ML20211N625
| ML20211N625 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/23/1986 |
| From: | Burdick T, Dave Hills, Lanksburg R, Mcmillen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20211N614 | List: |
| References | |
| 50-409-OL-86-01, 50-409-OL-86-1, NUDOCS 8607030170 | |
| Download: ML20211N625 (93) | |
Text
r U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-409/0L-86-01 Docket No. 50-409 License No. DPR-45 Licensee: Dairyland Power Cooperative 2615 East Avenue - South Lacrosse, WI 54601 Facility Name:
Lacrosse Boiling Water Reactor Examination Administered At:
Lacrosse Examination Conducted: May 14-15, 1986 Examiners-
. D.
(./ 2.1/M.
Date s
Hi s
Date 96i J. I. McMillen 2 I/
'Date
[Z 7/d Approved By:
r Operating Licensing Date Section Examination Summary Examination administered on May 14-15, 1986 (Report No. 50-409/0L-86-01). )
Examinations were administered to one reactor operator candidate and one senior reactor operator candidate.
In addition, a written examination was administered to one additional reactor operator' candidate.
Results: One reactor operator candidate passed the written and operating examinations.
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8607030170 860627 PDR ADOCK 05000409 V
e.
REPORT DETAILS 1.
Examiners R. D. Lanksbury, NRC D. E. Hills, NRC J. I. McMillen, NRC 2.
Examination Review Meeting At the conclusion of the written examination, the questions and answers were given to the facility training staff for review and comment.
Subsequent to this, on May 21, 1986, the licensee provided their comments to the Reactor Operator and Senior Reactor Operator written examinations.
The facility comments and the examiner's resolutions to each are on the enclosed attachment.
3.
Candidate Qualifications During the review of the candidates NRC Form 398, Personnel Qualifications Statement-Licensee, questions were raised as to the qualifications of the two Reactor Operator (RO) candidates to sit for the exam. The questions related to the training these candidates had received as reported on NRC Form 398. The licensee was requested to submit detailed documentation substantiating the reported training.
The requested information was submitted, however, upon review by the examiner it was found that it did not provide sufficient data to corroborate the reported number of hours of classroom training. Subsequently, the licensee was invited to come to the regional office to describe how the candidates met the minimum requirements for training to be licensed as a RO.
This working level meeting took place on May 7, 1986. The licensee presented an overview of their training program and their explanation as to how the candidates met the minimum qualifications to sit for a R0 exam.
Region III presented an overview of what information was to be provided on NRC Form 398. At the conclusion of the meeting, the Region III personnel attending indicated that sufficient information had been provided to clarify the two RO candidates qualifications to sit for the exam.
During the course of the operating exam for the Senior Reactor Operator (SRO) the examiner questioned the candidate as to the amount of on-the-job training (0JT) he had received as a Shift Supervisor (SS) and as an SRO.
The candidate indicated that he had received approximately four hours DJT as a SS and none as an SRO. This amount of DJT appeared to differ substantially from that reported on the original NRC Form 398 for this candidate. As a result of this finding, additional inspection in the licensed training area was conducted. The results of this inspection are documented in Inspection Report No. 50-409/86007. As a result of this additional inspection, it was determined that the SRO was not qualified to sit for the SR0 examination.
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r-4.
Exit Meeting On May 7, 1986, a working level meeting was held in Region III as described in Paragraph 3.
The following licensee personnel were present at this meeting.
L. Kelly, Training Supervisor J. Parkyn, Plant Superintendent The following NRC personnel were present at this meeting:
C. Hehl, Chief, Operations Branch T. Burdick, Chief, Operator Licensing R. Lanksbury, Chief Examiner S. Hare, Reactor Inspector On May 15, 1986, an exit meeting was held. The following licensee personnel were present at this meeting:
L. Kelly, Training Supervisor J. Parkyn, Plant Superintendent -
The examiners briefly discussed the apparent problem with the training of the SRO candidate.
(See Section 3 above).
3
r-FACILITY EXAMINATION COMMENTS AND EXAMINER RESOLUTION FOR LACROSSE EXAMINATION ADMINISTERED MAY 14, 1986 1.
Question 1.11 Facility Comment:
Explain why it is necessary to continue cooling the core following a shutdown from Power Operation.
Answer 1.11 Should be: to remove decay heat.
Early in core life, LACBWR does not have enough decay heat to damage the core.
Resolution:
Comment accepted - Possibility of actual damage to core if decay heat is not removed depends on core conditions. Therefore, will accept "to remove decay heat" for full credit.
2.
Question 2.04 A Facility Comment:
The answer states as one of the systems is Seal Injection emergency backup.
This was removed in April 1986 by Facility Change 69-85-4.
Resolution:
Comment accepted - Documentation submitted with comment shows seal injection emergency backup removed in April 1986.
Since question only required listing two systems, removal of this part of answer does not affect overall grading on question.
3.
Question 3.02 0 Facility Comment:
Your answer states a partial scram for undervoltage on 2400-volt bus IA and
- 18. An undervoltage condition on IA and IB bus generates a full scram.
Reference:
LACBWR Operating Manual Volume 4, Page 6-4.
Resolution:
Comment accepted - A partial scram will occur on undervoltage on just one of these buses. Undervoltage on both buses will cause a full scram.
Therefore, the correct answer to this question is full scram.
4.
Question 3.09 Facility Comment:
Due to an error in the training manual, the Liquid Waste Monitor sample point is stated incorrectly.
It should be as per LACBWR Operations Manual Volume 10 Page 5-5.
"The full volume of liquid waste discharge passes through the sample chamber" prior to discharge into the service water line.
4
(
l i
Resolution:
Comment accepted - Documentation submitted with comments shows sample prior to discharge into service water line. This is contrary to what is contained in the training manual. Therefore, the correct answer will be
" Liquid Waste Discharge prior to discharge into service water line."
5.
Question 4.03 8 Facility Comment:
The stated answer is correct for flooding only. A rupture downstream of the redundant check valves generates a LOCA. Which is the worst possible failure?
Resolution:
Facility comment did not provide documentation that worst possible failure would be a rupture downstream of the redundant check valves generating a LOCA.
LACBWR Operating Manual, Volume 1, Page 4-35 states that "the worst potential failure of the Alternate Core Spray System is regarded to be a failure in the Electrical Penetration Room which causes loss of Electrical Switchgear Bus 1A."
However, this statement is in the equipment flooding procedure so as to make it a matter of interpretation as to whether the statement applies to only flooding type problems. Therefore will also accept facility answer although documentation was not provided.
6.
Question 4.08 A&B Facility Comment:
As per LACBWR Operating Manual Volume 1, Page 2-6, the high voltage is secured prior to saturation, and withdrawn as per procedure to prevent damage.
Resolution:
Comment not accepted -' Reference provided by facility with comment just mentions that high voltage is secured and source range monitors are withdrawn. This reference does not state that this is done to prevent damage as the comment states. The reference given by the original answer does indicate the reasons and the question was specifically worded to apply to these reasons.
7.
Question 5.08 Facility Comment:
Your answer is correct for relative rod worth. You did not ask for that.
Your answer to Question 1.04 is " Control rod worth is dependent upon neutron flux.
Reference:
Standard Nuclear Theory.
5
3 Resolution:
Control rod worth is dependent upon neutron flux. Obviously, if there was no neutron flux, rod worth would be /. The control rod worth is actually proportional to the square of the neutron flux. The question did not ask the candidate to discuss differential or integral rod worth, but rather to explain why a statement given about rod worth was incomplete.
Local neutron flux is not the only factor that comes into play in determining control rod worth, though it is very important.
If overall neutron flux in the core is high, such as would be the case at 100% power, the local flux around any given control rod will also likely be high, but the control rod worth would likely not to be very high.
This is why on the GE plants the Rod Worth Minimizer is only required to be operable when power is below 20%. The answer stands.
8.
Question 6.02 Facility Comment:
Answer No. 2 is no longer correct.
Reference:
Operation Manual Volume 2, Page 3-12.
Resolution:
Comment accepted. Answer was charged to reflect only three possible answers and individual point values were increased so that overall question value remained the same.
9.
Question 7.01 Facility Comment:
As per Administrative Control Procedure (ACP)-06.2 Page 3, the immediate actions of this procedure are not required to be committed to memory.
Resolution:
Examiner Standard 402, " Scope of Written Examinations Administered to Senior Reactor Operators - Power Reactors," Paragraph A.3, states in part that a candidate must demonstrate complete knowledge and understanding of the symptoms, automatic actions, and immediate action steps specified by offnormal or emergency operating procedures. Question 7.01 required the candidate to know the immediate action of an emergency operating procedure.
Licensee administrative procedures that contradict the Examiner Standards are not adequate justification for not implementing those standards, therefore the question stands.
- 10. Question 7.04 A.1 Facility Comment:
As per 10 CFR 20, 3.0 Rem / quarter if NRC Form 4 available.
6
Resolution:
The above additional answer is more than we required for full credit and the candidate would not lose any points for stating this in addition to the required answer.
- 11. Question 7.05 B j
Facility Comment LACBWR Operating Manual Volume 1, Page 2-11, Channel 3 and 4 period bypass keys are placed in bypass upon logic change.
Resolution:
Question 7.050 required the candidate to basically know the second condition of two possible, (i.e., when the IRC's should be bypassed as specified in LACBWR Operating Manual, Volume IV, Pages 5, 6, and 7). This was specified in the question. The startup procedure may specify that it can be done sooner, but that was not the question. The answer stands, 7
EXAMINER COMMENTS 1.
During the grading of the exam, the point value for the answer to the second half of Question 5.12.B was increased from 0.66 to 0.67 to make the total point value 1.0.
2.
During the grading of the exam it was noted that the partial credit values for the answer to Question 8.07.B exceeded the question point value by 0.25.
The 0.25 partial credit at the end of the answer was deleted as extraneous thus bring the total partial credit point value in line with the question point value.
3.
During the grading of the exam it was determined that the partial credit point values for the answer to Question 8.11.8 were inappropriately divided. The first 0.5 points were deleted and the remaining two 0.5 partial credit points were increased to 0.75.
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4 U.
S.
NUCLEAR REGULATORY COMMISSIDN REACTOR OPERATOR LICENSE EXAMINATION I
10 FACILITY:
_LAQRQgSg______ _ _______
() (I REACTOR TYPE:
_gWB-@fG_________________
DATE ADMINISTERED: _@h49 Ell 4_'_______________
EXAMINER:
_HILLSz_Dz_______________
APPLICANT:
INSIBMGIl9NS_IQ_8EE6199 nil Write answers on one side only.
Ura separate paper for the answers.
Stcple 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 least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
- /. OF CATEGORY
% OF APPLICANT'S CATEGORY
__YBLUE_ _I9196
___gCQBE___
_yO6UE__ ______________C81gGQBy_____________
_2Ez99__ _EEzQQ 1.
PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW
_2Ez99__ _2Ezgg
________ 2.
PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
_2Ez99__ _2EzQ9
________ 3.
INSTRUMENTS AND CONTROLS
_29z99__ _2Ezgg
________ 4.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIGLOGICAL CONTROL 100.00 100.00
________ TOTALS FINAL GRADE _________________%
All work done on this examination is my own. I have neither given nor received aid.
EPPL5CEUT 5~55555iUR5~~~~~~~~~~~~~~
T i
MC RULES NE) W10ELIIES FOR LICENSE EXAMINATIONS l
During the a kinistration of this examination the following rules apply:
Cheating on the. examination means an automatic denial of your application 1.
t and cou,d result in more severe penalties.
Restroom trips am to be limited and only one candidate at a time may 2.
leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
Use black ink or dark pencil gly to facilitate legible reproductions.
I 3.
4.
print your name in the blank provided on the cover sheet of the examination.
Fill in the date on the cover sheet of the examination (if necessary).
5.
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.
- as 8.
Consecutively number each answer sheet, write "End of Category 1
appropriate, start each category on a new page, write Jon1 one sTde of the paper, and writ'e "Last Page" on the "Tast answer sheet.
9.
Number each answer as to category and number, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
i
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
l
- 12. Use abbreviations only if they are commonly used in facility literature.
The point value for each question is indicated in parentheses after the 13.
question and can be used as a guide for the depth of answer required.
l
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
l
- 15. Partial credit may be given. Therefore ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
If parts of the examination are not clear as to intent, ask questions of 16.
i the examiner only.
You must sign the statement on the cover sheet that 1. dig:stes that the n
17.
work is your own and you have not received or been given assittance in completing the examination. This must be done after the examination has i
been completed.
f e
- 28. tihen you complete your examination, you shall:
a.
Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam sids - figures, tables, etc.
(3) Answer pages including figures which are a part of the answer.
f b.
Turn in your copy of the examination and all pages used vo answer the examination questions.
Turn in all scrap paper and the balance of the paper that you did c.
