ML20116E814
| ML20116E814 | |
| Person / Time | |
|---|---|
| Site: | Reed College |
| Issue date: | 10/08/1990 |
| From: | Joseph E Pollock REED COLLEGE, PORTLAND, OR |
| To: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML20116E723 | List: |
| References | |
| FOIA-92-35 NUDOCS 9211100031 | |
| Download: ML20116E814 (2) | |
Text
. . _- _ _ _ _ . _
. \
'Portind, Oregon srsos.use -
R E E D-' C O L L E O E 1
I stactoa racettry 3 October 8,1990 I
John B. Martin, Administrator, Region V ;
U. S. Nuclear Regulatory Ocmmission '
1450 Maria Lane, Suite 210 !
Walnut.Or5ek, CA 94596 RE: Docket 50 P.88, Lloonso R-112
Dear Mr. Martin:
This letter is written as notification of the coeurrence of anomalous behavior of the linear power channel and/or the automatio mode of
- eeration on the Reed Reactor Facility, TRIGA Mark I Reactor.
Jn October 7,1900, beginning at about 15:40, the reactor power was increased to 150 kW as part of a flux monitoring experiment. The apparently anomalous behavior occurred immediately after the reactor was swltched into the automatic mode of operation C 15:49.
Por a periori of approximately 6 minutes, the signal reaching the linear power chart recorder led the operator to believe that the reactor power was 150 kW. Other factors, now believed to be kaccurate, Indicate that during this time, the actual power climbed At 15:55, '1
- slowly and stabilized at a level of approximately 240 kW.
the linear channel resoing increased sharply to match the 240 kW level and the automatic system immediately adjusted the regulating rod to return the reactor power to 150 kW.
At no time did any of tite reactor channels Indicate that-the reactor power exceeded our licensed lovel of 250 kW.. In adoitton, da!!y instrument calibration and safety ' system checks, routinely-performed prior to the operation, _werw.. normal.
The tests of the linear channel were repeated by the operator'following the occurrence and were again r.1,tal. Con muently, we do not believe, a t tty that there was a fallere ofFe"rew system or a violation of the
- we are notifying you because we Technical Specifications.
/ 9211100031 920414 C D92 '{Df,g Tehphene f s D R i 711 1112
- zon 97202 8199 h 3203 Soutkrast Woodstock Bostleear. , ,. I ,
4 w ;r v.n.c )
s-
. ,- n .,. ~ - ::
- feel that an.omalous behavior of any _ safety system should be
! Investigated fully and we welcome your input.
All reactor operation has been suspended. . initial conversations with General Atomic have been held. Testing of the linear channel and automatic mode in an attempt to leolate-ths problem will be conducted following additional consultation with . General Atomic. ! l will be contacting your staff after the holiday wookend or they may i contact me at (503)777-7222 for more_ information. A full report ,
will be prepared when testing is completed.
Sincerely, J. Michael Pollock AqtinD Director ,
cc. D. Bennett, Provost D. Gerrity, Chair, Reactor Cversight Committees ,
D. Stewart-Smith, ODOE y .S e
O- O O- A- r . _ , , . ,v. , , . ... , ., ,s 4 -
r' -= - - -vw++-- .- y
] ,
- alMdUd REED C0LLEGE T 'Ponhusd, Orjgocs 91202-8199
+
y
" " " ^ * " November 5,1990 John B. Martin, Administrator, Region V U. S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek, California 94596 ,l ,, .
RE: Docket 50 288, License R 112 0,~
7 . .
Dear Mr. Martin:
.-j -
A a
This letter is written as a follow-up report on the anomalous behavior of t@
linear power channel on the Reed Reactor Facility, TRIGA Mark i Reactor previously reported on October 8.
Attached are copies of a report prepared for the Reactor Operations Committee (ROC) following testing and repair of the system and the minutes of the ROC meeting where this was considered and authorization to resume operation granted.
The direct cause of the problem is attributed to either dirt and corrosion on the mode switch and circuit board cordacts and/or incorrect adjustment of the linear power amplifier. Complete cleaning of these contacts and calibration of the entire linear channel has eliminated both the reported problem and additional electronic r.oise problems in the console.
The root cause is that the console electronics is aging. I have contacted GA to request a cost estimate to hire an electronics technician, experienced in working with the TRIGA console. This individual would work with, and train. Reed's new electronics technician in conducting e complete calibration and cleaning of all electronic systems.
In addition, we have begun an evaluation of both the technical details and the 10CFR50.59 questions involved in modifying the reactor "y installing a second linear power channel (wide range power channel donated to Reed by Northrup, Inc.) to provide us with a third scram-capable power channel. This system would replace u the mode switch and the automatic servo control system which have given us minor l problems for many years, if you or your staff have any further questioc.s, please contact me.
Sincereiy,
[- u n / fjf .
/g-/y,
/ J. Michael Pollock Ac ing D; rector
'ec: Douglas Bennett, Provost r David Stewart Smith, ODOE Junaid Razvi, General Atomics f e): \; ( f f f 0
$y Portland. Oregon 97202-8199 Telephone i503 o 7 7I 111:
3203 Ecutheast Woodstock houirtard ..
/ ,., ./;d AD
).
REED COLLEGE ) ' Portland, Orkn mnsm Y %
Ct af'108 f Afli tT)
Minutes of Special Reactor Operations Committee Meeting October 24, 1990
Subject:
Problems with linear channel Attendance: Johnny Powell, Chair, Reactor Operations Committee; Committes Members: D. Griffiths, P. Terdal, L. Church; Dan Gerrity, Chair, Reactor Oversight Committee; Staff Members:
M. Pollock, Acting Director, S. Herbelin, Reactor Supervisor, M.
Begel, SRO Meeting was called to order by Chairman Powell at about 17:30.
Copies of the attached October 19,1990 report of the Acting Director had been distributed in advance of the meeting. Copies of the October 8 letter to Mr. John Martin, USNRC were distributed.
As the first order of business, Mr. Junald Razvi, Director of the Radiation Facility at General Atomics, was available by telephone to present his evaluation of the problem and response to it and to answer questions from committee members.
Mr. Razvi began by emphasizing that he could officially comment only on technical matters reiated to the reactor hardware and that it was Reed's responsibility to address requirements of the NRC and Technical Specifications. He indicated that the problem appeared to result from infrequent cleaning and maintenance of the electronic components of the reactor. It was his opinion that the staff had taken the proper action to correct the problem and that, with proper maintenance it should not recur. He indicated, however, that as the facility ages, similar problems are lik91y to increase. He also indicated that GA conducts, as part of their semi annual maintenance, electronic calibrations of their primary power channels and includes cleaning of the inside of the console and certain switches.
Mr. Terdal questioned Mr. Razvi with respect to whether the lic,sar power scram would have functioned had the operator chosen to raise the power curing the 6 minutes the system was apparently operating anomalously. Mr. Razvi indicated that, although that would be hard I
Portland. Ortpn 97202.E199 Telep h o n t t503o :7I.111:
3203 Southeast Woodstock Boulet ord
I to say for sure, it is possible that the actual reactor power would have exceeded 275 kW without a linear power scram. He emphasized however, that this is the reason for redundant safety systems on the TRIGA reactor and that, in his opinion, the incident raised no unreviewed safety questions. Specifically, he commented that:
- 1) The percent power channel, which is completely independent and unaffected would have scrammed the reactor;
- 2) Even if the percent power channel had failed as well, the recctor power would have stabi'hed at about 300 kW, well within the limits discussed in .he safety analysis report; 3)
All concern could be prevented in the future by installing a 3rd scram channel (the wide range channel donated by Northrup); and
- 4) Training operators to carefully observe all power channels, is the best way to detect anomalies and prevent overpower operation.
There being no more questions from the committee, Mr. Razvi was thanked and excused.
Motion (L. Church, seconded by P. Tordal) Limit the discussion at '
this meeting to questions related to restarting the reactor; postpone long term maintenance decision 3 for the next regular meeting when the staff will bring recommendations to the committee. Passed unanimously.
Motion (L. Church, seconded P. Terdal) The reactor be considered to be back in operation.
Mr. Tordal questioned whether a 10CFR50.55 review was required for restarting the reactor. Mr. Pollock pointed out that, although 10CFR50.59 specifically refers to making changes in the facility or procedures, committee approval to restart the reactor should imply a determination that no unreviewed safety questions exist and that the reactor can be operated safely.
Mr. Church indicated his belief that the staff had conducted the appropriate tcsting including attempts to repeat the anomaly.
The question was called and the motion passed unanimously.
The Acting Director indicated that he would be immediately issuing a notice to operators on this incident and would be filing a follow-up report with the NRC.
, q'
'f Octc vr 19,1990 Yo: Reacu Operations Committee Tom: J. Michael Pollock, Acting Director
Subject:
Anomalous linear channel behavior in automatic mode Or, October 7,1990, b9 ginning at about 15:40, the reactor power was M
. Increased to 150 kW at: part of a flux monitoring experiment. The
- apparently anomalous behavior occurred immediately after the nactor was switched into the automatic mode of operation at 15. ~ ,
t ly 6 minutes e signal reaching the %
For a period oi approxiri linear power chart recom ied the @ereu.' ~.o believe that the reactor power was 150 kv.. Other obswt rm suggest that the actu' e climbed slowly and stabiliz+o at a power level betwe M5 and 240 0.W; specifically, the log recorder indicated a power level matching its normal reading at 240 kW, the recorded
- core excess of $1.29 corresponds to a power level of 240 >JN (see attached core-e,< cess record), and a 15 second flux measurement made in the pneumatic transfer system during this interva' indicated a flux, based on measurements during previous experiments, of about 215 kW. The percent power reading, which was recorded sornetime during this interval, was recorderf at 65%,
however, it is uncertain when during that period it was recorded.
At 15:55, the linear channel reading increased sharply '.o aimost 97%
L (249..kW.).Jad the autprpatic system immediately adjusied the regulating rod so return the reactor power to 150 kW. At this point the reactor core 4rcess of 51.70 was normal.
- At no time did any of the reactor channels indicate that the reactor powc; exceeded our licensed level of 250 kW. !n addition, daily instrument calibration and safety system checks, routinely performed prior to the operation, were normal. The tests of the linear channel were repeated by the operator following the occurrence and were again normal.
..,a
'-~~---
ACTIONS TO IDENTIFY AND CORRECT THE PROBLEM
- 1) Reactor operation was suspended. General Atomics (GA) was consulted and, although there was not believed to have Daen a violation of the Technical Specifications, the NRC was notified of the anomalous behavior. Bob Ormond, the new campus electronics technician was contacted. Mr. Ormand spent several days familiarizing himself with the reactor docurr 4 nts in preparation for testing.
e 2) Review of the chart record by GA and subsequent phor,e consultation resulted in several recommendations: 1) to test temperature effects on germanium transistors with a low temperature heat gun; c to remove all circuit boards from the linear channet for cleaning; 3) to evaluate continuity through any
" cold soidered" contacts; and 4) to check and adjust the amplifier gain to match tne specifications found in the Instrumentation Maint nar .e Manual.
3)~ Past mair.ienance records were searched and problems related to the mode selector switch (including associated linear power scrams), automatic mode operation, and calibrations and alterations praiously performed on the linear channel were noted and discussed.
- 4) Attempts ware made to recreate the problem with reactor at zero po,ver, ie. Lil control rods in core. Switching the mode switch caused " spikes" on both linear and log channels. On two occasions, ;
turning the switch produced a rcpid rise in linear recorder indication similar to the occurrence on 10/7. The first event was interrupted L before_arl automatic SCRAM occurred by the operator switching ranges, and then a manual SCRAM. The second event was allowed to 4'$
continue until a reactor SCRAM occurred (about 110% of 0.1 W but it happened too rapidly to be any .nore precise) and was corrected by ~(
i
" jiggling" the mode switch but allowirig it to romain in automatic.
Joise spikes wers noted on the log channel but no " loss of signal" occurred.
The following maintenance operations were conducted by Michael 5)
Begel, SRO, and Bob Ormond, electronics technician, with assistance I and consultation from Michael Pollock, Acting Director. _
a) Heat testing was conducted by gently Plowing warm air over circuit boards. Calibrations were noted to fluctuate i
t
-- -~ _ _
Y s
. I significantly but in a smooth manner with changes in temperature.
No sudden failures or instantaneous changes were noted.
The Mode Switch was inspected for corrosion, dirt, an2 b)
} electrical continuity. Minor copper oxide deposits were noted; The switch w-- cleaned, 1
considercble dirt was present. vere removed and e
c) All circuit bos.rds in the linoer channt the contacts cleaned.
Linear Channel testing and calibration was perforrned d) following the GA procedure (Chapter 5, instrumentation Maintenance
} Included in this calibration and testing was an adjustment Manual). The initial voltage measured
=
of the gain on the 10 KC Amplifier. ~
between TP1 and TP3 (ground) during the test was 1320 mV compared to the 1215 mV called for in the procedure (See N maintenance log book).
f Following this maintenance, shutdown testing indicated virtually 6)
( no spil;es on either the linear or log pens when the mode switch was g
changed.
Followho consultation with Dan Gerrity, Chair, Reador 7)
Oversight Committee, and informal phone discussion with Mr.
Michael Cillis. USNRC. testing with the reactor crhical was performed by d,ichael Begel, SRO, Michael Pollock, SRO At 5 watts theated Acting
~ Director, and Bob Orma id, electronics technician.
