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 NO.
- 50-020/88-01(0L) FACILITY DOCKET NO.
50-020-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 . Norris.
Date Senior perations Engineer.
- 3 Yd ' APPROVED BY: _ groth, Chief Date - Peter V. E 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) candidate.
All candidates passed the examinations and received their licenses.
8803220165 880316
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
l l'
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2/0 1.
l Written l l l ~l 'l i I I L l Operating l 1/0 l 2/0 l l l I I l I
I .l Overall l 1/0 l 2/0 - l - l
I I ' CHIEF EXAMINER'AT SITE: B. S. Norris (USNRC) 0THER EXAMINERS: M. O. Bishop (EG&G) W. S. Rosener (EG&G) . - 1.
.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 programs.
No licensee response is required.
- 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 close.
2.
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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' 1.' -Summary of Comments Made at Exit Meeting:
. a.
It was noted by the NRC that the lead seal on the Emergency Decontamination locker. in 'the control room was not properly attached.
The facility stated that the contents would be reinventoried immediately and that the locker would be sealed, b.
The NRC noted that there is no accountability system in place to control the issuance of keys from the key locker in the control' room.
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 situation.
4.
The written examination questions and answers were reviewed by the facility after all candidates had completed the examination. The primary reviewer.was K. Kwok.
Attachments: 1.
R0 Written Examination and Answer Key 2.
SRO Written Examination and Answer Key 3.
Facility Comments and NRC Resolution on R0 Written Examination 4.
Facility Comments and NRC Resolution on SR0 Written Examination
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A Wac[geg /
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NUCLE.\\R REGULATORY C0MMI55;UN _ hEACTOR VEEhATOR LICEN5E EXAMINATION ' FACILITY: MIT
REACTOh TYPE: EsEEARCH DATE ADMIN 3TERED: 43/01/25 EXAMINER: BISHOF. M __
CANDIDATE l' 5IFUCTIGE5 TO CA' DIDAIEi l d Use separate paper for the answers.
Write answers on one side only.
Staple questi.'n sheet on top Of the answer shu+ts.
Foints for each question are indicated in parentneses arter tue question.
The passing grade requires at least 707 in eacn category.
Examination papers will
oc picked up six (6) hours after the examination starts.
- 4 OF CATEGORY
?; OF CANDIDATE'S CATEGORY ' 'JA LUE TOTAL 3 CORE VALUE CATEGORY 15 _._' 5 0 __15.50 A.
FRINCIFLEE OF REACTOR OPERATION 12.50 12.50 B.
FEATURES OF FACILITY DESIGN 14.50 14.50 C.
GENERAL OPERATING CHARACTERISTICS 15.00 15.00 D.
INSTRUMENT 3 AND CONTROLS 14.00 14.00 E.
SAFETY AND EMERGENCY SYSTEMS 14.50 14.50 F.
STANDARD AND EMERGENCY OPERATING FROCEDURES 14.00 14.00 G.
RADIATION CONTROL AND SAFETY 100.0 % Totals je - All work done on this examination is my own.
I have neither given ner received aid.
Candidate's 5ignature , MASER EPY ,
. . . .- . . 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 penalties.
2.
Restroom' trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date on the cover sheet of the examination (if necessary).
6.. Use only the paper provided for answers.
7.
Print your name in the upper right-hand corner of the first page of each section of the answer sheet.
8.
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 sheet.
9.
Number each answer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
' 11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. 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 required.
14. Show all calculations, methods, or assumptions used to obtain an answer , to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
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 examination.
This must be done after the examination has been completed.
_.. _ _
_ _ _ _ _ _ _ - _ _ _ l . .
. 16. When you complete your examination, you shall: As'semble your examination as follows: a.
I (1) Exam questions on top.
(2) Exam aids - figures, tables, etc.
(3) Answer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer the examination questions, c.
Turn in all scrap paper and the balance of the paper that you did not use for answering the questions, d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still , in progress, your license may be denied or revoked.
J' -
. . A.
PRfNCIPLE3 OF REACTOR OPERATION Page
. . 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 work.
. 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 FALSE.
An increasing concentration of Xe-135 reduces the thermal utilication a.
factor, f, and the multiplication factor, Keff, of the reactor core, b.
Xe-135 is produced both directly as a fission product and as the result of a decay chain from other fission products.
r , A good approximation for determining the production of Xe-135 is to c.
. assume that the Xe-135 is produced from the decay of Cs-135.
d.
The removal rate of Xe-135 is due to the neutron absorption rate in Xe-135 atoms and the radioactive decay of Xe-135 atoms.
(***** CATEGORY A CONTINUED ON NEXT PAGE *****)
_ - - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _, _ _ . _ _ _ ___ -, A.
PRINCIPLE 5 OFJEACTOR OPERATION Page
_ . . 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/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 different delta K changes.
LIST four of the possible delta K changes.
, 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 figure.
. J' = (***** END OF CATEGORY A *****) . . ... -
- __ . B[ FEATURE 3 0F_ FACILITY DESIGN Page
. . QUESTION B.01 (3.00) Refer to Figure 2, Reactor Core Tank Support, in back of test.
Identify each component labeled A through F on.he figure.
QUESTION B.02 (2.00) . Briefly explain HOW the valve design and location allows the anti-syphon valves to prevent syphoning water from the primary tank.
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: a.
What would be the physical consequence or overheating the shield? I b.
How many shield cooling regions is the shield divided into for cooling purposes? c.
What two interlocks protect the shield by not allowing reactor operation with the shield cooling system shutdown? (Setpoints not required).
J' - QUESTION B.05 (2.00) What are the four design functions of the D20 Cleanup System while it is i r.
its normal configuration? . (***** CATEGORY B CONTINUED ON NEXT PAGE +****)
-. . _ _ -. =. .. ..- 3: jf- , ..p , , 't '-B[ ' FEATURES OF-FACILITY-DESIGN pegs.
7- . . . 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 Console.
,
. . W il I i l l r i l l - - w i l l l l l L (***** END OF CATEGORY B *****) .. -.- - - - - -
i . C.' GENERAL OPERATING CHARACTERISTICS Pego
- ' . 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 answer.
QUESTION C.02 (2.50) a.
Briefly explain WHAT would be the effect on the cooling tower water 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) b.
What prevents the cooling tower basin from overflowing into the yard if the makeup valve sticks open? (0,5) c.
(TRUE or FALSE) Low secondary flow will cause a reactor scram if the reactor is operating and the low flow setpoint is reached.
(0.5) x QUESTION C.03 (3.00) a.
Briefly Explain WHY the reactivity effect of dumping the radial heavy water reflector varies with the position of the shim blades.
(1,0) , b.
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 shin.
bank Full-Inserted? c.
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 up.
- i (***** CATEGORY C CONTINUED ON NEXT PAGE *****) __ _ . _. _ . _. _ _ _ ,
_ - _. _ _. -$
g.
~C.
GENEFAL 0FERATING CHARACTERISTICS Pago
. . QUESTION C.04 (1.50) Answer each of the following TRUE or FALSE.
a.
It is possible and permissible to operate the reactor with no forced flow in the primary coolant system, b.
With the reactor at full power the pneumatic tube temperature will increase to 100 degrees C in approximately five minutes if cooling air is lost.
c.
Total thermal power output of the reactor.is the sum of Primary fewer,. Reflector Power, Shield Power, and Cooling Tower Loss.
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 power.
a.
Heavy Water System b.
Primary Coolant c.
Secondary Coolant d.
Shield Coolant l' QUESTION C.07 (2.00) Answer the following in regard to performing a reactor startup.
a.
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 test.
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 initiate.
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, a.
What has caused this circuit to activate? (1.0) b.
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 exam.
. (***** CATEGORY D CONTINUED ON NEXT PAGE *****)
. , ID. _INSTRUjENT3 AND COMTROL5 Pass 11 . . 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"? . QUESTION D.07 (2.00) Briefly explain WHY the reactor must be in "Neutron Kenetic Equilibrium" 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" condition.
. )* . - w (***** END OF CATEGORY D ****4) . - -_ - .- -.- . - -. - - - _ - _ _ - - - - --_ -...- _.
... --
E '_ SAFETY AND EMERGENCY SY3TEMi Pcg3 la . . 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: a.
When normal electrical power is lost and then regained 1 hour later, the Emergency Power MG set must be manually shutdown.
, b.
When normal electrical power is lost, the emergency-power MG set does not start for approximately 12 seconds.
Natural circulation can provide adequate core cooling if normal and c.
emergency power is lost.
d.
Primary coolant auxiliary pump, MM2 breaker cannot be shut unless normal power is available.
QUESTION E.03 (2.50) Answer the following in regard to the Containment Pressure Relief System.
a.
Describe the filters in the exhaust lines.
Include the TYPES of the filters and their positions relative to the flow stream.
(1.5) b.
What is the rated system flow? c.
If the system is lined up to relieve pressure, what determines the system flow rate? r - (***** CATEGORY E CONTINUED ON NEXT PAGE *****) . - . - _ - . _ -. - - . _ _ _ -.. _ - _ _ - _ _ _. --_
. E." SAFETY AND EMERGENCY SYSTEMS .Pago 13 . . QUESTION E.04 (3.00) a.
What four automatic actions are initiated when a major scram is initiated manually from the control console? (2.0) b.
When is the scram pushbutton on the medical therapy console operable? (0,5) c.
Is the medical therapy console scram a "minor" or "major" scram? (0.5) . GUE5 TION E.05 (2.00) a.
What is the purpose of having two vacuum breakers in each line of the reactor building negative pressure protection system? b.
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: a.
How many modes of emergency cooling are there? b.
What are the two basic criteria the system is designed to accomplish? c.
