IR 05000020/1988001
| ML20148B850 | |
| Person / Time | |
|---|---|
| Site: | MIT Nuclear Research Reactor |
| Issue date: | 03/10/1988 |
| From: | Eselgroth P, Norris B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20148B824 | List: |
| References | |
| 50-020-88-01OL, 50-20-88-1OL, NUDOCS 8803220165 | |
| Download: ML20148B850 (109) | |
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U.S. NUCLEAR REGULATORY' COMMISSION REGION.I- - OPERATOR LICENSING EXAMINATION REPORT _ EXAMINATION REPORT N /88-01(0L) FACILITY DOCKET N FACILITY LICENSE N0.: R-37
' LICENSEE: Massachusetts Institute of Technology 138 Albany Street Cambridge, Massachusetts 02139 FACILITY: MIT Research Reactor EXAMINATION DATriS: --January.25 26, 1988
' CHIEF EXAMINER: 4/[m 3W/5+ f B'arr . Norri Date Senior perations Enginee APPROVED BY: -
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3 Yd ' Peter V. E _ groth, Chief Date PWR Secti , Division of Reactor Safety _ SUMMARY: Written and operating examinations were administered to two Senior Reactor Operator (SRO) candidates and one Reactor Operator (RO) candidat All candidates passed the examinations and received their license PDR ADOCK 05000020 V DCD ,,
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DETAILS-
- TYPE ~0F. EXAMINATIONS: Replacement-EXAMINATION RESULTS:
r l R0 l SR0 l-l . Pass / Fail l Pass / Fail l l l l 1 - 1 l l' l Written l 1/0 1 2/0 l l l ~l
'l i I I L l Operating l 1/0 l 2/0 l l l I I l I 1 I
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CHIEF EXAMINER'AT SITE: B. S. Norris (USNRC) 0THER EXAMINERS: M. O. Bishop (EG&G) W. S. Rosener (EG&G)
. - .The following is a generic deficiency noted during the operating . . examinations. This information is being provided to aid the licensee in upgrading license _and requalification training program No licensee - response is require A lack of in-depth knowledge of the operation of the automatic containment isolation valve was noted. In general, candidates were not able to clearly describe the operation of the valve beyond the fact that a major scram would cause it to clos . Personnel Present at Exit Meeting:
l NRC Personnel L B. S. Norris - Chief Examiner l NRC Contractor Personnel ' M. O. Bishop - Examiner, EG&G iv. S. Rosener - Examiner, EG&G , Facility Personnel J. A. Bernard - Superintendent, MITR L. Clark - Director, Reactor Operations 0. V. Harling - Director, NRL K. Kwok - Assistant Superintendent, MITR
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13.' -Summary of Comments Made at Exit Meeting: It was noted by the NRC that the lead seal on the Emergency Decontamination locker. in 'the control room was not properly attache The facility stated that the contents would be reinventoried immediately and that the locker would be sealed, The NRC noted that there is no accountability system in place to control the issuance of keys from the key locker in the control' roo The facility stated that-they did not think that co'ntrol of the keys was-required for safe-operation, but that they would reevaluate the situatio . The written examination questions and answers were reviewed by the facility after all candidates had completed the examination. The primary reviewer.was K. Kwo Attachments: R0 Written Examination and Answer Key SRO Written Examination and Answer Key Facility Comments and NRC Resolution on R0 Written Examination Facility Comments and NRC Resolution on SR0 Written Examination
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U. S. NUCLE.\R REGULATORY C0MMI55;UN _ hEACTOR VEEhATOR LICEN5E EXAMINATION
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FACILITY: MIT l
REACTOh TYPE: EsEEARCH DATE ADMIN 3TERED: 43/01/25 EXAMINER: BISHOF. M 4__ CANDIDATE l' l5IFUCTIGE5 TO CA'd DIDAIEi Use separate paper for the answer Write answers on one side onl Staple questi.'n sheet on top Of the answer shu+t Foints for each question are indicated in parentneses arter tue questio The passing 4 grade requires at least 707 in eacn category. Examination papers will oc picked up six (6) hours after the examination start :4 OF ' CATEGORY ?; OF CANDIDATE'S CATEGORY
'JA LUE TOTAL 3 CORE VALUE CATEGORY 15 _._' 5 0 __15.50 FRINCIFLEE OF REACTOR OPERATION 12.50 12.50 FEATURES OF FACILITY DESIGN 14.50 14.50 GENERAL OPERATING CHARACTERISTICS 15.00 15.00 INSTRUMENT 3 AND CONTROLS 14.00 14.00 SAFETY AND EMERGENCY SYSTEMS 14.50 14.50 STANDARD AND EMERGENCY OPERATING FROCEDURES 14.00 14.00 RADIATION CONTROL AND SAFETY 10 % Totals je -
All work done on this examination is my ow I have neither given ner received ai Candidate's 5ignature
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NRC KULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the, administration of'this examination the following rules apply: l '. Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . Restroom' trips are to be limited and only one candidate at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date on the cover sheet of the examination (if necessary). Use only the paper provided for answer . Print your name in the upper right-hand corner of the first page of each section of the answer shee . Consecutively number each answer sheet, write "End of Category __' as , appropriate, start each category on a new page, write only on one side of the paper, and write "Last Page" on the last answer shee . Number each answer as to category and number, for example, 1.4, 6.3.
' 10. Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a guide for the depth of answer require , 14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be give Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN r 16. If narts of the examination are not clear as to intent, ask questions of the examiner only.
i' 17. You must sign the statement on the cover sheet that indicates that the work is your own and you have not received or been given assistance in completing the examinatio This must be done after the examination has been completed.
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16. When you complete your examination, you shall: As'semble your examination as follows: I (1) Exam questions on to (2) Exam aids - figures, tables, et (3) Answer pages including figures which are part of the answe Turn in your copy of the examination and all pages used to answer the examination questions, Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, Leave the examination area, as defined by the examine If after
, leaving, you are found in this area while the examination is still in progress, your license may be denied or revoked.
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. . PRfNCIPLE3 OF REACTOR OPERATION Page 4 . .
QUESTION A.01 (2.50) If the reactor is on a stable 25-second period HOW long will it take to change the power level by 0 decades? SHOW all wor . QUESTION A.02 (2.00) Briefly describe the t.wo phenomena that contribute to the moderator temperature coefficient of reactivity for MITR-II.
, QUESTION A.03 (2.00) The MITR-II reactor produces a relatively fast response for a given reactivity !.nput. Explain this response in terms of neutron generaticn time and delayed neutron fraction (BETA) at MITR-II.
QUESTION A.04 (2.00) INDICATE whether each of the following statements are TRUE or FALS An increasing concentration of Xe-135 reduces the thermal utilication factor, f, and the multiplication factor, Keff, of the reactor core, Xe-135 is produced both directly as a fission product and as the result of a decay chain from other fission product r , A good approximation for determining the production of Xe-135 is to
. assume that the Xe-135 is produced from the decay of Cs-13 The removal rate of Xe-135 is due to the neutron absorption rate in Xe-135 atoms and the radioactive decay of Xe-135 atom (***** CATEGORY A CONTINUED ON NEXT PAGE *****)
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PRINCIPLE 5 OFJEACTOR OPERATION Page 5
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QUESTION A.05 (3.00) HOW much reactivity has been added to a suberitical reactor if the count rate has increased from 100 ops to 150 cps and if the initial value of Kof:
-was 0.957 SHOW ALL WORK and express your answer in percent delta K/ QUESTION A.06 (2.00)
When calculating an estimated critical position for reactor startup, the operator uses the previous week's position and corrects for five possible
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different delta K change LIST four of the possible delta K change QUESTION A.07 (2.00) Refer to Figure 1, Regulating Rod - Control Blade Assembly, in back of test. Identify the components labeled A though D on the figur . J'
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B[ FEATURE 3 0F_ FACILITY DESIGN Page 6
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QUESTION B.01 (3.00) Refer to Figure 2, Reactor Core Tank Support, in back of tes Identify each component labeled A through F on .he figur QUESTION B.02 (2.00)
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Briefly explain HOW the valve design and location allows the anti-syphon valves to prevent syphoning water from the primary tan l QUESTION B.03 (2.00) l l What are four methods of increasing the cooling tower water outlet l temperature during reactor operations ? QUESTION B.04 (2.00) Answer the following in regard to shield cooling: What would be the physical consequence or overheating the shield? I How many shield cooling regions is the shield divided into for cooling purposes? What two interlocks protect the shield by not allowing reactor operation with the shield cooling system shutdown? (Setpoints not required).
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QUESTION B.05 (2.00) What are the four design functions of the D20 Cleanup System while it is i its normal configuration?
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. QUESTION B.06 (1.50)
Briefly explain how you would determine which tape was causing the alarm ir' the event of an alarm on the Leak Alarm Consol ,
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C .' GENERAL OPERATING CHARACTERISTICS Pego 6
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I QUESTION C.01 (1.50) If you increase reactor power from 100 KW to 4.9 MW, WHEN would you run a heat balance to confirm that the reactor was at 4.9 MW, Briefly explain your answe QUESTION C.02 (2.50) Briefly explain WHAT would be the effect on the cooling tower water
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system if the reactor were operated for thirty days at 4.9 MW with no blowdown from the cooling tower? Include WHY this effect occurs (1.5) What prevents the cooling tower basin from overflowing into the yard if the makeup valve sticks open? (0,5) (TRUE or FALSE) Low secondary flow will cause a reactor scram if the reactor is operating and the low flow setpoint is reache (0.5) x QUESTION C.03 (3.00) Briefly Explain WHY the reactivity effect of dumping the radial heavy water reflector varies with the position of the shim blade (1,0)
, How does dumping the heavy water reflector effect reactivity with the shim bank at Top of Active Core as compared to dumping it with the shi bank Full-Inserted? What position must the shim bank be in prior to pumping up the radial heavy water reflector following a dump of the refle6 tor? Briefly explain WHY this position is required prior to the pump u i (***** CATEGORY C CONTINUED ON NEXT PAGE *****)
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_ - _ . _ _ . -$ 1 ; ~ GENEFAL 0FERATING CHARACTERISTICS Pago 0
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QUESTION C.04 (1.50) Answer each of the following TRUE or FALS It is possible and permissible to operate the reactor with no forced flow in the primary coolant system, With the reactor at full power the pneumatic tube temperature will increase to 100 degrees C in approximately five minutes if cooling air is los Total thermal power output of the reactor.is the sum of Primary fewer,. Reflector Power, Shield Power, and Cooling Tower Los QUESTION C.05 (2.00) During reactor operation, primary system ph MUST be maintained between-a- and -b- , however, the DESIRED ph is-c- to __ -d- . QUESTION C.06 (2.00) List the normal flow for the following system with the reactor at full powe Heavy Water System Primary Coolant Secondary Coolant Shield Coolant l' QUESTION C.07 (2.00) Answer the following in regard to performing a reactor startu As shim blades are raised, WHAT TWO indications will you observe to verify the reactor is approaching criticality and HOW will the parameters react? l l
(***** END OF CATEGORY C *****)
. 'D." _INSTh0MENTS _ AND' CONTROI;S Pcgo 10 . .
GUESTION D.01 (2.00; , What are two reasons WHY Ion Chambers do not need to be compensated when used for Reactor Power Indication at full power? QUESTION D.02 (2.00) Refer to Figure 3 Period Channels, at back of tes Identify each of the components marked "A" through "H".
QUE5 TION L.03 (2.00) Luring reactor operation at 100% power you attempt to initiate automatic control of the regulating rod and it will not initiat WHAT four items would you check to assure the "Automatic-Control-Permit" circuit is not preventing initiation? QUESTION D.04 (3.00) During reactor operation at near full power the "Automatic Rundown Circuit" for the control rod drive system is activated, What has caused this circuit to activate? (1.0) What four automatic actions and/or indications are initiated by this circuit? (2.0) r QUESTIQE D.05 (2.00) Identify each item marked "A through H" on Figure 4 Functional Block Diagram of Absorber Control System, at back of exa .
(***** CATEGORY D CONTINUED ON NEXT PAGE *****)
, . ID. _INSTRUjENT3 AND COMTROL5 Pass 11
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QUESTION D.06 (2.00) What are the two reasons'for performing the once-a-week comparison of the "Reactor Thermal Fower Balance" and the "Output Signal of the Neutron Level Channels"?
