IR 05000423/1986010
| ML20207B191 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/02/1986 |
| From: | Barber S, Keller R, Kister H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20207B188 | List: |
| References | |
| 50-423-86-10, NUDOCS 8607170451 | |
| Download: ML20207B191 (112) | |
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{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO.: 86-10 (0L) . FACILITY DOCKET NO.: 50-423 FACILITY LICENSE N0.: NPF-49 LICENSEE: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141 FACILITY: Millstone Nuclear Power Station Unit 3 EXAMINATION DATES: Mar to ril 4, 1986 CHIEF EXAMINER: '
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S W Re r Engineer (Examiner) ' DATE REVIEWED BY: _ _ M. Keller, hief
/ [[2 Nd DATE Pro' t Section 1C b '
APPROVED BY: H. B. K s , Chief Projects Branch No. 1 / DATE : SUMMARY: Oral, written and simulator exams were administered to three reactor operator and nine senior reactor operator candidates. All reactor operator candidates passed all portions of their examinations and will be issued licenses. Six senior reactor operator candidates passed all portions of their examination and will be issued licenses. Of the three senior reactor operator candidates that did not pass: 1 failed the written only, 1 failed the simulator and oral only and 1 failed the written, oral and simulato Lack of proper examination security continues to be a proble , l
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REPORT DETAILS TYPE OF EXAMS: Initial EXAM RESULTS: 1 I I I RO l SR0 I l Pass / Fail l Pass / Fail l l I I I I I I IWritten Exam I 3/0 l 7/2 l l l 1 l 1 1 I I I I I I l Oral Exam I 3/0 1 4/2 l l l l l l ' I I I I I I I l Simulator Examl 3/0 1 4/2 l l l l l l l 1 1 1 I I I l0verall l 3/0 l 6/3 l
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1 I I I I l_ l l CHIEF EXAMINER AT SITE: D. Cc 'S. Barber, NRC I OTHER EXAMINERS: J. Hannon, NRC R. Ke'.ler, NRC F. Jaggar, INEL II Summary of generic strengths or deficiencies noted during the operating exam: Candidates were unable to explain the use and construction of area and process Radiation Monitoring Syste Candidates with Millstone 2 experience were unable to properly explain the operation of the Millstone 3 rod control system and the cold leg recirculation flow path.
I Simulator Deficiencies: The A Instrument Air Compressor can not be placed in constant run unless the B compressor is faile The Containment CAR fan monitor (CMS-22) alarmed for no reason during a steam generator tube ruptur Generic weaknesses noted during the grading of written examination RO candidates were unable to adequately explain the following: 1). How core' delta T is affected on a loss of natural circulatio ). The flowpaths for Reactor Coolant Pump #1 seal leakoff during a safety injectio ). The automatic actions that result from high activity due to a fuel drop acciden ). Red path entry conditions for a Response to Loss of Secondary Heat Sink, FR- SRO candidates were unabic to adequately explain the following: 1). The design attributes that cause core uncovery to be more likely for a cold leg break than for a hot leg brea ). The proper use of plant pump data to estimate flowrates at a given pressur ). Conditions that require entry into Inadequate Core Cooling Procedure ). The reasons for requiring a maximum reactor vessel venting time calculatio ). How to determine that Technical Specification leakage limits are being exceede ). The proper interpretation of the operability of the Reactor Coolant System's power operated relief valve V Training / Reference Material: The reference material supplied by the facility did not contain:
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Operating Procedure Index
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Emergency Procedure Foldout Page
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Critical Safety Function Status Trees Simulator Initial Condition List VII. Personnel Present at Exit Interview: E
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NRC Personnel D. Coe, Reactor Engineer (Examiner) J. N. Hannon, Section Leader, Operator Licensing J. T. Shediosky, Senior Resident Inspector J. Grant, Resident Inspector NRC Contractor Personnel F. S. Jagger, Contract Examiner Facility Personnel W. D. Romberg, Station Superintendent J. D. Crockett, MP3 Superintendent H. F. Haynes, Manager, Operator Training R. G. Stotts, MP3 Training Supervisor VII Summary of NRC Comments made at exit interview: The chief examiner reviewed the number and type of examinations administered during the previous week and presented _ generic weaknessestobserved during the simulator and oral examination Examination security was lax. A security officer passed through a barrier to use a restroom reserved for the candidates. This was the third occurrence of this type. In addition, personnel entered and exited the simulator without prior approval. The licensee committed, in a letter dated June 18, 1986, to station a security guard during the written examination and to post conspicuous signs during the simulator examination. These actions should be adequate to ensure examination securit Initially security access to the plant was delayed. However, attention by management improved access and reduced delays in later entries into the plant.
I Examination Review An examination review was conducted after the completion of all operating exam Items from Attachment 3 were discussed on a line item basis. All items were considered during grading but not all items resulted in a change to the master exam. Attachment 4 details the significant changes to the examinations.
Attachments: Written Examination and Answer Key RO Written Examination and Answer Key SR0 Facility Comments on Written Examinations Changes to Written Examination .
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Re.utecoers N4k g*[l U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION b C FACILITY: MILLSTONE _3 Wh REACTOR TYPE: PWR-WEC4 h M bnMc DATE ADMINISTERED: 86/04/01 EXAMINER: JAGGA APPLICANT: k<);A(T m Po^'~
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d's/%/d u n Nat/d s INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after l the examination start .
% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL S_QORE VALUE CATEGORY 25.00 25.00 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, REAT TRANSFER AND FLUID FLOW
, 25.00 25.00 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 25.00 25.00 INSTRUMENTS AND CONTROLS 25.00 25.00 FROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS FINAL GRADE % All work done on this examinaticn is my own. I have neither given nor received ai 'PPLICANT'S SIGNATURE A
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' * * .. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 2 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (1.00) The reactor coolant pumps require about 30% more power to operate in cold shutdown <200 F than at 530 F (hot shutdown) .
Why are the power requirements different for each condition? (0.5) If the VELOCITY of the fluid is the SAME at 200 F and at 530 F, will the MASS FLOW RATE also be constant at these two different conditions? Explai (0.5) QUESTION 1.02 (3.00) After operating in natural circulation for 2 hours, a complete loss of natural circulation flow occurs. How will the following parameters be affected initially (INCREASE, DECREASE, or NO CHANGE )? Briefly explain your answe (assume no further operator action)
a. Core delta T b. Core thermocouple temperature c. Steam generator pressure h RCS-f4e r ' QUESTION 1.03 ( .70) Following a reactor trip from 100% power, how long should it take before the source range instrumentation would be automatically energized?
(Choose the one most correct answer). minute minute minute minute minute .
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. * PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 3 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (3.00)
Explain HOW and WHY the following parameters change as reactor power level increases at MOL, a. Fuel Temperature Coefficient (delta pcm/ F) b. Mod.erator Temperature Coefficient QUESTION 1.05 (3.00) Briefly explain how the reactivity worth of a control rod is affected by each of the following? Moderator temperature increase ., Reactor power increases, Other control rod The radial position in the cor QUESTION 1.06 (1.50) Compare the effects of a 0.5% positive reactivity addition to a subcritical reactor if the reactor was slightly suberitical
[ shutdown margin = 1%] as compared to greatly subcritical [ shutdown margin = 5%] for the following two item (No calculations are required). The change in count rate The time to reacn a stable count rat QUESTION 1.07 (2.00)
After operation at 100% power for several weeks near the end of cycle, power is reduced to 75% using rods onl Describe the Xenon transient in terms of production and removal rates from the time power reaches 75% over the next 40 hours. Include the effects from each production and removal term .
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. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 4 THERMODYNAMICS. HEAT TRANSFER _AND FLUID FLOW QUESTION 1.08 (1.50) If the reactor is operating in the power range, how long will it take to raise power from 20% to 40% with a +0.5 DPM Start-up rate?
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1. 12 se . 21 se . 36 se . 54 se b. How long will it take to raise power from 40% to 60% with the same +0.5 DPM Startup rate? 1. 12 se . 21 se . 36 se . 54 se . . . QUESTION 1.09 (2.00) Describe the relationship be tween discharge flow rate and the following, for a centrifuga; pum . Pump spee . Pump discharge hea (1.0) Define the following term . Pump Runout Shutoff hea (1.0)
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. ' '. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 5 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (2.00) Since fuel temperature cannot be measured, what power distribut-ion limit at Millstone prevents exceeding the fuel temperature limit? (0.5) What limit must be observed to prevent exceeding the clad temperature limit? (0.5) Why will the clad surface temperature peak towards the top of the core rather than the location of peak actual heat flux? (1.0)
QUESTION 1.11 (1.80) True or False The production of Xenon from Iodine is FASTER than the decay of Xenon to Cesiu As a result of an increase in power from equilibrium Xenon conditions, Xenon concentration initially DECREASE SLOWING the rate of a power decrease LOWERS the height of the resultant Xenon pea QUESTION 1.12 (2.00) The control rods must be maintained above the Rod Insertion Limits during power operatio List THREE reasons for the Rod Insertion Limit HOW and WHY does the limit change as power is increased? QUESTION 1.13 (1.50) During a cooldown of the RCS using RHR, assuming a constant decay heat rate, how will RHR flow through the heat exchanger have to be adausted to maintain a constant cooldown rate from 350 F to 150 F? Explain your answe .
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. * PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 6 QUESTION 2.01 (1.00)
Indicate whether the following statements are TRUE or FALSE concerning the Reactor Protection syste . The 109 % Power Range nuclear flux trip does not provide protection until the low range trip is manually blocke The Intermediate Range high nuclear flux trip can be blocked if 1 of 4 Power Range channels is above 10 %. QUESTION 2.02 Reaclou 'Plc.(,T2.00) With theAComponent Cooling Water Pump Control Switch in the " AUTO" position, what are three conditions which will automatically start the standby pump? (setpoints not required.) (0.75) Under what condition will component cooling water automatically be isolated to the RCP's? .. (0.5) What is the approximate (+/- 50 psig) setpoint of the CCW relief valve downstream of the thermal barrier heat exchanger? Why is it set at this value? (0.75) QUESTION 2.03 (1.50) Describe how the 120 VAC System ensures power is available to the Vital Instrument AC panel (VIAC-1) under the following condition Inverter (INV-1) failur Rectifter failur Loss of normal 480 VAC input to the inverte I
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_ _ _ . . . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.04 (2.00) State the power supplies to each of the motor-driven AFW pump (0.5) From which steam headers does the turbine driven'AFW pump receive its steam supply? (0.5)
, State the signals, logic, and coincidence that will cause each of the AFW pumps to start automaticall (1.0)
GUESTION 2.05 (1.50) Briefly describe the system used to detect leakage past the Reactor vessel head 0-ring A brief drawing may be used. Include in your description the NORMAL positions of the valves during power operatio ,
(1.5)
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QUESTION 2.06 (3.00) Briefly describe what happens to No. 1 RCP shaft seal when the injection pressure increases by 50 psig over normal pressur (1.5) What are the flowpaths for the RCP #1 seal leakoff during a safety injection? (1.0) What two parameters determine the differential pressure across the RCP al seal? (0.5)
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-- . . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYST/ES PAGE 8 CUESTION 2.07 (2.00)
State the valve positions (OPEN or CLOSED) for the following RHR System Valves for the indicated emergency conditions. Place your answers on your answer pape COMPONENT INJEC COOLDOWN CL RECIRC HL RECIRC
---------------------------- --------- -------- --------- --------- RWST Suction Valves 00r41
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-b . Sump-Suction-Va-1veW +0.4) "=
O EY Hot Leg Suction Valves LD<4) Cold Leg Discharge Valves (8d) o.[ Hot Leg Discharge Valves (Sc()
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QUESTION 2.08 (2.50) The following pertain to the plant air system __ _ When an Instrument Air Compressor control switch is in the " AUTO" position, at what receiver pressure will the compressor start and stop? (1.0) y In the-event of a loss -of Containment Instrument Air- Compressors, C-s how 'will' air be-supplieduto the-Containment Alr-Sfstem? IndI6de
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applicabls'seEpoint (1<0) True or False l I Unit 3 Instrument Air System may be cross-connected with the 1 Unit 2 Instrument Air System via a spool piec (0.5) l QUESTION 2.09 (1.00) l Briefly describe how the Charging Pump and Safety Injection Pump Net ) Positive Suction Head (NPSH) is maintained during ECCS Cold Leg l Recirculatio .
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QUESTION 2.10 (2.50) How long after a Containment Depressurization Actuation (CDA) signal does the Containment Recirculation System (CRS) actuate? (0.5) What is the reason for the time delay? (0.5) After the required time delay the CRS enters MODE I and later MODE I Describe the difference between CRS MODE I and MODE II. (1.5) QUESTION 2.11 (2.00)
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Briefly describe how the letdown pressure control valve (PCV-131) Performs its function during both normal and solid plant operation QUESTION 2.12 (2.00) State two functions of the Containment Recirculation System (CRS).
QUESTION 2.13 (2.00) Which of the four reactor trip switchgear breakers (RTA, RTB, BYA, and BYB) receive a trip signal upon each of the folleair.g occurrence In Mode 1, train B shunt trip signal is received from the control room switc In Mode 1, an automatic reactor trip signal is received from Train In Mode 3 with the shutdown banks fully withdrawn, both bypass breakers are connected and closed, simultaneousl .
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QUESTION 3.01 (2.00) What safety limits are the following RCS trips designed to protect against? Overtemperature Delta T (0.5) Overpower Delta T (0.5) Why does the Overtemperature Delta T circuit have pressure as an input whereas the Overpower Delta T circuit does not? (1.0) QUESTION 3.02 (1.50) Why is " Bank Overlap" desired? (0.75) BRIEFLY EXPLAIN why the Rod Control Startup Reset Switch is not used when recovering from a dropped control rod at powe (0.75) i GUESTION 3.03 ( .70) Which of the following expresses the combined error signal used by the Reactor Control System to generate rod motion? NOTE: Outward rod motion is positive.
' (Impulse Pressure - Nuclear Power) + (Tref - Tavg) (Nuclear Power - Impulse Pressure) + (Tref - Tavg) (Nuclear Power - Impulse Pressure) + (Tavg - Tref) (Impulse Pressure - Nuclear Power) + (Tavg - Tref)
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QUESTION 3.04 (3.00) What input signal is used to adjust programmed level for the pressuriser level control system? (0.4) What is the normal programmed pressurize'r level at no load and full load? (0.6). If the pressurizer level control channel fails high during 100% power operation, what Reactor Protection signal will cause the Reactor to Trip? Provide a brief explanation of 4 why the Trip occured and the SEQUENCE of events that led to the tri (Assume no operator action) (2.0) l QUESTION 3.05 (1.50) While operating at 100% power, the steam pressure compensation signal to the Steam Generator Water Level Control System fails low. What is the immediate response of the feedwater flow? Explain your answe . . QUESTION 3.06 (2.40) State the three Reactor Protection System (RPS) signals and ,, associated logic and coincidences that will close the Feedwater ! Regulating Valves automaticall (0.8)
! Which RPS train (A or B) actuates the automatic closure of the Feedwater Regulating Valves? (0.3)
l True or False The Manual reset buttoms on MB2 give the operator a means of resetting any Feedwater Isolation Signal while the condition ' t causing the signal still exist (0.5) , How do the Feedwater Regulating Valves and their bypass valves fail upon loss of: ,
(0.8)
' Pneumatic pressure? Electric power?
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QUESTION 3.07 ( .70) During an RCS cooldown, INDICATED pressurizer level:
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1 is less than actua Q h, c is greater than actua A33w-et is the same as actua is less than actual at the start of the cooldown but is greater than actual when cooldown is complet QUESTION 3.08 (1.20) Describe the automatic opening and closing functions that are associated
: with the motor-operated block valves upstream of the Pressurizer l PORV Logic not require !
QUESTION 3.09 (2.00) The following concern Pressurizer Pressure control State the setpoints and coincidence (ie. 2/3, 2/4, etc.) for the following RPS input . High pressure reactor tri . Enable manual block of S . Low pressure reactor tri In addition to the RPS trips listed above, state 2 additional RPS inputs from the Pressuriser Pressure detectors.
