IR 05000324/1992301
| ML20127F385 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 01/06/1993 |
| From: | Ernstes M, Holbrook B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20127F347 | List: |
| References | |
| 50-324-92-301, NUDOCS 9301200225 | |
| Download: ML20127F385 (327) | |
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.o NUCLEAR REGULATORY COMMISSION REGION il-5 ..$ 101 M Af4tETTA ST"1E ET. ' [, ! ATL ANTA. G EORGI A 30323 s?******/ ENCLOSURE 1 EXAMINATION REPORT NO. 50-324/92-301 Facility Licensee: Carolina Power and Light Company facility Name: Brunswick Steam Electric Plant Facility Docket Nos.: 50-325 an'd 50-324 L Facility License Nos.: DPR-71 and DRP-62 Examinations were administered at the Brunswick Steam Electric Generating Plant near Southport, North Carolin Chief Examiner: W & de //6/97
_ Bobby + Holbr oo'k ' Dat6 tTgned V
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Approved By: It [. [ ////5J Michael E. Ernstes, Chief Date Signed Operator Licensing Section 2 * Operations Branch Division of Reactor Safety l SUMMARY-Scope: Written examinations and operating tests were administered to four Reactor Operator (RO) and nine Senior Reactor Operator (SRO) applicants durin the week of December 7.-1992. Additionally,.one operating crew was administered the simulator performance portion of a requalification retake examination.
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' Results: . All applicants, four R0s and nine Pb, passed these examination The operating crew and all individuals passef u i simulator portion of the
- requalification retake _ examinatio A strength;was noted in the quality of the examination prereview (para 4.b),
and with the' candidates' use of the Emergency Procedure Flow Charts-(para 4.c.4). A weakness was identifled.in'regards-to-the operators' hesitance . to insert a manual half scram when conditions warrant (para 4.c.2), and the l , ' SR0's physical posture which inhibited monitoring of the control room panels '
(para 4.c.3). t 9301200225 930108 PDR ADOCK 05000324 -.
_ REPORT DETAILS Persons Contacted Licensee Employees
*K. Ahern, Operations Manager, Unit 2 *H. Bradley, Manager, Nuclear Assessment *M. Brown,-General-Manager, Brunswick Steam Electric Plant, Unit 2 *W. Geise, Manager, Simulator Support *M. Jones, Manager, Training *D. Morgan, General Manager, Brunswick Steam Electric Plant, Unit 1 ' *B. Poulk, Manager, Licensed Training
- Examiners
*B. Holbrook, Chief Examiner, Region II *H Ernstes, Chief, Operator-Licensing Section 2 ,
D. Prawdzik, Examiner, INEL C. Tyner, Examiner, INEL i NRC Personnel Attending Exit
*P. Byron, NRC Resident Inspector * Attended exit interview Discussion Reference Material The NRC retains a set of Brunswick reference materials in the i regional office which is periodically updated'by the license However, when contract examiners are tasked with examination development, the licensee submits updated reference material for their use. Enclosure 1- of the 90-day letter, " Reference Material Requirements for Reactor / Senior Reactor Operator Licensing Examinations," contains a list that should be submitted to the _NRC for examination preparation. Although the training staff was very
, responsive in providing additional materials,' the following are examples of deficiencies identified pertcining to the reference material:
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The procedure index was not complet The radiation protection material was incomplete'but was
- updated with the second mailin Standing Instructions and Daily Instructions were sent with the second mailing.
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Report Details 2
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Simulator initial conditions, override capabilities, modeling deficiencies, and plant and simulator differences were sent with the second mailin Lesson plans for the Abnormal Operating Procedures (A0P) were sent with the second mailin Operating Instruction (01)-41, Operator Aids, was not sen Operations Study Material (OSM) for the Abnormal Operating Procedures was sent with the second mailin OSM 20-H is missing the Objectives page (page vi).
-- OSM 15-2A, " Primary Containment", was sent with the second mailin A0P, lesson Plan 07-M, was missing pages 25 - 3 Emergency Operating Procedure (E0P) for a reactor scram was sent with the second mailin Local Emergency Procedure (LEP)-04 was sent with the second mailin OSM Books 4 and 5, containing administrative topics, have not been revised since 1984 and 198 OSM-705 is not up to date with OG-01 and E&RC-0261 in regard to oxygen requirements to enter the drywel Administrative Instruction (AI)-58, " Equipment Clearance Procedure", Section 5.3.3.lb, states for a " Fail Closed" air operated valve to " tag the valve." The procedure does not give specific instructions on what actually should be tagge Many inconsistencies exist with the title of various shift positions. For example, OSM-702, calls the second SRO on shif t the " Shift Foreman" while 01-01, calls him the " Unit SRO." Also, the titles of the Shif t Operating Supervisor, Shift Supervisor, and Assistant Shift Operating Supervisor are frequently interchanged. This confusion also applies to the Operations Manager positions. These inconsistencies are evidenced by the number of incorrect responses on the written examinatio ____ __ - __ _ __ _ _ __ _ _ _ - - _ - -___-_ __ _ - - _ - _ _ - _ _ -
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Report Details 3 Written Examination The average grade for the R0 and SR0 written examinations were 8 percent and 87.1 percent respectfully. However, there were several questions missed by the majority of the candidates. The f allowing ' list indicates deficient knowledge in the areas examined by these question The question numbers are in [].
(1) Responsibilities / Conduct of Operations [R0-1]
Emergency Core Cooling System Isolation [R0-35] Rod Block Monitor functions [RO-48] Basis for Emergency Procedure actions [RO-82] -
(2) Responsibilities / Conduct of Operations [SRO-1]
Responsibilities During Emergencies [SR0-14] Maintain Containment Integrity [SRO-28] A strength was noted in the quality of the examination prerevie The facility members assigned to the prereview, which included two i training staff members and an active R0 and SR0 from the operations I staff, took a proactive approach to ensure the questions were valid, technically accurate, and worded using terms familiar to the candidates, There were no post-examination comments submitted by the licensee.
I i Operating Examination (1) Simulator Performance (Requalification Retake) The NRC administered a requalification retake examination, simulator performance portion, to members of operating crew
"D-1." The crew did not successfully pass the last NRC admin-istered'requalification examination that was conducted during the week of April 27, 1992. The licensee proposed scenario included evolutions and tasks for weaknesses that were identi--
fied during the last examination and topics that were covered during the remediation process. The crew and all individuals passed this examination. The crews' Communications Command and Control and Emergency Operating Procedure flow chart usage was noteworthy. .Due to the small sample size there was no program evaluatio s (2) Simulator Performance (Initial Examinations) There were several events during the simulator portion of the examination in which the Reactor Protection System (RPS) failed to initiate a reactor half scram when conditions warrante During one scenario, Average Power Range Monitor (APRM) channel
"C" failed high giving all the appropriate alarms, indications, and recorder responses to indicate a valid reactor half scram
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Report Detail should exist. The R0 candidate recognized the failure'to half scram and reported the failure to the crew and SRO. The RO-asked the SR0 if he should insert a manual half scram. The SR0's response was "no wait, I want-to check-T/S and 01-18."
The SR0 candidate took 15 minutes to review T/S and 01-18 prior to directing the crew to initiate a manual half scram. Another crew, under similar conditions, took six minutes to initiate a manual half scram. During follow-up questioning the candi'.,ates indicated that they should use 01-18 to verify (or prove) the instrument had in fact failed and to determine what action should be taken to comply with the T/S Limiting Condition for - Operation (LCO) for the failed instrument. The NRC's position is that the candidates should recognize the failure of RPS to initiate a reactor half scram, immediately verify all automatic actions have occurred, and manually initiate any action that' did not occur. This would include initiating a manual- half scram. During the time that the reactor half scram condition was not initiated, it was unknown if any scram condition would initiate a reactor scram when required. The Manager of License Training stated that similar instrument failures and events would be included in Licensed Operator Requalification (LOR) training, beginning in January 1993, to ensure the operators had a clear understanding of the expected actions. The
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training instructors will be briefed on the Operations ' Department expectations for this type of failure.
(3) The examination team noted that during the examination process, the SR0 candidates, while using reference material and making outside notifications, spent an-inordinate amount of time with their backs to the main control room panels. When events or conditions required the candidates to use reference material, the candidate walked to the front of the SR0 desk, stood with .
I his back to the control boards, and proceeded to use the refer- ! , ence material. During this time the candidate was not in a position to scan the control boards for predictable or abnormal trends, or to provide oversight of crew actions. Even though i there was no specific parameter-or trend identified as being missed, this posture does not lend itself to the crew concept, makes monitoring the control boards virtually impossible, and makes Communications Command and Control more difficul j (4) The simulator . scenarios provided an opportunity to evaluate the candidates' ability to recognize E0P entry conditions, execute various E0P flow charts, and_ transition within the various legs of the flow charts. The examination team noted that the can-didates skills, knowledge, and abilities in relation to the-various aspects of the E0Ps were very good.
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Report _ Details 5_
' (5) .Some candidates used reference material excessively when '
addressing Job Performance Measure (JPM) and administrative questions. One candidate took an. hour _to answer one JPM-question. Some candidates took excessive time paging through T/S and determining the appropriate LCOs. Others did not reference or apply the T/S interpretations. This indicates a lack of familiarity with come important topic . Performance of the Simulator There were c.everal simulator performance problems encountered during the-examination process. .While most of the problems did not hinder the examinations, one problem delayed the examination for up to 20 minute On three different occasions the simulator " locked up" and required a complete reboot. 1his required the exam team to stop the examination, remove the candidates from the simulator area, restart the simulator, and again start the examination. This delay did not appear to-affect the-crews' performance. While performing the High Pressure Coolant Injection (HPCI) System Performance Test (PT), it was determined that the HPCI valves would.not pass the PT acceptance criteria for stroke time. =Twice during the examination the' Rod Worth Minimizer " locked up" and would not - allow the candidates- to perform manual rod movements. 'For three simulator Initial Conditions (IC), the rod pattern was not correct'for that IC. This required the candidates to obta'in assistance from j personnel outside the control roo . Exit Meeting l At the conclusion of the site visit, the examiners met withfrepresenta-tives of_the plant staff,-indicated in paragraph 1 above, to discuss the results of the examination. The licensee did not identify!as! proprietary-any material provided to or reviewed b;y the-examiners.
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i ENCLOSURE 2 SIMULATOR FIDELITY REPORT Facility Licensee: Carolina Power and Light Company Facility Docket Nos.: 50-325 and 50-324 Operating Tests Administered On: December 7 - 11, 1992 This form is used only to report observations. These observations do not _- constitute, in and of themselves, 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 simulater facility other than to provide information which may be used in future evaluations. No licensee action is required solely in response to these observation During the conduct of the simulator portion of the operating tests, the following items were observed: Item . Description HPCI Operability Test (PT-09.2) Valves would not pass timing test Rod Worth Minimizer Locked up twice and prevented rod movement Simulator Locked up three times, required reboot Rod Pattern Was not correct for three Initial 5
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,NRC 00ESTION RESOLUTION There were no facility comment R0 Question 10 SR0 Question 11 l
Due to discrepancies between OSM-705 (dated September 16,1985) and the-current OG-01 and E&RC-0261, in regards to oxygen concentrations to enter the drywell, there is no correct answer to the question. NRC deleted the question from the examinations. There were no pass / fail decisions based on this questio . b' a
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. . . ... . NRC Official Use Only 912C AAASTE L Nuclear Regulatory Commission Operator Licensing Examination _ This document is removed from Official Use Only category on date of examinatio NRC Official Use Only
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! U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 2 CANDIDATE'S NAME: _ FACILITY: Brunswick 1 & 2 REACTOR TYPE: BWR-GE4 DATE ADMINISTERED: 92]12/07 INSTRUCTIONS TO CANDIDATE: Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passincy 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 %
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% TOTALS TINAL GRAliE All work done on this examination is my ow I have neither given nor received aid, cadardate's Signature l
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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 CHO1CE 023 a b c d 001 a b c d 024 a b c d 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
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.ANEWER SHEET Multiple' Choice (Circle-or X-your choice)
If-you change your answer, write your selection in'the blan .046 'a
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'047 a b c d 070 a b c- d 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 o b c d 074 a b c d 052 a b c d 075 a b c d _
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.A N:S WE R S H E'E T Multiple choice (circle or X your choice)
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Ir you change'your= answer, write'your selection in the blan a .b c
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098 a b c d 099 a b c d-100 a b c d
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- NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS - During the administration of this examination the following rules app?y:
1. Cheating on the examination means an automatic' denial of your application and could result in more severe penaltie . After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completin This must be done after you complete the examination. g the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all. contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner-of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer-sheet,- including any additional pages inserted when writing your answers on the-examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer to any short answer question . Partial credit may be given except on multiple choice questions. Therefore,- ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that-is provided mai count against yo For example,-if a question is worth one point and asks for four responses, each of which is worth 0.'25-points, and you give five responses, each of your responses will be worth 0.20 point If one of your five responses is incorrect, 0.20 will be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct. answer . If the intent of a question is unclear , ask questions of the examiner only.
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_ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ Page 6 i 14. When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets. In addition, turn in all scrap pape . Ensure all information you wish to have evaluated as part of your answer is on your answer shee Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your licence may be denied or revoke ._ e
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. QUESTION: 001 (1.00)
Which ONE of the following positions is the MINIMUM level of approval-required to transfer a safety-related item from Unit 2 to Unit I? , Plant General Manager I Manager - Operations Operations-Manager.- Unit 2 Unit 2 Shift Supervisor QUESTION: 002 (1.00) Which ONE of the following positions must approve securing the-Unit 1 Balance of Plant RO (BOP RO) with the unit shutdown? Plant General Manager Manager - Operations Operations Manager - Unit 1 d.- Unit 1-Shift Supervisor
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QUESTION: 003 (1.00) , As required by 10 CFR 26, " Fitness for Duty. Programs", wh ich ONE of _ the following is the MfNIMUM time an operator must abstain fram-the-consumptxon of-ascohol prior to any SCHEDULED shift? !- hours hours 4 hours
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. SENIOR REACTOR OPERATOR Page 8-l-QUESTION: 004 .(1.00)
Which ONE of the following annunciator window colors-designates a setpoint-important to reacter safety? Amber annunciator-window-with a red bar, Amber annunciator window with a blue ba Red annunciator window with a red bar, Red annunciator window with a blue ba ' QUESTION: 005 (1.00)
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A multiple input annunciator is in continuous alarm.due to one-invalid input
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The annunciator card has been pulled and the invalid-input signal defeated
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The annunciator card has been reinstalled Which ONE of the following colored dots should be placed on the annunciator window? ' Blue Yellow Red Black 1-
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SENIOR REACTOR OPERATOR l Page 9
- . QUESTION: 006 (1.00)
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motor-operated valve (MOV)-has been manually closed and torqued by-procedur Which ONE of-the following informs the operator that this MOV is in this r condition and must be untorquod prior to opening with the motor operator? A tracking LCO written for the valve, A caution tag placed on the RTGB.
4 A danger-tag placed on the RTG An entry made in the unit control Operator's Log.
I QUESTION: 007 (1.00) Which~ONE of the following radiation exposure guidelines is the MINIMUM . exposure at which the Shift Supervisor may waive the independent verification requirement for a system valve lineup? . Total-exposure expected during the independent verification is-in excess of 10 m > Total exposure expected during the independent verification is C ' in excess of 100 m General area radiation levels are in excess of 10 mr/h General area radiation levels are in excess Of 100 mr/hr.
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SENIOR REACTOR OPERATOR Page 10 QUESTION: 008 (1.00) 10 CFR 50.54(x) specifically allows " reasonable action that departs from a license condition or a technical specification in an emergency when this action is immediately needed to protect the public health and safety ...." Which ONE of the following approvals MUST be received prior to these actions being taken? These actions must be approved by: Plant General Manager _ NRC Senior Resident Inspector Manager - Operations Senior Control Operator QUESTION: 009 (1.00) The Control Room has received a report of a fire INSIDE the Protected Are Which ONE of the following REQUIRE the Shift Supervisor to call off-site fire department (s) for assistance? a. The fire is affecting access to, or the reliability of, safety-related plant equipmen _ b. The fire has NOT been extinguished within 10 minutes of the first repor c. The fire is in a contaminated area resulting in a 10 percent increase in airborne activity levels, The fire was NOT able to be immediately extinguished by the employee making the repor _ _ _
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Under which ONE of the-following sets of conditions may- a---FAIL CLOSED pneumatic valve be used as an isolation point for a clearance? I The valve is checked closed and the air supply isolation valve I to the valve is tagged close The air supply line to the valve is removed and tagged and the valve is tagged and double verified close The air supply isolation valve is tagged closed and the valve operator vent valve is tagged open.
' The valve has a gagging device or clamp installed and this H device is tagge l , l ESTION: 011 (1.00)
1 Unit \1 is making preparations for an drywell entr There are no self- l contai'ned breathing apparatus readily availabl Which Of of the following oxygen levels is the MINIMUM allowed for an entry underN these conditions? % ' % \ % p g/g[ /2_[ ( p L [d ' % 'N
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SENIOR REA'CTOR OPERATO Page 12 QUESTION: ~ 012 ( l'. 0 0) _ , LGiven the following conditions: ,
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Preventive maintenance is scheduled on the power supply for a - plant security compute A BSEP Equipment Clearance Form is being prepared and the-question answered "Yes
"Is p'lant- security ef fected?" on the -form has been
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Based on these conditions, which ONE of the following actions are required? The Security Shift Supervisor must be one of the two approval-signature The Security Shift Supervisor must be made aware of the work to be complete An independent verification must be done upon removing the power supply from servic A double verification must be performed upon restoring the power supply to servic QUESTION: 013 (1.00) Unit 2 is shutdown with around-the-clock fuel handling / core alterations in progres Which ONE of the following is the MAXIMUM time the fuel handling crew members are allowed to work during~a"conEinuous twenty-four (24) hour.
> period? .8-hours
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SENIOR REACTOR OPERATOR Page 13 QUESTION: 014 (1.00) Given the following conditions:
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A Loss of Coolant Accident has occurred on Unit 2
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A General Emergency has been declared
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The Emergency Operation Facility has been activated
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On-site and off-site iodine releases are occurring Which ONE of the following positions is responsible for the decision to administer Potassium Iodide to CP&L radiation worker personnel? Site Emergency Coordinator
~ The CP&L company designated physician Emergency Response Manager Radiological Control Director QUESTION: 015 (1.00)
Prior to a startup on Unit 1 with the plant cold, the operator adjusts the Control Rod Drive CRD) Pressure Control Valve (PCV) to maintain a 260 psid between Drive (Water Header pressure and reactor pressur Which ONE of the following describes how this pressure differential is maintained as reactor pressure increases during the ensuing startup? As reactor pressure increases during the startup: the operator will periodically adjust the Pressure Control ' Valve to maintain the required differential pressur the Flow Control Valve automatically opens to maintain constant flow, therefore a constant d/p across the PC the Pressure Control Valve automatically operates to maintain CRD system pressure above reactor pressure, the operator will periodically adjust the Flow Control Valve to maintain CRD system flow / pressure above reactor pressur _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - -
SENIOR REACTOR OPERATOR Page 14 QUESTION: 016 (1-00)
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Following a scram from 100% power the ball check valve in the insert port for one control rod drive mechanism malfunctions and fails to unseat and shift positio Which ONE of the following is the effect on that control rods' ability to scram? The control rod will: not insert until reactor and CRD pressures equaliz ~ hydraulically lock-up and remain stationar will fully insert at slower than normal spee will fully insert at higher than normal spee _ l _ -- - _ _
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.. . SENIOR REACTOR OPERATOR Page 15 QUESTION: 017 ( 1. 00) While at 100% power the CRD drive water pressure control valve is inadvertently close Which ONE of the following describes the system response to this valve closure? NOTE: P&ID (4 sheets) are included for referenc System flow decrease Charging pressure increase Drive water differential pressure increase Cooling water differential pressure decrease System flow decrease Charging pressure decrease Drive water differential pressure decrease Cooling water differencial pressure decrease System flow increase Charging pressure increase Drive water differential pressure increace cooling water differential pressure increase System flow increase Charging pressure decrease Drive water differential pressure decrease cooling water differential pressure increase _
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. QUESTION: 018 (1. 00 )- 1
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Which ONE of the following describes the operation of the auxiliary -I relay timer in the Reactor Manual Control System? -l The auxiliary relay timer: ; ' will halt the selected control rod witharawal upon a failure of the solid state master timer, controls rod sequence timing when rod withdrawai is demanded by the Rod Out Notch Override Switch.
' bypasses all interlocks except P.ud Wortn Minimizer insert l blocks during emergency control rod insertions, controls rod sequence timing when continuous rod insertion is i demanded by the Rod Movement Control Switc I
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Q!!ESTION: 019 (1.00) l Which ONE of the following is used to generate the Low Power Alarm Point l (LPAP) for the Rod Worth Minimizer? l l
' The main turbine first stage steam pressur The total steam flow signal from the Feedwater Control System, The total feedwater flow signal from the Feodwater Control l syste ; The highest reading Average Power Range Monitor input to the :
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flow converter .
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- QUESTION
- 020 (1.00)
Given the following plant conditions for Unit 1: ' Loop "A" Det pump flow - 45 millon Ibm /hr Loop "B" get pump flow - 5 millon Ibm /hr
"A" recirculation pump speed -
35%
"B" recirculation pump speed -
0% ' Loop "A" discharge valve - open Loop "B" discharge valve - open l l Which ONE of the following is the value for TOTAL CORE FLOW for these conditions? millon Ibm /hr millon Ibm /hr , millon lbm/hr b millon Ibm /hr
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l-l l QUESTION: 021 (1.00) l Brunswick Unit 2 Tech Specs regarding Recirculation System Jet Pump i OPERABILITY require a plant shutdown within 12 hours lf a jet pump is , found to be INOPERABL Which10NE of the followin is the concern for continued plant operation-with-an inoperable _(or fa led)-jet-pump? - Unbalanced neutron flux across the core due to flow variatioit Physical core damageEfrom a piece of a ' imaged jet pump, Invalid APRM Flow Biased SCRAM setpoints due to the change in flow through a failed jet pump.
( A reduction-in core reflood capabilities and increased blowdown , area during a Loss of Coolant Acciden .
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SElllOlt ilEACTOlt OPEllATOlt Page 18 ! l QUESTIOll: 022 (1.00) Which 011E of the following plant conditions will DECilEASE the AVAILABLE recirculation pump tiet Positive Suction llead (11PSil) ? Itecirculation pump speeds are simultaneously increased fron 2St to 40%. lleactor water level in stable and a high pressure feedwater heater isolate Itecirculation pump speed in at minimum and feedwater flow in increaned, Iteactor power in increaned irom 6St to BSt by withdrawing control rod QUEST 1011: 023 (1.00) A "Ileci rc 14G Speed Contral Signal railure" alarm han annunciated for the
"A" itecirculation pump on Unit Which OllE of the following in the expected effect on operation of the "A" Itecirculation pump?
The "A" Recirculation Pump speed: will run to mini.aum due to the low output signal from the controller, can only be changed by the individual pump controller in - manua will remain at its existing value until the scoop tubo lock can be rese will not change due to loco of power to the scoop tube positione _ _ _ _ _ _ - - _ _ _ _ - _ -
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SENIOR REACTOR OPERATOR Page 19 QUESTION: 024 (1.00) Given the High Pressure Coolant Injection (HPCI) system is in its normal standby lineu Which ONE of the following describes the position of the Minimum Flow Valve (F012) and the reason for that position? ! In the standby lineup, F012 ist open to provide the required system flowrate signal to the HPCI control syste closed to provide the required shutoff head for the HPCI. pump as it comes up to spee ; open to prevent pump overheating by providing an immediate flow ! path on startup, closed to prevent draining the Condensato Storage Tank to the suppression poo QUESTION: 025 (1.00)
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The Standby Liquid Control-(SLC) system monthly operability test on the Unit-1 "A" SLC pump is about to be performe Which ONE of-the following describes how the Reactor' Water Cleanup (RWCU) system isolation is avoided during this test? Starting the "A" SLC pump with the local pump control switch bypasses the RWCU isolation signa I The RWCU isolation signal is initiated from the.SLC squib valve firing circuit ~ The breaker for the appropriate RWCU isolation valve:will be opened prior to running the SLC pump, The RWCU system must be shutdown and the appropriate isolation-valves-closed before running the SLC pump, f B-
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SENIOR REACTOR OPERATOR Page 20 ;
! -QUESTION: 026 (1.00) .
The following conditions exist for unit The reactor has scrammed and the modo switch is in SHUTDOW The problem-has been identified and correcte , l Alarm "DISCil VOL llI LVL CRD TRIP" is actuate Which ONE of the following describes system response when the operator places the Scram Discharge Volume liigh Level scram Keylock switch to BYPASS, turns the scram reset switch to both directions and then places the mode switch to STARTUP? No cystem response for the present plant condition The scram will reset and remain reset, e The reactor /wYcRo ^Nll reset and again scram, The scram will reset when the scram discharge volume drain QUESTION 027 (1.00) , Which ONE of the following describes how use of the shorting links during special testing will affect the Reactor Protection System (RPS)? , Installation of the shorting links activates the SRM, IRM and APRM scrams in a coincident logic scheme,
, Installation of the shorting links activates the SRM scrams and L bypasses the IRM and APRM scram Removal of the shorting links activates the SRM scrams in a coincident (one-out-of-two-twice) logic scheme, Removal of the shorting links activates the SRM, IRM and APRM scrams in a non-coincident logic schem .,
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QUESTION: 02E (1.00) During normal plant operation the Unit 2 Traversing In-Core Probe (TIP) detector becomes stuck out of its in-shield positio Which ONE of the following describes how this will affect Primary Containment Integrity? Primary Containment Integrity: l a. can be established by closing the ball valve cutting the detector cabl I b. is not affected if the detector is stuck in a position outside the reactor vesse c. can be established by firing the TIP shear valv d. cannot be established until the detector is fully withdraw QUESTION: 029 (1.00) Which ONE of the conditions below is the point at which there are NO Source Range Monitor (SRM) control rod blocks preventing rod motion? a. IRM range switches are selected to Range 8 or abov b. SM1 reactor protection shorting links are remove c. Reactor Mode Switch is in NOT in "Run".
d. SRM power level is greater than 2.0 E5 counts.
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'l .0F F ai- if- 2 :RATOR Page 22 .
QUESTION: 030 (1.00) i
Given the following conditions for Unit 2:
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The "B" Recirculation Loop is isolated ; ,
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Reactor power is 22%
-- "A" Loop recirculation flow is 43% of rated *
Which ONE of the following is the Tech Spec APRM Plow Biased Simulated # ' Thermal Power Trip sotpoint for these conditions? (Choices are rounded to the nearest tenth.) . 1 .0 percent .4 percent .9 percent .5 percent _ QUESTION: 031 (1.00) Unit 1 is performing Average Power Range Monitoring CAPRM) calibrations and gain adjustment factor (GAF) adjustments as requ:. red by Tech Specs, , Which ONE of the following sets of conditions do not require adjustment? i APRM channel "B": indicated power is: 47% calculated power is: 46.5% , ' indicated power is: 76% calculated power is: 76.8% Indicated power is: 35% calculated power is: 35.4% indicated power is: 31% calculated power is: 34.1%
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SENIOR REACTOR OPERATOR Page 23 l i QUESTION: 032 (1.00) Which ONE of the following reactor vessel water level instruments will provide the most RELIABLE level indication during an accident resulting in high drywell temperatures? Narrow Range Instrument (N004A) Wide Range Instrument (N026B) Fuel Zone Instrument (NO37) Shutdown Range Instrument (N027A) QUESTION: 033 (1.00) Which ONE of the following reactor pressure conditions MAY cause
"NOTCllING" of the reactor water level instruments? Below 450 psig Between 500 psig and 625 psig Between 650 psig and 725 psig Above 775 psig
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_ - _ _ _ _ _ _ _ _ _ _ _ . SEllIOR REACTOR OPERATOR Page 24 ! QUESTION: 034 (_.00) Given the following Unit 2 plant conditions:
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Unit 2 has experienced a Loss of all AC Power
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The Reactor Core Isolation Cooling (RCIC) system started on a valid Level 2 signal
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An Auxiliary Operator reports to the control room that there is a large steam leak on the RCIC turbine
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The Shift Supervisor directs the Balance of Plant RO to isolate RCIC Which ONE of the following valves will CLOSE when the RO depresses the manual isolation pushbutton? _ a. F045 -- Turbine Steam Supply Valve b. F007 -- Steam Supply Inboard Isolation Valve F010 -- Condensate Storage Tank Suction Valve F029 -- Suppression Pool Suction Valve _ _~m._.-__-_m.._..-___.__ _ _ . _ _ _ . _ _
_ SENIOR REACTOR OPERATOR Page 25 I QUESTION: 035 (1.00) The following Unit 1 plant conditions are given:
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Reactor power is 65%
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Periodic Test 10.1.1 "RCIC System Operability Test - Flow Rates at 1000 PSIG" ks in procJress
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Suppression Pool temperature is 78 degrees F
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Suppression Pool temperature is increasing at 2 degrecs F every 12 minutes
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Suppression Pool cooling is in service Which ONE of the following is the MAXIMUM time the test may continue without violating Tech Specs? minutes minutes minutes minutes QUESTION: 036 (1.00) During a loss of coolant accident the Automatic Depressurization System (ADS) has initiated and is depressurizing the reacto Which ONE of the following conditions will close the ADS safety relief valves (SRV)? _ The SRV actuating air supply pressure is within 50 psi of - containment pressur Containment and reactor pressures are within 50 psi, Reactor pressure has decreased.to approximately 100 psi The SRV actuating nitrogen supply pressure has decreased below 100 psi _--_---______________-_-____mm______m_,_,,__._______ __.___ _ _ _ . _ , , _ . _ ,_
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SENIOR REACTOR OPERATOR Page 26 QUESTION: 037 (1.00) Which ONE of the following is indication that a Unit.2 Safety Relief Valve [SRV) vacuum breaker has failed in the open position.during SRV operat:.on? This failure will result in: direct pressurization of the suppression chamber air space each time the SRV is opene < steam bypassing the relief valve T-quenchers with a direct discharge path into the suppression poo an increased heat load on the drywell_ coolers each time the SR is opene suppression pool water being drawn up into the SRV discharge line after the SRV is closed.
QUESTION: 038 (1.00) Unit 1 has lost power to emergency bus E2 coincident with a loss of coolant acciden Which ONE of the following is the time delay from the Diesel Generator
#2 start signal to the "1B" Core Spray Pump starting? (Assume all systems function as designed.) seconds seconds seconds seconds -..- . . . . - . - . _ - ...m.--
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__ SENIOR REACTOR OPERATOR Page 27 QUESTION: 039 (1.00) Which ONE of the following describes the DIFFERENCE in the automatic start actions between the Unit 1 and Unit 2 Standby Gac Treatuent Systems (SBGT)? On a valid SBGT start signal: the Unit 1 SBGT train inlet and outlet valves do not automatically ope the Unit 1 purge system exhaust fans must be stopped by the operato ~ the Unit 2 purge system exhaust outlet valves do not automatically clos the Unit 2 SBGT train inboard and outboard ventilation dampers automatically clos QUESTION: 040 (1.00) Given the following conditions on Unit 2:
-- Unit 2 has had a large Recirculation Suction line break --
All Emargency Core Cooling Systems (ECCS) are functioning as designe All low pressure ECCS pumps are currently running to reflood the core to two-thirds heigh Which ONE of the following is the number of low pressure ECCS pumps required to MAINTAIN level at two-thirds height once established? - four three two one
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SENIOR REACTOR OPERATOR Page 28 l I QUESTION: 041 (1.00) Which ONE of the following, pressures is the point at which the Residual Heat Removal System operating in the Low Pressure Coolant Injection (LPCI) mode will begin to inject water into the reactor recirculation system? LPCI injection will begin at approximately: psig psig psig __ psig QUESTION: 042 (1.00) Which ONE of the following prevents a loss of main condenser vacuum when the Reactor Water Cleanup (RWCU) system is lined up for reactor reject operations? a. Administrative controls on the operation of RWCU Reject To Cor.deneer Valve (F034) and RWCU Reject to Radwaste Valve (F035). Interlocks preventing the simultaneous opening of RWCU Reject To Condenser Valve (F034) and RWCU Reject to Radwaste Valve (F035).
c. The RWCU system high differential flow automatic system - isolatio The automatic closing features of RWCU Reject Flow Control Valve (F033).
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SEH10R REACTOR OPERATOR Page 29 I
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QUESTION: 043 (3.00)
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Which ONE oftothe {' lilock Monitor ho following OPERAllLE p?lant conditic'io REQUIRE the Rod _, 110TE: Soo attached reference materini and P !
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, Reactor power -- 96% i
! Ronctor power -- 921 ' Reactor power --
89% Reactor power -- 100% f
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QUESTION: 044 (1.09) Which ONE of the following describen the function of the " Gain chango Circuit" in the Rod Illock Monitor (RBM) syste The Gain' Change Circuit: , initiates a rod block if the LPRM average i nput 10 higher than the APRM reference signal, provents rod movement until the RitM output in adjusted to equal the APRM referenco signal. _ t Initiates a " null acquenco" if the difforence between the-local average power and the APRM referenco nignal 10 too larg modifies the R11M ampliflor output-if the average of tho.LPRM inputa la lower than the APRM reference algna !
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QUESTION: 045 (1.00) Which ONE of the following is a properly or.iontated fuel bundle? , The channel spacer buttons are facing towards the control rod , blade passage are ; The identification boss on the ball handle is pointing away from the control rod blad ,
, The bail handle serial number is readable-from the outside edge of bundle in towards the control rod, The channel fastener is located on the outsido ec11 edge 180 degrees away from fuel cell cente QUESTION: 046 (1.00)
Which ONE of the following conuitions is the fuel tag board used as the PRIMARY reference for core fuel location during a refueling outage? The primary reference is the: I refueling floor tag board which is updated prior to the end of each shif refueling floor tag board which is updated after each core alteration, Control Room tag board which is updated prior to the end of each shif Control Room tag board which is updated after each core alteratio ,
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l . SENIOR REACTOR OPERATOR Page 31 QUESTION: 047 (1.00) " Which ONE of the following describes how the reactor core design ensures-that each fuel bundle receives adequate cooling flow? Adequate fuel bundle cooling is assured by: a, the fuel support pieces installed on the core peripheral fuel bundles to divert flow away from the outer bundle by placing the highest power bundles in the main discharge flow from the recirculation system jet pumps, minimizin' ' del bundle two phase flow resistance by limiting-the overala bundle Linear Heat Generation Rate ( LHGR) . bottom core-plate design which diverts flow to the cores centrally located highest power bundle QUESTION: 048 (1. 00) The following conditions exist for unit Reactor power 10 % MSIV 1821-F022C is stuck at 25 % ope Which ONE of the following MSIVs, if closed will cause a Half Scram signal? B21-F022D B21-F028C
._ B21-F022B F028B'
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SENIOR REACTolt OPEltAToll Pago 32
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i-QUESTIO!1: 049 (1.00) Which ONE of the following combinations of Main Steam Line Itadiation t Monitor tripa will cauno a PULL Main Steam Inolation Valvo (MSIV) closuro? (Do not consider any other actions from thone tripn.)
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? Channel "B" --
Downucale Channel "C" -- Inop Channel "C" -- Inop channel "D" -- liiyh - liigh Channel "A" -- liigh Channel "D" -- liigh - liigh Channel "A" -- liigh - liigh Channel "C" -- liigh - liigh OUESTION: 050 (1.00)
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During IIP turbino shall warming, the operator la cautioned not to'excood 155 poig first otage preneur Which ONE of the following is the result of excoading this limit? First etage preasurou above thin value: may reach the notpoint that removen the Group Notch control notch rentraint 1 imitations.- will cause the main turbino to roll off the turning year without receiving a propor warmu may heat the LP turbino exhaust hoods to the polnt where hood opray automatically initiaton cooling, Will place the plant cloan to the notpoint which_ arms the i Turbino Trip acram circuit.- ..5 - _. . _ _ ._.... _ _ _ _. - .. _ .. . 2 -
SENIOR REACTOR OPERATOR Page 33
.QUESTIONt 051 (1.00)
Which ONE of the following is the MAXIMUM power allowed during steady state, single loop operation following the trip of a Recirculation Pump? % % % %
, QUESTION: 052 (1.00)
Which ONE of the following is the MINIMUM total core flow at which the ' plant can operate and still be assured of avoiding power-oscillations or instabilities? million Ibm /hr million Ibm /hr . million Ibm /hr million Ibm /hr
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> i SENIOR REACTOR OPERATOR Page 34 _ i
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QUESTION: 053 (1.00)
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Which ONE of the following dr well leakage conditions REQUIRE the Shift . Supervisor to CONSIDER direct ng an immediate manual reactor scram? The drywell floor and equipment sump pumps are running > continuously, and capacity has been exceede The leak is from the recirculation pump seals and is above Technical Specification-limit ~ ~ The unidentified leak rate has increased by 3 gpm in the 1ast : 18 hours time perio The total leak rate is 30 gpm_as averaged over the last 24 hours time perio .
