ML20133D905
| ML20133D905 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/02/1985 |
| From: | Cooley R, Graves D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20133D894 | List: |
| References | |
| 50-298-OL-85-03, 50-298-OL-85-3, NUDOCS 8507220302 | |
| Download: ML20133D905 (70) | |
Text
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EXAMINATION REPORT 50-298/0L-85-03 l
l Facility Licensee:
Nebraska Public Power District l
P. O. Box 495 Columbus, Nebraska 68601 Facility Docket No:
50-298 Facility License No:
DPR-46 Examination administered at Cooper Nuclear Station (CNS) l 7/t/p r Chief Examiner:
David N. Graves Date M
l/d/f[
Approved by:
R. A. Cooley, Section@hief
[Ta te' Summary Examination on June 11-14, 1985 Examinations were administered to seven reactor operator candidates and one senior reactor operator candidate. All candidates passed.
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8507220302 850705 PDR ADOCK 05000298 PDR O
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DETAILS 1.
Persons Examined Reactor Operator Candidates Written examinations and oral operating examinations were administered to seven initial reactor operator candidates.
All candidates passed both portions of the examination.
Senior Reactor Operator Candidates The written and oral operating examinations were administered to one senior reactor operator upgrade candidate.
The candidate passed both portions of the examination.
I 2.
Examiners D.'N. Graves (Chief Examiner), NRC J. L. Pellet NRC 3.
Examination Report This Examination Report is comprised of the following sections:
a.
Examination Review Comments and Resolutions.
b.
Exit Meeting.
c.
Master copies of the Reactor Operator and Senior Reactor Operator written examinations.
I A.
Examination Review Comments and Resolutions This section reflects the comments made by the facility-during the i
examination review conducted following the written examination.
The comments accepted by the NRC reviewers have been incorporated into the master copies of the examinations. ~The following personnel participated in the review:
NRC Utility i
i D. N. Graves W. F. Gilbert A. E. Wilson J. L. Pellet D. A. Shallenberger D. W. VanDerKamp T. R. Sandner (GE)
K. P. Patek (GE)
K. E. Cigler (GE)
M. W. Parrish (GE)
Comments and resolutions are listed by section question number.
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COMMENTS l
1.
Question 1.03b and 5.04b The answer should read the fraction of l
fission neutrons which are born delayed instead of the number of fission neutrons j
that are born delayed.
Resolution:
Agree.
Answer key modified.
[
2.
Question 1.05e Samarium-149 is also shown as a direct fission product in the chart of the l
nuclides.
(Contribution to total
-Samarium-149 is minor) l Resolution:
Not sufficient to change answer key.
l Question at least asks for major contributor.
Direct fission product not accepted.
3.
Question 1.06(a&b)
Answer may be that steam flow indication is downstream of relief valves, so indicated steam flow will decrease by the amount flowing through the relief valves.
Resolution:
Agree.
Answer key modified.
4.
Question 2.03c and 6.07c The answer to the first part of Part C states that the flow control valve position remains unchanged. We feel that the flow control valve will change in.the open direction from its initial position, because the stabilizing valve failing closed causes the system flow to decrease slightly causing the flow l
control valve to open slightly to maintain the same flow.
Resolution:
Above accepted if candidate explained why flow reduced.
Key reflects this.
5.
Question.2.05c and 6.09 The student text, page CS-6, indicates that the white light will come on if the valve is closed with an initiation signal present.
This answer should also be accepted as correct.
l Resolution:
Agree.
Key does not need to be changed, i
6.
Question 2.06a The student text, page DG-12, discusses the trips that are not bypassed following an automatic diesel start as incomplete 1
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sequence, overspeed, generator lockout, and the manual trip.
The generator l
lockout encompasses the last five trips
. listed in the answer key.
Students L
may list generator lockout as one trip and thus fail to come up with five
- answers, t
L Resolution:
Not accepted.. Answer key stands as written.
I' l.
7.
Question 3.02a and 6.04 The student text, page RBM-15, lists
" power level less than.30%" as an auto bypass.
The answer should accept i
EITHER reference APRM less than 30%
l (answer key). or power level less than 30% (as it is listed in the student
. text).
Resolution:
Agree.
Key modified.
8.
Question 3.04b The high water level trip does not
- energize the trip solenoid.
It shuts the' steam inlet valve.
This question should be changed to accept 3 of 4 1-right answers for full credit.
Resolution:
Question not changed.
Answer key was
~
modified to remove high water level
' trip as an acceptable answer.
9.
Question 3.04c The local trip lever will also have to i
be reset locally and should be an acceptable answer.
Resolution:
Comment had no bearing on grading this' exam.
Question will be changed to ask for Automatic trip that must be l
locally reset.
- 10. Question 3.06a A'new MDC-has-added an' inhibit switch to the ADS logic. -This additional switch adds a third' answer of " Inhibit' switch to inhibit."
l Resolution:
Agree.
Answer key modified.
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- 11. Question 3.06b
- Drywell high pressure initiation signal
[
has been-removed from the ADS circuit and should-not be required to,be part of,the answer.
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5 Resolution:
Agree.
Answer key modified.
- 12. Question 4.01 and 7.04 The procedure does not state a preferred method.
HPCI is the first method that is listed in the procedure.
Candidates shuuld not be required to state a pre-ferred method if the procedure does not specify.one.
Resolution:
Disagree.
Other methods are used only if HPCI is unable to provide the desired control.
Answer key unchanged.
- 13. Question 5.09 a,b,&c Students may answer in terms of increasing quality causing a decrease in critical power and vise versa.
(Quality and enthalpy are directly related).
Answers which are expressed in terms of quality should be accepted.
Resolution:
Candidates answers would have been graded according to their explana-tions.
" Quality" answers would have been considered.
No candidates responded in those terms.
- 14. Question 5.10a Heat transfer through the gap may be ascribed to natural convection, radiation, or conduction.
(The gap is very narrow, allowing little i'
room for convection).
Conduction
-should also be allowed as an acceptable answer.
1 Resolution:
Material directly from plant training material.
Answer stands as written.
- 15. Question 6.01a The DEH pressure setpoint that the computer is using will not change.
l The pressure setpoint is the signal that is entered by the operator.
The actual setpoint will change due to a higher bias signal on the other pressure controller.
This 3 psig bias signal will cause the actual setpoint to increase by 3 psig but the pressure setpoint has not been changed by the failure.
The question is written in a manner that "no change" (925 psi) could be the answer.
l-6 Resolution:
Agree.
Answer key changed to 925 psig.
[
- 16. Question 6.02a The electrical fire pump C will auto start at 68 psig.
Reference:
Surveillance Procedure 6.4.5.3, page 4 of 5.
Resolution:
Agree.
Key modified.
17[ Question 7.06 AOP should be AP for Abnormal Procedure.
Resolution:
Agree.
Question and answer modified.
- 18. Question 7.07 and 7.08 These questions require the students I
to memorize subsequent actions of procedures which is contrary to the Operator Licensing Examiner Standards (NUREG-1021 of 10/83, Section ES-202 B.4) for abnormal and emergency procedures the candidates must demonstrate knowledge of the symptoms, automatic actions, and immediate actions.
It does not specify requirements concerning the subsequent actions.
Training policy has been for the candidates to understand the subsequent actions and not to memorize them.
Resolution:
The questions pertained to actions that have to be taken within either a specified period of time (less than 15 minutes) or a "short" period of time.
Actions are pertinent and time-limited.
Questions and answers stand.
- 19. Question 7.08 The procedure list two actions in one step and not separate steps as ifsted in the answer key.
Resolution:
Noted.
- 20. Question 8.04 A recent revision (Procedure 0.9, Page 2&3) was issued on May 15, 1985, to this procedure.
The candidate may or may not have been aware of this change.
i.
_b.
O 7
Answer key should allow for an answer from Revision 1 or Revision 2.
Resolution:
Agree.
Either accepted.
Master answer key will reflect Revision 2.
- 21. Question 8.05 Days 3 to 9 and 4 to 10 also exceed the greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a seven day period requirement and should also be acceptable as answers.
Reso l ut ior.-
Agree.
Key modified.
- 22. Question 8.07 Requiring the candidate to state
" Operations Center" is too specific and unnecessary knowledge.
The operators realize that contact is made to the NRC using the red phone.
Resolution:
Agree.
Key modified.
- 23. Question 8.10-Concepts which express the require-ments should also be allowed because the candidate demonstrates the requirements in addition to just stating the requirements.
Resolution:
. Agree.
Considered during grading.
- 24. Question 8.12 Technical Specifications do not express the requirements that the plant uses for manning requirements.
The requirement that the plant uses for crew manning are expressed in Conduct of Operations Procedure 2.0.3 and the Federal Register, Volume 48, No. 133 (July 1983).
j This question is confusing in that it requires the candidate to state requirements that are less stringent than those required by other plant r
procedures.
I f
Resolution:
Agree.
Considered during grading.
I
- 25. Question 8.03 Knowledge of what the hazardous materials are~and where the require-ments concerning them are located is
8 sufficient knowledge for someone not directly involved in the use of the materials. We feel that detailed knowledge of hazardous chemicals is inappropriate for a SRO candidate.
It does not reflect on knowledge level that is required of the candidate to be the supervisor of reactor operations.
Resolution:
Examiner feels that with the SR0 being the senior individual on site many times, he should be aware of the general hazards associated with materials routinely used. Detailed chemistry of hazardous material should not be required and the one answer to that effect will be deleted and point values reassigned.
B.
Exit Meeting Summary At the conclusion of the site visit, the examiners met with utility representatives to discuss the results of the examinations.
The following personnel were present for the exit meeting:
NRC Utility D. L. DuBois (Senior Resident Inspector)
R. D. Black D. N. Graves R. A. Jansky J. L. Pellet K. P. Patek J. L. Peaslee D. L. Reeves, Jr.
D. W. VanDerKamp D. A. Whitman A. E. Wilson Mr. Graves started the meeting by announcing preliminary results of the oral operating examinations with seven of the eight candidates being clear passes. One reactor operator candidate was not a clear pass as of the exit meeting.
The utility was informed that this did not mean the candidate failed the examination, but that a more detailed review of his weaknesses must be made.
Weaknesses, deficiencies, and items of concern noted during the exam were discussed.
The utility was informed that this information was for their use and benefit and not as items to be specifically followed up on by the NRC as open items.
Items discussed were as follows:
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i 1.
Candidates had difficulty with multiple casualty discussions when several procedures were in use simultaneously.
2.
Several procedures identified valves by valve number only, without verbal description, and candidates could not find or identify the valves.
3.
Candidates gave differing answers as to who the Fire Brigade l
Team Leader is on each shift.
Examiner observed that one candidate could not find where it was written as to who would lead the Fire Brigade.
4.
Examiner noted step off pads, normally associated with surface-contaminated areas, in areas that were not contaminated.
A contaminated clothing bin was provided but no I
I requirements were posted.
No generic or widespread candidate weaknesses were noted during the site visit and the utility was so informed.
The utility asked in what area (s) was the one not-clear passed candidate weak, or how could his performance be improved, to which the examiner responded that while knowledge weaknesses were noted, the candidates manner of answering made it very difficult to determine his actual depth of knowledge and understanding of system operation and concepts.
The meeting concluded with the examiners thanking the utility staff for their cooperation and efforts during the examination visit and informing them that final results would be. forthcoming as soon as possible.
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U.
S.
NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_QQQEEB__________________
REACTOR TYPE:
_RWB-gel _________________
DATE ADMINISTERED:_RilQh411________________
EXAMINER:
_QB8 Ment _Qt______________
APPLICANT:
IN118UQIl081_IQ_8EELIQ6BIl Uno separate paper for the answers.
Write answers on one side only.
Sttple question sheet on top of the answer sheets.
Points for each qu;stion are indicated in parentheses after the question. The passing grcde requires at least 70% in each category and a final grade of at locst 80%.
Examination papers will be picked up six (6) hours after tho examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY
__YALUE_ _IQIaL
___1GQBE___
_V8LUE__ ______________Q8IEQQBI_____________
_211Q0__ _212QQ
________ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_211Q0__ _21100
________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_21AQQ__ _21xQQ
________ 3.
INSTRUMENTS AND CONTROLS
_25tDQ _ _21tDQ
________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL j
CONTROL 1QQtDQ__ 1QQzQQ
________ TOTALS FINAL GRADE _________________%
All cork done on this examination is my own. I have neither givon nor received aid.
APPLICANT *S SIGNATURE i
11__EElb;1ELES_9E_UU2LE6E_EDEEE_EL6bl_0LEEL110W FAGL IbEBUQQ10601 Git _bE61_lB661EEB_6UD_ELu1R_ELOW QUESTION 1.01 (1.50) o.
Are BWR cores designed to be OVERMODERATED or UNDERMODERATED?
(0.5) b.
What does the state of moderation, over-or under, assure about the reactivity control design of the core?
(1.0)
QUESTION 1.02 (2.00) c.
Which reactivity coefficient is of greatest significance during normal power operation?
(0.5) b.
Which reactivity coefficient has great safety implications during a very rapid increase in power?
EXPLAIN.
(1.5) i QUESTION 1.03 (3.00) o.
On large down power transients, such as a scram, what limits the rate of power decrease?
(0.5) b.
Describe the relationship between B (Bets), B "bar", and B "bar" effective.
(1.5) c.
Describe how and why the response of the reactor to a positive reactivity change would be different if no delayed neutrons were
{
present in the core.
(1.0) i QUESTION 1.04 (2.00)
Indicate HOW (increase, decrease, unaffected) control rod worth changes for each of the situations listed below.
EXPLAIN WHY the J
rod worth is affected, if applicable.
1 o.
The reactor is heated from 100 deg F to 200 deg F (1.0) b.
Reactor power-is increased from 20% to 40% by control rod withdrawal (1.0)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1 __EBIBCitLE5_9E_GUELL6E_LOWL5-M AN' nN U110%
Ft GE ISEBUQQ10601C1&_BE61_lB651EEB_6UD_ELUIR_ELQW QUESTION 1.05 (3.00)
Complete the following:
(Blanks may require more than one word)
Xenon 135 has two (2) methods of production.
About 95% of the Xenon is produced by
___(a)___ and the remaining 5% is produced by ___(b)___.
Xenon also has two (2) removal methods,
___(c)___
cnd ___(d)___.
Samarium 149 is produced only by ___(e)___
and removed only by ___(f)___.
(3.0)
QUESTION 1.06 (3.00)
The reactor is operating at 75% power when ONE SRV inadvertently OPENS AND STAYS OPEN.
HOW would you expect the following parameters to change, and WHY?
NOTE:
Ensure your answer begins with the initial change and continues to a stable condition.
o.
Total INDICATED steam flow (1.0) b.
Turbine steam flow (1.0) c.
Reactor pressure (1.0)
QUESTION 1.07 (3.00)
For each of the following events, or changes in plant status, state chether the change will bring the system CLOSER TO, FURTHER FROM, or HAVE NO EFFECT ON the point at which the Reactor Recirculation pumps eill cavitate. GIVE A BRIEF EXPLANATION FOR EACH.
Assume all other parameters remain unchanged.
]
- o. Increase in reactor water level (1.0) b.
Loss of a feedwater heater (1.0) a
- c. Increase in Recirculation Pump speed (1.0)
(*****
CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1-PR1GLIELE1_0E_UUCLE6B_E9Esb EL6bl 0LLE61100, FAGE 4
IBEBUQQ1U601C1&_BE61_1860kEEB_6UD_ELulD_ELQW QUESTION 1.08 (2.00)
Why is nucleate boiling a better heat transfer condition than oubcooled heat transfer (2 reasons required)?
(2.0)
QUESTION 1.09 (2.00)
Match each of the four lettered items with one of the numbered items.
A letter-number sequence is sufficient.
(2.0) 1.
MAPRAT 5.
PCIOMR 2.
APLHGR 6.
CHF 3.
CPR 7.
TOTAL PF 4.
FLPD 8.
_____a.
Parameter by which plastic strain and deformation are limited to less than 1%.
_____b.
Ratio of bundle power required to produce onset of transition boiling somewhere in the bundle to actual bundle power.
_____c.
Parameter by which peak clad temperature is maintained less than 2200 degrees F during postulated design basis accident.
_____d.
Contains guidelines restricting power ramp rates above the threshold power.
QUESTION 1.10 (2.00)
Briefly describe the mechanism of natural circulation in a BWR, including the flow path.
(1.0)
QUESTION 1.11 (1.50) i A sume the plant is at 100% power and flow.
Explain WHAT HAPPENS to core flow and WHY es control rods are inserted.
Assume CONSTANT recirc pump speed.
(1.5) i l
I
(*****
END OF CATEGORY 01 *****)
2:__LL6Ul_DEllEU_lblLV21'E EUill AND_EUEE2ENCY Sill [UL I /4!
QUESTION 2.01 (3.50) c.
From where does the RHR Service Water Booster system take its suction and to where does it discharge?
(0.5) b.
HOW is an inadvertent admission of RHR Service Water into the RHR system prevented (physically, not administratively)?
(0.5) c.
WHAT is the relationship between RHR Service Water system pressure and RHR system pressure and WHY?
(1.0) i d.
HOW is'the relationship in "c"
above controlled or varied?
(0.5) o.
A RHR Service Water Booster Pump control switch is positioned to START and the switch stays in this position.
Is this desirable?
EXPLAIN.
(1.0)
I i
QUESTION 2.02 (1.50)
During normal operation, there is only one way that a reactor scram can be caused by actuation of a single component.
O.
What is this component?
(0.5) b.
What are the functions of the two (2) timers associated with a scram initiated by this csmponent (include the length of time associated with each timer)?
(1.0) l 1
QUESTION 2.03 (3.50) o.
HOW and WHY does the on-line CROH flow control valve respond following a scram?
(1.0) b.
What prevents the CR0 pumps from reaching a "r unoutt' condition following a scram?
(0.5) c.
A control rod is inserted one notch.
The insert stabilizing valve does not reopen after the rod movement.
How does this affect the indicated drive water differential pressure and on-line flow control valve position?
EXPLAIN.
(2.0)
(*****
CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2:__CL6bl_D[llkb lbCLUDJbt_k6LLli AND_EUEEOLUCl_!1klL5s F' A G E t
QUESTION 2.04 (3.00) o.
Describe the arrangement of components that ensures no portion of the i
primary containment exceeds the maximum design EXTERNAL pressure?
Include the maximum design EXTERNAL pressure.
(1.5) b.
The secondary containment system utilizes four different features to mitigate the consequences of a postulated loss of coolant accident or refueling accident.
LIST three (3) of these features.(1.5)
QUESTION 2.05 (3.50)
O.
The reactor is at 400 psig.
Describe how the Core Spray inj ect ion valves (MO-ll and MD-12) must be operated to have them both open at the same time.
(0.5) b.
The reactor is at 500 psig.
Can the Core spray injection valves (MO-ll and MO-12) both be opened at the same time?
If so, describe how they must be opened.
If not, can EITHER of the two valves be opened with no inj ection signal present?
EXPLAIN.
(1.0)
I o.
What is the meaning of the white light above the MO-11 and MD-12 control switches?
Once illuminated, what THREE (3) conditions will cause the condition to clear?
(2.0) i f
QUESTION -2.06 (3.00)
I o.
Once en automatic start signal has been received by the Emergency Diesel Generators, the safety lockout solenoid is energized and locks out some of the diesel generator trips.
LIST FIVE (5) specific AUTOMATIC trips or conditions that are NOT bypassed (still functional) following an automatic diesel start.
(1.5) b.
Where are the two (2) emergency shutdown pushbuttons for each diesel generator located?
BE SPECIFIC as to location.
(1.0)
I c.
Describe the overload capability of the diesel generators.
(0.5) r
(*****. CATEGORY 02 CONTINUED ON NEXT PAGE *****)
2t__CL6b1_DLi10b_lbiLUDIN0_E6LLll 60L'_LULEELUGl_EllllO!
FAGE QUESTION 2.07 (3.00)
O.
What is Care) the source (s) of operating gas to the Inboard and Outboard Main Steam Isolation Valves?
Include NORMAL and BACKUP supplies as applicable.
(1.0) b.
The pneumatic accumulator associated with each MSIV is sized to allow HOW MUCH valve operation?
(0.5) c.
The two normal operating solenoids for each MSIV are powered by WHAT TYPE (S) of power (AC, DC, 24V, 48V, 120V, etc.)?
(0.5) d.
How would the MSIV respond to 1.
A total loss of power to the valve (0.5) 2.
A total loss of pneumatics to the valve (0.5)
QUESTION 2.08 (4.00)
For each of the HPCI (High Pressure Coolant Inj ection) System component foilures listed below, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel, IF IT WILL NOT INJECT WHY, AND IF IT WILL INJECT, provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of system operation cith the failed component.
A sume NO OPERATOR ACTION, and the component is in the failed condition at the time HPCI receives the auto initiating signal.
o.
The GLAND SEAL EXHAUSTER fails to operate.
(1.0) b.
The turbine AUXILIARY LUBE OIL PUMP fails to operate.
(1.0) c.
The MINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when system conditions require it to be open.
(1.0) d.
The HPCI PUMP DISCHARGE FLOW ELEMENT OUTPUT SIGNAL to the HPCI flow controller is FAILED AT ITS MAXIMUM output.
(1.0) d
)
(*****
END OF CATEGORY 02 *****)
3t__lb116UDENTS AND_CDUlBOLE PAbl t
QUESTION 3.01 (2.50) c.
What characteristic of the events occurring in a fission chamber detector makes gamma discrimination possible?
(0.5) 6.
Describe how the IRM range switch position effects the SRM rod blocks.
Include the rod blocks affected and how the IRM's affect each one.
(2.0) j QUESTION 3.02 (2.50) 0.
List three (3) ways that the Rod Block Monitor (RBM) may be bypassed.
Include automatic and/or manual bypasses.
(1.5) b.
How does the RBM utilize the input from a LPRM detector that is failed HIGH or failed LOW 7 DISCUSS BOTH cases, but limit your answer to how the LPRM input is considered in the aversging and counting circuits.
Assume the LPRM recently failed and has NOT been bypassed with its individual bypass switch.
(1.0)
QUESTION 3.03 (2.00)
The reactor is operating at 100% under steady state conditions.
An instrument technician mistakenly isolates and equalizes the pressure ccross one of the MAIN STEAM line flow transmitters.
DESCRIBE the response of the Feedwater Control System until steady state conditions are again established.
ASSUME 3-ELEMENT CONTROL.
(2.0)
QUESTION 3.04 (2.50)
O.
The RCIC flow transmitter providing the input to the RCIC flow controller has failed to minimum (0 gpm output).
RCIC is in a normal standby lineup.
How will the RCIC turbine respond if an initiation signal is received?
