ML20153B944
| ML20153B944 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/04/1986 |
| From: | Dimmock L, Lang T, Mcmillen J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20153B948 | List: |
| References | |
| 50-373-OL-85-02, 50-373-OL-85-2, 50-374-OL-85-02, 50-374-OL-85-2, NUDOCS 8602180205 | |
| Download: ML20153B944 (69) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-373/374/0L Dockets No.' 50-374; 50-374 License No(s).
Licensee:
Commonwealth Edison Company Post Office Box 767-Chicago, IL :60690 Facility Name:
LaSalle County Nuclear Station Examination Administered At:
LaSalle County Nuclear Station Examination Conducted:
November 18,19,20,21 and 22, 1985 f/((
Examiners:
ag Dat'e k
- y. Dinmock Mf//d f
D'atii!
f Millei J/f/)'d
[Tay v
Approved By: -J cM 1 fef 8/f b
Operating Licensing Section Date Examination Summary Examination administered during'the week of November 18, 1985 (Report No. 50-373/374/0L-85-Q To seven senior reactor operator and five reactor operator candidates.
Results:
Six Senior operator and three reactor operator candidates passed the examination.
8602180205 860204 PDR ADOCK 05000373 V
[
REPORT DETAILS 1.
Examiners T. Lang, NRC Region III L. Diamock, NRC Region III
'J. McMil19n, NRC Region III M. Spencer, EGG 2.
Examination Review Meeting Copies of the examinations and answer keys were given to facility personnel after the administration of the written examination.
The facility provided written comments to the examiners on November 22, 1985.
These comments are. enclosed as attachment 1 to this report and resolution of the comments is' included as attachment 2.
3.
Exit Meeting Ron Crawford, Training Supervisor, was informed on November 22, 1985, of those persons who clearly passed the oral / operating examination.
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ATTACHMENT 1 FACILITY COMMENTS
.R0 Exam, Novembe'r' 18, 1985 Reviewed by:
L. Melander, SRO Training R. Roy, SRO Cert. Training
'W..Jakielski, SRO Operating
^
SUMMARY
OVERVIEW The following general observations are offered, intending to provide constructive feedback _ to the exam writer.for consideration when grading applicant written exams.
No LaSalle references were provided for answers to questions in Section 1.
.Some had no references at all.
Many questions had unusually large point weight value when compared to the relative importance of the topics.
Point distribution for answers to essay questions indicate a philosophy shift; i.e., more emphasis on memorization of details-:in lieu of. straight forward testing of concept understanding.
Many questions involved confusing sentence and wording factors.
Again this appears to depart from the straight forward approach of testing understanding and comprehension of operational principles.
Numerous discrepancies exist for the point values between the exam and answer key.
' In summary, it is felt.these factors significantly. detract from.the quality of
<this examination and its intent to test the' Operator's knowledge and ability to safely operate ai reactor plant.
Specific comments, which support these observations, follow this summary.
SECTION-1 1
- 1.1 Disparity in the number of points between the exam and answer key.
The test indicated 2.0 points and the answer key indicate'd 1.0 points.
For the question asked,.l.0 points.is a more realistic value for the subject
'~
being tested.. In _ addition, the phrase, "...the feedwater energy in
.(1bs/hr)(BTU /lba) is compared'with the' steam energy out'(1bm/hr)
(BTU /lba)...",should not be required for full credit - the question is more_than adequately' answered by the remaining paragraph.
1.'3a Student answers that indicate a stable critical neutron population'is achieved, but continue on stating that without further' Operator. action,
~
3 x
the neutron population will decay off (due to poison building) should also be accepted for full credit.
1.3b Again, there is a-disparity in the number of points on the test and the key.
The test indicates part b is worth 2.0 points while the key indicates it is worth I 5 points. The concept tested in part b is not substantially more impo: ant than that tested in part a and as such, should also be worth oni 1.0 point.
The wording of question 1.3 is very confusing. The lead in scenario is sketchy and does not clearly identify the condition which is a reactor startup with power levels below the SRM range.
The question in part b hy itself is not bad, but when it follows the question asked in part a causes considerable confusion as to what is the correct answer for either part due to the similarities in the situations.
The students were confused uy this question as well as most of the facility SRO licensed exam reviewers.
Specifically, the question reactor response if rods were pulled to one notch above the critical position.
As worded, this could be construed to mean that rods were stopped one notch before_ criticality is' achieved, which drastically changes reactor response.
Either question 1.3b should be deleted or. answers should be accepted that address the possibility that the reactor is subcritical instead of supercritical.
Also, if the question is not deleted, the following additional key change should be made.
The weighting in the answer key needs to be changed.
The important aspects in the answer should be that a supercritical condition is achieved causing reactor power to increase on a positive period until either an IRM scram is received, or, if the Operator ranged up, until negative reactivity (temp. or voids) turns power.
(See attached corrected key).
1.4b ^K/K is not a proper unit for keff.
Keff as a fraction is unitiess.
By adding ^K/K to the value of keff = 0.9 caused considerable confusion to some students.
Compounding this, is the fact that an equation ~ sheet was not provided to the students (which is required).
In accordance with NUREG 1021, ES-202, page 6 of 6, "All equations required to answer parts of the examination should appear in the equation sheet or as part of the question."
Based on the confusion generated by the misplaced units and the lack of an equation sheet, the student may have calculated the reactivity value as follows:
Kefft = 0.9 Keff2 = 0.95
- kefft = 0.95 - 0.9 = 0.0526 K/K = keff2 keff 0.95 2
As compared to the desired calculation:
t - 1 = 0.9 - 1 = _ }=_,777 p, = keff J
keff 0.9
.9 t
e 4
p = keff2 1 _ 0.95 1 = _
.05 = - 0.0526 g
keff 0.95 0.95 2
P = 0.585 As such, the answer key should be changed to accept answer b or c as acceptable or question 1.4b should be deleted.
1.5b A nonLaSalle reference was used in the answer key. According to the LaSalle Reactor Physics review lesson plan submitted with the exam reverence material, Xenon peak on a down power transient would occur 4-6 hours after the power change.
The answer key lists #2 as the correct answer (2-6 hours).
However, answer #3 (5-7 hours) fits this time frame as well as #2 did and should be counted as a correct alternate answer.
REF:
LaSalle Physics Review Lesson Plan, page 216 1.10 Question is somewhat tricky in that it asks for the primary contribution
.to heat energy added to the coolant with a subcritical reactor (decay heat).
Tiiree answers are listed that all deal with slowing down of fission products in the fuel and an Operator is trained that from the fission mechanism - the ionization by fission fragments is the biggest heat input.
Also, radioactive decay of fission products occurs also by gamma and neutron emission not solely by Betas and that the total fission energy associated with decay gammas and decay Betas is very close.
In addition, after the first refuel outage, neutron sources can be pulled
.from the core as the buildup of transuranic elements, such.as Curium produce, enough source neutrons to maintain a stable on-scale count rate.
These neutrons are produced by spontaneous fission and these fissions are a major contributor to decay heat after a long shutdown.
The major contributor to heat energy from these spontaneous fission events is the ionization by the resultant fission fragments.
Based on the misleading nature of the question and the evidence presented, the answer key should be changed to accept c as a correct answer for half credit.
REF:
LaSalle Reactor Physics Review Lesson Plan, pages 12 and 58 1.11 The question is an SR0 level question.
The subject matter in itself is testable, but the manner in which the question is asked, requires the student to have considerable reading skills to interpret the two cases being given and contrast each of their effects in order to determine the correct' answer.
According to NUREG 1021, ES-202, page 4:
"13.
Practical realistic questions that relate to operator knowledge and required operating practice should be used."
As such question 1.11 should be deleted as it does not challenge the Reactor Operator's knowledge directly, but instead concentrates on his ability to interpret a complicated and somewhat vague written description in order to answer a basic concept regarding samarium changes with power.
This is more appropriate for an SR0 rather than an R0.
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SECTION 2 2.2 Also accept raergency makeup water injection source to the reactor.
Its use in this capacity is described in LGA-04, page 3 of 21, and also in LOA-SC-03.
2.7b Typographical error - should be high neutron flux or high pressure.
Either answer should be acceptable for full credit.
2.11 Whether the Operator has memorized all loads supplied by non safety-related 125 VDC buses has little effect on his ability to operate the plant safely.
Plant procedures provide reference lists so he can determine what equipment has been effected by the power failure.
In addition, reviewing the System Description, Chapter 43, objectives and review questions, the trainee is required to know loads on the vital DC buses, but not the non safety-related buses (111X, 112X, 121X, 211X, 212X, 221X).
Also, NUREG 1021, ES-202, page 4 states:
" Questions on detailed system characteristics or instrumentation, such as annunciator logic or setpoints, should be avoided unless required for safety system operation."
Based on this, question 2.11 should be deleted.
If not deleted, only those answers dealing with the safety-related DC loads should be tested, namely b, c and f.
REF:
NUREG 1021, ES-202; LaSalle System Description, Chapter 43 SECTION 3 3.1 Question does not ask for how each unit at LaSalle is effected.
At the present time, only Unit 2 has a CRD charging header low pressure scram feature and as such an answer is only required for Unit 2.
No credit should be lost if the st.: dent didn't say Unit 1 was effected as it is intuitively obvious as it doesn't have the feature and the question didn't ask for how both units would be effected.
3.2 Question needlessly confuses student by introductory statement stating RWM will generate an insert block upon receipt of the third insert error.
This is extraneous and confuses the student as to what is being asked for in parts a and b.
If a student assumed that the insert block had been applied by the lead-in statement, then no other rods than error rods can be moved and only in the direction necessary to clear the error.
Based on this, a student could answer the question as false.
For this reason, the question should be deleted.
REF:
LaSalle System Description, Chapter 18, page 26-27 3.3a In order for the recirc pump to shift to fast speed, the following interlocks must also be satisfied:
low RPV water level (<12.5"),
inadequate subcooling as indicated by low ^T between steam space and recirc pump suction (<10.1 F), and E0C-RPT not actuated.
Any of these three answers should also be accepted for full credit.
REF:
LaSalle System Description, Chapter 5, page 84-85 6
3.3b The question did not ask for what is cavitation - it asked for the basis of the interlocks which is to prevent cavitation in the recirculation system components. As such, this is a sufficient answer for full credit.
3.4 Disparity.again in point totals between the test and the answer key.
For part a, the exam indicated it.was worth 0.5 and part b was worth 2.0.
The answer key didn't have a point total for part a.
Part a should be worth 1.0 points - 0.5 points for each answer.
Part b is probably overweighted and should be reduced to 1.5 points - 0.75 points for each answer.
3.'5c Question asks for what 2 alarms are generated by the flux estimator not what conditions cause each alarm.
As such the correct answer should.also include." Flux Estimator Failure and Flux Estimator Needs Maintenance",
which are the alarms.
The answer in the key is the setpoints or conditions causing these alarms.
3.7 Question does not indicate that the component is part of RPIS/RSCS system and therefore, is ambiguous and easily confused.
