ML20133P317
| ML20133P317 | |
| Person / Time | |
|---|---|
| Site: | McGuire  |
| Issue date: | 07/30/1985 |
| From: | Rogers T, Wilson B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20133N905 | List: |
| References | |
| 50-369-OL-85-02, 50-369-OL-85-2, NUDOCS 8508140260 | |
| Download: ML20133P317 (175) | |
Text
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ENCLOSURE 1 EXAMINATION REPORT 369/0L-85-02 Facility Licensee:
Duke Power Company 422 Soutn Church Street Charlotte, NC 28242 Facility Name:
McGuire Nuclear Station Facility Docket Nos. 50-369 and 50-370 Written, oral and simulator examinations were administered at the McGuire Nuclear Station near Cornelius, North Caro ina.
,nu e ni m 7 73 ((
Chief Examiner:
Thomas Rogers '
d Date Signed vbce So fb' Approved by:
LLA R Bryce A. Wilson, Section Chief Date Signed Summary:
Examinations on May 21-24, 1985 Operating and written examinations were administered tc two senior reactor operator candidates, both of whom passed.
Operating and written examinations were administered to seven reactor operator candidates, five of whom passed.
8508140260 850731 PDR ADOCK 05000369 G
pgp
REPORT DETAILS 1.
Facility Employees Contacted:
- Cage, G. W., Superintendent of Operations
- Frye, S. R., Director of Operator Training (TTC)
- Gilbert, G., Operations Engineer
- Griffin, B., Training Instructor (TTC)
- McCraw, N., Compliance Engineer
- Phillips, R., Unit Scheduling Engineer
- Reeside, B., Operating Engineer
- Travis, B., Superintendent Integrated Scheduling
- Attended Exit Meeting 2.
Examiners:
Gruel, R.
- Rogers, T.
Schrieber, R. E.
- Chief Examiner 3.
Examination Review Meeting At the conclusion of the written examinations, the examiners met with Joe Iddings, Ray Phillips, John Sadler, and Bill Reeside to review the written examination and answer key.
John Sadler was replaced by Bill Griffin during the review.
The following comments were made by the facility reviewers:
a.
SR0 Exam (1) Question 5.17 Facility Comment:
McGuire's Lesson Plan does reflect the answer as stated by the NRC.
However, the McGuire Data Book curve for differential boron worth as a function of core age and boron concentration shows that choice (c) is also correct. We ask that answers (b) and (c) be accepted as correct.
Training Services will correct the lesson plan and have it reflect the actual core conditions.
See Attachment 1.
NRC Resolution:
Choice (b) or (c) is acceptable for full credit in accordance with the Differential Boron Reactivity Worth Curve.
2 (2) Question 6.06 Facility Comment:
Excess letdown and normal return can be diverted to the Reactor Coolant Drain Tank by operator action (repositioning NV278).
This was answer (a) and is the answer on the answer key.
However, with no operator action, relief valve NV93 will lift and excess letdown / seal return flow divert to the Pressurizer Relief Tank, answer (b).
We ask that both (a) and (b) be accepted as correct answers.
See Attachment 2.
NRC Resolution:
Since the question did not specify automatic or manual diversion, choice (a) or (b) is acceptable for full credit.
(3) Question 6.32 Facility Comment:
The answer key is incorrect for this question.
The correct answer is:
2/3 undervoltage on bus Load shed actuated 860 (lockout relays) reset Diesel generator speed >95%
The answer key includes the condition " Sequencer not in Test".
This answer is drawn from PSM drawing, EQA-9 (Attachment 4), which is incorrect.
Training Services will correct this drawing.
Our documentation for " Sequencer not in Test" being wrong is Attach-ment 3 and for 86D being correct is Attachment 5.
NRC Resolution:
86D (lockout relays) reset is required for full credit.
" Sequencer not in Test" will be graded as an incorrect response.
(4) Question 7.05 Facility Comment: Our turbine generator startup procedure (Attachment 6) has us latch, trip and re-latch the turbine before rolling the turbine to 1800 rpm.
The valves that open upon latching are different each time, depending on whether the DEH is j
limiting-the governor valves' position.
We ask that answers (b) and (d) be considered correct.
See pages 4 and 5 of Attachment 6.
NRC Resolution:
Since the question did not specify the condition of the governor valve limiter, choice (b) or (d) is acceptable for full credit.
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(5) Question 7.21 Facility Comment: The answer key is incomplete.
It should read:
Stop both DEH Pumps Place turbine in manual and close Governor Valves in fast action Locally trip turbine Close all S/G SM Isolation Valves See page 3 of Attachment 7.
NRC Resolution:
Closing all S/G SM Isolation Valves is not the appropriate operator action for an untripped turbine from the control room.
It is the appropriate operator action if the turbine will not trip locally in accordance with the referenced procedure.
Since the question is specific to the control room manual trip, closing all steam generator steam isolation valves is incorrect and will be graded accordingly.
(6) Question 8.06 3
Facility Comment:
This question asks the candidate to pick which parameter is different between Unit 1 and Unit 2.
McGuire's Technical Specification revision of March 1985 makes all of the possible answers incorrect.
We ask that all answers be accepted.
Reference:
McGuire's Technical Specifications, Attachment 11.
NRC Resolution:
Since Technical Specifications have been revised such that they are now identical with respect to this question, the question has been deleted from the examination.
(7) Question 8.13 Facility Comment: This question asks the candidate for Shutdown Margin LCOs for both units.
McGuire's Technical Specification revision for the Unit 2 refueling outage resulted in both units' LC0 being the same value.
We ask that both the "old" and "new" value for Unit 2 be accepted as correct answers.
NRC Resolution:
Since Technical Specifications had been revised two months prior to administration of the written examination, the candidates for licensing have had sufficient time to be aware of this significant change.
The answer key has been changed to reflect the new minimum shutdown margin values and the written examinations will be graded accordingly, i
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(8) Questions 6.31, 7.35 and 8.24 i
Facility Comment: We feel the above questions are misleading, l
l confusing or inappropriate and should not be used on future exams:
f NRC Resolution: The referenced questions were reviewed and considered to be within the scope of written examinations administered to senior reactor operators in accordance with i
NUREG-1021, ES-402 and are specific to the McGuire Nuclear Station.
(b) R0 Examination:
(1) Question 2.04 Facility Comment:
The answer key has FALSE as the correct answer.
We ask that if the candidate put TRUE and qualified the answer reflecting his knowledge of the connecting valve between VS and VI, TRUE would be an acceptable answer also.
i NRC Resolution:
If candidate responds TRUE with an appropriate explanation, full credit will~be awarded.
(2) Question 2.07 Facility Comment:
The answer key gives (c) as the correct answer.
1 This is incorrect, as the totalizers do not determine or control the boron concentration of the makeup.
Answer (b) is more correct
- see Attachment 8.
We ask that answer (b) be accepted as correct or that this question be removed from the section.
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NRC Resolution:
The answer key has been changed from (c) to (b).
j (3) Question 2.11 1
Facility Comment: We request that answer (a) be accepted as correct, in addition to answer (d).
During exam review, the NRC agreed to this request.
NRC Resolution:
The answer key has been changed to accept (a) or (d) for full credit.
1 i
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5 i-(4) Question 3.07 Facility Comment:
The answer key is incorrect on the logic for Low Steam Pressure Safety Injection. We showed the examiners the i
correct logic as listed in Technical Specifications.
NRC Resolution:
Review of the answer and reference indicates that 1
the McGuire Lesson Plan on the Safety Injection System is incorrect.
The answer key was changed to be inaccordance with Technical Specifications.
(5)
Question 3.15 Facility Comment:
The answer key is incorrect.
Answer (a) is correct - see Attachment 9.
NRC Resolution:
The answer key has been changed to (a) in 1
accordance with the reference provided.
(6) Question 3.22 i
Facility Comment:
The NRC answer key is incomplete.
Loss of one CF Pump, with Impulse Pressure >56% will also give turbine j
runback.
See Attachment 10.
i NRC Resolution:
The answer key has been changed to accept " Loss of one CF Pump with Impulse Pressure >56% as one of the three I
requested responses, in.accordance with the reference provided.
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(7) Question 4.19 Facility Comment: We ask that this question be removed from the test.
The Control Room copy of the Critical Safety Function Trees is in color, therefore, there is not need for operators to l
identify them by the type of line connecting the logic boxes.
In 1
addition, all license candidates must pass a color-blindness test l
as part of their application.
i NRC Resolution:
Though the candidates may know the Critical Safety Function Trees by color, they are also identifiable by the type of connecting line.
Since the question is in a matching format, it does test the candidates on their familiarity with the j
Critical Safety Function Trees and therefore has been included as i
part of the examination as presented at the facility review.
(8) Question 4.23 Facility Comment:
This question asks the candidate to pick which parameter is different between Unit 1 and Unit 2.
McGuire's Technical Specification revision of March 1985 makes all of the possible answers incorrect.
We ask that all answers be accepted, i 1 gives examples of Technical Specification revisions i
made for the Unit 2 fuel load this year.
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NRC Resolution:
Since Technical specification have been updated i
such that no correct answer is provided for this question, the question has been deleted from the examination.
(9) Questions 1.23, 2.10, 2.15, 2.22, 2.25 and 2.27 Facility Comment: We feel the above questions are misleading, 4
confusing or inappropriate and should not be used on future exams.
i NRC Resolution:
Except for the questions in this list previously commented on, the questions were reviewed and determined to be l
within the scope of a written examination administered to reactor operators in accordance with NUREG-1021, ES-201 and will remain as part of the examination as presented to the facility reviewers.
4.
Exit Meeting At the conclusion of the site visit the examiners met with representatives of the plant staff to discuss the results of the examination.
Those individuals who clearly passed the operating examination were identified.
i There were no generic weaknesses during the oral examination.
r The cooperation given to the examiners and the effort to ensure an atmosphere in the control room conducive to oral examinations was also noted il and appreciated.
1 i
The licensee did not identify as proprietary any of the material provided to i
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or reviewed by the examiners.
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ENCLOSURE 3 U.
S.
NUCLEAR RECULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
MCGUIRE 1&2 REACTOR TYPE:
PWR-WEC4 DATE ADMINISTERED: 85/05/21 EXAMINER:
TON ROGERS APPLICANT:
INSTRUCTIONS TO APPLICANT:
Use separate paper for the answers.
Write answers on one side only.
i Staple question sheet on top of the answer sheets.
Points for each l
question are indicated in parentheses after the question. The passing I
Stade requires at least 70% in each category and a final grade of at least 80%.
Enamination papers will be Picked up sin (6) hours after the examination starts.
"; 0 F CATEGORY
% OF APPLICANT'S CATEGOPY VALUE TOTAL SCORE VALUE CATEGORY 40.00 25.00 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 40.00 25.00 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 40.00
__'5.00 7
PROCEDURES - NOPMAL, ABNOPnAL, l
i EMERGENCY AND RADIOLOGICAL CONTROL I
4b40 25.00 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS
'T/
4+4.00 100.00 TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given not received aid.
5PPLECdUTIS 555Uh5URE
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THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIOSr AND PAGE 2
QUESTION 5.01 (1.00)
The units of neutron flu:< ar e s.
neutrons /sec.
b.
(neut.rons-cm)/(em cubed-sec).
c.
neutrons /cm squared.
- d. neutrons /(cm-sec).
QUESTION 5.02 (1.00) 0.00733 delta k per k is the same as a.
0.733 pcm.
- b. 7.33 pcm.
c.
73.3 pcm.
d.
733 pcm.
QUESTION 5.03 (1.00)
Which of the following conditions would cause a 1/d plot to be non-conservative during fuel loading?
a.
Fuel being loaded closer to a source range detector than to the I
neutron source.
b.
Loading fuel in the order of high reactivity worth to low reactivity worth.
c.
Loading poison rods between the source range detectors and spaces to be filled by fuel assemblies, d.
Increasing the boron concentration in the moderator.
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THEORY OF NUCLEAR POWER PLANT OPERATIONr FLUIDS, AND PAGE 3
3 g--------------------------------------
GUESTION 5.04 (1.00)
A constant suberitical multiplication count rate does not meet the definition of criticality (Keff=1.0) because a.
there are no delayed neutrons when soberitical.
b.
there is a decrease in neutron population that is not detected by the source range detectors.
c.
it has no effect on the SUR indication.
d.
the definition of criticality excludes source neutrons.
QUESTION 5.05 (1.00)
The required Shutdown Mar 3in is less for Mode 5 than it is for Mode 1 because a.
negative reactivity from baron has replaced the negative reactivity from xenon by the time hode 5 is reached, thus contributing to a more stable reactivity condition.
b.
the negative reactivity worth from samarium increases following a plant shutdown.
c.
the reactivity transients resulting from a steam line break cooldown are minimal.
d.
all control bank rods are inserted.
QUESTION 5.06 (1.50)
List the three bases for the required rod insertion limits.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 4
3 g--------------------------------------
GUESTION 5.07 (1.00)
Reactor buckling has the greatest effect on the a.
neutron leakage from the core.
b.
resonance escape probability.
c.
fast fission factor.
d.
thermal utilization factor.
QUESTION 5.08 (1 00)
What materials in the control rod are resonance absorbers?
a.
As and Cd only.
b.
In and Cd only.
c.
As and In only, d.
Agr In, and Cd.
QUESTION 5.09 (1.00)
The xenon peak that occurs after a reactor trip is higher following a 4
100% power equilibrium xenon condition than a 25% power equilibrium condition because
+
a.
the fission yield for xenon is higher at 100% power.
b.
there is more iodine in the core at the time of a trip from 100% power.
c.
there are more thermal neutrons in the core at 100% power.
d.
there are more delayed neutrons in the core at 100% power.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 5
i QUESTION 5.10 (1.00)
Which of the following statements describes a condition occuring in a fuel assembly containing a control bank rod that has just dropped from a 100% power equilibrium condition?
a.
The xenon concentration begins to decrease.
b.
The samarium concentration begins to decrease.
c.
The DNBR begins to increase.
d.
Beta effective begins to decrease.
QUESTION 5.11 (1.00)
If reactor power increases from 1000 cps to 5000 cps.in 30 seconds, what is the SUR?
a.
1.0 DPM.
4 b.
1.2 DPM.
c.
1.4 DPM.
t d.
1.6 DPh.
I OUESTION 5.12 (1.00)
A suberitical reactor with a neutron source strength of 20 cps and a source range count rate of 200 cps has a keff of (assume a prcportional-ity constant of 1.0) a.
0.85.
b.
0.90 c.
0.95 d.
0.99.
.i
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 6
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QUESTION 5.13 (1.00)
Which of the following factors of the six factor formula is greater J
than 1.0 for the McGuire Nuclear Plants?
l a.
The fast fission factor.
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b.
The resonance escape probability.
c.
The thermal non-leakage probability.
d.
The thermal utilization factor.
QUESTION 5.14 (1.00) 4 Which of the following six factor formula terms increases on a power escalation to allow reactor power to match turbine power?
a.
The fast fission factor.
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b.
The thermal utilization factor.
j c.
The reproduction factor, d.
The doppler effect.
QUESTION 5.15 (1.00)
The effective delayed neutron fraction decreases over core life partly because 3.
the number of delayed neutron precursor groups increase.
't b.
the fission yield for Pu-239 incroases.
c.
of a buildup of Pu-239 in the core.
d.
soluble boron is removed from the core.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 7
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THERMODYNAMICS i
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OUESTION 5.16 (1.00)
Neutron capture as a result of resonance is more significant at EOL that. BOL primarily due to i
3.
the reduction of fuel to clad gap distance.
b.
the reduction in the moderator's baron concentration.
c.
the increase of Pu-240 in the core.
j d.
the increase in thermal neutron flux.
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GUESTION 5.17 (1.00)
Which of the following statements correctly describes the changes in differential boron reactivity worth over core life?
l a.
It increases then decreases.
1 b.
It decreases then increases.
c.
It continually decreases.
i d.
It continually increases.
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QUESTION 5.18 (1.00) i The reactivity worth of a control rod increases
- a. ss Tave increases from 150 degrees F to 500 degrees F.
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b.
as reactor power is reduced from 100% to 50%.
c.
as a result of fission product buildup.
i d.
when the soluble boron concentration incresses.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 8
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4 GUESTION 5.10 (1.00)
The doppler power coefficient tends to become less negative over core life as a result of I
a.
flux hardening.
b.
clad creep.
c.
fuel densification.
I d.
gap putrification.
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QUESTION 5.20 (1.00)
In which of the following conditions is the Moderator Temperature Coefficient most negative?
4 a.
BOL, high temperature.
b.
BOL. Iow Temperature.
c.
EOL, high temperature.
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d.
EOL, low temperature.
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QU.STION 5.01 (1.00)
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The Power Coefficient does NOT include a.
the Moderator Temperature Coefficient.
b.
the Doppler Temperature Coefficient.
I c.
the Control Rod Coefficient.
d.
the Void Fraction Coefficient.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 9
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QUESTION 5.22 (1.00)
Which of the following describes the behavior of Tave when a SG Atmospheric Dump fails open when operating at 50% with the controls rods ir. Manua l and no operator action?
a.
Decreases and remains there.
b.
Decreases, then returns to it's original value.
c.
Increases and remains there.
d.
Increases, then returns to it's original value.
QUESTION 5.23 (1.00)
Delayed neutrons allow an operator more control of the reactor because a.
there are more delayed neutrons than prompt neutrons.
b.
delayed neutrons are born at a higher energy level than prompt neutrons.
c.
delayed neutrons decrease the average neutron generation time.
d.
delayed neutrons increase the average neutron generation time.
QUESTION 5.24 (1.00)
Which of the following statememts describes the relationship between integral and differential rod worth?
a.
Integral rod worth (at any location) is the slope of the differential rod worth curve at that position.
b.
Integral rod worth (at any location) is the total area under the differential rod worth curve from the end of the rod to that locatton.
c.
Integral rod worth (at any location) is the square of the differential rod worth at that location.
d.
There is no relationship between integral and differential rod worth.
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CATEGOPY 05 CONTINUED ON NEXT PACE
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50 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 10 GUESTION 5.25 (1.00)
The units of heat flux are a.
BTUs.
b.
KW/FT.
c.
KW/HR.
d.
BTU /(HR-FT SOUARED).
QUESTION 5.26 (1.00)
The 2200 degrees F maximum peak cladding temperature limit is used because a.
it is 500 degrees F below the fuel cladding melting point.
- b. any clad temperature higher than this correlates to a fuel center line temperature at the fuel's melting point.
c.
a :ircalloy-water reaction is acceler ated at temperatures above 2200 F.
d.
the thermal conductivity of :trea'lloy decreases at temperatures above 2200F causing an unacceptably sharp rise in the fuel eenter1ine temperatore.
QUESTI0tl 5.27 (1.00)
The Fuel Maneuvering Limits are r,e qu i r e d following a refueling outage to prevent undue stresses to the a.
fuel of the new fuel assemblies.
b.
cladding of the new fuel assemblies.
c.
fuel of irradiated fuel assemblies.
d.
cladding of irradiated fuel assemblies.
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS. AND PAGE 11 TEEREUU5 NAE C5~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
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l GUESTION 5.28 (1.00)
Which of the following is a correct statement concerning the Bases for the Tech Spec Pressure / Temperature Limits?
a.
During cooldown, the most limiting location is at the outside of the vessel wall because of the thermal gradients produce compressive stresses there.
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b.
During cooldown, the most limiting location is at vessel wall because the thermal gradients produce tensile stresses there.
j c.
During heatop, the most limiting location is at the inside of the vessel wall because the thermal gradients produce compressive stresses there.
d.
During heatup, the most limiting location is at the outside of the I,
vessel wall because the thermal gradients produce compressive stresses there.
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OUESTION 5.29 (1.001 The most serious problem with reaching the critical heat flu: is caused by j
a.
the poor thermal conductivity of steam.
i b.
the blockage of flow through the core when steam bubble formation j
becomes significant.
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the displacement of baron from the core as steam bubbles formation
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becomes significant.
d.
the hijh pressure surges in the Reactor Coolant System caused by l
steam bubble formation.
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 12
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QUESTION 5.30 (1.00)
The allowable heat flux at the top of the core is lower then the bottom j
of the core because a.
the water at the top has the highest enthalpy.
b.
the top of the core is under a lower pressure.
c.
in the event of a design bases LOCA, the top of the core is l
uncovered longer.
j d.
the fuel pellets loaded at the top of the fuel rods are not designed j
for a high power output.
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GUESTION 5.31 (1.00) j The Reactor Protection System would become unreliable for DNB protection from the OT delta T trip if voids were allowed to form in the Reactor j
j Coolant System because 1
a.
the heat transfer coefficient of the cladding is reduced significantly.
l b.
the specific heat capacity of the reactor coolant inventory changes when voiding occurs and is not measurable by the RTDs.
I c.
the critical point if water is reached and is not measurable by the l
RTDs.
d.
entropy becomes more limiting than enthalpy, which is not within the j
design considerations of the Reactor Protection System.
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GUESTION 5.32 (1.00) i The rod bow penalty described in Tech Specs is necessary on Unit 2 to protect the a.
KW/FT limit.
i b.
DNB limit.
c.
fuel temperature limit.
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d.
cladding temperature limit.
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50 THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 13 GUESTION 5.33 (1.00)
If the Reactor Coolant System is to be maintained at 350 de3rees F, then the Steam Dump pressure controller should be set to the equavilent of a.
135 psia.
b.
205 psia.
c.
295 psia.
d.
350 psia.
QUESTION 5.34 (1.00)
Which of the following meets the definition of Ovadrant Power Tilt Ratio?
a.
Minimum power range upper detector output divided by the average upper detector output.
b.
