IR 05000220/1987016
| ML20236B595 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/15/1987 |
| From: | Florek D, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20236B586 | List: |
| References | |
| 50-220-87-16OL, NUDOCS 8710260253 | |
| Download: ML20236B595 (54) | |
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U.S.~' NUCLEAR' REGULATORY COM' MISSION REGION I OPERATOR-LICENSING EXAMINATION REPORT' EXAMINATION REPORT N0.: 87-16(0L) ! l FACILITY DOCKET NO.:
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50-220 FACILITY LICENSE NO.: DPR-63
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LICENSEE: . Niagara Mohawk Power Corporation
< ' FACILITY: 'Nine Mile Point Nuclear Station, Unit 1 EXAMINATION DATES: August 4-6, 1987 CHIEF EXAMINER: 6 -
W- jd -/M f 7_ D, Lange, be(pior OperatRons Engineer Date APPROVED BY:
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/ /5b7 D.>Florek, Acting Chief, BWR Section / Da/te !
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Operations Branch Division of Reactor Safety .!
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J SUMMARY:~ Written examinations and operating tests were administered to three- ; senior reactor operator (SRO) candidate The candidates passed both the !
' written-and operating examination l
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4 DETAILS TYPE OF EXAMINATIONS: Replacement EXAMINATION RESULTS:
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l SRO l l Pass / Fail l l l 1 I i l Written l 3/0 l l 1 l l l l l Operating l 3/0 l 1 _l 1 i l l l Overall l 3/0 l 1 l l 1 I I CHIEF EXAMINER AT SITE: D. Lange, Senior Operations Engineer OTHER EXAMINERS: A. Howe, Operations Engineer B. Hajek, USNRC Consultant ; The following is a summary of generic strengths or deficiencies noted on operating tests. This information is being provided to aid the licensee in upgrading license and requalification training programs. No licensee response is require STRENGTHS a' . Teamwork and shift communications Application of Technical Specifications and administrative require-ments Application of the Emergency Plan Control panel manipulations I
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l DEFICIENCIES No generic deficiencies were note . The following is a summary of generic strengths or deficiencies noted from the grading of written examinations. This information is being provided ', to aid the licensee in upgrading license and requalification training program No licensee response is require l STRENGTHS Knowledge of reactivity effects, xenon effects on rod worth, and pump net positive suction head, Knowledge of the feedwater control system, shutdown cooling, and diesel generator i Emergency Operating Procedure entry conditions, dropped fuel assembly procedure', stuck open electromatic relief valve procedur DEFICIENCIES Understanding of conditions affecting MAPLHGR. Knowledge of adequate core cooling condition Knowledge of turbine vacuum trips and APRM scram conditions while in ;
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the RUN mod Knowledge of refueling procedures, reactor water cleanup operations, and an understanding of Level Restoration Time in the Emergency Operating procedures, Knowledge of the requirements associated with Senior Reactor Operator presence in the control room and with who may relieve the Station Shift Supervisor of Emergency Director dutie , Simulation Facility Fidelity Repor During the conduct of the simulator portion of these operating tests, the following performance and/or human factors discrepancies were observed: The drywell leakage instrumentation did not properly display the increased drywell leakage during one scenario, Drywell pressure spuriously increased during one scenario.
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., p 3 ; Personnel Present at Exit Interview: NRC i'ersonnel D. Lange, Chief Examiner A. Howe, Examiner , N. Perry, Reactor Engineer -l i Facility Personnel T. Perkins, General Superintendent R. Zollitsch, Training Superintendent M. Dooley, Training Supervisor Unit 2 R. Morey, Training Unit 1 T. Roman, Station Superintendent Unit 1 Summary of NRC comments made at exit interview: 1 The Chief Examiner stated that examination results would be available in approximately 30 days. He noted that the examiners experienced no delays in site access and that the plant appeared to be very clea The generic strengths and weaknesses noted during the operating tests were discusse The simulator performance was discusse The policies concerning NRC requalification examinations were discusse The Chief Examiner also discussed the examination results concerning the Nine Mile Point Unit 2 requalification examinations administered by the NRC in July, 198 i 8, ' Implementation of Emergency Operating Procedures (EOP's) During the simulator scenarios, the examiners evaluated the candidates' ability to satisfactorily implement the Emergency Operating Procedures i
. (E0P's). During emergency evolutions they were familiar with their individual and team responsibilities and were able to execute the E0P's with the minimum shift staff identified in the facility Technical Specif-ications. The candidates did not physically interfere with each other nor did they duplicate efforts (unless required). They were able to transfer from one E0P to another and to enter and exit as required while assuring all necessary precautions were met and steps were complete Attachments: Written Examination and Answer Key (SRO) Facility Comments on Written Examinations after Facility Review q NRC Response to Facility Comments I
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, .. '- U. S. NUCLEAR REGULATORY C0MMISSION 'I SENIOR REACTOR OPERATOR LICENSE' EXAMINATION 4-
,t-FACILITY: NINE MILE POINT ; REACTOR TYPE: BWR-GE2 DATE ADMINISTERED: f7/08/04 ,
.s EXAMINER: HAJEK, CANDIDATE: /)I/Mb7d'/2 '
INSTRUCTIONS TO CANDIDATE: Use separate- paper for the answers. Write answers on one side onl ; Points for each
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Staple question sheet on top of the answer sheet question are' indicated in parentheses after the questio The passing grade requires at least 70% in each category and a final g.ade of at least 80%. Examination papers will be picked up six (6) hours after ; i the examination starts.
f % OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE_ TOTAL SCORE VALUE CATEGORY I 25.00 25.00 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND ! THERMODYNAMICS 25.00 25.00 PLANT SYSTEMS DESIGN,-CONTROL, AND INSTRUMENTATION 1 1 .00 25.00 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL 3
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* CONTROL 25.00 25.00 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 100.00 % Totals Final Grade l
All work done on this examination is my ow I have neither given ] nor received aid, i
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Candidate's Signature .
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS [*'
.. During the administration of-this examination the following rules apply: Cheating on the examination means an automatje denial of your application and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leave. 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 on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee ' Consecutively number each answer sheet, write "End of Category __," as appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe ]
11. Separate answer sheets from pad and place finished answer sheets face down on your: desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE l QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ack questions of the examiner onl . You must sign the statement on the cover sheet that indicatee that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
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..4- - : ( l 18. When you co2plets your oxt:1 nation, you shall: l l =** Assemble your examination as follows: ']
i (1) Exam questions on to I (2) Exam aids - figures, tables, et j l
(3) Answer pages including figures which are part of the answe i Turn in your copy of the examination and all pages used to answer J the examination question .I 1 Turn in all scrap paper and the balance of the paper that you did ]
not use for' answering the question { l Leave the examination area, as defined by the examiner. If after I leaving, you are found in this area while the examination is still )
.in progress, your license may be denied or revoke .q
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,. ', , THERMODYNAMICS . js QUESTION 5.01 (2,00)
Use the attached Figure 27-4, Steam Flow versus Reactor Pressure. to answer this question concerning the
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I operation of the Mechanical-Hydraulic Control System for the Main Turbin , Explain why the turbine chest pressure and the reactor vessel pressure change as reactor power is increased from startup to 100 percent while the . pressure setpoint is maintained constant at 950 psi (1.0)
! Why is the line labeled TURBINE CHEST PRESSURE linear while the line labeled REACTOR PRESSURE non-linear? (1.0)
QUESTION _ 5.02 (1.50) N1-SOP-2, Unexplained Reactor Power Change, lists eight items you should check if an unexplained increase or decrease in reactor power has occurred. If an unexplained power: INCREASE has occurred, state the initial change you'would expect in the following Consider each parameter independentl ' parameter Reactor Pressure (0,5) Feedwater Flow (0,5) Turbine-Generator Load (0.5) ,
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I i l QUESTION 5.03 (1.50) 1 l The Emergency Power Reduction section of N1-OP-43, ) Startup and Shutdown Procedure, instructs you to insert i control rods identified in the Reactor Analyst l I Instructions and cautions you not to withdraw these control rods once you have inserted the Name f two adverse core power distribution effects that could l occur if you do not follow this cautio (1.5) > l
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QUESTION 5.04 (3.00) g' For each of the following, state whether the MAPLHOR will. increase, decrease, or remain the same, AND briefly state WH s
! Near the beginning of life, cracks develop in the l fuel, and the fuel comes into contact with the j claddin (0.75) ' As fuel exposure increases, fission gases form and increase the gas pressure on the clad wal (0.75) As fuel exposure increases, the fission gases mix i with the helium fill gas and cause a change in the l heat transfer coefficient of the gases in the fuel
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pi (0.75) i As the fuel ages, deposits form on the exterior surface of the fuel pin (0.75) .