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QUESTION D.07 (2.00) Briefly explain WHY the reactor must be in "Neutron Kenetic Equilibrium"
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prior to performing the weekly comparison between the "Reactor Power Thermal Balance" and the "Output Signal of the Neutron Level Channels?" Include the effect on the "Neutron Level Channel" indication and WHAT causes this effect if the reactor is not in the "Neutron Kenetic Equilibrium" conditio .
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Pcg3 la E '_ SAFETY AND EMERGENCY SY3TEMi
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QUESTION E.01 (3.00) What are six loads automatically supplied EMERGENCY POWER through panel 1 if normal power falls? QUESTION E.02 (2.00) Answer each of the following TRUE or FALSE: When normal electrical power is lost and then regained 1 hour later, , the Emergency Power MG set must be manually shutdow When normal electrical power is lost, the emergency-power MG set does not start for approximately 12 second Natural circulation can provide adequate core cooling if normal and emergency power is los Primary coolant auxiliary pump, MM2 breaker cannot be shut unless normal power is availabl QUESTION E.03 (2.50) Answer the following in regard to the Containment Pressure Relief Syste a. Describe the filters in the exhaust line Include the TYPES of the filters and their positions relative to the flow strea (1.5) What is the rated system flow? If the system is lined up to relieve pressure, what determines the system flow rate? r
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E ." SAFETY AND EMERGENCY SYSTEMS .Pago 13
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QUESTION E.04 (3.00) What four automatic actions are initiated when a major scram is initiated manually from the control console? (2.0) When is the scram pushbutton on the medical therapy console operable? (0,5) Is the medical therapy console scram a "minor" or "major" scram? (0.5)
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GUE5 TION E.05 (2.00) What is the purpose of having two vacuum breakers in each line of the reactor building negative pressure protection system? What is the design internal pressure (positive and negative) of the reactor building? j GUESTION E.06 (1.50) Answer the following in regard to emergency cooling: How many modes of emergency cooling are there? What are the two basic criteria the system is designed to accomplish? What criteria must be satisfied in order to determine the system
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operable per the T.S. LC07 SETPOINTS REQUIRE . m I
(***** END OF CATEGORY E *****)
. 'FI 3TANDARD AND EMERGENCY OPERATING FROCEDURES Pega 14 . .
QUESTION F.01 (3.00) With fuel in the reactor, WHAT are six conditions / requirements that must be met for the reactor to bv in the "SECURED" conditon? QUESTION F.02 (1.50) Answer each of the following TRUE or FALCE: When the reactor is in the "Non-Operating but Attended" condition the
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control room must be continuously manned F( r4 M The console operator and dn nreactor supervisorsmust be in the c ntro' . room during a reactor startu The regulating rod is pulled to the estimated critical position prior to pulling any shim blades during a reactor startup, QUESTION F.03 (2.50) Answer the following in regard to bypassing a safety function NOT required by Technical Specifications: Can the reactor be operated with such a safety function bypassed?
(0.5) If not part of an approved procedure, WHO must authorice such a bypassi '
(Three required) (1.5) i{ hat do you do with a BYPASS LOG SHEET that is completed (filled up) or '
your shift? (0,5) r
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F .' STANDARD'AND EMERGENCY OPERATING FR0CEDUREs Faga 15
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QUESTION F.04 (2.00) A Licensed Reactor Operator has the authority to take reasonable action that departs from a license condition or techneial specificatione, What four conditions must be met prior to your taking such action? QUESTION F.05 (1.50;
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You are the operator at the control console with the reactor at full powe What are three immediate actions required if a "Withdraw Fermit Circuit Open" alarm is receivod? QUE3 TION F.06 (2.00) a. During performance or AOP 5.2.4 "Low Flow Primary Coolant," if the cause ofthe low flow is a tripped primary pump, the pump discharge valve, the inlet to HE-15, and the heat exchanger inlet valve that * corresponds to thetripped pump are shu BP.IEFLY explain WHY these actions are taken.(1,5) b. TRUE OR FALSE If a primary pump trips with the reactor at power, all shim blades and the regulating rod must be inserted prior to restarting the pump.(0.5)
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QUESTION F.07 (2.00) Dusing performance of AOF 5.8.9, Malfunction of a Shim Blade / Regulating Rod, operators are instructed to secure electrical power to the drive motor' of any stuck shim blad Briefly explain WHY this action is take i
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QUECTION G.01 (3.00) In regards to the Operational General Safety Rules, prior to entry, what
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are three joint responsibilities of ths operator-in-charge and any personnel entering either the reactor top, the medical therapy room, or the : equipment room when the reactor is operating?
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QUESTION G.02 (1.00)
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Under what conditions may come one be authorized to incur radiation exposures in excess of the 10 CFR 20 lim 4.ts? QUESTION G.03 (3.00) If the Reactor Floor Ar-41 Monitor gives an "High Level Radiation Monitor" alarm, where are four likely places the AR-41 originated.(2.0) Briefly discuss WEAT is done to help prevent the production of AR-41 at MITR-II. Include in your answer the preventive measur3s taken (2 required) AND HOW these neasurers nelp prevent production of AR-41.(1.0) QUESTION G.04 (1.00) Briefly explain why heavy water is a radiological conceYnt ' Include the isotope of concern AND the type of radiation emitte i i QUESTION G.05 (2.00) l What two types of dosimetry are all personnel working at the MITR-II reactor required to wear? (2.0) l f , ! (*4*44 CATEGORY G CONTINUED ON NEXT PAGE *****) L__
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U; ic . RADI ATION C011 TROL At3D _SAVET,_Y_ FCgo 17
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QUESTION G.06 (2.00) Answer each of the fellowing TRUE or FALSE:
. Personcel with BLUE film badge holders must be escorted at ALL times while in the Reactor Building to assure no radiation exposur Personnel with RED film badge holders must be under the direct supervision of a licensed SRO/R0 while conducting experiments involving radiatio Fersonnel with YELLOW film badge holders are all membrs of Operations or RF0 staff with current quarter exposure less than 3 Re Only personnel with YELLOW film badge holders may guide members of general public through the reactor buildin QUESTION G.07 (2.00) What radiation detection device results become the official record of your exposure? (0,5) You are assisting in a maintenance job and notice you have accumulated 60 mrem on your dosimete What two actions are you required to perform as a result of this exposure? (1.0) (TRUE or FALSE)
Frctective clothing used within the Restricted Area can NEVER be worn outside the Restri:ted Area. (0,5) r
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(********** END OF EXAMINATION **********) --
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EQUATION SHEET
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f = ma v = s/t v = ms s = v,c + lat s 2 Cycle efficiency = "* 'I E = aC a = (vg - v,)/t Et = lsuv
,, ,vg + ,t A . xy 4 . g ,-At PE = ath a = 6/t 1 = la 2/tg = 0.693/tg ,
W = v&P- (g)(q) ,,
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P=P 10 M IC)
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?=P et /T RVL = 0.693/v
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~SUR = 26.06/7 7 = 1.44 DT SCR = S/(1 - K,gg) /1 8 o SUR = 26 CR x = S/(1 - K,ggx)
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T = (1*/o ) + [(i- ' o)/1,ggo ] 1 eff 1 " eff 2
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y . g*/ (, ,, p M = 1/(1 - K,gg) = CR g/CRg I"( ~ 8)! eff # M = (1 - K,gg) /(1 . gaff)1 8*I aff'I)I aff * #eff eff !E SDM = (1 - K,gg)/K,gg a= [1*/TK,*gg .] + @/(1 + 1,ggT )] 1* = 1 x 10" seconds
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l 7 = I(v/(3 x 1010) 1,gg = 0.1 seconds A
: = ne l Idgg=1d22 VATER PARAMETERS Idg =Id2 -
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1 gal. = 8.345 lba 1/hr = (0.5 CE)/d2 4,,,,,,)
- 1 ga .78 liters R/hr = 6 ct/d2gg,,,)
1 ft = 7.48 ga MtsCEt.t.ANEOUS CONVERSIONS
Density = 62.4 lba/f t 1 Curia = 3.7 x 1010dps Density = 1 gm/cm 1*kg = 2.21 1ha Heat of vagorization = 970 teu/lba 1 hp = 2.54 x 10 3BTU /hr Heat of fusica = 144 Btu /lba 1 N = 3.41 x 100 Btu /hr 1 Atm = 14,7 psi = 29.9 in,l' Btu = 778 ft-lbf
1 f t. H 2O = 0.4333 lb f /in g inch = 2.54 cm T = 9/5"C + 32
"C = 5/9 (Or - 32) -- . - . _ _ - _ ... - - . , _ - . . . .__._- - _ - _-. - ___. -,. _ - - _ _ _ _ _ . - _ . - _ _ _ . -
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FUNCTION AL PI_OCK DIAGRAM OF ABSORBER CONTROL SYSTEM FIGURE 4
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. FRINCIFLE.3 0F 3EACTOR_ OPERATION Pcga 18 .
ANCWER A,01 (2.50) Froa equation sheet: P = ? e t/T (0.5) o t/25 sec
?/P = 100 = e (0,5)
o In 100 = t/25 see (0,5) t = (25 see) (In 100) (0.5)
: 115.13 seconds = 1.92 min 2tes (0.5)
REFERENCE MIT: Glasstone and Sesonske n1T Training Program Reference), PM 1.16.2 pp. 1 AN3WER A.02 (2.00) The first is the temperature rise of the light water in the reactor cor (0.5) Any such temperature rire will insert negative reactivity by causinc a hardening in the neutron spectru (This means that the average neutron takes longer to thermalise so there are fewer fissions. (0,5) The second phenomenon is the radiation heating of the heavy water reflector. (0.5) Temperature rises here add negative reactivity by allowing more neutron leakag (This second prccess lags the temperature rise of the light water in the core proper.) (0,5) REFERENCE MIT-RSM 1 /'
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ANCWER A.03 (2.00) The sensitive response is due to the short neutron generation time for the MITR-II, even though its delayed neutron fraction is large (beta-bar = 0.00786). (1.0) The large Beta effective is predominately due to(1.0) a large source of "slow born" photo neutrons developed in the reflecto l l
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Paga 19 A$~ PRINCIPLE 3OFREACTOROPERATION
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REFERENCE' MIT-RSM 1 ~ ANSWER A.04 -( 2. 00 ) ' True
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[t0.5] each e
REFERENCE MIT-RSM 1 ANSWER A.05 (3.00) crl/cr2 : (1-Keff2) / (1-Keff1) (0.5) 100/150 : (1-Keff2) / :1 0.95) (0.5) 1-Keff2 : (10/15 x 0.05 Keff2 = 0.967 (0.5) Change in reactivity : (Keff2-1) /Keff2 -
(Keffi-1) /Keffi (0.5) = (Keff2 - Keffi) / (Keffi x Keff2) (0.5) = (0.967 - 0.95) / (0.95 x 0.967) = 1.85% delta k/k (0.5)
REFERENCE MIT: Reactor Physics Notes (Reactor Suboritical Multiplication).a Eight pages from the front of book.(No page numbers in book)
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A[ PRINCIPLEG_OF__ REACTOR' OPERATION- Page 20
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ANSWER A 06 -(2.00) Delta K due to temperature change Delta K due t sample loading Delta K due to xenon Delta K due to fuel loading Delta K due to burnup (Any 4 at 0.5 each) -REFERENCE MIT-FM 3.1.1.2 page 13
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ANSWER A.07 (2.00) Magnet Coupling Armature Blade Bottom Permanent Magnet Guide Tube REFERENCE MIT-RSM Figure 1.12
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BI FEATURES OF FACILITY DESIGN _ Pags 21
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ANSWER B.01 (3.00) Core Shroud Core Tank Top Shield Lid Upper Shield Rin Lower Annular Ring Reflector Tank (0.5 each) REFERENCE , MIT-RSM Figure 1.11
ANSWER B.02 (2.00) y - St t * The valve is installed at the top of the core shroudA(0.5). Primary flow closes the ball valve during reactor operation (0.5). Loss of flow (0.5) allows the ball valve to open, breaking any syphon path (0.5).
REFERENCE MIT-RSM ANSWER B.03 (2.00) Bypas s booster pump Bypass Tower Operate C.T. fan at 1/2 spee /' Vary pitch of fan blade .., Restrict air admitted to tower (rearrange external boards and flaps).
