IR 05000317/1988018
| ML20153G711 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/24/1988 |
| From: | Eselgroth P, David Silk NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20153G702 | List: |
| References | |
| 50-317-88-18OL, 50-318-88-18OL, NUDOCS 8809080344 | |
| Download: ML20153G711 (58) | |
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DETAILS Examination Report Nos. 50-317/88-18 50-318/88-18 Facility Docket Nos. 50-317/318 Facility License Nos. DPR-53 and DPR-69 Licensee- Baltimore Gas & Electric C P.O. Box 1475
- Baltimore, MD 21203 Facility: Calvert Cliffs Nuclear Powar Plant Units 1 and 2 Examination Date: June 22, 1988 Chief Examiner: t+h8 David M. Silk hkW Date Operations Engineer Examiner ApprovedBy:/
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Peter y Eselgro W / Chief Date Y 88 f PWR Section, Operationr Branch Division of Reactor Safety Summa ry: On June 22, 1 0 8, a written examination was administered to one i
Senior Reactor Operator candidate from your staff who had previously passed the operating examination and had beea granted a waiver. The candidate passed the written examination. No strengths or weaknesses in the training program could be identified because only one candidate was examined. No exit meeting was held because the written examination was administered in the Regional Offic l l
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8809080344 880825 PDR ADOCK 05000317 Y PDC >
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TYPE OF EXAMINATIONS: Replacement
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EXAMINATION RESULTS: *
l SR0 l l Pass / Fail l l l l l l jWritten Exam l 1/0 l l
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[ Operating Examl Waived l t i I 1 l l l l0verall l 1/0 l l l l l
Attachments: I Written Examination and Answer Ke . NRC RESOLUTION OF FACILITY C0KMENTS FOR THE CALVERT CLIFFS SENIOR REACTOR 1 OPERATOR EAAMINATION ADMINISTERED ON JUNE 22, 1988, BALTIMORE GAS AND ELECTRIC COMPANY LETTER DATED JUNE 28, 1988,
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U. S. NUCLEAR REGULATORY COMMISSION
. SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: C&YERT CLIEES _
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REACTOR TYPE: EMB-QE DATE ADMINSTERED: 68/06/22 EXAMINER: JAGGA ' ' '
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CANDIDATE __
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INSTRUCTIONS TO CANDIDATEL l
! Use separate paper for the answers. Write answers on one sida only.
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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 the examination start % OF ,
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CATEGORY % OF CANDIDATE'S CATEGORY i VALUE_ _ TOTAL SCORE VALUE CATEGORY
- i 25.00 25.00 THEORY OF NUCLEAR POWER PLANT .-
OPERATION, FLUIDS.AND
.I THERMODYNAMICS l
25.00 25.00 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION i
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_25.00 _25.00 PROCEDURES - NORt!AL, ABNORMAL,
-EMERGENCY AND RADIOLOGICAL '
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.25._00 25.00 _ ADMINISTRATIVE PROCEDURES, l CONDITIONS, AND LIMITATIONS [
l 10 _ % Totals I j Final Grace i All work done on this examination is my ow I have neither given j nor received aid, i
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Candidate's Signature I
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1.- Cheating on the examination means an automatic denial of your application and could result in more severe penaltic . Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer Print your name in the upper right-hand corner of the first page of each section of the answer shee l Consecutively number each answer sheet, write "End of Category __" as appropriate, start each catego.y on a new page, write only on one side ,
of the papsr. and write "Last Pasu" on the last answer shee '
' Number each answer as to category and number, for example, 1.4, 6, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
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12. Uso abbreviations only if they are commonly used in facility literatur l 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 ,
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14. Show all calculations, methods, or assumptions used to obtain an answer j to mathematical problems whether indicated in the question or no !
l i 15. Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the exemination are not clear as to intent, ask questions of i the examiner onl '
17. 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 ar istance in completing the examination. This must be done after the examination has i
been complete !
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! 1&, When you com?lete your examination, you shall:
< a. Assemble your examination as follows: !
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(1) Exam questions on to (2) Exam aids - figures, tables, et l (3) Answer pages including figures which are part of the answe '
i b. Turn in your copy of the examination and all pages used to answer the examination question I c. Turn in all scrap paper and the balance of the paper that you did not use for answering the question i d. Leave the examination area, as defined by the examine If after I leaving, you are found in this area while the examination is still in progress, your license may be denied or revoke t
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pilEQBY OF NUCLEAR POWEILELANT OPERATION, Pcgo 2 FLUIDS.AND THERMODYN6MICS
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QUESTION 5.01 (1.50) Why is the Shutdown Margin requirement greater in modss 1-4 than in mode 57 (1,0) With the plant operating at 85% power and all systems in a normal configuration, the operator borates 100 PPM. Shutdown Margin will ... 3 1. Increas l 2. Decreas '
3. Remains unchange (0,5)
QUESTION 5.02 (2.00) How will individual CEA rod worth change as moderator temperature I
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is increased? (0,5) If the worth of an individual CEA rod is measured at a flux level !
of 10(EE-5) % power, how is the worth of the CEA affected if flux l level is raised to 10(EE-4)4 power AND there is NO Change in '
relative flux distribution? Explai (1.0) Beside temperature and neutron flux, what are TWO design factors l affecting the worth of an individual CEA? (0.5)
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QUESTION 5.03 (1.50)
At BOC, power is reduced frvam 100% to 50% and stabilized, briefly
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explain HOW and WHY each of the following plant parameters will be affected over the next 5 hoar Assume all systems are in automatic, rod control is in manual sequential, and no operator action is take RCS temperature ,
i RCS pressure i Turbine Generator control valve position l l
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5 THEQBY_OF NUCLEAB.EQWEB_ELABI-OPERATIOH2 Pogo 3
- ELU1DS.AND THEBdQDXEAMIQS
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QUESTION 5.04 (1.50)
During end of cycle operation with the reactor at 100% power, all rods out, boron concentration at 10 ppm in the RCS, and Tavs 10 F less than Tre :
EXPLAIN HOW and WHY reducing the power to 95% would affect ASI for each of the following means of reducing power. Consider each separately.
' Control rod insertion.
. Boron additio I t Raising Tav !
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. QUESTION 5.05 (2.00) i
' If a trip from 100% power occurs with Xenon at equilibrium, what is the approximate time interval after the trip that Xenon l
i will again be at the 100% equilibrium value? (0.5) ; How would this approximate time interval compare if the trip [
j occurred from 50% equilibrium conditions? (0.5)
r i When a reactor is returned to 100% power from peak Xenon i j conditions, why does the Xenon reactivity value trend become less I j than the 100% equilibrium value for a short period of tirae? (1,0)
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_ _ _ _ _ _ _ THEQRY OF NUCLEAR POWEB_ELABI_QEEBATIQl Pcga 4 Ekt)ll!S. AND THERMODYE&dLQE
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QUESTION 5.06 (2.00)
TRUE or FALSE 7 Critical rod height does NOT depend on how fast control rods are withdrawn, Critical rod height dictates the reactor power level when
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criticality is achieve The SLOWER the approach to criticality, the LOWER the reactor power level will be when reaching criticalit When reactor power is in the SOURCE range, changes in power level do NOT cause a change in Tav QUESTION 5.07 (2.00)
l Unit i has just restarted following a refueling outage while Unit 2 is near EO Annwer the following regarding the differences in plant response between the two units (explain your answers): At a steady power level of 10EE(-8) amps during a startup, equal reactivity additions are made (approximately 100 pcm). Which Unit will have the higher steady state startup rate? At 50% power, a control rod (100 pcm) drops. Assuming NO AUTOMATIC or OPERATOR ACTION, which Unit will have the lower steady state Tavg?
QUESTION 5.08 (1.50)
For the changes listed below (treat each one independently) indicate whether the Moderator Temperature Coefficient will become HORE NEGATIVE, LESS NEGATIVE or have NO EFFECT. (Assume all other parameters are constant) Neutron flux peak shifts radially inward from the eage of the core, Boron concentration decreases 100 ppm while core is at MO Increased nur.ber of burnable poisons are inserted into the cor (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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QUESTION 5.09 (2.00)
Indicate whether each of the following will INCREASE, DECREASE. or NOT AFFECT the Departure from Nucleate Boiling Ratio (DNBR). Assume all other parameters remain constan Primary coolant temperature decreases.