What criteria must be satisfied in order to determine the system operable per the T.S. LC07 SETPOINTS REQUIRED.
/' . 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: a.
When the reactor is in the "Non-Operating but Attended" condition the M control room must be continuously manned F( r4 , b.
The console operator and dn nreactor supervisorsmust be in the c ntro'. room during a reactor startup.
c.
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? a.
(0.5) b.
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 ' c.
your shift? (0,5) r = l t I . l l (***** CATEGORY F CONTINUED ON NEXT PAGE *****) - -
i,Il
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. F.' STANDARD'AND EMERGENCY OPERATING FR0CEDUREs Faga 15
. . - 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; ~ You are the operator at the control console with the reactor at full power.
, 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 shut.
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) >' 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 blade.
Briefly explain WHY this action is taken.
i , (***** END OF CATEGORY F *****) - - - - - - . . -
_ _ _ . ..- , ~ G.
'RADI ATIQN CONTROL AND_ SAFETY Pego 16' . . QUECTION G.01 (3.00) In regards to the Operational General Safety Rules, prior to entry, what - 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? . QUESTION G.02 (1.00)
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 a.
Monitor" alarm, where are four likely places the AR-41 originated.(2.0) b.
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 emitted.
' - i QUESTION G.05 (2.00) i 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__
_ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ - - _ - ___ U; ic . G.
RADI ATION C011 TROL At3D _SAVET,_Y_ FCgo 17
. QUESTION G.06 (2.00) Answer each of the fellowing TRUE or FALSE: .a.
Personcel with BLUE film badge holders must be escorted at ALL times while in the Reactor Building to assure no radiation exposure.
b.
Personnel with RED film badge holders must be under the direct supervision of a licensed SRO/R0 while conducting experiments involving radiation.
c.
Fersonnel with YELLOW film badge holders are all membrs of Operations or RF0 staff with current quarter exposure less than 3 Rem.
d.
Only personnel with YELLOW film badge holders may guide members of general public through the reactor building.
QUESTION G.07 (2.00) a.
What radiation detection device results become the official record of your exposure? (0,5) b.
You are assisting in a maintenance job and notice you have accumulated 60 mrem on your dosimeter.
What two actions are you required to perform as a result of this exposure? (1.0) c.
(TRUE or FALSE) Frctective clothing used within the Restricted Area can NEVER be worn outside the Restri:ted Area. (0,5) r
1 (***** END OF CATEGORY G *****) (********** END OF EXAMINATION **********) --
. . EQUATION SHEET . f = ma v = s/t - lat Cycle efficiency = "* 'I
v = ms s = v,c + s E = aC a = (vg - v,)/t
,,,v +,t A. xy 4. g,-At Et = lsuv g PE = ath a = 6/t 1 = la 2/tg = 0.693/tg , W = v&P-(g)(q) , ,, AZ = 931Aa 4*
- (tg + t ) b 6=AC,at . I. 1,e ' -rx Q = UAAT g y,m Pvr = Wgn I=I 10"* M IC)
P=P 10 TVL = 1.3/u , e /T RVL = 0.693/v t ?=P ~SUR = 26.06/7 , 7 = 1.44 DT SCR = S/(1 - K,gg) /1 o
SUR = 26 CR = S/(1 - K,ggx)
x T = (1*/o ) + [(i- ' o)/1,gg ]
eff 1 " eff 2 ~ o y. g*/ (,,, p M = 1/(1 - K,gg) = CR /CR
g g 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 [1*/TK,*gg.] + @/(1 + 1,gg )] 1* = 1 x 10" seconds T a= l 7 = I(v/(3 x 1010) ' 1,gg = 0.1 seconds A
- = ne l
Idgg=1d22 VATER PARAMETERS Id =Id
g
-
1 gal. = 8.345 lba 1/hr = (0.5 CE)/d 4,,,,,,)
1 gal.
3.78 liters R/hr = 6 ct/d gg,,,) - 1 ft = 7.48 gal.
MtsCEt.t.ANEOUS CONVERSIONS
10 Density = 62.4 lba/f t 1 Curia = 3.7 x 10 dps Density = 1 gm/cm 1*kg = 2.21 1ha
Heat of vagorization = 970 teu/lba 1 hp = 2.54 x 10 BTU /hr
Heat of fusica = 144 Btu /lba 1 N = 3.41 x 10 Btu /hr 1 Atm = 14,7 psi = 29.9 in,l's.
1 Btu = 778 ft-lbf
1 f t. H O = 0.4333 lb f /in g inch = 2.54 cm
T = 9/5"C + 32 "C = 5/9 (Or - 32) -- . -. _ _ - _ ... - -., _ -... .__._- - _ - _-. - ___. -,. _ - - _ _ _ _ _. - _. - _ _ _. -
.. -_ '
CONTROL BL A w., , (n) - - i h~ . wx 9- - , K f p%WN6'a7 _ g.. _.
W . . . . . _____ n/
j u s J/ mwwca _ lfi . l-l ' , (Ob (c) ' WESSED OFFSET . - " ' ' " '
- - r i / T N v" I ,
i j' WEleTED N / --EE:#, V.
Gul0E ROC X - - 3( , j i )i I p) r//1. v-}%, , . W.
' ' ' 1 J n ' f / / /,' y~s '_.. ' / / //f d. V;Q,v.t j .,. C ONN ECTING ARM SUIDE TUSE . CONTROL BLADE ASSEMBLY REG.R00 N ' n
. . . . r J' REG RCD E X T E N SION
m j y-G- -O g - Yfl \\
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... - - - ... ... (0) REGUL ATIN G ROD ASSEMBLY . FIG. I _ . _ _ _ ___ _____-_ _--_- - --_- - _..__- . __-_.___,
. . . . . ' . (e > f--UPPER SHIELO ACCESS / RING
- '
p . , .[ - - UPPER ANNUL AR RING ..- .., _ -y -....- l 'l l SUPPORT SLOCKS i --l S L .. =.e ! f] f, .. s: . 6_ - - - g CORE TANK (TOP SECTION) ,, e , /.' ' - E E g '
- - - . . ).Y e a%
==
j (v) '
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8 THERM AL SHIELD 1 (" OETAIL i >IL \\ . _ . /j - - .j ' , REACTOR CORE TANK SUPPORT FI. l - i . P,tm f on CIIANh...S . [ Power Supply _ .1 J.
x
. T k h c
- -
' Keathly Pico-Amp C i-1 i-i - Level Period U ' - g I _.
' b ! O I' , I Log-N
[ Recorder , . Counter ]G Scalor Select ,., . ! Lin/ Log Audio [ , , i Pro 17-y .- . l L I L, Iligh liigh
Volt Volt . Channel f2 b 'V l L - Irip Trip - , + ' t- > , s i %3 . - - - - .
'. . . I ,- NUCLEAR SCRAM (O ffgph ($) CIRCUITS ' y . MAGNET CURRENT t , AMPLIFIERS . U (s) v (s) ' ' AUTOMATIC CONTROL PERMIT ' f g - CIRCulT ,, ~ " f j t V V U ' RUN-DOWN SHIM BLADE DROP-OUT CIRCUIT CIRCUITS (6) . R , ROD i MOTOR [ ' l _ l
i BRAKE
l ,f '- FUNCTION AL PI_OCK DIAGRAM OF ABSORBER CONTROL SYSTEM FIGURE
- . A.
FRINCIFLE.3 0F 3EACTOR_ OPERATION Pcga 18 . ANCWER A,01 (2.50) Froa equation sheet: P = ? e t/T (0.5) o t/25 sec ?/P = 100 = e (0,5) o t/25 see (0,5) In 100 = t = (25 see) (In 100) (0.5)
115.13 seconds = 1.92 min 2tes (0.5) REFERENCE 1.
MIT: Glasstone and Sesonske n1T Training Program Reference), PM 1.16.2 pp. 1 AN3WER A.02 (2.00) The first is the temperature rise of the light water in the reactor core.
(0.5) Any such temperature rire will insert negative reactivity by causinc a hardening in the neutron spectrum.
(This means that the average neutron takes longer to thermalise so there are fewer fissions.
(0,5) The second phenomenon is the radiation heating of the heavy water reflector.
(0.5) Temperature rises here add negative reactivity by allowing more neutron leakage.
(This second prccess lags the temperature rise of the light water in the core proper.)
(0,5) REFERENCE MIT-RSM 10.8 /' = ANCWER A.03 (2.00) The sensitive response is due to the short neutron generation time for the MITR-II, even though its delayed neutron fraction is large (beta-bar = 0.00786).
(1.0) The large Beta effective is predominately due to a large source of "slow born" photo neutrons developed in the reflector.
(1.0) l (***** CATEGORY A CONTINUED ON NEXT PAGE *****) _
.. . A$~ PRINCIPLE 3OFREACTOROPERATION Paga 19 - , REFERENCE' MIT-RSM 10.5 ~ ANSWER A.04-( 2. 00 ) ' a.
True-b.
True c.
False d.
True [t0.5] each e REFERENCE MIT-RSM 10.6 ANSWER A.05 (3.00) (1-Keff2) / (1-Keff1) (0.5) crl/cr2
(1-Keff2) / :1 0.95) (0.5) 100/150
(10/15 x 0.05 1-Keff2
Keff2 = 0.967 (0.5) (Keff2-1) /Keff2 - (Keffi-1) /Keffi (0.5) Change in reactivity
(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) ~ . (***** CATEGORY A CONTINUED ON NEXT PAGE *****) . , , -. - -. ,w,-- ,,, - +,,,,,, - , - - , , , , --
._ A[ PRINCIPLEG_OF__ REACTOR' OPERATION-Page 20 . . ANSWER A 06-(2.00) 1.
Delta K due to temperature change 2.
Delta K due to. sample loading 3.
Delta K due to xenon 4.