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QUESTION 3.10 (3.00) The following concern RCS instrumentatio Wide Range temperature instruments have a control input to the Cold Overpressure Protection System (COPS). Why are the Narrow Range temperature instruments NOT used in the COPS? (0.5) In addition to the COPS,.what other control inputs do the Wide Range temperature instruments supply? (0.4) List seven control inputs provided by the Na,rrow Range temperature instrument (2.1) QUESTION 3.11 (3.00) The Millstone 3 unit has been operating at 65% with all control systems in automati For each of the following conditions, give the direction of initial rod motion and EXPLAIN why the rods mov A Steam Generator PORV fails ope (0.75) A feedwater heater string is bypasse (0.75) A lower detector of a power range channel fails high (N44). (0.75) "B"(2) RCP trips of (0.75)
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s - QUESTICN 3.12 (1.50) Describe the opening interlocks associated with MV 8812A&B (RHR suction valves from RWST). What is the purpose of the interlocks? QUESTICN 3.13 (2.50) During a fuel' drop accident, how would the radioactivity release be detected? (1.0) What automatic actions occur on detection of high activity as a result of a fuel drop accident? (1.0) What detects radiation that may escape from the containment atmosphere? (0.5)
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QUESTION 4.01 (3.00) Answer the following questions regarding EOP 35 FR-H.1, RESPONSE TO LOSS OF SECONDARY HEAT SINK: List the two specific symptoms which warrant entry into this procedur (2.0) What is required if, during this procedure, the RWST level decreases to less than &2Or410-gallons? (1.0)
.saci uco QUESTION 4.02 (2.00) Why is the RHR Pump placed on miniflow recirculation for a minimum of 5 min. prior to placing the train in service for plant cooldown? (0.5) How is a low boron concentration in an RHR train (to be placed in service) corrected? -
(1.0) Would starting an RER pump, with the CVCS letdown pressure control valve (PCV 145') in automatic, result in a pressure INCREASE OR
, DECREASE in the Reactor Coolant System (RCS) during solid plant
~. operation? /31 (0.5)
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QUESTION 4.03 (3.50) Assume the plant is operating at full power and the Axial Flux Difference (AFD) has been outside the target band for the last 5 minutes. What are the TWO actions specified which you may choose between to meet the Technical Specification requirements? Include time limitation (1.5) Assume that it is 0310 on 03/18/86 and the plant is presently at 45% powe Considering the AFD penalty history below, at what date and time may power be increased above 50%? EXPLAI (Show all calculations). Assume no deviation outside the band after 0310 on 03/18/8 TIME WENT OUT TIME BACK !
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DATE OF BAND IN BAND FOWER 03/17/86 0310 0318 85% 03/17/86 1557 1637 65% i 03/18/86 0148 0310 45% (2.0)
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i QUESTION 4.04 (2.00) . , _ A leak has developed in a CVCS Letdown piping component located - outside the containment building and may be manually isolate The general area radiation level in the area where the leak is to be isolated is 600 millirem per hour. The one available person to perform the work informs you that his present quarterly exposure and lifetime exposure levels are 2.90 Rem and 54.75 Rem respectivel Using only 10 CFR 20 whole body exposure limits as a guide, how long may this person work in the area before the quarterly exposure limit is exceeded? (SHOW YOUR WORK) (1.5) What is the minimum age that this person may be to perform the work? (SHOW YOUR CALCULATIONS)
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QUESTION 4.05 (2.00) Assume a peripheral control rod drops into the core at 100% powe According to AOP 3552 " Malfunction of the Rod Drive System". How is Tavg matched with Tref? During recovery why is a Rod Urgent Failure alarm received? Under what circumstances must the AUTO / MANUAL switch at the Pulse / Analog (P/A) Converter be held in MANUAL during rod recovery? Who(m) must be consulted if the rod has been inserted for more than i hour? QUESTION 4.06 (1.00) State the two entry conditions (symptoms) of high RCS activity as listed in AOP 3553 "High Reactor Coolant System Activity".
.. QUESTION 4.07 (1.50) The following pertain to tagging procedures as stated in ACP-QA-2.06A
" Station Tagging". Who are the only personnel that are authorized to issue Safety Tags to listed qualified individuals? What is the purpose of a " blue tag" found on a valve? True or False A tag of no other color may be attached to a switch or device bearing a blue ta l I .-
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QUESTION 4.08 (2.00) The following concern ES-0.2 " Natural Circulation Cooldown". Under what conditions is switchover to AFW pump supply to alternate water sources required? While maintaining a cooldown rate of <50 F/hr, what RCS parameter-is utilized to monitor the cooldown rate? Describe the method used to lower RCS pressur While maintaining RCS subcooling >80 F, what RCS temperature indication is utilized to monitor the subcooling?
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QUESTION 4.09 (1.00) I l What are the Millstone 3 administrative whole body limits according to SEP 4902 " External Radiation Exposure Control and Dosimetery Issue" for the following: With NRC Form 4?
' With NRC Form 4 and approved exposure upgrade? -.
QUESTION 4.10 (2.50) The following concern precaution statements listed in OP 3201 " Plant Heatup" The shutdown banks must be fully withdrawn whenever positive reactivity is being added by Baron or Xenon changes, RCS temperature changes or control bank movement except during 2 condition State the two condition Why must pressurizer pressure not be allowed to exceed 1985 psia until Steam Generator pressure is >585 psig? l Which RCP is the preferred pump for single pump operation? If RCP-A(1) is stopped, why must loop A(1) spray valve be closed?
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QUESTION 4.11 (2.50) State 4 of the 5 entry conditions for "Immediate Boration" AOP 356 Under what conditions may Immediate Boration be stopped?
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QUESTION 4.12 (2.00) State the four Immediate Operator Actions listed in ECA 0.0 '* Loss of All AC Power" INCLUDE how these actions are verified. NOTE: Sub steps are required.
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'
7 = ( L*/s ) * ((3 - s'/ Is } 7 = t/(s - 3) :4 = (1 . g }/(1 - Xg))
<
T = ( 3 - s )/(Is ) SC4=( . gf)/Kdf
<
t' = 10 sec:ncs a = (Kgf-1)/Xg= * M df*df / i = 0.1 seconcs*I l
=((t*/(TAdf Il * bTdf /(I * TIII ~
I;a; = i e2 =z d
I4 7 I.22 7 = (:4V)/(3 x 1010) R/hr = (0.5 CZ)/c 2(=,g,73 y t = sd A/hr = 5 CE/d (feet) ,
'415cs11aneous C:nveesfens 'datar dar*. mete's i gal. = 3.345 le curie = 3.7 x 1010:33 1gaj.=3.7811 tars 1 '4g==2.54 I nc 2.21xtem 10 } 3tu/Mr 1 = 7.48 gal .
Gensity = 62.4 lem/ft3 1 mw = 3.41 x 100 3tu/hr Censity = 1 ;:s/c9 lin = 2.54 ::s , l
? = 9/5'C + 32 Heat of vacoritaticn = 970 Stu/ lcm 'C = 5/9 (?-32)
Meat of fusion = 144 Stu/ltm 1 At2 = 14.7 psi = 29.9 in. H . I aiu = 778 ft.lbf I ft. H O 2
= 0.4335 luf/i ..
> . * . PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 19 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, F.
I i
, _
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ANSWER 1.01 (1.00) At 200 F density of water is greater; (more Ibm /ft3) therefore, it takes more power to move each ft3 of water the same distanc (0.5) m at 200 F is greater than m at 530 Example: m=( v) (p) (3600/1hr) m/p at 200 F = m/p at 530 F s If p increases m increase if p decreases m decrease (0.5) l i REFERENCE GP HT&FF pp. 288-289, 285
, !
ANSWER 1.02 (3.00) Increase [0.25]. Th will increase whil'e Tc remains relatively constant [0.5]. Increase [0.25]. Heat is no longer being removed at the same rat [0.5]. Decrease or no change [0.25] less primary to secondary heat transfer
< [0.5].
. d- Decrease W 5]. H e at-s i nk-removed -d riv i n g-head -redu ced -fo-Sib REFERENCE MP3 GP HTFF p. 357
. 000056A248 000056A257 000056K101
.
ANSWER 1.03 ( .70) ;
; I REFERENCE MP3 Reactor Theory RT-11 p. 4 :
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. . ~ . .-- - . .. . _
. ~ . . . . _ . - . _ _
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. PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 20 ~
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MILLSTONE 3 -86/04/D1-JAGGAR, A301
ANSWER 1.04 (3.00) 0.7f a. The Fuel Temperature Coefficient becomes less negative as power The resonances associated with Doppler broaden
~
increases L0c5]. and overlap ,with 0<51
,,;- increasing fuel temperature. [i).1S~
b. The MTC becomes more negative as power increases because Tave increases.J0:El Since the MTC is principally a function of moderator density change per F, at higher temperatures the change in density per F is greater. .LO<5] 4 c. W REFERENCE MP3 Reactor Theory RT-13 p. 4; RT-12 pp. 6-8 00100K549
'
ANSWER 1.05 (3.00) Increases -------- as moderator temperature increases, density
[0.25] decreases allowing neutrons to travel further thus having a higher probability of reaching -
a control rod [0.5]. Increases -------- sk h caused by an increase in the neutron flux
[0.25] level- in the cor [0.5)
d i r.h . bd'm Decreases -------- due to rod shadowing- the other rods absorb the
[0.25] neutrons and reduce flux for the rod in question [0.5]. t~.li who accep bw . fAJ g NJg.$ " Central control rods have a higher worth then those on the edges [0.25]
because relative flux tapers off at the edges of the cor [0.5] w.tl c.cecp f ddemed a@ ode - sons REFERENCE (% g ,t e,l5,} o p cy,c, m,, by. ,4 L ge, MP3 Reactor The'ryo RT-14 pp. 2-5 g g 001000K502 , 001010K504
/
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t PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 21 IBERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ' ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER 1.06 (1.50) The slightly (greatly) suberitical reactor will have a larger (smaller) increase in count rate. (0.75) The slightly (greatly) s'uberitical reactor will take a , longer (shorter) time to reach a stable count rate. (0.75) l REFERENCE MP3 Reactor Theory RT-8 pp. 2-5 015000K506 't i ANSWER 1.07 (2.00) After the power decrease, the production of xenon from fission [0.3] j and from the decay of iodine [0.4] is greater than the removal by decay of xenon [0.4] and burnout by flux. [0.3] After five hours, the removal rate is greater than the production [0.3] and positive reactivity is being added until equillbrium at about 40 hours. [0.3] REFEFINCE ._ MP3 Reactor Theory RT-16 pp.4-5 001000K538 i . ANSWER 1.08 (1.50) [0.75 ea.] REFEPINCE MP3 Reactor Theory RT-10 pp. 3-8 .. 001000K547
_ _ _ _ _ _ _ _ - _ - _ _ _ - - _ _ _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ . - _ - - - - - __..____ _ _ _ . _ _ _ _ _ .
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.- s PRINCIPr.rR OF NUCr. WAR POWER PLANT OPERATIO PAGE 22 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW _ ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER' 1.09 (2.00) . As pump speed increases, discharge flow rate increases-proportionally. [0.5] As pump discharge head increases, discharge flow decreases (according to the pumps characteristic curve). [0.5] w,it Ao ourp+ diuussio- A O,n,e0_n s?dscumm p a cu.,,ci . Pump runout is the condition when a centrifugal pump is pumping at its maximum capacity. (Greater than the design flow rate). [0.5] When a pump at shutoff head is pumping against a shut discharge valv (Max head the pump can deliver) [0.5] tu. tl a.lz. o a ccep f "s3 d % p ec: s w r e o b o v e e Aick h e e c, , p vi l l REFERENCE not p redu ce Clod . GP HTFF pp. 320, 322, 328 COMPONENT-PUMPS ANSWER 1.10 (2.00) w.t\ a.:,o a u rpY P( $. Voa , ~~ Q N' D '* 0 n . Local power density-KW/FT. [0.5] P,3.:.t-! gn,- c.D-1 ( C o ,Tc.vg s f <u c ') DNB (DNBR) [0. 5] M:o acc.epb kw[g y Clad surface temperature is a function of heat flux and moder-ator temperature. [0.5] Moderature temperature is higher at the top of the core. [0.5] REFERENCE MP3 GP HTFF pp. 224-228, 243-244 002000K510 002020K511 ANSWER 1.11 (1.80) TRUE TRUE TRUE j l l l
. . - - -
-e
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,. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATIO PAGE 2 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW _ ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, REFERENCE MP3 Reactor Theory RT-16, pp. 1-6 001000K533 ANSWER 1.12 (2.00) . Ensure the capability to insert adequate negative reactivity such that sufficient shutdown margin exists on a reactor tri . Minimine the amount of positive reactivit{(ap M ejected ch i o V yod
"^- can add to the core.^'2* "'W+ ,6t M*Aa w. t hm < n r o n a n oc mi+2 ccieret eminus Ensure acceptable power distribution limits are maintaine (1.0) The rod insertion limits increase as reactor power (dT) increases [0.5].
As power increases, the power defect inserts negative reactivity
[0.25]. On a reactor trip, this negative reactivity is removed as pcwer decreases and Tavg decreases to its no-load value [0.25].
to.n ci a ao wt - Maa d' w yer a , % y~ h Rtt w (,_ g,y] REFERENCE MP3 Topic 6 Lesson 3, pp. 26-27 001000K504 ANSWER 1.13 (1.50) As the RCS temperature drops, the temperatuare difference across the RHR heat exchanger also drops [0.25]. The cooldown rate will be lowe [0.25). As a result, the flow through the HX must be increased to maintain the same heat removal rate (0.5] . REFERENCE GP HTFF P. 178 005000A102 l
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l l
,, - - - -- - .e,- --,m- e
. * - .- .- , PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24
' ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER 2.01 (1.00) F F REFERENCE MP3 Topic 6 Lesson 4 pp.16, 22 015000K407 ANSWER 2.02 (2.00)
)Jo%I C.C.u) pw-p wok ru nw m3 Lee pump discharge-pressure-(-80 psigt# -
Blackout Sequence SI Sequence [.75] Phase "B" isolatio [0.5] psig. v/- 50 psig [.25] The piping it protects may be subjected to full RCS pressure if a thermal barrier.HX leak develop [.5] REF[RENCE MP3 Topic 4 Lesson 1 pp. 11, 15, 16 00800CK401 008030A304-i ANSI 4ER 2.03 (1.50) The Static Transfer Switch automatically transf rs to the alternatesourceof480VACpowerfromSus(32-2 [0.5] The DC supply to the inverter is available from the Battery charger and/or battery (301A-1). [0.5] Same as [.0 . 5 Mil als.o ace cet ]pcwer is Aveddcle_ kr% oO#"" P N CC Q*f2<"J"]- REFERENCE MP3 BOP Vol. 1 120 VAC System pp. 10 & 16 062000K410
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' - . . PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER 2.04 (2.00) PIA--34C PIB--34D [0.25 each] Main steam from headers A, B, & D [0.5]
O.% P1A & P1B--2/4 [D<f] lo-lo level from 1 S/G [0. 2] -if-stop--valvesA - open--[0 .-1LA-
--LOSP, SIS, CDA [0.3]
P2---------2/4 [0.1] lo-lo level on 2/4 S/Gs [0.2] REFERENCE MP3 Topic 4 Lession 2 pp. 18, 19, 22, 23 061000K202 061000K414 ANSWER 2.05 (1.50) Leakage through the 0-rings is collected by two leak-off lines. One line
~
penetrates the area between the inner'and outer o-ring, the other starts just outside the outer 0-rin Both lines have manually operated valves' . Downstream of the isolation valves. the individual lines join to form a common heade The common header' esn be isolated with an air operated i valv .s n%, ILAc. fw, go>a<q: G 8031 T*" l
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1 D*N Gaseaus f % Dem,n s \ 8o69 A p__ 8069A closed--Outer seal isolation valve 8069B open--Inner seal isolation valve , 8032 open----Common Valve to drain (8076 closed--Blind flange isolation) [0.2 ea.] REFERENCE l MP3 Topic 1 Lession 1 K405
. -
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- MILLSTONE 3
-
-86/04/01-JAGGAR, ANSWER 2.06 (3.00) As pressure increases, a closing force is exerted on the seal rin [0.5] The narrowing between the seal faces restricts the flow and increases the pressure felt on the underside of the seal face. [0 *. 5 ]
The increased pressure pushes the seal ring back up, opening the flow passage which allows more flow to escape, [0.5] thus re-establishing a correct equilibrium positio CUTF Through #2 seal to thelECDT; [0.5] and the #1 seal return line relief valve to the PRT.70.5] RCS pressure and the backpressure created by the VC [0.5] REFERENCE MP3 Topic 1 Lesson 2 pp. 7, 8, 25 Topic 2 Lesson 1 p. 28 002000K602 ANSWER 2.07 (2.00) INJECTION COOLDOWN CL RECIRC HL RECIRC _________ ________ _________ _________ OPEN CLOSED CLOSED CLOSED
/
b r CLOSED --CLOSED OPEN- --_OPEN # ' CLOSED OPEN CLOSED CLOSED OPEN OPEN CPEN' CLOSED c.t.45 E. D CLOSED CLOSED CLOSED OPEN QL-1EACH]
- REFERENCE MP3 Topic 3 Lesson 4 pp. 62-68 005000K408
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- MILLSTONE 3 -86/04/,01-JAGGAR, ANSWER 2.08 (2.50) Compressor starts--90 psig in receiver f f93D Compressor stops---110 psig in receivertsP3D[1.0]
' The Inst'rument Air System will s supply pressure via ( 31 AS-PV15 ) 'swhen Containment -Instrument-Air 3ressure drops ~to'100 psia'.v[1\O-]d D
e True [0.5] REFERENCE MP3 Instrument Air System Description p , 16 078000K401 078000K402
.-
ANSWER 2.09 (1.00) Valves 8804A & B are opened from tha Containment Recire. Pump discharge to the CCP and SI pump suction REFERENCE MP3 Topic 3 Lesson 4 pp. 67-68 000074K309 -- 006000K406
..