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.
e s
PAGE 2
.li__LBING1ELEE_DE_NWGLE9B_EQWEB_EL9NI_QEEB811QN a
IBEBdQDYNed1GEa_ME01_IB9NEEEB_9ND ELWID ELOW QUESTION 1.01 (1.50)
Indicate whether each of the f ollowing statements cancerning fission product poisons are TRUE or FALSE.
The equilibrium xenon concentration increases with increasing a.
power level while the equilibrium samarium concentration (0,5) significantly decreases with increasing power level.
b.
A reactor startup several days af ter a scram from extended high power operation is considered to be xenon and samarium (0.5) free.
c.
The value of peak xenon following a scram will depend directly upon the concentration of xenon-135 and iodine-135 present in (0.5) the reactor at the time of the scram.
DUESTION 1.02 (2.00) a.
For each of the following reactivity coefficients, indicate whether it will become more or less negative with increasing core age.
(0.5) 1.
Doppler Coefficient (0.5) 2.
Temperature Coefficient (0.5) 3.
Void Coefficient b.
Indicate which one of the above reactivity coefficients is the more important shutdown mechanism for step increases of reactivity above the prompt critical value and for large ramp increases (0.5) of reactivity (very short transients).
OUESTION 1.03 (1.00)
Sel ect the correct word in parenthesis for each of the f ollowing statements.
- a. The closer the reactor gets to critical, the (l onger, shorter) the wait must be to allow the subcritical neutron density to (0.5) reach equilibrium.
- b. During reactor heatup (more, less) outward rod movement will be required at higher temperatures for a given coolant temperature (0.5) increase.
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]
PAGE 3
I Az__EBING1ELEf_DE_NyCLgeB_EQwgB_EL9NI_QEgBBIlgN 2
IBEBdQDYN8dlGS2_Ug@l_IB@NSEgB_@ND_ELUID_ELQW I
i QUESTION 1.04 (1.00)
The reactor scrams f ollowing extended operat_ ion at high power levels.
Explain how the worth of a control rod positioned near the center of the core changes over the five hours immediately following the scram (1.0) cnd indicate what is the predominate effect causing this change.
QUESTION 1.05 (1.00)
Cc1culate a reactor cooldown rate (deg F/hr) for a vessel pressure dacrease from 800 psig to 350 psig in one hour.
Show all work for (1.0) f ull credit.
QUESTION 1.06
(.50)
Indicate which one of the f ollowing statements best describes why (0.5.
the condenser is operated at a vacuum.
a.
Less energy is extracted f rom the steam but overall plant efficiency is increased.
b.
As the vacuum is increased, the saturation temperature of the steam is decreased, allowing more energy to be extracted, c.
As the vacuum is increased, the saturation temperature of the steam is increased, allowing more energy to be extracted.
d.
The amount of energy extracted from the steam is not dependent on condenser vacuum.
)
I
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I
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eit__E.61NGIELED_DE_HWELEBB_EDNEB_EL9NIaQCEB011Q6, IBGB59DYN951GEi_BEBI_IB9NEEEB.9ND_ELWIR.ELDW O
QUESTION 1.07 (3.50) c.
Explain how f eedwater heating improves the ef ficiency of the (1.0) power plant?
b.
The reactor is at 100% power and one of the f eedwater heaters i s bypassed.
Explain how reactor power changes as a result (increase, decrease, Dr remain constant) and WHY.
(Justi f y 41.0) your answer.)
c.
Explain how bypassing this feedwatef heater affects iMe Net Positive Suction Head (NPSH) available to the f orced circulation pumps (i ncr ease, decrease, or remain constant) and WHY.
(Justif y (1.0) your answer.)
d.
Ensuring adequate NPSH is important in preventing what condition which could prove detrimental to the fqrced circulation pumps.
.t O. 5 )
DUESTION 1.05 (1.00) a.
Indicate which one of the f ollowing properties is corisidered an important aspuct of a good moderator.
(0.51 (1) low mass number (2) large absorption cross section (3) small scattering cross section b.
Explain why a moderator is needed in the reactor core.
(0.5)
ObESTION 1.09 (2.00) a.
Explein why a reactor core is designed to have excess
( 1. fM
]
reactivity.
b.
A reactor core is designed to have an effective multiplication l
factor (k-eff) of 1.091 delta k/k.
What is the value of the excess reactivity designed into this core?
(0.5) c.
If this same reactor core has a shutdown margin of 9.5 % delta k/k, what is the minimum total control rod worth required for beginning of core life conditions.
(0.5)
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PAGE
/5
.1,.__EBINQlELEL 9E_t!QQLg68_EQWE8_EL6HI_gggB8119h ISEBd92YNed1GE2_BEOI_IB9N9EEB_86D_ELWID_ELDW e
CUESTION 1.10 (2.00) of D21ayed neutrons are very important.because they allow control
]
tha reactor.
(1.0)
Explain how delayed neutrons af f ect reactor period.
a.
- b. Explain how delayed neutrons af fect the flux decay following (1.0) a f ull power scram.
OUESTION 1.11 (1.00)
Explain why it is necessary to continue cooling of the core following (1.0) a chutdown from power operation.
DUESTION 1.12 (2.00) a.
Define the term Minimum Critical Heat Flut< Ratio (MCHFR).
(0.75) b.
Est ab l i shmer.t of this limit is intended to protect against what (0.5) type of occurence?
c.
List three f actors on which the actual value of MCHFR would (0.75) depend.
OUESTION 1.13 (1.00)
Explain wiset is meant by the term " shadowing" as it pertains to (1.0) control rodc.
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J 1z__EBING1ELEE_DE_NWGLEBB_EDNEB_ELBNI_DEEBBIIDN, PAGE o
INEBN99XUBBIG52_NE91_IB9NSEEB_8ND_ELWID_ELDN 1
QUESTION 1.14 (1.00) l Differential pressure measurements can be used to determine level or flow.
For each of the following in COLUMN A, select the appropriate type of relationship that exists f rom COLUMN B.
(1.0)
COLUMN A (Item)
COLUMN B (Relationship)
- a. Level
- 1. Proportional to differential
- b. Flow pressure plus a constant.
2.
Proportional to differential pressur.e alone.
- 3. Proportional to the inverse of differential pressure.
4.
Proportional to the square of differential pressure.
- 5. Proportional to the square root of differential pressure.
OUESTION 1.15
(.50)
Adding latent heat to e liquid at seturated conditions will... (CHODSE ONE)
(0.5) a.
... increase the temperature of the water b.
... change the water to steam at the same temperature c.
... change the water to steam at a slightly higher temperature d.
... decrease its subcooling by making it boil DJESTION 1.16 (2.00) e.
The reactor is operating at 100*/. power and flow.
E :pl ain what happens to core flow and WHY with a reduction in power by control rod insertion.
Assume forced circulation pump speed remains constant.
(1.0) b.
During heatup, an increase in reactor power by control rod withdrawal will (increase, decrease, or not change) flow through the core.
Choose the correct answer and provide justification for you answer.
(1.0)
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PAGE 7
4 __EBINGIELEH_DE_HWGLE98_EDNEB_ELeNI_DEEBBIIDN, INEBNQDYN8HIGH2_UE9I_IBeNHEEB_0ND_ELUID_ELDN QUESTION 1.17 (2.00)
The reactor is operating at 100% power.
The turbine inlet valves or stop valve close as a result of a turbine trip or generator loss of load.
List two challenges that the transient presents to Safety a.
Level 1.
(1.0) b.
List four plant features available to turn back the (1.0) challenges.
1 l
(***** END OF CATEGORY 01
- )
Ez__DLBNI_Qg51@N_lNGLyp1NG_g8Egly_8NQ_gggBggNQy_gygIgd5 PAGE O
QUESTION 2.01 (2.00)
For each of the forced circulation pump interlocks below, indicate which one of the following types of protection that the interlock is intended to provide TYPES OF PROTECTION (1) pump cavitation protection (2) fluid piston bearing protection (3) excess positive reactivity protection (4) speed control actuator protection c.
When the individual pump manual speed control switch is moved f rom the pullout position, pump speed automatically increases to 40%.
(0.5) b.
Pump start is preventud unless it is at its minimum speed setting.
(0.5) c.
Increasing pump speed is prevented whenever the Rod Control Switch is in the " WITHDRAW" position.
(0.5) d.
A scram i nterlock automatically redures the pump speed to 80% if the e>:1 sting speed e>:ceeds the 80% value.
(0.5)
QUESTION 2.02 (2.00) a.
Explain why the Boron Tant temperature is maintained by immersion heaters and the supply and drain lines insulated and traced with electric heating cable.
(1.0) b.
Four control valves in the Baron Injection System will open allowing the boron solution to be pumped by the Core Spray Pumps.
What design f eature e>:ists to ensure thet at least one of these 1
two sets of control valves will open to orovide boron injection l
capability?
(1.0)
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PAGE 9
2z__EkeNI_DEE1EN_INGLWD1W9_S9EEIY_9ND_EUEBGENGY_SYEIEd5 t
QUESTION 2.03 (2.00) c.
List two signals which will automatically close the Decay Heat (1.0)
Cooling System blowdown line valve to the Main Condenser.
b.
Explain what design f eature exists to control reactor water level during startup and shutdown if for some reason water cannot be blown down to the Main Condenser through the Decay Heat (1.0)
Cooling System.
QUESTION 2.04 (1.50) a.
Assuming the Overhead Storage Tank level is decreased to 25 inches above the bottom of the tank, which systems can still be supplied (1.0) with water from this tank?
b.
Decontamination of the Overhead Storage Tank water to acceptable levels is accomplished by use of what system?
(0.5)
OUESTION 2.05 (2.50) a.
List three conditions that will automatically place the shutdown (1.Si condenser into operation.
b.
Explain why upon automatic initiation of the shutdown condenser system, the condensate valves open ten seconds after the steam inlet valves and the of f-gas valve.
(1.0)
OUESTION 2.06 (2.50)
Indicate which system is the NORMAL source of cooling water for each of the following components.
(0.5) a.
Shutdown Condenser (0,5) b.
Decay Heat Cooler (0.5)
Primary Purification System Regenerative Cooler c.
(0.5) d.
Component Cooling System Coolers (0,5) e.
Gland Steam Condenser
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2&__ELBUI DEE10N_1HGLl101HQ_SBEEIY_0ND_EDEBGENGY_EYEIEME PAGE 10 QUESTION 2.07 (1.50)
Indicate whether each of the following statements concerning the Emergency Core Spray System is TRUE or FALSE.
a.
When the control switch for the " Low Pressure Emergency Core Spray" Valve 53-25-001 is placed in "AUTD", the valve will open upon a reactor water low level and reactor low pressure condition allowing flow directly to the core spray header.
(0.5) b.
Boron injection will automatically isolate the Core Spray Pumps from the Emergency Core Spray System so that they can be used for the purpose of injecting boron into the reactor coolant.
(0.5) c.
The Emergency Core Spray System and the Baron Injection System will inject directly into the reactor through the core spray header.
(0.5)
'l DUESTION 2.08 (1.50)
List two systems which contain valves supplied by the Hydraulic a.
Valve Accumulator System.
(1.0) b.
Indicate where the water supplied by the Hydraulic Valve Accumulator System goes after performing work at the actuators of these valves.
(Be specific by listing the exact component within the system.)
(0,5)
DUESTION 2.09 (2.00)
Indicate which system or component is the PRIMARY source of water for each of the following systems, a.
Reactor Building Spray (0.5) b.
Seal Injection System (0,5) c.
Low Pressure Service Water System (0.5) d.
High Pressure Service Water System (0,5)
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PAGE 11 2z__ELONI_ DESIGN _1HGLUDibs_E9EEIY_9ND_EDEBGENGY_EYSIEUS QUESTION 2.10 (2.00) o.
Explain why Feedwater Heater No. 3 has a motor-operated shutoff valve on its normal overflow line while corresponding valves on Feedwater Heaters No. 1 and No. 2 are just manual valves.
(1.0) b.
A recombiner is used in the Gaseous Waste Disposal System to recombine hydrogen and oxygen in the waste gas.
Describe two distinct reasons why the hydrogen and oxygen are recombined.
(1.0)
QUESTION 2.11 (2.00)
For each of the f ollowing motor control centers, indicate whether or not it can receive emergency power from the diesel generators.
If amergency power can be supplied f rom the diesel generators, indicate which diesel generator (1A or 1B) is the NORMAL supply.
(Assume no cross-connect between 480-V Essential Buses.)
(0.5) a.
Turbine Building MCC 1A (0.5) b.
Reector Building MCC 1A (0.5) c.
Reactor Building MCC 1B (0.5) d.
Diesel Building MCC OUESTION 2.12 (1.50)
List three methods available for DAILY determination of primary (1.5) system leakage.
DUESTION 2.13 (2.00)
Explain what methods are available to backup the High Pressure a.