Continue your answer until the turbine is in a stable condition.
(1.0) b.
List four (4) RCIC system turbine trips that actuate by ener-gizing the turbine trip solenoid.
SETPOINTS NOT REQUIRED.
(1.0) c.
What is the only automatic RCIC turbine trip that must be reset locally?
SETPOINT NOT REQUIFED.
(0.E)
(*****
CATEGCRY 03 CONTINUED ON NEXT PAGE *****)
2___lG21bVOLUlE_600_CQUlBOLE IM4 9
QUESTION 3.05 (3.00)
For each of the situstions listed below, describe how the rocirculation and flow control systems will respond and why.
Include which component is controlling speed and whether both or only one pump is affected.
o.
28% power, recirculation pumps at minimum speed with both M/A transfer stations in manual.
Recirculation pump "A"
M/A transfer station is switched to AUTO.
(1.0) b.
75% power, recirculation flow control is in master manual when the MG set "A" tachometer output fails to a zero output.
(1.0) c.
75% power, recirculation flow control is in master manual when one reactor feed pump trips.
(1.0)
QUESTION 3.06 (3.50) o.
All parameter inputs to the ADS are present for initiation and the timer is counting down.
What are two (2) ways the operator can prevent the ADS valves from actuating?
(1.0) b.
Which of the four initiation logic signals intere) sealed in?
(0.5) c.
With NO ADS initiation signals present, what is the status of the red and blue lights on the 9-3 panel associated with the ADS valves?
(0.5) d.
What do the red and blue lights on the 9-3 panel indicate i
about the state or status of the ADS valves?
(1.0)
I o.
Experience has shown that the SRV's can inadvertently open if what condition occurs?
Assume the logic is functioning properly and there are no breaks in the system.
(0.53
(*****
CATEGORY 03 CONTINUED ON NEXT PAGE *****)
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21__IULIEU5LUIE_602_C001h0LS FAUL 10 QUESTION 3.07 (3.50)
O.
What provisions are made in the Offges System to DETECT en explosion or burn-back?
(1.0) b.
If an explosion or burn-back is detected, what automatic action takes place in the Offges System?
(0.5) c.
What are three ways that an Offgas System isolation can be accomplished (with the exception of individually closing each isolation valve)?
Include any time delays that occur as appropriate.
(2.0)
QUESTION ~3.08 (2.00)
Dascribe the response of the Main Turbine Control System (DEH) to occh of the events described below.
Assume ull control systems function normally and NO RPS TRIP is generated.
o.
One steam bypass valve partially opens while at 70% power (DEH in Mode 4)
(1.0) b.
Reactor power is manually increased above the load reference setpoint (DEH in Mode 4)
(1.0)
QUESTION 3.09 (1.50)
Answer the following with regard to the DEH System' O.
When does the turbine enter Mode 2 - Turbine Start?
(0.5) b.
What valves reposition as a result of entering Mode 27 (0.5) c.
Mode 3 - Turbine Load Control, is initiated when what event takes place?
(0.5)
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L
2:__lb!IB'0Eblf_6bO_CQblB9Li FACE 11 2
QUESTION 3.10 (2.00)
There are'six motor operated valves associated with the feedwater cystem that are interlocked with a main turbine trip.
Give the following for these valves on a main turbine trip:
(2.0) c.
Valve description (name or number) and Valve action (opens or closes) b.
Reason for the interlock c.
Any bypasses l
l
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END OF CATEGORY 03 *****)
62__tBOCLQUEEL_ _bOS06Lt_eEURE56L&_LULBkEUC1_ BUD F'd i;
86010LQQ1G6L_CQUlBQL QUESTION 4.01 (2.00)
The reactor was operating at high power when an MSIV isolation Cnd resulting scram occurs.
c.
Per procedure 2.1.8, Scram Recovery During Power Operation MSIVs Closed, what are two (2) methods by which the control rods are verified inserted?
(1.0) b.
What is the preferred method of controlling the reactor pressure and couldown?
(0.5) c.
If the preferred method causes the cooldown rate to be excessive, what system should be used?
(0.5)
I QUESTION 4.02 (1.50) i What is " Hot Standby" condition?
(1.5)
I QUESTION 4.03 (2.50) o.
Annunciator windows on panel g-5 having a RED background are associated with WHAT CONDITION?
(0.5) b.
Indicate the color of the annunciator windows that denotes each of the following conditions:
1.
Priority 1 - A critical condition that requires immediate (0.5) operator action 2.
Priority II - An off-normal condition which could rapidly (0.5) develop into a critical condition 3.
Priority III - An off-normal condition which requires (0.5) operator followup c.
WHO must give his/her approval before any annunciator card may be removed (identify by minimum requirements, NOT name)?
(0.5)
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CATEGORY 04 CONTINUED ON NEXT PAGE *****)
I S z _ _ E B 9 C El'L'E L1_: _U D BU 6 L t _6 EU R Bt 6 L t LU L B 0 E U21. ES2 f' AG E 1 :'
88Q19LQQ1G8L_GQUlBQL QUESTION 4.04 (2.00)
What are two (2) undesirable consequences of a loss of both Control Rod Drive Hydraulic pumps?
Include WHY the consequences are undesirable.
(2.0)
QUESTION 4.05 (1.50)
An unexpected and unexplained increase in reactor power has cccurred.
AP 2.4.1.3, Unexplained Increase in Reactor Power, requires reactor power to be reduced.
HOW does the procedure Gay to reduce power and WHY is that method used?
(1.5)
QUESTION 4.06 (1.00)
Match the class of fires listed below (A - 0) with the materials (1.0) involved (1 - 4).
CLASS OF FIRE MATERIALS INVOLVED
______A.
Alphs 1.
flammable liquids, gases, or gresses
______B.
Bravo 2.
combustible metals
______C.
Charlie 3.
ordinary combustibles (paper, wood, etc.)
______D.
Delte 4.
energized electrical equipment 1
l
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st__L bk'L LDL'BLi_.UDEU6Lt 6000bt6L t _l0L EElhC) A N L' i /4 E B6019LQQ1G6L_CQUlBQL QUESTION 4.07 (4.00)
A loss of all site AC power has occurred. Answer the folicwing questions concerning E0P 5.2.5.1, Loss of All AC Power Station Bisckout.
c.
How is the reactor scram verified?
(0.5) b.
FILL IN THE BLANKS.
If it can be ascertained that AC power WILL BE restored within __(1)_
minutes, maintain reactor level with __(2)__ or __(3)
(1.0) o.
Describe how reactor pressure is controlled if the condition of part "b" applies (power will be restored within X minutes).
(1.0) d.
Describe how reactor pressure is controlled if AC power CANNOT be restored within X minutes.
(0.5) o.
Why should reactor pressure NOT be reduced below the saturation pressure corresponding to the maximum drywell temperature?
(1.0)
QUESTION 4.08 (2.50) c.
What are three (3) areas or locations that will be manned by operations personnel following an evacuation of the control room?
(1.5) b.
If the reactor was not scrammed prior to leaving the control room, what are two (2) alternate methods of scramming the i
reactor, per E0P 5.2.1, Shutdown from Outside the Control Room?
(1.0) l QUESTION 4.09 (2.00)
A loss of the Service Water Systim has occurred and it is not cxpected that the system will be restored within a short period of time.
What are four (4) actions that must be performed?
(2.0)
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E 1
i St-Eb2CEDL' ELL : bOBU6Li_6000btLLt LDEEELuil 6bD rett B601QLQQlC6L_CQUIBQL l
(
QUESTION 4.10 (1.00) 1 List the four classes of Emergency Action Levels that have been catablished at Cooper Nuclear Station in order of INCREASING SEVERITY.
(1.0) i l
QUESTION 4.11 (2.50)
O.
What causes " rotor short" and " rotor long" conditions?
(1.0) b.
HOW should steam flow be changed (increased or decreased) to l
correct " rotor short" and " rotor long" conditions and WHY does the change in steam flow correct the situation ?
(1.5) l QUESTION 4.12 (2.50)
O.
Who has the ultimate responsibility for an individual's safety with regard to radiation protection?
(0.5) b.
What are the Cooper Nuclear Station ADMINISTRATIVE limits for whole body penetrating radiation exposure for a NPPO employee?
(2.0) l l
l 1
1 I
l 1
l l
l i
l l
l 4
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END OF CATEGORY 04 *****)
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ANSWERS -- COOPER
-85/06/11-GRAVES, D.
i l
l l
l ANSWER 1.01 (1.50) o.
Undermoderated (0.5) b.
This ensures that the moderator temperature coefficient (0.25) and the void coefficient (0.25) are negative (0.5).
REFERENCE GE Reactor Physics Review, pg 26, Fig 33 CNS Question Bank, Question VI-A-8 ANSWER 1.02 (2.00) o.
Void coefficient (0.5) b.
Fuel temperature coefficient (0.5).
On a power increase, fuel i
temperature increases almost immediately (0.5) adding negative reactivity (0.5).
The fuel temperature coefficient can change reactivity before any of the other reactivity coefficients can respond.
REFERENCE GE Reactor Physics Review,Section I, Coefficients of Reactivity l
ANSWER 1.03 (3.00) o.
The rate of power decrease is limited by the longest lived (0.25) l delayed neutron precursor (0.25).
l l
b.
B - The fraction of fission neutrons born delayed (0.25) for a l
given isotope (0.25).
8 "bar" - Weighted average of the various fuels' Bs (0.5).
8 "bar" effective _ The fraction of the thermal neutron population l
that was born delayed (0.5).
c.
With no delayed neutrons present, the neutron generation time would be the same as the prompt neutron generation time, and l
the rate of power change would be uncontrollably rapid (1.0).
REFERENCE GE Reactor Physics Review, pg 21-22 l
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-85/06/11-GRAVES, D.
ANSWER 1.04 (2.00)
O.
Increase (0.25).
As the moderator density decreases, the neutrons travel a greater distance and are more likely to interact with a control rod (0.75).
b.
Decrease (0.25).
The voids depress the thermal neutron flux which in turn decreases the rod worth (0.75).
REFERENCE GE Reactor Theory Review, pg 37 ANSWER 1.05 (3.00) c.
decay of Iodine b.
directly from fission c.
neutron absorption or burnout d.
radioactive decay o.
decay of fission products (Nd or Pm) f.
neutron absorption or burnout (0.5 each)
REFERENCE GE Reactor Physics Review, pg 45-46 ANSWER 1.06 (3.00) 4 0.
As the OEH closes the CVs due to the reactor pressure decrease from the RV opening, total indicated steem flow will decrease OR the SRV l
opening diverts steem flow away from the flow indicator sensor (1.0).
b.
As pressure decreases, the OEH will shut down on the CVs and turbine steem flow will decrease OR the SRV opening will divert steem flow away from the turbine (1.0).
c.
Will initially decrease, then will stabilize et a pressure slightly lower than what it was prior to the RV opening (1.0).
REFERENCE BWR-4 Transients
li__EB10CIELE t_0E UVEL( 65_E0VE b_EL6bl_ DEL E 611051 6/M l'
IBEBDQQ19601G1t_BL61_lB691'IB_6bD_ELV10 ELQW ANSWERS -- COOPER
-85/06/11-GRAVES, D.
1 ANSWER 1.07 (3.00)
- o. Farther from cavitation (0.5). As the reactor water level increases, the static head of water component in the NPSH determination is also increasing which adds NPSH. (0,5) b.
Forther from cavitation (0.5). If a feedwater heater is lost, then the temperature of the water entering the reactor is lower, which brings the water farther from the saturation temperature. (0.5) c.
