ML20206N039
| ML20206N039 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/18/1986 |
| From: | Burdick T, Dave Hills, Mark King, Sherman J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20206N030 | List: |
| References | |
| 50-263-OL-86-01, 50-263-OL-86-1, NUDOCS 8608260067 | |
| Download: ML20206N039 (101) | |
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- t U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-263/0L-86-01 Docket No. 50-263 License No DPR-22 Licensee: Northern States Power Company Monticello Nuclear Generating Plant Monticello, MN 55113 Facility Nan'e: Monticello Nuclear Generating Plant Examination Administered At: Monticello Nuclear Generating Plant, Monticello, Minnesota Examination Conducted: July 22-25, 1986 Examiners: /X1 _ ___,
8/8f7 Date D. Hills blo bu f.St(erman J 6 ITate folk , 8 Sate' M. Ki g \-
i Approved By: h / ,
E8 T. M. Eurdick, CtKef Date '
Operator Licensing Section
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Examination Suninary 1986 ( 50-263/0L-86-01)
Examination administered Written, oral,~and simulatoFon'e~xaiiinstion July 22-25,,siie~r,e,Repprt
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Noadniiiii'sYeTed 't'o TiVeTehior
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Reactor Operator (SR0) candidates and five Reactor Operatcr (RO) candidates.
In addition, one simulator examination was administered to an R0 candidate.
Rasults : All candidates passed these examinations.
8608260067 860819 PDR ADOCK 05000263 V PDR
- REPORT DETAILS
- 1. Examiners D. Hills, NRC Region III - Chief Examiner M. King, INEL J. Sherman, INEL
- 2. Examination Review Meeting Specific facility comments concerning written examination questions, followed by the NRC response, are enumerated in Attachments 1 and 2.
- 3. Exit Meeting At the conclusion of the examinations, an exit meeting was held. The following personnel attended this meeting:
Facility Representatives E. Earney, Training Superintendent B. McGillic, Operations Training Supervisor
_NRC R_epresentatives D. Hills, Chief Operator Licensing Examiner, NRC Region _III M. King, Operator Licensing Examiner, INEL J. Sherman, Operator Licensing Examiner, INEL The following observations and generic issues were discussed:
- a. Applications submitted by the licensee were found to be incorrectly completed and to contain insufficient information needed to determine the candidate's eligibility. This necessitated that the NRC request and wait for submittal of additional information prior to continuation of application processing.. The examiners stressed the importance of providing accurate and complete applications.
- b. The examiners noted that the relative ease of gaining access to the protected area contributed to the efficient conduct of the oral examination process. However, the examiners also noted that the unexpected policy of allowing only one oral examination to be conducted in the main control room at any one time tended to inhibit the examination process. The e).aminers requested that in the future, if this restriction is to be imposed, that they be informed prior to the examinations so that scheduling of the oral examinations can be planned accordingly.
- c. The examiners noted that during the simulator examinations multiple restart attempts on tripped pumps without first trying to determine the cause of the trip were prevalent.
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- d. The examiners noted that although procedure use during the simulator examinations was adequate, it could be improved. In particular, the examiners were concerned with a common tendency to not recheck completion of immediate actions after the procedures were taken out. In addition, the examiners stressed that procedure format and organization were poor which tended to inhibit their most efficient use.
- e. The examiners noted a general weakness among personnel who are not part of the fire brigade concernirg knowledge of available fire fighting methods.
- f. The examiners noted that the main control room copy of the technical specifications was in poor physical condition and thus needing replacement.
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l ATTACHMENT 1 RESOLUTIONS TO MONTICELLO'S COMMENTS ON NRC R0 LICENCE EXAMINATION OF 7/22/86 Facility comment on question / answer 1.01:
b, The Period equation may also be used to explain the power decrease.
(Reference M 8102 L-007 Rev 2 p. 39)
Examiner resolution:
Agree, with the condition that the candidates' answer includes an EXPLANATION as required by the question. The equation, by itself, does not answer the question.
Facility comment on question / answer 1.10:
Figure 1 utilizcd'for this question is not plant specific and differs with the Monticello power-flow operating map. (Attached). f Confusion may arise in two areas:
- 1. Examinees are not familiar with the provided power flow map.
- 2. Operation in the crosshatched area is prohibited at Monticello, but is allowed on the provided map.
(Reference C.2 - 35)
Examiner resolution:
- 1. Disagree, in question 1.10a,1.10b, & 1.10c only a general knowledge of a power to flow map is required. These questions may be answered from the provided figure.
- 2. Agree, question 1.10d, the use of crosshatched areas does differ between the supplied figure and the referenced figure. The answer key will be changed to accept either " Operation restricted" or
" Operation prohibited."
Facility coment on question / answer 2.01:
This answer could be referenced from two separate sources causing answers to differ somewhat from the given reference. Auto closure of the diesel generator output breakers is also covered in 4.16 KV station auxiliary system.
(Reference B.9.6-0008-0010)
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Examiner resolution:
Agree, the difference in answer key and the supplied reference is in the depth of the answer. Either source is adequate for this question.
The answer key will be expanded to reflect both references.
Facility comment on question / answer 2.04:
Vacuum breakers could be listed in reverse order.
i.e. Rx Building to torus Torus to drywell Examiner resolution:
Comment noted, The answer key is not order specific. No change to answer key required.
Facility comment on question / answer 2.05:
Student may state that the affected core spray pump be placed in " Pull to lock".
(Ref. C.4-0130)
Examiner resolution:
Agree, as discussed during the exit meeting, the answer is required per answer key. No credit lossed or gained for "... place pump in pull to lock.".
Facility comment on question / answer 2.07.1 Answer could also include cooling to the steam packing condenser.
(Ref. B.6.5-0005) .
Examiner resolution:
Agree, Answer key changed to: Provide cooling (.33) for SJAE (.33) and
' steam packing condenser (.33).
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Facility comment on cuestion/ answer 2.11:
Two additional requirements to auto open low-low set SRV's.
- 1. SRV control switch in auto.
- 2. < 50 # d/p tailpipe to drywell pressure.
New setpoints for low-low set SRV actuation (Just Incorporated)
Student may use the either the old or new setpoints.
H --> 1052 psig.
G --> 1062 psig.
E --> 1072 psig.
(Reference B.3.3-0013)
Question is misleading in that it does not specify which logic train.
Either Div. I or Div. II will actuate low-low set in a one out of two taken once logic. A reactor scram is required which uses a one out of two taken twice logic scheme. Confusion may arise as to which logic scheme is being asked for.
(Reference B.3.3-0013)
Examiner resolution:
Agree, two additional requirements to auto-open SRV's added 9 0.25 to answer points ea.
key. The question will be rescored with 4 answers for the signals and the logic # 1.0 point. Question value will remain at 2.0 points.
Disagree, the question asks for LOW-LOW SET LOGIC, not the scram logic or div 1 logic. No changed to answer key required.
Examiner correction to question / answer 2.11:
Removed the second "the" from qu)stion.
I Facility comment on question / answer 3.05:
Part a:
If the individual m/a station is shifted to auto, the recirc pump will increase to the speed demanded by the master controller at a rate limited to 2% per sec. The rate limiter limits the rate of change of the demand signal, therefore the mismatch summer never sees an error in excess of 10%.
(Reference B.5.8 - 3.4.9)
Part b:
Recire drive motor breaker will trip if indicated discharge valve position is less than 10% open greater than 66 seconds after pump start.
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(Reference B.1.4-29)
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Examiner resolution:
l Part a Agree, reference material provided and executing the transient on the Monticello simulator demonstrates the answer is according to the facility conment.
Part b Disagree, the reference material supplied did not support the comment that the recirc pump would trip. Futher telephone conversation with the utility confirmed this. This portion of the comment was withdrawn.
The removal of the discharge bypass valves is noted. The removal has no effect on the answer.
I Facility comment on question / answer 3.06:
Question is misleading in asking for the four conditions which will illuminate the light, when actually there is only one condition that will illuminate the light, SRM > 100 cps. Examinees may strive for
! more then one answer due to the question requesting more.
Reference B.5.1.1 - 0007)
Examiner resolution:
I Agree, The conditions listed in the answer key will bypass the SRM RETRACT NOT PERMITTED ROD BLOCK but not light the retract permit light.
The answer is changed to "SRM > 100 cps." and question value reduced from 2.0 to 1.0.
l Examiner correction:
On first line changed " green light" to " white light".
On second line changed " List four conditions ..." to " List the condition ...".
Facility comment on auestion/ answer 3.07:
- Question does not solicit the given answer. Initial plant conditions were not given nor was the system specified. Answers could vary
' depending on plant conditions and system the examinee assumed.
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Reference:
- 1. B.3.1 - 0008, 0009 l
I l 2. Cold Shutdown Valve Operability Tests Steps 1-4, 6-7, 10. 15, 21-22)
Examiner resolution:
Agree, the answer key is correct, according to Monticello's supplied information, but is not correct to "as built" plant. The various actual nethods of testing check valves and different plant conditions required invalidate this question. This question is deleted from the
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Facility comment on question / answer 3.13:
ATWS is also initiated at -47" for greater than 9 seconds.
(Reference B.5.6 - 0048)
Examiner resolution:
Agree, ATWS will be added to answer key as an acceptable answer.
Facility comment on cuestion/ answer 4.01:
Students could also use C.1 as a reference for a temporary change:
Temporary change reviewed and initiated by a Shift Supervisor and one other licensed SRO.
It does not conflict with the Technical Specifications or interfere with the safe operation.
(Reference C.1 - 0005)
Examiner resolution:
Agree, as dicussed during the exit meeting an RO may reference C.1 as it is more the SRO's responsibility to insure the intent of a procedure is not changed. The answer key will be changed to accept the exixting answer or the Facility referenced answer from C.I.
l Facility comment on question / answer 4.02:
l Candidate Brian Koenig has indicated that he placed procedure steps in their proper order of occurrence as per C.1 on the answer sheet, but failed to transfer the pressure that each will be performed or occur from the question sheet where he had listed them.
Examiner resolution:
Agree, the entire exam is treated as limited access material, therefore answers inadvertanly placed on the question sheet may be given full credit if correct.
Facility comment on question / answer 4.06:
Students could also indicate that Low-Low set SRV's would auto-matically control pressure since they do not rely on offsite power for actuation.
(Reference B.3.3-0015)
I Examiner resolution:
Disagree, the procedure requires pressure to be controlled MANUALLY !
between 800 and 1000 psia, which is below the SRV low-low setpoint of i 1052.
Facility comment on question / answer 4.08:
C.4 does state that the three methods are listed in preferred order.
However, it also states that the main objective is to insert as much negative reactivity in as short a time as possible. It also states that it is impossible to predict which method will produce the desired objective for all situations. From an operational standpoint, it is better to know the three methods available and then determine the preferred order on a case by case basis.
(Reference C.4 - 0009)
Examiner resolution:
Disagree, it may be impossible to predict which method will accomplish the disired objective for all cases, however this does not imply MOST cases. The preferred order would insert maximum reactivity in the least time for most cases. If a method is not available, (i.e. the scram cannot be reset) the operator is not changing the priority but using the methods available.
Agree, (last sentence of comment) It is more important to known the three methods than the methods and preferred order. This was part of the intent of the question. To clarify this the point breakdown will be identified on the answer key as .75 for each method and .75 for the correct (preferred) order. Point value for the question remains unchanged.
Facility comment on question / answer 4.10:
CROH, LPCI, core spray, condensate service or RHR/CS pressurizing stations could also be used for makeup depending on plant conditions.
i HPCI, RCIC, LPCI or core spray could be used for cooldown depending on plant conditions.
(Reference C.4 - 0203, 0204) l l
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Examiner resolution:
Agree, If the candidate identifies an available high pressure system or assumes the plant is at low pressure in the answer a low pressure system will be accepted.
Agree to HPCI and RCIC for use as cooldown. LPCI & CS accepted only if the candidate assumes the plant is at low pressure in the answer.
Facility comment on cuestion/ answer 4.12:
Two additional indications to confirm SLC flow are:
- 1. Squib indicating light extinguishes.
- 2. Steam flow from the vessel drcreases.
(reference 8.3.5 - 0026, 0027)
Examiner resolution:
l Agree, the two additional indications will be added to the answer key.
The question will be graded as 3 of 5 required. No change to question value. ,
Facility comment on Question / answer 4.14:
Question could also be interpreted to refer to the Emergency Power
- Reduction Procedure as the correct answer since one of the abnormal l
conditions listed that might require emergency power reduction is low condenser vacuum.
(Reference C.4 - 0198)
J Examiner resolution:
Disagree, the question states "... a slow loss of vacuum". The Emergency Power Reduction procedure does not adress the requirments of B.6.3. Running the RECIRC pumps back will be acceptable vice " reduce power" 1
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ATTACHMENT 2 RESOLUTIONS TO MONTICELLO'S COMMENTS ON NRC SR0 LICENSE EXAMINATION OF 7/22/86 Facility comment on cuestion/ answer 5.06:
eff.
May be explained using the period equation T = 1/
Examiner resolution:
Agree, with the condition that the candidates' answer includes an EXPLANATION as required by the question. The equation, by itself, does not answer the question.
Facility comment on question / answer 5.10:
Trainee may only state that the higher powered bundles are in the center not that the center bundles are approximately double.
4 Examiner resolution:
Agree, the answer requires only that the power in the center bundles is greater than power in the edge bundles. As explained during the EXIT MEETING, information inside parens in the answer key is not required for full credit but is to aid grading of the exam if that information is given by the candidate.
Facility comment on question / answer 5.11:
Question asks for turbine steam flow and the answer key discusses reactor steam flow, however, the same explanation applies.
Examiner resolution:
Agree, The use of reactor steam flow or turbine steam flow in the answer of this question is acceptable.
Facility comment on question / answer 5.12:
Trainee may include Pu-241 buildup or just say Plutonium buildup for part b.
Reference M 8102 L-007 Rev. 2 Examiner resolution:
Agree, Pu-239, Pu-241, or Pu will be acceptable as ONE answer for the question. The other answer remains "U-235 depletion".