According to NUREG 1021, ES-202, E.6:
" Questions on detailed system characteristics or instrumentation, such as annunciator logic or setpoints, should be avoided unless required for safety system operations.
As such, this question should be deleted.
3.8 The point distribution in the answer key is overly detail oriented.
Althour the answer key contains a word-for-word description of the operat1on of the drywell equipment drain sump pumps and input to alarms, that type of detail should not be required in the student's answer.
NUREG 1021, ES-202, Section E, points 6, 13 and 18, indicate that this type of question should be avoided.
Based on this, question 3.8 should be deleted.
If not deleted, the following answer key changes should be made:
For part a - the following answers should be acceptable for full credit:
1.
Excessive pump run time 2.
Excessive pump restarts or short restart time 3.
High drywell equipment drain sump temperature REF: LaSalle System Description, Chapter 73, pages 26 and 28 Also, the point value for this trivial interlock is excessive - if not deleted, decrease the point value of part a to 1.0 points.
3.11 For corner room spot coolers - in lieu of Unit 1 and 2 NE and NW area coolers also accept either:
1.
Division I ECCS pump corner room area (or spot) coolers 2.
LPCS/RCIC Room area cooler and A RHR Pump Room area cooler 7
Also, as a point of interest, according to NUREG 1021, ES-202, Section E,
. point 18, if a certain number of responses are expected,'n an answer,
-then the specific number should be included as part of the question.
3.12 Power available should not be necessary for full credit as having electricity available to an electric motor operated valve in order for it to operate is intuitive. Also, either +1.69# drywell pressure or -129 RPV level will cause the valve to operate.
REF:
LaSalle System Description, Chapter 38 3.14 Question is double jeopardy with question 3.2 and according to NUREG 1021, ES-202, Section E.15, should be avoided.
Also, no point value on exam or key for question and doesn't count in. point total.
Also, the question is tricky as it is a play on words, asking if the RWM will generate a rod block, when the fourth insert error is received.
The question could be misunderstood and the Operator answer true as the rod
~ block was generated on the third error and would also be in effect on the fourth error.
Consider deleting the question for this reason.
If not, the question should be worth no more than 1.0 points.
3.lb Answers 1 and 4 should be separated and counted as independent answers.
This makes 6 possible answers.
SECTION 4 4.1 Question was very confusing, part of this being due to the exam not delineating what part was part b.
Question wording was unclear and the students weren't sure IT they had to answer each of the 4 selections or pick out the best one.
Second, according to NUREG 1021, ES-202, Section E.23, states that diagrams should be included with questions such as this.
The lack of such a drawing further complicated the question.
As such, the question should be deleted.
If not deleted, grading should consider the conf.usion inherent in the question.
4.2 The Point value for this question is excessive, neither part should be worth more than 1.0 points.
4.3 According to the students, the examiner indicated that they should chose the single most correct answer, and yet the answer key lists two answers.
Also, the wording'for this question utilizing the word which causes the student to try and select the single most correct answer.
On this basis, this question should be deleted.
However,.if not deleted, question does not describe the type o'f failure that the CRD flow control system experienced.
Therefore, three of the conditions described could occur as follows:
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a.
If CRD FCV failed closed, cooling water would be lost to the drives resulting in M temperatures.
b.
'If CRD FCV failed open, cooling water flow would be very high causing drive' temperatures-to cool off quite a bit, potential thermal shock to drives would be worse than normal on a scram.
c.
Changing header pressure flow, if the CRD pumps trip off or the flow control valve fails wide open, flow to the charging header would decrease potentially causing a low pressure problem.
As such, answer a should be counted as a correct alternate answer to this question.
RED:
LOA-RD-04, page 1 4.5 The question is included in the procedures section and does not ask for a detailed explanation of the reactor physics associated with why notch override restrictions are lifted when a bypass valve is opened.
An acceptable answer should be on that says_the open bypass valve stabilized void content which is more effective in.providing negative feedback to changes in reactor power.
Also, redistribute points as indicated on corrected key.
4.12 Procedural guidance for injecting boron into the reactor has changed significantly since the NRC exam reference material had been submitted.
Guidance for injecting boren is contained in LGA-ATWS-01 and is required whenever the reactor canrat be shut down before Suppression Pool temperature reaches the Boron Injection Temperature limit (BIT curve).
.Also acceptable should be answers that correlate the points on the curve which would apply, namely greater than 5% reactor power with Suppression Pool temperature at 110 F.
During the test, the examiners indicated that the students should put down the old answer, the new answer, or both.
As such, any of these choices should be accepted for full credit.
REF:
LGA-ATWS-01, page 5 of 13 and curve.LGSA-G7 L
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FACILITY COMENTS '
SRO Exam, November-18, 1985 Reviewed by:
!J.~Harrington,;R0/SR0 Cert. Training.
R. Armitage, SRO Operating D..Crowl, SRO Operating
~S. Harmon, SRO Training s
The facility' reviewers feel overall', that this was a well written examinatio'n.
For the most part, questions were operationally oriented,
' written' clearly, and.were a good test of knowledge. We felt this showed improved effort on the part of the et.aminer, to produce a better quality examination.- However, even more effort is needed.to make sure questions are written clearly enough to illicit.the " correct answer" as listed on exam key.
Also, the exam keys should'be more clearly reviewed so answers correspond to.
the question asked.
5.03a-
-Increase 5.03b-Increase 5.03c Decrease Full credit.should be-given because this " literally explains".the ef.fect on critical power due to the change of the given plant.
. parameter..The question.does not ask "why" critical power changes v
'and therefore, the student should not-be' penalized for not doing so.
'5.06a Rod worth will increase because of the density change in the moduator as temperature increases.
Decreased density causes longer slowing down and thermal diffusion lengths.
Hence, thermal neutrons have a higher probability of reaching a control rod from the fuel.
'5.06b Rod worth will decrease as voids increase,-because voiding.
suppresses thermal neutron flux.
With decreased thermal flux (lower power) it will also: result in decreased rod worths, as control rod blades' absorb thermal neutrons.
4 J
5.11a Full credit should be given for part a, if the applicant states that the change is due'to buildup of fission product. poisons in the core.
REF:
License Requal Physics Lesson Plan, Revision 1, page 194 5.12-
. Equation. sheet given, does not have an applicable formula to work
~
this' count rate problem.- 'If' applicant assumes-a formula and works math correctly,tcredit should.be given.
-5.13b Clarify that stating, "without condensate depression, the condensate pumps would cavitate because of inadequate,NPSH"_, is the same as saying."due to the water at the eye of the pump being at saturation temperature."
'6.01a Question could be answered by applicant referencing surveillance requirements for jet pump operability according to T.S. 4.4.1.2.1.
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No clarification given for type of jet pump failure; therefore, if applicant assumes a type of failure and lists indications that support this assumed failure, full credit should be given.
6.04' Containment water volume is also used to' suppress steam from reactor upon safety relief valve actuation and should be given credit.
6.05 Part one should be deleted, because the normal drain to 14A heater from 15A is open before, during and after the hi hi level trip.
This valve does not reposition.
Our procedures, LOA-FW-01, " Loss of Feedwater Heater," or LOP-HD-02, " Placing a Feedwater Heater Into Service," do not reference the normal drain on a hi hi level trip because this valve stays open.
6.09 The answer key has more to do with internal RSCS program interlocks.
The question asks for system response at two conditions:
100% power and 1% power.
Full credit should be given if the applicant gives an accurate operational response, as given on-the corrected answer
-key.
Also, at 1% power we are <75% density not greater than.
REF:
RSCS Lesson Plan, pages 4 and 5 7.01c Technical Specifications also address trainir.g startups in T.S.
3.10.6.
This states that reactor coolant temperature must be 1200S F.
The question does not specifically state IAW LGP-1-1, therefore, 1200 F is acceptable.
7.02b "Until one bypass valve is open OR main generator-is on line." This statement should be given partial credit, as this is a pertinent part of notch override restrictions A50% rod density.
7.04a Clarify that the FW003 valve is the FRV Inlet Stop, which is manually throttled, and full H edit should be given if this normal plant terminology is used.
7.04c Full credit should be given if applicant states " maximize blowdown or reject flow." Operationally this is how we eliminate "feedwater on-off cycles by maximizing blowdown to reject excess water during heatup, which prevents the FRV from opening and closing due to level oscillations.
This in turn prevents colder condensate water from cycling "on-off to the feedwater spargers.
REF:
LGP-1-1, page 10 7.07 LOA-NB-02, Stuck Open Safety Relief Valve, also states to:
1.
Start Suppression Pool Cooling.
2.
If pool temperature exceeds 100*F, refer to LGA-03.
3.
If valve remains open after 4 attempts to close @ if:
pool temperature exceeds 110 F @ if:
two 2 minutes have elapsed since valve opened, then scram the reactor.
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a These are^all. acceptable answers, because the. question asks for Operator actions on a stuck open SRV. The' answer key addresses how to close the SRV. These are'all part of LOA-NB-02, immediate:
Operator actions and therefore, should be acceptable for. full credit.
7.08 LGA-ATWS-01 directs Operator to_ inject SBLC if the Boron Injection ~
Temperature (BIT) Curve-isexceeded. Full credit to applicant for
~ his response should be given.
t 7.10.
Point values for this question are excessive; however, a good a' & b
_ question. However, in 7.10b, the setpoints as _ listed on the answer
~ key are not specifically asked for and should not be needed.for full credit. Also, in 7.10a, 71%.in 2 minutes and 23% in_3.5 minutes are' acceptable answers'versus exact amperage.
-REF: System Description, Chapter 45, page 11
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18.05b A second licensed Operator _ specifically means another licensed Reactor Operator, {LR Senior Reactor Operator.
l-4 12
ATTACHMENT 2 RESOLUTION OF FACILITY COMMENTS SUMMERY OVERVIEW Some of the general observations are not technical and indicate only a-difference in interpretation of the Examiner Standards between the reviewers and the examiner. The examination was prepared and reviewed in accordance with the guidar provided in NUREG-1021. Resolution of specific comments about questions and answers are treated-below.
1.1 Question was considered to be worth two points. Answer key was adjusted to show value.
1.3a Consideration was given to answers that demonstrated an understanding of the concept of criticality. A just critical reactor would remain at a very low stable condtlion without significant buildup of poisons or an increase in temperature.
1.3b Question was corrected to indicate proper point value. The question is a standard question from the examination question bank and has been used many times without difficulty. The comments concerning the confusing nature of the wording, will be considered in rewriting the question.
1.4b Comment accepted. Credit was given for either 2 or 3, 1.5b Consent accepted. This question is from the examination question bank.
NUREG-1021 does not prohibit the use of non-plant specific references in Section 1 and 5 of the exarnination. The answer key will' be changed to eliminate any confusion.
1.10 The examiner does not consider the question to be " tricky." The question is from the examination question bank and has been used in the past.
Based on the abovc explanation and a review of _ the reference, the alternate answer is acceptable.
1.11 The question deals with a principle of theory of the operation of a reactor and is proper material for use in an examination. Question stands.
2.2 - Comment accepted.