Closer to cavitation (0.5). As pump speed increases, the pressure in the eye of the impeller decreases, which will cause the pump to cavitate earlier with the some NPSH. (0.5)
REFERENCE CNS Recirculation System Description, pg 25 GE Thermodynamics, Heat Transfer, and Fluid Flow pg 7-93 through 7-96 ANSWER 1.08 (2.00)
A large amount of heat is removed in changing the water to steem (heat of vaporization) (1.0).
The production of bubbles breaks up the Icminer layer, creating agitation, which provides better mixing (1.0).
REFERENCE GE Thermodynamics, Heat Transfer, and Fluid Flow, pg 9-8 ANSWER 1.09 (2.00) c.
8 b.
3 c.
2 d.
5 (0.5 each) l REFERENCE GE Thermodynamics, Heat Transfer, and Fluid Flow, Chapter 9 l
l i t _ _ L E l G C I L L EI. 0 E _bL'C L L 6 E _ E DVE B _ E L 601_0 L E L L 11 L L.
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ANSWERS -- COOPER
-85/06/11-GRAVES, D.
i ANSWER 1.10 (2.00)
A3 the water in the core heats it expends and becomes less dense (0.5).
When this happens, the weight or downward force of the C3re dense cooler water forces water through the jet pumps and into the bottom of the vessel (0.5).
The flow path is through the jet pumps into the bottom of the vessel, through the core, into the moisture seperators, and back into the annulus or jot pump area (1.0).
(2.0)
REFERENCE SE Thermodynamics, Heat Transfer, and Fluid Flow pg 9-117 through 119 l'
ANSWER 1.11 (1.50)
Core flow will increase (0.5).
Decreasing power by inserting control rods decreases the two phase flow resistence (1.0)
REFERENCE CNS Recirculation System Description, Figure 11 GE Thermodynamics, Heat Transfer, and Fluid Flow pg 9-47 1
I l
2 A__ E s hbl_DlillS ISCLVOISL' 16Elll_6bl'_l bll!:i M ' !'kil5' fl e t 2
ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 2.01 (3.50)
O.
suction - Station Service Water headers (0.25) discharge - Circulating Water discharge consi (0.25) b.
Two manual, locked-closed valves are in the line (0.5).
Number not required.
c.
RHRSW pressure is maintained greater than RHR system pressure (0.5) to prevent any radioactive leakage from the RHR system into the Service Water system (0.5).
d.
The pressure relationship is maintained by varying the position of the RHR HX Service water outlet valve (0.5).
o.
No (0.25).
The RHR HX Service Water outlet valve will continue to open and RHRSW pump runout could occur (0.75).
REFERENCE CNS Service Water System Description, pg 13-15 l
ANSWER 2.02 (1.50) o.
Mode switch (0.5) l b.
2 second timer (0.25) auto bypasses the scram signal initiated by taking the mode switch to shutdown (0.25).
The 10 second timer (0.25) ensures the rods are fully inserted before the scram can be reset (0.25).
REFERENCE CNS Reactor Protection System Description, og 5,7,8 l
i j
i
2 x _. 0 L Eb l _ D L 1105_ lU C L L'0100.16 L L il _600_ l L E !1 L G 01_ L l t i t O L F A6t
.1 ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 2.03 (3.50) o.
The flow control valve will go to its minimum position (0.5).
The sensed flow increases as charging flow to the accumulators increases.
The increased flow causes the FCV to close, diverting most of the flow to the accumulators (0,5).
b.
A flow orifice in the charging line limits the pump flow to prevent reaching a runout condition (0.5).
c.
Flow control valve position remains unchanged (0.5) because total system flow has not changed (0.5) OR Flow control valve will open slightly (0.5) to overcome the increased differential pressure (0.5).
The indicated differential pressure will increase (0.5) because flow previously bypassing the drive water pressure control valve is now passing through it (0.5).
REFERENCE CNS CR0 System Description, pg 5,6, Figure 1 ANSWER 2.04 (3.00) o.
Suppression pool to drywell vacuum breakers (0.6) and reactor building to suppression pool vacuum breakers (0.6) prevent exceeding the maximum design external pressure limit of 2 psi (0.3).
negative pressure barrier which minimizes the ground level b.
release by exfiltration low leakege containment volume which provides a holdup time for fission product det ay pr ior to release removal of particulates and iodines by filtration prior to release
- exhausting of the secondary containment atmosphere through an elevated release point (3 required at 0.5 each)
REFERENCE CNS Containment System Description, pg 4-5,10
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ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 2.05 (3.50) o.
MO-11 must be opened first (0.5) b.
No (0.25), Yes (0.25), either of the two valves can be opened, but both cannot be open at the same time (0.5).
t c.
The white light means that only manual operation of the valve is allowed (0.5).
Also accept that the valve is shut and an initiation signal is present.
The condition can be cleared by :
- termination of the auto initiation signal (0.5) reactor pressure increases to ' > 450 psig (0.5) loss of power to the pump motor bus (0,5)
REFERENCE CNS Core Spray System Description, pg 6 ANSWER 2.06 (3.00) incomplete sequence O.
- generator differential current
- generator phase overcurrent
- loss of field
- generator overvoltage reverse power l
I (5 required at 0.3 each) b.
One on the local control panel (0,5)
One on the west side (0.25) of the diesel engine (0.25) c.
It is permissible to run the diesel generator at 110% rated I
load (0.25) for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> out of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (0.25).
REFERENCE CNS Emergency Diesel Generator System Description, pg 13,14,17 l
j I
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ANSWERS -- COOPER
-85/06/ll-GRAVES, D.
ANSWER 2.07 (3.00) o.
Inboard MSIV - normally operated with nitrogen (0.33) but can also be operated with reliable instrument air (0.34) upon loss of ni-trogen.
Outboard MSIV - supplied only by reliable instrument air (0.33) b.
The accumulator volume is adequate to provide full stroking of the valve through one-half cycle (0.5).
c.
120 VAC (0.25) 125 VDC (0.25) d.
1.
Valves go shut (0.5) 2.
Valves go shut (0.5)
REFERENCE CNS Main Steam System Description, pg 3-4 ANSWER 2.08 (4.00)
O.
Will inject (0.25).
Turbine seal leakage resulting in potential air-borne activity in the HPCI room (0.75).
I b.
Will not inject (0.25).
Turbine stop and control valves will not open (0.75).
c.
Will inject (0.25).
Pump overheating and seal damage may result during low or no flow conditions (0.75).
d.
Will not inj ect (0.25).
Maximum signal from the flow element will cause the controller to keep turbine speed at minimum (0.75).
REFERENCE CNS HPCI System Description, pg 9,12,14 o
2z__IULIEUULUIE_6b2_C001E2Li FAGE 24 l
l ANSWERS -- COOPER
-85/06/11-GRAVES, D.
l l
ANSWER 3.01 (2.50) c.
The pulses generated by neutrons are six times larger than pulses generated by gammas (0.5).
"Six" not required.
b.
SRM Downscale (0.5) and Retract Not Permitted (0.5) rod blocks are effective only below IRM Range 3 (0.25).
SRM Upscale (0.5) and SRM Inop (0.5) rod blocks are effective only below IRM Range 8 (0.25).
REFERENCE CNS SRM System Description, SRM Interlocks and Trips Table I
i ANSWER 3.02 (2.50) c.
- Manual operation of the RBM BYPASS switch (0.5) l
-< 30% power (0.5)
- Edge rod selected (0.5) b.
Failed Low:
Removes the LPRM input from the averaging circuit (0.25) and provides s' signal to the counting circuit (0.25).
Failed High:
The higher input is averaged with the other inputs and processed as if it were a valid signal (0.5)
REFERENCE CNS Rod Block Monitor System Description, pg 5
(
ANSWER 3.03 (2.00) o.
Total steam flow would indicate 75% instead of the actual 100% (0.5).
The FWLCS would assume steam flow has decreased and reduce feed flow accordingly (0.5).
Level will begin to decrease.
As level con-tinues to decrease, a level error signal is generated which will increase feed flow back to the 100% value (0.5).
A new steady state level will be established lower than the original level.
Will accept any similar explanation (0.5).
REFERENCE CNS Feedwater Control System Description, pg 4,5 L
2t __IULIEL'UEUIL _639_99 BIS 9L L PAGE It ANSWERS -- COOPER'
-85/06/11-GRAVES, D.
l 9
ANSWER 3.04 (2.50)
- 0..
RCIC tusbine speed will ramp up on the ramp generator (0.5), then trip on overspeed (0.5).
b.
Manual trip High turbine exhaust backpressure (25 psig)
Low RCIC pump suction pressure (15" HG)
Any RCIC isolation' signal (4 required at 0.25 each) c.
Overspeed (0.5)
'. REFERENCE CNS RCIC System Description,-pg-12,13',15 ANSWER 3.05 (3.00)
's.
"A" pump speeds' up to 45% (0.5), the low setting on the dual
{
speed control limiter (0.25).
"B" pump speed is unaffected (0.25).
b.
"A" MG will trip on field undervoltage (0.5).
The tachometer output is the ref erence signal to 'the voltage regulator (0.25).
"B" pump speed is ur.af f ected (0.25).
c.
No change until reactor vessel level reaches 27.5" (0.25), then both pumps (0.25) runback to 45% speed (0.25) due to the #2 speed limiter (0.25).
REFERENCE CNS Recirculation System Description, pg 12,13, Figure 3, Figure 4 l
l
2 t __1UElEL'OLUlk_660_ CDUIE 9L S 5%0L et ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 3.06 (3.50) c.
Depress both logic RESET pushbuttons Turn off all running CS and RHR pumps Inhibit switches to INHIBIT (2 at 0.5 each) b.
120 second timer (0.50) c.
With no initiation signal preuent, the red light should be 8
off (0.25), and the blue ligh; should be on (0.25) d.
The red light indicates on when the solenoid is energized for valve opening (0.5).
The blue light indicates on when a a high pressure in the tail pipe does not exist (0.5).
o.
Excessive air supply pressure (0.5) can cause inadvertent opening.
REFERENCE CNS Nuclear Pressure Relief System, pg 4,6-g ANSWER 3.07 (3.50) c.
Pressure (0.33) and temperature (0.33) sensors in the SJAE discharge piping (0.33).
b.
The SJAE inlet isolation valves shut (0.5).
c.
Offgas rad monitors (0.5) with a 15 minute time delay (0.2S)
Oilution fan low flow (0.5) with a 5 minute time delay (0.25)
Offgas timer switch to CLOSE (0.5) with no time delay.
REFERENCE CNS Offges and Augmented Offgas Systems Description, pg 5-7 n..
PAGE 2t__1GLIEUtiEUli_60D_cggIEgLE ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 3.08 (2.00) e.
Increased steam flow causes steam header pressure to decrease.
DEH closes down on the governor valves to maintain the steam pressure setpoint (1.0) or similar explanation.
b.
Increased reactor power causes steam header pressure to increase.
DEH opens the governor valves until the load reference limit is reached.
At this point the increase in steam header pressure would overcome the bypass valves close bias and DEH would start to open the bypass valves (1.0) or similar explanation.
REFERENCE CNS DEH System Description, pg 7 ANSWER 3.09 (1.50) e.
Mode 2 is initiated when the turbine autostop oil system is latched (0.5).
b.
The turbine stop valves (0.2), intercept valves (0.2), and the reheat stop valves (0.1) open fully.
c.
