ML20057D876
| ML20057D876 | |
| Person / Time | |
|---|---|
| Site: | Washington State University |
| Issue date: | 09/24/1993 |
| From: | Caldwell J, Issac P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20057D863 | List: |
| References | |
| 50-027-OL-93-03, 50-27-OL-93-3, NUDOCS 9310060112 | |
| Download: ML20057D876 (50) | |
Text
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i ENCLOSURE I U. S. NUCLEAR REGULATORY COMMISSION OPERATOR LICENSING RE0VALIFICATION EXAMINATION REPORT REPORT NO.:
50-27/0L-93-03 FACILITY DOCKET NO.:
50-27 FACILITY LICENSE NO.:
R-76 i
i FACILITY:
Washington State University EXAMINATION DATES:
August 16-17, 1993 i
i EXAMINER:
Patri,chJ.Isapc, (Chief Examiner) f F!fl SUBMITTED BY:
/a7 ov-Chie:f ExaQner /7 Date APPROVED BY:
ANM4 v. [
Ja,eVCaldweli, thief Date N -Power Readtor Section 0 erator Licensing Branch ivision of Reactor Controls and Human factors Office of Nuclear Reactor Regulation l
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SUMMARY
Written and operating requalification examinations were administered to one Senior Reactor Operator (SRO).
The SR0 passed the examinations.
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9310060112 930924 PDR ADOCK 05000027 V
, ENCLOSURE-1 REPORT DETAILS 1.
Examiner:
Patrick J. Isaac, Chief Examiner 2.
Examination Results:
R0 SR0 TOTAL (Pass / Fail)
(Pass / Fail)
(Pass / Fail)'
NRC Grading:
N/A 1/0 1/0 Facility Grading:
N/A N/A N/A 3.
Written Examination:
The written examination was administered on August 16, 1993 to one Senior Reactor Operator. He passed the written examination.
The facilities written examination comments and the NRC's resolution to those i
comments are found in Enclosure 2.
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4.
Qperatino Tests:
Operating Tests were administered on August 17, 1993.
During the reactor start-up the examiner identified that the rod pull sheet i'
used by the candidate (Appendix "A" to S.0.P. #4) was the wrong revision (dated 1991 with the current revision being 1993).
In addition, modifications to the control element pull sequence had been made without the proper reviews and authorization as required by the facility's Administrative Procedure #2, Approval, Revision, and Review of Star,dard Operating Procedures.
A determination was made that the pull sheet was correct for the 33-X core and the start-up proceeded.
The SR0 passed the operating tests.
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t ENCLOSURE-1 5.
Exit Meetino:
Personnel attending:
Dr. Gerald Tripard, Director Jerry Neidiger, Rx Supervisor Brian Bunce Dr. Howard Miles James L. Caldwell, NRC Patrick J. Isaac, NRC t
The exit meeting was conducted on August 17, 1993. The facility examination comments were discussed as noted in Enclosure 2.
The NRC and the facility i
staff also discussed the concern raised during the operating tests as addressed in paragraph 4.
The licensee acknowledged the problem and informed the NRC that corrective actions will been taken.
Region V will review the licensee's corrective actions during the next scheduled inspection.
The NRC thanked the Washington State University staff for their cooperation and assistance during the examinations.
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ENCLOSURE 2 NRC RESOLUTIONS - WRITTEN EXAMINATION Note: The facility comments are attached to this enclosure SECTION A QUESTION A.4 The staff agrees with your comment and the question will be deleted.
Although the question and answer are technically correct in a simplified manner, they were not appropriate for your facility without some additional qualifications in the stem of the question.
Since there are no correct answers for your facility as the question was written, the question "A.4" will be deleted.
QUESTION A.5 The staff understands your comment, both "a" and "b" will be considered correct answers. The most correct answer is "a" as is stated in the Washington State University reference material as well as reference material from other facilities with TRIGA fuel.
flowever, the use of " Moderator Temperature" as distractor "b"
instead of Bath Temperature appears to have caused confusion, since the hydrogen in the Zirconium hydride is also a moderator.
Therefore both "a" and "b" will be considered correct.
QUESTION A.15 The staff agrees with your comment, since there are no correct answers, question A.15 will be deleted.
QUESTION A.19 Comment is noted, no change to the examination grading is required.
SECTION B OVEST10N B.2 Comment is noted, no change to the examination grading is required.
OVESTION B.5 The staff understands your comment and question B.5 will be deleted.
The purpose of the question was to test the candidate's understanding and knowledge of your facility's procedures.
The reference material provided for the purpose of preparing
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the examination used the terms N. R. C. Form No. 1, 2, 3 and 4 to distinguish between the different forms required for different authorizations routinely used by an operator.
In a multiple choice test the exact titles of the forms could not be used because they included the answer within them.
The NRC staff did not know that you do not require the operators to be familiar with the different form numbers and their purposes.
Since the operators are not required to be familiar with the form numbers question "B.5" will be deleted.
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i a ENCLOSURE 2 DUESTION B.8 The staff does not agree with your comment. Answer "c" is the ONLY correct answer.
In accordance with the Washington State University Technical Specifications there are only two Safety Limits. These limits are 1150 Degrees C for FLIP fuel and 1000 Degrees C for standard fuel. Answer "c" which was FLIP-Type fuel temperature = 1150
(
Degrees C is the only correct answer. The other distractors were Limiting Conditions of Operation (LCO) specifications not safety limits. The Technical i
i Specification defines Safety Limits as " limits on important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity."
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i where LCOs are those requirements which must be met by the regulations to allow the i
reactor to be operated. Although both Safety Limits and LCOs are important to the safe operation of the reactor, Safety Limits are the most important limits placed on reactor operation and exceeding those limits would result in the NRC questioning the i
continued operation of the facility.
The operators should know the difference
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between a safety limit and a LCO.
Answer "c" is the correct answer, no change to the examination grading is required.
QULESTION B.13 i
Comment is noted, no change to the examination grading is requi>ed.
This question i
has been used many times on non-power reactor examinations.
The purpose of the question is to determine if a candidate understands the function and purpose of a channel check as opposed to a channel test or a channel calibration.
Channel checks are routinely performed by the operators and therefore operators should understand j
how to conduct them and their purpose.
No change to the examination grading is required.
QUESTION B.17 U
Comment noted, no change to the examination grading is required.
l SECTION C DUESB ON C.]
The staff does not agree with your comment, no change to the examination grading is i
required.
In accordance with the information provided in the reference material, S.O.P. #19 indicates that only the Exhaust Gas Monitor has an alarm with a steady high-pitched tone.
No other alarm, which requires any of the actions given as distractors to the question, is characterized as a steady high-pitched tone. No change to the examination grading is required.
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i DULSTION C.2 The staff agrees with your comment, the individuals taking the examination were not penalized if they did not provide an answer for C.2h.
Although a blank was provided for C.2h. under the stem of the question, the blank for C.2h. was not provided on I
l the answer key.
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" ENCLOSURE 2 I
OVESTION C.3 The staff agrees in part with your comment, credit wi.'l be given for both "c" and j
"d" as correct answers. There were multiple correct aaswers but there were also two incorrect answers "a" as you suggested and "b."
Answer "b" was incorrect because the question asked for the " normal make-up path" and the distractor in "b" describes a path which is opposite of the normal path.
Credit will be given for either "c" or "d."
l OVESTION C.8 Comment noted, no change to the examination grading is required.
QUESTION C.9 The staff understands your comment, both answers "a" and "b" will be considered correct. The term fuel used in the stem of the question was not intended to mean the fuel meat but the core itself.