) Primary coolant pressure decrease , Primary coolant flow decreases.
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' Reactor power decrease '
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, QUESTION 5.10 (2.00)
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I Stable natural circulation conditions exist within the RCS with the t
! following parameters.
! Thot - Teold = 25 F f
i SG pressure : 885 psis
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Thot Subcooled Margin indicates 40 F subcooled
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Determine RCS pressur (Show steps used to arrive at your answer). !
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l What will be the temperature of the PORV discharge line if a PORY !
I opens when there is a steam bubble in the pressurizer, quench tank l pressure is 20 psig, and pressurizer pressure ist j psig? ;
j psig?
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ELLIIDS, AND THEBtiODYN6tiLGE
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QUESTION 5.11 (1.50)
What effect would each of the following failures have on a natural i
circulation cooldown which is underway at 490 Explain your answers i
- and consider each failure independently,
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j The steam dump valve which is throttled to control cooldown rate
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' Level is lost in the pressurise ' The auxiliary feedwater valve to one of the SG's falls shu ,
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QUESTION 5.12 (1.50)
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HOW does the coast-down flow provided by the reactor coolant pump
flywheels affect establishing natural circulation flow conditions, i and WHY?
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- QUESTION 5.13 (1.50) (
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- Would fuel center line temperature INCREASE, DECREASE, or REMAIN l j THE SAME in each of the following situations? BRIEFLY EXPLAI i
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j a. Power decreases with constant Tav j l
b. Core age increasen with constant powe l ,.
c. Pressurizer pressure increases with constant powe '
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3 THEORY OF_NMELEAR POMEB PLANT OEERATIO Page 7
' F L U I D S . A N D _. T H E B M O D Y N A M E S
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QUESTION 5.14 (2.50) ;
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! The following questions assume that a spurious reactor trip has just
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occured from extended 100% power operation (an inadvertent trip due to r
- instrument malfunction) and that a pressurizer safety valve lifte and l sticks ope Reactor Coolant Punpa are tripped according to procedure i
{j and RCS pressure drops to 1000 psia. Answer each part independentl I L j If natural circulation flow could NOT be attained, what means I of core cooling exists and will it be sufficient to cool the
! core? (1.0) e i CETs are relatively stable (not increasing) and read 585 F, and ;
a small constant feeding and steaming rate is occuring in the !
- steam generator In what regions of the RCS would the following l l heat removal mechanisms be used?
i j 1.) Boilin ;
j 2.) Superheatin i t j 3.) Condensatio (1.0) r
- Describe how auxiliary spray flow could be used to determine i
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or confirm the presence of RCS voidin (0,5)
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6., PLANT SYSTEMS DESIGN. CONTBQL. AND INSTRUMENTATION Pego 8
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QUESTION 6.01 (1.75)
I State TWO conditions that will cause ALL charging pumps (except those in PULL-TO-LOCK) to start automaticall (1.0) What are the THREE mechanisms employed in the CVCS system to f reduce the Reactor Coolant System activity prior to cooldown? (0.75) '
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4 PLAyT SYEIEdR_ DESIGN COHIBQL AND INSTRQUENTATION Pago 9
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QUESTION 6.02 (1.75)
For each of the following conditions (a.. b., and c.), LIST those actuation signals (1. through 6.) that should have triggere Consider each condition separatel Each condition may have for its answer none, one, or several actuation signal (s). SIAS CSAS CIS SGIS RAS AFAS A steam-line break has occurred in containment and containment pressure : 7 psig contsinment radiation : background S/G levels : -145 inches and -185 inches PZR pressure : 1800 psia RWT level = 30 ft S/G pressure : 680 and 580 psia i A small-break LOCA has occurred and containment pressure : 2 psig containment radiation : 6 R/hr S/G 1evels : -50 inches and -50 inches PZR pressure : 1800 psia RWT level : 20 ft
, S/G pressuce : 700 and 700 psia A feedwater problem has occurred and containment pressure : 1 psig containment radiation : background S/G 1evels : -50 inches and -150 inches PZR pressure : 2200 psia RWT level 30 ft S/G pressure : 720 and 650 psia
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$2 r PLANT SYSTEdg_DSSIGN. CONTBOL. AND INSTRUMENTATION Pego 10
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QUESTION 6.03 (2.80)
The plant is at 100% power and all controls systems are in automati The controlling pressurizer level channel fails lo What system responses will occur and what reactor trip signal, if any, will be generated if no operator action is performed. Setpoints are not required.
QUESTION 6.04 (1.75)
The following pertain to the excore neutron detectors and associated instrumentation for Unit LIST THREE control / protective functions of the wide range channel (0.75) l i What are TWO functions (inputs to other systems) of the control l channels? (1.0)
QUESTION 6.05 (2.50)
What automatic action (s) (other than alarms) occur (if any) when the following process radiation monitors exceed their High Levi setpoints?
Consider each of the three monitors separately, Liquid radwaste discharge monito (0.5) Steam generator blowdown tank radiation monito (0,5) Blowdown Recovery radiation monito (1.5)
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Et, . PLANT SYSTEMS DESIGN COHIBQL, AND INSTRUMENTATION Pcgo 11
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QUESTION 6.06 (2.00)
Will the plant trip as a result of the following instrument failures?
Assume no operator action. Justify your answer, SUR channels A and B fail HIGH during a reactor startup when the reactor is critical at 10E-6% power, SG#11 level channel A fails LOW followed by SG#12 level channel B failing HIGH while at 80% powe Loop #11 To channel A fails HIGH followed by loop #12 Th channel B failing HIGH while at 100% power, The lower UIC detector for Safety channel B fails HIGH followed by Safety channel D upper detector failing HIGH at 80% power.
QUESTION 6.07 (1.70)
The following pertain to the Shutdown Cooling System (SDC), STATE the pressure setpoint that interlocks both SDC suction header isolation valves preventing them from operatin (0.2) What are the functions (bases) of the TWO relief valves (2485 psig and 315 psis) on the SDC suction header? (1.5)
QUESTION 6.08 (1.00)
Indicate whether each of the following is TRUE or FALSE concerning 480 volt Motor Control Center Operation, In the event that 125 VDC control power is unavailable to a 480 VAC disconnect, it can be manually closed by charging the closing spring and depressing the close ba The "Bresker Improper Lineup" alarm is actuated when a third unit disconnect switch is closed with the associated 480 VAC circuit breaker racked ou (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****) PLANT SYSTEME_QESIGN, CONTROL. AND IEEIBUMENTATION Pcgo 12 f
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l QUESTION 6.09 (2.50)
a , State FOUR regulating control rod interlocks / limits that are ,
in effect when the system is in MANUAL INDIVIDUAL control? (1.2) WHAT are the TWO instrumentation signals / conditions that could provide a DROPPED ROD annunciator? (0.8) If a loss of 125 VDC control power to the trip circuit breakers j (TCBs) occurs, what component, if any, will ensure that the TCBs i open if a trip signal is generated? (0.5)
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j QUESTION 6.10 (2.50) i a
I What effect (INCREASE, DECREASE, NO EFFECT) will the following events have on the Thermal Margin / Low Pressure Trip SETPOINT, l Consider each separately. Assume the plant is at 100% power, Teold Loop 1 fails LOW !