2.11 Operator _ knowledge of systems that are affected by the loss of a bus is considered important to the operation of the plant. Question stands.
3.1 Both answers were listed in the answer key in order to give proper credit as long as the candidate specified which unit was being discussed.
3.2 Examiner reviewed the material and agrees that there might be some confusion concerning the question, although only one candidate gave an incorrect answer to part "b".
C e<'t was given for this answer based on a review of the reference mate:ial 13
3.3a Coment accepted. Credit was given for the alternate answers.
3.3b Coment accepted. Question will be clarified in future examinations.
~3.4 Reviewers opinion of_ point values is rejected.
3.Sc Comment accepted.
3.7 Examiner believes knowledge of this system is an acceptable topic for examination, but question is deleted because of lack of clarity.
3.8 Examiner does not believe the question violates the guidance presented in the Examiner Standards. Details in the answer key were presented to facilitate grading. Credit was given for answers.that demonstrated a knowledge of the alarms.
3.11 Alternate answers accepted. Coment concerning number of responses is noted, but had no effect on grading of question.
3.12 Comment accepted. Details are presented in answer key to facilitate grading'and prevent being criticized for~ not including intuitive conditions.
3.14-Question deleted.
3.15.No change required. Credit given for 4 correct answers.
4.1 Although reviewer states that question was confusing, all candidates answered it correctly. Question will-be clarified in future examinations.
4.2 Comment is not technical. No change to point value.
4.3 Examiner has no recollection of advising candidates to select the single most correct answer and has no notes to this effect on his copy of the examination. A review of the reference material indicates an error by the examiner in developing the question, therefore, it is deleted.
4.5 The question asks for the basis for lifting the restriction at this time. Detailed answer is given to facilitate grading. Credit was given for answers that indicated an understanding of the concept.
4.12 Comment accepted. Question graded according to material supplied.
5.02 Although there were no coments made in regards to this question, a change was made to the answer key. The answer key had two answers, either of which would be acceptable. When the point value for the question was assigned, it was inadvertently split with half credit for each answer. Each answer was given full credit if applicable.
5.03 Disagree with coment.
The word " Explain" means to relate the "cause."
Answer key remains unchanged.
14
j.
5.06a&b Coment unnecessary. The coment only reworded the answer given.in the key. Answer in the key was taken directly out of the reference material and will remain unchanged. Credit was given for reasonable answers.
-5.11a Disagree with coment. There are other fission product poiso.ns that buildup, but they do not have the same effect that samarium has,
~
5.12 Coment accepted 5.13b Comment accepted. Credit was given for answers that demonstrated an understanding of the concept.
6.01a The question asks for indications of failure at full power. The reviewer gave no reference for the different types of failures that would give d1f ferent indications.
Indications listed in the. answer key would lead to the use of the surveillance requirerrents in order to determine whether the indications are valid. Answer key stands.
6.04 Reviewer did not provide a reference for the comment. No change to answer key. Only one candidate received less than full credit for answer.
6.05 Comment accepted.
Points were not taken off for Item 1 in answer key.
6.09 Coment accepted. Credit given for-alternate answer.
7.01c Coment accepted.
It should be noted that the Technical Specifications list <200 F, while the LGP lists <212 F.
An operator could comply with the LGP and be in violation of the Technical Specification.
7.02b Coment accepted, although no reference is given.
7.04a&c Comment accepted.
7.07 Coment accepted.
7.08 Coment accepted.
7.10a&b Alternate answers are acceptable. Coment concerning point value is an opinion and has no technical basis.
8.05b Coment unnecessary. Credit given for answers that indicate an understanding -that the stand-in must be a licensed individual.
Interpretation of the Technical Specification based on the coment might lead one to believe that the SR0, when he stands in for the R0 could no longer be counted as one of the required SR0's on shift.
15
MASTER COPY
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U.S. NUCLEAR REGULATORY COMMISSION REACTOR OPERATOR LICENSE EXAMINATION i
FACILITY.
LaSalle i
REACTOR TYPE BWR DATE ADMINISTERED:
November 18, 1985 EXAMINER:
J. I. McMillen APPLICANT:
INSTRUCTIONS TO APPLICANT:
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%.
% of Category
% of Applicant's Category Value Total Score Value JY
. - 25 1.
Principles of Nuclear Power Plant Operations, Thermodynamics, Heat Transfer and Fluid Flow.
25 25 2.
Plant Design Including Safety and Emergency Systems 23 25' 25 3.
Instruments and Controls 25 25 4.
Procedures - Normal, Abnormal, Emergency and Radiological Control 91 160 100 TOTALS Final Grade All work done on this exam is my own, I have neither given nor received aid.
Applicant's Signature
Section 1 - Principles of Nuclear Power, Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 1.1 The most common unit of flow is gallons per minute (gpm).
In a BWR, feedwater and steam flow are measured in pounds per hour (1bm/hr). Why is 1bm/hr used rather than gpm?
(2.0) 1.2 What are four (4) design or operational factors that ensure Net Positive Suction Head (NPSH) for the recirculation pumps?
(2.0)
' 1.3 A. reactor startup is incorrectly attempted without a sufficient neutron source to give indications on the nuclear instrumentation:
a.
Describe the response of the reactor if the operator stopped pulling rods at-the just critical position.
(1.0) b.
Describe the response of the reactor if the operator stopped pulling rods one notch above the critical
/4 position.
(4re) 1.4 The reactor is shutdown with a Keff of 0.9Ak/k and the source range monitor indicates 100 cps.
Rods'are withdrawn and the source range now indicates 200' cps.
Choose the correct answer for each of the.following questions.
a.
The new Keff will be:
(1.0) 1.
0.93Ak/k 2.
0.95Ak/k 3.
0.97Ak/k 4.
0.99Ak/k b.
The amount of reactivity added was:
(1.0) 1.
0.0449Ak/k 2.
0.0526Ak/k 3.
0.0585Ak/k 4.
0.0635ak/k c.
If the same amount of reactivity were added again, the reactor would be:
(1.0) 1.
Subcritical 2.
Critical 3.
Supercritical' l
4.
Prompt critical i
1.
1.5 Using the following figures, choose the correct answer for each of the~three (3) questions asked below:
100 rene I
o Time A1 A2 112 Eenon Cl 01 h1 C'
Time a.
What is the approximate time from Al to A2?
(1.0) 1.
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 2.
30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> 3.
50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> 4.
70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> b.
What is the approximate time from B1 to B2?
(1.0) 1.
1-3 hours 2.
2-6 hours 3.
5-7 hours 4.
6-9 hours c.
Why does Xe concentration decrease from Al to D1?
(1.0)
~
1.
Xenon decay is equal to iodine decay 2.
Xenon burnout is equal to iodine decaying to Xenon.
3.
Xenon burnout is greater than iodine decaying to Xenon 4.
Xenon decay is greater than iodine decay 1.6 Answer the following statements about the Doppler Coefficient, True or False.
a.
Doppler Coefficient becomes ",, ore negative from 0 to 100% power due to the increased overlapping of resonance peaks at higher fuel temperatures.
(1.0) b.
Doppler Coefficient becomes more negative over core life due to the buildup of Pu-240 and fission products with large resonances in the epithermal range.
(1.0) 2
1.7 A reactor startup is in progress.
You have been given the estimated critical rod position for the conditions at 1200.
You start to pull rods at 0900 for the approach to critical.
How would each of the following conditions or events affect the actual critical rod position compared to the estimated critical rod position?
(Each condition or event should be considered independently of all others.) Answer as: more rod withdrawal, less rod withdrawal, or no significant effect.
a.
One reactor recirculation pump is stopped (0.5) b.
Shutdown cooling is stopped (significant decay heat)
(0.5) c.
Reactor head vent is inadvertently closed (0.5) d.
Moderator temperature gradually decreasing (0.5) e.
Reactor water cleanup system isolates (significant decay heat)
(0.5) 1.8 Match the following terms with the proper definition terms.
.a.
Natural circulation (0.5) b.
Saturated liquid (0.5) c.
Enthalpy (0.5) d.
Departure from nucleate boiling (0.5)
Definitions 1.
The point of which partial film boiling begins 2.
The point on the boiling curve where heat transfer surface temperature will rise sharply with little or no increase in heat flux 3.
The movement of a fluid base and the difference in density of the fluid caused by a differential temperature 4.
The total energy of a substance per unit mass i
~
5.
A~1iquid that cannot absorb any more energy without starting to vaporize 6.
The internal energy of a system tha!; is no longer available to do work 3
1.9 Answer the feilowing TRUE or FALSE a.
The. differential temperature necessary to transfer heat is inversely proportional to heat flux.
(0.5) b.
Pump runout is the term used to describe a centrifugal pump when it is operating with its discharge valve closed.
(0.5)
The-latent heat of vaporization is.another tefm for c.
the latent heat of condensation. hp tt* ')
(0.5) d.
One of the pump laws for centrifugal pumps states that power required by the pump motor is directly proportional to the square of the pump speed.
(0.5) e.
The faster a centrifugal pump rotates, the greater the NPSH required to prevent cavitation.
(0.5) 1.10 Decay heat is primarily produced from:
(0.5) a.
Neutron sources b.
Delayed neutrons slowing down in the fuel c.
Fission products slowing down in the fuel d.
Beta particles slowing down-in the fuel 1.11 Consider two operating history cases for~the same plaat.
In Case "A" the year is 1986 and the plant operates for 11 months at 50% power.
In Case "B" the year is 1987 and the plant operates for 11 months at 100% power.
a.
The samarium level in Case "A" is times that in Case "B."
(Select the answer)
(1.0) 1.
1/2 2.
4/5 3.
1 4.
2 b.
If the reactor is refueled arior to each of these operating history cases witi 40 new assemblies.
It would take
~in Case "A" compared to the time in Case "B" to reach the equilibrium samarium value for the new elements.
(1.0) 1.
half as long 2.
the same time 3.
twice as long 4.
four times as long END OF SECTION 1 4
Section 2 - Plant Design Including Safety and Emergency Systems 2.1 Give four reasons for using bottom entry control rods in a Boiling Water Reactor.
(GE likes the design is not one.)
(2.0) 2.2 The Standby Liquid Control line' serves several functions other than poison injection.
List four of these functions.
(2.0) i 2.3 a.
The scram discharge volume systems for Unit 1 and Unit 2 are different.
What are these differences?
(1.0) b.
What recent modifications have been made to the instrumentation on the scram discharge volumes?
Explain why this was done.
(2.0) 2.4 The signals to the RPS for Unit 1 and Unit 2 are different.
Explain this difference.
(1.0) 2.5 Each fuel bundle is surrounded by a fuel channel and these channels serve several functions.
List four of these functions.
(2.0) 2.6 Explain why it is necessary to have the turbine " slaved to" the reactor.
(1.0) 2.7 With the plant operating at 100% power what would be the effect on the plant if a.
the load limit is reduced to 75%
(.
3) b.
the max combined flow is reduced to 75%
(1.5) 2.8 The reactor is in cold shutdown.
The Reactor Water Cleanup System (RWCU) and Recirculation System thermocouples are indicating 180*F and 190 F, respectively.
