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 Co.
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 hkW David M. Silk Date Operations Engineer Examiner
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Y 88 ApprovedBy:/
Peter y Eselgro W / Chief Date 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 Office.
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PDR ADOCK 05000317 Y
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TYPE OF EXAMINATIONS:
Replacement
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EXAMINATION RESULTS:
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jWritten Exam l 1/0 l
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[ Operating Examl Waived l
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l Attachments:
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Written Examination and Answer Key.
2.
NRC RESOLUTION OF FACILITY C0KMENTS FOR THE CALVERT CLIFFS SENIOR REACTOR
OPERATOR EAAMINATION ADMINISTERED ON JUNE 22, 1988, 3.
BALTIMORE GAS AND ELECTRIC COMPANY LETTER DATED JUNE 28, 1988,
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION
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FACILITY:
C&YERT CLIEES
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REACTOR TYPE:
EMB-QE DATE ADMINSTERED:
68/06/22 EXAMINER:
JAGGAB.
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CANDIDATE
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i INSTRUCTIONS TO CANDIDATEL
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Use separate paper for the answers.
Write answers on one sida only.
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Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after.the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF CANDIDATE'S CATEGORY
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i VALUE_ _ TOTAL SCORE VALUE CATEGORY
i 25.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT
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OPERATION, FLUIDS.AND
.I THERMODYNAMICS l
25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION i
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_25.00
_25.00 7.
PROCEDURES - NORt!AL, ABNORMAL,
-EMERGENCY AND RADIOLOGICAL
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CONTE 0'
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.25._00 25.00
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ADMINISTRATIVE PROCEDURES, l
CONDITIONS, AND LIMITATIONS
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Totals I
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j Final Grace i
All work done on this examination is my own.
I have neither given j
nor received aid, i
Candidate's Signature
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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 penaltics.
2.
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 cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.
Use only the paper provided for answers.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
l 8.
Consecutively number each answer sheet, write "End of Category __" as appropriate, start each catego.y on a new page, write only on one side
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of the papsr. and write "Last Pasu" on the last answer sheet.
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9.
Number each answer as to category and number, for example, 1.4, 6,3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Uso abbreviations only if they are commonly used in facility literature.
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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 required.
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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 not.
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15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the exemination are not clear as to intent, ask questions of the examiner only.
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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 top.
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(2)
Exam aids - figures, tables, etc.
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Answer pages including figures which are part of the answer.
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i b.
Turn in your copy of the examination and all pages used to answer the examination questions.
I c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
i d.
Leave the examination area, as defined by the examiner.
If after I
leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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pilEQBY OF NUCLEAR POWEILELANT OPERATION, Pcgo
FLUIDS.AND THERMODYN6MICS
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QUESTION 5.01 (1.50)
a.
Why is the Shutdown Margin requirement greater in modss 1-4 than in mode 57 (1,0)
b.
With the plant operating at 85% power and all systems in a normal configuration, the operator borates 100 PPM. Shutdown Margin will...
1. Increase.
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2. Decrease.
3. Remains unchanged.
(0,5)
QUESTION 5.02 (2.00)
a.
How will individual CEA rod worth change as moderator temperature I
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is increased?
(0,5)
b.
If the worth of an individual CEA rod is measured at a flux level
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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
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relative flux distribution?
Explain.
(1.0)
c.
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 explain HOW and WHY each of the following plant parameters will be
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affected over the next 5 hoars.
Assume all systems are in automatic, rod control is in manual sequential, and no operator action is taken.
a.
RCS temperature
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RCS pressure i
c.
Turbine Generator control valve position l
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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THEQBY_OF NUCLEAB.EQWEB_ELABI-OPERATIOH2 Pogo
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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 Tref.
EXPLAIN HOW and WHY reducing the power to 95% would affect ASI for each of the following means of reducing power.
Consider each separately.
1.
Control rod insertion.
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2.
Boron addition.
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3.
Raising Tavg.
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QUESTION 5.05 (2.00)
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a.
If a trip from 100% power occurs with Xenon at equilibrium, what is the approximate time interval after the trip that Xenon l
will again be at the 100% equilibrium value?
(0.5)
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How would this approximate time interval compare if the trip
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(0.5)
r When a reactor is returned to 100% power from peak Xenon i
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conditions, why does the Xenon reactivity value trend become less j
than the 100% equilibrium value for a short period of tirae?
(1,0)
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CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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THEQRY OF NUCLEAR POWEB_ELABI_QEEBATIQlb.
Pcga
Ekt)ll!S. AND THERMODYE&dLQE
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QUESTION 5.06 (2.00)
TRUE or FALSE 7 a.
Critical rod height does NOT depend on how fast control rods are withdrawn, b.
Critical rod height dictates the reactor power level when criticality is achieved.
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c.
The SLOWER the approach to criticality, the LOWER the reactor power level will be when reaching criticality.
d.
When reactor power is in the SOURCE range, changes in power level do NOT cause a change in Tavg.
QUESTION 5.07 (2.00)
l Unit i has just restarted following a refueling outage while Unit 2 is near EOC.
Annwer the following regarding the differences in plant response between the two units (explain your answers):
a.
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?
b.
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, a.
b.
Boron concentration decreases 100 ppm while core is at MOC.
c.
Increased nur.ber of burnable poisons are inserted into the core.
(***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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THEORY OF NUCLEAR POWER PLANT OPEBAI1QH2 Pcgo
ELyfDS.AND THEBdQDXHMISS
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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 constant.
a.
Primary coolant temperature decreases.
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b.
Primary coolant pressure decreases.
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c.
Primary coolant flow decreases.
d.
Reactor power decreases.
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I Stable natural circulation conditions exist within the RCS with the t
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following parameters.
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Thot - Teold = 25 F f
i SG pressure 885 psis
Thot Subcooled Margin indicates 40 F subcooled
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Determine RCS pressure.
(Show steps used to arrive at your answer).
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What will be the temperature of the PORV discharge line if a PORY
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pressure is 20 psig, and pressurizer pressure ist j
1.
1700 psig?
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700 psig?
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CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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E%_IBEQBY OF NUGLE68_EQWE[LEL6HT OPEB611Qth.
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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 F.
Explain your answers i
and consider each failure independently,
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The steam dump valve which is throttled to control cooldown rate
fails open.
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b.
Level is lost in the pressuriser.
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c.
The auxiliary feedwater valve to one of the SG's falls shut.
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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
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BRIEFLY EXPLAIN.
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Power decreases with constant Tave.
