ML20213G688
| ML20213G688 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/08/1987 |
| From: | Burdick T, Hare S, Lennartz J, Reidinger T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20213G665 | List: |
| References | |
| 50-282-OL-87-01, 50-282-OL-87-1, 50-306-OL-87-01, 50-306-OL-87-1, NUDOCS 8705180404 | |
| Download: ML20213G688 (112) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Reports No. 50-282/0L-87-01; 50-306/0L-87-01 Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Hall Minneapolis, MN 55401 Facility Name: Prairie Island Excmination Administered At: Prairie Island Examinations Conducted: Senijr Reactor Operator and Reactor Operator o
n _ /. \; ~ -
Examiners: ,S. M. Hare f-8-87 Date
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l 0 d~ , g , y 2 J. A. Lennartz Date
' M '
. D. Reidinger ' b< 'b Date Approved 8y ur k b Chief, Operator Licensing Section Date Examination Summary Examination administered on March 23-27, 1987 (Reports No. 50-282/0L-87-01; No. 50-306/0L-87-01)
Areas Inspected: Written and operating exams were administered to six reactor operators and six senior reactor operators.
Results: Two senior reactor operators and five reactor operators passed the examinations.
8705190404 Er70511' PDR ADOCK 05000282 V PDR<
DETAILS
- 1. Examiners I- S. M. Hare NRC T. D. ReidInger, NRC J. A. Lennartz, NRC o T.-Guilfoil, Sonalysts
~I. Kingsley, Sonalysts
- 2. Examiner Observations During the administration of the operating exams, tne examiners observed both generic deficiencies and strengths on the part of the SR0 and R0 candidates. In addition, one concern was identified by the examiners regarding the verification methodology used when tripping instrumentation bistables.- In the simulator portion of the operating exam the following deficiencies in candidate performance were observed:
- SR0's in general, did not take command of control room activities during abnormal or emergency conditions.
- R0's were inattentive of control board conditions and as a result, their actions were generally reactive in nature, not preventative.
Abnormal board conditions existed for extended periods of time and were not addressed until the condition degraded to an alarm setpoint.
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- Operators were time conscious of simulated events. Instances were noted where candidates stated "The ATWS should be over by now" and "offsite power should be back by now" which indicated to the examiners that the candidates were trained in such a way as to expect transients to conclude in a set period of time. At the exit interview, the Training Supervisor stated that in an ATWS condition "I could leave the control room and trip the Rod Drive MG sets in 10 seconds." Since the Training Supervisor believed that an ATWS may easily and quickly be remedied, the examiners observation that the candidates were trained to expect transients to conclude in a set period of time may be valid.
- Operators did not conform to Red Path Procedure FR-S.1, ATWS.
Specifically, step 1 of FR-S.1 requires the operator to drive rods into the reactor core. Step 4 requires the initiation of emergency boration. Several instances were noted where the operator driving the rods, stopped driving rods several times to initiate emergency boration while the reactor was still at power. The examiners discussed with the licensee their position that the R0 should remain at the console driving rods, and the Lead R0 should initiate
. emergency boration.
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The other area where procedure FR-S.1 was not followed was the way in which the candidates tripped the Rod Drive MG sets from the control room. If the auxiliary operator that had been dispatched to locally trip the MG sets took an excessive amount of time to accomplish this-function (one instance this was 5 minutes), the candidates rapidly opened then closed breakers 13-3 and 14-2. This deenergized the busses supplying the Rod Drive MG sets and dropped' the rods, effectively terminating the ATWS situation. There are two potential problems with this, one being that it is not in the procedure. If this is a viable option that could be taken it should be a step in the procedure. The second problem with deenergizing the busses is the potential loss of all other equipment powered off the busses if that bus failed to re-energize.
- 'A0's and R0's generally did not communicate well during the sia.ulator scenarios.
In the operating exams the following strengths in candidate performance were observed:
- . Candidates were strong in their overall plant knowledge as demonstrated in the walkthrough portion of the operating exam.
- - Candidates were' strong in administrative areas as demonstrated in the simulator and control room portion of the operating exam.
One general concern over the licensee's control of tripping bistables was identified by the examiners. There is a need for a procedure that would provide formal documentation and verification that the proper bistables are being tripped following an instrument failure. It was suggested that some form of procedural sign-off with SR0 verification be utilized, which would provide positive control of the evolution.
- 3. Exit Meeting An exit meeting was held following the examinations to discuss the
. aforementioned observed weaknesses on March 27, 1987 with facility management and training staff representatives.
NRC Representatives in attendance were:
NRC i T. Reidinger, Chief Examiner S. Hare, Examiner
~J. Lennartz, Examiner J. Hard, Senior Resident Inspector M. Moser, Resident Inspector i
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. o-Facility Representatives in attendance were:
E. Watzl, Plant Manager L. Eliason, General Manager Nuclear Plants M. Sellman, General Superintendent of Operations D. Mendele, Plant Engineering and Rad Protection J. Gonyeau, Manager, Production Training T. Amundson, Superintendent of Training D. Reynolds, Operation Training Supervisor R. Lindsey, Assistant Plant Manager M. Balk, Superintendent of Operations M. Werner, Training Center Instructor R. Wirkkala, Training Center Instructor D. Westphal, Training Center Instructor M. Gardzinski,ining D. Benner, Tra Center InstructorTraining Center Instructor W. Bell, Training Center Instructor M. Reddmann, Technical Support Training Supervisor The facility management acknowledged the examiners observations discussed in section 2 of this report.
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ATTACHMENT A At the conclusion of the written examinations the facility is given a copy of the examination and the answer key for both the SR0 and R0 examinations.
- The. facility then has five working days in which to provide written comments
-concerning the examinations to the NRC.
The following paragraphs contain the facility comments concerning the examinations followed by the NRC comments.
REACTOR OPERATOR'S EXAMINATION QUESTION 1.02 Facility Comment: Key specifies 0.5 points for proper method of calculating the mass flow rate. Methods of calculating the flow rate other than that in the key should also receive credit, for example using the substitution method or converting percent to actual plant values to make the calculation and then converting back to percent.
NRC Respon'se: Concur. Any of several methods for calculating mass flow rate will be acceptable providing it is properly applied.
No answer key modification is necessary.
-QUESTION 1.03 Facility Comment: The answer key for part c. of the question allows either
" REMAIN THE SAME" or " INCREASE if Tave is assumed to decrease." Since an explanation is not required by the question, either answer should be acceptable without explanation.
NRC Response: Concur. Either answer will be accepted. -The-answer-key has been modified.
QUESTION 1.12 Facility Comment: The answer key for part c. of the question should also accept "due to a smaller power defect" as a reason why Unit 1 has the greater shutdown margin.
NRC Response: Concur. A small power defect will be accepted in lieu of a less negative MTC as they are cause/effect related and are essentially equivalent. No answer key modification is necessary.
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QUESTION 1.14 Facility Comment: Part a. of the question asks for the affect on MTC of withdrawing rods 20 steps. In addition to the response in the key, pages 3-22 through 3-25 of the reference (attached) states another reason for the change in the increased " sphere of influence" of the rods. At higher temperature, travel neutrons further and with rods out, fewer neutrons are absorbed. Credit should also be given for this response.
NRC Response: Do not concur. The question states the assumption that rod withdrawal is the only factor which affects MTC. This means that any changes in moderator temperature should be ignored. Therefore, the increased " sphere of influence" will not be acceptable for part a. No answer key modification is required.
QUESTION 2.01 Facility Comment: A change made to part a. during the examination requiring setpoints and coincidences. These values are not included in the key. The following setpoints and coincidences should be included in the key:
- S/G low-low level, 67%, 2 out of 3
- UV on buses 11 and 12, 75%, 1 out of 2 twice NRC Response: Do not concur. The change made during the examination was intended to ensure the question objective was understood. No additional information is required for full credit. No answer key modification is required.
QUESTION 2.02 Facility Comment: Part b. of the question asks for two conditions required to be present to stop the diesel generator using the Engine Control Transfer Switch. The question was modified during the' exam to indicate it was referring to stopping the diesel from the control room. However, the Engine Control Transfer Switch is located in the diesel room.
This may have caused some confusion by the examinees.
Referring to logic diagram NF-40325-1, placing the Appendix R isolation switch (listed on the diagram as Irolation Panel CS 55413) in local will prevent stopping the diesel from the control room as will either generator output breaker being closed. As long as the control room control switch is held in stop, these are the only two conditions necessary to stop the diesel. However, if an SI signal or bus undervoltage condition exists, as soon as the switch is released, the diesel will restart.
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m Again referring to the logic diagram, an SI signal actuates the MCA relay such that even if SI is reset, the MCA relay must.be reset also to allow the diesel generator to be stopped. Some examinees may refer to the MCA reset instead of an SI signal.
Thus for stopping the diesel ~ generator from the. control room, any two of the following conditions should be acceptable:
- no SI signal present (or MCA reset)
- diesel generator breakers open
- isolation switch in remote (or not.in local)
- no undervoltage condition on safeguard bus
REFERENCE:
Logic Diagram NF-40325-1 (attached)
NRC Response: Partially concur. The reference material provided does not clearly support the four answers requested for
- acceptance. However, the clarification made during the examination led the candidates to assume the diesel generator was being stopped from the control-room by placing its control switch in STOP. Therefore, the two answers provided in the facility comment for this condition will be accepted for full credit. The answer key has been modified.
QUESTION 3.02.b Facility Comment: Part b. of the question asks for immediate actions occurring as a result of the ' instrument failure. Credit should also be given for indicating an increase in pressurizer program level.
NRC Response: Concur. An increase in pressurizer program level will be accepted in lieu of an increase in charging pump speed as that are cause/effect related and are essentially equivalent. No answer key modification is necessary.
QUESTION 3.02.c Facility Comment: Part c. of the question asks for control and instrumentation systems affected by the Tave and Delta-T Defeat switches. The Delta-T Deviation Alarm is a required answer by the key. Since the question did not specify alarms, only a change in the calculated Rod Insertion Limit should be require for full credit.
NRC Response: Do not concur. The plant annunciator system is part of the plant instrumentation provided for operation of the plant. No answer key modification is required.
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QUESTION 3.08 Facility Commenti Due'to'an error in the reference material, the key is
- wrong ,in the indication for the full range indication of vessel level during normal full power operations. It should also be "off-scale low". Thispositiononthe meter may also be indicated as "RCP ON' as indicated in the RVLIS technical manual. This response should also be acceptable.
REFERENCE:
RVLIS Technical Manual, page 5-24 (attached)
NRC Response: Concur. The answer key has been revised to incorporate the facility comment. The utility is cautioned to correct the faulty reference material prior to the June 1987 examinations.
QUESTION 4.01 Facility Comment: Part a. of the key should also accept one safety injection pump running and RCS pressure <1500 psig with adverse containment as conditions for tripping RCP's.
REFERENCE:
E-0 Information Page-(attached)
.NRC Response: Concur. According to E-0 Information Page, the 1200 psig criteria is equivalent to 1500 psig with adverse containment.
'The answer key has been modified.
QUESTION 4.03 Facility Comment: Part d. of the key is in error for the maximum differential temperature between the pressurizer and the RCS. The correct answer is 320 F according to the Technical Specifications.
The value stated in the key (50 F) is the minimum temperature difference.
REFERENCE:
Technical Specification 3.1.B (attached)
NRC Response: Concur. The reference provided supports this comment.
The answer key has been modified.
SR0 EXAMINATION QUESTION 5.03 Facility Comment: Part a. of the question asks for a manual calculation ofQPTR. Althou of the question,gh thiswe have type no concerniswith of calculation the quality not done 4
b by operators at Prairie Island; they do not have the appropriate normalization factors to make the calculation, nor a procedure for doing it.
NRC Response: .Do not concur. The question requires only rudimentary knowledge (Tech Spec definition) as to how the QPTR should be calculated. Further, no normalization factor (p) is necessary since all full power readings are identical.
-This question is su) ported by Job Task Analysis used.in the generation of NJREG 1122 and is assigned a K/A of 19200SK113 with an importance rating of 3.3.
QUESTION 6.02 Facility Comment: This same question appears on the Reactor Operator examination as question 2.02. Question 2.02 was modified during the exam. However, the SR0 exam provided for review does not indicate whether the same modification to the question was made on the'SR0 exam. We assume the same change was made to the SR0 exam. -
E Part b. of the. question asks for to conditions required
.to be present to stop the diesel generator using the Engine Control Transfer Switch. The question was modified to indicate it was referring to stopping the diesel from the control room. However,_the Engine Control Transfer Switch is located in the diesel room. This may have caused some confusion by the examinees.
Referring to logic diagram NF-40325-1, placing the Appendix R isolation switch (listed on the diagram as-Isolation Panel CS 55413) in local will prevent stopping the diesel from the control room as will either generator output breaker being closed. As long as the control room control switch is held in stop, these are the only two conditions necessary to stop the diesel. However, if an SI signal or bus undervoltage condition exists, as soon as the switch is released, the diesel will restart.
Again referring to the logic diagram, an SI signal actuates the MCA relay such that even if SI is reset, the MCA relay must be reset also to allow the diesel generator to be stopped. Some examinees may refer to the MCA reset instead of an SI signal.
Thus for stopping the diesel generator from the control room, any two of the following conditions should be acceptable:
- no SI signal present (or MCA reset)
- diesel generator breakers open
- isolation switch in remote (or not in local)
- no undervoltage condition on safeguard bus 5
REFERENCE:
Logic Diagram NF-40325-1 (attached)
NRC Response: Partially concur. No clarification was provided durIng the examination. However, part B.1 of the answer key has been modified to accept the F.P. local control switch in remote.
. QUESTION 6.05 Facility Comment: Question asks for three locations core exit thermocouples can be read. Due to recent plant modifications, the thermocouples can also now be read on the Emergency Response Computer System (ERCS) and the Inadequate Core Cooling Monitor (ICCM) in addition to those locations given in the key.
REFERENCE:
B41, page 17,.ICCM Technical Manual, page 1-10 (attached)
NRC Response: Concur. Alternate answers in facility comments accepted.-
QUESTION 6.13 Facility Comment: Although the question is valid, the way it is presented tends to mislead the examinee. Capitalization of words in a question should be used to clarify the question, not mislead. A more appropriate way to ask the question would be "The OT Delta T rod stop setpoint is equal to the OT Delta T trip setpoint - 5%, and it blocks both manual and automatic rod INSERTION."
NRC Response: Comment noted.
QUESTION 6.14 Facility Comment: Part b. of the question asks for two actions to clear a lockout on a safety injection pump. Although the key contains appropriate responses for physically clearing the lockout, some examinees may have interpreted the question to be asking procedurally the actions necessary to clear-the alarm. -Therefore, responses which include determining the cause of the lockout and effecting necessary repairs, which are specified in the Annunciator Response Procedure for an SI pump lockout (attached), and contacting the Electrical Department prior to resetting the lockout as required by procedure C20.5 (attached). Credit should be given for these responses also.
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REFERENCE:
Annunciator Response Procedure for SI Pump.
Lockout, C20.5, page 63 NRC Response: Do not concur. The question was not contained in the procedural section.of the examination; therefore, answers referencing procedural action are not acceptable.
QUESTION 7.01 Facility Comment: Part b, of the question asks why the RCS must be depressurized on a loss of all AC. The key contains a correct response; however, a response "to add the SI accumulators to the system" should also be acceptable as was allowed on question 4.02 of the Reactor Operator exam.
REFERENCE:
ECA-0.0 Background Information, page 3 (attached)
NRC Response: Concur. The answer in the facility comments has been added to the answer key.
QUESTION 7.03 Facility Comment: .The question asks for the basis for tripping the turbine during an ATWS. The response required by the key is a correct answer; however, the examinees may also state the basis listed in the USAR, pages 14.8-10 and 14.8-11 (attached) and Westinghouse Owners Group-Emergency Response Guidelines Low Pressure Version Background FR-S/C/H, pages 76 and 77 (attached) which is to prevent a loss of heat sink or maintain SG inventory. Credit should also be given for this response.
REFERENCES:
USAR, WOG E0P Background NRC Response: Partially concur. The question does not address an ATWS concurrent with a loss of feedwater as the references provided infer. However, credit will be awarded for those who answer the question in this way provided they state that loss of feedwater was an assumption.
QUESTION 7.04 Facility Comment: The question asks for three reasons for " maintaining a minimum" feed flow to the S/G's during an uncontrolled depressurization. The responses given in the key were taken from the background information for step 2 of the procedure ECA-2.1. However, as stated in the sentence preceding the three reasons in the background section, these reasons are for "a reduction of feed flow". The 7
1 correct responses _to the question are found in the-background information for the caution before step 2 which states " Maintaining a minimum verifiable feed-flow.to the SG allows the components to remain in a
" wet" condition, thereby minimizin effects if feed flow is increased.g anyCredit thermal, shock should be
-given for a similar response.
REFERENCE:
ECA 2.1 Background Information, page 2 (attached)
.NRC Response: Concur. -The.second answer in the key has been modified to address the facility comment.
QUESTION 7.05-Facility Comment: Part a. of the question asks why high accumulator pressure is something you should worry about. As indicated in the key, injection-of nitrogen into the RCS is the correct
-response. The other two parts of the key contain responses which are a result of adding nitrogen to the RCS and should not be required for full credit since they were not asked for.
Part b. of the question asks for caution (s) to be observed if you reduce accumulator pressure. The responses stated in the key require memorization of general precautions in a non routine procedure. Job and Task Analysis does not indicate this is required. A more-appropriate caution would be to ensure pressure is not reduced below the Technical Specification minimum accumulator pressure.
'NRC Response: a. Do not concur. The question asks specifically for an explanation of why excessive accumulator pressure
-is to be worried about. The facility comment implies
- i. hat a simple answer is required, not an explanation.
Because'the consequences of nitrogen injection into the RCS are the primary concern over Excessive Accumulator pressure, the answer key will not be revised.
- b. Partially concur. The answer key has been modified i to also accept " ensure pressure is not reduced below the Tech Spec limit."
QUESTION 7.06 Facility Comment: Part b. of the question was revised during the examination.
The new question ~ asks for the " adverse consequences to the RCS". The examinees state that the question was worded
" adverse consequences to the RCP's" when the question 8
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was revised. The key also supports the affects on the RCP's not the RCS. Consideration should be given to this when grading.
NRC Response: Noted.~ The candidates were asked for the adverse effect on the RCP's.
QUESTION 7.08 Facility Comment: The second response in the key should say " Assure there is flow to both S/G" instead of " pressure".
NRC Response: . Concur.
QUESTION 7.09 Facility Comment: Part b. of the question asks why containment hydrogen concentration is a concern during the performance of Response to High Containment Pressure. Per.the reference given in the key in the background information for step 6 of the' procedure, " hydrogen concentration is a concern since a flammable mixture can burn if an ignition source is available and cause a sudden rise in containment pressure." The second response given in the-key should not be required for full credit.
REFERENCE:
FR-Z.1 Background Information, page 2 (attached)
NRC Response: Concur. Answer key corrected.
QUESTION 7.10 ,
Facility Comment: The question refers to an event that happened at Prairie Island in 1986. Part a. asks for indications that could be used to identify a problem with CCW flow to an RHR HX.
The key provides the two indications that are mentioned in the SOE. Credit should be given for other plausible indications of the problem such as total CC flow not increasing as expected or RHR HX outlet temperature higher than expected.
Part b. of the question asks for an explanation of why the RHR train should be declared operable or inoperable.
Per the key, the train should be declared inoperable, which is the conservative answer based on the information given in the question. However, as stated in the SOE, RHR was not declared inoperable since it was determined that the valve would have opened if required during an accident. Credit should be given for either response if adequately explained.
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REFERENCE:
50E-2-86-2 (attached)
NRC Response: a. Partially concur. Abnormal RHR HX outlet temperature would not be valid indication because a surveillance test was being performed. An insufficient increase in overall CCW flow is an acceptable _ alternate answer.
