ML20154F030
| ML20154F030 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 08/31/1988 |
| From: | Mccrory S, Pellet J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20154F020 | List: |
| References | |
| 50-285-OL88-03, 50-285-OL88-3, NUDOCS 8809190226 | |
| Download: ML20154F030 (96) | |
Text
J x
APPENDIX U.S. NUCLEAR REGUALTORY COMISSION REGION IV Operator Licensing Exam Report:
50-285/0L88-03 Operating License: DPR-40 Docket No: 50-285 Licensee: Omaha Public Pa.ver District (OPPD) 1623 Harney Omaha, Ne 68012 Facility Name-FortCalhounStation(FCS)
Examination at:
Fort Calhoua Station b *f/
f 3//(I Chief Examiner:
, /
- 5. LY McCrof'y, Lead E D6te '
Operator Licensing S(fdminer,
- ction, Division of Reactor Safety Approved by:
M be 9(/5/ [
- Chief, D6te adyy,d.vt.' Pellet,SectipSection,
/,c /
Operator Licensing Division of Reactor Safety Sumary NRC Administered Examinations Conducted During The Week of July 26, 1988 (Report 50-285/0L88-03)
NRC administered examinations to one Reactor Operator and four Senior Reactor Operator applicants, All applicants passed all portions of the examination and have been issued the appropriate license.
8809190226 800909 PDR ADOCK 05000205 V
pg
r
'd 2
DETAILS 1.
Persons Fxamined o
SR0 R0 Total License Examinations:
Pass -
4 1
5 Fail -
0 0
0 2.
Examiners g
S. L. McCrory, Chief Examiner F. Jaggar 3.
Examination Report Perfomance results for individual examinees are not included in this report as it will be placed in the NRC Public Document Room and these results are not subject to public disclosure.
o a..
Examination Review Coment/ Resolution l
In general, editorial comments or changes made during the examination, or subsequent grading reviews are not addressed by this resolution section. This section reflects re3olution of substantive comments made by FCS.
The only coments addressed in this section are those which were not accepted for incorporation into the 3
examination and/or answer key.
Those comments accepted are incorporated into the master examination key, which is includad in this report.
Comments may be paraphrased for brevity.
The full text of the comments is attached.
(3.09)
Part b asks for trip logic, not the number of l
channels required for a trip.
Response NRC considers the trip logic to include both the number of detectors needed to trip a channel and the number of channels needed to initiate a reactor trip signal.
(4.03)
The question should be deleted because there is an abomal difference between CETs and RCS T-hot in the answer i
indicated in the key and no other answer is correct.
Response Due to the low flow rates during natural circulation and therelativesensitivitiesofRTDsandCETs,(itislikely l
that RTO indication will appear to stablize temporarily) while CETs show a temperature decrease.
(4.06)
Expand the answer key for Part c to include: Meet MTC values as prescribed by the Reactor Engineer.
Response The question deals with restrictions on rod insertion.
This response is concerned with rod withdrawal restrictions.
4 1
7 3
b.
Site Visit Summary (1) At the end of the written examination administration, the facility licensee was provided a copy of the examination and answer key for the purpose of commenting on the examination content validity.
It was explained to the facility licensee that regional policy was to have examination results finalized within 30 days. Thus, a timely response was desired to attain this goal.
(2) The Chief Examiner comunicated to the facility licensee via the senior resident inspector a concern that Abnormal Operating Procedures (A0Ps) 6 and 7 lacked adequate guidance to the shift supervisor for implementation of the emergency plan. AOPs 6 and 7 provide guidance for plant control when evacuation of the l
control room is required. The procedures simply direct the shift supervisor to implement the emergency plan but do not instruct him as to where he should go to accomplish this, it was communicated to the Chief Examiner during the course of examination that shift supervisors had informally agreed to go to the technical support center to initiate the emergency plan when required to evacuate the control room.
To avoid possible confusion, the AOPs should specify where the shift supervisor reports to initiate the emergency plan following control room evacuation.
t c.
General Comments During examination development, it was determined that no lesson andAbilities(providedforcertainsystemscoveredintheKnowledge plans had been KAs) catalog, NUREG-1122.
Further, some of the questions provided in the facility licensee examination bank referenced lesson plans which had not been provided to NRC for examination development. OPPO should review the systems listed in NUREG-1122 i
against existing lesson plans and training material and provide the NRC with any lesson plans or other training material related to those systems. Where lesson plans or training material do not exist for a specific system listed in NUREG-1122, OPPD should provide justification why licensed operators are not trained on those systems.
Similarly, the Plant-Wide Generic KAs (194001 series) should be J
reviewed to determine existence of lesson plans or training material vr a justification as to why it is not needed.
Lesson plans currently held by the NRC RIV examiner staff are referenced in OPPD Letter LIC-88 385 of May 23, 1988.
d.
Master Examination and Answer Key Master copies of the FCS license examinations and answer keys are attached.
The facility licensee coments, which hava been accepted, i
l are incorporated into the answer key, i
f 4
- - - -, -. -,, - ~,,
4 e.
Facility-Examination Review Coments The facility licensee coments regarding the written examination are attached. Those coments which were not acceptable for incorporation into the examination answer key have been addressed in the resolution section of this report.
F I
U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE.EXAMINA110N FACILITY:
_EI._CALWQubl____ _ _
REACTOR TYPE:
_RWR=CF DATE ADMINISTERED _88Zd2224 EXAMINER:
_MCCRORY,_S.
CANDIDATE:
lblSIEUCIl0blS_IO_CANQ10eIEJ.
Une. separate paper for the answers.
Write answers on one side only.
-Staple question sheet uon top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category _and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
X OF CATEGORY
% OF CANDIDATE'S CATEGORY
__VALUE_ _IDIe1. __Srnor
_gan iin
_ rr r nojo ____________
_25.00__ _25.00
___ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
.25.QQ__ _25.QQ
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_25.QQ__ _25.QQ
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
.25.QQ__
25.QQ
________ 8.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 1Q0.0Q__
Totals l
Final Grade i
All work done on this examination i s my own.
I have neither given i
nor received aid.
Candidate's Signature i
{
l l
I
(
i
5.
Ti&ORY OF NLICIFAR POLIFR RfANT OEFRATION. F1 li f DSeeND PAGE 2
IHFRMODYNOMICS QUESTION 5.01 (1.00)
During a reactor startup, an initial reactivity addition causes power to increase from 20 to 40 CPS.
A second reactivity addition causes power to change to 80 CPS.
Which of the f ol l owi ng s t a temen t s is correct?
i n.
The first reactivity addition was smaller.
b.
The second reactivity addition was smaller.
c.
The first and second reactivity additions were equal.
d.
There is insufficient data to determine a relationship between the reactivity additions.
(1.0)
ANSWER 5.01 (1.00) e b.
The second reactivity addition was smaller.
(1.0)
^
REFERENCE l
E08 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.13.1 001010K536 015000K506
...(KA'S)
I QUESTION 5.02 (1.00)
With the reactor critical at 5000 CPS, CEA withdrawai is used to inc^ ease power to 10000 CPS.
Select the statement below that correc tly describes l
the position of the CEA's after power is stabilized at 10000 CPS.
(1.0) a.
The rod position will be higher than previous as more fuel must be l
exposed to the neutron flux to maintain the higher power level.
b.
Rods will be further withdrawn to compensate for the temperature I
defect at the higher power.
c.
The rod position will be the same because the outward motion needed to j
attain a given startup rate equals the inward motion necessary to l
Kill the startup rate.
l d.
The final position will be lower due to the increasing delayed neutron population associated with the higher power level.
r 1
,s.
.Turney nr wiirt ran payre pl un' nps pat f nN - El 11i nR ; ablD PAGE 3
'IHFRMnDYNAMICS f
ANSWER 5.02 (1.00) c.
Rod position will be the same (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS'LP 7-5-3, LO 3.4.3 001000K510
...(KA'S)
QUESTION 5.03 (1.00)
The moderator temperature coefficient (MTC) varies with certain plant conditions.
The MTC Cchoose one correct answer) a.
Becomes more negative as boron concentration is increased.
b.
Varles due to temperature (Tavg) because of the non-linear density changes of water as temperature changes.
c.
Causes axial flux distribution to be til ted toward the ' top of the core at the beginning of life.
d.
Wo9l d be expec ted to introduce a large negative reactivity in the event of a major steam line break.
ANSWER 5.03 (1.00)
B (1.0)
REFERENCE i
E0B GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.6.8 d
192004K103
...(KA'S)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
4 i
5-THEORY._QE tduCI N POWER _ELatfL.QEEBaI1Qti ELu1DS AND PAGE 4
IBERMODYNeMlCS OUESTION 5.04 (1.00)
In the event of a rod ejection accident, which one will be the first reactivity coefficient to insert negative reactivity?
a.
Doppler coefficient, b.
Pressure coefficient.
c.
Void coefficient.
d.
Moderator temperature coefficient.
ANSWER 5.04 (1.00)
[
A (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.7.5 192004K107
...(KA'S) 1 QUESTION 5.05 (1,00)
It takes less reactivity to go prompt critical at:
a.
BOL because of the higher value of beta effective.
b.
BOL because of the l ower val ue of be ta ef f ec t ive.
i c.
EOL because of the higher value of beta effective.
d.
EOL because of the lower value of beta effective.
ANSWER 5.05 (1.00)
D (1.0)
REFERENCE l
E0B GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-4, LO 4.5 i
192003K106 152003K107 192003K108
...(KA'S)
I i
(***** CATEGORY 05 CONT!aVED ON NEXT PAGE * * * **)
f
a 5._aIHEOnv nr NUCLEAR _EQldER_ELablI_QEERGIlot4_ELU1DS: mn PAGE 5
IBERdOQYNad1CS QUESTION 5.06 (1.00)
Which one of the f ol l owi ng statemen ts is correct concerning the change in differential boron worth (% delta K/k) with RCS boron concentration (range of 0 to 1600 ppm) and Tave (range of 532 degF to 580 degF)?
a.
It increases as Tave and RCS boron concentration increase.
b.
It decreases as RCS boron concentration increases but is constant as Tave increases, c.
It decreases as Tave and RCS boron concentration increase.
d.
It increases as Tave increases but is constant as RCS boron concentration increases.
ANSWER 5.06 (1.00)
C (1.0)
REFERENCE EGB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.5.4 192004K109 192004K110
...(KA'S)
QUESTION 5.07 (1.00)
Volding has occurred in the RCS, in the vicinity of the reactor vessel during a natural c,lrculation cooldown.
Which of the f ol l owi ng CORRECTLY characterizes the process of collapsing the void?
a.
The void will superheat if an attempt is made to collapse it too fast.
The rate of collapse will be governed largely by ambient heat loss from ths void.
b.
The void will collapse at a rate equivalent to the rate of HPS! f l ow I therefore, full HPSI should be run until the void is fully collapsed.
c.
The vold will be composed largely of hydrogen gas, and will therefore require degasifying of the RCS in order to begin collapsing it.
d.
The void will collapse immediately upon increasing the pressure above the local saturation pressure; the main concern i s wa t e r h amn.'r.
l r
b
, L. THEORY 'OF NilFLEAR ELLIER_.ELANT DEEReIInN. F1tlIDS eblD PAGE d
IwreMnnyNaMrcS ANSWER
'5.07
.(1.00)
A ( 1. 0 )'
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR SECY 82-475 193007K104
...(KA'S)
QUESTION 5.08 (1.00)
The ratio of Pu-239 and Pu-240 atoms to U-235 atoms changes over core Ilfe.
Which one of the pairs of parameters below is most affected by this change?
a.
doppler coefficient and beta b.
moderator temperature coefficient and doppler coefficient c.
beta and moderator temperature coefficient
- d..
moderator temperature coefficient and neutron generation time ANSWER 5.08 (1.00)
A (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.7.5 192004K107
...(KA'S)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
~
i
,5 d WEDEX 0F Alltrl FAQ PGdER_ELebfI_QEESAIladrELULOS,_ablO PAGE 7
IMEEtdQD'Ibletd1CS QUESTION 5.09 (1.00) i A general rule is often stated "doubling the count rate halves the margin to criticality".
This is mathematically stated by the equation:
CR1/CR2 = (1-Keff2)/(1-Keff1).
Which one of the. f ollowing statements is CORRECT concerning the above statement and equation?
l c.
Equal changes in Keff result in equal changes in subcritical multiplication level.
b.
Both Kefft and Keff2 have to be less than 1.0.
c.
The equation only approximates the instantaneous change in count rate; once-the equilibrium value is reached, the count rate will be higher.
d.
A second doubling of the count rate will result in the reactor becoming critical or supercritical.
ANSWER 5.09 (1.00) 8 (1.0)
REFERENCE t
E08 GENERIC FU4DAMENTALS FOR CE PWR FCS LP 7-5-4, LO 5.1 001010K516 192003K101
...(KA'S) i i
i i
(
i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
L_.IHEDRY OF NilffFAR_RDWEB_ELANT DEEBellfLN; FI lif DR. AND PAGE 8
IHFRMODYNAMfCS
?
. QUESTION 5.10 (1.00)
Which one of the following correctly describes the response of core del ta T if the RCS cool an t fl ow rate is reduced while the steam system is controlling to maintain 80% electrical output?
The plant has been at 80%
of full power, steady state, for 10 days.
All control rods are fully withdrawn (ARO).
a.
Delta T will increase but nuclear power remains the same to provide the same megawatt thermal output.
b.
Delta T will remain the same but nuclear power will increase to provide the same megawatt thermal output.
c.
Delta T will remain the same since the turbine / generator output power has not changed.
d.
Del ta T will increase and nuclear power must increase to provide more steam flow to the turbine.
ANSWER 5.10 (1.00)
A (1.0)
REFERENCE EOB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-7-22, LO 7.2-2 193007K108
...(KA'S)
QUESTION 5.11 (1.00) 1 The reactor is critical at 10E-3 % power when a S/G Atmospheric dun.p valve fails open.
Assuming BOL conditions, no rod motion, and no reactor trip, choose the answer below that best describes the values of Tavg and nuclear 1
power for the resulting new steady state.
(POAH = point of adding heat).
r a.
Final T avg greater than inttial Tavg, Final power above PDAH.
b.
Final Tavg greater than initial Tavg, Final power at PDAH.
c.
Final Tavg less than initial Tavg, Final power above POAH.
d.
Final Tavg less than initial Tavg, Final power at POAH.
i i
b i
d I
. L_ Twsney or NortFae EIgre ptANT gesRATinN. FI lif DS, AND PAGE 9
THERMnnyNaMicS ANSWER 5.11 (1.00)
C (1.0)
REFERENCE EQB GENERIC ':UNDAMENTALS FOR CE PWR FCS LP 7-5-4, LO 1.3 192008115
...(KA'S)
QUESTION 5.12 (1.00)
Which one of the following best supports the reason why Xenon reactivity increeses sharply after a trip following 1000 hrs. of operation at 100%
power?
a.
Iodine half-life is much shorter than Xenon half-life.
b.
Xenon decays less rapidly due to a r4 duction in the neutron flux.
c.
Iodine production is greatly reduced and Xenon production is greatly increased due to the reduction in neutron flux.
d.
Due to reduced neutron absorption, Iodine concentration increases, and Xenon decays directly from lodine, thus Xenon increases.
i ANSWER 5.12 (1.00)
A (1.0) s REFERENCE E0B GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3. LO 3.12.6 192006K107
...(KA'S) l I
l i
I
(***** CATEGORY 05 CONTINUED ON NEX1 PAGE
- )
, 5._,IHEORY OF NUCLEaE_EDWER_ELeblI_DEEEellatLELU1DSJQ PAGE 10 IHEBUOQYNAMICS QUESTION 5.13 (1.00)
Which one of the belav requires the most heat removal during condensation?
(assume steam is of equal quality) a.
two pounds of steam at 1200 psia.
3 b.
two pounds of steam at 600 psia.
c.
one pound of steam at 300 psia.
d.
one pound of steam at 0 psia.
ANSWER 5.13 (1.00)
B (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR steam tables FCS LP 7-7-16, LO 1.6-1 193003K125
...(KA'S)
QUESTION 5.14 (1.00)
As core age increases during a cycle, the integral rod worth at 100% full power of a regulating CEA located near the center of the cores a.
Increases due to the fact that temperature of the CEA is higher.
b.
Increases due to the fact that the CEA is operating in a region of higher neutron flux.
[
c.
decreases due to the fact that the temperature of the CEA is higher.
d.
decreases due to the fact that the boric-acid concentration in the core is less.
I ANSWER 5.14 (1.00) 8 (1.0) j i
4
.5.
.TMFORY OF NilciFAR Pntsk PiANT OPFRATI 4 ELUIDS, AND PAGE 11 INFRMnnyhed1CS REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.4.6 192005K107
...(KA'S)
QUESTION 5.15 (1.00)
The term "critical heat flux" refers to that heat flux.between a fuel pin cnd the bulk of the primary coolant:
a.
at which the heat-transfer mechanism changes between natural convection and nucleate boiling.
b.
at which DNBR would equal 1.3.
c.
at which the heat-transfer mechanism changes between nucleate boiling and partial film boiling.
d.
which is 77% of that which would cause clad failure.