In this case as emphasized throughout the reference material the correct answer is "b" Natural Convection.
However since there was some confusion among your staff because it was not clear that fuel meant i
the core than the answer "a" Natural Conduction will be considered acceptable. Both answers "a" and "b" will be considered acceptable.
QUESTION C.ll The stated question on which you raised a comment was C.10.
However, based on a review of the exam questions the staff determined that you were commenting on I
question C.11.
Therefore the staff will respond to your comment based on question C.ll.
The staff understands your comment and both "b" and "c" will be considered acceptable answers.
The intention of the question was to describe a startup in which a pulse was going to be conducted.
In this case the pulse rod would not have been withdrawn and consequently answer "c" would be the correct answer.
This startup sequence was described using S.0.P. #4 Appendix A, Withdrawal Sequence for Pulsing Operation.
Although, the stem of the question was clear and never implied the startup was being conducted to calibrate the pulse instrumentation, it appears that your staff was confused by the stem and believed that the startup was being conducted for the purposes of a pulse instrument calibration.
In this case the pulse rod would have been withdrawn and answer "b" would be the correct answer.
Both answers "b" and "c" will be considered acceptable.
QUESTION C.14 Comment is noted, no change to the examination grading is required.
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. ENCLOSURE 2 OVESTION C.16 The staff does not agree with your comment.
As stated in the information provided by you concerning the 33-X core, control blade #2 is listed as having the maximum reactivity insertion rate. As you stated in your comment the distractor "c" Blade
- 3 does not exist.
Control element #3 is the transient ROD and not a blade and the stem of the question asked "which of the following control blades (emphasis added) has the maximum reactivity insertion rate" therefore "c" is an adequate distractor.
The maximum reactivity insertion rate is defined in the Washington State University Safety Analysis Report, section 6.3.2 as " equal to the product of the maximum differential worth of the most reactive element in dollars per inch times the element speed in inches per second."
In the case of C.16, the question was clearly asking for the control BLADE not rod with the maximum insertion rate which is answer "b" Blade #2. No change to the examination grading is required.
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1 Thursday, August 19,1993 Mr. James L. Caldwell Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Caldwell:
On behalf of the facility and the examinees, Brian Bunce, Howard Miles and myself, I would like to thank you and Mr. Patrick Isaac for providing the operator exams on August 16 and 17,1993.
Jerry Neidiger (the reactor supervisor) and I went over the exam shortly after its completion. We would like to make the following comments on some of the questions:
Section A:
A4. We recognize that the purpose of the question is to see if a candidate knows where and how to use the exponential power increase formula. However by making reference to the WSU reactor, the question is presuming a real world reactor rather than some hypothetical, artificial reactor that doesn't have some of the real world constraints,like a prompt negative temperature coefficient. One of the real accidents considered in our S.A.R. is a sudden insertion of reactivity at full power (1MW) by either an inadvenent transient rod pulse or the dropping of a fuel cluster onto/into the core (Safety Analysis For Conversion to FLIP FUEL, May 1979 page 54). The S.A.R estimates 1142 deg C for a such a pulse. The problem with using the standard formula for power increase is that it fails to take into account the very large prompt negative temperature coefficient that permits us to pulse our reactor. One of the equations that describes the behavior of our reactor is initial pulsing reactor period, T = 5.6 msec /(reactivity - 1)
This means that for an initial pulsing reactor period of 10 msec the reactivity insenion is less than 51.50, which is a relatively small pulse. If one recognizes this feature, the best answer would be "a. 220 MW" On the other hand if one just blindly applies the power formula, one gets the given answer "b. 2.75 x 10"8 MW" I recommend that both answers be accepted and that if a similar question is asked in the future that some way be found to put something in the initial question to discriminate between these two answers.
AS. The way this question is worded, the answer "b Moderator Temperature" distractor becomesjust as credible an answer as "a Fuel Temperature" coefficient. It is actually the hydrogen in the Zirconium hydride that provides the prcmpt negative coefficient that terminates a power excursion. It is mixed in with the fuel so that when the fuel temperature rises, the hydrogen gets hot immediately. It is the hot hydrogen that causes the hardening of the neutron flux resulting in the k-effective reduction. The water between the rods is the main contributor to the " bath coefficient". We generally don't think of the water between the rods as the " moderator" in the context of power excursions. 'Ihat is why, depending on what one imagines the examiner to be looking for, both a and b would look like correct answers.
l A15. According to the training manual, Unit 11, page 18, the peak power is proportional to:
(reactivity - 1.00)2 The reason for the necessity of subtracting the 1.00 of reactivity is that the delayed neutrons don't have time to contribute to the pulse. In the training manual, unit 7, page 6-43 there is a graph of peak power element versus the square of the prompt reactivity insertion, 2
((Sk/ )-1)2 in dollars. It is all discussed in the section 6.4.2 called Transient Experiments. The correct answer to this question should be 262.5 MW.
A19.... takes into account which ONE..
Section B:
B2. Leave out "by $1.25" because this amount would violate our procedures.
B5. I strongly petition for having this question removed from the exam for several reasons. As I expressed in our exit interview, this question "might" be appropriate in an open book exam or in an oral exam, since I would expect any licensed person to know where the file drawer is that holds these forms and how to recognize the correct form, either from reading the SOP or by looking at the titles on the forms in the file drawer.
What I would never expect anyone to necessarily know is the form number! Form numbers have no greater significance than page numbers. The purpose in having these forms numbered is not so that anyone should ever commit them to memory but they are numbered for purposes of organization and filing. IfI read in the SOP's that form number 4 is required, I can quickly reach for the fourth file folder in the file drawer.
Secondly,I want to see this question scuttled because I believe it would reflect badly on the US NRC to leave it there. I don't want to ever have to tell trainees to expect a question like that from the NRC. I would hope to be abic to tell trainees that they should expect to memorize numbers that are important to the safe, proper and orderly operation of the reactor.
Finally, as proof that this question is irrelevant, not one person at the facility knew with any confidence what the form number was. Jerry Neidiger, who composed most of the forms and regularly reviews them, didn't know what number it was. I recently worked with Jerry on these forms to make them more detailed and up to date and I didn't know what number it was. Brian Bunce, who taught the trainees recently about these forms, had to look it up, because he didn't know what number it was. He emphasized the content and the signing authority, not the number on the form. We have a lot of forms here. They all have numbers. No function that I can think ofis in any way compromised by not knowing a form number.
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I hope you can tell me that this question, on second thought, was not really appropriate for a written, closed book exam.
B8. A good argument could be made that all four answers are correct. The only thing to commend answer c is that it is one of the first safety limits mentioned in the technical specifications and that many of the other safety limits are indirectly tied to this number. It happens to be the first mentioned Fuel Temperature limit but hardly should be considered the only safety limit. All of the exam options are included in the list of Limiting Conditions of Operations and if violated are reponable offenses primarily because they are in fact safety limits. I am including an excerpt from one of the old appendices attached to the emergency plan to show how others have considered some of these answers as safety limits (see attached page,"Old Emergency Plan Appendix").
B13. This is another question that we would prefer to see in an oral or open book exam.
B17. I think more information should be provided with this question in the future. I realize that it was used in the past but it actually requires very specialized knowledge about the decay mode of Cobalt-60. One needs to know that they are simultaneous gamma rays rather than independently emitted gamma rays with say a 50-50 branching ratio. Just adding the word " simultaneously" after the word " emits" would improve this question.
Cl. Written description of tone is too elusive. Other alarms are also characterized this way. The question would be better stated if a specific alarm is given (EGM, for example).
C2. Missing answer slot.