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QUESTION: 054 (1.00) I Which ONE of the following is the basis for requiring a reactor scram with no Control Rod Drive (CRD) pumps running during a plant startup? The reactor is scrammed: in anticipation of tripping the Recirculation-Pumps due to potential pump seal-damage from the loss of CRD flo to ensure there is no loss of scram capability as the scram accumulators depressuriz , to ensure all rods are fully inserted before overheating can affect the mechanism seals and impact scram times.
' due to the loss of normal reactivity' control and an alternate means of SLC injectio F
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_ _ _ __ . _ . _ _ _ _ _ - _ . . _ . _ _ -. - -_ _w_.-. _ _ _ _ _ _ _ . - ~ . - _ _ _ _ _ _ . _ _ _ SENIOR REACTOR OPERATOR Page 35-s l QUESTION: 055 (1.00) With 2 hours remaining in the shift it has been-determined that average I core megawatts thermal (CMWT) for the previous hour was 2446 CMW ; Which ONE of the following is the REQUIRED action? Power must be reduced to, or below, 2426 CMWT for at least one hou > Power must be reduced to, or below, 2426 CMWT for the remainder of the shif Reduce power to less than 2436 CMWT for the same amount of time it was above that valu , Reduce power to less than 2436 CMWT and ensure it does not exceed that limit for the remainder of the shif QUESTION: 056 (1.00)
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Which ONE of the following defines " Adequate Core Cooling"? Plant systems are removing sufficient heat from the core such that no Linear Heat Generation Rate (LHGR) limits are being exceede Plant systems are available and operating to remove long term decay heat under all possible combinations of Loss of Electrical Powe , The heat removal from the reactor is sufficient to restore'and maintain peak fuel cladding temperature below the point of failur Plant systems are available and operating to remove the sensible and decay heat during a postulated Losc of Coolant Accident.
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SENIOR RFACTOR OPERATOR Page 36 , F L QUESTION: 057 (1.00) Given the following plant conditions on Unit 2:
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Refueling nperations are in progress
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A sport fuel bundle is dropped and is heavily damaged
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PROCEES RX BLOG VENT RAD HI-HI annunciator is alarming . Which ONE of the fallowing is NOT an immediate operator action for the above conditions? Stop all tuel movenant Evacuate the Reactor 5411 ding Verify isolation cf Reactor Building ventilation Evacuate the Refueling floc QUESTION: 058 (1.00)
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Given the following conditiores for Unit 1:
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The plant has just experienced a complete loss of all means of Shutdown Cooling
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Temperature readings indicate a 1.5 degree F INCREASE'in bulk water temperature every 12 minutes
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Assume the reactor vessel head is installed
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No other parameters change
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Current temperature is 166 degrees F > Which ONE of the following is the time-allowed before primary containment integrity MUST be established? minutes
- minutes minutes minutes
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EENIOR REACTOR OPERATOR Page 37 l l l l QUESTION: 059 (1.00) j i
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Following a loss of power the Unit 1 " Torus to Reactor Duilding Vacuum-Breakers" must continue to receive nitrogen for operatio Which ONE of the following describes how the system is operated to accomplish this? The vacuum breaker accumulators will supply nitrogen for all postulated required operation The Abnormal Operating Procedure directs immediate manual ; opening of nitrogen supply valves, The solenoid valve supplying backup nitrogen to these valves fails open on loss of powe The backup nitrogen supply is unisolated by a check valve that opens as PNS pressure decrease QUESTION: 060 (1.00)
-Unit 2 is rapidly losing pressure in both the Service Air and Instrument Air systems.
, Which ONE of the following actions WILL be taken to cross-connect from the Unit 1 Service Air system? The Unit 2 Shift Supervisor will direct the cross-connect valve to be opened as soon as service air automatically isolates, b.- The cross-connect--will be-opened if-it will not affect Unit 1 air systems and the Unit 1 Senior Control Operator approve Either Unit 2 on-shift Senior Reactor-Operator may approve opening the cross-connect valve in order to prevent a Unit 2 scra If approved by the Unit 1 Plant Monitor RO, the cross-connect will be opened upon receiving the service air low pressure alar . - _ _ . ~ . - _ _ . _ . _ _ _ _ _ - _ - - _ . _ _ _ . _ _ _ . , . . - ._ _ . . _ . _ _ _ _ . --
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' QUESTION: 061 (1.00) , Which ONE of the following personnel becomes the Site Emergency coordinator during a control Room evacuation AND where will this person be located? a. Unit 2 Shift Supervisor at the Unit 1 Remote Shutdown Panel b. Unit 2 Shift Supervisor at the Unit 2 Remote Shutdown Panel c. Unit 1 Shift Supervisor at the Unit 1 Remote Shutdown Panel d. Unit 1 Shift Supervisor at the Unit 2 Remote Shutdown Panel QUESTION: 062 (1.00) Which ONE of the following describes when the Unit 1 Mode switch is REQUIRED to be placed in " Shutdown" during a control Room evacuation? The Unit 1 Mode switch is placed in " Shutdown": a. immediately prior to tripping the recirculation pumps b. aftor total steam-flow decreases to less than 3.0 E6 lbm/h c. after reactor pressure decreases to less than-700 psi d. immediately prior to clos!.ng the Main Steam Isolation Valves.
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SENIOR REl.CTOR OPERATOR Page 39 QUSSTION: 063 (1.00)
-Unit 2 is experiencing lowering main condenser vacuu Which ONE of the following conditions allow the operator to JNCREASE-power in an attempt to correct the lowering vacuum problem?
Increasing power to correct the vacuum problem may be used if: main condenser hotwell level had decreased to less than -7 inches and is now recoverin turbine exhaust hood sprays had been. operating to cool the Lp turbine hoods at low power level the lowering vacuum occurred simultaneously with an operator initiated generator load reductio the SJAE Condensate Recirculation Valve had been throttled open to improve air ejector efficienc QUESTION: 064 (1.00) Unit 1 main condenser vacuum is lowering. The-Shift Supervisor is directing operations in accordance with AOP-3 Which ONE of the following vacuum readings corresponds to the lowest value at which Feedwater Pumps will be able to maintain reactor water level? [>/g g y pT ;r73 inches Hg vacuum inches Hg vacuum inches Hg vacuum .4 inches Hg vacuum -- 1 u _ ._ ~ . . _ . . - _ - - - . _ _ . . . . - _ _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ - _ __ -
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SENIOR REACTOR OPERATOR Page 40 i
; -QUESTION: 065 (1.00)
Which ONE of the following is the reason why AOP-22.0, " Low System i Prequency", directe the operator to keep the unit on the line and , increase output, if possible, when system frequency is lowering? 1 If the unit is taken'off line with these conditions present, the result may be a loss of off-site powe .) I Increasing output will raise system frequency minimizing plant- , equipment damage from operation at less than 60-H i Maintaining the Brunswick units on line will be the most I reliabic means of stabilizing system frequenc The North and South Carolina portions of the grid do not carry . a rolling reserve capable of restoring system frequenc , t QUESTION: 066 (1.00)
_ Which ONE of the following is the reason a reverse power trip is not preferred means to take the main turbine off-line? Allowing an automatic reverse power trip of the main turbine: will result in excessive arcing'in the main generator output I breaker may cause a pressure spike sufficient to rupture the LP turbine ' relief diaphragm will result in an automatic cold start of that unit's diesel generator:,. may place an unnecessary load on the main turbine thrust bearin I s
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- SENIOR REACTOR OPERATOR Page 41 QUESTION: 067 (1.00)
Following a reactor scram, E0P-01, " Reactor Vessel Control Procedure", has been entered. An IMMEDIATE determination must be made regarding any required transition to the " Level / Power Control" flowchar Which ONE of the following critoria, other than all rods fully inserted, is used to make the determination that the reactor will remain shutdown under all conditions without boron? A Nuclear Engineer calculation of shutdown margi All rods jnserted to or beyond to position 0 All APRM channel "Downscale" alarms i ' All IRM channels less than 25 on Range 3
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QUESTION: 068 (1.00)
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Given the following plant conditions on Unit 2: -
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The reactor has scrammed >
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The Main Steam Isolation Valves have closed ,
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Continuous safety relief valve pneumatic supply is NOT available
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Reactor cooldown is not required at this time Which ONE of the following is the guidance for safety relief valva (SRV) operation for these conditions? The SRVs should be allowed to control pressure automatically in the safety mod . The SRVs should be opaned in sequence for no longer than 3 minutes each time, The SRV opened manually for pressure control should lu3
. substained opening The non-ADS SRVs should be used in sequence for pressure control.
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_ . . . . __ _ _. _. _ _ _ ...-._. _ _ _. _ .__ _ _ _ _ _ -.._.. .- _ _ - _ - _ . - _ _ _ . _ _ . . _ _ _ SENIOR REACTOR OPERATOR page 4 QUESTION: 069 (1.00) EoP-02, " Primary Containment Control Procedure", primary containment pressure control path, directs the primary containment to be vente Which-ONE of the following is the result of venting via the suppression chamber instead of the drywell?
?
Using the suppression chamber as the vent path will: [ i ~ allow better control of the release rate due to the sizing of the path's piping and valves, allow a more rapid decrease in suppression chamber pressure to below the Pressure suppression Pressure curv help absorb-some of the high energy in the primary containment by passing it through the suppression pool water, reduce the-levels of radioactivity. released as it passes through-the water in the suppression poo , t QUESTION: 070 (1.00) Which ONE of the following is the basis for securing suppression pool and drywell sprays when pressure has decreased to 2.5 psig? Continuing sprays below this pressure will adversely affect the net positive suction head of-the pump .This pressure ensures that downcomer/ ring header joint stresses do not reach the point of complete failure.- Maintaining a positive pressure ensures air-will not be drawn ;
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into, and de-inerting, the primary containmen Securing sprays at this point allows redirecting the Residual Heat Removal pumps to assure adequate core-coolin , l
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SE!1IOR. REACTOR OPERATOR Page 43 QUESTIO!1: 071 (1.00) With Unit 2 at power a spurious Group 1 isolation signal occur ALL Group 1 isolation valves clos Which O!JE of the following signals was the cource of the isolation? Main steam line high radiation Main steam line high flow J Main condenser low vacuum Turbine building high area temperature QUESTIOll: 072 (1.00) While operating in EOP-02, " Primary Containment Control-Procedure", the operator.is directed to perform an emergency depressurization when plant conditions "cannot be restored and maintained in the ' Safe' region of the SRV Tail Pipe Level Limit."
Which Ol1E of the following plant conditions must be evaluated along with Suppression Pool level in order to make the decision to depressurize? Reactor-pressure Suppression pool temperature > Suppression chamber pressure Delta T he l' i , F l l F _ _
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SENIOR REACTOR OPERATOR Page 44 l l
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h QUESTION: 073 (1.00) Given Table 2 from EOP-1, " Reactor Flooding Procedure" and the Condensate and Condensate Booster Pumps running and slowly injecting into the RPV: Number of Pressure Open SRVs (psig) ___...... ________ . 7 100 Table 2 -- 6 115 Minimum Alternate 5 145 Flooding Pressure 4 180 # 3 245 - 2 375 1 765
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Which ONE of the following plant conditions will CONFIRM adequate core cooling? t SRVs open RPV pressure is 100 psig SRVs open RPV-pressure is 325 psig SRVs open .
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RPV pressure is 300 psig SRV open RPV pressure is 760 psig
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QUESTION: 074 (1.00) Which ONE of the following is the MINIMUM reactor vessel water level-that will confirm adequate core cooling WITH injection systems operating? . +45.0 inches
- .0 inches - .5 inches - .O inches . ._ - . - - _ . _ -._ _ . _ . . , J., . . _ . _ . _ . _ . _ . . . . - - . _._u:--..-,_..-.,..-. _..__ _ ._. _ _ _ ._ _ _ _ _ __ .__ _ . _ _ _ _ - . _ _ _ . _ _ _ - . . . _ _ ,
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' SENIOR REACTOR OPERATOR Page 45 QUESTION: 075 (1.00)
K1 5 ONE of the following is the basis for operating within the limits cs "Jae Maximum Core Uncovery Time Limit Graph? The Maximum Core Uncovery Time Limit curve: is the time water level can be below bottom of active fuel with no cooling and clad temperature will not exceed 1500 degrees is the time limitation imposed for a level indication response once all injection is stopped and water level is lowered, determines how long RPV pressure must be maintained above 50 psig with Safety Roljef Valves open during RPV Floodin provides time to complete RPV-fl 7 ding prior to exiting this
- procedure and entering Primary Containment Flooding.
, l QUESTION: 076 (1.00) l l Which ONE of the following is the justification for limiting primary containment pressure to 70 psig? l As containment pressure approaches 70 psig: the primary containment vent valves' seat leakage exceeds Tech . Spec allowable levels compromising containment integrity.
l l the stroking speed of the primary containment vent valves-will' i increase to the point of causing valve damage during operation, there is inadequate differential pressure across the safety relief valves to provide an acceptable-flowrate during steam coolin the differential. pressure across the safety relief valve pilot valves decreasas1to the point where the valves may not operate.
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SENIOR REACTOR OPERATOR Pa7e 46 QUESTION: 077 (1.00) Which ONE of the following alternate rod insertion methods REQUIRE the Reactor Protection System to be reset? Control rod insertion by: venting the drive mechanism over-piston are venting the scram air heade using the Reactor Manual Control Syste increasing cooling water header pressur QUESTION: 078 (1.00) Which ONE of the following indications is used to monitor plant heatup/cooldown rate while in Alternate Shutdown Cooling? Recirculation loop suction line temperatur Steam dome pressure using steam table conversion Safety relief valve tailpipe temperatur The running ECCS pump local suction temperatur _
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- . ' .SElilOR REACTOR' OPERATO Page 47- .QUESTIOti: 079 (1.00)
The temperature control leg of EOP-03, " Secondary Containment Control
' Procedure", asks if " a primary system is discharging into the Reactor Building".
Which ONE of the following is a Primary System as referenced in this step? , A Primary System is: any plant safety-related system required to be operable in Modes 1, 2 and/or a system whose leak rate will-decrease as reactor pressure decreases, a system required to shutdown the reactor.or provide long-term core cooling, any plant system whose pressure will change with drywell and/o torus pressur _ _ ._-- _ _
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d SENIOR REACTOR OPERATOR Page-48
. QUESTION: 080 (1.00)
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. Given the following plant conditions on Unit 2:
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Loss of coolant accident in progress
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A Reactor Building entry is needed to complete safe shutdown actions
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Radiation level in the specific area does not allow personnel access to that area Which ONE of the following describe the Reactor Building radiation levels? Reactor Building radiation levels are: greater than the level necessitating a Site Area Emergency, beyond the levels required for an emergency depressurizatio above the maximum acceptab?.e operating limi in excess of the maximum safe operating valu QUESTION: 081 (1.00)
"
Which ONE of the following sets of conditions REQUIRE entry into EOP-02,.
" Primary Containment Control Procedure"? Primary containment oxygen concentration is 4.45% and hydrogen '
concentration is-1.25%. During the High Pressure Coolant Injection system operability.
, test suppression pool water level reaches -28 inche During the Reactor Core Isolation Cooling operability test suppression pool temperature reaches 107 degrees F.
, Malfunctioning drywell coolers require drywell venting.to maintain pressure steady at 1.65 psig.
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' SENIOR' REACTOR OPERATOR Page 49
- QUESTION: 082 (1.00)
Unit 1-is performing a startup following a reactor-scra Reactor Water Cleanup (RWCU) is lined up to reject-to the condenser to lower RPV water leve Which ONE of the'following parameters. limits the RWCU reject flow ' rate for these plant conditions? Nonregenerative heat exchanger outlet temperature, Rcgenerative heat exchanger outlet temperatur Reject Flow Control Valve downstream _ pressur . Filter-demineralizer differential pressur QUESTION: 083 (1.00) The Hot Shutdown Boron Weight (HSBW) is the amount of boron-that will maintain the reactor shutdown under hot standby conditions with the following assumptions:
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All rods are withdrawn to the maximum rod block limit
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No Xenon in the core
--
No voids-in the core
--
Reactor water temperature at saturation temperature for lowes SRV lift setpoint Which-ONE of the following.is the reactor water level assumed for HSBW? _
- /3 core heigh The Top _of. Active Fuel (TAF).
~
. ls . The low waterLlevel scramLsetpoin The high level trip setpoin .
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SENIOR REACTOR OPERATOR Page 50 QUESTION: 084 (1.00) A failure to scram (ATWS) has occurred and reactor power is approximately 35%. Which ONE of the following actions should be taken to prevent a main turbine trip on high reactor water level for these plant conditions?- a. Place the Recirculation Pump MG Set control switches to "Stop".
b. Place both Recirculation Pump speed controllers to 10%. c. Place both Recirculation Pump speed controllers to 28%. d. Place both Recirculation Pump speed controllers to 45%. QUESTION: 085 (1.00) Given the following conditions for Unit 2:
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A failure-to-scram condition (ATWS) exists '
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Reactor power is 45%
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The Main Steam Isolation Valves are shut
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Core cooling is adequate Which ONE of the following plant components is being most severely challenged for these conditions? a. Fuel cladding b. Reactor vessel c. Primary containment d. Injection systems
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SEN10R REACTOR OPERATOR Page 51
I l QUESTlON: 086 (1.00) EOP-02, " Primary Containment Control Procedure", Step SP/L-36 directs securing "HPCI 1rrespective of adequate core cooling ' if suppression pool level cannot be maintained above -6.5 fee Which ONE of the following describes the affect of continued operation of the High Pressure Coolant Injection (HPCI) system? Continued operation: will cause HPCI turbine damage due to the high temperature cooling water to the oil syste __ will cause damage to the turbine exhaust check valve due to high steam flow, will reduce suppression pool level below the safety relief valve T-quencher result in the failure and loss of the final fission product barrie QUESTION: 087 (1.00) Given the following plant conditions on Unit 2:
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A high energy break has just occurred in the Reactor Building
--
Reactor Building area radiation monitors are alarming
--
Reactor Building ventilation exhaust rad level is 6 mr/hr
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The Shift Supervisor has entered EOP-03, " Secondary containment - Control" and AOP-0 Which ONE of the following is NOT an immediate operator action for these conditions? Verify Reactor Building HVAC in operation, Within 15 minutes isolate the Reactor Building sprinkler syste Control access to the affected area, Ensure E&RC survey and post the area __ _ _______-_____--_ -_ - _ _ _
SENIOR REACTOR OPERATOR Page 52 QUESTION: 088 (1.00) Which ONE of the following is the maximum load allowed on a diesel generator while operating with the emergency buses cross-tied between the units during a loss of off site power? KW KW KW KW QUESTION: 089 (1.00) In accordance with Unit 2 Tech Specs, the reactor was depressurized due to suppression pool water temperature being greater than 120 degrees Which ONE of the following limitation is in effect? NOTE: Reference material attached Suppression pool temperature must be: less than 95 degrees F before changing Operational Condition less than 105 degrees F prior to opening the MSIV less than 110 degrees F prior to exceeding 1% powe less than 120 degrees F for 24 hour _ _ _ _ _ . . _ . _ ___ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .___
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, , - SENIOR REACTOR OPERATOR Page 53 QUESTION: 090 (1.00) l l Given the following plant conditions on Unit 2:
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Power is 45%
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The "A" Reactor Feed Pump is running j -- UPS power to the feed control system is lost Which ONE of the following describes how the Reactor Feed Pump will be affected? The Motor Gear Unit increases feed pump speed to approximately 6000 rp The Motor Gear Unit reduces feed pump speed to approximately 2800 rp The Motor Speed Changer will increase feed pump speed to about 6000 rp The Motor Speed Changer will reduce feed pump speed to about 2800 rp QUESTION: 091 (1.00) Ten (10) valves in a Tech Spec system are undergoing timed stroke testin Which ONE of the following actions must be taken if the one of the valves exceeds in maximum allowed stroke time? _ Stroke the valve a second time to provide confirmation of the - stroke tim Immediately declare the valve inoperable and generate the required LCO Exercise the valve through several stroke cycles then stroke once more for time, Log the stroke time, complete all valve testing and then take required corrective action _ _ _ _ _ _ _ _ _ _ _ _ _ - _
,, . _ . - - . . _ . . ~ SENIOR REACTOR OPERATOR Page 54 QUESTION: 092 (1.00) . .
With the plant at 65% power a failure in the "B" Recirculation Pump controls causes a slow speed increas Which ONE of the following describes when the "B" Recirculation Pump is required to be tripped by the AOP for the above conditions?
-The "B" Recirculation Pump will be tripped: if its speed _ continues to increase after the' scoop' tube is unlocke as soon as the speed difference between the Recirculation Pumps-exceeds 15%. if more than two APRM or IRM Upscale alarms are received simultane:usl if lockin increas the scoop tube on the pump does not stop the speed . QUESTION: 093 (1.00)
Which ONE of the following is NOT required in order for the High Pressure Coolant Injection (HPCI) system to be OPERABLE? Auxiliary Oil Pump Condensate Storage Tank . l Minimum Flow Valve (F012) Suppression Pool i
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SENIOR REACTOR OPERATOR Page 55 . QUESTION: 094 (1.00) The plant in at normal operating temperature and pressure when a break occurs in the Core Spray piping inside the reactor vesse Which ONE of the following indications will confirm this break has occurred? The core plate - core spray sparger differential pressure indication will: move from negative pressure towards positive pressur move from positive pressure towards negative pressur move from negative pressure towards more negative pressur more from positive pressure towards more positive pressur QUESTION: 095 (1.00) Which ONE of the following conditions is NOT required to be met prior to lining up the Core Spray system with the Condensate Storage Tank as its suction source, to ensure the system remains operable per Technical Specificatons? Condensate Storage Tank level is greater 150,000 gal, b. The Suppression Pool is determined to be INOPERABL c. The reactor is in Operational Condition 4 or 5*. The reactor vessel head is removed and the core defuele _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ . .a
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[ SENIOR REACTOR. OPERATOR Page56 QUESTION: _096 (1.00)
Unit-.2 has experienced a loss of 480 V MCC 2X Which-ONE of the following is the effect on the Standby Liquid Control (SLC) system? Adding neutron absorber solution to the storage tank will not be possible due to a loss of the 40 kW heater, Boron may start to precipitate out of solution due to the loss of the piping heat trace system, The "A" Squib Valve will not fire if the pump start _ switch is placed in " Pump A Start". The "A" Squib Valve will fire if the pump start switch is placed in " Pump A Start".
QUESTION: 097 (1.00) Conditions exist that indicate a fire in the Unit 2 turbine buildin > Which ONE of the following is NOT a responsibility of the control room operator'for these conditions? a. Manually start the fire pumps , Open the appropriate standpipe deluge valve from the control room Fire Protection. panel Monitor the fire detection instruments 1that are located-in the main control room Report the location of the fire so non-essential: personnel can evacuate I
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-SENIOR REACTOR OPERATOR Page 57
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r QUESTION: 098 ' (1. 0 0) Unit 2 is in a condition requiring.an Emergency Depressurizatio The Emergency Operating Procedures direct all 7 Automatic Depressurization-System (ADS) safety relief valves (SRV) be opene Which ONE of the following-is the MINIMUM number of SRVs that must be opened to meet the requirements for an Emergency Depressurization? QUESTION: 099 (1.00) Which ONE of the following is the purpose of the " white" indicating light associated with each Core Spray Pump? The " white" light indicates: that pumps' 4160 VAC breaker is closed with control power available, the pump was stopped while an " Initiation Signal" was still' presen the pumps' 4160 VAC breaker was-locally closed with emergency diesel generator' power being supplied.
4 the pump has tripped on fault with a valid initiation signal presen d- _ _ . b
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SENIOR REACTOR OPERATOR Page 58 _ QUESTION: 100 (1.00) With-Unit 2 at 85% powe LUsing the water chemistry report and reference material attached determine: Which ONE of the following actions must be taken? Commence an immediate plant shutdown and cooldown to less than 212 degrees Reduce the reactor water chemistry specifications to below Action Level 0 value within 24 hours, Reduce the reactor water chemistry specifications to below Action Level 2 value within 96 hours, Reduce the reactor water chemistry specifications to below Action Level 1 value within 120 hour t l: l.
i-l l ,
-(********** END OF EXAMINATION **********)
i - -_ . _ . _ -- a
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[Sl*NIOR REACTOR OPERATOR ,
4,/ I2,C : AiA N Page 2- -;<
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ANSWER- SHEET-Multiple Chcice -(Cir.cle or X your-choice)~ If you change your answer, write your selection in the blan , MULTIPLE CHOICE 023 a b @-d 001 a hc d 024 a b .c h 002 a hc d 025 @ b c d 003 a @ c d 026 a b hd 004 a .b @ d 027 a b c @. 005 a b c h 028 a b 'c h 006 a c d 029 @ b c d 007 h' b c d 030 a hc d 008 a b c h 031 @ b c- d-009 a hc d 032 a @ c d 010 .h b c d 033 hb c d-a% h
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,_ . SENIOR-REACTOR OPERATOR' .Page, 3 .< .
TANSWER S.li E E T-~
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Multiple Choice (circle or X your choice) ,
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If.you change your_ answer, write your selection in the-blan _
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046 a- -b c h 069 a b' c' @ b c d 070 a b @- d ' LO48 @ b c d 071 @ b -c
.
d 049 a hc d 072 hb c d 050 a 'b c @ 073 a @ - 'c d 051 a @ c d 074 b h' -d 052 a b hd 075 hb _c-
: d - 053 @ b c d 076 a- b c h 054 a @ c d 077 a b c h 055 @ b c d 078 a b' hd -
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SCNIOR REACTOR OPERATOR' ; Page- 4 ANSWER S.H E E T Multiple Choice- (Circle or X your choice) If you change your answer, write your selection in the blan , 092 @ b c d 093 a h c d 094 @ b c d 095- a b c @ 096 a b Q d 097 a @ c d-098 a @ c d 099 a hc d 100 a b c h l h
(********** END'OF EXAMINATION **********) :
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SENIOR REACTOR OPERATOR Page 59 ppC_ p AS/^j(
, ANSWER: -001 (1.00) REFERENCE: ,
01-01', " Operating Principles and Philosophy", Rev. 047, Page 29 CSM 7D2, " Administrative Documents", L.O. - 2
[2.7/3.7)
294001A103 ..(KA's) ANSWER: 002 (1.00) REFERENCE: 0I-01, " Operating Principles and Philosophy", Rev. 047, Page 25 OSM 7E, " Conduct of Operations", L.O. -2
' [2.7/3.7]
294001A103 ..(KA's) L l~
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ANSWER: 003 (1.00) _ j .~ l .. l- i
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. SENIOR REACTOR OPERATOR - Page :6P-l REFERENCE:
'10 CFR 26.20, " Written Policy and Procedures"'
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No Facility Specific Procedure or Learning Objective-Identified-
- (2.7/3.7] <
294001A103 ..(KA's) , ANSWER: 004 (1.00) , 1 REFERENCE: OI-05, " Annunciator Status", Rev. 020, Pages 1 & 2 OSM 7E, " Conduct of Operations", L.O. -5 (4.5/4.3] , 294001A113- ..(KA's) 4 ANSWER: 005 (1.00) REFERENCE: l 01-05, " Annunciator Status", Rev. 020, Page 6 L OSM 7E, " Conduct of' Operations", L.O. -5 (4.5/4.3] , 294001A113 ..(KA's)
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SENIOR REACTOR OPERATOR Fa;ie - 61 .
.-ANSWER: 006 (1.00) .
: REFERENCE:
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01-13, " Valve & Electrical Lineup Administrative Controls",_Re , Page 2 No Facility-Specific Learning Objecti
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- .[3.9/4.5)
294001K102 .~(KA's) ANSWER: 007 (1.00) REFERENCE: 01-13, " Valve & Electrical Lineup Administrative Controls", Re , Page 7 No Facility Specific Learning Objective Identified
' [3.7/3.7) .294001K101 ..(KA's)
ANSWER: 008 (1.00) d.
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. SENIOR LREACTOR: OPERATOR Page ; 62. - . . -REFERENCE:
10CFR50,54(x,y) No Facility Specific Procedure or Learning Objective Identified
[3.3/3.4]
294001A111 . . (KA's) ANSWER: 009- (1.00) REFERENCE: PFP-013, " General Fire Plan", Rev. 010, Page 7 No Facility Learning Objective Identified
- {3.6/4.2]
294001A110 . . (KA's) ANSWER: 010 (1.00) . , a.
,
. REFERENCE:
AI-58, " Equipment Clearance Procedure", Rev. 38, Page 21-
- OSM 7D2, " Administrative Documents", (3.9/4.5]
294001K102 . . (KA's)
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SENIOR REACTOR OPERATOR Page 63 ANSWEh 011 (1.00) N x .b # j,~g (Mt\ / 2/Y9L REFERENCE: N .
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d4A ,v>w*S
,4 y OSM 7DS, "E&RC Documents' e 2, [3.2/3.6)
294001K113 ..(KA's) N ____ ANSWER: 012 (1.00) REFERENCE: Al-58, " Equipment Clearance Procedure", Rev. 38, Page 29 OSM 7D2, " Administrative Documents", [3.9/4.5) 294001K102 ..(KA's) _ ANSWER: 013 (1.00) _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ - _ _
_ _ - . _ . . . ..___ _ . _ _ - - _ _ . _ _ _ . - . _ _ . . - . _ - - - . .. ._
< SENIOR REACTOR-OPERATOR-- - Page'. 64 ' ,
- REFERENCE: ,
.i FH-11, " Fuel Handling Procedure", "Rev. 045, Page;7=
OSM 7E, " Conduct of Operations", L.O. -1
[3.5/4.2)
-i-294001A112 ..(KA's)
~ ANSWER: 014 (1.00) REFERENCE:
,
PEP-03.8.3, " Administration of Radioprotective Drugs", Rev. 006, . Page 1 + No Facility Learning Objective Identified (2.9/4.7] 294001A116 ..(KA's)
-
ANSWER: 015 (1.00)- b.
.
a * -
- . .- - . . - - . . , . . . - . . . ~ - . . . _ . . . . . . - - . .- . ? SENIOR REACTOR OPERATOR -Page 65- ' REFERENCE:
SD-08, "CRD Hydraulic System", Rev. 016,-Pages 10 & 11 OSM 09-B, " Control Rod Drive Hydraulic System", L.O. - 3.d-&: [3.1/3.0).
201001K408 ..(KA's) j ANSWER: 016 (1.00) REFERENCE:
'
SD-07, " Reactor Manual Control System", Rev. 008, Pages 4 &-15 OSM 9A, "CRD Mechanism", Page 31 & Figure 6, L.O. -d
[3.6/3.7] -201003K404 ..(KA's)
ANSWER: 017 (1.00) REFERENCE: a OSM 9B, "CRD", L.O. - 6c,6f,6g
[3.3/3.3]
.
201003K601 ..(KA's)
,
( " ! l
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. . . . . ~ . . . - . . . .-.. . . . ~ - . _ - . . . . . _ . . .- - - . -_ .- ,
SENIOR REACTOR OPERATOR- Page 66 ! ANSWER: 018 (1.00)' e
- !
REFERENCE: SD-07, " Reactor Manual Control-System", Rev. 008, Pages 7 &.8' ' OSM 27-2A, " Reactor Manual Control System",-Pages 26 & 27, a
. (3.2/3.3]
201002K106 . . (KA's) ANSWER: 019 (1.00)
-> . REFERENCE:
SD-32.2, "Feedwater Control System",-Rev. 008, Page 17 OSM 27-2B, " Rod Worth Minimizer", Page 17, [3.1/3.2) 201006K104 . . (KA's) , ANSWER: 020 (1.00)
.
C.
I i
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w , S-- e e vg---v1 -- n-
. - _ _ _ . . ..
. , . . . . - _ . . . - . . . - . . . . . . - - .-. -.. -.-.- - . .
- SENIOR REACTOR OPERATOR .Page 67
: REFERENCE:
.i SD-02, " Reactor Recirculation System", Rev. 021,.Pages 23 &'24
'
OSM 10-2A, " Reactor Recirculation Systems", Figure 15, L.O. - 9' , 10-
& 12 [3.9/3.8)
202001A412 ..(KA's) ANSWER: 021 (1.00) REFERENCE: Brunswick Unit 2 Tech Spec 3.4.1... and bases, Pages 3/4 4-2-& B 3/4 4-1 OSM 10-2A, " Reactor Recirculation Systems", Figure 15, L.O. - 34
[3.5/3.7)
202001K601 ..(KA's) ANSWER: 022 (1.00) , a, , i-REFERENCE: OSM'10-2A, " Reactor Recirculation Systems", Page 53, L.O.'-~31
[3.2/3.2) =
202002A101 ..(KA's)
'
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l l
..
i- . . .
.
. . ,
., . .._ - = . - . - . - . . . . . . - - - - . . . . . . . . - . . . . - .
SENIOR REACTOR' OPERATOR- Page _68
' ANSWER: 023 (1.00) REFERENCE:
2 APP A-06 5-1, Re , Page 59 OSM 10-2A, " Reactor Recirculation Systems", L.O. - 1 [3.2/3.2) , 202002K305 ..(KA's) . ANSWER: 024 (1.00) REFERENCE: * .,
- - 2 OP-19, "High Pressure Coolant Injection System", Rev. 73,_ Pages-
- 8, 10 L 34-OSM 14B, "High Pressure Coolant Injection", I
[4.1/4.2)
206000G004 ..(KA's)
- --ANSWER
- 025 (1.00) . *
l i _ .- - - . .. .. . .- .
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ SE!110R REACTOR OPERATolt Page 69 l REFERE11CE: l l OSM 14G, " Standby Liquid Control System", Rev. O, Page 10, I I (3.7/3.8) 211000A108 ..(KA's) A!1SWER: 026 (1.00) ItE FEREt1CE : i OSM 28-2-A, " Reactor Protection System", Re , Page 46 L.O.-10 (4.0/4.1) 212000A216 ..(KA'n) AtJSWER: 027 (1.00) REFEREt4CE: OSti 28 2-A, " Reactor Protection System", Re , Pages 40 & 41, {3.3/3.4] 212000K502 ..(KA's) - A11SWER: 028 (1.00) _ _ - - _ . - . _ - - - . . - - _ - - - - - - - _ _ _ _ - .
. .,. _ . _ . . _ . _ . _ - -
-- - _ , _ . . . _ - . _ _ _ _ _ . . _ _ _ _ _ . _ . _ _ . - _ _ _ .
!
SENIOR REACTOR OPERATOR Page 70 ;
:
REFERENCE: , i OSM 25-F, " Traversing In-Core Probe System", Page 20, e & 9 l
(3.1/3.4) . .i 215001K604 ..(KA's)
:
ANSWER: 029 (1.00) I !
!
, REFERENCE: ; OSM 25-A, " Source Range Monitoring System", Re O, Page 14, I 3a j>004A3043.6/3.6)
.
..(KA's) i ANSWER: 030 (1.00)
$ REFERENCE:
OSM 25-D, " Average Power Range Monitoring System", Rev.'O, Pages 4, 5& 28, L.O. - 3b -;
[3.6/3.6)
21500SK505 ..(FA's)
- ANSWER: 031 (1.00) ,
p-
<
r,-w%,~ne--en-,nn,-..-n.,w-.. ---..,w.~ ,~.-..e,, a w...n.~ _- . ___ __ _ _ _ _ - _ _ _ _.-__ _ _ _ _ - __,
;
'
SENIOR REACTOR OPERATOR Page 71 l
-
. REFERENCE: i BTU Exam Bank Question #25-D, 387 (Examiner Modified)
!
OSM 25-D, " Average Power Range Monitoring System", Re O, ,
& 10 ,
[3.0/ A107) . . (KA's) l
.
ANSWER: 032 (1.00) , REFERENCE: OSM 26-2A, " Instrumentation", Rev. 5, Pages 6 & 7, No Facility l' Learning Objective Identified BTU Exam Bank Question #26-A, 969 (Examiner Modified) ,
. [3.6/3.8)
k , 216000K507 . . (KA's) I ANSWER: 033 (1.00) : REFERENCE:
'
NRC Information Notice 92-04 Dated ~8/18/92 " Effects of Non-condensable gases on BWR cold leg water level instruments" No Objective Identified , s
[3.4/3.6) ,
216000K506 . . (KA's)
.