Repeated switching between core excess was norma at $2.77.
manual and steady-state operation at 5 watts produced no " spikes" or anomalous readings. Attempts were made to turn the switch too slowly or incompletely since some problems have been noted in the switch and the range _ switch which were past with both the modo The related to poor electrical contact being made during switching.
power level was then increased in r nual Themode topower reactor 80% of 100 kW was aad the mode switch operated repeatedly.
y
' lowered to 60% of 100 kW in automatic and the mode switch tested again. Since the original problem occurred with the % demand potentiometer at about 60%, this pot was adjusted up and down through the range from 58-65 % to see if any anomalous behavior was noted. The reactor was een taken to 60% of 250 kW and the operation where the anomalous behavior The modeoriginally switchoccurred was againwas duplicated without any abnormality.
operated repeatedly at this level with no spikes or anomalous behavior occurring. Core-excess measurements during these tests ware normal and are included on *r e attached sheet.
-D 1
- 8) Atthough the anomalous behavior of the linear channel seems to have been fixed and system noise resulting from the mode switch significantly reduced, the reactor was again shutdown pending approval of the Operations Committee to resume normal operaticns.
CONCERNS RELATING TO FUTURE OCCURRENCES
- 1) Periodic minor problems in the facility, and especially in the console, must be addressed 'n a more timely fashion. Tt is the second situation which we have reported to the NRC recently which could probably have been avoided by a better preventative maintenance program. The first, the sticking "up button" on a control rod occurred after severta occurrences of sticking "down buttons" which proved to have the identical cause. The current problem occurred following several years of noisy operation of both the log and linear channels occurring during and immediately following the switching into the automatic mode of operation. This
" noise", including associated linear power SCRAMS but not the identical problem seen on this occasion, has plagued the facility since 19681' However, reducing these problems sooner would most likely have prevented this occurrence.
- 2) The scope of console checkout included in the annual checkout should be carefully examined. Annual testing of all instrument circuits, voltages, calibrations, etc. could improve operations; however, attempting to "fix" something which is "not broken" could cause additional problems. Mr. Begel has suggested that a procedure be irafted of determining voltages, etc. at test points throughout the console; without removing boards or making adjustments in the system. Criteria could then be established which would trigger more detallad tasting and adjustments.
3)
A better method needs to be developed to insure that information about previous problems and the k rwiedge gained from correcting them is transferred to tuture manag' .ient and operators. It was difficult to obtain information from past related rnalntenance operations. The Acting Director will evaluate various possibilities related to this itein.
- 4) The SOP-01 instructions on checking the calibration of the log channel need to be revised to match the GA Manual Procedure to improve the calibration of this channel. This change will also prevent the log and linear pen from overlapping at full power.
\
4
~
i
- 5) Operator training / retraining needs to be upgraded to address the handling of "off-normal" occurrences in much more detal than it currently is. The first priority is to have the instruments operate as well as we possibly can by conducting periodic preventative maintenance, and repairing or at least evaluating fluctuations as they arise. However the console is old and these situations will probably become more common. New and unusual situations will arise which require operators to make rapid judgements and any errors must by on the conservative side.
6)
Serious consideration needs to be given to installing the wide-range channel donated by Northrup. This would provide a second, completely independent, linear power channel with SCRAM capability reducing the possibility of a Tech Spec violation in case of a failure.
More importantly, it would replace both the mode selector switch and the automatic sorvo cnntrols (original equipment), which have given us headaches for years, with circa 1980 afectronics.
- 7) At the suggestion of Mr. Ormond, we are looking for a strip chart recorder to connect temporarily to the percer't power channel. This would provide additional information for diagnosis in the rennte possibility that there should be a recurrence of this anomaly.
- 8) The experience of the w, tor staff working witn Mr. Bob Ormond on this matter Indicates tis. nis addition to Reed will allow all of the scit7ce Jepartments to function more smoothly, and in the case of the reactor at least, more safely in the future. His presence allows us to believe that facility improvements such as installation of a wide-range channel are possible, c
- ~ '
i 'f . 'h
+
q 001 1- 0 19M Docket No. 50 288 Dr. J. . Michael Pollock, Acting Director 1 Reed Reactor Facility r Reed College.
3203 SE Woodstock Blvd. l' Portland,- Oregon 97202 Ii f. $ .. ., .
Gentlemen: ;
SUBJECT:
' INITIAL EXAMINATION REPORT NO. 50-288/0L-91-01 '
the NRC administered initial examinations to employees of On August-26, 1991, your facility who'had applid for a license to operate your Reed Reactor' Facility. Thi evaluation was conducted in accordance with NUREG-1021,At the cor.
" Operator Licensing Examiner Standards,' Revision 6.
examination, the examination questioris and preliminary findings;were discussed with those members.of your staff identified in the enclosed report.
In accordance with 10 CFR 2.790 of the Commission's regulations, a copy of--
this letter and the enclosures vill be placed in the Public Document Room.
Should you have any questions concerning this examination, please contact me on (301).492-1031.
Sincerely, k l -
Robert H. Gallo, Chief m
Operator Licensing Branch.
Division of Licensee Performance and Quality Evaluation, NRR
Enclosures:
- 1. Initial Examination Report No. 50-288/0L-91-01
- 2. Facility comments and NRC resolution of comments
- 3. Examination and answer key (R0/SRO). Y cc w/encls:
' See taxt page -
Distribution: OLB RF JCollins i:
JWRoe PMQualls, RV PIsaac COThomas- PDR l
RGallo LPDR JCaldwell TMichaels, PM WDean . Document Control. .
l PDoyle .
LTrember,LFDCB.(MNBB-4503)
Facility File (DMcCain) ; e OLB,Non-Power RF (DMcCain) t >
0 .DLP0 /
0LB:Djfg ),
L ..
Q - JCaldwell RGa p N
" ai :dlm -PDoyle 10/ef/91 l 10/ 7 /911 10/7 /91- C; g J4MTC%, 3_10/4LQ1 N
L k p -= = ,n til'~ e A w / iA#5 ..~ ), 3
. -. . '- --~ a. ,
Docket No. 50-288 Reed College CC:
Director, Gregon Department of Energy 528 Cottage Street, N. E.
Salem, Oregon 97310 Mayor of City of Portland 1220 Southwest 5th Avenue Portland, Oregon 97204 Administ rator Siting and Regulation Oregon Department of Energy Labor and Industries Building Room 111 Salem, Oregon 97'H 0
ENCLOSURE 1 L S.- NUCLEAR REGULATORY COP 91ISSION OPERATM LICENSING INITIAL EXAMINATION REPORT REPORT _NO.: 50-288/0L-90-01 FACILITY DOCKET NO.: 50-289 FACILITY LICENSE NO.: R-112 Reed Reactor facility FACILITY:
EXAMINATION DATES:
August 26-28, 1991 Paul V. oyle, Chief 'xaminer EXAMINER:
SUBMITTFD BY: R 1/ &
ul V.'D:iyle,Tpief xaminer M ~ 7' i/
Date' APPROVED BY:
bez .
aldwell, Chief jo-9-9/
Date' Jpes .
Non- Reacter-Section Operator Licensing Branch Division of Licensee Performance and Quality Evaluation, NRR
SUMMARY
NRC administered written and operating (RO) applicants.
examinations to th The three R0 three Senior Reactor 0perator Upgrade (SROU) candidates. The three SROU-applicants failed the written portion of the examination. candid issued the appropriate licenses.
?
A
-c' gaW Y
~, W"ga.
g,, o.- gg1; m 4
.p. !
REPORT DETAILS
- 1. Examiners:
Paul V. Doyle, Jr., Chief Examiner Frank Collins, Examiner Patrick J. Isaac, certification
- 2. Results:
R0 SR0 Total (Pass / Fail) 1Efss/ Fail) JPass/ Fail)
NRC Grading: 0/3 3/0 3/3
- 3. Written Examination:
The written examination was administered on Auguct 26, 1991 to three R0 candidates. One candidate failed one section and the other two candidates failed the whole examination.
The facility's written examination comments and the NRC's resolution to those comments are found in Enclosure 2.
> 4. Operating Examinations:
Operating Examinations were administered on August 27 and 23, 1991 to One R0 candidate failed the three R0 and three SRO Upgrade' candidates.All the other candidates passed operating examination.
the examination.
S. Exit Meeting:
Personnel attending: Douglas Bennett, Provost, Reed College Michael Pollock, Director, Reed Reactor Facility
-Paul V. Doyle, Jr., Chief Examiner Frank Collins, Examiner Patrick J. : Isaac, OLB Headquarters The examiners identifiec two generic program weaknesses.
l .
First, several of the candidates displayed an unfamiliarity with_
L Technical Specifications, in that'they could not read'ly locate L
survelliance requirements in Tech Specs.
Second, the-examiners found that the majority of.the initial applicants showed some weakness in their comprehension of Reed's administrative procedures and practices that govern reactor operation and
' documentation.
The licensee's management acknowledged these findings and committeo to resolve these issues.
a ENCLOSURE ?
FACILITY COMMENTS AND NAC RESOLUTION OF COMMENTS SECTION A Dggition 001:
Removing the The reactor has been stable at 20 watts for about an hour.
source from the core causes reactor power to:
- a. Decrease since the reactor is undermoderated
- b. Increase due to an increase in the amount of moderator
- c. Stay the same due to Keft being constant
- d. Decrease due to fast neutrons non-leakage probability increases Answer 001:
0 Reference 001:
Reed R0 Requal Exam.
Facility Comment ?>01:
This question was taken from Reeo's 1991 requal exam and, while the answer given is a correct description of reactor behavior, operators knowing this.
reactor is "never critical" (see attached letter from BothL.ofRuby tneseto NRC) and 2) a vare prohibition on removing the source from the core. incorporated Thus, the in in effect throughout the careers of most of our current operators.
l anticipated answer would be that "the reactoralthough Unfortunately, power would decrease due to our internal of source neutrons from the core."
which should have been omitted, this opinion was not i
! along with the facility material for this NRC exam.
!- Suggested reso'ution: Delete this question from the exam, i
NRC Resoluttor_QQl:
Comments accepted. This question will be deleted from the written examination and the EQB.
i i
l-
3ECTION A Ouestinn 007:
Why does the effect on reactivity by the Fuel Temperature Coefficient (FTC) decrease as fuel temperature incre'ses?
- a. The water density decreases, causing the neutron resonance escape probability (p) to decrease.
- b. The neutron thermal utilization factor (f) predominates in its effect on the neutron life cycle,
- c. The neutron energy resonance absorption peaks broaden less for the same degree of fuel temperature change.
- d. More neutrons leak out of the core.
Answer 007:
C Seference 007:
RRF Training Manual (pg 12-6 and 12-7, Facility comme't n 007:
The wording on this question is sufficiently confusing to require In our training seminar, we stress the consideration of a second answer. A large prompt Negative Temperature Crefficient of the TRIGA raattor.
percentage of this PNTC results from the H in the U,2r-H increasing neutron learsge (SAR Section E). Since we often refer to the " fuel" in the TRIGA reactor as consisting of an intimite mixture of U and Zr hydride, it is quite likely that some operators wovid equate " feel temperature coefficient" with the PNTC and thus respard with answer d.
Suggested Resolution: Accept either c) or d).
NRC Resolution 007:
Agree. Answer key has been modified to accept "c" and "d" as correct.
SECTION A Duett.iooJdi:
is 0.006 delta-X/K at full power, which one of the Assuming the Samariun worthfollowing is the Samarium worth 10 days a
- a. Estentially zero
- b. Essentially 0.006 delta-K/K
- c. Less than 0.006 delta-K/K but greater than zero
- d. Greater than 0.006 delta-K/K Answer 011:
d Reference 011:
RRF Training 14anual (pg.12-7-21) f3cility Comment 01_l:
None NRC Comment Oll:
Equilibrium Pm-149.
Samarium peaks following a reactor s Answer b is also correct.
Essentially Samarium worth does not change.
NRC Reso]ution 0}.1: Answer key has been modified to accept both answers.
Accept either b er o.
ENC:.05URE 3 NR", Official Use Only L
Nuclear Regulatory Comission Operator Licensing Examination This documert is removed from Official Use Only category on date of examination.
NRC Official Use Only
9 E
U. S. NUCLEAR REGULATORY COM41SS10N l NON-POWER REACTOR LICENSE EXWINATION l
FAC1111Y: Reed Ins':.
REACTOR TYPE: IRIGA DATE ADMINISTERED: 91/08/26 REGION: 5 CANDIDATE:
LICENSE APPLIED FOR: .
INSTRUCTIONS TO CANDIDATE Answers are to be written on the answer sheet provided. Attach any answer sheets to the examination. A 70% in each section is required to pass the examination. Examinations will be picked up three (3) hours after the examination starts.
% OF CATEGORY % OF CANDIDATC'S CATEGORY -
SCORE VALUE CATEGORY VALUE TOTAL A. REACTOR THEORY, THERMODYNAMICS 20.00 33.33 -
AND FACILITY OPERATIMG CRARACTERISTICS 20.00 33.33 B. NORMAL AND EMERGENCY OPERATING PROCEDURES AND RADIOLOGICAL CONTROLS t
33.33 C. PLANT AND RADIATION MONITORING 20.00 SYSTEMS TOTALS 60.00 FINAL GRADE All work done on this examination is my own. I have neither given nor received aid.
Candidate's Signature
L l
NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS DJring the administration 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. After the examination has been completed, you must sign the statement on the cover sheet indicating that the work is your own and you ht.e not received or given assistance in completing the examination. This must be done after you complete the cxamination.
- 3. Restroom trips are to be limited and only one candidate at a time may leave. You must avoid all contacts with anyone outside the examination rots to avoid even the appearance or possibility of cheating.
- 4. Use black ink or dark pencil only to facilitate legible reproductions.
S
- 5. Print your name in the blank provided in the upper right-hand corner of the examination cover sheet.
- 6. Fill in the date ca the cover shest of the examination (if necessary).
- 7. The point value for each question is indicated in parentheses after the question. The amount of blank space on an examination question page is NOT an indication of the depth of answer required.
B. If the intent of a question is unclear, ask questions of the examiner only. -
- 9. When turning in your examination, assemble the completed examination with In addition, examination questions, examination tids and answer sheets.
turn in all scrap paper.
