Text

Final Response to FOIA Request Re Insp,Exam & Annual Repts. Forwards Documents Listed in App A.Documents Also Available in PDR
	ML20155G648
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	09/27/1988
From:	Grimsley D; NRC
To:	Walthers C; THE TAB NEWSPAPERS, NEWTON, MA
Shared Package
ML20155G653	List: IA-88-450, Final Response to FOIA Request Re Insp,Exam & Annual Repts. Forwards Documents Listed in App A.Documents Also Available in PDR; ML20155G648;
References
	FOIA-88-450 NUDOCS 8810170293
	Download: ML20155G648 (4)
	v • d • e

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s RESPONSE TO FMEEDOM CF ,X l'* ! I 's"* f

\ INFORMATION ACT (FOlA) REQUEST c eit v ote S,. % .ee, W 2 i 1300 l

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I i items 1 through 5 of Appendix A are enclosed. j Items 6 through 27 of Appendix A can be obtained from the PDR.

' The new PDR is located at 2120 L Streete N. W.o lower-Levels Washington D. C.o 20555. e 1

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APPENDIX A 00CUMENTS TO BE RELEASED FOIA REQUEST NUMBER 88-450 DATE ORIGINATOR RECIPIENT DESCRIPTION Inspection Reports I

1. 3/21/85 T. Martin L. Clark, M.I.T. Inspection No.

Region 1, NRC 50-20/85-01 (9 pages) 2, 4/9/85 T. Martin L. Clark, M.I.T. Acknowledgement of i Region I, NRC response letter dated 2/27/85 (copy attached)

(6 pages)

E. Wenzinger L. Clark, M.I.T. Examination Report No.

3. 11/21/85 ,

Region I, NRC 50-20/85-03(OL)

(77 pages)

4. 11/27/85 T. Martin L. Clark, M.I.T. Inspection No.

Region I, NRC 50-20/85-02 (4 pages)

5. 4/25/86 E. Wenzinger L. Clark, M.I.T. Inspection No.

Region I, NRC 50-20/86-01 (13 pages)

6. 8/21/86 T. Martin L. Clark, M.I.T. Inspection No. !

Region I, NRC 50-20/86-02 (5 pages)

7. 8/29/86 J. Kinneman A. Ducatman, M.I.T. Inspection No.

Region I, NRC 86-01 (3 pages) 1

8. E. Wenzinger L. Clark, M.I.T. Examination Report No.

10/27/86 Region I, NRC 50-20/86-03(OL) l I .

(81pages) l l 9. 10/28/86 J. Kinneman A. Ducatman, M.I.T. Acknowledgement of !

Region I, NRC 9/23/86 response (copy i

i attached) to Inspection No. 86-01 (3 pages) j T. Martin L. Clark, M.I.T. Cembined Inspection Nos.

l 10. 9/9/87 Region I, NRC 50-20/87-02 and ;

70-938/87-02 (9 pages) l

11. 9/9/87 S. Collins L. Clark, H.I.T. Examination Reoort No.

Reg'on I, NRC 50-20/87-01(0L) (44 pages) i

! I

l APPENDIX A DOCUMENTS TO BE RELEASED FOIA REQUEST NUMBER 88-450 OATE ORIGINATOR RECIPIENT DESCRIPTION ,

12. 9/16/87 T. Martin F. X. Masse M.I.T. Combined Inspection Nos.

Region I, NRC 30-763/87-01 and 70-938/87-01 (12 pages)

13. 11/5/87 L. Clari, M. I .T . USNRC Letter

Subject:

Open Iten No. 84-01-02, Facility Operating License R-37, Occket 50-20 (7 pages)

14. 11/27/87 T, Martin F. X. Masse, M.I.T. Acknewledgement of

,l Region 1, NRC 10/16/87 response (copy attached) to inspection 30-763/87-01 and 70-938/87-01 (4 pages)

15. 12/30/87 T. Martin L. Clark, H.I.T. Inspection No.

Region I, NRC 50-20/87-03 (8 pages)

16. 3/3/88 R. Bellamy L, Clark, M.I.T. Inspection No,

Region 1. NRC 50-20/87-03 (3 pages) l 17. 3/16/88 R. Gallo L. Clark, M.I.T. Examinatic.n Report No.

Region I, NRC 50-20/88-01(OL) (5 pages) 18, 6/8/88 G. Sjoblem P. Powell, M.I.T. Inspection No.

Region I, NRC 30-763/88-01 (60 pages)

19. 7/18/68 J. Viggins J. Bernard, M.I.T. Inspection Report No.

Region I, NRC 50-20/88-02 (10 pages)

20. 8/30/88 R. Bellamy J. Bernard, M.I .T. Inspection No.

Region I, NRC 50-20/8S 03 (6 pages)

21. 9/6/88 J. Bernard, USNRC Letter

Subject:

NRC M.I.T. Region ! Inspection No.

50-20/88-02 (1 page)

3 APPENDIX A DOCUMENTS TO BE RELEASED FOIA REQUEST NUMBER 88-450 DATE OPIGINATOR RECIPir, DESCRIPTION Annual Reports I 22. 8/30/85 L. Clark, M.I.T. Dr. Murley, Region I, Letter

Subject:

Annual i NRC Report Period Covered:

I 7/1/84-6/30/85 (26 pages) l 23. 8/29/86 L. Clark, M.I.T. Dr. Murley, Region I, Letter

Subject:

Annual NRC Report Period Covered:

7/1/85-6/30/86 (25 pages) l j 24, 8/29/87 L. Clark, M. I .T. USNRC Letter

Subject:

Annual l

& K. Kwok, M.I.T. Report Period Covered:

! 7/1/86-6/30/87 (26 pages) l 25. 4/21/88 L. Clark, M.I.T. USNRC Letter

Subject:

Evaluation of Unresolved Safety l Question, 10 CFR 50.59 l

(b)(2), MIT Reactor Lic. R-37, Docket 50-20 (119 pages)

26. 8/29/83 K. Kwok & USNRC Letter

Subject:

Annual l J. Bernard, M.I.T. Report Period Covered:

7/1/87-6/30/88 (28 pages) l

27. 8/30/88 K. Kwok & USNRC Letter

Subject:

Revision J. Bernard, M.I.T. to Annual Report Period Covered: 7/1/87-6/30/88 l

(29 pages)

M R 211985 4

50-20 License No. R-37 Docket No:

Massachusetts Institute of Technology Research Reactor ATTN: Mr. Lincoln Clark, Jr.

Director of Reactor Operations 138 Albany Street ,

Cambridge, Massachusetts 02139 Gentlemen:

Subject:

Inspection No. 50-20/85-01 l A routine safety inspection was conducted on February 13-15, 1985 of the radiation protection program at the Massachusetts Institute of Technology '

Research reactor. Areas that were reviewed included implementation of radi-ation protection controls, equipment and instrur.antation, and environmental monitoring.

^

This inspection indicated that one of your activities was conducted in violation of NRC requirements. Details are provided in enclosure Appendix A and in the accompanying inspection report. Your immediate corrective actions and actions to prevent recurrence have been provided in a letter to Dr. Thomas Murley, Regional Administrator, dated February 27, 1985. Therefore, no addi-tional reply is required.

Your cooperation with us in this matter is appreciated.

Sincerely,

' Original Signed By:

N ' Thomas bbk DivisionT.ofM4rtin, Radiation Director Safety

' and Safeguards Enc'osures:

1. Appendix A, Notice of Violation

2. NRC Region I Inspection Report No. 85-01 _

cc w/encis:Dr. O. K. Harling, Director of the Reactor Laboratory

! Public Document Room (POR)

Local Public Document Room (LPOR)

Nuclear Safety Information Center (NSIC) h Commonwealth of Massachusetts (2) ounonu~,m,

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Region I Docket Room with concurrences)

Senivr Operations Officer (w/c :ncis) l l

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APPENDIX A NOTICE OF VIOLATION Massachusetts Institute of Technology Docket No. 50-20 Cambridge, Massachusetts License No. R-37 As a result of the inspection conducted on February 13-15, 1985, and in accord-ance with the revised NRC Enforcement Policy (10 CFR 2, Appendix C), published in the Federal Register on March 8, 1984 (49 FR 8583), the following violation was identified:

J 10 CFR 71.5 requires each licensee, who transports licensed material outside of the ceafines of its plant, to comply with the applicable requirements of L the regulations appropriate to the modes of transport of DOT in 49 CFR Parts 170 through 189. 49 CFR 172.203(d) requires that the description for a ship-ment of radioactive material must include: (1) the name of each radionuclide; (ii) a description of the physical and chemical form of the material; and

' (iii) the activity contained in each package in terms of curies, millicuries, or microcuries.

f.t Contrary to the above, on September 13, 1984, a package containing 281 milli-curies of rhenium-186 and 824 millicuries of rhenium-188 as rhenium wire was

f labeled and shipped with the incorrect description of radionuclide, physical and chemical form, and activity for the package. This material was shipped y

to Massachusetts General Hospital, identified as 8 millicuries of chlorine-38

., in the chemical form of calcium chloride salt.

} This is a Severity Level IV violation (Supplement V).

Pursuant to the provisions of 10 CFR 2.201, Massachusetts Institute of Technology is hereby required to submit to this office within thirty days of the date of the letter which transmitted this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been

).

'- taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be

^ achievea. Where good cause is shown, ~ ideration will be given to extend-ing this response time.

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U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-20/85-01 Docket No. 50-20 License No. R-37 Licensee: Massachusetts Institute of Technology Researen Reactor 138 Albany Street Camericoe, Massacnusetts 02139 Facility Name: MIT R ,

Inspection At: Cah: bridge, Massachusetts r

Inspection Conducted: February 1 715 1985

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Inspectors: / // s / J' Jea/A. Cio7fi .' ~Radijgion Specialist '

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(/ l l John R. 4 Kite, Senfor Radiation I/2 Yf

' cate Streciali st PWR Radiation Protection Section Approved by: 4 'A I2v M.' Shanbaky, Chief ~/

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Inspection Summary- i Inspection on February 13-15, 1984 (Report No. 50-20/85-01). '

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Areas Inspected: Routine, unannounced safety inspection of the radiation i

protection program, including: the status of previously identified items; radiological surveys, postings, material labeling, and controls; equipment,

] instrumentation, and leak tests; environmental monitoring.

u, The inspection involved 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months on-site by two region-based inspectors.

Results: Of the areas inspected, one violation of transportation requirements was icentified, i.e., failure to properly label a radioactive shipment ,ith respect

' p.- to radionuclide identity, physical and chemical form, and correct activt y as

,g; required by 49 CFR 172.203(d), paragraph 5.0.

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DETAILS 1.0 Persons Contacted P E. Karaian, Reactor Radiation Protection Officer g.

L. Clark, Jr., Associate Director, Nuclear Reactor Laboratory

) P. Coggio, Reactor Radiation Protection Technician 2.0 Purpose The purpose of this routine, unannounced, safety inspection was to review the licensee's radiation protection program with respect to the following

! elements:

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Status of Previously Identified Items

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Radiological Surveys, Posting, Material Labeling, and Controls Radioactive Material Identification f

Equipment, Instrumentation, and Leak Tests Environmental Monitoring s

v 3.0 Status of Previously Identifiec Items .

f 3.1 (Closed) Violation (S3-02-01). Failure to post hot cell on reactor 3

top as high radiation area and to control personnel access to the area. The licensee's corrective actions, as stated in Inspection

> Report 33-02, Section 5b, were reviewed to verify their implementation.

g; Implementation of the corrective 4.ctions appeared to ce adequate to

. prevent recurrence, e

[< ~ 3.2 (Closed) Follow-up (83-02-02). Radiation protection to control use

,, of radiation barricades and signs to avoid misuse. For other than f radiological control purposes, the licensee has purchased white ropes for the researchers to use to enclose their equipment and experiments.

3.3 (Closed) Follow-up (83-02-03). Radioactive contamination control by individuals working in materials laboratory section of Engineering i laboratory. The floors in front of Hoods 1 and 2 of tN12-139 are surveyed daily for contamination. Monthly surveys are performed for the entire laboratory area. Contaminated areas are cleaned immediately.

4.0 Radiological Surveys, Postinos, Material Labeling, and Controls The licensee's program for surveys, postings, labeling, and controls was j

reviewed against the criteria contained in 10 CFR 20.105, 20.106, '0.201,

' s r ,; 20.203, 20.204, 20.207, and 20.401.

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The licensee's performance relative to these criteria was evaluated by:

a. Examination of records of daily and monthly surveys for 1983 and 1984;

b. A tour of the facility;

c. Observation of signs and postings on equipment, in laboratories, in aallways, and on doors;

d. Direct radiological measurements of areas in the facility with a GM detector and a "Juno" ionization chamber;

e. Observations of access controls for the reactor building, and for monitoring activities within the reactor building; and

f. Discussions with licensee representatives.

The inspector noted the following:

Gamma surveys and smears are taken daily on floors and in common areas.

Monthly surveys are performed in laboratories and near equipment and radioactive waste storage areas. The Radia ion Protection Officer is informed daily of any contaminated areas through the use of a daily status sheet, which identifies the contamir,ated areas and states the corrective actions taken.

Access to the reactor building and auxiliary facilities, such as the

. radwaste storage areas and laboratories, is controlled with a key card.

Personnel entering the reactor building are required to call the control room and notify the operator of their intent to enter. The entrance to the reactor building is observed in the control room by a closed circuit TV camera. The TV camera can also be moved to observe approximately half of the reactor floor and the top of the reactor where a hot cell is located.

There were no violations identified in review of this area.

5.0 Radioactive Material Identification The inspector investigated two incidents, which occurred on September 4, 1984 and on September 13, 1984, in which rhenium wire samples were mistaken for other radioactive samples. On September 4, 1984, a reactor operator was directed to package and release a strontium chloride sample by an experimenter. The wrong sample was mistakenly packaged, but not shipped when it was fortuitously determined that the sample was erroneously identified.

In the incident occurring on September 13, 1984, a rhenium wire consisting of 281 millicuries of rhenium-186 and 824 millicuries of rhenium-188 was packaged, labeled, and shipped to Massachusetts General Hospital. The package was labeled and shipped as 8 millicuries of chlorine-38, in the 3 form of calcium chloride salt. This incident constituted a violation of 49 CFR 172.203(d), which states that each package of radioactive material must be identified as to radionuclide identity, physical and chemical form, and amount of activity.

As a result of this occurrence, one Massachusetts General Hospital

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employee sustained minor unplanned exposure to the wrist and whole body,

. of 150 millirads and 25 millirads, respecti.T v. 1 The licensee determined that the cause of this occurence was

, misidentification of the samples on the sample storage mao, located on the outer wall of the hot cell. As a result, the licensee initiated the corrective actions listed below:

a. Two internal memos were circulated emphasizing the imoortance of accurately marking the identity of samples on the sample map located outside the hot cell. The memos also stated that beta surveys were to be performed on all samples in addition to gamma surveys. All reactor operators initialled the memo to verify that they read it.

b. A lead container was placed in the hot cell and labeled "Rhenium Only," for the placement of the rnenium wire samples. Rhenium wire samples are now placed in this container only.

The inscector interviewed one reactor operator to evaluate the effective-

. ness of this corrective action. The reactor operator stated that the beta survey was not being performed. All otner corrective actions were being implemented.

The inspector discussed the failure to perform the oeta survey with the Radiation Protection Of ficer (RPO). The Radiation Protection Officar stated that he disagreed with the newly instituted reovirement to perform beta survey of the sampies because tnis practi:e would increase exposure to the operators wnicn nou'd not be consistent witn good ALARA practices.

On February 21, 1985, the Radiation Protection Officer and the Associate Director, Nuclear Reactor Laboratory, telephoned the inspector te present new corrective actions. Tne following actions were discussed with the inspector,

a. The requirement to beta survey would be eliminated;

b. A specific procedure will be written for work in the hot cell. The identity of the sample will be specified in one or more of the follow-ing ways:

(1) Use of the sample reference number; (2) Use of any distinguishable mari- or, the sampie and the mark

$ recorded on the work form (Part II); and (3) Use of any unique shape of the samples and the shape recorded on f"[t.' . the work form (Part II).

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c. The gamma dose will be verified on the work form; and

,- d. The importance of confirming the identity of the sample with the work form (Part II) will be reemphasi:ed.

The effectiveness of these corrective actions will be examined in a future inspection (85-01-01).

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6. Eauicment. Instrumentation, and Leak Tests '

Tne licensee maintains logs of all instrument calibrations. Survey instruments and monitors are calibrated quarterly. Effluent radiation g monitors are calibrated yearly and checked on a quarterly basis for response to a radioactive source.