Y P
, w +Ew .uw s.e r = w- a -- er e w w w. m w -nr-mMe .e e----me.-. . .-,2%., .m.,.-re- .v wvv- , ,w e ev= +- m. --v-,,em ,--e,--ei+-+-ee- --e
._ _ --____- - _ - -
_ . _ _ - - _ - _ _ . _ . . . _ _ _ _ . _ . _ _ . . ._ _ _ _ . . _ _ . , _ i l
- l SE!110R REACTOR OPERATOR Page 72
)
'
,
A!1SWER: 034 (1.00) ' s REFERE!1CE: i OSM 14-C, "RCIC", Re , Pages 10, 18 & 21, [3.4/3.5) 217000K601 ..(KA's) A11SWER: 035 (1.00) _
, REFERE11CE:
Brunswick Unit 2 Tech Spec 3.6.2.1, Page 3/4 6-9 , OSM 14-C, "RCIC", L.O. - 2 &9 ,
(3.5/3.6)
,
.217000A219 ..(KA's)
-
,
, ANSWER: 036 (1.00) ,
i
I
- ,- m. w, - - . -,,y.-- y- y ,r . - - , , , - - p
,
SEllIOR REACTOR OPERATOR Page 73 i REFEREllCE: i
'
OSM 14-F, " Automatic Depressurization System", Re O, Page 14, ' SD-20, " Automatic Depressurization and Safety /Rollef Valvo System", Rev. 034, Page 6 (4.2/4.3) l 218000A308 . . (KA's) i
.i A!1SWER: 037 (1.00) '!
REFERE!1CE: SD-20, " Automatic Depressurization and Safety /Rollof Valve System", Rev.- 014, Pago 4 ; OSM 14-F, " Automatic Depressurization System", Rev. O, Pago 16, .o
[3.0/3.2)
"
239002K605 . ,
(KA's)
t ,. At1SWER: 038 ( 1. 00) -i
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;
SEllIOR REACTOR OPERATOR Page 74 ,
,
REFERENCE: ,
SD-39, " Emergency Diesel Generator System", Rev. 10, Page 61 USM 20-2D, " Diesel Generators", Rev 5, Pages 17 & 41, L.O..- 23
'I (3.4/3.5)
.
264000K506 ..(KA's) ANSWER: 039 (1.00) a.
, REFERENCE:
, 'SD-10, " Standby Gas Treatment System", Rev. 009, Page 4 OSM 15-F, " Standby Gas Treatment", ; [3.7/3.8) l 261000K401 ..(KA's)
ANSWER: 040 (1.00)
' REFERENCE:
SD-17, " Residual Heat Removal System", Rev. 017, Page 3 OSM 14-2D, " Residual Heat Removal", L.O. -5 (4.2/4.3) z 203000A101 ..(KA's)
.
,
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i i
,
SENIOR REACTOR OPERATOR Pago 75-
;
,
!
' ANSWER: 041 (1.00) ,
:
- !
REFERENCE: T 2 OP-17, " Residual Heat Removal System", Rev.-99, Page 19 ' OSM 14-20, " Residual lleat Removal", L.O. -3 &8 (3.8/3.7) 203000A304 . . (KA's)
,
,
ANSWER: 042 (1.00) , REFERENCE:
"
~
2 OP-14, Reactor Water Cleanup-Systemh, Rev. 82, Page 13 OSM 11A, " RWCU", [3.4/3.3) __ 204000G007 . . (KA's)
,
ANSWER: 043- (1.00) k t w I
,
b 7~--++ r , Yw m :--- --
-
r m e ,m .-,-e .e,a, ,e -
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__ - - . . _ _ _ _ _ _ - _ _ . . _ _ _ _ _ . . _ _ _ _ _ . _ . - _ _ . - _ _ _ _ _ - . _ _ . _ . . _ .
:
SENIOR REACTOR OPERATOR Page 76 REFERENCE: ! Brunswick Tech Spec 3.1.4.3, Page 3/4 1-1 ,.; OSM 25-E, " Rod Block Monitor System", .
! [3.5/4.3)
215002G011 . . (KA's)
>
> AtlSWER: 044 (1.00)
, '
REFERENCE: SD-09, " Power Range Neutron Monitoring System", Rev.-009, Pages 22_-
& 23
,
OSM 25-E, " Rod Block Monitor System", Re O, Page 6,- .0 ( 3 . 8,' 3 . 8 ) 215002G007 . . (KA's)
'
ANSWER: 045 (1.00) REFERENCE: FH-11, " Fuel Handling Proceduro", Rev. 045, Page 4 -
. OSM 29-A, " Fuel Handling", Rev. 4, Page 46, [3.0/3.7) . >
234000K505 . . (KA's)
& ..y 4 2.,p.'[~,,..,.4O 1_ . , . . ,~..L , ,d .,%-, ' _ ..- ,., y ...L,w.,,,., [. J,,, y ^ ' '
_,_s...,m,,ym'_,L,,_,_,, J
-
_._,,,.,,%,m.,, , , - . , , , . . .
l SEllIOR REACTOR OPERATOR Page 77 ! AllSWER: 046 (1.00) REFERE!1CE: OSM 29-A, " Fuel llandling", Rev. 4, Page 37, L.O. - 5
[3.4/3.8)
234000G001 ..(KA's) _ A11SWER: 047 (1.00)
" REFERENCE:
50-01, " Nuclear Boiler System", Rev. 025, Page 4 OSM 8-A, " Reactor Vessel Internals", Re , Pages 19 7 20, & 5
[3.2/3.3)
290002K403 ..(KA's) _ ANSWER: 048 (1.00) ..
. - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ .
_ _ . . _ . - _ _ . . . . . _ _ _ _ _ . _ _ . . _ _ _
.
_ . . . . _ . _ . -~ _ _-.. _.._._ ....._._ _ .- SENIOR REACTOR OPERATOR Page 78 , t
!
REFERENCE: SD-25, " Main Steam System", Rev. 12, Page 5 OSM 17-2A, " Main Steam", Re , Pages 12 - 14, L.O. -3 (4.0/4.1)
"
239001K127 . . ( FA ' s ) ANSWER: 049 (1.00) REFERENCE: SD-11, " Process Radiation Monitoring System", Re , Page 39 OSM 17-2A, " Main Steam", .3
[3.6/3.8)
272000K101 ..(KA's)
, ~ ANSWER: 050 (1.00) REFERENCE: -2 OP-26, " Turbine. System Operating Procedure", Rev. 55, Page 18 OSM 18-2A, " Turbine", [3.6/3.7)
245000K104 ..(KA's)
,
= , - ,e,--- - - - . , -,,,,-.-.,,.. e.--,,,,.ww_,,- ,w,-i,,s..w-w,_.y_.,r ,.,.,,--n~n- , .me..-, , . , - , , --tw -
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. _ _ _ _ . . . _ . . . _ . _ _ . _ _ . . - _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . . . _ _ . . . . . _
SEllIOR REACTOR OPERATOR- Page 79 i ANSWER: 051 -(1.00) , REFERENCE: .
-
,
AOP-04.3, " Recirculation Pump Trip", Rev. 4, Page 4 ,
'
No Facility Specific Learning Objective Identified (3.8/3.8) 295001K206 ..(KA's) ANSWER: 052 (1.00) REFERENCE: Brunswick Tech Spec 3.4.1.1, Figure 4.1.1-1,-Page 3/4 4-lb OSM 10-2A, " Reactor Recirculation System", L.O. - 13 (3.5/3.8)
. .295001A201 ..(KA's)
,
ANSWER: 053 (1.00) , i l I^
,
'
t i. _~ _ . _ _ . . ~ _ _ ._
- . - . . ~ - . . . - . . . . . . - - . - ~ . - . - . ~ . - . _ . - _ _ - . ~ . . . - ~ - . . . . . . - - - - - .
SEMIOR REACTOR OPERATOR Page 80 l
. REFEREliCE: f
,
AOP-01.0, " Abnormal Drywell Leakage", Re , Page 4 f lio Facility Specific Learning Objective Identified I (3.4/3.8) i a 295010A201 ..(KA's) ;
,
AliSWER: 054 (1.00) - , !
~ REFERE!1CE:
i
'
SOP-02.1, " Inability To Move Control Rods", Re , Page 3 OSM 09-B, " Control Rod Drive liydraulic System", Rev. O, Pages 9 & 18, .
(3.7/3.9)
295022K301 ..(KA's)
-
AllSWER: 055 (1.00) , o ) h
h
.
- , . - . .- . . - . . , --.,,--.:,-........ --.-_.~.--.--...a...--.--.-- -__..__--2
.. ..
SENIOR REACTOR OPERATOR Page 81 i
;
REFERENCE:
-l
;
2 OP-02, " Reactor Rocirculation System", Rev. 68, Page 35 ' No Facility Specific Learning objective Identified ,f
[4.1/4.2)
295014A201 ..(KA's)
.
.
ANSWER: 056 (1.00)
, REFERENCE: l
;
,
OSM 07-2K.01, " Primary Containment Control", Re , Page 160,- .1
[4.0/4.1) ,
295014A107 ..(KA's) 4
,
,
ANSWER: 057 (1.00)
.i .;
-REFERENCE: AOP-07.0, " Spent Fuel Damage", Re . 2, Page 3 No Facility Specific Learning Objective Identified
[3.8/3.9)
295023G010 ..(KA's) .
') ... _ . _ , - , . . _ , - - . . . _ , . ._,......-..,,.-~,,_.._.-...,.....4,-.- --,-,_.-u-.-..
_ . ._. __,_ _-___ _-_- -_ . _ _ . - _ . _ _ _ _ _ _ _ . . . _ . _ . ~ . _ _ . _ . _ . i SE!?IOR REACTOR OPERATOR Pago 82 :
:
,
ANSWER 05 (1.00)
-
4 r REFERENCE: [ Brunswick Tech Spec Table 1.2, Page 1-11 Brunswlck Tech Spec 3.6.1.1, Pago 3/4 6-1
:
OSM 15-2A, " Primary Containment", L.O. -8 & 13
[3.6/3.8).
, 295021K101 ..(KA's) ANSWER: 059 (1.00)
< REFERENCE:
SD-46, " Instrument And Service Air System", Rev. 009, Page 11 OSM 21-2A, " Compressed Air",-Re , Page 24, [ 3 .- 5/ 3 . 3 ) r 295019A101 ..(KA's) ANSWER: 060 (1. 00) b.
l-l
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.
- ,
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e SI'1110R REACTOR OPERATOR Page 83 REFERE!1CE: AOP-20.0, " Pneumatic System Failuren", lle v . 9, Page 4 OSM 21-2A, "Compronced Air", Re , Page 24, L.O. - 21 (3.7/3.4) 195019G010 . . ( l' A ' n ) A11SWER: 061 (l.00) .____ _- _ - _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ -_ _ _ - - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - - _ _ _ _ - - .. --
!
'
, REFERE!1CE - r i AOP-32.0, "P,lant Shutdown From Outside Control Room", Rev. 19 , - i Pages 4 & 23 , 11o Facility Specific Learning Objective Identiflod (4.4/4.5) 295015K201 . . (IG 's) 1
<
I A!1SWER: 062 (1.00) L , e
!
>
,
i
-
_ . - . . . .
-__ __ _ . . _ _ . . . . _ _ . . _ . , _ _ _ _ _ . _ . . _ _ . . . . - . _ __..__;.___.;.____,__,,_,. ,.:_.,
. - - - - - . . - - - - . _ - ~ . - . - - - - - - . - . - - . . . - - - - _ . - - . - . . - .
.
.
SENIOR REACTOR-OPERATOR Page 84 I
!
;
REFERENCE: *
,
,
- - -
AOP-32.0, " Plant Shutdown From Outside Control Room", Rev. 19, Page !
No Facility Specific Learning Objectivo Identified (3.8/3.6) {
:
;
E 2950160010 ..(KA's) {
.
ANSWER: 063 (1.00)
, ,
REFERENCE:
!
AOP-37, " Low Condenser Vacuum", Rev. O, Paga S APP UA-23 3-1, "Exh Hood B Vacuum Low", Rev. 23, Pages 38 & 39 , No Facility Specific Learning Objective Identified
* (3.2/3.2)
295002G005 ..(KA's) ; 064 (1.00)
'
ANSWER: REFERENCE: AOP ,*/, " Low Condenser Vacuum", Rev. O, Page 5 _
'
OSM 17-2A, " Main Steam", L.O. --3 (3.5/3.6) 295002K203 . . ( KA_' s )
-
i i
, , . - ~ , . - . , . . . . . . . . . - - - - , - , ,-
'
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e u ..- , , - - - , . ~ . . ., ,.--.e., ~,,n.., . -.,w',...,-..,,n-,,.Sa,v,,n_.,, ,n.- .,m.v.- -~,5,,-,. m-w e ,- .
~ - . . - ~ . . - - . - . - , . . . - - . . . . - - - - . - . . . . . . , - . . - . - . . . - . . . - - - . . .
BENIOR REACTOR OPERATOR Page 85 l ANSWER: 065 (1.00) ,
1 j l REFERENCE: i i a AOP-22.0, " Low System Frequency", Rev. 3, Page 4 , No Facility Specific Learning Objectivo identifled
[3.9/4.2) ;
295003A205 ..(KA's)
- ANSWER: 066 (1.00) REFERENCE:
EOP-01, "Roactor Scram Procedure", Step RSP-006 01-37, " Preparation and Review of the PSTG", Rev. 017, Page AA-5' OSM 18-2A,-" Main Turbino", b.O. - 5
. -(3.2/3.3)
295005A208 ..(KA's) 1.
' ANSWER: 067 (3.00) l , l . l l L- , r D
l I
.
. ,
.. .. . . .
.
.
_ . _ _ _ __. _ __ _ _ . . _ _ _ _ _ . . _ . _ _ _ _ _. _ _ . _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ . _ .
!
I SE!1IOR REACTOR OPERATOR Page 86 REPERENCE i
!
,
EOP-01, " Reactor Vessel Control Procedure", Rev. O, Flowchart Step
- 006 OSM 07-2K,05, " Reactor Vesnol Control Proceduro" -
, .
!
i
[3.8/4.4)
295006G012 ..(KA's) { t
'
AI'SWER: 068 (1.00) > e i REFERENCE:
:
EOP-01, " Reactor Scram Procedure" , Step RSP-065- - t OSM 07-2K.04, " Scram Procedure Flowchart", Re , Pages 11, L.O.--
(4.0/4.1)
295007K304 ..(KA's)- At1SWER: 069 (1.00) ,
>
-,3d , * v ,- E v er-., ~- w .- . , - , . - . , - .v ,,,,4w w- -~,a- v, -- --<, ,-,-----<mr------,-e-- - - - - , m s---,4+---. - - - + < - - - , - - . , % -. +t
- . . - - - . . . - . . - - - - . - - . - ~ _ - - - . . . - - - ~ - . - _ ~ _ - . . - . -
!
!
'
SENIOR REACTOR OPERATOR Page 87 :
;
i
'
REFEREllCE:
,
-;
EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchart ' Ftop PC/P-17 l OSM 07-2K.08, " Supplemental Emergency Procedure", Re , Page 12, L.O. -3 ; , ! '
[3.8/4.0)
i 295010K301 ..(KA's) A11SWER: 070 (1.00)
. REFERE!1CE:
-
'
EOP-01-SEP-03, " Suppression Pool Spray Procedure", Re , Page 3
.OSM 07-2K.08, " Supplemental Emergency Procedure", Re , Pages 30
, & 31, L.O. - [3.9/3.9) > 295024A117 ..(KA's)
.
ANSWER: 071 (1.00) i e
.!
_ _ _ _ _ _ _ _ _ _ _ SENIOR REACTOR OPERATOR Page 88 REFERENCE: OSM 15-2E, " Primary Containment Isolation System", Re , Page 5, & (3.4/3.8) 295020A206 ..(FA's) ANSWER: 072 (1.00) a.
REFERENCE: EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchart Step SP/L-2 0 Containment Limits, SRV Tail Pipe Level Limit OSM 07-2K.01, " Primary Containment Control", L.O. - (3.5/3.6) 295029A202 ..(KA's) ANSWER: 073 (1.00) REFERENCE: OSM 07-2K.11, " Reactor Flooding Procedure", Re , Pages 40 & 41, L.O. -8 (4.6/4.7] 295031K101 ..(KA's) l _ ..
__.-_ ____ __ ____
. . . . _ . .
SE!1IOR REACTOR OPERATOR Page 89 AllSWER: 074 (1.00) REFERE!1CE: OS!4 07-2K.10, " Level / Power Control", Re , Page 94 llo Facility Specific Learning Objective Identified
[4.6/4.8)
295031A204 ..(KA's) A!1SWER: 075 (1.00) , REFERE!1CE: EOP-01, " Reactor Flooding Procedure", Re , Flowchart Figure 1 OS!4 07-2 K 11, " Reactor Flooding Procedure", Re , Page 37, L.O. - [3.7/4.0) 295031G007 ..(KA's) A!!SWER: 076 (1.00) d.
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ . _ . _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ _ - - _ _ _ _ _ - - _
_ _ _ _ - _ - _ _ _ _ _ _ ._ _ _ __ _ _ _ _ _ - _ - _ _ _ _ _ _____- _ __ - _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ i
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SENIOR REACTOR OPERATOR Page 90 l
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REFERENCE: f i OSM 07-2K.01, " Primary containment control", Rev. 5, Pages 139 & , 140, &7 l t !
[4.0/4.1) i
-
.
I 295024K208 . . (KA's)
:
I t ANSWER: 077 (1.00) , d.
REFEREllCE: ! EOP-01-LEP-02, " Alternate Control Rod Insertion", Rev. 012, Page 17 I . OSM 07-2K.07, " Local Emergency Procedures", &6 .
[4.0/4.2)
295015A102 . . (KA's)
!
078
"
ANSWER: (1.00)
. REFERENCE: l AOP-15.0, " Alternate Shutdown Cooling Methods",-Re , Page 4 NO-OSM or Facility Specific Learning Objectives Identified-(3.5/3.6)
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l SEldIOR REACTOR OPERATOR Page 91 .
:
i i AllSWER: 079 (1.00)
, REFERE11CE:
EOP-01-UG, " User's Guido", Rev. 15, Page 53 OSM 07-2K.02, " Secondary Containment Control", L.O. - (3.8/3.9) T 295032K303 ..(KA's) A11SWER: 080 (1.00) , REFERE! ICE: OSM 07-2K.02, " Secondary Containment Control", L.O. - BTU Exam Bank Question #07-K.02, 532 (Examiner Modified)
(3.9/4.2)
295033K102 ..(KA's)- 7 A14SWER: 081 (1.00)
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.
EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchar Step PCCP-2 * OSM 07-2K.01, " Primary Containment Control", Re , Page 10, '
(4.1/4.4)
:
295013G011 ..(KA's) t i AllSWER: 082 (1.00)
. ,
REFEREllCE:
;
OP-14, " Reactor Water Cleanup", Rev. 82, Page 20 OSM 11A, " Reactor Water Cleanup" , Re , Page 16, L.O. -2&3 (3.3/3.3] ,
295008A102 ..(KA's) -
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AtiSWER: 083 (1.00) : i t
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SE!$10R REACTOR OPERATOR Page 93 { REFERENCE: OSM 07-2M.10, " Level / Power control", Rev. 3, Page 93 i i
'i OSM 14-G, " Standby Liquid Control", &4
'
(3.4/3.6)
,
295037K104 ..(KA's)
. i ANSWER: 084 (1.00)
.
, REFERENCE: EOP-01, " Lovel/ Power Control Procedure", Re , Flowchart Step RC/Q-06 ,
'
OSM 07-2K.10, " Level / Power Control", Rev. 3, Page 86 OSM 10-2A, " Reactor Recirculation System", [4.1/4.2] r 295037K301 ..(KA's) !
'
ANSWER: 085 (1.00)
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I SEllIOR REACTOR OPERATOR Page 94
,
REFERI.NCE:
,
t EOP-01, "Lovel/ Power Control Procedure", Rev. 8, Flowchart Step ! RC/ P-06 & RC/P-07 OSM 07-2K.10, " Level / Power Control", Rev. 3, Pages 53 & 54, ;
' (3.8/4.1) P 295037K306 ..(KA's)
A!1SWER: 086 (1.00) , REFERENCE: EOP-02, " Primary Containment Control Proceduro", Rev. 4C, Flowchart Step SP/L-36 OSM 07.2K.01, " Primary Containment Control", Re , Page 69, [3.8/3.9) , 295030K201 ..(KA's)-
.AHSWER: 087 (1.00) , ._
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_- ! l SENIOR REACTOR OPERATOR Page 95 REFERENCE: AOP-05.0, " Radioactive Spills, High Radiation and Airborne Activity", Rev. 3, Page 3 l LP 07-M, " Abnormal Operating Procedures", L.O. -1 I
(3.9/3.6) .j
295034G010 ..(KA's)
!
I i ANSWER: 088 (1.00) , 1 i l l ' ;
!
REFERENCE: !
.
i AOP-36.1, " Loss of Any 4KV Buses", Re , Page-17 I OSM 20-2D, " Diesel Generators", L.O. -1&2
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(4.2/4.3)
:
295003A102 ..(KA's) !
,
ANSWER: 089 (1.00) -
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V SEil20R REACTOR OPERATOR Page 96 ItEFERENCE: 7~ s. ' ick Tech Spec 3.6.2.1, Page 3/4 6-9 Ct A, " Primary Containment", L.O. -
[4.1/4.2)
295026A201 ..(KA's) ANSWER: 090 (1.00)
- REFERENCE:
OSM 17-2B, " Condensate & Feodwater", Rev. 6, Page 29 17-211 Digital Feedwater Control BTU Exam Bank Question #20-F, 943 (Examiner Modified)
[4.0/4.0)
295009A102 ..(KA's) ANSWER: 091 (1.00) _ REFEREllCE: 01-04, "LCO Evaluation and Follow-Up", Rev. 043, Pages 43 & 44 OSM ~7 D3, " Operations Documents", L.O. - 3
[4.2/4.2)
294001A102 ..(KA's)
, - _ _ _ . _ _ _ _ _ _ _ _ _ . _ . _____ .. _
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.
SENIOR REACTOR' OPERATOR Page 97
. ANSWER:- 092 (1.00) REFERENCE:
AOP-04.1, " Recirculation Flow Control Failure - Increasing Flow", Rev. 003, Page 3 OSM 07-M, " Abnormal Operating Procedures", No Facility ~ Learning Objectives Identified
[3.8/3.7)
295001G010 ..(KA's) ANSWER: 093 (1.00) REFERENCE: Brunswick :i' Tech Spec 3.5.1, Page 3/4 5-1 ' OSM 14B, "It .., . ;'ressure Coolant Injection", L.O. - 8 ,
[3.7/4.4]
206000G011 ..(KA's) ! ANSWER: 094 (1.00) l l a.
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SENIOR. REACTOR: OPERATOR- .Page 98 REFERENCE: a i l SD-18, " Core Spray System",-Rev4 013, Page 8' i l
.OSM 14-2E, " Core Spray",-L.O. -8 ' [3.0/3.2)
209001K404 ..(KA's)
-ANSWER: 095 (1.00) REFERENCE:
Brunswick Unit 2 Tech Spec 3.5.3.1, Page 3/4 5-4 2 OP-18, " Core Spray System", Rev. 40, Page-16 ' OSM 14-2E, " Core Spray",-L.O. -6 &7
-[3.1/3.1)
209001K101 ..(KA's) ANSWER: 096 (1.00) _ __ _ __
--REFERENCE:
SD-05,-" Standby _ Liquid Control System", Rev.1008, Page-6 OSM 14G, " Standby Liquid Control System",.Rev. O, Page-6,; .b
-
L(3.1/3.2] l~ 211000K202 ..(KA's)-
:
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., _ . . , . , . . .
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' SENIOR REACTOR OPERATOR- Page 99" -ANSWER: 097 ( l' . 0 0 ) REFERENCE:
OSM 32 2-A, " Fire Protection", L.O._- f,page 4,35
-[3.'8/4.0)
l 286000G001 ..(KA's) ANSWER: 098 (1.00) REFERENCE: EOP-01-LPC, " Level Power Control", Rev. 8,' Flowchart Step RC/P-30 OI-37, " Preparation & Review of PSTG", Rev.-10' Page Y-28'
,
OSM 14-F, " Automatic Depressurization System", L.O. - 1 [4.5/4.6] 218000K302 ..(KA's)
.
ANSWER: 099 (1.00) b.
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;. SENIOR REACTOR-OPERATOR Page100 REFERENCE:
SD-18, " Core. Spray System", Rev. 013, Page 5 OSM 14-2E, " Core Spray", L.O. - 10 (3.8/3.6) r 209001A401 ..(KA's) N7Eia , luu- ( 1. u G7 - M EFERENCE: - - ~ t;OPr0T;T;PC-*tevelivwt.u Contro1*, Kev. 8, nowcfrart-StsQ_ 1r-77 , " Preparation 6 xev1wof-PS-TGttNev . 10, Page-Y-28- ~ 9sM 14-F, " Automatic 1cpressur trativu System M ,.O- - 13 = a._ , -f 4 T5/ 4 r6-}- -~~ ,
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l l l-l (********** END OF EXAMINATION **********) 1: l _ _ l.
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*QNUM *HNUM * ANUM l *QCHANGED TRUE *ACHANGED TRUE - *QDATE 1992/12/07 *FAC 325 BRUNSWICK 1 &2 *RTYP BWR-GE4 *EXLEVEL S *EXMNR CHUCK *QVAL *SEC *SUBSORT *KA 94001A114 * QUESTION JOD With Unit 2 at 85% powe Using the water chemistry report and reference material attached
- determine
' Which ONE of the following actions must be taken? ' Commence an immediate plant shutdown and cooldown to less than 212 degrees F.
, ' Reduce the reactor water chemistry specifications to below Action Level 0 value within 24 hours.
I Reduce-the reactor water chemistry specifications to below Action Level 2 value within 96 hours, l l Reduce the reactor water chemistry specifications to below Action Level 1 value within 120 hour * ANSWER lOO d '. ! * REFERENCE l AI-81, " Water Chemistry Guidelines", Rev. 11, Pages 6 & 12 No Facility Learning Objective Identified (2.9/3.4) !
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1 * gpp ,y A sTE/t U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION REACTOR OPERATOR LICENSE 3 REGION 2
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CANDIDATE'S NAME: , FACILITY: Brunswick 1 & 2 REACTOR TYPE: BWR-GE4
~
l DATE ADMINISTERED: 92/12/07 _ 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 %
$PP, c) .
i 100rOO % TOTALS FINAL GRADE All work done on this examination is my ow I have neither given nor received aid, i candidate's Signature
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-REACTOR OPERATOR Page 2
:s ANUWER S H E E~T
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'Hultiple Choice- (circle or X your choice)
If you change-your answer, write your selection-in the-blan MULTIPLE CHOICE 023 a b c' d 001 a b c d 024 a b c d 002 a b c d 025 a b c d
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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-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 .012 a b c d 035 a b c d f 013 a b c d 036 a b c d 014 a b c d 037 a d 015 a b c d 038' a- b d: 016 a b c d 039 a b c d 017 a b c d 040 a b c d- ' 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
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REACTOR OPERATOR Page 3
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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 069 a b c d 047 a b c d 070 a b c d 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 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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REACTOR OPERATOR Page 4 A N~S W E R SHEET Multiple Choice (Circle or X your choice)- If you change your answer, write your selection inLthe blan a b c d
.093 a b c d 094 a b c d 095 a b c d 096 a b c d 097 a b c d 098 a b c d 099 a b c d 100 a b c d (********** END OF EXAMINATION **********)
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Page 5
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: 1. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completin This must be done after you complete the examination. g the examinatio . Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer shee . Mark your answers on the answer sheet provide USE ONLY THE PAPER PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAG . Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question pag . Use abbreviations only if they are commonly used in facility literatur Avoid using symbols such as < or > signs to avoid a simple transposition error resulting in an incorrect answer. Write it ou . The point value for each question is indicated in parentheses after the questio . Show all calculations, methods, or assumptions used to obtain an answer to any short answer questions, 11. Partial credit may be given except on multiple choice questions. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . Proportional grading will be applied. Any additional wrong information that is provided may count against you. For example, if a question is worth one point and asks for four responses, each of which is worth 0.25 points,ints.and 0.20 po youone If give offive responses, your each of five responses isyour responses incorrect, 0.20will willbe worth be deducted and your total credit for that question will be 0.80 instead of 1.00 even though you got the four correct answer . If the intent of a question is unclear, ask questions of the examiner onl t
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Page 6 14. When turning in your examination, assemble the completed examination with examination questions,_ examination aids and answer sheets. In addition, turn in all scrap pape . Ensure all information you wish to have evaluated as part of your answer is on your answer shee Scrap paper will be disposed of immediately following the examinatio . To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio ,
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18. When you are done and have turned in your examination, leave the examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is still in progress, your license may be denied or revoked.
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_ _m.._ _ . . . = _ _ . - _ _ . _ . . REACTOR OPERATOR Page 7 QUESTION: 001 (1.00) Which ONE of the following positions must approve securing the Unit 1 Balance of Plant RO (BOP RO) with the unit shutdown? Plant General Manager Manager - Operations Operations Manager - Unit 1 , Unit 1 Shift Supervisor
; QUESTION: 002 (1.00)
As required by 10 CFR 26, " Fitness for Duty Programs", which ONE of the-following is the MINIMUM time an operator must abstain from the consumption of alcohol prior to any SCHEDULED shif t?
- hours hours hours hours QUESTION: 003 (1.00)
Which-ONE of the following annunciator window colors designates a-setpoint-important to reactor safety? Amber annunciator window with a red bar, Amber annunciator window with a blue-ba Red annunciator window with a red bar, Red annunciator window with a blue bar.
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i-REACTOR OPERATOR Page 8
- - . QUESTION: 004 (1.00)
Given the following conditions:
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A multiple input annunciator is in continuous alarm due to-one invalid input
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The annunciator card has been pulled and:the invalid input signal defeated
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The annunciator card has been reinstalled Which ONE of the following colored dots should he plac.3lon the annunciator window? Blue Yellow Red Black QUESTION: 005 (1.00)
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A motor-operated valve (MOV) has been manually. closed and torqued by procedur Which ONE of the following informs the. operator that this!MOV is in this : condition and must be untorqued prior to opening with the moto . operator? , A tracking LCO written for the valv A caution tag placed on the RTG A danger tag placed on the RTG An entry made in the nnit Control Operator's Lo P
REACTOR OPERATOR Page 9
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QUESTION: 006 (1.00) Which ONE of the following radiation exposure guidelines is the MINIMUM exposure at which the Shift Supervisor may waive the independent verification requirement for a system valve lineup? Total exposure expected during the independent verification is in excess of 10 m Total exposure expected during the independent verification is in excess of 100 m General area radiation levels are in excess of 10 mr/h General area radiation levels are in excess of 100 mr/hr.
QUESTION: 007 (1.00) 10 CFR 50.54(x) specifically allows " reasonable action that departs from a license condition or a technical specification in an emergency when this action is immediately needed to protect the public health and safety ...." Which ONE of the following approvals MUST be received prior to these actions being taken? These actions must be approved by: Plant General Manager - NRC Senior Resident Inspector - Manager - Operations Senior Control Operator _ _ - _ - _ _ _ _ .. _ ._._,._ _..._ _._.-._-.~. . . - _ _ . - _ _ _ . _..- .._ __ .._._. _ _ -_ REACTOR OPERATO Page.l'O QUESTIOt(: 008 '( 1. 00 ) A total loss of offsite power has occurred on both units. DG's #1 and-
#2 have failed to start. DG's #3 and #4 started and loaded normall .
Which one of the following describe the response of the Motor driven fire pump? a. _The power supply automatically transfers from E-1 to E- b. The power supply automatically transfers from E-2 to E- , c. The pump motor remains powered from E-3, its normal supply . d. The pump motor remains powered form E-2,_its normal supply and is unavailabl QUESTION: 009 (1.00) Under which ONE of the following sets of conditions may a FAIL CLOSED pneumatic valve be used as an isolation point for a clearance? a. The valve is checked closed and the air supply isolation valve to the valve is tagged closed, b. The air supply line to-the valve is removed and tagged'and the valve is tagged and double verified close c. The air supply isolation valve is tagged closed and the valve operator vent valve is tagged ope d. The valve has a gagging device or clamp installed-and this device is tagged.
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_ - . . _ _ _ _ . _ . - ,. . . , _ . . . , _ _ , . _ . _ . . _ . - . . _ . . _ . _ _ _ , . . . . , , , l REACTOR OPERATOR - Page 11 QUESTION: 010 ' (1.00) U -1 is making preparations for an drywell entr There are no self- > conta n'd breathing apparatus readily availabl ~Which ONE of h Q ollowing oxygen levels is the MINIMUM allowed for an entry under these con itions? % , % - % b9gy g c2
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QUESTION: 011 (1.00) Which ONE of the following conditions REQUIRE an independent verification when canceling a clearance? An independent verification must be performed if: a double verification was performed when the clearance tags were hun an independent verification was performed when the clearance tags were hun . i- the equipment being verified is " common" or shared between Unit 1 and Unit the equipment being verified will be inaccessible during plant operation.- o ! !
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_ . __ ,. REACTOR OPERATOR Page 12
- QUESTION: 012 (1.00).
Given the following conditions: ,
--
A fully trained, male radiation worker 32 years ol He is preparing to work in a 150 mrem /hr radiation fiel He DOES NOT have a NRC FORM 4 on fil His current quarterly exposure is 975 mro How long can this individual work in this area without exceeding'a Federally Regulated whole body radiation exposure limit? minutes minutes minutes minutes QUESTION: 013 (1.00) With the plant at normal operating pressure and temperature the' operator partially closes the Control Rod Drive Pressure Control Valve (F003).
Which ONE of the following describes the effect _that this will have cnr control rod speeds? s Partially closing the Pressure Control Valve: decreases control rod insertion spee increases control rod withdrawal speed, decreases control rod scram spee increases control rod scram spee s
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_ , . . . _ , ,_ . _ _ ._.. _ _... _. ..._. . _ . .._..._, . ..__.. .- . _.._ _. ..... ___ . _ _ . . _ _ _ _ _ . . . . _ _ _ . _ _ _ _ _ REACTOR OPERATOR Page 13 l QUESTION: 014 (1.00) -j Following a scram from 100% power the ball check valve in the i nsert port for one control rod drive mechanism malfunctions and fails to-unseat and shift positio Which ONE of the following is the effect on that control rods' ability .l to scram? The control rod will: ) 1 not insert until reactor and CRD pressures equalize, hydraulically lock-up and remain stationary.
. will fully insert at slower than normal speed.
' will fully insert at higher than normal speed, i L ,...
