IR 05000166/1987002
| ML20215B369 | |
| Person / Time | |
|---|---|
| Site: | University of Maryland |
| Issue date: | 06/04/1987 |
| From: | Coe D, Collins S, Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20215B336 | List: |
| References | |
| 50-166-87-02, 50-166-87-2, GL-83-17, NUDOCS 8706170305 | |
| Download: ML20215B369 (17) | |
Text
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . U.S. NUCLEAR REGULATORY COMMISSION
REGION I
REQUALIFICATION EVALUATION REPORT Evaluation Report No.
50-166/87-02 (OL) Facility Docket No.
50-166 Facility License No.
R-70 Licensee: University of Maryland College Park, Maryland 20742 Examination Date: March 29, 1987 Chief Examiner: D ? '/
D. Cce, Lead Reac or Engineer date Reviewed By: M) 4/'//f7 R. Keller, Chief, Projects Section 1C date Approved By: IM bh @ _ J. Col' ins, Deputy Director date Division of Reactor Projects Summary: The NRC audited the facility licensee's administration of its 1987 requalification written examination to five out of six licensed operators.
Twenty one percent of the examination was replaced with NRC questions, and all examinations were reviewed for adequate facility grading practices. All opera-tors achieved a passing overall score and facility administration of this examination was determined to be satisfactory.
f Y Y i i
,
- - i . DETAILS 1.
EXAMINATION RESULTS: R0 SR0 TOTAL Pass / Fail Pass / Fail Pass / Fail Written Examination 0/0 5/0 5/0 Evaluation of Facility Written Examination: Sati sfactory Grading: Satisfactory 2.
SCOPE The it. tity prepared requalification written examination was reviewed by the. ' C prior to its administration.
The examination was administered to five canior operators.
There are presently no licensed reactor operators at tnis facility. The NRC was present to monitor the proctoring methods
used by the facility.
Twenty one percent of the total examination value j was replaced with NRC written questions to provide an independent check on examination validity.
The NRC reviewed all facility graded examinations for adequate grading practices.
3.
WRITTEN EXAMINATION The facility written requalification examination consisted of a 100 point exam comprised of four equally weighted sections.
Its format was iden-tical to an NRC power reactor SRO written examination.
The review deter-mined that the facility examination was adequate to determine individual as well as generic knowledge deficiencies.
However, three questions in Section 8, Administrative Procedures, Conditions, and Limitations were not operationally important and were replaced with NRC questions.
4.
CONCLUSIONS: This audit found that the facility properly wrote, administered, and , graded a satisfactory requalification written examination for its licensed ' operators.
Operator performance on NRC substituted questions was not , significantly different than that on facility written questions, thus pro-viding an independent check or< the validity of the examination results.
The facility commitment to continue to administer its requalification examinations on a proctored basis and in accordance with the NRC policy expressed in Generic Letter 83-17 is noted and will be subject to future inspection and audit.
Since all operators fully passed the examination, it 'can be concluded that pievious unproctored examinations did not adver-sely affect the overall knowledge level of the licensed operators.
Attachment: 1987 Written Requalification Examination and Answers
- .. - _ _ _ - _ _ _ _ _ _ _ _ _ _. y f;( - * ' ATTACHMENT 1-
, Requali fit at ion Ex arn for 1987 Category V - Theory of NucIgear Reactor Operation 1. (3 pts) Using the six factor f or raul a, describe the various processes which roay happen to a neutron in one generation.
2.
(5ce.LeIc.~.) - . 3.
(2 pts) Describe what happens and explain t wo r easons for the behavior when the source is withdrawn f r orn the exactly critical - ~ condition et 10mW c orop ev ed to 100kW.
- q.
( S e e.
he le :- ) , 5. (2 pts) Explain how you could tell that you are exactly critical at 10rnW c oropar ed t o 100kW with the source inserted.
6.
(2 pts) Descibe an e x p er i rnent you could per f or ra to rneasure the rna x i rnurn shutdown Xe reactivity in the MUTR.
7.