M a:< i m u m power range upper detector output divided by the average upper detector output.
c.
Minimum power range upper detector output divided by the average lower i
power range detector output.
d.
Ma:: 1 mum power range upper detector output divided by the average lower power range detector output.
QUESTION 5.35 (1.00)
The heat transfar mechanism from the fuel to the coolant becomes film-boiling 6t what D N E:R ?
a.
0.0 b.
1.0 c.
1.3 d.
1.7
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THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 14 GUESTION 5.36 (1.00)
The reactor is at 80% power with a core delat T of 48 degrees F and 3 mass flow rate of 100% when a station blackout occurs. Natural circulation is established and core delta T goes to 40 degrees F. If decav heat is appro:<imately 2% of full power, what is the mass flow rate (% of full flow)'
a.
1.9%
b.
2.1%
c.
2.4%
d.
3.0%
GUESTION 5.37
(.50)
Overall plant thermal efficiency will decrease if Lake Norman temperature increases. TRUE or FALSE?
OllESTION 5.30 (1.00)
Which of the following statements is correct if the discharge valve from 3 centrifugal pump is being partially closed from the full open position) 2.
Pump head decreases as head loss decreases, b.
Pump head increases as head loss inctoasos.
c.
Volume flow rate increases as head loss decreases.
d.
o l om e flow rete decreases as ho3d loss decreases.
I*****
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5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 15 t
e GUESTION 5.39 (1.00)
During a plant heatop with the pressuricer pressure at 1000 pstg, failure of the air supply solenoids allows the PZR PORV to open slightly to a throttling position. The ma:<imum pressure reached downstream of the valve is a pp r ot:i m a te ly the PRT pressure of 5 psis. What would be the condition of the fluid downstream of the valve?
a.
Superheated steam b.
Dry saturated steam c.
Wet steam d.
Subcooled liquid QUESTION 5.40 (1.00)
Which of the following describes the changes to the steam that occur between the inlet and outlet of a real (not ideal) turbine?
- a. Enthalpy decreases, entropv decreases, quality decreases.
b.
Enthalpy increases, entropy i n c r e.a s e s, quality increases.
c.
Enthalpy constante entropy decreases, quality decreases.
d.
Enthalpy decreases, entropy increases, quality decreases.
(****.s END OF CATECORY 05 sasast
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 16 GUESTION 6.01 (1.00)
Which of the following Reactor Coolant System temperature instruments are used by the Reactor Protection System?
a.
The pressuri:er steam space temperature detectors.
b.
The hot-les narrow range temperature detectors.
c.
The cold-les wide range temperature detectors.
d.
The core e:<it temperature detectors.
QUESTION 6.02 (1.00)
A pressuri:er low level signal will isolate Letdown and a.
trip the pressuri:er heaters.
b.
start a second charging pump.
c.
trip the reactor.
d.
initiate safety injection.
QUESTION 6.03 (1.00)
Which of the following conditions will result in a reactor trip?
I a.
Low flow on 2/3 detectors in 1/4 reactor coolant loops when ?P-7 and c.P-0.
b.
Low flow on 2/3 detectors in 1/4 reactor coolant loops when ?P-8.
I c.
Low flow on 1/3 detectors in 4/4 reactor coolant loops when
'P-7 and
<P-9.
d.
Low flow on 1/3 detectors in 4/4 reactor coolant loops when iP-0.
(***** CATEGORY 06 CONTINUED ON NEfT PAGE
- )
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 17 GUESTION 6.04 (1.00)
Match the following reactor coolant pump seal flow paths with the appropriate designed flow rate. The flow rates may_be used more than once.
COLUMN A COLUMN B 1.
Flow'down the shaft into the Reactor Coolant a.
0.0 cc/hr System.
b.
10 cc/hr 2.
Flow up the shaft to the 42 seal.
c.
100 cc/hr 3.
41 seal discharge back to the VCT.
d.
3 sph 4.
62 seal discharge to the standpipe.
e.
3 spm 5.
43 seal discharge to the Reactor Coolant Drain Tank.
f.
5 spm OUESTION 6.05 (1.00)
The Chemical and Volume Control System removes encess Lithium via the 3.
cation bed dominerali:er.
b.
mixed bed dominerali:or.
c.
BTFS deminerali:ers.
d.
volume control tank vont to the Waste Cas System.
QUESTION 6 06 (1.00)
If the excess letdown normal seal return to the VCT is isolated, then the return flow is diverted to the a.
Reactor Coolant Drsin Tant.
b.
Pressure Rel:cf Tank.
c.
Containment Sump.
d.
Peactor Coolant Sampling System.
(*aaa* CATEGORY 06 CONTINUCD ON NE,ti PAGE *esas)
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTPUMENTATION PAGE 18 OVESTION 6.07 (2.00)
List all protection si nals that will automatically initiate the a
Safety Injection System. Include lo.jic and setpoints.
QUESTION 6.08 (1.00)
The Unit 1 Safety Injection pump A is powered from a.
1 ETA.
b.
1ETB.
c.
2 ETA.
d.
2ETB.
OljESTION 6.09 (1.00)
The makeup water to the Reactor Coolant System provided from the SSF controls is supplied by
.), t h e N') Systom positivo displacomont pump.
b.
the ttatn A NV Systen. contetrujol charJinj punip.
c, a 9sF dedicated positivo di3 placement pump.
d.
- SSF dedicated centrifogr] thar31n3 puner.
OllES TION
- s. t0
't.no:
uhtch of the followinj Fesiducal Heat Removal Ovstem components can be opersted from the EGF7 3.
NC ;A i flC loop
'C' to NO 15clation valve n, b.
(IB lO ' ilC toop to N0 pump suction cont. 1:olation valvei.
c T r.3 t o A fl0,. ump.
d.
T r a i n f: fl0 pump.
/***e*
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i 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMEN TATION PAGE 19 OUESTION 6.11 (1.00)
ND-2A (NC loop
'C' to ND isolation valve) 15 interlocked so it cannot be opened unless NCS pressure is less than
, pressort er vapor space temperature is less than ______.
FW-27 and NI-185 are ______.
a.
385 psis, 475 F, or, closed.
b.
555 psig, 475 F, and, open.
c.
555 psig, 475 F, or, open.
s d.
305 psta, 475 F, and, closed.
QUESTION 6.12 (1.00)
The Reactor Coolant System radtstion detector (EMF-40) monitors the a.
neutron radtstion level.
b.
gamma radtation level.
c.
beta radiation lovel, if. glpha radration level.
l GUESTIDH A.10
(
.501 A CPCS signal jreater than 0.25 psij must e:i s t to manuallv start the Containment Gpray Pumps. TRUE or FALSE?
Gile 1 T ION t.14
/.50)
A CPCS 11jnal greater than 3.0 psig must o' tit to manoilly open the Contatomont Spray Discharge Spray Valves. TFUE or FALSE?
00ESTION 3.15
't.00)
Lt;t the four Avniliary Feedwater Pump suction supplios.
(sasas CATEGOPf 06 CONTINUED ON NEXT FAGE suas
)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 20 GUESTION 6.16 (1.00)
List the four Av::iliary Feedwater motor driven pumps automatic start signals.
QUESTION 6.17
(.50)
The Steam Dump System will allow a turbine and reactor trip from 100% load without lifting tha ASME Code main steam safety valves and PORVs. TRUE or FALSE?
GUESTION 6.18 (1.00)
The Steam Dump System will select the Plant Trip Controller when a reactor trip is sensed by a.
the reactor trip breakers position.
b.
the individual rod position Indication.
c.
the turbine impulse pressure.
d.
turbine EHC oil pressure.
r.ttE S TIO N 6.19 (1.50)
List three different conditions that will automatically close the main steam isolation valves.
QUESTION 6.20 (1.00)
The Instrument Air after coolers'are cooled by the J.
Low Pressure Service Water System.
(
b.
Nuclear Service Water System.
- c. Component Cooling Water System.
d.
Recirculating Water System.
(as***
CATEGORY 06 CutJ TINUED ON NEX T P AGE a s s e e )
6..
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 21 i
4 QUESTION 6 21 (1.00)
Which of the following is NOT a water source for the Gpent Fuel Pool?
a.
FWST.
b.
RMWST.
c.
Nuclear Service Water.
d.
Low Pressure Service Water.
QUESTION 6.22 (1 00)
Both Main Feedwater pump turbines will automatically trip a.
on HI-HI water level in the outer h
do3 ouse.
b.
if two condensate booster pump breakers open.
c.
if two FDWP suction pressure switches drop below 230 psig.
d.
on a LOW-LOW hotwell level sijnal.
QUESTION 6.23 (1.00) l If D-ata A and 8 fillures occur in the Digital Rod Position Indication l
System for the simo rod, which of the followinj todication will be recteved?
i An Organt Alarm, a Pod Pottom Light. a General Warning Light, and a 4.
Non-Urgent Annunciator.
b.
All the indication given in choico eacept the Rod Bottom Lijht.
c.
All the indication given in enoice
'a' o cept the Non-Urgent Annunctator.
]
.l.
An Urgent Al ar $ and a General Warntng Light ontv.
t 1
OUESTION 6.24 I.50) uhy most the RN system be aligned to discharge back to the SNSWP whnnever it draws a suction from it?
(::::: CATEGOPY 04 CONTINUED ON NExi PAGE esses) i i
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 22 QUESTION 6.25
(.50)
The RN non-essential supply crossover will be isolated automatically on a blackout or a LOCA signal. TRUE or FALSE?
QUESTION 6.26 (1.00)
List five different a v:< i l i a r y systems supplied or cooled by the Main Circulating Water System.
GUESTION 6.27 (1.00)
The Component Cooling Water Reactor Building Non-Essential Supply Header Isolation Valves (HC-230, HC-228) will automatically shut on a.
a Sp signal.
b.
3 Ss signal, c.
a St stjnal.
d.
a HI-HI flow signal through the applicable valve.
QUESTION 6.28 (1.00)
The G e n e r.s t o r Hydrogen cooler's are normally cooled by the a.
Fecirculating Water System.
b.
Condensate System.
C.
Nuclear Service Water System.
d.
Low Pressure Service Water System.
I a
(esses CATEGORY 06 CONTINUED ON NErr rAct uses )
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 23 QUESTION 6.29 (1.00) is added to the ice in the Containment Ice Condenser to aid in iodine removal in the event of a LOCA.
- a. Potassium Chromate.
b.
c.
Sodium Tetraborate.
d.
Sodium H y d r o::i d e.
QUESTION 6.30 (1.00)
When synchron:ing the generator to the gride the procedure directs the operator to regulate turbine speed to slowly rotate the synchroscope in the fast (clockwise) direction. Which choice below correctly gives the two parai.eters that the synchroscope is indic. tin 3?
- a. Current and voltage differences.
b.
Current and frequency differences.
- c. Voltage,3nd phase differences.
d.
Frequency and phase differences.
QUESTION 6.31 (1.00)
Which of the following will automatically start the 1A Standby Diesel Generator?
- a. A signal from the 1A sequencer,
- b. A SS si3nal from the SSPS.
c.
2/3 undervoltage signals on the ICTA.
d.
A Sp signal from the SSPS.
QUESTION 4.32 (1.0s)
List the four conditions necessary for tho '3 t i n d b y Diesel G.tnarator feedor br eak er to close automatically, isess: CATCGORY 04 CCtlTINUE0 Ori tJEYT TAGE evess)
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6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 24 QUESTION 6.33 (1.00)
If the sequencer reset botton is depressed while sequencing is in progress during a blackout.
a.
no futher load groups will be applied.
b.
the sequencer will start over again delaying the loads vet to be sequenced on.
c.
all loads on will load shed.
d.
will lock in the blackout loads as first priority over LOCA loads.
QUESTION 6.34 (1.00)
Zf a 125 VDC vital battery charger is lost, the 120 VAC vital instrument loads on the associated channel will be a.
lost with no operator action.
b.
picked up automatically by a spare battery charger.
- c. picked up automatically by a battery.
d.
picked up automatically by another operating battery charger.
QUESTION 4.35 (1.00)
When an urgent failure occurs in a Rod Control System power cabinet, the ofrected rods are provented from moving by 2.
supplytng a current to 'he Novablo and Stationary Gripper Coils simultaneously.
b.
m auton.atic bus transfer to the DC Hold Supply Cabinet.
e, erier g t::ing blocking diodes that block pulses from the slavo eveler.
d.
opentnj contacts supplying th.* Lift and Mavable Gripper Coil moltLplening thyrinters.
(*ss**
CATCCOPt O' CONTINUED 0.1 NEXT PAGE **sse)
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 25 j
GUESTION 6.36 (1 00)
If power range detector N44 fails high, the Reactor Control System will respond when in AUTO by driving rods in a.
until the power mismatch error signal balances out with the temperature mismatch error signal.
b.
then back out until the temperature error si3nal is within it's allowable band.
- c. until the power mismatch error signal is within it's allowable band.
d.
Until the operator defeats the N44 signal to the Reactor Control System or all control rods are fully inserted.
i OUESTION 6.37 (1.00)
I Which of the following is a correct statement concerning the source range audio count rate circuit?
3.
If the control room audio count rate is lost due to a detector failure, then the containment audio count rate is lost at the same time.
b.
If the control room audio count rate is lost due to a detector failure, the audio count rate can be restored by a celector switch behind the audio count rate instrument cabinet back panel.
c.
If the control room and10 equot rete is lost due to a detector failure.
the audio count rate can be restored by a selector switch on the front of the audio count rate panel but then the containment audio count rate indte.ation will be lost.
d.
If the control room avdto count rate is lost due to an amplifter fattore. th a control room audio count rate will be lost untti the amplifter is repaired.
OllESTION 6.30 (1.00)
List the 'At Power' reactor trips. Logico and wtpoints are not required.
tsees* CATCGOPt 06 CONTINUED OH NEXT PAGE sesse>
l
l 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 26 l
GUESTION 6.39 (1.00)
Which of the following will cause a turbine runback when at 75 *: Power?
- a. Low stator water cooling pressure.
l b.
High steam generator level.
- c. Decreasing condenser vaceum.
d.
A condensate pump breaker opening.
QUESTION 6.40 (1.00)
List the five sensed pararneters used by the SG Water Level Control System.
QUESTION 6.41 (1.00)
If the turbine has just been latched for a plant startup, placing the OPC Key Switch in the TEST position will a.
trip the governor valves only.
b.
trip the governor and intercept volves.
c.
trip the throttle valves only, d.
trip the throttle and intercept valves.
(s:::s END OF CATCCORf 06 sasse) l l
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i 7.
PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 27
~~~~~~~~~~~~~~~~~~~~~~~~
~~~~ d65ULU55 DEL"6ENTRUL R
GUESTION 7.01 (1.00)
A reactor coolant pump should not be started unless the No. 1 seal delta P is greater than a.
100 psi.
b.
200 psi.
c.
300 psi.
d.
400 psi.
QUESTION 7.02 (1.00)
Prior to transferring rod control from MANUAL to AUTO during a reactor startup, Tave and Tref should be within a.
1 degree F.
b.
3.5 degrees F.
c.
5 degrees F.
d.
10 degrees F.
OllESTION 7 03 (1.00)
Indicate whether each of the following questions are TRUE or FALSE as they pertain to the Estimated Critical Rod Position Pu3ctivity Calance Caleviation procedure.
- i. The current fuel burnop used in the calcolation can be obtained from the OAC.
b.
The baron concentration value used in the calculation can be obt11ned from a Chemistry sample analysis or the boronometer roading.
(sis ** CATECORY 07 CONTINUED ON NEXT PAGE eessa) l l
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l 7.
PROCEDURES - NORMAL, A0 NORMAL. EMERGENCY AND PAGE 20
~~~~~"~~
~
RdD5blUU5ddE~CUNiR L
~~~~
I l
OUESTION 7.04 (1 00)
Which of the following statements is correct concerning the status of the Nuclear Instrumentation Recorder prior to withdrawing control banl-rods for a reactor s t a r tup")
a.
The highest reading source range channel and the highest reading intermediate range channel is selected and the NR-45 chart speed is set to
'H1' speed.
b.
The highest reading source range channel and the highest readinj intermediate range channel is selected and the NR-45 chart speed is set to 'Lo' speed.
c.
The highest reading source range channel and the lowest reading intermediate range channel is selected and the NR-45 chart speed is set to
'Hi' speed.
d.
The highest roadtng source range channel and the lowest reading intermediate range channel is selected and the NR-45 chart speed is uot to
'Lo' 1 peed.
QUESTION 7.05 (1 00) uhon an or, orator depresses the 'LJtch' PU5hbutton f0r a turbine i
s t er top, ho should vorify that tho a.
intercept and reheft stop >ilves and the jovernor and throttle valves open.
i,.s. trtercept end rehest stop valves opon. Jnd the jovernor and theottla valvos r em a t ri c lowd.
c.
totorcept 2nd ciheit Stop valves remath closed, and the jovernor and theettlo valves open.
d.
Intercept snd cohost stop v41ves and tho jovornor " 1ves opone 4nd the throttle valves remain clowd.
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l 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 29 RA656EU55CAE~C6 TR E~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
OUESTION 7.06 (1.00)
Match the following Unit i limits with the appropriate temperature.
The temperature selections may be used more than once.
Column A Column B
- 1. Normal reactor coolant system heatup a.
25 degrees F.
rate limit per hour.
1 b.
50 degrees F.
- 2. Normal pressuri:er heatup rate limit per hour.
c.
70 degrees F.
- 3. The maximum reactor coolant system d.
100 degrees F.
temperature without any reactor coolant pumps in service.
e.
160 degrees F.
4.
The manimum allowable temperature f.
320 degrees F.
differential between the pressurizer and the pressuri:er spray flutd.
5.
The minimum SG temperatore limit when the Reactor Coolant Systein pressure is above 200 psig OUESTION
' 07 (1.00)
A inainreed pump turbine trip test shov1d not be performed unless the iinst load is less than (select t h e ih a : tmum allowablan a.
10%.
b.
15 ';.
c.
5 0 *;.
d.
70%.
Iseres CATECOPr 07 CONTINUCD ON NEXT PAGE sessv)
k 7.
PROCEDURES - NORMAL,. ABNORMAL, EMERGENCY AND PAGE 30
~~~~R Bi5E5515AL CBETR E~~~~~~~~~~~~~~~~~~~~~~~~
l QUESTION 7.08 (1.00)
The perferred method to reduce the oxygen concentration in the reactor coolant when the plant is shutdoun and less than 200 degrees F is by a.
purging the pressurizer with hydrogen.
b.
placing a nitrogen blanket on the volume control tank.
c.
feeding and bleeding the reactor coolant system-.
d.
injecting hydra:ine into the reactor coolant system via the chemical and volume control system.
OUESTION 7.09 (1.00)
During refueling operations, one difference between removin3 a fuel assembly from a position surrounded by fuel assemblies on all sides vs. a fuel assembly with a vacancy on one side is a.
The dog Pendant Control is not used for the fuel assembly surrounded on all sides.
b.
The Fuel Holst Control is not use'd for the fuel assembly with an adjacent vacancy.
c.
The fuel assembly with an adjacent vacancv is offset 1/2 an inden prior to complete sssembly removal from the core.
d.
The fuel as sembly sur r ounded ori all sides is not withdrawn past the Fuel Slow Zone interl,ock.
QUESTION 7.10 (1.00) when draining the cold les accumulators, the water is stored in the a.
containment sump.
b.
refueling cavity.
fuel-ins water storage tank.
c.
d.
reactor coolant drain tank.
(*****
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7o PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 31
~~~~Ed655L6555IL EUNTE5L
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~~~~~~~~~~~~~~~~~~~~~~~~
GUESTION 7.11 (1.00)
The transfer of control of the pressurizer heaters (sub-bank of Group D) to the Safe Shutdown Facility is performed by a.
swapping power supplies of the motor control centers from normal to alternate in the ETA rooms.
b.
swapping a plus type connector from the normal connection to the alternate connection in the SSF.
c.
placing the Pressurizer Heater Selector Switch in the main control room to the ' Remote-SSF* Position.
P acing the Pressurizer Heater Selector Switch in the SSF l
d.
to the ' Local" position.
QUESTION 7.12 (1.00)
Which of the following conditions would warrant an emergency boration of the reactor coolant system?
a.
'Lo Insertion Limit' alarm during Mode 1 operations.
b.
One control rod stuck fully out on a reactor trip.
c.
A stvek open pressurl:et PORV during Mode 1 operations.
d.
Less than 2000 ppm baron in the coolant during Mode 6 operations.
l l
GUESTION 7.13 (1.00) l I
If a " Rod Control Urgent Failure' alarm occurs due to a failure j
in the logic cabinet, the Tave/ Tref mismatch is maintained by e.
controlling turbine load.
b.
taking manual control of individual control rod banks.
c.
taking manual control of individual control rod groups.
d.
boration and dilution of the reactor coolant system.
(*****
CATEGORY 07 CONTINUED ON r4 EXT PAGE *****)
I
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 R 656E66 EdL'EUUTRUE~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
GUESTION 7.14 (1.00)
The immediate operator action for a confirmed loss of a reactor coolant pump seal injection flow is a.
stop the applicable reactor coolant pump, then manually trip t h e t'e a c t o r,
b.
reduce reactor power below P-8, then trip the applicable reactor coolant pump.
c.
trip the reactor coolant pump and initiate safety injection.
d.
shutdown the reactor, then trip the applicable reactor coolant pump.
QUESTION 7.15 (1.00)
Upon Indications of a loss of condenser vacuum, a manual turbine trip is required when condenser vacuum becomes less than______
inches Hg or the exhaust hood temperature becomes greater than______ degrees F.
QUESTION 7.16 (1.00)
The trip bistables of a failed power range detector is placed in the trip condition by a.