_ . REACTOR OPERATOR _ Page114 QUESTIONi 015 (1.00) While at 100%. power the CRD drive water pressure control valve is inadvertently close Which ONE of the following describes the system response ~to this valve closure? NOTE: P&ID (4 sheets)_are included for reference, System flow decrease Charging pressure increase Drive water dif ferential pressure increase Cooling water differential-pressure decrease System flow decrease Charging pressure decrease Drive water differential-pressure decrease Cooling water differential pressure decrease System flow increase , Charging pressure increase Drive water differential pressure increase cooling water differential pressure increase System flow increase Charging pressure decrease Drive water differential pressure decrease cooling water differential pressure increase
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REACTOR OPERATOR Page 15 l QUESTION: 016 (1.00) Which ONE of the following describes the operation of the auxiliary relay timer in the Reactor Manual Control System? The auxiliary relay timer: will halt the selected control rod withdrawal upon a failure of the solid state master time controls rod sequence timing when rod withdrawal is demanded by the Rod Out Notch override Switch, bypasses all interlocks except Rod Worth Minimizer insert - blocks during emergency control rod insertion controls rod sequence timing when continuous rod insertion is demanded by the Rod Movement Control Switc QUESTION: 017 (1.00) Which ONE of the following control rod blocks is DISABLED " when the Reactor Mode Switch is switched FROM "Run" TO Startup"? Intermediate Range Monitor " Upscale" Recirculation flow unit " Upscale / Inoperable" Average Power Range Monitor " Inoperable" Rod Block Monitor "Downscale" _ l i _ _ . _ _ _ _ _ _ . __._____m_______.m_ __.______.____m __,__,_,_,
REACTOR OPERATOR Page 16 QUESTION: 018 (1.00) Which ONE of the following is used to generate the Low Power Alarm Point (LPAP) for the Rod Worth Minimizer? The main turbine first stage steam pressur The total steam flow signal from the Feedwater Control System, The total feedwater flow signal from the Feedwater Control syste The highest reading Average Power Range Monitor input to the flow converter _ QUESTION: 019 (1.00) Given the following plant conditions for Unit 1: Loop "A" jet pump flow - 45 millon lbm/hr Loop "B" jet pump flow - 5 millon lbm/hr
"A" recirculation pump speed -
35%
"B" recirculation pump speed -
0% Loop "A" discharge valve - open Loop "B" discharge valve - open Which ONE of the following is the value for TOTAL CORE FLOW for these conditions? millon Ibm /hr _ 45 millon Ibm /hr
- millon Ibm /hr millon lbm/hr
._ _ _ . _ - _ . . _ - . . . _ . . - _ . _ _ _ - . _ _ .. _ _.-._. _ __ . _ . _
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QUESTION: 020 (1.00) Brunswick Unit 2 Tech Specs regarding Recirculation System Jet Pump-CPERABILITY require a plant 1 shutdown within: 12 hours if a jet-pump is found to be INOPERABL Which ONE of the following is the concern for continued plant operation with an inoperable-(or failed) jet pump? ' Unbalanced neutron _ flux across the core due to flow variations.- Physical core damage from a piece of a damaged jet pum Invalid APRM Flow Biased SCRAM setpoints due to the change in flow through a failed jet pum ? A reduction in core reflood capabilities and increased blowdown area during a Loss of Coolant-Acciden ' QUESTION: 021 (1.00) Which ONE of the following plant conditions will DECREASE the AVAILABLE recirculation pump Net Positive Suction Head (NPSH)? Recirculation pump speeds are simultaneously increased from 25%'- to 40%. Reactor water level is stable and a high-pressure feedwater heater isolate Recirculation pump speed is at minimum and feedwater flow is increased.
L . Reactor power is increased from 65% to 85% by withdrawing L control rod , l l.
L L i l-l = . . _ _ _ . _ _ . _ . . - . _ . . . _
, _ _ _ . . _ _ _ _ - _ _ . _ _ _ _ _ . . _ -__~ .__.,_m ____m ' REACTOR OPERATOR- Page 18 QUESTION: 022 (1.00) l-A "Recirc MG Speed Control Signal Failure" alarm'has annunciated for the "A" Recirculation pump on Unit '
Which ONE of the following is the expected effect on operation of the-
"A" Recirculation pump?
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L The "A" Recirculation Pump speed: will run to minimum due to the low output signal from the controller, can only be changed by the. individual pump controller in manual will remain at its existing value until the scoop tube lock can be reset, will not change due to loss of power to the scoop tube positione QUESTION: 023 (1.00) The Reactor Recirculation System operating procedure cautions against excessive-operation of the Recirculation Pumps with suction pressures less than 300 psi Which ONE'of the following is the concern with operating the pumps below this pressure? Pump operation with suction pressure less than 300 psig: l i can place the pump motor close to it's maximum-allowed _ amperage limit,
~ Will place the pump close to its net' positive-suction head
, - limit i' .will-reduce the effectiveness of_the pump; thermal bushing.
!. , can affect the life of the pump seal cartridge-assembly, i I L -
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! REACTOR OPERATOR Page 19 QUESTION: 024 (1.00)~
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Which ONE of- the following describes operation of the liigh Pressure
. Coolant Injection (llPCI) Turbine Stop Valve (V8) and the-Turbine Control Valve (V9) during an automatic start sequence?
a.
' V8 -- fully opens after the Auxiliary Oil Pump starts V9 -- throttles open after the Auxiliary oil Pump starts V8 -- throttles open after the Auxiliary 011 Pump start ' V9 -- fully opens after the Auxiliary 011 Pump starts V8 -- fully open in the standby lineup V9 -- throttles open after the Auxiliary 011 Pump starts V8 -- throttles open after the Auxiliary Oil Pump starts V9 -- fully open in the standby lineup l i _ . - . - _ . _ _ . - - , _ . . . , - - . _ , . . . . _ . . . . . _ . _ . . _ , , _ . . - , , . -
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QUESTION: 025 (1.00).
Given the High Pressure Conlant Injection (HPCI) system is in its normal standby-lineu Which ONE of the following describes the position of the Minimum Flow Valve (F012) and the reason for that position? In the standby lineup, F012 is: open to provide the required system flowrate signal to the HPCI control syste closed to-provide the required shutoff head for the HPCI pump as it comes up to spee open to prevent pump overheating by providing an:Immediate flow path on startu closed to prevent draining the Condensate Storage Tank to the suppression poo QUESTION: 026 (1.00) The Standby Liquid control (SLC) system monthly operability test on the Unit 1 "A" SLC pump-is about to be performe Which ONE of the following describes how the Reactor: Water Cleanup (RWCU) system isolation is avoided during-this test? Starting the "A" SLC pump with the local pump control-switch bypasses the RWCU isolation signal, The RWCU isolation signal is initiated from'the_SLC squib. valve-firing circuit .The breaker for the appropriate-RWCU isolation valve will 1me
- -
opened prior to running the SLC pump,
- The RWCU system must be shutdown.and the appropriate isolation valves closed before running the SLC pum .
T e --,-n --n.-rg- - - -
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REACTOR' OPERATOR Page-.21 .
, QUESTION: 027 (1.00)
The-'following conditions exist-for unit > The reactor has scrammed.and the mode switch is in-SHUTDOW j The problem has been identified and correcte l Alarm "DISCH VOL HI LVL CRD TRIP" is actuated, j l Which ONE of the following describes system response when the operator places the Scram Discharge Volume High Level scram Keylock switch to-BYPASS, turns the scram reset switch to-both directions and then places the mode switch to STARTUP? No system response for the present plant condition The scram will reset and remain rese The reactor will reset and again scra The scram will reset when the scram discharge volume drain , QUESTION: 028 (1.00) Following a reactor-scram signal there is a 10 second time delay before-the scram may be rese Which ONE of the following describes the purpose of-this time delay?
'
During this 10 seconds: all of the scram relays are verified.deenergize the Scram Discharge Volume 1 completes' draining, the control rods reach the fully inserted positio all of the Scram Valves complete their stroke.
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-REACTOR OPERATOR Page.22 QUESTION: 029 (1.00)
:A: Traversing In-Core Probe -(TIP) trace is being_taken when an instrument technician error causes a containment isolation signa _l Which ONE of the following describes the automatic response of the-TIP system?
a. The TIP ball valve closes, cutting the detector cable and sealing the guide tub b. The TIP shear valve fires to cut the detector cable and sealing _ the guide tub c. The TIP drive shifts to manual reverse withdrawing the detector to the in-shield position, then the shear valve fire The TIP drive shifts to manual reverse withdrawing the detector to the in-shield position, then the ball valve close QUESTION: 030 (1.00) Which ONE of the conditions below is the point at which there are NO Source Range Monitor (SRM) control rod blocks preventing rod motion? a. IRM range switches are selected to Range 8 or abov b. SRM reactor _ protection shorting links are remove c. Reactor Mode Switch is in NOT in "Run", d. SRM power level is greater than 2.0 E5 count . . . , - _ - ,
_._.-_-_._ _.__- ___ - -
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QUESTIO!It 031 (1.00) i s Given the following conditionc for Unit 2:
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The "B" Recirculation Loop is isolated
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Reactor power is 22% ,
-- "A" Loop recirculation flow is 43% of rated ,
Which ONE of the following in the Tech Spec APRM Flow Biased simulated Thermal Power Trip notpoint for these conditions? (choices are rounded to the nearest tenth.)
i .0 percent i .4 percent
! .9 percent .5 percent -
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QUESTIOll: 032 (1.00) ,
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, Unit 1 is performing Average Power Range Monitoring [APRM) calibrations and gain adjustment factor (GAF) adjustments as requ; red by Tech Spec , . Which OllE of the following sets of conditions do not require adjustment? i APRM cha nnel "!3": , ' indiceted power is: 47% . calculated power is: 46.5% indicated power is: 76% ; 76.8% ' calculated power is: indicated power is: 35% calculated power ist 35.4% indicated power is:- 31% calculated power ist 34.1%
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REACTOR OPERATOR Page~24 . QUESTION: 033 (1.00) Which ONE of the following reactor vessel water level instruments will provide the most RELIABLE level indication during an accident resulting in high drywell temperatures? Harrow Range Instrument (N004A) Wide Range Instrument (H026B)
- Fuel Zone Instrument (NO37) Shutdown Range Instrument (N027A)
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QUESTION: 034 (1.00) Which ONE of the following reactor pressure conditions MAY cause
"NOTCllING" of the reactor water level instruments? Below 450 psig Between 500 psig and 625 psig Betweer 650 psig and 725 psig Above 775 psig
_ _
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_ _ .w.__ _ __ _ _ . _ _ _ . _ . _ _ . . _ _ ._m_..__ - _ _ _ . _
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REACTOR OPERATOR page ,4 * I QUESTION: 035 (1.00) ; Given the following Unit 2 plant conditions: f
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Unit 2 has experienced a Loss of all AC Power ,
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The Reactor Core Isolation Cooling (RCIC) system started on a ' valid Level 2 signal-
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An Auxiliary Operator reports to the control room that there is ' a large steam leak on the RCIC turbine
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The Shift Supervisor directs the Balence of Plant RO to isolate ! - RCIC
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Which ONE of the following valves will CLOSE when the RO depresses the manual isolation pushbutton? F045 -- Turbine Steam Supply Valve ; F007 -- Steam Supply Inboard Isolation Valvo F010 -- Condensate Storage Tank Suction Valve F029 -- Suppression Pool Suction Valve QUESTION: 036 (1.00) Which ONE of the following is the PREFERRED method to. terminate and prevent injection from the Unit 1 Reactor Core Isolation-Cooling LRCIC) system during an ATWST (Assume RCIC was initiated from a valid s:.gnal.) Close the RCIC Steam Supply Inboard & Outboard Isolation Valves , Close the RCIC' Steam Supply valve l' Depress the RCIC turbine manual trip pushbutton
' Depress the RCIC system manual isolation'pushbutton t
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- -___.___ _ _ _ _ _ _ _ _ _ _ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ . .. - _ _ _ _ _ _ _ _ _
i i REACTOR OPERATOR Page 26 (
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-QUESTION: 037 (1.00) {;
During a loss of coolant accident the Automatic Depressurization System !
(ADS) has initiated and is depressurizing the reacto >
Which ONE of the following conditions Will close_the ADS safety relief ! valves (SRV)? , The SRV actuating air supply pressure is within 50 psi of containment pressur Containment and reactor pressures are within 50 psi.- f
: Reactor pressure has decreased to approximately 100 psi ! The SRV actuating nitrogen supply pressure has decreased below 100 psi QUESTION: 038 ( 1. 00)
Safety Relief Valve (SRV) F013G is being operated from the-main Control-Roo Which ONE of the following provides the input to_the " Red" light on the control panel for that SRV? The " Red" light comes on when: the SRV pilot valve solenoid is energize steam flow begins through the SRV tail pipe, the SRV main valve lifts off its closed sea SRV discharge piping temperature reaches 295 degrees .. V g-- v--a+www- y s v-sw gT W *#"N*W t'-Nw-*-PC1-tP7'4*w c- e'mt w rwhg e + w- e w wr e' s W * ,
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REACTOR OPERATOR page 27 :
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QUESTION: 039 (1.00)
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Which ONE of the following is indication that a Unit 2 Safety Relief Valve [SRV) vacuum breaker has failed in the open position during SRV .: operatton? This failure will result in: direct pressurization of the suppression chamber air space each time the SRV is opene steam bypassing the relief valve T-quenchers with a direct-discharge path into the suppression pool, an increased heat load on the drywell coolers each time the SRV is opene , suppression pool water being drawn up into the SRV discharge line after the SRV is closed.
l QUESTION: 040 (1.00) Diesel Generator #3.is running supplying 4160 V emergency bus E3 in
<
parallel with the grid for periodic testin The Control Room operator takes the Diesel Generator #3 Voltage Adjusting Rheostat to the " Raise" positio Which ONE of the following would be the expected response of the. diesel , i generator indications?
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Diesel Generator #3: output'MVARs increase, output voltage increases, output KWs increase.
, output frequency increase [
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REACTOR OPERATOR Page 28 f
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QUESTION: 041 (1.00) l Unit I has lost power to emergency bus E2 coincident with a loss of coolant accident Which ONE of the following is the time delay from the Diesel Generator
#2 start signal to the "1B" Core Spray Pump starting? (Assume all systems function as designed.)
, seconds seconds , seconds
, seconds
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QUESTION: 042 (1.00) Which ONE of the following describes the DIFFEtlENCE in the automatic start actions between the Unit 1 and Unit 2 Standby Gas Treatment Systems (SBGT)? On a valid SBGT start signal: the Unit 1 SBGT train inlet and outlet valves do not automatically ope the. Unit 1 purge system exhaust fans must be stopped by th operator, the Unit 2 purge system exhaust outlet valves do not automatically clos < the Unit 2 SBGT train inboard and outboard ventilation dampers automatically-close.
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. . - .-. - .. -.- ...-_ -._.-
REACTOR OPERATOR Page 29 QUESTION: 043 (1.00) Which ONE of the following pressures is the point at which the Residual Heat Removal System operating in the Low Pressure Coolant Injection- -
(LPCI) mode will begin to inject water into the reactor recirculation system?
LPCI injection will begin at approximately . psig psig psig psig
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, QUESTION: 044 (1.00) The following plant conditions exist: Reactor pressure 900 psig Reactor level 185 inches l Drywell Pressure 2.9 psig
- Which ONE of the following is the result of placing the Residual Heat Removal (RHR) system "Two-Thirds Core Height Water Level Inhibit" switch '
to " Manual Override" followed by placing the " Containment Spray Valve Control Switch (Think Switch) to Manual? The Low Pressure Coolant Injection (LPCI) injection valves may be overridden closed, The Low-Pressure Coolant Injection (LPCI) injection valves.will F not automatically ope c.- The-RHR Containment Spray and Cooling valves automatic opening features are enabled, The RHR Containment Spray and Cooling valves may-be manually opened from the RTGB.
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REACTOR OPERATOR Page 30 J
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QUESTIO!1: 045 (1.00) Which -OllE of the following plant conditions will cause the Reactor Water-Cleanup (RWCU) system Inlet Inboard Isolation Valve (F001) to close?
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- A trip of either of the RWCU Recirculation pump , Excessive temperatures in the RWCU system area The Standby Liquid Control system switch is placed in " Pump A Start", Any signal that closes the Inlet Outboard Isolation Valve (F004).
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-QUESTIOll: 046 ( 1. 00)
Which ONE of the following prevents a loss of main condenser vacuum-when the Reactor Water Cleanup _(RWCU) system is lined up for reactor reject operations? Administrative controls on the operation of RWCU Reject To Condenser Valve (F034) and RWCU Reject to Radwaste Valve (F035). Interlocks preventing the simultaneous opening of RWCU Reject To Condenser Valve (F034) and RWCU Reject to Radwaste Valve '
(F035), The RWCU system high differential. flow automatic system isolatio , The automatic closing features of RWCU Reject Flow Control Valve (F033). ,
t
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. _ .___.._;_._.______._...._._ _ -
_..____..u....-- _ c.- , _ . _ . - - -.
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QUESTION: 047 (1.00) e Which-ONE of the following describes the function of the " Gain Change Circuit" in the Rod Block Monitor (RBM) syste The Gain change circuit: Initiates a rod block if the LPRM average input is higher than the APRM reference signa prevents rod movement until the RBM output is adjusted to equal the APRM reference signal, initiates a " null sequence" if the difference between the-local average power and the APRM reference signal is too larg modifies the RBM amplifier output if the average of the LPRM inputs is lower than the APRM reference signa QUESTION: 048 (1.00) Given the following conditions:
-- ' Unit 2 is at 45% power --
APRM "D" fails "Downscale"
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No operator actions have been taken -
-Which ONE of the following is the expected response of the Rod Block t Monitor (RBM) system?- (Do not consider the response of any other plant system.) RBM channel "B" sends a rod withdrawal. block to RMC RBM channel "A" sends a rod select block to RMC :RBM channel "B" rod block logic is bypasse RBM channel "A" automatically shifts to APRM "F".
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r f QUESTION: 049 (1.00) , Which ONE-of the following SRM combinations meet the MINIMUM operability requirements for CORE-ALTERATIONS? (Assume SRM locations are appropriate for the next fuel movement.)
(BYP = Bypassed DWNSCL = Downscale) ! SRM "A" SRM "B" SRM "C" SRM "D" I
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t Counts: 5 DWNSCL 2 5 i Inserted: Yes Yes Yes No
, Counts: BYP 4 DWNSCL 6 Inserted: Yes Yes No Yes ' Counts: 1 BYP DWUSCL 10 Inserted: Yes Yes Yes Yes- - - Counts: 5 BYP 10 2 Inserted: Yes Yes No Yes .
QUESTION: 050 (1.00)
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Which ONE of the following conditions is the fuel tag board used as th PRIMARY reference for core fuel location during a refueling outage? The primary reference is the: refueling floor tag board which is updated prior to the-'end of each shif , refueling floor tag board which is updated after each core alteratio Control Room tag board which is updated prior to the end of each shif Control Room tag board which is updated after each core alteratio _
- TM73'P- = = -
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REACTOR OPERATOR .Page 33 l QUESTION: 051 (1.00) Which ONE of the following describes how the reactor core design ensures that each fuel bundle receives adequate cooling flow? Adequate fuel bundle cooling is assured by: the fuel support pieces-installed on the core peripheral fuel bundles to divert flow away from the outer bundle by placing the highest power bundles in the main discharge flow from the recirculation system jet pump ; minimizing fuel bundle two phase-flow resistance by limiting the overall bundle Linear licat Generation Rate (LHGR). bottom core plate design which diverts flow to the cores -l centrally located highest power bundle ,
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QUESTION: 052 (1.00) A Unit 2 4160 VAC breaker has been racked in following maintenance on its associated pump. The operator racking in the breaker installed the control power fuses but failed to position the breaker toggle switch to the "on" positio . Which ONE of the following describes the affect on breaker operability? The breaker will'have: normal control room indications but no remote operation will be possibl normal control room indications but can be only be closed by local pushbutton, no control room indications and cannot be operated by the
: remote controls.
I L no control room indications but may be closed with the local j pushbutto __ l L l
, . ~ .. . .- ---_-=-: . - - . . - =.-.--.- - .. .. - -w
_ .. .. . - _ _ _ . . _ _ _ _ _ . . . . . . _ _ _ . _ _ _ _ _ _..__ _ _ __ _.__...__ , l REACTOR OPERATOR Page 34 ; I QUESTION: 053 (1.00)
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l A fire has been reported in the 2B RHRSW Booster Pump Moto Which ONE of the following is required concerning the Reactor Building , Standpipe Deluge valves? No action is required, the valves automatically open.
. An (AO) must be sent to manually open the valve The valves must be opened at the Fire Protection Panel in the main' control loo , i No action is required, the valves are normally ope : QUESTION: 054 (1.00) ,
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The following conditions exist for unit Reactor power 10 % - MSIV 1821-F022C is stuck at 25 % ope Which ONE of the following MSIVs, if closed will cause a Half Scram signal? F022D B21-F028C
- B21-F022B F028B 1 L
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_ _ _ _ _ . REACTOR OPERATOR Page 35 QUESTIO!1: 055 (1.00) Which ollE of the following combinations of Main Steam Line Radiation Monitor trips will cause a FULL Main Steam Isolation Valve (MSIV) closure? (Do not consider any other actions from these trips.) Channel "B" -- Downscale Channel "C" -- Inop Channel "C" -- Inop channel "I"' -- liigh - liigh hannel
. "A" --
liigh _ channel "D" -- liigh - liigh Channel "A" -- liigh - High channel "C" -- liigh - liigh QUESTIoll: 056 (1.00) During !!P turbine shell warming, the operator is cautioned not to exceed 155 psig first stage pressur Which 011E of the following is the result of exceeding this limit? First stage pressures above this value: may reach the setpoint that removes the Group !!otch Control notch restraint limitation _ will cause the main turbine to roll off the turning gear without receiving a proper warmu may heat the LP turbine exhaust hoods to the point where hood spray automatically initiates coolin will place the plant close to the setpoint which arms the Turbine Trip acram circui _ -- _ ___ _______-______ ______________ - _________ -
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REACTOR OPERATOR Page 36 QUESTION: 057 (1,00) Unit 2 has experienced a loss of coolant accident (LOCA) due to a recirculation loop brea Which ONE of the following prevents the operator from IMMEDIATELY-shifting one loop of Residual Heat Removal (RHR) from the injection modo to suppression pool cooling? The RHR Service Water system capacity will not immediately support the additional heat load of suppression pool cooling, i The valvo lineup required for suppression pool cooling is prevented from being completed until after a time delay, The Emergency operating Procedures require. maximum injection flow for the first ten (10) minutes following the LOC The RPR pumps cannet be stopped to transfer to suppression pool l cooling for ten (10) minutes following the initiation signa QUESTION: 058 (1.00) Which ONE of the following is the result of having a designed " holdup 4 time" in the main condenser hotwells? , The holdup time: , reduces overall radiation levels in the condensate and feedwater system allows radioactive and non-radioactive corrosion products to settle out in the hotwel provides some initial removal of dissolved oxygen from the condensat increases NPSH for the Condensate Pumps by increasing the subcooling of the condensat _
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_ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ .. - - .____ ..___ ____ REACTOR OPERATOR Page 37
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t QUESTIO!1: 059 (1.00)
Which OllE of the following is the effect of operating the Mechanics 1 '
-Vacuum Pumps above St. powe Mechanical Vacuum Pump discharge radiation levels will be excessive due to a limited discharge path hold up tim * operating temperatures will increase as the water injection heat exchanger capacity is exceeded, discharge hydrogen will increase to a concentration that could result in an explosio motor current limitations will be exceeded if operated for i extended period !
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QUESTIO!1: 060 (1.00) Which 011E of the following conditions will totally isolate the control Room from the outside atmosphere? Either emergency air filtering train is starte Any Control Room area smoke detector alarms, liigh radiation levels on control Room area rad monitor Abnormal chlorine levels at the Control Room air intake , _- . _ .
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REACTOR OPERATOR Page-38 ] i
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QUESTION: 061 (1.00) i Gjven the following conditions on Unit 2: f
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Initial Reactor Power 100% a PT,itI&C caused a failure of
--
During recirculation the performance pumps speedofcircu causing a speed reduction in ; both pump ,
--
Total core flow is 34 million Ibm /hr i
--
After transient Reactor power is 52%
--
APRM oscillations are 8% (peak to peak) Which ONE of the following 10 the proper operator response? NOTE: Reference is attached: , Manually scram the reacto Increase recirculation flo f Decrease recirculation flo Insert control rod , QUESTION: 062 (1.00) Which ONE of the following is the MINIMUM total core flow at which the ' plant can operate and still be assured of avoiding power oscillations or instabilities?
' million Ibm /hr - million Ibm /hr million Ibm /hr i million Ibm /hr -
>
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_ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ - _
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,
REACTOR OPERATOR Page 39 t I QUESTION: 063 (1.00) i The Unit 2 Plant Monitor RO receives the "CRD ACCUM LO PRESS /HI LEVEL" annunciator on Panel A-0 ! Which ONE of the following describes what the hydraulic control unit l (llCU) panel will be indicating if the fault is low accumulator pressure? t At the local !!CU panel the operator must depress the red indicating light for the affected HCU. The fault is low pressure if: a. the light was out and stays out when depresse , , b. the light was out and comes on when depresse , c. the light was on and goes out when depresse , d. the light was on and stays on when depresse i QUESTION: 064 (1.00) With 2 hours remaining in the shift it has been determined that average core megawatts thermal (CMWT) for the previous hour was 2446 CMW , Which ONE of the following is the REQUIRED action? a. Power must be reduced to, or below, 2426 CMWT for at least one hou b. Power must be reduced to, or below, 2426 CMWT for the remainder- , of the shif c. Reduce power to less than 2436 CMWT for the same amount of time it was above that valu d. Reduce power to less than 2436 CMWT-and ensure it does not exceed that limit for the remainder of the shift.- i e I f I _ _ . _ . . _ _ _ _ . . _ . _ _ _ . _ . _ . . _ . _ _ . _ _ . _ , _ _
REACTOR OPERATOR Page 40 QUESTION: 065 (1.00) Which ONE of the following defines " Adequate Core Cooling"? Plant systems are removing sufficient heat from the core such that no Linear Heat Generation Rate ( LliGR) limits are being exceede Plant systems are availai and operating to remove long term decay heat under all post ale combinations of Loss of Electrical Powe The heat removal from the reactor is sufficient to restore and maintain peak fuel cladding temperature below the point of failure, Plant systems are available and operating to remove the sensible and decay heat during a postulated Loss of Coolant Acciden QUESTION: 066 (1.00) Given the following plant conditions on Unit 2:
--
Refueling operations are in progress
--
A spent tuel bundle is dropped and is heavily damaged
-- PROCESS RX BLDG VENT RAD HI-HI annunciator is alarming Which ONE of.the following is NOT an immediate operator action for the above conditions? Stop all fuel movement Evacuate the Reactor Building Verify isolation of Reactor Building ventilation Evacuate the Refueling floor,
________________________-__ - _ _ _ -
_ . _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ . _ . . . . _ . _ _ _ - _ ..__._.. . _ _ . . , REACTOR OPERATOR Page 41 QUESTION: 067 (1.00) Given the following conditions for Unit 1:
--
The plant has just experienced a complete loss of all means of Shutdown Cooling
--
Temperature readings indicate a 1.5 degree F INCREASE in bulk water temperature every 12 minutes
--
Assume the reactor vessel head is installed
--
No other parameters change .
--
Current temperature is 166 degrees F j
.
Which ONE of the following is the time allowed before primary ' containment integrity MUST be established? j minutes minutes i c, 432 minutes minutes
-
'
QUESTION: 068 (1.00) Which ONE of the following cooli:.g water system conditions REQUIRE a manual reactor scram if pressure cannot be immediately restored? Conventional Service Water pressure is steady at 36 psig with one TBCCW Service-Water Supply Valve (SW-V4) in its throttled closed position.- Nuclear Service Water pressure is 25 psig and decreasing and the Diesel Generator Building service water header was left l isolated, All Turbine Building closed Cooling Water system pumps are running and pressure is 45 psig and increasing, d. . All Reactor Building-Closed Cooling Water: system. pumps are-running and pressure is 65 psig and decreasing.
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REACTOR OPERATOR Page 42 QUESTION: 069 (1.00) Following a loss of power the Unit 1 " Torus to Reactor Building Vacuum Breakers" must continue to receive nitrogen for operatio Which ONE of the following describes how the system is operated to accomplish this? The vacuum breaker accumulators will supply nitrogen for all postulated required operation The Abnormal Operating Procedure directs immediate manual opening of nitrogen supply valves, The solenoid valve supplying backup nitrogen to these valves fails open on loss of powe The backup nitrogen supply is unisolated by a check valve that opens as PNS pressure decrease , QUESTION: 070 (1.00) Followinc an evacuation of the Control Room, reactor pressure control on Unit 2 15 by manual operation of the safety relief valves (SRV).
Which ONE of the following is the guidance for operation of the SRVs jn this condition? The SRVs should be used in sequence and each opened until pressure has decreased by 150 psig or 3 minutes have elapsed, Each SRV should be used to maintain reactor pressure less than 1025 psig as necessary until its accumulator is depressurized- SRV sequential operation is used with no more than 3 minutes or an increase of 150 psig between individual valve openings, SRVs should be allowed to control pressure automatically until suppression pool temperature reaches 95 degrees F.
l l l ! l
. .. .- .----.-----.:--..----,---.:.--;.-..,u.--.--.... . - . . . - . - - . . - - - - - . .
_ . . -
REACTOR OPERATOR Page 43 i l l , QUESTION: 071 (1.00) l l
'
Which ONE of the following describes when the Unit 1 Mode switch is REQUIRED to be placed in " Shutdown" during a Control Room evacuation? The Unit 3 Mode switch is placed in " Shutdown": immediately prior to tripping the recirculation pumps j after total steam flow decreases to less than 3.0 E6 lbm/h after reactor pressure decreases to less than 700 psi immediately prior to closing the Main Steam Isolation Valve t QUESTION: 072 (1.00)-
.
Unit 2 is experiencing lowering main condenser vacuu Which ONE of the following conditions allow the operator to INCREASE power in an attempt to correct the lowering vacuum problem? Increasing power to correct the vacuum problem may be used if: main. condenser hotwell level had decreased to less than -7 inches and is now recoverin turbine exhaust hood sprays had been operating to cool the LP turbine hoods at low power level the lowering vacuum occurred simultaneously with an operator initiated generator' load reductio the SJAE Condensate Recirculation Valve had-been throttled open to improve air ejector efficienc , v+, ,c.--, ..wJ,-,,-3,,,, ,,,-.v' m-,,-,-,,,--. ,, .my- - , , ,,m, .,,.-E,w--#, ,, , - , ,, , - .,7.,y,+re-.E---w.,y-m,.-,, , - ,E,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - ._______ --_ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ ___
,
i REACTOR OPERATOR Page 44 l
,
QUESTION: 073 (1.00) v Unit 1 main condenser vacuum is lowering. The Shift Supervisor is '
! -directing operations in accordance with AOP-3 Is a ccortoauca w.'TA T4 APh 3 Which ONE of the following vacuum readings corresponds to the lowest '
value at which Feedwater Pumps will be able to maintain reactor water
'
level? inches Hg vacuum inches Hg vacuum inches Hg vacuum .: .4 inches Hg vacuu pgg 4F MoTT/mJ QUESTION: 074 (1.00) Given the following plant conditions:
--
Unit 1 off-line for refueling outage
--
Unit 2 at 75% power 3
~
--
Severe weather is causing degradation of the grid
-- System frequency is down to 58.3 Hz Which ONE_of the following is the required operator action? Coordinate with the load dispatcher and isolate unnecessary offsite lines.
, b.- When the allowable _ time limit for 58.3LHz is exceeded. increase _ generator output by 10 %. If definite indications of equipment damage are present scram the reactor, trip the turbine and enter EOP - 01, If the' maximum allowable time for 58.3 Hz is exceeded trip the turbin _
-
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REACTOR OPERATOR Page 45 l QUESTION: 075 (1.00) Which ONE of the following in the reason a reverse power trip is not preferred means to take the main turbine off-line? Allowing an automatic reverse power trip of the main turbine: will result in excessive arcing in the main generator output breaker may cause a pressure spike sufficient to rupture the LP turbine relief diaphragm will result in an automatic cold start of that unit's diesel generator may place an unnecessary load on the main turbine thrust bearin QUESTION: 076 (1.00)
"
Following a reactor scram, EOP-01, " Reactor Vessel Control Procedure", has been entered. An IMMEDIATE determination must be made regarding any required transition to the "Lovel/ Power Control" flowchar Which ONE of the following criteria, other than all rods fully inserted, is used to make the determination that the reactor will remain shutdown under all conditions without boron? A Nuclear Engineer calculation of shutdown marg # All rods inserted to or beyond to position 0 All APRM channel "Downscale" alarms in, All IRM channels less than 25 on Range 3 _ _ _ - _ _ - _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ - _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ - _
.. . _ _ _ _ _ _
REACTOR OPERATOR Page 46 QUESTION: 077 (1.00) E0P-02, " Primary Containment Control Procedure", primary containment pressure control path, directs the primary containment to be vente Which ONE of the following is the result of venting via the suppression chamber instead of the drywell? Using the suppression chamber as the vent path will: allow better control of the relence rate due to the sizing of the path's piping and valve allow a more rapid decrease in suppression chamber pressure to below the Pressure Suppression Pressure curve, help absorb some of the high energy in the primary containment by passing it through the suppression pool water, reduce the levels of radioactivity released as it passes through the water in the suppression poo QUESTION: 078 (1.00) With Unit 2 at power a spurious Group 1 isolation signal occur ALL Group 1 isolation valves clos Which ONE of the following signals was the source of the isolation? Main steam line high radiation - Main steam line high flow Main condenser low vacuum Turbine building high area temperature
. . .
__-_ _________ REACTOlt OPEllATOR Page 47 ! QUESTIOll: 079 (1.00) While operating in EOP-02, " Primary Containment Control Procedure", the operator la directed to perform an emergency depressurization when plant conditionu "cannot be restored and maintained in the ' Safe' region of the SitV Tall Pipe Level Limit."
Which O!1E of the following plant conditions must be evaluated along with Suppression Pool level in order to make the decision to depressurize? l<eactor pressure Suppression pool temperature Suppression chamber pressure Delta T he QUESTIOll: 000 (1.00) While at power the controlling Electro-llydraulic Control ( EllC) pressure regulator fails "high" causing the turbine bypass valves to ope Which O!1E of the following is the operator action that will close the turbine bypass valves? Place the non-affected pressure regulator in service and isolate the affected regulato h. Take the Maximum combined Flow Limit potentiometer to the
" decrease" directio Depress the " decrease" switch on the Main Steam Pressure Setpoint devic Place the bypass valves under control of the jack and select "Close".
_ _ _ - - - . - - _ _ - _ _ _ _ _ _ _ . _
._m_ - _ _ _ _ _ _ _ - - .__ _ _ - _ _ _
REACTOR OPERATOR Page 48 * i
;
!
QUEST 10ll: 081 (1.00) .
,
Given Table 2 from EOP-1, 6 Reactor Plooding Proceduro" and the condensato and Condensate-Booster Pumps running and slowly injecting ! jnto the RPV: Number of Pressure open SRVs (psig)
--....--. --------
;
7 100 Table 2 -- 6 115 Minimum Alternate 5 145 Plooding Pressuro 4 180 , 3 245 2 375 1 765
......... ....---- ....... ..... ...... -- -- . . .....----------...---
Which OllE of the following plant conditions will C0!! FIRM adequate core cooling? SRVs open RPV pressure is 100 psig SRVs open RPV pressure is 325 psig , FRVs open RP1 pressure is 300 psig SRV open RPV pressure is 760 psig
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_ - _ _ _ _ _ _ _ REACTOR OPERATOR Page 49
i i QUESTIOll: 082 (1.00) Which O!JE of the following is the basis for o erating within the limits of the Maximum Core Uncovery Time Limit Graph The Maximum Core Uncovery Time Limit curve: is the time water level can be below bottom of active fuel with no cooling and clad temperature will not exceed 1500 degrees is the time limitation imposed for a level indication response once all injection is stopped and water level is lowere determines how long RPV pressure must be maintained above 50 _ poig with Safety Relief Valves open during RPV Floodin provides time to complete RPV flooding prior to exiting this procedure and entering Primary Containment Floodin QUESTIOll: 083 (1.00) Which OllE of the following is the Primary Containment Pressure Limit while operating in EoP-02, " Primary Containment control Procedure"? The pressure corresponding to 68.5 feet of wate The pressure corresponding to 63 feet of water, psig psig _ ___._________.____m___ _ _ _ _ __
_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . . . . . . . . REACTOR OPERATOR Page 50 I i QUESTIO!J: 084 (1.00) Which 011E of the following alternate rod insertion methods REQUIRE the Reactor Protection System to be reset? Control rod insertion by: venting the drive mechanism over-piston are venting the scram air header, using the Reactor Manual Control Syste increasing cooling water header pressur QUESTIO!!: 085 (1.00) Which 011E of the following indications is used to monitor plant heatup/cooldown rate while in Alternate Shutdown Cooling? Recirculation loop suction line temperatur Steam dome pressure using steam table conversion Safety relief valve tailpipe temperature, The running ECCS pump local suction temperatur _
_ __ . _ _ _ . _ . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . ___
.. ___ _ _ _ _ . . . _ _ _ _ . _ _ . . . - . - . . . . ~ _ . _ _ _ _ - __._....._.__.m...,_. _ . . _ _ _ . . . . _ . . . . . ._
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.
REACTOR OPERATO Page 51
,
OUESTION: 086 (1.0s, l The terwerature control leg of EOP-03, " Secondary Containment Control-Proceduta", asks if " a primary system is discharging into.the Reactor-Building".
