ML20137N863
| ML20137N863 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/24/1986 |
| From: | Cooley R, Mccrory S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20137N856 | List: |
| References | |
| 50-368-OL-86-01, 50-368-OL-86-1, NUDOCS 8602040270 | |
| Download: ML20137N863 (150) | |
Text
.. . _. _ _.. _ _ .. .
ANO 2 EXAMINATION REPORT Report Number: 50-368/0L-86-01
, Docket No: 50-368 License No.: NPF-6 Licensee: Arkansas Power & Light Company P.O. Box 551 Little Rock, Arkansas 72203 Examinations administered at Arkansas Nuclear One Unit 2 (ANO 2)
Chief Examiner: / e ,# / 6 5.ML.RcCf 6rf, Examiner pte' Approved by: s L/31f r R.A. Cooley,Se$ionChief I[Dat624/ hb ,
i ,
Sumary Examinations conducted on December 10, 1985.
Written and oral examinations were administered to six (6) Senior
-Reactor Operators and ten (10) Reactor Operators. All candidates except 1
one Senior Reactor Operator passed these examinations, s-i fDR602040270 860129
] g ADOCK 05000360 PDR
)
N ANO 2 EXAMINATION REPORT No. 50-386/0L-86-01 Report Details
- 1. Examination Results SRO Candidates R0 Candidates Total Pass Fail % Total Pass Fait %
6 5 1 83 10 10 0 100
- 2. Examiners S. L. McCrory, Chief Examiner, NRC J. Pellet, NRC
- 0. Graves, NRC J. Whittemore, NRC R. Cooley, NRC
- 3. Examination Report This Exaniination Report is composed of the sections listed below.
A. Examination Review Comment Resolution B. Exit Meeting Minutes C. ANO 2 Examination Key (SR0/R0 Questions and Answers)
Performance results for individual candidates are not included in this report because examination reports are placed in NRC's Public Document Room as a matter of course.
A. Examination Review Comment Resolution In general, editorial connents or changes made during the examination, the examination review, or subsequent grading reviews are not addressed by this resolution section. This section reflects resolution of substantive comments made during the examination review. Attachment 1 contains the facility comments on both the Reactor Operator and Senior Reactor Operator examinations.
Conments, with which NRC did not agree, will be specifically addressed in this section. All other comments are incorpcrated into the master examination key (as requested or with only minor modification) which is provided elsewhere in this report as are all other changes mentioned above but not discussed herein.
L
ANO 2 EXAMINATION REPORT No. 50-386/0L-86-01 COPNENTS (1) 1.4/ REJECT. The 1/M value of 0.0129 is based on the 5.6 actual source strength which is not normally used to develop 1/M plots. Additionally, the proposed method for calculating this value for 1/M cannot be done without first going through the calculations of parts A and B.
(2) 1.6 REJECT. This question was taken from the facility developed unit 2 examination bank and is consistent with the original answer. Further, an increase of 40%
is not considered slight.
(3) 1.11 REJECT. The term "Xe-eq" is defined in part A as
" equilibrium Xenon reactivity". Reactivity worth is usually applied to incremental reactivity change such as reactivity per inch for CEA motion and reactivity per ppm for boron concentration changes. Equilibrium reactivity is an integral or total value. Using boron concentration changes over core life to justify most reactivity changes of the other materials in the core is poor. As a core ages several changes are taking place which cause the macroscopic cross-sections for absorption to change. However, the net effect is that the neutron flux density must increase to sustain a constant power generation level as the core ages. As the flux density increases, the reactivity impact of the poisons in the core increases.
(4) 5.8 The additional step proposed goes beyond the scope of what is required for a full credit answer and will not be credited either for or against the candidate.
(5) 7.8 REJECT. The minimum operational steam inlet pressure for the turbine driven EFW pump is 60 psi which equates to a steam generator temperature of about 3000F. Unless the RCS becomes decoupled from the steam generate s, its temperature would also be about 3000F when steam pressure became too low to operate the EFW pump. If this were the case, the RCS cooldown that was to be avoided originally has already occurred. Finally, whether the EFW pump can be operated or not has no bearing on avoiding cooldown only in controlling cooldown.
ANO 2 EXAMINATION REPORT No. 50-386/0L-86-01 (6) 8.10 REJECT. If the statement is not true for all cases then it is false.
B. Exit Meeting Summary At the conclusion of the exam period, examiners met with !
representatives of the plant staff to discuss the results of the
. examinations. The following personnel were present for the exit interviews:
i First Week :
t NRC UTILITY if7 L. McCrory J. Vandergrift D. Graves W. Perks J. Whittemore S. Gulick ,
W. Johnson L. McClure C. Anderson NRC informed facility staff that all candidates were clear passes )
on the oral examinations. NRC reported that some candidates had difficulty determining when, during the course of accident recovery, it was possible to secure safety injection. Procedures provided little or no guidance addressing the situation.
Generally, all candidates performed well to exceptionally well on the oral examinations.
Second Week NRC UTILITY RT A. Cooley J. Vandergrift J. Pellet W. Perks C. Harbuck L. McClure A. Elliot S. Gulick C. Anderson NRC reported that all candidates were clear pass on the oral examinations and that there were not signficant generic weaknesses observed.
C. ANO 2 Examination Key Date Administered: December 10, 1985 Exam Type: Reactor Operator and Senior Reactor Operator
i-U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION Facility: ANO Unit 2 -
Reactor Type: CE-PWR Date Administered: 12/10/85 Examiner: S.L. McCrory Candidate: i
- INSTRUCTIONS TO CANDIDATE:
READ THE ATTACHED INSTRUCTION PAGE CAREFULLY. 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. ;
2 .
. % of 1
Category - % of Candidate's Category Value Total Score Value Category 25 25 1. Principles of Nuclear !
j Power Plant Operation,
. Thermodynamics, Heat Transfer and Fluid Flow Plant Design Including 25 25 2. !
Safety and Emergency
! Systems 25 25 3. Instruments and Controls 25 25 4. Procedures - Normal,
} Abnormal, Emergency, and Radiological Control -
-100 TOTALS Final Grade %
( All work done on this examination is my own. I have neither given nor received aid.
i Candidate's Signature i
l I
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i IEtC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
l 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave.
, You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
l 3. Use black ink or dark pencil only to facilitate legible reproductions.
! 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
! 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
! 8. Consecutively number each answer sheet, write "End of Category _" as e appropriate, start each category on a new page, write on only one side of the j paper, and write "Last Page" on the last answer sheet, j 9. Number each answer as to category and number, for example, 1.4, 6.3.
- 10. Skip at least three lines between each answer.
l 11. Separate answer sheets from pad and place finished answer sheets face down on j -your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the l question and can be used as a guide for the depth of answer required.
- 14. Show all calcualtions, methods, or assumptions used to obtain an answer to 4 mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND 00 NOT LEAVE ANY ANSWER BLANK.
l
- 16. If parts of the examination are not clear as to intent, ask. questions of the i examiner only.
! 17. You must sign the statement on the cover sheet that indicates that the work is
- your own and you have not received or been given assistance in completing the j examination. This must be done after the examination has been completed.
j 18. When you complete your exa tination, you shall:
. a. Assemble your examination as follows:
(1 Exam questions on top.
(2 Exam aids - figures, tables, etc.
Answer pages including figures which are a part of the answer.
) (3
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
i d. Leave the examination area, as defined by the examiner. If after leaving, i you are found in this area while the examination is still in progress, your i license may be denied or revoked.
i
- e. Do not dicuss the examination with other licensee staff personnel until the formal examination review is complete.
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERM 0 DYNAMICS, HEAT TRANSFER AND FLUID FLOW 1.1 Given the conditions below, calculate the time required for reactor vessel bulk water temperature to reach 2120F. STATE ALL ASSUMPTIONS AND SHOW ALL WORK FOR FULL CREDIT. (3.0)
Conditions:
- 1. The reactor has been shutdown for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter a 250 day run at 100%.
- 2. ALL means of removing heat from the vessel are lost.
- 3. No circulation to the primary loops occurs.
- 4. Reactor vessel water is initially at 1120F and well mixed.
- 5. The reactor vessed head is de-tensioned but still sealed.
ANS:
ASSUMPTIONS A. Vessel water volume = 3000 - 5000 cu. ft. (0.5)
B. Decay heat load = 0.1 - 0.5% rated thermal power (0.5)
C. Rated thermal power = 2750 - 2850 MW (0,5)
D. Water density = 59.8 - 61.8 lbm/cu.ft. (.25)
E. 57,000 Btu / min = 1 MW (.25)
F. 1 Btu will raise 1 lbm water 10F (.25)
SOLUTION G. Water Mass = AxD = 209,300 - 309,000 lbm (0.1)
H. Heat Load = 8xCxE = 2-15 MW = 1.14 - 8.55x105 Btu / min (0.1)
I. Heat Required = FxGx(dT=1000) = 21 - 31x106 Btu (0.1)
J. Time Required = I/H = 24 - 272 min = 0.4 - 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (.45)
KEY:
HTTRANS REF:
BASIC CE REACTOR DESIGN, STEAM TABLES, BASIC HT&T VAL:
as indicated in the answer.
1.2 A. What is the minimum reactivity that must be added to a critical reactor for it to be prompt critical? (1.0)
B. How much (by what factor) would power increase in one second at AND 2 if it were prompt critical? (2.0)
ANS:
A. p (reactivity) GE beta effective (beta value = 0.005 -0.007)
B. T = 1*/p + (B-p)/tp (0.5)
So for prompt critical neglect the delayed term so that T = 1*/p (.25) 1* = 10 10-5sec (.25) p = 0.005 - 0.007 (.25)
T = .0014 - 0.02 sec (.25)
P/Po = et/T = e(50 - 700)/1 sec (0.5)
KEY:
RXTH CORE 0PS REF.
BASIC REACTOR THEORY VAL:
1 pt for A, and as indicated for B.
1.3 A. Explain how neutron production and indicated count rate would change if the neutron sources were removed from the reactor while it was subcritical (Keff less than 1). (2.0)
B. Explain how long it would take to reach a steady-state count rate when Keff is increased from 0.990 to 0.999 if it took one minute to achieve a steady-state count rate when Kef was increased from 0.90 to 0.99. 1.0)
ANS:
A. The count rate would decrease to a small value since the reaction is not self-sustaining (1.0). Neutron production would not go to zero since spontaneous and cosmic fissions still occur, but the indicated count rate could be 0 due to instrument limitations (1.0).
B. Longer (0.7) - 10 minutes OR 10 times as long (0.3).
KEY:
RXTH NEUT REF:
BASIC REACTOR THEORY VAL:
As indicated in the answer.
i i
- _ . . _ _ . . _ _ _ . _ _ . . ... ____. __.~,___ _., . _
1.4 During a reactor startup, the operator stops regulating group CEAs at 144 steps on group 3. The source range count rate levels off at 1857 cps. The initial count rate was 400 cps at 0 steps withdrawn on regulating group 1, with Keff = 0.940. (1.5)
A. Calculate the 1/M value for this control position.
, B. What is the new value of Keff at this condition?
ANS:
A. 1/M = CR1 /CR2 = 400/1857 = 0.215 (The candidate may calculate the source term using CR =
S/(1-Keff) and then calculate 1/M using 1/M = l-Keff. This will generate a 1/M = .0129 based on a CR = 24. This is not
, an operationally meaningful value since Keff must = 0 before l the source level could be seen on the instruments. Give credit only if the candidate states that 1/M is based on source level.)
B. 1/M = 1 - Keff 2 1 - Ke ft 1 - Keff2= 1 0.940) x 0.215 = 0.9871 KEY:
COREOPS REAC REF:
0.5 pt for A 1 pt for B
1 1.5 Explain why a relief valve body-to-bonnet leak will produce superheated steam if it is located on a steam generator but will not if located on the pressurizer. Use the mollier diagram provided and assume approximately normal full power values for pressures.
(2.0)
ANS:
SEE ATTACHED MOLLIER DIAGRAM Leakage from a valve is an isenthalpic process. Using mollier diagram, assume S/G pressure of 1000 psia and primary of 2200 psia and draw a straight line (constant enthalpy) we find 800F superheat for 100 psia at 14.7 psia, but not when 14.7 psia is reached for 2200 psia.
KEY:
FLUID REF:
AN0 2 HTFFM CH 2 & 4 VAL:
1 pt ea for use of mollier and explaination l
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18 0 3 : : : :
ENTROPY
1.6 TRUE or FALSE: (1.5)
A. If the system temperature difference that is driving natural circulation flow is doubled, the heat removal rate will go up by slightly greater than a factor of two.