Placing the applicable bistable test switch in the ' Test' position in the Reactor Protection Cabinet.
b.
removing the applicable control and instrument power fuses on the power range drawers.
c.
Placing the applicable Power Mismatch Bypass switch to the failed position at the Miscellaneous Control and Indication Panel.
d.
placing the applicable Comparator Channel Defeat switch to the failed channel position at the Detector Current Comparator Panel.
(xxxxx CATEGORY 07 CONTINUED ON NEXT PAGE ***xx)
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PACE 33 i
TROL RADIOLOGICAL CON i
f GUESTION 7.17 (1.00) i If a high activity in the reactor coolant has been identified as a crud burst, the corrective action is to s.
reduce reactor power.
1 i
b.
isolate the mixed bed demineralizer from service and place the cation bed deminerali=er in service.
i l
c.
isolate the mixed bed demineralicer from service and open it's bypass valve.
d.
increase letdown flow to 120 spm.
l l
l GUESTION 7.18 (1.00)
In the event that a Unit 2 load rejection occurs due to PCBs 58, 59, 61, and 62 opening, select the sequence of the given procedural steps to place the unit back on the grid.
)
1.
Place the 6.9KV Mode Select Switches in ' MAN'.
I 2.
Place the 6.9KV Mode Select Switches in 'AUT0*.
j 3.
Transfer the 6.9KV suitchgear loads to normal lineup supplies.
l 4.
Transfer the 6 9KV switchgear loads to 2ATA.
5.
Open 2B generator breaker.
6.
Close PCB 61 and 62.
7.
Close PCB 58 and 59.
l 8.
Depress and hold the " SYNC' pushbutton for the generator i
breaker to be closed.
I o.
Match incoming and running voltage.
l
- 10. Adjust generator speed to rotate the synchroscope slowly 2
in the ' FAST' direction.
- 11. AT the 5 minutes before 12 o' clock position on the synchroscope, close the generator break er.
i a.
2,4 6,7,5,8,9,10,11,1,3.
b.
Ir4,5,6,8,9,10,11,7,3,2.
c.
1,4 5r8,9,10,11,6,7,2,3.
d.
1,4,5,0 10,11,9,6,7,3,2.
)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 34 i
RADIOLOGICAL CONTROL
(
1 GUESTION 7.19 (1.00) i If one reactor coolant pump is lost at 20% reactor power, the j
operator should i
a.
trip the reactor.
i b.
verify a reactor trip has occured, c.
attempt to restart the reactor coolant pump if the cause j
j of the trip is immediately identified and corrected without performing a unit shutdown.
4 l
d.
shutdown the unit prior to any pump restart attempt even if the cause of the trip is immediately identified and l
corrected.
OUESTION 7.20 (1.25) f List the immediate operator actions for a reactor trip.
4 i
00ESTION 7.21
(.75)
List the immediate action steps for a turbine trip when the j
niain control room manual trip does not trip the turbine.
DUESTION 7.22 (1.00)
List the immediate actions required when the reactor cannot be tripped using the control room trip switches.
GUESTION 7.23 (1.25)
List the immediate operator actions for a safety injection.
1 GUESTION 7.24 (1.25)
List the immediate steps for a loss of all AC power.
t f
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35 l
RAD 5ULd55EAL'UUUTR5L
~~~~
i QUESTION 7.25
(.50) a An operator is required to perform the immediate a'ctions of the
}
Safety Injection Procedure for a spurious safety injection. TRUE or FALSE.
1 f
GUESTION 7.26 (1.00)
Which of the following conditions would require all NC pumps to be tripped during a valid safety injection?
a.
If an ND pump is running and the NC system subcooling is 10 degrees F.
l b.
If an NI pump is running and the NC system subcooling is 10
}
degrees F.
c.
If flow is verified from an auxiliary feed water pump and NC system subcooling is less than 0 degrees F.
l d.
If flow is verified from an NI pump and NC system subcooling is less than 0 degrees F.
GUESTION 7.27 (1.00)
Which of the following conditions satisfy the SI termination criteria of the Safety Injection procedure EP-Ol?
NC System Subcooling P2R Level NC System Pressure SG Total Flow 1
a.
25 degrees F 10%
Stable 500 spm b.
60 degrees F 7%
Increasing 550 spm c.
50 degrees F 7%
Stable 400 3pm i
d.
50 degrees F 10%
Increasing 350 spm i
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)
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PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 36
~~~~~~~~~~~~~~~~~~~~~~~~
R 5 5L55i5AL C5 TR L
~~~~
QUESTION 7.28 (1.00)
Which of the following is a characteristic of natural circulation?
I l
- a. SG steam pressure-STABLE or DECREASING.
f
- b. NC system subcooling-LESS THAN O DEGREES F.
- c. NC loop hot les temperatures-STABLE or INCREASING.
- d. NC loop cold les temperatures-GREATER THAN SATURATION TEMPERATURE FOR SG PRESSURE.
QUESTION 7.29 (1.00)
If natural circulation cannot be verified, the operator should a.
increase the steam generator level.
b.
increase the steam dumping rate.
c.
decrease NC system pressure.
d.
start an NC pump.
i QUESTION 7.30 (1.00) i j
During a natural circulation cooldown, the preferred method of NC l
system depressuri:ation is by a.
opening the normal spray valve.
i b.
opening the NV auxiliary spray valve.
j c.
opening the pressurizer PORVs.
f d.
opening the reactor vessel vent valves.
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 37
~~"E5656L66565L 6 ETR6L GUESTION 7.31 (1.00)
If the following colors appeared on the control room critical safety function status tree display, which color coded path has the highest priority?
a.
RED b.
ORANGE c.
YELLOW d.
GREEN QUESTION 7.32 (1.00)
If the following critical safety functions were all displayed oranger which one has priority?
a.
Soberiticality.
b.
Heat Sink.
c.
Integrity.
d.
Inventory.
QUESTION 7.33 (1.00)
A hydrogen bubble formed in the reactor vessel is eliminated by
- a. increasing pressurizer tenperature above core thermocouple readings.
- b. Injecting oxygen into the reactor coolant system via the chemical and.olume control system.
- c. maximizing coolant flow by running all reactor coolant pumps, increasing letdown flow to 120 spm, and placing the cstion bed demineralizer in service in parrallel with the mixed bed demineralicer.
- d. venting the reactor vessel head to the PRT.
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j 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 38 l
--- E 5i5E55fCIE 55 iE5E------------------------
i i
1 QUESTION 7.34 (1.00)
If a 500 mrem pocket dosimeter reads ______ or more, it should be returned to HP for re:eroing.
i a.
120 mrea f
b.
200 mrem c.
250 stem i
d.
300 mrem i
i GUESTION 7.35 (1.00) 1 If you recieve no whole body radiation exposure for the first seven weeks of a calendar quarter, tnen you may recieve up to ______
during the eight week without exceeding the normal whole body l
administrative exposure limit.
a.
100 mrem i
j
- b. 500 mrem j
c.
800 mrem i
d.
1000 mrem i
GUESTION 7.36 (1.00)
Which of the following is a 10 CFR 20 exposure limit?
a.
5 rem / year-whole body.
4 i
b.
1 rem / quarter-whole body.
t c.
13.75 rem / quarter-hands, l
j d.
7 rem / quarter-skin of whole body.
l j
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r_ _ _.,,,,. _ -. -.,,. _, -..,
..._..m._,,
,_,_,_...,_.,.-._s...
-. -. _ _ _ _.. -. - ~. -... -. - - _.
4 l
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7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 39 i
i RADIOLOGICAL CONTROL QUESTION 7.37 (1.00) i Which of the following actions does NOT achieve the goals of the ALARA program?
a.
Performing a trial run on a mock-up for a job that will be l
done in a high radiation area.
b.
Donning radiological respiratory protective equipment prior j
to entering an area of high airborne activity.
l
- c. Moving away from hot spots whenever possible when inside the RCA.
d.
Setting up a
- tag-team
- to perform a job in a high radiation area in order to divide the time spent in the area between two or more people.
QUESTION 7.38 (1.00)
If a portal monitor alarms when you're being counted upon exiting the RCA. you should i
s.
reset the alarm and be counted a second time.
b.
immediately notify Health Physics personnel.
c.
go to the nearest frisker and perform a whole body frisk.
d.
So to the hot change room.
QUESTION 7.39 (1.00) l Radiation Work Permits are issued a.
for routine operations.
b.
for a specific job in a particular area.
l c.
to specify personnel stay times.
d.
to prevent personnel from entering a contaminated area.
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i 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 40 i
~~~~R 656E655EIE EdOTEUE~~~~~~~~~~~~~~~~~~~~~~~~
j GUESTION 7.40 (1.00)
Which of the following statements regarding access ' keys to E::tr a High Radiation Areas is correct?
)
Extra High Radiation Area keys shall be issued to HP personnel a.
i only, during non-emergency conditions.
l b.
Extra High Radiation Area keys are issued on a daily bases and are only transferable to your shift relief.
l c.
The Extra High Radiation Area key locker includes keys for i
Radiographic Exposure Devices.
i f
d.
To be issued an Extra High Radiation Area key, a signed statement for issue approval by an HP supervisor must be l
documented on your Daily E::posure Time Record.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 41 4
4 i
GUEST'.ON 8.01 (1.00) l Controlled leakage as defined by Technical Specifications refers to t
i a.
letdown flow.
i i
b.
liquid radwaste release flow.
j c.
reactor coolant pump seal water flow.
l d.
excess letdown flow, i
4 QUESTION 8.02 (1.00) l Match the following Tech Spec defined leakage with the associated f
leak rate limit. Assume normal operating reactor coolant system j
temperature and pressure. The limits may be used more than once.
l COLUMN A COLUMN B 1
1.
a.
O spm.
2.
Controlled leakage.
b.
1 spm.
3.
Identified leak age.
c.
10 gpm.
4.
d.
40 3pm.
)
5.
Primary-to-secondary e.
500 opd.
leakage through one SG.
GUESTION 8.03 (1.00) 7 j
To prevent entering a Technical Specification action statement, j
the Guadrant Power Tilt Ratio shall not exceed ______ when reactor power is above 50%.
a.
1.00 b.
1.02 1
)
c.
1.05 l
d.
1.09 i
j
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 42 Bo j
l4
(
GUESTION 8.04 (1.00)
I If the reactor coolant system pressure exceeds 2735 psig when in Mode 3, Technical Specifications requires the pressure to i
be reduced to within the limit within j
I a.
5 minutes.
I b.
15 minutes.
a
- c. 30 minutes.
- d. one hour.
QUESTION 8.05 (1.00) r j
The specification requiring the upper head injection accumulator j
isolation valves to be opened applies f
j a.
in Mode 1 only.
1 j
b.
only when reactor power is above 467..
?
c.
in Modes 1 and 2 only.
e i
d.
in Modes 1,
2, and 3, if pressuri er pressure is above 1900 psig.
i OUESTION 8.06 (1.00)
Which of the following cold les accumulator specifications are different between Unit 1 and Unit 2?
I a.
The nitrogen cover-pressure minimum pressure.
b.
The minimum boron concentratidn.
c.
The number of water level channels required to be operable.
1 d.
The number of pressure channels required to be operable.
4 I
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~. _ _ _ _ _ _
, _ _. ~. _ _
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 43 l
1
)
GUESTION 8.07 (1.00) l The Unit 2 acceptable Axial Flux Difference target band changes
- a. with core age.
b.
when changing from base load to non-base load operation.
l 1
i
- c. when changing from relaxed axial offset control to non-
{
relaxed axial offset control.
l
- d. when power level is increased from 50% to 75%.
1 t
GUESTION 8.08 (1.00) j If control power is lost to a pressurizer power operated relief valve while in Mode 1,
a.
no action is required by Tech Specs provided another PORV I
is operable and all pressurl:cr code safety valves are operable.
i b.
Tech Specs require the the power supply to be removed from the associated block valve after verifying it to be open, if the PORV is not made nperable within one hour and continuous j
operation is desirable.
c.
Tech Specs require the associated block valve to be shut l
and its' power removed if the PORV is not made operable l
{
within one hour and continuous operation is desirable.
l d.
Tech Specs require action to be initiated within one hour 4
I to place the plant in at least HOT STANDBY within the i
following hour if the PORV is not made operable.
i GUESTION 8.09 (1.00)
The Technical Specification maximum heatup rates per hour for the Reactor Coolant System are a.
100 degrees F for Units 1 and 2.
i e
b.
100 degrees F for Unit 1 and 60 degrees F for Unit 2.
I
~
c.
60 degrees F for Unit 1 and 100 degrees F for Unit 2.
I d.
60 degrees F for Units 1 and 2.
i
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 44 GUESTION 8.10
(.50)
Tech Specs does NOT require any action if one control rod is immovabler provioed the immovable rod is within 12 steps of it's group step counter demand position. TRUE or FALSE?
QUESTION 8.11
(.50)
The shutdown bank demand position indication CANNOT be used as the one required Rod Position Indication system when in Mode 3 with the shutdown banks withdrawn. TRUE or FALSE?
OUESTION 8.12 (1.00)
If the lowest operating loop Tave drops below
, Tech Specs allow to restore Tave within the limit or proceed to place the unit in hot standby.
t a.
557 degrees F, 30 minutes b.
557 degrees F.
15 minutes c.
551 degrees F, 30 minutes d.
551 degrees F, 15 minutes GUESTION 8.13 (1.00)
The Shutdown Margin shall be greater than or equal to ______ %
delta k /k for Unit i and ______
% della k/k for Unit 2 when both units are in Mode 1.
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 45 i
i OUESTION 8.14 (1.00) l If one overtemperature delta T instrument channel is inoperable j
and placed in the trip condition, then l
a.
the plant must be shutdown when the survie11ance testing for j
the operable channels become due, if it is not restored to an operable status.
1 b.
the operable channels must be tested one at a time to prevent j
a reactor trip from occurring.
c.
the inoperable channel may be bypassed in order to perform the surviellance testing on the operable channels.
j d.
the surviellance test interval may be extended up to 100%
i to allow time to restore the inoperable instrument to an i
operable status.
j i
t i
j QUESTION 8.15 (1 00) l Turbine overspeed protection is required by Technical Specifications j
i because an overspeed condition could cause i
a.
turbine components to become missles which may penetrate the turbine casing, allowing higher than allowable off-site i
doses with an assumed maximum allowable fuel cladding and j
primary-to-secondary leakage.
i b.
turbine components to become missiles, which may damage.
i safety-related equipment.
i c.
the reactor thermal power to e::ceed the limits in the 1
Unit's License.
i j
d.
the reactor to go prompt critical.
l 3
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I (unxxx CATEGORY 08 CONTINUED ON NEYT PAGE
- )
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,--,y
,-n
,-,,,,,,_,mrm
,--,-,--,.,_- _ m
,.-,,m.,
m,-...- ---- +.
,m
, _.,~ _,- ~ - - -
_L.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 46 GUESTION 8.16 (1.00)
When a fire supression Spray / Sprinkler System is declared inoperable for a portion that protects an area containing redundant safety-related equipment, the required action is to s.
commence a Unit shutdown for the applicable unit within one hour.
b.
establish an hourly fire watch patrol for the affected area.
c.
establish a continuous fire watch with backup fire supression equipment in the affected area within one hour.
- d. los ambient temperature readings for the affected area hourly.
QUESTION 8.17 (1.00)
Which of the following is NOT a required specification to meet the operability requirements for the Boric Acid Storage System?
a.
A maximum boric acid solution temperature.
b.
A minimum borated solution volume.
c.
A maximum baron concentratton in solution.
d.
A minimum boron concentration in solution.
QUESTION 8.18 (1.00)
Indicate whether each of the following statements is TRUE or FALSE.
a.
When in Mode 6, Tech Specs requires the Source Range Neutron Flux nonttors to have their Alarm Setpoints at 0.5 decade above the steady-state count rate, to perform Core Alterations.
b.
When in Mode 6, Tech Specs requires only visual indication of neutron flu: from the Source Range Monitors in the Control Room, and only audtble Source Range indication in the Containment, to perform Core Alterations.
(***** CATEGORY OS CONTINllED ON NEXT PACE
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 47 QUESTION 8.19 (1.00)
If the Containment Purge System Noble Gas Activity Monitor (EMF-39) is declared inoperable, then a.
no purging or venting can be done via this pathway.
I b.
two independent containment air samples must be analyzed and two qualified people must verify the purge lineup to allow pur3 n3 via this pathway.
i c.
the flow rate must be verified at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to allow purging via this pathway.
d.
grab samples must be taken and analyzed at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to allow purging via this pathway.
QUESTION 8.20 (1.00)
How many 125 volt DC channels are required by Tech Specs to be operable and energized to prevent entering an action statement when in Mode 1?
a.
1.
b.
2.
c.
3.
d.
4 QUESTION 3.21 (1.00)
A quar +erly surveillance requirerrtent of Tech Specs may be extended up to ______ days without declaring the component inoperable due to the surveillance testing not being performed.
a.
O b.
23 c.
32 d.
41
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80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 48 GUESTION 8.22 (1.00)
Provide the minimum number of individuals required by Tech Specs for the following positions to operate both Units at full power.
a.
Shift Supervisor (s).
b.
______ SR0(s) (not includinj the Shift Supervisor (s)).
c.
R0(s).
- d. ______ A0(s).
e.
STA(s).
QUESTION 8.23 (1.00)
Which of the following is the responsibility of the Shift Supervisor in reguard to the use of operating procedures performed during the shift?
a.
Ensures the Working Copy File has a sofficient number of unused copies available.
b.
Initiates all procedure changes required during the shift.
c.
Reviews and approves all completed procedores.
d.
Compares the Working Copy to the Control Copy to ensure all changes are entered prior to use.
QUESTION 3.24
't.00)
Deviations from a procedoro during normal plant operations 3.
ara not allowed.
b.
can be made if the original intent of the procedure is satisfied.
c.
can be made with verbal approval of two licensed operators.
d.
can be made with verbal approval of two licensed operators if at least one of them is SRO licensed.
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Bo ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 49 GUESTION 8.25 (1.00)
If a step in an operating Fracedure with a signoff is not performed, the step should be marked N/A and initialed by a.
the operator performing the procedure.
b.
any operator on shift with an RO or SRO license.
c.
only an operator on shift with an SRO license.
d.
only a supervisor with an SRO license.
QUESTION 8.26 (1.00)
Which of the following is NOT in accordance with the Operations Managemen*
Procedures for transferring initials from a completed Working Copy of a procedure to an updated issue of the same procedure?
a.
The transfer of initials must be requested by the Shift Supervisor.
b.
Any licensed operator can transfer the initials.
c.
Oniv the initials of the person making the transfer need to be on the new Working Copy.
d.
The person making the transfer can initial both the Valve Check list and the Independent Verification Valve Checklist of the new Working Copy.
QUESTION 3.2'
(.501 If a Control Copy of a procedure has no change forms attached, then the procedure has had no changes made to it. TRUE or FALSE?
1
(***** CATEGORY 08 CONTINUED ON NFYT FAGE *****'
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 50 QUESTION 8.28 (1.00)
The reactivity curves are administratively controlled to assure the correct curves are used for each Unit by keeping separate Data Books -one for Unit 1 and one for Unit 2.
a.
b.
color coding the pages of Unit 1 curves white and Unit 2 curves pink.
c.
the operator initialling a checklist whenever reactivity curves are used.
d.
the Shift Supervisor's signature of authart:ation in the Reactivity Balance Calculation Procedure which specifically author 1:es the use of curves by Unit Number and Revision Number.
QUESTION 8.29 (1.00)
Information Stickers placed on the Main Control Boards are controlled by a.
an Information Sticker Logbook.
b.
a section in the Work Request Sticker Logbook.
c.
a section in the White Tag Sticker Logbook.
d.
a section in the Red Tag Sticker Logbook.
QUESTION 8.30
(
.50)
The position of valves essential to safety do NOT have to have their positions independently verified. TRUE or FALSE?
l QUESTION 3.31 (1.00)
Entries made in the Operators Logbook concerning equipment shared by both Units should be made in a.
the Unit 1 R0 Logbook.
b.
the Unit 2 R0 Logbook.
c.
both Unit 1 and Unit 2 RO Logbooks.
d.
the Common Unit Logbook.
l
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 51 GUESTION 8.32 (1.00)
The ' Buddy System' for entering the Reactor Building is required as soon as a.
the Reactor Coolant System temperature exceeds 200 degrees F.
b.
the Reactor Coolant System temperature exceeds 350 degrees F.
c.
the reactor power level exceeds 5%.
d.
the reactor power level exceeds 10%.
QUESTION 8.33 (1.00)
Post-Trip Reviews are performed by 3.
any personnel licensed by the NRC on the applicable Unit.
b.
personnel from a list of qualified reviewers maintained by the Reactor Engineer only.
c.
STAS.
d.
STAS, only if they hold a valid SPO license.
l QUESTION 8.34 (1.00)
The person responsible to maintain accountability of Operations personnel during an emergency plant condition requiring a Site Assembly is a.
the Shift Supervisor.
b.
the Superintendent of Operations.
c.
the Station Manager.
d.
the Fecovery Manager.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND
- 1. IMITATIONS PAGE 52 QUESTION 8.35 (1.00)
During a General Emergency, off-stte doses are projected for areas as far as ______ miles from the plant.
a.
2 b.
5 c.
7 d.
10 QUESTION 8.36 (1.00)
List the two Evacuation Sites.
QUESTION 8.37 (1.00)
The highest allowable planned whole body e:tposure that may be received, with appropriate approval, by a volunteer to save a life is a.
25 rems.
b.
75 rems.
c.
150 rems.
d.
375 rems.