ANSWER 5.15 (1.00)
C-(1.0)
REFERENCd EQB GENERIC FUNDAMENTALS FOR CE PWR FCG LP 7-7-18, LO 3.3-2 193008K106
...(KA'S) l t
OUESTION 5.16 (1.00)
A relief valve on a pipe opens at 885 psig. The temperature of the steam exhausted to the atmosphere is 320 deg. F. The fluid within the pipe is saturated steam. Using the steam tables provided, what is the approximate temperature of the fluid within the pipe, in the immediate vicinity of the relief valve?
(Select one response from below.)
l a.
530 deg. F l
b.
400 deg. F c.
320 deg. F V
d.
212 deg. F i
., L..IMEORY OF SLjctFAR P&ER PiANT OPFRAT f DN,F-LU f DS _eND PAGE 12 4
IMFRMnbyNaMICS ANSWER 5.16 (1.00)
A (Gaturated steam at 885 + 15 psia.) (1.0)
REFERENCE E0B OENERIC FUNDAMENTALS FOR CE PWR CE Steam lables FCS LP 7-7-17, LO 2.3-1 193004K115
...(KA'S)
GUESTION 5.17 (1.00)
Which of the f ollowing statements concerning power def ec t is correct?
c.
The power defect is the difference between the measured power coefficient and the predicted power coefficient.
b.
The power defect increases the rod worth requirements necessary to maintain the desired shutdown margin following a reactor trip.
l c.
Because of higher boron concentration, the power defect is more negative at the beginning of core Ilfe.
d.
The power defect necessitates the use of a ramped Tavg program to l
maintain an adequate subcooling marqin.
ANSWER 5.17 (1.00) 8 (1.0)
REFERENCE E0B GENERIC FUNCW1ENTALS FOR CE PWR l
FCS 7-5-3, LO 3.11.3 001000K549
...(KA'S) t i
t 4
6 l
l i
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE *****)
i
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. 5._.lBEORY._DE_tJUCLE6R_EOWER_ELatJI_DEEReIlotJ ELU1DS,_auQ PAGE 13 IBERd0 DYNAMICS QUESTION 5.18 (1.00)
With RCS pressure at 2100 psia, one of the seals in a RCP assembly fails.
Assume original leakoff flow was 1.0 gpm and leakoff backpressue is 50 psia.
Which DNE of the following is correct concerning both the pressure drop across each operable seal and leakoff f l ow?
a.
1050 psid and 1.5 gpm b.
1050 psid and 1.2 gpm c.
1025 psid and 1.2 gpm d.
1025 psid and 1.5 gpm ANSUER 5.18 (1.00)
C 1025 psid and 1.2 gpm (1.0)
REFERENCE EQ8 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-11-20, LO 1.7d 003000A201
...(KA'S)
OUESTION 5.19 (1.00)
Which of the following statements about burnable poisons is NOT true?
c.
As core age increases burnable poison effects partially compensate for fission product buildup effects.
b.
Including burnable poison in the fuel affects the moderator temperature coefficient.
c.
As cor e age increases burnable poison effects partially c omp e n s a t e for fissile depletion of the fuel.
d.
Including burnable poison in the fuel affects the doppler coefficient.
ANSWER 5.19 (1.00) 0 (1.0)
REFERENCE E00 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.3.2
}
. L IBEDELDE_NUCL FAR EDWER_ELANT DPERAT f DN. FLUJ.DS auD PAGE 14 IwreMnnYNAMICS 192007K101
...(KA'S)
QUESTION 5.20 (1.00)
The reactor trips from full power, equilibrium xenon conditions.
Six (6) hours later the reactor is brought critical at 5x10E-4%.
If power level is maintained at 5x10E-4%, which of the f oll owing statemen ts is COFREC1 concerning control rod motion?
c.
Rods will have to be withdrawn since xenon will closely follow its normal build-in rate.
b.
Rods will approximately remain as is as the xenon estaL11shes its equilibrium value for thi s power level.
c.
Rods will have to be inserted since the critical reactor will burnout
- xenon, d.
Rods will approximately remain as is since the xenon concentration is independent of time.
ANSWER 5.20 (1.00)
A (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-5, LO 5.3.11 192006K114
...(KA'S)
QUESTION 5.21 (1.50)
For each of the following events, descr ibe HOW and WHY RCP ' A' motor emperage will vary from its initial magnitude prior to the event.
a.
It is running with two (2) other RCPs and the fourth RCP is started.
I b.
I ts pump becomes mechanically uncoupled from the motor.
c.
Bearing degradation in the pump results in partial binding.
1
~
- 5. _.IHEORY. DE tdLICLEAR_ Pol #9 PiANT QPEReLL E _EL111DS.,_ablD PAGE 25 IHFRMODYtJaMI CS ANSWER 5.21 (1.50)
O.
The amps will decrease (0.2) because the flow through the pump decreases due to the increased backpressue from starting the fourth pump (0.3) b.
Amps will decrease (0.2) because the motor no longer has a load (0.3).
c.
Amps will increasa (0.2) because friction is increasing the load on the motor (0.3).
t 1
REFERENCE E08 GENcRIC FUNDAMENTALS FOR CE PWR FCS LP 7-7-21, LO 6.1.3
}
191004K107
...(KA'S)
QUESTION 5.22 (1.00)
The amount of heat being added by the reactor coolant pumps (choose one correct answer) a.
Is less than the RCS heat loss to ambient at operating temperature.
1 b.
Is less than the ri.ount of heat being lost to letdown at operating I
temperature.
c.
Is insignificant at normal operating temperature.
i l
I r
d.
Causes total S/G thermal output *o be greater than the thermal output of the core itself.
ANSWER 5.22 (1.00)
I D or C (1.0) i REFERENCE E0B GENERIC Fu1DAMENTALS FOR CE PWR FCS LP 7-11-20 i
003000K506
...(KA'S) i i
i f
I 1
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~,
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1 9
M-
'NQE'4_QEJUC1. EAR PQWER_PL*LQEEBell0N+_ELU1DS+_eNQ PAGE.16 IWERMQQYued1CS j
QUESTION 5.23 (1.00)
During a LOCA with a resultant loss of subcooling margin, Reactor Coolant Pumps (RCPs) are cecured for which one of the f ol l owi ng reasons.
I a.
To prevent pump damage resulting from operation under two phase conditions.
f b.
To prevent core damage resul ting from phase separation upon subsequent loss of RCS flow.
3 c.
To reduce RCS pressure by removing the pressure head developed by the RCPs.
d.
To remove the thermal heat being added to the RCS by the cperating RCPs.
2 ANSWER 5.23 (1.00)
A or 8 (1.0)
REFERENCE EGB GENERIC PUNDAMENTALS FOR (A P'JR l
000009K323 193008K114
...(KA'S) i I
00ESTION 5.24 (1.00)
Which of the following stat =ments about temperature measurement is NOT 4
CORRECT?
i 1
a.
If the sensing wire of an RTD breaks,
- 6. h e instrument will read l
offscale high.
l If a thermocouple wire breaks, the instrumeit
- ?'
- ead off scale low.
5 The temperatur> range which can be measured by an RfD is smaller than
[
that for a thermocouple.
d.
RTDs respond faster to temperature changes than thermocouples.
I s
I t
i i
.L.THEDRY OF NHet rap gnuro pl our DREnaTf Dt4 EL1.llDAED
.PAGE 17 IWFRMnnyhieMiel i.I
. ANSWER 5.24 (1.00)
D (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR r
191002K114
...(KA'S)
UUESTION 5.25
<.50)
As Keff approaches unity, for the same increase in Keff, a greater length of time is regt red to reach the new equilibrium level.
This is a charac-teristic ois O.
Moderator effects o:- neutron leakage.
1 b.
Inc.reasing population of celayed neutrons.
c.
Reactivity criange on power increases.
d.
Subcritical multiplication.
ANSWER 5.25
(.50)
D (0.C)
REFERENCE l
E90 GENERIC FUNDAMENTALS FOR CE PWR FCS IP 7-5-4, LO 5.1 192043K101
...(KA'S) i i
(
,o*-
J'
%TEGORY 05 *****)
i L
1 L
.m
~~
- 6. J:1.6tlI.SYSIEtiS_QESIGtJ._CatlIROL._atJD_1t1SIRt.tdEtJIGIlOtl PAGE 18 QUESTION 6.01 (2.50)
Using figure 2-1, show and label all major penetrations into the Reactor Coolant System.
DO NOT include instrument or test penetrations.
(Eighteen properly placed and labelled penetrations are required for full credit.)
ANSWER 6.01 (2.50)
/---RCP D---\\
/---RCP B---\\
/
\\<---SI CHG--->/
\\
/
\\
/
\\-->RCDT RCDT(--/
\\
VENT SI-->/
\\ drain drain
/
\\
/!\\
/
\\
/
N
/
SURGE
\\
/
N----!--------/
N SG 28<-----!--------!-------!
REACTOR
!-----!----!----!------>SG 2A
\\
\\!/
\\'/
/-------------N
\\!/
\\!/
/
\\
/<--SI
\\
RCDr SL
/
\\
/
\\ drain
/
drain
\\
/---> SPRAY
\\--->SI
/
RCDT<-!---s
/
\\
/--> RCD 1
\\
/ (- --CHG SPRAY--->\\
/
drain LO<---!
N---RCP C---/
N---RCP A--- /
(EIGHYEEN PENETRAT!ONS AT 0.139 EA)
REFERENCE FCS LP 7-11-20, OBJ. 1.2 002000K!00
...(KA'S)
(***** CAlEGORY 06 CONTINUCD ON NEXT PAGE
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. L..EletdI_SYSIEt1S DES 10blyCDtJIROL,_ AND_INSIRLMEblIeIl0td PAGt!
19 I
I QUESTION 6.02 (2.00) o.
Uhat two (2) conditions will geizerate a "Containment Vent Fan Off Normsl*?
(0.$)
b.
What is the failed position for'the face dampers and bypass dampers on the containment cooling and filtering units?
(0.5) c.
When the handswitch2s for the dampers in part b are placed in the UNFILTERED position, what provides the motive force to position the dampers?
(0.5) d.
Why is it important to maintain charcoal bed temperatures below 360 deg.F even after containment air temperatures are returned to normal following a large break LOCA?
(0.5) i ANSWER 6.02 (2.00) a.
control switch in PULL-0UT 69 permissive switch not in AFTER CLOSE position fan breaker not racked !n (any 2 at 0.25 ea) b.
face dampers fall open (0.25) bypass dampers fall shut (0.25) c.
Control air (instrcment air) (operates an air-pit'on to position the dampers for unfiltered operation.) (0.5) l d.
Prevent the release of cortaminants abscrbed by the charcoal. (0.5)
REFERENCE I
LO 1.2, 1.6, 1.8 022000S010 022000SG7 103000A401
...(KA'S) t 1
f QUESTION 6.03 (3.00)
Using the CVCS line diagram provided, circle each component which receives
[
a control signal as a result of a PPLS and indicate the expected response l
j of the component to the control signal.
l r
I i
I i
t i
[
o LEl.ANI_SYSIEtiS_ DES 10W._C211ROL+_AND_1NSIRik1EblIAI1CH PAGE 20 ANSWER 6.03 (3.00)
The following valves shut:
HCV 204 HCV 241 HCV 206-LCV 218-2 HCV 264 HCV 257 I
(0.2 ea)
The following valves opent i
[
HCV 268 (0.2 ea)
All 3 charging pumps star t (0.4)
Boric acid pumps A & B start (0.2 ea)
REFERENCE FCS LP 7-11-2, LO 1.4 000009A104 004000A202
...(KA'S) r QUESTION 6.04 (2.50) a.
List three (3) l oc a t i on s f r om wh i ch AFW p ump FW-6 may be started I
manually.
(0.75) b.
With the plant in hot shutdown and SG 1evel being maintained by FW-6 via normal feed regulator valves, describe the response of AFW 5
components to a prolonged loss of instrument air header pressure.
t include the impact on SG 1evel.
(Assume NO operator action.)
(1.?b)
(clarification made during examt normal = bypass) j k
p i
l l
6 A
L_.Pt MIT RYSIE11S_DESIE CDUIRDL_eUD_lNSIRI NntTellDN PAGE 21 ANSWER 6.04 (2.50) a.
1.
CB-10 (Control Room) 2.
Al-66A (Control Room) 3.
Local station at the pump 4.
4160 VAC switchgear bus 1A3 (pump motor breaker)
(any 3 at 0.25 ea) b.
1.
The recirc valves (FCV 1368 & 1369) fall open. (0.25) 2.
The AFW control valves (HCV 1107A,8 & 110BA,B) fall open. (0.5) 3.
The FW-10 Steam stop valves (YCV-1045 & 1045A/B) eventually open causing FW 10 to start and go on governor control. (0.$)
4.
SG level will increase until the SG is full (because the valves will not respond to the AFWAS reset signal at 60% level). (0.5)
REFERENCE FCS LP 7-11-1, LO 1.0, 1.8 061000A207 061000$G9
...(KA'S)
QUESTION 6.05 (2.50)
On the figure representing Al-41A, indicate the expected meter reading and light indications for a normal DC system line up and operation.
Use arrows to mark the meter scales and indicate light status as ON, OFF, BRIGHT, or D1M.
ANSWER 6.05 (2.50)
I SEE FIGURE v.!-41A Give full credi t f or vol tages indicsted between 130 - 140 VDC.
Give 1/2 ceridit for voltager indicated between 125 -130 VDC.
Charger current must inos cate between 170 - 200 amps.
Battery current should read 0 or slightly to the left of 0.
Both ground indication lights must be labelled as DIM Both the power available lights (Just below the ground lights) must be labelled as ON.
The "Nor ma l Source" light above the selector push button must be labelled as ON and the adjacent "Emer Source" labelled as OFF f
REFERENCE FCS LP ?-13-4, LO 1.4 063000A30.
...(KA'S)
($****
CATEGORY 06 CONTINUED ON NEXT PAGE
- )
6
.Elet1I_S'ISIEt1S_DESIGtl+_C0tlIROL _etJD_1tJSIRLt!EtJIGIlQtJ PAGE 22 OUESTION 6.06 (1.50)
A malfunction in the pressurizer pressure measurement "A"
channel (A/P1A-102Y) causes the outout to go to minimum.
c.
List two (2) RPS trip units which receive a signal from this channel and HOW each is affected by the malfunction.
(0.75) b.
Explain why the reactor will or will not trip (0.75)
ANSWER 6.06 (1.50) a.
High Pressure and TM/LP (0.25 ea)
The high pressure trip unit will not be affected (since minimum output equates to a low pressure condi tion). (0.25)
The TM/LP trip unit will go to the trip condition initiating a channel "A"
trip and a pre-trip alarm. (0.25) b.
A reactor trip will not occur (0.25) since the 2/4 logic is not satisfied by the single channel trip. (0.5)
REFERENCE FCS LP 7-12-25, LO 1.03, 007-12-26-H-25-1.03-0001 000027A216 012000A202
...(KA'S)
OUESTION 6.07 (2.00)
Pressurizer pressure is 2270 psig decreasing and pressurizer level is -3.0%
from program level.nnd decreasing.
Indicate the status of each i tem bel ow with OPEN, CLOSED, ON, OFF, MINIMUM, or MID () minimum but < maximum) as cppropriate.
Assume that the reactor is operating at 100% and the charging pump Standby Switch is selected to CH-1B-1C.
NOTE:
Other terms may be used to describe component status if desired.
a.
Spray valves b.
Proportional heaters c.
Backup heaters d.
Letdown flow contrel valves o.
Charging pumps B and C (answer f c.r each)
6 uBLaUI_S'ISIEtiS...DESIGtbC0t1I~01, uJD_itJSIRLt:!EUIaILOU PAGE 23 ANSWER 6.07 (2.00) a.
Open b.
Off c.
Off d.
Minimum e.
B - On C - Off (6 at 0.333 each)
REFEDENCE FCS LP 7-11-2, LO 4.1 010000K603 011000A102
...(KA'S)
OUESTION 6.08 (2.00)
The qualified safety parameter display system (OSPDS) calculates and displays three (3) saturation margin monitoring (SMM) values.
o.
List the three (3) saturation margins and the temperature sensors (inputs) used to calculate each.
If redundant sensors are available, state the criteria which determines the input used by the SMM (1.S) b.
What signal is used to provide the saturation temperature?
(0.5)
ANSWER 6.08 (2.00)
Upper head saturation margin - max of the top three UHJTCs CET saturation margin - representative CET b.
Pressurizer pressure (0.5)
"EFERENCE FCS LP 2-12-23, LO 1.7, OI-OSPDS-t-9, REV 0 194001A115
...(KA's)
OUESTION 6.09 (2.00) a.