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b. Core age increasen with constant power.
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c. Pressurizer pressure increases with constant power.
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THEORY OF_NMELEAR POMEB PLANT OEERATION.
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QUESTION 5.14 (2.50)
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The following questions assume that a spurious reactor trip has just occured from extended 100% power operation (an inadvertent trip due to
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instrument malfunction) and that a pressurizer safety valve lifte and l
sticks open.
Reactor Coolant Punpa are tripped according to procedure i
{j and RCS pressure drops to 1000 psia.
Answer each part independently.
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If natural circulation flow could NOT be attained, what means I
of core cooling exists and will it be sufficient to cool the
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core?
(1.0)
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b.
CETs are relatively stable (not increasing) and read 585 F, and
a small constant feeding and steaming rate is occuring in the
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In what regions of the RCS would the following l
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1.)
Boiling.
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Superheating.
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Condensation.
(1.0)
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Describe how auxiliary spray flow could be used to determine
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or confirm the presence of RCS voiding.
(0,5)
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PLANT SYSTEMS DESIGN. CONTBQL. AND INSTRUMENTATION Pego
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QUESTION 6.01 (1.75)
I a.
State TWO conditions that will cause ALL charging pumps (except those in PULL-TO-LOCK) to start automatically.
(1.0)
b.
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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CATEGORY 6 CONTINUED ON NEXT PAGE, *****)
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6.
PLAyT SYEIEdR_ DESIGN COHIBQL AND INSTRQUENTATION Pago
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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 triggered.
Consider each condition separately.
Each condition may have for its answer none, one, or several actuation signal (s).
1.
SIAS 2.
CSAS 3.
CIS 4.
SGIS 5.
RAS 6.
AFAS a.
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 b.
A small-break LOCA has occurred and i
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
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c.
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 (***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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$2 PLANT SYSTEdg_DSSIGN. CONTBOL. AND INSTRUMENTATION Pego 10 r
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QUESTION 6.03 (2.80)
The plant is at 100% power and all controls systems are in automatic.
The controlling pressurizer level channel fails low.
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 1.
a.
LIST THREE control / protective functions of the wide range channels.
(0.75)
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b.
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, a.
Liquid radwaste discharge monitor.
(0.5)
b.
Steam generator blowdown tank radiation monitor.
(0,5)
c.
Blowdown Recovery radiation monitor.
(1.5)
(*****
CATEGORY 6 CONTINUED ON NEXT PAGE *****)
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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, a.
SUR channels A and B fail HIGH during a reactor startup when the reactor is critical at 10E-6% power, b.
SG#11 level channel A fails LOW followed by SG#12 level channel B failing HIGH while at 80% power.
c.
Loop #11 To channel A fails HIGH followed by loop #12 Th channel B failing HIGH while at 100% power, d.
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),
a.
STATE the pressure setpoint that interlocks both SDC suction header isolation valves preventing them from operating.
(0.2)
b.
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, a.
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 bar.
b.
The "Bresker Improper Lineup" alarm is actuated when a third unit disconnect switch is closed with the associated 480 VAC circuit breaker racked out.
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CATEGORY 6 CONTINUED ON NEXT PAGE *****)
6.
PLANT SYSTEME_QESIGN, CONTROL. AND IEEIBUMENTATION Pcgo 12 f
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l QUESTION 6.09 (2.50)
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a.
State FOUR regulating control rod interlocks / limits that are
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in effect when the system is in MANUAL INDIVIDUAL control?
(1.2)
b.
WHAT are the TWO instrumentation signals / conditions that could provide a DROPPED ROD annunciator?
(0.8)
c.
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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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, a.
Teold Loop 1 fails LOW
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A RCP trips
c.
RCS pressure increases 25 psig
d.
A Linear Power Range Channel (Safety) fails high i
e.
Delta-T PWR Calibrate pot is reduced l
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l QUESTION 6.11 (1.75)
a.
Assume that an inadvertant trip of the AFAS "A" channel l
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j is active for a period of only 15 seconds and then clears.
Briefly describe how the AFAS signal is reset and the associated I
equipment returned to a normal configuration.
(1.5)
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How much time must have elapsed before the AFAS signal will i
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(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 / systems.
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 1.
Main Feedwater Regulating Valves 2.
Pressurizer spray valves 3.
Letdown 4.
Atmospheric Dump Valves 5.
AFW regulating valves 6.
EDG service water supply valves 7.
Auxillary Spray valve 8.
Turbine AFW pump speed (if operating)
(2.0)
b.
Describe TWO means of interconnecting the IA system with backup sources of air pressure. Indicate automatic setpoints, if any.
Answer this question independently of part "a." above.
(1.0)
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(***** END OF CATEGORY 6 *****)
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t 7. '__P.ROCEDURES - HOBHALt_A38QBti&L. EMEBQEHQX Pess 14 (
AHE.B&M!'hQQ1GALC.QHIRQL
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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
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used on the Control Element Drive System?
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Withdrawing Shutdown Group CEA's during a reactor startup, i
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Inserting Regulating Group CEA's during reactor shutdown, i
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Recovering a droppped CEA.
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QUESTION 7.02 (2.00)
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j The following questions pertain to information found in EOP-6, Steam (
j Generatur Tube Rupture.
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a.
State FOUR of tte FIVE trends that can be used to identify the i
affected Steam Generator.
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List the FOUR RMS monitors that can be used to verify that a
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Steam Generator Tube Rupture has occurred, i
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QUESTION 7.03 (2.50)
a The following questions pertain to information found in OP-5, Plant I
Shutdown from HSB to Cold Shutdown.
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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
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(1.5)
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b.
What is the minimum pressure for the RCS prior to blocking SIAS?
j (0,5)
8 c.
What is the minimum allowable pressure in the Steam Generator j
Prior to blocking SGIS?
(0.5)
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PRQQEDMBES_._J_QRMAL. ABHQBMAL, EMEBQENCY Pego 15
~ AND*BADIOLOGICAL._EQHIBQL
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QUESTION 7.04 (2.50)
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According to EOP-5, Loss of Coolant Accident:
a.
State FOUR of the SIX Entry Conditions.
(1.0)
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b.
State THREE of the FOUR positive indications of Core and/or RCS l
voiding.
(0.75)
j c.
What are THREE of the FOUR conditions / parameters the operator
{
must confirm / satisfy prior to stopping or throttling a safety injection train?
(0.75)
r I
,
QUESTION 7.05 (2.00)
a.