QUESTION 3.38 (1.00)
Potassium Iodide tablets are made availabe to a.
reduce the effects of beta rad 1ation to the lens of the eye.
b.
reduce the effects of neutron radiat. ion to soft body tissue.
c.
reduce the a.n o u n t of krypton collected in bone marrow, d.
reduce the 3 mount of radiciodine collected in the thvroid gland.
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ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 53 i
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QUESTION 8.39 (1.00)
The Special Orders Logbook provides the shift personnel i
a.
a list of priority work items to be performed during the shift.
f b.
special instructions and guidlines to perform certain procedures 1
and evolutions.
i I
l
- c. with a means to write a temporary procedure.
d.
with a way to keep track of Tech Spec Action Statements entered.
[
}
l QUESTION 8.40 (1.00) l 1
t l
Which of the following 15 NOT a required one hour or four hour reportable
[
event?
I j
a.
The initiation of a plant shutdown in accordance with a Tech Spec i
j1 Action Statement.
~
b.
A routine radioactive waste gas release.
j
- c. A deviation from Tech Specs required in an emergency situation to l
protect the public health and safety when there is no action i
i provisions in the license conditions or Tech Specs to adequately l
provide the protection.
[
d.
Transporting a contaminated person to a hospital.
[
OVESTION 8.41 (1.00) i When the dispatcher ordera switchyard breakers to remain opent they are identified so by tagging them with
?
I 1.
red t,ags.
I b.
yellow tags.
j c.
white tags.
[
j d.
blue tags.
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8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 54 GUESTION 8.42 (1.00)
Tcmporary modifications must be identified with a tag when they are installed a.
for more than one hour.
b.
past a shift change.
c.
by a Work Request in lieu of a procedure.
d.
on safety-related equipment.
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(***r*********
END OF EXAMINATION
- )
l 1
l TABLE D Ib Properties of Dr3 Saturated Steam (amta= ed; i
Temperature l
5pecific soi me Enthalp>
Enteop>
i Temp.
N'
$at
$at 5at Sai Sai Sat Esap
- F Esap psia
,,,, g g,,,g y,,,4 s
r r
i a
e s1
- e
- e h
h h
e 32 0.08854 0 01602 330t 0 00 1075 8 1075 8 0 0000 2.1877 2.1871 35 0 09995 0.01602 2947 3 02 1074.1 1077.1 0.006l 2.1709 2.1770 40 012J70 0.01602 2444 8 05 1071.3 6079.3 0 0162 2.1435 2.1597 45 0.14752 0 01602 20 M 4 1106 1068 4 1081.5 0 0262 2.1167 2.1429 50 0 17811 0.01603 1703 2 18 07 1065.6 1083.7 00M1 2.0903 2.1264 60 0.2M3 0.01604 1206.7 28 06 1059.9 1088.0 0.0$$$
2.0393 2.0948 10 0 3631 0.0160u 867.9 38.04 1054.3 1092.3 0.0745 1.9902 2.0647 to 0.5069 0 01608 6331 48 02 1048.6 1096 6 0.0932 1.9428 2.0h0 90 0.6982 0 016l0 468.0 57.99 1042.9 1100.9 0.1115 1.8972 2.0087 100 0.9492 0.01613 350.4 67.97 1037.2 l105.2 0.1295 f.8531 l.9826 110 1.2748 0 01617 265.4 77.94 1031.6 1800.5 0.1417 l.8106 l.9577 120 1.6924 0.01620 20127 87.92 1025.8 t il).7 0.1645 1.7694 l.9339 130 2.2225 0 01625 157.34 97.90 1020 0 1817.9 0.1816 l.7296 l.9112 140 2 8886 0 01629 123 Of 107 89 1014.1 1122.0 0.1984 l.6910 13894 150 3.718 0 01634 97 07 117.89 1006.2 1126.1 0.2149 I6537 1.8685 160 4.741 0.01639 77.29 127.89 1002.3 1830.2 0.2311 1.6174 1.8485 170 5.992 0.01645 62.06 137.90 996.3 1134 2 0.2472 1.5822 1.8293 ISO 7,510 0.01651 50.23 147.92 990.2 1838.1 0.2630 1.5400 1.8109 190 9.339 0.01657 40 96 157.95 984.1 1842.0 0.2785 1.5147 f.7932 200 11.526 0 01 % )
33.64 167.99 977.9 1145.9 0.2938 1.4824 1.7762 210 14.123 0 01670 27.82 178 05 971.6 1149.7 0.3090 1.4508 1.7598 212 146%
0 01672 26 to 180.07 970.3 1150 4 0.3120 1.4446 1.1546 220 17.186 0 01677 23 15 188.13
%5.2 1853 4 0.3239 f.4201 1.7440 230 20.780 0.01684 19 382 198.23 9588 1157.0 0.3387 1.3901 1.7288 240 24 969 0.01692 16 323 208.34 952.2 1160.5 0.3531 1.3609 f.7140 e
250 29 825 0 01700 13.821 216.48 945.5 1164 0 O M75 l.3323 1.6998 260 35 429 0.0 709 11.763 228.64 9 38.7 1167.3 0.3817 1.3043 1.6880 270 41.858 0.04717 10.061 238 84 931.8 1870.6 0.3958 1.2769 I6727 280 49.203 0.01726 8645 249.06 924.7 1873 8 0 4096 1.2501 1.6587 290 57.5 %
001735 7.461 259.31 917.5 1876 8 0 4234 1.2238 1.6472 300 67.013 0 01745 6 466 269.59 910.1 1879.7 04M9 1.1980 1.6350 310 77.68 0.01755 5.626 279.92 902.6 1182.5 0 4504 1.1727 f.6231 320 39 66 0 01765 4 914 290.28 894.9 1185.2 04.37 1.l478 1.6115 330 103 06 0.01776 4 307 V10.68 887 0 1887.7 0 4769 1.1233 f.6002 340 118.01 0.01787 3.788 311.l3 879 0 18901 0 4900 10992 1.5891 l
i l
l
TABLE D-lb Properties of Dr> Saturated Steam tennimurd.
Temperature 5 pudic iolume E nth alp >
kntrops w
Temp.
I*'
- F Sat Sai Sat Sai Sai Sat Etap hquid sapor liquid
'8P sapor liquid
,,p,,
t n
h h
h s
s.
n o
e e
a u
o 350 134 63 001799 3.342 32163 870 7 1192.3 0.5029 1.0754 1.5783 360 153 04 0 01811 2.957 332.18 852.2 1194 4 0 5158 10519
- 1. % 77 370 173.37 0 01823 2.625 342.79 853 5 1196 3 0 5286 1.0287 1.5573 340 195.77 0 018 %
2.335 353 45 844 6 18981 0.5413 1.0059 1.5471 390 220.37 0 01850 2 0836 364.17 835 4 1199 6 0 5539 0 9832 1.5371 400 247.31 0018W 1 8633 374 97 826 0 1201.0 0 5664 0 9608 1.5272 410 276.75 0 01878 1.6700 385 83 816 3 1202.1 0 5788 0 9346 1.5174 420 308 83 0 01894 1.5000 396.77 806 3 12031 0 5912 0.9166 1.5078 430 343 72 001910 1.3499 407.79 796 0 12038 06035 0 8947 1.4982 440 381.59 0 01926 1.2171 418 90 785 4 1204 3 0 6138 0 8710 14887 450 422 6 0 0194 1.0993 4301 7745 1204 6 0 6280 0 8313 14793 460 466.9 0 01 %
0 9944 441 4 763.2 1204 6 0 6402 0 4298 1.4700 470
$14 7 00198 0 9009 452 8 751.5 1204 3 0 6523 0 8083 1.4606 440
%61 0.0200 0 8172 4M 4 739 4 1203 7 0 6645 0 7868 1.4513 490 621.4 0.0202 0.7423 476 0 726 8 1202 8 0 6766 0 7653 14419 500 680.8 0 0204 0 6749 487 8
?l) 9 1201 '
0 6887 0 7438 I4325 520 812.4 0 0209 0 5594 511.9 686 4 l198 2 0 7130 0.7006 1.4136 l
540
%2.5 0 0215 04649 536 6 656 6 1193 2 0.7374 06%8 l.3942 560 1833 1 0 02:1 0 3868
%2.2 624.2 1186 4 0.7621 0 6121 1.3742 580 1325 8 0 0228 0.3217 588.9 588 4 1177.3 0 7872 0 % 59 1.3532 600 1542.9 0 0:36 0 2668 610 0 548.5 1165.5 0 8131 0.5176 1.3307 620 1786 6 0 0247 0.2201 646.7 503 6 1150 3 0 8398 0.4664 1.3062 MO 2059.7 0 0260 0.1798 678.6 452 0 1130 5 0 8679 0 4110 1.2789 660 2365 4 0 0278 0.1442 714.2 390 2 1104 4 0 8987 0 3485 1.2472 640 27081 00305 0.1115 757.3 309 9 1067 2 09351 0 2719 1.2071 700 3093 7 0 0369 0 0761 823.3 172.1 995 4 0 9905 0 1484 1.1389 705 4 3206 2 0 0503 0 0503 902.7 0
902.7 1.0580 0
1.0580 l
e l
O
TA Bt.F D-1.i' Properties of Dr3 Saiur.ned Steam +
Pressure specires soi me E nthstr>
Enitor>
Temp.
l Pr'55 g
g g
g g
g P5'*
- F ngg
,,g, g,,,, g Ever g,,, g Evap e
a i,
4, 4,
4, 8,
s.
8, l.0 101.74 0.01614 333 6 69 70 1036.3 1106 0 0.1326 1.5456 1.9782 2.0 126 08 0.01623 17373 93 99 1022.2 1116.2 0.1749 I.7451 1.9200 30 141 48 0 01630 Ils 71 109 37 1013.2 1122.6 0 2006 1.6855 I.8863 40 152.97 0 016 %
90 63 120 b6 1006 4 l 127.3 0.2198 1.6427 1.8625 5.0 162.24 0.01640 73.52 130 13 1001.0 l 131.1 0.2347 2 6094 1.8441 60 170.06 0 01645 61 98 137 96 996.2 l134 2 0.2472 1.5820 1.8292 70 176 85 0.01649 53 64 144 76 992l i136 9 02581 1.55 %
1.8167 8.0 182.86 0 01653 47.34 150 79 988.5 1139.3 0.2674 1.5383 I.8057 9.0 188.28 0 01556 42.40 156 22 985.2 l 141.4 0.2759 1.5203 1.7962 10 193.21 0 01659 38 42 168.17 982.1 1943.3 0.2835 1.5041 1.7876 14 696 212.00 0 01672 26 80 180 07 970.3 iI50 4 0 3l20 1.4446 1.7M6 15 213 03 0 01672 26.29 181.11
%9.7 1850 8 0.3135 1.4415 1.7549 20 227.%
0 01683 20 089 196.16 960.1 1156 3 0.3356' l.3962 l.7319 25 240 07 0 01692 16.30) 20k42 952 1 1160 6 0 3533 1.3606 I.7139 30 250 33 0 01701 13.746 218 82 945.3 1164 1 0 3680 1.3313 8.6993 35 259.28 0 01708 11.898 227.91 939.2 Il67.I 0.3807 1.3063 1.6810 40 267.25 0 01715 10 498 236 03 933.7 l169.7 0.3919 1.2844 1.6763 45 274 44 0 01721 9 401 243 36 928 6 1172.0 0 4019 1.2650 I.6669 50 281.0 L 0 01727 8 515 250 09 924 0 1174.1 0 4110 1.2474 1.6585 55 287.07 0 01732 7.787 256 30 919 6 l175.9 0 4193 1.2316 1.650s 60 29271 0.01738 7.175 262.09 915.5 1177.6 0.4270 1.2168 1.6438 65 297.97 001743 6 655 267.50 911.6 1179. t 04342 1.2032 1.6374 70 302.92 0.01748 6.206
,272 61 907.9 l180 6 0 4409 1.1906 1.6315 75 307.60 0 01753 5 816 277.43 9045 l 181.9 0 4472 I.1787 1.6259 80 31203 0 0l757 5.472 282.02 901.1 l 183.1 0 453I 1.167e 1.6207 85 316.25 0.01761 5.168 286.39 897.8 l184 2 0 4587 1.1571 1.6158 90 320 27 0.01766 48%
290.56 894 7 1185.3 0 4641 1.1471 1.6812 95 324 12 0 01770 4 652 294 56 891.7 I I 86.2 0 4692 1.1376 1.6064 800 327.8l 0 01774 4 432 294 40 8888 1887.2 0 4740 1.1286 1.6026 ll0 334 77 0 01782 4 049 305.66 883.2 l188 9 0 4832 1.1117 1.5948
l 1
TABLE D-la Properties of Dry Saturated Steam teontinued)
Pressure Specifie volume E mh alp >
t mrop.s g
- Temp,
'F Sai Sat sat
$st Sat 5st Essp Evap hout vapor Nuid
,,po, gg h
h.
A, s
s s,
r t
e e
r e
r 120 341.25 0.01789 3.728 312.44 877.9 1190 4 0 4916 1.0962 1.5878 130 347.32 0 01796 3 455 318.81 872.9 1191.7 0 4995 1.0817 1.5812 140 353 02 0 01802 3.220 32482 869.2 1193 0 0.5069 1.0682 1.5751 150 358 42 0.01809 3.015 330 51 863.6 Il941 0 5138 1.0556 1.5694 160 M3.53 0 01815 2.834 335.93 859.2 Il95.1 0 5204 1.0436 1.5640 170 4 8.41 0 01822 2.675 341.09 854 9 11 % 0 0.5266 1.0324 1.5590 180 373 06 0.01827 2'532 346 03 850 8 1896 9 05325 1.0217 1.5542 190 377.51 0 01833 2.404 350.79 846 8 1l97.6 0.5381 1.0116 1.5497 200 381.79 0.01839 2.288 355.36 843 0 1198 4 0 5435 1.0018 1.5453 250 400.95 0 01865 1,8438 376 00 825.1 1201.1 0.% 75 0.9588 1.5263 300 417.33 0 0l890 1.5433 393 84 809 0 1202.8 0.5879 0.9225 1.5104 350 431.72 0 01913 1.3260 409.69 794.2 1203.9 0 60 %
0 8910 1.4966 400 444.59 0 0l93 1.1613 424.0 780.5 1204.5 0.6214 0.8630 1.4844 450 4 %.28 0 0l95 1.0320 437.2 767.4 1204 6 0 63 %
0 8378 1.4734 500 467.01 0.0197 0.9278 449.4 755 0 1204 4 0 6487 0.8147 I.4634 550 476.94 0.0199 0.8424 460.8 7431 1203.9 0.6608 0 79M l.4542 600 486.21 0 0201 0.7698 471.6 731.6 1203.2 0.6720 0.7734 1.4454 650 494.90 0 0203 0.7083 481.8 720.5 1202.3 0 6826 0.7548 l.4374 700 503.10 0.0205 0 6554 491.5 709 7 1201.2 0 6925 0.7371 1.4296 750 510.86 0.0207 06092 500 8 699.2 1200.0 0 7019 0.7204 1.4223 800
$18.23 0 0209
- 0. % 87 509.7 688.9 1198 6 0.7108 0.7045 1.4153 850 525.26 0.0210 0.5327 518.3 678.8 1197.1 0.7194 0.6891 1.4085 900 531.98 0.0212 0 5006 526 6 668 8 1895 4 0.7275 0 6744 1.4020 950 538 43 0 0214 0 4717 534 6 659.1 1193.7 0.7155 0 6602 1.3957 1000 544 61 0 0216 0.44 %
542 4 649.4 1191.8 0.7430 0.6467 1.3897 1100 5 % 31 0.0220 0 4001 557.4 630 4 II87.7 0.7575 0 6205 1.3780 1200 567.22 0.0223 0.3619 571.7 611.7 1183.4 0.7711 0.59 %
l.3667 1300 577.46 0 0227 0.3293 585 4 593.2 1178 6 0 7840 0 5719 1.3559 1400 587.10 0 0231 0.3012 598.7 574.7 1873 4 0.7963 0 5491 1.3454 1500 5%.23 0.0235 0 2765 611 6 5%.3 1167.9 0 8082 0 5269 l.3351 2000 635.82 0 0257 0.1878 671.7 463 4 18351 0 8619 0.4230 1.2849 2500 668.13 0.0287 0.1307 730 6 360.5 1091.1 0 9126 0 3197 1.2322 3000 695.%
0.0346 0 0858 802.5 217.8 1020.3 0 9731 0.1885 1.1615 3206.2 705.40 0.0%)3 0.0503 902.7 0
902.7 1.0580 0
1.0580 1
I l
l ai.i w
Properues of Superheated Steam
- t.t prem, Tempw sweM (Saa. wYp.*F) 200 300 400 300 400 MD 900 900 1000 1100 8300 1400 v.
392.6 452.3 512.0 571.6 631.2 690.8 750 4 409.9 869.5 929.1 988.7 1107.8 l
A..
1850.4 1195.8 1241.7 1288.3 1335.7 1383u 1432.8 1482.7 1533.5 1585.2 1637.7 1745.7 (101.74) s.
2.0512 2.1153 2.1720 2.2233 2.2702 2.3137 2.3542 2.3923 2 4253 2.4625 2.4952 2.55M e.
78.16 90.25 102.26 114.22 126.16 138.10 150 03 161 95 173 57 185.79 197.71 221.6 5
A.
1148.8 1195.0 1241.2 1288.0 1335.4 1383 6 1432.7 1482 6 1533 4 1585.1 1637.7 1745.7 (162.24) s.
l.8718 1.9370 1.9942 2.0456 2.0927 2.1361 2.1767 2.214n 2.2509 2.2851 2.3178 2.3792 38.85 45.00
$l.04 57.05 63.03 69 01 74 9s so 95 no 92 92 88 98.84 110.77 10 A.
1846.6 1193.9 1240 6 1287.5 1335.1 13814 1432.5 1482.4 1533 2 1585 0 1637.6 1745.6 (193.21) s.
1.7927 I8595 1.9172
- 1. % 89 2.0160 2.0596 2.1002 2.1.tn i 2 5744 2.2068 2.2413 2.3028 y,
30.53 34 68 38.78 42.86 46 94 51.00 55 07 59 13 63.19 67.25 75.37 14.696 A.
1192.8 1239.9 1287.1 1334 8 1383.2 1432.3 1482.3 1533.1 1584 8 1637.5 1745.5 (212.00) s.
1.8160 1.8743 1.9261 1.9734 2.0170 2.0576 2.0958 2.1319 2.1662 2.1989 2.2603 22.%
25.43 28 46 31.47 34 47 37.46 40 45 43 44 4 42 49.41 55.37 20 A.
l 191.6 1239.2 1286 6 1334.4 1382.9 1432.1 1482.1 1533 0 1584 7 1637.4 1745.4 (227.%) s.
1.7808 1.8396 1.8918 1.9392 1.9829 2.0235 2.0615 2 0978 2.132l 2.1648 2.2263 II.040 12.628 14 168 15.688 17.198 18.702 20.20 21.70 2120 24.69 27.68 40 A.
1186.8 1236.5 1284.8 1333.1 1381.9 1431.3 1481.4 1532 4 1584 3 1637.0 1745.1 (267.25) 5.
1.6994 l.7608 1.8140 1.8619 1.9058 l.9467 1.9850 2.0212 2.0555 2.08u3 2.1498 v.
7.259 8.357 9 403 10.427 11.441 12.449 13.452 14 454 15 453 16.451 18.446 to A.
1881.6 1233.6 1283.0 1331.8 1380.9 I410.5 1480 8 1531.9 15ni n 16M 6 1744.8 (292.71) s.
l.6492 1.7135 1.7678 1.8162 1.8605 1.9015 1.9400 1.9762 2.0100 2.0434 2.1049 y.
6.220 7.020 7.797 8.562 9.322 10.077 10.830 II.5n2 12.332 13.830 40 A.
1230.7 1281.1 1330.5 1379.9 1429.7 1480.1 1531.3 15n3 4 16 %.2 1744.5 (312.03) s.
1.6791 1.7346 1.78 %
1.8281 1.8694 1.9079 1.9442 1.97n7 2.0115 2.0731 v.
4 937 5.589 6.218 6.835 7.446 8.052 8.6%
9.259 9.860 1l 000 100 A.
1227.6 1279.1 1329.1 1378.9 1428.9 1479.5 1530.8 1582.9 1635.7 1744.2 (327.81) s.
l.6518 1.7085 1.7581 1.8029 1.8443 1.8829 1.9193 1.9538 1.9867 2D484 4 081 4.6%
5.165 5.683 5.195 6.702 7.207 7.710 8.212 9.214 120 A.
1224 4 1277.2 1327.7 1377.8 1428.1 1478.8 1530.2 1542.4 1635.3 1743.9
. (...... {.......j l.6281 [ l.6469 l 7310 l.7822
, l.8237 2.4625 1.8980 1.9335 1.9664 2.0281 041.2% a v.
3 468 3.9 54 4.413 4.861 5 301 5.738 6.172 6.404 7.035 7 894 140 A.
1221.1 1275.2 1326 4' 1376.8 1427.3 1478.2 1529.7 1541.9 1634.9 1743 5 (353.02) s.
l.6087 1.6683 1.7190 1.7645 1.8063 1.8451 1.8817 1.9163 1.9493 2.0110 O
v.
3.008 3.443 3 849 4.244 4.631 5.015 5.3%
5.775 6.152 6.906 160 h.
1217.6 1273.1 1325.0 1375.7 1426.4 1477.5 1529.I 1581.4 1634.5 1743.2 (363.531 s.
1.5908 1.6519 l.7033 1.7491 1.7911 1.8301 I.8667 1.9014 1.9344 1.9962 i
v.
2.649 3 044 3.411 3.764 4.110 4 452 4.792 5.129 5.466 -
6.1%
180 4.
