IR 05000498/1993034
| ML20059H489 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 10/26/1993 |
| From: | Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20059H471 | List: |
| References | |
| 50-498-93-34, 50-499-93-34, NUDOCS 9311100100 | |
| Download: ML20059H489 (280) | |
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{{#Wiki_filter:. .- - - _ - - - _ . . _ . -- . APPENDIX A U.S. NUCLEAR REGULATORY COMISSION REGION IV , Inspection Report: 50-498/93-34 ! , 50-499/93-34 l l Licenses: NPF-76 i NPF-80 , l Licensee: Houston Lighting & Power Company l P.O. Box 1700 , ! Houston, Texas t
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l Facility Name: South Texas Project Electric Generating Station, Units 1 and 2 Inspection At: South Texas Project, Bay City, Texas Inspection Conducted: September 24 through October 1, 1993 ; I Inspectors: Ryan E. Lantz, Chief Examiner, Operations Section i Division of Reactor Safety l' Mark A. Satorius, Project Engineer, Project Section A Division of Reactor Projects i l Accompanying Personnel: Art Lopez, Examiner, Contractor i Battelle Pacific NW Labs Nancy Maguire-Moffitt, Examiner, Contractor ! Battelle Pacific NW Labs ; i l i i Approved: Qrkw Dh l0 2.b CA : John L. Pellet, Chief, Operations Section Date ' Division of Reactor Safety Insoection Summary Areas Inspected (Units 1 and 21: Routine, announced inspection of the qualifications of applicants for operator licenses at the South Texas Project facility, which included an eligibility determination and administration of comprehensive written and operating examinations. The examination team also observed the performance of on-shift operators. and plant conditions incident to the conduct of the applicant evaluations. The examiners used the guidance provided in NUREG-1021, " Operator Licensing Examiner Standards," Revision 7, Sections 201, 202, 203, 301, 302, 303, 401, 402, and 403, issued January 199 DR ADDCK O .'
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' l Results (Units 1 and 21:
* Four of the six applicants for reactor operator licenses satisfied the requirements of 10 CFR 55.33(a)(2) (Section 1.6).
- Eight of the nine applicants for senior reactor operator licenses ;
satisfied the requirements of 10 CFR 55.33(a)(2) (Section 1.6). !
* The reference material provided by the training department for i examination development was adequate (Section 1.1). ] * All applicants passed the written examinations, with scores ranging from I a low of 82 percent to a high of 94 percent with averages of 86 percent for reactor operator applicants, 90 percent for senior reactor operator applicants, and 88.4 percent overall (Section 1.2).
- The crews examined exhibited generally effective, formal communications, !
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with effective command and control on the part of crew supervision, with noted exceptions (Section 1.3.1).
- The applicants demonstrated a generic performance weakness which involved a hesitancy to secure equipment when abnormal conditions were noted immediately following equipment startu '
The applicants demonstrated a second generic performance weakness which ! involved a general unfamiliarity with low power and shutdown procedures ,
(Sections 1.3.1,1.3.2).
- Procedural guidance for loss of primary reactor coolant accident scenarios while shutdown was unclea Procedural guidance for abnormal response of a reactor coolant pump when starting was lacking (Sections 1.3.1,1.3.2).
- Poor plant labeling was observed to adversely impact operator performance and was consistent with prior NRC inspection reports (Section 1.3.2).
- General observations were made of on-shift control room operators and plant material conditions (Section 1.4).
Summarv of Inspection Findinas:
* There were no findings that were assigned a tracking number identified during the course of this inspectio l I
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Attachments:
- Attachment 1 - Persons Contacted and Exit Meeting
* Attachment 2 - Simulation Facility Report
* Attachment 3 - Written Examination Keys I
* Attachment 4 - Facility Post-Examination Review Comments
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I LICENSED OPERATOR APPLICANT INITIAL QUALIFICATION EVALUATION (NUREG-1021) ) I During the inspection, the examiners evaluated the qualifications of ; 15 license applicants; 6 for reactor operator (RO), 5 for senior reactor ; operator (SRO) currently licensed as R0s, and 4 for instant SRO. The ,
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inspection assessed the eligibility and administrative and technical l competency of the applicants to be issued licenses to operate and direct the ) operation of the reactivity controls of a commercial nuclear power facility in a
REGION 4 l CANDIDATi'S NAME: FACILITY: South Texas 1 & 2
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REACTOR TYPE: PWR-WEC4 DATE ADMINISTERED: 93/09/27
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INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination start ; CANDIDATE'S ; TEST VALUE SCORE % 100.00 % TOTALS l FINAL GRADE All work done on this examination is my ow I have neither given nor received ai Candidate's Signature i
I REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan MULTIPLE CH0 ICE 023 a b c d 001 a b c d 024 a b c d
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002 a b c d 025 a b c d 003 a b c d 026 a b c d 004 a b c d 027 a b c d 005 a b c d 028 a b c d 006 a b c d 029 a b c d i 007 a b c d 030 a b c d 008 a b c d 031 a b c d 009 a b c d 032 a b c d 010 a b c d 033 a b c d 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d l 016 a b c d 039 a b c d l 017 a b c d 040 a b c d l 018 a b c d 041 a b c d 019 a b c d 042 a b c d 020 a b c d 043 a b c d 021 a b c d 044 a b c d 022 a b c d 045 a b c d l l
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ANSWER SHEET l l Multiple Choice (Circle or X your choice) i If you change your answer, write your selection in the blan l 046 a b c d 069 a b c d
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047 a b c d 070 a b c d
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048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d 051 a b c d 074 a b c d 052 a b c d 075 a b c d 053 a b c d 076 a b c d
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054 a b c d 077 a b c d 055 a b c d 078 a b c d
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056 a b c d 079 a b c d ! 057 a b c d 080 a b c d i 058 a b c d 081 a b c d 059 a b c d 082 a b c d j 060 a b c d 083 a b c d 061 a b c d 084 a b c d l l 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b c d l 065 a b c d 088 a b c d 066 a b c d 089 a b c d 067 a b c d 090 a b c d 068 a b c d 091 a b c d , I
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ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan a b c d 093 a b c d
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094 a b c d 095 a b c d 096 .a b c d 097 a b c d 1 098 a b c d l 099 a b c d , 100 a b c d
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Policies and Guidelines for taking NRC Written Examinations During the administration of this examination the following rules apply: 1. Cheating on the examination will result in a denial of your application and could result in more severe penaltie . After you complete the examination, sign the statement on the ' f cover sheet indicating that the work is your own and you have not
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- received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80% or greate r
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4. The point value for each question is indicated in parentheses after the questio . There is a time limit of four (4) hours for completion of the examinatio . Use only black ink or dark pencil to ensure legible copie . Print your name in the blank provided on the examination cover " sheet and the answer shee . Mark your answers on the answer sheet provided and do not leave any question blan . If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted. but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids and answer , sheets and give it to the examiner or procto Remember to sign the statement on the examination cover shee . After you have turned in your examination, leave the examination area as defined by the examine _ __
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> REACTOR OPERATOR Page 7 . i I QUESTION: 001 (1.00) WHICH ONE (1) of the following describes the AUTOMATIC response of the Fuel Handling Building (FHB) HVAC system upon receipt of an FHB Exhaust HIGH RADIATION alann on RT-8035? All Main Supply fans START and the Relief Supply dampers CLOS All Main Exhaust and Exhaust Booster fans STOP and the Exhaust ' System dampers CLOS All Pain Exhaust and Exhaust Booster fans START and the Exhaust - I System shifts to align two filtering unit train All Main Supply fans START and.the Relief Supply dampers shift to ; align two filtering unit train QUESTION: 002 (1.00) WHICH ONE (1) of the following is the preferred method for transcribing initials from one working copy of a procedure to another when an original corking copy of a procedure becomes contaminated? The Unit supervisor should initial the new procedure for the ! individual who entered the data on the original working cop The person who entered the data on the original working copy should initial the new procedur j The on duty RO should initial the new procedure after another operator locally independently verifies the procedure is complet The Shift supervisor should initial the new procedure for the I individual who entered the data on the original working cop ' I
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I ; QUESTION: 003 (1.00) i
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! WHICH ONE (1) of the following is a proper method for independently l-verifying the position of a valve that is required to be in a THROTTLED ! position?
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., The valve should be moved slightly in the open direction and then j back to its original positio .
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! The valve should be moved only in the closed direction and then.
, back to its original positio l The valve position should be adjusted while monitoring flowrate , I through the valve from the control roo i The valve should not be moved, the position is verified by - observing the valve stem positio j
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QUESTION: 004 (1.00) : j WHICH ONE (1) of the following conditions requires the use of double valve protection when isolating a system or component? l
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' The DWST inlet valve has a packing leak that must be repaire $ A spent fuel pool filter is to be isolated for replacemen The hydrogen side seal oil (HSS0) cooler is to be isolated for a f s
tube lea A main steam line pressure transmitter is isolated for replacemen ! .
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REACTOR OPERATOR Page 9 QUESTION: 005 (1.00) , WHICH ONE (1) of the following individuals or positions is responsible for notifying the Issuing Authority that a clearance has been released? Unit Supervisor ' The Acceptor's imediate superviso ~ The first operator to perfonn duties on the EC The last operator to perform duties on the EC , QUESTION: 006 (1.00) Given the following: I - Unit 1 is in MODE A Reactor Containment Building (RCB) entry is planned.
f ' - OPGP03-ZO-0032, " Reactor Containment Building Entry" has been initiate WHICH ONE (1) of the following is the MINIMUM number of personnel that are l l required to be assigned for the RCB entry? ! i l l l f !
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Page 10 REACTOR OPERATOR- ] I i I QUESTION: 007 (1.00) t WHICH ONE (1) of the following evolutions requires the Shift Supervisor's j l presence in the "AT THE CONTROLS" area per the Operations Policies and Practices Manual?- I i Synchronization of the main generator.- l l Perfoming an SSPS surveillance tes ,.
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: Placing the Main Turbine First Stage Impulse-Pressure loop into servic ; Placing a Main Feedwater pump in servic ;
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i WHICH ONE (1) of the following methods of communication is the Fire Brigade l leader required to use when communicating with the control room during a j fire? .l ;
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- Portable Radio Channel 1 l
! Sound Powered Phone ;
I Emergency Phone Circuit J Plant Telephone .
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l Page 11 , REACTOR OPERATOR QUESTION: 009 (1.00)
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Given the following:
- Unit 2 is shutdown, maintaining vacuum in the main condense Main condenser hotwell samples indicate a HIGH Dissolved Oxygen conten ,
WHICH ONE (1) of the following chemicals should be added to the system to , ' reduce the dissolved oxygen content? , Hydrazine Ammonia Hydrogen Ammonium Hydroxide QUESTION: 010 (1.00) WHICH ONE (1) of the following colors on the plant computer is used to indicate when a component is in an energized (operating) condition? Cyan Magenta Red , i Blue j l i I l l l
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QUESTION: 011 (1.00)- t , WHICH ONE (1) of.the following is the time limit for notifying the State of- j - Texas and Matagorda County after the Emergency Director has initially ; classified an event? ! l minutes j minutes , ' i I hour i
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l QUESTION: 012 '(1.00)- ! l WHICH ONE (1) of the following is used to calculate the Tref signal? l
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i Auctioneered high T-ave, j
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, Averaged T-ho Turbine first stage impulse pressur .] s QUESTION: 013 (1.00) WHICH ONE (1) of the following rod control system conditions will cause a NON-URGENT failure alarm at the power cabinet? Lift coil disconnect switch OPE Loss of multiplexer signal, Loss of one (1) power suppl Bridge thyrister faul j - 6 --- , - - _ . r,#-- .-mi -.-, ,,w m_ rv - , y y e
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QUESTION: 014 (1.00) Given the following: f
- Unit 1 is at 100 % powe The Rod Control STARTUP switch is inadvertently placed in the :
RESET positio !
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- Rod Control is in AUTOMATI i
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WHICH ONE (1) of the following describes how this action will affect the . Rod Control System? Rod stop interlock C-11 cannot prevent further control bank D j outward rod motio I The plant will not respond to a dropped rod in control banks B,
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C, or I Any " Urgent Failure" alarms currently active will clear.
1 : The " ROD BOTTOM" alarm actuate . QUESTION: 015 (1.00) , WHICH ONE (1) of the following is the reason that all the RCP breakers OPEN ' on underfrequency? , i To protect the RCP from damage to_the anti-rotation device due to abnormal coastdown.
' To preserve the RCP flywheel kinetic energ To avoid water hammer transients in the RCS induced by rapid RCP speed change.
! To reduce the probability of a stress induced RCP sheared shaft accident.
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.Page 14 REACTOR OPERATOR-QUESTION: 016 (1.00)
Given the following:
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- Unit 1 is in MODE l
- RCP 1A motor is uncoupled from the pum Loop 1~is ful j
- Maintenance is working on 1A RC WHICH ONE (1) of the following prevents leakage of reactor coolant.up the
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RCP shaft? ! Nozzle dam installation prevents RCS water from entering the RCP l shaft are j Seal leakoff collects any RCS leakage up'the shaft and directs it l l back to the VC : The pump shaft mates with the top of the thennal barrier assembl ; Seal injection is maintained during this conditio j i
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j QUESTION: 017 (1.00) l
Given the following: 1
- Unit 2 has just tripped from 100% powe l - A spurious Phase A Containment Isolation Signal has actuate l - Charging header pressure is 1000 psi !
WHICH ONE (1) of the following RCP containment isolation valves will CLOSE? Seal Water Injection Isolation Valve CV-033 I CCW RCP Inside Containment Isolation Valve MOV-54 l l Seal Water Return Isolation Valve CV-07 I CCW RCP Outside Containment Isolation Valve FY-449 . _
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QUESTION: 018 (1.00) l WHICH ONE (1) of the following leaks could result in a DILUTION of the RCS? r Letdown Heat Exchanger leak Excess Letdown Heat Exchanger leak ! Reactor Coolant Drain Tank Heat Exchanger leak , Seal Water Heat Exchanger leak i f QUESTION: 019 (1.00) ' WHICH ONE (1) of the following is the reason for maintaining a MINIMUM PRESSURE of 15 psig in the VCT during normal at power operation? To ensure adequate hydrogen concentration in the RCS coolan l :
' To ensure coolant flow to RCP seal # To prevent tank collapse due to internal vacuum during a multiple CCP star To provide adequate CCP recirculation backpressure during normal ope ations.
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' QUESTION: 020 (1.00) !
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Given the following: ! i
- No ESF equipment is out of servic l - Unit 2 experiences a Safety Injection (MODE 1) signa j i - ESF Status Monitoring is indicating a lit component lamp and a l
flashing annunciato All other status monitoring indications are NORMAL for Safety , l Injection (MODE 1).
f WHICH ONE (1) of the following is the status of the ESF component? j J, l The component has been bypassed by the operato !
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! The component failed to actuate or properly positio The component has been reset by the RESET pushbutton and a valid l l l l signal is still presen ! The component is inoperable and the alarm has been acknowledged { l by the operato I
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i QUESTION: 021 (1.00) i Given the following: , l A MODE III ESF signal exists on UNIT ;
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- The operator wants to reset the sequencer on panel CP00 WHICH ONE (1) of the following conditions is NOT necessary to reset the sequencer?
l SI rese Undervoltage condition cleare Loading sequence complet Placing the sequencer control in MANUAL REMOTE TEST position.
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QUEST 80N: 022 (1.00) , Given the following: ,
- A LOCA occurred on Unit , - Normal offsite power is availabl Containment pressure is 5 psi [
WHICH ONE (1) of the following conditions will cause the sequencer to start , the Containment Spray pumps? l Containment Pressure ! Time after sequencer MODE 1 signal initiation , seconds 7.5 psig seconds 10 psig ( l ' seconds 7.5 psig seconds 10 psig . t i f i
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l l l ' REACTOR OPERATOR Page 18 QUESTION: 023 (1.00) Given the following:
- A reactor startup is in progres Reactor power is steady at 1X10E-9 amps (intemediate range).
- The reactor operator has depressed the "S" train BLOCK pushbutton for the Intermediate Range Nuclear Instrumentation (IRNI), i
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WHICH ONE (1) of the following is the reason for the reactor trip when IRNI channel N35 blows an instrument power fuse?- j The trip bistable DEENERGIZES and the Intermediate Range trip is {' active because the "R" train BLOCK pushbutton has not'been pushe , The trip bistable DEENERGIZES and the Intermediate Range trip has not been MANUALLY blocked since 2 of 4 Power Range channels are : l not above P-1 I I l The trip bistable ENERGIZES and the Intermediate Range trip is j active because the "R" train BLOCK pushbutton has not'been ) pushe j i The trip bistable ENERGIZES and the Intemediate Range trip has ') not been MANUALLY blocked since 2 of 4 Power Range channels are ; not above P-1 l l I
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QUESTION: 024 (1.00) Given the following:
! - Unit 1 is in MODE I at 100's powe I WHICH ONE (1) of the following is the most accurate method of determining reactor thermal power? Manually calculated average of four Power Range Nuclear Instrument Turbine generator output loa Proteus calculated secondary calorimetric powe Run an incore power flux ma i l
l REACTOR OPERATOR Page 19 ) { QUESTION: 025 (1.00) Given the following:
- Unit I has been operating at 80% for the last 48 hour No AFD penalty time has been accumulate The AFD monitor alarm was disabled at 060 Target AFD for 100% power is + 3.75%. - Indicated AFD has been continually monitored and logged as '
follows: TIME N41 N42 N43 N44 0600 + + + + + + + + + + + + + + + + ~ 0700 + + + + At 0700, WHICH ONE (1) of the following is the cumulative penalty ,
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deviation time for the previous hour? i a. 5 minute b. 21 minute . c. 26 minutes.
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REACTOR OPERATOR Page 20 QUESTION: 026 (1.00) Given the following:
- Unit 1 is at 100% power ' - The Digital Rod Position Indication System (DRPI) has experienced a failure of the Data "B" cabine .
WHICH ONE (1) of the following is the accuracy of a lit rod position LED ' for this condition? +4 to -4 steps +4 to -10 steps ; +10 to -4 steps P +10 to -10 steps
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QUESTION: 027 (1.00) Given the following:
- A LOCA has occurred on Unit All systems are in their normal operating alignmen , -
WHICH ONE (1) of the following is providing power to 11A and 12A Reactor Containment Fan Coolers (RCFC)? Load Center E1C
, MCC Bus IF l I Lead Center EIA MCC Bus 1G5 ,
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REACTOR OPEPATOR Page 21 QUESTION: 028 (1.00) ! :
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Given the following: i
- Reactor power is 13%. -
S/G 1A level is 78%. 1 l
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S/G 18 level is 88%. i
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S/G IC level is 76%.
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S/G ID level is 75%. , j WHICH ONE (1) of the following lists AUTOMATIC actions that should result l l from the above situation? i Turbine trip, Reactor trip, Feed Pump tri ! Turbine trip Feedwater Isolation, Feed Pump tri ! Reactor trip, Feedwater Isolation, Feed Pump tri .! i l Turbine trip, Reactor Trip, Feedwater Isolatio ;
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QUESTION: 029 (1.00) i
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Given the followirg: : i
- Unit 1 is operating at full powe ;
- The Auxiliary Feedwater and Main Feedwater (MFW) Systems are in j their normal system lineu ;
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WHICH ONE (1) of the following parameters is used by the Steam Generator ! Water Level Control System to control MFP speed? ,j Main steam header pressur MFW control valve positio MFW total flo Steam generator water level.
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! QUESTION: 030 (1.00) WHICH ONE (1) of the following will prevent ANY automatic start of a motor ! t
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- driven AFW pump? i l, Low pump suction pressur t
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- The pump discharge valve is close . I The pump control switch is in STOP on the MC l
: The transfer switch'is in the ASP position.- ;
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l QUESTION: 031 (1.00) ; , " Given the following: . i
- Unit I was in MODE 2 when a Loss of Offsite Pcwer-(LOOP) l occurre . , ' ; - The AFW pumps are running on recirculatio i
] WHICH ONE (1) of the following will cause the AFW pumps to ' start injecting f ' into the steam generators? l
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t Steam Generator 1A pressure decreases to 700 psi ' l l
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3 Steam Generator IB level decreases to 43%.
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! RCS pressure decreases to 1900 psi >
i Turbine driven AFW pump accelerates to -1800 rp . J J
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! ; QUESTION: 032 (1.00) l WHICH ONE (1) of the following is the purpose of the loop seal in the Auxiliary Feedwater Storage Tank (AFWST)? *
. Increases the useable volume of water in the tan Minimites oxygen intrusion into the tan ,
' Provides the recirculation flowpath for AFW pump testin ,
I Ensures adequate NPSH is maintained for the AFW pump l
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QUESTION: 033 (1.00) !
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WHICH ONE (1) of the following components-is the largest source of ; radioactive gas sent to the Gaseous Waste Processing System (GWPS), during .i normal full load operations?
: Volume Control Tank Boron Recycle Evaporator , Reactor Head Degassing Decay Tank i Reactor Coolant Drain Tank j
! i QUESTION: 034 (1.00) WHICH ONE (1) of the following events would result in an AREA radiation monitor alarm? RCS Leak at the Incore Seal Table Main Steam Line Break inside Containment Steam Generator Tube Rupture Pressurizer PORV seat leakuge l l l
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l QUESTION: 035 (1.00) WHICH ONE (1) of the following combinations of radiation types are detected by an Area Radiation Monitor?
; Beta and Alpha Alpha and Neutron
~ Beta and Gamma
! Gamma and Neutron l l QUESTION: 036 (1.00) ; Given the following: ! - Loop 1, 3, and 4 Tavg meters indicate 593 degrees Loop 2 Tavg meter indicates off scale HIG r
- Loop 1, 3, and 4 Delta T meters indicate 100%. , -
Loop 2 Delta T meter indicates 0%. : t WHICH ONE (1) of the following is the cause of these indications?
! Loop 2 Tcold failed LO Loop 2 Tcold failed HIG Loop 2 Thot failed LO .
I Loop 2 Thot failed HIG i
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j QUESTION: 037 (1.00) Gieen the following:
- Unit I has tripped from 100% power due a Loss of Coolant Accident (LOCA) with a loss of Offsite Power (LOOP).
- RCS pressure is 375 psi All RWST level indicators are at 15%.
- Emergency Diesel Generator #11 has failed to star ' - Low Head Safety Injection pump 1C (LHSI) has tripped on overcurren N0 flow is indicated for LHSI pump 1 All other equipment is operating normall WHICH ONE (1) of the following is the reason LHSI Pump 1B has N0 flow indication? LHSI pump Containment Emergency Sump Section valves have failed to open automatically on low RWST leve .16KV Bus ElB has no powe RCS pressure is above LHSI pump shutoff hea l Load sequencer is transitioning from Mode 11 to Mode III.
! QUESTION: 038 (1.00) WHICH ONE (1) of the following is an input to the Cold Overpressure
Mitigation System (COMS) for PCV-655A? Auctioneered high wide range Tho Auctioneered high wide range Tcol RCS pressure transmitter PT-40 RCS pressure transmitter PT-403.
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QUESTION: 039 (1.00)
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i Given the following: .l
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- All pressurizer heaters were on'IN AUTO prior to actual pressurizer level dropping below the heater low-level cutoff setpoin Pressurizer level is restored to above the low level reset poin Assume no operator action other than refilling the pressurize '
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- Assume no Safety Injection or LOOP signals presen Assume pressurizer pressure is 2200 psi l l
WHICH ONE (1) of the following describes how the heaters will respond? } Variable and backup heaters will come on AUTOMATICALL I Variable and backup heaters will remain of Variable heaters ONLY will come on AUTOMATICALL l l Backup heaters ONLY will come on AUTOMATICALL l I l l I ! l t '
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Page 27 REACTOR OPERATOR
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QUESTION: 040 (1.00) Given the following: l
- Reactor power is 99%. -
Pressurizer level is 58%.
- Letdown flow 75 gpm (One Orifice).
- A charging line leak develops near the charging line containment penetration that diverts ALL charging flow from the lin ,
- Normal seal injection is maintaine Assume N0 operator action is take WHICH ONE (1) of the following statements describes the FINAL pressurizer response? Decreasing pressurizer level to 17%, letdown isolates and pressurizer level increases leading to high level tri l
~ Pressurizer pressure increases to the high pressure trip setpoint following loss of pressurizer spray and the auto start of pressurizer backup heaters due to level deviatio i Stable lower pressurizer level following reduction of letdown flow, due to steam flashing in the regenerative heat exchange Stable higher pressurizer pressure due to heating of the water in the pressurizer, a result of losing the cooling effect of charging flo i l
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REACTOR OPERATOR-i
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QUESTION: 041 (1.00) Given the following: l
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- Unit 1 is at 80% powe _
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- A surveillance is in progress to test the reactor trip breaker i
- An operator calls the control room and reports that he found ALL reactor trip and bypass breakers RACKED IN and CLOSED.=
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WHICH ONE (1) of the following actions should be taken? .
; Immediately OPEN one trip or bypass breaker and continue.with the j surveillanc l' Suspend the surveillance in progress and restore nomal trip - ;
breaker configuratio i
! Immediately OPEN one trip and one bypass breaker on each train in a manner that will not cause a reactor trip.- -l
.. Trip the Reactor and enter OPOP05-E0-E00 !
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l QUESTION: 042 (1.00)
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! Given the following:
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- A plant startup is in progres :-
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l - Permissive status light " MANUAL BLOCK INT /LO PWR RANGE RX TRIP PERMITTED" is NOT ILLUMINATE ! Annunciator SM-024-A3, "P-13 TURB LOAD LESS T;iAN 10%" is j
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ILLUMINATE . WHICH ONE (1) of the following AUTOMATIC reactor trips could actuate and trip the reactor in the above situation? Pressurizer low pressure.
! RCP low flo Turbine tri Pressurizer high pressure.
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REACTOR OPERATOR Page 29 , QUESTION: 043 (1.00) WHICH ONE (1) of the following describes the reason Auctioneered High Nuclear Power is used as a control input to the Low Power feedwater Control Valves? Required feedwater flow is directly proportional to nuclear power at low power level It anticipates changes in heat flux into the SG at low power ' level Steam flow signals are not accurately pressure compensated at low power level Feedwater valve control based on level only will cause large flow oscillation QUESTION: 044 (1.00) WHICH ONE (1) of the following combinations of Spent Fuel Pool (SFP) makeup sources ensures if water was that added toboron concentration would be MAINTAINED or INCREASED the SFP? Reactor Makeup Water and Boron Recycle Syste Boron Recycle System and Demineralized Wate RWST and Reactor Makeup Wate Boron Recycle System and RWS . - _ - _ - _ _ _ - - - - -
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L ! REACTOR OPERATOR Page 30 j , l ! l l I QUESTION: 045 (1.00) ;
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Given the following:
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A Loss of Offsite Power (LOOP) has occurred on Unit- ,
- All control switches are in their normal alignmen EDG 23 output breaker tripped while the MODE II sequence was ;
executin MEAB operator reports no apparent reason for the breaker tri , l
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WHICH ONE (1) of the following will allow the MODE II sequence for EDG 23 to resume? I
. Depress the RESET pushbutton for the sequencer. . -( Place the MANUAL LOCAL TEST switch in the LOCAL TEST positio I
! Trip 23 DG and manually actuate a Safety Injection (SI) signa ,
)
,, Reset the MODE sequence using the key-operated MODE II logic ! ! reset switc i-i i l QUESTION: 046 (1.00) , WHICH ONE (1) of the following is the MAXIMUM length of time that the 125 VDC batteries are designed to supply ESF loads and emergency lighting , without the battery chargers? ! hours f hours-j i hours l hours _ _ _ - _ _- _ . _ .
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REACTOR OPERATOR Page 31 l QUESTION: 047 (1.00)
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WHICH ONE (1) of the following describes the power supplies to the ESF 4.16KV bus during the specified mode of electrical system operation? Preferred - Associated 13.8KV standby bus Standby- ESF DG, Emergency - Switchgear I Preferred - Associated 13.8KV standby bus, Standby- Switchgear '
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IL, Emergency - ESF D Preferred - Switchgear IL, Standby - ESF DG, Emergency - Associated 13.8KV standb Preferred - Switchgear IL, Standby - Associated 13.8KV standby bus, Emergency - ESF D . QUESTION: 048 (1.00) WHICH ONE (1) of the following types of smoke / fire detectors is used to detect a fire in the carbon filters of the HVAC units throughout the plant? Thermostatic line heat detecto Ionization smoke detecto Optical flame detector i Thermistor heat detector.
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l Page 32 . REACTOR OPERATOR l ! I r ' QUESTION: 049 (1.00) Given the following:
- Unit 1, Fire Pump #1 is in AUT All control panel switches are in their normal position BATT "A" CONNECTED and BATT "B" CONNECTED Blue lights are LI Fire system pressure is 130 psi An operator has pushed the STOP/ RESET pushbutton to stop the ,
pum The pump restarted when the pushbutton was release WHICH ONE (1) of the following conditions could be the cause of the fire , pump restart? Low fire system pressur Unit 2 has actuated a re-te start signa The control panel has lost AC power.
i The Weekly Test Timer (WTT) has actuated.
l l , l QUESTION: 050 (1.00) WHICH ONE (1) of the following actions must be taken within I hour if containment pressure is at +0.5 psig according to Technical Specification 3.6.1.4 " Containment Systems - Internal Pressure"? Take action to place the unit in a MODE that the specification - does not appl l l Initiate a containment normal purg Restore pressure to within limit Perform an RCS leak rate calculatio ! l
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- . l QUESTION: 051 .
(1.00)
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! WHICH ONE (1) of the following describes. the action associated with a HIGH - - alarm on Liquid Waste Processing System (LWPS) discharge monitor RT-8038 l
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during a release?
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l LWPS Diversion Va'.ve (FV-4077) closes and terminates all i i discharg ;
' LWPS Diversion Valve (FV-4077) diverts flow back to the Waste Monitor Tank Terminates flow to the Waste %nitor Tank , Trips the Waste Monitor Tank pump l l
QUESTION: 052 (1.00) Given the following:
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- Unit 1 is in MODE l I - RCS temperature is 325 degree i j - RCS pressure is 340 psi i
- RHR is in servic An unisolable leak in the Instrument Air (IA) system has occurre .. - IA system pressure is 60 psig and decreasin l WHICH ONE (1) of the following describes how the RHR system will respond?
! RHR heat exchanger bypass valves FCV-851/852/853 will fail OPEN l
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l and cause RCS temperature to DECREASE.
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. RHR heat exchanger bypass valves FCV-851/852/853 will fail CLOSED
" and cause RCS temperature to INCREAS RHR heat exchanger flow control . valves HCV-864/865/866 will fail- - OPEN and cause RCS. temperature to DECREAS RHR heat exchanger flow control valves HCV-864/865/866 will fail
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CLOSED and cause RCS temperature to INCREAS ,
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Page 34 REACTOR OPERATOR l l
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QUESTION: 053 (1.00) WHICH ONE {1) of the'following Service Air / Instrument Air (SA/IA) system valves will be the LAST to reposition on a DECREASING Instrument Air header l pressure? 1 SA to IA Crossover .j IA to Yard Isolation , ; SA Dryer Bypass 5 IA Containment Isolation
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-t QUESTION: 054 (1.00)
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WHICH ONE (1) of the following satisfies the conditions that must exist in order for the containment Normal Purge supply and exhaust isolation valves , to be considered " sealed" in accordance with PSP 03-SI-0016, " Containment . l Integrity Checklist"? .
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! The valves are CLOSED and the power supply breakers are LOCKED ; OPE l
: The valves are CLOSED and the power supply breakers are LOCKED l CLOSE .; The valves are LOCKED CLOSED and the power supply breakers are LOCKED CLOSE The valves are LOCKED CLOSED and the power supply breakers are'. l r
LOCKED OPE i l i i
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.j QUESTION: 055 (1.00)
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WHICH ONE (1) of the following is the reason for adding Trisodium Phosphate !
(TSP) to the containment during a Containment Spray Actuation? -{
: TSP washes iodine from containment walls during recirculation i phase of LOC .j
, l TSP prevents the sump pH from exceeding :i TSP minimizes the stress corrosion cracking of certain metallic f components inside containmen j
.) TSP eliminates the formation of hydrogen gas generation in the j post LOCA environment, e '? QUESTION: 056 (1.00)
l J WHICH ONE (1) of the following combinations of Residual Heat Removal (RHR) i loops may be used for low pressure letdown operations? j
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- Loop "A" and "B". l i
4 Loop "A" and "C".
. Loop "B" and "C".
i Loop "B" and "D". .
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> REACTOR OPERATOR- Page 36 j i I QUESTION: 057 (1.00) )
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' Given the following: ;
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- Unit 1 is at 100% powe All systems are in their normal alignmen l
- CCW Train IA is in servic !
- CCW level indicates 64% and is stabilizin j
- The STANDBY CCW pump has. starte '
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- ASSUME all automatic actions have occurred as designe ,
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WHICH ONE (1) of the following CCW system valves will.be closed at this
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indicated CCW surge tank level? " SUPPLY ISOLATION MOV-0768". " BRANCH ISOL MOV-0297". , " RETURN X-CONN FV-4657". ! "CCW RET HDR ISOL MOV-0192". f
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a $ QUESTION: 058 (1.00) ;
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Given the following: l t
- Unit 1 is at 100% powe l
- Essential Cooling Water (ECW) pump "A" is RUNNING.
- - ECW pump "B" control switch is in AUTO and the ECW/CCW TRAIN l
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2 SELECTOR switch is in STANDBY.
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ECW pump "C" is NOT RUNNING (AUTO after STOP).
~ WHICH ONE (1) of the following will automatically start ECW pump "B"? l
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1 CCW common discharge header pressure of 70 psi ECW train "A" and "C" pressures of 50 psi Placing ECW pump "B" switch on panel ZLP-654 in LOCA Essential Cooling Pond Makeup Pump "A" trip i
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I l QUESTION: 059 (1.00) WHICH ONE (1) of the following is' indicated when the " LOAD CHANNEL" ligh I on the EHC indicating panel is energized? . . A failure in the Impulse Pressure Feedback loop has occurred and the EHC system is in the IMP OUT mod A failure in the Impulse Pressure Feedback loop has occurred and ' the EHC system is in the Turbine Manual mod A failure of the Turbine Actual Reference counter has' occurred and the EHC system is in the Turbine Manual mod A failure of the Turbine Actual Reference counter has occurred and the EHC system is in the IMP OUT mod QUESTION: 060 (1.00) Given the following:
- Unit 2 has tripped from 100's powe All systems are operable and in automati Actual Tavg is 13 degrees F greater than no-load Tavg.,
WHICH ONE (1) of the following describes the response of the steam dump valves? Bank 1 Bank 2 Bank Bank 4- , !- Full open Full open Full Open Modulating Full open Full Open Modulating Closed i Full open Full Open Closed Closed Modulating Closed Closed Closed j l _ r*g y . e p .3g- --- sr *- e*+ v g
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~ QUESTION: 061 (1.00)
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! WHICH ONE (1) of the following design features provides the interlock. that prevents raising irradiated fuel with the New fuel elevator? j The upper limit switc ; The lower limit switc . A load sensing rela A radiation detecto y
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QUESTION: 062 (1.00) i Given the following: l
- Unit 1 is at 30% powe Power escalation is in progres l - Control rods are in MANUA ; - After a control rod withdrawal of several steps, the rods i i
continue outward when the IN-HOLD-0VT lever is returned to the neutral positio f WHICH ONE {1) of the following actions should be taken? -{; Place the Rod Bank Selector Switch in AUT0'and check for ! continued rod motio Trip the Unit and enter OPOP05-E0-E000, " Reactor Trip or Safety' l Injection". Check for failed instruments and select or block appropriate instrument input Hold the IN-HOLD-0VT' lever to the IN position and check for continued rod motio l l l l ! l L _ _ ._ _ . . .. l
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QUESTION: 063 {1.00) l Given the following: !