Given a LOCA condition, how soon after SIAS a=>tuation must
'
"core flush" commence?
(0.25)
,
i b.
State the TWO methods that are available for "core flush".
,
Include a brief description of flow paths for each method.
(1.5)
l c.
What is the reason that "core flush" is established?
(0.25)
QUESTION 7.06 (2.00)
,
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
!
conditions.
Consider each situation separately.
t
.
I a.
Small break LOCA with RCS pressure decreased to 1775 psia.
'
I b.
Small break LOCA with RCS pressure decreased to 1800 psia and a
i CIS actuated.
l c.
Total Loss of Feedwater event with RCS pressure decreased to
,
1800 psia.
,
d.
Large break LOCA with RCS pressure decreased to 1400 psia.
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PBQQEDMBES - NORMAL 2_A.SEQBdAL. EMEBGEEDY Pego 16 AND BADIOLQ919AL_90BIBQL
.
.
QUESTION 7.07 (1.50)
The following pertain to information found in the Calvert Cliffs Technical Specifications.
a.
State ALL dose rates that apply to designating an area as a High Radiation Area?
(0,5)
'
b.
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 control.
If CEAs cannot be inserted or drivon in, reactor shutdown may be accomplished by boration, a.
List the TWO flow paths for charging water to the RCS.
b.
List the TWO flow paths for supplying boric acid to the charging pumps.
!
QUESTION 7.09 (2.50)
The following pertain to information contained in EOP-4, Excess Steam Demand.
a.
What THREE parameters are used to identify the affected S/G7 (0.75)
b.
state SIX of the EIGHT actions taken to isolate the S/G's.
(1.5)
c.
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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CATEGORY 7 CONTINUED ON NEXT PAGE *****)
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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:
-
!
a.
State FIVE of the SEVEN indicationu/ alarms of a "Dilution While
!
Critical".
(1.0)
,
!
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!
b.
TRUE or FALSE?
Boration and CEA movement may be used CONCURRENTLY to maintain reactor power level constant.
(0.5)
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c.
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 QUESTION 7.11 (1.50)
l
!
I j
The following pertain to information found in AOP-1B, CEA Malfunctions.
i a.
How is programmed To maintained following a dropped CEA?
I j
b.
Based on Technical Specifications Fig. 3.1-3, Allowable Time to l
Realign CEA's.....
what is the MAXIMUM allowable time to realign a dropped CEA?
i c.
During CEA reclignment, what method is used to ecmpensate for
reactivity changes?
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EBQQEQQBES
__ M Lt_AESQBtie k EtfEBQEtiC1 Pogo 18
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AEQ_B&Q1QLQQICAL COSIBQL
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^
QUESTION 7.12 (2.00)
,
t j
The following pertain to information found in OP-1A.
l
,
a.
State the limit for each of the listed parameters where the
'
Reactor Coolant Pump (RCP) must be stopped.
<
--Lower sen.1 temperature.
I
"
--Motor thrust bearing temperature.
(0.8)
.
b.
What is the maximum time a RCP be run following a loss of CCW? (0.4)
.
,
c.
Given the following RCP seal cavity pressures, state which seal l
(UPPER. MIDDLE, or LOWER) has probably failed.
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)
[
I f
f f
!
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
!
100 MRAD /HR fast neutron.
You will be accompanying the maintenance
!
q man into the work area. Assums you are 27 years old and have a
!
<
lifetime exposure through last quarter of 29 REM on your NRC Form 4.
I i
So f ar this quarter, you have accumulated 1.0 REN.
i l
I How long nay you stay in this area without exceeding the Calvert
,
)
Citffs administrative limits?
Show all assumptions and caletintions.
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8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Paga 19
' AHD' LIMITAflQtLE
_
.
.
QUESTION 8.01 (2.50)
Explain why the following leakages are IN or OUT of compliance a.
with the Limiting Condition-for Operation.
Consider each INDIVIDUALLY and SEPARATELY.
1.
Unidentified-----------------------------------0.65 GPM 2.
Steam Generator Tube Leakage:
A----------------------------------------------0.45 GPM B----------
0.50 GPM 3.
Variou.
anual vent and drain valves seat and pack., gland leakage.----------------------4.6 GPM 4.
Back leakage through SI check valves detected
'
by leak check on previous shift.----------------3.9 GPM (1.0)
b.
Evaluate Technical Specification compliance if the leakage in
"a." is cumulative. (0.75)
Calvert Cliffs Technical Specifications require that systems be c.
operable for detecting leakage from the RCS.
List the 3 systems or methods used to detect leakage to the RCS.
(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-101J.
a.
Emergency Operating Procedures.
b.
Operating Instructions.
c.
Nuclear Engineering Operating Guids.
(*****
CATEGORY 8 CONTINUED ON NEXT PAGE *****)
8.
ADMINISTRATIVE PROCEDUBES. CONDITIONS.
Paga 20 AND' LIMITATIONS
.
.
QUESTION 8.03 (3.00)
a.
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 Limit.
(1.5)
b.
What is the objective or basis for the Technical Specification Reactor Core Safety Limit?
(0.5)
c.
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 exceeded.
(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 F.
In accordance with the Unit 1 Technical Specifications:
a.
What actions must be taken?
b.
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 1.
a.
What actions must be taken if operating with a minimum crew composition and one of the Reactor Operators becomes
incapacitated?
(1.5)
b.
Is the following action proper by the Shif t Supervinor?
Justify your answer.
Fifteen minutes before scheduled arrival of the on-coming shift, one of the four on-coming RO's calls in sick.
The on-shift SS decidos, due to the shift's present overtime condition, to i
call in a replacement.
The SS also decides that since the replacement should arrive shortly after shift change (about 30
)
minutes) to send his people home and let the next shift start with three R0s.
(1.0)
(*****
CATEGORY 8 CONTINUED ON NEXT PAGE *****)
8.
ADMINISTRATIVE PROCEDURES. CONDITIONS, Pcgo.21 AND' LIMITATIONS
.
.
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 4.
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 head.
After the pool has been drained, 11 LPSI pump develops a breaker problem and must be removed from service.
a.
What actions must be taken AND in what time period according to Technical Specification 3.9.8.27 b.
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. F.
.a.iple component tagout, when are the control board s
v cv be tagged /untagged?
(*****
CATEGORY 8 CONTINUED ON NEXT PAGE *****)
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8.
ADMJHISTRATIVE PROCEDUBES. CONDITIQNS._
Paas 22 AND LIMITATIONS
.
.
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 EAL.