ANSWER 2.10 ( 2. 5Q 6Go secoJs (11 m mular)) mir.utes- [0.5] Allow sum time for sufficient water to collect in the containment
[0.5)
l MODE I-- all four pumps take suction on the containment sump [0.5] ; and discharge to spray header [0.5) MODE II-Two pumps lined up to spray headers. [0.25] Two pumps realigned to supply low pressure safety injectio [0.25] RNFERENCE MP3 Topic 3 Lesson 4 pp. 70-71, 86 0060C0K102 006000K405,
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. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 l
-
ANSWERS -- MILLSTONE 3 -86/04/,01-JAGGAR, ANSWER 2.11 (2.00) cScd During normal plant operation) PCV-131 is used to maintain a constant backpressure on the letdowrF21owmontrol-valveF(This prevents the letdown flow from flashing to steam in the letdown lines or letdown heat exchanger.) [1.0] When.the plant is solid, water can be diverted from the RHRS to provide letdown upstream of the letdown heat exchanger. PCV-131 can be manual 4y' set to maintain desired RCS pressure by controlling RHR pump discharg [1.0]
! REFERENCE MP3 Topic 2 Lesson 1 pp. 11-12 I
004010K505 004020K612
.
ANSWER 2.12 (2.00) Condenses steam E0:5-] thereby depressurizes the containment atmosphere. EO - 5 ] C .o _ Removes core decay heat (over the long term by injecting water into the core). [1.0] REFERENCE MP3 Topic 3 Lesson 4 p. 69 i 006050G004 ANSWER 2.13 (2.00) RTA RTB BYA BYB ____ ____ ____ ____ TRIP -- TRIP TRIP -- -- TRIP
'c . TRIP TRIP TRIP TRIP [12 @ 0.17 ea.]
REFERENCE MP3 Topic 7 Lesson 1, 2, & 3 pp. 20-21 012000A307
, - - - n , - --,,-,,--,n---+m - , , - - - - - - , - e - ---c-
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~ INSTRUMENTS AND CONTROLS PAGE 29 ANSWERS -- MILLSTONE 3 -86 /04 /,01-JAGG AR , ANSWER 3.01 (2.00) . DNB Excessive fuel power (KW/ft) [0.5 ea.] DNB is a function of pressure whereas KW/ft is not related *
to pressur [1.0] REFERENCE MP3 Topic 7 Lesson 1, 2, & 3 pp. 44-49 012000K402 012000K403 ANSWER 3.02 (1.50)
'
a. Maintain an even flux distribution or prevent flux peaking. [0.75] j
'
b. (Only one group of rods needs to be reset). The Startup Reset Switch will reset all control rod group [0.75] REFERENCE MP3 Topic 6 Lesson 2 pp. 42, 25, 11, 46 - ~ -
: . . _ .
001010K403 _ 001010K501 *** - - 000003A102 ANSWER 3.03 ( .70) REFERENCE MP3 Topic 6 Lesson 2 pp. 19-20 001000K403
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, INSTRUMENTS AND CONTROLS PAGE 30 ANSWERS -- MILLSTONE 3 -86/04/,01-JAGGAR, F.
. ANSWER 3.04 (3.00) Tave (auct high) [0.4]
- C2.r5] C t .r *,2 25.0%,^s61.6% A [0.6]
' High pressuriser level trip [0.4] charging flow decreases [0.4]
' pressuriser level decreases [0.4] letdown isolates [0.4] and pressuriser level increases [0.4] , REFERENCE MP3 Topic 8 Lesson 4 I & C Failure pp. 61-62 Topic 6 Lesson 6 & 7 p. 25 011000K404 011000K104 I ANSWER 3.05 (1.50)
'
Feedwater flow would decrease [0.5]. The failed compensation signal would cause a decrease in the steam flow signal [0.75] causing a SF/FF mismatch [0.25] causing feed flow to decreas REFERENCE _
'
MP3 Topic 6 Lesson 9 p. 18 035010A301 ANSWER 3.06 (2.40) i A/4 . P-14 [0.1] J/f (0.1] levels on 1/4 [0.1] Steam Generator . Safety Injection [0.1] 1/1 l0.1] , . P-4 [0.1] 1/2 [0.1] trip breakers open in coincidence with a j
' low Tavg in 2/4 [0.1] loop j A [0.3] False [0.5] + . Closed Closed [0.4 each]
REFERENCE MP3 Topic 6 Lesson 9 pp. 31-32 1 059000K419 l 059000A212 i 059000A411 i _ -_ _ __ __ , __ . - - - -
- - - .. . . . _ . . _ . . . . . . . .
_
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3 '. INS'TRUMENTS AND CONTROLS PAGE 31 ANSWERS -- MILLSTONE 3
-
-86/04/01-JAGGAR, ANSWER 3.07 .( .70) ,
' REFERENCE MP3 Topic 6 Lessons 6 & 7, p. 23 3 011000K403 , ANSWER 3.08 (1.20) c> . G i Open-->2200psig IDd1
--(control switch in OPEN) <- --COPS ARM / BLOCK to ARM 10-c4]
Glosed--<-2200psig [Ov4-3 D-REFERENCE MP3 Topic 6 Lesson 6 & 7 p. 17 010000A403 i I- _ . . . I ANSWER 3.09 (2.00) 5
-
1 'A T'! O n , pr 9 4 . .2400' psig # 2^[4 psig 2/3 o- ; ~) f 1 ( "- IS60'psig 2/4 (Setpoint varies as lead-lag circuit) IEYr o- rg,o pug [0.2 each] Lew pressure Safety Injection
Loop m l \ shOTdT u a te ep setgoint - 'PoRVcircuit
[0.4 ea.]
Clot.h. G 1*uro e sm" REFERENCE O- S4 c- A MP3 Topic 6 Lesson 6 & 7 pp. 17-18, 6 010000K101
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INSTRUMENTS AND CONTROLS PAGE ANSWERS -- MILLSTONE 3 -86 /04 /,01-JAGG AR, ANSWER 3.10 (3.00) When the COPS is in service, the RCPs are not operating, therefore, the Narrow Range instruments which are located in the bypass manifold would not be reliable. Will also accept narrow range instruments range does not extend to'450 [0,5) Loop stop valve logi [9.43 c Rod control OPdT OTdT Steam dump Feedwater Isolation Pressurizer level A* '% I d' Rod insertion limits d0.3 each] E. C-!G in pet REFERENCE - MP3 Topic 6 Lesson 8 p , 19 010000K403 ANSWER 3.11 (3.00) If a S/G PORV opens, the increased steam demand on the S/G wil i cause Tavg to drop.[0.25] The control rods will move out to match Tavg with Tref. [0.5] The overall feedwater temperature to the S/G will dro [0.5] Tavg will decrease and rods will move ou [0.25] The N44 signal to the rod control system increases. This ) indicates a rate of mismatch between reactor power and turbine l loa [0.25]. The mismatch circuit will cause rods to drive i l
[0.5] All rods will trip in [0.25] because of a single loop loss of flow ,
tri l
[0.5]
REFERENCE MP3 Topic 6 Lesson 2 001000K602 l l l
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.. . INSTRUMENTS AND CONTROLS -
PAGE 33 ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, , ANSWER 3.12 (1.50) i Valves cannot be opened unless 8804, 8837, 8838A are in the fully closed positio [0.75] _ Prevents a potential flow path between the RWST and the Containment-Recire. Pump suction when either RHR or Containment Recire. system is operatin [0.75] REFERENCE MP3 Topic 3 Lesson 4 p. 63 005000K407 ANSWER 3.13 (2.50) Detectedby(RE41&42)monitorsinsidecontainmen [1.0) {ClosesCTV32A, B and 33A, B; stops containment purge supply
,m and exhaust fan [1.0] __ / Detected by the ventilation vent (extended range) monitor RE-10. [0.5]
tyU hval se w r v- nu9 %( o a49 VEL i$ &cem s,4 \..w o n, , ,
'
j # S ' * ' ' " 3 " OP ' "*" MP RM A401 072000K401 073000K401 m
; \ C. Lag e I.o rec.A-LE QQ. d meh DOLE L bV mQR e
- 6 m p ,- clo w o d oe t eybp i+ k%"
1 3 hAwak S~ Sk o 5 09 k by } 4"%I
- Co .7r Cr ac3 =.5 , o .a r b squ enc e']
*
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r-'. , - - - --, +3-4 . - + + w
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.. . ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 34 j RADIOLOGICAL CONTROL
.
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-86/04/01-JAGGAR, '
ANSWERS -- MILLSTONE 3 ANSWER 4.01 (3.00) ) From EOP 35 E-0 " REACTOR TRIP OR SI" when minimum AFW flow is not verifie [1. 0] 3 H Yo cd 2) S/G NR level < h with total FW flow to S/G's <(A3)gpm (red path condition). .0] C' b. ECCS should be aligned for cold leg recir [1.0] REFERENCE MP3 EOP 35 FR-H.1 pp. 2,9 000054G010 . l l ANSWER 4.02 (2.00) l To mix water for sampling <
. CecAne b e- coa c ,w.d %[0.5]
r. n c t :s *, RA.> Flowpath aligned from RWST o CVCS-letdown-(without-exceeeding_r._,~ AO-gpm-thru-4?TDN-HX4 unti boron concentration is equal to or . greater than RCS boron con entratio [1.0] Decrease [0,5] * 9"Y " REFERENCE MP3 OP-3310A p. 11 OP-3208 p. 7 005000G012 005000K109
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: PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 35 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, F.
. ANSWER 4.03 (3.50) Within 15 (or next 10) minutes [0.5] either 1. Restore the indicated AFD to within the target band [0.5], or 2. Reduce the thermal pcwer to <90% of rated thermal power.[0.5] a Accumulated penalty over the past 24 hours is 89 minutes.[1.0] The penalty will be reduced to 60 minutes at 1618 minutes on 03/18/85 and then power may be increased.[1.0] 85% 0318-0310 = 8 [0.25] 65% 1637-1557 = 40 [0.25] 45% 0310-0148 = 82/2 = 41 [0,5] 89 min total penalty 03/17/86, from 1557; 81 min left -60 = 21 min -> 1618~03/18/86 [1.0] REFERENCE MP3 Technical Specifications 3/4.2.1 p. 1-2 001000A303
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, ANSWER 4.04 (2.00) mrem = 100 mrem dose remaining [0.5] l 600 mR/Hr X 1 Hr/60 minutes = 10 mrem per minute [0.5] 100 mR/10 mR/ minute = 10 minutes [0.5] ' (N-18) = 54.75 + 0.10 [0.3] N - 18 = 54.85/5 N - 18 = 11 N = 11 + 18 = 29 [0.2] REFERENCE MP3 SEP 4902 p. 6 ,, 9990150000
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, , , * PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 36
. RADIOLOGICAL CONTROL 4 ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER 4.05 (2.00) 0 2f' By adjusting, turbine load. 10 57 o r b M m a l n * < c.sJ c r q C.o . .r] An attempt has been made to move a group of rods with all lift coils disconnecte [0.5] to.d aho ou el "t4.y d R Js < d u b r +b.h w G,J s.,. If the dropped rod is a control rod.' [6.5] Reactor Engineer (ing). [0.5] REFERENCE MP3 AOP 3552 App. A 001000A203 . ANSWER 4.06 (1.00) High activity based on chemistry sampl [0.5] Failed fuel monitor alar [0.5]
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REFERENCE
'MP3 AOP 3553 p. 2 0000076G010 ANSWER 4.07 (1.50) ;
l The onshift SS/SCO/SR [0.5] j The valve is to be cperated only by order of the individual to whom the tag is issue [0.51 Tru [0.5] REFERENCE MP3 ACP-QA-2.06A p , 15 9990140000 , l _ _ - . . - . _ ,
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U PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 37 RADIOLOGICAL CONTROL ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, ANSWER 4.08 (2.00) . When DWST level decreases.to <80,000 ga [0.5] RCS Cold leg temperatur [0.5] Letdown (or Par PORV) and aux. spra [0.5] Core exit TC [0.5] REFERENCE MP3 ES-0.2 pp. 4-6 000074K311 ANSWER 4.09 (1.00) mr/ quarter [0.5] mr/ quarte P -for53 '
.ts -t ; c REFERENCE <w LL i MP3 SHP 4902 p. 9 frcm & o .3 cw - r deu d < J3 p e - na n < 5 9990150000 d osi m e.43 r,cJ q . C. o . a.C i
ANSWER 4.10 (2.50) . The RCS has been borated to the cold shutdown concentratio . The RCS has been borated to the hot Xenon free concentration and is being maintained at no-load Tav [1.0] To prevent inadvertent Safety Injectio [0.5] B(2). [0.5) To prevent "short-cycling" spray flow from the other loop. [0.5] REFERENCE MP3 OP-3201 pp. 10, 13, 24, 27 002020G007
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ANSWER 4.11 (2.50) I . Control rod bank height below the rod bank low-low limit alarm.setpoint with the reactor critical, t ! Failure of one or more control rod clusters to fully insert i following a reactor trip or shutdown, indicated by digital rod position indication syste . Uncontrolled cooldown of the reactor coolant following a reactor trip or shutdow i
' Uncontrolled or unexplained reactivity increase, indicated by abnormal control rod bank insertion, increasing Tavg or increasing nuclear powe . Failure of the Reactor Makeup Control System to the extent that the makeup system must be bypassed to a. :omplish boration of the Reactor Coolant Syste '
[4 of 5 required. 0.5 each] When the entry condition is satisfie [0.5]
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: . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 39 RADIOLOGICAL CONTROL
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ANSWERS -- MILLSTONE 3 -86/04/01-JAGGAR, . ANSWER 4.12 (2.00) Verify reactor tri . Trip and bypass breakers ope Neutron flux decreasin . Verify turbine tri All turbine stop valves close . Check RCS isolate Pressuriser PORVs close Letdown isolation valves close d Excess letdown isolation valves close i f 9I Verify AFW flow >S25 gpm per intact S/ [1C at 0.2 ea] , v . REFERENCE E" # " * <# >NSf' o 4* - Y ""3 ' '
MP3 ECA 0.0 pp. 3&4 000055K302
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MILLSTONE 3 _________________________ REACTOR TYPE: PWR-WEC4 _________________________ DATE ADMINISTERED: 86/03/25 _________________________ EXAMINER: HANNDN, _________________________ APPLICANT: _________________________ _ INSTRUCTIONS TO APPLICANT: __________________________ Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer sheet Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Enamination Papers will be picked up sin (6) hours after the examination start % OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY _____-__ ______ ___________ ________ ___________________________________ 25.00 25.00 ________ ______ ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND T HERMOD YN AMJ CS 25.00 25.00 ________ ______ ___________ ________ PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUhENTATION 25.00 25.00 ________ ______ ___________ ________ PROCEDURES - NORMAL, ABNORMAL, CHERGENCY AND RADIOLOGICAL 22.00 CONTROL 25700- 25.00 ________ ______ ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS f,G.0 0 4 0G+00- 100.00 10fALS ________ ______ ___________ ________ FINAL GRADE _________________% All work done on this examination is my own. I have neither given nor received ai ~~~~~~~~~~~~~~ 5PFLEC5hTIU~555U55UFIE
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.THERH0 DYNAMICS
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l QUESTION 5.01 (1.50)
; Consider the MP3 reactor operating at constant ipower and temperatur Will the neutron flux sensed ;by the excore power range nuclear instruments'at EOL i ;be GREATER THAN, SMALLER THAN, or THE SAME AS that !