(Any 4 at 0.5 each)
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- . FEATURES-OF FACILITY DESIGN page 22 . . . REFERENCE'
MIT-RSM 3.10 ANSWER B.04 (2.00) Melting the lea (0.5) Four (0.5) . Low water flow from shield (PF-1) (0.5) Low pump discharge pressure (PPS-1) (0.5)
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REFERENCE MIT-RSM 3.13 paragraph 3.5.1, Shield Coolant System ANSWER B.05 (2.00) Maintains the purit . . Maintains Reflector Tank Leve . ' Provider, a surge volumn (to compensate for heatups - cooldown). Provides dump volumn for emergency reactor shutdow (0.5 each) REFERENCE MIT-RSM Section 3.3.3, Cleanup System, page r l ANSWER.- B.06 (1.50) Depress the pushbuttons, directly above the light for the alarming channel, (0.5) one at a time until the alarm light goes ou (0.5) Determine the location by utilizing the Leak Tape Location Lis (0.5) l I i ( ** * ** C ATEGC NY B CONTINUED ON NEXT PAGE *****)
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REFERENCE . I MIT-RSM paragraph 33.6, Leak Detection System, page 3.10
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C .' GENERAL OPERATING CHARACTERISTICS Pago 24
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ANSWER C.01 (1.50.', Twenty.four hours after reaching 4.9 MW(0.5). This wait is necessary because of the large heat capacity (0,5) of the graphite reflecto (0.5) REFERENCE MIT-RSM Section 6. ' ANSWER C.02 (2.50) The dissolved solids in the makeup water (0.5) would be concentrated (0.5) due to evaporation out of the tower (0.5) An overflow stand pipe drains to the sewe (0.5) True (0.5) REFERENCE MIT-RSM paragraph 3.4.2, Main Flow System, page 3-11 ANSUER C.03 ( 3. ';0 ) The shim blades operate in the region between the core and the heavy water reflector (0.5) thereby exerting a shadowing influence on the reflecto (0.5) l A dump with the bank fully inserted is worth about two thirds .the worth of a dump with the bank at Top of Active Cor (1.0) . c.The shim bank must be Fully Inserted (0.5). This position is required to ensure the reactivity inserted during the pu'mp up does not occur when the reactor is or could go critica (0.5)
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- . GENERAL OFEFATING_ CHARACTERISTICS Page 25 .
REFERENCE' MIT-RSM-1 .
. ANSWER C.04 (1.50) True True False REFERENCE MIT-FM-2.2, FM-5.5.1, FM-2.4 page 5 ANSWER C.05 (2.00) .5 .9 .5 .5 (0.5 each)
REFERENCE MIT FM 3.1, Startup Checklist, Section 3.1.1.1, Two Loop Mechanical page 13 of 15 r . l ANSWER.- C.06 (2.00) . gpm (+ or - 10 gpm)
- gpm (+ or - 50 gpm) gpm 4_ (+ or - 50 gp 90 - 110 gpm g,,g y jg,o y 4f
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GENERAL OPERATING CHARACTERISTICS Page 26
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REFERENCE' MIT-FM-3.1.1.2_page 14 of 15
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ANSWER C.07 (2.00) . Neutron level (0.5) will be increasing (0.5). Reactor period (0.5) will exhibit shorter transient period (0.5) gg ,r ki; ti * ts * u.yr, S a ,,..;- o -P th~!r 1&+~ d sy' -,L p,;.s REFERENCE Ol/ " ' MIT-FM 2.3, Reactor 5afety Procedures, page 2 of 7 r w
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D INSTRUMENTS AND CONTROLS Page 27
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ANSWER- D.01 (2.00) Neutron flux is several orders of magnitude greater than gamma flu . Gamma flux is proportional to fission rate and reactor powe REFERENCE MTI RSM-5.4 Paragraph 5.2.3, Uncompensated Ion Chambers ANSWER D.02 (2.00) Mag. Am Scram Am Off Scale Trip Period Network Pulse to D.C. Converter Lescriminator Amplifier Fietio } Chamber Ion Chamber _ REFERENCE MIT RSM Figure 5.1, Period Channels ANSWER D.03 (2.00) ALL shim blades must be above suberitical interlock positio . The power-set / actual power deviation must not exceed 1.5%. The regulating rod control switch must be in neutral positio . The regulating rod must be withdrawn beyond its neaf-in position l (1.6 inches).
(0.5 each) l l
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EI. INSTRUMENTS AND CONTROL Fago 28-
. . REFERENCE MIT-RSM- ANSWER D.04 (3.00) The regulating rod has been at its near-in position (0.5) (1.6 inches) for thirty seconds?yrvM(0,5) B. Hr. ,:- mtw+ a %I, . Red light comes o . busser sounds ,_ Reactor control shifts to manual (after 30 seconds). Selected shim blade drives in (prior to 1.5% deviation).
(0.5 each) REFERENCE MIT-RSM-4.5 (second paragraph) ANSWER D.05 (2.00) Withdraw Permit Circuit Shim Blade Magnets Automatic Run-Down Circuit Start-Up Interlocks ' Suboritical Interlock & Override Permit Circuit Regulating Rod Control Circuit All Absorbers in Circuit Shim Blade Motors and Brakes l' REFERENCE MII RSM Figure 4.1, Functional Block Diagram of Absorber Control System l l ' f (***** CATEGORY D CONTINUED ON NEXT PAGE *****) ,
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D.I _I_NSTRUMENTS AND_ CONTROLS Pago 29
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ANSWER D.06 (2.00) It'is the basis for calibrating the neutron level channel . It is the basis for setting the safety channel level trip point (1.0 each) REFERENCE MIT RSM - Item 6.3.4 page . ANSWER D.07 (2.00) The axial neutron flux shape would vary if this condition was not present (1.0) Consequently the neutron detectors output would vary (1.0).
REFERENCE MIT RSM Item 6.3.4 page 6.5, first paragraph
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. . ~ AFETY AND EMERGENCY SYSTEMS Pago 30 . .
ANSWER E.01 (3.00) Safety Amps Recorders Rad Monitors Clock Intercom System Servo Unit Front Panel Outlets Ann. Panel Indicators-Du O N W Rod Control
'Med. Rm. Recp Pneumatic Tube Magnet Power Evac. Alarm (Any 6 @ 0.5 each)
EEFERENCE MIT RSM Table 8-8B page 8.34 ANSWER E.02 (2.00) False False True False REFERENCE MIT RSM paragraph 8.8.2 Emergency Room Dist. System page 8.37 & 8.38 ANSWER E.03 (2.50) r Three filters in series in each line, Absolute Filter - Charcoal Filter - Absolute Filter (3 Filters @ 0.25 for type & 0.25 for location or ea). cfm (0.5) Building / atmosphere delta pressure (0.5):
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s i SAFF._TY_AND__ EMERGENCY SYSTEMS Pago 31
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REFERENCE' MIT-RSM-paragraph 8.4, Containment Pressure Relief System page 8.23 and Figure ANSWER E.04 (3.00) . Ventilation System Secure . Containment Shell is Seale . D20 Reflector is Dumpe . Withdraw permit circuit deenergized (shim blades drop).
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(0.5 each) When the Medical Therapy's console key switch is ON.(0,5) Minor (0.5)
REFERENCE MIT RSM paragraph 9.3 page ANSWER E.05 (2.00) Prevent leakage out of the building in the event of internal overpressur (1.0) .0 psig greater than atmospheric - positive (0.5) 0.1 psig less than atmospheric - negative (0.5) l' REFERENCE MIt RSM paragraph 8.1 page . l l l
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E ~. SAFETY AND EMERGENCY SYSTEM 5 Paga 32
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AN5WER E.06 -( 1. 50 ) (0,5) Ke'ep core covered (0.25) and core temp < boiling (0.25). gpm (0.25) within 5 minutes of activation (0.25).
REFERENCE
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MIT-RSM paragraph 3.2.7, Emergency Cooling page 3.4 , l' w e i
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't F [. STANDARD AND EMERGENCY OPERATING _ PROCEDURES Page 33 _
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ANSWER F.01 (3.00) Reactor Shutdown Console Key Switch "OFF". Console Key Switch Removed and in Proper Custod . No work in progress within main core tank involving fuel or experiment . No maintenance of the core structur . No maintenance of installed control blade . No maintenance of control blade drives unless visibly decoupled from the control blad ,
(Any 6 @ 0.5 each)
REFERENCE MIT-FM 2.2 page 3 of 11 ANSWER F.02 (1.50) False :m' False R.hu (hr. lbA bd N)') REFERENCE MIT PM 2.2 pp. 3 of 11 Item 16, PM 2.3 page 1 of " ' tem 2.3. J' ANSWER ,, F.03 (2.50) Yes (0,5) Reactor Superintendent, Duty Sbift Supervisor, and Electronics i
<,/.tM N. ( A en um -+u i Supervisorg e (O.5 each)
File it in the front or the reactor console lo ,.
(0.5)
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REFERENCE' MIT FM 1.9 Paragraph 1.9.2 page 1 of 2 ANSWER F.04 (2.00) An emergency where_such action is needed immediately to protect the public health and safet . No action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent.
.- Must be approved by licensed SRO prior to taking the actio . NRC notified if possible (prior to taking the action).
(0.5 ea) REFERENCE MIT-FM 1.3 page 1 of 2 and 2 of 2 ANSWER F.05 (1.50) Acknowledge the alar . Scram the reactor (minor) if not already scramme . Verify reactor power decreasin . Notify reactor shift superviso (Any 3 @ 0.5 each) REFERENCE d' MIT AOP 5.1.2, Withdraw Permit Circuit Open
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. . 1 ~ STANDARD AND EMERGENCY OPERATING PROCEDURES Page 35 l . .
ANSWER F.06 (2.00) a. This action prevents pumping vibration (0.5) which may occur if one pump is left running (0.5) with certain HX lineups.(0,5) b. True (0.5) . REFERENCE
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MIT AOP 5.2.4, Low Flow Primary Coolant, Immediate Action No. 4 ANSWER F.07 (2.00) This action will prevent damage to the drive motor if a scram occurs (1.0) since the motor wculd attempt to drive in and being unable to move the motor would burn out.(1.0) op fr av 4 sa S~ rw GY7 e A/3 4 (t.o) SA .ud / L bl h n : f y > q dly 1/ pdQ #sd . (i.e) REFERENCE MIT AOP 5.8.9, Malfunction of a Slim Blade / Regulating T.od, Followup Action No. 4 i l'
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G'. RADIATION CONTROL AND SAFETY Pcgs 36
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AN5WER G.01 (3.00) To determine that normal radiation levels exist based on control room and/or local instrumentatio . To assess the need for a radiation survey with a portable detecto . To evaluate the potential for dose rate changes during occupanc (3 at 1.0 ea) REFERENCE
MIT-FM 1.14, pg. 6 Item No. 9 ANSWER G.02 (1.00) To save a human life (0.5) or to insure nuclear safety (0.5).
REFERENCE MIT-FM 4.3, page 14 ANSWER G.03 (3.00) High flux regions such as the thermal column, pipe tunnel, lid space, experimental port, instrument lead boxes, reactor floor hot cell, 36V's if not sealed, or a drop in building temperature r
(any 4 at 0 0.5 each.)
b.The high flux regions are sealed and/or flooded with carbon dioxide (0,5) or helium in order to exclude as much air as possible since AR-4C is present in ai (0.5)
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G: RADIATION CONTROL AND SAFETY Pass'37
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REFERENCE-RSM page 7.5 paragraph 7.3, Reactor Floor Argon-41 monitor ANSWER G.04 (1.00) id~ Because of the Tritium (0.5) which is angalrh emitter (0,5).. REFERENCE MIT-FM 4.5, page 4
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AN5WER G.05 (2.00) Beta-Gamma Monitoring Badge Pocket Dosimeter (gamma)
(1.0 ea)
REFERENCE MIT-FM 2.5, page 1 ANSWER G.06 (2.00) False , False l' False d. True (0.5 ea) -
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.. . . . . . " RADIATION CONTROL'AND SAFETY Pasa 38 .
