ML20138B322
| ML20138B322 | |
| Person / Time | |
|---|---|
| Site: | University of Maryland |
| Issue date: | 12/02/1985 |
| From: | Strosnider J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20138B310 | List: |
| References | |
| 50-166-OL-85-04, 50-166-OL-85-4, NUDOCS 8512120267 | |
| Download: ML20138B322 (41) | |
Text
{{#Wiki_filter:. . _ _ ._- _. . . . _ _ . _ _ _ _. -_ .. . . _ - _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ . _ _ . . o s i - U.S. NUCLEAR REGULATORY COMMISSION REGION I OPERATOR LICENSING EXAMINATION REPORT EXAMINATION REPORT NO. 50-166/85-04 i FACILITY DOCKET NO. 50-166 i
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FACILITY LICENSE N0. R-70 I LICENSEE: University of Maryland [ College Park, Maryland 20742 ' FACILITY: . Maryland University TRIGA Reactor J I EXAMINATION DATES: October 22-23, 1985 , i l PREPARED BY: \h/ w Reactor Engineer (E hM 1 i
#M/2rC
'ner) / Da'te REVIEWED BY: _)
Chie',' Projects' Section IC
,j // [7-9[D , Date APPROVED BY: 2c h/ MS f'W /2 [85' ief, Projects Branch No. 1 Date
SUMMARY
- Two instant SRO candidates were examined. Both candidates passed i
the oral examination while one failed Section J of the written i examination. I e '] - 8512120267 851204 PDR ADOCK 05000166 O PDR
l l , I i REPORT DETAILS TYPE OF EXAMS: Replacement X EXAM RESULTS: l Pass / Fail l I Written Exam i 1/1 l l Oral Exam l 2/0 l i Overall l 1/1 1
- 1. CHIEF EXAMINER AT SITE: John D. Smith, PNL
- 2. Summary of generic strengths or deficiencies noted from grading of written exams Candidates displayed the following weaknesses:
Response to a radioactive spill Knowledge of component response to a ventilation isolation System response to instrument failures
- 3. Personnel Present at Exit Interview:
NRC Contractor Personnel venn D. Smith, PNL Facility Personnel Dr. Ralph Belcher, Reactor Director Dr. Frank Munno, Director
3
- 4. Summary of Comments made at exit interview:
Candidates clearly passed the oral examination. No generic weaknesses were found. Areas in which the candidates exhibited good training and knowledge were facility and control room familiarization and console operation.
- 5. Attachment 2 contains all facility comments with regards to the written examinatice but only the following changes were made to the written exam from the examination review:
Question No. Change Reason H.07 Delete fifth and sixth Modify answer key, response on answer key and add " Convection of cooling water removes heat in heat exchanger." H.08 The word " blade" should " Rod" is proper be changed to " rod." term. I.05 Answer should be Math error.
.656 mR/hr instead of 6.56 mR/hr.
I.06 Delete third response Modify answer key. in answer and re-distribute points. J.02 Delete third response Modify answer key answer and redistribute points. J.06 Will accept 33% for Expand answer key, answer to b). and d). J.07 Change answer b). to Expand answer key.
" Reactor scram caused by Taking Safety Switch off operate."
K.02 Change " element" to " Cluster" is
" Cluster." proper term.
l t
4 Question No. Change Reason K.08 Additional possible Expand answer key, answer to a).: " Gamma ray analysis of the pool water." Attachments:
- 1. Written Examination and Answer Key (SRO)
- 2. Facility Comments on Written Examination I
. . - . . ~_ .... --. _-. . - _ - _ ..- - _.-
[ M5 6 /f Li {. U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAHINATION
; FACILITY: UNIVERSITY OF MARYLAND
[ REACTOR TYPE' TRIGA DATE ADMINISTERED: BS/10/24 , EXAMINER: DUDLEY APPLICANT: _________________________ INSTRUCTIONS TO APPLICANT: Use separate paper for the answers. Write answers on one side only.
- Staple question sheet on top of the answer sheets. Points for each 1
question are indicated in parentheses after the question. The passing
- grade requires at least 70% in each category and a final grade of at ;
l least 80%. Examination papers wil." Je picked up six (6) hours after [ the examination starts.
% OF ; ; CATEGORY % OF APPLICANT'S CATEGURY l VALUE TOTAL SCORE VALUE CATEGORY 20.00 20.00 4
________ ______ ___________ ________ H. REACTOR THEORY 20.00 20.00 ________ ______ ___________ ________ I. RADI0 ACTIVE MATERIALS HANDLING 4 DISPOSAL AND HAZARDS 20.00 20*00 ________ ___ __ ___________ ________ J. SPECIFIC OPERATING CHARACTERISTICS ! 20.00 20.00
, ________ ______ ___________ ________ K. FUEL HANDLING AND CORE l . PARAMETERS
- 20.00 20.00
________ ______ ___________ ________ L. ADMINISTRATIVE PROCEDURES, - CONDITIONS AND LIMITATIONS ! 100.00 100.00 TOTALS. ' t FINAL GRADE _________________% All work done on-this examination is my own. I have neither 4 i Sivon nor received aid.