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Facility comment on cuestion/ answer 6.03. l The rod need not be disarmed if reactor is sufficiently suberitical such that it would remain substantially subcritical with the uncoupled rod fully withdrawn.
Reference C.4.III.f Page 67 Examiner resolution:
Disagree, The referenced information applies to the "... actions to prevent a recurrence must be completed ..." portion of the procedure.
Facility comment on question / answer 6.04:
Part a If the individual M/A station is shifted to auto, the recirc. pump will increase to the speed demanded by the Master controller (not necessarily set at minimum) at a rate limited to 2 %/sec. The rate limiter limits the rate of change of the demand signal. Therefore, the mismatch summer never sees an error in excess of 10%.
Reference B.S.8 pp. 3,4, and 9 Part b Recirc. MG set B drive motor breaker will trip if indicated discharge valve position is less than 10% open greater than 66 seconds after pump start. Also, the discharge bypass valves M0-2-54A and B were removed during the recirc piping outage of 1984.
Reference B.1.4 P. 29.
Extminer resolution:
Part a Agree, reference material provided and executing the transient on the Monticello simulator demonstrates the answer is according to the j facility comment.
I Part b Disagree, the reference material supplied did not support the comment that the recirc pump would trip. Futher telephone conversation with the utility confirmed this. This portion of the comment was withdrawn.
The removal of the discharge bypass valves is noted. The removal has no effect on the answer.
q Facility comment on question / answer 7.08:
Question "a" does not solicit the answer. Trainee may a;swer in regards to steam flow signal automatically placing the RWM in service.
Extminer resolution:
Agree, Answer to part "a" changed to "Placed in service automatically when power decreases below 35% as measured by steam flow.". Point value is reduced from 1.5 to 0.5 and question value is reduced from 2.5 to 1.5.
Extminer correction to question / answer 7.09:
The word " channels" in the second line of the question was changed during the exam to "AGAF's". This changed was announced to all candidatas and wrote on the chalk board.
Corrected a typo in the answer. "(#3 & #5)" was changed to
"(#4 & #5)".
Facility comment on cuestion/ answer 7.10:
Part b Trainees may not list the possible methods which may be used to cooldown as low as practical prior to placing shutdown cooling into service due to the wording of the question.
Extminer resolution:
Agree, Methods of cooldown deleted from answer and replaced with
" ... prior to placing shutdown cooling into service.". Points (0.33) redristributed in question 7.10b.
Facility comment on question / answer 8.04:
The trainee may list authorization to use potassium iodide.
Reference A.2-304 p. 1 Extminer resolution:
Agree, "Authization to use potassium iodide." added to answer key.
Facility comment on cuestion/ answer 8.11:
The C.4 manual does state that the three methods are listed in preferred order, however, it also states that the main objective is to insert as much negative reactivity ;in as short a time as possible. It
also states that it is impossible to predict which method will produce the desired ebjective for all situations. From an operational stand point it is better to know the three methods available and then determine the perferred order on a case by cases basis.
Reference C.4.I.C page 9 and C.4.If page 15.
Examiner resolution:
Disagree, it may be impossible to predict which method will accomplish the disired objective for all cases, however this does not imply ~MOST cases. The preferred order would insert maximum reactivity in the least time for most cases. If a method is not available, (i.e. the scram cannot be reset) the operator is not changing the priority but using the methods available.
Agree, (last sentence of comment) It is more important to known the three methods than the methods and preferred order. This was part of the intent of the question. To clarify this the point breakdown will be identified on the answer key as .75 for each method and .75 for the correct (preferred) order. Point value for the question remains unchanged.
Facility comment on question / answer 8.12:
Duplicate of 7.12 should be deleted from section 8.
Examiner resolution:
! Agree, Question 7.12 and 8.12 are duplicate questions. The facility was given the option to delete the question from either section.
i I Facility comment on question / answer 8.13:
Trainee may list loss of turbine protective logic, main generator protective logic, ECCS logic, etc. for the loss of auto-protective features. May also discuss of lube oil to recirc MG's, etc.
Reference C.4.III. pp. 51-56.
Examiner resolution:
Agree, as stated in answer key, other answers with justification are acceptable. The primary objective is the understanding of why the loss of a 125 Vdc system is a serious condition requiring a difficult shutdown if not recovered, not the identification of individual loads that are lost.
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U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MONTICELLO REACTOR TYPE: BWR-GE3 DATE ADMINISTERED: 86/07/22 EXAMINER: KING. M.
CANDIDATE:
INSTRUCTIONS TO CANDIDATEL Use separate paper for the answers. Write answers on one. side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. grade The passing crade requires at least 70% in each category and a final of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 28.00 26.79 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 27.00 25.84 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 25.00 23.92 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 24.50 23.44 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 104.50 Totals Final Grade All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil only to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a ngw page, write only an one side of the paper, and write "Last Page" on the last 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.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
- 15. Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY AN3WER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the s m.iner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer,
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all scrap paper and the balance of the paper that you did not use for answering the questions.
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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- PAGE 2
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS QUESTION 5.01 (2.50)
Describe the operating principles of the jet pump used in your recirculation system.
QUESTION 5.02 (3.00)
Indicate HOW each of the coefficients are effected (Increase, Decrease or Remain the same] by each of the three parameters listed? Consider each parameter seperately.
- a. Rod Worth (delta K/K/ Bank) by:
- 1. Moderator temperature INCREASE 3
- 2. Voids DECREASE
- 3. Fuel temperature INCREASES [3 0 0.33 ea]
- b. Alpha Dopple r (delta K/K/ F fuel) by:
- 1. Core age INCREASES
- 2. Fuel temperature DECREASES
- 3. Voids DECREASE [3 @ 0.33 ea]
- c. Alpha Voids (delta K/K/ % voids) by:
- 1. Fuel t(mperature INCREASES
- 2. Core age INCREASES
- 3. Control Rod Density INCREASES [3 @ 0.33 ea]
QUESTION 5.03 (2.00)
Explain WHY it is desirable to maintain the axial flux peak low in the core during BOL and WHAT can happen if this is not done. (2.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 3 THERMODYNAMICS QUESTION 5.04 (1.00)
The 8x8 fuel has a thermal time constant of approximately 5 to 6 sec-onds. This means that in 5 to 6 seconds following a sudden power in-crease: (Choose ONE answer below) (1.0)
- a. The fuel centerline temperature will reach its maximum (final) value.
- b. Clad surface temperature will reach its final value.
- c. Fuel centerline temperature will reach approximately 2/3 of its final value.
- d. Fuel centerline temperature, clad surface temperature and coolant temperature have each reached their equilibrium (final) values.
- e. Clad surface temperature will reach approximately 63% of its final value.
QUESTION 5.05 (1.00)
During high power operations (>60%), WHY is it more desirable to change power with recirculation flow than with control rods? -
(1.0)
QUESTION 5.06 (2.00)
Three (3) minutes following a reactor scram from high power, indicated reactor power is 75 on range 5 and decreasing.
[
- a. What will INDICATED power (level and scale) be one (1) minute later?
(fihow calculations) (1.0)
- b. Explain why power decreased at this rate. (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 4
- 5. 'THEQBY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS QUESTION 5.07 (1.00)
If the highest pressure feed heater is removed from service (extraction steam isolated), what happens to Megawatt output of the generator and why? (1.0)
QUESTION 5.08 (2.50)
With the reactor operating at 75% power, recirculation flow control fails, rapidly increasing flow,
- a. Which reactivity coefficient WILL ACT FIRST TO LIMIT the power excursion? (0.5)
- b. Explain your choice of coefficient. Insure your explanation addresses all other coefficients. (2.0)
QUESTION 5.09 (1.50)
MATCH the Failure Mechanism from column (1) AND the Limiting Condition from column (2) WITH the associated Power Distribution Limits (a-c) below.
- a. Linear Heat Generation Rate (LHGR)
- b. Average Planer Linear Heat Generation Rate (APLHGR)
- c. Minimum Critical Power Ratio (MCPR) 1 - FAILURE MECHANISM 2 - LIMITING CONDITION
- 1. FUEL CLAD CRACKING DUE TO LACK 1. 1% PLASTIC STRAIN OF COOLING CAUSED BY OTB
- 2. FUEL CLAD CRACKING DUE TO HIGH 2. PREVENT TRANSITION STRESS FROM PELLET EXPANSION BOILING
- 3. GROSS CLAD FAILURE DUE TO DECAY 3. LIMIT CLAD TEMP HEAT & STORED HEAT FOLLOWING TO 2200 F A LOCA (1.5)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AFD PAGE 5 THERMODYNAMICS QUESTION 5.10 (2.50)
A design feature in the reactor vessel ensures proper flow distribution through the core fuel bundles.
- a. What is this feature ? (0.5)
- b. EXPLAIN what would happen on a power increase with "NO CHANGE IN MEASURED CORE FLOW" if this feature were eliminated. (2.0)
QUESTION 5.11 (3.00)
Assume a normal power increase of 10% is made with the recirculation pumps.
Plot and explain each of the following parameters from the beginning of the transient to the final power level (+10%).
- a. reactivity EXAMPLE: l FLOW :...........................
- b. void fraction !
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- c. reactor pressure _ TEMP l...........................
- d. reactor power l PRES l...........................
- e. reactor period l l
- f. turbine steam flow LEVEL !...........................
l l (6 @ 0.5 ea.)
TIME QUESTION 5.12 (2.50)
The Core Average Delayed Neutron Fraction changer from core BOL to core EOL.
A. Does it INCREASE or DECREASE? (0.5)
B. Give TWO reasons for the change. (1.2)
C. Briefly explain how the change affects operational control of the reactor. (1.0)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
PAGE 6
- 5. ' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS QUESTION 5.13 (1.50)
During a lecture an instructor states "For a constant reactor reactor period, it takes the SAME AMOUNT OF TIME to change reactor power from 1% to 5% as it does to change it from 10% to 50%."
Show the calculations needed to prove or disprove his statement.
QUESTION 5.14 (2.00)
For each condition given below, indicate whether it will cause an INCREASE, a DECREASE, or have NO EFFECT on CRITICAL POWER:
A. Local peaking factor DECREASES B. Axial power peak shifts from BOTTOM to TOP of channel.
C. INCREASE in fuel bundle flow.
D. DECREASE in inlet subcooling.
(***** END OF CATEGORY 05 *****)
PAGE 7
-6. ' PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION QUESTION 6.01 (3.50)
A. How is the HPCI Turbine exhaust line protected against overpressure? (Identify two means, include setpoints). (1.5)
B. Identify which of the following are direct HPCI turbine trips and which are HPCI system isolations: (2.0)
- 1. HPCI steam line low pressure
- 2. Reactor high water level
- 3. HPCI steam line area high temperature
- 4. Pump suction low pressure QUESTION 6.02 (3.00)
Considering the Main Steam Line Radiation Monitoring System:
- a. How many, what kind, and where are the detectors located? (1.5)
- b. What three (3) automatic actions (excluding GP 1 isolation and.all alarms) take place when the system logic is tripped? (1.5)
QUESTION 6.03 (3.00)
- a. What are two (2) plant systems or components that receive rod position information from the RPIS, OTHER THAN the full core and 4 rod group displays? (1.0).
- b. With a selected rod at notch position 18, and its 02 notch position reed switch stuck shut, what will the 4 rod group display indicate for the selected rod's position? (0.5)
- c. What indications will result from a control rod moving to the overtravel position during a coupling check ? (0.5)
- d. After receiving the above indications (control rod in the overtravel position), What must be done per C.4.III.F. Control Rod System Failure?
(1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
PAGE 8
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION i
QUESTION 6.04 (2.00)
The reactor is operating at power. What will be the effect on/of the recirculation flow control system due to the following -
conditions:
- a. The reactor is at 30% power when an operator inadvertently shifts the "A" recire pump M/A transfer station to AUTO. (1.0)
- b. The reactor is at 95% power with recirculation flow control in master manual when full open indication on recirc pump "B" discharge valve (MO 2-53B #12) is lost. NOTE: MO-2-54B the recire pump #12 bypass discharge is open (1.0) i 4
QUESTION 6.05 (1.00)
Main steam pressure is being controlled by a normal lineup with the EPR in control at 65% power. Recire control is in master manual.
Assume a turbine bypass valve has failed open. What must be done per C 4 III.J, Turbine System Failure. (1.0)
QUESTION 6.06 (3.00)
The Core Spray System receives a valid initiation signal. One pump fails to start:
- a. WHY must that core spray loop be isolated? (0,5)
- b. HOW is the isolation accomplished (be specific)? (1.0)
- c. What valve in the Core Spray system is used by the operator (0.5) to throttle injection flow.7
- d. If the torus temperature reaches 130 F, the core spray system must be throttled. Explain why and to what flow rate.
(Assume normal system operation) (1.0)
QUESTION 6.07 (1.00)
Why are APRM channels 2 and 5 or 1 and 6 left in the bypassed state when not otherwise needed? (1.0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
" - --wv- .-w-i._< -,-.--,,m-y * -s. ,,- , , , ,s w 9,,, -,.-y-p -.-m ,pq,----_,,-p3 . ,,,,,,..--%* ,.g ,n,g e,m- ~q~w cw4m...w- p.,.y -
- 6. ' PLANT SYSTEMS DESIGN. CONIROL. AND INSTRUMENTATION PAGE 9 QUESTION 6.08 (3.50)
The Residual Heat Removal system can take a suction from four (4) sources cnd discharge to nine (9) areas.
- a. WHAT are three (3) of the auction sources? (1.5)
- b. Excluding the #11 and #12 recirc loop and with the crosstie open, WHAT are four (4) areas that each loop is capable of suppling water? (2.0)
QUESTION 6.09 (3.00)
With regard to the Main Steam System:
- a. Explain what phy=ically causes the MSIV closing speed during EXERCISING to be much slower than the normal closing speed?
Your answer should include HOW the valve is closed (motive force). (1.5)
- b. Explain HOW/WHY a relief valve discharge pipe (tail pipe) could be damaged due to its vacuum breakers sticking shut during repeated actuation (lifting) of the relief valve. (1.5)
QUESTION 6.10 (3.00)
For each of the HPCI (High Pressure Coolant Injection) System component failures listed below, STATE WHETHER OR NOT HPCI WILL AUTO INJECT into the reactor vessel, IF IT WILL NOT INJECT WHY, AND IF IT WILL INJECT, provide ONE POTENTIAL ADVERSE EFFECT OR CONSEQUENCE of system operation with the failed component.