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Unit 1 is at 305 powe ! ! - Control rods are in MANUA i , WHICH ONE (1) of the following is a symptom of an immovable control rod per i OPOP04-RS-0001, " Control Rod Malfunction"? ; l ,
.. , Tavg - Tref deviation 'of 3 degrees ) " DELTA T ROD WITHDRAWAL BLK ALERT"-alan I One or more RPIs disagree..by +8 step !
y " ROD CONTROL URGENT" alar l
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QUESTION: 064 (1.00) WHICH ONE (1) of the following represents the MINIMUM values for RCP l ' bearing temperatures that require the operator to trip the reactor and any ; affected RCP, per OPOP04-RC-0002, " Reactor Coolant Pump Off Normal"? ;
, degrees F (radial bearings) or 230 degrees F (seal !
water /bearingtemperature). !
; degrees F (radial bearings) or 187 degrees F (seal l water / bearing temperature). ;
l degrees F (radial bearings) or 187 degrees F (seal ! water / bearing temperature). ! degrees F (radial bearings) or 195 degrees F (seal j water / bearing temperature). ,
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' QUESTION: 065 (1.00) J
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- Given the following
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- - A LOCA has occurred on Unit 1.
' - OPOP05-E0-E000, " Reactor Trip or Safety Injection" is being
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performe , WHICH ONE-(1) of the following indications has an alternate limit during
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, adverse containment conditions?- ,
. RCS pressur i i AFW flo l SG pressur .[ SG narrow range level, j
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' QUESTION: 066 (1.00)
WHICH ONE (1) of the following is the reason for reducing Tave to less than j
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500 degrees F following a shutdown required by a Dose Equivalent I-131 : level of greater than 1 microcuries per gram? Slows coolant / fuel reaction rate, immediately reducing the. source-term of the activit Prevents the release of activity following a steam generator tube ruptur Minimizes the temperature related degradation of'the CVCS-demineralizers while the RCS~ clean-up is in progres Minimizes of the iodine spiking phenomena which occurs due to the-4 large change in THERMAL POWER level caused by the unit shutdown.
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y W f 'W- ' r*W--N g 1 -- W ec ,tme=wm+4y # 2P- '.**-"T7 T=?
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_l QUESTION: 067 (1.00) i j WHICH ONE (1) of the following is the criteria used to detemine if the _
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alternate mixed bed demineralizer should be placed in service during a high
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reactor. coolant activity event per OPOPO4-RC-0001, "High Reactor Coolant j Activity"? ! Chemical analysis detemines the Decontamination Factor (DF) is j inadequat ., j CVCS Letdown Failed Fuel Monitor RT-8039 in alar Letdown flow has been increased to maximu . .
.I 1 RCS gross activity exceeds 50% of the limit specified b Technical Specification i i
QUESTION: 068 .(1.00) WHICH ONE (1) of the following is the MINIMUM flow rate required to be .: established when Emergency Boration is established from the RWST7 'l t gpm gpm gpm gpm l
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QUESTION: 069 (1.00) Given the following:
- Unit I has experienced an ATWS from 100% powe Emergency Boration has been initiated in accordance with FRS1, ' " Response To Nuclear Power Generation /ATWS".
l l WHICH ONE (1) of the following is the TERMINATION criteria for Emergency ,
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i Boration? RCS boron concentration is greater than that required by the shutdown margin limit curve for cold shutdown.
l All rods indicate on the bottom.
l The reactor is subcritica , Emergency Boration flow has been maintained at MAXIMUM flow for a thirty (30) minute time period for all conditions.
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, QUESTION: 070 (1.00) Given the following:
- A reactor startup is in progres .
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- P-10 status light is NOT li WHICH ONE (1) of the following 120 VAC Vital bus failures will result in a reactor trip?
i DP-001 ! l l l DP-002 l DP-1203
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l DP-1201 l
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Page 43 REACTOR OPERATOR l I
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, QUESTION: 071 (1.00) J Given the following: l
; - A Main Steam Line Rupture occurred 60 minutes ag ' - Initial Tcold was 300 degrees Current Tcold is 180 degrees Current RCS Pressure is 500 psi '
WHICH ONE (1) of the following is the Critical Safety function path the plant is in due to the above conditions? Operational Limits Curve is attache Green Yellow Orange
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QUESTION: 072 (1.00) i WHICH ONE (1) of the following is an indication of less than adequate core cooling POST LOCA? Consider each case separatel RVWL plenum level indicates 25 t , Core exit thermocouples indicate 650 degrees RCS subcooling indicates 20 degree Two (2) RCS hot leg temperatures indicate 345 degrees l l
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l QUESTION: 073 (1.00) ;
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t'HICH ONE (1) of the following is the reason for stopping steam oenerator ' depressurization at approximately 231 psig when responding to an inadequate core cooling event? To prevent the steam generator tubes from being uncovered during depressurizatio '
! To prevent injecting nitrogen from the accumulators into the RC ; To minimize potential radioactive releases to the atmosphere- l
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during depressurizatio : To maintain RCS/ Secondary Delta T at values allowing RCP restar t
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QUESTION: 074 (1.00) .I
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Giuen the following:
- A Control Room Evacuation has been ordered by the Shift Superviso i
- None of the IMMEDIATE Action Steps where able to be performed prior to evacuating the Control Roo WHICH ONE (1) of the following is the FIRST action that is required to be - l performed per OPOP04-ZO-0001, " Control Room Evacuation"?
. Direct the Operator in ESF Train C switchgear room to verify the- l reactor trip and bypass breakers are ope i l
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! Establish communications with various control stations on sound I powered phone circuits.
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l Place the Auxiliary Shutdown Panel (ASP) switches in the ASP positio Announce Reactor Trip and Control Room Evacuation using the plant paging system.
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l ' , , l ! l l QUESTION: 075 (1.00) f Given the following: ;
- Unit 1 is at 90% powe " MAIN COND VACUUM LO" alarm is li "LP TURB EXH HOOD TEMP HI" alam is li Condenser vacuum is 22 inches Hg and DECREASING slowl '
WHICH ONE (1) of the following actions should be taken? Commence an orderly shutdown of the Main Turbin Trip the Reactor, then trip the Main Turbin Trip the Steam Generator Feedpump turbines (SGFP). Increase turbine load to allow increased exhaust hood coolin j ! QUESTION: 076 (1.00) l l Given the following: l
- Unit 1 is operating at 30% power.
i - All control systems are in AUTOMATI The "1A" Reactor Coolant Pump has just trippe WHICH ONE (1) of the following is the overall plant response? The reactor trips on a LOW RCS FLOW conditio Unit power is reduced to approximately 22% power (1/4 of original power level).
! l Unit power remains the same with steam flow increasing on the other three steam generator The reactor trips on HIGH steam generator level when "A" Steam ,
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generator level swell ! I i I
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- 1-l QUESTION
- 077 (1.00) .
.q Given the followin j - Unit 1 is at 100% powe _ - Pressurizer pressure selector switch is in the P457/456 positio ; - Pressurizer pressure control is in automati WHICH ONE (1) of the following describes the pressurizer PORVs response if ,_ l j
Pressurizer Pressure channel 457 fails HIGH? , PCV 655A will not close IF RCS pressure decreases belov 2185
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psi PCV 656A will not close UNTIL RCS pressure decreases below 2185 ,
:
psi PCV 655A will ope PCV 656A will open.
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QUESTION: 078 (1.00)
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WHICH ONE (1) of the following is the reason for the order that the valve positions are checked in Step 3, " Check if RCS is Isolated" of OPOP05-EO-EC00, " Loss of All AC Power"? They are listed according to control board locatio Those most likely to fail in a loss of AC power are listed'firs Those most likely to have an RNO corrective action outside_the control room are listed las They are listed according to capacity and potential for inventory los .
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QUESTION: 079 (1.00) i i WHICH ONE (1) of the following conditions will assure " acceptable" consequences following an Anticipated Transient Without Scram (ATWS) Event? -l Turbine is tripped within thirty (30) seconds and Auxiliary .l Feedwater is established within sixty-(60)' second l
. Turbine is tripped within sixty (60) seconds and Auxiliary , !
Feedwater is established within sixty (60) second l
! Turbine is tripped within sixty (60) seconds and Auxiliary ,
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Feedwater is established within ninety (90) second Turbine is tripped within ninety (90)J seconds and Auxiliary Feedwater is established within ninety (90) second l l t
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QUESTION: 080 (1.00) Given the following:
- The RCS has had a stuck.open Pressurizer safety valv The reactor tripped and safety injection initiate The RCS rapidly depressurized to saturation condition Pressurizer level initially dropped and then began to rise rapidly. WHICH ONE (1) of the following characterizes the relationship between pressurizer level and RCS inventory under these conditions? Level is an accurate indication of inventory, because voiding 1 i
would occur first in the pressurizer due-to the high temperature of the pressurizer wall Level is an accurate indication of inventory, because hydraulic pressure would force any voids to the pressurizer steam space and l
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out the safet Level is NOT an accurate indication of inventory, because RCS voiding may result in a rapidly increasing pressurizer.. leve Level is NOT an accurate _ indication of inventory,-because at higher temperatures the cold calibrated pressurizer level I channels falsely indicate high.
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i ! QUESTION: 081 (1.00)- WHICH ONE (1) of the following is the significance of blowing the loop seal , during a cold leg small break LOCA? System mass loss will continue prior ~to the steam vent path being established from the core to the brea Single phase cooling is enhanced while the. loop seal is still ' intac Reflux cooling becomes a viable cooling means following the loss of the loop seal.
i i The heat sink effect of the water in the. intermediate leg is lost i ! when the loop seal is lost, degrading core cooling.
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j QUESTION: 082 (1.00) l l
- WHICH ONE (1) of the following' is the reason the RCP's are required to be i tripped at 1477 psig during a small break LOCA? l
' To eliminate RCP heat input into the RCS. .
; To avoid RCP impeller damage as dissolved gases come' out of '
. solution at lower pressure l I To avoid the higher clad temperature consequences of RCP trip :
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- later during the acciden !
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f To prevent RCP seal damage during subsequent RCS l depressurizatio ;
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! REACTOR OPERATOR -Page 49 l QUESTION: 083 (1.00) WHICH ONE (1) of'the following is the reason for promptly closing the seal' leakoff isolation valve for a RCP with a high number 1 seal leakoff? Prevention of damage to the bearing and seal package due to high-temperatur Prevention of damage to the themal barrier due to high' flo , j Minimize the amount of RCS water that is routed to containment- l sum Assure a minimum back pressure is maintained on the number 3 I sea !
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QUESTION: 084 (1.00) ! Given the following: }
: - Unit I has been shutdown for 10 days following 100 days at 100%
powe The RCS temperature is 120 degrees The RCS is at midloo i
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A total loss of RHR occur No core cooling is re-established.
l l ' WHICH ONE (1) of the following is the MINIMUM time assumed per PGP03-ZO- , 0035, " Reduced RCS Inventory Operation," for the RCS to reach boiling .i conditions? ! I a. 5 minutes j
l minutes ] 1 minutes j , minutes !
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QUESTION: 085 (1.00)
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WHICHONE(1)ofthefollowingsetsofparametersaffectsvortexinginthe :l RHR suction piping? (reduced inventory operations) i RHR flow rate AND RCS leve ; I RHR flow rate AND RCS pressur ',
' Number of RHR pumps running AND RCS pressur .
I Number of RHR pumps running AND RCS level. .
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-i QUESTION: 086 (1.00)- )
i Given the following-
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Unit 2 is in MODE l
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Reactor power is 1E-08 amp Critical data is being take N-35 intermediate range control power fuses blow due to'an , internal faul ~l WHICH ONE (1) of the following is the appropriate action? j
! Enter OPOP05-E0-E000, " Reactor Trip or Safety. Injection," Step Hold power at IE-08 amps until repairs are mad l Insert rods to lower neutron flux level. until both source range ;
NIs energiz ; i Place the N-35 Level Trip switch in the bypass positio , l l I l l l l l l , - . ._ . .
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QUESTION: 087 (1.00) !
^
Given the following: l
- A steam generator tube rupture has occurred in Unit- .
- The RCS is at 520 degrees Containment pressure and temperature' are nonna !
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- The ruptured steam generator has level indication in the-control !
room of: - 99% narrow range - 85% wide range , WHICH ONE (1) of the following is the reason for the wide variance in level-indication?
, Due to rapid pressure fluctuations experienced in the ruptured ' steam generator, the water has been drawn out:of the narrow range j reference le J j Temperature stratification in the steam generator is exposing the transmitters to different density water,-. causing inaccurate level' indicatio l t i The wide range reference leg condensing pot has been overfilled due to the high level condition resulting in an erroneously low l indicatio , The wide range is cold calibrated and does not compensate for the ;
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density differences between the SG water and the reference le !
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QUESTION: 088 (1.00) ; ,
!
WHICH ONE (1) of the following determines the temperature at which the RCS' cooldown is terminated following a Steam generator tube rupture when using ~ 1 OPOP05-EO-E030, " Steam Generator Tube Rupture"?
- RCS pressur Maximum RCS temperature for RHR initiatio Ruptured S/G pressur RCS subcooling.
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REACTOR OPERATOR Paga 52 i QUESTION: 089 (1.00) WHICH ONE (1) of the following is the MINIMUM necessary to verify that the , turbine has tripped following a reactor trip per OPOP05-E0-E000, " Reactor l l Trip or Safety Injection"? Turbine Throttie valves Turbine Governor Valves j l Open Open . I Closed Open l Open Closed Closed Closed QUESTION: 090 (1.00) i Given the following:
- Unit 1 is operating at 100% powe ! - All controls are in the nomal power operation lineu Pressurizer level is DECREASIN VCT level is INCREASIN . -
SEAL WTR LO FLOW alam is li l i
-
REGEN HX LETDN HI TEMP alarm is li LETDN HX OUTLET HI TEMP alam is li )
' -
CHARGING FLOW Hi/LO alam is li WHICH ONE (1) of the following explains the given conditions? Pressurizer PORV failed open, Loss of chargin Small break LOC Letdown isolatio ,
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i REACTOR OPERATOR Page 53
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l QUESTION: 091 (1.00) WHICH ONE (1) of the following is correct regarding operation of a 4.16KV !;
.staitchgear breaker that has been OPENED LOCALLY following a loss of DC control power? j i Breaker can be closed remotely one time if the breaker is in the TEST positio Breaker can be closed one time locally using the switch on the ;
cubicle doo ! I l Breaker cannot be closed again until the closing springs are manually charged, i Breaker cannot be closed until the breaker is reset by removing f and then replacing the fuse bloc ; I !
!
QUESTION: 092 (1.00) !
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WHICH ONE (1) of the following differentiates between an unisolable i feedline break and an unisolable steam line break of the same size? ; l RCS heat removal would be greater from the steam line brea Containment' pressure would be greater for a feedline brea !
! Containment radiation levels would be higher from a steam line ;
brea RCS depressurization would be greater from a feedline brea !
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-. . = - - - -. - - - .
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REACTOR OPERATOR .Page 54 j
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QUESTION: 093 (1.00) t Given the following: .; i
- Unit I has tripped from 100% powe .
i
- The operator initiated Safety Injection due to rapidly decreasing .l pressurizer pressur l
- The following plant conditions were noted after the trip: i
- Pressurizer level 60% and increasin l
- RCS pressure 1700 psig and decreasin {
- Containment Pressure 2 psig and increasin !
- Containment radiation 0.5R/hr and increasin !
- Containment sump levels increasin ;
i WHICH ONE (1) of the following events has occurred? l l Main Steamline break inside containmen i
?
? RCS cold leg brea i i Steam generator tube ruptur i Pressurizer vapor space LOC :
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, QUESTION: 094 (1.00) I l Given the following: l
- Unit I was operating at 100% power when offsite power was los i
- Control room personnel have just transitioned to OPOP05-E0-E502,
" Natural Circulation Cooldown". j
:
WHICH ONE (1) of the following actions should be taken to establish natural circulation cooldown if a reactor coolant pump cannot be restarted? l , Dump steam from all intact steam generators using SG'PORV l l Place steam dump controls in MANUAL and dump steam to the . ! ' condense ) Start one reactor cavity and supports vent supply and exhaust fa Depressurize the RCS using auxiliary spray.
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-. . - . . - -. - - -. ..
REACTOR OPERATOR Page 55 I
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QUESTION: 095 (1.00) Given the following: .l
- Unit 1 is at 100% powe l - All control systems are in their nomal power operation l alignmen WHICH ONE (1) of the following results from Pressurizer Level channel. 466-I failing LOW?
Charging flow control valve (FCV-205) OPEN All pressurizer. heaters deenergiz ;
! Letdown isolation valve (LCV-465) CLOSE . Actual pressurizer level INCREASES.
! l-QUESTION: 096 (1.00) [ i Given the following
- A spent fuel assembly is being raised from its slot in the [
storage pool for return to the reacto t
- Gas bubbles are coming to the surface of the poo ! - Radiation levels in the Spent Fuel Pool area are increasin !
WHICH ONE (1) of the following actions is required of personnel in the Fuel l Handling Building (FHB)? l l Notify the control room to sound the Containment. Evacuation Ala ;
!
! Immediately evacuate all personnel from the Fuel Handling I Buildin l l
- Notify the Core Loading Supervisor, then replace the damaged fuel l assembly in its storage locatio l l Move the. fuel assembly into the RCB and notify the control room l to initiate containment isolatio ;
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REACTOR OPERATOR Page 56 l QUESTION: 097 (1.00) WHICH ONE (1) of the following is the MAY.iMUM length of time that a Specific Radiation Work Permit is valid, without any special approvals? day I week
' I month
, quarter
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t QUESTION: 098 (1.00) WHICH ONE (1) of the following is the MAXIMUM number of visitors one individual is allowed to escort inside Vital Areas per OPGP03-ZS-0001,
" Personnel Access Control"? Assume no special authorizations are requeste i l
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l REACTOR OPERATOR Page 57
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QUESTION: 099 (1.00) Given the following:
-
While touring through the MAB, you pass a valve room that is not posted with any radiation area sign ,
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- A Radiation Protection Technician infonns you that a valve five l (5) feet inside the valve room is producing a 1040 mrem field at l l
18 inches from the valv ,
. WHICH ONE (1) of the following is the proper posting / method of control for l this room? I Radiation Area High Radiation Area Iocked High Radiation Area High Contamination Area QUESTION: 100 (1.00) WHICH ONE (1) of the following controls / instruments is found only on Unit 2? Low noise Source Range Nuclear Instrument preamplifie The Central Alarm Statio Wind direction and speed indicatio The Seismic Monitoring Pane (********** END Of EXAMINATION **********)
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Attachment 3 NRC Official Use Only
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Sao ANS !
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Nuclear Regulatory Commission Operator Licensing Examination ;
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I SCO 09CN N 6560 > CO70 2 13 ' i i , This document is removed from I Official Use Only category on date of examinatio NRC Official Use Only 4
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SENIOR REACTOR OPERATOR Page 1 ANSWER KEY
@ ( Ec74f)
3(Eo) 57 5a0)
[%'
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MULTIPLE CHOICE JJ23 b- C, -W d 2 f oM t'a 001 c 024 d 025
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002 c b
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003 b 026 c 004 d [ 027 b 005 d / 028 d 006 dora fll 029 c 007 a 030 a 008 D { 031 a 009 c 032 c l 010 b,C,d - MA . 033 b 011 c 034 c 012 a 035 d 013 b 0 ly]' { V n ?Pl e o l)ff f 014 a 037 d 05 a
\ \ \l 038 b 016 b 09 b /
017 c 040 d / 018 d 041 c / OLb 042 a / l 020 b 043 d /// l 021 d 044 b i W I 022 c 045 c 1 - l l l
SENIOR REACTOR OPERATOR Page 2 ANSWER KEY l I l 046 b 069 b
//[[
047 c 070 d // ' 048 _ d fg /l/ M 53 no 071 b 049 d [g 9 072 c 050 a 073 c I ! 051 b 074 d l 052 a l[/ 075 a /
053 b 076 a 054 d l 077 a 055 b ff 078 c 056 a 079 a 1 080 c 057 c
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OLb
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058 d l / 059 d 082 a 060 b 083 a / _ 061 a lll 084 d 062 d 085 c 063 a 086 d 064 d 087 b l 065 d 088 c 066 c 089 a
} /
067 b 090 c l / / 068 a 091 d l
i J ., SENIOR REACTOR OPERATOR Page 3 i
ANSWER KEY
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092 b i 093 a f
- l 094 d !
l 095 b j ' 096 d [ i i 097 d
l 098 b f 099 b l , 100 a f i l
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TEST CROSS REFERENCE Page 1 l SR0 Exam PWR Reactor
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0rganized by Question Number QUESTION VALUE REFERENCE 001 1.00 32876 002 1.00 8000001 003 1.00 8000002 004 1.00 8000005 . 005 1.00 8000007 006 1.00 8000009 007 1.00 8000010 008 1.00 8000012 009 1.00 8000013 010 1.00 8000014 011 1.00 8000015 012 1.00 8000016 013 1.00 8000018 014 1.00 8000019 015 1.00 8000022
,, 016 1.00 8000023 017 1.00 8000024 018 1.00 8000026 019 1.00 8000027 020 1.00 8000028 021 1.00 8000030 022 1.00 8000032 023 1.00 8000033 024 1.00 8000034 025 1.00 8000035 026 1.00 8000036 ,
027 1.00 8000037 ! 028 1.00 8000039 029 1.00 8000042 030 1.00 8000043 031 1.00 8000044 032 1.00 8000045 033 1.00 8000046 034 1.00 8000047 , 035 1.00 8000048 036 1.00 8000049 037 1.00 8000051 038 1.00 8000052 039 1.00 8000054 040 1.00 8000055 041 1.00 8000057 042 1.00 8000059 043 1.00 8000060 044 1.00 8000061 045 1.00 8000062 046 1.00 8000063 047 1.00 8000064 , 048 1.00 8000065 l 049 1.00 8000066 ) l
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TEST CROSS REFERENCE Page 2 SR0 Exam PWR Reactor 0rg_anized by Question Number QUESTION VALUE REFERENCE 050 1.00 8000068 051 1.00 8000069 052 1.00 8000070 053 1.00 8000074 , 054 1.00 8000075 055 1.00 8000076 056 1.00 8000077 057 1.00 8000078 058 1.00 8000079 059 1.00 8000080
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060 1.00 8000081 061 1.00 8000082 062 1.00 8000083 063 1.00 8000084 064 1.00 8000085 065 1.00 8000086 066 1.00 8000087 067 1.00 8000088 068 1.00 8000089 069 1.00 8000090 070 1.00 8000091 071 1.00 8000092 072 1.00 8000093 ! 073 1.00 8000094 074 1.00 8000095 > 075 1.00 8000096 076 1.00 8000097 077 1.00 8000098 078 1.00 8000099 079 1.00 8000100 080 1.00 8000101 081 1.00 8000103 082 1.00 8000104 4 083 1.00 8000105 084 1.00 8000106 085 1.00 8000107 086 1.00 8000108 087 1.00 8000109 088 1.00 8000110 089 1.00 8000111 090 1.00 8000113 091 1.00 8000114 092 1.00 8000116 093 1.00 8000118 094 1.00 8000119 095 1.00 8000121 096 1.00 8000123 097 1.00 8000124 098 1.00 8000127
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TEST CROSS REFERENCE Page 3 SR0 Exam PWR Reactor - 0rganized by Question Number QUESTION VALUE REFERENCE 099 1.00 8000129 100 1.00 8000130
...... .
100.00
......
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TEST CROSS REFERENCE Page 4 SR0 Exam PWR Rea'ctor Organized by KA Group - i l PLANT WIDE GENERICS ll
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QUESTION VALUE KA 003 1.00 194001A102- l 096 1.00 194001A103 '
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007 1.00 194001A103'
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005 1.00 194001A106 " 097 1.00 194001A109 ; 012 1.00 194001A114 , 013' 1.00 194001A116 ; ' j 014 1.00 194001A116' 098 1.00 194001K102 .I 1 004 1.00 194001K102 " 095 1.00 194001K103 002 1.00 194001K103 006 1.00 194001K105 010 1.00 194001K114 j 009- -1.00 194001K114 l 011 1.00 194001K116 ! 008 1.00 194001K116 j ______ ; PWG Total 17.00 l l
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PLANT SYSTEMS Group I l QUESTION VALUE KA 015 1.00 001000K103 l 100 1.00 001000K602 l 016 1.00 003000G007 017 1.00 003000K110 1 018 1.00 004000K118-019 1.00 004000K304 021 1.00 013000A102 ! 020 1.00 013000A302 l 025 1.00 014000A102 ) 024 1.00 014000G005 .l 023 1.00 015000A105 022 1.00 015000K504 < 026 1.00 022000K201 ! 027 1.00- 059000A412 1 028 1.00 061000K406 l l 029 1.00 068000G005 030 1.00 071000A405 -l 031 1.00 072000A101 ! 032 1.00 072000K501 1 ______ _ , _ - . . - . . . _ . _ . _ . _ . - _ - - ~ _ .i
TEST CROSS REFERENCE Page 5 SR0 Exam PWR Reactor 0rganized by KA Group PLANT SYSTEMS Group I QUESTION VALUE KA
.
PS_I Total 19.00 Group II QUESTION VALUE KA 033 1.00 002020K509 034 1.00 006000K603 035 1.00 010000K403 036 1.00 011000K401 037 1.00 012000A206 038 1.00 012000K611 049 1.00 029000G010 040 1.00 033000K405 039 1.00 035010K401 ' 099 1.00 062000A401 042 1.00 062000K201
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041 1.00 064000K101 046 1.00 073000K401 !
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048 1.00 079000K101-044 1.00 086000A301 043 1.00 086000K604 ! 045 1.00 103000G005 l PS-II Total 17.00 Group III QUESTION VALUE KA 050 1.00 005000K109 051 1.00 008000A301 052 1.00 076000K402 047 1.00 078000K302 ______ PS-III Total 4.00 ______ ______ PS Total 40.00 EMERGENCY PLANT EVOLUTIONS i Group I : I j
_ _ TEST CROSS REFERENCE Page 6 , t SR0 Exam PWR Reactor ,
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" 0rganized by KA Group EMERGENCY PLANT EVOLUTIONS Group I i QUESTION VALUE KA ,
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053 1.00 000001G010 055 1.00 000003G003 054 1.00 000005A101 058 1.00 000011A213 , 059 1.00 000011G011 057 1.00 000015A122 056 1.00 000015A208 072 1.00 000015K104 062 1.00 000024A125 063 1.00 000024A205 1 076 1.00 000029G010 077 1.00 000029G012 065 1.00 000040G011 071 1.00 000051A202 075 1.00 000055A203 074 1.00 000055K302 064 1.00 000057A219 070 1.00 000068G010 069 1.00 000068K201 066 1.00 000074A201 068 1.00 000074G012 067 1.00 000074K311 061 1.00 000076A202 060 1.00 000076G004
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______ EPE-I Total 24.00 Group II QUESTION VALUE KA 086 1.00 000007G010 091 1.00 000008A220 078 1.00 000008K301 079 1.00 000009K101 080 1.00 000009K323 087 1.00 000022A201 082 1.00 000025A102 081 1.00 000025K101 073 1.00 000027A101 083 1.00 000033G011 090 1.00 000037G011 084 1.00 000038A111 085 1.00 000038A136 089 1.00 000054K101
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l l TEST CROSS REFERENCE Page 7 ' l l SR0 Exam PWR Reactor
i 0rganized by KA Group 1 ; i l ! ! EMERGENCY PLANT EVOLUTIONS l l , Group II QUESTION VALUE KA
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088 1.00 000058G006 l 001 1.00 000060K202 j
....__ 1 EPE-II Total 16.00 l !
Group III l QUESTION VALUE KA ; l l 093 1.00 000028K305 ! 094 1.00 000036G008 l l 092 1.00 000056G010
..____
EPE-III Total 3.00 ______
._____
l EPE Total 43.00 l
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l l Test Total 100.00 l I i
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ANSBER: 001 (1.00) [+1.0] REFERENCE: STP: OPOP04-RA-0001, p. 6-10 & 2 STP: LOT 202.38, LO# , STP: South Texas 1992 exa , KA: 000060K202 (2.7/3.1) 000060K202 ..(KA's)
;
ANSWER: 002 (1.00) l [+1.0] ,
!
REFERENCE:
!
STP: LOT 503.02, pp. 18-19, LO 08 & 0 STP: Technical Specification 6.12, p. 6-25 KA: 194001K103 (2.8/3.4) 1992 Exam Modified Question SRO Only l l 194001K103 ..(KA's) ANSWER: 003 (1.00) l l l [+1.0] 4 REFERENCE: l STP: OPGP03-ZA-0010, Plant Procedure Adherence and Implementation and Independent Verification, p.1 STP: LP LOT 504.01, LO KA: 194001A102 (4.1/3.9) 194001A102 ..(KA's)
i l ! SENIOR REACTOR OPERATOR Page 58 , l ANSUER: 004 (1.00) , l [+1.0]
:
REFERENCE:
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STP: OPGP03-ZO-0039, Operations Configuration Management, p. 4 STP: LOT 504.01, LO 05, and LOT 802.31, LO 2 . i l KA: 194001K102 (3.7/4.1) SR0 Only l >
194001K102 ..(KA's) ANSSER: 005 (1.00)
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REFERENCE: l STP: OPGP03-70-0003, Temporary Modifications, p. STP: LOT 802.10, LO 0 ' KA: 19400A106 (3.4/3.4) 194001A106 ..(KA's) ,
l l ANSWER: 006 (1.00) . l
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l d ,a. . [+1.0]
l l REFERENCE: STP: OPGP03-ZO-0032, Reactor Containment Building Entry, p. STP: LOT 504.01, LO I KA: 194001K105 (3.1/3.4)
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194001K105 ..(KA's)
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Page 59 SENIOR REACTOR OPERATOR l t , ANSBER: 007 (1.00) [+1.0] l
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REFERENCE:
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STP: Operations Policy and Practices Manual, p. STP: LOT 504.01, LO , STP: Modified Exam Bank Question L50401-92186-2 KA: 194001A103 (2.5/3.4) .
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194001A103 ..(KA's) . L ANSWER: 008 (1.00) [+1.0] REFERENCE: I STP: OPGP03-ZF-0011, STPEGS Fire Brigade, p. STP: LOT 504.01, LO KA: 194001K116 (3.5/4.2) 194001K116 ..(KA's) , ANSWER: 009 (1.00) i [+1.0] l
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REFERENCE: STP: OPGP03-ZI-0007, Confined Space Entry Program, p. STP: LOT 504.01, LO KA: 194001K114 (3.3/3.6) 194001K114 ..(KA's)
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! ' ANS' R: 010 (1.00)
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REFERENCE:
- STP
- OPGP03-ZI-00 Confined Space Entry Program, p.1 .
l STP: LOT 802, LO , KA: 194001K114 ( ) ! 194001K114 ..(KA's) F l ANSWER: 011 (1.00) ! [+1.0]
! REFERENCE:
- STP
- OPGP03-ZF-0006, " Control of Ignition Sources".
I STP: LOT 802.05, LO 3.
l KA: 194001K116 (3.5/4.2) l 194001K116 ..(KA's) ANSWER: 012 (1.00) [+1.0] , REFERENCE:
,
STP: OPGP03-Z0-0012, Plant Chemistry Specifications, p. 24 i STP: LOT 504.01, LO KA: 194001A114 (2.5/2.9) 194001A114 ..(KA's) l l l l l l
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ANSWER: 013 (1.00)
! [+1.0]
l REFERENCE:
STP: OREP01-ZV-1N02, Notifications to Offsite Agencies, p. KA: 194001A116 (3.1/4.4) . 194001A116 ..(KA's)
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l ANSWER: 014 (1.00) l [+1.0] REFERENCE: i STP: OREP01-ZV-IN02, Notifications to Offsite Agencies, p. 3.
l KA: 194001A116 (3.1/4.4) i 194001A116 ..(KA's) i l ANSWER: 015 (1.00) [+1.0] ) i REFERENCE:
STP: LOT 201.19, p. 27, LO# STP: LOT 201.18, p. 29-30, Lo# 1 & 2.
l VA: 001000K103 (3.4/3.6) 001000K103 ..(KA's) , ANSWER: 016 (1.00) [+1.0] l
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REFERENCE: STP: LOT 201.20, p. 44, LO # STP: Modified STP 1992 Exam Qucstion KA: 003000G007 (3.2/3.3) 003000G007 ..(KA's) ANSWER: 017 (1.00) . [+1.0] REFERENCE: STP: LOT 201.05, p. 8, LO #2 & KA: 003000K110 (3.0/3.2) 003000K110 ..(KA's) ANSSER: 018 (1.00) [+1.0] REFERENCE: ,
,
STP: LOT 201.06, p. 32, LO# STP: LOT 201.12, p. 3 KA: 004000K118 (2.9/3.2) 004000K118 ..(KA's) ANSWER: 019 (1.00) [+1.0] REFERENCE: STP: LOT 201.06, p.21, L0#19 KA: 004000K304 (3.7/3.9) 004000K304 ..(KA's) i ! SENIOR REACTOR OPERATOR Page 63 l l l I
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ANSWER: 020 (1.00) [+1.0] < REFERENCE: STP: LOT 201.22.H0.01, p. 11-12, LO#1 ; KA: 013000A302 (4.1/4.2)
.
