (1.5) Any pump may now be used for S/D Cooling.[0.753 (0.75) REFERENCE SSEG 12/12/03 exam & RHR, SYO17 C-1 rev 1, attachment A, pg 2 -
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0 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25
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ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, M.
' ANSWER 6.08 (3.00) Liq. Radwaste trips: L 1. Hish Rad i 2. Downscale/Inop 3. Low sample. flow 4. Low flow from cooling tower browdown line (3 required 0 0.5 each) b. Actions on Refuel Floor Wall Exhaust trip:
- 1. RB Zone III recire fan A/B starts 2. SBGT train A/B starts
- 3. Control Room Emergency Outside Air fan A/B starts 4. Isolation of RB Zone III dampers 5. Open RB Zone III Recire Damper (3 0 0.5 each)
i REFERENCE l SSES Process Rad Monitoring Sys, SYO17 B-2, pg. 4&8
ANSWER 6.09 (1.00) " 1. SLCS initiation , Non-re3enerative HTX outlet temp HI (140 F) i REFERENCE j SSES RHCU LP SYO-17 L-1 pg. 13 i i ANSWER 6.10 (1.50) A. yes (0.50) UI * tLO
!
- Demineralized water storas (1 $ l 65 '
2. Refuelins water storag . RHR service wate gn3c,, c c c < P3.*t 4'a rapo,
4. Emersency service wate . Fire protection system usin3 fire hose (4 reqd. e 0.25 each) .4 REFERENCE SSES, Lesson Plan , SYO17, Fuel Pool Cooling & Cleanup. Specific learning objectives 4 4 & l l
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. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 26
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~~~~Rd65ULUU5UIL CUUTR5L ____________________ ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, M.
ANSWER 7.01 (3.00) . Bus 1A (1A201) 2. Bus 2A (2A201) 3. Bus 1C (1A203) 4. Bus 2C (2A203) (4 0 0.25 ea) (1.0) B. 1. All tie bus lockouts reset 2. Control switches in NAT position 3. Unit auxilary busses 11A & 11B not already being feed by SU bus 1 (3 0 0.33 ea) (1.0) C. 201 Bus 1A bkr & 2A bkr 211 Bus 1C bkr & 2C bkr (4 0 0.25 ea) (1.0) REFERENCE Susquehanna, Loss of Startup Bus 10, Automatic Actions, ON-003-001, Rev 0, pg 2 & 3 ANSWER 7.02 (1.50) A. Take action to reduce power to within the license limit (3293 MWth) (0.75) B. 15 minutes (0.75) REFERENCE SSES ON-00-004 REV 3 PC2 ANSWER 7.03 (2.00) Restart of an idle loop recirculation pump while at power will result in an insertion of positive reactivity. Consult with the Reactor Engineer to ensure that PCIOMR and thermal limits will not be violated during the power increas (0.75) To reduce the probability of thermal binding and/or pressure lockin (0.75) Ventilation must be secured to prevent moisture accumulation in the MG set (0.5) REFERENCE SSES OP-164-001 Rev 4, 3.3.2 3, pg 8 3.3.2.h, p3 9
]
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. . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 27
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~~~~RAU56LUU UAL'UUNTRUL
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ANSWERS -- SUSQUEHANNA 182 -86/01/14-KING, .4.13, pg 14 y 9975 C ., ,f e e ' rj f- T (? - " ~ ANSWER 7.04 (3.00) casd doje assv av s a D Clad failure (0.5) gn(d Jo eg3 M " "' * \l" #"\~"" #_ # ATWS-Failure to SCRAM from Main Steam Line High Radiation (0.5) (1.0) EO-100-101, Entered on any reactor scram or scram si3nal 6 "Y , EO-100-102, Entered because scram si3nal did not scram rods a o i /A 5 y EO-100-105, Entered because of valid radioactive releaso a v s p or i ; Other procedures may be applicable and will be considered ,50 c1 OV justification for entry is correc Crt W r o lo i ro i tA (3 0 0.66 ea.)
REFERENCE SSES EO-100-101 EO-100-102 EO-100-1051 er f-Alarm Response Window Box 03&O4 7 gE gb d d.2 RJ t i- w a ie s Iwo ucce ANSWER 7.05 (3.00) u,m s, c y, d, r----ll~ c d [< e d W . TAF is approximately 117 feet. Core should be submerged.[0.75] ( g Possible breach of primary containment boundary due to increased ' loading from the added water and flooding above the highest containment vent elevation.EO.753 "c r- (1 5) The active fuel is covered with liquid or a two phase mixtur ECCS flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembl Steam flow is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembl (2 of 3 at 0.75 each) (1.5) REFERENCE SSES EO-100-103 rev 0, pg 15 EO-100-102 rev 0, pg 16
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1 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 28
~~~~EA656EU55EAE C5 TRUE~~~~~~~~~~~~~~~~~~~~~~~ ; ____________________
ANSWERSE-- SUSQUEHANNA~1&2 -86/01/14-MING, ANSWER 7.06 (2.00).