Argon sampling and monitoring are performed continuously using a GM detector which views a known volume of gas. A strip chart records 7.11 ;

data. Counts are summeo over one week. Additional air sampiing quipment ,

is mounted cn carts and moved to various locations, such as port openings, when needed, t

Leak tests are performed quarterly and semiannually, depending on the type of source. The licensee has determined that the lower level of detection for their leak tests is 1.2 x 10 -6 microcuries alpha, and 9 x 10 -6 micro- !

curies beta. Accurate recards are kept of leak tests with a clear oescription of the type of wipe (i.e. , dry or wet).

] Within the scope of this review, no violations were observed.

7. Environmental Monitorino The licensee's program for environmental monitoring was reviewed against criteria contained in 10 CFR 20.106 and Appendix B, Table II.

1 The licensee's performance relative to these criteria was evaluated by:

a. Visual inspection of two separate environmental monitoring stations for working instrumentation, and i oc7 tion with reference to the reacter stack;

b. Discussions with the Reactor Hesith Physics technician and the Radia-tion Protection Officer on the calibration of the instrumuntation and data collection and calculations; and

c. Review of the annual reports for 1983 and 1984 The licensee uses GM tubes for their environmental monitoring stations.

Each GM tube is connected to a count rate meter located inside a sheltered area. The signal from the count rate meter is sent through telephone transmission lines to strip chart recorders located inside the Health Physics office at the Nuclear Reactor Laboratory. Data on the strip chart recorders is collected daily and summed monthly.

f Within the scope of this review, no violations were observed, i

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8. Exit Interview The inspector met with licensee management at the conclusion of the inspection of February 15, 1985 to discuss the scope and findings of the

, inspection as detailed in this report. At no time during this inspection effort was written material provided to the licensee by the NRC inspector.

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APR 0 91985

Docket No. 50-20 Massachusetts Institute of Technology ATTN: Lincoln Clark, Jr.

Director of Reactor Operations 138 Albany Street Cambridge Massachusetts 02139 Gentlemen:

Subject:

Inspection Report No. 50-20/85-01 This refers to your letter dated February 27, 1985, in response to our telephone conversation of February 21, 1985.

Thank you for informing us of the corre:tive and preventive actions documented in your letter. These actions will be examined during a future inspection of your licensed program.

Your cooperation with us is appreciated.

Sincerely, 0 inal 58.gned BYt

'Q 7U ThmasT. Martin,Directo['

[j): Division of Radiation Safety and Safeguards CC:

Public Document Room (PDR) 1 Nuclear Safety Information Center (NSIC)

Commonwealth of Massachusetts (2) /

Or. O. K. Harling Director of the Reactor Laboratory /

Local Public Document Room (LPOR) i bec:

Region ! Docket Room (with concurrences) f Vh C40ffi ShaIaky, (sj ,(4 ( }h- l/@

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MASSACHUSETTS INSTITUTE OF TECHNOLOGY O K HAALING 138 Albany Street Camcridge. Mass C2139 L CLAAK. JR.

Director (617)253 4202 Director of Aeacter coerations s Februa ry 27, 1985 Dr. Thomas E. Hurley, Adminis tra tor Region 1, U.S. Nuclear Regula tory Commission 631 Park Avenue King of Prussia, PA 19406

Subject:

Inspection No. 50-20/85-01

Dear Dr. Murley:

In connection with the above health physics inspection at the MIT Research Reactor on February 13-15, 1985 by Ms. Jean Cioffi and Mr.

John White, it was requested that MIT provide by mail the following two items of informa tion, since they were not available at the time of the inspection.

The first item is the paperwork associated with a radioiaotope shipment that was made from the MITR to Massachusetts General Hospital (MGH) on September 13, 1985. The enclosed documents (Irradiation Informa tion Form - Part II and Isotope Shipping Memo) relate to a shipment that was intended to be chlorine-38, but a rhenium-186 wire sample pas inadvertently packaged and shipped instead. It should be recognized that the MITR has properly shipped more than three hundred chlorine samples to MGH and many thousands of other samples to other facilities. This ins tance is the only time tha t the wrong ma terial has been released.

The second item is related to the preparation of a document covering the procedures that MITR operators are instructed to follow for the release of the irradia ted ma terials for shipment. The procedure as now written specifies that:

1) Samples to be irradiated in the pneuma tic tubes be identified by some distinctive feature or marking and, prior to the release of an irradiated sample, its identity be verified against this distinctive f ea ture.

3 In addition, the procedure incorporates long-s tanding existing practice which was and is tha t:

2) A chart be maintained showing locations of irradiated samples stored in either of the pneumatic tube sample changing tress, f $ Il D h M l - u < w 17 s s' W f v- .

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3) Prior to release of any irradiated sample, it be identified by reference to the chart and to the Irradiation Information

,, Form - Part II, its expected gamma dose be confirmed, the sample be packaged as required by DCT and/or NRC regulations, necessary papr;rwork be completed, and the

, s torage chart be upda ted.

The written procedure covering the above was approved and issued on February 27, 1985.

Please contact me if any further information is required in this regard.

Sincerely, G w . ,e d - (_J ew A. J'~

Lincoln Cis.rk, Jr.

Director of Reactor Operations cca J. Bernard, MIT J. Cioff1, NRC Region 1 E. Karaian, MIT

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$ b.'fthI wry h& .s /N . 3 h /'r' d 2L4 s - ) ,s q N' at

'N M MW Reactor down time Total Time .

Activities:

! l(c <W C mr/hr at [O Gamma activity 7% %dmNbd (M7h,J op hg/

Sampie contamination: wipe activity Date S S f'I'N M B'v '

'M "'

/A'/M /o/ i Time / hus: 'll Shippinr:

Container radiation levels:

I'"

/

mr/hr: at I meter

'Y (d- er surf ace .- 15 / -

!*;,gI'g ~

con'.ainer wipe test activity bY index label: principal isotopes _

date time _

Release : or other disposition, by Waste Disposal Uncer License No.

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_ Mud -

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9- / I -f 4 Signature of person accepting sample

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Charre :

Orgsnt:ation

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.1 C0 Handhn;! ~' #' -

Total $

Facilitv $ [ / !d,f $8 f 1

Account No._ /yj

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..AM~

R 3

faNilCLEAR REACTOR LABORATORY

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v w* i (3'597

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Massachusetts Institute of Technology

" --. l. ;-

138 Albany Street Cambridge, Mass'. 02139 '

N9 628 )

IOTOPE SHIPPING MEMO Ship to (1f dliferent from invo6ce to)

Invo6ce to:

nt. e.Hocp NM Gsv&o.a. ho.sp,Tn-AITR irradistle..i Ref. No.

Terms Lustomer Order No.

Customer License No.

20-o3d/9 - Jo

' 70-9P /or Total Wolght 5hl ped in desk No. & Type Seal , outing Dacks,ges Wolg@hVPtc. @

pgM '7/pd A d.[ M Medical O Industrial _

Physical Fonn Chemical Form

' Lot No. Product Wolght or Volume N

M w ee Cscl>

Cl" o.1%

Unit Price Amount Calibrated Acth'ty tTotal Activity Shipp+d Specific Acthrity Cal. Date & Time Shipped Aa $ hipped Ordered >

1 __

i Other Charges (tist) l l

l Totsi Special instructione- A A0i ATION SU AVE 1 This survey is tne result of reacings tamen af ter the matenalis comUetely cacua;ed reacy snioment.

YSmearacle ecstamination meets U S.O.C.T. - recuirements.

M r/ H f Maximum surisce ts1:stion' Transoortinces(MrfHr C 3 f t.)

/ f C EsclustvoUseVehicle Lades af finee. O h ts1. f.Yehow it, O '/*Hed Hf. I NR S:rtece MtrH r

_ M rtHr "" Vehjcie placarcoc At 8 f eet dnis is to certify that the named articles are crecerly cosented. cackaged, marted Psenacing & Sniecing information ano isovied and are in crecer condition f or transportation and en accorcance with aceticaole regulations of the U.S.O.O T. and I AT A restnetec article regutatices.

This sn<pment is *ithin the limits crescritec for gassenger carrying a>rcraf t Desenotion act: active Malenal N O $ M 7/'/-

{

Aasiation $yrvey Dy _

> acioactive Material 5ea-o'- Cavi.o, Tn,s ,natemai Aja eo, coe, sie,iii,ec o, reni,,ec .., meecal u present form.

Transport Group wb. ^

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I MIT-NRL Shipping Checklist for DOT 7A-Type A Reusable Containers t

This checklist and an MIT-NRL shipping me=o should both be completed prior to re-

i. leasing any non-liquid radioactive sample in a DOT 7A-Type A reusable container.

These for=s document the mechanical integrity of the container. Refer to the isotope shipping memo for radiation levels and required labels.

i MITR Irradiation Ref. No. SC~ M3- /01 .

Isotope container integrity satisf actory with product packaged in 2R inner container with packing material sufficient to prevent r.ovemann and/or ge and d in place and cap secured on 2R inner container.

Threads of 2R container lubricated with graphite, teflon tape etc.

2R , container cover screwed on with at least five turns.

/

V Packing material in place about 2R container. Top section of packing is within two inches of drum ccver but not so close as to inhibit proper at chment of cover / bolt assembly.

h f D;ui cover and gasket intact.

V Drum cover and bolt assembly in position over cover and drum rims with bolt readed through ring eyes with ring separation at ends less than one inch.

S rity seal wouad around bolt eyes end to end and through bolt hole.

> _ _ Security seal sealed with cri=per.

(

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Signatute Date /

p h

File with Operations copy of the Irradiation Reference Form, i.

c;,i.".gc_5 June 2:. 1982

s p

'i O O .

l NOV :: r;35 .

4 Docket No.50-020 Massachusetts Institute of Technology Research Reactor

ATTN: Mr. Lincoln Clark, Jr. l Director of Reactor Operation 138 Albany Street Cambridge, Massachusetts 02139 :

Gentlemen:

SUBJECT:

EXAMINATION REPORT NO. 50-20/85-03 (OL)

! This transmits the Examination Report of Operator Licensing Examinations conducted by USNRC Region I at the MIT Facility the week of September 30, ,

t 1985. At the exit interview held with Mr. J. Bernard and Mr. K. Kwok on Octo'oer 1,1985, the preliminary results of these examinations were discussed.

No reply to tr.ds letter is required. Your cooperation in this matter is appreclated.

Sin erely, i

Cric3 S- .J.:dly l Edward C. Wenzinger, Chief !

Projects Branch No. 3 l i Division of Reactor Projects i

Enclosure:

i Examination Report No. 50-20/85-03 w/ attachments 1, 2 cc: w/ enclosure and attachments 1, 2 i Dr. O. K. Harling, Director, Reactor Laboratory 4

John Public A. Bernard, Document RoomTraining)

(PDR Coordinator h

, Local Public Document Room (LPOR)

Nuclear Safety Information Center (NS!C)

State of Massachusetts L

OFFICIAL RECORD COPY OL MIT ER - 0001.0.0 l 10/29/85 g$ >l

-Sn =4 % g Ch? '

u

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o O 4 Massachusetts Institute of Technology 2 h

l 3 bec: w/ enclosure, w/o attachments 1, 2 )

DRP Section Chief I D. Silk, Examiner l

Chief, OLB/DHES,NRR 1 OL File 12.0 Region I Docket Room (w/ concurrences)

Master Exam File

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f. OFFICIAL RECORD COPY OL MIT ER - 0002.0.0 10/29/85

,m . - _ _ _ - ~ ~ ~ _ - - - - - - - - - - * - - - * " - ' * * - # #'

A i

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REPORT DETAILS TYPE OF EXAMS: Initial __ Replacement _X_ Requalification EXAM RESULTS:

l RO l SRO l l Pass / Fail l Pass /Faill

l l l Written Exam l 2/0 l 1/0 l l l l r Oral Exam l 1/1 1 1/0 l ;

I I I ;

Overall l 1/1 l 1/0 l l l l

1. CHIEF EXAMINER AT SITE: David M. Silk

2. OTHER EXAMINER: Robert M. Keller

1. Summary of generic strengths of deficiencies noted on oral exams:

Canaidates displayed a good understanding of the plant. SRO candidate displayed a weakness in not assuming all responsibilities assigned to SRO by transferring responsibilities to plaat management personnel who hold l

SRO licenses.

2. Summary of generic strengths or deficiencies notad from grading of I written e ams: ,

Candidates were not familiar with:

relationship of early xenon peaking to harder neutron spectrum modet of operation for the 1-inch pneumatic tube system the horard of drying out charcoal filters how to seal beam ports

3. Comments on availability of, and candidate familiarization with plant reference material in the control room:

Candidates were familiar with plant reference material.

]Y Q' 94.LSe frtv% ;a- h

s 4

. 3

4. personnel present at Exit Interview.

NRC Personnel David Silk Facility personnel John Bernard Kwan Kwok

5. Summary of NRC Comments made at exit interview:

Two of the three candidates were clear passes on the oral examination.

Facility training material provided for examination preparation was well organized.

6. CHANGES P.ADE TO WRITTEN EXAM DURING EXAMINATION REVIEW:

Question No. Change Reason A.2 Delete question. The question called for a compsrative knowledge of reactor tyces.

B.2 Include in answer Expands answer "Verify system pressure". Key.

B.6 Delete from answer Inoperable at present.

"the pitch of fan blades can be changed".

C.1 Delete from question Can mislead cand'.date.

"The reactor has just Clarifies question, been started".

C.7 Also accept Answer a). Unusual Occurrence Report #81-4 justifies answera).

0.5 Delete question. This experiment (FCE) has been out of the reactor for two years.

O O 4

Question No. Change Reason E5 Also accept Answer c). If some loads are shed, the battery could supply power for about 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

E.7 Include in answer These alarms will Weekend transmit a signal to Intrusion the campus Patrol Alarm (Interior / System.

Exterior)

Fuel Vaults Operator Incapacitated Panic Button (In control room or receptionist desk)

G.1 Include in answer Expands answer "Check radiation levels" Key.

"Order personnel out" G.3 Also accept Answer c). Surface contamina-tion includes beta radiation.

G.8 Include in answer Expands answer "Gas monitor on reactor Key, ficor by main airlock".

G.11 Include in answer- Expands answer "To prevent nitric acid Key, formation from nitrous oxide".

J.2 Include in answer e). Expands answer Reactor floor het Key.

36V's if not sealed A drop in building temperature Include in answer b).

Use helium gas

- Seal ports Attachments:

1. Written Examination and Answer Key (RO)

2. Written Examination and Answer Key (SRO) s

e U.S. HUCLEAR REGULATORY C011115 N MRS E/f

~

REACTOR OPERATOR LICENSE EXAnlHATION Facility: MITR-11 Reactor Type: HWR/ LWR Cooled / Moderated Date Administered: October 1.1985 Examiner: W. J. Apley / J. C. Huenefeld Candidate: Answer Key INSTRUCTIONS 10 CANDIDATE:

Use separate paper for the answers. Write. answers on one side only. Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses af ter the question. The passinD grade requires at least 70% in each category. Examination papers will be picked up six (6) hours af ter the examination starts.

Category 1 of Candidate's 1 of Yalue Total Score Cat. Value Category 15.0 14.9 A. Principles of Reactor Operation 14tT' 13.9 B. Feetures of Facility Design 14.5 14.4 C. General Operating Characteristics 15.0 14.9 D. Instruments and Control s 15.0 14.9 E. Safety and Emergency Systems 13.5 13.4 F. Standard and Emergency Operating Procedures 14.0 13.9 ~

G. Radiation Control and Safety

, 10 0 lektr' TOTALS j Final Grade 1 All work done on this exam is my own. I have neither given nor received aid.

Candidate's Signature e

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MITR-!!

October 1, 1985 A. PRINCIPLES OF REACTOR OPERATION (15.0)

Points Available 00ESTION A.1 When calculating an estimated critical position, the operator uses the previous week's position and corrects for five different delta K changes. List four (4) of those delta K changes. (2.0)

ANSWER A.1 Delta K due to temperature change ,

due to sample loading due to Xenon due to fuel loading due to burnup (4 of 5 for full credit)

REFERENCE A.1 PM 3.1.1.2, p.11 QUESTION A.2 The d'!TR-II reactor produces a relatively fast response to a given reactivity input. Explain that response in terms of what the values of neutron generation time and delayed neutron fraction are at MITR-!!. (i.e., are both Beta and generation time small. one j small and the other large, etc.) (2.0)

ANSWER A.2 l The sensitive response is due to the short neutron generation time for the MITR-!!. even though its delayed neutron fraction is large .