Stea:n is issuing from the reactor head vent.
Explain how this could happen.
(Assume all instrumentation is properly calibrated.)
(2.0) 2.9 A Primary Containment Group Isolation signal is received and all but one of the associated isolation valves begin t close.
How does the PCIS status display-indicate that a valve is stuck full open?
(1.0) 2.10 For the following Reactor Core Isolation Cooling (RCIC)
System motor-operated valves, indicate whether the power supply is AC or DC.
(See attached drawing) a.
Inboard Isolation Valve (E51-F063)
(0.5) b.
Outboard Isolation Valve (E51-008)
(0.5) c.
Turbine Exhaust to Suppression Pool (E51-F068)
(0.5) d.
Warmup Bypass Isolation Valve (E51-F076)
(0.5) e.
Turbine Steam Supply Valve (E51-F045)
(0.5) f.
Pump Discharge Valve (E51-F013)
(0.5)
L
2.11 Match the following events with the loss of the appropriate breaker.
(1.5)
Events a.
Loss of Control Power (close/ trip) to Generator Excitation Field Breaker (0.25) b.
Loss of Narrow Range Pressure Indication on H13-P603 (0.25) c.
Loss of Main Steam Line "D" Flow Signal to Total Steam Flow Recorder (0.25) d.
Disables Narrow Range "B" Level Instrumentation (indication and signal to Reactor Level Control System)
(0.25) e.
Loss of Generator / Transformer Trip System 1 (0.25) f.
' Disables Backup Scram Valve F110A (0.25)
Bus 1.
111x 2.
211y 3.
212x 2.12 Regarding the end-of-cycle recirculation pump trip (E0C-RPT):
a.
What two conditions will initiate this trip?
(1.0) b.
What is the basis for the E0C-RPT?
(1.0) c.
When is this trip bypassed?
(0.5) 2.13 Unit 1 is operating at 80% power with the "A" TORFP and the MDRFP controlling level in three element automatic operation.
The "B" TDRFP is in manual and the operator depresses the automatic pushbutton on the M/A station for the "B" pump.
l a.
-Explain the effect this will have on the Reactor Water Level Control.
(0.5) b.
Why does this happen?
'(1.0)
END OF SECTION 2 2
i 1
7-s Section 3 - Instruments and Controls 3.1 True or False A low charging water pressure will initiate a reactor scram if the " test" switch is placed in the TEST position with the mode switch in " REFUEL-MODE."
(0.5)-
3.2 True or False The RWM will generate an " insert block" upon receipt
'of the third insert error, under the following
-conditions:
a.
When a withdraw error is received, resulting in a rod block and a rod other than the error rod is inserted (0.5) b.
During power down maneuvering, when steam flow drops below 20% for at least 5 seconds and the operating sequence is not yet latched (0.5) 3.3 The recirculation pumps have an interlock which prevents increasing pump speed above the low speed unless certain conditions are met.
a.
What are these conditions?
(1.0) b.
Explain the basis for this interlock?
(1.5) 3.4 Reactor Water Level No. 4 (31.5") has two functions.
a.
What are these two functions?
(0.5) b.
What is the basis for these functions?
(2.0) 3.5 a.
Why is the " flux estimator" used for the flux controller rather than the actual APRM signal?
(1.0) b.
What two conditions will automatically initiate a switchover to the actual APRM signal from the flux estimator? (Give applicable setpoints)
-(1.0) c.
What two alarms are generated by the flux estimator?
(1.0) 3.6 Match the ADS auto initiation signals (Column A) with the proper ~setpoint (Column B).
(2.5)
Column A a.
High drywell pressure
'k.
119 psig b.
Low Reactor Water Level 2.
-50 inches c.
Confirmatory low water level 3.
250 psig d.
LPCI/L995 discharge pressure
4.
105 seconds
e.
Time delay 5.
1.69 psig 6.
150 seconds
/#3 7.
-1249inchesd-
'h 8.
2.00 psig 4.
12.5 inches 7 *at it the fsnction of the Substitute Position
, Generator?
(0.5) 3.8 a.
Describe two conditions,.other than high level in the Drywell Equipment drain sump, that will cause the "DRYW EQUIP DRAIN SUMP TROUBLE" annunciator on panel IPM13J to alarm.
(2.0) b.
What do'es this alarm indicate to you.
(0.5) 3.9 When the reactor building exhaust Plenum Radiation Monitor system setpoint is exceeded, what automatic actions occur?
(1.5) 3.10 List six systems that can be controlled from the Remote Shutdown Panel.
(3.0)
Ju.
3.11 What are the loads supplied by Division I of.the CSC -ESW Systam?
g J-il.5) 3.12 What conditions must be present to caus'e the LPCS
. injection valve to automatically open? Include setpoints.
(2.0) 3.13 The ADS initiation logic has a time delay in it. What is the purpose of this time delay.
(1.0) d'# /#[' 4M7 rue or False A withdraw block is applied to all rods not having insert errors when the fourth insert error is receive,d.
' 3.15 The Reactor Water Level Control System receives its level input from either the A or B transmitter.
The selected transmitter provides a level signal for several other functions. What are these functions? (4 required)
(1.0)
END OF SECTION 3 2
Section 4 - Procedures - Normal, Abnormal, Emergency and Radiological Contrel 4.1 a.
Which of the following conditions would indicate Nu a failure of the #1 seal, #2 seal or both seals? '
And indicate which seal or seals have failed.
(1.5) 1.
The #2 seal pressure higher than normal and staging flow through the second seal is unchanged.
2.
The #1 and #2 seal pressure is higher than normal and staging flow through the second seal is unchanged.
3.
The #1 and #2 seal pressures decrease and staging flow is lower.
4.
The #2 seal pressure is lower and staging flow through the second seal increases.
4.2 a.
What parameter changes would indicate that an off gas
'~
hydrogen explosion had occurred? (List 4 parameters with their expected change where applicable.)
(2.0) b.
What are four of your immediate actions if the explosion has occurred?
(2.0) 4.3 Which of the following would indicate a failure of the control rod drive flow control system?
(1.0) a.
Rod drive - High temperature b.
Rod drive - Low temperature c.
Charging water - Header pressure low j
d.
Charging water - Header pressure high 4.4 When operating in automatic level control and the Reactor Water Level Control system fails,.why is it necessary to l
switch all M/A stations to manual prior to switching to the backup level control instrument? (That is, explain the l
consequences of not performing the change to manual on the M/A stations.) (Assume failure was due to loss of level sensing instrument.)
(1.5) 4.5 During startup and heatup the procedurally enforced restriction on withdrawal of control rods is lifted after a bypass valve opens.
Explain the basis for lifting this restriction under those conditions.
(1.5)
\\
L
4.6 Which of the following Process Radiation Monitoring Systems utilize Geigh-Mueller detectors?
(0.5) a.
' Main Steam Line Radiation Monitors b.
Service Water PRM c.
Reactor Building Closed Cooling Water PRM 4.7 You observe that a SRV has spuriously actuated and is stuck open.
a.
Briefly explain two methods to close the valve as stated per LOA-NB-02.
(0.5) b.
How would you determine that the valve has closed?
(1.0)
(Four indications necessary for full credit.)
c.
A manual scram is required if a SRV is stuck open and any one of three conditions exist.
What are these conditions?
(1.5) 4.8 The Inboard Shutdown Cooling Containment Isolation Valve (E12-F009) is normally powered from Division II switchgear with the alternate source provided by Division I.
When using the alternate power source are the following precautions TRUE or FALSE?
a.
The Outboard Shutdown Cooling Isolation Valve (E12-F008) must be closed when using the alternate power supply for the inboard valves.
(0.5)-
b.
There is no protective c.ircuitry to prevent feeding the valve motor operator from both the normal and' alternate power supplier.
(0.5) c.
The F009 valve will not automatically isolate when the valve is powered form its alternate power source.
(0.5) 4.9 During an ATWS, LGA-4 directs you to maintain the water level using condensate /feedwater, control rod drive and/or reactor core isolation cooling systems.
a.
Why is the use of these systems preferred over the high pressure core spray, low pressure core spray and low pressure coolant injection systems?
(1.25) b.
Explain the possible adverse consequences involved with using the HPCS, LPCS,'or LPCI under ATWS conditions.
(1.25) 2
?
4.10 Arrange the followi.ig LGA-03 (Primary Containment Control) operator actions in order of INCREASING Primary Containment Pressure.
(1.5) a.
Emergency Depressurization of the RPV b.
Perform RPV Flooding per LGA-05 c.
Restart a VP Chiller per LOA-VP-02 d.
Initiate Suppression Chamber sprays e.
Shutdown RR Pumps, D/W fans and initiate D/W sprays 4.11 LOA-FP-04, " Loss of Both Diesel Fire Pun.ps With a Loss of Offsite Power," requires the starting of at least one service water pump as an immediate action. What are the four immediate action steps necessary to start a service water pump?
(2.0) 4.12 Although LOA-NB-09, " Transient With Failure to Scram,"
contains a note that says the operator is not required to verify the exact number of control rods not properly inserted prior to performing immediate actions, the System Training Manual says the operator has authority to inject boron (SBLC) under certain reactor conditions. What are these conditions?
(2.0) 4.13 One of the entry conditions for LGA-02, " Secondary Containment Control," is "Any of the following parameters above the maximum normal operating value."
What are these parameters? (Values not required)
(2.5) 4.14 LGA-03, " Primary Containment Control," directs the operator to vent the containment regardless of release levels when drywell pressure exceeds a.
What is this pressure?
(0.5)
~
b.
What component is the limiting factor in setting this pressure limit?
(0.5) i END OF SECTION 4 l
l END OF TEST l
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.5 -
ANSWERS - Section 1 -. Principles of Nuclear Power, Plant Operation, Thermodynamics, Heat Transfer and Fluid Flow 1.1 Steam is a vapor'and one gallon of steam contains only a f fraction of the mass of one gallon of water (4s49); w (/d ftablishanenergybalanceacrossthereactor(4seph d.ufdd the feedwater energy in (1bm/hr) (BTU /lbm) is compared + d # g
/
i(with the steam enerav out (1bm/hr) (BTU /lbm) (4s6),)
Establishing flows in lbg/hr allows a true comparison of the two flows and a s4mple determination of whether mass is increasing or decreasing in the reactor vessel (1>.46-).
/0 Ref.:
Standard Heat Transfer Principles 1.E 1.
Located far below the normal water line to provide a large static head 2.
Feedflow less than 30%; trip to slow speed 3.
At high power operation, feedwater subcooling provide adequate NPSH 4.
Low reactor water level trips, cavitation interlock 5.
Suction valve closed trip, cavitation interlock 6.
AT steamline pump suction < 10 F; trip to slow speed 7.
No transfer to fast speed unless feedwater flow
> 30% and flow control valve is at minimum q be @ 0.5'en.
Ref.:
Standard Fluid Flow Principle
' 1. 3 a.
A chain reaction will be initiated by the neutrons in the core and power level will increase (0.5) but stabilize as some very low level (0.5).
b.