Assume NO OPERATOR ACTICN, and the component is in the failed condition at the time HPCI receives the auto initiating signal,
- a. The GLAND SEAL EXHAUSTER fails to operate. (0.75)
- b. The turbine AUXILIARY LUBE OIL PUMP fails to operate. (0.75)
- c. The MINIMUM FLOW VALVE fails to auto open (STAYS SHUT) when system conditions require it to be open. (0.75)
- d. The RAMP GENERATOR fails low. (0.75)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
- PAGE 10
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION QUESTION 6.11 (1.00)
The Shutdown procedure states that following a shutdown /cooldown the condenser hotwell can be maintained above 100 F by using the RFP recire line instead of the condensate recirc lines. (1.0)
Why does RFP recirc prevent cooling of the condenser hotwell ?
(***** END OF CATEGORY 06 *****)
- 7. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11 RADIOLOGICAL CONTROL QUESTION 7.01 (3.00)
At what pressure will each of the following be normally performed or occur during a reactor startup from cold conditions per start up procedure C.1, Heatup and Pressuriation 7
- c. The mechanical pressure regulator is allowed to open the Main Steam Bypass valve #1 to verify regulator operation.
- b. The RCIC Automatic isolation signal is reset.
- c. The HPCI Automatic isolation signal is reset.
- d. The mechanical pressure regulator override is adjusted to open the #1 Main Steam Bypass valve 10 - 15%.
- e. Electric pressure regulator is verified to assume pressure control,
- f. The Air Ejector Suction Isolation Valve Control Switch is placed in the Auto postion.
(6 @ 0.5 ea)
QUESTION 7.02 (3.50)
In accordance with the approach to criticality steps in the cold startup procedure, C.1, answer the following.
- a. What are the RO's required actions if criticality does not occur within the predicted critical rod pattern band indicated on Predicted Critical for Plant Startup form #2159? (0.75)
- b. When is the reactor considered critical? (0.75)
- c. What four (4) items are recorded, in the reactor log and on the predicted critical form, when criticality is established? (1.0)
- d. What are three (3) ways, that reactor period may be determined? (1.0)
(Not read off period meter.)
QUESTION 7.03 (1.50)
With who's (position title) approval and with what plant conditions may the Plant Restart Checklist be implemented 7
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
~
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 12 RADIOLOGICAL CONTROL QUESTION 7.04 (3.00)
Due to a malfunction the process computer is not available. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> previous power was reduced from 90% to 65%. You are requested to return to 90% power. Briefly explain what must be considered, including what restrictions apply, to increase power per C.2 Power Operation (3 required). (Assume all equipment is operable, except process computer.)
QUESTION 7.05 (1.50)
Explain why C.2, Power Operations does not require APLHGR, LHGR, and MCPR to be checked when power is less than 25%.
QUESTION 7.06 (1.00)
Monticello is load following and is requested to reduce load.
What restriction may apply to the rate of power reduction due to environmental conditions ?
QUESTION 7.07 (1.50)
When increasing power above the preconditioned envelope the maximum rate is A [.5]. This rate is applied to:[1.0]
(select the correct answer)
- a. Fuel nodes rates increasing most rapidly.
- b. Average of fuel node rates.
k
- c. Fuel rods rates increasing most rapidly.
- d. Average of fuel rods rates. (1.5)
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
~
- 7. PROCEQUEES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL QUESTION 7.08 (1.50)
- a. A normal shutdown is in progress. At what power level will the RWM be placed in service and briefly explain how this is accomplished. (0.5)
- b. When will the RWM rod blocks become effective ? (0.5)
- c. What is required to bypass the RWM if it is inoperable during a shutdown ? (0.5)
QUESTION 7.09 (2.00)
The computer heat balance calculates plant power to be "95%". The An OD-3 printout shovs APRM AGAFs reading:
APRM 1 0.985 APRM 2 1.015 APRM 3 1.022 APRM 4 1.020 APRM 5 1.025 APRM 6 0.995
- a. Which APRM's require adjustment ? (1.0)
- b. Beside adjustment. Explain what actions are required due to the APRM readings (Include in your answer which APRM channel (s) require the actions) (1.0)
QUESTION 7.10 (2.00)
- a. During a cooldown with only 1 recirc pump running which loop (running or idle) is preferred for RHR injection AND why ? (1.0)
- b. Explain the procedural requirements if it is desired to use the non-preferred loop. (1.0)
QUESTION 7.11 (1.00)
Why is reactor level maintained between +50 and +55 by Shutdown procedure C.3 when the reactor is in cold shutdown ?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- I
- 7. PROCEDURES - NORMAL. ABNORMAL. EtfEEEENCY AND PAGE 14 RADIOLOGICAL CONTROL l QUESTION 7.12 (1.00)
A valid low-low level has been received. Under what conditions would the 106 second auto-blowdown timer be blocked ?
QUESTION 7.13 (2.50)
During Power Operation:
- a. What acticns must be taken upon receipt of an AGAF alarm? (1.0)
- b. What are three instances when a TIP scan should be performed?(1.5)
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f
(***** END OF CATEGORY 07 *****)
i
- 8. ' ADMINISTRATIVE PROCEDURES CONDITIONS, AND LIMITATIONS PAGE 15 QUESTION 8.01 (1.00)
Which ONE of the following events is a Significant Operating Event (SOE) per ACD-3.9, Operating Events ?
- b. An Unusual Event has been declared by the Shift Supervisor.
- c. An inadvertant initiation of HPCI by an operator during a surveillance injecting < 500 gallons of water.
- d. A small brush fire on Thompson Island is reported to be under control by security.
QUESTION 8.02 (3.00)
Briefly explain WHY each of the following RECIRCULATION SYSTEM LIMITATIONS are necessary.
- a. With both pumps running, the speed of the faster pump may not exceed 130 percent of the speed of the slower pump for a core power less than 80 percent. (1.0)
- b. The operating pump must be reduced to 50 percent speed or less prior to restarting the tripped pump. (1.0)
- c. Recirculation flow shall not be increased unless the coolant temperature difference between the bottom head region and upper region of the vessel is less than 145 degrees Fahrenheit. (1.0)
QUESTION 8.03 (2.00)
During performance of a procedure a change is required due te plant conditions.
- a. Under what restrictions may a temporary change be made ? (1.0)
- b. How is approval for the change obtained and documented ? (1.0) i
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
l
R. " ADMIllISTRATIYE_EBQCEDURES. CONDITIONS. AND LIMITATIONS PAGE 16 QUESTION 8.04 (2.00)
As Shift Supervisor you have declared an alert and have assumed the position of Emergency Director. List four responsiblities of the Emergency Director that mar NOT be delegated. (4 0 0.5 ea.)
QUESTION 8.05 (1.00)
To control the plant during an emergency an operator is performing steps from the EPIP's. The operator states if he continues he may conflict with a Tech. Spec. requirement. What should the operator be instructed to do ?
QUESTION 8.06 (1.00)
A loss of offsite power has occured. The diesel generators have auto-started and loaded. Reactor vessel level is being controlled by RCIC in manual, pressure is being controlled by manual operation of relief valves per procedure C.4.III, Electrical Systems Failures. Which of the following loads must be MANUALLY restored as soon as possible (Choose one)?
- a. RPG MG set #12
- b. Fuel pool cooling pumps
- c. Circulating water pumps
- d. RBCCW pumps QUESTION 8.07 (3.00)
When the reactor water temperature is greater than 212 F the the Tech. Spec. list four temp limits for the suppression pool.
For each temperature limit below identify the Tech. Spec.
limit / requirement. (4 @ 0.75 ea.)
e
- a. less than or equal to 90 F
- b. less than or equel to 100 F c, greater than 110 F
- d. greater than 120 F
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
8.* ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 17 QUESTION 8.08 (2.00)
When attempting to withdrawal a control rod during a reactor startup the rod worth minimizer (RWM) fails. Explain if the startup may continue AND what (if any) restrictions apply.
QUESTION 8.09 (2.00)
While operating at 100 % power a group I isolation and a reactor scram occur. DATA collected from the computer (and operators) indicates the following occured:
- a. The group I isolation was caused by an Technician error,
- b. The reactor scram was caused by high reactor pressure.
- c. Both feed pumps tripped.
- e. An operator secured HPCI, took manual control of RCIC and maintained reactor vessel level below 48".
- f. A feed pump was restarted and level control transfered to the feed pump, RCIC was secured.
- g. Continuing to follow Monticello procedures, the plant is placed in a normal shutdown condition.
Select and EXPLAIN the correct statement.
- a. Power operation cannot resume because HPIC auto-initiated and injected to the reactor vessel,
- b. Power operation cannot resume because a safety limit may have been violated,
- c. Power operation cannot resume because the feed pumps should rot have tripped.
- d. Power operation cannot resume until the MSIV's have been inspected due to the Group I isolation signal.
QUESTION 8.10 (1.00)
What is the procedural definiticn of at ATWS?
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
ADMINISTRATIVE PROCEDURES, CONDITIONS. AND LIMITATIONS PAGE 18 8.
QUESTION 8.11 (3.00)
An ATWS has occured. All scram valves have been verfied open. List 3 methods IN THE PREFERRED order to insert the control rods.
QUESTION 8.12 ( .00)
Question / answer / reference deleted from exam.
QUESTION 8.13 (2.50)
Procedure C.4.III, Abnormal Conditions-Loss of 125 V DC bus A or bus B, states: A loss of 125 V DC bus will not initiate any operational transient. Explain why this is still a severe failure (Three reasons required) AND briefly explain the final conditions of the plant due to this loss. ( Assume the bus CANNOT be recovered.)
QUESTION 8.14 (1.00)
Due to a fire the Monticello's fire department assistance is required. Choose the correct statement:
- a. Plant manager concurrence must be obtained prior to requesting off-site assistance,
- b. Notify security so an escort will be arranged,
- c. Only fire fighters with Monticello Radiation Training may respond to the request,
- d. The Security Dept. will decide if off site assistance is required dependent on the available of security personnel to fight the fire.
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMINATION ***************)
EQUATICN $HEET Cycle aH iciency = (Net G f = .na v = s/t cut)/(Energy in) 2
. = ng s = V3 t - 1/2 at s .
E =.nc- A = 1:4 A=Ae3 AE = 1/2 av a = (Vf - 13 )/t PE = agn
- e/t 1 = M2/t1/2 = 0.693/t1/2 Vf = V, + at
.g , , j A=
aDE 1/2*" " [(c"') , (t } j 4
- .E = 931 m
-n m = V,yAo !,Ieo Q = mCoat I = I g e'"*
d = UAa T ! = I ,10** ##l Pwr = W ,4h TIL = 1.3/u su gyg , ,g,gg37, P = Po 10 r(t)
? = P ,e*/ ' SG = S/(1 - X,g)
SUR = 25.06/T G, = 5/(1 - <,gx)
G;(1 - X,g;) = G 2II ~ *a#2)
SUR = 25s/t* + (s - o)T M = 1/(1 - X,g) = G;/G 3 T = ( l'/s ) + ((a - s '/ Io] M = (1 - X ,g ,)/(1 - K ,g7)
T = V(s - a) SCM = ( - K ,g)/K ,g T = (a - o)/(Ta) t- = 10 sec:nes a = (K ,g-1)/K ,g = t.X ,g/X,g I = 0.1 sec:ncs'I o = CD'/(T K,g)] * [T,g/(1 + IT)] Id Id l i *2 ,2 gd 2
Id j 22 P = (:sV)/(3 x 1010) 2 R/hr = (0.5 CE)/c (=eters)
- = :N R/hr = 6 CE/cI (feet) ,
Miscellaneous Ccnve-siens Water Partnetsas 10 333 I curia = 3.7 x 10 1 gal. = 8.345 lem. 1 kg = 2.21 lem 1 gja .==7.48 3.78gal.
litars Ino=2.54x103Stu/nr 1 f. 3 1 = = 3.41 x 100 5tu/hr Density = 62.41 erg /ft lin = 2.54 :s Censity = 1 gm/c::r' *F = 9/5'C + 32 Heat of vaccritation = 970 Stu/ lcm *C = 5/9 ('F-32)
Heat of fusion = 144 3:u/le:u - 1 STJ = 778 ft-lbf 1 Acm = 14.7 ssi = 29.9 in. Hg.
1 ft. H 2c
- 0*433! 1*f/I"*
~
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 19 THERMODYNAMICS ANSWIRS -- MONTICELLO -86/07/22-KING, M.
ANSWER 5.01 (2.50)
In the nozzle, the high static head is converted to a high-velocity jet at a low static pressure [0.6]. The low pressure at the nozzle discharge draws the surrounding fluid into the throat where it is mixed [0.6].
A pressure rise occurs in the mixer section due to velocity profile rearrangement and momantum transfer in the mixing process [0.6].
The fluid enters a diffuser section which slows the relatively high velocity mixture and converts the dynamic head into static head [0.7].
REFERENCE Monticello, Thermodynamics and Fluid Flow, ch 9 pg 38 ANSWER 5.02 (3.00) a.l. increase a.2. increase a.3. remains the same b.1. increase
< b.2. increase b.3. decrease c l. increase i
c.2. decrease c.3. increase [9 @ 0.33 ea]
REFERENCE Monticello, Reactor Theory L.P., # M8102L-043 Rev 0, Figure 43 pg 43 of 43 ANSWER 5.03 (2.00)
Keeping the peak low in the core decreases peaking problems at EOL
[0.5). At EOL all control rods are fully withdrawn and large flux '
peaks occur. Without proper burnout, the bottom would be excessively reactive and rods would have to be inserted to control peaking. These rods would be very difficult to withdraw later[1.5] (2.0)
REFERENCE Monticello M 8102-L-016, BWR Inherent Reactivity Coefficients, pg. 50.
l
- 5.
- THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 20 THERMODYNAMICS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 5.04 (1.00)
Clad surface temperature will reach approximately 63% of its final value. (e) (1.0)
REFERENCE Monticello BWR Thermo & Thermo Limits, M 8104-L-012, pg 5.
ANSWER 5.05 (1.00)
Changing power with recirculation flow changes total core power while keeping the flux profile relatively unchanged OR to avoid localized flux peaking. (1.0)
REFERENCE REACTOR PHYSICS TRANSIENT ANALYSIS LP, pg 14 ANSWER 5.06 (2.00)
- a. Using P = Po e to the t/T then P = 75 e to 60/-80 P = ?S e to -0.75 = 35 on Range 5 (1.0)
- b. After the initial prompt drop, power cannot decrease faster than the longest lived delayed neutron appears, which has about a (55.6) see half life. (1.0)
REFERENCE Monticello M 8102-L-007 rev 2, pg 18,19,39 & B.5.1.1 S/U range monitors, Jan 86, pg 24,25 ANSWER 5.07 (1.00)
Megawatt output from the generator would increase (0.5). Steam that was formerly being extracted now passes through the turbine to t'he condenser (0.5). (1.0)
REFERENCE Monticello M 8104-1-017, pg 14,15 & GE HT and FF Ch 5, Thermo Cycles and Cycle Analysis, pg 57
PAGE 21 5.'THEQRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 5.08 (2.50)
- a. The fuel temperature coefficient. (1.0)
- b. Due to the time constant of the fuel (generally 5 - 7 seconds) or the amount of time required for heat to be transfered from the fuel to the coolant.[0.b] The moderator temperator or the void coefficient would not have any effect for several seconds.[1.0] The fuel temp.
coefficient acts instantly to insert negative reactivity.[0.5] (2.0)
REFERENCE Monticello M 8102-1-016, BWR Reactivity Coefficients, Attachment 1, pg 17, sec. D.C.29.
ANSWER 5.09 (1.50)
Failure Mechanism Limited Condition (0.2 ea.) (0.3 ea.)
2 1 A. LHGR B. APLEGR 3 3 C. MCPR 1 2 (1.5)
REFERENCE Monticello M 8104-1-012, BWR Thermo & Thermo Limits, pg 7,6
PAGE 22
- 5.
- THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND THERMODYNAMICS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER. 5.10 (2.50)
- a. Will accept core orificing OR orificed fuel support pieces. (0.5)
- b. As power increases the amount of boiling (two phase flow) increases.
[0.5] The amount of power generated in a peripheral bundle is <
(approximately half) that of a center bundle; therefore boiling is greatest in the core center.[0.5] Two-phase flow restricts cooling water flow due to the boiling action.[0.5] This would cause the higher powered bundles to receive less cooling water as their higher resistance to flow, would divert flow to lower power fuel bundles
[0.5] starving the higher power bundles. (2.0)
REFERENCE Monticello System Manual B.1.1, pg B.l.1-0045.0 & GE HT & FF, ch 8 pg 8-45.
5.' THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 23 THERMODYNAMICS ANSWERS -- HONTICELLO -86/07/22-KING, M.
ANSWER 5.11 (3.00)
!!! The following text describes the power increase transient. !!!
!!! The Plots are from GE RT, Ch 7, pg 7-30 and will be attached !!!
!!! to the answer KEY, !!!
The operator begins increasing flow at pt. one. This causes a DECREASE in void fraction and adds' positive reactivity.
Power level begins to rise immediately, increasing fuel element temp.
This implies more heat transfer to the coolant, thus increasing steam generation.
- a. reactivity : increases due to void decrease then returns to zero as fuel temp inc. and voids return adding neg. react.
- b. void fraction : decreases due to recire flow increase then increases as power increases (stm generation inc.)
- c. reactor pressure : reactor press. will increase due to the design of EHC & turbine control system.
- d. reactor power : Power will increase due to + react. from recire flow increase (or void decrease). Power will stabilize
(~ 10% higher) as fuel temp and voids add neg. react.
- e. reactor period . period will become + (from infinite), peak, then decrease going slightly neg. as period returns to inf.
- f. reactor steam flow: steam flow will increase due to EHC/ pressure control system which will open turbine throttle valves to maintain set pressure.
(6 @ 0.5 ea.)
REFERENCE GE HT & FF, Ch 7, pg 7-29,30 e
. j Page 23A l
l 70* PLOW 888# EASE pgCapCULATioM ff
,/
NsTART PLOW seCREAsE T.E gett REACTtylTY (AMIE)
- O TlbtE VOe FRA.CTION y, _
T-REACTOR PRassuRE 8
, S Pe g 4 s _
REACTOR 3 POwsR LEVEL 1
Tadt 3
REACTOR PERICO
- 4 s REACTOR sTEAu PLOW
~
,[
- This The operator begins increasing recirculation flow at point 1.
l causes a decrease in void fraction since increased coolant flow past fuel elements tends to sweep away voids more rapidly than they are formed. This causes a positive reactivity addition by the void coerncient of reactivity.
Power level begins to rise immediately, increasing fuel element thus temperatures. This implies more heat transfer to the coolant, -
increasing steam generation.
l
Page 23B
~
At point,3, the increase in power level begins to take effect, with void Once again, increased steam fraction beginning to increase.
which sends generation is controlled by a pressure regulator, additional steam generation through the turtine, increasing generator power output.
Reactor power level continues to increase until point 4 is reached.
The negative reactivity insertion rate, due to inct easing void fraction and Doppler coefficient, will overcome the positive reactivity still present in the core.
The power level steadles out at point 5, with net reactivity once again equal to zero, and void fraction back to its original value.
t 9
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- 5.
- THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 24 THERMODYNAMICS ANSWERS -- MONTICELLO -86/07/22-KING, H.
ANSWER 5.12 (2.50)
A. DECREASE (0.5)
B. 1. U-235 depletion (0.5)
- 2. Fu buildup (Accept PU, Pu-239, or Pu-241) (0.5)
C. Observed period for a given reactivity addition will be shorter at EOL than at BOL. (1.0)
REFERENCE GE Reactor , Theory,pp. 3-29, 3-36 ANSWER 5.13 (1.50)
Using the equation P = Po e * (t/T) solving for time results in the equation:
t = T x in(P/Po)
From this it can be seen that since 5/1 yields the same value as 50/10, and since all other factors in the equation are equal, the time is equal.
(1.5)
REFERENCE MONTICELLO Lesson Plan M 8102-L-007, Rate of Power Change, sec 2.3 ANSWER 5.14 (2.00)
A. Increases B. Decreases C. Increases D. Decreases (4 9 0.5 ea.)
REFERENCE GE Heat Transfer and Fluid Flow, pp. 9-25 to 9-30
- 6. PLANT SYSTEMS DESIGN. CONTROL, AND INSTRUMENTATION PAGE 25 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 6.01 (3.50)
A. 1. Rupture dies on the exhaust line(.6) at 175 psig(.15)
- 2. Turbine trip (.6) at 150 psig(.15) (1.5)
B. 1. Isolation
- 2. Turbine trip
- 3. Isolation (4 @ 0.5 ea.)
- 4. Turbine trip REFERENCE Monticello HPCI SYSTEM, B.3.2, pg B.3.2-0013 &0003.4 ANSWER 6.02 (3.00)
- a. Gross gamma radiation is detected by four (4) [0.5] gamma sensitive ionization chambers [0.5] located adjacent to and immediately downstream of the outer MSIV's at DW penetration in the steam chase.[0.5]
- b. Logic trip will initiate a reactor scram [0.5] and if operating trip the main condenser vacuum pump [0.5] and suction valve closure.[0.5]
REFERENCE Monticello B.S.11, pg B.S.11-5,6,&7 ANSWER 6.03 (3.00)
- a. RWM and process computer (1.0)
- b. Display will indicate 02 and 18 (0.5)
- c. Blank full core and 4 rod display (for overtraveled rod)
Rod overtravel Alarm (2 9 0.25 ea.)
- d. Immediately insert [0.5] and electrically disarm [0.5] the i
control rod. (1.0)
REFERENCE Monticello RPIS, B.S.4, pg B.S.4-1,8, & 11 l
I l
PAGE 26 6." PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 6.04 (2.00)
- a. The control system will see a step increase to 45% demand signal from the dual limiter and will increase pump speed. The rate limiter will control the speed increase @ 2%/sec - (Assuming no scram occurs).
(1.0)
- b. Loss of open ind. for the discharge valve removes the bypass signal around the speed limiter and it outputs a 24% signal. The "B" pump will slow to demanded speed as an interlock prevents the mismatch summer from generating a " lockup". (1.0)
REFERENCE Monticello Recire Flow Control, B.S.8 pg B.5.8-2,3,4, & fig i Recire System, B.1.4, pg B.1.4-30,31 ANSWER 6.05 (1.00)
- 1. Reduce load as low as possible. (0.33)
- 2. Scram the reactor. (0.33)
- 3. Close MSIV to prevent excessive cooldown. (0.33)
REFERENCE Monticello B.S.9-7,8,9, & fig 2 C.4.III.J. pg 86 ANSWER 6.06 (3.00)
- a. To maintain primary containment integrity (0.5)
- b. Close the inboard isolation valve (.33), position the outboard isolation valve bypass switch to bypass (.33), then close the outboard isolation valve (.33) (1.0)
- c. Inboard isolation valve. (0.5)
- d. 1. To prevent pump cavatition. (0.5)
- 2. 3800 gpm. (3500 - 3900) (0.5)
REFERENCE Monticello CORE SPRAY, B.3.1, pg B.3.1-4a, 6 & 19
6 ' PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 27 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 6.07 (1.00)
The LPRM's of APRM's 1 and 5 and APRM's 2 and 6 are shared, it is possible to have a failure that can trip both scram channels.
(For instance, an LPRM that fails high in APRM 1 may fail high enough to bring both APRM's 1 and 5 to the scram setting.) (1.0)
REFERENCE Monticello, System Description, B.5.1.2, Power Range Monitoring, pg 16, 19 & 20 ANSWER 6.08 (3.50)
- a. 1. Torus Ring Header
- 2. Condensate Storage Tank
- 3. #11 Reactor Recirculation Loop
- 4. Fuel Pool Skimmer Surge Tank (Temporary spool piece required).
[3 G 0.5 ea] (1.5)
- b. 1. Reactor Vessel Head
- 2. Upper Drywell Spray Header
- 3. Lower Drywell Spray Header
- 4. Torus Spray Header
- 5. Torus
- 6. Radwaste Surge Tank
- 7. Fuel Pool Spargers (Temporary spool piece required)
[4 @ 0.5 ea] (2.0)
REFERENCE Monticello, System Description, B.3.4, Residual Heat Removal, pg 2 & 3 4
e
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PAGE 28
- 6. " PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 6.09 (3.00)
- o. The air supply to the MSIV air cylinder is interrupted [0.5] and the air cylinder vents through an exhaust restrictor [0.5.]. Since no air pressure is applied to the top of the air cylinder, the valve closing springs provide the main closing force [0.5]. (1.5)
- b. Following the first actuation of the relief, the steam in its discharge line would condense causing a vacuum in the line [0.5].
This would result in torus water being drawn up into the line [0.5]
which could cause overpressurization of the line on the next act-uation (0.5]. (1.5)
REFERENCE MONTICELLO B.2.4, pg B.2.4-12,13 ANSWER 6.10 (3.00)
- a. Will inject [.25]. Turbine seal leakage resulting in potential air-borne activity in the HPCI room [.5]. (0.75)
- b. Will not inject [.25]. Turbine stop and control valves will not open [.5]. (0.75)
- c. Will inject [.25]. Pump overheating and seal damage may result during low or no flow conditions [.5]. (0.75)
- d. Will not inject [.25]. Minmum signal from the RAMP GENERATOR cause the controller to keep turbine control valves closed.[.5] (0.75)
REFERENCE MONTICELLO HPCI, B.3.2 ANSWER 6.11 (1.00)
RFP recire is below the condenser tubes (condensate recire is above),
This avoids passing the water over the tubes which would remove heat and cool the condenser hotwell. (1.0)
REFERENCE MONTICELLO Shutdone Procedures, C.3, pg C.3-0023 4
r
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 29 RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 7.01 (3.00)
- c. 150 psig
- b. 80 psig
- c. 130 psig
- d. 500 psig
- o. 900 psig
- f. 200 psig [All pressures + or - 10%] [6 @ 0.5 ea] (3.0)
REFERENCE Monticello, Startup Procedure, C.1, Heating and Pressurization, pg 34, 35, 36, 40 & 41 ANSWER 7.02 (3.50)
- a. 1. Discontinue rod withdrawal
- 2. Maintain the reactor suberitical (3 @ 0.25 ea)
- 3. Notify the Shift Supervisor
- b. Neutron Flux rises [0.25] with a constant (stable) period [0.25]
without additional control rod withdrawal [0.25]. (0.75)
- c. 1. the time
- 2. rod position
- 3. period
- 4. reactor water temperature
- 5. srm reading (4 9 0.25 ea.)
- d. o Decade rise divided by 2.3 or multiplied by .435 o Doubling time divided by .693 or multiplied by 1.445 o Time for IRM scale reading to increase by a factor of 2.718 (3 9 0.33 ea.)
REFERENCE Monticello, Startup Procedures, C.1, Cold Startup, Approach to Critical, pg 22
t
- 7. PRQCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30 RADIOLOGICAL CONTROL ANSFERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 7.03 (1.50)
Requires the approval of the SUPERINTENDENT, OPERATIONS [0.5]
and may be implimented after a "short " duration shut down[.33),
when no major maintenance has been performed [.33] or after a scram, if the nature of the scram is known and the cause remedied [.33]. (1.5)
REFERENCE MONTICELLO Start Up Proc, C.1, pg C.1-0001 ANSWER 7.04 (3.00)
- a. Each thermal limit has a margin > 0. (or is less than action limit)
- b. Equalibrium power level, i.e. is 90% below the " previous highest equalibrium value"
- c. Was the plant at equilibrium prior to the power reduction.
- d. Rod withdrawal or power increase after rod movement requires Plant Technical Engineering Staff " guidelines". (3 @ 1.0 ea.)