State any assumptions used in making a classification.
a.
While operating at 100% power, steady state, a bomb threat is received stating that an explosive device has been placed in the control room.
Upon conducting a search of the control room, what appears to be such a deci.ce le discovered.
Examination of the device indicates that it may be triggered by radio control or movement.
(0,5)
b.
While moving fuel in the core, during refueling outage, all off-site power is lost due to line damage from high winds.
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 operation.
(0,5)
c.
While operating at 100% power, steaoy state, a double ended rupture of one RCS cold leg occurs.
(0,5)
d.
While operating at 100% power, steady state, a main steam line break occurs downstream of containment but upstream of the main steam isolation valve.
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 loop.
(0.5)
.
.
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ADMINISTRATIVE PROCEDURES. CONDITIONS.
Pago 23 AND'LIMITATIOND
.
.
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-118-5.
a. During instrument testing while at power, three pressurizer pressure safety channels are momentarily place in bypass.
b. While critical at 1% power on Unit 2, Tave drops below 515 F and then returns to normal.
c. Refueling water tsnk level falls below 400,000 gallons and cannot be restored.
I d. During surveillance testing an expected actuation of LPIS train A occurs.
(3.0)
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(********** END OF EXAMINATION **********)
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5.
THEORY OF NUCLEAR POWEB PLANT OPERATION._
Paga 24 FLUIDS.AND THERMODYNAMICS
-
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.
ANSWER 5.01 (1.50)
a.
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 1-4.
(1,0)
b.
1 (Increase)
(0.5)
REFERENCE R0-302-3-1; Calvert Cliffs: T. 5.
p 1-3
'
EO 8.5 3.5/3.9 004000K519 19*.004K107 191004K112
..(KA's)
ANSWER 5.02 (2.00)
a.
Worth will increase (0,5) with an increase in moderator temperature.
b.
No change. [0.5)
The absolute value of neutron flux will not change the worth.
A shift in flux distribution is required. [0.5)
(1.0)
'
c.
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, 4.9
'
2.9/3.4 2 5/2.8 192005K107 001000K502 193003K125
..(KA's)
ANSWER 5.03 (1.50)
<
a, Decreases (~15 F) [0.25] due to buildup of Xe (0.25)
b.
Held constant (0.25] by PPCS spray and heaters (0.25)
c.
Increases (0.25] due to lower S/G pressure (0.25)
(*****
CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION.
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 3.5/3.6 3.2/3.5 3.6/4.1 3.6/3.6
'
039000K508 002000K510 001000K533 192008K124 193008K105
..(KA's)
ANSWER 5.04 (1.50)
1.
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).
2.
A change in Th as power decreases is greater than the change in Tc.
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 core.
ASI becomes more negative.
3.
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 core.
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 not.
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, 2.3 3.3/3.7 2.8/3.2 192005K111 015020K503 103008K122
..(KA's)
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5.
THE0BY OF NUCLEAR POWEB PLANT OPERATIQH2 Pcgs 26
,
' FLUIDS.AND THERMODYEAMISS
-
.
.
ANSWER 5.05 (2.00)
a.
~ 24 Hrs. (accept 20 - 30 hrs.)
(0.5)
b. Time would be shorter.
(< 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.4.a RO-302-2-1 pp. 48,49 E0 12.6 3.1/3.1 3.4/3.4 3.4/3.4 192006K107 192006K106 192006K105
..(KA's)
ANSWER 5.06 (2.00)
a.
T b.
F c.
F d.
T
[0.5 ea.]
REFERENCE RO-302-3-1 p.
EO 7.9 3.8/3.9 192008K105
..(KA's)
ANSWER 5.07 (2.00)
'
a.
Unit 2 (0.5] due to a lower Beta effective coefficient at EOC (0.5]
b.
Unit 1 (0.5) due to MTC being less negative, (so Tavg must decrease more to add + reactivity). [0.5)
i (***** CATEGORY 5 CONTINUED ON NEXT PAGE *****)
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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 3.8/3.9 2.9/3.4 3.9/4.1 001000K510 001000K549 192008K120
..(KA's)
ANSWER 5.08 (1.50)
a.
Less b.
More c.
More (If assume same Boron conc.,will accept no effect)
[0.5 ea)
l REFERENCE RO -302-2-1 pp. 20-26 EO L.7 3.1/3.1 3.3/3.6 001000K526 192004K106
..(KA's)
ANSWER 5.09 (2.00)
a.
Increase b.
Decrease c.
Decrease d.
Increase (0.5 ea.]
REFERENCE RO-301-13-0 pp. 26-28
.
EO 6.3
'
3.4/3.6 i
i 193008K105
..(KA's)
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5.
THEGRY OF NUCLEAR POWER PLANT OPERATION, Pass 28 l
FLUlDS.AND THEBMODYNAMICE i
.
.
ANSWER 5.10 (2.00)
Teold = 503 F corresponding to saturation temperature for 700 psia c
a.
[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 b.
1.
-225 F (320 - GSO)
2.
d ie-F ( 40G - 430 )
[0.5 ea.]
Jto J2e aro REFERENCE Steam Tables R0-301-8-0 p.4 TO 8. P.
3.3/S.4 2.8/3.1 3.6/3.8 193003K115 193003K124 193003K125
..(KA's)
ANSWER 5.11 (1.50)
a.
Incr.se cooldown rate (0.2) since more energy is being removed frorc he primary. (0.3]
,
b.
May interrupt natural circulation (0.2] since hot legs maybe voided.
[0.3] OR No effect (0.2] if hot legs do not become voided (0.3)
c.
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)
EO 2.3 3.9/4.1 193008K123
..(KA's)
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5..
THEQRY OF NUCLEAR POWER PLANT OPERATION.
Pcga 29 FLUIDS AND THERMODYM&tiLGE
.
.
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 p.
EO 2.6 3.9/4.2 193008K121
..(KA's)
ANSWER 5.13 (1.50)
a.
Decrease (0.25), smaller delta T required to transfer more energy from RCS (0.25).
b.
Decrease (0.25), fuel swelling and clad creep reduce clad gap which reduces delta T across gap and lowers center line temp (0.25).
c.
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 boiling.
,
REFERENCE
,
'
RO-301-13-0 p.
EO 13.4, 13.3 RO-301-12-0 pp. 6,7 2.2*/2.4*
2.3*/2.4 2.5/2.5 193007K101 193008K130 193008K128
..(KA's)
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5.
THEORY OF NUCLEAR POWER _ PLANT OPEBATION.
