IR 05000410/1985041
| ML20155B042 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 03/26/1986 |
| From: | Keller R, Kister H, Lange D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20155B025 | List: |
| References | |
| 50-410-85-41, NUDOCS 8604100191 | |
| Download: ML20155B042 (100) | |
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EXAMINATION REPORT Examination Report No.
85-41 OL Facility Docket No:
50-410 Licensee: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility: Nine Mile Point 2
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Examination Dates: December 10-19, 1986 Chief Examiner:
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9 86 David Lange, Le BWQxaminer
'date Reviewed by:
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Robert Kell r, Chief date Projects Section IC d pf.[K Approved by:
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Harri K' te, Chi V
'datef Projects ch No. 1 Summary: This examination report contains the results of the Operator Licens-ing examinations given at the Nine Mile Point 2 Nuclear Station the weeks of December 9 and 16, 1985. Twelve (12) Senior Reactor Operator candidates and twenty (20) Reactor Operator candidates were examined. All R0 candidates passed the written and oral examinations; two (2) failed the simulator exam-ination. Of the SR0 candidates, two (2) failed the written examination, one (1) failed the simulator examination, and one (1) failed both the written and oral examinations.
8604100191 860401 PDR ADOCK 05000410 V.
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EXAMINATION REPORT Examination Report No.
85-41 OL Facility Docket No:
50-410 Licensee: Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Facility: Nine Mile Point 2 Examination Dates: December 10-19, 1986 Chief Examiner:
David Lange, Lead BWR Examiner date Reviewed by:
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Robert Keller, Chief date
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Projects Section 1C Approved by:
Harry Kister, Chief date Projects Branch No. 1 Summary: This examination report contains the results of the Operator Licens-ing examinations given at the Nine Mile Point 2 Nuclear Station the weeks of December 9 and 16, 1985. Twelve (12) Senior Reactor Operator candidates and twenty (20) Reactor Operator candidates were examined. All R0 candidates passed the written and oral examinations; two (2) failed the simulator exam-
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ination. Of the SRO candidates, two (2) failed the written examination, one (1) failed the simulator examination, and one (1) failed both the written and eral examinations.
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J REPORT DETAILS TYPE OF EXAMS:
Initial X
Replacement Requalification EXAM RESULTS:
l R0 l
SRO I
l Pass / Fail l Pass /Faill I
I I
I I
I I
IWritten Exam l 20/0 l
7/3 I
I I
I I
I I
I I
l0ral Exam I
12/0 l
11/1 I
I
_I I
I I
I I
l Simulator Exam l 10/2 l
11/1 l
I I
I I
I I
I I Overall l
18/2 l
8/4 l
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1.
Chief Examiners at Site:
David Lange, NRC Lynn Kolonauski, NRC 2.
Other Examiners:
Frank Crescenzo, NRC Allen Howe, NRC Brian Hajek, NRC Consultant Gary Sly, PNL William Cliff, PNL Lee Miller, NRC
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1.
Summary of generic strengths or deficiencies noted on oral exams:
Most candidates were well aware of the differences between the simulator and the plant.
A few of the simulator groups were deficient in communication skills and procedure usage.
2.
Summary of generic strengths or deficiencies noted from grading of written exams:
No generic strengths or deficiencies were noted on the R0 exam.
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An overall weakness was noted in Section 5 of the SR0 exam; more specific weaknesses included an unfamiliarity with the Safety Parameter Display System (SPDS) and procedural cautions for the Reactor Recirculation System.
3.
Comments on availability of, and candidate familiarization with plant reference material in the control room:
Most candidates were adequately familiar with plant procedures but several of the simulator groups used only a limited number of procedures.
The R0 candidates were weak in locating specific piping and instrumen-tation diagrams as requested by the examiners. The SR0 candidates, however, were very successful in locating and using the P& ids.
4.
Personnel Present at Exit Interview:
NRC Personnel David Lange, BWR Chief Examiner, Region I Allen Howe, Reactor Engineer Examiner, Region I Frank Crsscenzo, Reactcr Er.gineer Examiner, Region I Steven Hudson, Senior Resident Inspector Lee Miller, Operator Licensing Branch, HDQ.
FacilityPersonnj P. T. Seifried, Nuclear Training Assistant Superintendent M. D. Jones, NMP 2 Operations Superintendent G. L. Weimer, Associate Generation Specialist, Nuclear K. F. Zollitsch, Nuclear Training Superintendent T. J. Perkins, Nuclear Generation Superintendent
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5.
Summary of NRC Comments made at exit interview:
The examiners noticed that plant accessibility requirements changed daily; this example of inconsistent access control may be indicative of a plant security problem.
Several problems caused delays during the written exam:
1.
Numerous incorrect and confusing Tech Spec action statements caused approximately a one-half hour delay when answering and clarifying two questions on Section 8 of the SR0 exam.
2.
The training material (mainly the NMP Lesson Plans and the Q/A Bank)
sent to the examiners for preparing the written exam contained many inaccuracies. As a result, the questions prepared from this material were confusing and required a great deal of clarification.
Many of the candidates even asked which answer we were looking for -
"the one in the lesson plan or the one in the procedure?"
During the written exam review, the training department asked the exam author to accept both the "right" and the " wrong" answer, because the wrong answer was identified in the training material and the candidates had been exposed to it.
We feel accepting incorrect answers for the sake of the training material is not justified and contrary to the interest of safety.
3.
The SR0 candidates were told to use the TS handout to answer the questions in Section 8 only.
The examiners observed several can-didates using the TS handout as a " memory jogger" while working in the other exam sections. This resulted in wasted time while search-ing through Tech Specs and not using the handout as directed.
4.
The Learning Objectives identified in the lesson plans should be revised to better represent the operating procedures and actual job performance. Whenever possible, the written exam questions are referenced to learning objectives.
The NMP2 lesson plans, however, did not include an adequate amount of learning objectives suitable for this use.
5.
A number of errors were identified in the surveillance testing pro-cedures used during the simulator exams; the NMP training department was notified about the errors.
The examiners thanked the NMP training department for their cooper-ation during the exam period. The roem provided for the examiners next to the simulator was very useful and appreciated.
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Summary of facility comments and commitments made at exit interview:
The licensee agreed to send a copy of the Refueling On-the-Job train-ing schedule for the operators to Region I.
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7.
Changes made to written exam during examination review:
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The following attachment addresses the NRC resolutions of Niagara Mohawk comments on the NMP 2 examinations given on December 10, 1985.
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Written Examination and Answer Key (RO)
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Written Examination and Answer Key (SRO)
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NRC Resolution of Niagara Mohawk Cemments i
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NUCLEAR REGULATORY COMMISSION REA~ TOR OPERATOR LICENSE EXAMINATION FACILITY:
NINE MILE POINT 2
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REACTOR TYPE:
BVR-GE5 DATE ADMINISTERED: 85/12/10 EXAMINER:
G.A.
SLY APPLICANT:
NI8Id/ g8/
INSTRUCTIONS TO APPLICANT Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on< top of the answer sheets.
Points for each question are indicated in parentheses afters the question. The passing
, grade requires at least 70% in each categor,y.and a final grade of at-least 80%.
Examination papers wi!! be picked up six (6)
hours after the examination starts.
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% OF CATEGORY
% OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOV 25.00 25.00 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_25.00 25.00 3.
INSTRUMENTS AND CONTROLS
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25 00'
25 00 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 100.00 TOTALS FINAL GRADE
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AII work done on this examination is my own. I have neither given nor received aid.
APPLICANT'S SIGNATURE W
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"eP PRINCIPLES OF NUCLEAR POVER PLANT OPERATION, PAGE
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOV QUESTION 1.01 (2.00)
The reactor is operating at 75% power.
Recirculation flow is subsequently increased to provide 100% power.
Briefly EXPLAIN the reactivity transient caused by the flow / power increase with emphasis on the following:
(Your answer should include the initial effect, what happens dursng the power change, and the final steady state.)
a.
core void content (1.0)
b.
core reactivity (1.0)
QUESTION 1.02 (1.00)
t Concerning control rod worths during a reactor startup from 100% PEAK XENON versus a startup under XENON-FREE conditions,
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VHICH statement is correct?
(1.0)
a.
PERIPHERAL control rod worth will be LOVER during the q
PEAK XENON startup than during the XENON-FREE startup.
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b.
CENTRAL control rod worth will be HIGHER during the PEAK l
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XENON startup than during.t h e XENON-FREE startup.
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c.
BOTH control rod worths will be the SAME regardless of i
core Xenon conditions.
d.
PERIPHERAL control rod worth will be higher during ti e PEAK XENON startup than during the XENON-FREE startup.
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CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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PACE
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QUESTION 1.03 (1.50)
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s The Reactor has been scrammed following 100 days of full-power operation.
STATE whether the following statements concerning fission poisons are TRUE or FALSE.
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a.
A 25% power reduction from 100% power would have a LARGER r
Xenon peak than a 25% power reduction from 50% power.
(0.5)
b.
The Equilibrium Concentration of Samarium IS DEPENDENT on flux level (i.e.,
stable 100% power or stable 50% power).
- (0.5)
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c.
Upon restarting the reactor following a 4-month outzge,
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the Samarium Concentration will DECREASE to its 100% full RN'jc
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power concentration.
(0.5)
l QUESTION 1.04 (1.50)
During a routine startup, control rods are withdrawn, adding a
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q specific amount of reactivity.
Consider two (2) cases:
1)
i that the reactor was slightly suberitical (Keff 0.995), and
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2) that the reactor was greatly suberitical (Keff 0.95)
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CHOOSE the word.or words that best complete the sentence.
a.
The change in the count rate in the slightly suberitical j
reactor would be (GREATER THAN, LESS THAN, EQUAL TO) the change in the count rate of the greatly suberitical reac-tor.
(0.5)
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in the count Ia t e in the slightly suberitical The[ise b.
reactor' would be (FASTER THAN, SLOVER THAN, THE SAME AS)
the rise in count rate of the greatly suberitical reactor.
(0.5)
c.
The time required to reach the equfIlbrium,, count rate in the slightly suberitical reactor would be (SHORTER, LONGER,
'3h THE SAME AS) in the greatly subcritical reactor.
(0.5)
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=1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
PAGE
- THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOV
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QUE@ TION 1.05 (2.50)
Hine Mile Pt.-2 Reactor has just experienced a LOCA.
An operator wishes to use nuclear instrumentation to determine water level within the core.
Your answer should include VHAT nuclear instrumentation would be used, HOV you would use this nuclear instrumentation, VHAT indicattons you would see and VHY?
(2.5)
QUESTION 1.04 (2.50)
During your shift an SRV inadvertently opens from 100% power 1000 psia.
By using the Mollier Diagram or Steam Tables, a.
VHAT is the tailpipe temperature assuming atmospheric
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pressure,in the suppression pool?
(0.5)
b.
If the suppression pool pressure were to increase, VHAT would the tailpipe temperature do (INCREASE, DECREASE, or STAY THE SAME)?
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(0.5)'
[. c.
11 the reactor is then depressurized, VILL the tailpipe
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temperature initially (INCREASE, DECREASEe or STAY THE SAME)?
(0.5),
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d.
At VHAT pressure would the tallpipe temperature be at its
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maximum value and VHAT temperature is it?
(0.5)-
N e.
At VHAT / pressure would the tailpipe temperature be at its minimum value?
INCLUDE value and assume a saturated system.
(0.5)
QUESTION 1.07 (2.50)
STATE, for the following conditions, whether p mp ampere would INCREASE, DECREASE, or REMAIN THE SAME:
a.
the pomp suction valve is slowly throttled closed (0,5)
b.
increase in inlet subcooling (0.5)
c.
slow closure in the discharge valve of the pump (0.5)
d.
rotor lock-up t0.5)
e.
rotor failure hgoq(
(0.5)
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CATEGORY 01 CONTINUED ON NEXT PAGE
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.1 PRINCIPLES OF NUC, LEAR POWER PLANT OPERATION, PACE
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOV OUESTION 1.08 (1.50)
Increasing recirculation pump speed will c a ti s e WHAT 'c h a n 9 5 (INCREASE, DECREASE, or REMAIN THE SAME) in each of the following parameters?
a.
actual. bundle power (0,5)
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(G.S)
b.
critical power
//d c.
critical power ratio (0.$)
QUESTION 1.09 (2.00)
A " central" and " peripheral" bundle have been inadvertently placed in each others' location.
VILL the misplaced bundles power and f l ow be -(HIGHER THAN, LESS THAN, or THE SAME AS) the
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same type of bundle in the same area of the core?
y a.
Central bundle in peripheral location:
(1.0)
b.
Peripheral bundle in central location:
(1.0)
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FN1M,CIPLES OF NUCLEAR POWER PLANT OPERATION.
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THERMODYNAMICS, HEAT TRANSFER AND FLUID PLOV QUE$ TION 1.10 (2.00)
MATCH the most correct " parameter" listed below to the corres-ponding " fuel integrity item".
(2.0)
Parameter
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1.
LHCR 2.
Bulk boiling 3.
Total peaking factor (TPF)
4.
Onset of transition boiling (OTB)
5.
Critical quality 6.
CPR 7.
APLHGR 8.
Boiling length
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Tuel Integrity Item a.
Specified to protect against boiling transition.
b.
Specified to limit plastic strain and deformation of clad-ding to less than 1%.
c.
Specified to limit peak fuel cladding temperature during
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a LOCA to less than 2200 deg F.
d.
Specified as the point / time when the liquid film a l o r.g the rod's surface is evaporated and cladding temperature starts 16 risa rapidly.
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QUEGTION 1.11 ('2. 0 0 )
COMPLETE the following:
(Blanks A through D MAY have more
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(2.0)
X'-133 has two (1) methods of production.
A b o u't 95% of the Xe e
is produced by
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_ an1 the remaining 5% of Ie as (
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pr'oduced by (B)
Xe also has two {2) removal
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unthods; at high power levels (C)
is the major
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removal method, at low pcwer levels (p)_
predominant removal method.
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becomes the
QUEGTION 1.12 (1.00)
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Explain HOV and WHY excess reactlyity varies with core age.
(1 0)
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CATEGORY 01 CONTINUED ON NEXT PACE *****)
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.1 Ph!NCT_PLES OF NUCLEAR POWER PLANT.0PERATION, PAGE
YHERM0 DYNAMICS, HEAT TRANSFER AND FL QlO FLOW i
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QUCSTION 1.13 (3.00)
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i You are currently operating at 100% p' owe r DOL when you loose partial feedwater heating.
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fa.
If the same situatibn were to occur at EOL, VHAT would be the
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corresponding reactivity changes (MORE NEGATIVE, LEBE NEGATIVE, E l gi NO CHANGE) to each of the abret co'efficients (i.e.,
delta (T)-
t mod, d e 'I t a (% voids), delta (T))?
(1.5)
f%Ed b.
If the STA te!!s you that feedwater temperature decreased by
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10 deg F,
voids decreased by 2%, and reactivity returns to
'P zero, WHAT would be the corresponding tomberature changa to
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-the fuel temperature?
(Assuma no rod movement, recirculation j
flow changes )
(1.5)
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.2 RhANT DESfCN INELQDTNG MATETV AND EMERGENCY SYSTFMS PACE
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QUESTION 2.01 (2.50)
Concerning the Standby Cas Treatment System:
a.
ARRANCE the following components in flowpath order from the reactor.
(1.31 1.
f a r.
2.
demister
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3.
elsetric heater
4.
flow element (train)
radiation element b.
STATE whettisr the IcJ1owing signals would (INITIATE, ISOLATE or NOT CHANGE) the SDGT syctem.
(1.5)
1.
The receipt of a high temperature alarm in Train
"A".
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Assume Train
"A" running and Train
"B" had been manually stoppsd.
f. Answer for each train.)
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High radiation alarm at the front face of the turbine.
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3.
High radiation alarm in 'he HPCS pump room.
t 4.
Vater level equal to 105 inches.
5.
Drywell pressure equal to 1 65 psig.
QUESTION 2.02 (2.50)
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a.
VHAT are the differences in modes of operation for the
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RHS Loops A and BP (0.5)
b.
VHAT is the reason for the intettbek between the (1.C)
1.
shutdown cooling section valve and t h e,.t e s t return valve?
2.
pressure control valve bypass valve (MOV-23A) and Rw pressure?
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c.
If a LPCI auto initiation function (high drywell) was overridden to realign the system in thutdown cooling mode and another LPCI signal (triple low level) was to come in, VOULD the RHS Loop realign from the shutdown cooling
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mode to the LPCI modes EXPLAIN.
(1.0)
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CATECORY 02 CONTINUED ON NEXT PAGE ***ma)
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PLANT DESICN INCLUDENQ APETY AND EMERCENCY lljTEMS PAGF
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OUESTION 2.03 (2.5C)
a.
The LPCS system design has features to pr6tect pipAng from overpressure.
To manual,1y open the LPCS inbaard and out-
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board l#alation valves, the correct sequenas is.
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(CHOOSE on6).
( As s um.e CS pumps are OFF initially a ttd both valves are closed.)