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- RCS pressure increases 25 psig
- A Linear Power Range Channel (Safety) fails high i Delta-T PWR Calibrate pot is reduced l I
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l QUESTION 6.11 (1.75)
, Assume that an inadvertant trip of the AFAS "A" channel l j is active for a period of only 15 seconds and then clear ; Briefly describe how the AFAS signal is reset and the associated I
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i equipment returned to a normal configuratio (1.5)
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- How much time must have elapsed before the AFAS signal will i j sesl in? (0.25) [
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Eu,_ELANT SYSTEMS DESIGNi CORIBQL, AND INSTRUMERIATION Peso 13
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QbESTION 6.12 (3.00)
The following questions pertain to a loss of instrument air assuming normal, at power, initial conditions, a. How would a loss of instrument air header pressure due to a rupture just downstream of IA-144/146 (Air compressor isolation from air header) immediately affect the following components / system Choose ONE of the following for each component / system:
A - fail open or flow maximum B - fail closed or flow stopped C - fail as is or flow cannot change D - no immediate effect or system functions normally Main Feedwater Regulating Valves Pressurizer spray valves Letdown Atmospheric Dump Valves AFW regulating valves EDG service water supply valves Auxillary Spray valve Turbine AFW pump speed (if operating) (2.0) Describe TWO means of interconnecting the IA system with backup sources of air pressure. Indicate automatic setpoints, if an Answer this question independently of part "a." abov (1.0)
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! QUESTION 7.01 (1.50)
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! For each operation listed, what is the manual mode of control NORMALLY [
used on the Control Element Drive System? !
l Withdrawing Shutdown Group CEA's during a reactor startup, i f- l Inserting Regulating Group CEA's during reactor shutdown, i
! Recovering a droppped CE !
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j The following questions pertain to information found in EOP-6, Steam j Generatur Tube Ruptur (
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- affected Steam Generato i a
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} Steam Generator Tube Rupture has occurred, i i
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a QUESTION 7.03 (2.50)
The following questions pertain to information found in OP-5, Plant I
, Shutdown from HSB to Cold Shutdow I
! Over a two hour period, the RCS steadily cooled down from 280 F l
- to 230 F. Explain whether or not a cooldown limit has been s i exceede (1.5) j l What is the minimum pressure for the RCS prior to blocking SIAS?
j (0,5)
8 What is the minimum allowable pressure in the Steam Generator j Prior to blocking SGIS? (0.5)
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~ AND*BADIOLOGICAL._EQHIBQL l
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QUESTION 7.04 (2.50) ,
According to EOP-5, Loss of Coolant Accident: State FOUR of the SIX Entry Condition (1.0) {
- State THREE of the FOUR positive indications of Core and/or RCS l voidin (0.75) j What are THREE of the FOUR conditions / parameters the operator {
must confirm / satisfy prior to stopping or throttling a safety injection train? (0.75)
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QUESTION 7.05 (2.00)
' Given a LOCA condition, how soon after SIAS a=>tuation must
"core flush" commence? (0.25) ,
i State the TWO methods that are available for "core flush". ,
Include a brief description of flow paths for each metho (1.5) l What is the reason that "core flush" is established? (0.25)
QUESTION 7.06 (2.00) ;
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The following questions concern RCP Trip Strategy as stated in the EOPs:
State the number of RCP's that should be STOPPED given the following !
condition Consider each situation separatel t
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I' Small break LOCA with RCS pressure decreased to 1775 psi I Small break LOCA with RCS pressure decreased to 1800 psia and a ;
i CIS actuate l l Total Loss of Feedwater event with RCS pressure decreased to ,
1800 psi ;
, Large break LOCA with RCS pressure decreased to 1400 psi i l
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7 Pego 16 PBQQEDMBES - NORMAL 2_A.SEQBdAL. EMEBGEEDY AND BADIOLQ919AL_90BIBQL
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QUESTION 7.07 (1.50)
The following pertain to information found in the Calvert Cliffs Technical Specification State ALL dose rates that apply to designating an area as a High Radiation Area? (0,5)
' How is entry controlled AND dose rates monitored for an area with a dose rate of 1500 mr/hr? (1.0)
QUESTION 7.08 (2.00)
The Functional Recovery Procedure. EOP 8, provides guidance for reactivity contro If CEAs cannot be inserted or drivon in, reactor shutdown may be accomplished by boration, List the TWO flow paths for charging water to the RC List the TWO flow paths for supplying boric acid to the charging pump !
QUESTION 7.09 (2.50)
The following pertain to information contained in EOP-4, Excess Steam Deman What THREE parameters are used to identify the affected S/G7 (0.75) state SIX of the EIGHT actions taken to isolate the S/G' (1.5) Why is it necessary to maximize RCS borction during the initial stages of an Excess Steam Demand event? (0.25) i
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L___FBQQEQQBES - NOBMAIu_ABNORMALu_EMEBQEtNI Pega 17 :
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&ND_ RADIOLOGICAL _CONIBQL
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QUESTION 7.10 (2.00)
l Concerning information found in AOP-1A, Inadvertant Boron Dilation: -
! State FIVE of the SEVEN indicationu/ alarms of a "Dilution While ;
! Critical". (1.0) ,
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! TRUE or FALSE?
Boration and CEA movement may be used CONCURRENTLY to maintain reactor power level constan (0.5) l l
. If CEA's have reached insertion limits as specified in Technical !
l Specifications, what time restrictions are placed on boron i j addition? (0.5) ,
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i i QUESTION 7.11 (1.50) l
! I j The following pertain to information found in AOP-1B, CEA Malfunction l
i How is programmed To maintained following a dropped CEA?
I j Based on Technical Specifications Fig. 3.1-3, Allowable Time to Realign CEA's..... what is the MAXIMUM allowable time to realign l a dropped CEA?
i During CEA reclignment, what method is used to ecmpensate for :
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reactivity changes?
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, EBQQEQQBES __ M Lt_AESQBtie k EtfEBQEtiC1 Pogo 18 AEQ_B&Q1QLQQICAL COSIBQL
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l QUESTION 7.12 (2.00) ,
- t j The following pertain to information found in OP-1A.
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' State the limit for each of the listed parameters where the
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Reactor Coolant Pump (RCP) must be stopped.
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--Lower sen.1 temperatur I
--Motor thrust bearing temperatur (0.8) ;
. What is the maximum time a RCP be run following a loss of CCW? (0.4)
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5 Given the following RCP seal cavity pressures, state which seal l
- (UPPER. MIDDLE, or LOWER) has probably faile Consider each of f i the cases separately, j i
1.) Upper -- 50 psia !
'
Hiddle -- 1400 psia
- 2.) Upper -- 700 psia q Middle -- 1975 psia (0.8) [
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! QUESTION 1.13 (1.00)
) !
In order to maintain the plant at 100% power, work must be performed r inside the containment in a radiation field of 100 MREM /HR gamma and !
q 100 MRAD /HR fast neutron. You will be accompanying the maintenance !
< man into the work area. Assums you are 27 years old and have a !
1 lifetime exposure through last quarter of 29 REM on your NRC Form I i So f ar this quarter, you have accumulated 1.0 RE i l
,
How long nay you stay in this area without exceeding the Calvert I
) Citffs administrative limits? Show all assumptions and caletintion '
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_ _ _ _ _ _ . ADMINISTRATIVE PROCEDURES. CONDITION Paga 19
' AHD' LIMITAflQtLE
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QUESTION 8.01 (2.50) Explain why the following leakages are IN or OUT of compliance with the Limiting Condition-for Operatio Consider each INDIVIDUALLY and SEPARATEL . Unidentified-----------------------------------0.65 GPM Steam Generator Tube Leakage:
A----------------------------------------------0.45 GPM B---------- -----------------------------------0.50 GPM Vario anual vent and drain valves seat and pack. , gland leakage.----------------------4.6 GPM Back leakage through SI check valves detected '
by leak check on previous shift.----------------3.9 GPM (1.0) Evaluate Technical Specification compliance if the leakage in
"a." is cumulative. (0.75) Calvert Cliffs Technical Specifications require that systems be operable for detecting leakage from the RC List the 3 systems or methods used to detect leakage to the RC (0.75)
.