Service Water System supply to the fire suppression water systems (1.0) in case the motor-driven pump f ails.
b.
For each of the below areas, explain what type of autnmatic fire suppression system is installed.
1.
1B Emergency Diesel Generator Room (0.5)
(0.5) 2.
Electrical Equipment Room
(***** END OF CATEGORY O2 *****)
PAGE 12 4t__iWEIBWUENI5_6HD_G9BIBQLE QUESTION 3.01 (2.00)
Explain the difference between control switch positions designated o.
as " AUTO" and as "MCA AUTO" f or the Alternate Core Spray diesel s.
(Be sure to include signals that would cause specific actions in (1.0) each switch position.)
b.
Explain why upon automatic start of these diesels the Alternate Core Spray System will not necessarily inject into the reactor.
(Be sure to include additional signals or reactor conditions which must be present and the specific system actions that result.
Assume that the Alternate Core Spray System is initially (1.0) in standby allignment f or automatic operation. )
DUESTION 3.02 (2.00)
Indicate whether each of the following signals or conditions will cause a partial scram, full scram, or no scram.
(Choose one.)
(Assume bypass switches are NOT in bypass.)
e.
Undervoltage on Reactor Building MCC 1A (0.5)
(0.5) b.
Turbine Main Stop Valve not fully open c.
Condenser Vacuum below low limit (0.5) d.
Undervoltage on 2400-volt Bus 1A and 1B (0.5)
OUESTION 3.03 (2.00) s.
List the Nuclear Instrumentation channels that use compensated son chambers to monitor neutron flux.
(1.0) b.
Explain what effect that undercompensation in this type of detector will have on indicated neutron flux level an hour after a scram from full power end why this same type of effect would not be readily seen just prior to the scram.
(1.0)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l PAGE 13 At__INEIBWUENIE_6ND_GQUIBDLE QUESTION 3.04 (2.50)
Explain what AUTOMATIC design feature has been instituted to c.
provide protection against an Anticipated Transient Without (1.0)
b.
Describe the logic involved in initiating automatic ATWS protective actions.
(Be sure to list the number and type of each channel input including setpoints and the number required for actual (1.5) initiation.)
QUESTION 3.05 (3.00)
The piceammeter for each Wide Range Nuclear Instrumentation Channels 5 and 6 is equipped with three' trip units.
E>:plein each of these trips including listing ALL setpoints and resulting system actions (3.0) associated with each trip.
DUESTION 3.06 (1.50)
List three conditions associated with Nuclear Instrumentation Source Ra'nge Channels 1 and 2 which will result in a Rod Withdraw Prohibit.
(1.5)
OUESTION 3.07 (2.00)
List four process variables which provide input to the Reactor (2.0)
Water Level Control System.
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
PAGE 14 Iz_.135IEWDENIS_9ND_GDNIBDLE QUESTION 3.08 (2.50) o.
Explain how the output from the Main Steam Bypass Wide Range and Narrow Range controllers is selected to provide actual control and under what operating conditions (startup, shutdown, testing, or normal power operation) each of these controllers would normally be the one providing control to the Main Steam (1.5)
Bypass Valve.
b.
Explain why a vacuum interlock exists to prevent the Main Steam Bypass Valve from opening until greater than 5 inches of vacuum has been achieved in the main condenser.
(1.0)
OUESTION 3.09 (1.50)
Describe which systems are monitored by each of the three liquid radiation monitors.
(Be specific by describing sample location within the system( 1. D for full credit.)
OUESTION 3.10 (2.00)
The Seal Injection Di f f erenti al Pressure Controller receives a signal from a differential pressure transmitter which is connected between the f orced circulation pumps suction header and the seal injection supply header.
Describe what other automatic actions, alarms or interlocl:s are associated with signals f rom this same transmitter. (Include
( 2. 0.
setpoints.)
DUESTION 3.11 (3.00) a.
Describe two conditions (other than equipment failure) which will cause the Control Room Ventilation System outside air intake damper to automatically close.
(1.0) b.
Describe four conditions (other than equipment failure) which will cause the Reactor Building Ventilation System isolation dampers to automatically close. U (2.0)
(
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
PAGE 15
,' gi__INSIBWUENIE_6ND_GQUIBQLS l
CUESTION 3.12 (1.00)
Indicate whether each of the f ollowing statements concerning the 120-V Non-interruptible Bus 1B are TRUE or FALSE.
c.
The normal main f eed power source is supplied by the static inverter and a reserve f eed power source may be supplied by the (0.5)
Turbine Building 120-V Regulated Bus.
b.
The non-interruptible bus power supply is manually transferred f rom the main f eed power source to the reserve f eed power source (0.5) in the event of low voltage on the bus.
<4 l
(***** END OF CATEGORY 03 *****)
l i
PAGE 16 iz__ESDGEDWBEE_:_NgBd864_8DNDBOOLa_EUEBDENGY_8HD BODIDLQGIget_ggNIgg6 QUESTION 4.01 (3.00)
It has been determined that a major primary system leak has daveloped.
What immediate actions are you required to take per tha " Major Primary System Leak" procedure, LACBWR Operating (3.0)
Minual, Volume I, Section 4.3.3.2 7 QUESTION 4.02 (2.00)
FALSE.
Indicate whether each of the following statements is TRUE or c.
In the event of a f ull scram the Forced Circulation Pumps automatically trip and station load transfers to the Reserve (0.5)
Auxiliary Transformer.
b.
In reference to a primary system leak, if a water l evel drop preceeded a pressure drop, a large water leak is indicated.
(0.5) c.
If a tornado is visually observed within ten miles of the plant and the plant is in the path of the storm, the operator is
( 0. 5 ;
required to shutdown the plant.
d.
When an alarm is received on the Fire and Smoke Detection System and the "Si l ence Al arm" switch is depressed, all other audible (0.5) zone alarms are disabled.
OUESTION 4.03 (3.00) a.
Explain why upon a total loss of Low Pressure Service Weter the Primary Purification Pump would be stopped by the operator.
(1.0) b.
In reference to a rupture of the Alternete Core Spray System, explain what is the worst possible f ailure of this system.
(1.01 c.
Explain where in the system an offgas system explosion is (1.0) most likely to occur.
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
st__BBQGEDWBES_:_UQBU96a _9BNQB5964_EDEBEENGY_9dD PAGE 17 i
l B9D1969EIG96_G9NIBDL CUESTION 4.04 (2.00)
Explain why each of the f ollowing automatic actions occur following a complete loss of control air to the Containment Building.
a.
Reactor scrams
.(1.0) b.
Standby domineralizer pump starts (1.0)
M QUESTION 4.05 (1.00)
Explain why the operator should place the control switch for any of the turbine oil system pumps to the RUN position if they startup in automatic.
(1.0)
OUESTION 4.06 (2.50) a.
The procedure for starting Condensate Pumps with the system drained and hotwell normal contains a special precaution saying "to close in on the discharge valve of the pump when starting."
Describe three adverse conditions thet could result from a failure to perform this action.
( 1. 5 :-
b.
This same special precaution also says that " venting must be thorough to ensure that no air pockets remain in the system."
Explain why air pockets remaining in the system would be a concern.
( 1. 0.
QUESTION 4.07 (1.00)
Explain what adverse consequence could result f rom valving the heater drains too rapidly in the procedure for bypassing the No. 1 or No. 2 Feedwater Heaters.
(1.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
l PAGE 10 S z__EBQGEDWBEE _:_WQB5eb a _GENQ658(a _g5gB@gNQ Y_gNQ BODIDbQGlGOL_GQNIBQL QUESTION 4.08 (2.00)
Explain why during reactor startup the Source Range detectors c.
may need to be partially withdrawn as the neutron flux increases. (1.0) b.
Explain why the high voltage to these detectors must be (1.0) de-energized when the channels saturate.
QUESTION 4.09 (1.00)
The reactor is at power with Forced Circulation Loop 1A already in service and Forced Circulation Loop 1B being returned to service.
In the process of returning the idle loop to service, the operator secures the purification pump and closes the primary purification (1.0) stop valve.
Expl ai n why this action was performed.
OUESTION 4.10 (1.00)
E::pl ain how a portable survey monitor 's meter range switch should be set when entering an area of unknown radiation level.
(1.0)
DUESTION 4.11 (3.00)
E>. plain the methods which the immediate actions would require yc_
to use for pressure control and for maintaining reactor water level f ollowing each of the bel ow occurences, a.
Full Reactor Scram per the " Scram Procedure", LACBWR Operating Manual, Volume I,
Section 4.1 (1.0) b.
Complete Loss of Electrical Power per the " Complete Loss of Electrical Power" procedure, LACBWR Operating Manual, Volume I (1.0)
Section 4.2 c.
Complete Loss of Electrical Power with a subsequent f ailure of the emergency diesel generators to start per the " Complete Loss of Electrical Power" procedure, LACBWR Operating Manual, Volume I,
Section 4.2 (1.0)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
PAGE 19 Az__EB9GEDW6EE_:_N9BdeL2_8DNQBd@(2_Edg8QgNQy_9ND BODI96991G86_QQNISQL QUESTION 4.12 (2.00)
Explain the differences, if any, in the immediate actions required by the " Equipment Flooding" procedure, LACBWR Operating Manual, Volume I, Section 4.9 upon both inlet lines rupturing or just one (2.0) inlet line rupturing in the Circulating Water System.
DUESTION 4.13 (1.50)
Dascribe the immediate action steps you are required to take per the " Equipment Flooding" procedure, LACBWR Operating Manual, Volume I, (1.5)
Section 4.9 upon a component cooling water system rupture.
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
LQUA110A SHEE1 Cycla effici o(Network I
f = es v = s/t W /( k m in cs = ag s = Vot + 5 att E = sc2 KE = 5 av2 a = (Vf - Vo)/t A = AN A = Age-At PE = egh Vf = V + at w = e/t x = an2/tg = 0.693/t5 W =.6 tgeff = [(t ) (t )3 t
b
[(tg) + (t )3 b
AE = 931 am g,g,e-zx k=aCpat 6=uAst I = Ic *
- e Pwr = Wr#
I = Io 10"* O TVL = 1.3h P = P 10sur(t)
HVL = -0.693h o
P = P et/T o
SUR = 26.06/T SCR = S/(1 - Keff)
CRx = S/(1 - Keffx)
CR (1 - Keff1) = CR (1 - keff2) 1 2
SUR = 25pla* + ( s-p )T T = (m*/p) + [(8 - p)/Ip3 M = 1/(1 - Keff) = CR /CRo 1
T = 1/(p - s)
M = (1 - Keffo)/(1 - Keffi)
T = (s - p)/(Ap)
SDM = (1 - Keff)/Keff A* = 10-5 seconds p = (Keff-1)/Keff = AKeff/Keff I = 0.1 seconds-1 p = [(a /(T Keff)] + [s ff/(1 + IT))
e
'I d11=1d22 2
P = (IeV)/(3 x 1010) 11d1 2=1d22 2
R/hr = (0.5 CE)/d (meters) z =.N R/hr = 6 CE/d2 (feet)
Miscellaneous Conversions Water Parameters 1 curie = 3.7 x 1010dps 1 gal. = 8.345 lbm.
1 gal. = 3.78 liters 1 kg = 2.21 lbm 3 8tu/hr 1 ft3 = 7.48 gal. -
I hp = 2.54 x 10 I sw = 3.41 x 106 8tu/hr Density = 62.4 lbm/ft Density = 1 gn/cm3 1 in = 2.54 cm Heat of vaporization = 970 Btu /1bm
- F = 9/5'C + 32 Heat of fusion = 144 Btu /1bm
'C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg.
1 BTU = 778 ft-1bf 1 ft H O = 0.433 1bf/in2 2
=
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Iz__EBING1ELES_DE_NWGLE96_EDWEB_EL9NI_DEEBOIlDN, PAGE 20 IBEBUQDYN951GS4_BEGI_IBONSEEB_96D_ELUID_ELDW ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 1.01 (1.50)
(After reaching equilibrium the samarium concentration c.
False (0.5) remains constant.)
b.
False (0.5) - (It may be xenon free, but samarium will increase following the scram.)
c.
True (0.5)
REFERENCE LACBWR Training Manual, p.
65-75.
ANSWER 1.02 (2.00) a.
1.
More negative. (0.5) 2.
Less negative. (0.5) 3.
Less negative. (0.5) 6.
Doppler Coefficient (0.5)
REFERENCE LACBWR Training Manual, p.
A-53-58, C-12.
ANSWER 1.03 (1.00) a.
longer (0.5) b.
more (0.5)
REFERENCE LACBWR Training Manual, p.
A-62,81,88,98,105.
4 ANSWER 1.04 (1.00)
Rod worth will decrease. (0.5)
Caused by buildup of xenon in the center of the core which reduces neutron flux in that region.
Control rod worth is dependent upon neutron flux. (0.5)
REFERENCE LACBWR Training Manual, p.