1214 0 1271 0 1323.5 1374.7 1425 6 1476 8 1528.6 1581.0 1634.1 1742.9 (373.06) s.
l.5745 l.6373 1.6894 1.7355 1.7776 1.8167 1.8534 1.8882 1.9212 1.9831 v.
2.361 2.726 3 060 3.380 3 693 4 002 4 309 4 613 4.917 5.521 200 A.
1210.3 1268 9 1322.1 1373 6 1424 8 1476.2 1528 0 1580.5 1633.7 1742.6 (381.79) s.
I.5594 1.6240 1.6767 1.7232 1.7655 1.8048 8.8415 1.8763 1.9@4 1.9783 v.
2.125 2.465 2.772 3 066 3352 3 634 3.913 4.191 4 467 5.017 -
220 4.
1206.5 1266.7 1320.7 1372 6 1424 0 1475.5 1527.5 1580.0 1633 3 1742.3 (389 86) s.
1.5453 l.6117 I6652 1.7120 1.7545 1.7939 1.830s 1.8656 1.8987 1.9607 e.
1.9276 2.247 2.533 2 804 3 06n 3 327 3.584 3 839 4 093 4.597 240 A.
1202.5 1264 5 1319 2 1371.5 1423 2 1474 8 1526 9 1579 6 1632.9 1742.0 1
(39737) s.
l.5319 1.6003 1.6546 1.7017 1.7444 1.7839 I8209 1.8558 1.8889 1.9510
{
2.063 2.330 2.582 2 827 3 067 3305 3.541 3.776 4.242 1
260 h.
1262.3 1317.7 1370 4 14223 1474.2 1526.3 15791 1632.5 1741.7 1
(404.42) s.
1.5897 1.6447 1.6922 1.7352 1.7748 i siis I.8467 1.8799 1.9420 l.9047 2.156 2392 2 621 2 845 3 066 3 286 3.504 3 938 280 A.
12e00 1316.2 1%9.4 1421.5 1473.5 1525.8 1578 6 1632.1 1741 4 (411.05) s.
1.57 %
1.6354 1.6834 1.7265 1.7662 1.8033 1.8383 1.8716 1.9337 y.
I.7675 2.005 2.227 2.442 2.652 2.859 3 065 3.269 3.674 300 A.
1260 0 1316.2 1368.3 1420 6 1472 8 1525.2 1578.1 1631.7 1741.0 (417.33) s.
l.5701 1.6268 1.6751 1.7184 1.7582 1.7954 1.8305 1.8638 1.9260 v.
l.4923 1.7036 1.8980 2 084 2.266 2.445 2.622 2.798 3.147 350 A.
1251.5 1310.9 1365.5 1418 5 1471.1 1523.8 1577.0 1630.7 1740.3 (431.72) s.
l.5451 1.6070 1.6563 1.7002 1.7403 1.7777 1.8130 1.8463 1.9086 l.2851 1.4770 16508 18161 1.9767 2.134 2.290 2.445 2.751 400 A.
1245.1 1306 9 1%2.7 1416 4 1469 4 1522.4 1575.8 1629 6 17395 (444 59) s.
I5281 1.5894 1.6398 1.6842 1.7247 1.7623 f.7977 15311 1.8916
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1.3 1.4 1.5 1.5 1.7 1.s Entropy
1 EQUATION SHEET Where og = m2
( densi ty) g ( vel oci ty)1( a rea ) g = ( den si ty) 2 ( vel oci ty)2 ( a rea )2 KE = mv2 PE + keg +PgVg = PE +KE +P V22 where V = specific PE = mgh g
2 2
~7 volume P = Pressure Q = ic (Tout-Tin)
Q = UA (T,y,-Tstm) 0 = m(hg.h )
p 2
P = P 10(SUR)(t) p. p,t/T SUR = 26.06 T = (B-pit o
o I
p delta K = (K,gg-1)/K,gg CRg(1-K,ggg) = CR II-Keff2)
CR = S/(1-K,ff) 2 M = (1-K,ggg)
SOM = (1_K,gg) x 100%
(1-Keff2I Keff 0.693 A = A e-(decay constant)x(t)
In (2) decay constant
=
=
g t
t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 3
1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /hr Density = 62.4 lbg/ft 1 MW = 3.41 x 106 3
Btu /hr Density = 1 gm/cm 1 Btu = 778 f t-lbf Heat of Vaporization = 970 Btu /lbn Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 f t-lbm/lbf-sec
MASTER 5.
THEOPY OF NUCLEAR POWER PLANI OPEFATION. FLUIDS, AND PAGE 55 ANSWERE -- MCGUIRE 1 !. 2
-85/05/21-TOM ROGEPS ANSWEP 5.01 (1.001 b
REFERENCE MNC OP-MC-SPS-RT-FP, p.10.
ANSWER 5.02 (1.004 d
REFERENCE MNS OP-MC-SPS-RT-NMF, p.48.
ANSWER 5.03 (1.00)
C REFERENCE hNS OP-MC-SPS-RT-SM, PP.23-24.
ANSWER 5.04 (1.00) d REFERENCE MNS OP-MC-SPS-FT-NMF, p.8.
A N S WE F:
5.05 (1.00) c REFERENCE nNS TE E: s s e s. 2/4.1.1.2.
1
5.
THEORY OF NUCLEAR POWER PLANT OPEPATION, FLUIDS, AND PAGE 56 ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS ANSWER 5.06 (1.50) 0 0.5 points each.
1.
?.
Limit + reactivitv on a rod ejection.
3.
Acceptable power distrubutilon limits.
REFERENCE MNS Core Performance, p.31.
ANSWER 5.07 (1.00) a REFERENCE MNS OP-MC-SPS-RT-NMF. p.35.
ANSWER 5.08 (1.00)
C REFERENCE MNS OP-MC-SPS-RT-NMF.
ANSuER 5.09 (1.00) t.
REFERENCE MNS OP-MC-SPS-RT-RP, p.24.
ANSWER 5.10 (1.00)
C REFERENCE MNS Core Performance, p.13.
t f
l t
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 57 i-THERMODYNAMICS i
ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROCERS 1
1 ANSWER 15. 1 1 (1.00) l C
REFERENCE MNS OP-NC-SPS-RT-RK, p.7.
i ANSWER 5.12 (1.00) b REFERENCE j
MNS OP-NC-SPS-RT-SM,p.11.
I 1
ANSWER 5.13 (1.00) 4 5
1 REFERENCE MNS OP-MC-SPS-RT-NMF, p.18.
i ANSWER 5.14 (1.00) b i
REFERENCE I
MNS OP-MC-SPS-FT-NMF, p.32.
1 ANSWER 5.15 (1.00) i c
ANSWER 5.16 (1.00)
C a
l 1
i l
i 1
4 i
I
i 4
So THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 58 TUERkbbEk bibs ~ ~
~
~~~~ ~
~
~~~~
7 ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS REFERENCE I
MNS OP-MC-SPS-RT-NMF, p.46.
ANSWER 5.17 (1.00) b O R d.
REFERENCE MNS OP-MC-SPS-RT-RCO, p.32.
ANSWER 5.18 (1.00) a i
I REFERENCE MNS OP-MC-SPS-RT-RCO, p.38.
ANSWER 5.19 (1 00) b REFERENCE MNS OP-NC-SPS-RT-RCO, p.28.
ANSWER 5.20 (1.00) d REFERENCE I
MNS OP-MC-SPS-RT-RC0r pp.18-19.
i
- i ANSWER 5.21 (1.00)
C i
REFERENCE i
MNS OP-NC-SPS-RT-RC0r p.15.
4 i
i
.t i
O
..e, m.,,.
--e
-..e
-y.m-
---,n.,.
r,,
,e.,,,-,-.,.,-,
,,,-.v.,aa
-,,.,w,-,
--nn,
-,,,.,a-g.-,
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 59 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 5.22 (1.00) a REFEPENCE MNS OF-NC-SPS-RT-RCO, p.5.
ANSWEP 5.23 (1.00) d REFERENCE MNS OP-NC-SPS-RT-RK, p.12.
ANSWER 5.24 (1.00) b REFERENCE MNS OP-MC-SPS-RT-RC0r F.37.
ANSWEP 5.25 (1.00) d REFEPENCE MNE Thermo-Core Performance, p.1.
AMSWEP G.26 (1.00)
C s
REFERENCE MNS Thernio-Core Performance, p.2.
l 1
1
.=.
i 3.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 60 THERMODYNAMICS I
ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS l
4 h
j ANSWER 5.27 (1.00) i d
l REFERENCE MNS Data Book, Sta. Dir 3 1.26, Sec.
1.3.
I ANSWER 5.28 (1.00) i b
4 REFERENCE MNS TS, B 3/4 4-15.
l ANSWER 5.29 (1.00)
I a
1 REFERENCE 2
MNS Core Performance, p.10.
ANSWER 5.30 (1.00) c REFERENCE MNS Core Performance, p.26.
ANSWER 5.31 (1.00) b l
REFERENCE i
MNS Thermo, para.
2.6.
4 1
1 1
e ii i
I a
i
't 1
. = _. -.,, _,,.... _,. -, _ _ - _.... _
4
~
i i
5.
THEORY OF NUCLEAR POWER PLANT OPERATIONe FLUIDS, AND PAGE 61 j
THERMODYNAMICS ANSWERS -- MCGUIRE 1&2
-85/05/21-TON ROCERS i
l ANSWER 5.32 (1.00) b l
REFERENCE t
HNS TS 3/4.2.3.
3 l
ANSWER 5.33 (1.00)
I a
l REFERENCE j
Steam Tables.
I i
ANSWER 5.34 (1.00) i i
b REFERENCE MNS Core Performance, p.20.
l l
l ANSWER 5.35 (1.00) b 4
l REFERENCE l
MNS Core Performance, pp.12-13.
}
ANSWER 5 36 (1.00) i d
i l
REFERENCE i
MNS Thermor Sec.
8.
4 i
i l
i i
+
l i
,,-,..--n,-
.--.,.,-.-,,.-,....,-..c....~...-
..,..--n.
i i
i l
]
5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND PAGE 62 THERs55isAniCs--------------------------------------
~~~~
4 i
ANSWERS -- MCGUIRE 1A2
-85/05/21-TOM ROGERS
.i 1
a l
ANSWER 5.37
(.50) l TRUE j
REFERENCE t
i MNS Thermo, Sec.
8.
1 ANSWER 5.38 (1.00) b i
i REFERENCE j
General Physics, HT&FF, p.328.
i ANSWER 5.39 (1.00) a j
Throttling process-constant enthalpy.
l h=1191.8 for P2R G 1000 psia hsat for 5 psis is 1131 REFERENCE Steam tables.
i j
ANSWER 5.40 (1.00) d FEFERENCE riNS OP-SS-HT-2, p.12.
1 4
i 4
l I
t E
1 1
i 1
I k
1 j
i
-4 1
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 63 ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS ANSWER 6.01 (1.00) b REFERENCE MNS OP-MC-SPS-SY-NC, p.24.
ANSWER 6.02 (1.00) a REFERENCE MNS OP-MC-SPS-SY-NC, p.27.
ANSWER 6.03 (1.00) l b
l REFERENCE MNS OP-MC-SPS-SY-NC, p.20 ANSWER 6.04 (1.00)
G 0.2 points each 1-f.
2-e.
3-e.
4-d.
5-c.
PEFERENCE NNS OP-MC-SPS-Si-NV, p.8.
ANSWEP 6.05 (1.00) s FEFERENCE MNS OP-nC-SPS-SY-NV, p.10.
1
4 l
l 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 64 ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS l
ANSWER 6.06 (1.00) et N
.l s
4 REFERENCE l
MNS OP-NC-SPS-SY-NV, p.19.
4 ANSWER 6.07 (2.00) 0 0.222 points each.
I 1.
Hi Containment Pressure, 2/3, 1.0 psis
(+,-0.1) j
- 2. Low Pressuri::er Pressure, 2/4, 1845 psis
(+,-
10.0 psis)
I
- 3. Low Steamline Pressure, g 585 psis
(+,- 5.0 psis)
REFERENCE Y3 dAMuwer.s ag t/V
$ 7E7Fweiaq*
i MNS OP-MC-SYF-SP-NI, p p.15.
j ANSWER 6.00 (1.00)
I 4
a s
REFERENCE MNS OP-NC-SYS-SP-NI.
ANSWER 6.09 (1.00)
C f
REFERENCE MNS OP-MC-SPS-SY-SSF. p.10.
ANSWER 6.10 (1.00) a 1
REFERENCE MNS OP-MC-SPS-SY-ND, p.16".
i 1
l i
1 i
t i
i
.____.._~_
1 l
i l
1 1
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 65 l
ANSWERS -- MCGUIRE 1A2
-85/05/21-TOM ROGERS
't ANSWER 6 11 (1.00) d REFERENCE MNS OP-MC-SPS-SY-ND, p.19.
i ANSWER 6.12 (1.00) i b
REFERENCE MNS PSM Rad. Mont., MC-IC-RM-3.
1 1
ANSWER 6 13
(.50) l TRUE.
REFERENCE j
MNS OP-MC-SPS-SY-NS, p.10.
i
]
ANSWER 6.14
(.50)
{
FALSE (0.25 psis).
REFERENCE MNS OP-MC-SPS-SY-NS, p.10.
[
i i
ANSWER 6.15 (1.00) l 9 0.25 points each.
j 1.
UST.
?.
CA Condensate Storage Tank.
3.
Condensate Hotwell.
)
4.
Nuclear Service Water.
REFERENCE MNS OP-MC-SPS-SY-CA, p.8.
f t
i 4
e
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 66 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 6.16 (1.00) 0 0.25 points each.
1.
Low low level in any SG.
2.
Trip of both main feedwater pumps.
3.
Ss signal.
4.
E:l a c k ou t signal.
REFERENCE MNS OP-MC-SPS-SY-CA, p.9.
ANSWER 6.17
(.50)
TRUE.
PEFERENCE MNS OP-NC-SPS-IC-IDE, p.6.
ANSWER 6.18 (1.00) a REFERENCE MNS OP-NC-SPS-IC-IDE, p.18.
ANSWER 6.10 (1.50)
G 0 5 points each.
1.
HI-HI Containment pressure.
2.
Low steamline pressure sP-11.
3.
High rate of pressure decrease :P- ' !.
ANSWER 6.20 (1.00) 2 FEFERENCE MNS OP-NC-SPS-SY-VI, p.8.
60 PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 67 ANSWERS -- MCGUIRE 1a2
-85/05/21-TOM ROGERS ANSWER 6.21 (1.00) d REFERENCE MNS OP-MC-SPS-SY-KF, p.9.
ANSWER 6.22 (1.00) e REFERENCE MNS OP-NC-SPS-SY-CF, p.19.
ANSWER 6.23 (1.00) a REFERENCE MNS PSM MC-IC-PPI-6.
ANSWER 6.24
(.50)
Vo prevent draining it.
REFERENCE MNS OP-MC-SPS-SY-PN, p.9.
ANSWER 6.25
(.50)
TRUE.
REFERENCE nNS OF-MC-SPS-EY-RN, p.13.
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 68 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 6.26 (1.00)
Any five G 0.2 points each.
1.
FWPT condensers.
2.
Condensate coolers.
- 3. KR coolers. (Recirculated Coolin3 Water) 4.
RL System. (Low Fressure Service Eater) 5.
RF System. (Fire Protection) 6.
RN System. (Nuclear Service Water?
REFERENCE MNS OP-MC-SPS-SY-RC, p.8.
ANSWER 6.27 (1.00) a REFERENCE MNS OP-MC-SPS-SY-KC, p.11.
ANSWER 6.28 (1.00) b REFERENCE MNS OP-MC-SPS-SY-CM, p.12.
ANSWER 6.29 (1.00)
C REFERENCE MNS OP-nC-SPS-SY-NF, p.23.
ANSWER 6.30 (1.00) c
~ _ _
i O.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 69 l
ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS I
i REFERENCE Power System Operation, R.H. Miller, pp.22-24.
I i
ANSWER 6.31 (1.00) 1 i
a 4
i REFERENCE I
MNS Standby Diesel Generator Lesson Plan, p.3.
i ANSWER 6.32 (1.00) 0 0.25 points each.
1.
2/3 UV on 4160 Essential Bus.
2.
Load shed actuated.
3.
DG >95% speed.
}
- 4. & 7c i: 1 :t.
% 1) (lecarouT RELMS) VCW m
4 j
FEFERENCE MNS SDG Lesson Plan, p.4.
- l ANSWER 6.33 (1.00) l a
l REFERENCE
]
MNS EDG Load Seq. Lesson Plan, p.14.
ANSWER 6.34 (1.00)
I C
f*
REFERENCE i
MNS 125VDC/120VAC Vital Instrument & Control Power System Lesson Plan, p.3.
1 i
i ANSWER 6.35 (1.00) l t
i l
l i
i I
i i
i
?
i l
1 l
1*
6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 70 i
i ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS i
1 i
REFERENCE j
MNS PSM, MC-IC-FLR-1.
I ANSWER 6.36 (1 00)
L b
i REFERENCE I
MNS PSM, MC-IC-RCS-2.
I ANSWER 6.37 (1.00)
I 1
e i
REFERENCE MNS PSM, MC-FC-ENI-4.
1 ANSWER 6.38 (1.00) j 9 0.25 points each.
- 1. PZR h13h level'
- 2. PZR low pressure.
J 3.
Low NC flow.
i l
4.
i
- 5. NCP underfrequency.
l REFERENCE MNS PSM, MC-IC-IPE-4.
i ANSWER 6.39 (1 00) i a
^
I REFERENCE i
MNS PSM, MC-IC-DEH-9.
i l
i i
i l
i t
ii i
I' I
I 1
6o PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION PAGE 71 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 6.40 (1.00) e 0.2 points each.
1.
SG 1evel.
- 2. Steam flow.
3.
Steam pressure.
4.
Feedwater flow.
5.
Nuclear power.
REFERENCE MNS PSM, MC-IC-SGL-1.
ANSWER 6.41 (1.00) b REFERENCE MNS PSM, MC-IC-DEH.
I l
)
l l
l l
~ - - -.....
f I
j 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 72 RADIOLOGICAL CONTROL i
l ANSWERS -- MCGUIRE 142 85/05/21-TOM ROGERS ANSWER 7.01 (1 00) 1 e
l b
i i
REFERENCE j
MNS OP/1/A/6150/02A.
p.
1.
i 1
ANSWER 7.02 (1.00) j a
i l
REFERENCE MNS OP/1/A/6100/01 p.
19.
i ANSWER 7.03 (1.00) l 9 0.5 points each
- s. TRUE b.
TRUE 4
REFERENCE MNS OP/0/A/6100/06, p.
2.
i 4
ANSuEP 7.04 (1.00) a
't l
REFERENCE nNS OP/1/A/6100/01, pp.14-15.
j ANSWER 7.05 (1.00) b or
}
}
REFERENCE j
MNS OP/1/A/6300/01, Encl.
4.2, p.4.
4 1
i j
I i
i i;
5 u--
r,c.,,~~-
,..,.-n,.r.,---,.,--,-.
-., _ --,--,,,,- n
--..nn.-y,.n-m,,--n.,
-m--,..
n s - w
+
,y---,..,-
,-~,-r,,.,~,---..
-,a,,n,m.-..mg-,,,n,,,nm,w~.
_-.- -._-._ - - -.. -..-__. ~
2 1
i 7o PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 73 f
RA515E55iCAE 55ATR5E~~~~~~~~~~~~~~~~~~~~~~~~
~~~~
ANSWERS -- MCGUIRE.182
-85/05/21-TOM ROGERS i
[
ANSWEP 7.06 (1.00) i i
e 0.2 points each.
1-b.
2-b.
)
3-e.
}
4-f.
[
5-c.
l REFERENCE
]
MNS OP/1/A/6100/01, pp.
1-2.
I l
ANSWER 7.07 (1 00) l l
i C.
j REFERENCE i
MNS OP/1/A/6250/01, Encl.
4.2, p.2.
i h
i ANSWER 7.08 (1.00) i d
l REFERENCE j
MNS OP/1/A/6100/01, p.7.
i l
ANSWER 7.09 (1 00) i C
j' REFERENCE j
MNS OP/0/A/6550/04, Encl.
4.8, pp.
1-3.
l l
i i
ANSWER 7.10 (1.00) i C
i i
1
(
1
}
f i
l
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 74
~~~~ dDf6E6dfCAC 55NTEUE~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- MCGUIRE la2
-85/05/21-TOM ROGERS REFERENCE MNS OP/1/A/6200/09, Encl.
4.4, p.1.
ANSWEP 7.11 (1.00) d REFERENCE MNS OP/0/A/6100/17, Encl.
4.1, p.5.
ANSWER 7 12 (1.00) d REFERENCE MNS AP/1/A/5500/38, p.2.
ANSWER 7.13 (1 00) e REFERENCE MNS AP/2/A/5500/14, Case I, p.2.
ANSWER 7.14 (1.00) b i
REFERENCE MNS AF/2/A/5500/12, p.10.
i ANSWER 7.15 (1.00) 0 0.5 points each 20;250 REFEFENCE MHS AF/2/A/5500/23, p.2.
i 5
i 1
l 70 PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 75 i
l
~~~~EI656(6656AE~66NTR6C"~~~~~~~~~~~~~~~~~~~~~~~
CNSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS j
i h
4 ANSWER 7.16 (1 00) j b
i REFERENCE l
MNS AP/2/A/5500/16, Case IV, p.9.
f i
f i
I ANSWER 7.17 (1.00) 5 t
i d
l REFERENCE MNS AP/2/A/5500/18, p.2.
l l
l ANSWER 7.18 (1.00) l b
)
i REFERENCE AP/2/A/5500/03, Case I, pp. 7-8.
i
~
4 I
ANSWER 7 19 (1 00) i 1
e 1
l REFERENCE l
MNS AF/2/A/5500/04, p.4.
i i
ANSWER 7.20 (1.25) i l,
G 0.25 points each.
