IR 05000336/1987031
| ML20148Q689 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/25/1988 |
| From: | Eselgroth P, David Silk NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148Q680 | List: |
| References | |
| 50-336-87-31OL, NUDOCS 8804130229 | |
| Download: ML20148Q689 (116) | |
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V. S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMIflATION REPORT EXAMINATION REPORT NO.: 50-336/87-31 (OL) FACILITY DOCKET NO.: 50-336 FACILITY LICENSE NO.: DPR-65 LICENSEE: Northeast Nuclear Energy Company P.O. Box 270 Hartford, Connecticut 06141-0270 FACILITY: Millstone, Unit 2 EXAMINATION DATES: 1/11/88 - 1/14/88 CHIEF EXAMINER: / Day 1d M. Silk,
[. 3/d7!N Date '
Operations En ineer (Exami er) , APPROVED BY: , Peter W. Eselgptfth, Chief
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PWRSection,gperationsBranch,DRS
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SU!HARY: On the week of January 11, 1988, written and operating exaninations were administered to six Reactor Operator and two instant Senior Reactor Operator candidates. All candidates passed all portions of the examinatio l l
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i 8804130229 880328 PDR ADOCK 05000336 V DCD
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REPORT DETAILS TYPE OF EXMS: Replacement EXAM RESULTS: l RO l SRO I l Pass / Fail l Pass / Fail I l l l l 1 l l l Written Exam l 6/0 l 2/0 l I I i l l I I I l0ral Exam l 6/0 l 2/0 l l 1 I I I I I I l Sin:ulator Examl 6/0 l 2/0 l I I I I I I I I l0verall l 6/0 l 2/0 l I l l l CHIEF EXAMINER AT SITE: D. Silk OTHER EXAMINERS: B. Norris P. Isakson (EGSG) F. Jagger (EG&G) Summary of generic strengths or deficiencies noted on oral exams: SR0 candidates did not know the definition of the word "clearance". Summary of generic strengths or deficiencies noted from grading of written exams: This information is being provided to doce.ient areas of weakness which should aid the licensee in upgrading replacement training programs. No reply is required, l j
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l l Reactor Operator Examination Question N {ubject 1.02 Xenon production and re v, val at dif ferent power level .04c Reason for limit on load current if Bus 24E is supplied from its backup source 2.09a Design basis for AFW flow time dela .10b Response of an CEA to a demand signal if the lift coil has faile .09b RPS response to the channel 'A' SG differential pressure failing hig .05 Conditions that allow the operator to leave the surveillance area of the control room according to ACP.6.01, Control Room Procedur Senior Reactor Operator Examination 5.01a Effects of cooler feedwater on a secondary calorimetri .06c Effect of an RCP trip at 100% power on the Thermal Margin / Low Pressure trip setpoin .05 Operator actions to be taken if all feeawater ficw in lost for greater than 5 minutes and reactor power is greater than 5%. 8.06b Operator actions prior to resetting protective relays following a reactor tri .10 Conditions requiring both independent shutdown cooling loops to be operable.
1 Personnel Present at Exit Interview: NRC Personnel David Silk, Operations Engineer (Examiner) Lynn Kolonauski, Resident Inspector, Millstone Unit 1
Facility Personnel Stephen Scace, Millstone Station Superintendent : Brad Ruth, Manager, Operator Training Mike Wilson, Supervisor, Operator Training, Millstone Unit 2 Joe Parillo, Supervisor, Simulator Training, Millstone Unit 2 Dan Pantalone, Instructor, Millstone Unit 2 l
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. Summary of NRC Comments made at exit interview:
The NRC recapped the events and personnel involved during the wee Appreciation was expressed to the training department for their support and to the operating crew in the control room for attempts to minimize the distractions to the candidates. Also, the NRC commended the performance of the simulator in that it ran without malfunctioning except for the spurious loss of a 4KV bus at the end of one scenario. The NRC commented that the candidates generally performed well during the simulator examination with the exception of vague communication The NRC mentioned that there appeared to be no set policy regarding the wearing of dosimetry because during the week plant personnel were observed with dosimetry attached to pants pockets, belts and shirt tails. The NRC also observed an instance where a worker improperly frisked out (picked up the detector without first checking his hand). Summary of facility comments made at exit interview: lhe licensee felt that the NRC written and simulator examinations were within the bounds of the enabling objectives and the scope of the emergency operating procedure Attachments: Written Examination and Answer Key (RO) i Written Exanination and Answer -Key (SRO) i Resoiution of Facility Comments for RO Exam Given January 11, 1988 at Millstone Unit 2 Resolution of Facility Comments for SRO Exam Given January 11, 1988 at Millstone Unit 2
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U. S. NUCLEAR REGULATORY COMMISSION l REACTOR OPERATOR LICENSE EXAMINATION FACILITY: Millstone 2 ) REACTOR TYPE: EWB;CE DATE ADMINSTERED: 88/01/11 EXAMINER: LEAES_EN, CANDIDATE INSTRUCTIONS TO CANDIDATE _L 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 questio 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 j the examination start '4 OF i CATEGORY % OF CANDIDATE'S CATEGORY i VALUE_ _TOIAL ___ SCORE - VAL]E__ CATEGORY l 25.00 _2f200 PRINCIPLES OF NUCLEAR FOWER PLANT OPERATION, THERMODYNAMICS, l HEAT TRANSFER AND FLUID FLOW 25.00__ _RhuQq PLANT DESIGN INCLUDING SAFETY l __ AND EMERGENCY SYSTEMS I 25.00 _2htkQ _ INSTRUMENTS AND CONTROLS I 25.00 __2 5 . QQ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 10 % Totals Final Grade l All work done on this examination is my ow I have neither given nor received ai MM"90?Y CalidfdaTe' ~s"5f gfafure "
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS _During the administration of this examination the following rules apply: - Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . Restroom trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin '
, Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). l
. Use only the paper provided for answer ' Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side
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J of the paper, and write "Last Page" on the last answer shee : Number each ancwer as to category and number, for example, 1.4, ( 10. Skip at least three lines between each answe ! '
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j 11. Separate answer sheets from pad and place f.4 nished answer sheets face down on your desk or tabl ' 12. Use abbreviations only if they are commonly used in facility literatur . The point value for each question la indicated la parentheses after the
question and can be used as a guide for the depth of answer require L 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete I
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18. When you complete your examination "^" ahall:
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a. Assemble your examination as fo.- ut (1) Exam questions on to (2) Exam aids - figures, tables, et /
(3) Answer pages including figures which are part of the answe b. Turn in your copy of the examination and all pages used to answer the examination question c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, d. Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked, i
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l' ,' PRINCIPLES OF NECLEAR POSER PLANT OEEBATIO Page 4 < THERMODYNAMICS, HEAT _TR&UEEEB_AtlE_EkQLD FLOW
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d QUESTION 1.01 (1.00)
Increasing the boron concentration at low temperature has little effect ; on the Moderator Temperature Coefficient as compared to higher operating temperatures because:
(choose the correct answer)
a. water density does not change as much at low temperature, b; water les density is greater at lower temperatures so neutron leakage is ' water density is greater at lower temperatures so parasitic neutron absorption is greate boric acid is less soluable at lower temperatures.
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Which one of the following statements concerning Xenon-136 production and removal is correct?
' At full power, equilibrium conditions, about half of the Xenon is produced by Iodine decay and the other half is produced as a i direct fission produc . Following a reactor trip from equilibrium conditions, Xenon peaks because delayed neutron precursors continue to decay to Xenon
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while neutron absorption (burnout) has cease !
' Xenon production and removal increases linearly as power level e increases; i.e., the value of 100% equilibrium Xenon is twice that of 50% equilibrium Xeno i
! At low power levels, Xenon decay is the major removal metho !
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At high power levels, burnout is the major removal metho ! i
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QUESTION 1.03 (1.00) ,
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]' As the core ages, the ratio of PU239 atoms to U235 atoms increase ; This changing ratio causes the: Reactor startup rate (SUR) to increase, for the same reactivity additio : i Void Coefficient becomes less negative, Moderator Temperature Coefficient to become less negativ ; Delayed neutron fraction to increas QUESTION 1.04 (1.00) Which one of the following statements concerning Shutdown Margin (SDM) is cor. rect? The mtximum SDM requirement occurs at EOC and is based on a rod ' ejection acciden .
! The maximum SDM requirement occurs at EOC and is based on a steam
, itne break acciden The maximum SDM requirement' occurs at BOC and is based on having a positive moderator temperatur+t coefficient, , 1 The maximum SDM requirement occurs at BOC and is based on a rod I e withdrawal accident while in the-source range.
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QUESTION 1.05 ( 1. 00 ) ( A> orFt 9 M' ** The reactor is criti:*.1 at 10,000 cps when a S/G PORV fails ope Assuming BOC conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of Tavg and nuclear power for the resulting new steady stat (POAH = point of adding heat). Final Tavg greater than initial Tavg, Final power above POA Final Tavg greater than initial Tavg, Final power at POA . Final Tavg less than initial Tavg, Final power at POA Final Tavg less than initial Tavg, Final power above POA QUESTION 1.06 (1.50) The reactor is taken critical with Xenon concentration at zer Power is raised to 50% at 5%/ mi A trip occurs as power reaches 50%. Use one of the following choices to describe how Xenon concentration will be trending for each of the following situations (a,b, and ! Increasing i Decreasing At equilibrium One hour after the tri Four hours after the trip the reactor is taken critical and power raised back to 50%. If the reactor had NOT tripped and power was level at 50% for one hou QUESTION 1.07 (2.00) How is Shutdown Margin (SDM) affected (Increase, Decrease, or No change) by a 50 ppm boron addition while operating at 50% power?
(0,5) List THREr 'nctors, other than RCS boron concentration and rod positiot '..ch will affect SDM and are used in the SDM calculatio (1.5) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) . . -
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QUESTION 1.08 (1.50) One reason for the CEA Insertion Limit is to ensure sufficient SDM is availabl What are the other TWO reasons? (1,0) The CEA Insertion Limit is a function of .
(Fill in the blank, place your answer on your answer sheet.) (0.5)
QUESTION 1.09 (2.00) (vorwi m c.u - w c: e a _ u e 7-_ g a. Explain HOW and WHY ASI is expected to change as power is increased from 20% to 70%, during a normal power increase at EO b. What TWO steps / methods are taken to maintain ASI within limit AND WHY these actions are effectiv QUESTION 1.10 (2.00) At BOC, power is reduced from 100% to 50% and stabilised, briefly explain HOW and WHY each of the following plant parameters will be affected over the next 5 hours. Assume rod control is in manual sequential, all other systems are in automatic, and No operator action is take a. RCS temperature b. RCS pressure c. S/G pressure d. Turbine Generator control valve position (***** CATEGORY l CONTINUED ON NEXT PAGE *****) __
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QUESTION 1.11 (1.00) I
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I ! Which one of the following conditions would cause a 1/M plot to be NON- l 4 conservative during fuel loading?. ; i Fuel being loaded closer to a source range detector than to the neutron source, Loading fuel in the order of high reactivity worth to low reactivity l wort , Loading poison rods between the source range detectors and spaces to be filled by fuel assemblies, Increasing the boron concentration in the moderato ; i QUESTION 1.12 (1.00) 1,
A centrifugal pump is operating at rated flow when the dischargo ,
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j valve is throttled towards the shut direction. Which one of the l following statements BEST describes the parameter changes that will i occur 7 f Flow constant, discharce pressure constant, motor amps in-crease, NPSH increases, , b i
' Flow decreases, discharge pressure increases, motor amps in- !
crease, NPSH increase I l
n Flow decreases, discharge pressure increases, motor amps in-crease, NPSH decrease ! .
, Flow decreases, discharge pressure increases, motor amps de- ;
crease, NPSH increases.
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QUESTION 1.13 (1.00) TRUE or FALSE During a RCS heatup, as temperature gets higher, it will take a smaller letdown flow rate to maintain a constant pressuricer level.
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( Increasing condensate depression (subcooling) will cause BOTH a decrease in plant efficiency AND an increase in condensate (hotwell) pump available NPS ( QUESTION 1.14 (2.00)
Assuming you are operating at 85% power indicate how the following changes in plant conditions would affect DNER (increase, decrease, remain constant). Consider each case separatel . The operator withdraws control rods without changing turbine loa . ASI changes from 0% to -2%. Steam Generator PORV fails ope ( vorri g #d A A bV) Pressuriner heaters are inadvertantly left o . Reactor coolant pump speed decrease [0.4 each) ( !
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QUESTION 1.15 (2.50) a. What is the subcooling margin (SCM) of the RCS if the following conditions exist? Th :580 F Pressurizer pressure =2185 psig Tc :520 F Steam Generator Pressure :850 psig (1.0)' b. If power is raised from 50 to 100%, how AND why will SCM change (increase, decrease, stay the same)? (0.75) c. Which of the following would give a smaller SCM7 Briefly exp1hin your choic Assume identical RCS pressure (0.75) 1. SCM during a controlled natural circulation cooldown immediately following a reactor trip from loss of flo . SCM from continued operation at 5% powe . SCM produced when all RCP's are operated at normal no-load temperature after extended shutdow : QUESTION 1.16 (2.00) The plant is in a Natural Circulation Mode of core coolin As the fission product heat decays away, describe HOW (increase, decrease or remain constant) and WHY you would expect the following RCS parameters to change. Assume that S/G pressure is being maintained constant at 900 psia, a. Teold b. Thot c. Loop transit time I
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QUESTION 1.17 (1.50) At each of the following leak locations, indicate the state of the exiting fluid (subcooled, saturated, or superheated). Assume norr;ai 100% power initial operating plant condition PZR steam space to CTMT atmospher Steam Dump to the condense Main steam header to turbine building atmospher (***** END OF CATEGORY 1 *****)
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l . I QUESTION 2.01 (1.00) l l What TWO features ensure that the RWST cannot be siphoned below the ! minimum Technical Specification level if a rupture were to occur in the RWST recirculation system piping? Consider ruptures in both the suction and discharge piping of the recirculation pum QUESTION 2.02 (2.00) The following questions concern the Containment Air Recirculation and Cooling System (CARCS) and its associated support system What automatic actions should occur to the CARCS after an SIAS? Assume an initial normal lineu TWO require (1.0) b. If 100% is defined as the heat removal capacity necessary to limit containment pressure to less than design pressure following an accident, what is the total heat removal capacity, in percent, of all CAR fans?
(0.b)
c. What instrument system uses the Containment Air Recirculation and Cooling System as a means of normal cooling? (0.5)
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- QUESTION 2.03 (3.00)
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- The following questions concern a loss of instrument air j (assume normal, at power initial conditions): !
l a. How would a complete loss of instrument air outside containment ' j immediately affect the following components / systems? Chooso ONE of the following (A,B,C, or D) for each of the 8 component / systems:
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, A- fail open/ flow maximum .
B- fail closed / flow stopped
C- fail as is/ flow cannot change
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D- no effect/ system functions normally
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1. Main Feedwater Regulating Valves 2. Pressuriser spray valves
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4. Letdown I t 5. Atmospheric Dump Valves
6. AFW flow control valves , !! 7. EDG service water supply valves ;
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' b. Describe two means of interconnecting the IA system with backup
sources of air pressure. Indicate automatic setpoints, if any.(1.0) , i t
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I The following questions concern the operation of 4160V Bus 24 t , l ! j List THREE loads which are supplied directly from Bus 24E (other I
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j than electrical busses). (1.0) : ( b. Describe the Reactor Safety function of the Kirk Key interlock 1 I i associated with Bus 24 (1.0) I If Bus 24E is supplied from its backup source, a limit is placed on ~ the limit? load current supplied by that source. What is the reason for this (1.0) i ! ! I j (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) i I a _ _ _ _ - _ . - , - - _ _ _ . _ - _ , _ _ _ . . _ - _ . _ _ _ , _ _ _ . - _ . --. -_
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QUESTION 2.05 (1.00) - What design feature of the clean liquid rad waste system ensures that waste additions are not made to a monitor tank which is being discharged? QUESTION 2.06 (2.00) The discharge of both LPSI pumps passes through valve 2-SI-306 "SDC flow control valve" during shutdown coolin What is the potentially adverse consequence of this valve being closed during power operation? (0,5) b. What THREE PHYSICAL precautions are taken to ensure it remains open while the plant is at power? (1.5) QUESTION 2.07 (1.60) What are TWO conditions which will cause automatic transfer of the 120V vital normal instrument to emergency AC bus (VIAC) static transfer switch from power? QUESTION 2.08 (3.50) l State TWO design features which serve to prevent the loss of water inventory in the Spent Fuel Pool (SFP). (1.0) List THREE sources of makeup water to the SF (1.5) If the SFP was filled with UNBORATED water will the Technical Specification required Keff of <0.95 still be satisfied? your answe EXPLAIN l (1.0) i (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****)
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QUESTION 2.09 (4,00) State the reasons / design basis for the incorporation of the AFW flow time dela (1,0) What is the basis for the minimum required CST inventory? (1.0) What is the "alterncte" source of water supply to the AFW pumps? (0.5) What are the THREE modes of operation for the AFW Flow Control valves? INCLUDE the purpose for each of the three position (1.5) QUESTION 2.10 (2.00) The following concern the failure of a CEA lift coil with the affected CEA in a withdrawn configuration, assume the reactor remains at power, a. Explain why the affected rod will/will not dro Explain why the affected rod will/will not move on a demand signa QUESTION 2.11 (2.00) Describe what automatically happens in the Chemical and Volume Control system upon receiving a SIAS signal. Five separate actions require l l
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QUESTION 3.01 (3.00)
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The following questions concern the Control Element Drive and Position Indication System: What regulating control rod interlockn/lia.its are in effect when in MANUAL INDIVIDUAL control? (1.5) b. What system supplies rod posit 3on signals to the upper and lower CEA stops - (UCS, LCS)? (0.5) i c. What are the instrumentation signals /condit. ions that could provide a "Dropped Rod" annunciator? 4TWO (auired) (1.0) , QUESTION 3.02 (2.50) ' For the conditions listed below, Judicate the ESAS channels that should actuat Consider the conditions in part a, of this question seperate from the canditions in part S/G #1 pressure 800 ps3a S/G #2 pressure 5.00 psia I S/G #1 level 60%, nafrvw range S/G #E leval 45%, narrow range Pressurized pressure 1800 raia Press.irxzer level 201, narrow i Spent-fuel pool area r range monitors 150 nir/hr (1.0) ' ! S/G #1 pressub; 900 paia ~ S/G #2 prel.ture 900 psia S/G #1 level 68%, narrow range E'G $2 level 68%, narrow range
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Pressurizer pressure 1500 psia Prerf;urf ter level 33%
RWST level 8% , Cd:stainment pressure 30 psig Spent-f uel builcing radiat'an a sensor channels all reading 100 mr/hr (1.5) !