Mode 3 is entered when the generator output breaker is shut (0.5).
REFERENCE CNS DEH System Description, pg 4,6 i
ANSWER 3.10 (2.00) e.
(RF-31MV) RFP1A S/U valve inlet (0.1) opens (0.1)
(RF-32MV) RFP1A S/U valve outlet (0.1) opens (0.1)
(RF-33MV) RFP1B S/U valve inlet (0.1) opens (0.13 (RF-34MV) RFP1B S/U valve outlet (0.1) opens (0.1)
(RF-29MV) RF P1A discharge valve (0.1) closes (0.1)
(RF-30MV) rip 1D discharge valve (0.1) closes (0.1) b.
To prevent overfeeding of the reactor on a turbine trip (0.3) c..
The interlock is automatically bypassed 3 minutes after the turbine trip (0.5).
REFERENCE CNS Feedwater Text CNS Question Bank, Book II, Section J, Question 13
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<E B6D10LQQ106L_QQNIBQL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 4.01 (2.00) o.
- All green full-in lights on the 9-5 panel are illuminated (0.5)
- Place the reactor mode switch in REFUEL (0.25) and verify the white REFUEL PERMISSIVE light is illuminated (0.25) b.
The preferred method is starting HPCI in the test mode (0.5).
c.
RCIC (0.5)
REFERENCE CNS GOP 2.1.8, Scram Recovery During Power Operation MSIVs Closed, Rev 9, pg 2,3 ANSWER 4.02 (1.50)
Hot Standby is a condition where reactor pressure is less than 1000 psig (0.5), coolant temperature is > 212 deg F (0.5), and the reactor mode switch is in the Startup/ Hot Standby position (0.5).
REFERENCE CNS GOP 2.1.9, Hot Standby Condition, Rev 5, pg 1 ANSWER 4.03 (2.50) e.
Scram or half-scram (0.5) b.
1.
red 2.
green 3.
white (0.5 each) c.
An SRO licensed individual (0.5)
REFERENCE CNS Alarm Procedure 2.3.1, General Alarm Procedure, Rev 8, pg 4,7
d2__tBOCEDVBEE_:_UQBU6Lt_6EUQBU6Lt_EUEBEEUC1_600 PAGE 29 B6Q10LQQ1G6L_GQUlBQL ANSWERS -- COOPER
-85/06/ll-GRAVES, D.
i
~
ANSWER 4.04 (2.00)
- Failure of the CRD pumps stops cooling water to the drives (0.5),
shortening seal life (0.5).
- Failure of the CRD pumps also allows the CRD accumulators to slowly depressurize (0.5).
At low reactor pressures, the accumulators are required to ensure that the control rods are f ully scrammed within the required time (0.5).
REFERENCE CNS AP 2.4.1.1.4, Loss of CRD Pump, Rev 5, pg 2 ANSWER 4.05 (1.50)
Reactor power is reduced with recirculation flow (0.5).
Reduction of power by reducing recirculation flow is the only way to reduce neutron flux patterns all across the core OR Moving the control rods in other than an approved sequence may worsen the problem (either acceptable) (1.0).
REFERENCE CNS AP 2.4.1.3, Unexplained Increase in Reactor Power, Rev 8, pg 1,3 ANSWER 4.06 (1.00)
A.
3 B.
1 C.
4 D.
2 REFERENCE CNS E0P 5.4.1, General Fire Procedure, Rev 17, Attachment "B",
pg i f
Sz__EB9CEDMBEE_:_U9806Le_6809806Lt EUEB9EUCl_6SR FAGL 3D 88Q19LQQ1G8L_GQUlB9L ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 4.07 (4.00) c.
The ceram is verified by visual inspection of the HCU scram valves' position (0.5).
b.
1.
15 minutes (0.5) 2.
HPCI (0.25) 3.
RCIC (0.25) c.
Manually lift the relief valves in sequence (0.5).
Reduce the reactor pressure between 100 and 150 psig per actuation (0,5).
d.
Depressurize the reactor using all available pressure relief valves (0.5).
o.
The reactor vessel level reference legs will begin to flash when the pressure approaches the saturation pressure for drywell temperature, causing erroneous high reactor water level indication (1.03 REFERENCE CNS E0P 5.2.5.1, Loss of All Site AC Power Station Blackout, Rev 2, pg 1 ANSWER 4.08 (2.50) o.
- Reactor Building RCIC Pump Area
- Control Building Cable Spreading Area
- Turbine Building 4160V Switchgear Room
- Turbine Building Control Corridor, 882'
- Telephone Switchboard in the Administration Office Reactor Building 931'6" instrument racks (3 required at 0.5 each)
Deenergize the APRMs at the RPS Power Panels (0.5) b.
- Trip the Scram Discharge Volume High-High level switches (0.5)
REFERENCE CNS E0P 5.2.1, Shutdown From Outside the Control Room, Rev 12, pg 1,2
)
PAGE 31 6t__EB9CEQ9Bth_:_b2806Lt_6EUQBU6Lt_EUEB9EUEl_600 B6D10LQQ106L_QQUlB9L ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 4.09 (2.00)
- Scram the reactor
- Trip the turbine Isolate the RWCU system
-- Shut down both recirculation pumps and associated oil pumps when the MG sets have stopped (4 at 0.5 each)
REFERENCE CNS E0P 5.2.3, Loss of All Service Water, Rev 7, pg 1 ANSWER 4.10 (1.00)
Notification of Unusual Event Alert Site Area Emergency General Emergency (0.5 each)
REFERENCE CNS EPIP 5.7.1 ANSWER 4.11 (2.50) c.
" Rotor short" and " rotor long" conditions are caused by the difference in heating rates of the turbine casings and the turbine rotor (1.0).
b.
" Rotor short" - Increase the steam flow (0.5) to allow the casing to cool down to equal the rotor (0.25).
" Rotor long" - Decrease the steam flow (0.5) to allow the casing to heat up as much as the rotor (0.25).
REFERENCE CNS AP 2.4.9.1.9, Abnormal Turbine Temperature Expansion, Rev 6, p3 1,2
PAGE 3;
6t__EBQCEQUEEE_:_UQB56Lt_6EUQBU6Lt_EUEBEEUGl_6bD B6010LQQ196L_QQUIBQL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 4.12 (2.50) c.
The individual (0.5) b.
150 mrem / day 300 mrem / week 1000 mrem /qtr 3000 mrem /yr (0.5 each)
REFERENCE CNS HPP 3.1.1.1, Radiation Protection at Cooper Nuclear Station, Rev 4, pg 1 CNS HPP 9.1.2.1, Radiation, Contamination, and Airborne Radioactivity Limits, Rev 14, pg 5 l
1
U.
S.
NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
_CQQEEB__________________
REACTOR TYPE:
_RWB-QEi_________________
DATE ADMINISTERED: _RELQh411________________
EXAMINER:
_QB6VEft_Qt______________
i APPLICANT:
IN11BUCIl0N1_IQ_eEELICANIl one side only.
Use separate paper for the answers.
Write answers Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at.least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY
% OF APPLICANT'S CATEGORY
__V8LME_ _IQIaL
___1CQBE___
_V6LUE__ ______________Q8IEQQBl_____________
_25tD0__ _25100
________ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2110R__ _25tDQ
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_251Q0__ _25tDQ
________ 7.
PROCEDURES - NORMAL, t.BNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_25100__ _251QQ
________ 8.
ADMINISTRATIVE PROCEDURES,
' CONDITIONS, AND LIMITATIONS 1DQzaQ__ 10Qz00
________ TOTALS FINAL GRADE _________________%
-All work done on this examination is my own. I have neither given nor received aid.
APPLICAN1'S SIGNATURE v
~
- n.,,.,
,-,,-.c
-w.>
a
,,p.--,--,,---
St__lSE051_9E_buCLEEB_E9WLB_EL691_DEEB61100t_ELVIDLt_60D PAGE IBEBb90106blCL QUESTION 5.01 (1.50) c.
Are BWR cores designed to be OVERMODERATED or UNDERMODERATED?
(0.5) b.
What does the state of moderation, over-or under, assure about the reactivity control design of the core?
(1.0) 4 QUESTION 5.02 (2.00) o.
Which reactivity coefficient is of greatest significance during normal power operation?
(0.5) b.
Which reactivity coefficient has great safety implications during a very rapid increase in power?
EXPLAIN.
(1.5)
QUESTION 5.03 (3.00)
Indicate HOW (increase, decrease, unaffected) control rod worth changes for each of the situations listed below.
EXPLAIN WHY the rod worth is affected, if applicable.
O.
The reactor is heated from 100 deg F to 200 deg F (1.0) b.
Reactor power is increased from 20% to 40% by control rod withdrawal (1.0) c.
Reactor power is increased from 70% to 90% by increasing recirculation flow (1.0)
QUESTION 5.04 (3.00) c.
On large down power transients, such as a scram, what limits the rate of power decrease?
(0.5) b.
Describe the relationship between B (Beta), B "bar",
and B "bar" effective.
(1.5) c.
Describe how and why the response of the reactor to a positive reactivity change would be different if no delayed neutrons were present in the core.
(1.0)
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE *****)
Lt__IrLLEI_9E_UUCLL6t_LMbTE_EL601_0LEB61100t_LLUID11_6SD rAct lbEBU9Q106DICL QUESTION 5.05 (3.00)
The reactor has operated at 70% power for several weeks.
Power is then increased to 95% or rated using the recirculation system.
DOscribe the changes operators would have to make to recirculation flow over the next two days in order to hold reactor power at 95%
of rated.
Include WHEN the changes would have to be made and WHY.
(3.0)
QUESTION 5.06 (2.50)
The reactor is operating at full power when the RFP Master Controller colfunctions, resulting in a total loss of feedwater.
A reactor scram 10 expected to occur within seconds.
During this short period, is roactor power level expected to INCREASE or DECREASE?
Give TWO (2) rossons for your choice, including WHY each causes a change in in reactor power.
(2.5)
QUESTION 5.07 (2.00)
Why is nucleate boiling a better heat transfer condition than cubcooled heat transfer (2 reasons required)?
(2.0)
QUESTION 5.08 (2.00)
Match each of the four lettered items with one of the numbered items.
A letter-number sequence is sufficient.
(2.0) 1.
MAPRAT 5.
PCIOMR 2.
APLHGR 6.
GEXL 3.
CPR 7.
TOTAL PF 4.
FLPD 8.
_____e.
Parameter by which plastic strain and deformation are limited to less than 1%.
_____b.
Contains guidelines restricting power ramp rates above the threshold power.
_____c.
_____d.
(*****
CATEGORY 05 CONTINUED ON NEXT PAGE *****)
L 1_ _10 E 0 b1_9 E _U U C L ( 68_ E 0 b'E B _ E L 651_0 E E B 61129 t _ E L U 10 h t _600 PAGL 4
lbEBb99166b1Cs s
QUESTION 5.09 (3.00)
D: scribe HOW and WHY each of the following changes will effect roactor CRITICAL POWER.
If CRITICAL POWER will not be affected, otate this.
o.
Loss of extraction steam to a feedwater heater (1.0) b.
Mass flow rate through the core is increased (1.0) c.
Reactor pressure is increased (1.0)
QUESTION 5.10 (3.00) o.
Describe the heat transfer mechanism (s) in a typical BWR fuel rod at BOL (new fuel) and EOL (exposed fuel).
(2.5) b.
For the same heat flux, would the temperature at the fuel center-line be higher for new fuel or exposed fuel?