- b. Do not concur. Technical Specifications require that if all automatic valves in the system are not operable, that the system shall be declared inoperable. The correct answer for the question as asked, is the one contained in the answer key.
. QUESTION 8.01 Facility _ Comment: Part b. of the question asks for functions performed by the TSC when activated. Four discrete responses are required by the key with points assigned for each response. Students would be able to more adequately fully answer these types of guestions if the number of required responses were specified in the question.
NRC Response: Any one correct answer will be given full credit.
QUESTION 8.03 Facility Comment: The question asks for the required action if a safety limit is violated. The depth of response indicated by the key is not specified in the cuestion. Full credit should be given for reactor shutcown immediatel notifig, and consult Technical Specifications.y, NRC NRC Response: Concur. Answer key revised accordingly.
QUESTION 8.04 Facility Comment: The way this true/ false question is asked requires the examinee to discern between the words including and excluding. If the intent is .to determine if shift turnover time is included in the work time restrictions, it would be better to ask that.
NRC Response: Comment noted.
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QUESTION 8.05 Facility Comment: Part a. of the question asks under what conditions e procedure or procedural step may be deviated from. Since the question did not specify during emergency conditions, as the key does, credit should be given for responses which address the procedural deviation / change processing procedure (5AWI 1.5.1), i.e., does not change intent of procedure, two SR0's approve, etc.
REFERENCE:
5 ACD 1.5, 5 AWI 1.5.1 (attached)
NRC Response: Partially concur. An emergency is a condition in which a procedure or procedural step may be deviated from. The answer key will be modified to reflect the procedural /
change procedure 5AWI 1.5.1 as an acceptable answer.
QUESTION 8.11 Facility Comment: This question requires memorization of a basis from section 4.0 of Technical Specifications for a surveillance procedure. Prairie Island Job and Task Analysis does not indicate this is required knowledge. Full credit should be given for stating basis is to demonstrate operability.
NRC Response: Concur. Answer key corrected accordingly.
QUESTION 8.12 Facility Comment: Part a. of the question states "What provision (s) in Prairie Island Technical Specifications minimize the consequence / probability of a rod drop accident during low power physics tests?" Technical Specification 3.10 basis states "An evaluation has been made of anticipated transients and postulated accidents, assuming that they occur during the portion of this test when the reactor is critical with all but one full-length control rod fully inserted. Further, the withdrawn full-length rod is assumed not to trip. As a result of this evaluation, it has been determined that for a steam line break upstream of the flow restrictor, the possibility of core DNB exists.
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The overall basis for this Tech Spec is.to ensure trippable reactivity exists in the event of an accident which would add positive reactivity. It does this.by:
those items mentioned in the key. However, the wording of the question concerning a rod drop accident would not. lead to those responses since a rod drop accident is not a consideration. .We believe this question should-be deleted from the exam. If not, wide-latitude in the grading of this question should be given.
Part b. of the question asks why low power physics tests are performed. The response in the key (verify rod worth)-
is only a portion of the overall reason for low power physics tests. Other acceptable answers should include verify core design, measure temperature coefficients, shutdown margin, critical boron concentration and power distribution.
REFERENCE:
Technical Specification 3.10 Basis, page TS 3.10-15 (attached)
NRC Response: Concur. Part a. of this question is deleted. Alternate answers equivalent to the answer. key will be accepted for part b.; however, the answer key will not be changed because references supplied do not support alternate answers.
QUESTION 8.13 Facility Comment: Question asks for memorization of a basis from section 4.0 of Technical Specifications for a surveillance procedure.
Job and Task Analysis does not indicate this is a required knowledge. Wide latitude in the grading of this question should be given.
NRC Response: Do not concur. An SR0 should be aware of_the basis for primary coolant system isolation valves because of the potential for a loss of long-term cooling ability.
Further, Job Task Analysis which was used to generate NUREG 1122 support this question via K/A's 103000K1 and 103000G1 through G5 which have importance ratings l
greater than 3.5.
, QUESTION 8.15 Facility Comment: The question, in general, asks.for an overall summary of Technical Specification 3.7 on the Auxiliary Electrical System and their bases. The key requires specific memorization of a paragraph in the bases. Paraphrasing of the bases should be allowed and full credit given for reasonable phrasing of the listed paragraph.
NRC Response: Concur. Full credit will be allowed for the paraphrasing of the Technical Specification basis.
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4 MASTER U. S. NUCLEAR REGULATORY COMMISSION COPY REACTOR OPERATOR LICENSE EXAMINATION FACILITY: PRAIRIE ISLAND 182 REACTOR TYPE: PWR-WEC2 DATE ADMINISTERED: 87/03/23 EXAMINER: KINGSLEY, I.
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 grade requires at least 70% in each category and a final grade of at least 80%. Examination papers will be picked up six (6) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 -"-
25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW 25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS
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25.00 25.00 3. INSTRUMENTS AND CONTROLS 25.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL 100.00 % Totals Final Grade l
l All work done on this examination is my own. I have neither given nor received aid.
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' Candidate's Signature l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS l
l 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 a% 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 new page, write only on ons side of the paper, and write "Last Page" on the last answer sheet
- 9. Number each answer as to category and nunber, for example,1.4, 6.3.
- 10. Skip at least three lines between each answer.
- 11. Separate answer sheets from pad and place finished answo sheets face down on your desk or table.
- 12. Use abbreviations only if they are connonly used in fac 'lity 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 answr 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 ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
- 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examination. This must be done af ter the examination has been completed.
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- 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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, . , _ . y-__- ,.- . _ _ . _ , _......- ._,... .,- ,,-._ ,,,._ , , ,- . __ . ,__ .-- _.,-_. ..,,_.____,.,..._.,,..._____.__m._-
" 1. -PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 2
. THERMDDYNAMIC5, HEAT TRANSFER AND FLUID FLOW QUESTION 1.01 (2.50)
- a. Primary system flow rate is greater than secondary system flow rate while the heat transferred by the two systems is essentially the same.
EXPLAIN how this is possible. (1.50)
- b. Which DNE of the following describes the changes to the steam that occur between the inlet and outlet of the high pressure turbine? (1.00)
- 1. Enthalpy decreases, quality decreases.
- 2. Enthalpy increases, quality decreases.
- 3. Enthalpy increases, quality increases.
~
!- 4. Enthalpy decreases, quality increases.
QUESTION 1.02 (1.00)
The reactor is producing 100% rated thernal power at a core delta-T of 60 degrees and a RCS mass flow rate of 100% when a station blackout occurs. Natural circulation is established and core delta-T goes to 28 F.
If decay heat is 2% rated thermal power, what is the core mass flow rate l' in percent?
QUESTION 1.03 (2.00)
Will the Departure from Nucleate Boiling Ratio (DNBR) INCREASE, DECREASE, or REMAIN THE SAME for the following changes in plant conditions? The plant is initially operating at 85 percent. power with all control systems in AUTOMATIC (unless stated otherwise) and Bank D control rods at 150 steps. Consider steady state power operation unless stated otherwise.
Consider each change INDEPENDENTLY and assume the reactor does NOT trip.
- a. All pressurizer heaters are er ergized with spray valves in manual.(0.50)
I
- b. RCS flow increases by 10%. ,
(0.50)
- c. The RCS is inadvertantly boratec by 10 ppm. (0.50)
- d. Turbine load is increased to 100 ptrcent using RCS boron adjustment to i maintain a constant rod height. (0.50)
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
-1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 3 '
- ' ' THERMODYNAMIC 5, HEAT TRAN5FER AND FLUID FLOW QUESTION 1.04 (1.00)
The reactor is subcritical with Bank C at 46 steps. An ECP has just been calculated that snows 1300 pcm are needed to reach criticality. Use the appropriate Rod Worth curve to determine the required bank position.
Assume no change in boron concentration, xenon, or RCS temperature.
QUESTION 1.05 (2.00)
State whether the following statements regarding the Net Positive Suction 1 Head (NPSH) available to the RHR pumps during the LOW HEAD RECIRCULATION l- phase following a LOCA are TRUE or FALSE and EXPLAIN your answer.
- a. Raising RCS pressure from 50 psig (current) to 100 psig will decrease the available NPSH to the RHR pumps. (1.00)
- b. As containment sump fluid temperature decreases, available NPSH to the RHR pumps decreases. (1.00)
J QUESTION 1.06 (2.50) i
- a. - The reactor is subcritical by 2.5% delta-k/k. The count rate is 115 CPS. Af ter a positive reactivity insertion, the count rate increases to 345. How much reactivity was added to the core? (1.50) ,
- b. Why does it take longer, after each reactivity addition, for the neutron population to reach equilibrium as Keff approaches 1.0? (1.00) i
}
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 4
., THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.07 (1.00)
During a reactor startup, the first reactivity addition caused count rate to increase from 20 cps to 40 cps. The second reactivity addition caused count rate to increase from 40 cps to 80 cps. Which ONE of the following statements is CORRECT?
- a. The first reactivty addition was larger.
- b. The second reactivity addition was larger.
- c. The first and second reactivity additions were equal.
- d. There is not enough data given to determine relationship of reactivity values.
QUESTION 1.08 (2.50)
- a. List the three most significant contributors to total power coefficient in order of INCREASING magnitude at BOC. (1.50)
- b. How does total power coefficient vary as the core ages? (1.00)
QUESTION 1.09 (1.00)
! The plant is in Hot Shutdown with a cooldown in progress via the steam dump system. In order to maintain a 200 F subcooling margin in the RCS when i reducing RCS pressure to 1600 psig, steam generator pressure must be l
reduced to approximately: (Select the correct answer.)
l a. 405 psig
- b. 325 psig
- c. 245 psig
- d. 165 psig
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 'i 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 5 THERMODYNAMICS, HEAT TRANSFER AND FLUID FLOW QUESTION 1.10 (2.00)
Indicate whether the following situations result in SUBC00 LED, SATURATED, or SUPERHEATED fluid conditions. The plant is at 50 percent power.
- a. Pressurizer PORY relieving to the PRT (0.50)
- b. Steam generator safety valve relieving to atmosphere (0.50)
- c. Steam from a Moisturt Separator Reheater entering a low pressure turbine (0.50)
- d. Condensate exiting the condenser hotwell (0.50)
QUESTION 1.11 (1.50)
List THREE core-related effects which would cause the Power Range indications to increase over core life if no adjustments were made.
QUESTION 1.12 (3.00)
Prairie Island Units 1 and 2 are operating at full power with all control systems in automatic. Unit 1 is at BOC with RCS boron at 900 ppm and Unit i 2 is at EOC with RCS boron at 200 ppm. Each Unit has Bank D control rods -
at 200 steps.
- a. Which unit, if either, has the larger (more negative) Differential Boron Worth and WHY? (1.00)
- b. Which unit, if either, has the larger ('more negative) Moderator Temperature Coefficient and WHY? (1.00)
- c. Which unit, if either, has the greater Shutdown Margin and WHY? (1.00) l
(***** CATEGORY 01 CONTINUED ON NEXT PAGE *****)
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 6 THERMDDYNAMICS, HEAT TRAN5FER AND FLUID FLOW QUESTION 1.13 (1.00)
Choose the CORRECT response. Shutdown Margin" as used in Technical Specification 3.10 is the amount by which the reactor core would be sub-critical at hot shutdown conditions if all control rods were tripped, assuming:
- a. normal hot channel factors are maintained, and assuming no changes in xenon or boron concentrations.
- b. that the highest worth control rod assembly remained fully withdrawn, and assuming xenon-free conditions and no changes in boron concentrations.
- c. normal hot channel factors are maintained, and assuming xenon-free conditions and no changes in boron concentration.
- d. that the highest worth control rod assembly remained fully withdrawn, I and assuming no changes in xenon or boron concentration.
QUESTION 1.14 (2.00)
Explain HOW and WHY Moderator Temperature Coefficient is affected by the following changes. Assume each is the only factor affecting MTC.
- a. rod withdrawal (20 steps) with the reactor at power. (1.00)
- b. increasing RCS temperature. (1.00) l
(***** END OF CATEGORY 01 *****)
-- - -_ , __ ,--, ._% - , , , . y-y- ._,. ,._,,, ,y .,,_,..__,,. __-.____,, ,,,..,_ , .-_ ___y _ _ _ _ . _ . , , _ _ - _ , , . , - , ,
- . =- . .. - ._. _
2.- PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 7 QUESTION 2.01 (3.00)
- a. List the FOUR unique signals which will initiate a turbine-driven g auxiliary feedwater AUTO-START signal. () Cit)6 S2(gTg C0thtM 0)
- b. What are the NORMAL and BACKUP sources of water to the turbine-driven auxiliary feedwater pumps? (1.00)
- c. With an auto start signal initiated, HOW and under WHAT CONDITIONS will the Auxiliary Feed Pump water supply be shifted from the normal supply to the emergency supply? Include setpoints. (1.00)
- QUESTION 2.02 (3.00)
- a. What TWO conditions will cause an AUTOMATIC START of an emergency diesel generator? Consider Unit 1 only. (1.00)
- b. Placing the Engine Control Transfer switc in the STOP position will shutdown the associated diesel generator i TWO conditions are present. -
What are these TWO conditons? (1.00)
- c. Following a loss of offsite power with a safety injection signal present, which of the following abnormal conditions, if occurring separately, will result in a diesel generator trip? (More than one answer may be correct.) (1.00) i
- 1. Excessive vibration
- 2. Generator phase differential
- 3. Generator reverse power
- 4. Low lube oil pressure
- 5. Overspeed
- 6. High jacket water temperature
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 8 QUESTION 2.03 (2.50)
- a. Why are the following features designed into each RCP7 (1.50)
- 1. Flywheel
- 2. Anti-reverse rotation device
- 3. Thermal barrier heat exchanger
- b. What TWO INTERLOCKS must be satisfied in order to start an RCP when its control switch is placed in START? (1.00)
QUESTION 2.04 (2.00)
For each one of the following process and/or area radiation monitors, state ALL the automatic actions which will occur when the high level alarm setpoint is reached. If no automatic actions occur, state this in your answer. (0.25 for each)
- a. R-11,12 containment air
- b. R-15 Condenser air ejector
- c. R-18 Waste liquid
- d. R-19 SG blowdown
- e. R-22 Shield building stack -
- f. R-23,24 Control room ventilation
- g. R-30,37 Auxiliary building ventilation
- h. R-39 CC system
(***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
- 2. PLANT DESIGN INCLUDING SAFETY AND EE RGENCY SYSTEMS PAGE 9 QUESTION 2.05 (1.50)
Select the THREE LOADS from the following list which would be deenergized following the loss of the 4160 volt bus 13.
- a. 11 Coolir Water Pump
- b. 13 Circu.. ting Water Pump
- c. 11 FWP
- d. 13 Heater Drain Pump
- e. 11 Condensate Pump
- f. 12 Circulating Water Pump QUESTION 2.06 (1.50)
Describe THREE coolant flowpaths within the reactor vessel which effectively bypass the core (i.e., do not provide core cooling) during normal plant operation. ,
- QUESTION 2.07 (2.50)
Refer to FIGURE 2-1 (RHR System) to answer the following questions: -
- a. List the TWO interlocks associated with MV-32230 which prevent overpressurizing the RHR system. Include setpoints. (1.00)
- b. State whether the following valves are OPEN or SHUT during the Low Pressure Recirculation Phase following a LOCA. (Place responses on your answer sheet.) (1.50)
- 1. MV-32066 l
- 2. MV-32065 l 3. MV-32230
- 4. MV-32085 5.
- MV-32076
- 6. MV-32093 -
l (***** CATEGORY 02 CONTINUED ON NEXT PAGE *****)
L
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 10 QUESTION 2.08 (2.50)
State how the following components respond (FAIL OPEN, FAIL CLOSED, REMAIN FUNCTIONAL, DIVERTS TO .... ETC.) when instrument air pressure is lost with the plant at Iv0% power.
- a. Low preuurc letdown control valve (CV-31203) (0.5)
- b. Volume rontrol tank letdown divert to HUT valve (CV-31205) (0.5)
- c. Steam generator PORVs (0.5)
- d. Pressurizer PORVs (0.5)
- e. Cold leg accumulator injection isolation valves (0.5)
. QUESTION 2.09 (1.50)
List THREE REASONS for the 30 second time delay associated with a generator trip resulting from a turbine trip.
QUESTION 2.10 (2.50)
- a. List TWO possible symptoms of leakage into the Component Cooling ,,
System. Assume no alarm setpoints are reached and no automatic actions occur. (1.00)
- b. List SIX components cooled by the CC system that could leak into the
QUESTION 2.11 (1.50)
List THREE conditions which must exist for a letdown orifice isolation l
valve to open when its control switch is in the OPEN position. Include setpoints.
QUESTION 2.12 (1.00)
List TWO reasons for maintaining a minimum pressurizer spray flow during normal "at power" operations.
(***** END OF CATEGORY 02 *****)
'! 3. INSTRUENTS AND CONTROLS PAGE 11 QUESTION 3.01 (3.00)
- a. List FOUR conditions that will prevent manual rod withdrawal. Include SETPOINTS and COINCIDENCES. (2.00)
- b. Compare L(gic Cabinet " URGENT" and "NON-URGENT" alarms regarding their effect on the Rod Control System. (1.00)
QUESTION 3.02 (3.50)
The plant is operating at 80s power with all control systems in automatic.
- a. An instrument malfunction causes Channel III Tave and Channel III Delta-T indications to FAIL HIGH. Identify the INSTRUMENT and type of MALFUNCTION that caused these responses. (1.00)
- b. What IlttEDIATE AUTOMATIC action (s) occur (s) as a result of the malfunction? (1.00)
- c. The operator subsequently selects the failed channel on the Tave and Delta-T Defeat switches. What control and instrumentation systems are affected when these switches are operated? Be SPECIFIC as to the effects for EACH switch. (1.50)
QUESTION 3.03 (1.50) .
- a. The reactor is critical at 10E-8 amps during a reactor startup. A malfunctioning steam header pressure transmitter causes three steam dump valves to open. Assuming the reactor does NOT trip, at what average temperature (Tave) will the Reactor Coolant System stabilize?
(0.50)
- b. Describe the feature (s) which cause(s) Tave to stabilize at this value.
(1.00) 4 m
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
_ _ - . . - . . _ _ _ _ _ _ _ . _ _ -. _ . , _ _ - - . _ _ _ . _ , , _ _ _ . _ , . _ _ . - . -,__.,___ _-_,,_ ~_._- ,._-_ ,,
- 3. INSTRUENTS AND CONTROLS PAGE 12 QUESTION 3.04 (2.50)
- a. What is the reason for RESETTING SI after actuation? (0.50)
- b. What TWO conditions must exist for SI to RESET when attempted by an operator?' (1.00)
- c. What SI SIGNALS should be BLOCKED and WHAT CONDITIONS must be present for the blocks to be EFFECTIVE during an RCS cooldown and depres-surization. (1.00)
QUESTION 3.05 (2.00)
- a. Describe one IR instrument response if the circuitry is undercompensated during a reactor shutdown, including any effects on SR instrumentation. - Include any applicable setpoints- (1.00)
- b. What operator action (s) is/are required to continue a reactor shutdown if one IR channel has failed high? Influde setpoints. (1.00)
C
-QUESTION 3.06 (1.50)
Indicate which of the Excore Nuclear Instrumentation Ranges (SOURCE, INTERMEDIATE, or POWER), will correctly match with the following statements. More than one may apply to each.
- a. Provides a direct input to the Rod Control System for speed and direction control.
- b. Has a reactor trip function that is blocked at some time between startup and full power operation.
- c. Ulilizes'cf rcuitry which ensures- gamma contribution to output of instrument is minimized.
QUESTION 3.07 (1.50)
List A3 RPS tgs that are provided for DNB protection.
l
(***** CATEGORY 03 CONTINUED ON NEXT PAGE *****)
l
- 3. INSTRt#ENTS AND CONTROLS PAGE 13 QUESTION 3.08 (1.50)
List the THREE indications provided by the Reactor Vessel Level Indicating System including their expected readings during normal full power operations.