< 10. To pass the examination, you must achieve at letst 70% in each category.
- 11. There is a time limit of (3) hours for completion of the examination.
- 12. Wher you are done and have turned in ycur examination, leave the examin- $
ation area as defined by the examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoked.
L 1
.q
.~ -
Page 1 A. RX THEORY, THERM 0 & FAC OP CHARS .
ANSWER SHEET
~
Multiple Choice (Circle or X your. choice)
If you change your answer, write ycur selection in the blank MULTIPLE CHOICE b c d 001' a b c d __
002 a b c d _,_,,_
003 a-b c d ,
004 a b c d 005-' a b- c d
.006 a 007 a b c' d ___,__ -
b c. d 008 a ..,
b c d ._.___
009- a b c J __,_,,
010. a a b c 'd _.
011 a b c d' 012.
b c d __,,,__
. 013 a c' d 014 -a' b __,,,,,,,_ .,
b c d __,,,_
015 a b- c d _ _ ,___
- 016 a
- b c d __,,,_,,,,,,
017 a a b. c d ______
018 b e d . _,,__,,,
019 .' a b .c d.
020 . a.
- )
(*****.END OF CATEGORY A y , # y- h -.-,s , y *
. . ... . . - - . . . . - - . - - .. . . . . - - -. . - - .- . _ - . . _ - - . . - . . . . . . ~ -_
4 7
Page 2
- 8. NORMAL /EMERG PROCEDURES 1 RAD CON ANSWER SHEET Multiple Choice'(Circle or X your choice) write your selection in the blank If you change your answe-MULTIPLE CH01CE a b c d 001
' a b c d 002 a b c d 003 a b c d _
004 a b c d 005 006 a b c d __
b c d 007- a 008 a b c d __ _
009 a b c d __
a b c d 010 a b c d 011 b c ~d 012 a _
~
b c d 013 a b c d _.
014 a b c d 015 a 016 a -b c d.___ .
017 a b c' d-b c d 018 ;a:
b' c- d-019- a a
~
b c d-. _
020
.(***** END OF CAiiGORY B ***** ).
k 6
- ' ha. # , 5,. r-v+-. s ,e- s- 4 -e -
t
v Page 3 C. PLANT AND RAD MONITORING SYSTEMS ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blank HULTIPLE CHOICE a b c d 001 a b c d _
002 a b c d 003 _
b c d 004 a b c d 005 a b c d 006 a b c a 007 a b c d __
008 a b c d __
009 a b c d 010 a 011 a b c d_
b c d 012 a b c d 013 a b c d _
014 a b c d 015 a 016 a b c d_
b e d 017 a b c d _
018 a b c d 019 a
! b c d ___
020 a
(**'** END OF CATEGORY *****)
C END OF EXAMINATION*****)
(*****
e L
A. RX THEORY, THERMO & FAC OP CHARS Page 4 QUESTION: 001 (1.00)
The reactor has been stable at 20 Watt for about an hour. Removing the source from the core causes reactor powtr to:
- a. Decrease since the reactor is undermoderated
- b. Iacrease due to an increase in the amount of mod.-rater.
- c. Stay the same due to Keff being consttnt.
- d. Decrease duc to fast neutron non-leakage probability increases. ,
QUESTION: 002 (1.00)
As power level increases, the Prompt Negative Temp. Coefficient (PNTC) causes:
- a. 238U to absorb neutrons over a wider range, thus decreasing the number of neutrons available for fission with 235U.
- b. Doppler resonance effects to decrease.
- c. The hydrogen atoms in the ZrHE to slow down more neutrons.
d More thermal neutron absorption by the moderator.
QUESTION: 003 (1.00)
Which ONE of the following statements correctly describe; the influence of delayed neutrons on the neutron life cycle?
- a. Delayed neutrons decrease the average period of a reactivity addition because they thermalize more quickly than orompt neutrons,
- b. Delayed neutrons take longer to thermalize bec:use they are born at higher energies than prompt neutrons.
- c. Delayed neutrocs re;se the length of the average neutron generation time to increase.
- d. Delayed neutrons are born later than prompt neutrons 'and make up a larger fraction of the fission neutrons.
(***** CATEGORY A CONTINUED ON NEXT PAGE *****)
t A. - RX THEORY, THERMO & FAC OP CHARS.
Page 5 QUESTION: 004 (1.00)
Which alteration or change to the core will most strongly affect the thermal utilization factor.
- a. Build up of fission products in fuel.
- b. Removal of moderator. ,
- c. Addition of 238U
- d. Removal of a control rod.
QUESTION: 005 (1.00)
With the reactor on a constant period .which transient requires ~the. longest time to occur?
A reactor povsr change of:
- a. 5% power -- going from 1% to 6% pwr
- b. 10% power -- going from 10% to 20% pwr-
- c. 15% power -- going.from 20% to 35% pwr.
-d. 20% power -- going from 40% to 60% pwr i
l QUESTION: 006 (1.00)
What is the stable Rx-period which produces a power rise from 1. watt l to 5~KW in 186 sect?
- a. 10 secs.
- b. 22 secs.
- c. 30 secs,
- d. 116 secs.
f (***~~ CATEGORY A CONTINUED ON SEXT FAGE.*****)
9- g -_ _ - * - -
. . -. . - - = - -. . .. .
RX THEORY, THERMO & FAC OP CHARS Page 6 A.
QUESTION: 007 (1.00)
Why does the effect on reactivity by the Fuel Temperature Coefficient (FTC) decrease as fuel temperature increases?
- a. The water density decreases, causing the neutron resonence escape probability (p) to decrease,
- b. The neutron thermal utilization factor (f) predominates in its effect on the neutron life cycle.
- c. The neutron energy resonance absorption peaks broauen less for the same degree of fuel temperature change.
- d. More neutrons leak out of the' core.
QUESTION: OGe (1.00)
Which DNE of the following is the reason for the 80 second period following a reactor scram?
' a. U-235 affinity for source neutrons,
- b. Fuel temp, coefficient adding positive reactivity.
- c. Longest lived delayed neutron precursors dtcay constant.
- d. Amount of negative reactivity added on a scram. exceeds the shutdown margin.'
(***** CATEGORY A CONTINUED ON NEXT PAGE *****)
~
4 t
l 5
l
a- ;
Page 7 A. RX THEORY, THERNO & FAC OP CHARS QUESTION: 009 (1.00)
Which ONE of the following stateents describes Count Rate characteristics after a control rod withdrawal with the reactor subcritical? (Assume t reactor remains subcritical.)
- a. Count Level will rapidly increase (prompt jump) then gradually increase to a stable value.
- b. Count Level will rapidly increase (prompt jump) then gradually decrease to the previous value,
- c. Count level will rapidly increase (pronipt jump) tc a stable value,
- d. There will be no change in Count Level until criticality is achieved.
QUESTION: 010 (1.00)
In a subcritical reactor, Keff is increased from 0.861 to 0.946. Which ONE of the following is the amount of reactivity that was added to the core?
- a. 0.085 delta-K/K
- b. 0.220 delta-K/K
- c. 0.104 delta-K/K d.0.12bdelta-K/K QUESTION: 011 (1.00)
Assuming the Samarium worth is 0.006 delta-K/K at
- a. Essentially zero.
- b. Essentially 0.006 delta-K/K
- c. Less than 0.006 delta-K/K but greater than zero
- d. Greater than 0.006 delta-K/K
- )
(*****-CATEGORY A CONTINUED ON NEXT PAGE
7 Page 8-A. RX THEORY, THERM 0 & FAC OP CHARS QUESTION: 0'. 2 (1.00)
Which ONE of the following is a correct ',atement concerning the faitors affecting control rod worth?
- a. Fuel burn up causts the rod worth for periphery rods to decrease.
- b. Fuel burn up causes the rod worth to increase in the center of the core.
- c. The withdrawal of a rod causes the rod worth of the remaining inserted rods to increase,
- d. As Rx power increases rod worth increases.
QUESTION: 013-(1.00)
The Rx is shutdown by 5% delta-K/K with a count rate of 100 cps on the J start up channel. Rods are withdrawn until the count rate is 1000 cps.
Which ONE of the following is the condition of the reactor after the rods are withdrawn? .
- a. Critical with Keff = 1.0
- b. Suberitical with Keff = 0.995 ,
- c. Suber;tical with Keff - 0.950
- d. Supercritical with Keff - 1.005
(***** - CATEGORY A' CONTINUED Oh NEXT PAGE *****)
l l
Page g A. RX THEORY, THERM 0 & FAC OP CHARS QUESTION: 014 (1.00)
Assume the following rod wortns: Safety and Shims are $4.25 each, Reg, is
$1.75 and Core excess is $2.5 . Calculate the Shutdown Reactivity with the Shim rod stuck all the way out.
- a. $3.0
- b. 53.5
- c. $6.0
- d. $7.25 QUESTION: 015 (1.00)
The Reed College Triga Rx is slightly undermoderated. Which of the folic., wing statements correctly describes the reactor operating characteristic of "Undermodulated*?
- a. A decrease in core water temperature kill cause a negative _
reactivity response.
- 5. Reducing the amount of moderation will cause a positive reactivity response.
- c. An increase in core water temperature will cause a positive reactivity response.
- d. A decrease in core water dentity will cause a negative reactivity response.
(***** CATEGORY A CONTINUED ON NEXT PAGE *****)
L _ ______ _ _ _ _ _ _ _ _ _ _ _ _ __ _
..u w. - ,
~ > , -
(
l Page 10 A. RX THEORY, THERM 0 & FAC OP CHARS i QUESTION: 016 (1.00)
In a just critical reactor, removing one dollar worth of reactivity will.
cause:
- a. The resultant period to be a function of the prompt neutron lifetime.
- b. The prompt neutron term to become unimportant
- c. The reactor period to be equal to (#-s)
Ap
- d. A sudden drop in neutron flux.
QUESTION: 017 (1.00)
Which statement illustrates a characteristic o'f Subcritical. Multiplication?
~
- a. As Keff approaches unity (1), for the same. increase in Keff, a-greater increase in nettron population occurs.
- b. The r. umber of neutrons gained per. generatian gets larger for each succeeding-generation..
- c. The number of fission neutrons remain constant for each generation.
- d. The number of source neutrons decreases for each generation.-
J
- )
(***** CATEGORY A CONTINUED ON NEXT PAGE ,
4 i
w
I '
f
_f<> :.-, 0 z;s s. 9
,9O, o t O-IMAGE EV Al.U ATION ((#[, / '*
qpf y & TEST TARGET (MT-3) ,7f
'g/ y N k",ji ,
- t?gg
$ 4 1.0 :2" 3 yr g2 7r !RI 2.0 l,l mea FB==8 n -,
i.6
~ I.25 T 1.4L== ii == a!)[
4- 150mm I
6" --
4_____
?N'h r 4$ Ay
.h ~ + ? %p,o ga egy ,, %,. - .
g,6,s.sbf-
)
.~ ,
mw . gs e i gkD' h./ ) v m f((6 '
/ .
go
/ ;
.o;l
Q:.
A 4 g a \\\g q.
4e p @,
'0 s . IMAGE EV ALU ATION
//[ /
\ /,o// '. k' TEST TARGET (MT-3) f /f (p
'Ok 3
\
l.0 1" T
x22
. a n =.
j,l . - -mm
$1] l.8 hae=s
'l.25 ![ 1.4 11.6 s ,
= .!i, == p
b 4___.____._.
- 150mm
_ _ _ _ _ _ _ _ _ . _ . - - - 6" - - -
4 e
l
<!$ pyj\\ x
- j% 5 #%r4 + o
+ ;e y / >, f/ ; y Qnf4x v w, O - w(h9 Ghi'
<{:s a \
i l f- , f-;$$
., am ' , ~ :w.,ideil w)
I ,, .
A
/
e.
n
. q'E 0 -
, n.
'?"
~
4'.
~
IMAGE EVALUATION
,C %
I
\g-V g
TEST TARGET (MT-3)
/ <['/.,O,'
d ' 'V a,%,6,, '
&fgb
\\
4Q
- $' ll 4 l.0
? :22 -
3 Mi 2.0 j,j
~
p t p La em "1.25 i 1.4 il i 6 = tlw.6
- - - - - 15 O ni n, - - - - - - - -
4._____. - . - - -
4 _ .___ . . _ _ _ _ _ _ _ . _ _ . . -
6" ---
x% .e"s,Ao
+
,p; %' 4h., #p,
~ '4 ,,; ,. ,
O*\
-~. -
-(:: , .b , ,
, . q, j;, ,; / p _"a
^'~
4g4 .-
'r{- yv 0;, ,y
,,p ;. : g;;sh0 ._ _ ,
i I
j Page 11 l A. RX IHEORY, THERMO & FAC OP CHARS QUESfl0N: 018 (1.00)
Which statement best describes the heat transfer mechanism at the Reed College Reactor?
- a. From the fuel center line, heat is transferred to the surface of the fuel rod by convection and is carried into the coolant by conduction.
- b. Heat is transmitted to the fuel rod surface by thermal radiation and carried to the coolant bt conduction.
- c. Heat conducted to the surface of a fuel rod is carried into the coolant and out of the system by convection.
- d. The temperature distribution from the fuel center line to the coolant is linear.
QUESTION: 019 (1.00)
Which statement best describes Xe-135 behavior following a Rx Scram?
- a. Xenon inneentration decreases due to production rate from fiss M stops,
- b. Xent,r. concentration decreases due to production rate from I-135 decay increasing.
- c. Xenon concentration increases due to production rate from Pm-149 increasing.
- d. Xenon concentration increases due to I-135 decay exceeding Xe-135 decay.
- )
(***** CATEGORY A CONTINUED ON NEXT PAGE
RX THEORY, THERMO & FAC OP CHARS Page 12 A.
4 QUESTION: 020 (1.00)
The reactor was shutdown after an extended two week, high power, run at 200 kw to irradiate a specimen. How long will it take for the MAXIMUM Xenon poisou effect to occur?