}
i
- 1. Manually exercise reactor trip train A & E: switches.
l l
- 2. Verify a reactor trip has occured.
j 3.
Verify a turbine /sonerator trip has occured.
i 4
Ver if y ET A & ETE: are energi ed.
i
- 5. Verify SI is not required.
I REFEPENCE 7
MNE AP/2/A/5500/01, p.3.
4 I
i i
i l
]
I 4
i i
1
70 PROCEDURES - NORMAL, ABNORMALe EMERGENCY AND PAGE 76
~~~~~~~~~~~~~~~~~~~~~~~~
Rd6Ebl55fCAL C5NTE6L
~~~~
ANSWERS -- MCGUIRE 1A2
-85/05/21-TOM ROGERS ANSWEP 7.21
(.75)
@ 0.25 points each.
1.
Stop both DEH pumps.
j 2.
Place the turbine in neanval and close the governor valves in FAST a c t i o ri.
- 3. Locally trip the turbine.
REFERENCE nNS AP/2/A/5500/01, p.3.
ANSWER 7.22 (1 00) 0 0.25 points each.
- 1. Manually insert rods.
T. Locally open the trip breakers.
- 3. Locally open the MG sets output breater.
4.
Locally opere the MG sets motor breaker.
REFERENCE MNS EP/2/A/5000/11.1, p.2.
i ANSWER 7.23 (1.25) 0 0.25 peints each.
1.
hanvally e: teres se r eac tor trip train A e B switches.
- 2. Verify a reactor tr 1p has occured.
- 3. Verify a turbine / generator trip has occured.
4.
Verify ETA R ETB are energi:ed.
- 5. Vertfy the load sequencer has actuated.
REFERENCE MNS EF/2/A/5000/01, pp.2-3.
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 77
~~~~Ed656E6GiCIE~C6NTR6E---~~-------------------
ANSWERS -- MCCUIRE 1&2
-85/05/21-TOM ROGERS ANSWER 7.24 (1.25) 0 0.25 points each.
1.
Manually e::ercise reactor trip train A&D switches.
- 2. Verify a reactor trip has occured.
- 3. Verify a turbine / generator trip has occured.
4.
Try to energt:e the 4160V bosses with the DGs.
- 3. Dispatch an operator to initiate NC pump seal i n j e c t 1 ore from the SSF.
REFERENCE MNS EP/2/A/5000/09, pp.2-3.
ANSWER 7.25
(.50)
TRUE REFEFENCE MNS EF/2/A/5000/1.2.
p.1.
ANSWER 7.26 (1 00) d FEFERENCE NNS EF /2/A/5000/01 Foldovt paje.
ANSWER 7.27 (1 00) b
. s REFERENCE MNS EF /2/ A/5000/01 p.10.
AMSWER 7.28 (1 00) s k
i
1 i
7.
PROCEDURES - NORMAL, ABNORMAL, ENERGENCY AND PAGE 78
~~~~E3656E66fCdE~66NTR6E~~~~~~~~~~~~~~~~~~~~~~~~
ANSWERS -- NCCUIRE 1&2
-85/05/21-TON ROGERS l
FEFERENCE i
MNS EP/2/A/5000/01, p.
Encl. 1 Foldout.
ANSWER 7.29 (1.00) b REFERENCE MNS EP/2/A/5000/01, Foldout paJe.
ANSWER 7.30 (1.00) b REFERENCE NNS EP/2/A/5000/1 1, p.5.
ANSWEP 7 31 (1 00) a FEFERENCE NNS EP/2/A/5000/10, p.2.
ANSWEP 7.32 (1 00) a REFERENCE NNS EF/2/A/5000/10, p.2.
ANSWEF 7.33 (1 00) d REFERENCE NNS EP/2/A/5000/16.3
_m
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 79
~~~~ A6i6E65fEAE 56 sis 5E------------------------
R ANSWERS -- MCGUIRE 1&2
-85/05/21-TON ROGERS ANSWER 7.34 (1.00) d l
REFERENCE MNS Orientation Manual, p.27.
ANSWEP 7.35 (1.00)
C REFERENCE MNS HP Manual, p.2.1-3.
ANSWER 7.36 (1.00) c REFERENCE 10 CFR 20.101 ANSWER 7.37 (1.00) d REFERENCE MNS HP Manual, 3.1.2.
ANSWER 7.38 (1.00)
E REFERENCE MNS HF manual, 2.3.4 i
I
7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 80
~~~~ d6E6E66 6 E'66NTR6E~~~~~~~~~~~~~~~~~~~~~~~~
R ANSWERS -- MCGUIRE.182
-85/05/21-TOM ROGERS ANSWER 7.39 (1.00) l l
b l
l REFERENCE I
ANSWER 7.40 (1.00) a REFERENCE MNS HP Manual, 2.5.5.
1
80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 31 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 8.01 (1.00) c REFERENCE MNS TS, 1.8.
ANSWER 8.02 (1.00) 9 0.2 points each 1-a.
2-d.
3-c.
4-b.
5-e.
REFERENCE MNS TS, 3.4.6.2.
l l
ANSWER S.03 (1.00) l l
l ANSWER 8.04 (1.00) l 6
l l
ANSWER 8.05 (1.00) d REFERENCE hNS TS, 3.5.1.2.
,r
}
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 82 ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS ANSWER 8.06 (1.00) 4k 8
REFERENCE MNS TS, 3.5.1.1.
ANSWER 8.07 (1.00) a REFERENCE MNS TS, 3.2.1.
ANSWER 8.08 (1.00) c REFERENCE MNS TS, 3.4.4.
ANSWER 8.09 (1.00) b REFERENCE MNS TS, 3 4.9.1.
ANSWER B.10
(.50)
FALSE.
REFERENCE MNS TS, 3/4.1.3.
ANSWER 8.11
(.50)
TRUE.
/
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 83 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS REFERENCE MNS TS, 3.1.3.3.
ANSWER 8.12 (1.00) d REFERENCE MNS TS, 3.1.1.4.
ANSWER 8.13 (1.00) 0 0.5 points each.
1.3-Unit 1.
-fg* Unit 2.
REFERENCE MNS TS, 3.1.1.1.
ANSWER 8.14 (1.00)
C REFERENCE MNS TS, Table 3.3-1.
ANSWER 8.15 (1.00) b REFERENCE MNS TS, E:ases 3/4.3.4.
s ANSWEP 8.16 (1.00)
C REFERENCE NNS TS, 3.7.10.1.
,-r-
= _.
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 84 i
ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS
.i ANSWER 8.17 (1.00) a REFERENCE MNS TS, 3.1.2.5.
I ANSWER 8.18 (1.00)
& 0.5 Points each.
a.
TRUE.
b.
FALSE.
ANSWER 8.19 (1.00)
B REFERENCE MNS TS, 3.8.2.1.
, ANSWER 8.20 (1.00) d REFERENCE MNS TS, 3.8.2.1.
ANSWER 8.21 (1.00) b REFERENCE MNS TS, 4.0.2 8 4.0.5.
9 1
l m
Bo ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 85 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS ANSWER 8.22 (1.00) 0 0.2 points each.
c.
1.
b.
1.
c.
3.
d.
3.
e.
1.
REFERENCE MNS TS, Table 6.2-1.
ANSWER 8.23 (1.00) c REFERENCE MNS OMP 1-2, p.2.
ANSWER 8.24 (1.00)
REFERENCE MNS OMP 1-2, p.
15.
ANSWER 8.25 (1.00) d REFERENCE MNS OMP 1-2, p.7.
l ANSWER 8.26 (1.00) a REFERENCE MNS OhP 1-2, p.12.
4 s
4
,_.y
80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 86
_________________________________________o________________
ANSWERS -- MCGUIRE 1&2
-85/05/21-TOM ROGERS ANSWER 8.27
(.50)
FALSE (Must check upper right corner of-the Procedure Prep. Process Record.)
REFERENCE MNS OMP p.13.
ANSWER 8.28 (1.00) b REFERENCE MNS OP/0/A/6100/06, p.2.
ANSWER 8.29 (1.00) a REFERENCE MNS OMP 1-5, p.5.
ANSWER 8.30
(.50)
TRUE (Only required to assure power supplies are energized.)
REFERENCE OMP 1-6, p.6.
ANSWER 8.31 (1.00)
C i
REFERENCE OMP 2-4, p.3.
a
8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 87 ANSWERS -- MCGUIRE 182 85/05/21-TOM ROGERS ANSWER 8.32 (1.00)
G REFERENCE j
MNS Sta. Dir.
3.1.8, p.1.
ANSWER 8.33 (1.00) b REFERENCE MNS Sta. Dir. 3.1.10, p.5.
ANSWER 8.34 (1.00) a REFERENCE MNS Directive 3.8.1.
ANSWER 8.35 (1.00) d REFERENCE MNS RP/0/A/5700/04, Encl.
4.5, p.3.
ANSWER 8.36 (1.00) 8 0.5 point each.
1.
Cowan's Ford Dam.
?.
Training & Technology Center.
REFERENCE MNS Directive 3.8.1.
._,yy,-,
w.
y,
,-r
, +,,, -
80 ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 88 ANSWERS -- MCGUIRE 182
-85/05/21-TOM ROGERS l
ANSWER 8.37 (1.00) b REFERENCE MNS HP Manual, 18.4, p.2.
ANSWER 8.38 (1.00) d REFERENCE MNS HP/0/B/100?/16, p.1.
l ANSWER 8.39 (1.00) i I
b j
REFERENCE j
MNS OMP 2-1, p.2.
ANSWER 8.40 (1.00) b REFERENCE 10 CFR 50.72 ANSWER 8.41 (1.00) b REFERENCE HNS Sta. Dir. 3.1.19r p.1.
ANSWER 8.42 (1.00) b
}
n-
.,wn
,e-
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- - -. - - - = -, - - - -
L e
o 4
4 1
8.
ADMINISTRATIVE' PROCEDURES, CONDITIONS, AND LIMITATIONS PAGE 89 ANSWERS -- MCGUIRE 182
-85/05/21-TON ROGERS REFERENCE MNS Sta. Dir.
4.4.2, p.3.
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i l
i
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i i
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r U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION Facility:
McGUIRE Reactor Type:
Westinghouse PWR Date Administered:
MAY 21, 1985 Examiner:
Schreiber/ Gruel Candidate:
Mas +y c u re c.oy INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheet. Points for each question are indicated in parenthesis after the question. The passing grade requi.res at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
Category
% of Candidate's
% of Value Total Score Cat. Value Category 30 25
- 1. Principles of Nuclear Power Plant Operation, Thermo-dynamics, Heat Transfer and Fluid Flow 30 25
- 2. Plant Design Including Safety and Emergency Systems 30 25
- 3. Instruments and Controls 30 25
- 4. Procedures: Normal, Abnormal, Emergency, and Radiological Control 120 TOTALS Final Grade All work done on this examination is my own; I have neither given nor received aid.
Candidate's Signature
1 1.0 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, (30.0)
HEAT TRANSFER AND FLUID FLOW Points Available QUESTION 1.01, Consider a subcritical reactor with a source strength of 20 cps.
At one point during a rod withdrawal, the count rate levels off at 100 cps. After further withdrawal, the count rate levels off at 200 cps.
?
(1.0) a.
What is the final value of Keff (a) 0.7 (b) 0.8 (c) 0.9 (d) 1.0 b.
How much reactivity was added to raise the count rate from 100 cps to 200 cps?
(1.0)
(a) 0.100 aK/K (b) 0.111 AK/K (c) 0.125 aK/K (d) 0.139 AK/K ANSWER 1.01 a.
(c) fe4'(d) b.
Reference (s)
McGuire Reactor Theory, Subcritical Multiplication, p.11.
- Section 1 continued on next page -
2 Points Available QUESTION 1.02 In a reactor with a source, a non-changing neutron flux is indicative of criticality. TRUE or FALSE.
(0.5)
ANSWER 1.02 False Reference (s)
McGuire Reactor Theory, Subcritical Multiplication, p. 9.
QUESTION 1.03 To increase power level from 5 x 10-10 amps to 10-8 amps in two (2) minutes, the start-up rate (SUR) should be maintained at decades per minute (dpm).
(1.0) a.
0.65 b.
0.75 c.
1.25 d.
1.5 ANSWER 1.03 a
Reference (s)
McGuire Reactor Theory, Reactor Xinetics, p. 7.
- Section 1 continued on next page -
3 Points Available OUESTION 1.04
?
(1.0)
Which of the following does NOT affect Keff a.
Moderator pressure b.
Fuel temperature c.
Source strength d.
Core age ANSWER 1.04 c
Reference (s)
McGuire Reactor Theory, Neutron Multiplication Factor, p. 42.
QUESTION 1.05 Which of the following is NOT needed to determine estimated critical boron concentration?
(1.0) a.
Rod height b.
Xenon worth c.
Samarium worth d.
Shutdown boron concentration 4
ANSWER 1.05 d
Reference (s)
McGuire Reactor Theory, Reactivity Balance, p. 7.
- Section 1 continued on next page -
l
4 Points Available 00ESTION 1.06 A reactivity of 0.015% AK/K is equivalent to which of the following?
(1.0) a.
1.5 pcm b.
15 pcm c.
150 pcm d.
1500 pcm ANSWER 1.06 b
Reference ( s)
McGuire Reactor Theory, Neutron Multiplication Factor, p. 48.
QUESTION 1.07 Which of the following methods of energy release from a fission is by far the largest?
(1.0) a.
Kinetic energy of fission fragments b.
Kinetic energy of fission neutrons c.
Beta emission from fission products d.
Gama emission from fission products ANSWER 1.07 a
Reference (s)
McGuire Reactor Theory, Fission Process, p.14.
l
- Section 1 continued on next page -
5 8
D Points Available OUESTION 1.08 Delayed neutrons are born on the average of seconds after the fission.
(1.0) a.
10-14 b.
10-10 c.
10-4 d.
10 ANSWER 1.08 d
Reference (s)
McGuire Reactor Theory, Fission Process, p.15.
OUESTION 1.09 Which of the following statements regarding the use of burnable poisons is CORRECT?
(1.0) a.
Allows smaller core geometry for same generation capacity b.
Reduces the number of control rods required c.
Reduces the possibility of a positive MTC d.
Used to shape axial power distribution ANSWER 1.09 1
c Reference (s)
McGuire Reactor Theory, Reactor Poisons, p.12.
- Section 1 continued on next page -
1
6 Points Available 00ESTION 1.10 With regard to axial flux, which of the following can only shift neutron flux toward the top of the core?
(1.0) a.
Burnup b.
Control rods c.
Soluble baron d.
Xenon ANSWER 1.10 a
Reference ( s)
McGuire Core Performance, p.17-19.
QUESTION 1.11 If a reactor is at power and all full-length control rods are operable, minimum shut-down margin exists provided rod insertion limits are not violated. TRUE or FALSE.
(0.5)
ANSWER 1.11 True Reference ( s)
McGuire Reactor Theory, Reactivity Balance, p. 24.
i
- Section 1 continued on next page -
j
~
i
7 Points Available OUESTION 1.12 The reactor has been at full-power operation for one month before it is shut down. What change in negative reactivity has occurred for each of the following poisons from 100% power equilibrium con-centration to shut-down equilibrium concentration?
a.
Xenon (1.0)
(a) 6500 pcm increase (more negative)
(b) 3700 pcm increase (more negative)
(c) 900 pcm decrease (less negative)
(d) 2800 pcm decrease (less negative) b.
Samarium (1.0)
(a) 375 pcm increase (more negative)
(b) No change (c) 75 pcm decrease (less negative)
( d) 225 pcm decrease (less negative)
ANSWER 1.12 a.
(d) b.
(a)
Reference (s)
McGuire Reactor Theory, Reactor Poisons, pp. 22,29.
l
- Section 1 continued on next page -
ll J
8 Points Available QUESTION 1.13 Which of the following statements regarding rod worth is CORRECT?
(1.0) a.
Greater at 150*F than at 500*F b.
Greater at 50% power than at 100% power c.
Greater for a rod surrounded by inserted rods than for a rod surrounded by withdrawn rods d.
Greater at the middle of the core than at the top of the core ANSWER 1.13 d
i Reference ( s)
McGuire Reactor Theory, Reactivity Coefficients and Defects, pp. 35-41.
S
- Section 1 continued on next page -
i
_ _ _ ~
.__,w4-
-, ~ _ _ - - -
m
9 Points Available 00ESTION 1.14 Which of the following is a definition of quadrant power tilt ratio (QPTR)?
(1.0) a.
Minimum upper detector output divided by average upper detector output b.
Maximum upper detector output divided by average upper detector output c.
Minimum upper detector output divided by average lower detector output d.
Maximum upper detector output divided by average lower detector output ANSWER 1.14 b
Reference (s)
McGuire Core Performance, p. 20.
- Section 1 continued on next page -
10 Points Available OUESTION 1.15 For each of the following, state whether there is an increase or decrease in Critical Heat Flux (CHF). Consider each item separately, a.
Increased flow.
(0.5) b.
Increased temperature.
(0.5) c.
Increased pressure.
(0.5)
ANSWER 1.15 a.
Increase b.
Decrease c.
Increase Reference (s)
McGuire Core Performance, pp.11-12.
QUESTION 1.16 The heat transfer mechanism from the fuel to the coolant becomes film boiling (departs from nucleate boiling--DNB) at what DNB ratio?
(1.0) a.
0.77 b.
1.0 c.
1.3 d.
1.7 ANSWER 1.16 b
Reference (s)
McGuire Core Performance, pp.12-13.
- Section 1 continued on next page -
l
,,n m -..
11 Points Available QUESTION 1.17 The reactor is at 80% power, rods in manual, when an inadvertent boration occurs which increases RCS (NC) boron concentration by 10 ppm. Assume a baron coefficient (differential boron worth) of
-10 pcm/ ppm and a differential rod worth of 10 pcm/ step.
If power changes by 5% (assume nct rod motion), what is the a.
approximate value of the power coefficient (pcm/% power)?
(1.0)
(a)
-100 (b)
-20 (c)
-10 (d)
-5 b.
How must rods be moved to restore the initial power level?
(1.0)
(a)
Inserted 20 steps (b)
Inserted 10 steps (c) Withdrawn 20 steps (d) Withdrawn 10 steps ANSWER 1.17 a.
(b) b.
(d)
Reference (s)
McGuire Reactor Theory, Reactivity Coefficients and Defects, pp. 16-17.
- Section 1 continued on next page -
12 Points Available QUESTION 1.18 The reactor is operating at 65% power when a S/G SM PORV fails open.
a.
If rod control was in automatic during this transient, which of the following describes the behavior of controlling T-ave?
(1.0)
(a)
Increases and remains there.
(b)
Increases, then returns to original value.
(c) Decreases, then returns to original value.
( d) Decreases and remains there.
b.
If rod control was in manual during this transient, which of the following describes the behavior of controlling T-ave?
(1.0)
(a)
Increases and remains there.
(b)
Increases, then returns to original value.
(c) Decreases, then returns to original value.
(d) Decreases and remains there.
ANSWER 1.18 a.
(c) b.
(d)
Reference (s)
McGuire Reactor Theory, Reactivity Coefficients and Defects, pp. 30-31.
- Section 1 continued on next page -
1
13 Points Available OUESTION 1.19 Overall plant thermal efficiency will decrease if condenser cool-ing water inlet cooling temperature increases. TRUE or FALSE.
(0.5)
ANSWER 1.19 True Reference ( s)
McGuire Thennodynamics, Section 8.
OUESTION 1.20 Which of the following describes the exit condition of a leak in a RCS (NC) hot leg during at-power operation?
(1.0) a.
Superhetted steam b.
Saturated steam c.
Wet steam d.
Water ANSWER 1.20 c
Reference (s)
McGuire Thermodynamics, Section 5.
t
- Section 1 continued on next page -
i i
1
14 c
Points Available OUESTION 1.21 The work produced by a pump is proportional to which of the following?
(1.0) i a.
AP b.
d.
(aP)-1 ANSWER 1.21 b
Reference (s)
McGuire Themodynamics, Section 9.
OUESTION 1.22 Condenser vacuum of 20 in. Hg is equivalent to what absolute pressure (psia)?
(1.0) a.
1.1 b.
4.9 c.
9.8 d.
13.6 4
ANSWER 1.22 b
Reference (s)
McGuire Themodynamics, Section 3.2.
{
- Section 1 continued on next page -
15 i
Points Available j
OUESTION 1.23 The turbine on the auxiliary feedwater (#2 TD CA) pump is an example of a constant process.
(1.0) i a.
Pressure b.
Enthalpy c.
Entropy j
d.
Temperature ANSWER 1.23 i
C Reference (s)
J McGuire Thennodynamics, Sections 5 and 8.
i 1
l 1
{
- Section 1 continued on next page -
16 l
Points Available OUESTION 1.24 Answer the following statements regarding centrifugal pump opera-tion TRUE or FALSE.
a.
Two identical pumps in parallel, each at 50% flow capacity, consume less power than one of the pumps at 100% flow capacity.
(0.5) b.
Two identical pumps in parallel can produce a greater dis-charge pressure at a non-zero flow rate than can one of the pumps at the same flow rate.
(0.5) c.
Two identical pumps in series can produce twice the flow at pump runout conditions than can one of the pumps.
(0.5) d.
Decreased inlet temperature increases available net positive suction head (NPSH).
(0.5)
ANSWER 1.24 a.
True b.
True c.
Fal se d.
