Text

Exam Rept 50-407/OL-91-01 on 911209-11.Exam Results:All Five Candidates Passed Operating Portion of Exam.Overall Written Exam Raised Many Concerns That Prompted New Written Exam to All Candidates,Following Completion of Addl Training
	ML20210E221
Person / Time
Site:	University of Utah
Issue date:	06/12/1992
From:	Doyle P; NRC
To:	;
Shared Package
ML20210E217	List: ML20210E213; ML20210E221;
References
	50-407-OL-91-01, 50-407-OL-91-1, NUDOCS 9206180417
	Download: ML20210E221 (70)
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4 ENCLOSURE 1 U. S. NUCLEAR REGVLAT0RY COMMISSl_0N OPERATOR LICENSING INITIAL EXAMINATION REEQR1 REPORT NO.: 50-407/0L-91 01 FACILITY DOCKET NO.: 50-407 FACillTY LICENSE NO.: R 126 FACILITY: University of Utah EXAMINATION DATES: December 09-11, 1992 EXAMINER: Paul V. Doyle, Chief Examiner SUBMITTED.BY: Od _ 4 I.b

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PauT' V. Doyl f ' f Examiner Date APPROVED BY: i _ k // [ %

N n-Power Reactor Section 0 erator Licensing Branch Division of Licensee Performance and Quality Evaluation, NRR

SUMMARY

The NRC administered written and operating examinations to four Reactor Operator (RO) and one Senior Reactor Operator Instant (SR01) candidates. All five candidates passed the operating portion of the operator licensing examination. The overall results on the written examination prompted the NRC staff to perform a thorough review to check on examination validity. This review raised a number of concerns which are listed in the report details. A

- meeting was held on March 23 and 24 to discuss the written examination concerns and to review the training prograr At this meeting, it was concluded that a new written examination w. ' be administered to all five candidates, following the completion of adduional training and certification by facility management that the candidates have been evaluated and determined to be ready to take the NRC examination.

9206180417 920612 PDR ADOCK 05000407 V PDR g

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REPORT DETAlts

1. Examiners:- !

Paul V. Doyle Jr., Chief Examiner, NRC Marion Daniels, Examiner (Sonalysts, Inc.) !

2. Results:

NRC Grading:

All five candidates passed the operating portion of the examination but the results indicated there were problems with the written portion of

-the examination. As a result, the NRC conducted an extensive evaluation of the written. examination. Based on concerns identified during the-grading and evaluation of the written examination, the NRC has decided to disregard the-results of the_ written e m ination administered ;

December 9, 1991. A new written examination will be administered at a later date.

3. Written Examination:

The written examination was administered on December 9, 1991 to four R0 ,

and one SRO instant candidates. All five candidates had difficulties with the examination, raising concerns regarding examination validity and- the adequacy of the . facility's training program, ,

The-NRC performed extensive reviews of the written examination including-independent reviews by. Region IV examiners. These reviews revealed some potentially confusing and inappropriate questions. . Several of the problems were due to inaccurate and incomplete reference material' provided by the facility. Enclosure 3 contains the' facility licensee's-written examination comments and the NRC's resolution to those comments.

In addition, the NRC reviewed the facility's past performance on NRC written examinations. This review revealed a history of marginal performance by University of Utah candidates on NRC written

examinations.

As a result of the performance of candidates on the written examination administered on December 9,1991 along with the history of facility license candidates, the NRC decided to meet with the facility staff on March 23 and 24, 1992,- to review the licensed operator training-program, and the written examination results. -Enclosure 2 contains a list of the partit,ipants,. findings and proposed resolutions discussed during this second meeting held on-site during March 23 and 24, 1992.

l After considering the written examination results, the incomplete and inaccurate material provided by the facility for examination generation, and the history of marginal performance on written examinations by University of Utah personnel, the NRC decided to disregard the results

/ this written examination. A new written examination will be administered at a later date.

4. Operating Examinations:

Operating examinations were administered on December 10, 1991 to the four R0 candidates and on December 11, 1991 to the SR0 instant candidate. The SR0 instant and all four R0 candidates passed this portion of the examination.

5. Exit Meeting: (December 11,1991):

Personnel attending: Dr. Gary Sandquist Director University of Utah Reactor Facility Paul V. Doyle J:. Chie" Examiner, NRC The examiners identified the following generic program weaknesses.

1. Much of the material required to prepare the examination was inaccurate or incomplete. The facility did not provide the NRC any material relating to Category A (Reactor Theory and Therco-dynamics) of the written examination. The chief examiner informed the facility director that Section A of the examination was being prepared using " Nuclear Reactor Engineering" written by Glasstone and Sesonske and " Introduction to Nuclear Engineering" written by John R. Lamarsh. The facility director acknowledged the use of these books as acceptable. The contract examiner prepared Category A using copies of these textbooks.

Mc:h of the material sent by the facility for preparing the examination contained information on systems which no longer matched the actual equipment. The facility's Safety Analysis Report (SAR) and Safety Evaluation Report (SER) discussed an old (pre-TRIGA mark 1) control panel. The faciitty Modifications Authorizations Manual discussed a new (TRIGA mark Ill) control panel. The facility did not provide any guidance on what parts of each document were applicable to the plant as tested.

The facility provided either no or inadequate material for the Nuclear Instrumentation, Rod Control, Security and the Reactor Control panel systems. In addition the SAR contained information on a water to water heat exchanger to cool the primary water. The University of Utah facility has never had a water to water heat exchanger. Instead the facility uses a Freon compressor to co.,1 the primary water.

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i Prior to the examiaation, the facility provided the examiners with a copy of the Standard Operating Procedures. These procedures are DRAFT procedures not approved for use at the facility. The procedures presently used to operate the facility, the ' Facility Operating Hanual," were not provided to the examiners.

2. During the operating portion of the examination, two Reactor Operator candidates demonstrated some weaknesses in their knowledge of Technical Specifications, although they demonstra'ed sufficient knowledge to pass this part of the examination. Since facility Technical Specifications allow the reactor operator to be the only licensed person at the control panel during some operations, it is important for all reactor operators to understand the requirements of the Technical Specifications.

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4 ENCLOSURE _Z MARCH 23-24. 1992 ON-SITE MEETING PARTICIPANTS:

Gary Sandquist, Reactor Director, Univ, of Utah Kevin Crawford, Idaho State University (Former Reactor Supv. - V. of Utah)

Michael Slaughter, SROI candidate, Univ, of Utah Henry Moeller, R0 candidate, Univ. of Utah Byron Hardy, R0 candidate, Univ. of Utah k Paul V. Doyle Jr., Chief Examiner, NRC James L. Caldwell, Chief, Hon-Power Operator Licensing Section, NRC Marvin Mendonce, Senior Project Manager, NRC -

Blaine Murray, Section Chief, Region IV, NRC.

Mary , Ann Biamonte, Training and Assessment 3pecialist, NRC BACKGROUND:

The-University of Utah originally requested that an examination be given in October, 1991. A week before the examination they requested that the date be changed to give the candidates more time to prepare. The exam was conducted the week of December 9,1991.

All of the candidates passed the operating test but the written examination results were not typical for a non-power facility. Because of the atypical performance the staff conducted extensive reviews of the written examination including independent reviews from Region IV personnel. Numerous problems were identified by both the NPC and the facility, Several of these problems were due to the incomplete end inaccurate' reference material provided by the facility for the preparation , the examination. Following correction of the applicable problems, the candidate's grades improved but overall performance still indicated weaknesses. Additionally, the staff reviewed the facility's ~

past performance on NRC written examinations. The review revealed a history of marginal performance.

. A conference call was held with the facility on February 7,1992 to discuss the results of the examination, to express the staff's concerns regarding the adequacy o' the training program and to schedule a meeting at the site. The meeting we scheduled for March 23 and 21,1992.

After considering the written examination results, the incomplete and inaccurate material provided by the facility for examination generation, and the history of marginal performance on written examinations by University M Utah personnel, the NRC decided to disregard the results of this written examination. A new written examination will be administered at a later date.

k MEETING DETAILS:

The meeting with the facility was conducted at the University of Utah on i March 23 and 24, 1992. The following information concerning the facility's licensed operator training program was identified through discussions with facility management, operators and candidates:

1. The facility has not established a formal training program or learning objectives to prepare operator licensing c edidates.

2. The training program which was in place .w 6 recently been compared against 10 CFR 55 or ANS/AF I St 9 dard -

Sunerican National Standard for the Selection and Training r Personnel for Research Reactors." _

3. The training program that exists consisted of the following:

a. Completion of selected undergraduate and/or graauate level Nuclear Engineering classes to ensure adequate knowledge of reactor theory. However, credit was given for passing courses up to 2 years prior to the licensing examination without a formal reevaluation of the candidate's knowledge,

b. Presentation of 'a series of 14 informal 1 to 1-% hour lectures.