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: sensed .at BOL ? Justify your answe C1.03 l l What steps are taken to ensure that the encore power range nuclear instruments accurately represent core power throughout care lifetime ? CO.5]
QUESTION 5.02 (1.50) What is REFLUX ColLING and when would core cooling be provided by this mechanism ? OUESTION SiO3 (2.00)
:. List three reactor safety concerns if Rod Insertion Limits are violate C1.53 3 You commence a start-up with SR counts reading 2 X 10 counts per second. Assuming the SR has not de-energized, you would expect to go critical about the time SR count rate has reached (select one from the following): CO.53 3 X 10 counts per second 3 SX 10 counts per second 3 A .O counts per second
4 32 X 10 counts per second
1 l OUESTION 5.04 (1.00) j Which condition would result in a higher SUR: a rod ejection accident at DOL or EOL ? Why ? CO.753 Tha reduction in U-235 alone accounts for the change in Beta Bar with fuel bornu (TRUE/ FALSE) CO.253 (***** CATEGORY 05 CONTINUED ON NEXT PAGE mrmre )-
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QUESTION 5.05 (3.00) For the follouing conditions, will the CALCULATED ECP for a startup performed 4 hours after a trip from a 60-day 100 % power run, be HIGHER THAN, LOWER THAN, or the SAME as the ACTUAL contro' rod position at criticalit Treat each condition separately. Dr i ef ly e>:pl ai n your answer . DOL Rod Worth Curves were incorrectly used to calculate ECP uhen EOL conditions e >: i s t . E1.03 Previous reactor critical data was generated during a 100 % power run which lasted for 2 hours and was ,p n preceded by an outage of 12 hours. (A rrv m e g re /;e v; 6C- c/mf /M ,'* I E1.0] One reactor coolant pump is stopped three n.i n u t e s prior to criticalit E1.03 OUESTION 5.06 (1.00) Why is individual RCP flow higher for 3 loop than 4 loop operation? QUESTION 5.v/ (2.50) For DOL conditions, at what rip 3 a::i a l core locatiori, TOP, riID D L E , or DOTT0n. is the eritical heat flux at a ni a >: I m u m ? CO 253 Drtefly 'xplain ob [0.75] How does the magnitude of t h- ritical heat flu: change ( IrJ C RE A S E , DECREASE, STAT " SAME) as the following parameters are DECREASE Conside, cach separatel E1.53 Tavg R l'. 5 p r t! $ 5 o r P
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i OUESTION 5.08 (3.00) Explain the response of reactor power and Tave during and after 2 minutes of Emergency Boration at 100 % powe Asso.we rod control is in manua E1.5] i Explain the response of reactor power and Tave after 2 minutes of Emergency Boration at 10E-8 amps and no loso Tav E1.5] 00ESTION 5.09 (3 00) The plant is operating at 25% power when the 12 steam generator main steam isolation valve fails shut. Given the initial conditions belo9 indicate the direction (STAYS CONSTANT, INCREASES, DECREASES) of the final steady state values for the listeo p a r a n.e t e r s . Assume no operator actions, all control svstems in manual, and no reactor trip or SI occur INITIAL CONDITIONS " lave = 565 F Tstm = 550 F Core delta T = 15 F Th = 572 F Turbine pover E0.25] Tave for affected loop
[0.253 Tave for non-affected loors CO.75] S/G prossore for affected loop EO.75) .-
b .' G pressure for non-affected loop EO.751 Will S/G code safety valve (s> lift ? [0.25] GUESTiON 5.10 (2.50)
(ou are operating at 100 % pouer with RCS Tave at 567 F end s stean pressvee of 745 psig. What most Tave be changed to in o r d .> r to mcintein theso condations uith 20 % of the tubes in eact. steam geoorat.or plugge i ?
a n '. apeltccblo formula Shou all work, including i. t a a * * CATEGORY 05 CONTINUED ON HEXT PAGE *****)
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_--------_--_- DUESTION 5.11 (2.00) For the following situations, indicate whether the final stable power level uill be HIGHER, LOWER, or THE SAME.as the initial power leve Explain your ansvers. Assume the initial power level is at the point of adding heat follouing a normal reactor startup at the end of lif Consider each situation separatel Steam d u nip pressure setting is lowered by 20 psi ro.53 A small steam leak develops inside containment that is insufficient to initiate SI or CI, but is sufficient to cause an increase in steam floe for the affected steam generato [0.51 ppm of boron is added. with all systems in manua CO.5] The steam generator level detector for the SGWLC system fails hig CO.5] 00ESTION 5.12 (2.00) What is the startup rate if a reactor is initally critical and a rod withdrawal results in the intermediate range indications increasing from 2E -8 amps to 5 E -7 amps in 1 minute and 45 see? E0.83 Using the folloeing initial conditions, calculate the minimum number of steps of rod bank insertion required to ensure the reactor is soberttica E1.23
- SUR is 0.5 Deh e f f e c t.1 v e delayed neutron fraction = 0.005-average neutron precursor decav constant = 0.00 sec -1-rod bank uor = 5 pcm/ step (***** END OF CATEGORY 05 ******
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QUESTION 6.01 (1.00) The design of the RCS is such that the core is more likely to uncover for a cold les break than for a hot leg break of the same sit E>: plain uhv this is true.: QUESTION 6.02 (3.00) If 6.9 KV Bus 35A is lost chile operating at the following pouer levels, will the reactor trip? Justify your answer by providing the applicable trip setroint and required logic and coincidence. Consider each case separately and independently, a. 75% pouer [1.53 power C1.53 00ESTION 6.03 <2.00) The steam pressure detector for i 1 SG sticks at the 100% value ehen turbine load is decreased from 100% to 75%. Starting with the load decrease, e::p l a i n the SGWLC System signal processing, including ALL steps of the resulting transient and endin3 at the final stable condition Assume no operator intervention and all control systems in a u t o ns a t i c . QUESTION 6.04 iZ.007 Will a SG Safets Valvet;i lift to a tube r uptur e event if the atfected SG i- t s o l a l,a d , SI is actuated and primary pressure levels out et 1200 psi 3 ? E: r i e f l y e:: plain your answe El.03 In the abovo case the erimary system is belt 9 cooled doon with the una f f ec tori s t e a n, Jenerators and the primart svutom t, e m p e r . t o r n is .a t 565 dagrees The affected s t e a n. Jenerator t isolated and its temperature settles out at 570 dogrees RCPs are secured, SI is terminated. and affected primarv loop isolation valves r e n. a t n ope How uov i ri ths second4rv svntem interact uith'the primarv system undor these c o n rii t i o n s ? [1.03 (***** C ie F E G O R'r v6 C 0 rl T Ir1U E D O rt rJE X T PAGE *****) i
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j -QUESTION 6.05 (1.00)
- Given the following ECCS pump discharge design flou rates j at nominal discharge pressures, ESTIMATE the leak rate if j
i a Loss of Coolant Accident occurred, one train of safety injection fails to operate, pressuriner level stabilizes (Cdj , , ! on scare, and RCS pressure stabilizes at 1600 psis ? E0.5] Justify your answe CO.5] } I Containment Recirculation Pump 3950 gem Safety Injection Pump 440 3pm RHR Pump 4000 gem j Centrifugal Charging Pump 150 gpm / a a GUESTION 6.06 (2.25)
- Refer to the attached Figure 6-1, MAIN ELECTRICAL
; DISTRIBUTION and indicate the electrical lineup for each of the follouing conditions by listing in the below matri:: the position of each breaker shoen (0 for OPEN or C for CLOSED).
~ ~ Starting up number three unit (MODE 2) No. 3 riain Generator synchronized (MODE 1) } Station blackout (from MODE 1)
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{ PLANT CONDITION ------>> . 3.
j BREAKER DESIGNATION
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I \/ l 15C-30 } 35A-1 358-1 35C-1 1 35D-1 j 35D-2 ] 34A-2 i 34C-1T-2
1 14-U2 j 34C-2 '
: 340-2 1 15-U2
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OUESTION 6.07 (1.50) You are operating at 100 % steady state pouer with containment pressure channel IV ( PE: 934A) failed high 'Jith its associated CTHT PRES Hi-3 alarm annunciator li A technician troubleshooting the trip bistables inadvertently
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de-energines the instrument power tn the input relay bay for containment pressure channel I Will a Containment Depressurization Actuation occur? E0.51 WHY or WHY NOT? [1.01 QUESTION 6.08 (2.00) You are operating at 50 % power s l o ta l y raising power with rods in automati Controlling first stage turbine impulse pressure falls 10 What is the initial direction of rod motion ? CO.5] l What plant condition is sensed by the instrumentation l that vil1 cause the initial rod motion? E1.0] What is the m a >: i m u m speed of initial rod motion (steps per minute) ? CO.5] OUESTION 6.09 (2.00) What is the function of the motor-driven a v::l l i a r y feedwater pump manual start block signal? E1.01 L 1, t o ystem line up and one setLch contIguration for ohich a 109-109 s t e a generator ucter level oculd not ' l cause an au toma tic av::iliary feeduater pump start? [1.0] l (***** CATECopi 06 CONTINUED ON NEXT PACE *****) i
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4 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 9 l __.__.______ ___________________________________________ OUESTION 6.10 (2.50) What are two (2) conditions that require the use of the , high level easte drain header instead of the preferred lo 'W ( [1.03 r&
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Ar T.F f) The Containment Drain Sump inventorp.(da1FNJs requiredw to be j monitored periodically to verify t, h a t tne RCS leak rate is ' within limst (TRUE/ FALSE) [0.53 How can Unidentified RCS Leakage rate be estimated? [1.03 l l QUESTION 4.11 (2.00) 1 Describe an IR detector response if the circuitrv is l overcompensated during a reactor startu E0.753 Deser Ibe an IR detector response if the circuitry is undercompensated during a reactor shutdoen, including any effects on SR detector instrumentatio CO.753 What operator action is required to continue a reactor shutdown if one IR channel has failed high ? Include any applicable setroint CO.53
QUESTION 6.12 (3.00) For the following INITIAL CONDITIONS *
- P R E 55Uf:IZ ER level contrel selector switch in position I/III; i
- PRESSURIZER level in the program band eith 10 7 reactor powe PRESSURIZER level recorder selected to Channel II, !
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- and assuming no operator actionsi l D E S C R I E:E the effect on pressurizer level and 1 too tz' of the todicitions available- to the operator, IF r. . level Channel III falls HIG C1.03 l Channel Il.falled HIGH Instead of Channel II L1.03 Charinel I fatis HIGH t o s t. e a d at II or II E1.01 Treat each siLoation separatelv, i***** CATEGOR( M CUNTINUED ON NEXT PAGE **.***)
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QUESTION 6.13 ( .75) Five (5) components are sequentially loaded automatically onto the diesels during a Containment Depressurination Actuation (CDA) with Loss of Off-Site Pouer that are not loaded automatically durin3 a station blackout with no CDA presen Select these 5 components from the list below and arrange in the proper sequenc RHR Pump Au> Feed F' ump CRDM Cooling Fan Safety Injection Pump Containment Pecire Pump Charging pump A u :. Air Recire Fan Service Water Pump RPCCW Pump Ovench Sprav Pump
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______________ _____ 00ESTION 7.01 (2.00) l
. \ hat 'o (m condi ions m *ist b roc reyefhe A sp riot- SI etua 'on can e res t ? --EG . u a W Under 6!h a t three (3) conditions may an SI be terminated % -C+rtl--
be1 Pcou d ure ? GUESTION 7.02 (2.50)
- Explain how a total loss of"AC power can lead to a RCP seal failur E1.03 List two (2) indications of a failure of a RCP riumber 1 seal? E0.53 What a r er the tuo (2) reasons for stopping all RCPs in case of a massive RCP seal failure if primary l system pressure is less than 1435 psia (1700 psis for ADVERSE C0r4 T AINr1ENT ) ? E1.03 OUESTION 7.03 *1.00:s
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i While attempting a dropped rod recovery on shutdos!n bank A, the lift coil disconnect switches for all rods in the bank are placed in the ROD DISCONNECTED positio What
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i alarm 9111 initiate uhen an attempt is made to withdraw the l affected rod? E0.53 ) Choose the one (li proper action that must take place before re-attempttng -ecoverv if the lift coil disconnect switch a l i g n r..e n t is corrected one hour and 15 minutes after initial rod dror ocents li reovaluate Rod Cluster Control Assembly ' mi sali grement ,
(2) consult roactor engineering 13's verifv SHUTDOWN HARGIN I l i 4 :> rJ O i IF tha rJ RC CO.53
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QUESTION 7.04 (2.00) In the response to inadequate core cooling or degraded core cooling, 6!h a t t6ro (2) functionalh diverse system parameters are checked to verify adequate core cooling has been recovered? [1.03 In the event tt becomes necessary to vent the Reactor Vessel in response to potential void formation, under what l condition is it necessary to determine the ma:<imum allotrable venting time. and uhy is this necessary? [1.03 00ESTION 7.05 (3.00) What is the preferred core cooling method doctrr a loss of ; normal and emergency AC power > Include source of cooling uater snd heat sin [1.0] i l i Tne actions or EUP 35 ECA 0.0 LOSS OF ALL AC POWER l I require letdoirn to be isolate What is the purpose of this sction? CO.7) i ' _ l Select the correct sequence of p o &r e r source attempts ~ 9here trying to r e ?.t o r e power to an AC Emergency Dus! i1) Emergenev DieseI, RSST, NSST (2s RSST. NSST, Emergency Diesel i3) N5ST, Emergency Diesel, RSST l E0.6] Doring e r e c o v e r ', trom lost of all AC power, uhv is it ! necessary to vertrv that Inver ter o is onergized as soon as power is restored to ous 34 ;
[0.71 OUESTIGH 7.06 ( 1.00)
l lhe plant stertop pr at edo es have r ec esit ly been n.o d t f 1 e d to i require s t a r f. i n g a second condensste pump before stritching i over to the tr.o t o r dr ivers feed pump from the turbine driven mato feed p u n. p . Whv tr e3 , this change necessary?
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________________.____ OUESTION 7.07 (2.00) List the only two (2) actions that can be taken locally at the auxiliary shutdoun panel (ASP) to stop an uncontrolled cooldoun during a shutdown from outside the control roo QUESTION 7.08 t'3.00) Answer the following <30estions concerning the Radiological Protection Program and Fuel Haridling Pr ocedures . The Corporate ALARA program includes a total annual exposure guide of __________ mrem gammas and __________ mrem n e u t r o r.s per year (select best ansuer from below). , 000 , 500 , 300 , 500 CO.5] Why is it important to orientate a neu fuel assembly so that the serial number on the top of the upper no:cle block is in the South 9est corner uhen storing the new fuel in the storage pool? E1.03 Assembly end/or evacuation should be considered if personnel vill be exposed to radiation oose rates greater than ________ mrem /br (fill in the blank).
CO.53 Fuel movement a re the fvel storage pool area has been suspended because only one criticality-radiation level monitor is operabl i
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What must be done before continuing fuel movement operations? [1.01 OUESTION 7 . 0 '? <1.50i List 2 options the oper4tnr has u s i rig the ! P r i m a r' y p l a r. t heater procedure to m a i ri t a i n a*given heat-up sate (HUR).
[1.01 What is the maximum primary n *, tem unat up rate alloped by procedure?
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ QUESTION 7.10 (2.00) List four (4) major functions that must be performed by'the operator to recover from a SG tube rupture even C2.03 OUESTION 7.11 (2.00) / List two per E- (2) indications used to verify a reactor trip LO.53 SUBCRITICALITY Critical Safety Function Status Tree is RE Arrange the b e l o ta list of operator actions reo,nired in the Response to rJu c l e a r Power Generation /ATWS procedure i ri the proper sequenc . Initiate immediate boration of RCS rianual Trip Reactor Locally open Reactor Trip & Bypass brkrs l Trip Dus 32B and 32N load center supply brkrs C1.03 l Whv is it necessary to have taso (2) SR neutron flux ! monitors operable during refue1in; operations > E0.53 OUESTION 7.12 (3.00, Define the term A D'J E R S E C OrJ T AINNEN T as used in the E0P If the follo41nj critical E1.03 s, a t e t y functions tvere all ORANGE, ohtch has pr1ority; E1.03 Integrity Heat sink Inventoro 4 Containment, D r i e f l '. e::p l a i n '4hv loop stop salves are not used to respond'tu a c ha l l orego to 3 critical safetv t' unctio E1.03 l
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QUESTION 8.01 (2.00) The concentration of the boric acid solution in the Boric Acid Storage System must be verified once a week in accordance with Technical Specification 4.1. The chemist sampled the boron concentration'on the following schedul (All samples taken at 1200 hours).