(Other answers, with justification, will be acceptable)
REFERENCE MONTICELLO Power Operations, C.2, pg C.2-5,6 ANSWER 7.05 (1.50)
For a limiting value (thermal limit) to occur below 25% an unreasonably large peaking factor would be required.[75]. Following a control rod operating sequence prevents an " unreasonably large" peaking factor from occuring[.75]. (1.5)
REFERENCE MONTICELLO Power Operations, C.2, pg C.2-0007,0008,0009 Tech. Spec. Bases 4.11
- 7. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 7.06 (1.00)
When upstream river temperature is <50 F, power reduction will be limited by a maximum 5 F/hr temp. change in the discharge canal temperature.
REFERENCE MONTICELLO Power Operation, C.2, pg C.2-0025 ANSWER 7.07 (1.50) f A (rate) = 0.11 KW/FT/HR Mult. Choice a.
REFERENCE MONTICZLLO, Power Operations, C.2, pg C.2-0026 ANSWER 7.08 (1.50) ,
- a. RWM placed in service automatically at <= 35% power as measured by steam flow. (0.5)
- b. RWM rod blocks become effective @ 20%. (0.5)
If the RWM is inop checklist #2169 must be complete. (0,5) c.
I REFERENCE MONTICELLO Normal Shutdown, C.3, pg C.3-0006 l
l l
l ANSWER 7.09 (2.00) 1
- b. With TWO APRM's in a ch. (#4 & #5) greater than 1.02, a shutdown must immediately begin and continue until the APRM's have been adjusted. (1.0)
REFERENCE i MONTICELLO Power Operations, C.2, pg C.2-0032 1
- 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 7.10 (2.00)
- a. The running loop [.53 to minimize thermal stress [.5]. (1.0)
- b. With approval of the superintendent, operations [.33), cooldown to as low as practical [.33] prior to placing shutdown cooling into service [.33] (1.0)
REFERENCE MONTICELLO RER, B.3.4, pg B.?.4-0069, 0070 ANSWER 7.11 (1.00)
To allow a natural circulation path between the inside and outside of the shroud. (1.0)
REFERENCE MONTICELLO Shutdown Procedures, C.3, pg C.3-0037 ANSWER 7.12 (1.00)
If HPCI and/or feedwater systems can maintain and recover the water level.
(1.0)
REFERENCE MONTICELLO C.4.III, Primary Containment Isolation, pg C.4-0024
~
PAGE 33
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 7.13 (2.50)
- a. 1. Demand OD-3 to determine which APRM'S caused the alarm
- 2. Increase the gain of the deficient APRM'S
- 3. Re-demand OD-3 to verify that all APRM'S are reading properly
- 4. Control Room Log entry stating which APRM gains were adjusted and why.
(1.0)
- b. 1. Approximately once a month
- 2. Immediately following a rod sequence exchange
- 3. Upon request of the Nuclear Engineer
- 4. Following a detector replacement (to calibrate the detector)
- 5. More than 3 base criticals exist and need valid process computer determination of MAPLHGR, MFLPD and/or APRM scram and rod block settings.
(3 required @ 0.5 each) (1.5)
REFERENCE MONTICELLO C.2, pg C.2-3,10-16,32,33 l
l l
l _
PAGE 34 8." ADMINISTRATIVE FROCEDURES. CONDITIONS. AND LIMITATIONS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 8.01 (1.00)
D REFERENCE Monticello ADMIN 4-ACD-3.9, pg 2 & 4-AWI-3.9.5, pg 19 ANSWER 8.02 (3.00)
- a. To enhance the capability of the LPCI Loop Selection Logic to detect some limited low probability breaks in the recire loop. (1.0)
- b. To prevent excessive jet pump vibration. (1.0)
- c. To preclude excessive thermal stresses on the reactor bottom head-to-support skirt transition and/or CED stub tubes. (Either component for full credit.)
Cold water reactivity accident for 0.25 (partial credit). (1.0)
REFERENCE MONTICELLO Ops, Manual, B.1.4 ANSWER 8.03 (2.00)
- a. Temporary changes to approved procedures may be made when the change CLEARLY does not change the INTENT of the procedure (Except for emergency situations). (1.0)
- b. Such changes shall be DOCUMENTED [.2] and shall have the WRITTEN [.2) concurrence of TWO[.2] individuals holding SRO LICENSES PRIOR [.2] to implimentation. ONE of the two shall be the ON-DUTY SHIFT SUPERVISOR [.2]. (1.0)
REFERENCE Monticello ADMIN 4-ACD-3.11, pg 9 m -- - - - ,-- , - - - - , - . , _
~
~
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 35 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 8.04 (2.00)
- c. The decision to notify Off-site Emergency Management Agencies.
- b. Any Protection Action Recommendations made to Off-site Emergency Management Agencies.
- c. Classification of the event.
- d. The decision to require on site evacuation.
- o. The decision to exceed normal radiation exposure limits.
- f. The Authorization to use potassium iodide.
(4 @ 0.5 ea.)
REFERENCE MONTICELLO EPIP's, A.2-102, pg 2.
ANSWER 8.05 (1.00)
The operator should be instructed to continue the EPIP's but do NOT violate / exceed a Tech. Spec. limit or requirement. ("None of the actions specified in the EPIP's shall take precedence over the actions that are necessary to comply with the Tech. Spec.") (1.0)
(If CLEARLY STATED that actions may be taken to protect the public in cccordance with 10 CFR 50.54(x), but exceeding violating / exceeding Tech.
Spec. the answer will be accepted.)
REFERENCE MONTICELLO EPIP's, A.2-101, pc 1 (precautions). .
ANSWER 8.06 (1.00) d (a. Not until off-site power is restored
- b. When convenient
- c. Not until off-site power is restored.)
REFERENCE MONTICELLO, ELEC. SYS. FAILURE, C.4.III, pg C. 4-0039,40,41, & 42.
r PAGE 36
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 8.07 (3.00)
- a. Torus water temp required during normal operation.
- b. Torus water temp during test OPS which adds heat to torus.
(greater than 90 F less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.)
- c. IMMEDIATELY scram reactor.
- d. Cooldown and depressurize to less than 200 psig using normal cooldown rates. (4 @ 0.75 ea)
REFERENCE MONTICELLO Tech. Spec; 3.7.A.1 ANSWER 6.08 (2.00)
The startup CANNOT [0.5] continue if LESS THAN 12 [0.5]
control rods have been withdrawn. (1.0)
If greater than 12 control rods have been withdrawn the startup CAN [0.5] continue with a SECOND independent operator used to REPLACE THE RWM [0.5]. (1.0)
REFERENCE MONTICELLO Tech. Spec; 3.3.B.3(b)
ANSWER 8.09 (2.00)
- b. [1.0)
On a group I isolation the scram signal should come from the MSIV closure NOT steam line pressure. The REACTOR PRESSURE HIGH SCRAM pressure signal is not the primary signal and per Monticello Tech.
Specs. a safety limit violation MAY have occured. (1.0)
REFERENCE MONTICELLO Tech. Specs., 2.1.C pg 2.1/2.3 & BASES 2.1.C pg 12
& BASES 2.4, pg 24
PAGE 37
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS ANSWERS -- MONTICELLO
-86/07/22-KING, M.
ANSWER 8.10 (1.00) 2 or more adjacent rods or more than thirty control rods greater than 06 position with a valid scram signal REFERENCE MONTICELLO Reactor Scre.m C.4.I, pg C.4-0008 ANSWER 8.11 (3.00)
- c. Use the reactor manual control system to insert rods. (1.0)
REFERENCE MONTICELLO Reactor Scram, C.4.I, pg C.4-0008 & 0009 ANSWER 8.12 ( .00)
Question / answer / reference deleted from exam.
REFERENCE Question / answer / reference deleted from exam.
ANSWER 8.13 (2.50)
Severe failure because:
- a. Many (1/4) indicating lights / alarms are inoperative.
- b. Remote control for many pumps / switch gear is inoperative.
- c. Many (depending on bus) auto-protective features are inoperative.
(3 @ 0.5 ea.)
(Other answers with justification will be accepted)
To comply with Tech. Specs. the plant must be in cold shutdown (1.0) condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 -s
- 8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 38 ANSWERS -- MONTICELLO -
-86/07/22-KING, M.
REFERENCE MONTICELLO Abnormal Conditions - Loss of 125 V DC bus A or bus B, C.4.III, pg C.4-0051 ANSWER 8.14 f1.00) b REFERENCE MONTICELLO EPIP's, A.3, pg A.3-0003 & 0006.
l l
-,,.-,--,-m,-,-,,,.7---=.-,.,-.m-ne-- , . - , . . . , - - - , - - .--.-,.,-n,-,-. - - - - . - - - - - - . - - - - , - - - - - -
c4r 48 U. S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION FACILITY: MONTICELLO REACTOR TYPE: BWR-GE3 DATE ADMINISTERED: 86/07/22 EXAMINER: KING. M.
CANDIDATE:
INSTRUCTIONS TO CANDIDATE:
Use separate paper for the answers. Write answers on one side only.
Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing Orade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY ,
27.75 26.75 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.50 24.58 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS 24.50 ._23.61 3. INSTRUMENTS AND CONTROLS 26.00 25.06 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL-103.75- Totals Final Grade i All work done on this examination is my own. I have neither given nor received aid.
l Candidate's Signature l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply: i l
- 1. Cheating on the examination means an automatic denial of your application and could result in more severe penalties,
- 2. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 5. Use black ink or dark pencil onlr to facilitate legible reproductions.
- 4. Print your name in the blank provided on the cover sheet of the examination.
- 5. Fill in the date on the cover sheet of the examination (if necessary).
- 6. Use only the paper provided for answers.
- 7. Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
- 8. Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only an one side of the paper, and write "Last Page" on the last 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.
- 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table. .
- 12. Use abbreviations only if they are commonly used in facility literature.
- 13. The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer required.
- 14. Show all calculations, methods, or assn +mtions used to obtain an answer to mathematical collems whether ind!aated in the question or not.
- 15. Partial credit may be given. Thar: 17 r. . ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANLEAR t.MK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner cnly.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been completed.
o r
- 18. When you complete your examination, you shall:
- a. Assemble your examination as follows:
(1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
- b. Turn in your copy of the examination and all pages used to answer the examination questions.
- c. Turn in all' scrap paper and the balance of the paper that you did not use for answering the questions,
- d. Leave the examination area, as defined by the examiner. If after leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
9 e
g r-a - + - -~
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PAGE 2
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
' THERMODYNAMICS . HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.00)
.Three (3) minutes following a reactor scram from high power, indicated reactor power is 75 on range 5 and decreasing.
- c. What will INDICATED power (level and scale) be one (1) minute later?
(Show calculations) (1.0)
- b. Explain why power decreased at this rate. (1.0)
QUESTION 1.02 (2.00)
- a. -How does feedwater heating improve the efficiency of the power plant? (1.0)
- b. If the highest pressure feed heater is removed from service (extraction steam isolated), what happens to Megawatt output of the generator and why? (1.0)
QUESTION 1.03 (3.00)
With the reactor operating at 75% power, recirculation flow control fails, rapidly increasing flow,
- a. Which reactivity coefficient WILL ACT FIRST TO LIMIT the power excursion? (1.0)
- b. Explain your choice of coefficient. Insure your explanation addresses all other coefficients. (2.0) t i
l
! (***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
1, PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3
' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.04 (1.50)
MATCH the Failure Mechanism from column (1) AND the Limiting Condition f rom column (2) WITH the associated Power Distribution Limits (a-c) below.
- a. Linear Heat Generation Rate (LHGR)
- b. Average Planer Linear Heat Generation Rate (APLBGR)
- c. Minimum Critical Power Ratio (MCPR) 1 - FAILURE MECHANISM 2 - LIMITING CONDITION
- 1. FUEL CLAD CRACKING DUE TO LACK 1. 1% PLASTIC STRAIN OF COOLING CAUSED BY OTB :
- 2. FUEL CLAD CRACKING DUE TO HIGH 2. PREVENT TRANSITION STRESS FROM PELLET EXPANSION BOILING
- 3. GROSS CLAD FAILURE DUE TO DECAY 3. LIMIT CLAD TEMP HEAT & STORED HEAT FOLLOWING TO 2200 F A LOCA (1.5)
QUESTION 1.05 (1.00)
Fill in the blanks with one of the given choices in the paragraph below describing the inverse power response to rod movements.
As a shallow rod is withdrawn during power operation, the power in the region above the blade tip [1]
(INCREASES, DECREASES) causing a(n) [2] (INCREASE, DECREASE) in void concentration in the upper region of the bundle. This results in a [3] (POSITIVE, NEGATIVE) reactivity addition due to the void coefficient of reactivity.
If the reactivity from the change in void concentration exceeds the reactivity added by the withdrawn rod, power will [4]_
(INCREASE, DECREASE) slightly.
(4 9 0.25 ea.)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
L
, 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 4
' THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.06 (2.50)
A design feature in the reactor vessel ensures proper flow distribution through the core fuel bundles,
- a. What is this feature ? (0.5)
- b. EXPLAIN what would happen on a power increase with "NO CHANGE IN MEASURED CORE FLOW" if this feature were eliminated. (2.0)
QUESTION 1.07 (1.25)
Consider a real plant system (NON-IDEAL) with two identical pumps in parallel, one of which is running. The second pump is started. System flow will be: (choose the correct answer)
EXPLAIN YOUR CHOIC . (NOTE BOTH PUMPS OPERATING @ 1800 rpm) (1.25)
- a. Double the original flow
- b. Less than double the original flow
- c. Greater than double the original flow
- d. The same, only the discharge head changes QUESTION 1.08 (1.00)
During high power operations (>60%), WHY is it more desirable to change power with recirculation flow than with control rods? (1.0) i .
QUESTION 1.09 (3.00)
I Give ONE undesirable result for each of the following.(Be more
! specific than " pump failure"):
- a. Operating a motor driven centrifugal pump for extended periods of time with the discharge valve shut. (1.0)
- b. Starting a motor driven centrifugal pump with the discharge valve full j open. (1.0)
- c. Operating a motor driven pump under " PUMP RUNOUT" conditions. (1.0) i
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
l
)
, 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 5 :
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW j l
QUESTION 1.10 (2.50)
Answer the following questions using figure 1.