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; 14.2.7 4.0/4.6 4.0/4.4 4.5/4.9 000074K103 000074K311 000074A206
..(KA's)
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6.
PLA,NT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION Paga 31
'
.
f ANSWER 6.01 (1.75)
!-
a.
The pumps will automatically start upon:
1.
Pressurizer level deviation = -15' inches 2.
Receipt of an SIAS (0.5 ea.)
b.
1.
Degassification 2.
Filtration 3.
Demineralization (De-ionization)
[0.25 ea.]
(0.75)
REFERENCE
'
SD 6, Pp. 29, 20-22 Lesson Objectives 00603K505 00603K601
00604K501 2.9/3.8 3.0/3.9 3.6/4.0 3.8/4.0 004000K115 004000K101 004010A202 004000A101
..(KA's)
ANSWER 6.02 (1.75)
,
a.
SIAS, CSAS, CIS, SGIS (0.25 ea.)
,
,
b.
CIS, SGIS (0.25 ea.)
'
c.
None (0.25]
REFERENCE SD-63 pp. 45-60 R0-34-1-1 pg. 7
'
Lesson Objectives - unavailable
,
3.8/3.9 4.2/4.4 013000K101 012000A301
..(KA's)
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PL4EI_SYEIEMS-DESIGH CONTROL. AND INSTRUMENTATION Pega 32
.
.
.
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.
SD-62 Fig. 62-8, Fig. 62-11
,
Lesson Objectives 06202K404
06202K405 3.6/3.9 3.1/3.8 3.7/4.0 3.8/3.9 3.2/3.4 3.3/3.7 011000K401 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
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4.1/4.2 3.7/3.9 3.1/3.5*
015000A202 015000A302 015000K101
..(KA's)
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PLANT SYSTEMS DESIGN, CONTEQLt_AEQ_lEETRUMEEIAllgM Pega 33
.
.
ANSWER 6.05 (2.50)
a.
The liquid discharge isolation valves are closed.
(0.5]
b.
Surface and bottom blowdown isolation valves are closed.
(0.5)
c.
Diverts blowdown recovery flow to the HWS.
Closes blowdown isolation valve to Cire. water.
Closes blowdown isolation valve to the condenser.
(0.5 ea)
.
REFERENCE
-
SD-14B p.
SD-18 pp. 17, 27
RO-122-1-1
'
EO 1.3.1
.
4.0/4.3 073000K401
..(KA's)
ANSWER 6.06 (2.00)
!
a.
No trip (0.25]
'
The trip will not occur until power reaches 10E-4% power (0.25)
b.
No trip (0.25]
Only the auctioneered LOW signal is selected, therefore only i
channel A will trip (0.25)
c.
TRIP (0.25)
Due to Delta T decreasing and causing setpoint to change (0.25)
'
d.
TRIP (0.25)
Both APD channels will trip (0.25)
'
REFERENCE i
SD No 59, pgs 10-31 Lesson Objectives 05904K501 3.9/4.3 3.1/3.5 a
012000K603 012000K403
..(KA's)
J
!
i'
(*****
CATEGORY 6 CONTINUED ON NEXT PAGE *****)
i I
't
-
.
.
.
6.
PLANT SYSTEMS DESIGN. qQNTROL. AND INSTRUMENTATION Paga 34
.
.
ANSWER 6.07 (1.70)
a.
>/= 300 psia.
(0.2)
b.
2485 psig - Protect piping from overpressure due to a sudden temperature increase in containment.
(0.75)
315 psig - Protect piping from overpressure due to simultaneous operation of charging pumps and SDC with the pressurizer solid.
[0.73)
REFERENCE SD-7 pp. 42,43,47 Learning Objectives - not available 3.2/3.5*
3.2/3.5 3.8/4.1 2.9/3.1 005000K104 006000K409 005000K407 005000K402
..(KA's)
ANSWEP, 6.08 (1.00)
a.
true (0.5)
b.
true (0.5)
REFERENCE SD-53 pp. 19, 44 3.2/3.3
'
'
062000G009
..(KA's)
i ANSWER 6.09 (2.50)
'
'
a.
Upper electrical limit
'
Lower electrical limit
>
'
CWP CEA withdrawal prohibit CMI CEA motion inhibit (0.3 pts each)
b.
1.
Rod drop from Reed switch 2.
NI negative rate of power change from NI system.
[0.4 pts each)
c.
UV trip devives (0,5)
'
(***** CATEGORY 6 CONTINUED ON NEXT PAGE *****)
.
.
.
.
-
-
-
-
-
6.
PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUBENTATIQH Paga-35
.
REFERENCE
.
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 3.4/3.6 4.5/4.4 3.7/3.8 2.4/2.8 2.9/3.1 063000K201 001000X604 001000K407 001000K105 001000K103
..(KA's)
ANSWER 6.10 (2.50)
a.
No effect b.
Decrease (No Effect if only considering the change in flow)
c.
No effect d.
Increase e.
No effect
[0.5 pts each]
'
REFERENCE SD-59 p.
3.9/4.3 3.3/3.8 2.9/2.9 012000K611 012000K501 012000K402
..(KA's)
ANSWER 6.11 (1.75)
a.
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)
b.
The signal vill seal-in after 18 seconds.
[0.25]
,
REFERENCE RO-34-1-1 pp. 1-9 Enabling Objective 1.02 4.1/4.4 3.7*/4.0 4.3*/4.5*
013000K404 013000A206 013000K107
..(KA's)
(*****
CATEGORY 6 CONTINUED ON NEXT PAGE *****)
_
J
.
E.
PLANT SYSTEMS DESIGN.-CONTBQL AND INSTRUMENTATION Peca 36
.
.
ANSWER 6.12 (3.00)
a.
1.
C 2.
D 3.
D (0.25 each]
4.
D 5.
D 6.
A 7.
D 8.
A b.
1.
Auto valve to plant air system X-ties PA to IA at 85
,
psig IA pressure (0.5]
'
2.
manual X-tie valve to Saltwater system air compressors.
(0.5]
3.
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 3.2/3.5 078000K402 078000K302
..(KA's)
,
i f
l (***** END OF CATEGORY 6 *****)
i
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-.
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- - -
-
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-
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- -. -
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7.
PRO,CEDURES - NORMAL. ABNORMAL..EMEBGENCY Pega 37 AND RADIOLOGICAL _CQN.LTROL
-
,
.
ANSWER 7.01 (1.50)
a.
Manual Group b. Manual Sequential c. Manual Individual
[0.5 ea.]