A-59,68.
1 I
PAGE 21 Iz__EBING1ELEE_DE_NWGLE9B_EDNEB_EL9NI_QEgB@IlQN 2 IHEBUDDYN90lGE2_ME91_IB9NEEEB_9ND_ELUID_ELDW ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 1.05 (1.00)
Convert pressures in psig to psia. (0.25) 364.7 psia 350 psig + 14.7 psi
=
814.7 psia 800 psig + 14.7 psi
=
Obtain corresponding temperatures from the Steam Tables. (0.5) 364.7 psia = 435.5 deg F 814.7 psia = 520.3 deg F Datermine the temperature change for the hour. (0.25)
(520.3 deg F - 435.5 deg F)/1 hour = 84.8 deg F/hr REFERENCE LACBWR Training Manual, p.
J-56 and Steam Tables.
ANSWER 1.06
(.50) b.
As vacuum is increased, the saturation temperature of the steam is decreased, allowing more energy to be extracted.
(0.5)
REFERENCE LACBWR Training Manual, p.
J-46-59 and Steam Tables.
LACBWR Operating Manual, Vol.
3, p.
4-1.
ANSWER 1.07 (3.50) a.
The energy recovered in feed heating would otherwise be lost to the main condenser or less heat is required from the reactor to reach the desired condition.
(1.0) b.
Reactor power will increase (0.5) due to the increase in inlet subcooling.
(0.5) c.
NPSH available would increase (0.5) due to the increase in inlet subcooling.
(0.5) d.
Cavitation (0.5)
REFERENCE LACBWR Training Manual, p.
B-62, C-20, J-46-52.
PAGE 22 iz__EBING1ELEg_gE_NygLges_ggwgB_ELeuI_gEEB911Q02 ISEBUDDYN901952_MEBI_IB9NSEEB_9ND_ELVID_EL9W ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 1.08 (1.00) c.
(1) low mass number (0.5) b.
To slow down neutrons into the thermal energy range where there is higher probability of causing a fission.
(0.5)
REFERENCE LACBWR Training Manual, p.
A-14,20.
(2.bb)
ANSWER 1.09 a.
To compensate f or factors which cause losses of reactivity over core life.
(1.0) b.
E:: cess reactivity = 0.091 delta k/L or 9.1 % delta L/L (0.5) c.
Minimum control rod worth = 0.186 delta k/k or 18.6 % delta L/k (0.5)
REFERENCE LACBWR Tr eining Manual, p.
A-50,51,99.
ANSWER 1.10 (2.00) a.
increase reactor period (1.0) b.
Following the prompt drop in flux, the negative reactor period would be limited consistant with the longest lived decay constant l
causing a slower drop until leveling out in the source range.
(1.0) i i
ANSWER 1.11 (1.00) s r.d* e
- d
- c sy.
'T1 r e a s c. h ea ' 9 e m e n 4 e 4 by f's s. ' s o
N h=at gener ated by # i ni-cn product decw i; suff-icient to damage th=
"""a if racting in ist continued, (1.0) i j
REFERENCE LACBWR Training Manual, p.
A-100.
PAGE 23 iz__BBINgitLgg_gE_gygLges_EgggB_ekgNI_QEgB@llgda IHEB59DYM001GEz_UE01_ISBNSEEB_gND_ELUID_ELDW ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 1.12 (2.00)
Ratio of the burnout heat flux for the fluid conditions at a given c.
power level to the maximum heat flux in a fuel rod at any location in the core.
(0.75) b.
Fuel Cladding Failure (0.5) c.
Steam Quality Mass Flow Rate Pressure Flow Area Geometry Axial Power Distribution Power Level (3 required - O.25 pts. each)
REFERENCE LACBWR Training Manual, p.
A-104,105, B-29-33.
ANSWER 1.13 (1.00)
The ability of control rod movement to change the worth of other control rods due to changes in the flux shape in the reactor.
(1.0)
REFERENCE LACBWR Training Manual, p.
A-62,105.
ANSWER 1.14 (1.00) a.
2 (0.5) b.
5 (0.5)
REFERENCE LACBWR Training Manual, p.
J-3-12.
ANSWER 1.15
(.50) b.
... change the water to steam at the same temperature (0.5) g-e-
y e
s--
PAGE 24 12__EBING1ELEE_DE_NWGLEBB_EQWEB_ELONI_QEg68IlgN, IMEBdQDYN@digg,_dg@I_IB@NSEgB,@NQ_E(Qlg_E(QW ANSWERS -- LACRDSSE
-86/05/14-HILLS, D.
REFERENCE j
l LACBWR Training Manual, p.
J-22.
ANSWER 1.16-(2.00) a.
Core flow would increase (0.5) due to a decrease in two phase flow resistance.
(0.5).
b.
Core flow would increase (0.5) due to an increase in natural circulation.
(0.5).
REFERENCE LACBWR Training Manual, p.
J-41,53.
ANSWER 1.17 (2.00) a.
high reactor power (0.5) high reactor pressure (0.5) b.
turbine stop valve closure partial scram main steam bypass valve (MSBV) high power level scram high pressure scram (SD cond. operation) forced circulation pump high pressure trip relief valves (4 required - O.25 pts. each)
REFERENCE LACBWR Training Manual, p.
C-22.
1 Cz__ELBNI_DESIEM_INGLUDINE_E9EEIY_9ND_ENEBEENGY_EYSIEUS PAGE 23 ANSWERS -- LACRDSSE
-86/05/14-HILLS, D.
ANSWER 2.01 (2.00) c.
(2) fluid piston bearing protection (0.5) b.
(3) excess positive reactivity protection (0.5) c.
(3) excess positive reactivity protection (0.5) d.
(1) pump cavitation protection (0.5)
REFERENCE LACBWR Training Manual, p.
B-62,63.
LACBWR Operating Manual, Vol.
2, p.
3-11,12.
ANSWER 2.02 (2.00)
To prevent precipitaion of the sodium pentaborate solution.
(1.0) a.
6.
Two independent gas sources, nitrogen and plant air, assure independent operation of these valves.
(1.0)
REFERENCE LACBWR Training Manual,
p.
E-5-9.
LACBWR Operating Manual, Vol.
2, p.
9-1,2.
- ANSWEF, 2.03 (2.00) c.
low reactor water level (0.5) high containment building pressure (0.5) b.
A blowdown line to the Overhead Storage Tank is provided in the Primary Purification System.
(1.0)
REFERENCE LACBWR Training Manual,
p.
B-76.
LACBWR Operating Manual, Vol.
2, p.
6-2, 7-3.
i
1 l
Ez__EL8HI_DEElGN_INGLWQ1Ng_geEgIY_eyp_gMgsggNGY_syElgMS PAGE 26 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 2.04 (1.50) a.
Emergency Core Spray System Fuel Element Storage Well System
. 0c al :njectisa Cynt;- 'bechup caz. g;ncy ;;' cup!
(2 required - 0.5 pts. each) b.
Fuel Element Storage Well System (0.5)
REFERENCE LACBWR Training Manual, p.
E-15,17.
LACBWR Operating Manual, Vol.
2, p.
15-1,2,7.
ANSWER 2.05 (2.50) a.
Reactor Building Steam Isolation Valve Not Full Open Turbine Building Steam Isol ation Valve Not Full Open Reactor Pressure Channel 1 above setpoint Reactor Pressure Channel 2 above setpoint (3 required - 0.5 pts. each) b.
To provide sufficient time for the tube side of the shutdown condenser to be pressurized and purged of noncondensible gasses.
(1.0)
REFERENCE LACBWR Training Manual, p.
E-21.
LACBWR Operating Manual, Vol.
2, p.
5-5.
ANSWER 2.06 (2.50) a.
Demineralized Water System (0.5) b.
Component Cooling Water System (0.5) c.
Cold Purification System Flow (0.5) d.
Low Pressure Service Water System (0.5) e.
Condensate System (0.5)
REFERENCE LACBWR Training Manual, p.
B-79.
LACBWR Operating Manual, Vol.
2, p.
5-2,13-1,3.
LACBWR Operating Manual, Vol.
3, p.
7-1.
PAGE 27 Sz__ELONI_REEIGN_INGLUDING_E9EEIY_0ND_EMEBDENRY_EYSIEMS ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 2.07 (1.50) c.
True (0.5) b.
True (0.5)
(The Baron Injection System will discharge to the c.
False (0,5) forced circulation discharge header.)
i REFERENCE LACBWR Operating Manual, Vol. 2, p.
B-1,9-1.
ANSWER 2.08 (1.50) a.
Forced Circul ation System (0.5)
Main Steam System (0.5) b.
Water Return Sump Tank in the Hydraulic Valve Accumlator System (0.5)
REFERENCE LACBWR Operating Manual, Vol.
2, p.
14-1,3.
ANSWER 2.09 (2.00) a.
Overhead Storage Tank (0.5) b.
Seal Injection System Reservoir (Will also accept Condensate Demineralizer System)
(0.5) c.
River Water (0.5) d.
Low Pressure Service Water System (0,5)
REFERENCE LACBWR Operating Manual, Vol.
2, p.
10-1,15-1.
LACBWR Operating Manual, Vol.
5, p.
6-4,7-1.
i
PAGE 28 Et__ELONI_QESl@M_lNQLQQld@_$@EglX_6NQ_Ed[6@[NQX_EX@IEDS ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 2.10 (2.00) a.
In the event that the No. 3 heater normal regulator mal f unctioned in such a way as to cause it to open wide, the motor-operated velve could be quickly closed f rom the Cont. ol Room.
This is necessary because overheating of the condensate leaving No. 2 heater could result in excessive temperature to the f eedpumps which would cause severe pump damage.
(1.0) b.
To reduce the storage load of the waste gas system.
(0.5)
To reduce the possibility of a combustible gas mixture in the compressed gas system.
(0.5)
REFERENCE LACBWR Operating Manual, Vol.
3, p.
9-2,10-4.
ANSWER 2.11 (2.00) a.
Power supplied by diesel generators (0.25) - Diesel Generator 1A (0.25) b.
Power supplied by diesel generators (0.25) - Diesel Generator 1B (0.25) c.
Power not supplied by diesel generators (0.5) d.
Power supplied by diesel generators (0.25) - Diesel Generator 1B (0.25' REFERENCE LACBWR Operating Manual, Vol.
4, p.
15-2,3,7.
ANSWER 2.12 (1.50)
Activity Retention Tank Level Changes Health and Safety High Volume Air Sample Humidity (3 required - 0.5 pts. each)
REFERENCE LACBWR Operating Manual, Vol. 11, p.
5-1.
I I
i
' Es__ELBUI_ DESIGN _INGLWDING_E9EEIY_eND_ENEBEENGY_EYSIEMS PAGE 29 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
j 1
ANSWER 2.13 (2.00) c.
two diesel-driven auxiliary service water pumps (0.5) emergency service water supply system (0.5) b.
1.
carbon dioxide flooding system (0.5) 2.
halon suppression system (0.5)
REFERENCE LACBWR Operating Manual, Vol.
5, 7-1,8-1.
i
st__lUSIBWUENIS_eup_ggyIBg65 PAGE 30 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 3.01 (2.00) a.
"MCA AUTO" allows the diesel to start on high containment building pressure while " AUTO" allows the diesel to start on either high containment building pressure or low High Pressure Service Water pressure.
(1.0) b.
A low reactor water level signal must be present to open the automatic isolation valves.
(0.5)
Flow to the vessel will not commence until reactor vessel pressure drops to approximately 150 psig.
(0.5)
REFERENCE LACBWR Operating Manual, Vol.
2, p.
17-1.
ANSWER 3.02 (2.00) a.
Full Scram (0.5) b.
Partial Scram (0.5) c.
Full Screm (0.5) d.
F.-tiel Ccram (0.5)
VaelStern REFERENCE LACBWR Operatin; Manuel, Vol.
4, p.
6-3,4,5,6.
ANSWER 3.03 (2.00) a.
Intermediate Range Channels N3 and N4 (0.5)
Wide Range Channels N5 and N6 (0.5) b.
Indicated neutron f l u:- will be higher than it should en hour after the scram.
(0.5)
Just prior to the scram at that power level the gamma signal is overshadowed by the neutron signal and thus is not as predominate.
(0.5)
REFERENCE LACBWR Training Manual, p.
D-8,14,15.
r
'Iz__JNgIByDgNig_gNQ_QQN16965 PAGE 31 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
i 4
ANSWER 3.04 (2.50) o.
Trip of both recirculation pumps.
(1.0) b.
Reactor pressure greater than 1350 psi (0.25) or reactor water level less than -30 inches (0.25).
Water level signals are provided from each of three water level safety channels.
(0.25)
Reactor pressure signals are provided by safety channels (2 channels) and the mainsteam bypass narrow range (1 signal) transmitters.
(0.25)
Trip relays perform 2-out-of-3 logic for either reactor water level or reactor pressure.