QUESTION 8.02 (1.50)
Ste'.a the approval requirements, by job title / position, for changes to the following types of procedures accoruing to CCI-101 Emergency Operating Procedure Operating Instruction Nuclear Engineering Operating Guid (***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ADMINISTRATIVE PROCEDUBES. CONDITION Paga 20 AND' LIMITATIONS
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QUESTION 8.03 (3.00) What combination of THREE parameters shall not exceed the limits of Fig 2.1-1 located in the Calvert Cliffs Technical Specifications to prevent exceeding the Reactor Core Safety Limi (1.5) What is the objective or basis for the Technical Specification Reactor Core Safety Limit? (0.5) Assuming Mode 1 operation, state the action reqoired by the Technical Specifications to be taken within one hour if th+
Reactor Core Safety Limit is exceede (1.0)
QUESTION 8.04 (2.00)
Unit 2 is in the process of a plant startup, reactor power is about 2%.
You note that the loop temperatures are all 500 In accordance with the Unit 1 Technical Specifications: What actions must be taken? What are the FOUR bases for the Minimum Temperature for Criticelity?
QUESTION 8.05 (2.50)
Answer the following in accordance with Calvert Cliffs Technical Specifications and procedure CCI-140D "Shift Staffing". Assume both i units in Mode l What actions must be taken if operating with a minimum crew composition and one of the Reactor Operators becomes 1 incapacitated? (1.5) ! Is the following action proper by the Shif t Supervinor? Justify your answe Fifteen minutes before scheduled arrival of the on-coming shift, one of the four on-coming RO's calls in sic The on-shift SS decidos, due to the shift's present overtime condition, to call in a replacemen The SS also decides that since the i replacement should arrive shortly after shift change (about 30 )
minutes) to send his people home and let the next shift start with three R0 (1.0)
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ADMINISTRATIVE PROCEDURES. CONDITIONS, Pcgo.21 AND' LIMITATIONS
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QUESTION 8.08 (2.00)
What are the TWO reasons identified by Technical Specifications for limiting the blow down of one steam generator in the event of a Main Steam Line Rupture?
QUESTION 8.07 (1.50)
Commensurate with Calvert Cliffs Technical Specifications, Primary Conte.inment Integrity must be established in operational MODES 1,2,3,and What plant conditions constitute CONTAINMENT INTEGRITY 7 QUESTION 8.08 (2.00)
Unit 1 has been in Mode 6 for 30 days when the Refueling Pool is drained to facilitate installation of the reactor vessel hea After the pool has been drained, 11 LPSI pump develops a breaker problem and must be removed from servic What actions must be taken AND in what time period according to Technical Specification 3.9.8.27 What alternate action may be taken to comply with Technical Specifications?
i l
l QUESTION 8.09 (3.00)
a. Under what THREE conditions may the independent verification of safety tags be waived?
b. Wb t TWO people may authorize a waiver of the independent
'
fication of safcty tags? Indicate people by title, c . .a.iple component tagout, when are the control board s v cv be tagged /untagged?
(***** CATEGORY 8 CONTINUED ON NEXT PAGE *****) ADMJHISTRATIVE PROCEDUBES. CONDITIQN Paas 22 AND LIMITATIONS
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QUESTION 8.10 (2.00)
Given each of the events described below, along with ":eAL Criteria",
pages 14 and 15 of the EPIP, (included with the examination) classify each event separately to the correct EAL. Note that some events may not fall into any EA State any assumptions used in making a classificatio While operating at 100% power, steady state, a bomb threat is received stating that an explosive device has been placed in the control roo Upon conducting a search of the control room, what appears to be such a deci.ce le discovere Examination of the device indicates that it may be triggered by radio control or movemen (0,5) While moving fuel in the core, during refueling outage, all off-site power is lost due to line damage from high wind After discussion with power dispatcher, offsite power will be restored in 20-24 hours. Onsite power systems operate according to des. tan, with adequate fuel oil for at least 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> of operatio (0,5) While operating at 100% power, steaoy state, a double ended rupture of one RCS cold leg occur (0,5) While operating at 100% power, steady state, a main steam line break occurs downstream of containment but upstream of the main steam isolation valv Subsequently, calculated steam generator tube leakage increases from 0.2 gem at steady state to 40 gpm with no indication of secondary activity in the intact loo (0.5)
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3 ADMINISTRATIVE PROCEDURES. CONDITION Pago 23 AND'LIMITATIOND
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QUESTION 8.11 (3.00)
For each of the following everats explain briefly why the NRC SHOULD or SHOULD NOT be notified within i hr according to CCI-118- :
a. During instrument testing while at power, three pressurizer pressure safety channels are momentarily place in bypas b. While critical at 1% power on Unit 2, Tave drops below 515 F and then returns to norma c. Refueling water tsnk level falls below 400,000 gallons and cannot be restore I d. During surveillance testing an expected actuation of LPIS l train A occur (3.0)
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i l THEORY OF NUCLEAR POWEB PLANT OPERATIO Paga 24 FLUIDS.AND THERMODYNAMICS -
.
.
ANSWER 5.01 (1.50) In mode 5 the plant is completsly cooled down (belor 200 deg. F) so that there is no <.eed to have reserve shutdown margin to offset the positive reactivity addition resulting from a rapid cooldown accident while in modes 1- (1,0) (Increase) (0.5)
REFERENCE R0-302-3-1; Calvert Cliffs: T. p 1-3 '
EO .5/ K519 19*.004K107 191004K112 ..(KA's)
- ANSWER 5.02 (2.00) Worth will increase (0,5) with an increase in moderator temperatur No change. [0.5) The absolute value of neutron flux will not change the wort A shift in flux distribution is required. [0.5) (1.0)
' CEA size, CEA absorber material, CEA location. [any 2, 0.25 ea)(0.5)
REFERENCE CE Trng Ctr Rx Theory Notes, Pp. 181-183
.
R0-302-2-1
! EO 4.1, .9/3.4 2 5/ '
192005K107 001000K502 193003K125 ..(KA's)
< ANSWER 5.03 (1.50)
a, Decreases (~15 F) [0.25] due to buildup of Xe (0.25) Held constant (0.25] by PPCS spray and heaters (0.25) Increases (0.25] due to lower S/G pressure (0.25)
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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_ _ _ _ _ _ _ _ _ _ _ THEORY OF NUCLEAR POWER PLANT OPERATIO Paga 25
' ELQ1DS.AND THERMODYNAMICS
.
REFERENCE Calvert Cliffs; SD #5, RCS, p 25 SD #23-1, Turbine Control and Protection System, p 54 RO-302-3-1 pp.75 EO 7.14 RO-62-1-1 LO 06201K602 3.5/ .2/ .6/ .6/3.6
'
039000K508 002000K510 001000K533 192008K124 193008K105
..(KA's)
ANSWER 5.04 (1.50) Adding negative reactivity to the top of the core causes the ASI to become more positive (power is driven to the bottom of the core). A change in Th as power decreases is greater than the change in T With a -MTC, less negative reactivity is inserted in the top of the core than the bottom due to positive reactivity feedback. Also, the HTC is more negative at the temperatures at the top of the cor ASI becomes more negativ . As SG steam flow is decreased, Tc increases. The hotter Tc entering the core reduces reactor power. At the lower reactor power there is a smaller delta T across the cor The net result is that Th increases more than Th decreases. 4kk MTC is more negative at the top of the core than at the bottom so the effects are ximetalv ^# and ACI deee no algn4.ficantly S.._R$C 5W" f*-eedt4 * "Y * " ed han e 69dM
[0.5 ea.]
REFERENCE RO-302-4-0 pp. 5, 6, 7 EO 1.1, 2.2, .3/3.7 2.8/ K111 015020K503 103008K122 ..(KA's)
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_ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ THE0BY OF NUCLEAR POWEB PLANT OPERATIQH2 Pcgs 26 ,
' FLUIDS.AND THERMODYEAMISS -
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ANSWER 5.05 (2.00)
a. ~ 24 Hrs. (accept 20 - 30 hrs.) (0.5)
b. Time would be shorte (< 20 hrs.) (0.5)
c. Due to the time delay in Xenon production from the decay of Iodine. (1.0) (Will accept burnout as a correct response).