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H. REACTOR THEORY PAGE 2 QUESTION H.01 (1.50) , . What affect, .if any, will starting the primary coolant pump have i if the reactor is operating at 250 KW7 Explain. QUESTION 'H.02 (1.50) I Why might it be necessary to make fine adjustments to the regulating rod followins a decrease in power from 250 MW to 10 KW? QUESTION H.03 (2.00) ! 4
- a. Under steady-state operation, would there be any significant j difference between the control rod positions at 100 watts and 200' watts? Explain.
- b. Under steady-state operation, would there be any significant
, difference between the control rod positions at 100 KW and 200 KW? Explain. ] , 1 ! QUESTION H.04 (2.00) i
! During a reactor startup after the reactor is critical what affect 1 will removal of the startup neutron source have on ,
- a. Detector indication
- b. Core reactivity OUESTION H.05 (2.00)
List two characteristics of Zr Hx fuel which accounts for the large prompt negative temperature coefficient, and explain how each of them insert reactivity. p 4 n
- .(***** CATEGORY H CONTINUED ON NEXT PAGE xxxxx) !
I, f i 1 4
i 1 1 I H. REACTOR THEORY PAGE 3 QUESTION H.06 (2.00)
- a. Assuming the reactor is at 125 watts and on a stable sustained [
, period of 50 seconds, how long will it take to reach 250 KW? Show your work.
- b. Assuming the reactor is at 125 watts and is placed on a 50 sec period, explain how the reactor will respond if no further operator actions are taken?
GUESTION H.07 (3.00) Explain the different modes of heat transfer by which the heat of fission is removed from the fuel. Include major components involved in the heat removal process starting with the fuel and ending at the ultimate heat sink. QUESTION H.08 (3.00) The reactor operator is conducting a routine reactor startup after it has been shutdown for several days. Prior to withdrawing a shim rol b!:d: he reads a stable count of 50 cp's on the startup channel. ' Immediately after withdrawing this ble"; he reads a coent of 80 cps. ul c.J
- a. If he performed no b+ede motion for five minutes, would the count rate increase, decrease or remain the same? Explain, assumin3 the reactor is suberitical at 80 eps.
- b. After 5 minutes he withdraws another bl ?]_: the same distance -
j but the reactor is still suberitical. Would the change in count i rate (time and magnitude) be different than he saw in part (a) l above? Explain,
- c. What indications would the operator observe to determine when the reactor had gone critical?
OUESTION H.09 (3.00) e How much recctivity has been added to a suberitical reactor if the count rate has increased from 100 cps to 150 cps and if the initial value of Keff was .95? (***** END OF CATEGORY H xxxxx) T E
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1 . 3 I. RADI0 ACTIVE MATERIALS HANDLING DISPOSAL AND HAZARDS PAGE 4 a
- QUESTION I.01 (2.00) l Does the number of disintegrations per minute (dpm) from a radioactive J Source equal the counts per minute (cpm) obtained from a survey instrv-cent? Briefly explain. -
OUESTION I.02 (2.00) j a. What is the greatest potential radiological problem that
, may result from an extended shutdown of the ventilation
? system during power operations?
, b. What means or methods are used to control personnel , exposure from radioactive materials that are produced ;
l in the reactor? QUESTION I.03 (2.00)
- a. Why must a Quality Factor be used when calculating exposures from
, measurements taken in RADS? c ! b. What steps should be taken to avoid contamination while swipins i a potentially contaminated surface? ! QUESTION I.04 (2.00) What is the half life of the isotope contained in a sample which produces the following count rates? Show all calculations. Time, Minutes Counts per Minute f initial count 900 30 740 60 615 90 512 i 180 294 i QUESTION I.05 (2.00) l If the source for calibration of the radiation area monitors was left unshielded on a table, what would be the expected J dose rate two feet from the source? Portions of SP 205, Area Radiation Monitor Calibration, are provided. j (*xmum CATEGORY I CONTINUED ON NEXT PAGE *****) f i f i i d l l _ . . , _ . . . _ . . . ..~ - ,,,_,,_.- .._ . , _ . . , , _ . . . . . . . _ , . , _ , _ . _ _ _ _ . - - . _ _ . _ , _ _ , - - _ _ . _ _ _ _ , _ . ,
. . ~. .- . - . - . . .. - .. . _ _ . _ _ . - - - . _-
F ' SP 205 ' Revision 1 l Theory and Source Information
~
1.0 The following pertains to the calibration of the radiation area monitors in the reactor building.
. The Tracerlab Serial #598 Co-60 source is used. It I
was rated at 4.7 millicuries on Dec. 8, 1960. The half life of Co-60 is taken as 5.27 years. - Ref er to the attached table for source strength at current calibration date. l From page 545 of Glasstone, " Principles of Nuclear Reactor Engineering":
; Dose rate at distance 'f' feet 6CE
! for a 'C' curie source equals--- i f2 R/hr. E = 2.5 Mev for Co-60 A i a Y i { . i J
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Io RADIOACTIVE MATERIALS HANDLING DISPOSAL AND HAZARDS PAGE 5 GUESTION I.06 (3.00) What actions should be taken if a flask of irradiated liquid is dropped on the floor and breaks? GUESTION I.07 (3.00) A fuel element is suspended in the Reactor Pool approximately 1 meter under water. A radiation survey meter held at the surface _ of the water reads 100 mrem /hr.