(1.0)
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1.
valve differential pressure i 994 psig outboat'd opens o\\b
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-then inboard opens 2.
valve differential pressure ( 700 psig, inboard o p'e n s then outboard opens 3.
vilve differential pressure ) 700 psig, outbcard opens then inboard openo 4.
valve differential 'pressore ) 700 psig, inboar6 opens
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then cutocard op,4ns
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b.
D"SchiBE th,a oper6 tion of the core spray sparger break Catection systek.
I,NCLVOE in your answer WHERE pressura is physically senttdr and VHAT delta pressures _a r e sensed.
(1.5)
QUESTION Y.04 (1,06)
The Reactor Recirculation Pump sect cartridge nssemblies consist of two (2) sets 9f sealing surraces and breakdown buehlng assem-blies.
Failura of the No.
2 seal assembly at rated conoitions would result in,.
(CHOOSE o n.a. )
(1.0)
a.
an increkse in No.
2,saal cavity pressure (Tok j
approximately 500 os'ig.to.tpproximately 1000 psig
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a.d e c r e a s e in No. 2 seal cavity pressure from approximat=Iy 500 psag to approximately 0 pstg i
c.
an increase in No. 1 seat cavity pt.nssure from approrsmately 500 psig to approximately 1000 psig q
d.
i decrease in NJ.
1 seal cavity pressure from approximately
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500 psig to approximately 0 psig i
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PLANT _DESTQK INCLUDIMC SAFETY AND EMERCENCY SYSTEMS PACE
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QUESTION 2.0$
(2.00)
i There are two (2) Check Valvss loc 3ted in t h' e discharge line immediately upstream and downstreas of ths RCIC primary containment line penetration. $TAVE the two (2) purposes of t h e :s e Chack Valve.
(2.0)
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GUESTION 2.06 (2.00)
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ANSVER TRUE,or FALSE for the foliowing:
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a.
The CRD Water Header pressuse is normally maintained at 260 psig above reactor pregsure.
(0.5)
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The standby CRD pump auto starts whEn the running p timp -
trips. ff 0fd'] 10ag lrh &*W c$e~ <!do fN aq rb Cldrl.r '( ( O'. Q )
c.
CRDM Accumulators are charged with air frop the service
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and instrument air system.
(0.5)
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d.
Speed Control of the CROM is accomplished by
't h r'o t t l i n g valves in the hydraulic control units.
.
(0.5)
h1ifrlh refbh hcf un O:./.:$
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.n QUESTION 2.G7 (2.50)
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STATE the following operating temperatures for the Reactor Vater Cleanup System:
a.
RVCU pump suction temperature (0.5)
b.
NMHK outlet temperature (0.5)
c.
Filter-domineraliser high temperature alarm (G.5)
d.
Filter-demineraliser inlet system isolation temperature (0.5)
e.
Return to feedwater temperature.
(0.5)
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,jZ. _ PLANT DERIGN.TNCLUDING SAS'ETY AND EM' ERQ_ENCY SYGTEME PAGE
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GUESTION 2.08 (2.50)
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a.
VilAT conditions will cause the Giv. !!! (CSH) diesel generator to shutdown during u LOCA condition?
(3 required)
(1.5)
h.
E6 sides the fuel oil storage and transist system, VHAT are the 6ther live ($) auxiliary systems necese_ary for reliable and safe 6peration?
(1,0)
QUESTION 2,09 (1.50)
Concerning c~ombustible gas production fo116 Wing a Loss of
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Cooling Accident (LOCA):
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a.
STATE two (2) sources of hydrogen prbduction.
(1.0)
b.
STATE the single spyrce of oxygen production.
( 0. 5.)
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"1 QUESTION 2.10 (2.DC)
LIST ths tout (4) signals which will cause an a,biopatic necirculstioS puhp downsalft from fast to slew speed.
(2.0)
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GUEST!CN 2 11 (2.00)
Concerning the four (4) vacuum relief lines between.the drywall and the. suppression c h a m b e,r.
a l.
In VHICH direction is ths flow designed to go?
(0.5)
b.
VIIAT condition (s) do the vacuum rettet lines' limit br protect ayatnst?
11.5)
UUESTION 2.12 (2.00)
. LIST the four (4), non-electrical, trips associated with 4 i
3 reacter feed puap (2 0)
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INSTRUMENTS AND CONTROLS PAGE
QUESTION 3.01 (3.00)
VHAT are the four (4) anticipatory scrams, HOV is each sensed, and HOW is each one bypassed?
(3.0)
QUESTION 3.02 (2.00)
The reactor is at 100% power with the generator synced to the grid.
Electrohydraulic Control (EHC) load set is 105%.
By using the attached EHC diagram, EXPLAIN VHAT would happen (con-trol valve, bypass valve) in the following circumstances:
a.
load limit potentiometer reduced to 95%.
(0.5)
b.
maximum combined flow limit potentiometer reduced to 95%.
(0.5)
X /NV c.
"A" pressure regulatory (%-ipy ini) fails low.
(0.5)
d.
failure of two (2) bypass valves full open.
(0.5)
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QUESTION 3.03 (2.00)
ANSWER the following questions based upon the situation described below.
The RRCS is fully operational.
The RRCS receives a reactor water low level 4105 inches) signal in both complementary logics of an RRCS channel and remains in for 120 seconds.
It takes 100 seconds from the initial reactor water low level signal before the APRM level is downscale.
a.
VHICH of the four (4) logics integrated into RRCS are actuated at T = 0 seconds?
(0.5)
b.
VHICH logics are actizated at T 25 seconds?
(0.5)
=
c.
VHICH logics are actuated at T = 98 seconds?
(0.5)
d.
HOV LONG from T = 0 seconds is it before the RRCS can be reset?
(0.5)
.
(*****
CATECORY 03 CONTINUED ON NEXT PAGE
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-
_
_ _ _ _ _ _ _ - _ - _ _ _ _ - _ _ - _
...
I'STRUMENTS AND CONTROLS
3.
N
.,
PAGE
QUESTION 3.04 (2.50)
An automatic RCIC initiation has occurred.
Subsequently, RCIC injection was automatically terminated due to high reactor water level.
a.
VHAT component in the RCIC system functioned to terminate the injection?
(0.5)
b.
Assuming no operator action, HOV will RCIC-respond to a subsequent decreasing water level? (below high water isolation setpoint)
/
(0.5)
c.
If an RCIC " Turbine Test" had been in progress when the initial automatic initiation signal had been received, HOV would the system have responded?
(O*I)
-
/
d.
If, following the initiation, the RCIC turbine had tripped on overspeed, COULD it be reset from the Control Room?
(O.f)
e.
If the RCIC system were lined-uptin standby, VHAT would
.
be the functional result of depressing and releasing the
~ ~.
'
manual i so la t i on'.bu t t on ?
( U'
"
QUESTION 3.05 (2.00)
For each of the following situations, STATE whether the ADS valves will OPEN, CLOSE, or REMAIN IN THE SAME position.
Initial Condition Action / Event a.
ADS logic initiated with all Turn off all operating ADS valves open.
ECCS pumps.
(0.5)
b.
All ADS logic s,ignals initiated y Push Channel A High,DV 'lyh$ h 105 sec. timer timing out.
rJf' PRESS SEAL-.IN, puM butt'on then timer times out.
(0.5)
c.
ADS valves closed en-2 ;b. a Failure of the N2 supp.ty St::x!ine; All initiation system downstream of signals are in, 105 timer storage tank (TK4).
just timed out (0.5)
d.
SRV keylock control switch All initiation signal (PNL601) for ADS valve in come in and timer times off position.
out.
(0,5)
.
(*****
CATEGORY.03 CONTINUED ON NEXT PAGE
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.
$.
-
_-
. _.
_. ~.. _,, _,,.. _.,
- _ _ _,
.
_
-
_ _ _.
_ _ _,. _ _,, _,,.. _, _ _ _ _ _ _.,,.,, _ _.,.,, _,, _,..
.'
- ~
INSTRUMENTS AND CONTROLS 3.
PACE
QUESTION 3.06 (3.00)
EXPLAIN VHAT affect the following failures would have on reactor level.
VHY?
(Assume 3-element control and Channel A control-ling.)
a.
'C'
steam line flow signal fails Iow.
(0.75)
b.
Channel
'A'
reactor level detector signal fails low.
(0.75)
servomotor./h),C,p(
c.
'A'
pump
(0.75)
d.
Inadvertent activation of the setpoint setdown circuitry.
(0.75)
QUESTION 3.07 (1.00)
_
Concerning the four (4) rod display:
a.
A control rod is selected for motion and a double X (XX)
appears in the rod position window of the four (4) display panel.
VHAT does this mean?
(0.5)
b.
VHAT if a rod were selected and a position window on the
. four (4) rod display panel, NOT corresponding to the selected selected rod, indicated blank?
(0.5)
.
QUESTION 3.08 (3.00)
Concerning the I n t e r m e d i a t. e Range Monitors (IRM):
a.
If an IRM is reading 7 on Range 9 and the operator down-ranges to range 7,
VHAT will the channel reading be?
(0.5)
b.
VHAT would be the corresponding APRM power level and WHAT trips, if any, will occur?
(1.0)
c.
Briefly DESCRIBE HOW the IRM system discriminates for gamma signals.
Include in your answer the difference between this method and that used for the SRMs.
(1.5)
(*****
CATECORY 03 CONTINUED ON NEXT PACE
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_
.__
_ _.
.
...
.. _
- .,3.
I STRUMENTS AND CONTROLS PAGE
l QUESTION 3.09 (3.00)
-
\\
'
The Instrument and Service Air systems receive air from a common set of three (3) air compressors.
a.
The control switches must be in the auto after stop (green flag) position during normal operations.
If the standby compressor started, VHAT would be the consequences of matching the flag to the running status?
(0.5)
b.
If the air header pressure continued to drop after the standby compressor started, and the Service Air System isolated, VHAT action is required to restore Service Air?
(0.5)
c.
If Instrument Air were to be completely lost, in VHAT position would each of the following valves fail?
(2.0)
'
1.
scram inlet valves l
2.
reactor water cleanup filter /demin inlet and outlet valves l
3.
cooling tower level control valve
'
4.
condenser 4-inch make-up valve (LV-103, Normal make-up)
QUESTION 3.10 (2.00)
i WHAT five (5) conditions will cause the Loop Flow Controllers to automatically transfer from Automatic to Manual, when operating in Master Manual control?
(2.0)
.'
' ~
(*****
CATEGORY 03 CONTINUED ON NEXT PAGE
- ****)
.
m-
-
v-
,e,,e
.-
y y
w
--
pr.ne
3.
TNSTRUMENTS AND CONTROLS
.
PAGE
l QUESTION 3.11 (1.50)
Given the following data for APRM Channel C:
'
LPRM Level:
A B
C D
Number of LPRMs assigned:
5
5 Number of LPRMs bypassed:
4
0 a.
If APRM Channel C selector switch on the local (back)
panel was placed to the COUNT position, VHAT would be the expected meter reading?
(SHOW calculations.)
(0.5)
)
b.
Based on the above data, is APRM Channel C operable:
ANSVER YES or NO and EXPLAIN VHY.
(1.0)
. _..
-
u a
(*****
END OF CATEGORY 03
- )
..
-.
.
._.
- -. - - - -
-... -.
..-.- -- -
-~
- - _
- _ _ _ _ - _ _ _ _ _ _ -
.
.
.4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
RA_DIOLOGICAL CONTROL QUESTION 4.01 ( 2.-5 0 )
During a plant startup and heatup, several actions must be taken as a function of RPV pressure.
For EACH of the following actions, GIVE the approximate pressures by which, or above which, the action must be taken according to N2-IOP-101A, P l.a n t Startup.
a.
The ADS must be verified operable prior to re. actor pressure exceeding (0.5)
...b.
Condenser vacuum must be established prior to opening a bypass valve with the vacuum being maintained by the SJAEs.
The EHC will open a bypass valve at approximately (0.5)
.
i c.
Start a motor driven feedpump when reactor pressure reaches i
about (0.5)
d.
Transfer the Mode switch to Run after (among verification of
other parameters) the steamline pressure has been verified
'
to be greater than (0.5)
e.
RCIC must be determined. operable prior to exceeding a reactor pressure of (0.5)
OUESTION 4.02 (3.00)
ANSVER the following questions concerning the main generator and load changes.
USE the attached Power Factor Chart.
a.
WHAT would be the operating load (MWe, KVA) limit with a lagging power factor 0.9 and H2 pressure at 30 psig?
(0.5)
b.
You are operating at a 0.'95' lagging power factor with 75 psig H2 and the load dispatch'er orders' you to d r,,o p your power factor to a 0.9 lagging power factor but maintain maximum MVe output.
In general, HOV would you change your operating condition?
Include in your answer initial conditions (MWe,KVA), a brief discussion of the power change, and the final conditions (MVe, KVA)
(2.5)
s
~(*****
CATEGORY 04 CONTINVED ON NEXT PAGE
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.
-
-
,,
- -- -
-
--,
,,,,--.-,-.-v w-
,e--
,,-e w - ~
wn-,
-
!.
- .4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL QUESTION 4.03 (3.00)
According to the start-up procedure:
a.
HOW is the SRM/IRM 1/2 decade overlap supposed to be verified?
(1.0)
b.
If reactor power is 13% and the mode switch is in start-up, SHOULD the reactor have scrammed?
VHY?
(0.5)
c.
HOV is the reactor determined critical (3 conditions)?
(1.5)
+s QUESTION 4.04 (1.00)
During the " steam condensing mode" of RHS, EXPLAIN HOW reactor cooldown rate is controlled.
(1.0)
,-
?. F
,..g
.
..
QUESTION 4.05 (1.00)
VHAT reactor conditions and characteristics [four (4) required]
influence the point of criticality and the rate at which it is l
approached during a reactor startup?
(1.0)
QUESTION 4.06 (1.50)
A precaution in OP-92, Neutron Monitoring, states that."BVR cores typically operate with neutron 11ux noise.
Care sho'uld be taken when operating in this area."
,
a.
WHAT problem can this " noise" create?
(0.5)
b.
In VHAT specific operating condition is this applicable?
(0.5)
c.
WHAT actions are required if this condition exists?
(0.5)
.
(*****
CATEGORY 04 CONTINUED ON NEXT PAGE'*****.)
.
.
.
P'OCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PACE
".4.
R RADIOLOGICAL CONTROL QUESTION 4.07 (2.00)
For the CRD System:
a.
PROVIDE the four (4) indications of a sucessful coupling check.
(1.0)
b.
WHAT immediate operator actions are required on loss of all CRD flow?
(1.0)
QUESTION 4.08 (2.00)
ANSVER TRUE or FALSE to the following questions on the Rod Vorth Minimizer (RUM) System.
a.
If an insert block is present, then three (3) control rod insert errors HAVE occur'ed and ALL three (3) rods r
are positioned.two (2) even notches past their pull sheet
,
minimum limits.
(0.$)
b.
When changing Ru power into the RV5~ operable range, the Rod Group V. indow VILL display the highest group which has
,
l less than t h r e e.'.( 3 ) insert errors and at least one (1) rod withdrawn'past its minimum limit.
(0,5)
c.
The select error lamp VILL illuminate whenever the selected rod is not responsible for the current rod block.
(0.5)
d.
If Rx power is changed such that the RVM system becomes operational with greater than the maximum amount of insert i
and withdra'wal errors present, NO NORMAL rod movement is possible, u n !'e s s the grosp contains a control rod causing an insert / withdrawal error, ( O.'. $ )
.'
.
l
.
(*****
CATECORY 04 CONTINUED ON NEXT PACE manna)
.
___. - _
,
.-
.-
-
-
- _..
. _ _ -. _. - - _,. - -,..
. -. -
- _
- -
.
,
I
.4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE
)
RADIOLOGICAL CONTROL
,,
QUESTION 4.09 (3.00)
ANSVER the f o l'1 o w i n g questions concerning radiation and radiological control:
For a 20-year-old employee with an accumulated occupational dose of 8 rem.
a.
WHAT would be the employees maximun federal limit for the quarter?
(1.0)
b.
COULD this employee b'e eligible for a life saving action and not violate any federal limits.
EXPLAIN (1.0)
c.
If the above individual were assigned to assist in the charging of the CRD accumulator (predicted to take 3-hrs)
WOULD he/she violate any administrative limits?
Radiation
-
i Protection stated that a 25 mrem /hr dose exists in the area.
j (Answer YES or NO, and PROVIDE limit.)
(1.0)
QUESTION 4.10 (2.50)
'+
According to Procedure N2-EOP-RL, a.
VHAT precautions must be taken PRIOR TO placing an ECCS system in manual?
(1.5)
b.
VHAT precautions must be taken VHILE an ECCS system is in manual?
(1.0)
QUESTION 4.11 (2.50)
You have been operating at 60% power when one (1) recirculation loop trips.
You have been requested to restart the idle loop, a.
According to the Recirculation Procedure, VHAT are the thermal limits that apply to the restart of an idle loop?
(1.5)
b.
If the idle loop cannot be restarted, COULD you continue to operate with only one (1) recirculation loop for an j
. extended period of time, (i.e.,
greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)?
l EXPLAIN.