013000A302 ..(KA's) ! ! ANSWER: 021 (1.00) ! [+1.0] I i REFERENCE: ! STP: LOT 201.22, p. 16. L0 # 1 KA: 013000A102 (3.9/4.2) . 013000A102 ..(KA's) ANSWER: 022 (1.00) [+1.0] REFERENCE: STP: LOT 201.16, p. 50-57. LO # 1 KA: 015000K504 (2.6/3.1) 015000K504 ..(KA's) ANSUER: 023 (1.00)
-b- [+1.0] 1 ( , ;
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REFERENCE: i STP: LOT 201.16, p. 56, LO #1 STP: T/S 3.2.1, p. 3/4 2-1 & 4.2.1.1. & 4.2.1.2, p. 3/4 2- i KA 015000A105 (3.7/3.9) NOTE: Answer - Out at 06:34 (T/S interpretation
- logged value assumed to exist during the interval preceding each logging), In at 06:55 015000A105 ..(KA's) , ,
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ANSWER: 024 (1.00) [+1.0] REFERENCE: STP: LOT 201.19, p. 31, LO #7 STP: Technical Specification 3.1.3.2, " Position Indication System", p. 3/4 - 1-1 i KA: 014000G005 (3.1/3.7) 014000G005 ..(KA's)
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i ANSWER: 025 (1.00) i [+1.0]
!
t REFERENCE: ,
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STP: LOT 201.19, p. 5, LO # KA: 014000A102 (3.2/3.6) 014000A102 ..(KA's) ! i ANSSER: 026 (1.00) l 1 [+1.0] ,
SENIOR REACTOR OPERATOR Page 65 REFERENCE: STP: LOT 201.36.F1G.05, LO # KA: 022000K201 (3.0/3.1) 022000K201 ..(KA's) ANSMER: 027 (1.00)
. [+1.0]
REFERENCE: , ! STP: LOT 201.13, p. 27, LO #2 KA: 059000A412 (3.4/3.5) 059000A412 ..(KA's)
ANSUER: 028 (1.00) [+1.0] ! REFERENCE: STP: LOT 202.28, p. 8, LO# 2 & STP: Modified exam bank question L20228-80173-0 KA: 061000K406 (4.0/4.2) 061000K406 ..(KA's) I ANSWER: 029 (1.00) [+1.0] REFERENCE: STP: LOT 201.37, p. 15, LO #3.
i KA: 063000K201 (2.9/3.1) 068000G005 ..(KA's)
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i l SENIOR REACTOR OPERATOR Page 66 , ANSWER: 030 (1.00) [+1.0] l REFERENCE: STP: LOT 203.14, p. 7 & 11, LO # KA: 071000A405 (2.6/2.6) . 071000A405 ..(KA's) , ANSWER: 031 (1.00) [+1.0] , t REFERENCE: t STP: LOT 202.4;:, p. 9-12, LO # 7 KA:072000A101 (3.4/3.6) 072000A101 ..(KA's) ANSWER: 032 (1.00) , [+1.0] ! REFERENCE: l STP: LOT 202.42, p. 4, LO # STP: Modified Exam Bank Question L20242-11706-01 KA: 072000K501 (2.7/3.0)
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ANSWER: 033 (1.00) [+1.0]
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REFERENCE: STP: LOT 201.15, p. 27, LO # 1 KA: 002020K509 (3.6/3.9) ! 002020K509 ..(KA's)
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ANSWER: 034 (1.00)
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REFERENCE: STP: LOT 201.10, p 15, LO # ;' KA: 006000K603 (3.6/3.9) 006000K603 ..(KA's) l ANSWER: 035 (1.00) i [+1.0) l REFERENCE: l STP: LOT 201.15, p. 7-8, L0 # i l KA:01000K403 (3.8/4.1)
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010000K403 ..(KA's) l ANSt5ER: 036 (1.00) [+1.0] REFERENCE: l STP: LOT 201.14, p. 11-12, LO# 1 STP: Logic Z-4215 KA: 011000K401 (3.3/3.7) : 011000K401 ..(KA's) i ! l l
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REFERENCE: STP: LOT 201.20, p. 17, L0 # 1 ;
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KA: 012000A206 (4.4/4.7) , ! : l 012000A206 ..(KA's) , ! !
ANSWER: 038 (1.00) ; [+1.0] ! i REFERENCE: STP: KA: LOT 201.15, 2p(. 012000K611 14-15, 9/2.9) LO# K611 ..(KA's) ANSWER: 039 (1.r,0) [+1.0] REFERENCE: ' STP: LOT 202.15, p. 17, L0# KA: 035010K401 (3.6/3.8) ' 035010K401 ..(KA's) ANSWER: 040 (1.00) [+1.0] ! i
__ ' i SENIOR REACTOR OPERATOR Page 69 j l REFERENCE: KA: 033000K405 (3.1/3.3) ! l 033000K405 ..(KA's) i
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ANSUER: 041 (1.00) 1 [+1.0] . REFERENCE:
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STP: LOT 201.39, p. 13-14 & 77, L0f 1 STP: OPOP02-DG-0001, Emergency Diesel Generator, p. KA: 064000K101 (4.1/4.4) 064000K101 ..(KA's) ANSWER: 042 (1.00) [+1.0] I REFERENCE: , STP: LOT 201.36, p. 6-9, L0f STP: LOT Exam Bank, L201136-91864-0 KA: 062000K201 (3.3/3.4) , 062000K201 ..(KA's) ANSdER: 043 (1.00) [+1.0] i '
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SENIOR REACTOR OPERATOR Page 70 REFERENCE: STP: LOT 201.29, p. 13-15, LO# 1 KA: 086000K604 (2.6/2.9) 086000K604 ..(KA's) ANSWER: 044 (1.00)
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b, c . [+1.0] REFERENCE: STP: LOT 201.29, p. 21-24, L0 # KA: 086000A301 (2.9/3.3) , 086000A301 ..(KA's)
ANSWER: 045 (1.00)
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REFERENCE: STP: LOT 201.01, p. 32. L0f 10 & 1 KA: 103000G005 (3.3/4.1) 103000G005 ..(KA's) ANSWER: 046 (1.00) [+1.0] REFERENCE: STP: LOT 202.41.HO.02, p. 11, LO# KA: 073000K401 (4.0/4.3) 073000K401 ..(KA's)
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ANSWER: 047 (1.00) [+1.0] r REFERENCE: STP: LOT 201.09, p. 11, LO # STP: LOT 504.02, LO # . - STP: LOT 202.26.TP.17 KA: 078000K302 (3.4/3.6) 078000K302 ..(KA's) ANSWER: 048 (1.00) [+1.0] ; REFERENCE: STP: LOT 202.26, p. 12-15, LO #1 STP: Modified Exam Bank Question L202226-25608-02 KA: 079000K101 (3.0/3.1) 079000K101 ..(KA's) l
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ANSWER: 049 (1.00) ; J [+1.0] REFERENCE: STP: LOT 202.33, p. 19, LO# 2 KA: 029000G010 (2.9/3.1) 029000G010 ..(KA's) ANSSER: 050 (1.00) [+1.0] I!
SENIOR REACTOR OPERATOR Page 72 REFERENCE: , l STP: LOT 201.09, p. 6. LO# KA: 005000K109 (3.6/3.9) 005000K109 ..(KA's) l ANSWER: 051 (1.00) [+1.0] , l : REFERENCE:
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STP: LOT 201.12, p.12-6, LO# STP: OPOP04-CC-0001, Addendum 1, p. 1 ' KA: 008000A301 (3.2/3.0) 008000A301 ..(KA's) l ANSWER: 052 (1.00) [+1.0] REFERENCE:
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STP: LOT 201.13.H0.01, p. 15, LO# 3 KA: 076000K402 (2.9/3.2) 076000K402 ..(KA's) ANSWER: 053 (1.00) [+1.0] i REFERENCE: STP: OPOP04-RS-0001, p. STP: LOT 504.02, p. 4, LO# 4 ,
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SENIOR REACTOR OPERATOR Page 73 ANSWER: 054 (1.00) [+1.0] i REFERENCE: STP: OPOP04-RS-0001, p. STP: LOT 504.02, p. 4, LO# KA: 000005A101 (3.6/3.4) l 000005A101 ..(KA's) 055 (1.00)
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STP: OPOP04-RS-0001, p. STP: LOT 504.02, p. 4, LO# 7 & KA: 000003G003 (3.3/3.8) , 000003G003 ..(KA's) i ANSWER: 056 (1.00) [+1.0] REFERENCE: STP: LOT 504.02, p. 4. L0# STP: OPOP04-RC-0002, p. KA: 000015A208 (3.4/3.5) 000015A208 ..(KA's) ANSWER: 057 (1.00) [+1.0]
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ANSWER: 058 (1.00) [+1.0] REFERENCE: STP: LOT 504.09, LO# 6 & " STP: LOT 501.13, p. 6-10.- KA 000011A213 (3.7/3.7) Note: The lower steam pressure of a secondary fault will manifest itself as an increased delta T in the RC I 000011A213 ..(KA's)
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ANSWER: 059 (1.00) l [+1.0]
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REFERENCE: STP: LOT 504.04, p. 19, LO# 6 & 7, KA: 000011G011 (4.3/4.5) 000011G011 ..(KA's) l 1 ANSWER: 060 (1.00) i !
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STP: LOT 503.01, p. 14. LO# STP: Technical Specification 3.4.8, p. B 3/4 4- STP: Modified 1992 STP exam question (different distractors) KA: 000076K306 (3.2/3.8)
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ANSWER: 061 (1.00) [+1.0] REFERENCE: STP: LOT 504.02, LO# STP: OPOP04-RC-0001, p. KA: 000076A202 (2.8/3.4) 000076A202 ..(KA's) ANSWER: 062 (1.00) [+1.0] ! l ! REFERENCE: I STP: LOT 504.02, LO# 1 STP: OPOP04-CV-0003, p. KA: 000024A125 (3.4/3.3) l 000024A125 ..(KA's) l ANSWER: 063 (1.00) ( [+1.0] 1 1
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STP: LOT 504.20, L0f STP: OPOP04-EE-0002, p. KA: 000057A219 (4.0/4.3) 000057A219 ..(KA's) , ANSWER: 065 (1.00) [+1.0] REFERENCE: STP: LOT 502.07.T STP: LOT 504.38, p.3, LO# 1 KA: 000040G011 (4.1/4.3) 000040G011 ..(KA's) i ANSWER: 066 (1.00) [+1.0] l
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SENIOR REACTOR OPERATOR Page 77 REFERENCE: STP: LOT 504.30, p. 4, L0# STP: LOT 502.05, LO# KA: 000074A201 (4.6/4.9) 000074A201 ..(KA's) ANSSER: 067 (1.00) , [+1.0] REFERENCE: STP: LOT 504.30, p. 9. LO# KA: 000074K311 (4.0/4.4)
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000074K311 ..(KA's) ANSWER: 068 (1.00) [+1.0] l REFERENCE: ; STP: LOT 502.09, p. 67,H0.01 STP: LOT 502.02, L0# I
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KA: 000074A116 (4.4/4.6) 000074G012 ..(KA's) ANSWER: 069 (1.00) [+1.0] REFERENCE: l i STP: LOT 803.04, p. 4, LO# KA: 000068K201 (3.9/4.0) 000068K201 ..(KA's)
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STP: LOT 504.02, LO# ' l STP: OPOP04-CR-0001, p. 2- KA: 000051A202 (3.9/4.1) i i i 000051A202 ..(KA's) ANSWER: 072 (1.00) , [+1.0] l l REFERENCE: 1 l l STP: LOT 501.05, p.12-14, LO# ] KA: 000015K104 (2.9/3.1) 000015K104 ..(KA's)
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l SENIOR REACTOR OPERATOR Page 79 REFERENCE: STP: LOT 501.06, p. 25-29, L0# STP: LOT 201.14, p. 10, LO# KA: 000027A101 (4.0/3.9) 000027A101 ..(KA's) ANSWER: 074 {1.00) -
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I [+1.0] ! REFERENCE: STP: LOT 504.22, p.9, LO# KA: 000055K302 (4.3/4.6)
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000055K302 ..(KA's) , i l l ANSWER: 075 (1.00) [+1.0] , REFERENCE: STP: LOT 504.22, p.11, LO# KA: 000055A203 (3.9/4.7) 000055A203 ..(KA's) ANSUER: 076 (1.00) [+1.0] REFERENCE: STP: LOT 502.04, p.14, LO# KA: 000029G007 (3.8/4.0) 000029G010 (4.5/4.5) 000029G010 0000293007 ..(KA's) I
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SENIOR REACTOR OPERATOR Page 80 l l ANSWER: 077 (1.00) [+1.0] i
l REFERENCE: { STP: LOT 504.28, p.7, LO# KA: 000029G012 (4.1/4.2) 000029G012 ..(KA's) ANSWER: 078 (1.00) [+1.0] , REFERENCE: STP: LOT 501.13, p. 23-24, LO# KA: 000008K301 (3.7/4.4) 000008K301 ..(KA's) ANSWER: 079 (1.00) [+1.0] REFERENCE: STP: LOT 501.13, p. 20-22, LO# 1 & ; KA: 000009K101 (4.2/4.7) Note: Blowing the loop seal is a very important part of cooling the core during a SBLOCA; the KA for this question relates l to the core cooling. Blawing the loop seal determines what cooling mode is in effect for the core. Understanding this concept is fundamental to understanding how a SBLOCA can cause core uncover K101 ..(KA's) ANSWER: 080 (1.00) ) l [+1.0] i i i
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i STP: LOT 504.09, p. 8, LO# KA: 000009K323 (4.2/4.3) l 000009K323 ..(KA's) ANSWER: 081 (1.00) [+1.0] ; REFERENCE: STP: LOT 504.57, p. 7-8, LO# ,
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l KA: 000025K101 (3.9/4.3) Generic Letter 88-17 ' reveals if the candidate is sufficiently sensitive to the issue of loss of ~ RHR, and the very short time frame available to respond.to same. Industry events have occurred where RHR has been lost at reduced inventory, and one ' key issue is the operators were often not aware how little time was available until saturation was reached in the core. The question does not , , require detailed knowledge of the saturation vs time curve due to the very : large time frame of the incorrect distractors.... actual time from 120F to ,
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200F is 10 minute I 000025K101 ..(KA's) < I ANSUER: 082 (1.00)
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STP: LOT 201.16, p. 24, L0f 1 KA 000033G011 (3.2/3.4) 000033G011 ..(KA's) i ANSWER: 084 (1.00)
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REFERENCE: STP: LOT 502.02, p. 27-28, L0# 1 KA: 000038A111 (3.8/3.9) 000038A111 ..(KA's) ANSWER: 085 (1.00) [+1.0] REFERENCE: STP: LOT 504.15, p.16, LO# 12. OPOP05-E0-E030 KA: 000038A136 (4.3/4.5) , l
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000038A136 ..(KA's) i ANSWER: 086 (1.00)
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ANSWER: 087 (1.00) [+1.0] REFERE. ,'.E: STP: LOT 504.02, LO# STP: Simulator Malfunction Boo STP: Exam Bank question L50402-92106-0 KA: 000022A201 (3.2/3.8) 000022A201 . (KA's) ANSWER: 088 (1.00) [+1.0] '
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STP: LOT 504.02, LO# STP: OPOP04-DJ-0001, p. STP: LOT 201.31, p.9, LO# KA: 000058G006 (3.4/3.8) 000058G006 ..(KA's) ANSWER: 089 (1.00) [+1.0] REFERENCE: STP: LOT 501.11, p. 12-15, LO# STP: LOT 501.12, p.11-14, LO# KA: 000054K101 (4.1/4.3) 000054K101 ..(KA's)
SENIOR REACTOR OPERATOR Page 84 ANSWER: 090 (1.00) [+1.0] REFERENCE: STP: OPOP-RC-0004, p. 3- STP: LOT 504.02, p.3, L0f . KA: 000037G011 (3.9/4.1) 000037G011 ..(KA's) ANSWER: 091 (1.00)
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d. [ol .0] REFERENCE: STP: LOT 501.13.H0.01, p. 4, LO# KA: 000008A220 (3.4/3.6) 000008A220 ..(KA's) ANSWER: 092 (1.00) [+1.0] REFERENCE: STP: LOT 504.05, p. 12, LO# KA: 000056G010 (3.7/3.9) 000056G010 ..(KA's) ANSWER: 093 (1.00) [+1.0]
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REFERENCE: STP: LOT 501.06, p. 32, LO# KA: 000028K305 (3.7/4.1) 000028K305 ..(KA's) . l l ANSWER: 094 (1.00) l
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STP: LOT 801.01, p.14, LO KA: 000036G008 (3.1/3.7) l l i 000036G008 ..(KA's) l ANSWER: 095 (1.00) [+1.0] l
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REFERENCE: STP: OPRP08-ZA-0001, p. 3, STP: LOT 802.03, p.2, L0f KA: 194001K103 (2.8/3.4) 194001K103 ..(KA's) i ANSWER: 096 (1.00) [+1.0] l l l l
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SENIOR REACTOR OPERATOR Page 86 REFERENCE: STP: OPGP03-ZA-0010, p. 9-10.
! STP: LOT 802.07, p. 3, LO# KA: 194001A103 (2.5/3.4) 194001A103 ..(KA's) i l ANSWER: 097 (1.00) - [+1.0] l l REFERENCE: STP: OPGP03-Z0-0022, p. 3.
i STP: LOT 802.08, p.3, LO# i
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REFERENCE: STP: LOT 504.01, p. 2. LO# STP: OPGP03-Z0-0039, p. 39-4 KA: 194001K102 (3.7/4.1) 194001K102 ..(KA's) ANSWER: 099 (1.00) [+1.0] I REFERENCE: ' STP: LOT 203.21, p. 5, LO# KA: 062000A401 (3.3/3.1) 062000A401 ..(KA's)
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, ANSWER: 100 (1.00) , [+1.0] i , j REFERENCE: . STP: LOT 201.18.HO.01, p. 9 & H0.02, p. 2, LO# STP: STP Exam Bank Question L20118-02162-0 , KA: 001000K602 (2.8/3.3) 001000K602 ..(KA's)
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Nuclear Regulatory Commission Operator Licensing i Examination
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This document is removed from i Official Use Only category on date of examinatio NRC Official Use Only i
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I U. 5. NUCLEAR REGULATORY COMMISSION j SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 4 )
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CANDIDATE'S NAME: FACILITY: South Texas 1 & 2
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REACTOR TYPE: PWR-WEC4 r , i DATE ADMINISTERED: 93/09/27
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lINSTRUCTIONSTOCANDIDATE: Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in ! parentheses after the question. The passing grade requires a final grade of ! at least 80%. Examination papers will be picked up four (4) hours after the examination start CANDIDATE'S TEST VALUE SCORE % i i 100.00 % TOTALS FINAL GRADE i All cork done on this examination is my ow I have neither given nor j received ai ; Candidate's Signature I
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SENIOR REACTOR OPERATOR Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan MULTIPLE CH0 ICE 023 a b c d 001 a b c d 024 a b c d
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002 a b c d 025 a b c d , 003 a b c d 026 a b c d ' 004 a b c d 027 a b c d 1 ! i 005 a b c d 028 a b- c d l l 006 a b c d 029 a b c d 007 a b c d 030 a b c d l 008 a b c d 031 a b c d j 009 a b c d 032 a b c d
l 010 a b c d 033 a b c d , 011 a b c d 034 a b c d 012 a b c d 035 a b c d 013 a b c d 036 a b c d 014 a b c d 037 a b c d 015 a b c d 038 a b c d 016 a b c d 039 a b c d l 017 a b c d 040 a b c d 018 a b c d 041 a b c d l 1 019 a b c d 042 a b c d 020 a b c d 043 a b c d j 021 a b c d 044 a b c d 022 a b c d 045 a b c d
-_ _ - _ _ - . SEN10R REACTOR OPERATOR Page 3 ANSWER SHEET 1 Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan l 046 a b c d 069 a b c d !
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047 a b c d 070 a b c d ;
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048 a b c d 071 a b c d 049 a b c d 072 a b c d 050 a b c d 073 a b c d . i ! 051 a b c d 074 a b c d 052 a b c d 075 a b c d ! 053 a b c d 076 a b c d 054 a b c d 077 a b c d 055 a b c d 078 a b c d 056 a b c d 079 a b c d 057 a b c d 080 a b c d 058 a b c d 081 a b c d 059 a b c d 082 a b c d 060 a b c d 083 a b c d 061 a b c d 084 a b c d 062 a b c d 085 a b c d 063 a b c d 086 a b c d 064 a b c d 087 a b c d 065 a b c d 088 a b c d 066 a b c d 089 a b c d 067 a b c d 090 a b c d 068 a b c d 091 a b c d
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I SENIOR REACTOR OPERATOR Page 4 ANSWER SHEET - Multiple Choice (Circle or X your choice) If you change your answer, write your selection in the blan , 092 a b c d 093 a b c d
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094 a b c d
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097 a b c d 098 a b c d ,
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099 a b c d 100 a b c d s
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(********** END OF EXAMINATION **********)
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I i l Page 1 . Policies and Guidelines I for taking NRC Written Examinations During the administration of this examination the following rules apply: 1. Cheating on the examination will result in a denial of your application and could result in more severe penaltie . After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio ; r 3. To pass the examination, you must achieve a grade of 80% or ! greate ! 4. The point value for each question is indicated in parentheses , after the questio i
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5. There is a time limit of four (4) hours for completion of the i examinatio ! 6. Use only black ink or dark pencil to ensure legible copie ' 7. Print your name in the blank provided on the examination cover sheet and the answer shee *
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8. Mark your answers on the answer sheet provided and do not leave any question blan i If the intent of a question is unclear, ask questions of the examiner onl . Restroom trips are permitted. but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or I possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids and answer sheets and give it to the examiner or proctor. Remember to sign the statement on the examination cover shee . After you have turned in your examination, leave the examination area as defined by the examiner.
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SENIOR REACTOR OPERATOR Page 7 ; ,JESTION: 001 (1.00) WHICH ONE (1) of the following describes the AUTOMATIC response of the fuel Handling Building (FHB) HVAC system upon receipt of an FHB Exhaust HIGH , RADIATION alarm on RT-8035? ! All Main Supply fans START and the Relief Supply dampers CLOS All Main Exhaust and Exhaust Booster fans STOP and the Exhaust , System dampers CLOS i All Main Exhaust and Exhaust Booster fans START and the Exhaust ! System shifts to align two filtering unit train ;
, All Main Supply fans START and the Relief Supply dampers shift to align two filtering unit train l
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QUESTION: 002 (1.00)
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Given the following:
- Unit 1 is in Mode 5 for a short outag ;
- During a containment inspection, the Shift Supervisor notices some radiation barricade ropes up in the area of RCP 1 ,
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- A radiation sign on the ropes reads " Caution: High Radiation Area, RWP Required for Entry" and indicates a MAXIMUM radiation level of 1.12 Rem /hr inside the + ve WHICH ONE (1) of the following additional posting requirements and/or controls are required for this area in accordance with Technical Specification 6.12. "High Radiation Area"? The area should be fenced off, the door (s) to the area kept !
locked, and the keys kept under the administrative control of the j Shift Superviso j The area should be fenced off, the door (s) to the area kept l locked with the keys kept under the administrative control of the , Radiation Protection Superviso ] The area shall have a flashing light in the immediate area as a warning devic d The area should have a closed circuit TV monitor installed to provide radiation protection personnel with continuous monitoring capabilit i
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QUESTION: 003 (1.00) WHICH ONE (1) of the following is the preferred method for transcribin l initials from one working cany of a procedure to another when an original , working copy of a procedure oecomes contaminated? l 6 The Unit supervisor should initial the new procedure for the ! individual who entered the data on the original working cop '
' The person who entered the data on the original working copy
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should initial the new procedur t The on duty R0 should initial the new procedure after another , operator locally independently verifies the procedure is ; complet !
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l The Shift supervisor should initial the new procedure f'or the l individual who entered the data on the original working copy.
! i l QUESTION: 004 (1.00) Given the following: ; i
- Maintenance is making preparations to repack a valve-that is ;
using its backseat as the isolation boundar j
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They have determined that it will be necessary to remove the j
- valve handwheel to access the packing glan ;
Removing the handwheel will NOT move the valve off the backsea '
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l u WHICH ONE (1) of the following is required to com,-lete the job in : accordance with STPEGS procedures? ~!
l The valve shall be Caution tagged with a cleared position of .! backseate J The applicable work document shall bc returned to the " Acceptor" to verify adequate clearance requirement ;
! The Equipment Clearance Order (ECO) shall be modified to allow removal of the handwhee A Danger tag shall be placed in a visible location on the valve body.
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QUESTION: 005 (1.00) ; WHICH ONE (1) of the following is considered a Temporary Modification per i OPGP03-ZO-0003, " Temporary Modifications"?- ; i Modifications to equipment that is out of servic i Installation of.tygon tubing on a pump's drain lin f
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" Reinstallation of a normally installed blind flange on a system that is shown on a system P&I Gagging a relief or safety valv i l
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QUESTION: 006 (1.00) l
Given the following: l
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Unit 2 is in MODE I
- An INITIAL entry into the Reactor Containment Building (RCB) is l l
being planne .
- OPGP03-ZO-0032, " Reactor Containment Building Entry" has been l
- initiate Oxygen levels were yerified to be 21% 4 hours ag l-I WHICH ONE (1) of the following pieces of equipment is required for the team i to enter into the RCB7 l I ! An Oxygen monito , A Neutron Survey Instrumen ) ASelfContainedBreathingApparatus(SCBA).
I A Radiation Monitoring instrument with a IR/hr rang . ! l l l
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QUESTION: 007 (1.00) WHICH ONE (1) of the following evolutions requires the Shift Supervisor's presence in the "AT THE CONTROLS" area per the Operations Policies and Practices Manual? Synchronization of the main generato Performing an SSPS surveillance tes . Placing the Main Turbine First Stage Impulse Pressure loop into servic i Placing a Main Feedwater pump in service.
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QUESTION: 008 (1.00)
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Under WHICH ONE (1) of the following conditions may an Operator be designated as the fire brigade leader?
! Under emergency conditions when deemed necessary by the Chemical Operations Forema , The Chemical Operations Foreman does not meet the requirements and the Operator has completed the required trainin Only when the unaffected units fire brigade leader is unavailabl When the Fire Protection Coordinator has approved the Operrtor to t act as the brigade leade ;
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i l l QUESTION: 009 (1.00) l WHICH ONE (1) of the following is a duty of the Standby Attendant required i by the Confined Space Entry program? l
. Ensure that personnel entering the Pemit Entry Space are wearing i the proper protective equipmen ; Test the atmosphere of the Permit- Entry Space if forced ,
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ventilation is interrupte ! Maintain communications with personnel inside the Pemit Entry
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Spac l Recover any individuals who become incapacitated by pulling on : their safety harness or lifelin : F
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i QUES : 010 (1.00) i WHICH ONE of the following is an PERMIT ENTRY SPACE as defined.by the Confined Space try Program procedure? !
: Blowdown Va Cubicles l
! Steam Generator 2A l
- ; Settling Basin i A deJelek i Underground pipe ways
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QUESTION: 011 (1.00) 1 l i l WHICH ONE (1) of the following, per OPGP03-ZF-006 " Control of Ignition i Sources", describes the reason why the responsible supervisor / foreman must j notify the Control Room prior to commencing Tungsten Inert Gas (TIG) l t'elding within the power block?. l l TIG welding creates a large amount of heat that may-potentially damage vital equipmen , i TIG welding in the power block requires that local smoke and heat l detectors be covered to prevent activating fire systems.
l TIG welding may have an effect on radiation monitors and nuclear instrumentatio TIG welding gives off fumes that will activate.the Toxic Gas , Analyzers and cause HVAC systems to auto-start.- l
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l QUESTION: 012 (1.00) l
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Given the following: ;
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- Unit 2 is shutdown, maintaining vacuum in the main condense l - Main condenser hotwell samples indicate'a HIGH Dissolved Oxygen !
content, j , WHICH ONE (1) of the following chemicals should be added to the. system to l reduce the dissolved oxygen content? Hydrazine l Ammonia l Hydrogen l Ammonium Hydroxide
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l QUESTION: 013 (1.00) l onsibilities is NOT ; WHICH ONE transferred to (1) theof the following Technical emergency Support Center resp (TSC) or Emergency . Operations :
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Facility (EOF) when they are activated? Emergency Directo . Emergency Notification System Comunicator- . Offsite Agency Comunicator ' Assembly Area Coordinator , i i QUESTION: 014 (1.00) : I WHICH ONE (1) of the following is the time limit for notifying the State of' j Texas and Matagorda County after the Emergency Director has. initially' : classified an event? l t minutes j i i minutes I hour l
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SENIOR REACTOR OPERATOR Page 14 l QUEST 10N: 015 (1.00) Given the following:
- Unit 1 is at 100 % powe . - The Rod Control STARTUP switch is inadvertently placed in the j RESET positio Rod Control is in AUTOMATI &
WHICH ONE (1) of the following describes how this action will affect the Rod Control System? Rod stop interlock C-11 cannot prevent further control bank D outward rod motio The plant will not respond to a dropped rod in control banks B, C, or I Any " Urgent Failure" alarms currently active will clea The " ROD BOTTOM" alarm actuate ,
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QUESTION: 016 (1.00) WHICH ONE (1) of the following is the reason that all the RCP breakers OPEN on underfrequency? ! To protect the RCP from damage to the anti-rotation device due to , abnormal coastdow To preserve the RCP flywheel kinetic energ To avoid water hammer transients in the RCS induced by rapid RCP speed chang To reduce the probability of a stress induced RCP sheared shaft i
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QUESTION: 017 (1.00)
-; .Given the following: ; -
Unit 1 is in MODE RCP 1A motor is uncoupled from the pum l
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Loop 1 is ful l
- Maintenance is working on IA RC :
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WHICH ONE (1) of the following prevents leakage of.. reactor coolant'up the .;
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RCP shaft?
-i Nozzle dam installation prevents.RCS water from entering the. RCP ;
shaft' are :
, Seal leakoff collects any RCS leakage up the shaft and directs it f back to the VC j i The pump shaft mat'es with the top of the themal barrier !
assembl ;
, Seal injection is maintained during this conditio ,
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< QUESTION: 018 (1.00) WHICH ONE (1) of the following leaks could result;in a DILUTION of the RCS?
: Letdown Heat Exchanger leak - Excess Letdown Heat Exchanger leak l Reactor Coolant Drain Tank Heat Exchanger' leak
; Seal Water Heat Exchanger leak
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PRESSURE of 15 psig Lin the VCT during normal at power operation? To ensure adequate hydrogen concentration in th'e RCS coolan To ensure coolant flow to RCP seal # ,
' To prevent tank collapse due to internal vacuum during a multiple CCP star .i l To provide adequate CCP recirculation backpressure during nonnal )
operations, j
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' QUESTION: 020 (1.00) ; i Given the following: ;
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No ESF equipment is out of servic ;
- Unit 2 experiences a Safety Injection (MODE 1) signa !
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- ESF Status Monitoring is indicating a lit component lamp and a i
! flashing annunciato i
- All other status monitoring indications are NORMAL for Safety - !
Injection (MODE 1). ! ! : WHICH ONE (1) of the following is the status of the ESF component? ;
! The component has been bypassed by the operato . The component failed to actuate or properly positio The component has been reset by the RESET pushbutton and a valid y signal is still presen The component is inoperable and the alam has been acknowledged l by the operato ,
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.i QUESTION: 021 (1.00)
Given the following:
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- A LOCA occurred on Unit Normal offsite power is availabl Containment pressure is 5 psi ;
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WHICH ONE (1) of the following conditions will cause the-sequencer to start , the Containment Spray pumps? .l
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Time after sequencer MODE 1 Containment Pressure signal initiation ; seconds 7.5 psig i seconds 10 psig i seconds 7.5 psig seconds 10 psig ;
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i I l I QUESTION: 022 (1.00) : Given the following:
- Unit 1 is in MODE 1 at 100% power, ;
WHICH ONE (1) of the following is the most accurate method of determining j reactor thennal power? 7 l
- Manually calculated average of four Power Range Nuclear ,
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Instrument I ( Turbine generator output loa i Proteus calculated secondary calorimetric powe I ! Run an incore power flux ma l
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! Given the following: ,
- Unit I has been operating at 80% for the last 48 hour No AFD penalty time has been accumulate The AFD monitor alarm was disabled at 060 Target AFD for 100% power is + 3.75%. - Indicated AFD has been continually monitored and logged as .
follows: TIME N41 N42 N43 N44 0600 + + + + + + + + + + + + + + + + . l 0700 + + + + l
At 0700, WHICH ONE (1) of the following is the cumulative penalty l deviation time for the previous hour? a. 5 minute I b. 21 minute i c. 26 minute d. 38 minute ! l l l i
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! i l l QUESTION: 024 (1.00) Given the following: l
- Unit 2 is at 100 % powe One digital rod position indicator is inoperable.
I WHICH ONE (1) of the following actions is used to determine the position of the non-indicating rod per Technical Specification 3.1.3.2, " Position . Indication Systems"? Use the incore system with a magnetic proximity probe attachment to detect the rod tip positio Perform current measurement of coils on the CRDM.
I Monitor and analyze QPTR chang Use incore detectors to perform a flux map of the core.
! QUESTION: 025 (1.00) Given the following:
- Unit 1 is at 100% power - The Digital Rod Position Indication System (DRPI) has experienced a failure of the Data "B" cabinet.
! WHlCH ONE (1) of the following is the accuracy of a lit rod position LED for this condition? +4 to -4 steps +4 to -10 steps
{ +10 to -4 steps 1 +10 to -10 steps l
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QUESTION: 026 (1.00) l Given the following: l
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A LOCA has occurred on Unit i
- All systems are in their normal operating alignmen ;
WHICH ONE (1) of the followin is providing power to 11A and 12A Reactor ; Containment Fan Coolers (RCFC ? , Load Center EIC MCC Bus IF Load Center EIA ,
! MCC Bus 1G5 i
QUESTION: 027 (1.00) Given the following:
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Reactor power is 13%. ,
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S/G 1A level is 78%.
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S/G 1B level is 88%.
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S/G IC level is 76%. !
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S/G ID level is 75%.
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WHICH ONE (1) of the following lists AUTOMATIC actions that should result : from the above situation? l l a. Turbine trip, Reactor trip, Feed Pump tri l l b. Turbine trip, feedwater Isolation, feed Pump tri c. Reactor trip, Feedwater Isolation, Feed Pump tri I d. Turbine trip, Reactor Trip, feedwater Isolatio l
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SENIOR REACTOR OPERATOR Page 21 QUESTION: 028 (1.00) WHICH ONE (1) of the following will prevent ANY automatic start of a motor driven AFW pump?
! Low pump suction pressur The pump discharge valve is close . The pump control switch is in STOP on the MC The transfer switch is in the ASP positio :
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QUESTION: 029 (1.00) ! l l WHICH ONE (1) of the following is a loadon 125 VDC switchboard E1D11? ] Pressur Mer PORV 656 i DC control panel ZLP-101 (field flash). Auxiliary Feedwater Valve RCP switchgear control powe QUESTION: 030 (1.00) WHICH ONE (1) of the following components is the largest source of radioactive gas sent to the Gaseous Waste Processing System (GWPS), during normal full load operations? Volume Control Tank i Boron Recycle Evaporator Reactor Head Degassing Decay Tank i i Reactor Coolant Drain Tank , l l l l i
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QUESTION: 031 (1.00) WHICH ONE (1) of the following events would result in an AREA' radiation- l monitor alarm?
; .RCS Leak at the Incore Seal Table I Main Steam Line Break inside Containment ,
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: Steam Generator Tube Rupture
. Pressurizer PORV seat leakage l i
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1 { l l QUESTION: 032 (1.00) l WHlCH ONE (1) of the following' combinations of radiation types are detected ! by an Area Radiation Monitor? , i Beta and Alpha
! Alpha and Neutron Beta and Gamma
/j l l Gamma and Neutron ,
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QUESTION: 033 (1.00) Given the following: . -tl
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- Loop 1, 3, and 4 Tavg meters indicate 593 degrees j
- Loop 2 Tavg meter indicates off scale HIG '
- Loop 1, 3, and 4 Delta T meters indicate 100%.
- Loop 2 Delta T meter indicates 0%. ;
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WHICH ONE (1) of the following is the cause of these indications?- ] Loop 2 Tcold failed LOW.
- Loop 2 Tcold failed HIG Loop 2 Thot failed LO Loop 2 Thot failed HIG '
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i QUESTION: 034 (1.00) Given the following:
- Unit I has tripped from 100% power due a Loss of Coolant Accident j (LOCA) with a Loss of Offsite Power (LOOP). l -
RCS pressure is 375 psi !
- All RWST level indicators are at"15%. -
Emergency Diesel Generator ill has failed to star .