- a. Ignoring buildup, what radiation level would you expect if the fuel element broke the water? Assume an attenuation coefficient of 0.035 cm^-1. (1.0)
- b. If the radioactive isotopes in the fuel element had an average half life of 30 minutes, how long would it take for the radiation level at the surface of a one inch lead shield cask to drop to 20 mrem /hr?
Assume an initial contact dose of 2 R/hr for the fuel element and a tenth thickness of two inches for lead. (2.0) GUESTION I.08 (4.00) A 23 year old individual has accumulated a lifetime occupational dose of 24 rem of whole body exposure documented in accordance with 10CFR20 and has received no exposure during the present calender quarter.
- a. How long may he work in a 3 mrem /hr area if he works an 8 hour day Monday through Friday? Show your work.
- b. An individual in a restricted area may be allowed to receive a whole body dose in excess of the quarterly limit under certain conditions.
Name three conditions. (***** END OF CATEGORY I ***x*)
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=n, J. SPECIFIC' OPERATING CHARACTERISTICS .
PAGE 6
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GUESTION J.0,1 (1.50) , Why might switching the Reactor Powe'r Range Switch to the ' calibrate position while in the automatic mode result in regulating rod withdrawal? QUESTION J.02 ( 2 . 50' * -' What components should respond to'a manual ahtuation from one l of the four emergency v.Itilation thutdown locations. Include how each component should respond.' OUESTION J.03 (4,00) What actions, if anyrfshould be taken in each of the following situations. Assume thy reactor is initially at 200 KW. Analyse each situation individually.
- a. A sample in the^p"neumatic tube has a reactivity worth such that its removal could possibly icram the reactor on either safety channel.
- b. The Exhaust Area RadiatiorbMonitor causes a scram during removal of a sample 9 sing thy pneumatic pystem. ,
- c. A 0.'5 spm leak from t'he sealson the primary coolant pump is discovered.
~
- d. Power decreases during a rod witlidrawal.
QUESTION J.04 (3.00)
- What action, if any, shoulf'd a Seniot_ Operator take if an experimenter declares a buI1(iri ,eseFoencyo 3 dve to a perceived radiological problem? ,-
'd, -
a a (***** CATEGORY J CONTINUED ON NEXT PAGE mummm) kg h. d q# r :% 9m e- .4 eg k i ( .s
+ b v
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J. SPECIFIC OPERATING CHARACTERISTICS PAGE 7 QUESTION J.05 (3.00)
- a. Explain how each of the factors in the following formula is
- determined. This is the formula used as part of the procedure j to measure the thermal output of the core.
, P=bm e (dT/dt) (2.0)
- b. How many times must a value for P be calculated? Explain. (1.0) i i
QUESTION J.06 (3.00) s What level should the following meters / controls be reading / set if reactor power is 100 KW? Attachment 1 to OP -104 is provided. O a. Safety I channel meter i
- b. CIC (red pen) chart recorder
, c. Wide Range channel meter
- d. Demand controller
, e. Reactor power range switch - QUESTION J.07 (3.00) .
I What responser if any, should be expected for each of the following i instrumentation failures if the reactor is a 250 KW and the "
, regulating rod is in automatic? Explain your answers assumins no operator actions.
- a. The fuel temperature monitor leads are shortet oSether.
- b. A calibration signal is sent to the pre-amp of the fission
- chamber power ranse detector circuit for safe channel 1.
) c. The linear power channel compensated ion chamber loses all compensation.
- d. High voltage is lost on the uncompensated ion chamber for
. Safety Channel 2.
- l I
i i ! (***** END OF CATEGORY J *****) i a k 1 J i e
s 5 l OP- 104 Enclosure 1 Revision'- , t -
- l. s .
o . x . 1 k
% Pbwer o ~ #- - - so w ro n ~ roo ,,, ,,t u lill lili 1111 till till till 1111 11)) [Ill llll lll[ llll 1tll 1ll1 jjjl ,
O ZS 60 75 i toc., its eso 87 5 2co 125 1so 3co Power' (kw) a .i 1
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O o/o Pour - (O IO~I 10 10 10 10' IO" 10' ' l 80 goo l l ! ! l I I j ' l l J.5 I 25 Zso, g.s 25 ' 25o, ,z.s ts tso, i mW. , W kW Power f 1, . t i N 4 T 3 t i. i r 7 , . yr +--w $r-w,-g y-q v omywwww----gvg-one vWww err-e w ww p*wmtwrw w-w
-K. FUEL HANDLING AND CORE PARAMETERS PAGE 8 GUESTION K.01 ( .67)
Do the thermocouples in the instrumented fuel rod measure temperature on the (Choose the most correct answer.) A. Surface of the fuel rod claddins. B. Outer surface of the fuel. C. Interior of the fuel. D. Center of the circonium rod. QUESTION K.02 , ( .67) How many fuci ele r.t; must be removed from the core prior to removing a control rod? (Choose the most correct answer.) A. O B. 2 C. 3 D. 4 QUESTION K.03 ( .66) Why are the MUTR fuel rods slightly smaller in diameter than standard TEIGA fuel rods? (Choose the most correct answer.) A. To reduce excess reactivity on initial core loadins. B. To fit the fuel rods into the modified fuel end plates. C. To maintain the metal to water ratio. D. To allew control rods to fit correctly into the fuel rod positions. QUESTION K.04 (1.50)
- a. What is the reason for the requirement that the fuel rod straightness test be conducted under water?