True Reference ( s)
Westinghouse Thermodynamics, Chapter 10, pp. 32-60.
- Section 1 continued on next page -
1
17 Points Available OUEST10N 1.25 The reactor is at 80% power at a core AT of 48"F and a mass flow rate of 100% when a station blackout occurs. Natural cir-culation is established and core aT goes to 40*F.
If decay heat is approximately 2% of full power, what is the mass flow rate
(% of full flow)?
(1,0) a.
1.9%
b.
2.1%
c.
2.4%
d.
3.0%
ANSWER 1.25 l
d Reference (s)
McGuire Thermodynamics, Section 8.
l
- SectfMe 3 ontinued on next page -
18 Points Available OUESTION 1.26 Which of the following is NOT an indication of a natural circulation cooldown?
(1.0) a.
NC system subcooled b.
S/G SM pressure stable or decreasing c.
NC loop hot-leg temperature stable or decreasing l
d.
NC loop cold-leg temperature above saturation temperature l
for S/G SM pressure ANSWER 1.26 d
Reference (s)
McGuire Natural Circulation Cooldown Procedure, foldout page.
- End of Section 1 -
l 1
O
19 2.0 PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEM (30.0)
Points Available OUESTION 2.01 Select the cooling water system that meets the following criteria:
l Serves both the Auxiliary and Turbine buildings, and does not cool any system containing primary coolant.
(1.0) a.
Recirculated cooling water system b.
Component cooling water system c.
Condenser circulating water system d.
Nuclear service water system ANSWER 2.01 a
Reference (s)
McGuire System Descriptions, Recirculated Cooling Water.
1 l
l i
- Section 2 continued on next page -
20 Points Available j
00ESTION 2.02 The effectiveness of the Halon fire protection system depends on the following:
(1.0) a.
It is used in conjunction with the Mulsifyre system.
b.
Each discharge head is independently activated by heat.
c.
It is distributed where needed by the Ventilation system.
d.
Local thermal detectors generate alarms as well as release the gas.
ANSWER 2.02 d
Reference ( s)
McGuire System Descriptions, Fire Protection system, pp. 6 and 7.
OUESTION 2.03 The Air Handling system for the ESF Pump Room is classified as Nuclear Safety Related. TRUE or FALSE.
(0.5)
ANSWER 2.03 True Reference (s)
McGuire System Descriptions, Auxiliary Building Ventilation System, p. 8.
- Section 2 continued on next page -
Points
21 Points Available OUESTION 2.04 Station Air is independent of Instrument Air. TRUE or FALSE.
(0.5)
ANSWER 2.04 False ns
" t r<tiVic.ad-in Reference (s)
McGuire System Descriptions, Station Air, p. 5.
QUESTION 2.05 The Containment Sample Blower subsystem has which of the follow-ing as its purpose?
(1.0) a.
To sample the hydrogen buildup following a LOCA b.
To routinely sample for gases in the containment c.
To periodically check containment for radioactivity d.
To automatically start the containment purge system ANSWER 2.05 a
Reference ( s)
McGuire System Descriptions, Containment Sample and Purge,
- p. 3.
- Section 2 continued on next page -
22 Points Available OUESTION 2.06 If the Nomal Letdown path is lost during operation in Mode 2 at BOL, and it is necessary to depend on the Excess Letdown path-way for at least several days at low power, which of the follow-ing is of concern?
(1.0) a.
Ion exchanger availability for coolant cleanup.
b.
Boron Themal Regeneration System availability, c.
Crud burst from core can affect reactor coolant pump seals.
d.
Hydrogen concentration in coolant uncontrolled.
ANSWER 2.06 a
Reference (s)
McGuire System Descriptions, Chemical and Volume Control,
QUESTION 2.07 When the Reactor Makeup System is in automatic, how does it know what strength Boric Acid to provide?
(1.0) a.
The automatic controller uses infomation from the Boronometer.
b.
The latest analysis from Chemistry is read into the controller.
c.
The controller uses the most recent settings on the Totalizers.
d.
The controller depends on VCT exit flow rate as well as level.
ANSWER 2.07 X
b
- w c= cd c4
-7 Reference (s)
McGuire System Descriptions, Chemical and Volume Control, E. 20 & 2h ff 6. t i (> > /H C S b ~ otT7. a), / 2Np g*
- Section 2 continued on next page -
I
23 1
Points Available QUESTION 2.08 Mark each item TRUE or FALSE. The Condenser Circulating Water System is the normal source of water to each of the following systems:
a.
Nuclear Service Water System.
(0.5) b.
Fire Protection Jockey Pumps.
(0.5) c.
Feedwater Pump Turbine Condensers.
(0.5) d.
Conventional Low Pressure Service Water.
(0.5) e.
Steam Air Ejector System Condenser.
(0.5)
ANSWER 2.08 a.
False b.
True c.
True d.
True e.
Fal se Reference (s)
McGuire System Descriptions, Main Condenser Circulating Water, various pages.
i
- Section 2 continued on next page -
l
24 Points Available 00ESTION 2.09 During a Blackout of Unit 1 (no S Of S ), the "A" pump of the s
p Nuclear Service Water (RN) System supplies the Containment Ven-tilation units with cooling water because:
(1.0) a.
The low level intake gives more NPSH.
b.
The RY pump motors are without power.
c.
The RB nonessential header isolates.
d.
The RN for Unit 2 would be affected.
ANSWER 2.09 a
Reference (s)
McGuire System Descriptions, Nuclear Service Water, p.15.
QUESTION 2.10 The five safety valves on each Steam Generator are set at different relief pressures because:
(1.0) a.
Control of steam flow is smoother.
b.
This prevents chattering of valves.
c.
Steam pressure oscillation is maintained.
d.
Steam release in Doghouse is minimized.
ANSWER 2.10 b
e:o :_ _d d Reference (s)
McGuire System Descriptions, Main Steam, p. 9.
- Section 2 continued on next page -
25 Points Available OUESTION 2.11 Which of the following statements is CORRECT about the Steam Dumps?
(1.0) a.
40% of full load can be dumped to the atmosphere.
b.
The system can handle a 5% demand change or 10%/ min.
c.
The turbine bypass to condenser can handle 50% load.
d.
A 30% load dump allows Reactor Follow operation.
ANSWER 2.11 Gd 7
Reference (s)
McGuire System Descriptions, Main Steam, p.12.
k~ s tem L<. g.
- 08., sL, m 00ESTION 2.12 The Standby Shutdown Facility conta' ins systems designed to handle the following major event:
(1.0) a.
Rupture of the dams impounding Lake Norman and the SNSW pond.
b.
A' major fire disabling the equipment in one room of the plant.
c.
The Technical Support Cent'er is unavailable, d.
A LOCA that requires evacuation of the plant site for 30 days.
ANSWER 2.12 b
Reference (s)
McGuire System Descriptions, Standby Shutdown Facility, p. 7.
Section 2 continued on next page -
26 Points Available OUESTION 2.13 Which of the following is NOT a part of the Onsite Power System?
(1.0) a.
The 230 KV lines and switchyard b.
The Main Generators at each unit c.
The Auxiliary Transformers at each unit d.
The Batteries / Distribution lines ANSWER 2.13 a
Reference ( s)
McGuire System Lescriptions, Main Power Distribution, p.14.
QUESTION 2.14 Before racking out the 6.9 KV breakers, you should ensure which of the following?
(1.0) a.
The lockout relays are reset.
b.
HP and Chemistry are notified.
c.
Exciter air cooler temp. <112*F.
d.
DC control power fuses pulled.
ANSWER 2.14 d
Reference (s)
McGuire System Descriptions, Main Power Distribution, p. 34.
- Section 2 continued on next page -
27 Points Available 4
OUESTION 2.15 Which of the following is an important requirement of the Emergency Diesel Generator?
(1.0) a.
Must block manual control during automatic start b.
Must accept loads at 95% voltage and 75% frequency c.
Must have AC power available at all times d.
Must reach rated speed <1l seconds from ESF signal ANSWER 2.15 e,~of. v (b 8
RE S -:n ve-it e,46 7
Reference (s)
McGuire System Descriptions, Diesel Generator, p. 9.
QUESTION 2.16 The outside air intakes for the Control Room Ventilation system will automatically close for several conditions outside the building. Select the outdoor condition which will NOT result in vent closure.
(1.0) a.
Smoke b.
Chlorine c.
Radiation d.
Tornado ANSWER 2.16 a
Reference (s)
McGuire System Descriptions, Control Room Ventilation, p.10.
Section 2 continued on next page -
28 F
t i
Points Available OUESTION 2.17 The Strong Motion Accelerograph will be triggered into activity when it receives a minimum shock as low as which of the following?
(1.0) a.
0.0015 g b.
0.015 g c.
0.15 g 4
d.
1.5 g ANSWER 2.17 b
Reference (s)
McGuire System Descriptions, Seismic Monitoring, p. 6.
OUESTION 2.18 The Vibration and Loose Parts Monitoring system is to detect which of the following conditions?
(1.0 )
a.
Fuel rod vibration caused by baffle jetting b.
Unbalanced flows among the primary loops c.
Imminent seal failure in the coolant pumps d.
Hardware left in the S/Gs after repairs ANSWER 2.18 d
Reference (s)
McGuire System Descriptions, Vibration and Loose Parts Monitoring,
- p. 5.
- Section 2 continued on next page -
29 Points Available OVESTION 2.19 In the Liquid Waste treatment system, the Waste Monitor Tank's normal discharge is to which of the following?
(1.0) a.
Condenser Circulating Water b.
Initial / Final Holdup Ponds c.
Waste Evaporator Preheater d.
Reactor Makeup Water Tank ANSWER 2.19 a
Reference (s)
McGuire System Descriptions, Liquid Waste Recycle, Data Sheet MC-WS-WL-12.
QUESTION 2.20 In the Reactor Vessel, the Energy Absorber Assembly serves the following purpose:
(1.0) a.
Minimizes core damage in the event of a LOCA.
b.
Catches the lower internals if the core drops.
c.
Dampens the motion of control rods as they drop.
d.
Cushions the effect of an ejected control rod.
ANSWER 2.20 t
b Reference (s) l McGuire PSM, Reactor Vessel, Figure MC-CMP-RVI-1.
- Section 2 continued on next page -
l
30 Points Available QUESTION 2.21 The function of the Flywheel on a Reactor Coolant Pump is as follows:
(1.0) a.
Provides coast-down flow upon loss of power to the pumps.
b.
Resists counter rotation during sequential startup of pumps.
Prevents overspeed of motor if the impeller siezes/ shears the shaft, c.
d.
Provides continued rotation in the event the motor bearings sieze.
ANSWER 2.21 a
Reference ( s)
McGuire PSM, Reactor Coolant Pumps, Figure MC-CMP-MCP-1.
QUESTION 2.22 An excessive leak from the top of a Control Rod Travel Housing will have the following adverse effect on safety:
(1.0) a.
Causes the drive rod to separate from the control rod cluster.
b.
Shorts out the electromagnet assemblies, making the rod drop.
c.
Holds the rod up, preventing it from scramming on trip signal.
d.
Causes the RCC to vibrate, damaging the thimble guide tubes.
ANSWER 2.22 c
Reference (s)
McGuire PSM, Control Rods, Figure MC-CMP-CR-1.
1
- Section 2 continued on next page -
2
31 Points Available 00ESTION 2.23 Which of the following statements is CORRECT regarding Auto Power Transfer on the 6900 Volt Bus?
(1.0) a.
Transfer can be in either direction, STBY/NOR or NOR/STBY.
b.
If the two power sources are out of sync, load shed occurs.
c.
Fast transfer occurs in one second to prevent load shedding.
d.
Kirk-Key interlocks prevent the 86 relay from operating.
ANSWER 2.23 b
Reference (s)
McGuire PSM, Main Power Distribution, p. 6.
QUESTION 2.24 When placing a Battery Charger or Inverter in service, close the DC breaker first. Why?
(1.0) a.
This puts the selector switch in " float" mode.
b.
This causes the DC bus ties to remain closed.
c.
This prevents current surges on the syster..
d.
This equalizes the charge on the batteries.
ANSWER 2.24 j
C Reference (s)
McGuire PSM, Vital Instrumentation and Control Power, p. 3.
- Section 2 continued on next page -
32 Points Available OUESTION 2.25 During a test run of a Diesel Generator, you hear a loud knock toward one end of the machine but no other effect is visible.
Crankcase vacuum drops off and engine speed decreases.
What has happened?
(1.0) a.
Fuel oil has ceased to flow because a valve has closed inadvertently.
b.
A piston has failed, exposing the crankcase to combustion gasses.
c.
The supercharger gas turbine has failed, causing engine backfire.
d.
Jacket cooling water is spilling into the crankcase, making steam.
ANSWER 2.25 b
- . 1
- d.
- aa,,... 4 s.
Reference (s)
McGuire System Descriptions, Diesel Generators, p. 22.
l l
- Section 2 continued on next page -
33 Points Available OUESTION 2.26 Design features of the Pressurizer (PZR) allow it to be cooled much faster than the Reactor Vessel (RV). The difference between the P7R and RV that allows this is:
(1.0) a.
The PZR wall is thicker than the RV wall.
b.
The RV gets hotter than the PZR.
c.
The RV is embrittled by radiation.
d.
The PZR is limited by flaw size.
i f
ANSWER 2.26 i
l c
Reference ( s)
{
McGuire System Descriptions, Pressurized Themal Shock, p. 24.
i i
l
- Section 2 continued on next page -
34 Points Available QUESTION 2.27 In the list below are postulated LOCA conditions that correspond to the Emergency Core Cooling Criteria.
In each case, state whether the condition exceeds or does not exceed the ECC criteria.
The set of postulated conditions are not necessarily consistent.
a.
The peak cladding tenperature prior to quench is estimated to be between 2300*F and 2700*F.
(0.5) l b.
The maximum cladding oxidation, prior to reflood, is estimated to be between 10% and 12% of the wall thickness.
(0.5) c.
Hydrogen in containment is measured and computed to be equiv-alent to 3.5%-of the zirconium present in the core.
(0.5) d.
The in-ccre flux monitor can be inserted into all but 10 of the available positions, none of which are next to each other.
(0.5) e.
Core exit thermocouples read between 140*F and 150*F a week after the LOCA.
(0.5)
ANSWER 2.27 a.
Exceeds b.
Not exceed c.
Exceeds d.
Not exceed e.
Not exceed my en my Reference (s)
McGuire System Descriptions, Safety Analysis, p. 32. Al so,
l
- Section 2 continued on next page -
35 Points Available QUESTION 2.28 Which of the following statements is CORRECT regarding the Ccid
~
Leg Accumulators?
(1.0)
This is a passive ECC system, requiring no auto / manual action.
a.
b.
Hydrogen pressurization is used to reduce corrosion after St.
c.
Boron concentration greater 4% to prevent a restart accident.
d.
Check valves prevent inflow from the Refueling Storage Tank.
ANSWER 2.28 a
Reference (s)
McGuire PSM Cold Leg Accumulators, Figure MC-SYS-ACC-2.
I
- End of Section 2 -
i i
36 3.0 INSTRUMENTS AND CONTROLS (30.0)
Points Available OUESTION 3.01 For the Nuclear Instrumentation System, match the setpoints given in column A with the function performed at the setpoint given in column B.
Note that setpoints (column B) can be used more than once.
A B
5 a.
Source Range, 10 cps (1) High flux rate trip (0.5) b.
Intemediate Range,10-10 amps (2) Reactor Trip (0.5) c.
I.R., 20%
(3) Pemissive, P-6 (0.5) d.
Power Range /I.R., 25%
(4) Overpower rod stop (0.5) e.
P.R., 103%
(5) Pemissive, P-10 (0.5) f.
P.R., 10%
(6) High flux rod stop (0.5) g.
P.R., 109%
(7) Negative rate trip (0.5)
ANSWER 3.01 2/
($ai p c 16tI S Me
&[l
/ q g(y a.
n a ra. s. T
- d. 2g e.
4 f.
5
- g. 2/
Reference (s)
McGuire Summary %nual (MSM), Excore Instrumentation p. 2a and Figure 2.
- Section 3 continued on next page -
i 37 Points Available QUESTION 3.02 Which of the following signals or events would cause an auto-matic start of a Motor Driven Auxiliary Feedpump?
(1.0) a.
1/4 lu low Steam Generator level in 1/4 S/Gs b.
Trip of Main Feedpump A or Main Feedpump B c.
Failure of Auxiliary Transformers 1 ATA and 2 ATB d.
Engineered Safeguards Features signal ANSWER 3.02 d
Reference (s)
McGuire Lesson Plan on Auxiliary Feed System, p.13 and Les-son Plan on Electrical Main Power Distribution Systems, p.17.
QUESTION 3.03 The Recirculation Valves en the Main Feedpumps must meet the followir.g control criterion:
(1.0) a.
Fail shut to minimize bypass flow.
b.
Manually controlled if pump trips.
c.
Automatically controlled unless pump reset.
d.
Maintain minimum flow of 2000 gpm.
ANSWER 3.03 b
Reference (s)
McGuire Lesson Plan on Main Feedwater System, p.16.
4
- Section 3 continued on next page -
38 Points Available 00ESTION 3.04 Which of the following will automatically close the vent valve on the Component Cooling Surge Tank?
(1.0) a.
Radioactive water leaks into the system.
b.
One of the two trains loses its coolant.
c.
Nuclear Service Water System is lost.
d.
The corrosion prohibitor loses strength.
ANSWER 3.04 a
Reference (s)
McGuire Lesson Plan on Component Cooling Water System, p.10.
OUESTION 3.05 Which of the following is a continuously variable means of pres-sure control in the Pressurizer?
(1.0) a.
Sprays b.
Safety valves c.
Heaters d.
PORVs ANSWER 3.05 c
Reference (s)
McGuire Lesson Plan on Reactor Coolant System, p. 22.
- Section 3 continued on next page -
39 Points Available 00ESTION 3.06 Select the condition (interlock protection) that allows the Reactor Coolant System and the Residual Heat Removal System to be joined.
(1,0)
RCS flow <1000 gpm and Refueling Water Storage tank isolated, a.
RCS pressure <385 psig.
b.
RCS pumps off, RHR pumps on, valves FW-27 and NI-185 open, RCS temperature <425'F, pressure <555 psig.
c.
Containment sump and RWST isolated, RCS pressure <530 psig, temperature <475'F.
d.
PZR vapor space temperature <475'F, RCS pressure <385 psig, RWST and containment sump isolated.
ANSWER 3.06 d
Reference (s)
McGuire Lesson Plan on Residual Heat Removal System, p.19.
- Section 3 continued on next page -
40 l
Points Available OUESTION 3.07 Complete the table of Automatic Safety Injection Signal Setpoints, Logics and Interlocks.
(3.0)
Auto Signal Setpoint Logic Interlock
- a. Low Pressurizer Pressure
- b. Low Steamline Pressure ANSWER 3.07 1845 psig 2/4 P-11 585 psig P-11 Reference (s)
McGuire Lesson Plan on Safety Injection System, p.15.
T3 Tu 3.3 - 3 4
- Section 3 continued on next page -
..e--,-
,4
.--e
. ~ - -
.y
41 Points Available OUESTION 3.08 Af ter a LOCA that is confined to the containment, the Contain-ment Spray System will restart (pumps on, discharge valves open) only on wnich of the following pair of signals?
(1.0) a.
Containment pressure reaches design pressure level of 45 psig, and CPSC starts to go above 0.25 psig.
b.
The Containment Pressure Control Signal starts to go above 0.25 psig and ESF system is RESET.
c.
Containment Pressure Control Signal is 10.25 psig, and all ice is melted.
d.
High-high containment pressure setpoint of 3.0 psig is exceeded, and CPSC starts to exceed 0.25 psig.
ANSWER 3.08 d
I Reference (s)
McGuire Lesson Plan on Containment Spray System, p.11.
Section 3 continued on next page -
t
-e n
42 Points Available OUESTION 3.09 If the Boric Acid Flow Deviation Annunciator alarms on panel 1AD7, which of the following is NOT the reason?
(1.0) a.
Boric Acid Transfer Pump fails to start.
b.
Valve NV-267 (B/A to B/A blender) fails shut.
c.
Makeup control system fails to function normally.
d.
Valve NV-171 (B/A blender to VCT) fails shut.
fISWER3.09 d
Reference (s)
CVCS Composite Drawing in Plant Summary Manual, MC-SYS-NV-4.
1 00ESTION 3.10 The Seal Standpipe low-level alarm for Reactor Coolant pump A has been coming in several times during the past shift. What is a possible reason?
(1.0) a.
Seal #1 is failing b.
High NC drain tank pressure c.
Drain valve is open d.
Makeup valve is shut ANSWER 3.10 C
1 Reference (s)
)
Annunciator Response Procedure 1AD7-A4, McGuire Lesson Plan on Reactor Coolant System, p.11.
- Section 3 continued on next page -
i
43 Points t
Available 00ESTION 3.11 l
Select the statement which describes the effect of lead-lag cir-cuitry on Overtemperature AT trip of the reactor.
(1.0) a.
DNBR limit is reached before Rx trip if both setpoint and AT measurement have lead-lag, b.
Rx trip is reached before DNBR limit if both setpoint and AT have lead-lag.
c.
Rx trip and DNBR limit are reached at the same time if the setpoint only has lead-lag.
d.
Rx trip and DNBR limit are reached at the same time if r
aT only has lead-lag.
ANSWER 3.11 b
4 Reference (s) i McGuire PSM Reactor Protection, Figure MC-IC-IPE-11.
i l
I J
- Section 3 continued on next page -
I n
44 Points Available QUESTION 3.12 The round ball on the base of the portable neutron detector serves the following purpose:
(1.0) a.