- a. Imediately,
- b. 8 to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- c. 35 to 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />.
- d. I to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
(***** END OF CATEGORY A *****)
Page 13 B. NORMAL /EMERG PROCEDURES & RAD CON QUESTION: 001 (1.00)
To account for fluctuations which might take the power above the license limit, RRF Procedure 50P-04, Rx Operations, precautions the operator not to exceed a power level of:
- a. 240 KW
- b. 250 KW
- c. 210 KW
- d. 225 KW QUESTION: 002 (1.00)
As defined by Technical Specifications, which ONE of the following statements is NOT a definition for a shutdown reactor.
- a. The console key is in the "0FF" position and the key is removed from the console and under the control of a licensed operator.
- b. No work is in progress involving fuel handling or refueling operations or maintenance of its control mechanisms,
- c. The minimum shutdown margin, with the most reactive of the operable control elements withdrawn shall be 0.6 delta K/K,
- d. Sufficient control rods are inserted so as tc assure the reactor is suberitical by a margin greater than 0.7 delta K/K cold without Xenon. .
(***** CATEGORY B CONTINUED ON KEXT PAGE *****)
i Page 14 B. NORMAL /EMERG PROCEDVRES & RAD CON QUEST 10N: 003 (1.00)
A point sourca of gamma radiation measures 50 mr/hr at a distance of 5 ft.
What is the exposure rate (mr/hr) from the source at a distance of 10 ft.
- a. 25 mr/hr
- b. 12.5 mr/hr
- c. 6.25 mr/hr
- d. 17.5 mr/hr QUESTION: 004 (1.00)
In accordance with 10 CFR 20 (Standards for protection Against Radiation),
which ONE of the following is the radiation dose standard for individuals in restricted areas per calendar quarter 7 (Assume that NRC Form 4 is on file.)
- a. Whole body - 1.25 Rem Active blood forming organs 1.25 Rem Hands and forearms - 7.5 Rem Skin of whole body - 18.75 Rem
- b. Whole body - 3.75 Rem Active blood forming organs 1.25 Rem-Hands and forearms - 7.5 Rem Skin of whole body - 18.73 Rem
- c. Whole body - 1.25 Rem Active blood forming organs 1.25 Rem Hands and forearms 18.75 Rem Skin of whole body - 7.5 Rem
- d. Whole body - 3.75 Rem Active blood forming organs 3.75 Rem Hands and forearms - 18.75 Rem Skin of whole body - 7.5 Rem
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
l
\
Page IS l B. NORMAL /EMERG PROCEDURES-& RAD CON l!
r QUESTION: 005 (I.00) 10 CFR 20.105 sets " permissible" levels of radiation in unrestricted areas.
What is the 10 CFR 20 whole body MAXIMUM dose limit, in any period-of one calendar year 'n an " unrestricted" area? - ,
- a. 1.25 Rem
- b. 500 arem
- c. 200 mrem
- d. 100 mrem QUESTION: 006 (I.00)
Which DNE of the following statements is an expression of the ALARA program?
- a. Reduces radiation exposure to the public.
- b. Reduces the radiation exposure to radiological workers. T c - Reduces the chances of a nuclear accident occurring.
- d. Reduces the amount of radiation emitted during reactor operations.
QUESTION: 007 (I.00)
Which ONE of the following is the correct _ posting if the radiation level in the area is 75 mr/hr?
- a. CAUTION RADIA110N AREA
- b. CAUTION RADI0 ACTIVE _ MATERIAL (S)
I
.c. CAUTION AIRBORNE RADI0 ACTIVITY AREA
- d. CAUTION HIGH RADIATION AREA-l
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
L i
Page 16 B. NORMAL /EMERG PROCEDURES & RAD CON QUESTION: 008 (1.00)
You are the RO on duty during an experiment. You discover that the Core Excess Reactivity Worth is 2.30% delta-K/K.
In accordance with SOP-03, what actions should be taken?
t
- a. Scram the Rx / Notify the Rx Supervisor.
- b. Shutdown the Rx / Notify the Senior Health Physicist.
- c. Shutdown the Rx / Record the time of the shutdown in the Purpose Stamp.
- d. No action required. Technical Specifications have not been violated.
QUESTION: 009 (1.00)
Which ONE of the following conditions is a violation of Technical Specificttions, Reactor Pool?
- a. Conductivity of the pool water is 2.2 micrombos per centimeter averaged over one month,
- b. Radioactivity in the pool water is 0.2 micro Ci/ml.
- c. Pool water ph is 5.7
- d. Bulk temperature of the coolant is 45 degrees C during reactor operation.
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
B. NORMAL /EMERG PROCEDURES & RAD CON Page 17 QUESTION: 010 (1.00)
In accordance with Technical Specifications, which ONE of the following statements is TRUE?
- a. Each fuel experiment shall be controlled such that the total inventory of Iodine isotopes 131 thru 135 in the experiment is no greater than 1.5 curies,
- b. The reactivity worth of any individual in-core experiment shall not exceed $2.00.
- c. Experiments containing materials corrosive to Rx components shall not be irradiated in the Rx.
- d. Explosive experiments shall be doubly encapsulated. c QUESTION: 011 (1.00)
A reactor operator may be called upon to verify the proper labelir.g of a container of radioactive materials being prepared for shipping. One of the radiation levels recorded must be the Maximum External Dose Rate. -
How far must the operator hold the detector probe from the container surface for the SECOND radiation measurement?
- a. 0.5 inches
- b. 2 centimeters
- c. I foot
- d. I meter
(***** CATEGORY B CONTINUED CN NEXT PAGE *'d***)
B. NORMAL /EMERG PROCEDURES & RAD CON Page 18 QUESTION: 012 (1.00)
In accordance with SOP-04, Reactor Operations, which statement describes when a STATUS STAMP is to be completed during operations $2 constant power.
- a. Whenever 30 minutes pass without an intervening shift change.
- b. Whenever a Technical Specification limit is exceeded,
- c. Whenever the reactor operator is temporarily replaced to go to the bathroom.
- d. When rods are moved to maintain the flux profile.
QUESTION: 013 (1.00)
Which ONE person may authorize entry into the Rx Bay by visitors under 18 years of age?
- a. Reactor Supervisor
- b. Director, Reed Reactor Facility
- c. Any licensed operator 2
- d. Associate Director, Raed Reactor Facility QUESTION: 014 (1.00)
During a long (8 hr) reactor run, the operator notices that the GSM reading has been drifting slowly upward.
What action is the operator required to perform when the GSM goes into alarm,
- a. Notify Oregon Dept. of Energy
- b. Call Police Dept. (911 on Red phone)
- c. Immediately secure the main facility circuit breaker
- d. Scram the reactor
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
1
1 i
B. NORMAL /EMERG PROCEDURES & RAD CON Page 19 I
QUESTION: 015 (1.00) :
In the event of a pool level alarm and visual-indication of abnormal loss of pool water, which ONE of the following actions is the reactor operator ~
NOT allowed to perform if he is the Emergency Coordinator. 4
- a. Shtdown the reactor
- b. Isolate the Pool
- c. Notify a member of the ENCL
- d. Terminate the emergency QUESTION: 016 (1.00)
The Reactor Facility must be evacueted due to a fire. Per Reed's Emergency Implementation Procedures, where do Facility personnel assemble?
- a. Chemistry Laboratory
- b. Reactor back lawn
- c. Reactor parking area
- d. Chemistry building hallway QUESTION: 017 (1.00)
Which of the following does NOT require NRC approval for changes?
- a. License
-b -Requalification plan
- c. Emergency Implementation Procedures
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
B. NORMAL /EMERG PROCEDURES & RAD CON Page 20 l
QUESTION: 018 (1.00)
SOP-43, Power Calibration, stet 23 that during power calibrations, radiation levels in the Reactor Bay ray be elevated. Which stateraent best describes why?
- a. Reactor will be operated at higher than normal power.
- b. Fadioactive gases (N-16) might reach reactor room.
- c. The pool level is lower than normal.
- d. All rods are withdrawn from the core.
QUESTION: 019 (1.00)
No special experiment shall be performed until the proposed experimental procedure has been reviewed and approved by which ONE of the following?
- a. Reactor Operations Committee - Director RRF - Reactor Supervisor
- b. Reactor Supervisor - Director RRF - Reactor Safety Committee
- c. Director RRF - Reactor Supervisor - Reactor Health Physicist
- d. Licensed SRO - Reactor Operations Committee - Director RRF QUESTION: 020 (1.00)
Which ONE of the following scrams is NOT required by Technical Specifications?
- a. Linaar channel
- b. % power channel
- c. Manual
- d. Log channel
(***** END 07 CATEGORY B *****)
4
.3 r
[ .
C. PLANT E!D RAD MONITORING SYSTEMS Page 21' l
QUESTION: 001 (1.00)
How is water or condensation removed from the rotary specimen rack?
- a. The pool is periodically drained to allow the condensation to evaporate.
- b. Water absorbing material is placed in a perforated specimen tube, which is inserted in the rack.
- c. An electric heater is placed in an insulated specimen tube, which is inserted'in the rack.
- d. An inert gas is inserted in the rack to blow out the condensation.
QUESTION: 002 (1.00)
The pool water level must be maintained at a specified level for proper skimmer operation. Why is this pool level necessary?
- a. Above this pool level, the reactor water system pumps will lose suction and fail.
- b. Below this level, air entry could interfere with proper operation of the rabbit tube.
- c. Above this level, the-water pressure could damage the purification system filters.
- d. Below this level, air entry could damage the purification system demineralizers.
(***** CATEGORY C CONTINUED ON NEXT PAGE *****)
Page 22 C. PLANT AND RAD MONITORING SYSTEMS QUESTION: 003 (1.00)
+
The neutron absorber in Reed's reactor control rods is:
- a. Aluminum oxide
- b. Zirconium hydride
- c. Graphite powder
- d. Boron carbide QUESTION: 004 (1.00)
Which ONE of the following nuclear channels provides the operator with a continuous record of neutron flux from approximately one watt to full power?
- a. Period channel
- b. Log power channel
- c. Count rate channel
- d. Linear power channel QUESTION: 005 (1.00)
Which ONE is NOT an input to the rtegulating Rod Servo?
- a. Linear power channel
- b. % demand potentiometer
- c. Rod raising interlock
- d. Period channel
(*+*** CATEGORY C CONTINUED ON NEXT PAGE *****)
C. PLANT AND RAD MONITORING SYSTEMS Page 23 QUESTION: 005 (1.00)
Using the drawing of the heating and ventilation systems provided, determine which of the following is the appropriate flow path in the event of an accidental air contamination.
- a. Supply unit A shutdown - supply C operates - Exhaust B operates.
Automatic dampers direct air from the Rx room through absolute filters and up the ventilation stack.
- b. Supply unit C shutdown - Exhauster B shutdown - supply unit A operates to supply air to Rx room thru absolute filter.
- c. All fans operating. Automatic dampers reposition to admit two air changes per hour.
- d. Supply units A and C shutdown - exhauster B is operating taking suction tnrough damper 13 and exhausting up the vent stack.
QUESTION: 007 (1.00)
Limit switches mounted on each drive assembly provide switching for console lights.
idhich one of the statements is FALSE?
- a. The DOWN light indicates that the control rod and rod drive are at their lower limits.
- b. The UP light indicates that the control rod and rod drive are at their upper limits.
- c. When the CONT /0N pushbuttons are depresssd, the ON lights are extinguished.
- d. The CONT side light of the CONT /0N switch goes off less than one second after a scram occurs.
(***** CATEGORY C CONTINUED ON NEXT PAGE *****)
l
e C. PLANT AND RAD. MONITORING SYSTEMS Page 24l QUESTION: 008 (1.00)
What will be the effect of a high differential pressure across the filter on the reactor water pump and the domineralizer flows?
- a. Increase reactor water pump flow and increase domineralizer flow. '
- b. Increase reactor water pump flow and decrease domineralizer flow.
- c. decrease reactor water pump flow and increase demineralizer flow.
- d. decrease reactor water pump flow and decrease demineralizer flow.
QUESTION: 009 (1.00)
Which ONE of the following statements correctly describes the pneumatic transfer system path and operating pressure?
a.-The pneumatic transfer system exhausts to the reactor room and-maintains the transfer system at a positive pressure.
- b. The pneumatic transfer system exhausts to the building exhaust and maintains the transfer system at a positive pressure.
- c. The pneumatic transfer system exhau ts to the reactor room and-maintains the transfer system at a negative pressure.
- d. The pneumatic transfer' system' exhausts-to the building. exhaust and maintains the transfer system at a negative. pressure.-
QUESTION: 010 (1.00)
Which ONE of the following statements is NOT a reason for maintaining a-minimum reactor pool level during reactor operation?
- a. Ensure proper operation _ of the pool skimmer
~
b.-Provide Net Positive Suction Head (NPSH) to the reactor _. water pump
- c. Proper " Dash Pot"" action for the control rods during a scram- '
- d. Provide shielding from the core
(***** CATEGORY C CONTINUED ON NEXT PAGE *****)
=- . . _ . . --. . . -
W t
Page 25 C. PLANT AND RAD MONITORING SYSTEMS QUESTION: 011 (1.00)
Which DNE of the following statements correctly describes the purpose of the PULL ROD in the control rod drive assembly?
- a. Provides rod full out position indication,
- b. Provides a means for manually adjusting rod position by pulling rod out,
- c. Provides rod bottom indication.
- d. Automatically engages the control rod on a pull signal.
QUESTION: 012 (1.00)
Which ONE of the following statements correctly describes the purpose of the potentiometer in the control rod drive assembly,
- a. Provides rod position indication when the electromagnet engages the connecting rod armature.
- b. Provides a variable voltage to the rod drive motor for regulating control rod speed.