. REFERENCE' MIT-FM 1.12, Film Badge Classification page 1 of 5
' ANSWER G.07 (2.00) Film Badge (0.5) (0,5) . Record the exposure on the posted dosimeter shee . Re-sero the dosimete (0.5) False (0.5)
' REFERENCE MIT-FM 1.12 page 5 of 5 of June 8, 1983 (Memo Karaian to Rad Workers and Experimenters
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U. 3. NUCLEAR REGULATORY COMMISSICH JEliiUR EEACTOR OPERATOR LICEUdE EXAMINATICN
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FACILITY: Maid. INSTITUTE CF TEC . REACTOR TYPE: RESEARCH __ i
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- DATE ADMINSTERED: 88/ 5 EXAMINER: ROESENER, CANDIDATE
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INSTRUCTIONS TO CANDIDATE:
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Use separate paper for the answers. Write answers on one side onl Staple queetion sheet on top of the answer sheet Pointe <~or each qaestion are indicated in parentneses after the question. The passing rode r+ua rn at R.u ; 'O. 1:. each e s tescry . R =inaticr. pap rs wi l' . be picaed up six (r3) hvurs after the er. amination start % OF JATEGORY 4 0F CANDIDATE'S CATEGCRY VALUE TOTAL SCORE VALUE CATEGORY
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20.00 20.00 REACTOR THEORY 19.0 0 C.C'7 20.00 RADIOACTIVE MATERIALS HANDLING DISPCSAL AND HAZARDS 20.00 20.00 SFECIFIC OPERATING CHARACTERISTICS 20.00 20.00 FUEL HANDLING AND CORE PARAMETERS I9.00 ADMINISTRATIVE PROCEDURES, ' {." 7 20.00 CONDITIONS AND LIMITATIONS
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Final Grade
, All work done on this examination is my ow I have neither given l ncr received ai I.oyj g,pg g, egp )) J Candidate *s Signature
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NRC RULES AND GUZDEL1NES FOR LICENSE EXAMINATIONS
, Turing tne acministration of this examination the following rules apply: Cheating on the examination means an automatic denial of your application and could result in more severe penaltie . 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 cheatin . Use black ink or dark pencil only to facilitate legible reproduction . Print your name in the blank provided on the cover sheet of the examinatio . Fill in the date nn the cover sheet of the examination (if necessary). Use only the paper r ovided for answer Print ycur name in the upper right-hand corner of the first page of each
, section of the answer shee . Consecutively number each answer sheet, write "End of Category __" as appropriate, start each category on a new page, write only on one side of the paper, and write 'Last Page" on the last answer shee Number each :,nswer as to category and number, for example, 1.4, . Skip at least three lines between each answe . Separate answer sheets from pad and place finished answer sheets face down on your desk or tabl . Use abbreviations only if they are commonly used in facility literatur . The point value for each question is indicated in parentheses after the question and can be used as a guide f or the depth of answer require . Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or no . Partial credit may be given. Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLAN . If parts of the examination are not clear as to intent, ask questions of the examiner onl l' 17. You must sign the statement on the cover sheet that indicatos that the'
work is your own and you have not received or been given assistance in completing the examination. This must be done after the examination has been complete _ - . , __
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13. When you complete your examination, you shall: Assemble your examination'as follows:
(1) -Ex&m questions on to (2) Exam aids - figures, tables, et (3) hnswer pages including figures which are part of the answe * Turn in your copy of the examination and all pages used to answer the examination questions, Turn in all scrap paper ar.d the balance of the paper that you did ,
r.ot use for answering the question Leave the examination area, as defined by the examine If after leaving, you are found in this area while the examination is still in pregress, your license may ce denied or revoked.
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- !!. RE ACTUfL T!!E'.lRY Pago 4
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GUEGTION H.01 (2.00; Define EFFECTIVE Delayed Neutron Fractio (1.00) Hcw would the Effective Delayed Neutron Fraction of the MITR-II change (INCREASE, DECREASE, REMAIN THE SAME) if the deuterim reflector were replaced by a light water reflector? Briefly explain your answe (1.00; uGE3 TION H.00 (3.00)
* What are the three parameters needed to calculate the power being transferred from the fuel to the primary coolant? (1.00) To obtal:. the total power output of tae reactor what two other coolant str ans must be evaluated.' (1.0 Why does it take about 24 hours of constant power operation before thermal equilibrium operation is st,ained for the MITR-II reactor?
(1.001 GUESTION H.03 (2.00>
If the moderator and reflector temperature rises from 35 C to 40 C. how far would the regulating rod be moved in order to maintain criticality? See the attached curves for information needed to calculate the rod motion. Show all work. State all assumption r
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1 REACTOR THEQRY Page 5
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QUESTIOtt H.04 (3.00) The reactor operator is conducting a routine reactor startup after it has been shutdown for several days. Prior to withdrawing a shim blade he reads a stable count of 50 cps on the startup channe Im'm ediately af ter withdrawing this blade he reads a count of '80 cp If he performed no blade motion for five minutes, would the count rate illCREASE. EECREASE or REMA!!1 THE SAME? Explain, assuming the reactor is suberitical at 80 cp (1.50) After 5 minutes he withdraws another blade the same distance (assume the same amoun of reactivity is added as in "a" above) but the reactor is still subcritica Is the change in count rate
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for this second rod withdrawal GREATER THAN, LESS THAN or THE SAME As the change in count rate observe in the first case? Explain your answe (1.50) QUESTIO!3 H.05 (0.00) If the shim bank is at the height predicted by the estimated critical position (ECF), anc the ECP was done Ocrrectly in accordance with MITR-II procedures, what will be the
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indications seen by the operator? . GUESTION H.06 (1.00) What is the major source of neutrons used for routine startups of the MITR-II?
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QUESTION H.07 (2.00) State TWO reacons why the differential rod worth of the regulating rod peaks at only a few inches of withdrawal as compare &'to the near center of travel peak of the shim bank differential wort (The differential worth curves are included as an attachment to aid in the visuali ation
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H. . REACTOR THEORY Page 6 QUESTION H 08 (3.00) How long after a reactor shutdown.does the xenon concentration peak? Assume one week of full power operation prior to shutdow (1.00) Explain in terms of the two production and two removal processes why the xenon buildup occurs following the shutdow (2.00)
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QUESTION H.09 (2.00) Assuming the reactor to be just critical at 100 kW: Explain tne initial (prompt) response of the reactor power to a ten inch insertion of the regulating rod. A general explanation is desireo 110T a calculatio (1.00) Describe the behavior of reactor power at two minutes following the ten inch insertio (1.00) . J'
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' RADIOACTIVE MATERIAL 5~ HANDLING DISPOSAL Pago 7 , 'AND HA"ARPS l .
QUESTION I.01 (2.00) When the reactor is in normal operation, how often should the
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secondary water be sampled for Tritium? (0.50) What three actions must be taken if-secondary water tritium concentrations exceed one microcurie / liter? (1.50) QUESTICH I.02 (2.00)
* In the event of an MITR-II radiological emergency, who is responsible for decisions and coordination of all immediate actions? - (1.00) What two actions should always be taken to maximize emergency plan effectivenes (1.00)
GLESTION I.03 (3.00) According to procedure PM-4.7.2 the EAL for a General Emergency can be determined by using the following formulae: For stack release... (1.58E7)(Permissible Concentration) 2. For contairtment release...(5.26E7)(Permissible Concentration) Why is the limit for stack releases more restrictivethan the limit for containment releases? (1.00)
. The above formulae assume that the particular radioisotope being released is known. Explain how the EALs account for the fact that the limiting radioisotope being released may not be known?
(1.00)
r List FOUR of the radiation monitor levels that you would read to determine if a General Emergency should be inacted? (1.00)
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. AND HAZARD 5 . = QUESTION I.04 (1.00) ~
Multiple Choice:(Choose on1 one answer) . There must be no direct ontact with. fingers on the irradiated containers or samples ecause of the high probability of: gamma radiat Dels r ed beta radi- io surfac contamination,
, alp a contaminatio ,
GUESTION I.05 (1.50) List THREE independent measurements or indications use to monitor or detect heavy water leakage into the seconcary coolan . QUESTION I.06 (1.00) Why must the blowdown of the cooling tower basins be secured when the reactor is not operating? QUESTICN I.07 (1.00) ' What TWO automatic responses of the sewer radiation monitor are verified prior to a discharge from the waste storage tanks? r
;UESTION I.08 (1.50) State the THREE nutomatic responses you would verify after receiving an "High Radiation Level Core Furge" alar GUESTICU I.03 (1.00)
5 tate the purpose of the Core Purge (Off-Gas) Syste (***** CATEGCRY I CONTINUED ON NEXT PAGE *****)
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, _ _ .__ __ .... ... . .7 !. RADIOACTIVE. MATERIALS HANDLING DISPOSAL Pag 6 9 - AND HAZARDS .
QUESTION I.10 (2.50) Why must the, "DO NOT USE THIS EXIT." sign on the inner door of the main' personnel airlock be back. lit when the, "Trouble NWel2 Gamma Monitor," alarm actuates? (1.00) State the' TWO warning indications that will be automatically actuated outside of the control room when the "Trouble NW-12 Gamma Monitor.' scam alarm actuates, include in your answer the location of the warnings which occur outside the control roo (1.50)
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QUISTION I 11 (1.50) What TWO individuals should jointly authorine emergency exposures wnich may exceed 10CFR20 limits? (1.00) . If it is not possible to reach the above individuals in a timely manner, who 13 the next in line to make the emergency expcsure authorications? (0.50; CUE 3TICN I.12 (2.00) What is the radiological concern associated with continued reactor power operations during a containment isolation? (1.00) What must be done in order to continue operations if a single plenum radiatLon monitor becomes inoperative due to a plugged flow line and containment isolates? (1.00) r
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?ECIFIC CPERATING CHARACTERIGTICG .Page 10
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QUESTION J.01 (4.00) Regarding the Convection and Anti-Syphon valves within the core tank: , Give the purpose of each of these valve type (1.00) Describe how each valve type functions to carry out its purpos Include in your explanation a description of the valve's perfor.mance in BOTH the normal AND the accident position (3.00)-
' (A simplified sketch demonstrating these points is acceptable.)
CUESTION J.00 (1.50) Th;lE or FALdE? The main core tank level indicator ML-3A is an electrically driven transmitter which is supplied by emergency powe (0.50) ML-JB is a pneumatically pcwered system and indicates on 1:near scale meters that are mounted in the control room and the emergency instrumentation cabinet in the utility roo (0.50) A reactor scram on low main tank water level will occur only if both ML-3 anc ML-2 level probes are uncovere (0.50) QUESTION J.03 (2.00? When would it be necessary to utilise the containment pressure relief system? ( 1. 0 0 .' What TWO functions does the containment pressure relief system bicwer perform? (1.00) J' QUESTION J.04 (1.00) In accordance with Standard Operating Plans , FM 2.3.1, Normal Reactor Startup.* what should be your immediate response if the reactor goes critical at a position more than 0.5 inches below the ECP7 .
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. . ' . .- tPECI71C OPERATtNG CHAF.ACTER8STZC3- Page 11
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QUESTION J.05 (2.00) Answer the following questions for a "Normal Reactor Startup": a ., Explain how the transfer from the fission chamber mode to the ion chamber mode is mad (1.00)
. ' Explain the response of the channel to the actions in "a" above.
, . -(1.00)
GUESTION J.06 (1.00)
#
Luring a reactor startup, why is the reactor power held for five minutes after each MW of power increase? GUESTION J.07 (2.50) Following a three day shutdown, the reactor was started up and has been r. inning at full power for 10 hour Assuming that the regulating rod is in-automatic control and that no changes in the sample or experiment configurations are to be made, how many times will it be necessary to manually cycle the regulating rod over the next 30 hours. All reactivity curves you will need for this question are attached to the tes Show all your wor GUESTICN J.08 (3.00) During the first'few hours of power operation following a short (4 hour) shutdown from an extended period of full power operation, the operator
- makes the mistake of allowing the Xe transient to drive the regulating rod to the near-in positio ~4 hen the automatic rundown light energizes the operator notices i What should be his THREE immediate actions if he"desires to continue power operations? (1.50)
"' How long after the light energi:es will a rod start to move?
(0.50) In accordance with Standard Operating Plans, FM 2.4.1 "Full Power Operation," when can the operator no longer take the actions of step
'a" to recover? (1.00)
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QUESTION J.09 (3.00) List FIVE of the immediate actions that are the responsibility of tne Reactor Superviaor upon a loss of normal wiectrical power in accordance with the Abnormal operating Procedure, PM 5.8.4, "Loss of Norme.1~ Electrical Power."