- Low Head Safety Injection pump IC (LHSI) has tripped on ,
overcurren N0 flow is indicated for LHSI pump 1 ;
; - All other equipment is operating nomall WHICH ONE (1) of the following is the reason LHSI Pump 1B has N0 flow I indication? ,
l a.- LHSI pump Containment Emergency Sump _ Suction valves have failed l to open automatically on low RWST leve ' l .16KV Bus ElB has no powe : RCS pressure is above LHSI pump shutoff hea j Load sequencer is transitioning from Mode II to Mode'II i
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l , QUESTION: 035 (1.00) l WHICH ONE (1) of the following is an input to the Cold Overpressure Mitigation System (COMS) for PCV-655A7 Auctioneered high wide range Tho Auctioneered high wide range Tcol RCS pressure transmitter PT-40 RCS pressure transmitter PT-40 i
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QUESTION: 036 (1.00) Given the following:
- All pressurizer heaters were on IN AUTO prior to actual t
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pressurizer level dropping below the heater low-level cutoff- ! setpoin Pressurizer level is restored to above.the low level reset poin ;
- Assume no operator action other than refilling the pressurize ,
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- Assume no Safety Injection or LOOP signals present. . ; - Assume pressurizer pressure is 2200 psi !
WHICH ONE (1) of the following describes how the heaters will respond? Variable and backup heaters will come on AUTOMATICALL j Variable and backup heaters will remain of Variable heaters ONLY will come on AUTOMATICALL Backup heaters ONLY will come on AUTOMATICALL !
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QUESTION: 037 (1.00) Given the following:
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Unit 1 is at 80% powe A surveillance is in progress to test the reactor trip breaker An operator calls the control room and' reports that he found ALL : reactor trip and bypass breakers RACKED IN and CLOSE WHICH ONE (1) of the following actions should be taken? I a. Immediately OPEN one trip or bypass breaker and continue with the surveillanc b. Suspend the surveillance in progress and restore normal trip breaker configuratio c. Immediately OPEN 9e trip and one bypass breaker on each train in a manner that wi not cause a reactor + ri d. Trip the Reactor and enter OPOP05-E0-E000.
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QUESTION: 038 (1.00) WHICH ONE (1) of the following will cause the Overtemperature Delta T reactor trip setpoint to INCREASE? Tavg increase : Pressure increase . Delta flux increases.
'I Power increase QUESTION: 039 (1.00) WHICH ONE (1) of the following describes the reason Auctioneered High Nuclear Power is used as a control input to the Low Power Feedwater Control Valves?
' Required feedwater flow is directly proportional to nuclear power at low power level It anticipates changes in heat flux into the SG at low power levels, Steam flow signals are not accurately pressure compensated at low '
power level Feedwater valve control based on level only will cause large flow oscillation ,
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l l ! QUESTION: 040 (1.00) i
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WHICH ONE (1) of the following combinations of Spent Fuel Pool (SFP) makeup - sources ensures that boron concentration would be MAINTAINED or INCREASED j if t:ater was added to the SFP7 l l Reactor Makeup Water and Boron Recycle Syste j Boron Recycle System and Demineralized Wate . i RWST and Reactor Makeup Wate Boron Recycle System and RWS i i
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QUESTION: 041 (1.00) l i Given the following: j
- DG 11 is in parallel operatio }
l - The DG mode selector switch is in the PARALLEL positio l l
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A reverse power condition is sensed on bus EI l i I j WHICH ONE (1) of the following will occur? ! ! ! l The DG will increase loa l r
- The DG will overspeed and tri ; The DG output breaker will trip ope : The DG will only respond to the emergency stop " PULL TO TRIP" ,
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QUESTION: 042 (1.00) , WHICH ONE (1) of the following describes the power supplies to the ESF l 4.16KV bus during the specified mode of electrical system operation? Preferred - Associated 13.8KV standby bus, Standby- ESF DG, l Emergency - Switchgear I Preferred - Associated 13.8KV standby bus, Standby- Switchgear , IL, Emergency - ESF D Preferred - Switchgear IL, Standby - ESF DG, Emergency - Associated 13.8KV standby.
I Preferred - Switchgear IL, Standby - Associated 13.8KV standby l bus, Emergency - ESF DG.
! l l QUESTION: 043 (1.00) WHICH ONE (1) of the following types of smoke / fire detectors is used to i i detect a fire in the carbon filters of the HV/2 units throughout the plant? Thermostatic line heat detector, Ionization smoke detecto : l l Optical flame detector Thermistor heat detecto !
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QUESTION: 044 (1.00) l l Given the following-
- Unit 1, Fire Pump #1 is in AUTO.- ,
- All control panel switches are-in their normal position BATT "A" CONNECTED and BATT."B" CONNECTED Blue lights are LI ~
- Fire system pressure is 130 psi . .
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- An operator has pushed the STOP/ RESET pushbutton to stop the ,
pum The pump' restarted when the pushbutton was release j
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WHICH ONE (1) of the following conditions could be the cause of the fire ! pump restart? l l l Low fire system pressur ;
; Unit 2 has actuated a remote start signa l
, The control panel has lost AC powe j i The Weekly Test Timer (WTT) has actuate !
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- 045 (1.00) l f !
WHICH ONE (1) of the following actions must be taken within 1 hour if l l containment pressure is at +0.5 psig according to Technical Specification ) 3.6.1.4 " Containment Systems - Internal Pressure"? ! Take action to place the unit in a MODE that the specification i ! does not apply.
l Initiate a containment normal purge.
j Restore pressure to within limit Perform an RCS leak rate calculatio ! l
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QUESTION: 046 (1.00) WHICH ONE (1) of the following describes the action associated with a HIGH l alarm on Liquid Waste Processing System (LWPS) discharge monitor RT-8038 ) ' during a release? !
/ LWPS Diversion Valve (FV-4077) closes and terminates all-
discharg )
' LWPS Diversion Valve (FV-4077) diverts flow back to the Waste Monitor Tank ,
i Teminates flow to the Waste Monitor Tank i Trips the Waste Monitor Tank pump .
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QUESTION: 047 (1.00)
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Given the following:
- Unit 1 is in MODE I - RCS temperature is 325 degree RCS pressure is 340 psi ! -
RHR is in servic t
- An unisolable leak in the Instrument Air (IA) system has '
l occurre !
- IA system pressure is 60 psig and decreasing'.
WHICH ONE (1) of the following describes how the RHR system will respond? i RHR heat exchanger bypass valves FCV-851/852/853'will fail OPEN j and cause RCS temperature to DECREAS : I RHR heat exchanger bypass valves FCV-851/852/853 will fail CLOSED and cause RCS temperature to INCREAS ; RHR heat exchanger flow control valves HCV-864/865/866 will fail l OPEN and cause RCS temperature to DECREAS l t I RHR heat exchanger flow control valves HCV-864/865/866 will-fail ! l CLOSED and cause RCS temperature to INCREASE.
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QUESTION: 048 (1.00) , t
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WHICH ONE (1) of'the following Service Air / Instrument Air (SA/IA) system valves will be the LAST to reposition on a DECREASING Instrument Air header pressure? .SA to IA Crossover :
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l SA Dryer Bypass ! IA Containment Isolation i QUESTION: 049 (1.00) WHICH ONE (1) of the following satisfies the conditions that must exist in !
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order for the containment Nomal Purge supply and exhaust isolation valves l to be considered " sealed" in accordance with PSP 03-SI-0016, " Containment ! l Integrity . Checklist"? l The valves are CLOSED and the power supply breakers are. LOCKED . l ' OPE j
! The valves are CLOSED and the power supply breakers are LOCKED 1 CLOSE i
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l L! QUESTION: 050 (1.00) ] WHICH ONE (1) of the following_ combinations of Residual Heat Removal (RHR)- ! loops may be' used for low pressure letdown operations? :l
: Loop "A" and "B".. ]
l l Loop "A" and "C".
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i Loop "B" and "D".
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QUESTION: 051 (1.00)
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l Given the following:
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- Unit 1 is at 100% powe , - All' systems are in their normal alignmen ; - CCW Train 1A is in servic i - CCW level indicates 64%.and is stabilizin )
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- The STANDBY CCW pump has started.
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- ASSUME all automatic actions have occurred as designe l
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, WHICHONE(1)ofthefollowingCCWsystemvalveswill--beclosedatthis 1 indicated CCW surge tank level? ! i " SUPPLY ISOLATION MOV-0768". I
! " BRANCH ISOL MOV-0297". j i " RETURN X-CONN FV-4657". j l
i l "CCW RET HDR ISOL MOV-0192". :
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QUESTION: 052 (1.00) Given the following: I
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Unit 1 is at 100% powe Essential Cooling Water (ECW) pump "A" is RUNNIN ECW pump "B" control switch is in AUTO and the ECW/CCW TRAIN SELECTOR switch is in STANDB ECW pump "C" is NOT RUNNING (AUTO after STOP). , WHICH ONE (1) of the following will automatically start ECW pump "B"? ' CCW common discharge header pressure of 70 psi ECW train "A" and "C" pressures of 50 psi ; Placing ECW pump "B" switch on panel ZLP-654 in LOCA : Essential Cooling Pond Makeup Pump "A" trip QUESTION: 053 (1.00) l Given the following:
- Unit 1 is at 30% powe Power escalation is in progres Control rods are in MANUA >
- After a control rod withdrawal of several steps, the rods continue outward when the IN-HOLD-0VT lever is returned to the neutral positio WHICH ONE (1) of the following actions should be taken? l Place the Rod Bank Selector Switch in AUTO and check for j continued rod motio ;
' Trip the Unit and enter OPOP05-E0-E000. " Reactor Trip or Safety Injection". Check for failed instruments and select or block appropriate instrument inputs, Hold the IN-HOLD-0UT lever to the IN position and check for continued rod motio l l
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_ _______ _ SENIOR REACTOR OPERATOR Page 33 QUESTION: 054 (1.00) l Given the following:
- Unit 1 is at 30% powe Control rods are in MANUA WHICH ONE (1) of the following is a symptom of an immovable control rod per OPOP04-RS-0001, " Control Rod Malfunction"? ,
l l Tavg - Tref deviation of 3 degrees ' l l " DELTA T R0D WITHDRAWAL BLK ALERT" alar One or more RPIs disagree by +8 steps, " ROD CONTROL URGENT" alarm.
! QUESTION: 055 (1.00) Given the following:
- Unit 1 is operating at 95% powe One rod has dropped into the cor Actions are being taken per OPOPO4-RS-0001, " Control Rod Malfunction".
- QPTR is determined to be 1.0 l t Reactor power must be reduced to WHICH ONE (1) of the following per t Technical Specification 3.2.4, " Quadrant Power Tilt Ratio"? See attached Technical Specification sectio % powe % powe % powe % powe . ._ _ _ _ _ . _ _ i
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i i QUESTION: 056 (1.00)
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WHICH ONE (1) of the following represents the MINIMUM values for RCP i bearing temperatures that require the operator to trip the reactor and any ! affected RCP, per OPOP04-RC-0002, " Reactor Coolant Pump Off Normal"? . degrees F (radial bearin or230degreesF(seal ; water /bearingtemperature).
' f degrees F (radial bearings) or 187- degrees F (seal water / bearing temperature). .i
; degrees F (radial bearings) or 187 degrees F (seal l water / bearing temperature). l t
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, I degrees F (radial bearings) or 195 degrees F (seal water / bearing temperature). j i t QUESTION: 057 (1.00)
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Given the following:
- Unit 1 is operating at 100% powe .
i ' - RCP 1A NO. 1 SEAL LKF FLOW HI/LO alann is '_I THERMAL BARRIER FLOW / TEMP TRBL alarm is LI l
- RCDT level increasing at a higher rate than normal. . ; - No.1 Seal leakoff flow recorder indicates 6 gp ; - Make-up to the RCS has increased 40 gpm to maintain pressurizer l leve l WHICH ONE (1) of the following has occurred? -
l No. I and No. 2 seals have failed and the RCS pressure drop is l across the No. 3 seal.
! The No. 2 seal has failed and is allowing water _from the standpipe to flow out the No. I seal leakoff lin The No. I seal has failed and the RCS pressure drop is across the No. 2 sea The No. 2 and No 3. seals have failed and the RCS pressure drop is across the No.1 seal.
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t QUESTION: 058 (1.00) I WHICH ONE (1) of the following parameters discriminate between a large
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secondary loss of coolant accident, apJ a large primary loss of coolant ' accident IMMEDIATELY after the accideat occurs? Pressurizer level ; t RCS Thot . Pressurizer pressure RCS Delta T ,
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QUESTION: 059 (1.00) Given the following:
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A LOCA has occurred on Unit OPOP05-E0-E000, " Reactor Trip or Safety Injection" is being performe WHICH ONE (1) of the following indications has an alternate limit during adverse containment conditions? RCS pressur AFW flo SG pressur SG narrow range level.
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I i t i QUESTION: 060 (1.00)
WHICH ONE (1) of the following is the reason for reducing Tave to less than , 500 degrees F following a shutdown required by a Dose Equivalent I-131 .,
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level of greater than 1 microcuries per gram? i Slows cool' ant / fuel reaction rate, immediately reducing the source term of the activit ' Prevents the release of activity following a steam generator tube ruptur : i Minimizes the temperature related degradation of the CVCS , demineralizers while the RCS clean-up is in progres i i Minimizes of the iodine spiking phenomena which occurs due to the !
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large change in THERMAL POWER level caused by the unit shutdow i i
QUESTION: 061 (1.00) ; WHICH ONE (1) of the following is the criteria used to determine if.the l alternate mixed bed demineralizer should be placed in service during a high j reactor coolant activity event per OPOP04-RC-0001, "High Reactor Coolan l
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Activity"? I Chemical analysis determines the Decontamination factor (DF) is ; inadequat ( l CVCS Letdown Failed Fuel Monitor RT-8039 in alar l Letdown flow has been increased to maximu RCS gross activity exceeds 50% of the limit specified by Technical Specifications.
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SENIOR REACTOR OPERATOR Page 37 QUESTION: 062 (1.00)
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WHICH ONE (1) of the following is the MINIMUM flow rate required to be established when Emergency Boration is established from the RWST7 gpm , gpm
, gpm gpm QUESTION: 063 (1.00)
Given the following:
- Unit I has experienced an ATWS from 100% powe Emergency Boration has been initiated in accordance with FRS1, " Response To Nuclear Power Generation /ATWS".
WHICH ONE (1) of the following is the TERMINATION criteria for Emergency Boration? RCS boron concentration is greater than that required by the shutdown margin limit curve for cold shutdow All rods indicate on the botto The reactor is subcritica Emergency Boration flow has been maintained at MAXIMUM flow for a thirty (30) minute time period for all conditions.
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t ! ' SENIOR REACTOR OPERATOR Page 38 l i QUESTION: 064 (1.00) Given the following:
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A reactor startup is in progres P-10 status light is NOT lit.
, WHICH OaE (1) of the following 120 VAC Vital bus failures will result in a l reactor trip? , DP-001 DP-002 DP-1203 DP-1201
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! i QUESTION: 065 (1.00) Given the following: l - A Hain Steam Line Rupture occurred 60 minutes ago.
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- Initial Tcold was 300 degrees F.
I - Current Tcold is 180 degrees Current RCS Pressure is 500 psi WHICH ONE (1) of the following is the Critical Safety Function path the plant is in due to the above conditions? Doerational Limits Curve is attache Green Yellow Orange Red I i
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QUESTION: 066 (1.00) ; l WHICH ONE (1) of.the following is an indication of less than adequate. core- ! cooling POST LOCA? Consider each case separatel l l RVWL plenum level indicates 25%. l l ( i Core exit themocouples indicate 650 degrees RCS subcooling indicates 20 degree ! Two (2) RCS hot leg temperatures indicate 345 degrees F.
, i I i QUESTION: 067 (1.00) ;
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WHICH ONE (1) of the following is the reason for stopping steam generator j ! depressurization at approximately 231 psig when responding to an inadequate- I l core cooling event? To prevent the steam generator tubes from being uncovered durin depressurizatio ;
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j l To prevent injecting nitrogen from the accumulators'into the RC To minimize potential radioactive releases to the atmosphere ) during depressurizatio ! To maintain RCS/ Secondary Delta T at values allowing.RCP restar .
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i I SENIOR REACTOR OPERATOR Page 40 QUESTION: 068 (1.00) Given the following:
- The reactor has tripped due to a loss of Main Feedwate Auxiliary feedwater has failed to actuat A Red Path for Core Cooling is indicate WHICH ONE (1) of the following sets of core exit thermocouples are used to ,
determine the existence of a RED path per OPOP05-EO-FRC1, " Response to Inadequate Core Cooling"? , < Five (5) highest reading therocouples.
l ! Average of ten (10) highest thermocouple Average of all thermocouples.
l Four (4) highest reading thermocouples in each quadran QUESTION: 069 (1.00) WHICH ONE (1) of the following is an indication that is NOT displayed on the Auxiliary Shutdown Panel (ASP)? ECW pressur RHR pump curren RCS cold leg temperatur Charging flo l
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- A Control Room Evacuation has been ordered by the Shift )
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- None of the IMMEDIATE Action Steps where able to be performed ,
prior to evacuating the Control Roo l
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WHICH ONE (1) of the following is the FIRST action that is required to be -; performed per OPOPO4-Z0-0001, " Control Room Evacuation"? ! l Direct the Operator in ESF-Train C switchgear room to verify'the reactor trip and bypass breakers are ope , Establish communications with various control stations on sound . powered phone circuit , Place the Auxiliary Shutdown Panel (ASP) switches in the. ASP positio ,
$ Announce Reactor Trip and Control Room Evacuation using the plant i paging syste j i
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l QUESTION: 071 (1.00)
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Given the following: [
- Unit 1 is at 90% powe l - " MAIN COND VACUUM LO" alarm is li ! - "LP TURB EXH HOOD TEMP HI" ' alarm is li ( - Condenser vacuum is 22 inches Hg and DECREASING slowl WHICH ONE (1) of the following actions should be taken? !
! j a. Commence an orderly shutdown of the Main Turbine, b. Trip the Reactor, then trip the Main Turbin c. Trip the Steam Generator feedpump turbines (SGFP). Increase turbine load to allow increase; cxhaust hood coolin i
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I 1 SENIOR REACTOR OPERATOR Page.42 : , I l : i . QUESTION: 072 (1.00) Given the following: l
- Unit 1 is operating at 30% powe ! - All control systems are in AUTOMATI The "1A" Reactor Coolant Pump has just trippe t WHICH ONE (1) of the following is the overall plant response? ,- !
l The reactor trips on a LOW RCS FLOW conditio Unit power is reduced to approximately 22% power. (1/4 oi original l power level). l Unit power remains the same with steam flow increasing on the other three steam generator .
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l The reactor trips on HIGH steam generator level when "A" Steam j generator level swell ;
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QUESTION: 073 (1.00)
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Given the following: , i
- Unit 1 is at 100% powe l - Pressurizer pressure selector switch is in the P457/456 positio !
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- Pressurizer pressure control is in automati WHICH ONE (1) of the following describes the pressurizer PORVs response if Pressurizer Pressure channel 457 fails HIGH7 i i PCV 655A will not close IF RCS pressure decreases below 2185 !
psi ;
, PCV 656A will not close UNTIL RCS pressure decreases below 2185 1 psi l
, PCV 655A will ope PCV 656A will ope '
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QUESTION: 074 (1.00) 'l WHICH ONE (1) of the following is the reason for the order that the valve f positions are checked in Step 3. " Check if RCS is Isolated"'of " f l OPOP05-E0-EC00, " Loss of All AC Power"? a. They are listed according to control board locatio : i b. Those most likely to fail in a loss of AC power are listed firs . j
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c. Those most likely to have an RNO corrective action outside the j control room are listed last.
l d. They are listed according to capacity and potential for inventory - los ;
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t QUESTION: 075 (1.00) BHICH ONE (1) of the following is a control room indication that the l Technical Support Center (TSC) Diesel Generator should be started per- l l OPOP05-E0-EC00, " Loss of All AC Power"? ]
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i ' Positive Displacement Pump (PDP) indicating lights are OF ; f 4 l Essential lighting starts to go di l Annunciator " LOSS of AC VOLTAGE to BATTERY CHARGER" actuate ,
! Normal supply breaker to LC IW indicates OPEN.
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SENIOR REACTOR OPERATOR Page 44 - l l QUESTION: 076 (1.00) WHICH ONE (1) of the following conditions will assure " acceptable"- ; consequences following an Anticipated Transient Without-Scram (ATWS) Event? Turbine is tripped within thirty (30) seconds and Auxiliary- ., Feedwater is established within sixty (60) second !
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b. - Turbine is tripped within sixty. (60) seconds and Auxiliar ..- t Feedwater is established within sixty (60) second ; i Turbine is tripped within sixty (60) seconds and Auxiliary Feedwater is established within ninet Turbine is tripped within .ninety (90)y (90)and seconds second Auxiliary Feedwater is established within ninety (90) second : i
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QUESTION: 077 (1.00) l l WHICH ONE (1) of the following actions should be taken if a Safety ; Injection (SI) signal occurs while performing OPOP05-ES-FRS1, " Response To :
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Nuclear Power Generation /ATWS"? Perform FRS1 to completion then enter OPOP05-E0-E000, " Reactor f Trip Or Safety Injection". . Perform FRS1 and OPOP05-E0-E000, " Reactor Trip Or Safety i Injection", simultaneously until Reactor Trip Breakers are.0 PE i then exit FRS l 1 Perform FRS1 until Reactor Trip Breakers are OPEN then enter . OPOP05-E0-E000, " Reactor Trip Or Safety injection". I
! Perform FRS1 through completion of Addendum 2 " Verification'of' l SI Equipment Operation", then return to OPOP05-E0-E000. " Reactor ;
Trip Or Safety Injection" step . l
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QUESTION: 078 (1.00)
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- The RCS has had a stuck open Pressurizer safety valv The reactor tripped and safety injection initiate The RCS rapidly depressurized to saturation condition Pressurizer level initially dropped and then began to rise ,
rapidly. WHICH ONE (1) of the _ following characterizes the . : i relationship between' pressurizer level and RCS inventory under l these conditions? . Level is an accurate indication of inventory, because voiding would occur first in the pressurizer due to the high temperature ! of the pressurizer wall j Level is an accurate indication of inventory, because hydraulic -
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pressure would force any voids to the pressurizer steam space and l ' out the safet Level is NOT an accurate indication'of inventory, because RCS l voiding may result in a rapidly increasing pressurizer leve , Level is NOT an accurate indication of inventory, because at ' higher temperatures the cold calibrated pressurizer level channels falsely indicate hig :
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QUESTION: 079 (1.00)
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WHICH ONE (1) of the following is the significance of blowing the loop seal ; during a cold leg small break LOCA? l r
' System mass loss will continue prior to the steam vent path being established from the core to the brea Single phase cooling is enhanced while the loop seal is still intac Reflux cooling becomes a viable cooling means following the loss-of the 1 cop sea ; The heat sink effect of the water in the intermediate leg is lost I when the loop seal is lost, degrading core coolin i
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SENIOR REACTOR OPERATOR Page 46 i I , . QUESTION: 080 (1.00) .
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WHICH ONE (1) of the following is the reason the RCP's are required'to be tripped at 1477 psig during a small break LOCA? ; To eliminate RCP heat input into the RC i To avoid RCP impeller damage as dissolved gases' come out of : solution at lower pressure .
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; To avoid the higher clad temperature consequences of RCP trip l later during the acciden :'
, To prevent RCP seal damage during subsequent RCS depressurizatio .;
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QUESTION: 081 (1.00) i l Given the following: l l
- Unit I has been shutdown for 10 days following~100 days at 100% l powe : - The RCS temperature is 120 degrees ; - The RCS is at midloo ! -
A total loss of RHR occur !
- No core cooling is re-establishe j u WHICH ONE (1) of the following is the MINIMUM time assumed per PGP03-ZO- ~
0035, " Reduced RCS Inventory Operation," for the RCS to reach boiling- -! conditions? minutes i minutes I l minutes ] l minutes _-. - _ _, . .-. . . _ - .. -
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QUESTION: 082 (1.00) WHICH ONE (1) of the following sets of parameters affects vortexing in the RHR suction piping? (reduced inventory operations) l RHR flow rate AND RCS leve I RHR flow rate AND RCS pressure, j Number of RHR pumps running AND RCS pressur .I
! Number of RHR pumps running AND RCS level.
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QUESTION: 083 .(1.00) i ! Given the following:
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Unit 2 is in MODE ;
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Reactor power is 1E-08 amp ;
- Critical data is being take N-35 intemediate range control power fuses blow due to an internal faul WHICH ONE (1) of the following is the appropriate action? 1 Enter OPOP05-E0-E000. " Reactor Trip 'or Safety Injection," Step . l Hold power at 1E-08 amps until repairs are mad Insert rods to lower neutron flux level until both source range NIs energiz Place the N-35 Level Trip switch in the bypass positio l
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SENIOR REACTOR OPERATOR Page 48 QUESTION: 084 (1.00) l Given the following:
- A steam generator tube rupture has occurred in Unit !
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- The RCS is at 520 degrees Containment pressure and temperature are noma The ruptured steam generator has level indication in the control room of: - 99% narrow range - 85% wide range .
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WHICH ONE (1) of the following is the reason for the wide variance in level indication? ; Due to rapid pressure fluctuations experienced in the ruptured steam generator, the water has been drawn out of the narrow range reference leg.
I Temperature stratification in the steam generator is exposing the l l transmitters to different density water, causing inaccurate level j l indicatio The wide range reference leg condensing pot has been overfilled I due to the high level condition resulting in an erroneously low l indicatio : i The wide range is cold calibrated and does not compensate for the ! density differences between the SG water and the reference leg.
l l QUESTION: 085 (1.00) , i WHICH ONE (1) of the following determines the temperature at which the RCS cooldown is teminated following a Steam generator tube rupture when using OPOP05-E0-E030, " Steam Generator Tube Rupture"? RCS pressur Maximum RCS temperature for RHR initiatio Ruptured S/G pressur RCS subcoolin l i I
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QUESTION: 086 (1.00) WHICH ONE (1) of the following is the MINIMUM necessary to verify that the , turbine has tripped following a reactor trip per OPOP05-E0-E000, " Reactor i Trip or Safety Injection"? l, Turbine Throttle valves Turbine Governor Valves Open Open , Closed Open Open Closed I Closed Closed .! l
QUESTION: 087 (1.00) ,
Given the following: ) l
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Unit 1 is operating at 100% powe All controls are in the normal power operation lineu !
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Pressurizer level is DECREASIN j
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VCT level is INCREASIN i
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SEAL WTR L0 FLOW alarm is li REGEN HX LETDN HI TEMP alarm is li LETDN HX OUTLET HI TEMP alarm is li CHARGING-FLOW HI/LO alarm is li ; 1 , l WHICH ONE (1) of the following explains the given conditions?' l l ! Pressurizer PORV failed ope I l , Loss of chargin Small break LOC Letdown isolation.
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- 088 (1.00)
WHICH ONE (1) of the following is correct regarding operation of a 4.16KV l saitchgear breaker that has been OPENED LOCALLY following a loss of DC control power? Breaker can be closed remotely one time if the breaker is in the TEST positio . . Breaker can be closed one time locally using the switch on the cubicle door.
l l Breaker cannot be closed again until the closing springs are ; l manually charge . Breaker cannot be closed until the breaker is reset by removing and then replacing the fuse block.
! . QUESTION: 089 (1.00) i WHICH ONE (1) of the following differentiates between an unisolable feedline break and an unisolable steam line break of the same size? t RCS heat removal would be greater from the steam line brea Containment pressure would be greater for a feedline brea Containment radiation levels would be higher from a steam line brea RCS depressurization would be greater from a feedline break.
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! SENIOR REACTOR OPERATOR Page 51 l l l QUESTION: 090 (1.00) ' WHICH ONE (1) of the following conditions requires tripping the reactor per OPOPO4-RC-0004, " Steam Generator Tube Leak"? Consider each case separatel Pressurizer level is at 27%. VCT level is 17% and increasing in AUTO contro . Narrow range SG 1evel is 68% and increasing slowl SG Blowdown Isolation valve SB-FV-5019 closes due to high
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QUESTION: 091 (1.00) Given the following:
- Unit I has tripped from 100% powe The operator initiated Safety Injection due to rapidly decreasing pressurizer pressur The following plant conditions were noted after the trip: ) - Pressurizer level 60!s and increasin RCS pressure 1700 psig and decreasin Containment Pressure 2 psig and increasing.
! - Containment radiation 0.5R/hr and increasin Containment sump levels increasin WHICH ONE (1) of the following events has occurred? Main Steamline break inside containmen , l l RCS cold leg brea Steam generator tube ruptur I Pressurizer vapor space LOC i
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QUESTION: 092 (1.00) ! ! Given the following: ,
- Following a Reactor Trip OPOP05-E0-E000, " Reactor Trip or. Safety )
Injection" is entere l
- Step 3b. states " Verify Power to AC ESF Busses, AC ESF busses - l ALL ENERGIZED."
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- It is determined that ESF bus E1C is NOT energize , l WHICH ONE (1) of the following describes.the required action? ,
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l Enter OPOP05-E0-EC00, " Loss of All AC Power" and perform required ! actions through step 30 and then exit to the appropriate recovery procedur Continue on with OPOP05-E0-E000 " Reactor Trip or Safety - Injection" while attempting to restore power to ESF i bus E1 Enter OPOP05-E0-EC00, " Loss of All AC Power" and perform required actions until ESF bus E1C is energize ' i Stop OPOP05-E0-E000 " Reactor Trip or Safety Injection" at step 3 until ESF bus E1C is re-energized by its Diesel < Generato ; I l ! !
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l QUESTION: 093 (1.00) i Given the following: ; i
- Unit 1 is at 100% powe , - All control systems are in their nonnal alignmen WHICH ONE (1) of- the following Pressurizer level failures results in'the - l greatest increase in reactor coolant inventory and which action is required . ;
to mitigate the failure? ; LT-465 fails LOW - Manually control Charging Flow Control Valve ;
(FCV-0205). LT-466 fails HIGH - Manually control Charging Flow Control Valve (FCV-0205).
! LT-466 fails LOW - Restore Letdown to servic LT-465 fails HIGH - Restore Letdown to servic ]
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i QUESTION: 094 (1.00) 3 WHICH ONE (1) of the following refueling conditions requires immediate, ' boration per Technical Specification 3/4.9.1, " Boron Concentration"? Keff is less than 0.9 RCS temperature is 68 degrees Washdown of refueling equipment that results in a boron concentration change of 20 ppm between two consecutive sample RCS boron. concentration is 1700. pp l I
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QUESTION: 095 (1.00) :
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WHICH ONE (1) of the following cases would the use of a General Radiation j Work Pennit be allowed, per OPRP08-ZA-0001, " Radiation Work-Permits"? Shuffling of Spent Fuel Assemblies in the Spent Fuel Poo !
! Entry into a Radiation Area in the Mechanical Aux Building (MAB) !
for surveillance testin .
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I Entry into a High Contamination Area to check a pumps oil leve Overhaul of a Spent Fuel Cooling pump tagged out of service for 5 ' t day :
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QUESTION: 096 (1.00) WHICH ONE (1) of the following requires that the procedure be designated as ,
' "IN HAND" per OPGP03-ZA-0010. " Plant Procedure Adherence and Implementation and Independent Verification? ! Single-step activities which must be coordinated at multiple 1 location l Activities that are prescribed by admistrative procedure i Activities that which are considered to be routin j Complex activities which must be performed in specified sequence I at a single locatio ,
QUESTION: 097 (1.00) WHICH ONE (1) of the following is responsible for performing the post-trip j data review investigation and the determining the cause'of a reactor trip? ' Unit Supervisor Plant Operations Review Comittee Plant Manager Shift Technical Advisor
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' QUESTION: 098 (1.00) WHICH ONE (1) of the following is necessary to allow a " Fail'Open" ai operated diaphragm valve to be used as a boundary valve for isolation ! purposes? The valve is closed and its air supply is tagged in.the open position to maintain air supply to the valv The valve shall be mechanically blocked in the closed position, ,
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if physically possibl The valve is closed and a DANGER tag is placed on the control switc ! The valve is closed.and has a check valve that will prevent f system flo l i
.I l QUESTION: 099 (1.00)
WHICH ONE (1) of the following describes the difference between Unit 1 and Unit 213.8KV electrical switching controls? ; The switch from 13.8KV Auxiliary bus 2J that feeds load center j l 12M also feeds 12G i
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1 The switch from 13.8KV Auxiliary bus 1J that feeds load center 12M also feeds 12F The switch from 13.8KV Auxiliary bus 1H feeds only load center 12 The switch from 13.8KV Auxiliary bus 2H feeds.only load' cent'er 12G.
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_ SENIOR REACTOR OPERATOR Page 56 i QUESTION: 100 (1.00) l WHICH ONE (1) of the following components is used in the Rod Control Tavg Control unit to anticipate the change in Tavg based on the trend value of Tavg? 1 Lead - Lag Compensator Non - Variable Gain Unit . Variable Gain Unit Rod Speed Programmer i
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Ro # 71 ) 500 " 6 b~ REV. 0 OPOP05-EO-T004 1NTEGRITY CRITICAL SAFETY FUNCTION STATUS TREE PAGE 1 OF 1 l ADDENDUM 1 , . INTEGRITY OPERATIONAL LIMITS ' Integrity OperationalI.imits
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II II 2560 psig -2 244 F m e - 231F
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lll / 218 F 2050 psig
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POWER DISTRIBUTION LIMITS 3 /4,2. 4 00ADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION l 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION: , With the QUADRANT TILT RATIO determined to exceed 1.02: Within 2 hours reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess i of I and similarly reduce the Power' Range Neutron Flux-High Trip l Setpoint within the next 4 hour Within 24 hours and every 7 days thereafter, verify that F,(Z) (by F,y evaluation) and Fj are within their limits by performing Surveillance Requirements 4.2.2.2 and 4.2.3.2. THERMAL POWER and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3. SURVElttANCE REOUIREMENTS l 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: l Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable.
! 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT l POWER TILT RATIO at least once per 12 hours by either: l Using the four pairs of symmetric thimble locations (Specification 3.3.3.2.a does not apply), or Using the movable incore detection system to monitor the QUADRANT POWER TILT RATIO subject to the requirements of Specification 3.3. *5ee Special Test Exceptions Specification 3.1 I SOUTH TEXAS - UNITS 1 & 2 3/4 2-10 Unit 1 - Amendment No. 39 Unit 2 - Amendment No. 30
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, HMP 1000A (12 91) , , i Houston Lighting & Power Company i OFFICE MEMORANDUM ,
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To NRC EVALUATION TEAM October 1, 1993 Fmm GLEN WELDON Su!pect ! REBUTTALS TO THE WRITTEN EXAMINATION t l i Attached are the rebuttals to specific questions on the Written Examination 199 administered to License Class #6 on September 24, , {
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We respectfully request that the following questions be reviewed: SRO Exam: #6, #10, #23, and #44 RO Exam:
#25, #49, and #96 (#25 and #49 are duplicates of #23 and #44 on the SRO Exam, respectively) ,
l If you have 972-7967 or any Wes questions, please Young at (512) contact Carl Brewer at (512) 972-766 ad cunrmads acap/c Q# /o S20 / h[ dAnJcD '
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! , l l l SRO EXAM: Question #6 ; i Given the following:
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Unit 2 is in Mode 4.