The RCIC turbine should be tripped and restarted as required
< - an operated at greater than 2200 rpm to control level 13-to 54 inches. The RCIC-(and HPCI) turbine should (1.0)
not'be operated at less than 2200 rpm to insure proper (0 5) oil pressure and flo (0.5) i REFERENCE SSES EO-200-100 rev 0, ps 3 of 9
: '-
ANSWER 7.07 (2.50) ADS used for rapid depressurization (0.00) preferred * a. designed system use
, b. ADS valves.more reliable c. discharse distribution (temp. concerns)
, d. setpoints hisher (high prob of not yet used)
(3 0 0.15 ea.)
a SRV used to control RPV pressure (in alpha sequence) (0.80)
, preferred * -a. even heat distribution b. saves ADS accum. pressure c. SRV' actuation distributed between all SRVs (3 0 0.15 ea.)
- . -
REFERENCE -- SSES EOP EO-200-100 rev 0, pg 5
b ANSWER 7.08 (2.00)
' Supplies additional scram signal for (10 secs) for (1.0)
backup scram signal, Bypass 8618 HSIV closure signal to maintain steam flow (1.0) path to condenser and steam flow to RFP REFEREMCC SSES EDP EO-200-101 rev 0, ps 2 . l .
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. _ __ _ _ _ . . . . , : . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29 - ~~~~~~~~~~~~~~~~~~~~~~~~ ~~~~R dD 5t55iCAt C UTR5L
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-ANSWERS -- SUSOUEHANNA 1&2 -86/01/14-KING, M.-
/decdn JUe ocup h; -Q k .' ,
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ANSWER 7.09 (2.00) 0JU .5-decade (0.5) If a 0.5 decade overlap is not met, startup must be halted until overlap is achieve (1.0) b . Sht-ee (0.5) -- T A vr REFERENCE Susquehanno, Plant Startup and Heatup, GO-100-002, Rev 0, pg 14 & 15 ! ANSWER 7.10 (2.00) Removin3 Extraction Steam 1. To limit errosion and/or vibration damnase to the remainins operating heaters. (0.50) The system design will not accomodate the increased flows and feedwater temperature dro (0.50) 3. No isolation valves exist in the piping. (0.50) B. To limit possible feedwater temperature transie..t (0.50) i REFERENCE l OP-147-001, pg. 9-11.
ANSWER 7.11 (1.00) A. To assure 2/3 Core Coverage following a LOCA . (0.50) l B. To provide Turbine Overspeed Protectio (0.50) REFERENCE l SSES, Procedure GO-100-003, ps. 4; SSES T/S pg. B 3/4 4-1 ANSWER 7 12 (1.00) Flow must be increased to > 3000 spm as quickly as possible to prevent ' the minimum flow valve from opening, (10 sec. time delay), and pumping vessel water to the suppression poo (1.00) i l
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-7.- PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 30
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____________________ ANSWERS -- SUSQUEHANNA 182 -86/01/14-KING, u REFERENCE OP-149-002.ps. 17
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l 8.- ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31
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l ANSWERS -- SUSQUEHANNA 1&2 -86/01/14-KING, F ANSWER 8.01 (2.00) A. true C. false B. true D. true (.5 each) (2.0) j REFERENCE 10 CFR 50.72 ' ANSWER 8.02 (2.50) 1. Whole body dose rate: >/= 100 mrem /hr 2. Work in any area greater than 2.5 mr/hr which will result in significant exposure (0.02 man / rem) l 3. Work in a posted contamination area >/= 10,000 dpm/100 cm E2 ! Use of air operated equipment or tools, welding, cutting, scinding, sandin3, millin 3 etc. of contaminated >/= 1000 dpm/100 cm E2 l components or facilities.
i 5. Opening of systems or components that contain or have contained radioactive materials and/or jobs for which Health Physics surveillance or ALARA controls are necessar . Work in a posted airborne radioactivity area of >/= 25% Maximum Permissible Concentration (MPC)). , l 7. Use of vacuum cleaners on contaminated systems, equipment or components or' opening of such vacuu . As deemed necessary by Health Physic (5 0 0.5 ea) (2.5) REFERENCE l Susquehanna, Acess Control and Radiation Work Permit System, -
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AD-00-705, Rev 6, Sec 6.9.1, pg 15 l '
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. ' .. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 32 ----------------------------------------------------------
ANSWERS -- SUSQUEHANNA 182 -86/01/14-KING, , ~ , .L--
,
wre: 7.~ s . 3 3. s . .' s AI fp.m v,, c/;
/./. , Ai4 r e yf.p.c., n.g d h p
.