What is the function of the Diverse Scram System (DSS)?
(0.75) b.
What plant parameter (include setpoint value and logic) will actuate the DSS?
(1.25)
6._mELANI_S'ISIEdS_QESIGN _CONIEQL+_6UD_1NSIBLL1EUIeIl0U PAGE 24 ANSWER 6.09 (2.00) a.
Generate a reactor / turbine trip signal independent of the RPS under ATWS (failure of RPS to generate a trip signal) conditions. (0.75) b.
Pressurizer pressure (0.5) at 2450 psia (0.25) on 2/4 detectors (0.25) tripping 1/2 channels (0.25).
REFERENCE FCS LP 7-12-25, LO 5.1, 5.2 000029A209 10CFRSO.62
...(KA'S)
QUESTION 6.10 (2.00) a.
At what p owe r ievel is the extended range operation of the Wide Range Log Power Channels enabled AND what operator action is necessary to permit extended range operation?
(0.5) b.
Hav does extended range operation differ from normal operation of a Wide Range Log Power Channel?
(0.5) c.
When audio monitoring of count rate is required, what two (2) switches must be positioned to sctivate the audio circult?
(0.5) d.
Aside from possible alarms, what is the control room indication of a non-functioning Campbell circuit for a particular Wide Range Log Power Channel as power is raised from 10E-6% to 10E+'%?
(0.5)
ANSWER 6.10 (2.00) a.
Extended range operation is enabled below 4x10E-7% power. (0.25)
The operator must take the extended range cutout switch to 0FF (0.25) (at about 10 cps control room indicatior).
b.
Bothe chambers provide signals that are summed to the channel in extended range with one detector's output disconnected when above approximately 10E-5% power.
(0.5) c.
The selector switch on the au d i o amp mu s t be set to * : desired channel (0.25) and the toggle switch for the select J ehannel must be in the proper position (0.25).
d.
The Indication from the affected channel will stop at about 10E-2%
l power. (0.5)
(The remaining channels will continue to show normal p owe r increase.)
. REFERENCE FCS LP 7-12-18, LO 1.3, 1.5, 1.7, 1.9 015000A202 015000K401
...(KA'S)
QUESTION 6.11 (3.00)
F; 11 in the numbered blocks on the simpilfied ESC logic diagram with the proper initiation and actuation signals.
ANSWER 6.11 (3.00) 1.
CRHS 11.
RAS 2.
SGLS 12.
SIA0 3.
CPHS 13.
DG Start 4.
PPLS 14.
CIAS 5.
STLS 15.
00 Brkr Prot 0.R.
I 6.
SGLL 16.
AFWAS 7.
SOLP 17.
VIAS 8.
SGDP 18.
Sequencers 9.
SGIS 19.
OPLS 10.
CSAS 20.
SEA (0.15 EACH)
REFERENCE FCS LP 7-12-14, LO 1.1 013000A301
...(KA'S) l i
i l
i i
i
(*****
END OF CATEGORY 06 *****)
l
Z _ EROCEDUBES..=_tiO2t2L aBNDEtML _Et:lERGEUCLetJQ PAGE 26 RaQ10LOOLCaL_C0tfIROL QUESTION 7.01 (2.50)
During a plant hettup from cold shutdG,n in accordance with O!-RC-3:
a.
What are the two (2) conditions, one of which must be met in order to preclude an RCS pressure surge prior to starting a Reactor Coolant Pump (PCP)?
(1.0) b.
What are the reasons for the upper and lower pressure Ilmits for RCS pressure maintained at 250 deg's RCS temperature?
(1.0) c.
What are the two (2) specific indications to the operator that a steam bubble has been formed in the pressurizer even though level indication is still at maximum?
(0.5)
ANSWER 7.01 (2.50) a.
1.
A pressurizer bubble exists.
2.
S/G secondary it less than 50 deg's above RCS temp. [0.5 ea.)
b.
Pressure is maintained so as not to lift PORV's OR to ensure pressure /
temperature limits are not exceeded (Iow temperature over-pressuriza-t;on) (0.5) and to be sufficient for RCP operation (NPSH) (0.5).
c.
Letdown f l ow > charging flow,
[0.251 and pressure stable. [0.25]
REFERENCE 01-RC-3, Pp. 2-7 n!-RC-9, P.
1 005000K104 010000G005 0100000004 011000A101
...(KA'S)
OUESTION 7.02 (2.00)
During power changes a.
What are the two (2) documents referenced by OP-4 that place limits on rate of power change and CEA movement?
(1.0) b.
What must be done if it becomes necessary to change boron I
concentration to reposition CEA's during a load increase?
(0.5) c.
What are two (2) conditions that may restrict the operators choice to insert CEA s?
(0.5) l
__ __. =,.
r
.1 PROCEDLIBES. - MNI : eablDhl EMERGENN.eblD PAGE 27 RADIDLDDIFAI PNTEDL t
i ANSWER 7.02 (2.00)
F a.
Technical Data Book (0.53 and Post Refueling Core Physics Testing tind Power Ascension Procedure (SP-PRCPT-1) or O!-RR-1 (any 2 at 0.5 ea) l b.
The load change must be stopped. (0.53 I
c.
PDIL or ASI limits (0.25 ea.)
i REFERENCE f
TDB, P.
!!!-22 OP-4, Pp. 1,2 E
0010000005
...(KA'S)
{
QUESTION 7.03 (2.50) a.
During refueling, what must be done after the lasertion of a fuel assembly into the core, bef ore another assembly is loaded.
A correct i
answer should include the reason for the action specified.
(1.0)
[
b.
As part of the initial conditions required for fuel movement, what is I
done admin i s tra t ivel y to ensure that core flow is not interrupted during fuel movement.
(0.75) e c.
Curing withdrawal of a fuel assembly from the core, it is noticed that an unexpected sustained increase in count rate is occurring.
What is the required initial action, AND what should be done if the required initial action is not practical?
(0.75) l I
ANSWER 7.03 (2.50) a A count rate should be taken and a 1/M plot done (0.53 to ensure the
{
next assembly loaded will not cause criticality. (0.53 b.
Controls for both LPSI pumps (0.253 and valves in ti.e flow path (0.253 arv (cautioned) tagged. (0.253 c.
Insert the fuel assembly (0.25) or emergency borate. f0.51 REFERENCE OP-11, Pp.
4, 7, 9 000036K103
...(KA'S)
(***** CATEGORY 07 CONTINUED ON WIXT PAGE *****)
2...E80CEQUEES _ 12DRt2LJBt10EtML+_EtdEEGENCY_eUD PAGE 28 bad 10LOGLCaL CatlIROL QUESTION 7.04 (1.00)
What is the MAXIMUM exposure an operator could receive in an area posted
'CAUTI ON-RADI ATI ON AREA" if he/she remained there for 30 minutes.
Explain your answer.
(1.0)
ANSWEb 7.04 (1.00) 50 Millirem. (0.51 By definition, a RADIATION AREA is an area where a person could receive > 5 meem/hr. but < 100 mrem /hr.
(0.5 Hrs.)
(100 mrem /hr. = 50 mrem (0.51 REFERENCE l
FCS E00 19400tK103
...(KA'S)
QUESTION 7.05 (2.00)
According to the Radiation Protection Manuals c.
Describe the administrative method used to assure that OPPD personnel do not exceed a quarterly whole body exposure limit.
State any appropriate values in the description.
(1.0) 6.
An OPPD employee is providing escort for a visitor with documented exposure history during a refueling outage.
What are the daily and weekly Permissible Accumulated Doses (PAD's) for the visitor and the employee?
(1.0)
2._ EBOCEDURES_=_NDEt2L._ABUDEtML._EldEEGEUC'4_auQ PAGE 29 BaQ10 LOG 1CeL_COUICOL ANSWER 7.05 (2.00) a.
When individuals reach quarterly exposures or 1.0 R or 2.4 R with a FORM-4, [0.51 a restriction is i nvoked al l owi ng an addi t i onal exposure of 100 mrem until the TLD is processed.[0.25)
When an additional 100 mrem is reached, entry is forbidden into any controlled area until exposure can be assessed.
[0.25) b.
Employee:
Daily:
300 Mrem Weekly: 900 Mrem Visitor:
Daily 300 Mrem Weekly 900 Mrem (0.25 ea.)
REFERENCE Radiation Protection Manual, P. VII-8-1-1 194001K103
...(KA'S)
GUESTION 7.06 (2.00)
According to AOP-19, Loss of 'butdown Cooling, what are the "LAS1 RESORT" methods (Flow Paths) for resioual heat removal for each of the following conditions.
For each case, describe the complete open/ closer loop path, c.
The reactor head installed.
(1.0) b.
The reactor head removed, transfer canal open.
(1.0)
I ANSWER 7.04 (2.00) a.
S!PWT ) HPS! Pumps into primary system ) PORV's OR, Loop Drains > RCDT
> RCDT drain pump back to SIRWT.
( 0.15 for each component in flow l
path and 0.05 for each manipulation to attain correct order.)
l b.
Align Fuel Transfer Canal Drain Pump to take suction f r om SI RWf an d discharge to Spent Fuel Pool. [0.5)
Align SFP pumps to transfer water from SFP to S!RWT.
[0.5)
REFERENCE AOP-19, P.
3 LP 2-1-41, L.O.
1.0, 1.2 000025K101 000025K301
...(KA'S)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE ++++*)
. Z.
EROCEDURES_= NO N L: ANOnt, Et1ERGEtdCY_AblQ "AGE 30 RAQ10 LOGICAL _CCEIROL QUESTION 7.07 (2.00) a.
According to AOP-14 (Loss of Coolant), the automatic response of the ECCS may be inadequate for long-term core cooling for large cold leg breaks.
Describe an alternate flow path for providing HPS! flow to the core under these conditions, as suggested by AOP-14.
(1.0) b.
In AOP-14 and throughout the E0P's the Operator i s CAUTI ONED t o promptly terminate containment spray as soon as termination criteria is met.
What two (2) component / equipment categories warrant this consideration and WHf?
(1.0)
ANSWER 7.07 (2.00) a.
HPS! pump through the cross connect (0.2) into the charging header,
[0.21 through the auxiliary spray ualve, [0.21 into the pressurizer and surge line, (0.21 and into the RCS Hot Leg. [0.23 b.
Containment spray may affect proper operation CO.4) of the 1.
RCP's, 2.
non-qualified equipment 3.
Instrumentation inside the containment.
(any 2 at 0.3 ea)
REFERENCE AOP-14, Pp. 7, 14 & E0P-03, P.
14 000011E312 026020A203
...(KA'S)
QUESTION 7.08 (1.50) a.
Describe the action (system manipulation) an operator would perform if a head bubble was suspected during natural circulation cooldown.
(0.75) b.
What action is then taken to confirm the existence of a bubble in the head?
(0.75)
l
. 2_EROCEDURES_. - hinn!. AggottRL,_Et2ERGEblCY eblQ PAGE 31 l
Ra010 LOGICAL _CCLIIROL L
ANSWER 7.08 (1.50) c.
1.
Verify letdown isolated l
2.
Stop depressurization and, if required, repressurize RCS to i
greater than or equal to 20deg.F subcooling
)
3.
Pressurize adn depressurize RCS within the limits of Fig. 4-2 4.
If necessary, operate PZR vent and/or reactor vessel head vent j
to clear trapped non-condensible gases.
(0.1875 ea) b.
Monitor PZR level and/or RVLMS (.5) for trending of RCS inuentory(.25)
[
i i
REFERENCE i
002000A403
...(KA'S)
.l I
l GUESTION 7.09 (1,50) i 1
OI-VA-2 gives t h e op e r a t or gu i de l i n e s on h ow t o 10ENTI FY and I SOLATE a l e ak f'
in the Auxiliary Building which causes a Stack High Radiation Alarm.
State these guidelines.
(1.5)
I t
ANSWER 7.09 (1.50)
Shut down supply and exhaust fans. [0.53
!solate all rooms, t0.51 Start one exhaust fan and open dampers one at a time while observing monitor j
until leak is found. [0.33 (Area monitors may be of some help.)
j REFERENCE FCS E0B
[
000060A202
...(KA'S)
)
I l
I i
l l
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
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f t
t i
Z._.EROCEDURES_=_UDEt::al._68UDEtML EMEEGEUCY_68Q PAGE 32 RattlGLCOLCeL_CDUIROL
'"J E ST I ON 7.10 (2.00) c.
During the performance of E0P-06, Loss of All Feedwater, the operator is cautioned, "Do not overreact t o l ow S/G 1 e v e l s, add f?edwater s l ow l y ".
What are two (2) REASONS for this caution?
(0.$)
b.
IF RCP's have been stopped, E0P-06 directs the operator to restart RCP's if six (6) criteria can be met.
State four (4) of these criteria.
L;alues tre not required.
(1.0) c.
I f Fe e dwa t e r to at least one S/G cannot be restored, what is the preferred cooling method (name only, functional description is not cequired), and which Functional Recovery Procedure es utilized.
(0.5)
ANSWER 7.10 (2.00) a.
1.
Prevent excessive cooldown rate.
2.
Prevent excessivc PZR level and pressure transient.
3.
Prevent overfilling S/O's.
4.
Prevent water hammer damage.
CAny 2, 0.25 ea.)
6.
1.
At ' east one S/G available for heat removal.
2.
S/G temp. < (50 deg's, above) ic.
3.
PZR steam space () 60%)
4.
CCW available 6,
PZR > (30%) and not decreasing.
(any 4, 0.25 ea.)
c.
Once-through cooling, (HR-4) CO.253 found in the RCS and Core Heat Removal FRP (EOP-20), t0.253 REFERENCE E0P-06, Pp. 5, 9,
10 000054L102 059000G010
...(KA'S)
OUESTION 7.11 (1.50) a.
What are two (2) actions required of the operator if there it an indication that four (4) CEA's have not fully inserted following actuation of the manual reactor trip pushbutton?
(0,5) b.
Per E0P-03 (LOCA), what are four (4) indirect indications or alarms checked by the operator to verify that PORV's and PZR : ode safeties i
are not open?
(1.0) l i
u l
2.. dROCEDURES_=_tJ08t%L,_aStJORt2L _EMERGEUC'4_atJD PAGE 33 RADIOLOGICAL _ cot 1IROL ANSWER 7.11 (1.50) a.
1.
Open CEDM clutch power supply breakers.
2.
Emergency borate.
(0.25 ea.)
b.
1.
Relief header temperature al arms.
2.
Relief header acoustic flow alarms.
3.
Quench Tank i
pressure il temperature ill level (any 4 at 0.25 ea',
REFERENCE E0P-01, P.
2 E0P-03, P.
6 000029K309 000029K310 000029K311 000029K312
...(KA'S)
OUESTION 7.12 (1.50) c.
The STEAM GENERATOR TUBE RUPTURE Procedure (EOP-04) cautions the operator early on to avoid use of HCV-1040 and FW-10.
State the reason for this caution.
(0.5) 6.
The operator is directed to determine which Steam Generator has tne tube rupture by monitoring or checking four (4) items.
What are these four (4) items?
(1.0)
ANSWER 7.12 (1.50) a.
Avoid unmonitored release of activity to the environment. (0.51 b.
1.
S/G activity 2.
Steam line radiation 3.
S/G Blowdown radiation 4
3/G tevels (0.25 ea.)
REFERENCE E0P-04, P.
5 000009A202 000037K305
...(KA'S)
(***** CATEGORY 07 CONTINUEl ON NEXT PAGE
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L l
. 2._.ERDCEDURES_=_UQ8t2L _aBUQ8tML _EMERGENC'i_6UQ PAGE 34 EaQ10LOGICe'._COtIIROL f
QUESTION 7.13 (2.00)
According to E0P-01:
a.
How is the operator direc ted to confirm the diagr.osio of an uncompilcated reactor trip?
(1.0) b.
If plant conditions indicate that MORE than an uncomplicated trip has occurred, what are the two (2) subsequent action paths *Jallable, one of which the operator must take?
(1.0)
ANSWER 7.13 (2.00) a.
Verify that Safety Function Status Check acceptance criteria is satisfied.
(1.0) b.
1.
(Attempt event diagnosis) and go to the appropriate EOP.
(0.5) 2, (If event is not diagnosed), then go to Functional Recovery Procedure (EOP-20).
(0.5)
REFERENCE E0P-01, Pp. 15-!?
000011A201 000011E312
...(KA'S)
OUESTION 7.14 (1.00)
What are the two (2) requirements to be met before the operator implements the LONG TERM ACTIONS of the Emergency Operating Procedures?
(1.0) l ANEWER 7 14 (1.09) i Appropriate operator actions for all success paths in use have been performed (0.53 and ALL safety function acceptance criteria are being
~
satisfied. (0.5 REFERENCE E0P-20, P.4 0000110011 000011G012
...(KA'S)
(*****
END OF CATEGORY 07
- )
_2 S _AAM I N I SIg&IIVE_2RDOEDijgES.,_ODblDlIlONS.,_eND_Lll:11I&IlDNS PAGE 35 t
QUESTION 8.01
(. 00) a.