A chain reaction will be initiated and super critical multiplication will result from the pocitive reacting addition associated with the rod notch withdrawal (0.5).
Power will increase (0.25) on a period depending on the notch worth (0.25) until a safety function scram the reactor (0,25) or until sufficient negative reactivity (temp or void) (0.25) is added to stop the power increase.
Ref.:
Standard Nuclear Principles 1.4 a.
-2 b.
3 eta.
c.
3 each @ 1.o
Ref. :
Standard Nuclear Principles 1.5 a.
3 b.
2 vL 3 c.
3 eu.h @S0)
Ref.:
Standard Nuclear Principles 1.6 a.
False b.
True
<a o (/.J)
Ref. :
Standard Nuclear Principles 1.7 a.
No significant effect b.
More rod withdrawal c.
No significant effect d.
Less rod withdrawal e.
More rod withdrawal 9 g,5.)
Ref. :
Standard Nuclear Principles 1.8 a.
3 b.
5 c.
4 d.
1 enth @ to.()
Ref. :
Standard Heat Transfer Principles 1.9 a.
False b.
False c.
True d.
False e.-
True
_,,g g g,f)
Ref.:
'~
I #'
1.10 d dt O l
Ref.:
l -
1.11 a.
3 b.
3 Q
Ref.:
I-END OF SECTION 1 I
2
ANSWERS - Section 2 - Plant Design Including Safety and Emergency Systems 2.1 a.
Less time required for refueling
~
b.
Internal moisture removal and steam separation can be more easily accomplished
~
c.
With a large % of voids in the upper part of the core power is reduced in this area.
If rods entered from the top, flux would be overdepressed.
O' d.
~ Bottom entry rods can be used for axial power shaping.
Helps control flux peaking in local area of core and yields optimum burnup.
Ref.:
LaSalle Training Manual, Volume II, Chapter 7, p. 8 2.2 a.
Core AP measurement b.
HPCS line break detection c.
Jet pu:np AP tap d.
RWCU bottom head drain flow tap o#a a.
e CR0 drive water and cooling water AP tap yl
.4, e E
E*M Nm.DicQ-te A mM LaSalle Trainin Ref. : ' 4.6R - 0V f-S. g Manual, Volume I, Chapter 2, p. 56 2.3 a.
Unit 2 has 2 vent and 2 drain valves in series.
Cl 0) b.
Two additional scram level switches were added. These Q 0) are AP instruments to provide diversity ir, measurement.
Ref.:
LaSalle Training Manual, Volume II, Chapter 8, p. 38 2.4 Unit 2 receives a signal from the CRD system which is (l.c) effective in "Startup" and " Refuel."
Ref.:
LaSalle Training Manual, Volume II, Chapter 8, p. 70 2.5 1.
Channels the coolant flow upward (90% through the channel 10% bypassed) 2.
Bearing surface for control rod blades 3.
Protection for fuel rods during handling 4.
Primary resistance to lateral acceleration loading on the assembly 5.
Insures correct control rod clearance 6,y o.r,..
Ref.:
LaSalle Training Manual, Volume I, Chapter 4, p. 14
2.6 A pressure control system was necessary so that reactor pressure is first changed and then turbine power (0.25) because an increase in reactor pressure would cause voids to compress and collapse (0.25) increasing moderator content (add reactivity)
(0.25).
This power increase would cause pressure to increase further and cause a possible divergent transient (0.25).
(The opposite is also true', i.e., lower power decreases pressure more voids,etc.)
Ref. :
LaSalle Training Manual, Volume IV, Chapter 26, p. 12 2.7 a.
Turbine control valves close, bypass valves open to
(. l 0) pass 25%.
b.
Control valves close.
Bypass valves do not open.
Reactor k'
pressure increases; voids collapse, adding reactivity power goes up.
Cycle repeats.
Reactor scrams: q high neutron flux err high pressure.
e W
Ref. :
LaSalle Training Manual, Volume IV, Chapter 26, p. 12 2.8 It is possible to have thermal stratification in the core (1.0).
Lacking sufficient circulation, the RWCU bottom drain and recirculation loop thermocouples will read significantly lower than the temperature at the surface of the water in the vessel (0.5).
If insufficient heat is-removed, this condition will result in raising the surface temperature above the boiling point (0.5).
Alternate If vessel level is not above steam separator drain lines (1.0) and you decrease shutdown cooling flow (0.5), there may not be adequate flow through the core to prevent surface temperature from rising to boiling point (0.5).
Ref. :
LaSalle Training Manual, Volume I, Chapter 2, p. 84 l
2.9 A valve not moving would remain flashing indicating that it (s. o )
was stuck open.
Valves that reach 90% open position are solid amber.
Ref.:
LaSalle Training Manual, Chapter 49 l
2.10 a.
AC b.
AC c.
DC d.
AC e.
DC f.
DC g,g Ref.:
LOP-R1-01E 2
2.11'a.
111x I
-b.
211y 1 c.
211y J d.
212x f e.
111x i f.
211y J g [o,t()
Ref.:
LOA-DC-02, pgs. 5/10 2.12 a.
1.
Turbine trip (closure of Main Turbine Stop valves) 2*9 2.
Generator load reject (fast closure of the main (0, f)
Turbine Control RETS pressure low)
(Full credit will not be given if only items in one are listed) b.
Improve MCPR consequences for the above transients (i.e.,
g,0 )
pressure transients) or recover loss of thermal margin which occurs at end of cycle c.
< 30% reactor power (30% first stage pressure)
/0*C 2.13 a.
No change (o. d b.
Since "A" TDRFP and MDRFP are already in automatic and the RWLC system is interlocked so that no more
[/, al than 2 components can be controlled automatically in 3 element control the third component does not become part of the auto control.
Ref. :
System Training Manual, Chapter 31 END OF SECTION 2
- 1
?
3
ANSWERS - Section 3 - Instruments and Controls 3.1 True, Unit 2; False, Unit 1 (O'h Ref.:
LaSalle Training Manual, Volume II, Chapter 8, p. 40 (o4) 3.2 a.
True b.
True (a.C Ref.:
LaSalle Training Manual, Volume II, Chapter 18, p. 24 Feedflowf30%andFlowControlValveatminimum O'
3.3 a.
b.
Cavitation in flow control valve due to inadequate subcooling (0.5).
If fast speed were permitted there would be velocity increase (0.25) and a pressure decrease (0.25) at the valve.
Bubbles would form in the valve and collapse at they exit causing valve damage (0.5).
Ref.:
LaSalle Training Manual, Volume I, Chapter 5, pgs. 70-71 3.4 a.
1.
Low level alarm
[O.
2.
Recirculation flow control valve runback permissive b.
1.
Level below which steam carry under would significantly affect the subcooling available for the recirculation pump NPSH.
(1.0) 2.
With less than 2 feed pumps operating, the recirculation Flow Control Valve will auto close to a minimum position to reduce power within capability of one feed pump.
(1.0)
Ref.:
LaSalle Training Manual, Volume I, Chapter 3, p. 36
' 3. 5 a.
APRM signal is noisy and would cause the flow control (l.o) valve to oscillate or (flux estimator. signal is almost noise free providing steady signal to FCV).
(o.d b.
1.
Actual APRM signal reaches 110%
2.
Actual flux and estimated flux differ by more than 5%
@'c)
Mb c.
1.
APRM raw signal selected for continuous 20 minute / k (o,f) period If logic switches between raw APRM and estimated 8 uf'A $ j Mi k O
L' 2.
g signal more than 200 times in 5 minutes Ref.:
LaSalle Training Manual, Chapter 6, p. 16 & 22
3.6 a.
5 b.
7 c.
9 d.
1 7
g(e,[/
e.
4 b2My7 cat)
Ref.:
System Training Manual, Chapter 37, p. 26 3.7 lhis is a deviceps t% posiUun data Trom the RPIS f
(ide Anti ^ ~ use seie u ed red end red drivi,g @ ls) to generate the next expected position of the rod being driven.
Ref.:
Systems Training Manual, pgs. 40 & 41 3.8 a)
Pump starts; a timer actuates (0.25); timer interval corresponds to time for pump to pump 20 gpm leak (0.5);
if pump runs past this time alarm sound (0.25).
If pump stops prior to this time (0.25) a second timer starts (0.25). This timer looks at restart time (0.25).
If pump starts before timer runs out (0.25).
b)
Alarm indicates leak has progressed to 20 gpm (0.5).
Ref.:
Systems Training Manual, Chapter 73, p. 26 3.9 a.
Start SBGT and associated samples (0.5) b.
Isolate Reactor Building ventilation (0.5) c.
Close primary containment vent and purge (0.5)
Ref.:
LaSalle Systems Training Manual, Chapter 72, pgs. 94 & 95 3.10 RCIC; RHR Loca_B (does not include valves for Steam Condensing mode), RHR service water pumps C and D and B Hx outlet valve (688), three safety relief valves H, K, P; recirculation pumo A suction valve;RBCCW pump 1A; a"*1,4 sarvice watar jnckay map OA and strainer outlet valve
- g'j (1WS113); plant evacua. tion siren.
Ref.:
LaSalle Systems Training Manual, Chapter 74, pgs. 40 & 41 3.11 RHR Heat Exchanger A; UnFt 1 and 2 LPCS pump'; motor coole RHR pump A _ seal cooler Diesel y
'*1 aenerator "0" cooler; m
TJnit 1 and 2 NF an NW area cooler; fiisl ~p~561 emergency makeup Mb supply A.
Ref.:
LaSalle Systems Training Manual, Chapter 40, p. 52 & 53 3.12 Power available to 141y; Level 1 (-129"); 1.69 psig drywell pressure;lessNhan500psigreactorpressureand<500psig between F005 and F006.
(2.0)
L'Salle Systems Training Manual, Chapter 38, pgs. 52 & 53 Ref.:
a 2
4 3.13 Time delay is to assure the HPCS has enough time to operate Q'
yet not so long that the LPCS/LPCI. systems would be unable to cool the core.
Ref.:
LaSalle Systems Training Manual, Chapter 37, pgs. 52 & 53
]!
'3:14 False-[)jg(c-Ref.:
LaSalle Systems Training Manual, Chapter 18, p. 46 3.15 1.
Upset / narrow range recorders 2.
Recirculation runback permissive (alarm / trip) 3.
Pump down shift (alarm / trip)
Q
- 2. [m 4.
High/ low level alarm
]
Ref.:
LaSalle Systems Training Manual,. Chapter 3, p. 14 END OF SECTION 3 i
l l
l l
3
ANSWERS - Section 4 - Procedures - Normal, Abnormal, Emergency and Radiological Control M
b,q[
4.1 a.
4 b.
2 Ref.:
LaSalle Training Manual, Volume I, Chapter 5, p. 30 4.2 a.
Sudden abnormal change in prefilter hP, after' filter 4
AP and off gas system pressure; high off gas system flow; A4 (,o.d high pretreatment radiation monitor; high charcoal 48 adsorber vault temperature; high area radiation in off gas.
b.
1.
Initiate load reduction (LOA-GG-01) 2.