QUESTION 3.09 (2.00)
TRUE or FALSE
- a. Failure of only ONE pressurizer pressure channel high CANNOT result in opening a pressurizer PORY with the plant at normal pressure. (1.00)
, b. All Pressurizer PORVs can be PREVENTED from opening on any automatic signal by holding their control switches in the CLOSE position. (1.00)
QUESTION 3.10 (1.00)
Other than low pressurizer pressure, viiAT CONDITION will cause automatic energizing of the pressurizer backup heaters? Include setpoints.
QUESTION 3.11 (1.50)
Compare Bank Demand Position Indication to Individual Rod Position .
Indication by answering the following.
- a. Which is more RELIABLE? (0.50)
- b. Which is more ACCURATE? (0.50)
- c. Which is INCAPABLE of detecting a stuck rod? (0.50)
QUESTION 3.12 (2.00)
The plant is at 75% power with all control systens in automatic.
Safeguards logic testing is in progress with the proper safeguards racks in Test and TWO steam pressure low pressure SI signals inserted. The operator resets both trains of SI. Will a SI occur? Why or why not?
(***** CATEGORY 03 CONTINUED ON NEXT PAGE
- )
m u -. _ _
- 3. INSTRUENTS AND CONTROLS PAGE 14 QUESTION 3.13 (1.50)
Other than an "S" signal list THREE signals / conditions that will initiate a containment ventilation isolation signal. (Include coincidences in your answer. )
(***** END OF CATEGORY 03 *****)
.. _ . - - - - . - - . - - _ . . ~ . . - - - - - . - . -
$ 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 15 RADIDLD6ICAL CUNTROL QUESTION 4.01 (2.00)
- a. List TWO conditions that together require the operator to trip all RCPs in accordance with E-0.0, Reactor Trip or Safety Injection. (1.00)
- b. What condi: ion (s) require (s) the tripping of RCPs following loss of CCW and seal injection flow in accordance with E-12, Loss of RCP Cooling Water? (1.00) l QUESTION 4.02 (2.00)
During a loss of all AC power, ECA 0.0 has the operator depressurize the RCS using steam generator PORVs.
- a. Why must the RCS be depressurized? (0.50)
- b. Why shouldn't the RCS be depressurized below 390 psig? (0.50)
- c. Why shouldn't the RCS be cooled below 320 F? (0.50)
- d. Regarding the depressurization, how should the operator respond if reactor vessel upper head voiding occurs? (0.50)
QUESTION' 4.03 (2.00)
Supply the following limits which must be observed during plant operation in accordance with C1.2, Unit Startup Procedure.
- a. Maximum RCS heatup rate (0.50)
- b. Maximum pressurizer heatup rate (0.50)
[ (0.50) l
- d. Maximum differential temperature between pressurizer and RCS (0.50) l l
l
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- . . - . . . . . - . - - - _ - - _ _ - _ _ _ = _. - - _ - - . -. - . - ._
~
4 -
- 4. PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND PAGE 16 RADIDLDEICAL LURIKUL QUESTION 4.04 (1.50)
- TRUE or FALSE
- a. The position of a throttled manual valve muy be independently verified hy closins and then reopening the valve the proper number of turns.
(0.50)
- b. The position of a fully open manual valve may be independently verified hy attempting to open the valve hy using its manual operator. (0.50)
- c. The position of motor or air operated valves may be independently verified hy observing remote valve position light indication. (0.50) t QUESTION 4.05 (1.00)
Explain why it is permissible to disable ar, annunciator which frequently
, alarms and clears other than for repair.
QUESTION 4.06 (1.00)
! Under WHAT_ CONDITIONS may nonlicensed individuals operate the reactivity controls in the control room? ,
- QUESTION 4.07 (1.00) i
- Which DNE of the following CSF conditions would have the HIGHEST priority in requiring operator response?
- 1. RCS Inventory - Orange Path
- 2. Core Heat Sink - Red Path i 3. Containment Integrity - Red Path j 4. Subcriticality - Orange Path l 5. RCS Integrity - Red Path
- 6. Core Cooling - Orange Path l
I
( ***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
- 4. PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND PAGE' 17
. RADIOLOGICAL CONTROL
- (
QUESTION 4.08 (2.50)
- a. Emergency Boration flow to charging pump suction has been limited to lap gpm byBoration Emergency throttlingflow?
VC-11-58. What is the REASON for limiting (0.50)
- b. In accordance with C-12, CVCS, what THREE specific operator actions are required to initiate emergency boration via y}"gg .p) g g
- c. In accordance with FR-S.1, Response to Nuclear Power Generation /ATWS, under what TWO conditions may Emergency Boration be TERMINATED? (1.00)
QUESTION 4.09 (1.00)
Select the ONE group of indications which are characteristic of natural circulation in accordance with ES-0.1, Reactor Trip Recovery.
RCS SG CORE EXIT SUBC00 LING PRESSURES Thot Tcold THERMOCOUPLES
- a. 50 F Increasing Decreasing Constant Decreasing
- b. 60 F Decreasing Decreasing Constant Decreasing
- c. 70 F Constant Decreasing Decreasing Decreasing
- d. 80 F Decreasing Constant Decreasing Constant
- e. 90 F Increasing Constant Increasing Decreasing QUESTION 4.10 (2.00)
In accordance with E-0.0, Reactor Trip or Safety Injection, what FOUR CONDITIONS provide verification of a reactor trip?
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
I 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 18 RAD 10LU6ICAL CUNIMUL QUESTION 4.11 (1.50)
Unit 2 is operating at 100% power when a telephone call is received reporting a bomb in the control room. A quick search reveals a
" suspicious" package which prompts the Shif t Supervisor to order evacuation of the control room. What are the initial responsibilities of the Unit 2 operators (SS, LPE 8 RO, and PE 8 RO) after exiting the control room?
Assume no operator actions have been taken prior to exiting.
QUESTION 4.12 (2.00) ,
- a. Provide the following whole body exposure limits (non-emergency) for a licensed operator. (1.50)
- 1. Administrative quarterly limit without extension
- 2. Maximum administrative quarterly limit with extension
- 3. Administrative yearly limit without extension
- b. Who may authorize extension of the administrative yearly limit for whole body exposure? (0.50) j QUESTION 4.13 (1.00)
Coglete as Required .
Adverse Containment conditions, as defined in emergency procedures, exist when containment pressure exceeds (a) psig (0.50) or when containment radiation level exceeds (b) R7Rr. E50) f 4
4
(***** CATEGORY 04 CONTINUED ON NEXT PAGE *****)
i 1
- - ---,,.-...-.m., -,.--,--..~..,_--,,.n. , . , - , . . - - , , ,,,-,,,_,.,,--,.,,,n~,, -
$ .4. PROCEDURES - NORMAL, ABNORMAL, EERGENCY AND PAGE 19 RADIOLOGICAL CONTROL
( QUESTION 4.14 (2.00) l l Refer to figure 4-1 for the following question.
Given each of the following sets of indications, determine whether AXIAL FLUX DIFFERENC
- is being maintained WITHIN the appropriate LIMITS and explain briefly WHY or WHY NOT. (0.50 each)
POWER AFD AFD AFD AFD LEVEL CHANNEL 1 CHANNEL 2 CHANNEL 3 CHANNEL 4
- 1. 63% -17 -22 -16 -21
- 2. 80% -12 -13 -20 -12
- 3. 70% +14 +20 +19 +18
- 4. 55% Out of Service -20 -22 -25 QUESTION 4.15 (1.00)
Technical Specifications require the RCS accumulators to be operable. WHAT CONDITIONS must be met in order for the accumulators to be OPERABLE?
QUESTION 4.16 (1.50) .
What are the THREE BASES for maintaining ROD INSERTION LIMITS?
(***** END OF CATEGORY 04 *****)
(************* END OF EXAMINATION ***************)
EQUATION SHEET f = ma v a s/t Cycle efficiency = (Net work out)/(Energy in) 2 w = og s = V ,t + 1/2 at 2
E = ac ~
KE = 1/2 av a = (Vf-V,)/t A = AN A*Aeg PE = agn Vf = V,+ at * = e/t x = an2/t3fg = o.693/t1/2 y , , p. ,o r t
1/teff = C(tu,1(t ,)] i A= , [(t1/2I
- I*bIl aE = 931 as
""Y avA ' ~
I = I ,e Q = aCpat 6 = UAaT I*I'o pwr = Wgah I
- I,10,,gyg TVL = 1.3/u sur(t) HYL = -0.693/n P = P,10
( .
p , p ,t/T SUR = 26.06 g SCR=S/(1-Keff)
CR x
- 3/(I ~ Keffx)
SUR = 26,/s* + (s - p)T CRj (1 - K ,ff3) = G 2 0 - Ieff2)
T = (1*/s) + [(s - oV io] M = 1/(1 - K,ff) = CR /CR j , ,
T = s/(o - s) M = (1 - K ,ff,)/(1 - K,ffj)
T = (s - o)/(Is) SOM = ( -K,ff)/K,ff a = (K ,ff-1)/K ,ff = aK ,ff/K,ff s' = 10 seconds I = 0.1 seconds"I e = [(t*/(T K,ff)] + [a,ff /(1 + IT)]
Id lj=Id2 ,2gd2 P = (14V)/(3 x 1010) Idjj 22 2 I = eN R/hr = (0.5 CE)/d (meters)
R/hr = 6 CE/d2 (feet)
Water Parameters Miscellaneous Conversions I gal. = 8.345 lem. 1 curie = 3.7 x 1010dos 1 ga]. = 3.78 liters ] kg = 2.21 lbm 1 fte = 7.48 gal. 1 np = 2.54 x 10 Stu/nr Density = 62.4 lbT/ft3 1 mw = 3.41 x 10 Stu/hr Density = 1 gm/cma lin = 2.54 cm Heat of vaporization = 970 Stu/lom *F = 9/5'C + 32 Heat of fusion = I44 Stu/lbm *C = 5/9 (*F-32) 1 Atm = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf 1 f t. H2O = 0.4335 lbf/in.
1 Table 1. Saturated Steam: Temperature Table Entropy Enthalpy Temp j Specific Volume Sat. Sat.
Abs Press. Sat. Sal. Fahr l Sat. Sal. Liquid Evap Vapor Lb per Liquid Evap Vapor Temp Liquid Evap Vapor 58 l fahr SqIn. he h ig hs 5f SIS p vg vig v8 2.1873 2.1873 32 8 t 0.0179 1075.5 1075.5 0.0000 34.8 0 016022 33043 33043 1076.4 0 0041 2.1762 2.1802 32.8 0 08859 1.996 1074.4 2.1732 36 8 0 01602I 3061.9 3061.9 1077.2 0 0081 2.1651 34.8 0 09600 4.008 1013.2 2.1663 38.8 0 016020 2839 0 2839.0 1078.1 0.0122 2.1541 36 8 010395 6.018 1072.1 0 016019 2634.1 2634.2 38.8 0.ll249 0 0162 2.1432 2 1594 48.8 8.027 1071.0 1079.0 42.8 0 016019 2445.8 2445 8 1069 8 1079.9 0 0202 2.1325 2.1527 44 8 48 I I.12163 2272.4 10.035 0 0242 2.1217 2.1459 0 016019 2272.4 10683 10801 0 13143 12.041 46.8 42 8 0 016019 2112 8 2112.8 1067.6 1081.6 0 0282 2.1111 2.1393 44 8 0 14192 1965 7 14 047 0.0321 2.1006 2.1327 48 8 0 016020 1965 7 1066.4 1082.5 46 8 0 15314 1830 0 16 051 0 016021 1830 0 48 8 0 16514 0 0361 2.0901 2 1262 $4.0 18 054 1065.3 1083.4 52 8 0 016023 1704 8 1704 8 1084.2 0 0400 2.0798 2.1197 54 0 0 I1796 1589 2 20 057 1064.2 0 0439 2.0695 2.1134 54 8 0 016024 1589 2 1085.1 52 8 0 19165 1482.4 22 058 1063.1 0 0478 2.0593 2.1070 56.8 0 016026 1482 4 1086.0 54 8 0 20625 1383 6 24 059 1061.9 2.0491 2.1008 58.8 0 016028 1333 6 1086 9 0 0516 56 0 0 22183 1292.2 26 060 1060.8 0016031 1292.2 $8.8
- 58 8 0 23843 0.0555 2.0393 2.0946 28 060 10593 10871 2.0885 62.0 0 016033 1207.6 1207.6 0.0593 2.0291 0 25611 30 059 1058.5 1088.6 64 8 54 8 0 016036 1129.2 1129.2 1089 5 0 0632 2.0192 2.0824 0.27498 32 058 1057 4 66 8
$2 0 0 016039 1056.5 1056.5 1056.3 1090 4 0 0670 2.0094 2 0764 68 8 54.8 0 29497 989.1 34.056 0.0708 1.99 % 2 0704 0 016043 989.0 1055.2 1091.2 56 8 0 31626 926.5 36 054 0 016046 926.5 78 8 68 8 0 33889 0 0745 1.9900 2.0645 38 052 1054.0 1092.1 12.8 0.016050 868.3 868.4 1093.0 J J723 1.9804 2 0587 78 8 0.36292 814.3 40 049 1052.9 1.9708 /0529 74.8 .
0 016054 814.3 1093 8 0 0821 72 8 0.38844 764.1 42.046 1051.8 1.% 14 2.0472 76 8 0 016058 764.1 10941 0 0858 74 8 0 41550 717.4 44 043 10501 1.9520 2.0415 78 8 0 016063 717.4 1095.6 0 0895 78.8 0 44420 673.9 46.040 1049.5 0 016067 6 73.8 0 47461 88.8 .
78 8 1096.4 0.0932 1.9426 2.0959 633.3 48 037 1048.4 1.9334 2 030J 82.8 0 016072 633.3 1097.3 0 0969 88 8 0 50683 595.5 50 033 1047.3 1:9242 2.0248 84 8 0.016071 595 5 1098 2 0 1006 82 8 0.54093 560.3 52 029 1046.1 1.915I 2 0193 86 8 0 016082 560.3 1099 0 0.1043 84 8 0 57702 527.5 54 026 1045 0 1.9060 2.0139 88.8 0 016087 227.5 1099 9 0.1079 IE S 0 61518 4%8 56 022 1043 9 065551 0 016093 4%8 189F0 2 0086 98.8 88 8 10423 1100.8 0.1115 32.8 468I 468.1 58 018 0.1152 1.8881 2 033 0 69813 0 016099 60 Old 1041.6 1101 6 94.8 98 8 0 016105 441.3 441.3 IIO2 5 0 1188 1 8792 1 9980 97 0 0 74313 416 3 62 010 1040 5 I.8704 19928 96 0 0 016111 416 3 Il03 3 0 1224 94 8 0 79062 392.9 64 006 1039 3 1.8617 19876 98 8 0 84072 0 016117 3928 10382 11042 0 1260 96 8 370 9 37.0 9 66 003 0 89356 0 016123 98 8
1 Enthalpy Entropy Abs Press. Specific Volume Sat. Sal. Temp Sat. Sal. Sat.
Temp tb per Sal. Evap Vapor Liquid - Evan Vapor Fahr l Liquid Evap Vapor Liquid Fahr SqIn. hg s, sg t i 1
- v sg he h ig sl8
' I p vg V8 0.1295 1.8530 1.9825 108.8 350.4 67.999 1037.1 1105.1 0.94924 0 016130 350.4 0.1331 1.8444 1.9775 182.0 -
188 8 331.1 69.995 1035.9 1105.9 1 00789 0 016137 331.I 0.1366 1.858 1.9725 108.8 i 182 8 313.1 71.992 1034.8 1106.8 0.016144 313.1 0.1402 1.8273 1.9675 188 8 104 0 1 06 % 5 73.99 1033.6 1107.6 0 016151 296.16 296.18 0.1437 1.8188 1.9626 100.0 105 0 1.1347 75.98 1032.5 1108.5 0 016158 280.28 280.30 188.8 12030 1109.3 0.1472 1.8105 1.9577- IN.8 265.37 265.39 77.98 1031.4 112.0 ;
118.8 1.2750 0.016165 79.98 1030.2 1130.2 0.1507 1.8021 1.9528 0 016173 251.37 251.38 0.1542 11938 1.9480 . 114.0 i 112.8 1.3505 81.97 1029.1 1111.0 23822 1144 1.4299 0 016180 238.21 83.97 1027.9 1111.9 0.1577 13856 1.9433 116.0 0 016188 225 84 225 85 0.1611 .I 3 774 1.9386 118.8 186 8 1.5133 85.97 1026.8 1112.7 0 016196 214.20 214.21 118 8 1.6009 120.8 87.97 1025.6 1813.6 0.1646 13693 1.9339 0 016204 203.25 203.26 0.1680 13613 1.9293 122.8 120.8 1.6927 89.96 1024.5 1114.4 0 016213 192.94 192.95 0.1715 13533 1.9247 124.0 122 0 1.7891 91.96 1023.3 1115.3 0 016221 18323 183.24 0.1749 1.7453 1.9202 128.8 124 8 3.8901 93.96 1022.2 1116.I 0 016229 174.08 174.09 0.1783 1.7374 1.9157 128.0 125 O 1.9959 95.% 1021.0 1117.0 0 016238 165.45 165.47 '
128.0 2.1068 138.8 1019.8 1117.8 0.1817 13295 1.9112 157.32 157.33 97.96 13217 1.9068 132.8 138 8 2.2230 0 016247 10183 1118 6 0.1851 0015256 149.64 149.66 99.95 13140 1.9024 134.0 1328 2.3445 1017.5 1119.5 0.1884 0 016265 142.40 142.41 101.95 13063 1.8980 138.0 134 0 2.4717 1016.4 1120.3 0.1918 I
135 55 13557 103 95 1.6996 1.8937 138.8
' 135 0 . 2.6047 0 016274 1015.2 1121.1 0.1951 129.09 129.11 105.95 138 0 2.7438 0 016284 0.1985 1.6910 1.8895 NOS 107.95 1014.0 1122.0 0016293 122.98 123 00 0.2018 1.6534 1.8852 142.0 t 140 8 2 8892 109.95 1012.9 1122.8 144.0 3 0411 0016303 117.21 117.22 0.2051 1.6759 1.8810 142 8 111.95 10113 1823 6 146.8 0 016312 11114 Ill16 0.2084 1.6644 1.8769 144 8 3.1997 I13 95 1010.5 1124.5 0 016322 106.58 106 59 0.2117 l.6610 1.8727 140.8 146 I 3.3653 115.95 1009.3 1125.3 3.5381 0 016332 101.68 10110 I44 8 150.8 i 1008.2 1125.1 0.2150 1.6536 1.8686
- 0016343 97.05 97.07 117.95 1.6463 1.8646 152.8 158 8 33184 1007.0 1126.9 0.2183 :
0 016353 92.66 92.68 119.95 1.6390 1.8E06 154.0 152.0 3.9065 1005.8 11273 0.2216 0 016363 88 50 88 52 121.95 1.6318 1.8566 150.0 ,
154.8 4.1025 1000.6 1128.6 0.2248 0016374 84.56 84.57 123.95 1.6245 1.8526 150.0 l 156 I 4 3068 1003.4 1129.4 0 2281 -
i 0016384 80.82 30.83 125.96 158 8 4.5197 1.6174 1.8487 188 8 0.23I3 127.96 1002.2 1130.2 182.0 l 43414 0.016395 77.27 77.29 02345 1.6103 1.8448 188 8 129.96 1001.0 1131.0 4 9722 0016406 73.90 73.92 02377 1.6032 1.8409 184 0 162.8 7012 131.% 999.8 1131.8 108.0 5 2124 0 016417 7030 1132.6 0 2409 1.5961 1.8371 184 8 67.68 13397 998.6 188.0 54623 0016428 67.67 0.2441 1.5892 1.8333 ,
156 0 64 80 135 97 997.4 1133.4 5 7223 0 016440 6438 168 8 IN.8 !