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QUESTION 5.05 (2.00) State whether each of the followjng statements concerning suberitical multiplication is TRUE or FALS As'Keff approaches unity, a larger change in neutron level occurs for a given change in Kef (0.5) If ' source neutrons are present in a reactor that is just critical (Keff = 1), count rate will increase
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at an exponential rat (0.5) i I When the count rate has doubled after a single j notch withdrawal of a control rod, the reactor has I moved half way to critica (0.5) I As Keff approaches unity, it takes longer for neutron level to reach equilibrium for a given I change in Kef (0.5) l
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In N1-OP-43, Startup and Shutdown Procedure, a Caution statement in the Shutdown section, provides two methods
'to prevent thermal stratification, depending on conditions jn the vessel. What are these two methods? (2.0)
! QUESTION. 5.07 (3.00) Indicate whether the following will INCREASE or DECREASE reactivity during operation, and briefly EXPLAIN wh Moderator temperature increases during a reactor startu (0.75) Fuel-temperature increase (0.75) A feedwater heater is los (0.75) One reactor recirculation system pump trip (0.75) i QUESTION 5.08 .(3.00) For a startup immediately after a shutdown or reactor scram under conditions of peak xenon, control rod notch worths can change significantly from their values in the absence of xeno EXPLAIN WHY and WHERE in the core
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would be. expected, AND EXPLAIN HOW the neutron flux and xenon concentration interact to cause this to occu (3.0)
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QUESTION 5.09 (2.00) For each of the following changes in plant status, state whether the Net Positive Suction Head (NPSH) to'the reactor recirculation pumps will INCREASE, DECREASE, or REMAIN THE SAME, and briefly EXPLAIN wh Vessel water level increase (1.0)
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QUESTION 5.10 .(1.50) l What three conditions, existing individually or collectively, assure' adequate core cooling? ]
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QUESTION 5.11 (2.50) 1 i r. , What are two reasons for using burnable poisons in I the reactor core? (1,0) b .- Describe the change in reactivity that occurs from beginning of life to the end of life as a result of the combination of burnable poison and fuel burnup (1.5) QUESTION 5.12 (1.00) Indicate (YES or h0) whether the Flow Blasing Correction Factor is used to protect against each of the following events? Loss of a feedwater heater while the reactor is operating at 100 percent powe (0.25) Turbine trip without bypass at 65 percent powe (0.25) Feedwater flow controller failure calling-for maximum flow at 40 percent powe (0.25) I Recirculation System flow controller calling for increasing flow at full powe (0.25) l I l l I
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g.- l QUESTION _6 '.01 ( 3'. 00 ) The plant is. operating ~at 100. percent power. State how EACH.of the fo]Iowing components or functions will respond to a complete loss of Instrument Air, AND what-operator action will be necessary to maintain or correct i the status of the system if Instrument' Air cannot be j
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restored to the componen l l Feedwat3r flow contro (0.75) Recirculation System flow contro (0.75) Emergency condenser shell side makeup valve (0.75) RBCLC' heat exchanger flo (0.75) .] ' QUESTION 6.02 (3,00)' With regard to the Shutdown Cooling Syste What are the three interlocks,. including setpoints, that must be satisfied for a pump start to occur? (2.5) What are the three interlocks. including setpoints, that will either-shut the system. isolation valves or prevent them from opening? (1.5) QUESTION 6.03- -(2.00) Radiation monitors will Initiate an Off Gas System j isolatio i What is the purpose of the holdup time prior to monitoring of the gases? (0.5) l What are the trips associated with these monitors, and what signals will result in an Off Gas isolation? State all logic combination (1,5)
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QUESTION 6.04 (2.00) (~ The Feedwater Control System uses density compensated )
. steam flow, feed flow, and. level signal Consider the 1 reactor to be operating at 100 percent with the Feedwater Control System operating in three element Contro i bdflowistemperaturecompensated. How and why V kg / I'JI -
will the reactor level INITIALLY change lhs bin j ignal falls to zero? (1.0) i Steam flow is prescure compensated. How and why 4 will the reactor level INITIALLY change if this ! signal fails to zero? (1.0) J l J
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QUESTION 6.05 (3.00) Two vacuum trips provide turbine protectio For EACH trip, list the automatic trip setpoints AND the automatic valve closures that occur ar. a result of the tri (3.0) QUESTION 6.06 (2.50) Concerning the Diesel Generators, If electrical power is lost to the power boards supplying power to the Air Start System air l compressors (PB1618 and PB171B), will the diesels start on receipt of a manual start signal? Explain your answe (1.0) i What is the purpose of the output breaker interlock, how is the interlock bypassed, and why is this bypass function necessary? (1.5) 0, W to . 54dt t#w> & tu d a .wh La<1qq A A Fujuc. p - h kd>l4 wpahA, hub Ah. U uAcd A hmi to m fd 61 Rp a loe y G wwc.G s,w ,A 1 pn:,thd a >wuwd y\rd ep d (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****) _ _ _ _ _
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: QUESTION 16.07 '(3.00)
Regarding the Core Spray System, Describe the jnitiating signals and logic necessary L for starting the Core Spray pumps and causing injection into the vesse (2.0)- What will be'the Core Spray System response if an initiation signal occurs and then clears 10 seconds after it is received? Be sure to discuss pump start sequencing in your answer. Assume th reactor ~is operating at 60 percent powe (1.0) I QUESTION 6.08 (2.50) i
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If a HPCI initlation is received at a time when a loss of power has occured at both the normal and reserve 115
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1 Which_ pumps will start? -(1.5)' l
i c .~ Will HPCI be available under these conditions.if j one of the normal HPCI pumps is locked out for i maintenance? Explai (0.5) QUESTION 6.09 (1.00) If a loss of power to the Motor Driven Fire Pump has occurred, what provisions exist for the Diesel Driven Fire Pump to start if (Consider separately.) ; Air pressure in the air start system is ' decreasin Briefly explai (0.5) b.- DC control power to the Diesel Driven Fire Pump has ' been los Briefly explai (0,5) l l
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-QUESTION 6.10 (2.00)'
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! .With the Mode Switch in RUN. What are all the APRM Scram LTrip points? Include all appropriate setpoints or condition (2.0) _ QUESTION 6.11 (1.00) The. Reactor Protection System has a bypass for the Scram Discharge Volume high level scram function. When is this bypass used, and why is it necessary? (1.0)
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7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 20
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MA QUESTION 7.01 (1.00) Y What is the entry condition to N1-SOP-3, Alternate Control Rod Insertion? (1.0)
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QUESTION 7.02 (2.00) During emergency circumstances, work may be performed without first processing an RW According to S-RP-2, Radiation Work Permit Procedure, Who by title may modify the normal RWP processing?