- b. What direction should be given a fuel handler if he reports that a fuel element does not pass the straightness test?
GUESTION K.05 (2.00) How many people are required to move fuel and where must these people be locatod? (***** CATEGORY K CONTINUED ON NEXT PAGE *xxxx) l i l
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1
K. FUEL HANDLING AND CORE PARAMETERS PAGE 9 GUESTION K.06 (2.50) What requirements must be met to have the reactor secureo when 4 a standard core is installed? QUESTION K.07 (3.00) Estimate the number of fuel elements needed to so critical using the 2 followin3 data taken during a core loading. Graph paper is provided. Fuel (CR with ALL Rods Out) Elements Detector A Detector B 0 270 303 2 290 400 4 323 526
- 6 385 800
-
- 8 472 1250 QUESTION K.08 (3.00)
- a. Give the two most likely means of detecting a leaking fuel element in the core.
- b. What are two possible hazards involved with operation at power with a leaking fuel element?
] GUESTION K.09 (3.00)
- a. What TWO instruments should the control room operator observe
, during movement of fuel? (1.5)
- b. How does an operator assure that a fuel handling tool is latched onto the top handle of the fuel bundle prior to moving the fuel? (0.75)
- c. Where should the fuel bundle. movement be recorded? (0.75)
QUESTION K.10 (3.00)
- a. Why is a neutron reflector important to reactor operations?
- b. List.THREE things that act as radial neutron reflectors for the MUTR core.
I (***** END OF CATEGORY K *****) l l-I i
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7 4 4 .i l j' L. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 10 l
- QUESTION L.01 (1.50)
- If a senior reactor operator is on call how far from the control j room is that person permitted to be?
GUESTION L.02 (1.00)
, When may an intermediate checkout be conducted instead of-an initial startup checkout prior to conducting a reactor startup?
QUESTION L.03 (2.25) Whose approval is required before the following experiments may be performed? j a. Routine Experiments , b. Modified Experiments j c. Special Experiments QUESTION L.04 (2.00) Under what circumstances, if any, may the reactor be operated assumins' 4
- a. The Exhaust Radiation Monitor is inoperable.
b.' The Exhaust Radiation Monitor and the Bay Radiation Monitor are BOTH inoperable. ! QUESTION L.05 (2.25) What actions must be taken if fuel temperature increases above 4 1000 C ? I 3 (xxxxx CATEGORY L CONTINUED ON NEXT PAGE *****) f w. L l l l
. - _ . . . = _ , . . _ ._ _ . . ~ . _ _ . . . _ . . _ . - _ . _ _ _ _ _ _ _ . _ . . _ . _ . ,
- - = =. .. .. . . _. --_
d i L. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 11
- QUESTION L.06 (3.00)
For each of the following situations indicate whether the presence , of an SRO IS or IS NOT required.
- a. Steady state operations at 250 KW with rod control in manual.
- b. Experiments conducted at 100 KW with rod control in automatic where an experiment with a reactivity worth of $0.10 is being inserted and withdrawn from the core.
- c. An experiment with estimated reactivity worth of $0.90 is being reconfigured in the core while the reactor is shutdown.
- d. Maintenance is being conducted on control rod drive motors and the rods are disconnected from the drives.
I
- e. A reactor startup following an interruption of electrical power.
GUESTION L.07 (3.50) What are four responsibilities of a Senior Reactor Operator during_ a declared emergency as detailed in the Emergency Preparedness Plan? GUESTION L.08 -(3.00)* Classify each of the following events using the portions of the Emergency Preparedness Plan provided. y
- a. There is a peaceful demonstration against nuclear weapons scheduled outside the-facility.
- b. The reactor is shutdown and all water is lost from the pool,
- c. An experimenter trips while transporting an activated sample he has just taken from the reactor and lacerates his leg with broken glass from the sample flask.
- d. Durin3 fuel handling an irradiated element drops to the bottom of the pool and ruptures. The released fission products alarm the Bay Radiation Monitor.
4 (xxxxx CATEGORY L CONTINUED ON NEXT PAGE xxxxx)
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, n TABLE 5.1
^
Emergency Classification Guide l
. l l EMERGENCY CLASS ACTION LEVEL 1
i
- Personnel injury Personnel Emergency l
Unusual Event (Class 1) e Receipt of bomb l threat
- Report or observa-tion of severe natural phenomenon.
- Substained fire or minor explosion within the reactor building.
- = Any event that causes or likely to cause radiation levels as indicated on any radiation area monitais above 100 arem/hr.