(3 pts) Assurning that the MUTR has a posi ti ve rooder at or t eroper at ur e coef ficient of reactivity, explain why the power level increases sornewh a t a f ew rninutes a f ter the pri rnar y purnp in turned on at high power l evel s (e. g.
2OO':N:.. 8. (3 pts) Describe indications, if any, that rn a y be observed on the c ont r ol r eorn i nst r utnent a t i on if too incore portion of the p n e urna t i c tube brole whi1e inserting e e arnpl e during high power operation.
. , 9.
(3 pts) State three reasons why the pr irnar y. cool ant water is Lept as pure as it is.
10.
(2 pts) State and explain the sho*t t er ro and long terrn indications you could detect due to a fuel rod rupture.
is 0.95.
Your start-up .2, Your reactor is suberitical and Keff channel reads 20 CPS . When the ct.unt rate has increased tc 40 CPS, what is the a) (2) new K gf? (Show all work) e b) Would the reactor be suberitical, critical, or supercritical if enough reactivity was added to double the count rate again? (Show all wrrk) (1.0) it Would the control rod's or shim rod's worth vary: ' a) If another rod is placed adjacent to it.
Briefly explain.
(1.0) , b) If the moderator temperature increases? (1.0) Briefly explain.
' i - -- -~ --- ~- -, _ ,, _ ~ g __ __,
, _ _ ._ _ . a- ., ' f - / Category VI - Reactor Facility Design, Contrel and I n st r urnen t a t i on i n 1.
(2 pts) Describe and explain any differences you would see in the response of the CIC when going to low power operation following a sc r arn for a long period ( several hours) of operation at 2OOkW under two di f f erent conditi ons on the cornpensating ] voltage: - a.
Un de r c ornpen sat ed b.
Overcompensated I 2.
(2 pts) What setpoint' setting would you adjt.st the fuel t erop er at ur e so that it would seram at about the same power level as safety channels I & II. Would this violate the Tech Specs? j Why? . l 3. (3 pts) Fr orn the attached drawing of the instrumentation, give labels to the numbers 1->6.
e . -f 4.
(3 pts) From the attached dr awing of the MUTA e configuration, give Iqbels to the numbers 1->6.
(ps) w e > 5.
(4 pts) List the^ design features which are implemented to minimi: e the release of radioactive rnaterial to the envi r onrnent. [[ 6.
(4 pts) List describe and give the reasons for the
- ^)
i , , sequeacing of the up, down, contact, and inag on lights i f you drop the Reg. Rod f r orn full out. Assume operating in the steady state rnod e. 7.
(4 pts) Lable all the roajor actions on the attached chart recording such as st ar t up chec ks, red and blue pens, s c r ains, , sour c e r ernova l, e st i rna t ed power l evel s, etc.
8.
(3 pts) Calculate the positive reactivity which is maintained by the automatic control sy st ern when going from 100mW to 2OOkW.
. . $ .
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I - " Category VII - Procedures - Normal, Abnormal, Emergency and i Radiological Control I
. 1.
(4 pts) What conditions must be met prior to any fuel movement? How often must the fuel be inspected? What constitutes a damaged element? ' 2, (S et be le d ) . , ' l 3.
(4 pts) List the indications of a leak in the tube side of the first heat exchanger. What actions would you take if you,discovere it while operating at 200 kW with the cooling system on.
(pW - n.
4.
(3 pts) List three types of Emergency Plan emergencies and give one example of each.
j , l 5.
(3 pts) List the people who must be present during the removal - j of an experiment in the through tube. What precautions are to be observed? ) l ' 6.
(3 pts) Provide the following in accordance with 10 CFR 20 for radi ation wor hers per quarter.
a.
roaximum permissable dose to the whole body b.
maximum perrnissable dose to the hands c. rnaximum permissable dose to the skin L 7.
(4 pts) A radioisotope is known to emit a single 2 MeV gamma ray for each disintegration. An unshielded sample of the radioisutope is observed to cause a radiation level of 0.5 rad /hr , at a distance of 1 ft.
a.
How rnany Curies of the raaioisotope is present?
b.