- QUESTION 3.03 (1.50)
1 With Channel X selected for pressurizer level control, level instrument
! LT 110X fails low. What automatic actio.ts Lnd f .idication/elarms will i
initially result, including the effect on actual pressurizar level, With NO operator action? l l l
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QUESTION 3.04 (1.50) What are Power TripTHREE conditions Test Interlock which will actuate the individual channel (PTTI)? QUESTION 3.05 (3.00) The following questions concern the Reactor Protection System: a. Which power? reactor trip inhibits are automatically removed above 10E-4%
(1.5)
b. Which reactor trips are automatically bypassed below 15% power? (1.0) c. Which power? reactor power signal is used to bypass the trips below 15%
(0.5)
QUESTION 3.06 (2.00) Describe how AND why the Feedwater Control System (FWCS) reacts to the following TWO occurences. Assume an initial normal full power plant condition with the FWCS in MANUAL, and feed flow, steam flow, and level : transmitter selector switches on BOTH for each occurrence and consider each separatel !
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a. Turbine Trip l b. The A S/G Alternate level transmitter fails hig i QUESTION 3.07 (2.00) Which of the following monitor channels have automatic actions (other than indication and alarm) as. .ciated with them? Briefly describe the automatic actions, if any, a. Spent fuel pool area monito b, Radwaste Vent Monitor - Gateous.
I c. Cendensate recovery tank monitor, j d. "lean radwaste monito (***** CATEGORY 3 CONTINUED ON NEXT PAGE *****)
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l t i QUESTION 3.08 (3.00) The power plant is operating at 85% of full power with ASI: and with the control rods at 110 steps on Group 7. Explain how , AND why changes in the following parameters will affect the TM/LP ! pressure set point (increase pressure setpoint, decrease, or stay i the same). Assume the operators take action to maintain a constant i 85% electrical output and the operators reaintain programmed I operating limit Consider each item separatel Xenon buildup in bottom of core, Tavg increasin RCS boratio QUESTION 3.09 (2.50) During equilibrium, full-power operation, describe how the RPS will respond to each of the following Channel 'A' input failure Include !' meter response, alarms, channel pretrips and trip Also, indicate the RPS channels that should be bypassed by the operator for each failure, ' consider each failure seraratel Channel upper NI fails hig 'A'
(1.7) Channel 'A' Steam-Generator differential pressure fails hig (0.8) l l
QUESTION 3.10 (3.00) : i The following concern the Steam Dump / Turbine Bypass System: a. How many valves would open after a reactor trip from 75% power? Explain, b. How many valves should be open 5 minutes after the reactor trip? Explain, c. How would your answers to a and b, above, change if the reactor trip was caused by a loss of off-site power? Explai ;
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, INSTRUMENTS AND CQHIBQLE Page 19 j
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i ! QUESTION 3.11 (1.00) - i l Match the Emergency Diesel Generator alarm condition in Column A with l
- the proper action in Column Place answers on your answer shee '
i I i Column A Column B l l
- e- D/0 12') L vmble" None, D.G.-will not Trip [
l b. Lube Oil Temp. High D.G. will Trip l c. Jacket Coolant Pres Low D.G. will Trip unless Emergency , d. Engine Overspeed start signal is present e. C. Cv,. vi ol Poww4- F.11;n - t i
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* . d PROCEDJJRES - NORMALdBECEMAlu_EMERGENC1 Page 20 * ,
AED_BADIOLOGLC&L CONTROL
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l QUESTION 4.01 (2.00) 1 Indicate whether each of the following statements about the station tagging system is TRUE or FALSE, according to ACP 2.06, Station Tagging procedur , j a. Any number of Blue Tags may be attached to a switch or other device at any one time.
J b. A Green Stripped Hold Tag may be attached to a device on which a i Yellow Caution Tag is already attache , c. Lifting of Tags in order to perform testing is permitte ! i d. Any number of Red Tags may be attached to a switch or other device j at any one tim QUESTION 4.02 (2.00)
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List FOUR of the six ications that are used to verify that Natural
- Circulation has been established, according to AOP 2553, Plant Cooldown ,
using Natural Cooldown procedur ' QUESTION 4.03 (3.00)
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One of the operator actions of the Electrical Emergency (Loss of Normal ! Power) procedure, EOP 2528 is to place the Enclosure Building ! , Filtra'. ion System in service, i
L What are FOUR of the steps which must be taken to place the Enclosure ;
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Building Filtration System on tne line?
(2.0) l j b. Why must this system be placed in service? (1.0) ;
a QUESTION 4.04 (4.00) a. What are FOUR plant conditions, abnormalities, or emergencies which
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would require the Reactor Coolant System to be Emergency Borated, according to AOP 2558. Emergency Boration procedure?
b. List the steps required to be performed to initiate emergency boration ] (four required for full credit) "
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QUESTION- 4.05 (2.00) What are the TWO conditions that allow the operator at the controls to leave the surveillance area of the control room, according to ACP 6.01, Control Room procedure? QUESTION 4.06 (2.00) A caution in the Emergency Diesel Operating procedure (OP 2346A) states;
"If a LNP is initiated, do not reset undervoltage relays at the ESAS Cabinet (s) until immediately prior to paralleling the Diesel Generator with the RSST during restoration per EOP 2528 (Electrical Emergency)."
Briefly explain the reason for this caution and the consequence of not following this precautio QUESTION 4.07 (2.00) For EACH of the below conditions list FOUR parameters used in the immediate action steps of EOP 2525 - Standard Post Trip Actions to determine if the conditions exist a. PORVs and Pressurizer safeties are NOT open, b. Normal containment condition QUESTION 4.08 (1.00) During a reactor shutdown from outside the control room, AOP 2551 Shutdow From Outside The plant cooldown exceeds ControlaRoom directs specified that MSIVs are to be closed if limi unavailable, how are the HSIV's closed? Switch If the control room is require or component numbers not (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****) _ _ _ _ .
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QUESTION 4.09 (3.00) The following questions concern EOP 2532, Loss of Primary Coolant procedure:
% m .) What is*the criteria to manually initiate SIAS? (1.0) When are the RCP's required to be stopped? (0.b) A caution in this procedure states that P2R level may not Provide an accurate indication of ECS inventory. What other indication would validate that PZR level'was an indication that the core was covered? (0,5) How is boron precipitation control accomplished if the SDCS is not available? (1,0)
QUESTION 4.10 (3.00) The following questions concern AOP 2556, Dropped CEA Recovery procedure: ' What are FOUR indications that a CEA is misaligned? (1.0) What single condition would require the turbine and reactor tripped, assume the reactor is at power and only one CEA is misaligned? (0.5) After the misaligned CEA occurs and reactor power is stable at 90%, How far must power be reduced AND how is this power reduction accomplished? (1.0) What action is required if two or more CEA's have dropped? (0,5) QUESTION 4.11 (1.00) What is the RCS leakage criteria which requires the reactor to be tripped, according to AOP 2568, RCS Leak procedure? What would be an indication of RCS leakage into the Safety Injection system, according to the above referenced procedure?
(***** END OF CATEGORY 4 *****)
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EQUATION SHEET s f = ma
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y= s/t ! v = as s = v,t + at
Cycle efficiency = Net Work (out) * ' E ergy (in) E = aC * a = (vg - v,)/t gg = hay 1 v g,vg , ,t A . AN A = A,e' " PE = ash w = 6/t A = In 2/tg = 0.693/tg
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W = v&P-AE = 931Aa t q(eff) = (t,,)(ts) !
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k= BAT p ,
, I . I ,*IX I g
, ,,k=UAAT
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*Pvr = Wga I . I,,-uX ,
I.I to-x/ M
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g
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y.y toSM(t).
TVL = 1.3/u y.y O ,t/T
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HVL ' O.693/u
'SUR = 26.06/T _
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T = 1.44 DT SCR = S/(1 - K,gg)
/A ' c)
SUR = 26 i g CR, = S/(1 - K,ggx) i
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T = lt*/o ) + [(f-'o)/A,ggo ] 1 eff 1 * *2O -leff T2 ' t = t*/ (, . ;;
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M = 1/(1 - K,gg) = CR g/CRO
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T = (I - o)/ A *ff o M = (1 - K,gg)n/(1 - K,gg)g a = (K,gg-1)/K,gg = 3 ,gg/Keff p= SDM = (1 - Keff)/Keff
[1*/TK,'gg -) + @/(1 + A,ggT )] ,
1* = 1 x 10 seconds
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P = I6V/(3 x 1010) A,gg/= 0.1 seconds I = No -
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Idgg=1d22 UATER PARAMETERS Id g =1d2 1 gal. = 8.345 lba R/hr = (0.5 CI)/d 2 g,,,,,,) 1 gal. = 3.78 liters 2 R/hr = 6 CE/d (feet) - 1 ft = 7.48 ga MTSCELLANEOUS CONVERSIONS . Density = 62.4 lbs/ft 1 Curie = 3.7 x 1010 dps Density = 1 gs/cm 3 1 kg = 2.21 lba Heat of valorizations = 970 f tu/lba 1 hp = 2.54 4 103 RTU/hr l l Heat of fusicn = 144 Stu/lba 1 N = 3.41 x 106 stu/hr 1 Ata = 14.7 Psi = 29.9 in. I Stu = 778 ft-lbf I ft. H O = 0.433", Ibf/in 2 2 g inch = 2.54 cm F = 9/5'c + 32 l
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*C = 3/9 ('r - 32)
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- '. 1 __EBlEAELES_9LERGLE AB_ EREEB_ ELABI._9EEBAIl9th Page 23 . IBEEMQDYEAMI.CL_BEAI_IBANDEEB_A E_ELMID_EL99 M/STEROPY ANSWER 1.01 (1.00) REFERENCE M2-0P-RO-FUND-2116E, p 18-2 K106 ..(KA's)
ANSWER 1.02 (1.00) REFERENCE M2-OF-RO-FUND-2116F, p 11-1 K105 ..(KA's) ANSWER 1.03 (1.00) ( REFERENCE M2-OP-RO-FUND-2116C, p 8-11, 22-2 K106 ..(KA's) ANSWER 1.04 (1.00) REFERENCE M2 TS, p. B 3/4 1-1; M2-OP-RO-FUND-2116G, p 35-3 K114 ..(KA's) ANSWER 1.05 (1.00) (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****) M/5TERCOPY
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IEEBdQQ1HAdlGL__ HEAT TRAHEEER AHD_ FLU.1D_ELQE
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REFERENCE M2-OP-RO-FUND-2116G, p 13-15, -2116E, p 24-2 K114 ..(KA's) ANSWER 1.06 (1.50) Increasing, Decreasing, Increasing, [0.5 ea.]' ( REFERENCE
. M2-OP-RO-FUND-2116G, p 40-44 and TP-5, 192006K107 ..(KA's)
ANSWER 1.07 (2.00) SDM is increased. [0,5) [any 3, 0.5 each)
-RCS avg temp -Samarium-Fuel burnup -Power defect-Xenon concentration -Power level REFERENCE l
M2-OP-RO-FUND-2116G, p 35-3 K114 ..(KA's) t ANSWER 1.08 (1.50) ! . ! Core design peaking factors are not exceeded (acceptable power I distribution limits). !
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2. The reactivity associated with a CEA ejection accident is accept-able (within analysis),
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REFERENCE , M2 TS 3.1.3.6 and bases (B3/4 1-3) C-E Reactor Theory, pp 193-19 ! M2-OP-RO-FUND-2117, p 52-54 EO 1 K115 l
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ANSWER 1.09 (2.00) ' Moderator density becomes less at the top of the core (0.3] causing ! the flux peak to move down in the core. [0.23 (Since ASI is 1-u/1+u) it will become more positive as the power is increased. [0.5] l . Reduce power which creates less restrictive limits. [0.5]
(more neg. ASI due to density changes in moderator associated with i delta-T and Tc program)
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2. Use rods to change flux shape which changes the value of ASI.[0,5)
(M* W a REFERENCE
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M2-OP-RO-FUND-2117 EO le and 22; TS p 3/4 2- ; 192005K110 192005K114 ..(KA's) !
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ANSWER 1.10 (2.00) ! n
, Decreases (0.25] due to buildup of Xe (0.25] Held constant (0.25] by PPCS spray and heaters (0.25] ' Decreases (0.25) due to the decrease of Tavg (0.25] [ graded based upon answer in a, above] Remains the same (0.25] since load is lowered on valve i r
position limiter and no operator action is assumed.Lo.srp (2.0) ] REFERENCE
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M2-OP-RO-2117, EO 7;-2116 l 192006K106 ..(KA's) ' ANSWER 1.11 (1.00) CL rs c (***** CATEGORY 1 CONTINUED ON NEXT PAGE *****)
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REFERENCE M2-OP-RO-FUND-2116G,p$h<23, 192008K106 192002K114 ..(KA's) ANSWER _ 1.12 (1.00) ( REFERENCE GP HTFF p.328 M2-OP-RO-FUND-2121E, p 19-2 K105 ..(KA's) ANSWER 1.13 (1.00) FALSE ( TRUE ( REFERENCE General Physics HT&FF, pp. 155 and 320 and Subcooled Liquid Density Tables M2-OP-RO-FUND-2121H, p 7,24,2 e 193004K111 ! 193005K103 ..(KA's) i L ANSWER 1.14 (2.00) l DNBR decreases DNBR decreases ) DNBR decreases i DNBR increases DNBR decreases (0.40 each) ( REFERENCE H2-OP-RO-FUND-2121I. p 19-2 K105 ..(KA's)
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l ANSWER 1.15 (2,50) ! I From the C-E Stm Tables, Tsat for 2200 psia: 649.5 F SCM: Tsat-Th: 649.5-580: 69.5 F +/-1 F [0,5 ea) (1.0) ! decrease [0.25] Th increases (as unit delta T increases with powei)[0.5] (0.75) c W = -=.- = -n:-_-e =mm [0.25] Core delta T during natural circulation cooldown will approach {
""11 lesd delte Tc Thot is greater than in the other 2 cases.[0.b] i 30-J5'% r' (0. 76) '
REFERENCE i
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M2-OP-RO-FUND-21210, Objective 5. -2'Al :C P M '" 3 - ! GP HTFF pp.356; Steam Tables 193008K115 ..(KA's)
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ANSWER 1.16 (2.00) Teold will remain temperatur constant [0.16) Since it follows S/G saturation
[0.5) !
L b. Thot will decrease [0.16) since less fission product heat is being ! produced than is being removed by the steam generators. [0.5) i c. Loop transit time will increase [0.163 since the driving head for flow (core delta T) is decreasing. [0.5) ;
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REFERENCE i M2-OP-RO-FUND-2121J. p 6-13,1 K122 ..(KA's) l
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ANSWER 1.17 (1.50)
! Saturate ' Superheate !
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. IEEBMQultl6MLGEt_HEALIBAt!EEER_AtID_ELulu_ELQH REFEREt4CE Steam Tables, Molier diagram M2 Exam bank EO RO-05, item 1090, lesson plan 2121 K115 ..(KA's)
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* . . * . 2 ELARLDESIGtLidgLUDING SAEEILAUD_EtiEBGENCX Pace 29 , E1EIEde .
ANSWER 2.01 (1.00) Siphon breaker in suction line of recirculation pump.(is located at 91.4% level).[0.5) Recirculation Pump discharges to top of tank. [0.5) REFERENCE Dws No. 25203-20015 and HPSI S.D. p MP2 question 382 002000G007 ..(KA's) ANSWER 2.02 (2.00) a. Idle CAR fan starts on low speed Operating CAR fans switch from hi to low speed 10 inch RBCCW valves on cooler outlets open [2 required. 0.5 each) b. 133% (0.5) c. Nuclear Instruments (0.5) REFERENCE M2-OP-RO-PRI-2313, p 41-47.(Cont Ventilation S.D.)
022000A301 ..(KA's) ANSWER 2.03 (3.00) a. 1. MFWRV C- fail as is i 2. par spray valves D- no effect (entmt air revr charged) 3. S/G feed pump control C- fail as is 4. Letdown B- system flow stopped 5. ADVs B- fail closed 6. AFW flow control valves A- fail open 7. EDG service water supply A- fail open (0.25 each] 8. MSIVs D- no effect (independent air reciever) b.1) Auto valve to station air compressor X-ties SA to IA at 85 (t'-3) psig IA pressure (0,5) 2) Manual
[0,5)
X-tie valve to Unit 1 station air via Unit 2 station air l
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REFERENCE I l
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! M2-0P-RO-SEC-2332 A/B, EO 4,9,11. (Instrument Air S.D) : i M2-OP-SEC-2322. p 3. (AFW S.D)
ESAS handout Appendix A pg 10 (EDG SW valves) { j 078000K402 078000K302 ..(KA's) [ i i r i I i ANSWER 2.04 (3.00)
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a. Swing ($1 Service water pump, swing HPSI pump, swing (MRE ! j b. Prevents tieing both 4160 emergency busses together through Bus 24 ! s (1.0) I c. Prevents exceeding load limits on RSST b; .n' oil (15G-21S) and .ite associated bussin ! i (1.0) i i REFERENCE 1 ;
! M2-OP-ELECT-2342, p 3,4 & fig. (Elect Dist SD), of GkV3 P I?-
i 062000A404 062000A206 l' < 062000K401 ..(KA's) 1 !
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may be opened only if the corresponding inlet valve is shut.(0 5) l REFERENCE !