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X... FUEL HANDLING AND CORE PARAMETERS Page-13
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QUESTION .K.01 (2.00) ' List FOUR indications that would be indicative of a release of fission products from the fuel elements in the core tank? QUESTION K.02 (2.00) What FOUR abn>rmal conditions may cause a Spent Fuel Storage Pool alarm?
,
4JE5TIQ3 K.03 (3.00) After each refueling or change in core loading, the reactor shall not be operated above a power level of 1.0 KW unless an evaluation is made to er.aure that two Technical Specifications are satisfied, What are the two Technical Specifications? (2.00) What persons shall complete and approve these evaluations? (1.00) QUESTICN K.04 (4.00) During refueling, what are TWO designed safety features associated with the hold-down grid plate AND vhat is the design purpose of each? (4.00) GUESTION K.05 (1.00) What is the basis for the Technical Specification that, "Prior.to transferring an irradiated element from the reactor vessel to the transfer flask, the element shall not have been operated in the core at
a power level above 100 kW for at least four days?" QUESTION K.06 (3.00) What Xe condition and TWO adverse control rod positions are assumed unen calculating shutdown margins? (1.50) -
. Explain what variable reactivity is AND TWO factors which cause it to occu (1,50)
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'i . . . FUEL HANDLING AND CORE PARAMETERS Page 14 QUESTION K.07 (2.00)
.' According to your Technical Specifications, when is-your reactor considered secured if fuel is present in'the core?
.
QUESTION K.08 (1.00) According to MITR Technical Specifications, approximately how many
* MITR fuel elements would be required for criticality assuming *
optimum conditions? (0.50) Anat is the maximum numoer or fuel elements allowed by Technical Specifications to be cutside designated , storage areas at any one time? (0.5 . QUESTION K.09 (2.00)
' When loading a new 445 gram U-235 fuel element during refueling, which ring (A,B or C) will give you the highest m0/ gram of U-2357 (0.50; If you load a 506 gram element instead of a 445 gram element in the same fuel location, will the m3/ gram of U-235 for the 506 gram element be GREATER THAN, LESS THAN or THE SAME AS that of the 445 gram elemen Briefly explain your answe (1.50)
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. . . , . ADMINISTRATIVE PROCEDURES CONDITIONS Pago 16 -
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QUESTIOJ L.01- (1.50) In accordance with MITR-II Technical Specifications, who must be present (minimum level of qualification) if the reactor is not .. secured but is shutdown? (1.00) For the conditions mentioned above, what additional personnel must be available on site or on call? (C.50
, QUESTION L.02 (1.00)
TRUI or FALSE 7 W.'rk shal; not be conductec in the reactor building un1+ss a reactor supervisor or a reliable person appointeo by a reactor supervisor is presen (0.501 The shift supervisor may grant permission to an experimenter to irradiate acids or other corrosive liquid (0.50) QUESTICN L.03 (2.00) What TWO things must you do before making a tour outside the range of the intercom system if you are the on-duty senior reactor operator? QUESTION L.04 (4.00) Whose permission is required to post a warning tag on facility equipment and who may post a warning tad? (1.00) ' List FIVE requirements which must be observed when
"locking out" facility equipmen J' (3.00) . . (***** CATEGORY L CONTINUED ON NEXT PAGE *****)
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QUESTION L.05 (2.00) In accordance with MITR-II Technical Specification 7.8.3 and a. . Administrative Procedure. PM 1.5, "Procedures Adherence
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Temporary Change Method," in what case may a temporary change be: made to a class B procedure without the pre-approval of the Director of Reactor Operations? (1.00) Who must approve the temporary change discussed in "a" above? ~(Give level of qualification.) (1.00) e QUESTION L.06 (3.00) What are the FOUR interrelated variables associated with the core thermal and hydraulic performance on which Safety Limits are based?
- (2.00) What is the objective of the Safety Limits? (1.00' '
QUESTION L.07 (1.50) In the~ event of a required building evacuation following an emergency; Who must normally authorize re-entry into any portion of the reactor facility? (0.50) , Under what circumstances can the on-duty shift supervisor authorice ! re-entry? (1.00) ! , r l
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QUESTION L.08 '(3.00) Ut.ilise the attached EAL's to answer the following:
, For each of the events below, state the minimum emergency classification which may be declare (0.50 each)
1. A large crowd of protesters marching around the reactor buildin . A fire damaging-an experiment which causes the release of radio-active materials, b 3. A tornado damaging the containment buildin . A slow and uncontrollable decrease in core tank level such that level remains above the anti-syphon valves, What criteria ' used for classifyias emergency conditions? (1.00) oa le re d GUESTION L.09 (2.00) . 10CFR55 defines an operator as any individu'l a who manipulates a control of a facility: Define the term "control" as it applies to MITR-I (1.00) Under what conditions is the person physically manipulating a control not required to hold a valid cperator's license? (1.00;
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. . . EEACTOR THEORY Page 18 ANSWER H.0) (2.00) The fraction of the thermal neutron population that was born delaye (1.00) DECREASE-[0.50]
Loss of the large source of slow born photoneutrons created in the deuterium LO.50] results in the decrease in the effective delayed neutren fraction if the heavy water reflector-is replaced by light
'
wate (1.00) REFERENC Reactor Physics Notes, Reactor Kinetics and Control Rod Calibration by Reactor Period Measurement, p. Reactor Systems Manual, Ch. 10, Sec. 1 ,
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ANSWER H.02 (3.00) . Delta Temperature of'the primary coolant.[0.33) 2. Heat Capacity of the primary coolant.[0.33) 3. Flow rate of the primary coolant.[0.34) (1.00) . Deuterium Tank coolant system.[0.50) 2. Shield coolant system.[0.50) (1.00) Because the graphite reflector has a large heat capacit (1.00) ' l' REFERENCE
,, Reactor Systems Manual, Ch. 6 Sec. 6. . . (***** CATEGORY H CONTINUED ON NEXT PAGE *****) - - _ _ . .. . . _ , --. . . - . - ., ,.
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'h e- REACTOR TiiEORY Page 19
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ANSWE H.03 (2.00) ' Reactivity worth at 35 C --> -75 m0 (+/- 5) [0.33] I2 L Neactivity worth at 40 C --> -k15 m0 (+/- 5) [0.33) Change in reactivity --> h m0 - (-75 m3) = , h m2 (+/- 10) [0.33] Assuming regulating rod initially at 6 inches out (any value will be
.
accepted) its worth is 190 mp [0. 3). 190 m0 +jFs m2 : }&3 m2 which is equivalent to a rod height of in. [0.35] f/ A4/ Therefore the rod would move out,J~h) Yin. [0.33). (Must be consistent within .6 in. casea on reactivity insertion calculated for temperature change.) (2.00) . REFERENCE Reactor Systems Manual, Ch. 10, Sec. 1 Reactor Systems Manual, Ch. 10, Figure 10.18 "Reactivity Effects for Uniform Heating of Primary and Reflector".
FM 6.5.16.1 "MITR-II Integral Reg Rod Worth Curve".
. ANSWER H.04 (3.00) INCREASE [0.50]. Right after the rod motion ceases suberitical multiplication equilibrium level is not yet established [0.50]. The level will continue to increase until the new equilibrium is reached [0.50). (1.50) GREATER THAN [0.50). As the multiplication factor approaches one (or as the reactor approaches criticality) [0.501, the number of generations required to reach equilibrium increases (0.50] and
,,
therefore the change in count rate increase (1.50) OR The final equilibrium level is proportional to (1 -1/p).
As K approaches 1, p approaches zero from the negative side and each step change in reactivity causes (1 - 1/p) to chsnge (increase) by a larger amount resulting in a larger increase in the associated neutron level (or count rate).
(As p aprroaches 0 from the negative side (1 - 1/p) approaches infinity.) (Nots Bien: p: rho.)
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' . . REACTOR THEORY Page 20 . REFERENCE '
Reactor Physics Notes, Reactor Startup and Reactor Suberitical Multiplication, pp. 7 - 1 ANSWER H.05 (2.00) Slight positive period (about 50 seconds) (1.00] and a steadily
' increasing power level (count rate) (without blade motion) (1.00].
(2.00)
hEFERE::CE Standard Operating Plan 2.3.1 "Normal Reactor Startup", step 1 . ANSWER H.06 (i.00) Photoneutror.s . OR Gamma + Deuterium --> Hydrogen + Neutron (1.00)
.E FERENCE Reactor Physics Notes, Reactor Startup and Reactor Suberitical Multiplication, p. J' ; - .
ANSWER H.07 (2.00) - The full-in position of the reg rod is six inches above the bottom of the cor (1.00) The reg rod is heavily shadowed by the shim bank in its upper region of withdrawa (1.00) ,
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- _ _ _ _ _ _ _ _ _ . - - - - - - - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ , . > SEACTOR THEORY Page 21 REFERENCE Reactor Systems Manual, Ch. 10, Sec. 10.4.
T ANSWER H.08 (3.00) hour (1.00)
' Immediately following reactor shutdown, the production of xenon from fission and its removal by burnup are effectively stopped [0.50].
Since the production from fission is small compared to the production of Xetion by iodine decay [0.50] and since the burnup is large ccmpared to the loss from decay of xenon (0.50] the net effect is a large decrease in xenon loss and a small decrease in xenon production and-thus a large increase in xenon concentration (0.50).
(2.00) REFEPINCE Reactor Systems Manual Ch. 10, Sec. 1 .
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ANSWER H.09 (2.00) The reactor power will drop immediately [0.50] due to the quick response of the prompt neutrons to the change in reactivity [0.50]. i (1.00) At two minutes the reactor power will be decreasing [0.50] at a rate controljed by the decay of delayed neutron precursors (0.50). (1.00)
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REFERENCE
' Reactor Physics Notes, Reactor Kinetics and Control Rod Calibration by Reactor Period Measurement, p. 1 _
For a nore's $ in ser r E o ' 0 $ TC n InCSCC;7"bC )*CWCf 'Y'N b CJ ' YU (b f dwn im m edig rely [gSO] duc To Thc, yui tk rgsfcnSe o$ 7~ T*$ c f ro n es;r ron.S 7~o The. r e nf C, h ts t'] t if) rc o t TIvir~j G.7&,
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. . RADIOACTIVE MATERIALS IIANDLING DISPCSAL Paco 22
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ANSWER I.01 (2.00) a.- At least once every 24 hour .
(0.50)
The cooling Tower Spray must be secured (0.50), . . Any discharge to the sanitary sewer of secondary water must be stopped (0.50).
" The heavy water reflector heat exchanger must be isolated (0.50)
(1.50) '
REFERENCE Technical Specification ANSWER 1.02 (2.00)
, The senior NRC-licensed member on shif (1.00) . Shutdown the reacto . Isolate the containmen (1.00)
REFERENCE Emergency Plan and Procedures, Se .3.1. . r
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1 RADIOACTIVE MATERIALS HANDLING DISPOSAL Pags 23
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ANSWER I.03 (3.00) Because the dilution for stack releases is caluculated to be less yhan the dilution for containment release (1.00) The EALs are based on the assumption that the most limiting MFC is being release (1.00) Any four of the following at 0.25 each: 1. Plenum gas monito ' 2. Plenum particulate monito . Stack gas monito . Stack particulate monito . Stack area monito (1.00) g, t, e r< de. m e. ;, e s < * oire, b e d< P.EFERENCE Emergency Pltin and Procedures, Sec . 4. 7. 2. 2.1 & 2, dal 't. 7 2. t. // 't. 's if /Yg /.
#km 6<. poter ef maalev*~ < ** < * ** rre ri* * l* # < coe re in ma r re tu , ,, ;3 ,, , s, );sen,ce tw<y from rse Abtre. A < s necen) coa 3 sI.ers es rhe f *I* r *f *p *)'in oin ur m & ceco* e ece* rr<
en!'!* C.* ke**d
<<feb rker~ of conc er n e r rh e. sa.ne. h e >rm al ec le < ce r [o..re -),
ASSWER I.04 (1.00)
'. v e/< re <{
REFERENCE Administrative Procedures. Sec. 1.10.8.1.1-11.