Attendante at these lectures was NOT mandatory and there was no formal evaluation of the candidate's understanding following the lectures,

c. Performance of self study to complement the lectures. No formal evaluation of the candidates' use of self-study was done, as recommended by ANS/ ANSI 15.4.

d. Performance of on-the-job training and plant walk-downs whu.h were -

mostly student initiated. The candidates present at the meeting indicated that they had rore concern for the operating " t portian of the examination and therefore placed more emy .. sis on preparing for it than on the written examination.

4. The facility had no formal means of evaluating the knowledge of the candidates. The evaluation methods used were as follows:

a. Obtaining a grade of B or better on the college level Nuclear Engineering courses,

b. Performance on previous NRC examinations on file at the facility.

c. The facility Director's belief that the candidate was ready.

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RESOLUTIONS:

In the short term, the facility Reactor Director has committed to completing the following actions prior to the staff administering a new written

- examination.

1. Conduct additional training for the candidates who were examined in December 1991.

2. Perform evaluations of the candidates to ensure they are ready to take an NRC examination.

3. Provide the NRC with a set of complete and accurate reference material for the generation of the new examination.

4. - Provide the NRC with a letter listing the additional training performed to 'prepare the candidates and to certify that the candidates are ready for an NRC administered written examination.

For-the long term, the facility committed to performing the following corrective actions to improve their training program:

1. Develop a consolidated training manual for use by the candidates to prepare for NRC operator licensing examinations and by the NRC for use in preparing the NRC examinations. The facility has already started preparing this manual. This manual would provide the information necessary to compensate for information lacking in the SAR, the SER cnd the facility Modifications Authorizations Documents.

2. Conduct a review to ensure the operator licensing training program adequately addresses 10 CFR 55 and ANS/ ANSI Standard 15.4.

3. Develop a formal checklist of requirements / objectives to be used in evaluating candidates readiness to assume operator licensing duties.

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4. Develop evaluation methods necessary to objectively assess the knowlt.dge and understanding level of the candidates.

T ENCLOSVRE 3 SECTION A

1. Which ONE of the following will be the resulting stable reactor period when a $0.25 reactivity insertion is made into an exactly critical reactor core.

a. 50 seconds

b. 38 seconds

c. 30 seconds _

d. 18 seconds Facility comment:

The correct answer is both b and c. If the value for the U-235 one delayed neutron group decay constant is 0.00 1/s as given in Glasstone & Sesonske, is used then answer obtained is 37.5 s which is equivalent to answer b.

Resolution:

Agree. Change the answer key to recognize either b or c as correct.

10. Which ONE of the following is the purpose of the neutron source used in the core during a reactor startup?

a. To prevent a very short period from occurring without indication to the operator,

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b. To ensure adequate neutrons are available to initiate the fission chain reaction.

c. To provide delayed neutrons for added control of the reactor period.

d. To offset the negative reactivity added to the core by fission

product poisons.

Facility Comment:

Clearly "a" is a correct response as indicated in paragraphs 5.279 through 5.281 (pages 322-323) in Glasstone & Sesonske (Sec 5.261 quoted in examination solution references refers to pulse type ion chambers and is apparently an incorrect reference.

Resolution:

Agree. Change answer key to make a the correct answer.

E g i SECTION A 12.;Which ONE of the following describes the Core Delayed Neutron Fraction change as a function of core age (Burnup)?

a.- - Decreases due to depletion of U-238

b. Decreases due to build-up of Pu-239

-c. Increases due to depletion of U-235

d. Increases-due to build-up of Pu-240

Facility Comment:~ Request question be deleted.

The production of PU-239 in the TRIGA core is very small because of the 20%

enriched fuel and the resulting buildu) of fission products from U-235, -

particularly Sm-149-ano other stable aasorbers which affect energy dependent neutron absorption in a complicated manner. U-238 depletion is associated with-Pu-239 buildup, so both responses a and b are related and true. The

effect of' depletion of U-235 in a 20% enriched core is complex and can increase 1the delayed neutron fraction because of energy dependent neutron capture of the energy spectral differences between delayed and prompt neutrons. -The answers to this-question are subtle and somewhat ambiguous -- 1 request _that the_ question be eliminated. Further the Glasstone & Sesonske

= references given-in theLanswer sheet are inconsistent: viz pg. 93 does not contain-Section 2.169 -- Delayed neutrons are described on page 110, section 2.190 -2.193.

. Resolution:

- Disagree.- This is a-fundamental knowladge for a TRIGA reactor. Operators are

-expected to have-a good knowledge of-Reactor Physics.

A:

-15. Which ONE of the-following is the principal source of heat in the reactor after a. shutdown from extended operation at 100 Kw? -

a. Production of-delayed neutrons

b. Subtritical reaction of photo-neutrons

.c. Spontaneous fission of U-238

d. Decay of fission fragments.

Facility Comment:-

The correct answer _for our TRIGA reactor depends on the quantity of D 20 located'within the core. With the two of the refloctor tanks installed the

-production of photoneutrons 'from-gamma, n reactions in deuterium can actually-resultLin. sufficient fission reactions (described by both answers a and b) that subcritical fission can be a significant post shutdown energy source.

'Again the: answer sheet' references are incorrect. Fission product decay is described in.pages 120 - 125 in Glasst2.ie & Sesonske, not pages 99 - 102.

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SECTION A Resolution:

Disagree. The question asks for the PRINCIPAL source of heat in the reactor following a shutdown from extended operation. Even if there were sufficient fissions to be measured, at stated in the facility's comments, the PRINCIPAL source of heat would be decay of fission products.

16. A reactor startup is being conducted 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after a reactor scram.

Reactor power has been increased to 50 Kw. Which ONE of the following describes the response of reactor power witnout any further operator actions? _

a. Power increases due to the burnout of the Xenon remaining in the -

reactor.

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b. Power increases due to the burnout of the Samarium remaining in the reactor following the scram.

c. Power decreases due to the buildup of Xenon from the decay of lodine-135 and Uranium-235 fissions.

d. Power decreases due to the buildup of samarium in the reactor from the decay of Promethium-149.

Facility Comment: Request question be deleted.

The answer is ambiguous. The question states that the reactor power is increased to 50 kW implying that a lower power level was prevalent for some unknown time period. Suppose as part of the startup the reactor had been brought to I kW or higher to stabilized operation and remeva the neutron source. Depending on the length of time for this stabilization, Xenon-135 buildup could be more significant than burnup of Sm-149 which is a stable isotope. Furthermore, the contribution of Sm to the reactivity of the UUNEL TRIGA is about 1% of that from Xe-135 and is negligible for our oper;tions.

We request that the question be omitted because of its inapplicability to our operations.

Resolution:

Agree. The burnout of samarium is insignificant with respect to the buil 9 of Xenon. In addition the question stem is quite vague in the descriptio,. of the increase in power. The answer key has been changed to delete this question.

SECTION A

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18. A reactor startup is in progress. Reactor power is at 0.1 Kw and the operator has established a 30 second period. Assuming no operator actions are taken, reactor power will continue to increase on a 30 second period until power reaches:

a. 1 Kw, then the reactor period begins to DECREASE

b. 1 Kw, then the reactor period begins to INCREASE

c. 10 Kw, then the reactor period begins to DECREASE _

d. 10 Kw, then the reactor period begins to INCREASE Facility Comment:

The large prompt negative temperature coefficient of reactivity strongly influences the reactivity in the UUNEL TRIGA core. However, a delayed positive temperature coefficient also exists because our TRIGA core is overmoderated (excess H density in water over optimum moderation and absorption) and an increap in the moderator temperature decreases the moderator density which increases the reactivity. A thirty second period is in the transiant power range 4.here both prompt and delayed effects are prevalent. Thus both answers a and c are possible and c is also possible depending on precore status - i.e. cold clean critical, etc.

Resolution:

Disagree with comment. For a given time period any of the four answers could be correct. From a strictly Reacttir Physics point of view for a short time under the given conditions the reactor will experience a decrease in reactor period. This will very quickly change to an increase in reactor period with .

further temperature changes in the core. Based on the ambiguity of the question, the answer key has been changed to delete this question.

4 SECTION B

6. For the Technical Specification limits listed in column A SELECT the appropriate value from the values listed in column B. (Items in eclumn B may be used once, more than once, or not at all, and only one answer may occupy a space in column A.) (4 required at 0.50 each)

a. Maximum allowable temperature 1. 460*

in a stainless steel, hydride 2. 530*

fuel element. 3. 800'

4. 1000*

b. Maximum allowable excess reactivity 5. 50.30 for a cold critical Xenon free core. 6. $0.50

7. $1.18

c. Maximum Shutdown margin. 8. $2.80

' d. Maximum allowable temperature in an aluminum, hydride fuel element.