Which ONE of the following is-'a Primary System as referenced in this step? g A P*imary System is: any plant safety-related systein -required to be operable in Modes 1, 2 and/or a system whose leak rate will decrease as reactor pressure
~ decreases,
' a system required to shutdown the reactor or prc /ide long-term core cooling, any plant system whose pressure-will change with drywell and/or torus pressur l
. G
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REACTOR OPERATOR Page 52 l QUESTION: 087 (1.00) Given the following plant conditions on Unit 2:
--
Loss of coolant accident in progress
--
A Reactor Building entry is needed to complete safe shutdown actions
--
Radiation level in the specific area does not allow personnel access to that area Which ONE of the following describe the Reactor Building radiation levels? Reactor Building radiation levels are: greater than the level necessitating a Stte Area Emergency, beyond the levels required for an emergency depressurizatio above the maximum acceptable operating limi in excess of the maximum safe operating valu QUESTION: 088 (1.00) Which ONE of the following sets of conditions REQUIRE entry into eor 92,
"Pri try Containment Control Procedure"? Primary containment oxygen concentration is 4.45% and hydrogen concentration is 1.25%. During the High Pressure Coolant Injection system operability test suppression pool water level reaches -28 inches, During the Reactor Core Isolation Cooling operability test suppression pool temperature reaches 107 degrees Malfunctioning drywell coolers require drywell venting to maintain pressure steady at 1.65 psi _ _ _ _ _ _ -_ _ _ _ _ _ __ _ _ _ _ _ _ _ - _ _ _ - _ _ __-
REACTOR OPERATOR Page 53 QUESTION: 089 (1.00) Unit 1 is performing a startup following a reactor scra Reactor Water Cleanup (RWCU) is lined up to reject to the condenser to lower RPV water leve Which ONE of the following parameters limits the RWCU reject flow rate for these plant conditions? Nonregenerative heat exchanger outlet temperatur Regenerative heat exchanger outlet temperatur Reject Flow Control Valve downstream pressur _ _ _ Filter-demineralizer differential pressure.
QUESTION: 090 (1.00) The following conditions Exist on Unit 2:
--
A failure-to-scram (ATWS) condition exists
--
Reactor power is 21%
--
Standby Liquid Control (S LC) system pumps "A" and "B" are In]ecting
-- "A" SLC Squib valve will not fire Which ONE of the rollowing is the required operator action for this condition? SLC should be secured to prevent damage to check valves and -
accumulators Shutdown 1 SLC pump and monitor system parameters Secure SLC and enter " Alternate Boron Injection" LEP-03 SLC should be shutdown to prevent damage to pump seals
. _ _ _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ -
- . . . . _ . _ . _._ . _ _ _ _ .__.__ _ . . _ _ _ _ _._.._ .._ m .._ . . _ _ _ . . _ . _ _ _ . . . . . _ _ _ __
-REACTOR OPERATOR Page.54 QUESTION: 091 (1.00)
A failure to scram (ATWS) has occurred and reactor power is approximately 35%. Which ONE of the following actions should be taken to prevent a main turbine trip on high reactor water level for these plant conditions? Place the Recirculation Pump MG Set control switches to "Stop". Place both Recirculation Pump speed controllers to 10%. Place both Recirculation Pump speed controllers to 28%. Place both Recirculation Pump speed controllers to 45%. QUESTION: 092 (1.00) EOP-02, " Primary Containment Control Procedure", Step SP/L-36 directs securing "HPCI 1rrespective of adequate core cooling" if suppression pool level cannot be maintained above -6.5 fee Which ONE of the following describes the affect of continued operation of the High Pressure Coolant Injection (HPCI) system? Continued operation: will cause HPCI turbine-damage due to the high temperature cooling water to the oil system, will cause damage to the turbine exhaust check valve due to
high steam flow.
'- will reduce suppression pool level below the1 safety relief valve T-quencher result in the failure and loss of the-final fission product barrier.- i i n
.-. , , . . , _ _
._ ,
. _ _ _ , _ _ _ . - _ _- _ ._._ _ . _ , . _ _ _ . _ _ . . _ _ _ _ _ . _ .. _ ~.. . _._, _ _ _. . _ . _ _ - . . . _ .
I REACTOR _ OPERATOR -Page 55
= QUESTION: 093_ (1.00)
Given_the following plant conditions on Unit 2: ,
--
A high energy break has just occurred in the -Reactor Building
-- -Reactor Building area radiation monitors are alarming --
Reactor Building ventilation exhaust rad level is 6 mr/hr
--
The Shift Supervisor has entered EOP-03, " Secondary containment Control" and AOP-0 Which ONE of the following is NOT an immediate operator action for these conditions? _ Verify Reactor Building HVAC in operation, Within 15 minutes isolate the Reactor Building sprinkler-system, Control access to the affected are Ensure E&RC survey and post the area QUESTION: 094 (1.00)
'Which ONE of the following is the maximum _ load allowed on a diesel-generator while operating with the emergency buses _ cross-tied between the units during a loss of off site power? KW ' KW KW ' KW-(
.
. l- !' l j .. l- _-a._._-
. _ ,. _._ ___ _ _ _ . _-. .- _ _ _ . _ m . . . . __ - . . . _ _ - _ .
u REACTOR OPERATOR _ Page 56~
. QUESTION: 095 (1.00)
.
, Ten-(10) valves in a Tech Spec system are undergoing timed stroke testin Which ONE of the following actions must be.taken if the one of the , valves exceeds in maximum allowed stroke time? Stroke the valve a second time to provide confirmation of the stroke tim Immediately declare the valve inoperable and generate the required LCO Exercise the valve through several stroke cycles then stroke
.once more for tim Log the stroke time, complete alll valve testing and then take-required corrective action QUESTION: 096 (1.00)
With the plant at 65% power a failure in the "B" Recirculation Pump controls causes a slow speed increase.
- -
Which ode of the following describes when the "B" Recirculation-Pump is H required to be tripped by the AOP for the above conditions? The "B" Recirculation Pump will'be tripped: if its speed continues to increase after the scoop-tube is unlocke as soon as-the speed di'fference between the-Recirculation ~ Pumps exceeds 15%. if more than two APRM or-IRM Upscale alarms are received __ simultaneously.
, if locking the scoop tube on the pump does not stop the speed L increase.
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,
. .. .~ .., . . -.. - . . . . . _ - . _ _ . . . - ,_- -
I
REACTOR OPERATOR Page 57 QUESTION: 097 (1.00) The plant is at normal operating temperature and pressure when a break-occurs in the Core Spray piping inside the reactor vesse Which ONE of the following indications will confirm this break has occurred? The core plate - core spray sparger differential pressure. indication will: move from negative pressure towards positive pressur move from positive pressure towards negative. pressur move from negative pressure towards more negative pressur more from pos).tive pressure towards more positive prassur QUESTION: 098 (1.00)
-Unit 2 has experienced a loss of 480 V MCC.2X Which ONE of the following is the effect on the Standby Liquid Control (SLC) system? Adding' neutron absorber solution to the storage tank.will not be possible due to a loss of the 40 kW heate Boron may start-to precipitate out of solution due to the loss of the piping heat trace syste _ The "A" Squib Valve will not fire if'the pump start switch is placed in " Pump A1 Start".
L The "A" Squib Valve will fire if the pump start switch is l placed in " Pump A Start".
l-l !~
-
'
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- .
v- __ REACTOR: OPERATOR Page_58 f QUESTION: 099 (1.00) Which ONElof the following is the purpose of the " white"' indicating light associated with each Core Spray Pump? . The " white" light indicates: that pumps' 4160 VAC breaker is closed with control power availabl the pump was stopped while an " Initiation Signal" was still presen the pumps' 4360 VAC breaker was locally closed with emergency diesel generator power being supplie the pump has tripped on fault with a valid initiation signal presen QUESTION: 100 (1.00) Unit 2 is in a condition requiring an Emergency Depressurizatio The Emergency Operating Procedures direct all 7 Automatic Depressurization System (ADS) safety relief valves (SRV) be opene Which ONE of the following is the MINIMUM number of SRVs that must be opened to meet the requirements for an Emergency Depressurization? . (********** - END OF EXAMINATION **********)
, ' ,j pc sn A.s72 / * REACTOR OPERATOR Page 1 ANSWER- -KEY MULTIPLE CilOICE 023 d-001 b' 024 a-002 be 025 de !
003 o' 026 a-004 dr 027 c- !
- 005 b< a8 c, -
006 a/ 029 . i 007 de 030 a- ' 008 b< 031 b-009 a ' 03'2 ar-
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. - - - . . - - . . - . . . . . - . - . . . . - . - . - - - . . . . - - - - . ..
A-N S W E R:
'
K-E Y-046 a- 069 c-047 d- 07 a-048 c- 071 a-049 b- 072 c-050 d- 073 b~ 051 a< 074 d-052 a- 075 c-
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060- .d7 -083 c&
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063 _d- 086 -b-064 a# 087 d.- 065 c' 08 c-066 -b'- 089 av 067- d 090 : b,- 068- ar 091 b.-
-.
L 1
. _ _ _ . * REACTOR OPERATOR Page 3 ANSWER KEY l
092 db 093 a' 094 c-095 b-096 a< 097 ar 098 cr 099 b< 100 b-i (********** END OF EXAMINATION **********)
... . _ . - . . . . _ . . _ _ . REACTOR OPERATOR 'j2 (. Page 59
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' ANSWER: . 001 (1.00) REFERENCE:
OI-01, " Operating Principles and Philosophy", Rev. 047, Page 25 OSM 7E, " Conduct of Operations", L.O. -2 (2.7/3.7) 294001A103 . . (KA's) ANSWER: 002 (1.00)
)' REFERENCE:
10 CFR 26.20, " Written Policy and Procedures"
- No Facility Specific Procedure or Learning Objective Identified [2.7/3.7) . 294001A103 . . (KA's)
ANSWER: 003- (1.00)
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REACTOR OPERATOR Page 60 REFERE! ICE: 01-05, " Annunciator Status", Rev. 020, Pages 1 & 2 OSt4 7E, " Conduct of Operations", L.O. - 5
[4.5/4.3)
294001A113 ..(KA's) ]E A11SWER: 004 (1.00)
~ REFERE!1CE:
01-05, " Annunciator Status", Rev. 020, Page 6 OSM 7E, " Conduct of Operations", L.O. - 5
[4.5/4.3]
294001A133 ..(KA's) A11SWER: 005 (1.00)
- _ . _ _ _ _ . . _ . _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___
h REACTOR OPERATOR- Page 61-REFERENCE:- 01-13, " Valve & Electrical Lineup Administrative Controls"-, Re .034, Page 2 No Facility Specific Learning Objective Identified _
(3.9/4.5]
i 294001K102 ..(KA's) ANSWER: 006 (1.00)
=\ ' REFERENCE:
01-13, " Valve &' Electrical Lineup Administrative Controls", Re , Page 7 No Facility Specific Learning Objective Identified (.3.7/3.7) l 294001K101
~ ..(KA's)
ANSWER: 007 (1.00) REFERENCE:' 10CFR50.54(x,y)= No Facility Specific Procedure or Learning Objective Identified
[3.3/3.4]_
294001A111 ..(KA's) i
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bu REACTOR OPERATOR = - Page 62-ANSWER: -008 (1.00)
- REFERENCE:
SD-41 Rwev. 11 page 24-25 SSM 32-A'Rev. O page 52 No Facility Learning Objective Identified (3.5/3.8) 294001K116 ..(KA's).
L ANSWER: 009 (1.00) l a.
ll
' REFERENCE:
AI-58, " Equipment Clearance Procedure", Rev. 38,1 Page 21 OSM 7D2, " Administrative Documents", L.O. - 7
[3.9/4.5)
294001K102 ..(KA's)= ANSWER: 010 (1.00) b.
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-
(Y S ?s REACTOR. OPERATOR p J,I D M6A* 5 Page 63 O&
REFERENCE: g,P (
-
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sg OSM 7DS,~"ELRC_,~ Documents", Page 2, L.O. - 3 [N' 07'2 ' g f4 .g i
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_ (3.?/3.6] % g, o '~D A M
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294001K113 ..(KA's) h dDE gA i//6-h2-2 *%' %s ANSWER: 011 (1.00) .q% cr , g c ,f' t w . J REFERENCE: AI-58, " Equipment Clearance Procedure", Rev. 36, Page 29 OSM 7D2, " Administrative Documents", [3.9/4.S] 294001K102 ..(KA's) ANSWER: 012 (1.00) REFERENCE: 10CFR20.101, " Radiation Dose Standards for Individuals in Restricted Areas" No Facility Learning Objective Identified (3.3/3.8) 294001K103 ..(KA's) ANSWER: 013 (1.00) > $
-- . _ ~ . , , _ . . . . . . . . -. . , - - . -
_ . , . - - - , , , - . - .
,
'l-REACTOR OPERATOR Page 64-LREFERENCE:
SD-08, "CRD Hydraulic System'!, Rev. 016, Page-3 OSM 09-B, " Control Rod Drive Hydraulic System", .f '
(3.1/3.0)
..(KA's)
'
201001K408
' ANSWER: 014 (1.00) I REFERENCE: ,
SD-07, " Reactor Manual Control System", Rev. 008, Pages 4 & 15 OSM 9A, "CRD Mechanism", Page 31 & Figure 6, d
~ [3.6/3.7]-
201003K404 ..(KA's).
ANSWER: 015 (1.00) ..
.- ! . ... , _ _ . . - , .. . . , , . . . , , _ . _ . , , - . . . _ _ . . , . , _ , , , , - _ ,
_ _ _ . . _ _ _ . _ - ._, .
.lREFERE! ICE: =
OSM 9B,-"CRD", L.O. - 6c,6f,6g-
[3.3/3.3)
201003K601 ..(KA's)
.
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'
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.--- . ~ --=~ -. ~. .-
REACTOR OPERATOR -Page 65 ANSWER: 016 (1.00) a.
' " REFERENCE: SD-07, " Reactor Manual Control System", Rev. 008, Pages 7 & 8 _ OSM 27-2A, " Reactor Manual Control System, Pages 26 & 27, * 6a l
,
[3.2/3.3]
:
201002K106 ..(KA's) :
!
!
ANSWER: 017 -(1.00) ! REFERENCE: . SD-07, " Reactor Manual Control System", Rev. 008,- Page-5-
~
OSM 27-2A, " Reactor _ Manual Control System", Pages 26 & 27, [3.2/3.1] 201002A301 ..(KA's) ANSWER:- 018 (1.00) , , . - - . ,
REACTOR OPERATOR - Paga:-66 REFERENCE: SD-32.2, Feedwater Control- System", Rev. 008, Pago-17-
.
OSM 27-2B, " Rod Worth Minimizer",_Page 17, _-5
[3.1/3.2)
201006K104 ..(KA'a)
ANSWER: 019 (1.00) C,
-REFERENCE:
SD-02, " Reactor Recirculation System", Rev. 021, Pages 23 & 24 ') i OSM 10-2A, " Reactor Recirculation Systems", Figure 15, _ L.O. - 9, 10
& 12 (3.9/3.8]-
202001A41 ..(KA's) ANSWER: 020 (1.00) .j-l l
0 u-
.
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, i 1. __ m .,.i___ ,,. .. ~ . - . . . - - - . . . - ~ . - . ~ . . . . . - . . . - , . . . .
REACTOR; OPERATOR Page 67
:RFFERENCE: '
Brunswick Unit 2 Tech Spec 3.4.1.4 and bases, Pages'3/4'4-2 & B 3/4 4-1 OSM 10-2A, " Reactor Recirculation Systems", Figure 15, L.O. - 34 ,
[3.5/3.7)
202001K601 ..(KA's) , , ANSWER: 021- (1.00) REFERENCE: , l l OSM 10-2A, " Reactor Recirculation Systems", . Page 53, ' 3 [3.2/3.2) 202002A101 ..(KA's) ANSWER: 022 (1.00) 1.
l' c.. l- ! REFERENCE: i-i l 2 APP A-06 S-1, Rev. 8, Page 59 i ' OSM 10-2A, " Reactor Recirculation Systems", L.O. - 17.c l- . .
[3.2/3.3] '
, !. 202002K305 ..(KA's)-
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. - _ _ - _ _ _ _ -- _ _ _ _ _ _ _ __ . - _ _ _ _ _ _ _ _ _ _ _ _ - _ . ___- __. . _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ _ _ _ _ - _ _ _ - _ _ - _ _ - _ _ _
: REACTOR OPERATOR Page.68 _
-=
ANSWER: 023 (1.00)
- REFERENCE:
2 OP-02, " Reactor Recirculation System", Rev. 76, Page 9 BTU Exam Bank Question =(Examiner Modified), No Facility Learning Objective Identified
[3.5/3.7)
. 202001G010 . . (KA's)
' ANSWER: 024 (1.00)
-I . REFERENCE:
2 OP-19, " High' Pressure Coolant Injection System", Rev. 73, Page 15 OSM 14B, " High Pressure Coolant Injection", L.O. - 4
-[ 3. 9/3. 8 ] _ .206000A303: . . (KA's)
ANSWER: 025 (1.00)
' M
4 4Le ,W 1 m,m . ra - w w'- - .e-e w -t gu----mwg e,ir- er ,w wm 4.-mwe ,,+--p- w ->- .>,wir+ we in - f-f- =-+w e-+--
_ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ REACTOR OPERATOP Page 69 I REFERENCE: 2 OP-19, "lligh Pressure Coolant Injection System", Rev. 73, Pages 8, 10 & 34 OSM 148, "lligh Pressure Coolant injection", (4.1/4.2) 206000G004 ..(KA's) ANSWER: 026 (1.00) REFERENCE: OSM 14G, " Standby Liquid Control System", Rev. O, Page 10, .n
[3.7/3.8)
211000A108 ..(KA's) ANSWER: 027 (1.00) REFERENCE: OSM 28-2-A, " Reactor Protection System", Re , Page 46 L.O.-10
[4.0/4.1)
212000A216 ..(KA's) ANSWER: 028 (1.00) I I _ _ __-____________--__ _ ____ - _ _ _ __ _ _ -
-
. . . . . . . REACTOR OPERATOR Page 70 REFERENCE: OSM 28-2-A, " Reactor Protection System", Rev. 3, Page 47, L.O. - 10 (4.2/4.2] 212000K408 ..(KA's) ANSWER: 029 (1.00) REFERENCE: SD-09.2, " Traversing In-Core Probe System", Rev. 004, Page 5 OSM 25-F, " Traversing In-Core Probe System", Sb (3.4/3.5] 21S001K401 ..(KA's) ANSWER: 030 (1.00) REFERENCE: OSM 25-A, " Source Range Monitoring System", Re O, Page 14, L.O. - 3a (3.6/3.6) 215004A304 ..(KA's) ANSWER: 031 (1.00) _ _ _ _ - _ - - _
REACTOR OPERATOR Page 71 REFERENCE: OSM 25-D, " Average Power Range Monitoring System", Rev. O, Pages 4, 5& 28, L.O. - 3b (3.6/3.6) 215005K505 ..(KA's) ANSWER: 032 (1.00) a.
REFERENCE: _ BTU Exam Bank Question #25-D, 387 (Examiner Modified) OSM 25-D, " Average Power Range Monitoring System", Rev. O, L.O. -8
& 10
{3.0/3.4)
215005A107 ..(KA's) ANSWER: 033 (1.00) REFERENCE:
-
-
OSM 26-2A, " Instrumentation", Rev. 5, Pages 6 & 7, No Facility Learning Objective Identified BTU Exam Bank Question #26-A, 969 (Examiner Modified)
{3.6/3.8)
216000K507 ..(KA's)
...._ _ _
. _ ._.- . . . _ _ . . . . . . . _ _ . . . - _ - . _ - . _ . _ . _ _ . _ _ . . _ . _ . _ - . . _ . . -_ _ _ _ . _ . _ . . _ . . _ _ . . _ _ .
l REACTOR OPERATOR 'Page 72 ANSWER: 034 (1.00)- REFERENCE: 'l l l NRC Information Notice 92-04 Dated 8/18/92 " Effects of Non- l condensable gases on BWR cold leg water level instruments" l No Objective Identified I (3.4/3.6] i 216000K506 ..(KA's) l ANSWER: 035 (1.00) REFERENCE: OSM 14-C, "RCIC", Re , Pages 10, 18 & 21, L.O. -8 ,
[3.4/3.5)
l 217000K601 ..(KA's) ANSWER: 036 (1.00) _
.
C.
e
h
. _ . , ~ . . , . _. . - - _ . . _ . , . . - . , _ . . - _ . _ . _ . . . , , _ , - . . . . . . . . . . _ . . , . - - - - , . - - . -, - , _ .
REACTOR OPERATOR Page 73 REFERENCE: SD-16, " Reactor Core Isolation Cooling System", Rev. 17, Pages 9, 14 & 17 No Facility Learning Objective Identified BTU Exam Bank Question #14-C, 1899 (Examiner Modified)
[3.7/4.0)
217000G001 ..(KA's)
-
ANSWER: 037 (1.00) b.
REFERENCE: OSM 14-F, " Automatic Depressurization System", Rev. O, Page 14, L.O. - 6 SD-20, " Automatic Depressurization and Safety / Relief Valve System", Rev. 014, Page 6
[4.2/4.3)
218000A308 ..(KA's) ANSWER: 038 (1.00) i _ _ _ _ _ _ _ _ - _ - - _ _ - _ _ _ - - - _ __ - . _ _ _ _ -
. .. ..
. REACTOR OPERATOR Page 74 REFEREllCE: SD-20, " Automatic Depressurization and Safety / Relief Valve System", Rev. 014, Page 10 OSM 14-F, " Automatic Depressurization System", L.O. - 7.e & 10.b.(1)
[3.7/4.0)
218000A102 ..(KA's) ANSWER: 039 (1.00) REFERENCE: SD-20, " Automatic Depressurization and Safety / Relief Valve System", Rev. 014, Page 4 OSM 14-F, " Automatic Depressurization System", Rev. O, Page 16, L.O. - 1 [3.0/3.2] 239002K605 ..(KA's) ANSWER: 040 (1.00) . . _ _ _ _ _ _ _ _ _ _ _ _ _ ____
____ _ - _ _ _ _ _ _ _ _ _ - _ - _ - _ _ __ _ _ __ _ __ _ _ _ _ __ _ - _ - - - _-__ _ _-_-___ REACTOR OPERATOR Page 75 REFERE!1CE: ' O OP-50.1, " Diesel Generator Emergency PoWor Syr, tem Operation", Rev. 37, Page 95 OSM 20-20, "Diosul Generators", L.O. -3 (3.5/3.6) _ l 264000A201 ..(KA's)
.
,
A11SWER: 041 (1.00) , REFEREllCE:
.
SD-39, " Emergency Diesel Generator System", Rev. 10, Page 61: OSM 20-20, "Diosol Generators", Rev. 5, Pages 17 & 41, L.O. - 23 (3.4/3.5] i d 264000K506 ..(YA's) ANSWER: 042 (1.00)
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_
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. _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ - - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _
;
REACTOR OPERATOR Page 76 ,
! -REFERENCE:
l
.
!
SD-10, " Standby Gas Treatment System", Rev. 009, Page 4 i OSM 15-F, " Standby Gas Treatment", L.O. -4 .
(3.7/3.8) ,
;
261000K401 ..(KA's) ; ANSWER: 043 (1.00) , J REFERENCE: -i 2 OP-17, " Residual lleat Removai System", Rev. 99, Page 19 , OSM 14-2D, " Residual lleat Remov.sl", L.O. - 3-& 8 (3.8/3.7) . 203000A304 ..(KA'a) ANSWER: 044 (1.00) REFERENCE: SD-17, " Residual licat Removal. System", Rev. 017, Page 14 OSM 14-3D,." Residual llent Removal", L.O. - 10 (3.5/3.5) 226001A407 ..(KA'a) i I , v c,,. . . _ , , . . _.,,-._,-,..,.M___,m,-,_,__,_,,,_m,,,,,__,,,i_,_ , , . , _ . , - , , _ , _ _ . . , - - . . . _ , , . _ , , , , . . . . , _ , , , , , _ , , , , , , . _ _ ,-~_J~,,
._ ,,,__.i.m,, _ ,,.;__._..,.,,, REACTOR OPERATOR Pago 77 n
A!1SWER: 045 (1.00) REFERE!!CE:
"
)
SD-14, Reactor Water Cleanup System", Rev. 012, Page 12 ! OSM 11A, " RWCU", l (3.6/3.6) 204000A303 . . (KA's) A11SWER: 046 (1.00) REFERE!1CE: , 2 OP-14, " Roactor Water Cleanup System", Rev. 82, Page 13 OSM 11A, " RWCU", (3.4/3.3) 204000G007 . . (KA's) ANSWER: 047 (1.00) _
-
e , - ,myw -, -~,-g--prw-e-w-* e- ,4-m'm=W--- -rw- 4 m w- w -,e ee - - -e-- T- 4-- m- -+c+ft<v wW=v+-w**yM%- w h r'r-M' T
. _ _ _ _ _ _ _ ___ _. _ .. _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ .
REACTOR OPERATOR Page 78
= REFERENCE: j l " . SD-09, Power Range Her * Monitoring System", Rev. 009, Pages 22 & 23 OSM 25-E, " Rod Block Monitor System", Re O, Page 6, .e j (3.8/3.8) l l
215002G007 ..(KA's) E
- ANSWER: 048 (1.00) .i i REFERENCE:
SD-09, " Power Range Neutron Monitoring System", Rev 009, Pages 48 '
& SO OSM 25-E, " Rod Block Monitor System", Re O, Pages 17 & 19, & (3.6/3.5)
-- 215002A304 ..(KA's) ANSWER: 049 (1.00)
, >
$
J l l _ .___
' . e --i -- E. w , . .-:2 ~.-m.,--..--.m.-.-- #--.~. c- 3-. -o-4,- ,es,c..,-,.wn-. e r_~.e.rv.c.-m.-.-..-4_.-s..~ -._-.w-,.+ ,. w. e4 m --rw,. er ++.m.,. ,
. _ _ _ _ _ _ _ _ _ _ _ _ _ _
REACTOR OPERATOR Page 79 REFEREllCE: Fil-11, " Fuel llandling Procedure", Rev. 045, Pages 14 & 15 BrtinGWi c}; Tech Spec 3.9.2, Page 3/4 9-3 OSM 29-A, " Fuel llandling", Rev. 4, Page 46, L.O. -7
[3.7/3.9)
234000A401 ..(KA'n) A!1SWER: 050 (1.00) -- d, REFERE!1CE: OSM 29-A, " Fuel llandling", Re , Page 37, L.O. -5
[3.4/3.8)
234000G001 ,.(KA's) A!1SWER: 051 (1.00) _ . . .
. - -i....m
REACTOR OPERATOR Page 80 I REFERE14CE: I - SD-01, "tiuclear 13 oiler System", Rev. 025, Page 4 OSM 8-A, " Reactor Vessel Internals", Re , Pages 19 7 20, & 5
[3.2/3.3)
290002K403 ..(MA's) AliSWER: 052 (1.00) __ REFEREllCE: OSM 20-H, "4160 AC Volt System", Rev. O, Pages 12 & 13, tio Learning Objectives in the OS [3.4/3.5) 262001G007 ..(KA's) AliSWER: 053 (1.00) _ REFEREliCE: SD-041 " Fire Suppression System", Rev. 11, Page 23 pep 013, Rev.10 page 7
[3.4/3.4)
286000A301 ..(KA's) _ _ _ _ - _ _ _ _ - _ _ _ _ -
REACTOR OPERATOR Page 81 ANSWER: 054 (1.00) a.
REFERENCE: SD-25, " Main Steam System", Rev. 12, Page 5 OSM 17-2A, " Main Steam", Re , Pages 12 - 14, L.O. -
(4.0/4.1)
239001K127 ..(KA's) ANSWER: 055 (1.00) b.
REFERENCE: SD-11, " Process Radiation Monitoring System", Rev. 4, Page 39' OSM 17-2A, " Main Steam", L.O. -3 (3.6/3.8) l 272000K101 ..(KA's) i
'
ANSWER: 056 ( l'. 00 ) '!
!
-
. .
-1 REACTOR OPERATOR Page'82 ,
,
REFERENCE: !
!
2 OP-26, " Turbine System Operating Procedure", Rev. 55, Page 18 a OSM 18-2A, " Turbine", [3.6/3.7) 245000K104 . . (KA's) ; ANSWER: 057 (1.00) c
+ REFERENCE: .
,
2 OP-17, " Residual Heat Removal System", Rev. 97, Page 86 OSM 14-2D, " Residual lleat Removal", Re , Pages 28-& 29, L.O.--
[3.7/3.5)
,
219000A402 . . (KA's)
. - ANSWER: 058 (1.00) REFERENCE:
OSM 17-2B, " Condensate and Feedwater System", _Re , Page 2,. &
(3.2/3.2]
256000A206 . . (KA's) L j .- l l-l
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. _ _ ,_ _ _ _ -. _ . . __._. _ . -
.-_ _ _ _ _ _ __.. _._ _ __m_ - - _.. .m. REACTOR OPERATOR Page 83 I
. ANSWER: 059 (1.00) REFERENCE: . I
\
2 OP-30, " Condenser Air Removal & Off-Gas Recombiner System", Re l 48, Page 8 ' OSM 17-2D, " Condenser Air Removal", L.O._- 5 *
[3.1/3.2) i 271000G010 ..(KA's)
ANSWER: 060 (1.00) ,
,
! REFERENCE:
'
SD-37, " Control Building licating, Ventilation and Air-Conditioning system", Rev. 005, Page 9 OSM 36B, " Control Building !!VAC", d
.[3.1/3.2)
i 290003K401 ..(KA's) ANSWER: 061 (1,00) j_ d.
l l r I
, '
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REACTOR OPERATOR Page 84
- REFERENCE:
AOP-04.0, " Recirculation Flow Control Failure - Decreasing Flow", Re , Page 4 OSM 10-2A, " Reactor Rocirculation System" , .a (3.8/3.7) l l 295001G010 ..(KA's) l ANSWER: 062 (1.00) , i , . REFERENCE: Brunswick Tech Spec 3.4.1.1, Figure 4.1.1-1, Page 3/4 4-lb OSM 10-2A, " Reactor Recirculation System", (3.5/3.8)
.
295001A201 ..(KA's) , i ANSWER: 063 (1.00) d.
l_ . I ! i e
6
, "* n~s- v - ,+-s ,,,c,-e--,~ g ,-~-- ~- ,- we ~-nwnw,..w.-w~vw,~rawn~~--wn.~ r,e,--- -w6, wa w s,- era - , *- - - , e e- - , e-,- ~w,--
l REACTOR OPERATOR- Fage 85 REFERENCE: l APP A-07 6-1, Rev. 8, Page 46 OSM 09-B, " Control Rod Drive Hydraulic System", [3.3/3.2) ! I 295022G009 . . (KA's) . l ANSWER: 064 (1.00) l , REFERENCE: 2 OP-02, " Reactor Recirculation System", Rev. 68, Page 35 No facility Specific Learning objective Identified
, [4.1/4.2] :
295014A201 . . (KA's) ANSWER: 065 (1.00) ;
. '
REFERENCE: s d OSM 07-2K.01, " Primary Containment Control", Rev. 5, Page-160,- .1
[4.0/4.1) ;
, - 295014A107 . . (KA'a)
,
i
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REACTOR OPERATOR 'Page 86 ANSWER: 066 (1.00) REFERENCE: AOP-07.0, " Spent Fuel Damage ', Re , Page 3 tio Facility Specific Learning Objective Identified (3.8/3.9) 295023G010 ..(KA's) l ANSWER: 067 (1.00) I REFERENCE: l Brunswick Tech Spec Table 1.2, Page 1-11 i Brunswick Tech Spec 3.6.1.1, Page 3/4 6-1 >j OSM 15-2A, " Primary Containment", L. O . '- 8 & 13
[3.6/3.8) ;
295021K101 ..(KA's) l
.
ANSWER: 068 (1.00) I _ __
,
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:
REACTOR OPERATOR Page 87 [ l REFEREllCE: !
:
'
AOP-19.0, " Conventional Servico Water System railuro", Rev. 005, '
'
Pago 3 11o Facility Specific Learning Objective Identified
[3.4/3.6) 4
,
t 295018K202 ..(KA's) s
's A11SWER: 069 (1.00) >
,
REFEREliCE:
,
SD-46, " Instrument And Servico Air System", Rev. 009, Page 11 OSM 21-2A, "Comprensed Air", Re , Pago 24, L.O. - 12
- (3.5/3.3)
295019A101 ..(KA's) ANSWER: 070 (1.00) L l l
:
:
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-!
REACTOR OPERATOR Page 88 !
'
l REFERE! ICE: l l
AoP-32.0, " Plant Shutdown From outside control Room", Rev. 19, Page l
'> l i
llo Facility Specific Learning Objective Identified
[4.0/4.0) ;
!
295016A108 ..(KA's)
!
AllSWER 071 (1.00) l
' \
REFEREliCE:
,
!
AOP-32.0, " Plant Shutdown From outsido control Room", Rev. 19, Page 3 , lio Facility Specific Learning Objective Identified 1
{3.8/3.6)
295016G010 ..(KA'n) i A!1SWERt 072 (1.00)
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. - . _ . . _ _ . . - - . . - -.__._.__._ _ _ _ _ _ . _ _ _ - _ - - - - . . _ _
l HEACTOR OPERATOR page 89 REFERE!JCE: AOP-37, " Low Condensor Vacuum", Rev. O, Page 5 , APP UA-23 3-1, "Exh flood B Vacuum Low", Rev. 23, Pages 38 & 39 fio Facility Specific Learning Objective Identified (3.2/3.2) i i
295002G005 1 l
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..(KA's)
A!!SWER: 073 (1.00) REFERE!JCE: AOP-37, " Low Condenser Vacuum", Rev. O, Page 5 OS!4 17-2A, "!iain Steam", L. O . -
[3.5/3.6)
295002K203 ..(KA's) AllSWER: 074 (1.00) REFERElicE: AOP-22.0, " Low System Frequency", Re , Page 3 11o racility Specific Learning Objective Identified (3.9/4.1) 295003G010 ..(KA's)
. . . . - . . . .. . . - . - . - - . - - -.. - - - - . - -. ... _- - .. _. - - ~.
,
REACTOR OPERATOR Page 90 i
'
,
!
ANSWER: 075 (1.00) i ! i i REFEREllCE: EOP-01, " Reactor Scram Procedure", Step RSP-006
,
01-37, " Preparation and Review of the PSTG", Rev. 017, Page AA-5 { OSM 18-2A, " Main Turbine", L.O. -5
[3.2/3.3) i 295005A208 ..(KA'a)
ANSWER: 076 (1.00) REFERENCE:
!
COP-01, " Reactor Vessel Control Procedure", Rev. O, Flowchart Step 006 f OSM 07-2K.05, " Reactor Vesual Control Proceduro",.L.O. - 12 (3.8/4.4) 295006G012 ..(KA's) ' ANSWER: .077 (1.00)
, _ _
I-i e .~ . .,.- ;....._, .. . , _ . . . . - - - . . . . . _ _ - . - _ , _ _ ~ _ . . . _ . - . - . . . . . _ _ . . . . . _ _ . - - . . . . . . _ _ . _ . . - . . , . - . _ . - . - - . . -
_ __.. _ - _ _ . ._.. _ _ _ __ _ . _ . _ -._ - _ _ _ _ _ _ _ __._ _ .._ - ._.. _ .. _ ._.._ _ _.._ . ! REACTOR OPERATOR Page 91 '
:
:
REFERENCE: ;
'
EOP-02, " Primary Containment Control Procedure",- Rev. 4C, Flowchart Step PC/P-17
,
OSM 07-2K.08, " Supplemental Emergency Proceduro", Re , Page 12, (3.8/4.0) 295010K301 ..(KA's) ANSWER:- 078 (1.00)
- REFERENCE:
OSM 15-2E, " Primary containment Isolation System", Rev.-2, Page 5,- L.O. -4 & (3.4/3.8) 295020A206 ..(KA's) . ANSWERt 079 (1.00) , i
?
i
+- ,. ,w--. . . - - - y
- - . . - - -. - - . . - - . _ ~ - ~ . . _ - - -
_ - . - - - - _ - - . .
- - - . . . - - . - . - _ ,
REACTOR OPERATOR Page 92
REFERENCE: l EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchart Step SP/L-20 .l l Containment Limits, SRV Tail Pipe Level Limit OSM 07-2K.01, " Primary Containment Control", L.O. - - {3.5/3.6) 295029A202 ..(KA's) ANSWER: 080 (1.00) REFERENCE: , SD-26.2, " Electro-Hydraulic Control System, Re , Pages, 4, 5& .1 15 l
<
'
OSM 19-28, " Electro-Hydraulic Control Electrical", L.O. -6
[3.7/3.8) i 295007A105 ..(KA's) .
.