(1.0)
l-(*****
CATEGORY 04 CONTINUED ON NEXT PAGE nanan)
i
.. -
- - - -
-
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PAGE
,
6.
PROCEDURES -. NORMAL. ABNORMAL. EMERGENCY AND
.
RADIOLOGICAL CO!ITROL
QUEQTION 4.12 (1.00)
LIST the Entry Conditions for Reactor Pressure Vessel (RPV)
(1.0)
Vater Level Control.
l
1-we e
.itPte;:
c
\\
,
~
.
(*****
END OF CATEGORY 04
- )
(*******manen*
END OF EXAMINATION nenneme****aann)
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_______________________________________________________________
EQUATION SHEET
____ _____________________________________________________________
.
.
,
Where mi = m2 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2
__________________________________________________________..______
KE = mv2 PE = mgh PE +KE +P V 1 i = PE +KE +P Y-where V = specific i
i
~1f
2 22 volume
_____________.._________.____ ______________________________P
= Pressure
__________.
Q = mc (Tout-Tin)
Q = UA (Tave_Tstm)
-Q = m(h -h I p
i 2
______________________________________________________ ______....__
P = P 10(SUR)(t)
P = P e /T SUR = 26.06 T = (8-p)t t
o o
T
_ _.. _ _ _ _ _ _ _ _ _ _ '_,.____________________..____________.___________p
_..__.
delta K = (Kef f-1)
CR (1-Keff1) = CR II-Keff2)
CR = S/(1-Keff)
2
'
-
M = (1-Keff1)
SDM = (1-Koff) x 100%
(1-Keff2)
_________________________________..___________________.._________
decay constant In (2)
0.693 A1 = A e-(decay constant)x(t)
=
=
g t1/2 t1/2
--_______________________.____________________________________
Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 gallon = 3.78 liters 1 Curie = 3.7 x 1010 dps 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 103 Btu /hr
Density = 62.4 lbg/f t 1 MW = 3.41 x 106
'
Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporizatfor, = 970 Btu /lbm Heat of Fusion = 144 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 1 inch = 2.54 centimeters 1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec2
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- 1'
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
.
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOV AN; VERS NINE MILE POINT 2-8 5 /12 /10-C'. A. SLY
--
2.
NMP-2 Exam Bank Question, p.
Cat. 1/3.
>
ANSVER 1.07 (2.50)
a.
decrease b.
increase c.
decrease i
d.
increase
'
e.
decrease
(+0.5 pts for each)
REFERENCE 1.
NMP-2 Student Learning Objectives for Fluid Statics, Dynamics,
and Delivery No.
7, 10, 12, 14, pp.
15 to 17.
b
'
ANSWER 1.08 (1.50)
e- -
a.
actual bundle power increases (+0.5)
b.
critical power increases (+0.5)
..
c.
critical power ratio decreases (+0.5)
REFERENCE 1.
NMP-2 Student Learning Objective for BWR Thermodynamics and Thermal Hydraulic Limits, No.
7, p.
9.
2.
General Electric Thermodynamics, Heat Transfer, and Fluid
'
'
Flow, MTC, March 1983, pp.
9-81, 9-86, 9 - 9 2,,
l
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.
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.
.
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.
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.
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. _ _
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-
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.
- 1 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOV ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY
!d C,G l.70 (fx v)odb Q,N NonO
.$sn k O k t h a n
-
W-low lla.4 daavkfw(wtr.
..
ANSWER 1.09 (2.00)
b*b ' Sf d^t m '*" (8d tF a.
Central bundle in peripheral location:
power higher than (+0.5)
/
H19Uh di l' h % v X " Kgm gg iIow hHyMm than (+0.5)
f Ag b.
Peripheral bundle in a central location:
"
"
power lower than (+0.5)
is se A
A.u-dt.C.WAa4.
"
i, flow than (+0.5)
REFERENCE 1.
Thermodynamics Lesson Plan, BVR Thermodynamics and Thermal
-
Hydraulic Limits, p.
13 of 20.
2.
NMP-2 Examination Bank, Category 1.5, p.
73.
n ANSVER 1.10
,(2.00)
j
,
a.
6.
CPR - protect against transition boiling b.
1.
LHCR - maintain cladding than 1% plastic strain.
c.
7.
APLHCR - maintain peak cladding surface temperature to less than 2200 deg F following a LOCA d.
4.
,
(+0.5 pts. for each response)
REFERENCE
,
1, NMP-2 IntrodSction to Thermodynamics and Thermal Hydraulic Limits, Figure 4,
p.
6, Student Learning Objective Nos.
1, 2,
5.
'
.
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, _
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,-
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-
.
- 1'
P'RINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOV ANSWERS NINE MILE POINT 2-85/12/10-G.A. SLY
--
ANSWER 1.11 (2.00)
A.
1-135 decay (+0.5)
B.
Direct fission yield (+0.5)
C.
Burnout (+0.5)
D.
Xe-135 decay (+0.5)
REFE,RENCE
1.
NMP-2 Operations Technology, Module I,
Part 16, pp.
1-16-1 to I-16-3, Student Learning Objective No.
!
ANSWER 1.12 (1.00)
_.
i
!
Initially the excess reactivity of the core will decrease i
'
due to a buildup of fission product poisons (+0.25).
Once fission product poisons reach an equilibrium value, the excess reactivity will increase as burnable poison burnout exceeds fuel l
burnup (+0.25).
This increase continues to a maximum value where fuel burnup begins to exceed poison burnout (+0.25)
The value then decreases until refueling due to fuel burnup (+0.25).
l REFERENCE l
1.
NMP-2 Operations Technology, Module I,
Part 7,
p.
7-7, Student Learning Objective No.
3.
Provide K-excess Graph.
2.
NMP Examination Bank Question Category 1,
5, p.
39.
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-
- 1-P'R'I N C I P L E S OF NUCLEAR POWER PLANT OPERATION, PAGE
THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW ANSWERS NINE MILE POINT 2-85/12/10-G.A. SLY
--
ANSVER 1.13 (3.00)
a.
delta (ro)
alpha (x) [ delta (x)3
=
delta (ro)D = more negative (+0.5)
delta (ro)V less negative (+0.5)
=
delta (ro)M = less negative (+0.5)
b.
alpha (d)(delta (T) fuel)
alpha (m)(delta (T) mod) + alpha (v) delta (%V)
=
(0.25 pts for equation)
alpha (m)
-1.0 x 10E-5 delta K/K/deg-F (0.25)
=
-1. @ z 10E-5 delta K/K/deg-F (0.25)
alpha (d)
=
alpha (v)
=
-1.0 x 10E-3 delta K/K/%V (0.25)
\\.0 10 E-3 (-2 ) ) ] / (-1/! x del a(T) fuel =((-1 x 10E-4(-10))+(-1 x 10E-5)
- -
delta (T) fuel
= -250 deg F or 250 deg F increase in fuel temp. (0.5)
-
~300 REFERENCE 1.
NMP-2 Operations Technology Module I,
Part.13, pp.
13-5, 13-6, Student Learning Objectives No.
2.c, 3.
2.
NMP-2 Operations Technology Module I,
Part 12, pp.
12-5, 12-7, Figures 12-6, 12-7, Student Learning Objectives
,
No.
2.b, 3.a, 3.b.
i
!
1 l
!
.
<
l
I
.
.
.
b
-
_. _.
_
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
'
+2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
AN~WER3 -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 2.01 (2.50)
a.
2, 3,
1, 4,
(demister, electric heater, fan, flow element, radiation element)
(+0.2 for each)
sta CVas G PC ch,42.
[yfted k. -kig )
~
b.
1.
.ini+8= _e Train
"B",
isatrt: Train
"A" (+0.5)
2.
no change (+0.25)
3.
no change (+0.25)
4.
initiate (+0.25)
5.
initiate (+0.25)
REFERENCE
.
1.
NMP-2 Operations Technology, Module V,
Part 6,
SBCT, pp. 2.- 7, Student Learning Objective Nos.
2, 3,
5, 6.
j]z-rop-6tf, & 2 ANSVER 2J2 (2.50)
Mit C
'
Mohr ovevaktA V h te lwy'B'ml1 907 W Au "
a.
RHR*B' - head spray (+0.25)
g
- containment flooding (+0.25)
(+0.5 TOTAL)
b.
Prevent inadvertent 4' raining of the vessel (+0.5)
2.
Prevent exceeding RHS design pressure (+0.5)
(+1.0 TOTAL)
M ( +0 - serktan Yh clo he MP (No 4 ov*w on ky Fost Golf @
9 0.25), because the second LPCI initiation signal c.
will realign the system by reopening the LPCI infection valve.
(+0.75)
(+1.0 TOTAL)
REFERENCE
NMP-2 Operations Technology, Module IV, Part 5,
RHS, pp.
5, 9,
10, Student Learning Objective Nos.
1, 5,
6.
ANSWER 2.03 (2.50)
a.
"1" (+1.0)
'
b.
Differential pressure is sensed between the core spray infection line (+0.25) and the RHS A LPCI injection r.::: r l(Q (+0.25)
A break in the CS piping outside the shroud (+0.25)
but inside the vessel (+0.25) would cause the dp to i ncrease DD fothe pressure drop across the steam separators.(+0.5,
+1.5 Total)
> 0<- 7tb kwy / Cove
.
.
0
.
.
, _-
. -.,.,
-
. - - -
._
-
-
-
_ -_-
- - ___ -
1.
PCANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE
,
ANSWERS -- NINE MILE POINT 2-85/12/10-C.A. SLY REFERENCE I
1. NMP-2 Operations Technology, Module IV, CSL, pp 7,
'
.
ANSVER 2.04 (1.00)
b.
(+1.0)
REFERENCE 1.
NMP-2 Operations Technology, Module III, Recirc.,
p.
3, Fig. 6 ANSWER 2.05 (2.00)
Mved M
,
-
l Check Valve -
Assure a non-is la ble or non-servicable flow
'
path for RCIC (
) and provides for the inside and outside primary containment i solation valve.
(+1.b)
dd 50ht k C.V VT %3 ohr VdV4., W(U not yi- (sol %s
'
'
_
' REFERENCE 1.
NMP-2 Operations Technology, Module IV, RCIC, p.
3, Fig. 1 i
l
,
'f ANSVER 2.06 (2.00)
a.
TRUE (+0.5)
b.
TRUE (+0.5)
J c.
FALSE (+0.5)
d.
TRUE (+0.5)
. REFERENCE
.
l
NMP-2 Operations Technology, control Rod Hy.draulics Rev. 1 p.
10 of 14, Student Learning Objective No.
'i i
i i
.
.
!
.
!
.
.
L
.
, - - -.
,,-.,,,.,-..r-
.-
,,w,--
..-g
..,.-,--,,_,,..-_,,-n.
_,. - - -,,..,
a,
,,,,. -
-. - - - - -
., -, -
,--,~~w-
.
.
..
<2.
. PLANT DESIGN INCLUDING SAFETY AND EMERCENCY SYSTEMS PAGE
AN;VER3 -- NINE MILE POINT 2-85/12/10-C.A. SLY rk%
AN;VER 2.07 (2.50)
1000 yfta - SYS'P t S* p=
g a.
533 deg F +/- 5 deg F (NMChQ)
b.*v120 deg F N O. f- '37 IRy "altnAi-W ' F
c.
130 deg F (no tolerance)
d.
140 deg F (no tolerance)
e.
437 deg F +/- 5 deg F NOTE: Since students were taught that there is a 100 deg F difference between (a) and (e) above, accept a 100 deg F difference for full credit for (e)
(+0.5 for each value)
REFERENCE
'
1.
NP-2-OP-37, Reactor Vater Cleanup.
'
2.
NMP-2 Exam Bank
-,
ANSWER 2.08 (2.50)
Acc3 b OY 100 1 y@
a.
1.
Engine (or generator) overspeed (+0.5)
2.
Generator (fitferintial lockout (+0.5)
I* 5% (ief.MCAd IIT U.. "
3.
Manual stop (+0.5)
SwihA k
bD b.
1.
Fuel oil supply 1*
2.
Jacket water cooling /SW 3, ggpc5g g y g p q 3.
Starting air 4.
Lubrication k* N @d M'ldM j
,,
5.
Combustion air (+0.2 each,
+1.0 Total)
'
T* YWtCL WW REFERENCE 1.
NMP-2 Operation Technology, CSH Diese! Generator, Rev.
1, p.
3, 13 of 16, 4=ght
'ayrecN w ind mty p %
ybet wl ov:"
l
.
-
-
-
-
_--n
-
,
__
-
-
P. ANT DEBICN INCLUDING SAFETY AND EMERGENCY SYSTEMS PACE
AN;WER3 -- NINE MILE POINT 2-85/12/10-C.A. SLY
,
AN;VER 2.09 (1.50)
a.
radiolytic decomposition of water, sire-water reaction (+1.0)
b.
radiolytic decomposition of water (+0.5)
g.,
REFERENCE 1 O k f M cy\\ d f "Z'.lb t.
(h NM 1.
N2-IOP-62 Hydrogen Recombiner.
2_. Ou of-M b 12.W
,
k%. b Mceht ANSVER 2.10 (2.00)
.
g 4 ap $
_
~
a.
Steam line to pump suction temp. difference is ( 7 deg
,
b.
Total feed flow ( 30%
GQC-A f7~)
c TSV or TCV closure with power
>'30% of rated d.
R e a c.t. o r w a t e r (. level 3
/P7 (+0.5 for each,
+2.0 TOTAL)
E'
b g
.
kIMJI C( WMA'
REFERENCE k ; It.itCi (
%
1.
NMP-2 Operations Technology, Recircu~1ation System, Rev.
1, pg.
ANSVER 2.11 (2.00)
a.
Suppression pool to drywe!!
(+0.5)
>
b.
limits negative pressure differential (+0.5)
to prevent drawing water.up the downcomer from the suppression pool to the drywe!!
(+1.0)
Of*
YYYtt tk. 0 YW
.
REFERENCE
NMP-2 Operations Technology, Primary C o n t a l" nme n t, Rev. 1 pg.
N O "~ SM Q.b. O (
ANSVER 2.12 (2.00)
RPV high level D' eve
'
2.
Low suction w/TD
3.
Iow-low suction 4.
Iow tube oil pressure (+0.5 for each,
+2.0 TOTAL)
)
,
3
!
._. -,
,
_ _. _
.
--
. ~.
.
,
I
'
'. 2'.
PdANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE
l i
AN~ VERS -- NINE MILE POINT 2-85/12/10-C.A. SLY I
i i
I
.
REFERENCE 1.
NMP-2 Operations Technology, Feedwater Sys.,
p.
,
h a
.
t
'
,
' ds
-
s
.
%
B
'
}
$
b
?
e p
e
,
i i
6
-
-
.
.
.9 JdJI3pMENTS AND CONTROLS PAGE
ANSVERS -- MINE MILE POINT 2-85/12/10-G.A. SLY
.
I ANSWER 3.01 (3.00)
a T C S., TCV, MSIV' closure, plus the SDV high water level scrams.
(+1.0 TOTAL)
b.
TSV valve position (>5% closed)
-
TCV-EST fluid pressure ((530 psig)
MSIV closure-valve position ()d% closed)
'SDV-level switches (or transmitters) (25 ga l ) = 46, T "
(+1.0 TOTAL)
c.
TSV bypasses when (30% of rated (1st stage shell
-
pressure)
'
TCV same as above
-
M5IV closure - bypasses with the MSS out of run j
SDV scram - MSS in S/D or Refuel with the bypass switches in D/P (4 total)
'~
~~
0. l[ br Ok
REFERENCE
,
NMP-2 Operations Technology, Reactor Protection System, Rev.
1, Table 1 2.
NMP-2 Technical Specification Bases, RPS LSS.
,g ANSVER 3.02 (2.00)
4.
control valves close 5% (+0.25),
open one bypass valve (+0.25) (or similar answer on diagram). ~(+0.5 Total)
b.
control valves close 5% (+0.25),
reactor scram probable due to increasing pressure since bypass
'v a l,v e s will not be open (+0.25).
(+0.5 Total)
'd' via 6s/s c.
w i l 1, develop"'a, press re7 error of 800 pstd.
This wil1 be s
.
j c,y, c /,g a/ demand for mapimu opening offaTl valves.
However) due
- cy,,,
y t,o the acflon o !,\\ t h maginum combined flow limiter, c o n t r o l / coy,,f,,y,,,i valves'wi!! go to
%, demand and' bypass valves to g%. l
/.e A4 u,,
/(+0.5)
l oo */.
~
$ y'/, # efyp,c gyu,,s fttr*rd.
d.
control valves close to 90% (+0.25) to maintain Rx pressure i
at 920 psig (+0.25)
(+0.5 Total)
'
_
.
.
I l
-
--
--
,
..
,
S'.
INSTRUMENTS AND CONTRQ M PAGE
ANSVERS NINE MILE POINT 2-85/12/10-C.A. SLY-c REFERENCE 1.
NMP-2 Operations Technology, EHC, Rev.