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Facility Comment:

Based on the information provided in Technical Specifications Section 3.2(4),

the maximum allowable excess reactivity is $2.80. The value of $1.18 applied to a specific core which was_ referenced in the SER (1985) and has not been employed since then. Thus both 7 and 8 are acceptable answers for part b of this question.

Resolution:

Agree in part. The candidates should answer this question based on the core installed. Therefore the answer key has been changed to recognize 8 as correct for part b of this question.

8. A person is working 4 feet from a gamma point source emitting 8 R/hr at one foot. Which ONE of the following is the length of time that the person can work without exceeding the whole body QUARTERLY 10CFR 20 does limit?

a. 37 minutes

b. 1.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months

c. 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

d. 6.0 hours0 days 0 hours 0 weeks 0 months Facility Comment:-

Based on the calculation the dose rate is 0.5 R/hr at 4 ft. If one uses 1.25 R/qtr_as given in 10CFR 20.101(a) then you obtain 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the work time.

If you use 3 R/qtr as listen in 10CFR 20.101(b) then you obtain 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for

- the work time. Both answers respond to 10CFR20 requirements.

Resolution:

Agree. The answer key has been modified to recognize either e or d as correct.

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SECTION C

5. With the reactor at 100 Kw and the heat exchanger operating, which ONE of the following is the rate of rise of the pool temperature?

a. 1*F/hr

b. 2*F/hr

c. 5'F/hr

d. 7'F/hr Facility Comment: Request b be accepted as a correct answer.

Both b and c should be acceptable. The value of 5'F/hr was given in the SAR (which assumed a GA water-water heat exchanger) written in 1975 before the present heat exchanger was installed. With the present heat exchanger (a refrigerative exchanger) we experience a temperature rise of about 2*F/ hour.

Resolution:

Agree in part. The candidates should answer this question based on the core as installed. The answer key has been to recognize b as the correct response to this question.

7. Two alarms will provide sight and sound indication for high reactor radiation conditions at the campus security headquarters. Which ONE of the following specifies the activating devices for the two alarms?

a. The Reactor Tank Monitor and the Neutron Generator Monitor.

b. The Reactor Tank Monitor and the Reactor Pool Low Water Level Monitor.

c. The Monitor immediately over the core and the Neutron Generator Monitor,

d. The Monitor immediately over the core and the reactor Pool Low Water Level Monitor.

Facility Comment:

Answers b and d appear to be the same since.the reactor tank monitor and the monitor immediately over the core can be construed to be the same. Request that both b and d be considered acceptable.

Resolution:

Agree. The answer key has been modified to recognize either b or d as correct.

1 LLCTION C

8. For the items labeled a through d on figure 3 (attached) and listed in column A, select the correct component from the items ir, column B.

(Items in column B may be used once, more than once, or not at all. Only one answer may occupy a space in column A.)

a. 1. Speed increasing gear

b. 2. Motor and Reduction gear

c. 3. Limit Switches

d. 4. Drum

5. Helipot -

6. Torque-converter

7. Brake

8. Drum Facility Comment:

We currently have rack-and-pinion drives as shown in MA-2 (Modification Authorization). The trainees were not introduced to the winch drives because those drives are no longer used to move the control rods.

Resolution:

Agree. The answer key has been modified to delete this question.

10. Which ONE of the following describes the method of supplying power to the radiation monitors and the facility intrusion detectors?

a. Directly from the 120 VAC facility service transformer only,

b. 24 VDC from 120 VAC rectifiers normally and a 24 VDC battery on loss -

of AC power,

c. 12 VDC from 120 VAC rectifiers normally and a 12 VDC battery on loss of AC power.-

d.. Directly from a facility 12 VDC battery only.

Facility Comment:

D implies the radiation monitors are powered by batteries only, with no connection to an external power system, in reality, we have a 12 volt battery continuously charged by a trickle charger that is connected to our 120 VAC facility power supply. Thus both c and d are technically correct.

Resolution:

Agree . The answer key has been modified to accept either c or d as correct.

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SECTION C

17. From the nuclear instrumentation items labeled a through d on figure # 2 (attached) and listed in column A SELECT the component identification from column B. (Items in column B may be used once, more than once or not at all, and only one answer may occupy a space in column A.)

a. 1. Power Level amplifier

b. 2. Pre-amplifier

c. 3. Magnet Scram Control

d. 4. Log-n Amplifier

5. Compensated Ion Chamber

6. Fission Counter -

7. Log-Power Switch

8. Linear Power Switch

9. Linear Count Rate Counter Facility Comment:

We currently have a TRIGA Hark Ill console as shown in the MA-2 (Modification Authorization). The schematic provided during the examination was that of the old Mark I console which has been removed and dismantled. The applicants have not trained on that console and were not required to learn the old instrumentation diagrams.

Resolution:

Agree in part. The candidates should answer this question based on the facility as it is presently configured. However, because the material (fission chamber, log amp [ log-n amp), and power amp, that was referenced in this diagram did nql change appreciably from what is now installed no change to the answer key is required.

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ENCLOSURE 4 U. S. NUCLEAR REGULATORY COMMISSION NON-POWER REACTOR LICENSE EXAMINATION FACILITY: Univ, of Utah REACTOR TYPE: TRIGA-1 DATE ADMINISTERED: 91/12/09 REGION: 4 CANDIDATE:

LICENSE APPLIED FOR:

INS 1 RUCTIONS TO CANDIDATE:

Answers are to be written on the exam page itself, or the answer sheet provided. Write answers one side ONLY. Attach any answer sheets to the examination. Points for each question are indicated in parentheses for each question. A 70% in each section 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 21.00 32.31 A. REACTOR THEORY, THERMODYNAMICS AND FACILITY OPERATING CHARACTERISTICS -

20.00 30.77 B. NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS 24.00 36,97 C. PLANT AND RADIATION MONITORING SYSTEMS 65.00 % TOTALS FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.

Candidate's Signature l

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' A. RX THEORY. THERMO ~ & FAC OP CHARS Page 2 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

4 MULTIPLE CH0 ICE 001 a b c d 002 a b c d 003 a b c d 004 a b c d

'005 a b c d 006 a b c d 007 a b c d 008 a b c d 009- a b c d 010 a b c d 011 a b c d 012 a b c- d 013 a. b c d 014 a b c d 015 a b- c d

016 a b c d 017 a b c d 018 a b c d 019 a b c d 020 a b c d 021 a b c d

(***** END OF CATEGORY A *****)

'B. NORMAL /EMERG PROCEDURES & RAD CON Page 3 ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank.

MULTIPLE CHOICE 018 a b c d 001 a b c d 019 a b c d _

002 a b c d 003 a b c d 004 a b c d 005 a .b c d 006 MATCHING a

b C.

d MULTIPLE CH0 ICE 007 a b c d 008 a b c d 009 a b c d 010 a b c d 011' a b c d 012 a b c d _

013 a b c d 014 a b c d 015 a b c d 016 a b c d 017 a b c d

(***** END OF CATEGORY B *****)

C. PLANT AND RAD MONITORING SYSTEMS. Page 4

' ANSWER SHEET Multiple Choice (Circle or X your choice)

If you change your answer, write your selection in the blank, MULTIPLE CH0 ICE MULTIPLE CH0 ICE 001 a b c d 010 a b c d 002 a b c d 011 a b c d 003 a b c d 012 a b c d 004 MATCHING 013 a b c d a 014 MATCHING b' a C b _

d c MULTIPLE CH0 ICE d 005 a b c d MULTIPLE CH0 ICE 006 a b c d 015 a b c d 007- a b c d 016 a b c d 008 MATCHING 017 MATCHING' a a

'b b c c d d 009. MATCHING MULTIPLE CH0 ICE a 018 a b c d b 019 a b c d c

d

(***** END OF CATEGORY C *****)

(********** END OF EXAMINATION **********)

o , ?NRC RV_LES AND GUIDELINES FOR LICENSE EXAMINATIONS Duringithe administration of this examination.the following rules apply:

li Cheating on the examination means an automatic denial of your application .

and could resultnin 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 I 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 blackfink or dark pencil only to facilitate legible reproductions. 7

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 inLthe 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 after the question.

- 10. Partial credit will NOT be given.

Ell. If the intent of a. question is unclear,-'ask questions of the examiner only.

-12.'When you are done and have turned in your examination, 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.

'A. RX THEORY. THERMO & FAC OP CHARS Page 7 4

u QUESTION: 001 (1.00)

Which ONE of the following will be the resulting stable reactor period when a

$0.25 reactivity insertion is made into an exactly critical resctor core?

a. 50 seconds

b. 38 seconds

c. 30 seconds

d. 18 seconds QUESTION: 002 (1.00)

Reactor Corc Data:

Core Excess Reactivity $1.30 Safety Rod Worth $1.76 Shim Rod Worth $1.55 Regulating Rod Worth 50.46 Which ONE of the following is the estimated shutdown margin of a fully loaded core?

a. 52.47

b. 50.92

c. $0.71

d. $0.50 -

QUESTION: 003 (1.00)

Which ONE of the following is the definition of the term Lf in the formula K,,, = K in ,ini,, L,n L,?

a. The fast fission factor

b. The fast neutron slowing down length

c. The fast neutron diffusion length

d. The fast non-leakage probability

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. RX THEORY. THERM 0 & FAC 03 CHARE Page 8 QUESTION: 004 (1.00)

As reactor power increases, a core reactivity loss (power defect) occurs as described by Figure A-1, Estimated Reactivity Loss vs Power. (See attached figure.)