(beta-bar = 0.00786). The large Beta effective is predominately due to a large source of "slow born" photo neutrons developed in the reflector.

l REFERENCE A.2 RSM 10.5

-Section A Continued on Next Page-

~

j , ,

.c

2 r

l

. MITR-Il l October 1. 1985 ._

1 Points ,

Available '

OVESTION A.3 1

WA isn't the MTR type elements cladding thicker or thinner? (1.5)

ANSWER A.3 l

It's thick enough to retain fission products (+0.5), and thin enough to not introduce a long delay time for heat removal in the event of a fast transient (+0.5).

i i REFERENCE A.3 Tech Spec 5-4

]

QUESTION A.4 l

t Explain the two (2) ways that the control elements affect reactivity l 45 they are moved in the core. (1.5) l ANSWER A.4 i

When inserted in the annular space between the core and the core i housing assembly, these control elements decrease reactivity both l by the direct absorption of neutrons and, to a lesser extent, by l warping the core flux distribution thereby increasing neutron leakage.

(+1.0 for absorbtion/+0.5 for increasing leakage)

REFERENCE A.4 RSM 10.5 i

l 4

i

-Section A Continued on Next Page-g

. 4

o .

3 4

MITR-!!

October 1. 1985 Points Available QUESTION A.5 If the reactor is on a stable 25-second period, how long will it take to change power level 2 oecades (show calcuTation)? (2.0)

ANSWER A.5 from equation sheet:

Sur = Y = h = 1.0424 P = Po 10sur t ,

h=100=10surt o =

2 = sur t t = 2/1.0424 = 1.92 minutes if the candidate doesn't know about SUR (which is checked w/o calculation in A.1), then he can calculate using formula sheet.

P = Pn e t/T P/Po 100 = et /25 see in 100 = t/25 see t = (20 sec)(in 100)

= 115. 0 seconds = 1.92 minutes.

REFERENCE A.5 Glasstone and Sesonske (MITR Trng Progr. Ref.)

PM 1.16.2. p.1

-Section A Continued on Next Page-

a. e *

=- -

4 MITR-il '

October 1. 1985 Points ;

Available QUESTION A.6 TRUE or FALSE: Xenon peaks earlier in MITR-Il after shutdown due to a harder neutron spectrum. (0.5) l ANSWER A.6 !

True REFERENCE A.6 t RSM 10.7 .

QUESTION A.7 Describe the two (2) phenomena that contribute to the temperature l coefficient of reactivity for MITR-!!. (2.0)

ANSWER A.7 The first is the temperature rise of the light water due to an increase in the thermal output of the reactor core. Any such temperature rise will insert negativo reactivity by causing a ,

, hardening in the neutron spectrum. (This means that the average

{ neutron takes longer to thermal 12e so there are fewer fissions.)

The second phenomenon is the radiation heating of the heavy water reflector. Temperature rises of this type add regative reactivity by allowing more neutron leakage to increase. This second process i

legs the temper 3ture rise of the light water in the core proper.

l REFERENCE A.7 RSM 10.8 l

1 i

1 -Section A Continued on Next Page-c . -

L - .- :---_-_--__ _

~

. - O 5

0

.. . MITR-II October 1. 1985 Points Available QUESTION A.8" l

If heavy water leaks into the light water system, what type of reactivity effect will it have if:

A. The leakage of pure, uncontaminated heavy water is into either

the light water reflector above the top of the core, or the light water reflector below the top of the core that is formed by the annular space between the core and the si63s and bottom of the core tank. (0,5)

8. Leakage of heavy water is into the core proper. (0.5)

C. The in-leaking D2 0 orogressively replaced the entire light water system. (0.5)

ANSWER A.8 A. Positive reactivity B. Strong, negative reactivity

. C. Strong, negative reactivity REFERENCE A.8 RSM 10.11

.}

1 i

t l

-Section A Continued on Next Page-t . .

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MITR-ll Octobe? 1, 1985 Points ;

Available QUESTIONA.9 A nuclear reactor has a shutdown margin of 7% delta k/k and a neutron of tector is recording 20 cpm. What will this detector read ehen

] (gf-0.997 (2.0) y

- \s

. ANSWER A.9 1 - Kg

= 0.07

, Kg 1 - K) + 0.07 K) 1 - 1.07 K)

K) = 1/1.07 - 0.93 1-K 1 Cpp

,, 1-K 2 ' Uk 0.07 CR p OT " 2T CR2 - 140 com .

i l

! REFERENCE A.9 -

1 Generic: "Academic Program for Nuclear Power Plant Personnel."

Volume 11. pp. 5-6 through 5-13. General Physics corporation. ;

i

2. Glasstor.e and Sesonske (MITR Trng. Progr. Rev.) '

, PM 1.16.2. p.1 i

-End of Section A- i l

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MITR-!!

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l October 1. 1985 l

B. FEATURES r: FAClllTY DESIGN (14.0)

Points Availble JUESTION B.1 Describe the four (4) modes of operation for the 1-inch pneumatic tube system. (2.0)

ANSWER B.1 A. Insertion and removal at the hot cell or primary chem room in the reactor basement.

B. Insertion at the hot cell and transfer of the irradiated sample to the NW-13 hot lab via the connecting pneumatic tube.

C. Insertion from the NW-13 hot lab. into the reactor, and transfer of the irradiated sample back to the NW-13 hot lab.

D. Transfer of a rabbit from the basement hot cell to the NW-13

hot lab.

REFERENCE 3d PM 1.10, p. 7 QUESTION R.?

How does the operator verify that the secondary system is properly Tined up to cooling tower basins? (1.5)

ANSWER B.2 Verify secordary system is propery lined up to cooling tower basins by either checking HV-14 of HV-14A open or by checking HM-1A running with flow through HF-3 at 60% of scale. (Etther answer correct.)

REFERENCE B.2 PH 3,.1.1.1, p. 2

-Oection B Continued on Next Page-

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MITR-l!

October 1. 1985 Points Available OVEST10N B.3 2 :

What design safety feature ensures that fuel loaded into the core will normally have access to only one core position at a timi? (1.25)

ANSWER 8.3 Hold-down grid latch must be released and the grid rotated to -

permit core access. Grid design prevents multiple position access. ,

c REFERENCE B.3 ,

PM 2.7. p.3 t L

+

QUESTION B.4 If the pressure relief system's charcoal filters become submerged, what problems will exist during filter housing and exhaust dryout? (1.25) i ANSWER B.4 4

The charcoa) generates heat while drying out and may eduse spontaneous combustion.

REFERENCE B.4 PM 5.2.14. p. 2 i

1 6

j i

-Section B Continued on Next Page-

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9 4 .

MITR-!!

October 1. 1985 Points Available OVESTION B.5 Explain how the anti-syphon valves work. (1.0)

ANSWER B 5 Ball float valves installed at the top of the core shroud. Inlet flow forces ball up closing outlet at topt w/o flow gravity forces ball down to break syphon.

P l

REFERENCE B.5 RSM 1.7 00EST10N B.6 List three (3) ways to reduce the degree of cooling tower efficiency I t

on cold days. (2.0)

AhSWER B.6

)

The yard booster pumps may bt hypassed partially or conpletely. ;

j as may the towers themselves. One of the cooling tower fans may be '

operated at half-speed the pitch of the fan blades can be changed.

and the air admitted to the towers can be restricted by rearranging the external boards and flaps. !

l (Any three.)

REFERENCE B.6 RSM 3.12 i

1

^

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-Section B Continued on Next Page-i

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. ----- - --- . - , , - . - - , , . - - - - - , - . - - - -..n,, _ , , , _ _ - -

'O MITR-Il October 1. 1985 Points Available 00EST10N B.7 Hy are beam ports cealed? (1.5)

ANSWER B.7 A. A plug is placed in vort B. Gas seals i C. Gasketed cover bolted over beam port's opening l

l l REFERENCE B.7 RSM 2.4 00EST10N B 8 Assume a loss of external electrical power feeders occurred. .

When normal power is later restored, what will happen to all the transfer switches and the inotor generator set? (1.0)

ANSWER B.8 A. Transfar switches return to normal.

B. Relay at the motor-generator set is energized thereby stopping the unit.

REFERENCE B.8 RSM 8.32

-Section B Continued on Next Page-

. .-- 1

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O "

o MITR-l!

October 1. 1985 Points Available QUESTION 8.9 l.T j Draw a top view of the core including location of the: M A. Regulatt..g rod (.5)

B. Shim blades (.5)

C. Radial absorber plates (.25)

D. Hexagonal absorber plates (.25)

ANSWER 8.9 See next page.

'! REFERENCE B.9 -

See attachment.

1 I

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-Section B continued on next page-1 4

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Answer 8.0 Core Section H.J.T.R. Il Reactor Systems Manus) t l -End of Section B-l l

13 MITR-!!

October 1. 1985 C. GENERAL OPERATING CHARACTERISTICS (14.5)

Points Available OVESTION C.1 The reactor has just been started up. Explain why nuclear instrumentation must be frequently calibrated in terms of thermal power as short lived fission product poisons (such as Xenon) build up in the reactor core. (3.0)

ANSWER C,1 Compenstation for the negative reactivity associated with the building "

in of equilibrium menon is achieved by withdrawing the shim blades.

The out-motion of the shim bank causes the axial flux profile of the reactor to change with the point of maximum flux moving upward. That, in turn, alters the leakage flux which is what is viewed by the nuclear instrumentation. This affects reactor control in the following manner.

The automatic control system controls the reactor by maintaining a constant flux at the locction of the chamber that feeds the auto-control network. Hence, as the axial flux profile changes with shim bank ,

height the auto-control channel will detect a "power-change". In [

1 reality, vf course, there is no net change in power, but a redistribution of power within the core. This is why it is essential to determine the thermal power output of the reactor by means cf a heat balancs which is not affected by flux distribution.

(+1.0 - change in axial flux profile)

(+1.0 - auto-control "sees" power change) l

(+1.0 - need to re-calibrate to thermal, not distributed power) '

REFERENCE C.1 PM 2.4. p.1 4

l i

1

-Section C Continued on Next Page-

14 -

e . MITR-!!

October 1. 1985 Points Aytiloble QUESTION C.2 Whis the maximum amount of reactivity in percent of delta k/k that may be added to the critical reactor without causing damage to the fuel integrity by the resulting power transiont? (1.0)

ANSWER C.2 1.8%

l REFERENCE C.2

Tech Specs 3-8 OVESTION C.3 h does it .oe 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the reactor to be in thermal equilibrium, such that a heat balance can be conducted? (1.0)

ANSWER C.3 Graphite reflector has a large heat capacity and is slow to attain an equilibrium temperature distribution, l

REFERENCE C.3 RSM 6,4 j

1 i

j -Section C Continued en Next Page-4 i

, ,, . . -, . _ _ - - - _ _ - - - . - - - - - - - - - - , - ------.~--,e e ,

15 MITR-!!

Odtober 1, 1985 Points Available

QQESTIONC.4 WhJ is "blowdown" of the water in the Forced Oraft Cooling Towers required? (1.0) .

ANSWER C.4 Forced draft cooling towers concentrate the solids in the makeup water and collect atmospheric dust. Hence, a feed-and-bleed purge is taaintained while they ace in operation in order ta keep the level of dissolved solids within a factor of three to five times that of the makeup water. A small portion of the water is diverted through a flow accumulation meter directly to the sewer. This flow is called "blowdown",

i REFERENCE C.4 RSM 3.12

, t i

i i

I I

r I '

n

-Section C Continued on Next Page-l l l !

16 n O MITR-II Oc'tober 1, 1985 voints Available OUESTION C.5 A. Explain how the reactivity effect of dumping the radial reflector varies with the position of the shim blades. (1.5)

B. Why is the radial heavy water reflector pumped up with the

~

shim bank in the fully inserted position? (1.0)

ANSWER C.5 A. In as much as the shim blades also operate in the region between the core and the redial heavy water reflector, the reactivity worth of dumping this radial reflector is dependent

. on the position of the shim blade bank. This effect can be considered as being due to the shadowing influence that the

,. blade bank exerts on the reflector. These results show that the reactivity worth of dumping the radial heavy water reflector when the shim bank is fully inserted is about

- two-thirds that of the corresponding value when the bank is at the top of the active core.

(+0.5 for reason, +1.0 for knowing more reactivity with rods at top. )

B. Safety considerations dictate that the radial heavy water reflector be pumped up with the shim bank in the fully-inserted position. This ensures that the reactivity insertion for this process will not occur when the reactor is or could go critical.

REFERENCE C.5 RSM 10.6

~.

l

-Section C Continued on Next Page-l l

l I

l

17 N

. MITR-!!

October 1. 1985 Points Available QUESTION C.6 You receive a high temperature shield coolant outlet alarm.

The shield coolant outlet temperature is rising slowly, and there is no evidence of a loss of shield flow or level.

Operationally, what is the probable cause? (1.5)

ANSWER C.6 The secondary side of the heat exchanger is probably clogged with mud. (Will accept other answers alluding to degraded HX performance.)

REFERENCE C.6 PM 5.4.8 OVESTION C.7 Approximately how long after a failure of the pneumatic blower (at full power 7 will the temperature in the pneumatic tubes reach 100 degrees C (select best answer)? (1.0)

A. Instantly B. 5 Minutes C. 30 Minutes D. Never A*!SWER C. 7 B. 5 Minuto -

REFERENCE C.7 PM 5.5.1

-Section C Continued on Next Page-

. t 18

[

MITR-!!

October 1. 1985 -

~

j Points Available QUESTION C.8 l TRUE or FALSE: It does require bypassing a number of safety functions, but it is possible to operate in the 100 kw mode with no forced circulation of primary coolant. (0.5) 1 t

ANSWER C 8 True l REFERENCE C.8

\

PM 2.2 i ;

i t J l QUESTION C.9 t j Describe how to calculate the total thermal power output of the (3.0) i j reactor, ,

i ,

i ANSWER C 9 {

Primary Power = (2.62 10~4)(Primary Flow)(Primary delta T)

\ . L Reflector Power -

(2.91x10")(0 20 Flow)(D2 0 delta T)

] {

l Shield Power (2.62x10"#)(Shield Flow)(Shield delta T) !

. l l Total Power = Primary + Reflector + Shield Power 1

#s not important, just the parameters and three constituents of l i total power.

4

REFERENCE C.9

PM 2.4. p.5 ;

-End of Section C-j ..

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MITR-!!

October 1, 1985

0. INSTRUMENTS AND CONTROL (15.0)

Points Available '

00EST10N 0.1 What is the purpose of the AUTO TRANSFER ABORT switch in the (1.5) ,

reactor control room?

8 ANSWER 0.1 The AUTO TRANSFER ABORT switch in the reactor control room is used to eject a sample from t'he reactor, and cause it to exit into the reactor rabbit station, thus blocking its transfer to the NW-13 hot ,

l lab. The rabbit tube it controls (IPH1 or 2PH1) is determined by :

the position of tne AUTO TRANSFER SELECTOR switch at the r6(tit station. Also, in the case of IPH1. a sample which had been previously ejected and was being monitored at the stop pin could be exited into the station.

Full credit for answer li half-credit for 2 only.

REFERENCE 0.1 PM 1.10, p. 11 l

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MITR-l!

October 1 1985 Points Available QUESTION 0.2 If automatic reactor operation is desired, the power-set is adjusted to bring the power-setpoint deviation indication to zero at the desired power level. Why must the scale be adjusted on channel 59 (the automatic control channel) so that its signal is reading mid-range on the indicating meter? (2.0)

ANSWER 0.2 If this signal is at either the low or high end of the display meter, the automatic control will either not take control or be sluggish in its response.

REFERENCE 0.2 PM 2.3. p. 5 0UESTION 0.3 Small changes in po.er may be made through the automatic control system. This is done by slowly varying the setpoint of the power-set potentioreter and adjusting the scales of the other instruments as necessary.

What would happen if the operator moved the setting too rapidly? (1.5)

ANSWER 0.3 The deviation meter trip would be exceeded and reactor control would trip off automatic.

REFERENCE 0.3 FM 2.4. p. 4

-Section 0 Continued on Next Page-l 1

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.. MITR-!! !

1 October 1,1985 l

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Points :

Available !

00ESTION 0.4 1 !