(Two required.) (0.5) What special monitoring is required? (0.5) Following work done in this manner, what reporting and evaluations are required? (1.0)
QUESTION 7.03 (3.00) According to N1-SOP-4, Turbine Trip, if a turbine trip occurs, Explain what condition must exist for a reactor scram to occur als (1.0) Whe< v!!! be the rost-turbine-trip conditions of the 345 kv breakers (R915 and R925), MOD 18, and Power Boards 11 and 127 (1.0) What five systems can be used to control reactor pressure? (1.0)
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, . -QUESTION 7.04 (2.00)
According to N1-0P-34 Refueling Procedure, during refuelJng operations, Why shall the fuel pool key lock switch on "G" _ l panel be placed in the REFUEL position? (0 5) Who by title is responsible for directing all fuel ) movements? (0.5)- 1 Where must the Access Platform be positioned during all fuel movements? (0.5) j What is the minimum depth of water that must cover ) fuel when it is being moved? (0.5) i QUESTION 7.05 (3.00) l According to N1-SOP-5, 115 KV Power Failure, if 115 KV ; power is lost with the reactor operating at 100 percent, l What three 4160 KV busses WILL normal power be lost to? (0.6) i 1 Will PB 11 and PB 12 lose power? Why or why not? (0.4) What loads are NOT shed when power is lost to Power Boards 16B and 1787 (0,5)
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QUESTION 7.06 (2.50) l According to S-RP-1, Access and Radiological Control, What is the Radiation Limitation for personnel ! designated as Self Monitors? (0,5) What are four of the five conditions that will ! negate use of an Extended RWP by a Self Monitor? (2.0) l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) j l i
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QUESTION 7.07 (2.00) As SRO, give four of the actions that you would perform or direct as required by N1-SOP-12, Dropped Fuel Assembly, if an irradiated fuel assembly has been droppe (2,0)
-QUESTION 7.08 (2.00)
According to N1-OP-3, Reactor Clean-Up System what are the reasons for the precautions: Insure that both valves from the Clean-Up System to Radwaste and to the Condenser are not open at the same time 7 (1.0)
' During reactor startups, shutdowns, scrams, power changes, or rod swaps, maximize the Clean-Up System flow by putting the second pump in servic (Two reasons.) (1.0)
QUESTION 7.09 (3.00) l For each of the following conditions, indicate whether or not EMERGENCY OPERATING PROCEDURE entry is require If entry is required, state which procedure (s) to ente If entry is not required, state "None." Consider each sub part as a separate item. Assume no additional abnormal conditions are present for each individual ite RPV level is 55 inche (0.25) Reactor power is 12 percent, Startup mod (0.25) Reactor power is 93 percent seven minutes after a load reject, e h n /, M5 % % p g. s g (0.25) Power- per:ti-:-a., ^;; <, _ ::!: tie" -"-"r (0.25) Torus water level is 10.5 fee (0.25) Drywell pressure is 3.6 psi (0.25) Reactor Building Ventilation Exhaust 7.5 MR/H (0.25) Torus water temperature is 85 degrees (0.25) Reactor shutdown, RPV pressure is 1090 psi (0.25) Drywell average temperature is 160 degrees (0.25) CRD Module Area Radiation Level is 100 MR/H (0.25) Suppression Pool level is 11.7 fee (0.25)
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,. .. * . -QUESTION 7.10 (2.00)'
According to N1-OP-1, Nuclear Steam Supply System, Section G.5, Stuck Open ERV,_after receiving indications of that an ERV is open,
~ Where must the control switch be positioned while depressing the reset button to attempt valve closure? (0.5) ! If the computer fails to confirm or identify the-open valve, what action is required to obtain confirmation? (0,5) If depressing the reset buttons fails to close the ERV, why may pulling the fuses on the back of F panel serve to close the ERV7 (0.5) If all attempts to close the valve fall, and the reactor is shut down and depressurized, at what pressure should the valve close? (0,5)
QUESTION 7.11- (2.50) Figure N1-EOP-7.1 from RPV Flooding shows the Level Restoration Time as a function of time after reactor shutdow What is the meaning of " Level Restoration Time"? (0.5) Is the acceptable operating region above or below the curve? (0.5) What action is required if the time limit for the Level Restoration Time has expired? (0.5) What is the basis for establishment of the times specified in the curve? (1.0)
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QUESTION 8.01 (3.00) l
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* SPECIFICATIONS TO ANSWER THE FOLLOWING QUESTIO * FULLY REFERENCE ALL SECTIONS YOU USE, *
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AND STATE HOW THOSE SECTIONS APPL ' The reactor is operating at 100 percent power. APRM Channel 12 has been bypassed because it has been declared to be inoperative due to its two inputs in the LPRM string at core position 41,12 being bypassed. No other APRMs are either inoperable or bypasse An I & C technician has come to you and requested permission to perform the weekly Instrument Channel Test on APRM Channel 16. The. test will be overdue unless performed on this shif Can you permit the APRM Channel to be bypassed and the test to be performed? Using the Tech Specs, fully explain why or why not, and the consequences of not bypassing the channel for the test, or of not performing the tes (3.0) QUESTION 8.02 (3.00) l According to the Technical Specifications, Under what three conditions is a licensed Senior Reactor Operator required in the Control Room? (1.0) What is the minimum shift crew compositio., during normal full power operations? (1,0) What exceptions are allowed to the minimum shift crew composition? (1.0) l i l l l l l
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r- . . . QUESTION- 8.03 (2.50) D According to AP-2.0, Production and' Control of Procedures, answer whether.each of the following statements is either TRUE or FALSE. If the statement is i FALSE, provide a correct statemen ;
. A temporary change must be approved by either a department supervisor, or by a member of the site management staff holding a Senior Reactor Operator's licens (0.5) Each approved change shall be initialed and dated prior to use of the procedur (0.5) Within two weeks'of' implementation, the copy of the marked up procedure incorporating the changes, attached to the TCN, shall be delivered to the Station Superintendent for signature and assignment ,
of independent revie (0,5) ! QUESTION 8.04 (2.50)
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With the plant operating at steady state conditions, 100
: percent power, the I&C Supervisor informs you that part of the circuitry which sends an isolation signal to the Drywell Equipment Drain Line outboard isolation valve on l receipt of a reactor low-low level signal has been found j to be inoperable during the monthly instrument channel ]
test on the Low-Low' Reactor Water Level Channe It i will take three days to procure and replace the ] defective component IN THIS SITUATION, WHAT ACTION (S) ARE YOU REQUIRED TO TAKE IN ACCORDANCE WITH THE TECHNICAL SPECIFICATIONS? (2.6)
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QUESTION 8.05 (3.00) From the events listed below, identify which require either IMMEDIATE or ONE HOUR notification, or NEITHER, per the reporting requirements of either 20CFR20 or 10CFR5 , Two workers performing Reactor Water Cleanup System ~ backwashing in a field which had measured 300 mR/hr actua]Iy received 2 rem whole body after 45 minutes wor (0,5) Unidentified leakage is measured to exceed 10 gpm and total leakage is measured to be about 20 gp (0,5) An instrument technician, working in the Core Spray System Loop A logic cabinet, causes the inadvertent initiation of the A loo (0.5) The reactor has been operating for seven days in an LCO because of the Joss of one diesel generato Because the diesel cannot be returned to service, an orderly shutdown has been starte (0.5) A fire has been discovered on the 281' level of the reactor building, resulting in the loss of the Shutdown Cooling System, and declaration of an Alert in accordance with the Emergency Pla (0,5) The Reactor Protection System automatically initiates following a load rejec (0.5) QUESTION 8.06 (1.50) According to AP-3.3.2, Control of Equipment Temporary ! Modifications, what are the three purposes for which the l Station Shift Supervisor may give permission to perform l a temporary modification in cases of emergency? (1.5) 1 l l l
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- l *- *.