Which ONE of the following states the factors that contribute to this loss of core reactivity?

a. hoderator and fuel meat temperature increase.

b. Temperature defect ano menon poisoning.

c. Xenon and samarium poisoning.

d. Doppler broadening and heavy water density decrease.

QUESTION: 005 (1.00)

Which ONE of the following is.the Keff of a reactor that is supercritical on a stable 60 second period?

a. 1.001

b. 1.0015

c. 1.002

d. 1.0025

(***** CATEGORY A CONTINtT? ON NEXT PAGE *****)

'A. RX THE0RL. THERM 0 & FAC OP CHARS Page 9 QUESTION: 006 (1.00)

Which ONE of the following statements describes the suberitical reactor response as Keff approaches unity? For a given change in Keff, the reactor will show:

a. A LARGER change in neutron level and a SHORTER period of time is required to reach the equilibrium neutron level.

b. A LARGER change in neutron level and a LONGER period of time is required to reach the equilibrium neutron level.

c. A SMALLER change in neutron level and a SHORTER period of time 1.s required to reach the equilibrium neutron level.

d. A SMALLER change in neutron level and a LONGER period of time is required to reach the equilibrium neutron level.

QUESTION: 007 (1.00)

A reactor startup is in progress. The reactor has just become critical and reactor power is 0.1 KW. The operator establithes a 45 second period.

Which GNE of the following is the amount of time required for the reactor to reach 1 KW7

a. I second

b. 45 seconds

c. 104 seconds

d. 450 seconds

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

'3 RX THEORY. THERM 0 & FAC OP CHARS Page 10

(

QUESTION: 008 (1.00)

Which ONE of the following is the MAJOR contributor to the large negative temperature coefficient of reactivity?

a. Decreased reflection of thermalized neutrons due to the heavy water reflector temperature increase,

b. Decreased moderator density due to the rapid moderator temperature increase,

c. Increased doppler broadening of the U-238 due to the fuel meat temperature increase. -

d. Increased neutron leakage from the fuel meat due to the ZrHx temperature increase.

i QUESTION: 009 (1.00)

Which ONE of the following describes the MAJOR contribution to the production and depletion of xenon in the reactor?

a. Produced from radioactive decay of iodine and depleted by neutron absorption only

b. Produced from radioactive decay of iodine and depleted by radioactive decay and neutron absorption

c. Produced directly from fission and depleted by neutron absorption only

d. Produced directly from fission and depleted by radioactive decay and __

neutron absorption-(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A '. RX THEORYJTHERM0 & FAG _0P CHARS 'Page 11

-QUESTION: 010:'(1.00)'

-Which ONE:of the following 'is the purpose of the neutron source used in the-core during:a reactor startup?- _

- a'. 'To prevent a very short period from occurring without indication to the operator.

b. -To ensure adequate neutrons are available to initiate the fission chain reactior.-

c. To provide delayed .neotrons for added control of the reactor period.

ld. 'To~ offset _the negative reactivity added to the core by fission product poisons.

QUESTION: 011 ~ (1.00)'

LThe count rate hasLincreased from 70 cps to 100 cps in a subcritical reactor.

.Which ONE of the following is the reactivity added to the reactor if tha

. initial lvalue' of Keff was 0.957~

, a. 1.50% delta K/K

b. 1.64%-delta K/K

c. :1.93%- delta K/K

d. 2.43% delta K/K QUESTION: 012 (1.00)

Which ONE of _ the_ following describes the Core Delayed Neutron' Fraction change as:a-function'of-core _ age (burnup)?-

a. Decreases,due to depletion of U-238

b. Decreases due-to build-up:of Pu-239

c. LIncreases due to-depletion of:U-235

'd . Increases due_to buildup of.Pu-240

. (***** CATEGORY- A CONTINUED ON'NEXT PAGE *****)

'.- ..m _ _ __ _ _ _ _ . _ . . - _ _ _ _ _ _ _ _ . . . _ . _ _ . _ . . . . - _ . _ . . _ . _ _ , . . _

.A. RX THEORY. THERM 0 & FAC OP CHARS Page 12 QUESTION: 013 (1.00)

Which ONE of the following fuel element arrangements in the core will result in the highest fuel element reactivity worth?

a. A stainless steel clad fuel element placed in the C-ring.

b. A stainless steel clad fuel element placed in the D-ring.

c. An aluminum clad fuel element placed in the B-ring.

d. An aluminum .; lad fuel element placed in the E-ring.

QUESTION: 014_ (1.00)

The reactor has just become critical when the operator withdraws a control rod and adds 0.05% delta K/K to the core.

Which ONE of the following describes the reactor period immediately following the reactivity addition?

a. Longer than the subsequent reactor period due to the delay in the neutron population increase from both long-lived and short-lived delayed neutron precursors,

b. Longer than the subsequent reactor period due to the delay in the nat7cn population increase from the long-lived delayed neutron precursors.

c. Shorter than the subsequent reactor period due to the immediate increase in the neutron population from prompt neutrons.

d. Shorter than the subsequent reactor period due to the immediate increase in the neutron' population from the short-lived delayed neutron precursors.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. RX THEORY. THERri0 & FAC OP CHARS Page 13 QUESTION:-015 (1.00)

Which ONE of the following is the principal source of heat in the reactor after a shutdown from extended operation at 100 KW7

a. Production of delayed neutrons

b. Subcritical reaction of photoneutrons

c. Spontaneous fission of U-238

d. Decay of fission fragments s

QUESTION: 016 (1.00)

A reactor startup is being conducted 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after a reactor scram. Reactor power has been increased to 50 KW.

Which ONE of the following describes the response of reactor power without any further operator actions?

a. Power increases due to the burnout of the xcaon remaining in the reactor following the scram,

b. Power increases due-to the burnout of the samarium remaining in the reactor following the scram.

c. Power decreases due to the buildup of xenon from the decay of Iodine-135 and Uranium-235 fissions. -

d. Power decreases due to the buildup of samarium in the reactor from the decay of Promethium-149.

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

1

O l 'A. RX THEORY. THERM 0 & FAC OP CHARS Page 14 l

QUESTION: 017 (1.00)

Which ONE of the following describes the characteristics of Samarium after a reactor shutdown from full power operation?

a. Increases to a peak value and then decays to nearly zero after 1 week.

b. Increases to an equilibrium shutdown value after 2 weeks,

c. Decreases to an equilibrium shutdown value after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. Remains the same as the equilibrium power operation value indefinitely.

QUESTION: 018 (1.00)

A reactor startup is in progress. Reactor power is at 0.1 KW and the operator

,has established a 30 second period.

Assuming no operator actions are taken, reactor power will continue to increase on a 30 second period _until power reaches:

a. I KW, then the reactor period begins to DECREASE.

b. 1 KW, then the reactor period begins to INCREASE.

c. 10 KW, then the reactor period begins to DECREASE.

d. 10 KW, then the reactor period begins to INCREASE.

s

(***** CATEGORY A CONTINVED ON NEXT PAGE *****)

' - - - __--_.-___-____u_ _ _ . _ _ _ - - _ _ _ _ . _ _ - _ _ _ . _ _ _ _ . - - - _ _ _ _ _ _ _ _ _ _ _ . - _ . _ _ _ _ _ _ _ _ _ _ . _ . . _ _ . _ _ _ _ - _ _

. . , .- . . . . - . = - - - ~ - -. - - - - ~ . - - -- - --

": A ; RX-THEORY.-THERMO & FAC OP CHARS 'Page 15 LQUESTION: 019 (1.00)

Technical Specifications limit the TOTAL WORTH of an individual experiment e

if it-is not securely fastened into the reactor.

Which ONE of the following is the reason for this limitation?

a. To prevent centerline fuel hy>54de_ melting upon accidental rapid insertion at power.

~ '

- b. -To prevent inadvertent criticality upon reaching the cold xenon free condition following a reactor shutdown.

c. To prevent a radiation release in excess of 10 CFR 20 limits from a single experiment accidental exposure. ,

d. 'To prevent exceeding the fuel temperature safety limit-should a reactivity

. pulse insertion occur due to handling the 9xperiment.

QUESTION:-020 .(1.00)

Which ONE- of the' followingLreactor components provides axial reflection of ,

neutrons.for.the core?