B. Viscosities of liquids decrease when the temperature increases.
C. If the pressure in a leaking pipe is reduced by 50%, the leak rate should be reduced by approximately 50%.
! ANS:
A. FALSE (dT3/2 g g)
B. TRUE C. FALSE (flow 2 e dp)
KEY:
FLUID HTTRANS REF:
- AN0 2 HTFFM CH 5 & 10 4
VAL:
0.5 ea
1.7 What four basic conditions must exist in order to create continuous flow due to natural circulation? (2.0)
ANS:
- 1. Heat source
- 2. Heat sink
- 3. Flow path (hydraulic coupling)
- 4. Elevation difference - heat sink above heat source KEY:
-FLUID HTTRANS REF:
AN0 2 HTFFM CH 9 VAL:
0.5 ea s-5 I
i L
1.8 Concerning RCP's NPSH, as system temperature is increased, why must system pressure be increased? (1.0)
ANS:
NPSH = Psys - Psat, or absolute pressure minus Psat. As temperature increases, Psat increases, to maintian positive NPSH, pressure must be increased.
KEY:
FLUIDS REF:
1 pt
1.9 Although the U238 resonance capture peaks broaden and flatten with increased fuel temperature, the area under the peak remains the same. Why then is there an increase in neutron capture as the fuel temperature is encreased? (1.0)
ANS:
The neutron population at resonance energy (neutrons capable of being absorbed) increases. This is due to the resonance encompassing a wider energy range (reduction in self shielding).
KEY:
NEUT RXTH REAC REF:
1 pt.
1.10 A. True or False? After a reactor trip, the reactor will reach a stable negative startup rate of about 0.33 dpm (80 sec period). (0.5)
B. How does the change in Beta-effective over core life affect the STABLE reactor period after a reactor trip? (0.5)
C. How does Beta-effective change over core life? (0.5)
D. How does the change in Beta-effective over core life affect the TRANSIENT reactor period after a reactor trip (before reaching a stable period)? (0.5)
ANS:
A. TRUE B. NONE C. Betae ff gets smaller D. Transient period is shorter (due to faster response with a smaller Beta).
KEY:
RXTH CORE 0PS REF:
BASIC REACTOR THEORY VAL:
0.5 each 6
r 1.11 A. HOW does equilibrium Xenon reactivity (XE-eq) at hot full power change as a function of core age (EFPD)? (0.5)
- 8. WHY does Xe-eq change as a function of core age? (1.5)
ANS:
A. Xe-eq gets larger as a function of core age.
B. Xe-eq is a function of flux not power (0.75) and flux increases as a function of core age (0.75).
KEY:
RXTH POIS0NS REF:
BASIC REACTOR THEORY VAL:
As indicated.
1.12 The ratio of Plutonium atoms to U 235 atoms increases as the core ages. Explain what affect the ratio has on the following: (3.0)
A. Delayed neutron fraction B. SUR C. Doppler defect ANS:
A. Delayed neutron fraction decreases because beta is less for plutonium.
B. SUR would be larger because beta decreased.
C. Doppler defect would increase because of more capture in Pu240, KEY:
CORE RXTH REF:
ANO 2 RTTM CH 13 VAL 3 pts, 1 pt ea END OF CATEGORY 1 i
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 2.1 A. Descrive the instrumentation available to the operator for monitoring the condition of of the RCP seals. (1.0)
B. What indications would you have if a coolant pump first stage seal failed? (1.0)
ANS:
A. 1. Pressure indication on each of the seal cavities.
- 2. High control bleed off flow alarm and indication
- 3. High control bleed off temperature alarn,and indication.
B. Mid-stage seal cavity pressure would go to RCS pressure, upper seal cavity pressure would increase proportionally.
Potential for C.B.0. to increase flow and temperature.
KEY:
RCS PUMP DESGN IND SEALS REF:
AN0 2 STM-2-03 VAL:
A. 0.333 ea B. 0.333 per line
2.2 A. What equipment (5 items) discharges to the Quench Tank?(1.5)
B. What three alarms are associated with the quench tank? (1.0)
C. If the Quench Tank level is too high, where can the operators discharge the excess water? (0.5)
ANS:
A. (ANY FIVE)
- 1. RCP seal water relief
- 2. Pressurizer safeties
- 5. Reactor vessel head vent
- 6. Pressurizer high point vent
- 7. Nitrogen addition
- 8. Reactor makeup water addition B. 1. Level hi/lo
- 2. Hi pressure
- 3. Hi temperature C. Reactor coolant drain tank KEY:
RCS TANK IND PATH REF:
AN0 2 STM-2-03, FSAR 5-5 VAL:
A. 0.3 ea B. 0.333 ea C. 0.5
2.3 What indication does the operator have that the reactor vessel head inner gasket is leaking? (1.0)
ANS:
Leakage is detected whenever the temperature in the leakoff line reaches GT 1500F and the RCS RV head leakoff alarm sounds (on 2K10,04).
KEY:
RCS IND SEALS REF:
ANO 2 STM-2-02 VAL:
0.5 ea for temp and alarm i
t 6
a
2.4 What five systems does the RWT provide water for? (1.5)
ANS: (ANY FIVE)
~2. Lp safety injection
- 3. Refueling canal fill
- 4. Fuel pool makeup
- 5. CVCS makeup (emergency boration)
- 6. HPSI
- 7. SITS NOTE: ECCS may be substituted for any one of HPSI, LPSI, or Containment Spray.
KEY:
MAKEUP TANK PATH REF:
ANO 2 STM-2-08 VAL:
0.3 ea
2.5 Identify the voltage by its corresponding letter designation.
Write the answers on the paper p.ovided for answers. (2.0)
A. B 1. 6900 VAC B. A 2. 4160 VAC C. LA 3. 480 VAC D. H 4. 120 VAC E. D 5. 120 Inst. AC F. RA 6. 125 VDC G. Y 7. 120 VAC RPS & ESF dist.
H. RS 8. 125 VDC RPS & ESF dist.
ANS:
A. 3 B. 2 C. 4 D. 1 E. 6 F. 8 G. 5 H. 7 KEY:
ELEDST REF:
ANO 2 STM-2-32 VAL:
0.25 ea
2.6 Describe the operation of and reason for the " Kirk" key interlock for the 2A310 and 2A410 breakers (2A3 and 2A4 cross-ties). (2.0)
ANS:
Operation - Insert 3 of 4 keys into panel to release 2A3/2A4 cross-tie keys as follows;
- 1. Key for supply breaker from 2A1
- 2. Key for supply breaker from 2A2
- 3. DG1 outupt breaker
- 4. DG2 output breaker Reason - Prevent paralleling of power supplies through the cross-i ties. (or words to that effect)
KEY:
ELEDST BKR REF:
ANO 2 STM-2-32 VAL:
A. 0.2 ea for leadin and each step B. 1 pt R
j
2.7 List by name six valves that are supplied by the electro-hydraulic oil system. (General types of valves, noo individual valve numbers). (1.5)
ANS: (ANYSIX)
- 1. Main feed pump control valves
- 2. Turbine control valves
- 3. Turbine stop valves
- 4. Extraction relay dump valve
- 5. Main feed pump stop valves
- 6. Turbine intercept stop valves
- 7. Turbine intercept valves KEY:
EHC COMP PATH VALVE REF:
ANO 2 STM-2-24 VAL:
0.25 ea
i 2.8 With the plant at 15% power during a plant startup, the operator is maintaining steam generator levels with all FWCS Hand / Auto stations in " manual". (2.0)
A. What will the response of the FWCS be if a reactor trip occurs?
B. What are two problems this could create if there were no operator action to correct?
ANS:
A. Feed water flow will remain constant.
B. 1. Result in overfeeding the S/G
- 2. Overcool the RCS KEY:
MFW CNTRL VALVE REF:
AN0 2 STM-2-69 VAL:
A. 1 pt B. 0.5 ea
2.9 List the three sources of water for EFW and state when each is i' used. (3.0)
ANS:
- 1. S/U and B/D Demin. effluent - low power (LT 10%)/ shutdown
- 2. Condensate storage tanks - normal (GT 10%)
- 3. Service Water System - Emergency KEY:
AFW PATH REF:
AN0 2 STM-2-19 VAL:
0.5 ea
t 2.10 TRUE or FALSE (2.5)
A. The refueling machine hoist can not be lowered with the bridga in motion (bridge drives energized).
B. The mast bumper interlock is only operable when the bridge is over the core region of the refueling canal.
C. If the spreader is not retracted, bridge motion is prevented.
D. The bridge will not move into the upender area unless the upender is vertical.
E. The " cable slack" interlock will stop hoist down motion.
ANS:
A. TRUE B. FALSE C. TRUE D. TRUE
- i E. TRUE KEY:
REFUEL REF:
l ANO 2 STM-2-51 VAL:
0.5 ea
2.11 Explain the purpose of the bypass lines with the pressure control valves in the CCW system. (1.0)
ANS Maintain constant preset backpressure in each loop with varying loads on the system.
KEY:
CCW VALVE REF:
AN0 2 STM-2-43 VAL:
1 pt
---.-- ,------ ---- --- - -,---y-- - - - ---
-- - - - - - - - , - -,- 1-,.-
2.12 Draw a sketch of the Reactor Coolant System. Include all penetrations, major components, major instrumentation tap, and loose parts monitor detectors. (3.5)
ANS:
SEE ATTACHED FIGURE The following items are 0.2 pt ea Reactor RCP's (0.05 ea)
S/G's (0.1 ea)
Pzr Spray valves (0.05 ea) and Aux spray (0.1)
RDT drains, sample points (0.022 ea)
SIS /SDC, SDC connections (0.04 ea)
Charging and letdown (0.067 ea)
Pzr code safties (0.1 ea)
Quench tank RTD's 0.021 ea Th 9 per loop Tc 3&3 per loop surge line relief 2 quench tank pzr interface Pressure detectors 0.05 ea pzr 4 quench tank Level detectors 0.083 ea pzr 2 quench tank Vibration and loose parts monitors 0.083 ea SG 1 ea Reactor vessel KEY:
RCS PATH DESGN REF:
AS INDICATED END OF CATEGORY 2
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. n ero's
' JU(GE CINE 2.TD QT p P.T RI PER INTESNE RTb E
TY '
' AUX SPRAY Reusp gro's L. QUENCH rfJ TANK PZR ->RTD I
a' ~?ihR. Pr I ."
' 4 m i RDT gj P te. Lr SPRAY NT U '
SAMPLE = VALVES QT N CHARGING gx 2P32A tha ( LP(Era o 2P32C LETDOWN RC jo 9 9- Wd [RC 46 R%\
9RTDs STEAM 4 i
%8 g [,.P.g. GENERATOR '
RX f
- STEAM 2E24A " "
GENERATOR Vsa. ( L.PM.
l 1 gfM i
- RDT l
~ y' w. 2P328 "
NI
' o 2P32D CHARGING SAMPLE
- RCS ONE-LINE DIAGRAM REACTOR COOLANT SYSTEM LP TITLE
> ';" '-- "'i' e
'i '.I . REACTOR COOLANT SY9 TEM FIG ( T.2 r
"'E I"E" l l S 111 3 l 3.2 $ A-R,l n l
1 l
l
- 3. INSTRUMENTS AND CONTROLS 3.1 List the 12 setpoints with their control / alarm functions associated with the pressurizer level control system. (3.0)
ANS:
13.2% Hi Hi level alarm 12.5% Hi Hi level alarm clear 11.1% Letdown Max.
4.5% Hi level alarm All heaters on >
Backup signal to stop all backup charging pumps 3.8% All charging pumps off Clears signal for all heaters on Hi level alarm clears
- 0.0% Level set (Not a control / alarm setpoint)
-1.4% Letdown to min.
Stop 1st B/U charging pump
-2.0% Stop 2nd B/U charging pump
-3.1% Start 1st charging pump
-4.2% Clear signal to start all charging pumps
-4.8% Start 2nd B/V charging pump
-5.2% B/U signal to start all charging pumps Lo level alarm Clear lo level alarm 29.0% Lo-Lo level alarm (actual level)
Heater cutout NOTE: Tolerance for setpoint + 0.5%
KEY:
PZR CNTRL REF:
ANO 2 STM-2-03 VAL:
0.1 ea for setpoint 0.15 ea for single functions, where a setpoint has multiple functions, divide 0.15 evenly amoung the number of functions.
l l
I i
l l
l
9 A
i 1
,- 3.2 During normal power operation, the two temperature sensing E
elements on the outlet of the letdown heat exchanger fail hi h.