! - An INITIAL entry into the Reactor Containment Building (RCB) is being planne '
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OPGP03-ZO-0032, " Reactor Containment Building Entry" has been initiate .
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Oxygen levels were verified to be 21% 4 hours ag Which one (1) of the following pieces of equipment is required for I the team to enter into the RCB?
! An Oxygen monito A Neutron Survey Instrumen i A Self Contained Breathing Apparatus (SCBA). ,
i A Radiation Monitoring instrument with a 1 R/hr rang I KEY ANSWER: D i l Response to Ouestion Step 4.2.1 of OPGP03-ZO-0032, " Reactor Containment Building Entry", requires that a survey instrument " capable of measuring radiation , levels iD excess pf 1 R/hr" be use This makes the selected i answer of "D" incorrec ! l Step 5.5 (of the same procedure), requires that a sample of the i oxygen concentration be obtained "upon initial entry". . .not 4 hours */ l carlste? later as the question specifies. Therefore, "A" would be correct,
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except for the 4 hour time delay.
Based on this knowledge, we submit that there are no correct choices for this question.
j Ref: OPGP03-ZO-0032, Rev. 3
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Reactor Containment Building Entry OPGP03-ZO-0032-ms Rev. 3 Page 2 of 7 l l 1.0 Purpose and Scope i This procedure outlines the steps necessary for ALL entries into the Reactor Containment Building in Modes I and .2 This procedure outlines the steps necessary for INITIAL entries into the Reactor Containment Building (RCB) in Modes 3 through I
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2.0 Definitions 1 None , J 3.0 Precautions I I No entry SHA',L be allowed into an area in the RCB prior to a ' radiological survey by a Health Physics Technicia k VHEN oxygen content is suspected to be less than 19.51, cannot be i verified or airborne radioactivity levels dictate, THEN Self I Contained Breathing Apparatus (SCBA) shall be use l
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I Every effort SHALL be made to maintain personnel exposure As Low As l Reasonably /.:hievable (ALARA). )
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- The area under the reactor vessel (RCB Room 1) SHALL NOT be entered I with the unit in Modes 1 or l
; IF the most recent RCS sample indicates sustained levels of Tritium l greater than or equal to 1.0 uCi/m1, THEN SAMPLE RCB atmosphere for Tritium upon RCB entr '
4.0 Prerequisites A minimum of two personnel (one Health Physics Technician) has been assigned to each entry tea .2 Each team has the following equipmen [ 4. A ey instrument capable of measuring radiation levels in j'CXCiss O -* M 3 exces of 1 R/h h* * 4. Neutron Survey Instrument (only in Mode 1 & 2).
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Asv5%)c./* Cf ,. 4. An 02 m nito (only if oxygen levels have not been verified S-3 g u +s t R jna- within the 8 hours prior to entry.) * is ajeg -p M
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Reactor Containment Building Entry OPJP03-ZO-0032 Rev. 3 Page 4 of 7 , A RVP has been issued and activated that identifies all planned work or activities for the entr .7 HP has briefed all workers on radiological hazards prior to entr .8 The Shif t Supervisor has briefed the HP technicians on the scope of any containment closeout inspections require ~ The Shift Supe 1 visor has granted written permission on (-1) to enter ; the RC { i 5.0 ?rocedure
' The work group initiating the entry coordinates with the Unit Supervisor, Security, and Health Physics as to the time, reason, and activities for the entr . The Health Physics Technician obtains and completes the RCB Entry ,
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Checklist (-1). , ufo IT,necessary during an Emergency, THEN the Shift Supervisor may " AIT Ay , waive any or all of the prerequisites with the concurrence of th * _
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General Supervisor Radiation Protection or designe ho FM" Upon entering the RCB, PERTORM RCB handswitch line-up per
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1 POP 02-LI-000 l g [EfCES .-C OBTAIN a grab sample or breathing zone air sample and O 28 ample upon l TWirse g2[ pinitialentry,intotheRC .6 The Plant Engineering Department performs a RCS Pressure Boundary ; Inupections in accordance with the guidelines specified in ! OPGP03-ZE-003 .7 All personnel SHALL follow directions from the Health Physics ; Technician concerning ALARA concepts and remain together in a group l I while inside the RC .B Plant Operations is responsible f or perf omance of a Containment Closecut Inspection per OPSP03-XC-0002, unless no operators are involved in the entry with the plant in modes 1 or 2. In these instances, the Health Physics Technician entering the RCB completes this inspection for the areas designated by the Shif t Supervisor and I provides prompt notification to the Shift Supervisor of surveillance completio .9 Upon exiting the RCB, immediately NOTITY the Control Room that the entry has been completed.
l 5.10 CONFIRM with security that all personnel have exited containment.
! 5.11 ENSURE the personnel hatch is locked or guarded upon exiting.
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SRC EXAM: Question #10 WHICH ONE (1) of the following is an PERMIT ENTRY SPACE as defined l by the Confined Space Entry Program procedure? ! Blowdown Valve cubicles l Steam Generator 2A l
! Settling Basin ; Underground pipe ways
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KEY ANSWER: B
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Resoonse to Ouestion ' Step 2.3 of OPGP03-ZI-0007, " Confined Space Entry Prcgram", states f that undercround utility vaults and open-too sosaes more than four t feet deen such as pits, tubs, vaults, and vessels are considered, l by definition, a Permit Entry . Spac Step 2.1 of the same ! procedure defines a confined Space as "any enclosed or permit entry space". , l Based on this knowledge, we request that both "C" and "D" be ' considered acceptable answers, as well as "B". ! Ref: OPGP03-ZI-0007 (Rev. 5) 1 1
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I j Confined Space Entry Procram OPGP03-ZI-0007 l Rev. 5 Page 2 of 12 Purpose and Scope This procedure establishes the Confined Space Entry
. Program and establishes the methods to ;be used to obtain .
a permit, open, enter, or work within an ENCLOSED space or a PERMIT ENTRY space at the South Texas Project (STP) . ' This procedure specifies requirements for monitoring an ENCLOSED space or PERMIT ENTRY space for oxygen levels or combustible gas levels prior to personnel entry and i periodically while work is being conducted in the spac i l This procedure applies to all Departments involved with
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f STP and all STP personne i l Contractors and subcontractors snall provide their own ! instruments and training for confined space entries, unless granted written permission to do otherwise by the Industrial Safety & Health Division (ISH).
2.0 Definitions CONFINED SPACE: Any enclosed or permit entry-spac NOTE An ENCLOSED space should be identified with an Enclosed Space placard (Addendum 2) and a PERMIT ENTRY space should be identified with a Permit Entry Space placard -
(Addendum 3) unless prevented by engineering concern ' ENCLOSED SPACE: Any work-space location that is sufficiently enclosed to permit the existence of an oxygen daficient atmospher '
.l l PERMIT ENTRY SPACE: Any space having a limited means of 2 l
) egress, which is subject to the accumulation of toxic or q}[ l flammable contaminants or has an oxygen deficient atmo-i spher Confined spaces include, but are not limited c ! to: storage tanks, process vessels,_ binsu bsilers d ent-I _ilation or exhaust ducts, sewers,. underground utility > cg j ,y. d l vaults) tunnels, pipelines, and open-top spaces more than f Q* ) t f50f~ feet deep such as pits, tubs, vaults, and vessel .4 CREW LEADER: The person in charge, regardless of title 4T l or classificatio l
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RO EXAM: QUESTION #49 SRO EXAM: QUESTION #44 Given the following:
- Unit 1, Fire Pump #1 is in AUT All control panel switches are in their normal position BATT "A" CONNECTED and BATT "B" CONNECTED Blue lights are LI Fire system pressure is 130 psi ) - An operator has pushed the STOP/ RESET pushbutton to stop the'
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- The pump restarted when the pushbutton was release I I
WHICH ONE (1) of the following conditions could be the cause of the i fire pump restart? Low fire system pressur l Unit 2 has actuated a remote start signa c. The control panel has lost AC powe l i d. The Weekly Test Timer (WTT) has actuate > KEY ANSWER: B l l
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Response to Ouestion Fire Pump #1 auto starts on a Loss of AC power to the engine controller. This start signal ensures that the Fire Pump remains available even though the batteries may completely discharge during an extended loss of power. When the engine is running, the engine-mounted alternator, via rectifiers in the control panel, feeds the DC control circuits and charges the batterie Therefore, both "B" and "C" should be considered as acceptable , answer l l Ref. Vendor Manual 6004-00002PA LOT 201.29.HO.01 ggi l
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1. T11IS DPXlII:G SUPPLEMENTS THE CONTROLLER DIAGRAM DWG, NO. 770 I 2. TERMI!!AL A::D WIRE NUP.BERS CORRESPOND TO THOSE ON THE CONTROLLER REFERE::CE ::OTE NO. 3 ON DUG NO. 7702, 3. REFERE::CE D'.-:G NO. 7704 FOR COMPLETE DESCRIPTION OF SYMBOLS AND ALLT.EVIATION .
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LOT 201.29.HO.01 l PAGE 3 OF 28 1. Diesel Engine Driven Fire Pumps j The Fire Pumps are manufactured by Patterson Co. and are horizontal split casing single stage driven by a 285 horsepower Cnnmins diesel engine at 1170 rpm. These pumps are rated to deliver 2500 gpm at 125 psig. Each Fire Pump is located in a separate steel reinforced concrete room inside the fire pump house. The engine has a turbo charger, filter intake , silencer, jacket water cooling system, starter with an alternator that connects the starter to one of two batteries during cranking, an I alternator for recharging the.battaries and supplying control power when there is a loss of normal power, a governor with mechanical overspeed protection, and an electrical l overspeed protection tachomete The engine jacket water cooling system is , cooled by a heat exchanger supplied with raw l , ' cooling water from the discharge of the fire . pump via a pressure regulator and a l solenoid valve that opens or a pump start. The
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pressure regulator and solenoid valve can be bypassed by opening a normally closed manual ) valve. After passing through the heat exchanger the raw cooling water is routed to the drain syste Inside each pump room is the engine controller, dual starting battery banks, and a 550 gallon fuel tank. The fuel tank is filled from the Auxiliary Fuel Oil Storage and Transfer System via a normally closed manual fill valve. By design the fuel tank can be refilled within 8 hour . Fire Pump Controller The Fire Pump engine controller has a START pushbutton and a five position mode selector , i switch located behind a glass window in the controller cabinet door. Located on the door l front is a STOP/ RESET pushbutton. The l controller door is normally locked and entry I should only be made by obtaining authorized permission. The positions selectable by the l mode selector switch are T'ST, AUTO, MANUAId 7J - MANUAL B, and OFF. The function of these switch . positions are:
LOT 201.29.HO.01 PAGE 5 OF 28 manually vented after the fuel tank has been refilled.)
When the fire pump is running in the TEST mode, it can be manually stopped by placing the mode selector switch in the AUTO position and depressing the STOP/ RESET switch or placing the selector switch in the OFF positio If the latter is used, return the , switch to AUTO to place the engine in - standby readines B. AUTO - The green ready light illuminates on the controller and the engine will automatically start on: Low fire water header nressuro (This start signal.is set up to start the first engine after a short time delay and the second and third engines after subsequent delays to keep system pressure transients from causing inadvertent starts of the first engine and allow the header pressure to be recovered before starting the subsequent engines. The time delays are arranged for 5 seconds at 120 psig, 15'sleconds at 105 psig, and 25 seconds at 85 psig for starting Fire Pumps 11, 12, and 13 respectively.) Loss of AC Power to the engine controller. (The purpose of this start signal is to ensure Fire Pump availability as the batteries could self discharge between the time after loss of power and when needed. When the engine is running the engine alternator via rectifiers in the control panel feeds the CC control circuits and charges the batteries). . Remote start sional from either unit control roo (It is important to note that the Fire Pumps cannot be stopped from the control room.)
! e Weekly Test Jiper .'ir"I'T) on - em Automatically starts the engine once
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a week, runs it for 30 minutes, and l then stops i l
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, htle: Houston Lighting & Power als Customer P.O. No. 35-1197-6031 '
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Specification No. 9Q270MS021-E I rucisi jf" 1;f (5-3\ on.orr@
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m 5 I(C5 2 Controller - Fire Pumo 02 Y p2r se Eph '" Item No. 9-Q27-0-E-FCUO2-A 1 I ceAur@ luom! ~ E AII. A 12CR
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< . - -The Schematic Wiring Diagram Trouble Shooting Drawmg 6801 presents a complete schematic diagram that FIRST CHECK UIAT THE CONTROL IS PROPERLY SET '
UP TO OPERATE WIDI DIE ASSOCIATED ENGINE AND
.". illustrates a control equipped with all of the regularly cata-logued accessories. Most controls as shipped do not anclude all TO YlELD THE DESIRED OPERATIONS. READ THE ,
J GENERAL INFORMATION PARAGRAPHS AND CHECK
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A of these accesories, therefore, a separate schematic diagram is hf
made for each control to show that control as it is built. Each such diagram is identified by the senal number of the control THE SPECIFIC DRAWING FOR OPERATION SEQUENCE Throughout this booklet mention is made of the many I l that it represents and notations on the diagram show which of optional features that are available both in the origmal i the optional items are meluded and how the internal inte construction of the control and through a variety of inter- { connections are mad connections at the special terminal strip located on the rear ' The schematic wiring diagram is presented as a " ladder daa- side of the relay panel. Each controlis set up at the factory 1 l gram". The flow of cunent is from one " rail" of the ladde with the equipment called for in the purchase order, for use with a spectfic engine and to respond to selected startmg and
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5 ' via " rungs" of the ladder to the opposite " rail". As draw battery power comes in via Terminals Nos. 6 and 8, and is stopping signals, etc. Utis inforrnation is all shown on a conducted via diodes 3R and 4R,thence through a contact on schematac wiring diagram which is identified by the sertal ,
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the control switch to the upper rail, current then flows via number of the specific o.-trol. In any instance where the l ' relay coils and/or lamps to the lower rail, and back to the operation of the control is ..ther than what is desired, refe I " ence to this diagram may indicate that a change m intercon-hattenes via Terminal No.1 nections will be all that is required in order to obtain the For identification purposes each rung is numbered. These desired operating features. It may also be that the engin numbers appear below the lower rail and opposite each run monted accessory equipment differs from that for which the h Above the upper rail and opposite cach rung in which there is " e ntrol u set up. If this is the case,some rearrangement of the -
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a relay. coil, there are other numbers; m some cases as many as intercomenons of the internal terminals may be all that is four numbers appear. These are the rung numbers in which the needed to make the control operate with the engm " andmdual relay contacts wd! be found. For example, above Failure of Engine to Crank and Run When Signaled to
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k rung No. 4 m which the cost of relev 7Ctt is dnwn.rwmters, '
' mrt-d 6,73, L9,6.J appear. This codmg means that there is a contact Turn switch successively to " MAN. A" and to " MAN.
el of relay 7CR in each of these four rungs:that the contacts m r- rungs Nos. 6 and 73 are normally open contacts.and that the B". In each position press " START" burton and hold for four to six seconds. Observe whether engme spms l at contacts in rungs 59 and 61 are normally closed contacts, dl Underlining the number indicates a normally closed contac briskly since this is a good indication of battery co l x" > All contacts are drawn in the poution in which they rende dation. With the control selector switch in either of the te when the relay cod is de. energize "M AN." positions all of the relays are bypassed, so if the
* engine wdl not run, the problem is almost sure to be The relay numbers are arbitrarily chosen. Where the cucuit outade of the Control. Obvious!y.if the cranking is very
- ' I requires more contacts than are available in a single relay of
' weak the batteries are low and must be recharted or the type used, the additional relays carry the same number replaced. If cranking is brisk and aill the engine lads to es with a letter added.(Example: Relay 9CR,9A.CR and 98<R) run, the most likely place to look for trouble is m the to The abbreviation "CR" designates a conventional control relay fuel system. The engme manufacturer's instruction book d
that operates and releases without delay. "TR" deuenstes 3 should yield information in this regard.When the engme 'y j is to run, whether under automatic or manual control,
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timmg relay. subscript letters "SO" for " Slow Operate" mde cases that the time delay period before the contacts operate Termmal No.1 is energized. (Terminal No.12 also with starts when the relay is energized. Subscript "SR" for " Slow spark igmtson engmes). As a final check of the Control en * Release" indicates that the tame period commences when the test for voltage between Termina! Nos. I and 11 and 1 tes relay is de. energized. Relays that delay operation m both between Nos.12 and 11. With voltcge present here and ) of ducctions show "SO" and "SR". Alongude contacts of the the engme cranking briskly, the trouble IS NOT in the nY Ummg relays the notation -*TDC' indicates " Time Delay Contro Failure of Enfine to Stop Upon Cancellation of the
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Cloung" and "TDO" indicates " Time Delay Opemng". The
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age numbers, unless otherwise noted, m all cases sefer to the set Signal to Run: nd time m seconds. The ratchet relay is desienated "RR". This First be sure that the Control is set up for an automatic relay shtits both sets of its contacts each time its coil 85 stop. If it is so set up, the ugnal to stop comes via tirrung energized so it always rendes with one or the other set of its relay 4TR which may be adjusted for as long a time as at contacts close thirty mmutes, so check that 4TR has timed out. When is Wire identification numbers are used throuehout the diagram 4TR times out. its contact opens and thus breaks the
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and these same numbers are affixed to the wires in the contro circuit to 3CR to cause an immediate shutdown of a
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All points that are connected together bear the same number, diesel engme. A spark ignation engme fitted with a All points that connect to terminals (both for external and for solenoid valve in its fuel line will run until the fuel that is se5 downstream from this valve is used up.then wdl stop. lf intemal connection) use the number that the termmal use it is a spark ignition engine, be sure that this valve is Diodes are used in the circuit where power is to be permitted instaned. U there as no uncli valve. remove imemal mt , to flow in one duection only. When the two batteries are to connection between Termmals Nos. 33 and 34 sed ! feed a nngle load. as is the case with 3R and 4R where the CAUTION: IF THIS VALVE IS NOT INSTALLE power flows simultaneously from Terminals Nos. 6 and 8 and ENGINE SHOULD HAVE SOME OTHER METHOD OF is directed into circuit 38.two diodes are connected "back.t ACCOMPLISHING ANT 1-DIESELING. When relay back". rhe@s 1R 2R and 5R.6R are also examples of"back.t ~ back" connection. but m these two tnstances the power flow is 4TR ts used m the circuit there will befumpers runnmg from Termmal No. 49 to Termmal Sos. 51 and ~
- "in" from a single source and is divided to feed "out"to two Ternunal No. 49 connects to the contacts to the contact separate load of relay 4TR. When 4TR reaches its timed.out con.
" * The current flow is in the direction of the arrow head: .e.- dition, these contacts open. Thus. af a malfunction of E from poutive to negative. and Drawmg 6801 shows the diodes 4TR is suspected. temporarily remove the Jumper from d5 ;i as they are connected for common battery termmallNo. Il Terrninal No. 49 leavmg Si connected to 52. With , ell !. connected to the pontive pole of each battery (positive Termmal No. 49 disconnected. the engine should stop ge t ground).Negativeground systems have the diodes drawn with without
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their arrows pomtmg in the oppoute ducctio started st. delay upon cancellation of the signal tha
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J* v Itage at 53 ared I,8: and proceeding,you should find NOTE: BE SURE THAT ALL SIGNALS CONNECTED
$O START THE ENGINE ARE CLEA . fun mitage rt Terrrunal No. 70 on relay 9ACR and at 71 ;
on relay 6CR. With full voltage through to this poin *a 3, Persistent Crankmg of a Running Engine: turn the control switch to ** TEST". This should put If the engine-mounted device that senses th'e fact that voltage through to point 75 on 3ACR. Next check at - ' the engine is running fails to send its signal to the point 99 on relay 2TR.then point 35 which is a terminal j Control, or if the Control fails to properly acknowledge on the selector strip. Assuming that there is a jumper 35 this signal, then, even though the engine is runnmg. the to 36 voltage will also appear at points 36 and 74 on Control wdl engage the engme starter intermittently for relay SCR, but the voltage beyond point 75 wdl be a matter of about ninety seconds, then willlock itself present for only alternate ten second periods as relays out of service and burn its red " FAILED TO START" 1TR and 2TR alternate crank and rest periods. During lamp. Check to see that there is rated voltage from the ten second ON" intervals there should be voltage Terminal No. 2 to Termmal No.11 when the engine also on either Terminal No. 9 or Terminal No.10, runs. If there is no voltage here, then the trouble is in depending upon which side the ratchet relay happens to the engine-mounted equipment and the running signalis be residing, but NOT on both 9 and 10 at same tim not oeing sent. If there is voltage here, then relay 9ACR The foregoing discussion has dealt with the relays and is not operating properly, or an internal wire to the coil contacts that are directly in the circuit to the engine of 9ACR has become disconnected. Check for voltage on cranking motor controls. The circuit passes throuth the coil of relay 9ACR. Interchange 9ACR with 9BCR normally closed contacts of relays 9A CR, 6CR. 2TR, to check whether the relay itself is the source of the 13CR and 8CR. It passes also through relay 3A-CR but troubl in this case through normally open contacts, so 3A CR must be in its operated (picked up) position. Thus,in 4 Blue !.anip Goes Dark: the point * point check, if voltage is found at point 71 Each of the two blue lamps is associated with one of the of relay 3A CR but not at point 75, check first point 98 two storage batteries. If a lamp goes dark fassummg that n relay 3ACR since this point is a cod termmal and the lamp itself is not burned out). it means that its must have voltage present. Check next points 52 and 17 associated battery has been disconnected or that this n the selector switch. There must be voltage at all of battery is in bad condition or is badly in need of r chargmg. Turn the control switch to "OFF" and press the " RESET" button. Even a badly discharged battery '[ bl that the engine will crank and run when the control switch is turned to " TEST" but will not do so or one that is in bad condition will usually be able to up n operation of the pressure switch. This is because light the lamp. If the lamp burns when the " RESET" relays 3CR and 3A.CR are energized directly from the button is pushed and then goes dark when the button is test switch whereas the pressure switch directly operates released, the associated battery is useless. If the lamp relay ICR. which in turn operates relay 2CR. which in i remains bright after the button is released, turn the co tum e mpletes the circuit to relays 3CR and 3A.CR.By trol switch to the " MAN." position and try a manual bleeding off pressure in the line to the pressure switc start from the suspected battery. With a low battery the cranking wd! be sluggish and the blue lamp probably wi!! its contacts sho,,uld close. Now, with the control switch go dark and stay dark even after the start pushbutton is set at , AUTO. check for voltage at points 47. and 46 l{ , n relay ICR and at points 44 and 31 of relay 2CR and .- ' release via jumper (or jumpers) to point 48 and thence to 98 on , l NOTE: Even with a fu!!y charged battery the blue lamp relays 3CR and 3A C I
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will be somewhat dimmed while the engme is 6. Eneme Cranks Briskly But Fails to Run-bemg cranked. Read the paragraphs headed Assuming that engme is gettmg fuel, then in order for inoperative Battery Protection" for a more the Control to make it run, it is necessary only that complete discussio l Terrrunal No. I be energized. (Terminal No.12 also ifit i S . Engme Fails to be Cranked Upon Signal to Start: is a spark ignition engme with a solenoid valve in the fuel 1 Assummg that both battenes are connected and are at a line). With the selector switch set at " TEST". check at i good state of charge, that the Control selector switch is Terminal No. I for voltage. If voltage is present here, the ! l Control ts establishing the "run" circuit for a diesel set at " AUTO." and that the interconnections of inte nal ternunals is such that the control is intended to re engme and the ignition circuit for a spark ignition pond to the signal to start; a failure to commence engine. Where it is a spark igration engme with a sole- ; cranking upon r'eceipt of signal (or within 10 seconds noid fuel valve Terminal No.12 also must be energize l after s:gnal) is due to a bad connection or to a faulty If both Termmals I and 12 are energized. then the I
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relay. First determine whether the problem lies m the reason for fa !ure to run IS NOT in the Control but is Control or at the eneine. To do this. try cranking via the anciated with the engine. Terminal No. I is fed via a contact of relay 3CR and a selector switch contac start pushbutton with the control switch set at " MA Check for voltage at Terminals Nos. 98.53 and 34 on A" and " MAN. B" in turn. This will check out the i engme wiring and the engine. mounted equipment smce 3CR. then at 34 and I on the selector swstch to deter- i all of the relays in the Control are bypassed when the mme where the circuit discontinuity resides. If it is a switch is in either " MAN." position. If it is found that spark igniten engme. check that there are jumpers between Termm, als 12 and 31 and between 33 and 34 on cranking takes place and the engine runs as it is supposed . the internal connection strip and that there is voltage on to under manual operation. then there probably is an Terminals 53 and 31 of relay 3C "open" somewhere in the Control wiring or in one of the relavs. The cranking circuit is completed to external Engine Crankmg in Short Spurts: Termmals Nos. 9 and 10 attemately and interrruttently The en5me. mounted device that puts voltage on Te via several relay contacts. Before commenemg a detailed minal No. 2 must do so ONLY when the engme is fir ng checkout it is suggested ilat the hot cable be removed and is running above cranking speed. When this device from the starting motor on the engme so that actual puts voltage on Terminal No. 2 the control is signaled to crankine cannot occur. All other connections should stop crankmg. If voltage appears at Termmal No. 2
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remam m place. Use a voltmeter or a test lamp and BEFORE the engine is firing and runnmg. cranking will connect one side to Terminal No. II. Touch the power be discontmued too soon; then. as the eneme speed circuit breakers. then to points 38. 53 and 18 on the drops. its speed sensing device wdl open its circuit and selector switch. Retardless of selector switch positio the control will resume cranking. The remedy rests m vou should find fuli voltaFe at 6.8.66.68 and 38.With the engine-mounted speed sensing device. on' t m the the switch set at " AUTO." there should also be full Contro V/ __ __________
_ _ _ . I l RO EXAM: QUESTION #96 Given the following:
- A spent fuel assembly is being raised from its slot in the storage pool for return to the reacto Gas bubbles are coming to the surface of the poo Radiation levels in the Spent Fuel Pool area are increasin WHICH ONE (1) of the following actions is required of personnel in, the Fuel Handling Building (FHB)?
I Notify the control room to sound the Containment ! Evacuation Alar Immediately evacuate all personnel from the Fuel Handling Buildin Notify the Core Loading Supervisor, then replace the damaged fuel assembly in its storage locatio Move the fuel assembly into the RCB and notify the control room to initiate containment isolatio <
l I , KEY ANSWER: B
Response to Ouestion This question leads the examinee to believe that he is the Fuel Handling Machine Operator with refueling operations in progres In this position, he is required to inform the Core Loading Supervisor, Shift Supervisor, and HP of any impending radiological situation. Since the Fuel Handling Machine Operator is in constant communication with the Control Room, he would inform the Control Room of the conditions inside the FHB. The Control Room personnel , would, per the instructions of OPOP04-FH-0001, make an FHB j , evacuation announcemen (The Fuel Handling Machine Operator does
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I not have the capability to make the announcement).
l We believe that the safest place for the damaged fuel assembly to , I be stored would be back in the storage-rack location from which it was retrieve This reasoning is based on the following: ( The Spent Fuel Pool provides greater than 23 feet of borated water for radiation shielding and Iodine scrubbing.
, The storage rack is a structurally sound place to support l the damaged assembly without effecting other fuel l assemblies located in the poo . OPGP03-ZO-0002 " Qualifications and Conduct of Operators l l t l _
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(Question 096 continued)
for Cranes, Hoists, and Monorail Systems", Rev. 4; requires crane operators to land and release all loads prior to leaving any crane unattende . OPOP08-FH-0009 " Core Refueling" Rev. 8; requires operators to immediately notify the Core Loading Supervisor if fuel damage occurs during refueling operation . Based on this knowledge, we request that both "B" and "C" be considered as acceptable answer Ref: OPOP08-FH-0009, REV. 8 OPGP03-ZO-0002, REV. 4 OPGP03-ZA-0010, REV. 15 y b 0 l
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STP HOUSTON LIGHTING AND PO'w'ER COMPANY l
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SOUTH TEXAS PROJECF CONTROLLED COPY
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ELECTRIC CENERATING STATION , PIANT PROCEDURES MANUAL U NAY I 41992'
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SAFETY-REIATEDE - -
. Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 (General) l Page 1 of 5 '
APPROVED: A .l~o-97 f- / V -9 PIANT ERATION MANAGER DATE APPROVED DATE EFFECTIVE Purpose ' This procedure provides actions to be taken in the event a spent or new fuel assembly is dropped or damaged during fuel handling operation .0 Symptoms and Entry Conditions
, Damage to a spent fuel assembly accompanied by alarms (RM-11 or RM-23)
on radiation monitors located in the vicinity of the damaged assembly: : 2. Containment Building: 2.1.1 1 Containment Atmosphere Monitor; RT-80ll 2.1. Containment Purge Exhaust Monitors; RT-8012 and RT-8013 2.1. RCB 68 ft Area Monitors; RT-8099 and RT-8055 2. Fuel Handling Building 2.1. FHB HVAC Exhaust Monitors: RT-8035 and RT-8036 2.1. FHB Main Floor Area Monitors: RT-8081, RT-8089, RT-8090, RT-8091, and RT-8097 Cas bubbles originating from the damaged assembl .3 Visual observation of a fuel handling accident (e.g., dropped fuel assembly).
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t Fuel Handline Accident OPOPO4-FH-0001 Rev 2 i ' Page 2 of 5 E l i o High Radiation on RT-8012 or RT-8013 will - I actuate Contairunent Ventilation isolatio l
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o High Radiation on RT-8035 or RT-8036 vill l actuate FHB HVAC in the Emergency Mod { o The Shift Supervisor is responsible for _ overall performance of this procedure.- The ! Shift Supervisor MAY delegate applicable i i responsibilities to the Core h ading Supervisor as appropriat I I
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3.0 In:cediate Actions None Subsecuent Actions H a spent fuel handling accident has occurred in RCB, E EE DIRECT Control Room to sound the Containment Evacuation Alar .2 H the spent fuel handling accident has occurred in the FHB. EXE EVACUATE the FH .3 E a new fuel accident has occurred, THEN EVACUATE the area surrounding the damaged fuel assembl .4 NOTIFY Health Physics and Reactor Engineering of the fuel handling l acciden .5 PIFER to OERP01-ZV-IN01, Emergency Classification, for potential ! Emergency Plan implementatio .6 E a spent fuel handling accident has occurred in the RCB, EEE PERFORM the following: , i 4. ENSUPI Containment Normal Purge Supply and Exhaust fans secured: o "SPLY FAN llA(21A)" *EXH FAN llA(21A)" o "SPLY FAN llB(21B)" "EXH FAN llB(21B)"
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Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 Page 3 of 5 4. ENSURE Supplementary Purge Supply and Exhaust fans secured: o "SPLY FAN 11A(21A)" "EXH FAN 11A(21A)" o "SPLY FAN 11B(21B)" "EXH FAN 11B(21B)" - 4. ENSURE Normal Purge Isolation Dampers closed: o "SPLY OCIV MOV-0007" o "SPLY ICIV MOV-0008" o "EXH ICIV MOV-0009" o "EXH OCIV MOV-0010' I 4. ENSURE Supplementary Purge Isolation Dampers closed: o "SPLY OCIV FV-9776" ;
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o "SPLY ICIV MOV-0003" o "EXH ICIV MOV-0005" I o "EXH OCIV FV-9777" 4. If an automatic Containment Ventilation Isolation occurs, THEN SECURE sample pump f RT-80l I 4. IF radiation levels OR sampling results indicate that spent ,
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fuel assembly cladding has been breached IgEN PERFORM the following: 4.6. INITIATE recirculation through the Emergency Containment Charcoal Filters to reduce the l concentration of I-131 per 1(2) POP 02-HC-0001, Containment HVAC.
- , , , , 4.6. UHEN the following conditions are met, IEEE OBTAIN a l purge permit and COMMENCE a Normal Containment Purge per 1(2) POP 02-HC-0002, Normal Containment Purge
o Health Physics concurs that a purge is necessar o Sampling results indicate the I-131 concentration has been reduced to its non-occupational MP o Further recirculation of the containment atmosphere through the Emergency Containment Charcoal Filters will NOT substantially reduce the concentraion of existing-radionuclides.
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I Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 ,
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l JF a spent fuel handling accident has occurred in. the FHB, THEN ! PERFORM the following: , 4. ENSURE FHB HVAC operating in emergency mod ' N 4. ESTABLISH ONLY two trains nib exhaust fans (main and booster) i operatin [ e PIACE one non-operating train HIB exhaust fans in FULL TO IhCK. p
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4. { SECURE one HIB filter train by manually closing the outlet ! 4. Q dampe Q ,
; REQUEST assistance from Reactor Engineering to determine a plan of l action for inspecting the damaged fuel assembl ; WHEN Health Physics has verified that radiological conditions are ;
acceptable,1 HEN the Shift Supervisor may allow personnel to return to ! the accident area to implement recovery action l
4.10 WHEN approval has been obtained from the Shift Supervisor 6LTD_ Reactor Engineering Supervisor, THEN RESUME fuel movemen !
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5.0 References { i UFSAR Chapter 9.1- l Technical Specifications: . ! 5. T.S. 3.9.9, Containment Ventilation isolation System l 5. T.S. 3.9.12 FHB HVAC j OERP01-ZV-IN01, Emergency Classification ;
! (2) POP 02-HC-0001, Containment HVAC l l (2) POP 02-HC-0002, Normal Containment Purge j (2) POP 02-HC-0003, Supplementary Containment Purge SV129V00012, HVAC FHB Supply System l
> V129V00013. HVAC FHB Exhaust System Logic Diagrams 5.9 I 9Z41906, RCB Purge Isolation RMS 5. Z42113. Instrumentation Actuation Trains A, B, & C 5. , FEB HVAC Main Supply Fans 5. Z41601, FHB HVAC Main Exhaust Fans 5. Z41602, THB HVAC Exhaust Booster Fans 5. Z42125, FHB HVAC Emergency Operation 5. Z41617, FHB HVAC Exhaust Air Bypass Line Damper 5. Z41903, Spent Fuel Fool Exh Radiation Monitors 5. Z41608, FHB Exhaust Filter Outlet Damper
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Core Refueline OPOP08-FH-0009 , Rev. 8 !
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4.17 VHEN the cause of a valid high flux at shutdown alarm is identified AND the Core loading Supervisor deems that conditions are. safe for all personnel. THEN Core Alterations may continu .18 IE there is a malfunction E suspicion of malfunction of any fuel ; handling equipment. THEN all fuel movement SHALL be terminated M " the Core leading Supervisor SHALL be notifie ,
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j 4.19 Handling tools SHALL NOT be left attached to fuel during planned ] work stoppage , 4.20 VHEN moving fuel assemblies between the fuel transfer canal and the , spent fuel pool, THEN special care SHALL be used to ensure that' the ;
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fuel assembly does not make contact with the transfer canal gate wall i 4.21 IE, during the course of refueling, fuel damage occurs, Tli]Qi all l refueling operations SHALL be immediately stopped M the Core l Loading Supervisor notifie This step does not apply to d===ge identified during core offload visual check > ) 4.22 lE an accident involving fuel assemblies occurs , . THEN GO TO OPOP04-FH-0001, " Fuel Handling Accident". ; I " i 4.23 The Refueling Machine Operator SHALL NOT lower a fuel assembly into
the reactor vessel until the ICRR data from loading the previous fuel assembly has been evaluate .24 The Refueling Machine Operator SHALL NOT disengage from any assembly 3 in the core until nuclear instrumentation indicates safe conditions j AND approval to disengage is received from the Core Loading Supervisor.