1, pp.
2, 5 to 9 of 14, Stude.nt Learning Objective Nos.
5, 6,
3,
. including EHC Figure 3.
ANIVER 3.03 (2.00)
'
,
s.
alternate rod insertion (+0.5)
b.
none (+0.5)
c.
standby liquid control (+0.i)
d.
S h eonde (+0.5)
//$M*h.
0 4.
ll,, e 34 5 L. fSLG
1.
NMP-2 Operations Technology, Module VI, Part 8,
pp.
2, 4,
'
7, Student Learning Objective No.
4.
g 4 b adYydI;ien \\fTd vt*
~
S ANSWER 3.04 (2.50)
a.
closure o'Y'the _ turbine sto_ valve (MOV-120)
(+0.5)
va h[ and b.
open the turbine stop retnject (+0.5)
c.
align in RCIC mode and inject (+0.5)
mthmt -
.
d.
no -( l oca ! !y )
(+0.5) - O /-
(fE5 elt<3e.to ! ddLA e.
no change to system logic (+0.5)
REFERENCE
NMP-2 Operations Technology, Module IV, Part 6,
RCIC, pp.
4, 9,
10, Student Learning Objective Nos,.
1, 3a, 3b, 5,
6.
.
.
'
- L
_
-.
--
,
.
.
'
3.
INSTRUMENTS AND CONTROLS PAG'E
.
ANCWERS -- NINE MILE FOINT 2-85/12/10-C.A. SLY
.
ANOVER 3.05 (2.00)
a.
valves close (+0.5)
^'
b, valves open (+0.5)
/+ peas:
=-; - " # 4'"
'"
c.
valves open (+0.5)
d.
valves open (+0.5)
'
NMP-2 Operations Technology, Moduls IV, Part 3,
ADO, pp.
10, 12, 13, 16, Student Learning Objective Nos.
3, 4,
t.
4, 7a.
ANSVER 3.06 (3.00)
a. decrease (+0.25) due to steam / feed mismatch requiring less water (+0.5)
(+0.75 Total)
b, i ncrease (+0.25) due to level mismatch requesting more water (+0.5)
(+0.75 Total)
~"
c.
no change (+0.25),
servo would lock u'p valve as is (+0.5)
(+0.75 Total)
d.
decrease (t0.25), due to reduction in operator setpoint of one-half input value (+0.5)
(+0.75 Total)
.
REFERENCE 1.
NMP-2 Operations Technology, Module IX, Part 6,
pp.
4, 5,
Student Learning Objective Nos.
4, 7.
ANSVER 3.07 (1.00)
'
a.
bad position indication into the RPIS (+0.5)
b.
peripheral rod selected (+0.5)
REFERENCE 1.
NMP-2 Operation Technology, Module'V!, Part 6,
RMCB, p.
6.
.
,
5
4
4 r
.
.
m.
.
.
.
.
.
.
.
.
.
.
.
.
-
.
.
~
_
__
_
_ _s
..
-
..
PNETRMM(NTE AND_CONTRQL1
'3.
PACE
-
ANSWERO -- N1NE MkLC POINT
-85/12/10-G.A. SLY hh3VER 3.08 (3.GO)
a.
off-scale high, (40)
(+0.S)
b.
2.3% power (93.1 MVt)
d" rod withdrawal block trip
,
s UpscaJe alarm trip 11 PClv & A F4m Q downsudbICM t
upscale trip (/, p c., 4 )
roc) elec( g.3fy
/
(+0.25 for each,
+1.0 TOTAL)
c.
The IHH systat uses the mathod of Cambulling to eliminate
'
the ggama signal (+0.5) where as the SRM system uses a pulse height distriminator.(+0.5)
The Cambelling method, roughly,
squareg the signal and then chops the gamma out.(+0.5)
'
REFERENCE 1.
NMP-2 Operations Techtiology, Module VI, Part 2,
IRM, pp.
3, 4,
6, C,
10, Student Learning Objective No. 3 n
ANSVER 3.09 t3,00)
a.
The compressor would not shut down and if it did shut down, it woyld not automatically restart.
(+0.5)
,
,
b.
The isolation valve (AOV-171) must be locally reopened.
This 16 done by placing a local switch to open.
(It will r.o t m a l. *(f f the air header pressure is spring return to greater than 85 psag, the' valve will open.)
(+0.5)
c.
1.
open (+0.5)
D
N ck.
'
2.
shut (+0.5)
i 3.
shut (,0.5)
I 4.
cpen i+0.5)
(+2.0 TOTAL)
REFERENCE
'
i NitP2 lops 19, pp.
7-10.
I i
'
,
,
.
P
- - -
- - -
-
-
- -
- -
-
-
-
-
.
.
-
-
. -
-
.
-
+3.
1NSTRUMENTS AND CONTROLS PAGE
,
,
ANSVERS -- NINE MILE POINT 2-85/12/10-C.A. SLY ANSVER 3.10 (2.00)
1.
sAny initiation of high to low recirc pump speed transfer.
2.
.High Drywell pressure (1.69 psig)
k
'
d22)-
'. 2LIO 3.
. Loss of feedpump with concurrent vessel low water level 4.
-Excessive rate of change of the Flux Controller output (f4'h4)
5.
. Deviation of 1% between the Loop Controller input
.I^ U"Y i
snu manual output signal (tracking failure)
YlFI ( *&MP f o r each,
+2.0 TOTAL)
0.4 REF E R Et(C E 1-NHP-2 Lesson Plan for RRFCS, pp'. L4-24.
12 e f 16.
,
jN3VER 3.11 (1.50)
-[ *t.
'
a.
% (+0.5),
5% (volts) for each LPRM not bypassed (+0.5)
(+1.0 Total)
b.
fio (+0.5).
There are fewer than two (2) operable inputs
-
en Level B (+0.5).
(+1.0 Total)
'
86FERENCE
~1.
NMP-2 Operations Technology, Module VI, Part 4,
APRM, p.
5, Student Learning Objective Nos.
3, 4.
.
)
'
-
-
\\
.
- - _.. - -.
.
., -
- -
.
-
-
. _ _ _ _ _ _. _ _
. - - _ _ -. _
--- I
_
.
-..
.
.-
..
,.
,
- 4.
PROCEDURE 5 NORM A L.,_hRg2Blihlg EMORQ1MCY AND PAGE
.
M AD10L CC 1g L CONTROL ANIVERS NINE MILE POJNT 2-85/13/1'3-C.A. SLY
--
AN;VER 4.01 (2.50)
a.
100 psig b.
150 psig c.
200 psig below the condensate booster pump #1 ? charge pressure.
N d.
W S"psig e.
150 psig
<>0.5 for each,
+2.5 TOTAL)
~
REFERENCE
NMP-2, N2-!OP-101a, Sections E.
2.24, 3,26, 7.30, and 3.5 ANSVER 4.02 (3.00)
a.
From the Power Factor Chart or Precautions
0.813 MVe (+0.25) and 400,KVA/,(+0.25)
II
'
E
'
b.
Initial State t 280 MVe ( $a 2 r) and 420'KVAf(+0.25)
y
'
N,.
Reduce generator load by rectrc. or control rods to 1.21 MVo
.
.+0.5).
Then raise reactive load (VARs) by adfuckina the AC voltage regulator (+0.5). [To be done in this order as not to
,_
ceec operationas 11mits.(+0.5)
$
hy "f)0 gy gg Q @
Final State 1 21eMVe ( M 57,a n d 600 KVA/,(+0.25)
,
REFERENCE 1.
NMP-2 N2-IOP-68, Main Cen.,
p.5 and Figure 3 Provide Potre r Factor Chart i
s
.
b l
l
.
.
. -.
. _. _
__ __
- *4'.
PROCEDURES - NORMAL, ABNORMAL. EtlERGENCY_ AND PAGE
RADIOLOGICAL CONTROL ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY AN1VER 4.03 (3.00)
a.
By visually observing that all IRMs are above downscale before any SRM count rate is above 10E+5 cps with the SRMs fully inserted.(+1.0)
b.
No (+0.25),
setpoint is 15% power (+0.25)
(+0.5 TOTAL)
c.
1.
N e u 'o r o n count rate increasing at a logarithmic rate (+0.5)
2.
No contr'o! m o v'e m e n t (+0.5)
3.
A stable positive period.
(+0.5)
(+1.5 TOTAL)
REFERENCE 1.
N2-IOP-101A, Plant Startup, pp.
8, 13.
+
ANSVER 4.04 (1.001 Reactor cooldown rate is controlled by the RHS heat exchanger level (+0.5).
'If the level is reduced, more heat exchanger tubes are1 exposed, and the condensing of reactor steam increases (+0.25)
If the level in the Hu is increased, the condensing rate decreases (+0.25)
(i.e.,
cooldown rate decreases)
(+1.0 TOTAL)
REFERENCE 1.
N2-OP-31, Residual Heat Removal, H.4, Steam condensing Mode.
>
.
e
,
...,__w.4 e
" -
"
a 4'.
PROCEDURES - NORMAL. ABNORMAL. EMERCENCY AND PACE
RADIOLOGICAL _ CONTROL ANSVERS -- NINE MILE POINT 2
- 8 5 /12 /10-C. A'.
SLY ANSVER 4.05 (1.00)
Any four (4) of the following:
(+0.25 each, max.
+1.0)
1.
Xenon concentration 2.
Moderator temperature 3.
Control Rod position (axial)
4.
Order of Rod Withdrawal
"
C::: 5:; r -- :
-
REFERENCE 1.
N2-OP-101A, Plant Startup, Precautions, pp.
2, 3.
.
ANSVER 4.06 (1.50)
a.
High neutron flux alarm and/or scram (+0.5)
b.
At or near 100% rod line, minimum recirc. flow (+0.5)
c.
Insert control rods per Reactor Analyst or increase recir-eu!ation flow (+0.5)
REFERENC'E 1.
N2-OP, Neutron Monitoring, Precautions and Off-Normal Pro-cedures.
,'
e
.,
_,,,. _ _,,,, -
- - - -
- 4.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PACE
.
RADIOLOGICAL CONTROL ANSVERS NINE MILE POINT 2-85/12/10-G.A. SLY
--
ANSVER 4.07 (2.00)
'
a.
A coupling check is the application of a continuous withdraw signal with the control rod full out to check coupling mechanisms by observing the following:
1.
red rod " full-out" light remains on (+0.25)
2.
rod overtravel annunciator does not come in (+0.25)
3.
drive water (Iow decreases to " stall flow" (+0.25)
4.
rod remains at position 48.
(+0.25)
'
(+1.0 TOTAL)
b.
1.
Reduce Recirc Flow to minimum (+0.25)
2.
Scram the Plant (+0.25)
n 3.
Follow Procedures (+0.25)
4.
Notify SSS (+0.25)
(+1.0 TOTAL)
REFERENCE
N2-IOP-30, CRD, pp. 21, 22, 30.
'
2.
NMP-2 Exam Bank Cat.
4, CRD, No.
9 (Part A).
ANSVER 4.08 (2.00)
a.
TRtfr (+0.5)
"E N'
~
b.
P A I. R (+0.5) h c.
TRUE (+0.5)
d.
TfNte (+0.5)
M S-(.
REFERENCE 1.
NMP-2, Operation Technology, Module VI, Part 6,
RMCS, pp.
9, 11, Student Learning Objective No.
5.
2.
N2-IOP-95A, pp.
2, 4.
.
.--
-
- - - -, -
,,
, - - -,,,.
., _
- -, _ _. _, - -
,... - - -,
..p_,
_ _ _ - - _ _ _ _ - - _
.
.
PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE
- 4.
RADIOLOGICAL' CONTROL ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 4.09 (3.00)
a.
2 ren.
(+0.5)
due to 5(N-18) limit (+0.5)
'
b.
No (+0.5),
exceeds 5(N-18) limit ( + 0. 5 )-O# Ne # # 8" #. ' fg,,,
c.
No (+0.5),
administrative limits state you can receive 100 mrem per week.
(+0.5)
REFERENCE 1.
NMP-2, S-RP-1, Access and Radiological Control, pp.
1, 12,
2.
NMP-1, EPP-15, Health Pl.ysics Procedure, p.
3.
.
ANSVER 4.10 (2.50)
a.
Do not secure or place an ECCS in MANUAL mode unless, by at least two indepencent indications (+0.5),
1.
misoperation in AUTOMATIC mode is conft.med (+0.5) or
+
2.
ad&qu' ate core cooling is assured.
(+0.5)
(+1.5 TOTAL)
b.
If an ECCS is placed in MANUAL mode, it will not i nitiate automatically.
Make frequent checks of the initiating or controlling parameter (+0.5).
When manual operation is no longer required, restore the system to AUTOMATIC / STANDBY mode if possible (+0.5).
(+1.0 TOTAL)
,
REFERENCE i
1.
NM P - 2, N2-EOP-RL, RPV Vater Level Control, p.
3.
.*
.
.
e
.
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_ _ _ _ _ _
_
_ _ _
. _ _ _ _ _ _ _ _ _ _ _ _ _
- 4.
PROCEDURES NORMAL. ABNORMAL. EMERGENCY AND PAGE
.
-
RADIOLOGICAL CONTROL ANIVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY
,
AN!VER 4.11 (2.50)
a.
1.
Steam dome space to bottom drain less than or equal to 145 deg F (+0.5)
I 2.
Idle loop to operating loop less than or equal to 50 deg F (+0.5)
b. do-- dx s sb k.F 3.
Idle loop to apsreti..;
'aap less than or equal to 5 0 W p+-
--
- --
(+0.5)
(+1.5 TOTAL)
b.
N (+0.5),
required to shutdown (+0.25) due to ECCS performance criteria (i.e.,
(Iow imbalance, etc.)
(+0.25)
(+1.0 TOTAL)
(
it. 5(p (0Q f,. 5 h c(py-4
.
REFERENCE
Tech. Spec.
3/4.4.1.4, pp.
3/4 4-4 and B.3/4 4-1.
- e ANSWER 4.12 (1.00)
RPV water level less than 159.3 in.
2.
RPV pressure greater than 1037 psig 3.
Drywell pressure greater than 1.68 psig Ah MSN 4.
A condition that requires isolation g
5.
A condition that requires an Rx scram, AND Rs power is i
above 4% or cannot be determined.
(+0.2'each, max.
+1.0)
REFERENCE
NMP-2, N2-EOP-RL, RPV Vater Level Control, p.
1, Student Learning Objective No.
2.
l
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-
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.
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l
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TEOT CRO2S REFERENCE PAGE
QUESTION VALUE REFERENCE
________
______
__________
01.01 2.00 SLY 0000004 01.02 1.00 SLY 0000005 01.03 1.50 SLY 0000006 01.04 1.50 L'LY0000007 e
01.05 2.50 SLY 0000009 01.06 2.50 SLY 0000010 01.07 2.50 SLY 0000011
-
01.08 1.50 SLY 0000012 01.09 2.00 SLY 0000013 01.10 2.00 SLY 0000014 01.11 2.00 SLY 0000001 01.12 1.00 SLY 0000002 01.13 3.00 SLY 0000003
_____4 25.00 02.01 2.50 SLY 0000054 02.02 2.50 SLY 0000055
'
02.03 2.50 SLY 0000057 02.04 1.00 SLY 0000058 02.05 2.00 SLY 0000059 02.06 2.00 SLY 0000061 02.07 2.50 SLY 0000062 02.08 2.50 SLY 0000063 r
02.09 1.50 SLY 0000064 02.10 2.00 SLY 0000113 02.11 2.00 SLY 0000114 02.12 2.00 SLYOOOO115
______
25.00 03.01 3.00 SLY 0000039 03.02
- 2.00 SLY 0000040 03.03 2.00 SLY 0000041 03.04 2.50 SLY 0000042 03.05 2.00 SLiOOOOO43 03.06 3.00 SLY 0000044 03.07 1.00 SLY 0000045 03.08 3.00 SLY 0000047
,,
03.09 0.00 SLY 0000048 03.10 2.00 SLY 0000049 03.11 1.50 SLY 0000050
_____.
25.00 04.01 2.50 SLY 0000056 04.02 3.00 SLY 0000065 04.03 3.00 SLY 0000066 04.04 1.00 SLY 0000067 04.05 1.00 SLY 0000069
.
e
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-, - - - -
-
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- -
e f.'.
=
-
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TEST CROSS REFERENCE PAGE
13UESTION VALUE REFERENCE
,________
______
__________
04.06 1.50 SLY 0000070 04.07 2.00-SLY 0000071 04.08 2.00 SLY 0000072 04.09 3.00 SLY 0000073
"J 0 4 '. 1 0.
2.50 SLY 0000101
04.11 2.50 SLY 0000102
'04.12 1.00 SLY 0000104
______
25.00
______
______
100.00 mr
%
.
<M MF he f ";
.'
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d
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r -,
- - - -
-,.,.,,,,,, -. _ _, - _ - - - - - -_
, _
,,,,,
w-
,
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- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -.
,
$1TOCbmenf A
..
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U.
S.
NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
NINE MILE POINT 2 REACTOR TYPE:
BVR-GE5 DATE ADMINISTERED: 85/12/10 EXAMINER:
G.A.