C3. Multiple correct answers, only wrong answer is a.
C8. Change 4500 to 4350, otherwise OK.
I C9. Heat is removed from the fuel to the cladding by conduction. If you had put the word
" rod" after the word " fuel" then the answer key would be correct. Otherwise the only technically correct answer to this question is "a. Natural Conduction" C10. There is confusion in this question because one can not be cenain about which phase of the whole pulse operation is being refened to. In preparation for pulsing a regular stan up must be done in order to do a necessary calibration of the pulse power channel. If that is the case then this is no different than a regular stan up and the transient rod is
" energized" 'Re question implies that this is so because one cannot "de-energize" what has not been " energized" So, from the way the question is worded, one is either in the pre-pulse calibration of the power channelin which the rod is routinely energized and it becomes inadvenently de-energized, or one is in the final pulse mode in which one may have inadvertently energized the pulse rod. In either case, at the moment the rod is de-energized,it falls back in under gravity. Thus the only correct answer to this question is "b. It will drop via gravity into the core."
Cl4. Better wording would be to change " reason" to " basis" C16. Technically speaking, #3 is not a " blade" it is a transient control md. Therefore if we are being picky about a choice being correct only if it is referred to in the technically correct tenninology then one would have to rule c. out as an answer. If ycu wanted to include c.
anyway your should have refened to them all as " control elements" and then in answer c.
you could put " transient rod #3". This would still not remove an ambigt.ity between
answers b. and c. because the corect answer would still depend on whether you intended a positive or negative reactivity insertion. The positive reactivity insertion would give c. as the correct answer and a negative reactivity insenion by a scram would give b. as the correct answer.
We suggest calling b. and c. as both correct or throw the question out.
In summary:
I know a lot of hard work goes into creating these exams. I appreciate your efforts and I hope that these comments will be helpful.
Sincerelv Gerald E. Tripard.
Director, Nuclear Radiation Center enclosure: Old Emergency Plan Appendix L
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t EXCEEDING A SAFETY LIMIT 1.
General. In the event that a safety limit is exceeded, an Alert event shall be j
declared and the procedure specified by Section 6.7 of the facility Technical Specifications followed as enumerated below. Exceeding any of the below-listed limiting conditions of operation shall, for the purpose of this procedure, constitute exceeding a safety limit.
Steady state fuel temperature above 500 C.
a.
b.
Exceeding a steady state power level of 1.3 MW at any time.
c.
Shutdown margin falls below 0.25 5.
d.
Pulsing with an insertion over 2.50 S.
j t
Excess radioactivity over 8.00 S.
e.
f.
Pulsing from a steady state power level over I KW.
2.
Procedures.
Immediately scram the reactor by pressing the building evacuation a.
button.
b.
Evacuate the facility and mobilize the emergency organization.
Inform the Reactor Supervisor and facility management of the safety c.
limit that was exceeded.
d.
The main office area may be re-entered once it is demonstrated that no radiological hazard exists in that area.
Immediately report the incident to the chairperson of the Reactor e.
Safeguards Committee.
f.
Evaluate the effects that exceeding the specific safety limit is likely to have had on the reactor.
g.
The facility may be re-entered once it is demonstrated that no radiological hazard was created as a result of exceeding the safety limit.
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L@2hhDJ U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REQUALIFICATION REACTOR LICENSE EXAMINATION FACILITY:
WASHINGTON STATE UNIVERSITY REACTOR TYPE:
TRIGA DATE ADMINISTERED:
1993/08/I6 REGION:
5 CANDIDATE.
I INSTRUCTIONS TO CANDIDATE:
Answers are to be written on the answer sheet provided. Attach the answer sheets to the examination.
Points for each question are indicated in paren-theses for each question. A 70% overall is required to pass the examination.
Examinations will be picked up three (3) hours after the examination starts.
% OF CATEGORY % OF CANDIDATE'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 19.00 A.
REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS 19.00 B.
HORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 21.00 C.
PLANT AND RADIATION MONITORING SYSTEMS 59.00 TOTALS FINAL GRADE ALL THE WORK DONE ON THIS EXAMINATION IS MY OWN.
I HAVE NEITHER GIVEN NOR RECEIVED AID.
CANDIDATE'S SIGNATURE
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
1.. Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
- 2. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examination. This must be done after you complete the examination.
- 3. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
- 4. Use black ink or dark pencil only to facilitate legible reproductions.
- 5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.
- 6. Fill in the date on the cover sheet of the examination (if necessary).
- 7. Print your name in the upper right-hand corner of the first page of each section of your answer sheets.
- 8. Before you turn in your examination, consecutively number each answer sheet, including any additional pages inserted when writing your answers on the examination question page.
- 9. The point value for each question is indicated in parentheses aftei the question.
- 10. Partial credit will NOT be given.
- 11. If the intent of a question is unclear, ask questions of the examiner only.
- 12. When you are done and have turned in your exaninction, leave the examin-ation area as defined by the examiner.
If you are found in this area while the examination is still in progress, your license may be denied or revoked.
EQUATION SHEET i
Net Work (out)
Q - m c AT Cycle Efficiency -
p Energy (in)
Q - m oh SCR - S/(1-Keff)
Q - UA AT CR, (1-Keff), - CR (1-Keff)2 2
26.06 (A,,,0)
(1-Keff)o (6 - p)
(1-Keff),
SUR - 26.06/7 H - 1/(1-Keff) - CR /CRo 3
P-P 10* "
SDM - (1-Keff)/Keff o
P-P e"'"
Pwr - W, m o
B(1-p)
P-P, f, - 1 x 10~5 seconds B-p (f*/p) + [(s-p)/A,,,p]
r-f*/(p-8) 7-p - (Keff-1)/Keff A,,,- 0.1 seconds ~1 p - AKeff/Keff 0.693 B - 0.007 A
DR 0
- DR 0,2 DR - DR,e'"
2 33 z
6CiE(n)
DR -
DR =
, Ci a Curies, E a Hev, R = feet 1 Curie - 3.7x10'D dps I kg - 2.21 lbm I hp - 2.54x10 BTU /hr 1 Mw - 3.41x10' BTU /hr 3
1 BTU - 778 ft-lbf
- F = 9/5'C + 32 1 gal H O = 8 lbm
'C = 5/9 (*F - 32) 2
Section [ R Theory Thermo[Fac.OperatinaCharacteristics Page 4
- QUESTION (A.1)
[1.0)
.With the reactor on-a constant period, which transient requires the longest time to occur?
a.
5% power -- going from 1% to 6% power b.
10% power -- going from 10% to 20% power c.
15% power -- going from 20% to 35% power l
1 d.
20% power -- going from 40% to 60% power
- QUESTION (A.2)
[1.0]
The reactor is shutdown by 5% delta-K/K with a count rate of 100 cps on the startup channel. Rods are withdrawn until the count rate is 1000 cps. Which ONE (1) of the following is the condition of the reactor after the rods are withdrawn?
I Critical with Keff - 1.0 a.
b.
Subcritical with Keff = 0.995 c.
Subtritical with Keff - 0.950 d.
Supercritical with Keff - 1.005
- QUESTION (A.3)
[1.0)
Which ONE (1) of the following defines the minimum Shutdown Margin (SDM) and the maximum excess reactivity for the WSU reactor I
a.
SDM of 0.250% delta K/K & excess reactivity of 5.60% delta K/K b.
SDM of 0.250% delta K/K & excess reactivity of 8.00% delta K/K c.
SDM of 0.175% delta K/K & excess reactivity of 5.60% delta K/K d.