AOP-1B 4.0/3.7 3.5/3.4 00000SG006 001000A403
..(KA's)
-
ANSWER 7.02 (2.00)
a.
1.
Mismatch in feed flow prior to the trip.
2.
Unexplained increase in S/G level prior to the trip.
3.
4.
Post-trip S/G level changes.
5.
S/G samples.
[any 4 @ 0.25 ea.]
b.
1.
Condenser off-gas RMS.
2.
S/G Blowdown RMS.
,
3.
MainSteam Line RMS.
4.
Main Vent RMS.
[4 @ 0.25 ea.)
REFERENCE EOP-6 pp. 3,
'
4.4/4.6 4.5/4.8 3.9/4.2*
000038A204 000038A202 000038A203
..(KA's)
I i
,
,
(***** CATEGORY 7 CONTINUED ON NEXT FAGE *****)
'
,
-
- - -
-
-
-
- -
-
-
-
-
-
-
-
- - -
- -
-
.
7.
PROCEDURES - NORMAL. A]3E9EMAL. EMEEGEgg Paga 38 AND RADIOLOGICAL CONTBQL
.
ANSWER 7.03 (2.50)
a.
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 limit.
(1.0)
b.
1760 psia.
(0.5)
c.
720 psia.
[0,5)
REFERENCE OP-5 pp.
1,
3.4/3.9 002000SG10
..(KA's)
!
ANSWER 7.04 (2.50)
a.
1.
Unexplained decreasing presurizer level.
2.
Unexplained decreasing pressurizer pressure.
3.
Loss of RCS subcooled margin.
4.
High Containment radiation alarm.
5.
Increase in containment sump level.
6.
Increase in containment sump alarm frequency.
[any 4 at 0.25 ea.)
b.
1.
Letdown flow greater than charging flow.
2.
Rapid increase in pressuriser level during an RCS pressure reduction.
3.
Loss of subcooled margin as determined using CET temperatures.
4.
"REACTOR VESSEL WATER LEVEL LOW" alarm.
(any 3 at 0.25 ea.)
i c.
1.
RCS subcooling > or = 30 F.
2.
PZR level > 155 in., and etable.
3.
At least i S/G available for heat removal 4.
RVLMS indicates core covered.
(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*/3.5 000009A104 000011SG11 000009K310 000011A211
..(KA's)
I i
l (*****
CATEGORY 7 CONTINUED ON NEXT PAGE *****)
f
- -
-
- -
-
-
-
-
- -
. - _ -
7.
PROCEDUBES - NORMALu_AREQBtiAL' EMERGEdgl Paga 39
.
AND RADIOLOGICAL CONTRQh
.
.
ANSWER.
7.05 (2.00)
a.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(0.25)
b.
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)
c.
Prevent restriction of core flow due to boron crystallization. (0.25)
REFERENCE SD-7 pp.67, 68; EOP-5 p.
LO 201EK312 3.8/4.2 000011K313
..(KA's)
ANSWER 7.06 (2.00)
a.
None b.
All (4)
c.
All (4)
d.
Two (0.5 ea.)
REFERENCE
,
EOP-0 p.
LO201EK306 3.4/3.7 4.1/4.2 000011K314 000009K313
..(KA's)
ANSLfER 7.07 (1.50)
a.
>100 ar/hr <1000 mr/hr >1000 mr/hr (0.25 ea)
l
!
b.
SWP issued.
Radiation monitoring device indicating dose rate provided.
Louked barricades with keys maintained by Shift Supervisor or Supervisor-Radiation Control.
(0.33 ea)
t (***** CATEGORY 7 CONTINUED ON NEXT PAGE *****)
f
7.
PROCEDURES - NORMAL. ABNORMAL ~. EMERGEHQY Pcco 40 AND RADIOLOGICAL CONIB9L
,
REFERENCE CC Tech. Specs pp. 6-20, 6-21 2.8/3.4 194001K103
..(KA's)
ANSWER 7.08 (2.00)
a.
1.
Normal CVCS lineup.
2.
Charging thru the Aux. HPSI header.
[0.5 ea]
b.
1.
BAST-- via gravity feed and boric acid pumps.
2.
RWT-- to charging pump suction.
[0.5 ea.]
REFERENCE EOD-8 pp. 6-7 4.2/4.3 3.4/4.7 3.4/4.7 004000K123 004000K122 000029K311
..(KA's)
i ANSWER 7.09 (2.50)
a.
S/G pressures
T cold S/G Levels.
[0.25 ea.]
b.
1.
Shut MSIV's.
2.
Shut Feed Water isolation valves.
3.
Shut MSIV bypass valve's, 4.
Shut Blowdown valves.
5.
Shut AFW Steam supply valves.
G.
Shut AFW block valves.
7.
Shut Atmospheric dump valves.
8.
Shut upstream drains.
[any 6 @ 0.25 each)
c.
The cooldown adds positive reactivity and the boron adddition prevents a return to criticality.
[0.25)
(*****
CATEGORY 7 CONTINUED ON NEXT PAGE *****)
7..
PROCEDURES - NORMAL. ABNORMALJMERGEILQY Pago 41 AND' RADIOLOGICAL CONTROL
.
REFERENCE EOP-4 pp.5, 6, 9 4.5/4.7 4.3/4.3 4.1/4.4 000040K105 000040A104 000040K304
..(KA's)
ANSWER 7.10 (2.00)
a.
1.
Increasing power level on Nuclear Instruments.
2.
Increasing Tc.
3.
Decreasing boron concentration indication.
4.
Demin, water flow alarm.
5.
BA flow alarm.
6.
Low BA pump discharge pressure alarm.
7.
Automatic withdrawal prohibit alarm.
[5 at 0.2 ea.)
b.
TRUE (0.5)
c.
Borate in 5 sec. increments.
(0,5)
REFERENCE AOP-1A p.
LO 00603003; 00603K401 3.8/3.9 3.8/4.3 004020A206 004000SG15
..(KA's)
ANSWER 7.11 (1.50)
a.
With turbine load.
b.
60 min, c.
Regulating CEA motion.
(0.5ea.]
REFERENCE AOP-1B pp. 5-6 T.S. Fig. 3.1-3 3.8/4.1 000003K304
..(KA's)
(*****
CATEGORY 7 CONTINUED ON NEXT PAGE *****)
7.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY P2ga 42 AHQ RADIOLOGIQAL CONTROL
_
.
.
ANSWEk 7.12 (2.00)
a.