REFERENCE NEOG 7 Fig. 1-II.D. RO-302-2-1 pp. 48,49 E0 1 .1/ .4/ .4/ K107 192006K106 192006K105 ..(KA's)
ANSWER 5.06 (2.00) T F F
- T [0.5 ea.]
REFERENCE RO-302-3-1 EO .8/ K105 ..(KA's)
l ANSWER 5.07 (2.00) ' Unit 2 (0.5] due to a lower Beta effective coefficient at EOC (0.5] Unit 1 (0.5) due to MTC being less negative, (so Tavg must decrease more to add + reactivity). [0.5)
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5, , THEQBLOF NUCLEAR POWER PLANT OPERATIOL Pega 27 FLU' IDS.AND THEBMODYNAMICS
.
REFERENCE RO-302-3-1 pp. 65,66 EO 7.11 3.8/ .9/ .9/ K510 001000K549 192008K120 ..(KA's)
ANSWER 5.08 (1.50) Less More More (If assume same Boron conc.,will accept no effect) [0.5 ea)
l REFERENCE RO -302-2-1 pp. 20-26 EO .1/ .3/ K526 192004K106 ..(KA's)
ANSWER 5.09 (2.00) Increase Decrease Decrease Increase (0.5 ea.]
REFERENCE
'
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RO-301-13-0 pp. 26-28 EO .4/ i i
193008K105 ..(KA's) '
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- THEGRY OF NUCLEAR POWER PLANT OPERATION, Pass 28 l FLUlDS.AND THEBMODYNAMICE i
.
.
ANSWER 5.10 (2.00)
c Teold = 503 F corresponding to saturation temperature for 700 psia
[0.25]
. Thot = Teold + 25 = 528 F [0.25]
40 F subcooled = 528 + 40 = 568 F (0.25]
568 F corresponds to 1207.72 psia (0.25]
290 AW AW . -225 F (320 - GSO)
2. d ie- F ( 40G - 430 ) [0.5 ea.]
Jto J2e aro REFERENCE Steam Tables R0-301-8-0 TO 8. .3/ .8/ .6/ K115 193003K124 193003K125 ..(KA's)
ANSWER 5.11 (1.50) Incr .se cooldown rate (0.2) since more energy is being removed frorc he primary. (0.3] , May interrupt natural circulation (0.2] since hot legs maybe voide [0.3] OR No effect (0.2] if hot legs do not become voided (0.3) Decrease cooldown rate (0.2) since SG tubes will become uncovered reducing heat removal. (0.3] (Will accept: No longer feeding with cold water). .
REFERENCE RO-301-14-0 (CAF)
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_____ THEQRY OF NUCLEAR POWER PLANT OPERATIO Pcga 29 FLUIDS AND THERMODYM&tiLGE
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ANs..dR 5.12 (1.50)
Coast-down delays natural circulation'[0.75] because it takes longer to establish a significant core differential temperature (0.75).
REFERENCE i
RO-301-14-0 EO .9/ K121 ..(KA's)
ANSWER 5.13 (1.50)
- Decrease (0.25), smaller delta T required to transfer more energy from RCS (0.25). Decrease (0.25), fuel swelling and clad creep reduce clad gap which reduces delta T across gap and lowers center line temp (0.25). No change (0.25), pressure has little effect on heat transfer in subcooled fluids (0.25). Accept increase if the assumption is stated that increasing pressure decreases nucleate boilin ,
REFERENCE
,
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RO-301-13-0 EO 13.4, 1 RO-301-12-0 pp. 6,7 2.2*/2.4* 2.3*/ .5/ K101 193008K130 193008K128 ..(KA's)
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_ _ _ _ _ _ _ THEORY OF NUCLEAR POWER _ PLANT OPEBATIO Paga 30 FLlTIDS. AND THERMODYNAMES
.
J ANSWER 5.14 (2.50)
Y23 a. Once through cooling [0.5) .NO'[0.5]
b. Boiling in the covered portion of the core (0.33]
Superheating in the uncovered portion of the core (0.33]
Condensation in the S/G U-tubes (reflux boiling) [0.33)
c. Rapid increase in pressurizer level during Aux. spray [0.5)
REFERENCE a. CEN-152 Rev 3 pg 5-31 and 5-34 EOP-5 Rev 0 pg 3 b. CEN-152 Rev 3 pg. 5-34 to 5-35 and pg. 5-91 c. E0P-5 Rev 0 pg. 17 EO 14.2.1; 14. .0/ .0/ .5/ K103 000074K311 000074A206 ..(KA's)
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--. . PLA,NT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 31
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f ANSWER 6.01 (1.75)
!- The pumps will automatically start upon: Pressurizer level deviation = -15' inches Receipt of an SIAS (0.5 ea.) . Degassification Filtration Demineralization (De-ionization) [0.25 ea.] (0.75)
!
REFERENCE
'
SD 6, Pp. 29, 20-22 Lesson Objectives 00603K505
00603K601 00604K501 2.9/ .0/ .6/ .8/ K115 004000K101 004010A202 004000A101 ..(KA's) ;
ANSWER 6.02 (1.75) ,
, SIAS, CSAS, CIS, SGIS (0.25 ea.) , CIS, SGIS (0.25 ea.) '
None (0.25]
REFERENCE
'
SD-63 pp. 45-60 R0-34-1-1 pg. 7 Lesson Objectives - unavailable ,
3.8/ .2/ l 013000K101 012000A301 ..(KA's)
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ANSWER 6.03 (2.80)
PZR heaters deenergize Letdown flow control valves close to minimum Both backup pumps start Auto makeup to VCT initiates (due to charging / letdown mismatch)
Sprays initiate (to reduce pressure from compressing the vapor space)
Reactor trips on high PZR pressure PORVs open [0.4 pts each]
REFERENCE RO-62-1-1 p. 23
,
SD-62 Fig. 62-8, Fig. 62-11
Lesson Objectives 06202K404 06202K405 3.6/ .1/ .7/ .8/ .2/ .3/ K401 011000K301 011000K104 011000K103 011000K102 011000K101 ..(KA's)
- ANSWER 6.04 (1.75)
'
a. SUR trip to the RPS !
SUR enable
, Zero mode bypass [0.25 ea]
b. Input to PRCs Input to Internal Vibration Monitoring System. [0.5 ea] '
REFERENCE RO-57-1-2 SD-57 pp. 47-50, Fig. 57-14, pg. 52 !
Lesson Objectivos - none available '
4.1/ .7/ .1/3.5*
015000A202 015000A302 015000K101 ..(KA's)
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6.. PLANT SYSTEMS DESIGN, CONTEQLt_AEQ_lEETRUMEEIAllgM Pega 33
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ANSWER 6.05 (2.50) The liquid discharge isolation valves are close (0.5] Surface and bottom blowdown isolation valves are close (0.5) Diverts blowdown recovery flow to the HW Closes blowdown isolation valve to Cire. water.
.
Closes blowdown isolation valve to the condense (0.5 ea)
REFERENCE -
SD-14B p. 6 SD-18 pp. 17, 27
'
ANSWER 6.06 (2.00) !