Delta K due to fuel loading 5.
Delta K due to burnup (Any 4 at 0.5 each) -REFERENCE MIT-FM 3.1.1.2 page 13 . ANSWER A.07 (2.00) A.
Magnet Coupling B.
Armature C.
Blade Bottom Permanent Magnet D.
Guide Tube REFERENCE MIT-RSM Figure 1.12
~ . (***** END OF CATEGORY A *****) . , - - ,. .- - - . -... -.. . -. -. -.... - . . - - _ _,..-_ -
., _ . BI FEATURES OF FACILITY DESIGN _ Pags 21 . . ANSWER B.01 (3.00) A.
Core Shroud B.
Core Tank C.
Top Shield Lid D.
Upper Shield Ring.
E.
Lower Annular Ring F.
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 1.7 ANSWER B.03 (2.00) 1.
Bypas s booster pump 2.
Bypass Tower /' 3.
Operate C.T. fan at 1/2 speed.
4.
Vary pitch of fan blades.
5.., Restrict air admitted to tower (rearrange external boards and flaps).
(Any 4 at 0.5 each) (***** CATEGORY B CONTINUED ON NEXT PAGE *****) . . .. .. . _ _.
o . - . B.
FEATURES-OF FACILITY DESIGN page 22 . . . REFERENCE' MIT-RSM 3.10 ANSWER B.04 (2.00) a.
Melting the lead.
(0.5) b.
Four (0.5) c.
1.
Low water flow from shield (PF-1) (0.5) 2.
Low pump discharge pressure (PPS-1) (0.5) ' REFERENCE MIT-RSM 3.13 paragraph 3.5.1, Shield Coolant System ANSWER B.05 (2.00) 1.
Maintains the purity.
2.
. Maintains Reflector Tank Level.
3.
' Provider, a surge volumn (to compensate for heatups - cooldown).
4.
Provides dump volumn for emergency reactor shutdown.
(0.5 each) REFERENCE MIT-RSM Section 3.3.3, Cleanup System, page 3.7 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 out.
(0.5) Determine the location by utilizing the Leak Tape Location List.
(0.5) l I i ( ** * ** C ATEGC NY B CONTINUED ON NEXT PAGE *****) .. - - . . - - _.. . -- .-. . -... - - .
B.
FEATURE.3 OF FACILITY DESIGN pcso 23 , . REFERENCE . MIT-RSM paragraph 33.6, Leak Detection System, page 3.10
r . - (***** END OF CATEGORY B *****) . - - -
+ . C.' GENERAL OPERATING CHARACTERISTICS Pago 24 . . 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 reflector.
(0.5) REFERENCE MIT-RSM Section 6.3.4 ' ANSWER C.02 (2.50) a.
The dissolved solids in the makeup water (0.5) would be concentrated (0.5) due to evaporation out of the tower (0.5) b.
An overflow stand pipe drains to the sewer.
(0.5) c.
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 a.
water reflector (0.5) thereby exerting a shadowing influence on the reflector.
(0.5) l b.
A dump with the bank fully inserted is worth about two thirds.the worth of a dump with the bank at Top of Active Core.
(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 critical.
(0.5) (***** CATEGORY C CONTINUED ON NEXT PAGE *****)., . _ _ _ _ - -.- . -. -. . - _. _ - - _. ..
- . C.
GENERAL OFEFATING_ CHARACTERISTICS Page 25 . REFERENCE' MIT-RSM-10.6 . . ANSWER C.04 (1.50) a.
True b.
True c.
False REFERENCE MIT-FM-2.2, FM-5.5.1, FM-2.4 page 5 ANSWER C.05 (2.00) a.
5.5 b.
6.9 c.
5.5 d.
6.5 (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) r.
. a.
140 gpm (+ or - 10 gpm)
b.
2050 gpm (+ or - 50 gpm) c.
900 gpm 4_ (+ or - 50 gp g,,g y jg,o y 4f d.
90 - 110 gpm ' (0.5 each)
(***** CATEGORY C CONTINUED ON NEXT PAGE *****) . - - -. .. . __.
,- .- .. _ _- .
. - . ' . C.
GENERAL OPERATING CHARACTERISTICS Page 26 ' . REFERENCE' MIT-FM-3.1.1.2_page 14 of 15 . ANSWER C.07 (2.00) a.
1.
Neutron level (0.5) will be increasing (0.5).
2.
Reactor period (0.5) will exhibit shorter transient periods.
(0.5) gg,r u.yr, S a,,..;- o -P th~!r 1&+~ d sy' -,L p,;.s g.
ki; ti * ts * - Ol/ " ' REFERENCE MIT-FM 2.3, Reactor 5afety Procedures, page 2 of 7 r w - (***** END OF CATEGORY C *****) . _.. _... _ - _.. -- - . - -,. _ -. _ _ _ _.. _. _. - , - - _ - -.. _ _ _ -, _. _..
- . D INSTRUMENTS AND CONTROLS Page 27 . . . ANSWER-D.01 (2.00) 1.
Neutron flux is several orders of magnitude greater than gamma flux.
2.
Gamma flux is proportional to fission rate and reactor power.
REFERENCE MTI RSM-5.4 Paragraph 5.2.3, Uncompensated Ion Chambers ANSWER D.02 (2.00) A.
Mag. Amp.
B.
Scram Amp.
C.
Off Scale Trip D.
Period Network E.
Pulse to D.C. Converter F.
Lescriminator Amplifier G.
Fietio } Chamber E.
Ion Chamber _ REFERENCE MIT RSM Figure 5.1, Period Channels ANSWER D.03 (2.00) 1.
ALL shim blades must be above suberitical interlock position.
2.
The power-set / actual power deviation must not exceed 1.5%. 3.
The regulating rod control switch must be in neutral position.
4.
The regulating rod must be withdrawn beyond its neaf-in position l (1.6 inches).
(0.5 each) l l
(***** CATEGORY D CONTINUED ON NEXT PAGE *****) --. .- - - - . .. . - - - - -
. _ - . EI.
INSTRUMENTS AND CONTROL 3.
Fago 28- . . REFERENCE MIT-RSM-4.4 ANSWER D.04 (3.00) a.
The regulating rod has been at its near-in position (0.5) (1.6 inches) for thirty seconds?yr (0,5) vM B. Hr.,:- mtw+ a %I, b.
1.
Red light comes on.
2.
busser sounds 3.
Reactor control shifts to manual (after 30 seconds).
,_ 4.
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) A.
Withdraw Permit Circuit B.
Shim Blade Magnets C.
Automatic Run-Down Circuit D.
Start-Up Interlocks E.
Suboritical Interlock & Override Permit Circuit ' F.
Regulating Rod Control Circuit G.
All Absorbers in Circuit H.
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 *****) , ,_ _ _ - . _ _. _ _,., _ .. , _ _ _ _ _7.. _ _ t . . ID. _I_NSTRUMENTS AND_ CONTROLS Pago 29 . . ANSWER D.06 (2.00) 1.
It'is the basis for calibrating the neutron level channels.
2.
It is the basis for setting the safety channel level trip points.
(1.0 each) REFERENCE MIT RSM - Item 6.3.4 page 6.5 . 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
i /* - l l l (***** END OF CATEGORY D *****) l -. .. _.. _ _ _ _ _ - _ _. ._ _, .__ _ _ _. - _ _., . _. _., _. _ _
. . ~ E.
3AFETY AND EMERGENCY SYSTEMS Pago 30 . . ANSWER E.01 (3.00) Safety Amps Recorders T.V.
Rad Monitors Clock Intercom System Servo Unit Front Panel Outlets Ann. Panel Indicators-Du O N W 7.
Rod Control 'Med. Rm. Recpt.
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) a.
False b.
False c.
True d.
False REFERENCE MIT RSM paragraph 8.8.2 Emergency Room Dist. System page 8.37 & 8.38 ANSWER E.03 (2.50) r a.
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).
b.
800 cfm (0.5) c.
Building / atmosphere delta pressure (0.5): (***** CATEGORY E CONTINUED ON NEXT PAGE *****) _ _ - - --. _., .. . .-. _ _ - . - . . -. _.. - -
s - if.
SAFF._TY_AND__ EMERGENCY SYSTEMS Pago 31 . REFERENCE' MIT-RSM-paragraph 8.4, Containment Pressure Relief System page 8.23 and Figure 8.8.
ANSWER E.04 (3.00) a.
1.
Ventilation System Secured.
2.
Containment Shell is Sealed.
3.
D20 Reflector is Dumped.
4.
Withdraw permit circuit deenergized (shim blades drop).
. (0.5 each) b.
When the Medical Therapy's console key switch is ON.(0,5) c.
Minor (0.5) REFERENCE MIT RSM paragraph 9.3 page 9.8 ANSWER E.05 (2.00) a.
Prevent leakage out of the building in the event of internal overpressure.
(1.0) b.
2.0 psig greater than atmospheric - positive (0.5) 0.1 psig less than atmospheric - negative (0.5) l' REFERENCE MIt RSM paragraph 8.1 page 8.1 . l l l . (***** CATEGORY E CONTINUED ON'NEXT PAGE *****) t t __. _ _ _
. - .- - -- . , . E ~. SAFETY AND EMERGENCY SYSTEM 5 Paga 32 , . - . AN5WER E.06-( 1. 50 ) a.
(0,5) b.
Ke'ep core covered (0.25) and core temp < boiling (0.25).
c.
10 gpm (0.25) within 5 minutes of activation (0.25).
REFERENCE MIT-RSM paragraph 3.2.7, Emergency Cooling page 3.4 . , l' w e i . (***** END OF CATEGORY E *****) . - - - - - --.. - _.. - _, - _.. _. _.. -... _.. - _.. _. _ _.-, _ _ _...... - _.. _ _ _,. _,. _ _.. - - _
't _ [. STANDARD AND EMERGENCY OPERATING _ PROCEDURES Page 33 F
. ANSWER F.01 (3.00) 1.