ANSWER: 081 (1.00) ; REFERENCE: OSM 07-2K.11, " Reactor Flooding Procedure", Re , Pages 40 & 41, L.O.'- 8 (4.6/4.7) 295031K101 ..(KA's) _ _
-wm-g w-*,-.mme --W+++rir'--, ww*% *wy-w-v"y - w ---
+-w 4 v-wg+-' * = w- '
*-w' *
. _ . _ _ . _ _ _ _ _ . . _ . _ _ . - - . _ - - . - - _ - _ . - _ _ . - - - - - - - - - > - - . - - -
REACTOR OPERATOR Page 93 l r ANSWER: 082 (1.00) ,
,
REFERENCE:
'
EOP-01, " Reactor Flooding Procedure", Re , Flowchart Figure 1 OSM 07-2K.11, " Reactor Flooding Procedure", Re , Page 37, .d ,
[3.7/4.0)
295031G007 ..(KA's) . ANSWER: 083 (1.00) REFERENCE: EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchart Step PC/P-18
- OSM 07-2K.01, " Primary Containment Control", Re , Pages 139 &
140, .h __ _
(4.2/4.4)
295024A201 ..(KA's) ANSWER: 084 (l'. 0 0 ) d.
p
. . . . _ . . , - - _ - _ . _ - , _ . - . _ _ . . - , - _ _ . . _ . .
-
.
. _ . . . - - . . = . . . _ . . _ . _ . . . .._ -. . - _ _ . . . . -
_ - . _ _ _ . . - . _ - ~~. . _ _ . . _ . _ _ __.-___._____._____m.._ _
._ . . _ . . ,
f REACTOR OPERATOR -Page 94 f REFERENCE:
!
EOP-01-LEP-02,." Alternate Control Rod Insertion", Rev. 012, Page 17 OSM 07-2K.07, " Local Emergency Procedures", &6 ,
!
[4.0/4.2) .
.
295015A102 ..(KA's)
'
ANSWER: 085 (1.00) REFERENCE:
.
AOP-15.0, " Alternate Shutdown Cooling Methods", Re , Page 4 NO OSM or Facility Specific Learning Objectives Identified
[3.6/3.6) <
,
295021A201 ..(KA's)
.
ANSWER: 086 (1.00) REFERENCE:
'
EOP-01-UG, " User's Guide"_, Rev. 15, Page 53 OSM 07-2K.02, " Secondary Containment Control", L.O. - (3.8/3.9) 295032K303 ..(KA's) E
-)
i
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,,-,...._-..,,w.,-+,..,-w--.,,- , , - - . , - - - - -
-
-.-
__ REACTOR OPERATOR Page 95 i A!1SWER: 087 (1.00) REFERE!1CC: OSM 07-2K.02, " Secondary Containment Control", .c BTU Exam Bank Question #07-K.02, 532 (Examiner Modified)
[3.9/4.2)
-_
295033K102 ..(KA's) A!1SWER: 088 (1.00) REFERE!1CE: EOP-02, " Primary Containment Control Procedure", Rev. 4C, Flowchart Step PCCP-2 OSM 07-2K.01, " Primary Containmc.nt Control", Re , Page 10, (4.1/4.4) _ 295013G011 ..(KA's) ANSWER: 089 (1.00) . _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ __ _
. . _ -- _ _ . _ . . . _ _ _ - . - _ _ _ _ _ _ _ . . . _ . . _ . . _ . _ - - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . . - ._ _
,
i REACTOR OPERATOR Page 96 --
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-l f
REFERENCE: OP-14, " Reactor Water Cleanup", Rev. 82, Page 20-
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OSM 11A, " Reactor Water Cleanup", Re , Page 16,- & -3 (3.3/3.3) i
.
295008A102 . . (KA's) ANSWER: 090 (1.00) f < REFERENCE: *
.!
OSM 14-G, " Standby-Liquid Control", L.O. - 5 1,13,15,pago 5,12-
[4.5/4.5) -
295037A104 . . (KA's)
.
ANSWER: 091 (1.00) : >
>
i , S I l l
'
l l l P P
-- . . . - - = = - e...-~ .---,..-<w..r...w-3 - +--,------,-.N,-,.,..- ,w,_,wre.,. .,,%.,y,m...e., y-%.a y.,,-,w..e,-+-n, _..-.c.,e,.-y-. REACTOR OPERATOR Page 97
' REFERE!1CE: EOP-01, " Level / Power Control Procedure", Re , Plowchart Step RC/Q-06 OSti 07-2 K.10, "Lovel/ Power Control", Re , Page 86 OSli 16-2A, " Reactor Recirculation System", [4.1/4.2) 295037K301 ..(KA'c)
~
A11SWER: 092 (1.00) REFERE!1CE: a EOP-02, " Primary Containment Control Procedure", T.ev. 4C, Flowchart Step SP/ L-3 6 OSM 07.2K.01, " Primary Containment Control", Re , Page 69, [3.8/3.9) 295030K201 ..(KA's) A!1SWER: 093 (1.00)
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REACTOR OPERATOR Page 98 REFERE!1CE: AOP-05.0, " Radioactive Spills, liigh Radiation and Airborne Activity", Rev. 3, Page 3 LP 07-M, " Abnormal Operating Procedures", [3.9/3.6) 295034G010 ..(KA's) A!!SWER: 094 (1.00) c.
REFERENCE: AOP-36.1, " Loss of Any 4KV Buses", Re , Page 17 OSM 20-2D, " Diesel Generators", L.O. -1 & 2
[4.2/4.3)
295003A102 ..(KA's) ANSWER: 095 (1.00) _ . . . . . .
__ . .
.
REACTOR OPERATOR Page 99 REFERENCE:
;
01-04, "LCO Evaluation and Follow-Up", Rev. 043, Pages 43 & 44 OSM 7D3, " Operations Documents", '
[4.2/4.2)
294001A102 ..(KA's) ANSWER: 096 (1.00) ' REFERENCE: AOP-04.1, " Recirculation Flow Control Failure - Increasing Flow", Rev. 003, Page 3 OSM 07-M, " Abnormal Operating Procedures",llo Facility Learning Objectives Identified
[3.8/3.7)
295001G010 ..(KA's)
l
ANSWER: 097 (1.00) a.
l l REFERENCE:
'
l , SD-18, " Core Spray System", Rev. 013,.Page-8 OSM 14-2E, " Core Spray", L.O. -8
[3.0/3.2]
l 209001K404 ..(KA's).
I y
>
l l l" -
; -. - _ . -_._,.____:___.._. _. .. . . . _ . . _ . - . . .__ _ __ _ _ _ . _ , , _ ___ _ _ _ - . _ . . .
- - - - _ . - . . . - - - - - . _ . . . . - - - - - - - - ~ . . - - - - - - . . - - ~ - . - --
REACTOR OPERATOR Pago100
ANSWER: 098 (1.00) l I i REFERENCE: SD-05, " Standby Liquid Control System", Rev. 008, Page 6 OSM 14G, " Standby Liquid Control System", Re O, Page 6, .b (3.1/3.2] , 211000K202 ..(KA's) ANSWER: 099 (1.00) ;
.
i REFERENCE: l
!
SD-18, " Core Spray System", Rev. 013, Page 5 OSM 14-2E, " Core Spray", [3.8/3.6) t 209001A401 ..(KA's) ANSWER: 100 (1.00) b.
-, ..-..- _-_..-.,_.._-. ~ . . . _ . - . _ ___ _ . . _ _ . . _ _ _ . - . . . . _ _ _ _ . . _ . . _ . _ _ . _ _ _ _ _ _ . _ . _ . _ _ _ - . . _ .
. . . .. .. . ~ . . . ~ - . . .. - . . . _ ~ . . . . .~. . - . - . . . . . ~ . . . .
i REACTOR OPERATOR -Page101"
-. REFERENCE :
EOP-01-LPC, " Level Power Control", Rev.-8, Flowchart Step RC/P-30 - 01-37, " Preparation & Review of PSTG", Rev. 10, Page Y-28 OSM 14-F, " Automatic Depressurization System", .a
[4.5/4.6)
218000K302 . . ( LY s ) i !~
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.
I t
l l
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END OF EXAMINATION
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i DATE / - 5-92 TIME 0500 BRUNSWICK-1 CYC. 8 SEQ. N *** PERIODIC NSS CORE PERFORMANCE LOG *** LOCATION AXIAL REL PWR 1 '2 3 4 5 6 7 8 0.59'1.10 1.20 1.17 1.16 1.12 1.08 1.06 1.02 1.00 0.90 0.60 9 10 11 12 CNT XXX PCT PWR XXXX
* A m ty reuer frose
* test cpJestien #49_
g REGION REL PWR 0.88 1.05 0.88 1.05 1.17 1.05 0.88 1.05 0.88 GMWE XXXXX * selections RING REL PWR 0.79' 1.23 1.22 1.17 1.18 1.09 0.62 CMFCP 0.890 APRM GAF 1.00 0.99 0.99 0.99 1.00 0.98 CMFLP0 0.824
-
\ 3 4
.
5 6 7 8 9 CMAPR 0.820 2.084 REGION 1 2 CMSF
) MFLCPR 0.862 0.881 0.862 0.862 0.890 0.861 0.862 0.881 C.862 CAEQ 0.159 .k LOC - FLOW 17-14 0.0936 21-14 0.0927 35-14 0.0936 13-22 0.0942 25-22 0.0928-39-22 0.0942 17-40 0.C936 31-40 0.0927 35-40 0.0936 CAQA CAVF 0.147 0.432 PK .36 1.40 .1.36 1.37 1.41 1.37 1.36 1.40 1.36 CAPD 45 611 MFLPD 0.824 0.789 0.824 0.786 0,772 0.786 0.824 0.789 0.824 CRD 0.056 k %% LOC PKFL 13-14- 4 31-14- 4 39-14- 4 13-22- 5 29-20- 4 39-22- 5 13-40- 4 31-40- 4 39-40- 4 2.08 2.00 2,08 1.99 1.95 1.99 2.08 2.00 2.08 CRSYM PR .01 PSI A hMAPRAT 0.819 0.800 0.820 0.799 0.783 0.799 0.819 '0.800 0.820 DPC-M 13.27 17-14- 4 31-14- 4 35-14- 4 13-22- 5 29-20- 4 39-22 5 17-40- 4 31-40- 4 35-40- 4 DPC-C 18.11 NLoc PKFS 1.88 1.84 1.38 1.84 1.79 1.84 1.88 1.84 * . 88 RWL 187.19 DHS 23.47 WFW 10.19 FAILED SENSORS 1 3 4 13 WD 29.09 WTSUB 64.37 FAILED LPRM LIST BASE CRIT CODE WTHB -1.00 UT 65.03 1213,C,4 2013,B,2 4413,A,2 4413,D,4 PCTUTR 84.456 0421,C,4 1221,B,2 2821,0,4 4421,D,4 WTFLAG 2.000 1229,B,4 1237,B,2 1237,D,4 4437,D,4 ITER 1.000 '
IREC 0.000
/' '
RCAF IXYFLG 1.000 0.000 CMFLEX 0.918
. CAVEX 19428.500 CYEXP 7091.500 THE 12 MOST LIMITING BUNDLES LPRM % POWER 99.645 FOR MFLCPR FOR MFLPD FOR MAPRAT MFLCPR LOC MCPR CPRL!M MFLPD LOC MRPD'RPDLIM MAPRAT' LOC MAPLMGR LIMLHGR 0.890 25-22 1.434 1.275 0.824' 13-14- 4 11.87 14.40. 0.820 35-14- 4 10.69 13.05 0.889 27-32 1.434 1.275 0.824 39-40- 4 11.87 14 40 0.820 35-40- 4 10.69 13.05 0.889 27-22 1.434 1.275 0.824 39-14- 4 11.87 14.-J 0.819 17-14- 4 10.69 13.04 0.889 25-32 .1.434 1.275 0.824 13-40- 4 11.87 14.40 0.819 17-40- 4 10.69 13.04
+ 0.881 31-40 1.447 1.275 0.810 35-14- 4 11.66 14.40 0.818 39-40- 4 .10.70 13.08 0.881 .31-14 1.447 1.275 0.810 35-40- 4 11.66 14.40 0.818 39-14- 4 10.70 13.08 0.881 21-14 1.447 1.275 0.810 17-14- 4 11.66 14.40 0.818 13-14- 4 10.70 13.08 0.881 21-40 1.447 1.275 0.810 17-40- 4 11.66 14.40 0.818'13-40- 4 .10.70 13.08 0.871 23-20 1.464 1.275 0.802 13-36- 5 11.54 14.40 0.813'13-36- 5 10.64 13.08 0.871 29-34 1.464 '1.275 0.801 13-18- 5 11.54 14.40 0.813 13-18- 5 10.64 13.08
'O.871 23-34 1.464 1.275 0.801 39-36- 5 11.54 14.40 0.813 39-36- 5 10.64 13.08 0.871 -29-20.'1.464 1.275 0.801 39-18- 5 11.54 14.40 0.8'3 39-18 5 10.64 13.08 THE NUMBER OF BUNDLES WITH MFLCPR GREATER THAM 1.0 = 'O THE NUMBER OF BUNDLES WITH MFLPD, GREATER THAN 1.0 = ' O THE NUMBER OF BUNDLES WITH MAPRAT GREATER THAN 1.0 = 0
f W - < s +
. .
CONDENSATE TRAIN FOR BSEP UNIT # 2 DATE 05-dec-1992 POWER 85 TIME 07:00
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REACTOR WATER Conductivity (umho/cm) [<=2.0) CL (ppb) < 2.0 NO3 (ppb) < 2.0 I Actual 1.600 Indicated 1.700 SS (ppb) NA SO4 (ppb) < SIO2 (ppb) < 5. DO2 (ppb) 5 Sample Point: 2-RXS-17 Ph [5.6-8.6) NA TOC (ppm) NA
----------------------------------------------------------------------------- "A" RWCU F/D EFFLUENT l "B" RWCU F/D EFFLUENT Conductivity (umho/cm) [<=0.1) l Conductivity (umho/cm) [<=0.1)
Actual 0.056 Indicated 0.055 l Actual 0.058 Indicated 0.058 SiO2 (ppb) < 5. Flow (gpm) 10 SiO2 (ppb) < 5. Flow (gpm) 10 Hotwell Reject (gpm) FEEDWATER COND. PUMP DISCHARGE HTR. DRAIN HEAD Cond.umho/cm [<0.1) COND.umho/cm [<0.1) COND.umho/cm [<0.1) Ac .055 In .055 Ac .071 In .071 Ac .059 002 ppb [20 - 100) -DO2 ppb [20 - 100) In .057 Ac . In . Ac . In . SS ppb [<10) < 1 o Cl ppb NA * Cl ppb NA MAIN STEAM , SIO2 ppb NA ---------- COND.umho/cm [<1.0) o If cond.<= .25 C1 analysis required once/w Ac .062 Foodwater flow #/hr 8.50E+06 In .062
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COND. F/D INFLUENT DEEP BED INFLUENT DEEP BED. EFFLUENT
------------------ -----------------
.
#CL capacity COND.umho/cm [<1.0) COND.umho/cm [<1.0] COND.umho/cm [<1.0) remaining.**
Ac .072 Ac .062 A BED NA 274.90 In .083 In .072 B BED NA 268.66 C BED NA 268.79 COND. F/D EFFLUENT COND. POLISHING D BED NA 275.69
------------------ EFFLUENT E BED -NA 263.63 COND.umho/cm [<1.0) --------------- F BED N .64 A F/D NA COND.umho/cm [<0.1]
B F/D NA Ac .053 C F/D NA In NA D F/D NA ** should be changed at approx. 118# :CL capacit CST CHEMISTRY L Cond. (umhc/cr) [<=1.0 ] 0.479 CL (ppb) [ <100 ] < SiO2 (ppb) < Ph [5.8-8.5] 5.90 SS (ppb) [ <100 ] < - 10 . NO3 (ppb) < SO4 (ppb) < Level (Feet) 28. _(Inches) Sample Point: 2-CO-V137
-----------------------------------------------------------------------------
REMARKS: All recommendations are based on analysis trends, E&RC 1000, Tec Spec., and GE Fuel Warranty, NA: Not Applicable TECHNICIAN: REVIEWED BY :
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CAROLINA POWER 6 LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PIANT -i PLANT OPERATING MANUAL VOLUME I
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UNIT 0
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I PROCEDURE TYPE: ADMINISTRATIVE INSTRUCTION NUMBER: -0 AI-81 PROCEDURE TITLE: WATER CHEMISTRY GUIDELINES !
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REVISION 11
< < , ,
k I e L Approved By: P %-69, AA ~ Date:
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M . 7 L-- 1-g p al-Manager s i
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.Page 1 of 29 ]_
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LIST OF EFFECTIVE PAGES 0 Al B1 page(s) Revision 1-29 11 _
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W G 0-AI-81 Rev. 11 Page 2 of 29 {'
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l PURPOSE The purpose of this document is to provide Water Chemistry action levels at different modes of operatio These action levels support the Brunswick Chemistry department mission to: 1.1 Minimize the harmful effects of chemical impurity ingres .2 Support long term reliability of plant equipment and components - through reduced corrosio .3 Minimize in-plant radiation field buildup and releases of radioactivity to the environment.
- To be effective, this document should have the full support of all levels of plant and corporate management. This includes the provision of adequate resources in terms of staff, equipment, maintenance, and funds in a timely manne Without the support of management at all levels, these guidelines will be ineffectiv The plant Water Chemistry Control Program should be reviewed and updated on a continuous basis and changes implemented as needed on a timely basi The importance of establishing and maintaining good water chemistry in the BWR primary system has been well documente This water environment is aggressive and has led to stress-corrosion cracking and corrosion
,(.) problems which in turn have contributed to deposition on fuel cladding and plant radiation field buildu Two of the most significant effects of this aggressive environment have been the intergranular stress corrosion cracking of sensitized stainless steels (ICSCC) and the erosion corosion of carbon steel components. Reducing and maintaining low level ingress of impurities into the plant systems will minimize corrosion and -
cracking problems, benefit fuel performance, and reduce plant radiation - field buildu It has been determined that ICSCC can be mitigated by Hydrogen Water Chemistry (HWC). Electro Chemical Potential (ECP) measurement action levels have been included in this documen ,
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2.0 REFERENCES I BWR Hydrogen Water Chemistry Guidelines, 1986 Revision, EPRI ; NP-4946-SR, October 1987 i 2.2 BWR Hydrogen Water Chemistry Guidelines, 1987 Revision, EPRI l NP-4947-SR-LD, January 1988 !
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0-AI-81 Rev. 11 Page 3 of 29
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_ 2.0 REFERENCES 2.3 PLP-04, Corrective Action Program CP&L Memorandum to C. E. Robertson, File: 10510B9, Serial: 89NSS1, dated September 15, 1989, Subject: Chloride Action Levels for RBCCW and TBCCW Systems 3.0 RESPONSIBILITIES
. , Plant General Manager The plant General Manager is responsible for establishing and maintaining the chemistry program and shall assign specific responsibilities for the implementation of the progra .2 Manager - Operations The Manager Operations is responsible for ensuring that appropriate corrective actions based on this manual, the associated guidelines, specificacions, procedures, and instructions are followe The following should be considered:
3. An understanding of the impact of the water chemistry- , progra % 3. Establishment of corrective action prioritie ', ) 3. Implemantation of corrective action .3 Manager - E&RC The Manager E&RC provides overall direction for.the ch'omistry-program. He is the interface with regulatory agencies and other-audit organizations and has the responsibility to respond to their finding .4 Manager - Maintenance The Manager - Maintenance is responsible for assuring prompt and effective repair.of plant equipment used for controlling or monitoring chemistry parameter .5 Manager - Technical Support The Manager - Technical Support is responsible for ensuring that all plant modifications are reviewed for chemistry control considerations. Plant design and modifications, if not well , -
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considered, may have adverse effects on the ability to control' water
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chemistr . t
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0-AI-81 Rev. 11 - Page 4 off29
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3.0 RESPONSIBILITIES 3.6 Managers - Supervisors i It is the responsibility of managers and supervisors whose work may f impact power plant chemistry to ensure that appropriate personnel l under their supervision are cognizant of the Power Plant Chemistry- { Program and the associated guidelines, specifications, procedures, _ l' ' and instructions and that their personnel make every effort to perform their assigned duties in accordance with these guideline i Employee Responsibility l Each company employee and contractor employee working in a facility where there exists the potential for affecting chemistry control requirements and performance objectives shall make every reasonable ; effort to perform their duties in accordance with chemistry l guidelines, specifications, procedures, and instructions. If an abnormal trend is noticed or a result is not within established '
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limits, refer to PLP 04 to determine the severity of the event and .
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if the event is reportabl ; DEFINITIONS / ABBREVIATIONS Definitions 4. Control Parameters: Those parameters that are known to 3 affect the corrosion of plant materials, fuel performance,
, ' !
and radiation field buildup. These parameters include, ; but are not limited to, chloride, specific conductivity, l dissolved oxygen, and silic Technical,Sp,ecification chemistry limits which . result in a limiting condition of operation shall in every case supercede this manua '- i
.
4, Diarnostic Parameters: Either those parameters which-can affect corrosion of plant materials, fuel performance,'and radiation field buildup, but for which insufficient data l . is as yet available, or those parameters which may be.used i in interpreting control-parameter deviations. lThese l parameters may include cation conductivity, pH, fluoride, l sulfate, nitrate, carbonate, total organic carbon, and ' sodiu ' l E0IE: If an Action Level 2 or 3 value is exceeded such that a uni ;. shutdown is required but.not performed'(as-directed by the Plant i General Manager), a review of the incident shall be performed and if ,
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l ' !. no other report is required, then an E&RC. Experience Report should l' be written in accordance with E&RC Procedure 002 i i t
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j 0-AI-81 Rev. 11 - Page 5 of 29
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14.0 DEFINITIONS / ABBREVIATIONS 4. Action Level 0: This is the achievable value which-can be-maintained for each parameter by the_ application of good operating practices. Every effort should be made.to operate at or below these values. - In some cases, the
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parameter may be outside the Action Level 0 value and no l in an Action Level 1, 2, or In this situation, efforts [ should be made to return the parameter to the optimum Action L(vel 0 valu . Action Level 1: - Available data indicates that long-term system reliability is likely to be diminished if the plant is operated for long periods of time above this valu If a parameter exceeds this value, the following actions shall be initiated: 4.1. Reduce the parameter below the Action Level 1 value within 96 operating hours or follow the applicable footnotes in the table .1. If the parameter cannot be reduced to below this value within 96 hours,_the system engineer in conjunction with the plant Chemistry group shall
. evaluate the consequences of contihued operation while exceeding this paramete A formal-report is ( '; not required for exceedin6 an Action level 1 x- ! parame te _ .
4. Action Level 2: Available data indicates that significant damage could be done to the system within~a short-time frame. If a parameter exceeds this level, prompt correction of the condition is. required. The.following action shall be initiated if a parameter' exceeds the: Action Level 2 valu .1. Reduce the parameter below the Action Level l2 value within 24 hours or_ follow the applicable footnotes in the table . . HQIE: Directions provided by-the plant General: Manager shall be'documentedL in the appropriate Nuclear Shift _ Supervisor's Logbook,
~4.1. If the parameter has.not been_ reduced below the 'I Action Level 2 value within 24' hours,'an orderly unit- J!
shutdown shall be-initiated and the plant shall be brought to a cold shutdown condition within 16 hours unless directed otherwise by_ the Plant General Manager or his designee.
^ 4.1 Following a unit shutdown caused by-exceeding an L s Action Level 2 value, a review of the. incident shall' _j be: performed and appropriate' corrective measures- I taken before the unit.is restarted.
- 0-AI-81 Rev. 11 - Page 6 of 29'
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. . _ DEFINITIONS / ABBREVIATIONS 4. Action Level 3: Available data indicates that it is inadvisable to operate the unit when a parameter exceeds this value. If a parameter exceeds the Action Level 3 value, the following actions shall be initiated:
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N01E: Directions provided by the Plant General Manager shall be documented in the appropriate Nuclear Shif t Supervisor's Logboo .1. Follow t'.ie applicable footnotes in the tables or begin an orderly shutdown of the unit immediately with reduction of the coolant temperature to < 212*F (100'C) as rapidly as plant constraints permit unless directed otherwise by the plant General Manager or -- his desi ne .1. Following a unit shutdown caused by exceeding an Action Level 3 value, a review of the incident shall be performed and appropriate corrective measures taken before the unit is restarte . Operational Mode: Operational mode, as referred to in this document, is the same as defined in the technical specification . Chemistrv Control Tables: Chemistry control tables are
' presented in Attachment 1 to this procedure. Operation at or below the action-level values presented in these tables will not prevent corrosion of the components in the plant, but if followed, will help minimize corrosion, radiation field buildup, and crud depositio .2 Abbreviations _
N/A 5.0 GENERAL It is the policy of Carolina Power & Light Company to engineer, construct, and operate nuclear power plants without jeopardy to the , health and safety of the public and of its employee { The function of the Corporate Nuclear Chemistry Policy is to support the overall CP&L policy, to maximize the availability and operating lifetime of a plant by maintaining the integrity of the major components, and to minimize corrosion damage, fuel failure, and radiation buildup through an effective water chemistry progra It shall be the responsibility of all organizations at the plant and of all corporate support groups to provide the support needed to ensure that
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the Corporate Nuclear Chemistry Program is suecassfu ; i 0-AI-81 Rev. 11 Page 7 of 29 _ _ _ _ _ _ _ _ _ - _ _ _ - ______
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_ 5.0.. GENERAL
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The frequency of analysis is governed by Technical-Specification, vendor-recommendations, and good operating practices.- Frequencies are given in , the' applicable E&RC_ procedures, operating procedures, or Technical- [ Specification .0 PREREQUISITES-I Reliable in-line instrumentation and recorders are some'of the most important tools available_for the successful operation of the plan ; Therefore, maintenance of sample stations, sample station chillers, and in line instrumentation, shall be classified as a Priority 2 item, _ according to the Maintenance Management System, for_those instruments required for the monitoring of Technical Specifications parameters, action level parameters, and diagnostic parameter , t PRECAUTIONS AND LIMITATIONS 7.1 Precautions Operation at or below the action level values presented in this document will not prevent corrosion of the components of the plant.
' but if followed, will help minimice. corrosion, radiation field buildup, and crud deposition *.
- 7,2 Limitations ;
l I a N/A , 8.0 REAGENTS AND APPARATUS N/A
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= 9.0 - ACCEPTANCE CRITERIA -
N/A ' ,
'10.0 PROCEDURE STEPS =
The method of sampling will be controlled by the appropriate E&RC procedure-or operating procedur .0 DIAGRAMS / ATTACHMENTS / CALCULATIONS , . 11'.1 Diagrams-N/A- .
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11.2 Attachments =
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11. Attachment A, Reactor Water--Cold Shutdown / Refueling
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' 11.2.2- Attachment B~, Reactor Water--Startup/ Hot Shutdown -l-
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.0-Al-81 Rev.' 11 - Page 8 of 29
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- 11.0 DIACRAMS/ ATTACHMENTS /CALCU!ATIONS 11. Attachment C, Reactor Water--Power. Operation .
11. Attachment D Reactor Water Cleanup System Startup/ Hot Standby and Power Operation 11. Attachment E, Heater Drain Header--Prior to' Pumping-Forward 11. Attachment F, Heater Drain Header -During Steady State Operation (Forward Pumped) , 11. Attachment G Condensate Pump Discharge (CPD) 11. Attachment H. Final Feedwater 11. Attachment I, Condensate. Polisher Effluent (CPE) 11.2.10 Attachment J, Vaste Sample Tank Transfer to CST / CST Water Quality 11.2.11 Attachment K, Control Rod Drive Water 11.2.12 Attachment L, MUD Tank and Makeup Water Demineralicer System Effluent -All Modes
) 11.2.13 Attachment M, Spent Fuel Pool--All Modes-11.2.14 Attachment N, Turbine Building and Reactor Building Closed Cooling Water System--All Modes 11,2.15 Attachment 0, Diesel Generator Cooling Water for Water- . '
Jacket-- All Modes 11.2.16 Attachment P, Auxiliary Boiler .2.17 Attachment Q, Torus 11.2.18 Attachment R,- Cooling Towers 11.2.19 Attachment S, Service Water Chlorination 11.2.2 Attachment T, Condenser Air Inleakage
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11.3 Calculations N/A __
-0-AI-81 Rev. 11 'Page 9 of 29
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l Attachment A Reactor Water--Cold Shutdown / Refueling ACTION LEVELS Parameter 0 1 2 3 Conductivity 5 > > 10(*) pS/cm at 25'C Chloride ppb s 20 > 100 > 200(*)
(*) Restore the conductivity to < 10 pS/cm and the chloride to < 200 ppb within 48 hours (Technical Specification 3/4.4.4).
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_ Attachment B
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i Reactor Water Startup(*W liot Shutdown ACTION LEVELS- i Parameter 0 1 2 3 -
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Conductivity s 0,500 > . 0,500 >1 >2 pS/cm at 25'c Chloride ppb- s 20 > 20 > 100 > 200 Dissolved oxygenN Sulfate ppb s 20 > 20 > 100- > 200 f Boron ppb < 10 > 40 > 8000(*) _ l (a) Prior to going to power operation, these parameters should be below Action 2 level .
' . (b) Dissolved oxygen-should be < 200 ppb when reactor water temperatur has stabilized at 212* (c) Inform Shift Supervisor /E&C Supervisor, j ' ,
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0-AI-81-Rev; 11 Page 11'of 2 J n- a.wes.,.
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Attachment C Reactor Water--Power Operation ACTION LEVELS
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Parameter 0 1(*) 2 3 Conductivity (*) s 0.200 > 0.300 >-1 > 50) pS/cm at 25'C
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Chloride ppb s 15 > 20 > 10 > 200 Silica ppb s 100 > 200(*) Sulfate ppb s 15 > 20 >100 > 200 ECP of Type'304 s -0. 25(*)(d) > 0,23(*)(d) _ _ stainless steel in recirculation water (vSHE) Boron ppb < 10 -> 40 > 8000(*) _
.- (a)See Table ' (b)If the decision is made by the Plant Manager not to shut the unit down immediately, refer to Technical Specification 3/4.4.4 for additional - ^ 4 requirement >
s m (c)Uhen operating under hydrogen water chemistry.
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(d)If excessive radiation is experienced in the' plant, it may be necessary to operate at ECPs:somewhat greater than is desire (e) Inform Shif t Supervisor /E&C Superviso {
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Attachment D l~ ,-
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Reactor Water Cleanup System -Startup/ Hot Standby and Power Operation-ACTION LEVELS
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Parameter 0 19) 2 3 Effluent Conductivity s 0.080 > 0.08 or_10% pS/cm at 25'C of influent conductivity
+ 0.055 Effluent Silica s 100 > 100 or 30%
cf influent if greater than 100 (*)ln all cases , filter effluent water quality will be dependent on reactor power, reactor conductivity, filter integrity, and other factor For these reasons, it is not always possible to operate RUCU ' filters below the action level; therefore, E&RC chemistry shall continually track filter performance against reactor water quality and-identify the optimum time for filter changeou Chemistry will then relay that information to Operations for scheduling of filter backwash-and precoat.
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Attachment E Heater Drain Header--Prior to Pumping Forward ACTION l.EVELS Parameter 0 IN 2 3 Conductivity s 0,1 > pS/cm at 25'C Suspended Solids s 50 > 50 ppb (a) Prior to pumping forward, conductivity and suspended solids should be less than Action Level 1. If these cannot be met, then the effects - of pumping forward should be evaluate .- .
0-Al-81 Rev. 11 Page 14 of 29 __ _ _ _ _ _ _ - _ - _ - .
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Heater-Drain Header .During Steady State Operatiori (Forward Pumped) 6.C.JION Q LEVELS Parameter 0 1 2 3 Conductivity s > 0.1(*) pS/cm at-25'C Suspended Solids s 10 > 10("F ppb (a)If one of these action levels is exceeded for_ greater than~96 hours-
-during steady state operation, an engineering review should be performed to evaluate the effects of pumping forwar L P
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Attachment G I Condensate Pump Discharge (CPD) ACTION LEVELS Parameter 0 1 2 3 Conductivity s 0.080(*) > 0.100(*) > 10 pS/cm at 25*C (a)Not applicable when unit is shutdow _ t _
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l Attachment H Final Feedwater ACTION LEVELS Parameter 0 1 2 3 Conduc tivity(*) < 0.06 > 0.070 > 0.1(d) pS/cm at 25'C Dissolved oxygen (*) 20-50 < 10 or > 200 ppb Feedwater Total (*) s > Copper (ppb) Feedwater Total (*) s > 5. 0(') Iron (ppb) Feedwater Suspended (b) < 15 > 50 Solids (ppb)
(a)Do not open the feedwater inlet valves to the reactor until conductivity is below 0.1 pS/cm. This limit may be exceeded with permission from the Shift Operating Supervisor and the E&RC Chemistry Supervisio ; (b) Applicable during unit startup onl (c)Not applicable during unit startu (d)1f conductivity exceeds 0.1 pS/cm during steady state power operation, investigate cause of high conductivity and eliminate the source of impuritie _ (e)No special action is required if this parameter exceeds the Action Level 1 valu .
0-AI-81 Rev. 11 Page 17 of 29 I l
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Attachment I f condensa'te Polisher System Eff'luent (CPE)-
' ACTION LEVELS Parameter 0 1 2 3 Conductivity (*) s 0.060 > 0.070 > 1,0 pS/cm at 25'c (a)A deep bed demineralizer should be changed whenever the effluent conductivity exceeds 0,070 pS/cm or whenever bleed from the demineraliser is detecte . ,, , i
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I Attachment J l
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Waste Sample Tank Transfer to CST and CST Water Quality ACTION LEVELS (*) Parameter 0 1 2 3
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Conductivity s 0.80 > S/cm at 25'c Chloride ppb s 10 > 10 Silica ppb s 10 > 20 TOC ppb s 200 > 200 Sulfato ppb s 10 > 10 Total y activity s 10 > 10 (C1) (b)
(a)All water processed to'the condensate storage tanks should meet these limits. The approval of the-Radwaste Supervisor and Environmental and Chemistry Supervisor is required to transfer out of limit sample tanks to the condensate storage tan g (b) Applies to CST onl ( . .) %
0-AI-81 Rev. 11 Page 19 of 29
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Attachment K Control Rod Drive Vater N ACTION LEVELS Parameter 0 1 2 3 Conductivity s 0.10 > 0.10 S/cm at 25'c Dissolved oxygen N s 50 _
> 200 (a)These limits are applicable in the normal CRD configuration in which drive water is from the effluent of the condensate demineralizer (b) Dissolved oxygen values are from final feedwate ,
0-AI-81 Rev. 11 Page 20 of 29
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Attachment L
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MUD Tank ; All Modes ,
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ACTION LEVELS (*) -f
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Parameter 0 1 2 3
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Conductivity s 0.800 > pS/cm at 25'c TOC ppb s 200 > 200 Chloride ppb s 10 > 10 > Silica ppb s 10 > 20 Sulfate ppb s 10 > 10 s 10 -> 10 ' Total y activity (C1) .
, (a)All water processed to the CST should meet these limits. The approva of the Radwaste Supervisor and E&C Supervisor is required to transfer out of limit tanks to the CS .
Make Up Unter Demineralizer System Effluent *
' , ACTION LEVELS ( /
Parameter 0 1 2 ~3 Conductivity s 0.100 > 0.100
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-pS/cm at 25'C
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TOC ppb s 200 > 400(*F l
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(a) Rinse-in characteristic for TOC Action Level l1 differ due to initial .?- ,
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demineralizer TOC level being higher than the MUD Tank. This is an { economic balance to maintain water within the Action-Level for the mud
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tank.
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.0-AI-81 Rev. l'1 Page 21 of 29 .. - - . . .- - ,. , _...,m, y,_,- , . . .,, , . , , , , , v
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i Attachment M Spent-Fuel Pool All Modes
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i ACTION LEVELS Parameter 0 1(*) 2 3 l Conductivity s > pS/cm at 25'C Chlorido ppb s 100 a 200 Fuel poo s 1.0 E-03 > 1.0 E 03 N-- 'l activity yCi/cc-(*) Maximize fuel pool cleanup, s-M Applicable only for utilizing divers for mainto. nance in. fuel poo t
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, l Attachment N I Turbine Building and Reactor Building Closed Cooling Water Systems All Modes
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ACTION LEVELS Parameter 0 1 2 3 Nitrite ppm 500-1500 < 500 .. -- pH at 25'C 9.0 1 < 9.0 or > 1 Gross activity s 5.0 E-07 2 5.0 E 04 -- -- pCi/ml _ Cl ppm <5 t 10 t 20(*) --
(*)Take appropriate action to reduce the chloride concentration to below 20 pp Unit shutdown is not .equire _
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0-AI-81 Rev. 11 Page 23 of 29
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Attachment O Diesel Generator Cooling Water for Vater Jackets All Modes ACTION LEVELS Parameter 0 1 2 3 i Nitrite ppm 500-1500 < 500 -- -- pH 8.5- < 8.5 or > Cl* ppm <5 a 10 m 20(*) -
(*)Take appropriate action to reduce the chloride concentration to below 20 ppm. Unit shutdown is not required.