If a 3 GV hole that contains a Nuclear instrument detector is flooded. what will happen to the detector output? Enolain why. (1.5) (

E I i ANSWER 0.4 i Output will decresse (+1.0) due to the increased attenuation of the ;

i neutrons (+0.5). i l

l REFERENCE 0.4 f

PM 5.4.11 l ,

J QUESTION 0.5 i

i The fatique cracking experiment alarm is actuated. Name two (2) i j f of the four (4) abnormal conditions which could cause such an alarrn. (2.0) i

[

1 l

ANSWER 0.5 1

l Two of the four needed.

a. A high sample temperature !

b. A very high sample ter.perature

c. A GM counter alare l

d. Low air pressure -

l j l i

REFERENCE 0.5 i

PM 5.7.9 l

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.f,6 ' MITR-!!

October 1. 1985 Points Available l

OVESTION 0,6

. Once the reactor-ready lamp is on, the regulating rod can i be moved to any position of travel. However, shim blade i withdrawal motion is limited to 4 inches by'the "sub-critical positien" interlock circuit. What ars th3 I

three (3) reasons for the sub-critical position interlock circuit? (1.5)

ANSWER 0.6 i

! 1, To maintain the shim blade bank programmed at a uniform height during

final approach to criticality.

2. To establish a level, below the critical position, to which the j shim blades may bo individually withdrawn in ore step.

j 3. To ')"ovide a convenient reference point at which the operator can 4

pause to make a complete instrument check before bringing tfie reactor to criticality.

i 1 REFERENCE 0.6 1

I RSM 4.3 J

j l OUESTION 0.7 L .

1_

T 0E or FALSE: Channel 9 (automatic control) operates on a gamma-sersitive cetector, not a compensated ion chamber. (0,5) 1 ANSWER 0.7

]

! True REFERENCE 0.7 RSM 5.9 _

1 1

-Section 0 Continued on Next Page-l

23

. . . MITR-Il October 1. 1985 Points Available QUESTION 0.8 There are two (2) primary conlant conductivity cellst MC-1 and 2.

W3 is MC-1 normally selected? (1.5)

-l ANSWER 0.8 Conductivity cell MC.1. which is positioned in a filter line at the i inlot to the ion exchance column,_is normally selected. The other cor . MC-2, is positioned in the outlet filter return line. Obvious i inlet measures highest and most conservative conductivity, unless e thly j ion exchanger is leaching out.

l REFERENCE D.S RSM 6.1 i

OVEST10N 0.9 i

Hy are flows in the reflector se-ondary coolat.t and shield coolant

taasured? (1.0)

-)

ANWER 0.9

]

Orifice plates and d/p cells.

I 1

R FER_ENCE 0.9 l RSM 6.6 1

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October 1, 1985 Points Available QUESTION D.10 Explain how the reading on the linear N-16 monitor would change as reactor power increases. (2.0)

ANSWER D.10 N-16productionisdirectlyproportionaltothefastneutronfkum, and therefore if the primary flow was constant, the radiation reading on this monitor would directly indicate reactor power.

REFERENCE 0.10 RSM 7.3

-End of Section D-

'o & &

MITR-Il 0'c'tober 1, 1985 E. SAFETY AND EMERGENCY SYSTEMS (15.0)

Points Available QUESTION E.1 What are the three (3) major safety requirements associated with operating MITR-II (according to the Standard Operating Plan General Instructions)? (3.0)

ANSWER E.1 The first, and most important, is that the release of radioactive materials to the environment be restricted to the lowest practical amount. The second safety requirement is thet on-site personnel be protected from contamination and that exposure to radiation be kept as low as is reasonably achievable. The third requirement is that equipment. especially the reactor itself, be operated and maintained properly and that nothing be done that would jeopardize future reactor operation.

REFERENCE E.1 FM 2.1. p. 1 0' JEST 10N E.2 Whv must the react'r be shut do.n if the compressed air system is To's t ?

(2.0)

ANSWER E.2 If neither compressor is capable of maintaining system pressure, the dump valve will open, the pneumatic instrum.entation will be lost and all airlock gaskets will deflate once the air within them leaks out past system check valves. You'll eventually lose containment integrity.

REFERENCE E.2 PM 5.5.4

-Section E continued on next page-t

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MITR-!!

October 1. 1985 Points Available I

00ESTION E.3 For each of the three (3) cases below, describe how emergency core cooling would be made available. -

(3.0) [

a. Assumptions: 1 Loss of normal electric power supply from Cambridge Electric Company.

2. All process systems are normal except for ;

the loss of power. .

b. Assumptions: 1 Level in the core tank cannot be maintained at the overflow level. but it has been f determined that it is not dropping below the f reactor inlet penetration (intet penetration l at -52 inches). -

c. Assumptions: 1. Level in the core tank cannot be maintained at the level of the reactor inlet penetration.

2. The lost water is being collected in the :

equipment room '. ump and/or a source of makeup l other than city water is immediately available.

ANSWER E.3

a. The system will be aligned as per normal shutdown cooling except that MM-2 will be sucplied power from the facility's emergency '

po.er supply and HE-2 will be cooled by city water,

b. The systems will be aligned as per modes 3 and 4. but these modes mill not be initiated until required. As long as the i conditions assumed for mode 2 prevail, natural circulation up !

through the core and do n through the flow shroud check valves l will suffice. Heat will be lost to ambient, the reflector tank. l and the off-gas system. (

L

c. >H-2 will be aligned to take a suction on either the equipment !

room sump through tte portable hose and strainer, or the other !

source of makeup, and discharged directly to the 8 inch reactor {

inlet line through MV-60 or through the spiay nozzles at the top :

of the core tank. I I

REFERENCE E.3 t RSM 3.4.5

-Section E continued on neat page- !

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e MITR-!! l October 1. 1985 i Points Available QUESTION E.4 i Wh it t two (2) mechanisms add negative reactivity to shut down the reactor when dump valve DV-4 is opened? (1.5) i ANSWER E.4

. When contents of reflector "dumps" to dump tank. negative !

reactivity added due to increased leakage (loss of reflector)

(+1.0). :

. There is a microswitch on the vale which provides a SCRAM when I the dump valve is opened (+0.5).

REFERENCE E.4 -

i l RSM 3.8 !

i i OVEST10N E.5 f i

l H2 long would the emergency batteries provide expected instrument ;

and pump power following a loss of both external electrical power feeders? (Select best answer.) (0,5) j L

l a. 40 minutes l

{ b. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months >

C. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months f

)] d. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

j ANSWER E.5 -

b. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months i REFERENCE E.5 ;

d r RSM 8.31 !

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. . . MITR-!! ,

October 1. 1985 t Points Available OVEST!0N E.6 i ;

Explain the difference between a major and minor SCRAM. (2.5)

ANSWER E.6 t

All automatic reactor scrams cause the current to the magnats ;

holding the shim blades to be interrupted. This causes the ;

, absorber sections to drop into the core and shut the reactor down. '

This action is defined as a minor scram. A major scram is initiated by depressing a major scram pushbutton. This action secures the ventilation system, seals the containment shell dumps the top part of the 030 reflector, and interrupts the withdraw permit ei_rcuit thereby dropping the shim blades.

(+0.5 for minor scram definition) i

(+2.0 for major scram four parts. +0.5 each) :

REFERENCE E.6 l 3

R$M 9.8 3

QUESTION E.7 There are eight safety and emergency related alarm conditions that

) will transmit a signal to the Campus Patrol Alarm System. Name five (5). (2.5) l 1

ANSWER E.7 :

Any five of below i i i High Temperature Reactor Outlet. MTS-1

. Low level Core Tank l

, . Low Pressure HM-1A

] . High Level Radiation Monitor ;

, . Smoke Detector Systen

! . Waste Tanks

. . Low Pressure Heltum Supply , ,

) . Leak Primary and 0 20 System l REFERENCE E.7 !

RSM 9.15

-End of Section E- l i

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' . MITR-!!

October 1. 1985 F. STANDARD AND EMERGENCY OPERATING PROCEDURES (13.5)

Points Available OVESTION F.1 Both shim blades and the regulating rod can be driven under automatic control provided the associated reactivity is less [

l than I delta k/k. (1.0)

ANSWER F.1 1.8% delta k/k l

REFERENCE F.1 !

Tech. Spec 3.9 (recent change) .

QUESTION F.2 What increase in reactor power requires the authorization and !

witnessing by the duty shift supervisor? (0.75)

ANSWER F.2

>10:

REFERENCE F.2 PM 1.3. p. 2 QUESTION F.3 -

List five (S) entries made in the Reactor Console Log for criticality data during a startup. (2.5)

ANSWER F.3

1. time ti. reactor power and period ,

iii, shim bank and regulating rod positions iv. cc,re outlet temperature

v. reflector outlet temperature REFERENCE F.3 PM 1.8, p. 2

-Section F continued on next page-

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MITR-!!

i October 1. 1985 Points Available OVESTION F.4 i What maximum pH value in primary system water requires imediate corrective action? (0.75) .

l ANSWER F.4, i 7.0 j REFLRENCE F.4 i PM 3.1.1.1 p. 12 ;

OVESTION F.5 t 1

j Wh t three (3) requirements must be met for the reactor to be in !

a ECURED CONDITION"? (3.0) l l

ANSWER F.5 i 1

i 1 The reactor is shutdown.

2. The console key switch is off with the key removed and in the i j troper custody. '

, r j 3. No work is in progress within the main core tank involving *

fuel or experiments, or maintenance of the core structure.

f installed control blades, or .nstalled control blade drives i when not visibly decoupled from the control blade. ;

l REFERENCE F.5

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October 1. 1985 Points Available OVESTION F.6 TRUE or FALSE: As defined in the MITR-!! startup checklists.

l the ECP ts actually not calculated for the infinite-period critical position, but for a supercritical position with a positive 50-second

] period. (0.5)

, ANSWER F.6 l True i REFERENCE F.6 ;

PM 2.3. p. 2 QUESTION F.7

.l What are the four (4) emergency classifications addressed in your emergency plan (PM 4.4)? (2.0) 4 ANSWER F.7

1. Unusual Event

2. Alert [

3. Site Area Emergency j 4 General Ernergency l 4

REFERENCE F.7 !

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October 1. 1985 Points !

Available ,

QUE5il0N F.8 i

The reactor is crd tical. You receive an alarm indicating that the primary coolant level has dropped 4.0" below the overflow i point. List your required immediate actions. (3.0) ;

t ANSWER F.8 *

1. Acknowledge the alarm. (+0.25) j

2. Scram the reactor (minor) if it has not already scrammed.

Verify that reactor power is decreasing. (+0.5) i

3. Notify the reactor shift supervisor. (+0.5) i 4 Check the core tank level indicators. ML-3A and ML-38 both '

to determine the actual coolant level and to decide if it is ,

dropping or remaining constant. (+0.25)

5. Prepare to initiate emergency cooling. Install the quick- !

connect hoses located in the control room and in the utility !

room between valves MV-69/MV-70 and city water lines. (+0.25)

6. Refer to Procedure 4.4.4.1 (Safety Limit Exceeded). (+0.5) 7 Notify the Assistant Reactor Superintendent, the Superintendent. I and the Director of Operations. If a safety limit was exceeded, i notify the Reactor Rad 1ation Protection Officer. (+0.20) ;

)

REFERENCE F.8 PM 4.4.4.4. p. 1

-End of Section F-l l

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October 1. 1985 G. RA0!ATION CONTR0t. AND SAFETY (14.0)

Points Available ;

OjlESTIONG.1 I

Wh A t action should the Operator-in-Charge take if the rabbit i radiation monitor trips? (1.0) ;

ANSWER G 1 i

4 Inform the shift supervisor (before investigation and resolution). .

REFERENCE G.1 ,

I PM 1.10 '

i !

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i OVEST!0N G.2

! i i

, Whatisthebasisofthemaximumirradiationtimedimitonthe rabbit (60-megawatt hours at a neutron flux of 10 )? (1.5) !

i

ANSWER G.2 i

j Embrittlement of the polyethylene containers. !

I

< REFERENCE G.2 i i

l P!1 1.10 p. 10 f t

I l

I OUESTION G.3 j

! There must be no direct contact with fingers on the irradiated container or samples be:ause oft (Select best antver.) (1.0) !

1 i a. high probable gama radiation !

5 b. high probable beta radiation [

j c. high probable surface contamination i 2

d. high probable alpha conta.aination f ANSWER G.3

\

l

b. beta l 4

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I REFERENCE G.3 i

PM 1.10. p. 10 i

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34

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M!TR-!! l October 1. 1985 !

i Points ;

Available j QUESTION G.4

! WM two (2) types of dosimetry are all personnel working at the MIT reactor required to wear? (2.0) f i i i

ANSWER G 4 !

i

1. hta-Gemma Monitoring ladge f j 2. Pocket Dosi n ter (gamma) !

REFERENCE G.4 PM 2.5. p. 1

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00ESTION G.5 W3 is a spill of heavy water a radiological concern? (1,0)

ANSWER G.5 Tritium content REFERENCE G.$

PM 4.5. p. 4 OVESTION G 6 If the containment building's ventilation system fails, what is the principal radioactive gas that will buildup in containment? (1,0)

ANSWER G.6 Ar-41 1

R FERENCE G.6 PH 4.0. p. 5

-Section G continued on next page-

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. MITR-!!

_0ctober 1, 1985 Points Available QUESTION G.7 TRUE or FALSE: When washing contaminated skin, it is important to use hot water to open and clean out potentially contaminated pores.

(0.5)

ANSWER G.7 False REFERENCE G.7 -

PM 4.4.4.10, p. 4 QUESTION G.8 Operation of the Blanket Test Facility (BTF) will cause certain radiation monitor detectors to read higher than normal. Which of the radiation monitors are most affected by use of the BTF (1.0)

ANSWER G.8 Secondary Water Monitors REFERENCE G.8 eM 5.6.2, p. 1 OUEST!0?,'O.9 Explain the difference in extent of qualification for blue, red, and yellow film badgos. Which badged group (s) are permitted to escort members of the general public through the Reactor Building? (1.5)

ANSWER G.9, Blue - beginning experimental work, mus.t be supervised Red - allowed to operate experimnt by themselves Yellow - sufficiently knowledgeable to escort public REFERENCE G.9 PM 1.12, p. 1

-Section G continued on next page-

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e MITR-!!

October 1, 1985 Points Available QUESTION G.10 List three (3) independent measurements or indicators used to monitor or detect heavy water leakage into.the secondary coolant. (2.0)

ANSWER G.10

1. Thesecondarywatermonitorisagamma-sensitivescintilljgion detector. It cannot detect tritium but is sensitive to N and F-17. tiso present in the heavy water when the reactor is operating.

2. Daily sampling of the secondary water will allcw detection of very small leaks.

3. Because of the nature of the reflector system, any loss of 0 20 inventory will be reflected by a decrease in the 02 0 level in the dump tank.

REFERENCE G.10 Tech Specs, p. 3-30 QUESTION G.11 Why is the Thermal Column Hohlraum maintained under a carbon dioxide purge? (1.0)

ANSWER G.11 To prevent activation of argon that would result if air entered the facility.

REFERENCE G.11 RSM 2.2 4

-Section G continued on next page-

37 MITR-Il October 1, 1985 Points Available QUESTION G.12 TRUE or FALSE: The purpose of the shield coolant system is to remove the heat deposited in the lead thermal shields by neutron radiation. (0.5) ,

ANSWER G.12 False (gamma)

REFERENCE G.12 RSM 3.13 -

End of Section G-4 e

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. R3 / Efr U. S. NUCLEAR REGULATORY COMMISSION SENIOR REACTOR OPERATOR LICENSE EXAMINATION -- -

FACILITY: MASS. INSTITUTE OF TECH.

REACTOR TYPE: TEST DATE ADMINISTERED: 85/10/02 EXAMINER: SILK, D.

APPLICANT: _M.!%dC.lify..____.....__

INSTRUCTIONS TO APPLICANT:

Use separate paper for the answers. Write answers on one side only.

Staple question sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing Stade requires at least 70% in each category and a final grade of *t least 80%. E;: amination paper s will be picked up six (6)

hours after the examination starts.

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY

.I'O 00_I.___ _20_.I._00 .._________ ________ H. REACTOR THEORY 20 0 20 0

...I.0... __.I.0 _ ______...__ ________ I. RADIDACTIVE'HATERIALS HANDLING DISPOSAL AND HAZARDS 20 20

...I_00___ _..I_00 _ ......____. ________ J. SPECIFIC OPERATING CHARACTERISTICS 20.00 20 00

_______. ....__ __..______. ________ K. FUEL HI'DLING AND CORE PARAMETERS I

20.00 20.00 ADMINISTRATIVE PROCEDURES, l

..__.___ ______ __..___..__ -_______ L.

l CONDITIONS AND LIMITATIONS 200.00 100.00 TOTALS l FINAL GRADE ________________ %

l i All work done on this examination is my own. I have neither Siven nor received aid. ,

i -

APPLICANT'S SIGNATURE l

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. H. REACTOR THEORY PAGE O QUESTION H.01 (3.00)

How much reactivity has been added to a svberatical reactor if the count rate has increased from 10^ eps to 150 eps and if the initial value of Heff was .95? Show all calculations and assumptions.