The shape identifies the type of detector.
b.
The ball provides shielding for the user.
c.
The ball provides a source of neutrons for calibration.
d.
The ball moderates neutrons for easier detection.
ANSWER 3.12 d
Reference (s)
McGuire Radiation / Rad. Control PSM, Figure N-RAD-RCS-10, Fast Neutron Detector.
- Section 3 continued on next page -
q l
I
45 Points Available OUESTION 3.13 Your film badge contains four separate films to detect Betas, Gamas, Fast Neutrons, and Slow Neutrons. TRUE or FALSE.
(0.5)
ANSWER 3.13 False.
(It has two films, beta windows, and Cd screens to measure the four types of radiation.)
Reference (s) l McGuire Radiation / Rad. Control PSM, fig. N-RAD-RCS-18, Film Badge Construction.
I i
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1
- Section 3 continued on next page -
46 Points Available 00ESTION 3.14 Which of the following statements about the Rod Control Program is CORRECT?
(1.0) a.
The nonlinear gain program in the power mismatch part of the circuit precedes the variable gain.
b.
The auctioneered high nuclear power signal modifies the T-ave input to the T-ave - T-ref summer.
t c.
Auctioneered high T-ave input to the error signal is affected if an RTD measuring T-H fails low.
d.
The T-ref input to the error signal decreases as turbine impulse pressure increases with power.
ANSWER 3.14 a
Reference (s)
McGuire PSM, Reactor Control, Figure MC-IC-RCS-3.
I 1
- Section 3 continued on next page -
47 Points Available OUESTION 3.15 The Digital Rod Position Indicator is capable of the following accuracy:
(1.0) a.
Between 1 and 5 steps b.
Between 6 and 10 steps
}
c.
Between 11 and 15 steps d.
Between 16 and 20 steps ANSWER 3.15 lA Reference (s)
McGuire PSM LED Rod Position, p.14, 2/(3/ff) p ff, f ) gj gh -MC -S/M -LC-ES4 0UESTION 3.16 The failure of a fitting in a reference leg (equivalent to 0.25 inch diameter hole) of the Pressurizer Level Measuring sys-tem produces which of the following effects?
(1.0) l a.
Causes indicated level in pressurizer to fall b.
Causes true level in the pressurizer to fall c.
Has the same effect as a ruptured diaphragm d.
Has the opposite effect of a loss of AP ANSWER 3.16 C
l Reference (s)
McGuire PSM Pressurizer Level, Figure MC-IC-PLC-2.
- Section 3 continued on next page -
5
48 Points Available QUESTION 3.17 Select the CORRECT statement about the Steam Generator Level Control System.
(1.0) a.
Auctioneered High Nuclear Power is the main input to the S/G Level Control program.
b.
Nuclear power from N41 or N43 is one input to the Feed Reg.
Bypass Valve control.
c.
The S/G 1evel signal is modified by LEAD function before com-parison to program level.
d.
The Level Error is an input to both the Bypass Valve and the Main Feed Reg. Valve.
ANSWER 3.17 d
4.
Reference ( s)
McGuire PSM Steam Generator Level, Figure MC-IC-SGL-1.
I
- Section 3 continued on next page -
49 Points Available OUESTION 3.18 The Programmed D/P across any S/G and its Feed Reg. Valve increases with reactor power. TRUE or FALSE.
(0.5)
ANSWER 3.18 True Reference (s)
McGuire PSM Feedwater Pump Speed Control, Figure MC-IC-FWS-3.
QUESTION 3.19 A PORV on the primary system has begun leaking significantly.
List six of eight indications that will detect the leakage.
(3.0)
ANSWER 3.19 1.
Tailpipe tenperature 2.
Acoustic detector 3.
PRT level 4.
PRT temperature 5.
PRT pressure 6.
Primary system pressure decreasing (the four pressure be bp h w b
/i d a d %
channels count as one instrument) or 7.
PORY position indication light j
8.
PORY open alarm.
Score: 0.5 each Reference ( s)
McGuire Systems Descriptions, Reactor Coolant, various pages.
l
- Section 3 continued *on next page -
l
i 50 l
Points Available QUESTION 3.20 1
Which of the following radiation detectors would be most effective following a severe accident that caused local radiation levels of several thousand r/hr?
(1.0) a.
Geiger-Muller tube b.
Beta scint. counter c.
Ionization chamber d.
Gamma scint. counter ANSWER 3.20 c
Reference ( s)
McGuire PSM Radiation Monitoring, p. 5.
QUESTION 3.21 Which of the following interlocks relating to the Steam Dump System is intended to deal with the situation of a 60% Loss of Load?
(1.0) a.
C-9 b.
P-12 c.
C-7A d.
C-7B ANSWER 3.21 d
Reference (s)
McGuire PSM Steam Dump Control, p. 6.
Section 3 continued on next page -
i l
e -
., ~
51 Points Available QUESTION 3.22 If there is a Turbine Runback signal to the DEH control, list three indications you should check to see what generated the signal?
(1.5)
ANSWER 3.22 1.
OTAT 2.
OPAT 3.
Generator breaker open 4.
Stator cooling inadequate r6 L f. ~t -)
S.
Less,+ +c.ap-g (i+
,v.,
Reference (s)
McGuire PSM Digital Electrohydraulic Control System, p. 2.
, nc -p: ~ oc a - 9 OUESTION 3.23 Containment High-High Pressure signal actuates both the Annulus Ventilation system as well as the Hydrogen Purge system in Containment. TRUE or FALSE.
(0.5)
ANSWER 3.23 False.
(The purge system is manually operated based on hydrogen concentration.)
Reference (s)
McGuire Lesson Plans, Annulus Ventilation, pp. 9 and 10.
- Section 3 continued on next page -
k
\\
52 i
Points Available QUESTION 3.24 List three conditions that will prevent an Emergency Diesel generator from automatically starting.
(1.5)
ANSWER 3.24 1.
Local Stop 2.
Engine Overspeed 3.
Turning Gear Engaged DT(
.. c(, (( %M' Mh MM muud) 5.'
.. e.... a,...e....
w Reference ( s)
McGuire PSM Standby Diesel Generator, p. 6.
TS 3/4 - 2r->~
Sie.. F u s,,, Gu,as;l. + y s, t
e Ausd 19 ( (LJ 2-\\,, tw eb 5 5- ( wQ g)
- End of Section 3 -
h
53
4.0 PROCEDURES
NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL (30.0)
CONTROL Points Available 00ESTION 4.01 Which of the following is NOT a correct statement regarding appropriate methods used to accomplish independent verifications?
(1.0) a.
Observation of action being performed b.
Observation of signoff on appropriate procedure step c.
Observation of affected status lights d.
Observation of system response ANSWER 4.01 b
Reference ( s)
McGuire Operations Management Procedures 1-6, p. 5.
i
- Section 4 continued on next page -
i
54 Points Available 00ESTION 4.02 Which of the following statements is NOT correct?
(1.0) a.
A licensed Reactor Operator or licensed Senior Reactor Operator is required to be present at the controls at all times while fuel is in the reactor.
b.
The designated " Operator at the Controls," can serve as the responsible person for Control Room Operations during the absence of the Unit Supervisor from the Control Room.
c.
A relief licensed operator on shift may relieve the " Operator at the Controls" without a written turnover to allow the operator at the controls to be absent from the Control Room for short periods of time.
d.
The " Operator at the Controls" shall not, under any circumstances, leave the surveillance area without obtaining a qualified relief.
ANSWER 4.02 d
Reference ( s)
McGuire Operations Manual Procedures 1-3, pp. 2-4.
- Section 4 continued on next page -
55 Points Available OVESTION 4.03 You are acting as " Operator at the Controls" while the unit is at 100% power. Which of the following occurrences do NOT need to be entered into the Reactor Operator's Log Book?
(1.0) a.
Lifting of relief valve b.
A load reduction request by the dispatcher c.
Release of a load reduction request by the dispatcher d.
Time at which STA is notified of an abnormal situation ANSWER 4.03 d (Unit supervisor's log book only)
Reference ( s)
McGuire Operations Management Procedure 2-4, pp. 3-7.
- Section 4 continued on next page -
t l
l
56 Points Available
)
QUESTION 4.04 The unit has just exceeded technical specification limits for 1
a radioactive release through the unit vent. You have been asked to record meteorological data and note that the wind direction is 225 degrees from north. The SRO asks you in which direction the plume will travel. What is your response?
(1.0) a.
Northeast b.
Southeast c.
Southwest d.
Northwest ANSWER 4.04 a.
Reference (s)
McGuire HP/0/B/1009/09, Release of Radioactive Materials Through Unit Vent Exceeding Technical Specifications, p. 3.
1
- Section 4 continued on next page -
57 Points Available OUESTION 4.05 Which of the following statements pertaining to radiation expo-sure and control is CORRECT?
(1.0) a.
High radiation zone is defined as an area accessible to personnel where you can receive a whole body exposure in excess of 100 millirems in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
b.
The entrance to high radiation zones are open during day shift, but are locked on offshif ts and weekends.
The calendar quarter, for exposure limits, is defined as the c.
3-month period following your last exposure of 1.25 rems.
d.
One of the ALARA methods for reducing exposure to individuals is to use more personnel for shorter times, therefore each person will get a lower exposure.
ANSWER 4.05 a
Reference (s) 1 McGuire Health Physics Manual, Sections 2.5.2.A, 2.5.4.1, 2.1.4, and 3.1.2.
- Section 4 continued on next page -
58 Points Available QUESTION 4.06 Which of the following statements in reference to a reactor startup is CORRECT?
(1.0) a.
The reactor can be taken critical with the shutdown banks partially inserted into the core as long as the shutdown margin requirements are met.
b.
You are allowed to achieve criticality by either rod withdrawal or dilution.
c.
Within 15 minutes of achieving criticality, and at all times when the reactor is critical, NC system lowest loop T-ave is
>551*F.
d.
The reactor operator can exceed a 1.0 DPM sustained start-up rate with permission of the SRO.
ANSWER 4.06 c
Reference (s)
McGuire OP/1/A/6100/01, Unit Startup, pp. 1, 15.
- Section 4 continued on next page -
l i
59 Points Available 00ESTION 4.07 During normal startup, which of the following is NOT a requirement for opening the NC Pump Seal Bypass Return Header Isolation Valve (INV-92A)?
(1.0) a.
Reactor coolant system (NCS) pressure is greater than 100 psig but less than 1000 psig.
b.
Seal injection flow to each NC pump is greater than 6 gpm.
c.
No.1 seal leakoff flow rate is less than 1 gpm.
d.
No. I seal leakoff temperature exceeds 120*F.
ANSWER 4.07 d (200*F)
Reference (s)
McGuire OP/1/A/6200/1, Chemical and Volume Control System, pp. 1-2.
i
- Section 4 continued on next page -
60 Points Available 00ESTION 4.08 According to the limits and precautions concerning unit operation, T-ave should be maintained within how many degrees of T-ref if rod control is in manual?
(1.0) a.
- 1*F b.
- 2*F c.
13*F d.
i4*F ANSWER 4.08 b
Reference ( s)
McGuire OP/1/A/6100/03, Controlling Procedure for Unit Opera-tion, p.1.
QUESTION 4.09 According to the Unit i shutdown procedure, the cooldown rate of the NC (RCS) System should not exceed
'F/hr.
(1.0) a.
50 b.
100 c.
150 d.
200 ANSWER 4.09 a
Reference (s)
McGuire OP/1/A/6100/02, Controlling Procedure for Unit Shut-down, p. 1.
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61 Points Available OUESTION 4.10 In the event of a forced control room evacuation, one R0 is to innediately proceed to the Reactor Trip Switchgear. Where must the other RO immediately proceed to?
(1.0) a.
Aux S/D panel b.
Main turbine front standard c.
2SM-15 (SM to 2nd stg RHTRS) d.
CA local control panels ANSWER 4.10 d (Auxiliary feedwater)
Reference (s)
McGuire AP/2/A/5500/17, Loss of Control Room, pp. 2-3.
QUESTION 4.11 Which of the following is NOT a symptom of a steam generator tube leak?
(1.0) a.
"IEMF-36 UNIT VENT HI GAS RAD" alarm b.
"IEMF-33 COND AE EXH HI GAS RAD" alarm c.
"IEMF-10 11 12 13 STEAM LINE HI RAD" alarm d.
Increase in frequency of auto makeup to VCT ANSWER 4.11 a
Reference (s)
McGuire AP/2/A/5500/10, Reactor Coolant Leakage, p. 2.
McGuire OP/1/A/6100/10R, Panel IRA 02-A2.
- Section 4 continued on next page -
62 Points Available i
00ESTION 4.12 i
During steady state operation, which of the following is NOT a symptom of NC (RCS) leakage?
(1.0) i i
a.
Containment floor / equipment sump level increase b.
"PZR SAFETY DISCHARGE HI TEMP" alarm c.
Decline in T-ave d.
Containment temperature increase ANSWER 4.12 C
Reference (s)
McGuire AP/2/A/5500/10, NC (RCS) Leakage, p. 7.
QUESTION 4.13 Which of the following is NOT a synptom of a loss of contain-ment integrity?
(1.0) a.
An open contaimnent isolation valve b.
Open equipment hatch c.
An inoperable air lock d.
Containment leakage in excess of Technical Specification limits ANSWER 4.13 a (Inoperable)
Reference (s)
McGuire AP/0/A/5500/24, Loss of Containment Integrity, p. 2.
- Section 4 continued on next page -
e
63 Points Available OUESTION 4.14 a.
Under which of the following conditions is emergency boration NOT required?
(1.0)
(a) Failure of more than one (1) control rod to fully insert following a reactor trip (b)
Less than 2000 ppm boron is mode 6 (c) Uncontrolled or unexplained reactivity decrease (d)
Shut-down margin less than 1.3% AK/K l
b.
Which of the following methods will NOT enhance emergency boration?
(1.0)
(a)
Start second BA transfer pump (b)
Start second NV pump (c) Open NV pump recirculation valves (d)
Increase letdown to maximum.
ANSWER 4.14 a.
(c) (increase) b.
(c) (close)
Reference (s)
McGuire AP/1/A/5500/38, Emergency Boration, pp. 2-3.
- Section 4 continued on next page -
64
=
Points Available OUESTION 4.15 The reactor has been operating at full power when steam-line pressures indicate 500 psig and " Safety Injection Actuated" status light becomes lit. List.your immediate actions.
(3.0)
ANSWER 4.15 1.
Manually exercise Rx Trip Train 1A AND 1B switches.
2.
Verify Reactor Trip.
3.
Verify Turbine / Generator Trip.
4.
Verify 1 ETA AND 1ETB - Energized.
5.
Verify Load Sequencers actuated.
Reference (s)
McGuire EP/1/A/5000/01, Safety Injection, pp.1-3.
QUESTION 4.16 In order to provide a minimum secondary heat sink af ter a reactor trip, the CA ( AFW) flow shall be greater than gpm.
(1.0) a.
600 b.
450 c.
300 d.
150 ANSWER 4.16 b
Reference (s) 1 McGuire EP/1/A/5500/01, Reactor Trip, pp. 5-6.
- Section 4 continued on next page -
65 l
I Points Available OUESTION 4.17 Which of the following is the NC (RCS) pump trip criterion during post-LOCA cooldown and depressurization?
(1.0) a.
NC System pressure less than 1500 psig.
b.
NC System subcooling is lost.
c.
Both NV pumps are injecting through SI flow path.
}
d.
"FWST LEVEL LO" alarm is actuated.
ANSWER 4.17 b
Reference (s) l McGuire EP/1/A/5000/2.2, Post-LOCA Cooldown and Depressuriza-l tion, Enclosure 1.
i OUESTION 4.18 i
Which of the following conditions does NOT require manual initiation of SI?
(1.0) a.
PZR Level - 10%
b.
PZR Pressure - 1780 psig c.
S/G SM Pressure - 520 psig d.
Containment Pressure - 2 psig ANSWER 4.18 a
Reference (s)
McGuire EP/1/A/5000/1.2, SI Termination Following Spurious SI, Enclosure 1.
- Section 4 continued on next page -
66 Points Available 00ESTION 4.19 Match the correct type of line with the appropriate Critical Safety Function Tree branch color.
(2.0) a.
Green 1.
Dashed line b.
Orange 2.
Dotted line c.
Red 3.
Double line d.
Yellow 4.
Single solid line ANSWER 4.19 a.
3 b.
I c.
4
,_ 2 4,,,g d.
2 Reference (s)
McGuire EP/1/A/5000/10, Critical Safety Function Status Trees (CAF).
Section 4 continued on next page -
1
N 67 Points Available QUESTION 4.20 During a reactor transient, a check of the CRITICAL SAFETY FUNCTION STATUS TREES directs you to the Response to Degraded Core Cooling procedure. You have determined that the. core exit T/C temperatures are above saturation. Which of the following is NOT an indication of inadequate core cooling?
(1.0) a.
Core exit T/Cs 1300'F b.
Core exit T/Cs 400*F and increasing c.
Rx vessel (low-range level 37%)
d.
Rx power 8%
ANSWER 4.20 d
Reference ( s)
McGuire EP/1/A/5000/12.2, Response to Degraded Core Cooling,
- p. 12.
l
- Section 4 continued on next page -
68 Points Available 00ESTION 4.21 Provide the Technical Specifications Reactor Coolant System leakage limits for each of the following.
a.
gpm PRESSURE BOUNDARY LEAKAGE (0.5) b.
gpm total primary-to-secondary leakage through all steam generators (0.5) c.
gpm IDENTIFIED LEAKAGE (0.5) d.
gpm at an RCS pressure of 2235 psig from any RCS Pressure Isolation Valve (as specified in TS Table 3.4-1)
(0.5)
ANSWER 4.21 a.
O b.
I c.
10 d.
1 Reference (s)
McGuire Technical Specifications 3.4.6.2.
i
- Section 4 continued on next page -
l
69 Points Available 00ESTION 4.22 According to McGuire Technical Specifications, which of the fol-lowing does NOT have to be restored to an OPERABLE status within one (1) hour?
(1.0) a.
Upper Head Injection nitrogen bearing accumulator pressure at 1000 psig when at 80% power.
b.
Inoperable Safety Injection Pump when at 80% power.
c.
Inoperable Centrifugal Charging Pump when in Mode 4.
d.
Inoperable ice condenser door when in Mode 4.
ANSWER 4.22 Reference (s) i McGuire Technical Specifications 3.5.1.2, 3.5.2, 3.5.3, and 3.6.5.3 (2/25/85 Memorandum).
- Section 4 continued on next page -
l
1 70 Points Available OUESTION 4.23 Which of the following Technical Specification limits is the same for both Unit 1 and Unit 2 while in Mode 1?
(1.0) a.
Shut-down margin b.
Moderator temperature coefficient g,3fL M hc-c.
Cold-leg injection accumulator volume d.
Refueling water storage tank minimum volume ANSWER 4.23 d
glare
.e. A 54
(
Reference (s)
McGuire Technical Specifications 3.1.1.1, 3.1.1.3, 3.5.1.1, and 3.5.5.
A-vw e uj. 4 2. ( M / \\, hm[. R 3(LLO 2.3 OUESTION 4.24 Which of the following is CORRECT regarding the required number of 4160-volt emergency, 600-volt emergency, and 120-volt A.C.
Vital busses required for Mode 1 operation?
(1.0) a.
Four 4160V, Four 600V, Four 120V b.
Two 4160V, Four 600V, Four 120V c.
Two 4160V, Two 600V, Four 120V d.
Two 4160V, Two 600V, Two 120V ANSWER 4.24 b
Reference (s)
McGuire Technical Specification 3.8.3.1.
- Section 4 continued on next page -
k
71 Points Available QUESTION 4.25 Which of the following is NOT necessary to consider when deter-mining if a diesel generator is OPERABLE?
(1.0) a.
A day tank containing 120 gallons of fuel b.
An air-start receiver with 220 psig air pressure c.
A Fuel Storage System containing 28,000 gallons of fuel d.
A fuel transfer pump ANSWER 4.25 b
Reference (s)
]
McGuire Technical Specification 3.8.1.1.
- End of Section 4 -
END OF EXAMINATION l
~
EQUATION SHEET Where ml
- 82 (density)1(velocity)1(area)1 = (density)2(velocity)2(area)2 2
PE +KE +P V 1 i = PE +KE +P V22 where V = specific KE = mv PE = mgh 2
2 i
i T
volume P = Pressure Q = mc (Tout-Tin)
Q = UA (T
-Tstm)
Q = m(h -h )
p ave t 2 P = P 10(SUR)(t) p, p e /T SUR = 26.06 T = (B-p)t t
o g
T p
delta K = (Keff-1)
CR (1-K,ffi) = CR (1-Keff2)
CR = S/(1-KeffI 1
2 M = (1-Keff1)
SDM = (1-Keff) x 100%
(1-Keff2)
K eff 1 = A e-(decay constant)x(t) in (2) 0.693 A
decay constant
=
=
g t
t 1/2 1/2 Water Parameters Miscellaneous Conversions 1 gallon = 8.345 lbs 1 Curie = 3.7 x 1010 dps 1 gallon = 3.78 liters 1 kg = 2.21 lbs 1 ft3 = 7.48 gallons I hp = 2.54 x 10 Btu /hr 3
3 6
Density = 62.4 lbg/ft 1 MW = 3.41 x 10 Btu /hr Density = 1 gm/cm 1 Btu = 778 ft-lbf Heat of Vaporization = 970 Btu /lbm Degrees F = (1.8 x Degrees C) + 32 Heat of Fusion = 144 Btu /lbm 1 inch = 2.54 centimeters 2
1 Atm = 14.7 psia = 29.9 in Hg g = 32.174 ft-lbm/lbf-sec h