,- ,
, QUESTION' 8.07- (3.00)
*
* NOTE: USE THE ATTACHED' SECTIONS OF THE TECHNICAL l
*
* SPECIFICATIONS TO ANSWER THE FOLLOWING QUESTIO * FULLY-REFERENCE ALL SECTIONS YOU USE, *
* *
AND STATE HOW THOSE SECTIONS APPL l
!
~
With the reactor operating at 100 percent power, you are {
'
notified that while relay logic was being checked on PB 103, one of the undervoltage relays was found to be inoperative, and that the exact cause is not know j This loss will disable the fast transfer to D/G 203 on a i loss of the 115 kV sourc In accordance with the Technical Specifications, , what actions are required? (1.5) j
' How will loss of D/G 103 affect the operability and testing requirements for the Core Spray System? (1.5) l i
)
QUESTION 8.08 (1.50) According to EAP-1, Activation and Direction of Emergency Plan, what are the three responsibilities that the Emergency Director may not delegate?. (1.5) QUESTION 8.09 .(1.00) Several Emergency Plan procedures, including EAP-1, EAP-2 EPP-1, and EPP-2, state that the Station Shift Supervisor shall assume the duties of the Site Emergency Director until properly relieved. According to EAP-1, Activation and Direction'of Emergency Plan, who may relieve the SSS? (Two required.) (1.0)
. (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
_ - _ _ _ _ _
,. > .'
QUESTION 8.10 (3.00)
*
* NOTE; USE THE ATTACHED SECTIONS OF THE TECHNICAL
*
* SPECIFICATIONS TO ANSWER THE FOLLOWINO QUESTIO * FULLY REFERENCE ALL SECTIONS YOU USE, *
*
*- AND STATE HOW THOSE SECTIONS APPL Nine Mile 1 is at 50 percent power for performance of the weekly control rod exercise tests. You learn from the previous Station Shift Supervisor that control rod 19-30 was found to be immovable in the fully withdrawn position. The testing is scheduled to continue during your shif What actions are you required to take in accordance with the Technical Specifications if control rod 27-38 (See attached core map.) is also found to be immovable in the fully withdrawn position? (2.5) During Startup operations, if a single control rod was found to be immovable with normal Control Rod Drive pressure after being fully withdrawn, could you continue the startup to power operations? (0.5)
QUESTION 8.11 (2.00) According to Section 1.0 Definitions in the Technical Specifications, what are the conditions that define Primary Containment Integrity? (2.0) i i
,
l
,
I
.
l (***** END OF CATEGORY 08 *****)
(... ********* END OF EXAMINATION ***************)
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1 br 10hr 100 hr 1000 hr 6 12 30 4---- minu te s - --+ !
!
TIME AFTER REACTOR SHUTDOWN
. .
Level Restoration Time Limit Figure N1-EOP-7.1 ,
..
N1-EOP-7 -23 3/14/86
.
. _ . . . . a_,. . . . . . . . . . . .. .. ._ .
. . . . . . . .
l t i
JTHERW DM8AMICS '
.}y o j,
:\-'
ANSWERS -- NINE MILE POINT, -87/08/04-HAJEK, ' _p 1 ANSWER L5.01 '(2.00) i i As power increases, steam flow increases-[0.5] and causes a pressure drop to occur.-[0.5]_
... g The. TURBINE CHEST PRESSURE curve is linear since )
the pressure drop across the control. valves '] is directly proportional to flow. [0.5] j
'The REACTOR PRESSURE curve is non-linear because it !
reflects all-the pressure drops caused be the flow-restrictors, isolation valves, and piping. [0.5]
.
REFERENCE K&As 293006K103 2.5, 245000A104 2.8, 245000A105 NMP1.0ps Tech Chapter 27, ppg. 4 - 5, Fig. Objective l {
-245000A104 245000A105 293006K103' ...(KA'S)
l I IdNSWER 5.02 (1.50) Reactor pressure increase [ collapsing voids] Feedwater flow increase (causing more subcooling] Load would have decreased (causing a pressure j increase].
REFERENCE 295014K106'3.9, 295014K204 3.3, 295014K206 3.5,
~ .K&As 295014A106 3.4, 295014A107 4.1, 295014A203 OE BWR Academic Series - Reactor Theory, Chapter Objectives 1.5, A106 295014A107 295014A203 295014K106 295014K204 295014K206 ....(KA'S) ,
l l
!
l
!
l I
;
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3.-.-., 1,
'
THERN0DYNASICS ANSWERS!-- NINE MILE POINT
-
-87'08/04-HAJEK,
/ ANSWER 5.03 (1.50)~ High local power might occur [0.75)
1 Abnormal flux patterns may be experienced. [0.75) ~ REFERENCE: K&As 201003K502 3.3, 201003K504 3.4, 201003G010 OE BWR Academic Series - Reactor Theory, Chapter 5, Objective 1.6, N1-OP-43, Startup and Shutdown Procedure, pg. 1 ! 201003G010 201003K502 201003K504 ...(KA'S) .. ANSWER 5.04 (3.00) Increases (0.25] because the heat transfer coefficients at the points of contact are greate [0.5) ) Decreases [0.25] because the stress on the cladding is greater. [0.5) Decreases [0.25] because the heat transfer coefficients of the fission gases are lower than I
~that of heJium. [0.5) d
{ Decreases (0.25] because crud lowers the heat transfer rate from the surface. [0.5]
REFERENCE GE Heat Transfer & Fluid Flow, 2/85, pg. 9-2 Objectives 4.2, 4.3, 4.4, and K&As 293009K112 2.9/3.5, 293009K113 3.1/3.6, , 293009K116 2.4/2,8 )
'293009K112 293009K113 293009K116 ...(KA'S)
.j
. l
,
<
i
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,.', .), THERMODYNAMICS ANSWERS'--iN!NE MILE POINT -87/08/04-HAJEK, \
. ANSWER 5.05 (2.00) True False , i True True
' REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 3-5 - 1 Objectives 1.1, 1.2, K&As 292003K101 2.9/3.0, 292008K104 3.3/3.4-292003K101 292008K104 ...(KA'S)
ANSWER 5.06 (2.00) Maintain two recirc loop suction and discharge valves fully open [0.5] and at least one recirc pump in service. [0.5] OR i If the reactor is flooded above the MSLs and it is 1
. required that all recirc pumps be out of service,-
assure the Reactor Shutdown Cooling System is in ! service [0.5), and close all recire pump suction and discharge valves. [0.5] , i REFERENCE j K&As 202001K412 3.5, 2020010010 l In N1-OP-43, Startup and Shutdown Procedure, pg. 2 l NMP1 Ops Tech Chapter 4, Objective 1 G010 202001K412 ...(KA'S) l
i
.