What is the radiation dose rate at 4 feet? '{ What thickness of shielding will be required to reduce the j radiation level f r orn the sample to 50 mrern/hr at one foot? ' Assume a half value thickness of 1 cm.
k The following measurements are made from a beta-gamma point source.
2r/hr at six inches (1) 0.5 mr/hr at ten feet (1) I What are the relative fractions of betas and gammas emitted? (state assumptions and show caleviations). 2, You are in charge of loading one of several experiments.
r I into the reactor , a) What is the maximum allowable worth of this experiment? (1) b) How do you insure that the experiment is,within the limits of part a)? (l)
__ _ - - m.
_.
.__ _ _-
_ ' . .
Category VIII - Administrative Procedures, Conditions, and Limitations a 1.
(4 pts) Guppose an experimenter proposed to undertake a prompt jump experiment by removing a rabbit containing. cadmium at low power critical conditions. What steps must be undertaken before this experiment could be undertaken? ~ 2.
(3 pts) Upon proceeding to 2OOkW, you notice that the CIC pen on the chart is not operating properly. What course of action would you take? Then suppose instead you notice that the fuel temperature indicator is not operating correctly. What course of action would you take? 3.
(3 pts) During reactor operation, how may people must be present in the control room and what must their qual i fi c at i ons.
be? In the Chemical 8< Nucl ear Engineering Build,ing and their qual i fi c at i ons? 4' List the two Technical Specifications requirements that assure safe storage of your fuel when it is not in the core.
(2) 5.
(3 pts) During reactor operations, an experiment fails and releases substantial amounts of radioactive material into the reactor building. The evacuation alarm is sounded. What speci fi c actions and responsibilities do you have as an SPD in this l situation? g In accordance with your Technical Specifications, name (2) four types of " Reportable Occurrences."
'
Title 10, Part 55, Operators Licenses, defines an operator as any individual who manipriates a control of a reactor (15-) a) Define " cont.rol" b) When can an unlicensed operator manipulate a cont ol? (2) g, Describe I'mnediate actions required by procedures which must be perforned upon learning that a group of students are marching to the Reactor Facility to attempt a sit-in.
(2.5) According to your Technical Specifications, the rate of y' reactivity insertion by control rod motion can not be greater than $0.30 per second. Briefly explain why.
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' . . . Answers to 1987 ?.equali fication Exam s V.1.
Six factor formula k= epsilon *pteta*f*P P ft epsilon = fast fission factor p= resonance escape probability eta = neutron replication constant f= thermal utilization - Pf=f ast non-leakage probility P = thermal non-leakage probability t V.2.
Mi/M2=(i k2?/(1 k 1 ? = 1-/ 3-C 1 -k-2F/,-04=->k2=h987 i -.. i V.3.
The two parts are ) 1.
Drop due to removal of neutrons 2.
Drop due to insertion of negative reactivity since the-source is made of Pu-239 and is worth about +4 cents 010mW both effects are comparable but 0100kW only the reactivity effect is around. @10mW the reactor will continue with a long negative period to almost zero power, but at 100kW, the reactor power level will level off at a slightly lower power due to the { fuel temperature feedback effects.
] -Vr4r-Wor-th-i-s-dep+ndent--on-the-squ-ar+-of-the-therea-1-neutcr+n-f-lu x-whi-d c h = g : uitF 1 o e s ti o n r-- ] V.5.
O 10mW the power level would be increasing in a linear manner, but would appear to be constant 100kW.
V.6.
First measure the critical rod position at low power. Next, operate at max power for several days. Go to low power and l measure the critical rod position over about 15 hours. Max Xe
should occur at about 12 hours. Max reactivity measured as the difference in critical rod position before and af ter high power.
V.7.
Turning on the primary cooling pump increases water flow through the core which increases the heat transf er coef ficient, which lowers the fuel temperature which causes an increase in power level. The fuel effect is larger than the moderator effect.
V.8.
Power level would flucuate more due to the void coef ficient of reactivity, bridge monitor may increase due to Ar-41 level increase,... V.9.
1.
Peduce activation products 2.
Easier to detect leaking fuel rods 3.
Less corrosion 4.
Clearer water to the the core for inspection r ! V.10. Short term: 1.