996.2 !!34.2 0.2473 1.5822 1.8295 0 016451 62.04 62.06 137.97 1.5753 1.8258 1 72.0 i ING 5 9926 9950 1135 0 0.2505 0016463 5943 59.45 139 98 1.5684 1.8221 174.8 62736 t72 8 1135.8 0 2537 '
6 5656 0 016474 56 95 56 97 141.98 . 993 8 1836 6 0 2568 15616 1.81M IN.I 1748 54.61 143 99 992.6 178.8 6 8690 0016486 54 59 0 2600 1.5548 1.8147 175 8 145.99 991.4 1137.4 1.1840 0 016498 52.35 5236 i 17e s
1 1 .w ..
l i
i falhalpy Entropy i
Abs Press. Specific Volume Sat. Sat. Temp Sat. Sat. Sat.
Temp Lb per Sat.
Liquid Evan Vapor Liquid Evap. Vapor Fahr Liquid Evap Vapor I Fahr SqIn. h Ig hg 5, ;ig_sL v, vIg vg hI I p 990.2 1138 2 R2631 1.5400 1.811I . let8 50 21 50 22 148.00 182.0 IIe8 7.5110 0016510 150 01 989.0 1139 0 0.2662 1 5413 1.8075 7.850 0.016522 48.172 18.189 0.2694 1.5346 1 8040 184 0 1870 46 232 46.249 152 01 987.8 1139.8 lesa 184.8 8203 0 016534 154.02 906.5 1140.5 02725 1.5279 1.8004 44383 44.400 100 0 18E 8 . 8.568 0.016547 42.638 156.03 985.3 1141.3 0.2756 1.5213 13%9.
8.947 E016559 42.621 180.8 0.2787 1.5148 11934 198 8 40.957- 158 04 984.1 1142.1 1 340 0016572 40.941 02818 1.5082 13900 192.8 190 0 39.337 39.354 16&O5 982.8 1142.9 194.0 192 8 9 747 0.016585 37.824 162 05 981.6 !!43J 02848 1.5017 I.7065 10.168 0 016598 37.808 0.2879 1.4952 13831 195.0 194 8 36.364 164.06 980.4 1144.4 1& 605 0016611 36 348 0.2910 1.4888 1.7798 190.0 .
- Its 8 34.954 34.970 166 08 979.1 1145 2 190 8 11.058 0 016624 .
l 0 2940 1.4824 11764 290 8 31639 168.09 977.9 1146 0 11.526 0 016637 33.622 03001 14697 1 7698 204 0 1 280.0 31.151 172.11 975.4 11475 200 0 12 512 0016664 31.135 03061 1.4571 1 7632 I 204 8 28.878 176.14 9 72.8 1149 0 13.568 0 016691 28862 03121 1.4447 17568 212 0 2008 26399 180.17 9 70.3 1150.5 212 I 146% 0 016719 . 26182 967.8 1152.0 03181 1,4323 13505 216.0 0.016747 24 878 24.894 184.20 215 8 15.901 965.2 11514 03241 1.4201 17442 !!00 0.016775 23.131 23.148 188 23 13380 2248 ;
22t3 17.186 192.27 962.6 1154.9 03300 1.4001 18.556 0.016005 21.529 21.545 03359 13 %1 17320 220 0 2248 20.073 196 31 960 0 11563 232 0 20.015 0.016834 20.056 0 3417 13842 1.7260 '.
220 B 18318 200 35 957.4 1157.8 21 567 0.016864 18301 0 3476 ' 13725 IJ201 235 8 2328 17.471 204.40 954.8 1159 2 23.216 0.016895 17.454 238.8 2410 208.45 952.1 1160 6 03533 13609 11142 24.968 1016926 16 304 16321 03591 1.3494 17085 2440 ,
240 0 15.260 212.50 949.5 1162.0 26 326 0016958 15 243 03649 13379 11028 240 0 244 8 14.281 216 56 946.8 1163.4 240 s 28396 0 016990 14.264 944.1 1164 7 0 3706 13266 I6972 252.0 t 13.358 13.375 220 62 256 0 252 0 30 883 0.017022 224.69 941.4 1166.1 03763 13154 1 6917 33 091 0 017055 12.520 12.538 255.8 .
03819 13043 16562 260 0 '
.11.745 11.762 228 76 930.6 1867.4 264 0 2000 35.427 0.017009 11683 0 3816 1.2933 16808 11.025 11.042 232 83 935 9 268 0 ;
254 3 37.894 0017123 1870 0 03932 1.2823 1.6755 10 358 10375 236.91 933.1 '
250 8 40.500 0017157 117! 3 0.3987 12715 16702 2 72.0 9738 9355 240 99 930.3 275 0 $
212.0 43.249 0 017193 0.4043 1.2607 1.6650 9.162 1180 245 08 927.5 1172.5 275.8 46.147 0 017228 0 4098 1.2501 16599 200 0 8 627 8.644 249 17 924.6 Il738 204 0 r
200 0 49.200 0017264 253 3 9213 1175.0 0 4154 1.2395 16548 52 414 0 01730 8 1780 81453 0 4208 1.2290 1.6498 200 0 204 0 7.6807 2574 918 8 1176.2 200 8 55395 0 01734 7 6634 261.5 915 9 1177.4 0 4263 12186 I6449 291.0 i 59 350 0 01738 72301 L2475 0.4317, 1.2082 1 6400 296 0 292.8 0 01741 6 8259 6.8433 265 6 913.0 1178.6 295 0 63 084
- . - - . - . .. - -. .- . . - . ~ . . . ..
L .- -.
La i
Enthalpy Entropy Specific Volone . Sat. Temp ;
Abs Press. Sat. Sat.-
Sat. Sat. Evap Vapor Fahr e
tb per Sat. Liquid Evap Vapor Liquid Temp t i
Fahr Sqin. Liquid Evap Vapor h ig hg sg seg s, 4 Vig vg he I p vg 3ge8 11793 0.4372 11979 I6351 6.4658 269 7 919 0 16303 304 0 :
67.8 5 00l?45 64483 907.0 1880 9 04426 1.1877 '
3MS 60%5 61130 273 8 16256 300e 71.119 0 01749 904.0 1182.0 04479 1.1776 304.8 57655 51830 .278 0 16209 312 8 388 8 75.433 00l753 282 1 9010 1883.1 0 4533 1.1676
- 001757 5 4566 5.4742 0 4586 1.1576 16162 316 8 312.8 79 953 .286 3 897.9 1184.1 0 01761 5 1673 5 1849 316.O' M 688 320 0 894.8 1185 2 0 4640 1.1477 I.6116 4 8961 4.9138 290 4 324 8 3NS 39643 001766 391.6 1186 2 0.4692 1.1378 1.6071 4 6418 4 6595 294 6 320.0 3N.8 94826 001770 888 5 1187.2 0 4745 1.1200 1.6025 4 4030 4 4208 2987 11183 1.5981 -
332.0 328 8 100 245 00l??4 302 9 885.3 1188 2 04798 0.01779 4.1788 4.1966 0.4850 1.1086 1.5936 336 e 332.8 105 907 307.1 882.1 1189 1 0 01783 3 9681 3 9859 336 8 111 820 340 $ 'i 878 8 1190.1 0 4902 1.0990 1.5892 3 7699 31878 3 11.3 04%4 1.0894 1.5849 344e I17.992 0 01787 875.5 1891 0 340.8 3 5834 3 6013 315.5 1.0799 1.5806 348 8 124 430 0 01792 872.2 1191.1 05006 '
344 4 3 4078 3 4258 3191 1.0705 15763 352.0 131.142 001797 868 9 11923 0 5058 348 8 3 2423 3 2603 323 9 10611 1.5721 356 0 138 138 001001 865.5 1193 6 0$110 352.4 001806 30863 3.1044 328.I 3588 145 424 308.s 4 862.1 1194.4 0.5161 1.0517 1.5678 2 9392 2 M73 332.3 I.0424 I.5637' 364 s i 180.0 153 010 0 01811 358 6 1895 2 0 52I2 '
28002 2 8184 336 5 I0332 -1.5595 360 e 160 903 0 01816 855.1 IlM 9 0.5263 364.8 2 6813 340 8 3F2.0 169 113 00l821 2 6691 851 6 11961 0 5314 1.0240 15554 3ttI 2.5451 2.5633 345 0 1.0148 15513 375 8 373.0 177 648 00l826 349 3 848.1 1197.4 0 5365 001831 2 4279 2 4462 376 8 186 517 300 0 0.5414 1.0057 1.5473 2.3353 353 6 M4.5 1198 0 1.5432 384 8 1% 729 0 01836 2 3170 MOS 1198 1 0 5466 0 9966 308 8 2.2120 2.2304 3579 05516 0 9876 1.5392 388 8 344 8 205 294 001842 837.2 I199 3 2.1311 362 2 ' 392.8 3000 215.220 00lM7 2.1126 366 5 833.4 1199.9 0 5567 09186 1.5352 2.0lM 2.0369 0 5617 0 9696 1.5313 396.0 392 8 225 516 00l853 370 8 8291 1200.4 001858 1.929! 1.9477 396 8 236 193 400.8
! 0.5667 09607 1.5274 375.1 825.9 1201.0 484 s 001864 1.M44 1.8630 0.5717 0 9518 1.5234 488 8 247 259 379 4 822 0 1201.5 - 400.0 258125 001870 1.7640 1.7827 0 5766 0 9429 1.51M 404 0 383.8 818.2 1201.9 412.0 001875 1.6877 1.7064 0.5816 0.934I I.5157 .
488.8 270600 388 I 814.2 1202.4 elle 001881 16152 1.6340 0 5866 09253 1.5118 412.0 282 894 392.5 810.2 1202.8 0 0i887 1.5463 1.5651 416.0 295 617 470 8 0 5915 ' O.9165 1.5000 1.4997 3% 9 806 2 1203.1 424.0 001894 1.4808 05964 0 9077 1.5042 428.8 308130 401.3 802.2 1203.5 4288 001900 1.41M 1.4374 0 6014 0 8990 15004 424 0 322.391 1.3782 4057 7980 12033 1.4966 4320 ~
l 001906 1.3591 0 6063 0 8903 428 8 336 463 1.32179 4101 7939 1204 0 436 8 i
432 0 351 00 0 01913 1.30266 414 6 789 7 1204.2 0.61I? 08816 1.4928 001919 1.24887 1.26806
- 436 8 366 03 j 419 0 785 4 1204 4 0.6161 0 8729 I4890 440 8 0 01926 119761 1.21687 0 6210 0.8643 14853 4440 ,
440 8 381.54 423 5 781.1 12046 397 56 0 01933 1.14874 1.16806 0.6259 0 8557 14815 4484 444 8 4JLO 776 7 12047 1.12152 4534 414 09 001940 1.10212 772 3 1204 8 0 6308 08471 14778 '
44W 0 01947 1 95764 1077tl 4325 14741 essa 4 534 431 14 4370 767 8 I204 8 06 56 0 8385
- 442 H 0 01954 101518 I03472
1
)
i Enthalpy Entropy Abs Press. Specific Volume Sat. Sal. Sal. Sal. S.I. Temp Temp tb per Sat. Liquid Vapor Fahr l Evap vapor Liquid Evap Vapor Evan Fahr SqIn. Liquid I he h it hg sg sig sg I p v, veg vg 1204.8 0.6405 0 8299 1.4704 480.0 0 97463 0.99424 441.5 763.2 468 8 - 466 87 0 01961 0.6454 0.8213 1.4667 464.0 l 0 95557 4461 758 6 12043 454 8 48556 0 01 % 9 0 93588 1204.6 0.6502 0.8127 1.4629 est.I 504 83 0 01976 0.89885 0.91862 4503 754.0 472.0 45s 3 455 2 7493 1204.5 0.6551 0.8042 1.4592 524 67 0 01984 0 86345 0.88329 475.0 412 8 0.84950 459.9 744.5 1204.3 0.6599 0.7956 1.4555 415 g 545 11 001992 0 82958 0.6648 03871 1.4518 400.8 0 81717 464.5 739.6 1204.1
- 488 8 566 15 0 02000 0 19716 0.6696 0.7785 1.4481 484.0 0 02009 036613 0 78622 469.1 7343 1203 8
! 4848 587 81 473 8 7293 1203.5 0.6745 03700 1.4444 488.0 610 10 0 02017 033641 035658 492.0 4IB 8 032820 478.5 724 6 1203.1 0.6793 01614 1.4407 492 8 633.03 0 02026 0 70794 1202.7 0.6842 03528 1.4370 496.0 010100 483.2 719.5 i 496 8 656 61 0 02034 0.68065 0 67492 487.9 7143 1202.2 0.6890 03443 1.4333 500.0 500 0 680 86 0 02043 0.65448 709 0 12013 0.6939 03357 1.4296 584.0 4
0 02053 0.62938 0 64991 4923 500.0 584 g 705.18 0 62592 497.5 7033 1201.1 0.6987 03271 1.4258 50s e 731.40 0 02062 0 60530 1200.5 0.7036 03185 1.4221 512.0 0 60289 5023 698.2 512 s 75732 0 02072 0.58218 1199,8 0 7085 03099 1.4183 516.0 0.55997 0.58079 507.1 692.7 515 8 78436 0 02081 03133 0.7013 1.4146 529.I 0.55956 512.0 687.0 1199.0 520 8 812 53 0 02091 0 53864 03182 0.6926 1.4108 574.I 0 02102 0.51814 0.53916 516.9 6813 1198.2 528I l 574 5 841.04 521 8 675.5 1197.3 01231 0.6839 1.407C 870 31 0 02112 0 49843 051955 03280 0.6752 1.4032 512.0 J 578 8 526 8 669.6 1196 4 900 34 0 02123 0 47947 0 50070 03329 0.6665 13993 536.0 j 531 8 5313 653 6 1195.4 931.17 0 02134 0 46123 048257 535 8 03378 0.6577 1.1954 548.I .
046513 536.8 657.5 1194 3 5488 96239 0 02146 0 44367 6513 1193.1 0 7427 0.6489 13915 544.8 0 02157 0 42677 0.44834 541.8 5448 995 22 546 9 645.0 1191 9 03476 0.6400 1.3876 548.8 4
1028 49 0 02169 0 41048 0 43217 0.6311 13837 552.0 543 8 552 0 638.5 1190 6 0.7525 1062 59 0.02182 039479 0.41660 0.6222 1J797 556.8
! 552 0 557.2 632.0 1189.2 03575 1097.55 0 02194 0 37966 0 40160 '
- 5568
! 038714 562.4 6253 11873 0.7625 0.6132 13757 588.8 568 8 1133 38 0 02207 0 36507 618.5 1886.1 03674 0.6041 1.3716 564 8 0 35099 037320 5676 554 a 1170.10 0 02221 572 9 611.5 1184.5 03725 0.5950 13675 584 0 120732 0 02235 0 33741 035975 03775 0.5859 13634 572.5 568 8 578 3 604.5 11823 1246 26 0 02249 0 32429 0 34678 0.5766 1J592 576.8 512 8 5837 597.2 1180.9 03825 1285 74 0 02264 0 31162 033426 575 8 589I 589.9 '1179 0 03876 0.5673 IJ550 500.0 1326 17 0 02279 0 29937 0 32216 03927 0.5580 13507 584.0 5ss e 594 6 582.4 1176.9 13673 0 02295 0 28753 0 31048 0.5485 13464 588.8 584 8 6001 5743 1174 8 03978 1410 0 0 02311 027608 02991$ 0.8030 0 5390 1.3420 592.0 558 8 0 02328 026499 028821 6053 566.8 II72 6 i 597 0 1453 3 0 27770 611.4 558.8 1170.2 0 8082 0.5293 13375 596.0 596 0 1497.8 0 02345 0 25425
EnthalpY Entropy Abs Press. Specific Volume Sal Temp Sat. Sat. Sal. Sat.
Temp Lb per Sal. Vapor Liquid Evap Vapo'r Fahr
,i Evap Vapor Liquid Evap fahr SqIn. Liquid I i p v, vig vg he h gg hg sg sig st man a 550.6 11673 0.8134 0.5196 1.3330 0.26747 617.1 0.5097 1.3284 004.0 l seta 1543 2 0.02364 0.24384 622.9 542.2 1165.1 0.8187 1589 7 0 02382 0.23374 0.25757 0.8240 0.4997 13238 600.8 EN e 628.8 533 6 1162.4 16373 0 02402 0.22394 0.24796 0.8294 0.4896 1.3190 512.0 1 888 8 0.23865 634 8 5243 1159.5 816.8
$12 3 1686.1 0.02422 0 21442 515.6 1156 4 0.8348 0.4794 1.3141
- 0.22960 640.8 Elf E 17359 0 02444 0.20516 j R8403 0.4689 1.3092 820.0 012081 646.9 5063 1153.2 824.8 520 8 1786.9 0.02466 0.19615 406.6 1149.8 0.8458 0.4583 1.3041 0.21226 653.1 0.4474 1.2988 828.8
$24 8 1839 0 0 02489 0.18737 4863 1146.1 0.8514 0.20394 659.5 0.4364 1.2934 832.0 0 02514 0.17880
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Teenp Sat. Sat. Sat. Lb/Sg in.
Abs Press. Evap Vapor Liquid Evap Vapor Liquid Eva, Vapor Liquid p Lb/Sg in. Fahr hg s, s g, sg t vg vig vg hg hgg j p
! 2.1872 2 1872 000065 ;
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0 016407 13 515 .73532 130 20 13879 10 0 ;
50 162.24 982.1 1143.3 0.2836 15043 0 016592 38404 38.420 161.26 13560 14 096 193.21 10 0 970.3 1150 5 0.3121 1.4447 '
212.00 0016719 26 782 26399 180 17 13552 15.0 14 OIS 9693 1150.9 0.3137 1.4415 213 03 0016726 26.274 26.290 181.21 15.0 960 1 1156.3 03350 13962 13320 20 0 227.96 0016834 20 070 20 087 196.27 30 0 me 945.2 1164.1 0.3682 13313 16995~
250.34 0017009 133266 133436 218.9 1.6765 40 0 30 0 933.6 1169 8 0 3921 1.2844 267.25 0017151 10.4794 184965 236.1 1.6586 50 0 40 0 923 9 1I 74.I 0.4112 1.2474 281.02 0017274 8.4967 85140 250 2 1.6440 00 0 50 0 915.4 1177.5 0 4273 1.2167 29231 0 017383 7.1562 1.1736 262 2 1.6316 NO 00 0 -907.8 Il8R6 0.4411 1.1905 302.93 0 017482 6.1875 62050 2723 1.6200. 80 0 NO 2821 900.9 1183 1 04534 1.1675 312 04 &017573 5.4536 5.4711 1.1470 1.6113 00 0 00 0 2903 894.6 11853 0.4643 320 28 0 017659 4.8779 4 0953 908 84743 1.1284 1.6027 100.0 4.4133 4.4310 290.5 808 6 1187.2 100 0 327.82 0.017740 0.4834 1.1115 1.5950 110 0 4.0484 305.8 883.1 1188.9 lies 334.79 0 01782 4.0306 1190.4 0.4919 1.0960 15879 ING 33097 33275 312.6 877 8 1300 ING 341.27 0 01789 11913 0 4998 1.0015 1.5013 3.4364 3.4544 319 0 872 8 140 0 130 8 34733 0 017 %
868.0 1893.0 0.5071 1.0681 1.5752 353.04 001803 3 2010 3.2190 325 0 1.5695 150 0 140.0 358 43 801809 2.9958 3.0139 330 6 863.4 IIM.I 05141 1.0554 1.5641 100 0 t 150 0 859.0 1195.1 0 5206 1.0435
- 363 55 801815 2.8155 2.8336 336.1 15591 170 0 100 0 854.8 1896.0 0 5269 1.0322 3
368.42 001821 2 6556 2.6738 341 2 15543 100 0 171 3 8503 1196.9 0 5328 1.0215 i
373 08 401827 25129 25312 346.2 1.5498 IM O i 100 0 350.9 M63 1197.6 053M l.0113 37753 0 01833 2.3047 2.4030 i 1910 -
15454 200.0 301.80 EOl839 2.2609 2.2873 355.5 M2.8 Il9t3 a5438 1.0016 1.5413 210 0 l m00 2.18217 359.9 839.1 1I99 0 0.5490 0.9923 230 0 305.91 0 01844 2.16373 05540 R9834 1.5374 2N O 2.06779 2.00629 364.2 835.4 1199.6 2N O 309 08 001050 120RI 85588 09748 15336 230 0 1.97991 1.99M6 368 3 831.8 230 0 39330 001855 828 4 1200.6 0 5634 0 9665 1 5299 240 0 001060 1.89909 1.91169 3 72.3 250 0 240 0 39739 825 0 - 1201.1- 0.5679 0.9585 15264 400 97 001865 I82452 -1.84317 376.1 1.5230- 200 0 '
2500 821.6 1201.5 0.5722 ' 0 9500 001870 1 35548 137418 3 79.9 1.5197 270 0 2000 404 44 818.3 1201.9 0 5764 >09433 407.30 0 01875 169137 131013 383 6 1.5166 200 0 270 0 815.I 12023 0.5005 0 9361 .