How should mistakes in the official control room log be corrected?(.b) b.
Who is responsible for maintaining the official control room log?(0.5) l I
ANSWER 8.01 (1.00) a.
Mistakes are lined out and laitiated.
(0.5) b.
The Reactor Operator (assisted by the ARO).
(0.5)
OR f
The Licensed Senior Operator (LS0) or Licensed Operator (LO)
REFERENCE FCS Standing Order 0-24, p.
1-4 194001A106
...(KA'S) 00ESTION 8.02 (2.00) i The follmaing relate to Hold and Caution Tag requirements.
a.
Who will request the hanging and clearing of "Hold Order" tags?
(0.S)
]
b.
According to Standing Order 0-20, in what two (2) situations will "Temporary Clearance" tags be used?
(1.0) c.
Describe a general situation / condition in which the use of "Caution" tags is warranted.
(0.5)
. 8.. AQHibilSIBAIltlE EROCEQURES._CWQ1I1WS._WQ_ LIM 1IAIll3blS PAGE 36 l
t ANSWER 8.02 (2.00) i i
n.
System operator (system operations /906 personnel) (0.5) a b.
1.
When equipment is to be opei ated In a test posi tion or mode (such J
as trip checks, motor rotation checks. l imi torque-Opi..ivr
~
settings,_etc.)_(0,5)-
2.
When necessary to allow sys*vm/ equipment usage for unusual circumstances. (0.5) c.
1.
When special instructions in the operation of components must be exercised. (0.5)
OR 2.
During system turnover to specify components and systems under OPPD control. (0.5)
REFERENCE FCS LP 7-15-3 E0 1.1, 5.1, S.O. 0-20, rev 19, p 3.
l 19400.8K102
...(KA'S) i QUES 110N 8.03 (1.50) a.
For a procedure change that is required immediately (On-the-Spot),
I what two (2) signatures are required on the FC-68 (Setpoint/ Procedure Change)?
(1.0) b.
How many times can a specific Temporary (On-the-Spot) Procedure Change be used?
(0.5)
ANSWER 8.03 (1.50) a.
NPC Senior Licensed PRC member-Shift Supervisor PRC member (eny 2 at 0.5 ea) b.
Once (0.5)
REFERENCE FCS Standing Order G-30,rev. 19,pgs. 3,4 194001A101
...(KA'S)
(***** CATEGORY 08 CONTINUED ON NEX1 PAGE
- )
8._ d Dt:111J LCIEaI1U E_EEQC E D U E ES,_ C OU Q1Il0US,_au D _ Lit!1IGIlotJ S PAGE 37 OUESTION 8.04 (1.50)
During the performance of a valve lineup, it is discovered that a valve identification label (tag) i s mi ss i ng f r om a val ve.
c.
What corrective actTbWs should be taken?
(1.0) b.
What are the Shift Supervisor's administrative responsiblities?
(0.5)
ANSWER 8.04 (1.50) c.
A temporary tag should be filled out (0.5), the top half of the tag hung on the valve in question (0.25), the bottom half of the tag given to the Shift Supervisor (0.25).
b.
Log the valve in the Temporarily Tagged Equipment Log (0.25) and forward bottom half of the temporary tag to the Operations Engineer (0.25).
REFERENCE FCS Standing Order G-48,rev. 2,p 1.
194001K101
...(KA'S)
GUESTION 8.05 (1.50) a.
What members of the Plant Fire Brigade report i nme d i a t e l y to the fire area?
(0.75) b.
What is the minimum number of Fire Brigade members that must be onsite at all times and how long may there be less than the required minimum onsite?
(0.7S)
ANSWER 8.05 (1.50) a.
Fire Brigade Leader (LO*) (0.25)
Two other operators (equipment / aux operators) (0.25)
Two security force personnel (0.25) b.
5 (.25)
No more than two hours. (0.5)
REFERENCE FCS Standing Order G-28.rev 11.p 61 T.S.
5.2.2 194001K116
...(KA'S)
(*****
CATEGORY 08 CCt4TINUED Ct1 NEX1 PAGE *****)
B JDtditJ1SIBallt!E_ERO C EDU E ES._C QtJD1IlOtJS._etJ Q_LP11IGI1OU S PAGE 38 j
QUESrl0N 8.06 (1.50)
~
~~
a.
Which ~ shlft's Shift Supervisor is required to conduct the daily review of the Key Log for the Operations Key Depository?
(0.5) b.
In reviewing the Key Log Sheet, the Shift Supervisor discovers that a key from the Operations Key Depository has been lost. What action (s) must he take?
(1.0)
ANSWER 8.06 (1.50) a.
The afternoon Shift Supervisor. (0.5) b.
Notify the Supervisor - Operations (0.5) and the Supervisor Admi n i s tra t ive Servi ce s and Secur i ty. (0.5)
REFERENCE FCS Standing Order 0-26,rev 15,pgs 2,3.
194001K105
...(KA'S)
OUEST I ON 8.07 (2.00) a.
What are the two (2) purposes of the checklist items on the "Flame l
Cutting and Welding Permit" (Form FC-18)?
(0.75) b.
On:e a "Flame cutting and Welding Permit" is issued, how often must it be renewed and when does it expire?
(0.76) c.
Where are each of the two ' Flame Cutting and Welding Permit" copies maintained?
(0.5) l i
l
i
. S._ eDt111d1SIRAIlWE_EROCEDNFS, CCtdDlIlDtL%_etJD_ Lit 11IAIlit$
PAGE 3Y I
ANSWER 8.07 (2.00) a.
- 1. Prevent fires due to cutting and welding.(0.5) l
- 2. Provide instructions to assigned firewatches.(0.25) i b.
Each shift (0.25)
When the work described in the Maintenance Order or Design Change Order is complete (0.25), but not more than 2 weeks from the date Issued (0.25).
c.
One in the Control Room (.25), the other posted at the Job sight (.25).
ORFERENCE FCS Standing Order M-9,rev !!,pg 1.
REFERENCE FCS SO M-9, REV 11, pg 1 194001K116
...(KA'S)
QUESTION 8.08 (1.00)
What is the basis for the limit on steam generator coolant radioactivity?
ANSWER 8.08 (1.00)
Ensure that the resultant off-site doses will be well within the limits (of 10 CFR) (0.5) in the event of a steam Inne break (.25) and a concurretit loss of off-site power (.25).
REFERENCE FCS TS 2.20 035000SG6
...(KA'S) 00ESTION 8.09 (1.50)
Protection against low temperature overpressurization requires disabling of the HPSI pumps.
What three (3) conditions mitigate the requirement to disable the pumps with the plant in col d shu tdown?
B._ dDd1U1SI8eI12E_EROCEQUEES._CDUQ1I1DUS+_eUQ_ Lid 1IeIl0NS PAGE 40 ANSWER 8.09 (1.50)
Not.'equired ift 1.
a pressurizer safety removed 2.
a pressurizer PORV removed 3.
the reactor vessel head is removed 4.
no charging pump available to borate, tany three at 0.5 ea.)
REFERENCE TS 2.3 006050A201 006050K402
...(KA'S)
QUESTION 8.10 (1.50) o.
What is the purpose of the Safety Limits and what three (3) plant parameters are used to determine if a Safety Limit has been exceeded?
(1.0) b.
What action must be taken regarding facility operation in the event a Safety Limit is violated? (Include any applicable time limits).
(0.5)
ANSWER 8.10 (1.50) a.
Maintain the integrity of the fuel cladding.(0.25)
Pressurizer pressure ( 25), cold leg temperature (.25), reactor power level (.25) b.
Place the unit in hot shutdown (.25) within ene hour (.25).
I REFERENCE T.S.
1.1. T.S. 5.7.1 000027A204 002000SG5 002000SG6
...(KA'S)
GUEST I ON 8.11 (2.00) a.
What is the basis for the CEA Transient Insertion Limits? (Three (3) elements required for full credit.)
(1.0) b.
What two (2) options are available if the regulating CEA groups are inserted beyond the iransient Insertion Limits?
(1.0)
I
~w
. B ADMINISIRGIlEERDCEDURES,_CDtdDlHDtaS eblD t f MJTellDtAS PAGE 41 4
ANSWER 8.11 (2.00) a.
Ensure that 1.
acceptable power peaking factors are maintained 2.
the minimum available shutdown margin is maintained 3.
the potential consequences of a CEA ejec' ton accident are Ilmited 1
to acceptable levels.
(0.333 each) b.
Limits.(0.5)
- 2. Red' ice reactor power to the allowed level (0.25) which permits
{
coatinued operation above the Transient Insertion Limit using the l
existing CEA group position.(.25)
REFERENCE T.S.
2.10.2
,p 2-52, 2-50c.
f i
001000K504
...(KA'S) j t
4 i
QUESTION 8.12 (1.50) s i
a.
What is the Technical Specification limit for primary-to-secondary leakage?
(0.5) i b.
List two (2) of three (3) means of detecting primary-to-secondary
(
j leakage per Technical Specifications, one of which must be operable at l
4 l
all times.
(1.0) l
[
r I
1 i
ANSWER 8.12 (1.50)
(
a.
1 gpm total for both steam generators.(0.$)
b.
- 1. Steam Generator 81owdown Radiation Sample Instrument i
- 2. Condenser Off Gas Radiation Monitor
- 3. Periodic Secondary Sampl e s Anal yzed f or Ac t i v i ty f
l (any two at 0.5 each) 5 i
REFERB4CE j
TECH 11 CAL SPECIFICATION 2.1.4 l
l 0000375G3 0000375G4
...(KA'3) i 4
I l
l
(*****
CATEGORY 08 CONTINUED OJ NEX1 PAGE
- )
l t
I l
i
8 mitd1SIBAI1UE EROCE011RES._F d.I1CidS._auQ_ Lit:11IaIl0NS PAGE 42 I
i QUESTION 8.13 (1.50) a.
What type of Technical Specification instrument surveillance will include the adjustment of any alarm, interlock, or trip setpoints?
l (0.5) l b.
Why do the Technical Specifications state that at rated power the
[
minimum operable high power level channels is three (3) and what must be done immediately if the number of operable becomes two (2)?
Explain the reason for the required action.
(1.0) i i
?
L ANSWER 8.13 (1.50) i a.
Channel calibration (0.5) 4 b.
1.
To provide adequate power t il t de tec t ion. CO.53 1
j 2.
Reduce power to 70% CO.253 to prevent undetected flux tilt from i
exceeding design peaking factors. [0.253 OR l
Place 1 channel in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (0.253 to provide minimum j
degree of redundancy. [0.253 l
4 l
REFERENCE 3
FCS TS DEFINITIONS, 2.16 015000SG6
...(KA'S) i i
l QUESTION 8.14 (2.00)
During operations at full re ac tor power, the emergency f eedwater s tor age j
tank is declared inoperable and repairs are expected to be completed in about three days.
What are the Technical Specification requirements 4
concerning operation of the plant?
The Technical Specification for Steam l
and Feedwater Systems has been provided.
i l
l i
1 ANSWER 8.14 (2.00)
I The "motherhood" tech spec, 2.0.1(1) applies to this situation.
The plant j
must be in hot shutdown in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (1.0) and suberitical and < 300oF in the next six hours (1.0).
(There is no need to go to cold shutdown since the tech spec does not apply below 300cF.
If the candidate says that the plant i
must be in cold shutdown in an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> deduct 0.5 pt.)
l 1
1
8
.r.ADt:llbilSIBAI1UE_EROCEDURES,_CONQ1Il0uS,_eWQ_ Lit:11 TAIL 0NS PAGE 43 i
REFERENCE TS 2.0.1(1), 2.5 061000SG5
...(KA'S) i QUESTION 8.15 (3.00) 4 While the plant is operating at 100% reac tor power, your reac tor operator computes a total Steam Generator leak rate of 1.5 gpm.
At the same time you receive the radiochemistry report for the RCS which shows RCS activity to be 365 uCi/gm DOSE EQUIVALENT I-131.
Before you can decide on a course l
1 of action, multiple lightning strikes from a severe electrical storm hit in the swi tchyard causing a loss of o;f-si te power.
While verifying the cutomatic actions for a reactor trip an operator notices that Steam Generator Isolation has actuated but S/G pressure is continuing to decrease.
The other operator reports that pressurizer level, which had been recovering, is now dropping rapidly.
s a.
Using only the information available, identify each event which by itself requires declaration of an emergency classification.
(1.5) b.
Using EPIP-OSC-1, determine a single classification for your current condition and Justify your selection fully.
(1.5) 4 ANSWER 8.15 (3.00) j a.
1.
Fuel cl ad damage indicated by high RCS activity (>300uci/gm) ->
loss of fission product barrier. (alert)
{
2.
Loss of all offsite power for >15 min. (NOUE) 3.
Uncontrolled rapid secondary depressurization during hot shutdown. (NOUE) j (0.5 each) b.
A Site Area Emergency should be declared (0.75) on the basis of the
{
shift supervisor's application of NOTES 1.a-c. of EPIP-OSC-1.
Any one of the three guidelines or the barrier criteria may be used as Justification as long as the logic used supports the selection (0.75).
L Under the barrier criteria that is sufficient information to conclude i
that the RCS pressure boundary is being challanged or has failed leading to failure or challenge of two barriers.
(If the candidate uses incorrect logic to support his selection of guidelines deduct 0.5 pt.)
REFERENCE EP!P-OSC-1, TS 2.1.3, 2.1.4 i
194001A116
...(KA'S)
,i i
i i
U. S. NUCLEAR REGULATORY C0tti!SSION REACTOR OPERATOR LICENSE EXAMINATION FAC1LITY:
_EI._CALW01.Ll__________
REACTOR TYPE's
_E11Q:CE_______
DATE ADMINISTERED _884Q242A____ ________ _
l EXAMINER:
_t$CCROR4_S.___________
l CANDIDATE:
lblSIEUCI12JS_IQ_Cebl01DAIEA Use separate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question.
The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (4) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY
__L%LijE_ _IDIAL SCDRE__
_L'at 115 CAIEgnw
_25.00__ _25.00 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
.25.00__ _25.QQ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_25.00_
25.QQ
___ __ _ 3.
INSTRUMEffiS AND C0tO ROLS
_25.QQ,._
25.QQ
--____ _ _ _ 4.
PROCEDURES - NORt1AL, AENORMAL, EMERGENCY AND RAD 10 LOGICAL CCNTROL Totals LQQ.QQ__
Final Grade All work done on this examination i s my own.
I have neither given nor received aid.
Candidate's Signature
- 1._u2RiblDIELES_DE_bluN 28 EDWER_EteNI_DEERAIIDta.
PAGE 2
c I HF RMOBYbhtl114BEeI.IRaNEFER_eblD_ELUID_ELDW P
b QUESTION 1.01 (1.00)
During a reactor startup, an initial reactivity addition causes power to increase from 20 to 40 CPS.
A second reactivity addition causes power to change to 80 CPS.
Which of the following statements is correct?
a.
The first reactivity addition was smaller.
l b.
The second reactivity addition was smaller.
c.
The first and second reactivity additions were equal.
d.
There is insu#ficient data to determine a relationship between the l
reactivity ar di t i ons.
(1.0) i i
i ANSWER 1.01 (1.00) i b.
The second reactivity addition was smaller.
(1.0)
O i
l REFERENCE l
EGB GENERIC Fu1DAMENTALS FOR CE PWR l
FCS LP 7-5-3, LO 3.13.1 001010K536 015000K506
...(KA'S)
(
e I
r l
QUESTION 1.02 (1.00) l t
With the reactor critical at 5000 CPS, CEA withdrawal is used to increase l
i power to 10000 CPS.
Select the statement below that correctly describes l
the position of the CEA's after power is stabilized at 10000 CPS.
(1.0) 4 i
a.
The rod position will be higher than previous as more fuel must be i
l exposed to the neutron flux to maintain the higher power level.
[
i l
j b.
Rods will be f ur ther wi thdrawn to compent, ate for the temperature i
defect at the higher power.
[
d l
l c.
The rod position will be the same because the outward motion needed to attain a given startup rate equals the inward motion necessary to i
Kill the startup rate.
l l
d.
'. n e final position will be lower due to the increasing delayed neutron
{
population associated with the higher power level.
1 L
i i
[
I I
1s u2 Bib lCLELES DE blUCLEAB EDWEB 2LablI DEEBAILD%
PAGE 3
THFRMADYNAMICR, HFAT TRANRFFR AND Fi lff D Fi nu ANSWER 1.02 (1.00) c.
Rod position will be t h e s ame.
(1.0)
REFERENCE EQB GENERIC FLNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.4.3 001000K510
...(KA'S) i QUEST!3N 1.03 (1.00)
The moderator temperature coefficient (MTC) varies with certain plant conditions.
The MTCt [ choose one correct answer) l a.
Becomes more negative as boron concentration is increased.
b.