~

a. -The aluminum top ano bottom grid plates-

b. _The graphite plugs on _each end of the fuel element-

c. 1The heavy water. elements inserted-into the core

d. .The. heavy water tank surrounding the core '

t

(*****; CATEGORY A CONTINUED ON NEXT PAGE *****)-

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~.. , -. . . , . , _- ...-.-- -.. - - . . . - . - , - . - -.- . . . . - .

A RX THEORY. THERMO & FAC OP CHARS Page 16 QUESTION: 021 (1.00)

Reactor power has just been increased from 1 KW to 50 KW.

Which ONE of the following describes the effect of the subsequent increase in the moderator and fuel temperateres on the control rod worths (CRW)?

a. CRW ir. creases because neutron leakage out of the fuel elements increases,

b. C?W increases because the neutron slowing down time decreases,

c. C(W decreases because the neutron slowing down length increases,

d. CIW decreases because the moderator density decreases.

(***** END OF CATEGORY A *****)

'E. NORMAL /EMERG PROCEDVRES & RAD CQN Page 17 QUESTION: 001 (1.00)

Which ONE of the following describes the condition of the reactor when it '..

SHUTDOWN?

a. The console key switch is in the "0FF" position and the key is removed from the console

b. All control rods are fully inserted into the core and there is no G refueling or maintenance in progress

c. The reactor is sub-critical by at least $1.00 of reactivity even if all rods are not fully inserted

d. The reactor is sub-critical and the fuel and bulk water temperatures are -

less than 40*C QUESTION: 002 (1.00)

Which ONE of the following is the fuel element failure that is prevented by limiting the maximum temperature of the fuel element ?

a. Ductile failure of the fuel element cladding due to cladding creep and thermal stresses at elevated temperatures

b. Fuel element cladding failure due to-the pressure buildup in the fuel element from fission gases and fuel moderator disassociation

c. Fuel element centerline fuel melting due to exceeding the melting point of the fuel hydride

d. Cladding " hot spot perforations" due to the fuel to cladding interaction -

caused by excessive fuel swelling

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

-*B. NORMAL /EMERG PROCEDURES & RAD CON Page 18 QUESTION: 003 -(1.00)

Which ONE of the following is the minimum number and type of personnel required when the reactor is critical or approaching critical?

a. A licensed operator at the control console, another licensed operator within the reactor laboratory and a senior reactor operator on call

b. A licensed operator at the console and a senior reactor operator within the reactor laboratory

c. A licensed operator at the console,_another responsible person within the reactor laboratory and a senior reactor operator on call

d. A licensed reactor operator at the console and another responsible person within the reactor laboratory

.QUESTJN: 004 (1.00)

Which ONE of the following radicactive gases poses the most significant hazard at the University of Utah TRIGA reactor?

a. Argon-41 i

b. Nitrogen-16

c. Oxygen-18

d. Tritium QUESTION: 005 (1.00)

Which ONE of the following is the maximum permissible QUARTERLY exposure to radiation at the University of Utah TRIGA reactor?

a. 300 millirem

b. 1250 millirem l c. 3000 millirem

d. 5000 millirem l (***** CATEGORY B CONTINUED ON NEXT PAGE *****)

i

B. NORMAL /EMERGJRQ((9V.8ES & RAD CON Page 19 QUESTION: 006 (2.00) 1 for the Technical Specification limits listed in column A SELECT the appropri- l ate value from the values listed in column B. (Items in column B may be used i once, more than once, or not at all, and only one answer may occupy a space in

-column A).

(4 required at 0.50 each)

COLUMN A COLUMN B (LIMITS) (VALUES)

a. Maximum allowable temperature 1. 460'C in a stainless steel, hyJride fuel ele: '

2. 530'C

b. Maximum allowable excess reactivity 3. 800'C for a cold critical xenon free core '

4. 7006

c. Minimum shutdown margin.

5. $0.30

d. Maximum allowable temperature.in an aluminum, hydride fuel element .. 50.50

7. $1.18

8. $2.80 QUESTION: 007- (1.00)

In accordance with the Technical Specifications, which-0NE of the following is the longest period of tinie that the reactor may be operatti with the ventilation system shut down?

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

b. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

.c. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

d. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

(***** CATEGORY B CONTINUED ON NEXT pAGE *****)

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- .. . _ - . - _. - - _ . ~_ - . . . _ . _ - _ . _ -

'B. NORMAL /EMERG FROELDMB[S & RAD CqN Page 20 l QUESTION: 008 (1.00)

A person is working 4 feet from a gamma point source emitting 8 R/hr at one foot.

Which DNE of the following is the length of time that the person can work

without exceediri; the whole body QUARTERLY 10 CFR 20 dose limit?

a. 37 minutes

b. 1.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months

c. 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months

d. 6.0 hours0 days 0 hours 0 weeks 0 months i

~

QUESTION: 009 (1.00)

' Materials are to be transferred to an off campus user. The container has a RADIDACTIVE-YELLOW III label attached te it.

Which ONE of the following is'the expected radiation level at the surface of the container?

a. At least 0.5 mr/hr but less than 50 mr/hr

b. At least 50 mr/hr but less than 100 mr/hr

c. At least 100 mr/hr but less than 200 mr/hr

d. At least 200 mr/hr l

l (*** CATEGORY B CONTINUED ON NEXT PAGE ****')

1

'B. NORMAL /EMERG,PROCED4RES & RAD CON Page 21 QUESTION: 010 (1.00)

A cobalt-60 source has been dropped in the reactor laboratory. Thirty (30) feet from the source a radiation reading of 100 mrem /hr has been detected.

Which ONE of the following is the curic content of the source? (Assume a 1.2 and a 1.3 MeV gamma emission.)

a 90 curies

b. 30 curtes

c. 6 curf:.s

d. 2.5 curies QUESTION: 011 (1.00)-

An experiment has been removed from the reactor. A radiation reading of 1 Rem /hr was recorded at 3 feet when the experiment was removed. Fifteen (15) minutes later a reading of 750 mrem /hr was recorded at the same distance (3 feet).

Which ONE of the following is the length of time (AFTER REMOVAL FROM THE REACTOR) that will be required for the radiation level to decrease to 10 mrem /hr at one (1) foot?

a. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months

b. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months d 12-hours i

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(***** CATEGORY B CONTINUED ON NEXT PAGE *****) ,

1

. .- __ _ _ _ _ . _ . _ . _ _ _ _ _ _ _ __ __ _ _ _ _ _ _ _ _ ~ _ _.

B. NORMAL /EMERG PROCEDEQ h PAQl0N Page 22 QUESTION: 012 (1.00)

A special experiment has been prossed for irradiation in the reactor and involves the production of radioisotopes.

Which ONE of the following specifies the required evaluation and approval far the experiment?

a. The experiment must be evaluated by the Reactor Supervisor, reviewed by the Reactor Safety Committee, approved by the Health and Safety Committee and designated as Class I

b. The experiment must be evaluated by Radiological Health and Safety Committee approved by the Reactor Safety committee and designated as Class I

c. The experiment must be reviewed by the Radiological Health and Safety Committee, approved by the Reactor Supervisor and designated as Class 11

d. The experiment must be evaluated by the Reactor Supervisor, approved by the Reactor Safety Committee and the Radiological Health and Safety Conaittee, and designated Class 11 QUESTION: 013- (1.00)

Which ONE of the following is the maximum fixed alpha contamination allowed on exposed surfaces in UNCONTROLLED radiation areas?

a. 10 counts per minut per square centimeter

b. 10 counts per minute per 100 square centimeters

c. 100 counts per minute per 10 square centimeters

d. 100 counts per minute per 100 square centimeters l

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(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

'B. NORMAL.1EMERG PROCED9RES & RAD C0tj Page 23 QUESTION: 014-(1.00)

Which ONE of the following is the lowest level of approval for deviation fro a Standard Operating Procedure to accomplish a task in the safest and most efficient manner?

a. A Senior Reactor Operator

b. Laboratory Management

c. Reactor Safety Committee ;

d. Radiation Safety Committee QUESTION: 015 (1.00)

Which ONE of the following specifies the dosimetry requirements for visitors '

being escorted while.the reactor is SHU100WN?

a. Each visitor is required to carry a pocket dosimeter if the visitor is not wearing a badge

b. One person in the group of visitors must wear a pocket dosimeter, in addition to the dosimetry worn by the escort

c. Each visitor is required to wear both a badge and a pocket dosimeter

d. One person in the group of visitors is designated to wear dosimetry, but the escorts dosimetry fulfills this requirement.