(0.5)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
5-7(b).
ANSWER 3.05 (3.00)
Downscale Trip Al arm (0.5)
Less than 5% on 150% scale (0.25)
Less than 2% on 6O% scal e (0.25)
High Level Alare (0. 5 )
Greater than 110*/. on 150% scale (0.25)
Greater than 44% on 60% scale (0.25)
High Level Scram (0.5)
Greater than 115% on 150% scale (0.25)
Greater than 46% on 60% scale (0.25)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
4-13.
ANSWER 3.06 (1.50)
Peroid less then 7 seconds on either Channel 1 or 2 Mode switch not in Operate on either Chennel 1 or 2 Teet plug inserted in Per. Input.iect on either Channel 1 or 2 Key swithes bypassing either Source Range Channel Outputs before wi thdraw permi t has been obtained.
(3 required - 0.5 pts. each)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
4-6.
I
PAGE 32 Jz__INSIBWdENIE_eND_QQNIBQLS ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 3.07 (2.00)
R2 actor Water Level (0.5)
R2 actor Steam To Turbine Pressure (0.5)
Steam To Turbine Flow (0.5)
Fcedwater to Reactor Flow (0.5)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
B-1.
LACBWR Training Manual, p.
D-53.
ANSWER 3.08 (2.50) a.
Main Steam Bypass Valve's high signal selector tranmits the higher of the two signals (the one calling f or the valve to be further open.)
(0,5)
Wide Range - Startup, Shutdown, Testing (0.5)
Narrow Range - Norme.1 Power Operations (0.5) b.
To prevent overpressuri:ation of the condenser.
(1.0)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
9-1.
LACBWR Training Manual, p.
D-48.
LACBWR Operating Manual, Vol.
2, p.
4-6.
ANSWER 3.09 (1.50)
Component Cooling Water Moni tor sample f rom component cooling water stipply header.
(0.5)
I c y.. 'd w a s 4 d s s c "'vs ' P'* '* * ' '
" ^ -
die-b----
Liquid Waste Monitor sample from deunet-eme nf the liauid waete dicchsrge inte th; ;g7,d;nce-con 1_ inn wc 6 c i i l mi r.t 1inc.
... 'r b discun t in4-o Cerv e s akte t.'n t,
Turbine Condenser Cooling Water Monitor sample f rom cirulation water discharge line after service water has joined this line.
(0.5)
REFERENCE LACBWR Training Manual, p.
G-17,18.
LACBWR Operating Manual, Vol. 10, p.
5-4,6a.
It__lhDIEWDENIE_gNQ_QQN16QLE PAGE 33 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 3.10 (2.00)
Dif f erenti al pressure decreases below setpoint (set at less than or cqual to 140 psi) standby seal inject pump starts.
(0.5)
. Low seal inject pressure alarm.
(0.5)
. Differential pressure must be greater than 80 psi bef ore f orced circulation pumps can be started.
(0.5)
Once operating, if the differential pressure decreases below OO psi for cpproximately 10 seconds the f orced circulation pumps automatically trip.
(0.5)
REFERENCE LACBWR Operating Manual, Vol.
2, p.
10-13.
ANSWER 3.11 (3.00) a.
Electrical Equipment Room Halon Suppression System Alarm Ammonia Detector Alarm hcl Detector Al arm (2 required - 0.5 pts. each) b.
High Activity (0.5)
High Reactor Pressure (0.5)
High Reactor Building Pressure (0.5)
Low Reactor Water Level (0.5)
REFERENCE LACBWR Operating Manual, Vol.
5, p.
10-11,17a.
ANSWER 3.12 (1.00) a.
True (0.5) b.
True (0.5) j l
REFERENCE LACBWR Operating Manual, Vol.
4, p.
19-2.
e 3,__PBQGggyBgs_:_UQBueLa_99WDB56L2_EUEBGENGX_6NQ PAGE 30 68D196991G96_GQUIB96 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 4.01 (3.00)
Shutdown the reactor if it didn't scram Ensure Core Spray is in operation if water level reaches -12 inches Trip the Forced Circulation Pumps Evacuate the Containment Building Ensure Reactor Building Ventilation Dampers and 4-inch Vent Header Internal Valve have closed Close Reactor Vent Header External Isolation Valve Turn control switch for the Containment Building Ventilation Dampers to close Varif y that Containment Heating Steam Condensate Val ve, Retention Tank Discharge Valve, Containment Vessel HPSW 1 solation Valve, Containment Vessel Demin Water Isolation Valve, its bypass, and Decay Heat Blowdown Valve are closed Attempt to maintain feedwater supply to keep core covered (6 required - 0.5 pts. each)
REFERENCE LACEWR Operating Manual, Vol.
1, p.
4-8,9.
ANSWER 4.02 (2.00) c.
False (0.5) - (Forced Circulation Pumps do not automatically trip due to a full scram.
Forced Circulation Pumps will reduce to 80% speed if above that speed at time of scram.)
6.
True (0.5) c.
False (0.5) - (Plant is not required to be shutdown in this case.
Plant electrical load is reduced however.)
d.
True (0.5)
REFERENCE LACBWR Operating Manual, Vol.
1, p.
4-2,7,16,22.
PAGE 35
$t_mPS9GEDWBES_:_NDBdBLa_6BNQBBBL2_EUEBGENGY_9dD BOD 196991GBL_G9 NIB 96 ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 4.03 (3.00) c.
To reduce heat load on the Component Cooling Water System.
(1.0) b.
A f ailure in the Electrical Penetration Room which causes loss of Essential Swithgear Bus 1A (1. 0) wa t a(s a a ce.cp+ **p4ure. ddwas ke am c.
Offgas Recombiner (1.0) ofve.a h4egeme %ga jerg.4_;g A L o C A.
REFERENCE LACBWR Operating Manual, Vol.
1, p. 4-32,35,39.
ANSWER 4.04 (2.00) a.
Hi gh f l u:< caused by shutdown condenser operation or power to flow due to Forced Circulation Pumps tripping. (Will accept either answer for full credit.)
(1.0) b.
Shutdown condenser and Overhead Storage Tank makeup valves opening.
(Will accept either one f or full credit.)
(1.0)
REFERENCE LACBWR Operating Manual, Vol.
1, p.
4-38.
ANSWER 4.05 (1.00)
To provide the pump breater with full electrical protection.
(1.0)
REFERENCE LACBWR Operating Manual, Vol.
3, p. 5-11.
ANSWER 4.06 (2.50) a.
Rupturing of the pump discharge check valve or other piping components from severe hydraulic lock. (0.5)
Tripping of the pump on overcurrent. (0.5)
Pumping the hotwell dry, with a resulting loss of pump suction and possible damage to pump. (0.5) b.
Air remaining in the system could, through subsequent valving operations, enter the suction of a running feed pump and severely damage the pump.
(1.0)
PAGE 36 I
Oz_LCBQGEDQBEE_:_UQBd861_0BWQB50La_EDEBEEUGY_0ND l
,s bed 196DGIG06_GQNIBQL ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
REFERENCE LACBWR Operating Manual, Vol. 3, p.
10-14.
ANSWER 4.07 (1.00)
Could result in a hotwell level surge severe enough to open the hotwell overflow valves wide and trip the feed pumps on low suction pressure.
(1.0)
REFERENCE LACBWR Operating Manual, Vol. 3, p.
10-15.
ANSWER 4.08 (2.00)
To maintain the reading within range of the instrument.
(1.0) a.
b.
Damage to the detector could result upon saturation.
(1.0)
REFERENCE LACBWR Operating Manual, Vol.
4, p.
4-28.
ANSWER 4.09 (1.00)
To aid in heatup of the 1B Loop such that it is approximately the same temperature as the operating loop.
(1.0)
REFERENCE LACBWR Operating Manual, Vol.
2, p. 3-21.
ANSWER 4.10
.(1.00) a Should be place on a range which is higher than the~ expected dose rate level.
Enter slowly and switch to a scale which indicates dose rate before proceeding.
(1.0)
REFERENCE LACBWR Training Manual, p.
G-24.
(~
PAGE 37 92_LCBQGEDWBEE_:_NQBdeLa_8aNQB56La_EUEBGENGY_8NQ BOD 1969EIGOL_GQNIBQL ANSWERS -- LACROSSE
-86/05/14-HILLS, D.
ANSWER 4.11 (3.00)
Reactor Feedwater System maintains reactor water level (0.5) a.
Main Steam Bypass Valve controls pressure (0.5) level (0.5)
High Pressure Core Spray Pumps maintain reactor water b.
Shutdown condenser may need to be used for pressure control since main condenser vacuum car.not be maintained. (0.5) c.
The air operated valves for the shutdown condenser "OPEN" on low control air pressure.
Manually control reactor pressure by throttling the steam isolation valves and High Pressure Service Water cooling water supply.
(0.5)
If reactor water level cannot be maintained, pressure must be reduced, as the Low Pressure Core Spray Valve and Alternate Core Spray System are the only means of water addition.
(0.5)
REFERENCE LACEWR Operating Manual, Vol.
1, p.
4-2,4,5.
ANSWER 4.12 (2.00)
Both rupture - Shutdown both main circulators (0.5) and scram the reactor (0.5)
Only one ruptures - Close inlet to the associated water box (0.5) and reduce power to 607. (0.5)
REFERENCE LACBWR Operating Manual, Vol.
1, p.
4-33.
ANSWER 4.13 (1.50)
Attempt to isolate rupture and prevent spraying on adjacent equipment (0.5)
If unable to isolate break, shutdown non-vital equipment cooled j
by component cooling water (0.5) j Shutdown reactor and control temperature or cool down without use of decay heat cooler if component cooling water system must be isolated (0.5)
REFERENCE LACBWR Operating Manual, Vol.
1, p.
4-36.
, e-U.
- 8. NUCLEAR RESULATORY COMMISSIDN SENIDR REACTDR DPERATDR LICENSE EXAMINATION
-_A_C_ROS_S_E L
FACILITY:
g y
gy REACTOR TYPE:
_kWR-A/C_________________
f
,. (f)ghl L2 U
DATE ADMINISTERED: _@ 6 / 9 Ell 4__ _ _ _ __ _ _ _ __ _ _ _ _
_ LANK @ByRY _R.___________
EXAMINER:
[
yf 2
E APPLICANT:
INSIBQCIlgNS_IQ_8PELIC8 nit use eeparate paper for the answers.
Write answers on one side only.
Staple question sheet on top of the answer sheets.
Points for each question are indicated in parentheses after the question. The passing grade requires at least 70% in each category and a final grade of at least 80%.
Examination papers will be picked up six (6) hours after the examination starts.
% OF
@ATEGORY
% OF APPLICANT'S CATEGGRY
__Y669E_ _191@(
___SQOBE___
_y@(gE__ ______________C@l[Gggy_____________
_2Ez99__ _2Ez99
________ 5.
THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS
_2Ez99__ _2Ez99
________ 6.
PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION
_2Ez99__ _2Ez99
________ 7.
PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL
_2Ez99__ _2Ez99
________ B.
ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 199z99-_ 199199
________ TOTALS FINAL GRADE _________________%
'All work done on this examination is my own. I have neither
.givcn nor received aid.
~~~~~~~~~~~~~~
APPLIC55Y 5~555U55URE
t PAGE 2
L.__IHEQBY_DE_HUGLE98_EQUEB_EL8HI_QEEB8Il0Na_ELUIDEa_9ND IMEBDQQYN951GH QUESTION 5.01 (1.00)
Differential pressure measurements can be used to determine level or flow.
For each of the following in COLUMN A, select the appropriate (1.0) typa of relationship that exists f rom COLUMN B.
COLUMN A (Item)
COLUMN B (Relationship)
- a. Level 1.
Proportional to differential b.
Flow pressure plus a constant.
2.
Proporti onal to differential pressure alone.
3.
Proportional to the inverse of differential pressure.
4.
Proportional to the square of differential pressure.
5.
Proportional to the square root of differential pressure.
QUESTION 5.02 (1.50)
Indicate whether each of the following statements concerning fission product poisons are TRUE or FALSE.
A.
The equilibrium xenon concentration increases with increasing power level while the equilibrium samarium concentration significantly decreases with increasing power level.
(0.5)
B.
A reactor startup several days after a scram f rom extended high power operation is considered to be xenon and samarium (0.5) free.
C.
The value of peak xenon following a scram will depend directly upon the concentration of xenon-135 and iodine-135 present in the reactor at the time of the scram.
(0.5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE
- )
r PAGE 3
L.__IME98Y_QE_NWGLE96_EQWEB_E68HI_9EEBell&_EkWi&_9ND IMEBqQDyM9 DIGS QUESTION 5.03 (2.00)
A.
Define the term Minimum Critical Heat Flux Ratio (MCHFR).
(0.75)
Establishment of this limit is intended to protect against what D.
(0.5) type of occurence?