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M2-OP-NLCT-PRI-2335A, p 29. (Radwaste SD) OGB000K401 ..(KA's) l ! l l ANSWER 2.06 (2.00) i 1 a.
j The inability of the LPSI pumps to inject water into the core fellowin a LOC [0.5) l
i b. 1) keylock switch to SI positivu [0,5) ] 2) Remove fuseblock for 2-SI-306 (0,5) 3) Isolate air supply (0,5) 4) Manual operator on the opposite side of the valve shaft is pinned and locked to the handwheel. (0,5) 5) Handwheel is locked in position. (0.51 (any 3 at 0.5 each) i
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REFERENCE M2-OP-PRI-2307, p 9.(LPSI SD) and OP2310 Rev 8 Section 7. K108 ..(KA's) ANSWER 2.07 (1.50) Inverter output voltage lo . Inverter failur . Inverter load overcurren [twe required, 0.75 each] REFERENCE M2-OP-ELECT-2346, p A304 ..(KA's) ANSWER 2.08 (3.50) Stainless steel line No penetrations in pool wall below the normal water leve (except the fuel transfer tube) 3- Siphon breakers for penetrations above the the normal leve Seismic qualified components and equipmen (2 required, 0.5 each] Primary makeup water syste RWST(.via bypass and purification system) 3- RWST(via LPSI system.)
4- AFW syste [3 required. 0,5 each] { Yes (0.5], the center to center distance (spacing) between spent fuel assemblies is sufficient to maintain TS Keff requirements with unborated water (0,5).
REFERENCE M2-OP-PRI-2305, p 8,9,16,1 I 192002K110 192002K112 059000K405 033000K401 ..(KA's) l (***** CATEGORY 2 CONTINUED ON NEXT PAGE *****) j l i i
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ANSWER 2.09 (4.00) > To prevent a return to power situation from occurring during a DBA MSLB acciden [1.0] , To remove decay heat for 10 hours with steam discharge to atmosphere [1.0] OR Sufficient water available for cooldown of RCS-<300-F within 6 hours, in the event of a total loss of off-site power. [1.0) Plant fi a system. [0.5] a
' NORMAL [0.25]- allows the valve to be open fully for auto actuation. [0.25] '
OVERRIDE [0.25]- allows manual control of the valve after an auto actuation. [0.25]
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RESET [0.25]- to return the mode of operation back to normal.[0.25] R'EFERECY
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M2-OP-SEC-2322, p 2,3,6,10,11. (AFW SD); TS B3/4 7-2. oPGSA1, do P A5 79 8 061000G007 061000K411 061000K401 ..(KA's) ANSWER 2.10 (2.00) The rod will NOT drop [0.5) due to the action of the Icx r'L grippers [0.5]. The rod will NOT move (up or down) [0.5) since the lift coil is used to raise the upper gripper in either direction [0.5]. l REFERENCE f M2-OP-SO-I&C-2302A, p 20-2 , i 001000K402 ..(KA's) I
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l ANSWER 2.11 (2.00) a. 1. Boric acid pumps start [0.4] 2. Charging pumps start [0.4] 3. Boric acid storage tank is lined up to inject boric acid direct / gravity [0.4] 4. Boric acid storage tank is lined up through MOV to charging pump suction. [0.4] 4. VCT makeup stop valve shuts [0.4] VCT outlet valve shuts [0.4] 6. Letdown line loop isolation valves shut [0.4] RCP bleed off to VCT shuts. [0.4] [five required, 0.4 each] REFERENCE M2-OP-PRI-2304, fig. 2a. (CVCS SD) 004010A205 ..(KA's) p 9M W 4 m m C o c.H - ia . C h . 9. cw 4/ n + 4rf w n Wh) 6-c 4 'ic57 -9 iQW
= C W@ wA:' -- .
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ANSWER 3.01 (3.00) a. 1. Upper: 2. CEA wi"g lower
,hdrawal electrical prohibit limits --?a
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3. CEA motion prohibit [03'55ach] b. Computer position indication [0.5] c. 1. Rod drop from reed switch 2. NI negative rate of power change from NI system [0.5 each] REFERENCE M2-OP-SO-I&C-2304A, (CEDS SD), attachment ! MP2 question 704 000003G203 001050A204 061000K401 ..(KA's) ANSWER 3.02 (2.50) AEAS [+0.5] MSI [+0.5] SIAS CSAS EBAS SRAS ) CIAS
[+0.3] each REFERENCE M2-OP-RO-I&C-2384 (ESAS SD), TP 6& A401 ..(KA's)
ANSWER 3.03 (1.50) 1. Letdown to minimum 2. Both backup charging pumps start 3. Channel X PZR level Hi/Lo annunciator 4. Actual PZR level increases 5. PZR level Lo/Lo annunciator 6. All heaters de-energize 7. Selected controller (X) output signal to minimum [any 5, 0.3 each)
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REFERENCE M2_0P-I&C-2304A (PZRLC SD), p 5-10 and fig G005 ..(KA's) ANSWER 3.04 (1.50) 1. Linear power channel summer control switch out of the (A + B)/2 positio [0.5] 2. Linear power channel high voltage bistable tripped [0.5] 3. Reactor protection system calibrate and indication panel Delta T power calculator test switch out of the operate position. [ (or RPSCIP calibrate switch) 4. Zero-operate-calibrate switches (2) on NI Linear Power Channels out of operate [0.5]
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5. Trip Test switches (2) on NI Linear Power Channels out of off [0.5]
[any 3, 0.5 each]
REFERENCE M2-OP-RO-I&C-2380-2 (NI SD), p 21,35-3 A302 ..(KA's) ANSWER 3.05 (3.00) a. RC flow RCP speed Thermal Margin / Low pressure [0.5 each] b. Loss of turbine Local Power Density [0.5 each] c. Linear safety channel [0.5] REFERENCE M2-OP-RO-I&C-2380-1 (RPS SD) fig. 29,1 NI S.D. Power Range Linear Channel para K406 ..(KA's)
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ANSWER 3.06 (2.00) a.MFWRVrampsshut(for30sec#shouldfullyclose) and MFW Bypass valve ramps to 75% open (5% flow) [0.5] to help match feed flow with the reduced steam flow [0.5] b. No effect [0.5] since the BOTH position low selects the level signal to be use [0,5] REFERENCE M2-OP-SEC-2385 (FWCS SD), p 4-7, A104 059000K402 ..(KA's) ANSWER 3.07 (2.00) InitiateanAEAS(tominimizetheradioactivityreleasedtoatmosphere)
[0.5]
b. none [0.5] c. High alarm shifts condensate recovery tank discharge from aux stm feedwater surge tank to the aerated waste system which is isolate [0.5) d.Highalarmclosestwodischargeisolationvalves(hostopdischarge flow)[0.5) REFERENCE M2-OP-RO-I&C-2383 (RMS SD), and OP238 G008 073000K401 ..(KA's) ANSWER 3.08 (3.00)
: Increase [0.5] because ASI is decreasing (becoming negative) and that will increase a penalty factor in the TM/LP calculation [0.5] Inrease[0.5]becauseincreasingTavgmeans(increasingS/Gpressure an a higher Tc [0.5]. I Increase [0.5) due to change in ASI caused by rod withdrawal [0.5].
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REFERENCE M2-OP-RO-I&C-2380-1 (RPS SD), p 26-31 and fig. 2 K611 ..(KA's) ANSWER 3.09 (2.50) Power range subchannel deviation alarm [0.25] TM/LP trip [0.3] - bypass [0.1] High-power trip-[0.3] - bypass [0.1] LPD trip [0.3] - bypass [0.1] High power reading for upper NI [0.25] No alarms [0.4] Bypass low reactor coolant flow [0.4] REFERENCE M2-OP-RO-I&C-2380-1 (RPS SD) 012000A205 ..(KA's) ANSWER 3.10 (3.00) Six valves, - 2 at dumps & 4 TEV's due to QO signal [TT & Tavg
> 557 degrees F] ,
b. Only one - after the reactor trip core decay heat will drop rapidly, one valve has the capacity to remove the amount of decay heat after 5 minute Loss of offsite power will render condenser dumps inoperativ answer for a) is two and b) is two since both ADV's have same REFERENCE M2-OP-I&C-2386 (RRS SD), p 10-1 < 041020K417 ..(KA's) ' l ANSWER 3.11 (1.00)
. .. . < - AsA &mt b. 3 l c. 3 sN each]
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. * * ,1 INEIR!!MEllIE_AND CQEIRQLE Page 38
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~ REFERENCE M2-OP-RO-ELECT-2346 (EDG SD), p 109-11 K302 ..(KA's) (***** END OF CATEGORY 3 *****)
. . . . . ' . ~ 4- . PROCEDURES - NORMAku_AHEQBMAL, EMERGEEQ1 Page 39 .
AND RADIOhQQLGAL_QQETHQh
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ANSWER 4.01 (2.00) a. FALSE b. FALSE c. TRUE d. TRUE [0.5 each] REFERENCE M2 ACP 2.06A, p 12,1 K102 ..(KA's) ANSWER 4.02 (2.00) 1. PZR level at least 20% 2. Heat removal from at least one S/G 3. RCS Subcooling >30-F ' 4. Loop delta-T between 10 and 45-F 5. Tc const. or decreasing 6. Th const or decreasing [four required, 0.5 each] REFERENCE M2 AOP 2553, p K101 ..(KA's) ANSWER 4.03 (3.00) 2.On G u=-_u % 4 :--&12rE.AMS(. M 5W~b a.5 Open EBFS fan suction dampers ( 2-EB-[d- @0<r)cr 2 "O 40 41 '" 23 Start EBFS fan A or V Close Cond. Air Removal Fan Discharge to Unit i stack (1-EB-55 and 56).
4f Close Cond. Air Removal to U-2 stack (2-EB-57) ht Stop Cond. Air Removal Fans F55 A and [four required, 0.5 each) b. This prevents back draft of Unit i stack gases into Unit [1.0] REFERENCE M2 EOP 2528, p 1 K302 ..(KA's) i (***** CATEGORY 4 CONTINUED ON NEXT PAGE *****)
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. . . ;. . PROCEDT)RES - NORMAL. ABNORMAL. EMERGEPCY Page 40 , AND BADlQLQGlGAL CONTHQL .
ANSWER 4.04 (4.00) a. Exceeding PDI . Two or more CEA's do not move into the core follcwing a reactor tr Unanticipated reactor cooldown 4. Unexplained increase in reactivity when shutdown or refuelin [0.5 each] (2.0) b. Open boric acid pump discharge (2CH-514) 2. Start both boric acid pump . Close boric acid pump recire valves (1CH-510 and 511).
4. Establish ma charging flo [0.5 each] REFERENCE M2 AOP 2558, p 2, K302 000024K301 004000K015 ..(KA's) ANSWER 4.05 (2.00) 1. Verify receipt of annunciator . Initiate corrective actions resulting from an emergenc [1.0 each] REFERENCE M2 ACP 6.01, p A103 013000K001 ..(KA's) ANSWER 4.06 (2.00) If the U/V relays are reset when the EDGs are operating in a LNP condition and a SIAS subsequently occurs [1.0], all ESF loads required by SIAS initiation will be energized simultaneously rather than sequentially and overload the respective ED [1.0] REFERENCE M2 OP 2346A, p A010 ..(KA's)
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. PROCEDUBES - NQBMAh ABNORMAL, EMERGENCY Page 41 l
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ANSWER 4.07 (2.00) I _ , . . _ . - , 1 a. 1. QT level 2. QT pressure (E. 'MUA-?4 A p-r * --Q i 3. QT temperature 1 4. Acoustic monitor l
b. 1. containment pressure (<2 psig) l 2. containment temperature (<120-F) ' 3. containment radiation 4. containment sump level [0.25 each] REFERENCE M2 EGP 2525, steps 3.3d, K301 ..(KA's) l l l ANSl/ER 4.08 (1.00) By opening the 125vde supply breaker to the solenoid operated air valves (allowing the MSIV to fail shut).(#18 in 201A/B-IV(DV-20)) [1.0] REFER E y 1 M2 AOP step 4.13. , ,4 op p er,9 4 jusp y' 3 l 039000A010 ..(KA's) < O*CdPA*A)L'**A ANSWER 4.09 (3.00) )1 a. PZR presg, creasing to 1600 psia [0.5] OR cont. press increasing to 5 psir3 .5] , b. PZR press decreasing to 1600 psia [0.5] l c. Subcooled RCS [0.5] d. Via PZR aux. spray AND HPSI pump (facility 1) [1.0] REFERENCE M2 EOP 253 A101 013000A201 ..(KA's)
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l . PROCEMEES - NOEMAL, ABNORMAL. EMERGEILQ1 Page 42 l , AflLEADIOLOG.LGAL CONTROL i
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ANSWER 4.10 (3.00)
(o. k ed~ss> &-^ . m A.P5
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a. 1. Rod dropped NI alarm 7.ce4 m e - % - 2. Rod dropped Reed switch alarm P 'W W W # ~ NA 3. Green or yellow light on the core mimic display 4N M #^- * b/w 4. Metrascope '*-<-* W e-*t 5. PDIL alarm [four required, 0.25 each] o "j :"----g b. PZR level decreases below 20*4 [[0.5] M Q g" <70%, reducing turbine load and boration (NO rod motion allowed)L1.0) d. Commence an orderly shutdown per plant procedures (OP 2205&2206)[0.5] REFERENCE M2 AOP 2556, p 2, K304 ..(KA's) ANSWER 4.11 (1.00) Exceeding the capacity of the CVCS to maintain PZR_ level [0.5) Level increasing in one or more of the SIT's M'High pressure alarm on SIT side of a loop check valve [0./5]. REFERENCE
, M2 AOP 2568, p K015 002000K405 ..(KA's) (***** END OF CATEGORY 4 *****) (********** END OF EXAMINATION **********)
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. . ~:. ;, I U. S. NUCLEAR REGULATORY COMMISSION ! .
SENIOR REACTOR OPERATOR LICENSE EXAMINATION l
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FACILITY: _ t11 L L S T O NE _ ? _ _ _ .. _ _ _ _ _._ _ _ REACTOR TYPEr _ EWE'-GE__.,_____ _ ___. _ DATE ADMINISTERED: _D9f91/11______,_________ EXAMINER: _g]LK 2 _p3________________ CANDIDATE: __(1_4_5 D _r _ _(<>f __ _ _ _ _ _ _ _ _ _ JNSIBUCIJgN5_IQ_C@Np]991El Use separate paper for the answer Write answers on one side onl Staple question sheet on top of the answer s h e e t Points for each question are indicated in parentheses after the questio 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 the examinati]n start % OF CATEGORY % OF CANDIDATE'S CATEGORY __Ye6ue_ ._1g196 ___SCggE___ _Ye6UE__ ______________C@lEGg@Y_____________ _29199__ _2Dt99 ___________ ________ THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS _29199__ _29199 ___________ ________ 6, PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION _29199__ _29 99 ___________ ________ PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL l CONTROL _29299__ _29199 ___________ ________ ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199199__ ___________ ________% Totals 1 Final Grade l All work done on this examination is my ow I have neither given nor received ai ___________________________________ Candidate's Signature
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NRC RULES AND GUIDELtNF.C FOR LICENSE EXAMINATIONS
, Durf.ng the administration of thi s examination the f ollowing rules apply: ' Cheating on the exnminatiCn means an automatic denial of your applicaticn-ard could result in more severe penaltie . Restroom trios are to be limited and only one candidate at a time may leav You must avoid-all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil gnly to facilitate legible reproduction Frint your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write gnly gn gne sidg !
of the paper, and write "Last Page" on the lust answer shee . Number each answer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad ant place finished answer sheets' face down on your desk or tabl ' l 12. Use abbreviations only if they are commonl y used in f aulli ty 1_i t er atur _ 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require . Show all calculations, metnods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be give Therefore, ANSWER ALL PARTS OF THE i QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl . You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examinatio This must be done after the examination has been complete , , _ - _ . . . - ~ . . .- _ _ . . . . . . . ..
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180 When you complete your ex ami na t i on , you shall:
- . - Assemble your examination as f ollows: (1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are partaof the answer, Turn in your copy of the examination.and all pages.used to answer the examination question Turn in all scrap paper and the balance of the paper that you dic not use for answering the question Leave the examination area, as defined by the. examiner. If after leaving, you are found in this aree while the examination is still in progress, your license may be denied or revoke !
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S.'.' THEORY OF NUCLEAR POWER PLANT OPERATION 2 _FLUIpS1 _gNp PAGE 2
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THERMODYNAMICS
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QUESTION 5.01 (1.50) How will each of the f ol lowing affect the results of a secondary calori-metric power calculation? Limit your answer to CALCULATED LOWER THAN ACTUAL, CALCULATED HIGHER THAN ACTUAL, or CALCULATED THE SAME AS ACTUA Consider each case separatel Measured feed water temperature is 10 degrees lower than actual feed water temperatur Measured steam generature pressure is 30 psig lower than actual steam generator pressure, Measured feed water flow is IES lbm/hr higher than actual feed water flow, i GUESTION 5.02 (2.00) Answer the following questions TRUE or FALS Without reducing turbine load, machine heating on the main generator can be reduced by reducing VAR loading on the generato When paralleling two AC power sources, the synchroscope should be moving slowl y in the FAST directio t While in parallel operations, if the diesel voltage is raised to a higher value, the diesel generator will pick up a larger share of { the reactive load.
l If the diesel generator is carrying an isolated vital bus, the governor control is used to adjust bus loa GUESTION 5.03 (1.50) Stable natural circulation conditions exist within the RCS with the following parameters: Thot - Tcold = 25 F SG pressure = 685 psig i That Subcooled Margin indicates 40 F subcooled , I Determine RCS pressur '
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Sil THEORY OF NUCLEAR POWER PLANT OPERATION1 _FLUlpS 1_@Np PAGE' 3
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QUESTION 5.04 (3.00) HOW would the fuel center line temperature change (INCREASE, DECREASE, or REMAIN THE SAME) in each of the following situations? Justify your answe Consider each situation separ at el Power decreases with constant Tave Tave increases with constant power Core age increases with constant power PZR pressure increases with constant power i OUESTION 5.05 ( .75) For an unrodded, uninstrumented fuel assembly, the hold down force of the Ouel alignment plate is transmitted to the core support as described by which ONE of the f ollowings Fuel alignment plate to flower assembly, to hold down springs, to upper end fitting, to fuel pins, to lower end fitting, to core ; , support plat Fuel alignment plate to upper end fitting posts, to upper end fitting, to guide tubes, to core support plate Fuel alignment plate to flower assembly, to hold down springs, to guide tubes, to core support plate Fuel alignment plate to flower assembly, to hold down springs, to upper end fitting, to guide tube, to lower end fitting, to core support plate * i QUESTION 5.06 (2.50) WHAT TWO RCS conditions must be present for the value of MTC most to be posi tive?