- a. Describe the means (method) power would be increased from point A to point B. (0.75)
- b. Explain the curve from point B to point C. (0.5)
- c. Explain the curve from point D to point E. (0.5)
- d. When is operation allowed in the crosshatched area ? (0.75)
QUESTION 1.11 (3.00)
Assume a normal power increase of 10% is made with the recirculation pumps.
Plot and explain each of the following parameters from the beginning of the transient to the final power level (+10%).
- a. reactivity EXAMPLE: l FLOW ...........................
- b. void fraction l l
- c. reactor pressure ~ TEMP l...........................
l
- d. reactor power l PRES !...........................
- e. reactor perio'd l l
- f. turbine steam flow LEVEL !....................... ...
l l
l (6 @ 0.5 ea.)
TIME
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PAGE 6 PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
- THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.12 (3.00)
Indicate HOW each of the coefficients are effected [ Increase, Decrease or Remain the same] by each of the three parameters listed? Consider each parameter seperately.
- a. Rod Worth (delta K/K/ Bank) by:
- 1. Moderator temperature INCREASES
- 2. Voids DECREASE
- 3. Fuel temperature INCREASES [3 @ 0.33 ea]
- b. Alpha Doppler (delta K/K/ F fuel) by:
- 1. Core age INCREASES
- 2. Fuel temperature DECREASES
- 3. Voids DECREASE [3 @ 0.33 ea)
- c. Alpha Voids (delta K/K/ % voids) by:
- 1. Fuel temperature INCREASES
- 2. Core age INCREASES
- 3. Control Rod Density INCREASES [3 @ 0.33 ea) 4
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
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- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION 1 PAGE 7 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW QUESTION 1.13 (2.00)
Match each the following four definitions with the term it defines. (2.0)
DEFINITIONS:
- 1. The quantity utilized as a measure of how close the system fluid is to saturation conditions.
- 2. The pressure developed when a pump is filled with fluid to be pumped and operated at normal speed with its discharge valve shut.
- 3. When insufficient pressure at the inlet to a pump results in the static pressure being less than.the staturation pressure of the fluid, the liquid begins to boil, forming thousands of tiny vapor pockets, which collapse when in the region of higher pressure.
- 4. The resultant loss of back pressure on the pump causes the impeller to increase in speed. The pump motor amps increase and the high current can damage the motor windings.
TERMS:
- a. Cavitation b. Pump runout
- c. Head d. Shutoff head
- e. Water Hammer f. Recirculation Ratio
- g. Net Positive Suction Head h. Pipe Whip i
(***** END OF CATEGORY 01 *****)
(
- 2. , PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.01 (2.50)
List 5 of 6 conditions which must be satisfied to allow automatic closure of the the 4.16 KV source breaker (ACB 152-502) to supply Bus 15 from the # 11 diesel generator. (2.5)
QUESTION 2.02 (3.00)
During a reactor scram:
- a. WHY does the on-line flow control valve in the Control Rod Drive Hydraulic System go to its Minimum position and WHY is this DESIRABLE? (1.5)
- b. EXPLAIN how a control rod would respond if its HCU scram inlet valve sticks shut with the scram outlet valve open? Consider reactor pressure both (1) high at 1000 psig and (2) low at 300 psig in your answer. (1.5)
QUESTION 2.03 (3.50)
- a. How is the HPCI Turbine exhaust line protected against overpressure? (Identify two means, include setpoints). (1.5)
- b. Identify which of the following are direct HPCI turbine trips and which are HPCI system isolations: (2.0) l
- 1. HPCI steam line low pressure j 2. Reactor high water level t 3. HPCI steam line area high temperature
- 4. Pump suction low pressure l
QUESTION 2.04 (2.50)
- a. Two sets of vacuum breaker valves are provided on the primary
, containment. One set relieves from the 7 to the ? .
l The other set relieves from the 7 to the 7 . The set-l points for each set is 7 and 7 respectively. (2.0) l l
- b. Why are these vacuum breakers required? (0.5) i l (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
l
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 9 QUESTION 2.05 (2.00)
The Core Spray System receives a valid initiation signal. One pump fails to start:
- a. WHY must that core spray loop be isolated? (0.5)
- b. HOW is the isolation accomplished (be specific)? (1.0)
- c. What valve in the Core Spray system is used by the operator (0.5) to throttle injection flow.?
QUESTION 2.06 (3.50)
The Residual Heat Removal system can take a suction from four (4) sources and discharge to nine (9) areas.
- a. WHAT are three (3) of the suction sources? (1.5)
- b. Excluding the #11 and #12 recirc loop and with the crosstie open, WHAT are four (4) areas that each loop is capable of suppling water? (2.0)
QUESTION 2.07 (1.00)
During two pump operation the condensate pumps require a minimum flow of 5000 rpm. The recirculation setpoint is 6000 gpm. Why is the additional flow (minimum) required ? (1.0)
^
QUESTION 2.08 (2.25)
Following an auto initiation of RCIC at a reactor pressure of 800 psig, reactor pressure decreases to 400 psig. HOW are each the following parameters affected (INCREASES, DECREASES, REMAINS CONSTANT) by the l change in reactor pressure? BRIEFLY EXPLAIN EACH CHOICE.
ASSUME the RCIC System is operating as designed.
j a. RCIC flow to the reactor (0.75)
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
t _ _ _ ___
,2. _ PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.09 (1.25)
Listed below are a number of components of the Reactor Water Cleanup System. Place the components in order of normal FLOW PATH beginning with the components downstream of the inlet isolation valves. (One component MUST be used twice for full credit.)
- a. Filter demineralizer
- b. Pumps
- c. Regenative heat exchanger
- d. Non-regenative heat exchanger QUESTION 2.10 (1.00)
The safety / relief bellows leaking alarm is received due to a bellows failure on RV 2-71A. Explain the effects of this failure on ALL modes of operation.
QUESTION 2.11 (2.00)
Identify the signals required and describe the logic for a low-low set SRV to automatically open. (2.0)
QUESTION 2.12 (1.00)
The plant is operating at 80% power when the OFF-GAS RECOMBINER has a malfunction. It is decided, with proper approval, to use the 30 minute delay OFF-GAS system.
Which of the following is correct ?
l a. The change over to the 30 minute system may cause a decrease l
in vacuum and power must be reduced to ~60% before the ,
i change over.
- b. The compressed gas storage must be operable and in use to limit the release during use of the 30 minute system.
! c. Heath Physics Dept. must verify release rates are less than Tech. Spec. limits prior to spectacle flange operation.
l
- d. The change over to the 30 minute system will require a plant shutdown and changes in the radiation monitoring system.
I
(***** END OF CATEGORY 02 *****)
i
PAGE 11
, 3. INSTRUMENTS AND CONTROLS QUESTION 3.01 (3.50)
- a. List the signal (s) and the setpoint(s) that will initiate the LPCI. (0.9)
- b. Describe the response of the LPCI Loop Select logic system if only one recire pump is in operation at the time of logic actuation. (2.0)
- c. List 3 modes of operation, other than LPCI and RHR (shutdown cooling) of the RHR system. (0.6)
QUESTION 3.02 (3.00)
Considering the Main Steam Line Radiation Monitoring System:
- a. How many, what kind, and where are the detectors located? (1.5)
- b. iThat three (3) automatic actions (excluding GP 1 isolation and all alarms) take place when the system logic is tripped? (1.5)
QUESTION 3.03 (2.00)
Explain how RCIC turbine speed is controlled following an automatic initiation signal. Begin with the steam admission valve shut and continue until the system is injecting at rated flow. Include what signal is controlling speed initially and at rated flow. (2.0)
QUESTION 3.04 (3.00)
- a. What are two (2) plant systems or components that receive rod position information from the RPIS, OTHER THAN the full core and 4 rod group displays? (1.0)
- b. Fhat action occurs automatically upon receipt of an RFIS INOP7 (1.0)
- c. With a selected rod at notch position 18, and its 02 notch position reed switch stuck shut, what will the 4 rod group display indicate for the selected rod's position? (0,5)
- d. What indications will result from a control rod moving to the overtravel position during a coupling check ? (0.5)
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
'f P
- 3. INSTRUMENTS AND CONTROLS PAGE 12 QUESTION 3.05 (2.00)
The reactor is operating at power. What will be the effect on/of the recirculation flow control system due to the following conditions:
- c. The reactor is at 30% power when an operator inadvertently shifts the "A" recirc pump M/A transfer station to AUTO. (1.0)
- b. The reactor is at 95% power with recirculation flow control in master manual when full open indication on recire pump "B" discharge valve (MO 2-53B #12) is lost. NOTE: MO-2-54B the recire pump #12 bypass discharge is open (1.0)
QUESTION 3.06 (1.00)
On the SRM detector drive control a white light indicates a SRM detector may be withdrawn. List the condition when the light will be on. (1.0)
QUESTION 3.07 ( .00)
Question / answer / reference deleted from examination.
QUESTION 3.08 (1.00)
Which one of the following actions requires the operator to hold the control switch until the valve operation is complete ?
- e. Fully opening a recire pump discharge valve.
- b. Fully closing a recire pump discharge valve,
- c. Fully opening a recire pump suction valve.
- d. Fully closing a recire pump suction valve
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
INSTRUMENTS AND CONTROLS PAGE 13
.3.
QUESTION 3.09 (2.00)
Main steam pressure is being controlled by a normal lineup with the EPR in control at 65% power. Recirc control is in master manual.
Explain the actions of the pressure control system for the following:
- o. System (Grid) frequency decreases. (1.0)
- b. One turbine bypass valve fails open. (1.0)
QUESTION 3.10 (2.00)
List four interlocks required to be met before a reactor feed pump will start. (4 @ 0.5 ea.)
QUESTION 3.11 (1.00)
Which of the following are TRUE concerning the Standby Liquid Centrol System: (CHOOSE ONE.) s
- a. In the event a remote (outside control room) reactor shutdown is required, SBLC injection can be actuated by the local pump START switch.
- b. The pumps may be operated simultaneously if necessary to shutdown the reactor in an ATWS.
- c. If injected, the SLC system will previde at least 3% SDM and 900 ppm boron in the reactor vessel.
- d. Nitrogen-charged accumulators assure adequate suction pressure for the pumps.
l l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l I
- 3. INSTRUMENTS AND CONTROLS PAGE 14 QUESTION 3.12 (1.50)
State what specific RFS action (if any) will occur directly from MSIV
. position if the following main steam lines shut in the run mode:
- a. Lines B, C, and D
- b. Lines B and C
- c. Lines A and B QUESTION 3.13 (2.50)
What are FIVE automatic actions which occur at the Low-Low Water Level Trip 7 (5 9 0.5 ea.)
I i
l (***** END OF CATEGORY 03 *****)
l
- 4. _ PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 15 RADIOLOGICAL CONTROL QUESTION 4.01 (1.00)
During performance of a procedure a change is required due to plant conditions. Under what restrictions may a temporary change.be made ?
QUESTION 4.02 (3.00)
At what pressure will each of the following be normally performed or occur during a reactor startup from cold conditions per start up procedure C.1, Heatup and Pressurization 7
- a. The mechanical pressure regulator is allowed to open the Main Steam Bypass valve #1 to verify regulator operation.
- b. The RCIC Automatic isolation signal is reset,
- c. The HPCI Automatic isolation signal is reset.
- d. The mechanical pressure regulator override is adjusted to open the #1 Main Steam Bypass valve 10 - 15%.
- e. Electric pressure regulator is verified to assume pressure control.
- f. The Air Ejector Suction Isolation Valve Control Switch is placed in the Auto postion.
[6 @ 0.5 ea) (3.0)
QUESTION 4.03 (3.50)
In accordance with the approach to criticality steps in the cold 4
startup procedure, C.1, answer the following,
- a. What are the RO's required actions if criticality does not occur within the predicted critical rod pattern band indicated on Predicted Critical for Plant Startup form #21597 (0.75) 4
- b. When is the reactor considered critical? (0.75)
- c. What four (4) items are recorded, in the reactor log and on the predicted critical form, when criticality is established? (1.0)
- d. What are three (3) ways, that reactor period may be determined? (1.0)
(Not read off period meter.)
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
4,__ PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 16
~
RADIOLOGICAL CONTROL QUESTION 4.04 (2.00)
What are 4 responsibilities of the control room operator as outlined in the EPIP's procedure A.2-101, Classification of Emergencies ?
QUESTION 4.05 (1.00)
If the core spray system is initiated with station auxiliary power from transformer 1R, what loads will be lost (2 required) 7 (2 @ 0.5 ea.)
QUESTION 4.06 (2.50)
A loss of off-site power has occurred. How are reactor vessel level and pressure controlled per procedure C.4.III, Electrical Systems Failures ?
QUESTION 4.07 (1.00)
' What is the procedural definition of an ATWS7 QUESTION 4.08 (3.00)
An ATWS has occurred. All scram valves have been verified open. List 3 methods IN THE PREFERRED order to insert the control rods.
QUESTION 4.09 (1.00)
A valid low-low level has been received. Under what conditions would the 106 second auto-blowdown timer be blocked 7 QUESTION 4.10 (2.50)
A steam line break has occurred and the MSIV's have closed and isolated the break. Explain the flow path (s) (in and out) to be used for cooldown per procedure C.4.II Primary Containment isolation, l
l
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
(
- 4. PAGE 17 EBOCEDURES - NORMAL. ABNORMAL. EMERGENCY AND RADIOLOGICAL CONTROL QUESTION 4.11 (1.00)
Neither CRD pump will operate. What conditions require the reactor to be scrammed 7 QUESTION 4.12 (1.50)
A flow indicator is not provided for standby liquid control (SLC).
If actuated, per procedure B.3.5, how do you confirm flow (three required)?
(1.5)
QUESTION 4.13 (1.00)
What is the maximum reactor power level the rotary vacuum pump may be operatied at per procedure B.6.3.c, Main Condenser Startup 7 (1.0)
QUESTION 4.14 (2.30)
List the 8 actions to be taken (time permitting) on a slow loss of main condenser vacuum. (2.0) l
(***** END OF CATEGORY 04 *****)
(************* END GF EXAMINATION ***************)
........t.....