C.
List three f actors on which the actual value of MCHFR would (0.75) depend.
QUESTION 5.04 (2.00)
The reactor is operating at 100% power and flow.
Explain what A.
happens to core flow and WHY with a reduction in power by control i
rod insertion.
Assume forced circulation pump speed remains (1.0) constant.
B.
During heatup, an increase in reactor power by control rod withdrawal will (increase, decrease, or not change) flow through the core.
Choose the correct answer and provide (1.0) j usti f i cati on for you answer.
QUESTION 5.05 (2.00)
Th9 reactor is operating at 100% power.
The turbine inlet valven or stop valve close as a result of a turbine trip or generator loss of load.
A.
List two (2) challenges that the transient presents to Saf ety (1.0)
Level 1.
B.
List four (4) plant f eatures available to turn back the (1.0) challenges.
QUESTION 5.06 (2.00)
Given a constant fuel temperature, explain how and why alpha-D will (2.0) change with an increasing void f raction.
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE C
Dz__INEDBY_9E_NWGLE86_EDWEB_ELBNI_DEEBBI19N4_ELUIDE,_9BD IMEBUDDYN9BIRE QUESTION 5.07 (3.00)
A.
Af ter making a rod withdrawal with the reactor critical, you notice a 100 second period.
How much reactivity was added by the rod withdrawal assuming Beta =.007 and Lamda =.08 sec-17 Show all work.
(1.0)
D.
After a reactor scram from power the shortest stable period possible is
-80 seconds?
Explain this statement.
(1.0)
C.
Is the INITIAL period IMMEDIATELY following the scram shorter than -80 (1.0) seconds?
Explain your answer.
QUESTION 5.08 (3.50)
A.
A common misconception regarding rod worth is that, "if the neutron flux increases in the vicinity of a rod, the worth of that rod also increases."
Explain why this statement is incomplete in explaining (1.5) rod worth.
B.
List four (4) physical factors which help determine actual individual rod worth.
(2.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
N D
__ItE98Y_9E_tRJGkteB_t9t!EB_EL9t!I_9tEB8I19tf4_ELW1994_M It!EBt!99Yt!9t!IGH QUEDTION 5.09 (3.00)
Reactor 251 Power w
o
'I 3' l
l
=
e l
Al A2 43 Time (Hours)
A.
What is the approximate time from Al to A37 (1.0) 1.
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 2.
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> 3.
50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> 4.
70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> B.
What is the approximate time from Al to A27 (1.0) 1.
1-4 hours 2.
4-6 hours 3.
6-9 hours 4.
9-12 hours (1.0)
C.
Why does Xe concentration decrease f rom A2 to A3?
1.
Xe decay is equal to iodine decay 2.
Xe burnout is equal to iodine decaying to Xe 3.
Xe burnout is greater than iodine decaying to Xe 4.
Xe decay is greater that iodine decay i
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
i L.__INEQBY_QE_NWGLE88_EQNEB_EL8HI_QEEBBIl0Na_ELUIDEa_8NQ PAGE a
INEBUQQXN851GE QUESTION 5.10 (1.00)
Which of the f ollowing statements BEST describes what happens to a fluid cc it passes through a venturi?
(1.0) c.
Pressure remains constant, but the velocity increases as the diameter of the venturi decreases.
b.
Pressure increases and velocity decreases as the diameter of the venturi decreases.
Pressure decreases and velocity ren.ains constant as the diameter of c.
the venturi increases.
d.
Pressure increases, but the velocity decreases as the diameter of the venturi increases.
QUESTION 5.11 (2.00)
During a reactor startup, criticality is achieved when a positive pariod is maintained without further positive reactivity additions.
Th2 definition of critical states Keff equals 1.0 and reactivity Gquals 0.0 and period would theref ore be infinite.
WHY then is the reactor declared critical when the period is positive?
QUESTION 5.12 (2.00)
Answer the f ollowing regarding transient ef f ects on core boiling hact transfer when operating at power:
A.
Briefly EXPLAIN the immediate (instantaneous) effect of a sudden core flow INCREASE on the amount of NUCLEATE boiling at the clad surface.
(1.0)
B.
A sudden flow (INCREASE, DECREASE) in the core could cause film boiling.
Briefly, JUSTIFY your choice.
(1.0) i l
l i
I
(***** END OF CATEGORY 05 *****)
l
PAGE 7
h__EL9NI_gygIgeg_gggigNg_GQNIgg,_ege_1ggIgyngNIeIlgN QUESTIDN 6.01 (1.50)
Located above each of the FCP control switches on Bench E are green, white, cnd red indicating lights.
For eacn of the combinations given below state wh3ther or not the companion function is TRUE or FALSE.
If false, provide (1.5) tha correct combination or function.
LIGHT FUNCTION 1.
Green Light On Breaker auto tripped.
2.
White Light On Breaker manually tripped & closing power available.
3.
Red Light On Breaker closed & tripping power available.
QUESTION 6.02 (2.00)
List the 4 FCP interlocks that prevent the addition of excess positive (2.0) roactivity.
QUESTION 6.03 (1.50)
Answer the f ollowing questions TRUE or FALSE.
Explain any false answers:
(1.5) 1.
If one of the two solenoid valves f or a rotovalve in the Forced Circulation System should fail, the controls will still function in the normal manner.
2.
Failure of both solenoid valves will cause a rotovalve to fail closed.
QUESTION 6.04 (4.50)
List 9 of the 12 conditions that will trip the FCP's.
Provide setpoints (4.5) ao applicable.
l
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE O
h__E19NI_EYSIE5E_DEH19Na_GQUIBDLa_9ND_INRIBWDENIBIIDN QUESTION 6.05 (3.00)
A.
For the f ollowing valves in the Eme, gency Core Cooling System, describe their f ailed position on a loss of nitrogen / air or a loss of electrical control power to there respective soleniods.
If the final position is other than full open/ closed, indicate the final position.
(1.5) 1.
Low Pressure Emergency Core Spray valve (53-25-001).
2.
Core Spray Pumps Domin. Water Inlet valve (53-25-002).
3.
Core Spray Pumps Demin. Water Outlet valve (53-25-003).
4.
Core Spray Pumps Demin. Water Inlet Shutoff valve (53-25-008).
5.
Serv. Water to Emergency Core Spray Pumps valve (53-25-004).
B.
List the type (i e.
120-volt ac, noninterruptible 120-volt ac, 125-volt de, etc.) power supply for the soleniods of the above valves.
(1.5)
QUESTION 6.06 (5.00)
A.
What ef f ect (s) does the downstale trip unit on Nuclear Instrument Power Range Channels 7 & B have on the safety system?
Specify in your answer the number and the condition of the trip units required(2.0) to cause the effect(s).
B.
State the setpoint of the downstale trip unit on Nuclear Instrument (1.0)
Power Range Channels 7 & 8.
C.
List two (2) All Rod Scram trips associated with the Nuclear Instrumentation Intermediate Range Channels 3 or 4.
(2.0)
QUESTION 6.07 (2.00)
A.
List two signals which will automatically close the Decay Heat Cooling System blowdown line valve to the Main Condenser, (1.0)
B.
Explain what design f eature exists to control reactor water level during startup and shutdown if f or some reason water cannot be blown down to the Main Condenser through the Decay Heat (1.0)
Cooling System.
QUESTION 6.08 (1.50)
Lict three methods available f or DAILY determination of primary (1.5) cystem leakage.
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE 9
h __ELOUI_EYSIEME_DEE196 _GDNIB b _0ND_INEIBW5ENIGIlgN QUESTION 6.09 (2.00)
A.
Explain the difference between control switch positions designated as " AUTO" and as "MCA AUTO" for the Alternate Core Spray diesels.
(Be sure to include signals that would cause specific actions in (1.0) each switch position.)
B.
Explain why upon automatic start of these diesels the Alternate Core Spray System will not necessarily inject into the reactor.
(Be sure to include additional signals or reactor conditions which must be present and the specific system actions that result.
Assume that the Alternate Core Spray System is initially in standby allignment for automatic operation.)
(1.0)
QUESTION 6.10 (2.00)
A.
Explain what methods are available to backup the High Pressure Service Water System supply to the fire suppression water systems (1.0) in case the motor-driven pump fails.
B.
For each of the below areas, explain what type of automatic fire suppression system is installed.
1B Emergency Diesel Generator Room (0.5)
(0.5)
- Electrical Equipment Room 1
(*****
END OF CATEGORY 06 *****)
PAGE 10 Z,.__EBDGEDWBE5_:_NDBUBL4_9BNDB56La_EDEB9ENGY_BND 89DIA 991996_G9 NIB 96 QUESTION 7.01 (3.00) i What six Tha Crib House pumps have unexpectedly lost their water supply.
(6) immediate actions are you required to take per the " Loss of Communications Between the Mississippi River and Crib House Pumps" procedure, LACBWR Operating Manual, Volume I, Section 4.147 (3.0)
QUESTION 7.02 (3.00)
The reactor is operating at 90% power with the reactor feedwater flow controller in AUTO when reactor water level unexpectedly starts increasing.
What four (4) immediate actions are required to be taken by LACBWR Operating Manual, Volume I, Alarm C1-1 (Reactor Water Level (Hi))?
NOTE:
(3.0)
Some immediate actions may have more than one action step.
QUESTION 7.03 (2.50)
A.
What three (3) condition (s) must be met before high pressure service water can be injected into the reactor vessel?
(1.5)
B.
Is there any concern with injecting high pressure service water into the reactor vessel?
If so, what?
(1.0)
QUESTION 7.04 (2.50)
A.
List the total occupational dose limits per quarter imposed by (1.0) 10CFR2O for the followings 1.
Whole body 2.
Ex tremiti es 3.
Skin B.
A new employee who is 10 years old is assigned to work f or you and is to be a radiation worker.
What is the individuals allowable accumulated occupational dose (lif eti me) to the whole body and allowable whole body quarterly limit per 10CFR207 (1.0)
C.
Briefly explain the dif f erence between a " rad" and a " rem".
(0.5)
(*****
CATEGORY 07 CONTINUED ON NEXT PAGE *****)
1 PAGE 11 Z.__EBQGEDUBEE_:_tf0Bdeba_eRNQBd@(3_g5EBQgNGL@NQ 809196991G96_GONIBQL QUESTIDN 7.05 (3.50)
With regard to the Nuclear Instrumentation Systems A.
LACBWR Operating Manual, Vol. IV, Para.
4.5.2,
" Normal Operat i on ",
states that it is important that when testing power range channels during reactor operation that its corresponding power-flow channel be bypassed.
1.
What potential consequence (s) will result if this is not done?
(2.0)
Explain your answer.
2.
What power range channel corresponds to what power-flow channel?
(0.5)
B.
LACBWR Operating Manual, Vol. IV, Para.
4.5.2,
" Normal Operation",
states:
When approaching the top end of the scale on the Intermediate Range Channels, or ___________________________________,
the Intermediate Range Channels should be bypassed.
Supply the missing condition (the exact words f rom the procedure are not (1.0) necessary for full credit).
QUESTION 7.06 (1.50)
LACBWR Operating Manual, Vol. VI, "Retueling", lists three (3) principle hczcrds which may be encountered during fuel and control rod handling (1.5) opsrations.
List these.
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
- )
Z.__EBQGEDUBEE_:_HQBdeLi_etNQBMeLa_EMEBg[NGy_eUD PAeE 12 BOD 1960GIGeb_GQNIBQL QUESTION 7.07 (2.00)
Indicate whether each of the f ollowing statements is TRUE or FALSE.
A.
In the event of a full scram the Forced Circulation Pumps automatically trip and station load transf ers to the Reserve (0.5)
Auxiliary Transformer.
B.
In reference to a primary system leak, if a water level drop preceeded a pressure drop, a large water leak is indicated.
(0.5)
C.
If a tornado is visually observed within ten miles of the plant and the plant is in the path of the storm, the operator is (0.5) required to shutdown the plant.
D.
When an alarm is received on the Fire and Smoke Detection System and the " Silence Alarm" switch is depressed, all other audible zone alarms are disabled.
(0.5)
QUESTION 7.08 (2.00)
Explain the differences, if any, in the immediate actions required by the " Equipment Flooding" procedure, LACBWR Operating Manual, Volume I, Section 4.9 upon both inlet lines rupturing or just one inlet line rupturing in the Circulating Water System.
(2.0)
QUESTION 7.09 (1.50)
With regard to the Forced Circulation System procedure f rom LACBWR Opsrating Manual, Vol. II, Sec. 3.6.2, Transfer From Two-Loop to Single-Loop Operation:
A.
The procedure cautions not to allow the teenperature in the idle loop to decrease below 200 degrees F.
Why is this a concern?
(1.0)
B.
The procedure also cautions you to stay below 50% power (82.5 MWT) when in single loop operation.