Reactor Shutdown 2.
Console Key Switch "OFF".
3.
Console Key Switch Removed and in Proper Custody.
4.
No work in progress within main core tank involving fuel or experiments.
5.
No maintenance of the core structure.
6.
No maintenance of installed control blades.
7.
No maintenance of control blade drives unless visibly decoupled from the control blade.
(Any 6 @ 0.5 each) , REFERENCE MIT-FM 2.2 page 3 of 11 ANSWER F.02 (1.50) R.hu (hr. lbA bd N)') a.
False b.
7 :m' c.
False REFERENCE MIT PM 2.2 pp. 3 of 11 Item 16, PM 2.3 page 1 of " ' tem 2.3.1.3 J' ANSWER,, F.03 (2.50) a.
Yes (0,5) b.
Reactor Superintendent, Duty Sbift Supervisor, and Electronics <,/.tM N. ( A en um -+u i Supervisorg e (O.5 each) ,. i File it in the front or the reactor console log.
(0.5) c.
, (***** CATEGORY F CONTINUED ON NEXT PAGE *****) ! - _ _. - _.. - -.. _. . _, _.
_ . -. . _ _,.
. - _ [.--STANDARDANDEMERGENCYOPERATINGPROCEDURES Paga'34 . REFERENCE' MIT FM 1.9 Paragraph 1.9.2 page 1 of 2 ANSWER F.04 (2.00) 1.
An emergency where_such action is needed immediately to protect the public health and safety.
2.
No action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent.
3.
Must be approved by licensed SRO prior to taking the action.
.- 4.
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) 1.
Acknowledge the alarm.
2.
Scram the reactor (minor) if not already scrammed.
3.
Verify reactor power decreasing.
4.
Notify reactor shift supervisor.
(Any 3 @ 0.5 each) REFERENCE d' MIT AOP 5.1.2, Withdraw Permit Circuit Open - (**e*e CATEGORY F CONTINUED ON NEXT PAGE ***'<*)
. . F.
STANDARD AND EMERGENCY OPERATING PROCEDURES Page 35 ~ . . 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 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 av 4 sa S~ rw GY7 e A/3 4 (t.o) S.ud / L A motor would burn out.(1.0) op fr 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' - (***** END OF CATEGORY F <*e**) , . - . - -. _ - -..-. . -
. G'. RADIATION CONTROL AND SAFETY Pcgs 36 . . . AN5WER G.01 (3.00) 1.
To determine that normal radiation levels exist based on control room and/or local instrumentation.
2.
To assess the need for a radiation survey with a portable detector.
3.
To evaluate the potential for dose rate changes during occupancy.
(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) a.
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 air.
(0.5) (***** CATEGORY G CONTINUED ON NEXT PAGE *****)
, - G: RADIATION CONTROL AND SAFETY Pass'37 . 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 , AN5WER G.05 (2.00) 1.
Beta-Gamma Monitoring Badge 2.
Pocket Dosimeter (gamma) (1.0 ea) REFERENCE MIT-FM 2.5, page 1 ANSWER G.06 (2.00) a.
False , b.
False l' c.
False d.
True (0.5 ea) -
(***** CATEGORY G CONTINUED ON NEXT PAGE *****)
.. . . . . . " G.
RADIATION CONTROL'AND SAFETY Pasa 38 . . REFERENCE' MIT-FM 1.12, Film Badge Classification page 1 of 5 ' ANSWER G.07 (2.00) a.
Film Badge (0.5) b.
1.
Record the exposure on the posted dosimeter sheet.
(0,5) 2.
Re-sero the dosimeter.
(0.5) c.
False (0.5) ' REFERENCE MIT-FM 1.12 page 5 of 5 of June 8, 1983 (Memo Karaian to Rad Workers and Experimenters . ,.. m , (***** END OF CATEGORY G *****) (********** END OF EXAMINATION **********) . - . .. . - -. - . - - - - - _.
-.
. kf7AFeMMwr k - U. 3.
NUCLEAR REGULATORY COMMISSICH l JEliiUR EEACTOR OPERATOR LICEUdE EXAMINATICN FACILITY: Maid. INSTITUTE CF TECH.
' . REACTOR TYPE: RESEARCH __ i , DATE ADMINSTERED: 88/
- EXAMINER: ROESENER, S.
CANDIDATE $4J7ca INSTRUCTIONS TO CANDIDATE: ' - _ = Use separate paper for the answers.
Write answers on one side only.
- Staple queetion sheet on top of the answer sheets.
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 starts.
% OF JATEGORY 4 0F CANDIDATE'S CATEGCRY VALUE TOTAL SCORE VALUE CATEGORY . 20.00 20.00 H.
REACTOR THEORY 19.0 0 C.C'7 20.00 I.
RADIOACTIVE MATERIALS HANDLING DISPCSAL AND HAZARDS 20.00 20.00 J.
SFECIFIC OPERATING CHARACTERISTICS 20.00 20.00 K.
FUEL HANDLING AND CORE PARAMETERS j.
I9.00 - L.
ADMINISTRATIVE PROCEDURES, {." 7 20.00 ' CONDITIONS AND LIMITATIONS 1f.oo . ' L.v 7 % Totals . Final Grade
, All work done on this examination is my own.
I have neither given l ncr received aid.
I.oy O. e p )) J Candidate *s Signature g,pg j g, g / n 'f. / e9* ' . _ _ _
. . . NRC RULES AND GUZDEL1NES FOR LICENSE EXAMINATIONS - Turing tne acministration of this examination the following rules apply: , 1.
Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
2.
Restroom trips are to be limited and only one candidate at a time may leave.
You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of cheating.
3.
Use black ink or dark pencil only to facilitate legible reproductions.
4.
Print your name in the blank provided on the cover sheet of the examination.
5.
Fill in the date nn the cover sheet of the examination (if necessary).
6.
Use only the paper r ovided for answers.
Print ycur name in the upper right-hand corner of the first page of each , section of the answer sheet.
8.
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 sheet.
- S.
Number each :,nswer as to category and number, for example, 1.4, 6.3.
10. Skip at least three lines between each answer.
11. Separate answer sheets from pad and place finished answer sheets face down on your desk or table.
12. Use abbreviations only if they are commonly used in facility literature.
13. 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 required.
14. Show all calculations, methods, or assumptions used to obtain an answer to mathematical problems whether indicated in the question or not.
15. Partial credit may be given.
Therefore, ANSWER ALL PARTS OF THE QUESTION AND DO NOT LEAVE ANY ANSWER BLANK.
16. If parts of the examination are not clear as to intent, ask questions of the examiner only.
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 completed.
_ -. , __
.. . . . . 13. When you complete your examination, you shall: a.
Assemble your examination'as follows: (1) -Ex&m questions on top.
(2) Exam aids - figures, tables, etc.
(3) hnswer pages including figures which are part of the answer.
b.
Turn in your copy of the examination and all pages used to answer
the examination questions, c.
Turn in all scrap paper ar.d the balance of the paper that you did r.ot use for answering the questions.
, d.
Leave the examination area, as defined by the examiner.
If after leaving, you are found in this area while the examination is still in pregress, your license may ce denied or revoked.
-
. _ _ - __ _ _ - _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ ._.
__ _ _ _ _ _ _ _- - __ _ - _ _. . - _ _ .. . !!. RE ACTUfL T!!E'.lRY Pago
- __ _______________ . . GUEGTION H.01 (2.00; a.
Define EFFECTIVE Delayed Neutron Fraction.
(1.00) b.
Hcw would the Effective Delayed Neutron Fraction of the MITR-II change (INCREASE, DECREASE, REMAIN THE SAME) if the deuterima.
reflector were replaced by a light water reflector? Briefly explain your answer.
(1.00; uGE3 TION H.00 (3.00)
a.
What are the three parameters needed to calculate the power being transferred from the fuel to the primary coolant? (1.00) t.
To obtal:. the total power output of tae reactor what two other coolant str ans must be evaluated.' (1.00.
c.
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 assumptions.
r . . ($$'** CATEGGRY H CONTINUED CN NEXT PAGE A****) - - -_-. _.
-
. . . . 11.
REACTOR THEQRY Page
---. -- __
- .
. . 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 channel.
Im' ediately af ter withdrawing this blade he reads a count of '80 cps.
m a.
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 cps.
(1.50) b.
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 subcritical.
Is the change in count rate 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 answer.
(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 ' 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? . 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 worth.
(The differential worth curves are included as an attachment to aid in the visuali ation "" of this question.)
. (o*** CATE1 CRY H C0tJTINUED ON SEXT PAGE ** w*)
. . . . H.
. REACTOR THEORY Page
QUESTION H 08 (3.00) a.
How long after a reactor shutdown.does the xenon concentration peak? Assume one week of full power operation prior to shutdown.
-(1.00) b.
Explain in terms of the two production and two removal processes why the xenon buildup occurs following the shutdown.
(2.00) QUESTION H.09 (2.00) ' Assuming the reactor to be just critical at 100 kW: a.
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 calculation.
(1.00) b.
Describe the behavior of reactor power at two minutes following the ten inch insertion.
(1.00) . J' = , -> (*=**< END OF CATEGORY H r r**) _, _ _.. _ _. _ _ _ . __
l . , ' . 1.
RADIOACTIVE MATERIAL 5~ HANDLING DISPOSAL Pago 7 ' 'AND HA"ARPS , . QUESTION I.01 (2.00) a.
When the reactor is in normal operation, how often should the secondary water be sampled for Tritium? (0.50)
b.