Varles due to temperature (ravg) because of the non-linear density j
changes of water as temperature changes.
c.
Causes axia' flux distribution to be tilted toward the top of the core at the beginning of Ilfe.
d.
Would be expected to introduce a large negative reactivity in the event of a major steam line break, i
l t
1 ANSWER 1.03 (1.00) 1
)
B (1.0)
REFERENCE EQB GENERIC FW4DAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.6.8 l
192004K103
...(KA'S) l 4
i I
i 2
j l
(*****
CATEGORY 01 CONTINUED ON NEXT PAGE * * * * * )
3
]
~. -
1._uBR1UC1ELES_DE_UUCLEGE_EDWER_ELANI_QEEEaIlOU, PAGE 4
IB EEMODn.Jed1CS,_B EaI_IEau SE ER_au D_ELU1D _ELOW QUESTION 1.04 (1.00)
In the event of a rod ejection accident, which one Hill be the first reactivity coefficient to insert negative reactivity?
a.
Doppler coefficient, b.
Pressure coefficient.
c.
Void coefficient.
d.
Moderator temperature coefficient.
ANSWER 1.04 (1.00)
A (1.0)
REFERENCE E00 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-3, LO 3.7.5 192004K107
...(KA'S)
OUESTION 1.05 (1.00)
It takes less reactivity to go promp t cr i t ical att a.
BOL because of the higher value of beta effective.
b.
BOL because of the l ower val ue of beta effective.
c.
EOL because of the higher value of beta effective.
d.
EOL because of the l ower value of beta effective.
ANSWER 1.05 (1.00)
D (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-5-4, LO 4.5 192003K106 192003K107 192003K108
...(KA'S) f
(***** CATFGORY 01 CONTINUED ON NEX7 PAGE
+++++)
1 _u2RibiCIELES_QE_tduCLEAR_EQWER_EletdI_QEEBA110td.
PAGE 5
IndEatdQQ'Itdetf1CS _WEAI IBeblSEER_etdQ_ELU1D_ELOW i.
I QUESTION 1.04 (1.00)
Which one of the following statements is correct concerning the change l
In differential boron worth (% delta K/K) with RCS boron concentration L
(range of 0 to 1600 ppm) and Tave (range of $32 degF to 580 degF)?
a.
It increases as Tave and RCS boron concentration increase.
i b.
It decreases as RCS boron concentration increases but is constant as Tave increases.
c.
It decreases as Tave and RCS boron concentration increase, d.
It increases as Tave increases but is constant as RCS boron concentration increases.
t i
I ANSb'ER 1.06 (1.00)
C (1.0)
REFERENCE EQB GENERIC Fu1DAMENTALS FOR CE PWR j
FCS LP 7-5-3. LO 3.5.4 192004K109 192004K110
...(KA'S) l
[
QUESTION 1.07 (1.00) l 1
Volding has occurred in the RCS, in the vicinity of the reactor vessel l
{
during a natural c ircul a t ion cool down.
Which of the f ol l owi.ig CORRECTLY l
c,haracterizes the process of collapsing the vold?
t i
a.
The vold will superheat if an attempt is made to collapse it too fao.
The rate of collapse will be governed largely by ambient heat losn
]
from the vold.
i A
b.
The void will collapse at a rate eruivalent to the rate of HPSI f l ow;
}
therefore, full HPS! should be run until the void is fully collapsed.
I c.
The void will be composed l argely of hydrogen gas, and will therefore l
l require degasifying of the RCS in order to begin collapsing it.
l i
i 6
d.
The void will collapse inmediately upon increasing the pressure above i
j the local saturation pressure; the main concern is water hammer, j
l 4
l s
1 i
I I
l
1...2PRINCIPLf1.0F Ml} CLEAR P NER PLMil.QPERATION, PAGE
. 4 o
IMF DMnwhiet11CS.,_HEAI_ISeblSEER_ AblD_ELUID_ELDlJ
- se
[.
ANSWER 1.07 (1.00)
A (1.0)
REFERENCE EQB GENERIC FUNDAMEffTALS FOR CE PWR St!CY 82-475 193007K104
...(KA'S)
I QUESTION 1.08 (1.00)
The ratio of Pu-239 and Pu-240 atoms to U-235 atoms changes over core life, Which one of the pales of parameters belcw is most affected by this f
change?
I a.
doppler coefficient and beta b.
moderator temperature coefficient and doppler coefficient T
c.
beta and moderator temperature coefficient l
1 l
l i
d.
moderator temperature coefficient and neutron generation time r
i i
l t
l ANSWER 1.08 (1.00) 4 f
l A (1.0)
{
i t
l REFERENCE t
EQB OENERIC FUNDAMENTALS FOR CE PWR i
(
3 192004K107
...(KA'S) p i
i h
c 1
i I
I i
I j
(*****
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t 1
t 1
i I
i
. _ _ _ _ _ _._. _ _ _,____,_ - m __
m,__
1
. LJRINC1ELES_QE N!.lCLEAE_EQWER_ELANI_QEEBeIl0N.
PAGE 7
IWEtt100'Itht:11CS,_WEAI_IRANSEER_eWD ELU10_ELOW l
t QUESTION 1.09 (1.00)
A general rule is often stated "doubling the count rate halves the margin to criticality".
This is mathematically stated by the equation:
CR1/CR2 = (1-Keff2)/(1-Keff1).
Which one of the f oll owing sta temen ts is CORRECT concerning the above statement and equation?
l a.
Equal changes in Keff result in equal changes in subcritical i
multiplication level, b.
Both Kefft and Keff2 have to be less than 1.0.
]
c.
The equation only appioximates the instantaneous change in count ratel 7
j once the equilibrium value is reached, the count rate will be higher.
d.
A second doubling of the count rate will result in the reactor l
4 becoming critical or supercritical.
k i
j i
ANSWER 1.09 (1.00) l i
j B (1.0)
REFERENCE 4
E0B GENERIC FLNDAMENTALS FOR CE PWR v
001010K516 192003K101
...(KA'S) l i
l I
(
i i
i I
i r
I l
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
i l
l i
i
d
..L._aEaIMcIELES, DE_NUCL M ErLIFEr ELAN 1_DEERenW PAGE 8
T Hr aMnnXNeti1CS.,_MEaI_IRANSEER_ euD_ELinD_ ELDW 4
4 QUESTION 1.10 (1.00) 1 Which one of the following correctly describes the response of core del ta T if the RCS coolant flow rate is reduced while the steam system is controlling to maintain 80% electrical output?
The plant has been at 80%
l 1
of full power, steady state, for 10 days.
All control rods are fully withdrawn (ARO).
]
a.
Delta T will increase but nuclear power remains the same to provide the same megawat.t thermal output.
t b.
Delta T will remain the same but nuclear power will increase to provide the same megawatt thermal output.
c.
Delta T will remain the same since the turbine /generktor output power j
has not changed.
l 4
i i
d.
Delta T will increase and nuclear power must increase to provide more steam flow to the turbine.
}
[
i ANSWER 1.10 (1.00) k i
i A (1.0) i l
REFERENCE I
EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-7-22, LO 7.2-2 l
1 193007K108
...(KA'S)
(
I l
QUESTION 1.11 (1.00) l t
l The reactor is critical at 10E-3 % power when a S/G Atmospheric dump valve j
falls onen.
Assuming BOL conditions, no rod motion, and no reactor trip, g
choose the answer belxe that best describes the values of lavg and nuclear i
power for the resulting new steady state.
(POAH = point of adding 1
heat).
l a.
Final Tavg greater than ini t i al Tavg, Final power above PDAH.
j b.
Final Tavg greater than initial Tavg, Final p owe r a t PGAH.
l I
I c.
Final Tavg less than initial Tavg, Final p owe r above POAH.
i d.
Final Tavg less than initial Tavg, Final power at POAH.
l f
l I
i l
i
- 1.. PR1blCIELES DE_blijCLEAR PIIslER_2LMF DRERATIIll.
PAGE 9
I MFEMODN f P R ; _MEAI_I RAN A FER.eblD_ELljlD_ELDW t
i ANSWER 1.11 (1,00) i C (1.0)
QEFERENCE i
EQB GENERIC FUNDAMENTALS FOR CE PWR l
FCS LP 7-5-4, LO 1.3 192008115
...(KA'S) t i
i OUEST!ON 1.12 (1.00)
Which one of the following best supports the reason why Xenon reactivity increases sharply af ter a trip following 1000 hrs. of operation at 100%
e power?
l a
t 4
a.
!odine half-Ilfe is much shorter than Xenon half-life.
L b.
Xenon decays less rapidly due to a reduction in the neutron flux.
I c.
Iodine production is greatly reduced and Xenon production is greatly L
increased due to the reduction in neutron flux.
3 j
d.
Due to reduced neutron absorption, lodine concentration increases, and j
j Xenon decays direc tly from lodine, thus Xenon increases.
I i
I ANSWER 1.12 (1.00) l A
(1.0) l REFERENCE l
l E0B GENERIC FUNDAMENTALS FOR CE PWR l
192006K107
...(KA'S) r f
j l
i i
1 t
(+***+
CATEGORY 01 CCt4TINUED ON NEXT PAGE * * + + + )
1 i
l 1
. 1._.ER1NCIELES_DE_UUCLEAR_EDWER_ELauI_0EEEaIl0U.
PAGE 10 IBEEMODYNM11CS,_BEaI_IEauSEER_auD_ELU1D_ELOW QUESTION 1.13 (1.00)
Select the most correct statement from the f ol l ow i n g.
a.
If two cen tr i f ugal pumps are in parallel, the combined pump head will be approximately the sum of the individual pump h& ads, b.
If two centrifugal pumps are in series, the combined power requirements will be approxirately equal to the cube of the individual pump p ower.
c.
If two centrifugal pumps are in series, the flow from each 3 imp will be approximately equal to the square of the individual pump sieed.
d.
If two centrifugal pumps are in parallel, t h e c omb i n e d f l ow w i l l be approximattiy equal to the sum of the individual cump flows.
ANSWER 1.13 (1.00)
D (1.0)
REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS 7-7-21, LO 6.3-5 191004K109 191004K110
...(KA'*,)
QUESTION 1.14 (1.00)
Which one of the f ol l owi ng wi l l cause plant efficiency to increase?
a.
Total S/G bl owdown is changed from 30 gpm to 40 gpm.
b.
Level increase to higher than normal in a f eedwater heater.
c.
Steam quality c h a n g e s f r on 99. 74 t o 99. 9%.
d.
Absolute condenser PRESSURE changes from 1.0 psi to 1.5 psi.
ANSWER 1.14 (1.00)
C (1.0)
l u fRitdC1ELES OF NifftFAD _ PCLIER EleMT GPERATIN.,
PAGE 11 THERMODYNeMICS,_. MEAT..TRANAFFR AND F1 ill D F1 OLI REFERENCE EQB GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-11-20, LO 3.0 002000K501
... (ItA' S)
QUESTION 1.13 (1.00) i Which one of the below requires the most heat removal during condensation?
(assume steam is of equal quality) a.
two pounds of steam at 1200 psia, b.
two pounds of steam at 600 psia.
l l
c.
one pound of steam at 300 psia.
d.
one pound of steam at 0 psia, i
i t
I ANSWER 1.15 (1.00) t i
G (1.0) t 1
REFERENCE EOB GENERIC FUNDAMENTALS FOR CE PWR 5
j steam tables
(
FCS LP 7-7-16, LO 1.6-1 193003K125
...(KA'S)
I 4
j-QUESTION 1.16 (1.00)
I l
i As core age increases during a cycle, the integral rod worth at 1 0 0'/. f u l l pcwer of a regulating CEA located near the center of the cores j
i a.
Increases due to the fact that temperature of the CEA is higher, b.
increases due to the fact that the CEA is operating in a region of higher neutron flux.
j c.
decreases due to the fact that the temperature of the CEA is higher.
d.
decreases due to the fact that the boric-acid concentration in the f
core is less.
l I
I e
L-
1._t2SiblCigt rs or unri rAn EDWER_RLANI_DEFRATI%
PAGE 12 IWFeMnBYhikifIS.,_WEAI_IEeNRFER_ablD_ELLilD_ELOW 1
?
ANSWER 1.16 (1.00)
B (1.0) i REFERENCE i
EQB GENERIC FUNDAMENTALS FOR CE PWR l
192005K107
...(KA'S>
i QUESTION 1.17 (1.00) i j
The term "critical heat flux" refers to that heat flux between a fuel pin and the bulk of the primary coolant
{
a.
at which the heat-transfer mechanism changes between natural
[
convection and nucleate boiling, i
l i
b.
at which DNBR would equal 1.3.
c.
at which the heat-transfer mechanism changes between nucleate boiling and partial fiIm bol1ing.
I' d.
which is 77% of that which would cause clad failure.
[
i l
l
[
ANSWER 1.17 (1.00)
C (1.0)
REFERENCE E00 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-7-18, LO 3.3-2 193008K106
...(KA'S) l l
a
(***** CATEGORY 01 CONTINUED ON NEXT PAGE ++***)
i
?
3 l
. L._uRRitdC1ELES_DE_bluCLEAR_EDWER_EletdI_QEER&Il0bl PAGE 13 IBEat100Y.bletd1CS._BEAI_IRetJSEER_etdD_ELUID_ELOW 4
QUESTION 1.18 (1.00)
A relief valve on a pipe opens at 885 psig. The temperature of the steam exhausted to the atmosphere is 320 deg. F. The fluid within the pipe is saturated steam. Using the steam tables provided, what is the approximate temperature of the fluid within the pipe, in the immediate vicinity of the relief valve?
(Selec t one response f rom bel ow.)
a.
530 deg. F t
b.
400 deg. F i
c.
320 deg. F l
L d.
212 deg. F I
i ANSWER 1.18 (1.00) i A
(Saturated steam at 885 + 15 psia.) (1.0) i l
REFERENCE
[
j E08 GENERIC FUNDAMENTALS FOR CE PWR l
3 CE Steam Tables j
193004K115
...(KA'S) i i
OUESTION 1.19 (1.00) i j
Which of the following statements concerning power def ect is correct?
E i
a.
The power defect is the difference between the measured power I
coef f i c i ent and the predic ted power coef f i c i en t.
l i
b.
The power de f ec t increases the rod worth requirements necessary to j
j maintain the desired shutdown margin following a reactor trip, i
I c.
Because of higher boron concentration, the power defect is more i
L j
negative at the beginning of core life.
i d.
The power defect necessitates the use of a ramped Tavg program to f
l maintain an adequate subcooling margin, f
I l
l i
f i
i
.L_.ER1NC12LES_DE_N1101 FAR EDWER_ELetE_DEERaT IDti.
PAGE 14
.IMEat:10DXbkht:11DS.,_ MEAT TReuSEER_euD_ELUID_ELOW t
i ANSWER 1.19 (1.00)
B (1.0)
~ REFERENCE i
EQB OENERIC FUNDAMENTALS FOR CE PWR FCS 7-5-3, LO 3.11.3 001000K549
...(KA'S) t h
QUESTION 1.20 (1.00) f With RCS pressure at 2100 psia, one of the_ seals in a RCP assembly falls, Assume original leakoff flow was 1.0 gpm and leakoff backpressue is 50 psia.
Which ONE of the following is correct concerning both the pressure drop across each operable seal and leakoff flow?
i a.
1050 psid and 1.5 gpm i
b.
1050 psid and 1.2 gpm c.
1025 psid and 1.2 opm d.
1025 psid and 1.5 gpm l
l I
ANSWER t.20 (1.00)
C 1025 psid and 1.2 gpm (1.0)
REFERENCE EQB GENERI(. FUNDAMENTALS FOR CE PWR FCS LP 7-11-20, LO 1.7d 0030004201
...(KA'S)
[
t f
I I
l
[
[
(***** CATEGORY 01 CONTINUED Ct1 NEXT PAGE *****)
I
{
t
1._tER1NC1ELES_QE_UUCLEaR_20WER_EL6UI_DEEEeIl0U.
PAGE 15 IhiEEUOQYuad1CS._HEaI_IEauSEER_auD_ELU1D_ELOW QUESTION 1.21 (1.00)
Which of the f oll owi ng st atemen ts abou t burnable poisons is NOT true?
c.
As core age increases burnable poison effects partially compensate f or fission product buildup effects.
b.
Including burnable poison in the fuel affects the moderator temperature coefficient.
c.
As core age increases burnable poison effects partially compensate for l
depletion of the fuel due to fission.
d.
Including burnable poison in the fuel affects the doppler coefficient.
l l
ANSWER 1.21 (1.00)
{
D (1.0) l REFERENCE l
E0B GENERIC FUNDA'iENTALS FOR CE PWR
}
FCS LP 7-5-3, LO 3.3.2 192007K101
...(KA'S)
OUESTION 1.22 (1.00)
The reactor t r i p s f r om f u l l power, equilibrium xenon conditions.
Six (6) hours later the reactor is brought critical at 5x10E-4%.