(1.0) HOW does the value of MTC change as reactor power is increased? Pro-vide TWO reasons and EXPLAIN each reaso (1.5)
QUESTION 5.07 (1.25) WHAT is "split enrichment" and WHY is it used in fuel assemblies?
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QUESTION 5.08 (2.50) WHAT FOUR factors determine whether a Xenon oscillation will be con-vergent or. divergent? (1.0) HOW does an operator dampen a Xenon oscillation? Include how ASI and ESI are use (1.5) QUESTION 5.09 (3.00) The plant is at end of cycle operation, the react'or at'100% power, all rods are all the way out, boron concentration at 10 ppm i n the RCS, and T(avg) 10 degrees less than Treferenc EXPLAIN HOW and WHY reducing the power to 95% would affect ASI for the f ollowing means of reducing powe Consider each case separatel ' Control rod insertion (1.0) Boron addition (1.0) Rai si ng T(avg) (1.0) QUESTION 5.10 (3.00) Millstone Unit 2 i s i n Mode 3, BOC, boron concentration is 900 ppm, all shutdown groups are withdrawn, and the-actual reactivity present in the core is minus 4% delta-k/k. A dilution of the boron concentration in-creases source range counts from 100 CPS to 196 CP Concurrent with the dilution, Xenon reactivity changes have added + 1000 pcm to the cor CALCULATE the new boron concentratio Assume a constant differential , boron worth of 10 pcm/ pp State all assumptions and show all wor I l l (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
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'.* 'T.HEORY'OF NUCLEAR POWER _PL@NT_OPERAT[ON t _FLUlDSt _AND PAGE 5 THERMODYNAMICS ~ . . .
QUESTION 5.11 (2.00) Compare the CALCULATED Estimated Critical Posi t i on (ECP) to the ACTUAL control rod position for a startup to be performed 4 hours after a trip from 100% power if the f ollowing events /condi tions occurre Consider each independently. Limit your answer to HIGHER THAN, LOWER THAN, or SAME as the EC O,hh) One Occter ccc! nt pump .; stopped t-c -ir,ut a pric- to c-iticelit t The startup is delayed until 8 hours after the trip The condenser steam dump pressure setpoint is increased 25 psig Condenser vacuum is reduced by 3 inches of mercury All steam generator levels are being raised by 5% QUESTION 5.12 (2.00) Assume one Reactor Coolant Pump trips at 30% power without a reactor protective system actuatio Indicate whether the f ollowing parameters will INCREASE, DECREASE, or REMAIN THE SAM l 1 Flow in the reactor coolant loops with the RCPs still running d ' Reactor vessel delta P Core delta T The steam flow in the steam generator on the other sion l
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it_iP,(@@l_SYSIEOS_DESIGNt_CQNIB06t_@ND_lNS18UdENI@llQN PAGE 6
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QUESTION 6.01 (2.50) WHAT FOUR (4) regulating control rod interlocks / limits are in , effect when the system is in MANUAL INDIVIDUAL control? (1.2) ; WHAT are the two (2) instrumentation si gnal s/condi ti ons that i could provide a DROPPED ROD annunciator? (0. 8) If a loss of 125 VDC control power to the trip circuit breakers (TCBs) occurs, what component, if any, will ensure that the TCBs open if a trip signal is generated? (0.5) QUESTION 6.02 (2.80) The plant is at 100% power and all controls systems are in automati The controlling pressurizer l evel channel fails low. What system responses will occur and what reactor trip signal, if any, will be generated if no l operator action is performe Setpoints are not require i t QUESTION 6.03 (2.00) In the event of a loss of coolant accident (LOCA) that gradually depressurizes the RCS, STATE the order in which the emergency core l cooling systems (ECCS) will inject into the RC Setpoints are not require !
! During a LOCA, all automatic ECCS function properly, pressurizer level stabilizes at 30% and the RCS pressure stabilizes at 1000 psi WHAT is the approximate break flow rate? Justify your answer using Figure QUESTION 6.04 (2.00) What conditions must exist for the Reactor Regulating System (RRS)
signal to control the condenser steam dump and turbine bypass valves?
(0.5) When is the quick open permissive switch used to prevent a quick open signal from opening the atmospheric dump valves? (0.5) In regard to the steam dump / turbine bypass valve operation, why is RRS Channel X selected while passing through 557 F during either an increase or decrease in power? (1.0) (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
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QUESTION 6.05 (1.50) The powe feedwater control system is in manual with the plant holding at 80% The feedwater regulating bypass valve for Steam Generator Number 2 drifts fully ope Assuming no operator action, explain the feedwater system response until equilibrium conditions are reache State what reactor trip, if any, would occu QUESTION 6.06 (2.40) What effect (INCREASE, DECREASE, NO EFFECT) wi'11 the f ollowing events have on the Thermal Margin / Low Pressure Trip Setpoln Consider each separ-at el Assume the plant is at 100% powe Tcold Loop 1 fails LOW ASI changes from 0.0 to - A RCP trips
- RCS pressure increases 25 psig A Linear Power Range Channel (Safety) fails high Delta-T PWR Calibrate pot is reduced QUESTION 6.07 (2.00)
As a result of a complete loss of instrument air, how will the f ollowing valves respond (FAIL OPEN, FAIL CLOSED, NO EFFr.CT). Service Water Outlet Valves or the RBCCW heat exchangers(2-RB-13.1A/B) Atmospheric Dump Valves (2-MS-190A/B) Containment Isolation Valve (2-CH-515) Charging Header Supply Valves (2-CH-518,519) Primary Makeup Water Valve (2-CH-210X) Boric Acid Makeup Valve (2-CH-210Y) Dele h d ; " 2! ^ St;2- ;1 oti or '/ ; ' ' ; ' '2 MC '4^/^; RCP Control Bleedof f Valve (2-CH-506)
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GUESTION 6.08 (2.30) If the seal flow reaches 10 gpm on one RCP, WHY will the bleedoff ex-cess flow check val ve close? (O.S) Fill in the approximate values for pt'mp seal pressures while at normal operating temperature and pressure ie the RCS (No. I seal is the lower seal). (1.8)
: Lower : Middle l Upper l l Seal : Seal l Seal l l Fails Fails l Fails : ----__--_-____________--_______-______---___-__-____---
l Presure Below l l l l l Middle Seal l l ! l _-___-_--___-____-__-____-_-_-______--_____-_____-_____ Pressure Below l l l l l Urper Seal : i
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Pressure Below : : : : Vapor Seal : : l
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QUESTION 6.09 (2.50) Answer the following questions TRUE or FALSE regarding the Inadequate Core Cooling Syste The delta T between the heated a r. d the unheated thermocouple decrease if they are surrounded by steam, A shorted thermocouple, regardless of plant power, will provide a reading of approximately the normal 100% power readin I An unheated thermocouple at 600 F will generate a low l evel alarm, l l With two phase mixture in the RCS, indicated level will be greater than actual level if an RCP is running, A pressure margin display with a negative (-) sign indicates superhea (***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
5.' hLANT SYSTEMS DESIGN 1_ CONTROL 1_AND_ INSTRUMENTATION PAGE 9
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. QUESTION 6.10 (2.00) LIST FOUR (4) independent components which could potentially leak radioactive water in the Reactor Building Cl osed Cooling Water (RUCCW) Syste (1.0) WHAT TWO (2) indications would the operator have available to i denti f y a leak into the RBCCW system? (0,5) WHAT TWO (2) sets of valves in the RBCCW system open on an SIAS signal? (0,5) OUESTION 6.11 (3.00) Answer the f ollowing questione regarding the containment spray syste a. A large break LOCA occur Containment pressure quickly reaches 35 psig but no CSAS signal was generated. Assuming no operator action, WHAT system or components are used to reduce containment pressure? tt+r5+(0.6) After a large break LOCA, why are the containment spray minimum flow valves (SI-659,660) closed?
+975+(0.6) "^" .
centai ment spray pump .: bring tasted t o e., aw- m vg ci eL. lit 7 OC
#ile the plant is at 1 0 0*/ p c .. c c . ^
,Icss ci ac ,wl g c -,- iLNP; cccu ; and ; -inute lcter CS^C :-i g n a l is generatc CXPL^IN M W-thc tzsted pu-- WILL c- WILL NOT 5te; t I . v a, thu CS,^G nawio ii.Gi WHAT valve realignments automatically occur when a Sump Recirc Actuation Signal is initiated?
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7.- hROCEDURES - NORMAL t_ABNORMALt_ EMERGENCY _AND PAGE 10
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QUESTION 7.01 (1.00) Answer the following question in accordance with AOP 2551, Shutdown From Outside the Control Roo While shutting down from outside the control room and with automatic boration unavailable, WHAT is the source of makeup to the RCS and WHY is this source selected? QUESTION 7.02 (1.00) Answer the following question in accordance with 'OP 2201, Plant Heatu While placing the volume control system in service during a plant heatup, the operator is cautioned to maintain the RCS pressure below 265 psi a by operation of the back pressure control valve EXPLAIN WHY thi s is require QUESTION 7.03 (1.50) In accordance with AOP 2553, Plant Cooldor:n Using Natural Circulation, assume during a natural circulation cooldown that auxiliary spray becomes inoperabl GIVE three (3) alternate methods of depressurizing the PC QUESTION 7.04 (2.50) Answer the following questions regarding OP 2207, Plant Cooldown, Over a two hour period, the RCS steadily cooled down from 350 F to 280 Explain whether or not a cooldown limit has been exceeded.
Oelb !' :-1 7 tF cc "C"; crc ur-i.g, #ich c eab i r e t i mii wi
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U"== a.li c' ford th: greatant prc;;urizar ep. a , capebil.iy~ 11miH Prior to initiating auxiliary spray, why is letdown flowrate maxi-mized while charging flow is minimized?
'O.5) [3.0)
QM d . Whcr nrcurir; aux;11;ry ; pray, i" "' mu s t Chargir; l lc a d a r Cupply '/o l v c ;
' 2 C;; 510 e,- 2 C;; 517; Le micaed? M97td (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****) . - .- .~- .= . - - . ' '7. PROCEDURES - NORMAL t_ ABNORMALt_EMERGENCV_AND PAGE 11 '. ' ' , , '8@D1Q(QGlG86_CQUlBQL . .
QUESTION 7.05 (7,00) Answer the following questions regarding OP 2203, Plant Startu With the reactor at 530 F, an inadvertant opening of the condenser steam dump valves occurs resulting in a 20 F cooldown of the RCS before the valves are close One of two actions must be take WHAT are the TWO (2) actions? Reactor power is 6% when all feedwater flow was lost for 8 minute r
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When re-establishing feedwater, flow is to be limited to less than 600 gpm if steam generator temperature is greater than 212 F and and level is less than 45%. What operator action should be taken, if any, and what adverse plant condition may result, if any, if max-imum f water flow is initiated? i l QUESTION 7.06 (2.50) a, In accordance with AOP 2564, Loss of RBCCW, if a service water header r
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rupture nas resulted in a loss of cooling water to the RBCCW heat exchangers, WHAT THREE (3) conditions may develop which require a l
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1 The "A" service water header becomes inoperable and i s removed from , servic Appropriate RBCCW component realignments are being made to allow the "B" service water header to act as the heat sink, in accordance with Technical Specifications, WHAT restrictions, if any, are placed upon plant operation in Mode I while in this configuration if the "B" diesel generator was out of service for maintenance before
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I the "A" service water header became inoperable?
(1.0)
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[. kROCEDURES - NORMAL t_ABNgRMAlt_ EMERGENCY _AND PAGE 12
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QUESTION 7.07 (3.50) Answer the following questions regarding AOP 2569, Steam Generator Tube Lea If a tube leak has occurred, WHAT FOUR valves (sets of valves) will close? (1.0) WHEN during a tube leak'should the reactor be tripped? (0,5) WHAT TWO (2) limits, as verified by Chemistry Personnel, require a unit shutdown to commence? (1.0) If the turbine driven auxiliary feedwater pump or the steam gener-ator atmospheric dump valve for the affected steam generator are operated during a tube leak, WHY must the duration of operation of these systems be recorded? (0.5) If operation with a tuae leak continues long term, WHY must the num-ber of circulating pumps and blowdown flowrate be checked? (0.5) QUESTION 7.08 (2.00) Answer the following questions regarding AOP 2571 Inadvertent ECCS Initiation, If an actuation occurs during solid condition in the pressurizer, WHY must the charging pumps be secured? If an ESAS Signal has been overridden by use of the equipment hand-switch, WHAT effect, if any, will future ESAS Signals have on that equipment? WHY should the TBCCW heat exchanger manual inlet valves be throttled prior to overriding an SIAS Signal? After resetting an SIAS Signal and restoring the safety injection systems, WHY must the RWST recirculation header drain valve to the PDT (2-SI-661) be opened?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
1_j 69CggUBgg_;_yg503(2_8pyggd361_EdgBGgyCy_8dg PAGE 13
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8991969GIC96_C9 BIB 96 .
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QUESTION 7.09 (3.50) Answer the f ollowing questions regarding EOP 2532, Loss of Primary Coolan Provide the SIX (6) checks that the operator uses to-verify natural circulation flo (1.5) WHAT FOUR (^) condit. ions must exist before a HPSI pump can be stopped?
(1.0) If voids exist in the reactor vessel during natural ci rcul a t i on . HOW will pressurizer level and pressure respond if an RCP is star'ted?
(0.5) i WHAT personnel hazard exists while performing boron precipitation ;
control?
(0.5) _
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GUESTION 7.10 (3.50) I Answer Action the f ollowing questions regarding EOP 2525, Standard Post Trip ; i EXPLAIN HOW the operator would manipulate the controls of the CVCS to perform a boration if two CEAs failed to inser (1.0) After checking oressurizer level, WHY does the operator check RCS subcooling? ;
(0.4) GIVE TWO (2) indications HOW an operator verifies that a PORV or pr7ssurizer safety is not open? (0.8)
, If, after tripping the turbine generator megawatts does not go to !
zero, WHAT action should the operator perform? (0.4) t Provide THREE (3) reasons WHY f eedwater should be added slowly during low steam generator water leve (U.9) l (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
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. * '21__BBOggpUBE3_;_NgBD961_9BNgBU961_EDEBGENgy_9NQ PAGE 14 - . '6891969 GIG @6_QgNIBg6 ..
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QUESTION 7.11 (2.00) Answer the following questions regarding EOP 2540 D, Functional Recovery of Heat Removal, When, if at all, is it permi ssibl e to NOT isolate a ruptured steam nerator? (0.75) WHY ; restarting the RCPs cause pressurizer level and pressury to decrease? (0.75)
, To establish heat removal using Once Through Cooling, WHY are two high pressure trip bistables pulled and one bypass key installed?
(0.b)
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(***** END OF CATEGORY 07 *****)
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-dii-QDdldlSIB@llyE_PEQQEDURE@t_QQgDillgNgt_gggbidll@llgME . *
PAGE 15
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UUESTION 8.01 (2.20) Discuss the relationship between Limiting Condi ti ons f or Operations (LCO), Limiting Safety System Settings (LSSS), and Safety Limits in terms of pre-venting release of radioacti vi ty. to the environment.
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QUESTION 8.02 (1.00)
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- Below are the dates of the Quarterly Battery Surveillence (2736B-1): 7/6/86 10/5/86 1/5/07 5/6/87 { EXPLAIN yiETHER OR NOT a surveillance interval has been exceeded, and if so, WhlCH ON , f GUESTIJN B.03 (2.00) With the plant at 100*/. power , a re-anal ysi s of steam generator tube exam-ination data, col l ected f rom the l ast ref ueli ng outage, i der.t i f i es th at one tube which had contained a repairable defect had not been repaired ' prior to declaring the steam gene.*ator operable. WHAT action, if any, is required in regard to plant operatin7 mode and WHY? Refer to the attached { Technicial Specificatio ; i
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QUESTION 8.04 (3.00) Answer the following quastions TRUE or FALSE with respect ACP-DA-206 Station Taggin If the "Operator in Attendance" must leave for a duration of 30 min-utes, the valves and breakers shall be tagged as listed or, the Tag Loc Sheet, SF 21 Valves are red tagged shut and breakers are red tagged open unless the tag position is written on the red ta o,9d u . sur or= sur a w a vo r ur T ur niew wavan ina y w= 4 444=w au m icii wow = or e 4 =ced
:o; ir the cosa if nc estcretiGn 1- requii ed 'i.e. c., tended Outege cn
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5ystGOL nGt iG BarviCE Q1 i & f uel ling Dwiayc} aniw kile 3 3 tiltec k b tile isu Dicck end the reesc, is given ir thc "nestcretisn PericraeF -biec , A blue tag and a red tag may be rt,ttached to the same piece of squipment at the same tim Any number of red tags may be attached to the same pi ece of equipment < at the same tim Restoration involving switch or breaker position will always be in the "AUTO" or "OPEN" posi tion respectivel QUESTION 8.05 (1.50) WHAT actions and notifications must be completed if Unit 2, while in Mode
,
3, is experiencing a pressure of 2775 psig due to equipment malfunctions and operator neglect? OUESTION 8.06 (3.00) Answer the following questions regarding ACP 6.01, Control Room Procedure, Following a refueling outage, WHAT THREE individuals 'by position) may give authorization to take the reactor critcal? ( 1. 5 ) Prior to resetting protective relays following a reactor trip, WHAT , ) TWO (2) precautionary actions should be performed? (1.0) , With only one operator at the controin, WHEN, if at all, may the
- operator leave the surveillance area during Mode 1 operations? (0. 5 )
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6. ADMINISTRATIVE
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PROCEDURESt ,CQNDlTIQNSt ,AND, LIMITATIONS PAGE 17
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QUESTION 8.07 (2.00) At 0930, with the plant at 85% power, the B HPSI pump was taken out of service for maintenanc At 1315 an uncomplicated reactor trip occurs ,t because an I&C techni ci an tripped an incorrect bi stabl e. The HPSI pump is expected to be back in service within three hour EXPLAIN WHETHER OR NOT a reactor startuo may be commence QUESTION 8.08 (2.00) In accordance with Technical Specifications, temporary changes to procedures at Unit 2 may be made if WHAT THREE provisions are met? QUESTION 8.09 (2.00) l An IbC technician has been trouble-shooting a problem with the rod control system. The PPO requests permission from the SCO to allow the technician i to insert Group 7 rods one step to test the rod control system. With the I plant at 90% power, SHOULD the SCO allow the technician to operate the l controls? Justify WHY or WHY NO k QUESTION 8.10 (2.40) : j While in Mode 6, WHAT conditions would require both independent
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shutdown cooling loops to be operable?