- 4.25 VHEN irradiated fuel is being transported through the fuel transfer tube, THEN high radiation conditions may exist in the FHB penetration space in the vicinity of the fuel transfer tube. Access d to this area SHALL be controlled per OPRP07-ZR-0009-04, " Guideline S
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For Spent Fuel or 1rradiated Material Transfer".
0- . 4.26 Irradiated fuel assemblies freshly unloaded from the reactor vessel SHALL be stored in the Region I racks in the Spent Fuel Pool for at least 30 days prior to being transferred to a Region II rack (Reference 6.15).
4.27 The following SHALL be performed to prevent tools, equipment, .and i other foreign material from falling into the reactor vessel or refueling cavity:
4.2 All handling tools SHALL be thoroughly cleaned before insertion into the reactor vessel.
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l Core Refueline OPOP08-FH-0009 ; Rev. 8 : Page 3 of 30
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a The Health Physics Manager, or designee, SHALL be responsible for g i l providing appropriate radiological coverage during refuelin , 2, l e 1 3.0 Prerecuisites ! I Prerequisite Checklist (-1) is completed prior to core offload'OR ' core reloa ' I 4.0 Ep_t_es and Precautions
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) All Core Alterations SHALL be observed and directly supervised by i either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation (T.S. 6.2.2.d). This person is hereafter referred to as the Core Loading Superviso ,
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l l Where allowed in the procedure, the Core Loading Supervisor may delegate duties, provided that the designee meets the same ; qualifications as the Core Loading Superviso ! The Core Loading Supervisor may verbally authorize a Reactor 1 Operator or Reactor Engineer to inicial/date/ time Fuel Transfer - { Forms (FTF) for hi . JI any of the following conditions occur. THEN Core Alteration SHALL be suspended pending investigation by. the Core . Loading Supervisor: 4. An unanticipated increase in the neutron count rate, by a factor of two occurs on both source range NI channels during any single refueling ste . An unanticipated increase in the neutron count rate, by a factor of five occurs on either source range NI channel during any single refueling ste . There is an unexplained decrease in boron concentration greater than or equal to 25 ppm between two successive samples in either the operating RHR loop (-2) or the refueling canal (-3), i 4. The operating RHR loop shows an unexpected temperature change of more than 10 *F between two successive readings l (-2).
4. One or both of the plant source range NI channels are inoperabl . Communications be tween . the Control Room Operator and the personnel in either the WB or RGB are los i I l l _ _~ - _ _ , - . _ , _ . . . - _ . . . . . _ . , , . _ _
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PREREOUISITE CHECKLIST ) OPOP08-FH-0009-1 INITIALS l (Page 4 of 5) l l
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Refueling from Cycle- to Cycle Unit )
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2 ENSURE within 2 hours of the beginning of movement of fuel or control rods that the refueling cavity is filled to a level between 63 ft 0 in and 66 ft 6 in. The lower limit will ensure that Technical Specification 3/4.9.10 is met ;
(Reference 6.32). The upper limit will preclude water '
draining down into the HVAC duct exhaust (Reference 6.33) . Last Verified Date: Time:
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23.0 ENSURE 1(2) PSP 03-EA-0002, "ESF Power Availability" has been performed within 8 hours prior to the commencement of core ' alteration Last Performed Date: Time: 1(2) PSP 03-XC-0001, " Refueling Containment 24.0 ENSURE Penetration Checklist" has been performed within 100 hours prior to the commencement of core alteration Last Performed Date: Time: , 25.0 ENSURE an Analog Channel Operational Test has been . performed on both plant source range NI channels (N-31 and N-32) eithe l l 25.1 Vithin 8 hours prior to commencing core alteration OR OR 25.2 At least once per 7 days since the ACOTs were performed for the initial core alterations ( head lift). Reference Technical Specification 3/4. Time: Last Performed (N-31) Date:
(N-32) Date: Time:
26.0 ENSURE that OPSP03-FH-0001, " Refueling Machine Pull Test" has been performed within 100 hours prior to using the refueling machine or hoist for the movement of thimble l I plugs, drive rods, or fuel assemblies within the reactor l vessel (Technical Specification 3/4.9.6).
Last Performed Date: Time: 27.0 ENSURE direct communicatioos has been established between the control room operatn and personnel at the refueling l stations in the RCB and FHB within 1 hour prior to commencing core alterations (Technical Specification 3/4.9.5). Time: Last Performed Date: _ ! l' l l . . _ _ _ . - ..
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Oualifications and Conduct of Operators for ! Cranes. Hoists, and Monorail Systems OPGP03-ZO-0002 Re {
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Page 11 of 17 ! i NOTE l t Signals for crane operation should be given per OPGP03-ZI-0026 (Gener! voice communication (telephone, radio, etc.) is used. Signals should be visi at all trmes. Special abnormal signals should be clearly agreed upon by thei crane operator and should not conflict with standard signal ! l 5.2.2.14 -
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Activate installed warning devices prior to crane transit and intermittently when approaching personnel l
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5.2.2.15 Follow only approved safe load paths per OPGP03-ZA-0069 (Control of Heavy Loads).
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Land and release attached loads, place the crane controllers in *O and set the parking brake (if applicable) before leaving any crane imanendea 5.2.2.17 Place all crane controllers in "OFF' following a loss of power. Ve crane operation upon power restoratio .2.3 Conduct - Gantry Cranes 5.2. The Operator shall walk the rails prior to initial lifts to ensure the rai and path are clear and to become aware of other wor.k activitie the crane's path. Examples of initial lifts include the4 first lift by a particular Operator using a particular gantry crane, first lift after returning from break or lunch, et 'O 5.2. i The Operator may require additional personnel to M walk the r A crane travel to ensure the rails and path are clear. The Signalman shall employ such personnel as required by the Operato e i personnel are then responsible for idendf Such A notifying the Signalman in a timely manner. ying interferences and Such personnel are not required to be operator or rigger certifie ' 5.2.4 Conduct - Reactor Containment Building Polar Cranes i 5.2. Warm up the crane controller circuitry (power breakers closed) for minutes before operating any control mechanisms. (SER 85-02) 5.2.4.2 j Park the crane along a north-south axis when not in use. (Modes 1-4 only) 5.2.5 Conduct - Outdoor Cranes (including mobile cranes) 5.2. Tie down or otherwise secure all outdoor cranes when no
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i J k OPGP03-ZA-0010 Plant Procedure Adherence
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and Implementation and Rev 15 Independent Verification Page 4 of 22 3.0 Responsibilities The Plant Manager is responsible for establishing and maintaining the station standards and expectations for procedure adherence and implementatio , The Department Manager assigned responsibility for the preparation of each procedure is responsible for: 3. Determining the procedure use control required as described in steps 4.2.5 through 4.2.1 . Ensuring that provisions for verification are established within applicable procedures which require independent verificatio .3 All Supervisors assigned responsibility for performing a procedure regt' iring independent verification are responsible for ensuring that only knowledgeable individuals are assigned to perform independent verifications and that independent verifications are performed in accordance with this procedur y 4.0 Procedure , Adherence to Procedures
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I 4. Procedures SHALL be strictly adhered to when performing plant activities. For implementation of new or revised procedures, the Plant Manager may allow deviations (such as early use of new forms, continued use of superseded forms, etc.). Permission for such deviations shall be prescribed in writin _ _ _ _ - - _ - - - _ _ _ _ _ _
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OPGP03-ZA-0010 Plant Procedure Adherence 1 () and Implementation and Independent Verification Rev 15 Page 5 of 22 t i i 4. Anyone performing a ,t,ask under the direction of an STP Procedure (Engineers, Operators, Technicians, Craftsmen, or Anyone) SHALL perform the steps of that procedure as written, unless such performance vould violate the intent of the procedure as discussed in step 4. . l 4. When performance of a step in a procedure: a) Vill, or it is believed may, violate the intent of the procedure; or b) Could result in the plant / system being placed in a condition that is not consistent with good Maintenance, Engineering;-or-Operating practices; or c) Might result in a personnel or equipment hazard; > SIDE AMD Inforza supervision, management, and/or the control
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r.o,ce I When a procedure cannot be performed.as written the
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f) 4. procedure SHALL be changed via Field in an Change or emergency Revision (see
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prior to performance, exceptObvious typographical errors need not be step 4.1.9).
- corrected as long as the meaning of the erroneous entry is I
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clear and the error does not affect component numbers, acceptance criteria values, or other information where single character deviation could affect result special 4. Night orders, standing orders, operating orders, orders, letters, memoranda, etc. SHALL NOT be used to amend, revise, or delete an approved procedure, nor shall These they be used to violate an approved procedur documents shall only be used to give supplementary directions or orders relating to the conduct of operations (e.g. reduce power at 1200 hours in accordance with approved procedures and Technical Specifications).
4. Changes, revisions, and deletions to procedures and temporary procedures SHALL be made in accordance with OPGP03-ZA-0002 (Plant Procedures).
4. For purposes of determining adherence, the terms _used to distinguish a requirement f rom a recommendation or a i granting of permission are explained in Addendum /
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! j ) i ! 1 RO EKAM: QUESTION NO. 25 j } SRO EKAM: QUESTION NO. 23 .j
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! - Unit 1 has been operating at 80% for the last 48 hours, j ! - No AFD penalty time has been accumulate i
- - The AFD monitor alarm was disabled at 060 l
"j - Target AFD for 100% power is + 3.75%. !
- Indicated AFD has been continually monitored and logged as, I follows: j l
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- TIME N41 N42 N43 N44 l j 0600 + + + + l
- 0622 + + + + + + + + + + + + + + + + l
1
At 0700, WHICH ONE (1) of the following is the cumulative penalty j j deviation time for the previous hour? i l minute b. 21 minute minutes.
4 minutes.
! i j KEY ANSWER: B i h } Response to Ouestion ! We feel that the most conservative response to this question is 26 , minutes. This is based on the following facts:
- (a) The STP Core Operating Limits Report (COLR) for Unit _1, d
Cycle 5, states that the target band for AFD must be ! within +3% or -12% of the target value. Calculating 80% q ~ of the target value (3.75%) at 100% power gives a target value of 3.0%. Therefore, the maximum target band limit
- is 6.0% with reactor power at 80%. The STP Technical j Specifications (3/4 2.1, Surveillance Requirement
- 4.2.1.2.a) states that one (1) minute of penalty l 1 deviation shall be accrued for each minute that AFD is i i outside the target band when rated thermal power is > '
j 50%. Technical Specifications also states that when the AFD Monitor Alarm is inoperable, AFD must be logged once i i per hour for the first twenty-four (24) hours and at j least once per thirty (30) minutes thereafte This j Surveillance Requirement (4.2.1.1.a) continues to state i i
. _ _ _ . _ . . -. _ _ . _ _
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that the logged values of the indicated AFD shall be assumed to exist durina the interval crecedina each locain (b) Based on the statement made in the Surveillance Requirement 4.2.1.1.a, we must assume that AFD exceeded the target band limit of 6.0% at 0634, even though the log was recorded at 065 This "out-of-target-band" reading must be considered valid with the accumulation of penalty points until the next AFD reading is recorde with less than 2/4 channels reading outside the target , ban At 0700, only one (1) channel is indicating greater than the maximum allowed target band value of l 6.0% and the accumulation of penalty points is stoppe l Therefore, the maximum amount of time that AFD is outside i the target band is 26 minute l
Based on this knowledge, we request that the answer for Question ]
#25 be changed to "C". j i si <
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- -- .. i 3/4.2 POWER DISTRIBUTION LIMITS TSI -002 , 3 /4. 2.1 AXIAL FLUX DIFFERENCE LTHITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (COLR). - I APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* ACTION: With the indicated AFD outside of the above required target band, and with THERMAL POWER: greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either: a) Restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THEPRAL POWE . greater than or equal to 50%, but less than 90% of RATED l THERMAL POWER: a) POWER OPERATION may continue provided: 1) The indicated AFD has not been outsida of the target band for more than I hour cumulative penalty deviation during the previous 24 hours, and 2) The indicated AFD is within the Acceptable Operation limits specified in the COL Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux * - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hour b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1., provided that the indicated AFD is maintained within the Acceptable Operation Limits specified in the COLR. A total of 16 hours operation may be accumulated with the AFD outside of the Target Band during this testing without penalty deviatio *5ee Special Test Exceptions Specification 3.1 SOUTH TEXAS - UNITS I & 2 3/4 2-1 Unit 1 - Amendment No. 9, 27, 39 Unit 2 - Amendment No. I, 77, 30
- -. .- . - -- -- . - - - - . .- - . - -_
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) greater than 15%, but less than 50% of PATED THERMAL POWER: THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the target band for more than I hour cumulative penalty deviation during the previous 24 hour THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the target band, and the indicated AFD has not been outside of the target band for more than I hour cumulative penalty deviation during the previous 24 hour SURVEILLANCE RE0VIREMENTS 4.2. The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THEPyAL POWER by: i Monitoring the indicated AFD for each OPERABLE excore channel,at.
l least once per 7 days when the AFD Monitor Alarm is OPERABL Monitoring and logging the indicated AFD for each OPERABLE excore channel at least .once per hour for the first.24 hours, and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each loggin .2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of: One minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and One-half minute penalty deviation for each I minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWE , 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective full' Power Days. The
- provisions of Specification 4.0.4 are not applicabl .2.1.4 The target flux difference shall be updated at least once per 31 Effective full Power Days by either determining the target flux difference i
, ' SOUTH TEXAS - UNITS 1 & 2 3/4 2-2 Unit 1 - Amendment No. TI, 39 Unit 2 - Amendment No.17, 30
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, t i SURVEltLANCE REOUIREMENTS (Continued) i
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pursuant to Specification 4.2.1.3 above or by linear interpolation between the I
- most recently measured value and the predicted value at the end of the cycle life. The provisions of Specification 4.0.4 are not applicabl ,
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" UNIT 1 CYCLE S I l
- CORE OPERATING LIhuTS REPORT J l
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1 1 CORE OPERATING LIMITS REPORT l l l This Core Operating Limits Report for STPEGS Unit i Oycle 5 has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been - developed using the NRC approved methodologies specified in Technica! Specification 6.9. l
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~Ibc Technical Specifications affected by this report are: l
,
!
1) 3/4.L13 Moderator Temperature CoeHicient Limits 2) 3/4.1 Shutdown Rod Insertion Limit ! l 3) 3/4.1 Control Rod Insertion Limits ' 4) 3/4. AFD Ilmits
.
5) 3/4. Heat Flux Hot Channel Factor 6) 3/4.23 Nudear Enthalpy Rise Hot Channel Factor !
-
2 OPERATTNG LTMTIS
~Ibe cycle-specific parameter limits for the specifications listed in Section 1.0 are presented belo .1 MODERATOR TEMPERA ~RIRE COEFFICIE?TT (Specification 3.1.13)
i 2.1.1 The BOI, ARO, MTC shall be less positive than the limits shown in Figure l l i 2.L2 The EOI, ARO, HFP, MTC shall be less negative than -4.8 x 10 Ak/k/ *F.
I 2.13 "Ibe 300 ppm, ARO, HFP, MTCshall be less negath e than -3.9 x 10" Ak/k/ *"(300 ppm Surveillance Limit).
Where- BOL stands for Beginning of Cycle Life EOL stands for End of Cycle Life ARO stands for All Rods Out ; HFP stands for Hot Full Power (100% RATED THERMAL POWER)
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South Texas Unit 1 Cycle 5 ROD INSERBON LIMFTS (Specification 3.13.5 and 3.13.6) 2.2.1 The Control Rod Insertion limits are provided in Figure .2.2 Fully withdrawn for all Control and Shutdown Banks shall be 250 steps and above, but
-
j not e-vrWing 259 steps withdraw .2.3 All banb shall have the same Full Out Position (FOP).
23 AXIAL FLUX DIFFERENCE (Speci5 cation 3.2.1) i l 23.1 AFD limits as required by Technical Speci5 cation 3.2.1 are determined by CAOC l Operations with an AFD target band of +3,-12E l
.
23.2 The AFD shall be maintained within the ACLerTABLE OPERATION portion of Figure 3, as required by Technical Speci5 cation ~ ! l 2.4 HEAT FLUX HOT CHANNEL FACIDR (Speci5 cation 3.22)
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l 2.4.2 K(Z) is provided in Figure .43 'ihe F ylimits for RATED THERMAL POWER (F"") y within specific core planes shall be: i 2.4 less than or equal to 1.90 for all core planes containing Bank *D' ! l l control rods, and i 2.4 less than or equal to the appropriate core height. dependent value from Table I for all unrodded core plane .433 PFy= _ _ . .
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South Texas Unit 1 Cycle 5 ,
i These Fylimits were used to confirm that the heat flux hot channel factor Fo(z) will be limited by Technical Specification 3.2.2 assuming the most limiting axial power i distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying variations in the axial xenon and power !
* l distributions, as described in WCAP 8385. Therefore, these Fy limits provide assurance l that the initial condidons assumed iri the LOCA analysis are met, along udth the ECCS
-
acceptance criteria of 10CFR50.4 I
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2.5 EMTIALPY RISE HOT CHANNEL FACTOR (Specification 3.23) ; , 2.5.1 FjF=1.4 , 2.5.2 PF3n=0 l 3 Rbt ERENCES ' 3.1 GarrespondenceSerialNumberST-UB-HI 01176,LetterfromRobertC.Cobb(Westinghouse) to Dave Hoppes (HL&P), Core Operating Limits Repott, Unit 1, Cycle 5 11-06-9 ~ l 3.2 NUREG-1346, Technical Specifications, South Texas Project Unit Nos. I and l l
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South Texas Unit 1 Cycle 5 FIGURE 1 MTC versus Power Level
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(23.25s):122 step o<edap g9.259): 122 step o<ertap !
(23.2se):12t a p ov.dag ps 2ss):121 st p o% )
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(22,256):119 sep oveden p8.256): 11s st.p o<.rtap 260 , , (2t M : 117 Cap O<* tap ,
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0 ' I O 10 20 30 40 50 60 70 80 90 100 RELATIVE POVEfl (%)
' Control Bank A is aircady withdrawn to I ull Out Positio South Texas Unit I Cycle 5 -
RGURE3 ! AFD Umits versus Rated "Ihermal Power 120 i i 100 (-11.90) (11,90) ,,
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-50 -40 -30 -20 -10 0 10 20 30 40 50 Flux Differerice (61) %
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i l South Tcxas Unit 1 Cycle 5 FIGURE 4 K(Z) - Normalized Fn(Z) versus Core Height .
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South Texas Unit 1 Cycle 5 TABLE 1 Unrodded Fry's vs. Core Height Core Height Unrodded Core Height Unrodded (Ft.) Fxy (Ft.) Fxy
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0.000 2.993 7.200 1.841 0.200 2.542 7.400 1.834 0.400 2.091 7.600 1.829 0.600 L869 7.800 L825 OJ00 1.734 8.000 L824 ' LOOO 1.663 8.200 1.824 L200 L640 8.400 1.826 L400 L633 8.600 L830 L600 L638 8300 1.836 L800 L656 9.000 L843 2.000 L678 9.200 1.849 2.200 L702 9.400 IE61 2.400 1.729 9.600 L874 2.600 L756 9.800 L888 2300 ,L783 10.000 L899 3.000 L809 10.200 1.907 3.200 1E34 10.400 L912
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3.400 - L852 10.600 L921 3.600 L846 10.800 L926 . 3.800 L837 ILOOO L926 4.00 .834 IL200 1.920 '
. 4.200 1.846 11.400 1.907 4.400 1E42 11.600 1.890 4.600 L836 11.800 1E66 4.800 1E30 12.000 IE34 5.000 1E27 12200 1.806 5.200 1E25 12.400 L788 5.400 1.823 12.600 1.812 5.600 1E25 12.800 1354 SE00 IE27 13.000 1.913 6.000 1E31 13.200 1.906 6.200 1.837 13.400 1.%9 6.400 IE44 13.600 2.112 6.600 1E52 13.800 2.472 6E00 1.855 14.000 2E33 7.000 1.850 Core Operatine Umiis Hemn n. . n r n L---
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APPENDIX A U.S. NUCLEAR REGULATORY COMISSION
REGION IV
Inspection Report: 50-498/93-34 50-493/93-34 Licenses: NPF-76 NPF-80 , Licensee: Houston Lighting & Power Company P.O. Box 1700 liouston, Texas ; l , ' Facility Name: South Texas Project Electric Generating Station, Units I and 2 l
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Inspection At: South Texas Project, Bay City, Texas Inspection Conducted: September 24 through October 1, 1993 j l Inspectors: Ryan E. Lantz, Chief Examiner, Operations Section Division of Reactor Safety Mark A. Satorius, Project Engineer, Project Section A Division of Reactor Projects Accompanying Personnel: Art Lopez, Examiner, Contractor : Battelle Pacific NW Labs i Nancy Maguire-Moffitt, Examiner, Contractor Battelle Pacific NW Labs
Approved: (Q)Am fy)D4 W 10
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2.h C6~ l John L. Pellet, Ch'ief, Operations Section Date ' l Division of Reactor Safety Inspection Summary l Areas Inspected (Units 1 and 21: Routine, announced inspection of the qualifications of applicants for operator licenses at the South Texas Project facility, which included an eligibility determination and administration of comprehensive written and operating examinations. The examination team also , observed the performance of on-shift operators and plant conditions incident i I to the' conduct of the applicant evaluations. The examiners used the guidance provided in NUREG-1021, " Operator Licensing Examiner Standards," Revision 7, Sections 201, 202, 203, 301, 302, 303, 401, 402, and 403, issued January 199 ' l i ,
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Attachments:
- Attachment 1 - Persons Contacted and Exit Meeting ,
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* Attachment 2 - Simulation Facility Report
* Attachment 3 - Written Examination Keys
* Attachment 4 - Facility Post-Examination Review Comments i I
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_4 DETAILS ) ' l 1 LICENSED OPERATOR APPLICANT IMITIAL QUALIFICATION EVALUATION (NUREG-1021) During the inspection, the examiners evaluated the qualificaticns of 15 license applicants; 6 for reactor operator (RO), 5 for senior reactor ! operator (SRO) currently licensed as R0s, and 4 for instant SRO. The , inspection assessed the eligibility and administrative and technical competency of the applicants to be issued licenses to operate and direct the operation of the reactivity controls of a commercial nuclear power facility in a
REGION IV== Inspection Report: 50-498/93-34 50-499/93-34 Licenses: NPF-76 NPF-80 , Licensee: Houston Lighting & Power Company P.O. Box 1700 Houston, Texas ; Facility Name: South Texas Project Electric Generating Station, Units 1 and 2
!
Inspection At: South Texas Project, Bay City, Texas Inspection Conducted: September 24 through October 1, 1993 l Inspecters: Ryan E. Lantz, Chief Examiner, Operations Section Division of Reactor Safety l Mark A. Satorius, Project Engineer, Project Section A i Division of Reactor Projects l Accompanying Personnel: Art Lopez, Examiner, Contractor Battelle Pacific NW Labs ;
Nancy Maguire-Moffitt, Examiner, Contractor l Battelle Pacific NW Labs '
) D l
l Approved: Qfp l0 2.{f C3 John L. Pellet, Ch'ief, Operations Section
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Date ' Division of Reactor Safety , Inspection Summarv l l Areas Inspected (Units 1 and 21: Routine, announced inspection of the ! qualifications of applicants for ope, ator licenses at the South Texas Project facility, which included an eligibility determination and administration of comprehensive written and operating examinations. The examination team also observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluations. The examiners used the guidance provided in NUREG-1021, " Operator Licensing Examiner Standards," Revision 7, Sections 201, 202, 203, 301, 302, 303, 401, 402, and 403, issued January 199 ! i
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2 i i : j Results (Units 1 and 2): ! l : ' * Four of the six applicants for reactor operator licenses satisfied the .l ! requirements of 10 CFR 55.33(a)(2) (Section 1.6).
- Eight of the nine applicants for senior reactor operator licenses satisfied the requirements of 10 CFR 55.33(a)(2) .(Section 1.6).
>
* The reference material provided by the training department for *
, examination development was adequate (Section 1.1). ! i : , * All applicants passed the written examinations, with scores ranging from ! a low of 82 percent to a high of 94 percent with averages of 86 percent for reactor operator applicants, 90 percent for senior reactor operator : applicants, and 88.4 percent overall (Section 1.2).
'
, * The crews examined exhibited generally effective, formal communications, '
with effective command and control on the part of crew supervision, with l noted exceptions (Section 1.3.1).
) * The applicants demonstrated a generic performance weakness which i
! involved a hesitancy to secure equipment when abnormal conditions were noted immediately following equipment startup.
, The applicants demonstrated a second generic performance weakness which !
- involved a general unfamiliarity with low power and shutdown procedures ,
(Sections 1.3.1, 1.3.2).
!
* Procedural guidance for loss of primary reactor coolant accident ;
scenarios while shutdown was unclea ; Procedural guidance for abnormal response of a reactor coolant pump when ; starting was lacking (Sections 1.3.1,1.3.2). !
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* Poor plant labeling was observed to adversely impact operator performance and was consistent with prior NRC inspection reports
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(Section 1.3.2).
- General observations were made of on-shift control room operators and plant material conditions (Section 1.4). )
Summary of Inspection Findinos: "
* There were no findings that were assigned a tracking number identified
- during the course of this inspectio ,
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Attachments:
* Attachment 1 - Persons Contacted and Exit Meeting
* Attachment 2 - Simulation Facility Report
* Attachment 3 - Written Examination Keys
* Attachment 4 - Facility Post-Examination Review Comments
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-4-DETAILS
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1 LICENSED OPERATOR APPLICANT INITIAL QUALIFICATION EVALUATION (N'JREG-1021) i t During the inspection, the examiners evaluated the qualifications of 15 license applicants; 6 for reactor operator (RO), 5 for senior reactor 4 operator (SRO) currently licensed as R0s, and 4 for instant SRO. The , inspection assessed the eligibility and administrative and technical competency of the applicants to be issued licenses to operate and direct the operation of the reactivity controls of a commercial nuclear power facility in accordance with 10 CFR 55 and NUREG-1021, " Operator License Examiner ! Standards," Revision 7, Sections 200 (series), 300 (series), and 400 (series).
' Further, the inspection included evaluations of facility materials, , l procedures, and simulation capability used to support development and administration of the examinations. These areas were evaluated using the guidance provided in the areas of NUREG-1021 cited above. Additionally, the examination team also observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluation After completion of the evaluations, the examiners determined that four of the six applicants for R0 licenses and eight of the nine applicants for SRO licenses satisfied the requirements of 10 CFR 55.33(a)(2) and will be issued the appropriate licenses. Three applicants, two R0 and one SRO, failed the dynamic simulator portion of their operating examinations and their license applications were denied. Those applicants whose application for a license was denied may reapply in accordance with 10 CFR 55.3 Performance results for individual applicants are not included in this report > because inspection reports are placed in the NRC Public Document Room as a matter of cours Individual performance results are not subject to public disclosur .1 Facility Materials Submitted for Examination Development l The chief examiner reviewed the licensee's materials provided for development i of the examination, which included station administrative and operating procedures, lesson plans, question banks, simulator scenarios, and job l performance measures (JPM). The procedures and lesson plans were adequat i Some JPMs were not current with the latest procedure revision in the initial submitta The facility bank of written questions, dynamic simulator scenarios, and JPMs was adequate in scope, depth, and variety. It was used extensively in , l developing the examination There is no regulatory requirement for a facility to develop and maintain a bank of valid test items (questions, JPMs, and scenarios) for NRC use to I develop examinations. However, because of the significant savings in l development time, the NRC has expressed willingness to use such material if it is available and meets the standards of NUREG-102 l
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1.2 Written Examinations !
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The examination team developed comprehensive written R0 and SR0 examinations ; in accordance with the guidelines of NUREG-1021, Revision 7, Section 40 The R0 and SR0 examinations consisted of 100 and 99 multiple choice questions, respectively. During the week of September 13, 1993, members of the facility operations and training departments, under the provisions of NUREG-1021, which i require execution of a non-disclosure security agreement, reviewed the i examinations in the Region IV office. The NRC considers the pre-
#
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administration review of the examination by the facility as part of the l examination development process. Therefore, the specific comments resulting from that review are not reported or otherwise retained. The chief examiner ! incorporated the facility review coments and administered the examinations to , the license applicants on September 24, 199 , The chief examiner provided the facility training staff with a copy of the :
"as administered" written examination and answer key along with the pre- 2 administration review coments on September 24, 1993, imediately following :
the completion of the examination by the applicants. The facility took that ! opportunity to further review and comment on the written examination and informally provided written coments on September 30, 1993. The facility's , post-examination review coments are contained in Attachment ! After careful evaluation of the facility post-examination review coments, the ! chief examiner accepted all the facility recomendations for the reasons given ; except as follows. One question on the SR0 examination was deleted from the ; examination as required by NUREG-1021 since it had three correct answers l rather than accepting the additional correct answers as recomended by the ; facility. The chief examiner made the appropriate revisions to the ; examination key l l Overall, applicants performed well on the written examinations. Scores ranged I from a low of 82 percent to a high of 94 percent with averages of 86 percent for reactor operator applicants, 90 percent for senior reactor operator applicants, and 88 percent overall. All applicants passed the written examinatio The chief examiner reviewed applicant performance on individual questions and observed that the following questions were missed by 50 percent or more of the
- applicants responding to the question. The questions are referenced here only l by examination level and question number. Refer to Attachment 3 for the ,
! complete question and answe l Common questions (SR0/R0): 23/25, 36/39, 48/53 Questions on the R0 examination: 23, 45, 55 Questions on the SR0 examination: 13 The chief examiner concluded that no specific area of significant knowledge j weakness was apparent in the responses to the above questions. Therefore, the ! information is provided to the facility training staff for consideration as i feedback into future training need l '
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. . -6-1.3 ODeratina Examinations The examiners developed comprehensive operating examinations in accordance with the guidelines of NUREG-1021, Revision 7, Section 301. The operating 1 examinations consisted of two parts, a dynamic simulator scenario portion and ,
a control room / plant walkthrough portion. The chief examiner previewed and validated the various portions of the operating examinations in the Region IV office during the week of September 13, 1993, and onsite on September 22 through 24, 1993, with the assistance of facility training personnel under , security agreement. The examination team administered the operating examinations during the week of September 27, 199 . Dynami_ , Simulator Scenarios
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The examination team evaluated four crews (one consisting of four SR0 instant applicants rotating through three positions, one of one SR0 upgrade and two R0 applicants, and two of two SRO upgrade and two R0 applicants) on one to three scenarios (depending on crew composition) using the South Texas Project plant-specific simulation facility. The examiners compared applicants' actual ,
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performance during the scenarios with expected performance in accordance with the requirements of NUREG-1021, Revision 7, Section 303, to evaluate applicants' competency on this portion of the operating examination The examination team noted that communications among crew members were ! generally effective and formal, but with several instances of open-ended, l incomplete, and informal communication practices noted. Additionally, the j examination team concluded that the crews displayed effective but inconsistent : command and control attributes. Crew briefings by the SRO were generally ! weak, and not conducted at all in some crews. In one crew, a course of action , which would have potentially mitigated the casualty in progress was discussed " and agreed to by the crew; however, the SR0 never directed the action to be attempted and, therefore, it was not accomplished. In another instance, the i I SR0 was so soft spoken that the R0s had difficulty discerning his reading or reviewing the procedure out loud from his directing of actions. In another ' instance, the R0 stated that "we need to trip the reactor" to which the SR0 immediately replied " trip the reactor" before receiving a report as to why the l reactor needed to be trippe One generic weakness was noted during the conduct of the dynamic simulator i examinations. Three crews were evaluated during a shutdown scenario in which a pressurizer power operated relief valve and its associated block valve were failed open, creating a loss-of-coolant accident (LOCA). Of' the three crews, only one crew immediately recognized the entry conditions for the appropriate mitigating abnormal procedure, one crew was very slow to recognize the required entry conditions, and the third crew never recognized the entry l conditions and adopted an ineffective mitigation strategy. This displays the l applicants' general unfamiliarity with low power or shutdown abnormal operating procedures and entry conditions. Poor procedural guidance for shutdown LOCA also contributed to the crews' inconsistent performance on this scenari Each of the three crews correctly entered Procedure OPOP04-RP-005,
"COMS Actuation or Failure," but that procedure does not successfully mitigate the loss of inventory, nor does it direct entry into or reference
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Houston Lighting & Power Company -7-Procedure OPOP04-RC-0006, " Shutdown LOCA," when attempts to isolate the power operated relief valve fail. The " Shutdown LOCA" abnormal procedure gave adequate guidance to mitigate the event. After a pre-exit meeting with the training department, the licensee informed the inspectors that this weakness had been noted and added to the training curriculum for future license classes and the license requalification training progra , Twelve of the 15 applicants passed this portion of the operating examination.
1. Walkthrouah Examinations The examination team evaluated each of the R0 and instant SR0 applicants using ten JPMs relating to tasks within the scope of potential duties of a licensed R0 or SRO (which included non-licensed operator tasks outside the control room). The examination team evaluated the remaining upgrade SR0 applicants on five R0 or SRO tasks each. The applicants performed some of the tasks in the simulation facility in the dynamic mode. They simulated (through discussions) the remainder of the tasks in the plant integrated control room and at local operating stations throughout the plant. Immediately following the performance of each task, the examiners asked pre-scripted questions relating to the system involved in the task. The questions solicited "short-answer" responses and permitted the applicants to use operationally controlled references to aid in their responses, unless specifically annotated to require response from memory. The examiners combined the applicants' task performance and question responses in accordance with the guidelines of NUREG-1021, Revision 7, Section 303, to evaluate performance on this portion of the operating examinatio ! Overall, the applicants performed adequately. All applicants passed this ! portion of the operating examination with satisfactory overall performance on systems and tasks.
One generic weakness was observed during the conduct of the walkthrough examinations. One JPM involved starting a reactor coolant pump (RCP), and required the applicant to recognize abnormal post start parameters, requiring rapid securing of the RCP. More than half of the applicants tested were very slow to secure the RCP, resulting in pump damage that could have been avoided through prompt operator action. All of the applicants recognized abnormally high pump amperages and reported the observation to their unit. supervisor, but most of the applicants did not recommend securing the RCP until procedurally directed by indications of high vibration. The inspectors noted that neither Procedures OPOP03-ZG-0001, " Plant Heat-up," OPOP02-RC-0004, " Operation of Reactor Coolant Pump," nor 1(2)P0P04-RC-002, " Reactor Coolant. Pump Off Normal," gave the operator quar.titative guidance on RCP trip criteria based on high pump amperage, with the exception of~a caution in Procedure OPOP02-RC-0004, which states, "The RCP ammeter will peg high due to _ _ _ - - _ _ _ _ - _ _ _ .