' No trip (0.25]
The trip will not occur until power reaches 10E-4% power (0.25) No trip (0.25]
Only the auctioneered LOW signal is selected, therefore only i channel A will trip (0.25)
- TRIP (0.25)
Due to Delta T decreasing and causing setpoint to change (0.25)
' TRIP (0.25)
Both APD channels will trip (0.25)
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REFERENCE i
SD No 59, pgs 10-31 Lesson Objectives 05904K501 3.9/4.3 3.1/ a 012000K603 012000K403 ..(KA's)
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- PLANT SYSTEMS DESIGN. qQNTROL. AND INSTRUMENTATION Paga 34
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ANSWER 6.07 (1.70) >/= 300 psi (0.2) psig - Protect piping from overpressure due to a sudden temperature increase in containmen (0.75)
315 psig - Protect piping from overpressure due to simultaneous operation of charging pumps and SDC with the pressurizer soli [0.73)
REFERENCE SD-7 pp. 42,43,47 Learning Objectives - not available 3.2/3.5* 3.2/ .8/ .9/ K104 006000K409 005000K407 005000K402 ..(KA's)
ANSWEP, 6.08 (1.00) true (0.5) true (0.5)
REFERENCE SD-53 pp. 19, 44 3.2/3.3
' '
062000G009 ..(KA's)
i ANSWER 6.09 (2.50) '
' Upper electrical limit '
>
'
Lower electrical limit l CWP CEA withdrawal prohibit CMI CEA motion inhibit (0.3 pts each)
1 . Rod drop from Reed switch NI negative rate of power change from NI syste [0.4 pts each) UV trip devives (0,5)
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1 l l PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUBENTATIQH Paga-35
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REFERENCE
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SD-60 pp. 34, 39, fig. 60-18 SD-57 p. 43 COS AA-24,AB-24 Learning Objectives RO-60-3-0 0600201 0600103 3.4/ .5/ .7/ .4/ .9/ K201 001000X604 001000K407 001000K105 001000K103
..(KA's)
ANSWER 6.10 (2.50) No effect Decrease (No Effect if only considering the change in flow) No effect Increase No effect [0.5 pts each]
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REFERENCE SD-59 .9/ .3/ .9/ K611 012000K501 012000K402 ..(KA's)
ANSWER 6.11 (1.75) The AFAS signal will reset automatically.[0.5] MS-4071 will close when the signal clears, [0.5] the AFW pump must be manually stopped. [0,5) The signal vill seal-in after 18 second [0.25] ,
REFERENCE RO-34-1-1 pp. 1-9 Enabling Objective 1.02 4.1/ .7*/ .3*/4.5*
013000K404 013000A206 013000K107 ..(KA's)
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. PLANT SYSTEMS DESIGN.-CONTBQL AND INSTRUMENTATION Peca 36
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ANSWER 6.12 (3.00) . C D D (0.25 each] D D A D A . Auto valve to plant air system X-ties PA to IA at 85 ,
psig IA pressure (0.5] ' manual X-tie valve to Saltwater system air compressor (0.5] X-connect Units 1 and 2 plant air systems (0.5]
[any two for 0.5 each]
REFERENCE S.D. 32 pg 15
. AOP-7D pp. 1,3-6 S.D. 41 Fig A-7 to A-9 :
S.D. 39 pg 21 l 3.4/3.6 3.2/ K402 078000K302 ..(KA's) *
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' PRO,CEDURES - NORMAL. ABNORMAL..EMEBGENCY Pega 37 AND RADIOLOGICAL _CQN.LTROL
, -
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ANSWER 7.01 (1.50)
a. Manual Group b. Manual Sequential c. Manual Individual [0.5 ea.]
REFERENCE OP-2 p.11 OP-4 p. 6 AOP-1B 4.0/ .5/ SG006 001000A403 ..(KA's) -
ANSWER 7.02 (2.00) . Mismatch in feed flow prior to the tri . Unexplained increase in S/G level prior to the tri . Main Steam Line RM . Post-trip S/G level change . S/G sample [any 4 @ 0.25 ea.] . Condenser off-gas RM . S/G Blowdown RM , MainSteam Line RM . Main Vent RM [4 @ 0.25 ea.)
REFERENCE
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EOP-6 pp. 3, 10 4.4/4.6 4.5/ .9/4.2*
000038A204 000038A202 000038A203 ..(KA's)
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. PROCEDURES - NORMAL. A]3E9EMAL. EMEEGEgg Paga 38 AND RADIOLOGICAL CONTBQL
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ANSWER 7.03 (2.50) No [0.5]. Even though the RCS was cooling down at a rate of 25 F/hr, while below 250 F the RCS temperature did not exceed the 20 F/hr cool-down limi (1.0) psi (0.5) psi [0,5)
REFERENCE OP-5 p , 4 3.4/ SG10 ..(KA's)
!
ANSWER 7.04 (2.50) . Unexplained decreasing presurizer leve . Unexplained decreasing pressurizer pressur . Loss of RCS subcooled margi . High Containment radiation alar . Increase in containment sump leve . Increase in containment sump alarm frequenc [any 4 at 0.25 ea.) . Letdown flow greater than charging flo . Rapid increase in pressuriser level during an RCS pressure reductio . Loss of subcooled margin as determined using CET temperature . "REACTOR VESSEL WATER LEVEL LOW" alar (any 3 at 0.25 ea.)
i . RCS subcooling > or = 30 . PZR level > 155 in., and etabl . At least i S/G available for heat removal
, RVLMS indicates core covere (any 3, 0.25 ea.)
i
, REFERENCE EOP-5 pp. 21, 11, 4 LO 201GA09 3.9/4.3 3.4/3.6 4.3*/4.5* 3.7*/ I I
000009A104 000011SG11 000009K310 000011A211 ..(KA's) l l
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. - _ - PROCEDUBES - NORMALu_AREQBtiAL' EMERGEdgl
. Paga 39 AND RADIOLOGICAL CONTRQh
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ANSWE .05 (2.00) hour (0.25) Pressurizer injection [0.25]-- HPSI pump discharge to CVCS thru Auxiliary HPSI header to Aux. Spray to RCS thru surge line
'. loop 2 Th). [0,5)
Hot Les Injection (0.25]-- LPSI pump from Containment Sump thru recirculation line and SDC return header into RCS loop 2 Th. [0.5) Prevent restriction of core flow due to boron crystallization. (0.25)
REFERENCE SD-7 pp.67, 68; EOP-5 p. 28 LO 201EK312 3.8/ K313 ..(KA's)
ANSWER 7.06 (2.00) None ! All (4) All (4) Two (0.5 ea.)
REFERENCE ,
EOP-0 p. 5 LO201EK306 3.4/3.7 4.1/ K314 000009K313 ..(KA's)
ANSLfER 7.07 (1.50) >100 ar/hr <1000 mr/hr >1000 mr/hr (0.25 ea) l
! SWP issue Radiation monitoring device indicating dose rate provide Louked barricades with keys maintained by Shift Supervisor or Supervisor-Radiation Contro (0.33 ea)
t (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) f PROCEDURES - NORMAL. ABNORMAL ~. EMERGEHQY Pcco 40 AND RADIOLOGICAL CONIB9L
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REFERENCE CC Tech. Specs pp. 6-20, 6-21 2.8/ K103 ..(KA's)
ANSWER 7.08 (2.00) . Normal CVCS lineu . Charging thru the Aux. HPSI heade [0.5 ea] . BAST-- via gravity feed and boric acid pump . RWT-- to charging pump suctio [0.5 ea.]
REFERENCE EOD-8 pp. 6-7 4.2/ .4/ .4/ K123 004000K122 000029K311 ..(KA's)
i ANSWER 7.09 (2.50) S/G pressures :
T cold !
S/G Level [0.25 ea.] . Shut MSIV' . Shut Feed Water isolation valve . Shut MSIV bypass valve's, Shut Blowdown valve . Shut AFW Steam supply valve Shut AFW block valve l Shut Atmospheric dump valve . Shut upstream drain [any 6 @ 0.25 each) The cooldown adds positive reactivity and the boron adddition prevents a return to criticalit [0.25)
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7.. PROCEDURES - NORMAL. ABNORMALJMERGEILQY Pago 41 AND' RADIOLOGICAL CONTROL
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REFERENCE EOP-4 pp.5, 6, 9 4.5/4.7 4.3/ .1/ K105 000040A104 000040K304 ..(KA's)
ANSWER 7.10 (2.00) . Increasing power level on Nuclear Instrument . Increasing T . Decreasing boron concentration indicatio . Demin, water flow alar . BA flow alar . Low BA pump discharge pressure alar . Automatic withdrawal prohibit alar [5 at 0.2 ea.) TRUE (0.5) Borate in 5 sec. increment (0,5)
REFERENCE AOP-1A LO 00603003; 00603K401 3.8/ .8/ A206 004000SG15 ..(KA's)
ANSWER 7.11 (1.50) With turbine loa min, Regulating CEA motio (0.5ea.]