SDM of 0.175% delta K/K & excess reactivity of 8.00% delta K/K
Section A R, Theory. Thermo & Fac. Operatina Characteristics Page 5 l
'* QUESTION (A.4)
[1.0]
DELETEQ The WSU_ reactor is operating at 1 MW (100% power).1 A nuclear excursion causes a rapid increase with a period of 10 milliseconds. ' Assuming the reactor scrams on high reactor power with a scram delay time of 0.1 second, which ONE (1) of the following will be the peak power attained by the reactor?
a.
220 MW b.
2.20 X 10' MW c.
2.75 X 10' MW d.
3.36 X 10'3 MW
- QUESTION (A.5)
[1.0]
Which ONE (1) of the following coefficients will be the FIRST to start turning reactor power after a power excursion from full power?
a.
Fuel Temperature l
b.
Moderatee-Temperatwe piOeispifif0fd e
c.
Void d.
Power
- QUESTION (A.6)
[1.0]
Reactor power decreases on a stable negative period after a reactor scram, s
following an initial-prompt drop. Which ONE (1) of the following is the reason for this?
I All prompt neutrons decay during the prompt drop, and the subsequent a.
rate of power change is dependent ONLY on the half-life of the longest lived prompt gamma emitter, b.
This rate of power change is dependent on the MEAN lifetime of the shortest lived delayed neutron precursor.
This rate of power change is dependent on the MEAN lifetime of the c.
longest lived delayed neutron precursor.
d.
This rate of power change is dependent on the CONSTANT decay rate of prompt neutrons following a scram.
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Section A R Theory. Thermo & Fac. Operatina Characteristics Page 6
- QUESTION (A.7)
[1.0]
A short reactor period is of greater concern when the reactor is which ONE (1) of the following?
a.
Close to 1 MW b.
Close to 500 KW c.
Close to 10 KW d.
Close to source counts
- QUESTION (A.8)
[1.0]
A fuel loading is in progress. Using the data provided below, after how many fuel bundles loaded will criticality occur?
a.
2Jth bundle b.
22nd bundle c.
24th bundle d.
26th bundle Count Rate No. of Fuel Bundles 842 2
936 4
1123 7
1684 12 2807 16 1.0 0.9 0.8 0.7 1 0.6 M 0.5 0.4 0.3 0.2 0.1 0.0 2
4 6
8 10 12 14 16 18 20 22 24 26 28 30 32 NUMBER OF BUNDLES INSTALIID
Section A R Theorve Thermo & Fac. Operatina Characteristics Page 7
- QUESTION (A.9)
[1.0)
Which ONE (1) of the following expresses the relationship between differential rod worth (DRW) and integral rod worth (IRW)?
a.
D2W is the slope of the IRW curve at a given location.
b.
[RW is the area under the IRW curve up to a given location.
c.
ORW is the square root of IRW at a given location.
d.
DRW is not related to the IRW.
- QUESTION (A.10)
[1.0)
In a subcritical reactor Keff is increased from 0.865 to 0.995. Which ONE (1) of the following is the amount of reactivity added to the core?
a.
0.13 delta A/K b.
0.15 delta K/K c.
0.13% delta K/K d.
0.15% delta K/K
- QUESTION (A.ll)
[1.0)
Assume that shutdown occurs from equilibrium conditions. Which ONE (1) of the following statements concerning reactivity values of equilibrium (at power) xenon and peak (after shutdown) xenon is correct?
a.
Equilibrium xenon is INDEPENDENT of power level; peak xenon is INDEPENDENT of power level.
b.
Equilibrium xenon is INDEPENDENT of power level; peak xenon is DEPENDENT on power level.
c.
Equilibrium xenon is DEPENDENT on power level; peak xenon is INDEPENDENT of power level.
d.
Equilibrium xenon is DEPENDENT on power level; peak xenon is DEPENDENT on power level, j
l
Section A R T eory Th'ermo &'Fa. Operatini Characteristics Page 8 j
- QUESTION-(A.12)
[1.0]
l With. the reactor initially at a keff of 0.95, a certain reactivity change
.j causes the cour,t rate to double.
If this same amount of reactivity is again added to the core, which ONE (1) of the following will be the status of the I
reactor?
a.
Subtritical i
b.-
Reactor will trip
{
c.
Super Critical d.
Prompt Critical l
t
- QUESTION (A.13)
[1.0]
Which ONE (1) of the following is the energy region in which a neutron is most likely to be lost due to non-fission absorption.
c a.
Fast > 10,000 ev i
b.
Epithermal - 1 ev to 10,000 e.
c.
Thermal < 1 ev d.
Approximately the same likelihood in each energy region i
- QUESTION (A.14)
[1.0)
The WSU TRIGA reactor is required to pulse from low power levels (less than one KW). Which ONE (1) of the following is the reason for this limitation on l
power level prior to the pulse?
a.
To prevent exceeding the maximum power level limit b.
To prevent exceeding the fuel element temperature limit c.
To prevent exceeding the pool temperature limit d.
To prevent exceeding the reactivity insertion limits l
l i
I e
+
t i
/
[
Section A R Theory. Thermo & Fac. Operatina Characteristics Page 9
- QUESTION (A.15)
[1.0]
DELETED Based on the relationship between peak power during a pulse and the reactivity added, the peak power can be conservatively estimated for various reactivity additions. Given the following conditions estimate the peak power for the new pulse operation. Which ONE (1) of the following peak power levels corresponds to the estimate?
Peak Power (previous pulse)
= 1050 MW Reactivity added (previous pulse)
- $2.00 Reactivity added (new pulse)
- $1.50 a.
590 MW b.
653 MW c.
787 MW d.
922 MW
- QUESTION (A.16)
[1.0)
Given a stable reactor period of 30 seconds, calculate how long it will take for the reactor to go from 115 watts to 500 KW. Which ONE (1) of the following is the calculated time?
l a.
2.5 minutes l
b.
4.2 minutes
~
c.
42 minutes t
d.
251 minutes
- QUESTION (A.17)
[1.0]
Beta (B) and Beta-effective (B,,,) both describe the total fraction of delayed neutrons. The difference between the two is which ONE (1) of the following?
is smaller than B since delayed neutrons are born at a lower energy l
B,,,l than prompt neutrons a.
leve is larger than B since delayed neutrons are born at a lower energy B,,,l than prompt neutrons b.
leve B,,, is smaller than B since delayed neutrons are born at a higher c.
energy level than prompt neutrons is larger than B since delayed neutrons are born at a higher energy f
d.
B,,,l than prompt neutrons leve i
t i
i i
Section A F1 Theory. Thermo & Fac. Operatino Chara11 eristics Page 10
- QUESTION (A.18)
[1.0]
Which ONE (1) of the following describes a critical reactor?
a.
K - 0 & Delta K/K - I b.
K - 0 & Delta K/K - 0 c.
K - 1 & Delta K/K - 1 d.
K - 1 & Delta K/K - 0
- QUESTION (A.19)
[1.0]
K 'e' following? differs from K,n,in,,, in that K,,, takes into JEE666Q which ONE (1) of tE a.
Leakage from the core b.
Neutrons from fast fission c.
T.he effect of poisons d.
Delayed neutrons
- QUESTION (A.20)
[1.0]
Which ONE (1) of the following parameter changes will require a control rod INSERTION to maintain reactor power constant following the change?
a.
Samarium buildup b.
Xenon buildup c.
fuel temperature decrease d.
U " concentration decrease (fuel burn-up) 2 1
Section A R Theory. Thermo & Fac. Operatino Characteristics Page 11
[
W c (1 o he following $ he K,, for a reactor shutdown by $0.557 a.