SLY APPLICANT:
N N
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
Staple question sheet onetop 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 wi!! be picked up six (6)
hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY
_.
VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5.
THEORY OF NUCLEAR POVER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 25.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 25.00 7.
PROCEDURES - NORMAL, ABNORMAL,
' *
EMERGENCY AND RADIOLOGICAL CONTROL 25.00 25.00 8.
ADMINISTRATIVE PROCEDURES,
,
CONDITIONS, AND LIMITATIONS
- -
100.00 100.00 TOTALS
.
FINAL GRADE
%
AII work done on this examination is my own. I'have n e 'i t h.e r givon nor received aid.
APPLICANT'S SIGNATURE
'\\
J
'
,
3.
THEORY OF NUCLEAR POVER PLANT OPERATION, FLUIDS, AND FAGE
THERMODYNAMICS
.
.
QUESTION 5.01 (2.00)
i ANSVER if the fo!!owing Sm-149 statements are TRUE or FALSE?
12,4+
a.
It is REMOVED from an operating reactor by burnout and
,,
radioactive decay.
'"
l b.
WHEN a reactor is resta.ted after a temporary shutdown, Sm-149 concentration inc reases for several days.
a3/
c.
It has LESS etfect on reactor operation than Xe-135 due to its smaller fission yield and s m a '. g e r microscopic neutron
_ i cross section.
'#
d.
The equilibrium concentration of Sm-149 at 50% FP is about TVO-THIRDS that of the equilibrium concentration at 100% FP.
' F O,/
QUESTION 5.02 (2.00)
STATE whether the fo!!owing situations would (INCREASE, DECREASE or NOT CHANGE) control rod worth, a.
Restart 10 hr following a scram from 100% power condition (peripheral rod only)
(0.5)
7y,)'.
b.
Second -.rod in a rod group following the withdrawal of the
l, -. b5 first rod in that group.
(0.5).
, Change from a cruciform shaped rod to a cylindrical rod
,c.
,,,
ic(
',
of the same volume.
(0.5)
'
-
.-
b' '
d.
Localized voiding of region not previously voided.
(0.5)
OUESTION 5.03 (2.00)
-
.
The reactor is critical at 10E+6 cps.
A stable period of 60 M'c y
seconds is achieved.
If rods are inserted. continuously until
~j.C
'/
the period drops to infinity and then the rod insertion is immediately stopped. WILL the reactor be (critical, supercrittcal, j
or suberitical) in the time following the rod stoppage?
EXPLA!!!
(2.0)
i (*****
CATEGORY 05 CONTINUED -(M4 NEXT PAGE
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5.
' THEORY OF NUCLEAR POWER PLANT OPEFATION. FLUIDS. AND PACE
THERMODYNAMICS
,
,
QUESTIO'N 5.04 (2.50)
You are the SRO in ch'arge of the initial fuel loading process.
As part of your duties, Operations has asked you to verify the STAS prediction to criticality as fuel is being loaded, a.
From the following information, PREDICT the point of cri-ticality after the 6th fuel bundle.
(1.0)
Count Rate, (cps)
1/M Value S,
CRo 100 1.00
=
F1, CR1
= 100 1.00 F2, CR2 102 0.??
=
'F3, CR3 105 0.95
=
F4, CR4 110 0.91
=
F5, CR5 113 0.87
=
F6, CR6 125 0.80
=
F7, CR7 149 0.67
,
=
F8, CR8 200 0.50
=
F9, CR9
~
500 0.20
=
M-b.
HOV MANY fuel bundles may be loaded following the 6th fuel fuel bundle, prior to being required by ANS/ ANSI Standards to make another criticality calculation?
(ANS/ ANSI Standards state that
..the maximum fuel load increment is the greater
"
of one fuel assembly, or one-half the additional bundles which 7SO)
are predicted to p*oduce criticality.")
f
,
c.'
DO the initial six (6) fuel bundles of the 1/M plot indicate that fuel is being loaded (IDEALLY, AVAY FROM the detector or TOVARDS the detector)?
(0,5)
.
.
.
,#
(*****
CATECORY 05 CONTINUED ON NEXT PAGE *****)
-.
._.
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T 5.
'T H E O R Y OF NUCLEAR POWER. PLANT OPERATION. FLUIDS. AND PAGE
THERMODYNAMICS
,
,
.
.
.
.
QUESTION 5.05 (2.50)
Given a large vented tank 30 ft in diameter and 60-ft high w'ith a centrifugal pump taking a suction from its base.
The pump is located at a vertical elevation corresponding to the bottom of the tank and it requires 5 ft of net positive suction head (NPSH) to prevent cavitation.
The tank is entirely full of water and is maintained at 60 deg F by heaters.
The tank is designed such that it could withstand 15 psi differential pres-sure in e i t h e,r direction.
Assume the vent becomes totally
,
clogged while the pump is in operation.
ANSVER the following questions.
e a.
VHAT is the lowest pressure t,h a t the tank will drop to as
.the pump continues to remove water from the tank?
(0.5)
b.
VILL the pump loose NPSH and begin to cavitate prior to reaching a level of 5 ft in the tank?
EXPLAIN.
(State any assumptions.)
(1.0)
c.
COULD the pump continue to pump water at a level below 5 ft without cavitation if the vent were open?
EXPLAIN.
(Assume no vorteming.)
(1.0)
OUESTION 5.06 (2.00)
Given.the following two (2) conditions and using the supplied information, DETERMINE which condition is operating.MORE CLOSELY to its MCPR limit.
(Show all work and state any assumptions.)
' ' ' '
'
K-f graph is provided.
(2.0)
Condition 1 Condition 2 R.x dome pressure =
950 psig Ra dome pressure 980 psig
=
Core flow = 54.25 M1b/hr Core flow
81 M1bfhr Rx power
1660 MW Rx power 2490 MV
=
=
=
.
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE
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,
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_____
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE
THERMODYNAMICS
..
QUESTION 5.07 (2.00)
Vater enters the regenerative heat exchanger from the reactor at 538 deg F and exits to the NRHX at 233 btullbm.
a.
If water exists the demineralizers at 120 deg F,
VHAT is the temperature of t,h e water returning to the reactor?
Show all work and state all assumptions.
(1.25)
b.
If a (10%) leak developed downstream of the domineraliser, VHAT would be the temperature of the water returning to the reactor?
( 0. 7 5 )'
<
QUESTION 5.08 (2.00)
Vh tl e Nine Mile Pt-2 is operating at 90%, extraction steam to the highest pressure feedwater heater is remov,ed.
An engineer observed that the turbine load increased by 20.MV electric and concluded that this action has improved (increased). the plant's thermodynamic effic.iency (not heat rate)
IS this concidsion correct?
EXPLAIN your answer fully.
.-
(INCLUDE VHAT caused' electrical output to increase.)
(2.0)
GUESTION 5.09 (3.00)
You are currently operating at 100% power BOL when you lose partial
feedwater h e a t 'i n g :
j a.
If the STA tells.you that feedwater temperature decreased by H
10.deg F, voids decreased by 2%, VHAT would be.the corres-l ponding temperature change to the fuel temperature.
(Assume
j no rod movement, recirculation flow changes and the reactor reactivity returns to zero.)
(1.5)
b.
If the same situation were to occur at EOL VHAT would be the corresponding reactivity changes (MORE NEGATIVE, LESS NEGATIVE, NO CHANGE) to each of the above coefficients
(i.e.
delta (T) mod, delta (% voids), delta (T) fuel)?
(1.5)
.
(*****
CATECORY 35 CONTINUED ON NEXT PAGE
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.
aI
.
5.
'TREORY OF NUCLEAR POVER PLANT OPERATION, FLUIDS. AND PAGE
THERMODYNAMICS
,
.
QUESTION 5.10 (3.00)
As the reactor is taken from COLD SHUTDOVN to RATED OPERATING CONDITIONS, HOV are the following affected and VHY?
a.
The MAGNITUDE of the MODERATOR TEMPERATURE COEFFICIENT.
(1.0)
b.
DIFFERENTIAL CONTROL ROD WORTH.
(1.0)
c.
The MAGNITUDE of the FUEL TEMPERATURE COEFFICIENT (Doppler)
(1.0)
QUESTION 5.11 (2.00)
Three (3) minutes following a scram from 100% power, reactor power is 75 on IRM Range 4 and decreasing.
WHAT will the indicat,ed power be one (1) minute later?
SHOV calculation and EXPLAIN any assumpt' ions made.
(2.0)
-n
~
.'
(*****
END OF CATEGORY 05
- )
'
' PLANT SYSTEMS DESICN. CONTROL. AND INSTRUMENTATION PAGE
.
.
QUESTION 6.01 (3.00)
EXPLAIN,VHAT affect the following failures would have on reactor
, level.
VHY?
(Assume 3-element control and Channel A cantrol-ling.)
, '. _'
^$
_
a.
'C'
steam line flow signal fails low.
[ pang]
b.
Channel
'A'
reactor level detector signal fails low.
- 7d mo't o r. -T'o FC.1/, 4onJ)
c.
'A'
pump servo i
d.
. Inadvertent activation of the setpoint setdown circuitry.
(hh,7fj QUESTION 6.02 (2.50)
Concerning the Safety Parameter Display System:
a.
VHAT are the available level one (1) display (s)?
(0.5)
b.
VHAT are the available safety function blocks and VHAT
- "
parameters are used to determine the safety function status?
(2.0)
QUESTION 6.03 (2.00)
The reactor is at 100% power with the generator synced to the grid.
Electrohydraulic Control (EHC) load set is 105%.
By using the attached EHC diagram, EXPLAIN VHAT would happen (con-trol valve, bypass valve) in the following circumstances:
a.
load limit potentiometer reduced to 95%.
(0.5)
b.
maximum combined flow limit potentiometer reduced to 95%.
(0.5)
Je% &c c.
"A" pressure regulatory f r a + p a i n t) fails low.
(0.5)
d.
failure of two (2) bypass valves full open.
(0.5)
l
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CATECORY 06' CONTINUED ON NEXT PAGE
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DESIGN, CONTROL, AND INSTRUMENTATION PAGE
-
.
.
6.
- PLANT SYSTEMS
..
.
.
QUECTION 6.04 (1.00)
'
VHAT two (2) ep$bEdi conditions will cause the RSCS to apply a
.
.
_____
(1.0)
rod block to a control rod?
.
QUESTION 6.05 (2.00)
f ANSVER the following questions concerning the Standby Liquid C'ontrol (SLS) System:
a.
The minimum concentration needed to shutdown the reactor from rated conditions is ppm in a minimum (0,5)
of minutes.
b.
VHAT is the purpose (s) of the interface between the Instrument (1.0)
Air Systems with the SLS system.
c.
The auto start feature is interrupted by either a loss of (0.5)
offsite power or a.LOCA (TRUE or FALSE.)
t QUESTION 6.06 (2.00>
Concerning the CRD Hydraulic System:
a.
The reactor operator is going to increase drive pressure to the HCU.
VOULD you as the acting SRO direct him to OPEN or CLOSE the drive water pressure control valve?
(0.5)
,
b.
EXPLAIN HOV your actt'on in part has changed the following flow rates (INCREASE, DECREASE, NO CHANGE)
(1.5)
.
1.
scram valve charging flow 2.
CRD total system flow 3.
cooling flow i
.
(*****~ CATEGORY 06 CONTINUED ON NEXT PAGE *****)
l
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.
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PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE
.
'
QUESTION 6.07 (2.00)
ANSVER the following questions based upon the situation described below.
The RRCS is fully operational The RRCS receives a reactor water low level (105 inches) signal in both complementary logics of a RRCS channel and remains in f o r j!(J) [ s e c o n d s.
It takes 100 seconds from the initial reactor water low level signal before the APRM level is downscale.
a.
Which of the four logics integrated into RRCS are actuated at T 0 seconds?
(0.5)
=
b.
Which logics are actuated at T 25 seconds?
(0.5)
=
c.
Which logics are actuated at T 98 seconds?
(0.5)
=
d.
How long from T 0 seconds is it before the RRCS can be
=
reset?
(0.5)
QUESTION 6.08 (1.50)
The Generator Gas Control System provides the main generator with hydrogen to cool' the rotor windings and internal compo-nents.
For this system, three (3) parameters ~ot.information (purity, pressure, and temperature) are available in the Control Room concerning generator hydrogen, i
a.
HOV DOES each offect generator cooling capability if deviated from normal 100% power operations?
(Assume purity and pressure to decrease and temperature to increase.)
(0.75)
b.
You are in the process of purging the main generator with carbon dioxide.
STATE HOV the following failure would effect this operation (AUTO ISOLATE or NO EFFECT)
(0.75)
pipe-failure at the exit of the electric vaporizer heater 2.
Iow level in Storage Tank (TKi)
3.
Iow generator gas pressure ((2 psig)
(*****
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PLANT SYSTEMS' DESIGN', CONTROL, AND INSTRUMENTATION PAGE
!
l
!
i QUESTION 6.09 (2.50)
Both the SRM and IRM compensate their detector signals with a unique type of discrimination process.
Briefly DESCRIBE HOV each system, SRM/lRM, accomplish this a.
task.
(1.5)
b.
VHY is there a difference between the two (2) discrimina-tion processes?
(1.0)
OUESTION 6.10 (2.00)
An automatic HPCS initiation has occurred, Subsequently HPCS injection was automatically terminated due to high reactor water level
__
a.
VHAT component in the HPCS system functioned to terminate the injection?
(0.5)
.. -
b.
Assuming no operator action, HOV VILL HPCS rnspond to a subsequent decreasing water level?
(0.5)
c.
WHAT would be the response to decreasing water level if HPCS injection has been terminated mannually,by closing the injection valve?
(0,5)
d.
If the HPCS system had switched sources from the' CST to the supression pool due to low CST level and the CST. level j
had subsequently recovered, WILL the system automatically
'
switch back to the CST suction?
(0.5)
(*****
CATEGORY 06 CONTINUED ON NEXT PAGE
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.
6'.
' PLANT SYSTEMS DESICN, CONTROL, AND INSTRUMENTATION t
QUESTION 6.11 (2.50)
a.
WHAT are the differences in modes of operation (System Cooling Line-ups) for the RHS Loops A and B?
b.
VHAT is the reason for the interlock between the (
a 1.
shutdown cooling suction valve and the test retur, valve?
2.
pressure control valve bypass valve (MOV-23A) and F. x pressure?
~
c.
If a LPCI auto initiation function (high drywell) were over-ridden to realign the RHS system to the shutdown cooling mode and another LPCI signal (Iow level) were to come in, VOULD the RHS loop realign from the shutdown cooling mode te the (1.0)
LPCI mode?
EXPLAIN.
-
.a GUESTION 6.12 (1.00)
,
The plant is operating at 100% power.
APRM channels A and C
>.
have failed high.
You call the I&C Technician to investig-te A Plant Auxiliary Operator wants to shift RPS B power supply to its alternate power. source for training.
Vould you let him?
EXPLAIN VHY or VHY not.
Direct your answer toward system (s)
responses instead of administrative requirements.
(1.0)
QUESTION 6.13 (1.00)
VHAT condition (s) do the vacuum relief lines between the drywell and suppression chamber limit / protect against?
(1,0)
(*****
END OF CATEGORY 06
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7.
PR'dCEDURbS - NORMAL. ABNORMAL. EMERGENT
'ID PAGE
RADIOLOG7 CAL CONTROL QUESTION 7.01 (2.00)
VHY are each of the following Reactor reulation System precautions necessary (i.e.,
what do t
prevent or ensure when observed)?
a.
An idle recirculation. loop shall r.
re started unless the temperature differential between
eactor pressure vessel steam space coolant and tho tom head drain line coolant is less than or equal to :
feq F.
and:
(0.5)
b.
When both loops have.been idle, ur w
the temperature differential between the reactor c 5. n t within the idle loop to be started up and the cool in the reactor pressure vessel is less than or eq-to 50 deg F,
or (0.5)
c.
When only one loop k3s.been idle, rss the temperature differential boaween the reactor <
int within the idle
-
and operating recirculation loops
- ess than or equal to 50 deg F and the operating loop f!
rate is 'ess than or equal to SO% of rated loop tiew.
(0.5)
d.
TRUE or
.f'A L S E :
The operator is al-d two (2) recircu-lation motor starts from ambient t-rature with'a required 45-minute delay between-tart (0.5)
>
m:
.-
QUESTION 7.02 (2.00)
Assume a loss of Station Air has occur-a.
VHAT three (3) automatic acttons st td be verifted as having occurred if STATION AIR HEA'
pressure is observed to be at 82 psig?
(Setpoint requ2
)
(1.5)
b.
Under VHAT circumstances de the I mt te Operator Actions a
for a Lo*45 of St. tion and/or Contri
.r require the reactor to be manually scrammed?
(0.5)
i (*****
CATECORY 07 CONTINUED or NEXT PAGE
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)
.
,-
PAGE
,
.
AND EMERGENCY ABNORMAL,
'RQGEDURES NQRM AL,
CONTROL 1ADIOLOGICAL
-
(2.00)
7.03 (RVP)'
(0.5)
3T10N Permits standard RVP?