QUESTION: 016 (1.00)

Which ONE of the following statements describes the retention of Operations Records?

a. Both Procedure Logs and fuel Logs need only be retained for five (5) years

b. Procedure logs need only be retained for five (5) years, but Fuel Logs need to be retained indefinitely

c. Fuel' Logs need only be retained for five (5) years, but Procedures Logs need to be retained indefinitely

d. Both Procedures Logs and fuel Logs need to be retained indefinitely

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

-. -- . . ~ . _ _ . ... -_ _

'Bi_30.StiAl/EMERG PROCEDURES & RAD CON Page 24

-00ESil0N: 017. (1.00)

Which ONE of the following is the maximum rate of positive reactivity addition that can be inserted by withdrawing a shim rod from the core?

a. 0.64% delta K/K/ minute

b. 0.58% delta K/K/ minute

c. 0.37% delta K/K/ minute

d. 0.25% delta K/K/ minute QUESTION: 018 (1.00)

Which ONE of the following statements describes the MINIMUM personnel requirements for performing the Prestart Checklist?

a. A licensed operator (R0 or SRO) must perform the Checklist, but an SRO must be present while performing steps which require control rod movement or magnet activation, b.- Any authorized laboratory personnel may conduct the Checklist, but an SR0 must be present while performing steps which require control rod movement or magnet activation.

c. A licensed operator (R0 or SR0) must. perform the Checklist, but only an SR0 may perform the steps which require control rod movement or magnet activation.

d. Any laboratory authorized personnel may conduct the Checklist, but a licensed operator (R0 or SRO) must be present while performing steps which require rod movement or magnet activation.

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL /EMERG PROCEDURES & RAD CON Page 25 QUESTION: 019 (1.00)

Which ONE of the following describes the function of the helipots on the control rod drives?

a. The helipot senses rapid rod motion in the withdraw direction and inhibits rod withdrawal.

b. The helipot measures " stretch" on the control rod cable and activates an alarm if stretch exceeds preset values.

c. The helipot measures control rod position and provides control rod position to the console.

d. The helipot senses stress applied to the torque converter and initiates an alarm on high stress.

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(***** END OF CATEGORY B *****)

l

'C. PLANT AND RAD MONITORING SYSTEMS Page 26 l

QUESTION. 001 (1.00)

Which ONE of the following describes the required storage location for the Reactor Tank fuel Handling Tool and the Heavy Water Element Tool?

a. The Reactor Tank Fuel Handling Tool is stored in a locked cabinet in the ,

Radiochem Lab and the Heavy Water Element Tool is stored on the west wall of the reactor room floor,

b. Both the Reactor Tank Fuel Handling Tool and the Heavy Water Element Tool are stored on the West Wall of the reactor room floor,

c. The Reactor Tank fuel Handling Tool is stored on the west wall of the reactor room floor and the Heavy Water Element Tool is stored in a locked cabinet in the Radiochem Lab.

d. Both the Reactor Iank fuel Handling Tool and the Heavy Water Element Tool are stored in a locked cabinet in the Radiochem Lab.

' QUESTION: 002 (1.00)

Which ONE of the following describes the design of the dummy fuel elements?

a. Aluminum tubes with inserts that can be used for the insertion of irradiation capsules.

b. Aluminum tubes packed with powdered graphite or heavy water over the core length,

c. Radiation resistant plastic tubes with inserts that can be used for the insertion of irradiation capsules,

d. Radiatior, resistant plastic tubes packed with powdered graphite or heavy water over the core length.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

_ . _ .__ ~ _ _ _ . _ . _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . .

.Lu PL ANT AND RAD MONITORING SYSTEMS Page 27 QU t.STION: 003 (1.00)

Which ONE of the following describes the operation of the ventilation system in the event of an accidental release of particulate or gaseous activity?

a. The inlet isolation damper closes and the reactor room filter damper opens forcing air through the filter and maintaining a negative pressure in the I reactor room. :

b. The inlet and outlet dampers close and the dampers for the reactor room and hood filter open to route the building air through the filtration system,

c. The inlet and outlet dampers close and the air is bypassed thrcugh a .. EPA filtration system maintaining the reactor room at a negative pressure.

d. The outlet damper closes forcing air through the HEPA filter and maintaining the reactor room at a slight positive pressure in relation to the laboratory.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

. . , , .- -. . ~ . - - - - . - - .- ., -

C. PLANT AND RAD MONITORING _ SYSTEMS Page 28 QUESTION: 004 (2.00) from the items listed in column B, SELECT those which are major operating considerations for defining the reactor design basis, and list them in the spaces listed a through d in column A. (Items in column B may be used once or not at all; and only one answer may occupy a space in column A (Answers in column A may be in any order)

(4 required at 0.5 each)

COLUMN A COLUMN B (HAJOR CONSIDERATIONS) (OPERATING CONSIDERATIONS / CONDITIONS

a. 1. Reactor Water Temperature

b. . 2. Retention of Water in the Reactor Tank C.

3. Temperature of the Heavy

d. Water Loaded Reflector Elements

4. Reactivity of Experiments

5. Fuel Temperature

6. Argon-41 Generation

7. Reactivity of Control Rods

8. Reactor Power

9. Pool Water Conductivity

10. Fuel Clad Temperature

11. Pool Water pH

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. PLANT AND RAD MONITORING SYSTEMS Page 29 QUESTION: 005 (1.00)

With the reactor at 100 Kw and thc- heat exchanger operating, which ONE of the following is the rate of rise of the pool temperature?

a. l'F/hr

b. 2'F/hr

c. 5'f/hr i

d. 7'F/hr QUESTION: dO6 (1.00)

Which ONE of the following is the ph level to be maintained in the pool water?

a. 5.5 to 6.0

b. 6.0 to 6.5

c. 6.5 to 7.0

d. 7.0 to 7.5 QUESTION: 007 (1.00)

Two alarms will provide. sight and sound indication for high reactor radiation conditions at th) Campus Security Headquarters.

Which ONE of the following specifies the activating devices for the two alarms?

a. The Reactor Tank monitor and the Neutron Generator Monitor

b. The Reactor Tank Monitor and the Reactor Pool Low Water Level Monitor

c. The Monitor immediately over the core and the Neutron Generator Monitor

d. The Monitor immediately over the core and the Reactor Pool .cd Water Level Monitor

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

C. PLANT AND RAD MONITORING SYSTEMS Page 30 QUESTION: 000 (2.00) for the items labeled a through d on figure 3 (attached) and listed in column A, select the correct component from the items in column B. (Items in column B may be used once, more than once, or not at all. Only one answer may occupy a space in column A.)

(4 required at 0.50 each)

COLUMN A COLUMN B (FIGURE LABEL) (COMPONENTS) l a, 1. Speed increasing Gear

b. 2. Motor and Reduction gear

c. ' 3. Limit Switches

d. 4. Drum

5. Helipot

6. Torque-converter Coupling

7. Brake

8. Drum i

L 1.

C.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

.C. PLANT AND.RA U iDNITORING SYSTEMS Page 31 QUESTION: 009 (2.00) ,

From the list of nuclear instrumentation channels in column A, SELECT the TECHNICAL Si'ECIFICATION required trip setting frcm column B. (Items in column B may be used once, more than once or not at all. Only one answer may occupy a space in column A)

(4 required at 0.50 each)

COLUMN A COLUMN B

a. Startup Channel Rod Withdrawal 1. 150% of full power Block ,

2. 120% of full power

b. Linear Power Level Channel Scraml

3. 115% of full power

c. Percent Power Level Channel Scram

4. 50 counts per second d Reactor Tank Water Level Scram

5. 2 counts per second

6. 2 feet above top of active fuel

7. 2 feet below top of the tank

8. I foot below normal operating level

9. I foot above top of active fuel QUESTION: 010_ (1.00)

Which ONE of the following describes the method of supplying power to the radiation monitors and the facility intrusion detectors?

a. Directly from the !?0 VAC facility service transformer only

b. 24 VOC from 120 VAC rectifiers normally and a 24 VDC battery on loss of AC power

c. 12 VDC from 120 VAC rectifiers normally and a 12 VDC battery on loss of AC

-power

d. Directly from a facility 12 VDC battery only

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

t. PLANT AND RAD HONITORING SYSTEMS Page 32 QUESTION: 011 (1.00)

Which ONE of the following has been designated as the maximum hypothetical accident at the UUTR facility?

a. Total loss of coolant from the reactor tank

b. Rapid insertion of reactivity into the core

c. Breaching the cladding of a single fuel element during fuel handling

d. Accidental misplacement of a single experiment with maximum allowed reactivity worth QUESTION:.012 (1.00)

A small pinhole leak has developed in a fuel element in the core. '

,Which ONE of the following would be the first detectable indication that the failure had occurred?

a. The Constant Air Monitor (CAM) would show an increasing treno over a period of days

b. The monthly gamma spectroscopy of the CAM glass and charcoal filters reveals a concentration of Kr-85m

c. The detection of a 151 kev gamma on the germanium crystal attached to the peristaltic pump

d. The monthly gama spectroscopy of the tank water reveals a concentration of long lived soluble products

(***** CATEGORY C CONTINUED ON NEXT PAGE *****) ;

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C. PLANT AND RAD M98110RIRG_ SYSTEMS Page 33 QUESTION: 013 (1.00)

Which ONE of the following is the type of detector used for the startup '

channel?

a. Compensated ion chamber

b. Uncompensated ion chamber

c. Bf3 d, fission chamber QUEST!0N:,014 (2.00)

From the items labeled a through d on Figure #1 (attached) and listed in column A, select the component identification from column B. (!tems in column B may be used once, more than once or not at all, and only one answer may occupy a space in column A).