411.07 001880 1.63169 1 65049 387.1 1.5135 290 0 2000 812.0 1202.6 0.5044 0 9291 414 25 001885 1 57597 1.59482 390 6 2910 05882 09223 1.5105 300 0 1.54274 3940 808.9 1202.9 300 0 41735 0 01009 I52384 1204.0 0 6059 0.8909 1.4968 350 0 ,
794 2 001912 1 30642 IJ2554 409 8 350 0 43133 424 2 700.4 1204.6 0 6217 0.8630 'I.4847 400 e 444 60 0 01934 1.14162 1.16095 400 0
-_ _ _ _ _ ~ - - - - - - - _ _ . . . - - - _ _ .
Enthalpy Entropy Specific Vehene Sat. Sat. Abs Press.
Sat. Sat. $~at.
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Liquid Evap Vapor Le' gen'd LblSq In. Fahr hg hgg hg sg sig sg . p !
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1204.8 06360 68374 1.4738 esee 101224 1.03179 4373 M7.5 50se i
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&7734 14461 Seee Sees 4713 732.0 12031 16723 486 20 0 02013 074962 076975 01552 14381 650 0 sese 170843 481.9 720.9 1202.8 0 6828 650 0 494 89 0 02032 068811 06928 03377 1.4304 700 8 491.6 710 2 1201.8 503 08 002050 0 63505 0 65556 700 0 07022 03210 1.4232 750 9 060949 500.9 - 699 8 12007 250 0 510 84 0 02069 150880 1199.4 03111 07051 I4163 300 e 054809 056896 509 8 689 6 IIIB 518 21 0 02007 1198 0 03197 0 6899 1.4096 050 s 0 51197 053302 SI5 4 679 5 geee !
850 0 525 24 0 02105 1196 4 01279 06753 1.4032 041968 0.50091 526 7 6691 '
900 0 531.95 0 02123 1194 1 01358 0 6612 1J970 350 s '
0 45064 0 47205 5343 660 0 legge 250 0 538 39 0 02141 1192.9 03434 06476 1.3910 0 42436 144596 542.6 650 4 ;
3008 8 544 50 002159 1191.0 a7507 06344 13851 1850 0 040047 842224 55a1 640 9 1850 8 55t53 002177 631.5 !!89.1 07578 06216 1.3794 1800 0 002195 &37863 040058 557.5 13738 1150 0 110B 0 556 28 038073 564 8 622.2 1187.0 tM47 06091 1150 0 56182 002214 0 35059 1184 8 0.7714 05%9 13683 . 1200 0 634013 036245 571.9 613 0 3200 8 567.19 002232 I!82.6 01780 05850 13630 1250 0 0 32306 034556 578 8 603 8 I3577 1300 e 12508 57238 002250 5946 1180 2 07843 05733 002M9 030722 132991 585 6 f.3525 1350 0 1380 0 577.42 585 4 1177.8 03906 0 % 20 i 002298 029250 031537 5923 13474 1400 8 13500 58232 576.5 11M 3 03966 0.5507 0 02307 027811 030178 598 8 1.3423 1450 0 lege 8 58707 567.4 1172.8 0 8026 0 5397 0 02321 026584 028911 605 3 I3373 1500 8 ,
1458 8 59130 558 4 1170.1 08085 0 5288 596 20 002346 025372 027719 6113 05182 I3324 1558 0 1500 0 024235 626601 6180 549 4 !!674 0 8142 13274 1600 0 1550 0 600 59 0 02366 1164 5 08199 05076 013159 025545 624 2 5403 1J225 1658 8 slag 8 604 87 0 02387 1161 6 08254 0.4971 002407 022143 624551 630 4 531J 1.3176 lies e 1658 8 609 05 522.2 1158 6 0 8309 E4867 613.13 0 02428 021178 0.23607 636.5 I700 8 1155 6 08363 0 4765 f.3I28 1758 8 0 20263 822713 642.5 513.1 13079 lesse 17500 617.12 002450 5038 11523 08417 0 4662 621 02 002412 0 19390 021861 648.5 13030 1850 0 lasta 021052 654 5 494 6 1I49 0 0 8470 0 4561 1858 8 624 83 002495 018558 1145 6 0 0522 0 4459 1.2981 1980 s EI1761 0 20278 660 4 485 2 11931 1958 8 1980 0 629 56 002517 475 8 1142.0 0 8574 0 4358 .
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~
1.- PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 20 THEMMUUTNAMIC5, HEAT TRAN5ttK AND FLUID FLOW e
ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY,I.
lWSTER CC)PY ANSWER 1.01 (2.50)
- a. In the secondary system there is a phase change. (0.50)
A phase change requires a large delta h.(Description of why phase (0.50) change is significant.)
With the larger delta h of the secondary, the same heat can be transferred with a lower flow rate. (Relate phase change to lower (0.50) flow rate.)
. b. 1. (1.00)
REFERENCE Westinghouse T.H Principles, 7-47 through 51, 12-12 3.2/000/002/K5.01/3.1 COMPONENT-HX and COND/2.0 ANSWER 1.02 (1.00)
To determine flow in NC:
Q = E cp delta-T => 100 = 100
- cp
- 60 .
Q1=h1cpdelta-T=> 2=El*cp*28 100 100
- cp
- 60 2 ml
- cp
- 28 (0.50 for proper method) 214.3 50 = ---------
ml
$1 = 4.29 percent (4.0 4.5 acceptable) (0.50)
REFERENCE Westinghouse T-H Principles, 4-65 3.2/020/002/K5.01/3.1
=' .
. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 21
- THERMODYNAMIC 5, HEAT TRAN5PtR AND FLUID FLDW ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 1.03 (2.00)
- a. INCREASE (0.50 each)
- b. INCREASE
- c. REMAIN THE SAME w,e INCREASE u q;; ;; ,,;7 :a
,o a.c -- - - -
- d. DECREASE REFERENCE Westinghouse T-H Principles, 13-23 3.9/020/015/K5.09/3.5/3.7 ANSWER 1.04 (1.00)
Bank D at 140 steps. (130 to 150)
REFERENCE C1-A, Reactivity Calculations, Figure C1-4A, Page 1 of 2 3.1/001/000/K5.05/3.5/3.9 ANSWER 1.05 (2.00)
- a. False.(0.50) Raising RCS pressure reduces pump flow rate which (reduces head loss and) increases fluid pressure at suction of pump. .
(NPSH = Psuct - Psat) (0.50)
- b. False.(0.50) Lowering fluid temperature lowers saturation pressure of fluid at pump suction. (NPSH = Psuct - Psat) (0.50)
REFERENCE Westinghouse T-H Principles, 10.56,60 COMPONENT / PUMPS-CENTRIFUGAL /2.9/3.2
. 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 22 THERMUDYNAMIC5, HEAT TRAN5PLR AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 142 -87/03/23-KINGSLEY,I.
ANSWER 1.06 (2.50)
- a. Rhol = -2.5% delta-k/k Keffl = 1/(1-rho) = 1/(1-( .025)) = 0.9756 (0.50)
CR1/CR2 = (1-Keff2)/(1-Keff1) 115/345 = 1/3 = (1-Keff2)/(1 .9756) (0.50)
Keff2 = 0.992 Rho 2 = -0.806% delta-k/k Reactivity added = -0.806% - (-2.5%) =1.69% del ta-k/k (0.50)
(1.64 to 1.74 acceptable)
- b. More neutron generations will be required for the neutron level to reach equilibrium. (1.00)
REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, Chapter 8 - 51 3.9/015/000/K5.06/3.4/3.7 ANSWER 1.07 (1.00) a REFERENCE Westinghouse Reactor Core Control, 9-10 3.9/015/000/KS.06/3.4/3.7 .
ANSWER 1.08 (2.50)
- a. 1. Void coefficient (0.4 each, 0.3 for correct order)
- 2. Moderator temperature coefficient
- 3. Doppler power (or fuel temperature) coefficient
REFERENCE Westinghouse Reactor Core Control, Chapter 3 - 42 3.1/000/001/K5.49/3.4/3.7
- 1. PRINr.IPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 23 THERMODYNAMIC 5, HEAT TRAN5FER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY,I.
ANSWER 1.09 (1.00) c REFERENCE Steam tables COMPONENT-HX AND CONDENSERS /2.9/3.0 ANSWER 1.10 (2.00)
- a. Saturated B. Superheated
- c. Superheated
- d. Subcooled (0.50 each)
REFERENCE Steam Tables COMPONENT-HX AND CONDENSERS /2.9/3.0 ANSWER 1.11 (1.50)
- 2. Neutron flux shif ts within the core. (to outer edges) (0.50)
- 3. Total neutron flux increases. (compensates for fuel depletion) (0.50)
REFERENCE Westinghouse Reactor Core Control, Chapters 5 and 8 3.9/015/000/K5.04/2.6/3.1 3.9/015/000/A1.03/3.7/3.7
- 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, PAGE 24 THERMODYNAMIC 5,efEAT TRANSFER AND FLUID FLOW ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I. ,
ANSWER 1.12 (3.00)
- a. Unit 2 (0.50) due to lower competition (or less spectrum hardening).
(0.50)
- b. Unit 2 (0.50) due to r boron concentration (or 1 r thermal utilization). (0.50)
- c. Unit 1 (0.50) due to less negative MTC. (0.50) .
REFERENCE Westinghouse Reactor Core Control. 2-46, 3-18, and 5-16 3.1/000/001/KS.47/2.9/3.4 3.1/000/001/K5.49/3.4/3.7 ANSWER 1.13 (1.00) d.
REFERENCE TS 3.10-1 PWG-5/2.9/3.9 ANSWER 1.14 (2.00)
- a. Withdrawing control rods causes MTC to become more positive. (0.50)
Core size is effectively increased which reduces neutron leakage. With less initial leakage, any temperature change will result in a smaller reactivity change. (0.50)
- b. At higher temperature the rate of moderator density change becomes larger (0.50) which causes HTC to become more negative. (0.50)
(Alternate explanations may be accepted.)
REFERENCE Westinghouse Reactor Core Control, 3-16 and 20 3.1/001/000/KS.49/3.4/3.7 i
. - . ___.m,.. --_,____ -,_ r- _ _ . _ _ _ . . _ - - _ - - _ - . _ , - - . _ . - . - _ , . - . , . _ . _-
qq I 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 25
. ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY,I.
ANSWER 2.01 (3.00)
- a. 1. S/G low-low level on 1 steam generator (of associated unit)
, 2. Safety Injection Signal (from associated ESF train)
(breakers open)
- 3. Trip of both Main Feed Pumps )
- 4. UV on buses 11 and 12 (0.25 each for a total of 1.00)
I b. CST - normal (0.50)
Cooling Water System - backup (0.50)
- c. Manually, (0.50) when CST Lo-lo Level alarm setpoint is reached. (or at 5 to 6 feet CST level) (0.50)
REFERENCE SD 8-288, p. 4,11 3.2/000/013/K4.04/4.3/4.5 3.5/000/061/K4.02/4.5/4.6 3.5/000/061/K1.07/3.6/3.8 ANSWER 2.02 (3.00)
- a. 1. SIS at either unit. (0.50)
)
b.
2.
- 1. UY Ay.',;sib nal from m/.t.'. S.deitherOsa Arnerc
- ';xt ;rr r t J 0.50) feguards(A. bus. (0.50)upAkulyoseder gggp7gr ess
\
- 2. Emergency diesel gendTtor breakers are open. (0.50)
- c. 2 and 5 (0.50 for each) 7..
f,g g r .. ..
- - ~ . --gk .,
SD 8-38A, p. 8,9,12 3.7/000/064/K4.02/3.9/4.2 3.7/000/064/A3.01/4.1/4.0 3.7/000/064/A4.01/4.0/4.3 1
..-,-,,.,w.wwen ,a,,-,e,,, c- w---,nn,, ,,
PAGE 26 i !
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS I
. ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 2.03 (2.50) ,
- s. 1. Extends flow coastdown ( following loss of power to an RCP (O.
- 2. Minimizes starting current (or core bypass flow). (0.50)
- 3. Cools reactor coolant supplying seals upon loss of normal seal injection. (0.50)
- b. 1. The associated oil lift pump must be running. (0.50)
- 2. Oil lift pump discharge pressure must be > 350 psig. (0.50)
REFERENCE SD 8-3, p. 6,18,21,24 3.4/000/003/K4.03/2.5/2.8 3.4/000/003/K4.05/2.3/2.7 3.4/000/003/K4.10/2.3/2.5 ANSWER 2.04 (2.00)
- a. Containment ventilation isolation. (Ctmt and In-service purge isol.)
- b. None.
- c. Stops discharge (to river). (0.25 for each correct answer) '
- d. Stops blowdown.
- e. Containment ventilation isolation. (Ctat and In-service purge isol.)
- f. Starts cleanup fan (0.125), shif ts ventilation to Rectre Mode. (0.125)
- g. Shif ts to auxiliary building special ventilation.
- h. Shuts vent valve on surge tank.
REFERENCE
-SD B-11. Table B11-1 3.11/000/029/K1.02/3.4 071/K1.06/3.1 l 068/K1.10/2.5 3.10/000/008/K1.03/2.8 A2.04/3.1 3.5/000/039/K1.09/2.7 076/K1.17/3.6 l
l ,
l.
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 27
~
ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
l l
l ANSWER 2.05 (1.50) 8.
d.
- e. (0.50 each)
REFERENCE Prarie Island 4.16 KU Bus Loads, 10/09/86 3.7/000/062/K3.01/3.5/3.9 ANSWER 2.06 (1.50)
- 1. Flow through control rod guide thimbles.
- 2. Flow from inlet plenum directly into outlet nozzle.
- 3. Flow through baffle (and former) wall plates.
- 4. Flow deflected into vessel head. (any three at 0.50 each)
REFERENCE SD B-4A, p. 6 3.2/0000/002/K6.13/2.3/2.8 ANSWER 2.07 (2.50) .
- a. 1. Valve cannot be opened unless RCS pressure (0.25) < 425 psig. (0.25)
- 2. Valve will close automatically if RCS pressure increases to (0.25) 600 psig. (0.25)
- b. 1. Shut
- 2. Open
- 3. Shut (0.25 for each)
- 4. Shut
- 5. Open
- 6. Open REFERENCE SD B-15, p. 15,16, Figure B15-4 3.2/000/006/K4.08/3.2/3.5 3.2/030/006/K4.03/3.4/3.6
- 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS PAGE 28 ANSWERS -- PRAIRIE ISLAND 182
-87/03/23-KINGSLEY,I.
ANSWER 2.08 (2.50)
- a. Fail open
- b. Diverts to VCT (0.50 for each)
- c. Fail shut
- d. Remain functional (provided with backup air accumulators)
- e. Remain functional (motor operated)
REFERENCE SD B-4A, p.17 SD B-12, Figure B12A-2 SD B-18B, Figure B188-2 SD B-27, p. 5 3.8/000/078/k3.02/3.4/3.6 l
ANSWER 2.09 (1.50)
- 2. Ensures RCS flow is maintained temporarily in event RCPs are deenergized. (Minimizes reduction of DNBR.) (0.50) l 3. Prevents turbine overspeed. (Ensures turbine is tripped before unloading generator.) (0.50) l .
REFERENCE PI Exam Bank, Section 3, #2 3.5/050/045/K1.01/3.4/3.6 l
- 2. PLANT DESIGN INCLUDING SAFETY'AND EE RGENCY SYSTEMS PAGE 29 ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 2.10 (2.50)-
- a. 1. Increasing CC surge tank level. (0.50)
- 2. Increasing CC radiation levels. (0.50)
- b. 1. RHR HXers
- 2. RCP thermal barrier HXers
- 3. Lettown HXer
- 4. Excess letdown HXer
- 5. Seal water return HXer (any 6 at 0.25 each) ,
- 6. SG blowdown HXer
- 7. Feedwater sample HXer
- 8. RCS sample HXers
. 9. Spent fuel pit HXer REFERENCE PI Exam Bank, Section 3, #16 3.10/000/008/K1.04/3.3/3.3 ANSWER 2.11 (1.50)
- 1. Pressurizer level (0.25) > 14.8% (0.25)
- 2. Both letdown isolation valves open (0.50)
- 3. At least one charging pump running (0.50) l REFERENCE I SD B-12, p. 8 3.1/020/004/K4.03/3.0/3.4 ANSWER 2.12 (1.00)
- 1. Reduces thermal shock to spray nozzle. (0.50)
Maintains uniform water chemistry in pressurizer. (0.50) l 2.
REFERENCE SD B-4A, p. 30 3.3/000/010/K4.01/2.7/2.9
. . . - . , . ._ .,--_,...,m., _ _ , , _ . , . _ _ _ _ . _ ~ , .,_,,..,,_.,_.v,,_,_, . ~ , . , _ _ , -
_,.,___..-,._,,.,_,-m.. .y.., ,-
3.' INSTRUMENTS AND CONTROLS PAGE 30 t:45WERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 3.01 -(3.00)
- a. 1. Power range high flux (0.20) 105% (0.20) 1/4 (0.10)
- 2. Intermediate range high flux (0.20) 20% (0.20) 1/2 (0.10)
- 3. OT Delta-T (0.20) Setpoint - 5% (0.20) 2/4 (0.10)
- 4. OP Delta-T (0.20) Setpoint - 5% (0.20) 2/4 (0.10)
- b. Urgent alarm stops all automatic and manual rod motion. (0.50)
Non-urgent alarm has no operational effect. (Indicates failure of redundant feature) (0.50)
REFERENCE SD B-5, p. 15,16,23 3.1/000/001/K4.07/3.7/3.8 3.1/050/001/A2.01/3.7/3.9 ANSWER 3.02 (3.50)
- a. Channel III Thot failed high. (1.00)
- b. 1. Automatic rod insertion. (0.50)
- 2. Charging pump speed increases. (or charging flow increases) (0.50)
- c. 1. Tave Defeat Switch allows removal of I channel of Tave from use in Tave auctioneering circuit (0.25) which provides input to:
l (a) Rod Control System (0.25)
(b) Steam Dump System (0.25) l (c) Pressurizer Level Control System (0.25)
- 2. Delta-T Defeat Switch allows removal of I channel of Delta-T from providing input to:
(a) Rod Insertion alarms (RIL computer) (0.25)
(b) Delta-T Deviation alarms (0.25)
REFERENCE OP C-7, p.18,19,20 3.9/000/016/K1.01/3.4/3.4 3.9/000/016/K4.03/2.8/2.8 l 3.9/000/016/A2.01/3.0/3.1 3.2/020/002/K5.09/3.6/3.9 3.2/000/002/A1.03/3.7/3.8 i . .. .. . _. .. . . _ - _ - . .