Har 1 --- nar 8 --- Har 16 --- Mar 24 --- Mar 31 Explain uhv surveillance time interval requirements WERE or WERE NOT exceeded on har 1 [1.0] Explain why surveillance time interval requirements WERE or WERE NOT exceeded on dar 2 [1.01 QUESTION 8.02 (1.00) Which of the following statements is the correct basis for the limits on Axial Flur Deviation (AFD) ? ~ a. To keep .wion redistribution during slou plant thermal power increases within the envelop of peaking factors 9hich may be reached on subsequent return to Icuer power leve b. To keep nenon redistribution dur ing rapid ,>lant thermal power increases within the envelop of peaking factors which may be reached on subsequent return to lower pouer level, c. fo keep ,<enon redistribution during slow plant thermal power reduction uithin the envelop of peaking factors 9hich may ba reached on subsequent return to rated power leve To keep ,<enon redistribution during rapid plarit thermal poner reduction within the envelop of peaking factors which may be reached on subsequent return to rated power leve QUESTION 8.03 ( 1.50) What is the technical basis for the roquirement to reduce lavg to less than 500 degrees uhen specific activity limits o r,
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GUESTION e.04 (2.00) What are TWO of the THREE means of protection taken to prevent a low temperature over pressurization accident in MODE 5, in accordance with Technical Specifications? OUESTION
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8.0S (3.00) natch the lettered technical specifications with the best description (numbered) of their purpose; Limiting Condition for Operatio El.01 Limiting Safety System Setting E1.01 _ Safetv Limit [1.01 As long as automatic protection _ occurs prior to exceeding this specification, then the abnormal condition uill be corrected prior to exceeding any limit . The integrity of the physical barriers which guard against the uncontrolled release of radioactivity is protected as long as this specification is not violate . This srecifiestion indicates the lowest functional capability or performance level of equipment required for Ssfe operation of the facilit QUESTION 9.06 (2.00) Hoe many members are required on the fire brigade per Technical Specifications. Section 6? i , E0.5] Who may NOT be included as members of che brigade? E1.01 Who is r esponsible to function as the fire brigade leader? CO.51-(***** CATEGORY 08 CONTINUED ON'NEXT PACE *****)
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___________________ ._________________ ___________________ OUESTION 0.07 (2.00) The plant is operating at 75% power and the latest leak rate data shows: spm - Total leakage gem - Leakage due to a leaking charging pump relier vaive gpm - Leakage into the Primary Drains Transfer Tank gpm - Leakage th.ovgh 3-RHS-dV8701C, RCS Loop 1, i HOT LEG to RHR __ 0.0 gpm - Total primary to secondary leakage 4.2 3pm - Leakage past RCP seals What l i m 2 +. 3 , if any, have been exceeded? OUESTION 6.08 (2.50) Lased on IS, should each of the follouing requests be granted? Justify your decisio Lonsider each request separatel A request to conduct an approved modification on Power Range deut'ron Flo: Channel N44, uhich vill require deener 31:ing the detector. Over Pouer Delta T Channel II is trippe Oporational node CO.53 A request to commence neating i_sp the Reactor Plant from node 5 COLD SHUTDOWN eith all code pressuriner safety valves inoperabl CO.53 A request to replace the air start solenoid valves on Unit 3 Emer gency triesel Generator number 2 while Unit 3 Emergency Diesel Generator number 1 is 00 Operational node [0 53 An approved cork order to replace the seals on both air lock door s. one at a tim Oparational n o d .' CO.51 A maintenance request to simultaneously replace both T. rains of HEPA filters in the Auxiliary Building ventilation exhaust system on Unit Operational rio d e EO.5]
(***** CA1EGORi 08 CONTINUED ON HEXT PAGE *****) . .
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4 ACHINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 18
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00ESTION 8.09 (2.00) Technical Specification 3.4.4 states 'All pover-operated relief valves (PORV's) and their associated block valves shall be OPERADLE*. For the following situations state what actions are required to be taken within an hoor if operations at pouer are to continue, One or more PORV's inoperable because of e:< c e s s i vr seat leakage e C1.03 BOTH PORV's inoperable due to sticking valve stems
[1.03 QUESTION 8.10 (3.00) If containment integrity is not met at rtP3 in Modes 1 or 2 then uhat 2 optior s does the SRO have for corroctive action? Time perlods are not require E1.O] In what rt o de or Modes is containment integrity NOT required? EO.5] For each of the follouing cituationsr indicate (ES if containment 2ntegrity is preserved and NO if it is violate If containment integrity is violated. e::p l a i n what would be necessary to restore it to normal, fl.5]
1i The coulpment access hatch is properly closed du :ng refueling operation nalntenance is repairing a erimped closed penetration pressorination supp1v line to the doo ) The outsida door of the personnel air lock is properly closed. The ins 2de door is cedged open with a Zu4 to allow for a detail'd inspection of the knife idoor) seal ) The outside containment porgo exhaust valve 3 H'J U- V 5 has the operator r e m o v e ri uith the valve open, i * w r* * CATEGOR( 08 CorJTINUED ON NEXT PAGE *****)
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I ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 19
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i 1' OUESTION 8.11 (1.00) i To 30 from COLD SHUTDOWN to HOT SHUTDOWN (Operational -- _ ,
'
{- MODE 5 to MODE 4), what seneral condition must be satisfied?
' '
O(ESTION 8.12 (3.00) . p j Classify each of the follo6 ting occurances using the
'
State of.Colnecticut Classification Scheme
' (EPIP FORM 4701-5) provide Identify t! hat factor (s)
l ) uere used to determine the final classificatio a. A small plane crdshes on the river shore 6tithin the
[
j plant perimeter about a\ half mile East of the plant.
.
; The radioactive medical isotopes the plane 6tas carrying >
cannot be locate / }
<
,
5 While trorking trith the polar crane in'the containment during l Mode 5, the brake on the hook fails, dropping the hook onto '
, a piping ru The sample line from'the pressuriner liquid ; ; space ruptures and fuel drop accident' area monitor 3RMS-RE41 : '
reading increases to the alarm setroint of 15 mR/h7
,
/
' An accident in the containment results in an injury _to a worke , The 6?orker is in contaminated anti-C clothing and has' broken his-j le Due to the accident the worker receives 2 Rem.of radiation ,
Previously he had accumulated 500 mrem for the q ua'r t e r and yea i j I Load follouing operations at~100% poner result in an. Iodine
spike of 350 ve/sm (> 5 :: TS allotfable). A main steam line ruptures , j inside con t aT nnien t . C on t a l o nien t spray is initiated. .During the [
: transient'five (5) atmospheric relief valves on the affected steam ! ! ! generator lift when the MSIVs shut and two (2) fail-to resea '
Reactor to secondary leakage through the affected SG tras s f , 500 gpd prior to the even j > i i \ . . - . - . - - J ( .._ _
- - :
(**xas END OF CATEGORY 08 sxf**) , ., * * s e. m e: m * s * * * *. END OF EXAMINATION ************x**) - )
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APPROVAL: A v -
#
y I4/ DATE: N / U SORC HTG. NO.: /7'/f - STATE OF CONNECTICUT INCIDENT CLASSIFICATION SCHEME Incident Class / Incident Description Protective Actions / Posture Code Emergency Actions l GOLF Radioactive material * Limit spread contam-transportation acciden ination and initiate clean u FOX Lost radioactive material " Assist in source in excess of Title 10, recovery CFR30.71, Schedule B ei Quantitie . ECHO Minor event of general None l interest but no public ( ' hazard with no radioactive j release DELTA-ONE Incideni with NO unplanned No protective action (unusual even.t) radioactive releas required'for publi .
./ .
DELTA-TWO Incident with no currently No protective action (unusual event) existing public hazard but required for publi WITH unplanned radiological Corporate and Station releases such that site Staff should remain on boundary plume doses are less standby for performance than 0.005 REM to the whole of dose calculation body and/or less than 0.025 REM to the thyroid from plume exposure pathway . EPIP Forn 4701-5 Rev. O Page 1 of 3 _ _ _ _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _
- . . . .
Incident Description Protective Actions / Incident Class /
*
Emergency Actions Posture Code Incident which has a potential Station and Corporate CHARLIE-ONE for projected site boundary will activate Emergency (Alert) plume doses or has a radio- Response Facilitie active release with between State and local will 0.005 and 0.05 REM to the standby for Key Staf whole body and between 0.025 If appropriate they and 0.25 REM to the thyroi will actuate Emergency Staff /EOC, and monitor food / water / milk. Bring EBS to Standby Statu ,, Incident which has a potential Station and Corporate CHARLIE-TVO will Activate Emergency (Site Area for or has a radioaitive release with projected site Response Facilitie (, Emergency)
~
boundary plume doses of 0.05 State and local will to 1.0 REM to the whole body activate Emergency Staff / and/or 0.25 to 5.0 REM to the EOC. Monitor food / water / thyroi milk. Consider placing milk. animals on stored
, ,
feed. Alert EBS.
~' D Activate EBS and public
. warning if necessar BRAVO Incident which has a potential Station and Corporate (General Emergency for or has radioactive releases will Activate Emergency with projected site boundary Response Facilitie Without Containment plume doses of 1.0 to State and local will Breach) .
EPIP Fors 4701-5 Rev. O Page 2 of 3
.
MAY 21 E64
. . - , , , , -,
w . - ~ . . -,, , - - - , - ,
_ _ _ _ _ _ _ _ ,
* . - . . ' . . . \
t Incident Class / Incident Description Protective Actions / Posture Code Emergency Actions REM to the whole body and/or activate Emergency 5.0 to 25.0 REM to the Staff /EO Control thyroi food / water / mil Immediate take shelter / access control for 2-mile radius and 5-miles downwin Extend to 10 miles downwind if necessar Evacuate
- 2-mile radius if not ,
constrained. Alert EBS. Activate EBS and public warning as appropriat l
.-
ALPHA Incident which has a potential Station and Corporate (General Emergency for or has radioactive releases will Activate Emergency Response Facilitie ~ With Containment with projected site boundary Breach) plume doses of greater than State and local will 5.0 REM whole body and/or 25 activate, Emergency Staff /
~~ REM to the thyroi EOC. Control food / water /
milk. Immediate take shelter / access contrcl for 2-mile radius and 10 miles downwin Evacuate 2-mile radius and 5 miles downwind if not constraine Assess need for additional evacuation. Alert EB Activate EBS and Public warning as appropriat . EPIP Form 4701-5
-
.. '_ Rev. 0 ( Page 3 of 3 W_ ..p; _
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- MAY 21 1984 A
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - , / * . , Equations j
p= 1 + _S Q = Mah tK,7f I +1t e a Q = mcpaT SUR = 26 p t* + (6 p) t Q = UAat ,
,
C3 (1-K y ) = C II-K ) 2 2 h = KV2 p = po 10 sur (t) ~ ii = 3.14 p = po ,t/t e = 2.72 SUR = 26.06 CR = 5 t 1-K
.: -- ~ eff
** P--
1 = CR c-S __ S U Ps = M (_A P r h ) ~~ M r CR
{sc t - {- 1 = 0.693
, seconds 7= g ,
g-p
. -
th g, = 10 e eR Conversions I curie = 3.7 x 10"dps I kg = 2.21bs 1 gal = 3.78 liters I gm/cm' = 62.4 lbs/ft' 1 in = 2.54 co I ft' = 7.48 gal 1 yr = 2.15 x 10'.se I gal. = 8.3453 lb . 1 MW = 3.41 X 10' BTy/HR
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THEORY OF-NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 20 _ __ ______________________________________ ______________
-ANSWERS -- HILLSTONE 3 -86/03/25-HANNON, ANSWER 5.01 (1.50) GREATER THA E0.253 As fuel burns up, flux must increase to maintain the same power leve a,y{ E0.75]
, Periodic calibration ees- j gain adjusts 2 je r e ' c r :m o .
.[alorimetricsare performed,nt . rious pouer
' lae-' E0.5] - REFERENCE HP3 System Description Topic 6 Lesson 4 Excore Nuclear Instrumentation MP3 Reactor Theory RT- ANSHER 5.02 (1.50) i Reflux boiling occurs when steam exits the core and is condensed in the SG tubes, with the resultin3 condensate returning to the core via the hot les to repeat the-cycl E1.0] This type of cooling occurs with a voided core when no reactor coolant pumps are runnin E0.53 REFERENCE HP3 Mitigating Core Damage - Loss'of Coolant Accident an Post Accident Cooling. page 1.12 1.31 HP3 Heat Transfer Thermodynamics and Fluid Flow Fundamer tals Section III Part B, Chapter 3
.
ANSWER 5.03 (2.00) (1) Acceptable nuclear peaking factor [0.53 (2)-Adequate shutdoun margi [0.51
(3) Dound rod ejection acetdent analysis assumption CO.53 {4}
E0.53 (The role of thumb is 4 or 5 ocublings of the original count rate).
REFERErlCE MP3 TS, 8 3/4 1-3; HP3 Reactor Theory and Orecating Characteristics, Topic: Subcritical hultiplication. page RT- l
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% THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND
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ANSWERS -- rtILLSTONE 3 ~86/03/25-HANNON, ANSWER 5.04 (1.00) EOL E0.253 Because a louer E:e t a Dar at EOL results in a p e'ter // /6 #Ed 304 red 7 :_ , e . 1 sul, ad EO .53
'o . FALSE EO.253 REFERENCE nP3 Reactor Theory and Operating Characteristics, Topic!
Prompt a rid Delayed Neutron Fractions. Page RT- ANSWER 5.05 (3.00) t-OWEt;- M/ G H E g Re,d rW Weaso W wp%dca AL v>oald pe nue5W4weUhC-
roa we Thrmcorn ,c i ass--wrew,1 = t c - ; e a c t-witv uwo#4+or%t m u 5 3 '.une .He Sant E 0. 53 BOL- ilut_ f. o neik
?
la rp r--b.44--iwe-; f a c t u a l criticality at EOL g([ften uovld occur assumed. r no - sooner /l. nan predicted.g if DOL conditions vere CO.53
- - t-OMEtc T 6"=N EO.53 Previous RCD accounts for more ::enon poison than exists in the actual startup; thus the ECP uill indicate more rod motian than is necessary to achieve criticalit E0.53 SAhE CO.53 Insignificent temperatiire chang .53 REFERENCE ciP 3 Reactor Theorv Section, 15. 16, and 17 l
l ANGWER 5.06 (1.00) H.> a d lossec are loss in the 3 loor cont'tgurstio E1.03 REFERENCE nP5 Hoat T r a n s t s- r Thermodynamics ind flesid Floo Section III Fart 1: . Chapt r 1, pa3e 326 l l
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. . THEORY OF NUCLEAR. POWER PLANT OPERATION, FLUIDS,~AND PAGE 22
' ----- ---- ------------------------------------- . --------------
ANSWERS -- MILLSTONE 3 ~86/03/25-HANNON, ANSWER 5.07 (2.50)
. BOTTOM of the cor CO.253 The critical heat flux is inversely related to reactor coolant quality, which is essentially nil at the bottom of the core (before any steam bobbles have started to form on the heat transfer surface) CO.75] . INCREASE- DECREASE DECREASE CO.5 each] [1.53 REFERENCE MP3 Heat Transfer Thermodynamics and Fluid Flow Fundamentals
~
Figure 4-9 page 22 ANSWER 5.08 (3.00) Power decreases initially due to the boron addition.. E0.5] The primary to secondary mismatch causes Tave to decrease.- E0.53 The decrease in Tave' inserts positive reactivity and
!
restores reactor pouer to a slightly louer than-or the same as initial power level. E0.5] E1.5] Tave does not change due to the boration. [0.53 (Tave is determined by the amount of pump heat and the steam dump setting.), After the initial transient, pouer. decreases at a negative!I/3 DPH! rate to the multiplied source-level. E1.03 1 E1.53 REFERENCE MP3 Transient and Accident Analysis Chapter.4.10
. . .
. . . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 23
_____ __ _____ ______________________________________ ______________ ANSWErsS -- riILLS TONE 3 -06/03/25-HANNON, e ANSWER 5.09 (3.00) Turbine power - STAYS CONSTANT at 25 % pouer E0.253 Tavq- (affected loop) - INCREASES (final value equal to Th) E o'1 > Tave (non-affected loops) - DECREASES E0.75] S/G pressure (affected loop) - INCREASES CO.753 S/G pressure (non-affected loops) - DECREASES E0.753 Yes E0.25]
(in affected loop ehen pressure exceeds 1135 psig).