REFERENCE AOP-1B pp. 5-6 T.S. Fig. 3.1-3 3.8/ K304 ..(KA's)
(***** CATEGORY 7 CONTINUED ON NEXT PAGE *****) PROCEDURES - NORMAL. ABNORMAL. EMERGENCY P2ga 42 AHQ RADIOLOGIQAL CONTROL _
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ANSWEk 7.12 (2.00) F 195 F [0.4 ea) mi f.0.4) .) UPPER 2.) LOWER (0.4 ea.]
REFERENCE
RO-106-1-1 pp. 15-17 OP-1A p. 1 LO 00501K704 '~
00503K702 3.5/ .7/ A202 003000A201 ..(KA's) i l
AN.9WER 7.13 (1.00)
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5(N-18) = 45 REM [0.25) i i
Total lifetime to date : 29 + 1 = 30 REM (0.25) i Total lifetime available : 45 - 30 = 15 REM Total this quarter available = 2 - 1 = 1 REM i
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Quarterly limit is more restrictive than annual limit gamma + neutron = total
' 0.1 REM /HR + (0.1 RAD /HR)(10 QF) = 1.1 REM /HR dose rate (0.25)
1.0 R/1.1 REM /HR = 0.91 HRS = 53 MIN CO.25)
I I REFERENCE l
CCI-800B p. 0 3.1/ . 5 '3. 4- l
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1.122 r '
! 194001K103 194001K106f ..(KA's)
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_ - _ _ _ _ _ _ _ _ _ _ _ _ _ ADMINISTRATIVE PROCEDUBES. CONDITION Pago 43 AND LIMITATIONS
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ANSWER 8.01 (2.50) . In compliance, unidentified limit is 1 GP (0.25) In compliance, Total leakage less than 1.0 GPM [0.25] In compliance, identified leakage spec. is 10 GPM [0.25] In compliance, identified leakage spec, is 10 GPM [0.25] Not in compliance (0.5) as total leakage is 10.1 GPM. [0.25) (0.75) . Containment atmosphere pariculat ; Containment sump level alar . Containment atmosphere gaseou [0.25 ea.]
j REFERENCE CC TS 3/4. .6/ .6/ K401 002000SG06 ..(KA's)
, i ANSWER 8.02 (1.50)
t 4 Two SRO license holders, one of whom must be the Shift Supervisor or Supervisor- Procedural Developemen Same as a. abov Two members of the plant management staff, one of which holds an SRO license on the affected uni (0.5 ea.) ,
I REFERENCE CCI-101J pg. 4 !
3.3/ i
., 194001A101 ..(KA's)
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. - _ - ADMINISTRATIVE PROCEDURES. CONDITION Pcgo 44 AMD LIMITATIONS
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ANSWER 8.03 (3.00) Thermal Power, Pressuriser Pressure, highest loop Tc temperatur (0.5 ea.) : Maintain integrity of fuel cladding --OR--
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To prevent the release of significant amounts of fission products to the primary coolant --OR--
To maintain the DNBR greater than 1.21 (0,5)
i Be in HOT STANDBY; notify NRC Operations Cente (0.6 ea.)
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i REFERENCE l CC Technical Specifications p , B2-1, 6-13 3.6/ .6/ G006 002020G005 ..(KA's)
t ANSWER 8.04 (2.00) ca- , /' Restore Tavg to > 515F (Ar25) within 15 minutes (G<2$) C R ,
jk."' 0F be in Hot Standby (0,% 2S ) within the next 15 minutes (9 25)
o . Ensure MTC is within analyzed rang (0.25) Protective instrumentation is within normal operating rang (0.25] Pressurizer is capable of being operable with steam bubbl (0.25)
, Reactor vessel is above minimum RT-NDT temperatrue (0.25]
REFERENCE CC TS U-2 pas 3/4 1-7, B3/4 1-2 3.3/ .6/ .
a 002000G005 0020000011 ..(KA's) ,
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, SJDBJNISTRATIVE PRQQEDURES. CONDJ11QL Pega 45 ,
AND LIMITATIONS :
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ANSWER 8.05 (2.50) Obtain a qualified relief [0.75) within two hours (0.75] Yes (0.50) the number of RO qualified personnel present would still meet TS requirements (0.50)
REFERENCE
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CC TS, pg 6-5 CCI-140D pg i ;
j 2.5/ A103 ..(KA's)
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ANSWER 8.06 (2.00) Minimize the positive reactivity effects of the RCS cooldown associated with the blowdow (1.0) Limit the pressure rise w/in the containment in the event
! that the rupture occurs inside the containmen (1.0)
REFERENCE
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CC Technical Specifications B 3/4 7-3 3.7/3.7
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039000K405 ..(KA's)
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. ADMINISTRATIVE PROCEDUBES. CONDITION raga 46 AMD'LIMITAllQEE
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ANSWER 8.07 (1.50)
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Containment Integrity shall exist when:
(a) All penetrations requ. red to be closed during accident conditions are either:
(1) Capable of being closed by an operable containment automatic isolation valve system, or (0.25)
(2) Closed by manual valves, blind flanges, or deactivated automatic control valves secured in their closed positions or as specified in the specification (0.25)
(b) All equipment hatches are closed and seale (0.25)
(c) Each air lock is operable (0.25)
(d) Containment leakage rated are w/in limit (0.25)
(e) The sealing mechanism associated with each penetration is operabl (0.25)
REFERENCE CC Technical Specifications, 1-2 3.1/ .4/4.1*
000069G008 103000G011 ..(KA's)
ANSWER 8.08 (2.00) Initiate corrective action to return 11 LPSI pur, to '>orable status within one hou Sne Spent Fuel Pool Cooling loop may be lined up +o pr' .'ide cooling
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(***** CATEGORY 8 CONTIllUED ON NEXT PAGE *****)
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3 4D41HISTRATIVE_FB9CEDUBEL_CONDIII.DUL Pase 47 I AUD_LldlIAIIDES I
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M FERENCE CC TS 3.9. LER 87-001EO 11 2.7/ SG06 ..(kA's)
ANSWER 8.09 (3 00)
a. Emergency [0.4]
High radiation or hazards area [0.3)
Operationally, tested and not a locked valve (0.3]
s o e s . c.g w w.- 4,a 4 gi., r .,awa a3 b. Shi t t. Streevisor [0.5)
Sat +te -T4ggittg-Supexuiaoe [0. 5 )
c .- u .#, s n r. - . s < . ,e -
c. Tagged First (0.51
'jntaggea last r0,5)
REFERENCE Calvert Cliffs: CCI-112, .a. .7/ K102 ..(KA's)
ANSWER 8.10 (2.00) Alcrt (CR evacuation may be anticipated) (0,5) , None (0,5)
- . S. A. (Failure of 2 FPB; G.A.E. if explein challenge to Srd). (0.5) S. A. (Failure of 2 FPB; MSLB w/ SGTR & uncontrolled release). (0.5)
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REFERENCE a CC EPIP Sect. pp. 14-15 3.4/4.4* g
194001A116 ..(KA's)
i ANSWER 8.11 (3.00) ,
a. Should report (0.35] plant is operated outside design basis (0.4] !
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b. No report (0.35] needed when an action statement for LCO is entered (0.4]
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c. Should report (0.35] shutdown due to inability to meut LCO l action statement requirement (0.4]
j d. No report (0.35] for ESF actuation during surveillance l j
(3.0)
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testing (0.4]
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e REFERENCE -
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Calvert Clif f s Instructions 11d-5, p1 !
l T.S. Unit 2, p 3/4 1-7
) 4.1*/ A102 ..(KA's) .