0.00385 b.
0.550 c.
0.722 I
d.
0.996 t
(*** End of Section A ***)
Section B Normal /Emera. Procedures & Rad Con Page 12 i
- QUESTION (B.1)
[1.0]
Which ONE (1) of the following conditions would require an operator to scram the reactor by pressing the building evacuation button as an immediate action?
a.
Excess core reactivity of 6% Delta K/K b.
Peak fuel temperature of 800 Degrees C during a pulse c.
Conducting a pulse operation by inserting 1.75% Delta K/K d.
Steady state fuel temperature of 450 Degrees C
- QUESTION (B.2)
[1.0)
The reactor is subcritical by $1.25 and maintenance is being conducted on the transient rod drive. Which ONE (1) of the following describes the condition of the reactor in accordance with Technical Specifications?
a.
Reactor Secured b.
Reactor Operation c.
Reactor Refueling d.
Reactor Outage
- QUESTION (B.3)
[1.0)
The reactor is operating at 100KW and someone on the operating staff is preparing to conduct an experiment using the reactor. This individual must obtain permission from which ONE (1) of the following to commence the experiment?
a.
Facility Director b.
Reactor Supervisor c.
Senior Reactor Operator d.
Reactor Operator
Section B Normal /Emera. Procedures & Rad Con Page 13
- QUESTION (B.4)
[1.0)
Which ONE (1) of the following is a requirement for Shutdown Margin?
l a.
Reactivity < 5.6% Delta K/K b.
Reactivity < 0.175% Delta K/K c.
Reactivity < $2.50 l
d.
Reactivity < $5.00
- QUESTION (B.5)
[1.0]-
DELErrED Authorization for the performance of an operational experiment will be
^
documented on which ONE (1) of the following forms?
a.
N.R.C. Form No. 1 1
b.
N.R.C. Form No. 2 c.
N.R.C. Form No. 3 d.
N.R.C. form No. 4
- QUESTION (B.6)
[1.0)
Which ONE (1) of the following positions would be the lowest level which could approve a temporary change to an Operational Experiment which does not change the original intent of the experiment?
a.
Reactor Operator i
b.
Senior Reactor Operator c.
Reactor Supervisor d.
Reactor Safeguards Committee
(
i l
- QUESTION (B.7)
[1.0) l A Senior Reactor Operator is NOT required to be present at the facility during t
which ONE (1) of the following?
i a.
Reactor Startup b.
Control Element Calibration i
i' Significant Power Level Reduction c.
I d.
Changing fuel Positions in the Storage Racks l
igetion B Normal /Emero. Procedures & Rad Con Page 14
' QUESTION (B.8)
[1.0)
Which One (1) of the following is a Safety Limit?
a.
Reactor Power - 1.3 MW b.
Pulse Mode Fuel Temperature - 830 Degrees C c.
FLIP-Type Fuel Temperature - 1150 Degrees C d.
Shutdown Margin - $0.25 I
- QUESTION (B.9)
[1.0]
l Which ONE (1) of the following is an immediate action required in response to an Exhaust Gas Monitor alarm?
a.
Scram the reactor b.
Evacuate the building c.
Place the ventilation system in the dilute mode d.
Place the ventilation system in the isolation mode j
l
- QUESTION (8.10)
[1.0]
How often must a reactor operator review each of the Standard Operating Proct 4es?
a.
Monthly b.
Quarterly c.
Bi-annually d.
Annually
- QUESTION (B.11)
[1.0)
Which ONE (1) of the following is NOT required for fuel loading changes?
a.
Approval from the Director of the Radiation Safety Office b.
A completed Reactor Start-up Checkoff c.
Licensed Operator at the Reactor Control Console d.
One control blade or transient rod fully withdrawn and all other control elements fully inserted during fuel movement
Section B Normal /Emera. Procedures & Rad Con Page 15 cQUESTION (B.12)
[1.0]
You are operating the reactor at approximately 100% power when you feel a major movement of the building structure, which you believe to be caused by an earthquake. Which ONE(1) of the following is the immediate action you would take?
a.
Press the Manual scram button b.
Place the Mode Switch in RUNDOWN c.
Press the evacuation alarm button d.
Notify the Senior Reactor Operator
- dVESTION (B.13)
[1.0]
Which ONE (1) of the following is the regulatory definition of CHANNEL CHECK.
a.
The comparison of a known value of a parameter to the measured value from the measuring channel.
b.
The introduction of a signal into the measuring channel to check channel behavior.
c.
The check of the combination of sensor, interconnecting cables or lines, amplifiers and output devices that are connected for the purpose of measuring the value of a variable.
d.
The qualitative verification of acceptable performance by observation of channel behavior including comparison with independent channels measuring the same variable.
- QUESTION (B.14)
[1.0]
Which ONE (1) of the following channels is required for pulse mode operations?
a.
Transient Rod Control b.
Wide Range c.
Linear Power level d.
Log Power Level i
Section B Normal /Emera. Procedures & Rad Con Page 16
- QUESTION (8.15)
[1.0]
During the course of perforcing the daily Health Physics surveys an excessive contamination level would be defined as which ONE (1) of the following?
i 5 X 10 Micro Ci/cm from the sample preparation laboratories a.
b.
5 X 10'5 Micro C1/cm from the radiochemistry floors 2
5 X 10 Micro Ci/cm from the sample transfer equipment 2
c.
d.
5 X 10~5 Micro Ci/cm from the radiochemistry hood 2
- QUESTION (B.16)
[1.0]
What level of facility staff is RE0VIRED to decide if a specific procedure is required to be written for maintenance on the reactor control and safety systems? Assume that a procedure for this activity does not already exist.
a.
Reactor Safety Committee b.
Reactor Supervisor c.
Senior Reactor Operator d.
Reactor Operator
- QUESTION (B.17)
[1.0]
A two year old Co-60 source that was originally 20 curies is received by the i
facility. The source simultsneousif emits two gammas at 1.17 Mev and 1.33 Mev with a half life of 5.2 years ^.^ What will the dose rate from this source be today at a distance of 15 feet, a.
28 mrem /hr b.
102 mrem /hr c.
287 mrem /hr d.
1021 mrem /hr
- QUESTION (B.18)
[1.0]
In the event of an emergency at the facility which ONE (1) of the following individuals will be designated as the Emergency Director. Assume that all of the individuals listed below are at the facility at the time of the emergency.
a.
Reactor Operator b.
Senior SR0 c.
Facility Associate Director d.
Facility Director
Section B Normal /Emera. Procedures & Rad Con Page 17
- QUESTION (B.19)
[1.0]
Following the completion of the shutdown margin measurement the individual performing the second man review of the measurement must be (at a minimum) which ONE (1) of the following?
a.
Reactor Operations Staff Member b.
Reactor Operator I
c.
Senior Reactor Operator d.
Reactor Supervisor
- QUESTION (B.20)
[1.0)
The facility's requalification program DOES NOT requires which ONE (1) of the follcwing to be completed each quarter to maintain the qualification as a i
Reactor Operator?
l I
a.
One Reactor Shutdown b.
One Reactor Startup c
One Reactor Checkout d.
One Reactor Pulse I
l i
(*** End of Section B ***)
i I
Section C Plant and Rad Monitorina Systems Page 18
- QUESTION (C.1)
[1.0]
The reactor is operating at 100% power and you as the reactor operator receive an alarm which is a steady high-pitched tone, which ONE (1) of the following actions are you required to take.
a.
Evacuate the building b.
Manually scram the reactor c.
Isolate the ventilation system d.