What control functions occur? 2.0)
'ANS:
j heat exchanger will go full open (0.5).
- 2. 2TE-4805 fails high - this will isolate the letdown flow to the CVCS rad-monitors (0,5) and boronometer (0.5) and also '
, bypass letdown flow around the demineralizers to the VCT(0.5). (This feature protects these components from high temperature conditions which they are not designed to handle.) ;
KEY:
- CVCS .CNTRL A0P VALVE i REF
AN0 2 STM-2-4 i VAL:
AS INDICATED i
l
3.3 List six of the eight functions for which the 125 VDC Power System is designed to provide power. (2.0)
ANS: (ANY SIX)
- 1. 6.9 kV switchgear control
- 2. 4.16 kV switchgear control
- 3. 480V load control
- 4. Reactor control (control rod drive, reactor trip circuit breaker control)
- 5. Reactor instrumentation and protective system
- 6. Engineered Safeguards System
- 7. Inverters (vital 120 VAC)
- 8. Other equipment necessary for normal unit opertion, and normal and emergency shutdown. (any component operated with 125VDC)
KEY:
ELEDST CNTRL REF:
AN0 2 STM-2-32 VAL:
0.333 ea l'
3.4 In reference to the SDBCS quick-opening block signals: (1.5)
A. What are the two signals?
B. What valves are affected by each?
C. What is the purpose of the block signals?
ANS:
A. 1. Tave 10W
- 2. Reactor tripped B. 1. Reactor trip blocks 2 upstream and 1 downstream atmospheric dump valves.
- 2. Low Tave and reactor trip blocks remaining valves C. Prevents excessive cooling of the RCS KEY:
MNSTM VALVE CNTRL REF:
ANO 2 STM-2-23 VAL:
A. 0.25 ea B. 0.3 ea C. 0.4
1 l
l 3.5 If a CEA is stuck, what will give the most reliable indication of its position and why? (1.5)
ANS:
The reed switch position indication is the most reliable (0.5) because it uses a magnet on the CEA extension shaft to activate reed switches and show actual position (0.5). The pulse counter position indication only counts the pulses sent to the coils to change rod position and will continue to indicate changing rod position even if the rod is stuck (0.5).
KEY:
RODCNTRL IND REF:
AN0 2 STM-2-02 VAL:
AS INDICATED
3.6 A. List the three dedicated CPC indications provided in the control room (do not list the digital display or any alarm / status lights). (1.0)
B. List the three dedicated COLSS indications provided in the control room (i.e., items that are permanently displayed vice assigned points normally on display). (1.0)
C. Why is the COLSS system not used for the reactor protection system? (1.0)
D. Which is more accurate COLSS or CPC's? (0.5)
ANS:
A. 1. Margin to DNB
- 2. Calibrated neutron flux
- 3. Margin to LPD B. 1. Core power limit, based on margin to DNB
- 2. Core power limit, based on margin to LPD
- 3. Core power Note: Linear power KW/FT/DNBR limit meter (2J19042/9040) may be substituted for either 1 or 2 but not both.
C. Too slow for indicating actual plant conditions during transients.
D. COLSS KEY:
RPS IND DESGN REF:
ANO 2 STM-2-65,66 VAL:
A. 0.333 ea B. 0.333 ea C. 0.5 ea for "too slow" and " transients" D. 0.5
3.7 During full power operation, the CIAS inadvertently actuates.
A. What effect doe this have on continued plant operation and why? (1.5)
B. What is required to reset it? (1.0)
ANS:
A. CIAS actuation affects the operation of the plant by isolating the majority of auxiliary features fed into containment. Due to the requirements of CCW to cool the RCP's and seals (0.5), the plant will be tripped (0.5) and RCP's shutdown on receipt of a high temperature alarm from the RCP or after 5 minutes without CCW OR override CIAS for RCP CCW and restore flow to the pumps (0.5).
B. 1. Signal cleared
- 2. Reset PPS (trip path)
- 3. Reset ESFAS (actuation path)
KEY:
ESF A0P CNTRL REF:
ANO 2 STM-2-70 VAL:
A. AS INDICATED B. 0.33 ea i
3.8 On an SIAS condition, what basically happens to the Service Water / Auxiliary Cooling Water System? (8 items) (2.0)
ANS: (ANY EIGHT)
- 1. ACW supplies shut
- 2. SFPHX supplies shut
- 4. Cooling tower makeup M0V's shut (SW loop 1 ACW return isolation 2CV1543-1 and SW loop 2 ACW return isolation 2CV1542-2)
- 5. Return M0V's to emergency pong open
- 6. Return MOV's to lake shut
- 7. SW pumps receive start signal
Note: "SW returns shift to the pond" may be substituted for 5 and 6 both but not either singularly.
KEY:
ANO 2 STM-2-42 VAL:
0.25 ea
3.9 Describe the actions of the Feedwater Control System under the following conditions: (2.5)
A. High level everride B. Reactor trip override ANS:
A. 1. MFRV will shut
- 2. BFRV will shut-
- 3. Feed pump program uses higher signal - 0% flow demand from its FWCS (minimum speed) or flow demand from the other FWCS B. 1. Non-selected feed pump goes to minimum speed
- 2. MFRV shut
- 3. BFRV goes to 5% flow demand (enough to remove decay heat) (11-15% flow)
- 4. When flow signal less than 5%, auto returns to no override configuration.
KEY:
MFW CNTRL REF:
ANO 2 STM-2-69 VAL:
A. 0.333 ea B. 0.375 ea i
3.10 List the 11 valves in the CVCS which receive ES signals and indicate the type of ES signal that each valve is capable of receiving. Valves may be listed by name or number. (3.0)
ANS:
- 1. Letdown line isolation (2CV-4820-2) - SIAS
- 2. Regenerative Hx inlet (2CV-4821-1) - SIAS or CIS
~
- 3. Regenerative Hx outlet (2CV-4823-2) - CIS
- 4. CB0 isolation (2CV-4846-1) - CIS or SIAS
- 5. CB0 isolation (2CV-4847-1) - CIS or SIAS
- 6. VCT outlet (2CV-4873-1) - SIAS
- 7. BAMT B recirc (2CV-4915-2) - SIAS
- 8. BAMT A recirc (2CV-4903-2) - SIAS
- 9. BAMT to charging pump suction (2CV-4916-2) - SIAS
- 10. Boric acid gravity feed to charging pps (2CV-4920-1) - SIAS
- 11. Boric acid gravity feed to charging pps (2CV-4921-1) - SIAS KEY:
AN0 2 STM-2-04 VAL:
2,4, & 5 - 0.333 ea all others 0.25 ea
3.11 TRUE or FALSE: (1.5)
A. The CPC uses average Tc to generate calibrated neutron flux power.
B. The incore flux power can be read on the plant power digital meter.
C. As containment ambient temperature increased, pressurizer level indicates higher.
ANS:
A. FALSE B. FALSE C. TRUE KEY:
CPC NI PZR IND REF:
ANO 2 STM-2-65, 67, 70 VAL:
+
0.5 ea END OF CATEGORY 3
- - - , , + , - - - , - - - - - , - , - --- ,,w ---e,- ,--9 w y n ew - ~ - --- w-- y
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 4.1 For operation with Tave greater than or equal to 3000F, Technical Specifications require 2 independent ECCS subsystems OPERABLE.
What constitutes an OPERABLE ECCS subsystem? (2.0)
ANS:
- 1. One OPERABLE high-pressure safety injection pump
- 2. One OPERABLE low-pressure safety injection pump
- 3. An independent OPERABLE flow path (0.25) capable of taking suction from the refueling water tank (0.25) on a SIAS (0.25) and automatically transferring suction to the containment sump on a RAS (0.25).
KEY:
ANO 2 T.S. 3.5.2 VAL:
0.5 ea for 1 & 2 as indicated in 3
4.2 TRUE or FALSE: (2.0)
A. Positive manual reactivity additions by more than one method is acceptable below the point of adding heat.
B. All regualting CEA groups shall be withdawn in their prescribed sequence except for physics and surveillance tests.
C. The reactor operator can exceed 1.5 DPM startup rate only with the permission of the Shift Supervisor.
D. Criticality must be anticipated any time CEAs are being withdrawn or boron dilution operations are being performed.
ANS:
A. FALSE B. TRUE C. FALSE .
3 D. TRUE KEY:
N0P RODCNTRL LIMITS REF:
ANO 2 OP 2102.08 SEC 4 VAL:
0.5 ea k
4.3 Identify the procedural limits for the following plant operations: (2.0)
A. RCS cooldown rate above 225 0F.
B. Maximum pressurizer to RCS temperature differential.
C. Hydrogen concentration in the RCS during normal operations.
D. S/G levels during plant startup.
ANS A. Do not exceed 1000F/hr B. 2000F C. 25-50 cc/kg D. 60%
KEY:
N0P LMTS REF:
ANO 2 OP 2102.02 VAL:
0.5 ea b
s- -~ . , , -e - _. . . .- - -..- , - - . ,- -_, , , - - - ,--,-n -- ,
4.4 During plant heatup, the operator observes that RCS temperature goes from 3500F to 3620F over a six minute period. Explain why this does or does not violate Technical Specifications for heatup rate. (1.5)
ANS:
While the heatup rate observed is 1200F/hr, TS are not violated because the limit is 1000F in a one hour period and the time interval was only six minutes.
KEY:
TS LMTS NOP REF:
ANO 2 TS 3/4.4.9 VAL:
0.75 ea for TS limit and general discussion
4.5 During plant heatup, three graphs are to be plotted every 10 minutes (not to exceed 30 minutes). What are these graphs? (1.5)
ANS:
- 1. RCS pressure vs temperature
- 2. RCS temperature vs time
- 3. Pzr temperature vs time KEY:
NOP JOB LMTS REF:
ANO 2 OP 2102.02 VAL:
0.5 ea
4.6 During operation, you are informed by an auxiliary operator that the generatur transformer is running hotter than usual. List three possible reasons why a transformer may run hotter than normal and appropriate corrective actions for each case. (1.5)
ANS: (ANY THREE)
- 1. Hi current - reduce load
- 2. Fault - unit shutdown
- 3. Oil breakdown - unit shutdown
- 4. Hi ambient temp - reduce load
- 5. Fan failure - start standby fan / reduce load
- 6. Pump failure - start standby pump / reduce load KEY:
A0P ELEDST REF:
ANO 2 STM-2-32 VAL:
0.2 ea for reason 0.3 ea for correction action
4.7 Why should CEA withdrawal be in small frequent steps when operating above 50% reactor power? (1.0)
ANS:
Prevent fuel failure (0.333) due to larage local power density changes (0.333) in the vicinity of CEA fingertips (0.333).
KEY:
NOP CORE LMTS REF:
ANO 2 OP 2102.04 VAL:
AS INDICATED 1
0 r --+-- - c- 4 .-- ._e. .--,v_,, r,- ----, . ~ .- _ ,-f. __ ,_,--, -_,_.. __-__ _ _ _ ._ . _ ,.. ., ,, .__. . . _ , , . _ . . _ _ . _ . _ . . _ _
4.8 TRUE or FALSE: (3.0)
A. LD-50/50 and LD-100/30 pertain to the percentageof those individuals who will die after 30 days of receiving a chronic dose.
B. G.M. tubes make good instruments for setting dose rate.
C. Tritium is a strong gamma emitter.
D. Teletector will monitor beta radiation only on the lower three scales.
E. Instruments that operate in the ion chamber region are energy dependent and indicat dose rates in Rem /hr as opposed to exposure in Rad /hr.
F. The shorter the half life of an isotope, the more radioactive (unstable) it is.
ANS:
A. FALSE B. FALSE C. FALSE D. TRUE E. TRUE F. TRUE KEY:
RADCON JOB DET LMTS REF:
AN0 2 AA-52009-001 VAL:
0.5 ea
4.9 Give two reasons why the relative hazard from radiation is higher from internal sources than external sources. (2.0)
ANS: (ANY TW0)
- 1. No protection (shielding) from internal exposure
- 2. Particulate radiations (alpha & beta) have very high ionizing ability but travel only a short distance in tissue.
The damage done internally is localized to a small area around the source but a great deal of damage is done.
- 3. The different chemical characteristics of the radionuclides cause some of them to concentrate in certain body tissues (I-131 thyroids, Sr-90 bone, etc.).