- c. Provides potential voltage as required for resetting the -
electromagnet current.
- d. Provides the potential voltage to relatch the connecting rod.
QUESTION: 013 (1.00)
Which ONE of the following is the flow through the primary loop and the cleanup loop?
a.110 gpm total flow with 10 gpm through the cleanup loop b.120 gpm total flow with 20 gpm through the cleanup loop c.110 gpm total flow with 20 gpm through the cleanup loop d.120 gpm total flow with 10 gpm through the cleanup loop
(***** CATEGORY C CONTINUED ON NEXT PAGE *****)
4 I
- C. PLANT AND RAD MONITORING SYSTEMS Page 26 l QUESTION: 014 (1.00)
Which DNE of the following ranges of nuclear instrumentation utilizes an uncompensated ion chamber as the neutron detection device?
- a. Count Rate channel
- b. Log N channel
- c. Linear Power channel
- d. % Power channel QUESTION: 015 (1.00)
Which ONE of the following statements describes the drive speeds of the Shim rod, Regulating rod and Safety rod?
- a. The Shim rod drives at 24 inches per minute, the Regulating and Safety rods drive at 19 inches per minute.
- b. The Shim and Regulating rods drive at 24 inches per minute, the Safety rod drives at 19 inches per minute,
- c. The Safety rod drives at 24 inches per minute, the Regulating and Shim rods drive at 19 inches per minute,
- d. The Regulating rod drives at 24 inches per minute, the Safety and Shim rods drive at 19 inches per minute.
(***** CATEGORY C CONTINUED ON NEXT PAGE *****)
Il
9
"~ Page 27 C. PLANT AND RAD MONITORING SYSTEMS QUESTION: 016 (1.00)
Which DNE of the following describes the action of the rod control system to drive the magnet draw tube down after a dropped rod?
- a. Deenergizing the rod magnet initiates the rod down motion of the draw tube,
- b. Actuation of the MAGNET DOWN limit switch initiates the rod down motion of the draw tube.
- c. Actuation of the ROD DOWN limit switch initiates the rod down motion if the rod drive is withdrawn,
- d. Resetting the scram signal initiates the rod down motion of the draw tube.
QUESTION: 017 (1.00)
Which of the following instruments is used to detect High range Beta-Camma radiation during an emergency condition?
- a. Eberline Model E-140
- b. Model R0-2
- c. CD V - 700 model 6B
- d. CD V - 715 model IB QUESTION: 018 (1.00)
In what region of the Pulse Size vs. Applied Voltage characteristic curve does the fission chamber operate?
- a. Geiger Muller
- b. Limited Proportionality
- c. Proportional
- d. Ioni:ation
(***** CATEGORY C CONTINUED ON HEXT PAGE *****)
C. PLANT AND RAD MONITORING SYSTEMS Page 28 QUESTION: 019 (1.00)
Which DNE of the following will cause a HIGH conductivity reading at the inlet of the demineralizer?
- a. Failure of cooling water heat exchanger
- b. Pool water temperature low
- c. Reactor water system pressure greater than seconde v water pressure l
- d. High reactor water pump flow <
l l
QUESTION: 020 (1.00)
Which ONE of the following radiation monitoring systems will cause a ventilation confinement actuation?
- a. GSM l
- b. RAM
- c. APM i
- d. Portable Survey Meter (GM) l
(***** END OF CATEGORY C *****)
(**********ENDOFEXAMINATION**********)
O l l
raar..nu , L4some rouv j h , .To E% e. m 5T1:N57 wad riac, 2 M.
9@
?@ t v ..
_g 1 o
u s.w. .
I@ ! k@
} r c C
~G'
,, ;; .n.
^
a a
-= ,
- d. Pen.N vm Fou ov m pe g"g my 9,7rg yy A
AI R.
A S ve e ty owiT REACTom Room B ExsAus.T UwiT REAcTom Room _.
+j'E * -r -(-@ e C - SU PPLY UNIT-CouTRoL4 MECH. ROOM (,10, enu Ql. ' 518 D- Ae sotoTE. F i L.T re. R. .
g o
5L H E M AT i c Aia systems M g, Re.Eo REACTOR a m cuse' s/
S Y M f50 L5. I M/ .15 S u p p ty Q E Ti.> ct u s ! D - ft-REAcio2 Roou Q E. ACTOR. Room - o
, Re e .wra Rcom Mscu. Room e _ ,
. CouTaou moem cowtaou Room Rc>o M OuTsios Ascz lurAw.E.
t' MucH.
Ex,HAOST STACK
.m
- c.
EQUATION SHEET Network (out)
Q = a e, AT Cycle Efficiency -
Energy-(in).
Q = a Ah SCR = S/(1-Keff)
Q = UA AT CR 3 (1-Keff): = CR 2 (1-Keff):
26.06 (1.,,p) (1-Keff).
SUR - M=
($ - p) (1 Keff)3 SUR = 26.06/7 - M - 1/(1-Keff) - CR /CRo 3 P = P 10 8
- SDM = (1-Keff)/Keff P = P e"I Pwr-W,E A(1-p)
P= P, l' = 1 x 10'5-seconds -
0-9 '
i T = (f*/p) + [(T-p)/A.r p) T . = f */(p-1) ;
p = (Keff-1)/Keff A.re = 0.1 seconds'2 o . p = 4Keff/Keff' 'O.693-Tif A = 0.0077 A L DR3 03 8 - DR22 08 DR - . DR,e
1 Curie = 3.7x102' dps -l kg = 2.21 lbm'
.1 hp - 2.54x10' BTU /hr- 5
! Mw - 3.41x10 BTU /hr-1 BTU = 778 ft-lbf 'F = 9/5'C + 32.-
{-
- C =_.5/9-('F - 32)-
s
'I,.
y e _. .,,
4 1( [{
Page.29 A .- RX THEORY, THERMO & FAC.0F CHARS ,
i
_ ANSWER: 001 (1.00) b' delel4dL gge
'f*
REFERENCE:
Reed'R0 Requal Exam.
ANSWER: 002 (1.00) a
REFERENCE:
Reed R0 Requal Exam L
F ANSWER: 003 (1.00)
C
REFERENCE:
Reed Reactor Training Manual _(pg. 5-14) .
ANSWER: 004 (1.00)
Ld.
REFERENCE:
Reed Reactor Training Manual (pg. 4-18)
(***** CnTEGORY A CONTINUED ON NEXT PAGE *****)
. ._- .. .2 _ :-,,-.,- _ --
b-
-i RX THEORY,-THERM 0 i FAC OP CHARS'-
Page.30-
- A.
ANSWER: 005 (1.00)
- ANSWER a
REFERLt E:
Reed Training Manual (pg. 5-22)_ _
ANSWER: 006 (1.00)- .,
~}
REFERENCE:
Reed Training Manual (pg 5-15)
ANSWER: 007 (1.00) c or cl f(
REFERENCE:
~
RRF Training Manual (pg 12-6 through 12-7)
ANSWER: 008-(1.00) c
REFERENCE:
RRF Training. Manual
-(***** CATEGORY -A CONTINUED ON NEXT PAGE *****)
Page 31 A. RX THEORY, THERMO & FAC OP CHARS ANSWER: 009 (1.00) a
REFERENCE:
RRF Training Manual ANSWER: 010 (1.00) c
REFERENCE:
Reed Training Manual (pg 12-5-5)
ANSWER: 011 (1.00) d ov b !?El
i:
REFERENCE:
RRF Training Manual (pg 12-7-21)
ANSWEC: 012 (1.00) ,
C RffERENCE:
Reed Training Manual (pg 12-6-21)
(***** CATEGORY A CONTINUED ON NEXT PAGE*****)
4 i
Page 32-
.- A.. RX THEORY, THERMO & FAC OP CHARS -
ANSWER: 013 (1.00) b
REFERENCE:
Reed Training Manual (pg 12-4-21)
ANSWER: 014. (1.00) b
REFERENCE:
Reed Requal Exar 88 89
[
.015 (1.00)
ANSWER:
d
. ~
REFERENCE:
~ Reed Training Manual -(pg' 12-6 / 12-7) ,
i
-ANSWER: .016 (1.00)
~
a .- ;
REFERENCE:
l.-
Reed Operator Requal . Exam 1990 '
/(***** CATEGORY _ - A CONTINUED ON NEXT PAGE
- )
s.
_. [
s .
a , - . , ~r ,,n , - , . .
se p-
Fage 33-A.- RX THEORY, THERMO & FAC.0P CHARS 1 ,
J l
'l
- ANSWER: 017 (1.00) i a
d
REFERENCE:
ll RRF Training Manual-(pg 12-4-22) ,
)
n ANSWER: 018 (1.00)
C
REFERENCE:
Reed Requal Exam 1990 s
f ANSWER: '019- (1.00) d
REFERENCE:
RRF Training Manual (pg 12-7-23 / 12-7-24) l ANSWER: 020-(1.00) b
(
REFERENCE:
RRF Training Manual ('pg 12-7-25)-
(*****! D ' 0F CATEGORY A*****)
g y - ,-m-e -w ..n-, -w .
- r- aw vm--
f Page 34 B. NORMAL /EMERG PROCEDURES & RAD CON ANSWER: 001 (1.00)
REFERENCE:
Reed Tech Specs ANSWER: 002 (1.00)
C
REFERENCE:
}
~ RRF Tech Specs ANSWER: 003 (1.00) b
REFERENCE:
RRF Training Manual ch. 2
" ANSWER: 004 (1.00) c
REFERENCE:
10 Cfr 20 sect. 20.101
(***** CATEGORY B CONTINUED ON HEXT PAGE *****)
Page 35 B. NORMAL /EMERG PROCEDURES 1 RAD CON ANSWER: 005 (1.00) b
REFERENCE:
10 CFR 20 ANSWER: 006 (1.00) b
REFERENCE:
RRt Requal Exam 1991 ANSWER: 007 (1.00) a l
REFERENCE:
~
! ANSWER: 008 (1.00) a l
REFERENCE:
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
e
.i Page 36 B. NORMAL./EMERG PROCEDURES & RAD CON 4 l
l l
ANSVER: 009 (1.00) a
REFERENCE:
RRF Tech Spec ANSWER: 010 .(1.00) a
REFERENCE:
RRF Tech Spus ANSWER: 011 (1.00) d
REFERENCE:
SOP-54 AM, "R: 012 (1.00) .,
c
REFERENCE:
SOP-04
(***** CATEGORY B CONTINUED ON NEXT PAGE *****)
d l
Page 37
- . HORMAL/EMERG PROCEDURES & RAD CON ANSWER: 013 (1.00) b
REFERENCE:
Admini.trative Procedures ANSWER: 014 (1.00) d
REFERENCE:
Emergency Implemer.tation Procedures ANSWER: 015 (1.00) d REFEP.ENCE:
Emergency Irelementation Procedures ANSWiA: 016 (1.00)
C
REFERENCE:
Emergency Implementation Procedures
(***** CATEGORY B CONTINUED ON NEXT PAGE *****) )
9
- - - . _ _ , _ _ _ _ _ _ _ , _ . _ - _ . - - - - - - - _ _ . - - - - - - - - - _ - - _ - - _ ----_-_-.__,__.___-_.a._- - - _ - _ _ , _ , . _ _ . _ , _ - - - - _ _ , _ _ , _ - - _ - - - - - . - _ - -
- l Page 38-B. -NORMAL./EMERG PROCEDURES & RAD CON ANSWER: 017 (1.00)
C I
REFERENCE:
4 RRF 1990 R0 Requal Exam ANSWER: 018 (1.00) b
REFERENCE:
RRF R0 Requal Exam 1988
' ANSWER: 019 (1.00) - ,.
a
REFERENCE:
Administrative Procedures ANSWER: 020 (1.00) d
REFERENCE:
Tech Specs
(***** END OF CATEGORY B_*****) ,
8 m **
w- ~ , e y i-.p. .< ~ * - -- .- y . - - - <m
Page 39 C. PLANT AND RAD MONITORING SYSTEMS ANSWER: 001 (1.00) b
REFERENCE:
GA Triga Maintenance and Operating Manual ,
ANSWER: 002 (1.00) d
REFERENCE:
GA-Triga Maintenance and Operating Manual .
ANSWER: 003 (1.00) d
REFERENCE:
- GA Triga Maintenance and Operating Manual ANSWER: 004 (1.00) ,
b
REFERENCE:
Triga Mechanical Maintenance and Operating Manual (Appendix B)
(***** CATEGORY C CONTINUED ON NEXT-PAGE *****)
=
)
,,,.,-r,e --o,
Page 40 !
C. PLANT AND RAD MONITORING SYSTEMS ANSWER: 005 (1.00) <
C
REFERENCE:
GA TRIGA Mech. MOntenance Manual .
5 ANSWER: 006 (1.00) a
REFERENCE:
RRF SAR Sect. 4.4 ,
ANSWER: 007 (1.00) ,
b
REFERENCE:
GA-TRIGA Electrical Maintenance Manual ANSWER: 008 (1.00) .
d
REFERENCE:
GA TRIGA Maintenance and Operating fanual.
l l (***** CATEGORY C CONTINUED ON NEXT PAGE "***)
l-I-
l
_, . ._ _- _..._.m - , z. _ .. ,
f
- i i
Page 41 C. PLANT =AND RAD MONITORING SYSTEMS i
f ANSWER: 009 (1.00) d ,
REFERENCE:
RRF Description and SAR (pg 5 7) 010 (1.00)
ANSWER:
b
REFERENCE:
GA TRIGA Mechanical Maintenance and Operating Manual ANSWER: 011 (1.00) ,
C ,
e
REFERENCE:
RRF SAR (pg 5 8 thru 5-12)
ANSWER: 012 (1.00) a
.3 1
REFERENCE:
RRFSAR(pg5-6through5-11)
(***** CATEGORY C CONTINUED ON NEXT PAGE *****).
l- .