Alert (Class 2)
- Any event that '
causes radiation
.:' levels as indicated on any radiation area monitors above 500 ares /br.
l 5-2 i I
. )
Lo ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 12 GUESTION L.09 (1.50) What is the basis for limiting the reactivity worth of a single unsecured experiment to one dollar? (***** END OF CATEGORY L *****) (************* END OF EXAMINATION ***************)
i l
)
l
. 1 I l H. REACTOR THEORY PAGE 13 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER H.01 (1.50)
Reactor power will increase E0.63 due to the addition of cold water and MTC feedback. [0.93 + REFERENCE l OP 104, p 1 ANSWER H.02 (1.50)
.To maintain the reactor critical at a steady state level 00.83 due to the delayed neutron effect. [0.73 REFERENCE OP 104, p3
- ANSWER H.03 (2.00) i
- a. No difference CO.43 because the reactor is below heat ranse and the power coefficient does not add any negative reactivity. CO.63 4
~
- b. Rods would be higher at 200 KW E0.43 to compensate for the negative reactivity added by the power coefficient. CO.63 REFERENCE 170: Reactor Theory II B.4
, ANSWER H.04 (2.00)
- a. Slight reduction in count rate CO.73 due to removal of neutron source which was positioned near detector. CO.83
- b. Reduce core reactivity 00.73 due to removal of a source of neutrons. [0.83 REFERENCE OP 103 1
1 1
e H. REACTOR THEORY PAGE 14 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER H.05 (2.00) ZrHx shifts the neutron spectrum at elevated temperature CO.43 which increases the leakage of slow neutrons from the fuel bearing region. [0.63 Doppler broadening CO.43 which causes more absorption in the resonance absorption region at elevated temperatures. CO.63 REFERENCE ' SER, p 4-4 ANSWER H.06 (2.00)
- a. P = Po e(t/T) [0.43 250 KW = 0.125 e(t/50) CO.43 Int 20003 =1n[e (t/50)3 7.6 = t/50 t= 300 see or 6.33 min. [0.23
- b. Power coefficient would lengthen period and would eventually stabalize power. C1.03 REFERENCE 170: Reference Package, Reactor parameters Nuclear Theory II B.2 ANSWER H.07 (3.00)
Conduction through fuel. (Radiation across fuel sap) (ea ) Conduction transfer from fuel to coolant. (e () Forced convection to heat exchanger. (C d i Conduction across heat exchanger. A t) 1 Isrced ccavection to cssling towers.
, Ev;F;;;tipn tu etes;Fhere. '
0,5 r-= - J'l-Convaf:e n c1 ce l .g u.,te - rv.%.e e<1 hafach'$v^(u.e) REFERENCE
, Introduction to Nuclear Engineering,-chapter 8; J R Lamarsh ! RSM pgs. 3.1, 3.2, 3.10 to 3.12 t
i
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H. REACTOR THEORY PAGE 15 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DVDLEY ANSWER H.08 (3.00)
- a. Increase slightly then level out(0.6) due to soberitical multiplication (0.4).
- b. Larger increase (0.3) and longer to level out(0.3) due to greater number of generations to reach equilibrium (0.4).
- c. Steadily increasing count rate or slight positive period with no rod withdrawal. (1.0)
REFERENCE OP 103, p2 ANSWER H.09 (3.00) cri /cr2 = (1-Keff2) / (1-Keff1) [0.93 100/150 = (1-Keff2) / (1-0.95) [0.53 1-Keff2 = 10/15 x 0.05 Keff2 = 0.967 CO.1] Change in reactivity = C1-Keff2/Keff23 - [1-Keffi/Keff13
= Keff2 - Keffi / Keffi x Keff2 E0.93
= 0.967 - 0.95 / 0.95 x 0.967 CO.5]
= 1.05 % delta K/K CO.1]
REFERENCE Procedure Manual (PM) 2.3 ps, 1,2 i
~. - -- - - . - - . _ - - _ _ _ - . - . - - - - - . ..-.. .-- . . . - _ - . . . - - - .