- 4. Bone seeking isotopes chemically bond to bone tissue and stay there. If the half life is long, the bone gets exposed to a lot of radiation for a long time.
KEY:
RADCON JOB REF:
ANO 2 AA-52009-001 VAL:
1 pt ea 1
- . - , - - . . . r u .-- ..- . - --_-
a 1
i i 4.10 A. What are 3 indications or methods for detecting increased l radioactivity in the secondary side of a S/G7 (1.5) f
[ B. In an emergency, after a reactor trip and SIAS, ghat
, immediate action may be required due to decreasti:g RCS pressure? (0.5)
C. What two RCS parameters must be monitored to ensure adequate core heat removal immediately after a reacter trip? (1.0)
ANS:
! A. (ANY THREE)
- 1. Sample activity
- 2. Condenser off-gas activity
! 3. Sample cooler radiation
- 4. Main steam line radiation i 5. Increased secondary system radiation B. Secure RCP's
.C. RCS flow (forced or natural circ)
- Subcooling margin KEY:
E09 IND CNTRL l REF:
ANO 2 E0P 2202.01, pgs 2, 3, 10, 11 VAL:
0.5 ea i
i i
l 1
l i
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4.11 During blackout conditions, an Emergency Diesel Generator (EDG) starts but fails to pick up its associated ESF bus: (3.0)
A. What are four conditions that may have prevented the EDG output breaker from closing ~onto its assigned bus?
B. Why must the operator take action to quickly energize the bus or stop the diesel?
ANS:
A. (ANY FOUR)
- 1. Improper EDG voltage
- 2. Improper EDG speed
- 3. Normal feeder breaker not open
- 4. Cross-tie feeder breaker not open
- 5. Bus lockout relays picked up.
- 6. Loss of/no DC power to breaker
- 7. Breaker fault
- 8. DC lockout
- 9. Breaker racked down B. The bus must be energized to supply cooling water to the EDG to prevent damage.
KEY:
E0P EDG CNTRL REF:
ANO 2 E0P 2202.01, pg 35 VAL:
A. 0.5 ea B. 1 pt
4.12 During natural circulation cooldown. (2.5)
A. How will dT (Th -TC ) differ for a constant cooldown rate with a high versus a low decay heat load? Why?
B. Why is higher dT (T h-Tc) necessary at lower RCS temperature to maintain a constant natural circulation flow rate?
C. What are two reasons that the procedure forces RCS pressure reduction in steps while the RCS temperature is held relatively constant at 3500F?
ANS:
A Delta T will be higher for a higher heat load (0.5) to provide a higher flow rate due to a larger density difference (0.5).
B. At lower RCS temperature, a higher dT is required because the density change per OF decreases as temperature decreases.
C. 1. Use of Aux spray is minimized.
- 2. Makes voiding (Pressurizer level increase) easier to detect when Tave and pzr level are held steady.
KEY:
A0P LMTS IND REF:
ANO 2 A0P 2203.13, pgs 2, 3 VAL:
A. as indicated B. 0.5 pt C. 0.5 ea END OF CATEGORY 4 l
l
MC LICENSE EXAMINATION HANDOUT EQUATIONS. CONSTANTS, AND CONVERSIONS 6 = rii*Cp '*deltaT 6=U*A*deltaT P = Po*10sur*(t) p , po.,t/T SUR = 26/T T = 1*/p + (p-p)/I p T=1/(p-p) T = ($-p)/X p p = (Keff-1)/Keff = deltaXeff/Keff p=1*/TKeff+hff/(1+X.T) '
A = In2/tg = 0.693/tg X = 0.1 seconds-1 I = Io*e "*
CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lbm/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Stu/lb 1 atmosphere = 14.7 psia = g9.9 c inches Hg.
Miscellaneous Conversions 1 curie = 3.7 x 101U disintegrations per second 1 kilogram = 2.21 lbm I horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 105 Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Btu = 778 f t-lbf
U.S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION Facility: ANO Unit 2 Reactor Type: CE-PWR Date Administered: 12/10/85 Examiner: S.L. McCrory Candidate:
INSTRUCTIONS TO CANDIDATE:
READ THE ATTACHED INSTRUCTION PAGE CAREFULLY. Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up SIX (6) hours after the examination starts.
% of Category % of Candidate's Category Value Total Score Value Category 25 25 5. Theory of Nuclear Power Plant Operations, Fluids, and Thermodynamics 25 25 6. Plant Systems Design, Control and Instrumentation 25 25 7. Procedures - Normal, Abnormal, Emergency, and Radiological Control 25 25 8. Administrative Procedures Conditions, and Limitations 100 TOTALS Final Grade %
All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature 4
- -y -- - - - - - ~
MtC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
1
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category _" as appropriate, start each category on a new page, write on only one side of the i
paper, and write "Last Page" on the last answer sheet, i
- 9. Number each answer as to category and number, for example, 1.4, 6.3.
2
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
I 12. Use abbreviations only if they are commonly used in facility literature.
! 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calcualtions, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND
- 00 NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
i 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the
- examination. This must be done after the examination has been completed.
l 18. When you complete your examination, you shall:
! a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are a part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
! d. Leave the examination area, as defined by the examiner. If after leaving,
- you are found in this area while the examination is still in progress, your license may be denied or revoked.
l e. Do not dicuss the examination with other licensee staff personnel until the formal examination review is complete.
1
- 5. THE0RY OF NUCLEAR POWER PLANT OPERATIONS, FLUIDS, AND THERMODYNAMICS 5.1 Given the conditions below, calculate the time required for reactor vessel bulk water temperature to reach 2120F. STATE ALL ASSUMPTIONS AND SHOW ALL WORK FOR FULL CREDIT. (3.0)
Conditions:
- 1. The reactor has been shutdown for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after a 250 day run at 100%.
- 2. ALL means of removing heat from the vessel are lost.
- 3. No circulation to the primary loops occurs.
- 4. Reactor vessel water is initially at 1120F and well mixed.
- 5. The reactor vessed head is de-tensioned but still sealed.
ANS:
ASSUMPTIONS A. Vessel water volume = 3000 - 5000 cu. ft. (0.5)
B. Decay heat load = 0.1 - 0.5% rated thermal power (0.5)
C. Rated thermal power = 2750 - 2850 MW (0.5)
D. Water density = 59.8 - 61.8 lbm/cu.ft. (.25)
E. 57,000 Btu / min = 1 MW (.25)
F. 1 Btu will raise 1 lbm water 10F (.25)
SOLUTION G. Water Mass = AxD = 209,300 - 309,000 lbm (0.1)
H. Heat Load = BxCxE = 2-15 MW = 1.14 - 8.55x105 Btu / min (0.1)
I. Heat Required = FxGx(dT=1000) = 21 - 31x106 Btu (0.1)
J. Time Required = I/H = 24 - 272 min = 0.4 - 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (.45)
KEY:
HTTRANS REF:
BASIC CE REACTOR DESIGN, STEAM TABLES, BASIC HT&T VAL:
as indicated in the answer.
h
5.2 A. What is the minimum reactivity that must be added to a critical reactor for it to be prompt critical? (1.0)
B. How much (by what factor) would power increase in one second at AN0 2 if it were prompt critical? (2.0)
ANS:
A. p (reactivity) GE beta effective (beta value = 0.005 -0.007)
B. T = 1*/p + (B-p)/tp (0.5)
So for prompt critical neglect the delayed term so that T = 1*/p (.25) 1* = 10 10-5sec (.25) p = 0.005 - 0.007 (.25)
T = .0014 - 0.0g sec (.25)
P/Po = et/T = e(50 - 700)/1 sec (0.5)
KEY:
RXTH CORE 0PS REF:
BASIC REACTOR THEORY VAL:
1 pt for A, and as indicated for B.
5.3 A. Explain how neutron production and indicated count rate would change if the neutron sources were removed from the reactor while it was subtritical (Keff less than 1). (2.0)
B. Explain how long it would take to reach a steady-state count rate when Keff is increased from 0.990 to 0.999 if it took one minute to achieve a steady-state count rate when Ke f was increased from 0.90 to 0.99. 1.0)
ANS:
A. The count rate would decrease to a small value since the reaction is not self-sustaining (1.0). Neutron production would not go to zero since spontaneous and cosmic fissions still occur, but the indicated count rate could be 0 due to instrument limitations (1.0).
B. Longer (0.7) - 10 minutes OR 10 times as long (0.3).
KEY:
RXTH NEUT REF:
BASIC REACTOR THEORY VAL:
As indicated in the answer.
4
5.4 The ratio of Plutonium atoms to U235 atoms increases as the core ages. Explain what affect the ratio has on the following: (3.0)
A. Delayed neutron fraction B. SUR C. Doppler defect ANS:
A. Delayed neutron fraction decreases because beta is less for plutonium.
B. SUR would be larger because beta decreased.
C. Doppler defect would increase because of more capture in Pu240, KEY:
CORE RXTH REF:
3 pts, 1 pt ea l
I t
l l
1
5.5 Beta is the fraction of all neutrons released by fission which are delayed: (2.5)
A. When comparing the individual Beta's from thermal fission of U235, Pu239, and fast fission of U238, which is largest?
- 8. From BOL to E0L, does the average delayed neutron fraction increase, decrease, or remain the same? Explain.
C. Why is Betae ff less than Beta?
D. For equivalent positive reactivity additions to a critical reactor, will the SUR be larger or smaller at E0L compared to BOL?
ANS:
A. U238 B. Decreases - U235 goes down while Pu23990es up C. Beff = Bcore X IlX I. 2 14 and I2 are importance factors based on the relative ability of delayed neutrons to cause thermal fission (I I) and fast fission (1 2) the product of which is less than 1.
D. larger KEY:
NEUT RXTH REF:
AN0 2 RTTM CH 13 VAL:
0.5 pt ea for A, B, and D 1 pt for C
5.6 During a reactor startup, the operator stops regulating group CEAs at 144 steps on group 3. The source range count rate levels off at 1857 cps. The initial count rate was 400 cps at 0 steps withdrawn on regulating group 1, with Keff = 0.940. (1.5)
A. Calculate the 1/M value for this control position.
B. What is the new value of Keff at this condition?
ANS:
A. 1/M = CR1 /CR2 = 400/1857 = 0.215 (The candidate may calculate the source term using CR =
S/(1-Keff) and then calculate 1/M using 1/M = 1-Keff. This will generate a 1/M = .0129 based on a CR = 24. This is not an operationally meaningful value since Keff must = 0 before the source level could be seen on the instruments. Give credit only if the candidate states that 1/M is based on source level.)
- 8. 1/M = 1 - Keff 2 1 - Keff2 10.940) x 0.215 = 0.9871 KEY:
COREOPS REAC REF:
. 0.5 pt for A 1 pt for B
5.7 TRUE or FALSE: (3.0)
A. The boron coefficient (worth) is more negative at a high boron concentrations.
B. Xenon concentration decreases over core life but the xenon worth remains the same.
I C. One way to dampen a xenon oscillation is to insert a control rod into the region of the highest neutron flux.
D. The time it takes to achieve peak Xe conditions following a shutdown is independent of the initial equilibrium power level.
E. If the reactor period is cut in half, startup rate will be doubled.
1 F. The samarium worth never changes since it is flux independent.
ANS:
A. FALSE B. FALSE C. TRUE D. FALSE E. TRUE F. FALSE KEY:
POISION REAC REF:
0.5 pt ea
5.8 During cold shutdown conditions, the SIT's are filled and pressurized. Over ther next few days the plant is taken to 100%
power. Temperature in containment rises as does the temperature in the SIT's. Given the following information, find how much gas must be vented to maintain pressure. (1.0)
T1 = 680F T2 = 1000F Vtotal = 1850 ft3 V
PYate$15 psia
= 1480 ft3 P2 = 615 psia ANS:
P1 x V1 = P2 x V2 T1 T2 (PxV/T)1 = 615 x (1850-1480) = 431 460+68 V2 = 431 x (460 +100) = 392 615 392-370 = 22 ft3 to be vented off 4 KEY:
FLUID REF:
0.5 ea for formualtion and final answer i
l i l i
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l
5.9 Explain why a relief valve body-to-bonnet leak will produce superheated steam if it is located on a steam generator but will not if located on the pressurizer. Use the mollier diagram provided and assute approximately iiormal full pever values fcr pressures.
(2.0)
ANS:
SEE ATTACHED MOLLIER DIAGRAM Leakage from a valve is an isenthalpic process. Using mollier diagram, assume S/G pressure of 1000 psia and primary of 2200 psia and draw a straight line (constant enthalpy) we find 800F superheat for 100 psia at 14.7 psia, but not when 14.7 psia is reached for 2200 psia.