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QUESTION 8.11 (2.40) I According exist? to Technical Specifications, WHEN does containment integrity i QUESTION 8.12 (1.50) ' WHAT is the Technical Specification bases for establishing a limit for ' the minimum temperature for criticality? '
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I (************* END OF EXAMINATION ***************) . i ^
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~ ~ . EQUATION $HEET . . . . . ;. - -
f a ca v o s/t Cycle efficiency o (Networt
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out)/(Energy in)
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2 - e = og s = V ,t + 1/2 at
E = ac KE = 1/2 av a = (Vf - V,)/t A = AN A=Ae'*g PE = agn l Vf = V, + at w = e/t x = an2/t1/2 = 0.693/t1/2
y ,yg - 1/2'##*E(*U?)(t)3 h
-
[(t1/2) + (*b)3 '
aE = 931 an I'= I n e'*
, ,
Q = aCpat Q = UAat I * I,e'"* , Pwr = Wfah I=I n 10**/UL l TYL = 1.3/u P = P 10 sur(t) HYL = -0.693/u p = p et lT o -
$UR = 26.06/T SG = S/(1 - K,ff)
CR, = s/(1 - K,ff,) l
~
SUR = 25e/t' + (s - o)T CR j (1 - K,ffj) = CR 2 (I ~ IW2) ' - T = (1*/s) + [(s - e)/ o] M = 1/(1 - Kg ) = CR j/CR, T = 1/(o - 8) M = (1 - K,ff,)/(1 - K,ff)) T = (s - o)/(lo) SDM = (1 - K,ff)/K,ff a = (X ,ff.1)/X ,ff = 4K ,ff/K,ff t' = 10~6 seconds
-
_ T = 0.1 r wonds'I e = ((1*/(T K,ff)] + (f,ff (1 / + AT)] Idl1"Id2 ,2 gd
F = (r+V)/(3 x 1010) Idjj 22 l
t = oN ' R/hr = (0.5 CE)/d (meters) I R/hr = 6 CE/d2 (feet) Water Parameters Miscellaneous Conversions
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1 gel. = 8.345 lbe.- I curie = 3.7 x 1010 dps 1 ga:. = 3.78 liters 1 kg = 2.21 lbe
- 1 ft< = 7.48 ga I hp = 2.54 x 10 3 8tu/hr
Density = 62.4 lbs/ft3 1 es = 3.41 x 106 8tu/hr
'Oensity = 1 gn/cd .
lin = 2.54 cm Heat of vaporization = 970 Stu/lbm 4 = 9/5'C + 32 Heat of fusion = 144 8tu/lbe 'C = S/9 ( T-32) 1 Ata = 14.7 psi = 29.9 in. H STU = 778 ft-lbf 1 ft R 0 2
= 0.4335 lbf/i ._.. . . .
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pt: .ti BHP S.G. - :
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30 - 200 A;
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NPSH 20
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80 _
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10 : Efficiency r'
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60 F I
.h -
y Total Head
- = "
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30g
- 2000 -; i
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' ts e-20 - 1000 -
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- Pg 0- i i 1 i i g . . . . . 0 . W 700 Gallons Per Minute
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. - __.- . . - . - - - - - . , __ ._ - _ . . . . . ' .. .. May 12, 1o70 '
REACTOR COOLANT SYSTEM STEAM GENERATORS
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LIMITING CONDITION FOR OPERATI0'. 3.4.5 Each steen generator shall be OPERA 8L APPLICABILITY: MODES 1. 2 and ;
ACTION:
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With one or more steam generators inoperable. restore the . inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200* SURVE!LLANCE REQUIREMENTS
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4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following Augmented Inservice Inspection Progra , 4.4.5.1 Augmented Inservice Inspection Program 4.4.5.1.1 5 team Generator 5 anele 5 election and Insoection - Each steam
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generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4- . 4 . 5.1. 2 Steam Generator Tube 5emple Selection and Inhoeco,im - The steam generator tube minimum semple size. Inspection resu' t c' ass' f cation, and the corresponding action required shall be as specified in Table 4.4-6, the inservice inspection of steen generator tubes shall be perfonned at the frequencies specified in Specification 4.4.5.1.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification
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L.4.5.1.4. The tubes selected for each inservice inspection shall include i et least 31 of the total number of tubes in all steam generators; the tubes
- nelected for these inspections shall be selected on a random basis except
! Where experience in similar plants with sisthe water chemistry i indicates critical areas to be inspected, then at least 50% of
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,the ' tubes inspected shall be from these critical area The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:
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BILLSTONE - UNIT 2 3/4 4-5 __ _ -- __ __
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REACTOR COOLANT SYSTEM
. l SURVEILLANCE REQUIREMENTS (Continued) . l ["' ~' All nonplugged tubes that previously had detectable wall
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penetrations (>205). Tubes in those areas where experience has indicated potential problem . A tube inspection (pursuant to Specification 4.4.5.1.4.a.8) l shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current ] probe for a tube inspection, this shall be recorded and i an adjacent tube shall be selected and subjected to a tube Inspectio The tubes selected as the second and third samples (if required by Table 4.4 6) during each in ^rvice inspection may be subjected to a partial tube inspection provided: The tubes selected for these samples include the tubes .' from those areas of the tube sheet array where tubes with imperfections were previously fcun . The inspection include those portions of the tubes where imperfections were previously foun The results of each sample inspection shall be classified into one of the following three categories: Category Inspection Results
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C-1 Less than 55 of the total tubes inspected i are degraded tubes and none of the inspected tubes are defectiv I C-2 One or more tubes, but not more than 15 of the total tubes inspected are defective, or between 55 and 105 of the total tubes inspected are degraded tube '
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C-3 More than lot of the total tubes inspected are degraded tubes or more than 15 of the inspected tubes are defectiv Note: In all inspections, previously degraded tubes must exhibit significant (>105) further wall penetrations , to be included in the above percentage calculations.
1 MILLSTONE - UNIT 2 3/4 4-6
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, REACTOR C00taNi SYSTEM SURVE!LLANCL REQUIREMERTS (Continued) .
4.4.5.1.3 InsDettien Frequencies - The above required inservice inspections of steam generator tunes snall be performed at the following frequencies: The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more tnan 24 calendar months after the previous inspection. If two consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all insocc-tion results f alling into the C-1 category or if two consecutive inspections demonstrate that previcusly observed degradation has not continued and no additional degradation has occurred, tne* inspection ir.terval may be extended to a maximum of once per 40 month , If the results of the inservice inspection of a stene generator . conducted in accordance with Table 4.4-6 at 40 montn intervals . f all into Category C-3, the inspection frequency shall be t increased to at least once per 20 months. The increase in I inspection frequency shall apply until the subsequent insoec-tions satisfy the criteria of Specification 4.4.5.1.3.at the interval may then be extended to a maximum of once per 40 i
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month Additional, unscheduled inservice inspections shall be perforced on each steam generator in accordance with the first sample inspection specified in Table 4.4-0 during the shutdown sucsequent to any of the following conditions: Friamry-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welos) in excess of the limits of Specification 3.4. I
, ' .A seismic occurrence greater than the Operating Basis , Earthquak . A loss-of-coolant accident requiring actuation of the l engineered safeguard . . A main steam line or feedwater line brea MILLSTONE - UNIT 2 3/4 4-7 Amenoment No. 22: 27, ?? ,10) - . ._ _ __ _ . _ _ - .. _ - - -
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REACTOR COOLANT SYSTEM,
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SURVEILLANCE REQUIREMENTS (Continued) _ 4,4.5. Acceptance criteria
, As used in this Specification Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfection . Deoradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tub . Deoraded Tube means a tube containing imperfections- >20t of tne nominal wall thickness caused be degradatio ' , t Deeradation means the percentage of the tube wall thickness affectes or removed w'y degradatio . Defect means an imperfection of such severity that it exceedt i sne plugging limit. A tube or sleeve containing a defect is j ;
defectiv l Plugeino t.imit means the imperfection depth at or beyond which tne tune or sleeve shell be repaired because it may become I unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thiekness for tubes.* l Unserviceable describes the condition of a tube if it leaks 1 or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.1.3.c abov . Tube Inspection means an inspection of the steam generator
- tune f rom tne point of entry (hot leg side) completely around
' the U - Bend to the top support of the cold le The steam generator shall be determined OPERABLE after completing the corresponding actions (plug or sleeve all tubes exceeding the l plugging. limit and plug all defective sleeves) required by Table
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4.4- l MILLSTONI - UNIT 2 3/4 4-7a Amendment No. 22, 37, 32, 89
*The plugging limit for sleeves will be det, ermined prior to next refueling outag . -.
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. - . . . ' , _ April 9 1986 - - - l REACTOR C00LAhn SYSTEM ' ,
i SURVEILLANCE REQUIREMENTS (Continued)
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4.4.5.1.5 Reports
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, Following each inservice inspection of steam generator tube the number of tubes plugged in each steam generator shall be i
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reported to the Coerdssion within 15 day The complete results of the steam generator tube inservice l inspection shall be included in the Annual Operating Report for the period in which this inspection was completed. This report shall include- ) Number and extent of tubes inspecte l i Location and percent of well-thickness penetration for each indication of an imperfectio . Identification of tubes plugged or sleeve , i
' Results of steam generator tube inspections which fall into Category C-3 shall be reported pursuant to 10 CFR 50.7 In lieu of any report required pursuant to Specification 6.6.1, a Special Report pursuant to Specification 6.9.2 shall be ,
submitted prior to resumption of plant operation and shall provide a description of investigations conducted to determine
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the cause of the tube degradation and corrective measures taken i to prevent recurrenc ..
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3/4 4-7b Amendment No. 22.37,52.87.111.
MILLSTONE - UNIT 2
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TABLE 4.4-5
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3 MININLM NUPSER OF STEAM GENERATORS TO BE
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INSPECTED DURING INSERVICE INSPECTION
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Preservice Inspection Yes E .
m No. of Steam Generators per Unit Two First Inservice Inspection One Second & St6 sequent Inservice Inspections One I
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Table Notation: s
* * The inservice inspection may be limited to one steam generator on a rotating sc..edule encompassing 3 N % of the tubes (where N is the number of steam generators in the plant)
e if the results of the first or previous inspections indicate that all steam generators are performing in a like manner. Note that under same circumstances the operating con- . ditions in one or more steam generators may be found to be more severe than those in I other steam generators. Under such circumstances the sample sequence shall be modified I to inspect the most severe conditions.
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ob . e- . . April 9. 1986 TAet.E 4.4-6
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STEAM GUERATOR TISE INSPECTION u - 3ND SAMLE INSPECTION 2ND SAMPLE INSPECTION
$ IST SAM 1.E INSPECTION Action Required '
Sample Sire Nesult Action Required Nesult Action Required Nesult E 4 minimum of C-1 None N/A N/A N/A N/A
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S tubes per C-1 None N/A N/A
" C-2 Nepair defective tubes and inspect C-2 Nepair defective C-l Mone additional 25 tubes and inspect tubes in this S.G.* additicnal 45 tubes C-2 Repair defective in this S.G.* tubes *
C-3 Perform action for C-3 result of first
, 'd sample '
C-3 Perfom action for
. C-3 result of first N/A N/A *
u sample
*
C-3 Inspect all tubes All other N/A in this S.G., repair S.G.s are None N/A defective tubes and C-1 inspect 25 tses in some 5. Perfom action for C-2 but no C-2 result of second N/A N/A k= each other S.S.* additional sample Prompt notification 5.G. are C-3 i k= to NRC peersuant to Additional Inspect all tubes in 10 CFR 50.72 S.G. is each S.G. and repair l x C-3 defective te es.* N/A N/A
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Prompt notification
, " to NNC pursuant to . * 10 CFR 50.72 l ." = N .
5 lNiere N is the nesber of steam generators in the unit, and n is the nuder of steam generators inspected O S=3 n during an inspection
. ? * Repair of defective tubes shall be limited to plugging with,the exception of those tubes which may be sleeved.
, O Tubes with defective sleeves shall be plugged.
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, St_,ldEQQy_DE_NQC(E@S_EOWg6_E(@yI_QPERATIQNt _F(UIDS t_AND PAGE 10-J - ISE80QDy@@d1CS
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ANSWERS -- MILLSTONE 2 -88/01/11-SILK, D.
. ANSWER 5.01 (1.50) Calculated higher than actual [0.53 Calculated nigher than actual CO.53 Calculated higher than actual CO.53 REFERENCE HTFF 2121 B pg 17, 18 EO 5 Heat transfer 191003 K 1.06 3.1/ K 1.08 3.1/ .2 002 020 K 5.01 3.2/ OO7K106 OO2020K501 193OO7K108 ...(KA'S) ANSWER 5.02 (2.00) TRUE TRUE TRUE FALSE CO.5 pts each] REFERENCE Question Bank pg 112 Electrical Theory 2131 pg 43-63 EO 6,9,10 2.7/ .1/ A203 0640002020 ...(KA'S) I l
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. - . - . - - . . . . . .. . . -. *s .- THEORY-OF NUCLEAR POWER PLANT OPERATIONi_FLUI2Si _ANp PAGE 19: + ., 'TbgRdgDyN9DJGS ANSWERS -- MILLSTONE 2 -88/01/11-SILK, ! . .
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ANSWER 5.03 (1.50) Tcold = 503 F corresponding to saturation temperature for 700 psia (0.253 ! Thot = Tcold + 25 = 528 F CO.53 , 40 F subcooled = 528 + 40 = 568 F CO.53 l l 568 F corresponds to 1207.72 psia CO.25]
:
REFERENCE Question Bank pg 91 HTFF 2121 J pg 18 EO 11 ' HTFF 2121 C pg 9,10 EO 5 2.8/ .6/ ] 193OO3K124 193OO8K115 ...(KA'S)
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ANSWER 5.04 (3.00) Decrease [0.253 Smaller delta T required to transfer heat to RCS CO.53 Increase CO.253 Center line temperature responds to RCS temperature
- in order to maintain a constant delta T for heat transfer [0,53 j Decrease CO.253 Fuel swelling and clad creep reduces clad gap which improves the heat transfer across the gap and lowers fuel temp (0.53 a
.! Remains the same CO.253 Pressure has little effect on heat transfer in subcooled fluids CO.53 i J
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. REFERENCE f Question Bank pg 87 ITFF 2121 B pg 3-8
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EO 2 l 2.5/ .4/2.6 i 193OO7K101 193OOOK116 ...(KA'S) '
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THEORY OF NUCLEAR POWER PLANT OPERATIONt,FLUlQSt,AND PAGE 20
, '., ISERdQDYN@dlCS l - ;
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, . f
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' ANSWER 5.05 ( .75) i
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< CO.753 I REFERENCE Question Bank pg 75,76 Fuel Design / Core Power Distribution 2117 pg 11 EO 3 ; i 2.5/ K101 ...(KA'S) r I ANSWER 5.06 (2.50) - High boron concentration CO.53 l Low moderator temperature CO.53 MTC becomes less positive (or more negative) CO.53 i A decrease in baron concentration CO.253 deminishes the posi tive com- i
ponent of MTC CO.253 i As moderator temperature increases (0.253, i ts densi ty decreases and tends to increase reasonance absorption OR neutron leakage CO.253 REFERENCE Question Bank pg 61 - Characteristics pg 11-23 ! i EO 3 l 2.9/ .1/ OO4K103 192OO4K106 ...(KA'S)
ANSWER 5.07 (1.25) i
"Split enrichment" refers to using lower enrichment in the f uel pins currounding the CEA guide tubes than is used in the rest of the assembly CO.53. This is done to lower power peaking (peaking factors) caused by "water hole" peaking at the CEA guide tubes (0.753 REFERENCE Question Bank pg 75 ! Fuel Design / Core Power Distribution 2117 pq 16-20 EO 2 !
2.9/3.1
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4t_'_IUEQBY_QE_UyCLEBB_EQWEB,&(@NI_QEEB@l1QUt_E(y1Q$t_@NQ PAGE 21
..- . 'IUE6dggyNQuigS ' ANSWERS -- MILLSTONE 2 -88/01/11-SILK, D.
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.
192OO5K112 ...(KA'S) ANSWER 5.08 (2.50) Core size er c64 fo$iYN"
' Core age Value of MTC Value of FTC CO.25 pts each3 As power shifts to the top of the core, insert Group 7 CEAs As power shifts to the bottom of the core, withdraw Group 7 CEAs CEAs will be moved when ASI crosses the ESI (Equilibrium Shape Index)
CO.5 pts each3 REFERENCE Characteristics, 2116 F, pg 22,27 EO 6,7 3.4/ OO6K106 ...(KA'S) ANSWER 5.09 (3.00) Adding negative reactivity to the top of the core [0.53 causes the ASI to become more positive (power is driven to the bottom of the core) CO.5 . As power decreases, the change in T h is greater than the change in Tc CO.25 With a -MTC, less negative reactivity is inserted in the ] . top of the core than the bottom due to positive reactivity feedback CO.25 Also, the MTC is more negative at the temperatures at the
,
,
top of the core CO.25 ASI becomes negative CO.25 . As SG steam flow is decreased, T c increases and the hotter T c enter-ing the core reduces reactor power CO.2 At the lower reactor power there is a smaller delta T across the core [0.2 The net result is that T c increases more than T h CO.23. But MTC is more negative at the top of the core than at the bottom [0.23 so the effects are approximately off-setting and ASI does not significantly change CO.23.