If power level is maintained at 5x10E-4%, which of the following statements is CORRECT concerning control rod motion?
c.
Rods will have to be withdrawn since xenon will closely f ol l ow its normal build-in rate.
b.
Rods will approximately remain as is as the xenon establishes its equilibrium value for t h i s p owe r level.
c.
Rods will have to be inserted since the critical reactor will burnout
- xenon, d.
Rods will approximately remain as is since the xenon concentration is independent of time.
' l 2RlUCl2LES DE UUDLEAB 2DUEB 2 Lath DEEEAZlDth PA6E 16 IH M inD.YNaMICS., HFAT 186USEEB AND ELUlD ELDU ANSWER 1.22 (1.00)
A (1.0)
CEFERENCE E08 GENER!C FUNDAMENTALS FOR CE PWR FCS LP 7-5-5, LO 5.3.11 192006K114
...(KA'S)
QUESTION 1.23 (1.50)
For each of the f ollowing euents, describe HOW and WHY RCP ' A' motor emperage will vary f r om its initial magnitude prior to the euent.
c.
It is running with two (2) other RCPs and the fourth RCP is started.
b.
I ts pump becomes mechanically uncoupled from the motor.
c.
Bearing degradation in the pump results in partial binding.
l 4
ANSWER 1.23 (1.50) 6.
The amps will decrease (0.2) because the fl ow through the pump decreases due to the increased backpressue f r om s t ar t i ng the fourth pump (0.3) b.
Amps will decrease (0.2) because the motor no longer has a load (0.3).
l c.
Amps will increase (0.2) because friction is increasing the load on i
the motor (0.3).
REFERENCE E08 GENERIC FUNDAMENTALS FOR CE PWR FCS LP 7-7-21 LO 6.1.3 191004K107
...(KA'S) f I
(***** CATEGOR 01 CODINUED ON NEXT PAGE
- )
l l
l i
. A - E31NCIELES_0E_uuGLEAR_20WER_ELANI_DEEGIlotJ.
PAGE 17 IHERtdQQY.det$1CS. beal IRGUSEER_etJD F.Lu1D_ELOW 1
QUESTION 1.24 (1.00)
Assume the plant is operating at 100% power, with all systems normal, and the backup prescurizer heaters are energized.
a.
How and why will reessurizer pressure change?
(0.5) b.
How and why will pressurizer level change?
(0.5) i
{
ANSWER 1.24 (1.00) a.
Pressure will increase (0.2) due to the water phase change (0.3).
b.
There will be no significant level change (0.5).
OR 4
Level will decrease (0.2) due to the removal of siquid mass from the pressurizer and the reduced volume of the remaining mass at higher i
j pressure (0.3) t REFERENCE 1
EGB GENERIC FLtJDAMENTALS FOR CE FtJR FCS 01-RC-7, p.
43 l
010000A105
...(KA'S)
I I
QUESTION 1.25
(.50)
As Keff approaches unity, for the same increase in Keff, a greater length l
of time is regulPed to reach the new equilibrium level.
This is a charac-
[
teristic oft a,
Moderator effects on neutron leakage.
l b.
Increasing population of delayed neutrons.
i c.
Reactivity change on power increases.
I d.
Suberitical multiplication.
i i
l
}
I i
i i
1.a ERluCIELES DE,U1JCLEAR EDWEE ELeUI DEF. RAT _1 m :
PAGE 18 IHFRHnDElat11CS BEAT TRAUSEER td.JD ELDlD ELDW (44SWER 1.25
(.50)
D (0.5)
REFERENCE EGB GENERIC tut 4DAMEt4TALS FOR CE PWR FCS LP 7-5-4, LO 5.1 192003K101
...(KA'S) l
(*****
D4D OF CATEGORY 01
- )
2-ELaul_ DES 10d_13CLUD1UQ_SeEEIY_tsuD_EMEtlGEUC%.S%SIEUS PAGE 19 1
QUESTION 2.01 (2.50)
Te r.h n i c al Speci+1 cation 2.2 on the Chemical and Volume Control System requires that th&re be at least one flow path to the core for boric acid I
injection whenever fuel is in the reactor.
Describe two of the three flow paths which meet this requirement per Tech Specs.
(2.5) i i
ANSWER 2.01 (2.50) 1.
The boric acid pumps (0.5) can deliver boric acid tank (0.25) contents to the charging pumps (0.5).
I 2.
The safety injection pumps (0.5) can take suction from the S!RW (0.25) and inject into the core (0.5).
I f
3.
The charging pumps (0.5) can take suction by gravity from the
[
S!RW tank (0.75).
1 4.
The charging pumps (0.5) can t&Ke suction by gravity f r om t h e f
)
boric acid tanks (0.75).
l (Any 2/4 at point values shown)
I i
REFERENCE i
2-10, 19 L
000024SG8
...(KA'S) i 1
QUEST 10t1 2.02 (1.50)
Answer the f oll owi ng ques t i ons re l a t i ng to starting a main feed pump.
1
)
Assume this is the first pump, started cold.
f n.
How many condensate pumps must be operating?
(0.5) l b.
How many consecutive starts may be atttmpted?
(0.25) i i
q c.
How should the temperature of the seal water lir.es change when the f
feed pump goe s f r om s t andby to normal operation?
(0.25) 7 t
d.
What should be the status of the warmup valves bef ore starting the l
feed pump?
(0.b)
(
I 4
4 i
r 1
f
?
f l
2....E LAWI.Q ES10U.1U CLU Q1NG.SA E EILeU D.EU E R G EU CLSY SIEd $
PAGE 20 ANSWER 2.02 (1.50) a.
1 (0.5) b.
2 (0.25) c.
Temperature difference ( i r,-ou t ) should disappear. (0.25) d.
Isolated. (0.5) i QEFEREN*E FCS t.P 7-11-11, p. 47-53 QEFERENCE 059000S013
...(KA'S)
OUESTION 2.03 (3.00)
Using figure 2-1, show and label all major penetrations into the Reactor i
Cool ant System.
DO NOT include instrument or test penetrations.
(Eighteen properly placed and labelled penetrations are required for full credit.)
ANSWER 2.03 (3.00) 1
/---RCP D---\\
/---RCP b---\\
/
\\(---S!
CHG--->/
\\
/
\\
/
\\-->RCDT RCDT(--/
N VENT SI-->/
\\ drain drain
/
\\
/?\\
/
\\
/
\\
?
/
SURGE N
{
/
N----!--------/
N
?
?
I SG 2B(-----!--------!-------!
REACf0R
!-----!----!----!------>50 2A 1
?
\\
\\t/
\\!/
/-------------\\
\\t/
N!/
/
i'
\\
/(--S!
\\
/
\\
/
\\ drain
/
i drain
\\
/---> SPRAY N--->Si
/
RCDT(-?---N
/
\\
/-->RLDT l
N
/(---CHG SPRAY--->\\
/
dr ai n l
LD<---!
N---RCP C---/
\\---RCP A---/
l (E!GHTEEN PENETPATICt4S AT 0.166 EA)
REFERENCE i
002000K10?
...(KA'S) 1 I
i
(***** CATEGORY 02 CONTINUED ON NEXT FAGE
- )
l l
- 2._.r.ELatJI_QESIGU.1UCLUD1UG_SAEEIY_eUD.D:!ERGEUCY_SYSIEUS PAGE 21 1
1
(
QUESTION 2.04 (1.75) a.
Following a RPS trip initiated by automatic RPS trip modules, WHA 1 conditions must be satisfied to allow the RPS "M" coil to be i
re-energized and HOW is it re-energized?
(1.0) b.
How is the RPS "M"
coil re-energized folicaing a manual trip from Al-33?
(0.75) i l
l ANSWER 2.04 (1.751 l
a.
Once the initiating trip has cleared (0.1), the operator must wait for l
KTD1 and KTD2 relays to time out. (0.4)
The
'M" coil is then re-energized by pressing the trip reset pushbutton. (0.5) b.
Manually close the trip breakers (0.5), press the trip reset l
pushbutton (0.25) 1
\\
l REFERENCE I
FCS LP 7-12-25 EO 1.16 001000K603
...(KA'S) l l
l l
QUESTION 2.05 (1.25) l a.
What provides backup cooling capability to bring the plant to a safe I
shu tdown condi t i on if Component Cooling Water is not operable?
(0.5)
I b.
What is the potential public hazard associated with the use of this CCW backup system sven if there is no loss of RCS 6ntegrity?
(0.75)
L l
i i
l L
ANSWER 2.05 (1.25) a.
Raw Water (0.5) (supplied to v ar i ou s CCW cool ed c omponen t s i
individually)
{
b.
Failure to properly isolate the CCW system prior to initiating ku i
would release toxic chemicals into the river. (0.75)
L REFERENCE l
FCS LP 7-11-6 EO 5.3 0000?dA103
... < KA ' S.'
t I
l
(*++++ CATEGORY 02 CONTINUED tJ NEXT PAGE *****)
l l
\\
t
- 2 u 2LauI DESIGu_1UCLUD1UO.SAEEIY auD EUERGEUCY_SYSICUS PAGE 22 1
QUESTION 2.06 (1.25) i c.
HOW is the operation of the RM61 affected by placing it in the "Inusrsion" mode?
(0.S) b.
WHEN and WHY is the ' Inversion" posi tlen selected on the RM61?
(0.75) 449JER 2.06 (1.25) a.
Placing the switch in INVERSION increases both the alert and alarm setpoints by a fixed amount. (0.5) t k
b.
It is used when a radon inversion (temperature inversion) has been I
confirmed (0.5) since the radon will add to the background level j
increasing the possibility of erroneous alarms and component activation. (0.25)
REFERENCE FCS LP 7el2-3 EO 6.2 073000K101
...(KA'S) l l
QUESTION 2.07 (2.00) a.
What two (2) conditions will generate a 'Cor.t a inmen t Ven t F an Of f i
Normal"?
(0.5) i b.
What is the failed position for the face dampers and bypass dampers l
on the containment cooling and filtering units?
(0,5) i c.
When the handswitches for the dampers in part b are placed in the j
UNFILTERED position, what provides the motive force to position the dampers?
(0.b)
(
i d.
Why is it impor t ant to maintain charcoal bed temperatures below 360 deg.F even after containment air temperatures are returned to normal following a large break LOCA?
(0.5) l t
I l
f l
l
l L r.fLAf1T_DESIGtt.ltJCLUDitA0_SeEEIY_eUD_EUERDEllCL.SYS1EliS PAGE 23 ANSWER 2.07 (2.00) a.
control switch in PULL-0UT 69 permissive switch not in AFTER CLOSE position fan breaker not racked in (any 2 at 0.25 ea) l l
b.
face dampers fall open (0.25) bypass dampers fall shut (0.25) i l
c.
Control air (instrument air) (operates an air-p
, to position the dampers for unfiltered operation.) (0.5) d.
Prevent the release of contaminants absorbed by the char coal. (0.5)
QEFERENCE
}
FCS LP 7-14-2, LO 1.2, 1.6, 1.0 022000SG10 022000SG7 103000A40)
...(KA'S)
OUESTION 2.08 (2.50) a.
List three (3) l oc a t i on s f r evn wh i ch ARJ pump FW-6 may be started manually.
(0./5) b.
W6th the plant in hot shutdeun and 50 level being maintained by FW-4 via normal f eed regul ator valves, describe the response of AFW c omp on e n t s to a prolonged 'oss of instrument air header pressure.
Inclade the impact on SG 1evel.
(Assume NO operator act6cn.)
(.
75)
(clarification made during examt normalabypass)
2._ ELauI_DESIGu_W CLUQ WQ_SAEEIY AUD_EUERGEUCY_ SYSTEMS PAGE 24 l
ANSWER 2.08 (2.50) a.
1.
CB-10 (Control Room) 2.
Al-66A (Control Room) i j
3.
Local station at the pump 4.
4160 VAC switchgear bus 1A3 (pump motor breaker) i (any 3 at 0.25 ea) b.
1.
The recire valves (FCV 1368 & 1369) fail open. (0.25) 2.
The AFW control valves (HCV !!07A,B & !!0BA,B) fail open. (0.5) l 3.
The FW-10 Steam stop valves (YCV-1045 & 1045A/0) eventually open causing FW-10 to start and go on governor control. (0.5) 4.
SG level will increase untti the SG is full (because the valves will not respond to the AFWAS rese t signal at 60% level). (0.5)
REFERENCE FCS LP 7-!!-1, LO 1.0, 1.8 061000A207 061000SG9
...(KA'S)
OUEST10t1 2.09 (2.50)
On the figure representing Al-41A, indicate the expected meter reading and light indications for a normal DC system line up and operation.
Use ar r ows to mark the meter scales and indicate light status as ON, OFF, BRIGHT, or DIM.
ANSWER 2.09 (2.50)
SEE FIGURE Al-41A Cive full credit for voltages indicated between 130 - 140 VDC.
Olve 1/2 credit for voltages indicated between 125 -130 VDC.
Charger current must indicate between 130 - 230 amps.
6 Battery current should read 0 or singhtly to the left of 0.
Both ground indication lights must be labelled as DIM Both the power available lights (just below the ground lights) must be labelled as Ctd.
The ' Normal Source' light above the selector push button must be labelled as ON and the adjacent 'Emer Source' labelled as OFF QEFERENCE FCS LP 7-13-4, LO 1.4 063000A301
...(KA's)
(***** CATEGORY 0 2 CONTINUED 04 NEXT PAGE * * * * *)
1
- 2._1.2Ltd1_ DES 10tLECLUDWQ_SeEEIX_auD_EMERGEUCY_SYSIEUS PA0E 25 OVEST!Q4 2.10 (2.25)
Fol l owi ng a small break LOCA with S! initiation (PPLS and LPHS), autcmatif Recirculation Actuation occurs.
The operator is directed to initiate additional cooling via the shutdown heat exchangers ner E0P-03 inmediately f ol l owing the RAS.
Assume the f ol l owi ng 1.
No manual operation of any SI component has occurred since Si actuation and the subsequent Recirculation actuation.
2.
RCS pressure is 750 psia.
a.
Identify all components that must be manipulated by the operator to establish the desired cooling flow path with the given plant conditions.
(0.5) established as a result of this b.
Describe the complete $1 flow action.
(1.75) l l
)
ANSWER 2.10 (2.25) a.
Open HCV's 349 & 350 (SDHX outlet to HPSI suction) d ea)
I j
b.
Containnent Sump ) Containment Spray Pumps > SDKX > HPSI ) Loop Safety j
Injection Nozzles. (0.25 for ea component. 0.5 for sequence) 1 REFERENCE 1
FCS LP 7-11-22, LO 1.4 E0P-03 000009A100
...(KA'S) l QUEST!0ti 2.11 (1.50)
A malfunction in t h e p r e s s '. ' ' e r pressure measurement
'A' channel (A/ PIA-102Y) causes the output to go to minimum.
1
]
a.
List two (2) RPS trip units which receive a signal frcm this channel I
and HOW each is affected by the malfunction.
(0.75) b.
Explain why the reactor will or will not trip.
(0.75) l l
)
i j
l i
l
~ -
1 9
2._e RLebiI_DESIGtd_1blCLUDitJG_SeEEIY_atJD_EMERGEblCY_SYSIEMS PAGE 26 1
1 ANSWER 2.11 (1.50) a.
High Pressure and TM/LP (0.25)
The:high pressure trip unit will not be affected (since minimum output 4
equates to a low pressure condi tion). (0.25)
The TM/LP trip unit will go to the trip condlion initiating a channel "A"
trip and a pre-trip alarm. (0.25) b.
A reactor trip will not occur (0.25) since the 2/4 logic is not satisfied by the single channel trip. (0,5)
REFERENCE FCS LP 7-12-25, LO 1.03, 007-12-26-H-25-1.03-0001 000027A216 012000A202
...(KA'S)
QUESTION 2.12 (3.00) 7 Using the CUCS line c'l agr am prov i de d, circle each component which receives a control signal as a result of a PPLS and indicate the exp(eted response of *he component to the control signal.
ANSWER 2.12 (3.00) 1 The following valves shut l
(0.2 ea)
The following valves opens
]
HC) 268 I
(0,2 ea) l All 3 charging pumps start (0.4)
Boric acid pumps A & B start (0.2 ea)
REFERENCE FCS LP 7-11-2, LO 1.4 000009A104 004000A202
...(KA'S) l i
]
(***** END OF CATEGORY 02
- )
\\
I
^
l
3._ 1NSIEUMENIS_eUQ_CONIROLS PAGE 27 QUESTION-3.01 (2.00)
Pressurizer pressure is 2270 psig decreasing and pressurizer level is -3.0%
from program level and decreasing.
Indicate the status of each i tem bel ow with OPEN, CLOSED, ON, OFF, MINIMUM, or MID (> minimum-but < maximum) as appropriate.
Assume that the reactor is operating at 100% and the charging pump Standby Switch is selected to CH-18-1C.
NOTE:
Other terms may be used to describe component status if desired.
a.