KEY:
FLUID REF:
AN0 2 HTFFM CH 2 & 4 VAL:
1 pt ea for use of mollier and explaination 4
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ENTROPY
5.10 A centrifugal pump is maintaining a flow rate of 1,000,000 lbm/hr and a discharge head of 200 psig at 1000 rpm with a power input of 300 Hp. If the pump head is doubled, what is the new flow rate, speed, and power. Show all work for full credit. (1.5)
ANS:
Pump laws:
V=kN h=kN2 KW=kN3 N increases by the square root of 2 Vf = Vi x (SR 2) = 106 x 1,414213 = 1,414,213 lbm/hr Nf = Ni x (SR 2) =103 x 1.414213 = 1,414 RPM Hpf = Hpi x (SR 2)3 = 300 x 2.828 = 848.5 Hp KEY:
FLUID REF:
0.5 ea i
5.11 As the core ages, the delta T from the fuel centerline to the coolant changes. This will subsequently change the fuel temperature at full power. Provide five (5) factors that change over the life of the core which affect the heat transfer ability and subsequent full power centerline temperature. (1.5)
ANS: (ANY FIVE)
- 1. Fuel densification
- 2. Fuel pellet swelling
- 3. Clad creep
- 4. Clad corrosion
- 5. Crud buildup
- 6. Gas in the gap
- 7. Thickness (size) of the gap
- 8. Fuel pellet thermal conducitivity KEY:
HTTRANS REF:
0.333 ea END OF CATEGORY 5
- 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 6.1 A. What equipment (5 items) discharges to the Quench Tank?(1.5)
B. What three alarms are associated with the quench tank? (1.0)
C. If the Quench Tank level is too high, where can the operators discharge the excess water? (0.5)
ANS:
A. (ANYFIVE)
- 1. RCP seal water relief
- 2. Pressurizer safeties
- 3. LTOP's
- 4. ECCS vent
- 5. Reactor vessel head vent
- 6. Pressurizer high point vent
- 7. Nitrogen addition
- 8. Reactor makeup water addition B. 1. Level hi/lo
- 2. Hi pressure
- 3. Hi temperature C. Reactor coolant drain tank KEY:
RCS TANK IND PATH REF:
A. 0.3 ea B. 0.333 ea C. 0.5
6.2 What indication does the operator have that the reactor vessel head inner gasket is leaking? (1.0)
ANS:
- 1. Leakage is detected by observing a rise in temperature on the leakoff line temperature indicator (2T15-4662).
- 2. Leakage is also detected by the RCS RV head leakoff alarm (on 2K10, D4 setpoint 1500F).
KEY:
RCS IND SEALS REF:
AND 2 STM-2-02 VAL:
0.5 ea for temp and alarm
6.3 What five systems does the RWT provide water for? (1.5)
ANS: (ANY FIVE)
- 2. Lp safety injection
- 3. Refueling canal fill
- 4. Fuel pool makeup
- 5. CVCS makeup (emergency boration)
- 6. HPSI
- 7. SITS NOTE: ECCS may be substituted for any one of HPSI, LPSI, or Containment Spray.
KEY:
MAKEUP TANK PATH REF:
ANO 2 STM-2-08 VAL:
0.3 ea
6.4 With the plant at 15% power during a plant startup, the operator is maintaining steam generator levels with all FWCS Hand / Auto stations in " manual". (2.0)
A. What will the response of the FWCS be if a reactor trip occurs?
B. What are two problems this could create if there were no operator action to correct?
ANS:
A. Feed water flow will remain constant.
B. 1. Result in overfeeding the S/G
- 2. Overcool the RCS KEY:
MFW CNTRL VALVE REF:
AN0 2 STM-2-69 VAL:
A. 1 pt B. 0.5 ea 1
6.5 List the three sources of water for EFW and state when each is used. (3.0)
ANS:
- 1. S/U and B/D Demin. effluent - low power (LT 10%)/ shutdown
- 2. Condensate storage tanks - normal (GT 10%)
- 3. Service Water System - Emergency KEY:
AFW PATH REF:
ANO 2 STM-2-19 VAL:
0.5 ea i
i l
L
r 6.6 Describe the four flow paths or all fluids tnat are involved with each reactor coolant pump. (2.5)
ANS:
- 1. CCW through the outer tube of the seal cooling heat exchanger
- 2. RCS through the pump impeller
- 3. RCS pumped by the auxiliary impeller through the inner tube of the seal cooler.
- 4. Controlled bleedoff coming around the seals via pressure-reducing orifices after having baeen cooled by the auxiliary impeler and seal cooler function.
KEY:
RCS PUMP DESGN REF:
ANO 2 STM-2-03 VAL:
0.5 ea for 1-3 1 pt for 4 i
l l
i
6.7 During normal power operation, the two temperature sensing elements on the outlet of the letdown heat exchanger fail hi h.
What control functions occur? 2.0)
ANS:
- 1. 2TE-4815 fails high - CCW flow controller valve from letdown heat exchanger will go full open (0.5).
- 2. 2TE-4805 fails high - this will isolate the letdown flow to the CVCS rad-monitors (0.5) and boronometer (0.5) and also bypass letdown flow around the demineralizers to the VCT(0.5). (This feature protects these components from high 1
temperature conditions which they are not designed to handle.)
KEY:
CVCS CNTRL A0P VALVE REF:
- AN0 2 STM-2-4 VAL
AS INDICATED (sensor designators not needed for full credit)
6.8 List six of the eight functions for which the 125 VDC Power System is designed to provide power. (2.0)
ANS: (ANY SIX)
- 1. 6.9 kV switchgear control
- 2. 4.16 kV switchgear control
- 3. 480V load control
- 4. Reactor control (control rod drive, reactor trip circuit breaker control)
- 5. Reactor instrumentation and protective system
- 6. Engineered Safeguards System
- 7. Inverters (vital 120 VAC)
- 8. Other equipment necessary for normal unit opertion, and normal and emergency shutdown. (any component operated with 125VDC)
KEY:
ELEDST CNTRL REF:
ANO 2 STM-2-32 VAL:
0.333 ea f
I
6.9 A. List the three dedicated CPC indications provided in the control room (do not list the digital display or any alarm / status lights). (1.0)
B. List the three dedicated COLSS indications provided in the control room (i.e., items that are permanently displayed vice assigned points normal on display). (1.0)
C. Why is the COLSS system not used for the reactor protection
- system? (1.0)
D. Which is more accurate COLSS or CPC's? (0.5) 1 ANS:
A. 1. Margin to DNB
- 2. Calibrated neutron flux
- 3. Margin to LPD
. B. 1. Core power limit, based on margin to DNB
- 2. Core power limit, based on margin to LPD
- 3. Core power Note: Linear power KW/FT/DNBR limit meter (2J19042/9040) may be substituted for either 1 or 2 but not both.
i C. Too slow for indicating actual plant conditions during transients.
D. COLSS I
KEY:
RPS IND DESGN REF:
l AN0 2 STM-2-65,66 VAL:
A. 0.333 ea B. 0.333 ea C. 0.5 ea for "too slow" and " transients" D. 0.5 4
f i
~ .__ . _.
6.10 During full power operation, the CIAS inadvertently actuates.
A. What effect doe this have on continued plant operation and why? (1.5)
B. What is required to reset it? (1,0)
ANS:
A. CIAS actuation affects the operation of the plant by isolating the majority of auxiliary features fed into containment. Due to the requirements of CCW to cool the RCP's and seals (0.5), the plant will be tripped (0.5) and RCP's shutdown on receipt of a high temperature alarm from the RCP or after 5 minutes without CCW OR override CIAS for RCP CCW and restore flow to the pumps (0.5).
B. 1. Signal cleared
- 2. Reset PPS (trip path)
- 3. Reset ESFAS (actuation path)
KEY:
ESF A0P CNTRL REF:
ANO 2 STM-2-70 VAL:
A. AS INDICATED B. 0.33 ea
6.11 On an SIAS condition, what basically happens to the Service Water / Auxiliary Cooling Water System? (8 items) (2.0)
ANS: (ANY EIGliT)
- 1. ACW supplies shut
- 2. SFPHX supplies shut
- 4. Cooling tower makeup M0V's shut (SW loop 1 ACW return isolation 2CV1543-1 and SW loop 2 ACW return isolation 2CV1542-2)
- 5. Return M0V's to emergency pong open
- 6. Return M0V's to lake shut
- 7. SW pumps receive start signal
Note: "SW returns shift to the pond" may be substituted for 5 and 6 both but not either singularly.
KEY:
ANO 2 STM-2-42 VAL:
0.25 ea END OF CATEGORY 6
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY, AND RADIOLOGICAL CONTROL 7.1 For operation with Tave greater than or equal to 3000F, Technical Specifications require 2 independent ECCS subsystems OPERABLE.
What constitutes an OPERABLE ECCS subsystem? (2.0)
ANS:
- 1. One OPERABLE high-pressure safety injection pump
- 2. One OPERABLE low-pressure safety injection pump
- 3. An independent OPERABLE flow path (0.25) capable of taking suction from the refueling water tank (0.25) on a SIAS (0.25) and automatically transferring suction to the containment sump on a RAS (0.25).
KEY:
ANO 2 T.S. 3.5.2 VAL:
0.5 ea for 1 & 2 as indicated in 3 t
7.2 During plant heatup, the operator observes that RCS temperature goes from 3500F to 3620F over a six minute period. Explain why this does or does not violate Technical Specifications for heatup rate. (1.5)
ANS:
While the heatup rate observed is 1200F/hr, TS are not violated because the limit is 1000F in a one hour period and the time interval was only six minutes.
KEY:
TS LMTS NOP REF:
ANO 2 TS 3/4.4.9 VAL:
0.75 ea for TS limit and general discussion
7.3 During plant heatup, three graphs are to be plotted every 10 minutes (not to exceed 30 minutes). What are these graphs? (1.5)
ANS:
- 1. RCS pressure vs temperature
- 2. RCS temperature vs time
- 3. Pzr temperature vs time KEY:
N0P JOB LMTS REF:
ANO 2 OP 2102.02 VAL:
0.5 ea
,4 k_.
7.4 During operation, you are informed by an auxiliary operator that the generator transformer is running hotter than usual. List three possible reasons why a transformer may run hotter than normal and appropriate corrective actions for each case. (1.5)
ANS: (ANY THREE)
- 1. Hi current - reduce load
- 2. Fault - unit shutdown
- 3. Oil breakdown - unit shutdown
- 4. Hi ambient temp - reduce load
- 5. Fan failure - start standby fan / reduce load
- 6. Pump failure - start standby pump / reduce load KEY:
A0P ELEDST REF:
AN0 2 STM-2-32 VAL:
0.2 ea for reason 0.3 ea for correction action 1
7.5 Why should CEA withdrawal be in small frequent steps when operating above 50% reactor power? (1.0)
ANS:
Prevent fuel failure (0.333) due to larage local power density changes (0.333) in the vicinity of CEA fingertips (0.333).
KEY:
N0P CORE LMTS REF:
ANO 2 OP 2102.04 VAL:
AS INDICATED 1
7.6 TRUE or FALSE: (3.0)
A. LD-50/50 and LD-100/30 pertain to the percentageof those individuals who will die after 30 days of receiving a chronic dose. -
B. G.M. tubes make good instruments for setting dose rate.
C. Tritium is a strong gamma emitter.
D. Teletector will monitor beta radiation only on the lower three scales.
E. Instruments that operate in the ion chamber region are energy dependent and indicat dose rates in Rem /hr as opposed to exposure in Rad /hr.
F. The shorter the half life of an isotope, the more radioactive (unstable) it is.
ANS:
A. FALSE B. FALSE C. FALSE D. TRUE E. TRUE F. TRUE KEY:
RADCON JOB DET LMTS REF:
AND 2 AA-52009-001 VAL:
0.5 ea i
7.7 A. What are 3 indications or methods for detecting increased radioactivity in the secondary side of a S/G? (1.5)
B. In an emergency, after a reactor trip and SIAS, what imediate action may be required due to decreasing RCS pressure? (0.5)
C. What two RCS parameters must be monitored to ensure adequate core heat removal immediately after a reactor trip? (1.0)
ANS:
A. (ANY THREE)
- 1. Sample activity
- 2. Condenser off-gas activity
- 3. Sample cooler radiation
- 4. Main steam line radiation
Subcooling margin KEY:
E0P IND CNTRL REF:
ANO 2 E0P 2202.01, pgs 2, 3, 10, 11 VAL:
0.5 ea
7.8 A. During recovery from blackout conditions the operator is directed to avoid any RCS cooldown. What are three reasons for this? (1.5)
- 8. During blackout what are four suggested actions to limit the drain on station batteries? (2.0)
ANS:
A. 1. No makeup available
- 2. No way to borate (SDM)
- 3. No way to maintain RCS pressure (voiding head)
- 8. (ANYFOUR)
- 1. Vent generator H2 and stop DC seal oil pump.