Increase in bridge and exhaust monitor levels l 2.
Visual inspection 3.
Increase in conductivity , Long term: , l
0 l - . ......... ' - l
-- __._. - .4 L vc:.N m . :., m- ' 1. Increase in gross beta / gamma activity in pool water measurement a 2. Increase in gamma spectrum activity in pool water M measurement . VI.1.
a.
undercompensated .' 'CIC would be reading higher than the actual power,since.it . , is falsely including some gammas. It would indicate'a constant j . power due to the fission product, gammas until the actual neutron g$ signal became'large enough'to overide the f al se gamma ~ signal. e
This could occur at a power level of several" watts or'even ) ,.3 higher.
b.
overcompensated
, Just . the oposite of a.. CIC reading lower than actual epower, since it is subtracting gammas from the ' neutron signal. It;would' ' indicate a negative power up to.several watts or even higher.
VI.2. Set point of about 200 deg' Cent on the fuel would scram the reactor at about 3OOkW=120% licensed power.
This wouldn't vi ol at e the Tech Specs because it says'that the fuel temperature setpoint shall be<= 400 degrees Cent.
-VI.3.'i= source; count rate relay 2= chart recorder 3=CIC 4= Safety I 120% relay 5= Ionization chamber 6= Fission chamber VI.4.
1= shim 2 2= instrumented fuel rod 3= fission chamber i 4=B-10 lined ionization chamber j 5= graphite r e fl ec t or i 6= neutron source l VI.S.
1.
fuel =UZrH 2.
ss clad 3. automatic exhaust fan turn off 4. all 9 scram point settings 5.
etc.
VI.G. 1. mag current light off because switch interrupt, contact-light of f when rod is in because opposite conditions on carriage down and rod drown switches, up light off due to CRDM starting to be driven down.
2. mag current back on when drop rod button released.. 3. down light on and contact on when CRDM has driven All the , way down.
VI.7.
...... VI.B. rho $=1/(1+1ambda*T)=1/(1+0.12*15)=.36$ s
3.,.~._.s.,~........-.m_,....... .. u.
.. ...... ..... .. . . . .. .... . .. . .. - , IL-( , !
. ( VII.1. Conditions: a.
SRO on bridge a b. Reactor shutdown with control rods inserted c. RO or SRO in control room-d.
"u;1 ...v vemen t - e c e r d ed4+t-Feg -book-- Inspected once every two year s.
Bowing >1/8", length >1/4" . VII.2.
1.
21 pha- -2.
m eut r ons- -9.
betc-41-gamma-- l VII.3. The pool water level would increace and over flow in+n the ' sump tank. Also the conductivity would incYease. If it wat important to remain at power, I would turn off the primary n ump, valve off the city water to HX1 and sewer valve to HX1. Leave HX2 i Could operate in this condition f o'/ several hours but then on.
would have to shutdown.
VII.4.
1. personnel emergency = class O being cut on a piece of radioactive material 2. unusual event = class 1 = sustained fire with a minor explosion in the reactor b1dg.
3.
alert = class 2 = severe fuel damage with clad failure following an earthquake or explosion in reactor b1dg.
VII.S. SRO or RO and appropriate number of assistants.
Precautions: gloves, face mask, radiation monitors, rope of? area, monitor for radiation levels,.... VII.6.
a.
1.25 rem b.
18.75 rem c.
7.5 rem
- VII.7. D=6CE/d a.
41.7 mci b.
31 mrad /hr c. 3 3 cm VIII.1.
1.
approval by RSC and RD because a special experiment 2.
fill out irradiation forms VIII.2.
a.
Would not have to shutdown the reactor. Would try to fix the CIC pen & recorder , b.
Scram the reactor. This would be a reportable occurrence and would have to take the appropriate actions VIII.3. In cotrol room = licensed RO or SRO. In b1dg. another ' person that has some knowledge of reactor operations.
If RO in control room then SRO niust be on call.
-W-I-I. 4 a.
en+-r-equa-1-cyc1 e"2 - - y r-s.
-b. >-5 yrc
c.