REFERENCE . Characteristics 2116 G pg 40-43 EO 9 l . ' 3.3/ .2/ OO4000K515 192OO5K114 ...(KA'S) i J
-. . . .- - . . .. . _ . _ ~ ...-
l THEORY _QE_ NUCLEAR _PQ@E6_&(@N1_QEE8@llQN t _E(tjlD@t_@ND PAGE 22
.- THE80QDYN6DlCS
,
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, .
. .
ANSWER 5.10 (3.00) Rhol = (K1 -1)/K1 = -0.04 K1 = 1/(( 1 - (-0.04)) = 0.9615 CO.53 100(1 - 0.9615) = 196(1 - K2), K2 = 0.9804 Co.53 Rho 2 = (0.9804 - 1)/0.9804 = -0.02 CO.253 Rho 2 - Rhol = -0.02-(-0.04) = 0.02 = 2000 pcm . CO.53 ; 1000 pcm is due to Xenon, thus the remaining 1000 pcm is due to boronCO.S3 1000 pcm = 1000 pcm /(10 pcm/ ppm) = 100 ppm Co.53 900 ppm - 100 ppm = 800 ppm new boron concentration C0.253 REFERENCE l Characteristics 2116 B pg 9,10 t EO 4 l i 3.5/ .8/3.8 i 001000K528 192008K104 ...(KA'S) ;
!
.
ANSWER 5.11 (2.00) i
;. C^MC Gelchd HIGHER HIGHER SAME a5 ' LOWER [*Pr& pts each]
REFERENCE Characteristics 2116 G pg 3,4,8,9 I EO 4, 10b
,
3.4/ K101 ...(KA*S) '
i
, . . - , . . - - - _ . _ _ _ _ _. _ _ -.. _ _ _ _ __ _ _ . . . _ _ _ - - . . _ _ _ - - . _ _ _ - . -
*s . ' . THEORY OF NUCLEAR POWER PLANT 23
' t- OPERATIONt _ FLUIDS t_AND PAGE
- 'IdEQUQQYN@dlCS .
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, . ANSWER 5.12 (2.00) INCREASE DECREASE INCREASE INCREASE CO.5 pts each] REFERENCE HTFF 2121 J pg 20-22 EO 7 HTFF 2121 E pg 17-22 EO 6 2.4/ .6/ K109 192OOBK121 ...(KA'S) l
. .. - - -. - _ ', .- '
. , , pt__PQ@U1_SY@IEdg_DEgl@dt_CONTRO(t_AND INSTRUMENTATION PAGE. 24
.- . 1
. ANSWERS -- MILLSTONE 2 -88/01/11-SILK, I 1-
. *
l l i < l ANSWER 6.01 (2.50) Upper and Lower electrical limit , CWP CEA withdrawal prohibit >
'
CMI CEA motion prohibit (0.3 pts each3 . Rod drop from Reed switch .NI negative rate of power change from NI syste CO.4 pts each] UV trip devices CO.53 f l
'
REFERENCE CEDS pg 8,36,37 Fig 30 CRAB Panel COS AA-24,AB-24
, TPG CEDS EO 6,13 l l RPS EO te,1f l
1 3.4/ .5/ .7/ .4/ .9/3.1 l OO1000K103 OO1000K105 OO1000K407 OO1000K604 063OOOK201 I ...(KA*S) ! ; i ANSWER 6.02 (2.00) 4 l PZR heaters deenergize !
, '
Letdown flow control valves close to minimum j Both backup pumps start Auto makeup to VCT initiates (due to charging / letdown mismatch) ! Sprays initiate (to reduce pressure from r.ompressing the vapor space) ' ]. Reactor trips on high PZR pressure (
; PORVs open CO.4 pts each] !
REFERENCE PZR P&L Control System pg 4-21
'
i RPS pg 19,20 , I TPG PZR L&P Control EO 6,7,9 i
3.6/ .1/ .7/ .8/ .2/ .3/ f 011000K101 011000K102 011000K103 011000K104 011000K301 , 011000K401 ...(KA'S) { l
1, t f
1 : i i j i . !
! ;
I
,
;
- !
\.'
PAGE 25 . , ,,e t _ _P(QUI,5yglgdQ _Q(G lGN t _QQN16QL t _6G Q,1N616QUENIQllQN ANSWERS -- MILLSTONE 2 -88/01/11-SILK, D.
.
.
ANSWER 6.03 (2.00) Charging pumps, HPSI, SIT, LPSI CO.2 pts each, 0.2 pts for order] (Leakage would be the sum of the HPSI and PDPs) HPS1: 2 X 450,Ogpm = 8tOQOgpm (0.53 PDP: 3 X 44gpm = 132gpm CV.5]
-------------------------- --- (Will accept + or - 4M6 gpm for the HPSI pumps)
000 REFERENCE SIT system og 4 LPSI system pg 6 HPSI system pg 12 FIG 6 PZR P&L Control pg 8 TPG ECCS EO 12c,12d 14,18a,1Bc 4.0/ .3/ .6/ .5/ .4/ .0/ OO6000K605 OO6000K602 OO6000K506 OO6000K100 OO6000K103 OO6000A101 ...(KA'S) ANSWER 6.04 (2.00) Tan ) 540 *F ClO Turbine tripnedlad Normal vacuunDG 10.25 gin -.w i d Rt.e.. it.e e -d s t, t Le
-
eCisemi.vi i SG i O. 53 To M PU58 a"8 I '"0"'*J cmdessf> n from the volvh end n*Hices Co.D ii a , Prevents electronic noise generated by the quick open relay of Channel ; Y CO.51 from causing inadvertant and undesirabl e equipment and instru- ; ment response CO.5 ' REFERENCE l RRS pg 10,11,12,l't l TPG MSS EO 6a,61 3.3/ .1/ .9/ K102 039000K106 041020K401 ...(KA*S) l l
_ _ _ _ _ _
< . , ,,b._*.,P(@NI_@!SIEd5_QEgi h _QQUISQLt_@NQ_lyS16gdENI@llQN PAGE 26 4NSWERS -- MILLSTONE 2 -88/01/11-SILK, ANSWER 6.05 (1.50)
Steam generator level increases until the "Hi level override" CO.53 senos a signal to close the feed reg valve and feed reg bypass valve of the No. 2 SG C O. 5 3. With SF > FF a reactor trip from Lo SG 1evel will occu * CO.53 REFERENCE FW Control System pp. 9,10 TPG MFW/FW Control EO 4c,5j 3.1/ .4/ .3/ K103 059000K104 059000K402 ...(KA'S) ANSWER 6.06 (2.40) No effect Ccu uunc Incrce s Decrease (kt soni.lenzike cky H $ low w'l b=#t h'*5%bonWfo'*f) P No effect Oc.c rarr TnceNsf No effect (0.4 pts each] REFERENCE RPS Description pp. 28,29,30, FIG 24 TPG RPS EO 3b 3.9/ .3/ .9/ OOOK402 012OOOK501 012OOOK611 ...(KA'S) l l
i
!
__-_-____ _ _ _ _ _ _ _ _ _ _ _ l
,4 aam4 &~ , - + E-,-aT 3-# ,m ,4 - _ . +.AJ - ._, a # ', . i . /it_lBCON1_SYSIEUS_QESIGNt_CQNISQ(t,$NQ_lNSlgydENI@llQN - PAGE 27 , ,
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, t f
- ;
,
l
,
" ANSWER 6.07 (2.00) l I a. FO l
' FC
'
. FC FO o. FC I
' ^
- . t C : M u .. ., '.. Oele$ $ i!C C er96- p t s eac h 3 i O.25'
REFERENCE
, '
AOP-2563 pp. 4,5 CVCS pp. 6,17.23 j MSS pp. 8,9 I
TPG Compressed Air EO 11a,11d , t 2.9/ .4/ .4/ ! j 000065K303 078000K105 07BOOOK302 ...(KA'S)
!
!
.i ANSWER 6.08 (2.30)
! '
c. To prevent blockage of bleedof f flow from other RCPs (0.53 r 1100tioJ 1100 60 ; 601W 60 60 CO.2 pts each] l
REFERENCE ] j Question Bank p. 5 RCS p. 35 4 TPG RCS EO 5,6 3.7/ .1/4.2 J 002000K106 002000K113 078000K302 ...(KA'S) ! i l :
; ,
t !
- 8 i
I _ . ,_ . _ _ _ _ , . __ _ _ _ _ , _ - - _ . _ _ _ _ _._ _, _. __. _ __ . _ _ _ _ _ _ - _ . _ . . . _ . - . _ . _ _ _ _ _ _ . . - . . - _ . _ _ _
* .' PLANT SYSTEMS DESIGN1 _ggNTRgL 1_ANp_JNSTRUMENI@I]QN PAGE 20
..', * ANSWERS -- MILLSTONE 2 -88/01/11-SILK, . ANSWER 6.09 (2.50) FALSE FALSE FALSE TRUE TRUE CO.5 pts each] REFERENCE ICC LP pp. 4,7,12.17 TPG ICCS EO 1,4,5,6 4.2/ .3/ .5/ OOOO74A101 000074A113 OOOO74K208 ...(KA'S)
. ANSWER 6.10 (2.00) . RCP thermal barrier / seal coolers 2. Letdown huat exchanger 3. Shutcown cooling heat exchangers 4 Primary system sample coolers (0.25 pts each) Radiation M.>nitor intrense (or alarm)
Surge tank level increase CO.25 pts each) ESF room air cooling coil outlet valves CAR cooling unit 10" outlet valves CO.25 pts each] REFERENCE RBCCW system pp. 3,11 TPG BRCCW system EO 9,11
3.6/ .3/ l OOOO26K302 OOBOOOK104 ...(KA*S)
\
l l l l I i l l
*
.,* ! PLANT SYSTEMS DESIGN t CONTROL t AND INSTRUMENTATION 29
. ..,
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.---n------------------ PAGE i
>
I . ANSWERS -- MILLSTONE 2 -88/01/11-SILK, , a '
- I v . ,
.
! 1 ANSWER 6.11 -( 3. 00 ) l
i ' a. CAR Units (0.5] EW*63
!
t To prevent transferring the water in tNr sump to the RWST (and possibly : j causing an unmoni tored release via the i:WST vent) (0.03 Co.4 3 I j
'h; "^" ;
] CC" ,c; re; pend CO.53 bece ;e tne LNr ' ecked e t the ! j brieW tr en er 0 h e L. ee L e, ,, wet Le ,eeet L w e i;. CGI m i l m i o, L, Sump outlet valves (CS-16.1A/B) open !O.!: ! 3 Minimum flow valves (SI-659,660) close EG.55 i }
Outlet velves on the shell side of the SDC Hx open (0.43 [d.6piagack) !
-
r REFERENCE i CSS pp.4,6,7,11 TPG ECCS EO 5,10 l l AC Distr. EO Bb 3.4/ .9/ .2/ .1/ .7/ K201 026000K301 026000K401 026020K403 026020K404 j ...(KA'S) ,
+
l l i t I i , I i l l
i
f ) J l l
<
l
1 i i ! l
!
. '*
, ,,1t_,PAOCEDURES - NORMAL t_@@NQRM@bt_EMERQENQY_@NQ PAGE 30
. ,
509196991cet_QQN16QL ANSWERS -- MILLSTONE 2 -88/01/11-SILK, . ANSWER 7.01 (1.00) b8 N5 o r. &rls a tel J M $ d' Makeup to the RCS must be from the RWST CO.53 To insure that the boron concentration is greater than or equal to that of the RCS. CO.53 REFERENCE AOP 2551, step 4.20.4 p. 7 TPG 2551 EO 3 3.1/ .4/ OO4000G10 OO4000K123 ...(KA'S) ANSWER 7.02 (1.00) Prevents the shutdown cooling system from isolating CO.53 and the safety injection tank outlet valves from auto opening CO.5 REFERENCE OP 2201, step 5.1.3, TPG 2201 EO 3 3.5/ .1/ .4/ .9/3.! OO4000K102 OO4010G10 OO4010K101 010000K106 ...(KA*S) ANSWER 7.03 (1.50) Fill and drain the pressurizer to cooldown and thereby l depressurize the RCS, CO.53 Depressurize the pressurizer by ambient cooling. CO.53 Open a PORV as needed to reduce RCS pressur CO.53 l l REFERENCE AOP 2553, step 4.6, l ' TPG 2553 EO 2f, 3 3.7/ OOOO27K303 ...(KA'S) l l
--. . - . -- _ _ - ... . . _ - . - -
' i Z*t._PRQQEQQRES_~_NQRM@Lg_ABNQRMAl t_QMgRGENQY_ANQ PAGE 31 L
'
!
/'- 809196991C06_CQN1RQ(
1 ANSWERS -- MILLSTONE 2 -88/01/11-S!LK, , l . . ;
I 1 f
;
1 '
' ^ ANSWER 7.04 (2.50) -
( , No C O,25 }. Even though the RCS was cooling down at a rate of 3b F/hr, ; , '
while below 300 F the RCS temperature did not exceed t h e -fe- F / h r c ool- l down limit C O . 7 3 ' . [i.0) 30 OddtM ACC vr ABC CG.33 1
,
To minimize thermal shock when aux spray i s i ni ti a t ed 44%4FF Cl*01
j i d
!
.fghd Iw pr uv a de a fiwn pat is i vr iiir wiier visiQ puGP% tG pi& v wiii dsGsQ& iv Lii- pumg, CO.0F REFERENCE ! OP 2207 pp. 3,6,7,10,13,15 l
; TS 3.4. ,
EO 1
'
TPG 2207 i ! } 3.3/ .2/ l l 010G10 035G5 ...(KA'S) {
,
l i ANSWER 7.05 (2.00) . I Restore Tavg ) 515 F CO.53 or be in hot standby within 15 minutesCO.53 } Trip the reactor * '
'
N'"b1b l Water hammer [0,25) 0.5 pt; ;;;'. 2 l l ! l REFERENCE ! ! OP 2203 pp. 4.5,7 l ! TPG 2203 EO 3D,3d !
I i 3.6/ .1/ .6/ '
,
'
OOOO54K101 OOOO54K301 OO2OOOG5 ...(KA*S) ' i l t i I .
' * l l
l l
!
l
t l s (
* *.. PAGE 32 ZL__E60gEgySES_ _NOSd@(t_@BNOSdQ6t_EUESGEyqY_QNQ
. .'- - , 50D1069G1G06_GQN16Q( ANSWERS -- MILLSTONE 2 -88/01/11- ILK. D.
.
.
ANSWER 7.06 (2.50) RCP seal temperature exceeds 250 F RCP bearing oil temperature exceeds 195 F RCP controlled bleedof f temperature exceeds 195 F CO.5 pts each3 The unit vii.11" u v io uvid uiiuid un , n. ii u o 24 iiuur m due iu TS 3.03C1.O] Cehwel d(Wdtn :n molei j$ %+ ym;tfr/ REFERENCE AOP 2564 pp. 2,3 AOP 2565 OP 2326 pp. 16,17 TS 3.7. TS 3.05 TPG 2564 EO lb TPG 2565 EO 4 4.0/ .4/ .8/ .4/ OOOO26K303 ANSWER 7.07 (3.50) . SGBD isolation valves close (2-MS-220A,B) 2. DD tank discharge valve to circ water closes (2-MS-15) 3. BD quench tank discharge valves to circ water closes (2-MS-135) 4 SGBD sample discharge valves to secondary sample sink close (HV-4287,4288) CO.25 pts eath] If leakage exceeds the capacity of the CV' to ar.ihtain PZH levelCO.53 SG leakage > 0.15 gpm Dose equivalent 1-131 > 0.1 uCi/gm (0.5 pts each3 To allow estimation of any release CO.53 To ensure consistency with chemistry discharge calculations CO.53 REFERENCE AOP 2569 pp. 2,4,5,6 TPG 2569 EO la.b,c,d 5
'
3.7/ .2/ .6/ .0/ OOOO37K305 000037K307 073OOOK101 073OOOK401 ...(KA'S)
- - - _ - _ _ _ _ _
._ ~ . "
.. IL _6dQGEQWBEE_ _NQBd@(t_@@NQBd@(t_(b(@Q(NQY,6NQ PAGE 33
.- 68019L001 gel _G991696 ANSWERS -- MILLSTONE 2 -88/01/11-S1LK, .
ANSWER 7.08 (2.00) To prevent exceeding the RCS pressure / temperature limits CO.53 None (0.53
!
.4 To prevent run out of the service water pumps CO.53 j To bleed off excessive pressure in the injection headers ( 0. 5 ')
i REFERENCE : ADP 2571 pp.2,3,5 i TPG 2571 EO lb,2,5 2.4/ .9/ i 000G4 006050K401 ...(KA*S) l ANSWER 7.09 (3.50) !
! PZR level > 20% !
Heat removal from at least one SG : P/T limit of Fig 4.2 satisfied for That l Loop Delta T between 10 and 45 F , Tcold constant or decreasing l That constant or decreasing CO.25 pts each] , PZR level > 35% and constant or increasing P/T limit of Fig 4.2 satisfied for That
. Heat removal from at least one SG '
Rx vessel level above the top of hot leg (> 4 3*/. ) CO.25 pts each3 Both decrease (0.53 High radiation exposure [0.53 REFERENCE EOP 2532 pp. 8,9,10,16 TPG 2532 EO 4a,4e,65,12 4.4/ K312 ...(KA'S) I
!
l . . . _
, .. . . t l It__EBQQ(QQRES - NONMA(t,QgNQRMQ(c.gMgRQgNQY,ANQ PAGE 34 ; '.* - - BeQ1QLQQ1 gel _QQN16QL t . !