I 1
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i I. RADI0 ACTIVE MATERIALS HANDLING DISPOSAL AND HAZARDS PAGE 16 ! ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY f I i ANSWER I.01 (2.00) No. The epm must be corrected for efficiency of the detector and
- the geometry of the source in relation to the detector. (2.0) i '
REFERENCE RSM pgs 5.2, 7.1 i ANSWER I.02 (2.00) i- a. Buildup of Ar 41. [1.03 I b. Use of rigidly developed and reviewed procedures which use time, distance and shielding considerations. C1.03 ; REFERENCE SER, p 12-2 l ANSWER I.03 (2.00) s I a. RADS measure amount of energy deposited in material while exposure rates indicate amount of biological damage. CO.63 Different . l types of radiation produce greater biological damage which must { be corrected for by Duality Factors. CO.43
- b. Handle swipes with gloves [0.53 and wear appropriate anti-contamination i jl clothing. [0 53 4
i REFERENCE SER, p 11-2 ANSWER I.04 (2.00) A = Ao e{- lamda X t} E0.6] 294 = 900 e{-lamda X 180} CO.33 ' 180 landa = 0.327 j landa = 0.00623 min {-1} C0.13 1 t 1/2 = 0.693/landa CO.63 F
= 0.693/ 0 00623 [0.3] , = 111 minutes E0.13 1
I I l l i 4 I i,
-'m-+ -w-m.-p..-e w r , *w-*-+-ww-N v m 6'- e rm - - . , - - w**redM"'*~**wf-'UmW*'r'T*vv'Tw'T""***-rree go v we w my'.vg a.ary-,
I. RADI0 ACTIVE MATERIALS HANDLING DISPOSAL AND HAZARDS PAGE 17 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY REFERENCE Nuclear Physics, Reactor Theory and Core Operating Characteristic, p 20-21 ANSWER I.05 (2.00) C= 1 '5 :i J.75 x e; A C, CO.63 D = {6 X 2.5 Mev / (2 ft) ^ 2} XC E0.83
= 1.5 X 2.5 X 1.75 E0.33
= -d . 51 c r c:/" o.6c.d /g [0.33 REFERENCE SP 205 ANSWER I.06 (3.00)
Notify all persons to evacuate the room at once . /c.1) Make-no immediate attempt to clean up the spill. (0.7' F1; ;h ; k. i n , di; ;;a of clothin3 Switch off all fans and air conditoner s. (c.7) Vacate the room and prohibit entrance to contaminated area.(cl/ [0.' each] REFERENCE EP 404, p2 ANSWER I.07 (3.00)
- a. I = Io e{-ux}
100 mrem /hr = Io e{-0.035 cm(-1) 100 cm} Io = 3311 mrem E1.03
- b. I = Io 10 {-x/TVL} (TVL is tenth thickness)
Io = I 10 <x/TVL} = 63.25 mrem /hr [1.03 Io = Ii e{-(.693/ half life)t> (where Ii is initial contact dose) t = -(half life /.693) In(Io/Ii) t - -(30 min /.693) In(63.25/2000) = 149.5 minutes [1.03 REFERENCE Introduction to Nuclear Engineerins, pss 22, 23; J R Lamarsh a-
- . . _ . . . . . . - . - . - . ~_ . - . . ..-- -. _. - _ .. . - . . . . . . . . -
l . i l I. RADI0 ACTIVE MATERIALS HANDLING DISPOSAL AND HAZARDS PAGE 18 ' i ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER I.08 (4.00) ,
- a. 5(N-18) = 5(23-18) = 25 1
25 - 24 = 1.0 Rem = Max. Dose (1.0) ; l Max. Dose = Dose Rate X Time i 1.00 Rem = 0.003 Rem /hr X 0 hr/ day X No. of Days j No. of Days = 41.6 days (1.0)
- b. Provided that (1) He does not exceed 3 rem per quarter (.66) 2 i
(2) His radiatirn history is known and recorded on j the proper form (NRC Form 4) (.67) (3) The dose received when added to his radiation i history does not exceed 5(N-18) rems where y N= the person's age at his last birthday (.67) REFERENCE 10 CFR 20.101 i 't l I i i 1 l I I i
)
i l 4 1 1 i
+
J. SPECIFIC OPERATING CHARACTERISTICS PAGE 19 ANSWERS -- UNIVERSITY OF HARYLAND -85/10/24-DUDLEY ANSWER J.01 (1.50) A calibration signal to the comparitor maybe below the demand
; set point causing rod withdrawal in an attempt to raise reactor power. E1.53 REFERENCE OP 104, p2 SER, p 7-4 ANSWER J.02 (2.50)
Two intake louvers shut. [0.53 (u t ) Two roof-mounted fans stop. 20.5: ICL) Tw; ;ir ;;nditi;n; ; on th; _;;t w;11 er; ; ; ; ;_ . - d . [0.53 Two louvers from the west laboratory shut. [0.53 /e t 1 One air conditioner from west laboratory secures. 00.52 (L() REFERENCE SER, p 6-1 ANSWER J.03 (4.00)
- a. Scram reactor. [0.53 Remove sample from the core manually. [0.53
- b. Insert sample back into the core CO.63 and allow for decay. [0.43
! c. Scram reactor. [0.73 Repair leak. [0.33
- d. Secure reactor. E1.03 i
REFERENCE OP 105 EP 402, p2 EP 403 t a t
9 i d. SPECIFIC OPERATING CHARACTERISTICS PAGE 20 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DVDLEY l ANSWER- J.04 (3.00) Assure building is evacuated. Ascertain that emergency is real. Obtain portable monitoring equipment. Contact University H.P. and Director of Facility.
.Take remedial actions.
CO.6 each3 REFERENCE E0P 404, p5 1 ANSWER. J.05 (3.00)
- e. b is a conversion constant CO.23
- n. is the mass if water and is determined by measuring pool level CO.53 i
e is a constant for the neat capacity of aster CO.23 i T is the temperature as measured by an installed termometer CO.53 t is time over which the measurement is taken CO.33 . P is power in KW CO.33 i i b. Twice. [0.53 Once for the heatup rate and once for the cooldown i rate. CO.53 REFERENCE SP 202 4 ANSWER J.06 (3.00)
- a. 40%
b . 91HF 314 4 c. 75% f O . "b@M T3R
- f. 30 (X10) kw REFERENCE t OP 104, p5 4
i S
J. SPECIFIC OPERATING CHARACTERISTICS PAGE 21 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY
- ANSWER J.07 (3.00)
- a. No affect. CO.353 Thermocouple output should fail low. [0.43
- b. ":;d ;; tion :^.05: C;1157-t;;n -ign;1 _ill rep 1;;; d;t;-ter signal
.., __s ..:,, ..._.: 1 : __ - - _c --- __ _ __ _ __,: e- 1 :__
~
NiNn;5. 5 - N IN N N r575 h N r. l l N y S vs U [ c f W' ?