What three actions must be taken if-secondary water tritium concentrations exceed one microcurie / liter? (1.50) QUESTICH I.02 (2.00) a.
In the event of an MITR-II radiological emergency, who is
responsible for decisions and coordination of all immediate actions? (1.00) - b.
What two actions should always be taken to maximize emergency plan effectiveness.
- (1.00) GLESTION I.03 (3.00) a.
According to procedure PM-4.7.2 the EAL for a General Emergency can be determined by using the following formulae: 1. 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 b.
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 c.
determine if a General Emergency should be inacted? (1.00) , a (***** CATEGORY I CCNTINUED ON NEXT PAGE *****) d . . -, -, , , - - - -.-- - -. -, - ..,,.
. . ... ' 1'. 'RADIGACTIVE MATERIALS HANDLING DISPCSAL Pega
- ___
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: a.
gamma radiat n.
Dels r ed b.
beta radi-ion.
c.
surfac contamination, , d.
alp a contamination.
, GUESTION I.05 (1.50) List THREE independent measurements or indications use to monitor or detect heavy water leakage into the seconcary coolant.
. 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" alarm.
GUESTICU I.03 (1.00) 5 tate the purpose of the Core Purge (Off-Gas) System.
(***** CATEGCRY I CONTINUED ON NEXT PAGE *****) .
, _ _ .__ __ .... ... . .7 !. RADIOACTIVE. MATERIALS HANDLING DISPOSAL Pag 6
AND HAZARDS - . QUESTION I.10 (2.50) s.
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) b.
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 room.
(1.50) . QUISTION I 11 (1.50) a.
What TWO individuals should jointly authorine emergency exposures wnich may exceed 10CFR20 limits? (1.00) b.
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) 3.
What is the radiological concern associated with continued reactor power operations during a containment isolation? (1.00) b.
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 - (***=x END OF CATEGCRY I *****)
.. - . . , J.
3?ECIFIC CPERATING CHARACTERIGTICG .Page 10 _ _ - - - - - - - - - - - - - - -
.--- , < . . QUESTION J.01 (4.00) Regarding the Convection and Anti-Syphon valves within the core tank: , a.
Give the purpose of each of these valve types.
(1.00) b.
Describe how each valve type functions to carry out its purpose.
Include in your explanation a description of the valve's perfor.mance in BOTH the normal AND the accident positions.
(3.00)- (A simplified sketch demonstrating these points is acceptable.)
'
CUESTION J.00 (1.50) Th;lE or FALdE? a.
The main core tank level indicator ML-3A is an electrically driven transmitter which is supplied by emergency power.
(0.50) b.
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 room.
(0.50) c.
A reactor scram on low main tank water level will occur only if both ML-3 anc ML-2 level probes are uncovered.
(0.50) QUESTION J.03 (2.00? a.
When would it be necessary to utilise the containment pressure relief system? ( 1. 0 0.' b.
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 . . (=*e** CATEGORY J CONTINUED ON NEXT PAGE ***i*) , _ ~ - _
.. '. .- J.
tPECI71C OPERATtNG CHAF.ACTER8STZC3-Page 11 _ _ _ _ _ _ _ __ . . 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 made.
(1.00) .' b.
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 hours.
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 test.
Show all your work.
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 position.
~4 hen the automatic rundown light energizes the operator notices it.
a.
What should be his THREE immediate actions if he"desires to continue power operations? (1.50) b.
How long after the light energi:es will a rod start to move? "' (0.50) c.
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) . (***** CATEGCRY J CONTINUED CN NEXT PAGE Cr***) .. . } ~ JPE.C'FIC CPERATI:lG CHARACTER:3 TICS Pega ic s. _ _ _. _ - = _ _ -- _ - - - i . . 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."
. -. I . . . . > ,. , __ J . , f I (***** END OF CATEGOP.Y J **= *)
_ _._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
. < X... FUEL HANDLING AND CORE PARAMETERS Page-13 - -- 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, a.
What are the two Technical Specifications? (2.00) b.
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 a.
unen calculating shutdown margins? (1.50) - . b.
Explain what variable reactivity is AND TWO factors which cause it to occur.
(1,50) ' (*'*** CATEGORY K CONTINUED CN NEXT PAGE -++***)
_ _ _ _ _ _ _ _ _ 'i . . K.
. 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) a.
According to MITR Technical Specifications, approximately how many MITR fuel elements would be required for criticality assuming
optimum conditions? (0.50)
b.
Anat is the maximum numoer or fuel elements allowed by Technical Specifications to be cutside designated, storage areas at any one time? (0.50.1 . QUESTION K.09 (2.00) When loading a new 445 gram U-235 fuel element during refueling, ' a.
which ring (A,B or C) will give you the highest m0/ gram of U-2357 (0.50; b.
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 element.
Briefly explain your answer.
(1.50) r (**r*< END OF CATEGORY K * " '* * ) _ _ _ _ _ _ _ _ _ _ _. __
. . . , . L.
ADMINISTRATIVE PROCEDURES CONDITIONS Pago 16 - AND LIMITATICNS - . QUESTIOJ L.01-(1.50) a.
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) .. b.
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 a.
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 present.
(0.501 b.
The shift supervisor may grant permission to an experimenter to irradiate acids or other corrosive liquids.
(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) a.
Whose permission is required to post a warning tag on facility equipment and who may post a warning tad? (1.00) ' b.
List FIVE requirements which must be observed when "locking out" facility equipment.
J' (3.00) . . (***** CATEGORY L CONTINUED ON NEXT PAGE *****) _, - ' , ' [L.
ADMINISTRATIVE PROCEDURES, CONDITIONS Paga 16 .~ AND LIMITATICNS ~ . QUESTION L.05 (2.00) In accordance with MITR-II Technical Specification 7.8.3 and a.. Administrative Procedure. PM 1.5, "Procedures Adherence 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) b.
Who must approve the temporary change discussed in "a" above? ~(Give level of qualification.)
(1.00) e QUESTION L.06 (3.00) a.
What are the FOUR interrelated variables associated with the core thermal and hydraulic performance on which Safety Limits are based? - (2.00) b.
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; a.
Who must normally authorize re-entry into any portion of the reactor facility? (0.50) , b.
Under what circumstances can the on-duty shift supervisor authorice ! re-entry? (1.00) ! r , l - i i l \\ l (***** CATEGORY L CONTINUED ON NEXT PAGE *****) l L
-- , - . ~* L.
ADMINIGTRATIVE PF0CEDURES.-CCNDITIONS pogo 17 --- =_--__ = __ 'AND LIM' TAT:0NG , - . QUESTION L.08 '(3.00) Ut.ilise the attached EAL's to answer the following: a.
For each of the events below, state the minimum emergency , classification which may be declared.
(0.50 each) 1. A large crowd of protesters marching around the reactor building.
2. A fire damaging-an experiment which causes the release of radio-active materials, b 3. A tornado damaging the containment building.
4.
A slow and uncontrollable decrease in core tank level such that level remains above the anti-syphon valves, b.
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 who manipulates a control a of a facility: a.
Define the term "control" as it applies to MITR-II.
(1.00) b.
Under what conditions is the person physically manipulating a control not required to hold a valid cperator's license? (1.00; . e r m . (=**** END OF CATEGCRY L *****) (*===****** END OF EXAMINATION **********) _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ -
. . . H.
EEACTOR THEORY Page 18 ANSWER H.0) (2.00) a.
The fraction of the thermal neutron population that was born delayed.
(1.00) b.
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 water.
(1.00) ' REFERENCE.
Reactor Physics Notes, Reactor Kinetics and Control Rod Calibration by Reactor Period Measurement, p.
3.
Reactor Systems Manual, Ch. 10, Sec. 10.3.
, , ' ANSWER H.02 (3.00) a.
1.
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) b.
1. Deuterium Tank coolant system.[0.50) 2. Shield coolant system.[0.50) (1.00) c.
Because the graphite reflector has a large heat capacity.
(1.00) REFERENCE l' ' ,, Reactor Systems Manual, Ch. 6 Sec.
6.3.4.
. . (***** CATEGORY H CONTINUED ON NEXT PAGE *****) - - _ _ . .. . --. . . _, . -. - ., ,.
. . , 'h e-H.
REACTOR TiiEORY Page 19 . ANSWER.
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) =, h m2 (+/- Change in reactivity --> h m0 - (-75 m3) 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/ ,J~h) Yin. [0.33).
(Must be consistent Therefore the rod would move out within .6 in. casea on reactivity insertion calculated for temperature change.)
(2.00) REFERENCE . Reactor Systems Manual, Ch. 10, Sec. 10.7 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) a.
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) b.
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 increases.
(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.)
. (***** CATEGORY H CCNTINUED ON NEXT PAGE *****) .-... -- _ -. - . _ _ _ -.
. -. <- ' . H.
. REACTOR THEORY Page 20 . REFERENCE Reactor Physics Notes, Reactor Startup and Reactor Suberitical ' Multiplication, pp. 7 - 13.
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 12.
. 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.
7.
J' -
- . ANSWER H.07 (2.00) - 1.
The full-in position of the reg rod is six inches above the bottom of the core.
(1.00) 2.
The reg rod is heavily shadowed by the shim bank in its upper region of withdrawal.
(1.00) , (***** CATEGORY H CONTINUED ON NEXT PAGE *****) _________________________ _ -
- _ _ _ _ _ _ _ _ _. - - - - - - - - _ - - _ _ _ _ _ _ _ _ _ _ _ _ , . > H.
SEACTOR THEORY Page 21 REFERENCE Reactor Systems Manual, Ch. 10, Sec. 10.4.
T ANSWER H.08 (3.00) a.
6 hours.
(1.00) ' b.
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. 10.8.
. - - ANSWER H.09 (2.00) a.