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, , Ws)w 4. -).open.pdy la) alm a ce.scisn) tost] ! ,
-clou vcro41d sw U-w-col) lea 3 l
- %)<tr...;hU<c% q p-p to.31] l ANSWER 7.10 (3.50)
l Open BAP discharge valve to charging pumps (2-CH-514) l Start both BAPs ; Close BAP rectre valves (2-CH-510.511) , Start all available charging pumps CO.25 pts each] l i t j To ensure PZR level is valid (and voids are not forming in another j portion of the RCS) CO.43 .
- l
' Normal quench tank level, pressure, and temperature
'
No acoustic monitor alarms CAny two, 0.4 pts each3 Close the MSIVs CO.43 i To avoids excessive PZR level and pressure transient i
. excessive cooldown rate i overfilling the SG CO.3 pts each3 ;
i REFERENCE l EOP 2525 pp. 3,4,5,8,11 ,
'
Intro to EOPs p. 12 TPG 2525 EO 4a 4b,6b TPG Intro to EOPs ED 11,15,16,20 l 3.9/ .1/ .0/ l 000005K306 000008K303 000074K311 ...(K4'S)
!
!
ANSWER 7.11 (2.00) l
! If the ruptured SG i s thd+ only one available for heat removal CO.753 ,
or 78 St.s T, Ia m+ bel,w ne 'f k mn the pakkj 4, I(k3 % n.. genmhe 6 4 , v.lats ; Void collapse may be large enough to, drain the PZR CO.753 : l Gabl ! ~ 5 the PORVs CO.53 i i REFERENCE , i EOP 2540D pp.4,9,11,15,19 l TPG 2540D EO 5 - * l
;
l I 4.4/ .6/ ! 000054K304 000054K305 ...(KA'S) j
:
I
,
I l
{
l
m I
. ' ,
, 8t_'_SE01N1 SIS @IlyE_88QCEQg8E h _CQ6QlIlg & _@NQ,LidlI@IlQNS
.. . , PAGE 35 .
ANSWERS -- MILLSTONE 2 -88/Ot/11-SILK, ANSWER 8.01 (2.20) LCO's indicate lowest perf ormance level of equipment required for saSe operation of the facility CO.75 If improper automatic action occurs prior to reaching ari LSSS, then Saf ety Limits will not be exceeded CO.7 If Safety Limits are not exceeded then fuel and RCS integrity will be maintained CO.75 REFERENCE 10CFR50.36.c.1,2 TPG TS EO 5 3.6/ .6/ OO2020G6 OO2020G5 ...(KA*S) ANSWER 8.02 (1.00)
?. . . .D_5 ,,Y_, __
_ _ _ . ?,_ 'I' . ' Y_ ..._ -? .. _ ! ...?.
.~
_
. .. ,. _ .. . . . - , . -._
-
%e ish >el from ,fs/% 4. s%/ti (r%- three conserv+i<e surve@.nce ;nkevals) U.s]
edeeds the rey;<ed 2.zs +;mes ps, sury,;lk.nce REFERENCE )< fervsl (o.5]. LER 87-008 TS 4.02 TS 3.8. TS ACP-QA-10.01 TPG TS EO Ba 2.4/ .0/ OOOG5 063OOOG3 ...fKA'S) {
!
ANSWER 8.03 (2.00)
[0.5) Co.57 A plant shut down4,hould be started within one hour t+r4-3 as required by TS 3.03 CO.53 with the plant in cold shutdown [within24 hours)CO.5 REFERENCE LER 87-003 TS 3.03 TS 3. TPG TS EO !q.15a 3.2/ .6/ G6 010G5 ...(KA'S)
. _ - - . - _
. .-.
. . _ _ . . . .
, , . ,82_'_090] NJ gIB9I] VE_ PB9C EQUB E@i _ CgNp]IJ QN}2,9ND_ LJ DJI911gN@ PAGE 36
,. .
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, .
.
ANSWER 8.04 (3.00) a. FALSE b. TRUE 444d :. '9UC d. FALSE TRUE o, (, f. FALSE C4,4-pts each] REFERENCE ACP-QA-206A pp. 6,9,10,12,24,25 3.7/ K102 ...(KA'S) ANSWER 8.05 (1.50)
[o.53 Restore RCS pressure within its limits within 5 minutesb#1b Notif y NRC (wi thin one hour) (0.75]pt: :: 53 REFERENCE TS 2. TS 6. ACP 6.01 .5/ .2/ G5 010000G3 ...(KA'S)
ANSWER 8.06 (3.00) Station Superintendent Unit Superintendent Operations Supervisor CO.5 pts each3 Understand the cause of the trip Ensure no abnormal conditions exists that preclude reset Record relay position (f or further investigation)
[Any two, 0.5 pts each] Verify the receipt of annunciators CO.53 REFERENCE ACP 6.01 pp. 5,6,7 2.8/ _ . . . _ _ _ _ . _ _ _
. _ . _ _ --_ _ - _ _ ,
, .8_,.).f_eDdlN[ SIB @IlyE_P6QCEDyBES ,
t _CQNDillgN@t_@ND_(ldlI@IlgNS PAGE 37
. ..
ANSWERS -- MILLSTONE 2 -80/01/11-SILK, , A f ter+.4 for 107: The darfvpm 6e ceencn{ O o) bMec YI: 1940014111 ...(KA'S) CCCS ) 4:rJh frowdd bon 4 e J L h $1 ff# or<ff*f0l)Isl;yed0.s} ANSWER 8.07 (2.00) The startup should not commence C1.03 because a mode change cannot occur with reliance upon the action statement for the HPSI. pump [1.0 REFERENCE TS 3. TS 3.04 3.4/ .4/ K109 OO3OOOG5 ...(KA'S) ANSWER 8.08 (2.00) The intent of the original procedure is not altered CO.63 The change is approved by two me,tais of thz plant renogsmcnt stof', at t-omt vom vi :,u. iiul d a ou CCC 1 i c e. se [ 0 . 7 3 lu e >ed 5 4 5 /ro m +6e vaif ia*Ind at-less4 oneef sm sllbe the m ddy $5 lo.'l3 l The change i s docs'mented, review by the PORC/SORC, and approved by the Unit Superinten . / Station Superintendent within 14 days of implement- l ation CO.73 I REFERENCE TS 6. MP M 3.02 sechos G.' .3/ A116 ...(KA'S) ANSWER 8.09 (2.00) The SRO should not grrst permission CO.53 becausa only licensed operators CO.53 or individuals ii. training CO.53 under direct supervision of a licensed operator can operate the controls CO.5 REFERENCE ACP 6.01 pp. 10,11 10 CFR Part 55.4, 5 .5/ A103 ...(KA*S)
s.fj_eQUlNIS16@llyE_P8ggEQQ5ES_CQNgillgN@t_@NQ_LidlI@llgN@ t PAGE 38 , , ,
.. ,
ANSWERS -- MILLSTONE 2 -88/01/11-SILK, D.
.
ANSWER O.10 (2.40) Whenever all of the following conditions are not satisfied: Reactor vessel water level at or above the vessel flange Reactor vessel pit seal installed Combined available volume of water in the refuel pool and RWST exceeds 370000 gallons One LPSI' pump not in shutdown cooling service and aligned to take suction from the RWST and deliver flow to_the RCS is operable OR One HPSI pump aligned to take suction from the RWST and deliver flow to the RCS is operable
[O.6 pts each]
REFERENCE TS 3.9. TPG EC :: 50C5 Eo ll 3.2/ OO5000G5 ...(KA'S) ANSWER 8.11 (2.40) All penetrations required to be closed during accident conditions are either: IO d Capable of being closed by an operable containment isolation valve system [o $) Closed by manual valves, blind flanges or deactivated automatic valves secured in their cl osed posi ti ons [0.5) The equipment hatch is closed and sealed [AN The airlock is operable pursuant to TS 3.6.1.3 N N
- 0. 6 p t s wou;. 3 l
REFERENCE ; TS p.1-2 ! TS 3.6. ! TPG EO 13a 3.3/ OOOG5 ...(KA'S)
..,kaj_AydlN1@l6@llyE_EBOGEQUBE@t_QONQlllON@t_@NQ_Lidll@llQNS PAGE 39 00 , ANSWERS -- MILLSTONE 2 -88/01/11-SILK, ANSWER 8.12 (1.50) At BOC MTC may be slightly positive at operating conditions CO.53 and since it will become more positive at lower temperatures CO.53, the limit is provided to restrict reactor operation when Tavg is significantly below normal operating temperatures CO.5 REFERENCE TS 3.1. TS p. B 3/4 1-2 2.9/ OO1G6 ...(KA'S) I
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RESOLUTION OF FACILITY COMMENTS - FOR R0 EXAM GIVEN JANUARY 11, 1988 AT MILLSTONE UNIT 2 FACILITY COMMENT 1.05-1,2: We cannot currently read 10,000 cps on our Excore NIs. The circuitry is designed to shift power indication to % power whenever countrate goes above 1000 cps. References are provide ' There are no FEY.1 on the S/ There are Secondary Safety Valves, Atmospheric Dump Valves and Steam Dump and Bypass Valve NRC RESOLUTION: Comment noted, however, does not affect the candidates ability to correctly answer the question. A note will be added to the question for future reference to make the question more technically correc FACILITY COMMENT 1.05-3: . If the student assumes a +MTC at BOC conditions, then the cooldown will add negative reactivity. The resulting condition when equilibrium is reached would be: Final Tavg less than Initial Tavg, Final Power below POA This means that no correct choice was given in the questio Based on this above, credit should be given for any written answers which assume a +MTC, as well as for key answer 'd" which presupposes a negative MT FACILITY COMMENT 1.098: i The candidates may assume based on part a, that 70% power is to be ' maintained or that power is being increased to 70%. As such, reducing I power to control ASI within limits may not be recognized as an option by the candidate The answer key should accept for full credit two of the following three steps / methods for ASI control: Rod Insertion . Rod Withdrawal Power Reduction
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REFERENCE: OP 2393, Xenon Oscillation Band Control
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NRC RESOLUTION: Disagre The question does not state power. is to be maintained in part Technical Specifications (TS) may require a power decrease and candidates don't have the option to not comply with TS action statements. However, will accept the reference and modify the answer key to accept 2 of the 3 steps / methods for full credi FACILITiCOMMENT1.11: The students are only required to discuss the effect of source-detector geometry on the 1/m plot. They are not required to know how fuel loading, fuel enrichment or poison loading affects the 1/m plo (Theory objectives related to 1/m plots are attached).
There are two correct answers to this questio Both a. and c. can be Correc The key answer, "c", is technically a correct respons However, this choice describes an evolution which is not done at MP Based on these comments it is recommended that full credit be given for either "a" or "c".
NRC RESOLUTION: Comment accepted. Answer key modified to accept a or c for full credi FACILITYCOMMENT1.14(3): There are no S/G PORV's on Millstone Unit NRC RESOLUTION: Same as comment 1.05-1, FACILITY COMMENT 1.15c: The answer states that the core delta T during Natural Circulation approaches full load delta T. This is incorrect. The Natural
' Circulati6n delta T will be approximately one-half of full power delta T (this assumes maximum possible decay heat). (Reference attached).
Based on this, the phrase "Core delta T during natural circulation cooldown will approach full load delta T." should be removed from the answer ke NRC RESOLUTION: Answer key will be modified from "full core delta T" to "20-25 degrees-F" as indicated in your reference and pages 4 and 1,0 of OP-RO-FUNO-2121 .. . _ _ _ _ _ _ _ _ _ . .- __ .. -- ..
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FACILITY COMMENT 2.04c:
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The reference given (H2-0P-ELECT-2342, p. 3 & 4) does not state a specific reason for the 460 amp limit on bus 24E when it is being I supplied from 24 Procedure OP 2343 (reference book 4, section 14), step 7.22, caution il states, "Do not exceed load limits on RSST 15G-21S or its busing 3.0 MVA 460 amps." The identifier "15G-21S" is the designation for the Unit 1 RSST, not a breaker or disconnect. As the Unit 1 RSST is not limited to 460 amps and the Unit 2 operators have no
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real indication or control of the total load on the Unit 1 RSST, the 460 l
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' amp limit is understood to be based on the bus connecting the Unit 1 RSST to 24 (Reference excerpts are attached).
NRC RESOLUTION: Comment note The suggested revision is not clear, but the procedure l is. The answer will be mcdified to remove the word "breaker". , I FACILITY COMMENT 2.08b: l The question asks for three (3) sources of SFP makeup water but does not ; solicit a system flow path, j Therefore an answer stating the RWST as a possible source should be fully accepted as one of the three required answer NRC RESOLUTION: As explained during the exam review the RWST has two flow paths as a possible source and if the candidate uses the RWST as two sources then it was necessary to include the path. This was identified during the review and the answer key was modified to place the flow paths in parenthesi i
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FACILITY COMMENT 2.09d: The AFW Flow Control Valves have three modes of operation as stated in both OP 2322 (ref. book 4) and AOP 2579B (ref. book 7). .These modes are ' Auto, Manual and Manual (Local). If an Automatic Feedwater Actuation Signal (AFAS) is present .th.tn the "Reset-Normal-Override Switch" has three modes of selection which allow the operator additional modes of operation (AFW SD & OP 2322). As the question did not mention the switch by any name, nor indicate that an AFAS had occurred,jit is impossible for an examinee to determine which "modes of operation" the question is attempting to solicit. Therefore, discussion in either area of AFW Valve control should be accepted for full credi NRC RESOLUTION: Comment accepted. Answer key modified to accept additional correct answer. However, the original references uses the words "modes of operation" when referring to Reset Normal-Override Switc .-. . . - _ _ - .-
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FACILITY COMMENT 2.10a: The CEA, upon loss.of the lift coil, will be held in, place by the Upper
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gripper and/or the Lower gripper coil ' The key answer should be changed to allow full credit for mentioning either the upper gripper or the lower grippe NRC RESOLUTION: Cominent accepted. Answer key was modified during the exam review to eliminate the word "lower". ! FACILITY COMMENT 2.11: ! The following CVCS components also receive a signal on SIAS according to reference M2-0P-PRI-2304 Fig. 2a. (CVCSSD): (# 1-7 given in answer i key). I PMW to charging pump suction (2-CH-196) Precise reactivity control isolates (2-CH-909, CH-910) 1 Boric acid pump recire. isolations close (2-CH-510, CH-511) NRC RESOLUTION: * The precise reactivity control valve is 2-CH-936 and does not isolate on SIAS. However, valves 2-CH-909/-910 are the chemical metering discharge valves, they do shut on SIAS and are in the flow path for precise reactivity control. Answer key was modified to include the three additional correct answers, with correction to No. FACILITY COMMENT 3.02a: Examinees may assume a cause for the given plant conditions (i.e. Excess Steam Demand Event in containment), in which case a SIAS, CIAS, and EBFAS could also occur on a high containment pressur As an EBFAS would override the AEAS signal, credit should be given if these actuatioas are assumed to occu NRC RESOLUTION: -
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Disagre Containment pressure was not given as a condition to be evaluated in part a, which would not support a MSLB inside containmen However, candidates will be graded for assumptions given and no change will be made to the answer ke FACILITY COMMENT 3.07: The question did not clearly solicit a reason for each of the automatic actions, but the answer key requires a reason for full credi Therefore, the reason part of the key answer should not be required for full credi . 4 . .. ___ .
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Coment accepted. The reasons in part a and d of the answer key were modified by placing the reasons in parenthesis during the exam revie FACILITY COMMENT 3.08b: The phrase "increasing S/G pressure" in the key answer should not be required for full credit. Steam Generator pressure is not an input used to TM/LP calculation NRC RESOLUTION: Coment accepte The answer key was already modified during the exam review to have parenthesis around "increasing S/G pressure".
FACILITY COMMENT 3.10b: An explanation that deals with the actual system res)onse that will close the other five steam dump valves, should also ae accepted for full credit. This includes Tavg less than the setpoint for the 8, C, and D steam dumps or steam pressure less than the atmospheric dump setpoin NRC RESOLUTION: Coment note The actual system response was not required for full credit, however, it would also support the response in lower decay heat removal requirement and will be considered as additional correct information, if supplied, no change in the answer ke FACILITY COMMENT 3.114. and e.: The D/G 12 U Trouble Annunciator will result from any one of 30 different conditions (reference attached). Depending on which condition caused the Annunciator to alarm, any one of the answers given in Column B could be correc Examole: If the annunciator alarms due to a "Lube Oil Level Low" condition, the D/G will not trip and answer number 1. is correc . If the annunciator alarms due to an "Engine Overspeed"rcondition, the D/G will trip under any condition and answer number 2. is correc . If the annunciator alarms due to a "Lube Oil Temp. High" condition, the D/G will trip, unless it had received an Emergency Start signal, and answer number 3. is correc Based on this, it is r'ecomended that part a. be deleted. '
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3.11 A DC Control Power Failure will either 1) Not Trip the D/G, 2) Trip the D/G unless an Emergency Start Signal is present or J) Cause the D/G to come up to speed on the mechanical governor with ng trip protection (exceptoverspeed). What Mill happen depends upon what portion or portions of DC Control Power is los To determine what will happen to the D/G, refer to t'he circuit diagrams supplie ' Figure 8.3-2 Sheet 12 shows that the "Loss of Control DC," Annunciator is caused by one of four relays: CR1, CR2, CR3, and CF . Identifying each one of these relays we find that: CR1 is the relay that indicates a loss of power to the starting portion of the D/G Control Circuitry. Refer to Figure 8.3-2 Sheet CR2 is the relay that indicates a loss of power to the shutdown and local starting portion of the D/G Control Circuitry. Refer to Figure 8.3-2 Sheet CR3 is the relay that indicates a loss of power to the automatic tripping and emergency shutdown portion of the D/G Control Circuitr Refer to Figure 8.3-2 Sheet CF4 is the relay that indicates a loss of power to the Exciter Control Circuitry. Refer to Figure 8.3-2 ' Sheet 1 . A loss of power to CR1, CR2 or CF4 will produce a D/G Trip signa Refer to Figure 8.3-2 Sheet 6. Note that a loss of power to CR4 will ngi produce a DG trip signal since the trip circuitry must be de-energized for this relay to lose power. A loss of CF3 will produce a "Loss of DC Control Power" annunciator, however, as shown on Figure 8.3-2 Sheet 1 . A Loss of power to CR1, CR2 or CF4 will agi produce a D/G trip if an Emergency Start Signal is present. This is shown on' Figure 8.3-2 Sheet 7. On an Emergency Start, the indicated ESS contact opens which prevents the Loss of DC Power Trip Signal from energizing the Shutdown Relay (SDR). In addition to the above, a loss of power to the circuitry monitored by either CR1 or CR2 will fail open the Air Start Valves and roll the D/G with air. If the Trip Circuitry (CR3) also loses power, the D/G will come up to speed and run with no trip protection available (except the overspeed trip) and the "Loss of DC Control Power" Annunciator in alar . 6 , _ _ _ _ _ _ - - _ ._ ,_ _ .___ -_
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- The difference in response of the 0/G to partia'l loss of control power vice total loss of control power is covered on pages 106 and {
107 of the D/G Instructor Guide (lesson plan excerpts attached).