Warn all personnel to stay clear of off gas areas 3.
Notify shift supervisor and Load Dispatcher d,
4.
Check "Adsorber Bed High Temperature" alarm (clear)
A,ha.$ )
5.
Check all off gas area radiation monitors 4.*
Ref.:
LOA, Volume 2, LOA-0G-02 4.3 Ref.:
LOA, Volume 2, LOA-R0-04 4.4 Switching to other level instrument prior to transferring all M/A stations to manual will cause excessive reactor
-level swings (0.5) and feedwater pipe vibrations (0.5).
This is due to the large variation that exists between the failed instrument and the backup instrument (0.5).
Ref. :
LOA, Volume 2, LOA-RL-01
. 4. 5 At that condition the core would contain voids (0.25) and the voids suppress the flux (0.25) thus improving the peak to average flux ratio (0.25).
This. reduces the notch worth of all rods (0.25).
The open bypass valve stabilizes pressure and the void' content is also more stable (0.25) and much more effective in providing negative reactivity feedback (0.25).
Re f. :
LaSalle Tiaining Manual, Volume III, Chapter 19, pgs. 40 & 41 4.6 c.
(o,S)
Ref.:
LaSalle Training Manuai, Chapter 72, p. 32
4.7 a.
1.
Cycle control switch from auto to open and (2Q back to auto 2.
Remove fuses for affected valve to deenergize
(.2b control circuit
-Alternate 3.
Operate keylock for ADS valve 4.
Hand switch at remote panel for HKP b.
Valve position indication (clear); leak detector;
,g generator load; steamflow; no rumbling noises; down 4pf comer piping temperature normal; suppression pool normal; radiation levels normal.
1.
Valve remains open for 2 minutes fo4) 2.
Valve remains open after a maximum of 4 attempts (0 5')
to close 3.
Suppression pool temperature reaches 110 F Ref.:
LOA, Volume I, LOA-NB-02 4.8 a.
F h(0**d b.
T p
c.
T Ref. :
LOP-RH-07, p. 4 Volume VIII, Chapter 49, pgs. 40 & 62 4.9 a.
Condensate /feedwater, control rod drive and reactor core isolation cooling inject outside the shroud (0.5) which allows for mixing of the cold unborated water being injected (0.5) with the warm water in the down comer prior to entering the core (0.25).
b.
Use of HPCS, LPCS or LPCI injects cold water directly into the core (1.0) and could cause a power surge (0.25).
4.10 c, d, e, a, b-or c, d, a, b, e edch (d'h Ref.:
LGA-03, Step C.4 4.11 a.
Dispatch operator to close discharge valve on selected
/o d pump.
b.
Restore power to bus by closing cross tie breakers
@h
[141x(241x) and 142x(242x) - ACB 1415(2415) and
.ACB 1425(2425)].
2
4, c.
Start selected service water pump.
(Select pump (b4)
I according to DG availability.
Do not exceed amp 494 or load 2860KW.)
@'b
- d.
Open service water pump discharge isolation valve.
Ref.:
LOA-FP-04, pgs. 1-2 4.12 Rx water level not maintained (0.5). or suppression pool water temperature reaches 110*F (0.5); an W ither a.
5 or more adjacent rods not inserted to at least 06 or (0.5) b.
30 or more rods not inserted to at least 06 (0.5)
Ref.:
LOA-NB-09, p. 2 LaSalle Systems Training Manual, Volume II, Chapter 10,
- p. 48 4.13 Area temperature (0.5) area differential temperature (0.5) area radiation level (0.5). Area water level (0.5); Rx building or Fuel Pool Cooling ventilation exhaust radiation level (0.5).
Ref.:
LGA-02, p. 1 4.14 a.
60 psig ed(84) b.
SRV actuators Ref.:
LGA-03, p 15 END OF SECTION 4 3
s.
k &
U.S. NUCLEAR REGULATORY COMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION FACILITY:
LaSalle REACTOR TYPE BWR DATE ADMINISTERED:
November 18, 1985 EXAMINER:
T. E. Lang APPLICANT:
INSTRUCTIONS TO APPLICANT:
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%.
% of
~
"~
Category
% of Applicant's Category Value Total Scorc Value 25.5 25.8 5.
Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 25 25.2 6.
Plant Design, Control and Instrumentation 25 25.2 7.
Procedures - Normal, Abnormal Emergency and Radiological Control
-23.5 23.8 8.
Administrative Procedures, Conditions and Limitations 99 100 TOTALS Final Grade All work done on this exam is mj own, I have neither given nor received aid.
Applicant's Signature
Theory of Nuclear Power Plant Operations, Fluids and Thermodynamics 5.01 For the following condition, state if the Fuel Temperature Coefficient becomes more or less negative.
a.
Increase in fuel temperature (0.5) b.
Increase in moderator temperature (0.5) c.
Increase in void fraction (0.5) 5.02 Define the following items:
a.
Latent. Mat of vaporization (1.0) b.
Critical Point (1.0) c.
Satura;ed Liquid (1.0) 5.03 Explain the effects of increasing the following core parameters oa steady state critical power.
a.
Core flow (1.0) b.
Inlet subcooling (1.0) c.
Reactor pres:ure (1.0) 5.04 If equilibrium xenon is obtained (the reactor has been operating at constant power for many hours), and the reactor power is doubled, will the new equilibrium xenon concentration be twice as great? Explain your answer.
(2.0) 5.05 You increase core power by pulling control rods around the center fuel bundle. Assuming that recirculation flow is kept constant would the flow through the center bundle increase, decrease, or stay the same? Explain your answer.
h..,
4 g4 (2.0)
-5.06 Explain how and why Rod Worth changes for the following conditions.
a.
Increase in moderator temperature (1.0) b.
Increase in void fraction (1.0)
t 5.07 Your reactor operator informs you that MAPRAT is 1.02
).
a.
Is the MAPRAT, as stated, conservative?
Explain your answer.
(0.75) b.
In regards to MAPRAT, which of the following (0.75) statements are True and which are False?
I 1.
MAPRAT maintained within limits ensures that transition boiling will not occur in 99% of the fuel bundles.
2.
Maintaining MAPRAT limits ensures that the MFLPD limits are met.
3.
Maintaining MAPRAT limits ensures that peak clad temperature will not r,each 2200'F during a LOCA.
5.08 a.
What are three coefficients of reactivity and what is a typical value of each?
(1.5) b.
How are the above coefficients of reactivity 4
affected by core age? (Less negative, more. negative or remain the same)
(1.5)
J 5.09 A fluid is at 400*F and 600 psig.
Is it subcooled or super heated and by how many degrees?
(1.0) 5.10 If reactor power level is increased from 50 MW to 370 MW in two minutes, what is the doubling time? (Show all work)
(2.0) 5.11 Using.the attached enrve:
~
a.
Why does core reactivity decrease from point A to point B?
(0.5) b.
Why does core reactivity increase from point B to point C?
(0.5) c.
Why does core reactivity decrease from point C i
to point D?
(0.5) 5.12 With an initial Keff of 0.9, sufficient control rods are withdrawn to increase the count rate by a factor of 10. What is the final Keff of the core?
(2.0)
I l
I 2
1 5.13 a.
Define " Condensate Depression."
(0.5) b.
Is it necessary for plants to operate with
" Condensate Depression?" Explain your answer.
(0.5) i
~
END OF SECTION
~
9'
+w
's@
i t
3
b Plant Design, Control and Instrumentation 6.01 a.
What are four indications of a failed jet pump?
Assume full power conditions.
(2.0) b.
Why is jet pump integrity important in regards to
~
accident conditions?
(0.5)'
6.02 What are the five ranges of vessel level instrumentation?
Be specific and include setpoints. 4/f 4 A h % 4 d (2.5)
LJ w w=
-A.
6.03 In regards t} Recirculation Pump Cavitation Interlocks:
a.
What are the two cavitation interlocks which will trip the pumps from 100% to 25% speed?
(2.0) i b.
. Are the interlocks above actuated immediately once their respective setpoints are reached? If not
-why not?
(0.5) 4 6.04 The design of the suppression chamber incorporates a free air volume and a containment water volume. What is the purpose or function of each?
(2.0) 6.05 What will happen to the inlet and outlet flow paths for the shell side of the 15A feedwater heater on an i
increasing level condition in the 15A heater? Assume level continues to increase to the trip points.
(3.0) 6.06 Reactor Water level can be automatically controlled in the two modes.
(Single Element and Three Element) a.
For each mode of operation what input signals are used to control level.
(2.0) b.
What components are controlled or regulated in each mode of operation? Be specific, include how many components can be controlled in each mode.
(2.0)-
4 6.07 For each of the following, state whether a Rod Block, Half-scram, or No Reactor Protection System Action is generated for that~ condition. Note:
If two or more actions are generated, i.e., rod block and half-scram, state the most severe, i.e., half-scram.
a.
APRM B downscale, Mode Switch in Run (0.5) ngstoA C, Mode Switch in Startup (0.5) b.
12 L 7Bupscale(>108% slow)ModeSwitch c.
Flow Units A an in Run (0.5) d.
Reactor water level 55", Reactor Power 18% Mode Switch in Run (0.5) d
..,n---.n.-.-.,-r---,.-.
,.vr w-,-
--.-..-.-m,--
6.08 There are two means of bypassing the operation of the R8M automatically. What are these two means?
(2.0) 6.0iHowdoestheRSCSrespondwhenanRPISfailureoccurs?
Assume both high power (100% power) and low power (1% power) conditions.
(2.0) 6.10 In regards to the Control Rod Drive System:
a.
What prevents a control rod drive. pump from going into " Runout" following a scram?
(1.0) b.
How does the on-line FCV respond following a scram?
(0.5)
.6.11 The SBLC system has a minimum injection time of 50 minutes to prevent reactivity " Chugging." What is_ reactivity ~
" Chugging," and why is it disadvantageous?
(1.0)
END OF SECTION e
2
Procedures - Normal, Abnormal, Emergency and Radiological Control 7.01 During a training startup with one RHR loop in the Shutdown Cooling Mode, three items are required to be verified once per hour. What are these three items?
(1.5) 7.02 What restrictions apply to the use of " Notch Override" switch during a unit startup?
(2.5) 7.03 What four items must be logged once the reactor has (1.0) been brought critical?
7.04 What actions must be performed to the Feedwater System and the RWCU System to prevent or minimize feedwater nozzle, sparger, and header, thermal stress at low flow conditions? Two actions required for full credit.
(2.0) 7.05 During a normal unit shutdown you are cautioned not to open the turbine vacuum breaker at high RPMs. Why is
- this caution necessary?
(1.0) 7.06 Define the following terms:
a.
High Radiation Area (1.0) b.
Radiation Area (2.0) 7.07 What action must be taken on a stuck open safety relief valve?
(1.5) 7.08 The operator is not required to verify the exact number of control rods. not properly inserted before taking the immediate actions on an ATWS but there are some guidelines d
@t are these guidelines?
(3.0)
Wha as to when to inject SBLC.
dw.