QUESTION H.02 (3.00)

If heavy water were siixed with light water cooling the core:

a. Nov1d the neutron lifetime increasee decrease, or remain the same? (0.7)

b. Would the migration length increase, decrease, or remain the same? (0.7)

c. What is the overall reactivity effect? Explain. (1.6)

QUESTION H.03 (3.00)

Explain the different modes of heat transfer by which the heat of fission is removed fron, the fuel. Include major components involved in the heat removal process starting with the fuel and ending at the ultimate heat sink. (3.0)

DUESTION H.04 (1.00)

Why are delayed neutrons important?

OUESTION H.05 (3.00)

Explain the effect of the temperature coefficient on reactivity if the thermal power of the MITR II core increases. Include both light and heavy water effects.

(zazza CATEGORY H CONTINUED ON NEXT PAGE xxxxx)

2. .

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1 H. REACTOR THEORY PAGE 3 i i

QUESTION H.06 (3.00)

The reactor operator is condveting a r outiree r eactor startup after it has beers shutdowr, for several days. Prior to withdraw 1 rig a sham blade he re. ads a stable count of 50 eps on the startup channel.

Immediately after withdrawing this blade he reads a courit of 80 eps.

a. If he performed no blade motion for five minutes, would the c ourit rate i rie r e a s e , decrease or remain the same? E >: p l a i r ,

assuming the reactor is soberitical at 80 eps.

b. After 5 mirevtes he withdr aws ariother blade the same distarice but the reactor is still suberitical. Would the change in count rate (time arid ma3rii tude ) be different there he saw in part (a) above? Explain.

c. What indicatiores would the operator observe to determine when the reactor had g orie c r i t i c a l ?

QUESTION H.07 (4 00)

Xenon and Samarius, are two poisons which have a sigreificant ef fect ora reactor op e r a t i o ris . Discuss arid compar e these two poisores f or the f ollowires :

a. Sovrees of the poisores in the core (1.0)

6. He ares of removal from the core (1.0)

c. Effect o ra reactor operations after shutdown (2.0) gxxxxx END OF CATEGORY H m2xxx)

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- I. RACI0 ACTIVE NATERi .S HANDLING DISPOSAL AND H4 CAR

.,___________________________________________________DS __

PAGE 4 '

GUESTION I.01 (4.00)

A 23 year old i r.c i v i d u a l has accumulated a lifetime occupational dose of 24 rem of whole body exposure d o e v aie rit e d in accordance with 10CFR20 and has,,r.eceived no exposure dur ing the preseret calender quarter.

a.

How long may he work t re a 3 meen/hr area if he works an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months day Monday through Friday? Show your work.

b.

An individual iri a restricted area may be allowed to receive a whole body cose in e:: cess of the quarterly limit under certain conditions.

Name three conditions.

QUESTION I.02 (2.00)

A mixed gamsia and beta source in liquid form spills on the floor. Readings at to feet indicate 1.0 a. rem /hr o re a beta samma survey meter. If beta's are not detected further than six feet from the spill and if the combined beta samma dose rate at one foot is 120 mees /hr, what is the beta to samma ratio? Show your c a l c u l a t i o re s .

QUESTION I.03 (3.00)

c. Does the biological effect of a 100 REM dose depend on whether it is a rieutr on or gamma dose? Explain.

b. Does the biological effect resulting from bodily intake of a given quantity (in terms of microcuries) of a radioactive material depend ori which partievlar isotope is involved?

Explain.

(maxxn CATEGORY I CONTINUED ON NEXT PAGE xxxxx) 1

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I. RADIDACTIVE NATERIAto HANDLING OISPOSAL AND HAZr.RDS PAGE 5

. ___w___________________________________________________

DUESTION I.04 (3.00)

A fuel element i s suspended in the Reactor Pool approximately 1 meter under water. A radiation survey meter held at the surface of the wate.r reads 100 arem/hr.

s. IgnorinS buildup, what radiation level would you expect if the fuel element broke the water? Assume an attenuation coefficient of 0.035 cm^-1. (1.0)

b. If the radioactive '.sotopes in the fuel element had an average half life of 30 minutes, how long would it take for the radiation level at the surface of a one inch lead shield cask to drop to 20 mren/hr?

Assume an initia1' contact dose of 2 R/hr for the fuel element and a tenth thickness of two inches for lead. (2.0)

DUESTION I.05 (3.00)

To assure that experiments in the reactor do not affect the safety of the reactor, Technical Specifications demand that all experiments within the reactor shall confirm to a set a conditions. List six of the seven cond-itions set forth in the Technical Specifications.

QUESTION I.06 (3.00)

For the case of a radiological emersency, list seven immediate actions that the on-shift supervisor must ensure have been completed. (Assume no medical assistance and no radiation surveys by Campus Police are required).

QUESTION I.07 (2.00)

Does the number of disintegrations per minute (dpa) from a radioactive source equal the counts per minute (epm) obtained from a survey instru-ment? Briefly explain.

i (xxxas END OF CATEGORY I xxxxx)

G W m 6 e e

-7 _

J. SPECIFIC OPERATINd HARACTERISTICS PAGE 6 QUESTION J.01 (3 00)

What three actions must be taken when 1 microcurte/ liter of tritive. is present in the secorida r y coolant water?

QUESTION J.02 (3.00)

a. If the Reactor Floor Ar-41 Monitor gives an 'High Level Radiation Non-itor' alarm, where are five likely plcees for the Ar-41 to originate? .

(2.0)

b. What is done to prevent the preduction of Ar-41? (1 0)

GUESTION J.03 (2.00)

Briefly describe the natural convection valves, how they work, and what is their function?

QUESTION J.04 (3.00)

What does the 'suberitical position' interlock cirevit do and give three reasons why it is incorporated into the shim blade control cirevit.

QUESTION J.05 (3.00)

Figure 1 shows the differential regulating rod worth curve for your reactor. Give two reasons why the curve peaks at the location shown.

QUESTION J.06 (3 00)

a. Driefly explain why the reactivity worth of the 020 Reflector Duap is dependent on the position of the shim blade bank.

b. What is the required position of the shim bank when the radial heavy water reflector is pumped into place? Briefly expla).n why.

(zmxxx CATEGORY J CONTINUED ON NEXT PACE musum) a

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e J. SF ECIFIC OF ERATING HARACTERISTICS 7

, --_...____..___'___..__.........______. F' A G E DUESTION J.07 (3.00)

E:r t e f ly e::pl a t ri the n.os t r eliable method of deter s. trit res the steady state power at full F.ower arid wner this niethod cari be used. ,

(xxxxx END OF CATEGORY J xxxxx)

a

. H. FUEL HANDLING AND .JRE F PAGE 8

____________________________'ARAMETERS _________

GUESTION I:.01 (3.00)

After each refueling or chan3e in core loadinge the reactor shall not be operated above a power level of 1.0 KW unless an evaluation is made to ensure that two Technical Specifications are satisfied,

a. What are the two Technical Specifications? (2.0)

b. What persons shall complete and approve these evaluations? (1 0)

DUESTION K.02 (3.00)

Give the basis for the following specifications:. ,

s. The reactivity worth of the regulating rod connected to the auto-matic control system is less than 0 7% delta k/k.

b. The maximum controlled reactivity addition rate is no more than 5x10^-4 delta k/k /sec. .

c. The reactivity worth of the D20 reflector dump is greater than the reactivity worth of the most reactive shim blade.

QUESTION K.03 (4.00)

During refueling, what are two designed safety features associated with the hold-down grid plate and what do they prevent?

QUESTION K.04 (3.00)

a. Under what condition, during refuelings is the heavy water reflector not dumped? (2.0)

b. What Technical Specification requirement must be checked if the heavy vater reflector is not dumped? (1 0)

DUESTION K.05 (2.00)

What two Technical Specifications requirements must be met before approval is given to remove the spent fuel from the reactor vessel to the transfer flask?

(mzans CATEGORY H CONTINUED ON NEXT PAGE suzza) l

=

l

K.

o FUELHANDLINGA()COREPARAMETERS FACE

.. ~*

QUESTION K.06 (3.00)

According to your Technical Specifications what safety channels a.ust be operable to s.ove fuel in the core and what are the set points, if any?

QUESTION K.07 (2.00)

According to your Technical Specificationse when is your reactor considered secured?

/

l I

l (m**ma END OF CATEGORY K xxxxx) k

'L. ADMINISTRATIVE FR .EDURES, CONDITIONS AND LIhsTATIONS PAGr GUESTION L.01 (2.50)

In accordance with your Administratiore procedures:

a. Ortefly desertbe the a d m i ri t s t r a t i v e procedures followed if a safety f unc t i ore required by Technical Specifications as a Limiting Condition for Operation is to be temporarily bypassed (assume it is not a part of an approved procedure). Include in your answer who may authori e the bypasse conditiore of the reactor and r ecor ding r equir emerits. (1 5)

b. What additional requirements are necessary if a jumper is used? (1.0)

GUESTION L.02 (4.50)

Indicate whether or not each of the following is a violation of procedures and/or Techr.ical Specifications. Briefly explain why it is or it is not a violation.

a. Operating with five shisi blades, the sixth shim blade is fully inserted

b. Operatin3 at 2 MW with oree primary pump and 1000 spa primary coolant flow rate

c. Operating at 150 MW with the emergency cooling system inoperable

d. Operating at 100 MW without emergency power available

e. Operating at full power with one of the three reactor floor area r adi a tiori morii tor s ir operative

f. Increasing the reactor power fror. 200 KW to 300 MW with the duty shift supervisor t re the Utilities Room.

(0.75 each)

GUESTION L.03 (3.00)

Any change to a component or system which involves an 'unreviewed safety q u e s t i e rs ' is a ' Class A' prop'esal. A proposal change 'shall be deemed to involve an unreviewed safety question' if what three criteria are met?

(musur CATEGORY L CONTINUED ON NEXT PAGE xxxxx) l i

~

l l

e i

~

o L. 00MINISTRATIVE I .CEDURESe CONDITIONS AND L. ITATIONS PAGE !!

QUESTION L.04 (1.50)

List five of the services that the Reactor Radiattor. F r o t e c t i o r. O f f i c e as responsible for providing for radiattore p r o t e c t i ore ared compli anc e with g ov e r nm ere t a l regulations.

00ESTION L.05 (1 00)

Under what conditions may soms one be authori=ed to incur radiation exposures in excess of the to CFR 20 limits?

QUESTION L.06 (3.00) -

In regards to General Safety Rules, once permission is granted, what are three joint responsibilities of the operator-in-charge and the personnel entering either the reactor tope the medical therapy roome or the equip-ment roos, when the reactor is operatin3?

OUESTION L.07 (2 00) a, What are four variables associated with the core thermal and hydravile pe r f or m a rec e ?

b. What i s the objective of the Safety Limits?

QUESTION L.08 (2.50)

a. Given the events below; state which emergency classification should be declared. (0 5 pts each)

1. A large crowd of protesters marching around the reactor building.

2. A fire damagsres an experineret which causes the release of radio-active materirls.

3. A tornado damaging the contairement building.

4. A slow and vr controllable decrease in core tank level such that level remains above the anti-syphon valves.

b. What criteria is used for cla ssi f ying eme r gency coredit t oris ? (0.5)

(zazza END OF CATEGORY L manas) 1 (musussaransas END OF EXAMINATION masasazzuxxazza)

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Table 4.5.3-1: If ts for ?'otif icat ion nf linusual T. vents

1. Confirmed abnornal radiation levels leading to actuai or projected radiological effluents at the site boundary exceeding 10 M?C for unrestricted areas when averaged over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This level corresponds to an ervosure of 15 mren whole body accumulat ed over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (P)' 4.4.4.3 5h)

2. Report or observation that .evere natural obenomena are either f erinent or ex-isting. These include scorns with tornado or hurricane force winds that could strike the facility, earthquakes that could adversely affect the reactor's safety systems, and floods that could adversely af fect the reactor's safety systecs. (PF 4.4.4.2)

3. Threats to or breaches of security. (Pit 4 4.4.5/4.4.4.6) 4 A reactor safety linit's being exceeded such that a fuel danage accident that ceuld release radionuclides to the containment building is possible.

(PM 4.4.4.1)

5. A fire within the containment building that lasts beyond the incipient stage or for more thin ten rinutes. (?F 4.4.4.3)

6. Receipt of a bonb t hr ea t . (P!! 4.4.4.7)

SRW-0-82-19 AUG 6 1982

_- ___-. . - . _. - . _-. _ _ _ = - _ _ .

. Table 4.5.3-2 EALs for an Alert . ;

i

1. Confirmed abnornal radiation levels leading to actual or projected radiologica) effluents at the site beundary eveeeding Sn MPC for unrestricted areas when sveraged over 24 Fours. This level corresponds to an exeosure of 75 nren whole body eccumulated over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (PM 4.4.4.15b)

2. Same as #1 except the effluents could cause an integrated exposure of 100 mrer thyroid. (PM 4.4.4.15b)

3. Radiation levels at the site boundary of 20 arem/ hour sustained for one hour.

(PM 4.4.4.14b/4.4.4.11)

4. Abnormal loss of prinary coolant such that the core tank level remains at or above he anti-syphon valves. (PP 4.4.4.4)

5. Loss of radioactive naterial control that causes radiation dose rates or air-borne radionuclides to increase above permissible exposure levels by a factor of 1000 throughout the containrent building. (P.* 4.4.4.12)

6. Radiation dose rates throughout the containment building in excess of 100 ares /

hour sustained for one hour. There levels would necessitate evacuation of all personnel. (PH 4.4.4.12)

7. A fire leading to loss of radioactive material control within the containment building. (PM 4.4.4.3)

8. An ' imminent or existing hazard such ast (a) Missile (s) iepacting on the containment building.

(b) An explosion that a ffects f acility operation.

(c) An uncontrolled release of toxic or flammable gases into the containment bu ilding. (PM 4.4.4.9)

SRJ-0-82-19 AUG 6 1982

{-

O Table 4.5.1 1: _TA?.s fer a Site Area Feerrency 4 sp

1. Cenfirmed abnornal raillation levels leadinr to actual or ornjected redintori-I cal effluents at the si:e konndarv exceedinr. 250 FPr for unrestricted areas when averaced over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This level corresnonds to an exposure of 375 area whole body accunulated over 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . (PP 4.4.4.15b)

2. Sane as #1 except the effluents could cause an integrated exposure of 500 l

mrem thyroid. (PP 4.4.4.15h)

3. Radiation levels at the site boundary of 100 nree/ hour sustained for one hour.

(PM 4.4.4.14b/4.4.4.11) 4 Abnormal loss of primary coolant such that the core tank level drops below the anti-syphon valves. (Note: This accident is not considered credible, but procedures exist for cooing with it.) (PP 4.4.4.4)

5. Inminent loss of physical control of the reactor. (PF 4.4.4.6)

6. Severe natural events Feint. experienced. These include:

(a) An earthquake that is causing observable damage to the reactor safety equipment within the containrent building.

(b) A flood that is affecting the ooerability of any reactor safety system.

(c) Tornado or hurricane force vinds that are damaging the containment building. (PM 4.4.4. 2) l l

l l

SRf-0-82-19 AUG 6 1982

, ts. AJ ct Tabl e 4. 5.3-4 : EALs for a General Emerrency -

1. Actual or projected doses at the sit e bounda ry in the exposure pathway of 1 rem whole body or 5 ren thyroid for unrestricted areas when averar.ed over one hour.

Note: Pigure 4.7.2.2-1 lists the conditions and instrument readings correscend-ing to a projected of f-site dose of 1 ree/ hour. (PF 4.4.4.15a)

2. Sustained actual or projected radiation levels at the site boundary of 500 nrer/

hour whole body. (PM 4.4.4.14 a/4. 4.4.11/4.4.4.12)

3. Blockage of fuel element channels thereby causing a loss of coolant to the af f ec ted channels and a f uel relt. This is the design basis accident.

(PH 4.4.4.15a) 4 Loss of physical control of eitber the containment building which includes the control roon or of auxiliary areas that house vital equipment. (PF 4.4.4.5/

4.4.4.6).