Immediately reduce power level
- QUESTION (C.2)
[2.0]
Refer te Figure 1 and match the core components in column 2 to the core location in column 1.
There is only ONE (1) component for each location. No component is used more that once, and NOT all components are used.
a.
1.
Neutron Source b.
2.
Transient rod c.
3.
Instrument fuel rod d.
4.
Reflector e.
5.
Log-n Fission Chamber f.
6.
No. 1 CIC g.
7.
No. 2 CIC h.
8.
Pulse Gamma Chamber 9.
Standard fuel bundle 10.
FLIP fuel bundle I
- QUESTION (C.3)
[1.0]
When make-up water is required in the reactor pool which ONE (1) of the following describes the normal make-up path, a.
Through a manual valve operated by the reactor operator b.
Through the mixed-bed ion ex;. hanger and recirculation pump c.
Through the Culligan deionizer and recirculation pump d.
Through a solenoid valve operated by a float switch in the pool l
i o
Section C Plant and Rad Monitorina Systems Page 19
- QUESTION (C.4)
[1.0)
When the sampling tank becomes full which ONE (1) of the following is the normal line-up required to prepare the tank for taking a sample.
a.
3-way valve in RECIRC. position & MODE SWITCH in RECIRC. position b.
3-way valve in RECIRC. position & MODE SWITCH in DUMP position c.
3-way valve in DUMP position & MODE SWITCH in DUMP position t
d.
3-way valve in DUMP position & MODE SWITCH in RECIRC. position
- QUESTION (C.5)
[1.0]
An in-core experiment failure has occurred at the facility which resulted in -
significant increase in the CAM readings. Which One (1) of the following ventilation fan and damper line-ups will be used to clear the pool room of airborne radionuclides.
^
a.
F-1 & F-3 0FF, F-4 ON - D-1 & D-3 CLOSED, D-2 & D-4 OPEN i
b.
F-1 & F-4 0FF, F-3 CN - D-1 & D-4 CLOSED, D-2 & D-3 OPEN c.
F-4 & F-3 0FF, F-1 ON - D-2 & 0-4 CLOSED, D-1 & D-3 OPEN d.
F-1 & F-4 0FF, F-3 ON - D-2 & D-4 CLOSED, D-1 & D-3 OPEN l
- QUESTION (C.6)
[1.0]
i With the reactor secured and the console control power turned off which ONE (I) of the following alarms will result in an automatic activation of the Building Evacuation alarm?
a.
High ARM alarm i
b.
Fire alarm i
c.
Low pool water level i
d.
High neutron flux
- QUESTION (C.7)
[1.0]
The present reactor core configuration is core # 33-X.
How many of the fuel clusters installed in the core contain fuel rods with a uranium content of 8.5 wt% and an erbium content of 1.5 wt%7 a.
13 b.
17
-I c.
26 d.
30 3
i Section C Plant and Rad Monitorina Systems Page 20
- QUESTION (C.8)
[1.0]
Which ONE (1) of the following Modes of the operation of the reactor 1
ventilation system provides approximately 4500 9 50 CFM of treated external
~
air into the pool area?
i a.
Normal mode I
i b.
Emergency mode c.
Isolation mode d.
Dilution Mode
- QUESTION (C.9)
[1.0)
During reactor operation,b by which ONE (1) of the following?the heat produced fro removed from the fw.4 E6r l
a.
Natural Conduction a
b.
Natural Convection l
c.
Heat exchanger l
d.
- QUESTION (C.10)
[1.0]
The primary neutron absorption portion of the Servo-blade is composed of which ONE (1) of the following materials?
I a.
Boral j
b.
Borated graphite
'[
c.
Aluminum i
i d.
Stainless Steel i
- QUESTION (C.ll)
[1.0]
While withdrawing Blade #2, during a reactor start-up 4n preparat4en-fe E3 Eondbit a pulse operation, the solenoid valve located between the accumulifor r
thk"Ed the transient rod cylinder is inadvertently de-energized. Which ONE (1) of the following situations best describes the effect on the transient j
rod?
a.
It will move to the maximum upper limit l
t b.
It will drop via gravity into the core l
c.
It will remain in its present position (as is) f d.
It will move to the maximum upper limit then drop into the core i
l
Egetion C Plant and Rad Monitorina Systems Page 21
- QUESTION (C.12)
[1.0)
The WSU reactor incorporates both a moderator and a reflector in its design.
Which ONE (1) of the following describes the primary difference between a moderator and a reflector?
a.
Moderators decrease core leakage while reflectors provide a neutron shield b.
Moderators decrease core leakage while reflectors slow down neutrons c.
Moderators slow down neutrons while reflectors provide a neutron shield d.
Moderators slow down neutrons while reflectors decrease core leakage
- QUESTION (C.13)
[1.0]
Which ONE (1) of the following instruments provides an input signal to Safety Channel #1 a.
CIC #1 b.
CIC #2 c.
Pulse Gamma Chamber d.
Fission Chamber
- QUESTION (C.14)
[1.0]
What is the DESIGN reason basis for the WSU reactor being limited to a steady state power level of ONE Hegawatt?
a.
Design limit on convection cooling b.
Design limit on secondary cooling system c.
Design limit on Standard TRIGA fuel d.
Design limit on FLIP TRIGA fuel
- QUESTION (C.15)
[1.0]
The reactivity worth of Standard TRIGA fuel and FLIP fuel are best described by which ONE (1) of the following?
a.
FLIP fuel reactivity is much greater that Standard fuel b.
Standard fuel reactivity is much greater than FLIP fuel c.
The reactivities for both fuels are about equal d.
The reactivity of each type fuel is based only on U-235 %
i l
Section C Plant and Rad Monitorina Systems Page 21 cQUESTION (C.12)
[1.0]
The WSU reactor incorporates both a moderator and a reflector in its design.
Which ONE (1) of the following describes the primary difference between a moderator and a reflector?
a.
Moderators decrease core leakage while reflectors provide a neutron shield b.
Moderators decrease core leakage while reflectors slow down neutrons c.
Moderators slow down neutrons while reflectors provide a neutron shield d.
Moderators slow down neutrons while reflectors decrease core leakage
- QUESTION (C.13)
[1.0]
Which ONE (1) of the following instruments provides an input signal to Safety Channel #1 a.
CIC #1 b.
CIC #2 c.
Pulse Gamma Chamber d.
Fission Chamber
- QUESTION (C.14)
[1.0)
What is the DESIGN reason basis for the WSU reactor being limited to a steady state power level of ONE Megawatt?
a.
Design limit on convection cooling b.
Design limit on secondary cooling system l
c.
Design limit on Standard TRIGA fuel I
d.
Design limit on FLIP TRIGA fuel i
- QUESTION (C.15)
[1.0) i The reactivity worth of Standard TRIGA fuel and FLIP fuel are best described l
by which ONE (1) of the following?
a.
FLIP fuel reactivity is much greater that Standard fuel b.
Standard fuel reactivity is much greater than FLIP fuel c.
The reactivities for both fuels are about equal d.
The reactivity of each type fuel is based only on U-235 %
Section C Plant and Rad Monitorina Systems Page 22
- QUESTION (C.16)
[1.0)
Which ONE (1) of the following control blades has the maximum reactivity insertion rate?
a.
Blade #1 b.
Blade #2 c.
Blade #3 d.
Blade #4 c
- QUESTION (C.17)
[1.0)
Which ONE (1) of the following effects is the most significant factor in.
causing the large prompt negative fuel-temperature coefficient in the TRIGA fuel?
a.
Cell and inhomogeneities b.
Doppler c.
Core leakage d.