- 3. INSTRUENTS AND CONTROLS ~PAGE 31 ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 3.03 (1.50)
- a. 540 F (0.50)
- b. When Tave decreases to the low-low Tave setpoint (540 F) all steam dump valves will close (0.50). The affected steam dump valves will open when decay heat and/or pump heat increases Tave above 540 F and close again at 540 F. (0.50)
REFERENCE SD B-7, p. 46 Westinghouse Logics, Sheet 11 3.5/020/041/K4.09/3.0/3.3 ANSWER 3.04 (2.50)
- a. To return control of safeguards equipment to the operator. (0.50)
- b. 1. Reactor trip (P-4) (0.50)
- 2. (90 second) time delay has timed out. (0.50)
- c. 1. Low pressurizer pressure; (0.25) when RCS < 2000 psig (0.25)
REFERENCE SD B-18C, p. 16,17,29,30 -
3.9/000/012/K6.10/3.3/3.5 3.9/000/012/A4.04/3.3/3.3 ANSWER 3.05 (2.00)
- a. Undercosipensation results in a higher than actual reading, (0.50) and if > 10E-10 amps will prevent the SR detectors from automatically energizing (0.50).
- b. The operator must manually energize the SR detectors with the Source Range Manual Reset switches (0.50) when the operable IR channel drops below the P-6 setpoint(0.50).
REFERENCE Westinghouse Logics, Sheet 3 SD B-9 .
3.9/000/015/K6.02/2.6/2.9 3.9/000/015/A2.02/3.1/3.5
- 3. INSTRUMENTS AND CONTROLS PAGE 32 ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 3.06 (1.50)
- a. POWER
- b. SOURCE, INTERMEDIATE and POWER
- c. SOURCE and INTERMEDIATE (0.25 each)
REFERENCE SD B-9 3.9/000/015/K5.02/2.7/2.9 3.9/000/015/K5.03/2.3/2.6 3.9/000/015/Kl.03/3.1/3.1 ANSWER 3.07 (1.50)
Power range high flux - high setpoint Power range high flux rate - positive Power range high flux rate - negative OT Delta-T
- Low pressurizer pressure RCS low flow \ L.eop (0.25 each) 6ee74N Ac5 Lad kow 2.\eops REFERENCE SD B-8, Table B-8-1 3.9/000/012/K4.02/3.9/4.3 ANSWER 3.08 (1.50)
- 1. RV Dynamic Head Indication (0.25), 100% to 108% (0.25) # de A A AA g
- 2. RV Upper Range Indication (0.25), off scale low (0.25) ? ^
- 3. RV Full Range Indication (0.25), off scale i
- REFERENCE (0.25) J C"A KCP 0^" " . " ' '
SD B-48, p. 10 3.9/000/016/K4.02/2.3/2.7 lod 3.9/000/016/A3.02/2.9/2.9 ANSWER 3.09 (2.00)
- a. TRUE (1.00)
- b. TRUE (1.00) t
, ;~ 3. INSTRUMENTS'AND CONTROLS PAGE 33
- ANSWERS'-- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
REFERENCE Westinghouse logics, sheet 12 3.3/000/010/K4.03/3.8/4.1 ANSWER 3.10 (1.00)
Pressurizer high level deviation (0.50) at 10% > program level (0.50).
REFERENCE Westinghouse logics, sheet 13 3.2/000/011/K5.13/3.2/3.4 ANSWER 3.11 (1.50)
- a. IRPI (0.50)
- b. BDPI (0.50)
- c. BDPI (0.50)
REFERENCE SO B-6, p. 4,5,6 3.1/000/001/K4.01/3.5/3.8 ANSWER 3.12 (2.00)
Yes (1.00) Resetting SI removes safeguards racks from the Test mode (0.50) and the two steam pressure low pressure SI signals still remain. (0.50)
REFERENCE P-RE-2-86-3, p. 1,2 -
3.9/000/012/K4.01/3.7/4.0 3.9/000/012/K4.02/3.9/4.3 3.9/000/012/K4.08/2.8/3.3 3.2/000/013/K4.10/3.3/3.7 l
ANSWER 3.13 (1.50) l (0.50 each) l
- 1. High radiation from either R-11, R-12, or R-22
- 2. Operation of 1 out of 2 manual containment isolation switches.
- 3. Operation of 2 out of 2 manual containment spray switches.
l l
i
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PAGE 34
,. 3. INSTRUMENTS AND CONTROLS ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
REFERENCE Prairie Island System Description B-18C, p.12 3.6/000/103/K4.06/3.1/3.7
-s
- 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND PAGE 35
- RADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 4.01 (2.00)
- a. At least one Safety Injection Pump running, (p.50) and RCS pressure
< 1200 psig. (0.50) g < fy ppf ffy g Mgj
- b. 200 F (0.50) radial bearing temperature. (0.50)
REFERENCE E-0.0, Information Page E-12, p. 3 3.4/000/003/A2.02/3.7/3.4 3.4/000/003/PWG/10/4.1/4.4 ANSWER 4.02 (2.00)
- a. To minimize RCS inventory loss (via the RCP seals). (Also accept, to cause accumulator injection) (0.50)
- b. To prevent injection of accumulator nitrogen into the RCS. (0.50)
- c. To prevent possibility of recriticality. (0.50) l l d. Continue depressurizing. (0.50) l '
l REFERENCE ECA-0.0, p. 7 3.7/055/000/EK3.02/4.3/4.6 3.7/055/000/PWG/7/3.5/4.0 l
ANSWER 4.03 (2.00)
- a. 60 F/hr
- b. 100 F/hr
- c. 50 ppm
- d. 66*F (0.50 each)
J20 REFERENCE C1.2, p. 3,5 PWG/7/3.5/4.0
- - . - - - - - - - - - ,-r .- ,, .. -- .-r - -
,. 4. PROCEDURES NORMAL, ABNORMAL, EMERGENCY AND PAGE 36 RADIDLDGICAL CONTROL
- ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 4.04 (1.50)
- a. False
- b. True (0.50 each)
- c. True REFERENCE i AWI, 3.10.1, p. 2,5,6,7 PWG/13/3.7/4.0 ANSWER 4.05 -(1.00)
The Intermittent alarm can reduce the operators' awareness of other plar.t alarms or indications.
REFERENCE SWI-0-14, p. 1 PWG/9/3.8/4.0 ANSWER 4.06 (1.00)
~
Only when under the direction of and in the presence of a licensed operator (0.50) for the purpose of training for an R0/SR0 license. (0.50) .
REFERENCE SWI-0-2. p. 7 10CFR50.54 10CFR55.9 PWG/17/2.9/3.5 PWG/31/3.1/4.7 ANSWER 4.07 (1.00) 2 REFERENCE F-0, Background Information PWG/10/4.1/4.5
- 4. PROCEDURES NORMAL, ABNORMAL, EMERGENCY AND PAGE 37 RADIOLDGICAL CONTROL s
ANSWERS -- PRAIRIE ISLAND 182. -87/03/23-KINGSLEY, I.
ANSWER 4.08 (2.50)
- a. To prevent plugging seal injection throttle valves, or to prevent clogging seal injection filters, or to prevent loss of seal injection. (any one at 0.50)
- b. 1. Shift running BATP to fast (0.25)
- 2. pen nservic BAST recirc valve 50%. (0.25)
- 3. pen V-32086 Emergency Borate Valve. (0.50)
- c. 1. Reactor is subcritical, and (0.50)
- 2. Any uncontrolled cooldown is stopped. (0.50)
REFERENCE C-12, TM-86-60, p. 2 FR-S.1 3.1/024/000/EK3.01/4.1 3.1/024/000/EK3.02/4.2 PWG/11/4.3/4.4 ANSWER 4.09 (1.00) d.
REFERENCE ES-0.1, Attachment A 3.7/055/000/EK1.02/4.1/4.4 ANSWER 4.10 (2.00)
- 1. All reactor trip breakers indicate open.
- 2. Rod bottom lights lit.
- 3. Rod position indicators at zero.
- 4. Neutron flux decreasing. (0.50 each)
REFERENCE E-0.0, p. 3 PWG/11/4.3/4.4
- m ,-
, , 4. PROCEDURES NORMAL, ABNORMAL, EMERGENCY AND PAGE 38 RADIOLOGIGAL SUNTROL
. ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 4.11 (1.50)
SS - Proceed to the Hot Shutdown Panel - Auxiliary Feedwater pump room (0.25) and provide necessary assistance. (0.25)
LPE 8 RO - Proceed to Train B Remote Shutdown Panel (0.25) and prepare for control transfer. (0.25)
PE & R0 - Proceed to Unit 2 Rod Drive Control Room (0.25) and manually trip Unit 2 reactor. (0.25)
REFERENCE C1.8, p. 3 PWG/11/4.3/4.4 ANSWER 4.12 (2.00)
- a. 1. 1250 mrem
- 2. 3000 mrem
- 3. 5000 mrem (0.50 each)
- b. Superintendent of Radiation Protection (0.50)
REFERENCE ,
F-2, p. 11,12 PWG/16/3.4 ANSWER 4.13 (1.00)
- a. 5 (0.50)
! b. 10exp4 (0.50) l l REFERENCE l E-0.0, Information Page PWG/10/4.1/4.5 l
l I
(
1 I
p' .
PAGE 39
,, 4. PROCEDURES NORMAL, ABNORMAL, EMERGENCY AND
[g RADIOLDEICAL CONTROL i ANSWERS -- PRAIRIE ISLAND 182 -87/03/23-KINGSLEY, I.
ANSWER 4.14 (2.00)
- 1. No (0.25) Two operable channels exceed limits (0.25)
- 2. Yes (0.25) Less than two operable channels exceed limits (0.25)
- 3. No (0.25) Two channels exceed limits (0.25)
- 4. Yes (0.25) Less than two operable channels exceed limits (0.25)
REFERENCE PI TECHNICAL SPECIFICATIONS, p. TS.3.10-3 3.1/000/001/A3.03/3.6/3.8 3.1/000/001/A2.06/3.6/4.0 PWG/8/3.5/4.5 ANSWER 4.15 (1.00)
- 1. Isolation valves must be open.
- 2. Must contain 1270 +/- 20 cubic feet of borated water.
- 3. Minimum boron concentration of 1900 ppm.
- 4. Nitrogen cover gas pressure >/= 700 psig. (0.25 each)
REFERENCE '
PI TECHNICAL SPECIFICATIONS, p. TS.3.3-1 PWG/8/3.5/4.5 MSWER 4.16 (1.50)
- 1. Ensures adequate trip reactivity. (0.50 each)
- 2. Ensures meeting power distribution limits.
- 3. Limits the consequences of a rod ejection accident.
REFERENCE PI TECHNICAL SPECIFICATIONS, p. TS.3.10-15 PWG/5/2.9/3.9
,- i
>e -
L U. S. NUCLEAR REGULATORY COMMISSION
<- SENIOR REACTOR OPERATCR LICENSE EXAMINATION FACILITY: PRAIRIE ISLAND 1&2 REACTOR TYPE: PWR-WEC2 DATE ADMINISTERED: 87/03/23 EXAMINER: HARE. S.
CANDIDATE:
INSTRUCTIONS TO CANDIDATE: k 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 crade 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 _YALUE CATEGORY ws 25.00 25.DA 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS 96 25.00 25.OR 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRUMENTATION 95 -
25.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL g.
! M.00 07 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS 99 M. 00 % Totals Final Grade t All work done on this examination is my own. I have neither given nor received aid, i
Candidate's Signature I
fMSIER COPY
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 4 room to avoid even the appearance or possibility of cheating.
- 3. Use black ink or dark pencil 2nly to facilitate legible reproductions.
- 4. Print your name in the blank provide $ 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. P,rint 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 B2w Page, write onlY 2n QR2 Aide 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 l QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
- 16. If parts of the examination are not clear as to intent, ask questions of the REamiDar 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.
-w , .~,,,.r-- - - - - - - - , , , ,-- . , , - - . . - , _ _ _ - - - . . - - . , , . . , , - - _ - - . , . . . - - - .
- < - - ~- -
- 18. When you complete your examination, you shall:
4
- 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.
i 9
.-- . . - , _ . _ _ , _ -- , , ,,. - - - n- ..
PAGE 2
,5. THEQRY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND IHERMODYNAMLQE 4
. QUESTION 5.01 (2.00)
Concerning total power defect:
- a. List the three (3) defects that the total power defect is comprised of. (1.5)
- b. What is the largest contributor to total power defect at BOL? (.25)
- c. What is the largest contributor to total power defect at EOL7 (.25)
QUESTION 5.02 (1.50)
State whether SDM will initially INCREASE, DECREASE, or REMAIN THE SAME for each of the below listed events. Assume reactor is at 20%
power. (Consider each case separately)
- 1. Boron dilution of 20 ppm
- 2. Pull rods out in the controlling bank 10 steps
- 3. The reactor trips QUESTION 5.03 (2.00)
The following readings were taken off the power range detectors:
-(Note: Power Range instrument N43 is out of service.)
N41 N42 N43 N44 DET A (Upper) 374.4 356.6 0 360.0 DET B (Lower) 324.0 342.0 0 360.0 All readings are in microamperes. Full power current on all detectors ds known to be 400.0 microamperes,
- a. Compute the QPTR using the given data. (Show all work) (1.5)
- b. What is the Tech Spec limit for QPTR? (0.5)
QUESTION 5.04 (1.50)
Two identical reac.>rs are taken critical. Reactor A has a rod speed of 50 steps per minute, and Reactor B has a rod speed of 25 steps per minute.
(Assume continuous rod withdrawal)
- a. Which reactor will go critical first? (Briefly explain)
(.50)
- b. Which reactor will have the highest SR counts at criticality? (Briefly explain) (.50)
- c. Reactor A will have (SAME, HIGHER, LOWER) critical rod height than Reactor B. (Briefly explain) (.50)
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
- 5. THEORY OF NUCLEAR POWER PLANT OEEB& TION. FLUIDS. AHD PAGE 3 THERMODYNAMICS e
QUESTION 5.05 (1.00)
List two of three conditions necessary for brittle fracture of a carbon steel pressure vessel to occur.
QUESTION 5.06 (2.00)
An estimated critical-position (ECP) has been calculated for a reactor startup that is to be performed 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after a trip from a 60-day full power run. How would each of the following events or conditions (independently) affect the actual critical rod position compared to the ECP? In your answer, state whether the actual rod position would be:
HIGHER THAN ECP, LOWER THAN ECP, or NO SIGNIFICANT DIFFERENCE.
- a. The startup is performed approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> early. (0.5)
- b. Condenser vacuum decreases from 28" Hg to 24" Hg. (0.5)
- c. The steam dump pressure setpoint is increased by 100 psi above the no-load setpoint. (0.5)
- d. All steam generator levels were raised by 5% just prior to criticality. (0.5)
-QUESTION 5.07 (1.50)
- a. With RCS Tavg equal to 290 degrees Fahrenheit and a bubble in the par, what is the maximum allowable RCS pressure in psig which does not violate the 320 degree Fahrenheit differential temperature limitation between the RCS and the pressurizer? (.75) -
, b. What steam generator pressure must be established to ensure a 50
! degree Fahrenheit subcooling margin when RCS pressure is 1000 psig?
! (.75)
QUESTION 5.08 (2.50)
- a. State the TWO methods of xenon production, and the TWO methods of removal. (1.0)
- b. State the method of Samarium production and removal. (0.5)
! c. Compare Xenon and Samarium in regard to their variation in concentration following a power reduction from 100% to 50% and remaining at 50% for two weeks. (1,0) l l
l i (***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
L_
. 5.'* THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 4 THERMODYNAMICS
~
QUESTION 5.09 (2.00)
During refueling operations, you, as the SRO, are given the following information:
Boron sample on the refueling canal was 2027 ppm and the total reactivity in the reactor vessel is -4570 pcm after the last fuel bundle was loaded.
Give a brief explanation and reasons as to why the fuel movement should or should not continue. (Show all work.)
QUESTION 5.10 (1.50)
For each of the below listed parameter changes, state whether DNBR will INCREASE, DECREASE, c.r REMAIN THE SAME. (Consider each case separately.)
- a. Primary system pressure decreases 200 psig,
- b. Primary system Tave decreases 15 degrees F.
- c. Reactor power increases from 50% to 954.
QUESTION 5.11 -(1.00)
During a reactor startup, the first reactivity addition caused the count rate to increase from 20 to 40 cps. The second reactivity addition caused the count rate to increase from 40 to 80 cps. Which of the following answers is correct? (There is only one correct answer)
- a. The first and second reactivity additions were equal. -
- b. The first reactivity addition was larger.
- c. The second reactivity addition was larger.
- d. There is not enough data given to determine.
QUESTION 5.12 (1.00)
Which one of the following DOES NOT occur over core life, i.e, from BOL to EOL7
- a. Axial power distribution flattens
- b. Effective fuel temperature decreases
- c. MTC becomes less negative
- d. Reactor response time decreases
(***** CATEGORY 05 CONTINUED ON NEXT PAGE *****)
& .'5.* THEORY'OF'NUCM AR POWER PLANT OPERATION. FLUIDS. AND PAGE 5 THERMODYNAMICS t, -
QUESTION 5.13 . (1.00):
The calorimetric required at 100% power was just completed as per SP 1005,
- :and the Power Range nuclear instruments were adjusted accordingly. Upon review'of the calorimetric-paperwork, you realize that the. ACTUAL feedwater temperature was 15 degrees F greater than the feedwater temperature used to calculate the calorimetric.
In reference to the above, is INDICATED POWER reading GREATER THAN , LESS
'THAN, or EQUAL to ACTUAL POWER?
j' QUESTION 5.14 .(1.50)
- TRUE or. FALSE.
- a. . As Keff. approaches _ unity, a smaller-change in neutron level will result from identical changes-in.Keff. (0.75) i b. With Keff greater than unity, a constant positive startup rate with ,.
increasing neutron level will occur only if net REACTIVITY is
? NOT changing. (0.75) i- QUESTION ~ 5.15 (1.00)
I, The reactor is suberitical with Bank C at 46-steps. An ECP has just been
. calculated that shows 1300 pcm are needed to reach criticality. Use the
- appropriate Rod Worth Curve to determine the required bank position.
1 Assume no change in boron concentration, xenon, or RCS temperature.
i QUESTION 5.16 (2.00)
- State whether the following statements regarding the Net Positive Suction Head (NPSH) available to the RHR pumps during the LOW HEAD RECIRCULATION phase following a LOCA are TRUE or FALSE and EXPLAIN your answer.
S
- a. Raising RCS pressure from 50 psig (current) to 100 psig will decrease
- b. As containment sump fluid temperature decreases, available NPSH to the RHR pumps decreases. (1.0) 4 4
i l
l (***** END OF CATEGORY 05 *****)
i
- .er *m.- - ,,.r. .rw, - . - , , - ~ . -w.-.,-r - - - , . - - ,.ee,, - -w-we ,..c,- ,,c-- e,y-- .-,,.----r+-er,,-e -.n- w.-e,-- - - . - * -w. - - -
6.' PLANT SYSTEMS DESfGN. CQNTROL. AND INSTRUMENTATION PAGE 6 QUESTION 6.01 (1.50)
List the THREE REASONS for the 30 second time delay associated with a generator trip resulting from a turbine trip.
QUESTION 6.02 (3.00)
- a. What conditions will cause an AUTOMATIC START of an emergency diesel generator? (1.0)
- b. Placing the Engine Control Transfer Switch in the STOP position will shutdown the associated diesel generator if TWO conditions are present. What are these TWO conditions? (1.0)
- c. Following a loss of offsite power with a safety injection signal present, which of the following abnormal conditions, if occurring separately, vill result in a diesel generator trip? (More than one answer may be correct.)