(0.5)
(***** END OF CATEGORY 05 *****)
L1__ Lit 'n 5T LIEts_DEE100t_CDulBOLt_6BD_1bElhUDEU16110';
FAG QUESTION 6.01 (2.00)
The DEH system is operating in Mode 4.
The plant is at 100% power.
Pressure setpoint value is 925 psi, actual steam pressure is 955 psi.
A. failure occurs in the inservice pressure controller (A), whereby its output fails low.
After the small transient on tha plant occurs cnd the plant reaches steady state, what will be the approximate values for the following parameters?
2.
DEH pressure setpoint (0.5) b.
Steam pressure input signal to the pressure controllers (0,5) c.
Governor valve demand (0.5) d.
Bypass valve demand (0.5)
QUESTION 6.02 (3.00) c.
What is the automatic starting sequence of the pumps in the Fire Protection System including the pressures at which they (1.0) start?
b.
What area (s) are protected by the High Pressure CO2 Fire Protection System?
(0.5) c.
Briefly describe the three (3) means of manual actuation of the High Pressure CO2 Fire Protection System.
(1.5)
QUESTION 6.03 (1.00)
At a voltage of 3000 V on the 4160 V buses, motors are subject to failure of the control fuses in their AC control circuits cnd overheating of the motors.
Why would low voltage on the bus make these events more likely to occur?
(1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
L:__ L'L 6bl_ flk il'_4_0L110bt_C NM R 06 t _65L'_lbElE k'CLU161190 I I1!
QUESTION 6.04 (2.50) o.
List three (3) ways that the Rod Block Monitor (RBM) may be bypassed.
Include automatic and/or manual bypasses.
(l.S) b.
How does the RBM utilize the input from a LPRM detector that is failed HIGH or failed LOW?
DISCUSS BOTH cases, but limit your answer to how the LPRM input is considered in the averaging and counting circuits.
Assume the LPRM recently failed and has NOT been bypassed with its individual bypass switch.
(1.0)
QUESTION 6.05 (4.00)
For each of the HPCI (High Pressure Coolent Inj ect ion) System component failures listed below, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel, IF IT WILL NOT INJECT WHY, AND IF IT WILL INJECT, provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of system operation with the failed component.
Assume NO OPERATOR ACTION, and the component is in the failed condition at the time HPCI receives the auto initiating signal.
c.
The GLAND SEAL EXHAUSTER fails to operate.
(1.0) b.
The turbine AUXILIARY LUBE OIL PUMP fails to operate.
(1.0) c.
The MINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when system conditions require it to be open.
(1.0) d.
The HPCI PUMP DISCHARGE FLOW ELEMENT OUTPUT SIGNAL to the HPCI flow controller is FAILED AT ITS MAXIMUM output.
(1.0)
QUESTION 6.06 (3.50) a.
What provisions are made in the Offgas System to DETECT en explosion or burn-back?
(1.0) b.
If en explosion or burn-back is detected, what automatic action takes place in the Offgas System?
(0.5) c.
What are three ways that an Offgas System isolation can be accomplished (with the exception of individually closing each isolation valve)?
Include any time delays that occur as appropriate.
(2.0)
(*****
CATEGORY 06 CONTINUED ON NEXT PAGE *****)
1
~
Lt__EL6Bl_L1hlEUE_DESIEUt_00blEOLt 6BR_lbElEUGEU1611gy rAct QUESTION 6.07 (3.50) i 0.
HOW and WHY does the on-line CRDH flow control valve respond following a scram?
(1.0) b.
What prevents the CRD pumps from reaching a " runout" condition following a scram?
(0.5) c.
A control rod is inserted one notch.
The insert stabilizing valve does not reopen after the rod movement.
How does this affect the indicated drive water differential pressure and on-line flow control valve position?
EXPLAIN.
(2.0)
QUESTION 6.08 (2.00)
The reactor is operating at 100% under steady state conditions.
An instrument technician mistakenly isolates and equalizes the pressure across one of the MAIN STEAM line flow transmitters.
DESCRIBE the response of the Feedwater Control System until steady state conditions are again established.
ASSUME 3-ELEMENT CONTROL.
(2.0)
QUESTION 6.09 (3.50) e.
The reactor is at 400 psig.
Describe how the Core Spray inj ect ion valves (MO-ll and MO-12) must be operated to nave them both open at the same time.-
(0.5) b.
The reactor is at 500 psig.
Can the Core spray inj ect ion valves (MO-ll and MO-12) both be opened at the same time?
If so, describe how they must be opened.
If not, can EITHER of the two valves be opened with no injection signal present?
EXPLAIN.
(1.0) c.
What is the meaning of the white light above the MO-ll and MO-12 control switches?
Once illuminated, what THREE (3) conditions will cause the condition to clear?
(2.0)
(*****
END OF CATEGORY 06 *****)
N4E e
Zz__EEQCE09EES_ _URS06Lt_6EU985ELt_EtEEEEUC1_6BD B6D19LQQ1G6L_GQUlBQL QUESTION 7.01 (3.50) o.
Who has the ultimate responsibility for an individual's safety with regard to radiation protection?
(0,5) b.
What are the Cooper Nuclear Station ADMINISTRATIVE limits for whole body penetrating radiation exposure for a NPPD employee?
(2.0) c.
What is the MAXIMUM amount of whole body penetrating radiation the following individual can receive in 1985 (TOTAL DOSE for the year) per 10 CFR 20.
Assume all required paperwork and approvals are obtained.
Explain how you arrived at your answer.
(1.0) 35 year old male dose received so far this quarter:
500 mrem dose received so far this year:
1200 mrem lifetime dose:
3500 mrem QUESTION 7.02 (3.50)
Answer the following with regard to the primary containment:
e.
During a reactor and plant startup, when must the Oxygen concentration be less than 4% ?
(1.0) b.
The reactor shall be scrammed if suppression pool temperature reaches ________.
(0.5) c.
Power operation shall not be resumed until the pool temper-ature is reduced below (0.5) d.
What allowance is made for testing that adds heat to the suppression pool regarding pool temperature?
(0,5) o.
During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than ________ psig at normal cooldown rates if the pool temperature reaches ________.
(1.0)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Z:__EBQGEDUBEL_:_URBU6L _6BURB06Lt_EUEBREUGl_600 F' AGE 9
B6010LQElG8L_CQUlBQL QUESTION 7.03 (2.00)
MATCH each of-the events (a - d) with the pressure at which the ovent may be performed during a cold plant startup per GOP 2.1.1,
" Cold Startup Procedure".
Items may be used more than once or not at-all as appropriate.
(2.0)
_____a.
Begin placing feed pump in service 1.
25 psig
_____b.
Open HPCI steam isolation valves 2.
50 psig
_____c.
Steam seals placed in service 3.
100 psig
_____d.
Roset and unisolate RCIC 4.
150 psig 5.
350 psig QUESTION 7.04 (2.00)
The reactor was operating at high power when an MSIV isolation and resulting scram occurs.
a.
Per procedure 2.1.8, Scram Recovery During Power Operation MSIVs Closed, what are two (2) methods by which the control rods are verified inserted?
(1.0) b.
What is the preferred method of controlling the reactor pressure and cooldown?
(0.5) c.
If the preferred method causes the cooldown rate to be excessive, what system should be used?
(0.5)
QUESTION 7.05 (2.00)
What are two (2) undesirable consequences of a loss of both Control Rod Drive Hydraulic pumps?
Include WHY the consequences are undesirable.
(2.0) 4 QUESTION 7.06.
(1.50)
An unexpected and unexplained increase in reactor power has occurred.
AP 2.4.1.3, Unexplained Increase in Reactor Power, requires reactor power to be reduced.
HOW does the procedure sey to reduce power and WHY is that method used?
(1.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
Z.
PRocEDUBES_ _NQB06Lt_669QB06(t_[g[80[U(1_65;.
Pact B6010LQElQ6L_QQNIBQL QUESTION 7.07 (3.50)
A loss of all site AC power has occurred. Answer the following questions concerning E0P 5.2.5.1, Loss of All AC Power Station Blackout.
c.
How is the reactor scram verified?
(0,5) b.
FILL IN THE BLANKS.
If it can be ascertained that AC power WILL BE restored within __(1)__ minutes, maintain reactor level with __(2)__
or __(3)__.
(1.0) c.
Describe how reactor pressure is controlled if AC power CANNOT be restored within X minutes.
(1.0) d.
Why should reactor pressure NOT be reduced below the saturation pressure corresponding to the maximum drywell temperature?
(1.0)
QUESTION 7.08 (2.00)
A loss of the Service Water System has occurred and it is not cxpected that the system will be restored within a short period of time.
What are four (4) actiona that must be performed?
(2.0)
QUESTION 7.09 (2.50) c.
What causes " rotor short" and " rot.or long" conditions?
(1.0) b.
HOW should steam flow be changed (increased or decreased) to correct " rotor short" and " rotor long" conditions and WHY does the change in steam flow correct the situation ?
(1.5)
QUESTION 7.10 (2.50) c.
What are three (3) areas or locations that will be manned by operations personnel following an evacuation of the control room?
(1.5) b.
If the reactor was not scrammed prior to leaving the control room, what are two (2) alternate methods of scramming the reactcr per E0P 5.2.1, Shutdown from Outsid( the Contrcl Room?
(1.0) l
(*****
END OF CATEGORY 07 *****)
l t
e t_ _6051 h l118611 V E _ E B 9 C E D L'B E L t _90 G D 11100 E t _6U D _ L 1511i1190 L W4L i
I l
QUESTION 8.01 (2.00)
Give two (2) general examples of NONCOMFORMANCE conditions that could require completion of a Noncomformance Report (specific oxamples not required).
(2.0)
QUESTION 8.02 (2.50)
What are the four (4) basic types of air breathing apparatus Ovailable for use at Cooper Nuclear Station and WHICH one(s) coy be used in a NON-LIFE SUPPORTING ATMOSPHERE?
(2.5)
QUESTION 8.03 (3.00)
Match the following hazardous materials (a - d) with the applicable hazard (s) or characteristic (s).
Hazards or characteristics may be used more than once or not at all.
(3.0)
_____a.
Sulfuric acid 1.
As found in general use,
_____b.
Sodium hydroxide has a garlic-like odor.
i
_____c.
Acetylene 2.
May be absorbed through
_____d.
Mercury the skin 3.
Can cause severe skin damage on contact 4.
Is flammable or combustible i
QUESTION 8.04 (2.00) c.
When is an independent verification that a Clearance Order is implemented properly or returned to service properly generally required?
(1.0) i b.
Who determines if an independent verification is required?
(0.5) i c.
During outages when the reactor is in cold shutdown, the independent verification when returning the equipment to service will not be required on the Clearance order.
How will the requirement be fulfilled?
[0.5) 2-4
(***** CATEGORY.08 CONTINUED ON NEXT-PAGE *****)
D2_ 6D5101218611YE EB9CEDUBELt_00b0ll10Uit 6bO LIUll6110Uh N'
QUESTION 8.05 (2.00)
Given the following work schedule for en operator, identify the violations of the CNS guidelines for administering overtime.
DAY TIMES WORKED 1
0800 - 1600 2
0800 - 2000 3
0800 - 2400 4
1000 - 1800 5-8 0800 - 2000 9
0800 - 1800 10 0000 - 0800 11-14 0800 - 1600 15 0800 - 2000 QUESTION 8.06 (3.00)
O.
The Operations Department has two (2) key depositories.
What types of keys are in each of the depositories?
(1.0) b.
Who has control over key checkout in each of these depositories?