200 F 195 F
[0.4 ea)
b.
10 min.
f.0.4)
c.
1.)
UPPER 2.)
LOWER (0.4 ea.]
REFERENCE RO-106-1-1 pp. 15-17
OP-1A p.
LO 00501K704
'~
00503K702 3.5/3.9 3.7/3.9 003000A202 003000A201
..(KA's)
i l
AN.9WER 7.13 (1.00)
?
.
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
'
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)
'
I 1.0 R/1.1 REM /HR = 0.91 HRS = 53 MIN CO.25)
I REFERENCE l
CCI-800B p.
3.1/3.4 2. 5 '3. 4-1.122 r
,
'
!
194001K103 194001K106f
..(KA's)
'
i
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(***** END OF CATEGORY 7 *****)
,
-
8.
ADMINISTRATIVE PROCEDUBES. CONDITIONS.
Pago 43 AND LIMITATIONS
-
.
ANSWER 8.01 (2.50)
a.
1.
In compliance, unidentified limit is 1 GPM.
(0.25)
2.
In compliance, Total leakage less than 1.0 GPM
[0.25]
3.
In compliance, identified leakage spec. is 10 GPM
[0.25]
4.
In compliance, identified leakage spec, is 10 GPM
[0.25]
b.
Not in compliance (0.5) as total leakage is 10.1 GPM. [0.25)
(0.75)
c.
1.
Containment atmosphere pariculate.
2.
Containment sump level alarm.
3.
Containment atmosphere gaseous.
[0.25 ea.]
REFERENCE j
CC TS 3/4.4.6 2.6/3.8 3.6/3.8 002020K401 002000SG06
..(KA's)
i
,
ANSWER 8.02 (1.50)
t
a.
Two SRO license holders, one of whom must be the Shift Supervisor or Supervisor-Procedural Developement.
b.
Same as a.
above.
c.
Two members of the plant management staff, one of which holds an SRO license on the affected unit.
(0.5 ea.)
,
I REFERENCE CCI-101J pg. 4 3.3/3.4 i
194001A101
..(KA's)
.,
J A
J (*****
CATEGORY 8 CONTINUED ON NEXT PAGE *****)
-
-
-
. - _ -
8.
ADMINISTRATIVE PROCEDURES. CONDITIONS.
Pcgo 44 AMD LIMITATIONS
.
.
ANSWER 8.03 (3.00)
a.
Thermal Power, Pressuriser Pressure, highest loop Tc temperature.
(0.5 ea.)
b.
Maintain integrity of fuel cladding
--OR--
'
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 c.
Be in HOT STANDBY; notify NRC Operations Center.
(0.6 ea.)
i
'
REFERENCE l
CC Technical Specifications pp.
2-1, B2-1, 6-13 3.6/4.1 2.6/3.8 002020G006 002020G005
..(KA's)
t ANSWER 8.04 (2.00)
e.C ca-1/' Restore Tavg to > 515F (Ar25) within 15 minutes (G<2$) C R
,
a.
,
jk."' 0F be in Hot Standby (0,% ) within the next 15 minutes (9 25)
2S o
c.s b.
1.
Ensure MTC is within analyzed range.
(0.25)
2.
Protective instrumentation is within normal operating range.
(0.25]
3.
Pressurizer is capable of being operable with steam bubble.
(0.25)
4.
Reactor vessel is above minimum RT-NDT temperatrue (0.25]
,
REFERENCE CC TS U-2 pas 3/4 1-7, B3/4 1-2 3.3/4.0 3.6/4.1
.
002000G005 0020000011
..(KA's)
,
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,
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CATEGORY 8 CONTINUED ON NEXT PAGE *****)
.
i
,
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,
SJDBJNISTRATIVE PRQQEDURES. CONDJ11QL Pega 45
,
AND LIMITATIONS
.
ANSWER 8.05 (2.50)
a.
Obtain a qualified relief [0.75) within two hours (0.75]
b.
Yes (0.50) the number of RO qualified personnel present would still meet TS requirements (0.50)
REFERENCE
,
CC TS, pg 6-5 l.
CCI-140D pg i
j 2.5/3.4 194001A103
..(KA's)
ANSWER 8.06 (2.00)
'
a.
Minimize the positive reactivity effects of the RCS cooldown associated with the blowdown.
(1.0)
b.
Limit the pressure rise w/in the containment in the event
!
that the rupture occurs inside the containment.
(1.0)
REFERENCE
'
CC Technical Specifications B 3/4 7-3
'
3.7/3.7 039000K405
..(KA's)
,
,
!
,
(*****
CATEGORY 8 CONTINUED ON NEXT PAGE *****)
.-
.
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8.
ADMINISTRATIVE PROCEDUBES. CONDITIONS._
raga 46 AMD'LIMITAllQEE
,
.
.
ANSWER 8.07 (1.50)
,
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 specifications.
(0.25)
(b) All equipment hatches are closed and sealed.
(0.25)
(c) Each air lock is operable (0.25)
(d) Containment leakage rated are w/in limits.
(0.25)
(e) The sealing mechanism associated with each penetration is operable.
(0.25)
REFERENCE CC Technical Specifications, 1-2 3.1/3.9 3.4/4.1*
000069G008 103000G011
..(KA's)
ANSWER 8.08 (2.00)
a.
Initiate corrective action to return 11 LPSI pur, to '>orable status within one hour.
Sne Spent Fuel Pool Cooling loop may be lined up +o pr'.'ide cooling
-
' low.
(1.0 ea.]
(***** CATEGORY 8 CONTIllUED ON NEXT PAGE *****)
l I
c_ _
_ _ _ _ _ _ _ _ _ _
4D41HISTRATIVE_FB9CEDUBEL_CONDIII.DUL Pase 47 I
AUD_LldlIAIIDES I
.
M FERENCE CC TS 3.9.8.2 LER 87-001EO 11 2.7/3.6 005000SG06
..(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, L.O.
1.a.
1.b.
3.7/4.1 194001K102
..(KA's)
ANSWER 8.10 (2.00)
a.
Alcrt (CR evacuation may be anticipated)
(0,5)
,
b.
None (0,5)
- .
S. A. E.
(Failure of 2 FPB; G.A.E.
if explein challenge to Srd).
(0.5)
d.
S. A. T.
(Failure of 2 FPB; MSLB w/ SGTR & uncontrolled release).
(0.5)
!,..
7!
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f.
K0HLNlEIB4TIVE PROCEMRKL QQHDITIONS Pega 48 t
AND'.LIMITATIONE
,
.