REFERENCE Steam Tables ~~ nP3 System Description, Topic 1, Lesson 3, page 17 Transient & Accident A ri a l y s i s . Chapter 4, Page 4.11 ANSWER 5.10 t2.50) S/G heat transfer = 0 = UA(Tavg - Tstm) [1.03 0, a id Tstm remain constanti A1(Tavgi - Tstm) = A2(Tavg2 - Tstml Given. A2 = Al From Steam Tabies: inat for ~00 psia = 515 F E1.01
A1(567 - 515) = 0.0A1(Tav32 - 515> Tavg2 = 580 F (13 degree increase) E0.53 REFERENCE ciP3 Heat Transfor ihermodvnamics and fluid Flov Fundamentals Section 11, P a r t. Ch J, t o r & i i
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. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 24 ----- -- - -------------------------------------- -------------
ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, ANSWER 5.11 (2.00) HIGHER E0.253 steam dump pressure setting decrease causes RCS temperature to decreas MTC and FTC both add positive reactivity to increase power. CO.253 E0.53 THE SAME E0./53 the steam dump system 9ill compensate for steam leak by shutting valves to maintain demanded steam GPnerator pressure. CO.25] EO.53 THE SAHE CO.253 the negative reactivity eill be matched by the positive reactivitv added by MTC and FTC. causing no change i Pouer. CO.25] E0.53 gr onu vg v' w. d close wn h.l flos error
'
Lo w e K d.-H W ttE+ EO.253 RG6-tempeectvre tri-lHewea sc d :.c tm drn-trorm 1 c sel> le al e r<' fMeeter ver+,-Ndded to the_denam g,rnr=+or a Md-i-r+pe-t-+rt-ve r n c t i r @eemrey-po* r -te-ir.cr+a-t. . EO.253 , E0.53 Rr 4r;f s l\ c u.vr dea a.uhrd (csel decre m ( g 30 % "o n L/g ,de tc dvs REFERENCE dP3 Transtent and Accident Analysis Section 3 Normal b %t 5 6) . Transient Analysis
[q ?.? N$$ f 3 sL P lb b_ i h A $ S (a Ld LC $
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. . THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 25 ---- -------------------------------------- --_--__-_-_-_-
ANSWERS -- tiILLSTONE 3 -86/03/25-HANNON, ANSWER 5.12 (2.00) sur(t) F = Po(10) where t given in minutes CO.43-7 -0 (1.75)sur 5x 10 /2 x 10 = (10) CO.23 log (25)/1.75 = sur = .8 DPn CO.23 compute the reactivity represented by the stable SUR* rho = beta eff/ 1 + lambda bar( 26.06/ cur) CO.53
= 0.005/ 1 + 0.03 (26.06 / 0.5) CO.43 = 9o.7 pcm (0.13 t Steps = 96.7 pcm/ 5 pcm/ ster =
19 steps; I_L Sh, CO.23 REFERENCE rip 3 Transient and Accident Analysis Section 3 Normal Transient Analysis MP3 Reactor Theory RT-1 N E'3 Topic 6 Lesson 4 E::c o r e Nuclear Instrumentation
.
% w m ,_
. . . PLANT SYSTEMS DESIGNr CONTROL, AND INSTRUMENTATION PAGE 26
______________________________________________________ ANSWERS -- MILLSTONE 3 -06/03/25-HANNON, ... ANSWER 6.01 (1.00) -
-
Steam venting out the break occurs sooner E0.25] for a hot leg break than for a cold leg break because of the loop seal between the.S/G and RCP E0.253. Thus, inject'fon flou is more effective in the hot leg break because injection flou exceeds break flow when break fine shifts + from 2 phase to all steam E0.25 Therefore, less mass is lost from the system E0.25 E1.03 REFERENCE nitigating Core Damage LOSS OF COOLANT ACCIDENT AND POSTACCIDENT COOLING page 1.32 ANSWER 6.Q2 (3.00) (ES. [ 0 . 5'0] The reactor will trip if 2 out of 3 circuits CO.25] in any 1 out of 4 loops CO.253 sense 90% flow EO.253 with power > 39% [0.253 (P-8). E1.53 NO. [0.53 The reactor will not trip because the requir ed logic is 2 out of 3 circuits CO.253 sensing 90% flee C0.25] in 2 out of 4 loops E0.25] eith power betueen 10% (P-7) and 39% E0.25] (P-8). E1.53 REFERENCE MP3 Simulator dalfunction Cause & Effects Document - ED03 * hP3 NSSS Topic 7 RPSAS pages 54 and 66
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- . . . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 27 ------------------------------------------------------
ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, AHSWER 6.03 (2.00) As power decreases, steam flow detector delta-P decreases, . and steam generator pressure increase CO.43 Since steam flow uses a square root extractor corrected for ' steam pressure, and steam pressure is stuck at a lower value ' than actual, indicated steam flow will be lower than actual flo CO.43
'The steam-feed flow error signal will tend to close the FWR LO.43 As the level decreases, the level signal will open the FWR .43 Eventually, the level error will cancel the flow error and steam floe will equal feed flow at a lower leve E0.43 C2.03 REFERENCE HP3 NSSS Toric 6 IRC Systems Lesson 9 Steam Generator Water Level Contr ol ANSWER 6.04 (2.00) Ye CO.25]
With hSIV shut, affected SG will equalize with primary pressur First valve set at 1185 psi CO.753 SG would backfeed primary due to Psat being higher than primary (Psat fcr 570 F = approximately 1230 psia). C1.01 REFERENCE HP3 Transient and Accident Analysts Section 10 Stean. Generator Tube Ruptt. e ANSWER 6.05 e1.00) g 7g,m Estimated leakage would be the 4em-eA one centrifugal chargingPumphlE ' a - n ... w . - w v w ~ ., a t - ~
- 'i9fy gpm (*i50 5 C+ g p m ) [0,153 900 f 5'O C1.0]
E rfimA is made m am;mt W p d nSo x (pE :
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; . PLANT SYSTEMS DESIGN, CONTROL, AND. INSTRUMENTATION- PAGE - 28 ------------------------------------------------------
- ._ ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, J.
i REFERENCE a HP3 NSSS Topic 3 Lesson 4 ECCS ! i ANSWER 6.06 (2.25' ' A SSo minq $SS T* is,aMad PLANT CONDITION.------>> . ,
,
OREAKER DESIGNATION
\\// \/
15G-3U 0 C 0 35A-1 C C 0
, c e. _ , .- .- n n I 35B-2 0 0 0
{ 35C-1 C C 0 j 35D-1 C C 0
'
350-2 0 0 0 34A-2 C C 0 34C-1T-2 'C C O ) 14-U2 0 0' O 34C-2 0 0 0 340-2 0 0 0 15-U2 0 0 0 i 340-1T-2 C C D i 34B-2 C C 0 E2.25] REFERENCE
- r!P3 BOP Introduction to Electrical Systems j
hP3 hain One Line/ Phasing Diagram Power Distr,ibution System
Ccmposite S a W DWG NO. 12179-EE-1A-6 li lh93 SOf blecAriutA Diitr:bJHon 5ysten1 4*Ib0 VO $ 5 Sy 's ftm
! ANSWER 6.07 (1.50)
rJ 0
[0.5]
j The input-relay must,energire to actuate for an unsafe ) condition (to avoid inadvertent spray actuation in the event , of a loss of instrument pouer). E1.03
, REFERENCE .
] hP3 HSSS Topic 7 RPSAS p to
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______________________________________________________ ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, J.
. ANSWER 6.08 (2.00) Rods drive i [0.53 Sudden reactor / turbine mismatc E1.03 j steps per minute CO.53 REFERENCE
HP3 System Description Volume 4 Rod Control
ANSWER 6.09 (2.00)
- Prevent starting the motor-driven av
- <iliary f eed&!ater pumps from the main control board while the emergency generator load sequencer is operatin E 1. 0 3 - SWITCH CONFIGURATIONS:
(1) The TRANSFER stritch in LOCAL and the STOP-AUTO-START *~
switch on the switchgear bus in STOP i (2) The TRANSFER stritch in REMOTE and the i STOP-AUTO-START switch on the control board in STOP ) (3) The TRANSFER s6! itch in REMOTE and the STOP-AUTO-START switch on the control board in PULL-TO-LOCK (f) 5ejvencer te;tlaedg Cany one 0.53
, SYSTEh LINE UP: ' (1) The affected steam generator RCS-loop _is isolated (2.) A F C Pe * P f* d ;ca E0. 53 - E1.03
, REFERENCE WlJe$ LIO 5ed { M'] M 0-J , _ l J HP3 tJSSS Topic 4 Lesson 2 Au:<iliar y' Feedtra ter System i
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. . . PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 30
_________._____.___________________.__________ __________ ANSWERS -- HILLSTONE 3 -86/03/25-HANNON, ANSWER 6.10 (2.50) High radiation EO.53 To::ic chemicals [0.5] E1.0] TRUE E0.53 chvqing[Lg_fdoWo -f. low />a me rio n i t o r sump level and Pump operatio [1.0] REFERENCE R C.5 Maah eq (cmpai4^ p.v g w M cp R () rF3 BOP Reactor Plant Aerated Drains nP3 P ! ID Radioactive Liquid Waste 6. Aerated Drains Sh 3 of 3 is 4.4.o. AO? 3555 &earjor~ loole d Le d Ad5WER o.11 (2.00) Overcompensation results in a louer than actual reading E0.253 wito indicated level increasing at a higher than actual rate. -64,-243 g i v i n g -a false high startup rate indicatio tim?fr'f- [0553 - E0.753 Undercompensation results in a higher than actual reading. [0.253 and if 5: 10E-11 amps [0.253 will provent the SR detectors trom automatically energining E0.25]. [0.75] The oper ator can manually energine the SR detectors u l tii the SR Block-Reset seitch E0.253 uhen the operable IP
: v.JU, channel drops below { 10E-10 amps (P-6 setpoint)
'
E0.5] REFERENCE nP3 Transient and Accident Analysisr Instrumentation and Eontrol Fa11 ore Analysis pages 5.32 - 33 SPC rK S S iopic 6 t.esson 4 E ::c o r e dI page 18;[9
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. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 31
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ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, ANSWER 6.12 (3.00) Steady state operatio [0.5] High Level alarm and Channel III level indicating HIG [0.5] E1.0] Steady state operatio [0.5] High Level alarm, Channel II level indicating itIGH, and the recorder indicating HIGH (any tuo). E0.5J J1.0] FINAL RESULT: Actual level increases until high level trip (2/3)
(Logic not required) E0.5]
INDICATIONS: High level deviation alarm B a cl- up heaters turn on High error signal reduces charging flow to minimum t.ctual pressuriner level decreases (;niHall ) Lov level alarm y Letdown secures and all heaters turn off (any tuo of above indications) [0.5J E1.01 REFERENCE nP3 Transient and Accident Analysis Section 5, page 5.52
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ANSWER 6.13 c g r g;/13 Pw(ng.75) 3 Quench Spray Pump Safety Injection Pump RHR Pump
& ^ .. r: c c i r c Me--
Containment Recire Pump (0.1 each & 0.25 for proper sequence) E0.75] REFERENCE ct P ~ BOP Introduction to Electrical Systems MP3 nain One Line/ Phasing Diagram Power Distribution System Composite 5 d W DWG rJ 0 . 12179-EE-1A-4
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. ' . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 C6UTRUL'"' ~~~~~~~~~~ - '~'Rd65UL6G5CEL --------------------
ANSWERS -- MILLSTONE J -86/03/25-HANNON, ANSWER 7.01 (2.00) o P- sivei rlactor i s\t r ipped E .4] and 1 ECCS h adq, ar star ed Permi(Q [ (afto a presetxtime del -). . TO- 87 , > K . Adequate RCS subcooling Secondary heit sink ok Ch4-P O b EO J . RCS inventors ok(P2gLjfws)[A43 ' __Lt 2L REFERENCE HP3 NSSS Topic 7 RPSAS
[_LO )
Westinghouse Geners Grcup Emergency Response Guidelines E::e cu t i ve Volume - Generic Issues ANSWER .02 (2.50) (1> Loss of high pressure injection flow from CVCSI ( 2 ') Loss of cooling water flov to the RCP thermal [ barrier cooling system heat e:< changer s ; l t3) Continuous loss of RCS coolant occursi ~
(4) Seal overheating results in degredation and eventual failur [0.25 EACHJ E1.0] (1) RCP HI RANGE LKG FLOW HI a l t r in i (2) RCP LO RANGE LKG FLOW LOW alarm (followed by above alarm);
l i31 An u r.e::p l a i n e d increase in the RCP Seal Water o r ny pri Flov above 5 gPm eith a corresponding RCP Lover } E: ear ing water temperatore approaching 230
[ } (ANY TWO) E0.25 each] E O . "i J e c, h i n t ru i z e mass transfer otit the brea [0.53 Lassen the possibility of forminj two Phade flov in RCS vith resulting poor heat t.r a n a t e r pr oper ties and l potential for i n a rie qu a t.e core coolin [0.5] E1.0]
i i REFERENCE rip 3 rii t i gat ing Cor e Damage; Loss of Ai1 AC Pover, page EOP 35 E-1 Loss of Reactor or Secondarv Coolant i
\ rip 3 NSSS Topic 1 Lesson 2 Reactor Coolaret Pump AOP 3554 RCP TRIP or SEAL FAILURE l g) 4 2. RcP Jeyfe bkyt gew high alem l (5) A <t More% W A dpM prp rd MsA s yy fn~' %) y nra w & ud f Haw Reada o)
e) & w - es a y ea a w:en n l
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Cf) An & crea k _ Mh Flaw
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l PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 33 ________________________________________________ ) RADIOLOGICAL CONTROL
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ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, l l ANSWER 7.03 (1.00) ROD CONTROL URGENT FAILURE E0.5] (2) [0.53 I REFERENCE l AOP 3552 dALFUNCTION OF THE ROD DRIVE SYSTEM. DROPPED ROD j Technical Specification 3.1. nP3 NSSS Topic 6 Lesson 2 Rod Control l l Ar4 SW ER 7.04 ( 2 . 0 0 's (1) RVLh5 plenum l e v e l -E v5t ( G RE A T E R THAN OR EQUAL TO 19%)
'AND i (2) (at least teo) RCS hot les temperatures f0757 ~ I (LESS THAN 350 degrees F) OR E1.03 (3) cvxE ya;F TCs IF PRT not available, or PRT rupture disc fails while venting the Reactor Vessel to the PRT (either one)
l EO.5] l haintain Containment Hydrogen concentration LESS THAN _ 3% E0.53 E1.0] REFERENCE l EOP 35 FR-C.1 PESPONSE TO INADEGUATE CORE COOLING i EOP 35 FR-C.2 RESP 0rJSE TO DEGRADED CORE COOLING EOP 25 Fh-E.3 R E S P 0 tJ S E TO VOIDS EN REACTOR VESSEL n2tigating Core Damage LOSS OF COOLANT ACCIDENT AND POS T AC C [DErJT COOLING l
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. * . . PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 -- - - - - - - - - - - ~ ~ - - - - - - - - -
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ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, St1FbTIF.5 (if no opv h sdim}oY ANSWER 7.05 (3.00) The stejm driven AFW pump provides feed [0.51 and the s e c o n d a r y.J'O R V s dump steam to atmosphere [ 0.25 ] to promote natural circulat N [1.01 on(9.th.li$- w reS 4Dn0)
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Minimize RCS inventdry loss (and shrinkage due to cooldown). [0.7] (1) [0.63 Restore power to the plant process compute [0.73 REFERErJCE EOP 35 ECA 0.0 LOSS OF ALL AC POWER OF 3345A 120 Volt Non-Vital Instrument AC M iT16 e ri t!6 Cc AE D #i m 4 6 E CH4F7EA + ? +. S <r. i t ANSWER .06 (1.00) To avoid unplanned reactor trips due to lou-low-SC leveli (C pu2A bg Lof f op mS pec4{wxta p<wf f a ct; )
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C1.0] R EF ERE riC E Report of unplanned reactor-trip at 1042 hours on February 12, 1986; the 4th such trip at MP3 since initial criticalit ~332_i E c e d W u Y L'i ANSWER 7.07 (2.00i close SC atmospherte dumps -01 rob ,
' close atmospheric dump bypass valves _c i. or,L REFERENCE the, nie AN plouj tr 'g, y _ ")
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EOP 3503 5HUf0090 OUTSIDE CONTROL POOn
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% PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
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__..__________ ______ ANSWERS -- rtILLSTONE 3 -86/03/25-HANNON, ANSWER 7.08 (3.00) E0.53 This orientation must be maintained to ensure proper positioning of the fuel in the core and ease of SNM accountab112t E1.03 C-n,%.erht inopera ble WiNN^+ b Ser vi'E; o r CO 53 Provide an approprlate portable continuous monitor uith the required alarm setpoin { gjd he r pae}. [1.0] REFERENCE SHP 4902 Rev. 8 E:< t e r n a l Radiation E :< p o s e.i r e Control and Dosimetrv Issue page 10 EPIP 4010A Shtit Supervisor EPIP 4001 Director of Station Emergency Operations OP 3211A New Fuel Assembly Receipt and Inspection p 22 OP 3210D T5 Table 3 3-6 Action 28 ANSWER 7.09 (1.50's . Start /stop RCPs PZR B/U heaters ON/0FF RHP (Ice control iank t t'O ? E1.03 LOU deqTee? I In d r One hour erloG E0.53
# * Of h0 .QLgQ l h f'd If)
Or w.1 eage n ' i'y
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1ochnirai Opoc1iicatton 3. 4 Y 1 TS 3. 9'-Z a b a tt a ) l ( f2DWlbMk.b '
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' PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
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ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, '
)
ANSWER 7.10 (2.00) l i Diagnose and identify faulted SG ' I Isolate affected SG i Cooldoen primary system Depressurire primary system Terminate SI (any 4) E2.03 l REFERENCE ! HP3 Transient and Accident Analysis STEAh GENERATOR TUBE RUPTURE E0P 35 E-0 '
! AOP 3556 STEAM GENERATOR TUBE LEAK i
E0P 35 E-3 STEAM GENERATOR TUBE RUPTURE
I j ANSWER 7.11 (2.00)
' . Digital RPI at 2ERO Reactor trir and bypass breakers OPEN '
- DECREASING neutron f l u >. o n NIs (cny two* 0.25 each) [0.53 ,4,1,3 , -t
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3 C1.03 i Redundant)fdDCrs monitoring b btlliYcapability(Ar g) . rc giR(1 b E0.53 REFERENCE nP3 Mitigating Core Damage, Anticipated Transients Without
- Trip
] E0P 35 E-0 Reactor Trip or Safety Injection i E0P 35 FR- ; i TS Cases 3/4. l ! t _
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PAGE 37 RAUEULUUIUUL'UUUTRUt ____________________ ANSWERS -- HILLSTONE 3 -86/03/25-HANNON, ANSWER 7.12 (3.00)
* High temperature and/or radiation levels that affect instrument accurac E1.03 E1.03 Loop stop valves were not designed to be used for accident mitigation purposes. (OP 3205 allows shifting from 4 loop to 3 loop configuration in a controlled manner for normal plant operation.eith one loop out of service.)