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(********** END CF EXAMINATION *******'***)
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NRC RECOLUTION OF FACILITY COMMENTS FOR THE
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CALVERT CLIFFS SENIOR REACTOR OPERATOR EXAMINATION
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ADMINISTERED ON JUNE 22, 1988 Quostion Number: 5.04 Fccility Comment: 3) Raising Tav The analysis of temperature change across the core is correct. However, the result would be in an increased negative ASI, power moving toward the top, as a result of the temperature chang Tc increasing adds negative reactivity to the lower half of the cor Th decreasing adds positive reactivity to the top of the cor Therefore power shifts to the top of the core. The more negative MTC in the upper core aids this process causing a larger power (ASI) swin Reference: Sama as state NRC Resolution: Agree with commen ,
Reference stated supports facility comment and does not
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fully support answer ke Answer key is changed to reflect facility commen Quastion Number: 5.07 l Fccility Comment: The question used units of PCMS in lieu of % for reactivit Although this did not confuse the candidate, it was inconsistent with material presented to build the exam. The unit of PCM is not used at Calvert Cliff NRC Resolution: Comment Note References to the unit PCM will be deleted from the questio In addition, units of AMPS are also deleted from the question.
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Quoation Number- 5.10 Facility Comment: F (255-265) for 1700 psig F (320-380) for 700 psig This is a constant enthalpy expansion to 35 psi Additionally, units of psia & psig should remain consisten Reference: Steam Tables NRC Resolution: Agree with commen Answer key changed to reflect proper answe _ . _ . ,
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Quostion Number: 5.14 a.
Fkeility Comment: a. Once through cooling is. correct. However, the
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answer key states incorrectiv that this method is inadequate. The references stated do not apply to this method of cooling and therefore cannot be used
, to verify the answer. A phone conversation with D :
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William Dove of CE verified this e viable cooling
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method. The answer should be "yes".
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NRC Resolution: Agree with comment.
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Answer key changed to YE ! Qucation Number: 6.04 j
j Fccility Comment: "Control Channels" is not standard Calvert Cliffs E
- terminolog Therefore, this quastion could be j interpreted as reactor Regulating or the narrow range t safety NI channel If narrow range NIs are assumed, ;
the answer would change to any two of the three lissted i
belo l t
! Inhibit SUR trip !
I l Enable loss of load trip i Enable APD trip
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NRC Resolution: Disagree with comment.
4 Reference stated in facility comment (RO-57-1-3) was not supplied to the grude Reference stated in the j [
examination (RO-57-1-2 page 39) refers to the Power Range -
l Control Channel. Question premise refers I i to "excore neutron detectors and associated i
! instrumentation" During the examination the candidate ;
- asked "What is meant by Control Channels?", he was told i j they refer to Power Range NIs. No change to answer ke !
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Qucation Nu bar: 7.13
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Focility Comment: This knowlwdge is beyond that expected for an SRO at
CCNP Radiation Protection personnel are on shift to
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provide the necessary s.uppor We request this question be droppe NRC F.esolution: Comment noted, i
NUREG-1021, Operator Licensing Examiner Standards, lists items that are to be included in the scope of the written j examination for an SR Included in this scope are 1 1.) ... describe methods for performing maintence so that he, his crew, and the general public are protecte .) Candidate should be familiar with the concept of I ALARA and be able to demonstrate his knowledge regarding l this concept.
l NUREG-1122. K/A Catalog, lists, under Plant-wide .
I Generics, t 1.) K1.03 Knowlege of 10CFR20 and related facility !
radiation control requirements. --SRO value .) Kl.04 Knowledge of fscility ALARA program. --SRO l value The calculation of a stay time for performing maintence should be an evolution that a supervisor performs not only for himself but for the personnel who work for hi The Radiation Protection personnel on shift should support the supervisor in this functio No change to answer ke Comm2nt on Note: Recommendation noted for future consideratio ADDITIONAL COMMENTS MADE BY GRADER:
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Quoation dumber: 8.04 I Changed answar to require an OR response in order to I
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obtain the requested informatio Quastion Number: 8.09 j Added an additional correct answer; to reflect es rent facility informatio Quontion Number: 8.09 t i
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Changed the second answer; to reflect the common (
nomenclature of the facilit '
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l ATTACHMENT 2 NRC RESOLUTION OF FACILITY COMMENTS FOR THE CALVERT CLIFFS SENIOR REACTOR OPERATOR EXAMINATION ADMINISTERED ON JUNE 22, 1988 Question Number: b 04 Facility Comment: 3) Raising Tavg.
l The analysis of temperature change across the core is
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correct. However, the result would be in an increased
! negative ASI, power moving toward the top, as a result of
! the temperature change. Tc itereasing adds negative l reactivity to the lower half of the cor Th decreasing adds positive reactivity to the top of the cor Therefore power shifts to the top of the core. The more neg4tive MTC in the upper core aids this process causing a larger power ( ASI) swin Reference: Same as state NRC Resolution: /. gree with comen Reference stated supports facility comment and does not fully support answer key. Answer key is changed to reflect fr.cility commen Question Number: 5.07 Facility Comment: The question used units of PCMS in lieu of *. for r(activity. Although this did not confuse the candidate, it was inconsistent with material presented to build the exam. The unit of PCM is not used at Calvert Cliff !
NRC Resolution: Coment Note l References to the unit PCM will be deleted from the questio In addition, units of AMPS are also deleted from the questio Question Number: 5.10 Facility Comment: b . F (255-265) for 1700 psig F (320-380) for 700 psig This is a constant enthalpy expansion to 35 psi Additionally, units of psia & psig should remain consisten Reference: Steam Tables l
l NRC Resolution: Agree with comen Answer key changed to reflect proper answe I l
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Question Number: 5.14 a.
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Facility Comment: 1. Once through cooling is correct. However, the answer key states incorrectly that this method is inadequat The references stated do not apply to this method of cooling and therefore cannot be used to verify the answer. A phone conversation with Dr. William Dove of CE verified this a viable cooling metho The answer should be "yes "
NRC Resolution: Agree with comnen Answer key changed to YE Question Number: 6.04 Facility Comment: "Control Channels" is not standard Calvert Cl*ffs terminolog Therefore, this question could be interpreted as reactor Regulating or the narrow range safety NI channels. If narrow range NIs are assumed, the l
answer would change to any two of the three listed belo l Inhibit SUR trip Enable loss of load trip Enable APD trip Reference: R0-57-1-3 pages 22 & 23 NRC Resolution: Disagree with comm.en Reference stated in factitty comment (RO-57-1-3) was not supplied to the NRC. Reference stated in the examination (R0-57-1-2 page 39) refers to the Power Range Control Channel. Question premise refers to "excore neutron detectors and associated instrumentation." During the examination the candidate asked "What is treant by Control Channels?," he was told they refer to Po.er Range NIs. No change to answer ke )
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Question Number: 7.13 .
Facility Comment: This knowledge is beyond that expected for an SRO at i CCNP Radiation Protectic,n personnel are on shift to !
provide the necessarv support. We request this question
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NRC Resolution: Comment note '
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10 CFR 55.43 (b) (4), "Radiation hazards that may arise ,
during normal and abnormal situations, including l
! maintenance activities and various contamination conditions," is an area of knowledge which will be sampled -
i on an SRO written examination. Examiner Standard ES-402 A 3 j lists items that are to be included in the scope of the written examination for an SRO. Included in this. scope are
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1.) ... describe methods for performing maintenance so that [
1 he, his crew, and the general public are protected. 2.) l I Candidate should be familiar with the concept of Al. ARA and {
l be able to demonstrate his knowledge regarding this l
concep !
- NVREG-1122 K/A Catalog, itsts, under Plant-wide Generics, i j (1) K1.03 Knowledge of 10CFR20 and related facility '
j radiation control requirements. --SRO value ;
j (2) K1.04 Knowledge of facility Al. ARA program. --SRO ,
value l
- The calculation of a stay time for performing maintenance ;
j should be en evolution that a supervisor performs, or j
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reviews, not only for himself but for the personnel who t work for hi !
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j No change to answer key.
.i i Comment on Note: Recommendation noted for future consideration.
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! ADDITIONAL COMMENTS MADE BY GRADER: !
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Question Number: 8.04 l J 1 j Changed answer to require an OR response in order to ;
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obtain the requested informatio l
1 Question Number: 8.09 ;
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I Added an additional correct answer to reflect current facility informatiu1,
Question Number: 8.09 I I
Changed the second answer to reflect the common 1 nomenclature of the facility.
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