- 1. Excessive vibration
- 2. Generator phase differential
- 3. Generator reverse power
- 4. Low lube oil preat.ure
- 5. Overspeed
- 6. High jacket water temperature QUESTION 6.03 (2.00)
Technical Specification 3.3.A.6 requires that each RCS accumulator shall be operable when RCS pressure is greater than 1000 psig. What are the -
operability requirements of the accumulators? (List setpoints, if any.)
1 QUESTION 6.04 (1.50)
In regards to the Engineered Safeguards System, any "s" signal will initiate a ccntainment ventilation isolation. What other 3 signals / conditions will initiate a containment ventilation isolation signal? (Include coincidence in your answer.)
QUESTION 6.05 (1.50)
What are the 3 possible locations (system components) that core exit thermocouple temperatures can be read from?
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
(
f, 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 7 QUESTION 6.06 (1.50)
P7, which receives inputs from P10 and turbine impulse pressure, automatically blocks "at power" trips. What are these "at power" trips that automatically get blocked? (Coincidence and setpoints are not required.)
QUESTION 6.07 (1.00)
TRUE or FALSE.
- The component cooling water heat exchanger outlet temperature should not normally exceed 105 degrees F. However, a maximum temperature of 120 degrees F for a three hour period is acceptable when the RH system is operating for a unit cooldown.
QUESTION 6.08 (3.00)
- a. What two conditions / interlocks must be met in order to OPEN the RHR Heat Exchanger to SI Pump suction valves MV-32206 or MV-322077 (1.0)
- b. List FOUR reasons that the valves (in part a) cannot be opened prior to meeting the interlocks? (2.0)
QUESTION 6.09 (1.00) .
The Tref signal is derived from the high turbine 1st stage impulse chamber pressure. What two items specifically prevent the high failure of the pressure transmitter causing a false high Tref signal above the full power value?
QUESTION 6.10 (1.00)
Which of the following fuel handling tools require air for operation?
- a. Spent fuel handling tool
- b. RCC change tool
- c. New RCC handling tool
- d. Thimble plug handling tool
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
. '8. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 8 o
6 QUESTION 6.11 (2.00)
Indicate whether the OP Delta "T" trip setpoint will INCREASE, DECREASE, Consider each change or REMAIN THE SAME if the following changes occur.
independently. Briefly justify each answer,
- a. The pressurizer spray valves fail open for 10 seconds. (0,5)
- b. An overdilution of the RCS occurs with rods in manual at 100%
reactor power. (0.5)
- c. Reactor power is decreased from 100% to 50% in a controlled manner.
(0.5)
- d. The reactor operator borates 50 ppm with rods in manual at 20%
reactor power. (0.5)
QUESTION 6.12 (1.00)
Briefly explain how an Urgent Failure alarm is generated when recovering a dropped RCCA in a bank that has more than 1 group.
(1.0)
QUESTION 6.13 (1.00)
TRUE or FALSE.
The OT Delta T rod stop setpoint is equal to the OT Delta T trip setpoint
-54, and it blocks both MANUAL and AUTOMATIC rod insertion.
!; QUESTION 6.14 (2.00) 1 Shortly after the plant experiences a reactor trip and safety injection D due to a LOCA, the Reactor Operator reports that the "SI Pump Locked Out" alarm is actuated on the control board.
y a. What caused this alarm to activate? (1.0)
- b. What are 2 possible actions that must be taken to clear this alarm?
(1,0)
(***** CATEGORY 06 CONTINUED ON NEXT PAGE *****)
i
- 6. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 9 QUESTION 6.15 (2.00)
For each one of the following process and/or area radiation monitors, state all automatic actions which will occur when the high level alarm setpoint is reached. If no automatic actions occur, state this in your answer. (0.25 for each)
- a. R-11, 12 containment air
- b. R-15 Condenser air ejector
- c. R-18 Waste liquid
- d. R-19 SG blowdown
- e. R-22 Shield building stack
- f. R-23, 24 Control Room ventilation
- g. R-30, 37 Auxiliary building ventilation
- h. R-39 CC system I
t
(***** END OF CATEGORY 06 *****)
, ,7 PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 10 RADIOLOGICAL CONTROL
~
QUESTION 7.01 (2.00)
During a loss of all AC power, ECA 0.0 has the operator depressurize the RCS using the Steam Generator Power Operated Relief valves.
- a. What RCS cooldown rate should.be attained during depressurization?
- b. Why must the RCS be depressurized?
- c. Explain why the RCS shouldn't be depressurized below 390 psig?
- d. What should be done if, while depressurizing, pressurizer level is lost or vessel head voiding occurs?
QUESTION 7.02 (3.00) ,
Due to a catastrophic fire in the Relay Room, a decision has been made to evacuate the control room.
- a. Who is responsible for making that decision and what are that person's subsequent immediate actions? (1.4)
- b. List the immediate actions for the Unit 1 and Unit 2 Plant Equipment and Reactor Operators (not the Lead Plant Equipment and Reactor Operators)? (1.6)
QUESTION 7.03 (1.50)
In Procedure FR-S.1 entitled, " Response to Nuclear Power Generation /ATWS,"
after the operator has verified a reactor trip, he is directed to verify a turbine trip. Why is it important that the turbine also trips? .
QUESTION 7.04 (1.50) i While in Procedure ECA-2.1 entitled, " Uncontrolled Depressurization of Both Steam Generators," Step 2 has you maintaining a minimum feed flow to the steam generators.
f Provide three reasons for maintaining this minimum feed flow, i
l
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 11 L* RADIOLOGICAL CONTROL QUESTION 7.05 (2.00)
Assume you have just come on shift and taken turnover. The reactor is at 80% power and the previous shift had just completed raising accumulator level to 56 percent. While walking down the control room boards, you notice pressure in the affected accumulator increased to 800 psig.
- a. Explain why high accumulator pressura is something you should be
, worried about? (1.0)
- b. What cautions, if any, should be observed if you choose to reduce accumulator pressure? (1,0)
QUESTION 7.06 (2.50)
Assume that because of an instrument malfunction, the Reactor Makeup Control is unable to maintain the minimum water level in the VCT above the low low level.
- a. What automatic actions would take place in the first 15 minutes (after reaching the low low level setpoint) assuming no operator actions and ALL control systems in automatic. (1.0)
- b. What precautions, if any, should be taken with regard to the reactor coolant pumps. (0.5) Explain (1.0)
QUESTION 7.07 (3.50)
Procedure E12 describes the immediate and subsequent actions to be taken in the event of a loss of reactor coolant pump cooling water. -
- a. What are six indications (alarms, annunciators, observations) that the reactor coolant pump has lost either component cooling or seal injection water? (1.5)
- b. What operator immediate actions should be taken in the event that all reactor coolant pump cooling water is lost and procedure E12 has been entered? (Include all options) (2.0)
QUESTION 7.08 (1.00)
What are the required immediate operator actions if Procedure ES,
" Loss of Feedwater Supply" has been entered?
- (***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
. 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 12 RADIOh0GICAL CONTROL i-QUESTION 7.09 (2.50)
Procedure FR-Z.1, Response to High Containment Pressure requires that hydrogen concentration be monitored.
- a. What is the primary source of hydrogen in containment during a major loss of coolant accident? (0.75)
- b. Why is containment hydrogen concentration a concern while you are in this procedure? (1,0)
- c. How is hydrogen concentration reduced while you are in this Procedure? (0.75) 3 ir QUESTION 7.10 (2.00)
In June 1986, during an RHR surveillance test, with both units operating at 100 percent power, problems were experienced with the RHR HX component cooling water inlet valve such that insufficient CCW flow was available (500 gpm) to the RHR HX.
- a. What were/are tuo, control room indications that could be used to identify a problem with CCW for the affected HX7
- b. Explain why this train of RHR should be declared operable / inoperable?
QUESTION 7.11 (2.00)
Answer the following questions regarding the release of gaseous effluents.
- a. What is(are) the Technical Specification bases for limiting the -
quantity of radioactive material in any gas storage (decay) tank to less than 78,800 curies of noble gases?
- b. When making a gas decay tank release, wind velocity and direction should be checked to insure that unfavorable wind conditions do not e-ist.
- 1. What are " unfavorable wind conditions" per Technical Specifications?
- 2. What are the potential consequences of making a gaseous release during unfavorable wind conditions?
(***** CATEGORY 07 CONTINUED ON NEXT PAGE *****)
.w 7.' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 13 RADIOLOGICAL CONTROL
,4 4 v.
QUESTION 7.12 ( 1.' 50 )
Unit l'is shut down for a refueling outage and fuel movements are being perfermed. While transporting a fuel assembly from the Reactor Refueling
~ ' Canal to the Transfer Canal, the transfer car becomes stuck in the middle
- of the Fuel Transfer Tube.
La. What indication (s) or lack of indication (s) on the fuel pool control panel would indicate if the transfer car was stuck? (0.5)
.b. If the transfer car was stuck and its air driven motor was inoperable, explain how and in which direction the transfer car would be removed.
n s
a l
l l
l
(***** END OF CATEGORY 07 *****)
i
-i-- - .
e-. -w- --aw , --- - -+w --, . , - - - - .--,w .,,.r--
8.~ ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 14 QUESTION 8.01 (2.25)
- a. Under what Emergency Classification (s) is the TSC activated? (0.75)
- b. What function (s) does the TSC perform while activated? (0.75)
QUESTION 8.02 (1.00)
What is(are) the bases for the flow restriction orifice in the vent path from either the reactor vessel or the pressurizer? (1.0)
QUESTION 8.03 (2.00)
What required Tech Spec action (s) should be taken if a Prairie Island Safety Limit is violated?
=
QUESTION 8.04 ( .75)
TRUE or FALSE An individual shall not work or have worked more than a total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, including shift turnover.
QUESTION 8.05 (1.90) -
Answer the following questions related to procedural adherence.
- a. Under what condition (s) may a procedure or procedural step be deviated from? (0.6)
- b. Assuming a procedure or procedural step has been violated, what actions should take place to document the violation?
- c. How may data entered incorrectly in a procedure / checklist be modified?
- d. Under what conditions may procedures not be required at the work site?
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
a ,8:' ADMINISTRATIVE PROCEDURES. C'ONDITIONS. AND LIMITATIONS 1 PAGE 15 QUESTION 8.06 (1.50)
Unit 2 is operating at power and Unit 1 is being shut down for refueling.
Unit 1 status is as.follows:
Reactor cooldown proceeding with RCS temperature greater than 200 degrees F.
' Reactor coolant borated to refueling boron concentration (2221 ppm).
- 121 boric acid storage tank at 1950 gallons What section of Technical Specification 3.2 (Attachment 8.1) applies when-
'the. reactor is borated to refueling concentration and the unit is in the process of being cooled down (if any at all)? Justify your answer.
. QUESTION 8.07 (2.60)
Regarding control room shift organization, operation, and turnover:
- a. What information, as a minimum, shall be discussed with an incoming Shift Supervisor at shift turnover? (1.2)
- b. What is the minimum shift crew composition when both units are above cold shutdown? (0.8)
- c. What three requirements must be satisfied / performed prior to the-oncoming operator accepting the responsibilities for his position from the outgoing operator? (0.6)
-QUESTION 8.08 (1.50)
What are(is) the Technical Specification Bases for maintaining the specific activity of the secondary coolant system less than 0.10 uCi/gm Dose Equivalent I-1317 QUESTION 8.09 (1.00)
TRUE or FALSE
- During emergencies, the on duty Shift Technical Advisor is prohibited from
- assuming the overall plant decision making authority normally delegated to the Shift Supervisor.
(***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
,8.'* ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 16 QUESTION 8.10 (1.40)
What on shift personnel make up the fire brigade and what are their individual assignments / titles?
QUESTION 8.11 (1.00)
List 4-v vi -Lhe Technical Specification Bases for running each circuit of the control room ventilation system for a minimum of 15 minutes every month?
QUESTION 8.12 .50) a o dr ci
- b. Why are low power physics tests performed?
QUESTION 8.13 (1.00)
What are(is) the bases for the periodic leakage testing of primary coolant system pressure isolation valves to ensure valve integrity?
QUESTION 8.14 (1.00)
What are(is) the Technical Specification Bases for the minimum condensate storage tank volume of 100,000 gallons?
\
1 QUESTION 8.15 (1.50)
The auxiliary electrical systems Technical Specifications place restrictions on which and how many electrical power supplies are available for any given plant condition. If adhered to, these specifications assure l that a minimum number and type of power supplies are available to safely i shut down the plant. What type of. power supplies, as a minimum, must be available and what loads must they provide under normal and accident l
l conditions?
l 1
l l (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****)
l t
, ', ' 8 .' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 17 QUESTION 8.16 (1.60)
What (are)is the basis for having a minimum of 70,000 gallons' diesel fuel available in the interconnected storage tanks?
QUESTION 8.17 (1.50)
List all conditions in which the Auxiliary Building special doors can be left open.
.l
(***** END OF CATEGORY 08 *****)
(************* END OF EXAMT. NATION ***************)
5'. THEORY OF NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 18 THERMODYNAMICS ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 5.01 (2.00)
.a. Doppler-only power defect.
Moderator-only power defect.
Void-only power defect.
(.50 each)
- b. Doppler only power defect. (.25)
- c. Doppler only power defect. (.25)
REFERENCE Westinghouse Reactor Core Control For Large PWR's, Chapter 3, page 41 001000K549 ...(KA'S)
ANSWER 5.02 (1.50)
- 1. Decrease
- 2. Remains the same
- 3. Remains the same
(.50 each)
REFERENCE Westinghouse Reactor Core Control for Large PWR's, Chapter 7, pp 13-20 192002K114 ...(KA'S)
I MASIER CORY l
l l
S'.' THEORY OF NUCMAR POWER PLANT OPERATION. FLUIDS. AND PAGE 19
".* THERMODYNAMICS
. ANSWERS - PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 5.03' (2.00)
- c. DET A (normalized)
N41 N42 N44 374.4=.936 356.6=.8915 360=.9000 400 400 400 DET A (Avg) = 938 + .8915 + .9= .909 3
DET A QPTR = Max = (N41) = .936 = 1.03 (.25 pts)
Avg Avg .909 DET B (normalized)
N41 N42 N44 324.0=.81 342.0=.855 360=.900 400 400 400 DET B (Avg) = .81 + .855 + .900 = .855 3
DET B QPTR =_dax = .900 = 1.05 (.25 pts)
Avg .855 Core QPTR = 1.05 (.50 pts)
(.50 pts for work)
- b. QPTR shall not exceed 1.02 (also accept 2%) (.50 pts)
REFERENCE Westinghouse Reactor Core Control for Large PWR's, Chap. 8, p. 29-31 P.I. Technical Specifications 3.100 001050K5 192005K113 ...(KA'S) r 5
, . _ . _ . - _ . _ , _ _ _ _ _ . , . . _ . - . ~ -
- 5. THEORY OF' NUCLEAR POWER PLANT OPERATION. FLUIDS. AND PAGE 20 THERMODYNAMICS
. ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
f ANSWER 5.04 (1.50)
- a. Reactor A (.25), because critical rod height will be reached quicker. (.25)
- b. Reactor B (.25), because the slower rod speed allows time for suberitical multiplication to increase count level to a higher value. (.25)
- c. They sell be the same ( .25) because critical rod height is not.affected by rod speed because the same amount of reactivity needs to be added to both reactors. (.25)
REFERENCE Westinghouse Fundamentals of Nuclear Reactor Physics, Chapter 5, pp 19-25 and Chapter 7 Westinghouse Reactor Core Control for Large PWR's, Chapter 6, pp 10-13 and Chapter 9, pp 13-15 192008K104 ...(KA'S)
ANSWER 5.05 (1.00) 1.- Temperature below RT-NDT
- 2. Material defect or flaw
- 3. External stress on vessel (Any 2 9 .50 each)
REFERENCE Westinghouse Thermal-Hydraulic Principles, Chapter 13, pp 60, 61 193010K104 ...(KA'S)
ANSWER 5.06 (2.00)
- a. Lower than-ECP
- b. No Significant Difference
- c. Higher than ECP
- d. Lower than ECP REFERENCE Westinghouse Reactor Core Control, Chap. 7, pg. 25
, 001010A207 ...(KA*S) 4
..u
. , 5[ THEQRY OF NUCWAR JOWER PLANT OPERATION. FLUIDS. AND PAGE 21
. THERMODYNAMICS
. ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 5.07 (1.50)
- a. 1647 psig +/- 7 psig (.75)
- b. 642 psig +/- 7 psig (.75)
REFERENCE Steam Tables-193003K125 ...(KA'S)
ANSWER 5.08 (2.50)
- a. Xenon is produced by fission (0.25) and iodine decay (0.25) Xenon is removed by neutron absorption (burnout), (0.25) and Xenon ~
decay (0.25)
- b. Samarium is produced by Promethium decay (.25) and removed by neutron capture (burnout). (0.25)
- c. Xenon will peak (0.25) and then decrease to a new equilibrium value below the initial value (0.25).
Samarium will peak (0.25) and then return to the initial value (0.25)
REFERENCE Westinghouse Reactor Core Control for Large PWR's, Chapter 4, pp. 26-35 192006K103 ...(KA'S)
ANSWER 5.09 (2.00)
The fuel movement should not, continue (0.5), because Technical Specification (3.8.A.4) on refueling re. quires > 2000 ppm boron concentration or Keff < .95 (whichever is most restrictive) during fuel movement. (0.5)
If Total reactivity (p) = -4570 pcm then K = 1 / (1 p) = 1 / (1-(- 0.04570)) = .956 (0.5)
Since Keff is not less than or equal to .95, the fuel movement
- should be stopped. (0,5)
REFERENCE Prairie Island Technical Specifications 3.8.A.4 192002K112 ...(KA'S)
- 5. THEORY OF NUCW AR POWER' PLANT OPERATION. FLUIDS. AND PAGE 22
. THERMODYNAMICS
, ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 5.10- (1.50) ,
- a. Decreases
- b. Increases
- c. Decreases
(.50 pts each)
REFERENCE Westinghouse Thermal-Hydraulic Principles and Applications to the PWR, Chapter 13, pp. 17-24 193008K105 ...(KA'S)
ANSWER 5.11 (1.00)
- b. (1.0)
REFERENCE Westinghouse Reactor Core Control, Chapter 9, p. 10 192008K104 ...(KA'S)
ANSWER 5.12 (1.00)
- c. (1.0) .
, REFERENCE Westinghouse Reactor Core Control for Large PWR's, pp. 8-21, 2-45, 3-19, 2-23 192004K106 ...(KA'S)
ANSWER 5.13 (1.00)
Greater than.