C1.03 REFERENCE Westinghouse Emergency Response Guidelines, E::e c u t i v e Volume 9& E er e, iW.
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PAGE 38 ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, ANSWER 8.01 (2.00) Interval requirement not exceeded E0.5]. Eight days does not e::ceed 1.25 times the specified interval E0.5]. ' [1.0] Interval requirement exceeded E0.5]. The last 3 conseceitive - intervals exceed 3.25 times the specified interval E0.5 [1.0] REFERENCE TS 4. MP3 Topic 2 CVCS Lesson 2 Reactor nakeur System ANSWER U.02 (1 00) d
[1.03 REFERENCE SC 3/4 2-2 ANSWER 8.03 (1.50)
Prevents a release of activity in event of a SCTR E1.03 because the saturation pressure for 500 degrees is less than atmospheric steam relief valve setroint E0.5 [1.5] REFERENCE iS B 3/4 d-7 l l
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s' ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 39
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ANSWERS -- MILLSTO.iE 3 -86/03/25-HANNON, ANSWER 8.04 (2.00) ' Two PORV's operable with appropriate relief setpoints
- Two RHR suction relief valves (each with a setpoint of 450 PSIG)
.' RCS depressurized (with vent area at least 7 square inches)-
[TWO requiredi 1.0 each]
E2.03 REFERENCE TS RCS Overpressure Protection Systems 3.4. ANSWER 8.05 (3.00) a-3 b-1 E1.03 c-2 E1.03 m cl.o] REFERENCE 10CFR50.36 MP3 TS ANSWER 8.b6 (2.00) E0.53 Brigade shall NOT include the minimum shift cre9-required for safe shutdoun of Unit 3 E0.53 and any personnel required for essential functions during the fire. 00.53 .[1.03 One of the brigade members is designated as the leade E0.53 REFERENCE E0P 3509 Rev 3 Attachment C OP 3256
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6 ADMINISTRATIVE PROCEDURE 3, CONDITIONS, AND LIMITATIONS PAGE 40 . ------------------------ -------------------------------- ANSWERS -- MILLSTONE 3 -86/03/25-HANNON, ANSWER 8.07 (2.00) -5 ~*
!
RCS Pressure Isolation Valve Limits exceede E1.03 UNIDENTIFIED Leakage limits exceede E1.03
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REFERENCE rip 3 P&ID LOW PRESSURE SAFETY INJECTION SH 1 of 3 I Technical Specification 3.4. ANSWER 8.08 (2.50) - * Yes[0.23 no bistable for OPdeltaT from NI's. [0.33 No p[0.23Cuso acWF yes. Pre mim'. 2005 tJa.s not exceeA cannot enter Mode 4 unless applicable LCO's'are me [0.53
[0.33 [0.53 No E0.23 do not intentionally enter action statement for
maintenace unless other DG demonstrated OPERABLE. CO.33 E0.53 1 Yes CO.2[ maintenance and entry into action statement has been appr o ved sWapar tment-Head,. E0 6 EO.53 No E0.23 only if one train at a time is taken out of servic CO.33 CO.53
- REFERENCE l
I Functional Diagram Primary Coolant System Trip Signals sheet 5 T.S. 3.04 3. .6.1.3 ' ACP-0A-2.02C
1 ANSWER 8.09 (2.00)
! Either restore the'P'ORV(s) to operable status E0.53
or close the associated block valve (s) [0. 5 3 - E1.03 ' Either restore the valves to operable status E0.53 or close the block valves AND remove power from.the ' block ualves E0.53 (AND be in HOT STDY within'the next 6 hours and i n HOT SHUTDOWN uithin the next'- 4 hours:'. E1.03 i * l REFERENCE TS 3. , _ e en d9g- W ',W DAceT'994+$mJg$ep
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g ywp (qW4 *gW4%Pt *g-p sq q ve-Y g.-. u--p#, m g g-g ( w F-e+- "6 *T =-7'*
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i ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 41
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ANSWERS -- HILLSTONE 3 -86/03/25-HANNON, i ANSWER 8.10 (3.00) 4 Restore containment integrity (1 hr) CO.53 OR Go to cold shutdown (within the next 36 hours) [0.53 E1.03-b. Mode 5 COLD SHUTDOWN bdd. [p g[so a (g . M86 [0.53 4 ) YES NI AO 00W SU~ lOhdt (A)!O
/Qgg g g C0.53 2) YES M'~ fd6 AIM [ MOL/O E0.53
3) N The valve must be close #1 MI - E0,53
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REFERENCE T. S. Definitions and 3.6. ' ANSWER 8.11 (1.00)
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The full complement of equipment for HODE 4 must be operable, or another wav o f' saying this, LCO's for riODE 4 must be met without reliance on the ' Action Statement E1.03 f --- - - -- - _ : = ~ REFERENCE f _~~'- ~~ TS B 3/4 0-1 p[' $50 (2((Eg W/. m f _ _..__ - s
, {ANSWER 8.12 f . ( 2( /$ , b .
(3.00) GOLF 0.53 'does not effect plant ops, was not initiated by plant, and does not effect local travel pat: erns. [0.253 i
(Convart to FOX if scorce recovered).
00.753
} DELTA-TWO E0.53 due to leak rate EO.253
' CO.753 DELTA-ONE EO.53 due to transport of contaminated i individual. 00.253 E0.753
. CHARLIE-TWO tor higher) [0.53 Pri-See leak rete c o n.b i n e d
, with IODINE Spike E0.253 CO.751
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't E F E R Et1C E I EPIP 4701 4 rtP3 . 80 P Radiation rion t tor i ns Sys tems /
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Cyclo c Uiciency o (NItwrw
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w = ag s = Vo t + 1/2 at E i = :nc
* ,(E = 1/2 mv a = (V7 - Vn )/t A = AN A=Aeg PE = m9n ;
Vf = V, + a t w = e/t x = an2/t1/2 = 0.693/t1/2 l 9 ,yg - t 1/2*#f " b(ElnII*h)3 j ((tifg) + (t3 )] -
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aE = 931 cm I'= I g e'** i _ Q = mCpat Q = UAat I*Ieg Pwe = w :.5
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I = 1, 10'*/ M TVL = 1.3/u P = P 10 sur(t) HVL = -0.693/n
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, SUR = 26.06/T SCR = S/(1 - KeU) , x = S/(1 - X,ffx) CR SUR = 25a/ t' + (8 - o )T CR j (1 - K,g)) = CR Z (1 - keff2)' T = ( t*/a ) + [(s - a )/Eo] M = 1/(1 - X,ff) = CR j/CR, T = U(o - 3) M = (1 - K,gg)/(1 - K,ff)) T = (a - o)/(la) SDM = (1 - K,g)/Ke# o = (K,g-l)/K ,ff = M ,ffA eff t* = 10-5 seconds I = 0.1 seconds-I I o = [( t*/(T K,g)] + [i~ e U / (1 * T)] Idll"Id P = (r4V)/(3 x 1010) I j d) 2 =2 Id22 2 2 r = oN R/hr = (0.5 CE)/d (mecers) R/hr =.6 CE/d2 (feet)
Water Parameters Miscellaneous Conversions j 1 gal. = 8.345 lb I curie = 3.7 x 1010dps 1 ga]. = 3.78 liters 1 kg = 2.21 lba 1 ft3 = 7.48 ga I hp' = 2.54 x 10 3 Stu/hr Density = 62.4 lbm/ft3 1 m, = 3.41 x 106 Stu/hr Density = 1 gm/cm3 lin = 2.54 cm
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Heat of vaporization = 970 Stu/lbe 'F = 9/5'C + 32 Heat of fusion = 144 Stu/lba - 'C = S/9 (*F-32) _ 1 Atm = 14.7 psi = 29.9 in. H BTU = 778 ft-lbf I ft. H O 2
= 0.4335 lbf/i . -- ..
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TTRCbnset?f 3 I COMMENTS ON NRC WRITTEN EXAMS AND ANSWER KEYS REACTOR OPERATOR EXAM 1.02 d No explanation should be required since was given l in problem that flow was lost. Any reason given why this occured would be stri~ctly hypothetical.
! l 1.03 Either b or c should be allowed as correct, if
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I calculate: s Assume EOC conditions:
~4 I.R. at 100 percent power = 5 x 10 at energization.= 5 x 10 .11-amps Beta effective = 0.0054 Stable SUR after trip = 1/3 dpm Reactivity after trip = 7000 pcm surt P=P 10 Beta o
Beta - Rho t= 17.7 minutes Time to energize from typical shutdown curves from Westinghouse and evolution conducted on simulator (see attached curves) indicate approximately~ 15 minutes to source range energize .05 Rod worth is not affected by increase in the flux level in core, as stated in. ke Rod worth for
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power increase may increase, decrease, or remain constant, depending on flux distribution and location of control cod. (Ref. Large Pressurized Water Reactor Core Control by Westinghouse p )
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1.05 Ef fect on rod worth is dependent on location of rod in question relative to the rod move If rod inserted is close to one in question, the moved cod will depress the flux in the region and lesson the worth of rods in close proximity. If rod inserted is far away from the rod in question the flux in _ the region of this questioned rod will increase and-its worth will increas (Ref. Westinghouse Reactor Theory Review Text pgs. I-5.42 and I-5.43).
I 1.05 If a rod is at outer regions of core, but not at , edge, it will be at region of highest local flux and will have highest rod wort Therefore rod , worth at outer region of core can be higher than rod worth at inner region of cor !
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l 1.09 b.2. Also accept system pressure above which the pump will not produce any flo j
[ j ,v, c c' ? O , '. M i Y
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- ko 1.10 Accept RCS flow rate, nuclear enthalpyJr-ise-hot ,
channel (E)m2 heat-flurhottanrieL f actor-(FO(2})~-Ref. T.S. B ases for 3/4.2.2 and 3/4. pg B 3/4 2- Accept peak linear power density (KW/Ft) - Re T.S. Bases for 3/4.2 pg B 3/4 2-1 1.12 Also accept limits potential effects of rod misalignment on associated accident analysis - Re T.S. Bases for 3/4 1.3 pg B 3/4 1- .
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2.02 a. . Wording of the question can result in confusion between RPCCW and TPCCW
. The question is not consistent with the : normal operation of the system in that we do not operate with a RPCCW pump in standby, ready for an auto start. Both RPCCW pumps are running supplying each trains subheaders with the
, containment safety related sub-header split by having all inside containment sub-header cross-tie valves closed. The swing pump (PIC) is prevented from operating by electrical and mechanical interlocks with the other pumps operating. -Ref: NSSS Text RPCCW Chapter Page ; RPCCW OP 3330 . . There is no low discharge pressure auto start feature for RPCCW pumps Ref: NSSS Text RPCCW chapter Pages 9-1 . RPCCW Pumps are not sequenced on a LOP /CDA Signal Ref: BOP Text Sequencer Chapter Figure .04 c. Delete the "RCS Stop Valves Open" from answer ke This has been removed f rom the AFW Start Logics thus not requiring RCS Stop Valves to be Open to produce an AFW Pump Auto Star REF: functional logic diagram Figure 7.2-1 Sheet 7 and 15, Amendment 14 dated July 198 .05 Should also accept reactor vessel leakoff flange Temperature Alarm and Indication.- Ref: P&ID EM 102A-3 and EM 107 A-4, NSSS Text, Reactor Vessel and Internals Chapter PG 8- , - . _ _ - - . . .
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2.06 Correct terminology for RCDT in answer key would be containment drain. transfer tank (CDTT) 2.07 This part should be delected, as there is no alternate suction for the RHR System from the containment. The Containment Recirc System is the only system to take a suction on containment for long term cooling in recirculation mod Ref: NSSS Text RHR chapter, P&ID EM 112 A- .07 The RHR pump cold leg discharge isolation valves are closed, not open, while in Cold Leg Recir Ref: ES-1.3 Rev 1, Change 1 Page 4 Step . 2.08 The normal configuration for the Containment Instrument Air System is not to run the containment instrument air compressers af ter fuel load. The containment instrument air systea is supplied from
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the instrument air system through the containment . _ isolation valve Ref: OP 3332B Objectiv .10 The time delay till the containment recirc system starts has been changed to 660 seconds (11 min).
Ref FR 2.1, Rev 1, Page 6, Item .11 During normal operation, PCV 131 maintains a . constant backpressure on the letdown orifices, not letdown flow control valvs.$. The letdown pressure control valve is used in both manual- and automatic for solid plant pressure control, not exclusively in manual as indicatred in the answer key. Ref: NSSS Text CVCS Chapter Pages 8-13; OP 3201; OP 320 .
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. 3.01 Overpower Delta T does not protect ag ains t a safety limi nV(.o *
3.06 P-14 coincidence is 2/4 levels (Ref: Function Di ag r ams , Rx Trip Sys., Figure FSAR 7.2-1 (7 of 19) 3.07 Must anticipate candidate assumptions as to which PZR level xmtter he is looking'at, since cooldown OP has him use cold-cal level channel (LI-462) and g ,h
Figure 7.1 to monitor actual PZR level. Ref: OP gge b*
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gd 3308, step 5.10, Caution 2, Pg 16. Cold-cal fi
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y d instrument (LI-462) will read lower than actual.
J 3.08 Open - >2200 psia
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1 COPS Arm / Block to " ARM" Close - no AUTO signals to close these valve (Ref: FSAR Funct. Diag. Fig 7.2.1 (19 of 19)
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3.09 a. . psia or 2370 psig (REF: E.0, PLS) psia /1885 psig (REF: E-O, PLS)
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l b. Add PORV block, 2/4 detec., 2200 psia (REF: FSAR Funt. Diag. Fig 7.2-1 (19 of 19) & (6 of
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3.13 Rad element numbers should not be required in key.