Spray valves b.
Proportional heaters c.
Backup heaters d.
Le tdown fl ow con trol valves l
e.
Charging pumps B and C (answer for each) i 9
ANSWER 3.01 (2.00) 5 a.
Open b.
Off l
c.
Off
[
d.
Minimum e.
B - On C - Off (6 at 0.333 each) l REFERENCE i
010000K603 011000A102
...(KA'S) i QUESTION 3.02 (2.50)
The NR NIS power level provides indication at the RPS and CB-4 plus i
j actuation of the Level 1 and rod trip blstable on the drawer front.
List i
six (6) other uses of the power signal from the NR NIS.
i i
i t
i i
k i
I i
i
)
- 31..... INSTEle4FNTS AND rrw T e nt R PAGE 28 ANSWER 3.02 (2.50)
'(ANY 6 AT 0.416 EA)
-1.-
.Q power auctioneer 2.
APD calcul ator 3.
Comparator averager 4.
Deviation comparison and alarm 5.
Azimuthat tilt (Tq) 6.
Reactor Reg 7.
VOPT (HI PWR) 8.
Plant computer 9.
TM/LP cal cul ator REFERENCE FCS LP 7-12-19, LO 1.04 015000K100
...(KA'S)
I QUESTION 3.03 (2.00)
What signals, including power levels, type of channels, and number of channels, produce each of the f oll owing?
a.
Bypass of the "High Startup Rate" reactor trip.
(1.0) b.
Removal of voltage from the unshleided startup detectors.
(0.5) c.
Bypass of the "Loss of Lead" reactor trip.
(0,5)
ANSWER 3.03 (2.00) a.
Below 10E -4% power (0.25) on 2 wide range channels (0.25)
Above 15% power (0.25) on 3 of 4 power range safety channels (0.25) b.
Above 10E-5% power (0.25) on (that wide range channel) 2 wide range channels (0.25) i i
c.
Be l ow 15% power (0.25) on 3 of 4 power range safety channels (0.25)
REFERENCE FCS LP 7-12-18, LO 1.0 012000K406
...(KA'S)
L
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
I
{
3._a1USIEUMENIS_eUQ_CQUIRQLS PAGE 29 i
QUESTION 3.04 (2.00)
The qualified safety parameter display system (QSPDS) calculates and displays three (3) saturation margin monitoring (SMM) values.
c.
List the three (3) saturation margins and the temperature sensors (inputs) used to calculate each.
If redundant sensors are available, state the criteria which determines the input used by the SMM (1.5) b.
What signal is used to provide the saturation temperature?
(0.5)
ANSWER 3.04 (2.00)
Upper head saturation margin - max of the top three UHJTCs CET saturation margin - representative CET b.
Pressurizer pressure (0.5)
REFERENCE FCS LP 2-12-23, LO 1.7, 01-OSPDS-1-9, REV 0 194001A115
...(KA'S)
QUESTION 3.05 (2.50) a.
What three (3) conditions initiate the Feedwater Regulating system protective ramp function?
(1.S) 6.
Once initiated, how can the ramp be stopped, other than by use of the ramp stop button on the man / auto station?
(0.5) c.
What is the status of the feedwater regulating valve and controller at the conclusion of the rampdown phase?
(0.5)
.L :.ltdSIR121ENIS_eUD_DDtiIRDLS PAGE 30-l l
ANSWER 3.05 (2.50) a.
1.
Turbine trip 2.
Feedwater Regulating system in automatic (0.625 ea) b.
- 1..
Reset the turbine trip (0.25)-
2.
Place auto / inhibit switch in the inhibit position (0.25) 3.
Put the controller in manual (0.25) c.
Controller in auto (0.25) with valve approximately 8% open (0.25)
REFERENCE FCS LP 7-12-15, rev.
1, p.
34-36 059000K418
...(KA'S)
QUESTION 3.06 (2.50)
Three (3) level transmitters (LT-903X, LT-903Y, & LT-903Y-1) supply measurement of down c ome r level used by the Feedwater Regulating system (FRS) to control SG RC-2A' level.
(2.5) a.
What is one (1) of the two (2) u se s t h a t '.iey be made by the FRS of the level signal from LT-903X?
(0.$)
6.
What are three (3) of the four (4) uses that may be made by the FRS of the level signal f rom LT-903Y?
(1.5) c.
What is the single use by FRS of the level signal from LT-903Y-1?(0.5)
ANSWER 3.06 (2.50) a.
1.
Auto control using 3 element mode 2.
Indication for manual control (1/2 9 0.5) i b.
1.
High level override of reg valve 2.
Auto control of bypass valve 3.
Indication for manual control 4.
Level alarms (hl and lo)
(Any 3/4 3 0.5 e.; 1.5) c.
Indication (on AI-179)
(0.8)
REFERENCE FCS LP 7-12-28, rev.
1, p.
19, 20 035010K401 035010K408
...(KA'S)
(*****
CAT E GO RY 0 3 C ONT I NUED ON N EX T PAG E * * * * * )
s
- L :.lNSIRLD4FNTS eblD CDtHRDLS PAGE.31 QUESTION 3.07 (2.00)
What are the four (4) ENGINE trips for the emergency diesel generator which
.are overridden on an emergency start?
(2.0)
ANSWER 3.07 (2.00) 1.
Low lube oil pressure 2.
High crankcase pressure t-3.
High water temperature 4.
Generator lockout
( 0.5 ea.)
REFERENCE I
FCS OCN: 007-013-06-C-05-1.09-0002 064000K402
...(KA'S) 1 QUESTION 3.08 (2.50) a.
What three (3) conditions can cause a Rod Block?
b.
How can the operator reposition CEDMs as necessary to clear a Rod Block?
c.
What two (2) conditions / signals can cause a Rod Withdrawal Prohibit?
ANSWER 3.08 (2.50) a.
PDIL, Deviation, or Out of Sequence / Overlap. (0.25 ea) b.
8/ controlling the affected CEDMs in manual individual while holding the Rod Block Bypass Switch in "bypass". (0.75) l c.
Hi power pre-trip (0.5)
Hi SUR pre-trip (0.5) l REFERENCE FCS LP 7-12-26 E0 1.3 001050K401
...(KA'S) l
\\
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l i
- 3._21NSIEudEUIS_eUQ_CQUIEOLS PAGE 32 QUESTION 3.09 (2.00) c.
What is the function of the Diverse Scram System (DSS)?
(0.76) b.
What plant parameter (include setpoint value and logic) will actuate the DSS?
(1.25)
ANSWER 3.09 (2.00) c.
Generate a reactor / turbine trip signal independent of the RPS under ATWS (failure of RPS to generate a trip signal) conditions. ( 0. 75) b.
Pressurizer pressure (0.5) at 2450 vsla (0.25) on 2/4 detectors (0.25) tripping 1/2 channels (0.25).
REFERENCE FCS LP 7-12-25, LO 5.1, 5.2 000029A209 10CFR50.62
...(KA'S)
QUESTION 3.10 (2.00) a.
At what power level is the extended range operation of the Wide Range Log Power Channel s enabl ed AND what operator action is necessary to permit extended range operation?
(0.5) b.
How does extended range operation differ from normal operation of a Wide Range Log Power Channel?
(0.5) c.
When audio monitoring of count rate is required, what two (2) switches must be positioned to activate the audio circuit?
(0.5) d.
Aside from possible alarms, what is the control room indication of a non-functioning Campbell circuit for a particular Wide Range Log Power Channel as power is raised from 10E-6% to 10E+1%?
(0.5) l t
l l
l t
'3 _21USIRUbEUIS_eUD_CONIEDLS PAGE 33 ANSWER 3.10 (2.00) c.
Extended range operation i s e n abl e d be l ow 4x 10 E-7% p owe r. (0.25) lne operator must take the extended range cutout switch to 0FF (0.25) (at about 10 cps control room indication),
b.
Both chambers procide singals that are summed to the channel in extended range with one detector's output disconnected when above approximately 10E-5% power.
(0.5) c.
The selector switch on the audio amp must be set to the desired channel (0.25) and the toggle switch for the selected channel must be in the proper position (0.25).
d.
The indication from the affected channel will stop at about 10E-2%
p owe r. (0.5)
(The remaining channels will continue to show normal power increase.)
REFERENCE FCS LP 7-12-18, LO 1.3, 1.5, 1.7, 1.9 015000A202 015000K401
...(KA'S)
QUESTION 3.11 (3.00)
Fill in the numbered blocks on the simplified ESC logic diagram with the proper initiation and actuation signals.
ANSWER 3.11 (3.00) 1.
CRHS 8.
SG1S 2.
SGLS 9.
CSAS 3.
PPLS 10.
RAS 4.
STLS 11.
S!AS 5.
SGLL 12.
CIAS 6.
SGLP 13.
VIAS 7.
SGDP 14.
OPLS (0.214 EACH)
REFERENCE FCS LP 7-12-14, LO 1.1 013000A301
...(KA'S)
(+**** END OF CATEGORY 03 ++++*)
d _2RROCEDURES..=_MQmat ABNORt2L,_EMERGENCYED PAGE 34 BAD 10 LOGICAL _CONIROL QUESTION 4.01 (1.00)
Choose the correct answer belows a.
During a reactor startup, CEA motion inhibit due to a rod block is cleared by using the Manual Group Mode, b.
Manual Individual is used to return a group of CEA's to the proper sequence or overlap.
c.
It is not necessary to use the "Rod Block Bypass Switch" when inserting CEA's whose motion is inhibited by an inoperable CEA.
d.
"Bypass" operation is not necessary when the Rod Drive Power Interrupt alarm is in.
ANSWER 4.01 (1.00)
"b."
REFERENCE OI-RR-1, Pp.2,3 FCS LP 7-12-26, LO 1.3, 2.1 001000K401
...(KA'S)
QUESTION 4.02 (1.00)
Choose the letter which corresponds to the best answer.
The purpose of OPERATING INSTRUCTIONS (01's) is tot a.
Provide general guidelines for the operation of specific equipment.
b.
Provide de tail ed instruction for the integrated operation of the plant.
c.
Provide general guidelines for operation of a particular system.
d.
Provide detailed instructions for the operation of plant systems and equipment.
~ _
_4...zEEDCgDumEE NonMal 7,,ggb[QRt%l, Et1ERDENCY eND PAGE 35 RADIOL 0GICAL.fDtdIRDL l
il ANSWER 4.02 (1.00)
"d."
REFERENCE ~
Standing Order 0-17, Section 1.1 & Question Bank 001000G013
...(KA'S)
QUESTION 4.03 (1.00) i From the f ol l owi ng possi bl e combi na t i ons of par ame ters, sel ec t the one i
'c omb i n a t i on that most correctly describes characterl. tics of natural circulation cooling.
RCS SG Thot Core Exit Tcold Subcooling Pressure Thermocouples a.
20 F Decreasing Stable Decreasing Tsat for SG Press b.
20 F Stable Tsat for Decreasing Stable t
RCS Press c.
30 F Decreasing Decreasing Stable Stable l
.I l
d.
40 F Stable Decreasing Increasing Tsat for l-SG Press I
1 e
ANSWER 4.03 (1.00)
- a.'
i REFERENCE l
l E0P-20, P. 97 l
000038K103 193008K122
...(KA'S) r i
l L
l l
I t
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
j l
t 1
i
' A _ 2 E R O C E Dt.1R E S _ = _tJ O Rt e t, ABblO2t:1&LyEMERGEtJCY_eUD PAGE 36 R&D10 LOGICAL _CatJIROL QUESTION 4.04 (1.00)
An area where localized radiation levels are 5 times higher than the 2.5 mrem /hr general area, should be posted ass a.
Restricted area b.
Hot spot c.
Radiation Area p
d.
High Radiation Area ANSWER 4.04 (1.00) m e" REFERENCE Radiation Protection Manual, P. VII-3-2 194001K103
...(KA'S)
GUESTION 4.05 (2.50)
During a plant heatup from cold shutdown in accordance with 01-RC-3:
a.
What are the two (2) conditions, one of which must be met in order to preclude an RCS pressure surge pa8,r to starting a Reactor Coolant Pump (RCP)?
(1.0) b.
What are the reasons for the upper and lower pressure limits for RCS pressure maintained at 250 deg's RCS temperature?
(1.0) c.
What are the two (2) specific Indications to the operator that a steam bubble has been formed in the pressurizer even though level Indication
/
is still at maximum?
(0.5) l l
l l
. _4aEBDCEDUBES_ _UDRMAL,_eBUDRtML,_ EMERGENCY _eUD PAGE 37 EADlDLDBlCAL_CDUXROL ANSWER 4.05 (2.50) c.
1.
A pressur izer bubbl e exi sts.
2.
S/G secondary is less than 50 deg's above RCS temp. (0.5 ea.)
b.
Pressure is maintgained so as not to lift PORVs OR to ensure pressure /
temperature limits are not exceeded (Iow temperature over-pressuriza-tion) (0.5) and to be sufficient for RCP operation (NPSH). (0.5) c.
Letdown flow > charging flow, CO.25) and pressure stable. CO.251 REFERENCE OI-RC-3, Pp. 2-7 OI-RC-9, P.
1 FCS LP 7-11-23, LO 1.2 005000K104 010000G005 0100000004 011000A101
...(KA'S)
QUESTION 4.06 (2.00)
During power changes:
a.
What are the two (2) documents referenced by OP-4 that place limits on rate of power change and CEA movement?
(1.0) 6.
What must be done if it becomes necessary to change boron concentration to reposition CEA's during a load increase?
(0.5) c.
What are two (2) conditions that may restric t the operators choice to insert CEA's?
(0.5)
ANSWER 4.06 (2.00) a.
Technical Data Book (0.5) and Post Refueling Core Physics Testing and Power Ascension Procedure (SP-PRCPT-1) or 01-RR-1.
(any 2 at 0.5 ea) b.
The load change must be stopped. (0.5) c.
PDIL or ASI limits (0.25 ea.]
REFERENCE
- TDB, P.
III-22 OP-4, Pp. 1,2 001000G005
... (' KA ' S )
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
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' d._ *EROCEDURES ' - NnRMAL._eablORt2L,_EdERGEblC'f_ablQ PAGE 38 BAD 10 LOGICAL _COtJIROL QUESTION 4.07 (1.50)
A s t amp is placed on the front of a procedure, requiring a signature and date, prior to using the procedure.
i a.
What are the stamp and signature for?
(1.0) 6.
Whose signature is required on the stamped area?
(0.5) l ANSWER 4.07 (1.30) 4 c.
The stamp and signature indicate that the latest revision of the procedure is being used.
[1.0) b.
the individual using the procedure.
[0.5)
REFERENCE i
Standing Order G-7 194001A101
...(KA'S) 1 QUESTION 4.00 (1.50)
According to the Radiation Protection Manual, what are your four (4) specific immediate responsibilities should you observe a large spl11 of radioactive liquid material, WITHOUT trained HP personnel present?
4 ANSWER 4.08 (1.50) t.
Warn others in the area.
2.
Notify the control room.
3.
Notify RP 4.
Attempt to contain (prevent the spread of) the spill. 10.375 ea)
REFERENCE Radiation Protection Manual, P. VII-8-1-1 194001K103
...(KA'G)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE
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i
< d._.r.EBOCEQl. LEES _ __hlDat2Li_eabl08t2L._EtiEEGEtJC'f_eUQ PAGE 39 R&D10 LOGICAL _CONIROL QUESTION 4.09 (2.00)
According to AOP-19, Loss of Shutdown Cooling, what are the "LAST RESORT" c
methods (Flow. Paths) for residual heat removal for each of the following conditions.
For each case, describe the compl e te open/c l osed l oop pa th,
n.
The reactor head installed.
(1.0) b.
The reactor head removed, transier canal open.
(1.0)
ANSWER 4.09 (2.00) l a.
SIRWT > HPS! Pumps into primary system > FORV's OR, Loop Drains > RCDT t
j
> RCDT drain pump back to SIRWT.
C 0.15 f or each componen t i n f l ow path and 0.05 for each manipulation to attain correct order.)
b.
Align Fuel Transfer Canal Drain Pump to take suction from SIRWf and discharge to Spent Fuel Pool. (0.51 Align SFP pumps to transfer water from SFP to SIRWT.
[0,5)
REFERENCE AOP-19, P. 3 LP 2-1-41, L.O.
1.0, 1.2
{
000025K101 000025K301
...(KA'S)
QUESTION 4.10 (1.50) i a.
Describe the action (system manipulation) an operator would perform 4
if a head bubble was suspected during natural c i rcu l a t i on cool down.
t (0.76) b.
What action is then taken to confirm the existence of a bubble in the
[
head?
(0.75) l 1
i a
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i i
4 1
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i
A-.:.EROCEQL1RES_=_NORt2L.,._aStaQRtnL,_Et:1ERGENCY AND PAGE 40 RAQ10 LOGICAL _CQtEROL ANSWER 4.10 (1.50) a.
1.
verify letdown isolated 2.