(4 required at 0.50 each)

COLUMN A COLUMN B (FIGURE LABEL) (COMPONENT)

a. 1. Reactor Room Air input

b. 2. Roof Blower

c. 3. Hood Exhaust Air

d. 4. IlEPA filters

5. Rabbit and Neutron Generator Exhaust

6. Emergency Exhaust Damper

7. Air input Emergency Damper

8. Air Monitor

9. Air Monitor Rt.ay

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

. .. . .w

j C. PL ANV AND RATLMQNITORING SYSTEMS Page 34 QUESTION: 015 (1.00) l Which ONE of the following assures the integrity of the Reactor Tank to prevent inadvertent leakage?

~

a. The interior and exterior surf aces of the tank are coated with asphalt and polyethylene sheeting

b. The liner is fabricated of aluminum overlaid with stainless steel to prevent deterioration due to corrosion

c. The tank is of the double containment design with a slight nitrogen pressure between the liners to prevent leakaga

d. The liner-is of the double containment design with sand between tha liners to make the system more resistant to earthquakes QUESTION: 016 (1.00)

Which ONE of the following describes the effect of the reflector used in the University of Utah TRIGA reactor compared to a mostly graphite reflector?

a. =K is reduced resulting in the installation of additional fuel elements to achieve a proper operational loading

, b. $? is increased resulting in the installation of fewer fuel elements m :hieve a proper operational loading a' is reduced resulting in the installation of fewer fuel elements to c.iieve proper operational loading

d. =K is increased resulting in the installation of additional fuel elements to achieve a proper operational loading l

l

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

i'

.C. PLANT AND RADjiQNITORING SYSTEMS Pege 35 4

QUESTION: 017 (2.00)

From the nuclear instrumentation items labeled a through d on Figure #2 .

(attached) and listed in column A, SELECT the component identification from ;

column B. (Items in column B may be used once, more than once or not at all, and only one answer may occupy a space in column A).

(4 required at 0.50 each)

COLUMN A COLUMN B i

(FIGURELABEL) (COMPONENT)

a. 1. Power level amplifier

b. , 2. Pre-amplifier

c. 3. Magnet Scram Control

d. 4. Log n Amplifier

5. Compensated Jon Chamber

6. Fission Counter

7. Log Power Switch

8. Linear Power Switch l 9. Linear Count Rate Counter QUEST 10N:'018 '(1.00) i A cataclysmic event has occurred that resulted in the complete opening of both

, containment tanks.

! Which ONE of the following is the approximate time required for the water to drain to the bottom of the core. (a lo:s of 22 feet of water)?

l a. 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months I

b. 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months

c. 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months s-

d. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months L (***** CATEGORY C CONTINUED ON NEXT PAGE *****)

--,,,-r..-,, - , - - - - , , - c.-, .- ,- - , . - . ~ , - - - , - - - - - , , - r---,,-,.v- ~ - - - - - - - - -

'C. PLANT AND RAD M1NITORING SYSTEMS Page 36 QUESTION: 019 (1.00)

Which ONE of the following is the power range of the Log n instrument channel?

a. I watt to 5 x EE 3 watts

b. 5 x EE-4 watts to 5 x EE 5 watts

c. 5 x EE 3 watts to 5 x EE 5 watts

d. I watt to 5 x EE 5 watts .

(***** END OF CATEGORY- C *****)

(********** END OF EXAMINATION **********)

'A . RX THEORY. TH ELQ & FAC OP CHARS Page 37 l

l ANSWER: 001 (1.00)

C.

REFERENCE:

Fundamentals of Nuclear Engineering PP 62-64 Glasstone & Sesonske, pg 239, Sec 5.28 ANSWER: 002 (1.00).

c.

REFERENCE $ .

UUTR SER pg 4-2 and Table 4.1 ANSWER: 003 (1.00) d.

REFERENCE:

Nuclear Reactor Theory, Lamarsh, pg 301, Sec 9-4 ANSWER: 004 (1.00) b.

REFERENCE:

UUTR SAR Fig. 5.1.1, Estimated Reactivity loss vs Power Curve ANSWER: 005 (1.00) a.

l

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

l l

'A. RX THEORY. THERdp & FAC OP CHARS Page 38

REFERENCE:

NUCLEAR POWER PLANT OPERATORS COURSE, fundamentals of Nuclear Reactor Engineering Section 3, Reactor Period and Reactivity ANSWER: 006 (1.00) b.

I

REFERENCE:

l Equation Sheet, Subcritical Count Rate Nuclear Reactor Engineering, Glasstone & Sesonske, pg 322 ANSWER: 007 (1.00) c.

REFERENCE:

, Equation Sheet Nuc. car Reactor Engineering, Glasstone & Sesonske, pg 236, Sec 5.21 ANSWER: 008 (1.00).

d.

REFERENCE:

UUTR SER pg 4-8 y -ANSWER: 009 (1.00) i.

b.

l -.

i l

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

A. RX THEORY. THERMQ & FAC OP CHARS Page 39

?EFERENCE:

Introduction to Nuclear Reactor Theory, Lamarsh, pg 285, Sec 7-4 ANSWER: 010 (1.00) d.

REFERENCE:

Nuclear Reactor Engineering, Glasstone & Sesonske, pg 323, Sec S.261 ANSWER: '011 (1.00) b.

REFERENCE:

Equation Sheet C-E Reactor Theory pgs. 74-76 General Physics Nuclear Theory 11 B.1 EQB Question 4616 1987/08/22 ANSWER: 012 (1.00) b.

REFERENCE:

General Physics Academic Program for Nuclear Power Plant Personnel EQB Question 4619 1987/08/22 Nuclear Reactor Engineering, Glasstone & Sesonske, pg 93. Sec 2.169, Table 2.10; pg 237, Sec S.23, Table 5.2; pg 465, Sec 8.34 ANSWER: 013 (1.00) .

C..

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

b - . . - . . _ - . . - . . -

'A. RX THEORY. THERM 0 & FAC OP CHARS Page 40 -

REFERENCE:

UUTR SAR pg 5-6, 5-7, and fig. 5.1 ANSWER: 014 (1.00)

C.

REFERENCE:

4 Nuclear Reactor Engineering, Glassione & Sesonske, pg 242, Sec 5.36 ANSWER: 015 (1.00) .

d. !

REFERENCE:

Nuclear Reactor Engineering, Glasstone & Sesonske, pg 99 - 102.

ANSWER: 016 (1.00) b.

REFERENCE:

Introduction the Reactor. Theory, Lamarsh, pg 470 - 473 ANSWER: 017 (1.00) b.

REFERENCE:

Introduction the Reactor Theory, Lamarsh, pg 474 - 476

(***** CATEGORY A CONTINUED ON NEXT PAGE *****)

'A. RX THEORY. THERMO & FAC OP CHARS Page 41 ANSWER: 018 -(1.00) 5.

REFERENCE:

Nuclear Engineering, Glasstone & Sesonske, pg 257 - 260 UUTR SAR, pg 5-7 ANSWER: 019 (1.00) d.

REFERENCE:

Technical Specificat ons, 3.2 Basis ANSWER: 020 (1.00) b.

REFERENCE:

UUTR SAR, pg 5-29 and 5-30 UUTR SER, pg 4-6, Sec 4.1.1 ANSWER: 021 (1.00) a.

REFERENCE:

VVTR SER, pg 4-8, Sec 4.5 Nuclear Reactor Engineering Glasstone & 3esonske, pg 284, Sec 5.155

(***** END OF CATEGORY A *****) ,

.E,, ,,.,.m ..,.m.,,,y.. . - - - . , ., ,_ ,, o _ . . , . . . _,..<#_, ., .,. . . . ,, , , _ . . . . . . - , . . . . - - .