Radiation Vork a
verses Concorning issued (0 5)
RVP be RVP?
extended extended an of an WHEN would issuance length of a.
maximum the CRD VHAT is the recharge work force said that a to b.
has states Radiation protection Maintenance a
assign You are to in the area.
c.
exists dose 25 mrem /hr should take that the job or employee, (1.0)
with one
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> employees and VHY?
with two work
3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> f o,x the choose m
would you force VHICH work and (1.50)
rad 14 tion 7.04 concerning employee with an OUESTION question 20-year-old following for a the Answer control-of 8 rem.
radiological occupational federal dose limit for 40.5)
accumulated maximum employees be the VHAT would action 41.01 quarter?
life saving a.
a the for eligible EXPLAIN.
be limits?
employee federal COULD this any b.
violate and not has Dsrector (2.50)
Emergency may NOT be 7.05 the QUESTION Emergency authorities conditions.
(2.51 Plan, that List to the Site and/or emergency Accordingresponsibilities during ities.
subordinate certain delegated rekponsibilities/ author to a A
(5)
five these i
k l
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CATEGORY (*****
.
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--.
.
PRdCEDURES - NORMAL, ABNORMAL. EMERGENCY ANC PAGE
- RADIOLOGICAL CONTROL
.
l QUESTION 7.06 (2.00)
Following a required initiation of the S t a r ib y Liqu
'ntrol System you are directed by the level / power
- ntrol-dure to-
>
a
"Lewer RPV water level by terminating and prever
injection except from CRD and Boron snjection sy ms until either:
VHAT is the purpose for cwering water
"
level at this time?
(1.0)
d.
You are also directed to an7ect boron prior to e euppression p b o l. temperature reaching 110 deg F.
VHAT is tF eason for bcron infection pract to reaching t h ir, t e mp e r a t e-Limit 2 (1.0)
,
f i
QUESTIDM 7.07 (1.50)
Whale attempting to line up shutdown co01tnc mode et MR, you: reactor operator informs you that the ruction from the recarculation line is frozen in the closed tion, hecording to the Alternate Shutdown Cooling Procedu:
.(N2-EOP+C5),
in general, VNAT would be the titernate 2t
removei f16w path for eerforming the s h u t e m r.
cool-acti (1,5)
i I
i
'
QUESTION 7.06 (2.50)
Concerning the blewdown and r.e c i r c u l a t t e n t h o t shutdo modes
!
of the Reactor Vater C l e a n'u p (RVQU) System-i j
a.
The operator is cautioned to pIdee the EVCU sys Jato e
bicwdown mode prior-to starttng the CRD oumps.
,T 4s
-
the r.e a s o n fer this cautien?
(0.5)
b Vben operating in the bluNdown mode VHY s h o u l d r.
s j
diveit a'l the kVCU ilow ts Liquid Fad Vaste
<p r
!
Maan Condenser?
st.0)
A c
L'H E N and VHY would the hot st.utdown mode of thc syctem
<
most likely be used?
(1.0)
a N
(*****
CATEGORY 07 CON 1 1 Ht) C D Oil NEXT PAGE
- - ->
i
.
.e w,
e
.n4..
~4,p
,
.-
g
-, -
ww g-g
,g g g g 4ew
, -
-n~v
_
._ _
-
.
24'
PR"O'CEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE IS RADIOLOGICAL CONTROL QUESTION 7.09 (3.001 USE the attached figures from N2-EOP-SPL to ANSVER the fo!!owing questions:
a.
DETERMINE the minimum suppression pool level given a RPV pressure of 700 psig and suppression pool temperature of
160 deg F?
(1.0)
b.
WHAT is the basis for the Heat Capacity Level Lamit curve and VHICH area is the safe area of operation?
(Above or below the line.)
(1 0)
c.
EXPLAIN VHAT would happen if drywell spray were initiated I
above the Drywell Spray Initiation Pressure Limit?
(1.0)
l l
l QUESTION 7.10 (2.00)
Procedure N2-EOP-SPT (Suppression Pool Temperature Control)
directs the operator to " runback recire. and manually scram" 11 an SRV has been s t uck open and cannot be closed.
a.
VHY is recire. runback prior to reactor scram?
(0.5, b.
Following the reactor scram you are required to depressurize the reactor, if the suppression pool temperature cannot be maintained within the Heat Capacity Temperature Limit You are also cautioned not to "depressurize the RPV below 60 psig unless motor driven pumps sufficient to maintain RPV water level are running and available."
VHAT is the basis for the caution and VHAT system / components does it spocifically address?
(1,5)
l
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CATEGORY 07 CONTINUED ON NEXT PACE
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7..
PROCEDURES
' NORMAL, ABNORMAL, EMERGENCY AND PAGE
-
l nADIOLOGICAL CONTROL
l QUESTION 7.11 (2.00)
l ANSVER the following question concerning the main generator and
"
load changes.
USE the attached Power Factor Chart You are operating at a 0.95 laggang power factor with 75 pstg
!
H2 and the load dtspatcher orders you to drop your power factor
.
to a 0.9 lagging power factor but maintain maximum MVe output.
!
In general, HOV would you change your operating conditton?
{
(INCLUDE in your answer the initial conditicns (MVe, KVA),
a brief discussion of the power change, and the final conditions
!
(MVe, KVA).
(2.0)
!
,
QUESTION 7.12 (2.00)
'
l According to N2-ICP-21, Main Turbine, there are several pre-
'
!
cautions and time l i mi t a t i o n s associated with turbine startup to assure proper operation, warmup, and to preclude damage from excessive vibration.
e a.
VHY should shell warming'begin as soon as possible after steam seals are establi=hed, and VHAT might result if she!!
warming is e x -: e s s i v e. y delayed?
(1.0)
.,
b.
VHAT might occur if first stage pressure exceeds 90 psig
,
I during sheII warming?
I!.0)
I
,
,
!
+
!
r i
- a
'
s a
,
(*****
END OF CATEGORY 07
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l
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i
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- ~
-
-
-r-,
,
v
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r
,e r
-,-c,y
,-
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~-
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.-.-<m,---
,s
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aw.-
-sn,,
w
-
..
.-
.
._
.
...
...
.
..
O ADMIMISTRATIVE PROCEDURES. CONDITIONS. AND Lill!TATIONS PAGE
.
.
QUESTION
- 01 (2.00)
CONSIDER the following situations.
.
a
'According to Technical Specifications tr IT PERMISSIBLE to
go from startup to run if IRMs A,
B, and C are inoper=ble?
E1? LAIN.
- 1.0)
i b.
If the same IRMs were found inoperable chtte in run, WOULD w.a violate any Technical Specifications by:
f
Staying sn Run?
EXPLAIN.
f3.5)
Placing the mode switch in Startep?
EXPLAIN.
to.5)
i QUESTION 0 02 (3.00)
,
E
$
The Division 1 Diesel is operating and is 30 minutes into a surveillance test when the air starting system fails.
The
-
main +.enance repair team estimates a 2-day minimum repair time.
(USE the attached Tech. Spec. to explain your answers.)
a.
- S the Diesel Generator inoperabte according to Tech.
3pec.?
EXPLAIN.
(1.0)
,
b.
ARE all the Division 1 ECCS systems inoperable because ot
'he Diesel Generator problem?
EXPLAIM, (1.0)
,
t 6'
,
c.
It at the same time the Division 2 o +4py s p ra y pump is out
'
c( service, VhAT added implications does. this_have on your Tech. Spec. position?
(1.0)
,
'
QUESTION 8.03 (2.00)
The reactor operator is performing a surveillance of the Standby Ligntd control System and due to system modificaticns a procedural step becomes impossible to perform.
a.
Under this condition CAN a temporary change be issued?
(0.5)
b.
WHAT three (3) " key points" must be adhered to when issuing a temporary charge?
(1.5)
,
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CATECORY 08 CONTINUED ON NEXT PAGE
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l'
ADMINISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIOda, PAGE
,
-
-
.
.
QUESTION 8.'04 (3.00)
LIST the Nine Mile Pt. 2 Tech. Spec. Safety Limits required in Operational Condition 1.
(Setpoints required.)
(3.0)
,
QUESTION 8.05 (2.50)
A weekly surveillance, normally performed on Friday, was performed
,
on the (o11owing days due to manpower limitations over the Thanks-giving Holiday.
_,
,
Friday - November 22 Vednesday November 27 (5 days from last sury.)
~
-
Thursday - December 5 (8 days from last surv.)
Friday - December 13 (8 days from last sury.)
.
a.
HAVE the surveillance requirements been exceeded for this set of dates (YESIND)?
EXPLAIN your answer.
(1.5)
b.
VHEN is the maximum allowable date that the next surveillance
'
can legally be performed?
(INCLUDE HOV you determined this date.)
(1.0)
QUESTION 8.06 (2.50)
.
Vith the reactor p2 Ant in mode 1,
it is determined that four (4)
gallons per minute are being collected by the Drywell floor drain system.
Also, the Drywell equipment drain system indicates 22 gallons per minute (steady) collection rate, a.
VHAT are the maximum a!!owable plant leakage limits?
(1.5)
.'
b.
STATE the actions required by Tech. Spec. for the above condition (if any).
(1.0)
.
.
(***** CATECORY 08 CONTINUED ON NEXT PAGE
- 1
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-
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4.._ADMIdISTRATIVE PROC EDORES.. CDNDITI'GMB, AND LIMITATIONA PAGE
..l
-
QUESTION 8.07 (2.50)
Concerning shift compi ment and shitt t.urnovers:
a.
.HOV MANY SRO, RO, STA are tsquired in operation condition 17 (1.5)
b.
Vou come on s h i t't and find that one (1) RO has not reported I
for duty.
CAN yuur crew accept shift responsibilities in this condition?
VHY or VHY NOT7 (0.75)
QUESTION 8.08 (1.50)
Concerning the APRM setpoints for power distribution limits:
a.
CALCULATE the scram trip setpoint(s)
f the reactor is operating at 3 0 0 0 tSJ TH wi t h mo s t Ilmiting LHGR mode operating at 10 KV/ft.
ASSUME an LHGR limit of 13.4 KV/tt.
(C.75)
b.
DOES this result require any APRM adjustment?
VHV or i
VHY NOT?
(0.75)
er f
QUESTION 8.09 (1.50)
'
';
Technical Spe'cification 3.7.1.i requires two plant servies water i
pumps per loop t o '.b e operable and provides explicit action ge #[..
requirements if one service water pump par loop is inoperable.
j,
,
,
if both of the service watet pumps per loop were to become 6'M '
inoperabic, no sp.ecific action statement would apply.
,.
a.
VHAT would be your required action?
6.7[~(1.0)
.
b.
HOV SHOULD an operator intetpret tech specs in this
. instance and in other 'similst instances not directly provided
,,
for in the action statements to insure the i n t eit t of the
/
T specifications are met?
y("
(1.5)
i
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CATZCORY 08 CONTINUED ON NEXT PkCE
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8.
ADMfMISTRATIVE-PROCEDURES. CONDITIONS, AND LIMITATIONS PAGE
.
-
.-
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l
QUESTION 8.10 (2.00)
l Using the attached Technical Specifications, DETERMINE the maximum i
time that the reactor.may continue operation given the following aalfunctions. Reference the secticns of tech spnes used in deterkining your answer.
OM/(I a.
It is discovered that valve F048A (RHR A " H e a t Exchanger Bypass) is failed open and cannot be closed.
(1.0)
b.
Susequent to the malfunction in (a) above, it is found that RHR pump E is inoperable.
(1.0)
GUESTION 8.11 (2.50)
The RCIC outboard isolation valve (21CS-MOV121) motor controller has failed in the deanergied position and the valve won't shut Maintenance is cur.rently attending to the problem.
By using the
'Ittached Technical Spectfications:
<<<
a.
STATE which Tech. Specs. apply to this problem.
(G.5)
~.
STATE whether RCIC is OPERABLE cr INOPtRABLE and GIVE s
ANY necessary action rt.stement(s) required.
(2.0)
.
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END OF EXAMINATION
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7 i"' THEORY
.
.
.
.
S OF NUCLEAR POVER PLANT OPERATION, FLUIDS, AND PAGE
-
,
THERMODYNAMI.G1 ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANLVER 5.01 (2.00)
'
a.
FALSE (+0.5)
b.
FALSE (+0.5)
c.
TRUE (+0.5)
d.
FALSE (+0.5)
REFERENCE 1.
NMP-2 Operations Technology, Module 1,
Part 15, pp.
I-15-1, I-15-2, Student Learning Objectives No.
2,
,
ANSVER 5.02 (2.00)
i 4.
increase (+0.5)
b.
decrease-(+0.5)
c.
decrease (+0.5)
'
du-u ~-
d.
kne ease (+0.5)
_
s
,
REFERENCE
'"
-
1.
NMP-2 Operations Technology, Module 1,
Part 14, pp.
- -14-9 A
I-14-10, Student Learning Objective No.
4.
'
Q-C,.E. L & 7 M.1 rf. s-isu ANSVER 5.03 (2.00)
Supercritical (+0.5)
When the period reaches infinity, the reactor i s. exactly c r i t ical-on prompt neutrons.
. ( + 0. 5 )
A f't e r the rod insertion stops the. delayed neutron precursors which were formed in previous generations and at a higher power level tend to pull power back up (+0.5)_
Therefore, the reactor is still supercritical due to the latent effect of delayed neutrons (+0.5)
(+2.0 Total)
REFERENCE
NMP-2 Operations Technology, Module 1,
Part 11, p p,.
1-11-6.
'
j i
s
.
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE'
'
THERMODVAAMICS ANSVERS -- NINE MILE' POINT 2-85/12/10-G.A.. SLY i
'
ANSVER 5.04 (2.50)
a.
See Figure 8-11 in Reference material criticality predicted at pin 16 (+0.5 for plot and
+0.5 for usage)
b.
next reading 1/2(16 - 6) + 6 =
after the 11th fuel bundle
=
therefore 5 more fuel bundles may be loaded (j,o)
4 00 desst ) L f a v6z '>
c.
away from the detector ( 0.;f)
REFERENCE
NMP-2 Operations Technology, Module 1,
Part 8,
pp.
1-8-10 and 1-8-13, Figure 8-11, Student Learning Objec'eive No.
4.
.
+
ANSVER 5.05 (2.50)
s.
a.
The lowest pressure that the tank could drop to would be the saturation pressure for 60 deg F which is 0.25$) psia.
(+0.5)
f $4 Lud tr'
Alif35*-
0K
&//d da.O -f vare(w u = 7Js dd A,6i b.
Assuming head loss due to flow 7WETnegligible, the answer is no.
Cavitation would not begin until the 1e'el drops v
below 5 ft in the tank.
(+1.0)
c.
Yes.
he added pressure of 1999' psia at the pump suction would allow all of the water to be removed.
( + 0. 5 ')
REFERENCE
~
1.
NMP-2, SLO for Fluid Stati.cs, Dynamics and Delivery, No. 10, pp.
15, 16.
j.
.
W-
.,I i
E
.\\
'
.
.
.
g e.
= =
,, - -
-%,
-.. - -,~
-- --.=
/
5'.
THEdRY Of NUdLEAR POVER PLANT OPERATION. FLUIDS. AND PAGE
)
THERMODYNAMICS
-
ANSVERS -- NINE MILE POINT 2-85/12/10-C.A. SLY ANSVER 5.06 (2.00)
Assuming 100% core flow is 108.5 Mlb/hr (+0.25),
min MCPR (limit)
1.24 (+0.5)
=
g\\p '
hj For Condition 1:
% core flow =
54.25/108.5 50%
=
from Figure 3.2.3-1 Kf 1.175 (+0.25)
'
=
therefore the MPCR(limit)
1.24(1.175),= 1.457
=
t-tt(,gg3 delta (MCPR)
1.57 - 1.457
=
=
nC:
"i For Condition 2:
% core flow 80/108.5 74.6%
=
=
/, OT Figure 3.2.3-1 Kf = 1-445 (+0.25)
therefore the MCPR(limit)
1.24(1.005),= 1.25
=
delta.(MCPR)
b-TU*
Obs)
1.37 - 1.25 =
=
-
Condition Q,is closer to limits (+0.
)
(0.25 for math)
.
u REFERENCE
NMP-2 Student. Learning Objective for BVR Thermodynamics and Thermal Hydraulic Limits, No.
7, p.
9.
2 '.
' General E l e c t r i'c Thermodynamics, Heat Transfer and Fluid g
Flow, MTC, March 1983, pp.
9-96 to 9-99.
3.
NMP-2 Tech. Specifications 3/4.2.3, Minimum Critical Power Ratio. Figure 3.2.3-1 and Table B2.1.2-2.
}
l
'
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,
,
'
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~ -,
,,
.
-
-.r-
.--
-
.
.
.
.
,3.
' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 5.07 (2.00)
n.
1.
Set up equation
= M delta (h)
A.
MCp delta (T)
delta (h)
B.
Cancel M because both are equal, Cp delta (T)
=
\\
FTSiBlullbm (+0.25)
C.