. )* ASWER I.05 (1.50) Secondary water radiation monitorin . Daily sampling and analysis of the secondary water (for tritium). Monitoring of level in deu.terium dump tank (by a low level alarm or by site glass). (1.50)
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-* RADICACTI'/E MATERIALS hat 4DLING DISPOSAL Pego 24 AND HA'.:ARDS . .
REFERENCE Technical Specification . % ANSWER I.06 (1.00) Because tne secondary radiation monitors are not capable of detecting any primary to secondary leakage when the reactor is shutdow REFERENCE
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Reactor Systems Manual. Ch. 7 Sec. 7. ANSWER I.07 (1.00) The "High Level Radiation Moaitor" alarm actuate (Exact name of alarm is not required for full credit.) (0.50) The sewer pump (RM-3) stop (0.50) REFERENCE Operating Procedures Checklists. PM 3.6. "Waste Storage Tank Dump Procedure'. . Re unr- c he ny sefraceaGea. h <.s rh oje#e4/liry af rk se ee (* di srks er *< /re e e k alaj . h o. Tsis s. i d nee l re b h ,,/Ja,f t, 74;, A n swer in ras brer ANSWER I.08 (1.50) Closure of the intake valve (MV-83). . Closure of the core purge blo"er suction valve (the primary storage
- tank air discharge valve) (MV-64). The core purge blower trips of (1.50) (***** CATEGCRY I CONTINUED ON NEXT PAGE **=**)
. . . . . F.ADI0 ACTIVE MATERIALS HANDLING DISPCSAL Pega 20 . AND HAZAhDS .
REFERENCE Abnormal Operating Procedures, PM 5.6.5, "High Radiation Level Core Purge", p. Reactor Systems Manual Ch. 7 Sec. 7.7.
. ANSWER I.09 (1.00) fe e /.00)vf: The purpose of the core purge system is to (e piny one ofrksa kilm'<3,1-.. r.c '. y-;-; ;; > , 7 ,; i- + w .-- t.n<a * va +b. - 4 t. n .- w'
, Per << < r Ardr*3 e n >~ ild v REFERENCE 2, fee-* * r A cci a-4 rN f 4 'l 3. fre n < r n e e vn . l< &*< o+ N ".
c.eactor Systems Manual, Ch. 3. Sec. 3. Te d ni csl Speco*hcorim y 3.'f ANSWER I.10 (2.50)
' * To be conservative, it is assumed that the actual radiation level in the set-up area is high [0,50] and, until proven otherwise, personnel are not allowed to exit through the airlock [0.50).
OR To protect personnel [0.50] from potential overexposure [0.50].
(1.00) A blue light [0,50) and a warning bell [0.50] at the reception desk [0.50). (1.50)
REFERENCE Abnormal Operating Procedures, PM 5.6.1, "High Radiation Set-Up Vault", and FM 5.6.4, "Trouble NW-12. Gamma Monitor".
, Reactor Systems Manual, Ch. 7, Sec. .
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ANSWER I.11 (1.50)
; The Director of Reactor Operations [0.50) and the MITR Radiation Protection Officer (0.50). (1.00) The Se'nior licensed member of the NRL Staff on-sit (0.50)
REFERENCE Emergency Plan and Procedures, Sec. 4.3. . ANSWER I.12 (2.00) Buildup of Ar-41 in the containment buildin (1.00) Bypass the affected channel using the key switch [0.50) and reopen dampers and restart ventilation fans (restore containment ventilation to service) [0.50). (1.00) REFERENCE Abnormal Operating Procedures. PM 5.6.3, "Trouble Radiation Monitor".
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.s . .. . . . SPECIFIC OPERATING CHARACTERISTICS Pago 27
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ANSWER J.01 (4.00) a. The anti-syphon valves prevent syphon draining of the core tank following a break of the reactor coolant inlet line (0.50).
'The convection valves permit convection cooling during periods of . no forced flow (0.50). (1.00) The anti-syphon valve's ball is forced upward to cover the syphon break during normal primary pump operation (0.75). If the
- inlet line were to break the pressure holding the ball ball would be lost and gravity would cause the ball to drop which would ,
uncover the syphon break (0.75).
The convection valve's ball is forced upward to cover what is effectively a core bypass hole during normal primary pump operation (0.7b'. If primary coolant flow was Icst, gravity would cause the ball to drop which would uncover the core bypass hole completing a path that would allow for natural circulation (0.75).
CR Make sketches similar to those attached which demonstrate the above point (3.00) REFERENCE Reactor Systems Manual. Ch. 1. Sec. 1.5.3; Ch. 3, Sec 3. Reactor Systems Manual, Ch. 1, Figure 1.17 "Natural Convection Valves", and Figure 1.16 "Anti-syphon Valves".
. ANSWER J.02 (1.50) a. TRUE J' b. FALSE
"' c. FALSE (1.50)
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. ..- . . SPECIFIC OPERATING CHARACTERISTICS Page 28 ^
REFERENCE MITR RSM Re' actor Systems Manual Ch. 6. Se . ANSWER J.03 (2.00)
, If building pressure exceeds 2 psig, and if radiation levels and/or structural damage preclude opening the main or auxiliary damper (1.00) G a-o a.r c4.r m. s a c. = .n 2 <.: . .. .m....-.
^
.#i ., .v . v .'. . '. s . .
e. . . u an. vase cae enarw s..X s c. ro c., s 4 .*
'A=y r . ,+ '-r e#"s#fe8 "
//o# * ' :<3""<*r-REFERENCE e In s =l= s
<* a r'* /a a' a r M M * * a * -a / /** 4 re e . e. k s . e. < / fitre * ^
Reactor Syst.' ems I[ua^), #dn. 'b',"Te#c. o7 Abnormal Operating Procedures, FM 5.5.7 "Building Overpressure.
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ANSWER J.04 (1.00) Lower the shim bank [0.50] by at least 1.C inch [0.50). (1.00) REFEPENCE Standard Operating Plans, FM 2.3.1 "Normal Reactor Startup". Step 1 r AllSWER J.05 (2.00) The transfer is mcie by adjusting (reducing) the gai (1.00)
(* NOTE: Half credit is lost if the candidate includes adjusting the discriminator as this is no longer done.) Adjusting the gain downward causes a large decrease in the input to the channel [0,50] which results in a short duration negative period indication [0.60]. (1.00) (***** CATEGORY J CONTINUED CN NEXT PAGE *****)
. .- . PECIFIC OPERATING CHARACTERISTICS Page 29 ' .
REFERENCE Standard Operating Plans, FM 2.3.1 "Normal Reactor Startup". Step 1 Reactor Systems Manual. Ch. 5. Sec. 5. ~ .,' * Facility Assistant Superintendent's Explanation-during plant tou ANSWER J.06 (1.00)
*
To allow the reactor core and primary coolant to approach thermal equilibrium (0.50) thereby reducing stress on the fuel element cladding [0.50). (1.00) , REFERENCE Standard Operating Plans, PM 2.3.1 "Normal Reactor Startup", Step 21 & 'l
.
ANSWER J.07 (2.50) At 10 hours into an essentially Xe free startup the Xe reactivity is
- 1675 m2 -
At 40 hours the Xe reactivity is - 3900 m3 Therefore the change in reactivity due to Xe is - 2225 m2 [1.00) The regulating rods cycle between 2 and 10 inches, equivalent to 68 m0 to 248 m0 or a change of ISO m0 for each cycle. [1.00] 2225 m0/ 180 ma per cycle = 12.33 cycles (will accept 11 to 13 cycles)
[0.50) (2.50)
REEERENCE Standard Operating Plans, PM 2.4.1 "Full Power Operation", Step PM 6.5.16.1 "MITR-II Integral Reg Rod Worth Curve".
Reactor Syr.tems Manual, Ch. 10, Figure 10.16 "MITR-II Xenon Startup Transient" & Figure 10.17 "MITR-II Xenon Shutdown Transient".
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ANSWER J.08 (3.00) . Depress the rundown reset butto . Place the control rods on manua ., 3. Reshi (1.50) -
- second (0.50; When the rod being driven in is no longer within two inches of the the bank heigh (1.00; .
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REFERENCE Standard Operating Plans, FM 2.4.1 "Full Power Operation", Step 4 A:'S*ER J 09 (3.00)
,
Any five of the following at 0.60 each:
' Check that the MG set is running [0.30) and adjust its output voltage to 208 volts (0.30).
. Notify the MIT Campus Police and Physical Plant of the extent of the ' power outag . If possible, notify the occupants of NW-12 that the vault alarm was spuriou . Check the personnel accountability board and determine how many experimenters etc. are inside the containment buildin ' Bypass secondary coolant flow to the cooling tower basin . If necessary, open CV-33 to supply pressurized aCr to the. main ' personnel lock gasket ' Enter the containment via either personnel lock, check the reactor floor and basement for personnel and escort all non-essential personnel from the building via the main loc (3.00) VetESy co n s*l h s.: ew. e' l e r- J,is
$. t f ofererer-im m edi < r A s.r i .,s.
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l SPECIFIC OPERATING CHARACTERISTICS Page31!
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REFERENCE Abnormal Operating Procedures PM 5. "Loss of Normal Electrical -l
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Power", pp. 1 & A eaernsi epresrien feeseduma , fM S r b
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. JUEL HANDLING AND CORE PARAMETERS Page 32 ANSWER K.01 (2.00)
, Any FOUR of the following at 0.50 each: Increasing readings on core purge monitor (or ad arm).
Increasing readings on plenum air monitor (or alarm).
Increasing readings on N-16 monitor (or alarm).
Increasing activity in pool sample Increasing primary conductivity (or alarm).
7.ncreasing equipment room monitor , Increasing containment building area monitor (2.00) REFERENCE Abnormal Operating Procedures, PM 5.8.2 "Fission Product Detection in the Primary Coolant", .
*
ANSWER K.02 (2.00) Loss of power to the SFSF control and alarm pane . Leak Low SFSP leve . Low flow through the SFSP ion exchange (2.00) REFERENCE Abnormal Operating Procedures, PM 5.7.12 "Spent Fuel Storage Pool", p. )* , w ANSWER K.03 (3.00) . The ratio F F /d F is predicted to be less than 2.9 (1.00]. HC p ff The core is predicted to operate below incipient boiling at every point in the core [1.00). (2.00)
(1,00; Two Senior Reactor Operator As s ie r < s r~ S.ye r b reede * r' ^ <j 6 a ( The. s.feein ts- 4.r <.Jfe e rh a, < rha m . o r. x </ < /tce .,,; 3,rg, )
or a m ed din srly s in e r h ey (***'* CATEGORY K CONTINUED ON NEXT PAGE *****)
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. . FUEL HANDLING AND CORE PARAMETERS Page 33 REFERENCE .
Technical Specification, 3.1.3, aa J 7,2.
, ANSWER K.04 (4.00 , C**N, A, mo ef r-h s. fe l o wln g sr 2. 0 0 f * * * rs 1. yFeature - The grid iT designed so that there is normally access to
,
only one core position at a time [1.00].
'
Purpose - This limits the amount of water that can be in the core at any one time [0.50] by making it difficult, (though not impossible, for more than one core position to be defueled at time (0.50]. (2.00: 2. Feature - The grid's latch is interlocked with the primary coolant pumps so that if the latch is released, the coolant pumps stop and remain off until the grid is irtched again [1.00]. Purpose - This protects the fuel elements from damage and the
'
reactor as a whole from inadvertent criticalit OR This prevents core components frcm being expelled by hydraulic force [1.00). (2.00)
:
REFERENCE Standard Operating Plans, FM 2.7 "Fuel Handling", .
)*
ANSWER K.05 (1.00)
,' ' This prevents melting of the fuel element by afterhea (1.00) .
REFERENCE Technical Specification, 3.1 Ths c.s e n e r b cor, red unles.s th e s 4 ;a., f w e - yriel b(* de 3 < rc S //f i<sc=recI [f.vo), furfose - Tho's pre n o rs +xl p a ve m e n y- wj rk o s. g-ns e. xim um s i,v e dom re e widry from rhe Shim b la d e.s [ l. co '1 (**ot CATEGORY K CONTINUED CN NEXT PAGE ** * ** )
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. . FUEL HANDLING AND CORE PARAMETERS Page 34
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ANSWER K.06 (3.00)
' It is assumed that the core is Xe free [0.50) and that the regulating rod [0.50) and the most reactive shim blade are fully withdrswn [0.50). (1.50; Variable reactivity refers to reactivity changes that may occur during core life (0.50).