- c. No affect. CO.353 Compensation does not effect power level indication at this power level. [0.43
- d. Reactor scram. CO.353 Loss of HV provides a scram signal to the scram logic circuitry. CO.43 REFERENCE SER, p 4-7, 7-3 i
' L k
- t
K. FUEL HANDLING AND CORE PARAMETERS PAGE 22 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER K.01 ( .67) C. Interior of the fuel. REFERENCE SER, p 4-2 ANSWER K.02 ( .67) D. 4 REFERENCE Maintenance Procedure 304 ANSWER K.03 ( .66) C. To maintaire metal to water ratio. REFERENCE SER, p 4-1 ANSWER K.04 (1.50)
- c. Limits irradiation exposure. CO.753
- b. Take fuel element out of service by placing it in storase. [0.753 REFERENCE SER, p 9-2 ANSWER K.05 (2.00)
SRO CO.37] on reactor bridge CO.33 RO or SRO CO.373 in contro room [0.33 fuel handler CO.363 on reactor brid3e [0.33 REFERENCE Maintenance Pequest 303
t K. FUEL HANDLING AND CORE PARAMETERS PAGE 23 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DVDLEY ANSWER K.06 (2.50) Console is secured. CO.93 No work is in progress involving core fuel, core structure, or installed control rods. [0.83 No experiments in or near the reactor with worth greater than $1.00. [0.83 REFERENCE TS, p3 ANSWER K.07 (3.00) REFERENCE CE-Nuclear Physics, Reactor Theory and Core Operating Characteristics, p 140 ANSWER K.08 (3.00)
- a. Visual inspection CO.753 Radiation Monitor Alarms CO.753 (m.14 .- ( ,,,,,,,,, e y 7o,/ g j FAe (c,.l y,[ , }
- b. Radiatrion levels CO.753 Contamination [0.753 REFERENCE SP 201 ANSWER K.09 (3 00)
- a. Reactor Brid 3 e Area Radiation Monitor. CO.753 CIC power tracing. CO.753
- b. Shake the fuel handling tool. [0.753
- c. Control Room Los Book. C0.753 REFERENCE Maintenance Procedure 303, p3 s y se- ,,w- --,m wm- 8wwr*,-~=a-+---w
- ,w w,- -r-y-v,w--,mn-+mn,.r--+,,---e,g gm,_mn--e._p
. l I
K. FUEL HANDLING AND CORE PARAMETERS PAGE 24 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER K.10 (3.00)
- c. Reflectors reduce the amount of fissionable material needed to maintain critica111ty by reducing the thermal leakage from the core. (1.5)
- b. Graphite reflector Thermal colven Pool water CO.5 each]
REFERENCE SER, p 4-1
L. ADMINISTRATIVE FROCEDU3ES, CONCITIONS AND LIMITATIONS PAGE 25 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER L.01 (1.50) On the College Park campus [0.53 or within 10 miles from the facility CO.53 and can reach the facility within one half hour. [0.53 REFERENCE T.S., p2 ANSWER L.02 (1.00) For reactor runs conductcd after the initial reactor run of the same day. [1.03 REFERENCE OP 101, p1 OP 103, p1 ANSWER L.03 (2.25)
- a. Duty Senior Reactor Operator [0.753
- b. Facility Director CO.43 or designated alternate CO.353
- c. RSC [0.43 and Facility Director or designated alternate. CO.353 REFERENCE OP 105 ANSWER L.04 (2.00)
- e. May continue to operate with one of two rad monitors inoperable. C1.03
- b. May continue to operate. [0.43
, Reelace monitors with portable samma sensative intrument. CO.33 I Complete repairs in 48 hours. [0.33 REFERENCE TS, p 12
L. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 26 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY ANSWER L.05 (2.25) Shutdown reactor (no restart until authorized by NRC)CO.753 Report to Chairman and Reactor Safety Committee. CO.753 Report to NRC within 24 hours. [0.753 REFERENCE TS, p 26 ANSWER L.06 (3.00)
- a. IS NOT
- b. IS NOT
- c. IS
- d. IS NOT
- e. IS NOT CO.6 each]
REFERENCE TS, p 20 - ANSWER L.07 (3.50) Placing facility in a safe shutdown condition. Terminating or mininiizing release of radiactive material. Protecting the facility personnel and visitors. Assessing severity of the event. Notifying the Reactor Director. Cany 4 0 0.82 each3 REFERENCE
- Emergency Preparedness Plan, p 3-3 ANSWER L.08 (3.00)
- a. No classification
- b. Alert
- c. Personnel Emergency
- d. Unususal Event
[0.75 each] 1 I i I
L. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 27 ANSWERS -- UNIVERSITY OF MARYLAND -85/10/24-DUDLEY REFERENCE Emergency Preparedness Plan, Sec. 4 ANSWER L.09 (1.50) To prevent the safety limit CO.5] from being exceeded if the positive worth E0.5] of the experiment was suddenly inserted. [0.53 REFERENCE T.S., p 13 l I
4 . TEST CROSS REFERENCE PAGE 1 OUESTION VALUE REFERENCE
! H.01 1.50 DUD 0001019 H.02 1.50 DUD 0001020 4 H.03 2.00 0U00001021 H.04 2.00 DUD 0001022 i H.05 2.00 DVD0001023 H.06 2.00 DUD 0001024 i
H.07 3.00 0000001025 N.08 3.00 DUD 0001026 j H.09 3.00 0UD0001027 20.00 I.01 2.00 DVD0001028 l I.02 2.00 DUD 0001029
- I.03 2.00 DUD 0001030 I.04 2.00 DUD 0001031 d' I.05 2.00 DU00001032 i I.06 3.00 00D0001033 l I.07 3.00 DUD 0001034 j I.08 4.00 DVD0001035 i 20.00 l J.01 1.50 0UD0001036 J.02 2.50 0000001037 l J.03 4.00 DUD 0001038 l j J.04 3.00 0000001039 !