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-ANSWER 8.03 (3.00) sr'" e a-1.... ,3 r,- 3. g e
a. Yes.(0.5) Declare the Emergency Diesel Generator inoperablei g [fy 8 6 Because this undervoltage relay would disable the fast transfer to the other off-site circuit and the auto start feature of the associated diesel generator, Tech. Spec. 3.8. (0.5) would be applicabl The following action must be taken: 1. Demonstrate the operability of the remaining AC sources by performing Surv. Requirements 4.8.1.1.1.1 within one hour and 4.8.1.1.2.a.4, for one diesel generator at a time, eithin
'
four hours and at least once per 8 hours therafte Restore one of the above required AC sources to Operable status within 12 hours or be in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the followin3 24 hour Both off-site sources and all D/Gs most be restored to an Operable status within 72 hours from the time of the initial loss or be in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the following 24 hours. (action a) (1.0) ,
'
2. With one diesel generator of the above required A.C. electrical power sources inoperable, in addition to action a or b, verify
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within 2 hours that all required system, subsystems, trains, components and devices that depend on the remainin3 diesel generators as a source of emergency power are also operable; otherwise, be in at least hot shutdown within the next 12 hours and in cold shutdown within the following 24 hours. (action c) (1.0) REFERENCE Susquehanna, Technical Specifications 3/ ANSWER 8.04 (3.00) NO.(1.0) You must have Suppresion Pool water temp less than 90 degrees F before enterin3 Operational Condition 2(1.0) because Tech. Spec. 3. does not allow you to enter an operational condition while relying on an action statement.(1.0) REFERENCE Susquehanna, Technical Specification 3.0.4 and i Technical Specification 3.6.2.1 ,
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. . . ADMINISTRATIVE PROCEDURES, CONDITIONS,.AND LIMITATIONS PAGE 33 ---------------------------------------------------------- [ ANSWERS -- SUSQUEHANNA 182 -86/01/14-KING, i ANSWER 8.05 (3.001 , Technical Specifications, Table 3.3.1-1 require that there be a minimum of two (2) operable channels per trip syste In this situation, RPS Channel A does not meet the specification.(1.5)
The action sta_tement requi.res tha.t you be in Startup within A hour BUT - LCO 3.3.1 allows that if the minimum number of operable channels
is not met in one channel, that that channel may be placed in the
! tripped condition within one hour. (1.5)-
THEREFORE -~in this condition the plant may continue to operate with
- a half-scram in on RPS Channel 'A'.
'
REFERENCE 3 Susquehanna, Technical Specifications - Section 3/4.3, INSTRUMENTATION
,
' ANSWER 8.06 (3.00) Due to the f act -that this instrument functions as one of the Core Spray
! and RHR (LPCI HODE) Permissive signals, the minimum operable channels per trip system requirements of T.S. Table 3.3.3-2 for the Core Spary
, ' and RHR/LPCI systems cannot be met for one trip syste .Therefore, in accordance with T.S. 3.3.3.b, the action of Table 3.3.3-1 must be taken.
" This action requires that the Core Spray and RHR/LPCI systems be declared INO Having both Core Spray and RHR/LPCI INOP requires that the actions of T.S. 3.5.1 a.1 and b.3 be me Since these cannot be met T.S. 3.0.3, , requirin3 commencement of a shutdown within one hour, applie GRADING KEY: -1.0 points for identifyin3 the requirements of Table 3.3.3-1-1.0 points for idenifying the Action requirements of T.S. 3. .0 points for identifying the requirements and actions of i T.S. 3.5 & 3.0.3 l REFERENCE l SSES Technical Specifications 3.3.3-1 and 3 3.5
i J l I I i
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p ' ' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 34
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ANSWERS -- SUSQUEHANNA 1&2 s-86/01/14-KING, M.