REFERENCE a
CC EPIP Sect. 3.0 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]
.
-
b. No report (0.35] needed when an action statement for LCO is
'
entered (0.4]
.
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 testing (0.4]
(3.0)
'
<
!
l l
a
>
REFERENCE e
-
'
Calvert Clif f s Instructions 11d-5, p1 l
T.S. Unit 2, p 3/4 1-7
)
4.1*/3.9 194001A102
..(KA's)
.
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(********** END CF EXAMINATION *******'***)
.
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NRC RECOLUTION OF FACILITY COMMENTS FOR THE CALVERT CLIFFS SENIOR REACTOR OPERATOR EXAMINATION
-
.
ADMINISTERED ON JUNE 22, 1988 Quostion Number:
5.04 3.
Fccility Comment: 3)
Raising Tavg.
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 change.
Tc increasing adds negative reactivity to the lower half of the core.
Th decreasing adds positive reactivity to the top of the core.
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) swing.
Reference:
Sama as stated.
NRC Resolution:
Agree with comment.
Reference stated supports facility comment and does not
,
'
fully support answer key.
Answer key is changed to reflect facility comment.
Quastion Number:
5.07 l
Fccility Comment: The question used units of PCMS in lieu of % for reactivity.
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 Cliffs.
NRC Resolution:
Comment Noted.
References to the unit PCM will be deleted from the question.
In addition, units of AMPS are also deleted from the question.
'
Quoation Number-5.10 b.
Facility Comment: b.1.
260 F (255-265) for 1700 psig 2.
330 F (320-380) for 700 psig This is a constant enthalpy expansion to 35 psia.
Additionally, units of psia & psig should remain consistent.
Reference:
Steam Tables NRC Resolution:
Agree with comment.
Answer key changed to reflect proper answer.
_
_.
. _.
,
.
.
.
.
- _ _.
.
-
_. _.
.
-
__-
-
. __
_.
_
-
.
.
Quostion Number:
5.14 a.
Fkeility Comment:
a.
Once through cooling is. correct.
However, the
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 Dr.
,
William Dove of CE verified this e viable cooling
'
method. The answer should be "yes".
,
F
'
NRC Resolution:
Agree with comment.
Answer key changed to YES.
,
!
Qucation Number:
6.04 b.
j
j Fccility Comment: b.
"Control Channels" is not standard Calvert Cliffs E
terminology.
Therefore, this quastion could be j
interpreted as reactor Regulating or the narrow range t
safety NI channels.
If narrow range NIs are assumed,
the answer would change to any two of the three lissted i
below.
l
!
1.
Inhibit SUR trip
!
t l
2.
Enable loss of load trip i
I 3.
Enable APD trip
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Reference:
RO-57-1-3 pages 22 & 23
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NRC Resolution:
Disagree with comment.
Reference stated in facility comment (RO-57-1-3) was not
supplied to the gruder.
Reference stated in the
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j examination (RO-57-1-2 page 39) refers to the Power Range
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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 key.
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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
Radiation Protection personnel are on shift to provide the necessary s.upport.
We request this question
'
be dropped.
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 SRO.
Included in this scope are
1.)
... describe methods for performing maintence so that he, his crew, and the general public are protected.
2.)
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 3.4 2.)
Kl.04 Knowledge of fscility ALARA program. --SRO l
value 3.5 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 him.
The Radiation Protection personnel on shift should support the supervisor in this function.
No change to answer key.
Comm2nt on Note:
Recommendation noted for future consideration.
ADDITIONAL COMMENTS MADE BY GRADER:
'
Quoation dumber:
8.04 a.
I Changed answar to require an OR response in order to I
'
obtain the requested information.
Quastion Number:
8.09 a.
j Added an additional correct answer; to reflect es rent facility information.
Quontion Number:
8.09 b.
t i
,
Changed the second answer; to reflect the common (
nomenclature of the facility.
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L
_ _ _ _ _ _ _ _.
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....
3
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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 3.
Facility Comment:
3)
Raising Tavg.
l 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 change.
Tc itereasing adds negative l
reactivity to the lower half of the core.
Th decreasing adds positive reactivity to the top of the core.
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) swing.
Reference:
Same as stated.
NRC Resolution:
/. gree with coment.
Reference stated supports facility comment and does not fully support answer key.
Answer key is changed to reflect fr.cility comment.
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 Cliffs.
!
NRC Resolution:
Coment Noted.
l References to the unit PCM will be deleted from the question.
In addition, units of AMPS are also deleted from the question.
Question Number:
5.10 b.
Facility Comment:
b.1.
260 F (255-265) for 1700 psig 2.
330 F (320-380) for 700 psig This is a constant enthalpy expansion to 35 psia.
Additionally, units of psia & psig should remain consistent.
Reference:
Steam Tables l
l NRC Resolution:
Agree with coment.
Answer key changed to reflect proper answer.
I l
,...
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.
.
Question Number:
5.14 a.
Facility Comment:
1.
Once through cooling is correct.
However, the answer
'
key states incorrectly 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 Dr. William Dove of CE verified this a viable cooling method.
The answer should be "yes "
NRC Resolution:
Agree with comnent.
Answer key changed to YES.
Question Number:
6.04 b.
Facility Comment:
b.
"Control Channels" is not standard Calvert Cl*ffs terminology.
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 below.
l 1.
Inhibit SUR trip 2.
Enable loss of load trip 3.
Enable APD trip Reference:
R0-57-1-3 pages 22 & 23 NRC Resolution:
Disagree with comm.ent.
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 key.
)
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Question Number:
7.13
.
Facility Comment:
This knowledge is beyond that expected for an SRO at i
Radiation Protectic,n personnel are on shift to
!
provide the necessarv support. We request this question
be dropped.
.
.
NRC Resolution:
Comment noted.
'
)
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
'
1.)... describe methods for performing maintenance so that
[
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
concept.
!
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 3.4.
j (2) K1.04 Knowledge of facility Al. ARA program.
--SRO
,
value 3.5.
l
The calculation of a stay time for performing maintenance
j should be en evolution that a supervisor performs, or j
reviews, not only for himself but for the personnel who
!
t work for him.
!
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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:
!
i
,
Question Number:
8.04 a.
l J
j Changed answer to require an OR response in order to
obtain the requested information.
l
'
Question Number:
8.09 a.
I Added an additional correct answer to reflect current facility informatiu1, Question Number:
8.09 b.
I
I Changed the second answer to reflect the common
nomenclature of the facility.
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