Lookup h(f) for 538 deg F = logicfNg)
g (+0.5 pts for equation and
-
O
\\
'/ g y p/ g g 39 g
/
2.
Solve for delta (h)
A. f?.t &+4-2 3 3 ) Btullba
=
+0 Btullba B.
delta (h)
'
)
pgg7 3.
Solve for T(hot) to reactor f
delta (h)
T(cold))
Cp (T(hot)
=
Cdelta (h) / Cp3 3 gf.57 83/hw $ M/7/
-
T(hot 11(= T(cold k7
'
S beP S Wi> * 3 '!
"
/
_..
,,, -
b.
Solve for T(hot) to reactor
~
= M1 * deIta (h)
1.
M2 * Cp * (T ( h o t ) -T ( c o l d ) )
= 90% M1 I^IC 2.*42 Cp)]
T(cold) + Cdelta (h) /
(0.9 *
3.
T(hot)
=
464-deg FQO.f)
4.
T(hot). =
.
' f'1
'i l*' '/ 8 REFERENCE
Thermodynamics Lesson Plan, Heat Transfer and Heat Transfer Equipment, p.
13 of 13.
2.NMP-2 Examination Bank Category 1,5, pp.
67, 68, Student Learning Objective No.
3.
ANSVER 5.08 (2.00)
No (+0.25)thermo efficiency is a comparison of Energy In to Energy Out (+0.5)
The increase in output results from no steam being diverted to the high pressure feedwater heater (+0.5)
Because the feedwater is now cooler, more energy from the reactor is required to bring the water up to saturation temperature (+0.5)
thus thermo efficiency is down (+0.25).
(+2.0 Totat)
,.J, f,,
[ q (gg g p
,; ff.,'
REFERENCE
NMP-2 Power Plant Cycles, pp.
5-7, Student Learning Objec-tive No.
4.
2.
Genera 1 Electric Thermodynamics, Heat Transfer, and Fluid Flow, MTC, March 1983, pp.
6-38, 6-66.
,
. -. - -,
.. _
, _
_
_
--
-- -
--
.
51.
THEdRY Oh NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE
THERMODYNAMICS ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 5.09 (3.00)
alpha (d)(delta (T)) fuel alpha (m)(delta (T)) mod + alpha (v) delta (%V)
=
(0.2spts for equation)
alpha (m)
-1 x 10E-4 delta K/K/deg-F (0.25)
=
alpha (d)
=
-1.2 x 10E-5 delta K/M/deg-F.
(0 25)
alpha (v)
=
-1.0 x 10E-3 delta K/K/%V (0.25)
. _... -
delta (T) =[I-1 x 10E-4)(-10) +
(-1 x 10E-3)(-2)] /
(-1.2 x 10 E - 5 )4+ 5+-
delta (T) fuel
= -250 deg F or 250 deg F increase in fuel temperature (+0.5 pts). yeg,p /,ogjp$d b.
delta (ro)
alpha (x) * delta (X)
-
=
delta (ro)dop
= more negative (+0.5)
delta (ro) void less negative (+0.5)
=
deltatro) mod less negative (+0.5)
'
=
'
REFERENCE
NMP-2, Operations Technology, Module I,
Part 12, pp.
12.5,
.-
12.7, Fig.
12-6, 12-7.
Student Learning Objectives No.
2.c.
2.
NMP-2, Operations Technology, Module I,
Part 13, pp.
13.5[
13.6.
Student Learning Objectives No.
2.c,
ANSWER 5.10 (3.00)
a.
INCREASES [+0.253.
Because the change in density of water per degree F change in temperature increases with increasing temperature
[+0.753.
(+1.0 TOTAL)
b.
INCREASES [+0.253.
Because neutron leakage from the fuel cell to the volume around the control rod increases exposing the
,
rod to a higher thermal neutron flux [+0.753.
(+1.0 TOTAL)
'
c.
DECREASES [+0.253.
Because the amount of resonance broadening per degree F change fuel temperature decreases OR at higher fuel temperatures most of the broadening takes place at the higher energies where fewer and fewer neutrons exist [+0.753 (Either reason correct: for full credit,)*
,
(1.0)
)
-
.
.
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,
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/
5.
- THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE
THERMODYNAMICS
.
ANIVERS -- NINE MILE POINT 2-85/12/10-C.A. SLY REFERENCE 1.
NMP2, Operations Technology, Module I,
Part'12,13,14 pp.
,
I-12-3, I-12-4, I-12,5, I-13-2, I-13-3, I-14-6, I-14-7, S t u,d e n t Learning Objective No.
12-2, 13-2, 14-4a.
ANSWER 5.11 (2.00)
.
Using P
Po e ** (t/T)
(+0.5)
=
75 e ** (-60/80)
(+0.25)
=
35 on Range 4 (+0.25)
=
On a down power transient, with large negative reactivity i n s e r t i o n s, FP9-4+= t h e stable decay period is determined by the longest lived half-life.
(+0.5)
For this example it is assumed to be -80 sec.
(+0.5)
..
-
REFERENCE 1.
NMP2, Operations Technology, Module I,
Part 10, p.
I - 1 0 -.2.
Student Learning Objective No.
3.
.'
,
.
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6.
' PLANT SYSTEMS.
h
.
.
DESIGN, CONTROL, AND INSTRUMENTATION PAGE
~
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AN;WER3 -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSWER 4.01 (3.00)
a.
decrease (+0.25) due to steam / feed mismatch, requiring less water (-+ 0. 5 )
(+0.75 Total)
b.
increase (+0.25) due to level mismatch requesting more water (+0.5)
(+0.75 Total)
c.
no change (+0.25),
servo would lock up valve as_is (+0.5)
(+0.75 Total)
d.
decrease (+0.25),
due to reduction in operator setpoint of one-half input value (+0.5)
(+0.75 Total)
/$d" ell-
% % Nf o00 san /s ee) ssl(Nd J
1.
NMP-2 Operations Technology, Module IX, Part 6,
pp.
4,
Student Learning Ojbective Nos.
4, 7.
2.
NMP-2, IOP-7,'pp.
3.
- L..,
-. ~..
ANSVER 6.02 (2.50)
g/2/ti /s.,
Co tt- [h.J a.
Safety function. status display (+0.5) OI
/x %tu#W ffV a v o '
b.
1.
reactivity control - APRM status DA wc// /84/
f (0.1 es e h ~)
'.
core cooling - RPV level
.
3.
coolant system integrity - RPV pressure or drywell pressure.
l 4.
containment integrity - drywell pressure, drywell oxygen. concentration, or suppression pool temperature (0.25 each display,
+0.25 for each status pa r ame t e r-)
REFERENCE
,
1.
NMP-2 Operations Technology, Module VI, Part 12, SPDS, j
pp.43, 4 of 8,
Student Learning Objective Nos.
2,
)
,
e
.
_
-
-
.__
_
/
..
.
j 6.
' PLANT SYSTEMS DESICN. CONTROL. AND INSTRUMENTATION PAGE
=
'
.
AN;WERS -- NINE MILE POINT 2-85/12/10-G.A. SLY A'NSVER 6.03
'(2.00)
a.
control valves close 5% (+0.25),
open one bypass valve (+0.25)
(or similar answer on diagram)
(0.25)(reactor b.
Control valves close 5%,
scram probable due to increas'ing pressure since) bypass valves will not
.
g * j* dt{, '}
be open (+0.25)
,
(will N
develop a pres ure error of 800 psid This will be on c.
,
d hmalid_LgP;na x imum fg, ever, due ov/ U.
Qe.
m#1 control a
~
o ng a
v
to the action o f /th a n 'e om in f l'o wv'alNes'/,
I j
va Yes w'L1 1 tb 104%
emand and b pass go to
.
V fic 71,udle n40 d.
control valves close to 90% (+0.25)
to maintain uw pressure at.924 psig (+0.25).
REFh.RENCE 1.
NMP-2 Operations Technology, EHC, Rev.
1, pp.
2, 5 to 9 of 14, Student Learning Objective Nos.
5, 6,
8, including EHC Figure 3.
..
.. _
ANSVER 6.04 (1.00)
~
a.
If subs.titute position data has already been entered from the RSCS operators panel, that rod has been moved one notch, and good position data is still missing, then a rod motion insert and withdraw block will occur.
(+0.5)
b.
From 75% rod density to the LPSP, only notch rod movement is allowed between 00 and 12.
(+0.5)
I EFERENCE 1.
NMP2, Operation Technology, Module VI, Part 6,
p.
11; SLO 5.a.
j. ' ~ 3f'- %d a4 q6 I
O dcl comm,ss. J. a,e d,,o t,.,. sap.sdahl c], f.d a) odud ha k /,m,f
s.,ent a w,t% s~k !.2 4
y, gra a' ads fuJ f
.s. Ascs,;,ey
&"1 :lwe(.4 4-f II ared,%
.
-l
-
.
.
Y
.
.
.
6.
' PLANT SYSTEMS DESIGN, CONTROL. AND INSTRUMENTATION PAGE
i
,
,
ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 6.05 (2.00)
a.
660 ppm (+0.15),
50 (+0.25)
b.
Instrument Air - air to bubbler level indicator (+0.5)
sparging air for preparation of poison solution (+0.5)
c.
FALSE ( 4 0.5~ )
REFERENCE 1.
NMP-2 Operations Technology, Module VI, Part 9,
pp.
2, 5,
B, 9,
Student Learning Objective Nos.
2, 3,
4,5.
ANSVER 6.06 (2.00)
a.
close (+0.5)
b.
I no change (+0.5)
2.
.no change (+0.5)
3.
' --- --
(+0.5)
REFERENCE
-
1.
NMP-2 Operations Technology, Module III, Part 5,
pp.
4,5,6; Student Learning Objective No.
4.
ANSVER 6.07 (2.00)
a.
Alternate Rod Insertion, Recirculation Pump Trip b.
None A WO L4.T40L-ATsro < ,'O.c)
c.
Stan'dby Liquid Control go see. g4f
' l d.
34M //nrah
.30 C
-
to go,;, y fp,s.a f,. gc c.
(+0.5 for each,
+2.0 TOTAL)
j REFERENCE
NMP-2, Operations Technology, Module VI, Part 8,
pp.
2,4,7 Student Learning Ob)ectives No. 4
'
i
.
-,
,, - - -
w-.
- - - -
, -, -, --
--n,
,, - -.,, - -
., - - -,,,,,
,,, -,
,,,,-_e,-
-
.]r 6. PLANT SYSTEMS. DESIGN. CONTROL. AND INSTRUMENTATION PAGE
'
ANSVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY (
ANSVER 6.08 (1.50)
a.
Purity - as purity decreases, cooling capability decreases (+0.25)
I 2.
Pressure - as pressure decreases, cooling capability
!
decreases (+0.25)
3.
Temperature - as temperature increases, cooling cap-ability decreases (+0.25)
@ i g (6. b o n (.w ) lwd on MA b.
1.
Terminate purge due to single purge line (+0.25),
(-generator isolates)
'
ash) bokhb
.
2.
no.-etTve4 (Tank 1 isolates) (+0.25),(TK2 M /1Dd A l
supplies)
3.
no effect (+0.25),
(dump valve not opened or used during purge)
'~
REFERENCE
- ~ ~ 1-: NM P - 2 Operations Technology, Mo d u l e* -V 1-I I,-P a r t 6,
pp.
4, 6,
8, 10,. Student Learning Objective Nos.
2, 4,
5.
,
f.'2 - J.x,
- ?
.(&
(1. l. L.
~
-
ANSWER 6.09 (2.50)
a.
Pulse. height - neutron pulse larger than gamma pulse ( + 0. 2 5 )', ' pulse height discriminator (+0.25) chops gamma and only passes neutron pulses (+0.25)
Cambelling - neutron pulse is larger than gamma pulses (+0.25); Cambelling (+0.25) squares the two signals then gamma and passes only neutroAs (+0.25)
(chop
'
-% b e wna.s u n s.pd.' eu.&
(+1.5 Total)
b.
Due to the low number of events and greater sensitivity (+0.25),
the SRM deals with individual counts (pulses)
(+0.25) where the IRM deals with time averaged signals
~(+0.5)
(+1.0 Total)
f REFERENCE 1.NMP-2 Operations Technology, Module VI, Part 1,
SRM, pp.
7, 8,
Student Learning Objective No.
3.
2.NMP-2 Operations Technology, Module VI, Part 2,
IRM, pp.
3,
-
4, Student' Learning Objective'No.
3.
i
.
t
.
.
,, _ _
_ _
,
. _... _ _
.-
6.
' PLANT SYSTEMS DESIGN, CONTROL, AND TNSTRUMENTATION PAGE
~
,.
.,.
.
ANSWERS -- 'NINE' MILE POINT 2-83/12/10-G.A. SLY
.
?
.
ANOVER 6.10'
(*2. 0 0 )
a.
closure of the HPCS injection valve (MOV-107)
(+0.5)
M'pNn v e.hb s==ta- @t b.
Mu on the low-low setpoint (+0.5)
h.
c.
stay in manual bypass and not reinitiate (+0.5)
d.no (+0.5)
REFERENCE 1.NMP-2 Operations Technology, Module IV, Part 2,
HPCS, pp.
1, 2,
5, 6,
7, Student Learning Objective Nos.
2, 4,
7.
ANSWER 6.11 (2.50)
,ptog y Q, Msit. W'
a.
'B'
- Head Spray Mode (+0.25)
~
.
- IT\\
- Containment Flooding Mod 9 5)
-o/6-
_
b.
Prevent inadvertent draining of the vessel (+0.5)
2.
Prevent exceeding RHS design pressure (+0.5)
NC Yl)e S vc& Vu Ive MJV i W $ Ne S L i n $~ b hN l* )
--
c.
Yes(+0.2E), tre+au s-e t h e - t ecend-I,PC I irit!2 68^=
=ic==8 e l-H ht-trN y.= t em-by-r eopen-i-ng _t h e LPC!
!=j-<+ian valve.
~'(+0.75)
blu u To q ktu.3 v',h ycJ-l gg f g,yyp,4) iJpn& Q (x n REFERENCE 1.
NMP-2, Operations Technology, Module IV, Part 5,
RHS, pp.
5, 9,
10.
Student Learning Objectives No.
1, 5,
6.
'
g z-1D&b. e'f
,..
ANSWER 6.12 (1.00)
/
No (+0.25)
When transferring RPS power supplies, the RPS is momentarily deenergized because the transfer is break before make.
This would result in a scram due to the 1/2 scram j
already present (+0.25)
C REFERENCE NMP2, N2-IOP-97, RPS, p.
6.
. - ;.;g*
.., +
4.
' PLANT SYB'TEMS DESIGN. CONTROL. AND INSTRUMENTATION PACE
.
..
AN'iVE RS -- NINE MILE POINT 2-85/12/10-C.A. SLY ANSVER 4.13 (1.00)
~ & ~},;ll //oge hf h A5 P'Y - h b Limits negative pressure differential (+0.5)
to prevent drawing water up the downconer from the suppression pool Qh to the drywell.
(+0.5)
REFERENCE 1.
HMP-2 Operations Technology, Primary Containment, Rev.
1, p.
- A sec 6-3N-4.Y
>
)
..
J
- _ _ - - -.
_. -. _ _.
. _. _
.,
__.._,m_
,,,_,,,_,__,.__.___,_,-,_,,y_
_ _,,,
.__,,__,,,m,
,
_, _. _. _,,,.. _, _. _ _ _. - _..,.
_.,,
..
- - _ _
7.
PROCEDURiS NORMAL. ABNORMAL, EMERGENCY AND PAGE
RADIOLOGICAL CONTROL i
NINE MILE POINT 2-85/12/10-G.A. SLY AN;WERS
--
,
ANSVER 7.01 (2.00)
a.
Prevent undue stress on vessel.
b.
Prevent undue thermal shock in recirculation pump and nozzles.
c.
Pravent undue thermal stress on vessel nozzles and bottom head.
d.
False.
REFERENCE 1.
Technical Specifications 3/4.4.1.4 and B.3/4.4-1.
2.
NMP-2, N2-IOP-29, Recirculation, p.
3.
ANSVER 7.02 (2.00)
a.
1.
cond station air compressor on sta dby has started c88" at 100 psig.
(+0.5)
[or $O f 3 #h u h)
2.{Thirdb "$Mk c1w T
90 psig.
(+0.5)
( or OS f 5 sy )
f k. Sad station air compressor on s,tandby has started at
$tT f g,g Ir si Su s? e.o sir'
f.tA$ - A0Vril s
p rdC f'\\ d 3.
Station air isolation valve (M" p,' ;e 3
-
G1) closed at 85 psig.
(+0.5)
r( J.- 35
%
psig[,e f Sg,d a.E /an. h,h
'
b.
If
-
pressur'e reaches 60 REFERENCE NMP-2, N2-IOP-19, Instrument Air, pp.
6, 8,
?) s.3 ANSVER 7.03 (2.00)
a.
routine or repetitive work (+0.5)
b.