Factors which cause variable reactivity include:
* (any two fuel of the following at,o. fog burnup G5'each) [1.00)
xenon-(and samarium) changes
-
cample enanges experivant changes (1.50: _ REFERENCE Technical Specification, ANSWER K.07 (2.00) The reactor is shutdow (0.50) Console key switch off and key is in proper custod (0.50) No work in progress within the main core tank [0.50) involving fuel or experiments [0.20), or maintenance of the core structure [0.10), installed control blades (0.10] or installed control blade drives when not visibly decoupled from the control blade (0.10). (1.00) i
,*'
REFERENCE
,, Technical Specifications, ANSWER K.08 (1.00)
4 m *ia _
<. n < - 4 : tu.ovJ-d-. 9 f3 (fo/~t) (n,g y '
4, . Y (o se )
(***** CATEGORY K CONTINUED ON NEXT PAGE *****)
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, ' FUEL HANDLING AND CORE PARAMETERS Page 35!
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REFERENCE
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Technical Specifications, 3 .10 . -
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' ANSWER K.09 (2.00) A-ring because the flux is highest due to less leakag (0.50); LESS THAN-(0.50)
'
Due to the self shielding of the more heavily loaded elements.'(1.00).
(1.50) REFERENCE Reactor Systems Manual. Ch. 10 Sec. 10. ,
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-, . ADMINISThATIVE PROCEDURES, CCNDITIONO ~ Page 06 .AND LIMITATIONS s
ANSWER L.01- (1.50) . A licensed-SRO [0.50). Another person (qualification not specified) [0,50). (1.00) MITR Radiation Protection Officer (or his designated alternate).
- (0.50) REFERENCE
Technical Specifications 7. ANSWER L.02 (1.00) a. TRUE (0.50,
'
b. FALSE (0.50) REFERENCE Administrative Procedures, PM 1.14.2.3, Paragraph 6, and FM 1.14.2.1, Paragraph ANSWER L.03 (2.00) Obtain seme method by which the Operator-in-Charge may page yo p (1.00) Inform the Operator-in-Charg (1.00)
-
REFERENCE Administrative Procedures, PM 1.14.1, Paragraph .
. (***** CATEGORY L CCHTINUED ON NEXT PAGE *****) .
.. . . -' ' ADMINISTRATIVE PROCEDUEES, CONDITIONS oa go 37 . AND LIMITATIONS . .
ANSWER L.04 (4,00) . On duty console operator [0.50].
- * Any member of the NRL/RPO staff [0.50]. (1.00) Any five of the following at 0.60 each: .
1. SRO will witness lockout [0.30] AND verify the system is in a safe condition [0.30]. 2. Ferson performing work will perform lockou , 3. Person performing work will retain the key on his perso . A notation as to the system being locked out shall be made on the statas boar . The system must be tagged ou . Lockouts anall be removed under the direction of an SRO [0.30] by the the person who performed the work [0.30). (3.00)
'?. .% s e o hn *% I<emissio** +* en re s- e n -J 7 e , s ./a, e[en te REFERENCE Ad.ainistrative Procedures, PM 1.1 ANSRER L.05 (2.00) If the change does not change the intent of the origine.l. procedur (1.00) . An SR0 [0.50].
l . r l Any other member of the reactor staff [0.50] (1.00) REFERENCE Technical Specifications 7. I4elmin i s trarls freerder e , Pm t. f * Proerdves Adherence 7'enjeny C ha nye A1e rhod . !
(44544 CATEGORY L CONTINUED CN NEXT FAGE - ***) .
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, ADMINISTRATI'/E PROCEDURES, CONDITIONS Fage 33
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.AND LIMITATION 5 ANSWER L.06 (3.00)
. . Total reactor thermal power [0.50].
2. Reactor coolant total flow rate (0.50).
3. Reactor coolant outlet temperature (0.50).
4. Height of water above the outlet end of the heated section of the 4-hottest fuel channel [0.50). (2.00) To establish limits within which the integrity of the fuel clad is maintaine (1.00'
-
OR (will accept)
.
To prevent flow instabilitie REFERENCE Technical Specifications 2.1.
l i ANSWER L.07 (1.50) The Emergency Directo (0.50) U+ildir;c ; ce.uetiv.. 4.2wasivaved by vhe bu11uw vi iusca-si (6. 5v ; l - f ailu w. lveo sf .unLileties for routic.e smesens (i.e., less of I
,2'"-eite elac+rical reus* ^ * = * * [n sn3 (1.00)
J' ,' REFERENCE I l - Emergency Plan and Procedures; PM 4.3.3.3.
, I. when evee r h e- on du ry .Shif r Sqer iso r I.s the. s e r,'ny 6myency g;,e, y , co g, y X . wh en e ve - rs e. b uildin c < < c ~ < rk s w<s n e e e s so rs reat Th b ild -y o F Reg a n - W L o 2 r,1 h //*~ l') < hs r * wa ril< rk , {ve co -rin e. <e r .,o n geo.n s 3 ( s veh < s les s o S o#-s irt. eleerrie<l owee o <-
(*****CATEG[RY L CONT.5re1o.,)ON INU D NEXT PAGE * **** ) (l.00)
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. . . ADMINISTRATIVE PROCEDURES, CONDITIONS Page 39 AND LIMITATIOhS ,
ANSWER L.08 (3.00) Notification of~ Unusual Event
' . Alert . Site Area Emergency . Alert j (2.00) - # ogi[ai C / consequence rdct (1.00) Potential radi REFERENCE Emergency Plan and Procedures, FM 4.5, Tables 4.5.3-1,2,3 & 4 "EALs for Notification of Unusual Events. Alerts, Site Area Emergencies & General Emergencies", Pm 4.4, Sec 4. ANSWER L.09 (2.00) Apparatus and mechanisms of a nuclear reactor, the mar.ipulation .sf which d:rectly affect the reactivity or power of the reactor. (1.00) The individual manipulating the con?.rol must be under the direction of and in the presence of a licensed reactor operator (G.50) and the manipulation must be paet of Tiva individuals training as a ,
student [0.50). (1.00) REFERFNCE 10CFR55, Part 5 , 55.9 and 55.1 r
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TEST CROSS REFERENCE Page 1
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QUESTION VALUE REFERENCE H.01 2.00- ZZZ0000001 H 02 3.00 .ZZZ0000002 H.03 2.00 ZZZ0000003 H.04 3.00 ZZZ0000004
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Table 4.5.1-4: Eats for a General Emerrency q)g Actual or projceted doses at the site boundary in the exposure pathway of 1 rem whole body or 5 ren thyroid for unrestricted areas when averaged over one hou . Notet yigure 4.7.2.2-1 lists the conditions and instrument readings corresnend-ing to a projected off-site dose of 1 rem / hou (PM 4.4.4.15a) >
* Sustained actual or projected radiation levels at the site boundary of 500 mree/
hour whole bod (P M 4. 4. 4 .14 a/ 4 . 4. 4.11/4 . 4 .4 .12) Blockage of fuel element channels thereby causing a loss of coolant to the affected channels and a fuel melt. This is the design basis acciden (PM 4.4.4.154)
'
4 Loss of physical control of eitber the containeent building which includes the control room or of auxiliary areas that house vital eenipeen (PM 4.4.4.5/ 4.4.4.6).
{} Events that have esused or will cause r.assive facility and/or reactor system damage that cou13 lead to the reiting of fue (Pr 4.4.4.15a)
.
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M SR#-0-82-19 AUG 6 1982 _., . - _-. _ . _ _ _ . . _ _ _ . . _ . . _ _ . . _ . . _ ~ _ . . - . _ _ _ . . _ _ _ . - - _ _ _ _ _ - _ _ _ _ _
- . . . , , , ....- ..., . g, P" . Pa. 12 of 13 , .
Table 4.5.1-1: FAf.s for a . Cite Area Fmersenev
'? Confireed abnornal radiation Icvols leadinP to actual or orojected radiolort-cal offluents se the site kenndary exceedine. 250 FPC for unrestricted areas
-- uhen averar.ed over 24 hours. This level corresnonds to an exposure of 375 (PM 4.4.4.15b)
'
mrem whole body accumulated over 24 hour . Same as di except the ef fluents could cause sn inter,rsted exposure of $00 mren thyroi (PF 4.4.4.15h)
. Radiation levels at the site boundary of 100 nrem/ hour sustained for one hou (PM 4.4.4.14b/4.4.4.11) Abnormal loss of pri .sry coolant such that the core tank level drops helev the anti-syphon valve This accident is not considered credihie, Qtote:
but procedures exist for cooing vich it.) (PF 4.4.4.4)
' Inminent loss of physical control of the reacto (P" 4.4.4.6) )* . Severe natural events beine. experienced. These include: (a) An earthquake t hat is causing cbr.ervable danage to the reactor safety equipment vi:hin the containment buildin (b) A flood that is affecting the operability of any reaccce safocy syste (c) Tornado or hurricane force vinds that are daeaging the. containment '
build'in (PM 4.4.4.2)
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, Table 4.5.1-2: EALs for an Alert
' Confirmed abnormal radiation levels leading to actual or orojected radiological effluents at the site boundary exceeding 5n vPC for unrestricted areas when
.
averaged over 24 hours. This level corresoonds to an excesure of 75 nren whole body accumulated over 24 hour (?M 4.4.4.13b) Same as 81 except the af fluents could cause an integrated exposure of 100 mree thyroi (PM 4.4.4.15b) d Radiation levels at the site boundary of 20 mrem / hour sustained for one hou (PM 4.4.4.14b/4.4.4.11) Abnormal loss of prinary cooline such that the core tank level remains at or
*
above the anti-syphon valve (PM 4./ 4.4) , Loss of radioactive material control that causes radiation dose races or air-gs borne radionuclides to increase above oermissible exposure levels by a factor t.)
of 1000 throughout the containment buildin (PF 4.4.4.12) Ri.diatior. dose rates throughout the containment buildinr. ir. excess of 100 mrem / hour sustained for one hour. These levels would necessitate evacuation of all personne (PM 4.4.4.12) A fire leading to loss of radioactive material control within the containment buildin (PM 4.4.443) , An imminent or existing hazard such ast ./' ( l (a) Missile (s) impacting on the containment buildin (b) An explosion that af fects f acility operation.
[ '
. (c) An uncontrolled release of toxic or flammable gases into the containment buildin (?M 4.4.4.9)
s-i l-e, so AUG 6 1982 __ _ . - . _
. _ - _ . __, _
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. Confirmed abnormal radiation levels leading to actual or projected radioloeical s'ffluents at c'he site boundary exceeding 10 M'C for unrestricted aream vben averaged over 24 hoitrs. This icvel corresponds to an exposure of 15 eren note body accurut.ited over 24 hour (p" 4.4.4.15h) Report or observation that severe naturni nhenomena are either terinent or ex-4 istin These include storr.s with tornado or hurricane force winds that could strike the facility, earthquakes that could adversely affect the reactor's safety systees, and floods that coi:1d adversely af feet the reactor's safety system (P 4. 4.4. 2) Threats to or breaches of security. ~ (P" 4.4. 4. 5/4.4.4. 6) A reactor safety limit's being exceeded such that a fuel dar. age accident that h could release radionuclides to the containment building is possibl (PM 4.4.4.1) A fire within the cent.;in=ent building that lasts beyond the incipient stage or for -ere thin ten rinutt' (P." 4.4.4.3) Receipt of a bonb threa (PM 4,4.4.7)
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..
/ ? l ~fa c L u f 3 FACILITY COMMENTS AND NRC RESOLUTIONS FOR MASSACHUSETTS INSTITUTE OF
- TECHNOLOGY REACTOR OPERATOR EXAM ADMINISTERED ON JANUARY 25,198 .
QUESTION B.02 Facility Comment: Top of core shroud is -52 inches which is the inlet penetration for the primary piping. The anti-syphon valves are located at the inlet penetration. RSM 3.2.7, Mode 2 of ECS cooling NRC Resolution: Comment is valid. Credit will be given if this value is utilized in reference to the anti-syphon valve locatio "or -52 inches" will be added to first sentence of answer immediately after "core shroud"
,
QUESTION C.0 Facility Comment: The plugs in the cooling tower standpipe overflow must first be remove NRC Resolution: Comment is valid. No credit will be deducted if this step is included in the discussio QUESTION C.07 Facility Comment: Operators are taught to distinguish between steady (asympotic) and dynamic periods. See MITR Physics Note NRC Resolution: Comment is valid. Full credit will be given if a discussion of prompt and long term effects on reactor perio@'is included in the answer. The answer sheet will be modified to reflect this by adding "Will also accept correct
,,
discussion of steady state and dynamic periods for full credit".