! .J
, ;3 , - . . , ,m - . -- . . - - - - - , - - - - - . - - - . - --
V , ,
w >
i" i i ),:
Page 42. [
C. PLANT AND RAD MONITORING SYSTEMS i
ANSWER: 013 (1.00) b
REFERENCE:
GA TRIGA Mech. Maint. and Operating Manual ANSWER: 014 .(1.00) d
REFERENCE:
GA TRIGA Inst. Maint. Manual i
ANSWER: 015 (1.00)
-d REFE.tENCE:
GA TRIGA Mech. Maint. & Operating Manual L
e
' ANSWER: 016 (1.00) c
REFERENCE:
l GA TRIGA Mech. Maint & Operating Manual' L-(***** CATEGORY C CONTINUE 0 ON NEXT PAGE'*****)
ee-~a w
-I 1
C. PLANT AND RAD MONITORING SYST'MS Page 43 1
l j
ANSWER: 017 (1.00) i d
REFERENCE:
RRF introduction to Radiation and Rad. Inst. Manuals ANSWER: 018 (1.00) c ,
REFERENCE:
RRF Introduction to Radioactivity (pg 12-1-25 1 2-31) t ANSWER: 019 (1.00) a
REFERENCE:
GA TRIGA Maintenance and Operating Manual P b
ANSWER: 020.(1.00)
- ANSWER a ,
REFERENCE:
RRF Emergency Plan ,
(***** END OF CATEGORY C *****)
(* * * ** * * * ** END OF EXAMINAT ION * * *'** * * ** )
m $
+ - - - g - e - ei-.-
- c.- v t,
TEST CROSS REFERENCE Page 1 R0 Exam ??? Reactor Organized by Question Nuaber QUESTION VALUE REFERENCE 001 1.00 9000001 002 1.00 9000002 003 1.00 9000003 004 1.00 9000004 005 1.00 9000005 006 1.00 9000006 007 1.00 9000007 008 1.00 9000008 009 1.00 9000009 010 1.00 9000010 011 1.00 9000011 012 1.00 9000012 013 1.00 9000013 014 1.00 9000014 015 1.00 9000015 016 1.00 9000016 017 1.00 9000017 018 1.00 9000018 019 1.00 9000019 020 1.00 9000020 20.00 e
a M
TEST CROSS REFERENCE Page 2 i
-i R0 Exam ?7? Reactor :
l 0rganized by Question Nuaber ,
QUEST!DN VALUE REFERENCE 001 1.00 9000021 002 1.00 9000022 !*
003 1.00 9000023 004 1.00 9000024 ,
005 1.00 9000025 '
006 1.00 -9000026 '
007 1.00 -9000027 :
008 1.00 9000028-009 1.00 9000029- ,
010 1.00 9000030 011 1.00 9000031 012 1.00 9000032 013 1.00 -9000033 ,
014 1.00 9000034 i 015 1.00 9000035 016 1.00 9000036' 017 1.00- 9000037 018 1.00 9000038 019 1.00 9000039 020 1.00 -9000040-20.00 4
- y -v . ',,-~ g ,w ..-+%4--e-
TEST CROSS REFERENCE Page 3 R0 Exam ??? Reaetor 0rgan1 zed by Question Nuaber QUESTION VALUE REFERENCE 001 1.00 9000041 002 1.00 9000042 003 1.00 9000043 004 1.00 9000044 005 1.00 9000045 006 1.00 9000046 007 1.00 9000047 008 1.00 9000048 009 1.00 9000049 010 1.00 9000050 011 1.00 9000051 012 1.00 9000052 013 1.00 9000053 014 1.00 9000054 015 1.00 9000055 016 1.00 9000056 017 1.00 9000057 018 1.00 9000058 019 1.00 9000059 020 1.00 9000060
. 20.00 60.00 RO Exam ??? Reactor Organized by KA Group This summary only valid for PWR or BWR1
- g. Nov 25 91 11:57 No.008 P.02
- e UNITED eTATES f' g l
s .; NUCLEAR REGULATORY COMMISSION neosoN V
/ %, / wndvYcNtIc$LYoNA pas NDV 251991 CONFIRMATORY ACTION LETTER Docket No. 50-288 Li nse No. R 112 Reed College Portland, Oregon 97202 Attention: William Haden Acting President Gentlemen:
Pollock of your staff and Mr. Bob)y faulkenberry, Deputy Administrator, HRC Region V,.on November 24, 1991.
This conversation was in facility on Novemberfollow-up 23, 1991. to the Unusual Event declared at the Reed College following: Based on this discussion, we understand the (1) damaged at the reactor and comunicate this plan to Region V office. Reed College will not the plan without prior NRC concurrence. take any action to implement (2)
Reed College *will not manipulate the reactor control rods nor take the reactor critical until the cause of the radioactivity release is understood, comunicated to the NRC Region V office, and NRC concurrence is obtained.
imediately.If your understanding differs from that set forth above, please call me issuance of an order formalizing the above comitments or req actions on the part of Reed College.
Nor does it preclude NRC from taking the issuance of this letter. enforcement action for violations of NRC requirem 1
/ou Regional Adm 4tor j@
s, w, rw _v. u c t -
~
f/ _ _ _ --- ~^
e f [
p' \
,, , 7 '/ g NUCLEAR REGULATORY COMMISSION UNITED STATES g ? REGION V T . i 1450 MARIA LANE o , , , , , *g WALNUT CREEK. CAUFORNIA S4sWs368 (JAN- 2 %52 Docket No. 50-288 Reed College Portland, Oregon 97202-8199 Attention: Dr. William Haden Acting President Gentlemen.
SUBJECT:
HRC INSPECTION This refers to the special inspection conducted by Messrs. J. Reese, J. Melfi, and P. Qualls of this office on November 24 - November 27, 1991, of activities authorized by NRC License No. R-112 and to the discussion of our findings held by the inspectors with you and Dr. Bennett at the conclusion of the inspection. This report also refers to telephone conversation with Mr.
Pollock on December 4, 1991.
This was a saecial inspection resulting from the event at Reed Reactor Facility on iovember 23, 1991. Areas examined during this inspection are described in the enclosed inspection report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observations by the inspectors.
No violations of NRC requirements were identified within the scope of this inspection.
In accordance with 10 CFR 2.790(a). a copy of this letter and the enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning this inspection, we will be glad to discuss them with you.
Sincerely, f .f' ,
oss A. Scarano, Director Division of Radiation Safety and Safeguards
Enclosure:
Inspection Report No. 50-288/91-01 cc enclosure i Dr. Douglas Bennett, Reed College /J[
Mr. J. Michael Pollock, RRF ((,,1 Dr. William Vernetson, Director of Nuclear Facilities, j University of Florida ,
l D -Q.Lf df 0 4YA \Q h i
g
L i C
U. S. NUCLEAR REGULATORY COMMISSION REGION V l
4 Report No: 50-288/91-k Docket No: 50-288-Licensee No: R-112 Licensee: Reed College Partiand, Oregon 97202 i Facility Namt: Reed Reactor facility (RRF) -
Inspection at: Reed College,_ Portland, Oregon ;
Inspection Conducted: November 24 - November 27, 1991 and December 4, 1991 Inspectors: /2 /3, /1/
P. 9 calls, Reactor Inspector Date Signed J/ Me lfi, Trojan e t inspector l0/1d \ Jn ll2 l$2-
.J.- Ye de, tfiler / Dste Signed haf uards, Emergency Preparedness, ,
and o -Power R or nch Approved by: [
RT A. 5caranol Dirtctor-
- //i Dste'51gned z
Division of Radiation Safety'and Safeguards-Summary -[
Inspection on November 24 '- December 4,1991 (Report No. 50-288/91-01)-
Areas Inspected: Special inspection to review and evaluate the Reed College response to a fuel element leak. Inspection Procedure 92700 was used in the course of this inspection.
Results. In the areas inspected, no violations or deviations were. identified.
Licensee response was found adequate to protect the health and safety of the
.public. ,
h e
ogqL L&y((--
"L
1 I
DETAILS
- 1. Persons Contacted
- Dr. W. Haden, Acting President
- Dr. D. Bennett, Vice President-Provost
- Ms. H. Watson
- Mr. J. Pollock, Acting Acting Vice President Director RRF for Public Affairs Mr. M. Parrott, Health Physics Consultant Mr. P. Terdel Seni Mr.E.Argetsinger,orReactorOperatorSenior Reactor Operator Ms. A. Le Roux, Senior Reactor 0)erator Mr. J. Bradford, Senior Reactor Operator
- Denotes those individuals attending the exit interview
- 2. Follow up of Event (9270)
A. Overview On November 23, 1991, at 3:15 p.m. PST. operators at the 250 kwt Reed College TRIGA research reactor received a Gaseous Stack Monitor (GSM) Alarm and a Reactor Room Ventilation Isolation signal. Soon af ter the GSM alarm the operators tripped the reactor. The reactor had been operating at 96% power for five and one half hours prior to the trip. Following the reactor trip the continuous air monitor (CAM) alarmed.
The ventilation isolation signal isolated supply air to the reactor bay and switched exhaust air through a HEPA filter at approximately 100 cfm. The liccnsee initially estimated the release rate to be 15 microcuries per minute; believed to be all noble gases. The release continued into the early hours of November 24, as purging of the reactor bay continued.
ThelicenseedeclaredanUnusualEventat3:30p.m.licenseeNovember due to indications of a leaking fuel element. The 23, subsequently notified Oregon State Department of Energy and the Oregon State Radiation Health Office. Assistance was obtained from the Oregon State Radiation Health Office to conduct environmental radiation surveys. Direct radiation and airborne surveys conducted by the State (including iodine sampling) detected no radiation levels above background outside the facility.
The licensee notified the NRC of the event at 7:57 p.m. on November
- 23. TheNRCResidentInspectorfromtheTrojanNuclearPowerPlant in Ranier, Oregon was dispatched to observe initial licensee followup activities. Two inspectors from the NRC Region V staff -
arrived on campus the following morning to investigate the event and monitor licensee response.
l
. 2 B. Areas Reviewed
- a. Operations Review The inspector reviewed the operator's )
activities and the facility operating logs prior to the event, i It appeared to the inspector that only routine operations were ;
in progress at the facility and logs indicate no activities which might result in damage to a fuel element.
In August of 1991 the licensee experienced a leak in the reactor pool water heat exchanger. On August 23, 1991 the ;
Reactor Director issued Notice to Opera %rs No. 91-5 which ,
documented that the pool cooling system was isolated and directed that no operation in excess of 500 Kw total energy be-performed without additienal authorization.. This guidance was issued because the ability to remove heat from the pool was limited. Subsequent to August 23, the only occasion the reactor was operated in excess of this limit wcs on November 23, 1991 The ins)ectors verified that this operation had been authorized by the Reactor Director. The inspector determined that operation without the heat exchanger is acceptable as long as the pool maximum temperature of 50 C in Technical :
Specifications is not exceeded.
- b. Experiments The inspector reviewed the irradiation logs for i the samples being-irradiated. These were recorded on request nos. 2863, 2862, and 2861 and consisted of samples of teflon air filter for air pollution research, lake sediment samples, and geologic soil samples. The Reactor Director stated that these should contain no natural uranium which could cause an increase in fission gases in the reactor bay. A review of recent irradiations by the inspector indicated no experiments that were not' properly reviewed by the reactor staff and all were conducted in accordance with the technical. specifications. ;
- c. Power History The inspector reviewed the operating history as-recorded in the operatin 1991 until the November 23, 1 91 event were reviewed. The logs... Logs from noted no occasions where the facility operated continuously in excess of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at significant power levels. On January 16, 1991, at 60% during power, power or 150 calibrations,inuous Kw for 8.0 cont hours for a totalthe reactor wa run of 1200 Kw-br. On November 23, 1991, the reactor was l operated for 5' hours at 240 Kw for a total of 1200 Kw-hr. The inspector reviewed the power record for November 23 and found that reactor was exceeded. Thepower was inspector maintained concluded that ',he'stable recordsat 240 Kw and no l from-l the January operation would serve as satisfactory baseline to compare-with the November 23 event due to sim)lar power total. da
- i. on a short time period.
I d. Event Response The inspectors reviewed the response of the L operators t.; the event. The operators responded to the GSM alarm in accordance with their operating procedures and the Emergency Plan. The event was properly classified as an Unusual L
l
. 3 Even., -It appeared to the ins')ectors that the operators j responded in accordance with t1eir Emergency Plan Implementing '
Procedures (EPIP) guidance. Although several entries into the reactor room were made to take samples and to secure the primary water-system, no personnel contamination was detected and follow-up j whole body counter evaluations detected no internal contamination. l The licensee appeared to have made the proper notifications in - {
accordance with their EPIPs. ;
U
- e. Fuel Inspection Technical Specifications require the licensee to annually inspect one-fitth of the fuel elements such that all are inspected after five years. The inspector reviewed 50P 40, Inspecting and Hand'ing of fuel. The procedure re discolorations scratches, or corrosion be logged. quires Neither thethat any procedurenoriechnicalSpecificationsrequireanyevaluationof the data recorded. Fuel inspections were conducted in January, 1990 and in January, 1991. During-both inspections, the licensee recorded scratches and discolorations on numerous fuel elements.
The licensee stated that they had received no guidance from the vendor on evaluation criteria for these abnormalities after conducting +.he inspection, The inspectors recommended the licensee contact the vendor for guidance in this area. ,
- f. Fuel Pool Temperature. The reactor fuel pool temperature alarm had been alarming prior to the incident.. As a compensatory measure the licensee was checking three independent temperature measurements periodically during reactor operation. .The operators were checking the reactor water ion exchanger inlet and-outlet temperatures remotely at the reactor console and also using a calibrated temperature probe manually measuring the pool temperature at the location of-the alarm arobe. This was-being ,
performed at least hourly. The alarm pro)e has no temperature readouts to verify the set point. During the operation temaerature increased about 12*C to a maximum of 35*C. The Tecinical Specification limit is 50'C. The 12 C increase -
correlates with the. January 16. temperature increase of 10 C. The inspector did note that in January the maximum pool tem >erature reached 23 C and in November it reached 35'C but since )oth are well within design limits, the inspector could not attribute any -
causal factor to reactor pool temperature.