1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _
~ ~ ~ ~ ~ .
3 TH3!00%YZAMICS
-
i r .l ANSWERS -- NINE MILE POINT -87/08/04-HAJEK, : ..
;[4 ' -
ANSWE .07 L ~ (3.00) f ~ Decreases [0.25] due ':0 [the increase in neutron leakage from).the moderator temperature coefficient. ~
) [0.5] l i . . Decreases [0.25] due to the Doppler coefficient (and l increased neutron resonance absorption). [0.5] Increases'[0.25] due to [ increased subcooling (or colder water going into the core) and] the moderator j temperature coefficient. [0.5] ' Decreases [0.25] due to [ swelling of the volds and]
the negative void coefficient.-[0.5] . i REFERENCE GE BWR Academic Series - Reactor Theory, ppg. 4 - 8, 9, 16, 24, 34, 35, 3 , Objectives 1, 2,-3, 4, 5, i K&As 292004K102 2.5/2.6, 292004K110 3.2/3.2, ' 292004K105 2.9/2.9, 202001K303 3.9/3.9,
-259001K306 3.1/ !
202001K303 259001K306 292004K102 292004K105- 292004K110
...(KA'S)
l
.
5.08
'
ANSWER (3.00) High notch worths will occur in the periphery of the , core where the xenon Jevel is currentJy lowest and the ) flux is currently higher than in a non-xenon startup
[0.75]. I
'
Low notch worths will occur in the center of the core l where the xenon concentration is currently the highest and the flux is lower [0.75]. This is because xenon production is a function of the reactor neutron flux [0.5]. Therefore, the xenon concentration will be highest where the neutron flux was highest - that is, near the center of the core [0.5]. 1 This will depress the flux in the center, forcing the i high flux region to the exterior of the core [0.5]. ! REFERENCE i GE BWR Academic Series - Reactor Theory, ppg. 6-12.
, Objectives 2.5.2, 2. K&As 292006K107 3.2/3.2, 292006K108 2.8/3.2, }-
..
I
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i i
5 .- THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS, AND PAGE 23
.. , ,
THERMODYNAMICS , ANSWERS -- NINE MILE POINT -87/08/06-HAJEK, B.
.' 292006K110 2.9/2.9, 292006K114 3.1/ K107 292006K108 292006K110 292006K114 ...(KA'S) ANSWER 5.09 , (2.00)
~ NPSH increases. [0,5) As the water level increases, the static head component of the NPSH equation increases. [0.5] NPSH increases. [0.5] When feedwater heating is lost, inlet subcooling to the recire pumps increases (more subcooling), increasing the margin to saturation. [0.5)
REFERENCE K&As 293006K110 2.7/2.8, 291004K106 3.3/ K103 3.2/ GE Heat Transfer & Fluid Flow, 2/85, ppg. 6-76 - 8 Objectives 10.9, 10,10, 202001K103 291004K106 293006K110 ...(KA'S) ANSWER 5.10 (1.50) The active fuel is covered with a 11guld or a two phase mixture. [0.5] Core Spray is cooling each fuel assembly in sufficient quantity to remove all heat generated in the assembly. [0.5] Steam flow is cooling each fuel assembly in sufficient quantity to remcVe all heat generated in the assembly. [0.5] g K&As 295031K302 4.4/4.7, 295031K303 4.1/4.4, 295031K304 4.0/4.3, 295031K101 4.6/ MOCD Text, Chapter 3, pg. 3-1. Nov, 86. Objective K101 295031K302 295031K303 295031K304 ...(KA'S)
.
_ _ _ _ _ _ _ _ _ _ _ _ _
;-__-_ _ _ _ _ _ .
. .,. 5 1 l
< ANSWERS --'NINE MILE POINT .-87/08/04-HAJEK, !
*
. .
!
ANSWER _ 5.11 (2.50) . Flux shaping.[0.5] Extending core life (0,5) _ Initially, as a' result of the gadolinium burnup rate exceeding the fuel burnup rate, core ! reactivity increases. (0.75] Near middle of life, ; the gadolinium has essentially all burned up, and the fuel'burnup rate exceeds the Gd burnup rate, . j thus reducing core reactivity. [0.75]
{ Curve presentation also acceptable ] '
I REFERENCE K&As 292007K101 2.9/3.1, 292007K103 2.4/ GE BWR Academic Series - Reactor Theory, ppg. 6-1 Objectives 4.1, K101 292007K103 ...(KA'S) ANSWER 5.12 (1.00)
' No No Yes
, No j
.I REFERENCE l
'
K&As 293009K127 2.7/ GE Heat Transfer & Fluid Flow, 2/85, pg. 9-3 Objective 5.11 293009K127 ...(KA'S) 1 l l l l
. '
_ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . .
. _ - _ _ _ _ ' * '
ANSWERSL-- NINE MILE POINT -87/08/04-HAJEK, !
.'
ANSWER '6.01 (3.00)' l The FCVs will lock up [0.5] and local manual control will need to be taken. [0.25]-
. The MO speed controller will lock up [0,5) and local manual control will need to be taken..[0.25] The makeup valves will fail open-[60-17 and 18), [0.5] and they will need to be locally and manually throttled to maintain water level. [0.25] The heat exchanger flow will decrease as the TCV [TCV 70-137] fails open. [0,5) Local manual control of the TCV will be required. [0.25}
REFERENCE j K&As 202002A205-3.1, 259001K601 3.0, 259001A205 3.0, ; 207000K604 3.3, 295018A203 j NMP1 Ops Tech Chapter 30, objective ll, SOP- A205 207000K604 259001 A205 - 259001K601 295018A203
...(KA'S)
ANSWER 6.02 (3.00)' . All system isolation valves open [0,5] Suction pressure greater than 4 psig [0.5] Suction temperature'less than 350 F [0.5]' . RPV pressure greater than 120 psig [0.5] ! Vessel isolation at lo-lo (5 ") [0,5) System area temp greater than 120 F [0.5] I REFERENCE ! K&As 205000K401 3.4, 205000K402 3.8, 205000K403 i NMP1 Ops Tech Chapter 20a, ppg. 8 - 9, Objectives 7, K401 205000K402 205000K403 ...(KA'S) i
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~. ANSWERS -- NINE MILE POIN i
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-ANSWER 6.03 (2.00)
a'. To allow for decay,of the short lived isotopes
-[N-16 and 0-19] so background can be measured j
[0.25]fand'so that. fuel leaks can be detecte ,
[0.25] .Downscale, High, High-High [0.2 each] \
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Valve closure results if have two hl-hi signals
[0.45]
or j one hi-hi and one downscale signal. [0.45] REFERENC K&A '272000K102 3.5, 272000K305 3.7, 272000K402 4.1, 272000K403 3.9, 272000A201 i NMP1 Ops Tech Chapter 31b, ppg. 2, 1 Objectives 6, ! 272000A201 272000K102 272000K305 272000K402 272000K403 j
...(KA'S) q - ANSWER 6.04 (2.00) T N he mperature signal fails to zef.p. indicated Lead h Slo M'd5 feed flow w se for&]'aiiBTivel will u dg y) b t p ; nd, .