1 year (+0.5)
,
group / (+0.5)
due to ALARA program (+0.5)
c.
i REFERENCE
NMP-2, S-RP-2, RVP Procedure, p.
14.
2.
NMP-2, S-RP-7, ALARA, pp.
2,3.
/(
(,,,
g.y-f, $ (,,
i
.. - -
.
__-
,..
.- --.
- _. -. -
,
.
.
.
_.
..
-
-
,
7.
'PRQCEDURES NORMAL. ABNORMAL. EMERCENCY AND L.PAGE
'"
.
RA.DIOLOGICAL CONTROL ANIVERS -- NINE MILE POINT 2-85/12/10-G.A. SLY ANSVER 7.04 (1.50)
)
I
a.
2 re (+0.5)
due to 5(N-18)
,
.
b.
no (+0.5),
exceeds 5(N-18) limit (+0.5)-O't '[U 0 N 14A
""C-y l
REFERENCE 1.
NMP-2, S-RP-1, Access and Radiological Control, pp.
1, 12, 17.
2.
NMP-2, EPP-15, Health Physics Procedure, p.
3.
ANSVER 7.05 (2.50)
I 1.
Making decision to notify offsite emergency management l
agencies g
i 2.
Making protective action recommendations as necessary to l
offsite emergency management agencies 3.
Classification of the emergency event 4.
Determining the necessity for a site evacuation 5.
Authorizing emergency workers to exceed normal radiation exposure limits (+0.StEwd )
,
'
REFERENCE NMP-2, SEP Sec.
5, Organisational Control of Emergencies, pp.
4, 5.
l l
ANSWER 7.06 e.e. a d G O(2.00)M-60 b a.
Concentrate boron (+0.5)
enhance void generation (+0.5)
b.
Max. temp. at which SLC initiation will result in injection of hot shutdown boron wei before the supp. pool reaches ATVS,h.e.ght the HCTL in an assures shutdown prior to emergency depressurization.
(+1.0)
REFERENCE 1.
NMP-2, N2-EOP-C7, Level / Power Control, p.%, Student Learn-ing Objective Nos.
1, 3.
2.
NMP-2, N2-EOP-RG, RPV Reactivity Control, p.
2, 12 of 21
% %s I
_
_
._.
.-.
--
J
y
.
s 7.
- PROCEDURE'S - NORMAL. ABNORMAL. EMERGENCY AND PACE
.
RJLDI OLOG I C AL CONTROL AN" VERS -- NINE MILE POINT 2-85/12/10-C.A. SLY Student Learning Objective No.
3.
ANSVER 7.07 (1.50)
Following flooding of vessel (+0.25). the flow path would be:
Main steam lines to suppression pool via SRVs (+0.25)
Suppression pool to vessel via core spray (<0.25) or LPCI (O A ICCb M 5h (+0.25)
Heat is removed from suppression pool by suppression pool cooling mode of RHR (+0.5)
l REFERENCE l ;
NMP-2, N2-EOP-C5, Alternate Shutdown Cooling, p.
1.
Y"Y A
ANSVEP.
7.08 (2.50)
to h Na or a.
The CRD pump will increase water level (+0.25) and there hi p5I#
is no outlet flow path established (+0.25)
n b.
Because cooling flow is lost to the regeneratives heat US$
r
$Q exchanger WA.25) increasing the outlet temperature to cir54 t?
g I
h&* the NRHX (+0.25h possibly causing isolation of system (+0.5)
Yr'* ' "
s% e
.
W{
r c.
Hot shutdown with no recirculation pumps operating (+0.5)
minimizes the Hotstvbaly, -Co rstrati Afictionkofy,ssel j?,
ve water (+
.5)
i;
'
or r
oc n
c q
REFERENCE j f?% e[lE NMP2, Operations Technology, #CCR f
g.1-rop -37 pg. t{ G m Q t/
ANSVER 7.09 (3.00)
/Y/
a.
CAF (+1.0 TOTAL)
.:: *
b.
Above (+0.5)
assures sufficient heat capacity available to absorb the energy from RPV blowdown (+0.5)
c.
Spray initiation above this limit may, result in a containment
.
depressurization rate that exceeds the relief capacity of the drywell M c-
- ^ -
'-a vacuum breaker.
(+1.0 TOTAL)
a
.
.
I J
7.
PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND PAGE
I RADIOLOGICAL CONTROL AN2 VERS -- NINE MILE POINT 2-85/12/10-C.A. SLY REFERENCE g/L 1.
NMP-2 N2-EOP-PCF,.;-
.ud p.
8,
', - 1 '. of 18 Student Learning Objective No. 3 ANSVER 7.10 (2.00)
a.
To minimise the transient.
(+0.5)
b.
RCIC (+0.5) will isolate at 50 psig (+0.5)
and you want assurances that you have an injection mode available prior to depressurisation.
(+0.5)
REFERENCE
.
,, ___,,,.,
-- j/W 2., pt.-ScP-69 o t'y 10 * C LI.
r
...
..
n..-
1.
nnr-.-
--
__
r.
I 2.
NMP-2 N2-EOP-SPT, p.
2, 3 and p.
8, 10 of 13 Student Learning Objective No. 3 ANSVER 7.11 (2.00)
Initial State
.... - MV e and 05r MV (+0.25)
/ 2?f
'lW
.
~
You shou reduce generator load by recirc. or rods to 1.210 MVe (+0.
5),
then raise reactive load (VAR) by adjusting the AC voltage regulator (+0..)(5)
(+0.5 for order of steps)
1 Final State 1.210 MVe and 600 KVA4(+0.25)
REFERENCE 1.
NMP-2, N2-!OP-48, Main Cen.,
p.
5 and Figure 3
,
'
Power Factor Chart Provided ANSVER 7.12 (2.00)
a.
This is necessary to prevent uneven heating of the rotor I
.
(+0.5)
If it is not started, a rotot long condition could result.
(+0.5)
setpoint(sf the " Turbine Stop and Control Valve Closure b.
The exceeded)(+0.5)
Bypassed" annunciator)could be and a reactor scram would result.
(+0.5)
/
l
--
-. - _
7.
- PROCEDURiB NORMAL. ABNORMAL. EMERGENCY AND PAGE
RADIOLOGICAL CONTROL ANSWERS NINE MILE POINT 2-85/12/10-G.A. SLY
--
REFERENCE 1.
NMP-2 N2-!OP-21, Precautions 2,
3.
i
>
!
l
\\
-
,
- l t
- - - -
,, - - -.
- - - - -
.-_.._-,,
- - -
, -... - - - -, -- - - - -, -.,, - - -. -. - -, - _ - - - - -.. -. - -
.
_ _ _ _ _ _ - - _
_ _ _ _ _ _ _
a.
ADMINISTR'ATIVE PROCEDURES. CONDITIONE. AND LIMITATIONS PAGE
ANSWERS -- NINE MILE POINT 2-85/12/10-C.A. BLY I
t'
ANSVER 8.01 (2.00)
a.
Yes (+0.5),
following putting RPS trip System A in the tripped position (+0.5)
as per 3.3.1.a.
(+1.0 Total)
b.
1.
No (+0.25),
IRMs are not required in Condition 1 and you may stay there (+0.25)
(+0.5 Total)
2.
Yes (+0.25),
unless you had the RPS trip System A in the tripped position (+0.25).
Specification 3.0.5 is not applicable.
(+0.5 Total)
REFERENCE 1.
Tech. Specs, pp. 3/4 0-1, 3-1 to 3-4.
ANSVER 8.02 (3.00)
a.
Yes (+0.5),
due to failure of surveillance 4.8.1.1.2.7 air pressure greater than 225 psig (+0.5).
(+1.0 Total)
- O rt.- a. At4%4.$ys +tet$ a r e. n o + o fttstb it
.
f b.
No (+0.5),
due to Specification 3.0.3 which states you can be without emergency power source if ou have everything else (+0.5).
(+1.0 Total) 9 AI5c tu s'
geT u nWe' o f T,$. 3,N' I* *
Cr L~ I o s) E
'
c.
You would be in violation of Specification 3.0.3 (+0.5),
and must perform the action statement (+0.5)
(+1.0 Total)
's
,
ff* 3,$././.E $U$ClGLhh 7by M *A'
REFERENCE
- g,g _.
1.
Tech. Spec.,
pp. 3/4 0-1, 8-1 to 8-8.
g-rs. 3.s;/.d 1ett idq 4k.
/,
-
Gr'
'
c o-f.f 3.s p, a. Wr s.u& e4 a.IkL*h '
l
'f5.
s. l'./ e o./;,./
I I
l'
I
,
s.
ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE
.
- AN; VERS -- NINE MILE POINT 2-85/12/10-G.A. GT Y ANSVER 8.03 (2.00)
a.
Yes ($0.5)
b.
1.
The intent of the original procedure is not altered.
(+0.5)
2.
The change is approved by two (2) members of the plant management staff, at least one (1) of whom holds a Senior Reactor Operators License on the unit affected.
(+0.5)
3.
The change is documented, reviewed, and approved ". y the General Superintendent Nuclear Generation or designee within 14 days of implementation.
(+0.5)
REFERENCE 1.
NMP-2 Tech. Spec., Administrative Procedures 6.8.3.
2.
NMP-2 Exam BTnk.
ANSWER 8.04 (3.00)
1.
THERMAL POWER, Low Pressure or Low Flow Thermal Power shall not exceed 25% of Rated Thermal Power with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
(+1.0)
2.
THERMAL POVER, High Pressure and High Flow gl The Minimum Critical Power Ratio (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure
greater than 785 psig and core flow greater than 10% of j
rated flow.
(+1.0)
j 3.
REACTOR COOLANT SYSTEM PRESSUR,E The reactor coolant system pressure, as measured in the reactor vessel steam dome, sha!! not exceed 1325 psig.
6e l. 0)
REFERENCE
NMP-2 Tech. Spec., pp.
2-1, 2-2.
.
-. _. - -
.-.---
-
_.
- 1 ;1 ADMINISTRAYIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE
'
AN;UERS -- NINE MILE POINT 2-85/12/10-G.A. GLY ANLVER 8.05 (2.50)
a.
No (+0.5),
allowed to exceed weekly by 25% er...;
d=y (+0.25)
no restriction doing them early (+0.25),
Also did not exceed 3.25 times interval for three (3) consecutive surveillance (+0.5)
b.
Next surveillance would be Wednesday, December 19 (+0.5),
,
'
because you are limited by the three (3) consecutive interval limit (22 days) from November 27 (+0.5)
REFERENCE 1.
NMP-2 Tech. Spec., pp. 3/4 0-2.
,
ANOVER 8.06 (2.50)
.
a.
1.
No known Pressure Boundary Leakage (+0.5)
'!
2.
5 gpm unidentified leakage (+0.5)
t
3.
25 gpm total leakage averaged over a 24-hour period (+0.5)
!
b.
4 gpm
---% unidentified leakage 22 gpm --- identified leakage
.
24 gpm total leakage (+0.5)
..
Reduce the total leakage rate to less than 25 gpm within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least hot shRtdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown in the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(+0.5)
REFERENCE 1.
NMP-2 Tech. Spec., LCO, Reactor Coolant System, Operational Leakage.d.f.6/.7 LD
,
ANSVER 8.07 (2.50)
a.
SRO - 1 (+0.5)
(+0.5)
-
"
--===d
^ ^ - e '&e cs (+0.5)
1 - ' "
-
W b.
No (+0746), the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> asception does not apply during shift changes (+0.5)
( A 74-To444) _
f O VC V C f3
\\Q sf o n.5
rcA 'J'C!r
- G A
s 1. 0 5
S.
ADMINISTR'ATIVE PROCEDURES. CONDITIONS. AND 1METAT1DNS PAGE
"
NINE MILE POINT 2-85/12/10-C.A. CLV j
+ ANSWERS
--
l
\\
REFERENCE I
1.
NMP-2 Tech. Spec.,
3.6, p.
6-1 and Table 6.2.2-1.
AN;WER 8.08 (1.50)
j
'
a.
T=
FRTP/CMFLPD Both items defined in Tech. Spec.
'
Definitions (+0.25)
1.1 (+0.25 0.902/0.746 C(300/3323)/103/13.4 T
=
=
=
)
S is less than or equal to (0.66 V + 51%)
(+0.25)
,
b.
No (+0.25)
The Tech. Spec. require an APRM adjustment only if Tau is less than or equal to 1.
(+0.5)
,
REFERENCE 1.
NMP-2 Tech. Spec., 3/1 2.2, LCO, Power Distribution Limits,
,
APRM Setpoints, p.
2-5.
2.
NMP-2 Exam Bank.
ANSVER 8.09 (1.50)
~
r be in hat shut'do in 12 a.
Re' store on pump ~ ithin 7(hours'
hour'e ank e d sh do wiKhin 24 our A l'to take T 1'G,N O. 73 ~
g re\\uirgd k spec.
s
~
s.os.5.2NandN3.8.1.2..
(MW 0 7,3 s,
.
.
b.
T.S.
3.0.4.
delineates the measures to be taken for those circumstances not directly provided for in the action statements and whose occurrence would violate the lutent of the specification.
( + 144-TOTAL )
.e-0 7'T REFERENCE 1.
NMP2, T.S.
bases 3.0.4.,
3.5.2, 3.7.1.1, 3.8.1.2 ANSVER 8.10 (2.00)
a.
Restore within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in Hot Shutdown in 12 hrs.
T.S.
3.6.2.3 A (1.0)
b.
Be in at least HOT Shutdown in 12 hrs.
T.S.
3.6.2.3.h (1.0)
..
.
.-.
_
. __
_
_
_
J
-
a [*. ~ADNINISTM ATIVs PRr1CjpgQ, CONDLTLQNS. ANb L IHITATIOfG.
?AC3
[2
~
'
-
,
.
.t
-
i
.ANIVERS -- NINE MILE POINT 2-85/12f16-C.A. SLY
REFERENCE I
1.
NMP2, T.S.
3.6.2.3
,
I l
l l
ANSVER 8.11 (2.50)
'
a.
T.S.
3.6.3 --
.7.^
(+0.5)
-
CPt*<Lb lG b.
N4 (+0.5) RCIC can provide its intended function, but you
-
have violated Primary Containment Integrity requirements and must (+0.25)
demonstaate the inboard isolation valve operable and (+0.25)
i 2.
within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:
(+0.25)
i a.
restore the inop valve to operable (+0.25)
b.
isolate line (this makes RCIC inop) (+0.25)
3.
or be in Hot S/D in 12 hrs and Cold S/D in 24 hrs.
(+0.25)
REFERENCE 1.
NMP2 T.S.
3.6.3 and 3.7.4
I
.
b l
I
,_
._.
. - -,
-. _ _ _ - - -, -. -,, - - _ -
- - - -
..
,- g w--.
-
--
. - = -.
. - _...-
.-
'
'
TEST CROOS REFERENCE PAGE
.,.
dUE~110N VALUE REFERENCE
~____...
____._
__________
05.01 1.00 SLY 0000017 05.02 2.00 SLY 0000021 05.03 2.00 SLY 0000022 05.04 2.50 SLY 0000023 05.05 2.50 SLY 0000024 05.06 2.00 SLY 0000025 05.07 2.00 SLY 0000026 05.08 2.00 SLY 0000027 05.09 3.00 SLY 0000078
'
05.10 3.00 SLY 0000106 05.11 2.00 SLYOOOO111
______
25.00 06.01 3.00 SLY 0000028 04.02 2.50 SLY 0000027 06.03 2.00 SLY 0000030 06.04 1.00 SLY 0000031 06.05 2.00 SLY 0000032 06.04 2.00 SLY 0000033 06.07 2.00 SLY 0000034 06.08 1.50 SLY 0000035 06.09 2.50 SLY 0000036
'
06.10 2.00 SLY 0000037 06.11 2.50 SLY 0000051 06.12 1.00 SLYO000090
06.13 1.00 SLYOOOO116
______
25.00 07.01 2.00 SLY 0000091 07.02 2.00 SLY 0000092 07.03 2.00 SLY 0000093 07.04 1.50 SLY 0000094 07.05 2.50 SLY 0000095 07.06 2.00 SLY 0000098
-
07.07 1.50 SLY 0000099 07.08 2.50 SLY 0000100 07.09 3.00 SLY 0000109 i
07.10 2.00 SLY 0000110 j
07.11 2.00 SLY 0000120 07.12 2.00 SLY 0000121
______
25.00 08.01 2.00 SLY 0000080 08.02 3.00 SLY 0000082 08.03 2.00 SLY 0000083 08.04 3.00 SLY 0000084 08.05 2.50 SLY 0000086
m
-
.,,-
-
-- y
--
- - - - - - - --
,
m
.g,.3,
'
TEST CR2CD REFERENCE PAGE
!
')
'
-
w.
L Eb2CTT' N VALUE REFERENCE s
________
______
__________
08.06 2.50 SLY 0000087 08.07 2.50 SLY 0000088 08.08 1.50 SLY 0000089
!
08.09 1.50 SLY 0000107 08.10 2.00 SLY 0000108 08.11 2.50 SLY 0000112
_. ___
25.00
. ____
______
100.00 l
!-
I
&