The reactor power will drop immediately [0.50] due to the quick response of the prompt neutrons to the change in reactivity [0.50]. (1.00) i 6.
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) . ,. REFERENCE ' Reactor Physics Notes, Reactor Kinetics and Control Rod Calibration by Reactor Period Measurement, p.
11.
_ For a nore's $ in ser r E o ' 0 $ InCSCC;7"bC )*CWCf 'Y'N b CJ ' TC n YU f dwn im m edig rely [gSO] duc To Thc, yui tk rgsfcnSe o$ (b T*$ c f ro 7~ n es;r ron.S 7~o The. r e nf C, h ts t'] t if) rc o t TIvir~j G.7&, (***** END OF CATEGORY H *****)
' . . . I.
. RADIOACTIVE MATERIALS IIANDLING DISPCSAL Paco 22 ~ 'AND HA5ARDS- - . . ANSWER I.01 (2.00) a.- At least once every 24 hours.
(0.50) . . b.
1.
The cooling Tower Spray must be secured (0.50), 2.
Any discharge to the sanitary sewer of secondary water must be stopped (0.50).
3.
The heavy water reflector heat exchanger must be isolated (0.50) " (1.50) ' REFERENCE Technical Specification 3.8.
ANSWER 1.02 (2.00) a.
The senior NRC-licensed member on shift.
(1.00) , b.
1.
Shutdown the reactor.
2.
Isolate the containment.
(1.00) REFERENCE Emergency Plan and Procedures, Sec.
4.3.1.2.1.
. r . . t . (***** CATEGORY I CONTINUED ON NEXT PAGE *****) , c ._- , _ _., . . - - - _ - _. - ..,y ,,, _.
._, , -- -
. . .
RADIOACTIVE MATERIALS HANDLING DISPOSAL Pags 23 AND HAZARDS - .- ANSWER I.03 (3.00) a.
Because the dilution for stack releases is caluculated to be less yhan the dilution for containment releases.
(1.00) b.
The EALs are based on the assumption that the most limiting MFC is being released.
(1.00) c.
Any four of the following at 0.25 each: 1.
Plenum gas monitor.
' 2.
Plenum particulate monitor.
3. Stack gas monitor.
4.
Stack particulate monitor.
5. Stack area monitor.
(1.00) g, t, e r< de. m e. ;, e s < * oire, b e d< y.
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, rse Abtre. A < s necen) coa 3 rhe f *I* r *f *p oin r & co e e en C. ke << rke
);sen,ce tw<y from
- )'in u m ce * ec * rr< !'!* *
- d feb r~ of conc er n e r t.
sI.ers es 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) 1.
Secondary water radiation monitoring.
2.
Daily sampling and analysis of the secondary water (for tritium).
3.
Monitoring of level in deu.terium dump tank (by a low level alarm or (1.50) by site glass).
(***** CATEGORY I CONTINUED ON NEXT PAGE *****) . - _..__
, _ - - -. ,, ,
1.
RADICACTI'/E MATERIALS hat 4DLING DISPOSAL Pego 24 - * AND HA'.:ARDS . . REFERENCE Technical Specification 3.8.
. % ANSWER I.06 (1.00) Because tne secondary radiation monitors are not capable of detecting any primary to secondary leakage when the reactor is shutdown.
REFERENCE
Reactor Systems Manual. Ch. 7 Sec.
7.4.1.
- ANSWER I.07 (1.00) 1.
The "High Level Radiation Moaitor" alarm actuates.
(Exact name of alarm is not required for full credit.)
(0.50) 2.
The sewer pump (RM-3) stops.
(0.50) REFERENCE Operating Procedures Checklists. PM 3.6.
"Waste Storage Tank Dump Procedure'. p.
2.
oje#e4/liry af rke.
Re unr-c he ny e.
sefraceaGea. h <.s rh e.
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 brers.
ANSWER I.08 (1.50) 1.
Closure of the intake valve (MV-83).
j.
2.
Closure of the core purge blo"er suction valve (the primary storage tank air discharge valve) (MV-64).
- 3.
The core purge blower trips off.
(1.50) (***** CATEGCRY I CONTINUED ON NEXT PAGE **=**)
.. . . . L.
F.ADI0 ACTIVE MATERIALS HANDLING DISPCSAL Pega 20 AND HAZAhDS . . REFERENCE Abnormal Operating Procedures, PM 5.6.5, "High Radiation Level Core Purge", p.
1.
Reactor Systems Manual Ch. 7 Sec.
7.7.
. ANSWER I.09 (1.00) (e y one ofrks kilm'<3,1-.. )vf fe e /.00 : The purpose of the core purge system is to pin e.a O.s a r.c '. y-;-; ;; > 7,; i- + w.-- t.n<a * va +b.
- 4 m.
t.
n.- w' , , t.
Per << < r Ardr*3 e n >~ ild vf.
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.
3.2.5.
Te d ni csl Speco*hcorim 3.'f y ANSWER I.10 (2.50) a.
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) b.
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. 7.6.
. (***** CATEGCRY I CONTINUED ON NEXT PAGE *****)
. . ' ~ RADIOACTIVE-MATERIALS HANDLING DISPOSAL Pcgo 26 ' I, ~~~~ AND HAZARDS . . . . ANSWER I.11 (1.50)
a.
The Director of Reactor Operations [0.50) and the MITR Radiation Protection Officer (0.50).
(1.00) b.
The Se'nior licensed member of the NRL Staff on-site.
(0.50) REFERENCE Emergency Plan and Procedures, Sec.
4.3.3.2.
. ANSWER I.12 (2.00) a.
Buildup of Ar-41 in the containment building.
(1.00) b.
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".
m
r . w ! (***** END OF CATEGORY I *****) L
.s . .. . . . J.
SPECIFIC OPERATING CHARACTERISTICS Pago 27 __ __ _ . . 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) . b.
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 points.
(3.00) REFERENCE Reactor Systems Manual. Ch.
1. Sec. 1.5.3; Ch.
3, Sec 3.2.7.
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) . . (***** CATEGORY J CONTINUED ON NEXT PAGE *****) .. ..
_ _ _ - _ _ _ _ _ _ _ _ _.
. . ..- . J.
. SPECIFIC OPERATING CHARACTERISTICS Page 28 ^ REFERENCE MITR RSM 6.8 Re' actor Systems Manual Ch. 6. Sec.
6.5.1.
ANSWER J.03 (2.00) a.
If building pressure exceeds 2 psig, and if radiation levels , and/or structural damage preclude opening the main or auxiliary dampers.
(1.00) 6.
G a-o a.r c4.r m. s a c. = .n 2 <.:. c.c ^ . v.'.. '. s.. .. .m....-. .#i ., .v e... u an. vase cae enarwa. i.b.
s..X 'A=y r.,+ r e s fe //o s. :<3""<*r-c. ro c., s 4.* s REFERENCE '-
- " # 8 " # * '
<* a r'* /a a' a r M M * * a * -a / /** 4 re e r.
1.
e In s =l= s e. k s. e. < / fitre a.. Reactor Syst.' ems I[ua^), #dn. 'b',"Te# ^ c.
o74.
- Abnormal Operating Procedures, FM 5.5.7 "Building Overpressure.
- . 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 11.
r AllSWER J.05 (2.00) a.
The transfer is mcie by adjusting (reducing) the gain.
(1.00) (* NOTE: Half credit is lost if the candidate includes adjusting the discriminator as this is no longer done.)
b.
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 *****)
. .- . J.
4PECIFIC OPERATING CHARACTERISTICS Page 29 ' . REFERENCE Standard Operating Plans, FM 2.3.1 "Normal Reactor Startup". Step 15.
Reactor Systems Manual. Ch. 5. Sec. 5.3.1.
~ ., '
- Facility Assistant Superintendent's Explanation-during plant tour.
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 3.
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".
(***** CATEGORY J CONTINUED ON NEXT FAGE *****)
. ., ' J.- SPECIFIC OPERATING CHARACTERISTICS Pego 30 ---_ .
. . ANSWER J.08 (3.00) a.
1. Depress the rundown reset button.
2. Place the control rods on manual.
3. Reshim.
(1.50) - ., b.
30 seconds.
(0.50; - c.
When the rod being driven in is no longer within two inches of the the bank height.
(1.00; . , REFERENCE , Standard Operating Plans, FM 2.4.1 "Full Power Operation", Step 4 b.
A:'S*ER J 09 (3.00) Any five of the following at 0.60 each: , 1.
Check that the MG set is running [0.30) and adjust its output ' voltage to 208 volts (0.30).
. 2.
Notify the MIT Campus Police and Physical Plant of the extent of the ' power outage.
. 3.
If possible, notify the occupants of NW-12 that the vault alarm was spurious.
4.
Check the personnel accountability board and determine how many experimenters etc. are inside the containment building.
5.
Bypass secondary coolant flow to the cooling tower basins.
' 6.
If necessary, open CV-33 to supply pressurized aCr to the. main ' personnel lock gaskets.
'7.
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 lock.
(3.00) $. VetESy t f e.
co n s*le.
ofererer-h s.: ew. e l e r-J,is ' im m edi < rt.
A s.r i.,s.
. (***** CATEGORY J CONTINUED ON NEXT PAGE ***r*) _ _ _ - - .. - - . _ _. _ _ . _ - - _ . -_ .
' , . .,
, , l
J.
SPECIFIC OPERATING CHARACTERISTICS Page31! -- REFERENCE Abnormal Operating Procedures PM 5.8.1 "Loss of Normal Electrical - Power", pp. 1 & 2.
, A eaernsi epresrien feeseduma, fM S 0.
r b , t I t h I i )* w.