Initially _dao [m a t nse at -
lower Temperatures.] '\ b N ' N A'l3d c'. ! l ..;.1 b di I (Steam flow is pressure compensated using a reactor.( G - pressure signal). If the pressure signal falls to f,.g'> y,,,h,l . u 1 zero,' indicated steam flow will decrease [0.5] M k 1 * " U l
[since steam is more dense at higher pressures], .
"'"g"" ,
and vessel level will initially rise. [0.5] % ". U ##. .
* s c h ..uu m... I K 259001K607 3.8, 259001A207 3.8, 259002K603 3.1, 259002K604 ", A *
'*"a F* "[$*'
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NMP1 Ops Tech Chapter 23, ppg.7 - 8, Figure 23- MO*H % N "- '" '"' Objectives 13, 1 km AhA-f * % 259001A207 259001K607 259002K603 259002K604 . . . ( KA ' S ) p w.< t M., . *
* %4al .u.n) and0 ' nc
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' __m .-m__ - . _ _ _ . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27
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ANSWERS -- NINE MILE POINT -87/08/04-HAJEX, B.
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ANSWER 6.05 (3.00) Vacuum Trip * 1 High Exhaust Hood Temp 225 F - Low Condenser Vacuum 20" Hg relative to atmosphere (Below atmosphere) Back-up Overspeed Trip at 112 % Rated Shaft Pump Trip 95 psig Thrust Bearing Trip, 30 to 35 mils
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Generator Protective Circuitry High Reactor Water Level 95" High Moisture Separator Drain Tank Level 17" Closes the main stop valves, reheat stop valves, intercept, and control valves Vacuum Trip * 2 Low Condenser Vacuum below 10 " Hg relative to atmosphere Trips and maintains the Bypass valves close .2 for each trip, 0.24 for each valve actio REFERENCE K&As 245000K405 3.0, 245000A203 3.6, 245000G012 NMP1 Ops tech Chapter 27, pg. 1 Objective A203 245000G012 245000K405 ...(KA'S)
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ANSWERS - .NINE MILE POINT -87/08/04-RAJEK, B.
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ANSWER .6,06 (2.50) Yes. [0.5] There are five air receivers that provide sufficient storege for five engine starts-without replenishment. [0.5] . The interlock is provided to prevent having both the DG output breaker and the normal bus source breaker closed at the same time. [0.5] The interlock is bypassed by turning on the synchroscope [0.5] to allow paralleling the DG with j site power (or to close the DG output breaker). [0.5] REFERENCE K&As 264000K101.4.1, 264000K106 3.2, 264000K408 3.7, i 264000K505 3.4, 264000K601 NMP1 Ops Tech Chapter 34, ppg. 4, 16. Objectives 5, 7, K101 264000K106 264000K408 264000K505 264000K601
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E ANSWER 6.07 (3.00) H h a . -- .An initiation (pump start) will occur on receipt of
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EITHE a Rx lo-lo level signal [0.4]-
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OR [0.2] ra high drywell pressure signal [0.4].
! The valves will open when reactor pressure falls ,
l below 365 psig. [0.5] .
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The signal must be received in both RPS channels 1 (11 and 12].for the pumps to start and the valves .l to actually operate. [0.5] [This could be a 1o-1o I water level in one channel and a high D/W in the -l other,or it could be the same signal in both ) channels.] l l The pumps sequence on at ' seconds CS pumps 111 and 112 [0.1] 1 7 seconds Topping pumps 111 and 112 [0,1] 13 seconds CS pumps 121 and 122 [0.1] 20 seconds Topping pumps 121 and 122 [0,1] 1 Since the requence continues to the end once it has l started, ell the pumps will start even though the l signal hLs cleared. [0.4] The valves will not open .l s ir.ca reactor pressure will not be dropping. [0.2] REFERENC , K&As 209001K203 3.1, 209001K401 3.4, 209001K408 4.0, 1 209001K409 NMP1 Ops Tech chapter 17, ppg. 10 - I Objectives 5, 8, , 209001K203 209001K401 209001K408 209001K409 ...(KA'S) i ANSWER 6.08 (2.50) From Bennett's Bridge [6000 KVA generators through the Lighthouse Hill switching yard]. #13 Condensate, #13 Feedwater Booster, and #12 Feedwater. [0.5 each] Yes, [0.2] The non-preferrred corresponding component will start. [0.3] REFERENCE ,- K&As 206000K104 3.6, 206000K202 3.1, 206000K603 3.1, L 206000A108 ; l.-- ! )
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,: 'i 's NMP1 Ops Tech Chapter 16, pg. 1 Objectives 3, 4, A108 ;206000K104 '206000K202 206000K603 ...(KA'S)
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l ANSWER 6.09 ( 1.,00 ) l i
-) .The pump'will auto start.if air pressure falls to about 60'psig (per OP-21, pg. 6) or to 40 psig (per CAF SD 38a, pg. 7).
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' If DC power is lost, an operator can be dispatched 'i locally.to manually start the pum )
K REFERENCE J' K&As 286000K601'3.1, 286000K602 2.9. 286000A406 3.4, 286000G007 NMP1 Ops Tech Chapter 38a, ppg. 7, 1 Objective 4, 5, N1-0P-21, ppg. 6, 10, 286000A406 286000G007 286000K601 286000K602 =...(KA'S)
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ANSWER 6.10 (2.00) , L Downscale [0-25] at 5 Y of full scale [0.25] with the companion IRM upscale (0.25] or IRM Inop [0.25] Upscale [0.25] per Flow Blas [0.25] Inop [0.25] Less than 4 inputs [0.25]. i Channel switch not in operate (0.25] APRM module unplugged [0.25] REFERENCE ]i K&As 215005K101 4.0, 215005K402 4,2
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NMP1 Ops Tech Chapter 9d, pg. 1 Objective K101 21500SX402 ...(KA'S) l l ANSWER 6.11 (1.00) It Js used following a scram (0.3] to permit resetting of RPS [0.3] so the SDV can be vented and drained. [0.4]
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b REFERENCE K&As 201001K107 3.4, 201001K406 3.9, 212000K412 NMP1 Ops Tech Chapter Sa, pg. 9. Objective 6.
f 201001K406 212000K412 ...(KA'S) 201001K107 l ! I _-_-_ _ _ -
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l 7.01 (1.00) ggbn/ ' (:- ANSWER Any control rod is not inserted to position 157 following a reactor scra ~ jj REFERENCE j K&As 295015G011 N1-SOP-3 i 295035G011 ...(KA'S) 4
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ANSWER 7.02 (2,00)
. LThe Station Shift Supervisor (0.25] or when practical, the Chemistry and Radiation Protection Supervisor [0.25] !
! Continuous monitoring by a Radiation Protection l Technician
; All normally necessary documentation (0.3] I Radiological Incident Report (0.4]
Post ~'ob ALARA evaluation [0.4]. REFERENCE K&As Plant Wide Generics 294001K103 3.8, 294001K104 S-RP-2, Radiation Work Permit Procedure, pg. Objective K103 294001K104 ...(KA'S) i
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4 RADIOLOGICAL'CONYROL !. ANSWERS -- NINE MILE POINT -87/08/04-HAJEK, B.