I Suggest minor credit for trip of supply and exhaust f ans because their trips are a cascading ef fec , ie: _
- Rad element close dampers 32 A/B i - Damper closure stops exh. Fan -
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- Stop of exh. Fan 4A stops. supply unit HVUlA damper numbers should not be required in ke Candidate must assume release path is thru Au Bld For key to be correct. In addition, several of the process vent monitors available throughout
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the Aux.- Bldg. ,may indicte activity increase as well as HVR-10. . Rad element numbers should not be required in key. REF: LSK 22-lD& IJ, P&ID 148A & 153 ~ 4.01 Key is in error in 2 places for specific number , Key has 35% vice 34% for S/G NR level l 2. Key has 275 gpm vice 525 gpm for FW flow verificatio Answer key is correct, however, questions uses wrong level; level should be 520,000 vice 520, 410 gal .02 Key should also accept that' cunning pump for 5 minutes will equalize boron concentration to that of RWST (thus raising RHR boron concentration if low) ,
-l Per OP 3310A the RHR alignment is from-RWST thru RHR pump and back to RWS No mention is made of the flowpath from CVCS, which the key refers to (REF: OP 3310A, Step 7.2).
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4.05 Par AOP 3552 there are two possible ' answers to this'
'quGstion drptnding on whoro in the pcoceduro you .
refer t _
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. Possible answers are: I 1 , Per the key J r If QPTR > 1.02, do not increase turbine load, l
borate as necessary to minimize TAVG-TREF deviatio Key should also accept that the urgent failure is a
" Power Cabinet Urgent Failure" due to regulation f I
failur i l 4.09 Also acceptable: )1
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1500 & 0.5,(current quarterly permanent dosimetry l l . reading) l Up to a maximum of 250 (REF: SHP-4902) Step 8.1.2.7) l l 4.10 Key is correct; however question uses 585 psig I which is incorrect, should be 660 psig (REF: Op 3201) 4.12 Step 4 of ECA 0.0 says to " Verify AFW flow 525 GPM j per intact S/G" just as the key indicate : Some examiners may answer " Verify AFW Flow >525 GPM TOTAL." This answer should also be acceptable 1
since many examinee's are aware of an error in the i EOP' Attached are the pages from the ERG's that i indicate this step should correspond'to the value l i for AFW flow in the heat sink CSF status tree red
! path. Also attached is a copy of that status tree.
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GENIOR REACTOR OPERATOR EXAM 5.01 Calorimetrics are not done at different' power levels to adjust detectors following initial startup testing.
l 5.04 Question asks "which condition would result in a higher SUR; a cod ejection at BOL or EOL." Answer key says EOL results in a higher peak power leve This is not asked for in the questio A more correct anser would be the smaller B results in a-higher SUR at EOL
5.05 1. Question refers to improper use of rod worth curves yet answer does not even mention the Answer Ke;- discusses B from BOL to EOL which has , no bearing on this questio Rod Worth curves are not based on ,. A more correct answer should compare rod worth , at BOL to EOL and conclude that rod worth increases over life. Since BOL curves are used the reactor will actually go critical below the calculated EC I ' l The answer key discussion does not match the answe Key says criticality would occur sooner than predicted and answer is LOWER however j question asks to compare calculated ECP to I actual. Therefore your answer should be HIGHE Refer: Core characteristic's handout) -
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Wording of question is confusin Examinees may 9 answer LOWER, implying the actual criticality is lower than the ECP which is generally the way the two are compared.
l l 2. Insufficient' information is provided in this questio Examinee needs information about l power level before the 12 hour outage ~ discussed - in the question in order to estimate xenon for this conditio (Attached figure shows xenon ! l vs. time after trip). Essentially the previous RCD could be at any xenon condition depending on the power level before the outag Also, answer key conclusion is wrong based on reasoning provided. Again, like-the previous question the_ answer key reverses the comparison scheme called for in the questio " Actual v ECP", vice "ECP vs. Actual".
. 5.07 a. Key should also accept, Critical Heat Flux is the heat flux necessary for DNB to occur. Since
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temperature at core bottom is lower, and pressure is highest, greater margin to DNB, CHF is at its maximum. Discussion of steam quality should not be necessary for full credi ; l 5.11 d. Answer key is incorrect, if a level detector (controlling channel) fails high, the feed j regulating valve will close until level error equals flow error and steam generator level
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i stabilizes at a lowe'r' 'value. Reactor power will settle out to the same level as befor ' l L . l l
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i 6.05 With RCS pressure at 1600 psig the SI pumpw ould
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i not be injecting-(shut off head for SIH is 1580 >
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PSIG). An estimate of charging- pump flow would be approx. 400 gpm. at 1600# assuming it's 150 gpm at about 2250# and 550 gpm at 660#.
* (REF: MP-3 NSSS Trng. Topic 3 Lesson 4 Pg 45.)
6.06 For plant condition 3 (Sta. Blackout) the 34C-IT-2 and 34D-IT-2 breakers would be open upon Failure system to Fast Transfe In addition the EDG breakers are ope Also, condition 1 has two acceptable answers: RSST or NSS ; (REF: MP-3 Trng. BOP Chap. 4.16 kv.)
! 6.07 Techincally incorrec There is no instrument I power to the input relay bay.
j 6.08 Should accept " Tref less than Tavg causing Terror for rod insertion."
.. i 6.09 Can also block by sequencer test lineup or by AFW SYSTEM valves ' being close l 6.10 Estimates can also be made using RCS leakage computer program (or manual calculation), and/or charging / letdown Flow balanc (REF: EOP-3555.)
r , 6.11 No effect on SUR and indicated level won't increase at higher than actual rat .
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j 6.13 There is no load called " Aux ' Air Recirc ; Pan" - remove from answer ke If this was to mean Ctm Air Recire fans, they do ncs start on CDA.' Add
" charging pumps" to key, since during LOP the
, sequencer does not give them a start signal but it does during CDA. Also note that nothing is loaded on diesels during a blackout,.by definitio It is i not reasonable to expect students to know exact sequence " 7.01 The question asks what conditions must exist By Procedure" to reset a spurious SI actuatio The conditions to terminate SI "Per Procedure" are:
. RCS Subcooling
. Secondary Heat Sink
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. RCS Inventory (PZR Level / Pressure)
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I Credit should also be given for these condition (REF: EOP E-0, Rev 1, Page 10, Step 26) , 7.02 Answer Number 3 is the result of a seal failure, ' not a condition resulting from A loss of all AC that can lead to (produce) A seal failure. This ) answer should be deleated fromt he answer ke f (REF: Mitigating CORE damage, Chapter 4, Page 4.5)
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7.02 b. The following can also be indications of A.RCP #1 Seal failure and should also be accepted for full - credit: .
. #2 RCP Seal leakof f flow high alarm
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. An increase in the af fected pump Seal water supply flo . An increase in the Leakoff Flow Recorder (s).
! l . Immediate response to a major f ailure would be a decrease in PZR level and a PZR Level Deviation Alar i
. An increase in charging flo ' (REF: Dynamic response of'the simulato,r;_ NSSS Text, RCP Chapter; P&ID EM 103A- I !
c. No commen .04 a. Credit should also be given for exit thermocouples
<l200 (or <700 F). There are numerous kickouts in FR C.1 prior to this step that look at CORE Exit
- Thermocouple Temperature along with Plenum Level and Hot Leg Temperature to verify adequate CORE coolin (REF
- FR-C.1, STEP S&6, STEP 15 & 16).
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7.05 The only source of cooling is steam relief out the ,
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S/G safeties due to the loss of instrument air and > the loss of power to the atomospheric bypass valve - on the loss of power. Manual operator action would , . be required to. locally open the atomospheric bypass valves to bleed stea During the time delay to initiate manual dumping of steam via the atmospheric bypass valve, the S/G safeties will lift to remove CORE decay heat. (REF: Mitigating , CORE d amage, Chapter 4, Pages 4.8 - 4.12 7.06 Credit should be given for "To prevent a loss of main feedwater pump suction". The minimum flow . requirements for two feedwater pumps is greater than the capacity of a single condensate pump is why this condition can occur. This is consistent with the stated reason for the procedural chang (REF: OP 3321, REv 0, Change 5, Change 6) 7.07 Credit should also be given for " Controlling AFW flow to the S/G's" which can result in excessive cooldown, especially at the beginning of cycle.
j (REF: E-0, Rev 1, Page 8, Step 20 Response not obtained; ES-0.1, Rev 1, Page 3, Step 20 Response Not Obtained; et . 7.08 The answer (reason for procedure words) is to minimize schedule time for fueling sequence.
, 7.08 Credit should also be given for restoring the
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inoperable instrument to operation to meet the
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r minimum channels operable requiremen The question does not preclude this action as a possible solutio It should be noted that
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students are not required to memorize' Tech. Spec.
i action statement . - - _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ - - _ - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _
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7.09 Could also be " Temperature Control Valve" vise
" Flow Control Valve". (REF: OP 3201, PAGE 22,
! Caution).
b'. Credit should also be given for 60 F/H This is consistent with Precaution 4.3 of OP 3201 which requires pressure and temperature to be maintained
- in accordance with Tech. Spec. Figure 3.4-2 which l is based on a 60 F/Hr heatu (REF
- OP 3201, page 10; Precaution 4.3 and Tech Spec. Figure 3.4-2 Explanation note at top of Figure)
i , 7.11- Other possible answers: 1-
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Required by Tech. Spec i !
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Criticality moniotorin ; 8.02 No answers correc Limit ensures F (Z) envelope
- not exceeded during normal ops or in -event of xenon redistribution following power change The reference given for the key answer does not refer to this, but rather refers to why we allow j operation outside the target band within time
' limits. This prevents xenon redistribution from being excessive and causing envelope of peaking factors to be reache Any answer should be accepted as correct.
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to any further data give .If use other leakages which are given, identified leakage may be excessive. The number given for leakage past RCP
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seals does not define where it goes and could be
- taken to be either part'of controlled leakage or unidentified leakage. The identification of excessive RCS Pressure Isolation Valve Leakage would require memorization of a Surveillance table, which is not required of the operators.
8.08 Also allow yes with explanation that can't enter Mode 4 (i.e. can only heat up to 200 F) Department head can't approve entry into action statement .10 The question requires knowledge not required to be ' memorized by the operator (i.e. action statement of greater than or equal to one hour).
, Containment Integrity, as defined, is not required in Mode 6 either. However, a " modified i containment integrity" covered under Specification I 3/4 9.4 is required during core alterations or movement of fuel in containment. Therefore, Mode 6 should also be accepted as a correct answe c.2 This involves memorizing action statements and surveillance steps in two separate specification , l
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8.12 Not enough information to classif . I b. c. & d. Both questions asked to classify events based i on the given EPIP For The form is wholly inadequate to provide the classificatio Apparently, EPIP Forms 4701-3 and 4701-4 were intended to be given with the exam to classify the events and were accidentally left of These are the forms which the , answers in the key are based on. Since the required forms were not provided, any answer l justified by the candidate should be accepted for full credit. Incidents require EAL Tables to classif i
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3 D GAMMA-COMPENSATED CURVE st 9 eg a lon Current Meter R'eadings (Amperes) 2N v 10 ~4 J 10 -5 h Electrically Adjusted Compensated lon Chamber
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Typical Shutdown Curves 10 -7
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Neutron level decay due to 54 seconds 10 -8
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half life delayed neutron emitter (80 seconds negative period)
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10 - Correctly Compensated 11 _ Over Compensated
- 10 -12 I I I I I I '!I \ l I l 0 2 4 6 8 10 12 /, 14 16 18 20 22 24 i j i in unos i Time Af ter Reactor Shutdown (Minutes), jd ggfl y
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! i ATTACHMENT 4 RESOLUTION OF FACILITY COMMENTS REACTOR OPERATOR EXAM 1.02 d Question was deleted.
! 1.03 Answer key was c. hanged to accept "b" or "c".
1.05 b Answer key was changed to reflect a shifting of neutron flux distribution caused by the doppler effect, changes in moderator temperature and buildings of fission product poison .05 c Either a description or a discussion was accepted 1.05 d Answer. key was changed to accept the statement that rods at the outer regions of the core may be of higher worth than central rod .09 b(2) Answer key was changed to accept " system pressure above which the pumps will not produce flow."
1.10 a Answer key was changed to accept RCS flow, FQ(Z), and Figure 2.1-1 pa ramete rs . 1.10 b "KW/ft" was added to the answer ke .12 a(2) Answer key was modified to accept " limit potential effects of rods misalignment on associated accident analysis."
1.12 b Answer key was modified to accept " higher delta T input to RIL - Computer."
2.02 a Answer key was modified by deleting " Low pumps discharge pressure (80 psig)" and adding " Normal CC W pump not running." ' 2.04 c "Stop valves open" was deleted from the answer ke .05 " Alarm indication" was added to the answer key but was not required . for full credi l 2,06 b '
"RCDT" was changed to "CJTT" in the answer ke .07 b Question was delete .07 d Answer key for CL RECIRC was changed from "open" to " closed."
2.08 b Question was delete .10 a Answer key was changed from "5 minutes" to "11 minutes."
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2 Attachment 4 ! 2.11 Answer key was changed to reflect control of pressure on letdown orifices and manual and automatic control of the letdown pressure l control valve .01 a No change was made to the answer ke Overpower delta temperature ! protects the core from overpower conditions as stated in RPSAS l Lesson Plan page 4 .06 a(1) Answer key was changed from "2/3" to "2/4" coincidence, j 3.07 Candidates were directed to assume the " hot channel" level l indication during the examinatio .08 Answcr key was change to reflect current facility conditions by deleting the answer for closin .09 a Answer key was changed to reflect current facility setpoint .09 b "PORV block at 2200 psia" was added to the answer ke .13 a The element number was not required for full credi .13 b Answer key was restructured to include the cascading effect of the exhaust fan trip .13 c Additional answers were accepted if they could be supported by facility documentatio .01 Plant specific numbers were used in place of generic values.
l l 4 02 a Answer key was modified to accept alternate answer of " equalize l boron concentration in RCS with RWST."
l l 4.02 b Answer key was change to require the flow path from the RWST thru j the RHR pump and back to the RWST.
l ' 4.05 a Answer key was changed to include "or boration as necessary." l
4.05 b Answer key was modified to also accept " Urgent Failure due to regulator failure."
l 4.09 b Answer key was changed to require "2500 mr/qtr and 1500 + 0.5 current
- quarterly permanent dosimetry reading.
! i 4.10 b The value of 585 psig is in accordance with the reference material provided for examination preparatio .12 Answer was key modified to accept either total flow or intact 5/G flow.
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3 Attachment 4 SENIOR REACTOR OPERATOR EXAM 5.01 b "At various power levels" was deleted from the answer ke .04 a
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Answer key was changed to " higher SUR at EOL."
5.05 Answer key was changed to " higher."
2.. Answer key was changed to " higher." Candidates were directed to assume the same 60-day power history that was established in the stem statemen .07 a The statement concerning the formation of steam bubbles was not.re- ! quired for full credi .11 d Answer key was changed to " lower; bypass FRV will close until flow ! error offsets level erro RX trip will occur when actual level decreases (to 30% on 2/4 detectors in 1/4 SG)."
6.05 Answer key was changed to the approximate flow rate of one charging ' pum .06 Answer key was changed to include the assumption that the NSST is availabl .07 Answer key was not change .08 b Answer key was not change .09 Answer key was modified to include " sequencer test lineup" and "AFW pump suction valves closed."
6.10 c Answer key was modified to include " charging / letdown flow balance" and "RCS inventory computer program."
6.11 a The reference material does not support any change to the answer key, 6.13 Answer key was modified by replacing " Aux Air Recirc Fan" with
" Charging Pump." No reference material was provided to support this modificatio .01 a Question was deleted.
l 7.02 a No change was made to the answer key since all responses were considered correc .02 b Answer key was changed to accept five additional correct response .04 a Answer key was modified to accept " Core Exit Thermocouples."
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_ . -. 4 Attachment 4 7.05 a Answer key was changed to include:
" Secondary Safeties and PORV's."
7.06 Answer key was changed to include " caused by loss of main feedwater ' pump suction."
7.07- Answer key was changed to also accept " throttle AFW flow."
7.08 b No change was made since answer was in accordance with reference materia .08 d Answer key was changed to include " Restore the inoperable instrument to service or...." 7.09 a No change was made since answer was in accordance with reference materia .09 b Half credit was given for "60 F/hr."
t 7.11 c Answer key was changed to include " criticality monitor" and " required ; by technical specifications."
8.02 Question was graded to the answer key since no supporting documenta-tion was provide .07 Question was graded to the answer ke .03 b Answer key was changed to also accept "Yes; Provided 200 F was not exceeded."
. 8.08 d Answer key was changed by deleting the approval of department hea .10 a Answer key was not change .10 b No deduction was made for including " Mode 6" in the answe .10 c(2) Partial credit was given for FO .12 Question was delete .
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