> ! Y . , l i ! ! (***** END OF CATIGORY J n***) ! , . - -. - . -. -. . -. -. - -... - -. -. . - -. - - _ -... -. - - - _ -... - -
_ _ _ _ _ _ _ _. _ _ _ _ s . K.
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 samples.
Increasing primary conductivity (or alarm).
7.ncreasing equipment room monitors.
Increasing containment building area monitors.
(2.00) , REFERENCE Abnormal Operating Procedures, PM 5.8.2 "Fission Product Detection in the Primary Coolant", p.
1.
- ANSWER K.02 (2.00)
1.
Loss of power to the SFSF control and alarm panel.
2.
Leak 3.
Low SFSP level.
4.
Low flow through the SFSP ion exchanger.
(2.00) REFERENCE Abnormal Operating Procedures, PM 5.7.12 "Spent Fuel Storage Pool", p.
1.
)* , w ANSWER K.03 (3.00) a.
1.
The ratio F F /d F is predicted to be less than 2.9 (1.00]. HC p ff 2.
The core is predicted to operate below incipient boiling at every point in the core [1.00).
(2.00) b.
Two Senior Reactor Operators.
(1,00; e rh a.
As s ie r < s r~ S.ye r b reede * r' ^ <j 6 a ( The. s.feein ts-4.r <.Jfe or a m ed din srl s in e a.
r h ey a, <a.
rha m. o r.
x </ < /tce.,,; 3,rg, ) y (***'* CATEGORY K CONTINUED ON NEXT PAGE *****) .. - - - . . - . .
' , s . K.
. 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 criticality.
OR This prevents core components frcm being expelled by hydraulic force [1.00).
(2.00)
REFERENCE Standard Operating Plans, FM 2.7 "Fuel Handling", p.
3.
)* ANSWER K.05 (1.00) ,' ' This prevents melting of the fuel element by afterheat.
(1.00) . REFERENCE Technical Specification, 3.10.
c.s e n e r be.
cor, red unles.s th e s 4 ;a., yriel 3.
f w e. - Ths e b(* de 3 < rc S // i<sc=recI [f.vo), f 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.sK CONTINUED CN NEXT PAGE ** * ** ) [ l. co '1 (**ot CATEGORY __ . _ _ - - .
. . . . .
, . K.
. FUEL HANDLING AND CORE PARAMETERS Page 34 , ANSWER K.06 (3.00) a.
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; b.
Variable reactivity refers to reactivity changes that may occur during core life (0.50).
Factors which cause variable reactivity include: (any two of the following at,o. fog G5'each) [1.00)
fuel burnup xenon-(and samarium) changes cample enanges - experivant changes (1.50: _ REFERENCE Technical Specification, 3.3.
ANSWER K.07 (2.00) 1.
The reactor is shutdown.
(0.50) 2.
Console key switch off and key is in proper custody.
(0.50) 3.
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, 1.1.
ANSWER K.08 (1.00)
m
- ia
<. n < -
_
tu.ovJ-d-. 9 f3 (fo/~t) (n,g y ' 4,. Y (o se ) (***** CATEGORY K CONTINUED ON NEXT PAGE *****) . - - _ _ _ _
. _ . . . . . . .. K.
' FUEL HANDLING AND CORE PARAMETERS Page 35! , .
REFERENCE , Technical Specifications, 3.10. - , . ' ANSWER K.09 (2.00) a.
A-ring because the flux is highest due to less leakage.
(0.50); b.
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. 10.8.1.
, . . . . . " . i- . , . j, i.
, - , t ! l
i A . ! . (***** END OF CATEGORY K *****) i - - . -
' _ . -, L.
ADMINISThATIVE PROCEDURES, CCNDITIONO ~ Page 06 . .AND LIMITATIONS s ANSWER L.01-(1.50) a.
1.
A licensed-SRO [0.50).
2.
Another person (qualification not specified) [0,50).
(1.00) b.
MITR Radiation Protection Officer (or his designated alternate).(0.50) - REFERENCE
Technical Specifications 7.2.1 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 5.
ANSWER L.03 (2.00) 1.
Obtain seme method by which the Operator-in-Charge may page you.(1.00) p 2.
Inform the Operator-in-Charge.
(1.00) - REFERENCE Administrative Procedures, PM 1.14.1, Paragraph 2.
. . (***** CATEGORY L CCHTINUED ON NEXT PAGE *****) .
.. . -' . ' o go 37 L.
ADMINISTRATIVE PROCEDUEES, CONDITIONS a AND LIMITATIONS . . . ANSWER L.04 (4,00) a.
1.
On duty console operator [0.50]. 2.
Any member of the NRL/RPO staff [0.50]. (1.00)
- b.
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 lockout.
, 3.
Person performing work will retain the key on his person.
4.
A notation as to the system being locked out shall be made on the statas board.
5.
The system must be tagged out.
6.
Lockouts anall be removed under the direction of an SRO [0.30] by the the person who performed the work [0.30).
(3.00) e n -J 7 e, s./a, e[en ter.
'?. .% s e o hn *% I<emissio** +* en re s-REFERENCE Ad.ainistrative Procedures, PM 1.14.3.
ANSRER L.05 (2.00) If the change does not change the intent of the origine.l. procedure.
a.
- (1.00) b.
1.
An SR0 [0.50]. l r . l 2.
Any other member of the reactor staff [0.50] (1.00) REFERENCE Technical Specifications 7.8.3.
I elmin i s trarls e.
freerde e, Pm t. f
- Proerdves Adherence 7'enjeny
r C ha nye A1e rhod. ! (44544 CATEGORY L CONTINUED CN NEXT FAGE - ***) . n - w - - ~ m
. . , , , . L.
ADMINISTRATI'/E PROCEDURES, CONDITIONS Fage 33 , .AND LIMITATION 5 ' ANSWER L.06 (3.00) .a.
1. 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 hottest fuel channel [0.50).
(2.00) 4-b.
To establish limits within which the integrity of the fuel clad is maintained.
(1.00' OR (will accept) - . To prevent flow instabilities.
REFERENCE Technical Specifications 2.1.
l i ANSWER L.07 (1.50) a.
The Emergency Director.
(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.
, b.
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 The.
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.
<er.,o n geo.n s 3 ( s veh < s les s o S o#-s irt. eleerrie<l oweeL CONT.5re1o.,)ON NEXT PAGE * **** ) (l.00) (*****CATEG[RY o <- INU D W '
. .L.
ADMINISTRATIVE PROCEDURES, CONDITIONS Page 39 . AND LIMITATIOhS , ANSWER L.08 (3.00) ' a.
1.
Notification of~ Unusual Event 2.
Alert 3.
Site Area Emergency . . (2.00) 4.
Alert j / b.
Potential radi ogi[ai C rdct (1.00) -
consequences.
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 4.4.1.
ANSWER L.09 (2.00) a.
Apparatus and mechanisms of a nuclear reactor, the mar.ipulation.sf which d:rectly affect the reactivity or power of the reactor. (1.00) b.
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 55.4 55.9 and 55.13.
, r . i l ' l . l (*r**e END OF CATEGORY L *****) ' ( * * * * * * * * * * END OF EXAMIN ATION * * ** ** * * * * ) l l .. _, _ _ -. _, . _ _ _ ... _ __ .. _. . _ _.
. . . <. TEST CROSS REFERENCE Page
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a _- - . , PM 4.5 Pg. 13 of 13 '
' . Table 4.5.1-4: Eats for a General Emerrency q)g 1.
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 hour.
. Notet yigure 4.7.2.2-1 lists the conditions and instrument readings corresnend-ing to a projected off-site dose of 1 rem / hour.
(PM 4.4.4.15a) > 2.
Sustained actual or projected radiation levels at the site boundary of 500 mree/
hour whole body.
(P M 4. 4. 4.14 a/ 4. 4. 4.11/4. 4.4.12) 3.
Blockage of fuel element channels thereby causing a loss of coolant to the affected channels and a fuel melt. This is the design basis accident.
(PM 4.4.4.154)
Loss of physical control of eitber the containeent building which includes the ' control room or of auxiliary areas that house vital eenipeent.
(PM 4.4.4.5/ 4.4.4.6).
{} 5.
Events that have esused or will cause r.assive facility and/or reactor system damage that cou13 lead to the reiting of fuel.
(Pr 4.4.4.15a) . J' . ' . . M AUG 6 1982 SR#-0-82-19 _.,. - _-. _. _ _ _.. _ .. .. .. .. ~ .. -. .. . - - _ - _ _ _ _
.. - . , , g, ....-..., . , P" 4.5 Pa. 12 of 13 . , . Table 4.5.1-1: FAf.s for a. Cite Area Fmersenev '? 1.
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 -- ' mrem whole body accumulated over 24 hours.
(PM 4.4.4.15b) - 2.
Same as di except the ef fluents could cause sn inter,rsted exposure of $00 mren thyroid.
(PF 4.4.4.15h) . 3.
Radiation levels at the site boundary of 100 nrem/ hour sustained for one hour.
(PM 4.4.4.14b/4.4.4.11) 4.
Abnormal loss of pri.sry coolant such that the core tank level drops helev the anti-syphon valves.
Qtote: This accident is not considered credihie, but procedures exist for cooing vich it.)
(PF 4.4.4.4) Inminent loss of physical control of the reactor.
(P" 4.4.4.6) ' ) * 5.
6.
Severe natural events beine. experienced. These include: . (a) An earthquake t at is causing cbr.ervable danage to the reactor safety h equipment vi:hin the containment building.
(b) A flood that is affecting the operability of any reaccce safocy system.
(c) Tornado or hurricane force vinds that are daeaging the. containment ' build'in g.
(PM 4.4.4.2) /* . 'k AUG 6 1982 . co e_n_ a s _3 o . _. ...-. .-- ... - -.