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ANSWER 7.03 (3.00) ;
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. If reactor power is above 45 percent. [1.0)
~ .345 kv breakers [R915 and R925] tripped'
MOD 18 open PBs 11 and 12 supplied from reserve power
[0.33 each) Main turbine bypass valves .EC ERVs RWCU Main steam line drains [0.2 each)
REFERENCE ) K&As 295005K201 3.9, 295005K208 3.3, 295005K207 N1-SOP-4, Turbine-Trip Objective T. O. I K201 295005K207 295005K208 ...(KA'S)
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ANSWER 7.04 (2.00) To place the fuel pool high rad monitor on the refueling bridge on the emergency ventilation circuit. [0.5) The Chief Shift Operator [0.5] In the " full up" position. [0.5) Eight feet. [0.5] REFERENCE K&As 2340000001 3.8, 234000G010 N1-0P-34, Refueling Procedure, ppg. 4 (Description), 8 - ; 9 (Prerequisites).
NMP1 Ops Tech Chapter 37, Objectives 4, 1 G001 234000G010 ...(KA'S)
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7 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 i i RADIOLOGICAL CONTROL
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i ANSWERS -- NINE MILE POINT -87/08/04-HAJEK, ' . ANSWER 7.05 (3.00) PB 101, PB 102, and PB 103 will be los [0.2 each properly identified) . l l PB 11 and PB12 will be lost if they are being supplied from reserve power. [0.4) The CRD pumps. [0.5) and 171 -OR- emergency lighting [0.5) 162 and 172 -OR- continuous power [0.5) 167 -OR- computer supply [0,5) REFERENCE K&As 295003K102 3.4, 295003K203 3.9, 295003K302 3.1, 295003K303 3.6, 295003A204 3.7, 295003G011 N1-SOP-5, 115 KV Power Failure 295003A204 295003G011 295003K102 295003K203 295003K302 295003K303 ...(KA'S) l ANSWER 7.06 (2.50) These personnel are permitted to monitor themselves in beta-gamma fields up to 2500 mr/hr. [0.5) . Whole body dose rates in excess of 2500 mr/h . The area is posted "No Entry, No Self Monitors, contact Radiation Protection." i Excessive steam or water leakage from contaminated systems OR potential airborne hazar . Work beyond the scope of the extended RW . Entry into areas requiring neutron dosimetr [0,5 for any four of the above.)
REFERENCE K&As Plant Wide Generics 294001K103 3.3/ S-RP-1, Access and Radiological Control, ppg. 21 - 2 LP Objectives 5, 6 K103 ...(KA'S)
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ANSWER- .'7.07 (2.00) Activate the.E Plan, if require l Evacuate the refueling floo ~ i
* Evacuate the Drywel . Sound the station alar . Initiate EV . Check the ARMS for evidence'of gaseous release [0,5 EACH FOR ANY FOUR] )
REFERENCE K&As 295023G010 j N1-SOP-12,. Dropped Fuel Assembly l NMP3 Ops Tech Chapter 37, Objective 1 ~1 2950230010- ...(KA'S) ] l l ANSWER 7.08 '(2.00) Both valves open could result in loss of condenser vacuu i b '. This will help-minimize long term radiation level buildup'in the reactor recirculation {
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, piping by removing insolubles before they plate l out. [0,5) 1
] It will help reduce the magnitude of thermal j-cycling on the feedwater nozzles during the above periods of low feedwater demand. [0,5)
REFERENCE K&As 204000K102 3.0, 204000K120 3.5, 204000G010 N1-0P-3, Reactor Clean-Up System, pg. G010 204000K102 204000K120 ...(KA'S) l
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' PROCEDURES - NORKAL,' ABNORMAL; EMERGENCY'AND PAGE: 35' , .*. .
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'NSWERS'--~NINE' MILE' POINT A -87/08/04-HAJEK, l
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s ANSWER .7,09- (3.00) None
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b.- E0P-3 [ Failure to Scram] , j E0P-3 [ Failure to Scram] j
' E0P-3 .i None E0P-4, E0P-2 -[0.125 each]. E0P-5 (Secondary Cont. Control) E0P-4 [ Primary Cont. Control] E0P-2 [RPV Control] .j E0P-4 j
' E0P-5 ** Name or Number accepted ** l EOP-4 REFERENCE {
NI-EOP-1 - LP Enabling Objective K&As 295024-38, System Generic K/As 11 (4.3/4.5).
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ANSWER 7.'10 (2.00) t In the auto position- Dispatch an operator to the " Valve Monitoring-System" in the Auxiliary Control Roo Because the solenoids are normally energized to ; open_the ERVs, and pulling the fuses will de- ! energize the At 50 psi ! REFERENCE K&As 239002A203 4.2, 239002G014 N1-0P-1, Nuclear Steam Supply System, Section G.5, Stuck l Open ERV, pg. 3 j 239002A203 2390020014 ...(KA'S) l l i I a
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_7; PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND ' ' _j c T j--
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RADIOLOGICAL'CONTRO ANSWERS -- NINE MIL 3 PO?NTL -87/08/04-HAJEK, : ,
ANSWER . 7.1 '1 (2.50) :This is.the time'during which-level indication has
'been' lost [0.5] after termination of' injection has, begun.'[0.5] The. acceptable operating region is below the curv That is, action does not have to be taken until the time exceeds the time shown on the curv . Reflooding has to begin, i
The Maximum Core Uncovery Time Limit defines the
' time required.for the peak cladding temperature to reach 2200 F.with no. spray or steam cooling and the core completely uncovere REFERENCE'
K&As 295032K101.4.7, 295031K201 4.4, 295032K302 4.7, 295031A204 LP Enabling Objectives 2.3, A204 295031K101 295031K201 295031K302 ...(KA'S) j i
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' " ~ -87/08/04-HAJEK, A,NSWERS -- NINE MILE POINT 4 ..
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l" ANSWER ' 8 . -01 ;(3.00)
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Note (e) to Table 3.0.2.a states that only one APRM Channel in a given' core quadrant may be bypassed. [0.5] q Note (c).to Table 3.6.2.g [on pg. 223] is more -j-restrictive in.that it requires that no two LPRMs in an9 l radial location serving one APRM may be bypassed without- j inoping the channel. However, if this is the only-reason for the inop, then Channel #12 may be returned to service for the duration of the test. [0.5] . ,
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i Otherwise, Channel 816 cannot be bypassed, and the test will have to be performed with the RPS. Channel in the d
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tripped condition. [0.5]
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If the test is not performed, Channel #16 will have t be declared INOP, and a half trip will have to be taken anyway. [0.5] Alternatively, it may be possible to use a TIP to replace i one of-the failed LPRMs in Channel #12 to make the k channel operable per both notes. [0.5] ) The alternative is to shut down the reactor per TS ! 3.6.2.a.(1) when the next APRM has its ST come due. [0.5] REFERENC K&As 215005G005 4.2, 215005G010 j Tech Specs Section 3.6.2a and Tabics 3.6.2 a and j I
.LP Objective _V.1, G005 '215005G010 ...(KA'S)
ANSWER 8.02 (3.00) During power operations (0.33], hot shutdown
[0,33], and when the emerEency plan is activated
[0.33]. One SR0 [0.25] I Two R0 [0.25]
Two Unlicensed [0.25] One Aost Station Shift Supervisor serving the STA l function [0.25] 1 The shift crew composition may be one less than the minimum [0.4] not to exceed two hours [0.3].except that this does not apply during shift change to allow for a late arrival. [0.3] l
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REFERENCE K&As~ 294001K116 3.5/ Tech 1 Specs Section 6. K116 ...(KA'S) ,
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ANSWER 8.00 (1.50). False. Must have concurrent approval by the department supervisor concerned AND the plant-management person, either one of whom may hold an SRO li ense,
. True, c '. False. .On the first business day following implementation . . .