5. Events that have caused or will cause massive facility and/or reactor system damage that could lead to the neiting of fuel. (PP 4.4.4.15r) 6 se SRf-0-82-19 AUG 6 1982

~

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f o ca v s/t c"cle efficiency e (Networt

~

w t)/(Energy in) 2 w = og s = V,t + 1/2 at 2

E=x .

KE = 1/2 av a = (Vf - V,)/t A = AN A= '.ne ***

PE = agn yf = y, + a t *

e/t A = an2/t1/2 = 0.693/t1/2 g ,y, t 1/2##*EI*1M}I*b)3

((t1/2) * (*b)3 '

aE = 931 am

4*

t=ge

. . O Q = aCpat d = UAat c e""*

! = I pwr . w ah 7 ! = I,10** N TYL = 1.3/u p = p 10 sur(t) HYL = -0.693/u P = Po e*/ .

SUR o 26.06/T SCR = S/(1 - K,ff)

CR x =S/(1-Keffx)

SUR = 26o/t= + (s - o)T CRj (1 - K,ff)) = CR2 (1 - k,f.f 2) '

T = ( i*/s ) + ((8 - o )/ a ] M = 1/(1 - X,ff) = CR j/CR, T = 1/(o - s) M = (1 - K,ffa)/(1 - K,ff))

T = (s - o)/(lo) SDM = (1 - K,ff)/K,f, o = (X ,ff-1)/Kaff * '#eff/K eff 1* = 10-5 seconds T = 0.1 seconds" e u ((t*/(T K,ff)] + (i,ff /(1 + [T)]

Id li*Id Id 2 ,2 gd 2

P = (14V)/(3 x 1010) jj 22

2 1 = eN '

R/hr = (0.5 CE)/d (meters)

R/hr = 6 CE/d2 (feet)

Water Parameters Miscellaneous Conversions 1 gal. = 8.345 lbs. I curie = 3.7 x 1010gp, I

i gaj. = 3.78 liters 1 kg = 2.21 lbm 1 ft* = 7.48 gal. I hp~= 2.54 x 10 3 Btu /hr Density = 62.4 lbe/ft3 1 av = 3.41 x 106 Btu /hr i

'Oensity = 1 gm/cm3 , lin = 2.54 cm l Heat of vaportration = 970 Btu /lem 'F = 9/5'c + 32 l Heat of fusion = 144 Stu/lbe *C = 5/9 (*F-32) 1 Ata = 14.7 psi = 29.9 in. Hg. 1 BTU = 778 ft-lbf l 1 ft. H 2O = 0.4335 lbf/in.2 l _ - _ . _ _ _ _ _ _ _ _ ___ _____ __ _ -_ _ _ _ _ _ _ _ _ _

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ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILK, D.~'

ANSWER H.01 (3.00) erl /cr2 = (1-Heff2) / (1-Heff1) CO.93 100/150 = (1-Heff2) / ( 1 -- 0 . 9 5 ) [0 53 1-Heff2 = 10/15 x 0.05 Heff2 = 0.967 CO.13 Change in rcactivity = C1-Heff2/Keff23 - C1-Keff1/Heff13

= Heff2 - Keffi / Keff1 x Keff2 C0.93

= 0.947 - 0.95 / 0.95 x 0.967 C0.53

= 1.85 % delta W/H [0 13 REFERENCE Procedure Hanval (FM) 2.3 pg. 1e2 ANSWER H.02 (3.00)

a. Increase (0.7)

b. Increase (0 7)

c. The increased seigration length would tend to increase neutron lifetime and leakage and thus add negative reactivity. (1.6)

REFERENCE Reactor Systes.s Hanval (RSM) pg. 10.10 ANSWER H.03 (3.00)

Condvetion through fuel.

Condvetion transfer f r o n. fuel to coolant.

Forced convection to heat exchanger.

Condvetion across heat exchanger.

Forced convection to cooling towers.

Evaporation to atmosphere. (0 5 pts each)

REFERENCE Introdvetion to Nuclear Engineering, chapter 88 J R Las.arsh RSM pgs. 31, 3.7, 3.10 to 3.12

H. REACTOR fHEORY 0 .._______.-____..- PAGE 13

' ANSWERS -- NASS. INSTITUTE OF TECH. -85/10/02-SILHe D.

ANSWER H.04 (1.00)

The delayed neutrons increase generation time which increases the period and thus the reactor can be controlled.

REFERENCE Introduction to Nuclear Engineering. chapter 7 pg. 2453 J R Las.ersh ANSWER H.05 (3.00)

Increasing the temperature of the light water will insert negative reactiv-ity by causing the neutrons to take longer to thermali:e so there are fewer fissions (1.5). Heating of the heavy water reflector will add nega-tive reactivity by allowing neutron leakage to increase (1.5).

REFERENCE RSM P3 10.8 ANSWER H.06 (3.00) a.

Increase (0.4).

slightly then level out(0 6) due to suberitical avitiplication 6.

Larger increase (0.3) and longer so level out(0.3) due to greater number.

of generations to reach equilibrium (0.4).

c. Steadily increasing count rate or slight positive period with no rod withdrawal. (1 0)

REFERENCE PM 2 3 pg. 12 l

l

. H. REACTOR THEORY PAGE 14 ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILK, 0. ~'

ANSWER H.07 (4.00)

o. Both are produced directly fron. fission and fros. their respective decay chain. Te-135 decays to I-135 which decays to Xe-135. Nd-149

<ecays to Psi-149 which decays to Sm-149. (1.0)

b. Both can be removed from the core by neutron absorption. Xe-135 can also be removed by radioactive decay, whereas Ssi-149 is stable. (1.0)

c. When the reactor is shutdown, both poisons increase in concentration due to production from their decay chains and because neither are being removed by neutron absorption. Sm-149's increase is relatively small and reaches a maximum and remains there until the reactor is operated again. Xe-135 wi?1 increase to a peak and then decrease slowly as more Xe-135 is decaying than is being prodveed by the decay of I-135.

(2 0)

REFERENCE RSH pg. 10.6 to 10.8

,g- , - _ - - - - . - -

1.__ . . ___ ... . . )b. O__j AG 15

'ONSWERS -- N ASS . INSTITUTE OF TECH. -85/10/02-SILKe D. -

ANSWER I.01 (4.00)

o. 5(N-18) = 5(23-10) = 25 25 - 24 = 1.0 Rem = Hau. Dose (1.0)

Max. Dose = Dose Rate X Time 1.00 Rem = 0.003 Rem /hr X 8 hr/ day X No. of Days No. of Days = 41 6 days (1.0)

b. Provided that (1) He does not exceed 3 rem per quarter (.66)

(2) His radiation history is known and recorded on the proper form (NRC "orm 4) (.67)

(3) The dose received when added to his radiation history does not exceed 5(N-18) reas where N= the person's age at his last birthday (.67)

REFERENCE 10 CFR 20.101 ANSWER I.02 (2.00) dx (r)*2 = 0 x (R)^2 1 mr/hr x (10)^2 = D n (1)^2 0 =100 ar/hr C1.03 Leta dose = 120mr/hr -100 mr/hr

= 20 er/hr l Beta to samma ratio = 20/100 = 1/5 C1 03 REFERENCE Introduction to Nvelear Engineering, chapter 9 pg 409,410; J R Lamarsh F

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1. RADI0 ACTIVE MATEL .LS HANDLINC DISPOSAL AND .aZARDS PAGE 16 ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILKe D.

ANSWER I.03 (3.00)

c. No CO.53 A R e si dose aceovnts for the type and energy of radiation. C1 03

b. Yes CO.53 Internal dose depends on biolo3ical and physical T 1/2, referred or3an, type of radiation. [1.03 REFERENCE Introduction to Nuclear Engineering, chapter 97 J R Lamarsh ANSWER I.04 (3.00)

a. I = Io e<-ux) 100 aresi/hr = Io e<-0 035 cm(-1) 100 cn>

Io = 3311 a r e s. .ca

b. I = Io 10 {-x/TVL) (TVL te - 4

. u Io = I 10 <x/TVL) = 63.25 s.ren/hr Io = Il e<-(.693/ half life)t> (wher- la 1 ,. e ,

e) t = -(half life /.693) In(Io/Ii) t = -(30 min /.693) In(63.25/2000) = 149.5 minutes ,(3 REFERENCE Introduction to Nvelaar Engineeringe pas 22, 838 J R L- st ANSWER I.05 (3.00)

Reactivity Effects Thermal-Hydraulic Effects Chemical Effects Radiolytic Decoa. position Experiment Scram prototype Testin3 Radioactive Release (0 0.5 pts, any 1x i 3.0)

REFERENCE Technical Specifications (T.S.) 6.1, ps. 6-1 to 6-7

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. I. RADI0 ACTIVE MATER. .S HANDLING DISPOSAL AND hn/ARDS PAGE 17 ANSWERS .- NASS. INSTITUTE OF TECH. -85/10/02-SILK, D.

ANSWER I.06 (3.00)

a. The reactor is shut dowr.

b. The containment building is isolated

e. Experimenters are evacuated

d. Off-duty licensed and radiation protection personnel are notified

e. The MIT Campus Police are requested to stand-by

f. Radiation levels are monitored on-site and tracked off-site using

] '

g.

the MITR Radiation Protection Office's remote monitors Off-duty personnel are briefed as they arrive REFERENCE l Procedure Manual (PM) 4.3 pg 3 ANSWER I.07 (?.005 No. The cps must be correctec for efficiency of the detector and the geos.etry of the source in esistion te the detector.

REFERENCE RSM pgs 5.2, 7.1 r

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J. SPECIFIC OFERATIN. CHARACTERISTICS PAGE 18 CNSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILMe 0.

ANSWER J.01 (3.00)

The coolirig tower spray shall be shut dowre The secondary system water discharge shall be stopped The 020 reflector heat enchanger shall be isolated REFERENCE T.S. 3.8, pg. 3-26 ANSWER J.02 (3.00)

a. High f1va regions such as the thermal colvene pipe tunnele lid space, e xp e r i n e re t a l port and ires trumerit lead boxes. (2 0)

b. The high flun regions are sealed and/or flooded with carbon dioxide in order to enelude as much air as possible sir.ce Ar-40 is present in air. (1.0)

REFERENCE RSM pg. 7.5 ANSWER J.03 (2.00)

Natural convection valves are ball type pressure-operated check valves located ore the wall between the inlet and outlet of the core that are designed to open on a loss of primary pump pressure to allow natural convective flow a r ov rid the core.

REFERENCE SAR pg. 6.5 ar d 15 12 l

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'J. SPECIFIC OPER ATIlk OHAR ACTERISTICS

, ...................................... PAGE 19 CNSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILK, D.

cNSWER J.04 (3.00)

It limits shim blade withdrawel motion to four inches.

1. It ha"intains shim blade bank programmed at a uniform height during final approach to criticality.

2. It establishes a level, below the critical position, to which the shim bindes may be individually withdrawn in one step.

3. It provides a convenier.t reference point at which the operator can pause to make a complete instrument check before bringing the reactor to criticality

(.75 pts each)

REFERENCE RSM pg. 4.3 ANSWER J.05 (3.00)

The peak in the differential regulating rod worth occurs at low rod height bceause the full in position for the regulating rod is six inches above the bottom of the fuel elements and once the regulating rod is withdrawn any appreciable amount, it is heavily sh&dowed by the adjacer.t shim blades.

REFERENCE RSM pg. 10 6 ANSWER J.06 (3.00)

a. The blade bank exerts a shadowing influence on the reflector

b. Full in - this insures that the reactivity inset tiore for this process will not occur when the reactor is or could go critical REFERENCE RSM pg. 10.6 O

b.

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Jo SF'ECIFIC OFERATINL CHARACTER 3STICS

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p. A C E 20 ,

ANSWERS -- MASS. INSTITUTE OF TECH. -

85/10/02-SILKe D.

ANSWER J.07 (3.00)

A heat balance calculated froe the prisiarye reflectore and shield system flows and temperature rises once these systems are i n thermal equilibrium.

REFERENCE PM 2 4, pg. 2 i

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M. FUEL HANDLING AND CORE PARAMETERS 8' AGE 21 ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILM, D.

ANSWER K.01 (3.00)

a. The ratio F hc Fp /d F,g 4

is predicted to be less than 2.9 The core is predicted to operate below iricipient boiling at every point in the core.

b. Two Sersior Reactor Operators.

REFERENCE T.S. 3.le pg. 3-1 ANSWER K.02 (3.00)

a. The total vorth of the rod is tu be limited such that the complete withdrawal of the rod will not make the reactor prompt critical

b. This value is conservatively within the range of reactivity insertion rates normally accepted for reactor operation. Control systems in this range give asiple margin for proper human response during l approach to critical and power operations.

c. The additional indepetident capability for reactivity control provided by the 020 reflector dump gives added assurance that the reactor care be made soberitical under a re adverse condition of fuel loading or coritr ol bl ade m a l f ur.e t i or , .

REFERENCE T.S. 3.9. pg. 3-32 to 3-35 ANSWER H.03 (4.00) l 1. Tne grid is designed so that there is normally access to only one core position at a time (1.0). This Itaits the asiount of water that can be in the core at any one time by making it difficult, though not j impossible, for more than one core position to be defueled at time.(1 0) 4

2. The grid's latch is interlocked with the primary coolant pumps so that if the latch is releasede the coolant pumps stop and remain off until the grid is latched aga:,n (1.0). This protects the fuel elements from damage and the reactor as a whole from inadvertent criticality (1 0)

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K. FUEL HANDLING AND CORE PARAMETERS PAGE 22 ANSWERS -- NASS. INSTITUTE OF TECH. -85/10/02-SILK, O.

REFERENCE PM 2.7, ps. 3 ANSWER K.04 (3.00)

c. If dumping would cause the nuclear instrumentation startup channels to indicate less than 10 counts per minute. (2 0)

b. The shutdown margin would have to be checked. (1 0)

REFERENCE PM 2.7, ps. 3 ANSWER K.05 (2.00)

, 1. The element to be moved cannot be moved unless it has not been operated in the core at a power level above 100 MW for at least four days.

2. The K-effective of any storage area outside of the reactor core shall be less than 0.90 REFERENCE T.S. 3 10.4e pg. 3-37 ANSWER K.06 (3.00)

Safety channels operable Set points Period (2 channels) > 3 see Neutron Flu:< Level (2 channels) 100 kW D20 Dump Valve Selector Switch (1) -

Manual s.ajor scram (2) -

(0.5 each response) t

REFERENCE i T.S. 3.7.2, pg. 3-21,22 1

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e K. FUEL HANDLING AND CORE PARAMETERS PAGE 23 CNSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILK, D.

ANSWER g,07 (7,gg, P

a. The reactor is shutdown (.66)

b. Console key switch off and key is in proper custody (.67)

c. No work in progress within the main core tank involving fuel or experimentse or o.aintenance of the core structures i nstalled control blades or installed control blade drives when not visibly decoupled from the control blade (.67)

REFERENCE T.S. 1 1, P3 1-1 .

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o L. ADMINISTRATIVE PROCEDUFESe CONDITIONS AND LIMITATIONS PAGE 24 ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILK, D.

CNSWER L.01 (2 50)

a. 1. The reactor must be shutdown and the bypass must be removed before reactor startup (0.5)

11. Must be approved by Duty-Shift-Supervisor or Reactor Superintend-ent (0.5) 111. The bypass authori:or's initials must be recorded on the bypass los sheet (0.5)

b. If Jumpers are usede the jumper must be taggedl a warning te3 placed on the shim blade control handle stating that the reactea is not to be started up until the bypass is removed. (1.0)

REFERENCE PM 1.9e pg. 1 CNSWER L.02 (4.50)

o. Violation of Technical Specifications (T.S.)(.25)e sixth shis blade must be at the operating position or higher (except if < 1 KW for blade calibration) (0.5)

6. No violation (.25), with one pump 3.0 MW allowed and minimum of 900 9Fm '

(0.5)

c. Violatiori of T.S. (.25), power levels in excess of 100 MW require the emergency cooling system to be operable (0.5)

d. Violation of T.S. (.25), emergency power must be available whenever the i l reactor is operating (0.5) l o. No violation (.25). T.S. requires at least one area radition monitor on the reactor floor to be operating (0 5)

f. Violation of procedure (.25), the duty shift supervisor must authori:e and witness both startups and ircreases i re reactor power of greater r than 10% (0.5)

REFERENCE

a. T.S. pg. 3-32

b. T.S. pg. 2-5

c. T.S. pg. 3-19

d. T.S. pg. 3-21 i

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ADMINISTRATIVE F ROCEDURES. CONDITIONS AND LIMITATIONS PAGE 25 CNSWERS -- NASS. INSTITUTE OF TECH. -65/10/02-SILM, D.

o. T.S. ps. 3-27

f. PM 1 3, pg. 2 ANSWER L.03 (3 00)

1. If the orobability of ocevrence of the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be increased.

2. Ifthan a possibility for an accident or aalfunction of a different type sy evalusted previously i n the safety analysis report may be created.