. . . . . , . . . . . .. ,~. ,. .. ., ,
m
.........,...s
........ . .. . ... .. . . .s /
\ T rs, e
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, JLe:c.555 i i re AcW -
! ". k@\EV , -
1~ A kNV #
1 3 a
\
/g
- f Mf , J- -
ps-Ic
- * /
sm
, r / - - ,
}rs /A r
~
f .
/
.'f ) .. . .. . . .. . ...
D ...........
FIGURE ONE
i
. . j EQUATICN SHEET }
l C/cle afficiency = (1et M l f = ena v = 1/t cut)/(Energy in) 2
,=y s = V ,t + 1/2 ac E=* A
- 13 A*A'o 4E = 1/2 av 4 * (Vf - Vo )/t PE = agn
- = e/t i = sn2/tifg = 0.693/t1/2 yf = V, , at t aff = C(tti,)(~~)]
- 2 1/2
'd = v .P rD A= 7 ((cl/2I
- IbI
- ' # -m a = V,yAo t , !ot Q = mCaat ~
I = I,e "#
a = UAa.7 pe=ggf ! = I,10~* ##'
TYL = 1.3/u HVI. = -0.593/u P = P 10 sur(t)
P = P,e"7eo ,, ' ,
SG = S/(1 - 4,ff)
SUR = 25.06/T G, = S/(1 - <,ff x)
G j(1 - X,ff;) = G 25I ~ *eff2I SUR = 29s/t* + (s - o)T M = 1/(1 - K,ff) = CR;/G 3 T = ( t"/s ) + ((a - a l/ Io]
M = (1 - X ,ff,)/(1 - X ,ffj) 7 = a/(s - a)
SCM = ( - X,ff)/X,ff T = (a - o)/(Ta) sec:nes t' = 10 a = (X,ff-1)/X,ff = t.X,fgX,ff *I I = 0.1 sec:nds e = ((t*/(T X,ff)] * (I,ff /(1 + IT)]
Id lt"I#2,22 g#
Idj 22 P = (:4V)/(3 x 1010) 2 2/nr = (0.5 CZ)/c (meters)
- = :N R/hr = 6 CZ/d2 (f,,g) ,
Miscellaneous 0:nve'-siens Watar Darwetem 1 curie = 3.7 x 1010:33 1 gal. = 8.345 lem.
1 kg == 2.54 2.21xItm 10 ] 3tu/nr Iga}.=3.7811
= 7.48 gal. tars 1 no 1 f. 1 = = 3.41 x 100 5tu/hr Oensity = 62.4 les/ft3 lin = 2.54 c:n Censity = 1 gs/c9 *F = 9/5'C
- 32 Heat of vaccritation = 970 Stu/lem 'C = 5/9 (*F-32)
Heat of fusion = 144 Stu/lem . 1 87U = 778 ft-lbf 1 At:n = 14.7 osi = 29.9 in. Hg.
i 1 ft. H 2O = 0.4335 Itf/in.
- 1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION. PAGE 18 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 1.01 (2.00)
- a. Using P = Po e to the t/.T then P = 75 e to 60/-80 P = 75 e to -0.75 = 35 on Range 5 (1.0)
- b. After the initial prompt drop, power cannot decrease faster than the longest lived delayed neutron appears, which has about a (55.6) see half life. (1.0)
REFERENCE Monticello M 8102-L-007 rev 2, pg 18,19,39 & B.5.1.1 S/U range monitors, Jan 86, pg 24,25 ANSWER 1.02 (2.00)
- 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 conditions. (1.0)
- b. Megawatt output from the generator would increase (0.5). Steam that was formerly being extracted now passes through the turbine to the condenser (0.5). (1.0)
REFERENCE Monticello M 8104-1-017, pg 14,15 & GE HT and FF Ch 5, Thermo Cycles and Cycle Analysis, pg 57 ANSWER 1.03 (3.00)
- a. The fuel temperature coefficient. (1.0)
- b. Due to the time constant of the fuel (generally 5 - 7 seconds) or the amount of time required f or heat to be transfered from the fuel to the coolant.[0.5] The moderator temperator or the void coefficient would not have any effect for several seconds.[1.0) The fuel temp.
coefficient' acts instantly to insert negative reactivity.[0.5] (2.0)
REFERENCE Monticello M 8102-1-016, BWR Reactivity Coefficients, Attachment 1, pg 17, sec. D.C.29.
/
.PRINCIELES OF NUCLEAR FOWER PLANT OPERATION, PAGE 19 1
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 1.04 (1.50)
Failure Mechanism Limited Condition (0.2 ea.) (0.3 ea.)
- a. LHGR 2 1
- b. APLHGR 3 3
- c. MCPR 1 2 (1.5)
REFERENCE Monticello M 8104-1-012, BWR Thermo & Thermo Limits, pg 7,8 ANSWER 1.05 (1.00)
- 1. INCREASE [0.25]
- 2. INCREASE [0.25]
- 3. NEGATIVE [0.25]
- 4. DECREASE [0.25]
REFERENCE Monticello M 8102-1-016, BWR Reactivity Coefficients, pg 48.
ANSWER 1.06 (2.50)
- a. Will accept core orificing OR orificed fuel support pieces. (0.5)
- b. As power increases the amount of boiling (two-phase flow) increases.
[0.5] The amount of power generated in a peripheral bundle is <
(approximately half) that of a center bundle; therefore boiling is greatest in the core center.[0.5] Two-phase flow restricts water flow due to the boiling action.[0.5) This would cause the higher powered bundles to receive less water as their higher resistance to flow, would divert flow to lower power fuel bundles (0.5] starving the higher power bundles. (2.0)
REFERENCE Monticello System Manual B.1.1, pg B.1.1-0045.0 & GE HT & FF, ch 8 pg 8-45.
4
.-,._ny .___m. , . _ - _ _ . _ _ _ _ _ . . , _ _ _ , , , _ . , . . . , _ . _ _ . _ _ _ _ _ _ _ _ , . . _ . _ , _ . _ _ _ _ _ _ _ , , . .
. 1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION & PAGE 23 THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ;
l ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 1.07 (1.25) b [.5]
EXPLANATION:
(Less than double the original flow). When delivering water into a piping system that offers frictional resistance, 2 pumps operating in parallel will encounter greater resistance to flow.
The resistance lowers the total flow to less than twice the original flow.[0.75) (1.25)
REFERENCE Monticello M 8104-1-015, pg 3,6,16 ANSWER 1.08 (1.00)
Changing power with recirculation flow changes total core power while keeping the flux profile relatively unchanged OR to avoid locali=ed flux peaking. (1.0)
REFERENCE REACTOR PHYSICS TRANSIENT ANALYSIS LP, pg 14 ANSWER 1.09 (3.00)
- a. The pump will eventually add a sufficient amount of heat to the fluid.to cause cavitation. Also will accept overheating of the pump. (1.0)
- b. Could cause excessively long starting currents or water hammer if the downstream piping was not filled. (1,0)
- c. Causes excessive motor amps to be drawn and the high current could cause damage to the motor windings OR may cause cavitation. (1.0)
REFERENCE Monticello System Manual B.3.4, Figure 8. & GE HT & FF ch 6, pg 108,109.
PAGE 21
. 1. . PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW ANSWERS -- MONTICELLO 86/07/22-KING, M.
ANSWER 1.10 (2.50)
- a. Point.A to Point B is the 100% rod line. Power would be increased by recire flow control (increase). (0.75)
- b. Power decreases (causes other than recirc), void content decreases resulting in a increase in core flow due to the lowered flow resistance.
(0.5)
- c. Point D to Point E is the two pump min. flow line. With two recire pumps a min. speed CORE flow will follow this line as power increases. (0,5)
- d. Operation allowed only while building equilibrium Xenon. (0.75)
(Provided flow map was not plant specific. Accept " Operation not allowed".)
REFERENCE GE RT, Ch 7, pg 7-29
-,--y,n - . - - - , , - - - - - - - - - . - - . - - - - - , . ---,_.,,y,,, . - - , + .-g- , ,--.y* .,,.
- 1. PAGE 22 PRINCIELES OF NUCLEAR POWER PLANT OPERATION2 THERMODYNAMICS. HEAT TRANSFER AND_ FLUID FLOW ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 1.11 (3.00)
!!! The following text describes the power increase transient. !!!
!!! The Plots are from GE RT, Ch 7, pg 7-30 and will be attached !!!
!!! to the answer KEY, !!!
The operator begins increasing flow at pt. one. This causes a DECREASE in void fraction and adds positive reactivity.
Power level begins to rise immediately, increasing fuel element temp.
This implies more heat transfer to the coolant, thus increasing steam generation,
- a. reactivity : increases due to void decrease then returna to zero as fuel temp inc. and voids return adding neg. react.
- b. void fraction : decreases due to recire flow increase then increases as power increases (stm generation inc.)
- c. reactor pressure : reactor press. will increase due to the design of EHC & turbine control system.
- d. reactor power : power will increase due to + react. from recire flow increase (or void decrease). Power will stabilize
(~ 10% higher) as fuel temp and voids add neg. react.
- e. reactor period : period will become + (from infinite), peak, then decrease going slightly neg. as period returns to inf.
- f. reactor steam flow: steam flow will increase due to EHC/ pressure control system which will open turbine throttle valves to maintain set pressure.
(6 @ 0.5 ea.)
REFERENCE GE HT & FF, Ch 7, pg 7-29,30
i L
Page 22A i
top Plow ReCReAsa RECIRCULATION ft s
,/
w TART e Ptow weCRsAsa Tada legt REACTIVITY (AK/K) +
TIME VOID FRACTION y _
vus REACTOR PRESSURE e F.
se nus 4
REACTOR 3 power LEVEL ]
TIME 3
REACTOR PEstlOO
- 4 g
_ Tius 8
REACTOR STEAu Flow A TRIE The operator begins increasing recirculation flow at point 1. This causes a decrease in void fraction since Increased coolant flow past fuel elements tends to sweep away volds more rapidly than they are formed. This causes a positive reactivity addition by the vold coefficient of reactivity.
\
Power level begins to rise immediately, increasing fuel element temperatures. This impiles more heat transfer to the coolant, thus increasing steam generation.
, , . _ _ . _ , . . , . _ , , , _ _ . _ ._..______,~_,,.._.___+______m_._y,__-.__.-_.-._-,_m _. - _ . _ _ , - _ _ . , , , - , _ .
4
. 1 Page 228 At point 3, the increase in power level begins to take effect, with void fraction beginning to increase. Once again, increased steam generation is controlled by a pressure regulator, which sends additional steam generation through the turbine, increasing generator power output.
Reactor power level continues to increase until point 4 is reached.
The negative reactivity insertion rate, due to increasing void fraction and Doppler coefficient, will overcome the positive reactivity still present in the core.
The power level steadles out at point 5, with net reactivity once again equal to zero, and void fraction back to its original value.
+
9 m
-4 0
e, . - , . w- - - _ --., ,--n ., - - - - - - - , , - - - - - .-. .. ,-_-- - - - - - - - , - - - - - - - - . - - - - - . - - - - -
PAGE 23
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION.
THERMODYNAMICS. HEAT TRANSFER AND FLUID FLOW
-86/07/22-KING, M.
. ANSWERS -- MONTICELLO ANSWER 1.12 (3.00) a.1. increase a.2. increase a.3. remains the same ,
b.1. increase b.2. increase b.3. decrease
/
c.1. increase c.2. decrease c.3. increase- (9 @ 0.33 ea]
REFERENCE Monticello, Reactor Theory L.P., # M8102L-043 Rev.0; Figure 43 pg 43 of 43 ANSWER 1.13 (2.00)
- a. - 3.
- b. - 4.
- d. - 2.
- g. - 1. (4 @ 0.5 ea.)
REFERENCE Monticello Fluid Flow, M-8104-1-015, pg 2,3,4,5,6, & 7 O
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 24 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 2.01 (2.50)
For auto closure of ACB 152-502 must have:
- 1. Diesel 9 voltage (or within 10% of rated voltage)
- 2. #1AR transformer de-energized
- 3. All source breakers to the bus open (ACB-152-511,501, & 308)
- 4. Bus and breaker lockout relays reset (185-5, & 502)
- 5. Breaker control switch on C08 in automatic
- 6. Bus transfer lockout switch in set up NOTE: Will accept any of the following in place of #2 above
- 1. Loss of voltage on lowside of 1AR transformer OR
- 2. Loss of voltage on Essential Bus for > 10 seconds OR
- 3. Degraded voltage on Essential Bus for > 10 seconds (5 of 6 required at 0.5 each)
REFERENCE Monticello SEC. B.9.8. - 20 & B.9.6-0008,00010 ANSWER 2.02 (3.00)
- a. The flow controller sees a high flow from the flow element which is sensing charging flow to the accumulators [.75].
This directs most of the pump discharge to recharge the scram accumulators faster [.75] (1.5)
- b. 1. At high reactor pressure the rod will still scram but at slower rate than normal [.75]
- 2. At low reactor pressure the control rod would not scram [.75) (1.5)
REFERENCE Monticello CRD HYDRAULIC SYSTEM, B.1.3,pg 12 & 34 ANSWER 2.03 (3.50)
- a. 1. Rupture dics on the exhaust line(.6) at 175 psig(.15)
I 2. Turbine trip (.6) at 150 psig(.15) (1.5)
- b. 1. Isolation
- 2. Turbine trip
- 3. Isolation (4 @ 0.5 ea.)
- 4. Turbine trip I
r
. 2 _, PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25 ANSWERS -- MONTICELLO -86/07/22-KING, M.