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.0-Al-81 Rev. ll Page 24 of 29
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Attachment P
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Auxiliary Boilers
- ACTION LEVELS Parameter 0 1 2 3 Conductivity < 4000 > 4000 ps/cm at 25'C pH at 25'c 9.8 1 < 9.8 or > 11.0 N P < 50 S0 3 -200-400 < 200 SiO 2 s 100 > 100N Cross activity 5 LLD > 1,0 E-07 pCi/ml (a) Appropriate action will be taken as designated by the Chemistry group
to bring parameters within the suggested ranges. Due to intermittent operation of the boilers, these ranges are for operation, standby, and , lay-u 's x -1
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l Attachment Q ; Torus i i ACTION LEVELS I
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Paratnoter 0
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1 2 3 conductivity 5 a 1 pS/cm at 25'c
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Chloride ppb s 200 a 500
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Suspended solids 65 PP'n a TOC ppb s 1000 Julfate ppb s 200 t 500
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. - 0+Al d1 Rev. 11 . Page 26 of 29:
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Attachment R Co711n5 Towers , ,
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ACTION LfSIjd ; Parameter , 0 14) 2 3 l pit at 25'C 6.0 < 6.0 or > ' Langlier Index 05 - + < .0.5 or > 2.0 '
(*)Adj us t the chemical feeds as necessary to maintain these parameter ,
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i j , Attachment 5 i Service k'ater Chlorination l l ACTION LEVELS 0 1 2 3 Paratteter Free Available 2 < . 2 5(*) None chlocine (ppm) detected W i E011: Only applicabic when chlor! nation is is servic (a) Restore the free available chlorine to a .25 ppm within 96 hours or perforts an engineering evaluation to justify continued operatio Chlorination may not be required depending on season of the year and water temperatur (b)tiotify Radwaste Operations personnel, and restore chlorine level within 24 hour . _
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0-Al-81 Rev. 11 Page 28 of 29
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l l /sttactutent T l Condenser Air Inicahage ,
(Power Operation) ;
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ACTION LEVELS } I 0 1 2 3 Parameter Condenser Air < 30 > 60 ia 150N Inleakage (sefm)
(a)AOC Bypasses, continuous operation will impact Release Limit from sit Inleakage Testing Action Plan should be implemente :
0 AI-81 Rev. 11 Page 29 of 29
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SECTIONS 3.0 AND LIMITING CONDITIONS FOR OPERATION
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AND SURVEILLANCE REQUIREMENTS
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RETTPED TECH. SPEC Updated Thru. Amend. 78 -
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s 1_/ 4 LIMITING CONDITIONS FOR OPERATION AND SURVEILI.ANCT. RIOUIREKENTS T' AP Pt.IC ABILITY LIMITtNC CONDITION FOR OPERATION 1. t,tmittne Conditions for Operation and ACTION requirements shall be applicable d u r in,! the OPERATIONAL CONDITIONS or other states specified for each specificatio T. Adherence to the requirements of the limiting Condi t ion for Operation md issnciated ACTION within the specified time interval shall constitute mmp1 t ance with the specificatio In the event the Limiting Condition for nr eration is restored prior to expiration of the specified time interval, completion of the ACTION statement is not require . In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot he satisfied because of circumstances in excess of thn=e addressed in the specification, the unit shall be placed in at least HOT WTn0VN wit bin 6 hours and in COLD SHUTDOWN vi t b in t he following 30 hours unless correctiv- measures are completed that permit operation under the norminnible ACT10N statements for the specified time interval as measured, f rom initial discovery or until the reactor is placed in an OPERATIONAL CONDITION In sbich the specification is not applicabl r.x c e p t i on s to these enotrements shall he stated in the individual specification . n . '. rn try into an OPERATIONAL CONDITION or other specified applicability state shall not he made unlees the conditions of the Limiting Condition for neerition are me t without reliance on provisions contained in the ACTION statements unless otherwise excepted. This provi< ion shall not prevent passare t hrouch GPERATIONAL CONDITIONS requi red to compl e with ACTION requirement . When a system, subsystem, train, component, or device is determined to ne innnerable solely because its emergency power source is inoperable, or
<aleiv hecause its normal poser source is inoperable, it may be considered WERARLF for the purpose of satisfying the requirements of its applicable '.i m i t i ng Condition for Operation, provided: (1) its corresponding normal or amergenev power source is OPERABLE; and (2) all of its redundant system (s),
subsv<tems(s), train (s), component (s), and device (s) are OPERABLE, or likewise 4 it t if v the requirements of this specificatio Mnless both conditions (1) and ( 2) tre satisfied, the unit shall be placed in at least HOT SHUTDOWN within 6 hours, and in at least COLD SHUTDOWN vithin the following 30 hou r s . This s pe c i f i c a t ion is not applicable in Conditions 4 or 5.
. UeNSWICK - UNIT 2 3/4 0-1
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RETYPED TECH. SPEC Updated Thru. Amend. 78
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{ APPLICABILITY SURVEILLANCE REQUIREMENTS
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4. Surve t t lance Requirement's shall be a;>plicable during the OPERATIONAL CONDITIONS or other states specified f or individual Limiting Conditions f or Operation unless otherwise stated in an individual Surveillance Requiremen . Each surveillance Requirement snall be performed within the specified time int erva! wi t h a maximum all owable e xt ension not to exceed 25% of the surveillance interva . Performance of a Surveillance Requirement within the specified time -- interval shall constitute compliance with OPERABILITY requirements for a Limiting Ccndition f or Operatinn and associated ACTION statements unless otherwise required by the specificatio Surveillance requirements do not have to be performed on inoperable equipmen . Entry int o an OPERATIONAL COND1110N or other specif ied applicable state snall not be made unless the Surveillance Requirement (s) associated with the Limiting Condi tion f or Operation have been perf ormed within the applicable surveillance interval or as otherwise specifie . Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, i, and 3 components shall be applicable as follows: Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pres sure Vessel Code and applicable Addenda as required by 10 CFR 50, Sectton 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50 Section 50.55a(g) (6) _
(i).
BRUNSWICK - UNIT 2 3/4 0-2 Ame nd me n t No. 180
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([ 3/4.0 A PPLI CABI LI TY ah,
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SURVE!LLM/CE REQUIREMENTS (Continued ) I
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- Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice j inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as 1 f ollows in these Technical Specifications
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- ASME Boiler and Pressure Vessel Required frequencies Code and applicable Addenda for performing inservice terminology for inservice inspection and testing inspection and testing activities activities (
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Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once'per 92 days Semiannually or every 6 months At least once per 184 cays Yearly or annuall y At least once per 366 days The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activitie Perf ormance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
- Nothing in the ASME Boiler and Pressure Vessel Code shall be const rued
to supersede the requirements of any Technical Specification.
4 The Inservice Inspection Program f or piping identified in NRC Ceneric
Letter 88-01 shall be performed in accordance with the staff positions on schedule. methods & personnel & sample expansion inicuded in this letter.
l-BRUNSWICK - UNIT 2 3/4 0-3 Amendment No. 180
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3/ REACTIVITY CotRROL SYSTEMS
.o 3/4 SHUTDOWN MARCIN LEHITING CONDITION FOR OPERATION 3. The SHUTt0WN MARGIN shall be equal to or greater than 0.38% a k/ APPLICABILITY: CONDITIONS 1, 2, 3, 4, and ACT ION:
With the SHUTDOWN MARGIN less than 0.3B: a k/kg In CONDITION 1 or 2. reestablish the required SHUTDOWN MARCIN within 12 hours or be in at least HO" SHUTDOWN within the next.12 hour . In 00NDITION 3 or 4, immediately verif y all control rods to be f ully inserted, suspend all activities that could reduce the SHUTDOWN MARGIN, and demonstrate SELDNDARY COlHAINKE!R llRECRITY within 1 hour; reestablish the required SHUTDOWN MARGI . hs 00NDrIION 5, suspend 00RE ACIERATIONS and other activities that could reduce the SHUTDOWN MARGIN, f ully insert all insertable control rods, and demonstrate SECDNDARY 0)tRAINMENI INEEGRlTY within 1 hour;
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) reestablish the required SHUTDOWN MARGIN. The provisions of "* . Specification 3.0.3 are not applicabl SURVEILLANCE REQUIREMENIS l THE SHUTLOWN MARGIN shall be determined to be equal to or greater than-4.1.1 l
j 0.38% a k/k: l By measurement within 24 hours prior to or during the first start-up af ter completing CORE ALTERATIONS, and l By analytical determination within 12 hours af ter detection of a withdrawn control rod that is immovable or untrippable, except that the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable ro e i hn/ l ' BRUNSWICK - UNIT 2 3/4'1-1 RETYPED TECH. SPEC Updated Thru. Amend.-7[
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l REACTIVITY 00NTROL SYSTEMS s 3 / 4.1. 2 REACTIVITY ANOMALIES 4
LIMITING CONDITION POR OPERATION 3.1.2 The reactivity dif ference between the actual ROD DENSITY and the predicted ROD DENSITY shall not exceed 1% Ak/ APPLICARILITY: CONDITIONS 1 and ACTION: With the reactivity dif ferent hy more than 1% ak/k Perform an analysis to determine and explain the cause of the reactivity dif ference; operation may continue if the dif ference is explained and corrected, or Be in HOT SMUTDOWN vithin 12 hour Submit a Special Test Prcuram to the Commission describing the methods to be used to determine the cause and magnitude of the reactivity dif ferenc , ,. [,) .90RVEILLANCE REQUIREMENT 9 4. The ROD DENSITY shall be credicted and compared to the actual ROD DENSITY for selected operating conditions: During the first start-up following CORE ALTERATIONS, and At least once per ef fective full power month during POWER OPERATIO . O: BRUNSWICK - UNIT 2 3/41-2 RETYPED TECH. SPEC Uodated Thru. Amend.;78L
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REACf1VITY CONTR0f. SYSTEMS
- 3f,4.1,. 3 CONTROL RODS CONTROL ROD OPERAR1,tg L1HITING CONDITION FOR OPERATION ~.- ..._____ ,
3.1. All control rods shall be OPERABL APPLICABILITY: OPERATIONAL CONDITIONS 1 and ACTION: Uith one conerni rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to he untrippable: Within one hourt a) Verify that the inoperable control rod. if withdrawn, is separated from all other inoperable control rods by at least two control cells in all direction . b) Oisarm the associated directional control valves hydraulically by closing the insert and withdraw isolation valve . Otherwise, be in at least HOT SHUTDOWN within the next 12 hour I, ,} Restore the inoperable control rod to OPERABLE status within 48 1,s ' hours or be in at least HOT SHUTDOWN within the next 12 hour With one or more control rods inoperable for causes other than addressed in ACTION a, above ! If the inoperable control rod (s) is withdrawn, within one hour: a) Verify that the inoperable withdrawn control rod (s) is separated from all other inoperable control rods by at least two control cells in all directions, and b) Demonstrate the insertion capability of the inoperable withdrawn control rod (s) by inserting the control rod (s) a least one notch by drive water pressure within the normal operating range *. or c) Fully insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves either: 1) Electrically, or 2) Ilydraulically by closing the drive water and exhaust water isolation valve , 1-(4jjy *The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperabl BRUNSWICK - UNIT 2 3/41-3 RETTPED-TECH. SPEC . Updated Thru.-Amend.-70 _ ,_
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REACTIVITY CONTROL SYSTEHL L1 HIT 1HC CONDITION FOR OPER ATION (Continued)
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ACTIONI (Continued) If the inoperable control rod (s) is insertedt a) Within one hour disarm the associated directional control valves eithert 1) Electrically, or 2) Hydraulically by ' closing the drive water and exhaust water isolation valves, b) Otherwise, be in at least HOT SHUTDOWN within the next 12 hours, With more than 8 control rods inoperable, be in at least HOT SHUTDOWN within 12 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram ' discharge volume drain and vent valves shall be demonstrated OPERABLE at least once per 31 days byt*
[' ' '} Verifying each valve to be ope <
, 1., Cycling each valve at least one complete cycle of full trave .1.3.1.2 All withdrawn centrol rods not required to have their directional control valves _ disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notcht At least once per 7 days when above the preset power level of the RWH l and At least once per 24 hours when above the preset power level of the RWM and any control rod is immovable as a result of excessive l friction or mechanical interferenc .1.3. All withdrawn control rods shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE, when-the reactor -protection system logic is tested per Specification 4.3.1.2,~ b verifying that the drain and vent valvest Close within 30 seconds after receipt of a signal for control rods to scram, and Open when the scram signal is reset or the scram discharge-volume-trip is bypasse '!
. $dg/ *These valves may be closed intermittently f or testing under administrative control.-
'3/4.1-4 Ame ndme nt No. 175
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REACTIVITY CONTROL SYSTEMS CONTROL ROD HAXIMUM SCRAM INSERTION TIMES LtHITINC CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion tirae of each control rod f rom the f ully withdrawn position to notch position 6, based on de-energitation of the scram pilot valve solenoids as time zero, shall not exceed 7.0 second APPLICABILITY OPERATIONAL CONDITIONS I and ACTION With the maximum scram insertion time of one or more control rods exceeding 7.0 seconds, operation may continue and the provisions.of Specification 3. are not applicable provided that t The control rod with the slow insettion time is declared inoperable, The requirements of Specification 3.1.3.1 are satisfied, and If within the preset power level of the RWM, the requirements of Specification 3.1.4.1.d are also satisfied, and The Surveillance Requirements of Specification 4.1.3.2.c are l d.
} } performed at least once per 92 days when operation is continued with O- three or more cont,rol rods with slow scram insertion timest Otherwise, be in at least IIOT SHUTD0VN within the next 12 hour SURVEILLANCE REQUIREMENTS 4.1. The maximum scram insertion time of the control rods shall be demonstrated through measurementi For all control rods prior to TilERHAL POWER exceeding 40% of RATED-TilERHAL POWER following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days, For specifically af fected individual control rods f ollowing maintenance on or modification to the control rod or rod drive system which could affect the scram insertion time of those specific control rods, and For 10% of the cont rol rods, on a rotating basis, at least once pe days of operation.
MA. /4 1-5 Amendment No.175 BRUNSWICK - UNIT 2 spr- w i7 y r, y.e gr-- -- yM y-%r-Wy .--@ #4F<"pelyG-.9--'p 4 e A,e- a+ - m,awm- --. ,, _, e --m +--
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CONTROL ROD WERAGE SCRAM INSERTION TDIES LIMITING CONDITIONS TOR OPERATION 3.1. The average scram insertion time of all OPERABLE control rods f rom the fully withdrawn position, based on de-energi:ation of the scram pilot valve solenoids as time zero, shall not exceed any of the followingt Position Inserted Trom Average Scram Inse r-Fully WithdrSwn tion Time (Seconds)
46 0.31 36 1.05 ' 26 1.82
.6 3.37
_A_P.C L IC A M ILITY : OPERATIONAL CONDITIONS 1 and . ACTION: With the average scram insertion time exceeding any of the above limits, be in at least 110T SHUTDOWN within 12 hour .
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SURVEILLANCE REQUIREMENTS 4.1. All control rods shall be demonstrated OPERABLE by scram time testing q from the fully withdrawn position as required by Surveillance Requirement 4.1. .
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4 i Myf BRUNSWICK - UNIT 2 3/4 16 Ame ndment No. 93
2' , REACTIVITY CONTROL SYSTEMS FOUR CONTROL ROD CROUP SCR AM INSERT!ON TIMES
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LIMITING CONDITION FOR OPERATJON .. 3.1.3.4 The average scram insertion time, f rom the f ully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following: Position Inserted from Average Scram Inser-Fully Wi t hd ra wn tion Time (Seconds) 46 0.33 36 1. 12 26 1.43 6 3.38 APPLICABILITYi OPERATIONAL CONDITIONS 1 and ACTION With the average scram insertion times of controi rods exceeding the above .
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limits, operation may continue and the provisions of Specification 3.0.4 are
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L# The control rods with the slower than avers'ge scram insertion times are declared inoperable, The requirements of Specification 3.1.3.1 are satisfied, and If within the preset power level of the RWH, the requirements of - Specification 3.1.4.1.d are also satisfied, and The Surveillance Requirements of Specification 4.1.3.2.c are l performed.at least.once per 92 days when operation is continued with j three or more control rods with slow scram insertion times.
l Ot he rwi s e , be iri at least HOT SHUTDOWN within the next 12 hours.
! L SURVEILLANCE REQUIREMENTS '
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f , 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing
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l from the fully withdrawn position as required by Surveillance Requirement l 4.1.3.2.
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I 3/4 1-7 Amendment No. 175 BRUNSWICK - UNIT 2 l ____m-- L -_ _
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CORTROL ROD- SCRAH ACCUMULATORS 1
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LIMITING CONDITION FOR OPERATION 3.1. All control rod scram accumulators shall be OPERABL APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*. ACTIONI
> In OPERATIONAL CONDITION 1 or 2 with one control rod scram accumulator inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within 8 hours t The inoperable accumulator is restored to OPERABLE status, or The control rod associated with the inoperable accumulator is i
declared inoperable, and the requirements of Specification 3.1.3.1 are satisfie . And, if within the preset power level of the RWH, the requirceents of Specification 3.1.4.1.d are also satisfied.
, Otherwise, be in at least HOT SHUTDOWN within the next 12 hour .
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L) In OPERATIONAL CONDITION 5* with a withdrawn control rod scram accumulator inoperable, fully insert the affected control rod and electrically disarm the directional control valves within one hou The provisions of Specification 3.0.3 are not applicabl SURVEILLANCE REQUIREMENTS 4.1. The control rod scram accumulators shall be determined OPERABLE! -
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~ At least once per 7 days by verifying that the pressure and leak detectors are not in the alarmed condition, and At least'once per 18 months by performance of at CHANNEL FUNCTIONAL TEST of the-leak detectors, and
- CHANNEL CALIBRATION of the pressure detectors.-
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*At least the accumulator associated with each withdrawn control ro applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
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w 3,/4 1-8 Amendment No. - 175 BRUNSWICK - UNIT 2
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1 2' REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE COUPLINC _
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LIMITINC CONDITION FOR OPERATION 3.1. All control rods shall be coupled to their drive mechanism APPLICAlilLITY: CONDITIONS 1, 2, and $*. j ACTION: In CONDITION 1 or 2 wi,th one control rod not coupled to its
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associated drive mechanism, the provisions of Specification 3.0.4 are not applicable and operation may continue provided 2 Withis the preset power level of the RWM, the control rod is l ' declared inoperable and fully inserted until recoupling can be attempted with THERMAL POWER above the preset power level _of the * RWH and the requirements of Specification 3.1.4.1.d are l satisfie . Above the preset power level of the RWM, the control rod drive is inserted to accomplish recoupling. If recoupling is not accomplished on the first attempt, declare the control rod inoperable, fully insert the control rod, and electrically . f' T * disarm the directional control valves. .
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t/ The requirements of Specification 3.1.3.1 are satisfie In CONDITION $*, with a withdrawn control rod not coupled to its associated drive mechanism, insert the control rod to accomplish recoupling. The provisions of Specification'3.0.3 are not a ppli ca bl e .
SURVEILLANCE R6QLU REMENTS 4.1.3.6 'The coupling integrity of- a control rod shall be demonstrated by -3 withdrawing the ' control rod to the fully withdrawn position and verifying that-the rod does not go to the overtravel positioni i
*At least each _ withdrawn control . cod. Not applicable to control rods removed .
per Specification 3.9.10.1 or 3.9.1 :
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BRUNSWICK - UNIT 2 3/4 1-9 Amendment No. .175
' ' :.; ~ .:..u a ._: REACTIVITY CONTROL SYSTEMS . 'J:-
SURVEILLANCE REOUIREKENTS (Continued)
. Prior to reactor criticality af ter completing CORE ALTF. RATIONS that could have af fected the control rod drive coupling integrity, Anytime the control rod is withdrawn to the " Full out" position in subsequent operation, and Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive coupling integrit ,
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BRUNSVICK - UNIT 2 3/4 1-10 RETYPED TECH. SPECS.. .. Uodated-Thru. Amend. 7
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_ - _ _ _ _ - _ N REACTIVITY CONTROL SYSTEMS CONTROL ROD POSITION INDICATION LIMITINO CONDITION FOR OPERATION 3.1. .11 control rod reed switch position indicators shall be OPERABL * APPLICABILITY: CONDITIONS 1, 2, and 5*. ACTION:
! In CONDITION 1 or 21 With one or more control rod reed switch position indicators inoperable, including " Full-in" or " Full-out" indication, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided that within one hourt The position of the control rod is determined by an l 1)
alternate method, or 2) The control rod is moved to a position with an OPERABLE l reed switch position indicator, or 3) The control rod with the inoperable reed switch position l indicator is declared inoperable and the requirements of Sp-cification 3.1.3.1 are satisfied;
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4) And, if within the preset power level of the RWM, the requirements of Specification 3.1.4.1.d are also satisfiedt
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Otherwise, be in at least HOT SHUTDOWN within 12 hcur l In CONDITION $* with a withdrawn control rod reed switch position indicator inoperable, fully insert the withdrawn control ro The ~ provi sions of Specification 3.0.3 are not applicabl .
*At least each withdrawn control rod. Not applicable to control rods removed a per Specification 3.9.10.1 or 3.9.1 /4 1-11 Amendment No. 175 BRUNSWICK - UNIT 2
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* REACTIVITY CONTROL SYSTEMS SURVFILLANCE REQUIREMENTS ,
4.1. The control rod reed switch position indicators shall be determined l OPERAB1.E by verifyings At least once per 24 hours, that the position of the control rod is indicated, That the indicated control rod position changes during the movement ' of the control rod when performing Surveillance Requirement 4.1.3.1.2, and That the control rod reed switch position indicator corresponds to the control rod position indicated by the " Full-out" reed switches when performing Surveillance Requirement 4.1.3. .
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3/4 1-12 Amendment No. 175 BRUNSWICK - UNIT-2
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7^^ REACTIVITY CONTROL SYSTEMS
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CONTROL ROD DRIVE HOUSING SUPPORT LIMITINC CONDITION FOR OPERATION 3.1.3.8 The control rod drive housing support shall be in place when there is fuel in the reactor vesse APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and ACTIONt With the control rod drive housing support not in place, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hour SURVEILLANCE REQUIREMENTS _ 4.1.3.8 The control rod drive housing support shall be inspected after reassembly and verified to be in place prior to start-up any time it has been disassembled or when maintenance has been performed in the ca.1 trol rod <lrive housing support are .
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I 3/4 1-13~ Amendment No. 163 BRUNSWICK - UNIT 2 l-I _ _ _ . _ _ _ , _ .,. _ -
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[?' REACTIVITY CONTROL SYSTEMS
*i 3/4 CONTROL ROD PROCRAM CONTROLS r
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ROD WORTH HINIMIZER . LIMITINC CONDITION FOR OPERATION 3.1. The Rod Worth Minimizer (RWM) shall be OPERABLE when THERMAL POWER is less than 10% of RATED THERMAL POWE APPLICABILITY: OPERATIONAL CONDITIONS I and 2*.
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ACTIONI With the RWM inoperable after the first 12 control rods have been f ully withdrawn on a startup, operation may continue provided that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified ' member of the plant t echnical staf With the RWH inoperable before the first 12 control rods are withdrawn on a startup, one startup per calender year may_ be performed provided that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staf ?? , 20- With RWM inoperable on a shutdown, shutdown may continue provided that control rod movement and compliance with the prescribed BPWS control rod pattern are verified by a second licensed operator or qualified member of the plant technical staf With RWM operable but individual control rod (s) declared inoperable, operation and control rod movement below the preset power level of the RWM may continue providedt No more than three (3) control rods are declared inoperable in any one BWS group, and, The inoperable control rod (s) is bypassed on the RWM and control rod movement of the bypassed rod (s) is verified by a second licensed operator or qualified member of the plant technical staf With RWM inoperable, the provisions of Specification 3.0.4 are not j applicable.
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* Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is-permitted for the purpose of determining the OPERABILITY of the RWH prior to
, o withdrawal of control rods for the purpose of' bringing the reactor to 'ky criticalit BRUNSWICX - UNIT 2 3/4 1-14 Amendment No.175 l
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REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMEliTS 4.1.4. The RWM shall be demonstrated GPERABLE in OPERAT10t4AL CONDIT10tl 2, prior to withdrawal of control rods f or the purpose. of making the reactor critical and in OPERATIONAL CONDITION 1 when the RWH is initiated during control rod insertion when reducing THERHAL POWER by: least one Verifying proper annunciation of the selection error of at out-of-sequence c ont r ol rod , and Verifying the rod block function of the RVH by moving an out-of-sequence control ro ~ 4.1.4.1.2 The RWM shall be demonstrated OPERABLE by verifying the control rod Banked Position Withdrawal Secuence input to the RWM computer is correct f ollowing any loading of the sequence program into the compute _ 1:. Q,/ BRUNSWICK - UNIT 2 3/4 1-14a Amendment No. 175 l
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i' REACTIVITY CONTROL SYSTEMS . KOD SEQUENCE CONTROL SYSTEM P Pages 3/4 1-15 through 3/4 1-16 have been delete ,. 1 )
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' 3/4 1-15 Amendment No. 175
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REACTIVITY CONTROL SYSTEMS ROD BLOCE MONITOR LIMITINC CONDITION FOR OPERATION 3.1. Both Rod Block Honitor (RBH) channels shall be OPERABL APPLICABILITY OPERATIONAL CONDITION 1 with: l
, THERMAL POWER greater than 30% of RATED THERHAL POWER and less than J
90% of RATED THERMAL POWER and the MINIMUM CRITICAL POWER R (HCPR) less than 1 70, or THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER and the HCPR less chan 1.4 ACTION!
+ With one RBH channel inoperable, POWER OPERATION may continue -
provided that either
, ~ The inoperable RBM channel is restored to OPERABLE status within ; ; 24 hours, or W' The redundant RBM is demonstrated OPERABLE within 4 hours and at least once per 24 hours until ene inoperable RBH is restored to OPERABLE status within 7 day Otherwise, trip at least one rod block monitor channe With both RBM channels inoperable, trip at least one rod block i
monitor channel within one hou I l
SURVEILLANCE REOUIREMENTS l ! 4.1.4.3 Each of the above required RBH channels shall be demonstrated OPERABLE by performance of a CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATI at the frequencies and during the OPERATIONAL CONDITIONS specified in Table 4.3.4- .
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b Amendment No.'168-l- noimevne - imtt 9 3/4 1 17
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RE ACTIV!TY CONTROL SYSTEMS l 3/4.t.5 STANDBY LIQUID CONTROL SYSTEM LIMITtNO CONDITION FOR OPERATION 3. The standby liquid control system shall be OPERABLE with: An OPERAB12 flow path f rom the storage tank to the reactor core, ' containing two pumps and t.'o inline explosive injection valves, The contained solution volume-concentration within the limits of Figure 3.1. 5- 1, and The solutto, temperature above the limit of Figure 3.1.5- ' APPLICABILITY: CONDITIONS 1, 2, and ACTION: in CONDITION 1 or 2r
'. = With one pump and/or one explosive valve inope rable , restore v the inoperable pump and/or explosive valve to OPERABLE status >
. within 7 days or be in at least HOT SHUTDOWN within the ne xt 12 hour . With the standby liquid control system inoperable, restore the system to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours, In CONDITION 5: With one pump and/or one explosive valve inoperable, re s to re the inoperable pump and/or explosive valve to OPERABLE status within 31 days or suspend all operations involving CORE ALTERATIONS or positive reactivity change . With the standby liquid control system inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes and fully insert all insertable control rods within one . hou I The provisions of Specification 3.0 3 are not applicabl ' 6, W BRUNSWICK - UNIT 2 3/4 1-18
& T&T 8&DY RETYPED TECit. SPEC Updated.Thru. Amend.'78
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. . REACTIVITY CONTROL SYSTD45 SpVEILLANCE 3200IRDiENTS 4. The standby liquid control system shall be demonstrated OPERABLE At least once per 24 hours by verifying thatt The volume and temperature of the sodium pentaborate solution are within the limits of Figures 3.1.5-1 and 3.1.5-2, and The heating tracing circuit is OPERABL At least once per 31 days byt Starting each pump and recirculating demineralized water to the test tank, Verifying the continuity of the explosive charge, and Determining the concentration of boron in solution by chemica analysis. This test shall also be performed anytime wa:er or .
boron is added to the solution or when the solution temperature drops below the limit established in Figure 3.1.5- , j At least once per 18 months during shutdown byl
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' 1 4" Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path *
l from the pumps to the reactor pressure vessel is available by ! pumping demineralized water into the reactor vesse The ) replacement charge for the explosive valve shall be from the i same manufactured batch as the one fired or from another batch I which has been certified by having one of that batch ! successfully fired. Both injection test loops shall be tested j in 36 month . Demonstrating that the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to 1190 psig is me . Demonstrating that the pump relief valve setpoint is 1450 t 50 poi l h k:1 BRUN5VICK - UNIT 2 3/4 1-19 Amendment No. 143
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3/4.2 POWER DISTRIBUTION LIMITS 3/4. AVERACE PLANAR LINEAR HEAT CENERATION RATE LIMITINC CONDITION FOR OPERATION 3.2.1 During power operation, the AVERACE PLANAR LINEAR HEAT CENERATION RATE ( APLHCR) for each type of fuel as a function of axial location and AVERACE PLANAR EXPOSURE shall not exceed limits based on applicable APLHCR limit values that have been approved for the respective fuel and lattice type and When hand determined by the approved methodology described in CESTAR-I calculations are required, the APLHCR for.each type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the limiting value, adjusted for core
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flow and core power, for the most limiting lattice (excluding natural uranium) of each type of fuel shown in the applicable figures in the CORE OPERATING LIMITS REPOR , APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWE ACTION: With an APLHCR exceeding the limits specified in Technical Specification 3. initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the required limits within 4 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hour SURVEILLANCE REQUIREMENTS 4. All APLHCRs shall be verified to be equal to or less than the limits specified in Specification 3.2.1: At least once per 24 hours, Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHC .
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3/4 2-1 Amendment No. 168 BRUNSWICK - UNIT 2
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I . POWER DISTRIBUTION LIMITS 3/4.2.2 MININUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.2.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, core power, and cycle average exposure, shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT. The MCPR limits for ODYN OPTIJ.N A and ODYN OPTION B analyses, used in the above determination, shall be specified in the CORE OPERATING LIMITS REPOR APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER ACTION: With MCPR, as a function of core flow, core power, and cycle average exposure, i less than the applicable MCPR limit specified in the CORE OPERATING LIMITS l REPORT, initiate corrective action within 15 minutes and restore MCPR to within the applicable limit within 4 hours or reduce THERMAL POWER to less than 25% of RATED THERHAL POWER within the next 4 hour SURVEILLANCE REQUIREMENTS 4.2. MCPR, as a function of core flow, core power, and cycle average exposure, shall be determined to be equal to or greater than the applicable limit determined of Specification 3.2.2.1: At least once per 24 hours, Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is operating in a LIMITING CONTROL ROD PATTERN f or MCP l BRUNSWICK - UNIT 2 3/4 2-2 Amendment No. 168
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POWER DISTRIBUTION LIMITS 3/4.2.2 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B) LIMITINC CONDITION FOR OPERATION 3.2.2.2 For the OPTION B MCPR limits provided in the CORE OPERATING LIMITS REPORT to be used, the cycle average 20 (notch 36) scram time (r,y,) shall be less than or equal to the Option B scram time limit (tB), where T,y, and t 3 are determined as follows: a I y ave
. I"I n N "I i , where __
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i=1 i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle (including BOC), surveillance test, and N t == Number Average scram of rodstime tested in the to notch 36 ich for surveillance test i t; t =u+ .65 ( n ' "#" B
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i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including BOC), surveillance test N t = Number of rods tested in the ich Ny = Number of rods tested at BOC, - u = 0.813 seconds (mean value for statistical scram time distribution from de-energitation of scram pilot valve solenoid to pickup on notch 36), a = 0.018 seconds (standard deviation of the above statistical distribution) APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWE BRUNSWICK - UNIT 2 3/4 2-3 Amendment No. 168
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_ . . . . _ . . 2 _ . - _ _ _ _m -. . . _ . . . _ . - - . _ _ . _ . _ . . , _ __ _ . _ . _ _ l .. i POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION (Continued) , ACTION: 1 Within twelve hours after determining that t,y, is greater than tg, the operating limit MCPRs shall be either:
. Adjusted for each fuel type such that the operating limit MCPR is the maximum of the non pressurization transient MCPR operating limit specified in the CORE OPERATINC LIMITS REPORT or the adjusted pressurization transient MCPR operating limits, where the adjustment is made by: ,
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t t
'"* - MCPR option -
MCPR = MCPR option B
+ (MCPR ,
option A B) a d j.usted t g -t g where: t = 1.05 seconds, control rod average scram insertion time A limit to notch 36 per Specification 3.1.3.3, , MCPR option A * Specified in the CORE OPERATINC LIMITS REPORT, MCPR option B " S ecified P in the CORE OPERATING LIMITS REPORT, or, The OPTION A HCPR limits specified in the CORE OPERATING LIMITS-REPORT.
l l-SURVEILLANCE REQUIREMENTS
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l ' 4.2. The values of t and t shall be determined and compared each time l a scram time test is perfoImed, kherequirement for the frequency of scram time testing shall be identical to Specification 4.1. .
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e i 3/4.6 CONTAINHENT SYSTEMS 3/4. PRIMARY CONTAINMENT PRIMARY CONTAIRMENT INTECRITY LIMITINC CONDITION FOR OPERATION 3.6. PRIKARY CONTAINMENT INTECRITY shall be maintaine APPLICABILITY: CONDITIONS 1, 2. and ACTION: Without PRIKARY CONTAINMENT INTECRITY, restore PRIKARY CONTAINMENT INTECRITY within 2 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour SURVEILLANCE REQUIREMENTS
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4.6. PRIMARY CONTAINMENT INTECRITY shall be demonstrated: At least once per 31 days by verifying that all primary containment penetrations" not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in Table 3.6.3-1 or Specification 3. i i By verifying each primary containment air lock OPERABLE per Specification 3.6. By verifying the suppression pool OPERABLE per Specification 3.6. _
* Exrept valves, blind flanges, and deact: :ed automatic valves which are located inside the containment, the MS!' :t, the RWCU Penetration Triangle Room, or the TIP Room, and are locked, scaled, or otherwise secured in the closed position. Tha._ penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been de-inerted since the last verification or =cre often than once per 92 days. Those val'?es located above the dryvell head requiring head shield block removal for verification will be verified prior to each replacement of the shield block I SRUNSWICK - UNIT 2 3/4 6-1 Amendment No. 113
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i CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6. Primary contain=ent leakage rates shall be limited to: An overall integrated leakage rate of: Less than or equal to L , 0.5 percent by weight of the containment air per 24 $ours at P,, 49 psig, or Less than or equal to L , 0.357 percent by weight of the
- containment air per 24 hours at a reduced pressure of Pg ,
25 psi A combined leakage rate of less than or equal to 0.60 L, for all penetrations and all valves listed in Table 3.6.3-1, except for main steam line isolation valves *, subject to Type B and C tests when ' pressurized to P,, 49 psi *Less than or equal to 11.5 scf per hour for any one sain s: cam line inciation valve when tested at 25 psi APPLICA31LITY: '~
- hen PRIMARY CONTAISMENT INTEGRITY is required per Specification 3.6. ACTION:
With: The measured overall integrated primary containment leakage rate exceeding 0.75 L, or 0.75 Lg , as applicable, or _ The =easured combined leakage rate for all penetrations and all - valves listed in Table 3.6.3-1, except for =ain stea line isolation valves *, subject to Type 3 and C tests exceeding 0.60 L,, or The seasured leakage rate exceeding 11.5 scf per hour for any one sain steam line isolation valve, restore: The overall integrated leakage rate (s) to less than or equal to 0. 7 5 L, or 0. 7 5 Lg , as applicable, and The combined leakage rate for all penetra:1ons and all valves listed in Table 3.6.3-1, except for sain steam line isola: ion valves *, subjee: :o Type B and C tes ts to less than er equal to 0.60 L,. and
* Exemption to Appendix "J" of 10 CFR 5 BR.'JNSWICK - UNIT 2 3/4 6-2 Acendment No. 91
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CONTAIRMENT SYSTEMS LIMITINC CONDITION FOR-OPERATION (Continued) ACTION (Continued) The leakage rate to less than or equal to 11.5 sef per hour for any one main steam line isolation valve, prior to increasing reactor contant system temperature above 212* SURVEILLANCE REQUIREMENTS
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4.6. The primary containment leakage rates shall be demonstrated at the following schedule and shall be determined in~conformance with the criteria specified in Appendix J of 10CFR50: Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 1 10 month intervals during shutdown at P 49 psig, or P t, 25 psig, during each 10 year service period.,,The third test of each set shall be conducted during the shutdown for the 10 year plant inservice inspectio If any periodic Type A test fails to meet either 0.75 L, or 0.75 L the test scheduleforsubsequentTypeAtestsshallbereviewedank, approved by the Commissio If two consecutive Type A tests fail to meet 0.75 L, or 0.75 Lg, a Type A test shall be performed at each plant shutdown for refueling or every 18 months, whichever occurs first, until two consecutive Type A tests meet 0.75 L, or 0.75 Lg , at which time the above test schedule may be resume The accuracy of each Type A. test shall be verified by a supplemental test which: Confirms the accuracy of the test by verifying tnat the-difference between the supplemental data and'the Type A test data is=within 0.25 Lg or 0.25 Lt . Has duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental tes . Requires the quantity of gas injected into the containment or bled from the containment.during the supplemental test to be equivalent to at least 25 percent of the total measured leakage at P,, 49 psig or-P g, 25 psi . BRUNSWICK - UNIT 2 3/4 6-3 -Amendment No. 166
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SURVEILLANCE REOUIRDENTS (Continued) Type B and C tests shall be conducted with gas at P , 49 psig, at intervals no greater than 24 months except fer tests involving: Air locks, Main steam line isolation valves, Air locks shall be tested and demonstrated OPERABLE per Surveillance taquirement 4.6.1.3.