- g. Safety Rod Motor Control. - After the Scram, the safety control ,
rod motor rotated in the outward-direction thus did not reset as required . The safety rod dropped ~into'the core on the trip-signal as required. No imatoper rod movement thus occurred. -The licensee felt that the proslem was in the balance circuit Similar problems have occurred at'other Triga reactors. This-occurrence did not affect the event or.the licensee's actions.but is a concern which'should to be corrected prior to a future reactor '
start-up. This matter will be addressed.in a future inspection (lFI 50-288/91-01-01)..
4
- h. Similar Event. There have been no similar events at the Reed Heactor Facility. On July 26 1989 at General Atomics (GA)
TheMarkFreactor,a1.5Mwirigareactorexperiencedsimilar symptoms. These were finally traced to a fuel element with a pinhole leak. General Atomics did not evaluate the fuel element to determine the cause of the pinhole development.
C. Radiological Review
- a. Radiation Monitors In this area, the ins)ectors reviewed the response of the Gaseous Stack Monitor (G54), Air Particulate Monitor (f.PM), and the Continuous Air Monitor (CAM) to the event, compared the monitors response during the event to the monitors response to normal operations during the past year, traced the GSM and APM system to determine sample and detector location and reviewed the calibration of the detectors.
A brief description of the monitoring system follows. The GSM is an offline G-M detector which takes a sample of the effluent stream after the normal and emergency exhaust filters just prior to its release to the environment. The sample is passed through a particulate filter prior to reaching the detector in order to remove the p ciculates and monitor as pure a gaseous sample as possible.
Exhaust from the sample is inserted bac( into the effluent stream prior to the noraal filters. A remote readout of the monitor is located in the control room with essentially two alarm functions.
The first is the failsafe light. The failsafe light is normally on when the activity detected is below the actuation point (80% of the alarm setpoint). When this level is reached, the light is extinguished. No audible alarm occurs. The next level is termed the alarm. The alarm level indicates that the activity being released is approaching abnormal levels and that a problem may exist. The alarm level is indicated by both and audible and visual alarm.
The APM is also an offline G-M detector which takes a sample of the effluent stream and passes the sample through a particulate filter located near the detector. The detector is located within the reactor bay and is influenced somewhat by the direct radiation within the bay during reactor operations. The alarm function and readouts are similar to those described for the GSM.
The CAM takes its sample from above the reactor pool. The sample is passed through a particulate filter prior to entering the detector chamber. The particulate filter is monitored by the detector as well as the gas. Alarms and readouts for this monitor are all local at the monitor.
The inspectors reviewed the response trace of the monitors to the event and compared this response trace to that of similar operation on January 16, 1991, (considered normal operation .
During the January run, it was noticed that all three tracet i.. creased immediately following reactor startup. This increase continued for aparoximately 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the reactor power was constant at whic1 time the traces leveled and remained relatively constant until reactor shutdown (after approximately 1200 Kw-br
l 5
- of operation). After reactor shutdown, the activity decreases m.til background levels are reached. The response equates to the production, equilibrium release and termination of production and decay of Argon-41.
The traces during the November 23 event, began as would normally be expected with a constant increase immediately following reactor startup and leveling after aaproximately 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a constant power level. After about 3 sours (720 Kw hr) of operation, the CAM trace began to increase. Af ter about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (960 Kw-br) of operation, the GSM and APM also began to increase. The rate of increase for all three traces was sharper than that noted during the normal production and release of Argon following reactor startup.
After approximately 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of operation, the activity detected by the GSM, reached the alarm setpoint and the operators shutdown the reactor. For several hours following shutdown the traces continued to increase indicating that activity was still being released from the reactor pool. This continued for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following reactor shutdown when the traces began to irdicate a decrease. The decrease continued for about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at which time normal background levels were reached.
Based upon this review, the inspectors concluded that no evidence of previous leakage existed and that the leak only became evident after several hours of operation on November 23, 1991.
The calibration of the GSM, APM and CAM were reviewed. Calibrations are required annually by TS. The licensee calibrates the monitors semi-annually inpractice. Calibration of the GSM is performed by irradiating a sampie of Argon gas in the reactor, calculating the activity of the irradiated sample, utilizing a syringe to extract a known amount of the Argon and inserting that amount into the detector well. The detector response to the known amount of gas, in counts per minute (cpm) is then used to calculate an empirical detectorefficiency(microcuriespercubiccentimeter(uCi/cc)per cpm).
During the inspection review, it was determined that the GSM had received a calibration in March,1991, following the replacement of the detector. This calibration resulted in an efficiency near those efficiencies seen in previous calibrations. Another calibration, the routine semi-annual calibration was performed in Jul;,1991 This calibration resulted in an efficiency, an order of magnitude j higher than any found in 1.he past. At the time of this inspection, the July calibration was not signed by the M. son who performed the calibration nor by the supervisor responsible for the review. When questioned why the calibration paperwork had not been completed, the Acting Reactor Director stated that completion of the calibration and signing of the paperwork was pending a review of the cause of the large differences in the efficiencies. This review had not occurred at the time of this inspection. The inspector was informed that the old efficiency from the Pirch calibration was being used at the time of the event to determine the magnitude of the release.
- ~ - - - - - - - , - - = _ _ _- ._ ___-____ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- 6 The inspector informed the licensee that such large discrepancies in efficiencies is an indicatica that either the detector or its associated hardware may not be functioning correctly. As such, until the cause of the discrepancy is determined the monitor should notbeconsidered03erationalandanydeterminatIonofreleaseusing this monitor would >e suspect.
During the inspection the licensee performed an analysis of the two calibrations and discovered tr.at the calculations of Argon activity for the March calibration were in error. Correcting this error resulted in an efficiency very close to that found during the July calibration.
Review of the CAM calibration by the inspector revealed that the last calibration, performed in August, 1991, had not been completed.
The calibration had proceeded to the point of recording the initial "as found" settings. Since the calibration had not proceeded to the point where the monitor setuo based u1on the previous calibration (performedinJanuary)wasclanged,tieinspectordeterminedthe monitor was operational It appears that the licensee failed to realize that failure to complete a calibration within the specified time could cause the monitor to be considered technically out of service. It is recommended that the licensee place a higher priority on ensuring that calibrations are completed within tie specified time and that abnormalities are resolved in a timely manner.
During the interviews with operators, it was determined that during the event the APM failed to annunciate (by extinguishing the light) that the detected activity had exceeded the failsafe setpoint. The posted setpoint (on the front of the monitor readout) of 4,000 cpm was determined by the latest calibration. The inspectors were not able to determine the cause of the discrepancy at the time of the inspection. This matter will be reviewed in a future inspection.
(IFI 50-288/91-01-02)
, b. Surveys The inspectors observed and reviewed the licensee's radiological surveys for facility recovery. At the time of the event, the State of Oregon at the request of the licensee sent an environmental survey team to determine the consequences of the offsite release of radiation. The state's team performed dose rci.e measurements and sampled the air for particulate s and radiciodine around the reactor facility. Results of the surveys indicated no radiation detectable above background.
At the request of the inspectors the licensee performed a swipe-survey of the exit of tia effluent stack. The sample was scanned for contamination using a count rate meter. No contamination was found.
The inspectors independent of the licensee performed offsito surveys using an Eberline PRM-7, miro-rem meter and an Eberline E-530 countrate meter. The inspectors surveyed the area surrounding the reactor facility including the new Chemistry building immediately
7 adjacentandfoundnoindicationofradiationabovebackground.
The results of the reactor pool samples taken by tite licensee were revised. T.? licensee took a 2.5 liter sample the night of the event (approxu ately midnight on November 23). One liter was analyzed e t'ie gamma spectroscopy system and another liter wat passed through tie ion exchange resin. Then the resin was anslyzed by the gamma spectroscopy system. The results of these analysis indicated the wasence of relicactive fission nobb osses, fenon-135 (Xe-135), kryoton-85m Kr-85m (gases,) principally the
, krypton-88 (Kr-80 No radiolodine were found.'
The licensee discussed with the inspector i uir plan to terminate the NOUE. The licensee removed the fixed mt:r air sample from the CAM and sampled the reactor pool water to determine the amount of activity that remained in the reactor bay and the amount in the rear. tor pool water. Both samples were analyzed on the gamma spectroscopy system. The inspector reviewed the results of both samples which indicated no fission gases remained. The l'.censee initiated the reactor bay normal ventilation after these results were reviewed.
The inspectors observed the licensee's surveys performed to release the reactor bay from contamination control. The initial survey consisted of 5 swipes taken on various pieces of equbment in the reactor bay. The inspector reviewed the results of t1is survey with theActingReactorDirector(ARD)andinformedtheARDthatthe -
survey did not appear sufficient to characterize the radiological conditions in the bay. As a result the licensee performed a more
- comprehensive survey of the bay consisting of 6 swipes and 4 large area ocp swipes. The results indicated no loose surface contamination. ,
- c. Release The inspector reviewed the licensee's determination of the total amount of activity released during the event. Based upon conservative assum millicuries (mci) ptions-the licensee determined Thethat 113of noble gases activity breakdown based upon the ratios of the gases in the reactor pool water were: Xe-135, 59.89 mci; Kr-88, 33.9 mci; Kr-85m,19,21 mci.
The inspe': tor calculated the maximum offsite' dose based upon worse case meteorological conditions utilizing the methodology in Reg Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, 1982. .The-maximum dose that could have been received by an individual located at the edge of the exclusion area for 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> was calculated to be 0.07 millirem.
- d. Perseinel Exposures The inspectors reviewed the exposures received by the operators who initially responded to the event and those personnel involved in facility recovery. The operators on duty at the time of the event made a total of four entries into the reactor bay after the initial alarm was received on the-CAM. These entries were made for various reasons including, to obtain the reading on
(
, , -.---r-,, , e
- 8
' the CAM, obtain reactor pool water samples and change the filter on the CAM, isolate flowand,into ti,e reactor pool.inThe accordance with by exposure r6ceived emergency the plan operators, based upoa pocket ionization chamber readings was not i detectable. The inspecto.'s also noted that the operators initially I responding did not become contaminated by the noble gases upon entry. This would indicate that .e concentration in the reactor bay at the time of entry was low.
As noted above, the operators initially responding to the event made four entries into the reactor bay after the initial CAM alarm. These entries were made without respiratory protection. At the time of the entrieg. the constituents of the radiological airborne contamination i the operators were entering unknown had not beerconditions.
radiological tefined thus,The inspectors discussed with the licensee that entries into unknown con-Jitions should be made utilizing personnel protection (arotectiveclothing,respiratoryprotection)tothefullest extent possiale.
D. CONCLUSION The inspectors concluded that the most probable cause of the event was a leak in a fuel element; probably a snll pinhole leak. The similarity to the GA event and the composition of the gases released would support this conclusion. The camples being irradiated had not been removed from the reactor at the conclusion of the inspection. The licensee's operators reacted proaerly and the ventilation system functioned as designed to limited t1e release. It appeared that the health and safety of tie campus population and nearby residents and reactor operators were not affected by the release.
- 3. RECOVERY On November 25, 1991, the NRC Region V sent to the licensee a Confirmatory Action Letter (CAL) which required the licensee to develop a plan and get Region V concurrence prior to moving any fuel elements. The inspectors also discussed with the licensee that it appears that Technical Specification E.4 would prevent reactor operation until the leaking element is removed. The licensee indicated on December 3 1991 that a llan would be developed for recovery with assistance from iest Researc1 and Training Reactor (TRTR) groua and that it would rot be completed before February 1, 1992. The NRC will review th; licensee's action plan and recovery effort in a later inspection (50-288/91-01-03).
- 4. Exit Meeting On November 26, 1991, an exit meeting was held with members of the licensee's organization as indicated in paragraph 1. The items in paragraph 2 and 3 and the CAL and Technical Specifications in paragraph 3 were discussed at that time.
0 January 16, 1992 Lb ugo Au t Docket No. 50-288 ggy V Dr. J. Michael Pollock I9i2 y 7 g p3 {: 36 Reed College Reactor Facility Reed College '
3203 SE Woodstock Blvd.
Portland, Oregon 97202
Dear Dr. Pollock:
SUBJECT:
ISSUANCE OF AMENDMENT NO. 5 TO FACILITY LICENSE NO. R-112 - REED COLLEGE REACTOR FACILITY The Commission has issued the enclosed Amendment No. 5 to Facility License No. R-112 for the Reed College Reactor Facility.
The amendment consists of a change to the Facility License in response to your submittal dated December 5, 1991.
The amendment grants relief from certain surveillance requirements while Reed College is implementing a Recovery Plan as agreed to in NRC's Confirmatory Action Letter of November 25, 1991.
A copy of the related Safety Evaluation supporting Amendment No.
5 is enclosed.
Sincerely, Original signed by:
Marvin M. Mendonca, Senior ProjJet Manager Non-Power' Reactors, Decommissioning-and Environmental Project Directerate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amendment No. 5
- 2. Safety Evaluation cc w/ enclosures:
See next page DISTRIBUTION Docket. file- EJordan NRC & Local PDRs EHylton-PDNP rf GHill (4) DCrutchfield .ACRS (10)
WTravers GPA/PA AAdams MVirgilio TMichaels BBoger OGC JPartlow CGrimes MMendonca. Wanda Jones DHagan OC/LFMB Region V (4)' RZimmerman,=RV p P PDNP:PM M OGC P b ton MMendonca h S ft:5 1/ /92 1/j'd/92 / 1/g/92 1/( /92
.jh c;l[k
}W Yp%d
i
- Reed College Docket No. 50-288 CC*
Director, Oregon Department Of Energy 528 Cottage Street, N. E.