W w W
7.09 What automatic actions occur on a high reactor water level (level 8)? Four automatic actions required for
' full credit.
(2.0) 7.10 If the generator trips during a Stator Cooling Runback:
a.
What are two' conditions which could have initiated the trip?
(2.0) b.-
What are two initiating signals for the Stator Cooling Runback?
(2.0) 7.11 According to LaSalle " General Precautions," if directed to open the ADS SRVs or-if ADS initiates automatically, DO NOT close the ADS valves unless one of two conditions exist. What are these two conditions?
(2.0) 4
- - ~
.. ~,
---,-w--
,w,7-, - ------,-
4 7.12 Other than "As directed" what are the entry conditions for Level / Pressure control LGA-01? Be specific, include all setpoints.
(1.5)-
END OF SECTION i
.e
.e F
I 4
3 4
2
Administrative Procedures, Conditions and Limitations
^
8.01 Define the following terms according to your Technical Specifications.
t.-im ch a.
Li:;f t:d ontrol Rod Pattern (1.0) b.
Channel Check (1.0) 8.02 What is required in order to implement a " Temporary System Change" on backshifts and weekends? Include in your answer any specific requirements such as qualifications?
(2.0) 8.03 In order to remove a safety-related system from service the outage must have independent verification.
a.
Can anyone verify the. system is removed from service?
Explain your answer.
(1.0) b.
Under what conditions can the second verification be waived?
(1.0) 8.04 Other than on-shift personnel what other personnel have unlimited access during normal operations? Six required for full credit.
(3.0) 8.05 a.
10 CFR 50.54 states, "An operator or Senior Operator licensed pursuant to Part 55 shall be at the controls at all times during operation of the facility." What must an operator do to meet this requirement?
(2.0) b.
If a unit operator finds it necessary to leave the "at the controls" area who must stand-in for him?
(0.5)
"8.06 What three (3) actions must be completed within one hour if one control rod is determined inoperable due to being immovable,'as a result of excessive friction or mechanical interference, or known to be untrippable?
(3.0) 8.07 For each of the following conditions, state whether you would consider the system operable or inoperable for Condition 1 plant operation.
For each one you consider inoperable, briefly state why you determined the system to be inoperable (i.e., why it cannot perform its intended function).
a.
The LPCS " Keep Filled" pressure alarm has actuated and subsequent high point venting of the LPCS discharge piping does not result in water issuing from the vents.
(1.5) b.
Two Suppression Chamber Vacuum Breakers cannot be shut.
(1.5)
-.,,.n.
8.08 What is meant by Primary Containment Integrity?
(3.0) 8.09 According to Technical Specifications, what is the basis of the following scram.
a.
Main Steam Line Isolation Valve Closure (1.0) b.
Main Steam Line Radiation High (1.0) c.
Turbine Step __ Valve Closure (1.0)
END OF TEST m
2
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Tat.le 2: Saturated Steam: Pre.sure Tetrie U
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w=g-.s, Ao=.6023 x 10 , A=ato. ic weight A tt) = V E '$e Vn:=6CI:/d2 n = v/(1+d) 2(x) = 2o e **
6 g, y f g3,7xiot o)
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I CP=(CPbase} (
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/
1 in = 2.54 ca '
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~
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b 30 seer N) * '4'l T**~
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9 N
9
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i M
W*
ANSWERS - Theory of Nuclear Power Plant Operations, i
Fluids and Thermodynamics 5.01 a.
Less (0.5) b.
More (0.5) c.
More (0.5)
Ref.:
Reactor Physics Review, pgs. 150 & 160 5.02 a.
Latent Heat of Vaporization - The thermal energy which must be added to a substance to change its state from saturated liquid to saturated vapor.
(This heat must also be removed from a substance to change its state from saturated vapor back to saturated liquid.)
(1.0) b.
Critical Point (K) - The critical point represents the highest temperature (critical temperature) and pressure (critical pressure) at which a gas and liquid can exist in equilibrium as distinguishable phases.
(1.0)
... c.
Saturated Liquid - The state of a substance wherein (given a constant pressure):
Any further addition of thermal energy in the form of i
a heat will result in vaporization (boiling) of.the substance at a constant temperature (saturation temperature) until all of the substance is evaporated.
(0,5) augJ/,.
(/.b)
OR fo /4 ~
,~s.
Any further removal of thermal energy in the form of heat will result in lowering the temperature of the substance.
(9:#)
J Note:
Either a further addition or removal will be accepted for full credit.
Ref. :
Reactor Physics Review, Thermodynamics, p. 44 i
5.03 a.
Increasing the bundle mass flow rate also causes critical power to increase.
The higher mass flow rate provides a more efficient heat transfer mechanism.
Therefore, a greater coolant enthalpy increase is required to reach saturation conditions.
Thus, for the same bundle power conditions, the margin to OTB.has increased.
(1.0) l
i b.
An event that increases bundle inlet subcooling (such as removing a feedwater heater from service) causes an increase in critical power.
For a given flow rate, a greater coolant enthalpy rise is required to reach saturation conditions.
Therefore, the bundle power required to produce critical quality /0TB is also increased.
(1.0) c.
Increasing bundle pressure, in the normal reactor operating pressure range, causes critical power to decrease.
Consulting the steam tables, at high pressure (>800 psia), the latent heat of vaporization (hf ) decreases as pressure is increased.
Therefore, 9
the bundle power required to reach OTB also decreases as pressure increases.
(1.0)
Note:
(0.25) pts for direction either increase or decrease.
(0.75) pts for explanation.
Ref.:
Core Thermal Hydraulics, pgs. 56 & 58 5.04 No. The production rate is directly proportional to
~ power level, but removal rate is proportional to xenon concentration and it contains a power dependent term, thermal neutron flux.
Since flux is directly proportional to power level, the burnout term becomes more significant.
This results in an equilibrium xenon value which is higher than the original value, but not twice as high.
This is demonstrated in Figure 70 which is shown here.
(2.0)
Ref.:
Reactor Physics Review, p. 216 5.05 As fuel temperature is increased (due to Control Rod Pull) more voiding is created. Therefore, more back pressure which would mean less flow (1.0).
- However, because of core orificing back pressure is less significant and flow through the bundle is almost the same (1.0).
(2.0)
Ref.:
Core Thermal Hydraulics, pgs. 20 & 22 5.06 a.
Moderator Temperature Effects on Rod Worth.
- First, it must be realized that regardless of its dimensions, a control rod is still localized poison.
The neutrons which interact with a control rod, travelled a measurable distance to reach the control rod.
It stands to reason that the further a neutron travels while slowing and while thermal, the better the chances of interacting with a control rod. As moderator temperature increases during reactor heatup, the density decreases. This decreased density will yield t
longer slowing down lengths, and longer thermal l
2
diffusion length.
Hence, thermal neutrons from further within a fuel bundle will visit a control blade.
Rod worth will increase as the reactor heats up to operating temperatures around 545*F.
(1.0) b.
Void Effects on Rod Worth.
Recall from the discussion on voids and neutron flux, that a small increase in void fraction made a large change in moderator density.
It would seem that this effect would follow the guidelines of a temperature change. However, as with the previous discussions, void effects are much more pronounced. The key is that the increased slowing down and thermal diffusion lengths will result in a considerable increase in the average energy of the neutron flux. Thus, more fast and epithermal neutrons will result. The fast nonleakage probability decreases, and the resonances of U-238 and Pu-240 will have a greater to strip the epithermal neutrons from the generation.
The net effect is lower or decreased thermal flux. While decreased thermal flux will cause a decrease in power, it will also result in a decreased worth of the control rods.
Remember, if you have no
~
~
thermal neutrons, you cannot turn power in a reactor by the use of control rods.
In case of voids and control rods, power is turned because of the decrease in thermal neutrons available to cause thermal fission.
(1.0)
Ref.:
Reactor Physics Review, p. 190 5.07 a.
The process computer compares the APLHGR in a node to the MAPLHGR limit for the node in a ratio called MAPRAT.
MAPRAT _ APLHGR MAPLHGR limit As long as the largest MAPRAT is less than one, we are assured that we have not exceeded this thermal limit.
Therefore, the MAPRAT as stated is not conservative.
(0.75).
b.
1.
F (0.25) 2.
F (0.25) 3.
T (0.25)
Ref.:
Core Thermal Hydraulics, pgs. 46 & 50
~
5.08 a.
Voids 1 x 10 3 Ak/k (0.5)
Temperature 1 x 10,4 Ak/k (0.5)
Doppler 1 x 10 5 Ak/k (0.5) 3
b.
Less negative (0.5)
Less negative
,(0.5)
More negative (0.5)
Ref.:
Reactor Theory, p.168 5.09 400*F sat. pressure = 247.26 psia 600 psig = 614.7 psia: sat. temperature = 488.7*F Fluid is subcooled (0.5) by 88.7*F (0,5)
Partial credit given if 14.7 is forgotten.
(1.0)
Ref.: Thermodynamics / Steam Tables 5.10 P = Poe T/T 370 = 50e 120 sec/T T = 59.95 sec.
(1.'0)
... Doubling time =
T
= 41.49 sec (1.0) g Ref.:
Reactor Physics 5.11 a.
Decreases mainly due to samarium build-in (0.5) b.
Increases mainly due to gadolinia burnout.
(May also give credit for Pu-239 build-in)
(0.5) c.
Decrease due to fuel depletion.
(0.5)
Ref.:
Reactor Physics
..5.12 CR (1 - Keff ) = CR (1 - Keff )
2 i
2 2
CR /CR2 = 10 2
CR (1 -'Keff )/CR2 (1 - Keff )
1 i
2 (1 - 0.9)/10 = (1 - Keff )
2 (0.1)/10 = 1 - Keff2
.01 = 1 - Keff2 i
Keff2 = 1 - 0.01 Keff2 = 0.99 (2.0)
Ref.:
Reactor Physics Review, p. 72
+-
i 4
l
5.13 a.
The subcooling of condensate, i.e., the condensed liquid existing ~in the hotwell usually is about 8'F below saturation temperature.
(0.5) b.
Without " Condensate Depression" the condensate pumps would cavitate due to the water at the eye of the pump being at saturation temperature.
(0.5)
Ref.:
Reactor Physics Review END OF SECTION og
,,o e
l l
l l
?_
5
ANSWERS - Plant Design, Control and Instrumentation 6.01 a.
An increase in indicated core flow (0.5)
An increase in affected recirculation loop flow (0.5)
A decrease in generator output (0.5)
A decrease in core plate AP (0.5)'
b.
Will accept either of the answers below for full credit.
(0.5) 1.
During a line break, an intact jet pump will act as a flow restrictor, and limit blowdown, while a broken jet pump will increase the blowdown area.
2.
An intact jet pump acts as a standpipe which prevents water from draining below two thirds core height.
Ref. :
Reactor Vessel and Internals, Chapter 2
~
~
0 + 60 (0.5) 6.02 narrow range-wide range
-150 to +60 (0.5) upset range O to +180 (0.5) shutdown range O to +400 (0.5) fuel zone
-111 to -311 (0.5)
Ref.:
RPU Instrumentation 6.03 a.