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. ' -8-starting current but will drop back on scale within 20 seconds after the RCP is started." The operators' hesitancy to secure the RCPs after recognizing abnormal startup parameters was a generic weakness and had potentially safety related consequences with damage to the RCP windings, seal packages, and subsequent reactor coolant system leakage. The procedures for operating the RCPs provided minimal guidance to direct the operator to secure the RCP following startup with abnormally high running amperag The inspectors noted several examples of poor labeling which directly affected, the applicants' performance on the respective JPMs. While conducting tasks in the spent fuel pool filter and demineralizer bays (MEAB,10 foot level), the inspectors noted that several applicants had great difficulty in locating several valves, and once located, seemed unsure of the correct valve position due to poor labeling of the valves. Black marker was used in several instances to informally label valve stem positions, as well as valve number In another task involving local start of the emergency diesel generator after an overspeed condition, several applicants could not reset the overspeed trip due to an inability to locate the reset pushbutton on the overspeed governo The pushbutton was not labeled and was painted the same color as the governor block, where the pushbutton was located. A third example was noted when locally operating a steam generator power operated relief valve, several l operators mistakenly operated the solenoid isolation valves instead of the j open/ shut solenoid override knurled knobs. In an emergency situation, these 4 labelling problems could significantly slow or hamper recovery efforts by ,
l local operator The inspectors also noted an unfamiliarity with dynamic JPM methodology among some of the applicants. When questioned, one applicant indicated that he was not trained using dynamic JPMs. The licensee confirmed at the pre-exit , meeting with the training department that the applicants are given an option to perform the JPMs dynamically if desired, but it is not required by their . training program. The NRC previously noted a training program weakness in NRC ! Inspection Report 50-498/93-01, 50-499/93-01 in that the training program did not require validation of dynamic JPMs prior to use on examinations. The inspectors considered the corresponding lack of a program requirement to conduct dynamic JPMs a potential program weakness and challenge to consistent and effective operator performanc .4 Observations The examination team observed the performance of on-shift operators and plant conditions incident to the conduct of the applicant evaluations. These observations did not impact the evaluation of individual applicants and are included in this report for information onl * Material condition of the plant was essentially the same as observations made in January 1993, including inadequate / poor labeling of valves and cabinets and poor painting on doors and bulkheads.
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- Control room communications were weaker than those observed of the I license applicants, with more prevalent instances of open-loop and other informal practice l
- The operating crew on Unit I was preparing to run the 11 emergency i i
diesel generators and had the procedure with headphones on a procedure ' stand attached to the control panel, standing by at 8:00 am. Due to
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some delays, the test had still not commenced when the inspectors left r the control room at 4:00 pm, but the procedure stand, procedure, and ' headphones were still attached to the panel. Leaving the procedure , st;:nd at the panel for that duration, where it could easily obstruct ! control room operators, is a poor practic l
- A security guard was observed unlocking a door to a room labeled
" Storage Area," which, when opened, revealed two other guards already in l the room. When questioned, the security lieutenant stated that the room ;
was no longer used for storage, but was now used for security building l watch turnovers. Conducting these turnovers behind locked doors in a mislabeled room indicates a continuing insensitivity to appearanc l
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1.5 Simulator Fidelity
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During the preparation and conduct of the operating examinations, the examination team observed no discrepancies in simulator fidelit j
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1.6 Conclusions l The examination team concluded that the performance of 12 of the 15 applicants l for operator licenses satisfied the requirements of 10 CFR 55.33(a)(2) and i recommended that licenses be issued. The three applicants whose applications ; were denied may reapply in accordance with 10 CFR 55.3 ; In general, the examination team concluded that:
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* Individual applicants and crews performed marginally. Communications, although generally effective, were inconsistent, and senior operator direction of crews and command presence was only adequat * Poor plant labeling was observed to adversely impact operator ,
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performance and could significantly hamper local recovery efforts in an emergenc '
* The applicants demonstrated a generic weakness that involved an unfamiliarity with shutdown operating procedures and entry condition This was further affected by procedural guidance unable to compensate for the demonstrated weakness. Although the training department identified this area to be added to the license training program, this is an example of a weakness in self-assessment and timely corrective actions which challenges the effectiveness of licensed operator !
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* The applicants demonstrated a second generic weakness that involved failure to secure reactor coolant pumps when abnormal conditions were I noted. This weakness was compounded by a lack of quantitative procedural guidance and reliance on skill of the craft to determine a ,
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threshold at which the pumps should be secured. This displays a training program weakness and potential for unnecessary equipment damage , due to untimely operator response to abnormal condition l
* The facility material adequately supported the examination development '
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ATTACHMENT 1 1 PERSONS CONTACTED Licensee Personnel
*D. Bize, Licensing Engineer *J. Calloway, Staff Consultant *K. Christian, Manager, Unit 1 Operations U. Coffey, Senior Reactor Operator Instructor *W. Cottle, Group Vice President ', *M. Coughlin, Senior Licensing Engineer * Dowdy, Manager, Unit 2 Operations J. Goodwin, Supervisor, Training Support, Acting *J. Groth, Vice President, Nuclear Generation *S. Head, Department General Manager, Nuclear Licensing T. Hurley, Senior Reactor Operator Instructor * Kinsey, Vice President, Plant Support K. Kline, Simulator Configuration Coordinator K. Krueger, Operations Support Supervisor, Acting * Ludwig, Manager, Nuclear Training Department R. M*Annallay, Senior Reactor Operator Instructor l *L. Myer, Plant Manager, Unit 1 *K. Poling, Assistant Training Manager *J. Shepperd, General Manager, Nuclear Licensing *G. Weldon, Manager Operations Training 1.2 NRC Personnel K. Erickson, Contractor, Pacific Northwest Laboratories *D. Garcia, Resident Inspector S. Guenther, Office of Nuclear Reactor Regulations, Operator Licensing Branch *R. Lantz, Reactor Engineer * Satorius, Project Engineer J. Keeton, Resident Inspector D. Loveless, Senior Resident Inspector J. Pellet, Chief, Operations Section *J. Tapia, Reactor Engineer In addition to the personnel listed above, the examiners contacted other personnel during this inspection period.
l * Denotes personnel that attended the exit meetin EXIT MEETING An exit meeting was conducted on October 1, 1993. During this meeting, the examiners reviewed the scope and generic findings of the inspection. The examiners did not disclose preliminary results of individual evaluations since they are subject to change during the final review and approval process. The licensee did not identify as proprietary any information provided to, or reviewed by, the examiner. The licensee did not state any position on the findings presented during the exit meetin _
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ATTACHMENT 2 SIMULATION FACILITY REPORT Facility Licensee: South Texas Project, Units 1 and 2
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Facility Docket: 50-498, 50-499 Operating Tests Administered at: South Texas Project
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Operating Tests Administered on: September 27 - 30, 1993 These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information which may be used in future evaluations. No licensee action is required in response to these observation During the dynamic operation of the simulator in support of the operating tests, no previously unidentified simulator fidelity problems were observed.
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HL&P 1008A (12-91) . . Attachment 4 t
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Houston Lighting & Power Company I
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OFFICE MEMORANDUM '
- - To NRC EVALUATION TEAM October 1, 1993 l From GLEN WELDON i
Sutyea REBUTTALS TO THE WRITTEN EXAMINATION f I
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Attached are the rebuttals to specific questions on the Written ! Examination administered to License Class #6 on September 24, 199 !
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We respectfully request that the following questions be reviewe ! i SRO Exam: #6, #10, #23, and #44 !
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RO Exam: #25, #49, and #96 (#25 and #49 are duplicates of !
#23 and #44 on the SRO Exam, respectively) l If you have any questions, please contact Carl Brewer at (512)
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972-7967 or Wes Young at (512) 972-766 i
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SRO EXAM: Question #6 i Given the following: ;
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Unit 2 is in Mode An INITIAL entry into the Reactor Containment Building !
(RCB) is being planne ! -
OPGP03-ZO-0032, " Reactor Containment Building Entry" has i been initiate '
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Oxygen levels were verified to be 21% 4 hours ag Which one (1) of the following pieces of equipment is required for the team to enter into the RC y An Oxygen monito t A Neutron Survey Instrumen , A Self Contained Breathing Apparatus (SCBA). A Radiation Monitoring instrument with a 1 R/hr rang . I
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KEY ANSWER: D ; Response to Ouestion i
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Step 4.2.1 of OPGP03-ZO-0032, " Reactor Containment Building Entry", requires that a survey instrument " capable of measuring radiation ,
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levels in excess p_f 1 R/hr" be use This makes the selected ! answer of "D" incorrec ; i
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Step 5.5 (of the same procedure), requires that a sample of the l oxygen concentration be obtained "upon initial entry". . .not 4 hours ~/ ! carlier? later as the question specifies. Therefore, "A" would be correct, *
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except for the 4 hour time dela I
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Based on this knowledge, we submit that there are no correct choices for this questio Ref: OPGP03-ZO-0032, Rev. 3
- resolohin - acceff a. or d as corrcc-1
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l \- '* Reactor Containment Building Entry OPGPC3-ZO-0032 . ' g Rev. 3 f Page 2 of 7 :
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1.0 Purpose and Scope l
? This procedure outlines the steps necessary for ALL entries into the Reactor Containment Building in Modes I and .2 This procedure outlines the steps necessary for INITIAL entries into the Reactor Containment Building (RCB) in Modes 3 through ;
s 2.0 Definitions
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None l 3.0 Precautiens j No entry SHALL be allowed into an area in the RCB prior to a radiological survey by a Health Physics Technicia .2 VHEN oxygen content is suspected to be less than 19.51, cannot be , verified or airborne radioactivity levels dictate, THEN Self Contained Breathing Apparatus (SCEa) -hall be use .3 Every effort SHALL be made to maintain personnel exposure As Low As Reasonably /.chievable (ALARA).
, The area under the reactor vessel (RCB Room 1) SHALL NOT be entered I with the unit in Modes 1 or .5 IF the most recent RCS sample indicates sustained levels of Tritium greater than or equal to 1.0 uC1/ml, THEN SAMPLE RCB atmosphere for Tritium upon RCB entry.
l 4.0 Prerequisites A minimum of two personnel (one Health Physics Technician) has been assigned to each entry tea .2 Each team has the f ollowing +quipmen m s 4. A ey instrument capable of measuring radiation levels in ),&XCG$5 O 7 * e45 'A ) .' exces of 1 Rfhr.
' Jt: .~nkf A h 5 . 4. Neutron Survey Instrument (only in Mode 1 & 2).
4 A> 5 O cr G[ 4. An og monito (only if oxygen levels have not been verified
withan the 8 hours prior to entry.) * l$~)f ts&S I5I /n~e is a sj u p 4L
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Reactor Containment Building Entry OPGP03-ZO-0032 ' ; Rev. 3 ) Page 4 of 7 A RWP has been issued and activated that identifies all planned work or activities for the entr .7 HP has briefed all workers on radiological hazards prior.to entr j The Shift Supervisor has briefed the HP technicians on the scope of l any containment closeout inspections require ;
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, The Shift Supervisor has granted written permission en (-1) to enter .
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5.0 Procedure i The work group initiating the entry coordinates with the Unit Supervisor. Security, and Health Physics as to the time, reason, and i l activities for the entry.
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; The Health Physics Technician obtains and completes the RCB Entry ;
Checklist (-1). '% i
UfeM Ll IF necessary during an Emergency, THEN the Shift Supervisor may * N7 Ay , f waive any or all of the prerequisites with the concurrence of th _
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General Supervisor Radiation Protection or designe ( l hours f t' Ai *~
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5.4 Upon entering the RCB, PERPORM RCB handswitch line-up per IPOP02-LI-0001.
l y i bfCI 5 W 5.5 OBTAIN a grab sample or breathing zone air sample and 02888Ple upon I ' T N 7: r29 pj r/2 pinitialentry,intotheRC .6 The Plant Engineering Department performs a RCS Pressure Boundary Inspections in accordance with the guidelines specified in : OPGP03-ZE-003 L
; All personnel SHALL follow directions from the Health Physics . i Technician concerning ALARA concepts and remain together in a group I while inside the RC ! Plant Operations is responsible for performance of a Containment ]
Closecut Inspection per OPSP03-XC-0002, unless no operators are t involved in the entry with the plant in modes 1 or 2s In these { instances, the Health Physics Technician enterir4 um 5:'M completes ( this inspection for the areas designated by the @t." $ % ervisor and ) provides prompt notification to the Shift Supervisor et marveillance j completio t Upon exiting the RCB, immediately NOTIPT the Control Room that the entry has been complete l l l 5.10 CONFIEM with security that all personnel have exited containmen l 5.11 D45URE the personnel hatch is locked or guarded upon exiting.
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SRO EIAM: Question #10 { WHICH ONE (1) of the following is an PERMIT ENTRY SPACE as defined by the Confined Space Entry Program procedure? ! i Blowdown Valve Cubicles
! Steam Generator 2A !
J Settling Basin ) Underground pipe ways l KEY ANSWER: B .
J Response to Ouestion I Step 2.3 of OPGP03-ZI-0007, " Confined Space Entry Program", states l that underoround utility vaults and open-too spaces more than four j feet deen such as pits, tubs, vaults, and vessels are considered, by definition, a Permit Entry Spac Step 2.1 of- the same : procedure defines a Confined Space as "any enclosed or permit entry l space".
Based on this knowledge, we request that both "C" and "D" be l considered acceptable answers, as well as "B". j
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Ref: OPGP03-ZI-0007 (Rev. 5) { i l l
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Confined Space Entry Procram OPGP03-ZI-0007 ! Rev. 5 Page 2 of 12 Purpose and Scope This procedure establishes the Confined Space Entry l _ Program and establishes the methods to be used to obtain . :
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, a permit, open, enter, or work within an ENCLOSED space i ! or a PERMIT ENTRY space at the South Texas Project (STP) . ' i This procedure specifies requirements for monitoring an i ENCICSED space or PERMIT ENTRY space for oxygen levels or l combustible gas levels prior to personnel entry and i periodically while work is being conducted in the spac .3 This procedure applies to all Departments involved with ! l
- STP and all STP personne '
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! Contractors and subcontractors shall provide their own ! instruments and training for confined space entries, ; unless granted written permission to do otherwise by , l the Industrial Safety & Health Division (ISH). ;
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- Definitions
. CONFINED SPACE: Any enclosed or permit entry-space.
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t NOTE An ENCLOSED space should be identified with an Enclosed Space placard (Addendum 2) and a PERMIT ENTRY space ' should be identified with a Permit Entry Space placard .
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i (Addendum 3) unless prevented by engineering concern ; ENCLOSED SPACE: Any work-space location that is sufficiently enclosed to permit the existence of an . oxygen deficient atmospher , '.s i PERMIT ENTRY SPACE: Any space having a limited means of
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e egress, which is subject to the accumulation of toxic or gf flammable contaminants or has an oxygen deficient atmo- ; spher Confined spaces include, but are not limited to: storage tanks, process vessels,_hins u b_ oilers d ent- f _ilation or exhaust ducts, sewers,.un_derground utility > ;j, g 'd vaultsi. tunnels, pipelines, and open-top spaces more than vy 4g; foUr~ feet deep such as pits, tubs, vaults, and vessel ' CREW LEADER: 7.m person in charge, regardless of title OT or classificatio . - i l-l ,- RO EIAM: QUESTION #49 SRO EIAM: QUESTION #44 Given the following:
- Unit 1, Fire Pump #1 is in AUT All control panel switches are in their normal position BATT "A" CONNECTED and BATT "B" CONNECTED Blue lights are LI Fire system pressure is 130 psi An operator has pushed the STOP/ RESET pushbutton to stop the*
pum The pump restarted when the pushbutton was release WHICH ONE (1) of the following conditions could be the cause of the fire pump restart? ! l Low fire system pressure, b. Unit 2 has actuated a remote start signal.
l c. The control panel has lost AC power.
l d. The Weekly Test Timer (WTT) has actuated.
! KEY ANSWER: B Response to Ouestion - Fire Pump #1 auto starts on a Loss of AC power to the engine controlle This start signal ensures that the Fire Pump remains available even though the batteries may completely discharge during an extended loss of power. When the engine is running, the engine-mounted alternator, via rectifiers in the control panel, feeds the DC control circuits and charges the batterie Therefore, both "B" and "C" should be considered as acceptable answer Ref. Vendor Manual 6004-00002PA LOT 201.29.HO.01 b
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igy Title: Houston Lighting & Power
@v i i e Customer P.O. No. 35-1197-6031 .
67 Job No. CR-0241
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l 69 2 SP ecification No. 9Q270MSO21-E D
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pcM ncRx 1:cax ncgx Controller - Fire Pump 02
,3)@Ql,-$3/Q3 34) , ~
Item No. 9-Q27-0-E-FC002-A surY.2r ~ B(4 rrAnvRE'/su rgG
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HOTES - 1. THIS DP.c. I::C SUPPLU'ENTS THE CONTROLLER DIACRAM D'.'G. NO. 770 . TEFJ1INAL AND WIRE NUMBERS CORPISPOND TO T110SE ON THE CONTROLLER FITERE::cE ::0TE No. 3 ON DUG NO. 770 . REFERE::CI D*. G NO. 7704 TOR COMPLETE DESCRIPTION OF SYMBOLS
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1. Diesel Engine Driven Fire Pumps The Fire Pumps are manufactured by Patterson Co. and are horizontal split casing single stage driven by a 285 horsepower Ctmains diesel ' engine at 1170 rpm. These pumps are rated to l deliver 2500 gpa at 125 psig. Each Fire Pump i is located in a separate steel reinforced concrete room inside the fire pump house. The engine has a turbo charger, filter intake , ; stilencer, jacket water cooling system, starter l with an alternator that connects the starter to one of two batteries during cranking, an l mH=rpator for recharging the batteries and l supplying control power when there is a loss of normal power, a governor with mechanical l overspeed protection, and an electrical ) overspeed protection tachomete The engine jacket water cooling system is cooled by a heat exchanger supplied with raw l cooling water from the disc'arge of the fire i
' pump via a pressure regulator and a solenoid valve that opens or a pump start. The pressure regulator and solenoid valve can be bypassed by opening a normally closed manual
' valve. After passing through the heat exchanger the raw cooling water is routed to the drain syste ! Inside each pump room is the engine controller, dual starting battery banks, and a 550 gallon ! fuel tank. The fuel tank is filled from the Auxiliary Fuel Oil Storage and Transfer System via a normally closed manual fill valve. By design the fuel tank can be refilled within 8 hours.
, 1. Fire Pump Controller The Fire Pump engine controller has a START pushbutton and a five position mode selector switch located behind a glass window in the controller cabinet door. Located on the door front is a STOP/ RESET pushbutton. The controller door is normally locked and entry should only be made by obtaining authorized permission. The positions selectable by the mode selector switch are T'ST, AUTO, MANUAh MANUAL B, and OFF. The function of these switch . positions are: l i ,
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IDT2 01. 29.HO. 01 PAGE 5 OF 28
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manually vented after the fuel tank has been refilled.)
When the fire pump is running in the
- TEST mode, it can be manually stopped l
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by placing the mode selector switch .
' in the AUTO position and depressing the STOP/ RESET switch or placing the l selector switch in the OFF positio i If the latter is used, return the switch to AUTO to place the engine in -
standby readines B. AUTO - The green ready light illuminates on the controller and the engine will
automatically start on: Low fire water header pressure (This start signal.is set up to. start the first engine after a short time delay and the second and third engines after subsequent delays to keep system pressure transients from causing inadvertent starts of the first engine and allow the header pressure to be recovered before 3 starting the subsequent engines. The .
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time delays are arranged for 5 seconds at 120 psig, 15's'econds at 105 psig, and 25 seconds at 85 psig , for starting Fire Pumps 11,12, and 13 respectively.) Loss of AC Power to the engine controller. (The purpose of this start signal is to ensure Fire Pump availability as the batteries could self discharge between the time after , loss of power and when needed. When I the engine is running the engine alternator via rectifiers in the control panel feeds the DC control circuits and charges the batteries). Remote start sianal from either unit control roo (It is important to note that the Fire Pumps cannot be ) stopped from the control room.) Weekly TesA.If.pper 9TT) on - ,;. Automatically starts the engine once a week, runs it for 30 minutes, and - then stops i ..
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- The Schematic Wiring Diagram Troubb Shooting . Drawing 6801 presents a complete schematic diagram that FIRST CHECK THAT THE CONTROL IS PROPERLY SET - ~
i!!ustrates a control equipped with all of the tegularly cata- UP TO OPERATE WITH THE ASSOCIATED ENGINE AND logued accessories. Most controls as shipped do not include all TO YIELD THE DESIRED OPERATIONS. READ THE J of these accessor,es, therefore, a separate schematic diagram is GENERAL INFORMATION PARAGRAPHS AND CHECK Q
na for each control to show that control as it is built. Each such diagnm as identified by the serial number of the control THE SPECIFIC DRAWING FOR OPERATION SEQUENCE Throughout this booklet mention is made of the many that it represents and notations on the diagram show wiuch of optional features that are available both in the original the optional items are mcluded and how the internal inte construction of the control and through a variety of inter-camectens are mad connections at the special terminal strip located on the rear The schematie wiring diagram is presented as a " ladder dia- side of the relay panet Each control is set up at the factory < with the equipment called for in the purchase order, for use l gram". The flow of cunent is from one " rad" of the ladde via " rungs" of the ladder to the oppoute " rail". As draw with a specific engine and to respond to selected startmg and l l battery power comes in via Terminals Nos. 6 and 8, and is stopping ugnals, etc. This information is all shown on a j conducted via diodes 3R and 4R,thence through a contact on schemaue wiring diagram which is identified by the serial .
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the control switch to the upper rail; current then flows via number of the specific control. In any instance where the
* operation of the control is other than what is desired, refer-relay cods and/or lamps to the lower rail, and back to the " ence to this diagram may indicate that a change in interco hattenes vta Tenrunal No. I nections wdl be all that is requued in order to obtain the For identification purposes each rung is numbered. These desired operating features. It may also be that the engin numbers appear below the lower rail and oppoute each run mounted accessmy equipment differs frorn that for which the Above the upper rail and opposite each rung in which there is control as set up. If this is the case,some rearrangement of the "
a relay coil, there are other numbers; m some cases as many as imuconnectens of the irnunal tenninals snay be all that is
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four numbers appear. These are the rung numbers in which the needed to make the control operate with the engin " mdmdual relay contacts wdl be found. For example, above
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rung No. 4 in which the coil of relay 7CR is drawn, number . Failure of Engine to Crank and Run When Signaled to d ' 6. 73. ),2. 6.J sppear. This coding means that there is a contact Start: el i of relay 7CR m each of these four rungs; that the contacts in Turn swnch succesuvely to " MAN. A" and to " MAN-t- rungs Nos. 6 and 73 are normally open cotitacts.and that the B". In each position press " START" button and hold a contacts in rungs $9 and 61 are normally closed contact for four to ux seconds. Observe whether engme spms ll
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Underbrung the number indicates a normally closed contac briskly since this is a good indication of battery con- i x " > All contacts are drawn in the poution in which they reside daion. With the control selector swach in enher of the 1 le when the relay cod is de.enerF ize "M AN." poutions all of the relays are bypassed, so if the ! e- engsne wdl not run, the problem is almost sure to be ( The relay numbers are arbarardy chosen. Where the circuit outade of the Control.Obviously.if the cranking is very
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requires more contacts than are avaslable in a ungle relay of
l the type used, the addnional refsys carry the same number weak the battenes are low and must be recharfed or tePl aced. If crenking is brisk and still the engme lads to es wnh a letter added. (Example: Relay 9CR. 9A CR and 9B CR) run, the most likely place to look for trouble is in the to The abbreviation "CR" deugnates a convenuonal control relay fuel system. The engme manufacturer's instruction book ed jy that operates and releases wahout deby. "TR" destenates 3 should yield anformation in this regard.When the engme 3, tunmg relay subsenpt letters "SO" for -Slow Operate" md'- as to run, whether under automatic or manual control, cates that the time delay period before the contacts operate Termmal No. I is energized. (Terminal No.12 also with starts when the relay is energized. Subscript "SR" for " Slow spark ignition engmes). As a final check of the Control
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fn Release" indicates that the time period commences when the ten for voltage between Termmal Nos. I and 11 and 25 relay is de-energszed. Relays that delay operation m both between Nos.12 and 11. With voltage present here and 0 ducctions show "50" and "SR". Alongude contacts of the the engme crankmg brnkly, the trouble 15 NOT in the !
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tunmg relays the notation "TDC" indicates " Time Delay Contro l Cloung" and "TDO" mdicates " Time Deby Openmg". The Failure of Enfme to Stop Upon Cancellation of the l
.ge numbers, unless otherwise noted. in all cases refer to the set Signal to Run:
i nd ttme m seconds. The ratchet relay is designated "RR". This Fast be sure that the Control is set up for an automatsc I relay shdts both sets of us contacts each time its coal as stop. If it as so set up.the ugnal to stop comes via tmung i energued so n always reudes wnh one or the other set of us relay 4TR which may be adjusted for as long a time as
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et is c niacts close tlurty mmutes, so check that 4TR has tuned out.When l Wire identification numbers are used throughout the diagram 4TR times out, as contact opens and thus breaks the i
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and these same nombers are affhed to the wues in the contro cucun to 3CR to cause an immediate shutdown of a mt diesel engine. A spark ignaion engme fitted wah a All pomts that are connected together bear the same numbe All pomts that connect to termmals(both for external and for solenoid valve m its fuel line will run until the fuel that is 5e5 intemal connecuon) use the number that the termmal use downnream from this valve is used up.then wdl stop. If n is a spark :lmstion engme. be sure that this valve a Diodes are used in the circun where power is to be permuted
** - to flow in one ducction only. When the two batteries are to artstalled. If there is no such valve. remove mternal feed a ungle load. as as the case with 3R and 4R where the corm e etio n between Terminals Nos. 33 and 34 3ed 2 CAUTION: IF THIS VALVE IS NOT INSTALLE power flows amultaneously from Termmals Nos. 6 and 8 and ENGINE SHOULD HAVE SOME OTHER METHOD OF is directed into circuit 38.two diodes are connected "back-t ACCOMPLISHING ANTI-DIESELING. When relav back". Diodes 1R 2R and SR-6R are also examples of"back-t back" connecuon. but m these two instances the power flow is 4TR is used in the cucun there will bejumpers runnmg ' "m" from a smgle source and is drvided to feed "out" to two from Termmal No. 49 to Termmal .Nos. 51 and 5 separate load Terminal No. 49 connects to the contacts to the contact ~
The current flow is in the ducction of the arrow head; te' re y en 4R reae.Vs ps timeh e
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O from postrve to negat ve. and Drawing 6801 shows the dsodes'
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is suspected.ternporath renme the jumper n"om ds- 1, as they are connected for common batte'Y termmalINo. ill- Termmal No. 49 leavine 51 connected to 52. Wah
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. ell . connected to the poutive pole of each battery (positive Termmal No. 49 disconnected. the engine should stop ge t ground) Negauve ground systems have the diodes drawn wah without delay upon cancellatson of the signal that
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their arrows pomtang m the opposne ducctio started i ] 7 M .
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.- ' J voltage at 53 and I8: and proceeding,you should find ?
NOTE: BE SURE THAT ALL SIGNALS CONNECTED full v ltage at Termmal N2. 70 on relay 9ACR rnd at 71 ,, TO START THE ENGINE ARE CLEA on relay 6CR. With full voltage through 13 this point, 4 3, Persistent Cranking of a Running Engine: turn the control switch to " TEST". This should put ! . voltage through to point 75 on 3ACR. Next check at - l If the ensme-mounted device that senses th~e fact that point 99 on relay 2TR,then point 3S which is a terminal the engme is running fails to serxl its ugnal to the { Control, or if the Control fails to property acknowledge on the selector strip. Assuming that there is ajumper 35 this ugnal, then, even though the engine is running, the to 36 voltage will also appear at points 36 and 74 on Control wdl engage the engtne starter stermittently for relay BCR, but the voltage beyond point 75 wdl be a matter of about ninety seconds. then wdllock itself present for ordy alternate ten second periods as relays out of service and burn its red " FAILED TO START" ITR and 2TR alternate crank and rest periods. During . I Ismp. Check to see that there is rated voltage from the ten second "ON" intervals there should be voltage Termmal No. 2 to Terminal No.11 when the engine also on either Terminal No. 9 or Temunal No.10, runs. If there is no voltage here, then the trouble is in depending upon which side the ratchet relay happens to the engme-mounted equipment and the running signalis be rending, but NOT on both 9 and 10 at same tim not oemg sent. If there si voltage here,then relay 9ACR The foregoing discussion has dealt with the relays and is not operatmg properly, or an internal wire to the cod contacts that are directly in the etreurt to the engme of 9ACR has become disconnected. Check for voltage on cranking motor controls. The circuit passes throuch the coil of relay 9ACR. Interchange 9ACR with 9BCR normally closed contacts of relays 9A CR,6CR. 2T I to check whether the relay itself is the source of the 13CR and BCR. It passes also through relay 3A CR, but troubl in this case through normally open contacts, so 3A CR I must be in its operated (picked up) poution. Thus,in l S Blue I. amp Goes Dark- the point.to. point check,if voltage is found at point 71 i' Each of the two blue lamps is ass:>ciated with one of the of relay 3A CR but not at point 75, check first point 98 two storage batteries. If a lamp goes dark (assuming that on relay 3A CR since this point is a cod termmal and the lamp itself is not burned out), it means that its must have voltage present. Check next points 52 and 17 l associated battery has been disconnected or that this on the selector switch. There must be voltage at all of l i battery is in bad condition or is badly in need of r these point chargmg. Turn the control switch to "OFF" and press it is possible that the enFine will crank and run when the the RESET" button. Even a badly discharged battery control switch is tumed to " TEST" but wdl not do so I
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or one that is m bad condition wd! usually be able to UPon operation of the pressure sw tch. This is because nght the lamp. If the lamp burns when the " RESET- relays 3CR and 3A CR are energtzed directly from the button is pushed and then goes dark when the button is test switch whereas the pressure switch directly operates released. the associated battery is useless. If the lamp relay ICR. which m turn operates relay 2CR. which in I remams bright after the button is released, turn the co tum c mpletes the cucuit to relays 3CR and 3A CR. By trol switch to the " MAN." position and try a manual bleeding off pressure m the line to the pressure switch, start from the suspected battery. With a low battery the its contacts should close. Now, with the control switch . < erankmg wd! be sluggish and the blue lamp probably wd! set at AUTO. check for voltage at pomts 47. and 46 ( go dark and stay dark even after the start pushbutton is n relay ICR and at pomts 44 and 51 of relay 2CR and . release via jumper (or jumpers) to point 48 and thence to 98 on 9 NOTE: Even with a fully charged battery the blue lamp relays 3CR and 3A C . . . . _ . wdl be somewhat dimmed while the engme is Eneme Cranks Briskly But Fails to Run: bems cranked. Read the paragraphs headed Assuming that engme is gettmg fuel. then in order for '
"1noperative Battery Protection- for a more the Control to make it run, c. is necessary only that t complete discussio Terrrunal No. I be energized. (Termmal No.12 also if it ,
s . Engme Fails to be Cranked Upon Signal to Start: is a spark ignition engme with a solenoid valve in the fuel . line). With the selector switch set at " TEST". check at Assuming that both battenes are connected and are at a Terminal No. I for voltage. lf voltage is present here. the good state of charge, that the Control selector switch is Control ts establishing the "run' circuit for a diesel 1 set at "AUTC." and that the mterconnections of mte engme and the ignition circuit for a spark igmtion i nal temunais is such that the control is intended to re engme. Where it is a spark ignition engine with a sole- I pond to the sienal to start; a failure to commence
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noid fuel valve. Termmal No.12 also must be energize cranking upon receipt of signal (or withm 10 seconds if both Termmals I and 12 are energized. then the after signal) is due to a bad connection or to a faulty season for fadure to run 15 NOT in the Control but is relay. First determine whether the problem lies in the associated with the engme. Termmal No. I is fed via a Control or at the eneme. To do this, try crankmg via the contact of relay 3CR and a selector switch contac i start pushbutton wdh the control sw tch set at " MA Check for voltage at Termmals Nos. 98. 53 and 34 on ! A' and " MAN. B" in turn. This will check out the 3CR. then at 34 and I on the selector swrtch to dete engme winns and the engme mounted equipment smce mme where the circuit discontmuity resides. If it is a all of the relays m the Control are bypassed when the spark igmtion engme, check that there are jumpers switch is m either " MAN." position. If it is found that between Ternunals 12 and 31 and between 33 and 34 on crankme takes place and the engme runs as it is supposed the internal connection stnp and that there is voltage on to under manual operation, then there probably is an Termmals 53 and 31 of relay 3C "open" somewhere m the Control wirmg or m one of Engme Crankmg in Short Spurts-the relars The crankmg circuit is completed to external l Ternunals Nos 9 and 10 arternately and interrruttently The engine. mounted device that puts voltage on Ter-via several relay contacts. Before commenemg a detailed minti .Vo. 2 must do so ONLY when the engme is firine checkout at is surgested that the hot cable be removed and i.s runnmg above crankmg speed. When this device from the startmg notor on the engme so that actual puts vc/tage on Terminal No. 2 the control is signaled to stop craWng. If voltage appears at Termma' No. 2 crankms cannot occur. All other connections should
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BEFORE the engine is firing and runnmg. cranmg wdl remam m place. Use a voltmeter or a test lamp and connect one side to Termmal No.11. Touch the power be discontinued too soon; then. as the engme speed circuit breakers then to points 38. 53 and 18 on the drops. its speed sensmg device wdl open its circuit and wiector switch. h'eeardless of selector switch positio the control wdl resume crankmg. The remedy rests m the erigine. mounted speed sensinit device. not in the vou should find fud voltare at 6. 8. 66. 68 and 35. With control
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the switch set at "AUTOJ there should also be full O/
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RO EXAM: QUESTION #96 l
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Given the following:
- A spent fuel assembly is being raised from its slot in the i storage pool for return to the reacto l - Gas bubbles are coming to the surface of the poo i - Radiation levels in the Spent Fuel Pool area are increasin )
WHICH ONE (1) of the following actions is required of personnel in, the Fuel Handling Building (FHB)? Notify the control room to sound the Containment Evacuation Alar Immediately evacuate all personnel from the Fuel Handling Buildin Notify the Core Loading Supervisor, then replace the damaged fuel assembly in its storage locatio Move the fuel assembly into the RCD and notify the control l room to initiate containment isolatio ! KEY ANSWER: B l Response to Ouestion This question leads the examinee to believe that he is the Fuel Handling Machine Operator with refueling operations in progres In this position, he is required to inform the core Loading Supervisor, Shift Supervisor, and HP of any impending radiological situation. Since the Fuel Handling Machine Operator is in constant communication with the Control Room, he would inform the Control Room of the conditions inside the FHB. The control Room personnel would, per the instructions of OPOPO4-FH-0001, make an FHB evacuation announcemen (The Fuel Handling Machine Operator does ' not have the capability to make the announcement).
l We believe that the safest place for the damaged fuel assembly to be stored would be back in the storage-rack location from which it was retrieve This reasoning is based on the following: The Spent Fuel Pool provides greater than 23 feet of borated water for radiation shielding and Iodine scrubbin . The storage rack is a structurally sound place to support the damaged assembly without effecting other fuel assemblies located in the poo . OPGP03-ZO-0002 " Qualifications and Conduct of Operators I
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for Cranes, Hoists, and Monorail Systems", Rev. 4; ) , requires crane operators to land and release all loads l l prior to leaving any crane unattende ]
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l l OPOP08-FH-0009 " Core Refueling" Rev. 8; requires l operators to immediately notify the Core l Loading Supervisor if fuel damage occurs during refueling operation , I l Based on this knowledge, we request that both "B" and "C" be . considered as acceptable answer Ref: OPOP08-FH-0009, REV. 8 OPGP03-ZO-0002, REV. 4 OPGP03-ZA-0010, REV. 15 fbY0
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STP ! HOUSTON LIGHTING AND POWER COMPANY SOUTH TEXAS PROJECT CONTROLLED COPY ELECTRIC CENERATING STATION [ FIANT PROCEDURES MANUAL ll NAT I 4 l99[
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O DEPARTMENT PROCEDURE J,g gC V , SAFETY-REIATED (O) "
. Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 (General)
Page l'of 5 APPROVED: A)LJ R-47 f-/y-9L PIANT ERATION iMNAGER DATE APPROVED DATE EFFECTIVE Purpose This procedure provides actions to be taken in the event a spent o new fuel assembly is dropped or damaged during fuel handling operation .0 Symo ton s: a'nd Entry Conditions Damage to a spent fuel assembly accompanied by alarms (RM-11 or RM-23) j on radiation monitors located in the vicinity of the damaged assembly: 2. Containment Building: 2.1. Containment Atmosphere Monitor; RT-8011 2.1. Containment Purge Exhaust Monitors; RT-8012 and RT-8013 2.1. RCB 68 ft Area Monitors; RT-8099 and RT-8055 2. Fuel Handling Building 2.1. FHB HVAC Exhaust Monitors; RT-8035 and RT-8036 2.1. FHB Main Floor Area Monitors: RT-8081, RT-8089, RT-8090, RT-8091, and RT-8097 Cas bubbles originating from the damaged assembl .3 Visual observation of a fuel handling accident (e.g., dropped fuel j assembly).