REFERENCE Monticello HPCI SYSTEM, B.3.2, pg B.3.2-0013, 0003 & 0004 ANSWER 2.04 (2.50)
- a. Torus (suppression chamber) to the Drywell (0.8)
Reactor Building Atmosphere to the Torus (suppression chamber) (0.8) 0.5 psid and 10" water. Accept answers < 1 psid (0.4)
- b. The primary containment is not designed for a negative pressure differential. (.5)
REFERENCE Monticello Primary Containment, B.4.1, pg B.4.1-30,31 ANSWER 2.05 (2.00)
- a. To maintain primary containment integrity (0.5)
- b. Close the inboard isolation valve (.33), position the outboard isolation valve bypass switch to bypass (.33), then close the outboard isolation valve (.33) (1.0)
- c. Inboard isolation valve. (0.5)
REFERENCE Monticello CORE SPRAY, B.3.1, pg B.3.1-4a, 6, & 19
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 26 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 2.06 (3.50)
- o. 1. Torus Ring Header
- 2. Condensate Storage Tank
- 3. #11 Reactor Recirculation Loop
- 4. Fuel Pool Skimmer Surge Tank (Temporary spool piece required).
[3 @ 0.5 ea] (1.5)
- b. 1. Reactor Vessel Head
- 2. Upper Drywell Spray Header
- 3. Lower Drywell Spray Header
- 4. Torus Spray Header
- 5. Torus
- 6. Radwaste Surge Tank
- 7. Fuel Pool Spargers (Temporary spool piece required)
[4 9 0.5 ea] (2.0)
REFERENCE Monticello, System Description, B.3.4, Residual Heat Removal, pg 2 & 3 ANSWER 2.07 (1.00)
The additional cooling (flow) is to provide cooling (.33) for SJAE(.33) and steam packing condenser (.33) (1.0)
REFERENCE Monticello System manual B.6.5, Condensate and Reactor Feedwater, pg B.6.5-0005.0 & 0006.0
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 2.08 (2.25)
- a. Remains constant [.25]. Flow is controlled by the RCIC flow controller which will attempt to maintain a constant output flow regardless of reactor pressure [.50). (0.75)
- b. Decreases [.25]. The flow controller functions to maintain a constant flow, thus pump discharge pressure is decreased along with the decreasing reactor pressure to maintain constant flow.
OR Since the flow controller maintains a constant flow to the reactor, as reactor pressure decreases, the pump discharge head must decrease to maintain a constant flow (constant NPSH) [.50]. (0.75)
- c. Decreases [.25]. Since pump discharge head is decreasing to maintain a constant flow, turbine RPM must also decrease [.50]. (0.75)
REFERENCE Monticello Pumps and Fluid Flow, and System Manual, B.2.3 ANSWER 2.09 (1.25)
C,D,B,A,C (5 @ 0.25 ea)
REFERENCE MONTICELLO RWCU, B.2.2, Fig. at end of chapter.
ANSWER 2.10 (1.00)
Manual operation [.33] and ADPS operation [.33] will function.
Only the safety valve function is inoperable [.33] (1.0)
REFERENCE MONTICELLO ADPS, B.3.3, pg B.3.3-0003
_ = _ _ _ _ _. - _ _ ~ - _ _ - . _ . _ _ _ _ ~ ._ _ _ _. _ . _ _ . _ _ _
A
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 2.11 (2.00)
SIGNALS: 1. Scram
- 2. Rx pressure (1040, 1050, & 1060)
- 3. SRV control snitch in AUTO
- 4. < 50 psid (tailpipe to drywell) for 10 sec. (4 6 0.25 ea.)
LOGIC: "Two out of two, once"[.33]. This logic require TWO scram signals and TWO pressure signals in ONE channel " tripped" to activate the LLS-SRV's[.33]. There are two logic channels each channel operates the same three valves [.33]. (1.0)
REFERENCE MONTICELLO APRS. B.3.3, pg B.3.3-0013 ANSWER 2.12 (1.00) d (1.0)
REFERENCE MONTICELLO Gasous Radwaste, B.7.2, pg B.7.2-0201
PAGE 29
- 3. INSTRUMENTS AND CONTROLS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 3.01 (3.50)
(0.20 for each signal & 0.10 for each setpoint) (0.9)
- b. The logic system sends a trip signal to trip the operating pump [0.5]
The Logic then waits for reactor pressure to decrease to 900 psig to allow time for recire pump coastdown.[0.5] Logic then compares the jet pump riser D/P's to determine the broken loop.[0.5] The logic selects the loop with the greatest D/P or the # 12 loop if both are equal.[0.5]
(2.0)
- c. Containment spray Suppression pool cooling Fuel pool cooling Suppression pool to radwaste drain (3 req. @ 0.2 ea.)
REFERENCE Monticello RER, B.3.4, pg. B.3.4-0004,5,6,7, & 17.
ANSWER 3.02 (3.00)
- a. Gross gamma radiation is detected by four (4) [0.5] gamma sensitive ionization chambers [0.5] located adjacent to and immediately downstream of the outer MSIV's at DW penetration in the steam chase.[0.5]
- b. Logic trip will initiate a reactor scram [0.5] and if operating trip the main condenser vacuum pump [0.5] and suction valve closure.[0.5]
REFERENCE Monticello B.S.11, pg B.S.11-5,6,&7
PAGE 30 o 3. INSTRUMENTS AND CONTROLS ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 3.03 (2.00)
When the steam admission valve starts to open(.25), a ramp generator is triggered (.25). A low signal selector selects whichever signal is lower (.25), either the flow signal or the ramp generator (.25).
This signal is compared to actual turbine speed (.25) and a signal is generated to be sent to the control valve actuator to adjust turbine speed accordingly(.25). The ramp generator signal will be the controlling signal (.25) until the flow controller. output is below it. At rated flow, the flow signal will be the controlling signal (.25). (2.0)
REFERENCE Monticello RCIC System, 3.2.3, pg B.2.3-0010 ANSWER 3.04 (3.00)
- a. RWM and process computer (1.0)
(1.0;
- b. Rod select block
- c. Display will indicate 02 and 18 (0.5)
- d. Blank full core and 4 rod display (for overtraveled rod)
Rod overtravel Alarm (2 @ 0.25 ea.)
REFERENCE Monticello RPIS, B.S.4, pg B.S.4-1,8, & 11 ANSWER 3.05 (2.00)
- a. The control system will see a step increase to 45% demand signal from the dual limiter and will increase pump speed. The rate limiter will control the speed increase @ 2%/sec. (Assuming no scram occurs).
(1.0)
- b. Loss of open ind. for the discharge valve removes the bypass signal around the speed limiter and it outputs a 24% signal. The "B" pump will slow to demanded speed as an interlock prevents the mismatch summer from generating a " lockup". (1.0)
REFERENCE Monticello Recire Flow Control, B.S.8 pg B.5.8-2,3,4, & fig i Recire System, B.l.4, pg B.l.4-30,31
, 3. INSTRUMENTS AND CONTROLS PAGE 31 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 3.06 (1.00)
- 1. SRM >100 cys. (1.0)
REFERENCE MONTICELLO B.S.1.1-0007)
ANSWER 3.07 ( .00)
Question / answer / reference deleted from examination.
REFERENCE Question / answer / reference deleted from examination.
ANSWER 3.08 (1.00) a.
REFERENCE Monticello Recire. Sys, B.1.4, pg B.1.4-14,15 ANSWER 3.09 (2.00)
- a. The system cannot respond to a decrease in frequency. NO ACTIONS from pressure control system. (1.0)
- b. Reactor pressure would decrease due to the pwr/stm flow mismatch.
This would decrease the error signal from the EPR summer. The reduced signal would close the turbine control valves, reducing steam flow through the turbine. Reactor pressure would return to the original value with pwr/stm flow (total) equal. (1.0)
REFERENCE Monticello B.5.9-7,8,9, & fig 2
- 3. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 3.10 (2.00)
- 1. suction pressure (>85 psig)
- 2. lube oil pressure (>5 psig)
- 3. recire valve open
- 4. IW-98-1,2 open
- 5. reactor water level (<48 inches) (4 @ 0.5 ea.)
REFERENCE Monticello System Manual B.6.5, pg B.6.5-0009.0 ANSWER 3.11 (1.00)
C (1.0)
ANSWER 3.12 (1.50)
- a. Full scram
- b. None
- c. Half scram (3 @ 0.5 each)
ANSWER 3.13 (2.50)
- 1. Initiate HPCI
- 2. Initiate RCIC
- 3. Recirculation Pumps Trip
- 4. PCIS Group 1 Isolation
- 5. Diesel start
- 7. ATWS (with 9 sec delay)
(5 @ 0.5 ea.)
REFERENCE MONTICELLO Primary Containment Isolation, C.4, pg C.4-0023, & 0024.
& B.5.6-0048
-~
- 4. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCv AND PAGE 33 RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 4.01 (1.00)
Temporary changes to approved procedures may be made when the change CLEARLY does not change the INTENT of the procedure (Except for emergency situations). (1.0)
- OR**
Temporary change reviewed and initiaied by the shift supervisor and one other SRO.
The change does not conflict with the T.S. or interfere with the safe operation of the plant. (1.0)
REFERENCE Monticello ADMIN 4 ACD-3.11, pg 9 & C.1-0005 ANSWER 4.02 (3.00)
- a. 150 psig
- b. 80 psig
- c. 130 psig
- d. 500 psig
- e. 900 psig
- f. 200 psig [All pressures + or - 10%) [6 @ 0.5 ea] (3.0)
REFERENCE Monticello, Startup Procedure, C.1, Heating and Pressurization, pg 34, 35, 36, 40 & 41 i
1 I
, 4. PROCEDURES - NORMAL. ABNORMAL, EMERGENCY AND FAGE 34 RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 4.03 (3.50)
- c. 1. Discontinue rod withdraval
- 2. Maintain the reactor subcritical
- 3. Notify the Shift Supervisor (3 9 0.25 ea)
- b. Neutron Flux rises [0.25] with a constant (stable) period [0.25]
without additional control rod withdrawal [0.25]. (0.75)
- c. 1. the time
- 2. rod position
- 3. period
- 4. reactor water temperature
- 5. srm reading (4 @ 0.25 ea.)
- d. o Decade rise divided by 2.3 or multiplied by .435 o Doubling time divided by .693 or multiplied by 1.445 o Time for IRM scale reading to increase by a factor of 2.718 (3 @ 0.33 ea.)
REFERENCE Monticello, Startup Procedures, C.1, Cold Startup, Approach to Critical, ,
pg 22 !
i ANSWER 4.04 (2.00) [
- n. The control room operator shall immediately notify the Site .
Superintendent of any events that may be classified as emergency conditions. (0.5)
I
- b. The operator shall attempt to verify any indications. (0.5) i
- c. The operator shall assist the Site Superintendent in
- assessing the indications and determining the class-ification of emergency. (0.5)
- d. The operator shall take immediate actions as dictated by '
l Plant procedures and his general knowledge to control the event and place the plant in a safe condition. (0.5)
(The candidate's answers need not be verbatim to the above)
REFERENCE l
MONTICELLO EPIP's, A.2-101, pg 4.
l l
PAGE 35
.4. . PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND '
RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 4.05 (1.00)
- o. cooling tower fans
- b. cooling tower pumps
- c. drywell cooling recire f ans (2 @ 0.5 ea.)
REFERENCE MONTICELLO B.9.6, pg B.9.6-0013.
ANSWER 4.06 (2.50)
- a. MANUALLY initiate RCIC. (0.5)
- c. Control reactor pressure between 800-1000 psig by MANUAL INITIATION of relief valves. (1.0)
REFEPINCE MONTICELLO ELEC. SYS. FAILURES, C.4.III, pg C.4-0040 & 0041 ANSWER 4.07 (1.00) 2 or more adjacent rods or more than thirty control rods greater than 06 position with a scram signal.
REFERENCE MONTICELLO Reactor Scram, C.4.I, pg C.4-0008
l 1
- 4. PROCEDURES - NORMAL, ABNORMAL. EMERGENCY AND PAGE 36 EADIOLOGICAL CONTROL
-86/07/22-KING, M. l ANSWERS -- MONTICELLO ANSWER 4.08 (3.00)
- c. Use the reactor manual control system to insert rods. (0.75)
Correct order (as above) (0.75)
REFERENCE MONTICELLO Reactor Scram, C.4.I, pg C.4-0008 & 0009 ANSWER 4.09 (1.00)
If HPCI and/or feedwater systems can maintain and recover the water level.
(1.0)
REFERENCE MONTICELLO C.4.III, Primary Containment Isolation, pg C.4-0024 ANSWER 4.10 (2.50)
- a. MAKEUP (in)
CRD (Accept low pressure systems only if candidate assumes the pressare is below the shutoff head of selected pump.)
- b. REJECT (out)
Reliefs (initial)
RWCU to condenser (or radwaste) (2 @ 0.5 ea.)
d
- 4. PRQ.CEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 37 BADJOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
REFERENCE MONTICELLO Primary Containment Isolation, C.4.II, pg C.4-0029 ANSWER 4.11 (1.00)
When the second " accumulator low pressure" alarm is received. (1.0)
REFERENCE MONTICELLO Abnormal Condition - Control Rod System Failure, C.4.III, pg C.4-0069.
ANSWER 4.12 (1.50)
- a. decrease in reactivity.(power)
- b. SLC storage tank draindown.
- c. pump running indication.
- d. SQUIB indicating light out.
- e. Steam flow from vessel decreases. (3 of 5 @ 0.5 ea)
REFERENCE MONTICELLO SLC, B.3.5, pg B.3.5-0010 ANSWER 4.13 (1.00)
! 5% THERMAL POWER REFERENCE MONTICELLO B.6.3, pg B.6.3-21 A
J PAGE 38
, 4. PROCEDUEES - NORMAL. ABNORM 6L. EMEEGENCY AND RADIOLOGICAL CONTROL ANSWERS -- MONTICELLO -86/07/22-KING, M.
ANSWER 4.14 (2.00)
- 1. Check sealing steam to the turbine
- 2. Check steam pressure to the SJAE
- 3. Place second SJAE in sevice
- 4. Valve in standby after condenser trap
- 5. Check cire. water pumps.
- 6. (Check) water boxes completely vented
- 7. Check RFP seal drain tank level
- 8. Reduce power (8 @ 0.25 ea.)
REFERENCE MONTICELLO B.6.3, pg B.6.3-23a l
l i
l l
! - . _ .