- 3. Secure Main turbine DC powered lube oil pump
- 4. Secure plant computer if not required
- 5. Secure any unncessary DC lighting KEY:
E0P ELEDST CtHRL REF:
AN0 2 E0P 2202.01, pgs 56, 57 VAL:
0.5 ea i
7.9 During natural circulation cooldown: (2.5)
A. How will dT (Th -TC ) differ for a constant cooldown rate with a high versus a low decay heat load? Why?
B. Why is higher dT (T h-Tc) necessary at lower RCS temperature to maintain a constant natural circulation flow rate?
C. What are two reasons that the procedure forces RCS pressure reduction in steps while the RCS temperature is held relatively constant at 3500F7 ANS:
A Delta T will be higher for a higher heat load (0.5) to provide a higher flow rate due to a lt <" density difference (0,5).
B. At lower RCS temperature, a higher dT is required because the density change per OF decreases as temperature decreases.
C. 1. Use of Aux spray is minimized.
- 2. Makes voiding (Pressurizer level increase) easier to detect when Tave and pzr level are held steady.
KEY:
A0P LMTS IND REF:
ANO 2 A0P 2203.13, pgs 2, 3 VAL:
A. as indicated B. 0.5 pt C. 0.5 ea l
7.10 A. What is your 10CFR20 quarterly radiation exposure limit:
- 1. without form 47
- 2. with form 47 (1.0)
B. What is your AN0 recommended weekly exposure limit:
- 1. without form 4?
2, with form 47 (1.0)
ANS:
A. 1. 1.25 rem /qtr
- 2. 3 rem /qtr B. 1. 100 mrem / week
- 2. 300 mrem / week KEY:
RADCON JOB REF:
AN0 2 AA-52009-001 VAL:
0.5 ea i
i 7.11 Which of the following items are addressed in Technical Specifications? Indicate YES or N0 for each item. (2.5)
A. Fuel Temperature Coefficient B. Boration system heat tracing C. Wind Direction Instrument D. Lake Dardanelle Thermal Gradient E. Chlorine Detection System F. Loose stone (riprap) around the Emergency Cooling Pond G. Main Condenser Evacuation System H. 125 VDC Battery Terminal corrosion I. Quick open response time for Steam Dump Bypass Valves J. Spent Resin Tank curie content ANS:
A. N0 B. YES - 3.1.2.8 C. YES - 3.3.3.4 D. NO E. YES - 3.3.3.7 F. YES - 4.7.4.1 G. N0 H. YES - 4.8.2.3.1 I. NO J. NO REF:
ANO II TS AS INDICATED Question value is 2.5 pts, 0.25 per item.
L'
7.12 Explain how core exit thermocouples can provide indication o core uncovery.
ANS:
If indicated temperature is greater than Tsat for that pressure (then the core exit TC's will be indicating a superheated temperature indicating partial core uncovery).
KEY:
E0P CORE IND REF:
AN0 2 E0P's VAL:
1 pt END OF CATEGORY 7 L
- 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 8.1 TRUE or FALSE: (3.5)
A. A Shift Superviosr's responsibilities include the authorization of the initiation and restoration of temporary modifications prior to such acitivities taking place.
B. A 17-year old has a 5(N-18) = 0 accumulated exposure limit.
Therefore, his dose is limited to 5% of the basic permissible of 1000 mrem per quarter.
C. If a power-operated valve is to be tagged with a hold card, both the power supply and handwheel shall be tagged.
D. A shift administrative assistant is not allowed to fill out the Hold Card Request Sheet and Hold Cards.
E. Fuel handling operations may include personnel who are unlicensed.
F. With one pressurizer code safety valve operable, reactor power is limited to less than or equal to 25%.
G. When RCS temperature is less than 2000F, all reactor coolant pumps and shutdown cooling pumps may be de-energized for one hour.
ANS:
A. TRUE B. FALSE C. TRUE D. FALSE E. TRUE F. FALSE G. TRUE i
KEY:
TS N0P CNTRL LMTS RADCON REF:
AN0 2 OP's 1000.27, 1000.28, 1015.07, 2502.01 TS 3.4.3, 3.4.1.3 VAL:
- 0.5 ea l
8.2 A. During cold shutdown (mode 5), what is the minimum shift complement? (1.0)
B. List the personnel required to carry out fuel handling operations (include transferring from Aux. Bldg. to Reactor Bldg. in manual). (1.5)
ANS:
A. 1 each of the following:
SR0 R0 Non-licensed operator HP Tech B. 1 each of the following:
CR0 (Control Room Operator)
Spent fuel bridge operator (licensed)
Operator for upender Aux. Bldg.
Reactor fuel bridge operator (licensed)
Operator for upender in Reactor Bldg.
SR0 at the Reactor fuel bridge area (in charge of fuel ops)
KEY:
N0P TS JOB REF:
ANO 2 OP 1015.01 VAL:
0.25 ea t
8.3 What four conditions require an independent review of tagout lineup and installation? (2.0)
ANS:
Independent verification of all tagouts is required.
KEY:
TAG J08 REF:
AN0 2 OP 1000.27 VAL:
0.5 ea
8.4 For each of the following chemical analyses of the RCS, indicate:
- 1. Whether or not the concentration is within limits.
- 2. How to correct unsatisfactory conditions (i.e., outside normal operational limits).
- 3. The possible consequences of continued operation with the existing condition. (3.0)
A. Cl - 0.2 PPM F1 - less than 0.1 PPM 0
(2 - 0.1 PPMreactor power - 501.)
B. Cl - less than 0.1 PPM F1 - less than 0.1 PPM 0
(2 - 0.2atPPM reactor standby)
ANS:
A. 1. Cl - w/i transient limit F1 - sat 02 - sat
- 2. Increase.CVCS flow rate (letdown)
- 3. Cl stress corrosion (time dependent)
B. 1. Cl - sat F1 - sat ,
02 - w/1 transient limit
- 2. Increase H2 overpressure in VCT
- 3. 02 stress corrosion (general oxidation)
Reduction of protective oxide film (time dependent)
KEY:
TS LMTS REF:
ANO 2 TS 3/4.4.7 VAL:
0.5 each for the six subsets.
8.5 What is the basis for maintaining 23 feet of water above the core during refueling operations? (1.0)
ANS:
Ensures sufficient water depth is available to remove 99% of the assumed (10%) iodine gap activity (0.5) released from the rupture of an irradiated fuel assembly (0.5).
OR Ensures sufficient heat sink to allow time to shift to alternate shutdown cooling or initiate emergency cooling to the core on loss of the operating shutdown cooling loop.
KEY:
TS RADCON REF:
ANO 2 TS 3/4.9.9, Basis 3/4.9.8 VAL:
AS INDICATED 1
. . . . - . _ . - . -. . . . - . . .. . _ . - _ _ _ . - _ = .
8.6 What is the basis for the Technical Specification limit on primary coolant specific activity? (1.5)
ANS:
Ensures that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of part 100 limits i' following a S/G tube rupture accident in conjunction with an assumed steady state primary-to-secondary leak rate of 1.0 gpm y and a concurrent loss of offsite electrical power.
i 8
KEY:
TS RADCON LMTS RCS 4 REF:
j ANO 2 TS 3/4.4.8 VAL:
0.25 for each underlined section i
t i
k i
I 1
The following scenerio applies to questions 8.7 through 8.9. Use the Technical Specification extract provided to answer these questions.
The reactor has been operating for more than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at 100% power with all rods out. The core age is 100 EFPD. Operator error results in an inadvertant reactor trip with no actuation of ESF systems. Four hours after the trip reactor recovery has proceeded to the point that group 6 rods are at 134" when a minor CEA deviation alarm is received and the reactor operator notices that one of the group 6 rods is still at 130".
8.7 At this time would the rod be considered OPERABLE or IN0PERABLE?
Explain. (1.5)
ANS:
The rod is OPERABLE. As long as the deviation is less than 7",
the LC0 is met and the rod may be considered OPERABLE until further investigation shows otherwise.
KEY:
TS REF:
ANO 2 TS 3.1.3.1 VAL:
0.5 pt for OP. or INOP.
1 pt for explanation l
l l
l
8.8 Attempts to move the affected rod individually are unsuccessful and the rest of the group 6 rods are positioned at 130".
With the information available up to this point, select the Technical Specification which most adequately covers the known condition and justify your selection. (To get credit, the reference must be given to the smallest subparagraph division (Ex: 3.4.1.2.a.2.b)). (1.5)
ANS:
TS 3.1.3.1 At this time nothing is known about the rod except that it will not respond to in/out signals from the control panel. A0P 2203.03, CEA Malfunctions, states that a rod is not declared to be IN0PERABLE until it is known to be either untrippable or immovable. Therefore, no lI0s have been violated at this time.
KEY:
TS RODCNTRL REF:
ANO 2 TS 3.1.3.1 and NRC precedent VAL:
0.5 pt for TS and 1 pt for justification.
i L
8.9 The fault is an open in the lift coils for the affected rod.
A. What, if any, alarms or annunciators would have been activated in the control room which are specific to the fault? (1.0)
B. With this information, what is the least restrictive appropriate Technical Specification statement? Justify your selection. (1.5)
C. With this condition, is the plant permitted by Technical Specifications to be operated at full power for an unspecified period of time? (0.5)
ANS:
A. None. There no alarms or annunciators which would specifically indicate an open in the lift coil of a rod.
(This can only be determined by measuring the current across a shunt in the power line to each coil.)
B. 3.1.3.1.c - Since an electrical fault has been identified which would not have resulted from mechanical binding nor would interfere with tripping the rod, and alignment and insertion are in limits, this is the appropriate TS.
C. Yes KEY:
TS RODCNTRL IND REF:
ANO 2 TS 3.1.3.3 VAL:
A. - 1 pt B. - 0.5pt for TS and 1 pt for justify C. - 0.5
8.10 Which of the following conditions or events would result in an Emergency Action Level classification of " ALERT" or higher?
Indicate yes or no for each one. (4.0)
A. Site experiencing straight winds greater than 75 mph.
B. Evacuation of the control room anticipated with control of shutdown systems established from local stations.
C. RCS leakage greater than 44 gpm but less than 100 gpm.
D. Area Radiation Monitors in the Reactor Building reading 2,500 mR/hr with no explanation.
E. Steam line break with 12 gpm primary to secondary leakage.
F. Dose equivalent I-131 of the RCS is 350 uCi/ml, not due to spiking.
G. Loss of all annunciator power indefinitely.
H. Loss of both shutdown cooling trains with anticipation of recovering at least one train in about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
ANS:
A. N0 B. YES C. NO D. YES E. NO F. YES G. YES H. YES KEY:
EPLAN REF:
4 pts, 0.5 each l
I
8.11 A. What personnel make up the on-site fire brigade? (Give number of people from each source and the source.) (1.5)
ANS:
2 from affected unit 1 from unaffected unit 2 from security KEY:
JOB REF:
ANO 2 OP 1015.07 VAL:
0.5 ea END OF CATEGORY 8
l NRC LICENSE EXAMINATION HANDOUT EQUATIONS, CONSTANTS AND CONVERSIONS f
6=iii*C'*deltaT p 6=U*A*deltaT p . p,*1o sur*(t) P = P *et /T SUR = 26/T T=1*/p+(p-p)/Ip T=1/(p-$) T = ($-p)/X p p = (Keff-1)/Keff = deltaKeff/Keff p=1*/TXeff+hff/(1+1T)
- A = In2/tg = 0.693/tg X = 0.1 seconds-1 I = Io*e "*
CR = S/(1-Keff) 2 R/hr = 6*CE/d feet Water Parameters 1 gallon = 8.345 lbm = 3.87 liters 1 ft3 = 7.48 gallons Density 9 STP = 62.4 lb,/ft3 = 1 gm/cm3 Heat of vaporization = 970 Btu /lbm Heat of fusion = 144 Btu /lba 1 atmosphere = 14.7 psia = 29.9 inches Hg.