Salem, Oregon 97310 Mayor of City of Fortland 1220 Southwest 5th Avenue Portland, Oregon 97204 Administrator Siting and Regulation Oregon Department of Energy Labor and Industries Building Room 111 Salem, Oregon 97310 m_. ______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ - . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
8 omsg e jo, UNITED 5TATES 3 .$ [f ' }' ,g
- E NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
%....+/
BEED COLLEGE DOCKET NO. 50-288 ,
AMENDMENT TO FACILITY LICENSE Amendment No. 5 License No. R-112
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by Reed. College (the licensee), dated. December 5, 1991, complies with the standards and raquirements of the Atomic Energy Act.
of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules e and rer71ations of the Commission; C, There-is reasonable assurance: (1) that the-activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such with activities will be conducted the Commission's in regulations; compliance D. The issuance of this amendment will not be inimical ~to the common defense and security or to-the health and '
safety of the public;-
E. The issuanci of this amendment is in'accordance with 10 CFR Part 51 of the Commission's regulation and all-applicable requirements have been satisfied; and- .
F. Prior notice of this amendment was not required by 10 CFR 2.105(a)(4) and publication of notice for this amendment is not required by 10 CFR 2.106(a) (2) .
p*
z<i ,3 41;ba (7 l
'p f7 d, 1 cc ? .I .
- 2. Accordingly, the license is amended by adding the following paragraph 3.F to Facility License No. R-112 as follows:
F. Until completion of the Recovery Plan in response to the Commission's Confirmatory Action Letter of November 25, 1991, the time periodicity requirements for Technical Specifications surveillances E.3, F.2, F.9, and F.10 will be waived. However, these surveillance requirements must be completed prior to return to routine reactor operations.
- 3. This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION OA Seym ur-H. Weiss, Director Non-Power Reactors, Decommissioning and Environmental Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation Date of Issuance: January 16, 1992
+fpa aseg'o
,, UNITED STATES
- NUCLEAR REGULATORY COMMISSION 3'
- ;E W ASHINGTON, D. C. 20555 kg s...+
j! l SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l
SUPPORTING AMENDMENT NO. 5 TO FACILITY LICENSE NO. R-112 REED COLLEGE DOCKET NO. 50-288
1.0 INTRODUCTION
By letter dated December 5, 1991, Reed College (the licensee) requested a change to Facility License No. R-112 for the Reed College Reactor Facility. The requested change would waive the time periodicity requirements for Technical specifications surveillances E.3, F.2, and F.9 until completion of the Recovery Plan in response to the Commission's Confirmatory Action Letter of November 25, 1991. However, the requested change would also assure that the surveillance requirements must be completed prior to return to routine reactor operations.
2.0 BACKGROUND
As a result of an Unusual Event on November 23, 1991 due to high radiation indications, the licensee agreed to provide Region V with a Recovery Plan for NRC concurrence prior to implementation of this plan and return to operations. Because of the conditions of the licensee's agreements with Region V in accordance with the Confirmatory Action Letter, certain surveillance requirements can not be satisfied, e.g., visual inspection of the fuel and control rods, and control rod drop times. Thurefore, the licensee requested relief from certain surveillance requirements, and indicated that the surveillances would be complete prior to operation "for any purpose not specifically addressed in the Recovery Plan."
l 3.0 EVALUATION l
I During the review of the Technical Specifications, the NRC staff determined that in addition to the above described relief, the licensee would need relief from the requirements of section F.10 on annual thermal power calibrations of the linear power level channel. This was confirmed with the licensee and has been incorporated in the evaluation and amendment at the licensee s request.
- [3 V
Q h l%
, ,u, f '
Y.
/.MY ,
gp$S
_.[- .
i Because the reactor will not be operated, excapt as required by the Recovery Plan which will receive NRC concurrence, there will be no transient effects placed on the fuel or core. Therefore, it is highly unlikely that there will be any fuel or control degradation, and there is no need for control rod drop or thermal power indication. However, even in the unlikely event of fuel or control rod degradation during the shutdown period, no event would exceed the consequences previously analyzed and accepted for full power operations.
Therefore, the surveillance time periodicity requirements for fuel inspection (Technical Specification E.3), for control rod inspection (Technical Specification F.2), for control rod drop testing (Technical Specification F.9), and for thermal power calibration (Technical Specification F.10) can be acceptably waived until completion of the Recovery Plan.
3.0 EEIRONMENTAL t'ONSIDERATION This amendment involves changes in the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes in inspection and surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and there is no significant increase in individual or cumulative occupational radiation oxposure.
Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9) . Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
- 4. CONCLUSION The staff has concluded, based on the consideration discussed above, that: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, or create the possibility of a new or different kind of accident from any accident previously evaluated, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by the preposed activities, and (3) such activities will be conducted in compliance with the commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or the he'lth and safety of the public.
Principal Contributor: Marvin Mendonca Date: January 16, 1992 L
U e
e -
lbw_ l __ M 4
' 'l fu,. A M c_. ,c e y a x en, d , ) a f ,17. lc < ice c.~ rd
/ ', hr8 * * * * #*
- m- s , inn. A r 5
\fg %_ -
/
m - .
. c.;
N
/
- e'm -
i j
- s N p f- , - 9 4 c,.4. u a 24. -
un M'*
N_ ,, _ i'T g 8 PM' g ,/p g,,, p;,,g pgn g.< 5 h'
t N j.l -
'5 '" ***;as 1
N=__ij- ".
n, il
,.,j p
-" t'# w .. r g
NJ N
=
- J/ -d iss ,,
3.,; gei + m t
\ d $ / ius N ' 5""
g l s 9
H N 3-
- d 4.
N 1 sW/ V
(
C - ==""'# l .,
A PM ghl I ,p
~ "i i
. J ,6.i 4% - .- TN @! i k,,, ' '
MC .
t
% _ _ A w " 'r l; \[. - po
.~
cn :,
e u so ,
i hp%_
l
__-.s l! .
h(w -. .
,) n,, t 25 <r' if' 70 < r, ..c i un e ' , a t) is w. .t
\ [r 1 '
- 3) 3 r v. .t Mocr
- N' AN_- .-
r . yy t.. .t er .r , 6
'Wf
".,"' 3 .s l
N b' ~~ L ,
- u. n. + o n ,a
}
u NlL x_ - -
~
3
. !N.[
x '
,, q ii -_ _ , ,
-qj'p:'-k, sJr., . ,,,
[ .
'b 1A, Aig - i
(: . .1 iL 2
L
Ii n T, I, :-- -
N , :- .
- ,@ ,,d I'{,
i i WL Il Lbh i im ! m 7,2 .
n e . --
i O gr 1 .
REED REACTOR FACILITY hCOUAllFICATIDH PLAN 1 Introduction This Requalification Plan is developed in compliance with the requirements of 10CFR55.SS revised and published in the Federal Register on March 27,1987.
- 2. Schedule The Requalification Program shall be continuous and operate on an annue' cycle. For pur of tnis Plan, a year begins and ends on July 1; calendar quarters are a.r rollows:
FIRST QUARTER: Ju y 1--September 30 SECOND OUARTER: October 1-December 31 I THIRD OUARTER: January 1-March 31 FOURTH OUARTER: April 1--June 30
- 3. Lectures and Meetings 3.1. A minimum of Dve lectures shall be scheduled during each year, with a minimum of 2 ea semester. These shall be designated by the Director as Requailfication Lectures. The subjects of these lectures may be drawn from the topics listed below with the specific content be based on items identified as weaknesses in the training program or operator knowledg identified in the results on NRC licensing exams, facility administered Requalification exams described in ibis Plan, and/or observations of operator knowl a) Theory and p< nciples of reactor operation b) Reactor instrumentation, control, and safety systems '
c) Radiation control and safety; use and handu d) n g of radioactive materials Requalification and Emergency Plans, Administrative Procedures Title 10 Code of Federal Regulations.
e) Stardard and Emergency Operating Procedures 3.2. Reviews of Facility Changes:
shall be held to make them aware of facility design changes, and ch administrative requirements. This meeting may be held in conjunction with a Requalification lecture or may be held separately, at the discretion of the Director.
3.3. Any operator who is unavoidably absent from any Requalification Meeting shall revie notes, handouts, tape recordings or other reference material from the meeting. This materist shat! be summarized the operator has understoodin thewriting material.for the Director (or his designate) to assure the Facility that
- 4. On the-Job Training 4.1.
Each Reactor Operator and Senior Reactor Operator must actively perform the functions for which they are licensed for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> per calendar quarter. The Director shall maintain a list of thoe activities which may and may not be counted towards this requirement.
- - - - --- - ---- ~ al
- r. 4.2. Each licensed operator and senler reactor operator shall demonstrate an ability to conduct a variety of reactivity manipulations during the course of the requalification year. A minimum of 10 reactivity manipulations shall be completed during each requalification yeai. The Director shall maintain a list of those operations and the number of each, which qualify towards meeting this requirement.
4.3. Each Reactor Operator and Senior Reacior O: or shall be responsible for maintaining records, in a format approved by the Director, to auf compliance with sections 4.1 and 4.2.
- 5. Observation 5.1. Reactor administration, supervisory, and health physics and training personnel may conduct observations of the performance and competancy of licensed Reactor Operators and Senior Renctor Operators during routinely scheduled operations. These observations should normally be unannounced and shall be dovumented.
5.2. Documented ebservation of each operator shall occur at least once each year.
- 6. Evaluation 6.1. A comprehensive written and operating test shall be administered annually to each Reactor Operator and Senior Reactor Operator. The results of the test will be used to determine the areas in which retraining is need to upgrade licensed operator and senior operator know! edge.
1 6.2. The writter exam (s) shall be prepared by the Director or a Senior Reactor Operstor assigned by the Director. ;
6.3. All written examinations shall be administered on the same day and the exam shall be proctored. If any RO or SRO unavoidably misses the exam, a different exam shall be administered.
6.4. The operating exam shall be administered by a Senior Reactor Operator assigned by the Director.
6.5 The Senior Reactor exam administered by the NRC, shall suffice to satisfy the annual requalification exam requirement.
- 7. Accelerated Requalification Program 7.1. Any operator who has an identified weakness as described in this section shall be entered into an Accelerated Requahfication Program.
7.2. Any Reactor Operator or Senior Reactor Operator who falls to achieve a score of 70%
overall on the written exam OR 70% on the operating exam OR who,in the opinion of the Director, is clearly deficient in operating knowledge' or skills shall not operate the reactor i excent under the direct supervision of a Senior Reactor Operator until completion of an Accelerated Requalification Program.
7.3. Any operator who achieves an overall score of 70% on the written exam AND 70% on the operating test may continue to operate the re:ctor. However,if any Reactor Operator or Senior Reactor Operator fails to achieve a score of 80% on one or raore sections of the annual exam, an Accelerated Requalification Program must be completed
- r 7.4. The content of any accelerated requalifcation program shall be determined on a case by-case basis by the Director,in consultation with the operator, and shall be documented prior to
- Initiation it may include, but need not be limited to, attending or presenting lectures on the areas of weakness, Individual or group study sessions, or tutorials.
7.5. Successful completion of the Aceslerated Retraining Program shall be based on the passing -
- of an additional exam. This exam shall be administered within six months of the date of the original requalification exam. The extent of this additional exam will normally be limited to areas of identified weakness.
- 8. Records Originals of all records relating to the requalification requirements in this Plan shall be retained by the faciPty. Copies shall be provided to the RO or SRO at their request.
t Reactor Observation Record Page 1 Operators Name:
Operating Year:
This record shall be completed by the operator, When all requriements have been fulfilled, the operator shall submit it to the Director.
REQUIRED REACTIVITY MANIPULATIONS (in required.ach yearinciveing at least one from each of the 5 categories listed; include date and time on the blank provided)
- 1) Reactor start up: _ _ _
(including Start-up Checklist, manipulation of control rods from cold shutdown through c measurement of the " core excess")
- 2) Increase in reactor power: ___ _ _ __
(beginning with the reactor critical at a power level less than 1 kW and ending at a power level where the negative temperature coefficient is significant)
- 3) Decrease in reactor power:
(beginning with the cactor critical at a power level greater than 1 kW and ending with -
the reactor critical at a power level 1/10 or less of the initial power)
'4) Reactor shut down: _____________________________________________
(beginning with the reactor critical, ending with the reactor " shut-down' as defined in the Technical Specifications; - including Shutdown Checklist)
- 5) Special operations (one- required. each year):
a) Fuel elements inspection:
(as the individual on t'e inspection crew who actually manipulates the fuel handling tool with an element attached) b) Control rod calibration:
, (as the individual ' who actually performs the reactivity manipulations) c) Control rod Inspection:
(actual removal from or reinsertione of :he control rod into the core) -
d) Other (describe):
J-Reactor Operation Record Page 2 >
Hours-performing = duties of licensed operator-(show 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> / quarter on this sheet; continue record,11 desired, on continuation sheet)
Date # of Hours Descriotion of Activities FIRST QUARTER (July 1-Sept. 30)
SECOND QUARTER (Oct.1-Dec. 31)
THIRD OUARTER (Jan.1-Mar. 311 l
FOURTH QUARTER (Apr.1-June 30) i l
l-
+ ,
Operator Observation Record --
. Operator Name:- )
Observer Name and Title (printed):
Date and Time: 2 Operation (s) Observed:
Was operator either: a) familiar with SOP being used; OR b) referring to SOP, while conducting the operation (either acceptable)?. Comments:
Did operator exhibit a general familiarity with the reactor, .the facility, and its systems as they related to the operation performed? Comments:
Were all required log entries complete, accurate, and legible?
Other Comments (use additional sheets if necessary): ,
Additional information is attached i Observer Signature:
I J