. Main steam line isolation valves shall be leak tested at least once per 18 month All test leakage rates shall be calculated using observed data converted to absolute values. Error analyses shall be performed to select a balanced integrated leakage measurement syste The provisions of Specificatien 4.0.2 are not applicable to 2t. month and 40 t 10 month surveillance interval . BR;;s5 WICK . UNIT 2 3ft. 6-3A Amendment No. 91 _ _
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. . CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCKS u;MITING CONDIT!ON FOR OPERATION 3.6. The primary containment air lock shall be OPERABLE with: Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one rir
.ccx door snall be closed, and An overall air lock leakage rate of les i than or equal to 0.05 L3 at _
P ' '9 psi 3' a Ci_:- . OPERATIONAL CONDITIONS 1, 2* and T : '. . _ . Witn one pr. mary containment air lock door inoperable:
'.. Maintain at least the OPERABLE air lock door closed and either restere the inoperable air lock door to OPERABLE status within 24 hours or lock the OPERABLE air lock door close . Operation may then continue until performance of the next required overall air loct leakage test provided that the OPERABLE air lock decr is verified to be locked closed at least once per 31 day . Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour The prc'.tsions of Specification 3.0.4 are not applicabl :. Atta the primary containment air lock door interlock inoperable: Lock the inner air lock door close . Operation may then continue provided that the inner air lock door is .*erified to be locked closed at least once per 31 day . Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour The previsions of Specification 3.0.4 are not applicabl :. With the prim.ary containment ai r lock inoperable, except as a result of an in peraole air '.ock door or interlock, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour * See Special Test Exception 3.1 BRUNSWICK - UNIT 2 3/4 6-4 Amendment No. 158
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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6. Each primary containment air lock shall be demonstrated OPERABLE: By verifying the seal leakage rate to be less than or equal to 5 scf per hour when the gap between the door seals is pressurized to 10 psig*: Within 72 hours following each closing, except when the air lock is being used for multiple entries then at lease once per 72 hours, and Prior to establishing PRIMARY CONTA1VMENT INTECRITY wnen the air lock has been used and no maintenanct has been performed on the air lock, and When the air lock seal has been rep. ace By conducting an overall air lock leakat,e that at P , 49 psig, and by verifying that the overall air lock lea tage is with,in its limit: At least once per six months #, ar d Prior to establishing PRIMARY CCNTAINMENT INTECRITY when maintenance (except for seal replacement) has been performed on the air lock that could affect the air lock sealing capability.* By verification of air lock inter'.ock OPERABILITY: Prior to establishing PPlHARY CONTAINMENT INTEGRITY when the air lock has been u:ed, and Prior to and following a drywell entry when PRIMARY CONTAIUMENT INTEGRITY is required, and Following the performance of maintenance affecting the air lock interloc .
* Exemption of Appendix J of 10 CFR 5 # The provisions of Specification 4.0.2 are not applicable
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BRUNSWICK - UNIT 2 3/4 6-5 Amendment No. 158
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CONTAINMENT SYSTEMS l PRIMARY CONTAINMENT STRUCTURAL INTEORITY LIMITING CONDITION FOR OPERATION 3.6. The structural integrity of the primary containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6. APpLICASILITY: OPERATIONAL CONDITIONS 1, 2, and ACTION: With the structural integrity of the primary containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 212* SURVEILLANCE REQUIREMENTS _ 4.6.1. The structural integrity of the exposed accessible interior and
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exterior surfaces of the primary containment, including the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test by a visual inspection of those surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradatio ~ 4.6.1.4.2 Reports Any abnormal degradation of the primary containment structure detected during the above required inspections shall be reported to the Commission pursuant to Specification 6. This Special Report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions take .
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BRUNSWICK - UNIT 2 3/4 6-6 Amendment No. 110
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CONTAISHENT SYSTEMS t PRIKARY CONTAINMENT INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION 3.6. Primary containment internal pressure shall be maintained between-0.5 and 1 75 psi APPLICABILITY: CONDITIONS 1, 2, and ACTION: With the containment internal pressure outside of the specified limit, restore the internal pressure to within the limit within I hour or be in at least HOT SHUTDOWN vithin the next 12 hours and in COLD SHUTDOW within the following 24 hour . SURVEI Lt.ANCT. REOUIREMENTS _ 4.6.1.5 The primary containment internal pressure shall be determined to be within the limits at least once per 12 hours.
BRUNSWICK - UNIT 2 3/4 6-7 RETYPED TECH. SPEC Updated Thru. Amend. 78 _ __ _ . . _ _ _ . _ _ . . .. . .
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. . , ,s CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AVERACE AIR TEMPERATURE LIMITINC CONDITION FOR OPERATION 3.6. Primary containment average air temperature shall not exceed 135 0 F.* I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and l ACTION: With the primary containment average air temperature > 135 F*, reduce the l average air temperature to within the limit within 8 hours, or be.in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 12 hour SURVEILLANCE REQUIREMENTS 4.6. The primary containment average air temperature shall be the volumetric average of the temperatures at the following locations and shall be determined at least once per 24 hours Location Below 5' elevation, Between 10' and 23' elevation, Between 28' and 45' elevation, Between 70' and 80' elevation, and Above 90' elevatio .
* The primary containment average air temperature limit may be increased to 140*F until August 15, 1985, at which. time the limit will be returned to 135' BRUNSWICK - UNIT 2 3/4 6-8 Amendment No. 112 , _ _ _ _ ____.___.________i -
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. CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS ,
SJPPRESSION CHAMBER
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LIMITING COSDIT!0N FOP. OPERATION 3.6. The suppression chamber <nall be OPERABLE with: The pool water: Volume between S7.6003 f t and 89,600 ft3, equivalent to a level between -27 inches and -31 inches, and a Maximum average temperature of 95'F during OPERATIONAL CONDITION 1 or 2, except that the saximum average temperature cay be permitted t'o increase to: a) 105'F during testing which adds heat to the suppression chambe . b) 110'F with THERMAL POVER less than or equal to'1% of RATED TRERMAL POWE c) 120*F with the sain steam line isolation valves closed following a scra Two OPERABLE suppression chamber water temperature instrumentation
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channels- with a minimum of 11 operable RTD inputs per channe A total leakage f rom the drywell to the suppression chamber of less than the equivalent leakage through a 1-inch diameter orifice at a differential pressure of 1 psig.
I APPLICASILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION: With the suppression chamber water level outside the above limits, _ restore the water level to within the limits within 6 hours or be in at least HOT SKUTDOWN within the next 12 hours and in' COLD SMJTDOVN within the following 24 hour In OPERATIONAL CONDITION 1 or 2 with the suppression chamber average ' water temperature greater-than 95'F, restore the average temperature to less than or equal to 95'T within 24 hours or'be in at least M0T SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, except, as permitted above: I l
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BNJNSWICK - UNIT 2 3/4 6-9 Amendment No. 103
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. . CONTAINMENT SYSTEMS LIMITING CONDITIONS FOR OPERATION (Continued) ACTION: (Continued) With the suppression chamber average water temperature greater than iv5'T :aring testing which adds hest to the suppression chamber. stop all testing which adds heat to the suppression chamber and restore the average temperature to less than or equal to 95'T witnin 24 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour . With the suppression chamber average water te=perature' greater than 110*F manually scram the reactor and operate at leas t one residual heat removal loop in the suppression pool cooling mod . With the suppression chamber average water temperature greater than 120*F, depressurize the reactor pressure vessel to less than 200 psig within 12 hour With one suppression chamber water temperature instrumentation ~ channel inoperable, restore the inoperable channel to OPERA 3LE status within 7 days or verify suppression cha=ber water temperature to be - within the limits at least once per 12 hour With both suppressiot. chamber water temperature instrumentation channels inoperable, restore at least one inoperable temperature instrumentation channel to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 212* SURVEILLANCE REQUIREMENTS 4. 6. The suppression chamber shall be demonstrated OPERASLE: By verifying the suppression chamber water volume to be within the limits at least once per 24 hour BRUNSWICK - UNIT 2 3/4 6-10 Amendment No. 103
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J CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) At least once per 24 hours in OPERATIONAL CONDITION 1 or 2 by verifying the suppression chamber average water temperature to be
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le~ss than or equal to 95'F, except: , At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber
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average water temperature to oe less than or equal to 105' . At least once per hour when suppression chamber average water temperature is greater than 95'F, by verifying: a) Suppression chamber average water temperature to be less than or equal to 110*F, and b) THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWE . At least once per 30 minutes following a scram with suppression chamber average water temperature greater than 95'F, by
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verifying suppression chamber average water temperature less than or equal to 120* By an external visual examination of selected emergency core cooling system suction line penetrations of the suppression chamber enclosure prior to taking the reactor from COLD SHUTDOWN af ter safety / relief valve operation with the suppression chamber average water temperature greater than or equal to 160*F and reactor coolant system pressure greater than 200 psi By verifying at least two suppression chamber water temperature instrumentation channels OPERABLE by performance of a - CHANNEL CHECK at least once per 24 hour . CHANNEL FUNCTIONAL TEST at least once per 31 days, and (- CHANNEL CALIBRATION at least once per 18 months (550 days).
r I with the temperature alarm setpoint for high water temperature less: i than or equal to 95' At least once per.18 months by: , A visual-inspection of the. accessible interior of the suppression chamber and exterior of the suppression chamber enclosure.
l l . BRUNSWICK - UNIT 2 3/4 6-10a Amendment No. 160 l
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, , CONTAI.WENT SYSTEMS ' SURVEILLANCE REQUIREMENTS (Continued) Conducting a drywell-to-suppression chamber bypass leak test at !. an initial dif f erencial prer.sure of 1 psig and verifying that I the dif f erential pressure does not decrease by more than 0.45 taches of water per ninute for a 10 .inute perio .
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BRUNS WICK - UNIT 2 ,4 6-10b Amend:sent No. 10 .
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SUPPRESSION POOL C00LINC LIMITING CONDITION FOR OPERATION 3.6. The suppression pool cooling mode of the residual heat removal (RHR) system shall be OPERABLE with two independent cooling loops, each loop consisting of two pumps and one heat exchange APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and ACTION: . ,c With one RER suppression pool cooling loop inoperable, operation may
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continue and the provisions of Specification 3.0.4 are not applicable; restore the inoperable loop to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour With both RHR suppression pool cooling loops inoperable, restore at least ene loop to OPERABLE status within 8 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour SURVEILLANCE REQUIREMENTS
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4.6. The suppression pool cooling mode of the RHR system shall be demonstrated OPERASLE: At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct positio At least once per 92 days by verifying that each RER pump can be started from the control room and develops a flow of at least 7,700 gpm on recirculation flow through the RER heat exchanger and the suppression poo h ,W?
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BRUNSWICK - UNIT 2 3/4 6-11 Amendment No. 122 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
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l 4 . CONTAINVENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION ,,
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1.6.3 The primary containment isolation valves specified in Table 3.6.3-1 shall be OPERAB17. with isolation times as shown in Table 3.6.3-1, and the rmetor instrumentation system isolation valves shall be OPERABLE . APPt.ICABILITY: CONDITIONS 1, "!, and ACTION: With one or more of the primary containment isolation valves specified in Table 3.6.3-1 inoperable, operation may continue and tha provisions of Specificatior. 3.0.4 are not applicable prov,1ded that v least one isolation valve is maintained OPERABLE in each af f ected penetration that is open and either: The inoperable valve (s) is restored to OPERABLE status withia A hours, or Each af fected penetration line is isolated within A hours by use of at least one deactivated automatic valve secured in the isolation position, or Each affected penetration line is isolated within 8 hours by use of at least one closed manual valve or blind flang ' Otherwise, be in at least HOT SHUTD0b'N within the next 12 hours and in COLD SHUTD06N within the f ollowing 24 hour With one or more of the reactor instruentation system isolation valves inoperable, operation may conti cae and the provisions of Specifications 3.0.3 and 3.0.4 are n.t applicable provided that within A hours; The inoperable valve is returned to OPERABLE status, or The instrument line is isolated and the associated instrumet ' declared inoperabl Otherwise, he in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTD04N within the following 24 hour %t!NSW[CK - LNIT 2 3/4 6-12 RETYPED TECH. SPEC Updated Thru. Amend. 78 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ __ _
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CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.3.1 Each primary containment isolationvalvespecifiedinTjble3.6.3-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair, or replacement work is performed on the valve or its associated actuator, control, or power circuit by performance of the cyclin test and verification of isolation tim .6. Each isolation valve specified in Table 3.6.3-1 shall be demonstrated OPERABLE at least once per 18 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation positio .6. The isolation time of each power-operated or automatic valve , specified in Table 3.6.3-1 shall be determined to be within its limit when tested pursuant to Specification 4. .6.3.4 Each reactor instrumentation system isolation valve shall be demonstrated OPERABLE at least once per 18 months by cycling each valve through at least one complete cycle of full trave . l
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i i , l l l BRUNS * DICK - UNIT 2 3/4 6-13 Amendment No. 137 i
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- PRIMARY CONTAINMENT 190LATION VALVES Table 3.6.3-1 nas been delete Refer to Plant Procedure RCI-0 Pages 3/4 6-15 through 3 /4 6-17 have been delete (Next page is 3/4 6-18) l_
BRUNSWICK - UNIT 2 3/4 6-14 Amendment No. 179
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1 . . l CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF DRYWELL - SUPPRESSION CH AMBER VACUUM BREAKERS LIMITING CONDITION FOR OPERATION 3.6. All drywell-suppression chamber vacuum breakers shall be OPERABLE and in the closed position with: The position indicator OPERABLE, and An opening setpoint of less than or equal to 0.5 psi __ APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and ACTION: With no more than 2 dryvell-suppression chamber vacuum breakers inocerable for opening but known to be in the closed position, the provisions of Specification 3.0.4 are not applicable and operation may continue until the next COLD SHUTDOWN provided the surveillance requirements of Specification 4.6.4.1.a are performed on the OPERABLE vacuum breakers within 4 hours and at least once per 15 days thereafter until the inoperable vacuum breakers are restored to OPERABLE status; otherwise be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, l With one drywell-suppression chamber vacuum breaker in the open position as indicated by the position-indicating system, the provisions of Specification 3.0.4 are not applicable and operation may continue provided the surveillance requirements of Specification 4.d.4.1.a are performed on the OPERABLE vacuum breakers and the surveillance requirements of Specification 4.6.4.1.b are performed witnia 8 hours and at least once per 72 hours thereafter until the inoperable vacuum breaker is restored to the closed position; otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour With the position indicator of any drywell-suppression chamber vacuum breaker inoperable, the provisions of Specification 3.0.4 are not applicable and operation may continue, provided the surveillance requirements of Specification 4.6.4.1.b are performed within 8 hours and at least once per 72 hours thereafter until the inoperable position indicator is returned to OPERABLE status; otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour BRUNSWICK - UNIT 2 3/4 6-18 Amendment No. 159 _____ __-_____ - __- -______ - _
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. . CONTAINMENT SYSTEMS StTRVEILLANCE REQUIREMENTS 4.6. Each drywell-suppression chamber vacuum breaker shall be demonstrated OPERABLE:
. At least once per 31 days and af ter any discharge of steam to the suppression chamber f rom any source, by exercising each vacuum breaker through'one complete cycle and verifying that each vacuum breaker is closed as indicated by the position indication syste Whenever a vacuum breaker is in the open position, as indicated by the position indication systam, by conducting a test that verifies that the dif f erential pressure is maintained greater than 1/2 the initial delta P for one hour without N2 makeu At least once per 18 months during shutdown by: Verifying the opening setpoint, f rom the closed position, to be less enan or equal to 0.5 psid, Performance of a CHANNEL CALIBRATION that each position -
indicator indicates the vacuum breaker to be open if the vacuum breaker does not satisfy the delta P test in 4. 6. 4.1. . 4 t BRUNSWICK - UNIT 2 3/4 6-19 Amendment No. 103
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CONTAINMENT SYSTEMS SUPPRESSION POOL - REACTOR BUILDINC VACUUM BREAXERS LIMITING CONDITION FOR OPERATION 3.6. All suppression pool-Reactor Building vacuum breakers shall be { OPERABLE with: an opening setpoint of less than or equal to 0.5 psid ! I an OPERABLE Nitrogen Backup System consisting of two independent ! subsystems (one subsystem for each vacuum breaker). 3 APPLICABILITY OPERATIONAL CONDITIONS 1, 2, and ACTION:
. . With one suppression pool-Reactor Building vacuum breaker inoperable for- l opening but known to be in the closed position, restore the inoperable vacuum breaker to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour With one Nitrogen Backup System subsystem inoperable, verify the remaining subsystem is OPERABLE and restore the inoperable subsystem to OPERABLE status within 3I days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hour With both Nitrogen Backup System subsystems inoperable, restore at least one inoperable subsystem to OPERABLE status within 7 days; otherwise, be j in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN l within the following 24 hour !
SURVEILLANCE REQUIREMENTS 4.6.4. Each suppression pool-Reactor Building vacuum breaker shall be l demonstrated OPERABLE At least once per 92 days by: Manually verifying that each vacuum breaker check valve is free to-open, and Cycling each vacuum breaker butterfly valve through at lease one complete. cycle of full trave At least once per 13 months by Demonstrating that the force required to open each vacuum breaker L check valve does not exceed 0.5 psid.
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. t SURVEILLANCE REQUIREMENTS (Continued) . Demonstrating that the vacuum breaker butterfly valve opens at -0.45 1 0.05 psid, dryvell pressure going negative relative to Reactor Building-pressur . Visual inspection .6.4. The Nitrogen Backup System shall be demonstrated OPERA 3LE: At least once per 24 hours by verifying that each subsystem is pressuri:ed to greater than or equal to 1130 psi At least once per 18 months by verifying that each subsystem maintains system pressure with a leakage rate of less than or equal to .65 SCFM at a starting pressure greater than or equal to 1130 psi ! At least once per 18 months by performing a logic system functional test to ensure actuation of the nitrogen backup syste , BRUNS *4ICK - UNIT 2 3/4 6-20a Amendment No. 138
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CONTAIhMENT SYSTEMS,
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3/4.6.5 SECONDARY C0hTAIhWENT SECONDARY CONTAINMENT INTEGRITY LIMITING CONDITION POR OPERATION 3 6. SECONDARY CONTAIhWENT IhTEGRITY shall be maintaine APPLICABILITY: CONDITIONS 1, 2, 3, 5, and *. ACTION:
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Without SECONDARY CONTAINMENT INTEGRITY, restore SECONDARY CONTAINHENT INTEGRITY within 8 hours, or: In CONDITION 1, 2 OR 3, be in at least HOT SHUTDOWN within the nex' 12 hours and in COLD SHUTDOWN within the following 24 hour In CONDITION 5 or *, suspend irradiated fuel handling in the secondary containment, CORE ALTERATIONS, and activities which could reduce the SWUTDOWN MARGIN. The provisions of Specification h are not applicabl SURVEILLANCE REOUIREMENTS , 4.6.5 1 SECONDARY COh7AlhMEh7 INTEGRITY shall be demonstrated by verif ying : At least once per 92 days that each secondary containment isolation damper is OPERABLE or secured in the closed position per _ Specification 3.6 At least once per la months by operating a standby gas treatment system for I hour and maintaining > 1/4 inch of vacuum, water gauge , at a flow rate not exceeding 3000 CT *When irradiated fuel is being handled in the secondary containmen ! BRUNSWICK - UNIT 2 3/4 6-21 RETYPED TECH. SPEC Updated Thru. Amend. 78
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SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS LIMITING CONDITION FOR OPERATION 3.6.5.2 The secondary containment automatic isolation dampers shown in Table 3.6.5.2-1 shall be OPERABL APPLICABILITY: OPERATIONAL CONDITIONS 1. 2, 3, 5, and *. l ACTION: With one or more et the secondary containment isolation dampers specified in Table 3.6.5.2-1 inoperable, operation may c.ontinue and the provi sions of specification 3.0.4 are not applicable, provided that at least one isolation damper is maintained OPERABLE in each affected penet ration t ha t 1, 5 open, and: The inoperable damper is restored to OPERABLE status within d hours, or The atfected penet ation is isolated by use of a closed damper within 8 hours, or SECONDARY CONTAINMENT INTECRlfY is demonstrated within d hours and the damper is restored to OPERABLE stat us within 7 day Otherwise, in OPERATIONAL CONDITION 1, 2, or 3. be in at least HOT l SHUTDOWN wit hin the next 12 hours and in COLD SHUTDOWN within the following 24 hour Otherwise, in OPERATIONAL CONDITION 5 or *. suspend irradiated fuel l handling in the secondary containment, CORE ALTERATIONS, or activities that could reduce the SHUTDOWN MARCIN. The provisions et Specification 3.0.3 are not applicaol .
*When irractated tuel is being handled in the secondary containmen .
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BRUNSWICK - UNIT 2 3/4 6-22 Amendment No. 179
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CONTA!KKENT SYSTEMS SURVEILIANCE PIOU1REKENTS -- 4.6.5.2 Each secondary containment automatic isolation damper specified in Tabit 3.6.5.2-1 shall te demonstrated OPERABLE: At least once per 92 days by cycling each automatie isolation damper testable during plant operation through at least one complete cycle of full trave Prior to returning the damper to service af ter maintenance, repair, or replacement work is perf ormed on the damper or its associated actuator, control, or power circuit by perf ormance of the cycling test and verification of isolation tim At least once pe r la months during COLD SHUTDOWN or REFUELING by: Cycling each automatie damper through at least one complete cycle of full travel and measuring the isolation time, and Verifying that on a secondary containment isolation test signak each automatic damper actuates to its isolation positio ,
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BRUNSWICK - UNIT 2 3/4 6-23 RETYPED TECH. SPEC , Updated Thru. Ame nd . 7 6 I- ____
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C0KTAINMEKT SYSTEMS 3/4.6.6 CONTAINMENT ATHOSPHERE CONTROL STANDBY CAS TREATKENT SYSTEM LLMITING CJ:431T10S TOP. opt.u!!n 3.6. Two independent standby gas treatment .absyt tees shall be OPERABL l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2. 3, 5, a nd * . g ACTION : With one standby gas treatment subsystem inoperable: In OPERATIONAL CONDITION 1, 2, or 3. restore the inoperable l sybsystem to OPERA 3LE status within 7 days or be in at least HOT SHUTDOVN within the next 12 hours and in COLD SHUIDOVN witnin t following 24 hour . In OPEKATIONAL CONDITION 5 or *, restore the inoperable subsystem to l OPERABLE status within 31 days or suspend irradiated fuel handling in the secondary containment, CORE ALTERATIONS, or operations that could reduce the SHUTDOWN KARGIN. The provisions of Specification 3.0.3 are not applicabl With both standby gas treatment subsystems inoperable; In OPERATIONAL CONDITION 1, 2, or 3, be in at least HOT SHUTDOVN l within 12 hours and in COLD SHUTDOWN within the next 24 hour . In OPERATIONAL CONDITION 5 or *, suspend all irradiated f uel l handling in the secondary containment, CORI ALTERATIONS, or operations that could reduce the SHUTDOWN KAAGIN. The provisions of Specification 3.0.3 are not applicabl SURVEILLA.NCE REQUIREMENTS 4.6. Each standby gas treatment subsystem shall be demonstrated OPERABLE: At least once per 31 days by initiating, from the control room, flow through the REPA filters and charcoal adsorbers and verifying that the subsystem operates for at least 10 hours with the heaters on g aut omatic contro *W' hen irradiated t;uel is being handled in the secondary containmen BRUNSVICK - UNIT 2 3/4 6*25 Ame nd me nt No . 102
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CORTAINMENT SYSTL1S, SURVE11. LANCE PlQUlPlMENT$ (Continued) At least once per 18 oonths or (1) af ter any structural maintenance on the KEPA filter or chatcoal adsorber housings, or (2) following painting, fire, or enemical release 11 any isati.atte, tone comcunicating with the system by: Verifying that the cleanup system satisfits tne in piace testing acceptance criterta and uses the test procedures of Ragulatory Positions C.5.a. C.S.c, and C.5.d of Regulatory Guide 1.52, Revision 1. July 19 70, and the system flos rate is 3000 cfm 1 10%. Verifying within 31 days af ter removal that a laboratory analysis of a representative carbon sampia obtained in accordance with Regulatory Position C 6.b of Regulatory Guide !.52, Revision 1. July 19 76, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 1, July 19 7 , Verifying a system flow rate of 3000 cfm 110% outing system operation when tested in accordance with ANS1 N510-19 7 Af t e r eve ry 720 hours of charcoal adsorber operation by verif ying within 31 days af ter removal that a laboratory analysis of representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 1. July 1976, meets the laboratory l testing criteria of Ragulatory Position C.b.a of Regulatory Guide 1.52 Revision 1, July IV 7 At least once per 18 months by: Verifying that the pressure drop across the combined REPA filters and charcoal adsorber banks is last than 8.5 inches ' dater Gauge l while operating the filter train at a flow rate of 3000 ef a ; 10%. Verifying that the filter train starts on each secondary containment isolation test signa . Verifying that the heaters will dissipate at least 15.2 kw unen tested in accordance with ANS1 d510-19 7 BRUNSWICK - UNIT 2 3/4 6-26 Ame nd me nt No. 102
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CONTA!WENT SYSTEMS SURVEILLANCE PIQUIREMENTS (Continued)
.. After each complete or partial replacement of a REPA filter bank by verifying that the HEPA filter banks remove > 99% of the DOP when they are tested in-place in accordance with M45I N510-1975 while operating the system at a flow rate of 3000 cfm + 10%. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove > 99% of a halogensted hydrocarbon ref rigerant test gas when the7 are tested
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in-place in accordance with M4S1 N510-1975 while operating the system at a flow rate of 3000 cfm + 10%. ,
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RETYPED TECH SPEC BRUSSWICK - UNIT 2 3/4 6-27 Updated Thru. Arne nd . 78 _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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C0!C AI! MENT SYSTEMS CO!CA!!NE!C ATMOSPHERE DI!1" ION SYSTEM LIMITINO CONDI !ON FOR CPEF.ATION 3.6. The containment atmosphere dilution (CAD) system shall be OPEPABLI with:
' An CPEFABLE flow path capable of supplying nitrogen to the dryvell, and A minimum supply of 4350 gallons of liquid nitroge _
APP LICA3 ILITY: CONDITION 1*. ACT ION:
'. ith the CAD system inoperable, restore the CAD system to OPEFABLE status within 31 days or be in at least STAP.T"P within the next B hours. The '
provisicas of Specification 3 0.4 are not applicabl , SURVEIL 1ANCE F.ECU1REMI!CS 4.6. The CAD syste, shall be demonstrated to be OPEPABLE: At least once per 31 days by verif ying that: The system contains a minima of .4350 gallons of liquid nitrogen, and Each valve (manual, power-operated, or automatic) in the flow path not locked, sealed, or otherwise secured in position, is in _ its correct positio At least once per 18 months by: Cycling each power-operated (excluding aut:matic) valve in the ficw path through at least oce complete cycle.cf full travel, l and Verifying that each automatic valve in the flow path actuates to its correct position on a Group 2 and 5 solation test signa '
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CONTAIhWEtU SYSTEMS OXYCEN CONCENTRATI0M L1HITING CORDITION TOR OPEPATION 3.6.6.3* We primary containment atmosphere oxygen concentration shall be less than 4% by volume during the period f rom Within 24 hours af ter TilERMAL POVER > 15E of RATED THERMAL POVER, to Within 24 hours prior to a scheduled reduction of THERMAL POVER to ( 15% of RATED THERMAL POWE APPLICABILITY: CONDITION ACTION: With the oxygen concentration in the primary containment exceeding the limit, te in at least START-UP within B hour . SURVEII. LANCE REOUIREMENTS 4.6. The oxygen concentration in the primary containment shall be verif ted to be within the limit within 24 hours af ter THERMAL POWER > 15% of RATED THERMAL POWER and at least once per 7 days thereaf te *For the period commencing at 0630 on June 29, 1981, a temporary exemption is allowed to operate BSEP-2 in Condition I with containment oxygen concentration exceeding 4% by volume for 72 hour [ f G V7"d3
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BRUNSWICK - UNIT 2 3/4 6-29
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RETYPED TECH. SPLC Updated Thru. Amend. 78
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CAS ANALYZER SYSTEH$ LIMITING CONDITION FOR OPERATION 3.6.6.4 Two independent gas analyzer systems for the drywell and suppression
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chamber shall be OPERABLE with each system consisting of an oxygen analyzer and a hydrogen analyze * APPLICABILITY: OPERATIONAL CONDITION ACTION With one oxygen and/or one hydrogen analyzer inoperable, restore at least two oxygen and two hydrogen analyzers to OPERABLE status within 31 days or be-in at least STARTUP within the next 8 hours. The , provisions of Specification 3.0.4 are not applicabl , With no gas analyzer OPERABLE for oxygen and/or hydrogen, be in at least STARTUP within 8 hour ' SURVEILLANCE REQUIREMENTS 4.6. Each gas analyzer system shall be demonstrated OPERABLE at least once . per 92 days by performing a CHANNEL CALIBRATION using standard gas samples containing a nominalt Zero volume percent hydrogen, balance nitroge Seven to ten volume percent hydrogen, balance nitroge Twenty-five to thirty volume percent hydrogen, bal ance nitrogen.
7 Zero volume percent oxygen, balance nitroge Seven to ten percent oxygen, balance nitroge ,
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BRUNSWICK - UNIT 2 3/4 6-30 Amendment No. 160 _ . _ , . . _ _ _ . _ . . _ , . . _ . . _ . . - _ _ - _ , _ _ . _ .. . a _ u_-.___;-__ J. _[
* . i SPECIAL TEST EXCEPTIONS 3/4.1 PRIMARY CONTAINKERT INTEGRITY '
LIMIT!NG CONDITION FOR OPERATION 3.10 1 The provisions of Specification 3 6.1 1, Specification 3.613, and Table 1.2 may be suspended to permit the reactor pressure vessel closure head t o be removed and (LJ air lock doors to be open during low power PHYSICS TESTS with the THEPJiAL POWER < 5% of RATED THEPJiAL POVER and reactor coolant temperature < 212' APPLICABILITY: CONDITION 2, during low power PHYSICS TEST ACTION: With THERMAL POVER exceeding 5% of RATED THEPJ4AL POVER or the reactor coolant temperature > 212*F, immediately deenergire the scram solenoid valve . l SURVEILLANCE REQUIREMEllTS 4 10 1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once per hour.
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SPECIAL TEST EXCEPTIONS 3/4.10.3 SHUT DOWN MARGIN DEMONSTRATIONS LIMITING CONDITION FOR OPERATION 3.10.3 The requirements of Specifications 3.9.1, and 3.9.3, and Table 1.2 may be suspended to permit the reactor mode switch to be locked in the Start-up position and to allow two control rods to be withdrawn f or shutdown margin demonstrations provided at least the f ollowing requirements are satisfied: The source range monitors are OPERABLE with the RPS circuitry shorting links removed per Specification 3.9.2, The rod worth minimizer is OPERABLE per Specification 3.1.4.1 and is progracmed for the shutdown margin demonstration, and The " notch-override" control shall not be used during movement of the control rod AP P LI CAB ILITY : 0)NDITION 5, during shutdown margin demonstration ACT ION: , With the requirements of the above specification not satisfied, immediately restore the reactor mode switch to the Ref uel positio SU RV E I LLANCE R EC/J IREMENT S 4.1 Within 2 hours prior to the perf ormance of a shutdown margin demonstration verif y that: The source range monitors are OPERABLE per Specification 3.9.2, and The rod worth minimizer is OPERABLI with the required program, per Sp e c if icat ion 3.1. F RU NSW ICK - U NIT 2 3/4 10-3 RETYPED TECH. SPEC Updated Thru. Amend. 71
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l SPECIAL TEST EXCEPTION RECIRCULATION Ih0PS LIMITING CONDITION FOR OPERATION 3.10.4 The requirement of Specification 3.4.1.1 that two recirculation loops te in operation may te suspended for up to 24 hours during the performance of start-up and PRYSICS TEST APPLICABILITY: CONDITIONS I and ACTION: With the above specified time limit exceeded, decnergize the scram solenoid - valve SURVEILLANCF. PIOUIREMENTS 4.1 The time during which the above specified requirement has been . suspended shall be verified to be less than 24 hours at least once per hour during start-up and PHYSICS TEST BRUNSWICK - UNIT 2 3/4 10-4 RETf?ED TECH. SPF.C Updated Thru. Amend. 78 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - - _ - _
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I SPECI AL TEST EXCE PTIONS 3/4 10.5 PLANT SERVICE WATER LIMITING CONDITION FOR OPERATION 3.1 The service water conventional header required to be operating pe r Specification 3.7.1.2 ACTION b.3 may be removed from operation by stopping the pumps to permit isolating and draining the service water nuclear header for maintenance provided thatt The service water conventional header remains lined up to supply cooling water to the required ECCS load The draining / maintenance on the service water nuclear header _ will not affect the service water conventional system or lineuo described in a. abov Average coolant temperature is ~(100'T and the heatup rate is f 10'T per hou Two dedicated, qualified members of the unit operational staff are assigned to initiate the service water conventional heade'r pumps should any of the following occur: Ar.y event occurs which requires ECCS actuatio . Primary coolant temperature exceeds liO' . A loss of offsite power occur APPLICABILITY: CONDITIONS 4 and 5 with the nuclear header inoperabl ~~~ ACTION: With the requirements of the above specification not satisfied, as soon as practicable, restore the: - Service water conventional header to operating status per the requirements of Specification 3.7.1.2 ACTION b.3, or Service water nuclear header to OPERABLE status per Specification 3. 7. SURVEILLANCE REQUIREMENTS 4.1 When the service water conventional header is not operating as specified above: Prio r to securing all service wat e r pumps . verify that the service water conventional header is lined up to supply cooling water for ECCS by verif ying that each valve servicing safety-related equipment that is not locked in the proper position is administratively controlled in the proper positio BRUNSWICK - UNIT 2 3/4 10-5 RETYPED TECH. SPEC ' Updated Thru. Amend. 78
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SURVEILLANCE REOUIREMENTS l Every four hours , verif y that the primary coolant teoperature is less than or equal to 140' Prior to securine the service writ e r rue.r s and at least once per elrht hour s , ve ri f y two-way co m uniestions between the Control Room and the Service Vater Buildin . _ BRUN%'ICK - UNIT 3/4 10-6 RETYPED TECli. SPECS . Updated Thru. Amend. 78
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