Miscellaneous Conversions 1 curie = 3.7 x 10m disintegrations per second 1 kilogram = 2.21 lba 1 horsepower = 2 54 x 103 Btu /hr 1 mw = 3.41 x 10E Btu /hr 1 inch = 2.54 centimeters degrees F = 9/5 degrees C + 32 degrees C = 5/9 (degrees F - 32) 1 Stu = 778 ft-lbf
ERC-1853 12/12/85 AN0 2 EXAMINATION REPORT ARKANSAS NUCLEAR ONI UNIT TWO REACTOR OPERATOR EXAM OF 12/10/85 EXAM COMMEhTS Attachment 1 i .- _ - . _ - - - _ - - . . _ . _ - . -
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 1.4 During a reactor startup, the operator stops regulating group CEAs at 144 steps on group 3. The source range count rate levels off at 1857 cps. The initial count rate was 400 cps at 0 steps withdrawn on regulating group 1, with Keff = 0.940 (1.5)
- a. Calculate the 1/M value for this control position
- b. What is the new value of Keff at this condition?
ANS:
- a. 1/M = CR /CR2 = 400/1857 = 0.215 i
- b. 1/M = 1 - Keff2 /1 - Kefft 1 - Keff2 = (1 - 0.940) x 0.215 = 0.9871 REF:
ANO-2 RTTM Chpt. 15 REQUESTED CHAl[GE
- Key answer is correct if you assume the source counts are 400.
However, since Keff was already .94 and the new Keff was .9871, then 1/M = 1 - K, would result in 1/M = .0129 Either answer .215 or .0129 should be accepted.
1
- DESIGNATES REQUESTED CHANGE 1 Attachment 1 I
ERC-1853 12/12/85 ANO 2 EXAMINATION REPORT 1.6 True or False
- a. If the system temperature difference that is driving natural circulation flow is doubled, the heat removal rate will go up by slightly greater than a factor of two.
ANS:
- True 3/2 Q a AT IfATincreasesbyafactorofbQwillincreasebyafactorof23/2 or 2.8.
This should be considered "slightly" greater than 2 and answered True.
REF:
Heat Transfer Handbook Chapter 9 page 135
- DESIGNATES REQUESTED CHANGE 1
i l
2 Attachment 1
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 1.7 What four basic conditions must exist in order to create continuous flow due to natural circulation?
ANS:
- 1) Heat source
- 2) Heat sink
- 3) Flow path (hydraulic coupling)
- 4) Elevation difference (heat sink above heat source)
REF:
Heat Transfer Handbook Chapter 9 Page 133
- DESIGNATES REQUESTED CIIANGE I
t 3 Attachment 1
. _ . - _ _ . _ , - - , .- , y
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 1.9 Although the U2 as resonance capture peaks broaden and flatten with increased fuel temperature, the area under the peak remains the same.
Why then is there an increase in neutron capture as the fuel temperature is increased? (1.0)
ANS:
- Answer acceptable. Another answer should be "The reduction in self shielding" REF:
ANO 2 RTTM Chapter 17
- DESIGNATES REQUESTED CHANGE 4 Attachment 1
ERC-1853 ANO 2 EXAMINATf0N REPORT 12/12/85 1.11 A. HOW does equilibrium Xenon reactivity (KE-eq) at hot full power change as a function of core age (EFPD)? (0.5)
B. WHY does Xe-eq change as a function of core age?
ANS:
A. Xe-eq gets larger as a function of core age.
B. Xe-eq is a function of flux not power (0.75) and flux increases as a function of core age. (0.75)
- B. Question not clear. If Xe-eq reactivity worth is asked for, then it is larger because of reduced competition from boron. If concentration is being asked for, then the concentration decreases because of increased burnout.
- DESIGNATES REQUESTED CHANGE 5 Attachment 1
ERC-1853 ANO 2 EXAMINATION REPORT 12/?2/85 2.2A What equipment (5 items) discharges to the Quench Tank? (1.5)
'ANS:
A. 1) RCP Seal Water Relief (RCP Control Bleedoff Relief)
- 2) Pressurizer Safeties
- 3) LTOPs
- 4) ECCS Vent
- 5) Reactor Vessel Head Vent
- 6) Pressurizer High Point Vent
- 7) Nitrogen Addition
- 8) Reactor Makeup Water Addition REF:
P&ID M-2230 H1
- DESIGNATES REQUESTED CHANGES / ADDITIONS l
6 Attachment 1 I
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 2.3 What indication does the operator have that the reactor vessel head inner gasket is leaking? (1.0)
ANS:
- 1) Leakage is detected by observing rise in temperature on leakoff line temperature indicator. (2 TIS-4662)
- 2) Leakage also detected by RCS RV head leakoff alarm (set point 150 F on 2K10, D4).
- Question did not ask for alarm setpoint. .5 for indication .5 for alarm REF:
P&ID M-2230
- DESIGNATES REQUESTED CHANGE / ADDITION 7 Attachment 1
f.RC-1853 ANO 2 EXAMINATION REPORT 12/12/85 2.4 What five systems does the RWT provide water for? (1.5)
ANS:
- 2) LP safety injection
- 3) Refueling canal
- 4) Fuel pool makeup
- 5) CVCS makeup (emergency boration)
- 6) IIPSI
- 7) SITS
- ECCS may be listed instead of IIPSI, LPSI, and Containment Spray REF:
P&ID M-2231 & M-2236
- DESIGNATES REQUESTED CHANGE / ADDITION l
l 8 Attachment 1 t.
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 3.1 List the 12 setpoints with their control / alarm functions associated with the pressurizer level control system.
ANS:
13.2% HI-HI Level Alarm 12.5% HI-HI Level Alarm clear 11.1% Letdown max.
- 4.5% HI Level Alarm, all heaters on, backup signal to stop all backup charging pumps 3.8% Clear signal to turn on all heaters, HI Level Alarm clears
- 0.0% Level setpoint (no associated contro/ alarm functions)
-1.4% Letdown to minimum, stop 1st B/U charging pump
- -2.0% Stop 2nd B/U charging pump
-3.1% Start 1st B/U charging pump
-4.2% Clear backup signal to start all backup charging pumps LOW Level Alarm clears
-4.8% Start 2nd B/U charging pump
-5.2% B/U signal to start all charging pumps, LOW Level Alarm
- 29.0% (Actual Level) LO-LO Level Alarm & Heater Cutout
- Request change tolerance for setpoint to i 1.0%. Understanding of system response if more important than actual setpoint.
REF:
E-2704
- DESIGNATES REQUESTED CHANGE / ADDITION I
l I
l l
i l
l l
9 Attachment 1 L
ERC-1853 ANO 2 EXAM 1 NATION REPORT 12/12/85 3.2 During normal power operation, the two temperature sensing elements on the outlet of the letdown heat exchanger fail high. What control functions occur? (2.0)
ANS:
- DESIGNATES REQUESTED CHANGE 4
i 10 Attachment 1
ERC-1853 AN0 2 E X AMINATION REPORT 12/12/85 3.3 List six of the eight functior.s for which the 125VDC power system is designed to provide power.
ANS:
- 1. 6.9 KV Switchgear Control
- 2. 4.16 KV Switchgear Control
- 3. 480V Load Center Control
- 4. Reactor Control (Control Rod Device, Reactor Trip Circuit Breaker Control)
- 5. Reactor Inst. and Protective System
- 6. Engineered Safeguards System
- 7. Inverters (Vital 120VAC)
- 8. Other equipment necessary for normal unit operation, and normal and emergency shutdown. (Any component operated with 125VDC)
REF:
ANO 2 STM-2-32
- DESIGNATES REQUESTED CHANGE 11 Attachment 1
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 3.6 List the three dedicated COLSS indications provided in the Control Room (i.e. , items that are permanently displayed, vice assigned points normally on display)
ANS:
- Agree with key, however, student may have listed the core power limit based on margin to DNB or LPD AS Linear Power KW/FT/DNBR limit meter (2J19042/9040). This is the way it is labeled on Panel 2003.
REF:
Plant Panel 2CO3
- DESIGNATES REQUESTED CHANGE / ADDITION 12 Attachment 1
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 3.7 During full power operation, the CIAS inadvertently actuates.
A. What effect does this have on continued plant operation and why?
B. What is required to reset it?
ANS:
- A. Agree with key, however, we need to add capability to override and restore CCW to RCPs (as per DCP 84-2058)
B. 1) Signal cleared
- 2) Reset PPS (trip path)
- 3) Reset ESFAS (actuation path)
- 4) Reset components
- 4) " Reset Component" is not needed to reset a CIAS.
REF:
DCP 84-2058 OP2202.01 B.2 Page 70 of 138
- DESIGNATES REQUESTED CHANGE / ADDITION i
13 Attachment 1
-4 I
ERC-1853 ANO 2 EXRMINATION REPORT 12/12/85 3.8 On an SIAS condition, what basically happens to the Service Water /
Auxiliary Cooling Water System? (8 items)
ANS:
- 1) ACW supplies shut
- 2) SFPHX supplies shut i
- 4) Cooling tower makeup MOVs shut (the procedure names these however they are labeled
- a. SW Loop 2 ACW Return Isolation 2CV1542-2
- b. SW Loop 1 ACW Return Isolation 2CV1543-1)
- 5. Return MOVs to emergency pond oren
- 6) Return MOVs to lake shut
- 7) SW pumps receive start signal
An interlock will then open the SW valves associated with the equipment.
- #5 & #6 can be covered by saying that SW returns shift to the pond REF:
P&ID M2210 Sheet 1-3
- DESIGNATES REQUESTED CHANGE / ADDITION 14 Attachment 1
ERC-1853 AN0 2 EXAMINATION REPORT 12/12/85 3.9 Describe the actions of the Feedwater Control System under the following conditions:
A. High Level Override B. Reactor Trip Override ANS:
A. 1) MFRV will shut
- 2) BFRV will shut
- 3) Feed pump program uses higher signal - 0% flow demand from ITS FWCS (minimum speed) or flow demand from other FWCS
- B. 1) Selected MFP trips on turbine trip (this function not due to RTO occurs due to turbine trip)
- 2) Feed pump goes to minimum speed
- 3) MFRV shuts
- 4) BFRV goes to a position corresponding to a 5% flow demand signal (actual valve position ~ 11-15%, enough flow to remove decay heat)
- 5) When flow demand signal is less than 5%, auto returns to no override configuration REF:
AA52002-015
- DESIGNATES REQUESTFD CHANGES / ADDITIONS i
I 15 Attachment 1
ERC-1853 ANO 2 EXAMfNATION REPORT 12/12/85 4.3 Identify the procedural limits for the following plant operations:
A. RCS cooldown rate above 225 F B. Maximum pressurizer to RCS temperature differential C. Ilydrogen concentration in the RCS during normal operations D. S/G 1evels during plant startup ANS:
A. Do not exceed 100 F/hr
- B. 200 F or 350*F (see note)
C. 25-50 cc/kg
- D. 60%
- B. 200 F is the limit we attempt to maintain but exceeding this only requires logging the spray cycle. The question using the word maximum will probably be construed as 350 F due to notes in OP2102.02 Attachment "A". Due to wording of the question either answer should be acceptable.
- D. OP2102.02, this has been changed to 60% S/G level for startup due to S/G trip setpoints being changed allowing more room for operation between trip setpoints.
REF:
OP2102.02, Attachment "A"
- DESIGNATES REQUESTED CHANGE / ADDITION 16 Attachment 1
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 4.6 During operation, you are informed by an auxiliary operator that the generator transformer is running hotter than usual. List three possible reasons why a transformer may run hotter than normal and appropriate corrective actions for each case. (1.5)
ANS:
(Any three)
- 1) Hi current-reduce load 2} Fault - unit Shutdown
- 3) Oil breakdown - unit shutdown
- 4) Hi ambient temp - reduce load
- 5) Fan failure - start standby fan / reduce load
- 6) Pump failure - start standby pump / reduce load KEY:
A0P ELEDST REF:
ANO 2 STM-2-32 VAL:
0.2 each for reason 0.3 each for corrective action
- DESIGNATES REQUESTED CHANGE / ADDITION 18 Attachment 1
ERC-1853 ANO 2 EXAMlNATION REPORT 12/12/85 4.9 Give two reasons why the relative hazard from radiation is higher from internal sources than external sources. (2.0)
ANS:
(Any Two)
- 1) No protection (shielding) from internal exposure
- 2) Particulate radiations (alpha & beta) have very high ionizing ability but travel only a short distance in tissue. The damage done internally is localized to a small area around the source but a great deal of damage is done.