, ~ - - . , ' "* ,, . ?M 4.5 ?R. 11 of 13 ' . . Table 4.5.1-2: EALs for an Alert ,gg , ' 1.
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 hours.
(?M 4.4.4.13b) 2.
Same as 81 except the af fluents could cause an integrated exposure of 100 mree thyroid.
(PM 4.4.4.15b) d 3.
Radiation levels at the site boundary of 20 mrem / hour sustained for one hour.
(PM 4.4.4.14b/4.4.4.11) 4.
Abnormal loss of prinary cooline such that the core tank level remains at or above the anti-syphon valves.
(PM 4./ 4.4)
5.
Loss of radioactive material control that causes radiation dose races or air- , borne radionuclides to increase above oermissible exposure levels by a factor gst.)
of 1000 throughout the containment building.
(PF 4.4.4.12) 6.
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 personnel.
(PM 4.4.4.12) 7.
A fire leading to loss of radioactive material control within the containment building.
(PM 4.4.4 3)
3.
An imminent or existing hazard such ast ./' , ( l (a) Missile (s) impacting on the containment building.
(b) An explosion that af fects f acility operation.
[ (c) An uncontrolled release of toxic or flammable gases into the containment ' . building.
(?M 4.4.4.9)
s-i l-AUG 6 1982 e,4.3 19 so __ _. -. _ . _ - _. __, _ .
.. . - . . .... ~.:N ' an . , PM 4.5
- Pe. 10 o f 13 J
.. . Table 4.$.3-it EAI.s for "otification of (fnusual F. vents . 1.
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 hours.
(p" 4.4.4.15h) 2.
Report or observation that severe naturni nhenomena are either terinent or ex-isting.
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 systems.
(P 4. 4.4. 2) 3.
Threats to or breaches of security. ~ (P" 4.4. 4. 5/4.4.4. 6) 4.
A reactor safety limit's being exceeded such that a fuel dar. age accident that h could release radionuclides to the containment building is possible.
(PM 4.4.4.1) 5.
A fire within the cent.;in=ent building that lasts beyond the incipient stage or for -ere thin ten rinutt's.
(P." 4.4.4.3) 6.
Receipt of a bonb threat.
(PM 4,4.4.7) /' ~ l , ,,,,, c.oo, .. .....
..._-. __ _ -.. . .-. . - ... -.. - - - .-.
. _ __ -_ . . I j . t ,. RALL LIFTS . k
SY HY DRAULIC ' I l ' ' i ' \\ PRESSURE [ h .l '
BALL DROPS l s BY GRAVITY i j , , f ' .g [ O O s s s % i l
d CORE SUPPORT l l i . FL AN G E s-
- b M E\\ - l Yj I ' SV}i J [ ^ SV ,, Y S _ ' L '[
VALVE OPEN VALVE CLOSED
i
FIGURE I.17 ofa no._ o 'd-3 mars NATURAL C ONVEC TlON VALVES areno7zo M _____. 31.on,91 . .elt Arr'
v byde
- '
. %
, ' e
, - -, - - - - - - - -, -, - - -, -,. - -, - - - - - - -, -,, - . SYPHON BREAK g VALVE. BALL HELD UP T - BY INLET. PRESSURE ~ - >=( K(b a - i J J . , INLET FLOW f O O ' d-- O gO O o g g _,,/ 0,0 0' ' ' ~ .
, j Ong _,0 ' EFFECT O' O O' SYPHON O O0 g g O 0 0 0 ! CALL DROPS BY ,, l G R AVITY
W ' . ' ' ' \\ " \\
- -
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+ '%--
'%u? , , - VALVE OPEN $ VALVE CLOSED " ' . "# ^ " ~ "o-M- 3 ! ' FIGURE l'.16 - ~ ~ ""
- ^" " "d @1- -- si_gt ANTI-SYPHON VA LVES - , ' Q/AAPP'L_jg./,t_ J//e7Jt
/ ? l ~fa c L u f
-. .. , FACILITY COMMENTS AND NRC RESOLUTIONS FOR MASSACHUSETTS INSTITUTE OF - TECHNOLOGY REACTOR OPERATOR EXAM ADMINISTERED ON JANUARY 25,1988.
. 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 location.
"or -52 inches" will be added to first sentence of answer immediately after "core shroud" , QUESTION C.02.b Facility Comment: The plugs in the cooling tower standpipe overflow must first be removed.
NRC Resolution: Comment is valid. No credit will be deducted if this step is included in the discussion.
QUESTION C.07 Facility Comment: Operators are taught to distinguish between steady (asympotic) and dynamic periods. See MITR Physics Notes.
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 DM2.
NRC Resolution: Comment is valid. Answer will be changed to correct this ..... -n---,- . ----.-, , - - - - - - -, - -.. - -- - - - - - - -, - - - -,,.. - -, , - - ,- -- -
. . QhESTION E. 06.b ~ 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 answers.
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 confusing.
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 SRO.
KRC Resolution: Comment is valid. For this exam the Electronics Technician t l will be accepted as an acceptable substitute for Electronics Supervisor.
' 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 - - - _. - - - _ - . - - . - _ - .. -.
- QbESTION F.07
Facility Comment:
- Also, prevent sudden reactivity ef fect should blade move rapidly if suddenly freed.
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 answer.
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 Added " per pump or 1800 rpm total" immediately 1. QUESTIONS C.06(c) - after "900 gpa" in the answer. This will account for total secondary flow.
Added "with the reactor in automatic control" 2. QUESTION D.04(a) - to end of the answer. This will make the answer exact.
l )* ,
- l .. _. -. -. _. -. -,.... ~ . _. - -,.. _ _ _ -.. . .. . - _...
Ams uor Y ~ n _ 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 paragraphs.
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 accepted.
RSM 1.6.2.
' 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 l NRC-licensed member on shift if the director for operations is not present.
' NRC RESOLUTION: l In accordance with the "MITR-II Procedure Manual Chapter #4 Emergency j 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 accepted.
. I
'
l
. - ..._..- . _ _ _. - - - - - - -. - - ._ .
. . , OVESTION 1.03(a) . FACILITY COMMENT: The answer is not consistent with the answers from previous NRC exams.
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 public.
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 expected.
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: 1.
Stack gas monitor i 2.
Stack particulate monitor l 3.
Stack area monitor 4.
Portable monitoring l < ' REFERENCE: PM 4.4.4.14 page 1 and PM 4.4.4.15 page 2 which give thejEALS for general emergencies.
1RC 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
1 . _ _ .. .. __ .- - - . --
. . . . 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 answer.
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 bank.
@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 discharge.
So, there are three automatic responses.
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 buildup.
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 release.
' The answer is changed to read: ( ,
! ! ... _ _ _ _ _ -. ... -. . _ _ _ - _ . ..
>. .. , 'lhe purpose of +he core purge system is to (any one of the following . for1.00): ' 1.
Prevent hydrogen buildup.
4I 2.
Prevent the accumulation of A 3.
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 test.
(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: , 1.
Pressurizing the containment for the annual leak test.
2.
Cleaning the charcoal filter.
3.
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 button.
NRC RESOLUTION: The procedure clearly states that the rundown reset bettnn is to be pressed.
The answer is not changed.
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)
, --r , r.,,,,-s,-.--- - - ~ -,, - - y -,. - - - - - - y , w e - , .., ,,
.. . n NRC RESOLUTION: . , .(a) Is not listed in 5.8.4 and is not accepted.
(b) Is a generic response for all casualties and is accepted. The answer is changed as follows: I "8.
Verify the console operator has carried out his immediate actions."
ADO REFERENCE: "Abnormal Operating Procedures, PM 5.0."
l OVESTION L 9}, FACILITY COMMENT: The technical specifications state two senior operators.
But, it has
always been the Assistance Superintendent and the Superintendent.
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 directly.
Since they are the most senio.r licensed SRO's)."
ADD REFERENCE: "Technical Specifications, 7.2."
l 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)
NRC RESOLUTION: Agreed, the answer is changed as follows: - "Any two of the following at 2.00 points each: 1.
.....(As already written in answer) 2.
.....(As already written in answer) 3.
Feature - The grid cannot be rotated unless the shim blades are l fully inserted (1.00).
l l
l l . - - - -.. - .. _ _ _ - -, . _ _. . _- _ , _
. . e 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 typewriter.
NRC RESOLUTION: Agreed, answer separated will read as follows:
"a.
8 1/3 (+ 0/-1) b.
4" OVESTION l.07(b) FACILITY COMMENT: On-Duty Shift Supervisor may also authorize reentry to save a life.
(PM 4.4.4.13) - ' NRC RESOLUTION: The facility comment and the examination answer are both.too limiting.
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.")
.. - .. .. . - _-- . _ _. __ __ _.
- -
. . , The answer, in light of the above and in li ht of tha way in which the questionwasaskedischangedtoreadasfo$1ows: . ' 1.
"Whenever the on-duty Shift Supervisor is the acting Emergency Director (0.50).
2.
Whenever the building evacuation was necessitated by the buildup of Argon-41 (0.25) following a loss of ventilation for routine reasons (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".
,
, f r = l l
.
m w . . f f4 The following changes were made to the examination as a result of final review.
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 curve.
, QUESTION K.06 Change "0.05 each" to "O 50 each" Explanation: Editorial error.
. OVESTION l.04 Add the following: 7.
Must obtain permission from the on-duty console operator.
Explanation: Requirements for hanging the tag are also requirements i for conducting the lockout.
DUESTIONt.05 Add Reference: " "Administrative Procedure, FM 1.5, "Procedures Adherence Temporary Change Hothod".
Explanation: Additional reference as noted in question.
OVESTION L.08b r , Deleted < Explanation: Question was too vague,,and criteria was used instead of - ! the intended "criterion.
' !
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