Resonance absorption
- QUESTION (C.18)
[1.0)
Standard TRIGA fuel has a shorter useful core lifetime than FLIP TRIGA fuel.
Which ONE (1) of the following characteristics of Standard fuel is a reason for this SHORTER lifetime?
a.
Uranium content of 8.5 wt%
b.
Required position in the core c.
U-235 enrichment d.
Erbium content of 1.5 wt%
- QUESTION (C.19)
[1.0]
Compensating voltage for the Compensated Ion Chambers (CIC) is lost just after the reactor has become critical and is increasing in power. How will the power indications associated with the CICs change compared to actual power?
a.
Remain unchanged l
b.
Indicate higher than actual power c.
Indicate lower than actual power d.
May be higher or lower depending on actual power level
Section C Plant and Rad Monitorino Systems Page 23
- QUESTION (C.20)
[1.0)
The purpose of the Nitrogen-16 diffuser system is best described by which ONE (1) of the following?
a.
Decrease the activation rate of Oxygen-16 to Nitrogen-16 b.
Increase the transport time of Nitrogen-16 to the pool surface c.
Break up the oxygen-16 bubbles to reduce Nitrogen-16 production d.
Increase heat transfer from the core to the pool at high powers s
i
(*** End of Section C ***)
l
Section C Plant and Rad Monitorina Systems Page 23
- QUESTION The purpose o(C.20)
[1.0]
(1) of the following?f the Nitrogen-16 diffuser system is best described by which ONE Decrease the activation rate of Oxygen-16 to Nitrogen-16 a.
b.
Increase the transport time of Nitrogen-16 to the pool surface Break up the oxygen-16 bubbles to reduce Nitrogen-16 production c.
d.
Increase heat transfer from the core to the pool at high powers
(*** End of Section C ***)
Section A F1 Theory. Thermo & Fac. Operatino Characteristics Page 24
- ANSWER (A.1) a
- REFERENCE (A.1)
WSU Reactor Operator Training Manual, Unit 5, page 168
- ANSWER (A.2) b
- REFERENCE (A.2)
WSU Reactor Operator Training Manual, Unit 5 ??
.05 - Keff-1/Keff CR (1-Keff,) - CR (1-Keff)
Keff - 0.95 100 (1-0,.95) - 1000 (1z_ ff) 2 Ke Keff (5/1000) - O_915
- ANSWER (A.3)
C
- REFERENCE (A.3)
Technical Specifications 3.2 & 3.4 pages 8 & 9
- ANSWER (A.4)
DELETED c
- REFERENCE (A.4)
WSU Reactor Operator Training Manual, Unit 5, page 168 P = P e ' - 1.25 MW (2.2 X 10')
o p - 2.75 X 10'
- ANSWER (A.5) a&b
- REFERENCE (A.5)
WSU Reactor Operator Training Reactor, Unit 5
- ANSWER (A.6) t c
- REFERENCE (A.6)
WSU Reactor Operator Training Manual, Unit 5, page 178
- ANSWER (A.7) d
- REFERENCE (A.7)
??
1
- ANSWER (A.8) b
- REFERENCE (A.8)
WSU Reactor Operator Training Manual, Unit 5, Pages 130-138
- ANSWER (A.9) a
- REFERENCE (A.9)
??
e
Section A R Theory. Thermo & Fac. Operatina Characteristics Page 25
- ANSWER (A.10) b
- REFERENCE (A.10)
WSU Reactor Operator Training Manual, Unit 6 page 6-11 Delta Rho - (Keff, - 1)/Keff (Keff 1)/Keff z
i Delta Rho - (0.995-1)/0.995 - (0.865-1)i 865
/0.
Delta Rho - 0.151 delta K/K
- ANSWER (A.II) d
- REFERENCE (A.11)
WSU Reactor Operator Training Manual, Unit 5, Pages 149-157
- ANSWER (A.12)
C
- REFERENCE (A.12)
WSU Reactor Operator Training Manual, Unit 6, page 168 CR /CR, (1-K,,,3)/(1-K,,,2) 2 1-K,,,, - (1-0.95) /2 Keff2 (1-0.95)/2 - 0.975 Delta Rho - (K,# f 5-0.b'$})/(0.Nb))((K'.#hh))
-K
/(K Delta Rho - (0.9 0
Delta Rho - 0.0270 0.02/0 - (K
- 0.975)/0.975 K (0.0270)(0.67$)K
=K 0.67$
K,,,,2-0.975/(16,.#$270),,2-K
,,2 - 1.002
- ANSWER (A.13) b
- REFERENCE (A.13)
WSU Reactor Operator Training Manual, Unit 5, pages 52 & 53
- ANSWER (A.14) b
- REFERENCE (A.14)
Technical Specification 3.6.3 Basis
Section A R Theory. Thermo & Fac. Doeratino Characteristics
.Page 26 l
- ANSWER
'(A.15)
DELETED I
a
- REFERENCE (A.15) l WSU Reactor Operator Training Manual, Unit 7 page 38 P / Rho' - P / Rho 2
lb50/(2)2,a P P - 1050(1.5}f(1.5}a l
/(2)
- 590.63 MW 2
- ANSWER (A.16) b
- REFERENCE (A.16)
WSU Reactor Operator Training Manual, Unit 6 m
p.-p e i
t - in(P/P))(T) - In(500/0.115)(30) t - (8.377 (30) = 251 see or 4.2 minutes 3
- ANSWER (A.17) b
- REFERENCE (A.17) i WSU Reactor Operator Training Manual, Unit 6, Page 6-12 l
3 l
- ANSWER (A.18) d
- REFERENCE (A.18)
WSU Reactor Operator Training Manual, Unit 6, page 6-11
- ANSWER (A.19) a
- REFERENCE (A.19)
WSU Reactor Operator Training Manual, Unit 5, page 91 and Unit 6, page 6-7 i
?
- ANSWER (A.20)
\\
C i
- REFERENCE (A.20)
WSU Reactor Operator Training Manual, Unit 5, pages 149-157 and Unit 6, page 1-2 l
I
- ANSWER (A.21) d l
- REFERENCE (A.21)
WSU Reactor Operator Training Manual, Unit 6, page 6-12 i
i i
a l
t
..m..
Section B Normal /Emera. Procedures & Rad Con Page 27
- ANSWER (B.1) a
- REFERENCE (B.1)
Emergency procedures - Section on " Exceeding a Safety Limit" 6% - Beff $ =.007 ($)
.06/.007 - $8.57 > $B.00
- ANSWER (B.2) b
- REFERENCE (B.2)
Technical Specifications 1.1 - Definition of Reactor Operation
- ANSWER (B.3) d
- REFERENCE (8.3)
S0P #1, paragraph F, page #8
- ANSWER (B.4) b
- REFERENCE (B.4)
Technical Specification 3.2 0.175% = Beff ($)
$ = 0.175/0.007 - $0.25
- ANSWER (B.5)
D_ELETED d
- REFERENCE (B.5) 50P #3, Paragraph C " Authorization to perform Experiments, page 2
- ANSWER (B.6) b
- REFERENCE (B.6)
S0P #3, paragraph D.1, page 4 and Technical Specification 6.B
- ANSWER (B.7) c.
- REFERENCE (B.7)
S0P #4, Section A.3, Page 1
- ANSWER (B.8) c
- REFERENCE (B.B) n Technical Specification 2.l(2)
Section B Normal /Emero. Procedures & Rad Con Page 28
- ANSWER (B.9) d.