{ Based on the above information it is recommended that part e. be delete ' j NRC RESOLUTION: Comment accepte I The answer key and questions were deleted and points i for b, c & d revalued to 0.33 eac ' FACILITY COMENT 4.03a: The question requires the candidate to state four of the steps which , must be taken to place the Enclosure Building Filtration Systems (EBFS) l in service during an Electrical Emergency. The guidance for performing this is contained in E0P 2528, Electrical Emergency, step 3.11, Contingency Actions. As such, it is not expected that licensees perform this step from memory. The answer key lists five steps; the procedure lists six steps. Steps four and five of the key answer (EOP 2528 steps 3.11 e and 3.11 f) describe actions that remove the Condenser Air Removal System from Servic * Full credit should be given for describing actions taken which will start EBF NRC RESOLUTION: - Comment note Candidates are required to have system knowledge on design intent, construction, operation and interrelationships in accordance with NUREG 1021, ES-202, 8-2 Page ES-202 page 2 of The difference in the number of steps is due to the dampers being operated, the answer key will be modified to include "open enclosure building exhaust to EBFS" as an additional step. Credit will still be given for removing the Condenser Air Removal System from service since this is part of the action required in placing the EBFS system in service which is done to prevent back draft of Unit I stack gases into Unit FACILITY COMENT 4.04b: The key answer contains four steps which are performed to /nitiate emergency boratio In addition to these steps, credit should be given for stating that: o If the boric acid pumps fail to start, open the gravity feed valve o If the gravity feed valves are being used, close the volume control tank outlet valv . These steps are performed if the boric acid pumps fail to start during emergency boratio ~
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REFERENCE AOP 2558, Rev. O, steps 4.3 and NRC RESOLUTION: Comment not accepte The question does not solicit" response not obtained or contingency information. The additional steps referred to are contingency actions implying that an attempt has been made to start the boric acid pumps etc., a primary action which is required for full credit. The additional information, if given by the candidate is not required for full credit and if correct will not count against the candidates score, no change to answer ke FACILITY COMMENT 4.08: The key answer states the guidance that is contained in A0P 2551 for closing the HSIV's from outside the control room. In that the candidate did not have this procedure available when answering the question, full credit should be given for describing alternatives which result in MSIV closure. Two alternatives include: o closing the MSIV's from the bottle on panels, C70A and B (The bottle up panels are located outside of the control room) REFERENCE: AOP 2579A, Rev. 2, step . o locally isolating instrument air at the MSIV's and then bleed the air pressure from the operating cylinders and accumulators REFERENCE: Orawing attached NRC RESOLUTION: Comment accepted. Answer key mods.ied to reflect the two additional ways which result in MSIV closure from outside the control roo /* FACILITY COMMENT 4.10a: The key answer lists five indications of a misaligned CEA which are found in the Entry Conditions to AOP 2556, Dropped CEA Recover Additional indications, not used as Entry Conditions, should also receive credit. These include:
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CEA Group Gross Deviation annunciato . CEA Group deviation backup annunciato , CEA Motion Prohibit annunciato . Correct a discussion power mismatc of NSS and B0P parameter changes re REFERENCES:
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H2 0P RO-IAC-2380 2, pg 19 and 20 OP 2302A, Rev. 9, Sections 8.6, , 8.8, 8.9, 8.15, 8.25 NRC RESOLUTION: Comment accepte indications not identified in the Dropped CEA RecoverAnswe references expanded to include additional references.y Procedure and _ FACILITY COMMENT 4.llb:
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The, implyingquestion asks for that one answer "an" indication of RCS leakage . . . ., is require ,
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Coment accepte either indication for full credit. Answer key was modified during exam re
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. "ATTACHMENT 4" RESOLUTION OF FACILITY COMMENTS FOR SRO EXAM GIVEN JANUARY 11, 1988 AT MILLSTONE UNIT 2 FACILITY COMMENT 5.08a:
The answer key gives core size as one of the 4 factors which affect the convergence or divergence of a Xenon Oscillation. While this is true, the design of the core gives a fixed siz The "effective" size of the core can be changed, however, by the positioning of the Group 7 CEAs. Based on this, CEA position as well as core size should be accepted as an adequate answe NRC RESOLUTION: CEA position or core size will be accepted as one of the four factors which af fect convergence or divergence of a Xenon Oscillatio FACILITY COMMENT 5.10: This question gives reactivity in units of both delta k/k and pcm. At Millstone 2 the operators only use units of delta k/k (or % delta k/k) and are not required to use units of pcm. Based on this it is recommended that no credit is taken off for incorrect conversions between pcm and delta k/ NRC RESOLUTION: Comment accepte FACILITY COMMENT 5.11a: In this part of the question on ECP vs. Actual CEA position, it is stated that one RCP trips two minutes prior to criticalit If this did happen, a Reactor Trip due to RCS Low Flow would occur making the pull to criticality impossibl (Reference attached).
Based on this information, it is recommended that Part a. be delete NRC RESOLUTION: Comment accepte .1 points will be redistributed to parts b. through FACILITY COMMENT 6.03.b: The question requires the candidate to use a HPSI pump curve provided to determine HPSI flow rate at a given pressur .
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l The key answer allows for + or - 50 gpm when making this determinatio Based on the pump curve provided, which does not contain an accurate grid, this allowance should be increase Plus or minus 100 gpm is recommende NRC RESOLUTION: Comment accepted. Also, upon further investigation of the pump curve, I the answer was modified to 1032 gpm for the leak rat ) l FACILITY COMMENT 6.04.b: The key answer states that (the quick open permissive switch is used)
"when there might be radioactivity in a SG." The reference sited (RRS pg 10, 11, 12) includes this information for historical design purpose The reference uses the words "The switch was included to permit . ."
It is not currently used for this purpos A0P 2569, Steam Generator Tube Leak, contains no guidance concerning its us The answer key should be changed to accept, for full credit, a response l that indicates that this switch is used to protect perscnnel when draining condensation from the valves and muffler Reference: RRS System Description, page 1 NRC RESOLUTION: Comaent Accepte FACILITY COMMENT 6.06.b: The key answer indicated that the TMLP trip setpoint will "decrease" as i ASI changes from 0.0 to -0.1. This is incorrect. The TMLP trip setpoint will increase under this conditio NRC RESOLUTION: Comment accepte FACILITY COMMENT 6.06.c: If the candidate assumes that the plant trips due to the RCP trip (as it would), the key answer is correct, the TMLP trip setpoint will derrease to its floor valu If the candidate considers RCS flow as the only variable of concern when answering the question, then "no effect" is correct. Actual flow is not an input into the TMLP trip circuitr !
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NRC RESOLUTION: If the candidate states an assumption, the response will be graded accordingly; otherwise the response will be graded in accordance with the answer ke FACILITY COMMENT 6.06.e: The key answer indicates that the TMLP trip setpoint will decrease when a linear power range channel (safety) fails high. This is incorrect. The TMLP trip setpoint will increase under this conditio The key answers should be revised accordingl Reference: T.S. 2.2 Figures 2.2-3, 2.2-4 NRC RESOLUTION: Comment accepte FACILITY COMMENT 6.07.g: The key answer is "No Effect", describing the response of the MSIV's to complete loss of instrument air. The MSIV's are equipped with air accumulators which serve to hold the valves open for a period of time following a degradation in instrument air pressure. Without a time frame for consideration indicated in the exam question, the response could correctly be either "tw e f fect" w "f ails closed."
The key answer should be changed to accept for full credit either "no effect" or "Fail Closed."
Reference" Main Stream System Description, pg 8-9 NRC RESOLUTION: Question 6.07.g will be deleted and 0.03 points redistributed to parts through f, and FACILITY COMMENT 6.08.b: The key answer requires specific pressure values for various seal i conditions. No allowance is included in the key for variations from the specified value l
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The key should be changed to allow for i 15 psi for vapor seal pressure i and 1100 psi for all other pressur l
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NRC RESOLUTION: Comment accepte FACILITY COMMENT 6.11.c: The circuit that prevents the containment spray pump from responding under the conditions specified in the exam question is the "Main Generator Final Coastdown Circuit" (Reference: Containment Spray System Description, page 6). Detailed knowledge of this circuit is not required by our learning objective The exam should be changed to eliminate part NRC RESOLUTION: Comment accepte .0 point will be redistributed among parts a. , b. and FACILITY COMMENT 7.01: The question refers to a note contained within a procedural step entitled
"Boration without Boric Acid Pumps available". The title of the step was not made available to the candidate. Taking the note out of the context ,
prevents the examinee from interpreting the meaning of "automatic ' boration unavailable" Therefore, an answer giving the boric acid storage tanks as a source of makeup should be accepted for full credi Reference A0P 2551, pg. 6 & 7, step 4.2 NRC RESOLUTION: Comment accepte FACILITY COMMENT 7.03: The question asks that three alternate methods of depressuring the RCS be given if auxiliary spray is inoperabl The objectives listed in the TPG for A0P-2553 do not require the students l to memorize alternate actions. In fact, the objectives specifically i state that procedures must be used for two of the alternate methods mentioned in the procedure. (Reference attached).
Based on this it is recommended that any reasonable method of depressurizing the RCS should be accepted as an answe .
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NRC RESOLUTION: Comment accepte FACILITY COMMENT 7.04: The answer to Part a. states that the cooldown limit below 300 F is 20 F/Hr. This is incorrec The cooldown limit below 300 F is 30 F/H (Reference attached). Part b. of the question asks what combination of three RCPs will provide the highest spray flo The objectives listed in the TPG for OP-2207 do not require the students to memorize these pump combination (Reference attached.). Additionally, the key answer is incorrect (References attached).
Based on this it is recommended that part b. be delete . Part d. asks why the charging Header Valves must be closed when securing auxiliary spray. As a matter of fact, these valves must be opened when securing auxiliary spray in order to ensure that the charging pumps have a discharge flowpat (Reference attached).
Based on this it is recommended that part d. be delete NRC RESOLUTION: An>nci Ley nili Le cho %ue h v... G 'F/nr to 30 F/hr for pait e. Taits and d. will be deleted and 1.0 point redistributed among parts a. and FACILITY COMMENT 7.06 b: The question asks for the restrictions on plant operation in Mode I based i on the stated conditions. The question does not ask for Tech. Spec, l references or time limit Based on the information given in the question, both 0/G's are inoperable (T.S. 3.8.1.1). And both Service Water headers are inoperable based on l the provisions of T.S. 3. Both of the above technical specifications prevent continued operation in Mode Additionally, the time limits associated with the actions of these
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technical specifications do not require memorizatio i l Based on the above, the correct answer to the question should be that ' Continued Operation in Mode 1 is not Possible. No other information should be require .
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NRC RESOLUTION: Comment accepte FACILITY COMMENT 7.10a: The answer to this question fails to include the possibility of opening the Gravity Feed Valves to perform a boration of the RCS. This method of boration can be used based on the Emergency Boration Procedure: A0P 2558 and Emergency Operating Procedure 2540A (which is referenced in the boration step, 3.1, of E0P 2525).
Based on this information it is recommended that an additional correct answer would be:
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Start all available charging pumps NRC RESOLUTION: Comment accepted as an alternate answe FACILITY COMMENT 7.11: Part a. of this question ask3n hen it is permissible to nyt isolate a ruptured Stean, Generato The answer key only gives one answer: If the ruptured SG is the only one available for heat remova Based on the SGTR E0P, there are additional correct answers:
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If RCS TH is n t below 520 F, the faulted SG should not be isolate The faulted SG may be unisolated to prevert overfillin he faulted SG may be unisolated to cooldown the S Based on this information it is recommended that any of the above answers also be accepted for full credi (Reference attached). Part c. The additional correct information that the bypass key is installed to allow for control of the PORV should be accepte . I
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NRC RESOLUTION: Question 7.11 asked questions regarding E0P 2540 D, Functional Recovery of Heat Remova The first suggested alternate response to part a. will be accepted in that it is supported by E0P 2540 D. The second and third alternate responses are not supported by E0P 2540 D and therefore will not be accepted as alternate response . Comment accepte FACILITY COMMENT 8.01: The answer key should be changed to allow for different, correct responses which indicate relationships between LCO's, LSSS's, and Safety Limits. One such response would be: l If the safety limits are not exceeded, fuel and RCS integrity will be maintaine LSSS's serve to trip the reactor to ensure that safety limits will not be exceeded, assuming that LCO's are being me Reference: 10 CFR 50.36, M2-OP-RO-ADMIN-2001, T.S. 2.1 and Base NRC RESOLUTION: Comment Accepte FACILITY COMMENT 8.02: Technical Specification 4.02 b. specifies "the combined time interval for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval". The period between 7/6/86 and 5/6/87 incorporates 3 consecutive surveillance intervals and exceeds 3.25 times the surveillance interva The question asks "EXPLAIN WHETHER OR NOT a surveillance interval has been exceeded and if so WHICH ON Bases on this an acceptable alternative answer should be
"The interval from 7/6/86 to 5/6/87 (for 3 consecutive surveillance intervals) [0.5) exceeds the required 3.25 times the surveillance interval [0.5]".
NRC RESOLUTION: Comment Accepte ,
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FACILITY COMMENT 8.03: This question asks "What action, if any, is required . . . " "and why?" It does not ask for the time in which this action must be completed nor the reference, by paragraph, for this action. Therefore, the answer should read "A plant shutdown should be started (within 1 hour) [I.0] as required by Technical Specifications [0.5] with subsequent action to bring the plant to cool shutdown in accordance with T.S. [0.5]. NRC RESOLUTION: The 1 hour time limit will be required for full credit because T.S. 3.03 prescribes that actions must be performed within a specified time limi FACILITY COMMENT 8.04: TPG ACP-QA-2.06A EO #9d requires the operator to: "Given ACP-QA-2.06A and a request for clearing tags explain the conditions required for clearing including document 6 tion of restoration".
Question 8.04 c. specifically addresses the topic with regard to documentation of restoration but did not provide the candidate the requisite procedur !
Additionally the reference sited in the answer key specifies that 1
"Normally the "Restoration Performed" block should not be filled in when I the tags are issued". The question addressed a specific exception from this narr.;al practice as alivned by the cited referenc It is recommer.ded that 8.04 c. be delete !
NRC RESOLUTION: 1 Comment accepte FACILITY COMMENT 8.05: The question does not ask for the time frame during which "Actions and notifications must be completed" but only for "WHAT" actions and notifications must be completed".
Additionally MP2 learning objectives do not require memorization of one hour (immediate) reporting criteri The answer key should be changed to allow full credit if the candidate stated that RCS pressure must be restored to within its limits and that notifications are conducted in accordance with procedure ,
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NRC RESOLUTION: The time limit will be required for restoring RCS pressure, whereas no time limit will be required to notify the NR FACILITY COMMENT 8.07: This question specifies "B HPSI pump was taken out of service for maintenance". No reference is made to either A or C HPSI pumps. OP 2308 paragraph 7.5 provides operational guidance for removal of B HPSI pump from service that results in "Restoring HPSI pump A to service as Fac I pump or HPSI pump C as Fac II pum ..
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Therefore two HPSI pumps would be available on separate facilitie T.S. 3.5.2 requires "Two seperate and independent ECCS subsystem shall be operable with each subsystem comprised of one operable HPSI pump . . .
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Therefore operability of ECCS is not relying on the action statemen The answer should read "The startup can be commenced [1.0] because the ECCS is operable provided both A & C HPSI pumps are properly aligned
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Additionally, full credit should be awarded if the candidate assumed that the A or C HPSI pump was inoperable and answered in accordance with ke NRC RESOLUTION: Comment Accepte FACILITY COMMENT 8.08: ACP-QA-3.02 provides guidance for non intent changes in Section 6. , Since temporary changes to procedures meet the definition of non intent I changes as specified in paragraph 4.7, they could be treated as a non intent change. Therefore an acceptable alternative answer is: The change is a non intent change [.6]. 1 The change is approved by two licensed SR0's from the unit involved, ) at least one of whom shall be the on duty SS [0.7]. ' The change shall be reviewed within 14 days of implementation (by PORC/SORC) [0.7]. Additionally, in that temporary changes are only allowed if the intent of tha procedure is not altered, then credit should be awarded if the candidate describes those provisions which distinguish between intent and non intent change . t .
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NRC RESOLUTION: The criteria suggested by the Facility (especially No. 2) will be accepted as alternate responses. However, credit will not be given if a candidate describes those provisions which distinguish between intent rad non intent changes because the question did not ask for distinctions . between intent and non intent change l FACILITY COMMENT 8.10: The cited TPG states "Define non compliance per 3.02". No TPG for l Shutdown Cooling Refueling or Technical Specification requires ie ! licensed operator to know from memory those times / conditions i both trains of SDC are not required, when in mode The operator is thus encouraged and trained to reference Tech Specs and Procedures prior to removing a SDC loop from Servic Based on those points it is recommended that this question be delete NRC RESOLUTION: The enabling objective used in the referenca was intended to be TPG SDCS E0 11 (instead of TPG E0 11). This enabling objective states: "State the Technical Specification requirements for ths SDCS operation in Modes 5 and 6." Based on this facility-specific enabling objective, question 8.10 will not be delete FACILITY COMMENT 8.11: l The point distribution of the key answer in unclea l NRC RESOLUTION: l The point distribution was changed and cleariy labelled to aid in the grading proces l .
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