QUESTION E.01 Facility Comment: Believe there is a typo in the answer. Should be MM2 not DM NRC Resolution: Comment is valid. Answer will be changed to correct this
. . . . .
--
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. .
QhESTION E . 0 l I
~
Facility Comment:
* Also, provide core cooling in event of loss of off-site power (Mode I).
NRC Resolution: Comment is not valid. Although this function is provided by keeping the coolant temp. less than boiling it could not be considered acceptable for full credit. Partial credit will be given this response for one of the two required answer QUESTION F 02(b) Facility Comment: Answer.as given is correct, but sometimes we have both a duty shift supervisor (examples, MIT Physical Plant, customer
*
relations) and a shift supervisor (reactor operation only).
So, question could be confusin NRC Resolution: Comment is valid. For this test the answer will be changed to False and for future tests the question will be changed to read "Reactor Supervisor on Duty" instead of "Duty Reactor Supervisor".
QUESTION F.03(b) Facility Comment: The Electronics Technician is currently also the Electronics Supervisor, a licensed SR KRC Resolution: t Comment is valid. For this exam the Electronics Technician l will be accepted as an acceptable substitute for Electronics ' Superviso QUESTION F.04 )
,'
Facility Comment:
-
Question inappropriate for and RO exam. Decisions of this type would be made by the shift supervisor.
l NRC Resolution: l Comment is not valid. The question directly addresses authority limitations placed on the RO by 10CFR, it does not ask him to make the decision. However, since it would be l l reasonable to expect the RO to immediately bring such an action to the SRO for approval this will be taken into account during grading.
l l - - _ - - _ - . . - .
* 4 QbESTION F.07 ; *
Facility Comment:
* Also, prevent sudden reactivity ef fect should blade move rapidly if suddenly free NRC Resolution:
Comment is valid. This could be a possible consequence of not opening the breaker and will be added to the answer as an alternate acceptable answe QUE5 TION G.04 Facility Comment: Appears to be a typo. Should be "Beta" not "Alpha".
, NRC Resolution: Comment is valid. Answer will be changed to read "Beta" instead of "Alpha".
CHANGES MADE BY EXAMINER DURING THE EXAM GRADING 1. QUESTIONS C.06(c) - Added " per pump or 1800 rpm total" immediately after "900 gpa" in the answer. This will account for total secondary flo . QUESTION D.04(a) - Added "with the reactor in automatic control" to end of the answer. This will make the answer exact.
l )* ,
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~
_ Ams uor n Y
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MASSACHUSETTS INSTITUTE OF TECHNOLOGY REACTOR MITR-II
'
SENIOR REACTOR OPERATOR EXAMINATION JANUARY 25, 1988 Specific facility coments concerning the SRO examination, followed by the NRC resolutions, are listed in the following paragraph OVESTION H.09(a) FACILITY COMMENT: Question implies a step insertion of the regulating rod whereas in practice the regulating rod cannot be decoupled from its drive and its maximum speed of insertion would be a ramp rate at 4.25 inches per minute. Both ramp and step insertions of the regulating rod should be accepte RSM 1. ' NRC RESOLUTION: It is agreed that answers assuming ramp insertion should be accepted in addition to those assuming step insertions. The answer key is changed as follows:
"0R For a normal insertion of ten inches, the power will begin to ramp down immediately (0.50] due to the quick response of the prompt neutrons to the ramp change in reactivity [0.50]."
t OVESTION I.02(a) FACILITY COMMENT: Answer should also be the emergency director who may be the most senior NRC-licensed member on shift if the director for operations is not l ' presen NRC RESOLUTION: l In accordance with the "MITR-II Procedure Manual Chapter #4jEmergency Plans and Procedures," paragraph 4.3.1.2.1,
- - "The senior NRC-licensed staff member on shift...is responsible for i decisions and coordination of all immediate actions in an emergency..."
In light of the above procedural direction, the answer "Emergency Director" is not accepte . I 1 '
- !
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l
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_ - .. ._. .- . _ _ _ . ._ .
. . , . OVESTION 1.03(a)
FACILITY COMMENT: The answer is not consistent with the answers from previous NRC exam The correct answer should be that the point at which maximum reconcentration occurs for a stack release is far away from the reactor where the population is concentrated, whereas the maximum concentration for a containment building release is at the containment wall where a 60 foot distance is available before normal occupancy by the general publi NRC RESOLUTION: The question is not the same as on previous NRC exams. It is much more general and, therefore, a much more general answer was expecte However, a correct detailed answer explaining why the dilution factor is more for containment than fo: stack releases is acceptable. The
* following is added as an alditional correct answer: "Because the point of maximum concentration for a containment release is some distance away from the point of concern (where the public has access) [0.50] whereas the point of maximum permissible concentration and point of concern are the same for stack releases (0.50).
ADD REFERENCE:
"Emergency Plan and Procedures, PM 4.7.2.2.1."
@ESTION I.03(c) FACILITY COMMENT: Answer should be: Stack gas monitor i Stack particulate monitor l Stack area monitor Portable monitoring l < ' REFERENCE: PM 4.4.4.14 page 1 and PM 4.4.4.15 page 2 which give thejEALS for general emergencie RC RESOLUTION: , Agreed, additional answer will be added as follows: I "6. Portable monitors at site boundary."
ADD REFERENCE:
"Emergency Plan and Procedures, Section 4.4.4.14, page 1."
l l 2
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!
, OVESTION I.04 FACILITY COMMENT:
Both b and c are correct. MIT has pointed out the ambiguity of this question from previous NRC examinations. Either b or c was previously accepted as the correct answe NRC RESOLUTION: It is agreed that there are two good answers to the question and, there. fore, it is deleted from the test. The note, "Deleted due to ambiguity," will be added to the question in the exam ban @ESTION I 07 FACILITY COMMENT:
Change made to PM 3.6 after material was sent to MRC ADDED requirement to check for operability of the sewer radiation monitor prior to discharg So, there are three automatic response NRC RESOLUTION: Noted. Addad note to exam to go into exam bank as follows:
"Recent change to procedure has the operability of the sewer radiation monitor checked also. This will need to be included in this answer in the future."
OVESTION I,09 , FACli.1TY COWiENT: l Purposes of core purge syeter:
(a) Detect fission product gas releare. (PM 5.6.5) (b) Prevent hydrogen buildup from radiolysis. (Tech Spec 3.4, Basis) (c) Limit Ar-41 buildu NRC RESOLUTION: j, l It is agreed that the core blower is integral in the removal of H2 from - the air space at the top of the core tank. It is not agreed that the system is installed for the detection of fission product gas releas '
The answer is changed to read: ( ,
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.. . . . _ _ _ _ _ - . .. . - . _ _ _ - _ .
>. .. , . 'lhe purpose of +he core purge system is to (any one of the following for1.00): ' 1. Prevent hydrogen buildu . Prevent the accumulation of A 4I Prevent the accumulation of N56,.
ADO REFERENCE:
"Technic 61 Specification, 3.4".
OVESTION J.03fb) FACILITY COMMENT: Pressure relief blower is also used to pressurize the containment building for the annual building pressure and leakage tes . (PM6.1.2.1) NRC RESOLUTICN: , Agreed, the answer will be changed to read as follows: , "Any two of the following at 0.50 each: Pressurizing the containment for the annual leak tes . Cleaning the charcoal filte . Activating th.' charcoal filter."
. QMESTION J.08(a) FACILIT'i COMMENT: Placing reactor on manual control has the net effect of depressing the rundown reset butto NRC RESOLUTION: The procedure clearly states that the rundown reset bettnn is to be presse The answer is not change QUESTION.L.92 FACILITY COMMENT.: Also, supervir,or should verify that:
(a) Core is being adequately cooled. (PM 5.8.4) (b) Operator is following proper procedures. (PM 5.0)
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. NRC RESOLUTION: , .(a) Is not listed in 5.8.4 and is not accepte (b) Is a generic response for all casualties and is accepted. The answer is changed as follows:
I
" Verify the console operator has carried out his immediate actions."
ADO REFERENCE:
"Abnormal Operating Procedures, PM 5.0." l OVESTION L 9},
FACILITY COMMENT: 4 The technical specifications state two senior operator But, it has always been the Assistance Superintendent and the Superintenden NRC RESOLUTION: Since the normal method of operation is more restrictive than the Technical Specification, Assistance Superintendent and Superintendent will be accepted. The answer is changed as follows:
"(The Superintendent and/or the Assistant Superintendent may be named directl Since they are the most senio.r licensed SRO's)."
ADD REFERENCE: l
"Technical Specifications, 7.2."
OVESTION X.04 l l FACILITY COMMENT: There are three safety features. The third is that the blades must be fully inserted in order to rotate the grid. Insures reactor shutdown prior to refueling. (RSM 1.4)
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NRC RESOLUTION:
- Agreed, the answer is changed as follows: "Any two of the following at 2.00 points each: .....(As already written in answer) .....(As already written in answer) Feature - The grid cannot be rotated unless the shim blades are l fully inserted (1.00).
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, Purpose - This prevents fuel movement without maximum shutdown reactivity from the shim blades . (1.00]. (2.00)
ADO REFERENCE:
"Reactor Systems Manual, Chapter 1, Section 1.4." ,
OUESTION K.08 FACILITY COMENT: - Believe answer is correct (81/3 and 4) but it is not legible due to over printing by typewrite NRC RESOLUTION: 4 Agreed, answer separated will read as follows:
" /3 (+ 0/-1) "
OVESTION l.07(b) FACILITY COMMENT: On-Duty Shift Supervisor may also authorize reentry to save a lif (PM 4.4.4.13)
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NRC RESOLUTION: The facility comment and the examination answer are both.too limitin Reactor reentry for any valid purpose is authorized by, "The Emergency Director who is the Director of Reactor Operations or, in his absence, the Senior NRC-licensed staff member on site." Therefore, the on-Duty , Shift Supervisor could authorize reentry any time he was the Senior NRC-licensed staff member on-site (Reference - PM 4.4.4.13, "Rea~ctor ' Reentry") at whenever," the building macuation was necessitated by the buildup of Argon-41 following a loss of ventilation for routine reasons such as loss of off-site electrical power or steam," ~(Reference PM 4.3.3.3, "Authorization for Reentry.") f Also note that the Emergency Director is the Director of Reactor
- Operations but his function is carried out by the on-duty Shift Supervisor and then the on-site senior licensed member of the Reactor Operations Staff (who may be the on-duty SS) until properly relieved of this resporsibility by the Director of Reactor Operations, (Reference PM 4.3.2.1, "Emergency Director.")
.. - _-- . _ _. __ __ . . , 1 . The answer, in light of the above and in li ht of tha way in which the questionwasaskedischangedtoreadasfo$1ows: ' "Whenever the on-duty Shift Supervisor is the acting Emergency Director (0.50). Whenever the building evacuation was necessitated by the buildup of Argon-41 (0.25) following a loss of ventilation for routine reasons I (0.25) (such as loss of off-site electrical power or steam).
(1.00) ! REFERENCE:
"Emergency Plan and Procedures, PM 4.4.4.13 and 4.3.2.1".
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f The following changes were made to the examination as a result of final f4 revie OU$$TIONH.03 Change " 138" to "-126"
*
Change "-63" to "-51" Change "253" to "241" Change "10.7" to "9.5" Change "4.7" to "3.5" Explanation: Error made in reading temperature versus reactivity curv , QUESTION K.06 Change "0.05 each" to "O 50 each"
. Explanation: Editorial erro OVESTION l.04 Add the following:
7. Must obtain permission from the on-duty console operato Explanation: Requirements for hanging the tag are also requirements i for conducting the lockou DUESTIONt.05 Add Reference:
"Administrative Procedure, FM 1.5, "Procedures Adherence Temporary "
Change Hothod".
Explanation: Additional reference as noted in questio OVESTION L.08b , r Deleted <
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Explanation: Question was too vague,,and criteria was used instead of ! the intended "criterio ' !
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