3. Ifspecification the marSin is of safety reduced.

as defined in the basis for any technical REFERENCE PM 1 4e pg.2 CNSWER L.04 (1.50)

o. Registration and instruction of radiation workers

b. Personnel menatoring of radiation exposure

c. Radioisotope laboratory inspectionse radiation surveys, and area monitoring

d. Radioactive waste collection

o. Calibration and repair of radiation protection instruments f.

Calibration of reactor radiation detection instruments 3 Environmental monitoring

h. Leak-testing of sealed radioactive sources .

i. Advice in radiation emergenciese and special decontamination operations J. Maintenance of radiation protection records (any five, .30 pts each)

REFERENCE PM 1.11, pg. 1 m

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a- .

e' L. ADMINISTRATIVE PROCEDURESe CONDITIONS AND LIMITATIONS PAGE 26 ANSWERS -- MASS. INSTITUTE OF TECH. -85/10/02-SILM, D.

ANSWER L.05 (1.00)

To save a humar, life (0.5) or to insure nuclear safety (0.5)

REFERENCE PM 4.3, pg. 14 ANSWER L.06 (3 00)

1. fo determine that normal radiation levels exist based on control room and/or local instrusientation.

2. To asso.ss the need for a radiation surve'y with a portable detector.

3. To evaluate the potential for dose rate changes during occupancy.

REFERENCE PM 1 14, pg. 6 ANSWER L.07 (2.00)

a. Total reactor thermal power Reactor coolant total flow rate Reactor coolant outlet tesperature Height of water above the outlet end of the heated section of the i hottest, fuel ebennel (.25 pts each) ;

b. To establish limits within which the integrity of the fuel clad is maintained (1.0)

REFERENCE i T.S. 2.1. pg. 2-1 i

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o, f 0, a 4 L. ADMINISTRATIVE PROCEDURES, CONDITIONS AND LIMITATIONS PAGE 27 ANSWERS -- MASS. INSTITUTE OF TECH. -

85/10/02-SILK. D.

ANSWER L.08 (2 50)

a. 1. Notification of Unusual Event i

2. Alert

3. Site Area Energency

4. Alert (0.5 pts each)

b. Po t e rit i a l radiological consequences (0.5)

REFERENCE

c. PM 4.5, pas. 10 to 12 -

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b. PM 4.4, pg. 2 P

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. NOV 2 71985 Docket No. 50-20 i

Massachusetts Institute cf Technology Research Reactor ATTN: Mr. Lincoln Clark, Jr.

Director of Reactor Operations 138 Albany Street Cambridge, Massachusetts 02139 Gentlemen:

Subject:

Inspection No. 50-20/85-02 This refers to the routine, physical protection inspection conducted by

. Mr. William Hadden of this office on September R5,1985, at the Massachusetts L

Institute of Technology reactor facility of activities authorized by NRC License No R-37 and to the discussions of our findings held by Mr. Hadden with you at the conclusion of the inspection.

o g Areas examiaed during this inspection are described in the NRC Region I Inspection Report which is enclosed with this letter. Within these areas, the b inspection consitted of selective examin:tions of procedures and repres-f entative records, interviews with personnel, cnd observations by the inspector.

I' Within the scope of this inspect 19n, no violations were observed, l' - No reply to this letter is required. Your cooperation with us in this matter is appreciated.

1 3 Sincerely,

~

Ort sinal Signed By:

f2 James H. Joner

% Thomas T. Martin, Directee e Division of Radiation Safety g;; and Safeguards t*t

Enclosure:

NRC Region ! Inspection Report No. 50-20/85-02 b cc:

Public Document Room (POR)

Nuclear Safety Informatton Center (NSIC) y, , Comonwealth of Massachusetts (2) /

r '- -

Dr. O. K. Harling, Director of the s Reactor Laboratory / {

a OFFICIAL RECORD COPY MA INSTITUTE 50 0001.0.0 11/08/85

I W e s e 22 g F[f U, b b j

ll O O 2 NOV 2 71985

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Region I Docket Room (w/ concurrences)

Mariagement Assistant, DRMA (w/o enc;osure) f l

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4 O O U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-20/85-02 r

Docket No.50-20_ License No. R-37 Licensee: Massachusetts Institute of Technology 138 Albany Street Cambridge, Massachusetts 02139 Facility Name: MIT Nuclear _ Reactor Laboratory Inspection At: Cambridge, Massachusetts Inspection Coriducted: September 25, 1985 Date of Last Physical Security Inspection: September 27 - 28, 1982 Type of Inspection: Routine, Unannounced, Physical Security Inspector : %b //- /# -#f _

date William J. Madden, Physical Security Inspector

i Approved by: //- / # - #f f date R.R.Keifg ief, Safeguards Section, ORS &S /

Inspection Summary: Rovtino, unannounced, physical protection inspection on September 25,1985 (Report No. 50-20/85-02)

Areas Inspected: Implementation of the licensee's NRC approved physical secur-ity plan for the protection of special r.uclear material (SNM) of moderate strategic significance. The inspection involved 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months onsite by one NRC inspector.

(

Results: The licensee was in ecmpliance with N8tC requirements in the areas examined.

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REPORT DETAILS

1. Key persons Contacted

Lincoln Clark, Jr., Director of Reactor Operations

John Bernard, Superintendent, MIT Research Reactor Operations / Maintenance

Kwan Kwok, Assistant Superintendent, MIT Research Reactor Jerry McDade, Supervisor, Campus Security Systems

Denotes those present at the exit interview.

2. 30703 - Exit Interview

. The inspector met with the licensee representatives indicated in para-graph I at the conclusion of the inspection on Se:tember 25, 1935, and summarized the scope and findings of the inspectien.

{

3. 81480 - physical protection of SNM of Moderate Strategic Significance

, The licensee's program for the physical protection of SNM of moderate strategic significance was "eviewed by the inspector and was found to conform to 3RC requirements and the licensee's implementing procedures.

'(

Specific components of the program that were inspected inc1:Jded: records and reports; security organization; alarm response; key control; detec-tion aids; physical barriers; and written security procedures, e .

r APR 2 5 m Docket No. 50-20 Massachusetts Institute of Technology

ATTN; Mr. Lincoln Clark, Jr.

Director of Reactor Operations 138 Albany Street Cambridge, Massachusetts 02139

?

.p Gentlemen: j g.:

Subject:

Inspection No. 50-20/86-01 This refers to the routine safety of activities inspection authorized by NRC conducted by Mr. T. Foley o R-37 Licer;se No. d members office on February 25-28, 1986 and to the discussions of our findings held by Mr. Foley with Mr. Clark an of your staff at the conclusion of the irspection.

Areas examined during this inspection Within these areas, are described in the NRC Re the inspection in-d Report which is enclosed with this letter. consisted of selective examina terviews with personnel, and observations by the inspector.

Our inspector also verified the steps you have taken to correct the vio 1982, brought to your attention in the enclosure to our letters dated June 13,W October 25, 1983 and March 21, 1985.

action at this time.

Within the scope of this inspection, no violations were observed.

Your cooperation with us in this matter is No reply to this letter is required.

appreciated.

Sincerely, crisinal signed By:

Edward C. Wenzinger, Chief Projects Branch No. 3 i

' DivisionofReactorProjects 50-20/86-01 NRC Region I Inspection Report No.

Enclosure:

cc w/ encl:

- Dr. O. K. Harling, Director of Reactor Laboratory

Public Document Room (PDR) local Public Document Room (LPDR)

Nuclear Safety Information Center (NSIC)

-Commonwealth of Massachusetts (2) jf' pp

a .

Massachusetts Institute of Technolugy 2 UR S 5 W bec w/ enc 1:

Region 100cketRoom(withconcurrences) '

Management Assistant, DRMA (w/o encl)

ORP Section Chief T. Foley (w/cy of encl) l i

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U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No.: 86-01

. Docket No.: 50-20 License No.: R-37 Licensee: : Massachusetts Institute of Technology 138 Albany Avenue Cambridge, Massachusetts 02139 Facility Name: MIT Nuclear Laboratories !

Inspection At: Cambridge, Massachusetts Inspection Conducted: February 25-28, 1986 i Inspectors: T. Foley, Senior Resident Inspector, Calvert Cliffs C. Holden nior esident Inspector, Maine Yankee Approved by: . < .

M-~^ hDate T. C. Elsasser g ief, Reactor Projects Section 3C Summary: Inspection on February 25-28, 1986 (Report No. 50-20/86-01) j Areas Inspected: A routine uiiannounced on-tite inspection of licensee activities including: Action taken on Previous Inspection Findings, Facility Tour, Facility Operations, Requalification Training, Surveillance, Experiments, Radiation Protec-tions, Audits and Committees, and verification of reduced on-site storage of High Enriched Uranium (HEU). .

! Results: Although no violations were identified, two concerns were identified i regarding documentation of the licensee's bases for changes, test and experiments determined not to involve an unreviewed safety question (Paragraph 5.a), and cali-brations of dosimetry instruments (Paragraph 10). Stored quantities of HEU on site are minimal, and operation of the facility appears to be in conformance with ap- ,

plicable requirements.

I

! i 1

e L

i DETAILS

1. Key Persons Contacted "J. Bernard, Superintendent, MIT Research Reactor Operations and 4

. Maintenance

L. Clark, Jr. , Director of Reactor Operations

80. Harling, Director, Nuclear Reactor L:boratory

K. Kwok As MIT Research Reactor

E.KaraIan,sistantSuperintendentMITRadiationProtectIonOfficer "Denotes those present at the exit interview.

2. Licensee Action on Previously Identified Enforcement Items (Closed)(82-01-01) The failure to maintain at least 12 inches edge-to-edge separation of packages containing SNM was corrected by moving the Bir sub-assembly containing 1.1% enriched U0,, to a location not within 12 inches edge-to-edge of any other SNM. Additionally, within each storage location signs are posted with instructions specifically prohibiting storage within 12 inches of other SNM.

(Closed) (83-02-01) The corrective actions identified in Inspection Report 50-20/83-02, regarding the licensee's failure to adequately post the Hot Cell

. Area as a High Radiation Area, are still in place. The inspector verified the actions taken by the licensee identified in the above report, l (Closed)(83-02-02) The inspector verified that the licensee no longer uses i yellow and m genta ropes for barriers where radiation areas do not exist, and that Radiation Protection controls the use of radiation area barrier ropes.

4 (Closed) Violatien (85-01-01) The licensee's ccrrective actions to packaging 281 mil 11 curies et Rhenium-186 and 824 millicuries of Rhenium-188 wire and incorrectly labeling the package as 8 millicuries of Ct.lorine-38 for shipment i to Massachusetts General Hospital were as follows:

(1) a specific procedure for "Hot Cell" work was written,

. (2) the control of work was re-emphasized to Hot Cell workers, (3) specific references are now written on samples and pneumatic tube samples are identified, (4) specifically-shaped containers as indicated on Part II of the work form are used, (5) distinguishable markings on the samples are recorded on Part II of the work form, and (6) the gamma dose rate on the work form is verified.

The inspector verified thet these actions were performed and in effect during l . inspection of the facility.

3 (Closed) (84-01-01) Procedure should be clearly labeled with the title of the individual responsible for its implementation. The licensee has placed the responsibility on the console operator for all immediate actions of Abnormal Operating Procedures (AOP) and Emergency Operating Procedures (EOP). The

{ shift supervisor is responsible for review of the immediate actions and for follow up action. The inspector reviewed AOPs and found the procedures con-

]' tained the necessary direction to the licensed operators.

(Closed) (84-01-03) Incorporate Emergency Action Levels (EAL) into procedures such that classification of events is readily available. The licensee has incorporated into procedures the EAls listed for non-radiological emergencies.

EALs for 9xcess Radiation at the Site Boundary Resulting from a Contained Source" weie incorporated into the appropriate procedure. EALs are covered in procedures either as a sub part of major radiological emergencies or emer-gency procedures.

(Closed)(84-01-04) Accuracy of Procedural References. The licensee reviewed procedures and corrected the typographical errors which led to the inaccura-cies.

(Closed) (84-01-05) Provide high range dosimeters within the Containment building emergency lockers. The licensee located two high range dosimeters in the emergency locker in Containment. In addition, other high range dosi-meters are located outside of the Containment for use by other personnel as-sisting in emergency actic.s.

(Closed) (84-01-06) Provide guidance on supplying dosimetry to meoical per-j sonnel. Dosimeters will be issued to responding medical personnel if the in-jury involves radiation exposure or contamination. This action has been pro-ceduralized for medical emergencies.

t (0 pen) (84-01-02) Develop EALS based on specific instrument readings for each of the four classification levels specified in the Emergency Plan. The lic-ensee responded to this item in its reply to Inspection Report 84-01 dated July 25, 1984. This particular item was confusing since the licensee inter-preted the action necessary to close this item as being a rewrite of the 4

Emergency Plan. The licensee listed the actions it would need to accomplish I

a rewrite of the Emergency Plan and requested additional guidance. NRC Region i I responded on September 14, 1984 and forwarded this item to Headquarters for review. The inspector discussed the issue with Headquarters personnel and determined that resolution of this item does not require a rewrite of the 1

Emergency Plan. Additional discussions between the licensee and Headquarters were conducted. Documentation of the resolution of this item will be reviewed

{ in subsequent inspections. This item is open.

]

3. Facility Tour

. On February 25, 1986 at about 6: 00 p.m. the inspector arrived on site. Ob-

servation of physical security controls appeared adequate. The inspector met l the Assistant Superintendent and ascertained that shift staffing was in con-j

4 formance with Technical Specifications (TS). Subsequently a meeting was con-ducted with the Director of Reactor Operations regarding the scope and purpose

, of the inspection. A tour of the facility was conducted immediatel 4

after. General observations of security, health physics controls, yhousekeep- there-ing, staffing and back shift operations were noted. Control Room observations

' and Reactor Plant system parameters were monitored by the inspector and com-pared to Technical Specifications. No inadequacies were noted. Additional tours were made later during subsequent days of the inspection. Inspection tours included: Spent Fuel Pool, New fuel Vault, Reactor Vessel Head area, t Hot Cell, Rad Waste Storage areas, experimental laboratories, Blanket Testing :

Facility, and Administrative Offices. No inadequacies were identified.

4. Facility Operation Review The facility is used primarily by MIT graduate students for a variety of neu-tron activation experiments. The licensee continues to operate the reactor continuously from 8:00 a.m. Monday until Friday evening using a three shif t schedule. During the inspection t.he licensee performed various control rod

manipulations and demonstrated the "automatic control of reactor power and 4

reactivity constraints" experiment. The licensee demonstrated various reac-tivity limiting controls and safeguards associated with the reactivity control system. The inspectors reviewed shift staffing, Control Room logs and cb-served the operators' performance. Reactor coolant system parameters and system annunciators were discussed with the plant operators. General condi-tions as they applied to fire prevention and radiological cleanliness were observ6d. Although no discrepancies were noted in the above areas, the in-spector had the following comments:

(a) A review of the reactor start up and shutdown enecklists was conducted.

inv Ir.:p?ctor noted that several start up r.heckliste were not complete since some instrumentation was not checked. The inspector was able to i verify, through other documentaticn, the exact stat'.:: of the equipment.

The instrumentation in quistion did not impact on Technical Specification

) requirements. The licensee agreed that a more thorough review of check-lists was necessary.

4 l (b) The inspector reviewed the hourly calorisetric calculation performe.1 by j the operators. Additionally, the Estimated Critical Position (ECP) cal-

! culations were reviewed. The October 21, 1985 and February 18, 1986 ECPs did not have all blanks completed. However, the inspector determined th&t the blanks did '1ot apply to those startups. The licensee agreed that the ECPs should be annotated to show they are complete.

(c) The inspector also compared Technical Specification surveillance require-

. . ments with Operator Logs. The OF-1 flow recorder is bypassed during reactor start-up. The bypass is removed prior to increasing power above ,

1 a pre-set level. The inspector reviewed the Bypass Lot and determined l , that the operators were removing this bypass and signing for its removal, l

4 b

but the times and dates were not listed. This made verification of the reinstatement of the flow recorder difficult. The licensee agreed to study the problem.

Other operating documentation reviewed included the Job Workbook, Fuel Loading Permission, Shutdown Margin Calculations, and Operators' Logs. In general the licensee's record keeping was u ceptable. The filing of data in logs was orderly, and data were easily retrievable.