Why is this caution in the (0.5) procedure?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE
- )
PAGE 13 Z.__EBQGEQWBgs_:_t!gBt!eba_etNQBt!eba_Et!EBGEt!GY_eNQ BeP1960 GIG 06_GQNIBQL QUESTION 7.10 (1.50) 10 CFR 20 and 10 CFR 50 designates 15 types of events that must be report-Gd to the NRC at once (within one hour). List five seperate events that rGquire NRC notification within one hour. Note that listing more than one ovant that comes under the same heading or type will count as one.
QUESTION 7.11 (2.00)
A.
LACBWR Operating Manual, Vol. III, Sec. 12.4, Operating Procedure, describes a limitation on the startup of the reactor feed pumps f ollowing a f eed pump trip.
What is the limitation?
(1.0)
B.
Explain the purpose of the limitation and any conditions for which it (1.0) is waived.
l
(***** END OF CATEGORY 07
- )
~
PAGE 14 Es.__8Dt!1HIEIB011YE EBQGEDWBEEA-QQNQlllQNga_8NQ_ Lit l11gIlgNg QUESTION B.01 (2.50)
A.
What is the minimum shift crew composition required by your Technical Specifications for conditions 1, 2,
and 3.
Address the following positions in your answer - SS, SRD, RD, AD, STA.
(1.0)
D.
Under what condition (s) and for how long may the required shift crew compostion be less than the above.
Be specific as to the number less than required, length of time, immediate actions and any cases for which (these) provi si on (s) is (are) not applicable.
(1.5)
QUESTION 8.02 (1.50)
A.
True or False:
The Onsite Emergency Response Director (ERD) is the Cooperative management psrcon responsible f or overall direction of the response to an emergency ccndition after activation of the EDF.
(1.0)
B.
Who INITIALLY fills the role of the ERD?
(0.5)
QUESTION 8.03 (1.50)
List 3 of the 4 general types of procedures that are required by ACP-06.2, Procedure Adherence and Temporary Changes, to be present and followed (1.5) etcp-by-step.
QUESTION B.04 (1.00)
TGchnical Specification 2.3.3.2 requires that the rotoport valves on the diccharge of the Forced Circulation System pumps be capable of closing in 25 sec. and require a minimum of 4 min. to open.
What is the basis f or occh of the above times?
(1.0)
QUESTION B.05 (1.50)
Liot 3 actions which shall be taken in the event a safety limit is (1.5) violated.
(*****
CATEGORY OB CONTINUED ON NEXT PAGE *****)
i 9___eD51NIEIBBI1YE_EB9GEDJBEDi_G9ND1I19BEa_9ND_L151IeIIDNS PAGE 15 QUESTION 8.06 (1.50)
According to Tech Specs, containment integrity shall be maintained in cKnditions 1, 2, 3, and during:
Give 3 other conditions.
(1.5)
QUESTION B.07 (3.00)
A.
Per ACP-06.2, Procedure Adherence and Temporary Changes, under what circumstance (s) are LACBWR personnel authorized to deviate from an approved procedure.
Include any actions that must be taken.
(2.0)
B.
What authorization is required for Temporary Changes (to normal operating procedures) that do not change the intent of the approved procedure?
If there are exceptions, include these in your answer. (1.0)
QUESTION 8.08 (2.00)
What are your duties as the Duty Shift Supervisor per ACP-15.1, Installation and Removal of Jumpers, Lifting and Replacement of Leads 7(2.0)
QUESTION B.09 (2.50)
With regard to the requirements of ACP-17.1, Incident Reports Including Rsd Phone Notification to the NRC:
A.
You attempt to make a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification to the NRC as required by 10CFR50.72 and find that the red phone is inoperable.
What are the two (2) options available to you?
List in the order to be f ollowed.
(1.0)
B.
Define the term " incident".
(1.5)
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
i St__eDUINIEIB9I1YE_EBDGEDUBEE4_GDHD1IIDUE4_9ND_L15119I1QNE PAGE IS QUESTION 8.10 (3.00)
With respect to the requirements of ACP-15.2, Equipment Lock And Tag Centrol A.
In the event that it is impossible to comply with the procedure, who has the authority to allow the work to proceed?
(0.5)
B.
Who has the responsibility f or the placing, removing, and recording of tags and checks?
(0.5)
C.
What three (3) basic pieces of inf ormation are required to be entered in the Control Room Log each time a " Hold" is placed?
(1.0)
D.
Where are the completed copies of released " Hold" cards filed and f or how long must they be kept?
(1.0)
QUESTION 8.11 (3.00)
A.
When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable limiting condition for operation, provided:
What are the 2 provisions?
(1.5)
B.
If the system does not satisf y these provisions, what action (s) must be taken (with the plant)?
(1.5)
QUESTION 8.12 (2.00)
With regard to the requirements of ACP-02.12, Shift Technical Advisor:
A.
The STA is required to be 1
and within _____2_____ of the Control Room, unless releived by _____3_____,
when the plant is in Operating Condition (s) _____4_____.
(1.0)
B.
What are the two (2) requirements imposed on the Duty Shift Supervisor.
(1.0)
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
4 e
e PAGE 17 L _IHEQBYR_NWGLEBB_EQWEB_EkeNI_QEEBel10Na_ELUIREi_eNQ IHEBUQQYN001Cg ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 5.01 (1.00) c.
2 (0.5) b.
5 (0.5)
REFERENCE LACBWR Traini ng Manual, p. J-3-12.
ANSWER 5.02 (1.50)
A.
False (0.5) - (After reaching equilibrium the samarium concentration remains constant.)
B.
False (0.5) - (It may be xenon free, but samarium will increase following the scram.)
C.
True (0.5)
REFERENCE LACBWR Training Manual, p.
65-75.
ANSWER 5.03 (2.00)
A.
Ratio of the burnout heat flux f or the fluid conditions at a given power level to the maximum heat flux in a fuel rod at any location in the core.
(0.75)
B.
Fuel Cladding Failure (0.5)
C.
Steam Quality Mass Flow Rate Pressure Flow Area Geometry Axial Power Distribution Power Level (3 required - 0.25 pts. each)
REFERENCE LACBWR Training Manual, p.
A-104,105, B-29-33.
Dz__INE98Y_9E_NWGLE98_E9 WEB _E19BI_9EEB9I19N2_ELUIDEa_eND PAGE 10 INEBU99XN951gs ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
J ANSWER 5.04 (2.00)
A.
Core flow would increase (0.5) due to a decrease in two phase flow resistance.
(0.5).
B.
Core flow would increase (0.5) due to an increase in natural circulation.
(0.5).
REFERENCE LACBWR Training Manual, p.
J-41,53.
ANSWER 5.05 (2.00)
A.
high reactor power (0.5) high reactor pressure (0.5)
B.
turbine stop valve closure partial scram main steam bypass valve (MSBV) high power level scram high pressure scram (SD cond. operation) forced circulation pump high pressure trip relief valves (4 required - 0.25 pts. each)
REFERENCE LACBWR Training Manual, p.
C-22.
l ANSWER 5.06 (2.00)
Alpha-D will become more negative as void f raction increases.
As voids increase, the slowing down length and slowing down time are very long.
Since the neutrons spend a longer period of time in the resonance energy cpsctrum, more neutrons will be resonantly captured and alpha-D will be core negative.
REFERENCE LACBWR Training Manual, Sec.
A.9.3, pg. A-56.
-v
hz__INEQBY_DE_NWGLE88_E9 WEB _!196I_DEEBBIIDL_ELUIDL_eND PAGE 19 ISEB509YNed1G9 ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 5.07 (3.00)
A.
B-p B
T = -------
so p = -------
AP hT+1 p =.007/ (100) (0.08) +1 = 7.78 x 10 E-4 delta k/k l
B.
After the initial prompt drop, power cannot decrease faster than the f
longest lived delayed neutron appears.
(1.0)
C.
Yes.
The initial drop in power will only be due to the prompt neutrons.
(1.0)
REFERENCE LACBWR Training Man., Sec.
A.4.7, pg. A-9; Sec. A.14, pg. A-100.
ANSWER 5.08 (3.50) l A.
The worth depends on the neutron flux in its location compared to the average neutron flux in the core.
Thus, as power is increased, the flux also increases.
This will increase the worth of the rod if the local flux is increased more than the core average flux.
B.
1.
Voi ds l
t 2.
Fuel loading
)
l 3.
Control rod pattern 4.
Moderator temperature 5.
Exposure (fuel / control rods) 6.
Peaking f actors l
(4 9 0.5 each)
I l
REFERENCE LACBWR Training Man., Sec. 9.5, pgs. A 64 & Standard Nuc. Theory.
l 1
PAGE 20 Le__It!EDBY_9E_UUGLE98_EDNEB_EL9dI_DEEB9IIDN4_ELUIDS2_9ND IMEBU9DYN9dIGH ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 5.09 (3.00)
A.
3 D.
2 C.
'3 REFERENCE LACBWR Training Man., Sec. A.10.1, pgs A-69 & 70.
ANSWER 5.10 (1.00) d.
REFERENCE LACBWR Training Man., Sec.
J.A.1, pg. J-6 -11.
ANSWER 5.11 (2.00)
Criticality is achieved low in the poweb range (i. e.
source or intcrmediate range).
It is difficult and time consuming to 1.0) and subtritical distinguish between exactly critical (Keff
=
multiplication and/or the effects of source neutrons [1.03.
If a positive period is maintained without further positive reactivity cdditions, it is assured that criticality has been acnieved.
The reactor is in f act supercriticalE1.03.
REFERENCE LACBWR Training Man., Sec. A.12; Standard Nuclear Theory.
ANSWER 5.12 (2.00)
A.
The sudden flow increase causes the clad surf ace temperature to decrease EO.333 due to more efficient convection heat transfer [0.333, decreasing the amount of nucleate boiling (1.0)
[0.333.
7 B.
DecreaseEO.333.
Due to increased clad surface temperature [0.663. (1.0)
,_1
PAGE 21 3:1__INE98Y_DE_NWGLE98_E9 WEB _EL9NI_QEEBBIlQNa_ELWIDE2_6ND INEB599YN951gg ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
REFERENCE LACBWR Training Manual, Sec. J3 Standard Thermodynamics Theory.
e
,,n
PAGE 22 64.__EL9NI_BYSIEN5_DEEIGN2_G96IBQL,_9ND_INDIBWNENIBI19N
. ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 6.01 (1.50) 1.
False [0.33 - Breaker manually tripped & closing power available[0.33.
2.
Falseto.33 - Green & white light on = Breaker auto trippedCO.33.
3.
True[0.33.
REFERENCE LACBWR Op. Man., Vol. II, Sec.
3.5, pg. 3-10.
ANSWER 6.02 (2.00) l 1.
Pump cannot be started unless it is at its minimum speed.
2.
ruwe cennui we =Le Lau uniums tne speea control system in in menwel.
3.
Control rods cannot be withdrawn while the pump speed is increasing.
4.
Pump speed cannot be increased when the rod control switch is in the J
withdraw position 3. p 04 0 0.N each)
REFERENCE LACBWR Training Manual, pg. B-63.
1
\\
ANSWER 6.03 (1.50) 1.
TrueCO.53.
2.
FalseEO.53.
The valve will fail in the open position [0.53.
REFERENCE LACBWR Op. Man., Vol. II, Sec.
3.5, pg. 3-14.
f I
-.,....m..
' ht__EL 9BI_EYSIEDS_DEE1984_G9 NIB 964_9NQ_INSIBWWENIBIl9N PAGE 23
, ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 6.04 (4.50) 1.
Turning the pump control switch to "stop".
2.
High seal leakoff temperature (>150 deg.
F.).
3.
Low seal leakoff flow
(<0. 2 gpm).
4.
Low seal water reactor pressure diff. (<BO psi. ).
5.
Low lube oil pressure (<5 psi.).
- 6. - Suction valve f or pump fully closed.
7.
Low 2400-volt bus voltage.
B.
Motor overloaded.
9.
Short ciruit.
- 10. Electrical ground.
- 11. Reactor water level </= -30",
2 of 3 logic, on saf ety channels.
- 12. Reactor pressure >/= 1350 psig, 2 of 3 logic, on saf ety char nels and pressure transmitter off alternate core spray line.
(9 G O.5 each)
REFERENCE LACBWR Op. Man., Vol. II, Sec.
3.5, pg. 3-11.
ANSWER 6.05 (3.00)
A.
1.
Open 2.
Open 3.
Open, in an amount corresponding to the position of the handwheel.
4.
Open 5.
Closed (5 0 0.3 each)
B.
1.
noninterruptible 120-volt ac.
2.
125-volt dc.
3.
125-vol t dc.
4.
125-volt dc.
5.
125-volt dc.
(5 9 0.3 each)
REFERENCE LACBWR Op. Man., Vol. II, Sc.