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Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 Page 2 of 5 l
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l o High Radiation on RT-8012 or RT-8013 will . actuate Containment Ventilation Isolatio o High Radiation on ET-8035 or RT-8036 will I actuate DIB HVAC in the Emergency Mod o The Shift Supervisor is responsible for overall performance of this procedure.. The Shift Supervisor MAY delegate applicable responsibilities to the Core loading , Supervisor as appropriate.
l l l 3.0 Icmediate Actions j None ! Subsecuent Actions l l l E a spent fuel handling accident has occurred in RCB, ]]ig! DIRECT , Control Room to sound the Containment Evacuation Alar .2 H the spent fuel handling accident has occurred in the DiB. ED! i EVACUATE the FH j H a new fuel accident has occurred, ]]i.E,]! EVACUATE the area surrounding the damaged fuel assembl .4 NOTIFY Health Physics and Reactor Engineering of the fuel handling accident.
, REFER to OERP01-ZV.IN01, Emergency Classification, for potential l . Emergency Plan implementation.
( H a spent fuel handling accident has occurred in the RCB. THEN f PERFORM the following: 4. ENSURE Containment Normal Purge Supply and Exhaust fans secured: o "SPLY FAN llA(21A)" "EXH FAN llA(21A)" o "SPLY FAN 11B(21B)" *EXH FAN 11B(21B)" ,
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Fuel Handline Accident OPOPO4-FH-0001 Rev. 2 Page 3 of 5 i 4.6.2 ENSURE Supplementary Purge Supply and Exhaust fans secured: o "SPLY FAN llA(21A)" "EXH FAN llA(21A)* o "SPLY FAN llB(218)" "EXH FAN 11B(21B)" . 4.6.3 ENSURE Normal Purge Isolation Dampers closed:
o "SPLY OCIV NOV-0007" ; l o "SPLY ICIV MOV-0008" j
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o "EXH ICIV MOV-0009" o "EXH OCIV MOV-00lb" . 4.6.4 ENSURE Supplementary Purge Isolation Dampers closed: o "SPLX OCIV FV-9776" ! l o "SPLY ICIV MOV-0003" o "EXH ICIV MOV-0005" l o "EXH OCIV FV 9777" , ___ 4.6.5 If an automatic Containment Ventilation Isolation occurs. IHEN SECURE sample pump for RT-801 ; ___ 4.6.6 II radiation levels OR sampling results indicate that spent i fuel assembly cladding has been breached IHEN PERFORM the following: l 4.6. INITIATE recirculation through the Emergency Containment Charcoal Filters to reduce the concentration of 1-131 per 1(2) POP 02-HC-0001, Containment HVA .6. VHEN the following conditions are met. THEN OBTAIN a purge permit and COMMENCE a Normal Containment Purge per 1(2) POP 02-HC-0002. Normal Containment Purge: o Health Physics concurs that a purge is necessary, o Sampling results indicate the I-131 concentration has been reduced to its non-occupational MP o Further recirculation of the containment ; atmosphere through the Emergency Containment Charcoal Filters will NOT substantially reduce i the concentraion of existing radionuclide l l
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Fuel Handline Accident OPOPO4- m -0001 Rev. 2 Page 4 of 5
' J_E a spent fuel handling accident has occurred in the FHB, Elgi PERFDRM the following:
4. ENSURE EB HVAC operating in emergency mod j 4. ESTABLISH ONLY two trains MB exhaust fans (main and booster) operatin o , 4. PLACE one non-operating train NB exhaust fans in PULL TO IDCK. p . 4. SECURE one WB filter train by manually closing the outlet Q dampe IL REQUEST assistance from Reactor Engineering to determine a plan of action for inspecting the d==mged fuel assembl .9 UHEN Health Physics has verified that radiological conditions are acceptable, DIEN the Shift Supervisor may allow personnel to return to the accident area to implement recovery action i 4.10 VHEN approval has been obtained from the Shift Supervisor MfR Reactor Engineering Supervisor, THEN RESUME fuel movemen !
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l Fuel Handlinr. Accident OPOPO4-FH-0001 , ' Rev. 2 ) Page 5 of 5 j 5.0 References UFSAR Chapter .2 Technical Specifications: , , 5. T.S. 3.9.9, Containment ventilation Isolation System 5. T.S. 3.9.12 FHB HVAC OERP01-ZV-IN01, Emergency Classification (2) POP 02-HC-0001, Containment HVAC (2) POP 02-HC-0002, Normal Containment Purge (2) POP 02-HC-0003, Supplementary containment Purge SV129V00012 HVAC FHB Supply System V129V00013. HVAC FHB Exhaust System Logic Diagrams 5. , RCB Purge Isolation RMS l 5. Z42113. Instrumentation Actuation Trains A, B, & C l l 5. Z41600, FHB HVAC Main Supply Fans
5. Z41601, FHB HVAC Main Exhaust Fans , 5. , THB HVAC Exhaust Booster Fans l l 5. , FHS HVAC Emergency Operation 5. , FHB HVAC Exhaust Air Bypass Line Damper 5. Z41903, Spent Fuel Fool Exh Radiation Monitors l 5. Z41608, FHB Exhaust Filter Outlet Damper
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Core Refuelinc OPOP08-TH-0009 Rev. 8 Page 6 of 30 4.17 WEN the cause of a valid high flux at shutdown alarm is identified AND the Core Loading Supervisor deems that conditions are safe for all personnel, THEN Core Alterations may continu * 4.18 H there is a malfunction DR suspicion of malfunction of any fuel handling equipment, THEN all fuel movement SHALL be terminated aND , the Core Loading Supervisor SHALL be notifie l 4.19 Handling tools SHALL NOT be left attached to fuel during planned work stoppage .20 WEN moving fuel assemblies between the fuel transfer canal and the spent fuel pool, THEN special care SHALL be used to ensure that the fuel assembly does not make contact with the transfer canal gate wall .21 Jf, during the course of refueling, fuel damage occurs, "IllEN all refueling operations SHALL be immediately stopped MD the core ; loading Supervisor notifie This step does not apply to damage identified during core offload visual check l 4.22 H an accident involving fuel assemblies occurs , . THEN GO TO OPOP04-FH-0001, " Fuel Handling Accident" 4.23 The Refueling Machine Operator SHALL NOT lower a fuel assembly into the reactor vessel until the ICER data from loading the previous fuel assembly has been evaluate l 4.24 The Refueling Machine Operator SHALL NOT disengage from any assembly I in the core until nuclear instrumentation indicates safe conditions A.NJD approval to disengage is received from the Core Loading Superviso i 4.25 UHEN irradiated fuel is being transported through the fuel transfer tube, THEN high radiation conditions may exist in the FHB penetration space in the vicinity of the fuel transfer tube. Access d to this area SHALL be controlled per OPRP07-ZR-0009-04, " Guideline '2 For Spent Fuel or Irradiated Material Transfer". * C-4.26 Irradiated fuel assemblies freshly unloaded from the reactor vessel , ' SHALL be stored in the Region I racks in the Spent Fuel Pool for at least 30 days prior to being transferred to a Region II rack
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4.27 The following SHALL be performed to prevent tools, equipment, and other foreign material from falling into the reactor vessel or refueling cavity: 4.2 All handling tools SHALL be thoroughly cleaned before insertion into the reactor vessel.
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Core Refueline OPOP08-FH-0009 Rev. 8 Page 3 of 30
% .5 The Health Physics Manager, or designee. SHALL be responsible for g providing appropriate radiological coverage during refuelin o~
3.0 Prerecuisites Prerequisite Checklist (-1) is completed prior to core offload'VR core reloa .0 Notes and Precautions All Core Alterations SHALL be observed and directly supervised by either a licensed Senior Reactor Operator or licensed Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation (T.S. 6.2.2.d). This person is hereafter referred to as the Core Loading Superviso .2 L'he re allowed in the procedure, the Core loading Supervisor may delegate duties, provided that the designee meets the same qualifications as the Core Loading Superviso .3 Ihe Core Loading Supervisor may verbally authorize a Reactor Operator or Reactor Engineer to initial /date/ time Puel Transfer Forms (FTF) for hi . any of the following conditions occur, THEN Core Alterations SHALL be suspended pending investigation by the Core Imading Supervisor: 4. An unanticipated increase in the neutron count rate, by a factor of two occurs on both source range NI channels during any single refueling ste . An unanticipated increase in the neutron count rate, by a factor of five occurs on either source range N1 channel during any single refueling ste . There is an unexplained decrease in boron concentration greater than or equal to 25 ppm between two successive samples in either the operating RHR loop (-2) or the refueling canal (-3).
4. The operating RHR loop shows an unexpect.ed temperature change of more than 10 *F between two successive readings (-2).
4. One or both of the plant source range NI channels are inoperabl . Communications between the Control Room Operator and the personnel in either the FHB or RCB are los .- .
_ - Ps O 2 Uni 8 t _____ OPOP08-n{-00 (Page 4 09-RevPOP 8 0 -ni-00
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orENSURE wi e of Pag levecol ntro lthin ho2 ___ Refu ling f 5) 1 23 of 30 will e betwe r ds urs o rom Cycle (Refe nsur e en 63 that ft 0 theofthe e begin n i a renc that in e nr fu ling cng of _____, to Cycle Lastdr ining ow de 6.32). c a Te66hnic lavi movem ad ____,. IFJII1LI_ J 2 Verifiedn into The c ft 6 ent of ENSURE t Date he HVAC upper limiSpe a o ific tiin.totye is filled fu l n3 The lowe perfor 1(2)P duc t t a al e exhas will pr / is r limi t Laster m d wisp 03-E -000 c A - s t atio ns. thin 8 hou 2, *ESF uTime-t (Refer e lude w met 2 Perfor mde rs prio Pow enc ater DISURE Date: a e 6.33).
Pe r to the er Av ilabili Last netratio 1(2)P prior to c _,__ co mm encem ty*ha ent ofs be en _ 2 the Per fo r e n Che klist" hSP03-XC-000__________ mdcommence a s be Date- ment of en perfo 1, Time - c ore DiSURE perfo "Refu e ling N-32 ) r md e an on An log a c ore e alterma d wi th Containm _ either: both plaCha nnl e r tions.in 100 ho ent 25 1 V nt s Oper Tim :e urs ource r ithin 8 ho ationa l _____________, 2 At le perfo r at s urs prior to ange NI cha Te t s nnelsN-31 s
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_ head lmd e once com 3 r per OR men c ing c adn
Last Per/4. fo r e t 9.2 ifthe)foini days sinc ore al ter
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Refer ence tial c ore e the atio s EN m d (N-31) ate: D c alte r ACOTs w SURE (N-32) Dat Te hnic l ations a er _OR rhas efu been that OPSP e: _ _ ___________, c Spe ifica (tii . plugs,le ing c md 03-FH-0001 ,____ ves mperfo r e Tim e on e drivea hine o wi . s ods, rhni Time :: ________, La ts rl (Te cl Pe formd c e a Spe r hoistthin c ifor fu e ass fo r l houe the m rs100"Refu ing prio Ma c hi _l _ _ ___ r ne Pull _ co E dir e Date: ic at emblie s ov em to Tes t" ct tionsntro l ro om comm n _,_.______i n3 o wi ent of using the
/ . 6). thin the thimb ncin in the ope u rica tio _______
Tim * re c le 5). g core RCB ato r ns ha e a tor ad n s be en per Perfo r e alte rad n s sonnele tab md Date: atio ns DiB within c I at lished the e betw _ ho ur e een _________ a (Te r fu ling c hnic l __, prior Spe ific ti a to Time : on _,_,,_,_____,_, ____,,,__
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OPGPC Re O ratorss for Page la and Conduct ofS* stem _ ualifications and Monorail s Cranes Hoists less eralPJgging) unaudible NOTE or and 26 (Gen visible signalman r OPGP03 ZI00should ben by the -
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shou ld be given used.peradio, Signalsclearly etc.)istransit agreed and upo o peration e d signal crane s to crane [ Signals for voice comm unication (telephon i nals houldbenflict ,cialdevabnormal ices prior nne s with g standar at alltimes. Speand should not co AM9 (Control craneo perator installed wamingapproaching FF", persohs pe 5.2.2.14 Activateintermittently whenved safeload patcrane controllersin"Ony only appro crane v FoUo w ) l cethel) before Ica ing a f 5.2.2.15 of Hea y Loads . v attachedloads,p abrake of po (if applicab e wer. Verify , Land andreleasethe parkingwing aloss 5.2.2.16 and set FF' follo tmattende all cran e controllersin"Ower u o restoratio rails [
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Place operation pon p ensurethe g , 5.2.2.17 crane Cran es initiallifts towork liftby a activitie A lk the railsprior aware toof otherlude t the first lift af nduct Gantry come Co 5. The Operator clear shallwaand tobe of initiallifts ing in 5.2. Examplesusing a particular and the e path arecran c 's path. elto ak walktherailsor clear. The Signa lun h, etc lman dur partic u lar Operator f rombre Operato Such and returning additionalpersonnand ces path may require rails not TheOperator ensure the vel to h personnel as requiredidentifyingSuch pe 5.2. crane tra shall employ then responsible suc e fora tim ly manne certifie personnel are lmaninor rigger o ator notifying the Signarequired to be p ilding Polar Cranes closed)for 85 02) tainment Bu (po werbreakers c hanisms. (SER Conductu Reactor crane controller Con ircuitrying not in use. (Modeany co 5. o when Warm p theutesbefore perat 5.2. min n orth-south axis along a crane es Park the cran ) n ot i' only) obile when 5.2. es (inc luding m crancs utdoor or Cran all o secure w 5. Conduct . Outdo wn or other ise Tie do 5.2. .
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Core Refuelinn OPOP08-FH-0009 Rev. 8 Page 23 of 30 PREREOUISITE CHECKLIST OPOP08-ni-0009-1 (Page 4 of 5) INITIALS ,
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Refueling from Cycle to Cycle l Unit l
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l ENSURE within 2 hours of the beginning of movement of fuel
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2 l or control rods that the refueling cavity is filled to a level between 63 ft 0 in and 66 ft 6 in. The lower limit will ensure that Technical Specificacian 3/4.9.10 is met (Reference 6.32). The upper limit will preclude water draining down into the HVAC duct exhaust (Reference 6.33).
Time: Last Verified Date: 2 ENSURE 1(2) PSP 03-EA-0002, "ESF Power Availability" has been performed within 8 hours prior to the commencement of core l alteration ) Time: Last Performed Date: l 1(2) PSP 03-XC-0001, " Refueling Containment l 24.0 ENSURE ! Penetration Checklist" has been performed within 100 hours f prior to the commencement of core alteration l Iest Performed Date: Time: , 25.0 ENSURE an Analog Channel Operational Test has been l performed on both plant source range NI channels (N-31 and i N-32) either:
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25.1 Vithin 8 hours prior to commencing core alteration OR I OR 25.2 At least once per 7 days since the ACOTs were ! performed for the initial core alterations ( head lift). Reference Technical Specification 3/4. Time: Last Performed (N-31) Date:
(N-32) Date: Time:
26.0 ENSURE that OPSP03-FH-0001, " Refueling Machine Pull Test" has been perforced within 100 hours prior to using the ' refueling machine or hoist for the movement of thimble plugs, drive rods, or fuel assemblies within the reactor vessel (Technical Specification 3/4.9.6).
Last Performed Date: Time: l 27.0 ENSURE direct communications has been established between the control room operator and personnel at the refueling stations in the RCB and FRB within 1 hour prior to commencing core alterations (Technical Specification 3/4.9.5). Time: Last Performed Date: l l
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Qualifications and Conduct of Operators for -I Cranes. Hoists. and Monorail Systems OPGP03-ZO-0002 Re l Page II of 17
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NOTE , f,i voice communication (telephone, radio, etc.) is u I
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at all tunes. Special abnormal signals should be clearly agreed upon by th{ crane operator and should not conflict with standard signal j i 5.2.2.14 , t Activate installed warning devices prior to crane transit and intermittently when approaching personne l
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i 5.2.2.15 ; Follow only approved safe load paths per OPGP03-ZA-0069 (Cont of Heavy Loads).
i j 5.2.2.16 12.aa and release attached loads, place the crane controllers in i and set the parking brake (if applicable) before leaving any crane enumenmdarf i 5.2.2.17 i i Place all crane controllers in "OFF" following a loss of powe crane operation upon power restoratio ;
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5.2.3 Conduct - Gantry Cranes !
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i 5.2. ~ The Operator shall walk the rails prior to initial lifts to ensure the r! and path are clear and to become aware of other work activitie , the crane's path. Examples of initial lifts include the first lift byi particular Operator using a particular gantry crane, first lift after-i returning from break or lunch, et .2. i The Operator may require additional personnel to 'M walk the ra crane travel to ensure the rails and path are clear. The Signalman shall employ such personnel as required by the Operato e ; personnel are then responsible for identif Such A i notifying the Signalman in a timely manner. ying interferences and i Such personnel are not required to be operator or rigger certifie ! i
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5.2.4 Conduct - Reactor Containment Building Polar Cranes i
5.2. Warm up the crane controller circuitry (power breakers closed) for minutes before operating any control mechanisms. (SER 85-02) 5.2. Park only) the crane along a north-south axis when not in use. (Modes 1-4 5.2.5 Conduct - Outdoor Cranes (including mobile cranes) 5.2. Tic down or otherwise secure all outdoor cranes when not
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OPGP03-ZA-0010 Plant Procedure Adherence
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and Implementation and Rev 15 Independent Verification Page 4 of 22 +
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3.0 Responsibilities ' l The Plant Manager is responsible for establishing and maintaining i the station standards and expectations for procedure adherence and implementatio s The Department Manager assigned responsibility for the preparation
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3. Determining the procedure use control required as described j in steps 4.2.5 through 4.2.1 . Ensuring that provisions for verification are established ,
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within applicable procedures which require independent verificatio .3 All Supervisors assigned responsibility for performing a procedure r requiring independent verification are responsible for ensuring that only knowledgeable individuals are assigned to perform independent verifications and that independent verifications are performed in accordance with this procedur .+ l y 4.0 Procedure Adherence to Procedures , A A A * A A A A A A *A A A A A A A A A A A A A A A
* * CAUTION l *
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* * Failure to comply with procedural f * adherence requirements can result in *
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* systems. Strict compliance with *
* procedural adherence is the *
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* responsibility of all personnel * * performing activities governed or * * directed by plant procedure * * *AAAAAAAAAAAAAAAAAAAAAAA 4. Procedures SHALL be strictly adhered to when performing I plant activities. For implementation of new or revised procedures, the Plant Manager may allow deviations (such as early use of new forms, continued use of superseded forms, etc.). Permission for such deviations shall be prescribec ;
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OPGP03-ZA-0010 ; '
' () Plant Procedure Adherence and Implementation and Independent Verification Rev 15 Page 5 of 22 l
4. Anyone performing a , task under,the direction of an STP Procedure (Engineers, Operators, Technicians, Craftsmen, or Anyone) SHALL perform the steps of that procedure as written, unless such performance would violate the intent of the procedure as discussed in step 4. .
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4. When performance of a step in a procedures a) Will, or it is believed may, violate the intent of the procedure; or b) Could result in the plant / system being placed in a condition that is not consistent with good Maintenance, Engineering;-or-Operating practices; or c) Might result in a personnel or equipment hazard;
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SIEFAMD Inform supervision, management, and/or the control r,oo . When a procedure cannot be performed,as written the procedure SHALL be changed via Field Change or Revision
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prior to performance, except in an emergency (see f; step 4.1.9). Obvious typographical errors need not be '; corrected as long as the meaning of the erroneous entry is
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clear and the error does not affect component numbers, l acceptance criteria values, or other infonmation where l single character deviation could affect result . Night orders, standing orders, operating orders, special (' orders, letters, memoranda, etc. SHALL NOT be used to amend, revise, or delete an approved procedure, nor shall l they be used to violate an approved procedure. These documents shall only be used to give supplementary directions or orders relating to the conduct of operations (e.g. reduce power at 1200 hours in accordance with approved procedures and Technical Specifications).
4. Changes, revisions, and deletions to procedures and temporr:y procedures SHALL be made in accordance with OPGPO*-ZA-0002 (Plant Procedures).
4. For purposes of determining adherence, the terms used to distinguish a requirement from a recommendation or a granting of permission are explained in Addendum / I l
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l RO EXAM: QUESTION NO. 25 i SRO EXAM: QUESTION NO. 23 Given the following: !
; - Unit i has been operating at 80% for the last 48 hour No AFD penalty time has been accumulate The AFD monitor alarm was disabled ati 060 ; - Target AFD for 100% power is + 3.75%. - Indicated AFD has been continually monitored and logged as, j follows: :
I i TIME N41 N42 N43 N44 ; 0600 + + + + ;
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0622 + + + +3.9 l 0634 + + + + , l 0655 + + + +5.2 1 0700 + + + + t At 0700, WHICH ONE (1) of the following is the cumulative penalty deviation time for the previous hour? minutes, l f minute , minute i minute ! KEY ANSWER: B l t Response to Ouestion l i l We feel that the most conservative response to this question is 26 l l minutes. This is based on the following fact , l l l (a) The STP Core Operating Limits Report (COLR) for Unit 1, Cycle 5, states that the target band for AFD must be within +3% or -12% of the target value. Calculating 80% ) of the target value (3.75%) at 100% power gives a target value of 3. 0%. Therefore, the maximum target band limit is 6.0% with-reactor power at 80%. The STP Technical Specifications (3/4 2.1, Surveillance Requirement 4.2.1.2.a) states that one (1) minute of penalty deviation shall be accrued for each minute that AFD is outside the target band when rated thermal power is > 50%. Technical Specifications also states that when the AFD Monitor Alarm is inoperable, AFD must be logged once per hour for the first twenty-four (24) hours and at least once per thirty (30) minutes thereafte This Surveillance Requirement (4.2.1.1.a) continues to state I
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~l that the logged values of the -indicated AFD shall be assumed to exist durina the interval crecedina each i locoin (b) Based on the statement made in the Surveillance Requirement 4.2.1.1.a, we must assume that AFD exceeded the target band limit of 6.0% at 0634, even though the ,
log was recorded at 065 This "out-of-target-band" ! reading must be considered valid with the accumulation of ! penalty points until the next AFD reading is recorded., l with less than 2/4 channels reading outside the target ! l ban At 0700, only one (1) channel is indicating l greater than the maximum allowed target band value of 2 6.0% and the accumulation of penalty points is stoppe l Therefore, the maximum amount of time that AFD is outside i the target band is 26 minute j Based on this knowledge, we request that the answer for Question ;
#25 be changed to "C". !
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3/a.2 POWER DISTRIBUTION LIMITS TSI -002 4 3 /4. AXIAL FLUX DIFFERENCE tlHTTING CONDITION FOR OPERATION l 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target band (flux difference units) about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (COLR). . I APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* ACTION: l With the indicated AFD outside of the above required target band, [ and with THERMAL POWER: greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either: a) Restore the indicated AFD to within the target band , limits, or l b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.
l greater than or equal to 50%, but less than 90% of RATED THERMAL POWER: l
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a) POWER OPERATION may continue provided: 1) The indicated AFD has not been outside of the target i band for more than I hour cumulative penalty l deviation during the previous 24 hours, and i 2) The indicated AFD is within the Acceptable Operation limits specified in the COLR.
l ! Otherwise, reduce THERMAL POWER to less than 50% of RATED THEPyAL POWER within 30 minutes and reduce the Power Range Neutron Flux * - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hour b) Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1., provided that the indicated AFD is maintained within the Acceptable Operation Limits specified in the COLR. A total of 16 hours operation may be accumulated with the AFD outside of the Target Band during this testing without penalty deviatio "See Special Test Exceptions Specification 3.1 SOUTH TEXAS - UNITS 1 & 2 3/4 2-1 Unit 1 - Amendment No. 9, 27, 39 Unit 2 - Amendment No. I, T7, 30 l
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! ! POWER 015TRIBUTION LIMITS j QMITING CONDITION FOR OPERATION i ACTION (Continued)
- greater than 15%, but less than 50% of PATED THERMAL-POWER
l THERMAL POWER shall not be increased above 50% of PATED THERMAL . ' POWER unless the indicated AFD has not been outside of the i target band for more than I hour cumulative penalty deviation j during the previous 24 hour > .. ThldAL POWER shall nct be increased above 90% of RATED THERMAL l POWER unless the indicated AFD is within the target band, and the indicated AFD has not been outside of the target band for more than , I hour cumulative penalty deviation during the previous 24 hour SURVElltANCE REQUIREMENTS , 4.2. The indicated AFD shall be determined to be within its limits during .
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POWER OPEPATION above 15% of RATED THERMAL POWER by-t Monitoring the indicated AFD for each OPERABLE excore channel,at i least oz.ce per 7 days when the AFD Monitor Alarm is OPERABL '
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l Monitoring and logging the indicated AfD for each OPERABLE excore ' channel at least once per hour for the first 24 hours, and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall te assumed to exist during the interval preceding each loggin .2.1.2 The indicated AFD shall be considered outside of its target band when ! two or more OPEPABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band
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l shall be accumulated en a time basis of: One minute penalty oeviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and One-half minute penalty deviation for each I minute of POWER i OPEPATION outside of the target band at THERMAL POWFR levels between 15% and 50% of RATED THERFAL POWER.
' 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effectiva Full Power Days. The 4 provisions of Specification 4.0.4 are not applicable.
j 4.2.1.4 The target flux difference shall be updated at least once per 31 i Effective Full Power Days by either determining the target flux difference l
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SOUTH TEXAS - UNITS 1 & 2 3/4 2-2 Unit 1 - Amendment No. 77. 39 Unit 2 - Amendment No. 47, 30 l
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i POWER DISTRIBUTION LIMITS
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SURVEILLANCE REOUIREMENTS (Continued)
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pursuant to Specification 4.2.1.3 above or by linear interpolation between the 1 most recently measured value and the predicted value at the end of the cycle ' j life. The provisions of Specification 4.0.4 are not applicabl l l i
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South Texas Unit I Cyc!c 5 - STP r ,
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UNTT 1 CYCLE 5 : r i CORE OPERATING LIMEIS REPORT l i I
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l South Texas Unit I Cycle 5 i CORE OPERATING LIMITS REPORT I This Core Operating Umits Report for STPEGS Unit 1 Cycle 5 has been prepared in accordance l
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with the requirements of Technical Speci5 cation 6.9.L6. *Ihe core operating limits have been i developed using the NRC approved methodologies specified in Technical Specification 6.9. l
~Ihc Technical Specifications affected by this report are:
1) 3/4.L13 Moderator Temperature C# dent Limits l 2) 3/4.1 Shutdown Rod Insertion Limit 3) 3/4.1 Control Rod Insertion Limits 4) 3/4. AFD Limits 5) 3/4. Heat Flux Hot Channel Factor 6) 3/4.23 Nuclear Enthalpy Rise Hot Channel Factor
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2 OPERATING LIMTTS
~Ibe cycle-speci5c parameter limits for the specifications listed in Section LO are presented belo .1 MODERATOR TEMPERATURE CObi-HCIENT (Specification 3.a.13)
i i 2.1.1 The BOL, ARO, MTC shall be less positive than the limits shown in Figure .L2 'Ihe EOI, ARO, HFP, MTC shall be less negative than -4.8 x 10" Ak/k/ " l l l 2.13 The 300 ppm, ARO, HFP, MTCsha!! be less negath e than -3.9 x 10" Ak/k/ "F(300 ppm Surveillance Limit).
Where- BOL stands for Beginning of Cycle Life EOL stands for End of Cycle Life ARO stands for All Rods Out HFP stands for Hot Full Power (100% RATED THERMAL POWER) , ! , . _ _ _
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ROD INSERTION LIMITS (SpeciGcation 3.13.5 and 3.13.6) ! !
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2.2.1 The Control Rod Insertion limits are provided in Figure }
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2.2.2 Fully withdrawn for all Control and Shutdown Banks shall be 250 steps and above, but not Wing 259 steps withdraw t
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2.23 All banks shall have the same Full Out Position (EOP).
23 AXIAL FLUX DIFFERENCE (Specification 3.2.1) ; i
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23.1 AFD limits as acquired by Taahl SpeciScation 3.2.1 are determined by CAOC ; Operations with an AFD target band of +3,-12E , t
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23.2 The AFD shall be maintained within the ACCEPTABLE OPERATION portion of i
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Figure 3, as required by Technical Specification .
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I 2.4 REAT FLUX HOT CHANNEL FACIUR (Specification 3.2.2) 2.4.1 FRU q =2.5 !
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2.4.2 K(Z) is provided in Figure j
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2.43 The F ylimits for RATED THERMAL POWER (FRR) within speciOC Core planes shall be: : l I
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2.4 less than or equal to 1.90 for all core planes containing Bank *D' ! control rods, and i l l 2.4 less than or equal to the appropriate core height-dependent value from Table 1 for all unrodded core planes.
l 2.433 PFy =0.2.
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'Ihese Fylimits were used to confirm that the heat flux hot channel factor Fo(z) will be limited by Technical Specification 3.2.2 assuming the most limiting axial power distributions expected to result for the insertion and removal of Control Banks C and D during operation, including the accompanying variations in the axial xenon and po $
distributions, as described in WCAP 8385. Iherefore, these F,ylimits provide assurance that the initial conditions assumed in the LOCA anal3ds are met, along with the ECCS acceptance criteria of 10CFR5 < 2.5 EN1RALPY RISE HOT CHANNEL FACTOR (Speci5 cation 323) 2.5.1 FE=L4 .5.2 PFan=0 REFERENCES 3.1 CouaycadenceSerialNumberST-UB-HI 01176,IetterfromRobertC.Cobb(Westinghouse) to Dave Hoppes (HL&P), Core Operating Limits Report, Unit 1, Cyde 5 1106 9 . 3.2 NUREG.1346, Technical Specifications, South Texas Project Unit Nos.1 and 2.
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South Tcxas Unit 1 Cycle 5 i FIGURE 1 i
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MTC versus Power level
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-5 0 20 40 60 80 100 Relative Power (%)
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South Texas Unit I Cycle 5 ! FIGURE 2 Control Rod Insertion Lirnits Versus Power level (23.259):122 sep overtap U9.259): 122 step o<ersep (23.258): 121 step ovatap ps.2ss):121 st.p o%
(22,25Q: 119 Sep Ovedap 98.259: 119 Step overtap 260 ,
(213A' : 117 S m 0<*5ap - F7,254): t tr a, o,. .
(20.2sa:11sstepowwt p -
- ps.2sa:11ss po<w / (19,250):113 sep ovatap / Us.2sg:113 a.p ovatap
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I O 10 20 30 40 50 60 70 80 90 100 RElliTIVE POVEIl (%) ! l l i
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. Control Bank /s is aircady withdrawn to Full Out Position.
l i
. _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _
l
,
'
i South Tcxas Unit i Cycle 5
-
RGURE3 AFD Umits versus Rated Thermal Power l 120 I i 100 (-11.90) (11.90) , . I I
~
UNACCEPTABLE
~
UNACCEPTABLE
~
OPERATION OPERATION
~
/ \
/
/ h
{ 80 /
k m \ W / h 3: s ( 2 / \ : ' 4-I
/ '/ h ,
so ( a m / \ l
'h (-31,50)
I ACCEPTABLE _ _;
\_ (31,50) _ @ OPERATION F-k 40
4
0 l-50 -40 -30 -20 -10 0 10 20 30 40 50 Flux Difference (61) %
" . .
South Texas Unit 1 Cycle 5 FIGURE 4 K(Z) - Normalized Fo(Z) versus Core Height l l j
,
(7,1.0) 1 1 -L U
'
(13,0.936) --
~ ~ _ --- i
!
\ l
\
@ '
Ni O-n b
\ '
u_
-
u bt
' $ e (14, 0.68) __
E I ci-Z
$
M .2
0 2 4 6 8 10 12 14 16 Core Height (Ft) ~ .- A
_ _ . . - _
, , .-
s South Texas Unit 1 Cycle 5 TABLE 1 i Unrodded Fxy's vs. Core Height
'.
l Core Height Unrodded Core Height Untodded i
(Ft.) Fxy (Ft.) Fxy
'
0.000 2.993 7.200 1.841 , 0.200 2.542 7.400 L834 ! 0.400 2.091 7.600 1.829 ! 0.600 L869 7.800 1.825 ! 0E00 1.734 8.000 1.824 ;
'
l LOOO 1.663 8.200 1.824 I L200 L640 8.400 1.826 : L400 L633 8.600 L830 L600 L638 8.800 L836 L800 L656 9.000 1.843 i 2.000 1.678 9.200 1.849 : 2.200 L702 9.400 L861 F l 2.400 1.729 9.600 L874 l 2.600 L756 9.800 L888 i 2E00 L783 10.000 L899 l ! 3.000 'L809 10.200 1.907 l 3.200 124 10.400 1.912 ! 3.400 L852 10.600 1.921
'
3.600 1.846 f 10.800 1.926 . i 3.800 L837 ILOOO L926 I 4.00 .834 11.200 1.920 ' ! 4.200 L846 11.400 1.907 4.400 4.600 1E42 1236 11.600 L890 ( 11.800 1.866 l 4.800 1230 12.000 1E34 .
!
5.000 1.827 12.200 1.806 5.200 l 1.825 12.400 1.788 j 5.400 1.823 12.600 1.812 l j 5.600 1.825 12200 1E54 ' 5.800 1E27 13.000 1.913 6.000 1E31 13.200 1.906 6.200 1.837 13.400 1.969 6.400 1E44 13.600 2.112 6.600 1252 13.800 2.472 , 6.800 1.855 14.000 2.833 7.000 1.850 i n , Core Omrasine Umies Remn n r is j i
-
,_
}}