Stop depressurization and, if required, repressurize RCS to greater than or equal to 20deg.F subcooling 3.
Pressurize and depressurize RCS within the limits of Fig. 4-2 4.
If necessary, operate PZR vent and/or reactor vesssi head vent to clear trapped non-condensible gases.
(0.1875 ea) b.
Monitor P2R level and/or RULMS (.5) for trending of RCS inventory (.25)
REFERENCE
62 002000A403
...(KA'S)
QUESTION 4.11 (1.50)
OI-VA-2 gives the operator guidelines on how to IDENTIFY and ISOLATE a leak in the Auxiliary Building which causes a Stack High Radiation Alarm.
State t
these guidelines.
(1.5) i ANSWER 4.11 (1.50)
Shut down supply and exhaust fans. (0.5)
Isolate all rooms. (0.SJ Start t
one exhaust fan and open dampers one at a time while observing monitor until leak is found. (0.53 (Area monitors may be of some help.)
PEFERENCE I
FCS E0B 000060A202
...(KA'S)
QUESTION 4.12 (1.50) a.
Name the three (3) items that should be removed by control room personnel if the control room is evacuated due to a fire.
(0.75) b.
What three (3) inmediate actions should be performed by operators prior to leaving the control room if it is evacuated due to toxic gas?
(0.75)
.o
'A-
- RROCEQLIBES - NOBtBL._eBbl08tML._EMERGENCI_eblQ PAGE 41 RADIOLOGICAL _CONIROL l
-l l
ANSWER 4.12 (1.50) i a.
Key for ASP, Key for east swgr. room and transceiver. [0.25 ea.]
b.
1.
Trip the reactor 2.
Trip the turbine 3.
Start the turr.ing gear oli and lift pumps.
[0.25 ea.]
REFERENCE AOP-6, Pp2-4 AOP-7, P.
1 000068G006 000068G011
.. (KA'S)
QUESTION 4.13 (1.25) i
.The reactor has experienced a Reactor Trip.
An early step of E0P-01 has you trip the turbine.
a.
How do you verify that the turbine tripped?
(0.5) b.
What are the two (2) methods specified to trip or stop the turbine if It did not manually trip sod is causing continued cooldown?
(0.75) l h
ANSWER 4.13 (1.25)
[
a.
Verify all turbine intercept and stop valves closed. [0.5)
I b.
1.
Control fluid Pumps to Pull-to-lock.
2.
Close MS!V's (and Bypasses)
(0.375 ea.)
REFERENCE E0P-01, P.
4 000007A101 000007A107
...(KA'3) t y
L OUESTION 4.14 (1.75) 4 a.
What are two (2) actions required of the operator if there is an indication that four (4) CEA's have not fully inser ted f ol l owi ng actuation of the manual reactor trip pushbutton?
(0.75)
{
b.
Per E0P-03 (LOCA), what are four (4) indirect indications or alarms checked by the aperator to verify that PORV's and P2R code safeties are not open?
(1.0)
[
1 k
4
- L._1EROCEQllEES_=_blOBtnLJBbl0BtSL,_EtdERGEblC'f_ablQ PAGE 42 EaQ10 LOGICAL _CQblIROL ANSWER 4.14 (1.75) a.
1.
Open CEDM clutch power supply breakers.
2.
Emergency borate.
[0.375 ea.]
b.
1.
Relief header temperature starms.
2.
Rel i ef header acous t i c fl ow al arms.
3.
Quench Tank i
pressure il temperature ill level (any 4 at 0.25 ea)
REFERENCE E0P-01, P. 2 E0P-03, P. 6 000029K309 00002rK310 000029K311 000029K312
...(KA'S)
QUESTION 4.15 (1.50) a.
The STEAM GENERATOR TUBE RUPTURE Procedure (EOP-04) cautions the operator early on to avoid use of HCV-1040 and FW-10.
State the reason for this caution.
(0.5) b.
The operator is directed to determine which Steam Generator has the tube rupture by monitoring or checking four (4) items.
What are these four (4) items?
(1.0)
ANSWER 4.15 (1.50) l l
a.
Avoid unmonitored release of activity to the environment. [0.53 b.
1.
S/G activity 2.
Steam line radiation 3.
S/G Bl owdown radi at i on 4.
S/G 1evels (0.25 ea.)
REFERENCE E0P-04, P. 5 000C09A202 000037K305
...(KA'S)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- d._f.RROCEDURES_=_bl08tML._eabl08t2L+_Et:1ERGEblCL ANQ PAGE 43-bed 10 LOGICAL _CQblIROL l
l'
. QUESTION 4.16 (1.00)
What are the two (2) requirements to be met before the operator implements the LONG TERM ACTIONS of the Emergency Operating Procedures?
ANSWER 4.16 (1.00)
Appropriate operator actions for aII success paths in use have been performed (0.5) and ALL safety function acceptance criteria are being i
satisfied. (0,5) i REFERENCE l
E0P-20, P.4 000011G011 0000110012
...(KA'S) l i
-QUESTION 4.17 (1.50) i a.
State whether each of the following operations would tend to I
INCREASE, DECREASE, or NOT AFFECT the size of a void existing in the reactor vessel head.
l i
1.
Pressurizer spray operation.
2.
Pressurizer heater operation.
3.
Reactor vessel head vent operation.
(0.73) b.
Aside from reactor vessel level, what are three (3) criteria to be met before.the operator is allowed to throttle or stop Safety Injection flow?
(0.76)
?
ANSWER 4.17 (1.50)
I a.
1.
Increase t
2.
Decrease l
3.
Decrease (0.25 ea.]
[
b.
1.
RCS subcooling ) 20 deg's.
2.
PZR level > or = 45% and not decreasing.
3.
At least one S/G available for heat removal.
(0.25 ea.]
[
REFERENCE f
E0P-03, Pp. 17, 26 000009A238 000009K324
...<KA'S) t'
?
i r
r n.
_ - n,,
FORT CALHOUN STATION REACTOR / SENIOR REACTOR OPERATOR EXAMINATION ALMINISTERED ON JULY 26, 1988 REVIEW C0ftENTS i
OVESTION 1.11/5.11 The Fort Calhoun core often has a slightly positive MTC at 80L.
To eliminate possible confusion, it is recomended using MOL instead of BOL for future use of this question.
OVESTION 1.18/5.16
?
For future use it is recomended clarifying the quest %n as follows:
j A relief valve on a pipe opens at 885 psig.
The temperature of the steam downstream of the relief valve, exhausted to the atmosphere, is 320' F.
l The fluid within the pipe upstream of the relief valve is saturated steam.
[
Using the steam tables provided, what is the approximate temperature of the fluid within the pipe, immediately upstream of the relief valve?
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i OVESTION 5.22 i
i During nominal 100% power operation, the four reactor coolant pumps draw a total of approximately eight megawatts of power.
Rated core power is 1500 megawatts.
The power contributed by the reactor coolant pumps as compared to the power produced by the reactor core could be considered insignificant (8MW/1500MW=0.5%).
Since choice c. addresses "normal operating temperature,"
which irplies 100% power conditions, choice c. is a valid response.
The answer key should be revised to accept either response c. or d. for full credit, e
Reference:
SP-CTPC-1, revision 3 for calculation of NSSS calorimetric i
00ESTION 5.23 At the Fort Calhoun Station, the Emergency Operating Procedures consider adequate subcooling margin to be 20' F.
The RCP NPSH curve is more restrictive than the 20' F.
A loss of subcooling margin also means that RCP NPSH requirements have not been met, therefore pump cavitation (that is, two phase j
conditions) and pump damage could occur.
The answer key should be revised to f
accept either response a. or b. for full credit.
l l
Reference:
TDB page !!!. 25 and Figure 4-2 of the Emergency Operating Procedures l
1 00ESTION 2.03/6.01 The drawing provided for the answer incorrectly shows the relative location of i
a few penetrations. Full credit should be allowed for responding as indicated i
on the attached drawing.
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Reference:
Marked-up copy of answer 6.01 and P&l0 E-23866-210-110 l
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OVESTION 2.07/6.02 Instrument air, not service air, provides the motive force to aosition the dampers (HCV-724A/BandHCV-725A/B).
The question does not as( what type of 4
opere, ting mechanism is provided or how it functions.
For part c.,
it is suggested that full credit be given for stating that control air (instrument air) provides the motive force to position the dampers.
eference: P&ID 114M M4$4, ih. E of 5 Di'55T.0N 2.08/6. 1 Part b.3. of the ansMi den?.i m s FW10 steam stop valves YCV-1145 & 1145A/8.
The answer should be cas. c.d a reflect the correct valve numbers, YCV-1045 &
1945A/B.
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Reference:
Palb 11405-M-U2 OVESTION 2.09/6.05 i
Battery charger current is variable, depending on 125 VDC bus loading.
The battery charger are rated for maximum current of 400 amps each. LP 7-13-4 j
states that normai cattery charger current output is aooroximately 170 - 200 J
amps. Battery charger current on AI-41A indicated approximately 140 amps on 4
7/27/88. On 8/1/88, battery charger current on Al-418 indicated ap)roximately 225 amps.
It is recommended accepting 130 - 230 amps for battery clarger current.
I j
Reference:
LP 7-13-4, page 22, rev.2 i
OVESTION 3.01/6.07 l
The suggested responses provided in the question do not include "Full Open";
therefore, "Open" should be considered a correct answer for part a.
l OVESTION 3.02 The question states rod trip bistable.
It should say rod drop bistable.
This comment is provided for future use of this question.
OVESTION 3.03 The correct answers should be a.
Each RPS High SUR trip channel is bypassed below 10E-4% power on the respective wide range channel or above 15% power on the respective power range safety channel.
SUR trip must be bypcssed on 3 of 4 RPS channels to prevent High SUR reactor trip (since the RPS employs 2/4 logic),
b.
Above approximately 10E-5% power on that wide range channel.
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c.
Each RPS Loss of Load trip channel is bypassed below 15% power on the j-respective power ran e safety channel.
Loss of Load trip must be bypassed on 3 of 4 RPS channe s to prevent reactor trip (since the RPS employs 2/4 logic).
Reference:
LP 7-12-25, rev. 2, pages 39 & 61; LP 7-12-18, rev. 1, page 31 OVESTION 3.05 The Feedwater Regulating Syster,was modified during the previous refueling outage.
There are now only two (2) conditions that must be met to initiate rampdown and there is not a ramp stop button on the man / auto station.
f The correct answers should be:
a.
1.
Turbine trip 2.
Feedwater Regulating System in automatic b.
1.
Reset the turbine trip 2.
Place auto / inhibit switch in the inhibit position 3.
Put the controller in manual c.
Controller in auto with the valve approximately 8's open.
Reference:
LP 7-12-15, rev. 2, pages 12, 13, & 39 OVESTION 3.09/6.09 Portions of the Diverse Scram System were installed during the previous refuel-ing outage; however, the system is not yet operational.
The lesson plans were revised to reflect the installation of the system so the operators could be i
made aware of its purpose.
Part b. of the question asks for the setpoint value and logic for the parameter that actuates the DSS; it does not solicit how many 055 channels are required to cause a reactor trip.
It is recommended that part b, of the question be deleted since the OSS is not yet operable.
However, should it be decided to retain part b., the answer j
should read as follows:
Pressurizerpressure(0.5)at2450 psia (2435psig)(0.25))on2/4 detectors b.
(0.5).
(The correct setpoint is 2450 psia, not 2385 psia.
Reference:
LP 7-12-25, rev. 2, page 87 OVEST!0N 3.10.a/6.10.a The extended range is enabled at approximately 10*7's power.
The EXT RNG C/0
) mu switch (togleON/0Fg/10'ytbeOFFforthechanneltogointoextendedrange-administrat vely 10" which could correspond to 10 to 100 CPS.
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ft is recommended that the answer be revised as follows:
Extended range operation is enabled at approximately 10-7% power (0.25).
a.-
The operator must take the extended range cutout switch to 0FF (0.25).
Reference:
LP 7-12-18, rev. 1, pages 31-34 & 51 00ESTION 3.10.b/6.10.b l
It is recommended that the answer be revised as follows:
[
t b.
Both chambers provide signals that are summed to the channel in extended ranSewithonedetector'soutputdisconnectedwhenaboveapproximately 10 % power (the high voltage is never removed from the dual fission chambers.)
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Reference:
LP 7-12-18, rev. 1, pages 31-34 f
OVESTION 3.10.d/6.10.d The correct answer to this question depends on the definition of
[
"non-functioningj.
If the Campbelling leg of the circuitry is putting out zero signal above 10' % power, the answer is correct.
"Non-functioning" can also
(
be interpreted as signals generated above or below normal which will cause the l
channel to daviate from the remaining channels in indication.
It is recommended that the candidate's stated assumptions be considered when i
l evaluating his response.
(
OVEST!0N 4.03 An indication of natural circulation cooling is no abnormal difference between l
Since answer a. indicates Thot is stable but core exit t
thermocouples are decreasing, it shows an abnormal difference between CET's and t
RCS TH and, therefore, cannot be a correct answer.
t It is recommended that this question be deleted from the examination since there is no correct response.
(
Reference E0P-20, page 97 00ESTION 4.05/7.01 It is recomended that the answer for part b. be revised as follows:
[
b.
Pressure is maintained so as not to lift PORV's QR to ensure over-pressurization) (either answer is correct) (0.5)perature pressure / temperature limits are not exceeded (low tem and to be sufficient for RCP operation (NPSH).
(0.5)
Reference:
01-RC-3, CAUTION after step I!!.A., rev. 33 i
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e.,
OVESTION 4.06.a/7.02.a It is recommended that the answer be revised to accept any two of the following responses for full credit:
1.
Technical Data Book 2.
Post Refueling Core Physics Testing and Power Ascension Procedure (SP-PRCPT-1) 3.
01-RR-1
Reference:
OP-4, IV.A.2.f., rev. 26 OVESTION 4.06.c/7.02.c This question does not directly refer to OP-4.
Therefore, the answer should be revised to accept any two of the following responses:
1.
PDIL limits (or :hutdown margin requirements) 2.
ASI limits 3.
Meet MTC values as prescribed by the Reactor Engineer
Reference:
OP-3, III.1, rev. 23 OVESTION 4.10/7.08 It is recommended that the answer be revised as follows:
a.
1.
Verify letdown isolated 2.
Stop depressurization and, if required, repressurize RCS to greater than or equal to 20 F subcooling.
3.
Pressurize and depressurize RCS within the limits of Figure 4-2.
4.
If necessary, operate PZR vent and/or reactor vessel head vent to clear trapped non-condensible gases, b.
Monitor PZR level and/or RVLMS for trending of RCS inventory.
Reference:
E0P-02, 3.37 page 22, rev. 2 OVESTION 7.10 The following change is recomended for the answer to part c.:
Once-through cooling ((HR-4), found in the RCS and Core Heat Removal FRP c.
(found in E0P-20).
Identifying the appropriate procedure, either by number or title, should be acceptable.)
Reference:
E0P-20, rev. 2
,Qq QT' M 4,12 Answer b.3. should state "Start the turning gear oil pump and lift pumps.
Reference:
AOP 6 Page 5
l
- ,'r c
00ESTION 4.14/7.11 Each parameter listed as a correct response for part b. is equip)ed with indi-cation and alarm.
Specifying either alarm or indication should ae considered correct.
OVESTION 7.13 f
f
)
Part a. solicits specifically operator actions for confirming the diagnosis of an uncomplicated reactor trip. The Shift Technical Advisor perfor.ms the safety function status check and informs the Shift Supervisor of the results.
It is recommended that the answer be revised as follows, to require only operator i
actions l
i.
a.
Verify that the safety function status check criteria are satisfied.
l
Reference:
E0P-01, page 17, rev. 1 i
i OUESTION 8.01 l
The control room crew composition and titles have recently been changed.
The answer to part b. should be reYised as follows:
e i
t b.
ThaReactorOperator(assistedbytheAR0)
OR The..'censedSeniorOperator(LS0)orLicensedOperator(LO)(accepteither res o e for full credit) j
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s i
Reference:
FCS Exam Bank Question 007-015-48-H-07-4.04-0001 4
{
00ESTION 8.03 f
Part a. of the question asks "...what two (2) signatures are required...", when t
t j
in reality three (3) signatures are required for approval of an On-the-Spot procedure change.
The answer key should be revised to accept any two of the 4
following responses:
1.
PRC member i
2.
Senior Licensed PRC member 3.
Shift Supervisor l
Reference:
Standing Order G-30 Section 3.3.1, rev. 21
{
00ESTION_,$,di The answer for part a. should be revised to reflect the current revision of Standing Order G-28.
The following members of the Plant fire Brigade report to the fire areat FireBrigadeLeader(LO*)
i Two other operators (Equipment / Auxiliary Operators)
Two security force personnel 1
~eference:
Standing Order G-28, sect. 3.2.5.1 9-j and 3.2.5.2.b f, rev. 12, l
June 13,1988 j
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