' ANSWER 8.07 (1.50) Evaluate as to whether an occurence requires verbal / written NRC report or is of " Sis Public Interest' 2. Notify duty mana3er of occurences requiring Immeadiate or Prompt NRC notificatio . Makins the Immediate or Prompt notificatio (3 0 0.5 ea) REFERENCE SSES Admin, AD-0A-424 rev 4, ps 7 ANSWER 8.00 (2.00) 1. Safety Related or T.S. sys or component is required to be~out of servi _c . Blockins requested or permit holder 3. Electrical devices are used for blockin3 (2 9 1.0 ea.)
REFERENCE SSES Admin, AD-GA-306 rev 4, pg 7 ANSWER 8.09 (2.50) it An individual should not be permitted to work _more than 16 hours strai3ht-[0.53, exetuding shift turnover tim . An individual should not be permitted to work more than 16 hours in any.24-hour period E0.53, nor more than 24 hours in an 48-hour period CO.53, nor more than 72 hours in any seven day perio'd [0.53, all excludins shift turnover tim . A break of at least eight hours should be allowed between work periods [0.53, including shift turnover tim (2.5) REFERENCE Susquehanno, Conduct of Operations, AD-0A-300, Rev 4, Sec 6.1.3, p3 15 E.16
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. . ,, e' ~ . , .? ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 35
__________________________________________________________ ANSWERS - SUSQUEHANNA 182 -86/01/14-KING, ANSWER 8.10 (1.00) a..To ensure secondary containment integrity will not be compromised. (0.5)
'b. FALSE (0.5)
REFEREN CE-;;. . ,.. . Susquehanna,.Phimary Containment Access and Control, AD-GA-309, Rev. 6 Sec. 6.5, ps 11 and 1 ANSWER 8 11 (1.50) a. Indicates a condition when the platform is over the reactor, I a control rod is withdrawn, and the grapple is loade (0.50) b. This lamp lishts only if the normal maximum up limit fails and the hoist is stopped by the backup hoist limit switc (0.50) I c. This occurs when a fuel assembly load is on any hoist and I refuel switch,4L is activated when the refuelin3 platform is over the vesse (0.50) REFERENCE SSES. Fuel Handling System SYO17 M-2, pg. 18 and 19 ANSWER 8.12 (3.00) 3.3.4.2.a and reqd e 0.5 each
.** 3.2.3.a e 2 A '# " "
0.5 l 3.3. and acti .i 6 of table 3.3.1-1 action 6 of tab ~; 2 reqd 9 0.5 each { 6 most limitin3 (15 mins) REFERENCE SSES Tech Specs as noted in answer , l l t
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TEST CROSS REFERENCE PAGE 1 QUESTION VALUE REFERENCE
- ________ ______ __________
05.01 2.00 AXA0000073 05.02 2.50 AXA0000074 05.03 3.00 AXA0000075 05.04 3.00 AXA0000076 05.05 3.00 AXA0000089 05.06 2.50 AXA0000090 05.07 3.00 AXA0000091 05.08 2.50 AXA0000092 05.09 1.50 AXA0000093 05.10 2.00 AXA0000103 __.___ 25.00 06.01 3.25 AXA0000077 06.02 1.50 AXA0000078 06.03 3.00 AXA0000079 06.04 3.00 AXA0000080 06.05 3.00 .AXA0000095 06.06 2.50 TAXA 0000096 - 06.07 2.25 AXA0000097 06.08 3.00 AXA0000098 06.09 1.00 AXA0000115 06.10 1.50 AXA0000122 ______ 24.00 07.01- 3.00 AXA0000082 07.02 1.50 AXA0000099 07.03 2.00 AXA0000100 07.04 3.00 AXA0000101 07.05 3.00 AXA0000102 07.06 2.00 *~ AXA0000104 07.07 2.50 AXA0000105 07.08 2.00 AXA0000106
.07_.09 2.00 AXA0000116 07.10 2.00 AXA0000117 07.11 1.00 AXA0000119 07.12- 1.00 AXA0000119
___-__ 25.00
'
08.01 2.00 AXA0000083 08.02 " 2.50 AXA0000084 08.03 3.00 AXA0000085 08.04 3.00 AXA0000086 08.05 3.00 AXA0000087 08.06 3.00 AXA0000088 08.07 1.50 AXA0000108 08.00 2.00 AXA0000111 08.09 2.50 AXA0000114
, -
.J-a i ,
T t TEST CROSS REFERENCE PAGE 2 OUESTION VALUE REFERENCE ________ ______ __________ 08.10 1.00 AXA0000120 00.11 1.50 AXA0000121 00.12 3.00 AXA0000123 ______ 28.00 ______ ______ " 102.00
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