>-1 yr
. . . - d.
>= 5 yr e. l i f et i rne o f f acility VIII.5. secure reactor, secure venti 11ation, take log book and radiation roonitor, check to see i f averyone evacuated, prevent re-entry into reactor room, noti fy RD and RSO, assist'in erner gency , VII!.E. Net t 2ecfcrable, 0 +1 y fcr faci 1ity en t he-1-i+ ens c, col; -en cent e1e epe:1444d an 1-14+ns+r--su b.-jeet-t<+-NRG-r-ul-et .... :a.......e. .: p' ', ' =VIII.7. Oni-y-SRO=c, c, d;- r. n. -,.n __ u , f T bJ %.'. hr i + b.i -'- b# p h p 5
i ) i ) , ) . e
,
i
, . - 1-K ](,', [* C.R.2 effi
.95 .975
- g
=
, ,
- ff2 a)
" 1-K 1-Eeff2 C.R y ff2 , i i ' l 1-K
.975 eff2 .9825 .g =
i = b) y_g y_g egf3 - . eff3 eff3 - Therefore the reactor is still suberitical_.
, g-a) Yes, placing a second rod adjacent to a first will depress ie the flux in that area and decrease the worth of the first
i rod.
This is kncwn as rod " shadowing."
. b) The rod worth will increase as temperature increases.
This is because the neutron diffusion length increases as temperature increases, allowing the control rod to see more neutrons.
~1 i
e !. - ,~n,,,.., _,,n.e,e m ~ . - _ -. __ , -..,. -.,,,,,. -,,. ,
- . . _ _ _ - _ _ _ _ _ _ _ _ _ . - .. . . ,
. $ k ' Assume S vill not travel 10 f t in air therefore the 0.5 mr is all gamma's g
RD 0.5 (10)_
200 mr/hr y1 " = dose at 6 inches (0.5 ft)
"
- 25
! = (0.5)2 ' D2 - 2000-200 = 1800 ar/hr 200 ar/hr S = at 6 inches .y = 't.- 1800 ' = " S/y. ratio 200 . .. t , .s , $1.00 reactivity (Tech. Spec.pg.h)' k 2* a) insert Reg. Rod.
Go critical without experiment, Find p worth b) Add experiment, go critical on Reg. Rod.
from difference in Reg. Rod positions.
,. I-I i i - ! j I T
l i
'
E _.
__ _ ~ , _
. Q . Raf. Tech. Specs, pg 19 , /0 All fuel elements shall be stored in a geometrical array ' ' where the k,gg is less than 0.8 for all conditions of moderation.
, b) Irradiated fuel elements and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.
[, 'Ref. Tech. Specs.,p.p) (Any four) . 14 setal o rpewtokle.
Occurrence - An-c m 21 occurrence is any of the following h which occurs during reactor operation: a) Operation with any safety system setting less conservative than specified by the Limiting Safety System Settings b) Operation in violation of a Limiting Condition for Operation c) Malfunction of a required reactor or experiment safety system component which could render or threaten to render the system incapable 6f performing its intended safety function j d) Any unanticipated or uncontrolled change in reactivity greater than $1.00 e) An observed inadequacy in the implementation of either administrative or p'rocedural controls, such that the inadequacy could have cadsed the existence or development of a condition which could result in operation of the reactor outside the specified safety limits j f) Release of fission products from a fuel element }~. a) Control means apparatus and mechanisms the manipulation of which directly affect the reactivity or power level of the reactor b) Any individual may manipulate the controls of a facility as a part of his training to qualify for an operator license under the direction and in the presence of a licensed operator or senior operator, or as part of his training as a student in a nuclear engineering course under the direction and in the presence of a licensed operator -or senior operator.
k Ref. - EP 405, pg.1 a) Notify the Division of Public Safety Office b) Notify the Reactor Director c) Notify the Radiation Safety Office , d) Secure all doors to reactor building and containment area e) Activate the intrusion alarm system j $, Ref. Tech. Spec., pg. 3 Assures that power increases caused by rod motion will be terminated by the reactor safety system before the, fuel temperature safety limit is exceeded . }}