- REFERENCE (B.9) 50P # 19 paragraph C.e.2)a)
- ANSWER (B.10) d
- REFERENCE (R.10)
Administrative Procedure Section 2 " Approval Revision and Review of Standard Operating Procedure:" paragraph F, page 3
- ANSWER (8.11) a
- REFERENCE (B.ll)
S0P #7, Paragraph B; pages 1 & 2
- ANSWER (B.12) c
- REFERENCE (B.12)
Emergency Plan, Earthquake Emergency Procedure
- ANSWER (B.13) d
- REFERENCE (B.13)
Technical Specification 1.4, Definition of Channel Check
- ANSWER (B.14) b
- REFERENCE (B.14)
Technical Specification Table 3.2, Minimum Reactor Safety Channels 1
- ANSWER (B.15) a
- REFERENCE (B.15) i S0P #10 paragraph C " Routine Surveys" (C.1.C.2)
- ANSWER (B.16) c
- REFERENCE (B.16)
Administrative procedure #6 " Performance of Maintenance Activities" paragraph C, page I l
Section B Normal /Emero. Procedures & Rad con Page 29
- ANSWER (8.17) d
- REFERENCE (B.17)
WSU Reactor Operator Training Manual, Unit 4, Chapter III, page 21 C - C,e' *"3'#2 C - 20e.asnzus.2 n
C - 15.32 Ci E - 1.17 Mev + 1.33 Mev - 2.5 Mev 2
D - 6CE/d 0 - 6(15.32)(2.5/(15)2 - 1.021 Rem /hr or 1021 mrem /hr
- ANSWER (B.18) b
- REFEKENCE (B.18)
Emergency Plan (Emergency Organization Section)
- ANSWER (B.19)
C
- REFERENCE (B.19)
Administrative Procedure #5 " Surveillance Documentation Review" paragraph B, page 1
- ANSWER (B.20) a
- REFERENCE (B.20)
Reactor Operation Training Manual, Volume I, Introduction Section, Reactor Staff Requalification Program, paragraph 2 " Quarterly Operation Requirements
Section C Plant and Rad Monitorina Systems Page 30
- ANSWER (C.1) c
- REFERENCE (C.1)
S0P #19, paragraph C.2.d, page 5
- ANSWER (C.2)
- a. - 2.
- e. - 7.
- b. - 8.
- f. - 6.
- c. - 3.
- g. - 1.
- d. - 5.
- h. - 10.
- REFERENCE (C.2)
WSU Core Layout, Core No. 33-X, January 13, 1993
- ANSWER (C.3) c&d
- REFERENCt (C.3)
FSAR Section 4.10, page 4-30
- ANSWER (C.4) b
- REFrRENCE (C.4)
S0P #11 Section B pages 1 & 2
- ANSWER (C.5) b
- REFERENCE (C.5)
FSAR Figure 3.2-1 " Reactor Ventilation System" page 3-8 and Emergency Plan Section 8.2 pages 32, 33 and 34
- ANSWER (C.6) a
- REFERENCE (C.6) 50P #19 paragraphs B.5, page 3
- ANSWER (C.7) a
- REFERENCE (C.7)
Core layout drawing, core no. 33-X, date January 13, 1993
- ANSWER (C.8) a
- REFERENCE (C.8)
Emergency plan, Section 8.2 & Pool Room Ventilation System drawing, pages 32 &
33
- ANSWER (C.9) a&b
- REFERENCE (C.9)
FSAR Section 4.1, page 4-1
Section C Plant and Rad Monitorino Systems Page 31
- ANSWER (C.10) d
- REFERENCE (C.10)
FSAR Section 4.S, page 4-14
- ANSWER (C.11) b&c
- REFERENCE (C.11)
FSAR Section 4.7, page 4-19 & 21
- ANSWER (C.12) d
- REFERENCE (C.12)
WSU Reactor Training Manual Unit 14 " Glossary"
- ANSWER (C.13) d
- REFERENCE (C.13)
FSAR Figure 4.8-2, page 4-26
- ANSWER (C.14) a
- REFERENCE (c.14)
FSAR Section 6.3.4, page 6-6
- ANSWER (C.15) c
- REFERENCE (C.15)
FSAR Section 1.2, page 1-1
- ANSWER (C.16) b
- REFERENCE (C.16)
Information on the 33-X core provided by the facility
- ANSWER (C.17) a
- REFERENCE (C.17)
WSU Reactor Operating Training Manual Volume 11 Unit 7 paragraph 6.3.3, page 6-27
- ANSWER (C.18) c
- REFERENCE (C.18)
WSU Reactor Operator Training Manual Unit 11 " Introduction to the WSU Nuclear Reactor " page 5
Section C Plant and Rad Monitorina Systems Page 32
- ANSWER (C.19) b
- REFERENCE (C.19)
Operating characteristics of CIC's i
- ANSWER (C.20) b
- REFERENCE (C.20)
FSAR Section 4, paragraph 4-12, page 4-35 V
6
Section A R Theory. Thermo & Fac. ODeratina Characteristics Page 1 1.
a b c d 13.
a b c d 2.
a b c d 14.
a b c d 3.
a b c d 15.
a b c d 4.
a b c d 16.
a b c d 5.
a b c d 17.
a b c d 6.
a b c d 18.
a b c d _
7.
a b c d 19.
a b c d 8.
a b c d 20.
a b c d 9.
a b c d 21.
a b c d
- 10. a b c d 11.
a b c d
- 12. a b c d
Section B Normal /Emero Procedures & Rad con Page 2
~
1 5
1.
a b c d 11.
a b c d 2.
a b c d 12.
a b c d 3.
a b c d
- 13. a b c d 4.
a b c d
- 14. a.
b c d 5.
a b c d
- 15. a b c d 6.
a b c d 16.
a b c d 7.
a bc d 17.
a b c d 8.
a b c d 18.
a b c d 9.
a b c d 19.
a b c d 10.
a b c d 20.
a b c d f
s
Section C Plant and Rad Monitorina Systems Page 3 4
1.
a b c d 8.
a b c d
- 2. a.
9.
a b c d b.
10.
a b c d c.
11.
a b c d d.
12.
a b c d e.
13.
a b c d f.
14.
a b c d
- g. ___
15.
a b c d h.
1 3.
a b c d 16.
a b c d 4.
a b c d 17.
a b c d i
5.
a b c d 18.
a b c d i
6.
a b c d 19.
a b c d i
i
- 7. a b c d ___
- 20. a b c d i
Section A R Theory. Thermo & Fac. Operatina Characteristics Page 1 ANSWER KEY A. I a A. 2 b A. 3 c A. 4 c DELETED A. 5 a&b A. 6 c A. 7 d A. 8 b A. 9 a A.10 b A.11 d A.12 c A.13 b A.14 b A.15 a DELETED A.16 b A.17 b A.18 d A.19 a A.20 c A.21 d
Section B Normal /Emera Procedures & Rad con Page 2 ANSWER KEY B. I a
B. 2 b
B. 3 d
B. 4 b
B. 5 d DELETED B. 6 b
B. 7 c
B. 8 c
B. 9 d
B.10 d
B.11 a
B.12 c
B.13 d
B.14 b
B.15 a
B.16 c
B.17 d
B.lB b
B.19 c
B.20 a
1
1 Section C Plant and Rad Monitorino Systems Page 3 ANSWER KEY C. I c
C. 2 a,2 b,8 c,3 d,5 e,7 f,6 g,1 h,10 C. 3 d&c C. 4 b
C. 5 b
C. 6 a
C. 7 a
C. 8 a
C. 9 b&a C.10 d
C.11 c&b C.12 d
C.13 d
C.14 a
C.15 c
C.16 b
C.17 a
C.18 c
C.19 b
C.20 b
l i
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