(1.0) c.
During power operation, where is the key to the Reactor Mode Switch kept?
(0.5) d.
The Reactor Mode Switch key is under the direct control of whom?
(0.5) 1 QUESTION 8.07 (1.50) l A licensee may take reasonable action that departs from a condition or a Technical Specification in an emergency when this action is immediately need,ed to protect the public health and safety and no cction consistent with license conditions and Technical Specifications that can provide adequate or equivalent protection is immediately Cpparent.
c.
Such an action shall be approved, as a minimum, by whom?
(0.5) b.
What two (2) notifications should be made prior to the above action if at all possible, but always as soon as possible afterwards?
(1.0)
(*****
CATEGORY 08 CONTINUED ON NEXT PAGE *****)
6:__60DIb1118611YE_EBOCEDUBEht_C90Dil109ht_60D_Litll611Lb_
f*0L QUESTION 8.08
(.50) j i
Fill in the blank.
In accordance with 10 CFR 55, "if a licensee hts not been actively performing the functions of an operator or conior operator for a period of _____ months or longer, he shall, prior to resuming activities licensed pursuant to this part, demonstrate to the. Commission that his knowledge and understanding of facility operation and administration are satisfactory."
(0.5)
QUESTION 8.09 (1.50)
O.
Who is responsible for preparing Scram Reports?
(0.5) b.
When is a Scram Report required to be prepared?
(1.0)
QUESTION 8.10 (3.00)
Conduct of Operations Procedure 2.0.6, Reactor Post Trip Review cnd Restart Authorization Procedure, lists five (5) criteria that cust be satisfies prior to restarting the reactor following a coram.
LIST THREE (3) of these criteria.
(3.0)
QUESTION 8.11 (2.00)
Per the Technical Specifications Limiting Safety System Settings, chat are four (4) protective actions designed to prevent exceeding the Reactor Coolant System pressure safety limits?
SETPOINTS NOT REQUIRED.
(2.0)
QUF, i T I ON 8.12 (2.00)
State the LICENSED MANNING requirements per the CNS Technical Specifications for each of the following situations.
Indicate thether the individual (s) must be in the control room or just present at the station.
c.
Cold Shutdown (core loaded)
(1.0) b.
During a reactor startup.
(1.0)
(*****
END OF CATEGORY 08 *****)
1 R*************
END OF RXAMINATION ***************1
[
L t _ _ I B E D E l_ D E _ U U C L E 6 B _ E 0 h'E E _ E L 6 bl _0 E E E 611001_ E L U I D i t _60 p PA6L
- 4 lbEBUQQ19601C1 ANSWERS -- COOPER
-85/06/ll-GRAVES, D.
ANSWER 5.01 (1.50) o.
Undermoderated (0.5) b.
This ensures that the moderator temperature coefficient (0.25) and the void coefficient (0.25) are negative (0.5).
REFERENCE GE Reactor Physics Review, pg 26, Fig 33 CNS Question Bank, Question VI-A-8 ANSWER 5.02 (2.00) o.
Void coefficient (0.5) b.
Fuel temperature coefficient (0.5).
On a power increase, fuel temperature increases almost immediately (0.5) adding negative reactivity (0.5).
The fuel temperature. coefficient can change reactivity before any of the other reactivity coefficients can respond.
REFERENCE GE Reactor Physics Review,Section I, Coefficients of Reactivity ANSWER 5.03 (3.00) o.
Increase (0.25).
As the moderator density decreases, the neutrons travel a greater distance and are more likely to interact with a control rod (0.75).
b.
Decrease (0.25).
The voids depress the thermal neutron flux which in turn decreases the rod worth (0.75).
o.
Increase (0.25).
Rod worth decreases as void content increases.
A slight decrease in void content means the thermal flux is not quite as depressed and control rod worth subsequently increases (0.75).
REFERENCE GE Reactor Theory Review, pg 37
L1__lUE981_9E_UUCLL6b_E0WLE_EL6Sl_0 ELE 611kb2_LLUID1t_6UD PAM it IUEBUQQ1U601G1 ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 5.04 (3.00) c.
The rate of power decrease is limited by the longest lived (0.25) l delayed neutron precursor (0.25).
b.
B - The fraction of fission neutrons born delayed (0.25) for a given isotope (0.25).
B "bar" - Weighted average of the various fuels' Bs (0.5).
8 "bar" effective _ The fraction of the thermal neutron population that was born delayed (0.5).
c.
With no delayed neutrons present, the neutron generation time would be the same as the prompt neutron generation time, and the rate of power change would be uncontrollably rapid (1,0).
REFERENCE GE Reactor Physics Review, pg 21-22 ANSWER 5.05 (3.00)
During the first 4 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (0.5) recirculation flow would have to bo decreased (0,5) to compensate for the decrease in xenon concen-tration (0.5).
During the remainder of the transient, recirculation flow would have to be increased (0.5) as xenon builds up to its equilibrium value (0.5) over the next 40 - 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> (0.5).
REFERENCE 1
GE Reactor Physics Review, pg 45
l L __ISE981_0E_UUCLL6B_E0b'EB_EL6Bl_9tLB61100t _LLui m,_690 st u
IbEBdQQ166dlGE ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 5.06 (2.50)
DECREASE.(0.5]
REASONS:
1.
Immediately, the loss of feedwater flow causes a decrease in moderator subcooling which introduces negative dk/k into the core.(1.0]
2.
When feedwater flow drops below 20% (~3 sec into trans.)
the recirc pumps auto runback to 20%. The decrease in core flow causes an increase in voiding which also adds neg. dk/k into core.(1.03 REFERENCE I
CNS Recirculation System Desctiption GE Reactor Physics Review, Section 1, Coefficients of Reactivity i
ANSWER 5.07 (2.00)
A large amount of heat is removed in changing the water to steam (heat of vaporization) [1.0].
The production of bubbles breaks up the laminar layer, creating agitation, which provides better mixing (1.03 REFERENCE GE Thermodynamics, Heat Transfer, and Fluid Flow, pg 9-8 ANSWER 5.08 (2.00) o.
8 b.
5 c.
1 d.
4 (0.5 each)
REFERENCE GE Thermodynamics, Heat Transfer, and Fluid Flow, Chapter 9 i
+~
52__lbE9bl_9E_UVELl68_E9 WEE _EkiUl_90LFAlIUNt_LLUlDlt_6bD PAGE 1^
lbEEDQQ16601CL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 5.09 (3.00) e.
CRITICAL F0WER increases (0.25).
A greater enthalpy rise is required to bring the coolant to saturation conditions resulting in less steam generation for a given power level (0.75) or similar.
b.
CRITICAL POWER increases (0.25).
Same reason as above (0.75),
c.
CRITICAL POWER decreases (0.25).
A lower enthalpy rise is required to change a given mass of coolant from liquid to vapor at a higher pressure (0.75).
REFERENCE GE Thermodynamics, Heat Transfer and Fluid Flow, pg 9-85 through 9-87 ANSWER 5.10 (3.00)
O.
New fuel heat transfer mechanism:
Natural convection from pellet across helium filled gap to cladding (0.5), conduction through cladding (0.25), convection from cladding to coolant (0.5).
Exposed fuel heat transfer mechanism:
Conduction from pellet to cladding (0.5), conduction through cladding (0.25),
convection from cladding to coolant (0.5).
b.
Higher for new fuel (0.5)
REFERENCE GE Thermodynamics, Heat Transfer, and Fluid Flow pg 8-64
Ez__EL6Bl_EXElEUE_QEhlggt_COUIE9L _60D_lG1189BEU1611Qi, PAGE 16 ANSWERS -- COOPER
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ANSWER 6.01 (2.00) c.
925 psi b.
958 psi c.
100%
d.
0%
(0.5 each)
REFERENCE CNS DEH System Description, pg 7, Fig. 4 CNS Question Bank, Book II, Section D, Question 2 CNS AOP 2.4.9.1.11, DEH Pressure Controller Output Fails Low ANSWER 6.02 (3.00) o.
jockey fire pump - 120 psig electric fire pump E-110 psig diesel fire pump D - 105 psig electric fire pump C - 68 psig (0.1 for each setpoint, 0.2 for correct order, 0.1 for correct number of pumps) b.
D/G rooms (0.25)
Day tank rooms (0.25) c.
- Actuation of the MAIN manual pneumatic release bottle (0.5)
- Actuation of the RESERVE manual pneumatic release bottle (0.5)
- Actuation of the manual pneumatic release bottle in the D/G room (0.5)
REFERENCE o.
CNS SOP 2.2.30, Fire Protection System, Rev 23, pg 8 b.
and c.
_CNS SOP 2.2.2, Carbon Dioxide Systems, Rev 14, pg 2,6 ANSWER 6.03 (1.00)
The bus loads draw excessive current necessitated by the low bus voltage (1.0).
REFERENCE CNS AOP 2.4.6.2, Startup Station Service Transferrer Failure / Loss of 161 KV Line, Rev 7, pg 4 l
bz__ELEUl_11SIE01_0L119bt_COSIB0li_6UD_lSilEUGEU16110b PAGE l ':
ANSWERS -- COOPER
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ANSWER 6.04 (2.50) c.
- Manual operation of the RBM BYPASS switch (0.5)
- < 30% power (0.5)
- Edge rod selected (0.5) b.
Failed Low:
Removes the LPRM input from the averaging circuit (0.25) and provides a signal to the counting circuit (0.25).
Failed High:
The higher input is averaged with the other inputs and processed as if it were a valid signal (0.5)
REFERENCE CNS Rod Block Monitor System Description, pg 5 ANSWER 6.05 (4.00) a.
Will inject (0.25).
Turbine seal leakage resulting in potential air-borne activity in the HPCI room (0.75).
(1.0) b.
Will not inject (0.25).
Turbine stop and control valves will not open (0.75).
(1.0) c.
Will inject (0.25).
Pump overheating and seal damage may result during low or no flow conditions (0.75).
(1.0) d.
Will not inj ect (0.25).
Maximum signal from the flow element will cause the controller to keep turbine speed at minimum (0.75).
(1.0)
REFERENCE CNS HPCI System Description, pg 9,12,14 ANSWER 6.06 (3.50) o.
Pressure (0.33) and temperature (0.33) sensors in the SJAE i
discharge piping (0.33).
b.
The SJAE inlet isolation valves shut (0,5).
c.
Offgas rad monitors (0.5) with a 15 minute time delay (0.25) j Dilution fan low flow (0.5) with a 5 minute time delay (0.25) j Offces timer switch to CLOSE (0.5) with no time deley.
1
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-85/06/ll-GRAVES, D.
REFERENCE CNS Offgas and Augmented Offgas Systems Description, pg 5-7 i
ANSWER 6.07 (3.50) s.
The flow control valve will go to its minimum position (0.5).
The sensed flow increases as charging flow to the accumulators increases.
The increased flow causes the FCV to close, diverting most of the flow to the accumulators (0.5).
b.
A flow orifice in the charging line limits the pump flow to prevent reaching a runout condition (0.5).
c.
Flow control valve position remains unchanged (0.5) because total system flow has not changed (0.5) OR Flow control valve will open slightly (0.5) to overcome the increased differential pressure (0.5).
The indicated differential pressure will increase (0.5) because flow previously bypassing the drive water pressure control valve is now passing through it (0.5).
REFERENCE CNS CRD System Description, pg 5,6, Figure 1 ANSWER 6.08 (2.00) o.
Total steam flow would indicate 75% instead of the actual 100% (0,5).