REFERENCE Prairie Island Surveillance Procedure 1005 [2005]
015000K504 ...(KA'S)
L l
- 5. THEORY OF NUCLEAR POWER PLANT OPERATION FLUIDS. AND PAGE 23 l THERMODYNAMICS ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 5.14 (1.50)
- a. False (0.75)
.b. True (0.75)
REFERENCE Westinghouse Reactor Core Control for Large Pressurized Water Reactors, Chapter 2, pg. 5 192002K112 ...(KA'S)
ANSWER 5.15 (1.00)
B'ank D at 140 steps. (130 to 150)
REFERENCE
'C1-A, Reactivity Calculations, Figure C1-4A, Page 1 of 2 001000K505 ...(KA'S)
ANSWER 5.16 (2.00)
- a. False. (0.50) Raising RCS pressure reduces pump flow rate which increases fluid pressure at suction of pump. (NPSH:Psuct - Psat)
(0.50)
- b. False. (0.50) Lowering fluid temperature lowers saturation pressure of fluid at pump suction. (NPSH=Psuct - Psat) (0.50)
REFERENCE Westinghouse T-H Principles, 10.56,60 COMPONENT / PUMPS-CENTRIFUGAL /2.9/3.2 0200156K50 ...(KA'S)
6
- 6. PLANT SYSTEMS-DESIGN. CONTROL. AND INSTRUMENTATION PAGE 24
' ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 6.01 (1.50)
- 2. Ensures RCS flow is maintained temporarily in event RCP's are deenergized. (Minimizes reduction of DNBR.) (0.5)
- 3. Prevents turbine overspeed. (Ensures turbine is tripped before unloading generator.) (0.5)
REFERENCE ,ys Prairie Island Exam Bank, Section 3, #2 S '3 050045K101 ...(KA'S) \
Cf f*W fgi ANSWER 6.02 (3.00)
- a. 1. SIS at either unit
\cd'( I'V". (0.5)
- 2. UV signal from either safeguards bus (0.5)
- b. 1. Tus {I igne1 prr :nt- (0.5)
- 2. Emergency diesel generator breakers are open (0.5)
- c. 2 and 5 (0.5 for each)
REFERENCE Prairie Island System Description B-38A, p. 8, 9, 12 000064K402 ...(KA'S)
ANSWER 6.03 (2.00)
(0.5 pts each)
- 1. The isolation valve is open
- 2. Volume is 1270 (+/- 20) cubic feet of borated water
- 3. Minimum boron concentration of 1900 ppm
- 4. A nitrogen cover pressure of at least 700 psig REFERENCE Prairie Island Technical Specification 3.3.A.1.b 194001K8 ...(KA'S) l l
, , 6'. PLANT SYSTEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 25 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
i ANSWER 6.04 (1.50)
(0.5 each)
- 1. High radiation from either R-11, R-12, or R-22
- 2. Operation of 1 out of 2 manual containment isolation switches.
- 3. Operation of 2 out of 2 manual containment spray switches.
REFERENCE Prairie Island System Description B-180, p. 12 013000 ...(KA'S)
ANSWER 6.05 (1.50)
(0.5 pts each)
- 1. Plant Process Computer
- 2. Control Room Core Exit Thermocouple Display
- 3. Test Equipment readings from the core exit thermocouple junction boxes
- 4. P-3 5 c. ERCS REFERENCE Prairie Island Technical Specification 3.15.A Basis 194001 ...(KA'S)
ANSWER 6.06 (1.50) .
(0.5 each)
- 1. 2 loop low flow related trips (undervoltage, RCP breaker open, low flow)
- 2. Low pressurizer pressure
- 3. High pressurizer level REFERENCE Prairie Island Plant Info Summary, p. 1 194000 ...(KA'S)
ANSWER 6.07 (1.00)
False.
REFERENCE Prairie Island System Description B-14, p. 26 008030K7 ...(KA'S)
3 ,
. 6 .' PLANT SYSIEMS DESIGN. CONTROL. AND INSTRUMENTATION PAGE 26 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
~
ANSWER 6.08 (3.00)
- a. The associated RHR pump discharge pressure is less than 210 psig.
(0.5)
The associated SI pump suction isolation valve from the RWST is shut. (0.5)
- b. Protects the low pressure cross connect piping between the two systems. (0,5)
Prevents a loss of water back to the RWST. (0.5)
Prevents contamination of the RWST by water from Sump B. (0.5) i Prevents airborne radioactive release through the RWST vent. (0.5)
REFERENCE Prairie Island System Description B-15, p. 17; B-18B, p. 14 005000K407 . ..(KA'S)
ANSWER 6.09 (1.00)
The " low selector" (0.5)
The controller getsa input signal of 560 degrees F (0.5) co A s+A d k REFERENCE Prairie Island System Description B-7, p. 12 041020K411 . ..(KA'S) 1 ANSWER 6.10 (1.00) b.
REFERENCE Prairie Island System Description B-17, pp 36-39
, 194001 . ..(KA'S)
, . - . ,~.-,,-.n-- , - - - - - - - - , . - - - , . , , - - - , -
- 6. PLANT SYSTEMS DESIGN. CONTRdL. AND INSTRUMENTATION PAGE 27 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
'O ANSWER 6.11 (2.00)
- a. Remains the same (0.25). RCS pressure is not an input to the OP Delta T setpoint (0.25).
- b. Decreases (0.25) - Due to rising average RCS temperature. (0.25)
- c. Remains the same (0.25) - OP Delta T setpoint is constant for all steady state normal operating power levels (0.25).
- d. Remains the same (0.25) - A decreasing RCS temperature, due to the boration, will not cause OP Delta T setpoint to increase (0.25).
REFERENCE
-Prairie Island System Description B-8, pp 7-9 Prairie Island Technical Specification 2.3.2.e 012000K611 ...(KA'S)
ANSWER 6.12 (1.00)
When recovering a dropped RCCA in a bank that has more than one group, you have to open the lift coil disconnect switches for all the rods that you don't want to move in that bank (.50), so when you go to pull out the dropped RCCA, the rods in the opposite group demanding current will not match coil current (Regulation Failure) and you will get an Urgent Failure alarm. (.50)
REFERENCE Prairie Island System Description B-5, pp 16-18 Prairie Island Plant Operations Manual C5, Section 8.2
- 001000K409 001050A201 ...(KA'S)
! ANSWER 6.13 (1.00) .
False.
REFERENCE l Prairie Island System Description B-5, p. 22 001000K407 ...(KA*S)
6'. PLANT SYSTEMS DESIGN. CONTRO'L. AND. INSTRUMENTATION PAGE 28 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 6.14 (2.00)
- b. 1. Reset using local " Reset" pushbutton or (0.5)
- 2. Placing control board switch to stop. (0.5)
REFERENCE Prairie Island System Description B-18A, p. 12 006000K603 ...(KA'S)
ANSWER 6.15 (2.00)
- a. Containment ventilation isolation.
(Ctmt and In-service purge isol.)
- b. None.
- c. Stops discharge (to river),
- d. Stops blowdown.
- e. Containment ventilation isolation. (Ctmt and In-service purge isol.)
- f. 1) Starts cleanup fan, 2) Shifts ventilation to Recire Mode,
- g. Shifts to auxiliary building special ventilation.
- h. Shuts vent valve on surge tank.
REFERENCE SD B-11, Table B11-1 0.0008K103 000029K102 000068K111 000071K106 ...(KA'S) e I
l i
i
- 7. ?RQQEDURES - NOFMAL. ABNORMAL. EMERGENCY AND PAGE 29 MOLOGICAL cob TROL !.
V._
ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S. ?6 e f$
4t <S
{T o
' ANSWER 7.01 (2.00) iff
.O Maximum attainable rate (0.25) that is controllable (0.25). /f
c .- To prevent injection of accumulator nitrogen into the RCS (0.5).
- d. Continue depressurization (0.5).
REFERENCE ECA 0.0 000055 ...(KA'S)
ANSWER 7.02 (3.00)
- a. Unit 1 Shift Supervisor (0.4)
Notify SEC and STA (0.5)
Proceed with radio to Remote Shutdown Panels (0.5)
- b. Unit 1 or 2 PE & RO:
Trip reactors (0.2)
Close MSIVs (0.2)
Close Pressurizer PORVs (0.2)
Assure turbine trip (0.2)
Unit 1 PE & RO:
proceed with radio to D1 Diesel Room to isolate diesel control circuits from Control Room (0.2)
Unit 2 PE & RO:
Proceed with radio to each unit rod drive control room and verify reactor trip. (0.2)
Verify turbine trip at front standard. (0.2)
Proceed to screenhouse with radio to verify fire header pressure.
(0.2) i l
REFERENCE FS, Appendix B 000068 ...(KA'S) i l
l l
7 ~. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 30 BADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 7.03 (1.50)
To prevent further cooling of the RCS (0.75). This would result in a {
reduction in Tave which would add excessive reactivity reducing shutdown h M S) p M g, , gi,3 g g 4 , J <s7 REFERENCE WI CdAc er&4[ IOSS#f, [cn, M@d Flow Procedure FR-S.1 000029K306 ...(KA*S) _
ANSWER 7.04 (1.50) 1.' To minimize additional coolciown (0.5) ggj M.
Q _
- 2. To prevent steam generator tube dr ut"(0.5) M
- 3. To minimize the water inventory in the steam generators (0.5)
REFERENCE Procedure ECA-2.1 035010K301 ...(KA'S)
ANSWER 7.05 (2.00)
- a. Excessive pressure in the accumulator could result in the injection of nitrogen into the RCS (0.33) which could displace water in the S/G tubes (0.33) which would prohibit natural circulation (0.33).
- b. One operator is dedicated to operating containment isolation valves used to vent accumulators (0.5).
Assure persons are not working in vicinity of accumulator vent valves. (0.5)
"W
^
gg
&kW ~ (M ,
o r C-18 M TM @
006020A107 ...(KA'S)
, 7 .^ PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 31 RADIOLOGICAL CONTROL ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 7.06 (2.50)
(0.33). Controlling groups of control rods will travel out to maintain Tave with Tref setpoint (0.33).
- b. Sh;uld L., ...d mi ni ai sc f 1 ::- fi u m the T.; T -i G . 5 ) . Ceuldkoseseal injection flow (035) caused by the 12% boric acid-solidifying in the seal injection throttle valves or the seal injection filters (075).
REFERENCE Procedure C12, pp 83, 88 0000A104 000022K302 ...(KA'S) l l
l
- 7. PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 32 RADIOLOGICAL CONTROL
. ANSWERS -- PRAIRIE' ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 7.07 (3.50)
[Any 6 9 .25 each)
- a. Indication by #11 [21] and #12 (22] component cooling Pump switch module lights that the component cooling pumps are not operating.
Low level alarms from #11 [2] component cooling surge tank.
Low pressure alarm on the component cooling outlet header.
Low flow indication from the component cooling heat exchanger out'let header.
Low flow alarm in the outlet component cooling line from #11
[21] or #12 [22] RCP.
High flow of component cooling water to the reactor coolant pump (s) (line rupture).
High temperature indication / alarm for the reactor coolant pump motor radial and thrust bearings (greater than 185 degrees F)
- 11 [21] or #12 [22] RCP #1 Seal Water Low Delta P Alarm.
Indication on the charging pump switch module lights that the charging pumps are not running. ,
- 11 [21] or #12 [22] RCP Labyrinth Seal Low Delta P Alarm.
Seal Injection Flow Low Indication.
High temperature indication / alarm from the reactor coolant Pump (s) radial bearing (greater than 170 degrees F).
- b. Attempt to start any charging pump to restore seal injection.
(0.5)
Prior to reaching radial bearing temperature of 200 degrees, Perform the following (0.30).
- 1. Trip the reactor (0.3)
- 2. Step the RCP's (0.3)
- 3. Close the #1 Seal Leakoff Isolation Valves (0.3) for both pumps (0.3)
REFERENCE Procedure E12 -
000017A107 ...(KA'S)
L PROCEDURES . NORMAL. ABNORMAL. EMERGENCY AND PAGE 33 RADIOLOGICAL CONTROL
. ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 7.08 (1.00)
Verify turbine and motor driven auxiliary feedwater pump discharge pressures exceed the S/G pressure. (0.5)
AssurethereisT[Jou4-- ;uie to both S/G. (0.5)
REFERENCE Procedure E5 000054K304 ...(KA'S)
ANSWER 7.09 (2.50)
- a. Zirconium-water reaction. (0.75)
- b. Explosion / pressure spike (Oe67 which could breach containment. 40737
ANSWER 7.10 (2.00) l
, a. Dual light indication on HX CCW inlet valve. (0,5) -
CCW flow indication on HX CCW inlet valve (0.5)
- b. Because the flow is below TS allowable for operability (0.5), the RHR (i'
HX and RHR pump should be declared out of service. (0,5) l Os 0$rs W av4.4.9 Q. W -9 Q & 5 c^aua. M m @ <.A .(O 6 ))
REFERENCE ,
Significant Operating Event SOE-2-86-2, Procedure C14 l
008000K301 ...(KA'S) 1 i
I v
L
7 '. ' PROCEDURES - NORMAL. ABNORMAL. EMERGENCY AND PAGE 34 RAD 10 LOGICAL CONTROL
. ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 7.11 (2.00) i a. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, (0.5) the resulting total body exposure to an individual at the nearest exclusion area bou d r will not exceed 0.5 rem. (0.5) locFR10 n locFR 100 6%5c oJc^f
- b. 1. Wind from 5 degrees west of north to 45 degrees east of north
( 0 25 ) at less than 10 mph (025).
OR Wind from north to south in the direction of the cooling towers (0 j3 ) at less than 10 mph (0 25) .
(Accept either)
- 2. At low wind, velocities (below 10 mph) the gaseous activity
. released from the gaseous radwaste system could be at or near ground level near the cooling towers and remain long enough to be drawn into the circulating water in the tower. (0G5)
This specification minimizes the possibility of releases from the gaseous radwaste system from entering the river from tower scrubbing. (0 25 )
REFERENCE Tech Spec 3.9-5, SOP C21.3, Lesson Plan 85-33 071000 ...(KA'S)
ANSWER 7.12 (1.50)
- a. The conveyor at pit light (0.25) would not light (0.25).
- b. It may be freed or pulled to its spent fuel pit side position (0.25) via a hook and cable attached to the emergency cable (075)
A hook attached to a wire cable can be " fished" through the loop of the emergency cable. (Bee 9)
REFERENCE D.5., C17, Lesson 8181L-003 000036A104 ...(KA'S) 8
,8. ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 35 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 8.01 (2.25)
- a. Alert (0.25)
Site Area Emergency (0.25)
General Emergency (0.25)
- b. l Support the Control Room command and control functions M 1 Assess the plant status 64kM) andpotential of f site impact (^ . 25; y Coordinate ANY /$6+emergency r ss
. onse a&ctions (M hM REFERENCE E plan section 7 194001 ...(KA'S)
ANSWER 8.02 (1.00) i Orifice will limit the flow from an inadvertant actuation of the vent system to less than the flow capacity of the reactor coolant makeup system.
REFERENCE Technical Specification Bases 3.1-5 002020 ...(KA'S)
ANSWER 8.03 (2.00) .
1.
2.
Reactor shut down immediately NRC notified (0,5)
-( 07, /0
- 3. D rerted t entral Mert :: S cle er "le..i.o (0.25) and th: Chairmyn of j a Cefety f.udi+ Committee In 95) ^r their derip -teri =1+avnates.
- 4. A safa+y limit fic'snt' n (0.5) ,
speo shnLe cw a i+ 7a es ~ e4.. reps"k, ' ' hc(c.p.er . -- we) md .
REFERENCE Procedure C.1.1, pg. 3 194001 ...(KA'S)
ANSWER 8.04 ( .75)
False.
REFERENCE 5ACD 3.15, pg. 4 194001 ...(KA'S)
8.' ADMINLSIBATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 36 o
- ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 8.05 (1.90) 4.a)
- a. h :- vi o,,m . . c . . c y g (0. t), uv y m . ou i, . u .; m , t: r r ;....a ( c . I)',:
-": : r- 21 r u " C.:.,, vi a.m... 1: th: f :ility :;.I;
- b. Document in Reactor Log (s). (0.4) (o.05)
- c. Draw single line through (0.19), initial (0.19), and date it (0.16),
- d. Procedural steps committed to memory. (0.2)
Routine procedural actions frequently performed. (0.2)
Procedural steps communicated from a remote location. (0J)
REFERENCE Procedure SACD 1.5 194001 ...(KA'S)
ANSWER 8.06 (1.50) 0 Technical Specification 3.2.0.5 applies (0.5) It also states that two boric acid tanks shall have a minimum of 2000 when one unit is critical and the other unit is heated or maintained above 200 degrees F.
Even though the unit is being cooled down, it is still above 200 degrees F and falls within the Technical Specifications. (1.0)
REFERENCE Prairie Island Reportable Event P-RE-1-85-01 004020 ...(KA'S) ggg y dsp A o.8 b W b (o.4/)
w N o Sf0f Y (3'd)M) g k W"c sk Q OR Qg Q (l 9 . y a s 4u2(%) m a> m % "~K~9
. , , - - - ---..--.,,-.n-..-
8 '. ' ADMINISTRATIVE PROCEDURES. C'ONDITIONS. AND LIMITATIONS PAGE 37 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S.
ANSWER 8.07 (2.60)
- a. Current plant and system status Outstanding surveillance procedures Maintenance in progress or planned Plant operation evolutions in progress or planned i Tech Spec related problems New operational or administrative procedures or instructions (0.2 pts each)
$ licensed or unlicensed individuals (0.2 pts each)
- c. Receive complated status of the day.
Information tells outgoing operator "you are relieved".
Signs shift change status form (0.2 pts each) ;
REFERENCE SWI-0-2 194001 ...(KA'S)
ANSWER 8.08 (1.50)
Will assure the resultant offsite radiati dose will be limited (d>MEP to a small fraction of 10 CFR Part 100 limit n the event of a steam line rupture (0,5) concurrent with a 1 spm primary to secondary tube leak in the steam generator of the affected steam line. (0,5) i REFERENCE-T.S pp 3.4-3 .
000060 ...(KA'S)
ANSWER 8.09 (1.00) .
True.
REFERENCE SWI-0-2, pp. 8 194001 ...(KA'S)
. - _ - - . - _ _ - - . _ . . . - . . . . . . ,._-,-_._--m - - , - . _ - . _ _ - . _ _ _ _ _ , , - ~ . _ _ - -
, 8'. ' ADMINISTRATIVE PROCEDURES. CONDITIONS. AND LIMITATIONS PAGE 38 ANSWERS -- PRAIRIE ISLAND 1&2 -87/03/23-HARE, S. ,
ANSWER 8.10 (1.40)
Unit 2 SS (0.2) - Fire Brigade Chief (0.2)
Turbine or Aux. Bldg. APEO (0.2) - Assistant Chief (0.2)
BOP or excess CR Operators (0.2) - Fire Fighters (0.2)
Anyone, as designated - runners (0.2)
REFERENCE F5 194001K111 ...(KA'S)
ANSWER 8.11 (1.00)
Demonstrate operability (4-6.) fcD P--^"e er:;.eivo mviaLu o .Livh 111 Lulld uy vu th: 34--*k.c ro si REFERENCE T.S 4.14 194001 ...(KA'S)
ANSWER 8.12 ( 0. 50 ) .
^
. ih w t ro f 11y withd wn '
OL 00 u rt e m a A
- b. To verify rod worth -
REFERENCE Technical Specification 3.10 Basis 001050 ...(KA'S)
ANSWER 8.13 (1.00)
Assuring valve integrity reduces the probability of gross valve failure and consequent intersystem LOCA which bypasses containment (1.0).
REFERENCE T.S 4.3 103000 ...(KA'S)
. '89' ADMINISTRATIVE PROCEDUnn. CONDITIONS. AND LIMITATIONS PAGE 39 ANSWERS -- PRAIRIE ISLAND 18s2 -87/03/23-HARE, S.
ANSWER 8.14 (1.00)
Minimum amount of water available is sufficient to remove decay heat in the reactor for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
REFERENCE T.S 3.4 056020 ...(KA'S)
- ANSWER 8.15 (1.50)
One external (0.25) and one standby (0.25)
Normal: Safe shutdown (0.25)
Containment isolation (0.25)
Accident: Operate required ESF equipment (0.5)
REFERENCE Technical Specification 3.7-2 062000 ...(KA'S)
ANSWER 8.16 (1.60)
Operation of I diesel cooling water pump (0.7) and 1 diesel generator -
(0.7) for 14 days (0.2).
REFERENCE .
Technical Specification 3.7-3
(
064050 ...(KA'S)
ANSWER 8.17 (1.50)
Containment integrity not required (0,5). Intermittently, provided they are under strict administrative control (0,5) and can be shut within 6 minutes (0.5).
REFERENCE Technical Specification 3.6-1 103000 ...(KA'S) 0 ALTERNATE ANSWER for 8.15 Two independent Off-Site, power sources (.125)
Diesel Generators (.125)
Both Batteries and Associated Charges (.125) 4160 Safeguard busses (.125)
Normals Safe Shutdown Loads (.25)
Containment Isolation (Safeguards MCC) (.25)
Emergency: Operate Required ESF loads (.5)
. _ _