'B. NORMAL /EMERG PROCEDURES & RAD CQ!j Page 42 ANSWER: 001 (1.00) c

REFERENCE:

Technical Specificat.cn 1.1 ANSWER: 002 (1.00) b

REFERENCE:

Technical Specification 2.1 ANSWER: 003 (1.00)

C

REFERENCE:

Technical Specification 6.2 SAR 7 6.a ANSWER: 004 (1.00) a

REFERENCE:

SAR 8.4.l~

ANSWER: 005 (1.00).

l b

(***** CATEGORY B CONTINUED ON NEXT PAGE *****)

B. NORMAL /EMERG PROCEquBES & RAD CQN Page 43

REFERENCE:

Radiation Safety Manual Section II.c Table I ANSWER: 006 (2.00)

a. 4

b. 7

c. 6

d. 2 (4 required at 0.50 each)

REFERENCE:

Technical Specifications 2.1, 3.2 and 3.3.3 ANSWER: 007 ( 1. 0's) d

REFERENCE:

Technical Specification 3.5 ANSWER: 008 (1.00) d

REFERENCE:

10CFR20.101 ANSWER: 009 (1.00) d

(***** CATEGORY B CONTINVED ON NEXT PAGE *****)

- _ = . -

B. NORMAL /EMERG PROCEDURES & RAD CON Page 44 l

REFERENCE:

Standard Operating Procedure Vil, Health Physics Procedures, section G.2 9 ANSWER: 010 (1.00) c

REFERENCE:

Equation Sheet ANSWER: 011 (1.00) b

REFERENCE:

Equation Sheet ANSWER: 012 (1.00) d

REFERENCE:

SAR Section 9.3.5 ANSWER: 013 (1.00) d.

REFERENCE:

'SAR-Appendix V Radiation Safety Manual Section III

(***** CATEGORY B CONTINVED ON NEXT PAGE *****)

B..... NORMAL /EMERG PROCEDVRES & RAD CON Page 45 4

ANSWER: 014 (1.00) b.

REFERENCE:

Standard Operating Procedures Section C.a ANSWER: 015 (1.00) d.

REFERENCEi Standard Operating Procedures Section A.3, Access While the Reactor is Shutdown ANSWER: 016 (1.00) b.

REFERENCE:

Standard Operating Procedures Section B, Operations Records ANSWER: 017 (1.00) a.

REFERENCE:

SER 4.1.2, Control Rods ANSWER: 018 (1.00) d.

(***** CATEGORY B CONTINVED < w PAGE *****)

B. NORMAL /EMERG PROCEDURES & RAD CON Page 46

REFERENCE:

Standard Operating Procedures C.2 Prestart Checklist ANSWER: 019 (1,00) c.

REFERENCE:

SER 7.1.3, Control Rod Circuit L

(***** END OF CATEGORY B *****)

i.

=

F

9;.3 ,

C. PM LJND RAD MONITORING SYSTEMS Page 47 I

ANSWER: 001 (1.00) h r .

S' .,

,, . s;;9 rating Procedures F.2, P.aattor Tank Tools c

ANSA '. 902 (1.00) b.

h

. itEFERENCE:

.u SAR Secticn 5.1.2, Reflector '

e .

ANSWER: 003 (' 30) c

REFERENCE:

SAR 4.6.2, Ventilatior. Emergency Operation ANSWER: 004 (2.00)

a. 2 ,

b 5

c. 8

d. 10 (! requred at 0.5 each)

REFERENCE:

ANSWER: 005 (1.00) c.

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

.,7 C. PLANT AND RAD MONITORING SYSTEMS Page 48

REFERENCE:

SAR 5.3.0, Cooling System ANSWER:- 006 (1.00)

C.-

REFERENCE:

-SAR 5.3.9, Perification System ANSWER: 007 (1.00) d.

REFERENCE:

SAR 4.7.2, Area Monitors and 5.3.10, Radioactive Monitoring and Disposal ANSWER: 008 (2.00) a.LI-

b. 5 C. 7-

d. 6 (4 required at 0.50 each)

REFERENCE:

SER Figure 7.2 ANSWER: 009 (2.00)

a. 5

b. 2 C. 2

d. 8 (4 required at 0.50 each)

_(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

'C.. Pl, ANT AND RAD _ MONITORING SYSTEMS- Page 49 i

REFERENCE:

Technical Specification 3.3.3 ANSWER: 010 (1.00) d.

REFERENCE:

SER 8.2 Emergency Power ANSilER: 0111 (1.00)

C.

REFERENCE:

SER 14.1, Fuel Handling Accident ANSWER: 012 (1.00) d

REFERENCE:

' Standard Operating Procedures, Reactor Calibration, Surveillance and Maintenance Section III.G.5, Leaking Fuel Element L

ANSWER: 013 (1.00) d y:i

(***** CATEGORY C CONTINUED ON NEXT PAGE *****)

F: -

-C. PLANT AND RAD MONITORING SYSTEMS Page 50

REFERENCE:

Safety Evaluation. Report Section 7.3.1 4

ANSWER: 014 (2.00)

a. 8

b. 7 '

1

c. 4

d. 6 (4 required at 0.50 each)

REFERENCE:

l SAR 4.6 Figure 4.6.2

ANSWER: 015 (1.00) d

REFERENCE:

SAR Section 5..'.2.d ANSWER: 015 (1.00) a.

REFERENCE:

SAR 5.3.5, Reflector ANSWER: 017 (2.00)

a. 8

b. I

c. 6

d. 4 (4 required at 0.50 each)

'***** CATEGORY C CONTINUED ON NEXT PAGE *****)

-G. PLANT AND RAD MONITORING SYSTEtil Page 51

REFERENCE:

-SAR Section'5.3.7 Figure 5.3.7.5 ANSWER: 018 (1.00)-

d

REFERENCE:

SAR Section 8.6.1 ANSWER: 019 (1.00) d.

(***** END OF CATEGORY C *****)

(********** END OF EXAMINATION **********)

" A. -RX THEORY. THERM 0 & FAC OP CHARS Page ~l ANSWER KEY HULTIPLE CHOICE 001 c 002 c 003 -d 004 b 005 a -

006- b 007 c 008 d 009 b 010 d 011 b 012 b 013 c 014 c 015 d 016- b 017 b 018 a 019 d 020 b 021 a 1

l

(***** END OF CATEGORY A *****)

I i

B. NORMAL E MERG PROCEDURLS & RAD CON Page - 2 ANSWER. KEY MULTIPLE CHOICE 018 d 001 c- 019 c 002 b 003 c 004 a 005 b 006 MATCHING a- 4 b 7 c 6 d -2 MULTIPLE CH0 ICE 001 d 008 d 009- d.

010 c 011 b 012- d 013 -d 014 b

-015 d

~

016 b 017: a-1

(***** END OF CATEGORY B *****)

.- C. PLANT AND RAD MONITORING SYSTEMS Page 3 ANSWER KEY MULTIPLE CHOICE MULTIPLE CH0 ICE 001 a 010 d 002 b 011 c 003 c 012 d 004 MATCHING 013 d a 2 014 MATCHING b 5 a t c 8 b 7 d I c 4 HULTIPLE CH0 ICE d 6 005 c MULTIPLE CH0 ICE J06 c 015 d 007 d 016 a 008 MATCHING 017 MATCHING a 1 a 8 b 5 b 1 c 7 c 6 d- 6 d 4 009 MATCHING MULTIPLE CH0 ICE a 5 018 d b- 2 019 d c 2 d 8

(*****.END OF CATEGORY C *****)

(********** END OF EXAMINATION **********)

(---

jo TEST CROSS REFERENCE- Page _1 Io-;

S R 0- Exam 7.? ? Reactor 0raanize-d by- 0uestion Number OVES 110N YALE REFERENCE 001 1.00 9000293 002 1.00 9000294 003 1.00 9000295 004 1.00 9000296 005 1.00 9000297 006 1.00 9000298 007 1.00 9000299 008 1.00 9000300 .

-009 1.00 9000301 010 1.00 9000302

. 011 1.00 9000303 ,

012 1.00- 9000304 013 1.00 9000305 014 1.00 9000306 +

015 1.00 9000307 016 1.00 9000308 017 -1.00 9000309 018 1.00 9000310 019 1.00 9000?ll 020 1.00 9000312 021 1.00 9000313 21.00 001 1.00 9000314 002 1,00 9000315 4 003 1.00 9000316 004 1.00 9000317 005 1.00 9000318 1 006- 2.00 9000319 007 1.00 9000320 008 1.00 9000321 009 1.00 9000322 010 1.00 9000323 011 1.00 9000324 012 1.00 9000325 013 1.00 9000326 014 1.00 9000327-015 1.00 9000328 016 1.00 9000329 017 1.00 9000330 018 1.00 9000331 019 1.00 9000332

, 20.00 001 1.00 9000333 002 1.00 9000334 003 1.00 9000335

e P

TEST CROSS REFERENCE Page 2

> l SR0 -Exam ??? Reactor Oraanized by 0uestion Number i 2V151108 YALRE BffERENCE l 004 2.00- 9000336 005 1.00 9000337 006 1.00 9000338 007 1.00 9000339 008 2.00 9000340 009 2.00 9000341 010 1.00 9000342 011 1,00 9000343 012 1.00 9000344 013 1.00 9000345 014 2.00 9000346 015 1.00 9000347 016 1.00 9000348 017 2.00 9000349 018 1.00 9000350 4 019 1.00 9000351 24.00-

~

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1 i

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Text

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