8, pgs. B-3&4.
l
kz__kL9NI_gySIgdg_QEg1DN _GQNIBQL3_9ND_INSIBWHENIBIlQN PAGE 24 2
, ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 6.06 (5.00)
A.
When one of the two 15% trip units (one in each power range channel)
[1.03 is in the untripped condition, the safety system changes from a one-out-of-two to a two-out-of-four requirement [1.03 (two out of the four channels 5,6, 7 and 8 must send a scram signal to the safety system to cause a reactor scram).
D.
Trip unit will be in the tripped condition when the flux amplifier is indicating less than 15%.
C.
(2 of 3 for full credit) 1.
Period less than 3 seconds on either channel 3 or 4 (and these j
outputs not bypassed by key switches).
2.
Mode switch not in Operate on either Channel 3 or 4 (and these outputs are not bypassed by key switches).
3.
Test plug inserted in Input Test jack on either channel 3 or 4 and these outputs not bypassed by key switches.
REFERENCE LACBWR Op. Man., Vol. IV, Sec.
4, pgs. 4-11,15,16.
ANSWER 6.07 (2.00)
A.
Iow reactor water l evel (0.5) high containment building pressure (0.5)
B.
A blowdown line to the Overhead Storage Tank is provided in the Primary Purification System.
(1.0)
REFERENCE LACBWR Training Manual, p.
B-76.
LACBWR Operating Manual, Vol.
2, p.
6-2, 7-3.
ANSWER 6.08 (1.50) 1.
Activity 2.
Retention Tank Level Changes 3.
Health and Safety High Volume Air Sample 4.
Humidity (3 e 0.5 each)
REFERENCE LACBWR Operating Manual, Vol. 11, p.
5-1.
h__PL9NI_gIgIg53_gggigh _GQNIBL _9Np_INSIBybgNIgIlgN PAGE 25 ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 6.09 (2.00)
A.
"MCA AUTO" allows the diesel to start on high containment building pressure while " AUTO" allows the diesel to start on either high containment building pressure.or low High Pressure Service Water pressure.
(1.0)
B.
A low reactor water level signal must be present to open the automatic isolation valves.
(0.5)
Flow to the vessel will not commence until reactor vessel pressure drops to approximately 150 psig.
(0.5)
REFERENCE LACBWR Operating Manual, Vol.
2, p.
17-1.
ANSWER 6.10 (2.00)
A.
two diesel-driven auxiliary service water pumps (0.5) emergency service water supply system (0.5)
B.
1B Emergency Diesel Generator Room - carbon dioxide flooding (0.5) system Electrical Equipment Room - halon suppression system (0.5)
REFERENCE LACBWR Operating Manual, Vol.
5, 7-1,8-1.
l
PAGE 21 Za__tB9GEDUBEE_ _H985862_8BUDB5062_E5EB9ENGY_0ND B9D196991G96_G9 NIB 96
-86/05/14-LANKSBURY, R.
ANSWERS -- LACROSSE v
a ANSWER 7.01 (3.00) 1.
Ensure the reactor has scrammed.
2.
Follow " Scram Procedure".
3.
Ensure the MSIV's have closed.
4.
Place keys 4 & 9, " Reactor Bldg. Steam Isolation Valve Not Full Open Scram Bypass Nos. 1 & 2", in bypass & use the shutdown condenser in manual control to commence cooldown of the primary system at < or =
60 deg. F./hr..
5.
Shutdown non-vital equipment that uses CCW.
6.
Send an operator to the Crib House to insect the situation and determine the cause.
(6 0.5 each)
REFERENCE LACBWR D,m.
Man., Vol.
I, Sec. 4.14, pg.4-44.
ANSWER 7.02 (3.00) 1.
Observe reactor water level recorder and the reactor water level indicator to verify high water level alarm [0.53.
2.
Ensure all rods scram takes place if l evel reaches +19 inches [0.53.
3.
Ensure the reactor vessel level controller setpoint is in the "O"
position.
If flow is not decreasing, take manual control with the coupling controller & decrease the f eedwater flow to the reactor using the controller manual knob [0.53.
4.
In case of loss of control to the feedwater pump drive scoop tube actuators:
a.
Attempt to reduce f eedwater flow by cycling the Recirculation Valve for the operating feedwater pump [0.53.
b.
Try to regain feedwater flow control by throttling with the feed-water to reactor flow control valve [0.53.
c.
Prepare to scram the reactor if control of feedwater flow cannot be obtained [0.53.
REFERENCE LACBWR Op. Man., Vol.
I, Alarm C1-1, pg. 3-64.
l
PAGE 27 Ze__EB9GE9WBEE_:_N985863_0DN9Bd6La _gDEBggNGy_969 Be91E991G96_G9NIBA ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 7.03 (2.50)
A.
1.
Shift. Supervisor permission.
2.
Additional cooling is required.
3.
There is no other way to cool the core.
(3 0 0.5)
B.
Use of service water (river water) may cause blockage of the spray lines.
REFERENCE LACBWR Op. Man., Vol.
1, Sec.
4.3, pg. 4-12 & Vol. II, Sec.
8.2, pg.
B-2.
ANSWER 7.04 (2.50)
A.
1.
1.25 rem 2.
18.75 rem 3.
7.5 rem (3 9 0.33 each)
B.
Accumulated whole body limit = 5(N-18) = 0 rem.
(0.5)
(0.5)
Quarterly whole body limit = 1.25 rem.
The rad is a measure of the energy released in tissue by ionizing C.
radiation whereas the rem is a measure of biological effect.
REFERENCE 10CFR20.101.
O I
PAGE 28 Z.__EBQGEDUBEE_:_UQB5eLi_8BNQBueLa_EMEBGENGY_eup bed 1DLQEIGOL_GQNIBQL ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 7.05 (3.50)
A.
1.
If the scram bypass key is not used f or the appropriate power-flow channel scram bypass then when the power range channel is tested at its trip point this power level signal will also be sent to its correspond-ing power-flow channel where it will be used to compare to recircula-tion flow.
Depending on plant conditions at this time the power-flow channel may f alsely see power too high f or recirculation loop flow and initiate a full reactor scram.
2.
Power range channel 7 -> power-flow channel 1.
Power range channel 8 -> power-flow channel 2.
B.
When boiling noise introduces spurious period signals (exact words no required for full credit).
REFERENCE LACBWR Op. Man., Vol IV, Sec.
4, pg. 4-28; Sec.
5, pgs. 5-6t<7.
ANSWER 7.06 (1.50) 1.
Handling of irradiated fuel or control rods, or other irradiated components, with insufficient water cover.
2.
Damaging an irradiated fuel assembly, thereby causing a possible rupture of the fuel tubes resulting in contamination of the pool and other areas of the system and airborne contamination.
3.
Permitting components, such as the core spray tube bundle, to dry during transfer outside of the water, thus inviting the possibility of airborne activity within the Containment Building.
(3 9 0.5 each)
REFERENCE LACBWR Op. Man., Vol. VI, Sec.
1, pg.
1-6.
PAGE 29 Il__EB9GEDWBEE_:_UDBU L _099985 L _EUEB9ENGY_0ND B99196991C96_G9NIBD6 ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 7.07 (2.00)
(Forced Circulation Pumps do not automatically trip A.
False (0.5)
Forced Circulation Pumps will reduce due to a full scram.
to 80% speed if above that speed at time of scram.)
D.
True (0.5)
(Plant is not required to be shutdown in this C.
False (0.5) case.
Plant electrical load is reduced however.)
D.
True (0.5)
REFERENCE LACBWR Operating Manual, Vol.
1, p.
4-2,7,16,22.
r ANSWER 7.08 (2.00)
Beth rupture - Shutdown both main circulators (0.5) and scram the reactor (0.5)
(0.5) and Only one ruptures - Close inlet to the associated water box reduce power to 60% (0.5)
REFERENCE LACBWR Operating ManLal, Vol.
1, p. 4-33.
ANSWER 7.09 (1.50)
To ensure that the temperature in the loop stays above the loop NDT A.
(130 degrees F.).
B.
To ensure compliance with the Technical Specifications (4.2.2.9).
REFERENCE LACBWR Op. Man., Vol. II, Sec. 3, pg. 3-19.
r-
PAGE 30 Zi__EBDGEDWBEE_:_UDBdeba_9BNDB5eLa_E5EB9ENGY_9ND BBDIDLD91 gel _GDNIBDL ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 7.10 (1.50) 1.
Events defined by 10 CFR 20 involving:
- c. radiation exposure to personnel.
- b. radioactive rel eases.
c.
loss of facility operations d.
damage to property
- 2. Events defined by 10 CFR 50 involving:
- c. declaration of emergency classes
- b. plant shutdown required by technical specifications
- c. deviations from technical specifications in an emergency as necessary to protect the public health and safety.
any serious degradation of 'the nuclear plant including it's principal d.
safety barriers.
- o. unanalyzed conditions that significantly compromise plant safety.
f.
a condition that is outside the design basis of the plant.
- g. conditions not covered by the plant 's operating and emergency proce-dures.
- h. any natural phenomenon or other external condition that poses a threat to plant saf ety or significantyly hampers site personnel in the performance of duties necessary for safe plant operation.
- a. valid signal.
J. any event that results in a loss of emergancy assessment capability, of f site response capability, or communications capability.
- k. any event that poses an actual threat to the plant safety or significantly hampers site personnel in the performance of duties necessary f or the safe operation of the plant including fire, toxic gas releases or radioactive releases.
(5 9.3 each)
REFERENCE 10CFR20.403 & 10CFR50.72.
ANSWER 7.11 (2.00)
A.
On a reactor f eed pump trip, no more than two successive attempts will be made to restart the pump motor.
B.
If more than two attempts are made, severe damage to the motor may resultCO.53.
This caution is waived in an emergency [0.53.
REFERENCE LACBWR Op. Man., Vol. III, Sec. 12.4, pg. 12-6.
PAGE 31 91__6D51NISIB9IlYE_EBDGEDUBEE4_GDNplIlgNH4_9ND_LlulIGIl9NE
- ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER 8.01 (2.50)
A.
SS - 1, SRO - none, RD - 1, AD - 2, STA - 1.
(5 e.20 each)
B.
The AD or STA positien can be 1 less than required in order to accommodate unexpected absence of on-duty shift crew members [0.53 provided immediate action is taken to restore the minimum crew composi-tion within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> [0.53.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent [0.53.
(1.5)
REFERENCE LACBWR T.S. Tabl e 6. 2. 2-1.
ANSWER 8.02 (1.50)
A.
False (ECD).
B.
Duty Shift Supervisor.
1.
ANSWER 8.03 (1.50)
Plant Startup Startup From Partial Scram Normal Shutdown All Technical Specification Tests (3 9.5 each)
REFERENCE ACP - 06.2, pg. 3.
ANSWER 8.04 (1.00)
Th2 basis f or the required times is to prevent a cold water transient (4 f
min.) [0.53 but to allow the loop to be isolated quickly in an emergency (25 sec.) [ 0. 5 3.
i m
y.,
.-m_
PAGE 32 Ra__8DUIN1EIBBI1YE EBQGEDWBEEa_GQUQlIl0NEa_8NQ_L151I8Il0NE ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
REFERENCE LACBWR Op. Man., Vol. II, Sec. 3, pg. 3-29.
ANSWER B.05 (1.50)
(3 of 4 below for full credit) a.
The unit shall be placed in ct least Hot Shutdown within two hours.
- b. The NRC operations center shall be notified as soon as possible and in all cases within one hour.
The Assistant General M: nager Power Group and the SRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
- c. A safety limit violation report shall be prepared.
- d. The safety limit violation report shall be submitted to the Commission, the SRC and the Assistant General Manager--Power Group within 14 days of the violation.
REFERENCE Tach. Spec.
'6. 7.1., pg. 6-31.
ANSWER 0.06 (1.50) 1.
2.
Handling of irradiated f uel.
3.
There is fuel in the reactor and any control rod is withdrawn.
(3 0 0.5 each)
REFERENCE Tach. Spec.
4.2.1.1, pg. 28.
4 gi._$pd1NigIseIlyg_esggg1Negga_gggg111gsg,_ egg _kinIIeI1gsg PAGE 33 ANSWERS -- LACROSSE
-86/05/14-LANKSBURY, R.
ANSWER B.07 (3.00)
A.
1.
In order to prevent personnel injury, including the public, or damage to the facility [0.53, provided such changes are documented by recording circumstances in the Shif t Supervisors log [0.53.
2.
In cases where a component is not expected to perform as designed
[0.253 or a system is operating non-routinely [0.253, authority is granted to deviate from routine oper,ating procedures to enable the plant to be restored to normal operation, provided such one-time, non-routine operations are documented in the Shift Supervisors log and on Form 1 of ACP-06.2 (Procedure Tempcrary Change Log) [0.53.
B.
They must be approved by 2 members of the plant management staff, one of whom must hold a valid SRO license [0.53.
Security Control Procedures [0.253 and Ht
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