- 3) The different chemical characteristics of the radionuclides cause some of them to concentrate in certain body tissues (I-131 thyroids, SR-90 bone, etc.).
- 4) Bone seeking isotopes chemically bond to bone tissue and stay there. If the half life is long, the bone gets exposed to a lot of radiation for a long time.
- Other reasonable answers should be excepted i.e., continuous exposure due to inability to place distance between you and source.
KEY:
Radeon Job REF:
ANO 2 AA-52009-001 and AE-10101-042 VAL:
1 point each l
l l
- DESIGNATES REQUESTED CIIANGE/ ADDITION l
l l
18 Attachment 1 t
ERC-1853 ANO 2 EXAMINATION REPORT 12/12/85 4.11 During blackout conditions, and emergency diesel generator (EDG) starts but fails to pick up its associated ESF bus:
A. What are four conditions that may have prevented the EDG output breaker from closing onto its assigned buc?
B. Why must the operator take action to quickly energize the bus or stop the diesel?
ANS:
A. (Any four)
- 1) Improper EDG voltage
- 2) Improper EDG speed
- 3) Normal feeder breaker not open
- 4) Cross-tie feeder breaker not open
- 5) Bus lockout relays picked up
- 6) Loss of/no DC power to brk
- 7) Breaker fault
- 8) DC lockout
- 9) Breaker racked down
- The above should be added as possibilities for the brk not auto closing, since no specific reference was asked.
REF:
ANO 2 OP2202.01, Page 35, E-2100 Sheet I and E-2076
} VAL:
A. 0.5 each B. I pt
- DESIGNATED REQUESTED ClfANGES/ ADDITIONS 19 Attachment 1
.f .
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 ARKANSAS NUCLEAR ONE UNIT TWO SENIOR REACTOR OPERATOR EXAM OF 12/10/85 EXAM COT!ENTS Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 5.6 During a reactor startup, the operator stops regulating group CEAs at 144 steps on group 3. The source range count rate levels off at 1857 cps. The initial count rate was 400 cps at 0 steps withdrawn on regulating group 1, with Keff = 0.940 (1.5)
- a. Calculate the 1/M value for this control position
- b. What is the new value of Keff at this condition?
ANS:
- a. 1/M = CRt /CR2 = 400/1857 = 0.215
- b. 1/M = 1 - Keff2 /1 - Keffi 1 - Keff2= (1 - 0.940) x 0.215 = 0.9871 REF:
ANO-2 RTTM Chpt. 15 REQUESTED CHANGE
- Key answer is correct if you assume the source counts are 400.
However, since Keff was already .94 and the new Keff was .9871, thea 1/M = 1 - K, would result in 1/M = .0129 Either answer .215 or .0129 should be accepted.
- DESIGNATES REQUESTED CHANGE i
4
! 1 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 5.8 During cold shutdown conditions, the SITS are filled and pressurized.
Over the next few days the plant is taken to 100% power. Temperature in containment rises as does the temperature in the SITS. Given the following information, find how much gas must be vented to maintain pressure.
Ti = 68 F T2 = 100 F 3
Vtotal = 1850 ft V
water 4 0 ft3 P1 = 615 psia P2 = 615 psia 1
ANS:
P& ; Pn P n , 615(1850-1480) = 431 T1 T2 T1 460 + 68 i, = 431(460 + IOM = 392 615 392 370 = 22 ft3
- This answer is acceptable, however, a more accurate approach would consider the increase in water volume due to the temperature increase (1480 ft3) =
6 (1480) = 1487.7 which represents ~ 8 ft3 increase total gas vented = 22 + 8 = 30 ft3
- DESIGNATES REQUESTED CilANGES/ ADDITIONS 2 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 6.1 What equipment (5 items) discharges to the Quench Tank? (1.5)
ANS:
A. 1) RCP Seal Water Relief (RCP Control Bleedoff Relief)
- 2) Pressurizer Safeties
- 3) LTOPs
- 4) ECCS Vent
- 5) Reactor Vessel Head Vent
- 6) Pressurizer High Point Vent
- 7) Nitrogen Addition
- 8) Reactor Makeup Water Addition
- Agree with key, request addition of additional sources REF:
P&ID M-2230 H1
- DESIGNATES REQUESTED CHANGES / ADDITIONS 3
Attachment 1
ERC-1854 ANO 2 EXRMINATION REPORT 12/12/85 6.2 What indication does the operator have that the reactor vessel head inner gasket is leaking? (1.0)
ANS:
- 1) Leakage is detected by observing rise in temperature on leakoff line temperature indicator. (2 TIS-4662)
- 2) Leakage also detected by RCS RV head leakoff alarm (set point 150 F on 2K10, D4).
- Question did not ask for alarm setpoint. .5 for indication .5 for alarm REF:
P&ID M-2230
- DESIGNATES REQUESTED CHANGE / ADDITION i
4 Attachment 1
ERC-1854 ANO 2 EXAMINATION RtPORT 12/12/85 6.3 What five systems does the RWT provide water for? (1.5)
ANS:
- 2) LP safety injection
- 3) Refueling canal
- 4) Fuel pool makeup
- 5) CVCS makeup (emergency boration)
- 6) HPSI
- 7) SITS
- ECCS may be listed instead of HPSI, LPSI, and Containment Spray REF:
P&ID M-2231 & M-2236
- DESIGNATES REQUESTED CHANGE / ADDITION j
5 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 6.7 During normal power operation, the two temperature sensing elements on the outlet of the letdown heat exchanger fail high. What control functions occur? (2.0)
ANS:
- DESIGNATES REQUESTED CHANGE 6 Attachment 1
i ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 6.8 List six of the eight functions for which the 125VDC power system is designed to provide power.
ANS:
- 1. 6.9 KV Switchgear Control
- 2. 4.16 KV Switchgear Control
- 3. 480V Load Center Control
- 4. Reactor Control (Control Rod Device, Reactor Trip Circuit Breaker Control)
- 5. Reactor Inst. and Protective System
- 6. Engineered Safeguards System
- 7. Inverters (Vital 120VAC)
- 8. Other equipment necessary for normal unit operation, and normal and emergency shutdown. (Any component operated with 125VDC)
REF:
ANO 2 STM-2-32
- DESIGNATES REQUESTED CHANGE i
l.
7 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 6.9B List the three dedicated COLSS indications provided in the Control Room (i.e., items that are permanently displayed, vice assigned points normally on display)
ANS:
- Agree with key, however, student may have listed the core power limit based on margin to DNB or LPD AS Linear Power KW/FT/DNBR limit meter (2J19042/9040). ThisisthewayitislabeleionPanel2C03.
REF:
Plant Panel 2CO3
- DESIGNATES REQUESTED CHANGE / ADDITION 8 Attachment 1
ERC-1854 AN0 2 EXAMlNATION REPORT 12/12/85 6.10 During full power operation, the CIAS inadvertently actuates.
A. What effect does this have on continued plant operation and why?
B. What is required 'to reset it?
ANS:
- A. Agree with key, however, we need to add capability to override and restore CCW to RCPs (as per DCP 84-2058)
B. 1) Signal cleared
- 2) Reset PPS (trip path)
- 3) Reset ESFAS (actuation path)
- 4) Reset components
- 4) " Reset Component" is not needed to reset a CIAS.
REF:
DCP 84-2058 OP2202.01 B.2 Page 70 of 138
- DESIGNATES REQUESTED CRANGE/ ADDITION 9 Attachment 1
ERC-1854 ANO 2 EXAMlNATION REPORT 12/12/85 6.11 On an SIAS condition, what basically happens to the Service Water /
Auxiliary Cooling Water System? (8 items)
ANS:
- 1) ACW supplies shut
- 2) SFPHX supplies shut
- 4) Cooling tower makeup MOVs shut (the procedure names these however they are labeled
- a. SW Loop 2 ACW Return Isolation 2CV1542-2
- b. SW Loop 1 ACW Return Isolation 2CV1543-1)
- 5. Return MOVs to emergency pond open
- 6) Return MOVs to lake shut
- 7) SW pumps receive start signal
An interlock will then open the SW valves associated with the equipment.
- #5 & //6 can be covered by saying that SW returns shif t to the pond REF:
P&ID M2210 Sheet 1-3
- DESIGNATES REQUESTED CHANGE / ADDITION 10 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 7.4 During operation, you are informed by an auxiliary operator that the generator transformer is running hotter than usual. List three possible reasons why a transformer may run hotter than normal and appropriate corrective actions for each case. (1.5)
ANS:
(Any three)
- 1) Hi current-reduce load
- 2) Fault - unit Shutdown
- 3) Oil breakdown - unit shutdown
- 4) Hi ambient temp - reduce load
- 5) Fan failure - start standby fan / reduce load
- 6) Pump failure - start standby pump / reduce load KEY:
A0P ELEDST REF:
ANO 2 STM-2-32 VAL:
0.2 each for reason 0.3 each for corrective action
- DESIGNATES REQUESTED CHANGE / ADDITION 12 Attachment 1
ERC-1854 ANO 2 EXAMINAT10N REPORT 12/12/85 7.8 A. During recovery from blackout conditions the operator is directed to avoid any RCS cooldown. What are three reasons for this? (1.5)
B. During blackout what are four suggested actions to limit the drain on station batteries?
ANS:
(Any Three)
A. 1) No makeup available
- 2) No way to barate
- 3) No way to maintain RCS pressure
- 4) Steam pressure to low to run steam driven EFW pump (2P7A)
(Any Four)
B. 1) Vent generator 112 and stop DC seal oil pump
- 3) Secure main turbine DC powered lube oil pump
- 4) Secure plant computer if not required
- 5) Secure any unnecessary DC lighting KEY:
E0P ELEDST Catrl REF:
ANO 2 E0P2202.01, Page 56, 57 VAL:
0.5 each
- DESIGNATES REQUESTED CIIANGE 12 Attachment 1
ERC-1856 ANO 2 EXAMINATf0N REPORT 12/12/85 8.1D True or False A shift administrative assistant is not allowed to fill out the lloid Card Request Sheet and IIold Cards.
ANS:
- The shift administrative assistant is not specifically addressed in the latest revision of OP1000.27 Revision 4. The shift administrative assistant's qualifications vary and their ability to prepare tagouts is dealt with on a case to case basis. Question 8.1 may be answered yes or no.
REF:
OP1000.27 Rev. 4
- DESIGNATES REQUESTED CHANGE
\
v.
14 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 8.3 Khat four conditions require an independent review of tagout lineup and installation?
ANS:
- Revision 4 of 1000.27 dated 5/21/85 requires an independent verification for all tagouts.
REF:
OP1000.27 Rev. 4 Pages 9 & 11 (Included as Attachment 1)
- DESIGNATES REQUESTED CHANGES 14 Attachment 1 l
ERC-1856 ANO 2 EXAMINATION REPORT 12/12/85 8.5 What is the basis for maintaining 23 feet of water above the core during refueling operations?
ANS:
- T.S. Bases 3/4.9.8 " Shutdown Cooling and Coolant Circulatior." also contains an explanation of the 23 feet of water requiremen' -m d may be included in the examinee's response.
REF:
T.S. Bases 3/4.9.8
- DESIGNATES REQUESTED CHANGE 15 Attachment 1 i
ERC-1854 ANO 2 EXAMfNATION REPORT 12/12/85 8.8 Attempts to move the af fected rod individually are unsuccessful and the rest of the group 6 rods are position at 130".
With the information available up to this point, select the Technical Specification which most adequately covers the known condition and justify your selection.
ANS:
- At this time nothing is known about the rod except that it will not respond to in/out signals from the control panel. As per attached abnormal operating Procedure 2203.03 Rev. 2 dated 11/26/85, the rod would first be investigated and not assumed to be intrippable until verified to be so.
REF:
OP2203.03 Rev. 2 (New Procedure Revision) (Included as Attachment 2)
- DESIGNATES REQUESTED CHANGE 16 Attachment 1
ERC-1854 ANO 2 EXAMINATION REPORT 12/12/85 8.10 Which of the following conditions or events would result in an Emergency Action Level classification of " Alert" or higher?
ANS:
- Answers to question 8.10 are subject to interpretation. Answer to the questions should be accepted with the appropriate assumptions. For example, part 8.10D requires a loss of control of radioactive material and 8.10.G is not applicable above cold shutdown. The examinee's response may be yes or no depending on assumed situations.
REF:
OP1903.10
- DESIGNATES REQUESTED CilANGE 17 Attachment 1
'