i J.05 3.00 0U00001040 ! J.06 3.00 DU00001041 [ i J.07 3.00 DU00001042 ' l 20.00 i K.01 .67 0000001051 i 4 K.02 .67 00D0001052 ' K.03 .66 DVD0001053 K.04 1.50 DVD0001054 K.05 2.00 0000001056 i
, K.06 2.50 DVD0001057
{ K.07 3.00 DUD 0001058 { K.08 3.00 0000001059 [ K.09 3.00 0000001060 , j K.10 3.00 0U00001061 j ______ i j 20.00 L.01 1 50 0U00001043
-L.02 1.00 0000001044 i L.03 2.25 0U00001045 '
! L.04 2.00 0U00001046 1 i i
4 TEST CROSS REFERENCE PAGE 2 OUESTION VALUE REFERENCE L.05 2.25 DUD 0001047 L.06 3.00 DUD 0001048 L.07 3.50 DUD 0001049 L.08 3.00 DVD0001050 L.09 1.50 0U00001062 20.00 100.00
ATTACHMENT 2 s COMMENTS ON EXAMINATION dUESTIONS , H REACTORY TilEORY ! 11 . 0 5 In reference to doppler Broadening. The, presence of the Zrlix is not the reason for doppler broadening. This is a temperature effect dealing with the proporties of the fuel (that is the U-238) itself. I feel this part of the examiners response is either incorrect or incomplete. 11.07 There are no cooling towers at the University of l Maryland Facility. This is an incorrect response. I II.08 0.K. Comment: The word " Blade" should be changed to " rod". I RADIOACTIVE MATERIAL IIANDLING DISPOSAL AND IIAZARDS I.05 Answer given 6.56 mrm/hr is incorrect - correct answer is 0.656 rm/hr. I.06 Why flush skin and dispose of clothing? There is no assumption of splattering onto a person. I dis-agree with this part of response to question as asked. J SPECIFIC OPERATING CilARACTERISTICS J.02 Comment: Answer (3) not portinent. None of the west balcony air conditioners are tied into the emergency ventilation shut down switch. (2) Louveres located on the inner wall of the west balcony dn ! close, however. J.03c Could get by on our reactor simply hy turning off primary pump and isolating the pump section. If level has decreased then turn on makeup water and
~ ,
o scram only if low-low point is reached. J.05b This question is ambiguous. I can not determine i intent of examiner. Power is constant at some level . above zero duritig heat-up and zero during cool-down. Cool-down used to determine b, not P. . J.06 Note that b and e are dependent. Could be 100% on 100 KW settihg. Alno responue to d (demand controller) can lead to much discussion. Are we in auto or not.
! No indication in question. Question should be thrown 1 )
out, i J.07b In order to send a calibration original to the pre amp of the fission chamber, I need to take the safety 1 switch off operate. This will result in a " manual .i l scram". Examiners response makes sense only if he i is referring to the CIC or linear channel. i c I would have to " guess" different'Jy. Loss of comp. voltage will result in slightly higher indicated power. Rod will drive down to compensato. I have l 1 1
- no fooling for how significant so that the apparent recult might be "nu effect".
Parts (a) and (d) 0.K. K FUEL !!ANDLING AND COR3 Pt.Pp1ETERS *
'K.02 The word "elempnt" should be changed to " clusters".
In this core the correct answer is in clusters, j K.04 We are not required by our Tech. Specs, to perform this operation. Thorofore, question should be thrown f
1
. i 4
out. K.07 Answer not given on your answer sheet. The most conservative answer'given by the two sets of data should be used. K.08 One other answer should be considered. That is, i by gamma ray analysis of the pool water. L ADMINISTRATIVE PROCEDURES CONDITIONS AND LIMITATIONS No comments: This section O.K. as is. 4 i 4 i a d I
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