REFERENCE K&As 294001A103 2.7/ AP-2.0, Production and Control of Procedures, pg. 1 LP Objective E0-1, 294001A10 ...(KA'S) ANSWER 8.04 (2.50)
Section 3.6.2.a.(2) requires the isolation valves to be ! closed or the valves to be considered inop and l sections 3.2.7 and 3.3.4 to apply. (Only 3.3.4 will apply in this situation.) [0.5] .i Section 3.3.4.b requires that if any isolation valve becomes inop, the system can be considered operable if f at least one valve in each inop line is closed. [0.5] , If this cannot be met, the reactor must be cooled to
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" ' less than 215 F within lo hours. [0.5] Since operation with the DWEDS isolated will not allow leakage rates to be determined as required by Section y 4.2.5.a [0.5], the reactor will have to be taken to cold shutdown within 24 hours per Section 3.2.5 statement at end. [0.5]
' . REFERENCE K&As 223002G005 3.1/4.1, 223002G011 3.4/ Technical Specifications as noted in answe LP Objectives V.1, G005 2230020011 ...(KA'S)
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" ADMINISTRATIVE PROCEDt'RF.S. CONDT7JON4 A' 'A110NS PAGE' 39 I 3.*.,
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ANSWE .05 (3.00).
- Neither. [Imm. after 25 jo 24 1.- : . ;er 5 rem]. Immediate (because cf rec' i? r d det ic ,sion of '
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Unusual Event under Reactor Coolant F" stem Integrity. ;Viele.ucen of Tt ;h Spca
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.ujres one hour notification.) Neither [ 4 hour} l f 'One hour repor ,
I Immediat Neither (four hetu not i f i c t t j er:) i REFERENCE' K&As 294001K103 2.3f3.0. a . 's 01 A110 2 . c '4 . 7, j
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2640000003 4s? 2120UU0003 4.5 a 60010003 Reporting Requirements note Enci g l . Procedure EAP-2, 10CFh LT Objc: t ived Y.* V r. . 2090010003 212000G005 2 0 4 r. . . ' ' . 's -:001A110 '294001K103 j
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i ANSWER 8,06 (1.561 To prevent personnel inje: y [G.5) To prevent equil r..t.nt dar sc [0.5) To increase the trargin of safety (( ,
I REFERENCE ] K&As 294001A103 2.7/3.7, 204001A110 3.0/4.2, ! 294001A116 2.9/ ) AP-3.3.2, Control of Equipment Ten porary { Modifications, pg. I 294001A103 294001A110 294001A110 .(KA'S) l
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N _______1 - ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 40-87/08/06-HAJEX, ' '
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ANSWER 8.07 (3.00) TS 3.6.3.c requires that one DG may be inop provided two 115 kv external lines are available and the DG is returned to service within seven - l days. [0.5) l In the interim, the operable DG must be tested ' immediately [0.5) and once/ week thereafter. [0.5) TS 3.0 states that if a subsystem or component is determined inop solely due to the loss of its emergency power source, it may be considered operable for satisfying the LCO provided 1. Its corresponding normal source is available,
[0.5] and 2. All redundant systems, etc., are operabl [0.5)
Therefore no testing of Core Spray is necessary provided all components passed their most recent surveillance tests per 3/4.1.4. [0.5) REFERENCE K&As 264000G005 3.4/4.1, 264000G011 3.4/ G005 3.3/4.2, 2090010011 3.4/ Technical Specifications as noted in answe LP Objectives V.1, G005 209001G011 264000G005 264000G011 ...(KA'S) ANSWER 8.08 (1.50) Classification of the emergency [0.5) Determining the necessity for a site evacuation (0,5) Authorizing emergency workers to exceed normal radia; ion exposure limits [0.5) REFERENCE K&As 294001A116 2.9/ EAP-1, Activation and Direction of Emergency Plan, p . LP Objective EO-3, 294001A116 ...(KA'S)
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ANSWERS --- NINE MILE POINT
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ANSWER ;8.09 ' (1.00) l General Superintendent-Nuclear Generation (0.5] Acting General Sup.-Nuc. Generation [0.5]- ,
[or designee) ,
l REFERENCE j K&As 294001A116 2.9/ j EAP-1, Activation and Direction of Emergency Plan, p ! I 294001A116 ...(KA'S) ;
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ANSWER 8.10 (3.00) Declaro the rod inop [0.5] and valve the rod out of servic. [0.5) if according to,3.1.1.a(2) if it can i be demonstrated that the conditions of 3.1.1.a(1) l are met. [0.5) Then, per 4.1.1.a.(2), perform a rod exercise test at least once per 24 hours until the number of inop rods can be reduced to less than j two. [0,5) All actions are dependent on the failures not being due to collet housing failure If they were due to collet housing failures, the reactor would have to be shutdown within 48 hours. [0.5) Ye [As long as the conditions listed in Part a are adhered to,] {0.5) REFERENCE K&As 201003G005 3.3/3.9, 202003G011 3.2/ Technical Specifications as noted in answe LP Objectives V.1, G005 201003G011 ...(KA'S) i
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f.* s' ANSWER 8.11- (2.00) The Drywell and absorption chamber are closed.
I and [0.4) l All non-auto PCI valves which are not required to ' I be open for plant operation are closed. [0.4) At least one door in the airlock is closed and I sealed. [0.4) { l I All auto containment isolation valves are operable !. or are secured in the closed _ position. [0.4) All blind flanges and manways are closed. [0.4] 4 I l REFERENCE K&As 223002G006 2.9/3.9 , Tech Specs, pg. l LP Objective V. G006 ...(KA'S) l l l l i
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l QUESTION VALUE REFERENCE QUESTION VALUE REFERENCE { >' :,. l
05.01 2.00 BRH0000385 07.01 1.00 BRH0000374 !
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05.02 1.50 BRH0000386 07.02 2.00 BRH0000375 05.03 1.50 BRH0000387 07.03 3.00 BRH0000376 05.04 3.00 BRH0000388 07.04 2.00 BRH0000377 05.05 2.00 BRH0000389 07.05 3.00 BRH0000378 j 05.06 2.00 BRH0000390 07.06 2.50 BRH0000379 05.07 3.00 BRH0000391 07.07 2.00 BRH0000380 05.08 3.00 BRH0000392 07.08 2.00 BRH0000361 05.09 2.00 BRH0000393 07.09 3.00 BRH0000382 ) 05.10 1.50 BRH0000394 07.10 2.00 BRH0000383 j 05.11 2.50 BRH0000395 07.11 2.50 BRH0000384 { 05.12 1.00 BRH0000396
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06.01 3.00 BRH0000409 08.01 3.00 BRH0000397 j 06.02 3.00 BRH0000410 08.02 3.00 BRH0000398 j 06.03 2.00 BRH0000411 08.03 1.50, BRH0000399 06.04 2.00 BRH0000412 08.04 2.50 BRH0000400 06.05 3.00 BRH0000413 08.05 3.00 BRH0000401 06.06 2.50 BRH0000414 08.06 1.50 BRH0000402 -1-06.07 3.00 BRH0000369 08.07 3.00 BRH0000403 ) 06.08 2.50 BRH0000370 08.08 1.50 BRH0000404 i
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06.0 .00 BRH0000371 08.09 1.00 BRH0000405 06.10 2.00 BRH0000372 08.10 3.00 BRH0000406 06.11 1.00 BRH0000373 08.11 2.00 BRH0000407
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