The FWLCS would assume steam flow has decreased and reduce feed flow accordingly (0.5).
Level will begin to decrease.
As level con-tinues to decrease, a level error signal is generated which will increase feed flow back to the 100% value (0.5).
A new steady state level will be established lower than the original level.
Will accept any similar explanation (0.5).
REFERENCE CNS Feedwater Control System Description, pg 4,5 i
l l
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-85/06/ll-GRAVES, D.
~
ANSWER 6.09 (3.50) o.
MO-ll must be opened first (0.5) b.
No (0.25), Yes (0.25), either of the two valves can be opened, but both cannot be open at the same time (0.5).
c.
The white light means that only manual operation of the valve is allowed (0.5).
Also accept,that the valve is shut and an initiation signal is present.
The condition can be cleared by:
- termination of the auto initiation signal (0.5) reactor pressure increases to > 450 psig (0.5)
- loss of power to the pump motor bus.(0.5)
REFERENCE CNS Core Spray System Description, pg 6 l
y
')
..o
>+
t J
% /
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B6019LQE196L_CQGlBQL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 7.01 (3.50) o.
The individual (0.5) b.
150 mrem / day 300 mrem / week 1000 mrem /qtr 3000 mrem /yr (0.5 each) c.
1200 mrem this year minus 500 mrem this quarter means that 700 mrem was received in the first quarter.
This quarter he can receive 3000 mrem, as he can in each of the last 2 quarters.
He can end up with a total yearly dose of 9700 mrem maximum (1.0).
REFERENCE CNS HPP 9.1.1.1, Radiation Protection at Cooper Nuclear Station, Rev 4, pg 1 CNS HPP 9.1.2.1, Radiation, Contamination, and Airborne Radioactivity Limits, Rev 14, pg 4,5 ANSWER 7.02 (3.50) c.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (0.5) after placing the Mode Switch in RUN (0.5) b.
110 deg F (0.5) c.
normal power operational limit OR 90 deg F (0.5) d.
Pool temperature must be kept below the normal operational limit + 10 deg F (0.5).
o.
- 200 (0.5)
- 120 deg F (0.5)
REFERENCE CNS SOP 2.2.60, Primary Containment Cooling and Nitrogen Inerting System, Rev 27, pg 6,7 ANSWER 7.03 (2.00) c.
5 b.
3 c.
I d.
2 (0.5 each) l
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-85/06/11-GRAVES, D.
REFERENCE CNS GOP 2.1.1, Cold Startup Procedure, Rev 38, pg 8,9 ANSWER 7.04 (2.00) e.
- All green full-in lights on the 9-5 panel are illuminated (0.5)
Place the reactor mode switch in REFUEL C0.25) and verify the white REFUEL PERMISSIVE light is illuminated (0.25) b.
The preferred method is starting HPCI in the test mode (0.5).
c.
RCIC (0.5)
REFERENCE CNS GOP 2.1.8, Scram Recovery During Power Operation MSIVs Closed, Rev 9, pg 2,3 ANSWER 7.05 (2.00)
- Failure of the CR0 pumps stops cooling water to the drives (0.5),
shortening seal life (0.5).
- Failure of the CRD pumps also allows the CRD accumulators to slowly depressurize (0.5).
At low reactor pressures, the accumulators are required to ensure that the control rods are fully scrammed within the required time (0.5).
REFERENCE CNS AP 2.4.1.1.4, Loss of CR0 Pump, Rev 5, pg 2 ANSWER 7.06 (1.50)
Reactor power is reduced with recirculation flow (0.5).
Reduction of power by reducing recirculation flow is the only way to reduce neutron flux patterns all across the core OR M-sing the control rods in other than an approved sequence may worsen the problem (either acceptable) (1.0).
REFERENCE CNS AP 2.4.1.3, Unexplained Increase in Reactor _ Power, Rev 8, pg 1,3
2:___B00EDUBES_:_b2856L1_5EU9806Lt_EUEEEEUC1_600 FMr EsR10LQQ1G6L_CQUlBQL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 7.07 (3.50) o.
The scram is verified by visual inspection of the HCU scram valves' position (0.5).
b.
1.
15 minutes (0.5) 2.
HPCI (0.25) 3.
RCIC (0.25) c.
Depressurize the reactor using all available pressure relief valves (0.5), and maintain 100 to 150 psig using RCIC as long as possible (0.5).
d.
The reactor vessel level reference legs will begin to flash when the pressure approaches the saturation pressure for drywell temperature, causing erroneous high reactor water level indication (1.0).
REFERENCE CNS E0P 5.2.5.1, Loss of All Site AC Power Station Blackout, Rev 2, pg 1 ANSWER 7.08 (2.00)
- Scram the reactor
- Trip the turbine
- Isolate the RWCU system
- Shut down both recirculation pumps and associated oil pumps when the MG sets have stopped (4 at 0.5 each)
REFERENCE CNS EOP 5.2.3, Loss of All Service Water, Rev 7, pg 1
22__EBQCEDVBLE : UOBU6Lt SEb2EU6Lt EUEBGEUCl 6UD PAGL M
B6Q10LQQlR6L GQUlBQL ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 7.09 (2.50) o.
" Rotor short" snd " rotor long" conditions are caused by the difference in heating rates of the turbine casings and the turbine rotor (1.0).
b.
" Rotor short" - Increase the steam flow (0.5) to allow the casing to cool down to equal the rotor (0.25).
" Rotor long" - Decrease the steam flow (0.5) to allow the casing to heat up as much ai the rotor (0.25).
REFERENCE CNS AP 2.4.9.1.9, Abnormal Turbine Temperature Expansion, Rev 6, pg 1,2 ANSWER 7.10 (2.50)
Reactor Building RCIC Pump Area c.
- Control Building Cable Spreading Area Turbine Building 4160V Switchgear Room
- Turbine Building Control Corridor, 882'
- Telephone Switchboard in the Administration Office
- Reactor Building 931'6" instrument racks (3 required at 0.5 each) b.
- Deenergize the APRMs at the RPS Power Panels (0.5)
- Trip the Scram Discharge Volume High-High level switches (0,5)
REFERENCE CNS E0P 5.2.1, Shutdown From Outside the Control Room, Rev 12, pg 1,2 4
c-
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ANSWER 8.01 (2.00)
- A' deficiency in physical characteristics of a system or component which renders the item unable to perform to the design intent and operating license commitment for that system or component.
- Lack of required documentation to assure that the component in-service conforms to requirements.
- Noncomformance with an authorized operating procedure or instruc-tion.
- A deficiency in administrative cont.ols intended to meet the Q.
A.
Program commitments or NRC Regulatory criteria.
(Any 2 at 1.0 each, wording need not be exact, but concepts should be cimilar)
REFERENCE CNS Administrative Procedure 0.5, Nonconformance and Corrective Action, Rov 1, pg 1 ANSWER 8.02 (2.50)
- Self-contained (0.5) may be used in a non-life supporting atmosphere (0.5).
- Air line (0.5)
- Oro/ nasal respirator (0.5)
Full face canister respirator (0.5)
REFERENCE CNS Administrative Procedure 0.6, Personnel Safety, Rev 0, pg 6 ANSWER 8.03 (3.00) c.
3 b.
3 c.
1,4 d.
2 (G at 0.6 each)
REFERENCE CNS Administrative Procedure 0.7, Hazardous Meterial Control, Rev 0, pg 1-5
a tz__60010111ELIIVE_E80GEDUEESt_CQUDI110 bit _600_Llull61190L F%GE 27 ANSWERS -- COOPER
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l l
ANSWER 8.04 (2.00)
O.
When valves, breakers, switches, or other components in the main flow path of a safety related system are tagged in or out of service (1.0).
b.
Shift Supervisor (0.5) c.
The verification will be covered by the system valve lineup prior to startup (0.5).
REFERENCE CNS Administrative Procedure 0.9, Equipment Clearance and Release Orders, Rov 2, pg 2-3 ANSWER 8.05 (2.00)
- >14 consecutive days worked without a day off (0.5)
- Days 2 and 3:
More than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> worked in 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period (0.5)
- Days 9 and 10:
Less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> allowed between work shifts (0.5)
- Days 1 - 7 (0.1), days 2-8 (0.1), days 3 - 9 (0.1), and days 4 - 10 (0.13:
More than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> worked in a 7 day period (0.1).
REFERENCE CNS Administrative Procedure 0.12, Station Overtime and Recall of Standby Personnel, Rev 0, pg 1 ANSWER 8.06 (3.00) a.
One depository, #1, contains keys necessary for operational controls, area access doors, and panel doors (0,5) or similar type wording.
The other depository, #2, will contain keys necessary for operation of systems bypass switches normally within the control room (0.5).
b.
Depository #1 is under the control of the Shift Supervisor (0.5).
Depository #2 is under the control of the Shift Supervisor (0.25) and the Control Room Operators (0.25).
c.
The key will be inserted into the Mode Switch (0.5) d.
Reactor Operator (0.5)
REFERENCE CNS Administr ative Procedure 1.3, Key Control, Rev 0, pg 2
8t__6001UIEIB611YE_EB9CEDVBEEt_99BDil190Et_6UD_L1til611995 PAGE 20 ANSWERS -- COOPER
-85/06/11-GRAVES, D.
ANSWER 8.07 (1.50) c.
The action shall be approved by a licensed Senior Operator as a minimum (0.5).
b.
- Division Manager of Nuclear Operation (0.5)
- NRC (0.5) Operations Center REFERENCE CNS Conduct of Operations Procedure 2.0.1, Rev 1, pg 1,2 ANSWER 8.08
(.50)
Four (0.5)
REFERENCE 10 CFR 55.31 ANSWER 8.09 (1.50) c.
Shift Supervisor (0.5) b.
A Scram '.eport is required to be prepared following a scram I
when fuel is in the reactor (0.5) and more than one operable Control rod is withdrawn (0.5).
REFERENCE CNS Conduct of Operations Procedure, 2.0.2, Operations Logs and Reports, Rev 1, pg 5 ANSWER 8.10 (3.00) i The plant is in a-safe condition.
The cause of the scram is understood or it is attributed to a spurious trip and is unlikely to reoccur.
- Corrective action has been identified and appropriately implemented.
- The proper automatic operation of plant safety-related systems has been observed.
The Division Manager of Nuclear Operations approves the restart I
of the plant.
l (3 required at 1.0 each)
o.=*
Ex-6Dbibl118611k'E EBQCEDUBEEt G002111995t 6BD Lib 116119Uh PAC,E 20-ANSWERS -- COOPER
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REFERENCE
-Conduct of Operations Procedure 2.0.6, Reactor Post Trip Review and Restart Authorization Procedure, Rev 0, pg 6-7 ANSWER 8.11 (2.00)
- Reactor vessel high pressure scram
- Relief valve actuations
- Safety valve actuations Shutdown cooling valve isolation on high pressure (0.5 each)
REFERENCE CNS Technical Specification 1.2 CNS Technical Specification 2.2 ANSWER 8.12 (2.00) 1 RO (0.25) in the control room (0.25) o.
1 SRO (0.253 present at the station (0.25) b.
- 2 R0s (0.25) in the control room (0.25) 1 SRO (0.25) either in the control room or immediately available to the control room (0.25)
NOTE:
The above Technical Specification requirements are less restrictive than the manning requirements the plant actually uses (Conduct of Operations Procedure 2.0.3 and the Federal Register, Volume 48, No. 133 (July 1983).
REFERENCE CNS Technical Specification 6.1.3 i
-e v...
.,.