1.
If the temperature difference between steamline and pump suction is less than 10.1*F.
(1.0) 2.
If feedwater flow is less than 30%
(1.0) b.
These interlocks activate fifteen seconds after the condition is indicated in order to prevent spurious trips from temporary changes in temperature or feedwater flow.
(0.5)
Ref.:
Recirc., Chapter 5, p. 70 6.04 The containment water volume function is to provide a standby reservoir of water capable of quenching the steam resulting from a loss of coolant accident.
(1.0)
The suppression chamber gas volume, is of sufficient size to accommodate, under accident conditions, all the non-condensable gases within the drywell without exceeding the design pressure capabilities of the primary structure.
(1.0)
Ref.:
Primary Containment, Chapter 48, p. 20 6.05 1.
Normal drain to 14A heater full open.
-(0.5) 2.
Emergency drain valve opens.
(0.5) 3.
15A Extraction Steam Inlet Stop Valve closes.
(0.5) 4.
LP Heater 15A Drain Input from HP heater 16A Normal Drain Valve closes.
(0.5) 5.
LP Heater 15A Input from HP 16B Normal Drain Valve closes.
(0.5) 6.
LP Heater 15A/B/C Extraction Non-Return Check Valve closes.
(0.5)
Ref.:
Feedwater Heaters and Drains, Chapter 30, pgs. 58/60 level (0.5) 60.6 a.
Single Element level, feed flow Three Element steam flow (1.5) b.
Single Element one component (Turbine A-Valve-Turbine B)
(1.0) two components, two TDRFP or Three Element one TDRFP and FRV (1.0)
Ref.:
RWLC, Chapter 31, pgs. 6 & 8 6.07 a.
Rod block (0.5) b.
Half-scram (0.5) c.
Rod Block (0.5) d.
No Reactor Protection System Action (0.5)
Ref.:
APRMs, Chapter 14, p. 104 6.08 1.
If the Reference APRM downscale trip occurs (Setpoint < 30% power) the associated RBM is automaticaT1y bypassed.
(1.0) 2.
If an edge or peripheral control rod is selected, both RBM channels are bypassed automatically.
(1.0)
Ref.:
RBM, Chapter 15, p. 62 6.09 1.
< 75% rod density RPIS failure blocks all rod movement.
(1.0) 2.
> 75% rod density RPIS - N/A (1.0)
~~
Ref.:
RSCS, Chapter 19, pgs. 18 & 19 2
L
6.10 a.
After a scram the CR0 pumps will try to charge all 185 accumulators at once. To prevent " Runout" and pump damage a. series of restricting orifices in the charging line will limit flow to approximately 200 gpm.
(Setpoint not required for full credit)
(1.0)-
b.
After a scram the FCV diverts most of the water to recharge the accumulators.
The FCV does not shut fully, which minimizes water hammer on downstream piping, when the scram is reset.
(0,5)
Ref.: CRD Hydraulics, Chapter 8, pgs. 22 & 24 6.11 Reactivity " Chugging" is when a mass of water with a high sodium penetaborate concentration is periodically cycled in and out of the core.
(0.5)
This results in extreme power oscillations, which could damage fuel.
(0.5)
Ref.:
SBLC, Chapter 10, pgs. 30 & 31 END OF SECTION J
b 3
ANSWERS - Procedures - Normal, Abnormal. Emergency and Radiological Control 7.01 a.
Reactor Vessel shall be verified to be unpressurized.
(0.5) b.
Thermal power less than 1% of rated thermal power.
(0.5) c.
Reactor Coolant Temperature less than 212*F.
(0.5)
Ref.:
LGP 1-1, p. 8 7.02 a.
When 25% of the control rods have been pulled to position 48 (0.5).
Discontinue use of the " Notch Override" switch for control and withdrawals between positions 0 and 12 (0.5)
(1.0) b.
When 50% of the control rods have been pulled to position 48 (0.5).
Discontinue the use of the
~
~
" Notch override" switch between rod position 00 and 24 (0.5) while startup and reactor heatup are in progress (0.5).
(1.5)
Ref.:
LGP 1-1, p. 12 7.03 a.
Time (0.25) b.
Rod Position (0.25) c.
Coolant Temperature (0.25) d.
Reactor Period (0.25)
Ref.:
LGP 1-1, p. 13 7.04 To minimize feedwater nozzle, sparger, and header thermal stress at low flow conditions:
a.
FRV Inlet Stop will be manually throttled b.
Reactor Water Cleanup return flow will be maintained as high as possible.
c.
Feedwater on-off cycles should be eliminated Any two for full credit (2.0)
Re f. :
LGP 1-2, p. 7 7.05 Opening the vacuum breaker at high RPMs imposes excessive loads on the turbine last stage buckets.
(1.0)
Ref. :
Normal Unit Shutdown Procedure l
l
7.06 a.
High Radiation Area - Any area accessible to personnel in which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 100 millitem.
(1.0) b.
Radiation Area - Any area accessible to personne1'in which there exists radiation at such levels that a major portion of the body could receive in any one hour a dose in excess of 5 millires (1.0) or in any 5 consecutive days a dose in excess of 100 millirem (1.0).
(2.0)
Ref.:
Radiation Protection Definitions
,7.07 Attempt to close the valve as follows 1.
Cycle the SRV from auto to open and back to auto.
(0.5) 2.
Remove the valve fuses.
(0.5) 3.
Replace the fuse and observe the valve indication to see if the valve closed.
4.
Scram the reactor if the SRV did not close af ter two minutes.
(0.5)
Note:
Part 3 is implied in Part 2 and if not mentioned by the candidate no credit will be lost.
Ref.:
LOA-NB-02, p. 2 7.08 If five or more adjacent rods are not inserted to et least notch position 06 (1.0) or thirty or more rods are not inserted to at least notch position 06 (1.0)
AND Reactor Vessel Water level cannot be maintained above +12.5" (0.5) or Suppression Pool temperature reaches 110'F (0.5).
(3.0)
Ref. :
LOA-N8-09, p. 2 7.09 a.
Trip Main Turbine (0.5) i b.
Close the RCIC Turbine Steam Supply Valve (0.5) c.
Trip the Reactor Feed Pumps (0.5) d.
Close the HPCS injection valve (0.5)
Ref.:
LOA-N8-10, p. 2 7.10 a.
Stator amps less than 21,831 in two minutes (1.0) or stator amps less than 7057 in 3 1/2 minutes (1.0).
(2.0) b.
46 psig Stator Inlet Pressure (1.0) or 81*C Stator-Outlet Temperature (1.0).
(2.0) 2
Ref. :
LOA-GC-01 7.11 a.
Directed by an LGA caution to stay above 57 psig (1.0) b.
Misoperation in the auto mode is confirmed by at least two independent indications (1.0)
Ref.:
LGA-GP 7.12 Boron has not been injected into the RPV to shutdown the reactor (03 AND a.
RPV water level (0.2) below +12.5" (0.2) b.
RPV pressure (0.2) above 1043 psig (0.2) c.
Drywell pressure (0.2) above 1.69 psig (0.2)
(1.5)
Ref.:
LGA 0.1 END OF SECTION i
i 1
4 4
1 l
l 3
.._r....
,__._,_,___.____,_______,__.__.m_._______,.
ANSWERS - Administrative Procedures, Conditions and Limitations 8.01 a.
Limiting Control Rod Pattern - shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting valve for APLHGR, LHGR or MCPR.
(1.0) b.
Channel check shall be the qualitative assessment of channel behavior dudng operation by observation.
This determination shall include, where possible comparison of the channel indication and/or status with other indications and/or status derived from
~
independent instrument channels measuring the same parameter.
(1.0)
Re f. :
T.S. Definitions 8.02 On backshifts and weekends if an immediate " Temporary System Change" is required, it may be implemented provided the safety evaluation is completed (0.5) and reviewed by two SR0s (0.5) and the change is authorized by the Shif t
- Engineer (0.5). One of the SR0s must have an engineering degree such as the SCRE (0.5).
(2.0)
Ref.:
LAP 240-6, p. 3 8.03 a.
The person making the inspection for verification should be at least as qualified as the person placing the outage.
(1.0) b.
When the second verification involves high radiation conditions, the second verification can be waived.
(1.0)
Ref.:
LAP 900-4
~8.04 Any six for full credit.
1.
Station Manager 2.
Superintendents 3.
Assistant Superintendents 4.
Operating Engineers 5.
Technical Staff Supervisor 6.
Region III Resident Inspectors 7.
Station QA 8.
Station Security (3.0)
Ref.:
LAP 1100-12
8.05 a.
"At the controls" means that the unit operator will be in line-of-sight of the unit front panels, that he is in a position to monitor plant parameters, and that he is in a position to take immediate action, if required.
(2.0) b.
A second licensed operator, normally the center desk operator.
(0.5)
Ref.:
LAP 1600-2 8.06 a.
' Verify the inop rod is separated from all.other inop rods by at least two control cells in all directions.
(1.0) b.
Electrically or mechanically disarm the associated directional control valves.
(1.0) c.
Verify adequate shutdown margin.
(1.0)
Ref.:
T.S., pgs. 3/4 1-1 & 1-3 8.07 a.
Inoperable LPCS (0.5).
Potential for water hammer in discharge piping and possible discharge piping damage
~
(1.0).
(1.5) b.
Inoperable Vacuum Relief System (0.5). Would allow bypassing the suppression pool in case of an accident (1.0).
(1.5)
Ref.:
T.S., pgs. 3/4 5-4, 1-20 & 6-35 and B 3/4 5-1, 1-4 & 6-4 8.08 a.
All primary containment penetrations required to be closed are either:
1.
Capable of being closed by an operable primary containment automatic isolation system or; (0.25) 2.
Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position.
(0.25) b.
All primary containment equipment hatches are closed and sealed.
(0.5) c.
Each primary containment air lock is operable.
(0.5) d.
The primary containment leakage rates are within limits.
(0.5) e.
The suppression chamber is operable.
(0,5)
+
2
The sealing mechanism associated with each primary f.
containment penetration; e.g., welds, bellows or (0.5) 0-rings, is operable.
Ref.:
LaSalle Technical Specifications Main Steam Line Isolation Valve - Closure 8.09 a.
t The main steam line isolation valve closure trip was provided to limit the amount of fission product The MSIVs release for certain postulated events.
are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure. The MSIVs closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic (1.0)
Safety Limits.
b.
Main Steam Line Radiation - High The main steam line radiation detectors are provided When to detect a cross failure of the fuel cladding.
the high radiation is detected a trip is initiated to reduce the continued failure of fuel cladding.
At the same time the main steam line isolation valves are closed to limit the release of fission products.
The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel No credit was taken for operation of this cladding.
trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability (1.0) of the Reactor Protection System.
Turbine Stop Valve - Closure c.
The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that With would result from closure of the stop valves.
a trip setting of 5% of valve closure from full open, the resultant increase in heat flux is such that (1.0) adequate thermal margins are maintained.
Ref.:
T.S., B2-11 & 82-12 END OF SECTION 3
-