S. Audits and Committees A review of audit reports and committee activities was conducted,

a. Committey The committee charged with the oversight of reactor safe operation is the Reactor Safeguards Committee. The committee meets at least once each year and is responsible to the Administration of MIT. The committee chairman establishes subcommittees to assist the committee in conducting its review functions. The committee or an active subcommittee reviews and approves all operating procedures, emergency plans, proposed modifi-cations to the reactor, the use of reactor related experimental facili-ties and experiments, and all equipment and procedures involving the use of licensed radioactive material in the reactor building.

l Through a review of committee activities, the inspector attempted to

^ ascertain that the committee reviews abnormal occurrence and unusual occurrence reports, violations, categories of particular tests and ex-peri. tents, Technical Specification changes, potential unreviewed safety

questions (URSQ), emergency plans and security plans.

The inspector reviewed several unusual occurrence reports and associated

corrective actions related to licensee experiments, logs, and emergency plans, and determined that each was properly documented by the Safeguards

Committee. It was noted, however, that only "categories" of experiments are reviewed by the Safeguards Committee in order to determine whether

! an unreviewed safety question exists. The inspector further noted that

within a "category", there are experiments which have no safety analysis, t According to the licensee these other experiments do not require a review by the Safeguards Committee because they are considered "Class B" proce-i dures, i.e., they are described in the Safety Analysis Report (SAR) and i do not involve an URSQ. Instead, Class B procedures require a review by two licensed operators and the Director of Reactor Operations to de-

! termine, in part, whether a potential exists for an URSQ and consequently whether further review is required. The bases for this determination is not maintained. Similarly, bases are not maintained for other changes, I tests and experia.ents, which have previously been reviewed, and deter-l . mined not to involve a potential for an URSQ.

l l

6 The inspector stated that 10 CFR 50.59 Paragraph (a) (1) is permissive in that it allows the licensee to make changes to the ficility and its operation as described in the Safety Analysis Report (SAR) without prior approval, provided a change in Technical Specifications is not involved or an "unreviewed safety question" does not exist. Paragraph (b) requires that the licensee maintain records of changes made under the authority ofParagraph(a)(1). These records must include a written safety evaluation which provides the basis for determining whether an unreviewed safety question exists.

The inspector stated that this meant that any proposed change to a system >

or procedure, as described in the SAR either by test or drawings should bereviewedbythelicenseetodetermInewhetheritinvolvesanunre-viewed safety question, and in all cases, the safety evaluation must provide the basis for determination that the proposed change, test or experiment does not involve an unreviewed safety question.

The inspector determined that the licensee complies with the above for those changes, tests and experiments which have been reviewed and deter-mined to have a potential for an UR$Q, but not for those that have been determined not to involve an URSQ, in that the bases or reasoning for j

tiie "sorting out" (determination of why a potential for an URSQ does not exist) is not documented.

The licensee questioned the inspector as to what constituted "a change" and how other licensees resolve documenting the basis for changes which occur to system and procedures or drawings described in the SAR. The inspector discussed various acceptable alternatives and subsequently forwarded to the licensee the NRC Policy, Part 9800 of Inspection and I

Enforcement Manual "CFR Discussions" 10 CFR 50.59.

l The licensee agreed to further evaluate the requirement in light of the provided NRC interpretation / policy. This matter is unresolved pending the licensee's action to provide the documented bases or rational for those changes, tests or experiments which do not involve an unreviewed safety question (50-20/86-01-01).

b, Audits Audits of f acility operation > are performed primarily by the Reector Superintendent. These audits are quite thorough and comprehensive, i However, corrective action, recomendations and implementation are largely the responsibility of the Reactor Superintendent. The Super-i intendent completed audits of the following, during October through

December 1985

(1) Reactor Console Log Unusual or Abnormal Entries

, (2) Changes to procedures / checklists / manuals (3) Job Workbook Records (4) Test and calibrations

7 (5) Radiation Surveys and Environmental Monitoring Radioactive Effluent Records (6) Fefueling and Excess Reactivity (7) Fece'nmendation of Reportable Occurrence Reports and Unusual Occur-rence Reports

, (8) Trai.iing Files (9) QA Program /Tagouts/ License R-37.

The Reactor Superintendent perfctms these audits repetitively on three month cycles throughout the year in addition to his normal auties. The inspector reviewed the above audits for July through December 1985, and noted th M there were no substantive findings.

However, the inspector questioned the lack of independence and organiza-tional freedom provided by this method of auditing one's own work. The licensee had previously been concerned about this matter and subsequently initiated an annual independent audit by Mr. W. Fecych, a licensee con-sultant. Audits by Mr. Fecych for the 1984 and 1985 period were reviewed by the inspector and found to encompass outstanding items, operating logs, and dosimetry calibrations.

The inspector stated that although this independence provided more ob-ject M ty, the scope and depth of the audits was limiting and should be more comprehensive.

The licensee's Safety Analysis Report which described the Quality Assur-ance Plan, dated October 1970, provides justification for not requiring I the independence and organizational freedom required by 10 CFR 50, Ap-pendix B; however, Section 11.2.2 provides a list of activities which fall under the licensee's Quality Assurance Program, and as such should be included in a schedule to be audited on a periodic frequency. Al-though a clear requirement for audits addressing all aspects of the Quality Assurance Program is not evident, current regulations and indus-try standards do place more emphasis in this area. The inspector recom-mended that the licensee consider evaluating current requirements and provide additional independence to those areas within the defined Quality Assurance Program.

6. Technical Specification Surveillance The inspe: tor verified by review of plant surveillance and other records that the folicwing TS surveillance requirements met frequency and acceptance cri-teria:

TS No. Requirement 6.4.1.3 Helium Gas Holder Alarm 6.4.1.4 0 0 Helium System Alarm 3

)

i 8

TS No. Requirement a

6.4.15 Reflector Tank D 20 Level Scram 6.2.4 Period Level Indication Off Scale Scram 6.1. 4.1 Nuclear Safety System Pesponse Time 6.1. 4. 2 D2 0 Reflector Dump Time 6.1.4.4 Primary Coolant Flow Scram Time No inadequacies were identified.

7. Emergency P1anning The Massachusetts Institute of Technology Reactor Emer ency Plan was reviewed.

Drills and lectures a e periodically (at least annuall ) performed. Training records, changt.s in the plan and audits of emergency p anning activities were reviewed. The November 19, 1985 Emergency Plan Exercise consisted of MIT Reactor Operators, Radiation Protection Personnel and MIT campus police.

Local police, hospital and fire department agreements were verified to be up-to-date. The Emergency Plan is vto-date and being effectively implemented except as noted in paragraph 2, icensee Action on Previously Identified Items," Item 84-01-02, which remains open.

No inadequacies were identified.

8. Experiments ,

Experiments performed at the MIT Reactor are varied. Currently, neutron ac- '

tivation and analysis and automatic reactivity control experiments are in progress. Experiments are divided into the following categories: reactor operation exreriments, Beam Fort experi w nts incore experiments, thermal ,

columnexperimentsandmedicaltherapyexperleents. The licensee uses a "Proposed Experiment Review and Approval Form" in order to control the appro- ,

val process. The inspector reviewed the following experiments for approvals and safety analysis:

Use of Dry Ice in Pneumatic Tubes Sodium metal filled subassembly in the Blanket Test Facility !

-- Closed Loop Control of Reactor Power using Shim Blades and Regulating Rods simultaneously The use-of-dry-ice experiment and use-of-sodium experiments were not accom- ,

, panied by safety evaluations, however, they were reviewed and approved. The acceptability of these experiments was based on similarity to the other ex-periments which had previously been approved and which were accumpanied by i

l 9 a written safety evaluation. The inspector verified that in addition to re-views, approvals, and safety evaluations, predicted parameters were determined and ascertained within tolerance, irradiated items were properly controlled, and individuals conducting the er,periments were trained prior to using the facility (see Trainir;, paragraph 9.b).

No inadequacies were identified.

i 9. Training Review -

l

a. Requalification iraining A review was conducted of licensed operator training, examinations and I reactivity manipulation records, Schedules of lectures and samples of '

lesson plans were also reviewed. The inspe: tor ascertained that required I

records were maintained and that the licensee requalification training program was current and fully implemented.

A review of the 1984 and 1985 records indicated that five senior reactor operators had passed their requalification examinations. One reactor operator was upgraded by virtue of passing the SRO examination. One ;

reactor operator's license duties were suspended by the licensee for failure to take the requalification examination.

No inadequacies were identified.

b. Experiments and Student Trainino .

t l The inspector reviewed documents and discussed with various department

' staff the training of individuals wno conduct experiments. Personnel t are trained in the following areas: '

10 CFR Part 19 10 CFR Part 20

-- l Tables from 10 CFR Parts 20 and 30 USNRC Regulatory Guide 8.13 Procedures for Radiation Protection -

Facility Emergency Evaluation Procedure Film Badge Classification Procedure Radiation Exposure Record Application Exclusion Area Entry Permit Maximum Permissible Dose '

l Each person is given approximately three days to read the above material.

A one and one-half hour lecture is given on the same material followed by a question / answer session to determine students' knowledge of exposure i limits and restri<:tions. Twenty hours of classroom instruction is pro-

. vided on the use and handling precautions associated with the experi-mental facility and equipment prior to allowing each person to work or attend classes in the building.

I I

10 Retraining is given annually to persons who handle or receive radioactive !

materials. This retraining includes but is not limited to the following topics:

Permissible Radiation Doses :

Facility Organization Biological Effects of Radiation :

Facility Evacuation Plan l l No inadequacies were identified. l J t l 10. Radiation Protection Controls

, The inspector noted radiation postings and controls throughout the facility.

4 Radiation instruments were noted to be calibrated and source ch2cked regularly, i Reviews were conducted of radiation surveys, contamination surveys, exposure ;

4 records of experiments and MIT staff. (Generally, the radiation levels are '

i less than 5 mr/hr in most accessible areas.) Hot Cells were adequately posted as riigh Radiation areas. Some small areas around the Beam Ports had higher .

radiation intensities (as high as 15-25 mr/hr) wheress other areas around the l

Beam Ports were 1-2 mr/hr. The inspector indicated that 10 CFR 50, Appendix L I provider guides for maintaining dose to individuals as low as reasonably

! achievable. The licensee agreed to consider placing controls / signs in or ,

j around these areas where higher than normal (5 mr/hr) radiation levels could r j exist to make personnel aware of the potentially higher intensities and to !

aid personnel in minimizing their dose.

i

During review of dosimetry records and calibrations of instruments, the in- !

I spector determined that personnel pocket dosimeters were not being calibrated.

The inspector noted that 10 CFR 50, Appendir B requires that all devices used l l

to ensure quality should be properly calibrated. The licensee provided a !

1 quality assurance audit that previously had identified this same issue. The !

I licensee stated that programs would be established to calibrate all dosimetry. !

The inspector indicated that pending licensee action on the Quality Assurance !

j Audit, dated November 18, 1985, this item is unresolved (86-01-02). [

l i

11. Stored Quantities of Hio5 Enriched Uranium (HEU) On Site -

i In accordance with NRC Inspection and Enforcement Temporary Instruction 2545/1, l the inspector examined the quantity, storage and controls associated with HEU l

j on site.

< The inspector observed the new fuel vault contents to physically ascertain l what new fuel was accessible. Only one fuel element and a few miscellaneous components totalling less than 1 kg of HEU were in the new fuel vault. Safe- i guard controls associated with the vault are described in Safeguard Inspection i Report (50-20/84-02). l t

Through discussions with the licensee, review of operation history, and ob- l

servation of the Spent Fuel Pool, the inspector determined that the quantity l

of material exempt from the licensee's inventory of accessible HEU was greater l i !

l,. 11 i

than 100 Rem /hr at three feet. The current inventory of accessible fuel is of Low Strategic Significance. The current MIT policy is to maintain ' nun-o dreds" of grams of accessible HEU on site versus the thousands" of grams permitted, excluding the self protecting fuel, except just prior to fuel

transfer. This was documented in a letter to the Secretary of the Commission

, from L. Clark, October 19, 1984.

l The licensee is currently awaiting a fuel cask from DOE in order to reduce

! its inventory of spent fuel.

12. Exit Interview At the conclusion of the inspection on February 28, 1986 the inspector met with the director of the facility and reviewed the scope and findings (i.e.,

unresolved items in paragraphs 5 and 10). The inspector noted the licensee's candor and good cooperation. At no time during this inspection was written material provided to the licensee by the inspector.

n

AUG 211986 q:

V Docket Ko: 50-20 License No. R-37 1

Massachusetts Institute of Technology Research Reactor ATTN: Mr. Lincoln Clark, Jr.

Director of Reactnr Operations 138 Albany Street Cambridge, Massachusetts 02139 Gentlemen: -

Subject:

Inspection Report No. 50-20/86-02 ,

A routine, unannounced inspection was conducted on July 21-22, 1986 by Ms. Jean A. Ciofft of the Massachusetts Institute of Technology Research Reactor radiation protection program. The elements of the program reviewed I are described in the enclosed inspection report.

Within the scope of this inspection, no violations or deviations were observed, j

No reply to this letter is required. Your cooperation with us in this matter is appreciated.

Sincerely, kitu.n ut ::tys l.g k V j

6 as T. Martin 31 rector ,

4 vision of Radiation Safety l and Safeguards l

Enclosure:

NRC Region ! Inspection Report No. 50-20/86-02 j i

t l cc w/ enc 1. 1 vCr. O. K. Harling, Director of the Reactor Laboratory :

4 4 r. Alan Ducatman, Director, Environmental Medical Service Public Document Room (PDR) local Public Document Room (LPDR) l l Nuclear Safety Information Center (NSIC)

vCommonwealth of Massachusetts (2)

I I

1 i i t OFFICIAL RECORD COPY IR MIT 86 0001.0.0 08/18/86 O/nn 3

< v m W-e-9 wN< . A- 6-t _

3 The licensee determined that the cost-benefit ratio for all Argon

! reduction work was in the range of $550-1100 per man-rem. These figures compared favorably to the guidelines specified in 10 CFR 50, Appendix 1, section !!.D. Additional Argon reduction work took place in 1985 (see paragraph 7.0). This item is considered closed.

l 3.2 (Closed) 86-01-02 (!aspector Follow-up) Calibration of personnet pocket dosimeters. The licensea initiated the calibration of their pocket ion chambers. The dosimeters will be calibrated semi-annually using a 5 curie Cesium-137 s>uren. This item is considered closed.

4,0 Traintno and Qualification of Personnel The licenste's program for training and qualification of personnel was i reviewed with respect to criteria contained in

- 10 CFR 19.12. "Instructions to Workers";

- Technical Specification 7.10. "Radiation Protection Program."

l The licensee's performance with respect to the above criteria was deter-mined by:

I

- review of the "Massachusetts Institute of Technology Required

! Procedures for Radiation Protection,"

l

- discussions with licensee personnel, i

W1;hin the scope of this review, no violations were identified. The i

licensee appeared to be trait.ine and qualifying radiation workers in accordance with regulatory rer,Arements and the conditions of their l

license.

5.0 Imolementation of the Radiation Protection Program

! The licensee's program for controlling radioactive materials and con-

! tamination, providing surveillance and monitoring, and establishing and j maintaining administrative radiological work controls was reviewed relative to criteria and commitments in:

l

- 10 CFR 19.11, 19.12, 20.201, 20.203 and 20.401;

- Technical Specification 3.8, "Radioactive Effluents and Radiation j Monitors";

i

- Technical Specification 4.3, "Reactor Control, Safety, and Radiation i Monitoring System Surveillance"; and i '

! - Technical $pecification 7.10, "Radiation Protection Program."

l TDe licensee's performance related to the above criteria was determined i by:

\

4 4

facility tour on July 21, 1986 to observe work in pro v ass; postings, signs, and labels; and radiation monitoring instrumentation; review of calibration records for hand and foot monitors, survey ;

instrumentation, area radiation monitors; :

review of survey records for radiation, contar' nation, ard airborne L radioactivity; and discussions with licensee personnel.

Within the scope of this review, no violations or deviations were identified. The inspector witnessed the lifting of the reactor head and -

noted that licensee personnel observed the proper industrial safety pre- s cautions, and efficient contamination control techniques. The inspector 1 also observed the addition of signs to remind experimenters to survey their work areis for possible higher radiation intensities around beam ,

ports.

Two areas for improvement were identified for licenset attention.

The licensee maintained no implementing procedures for the reactor radiation protection program. For instance, there were no procedures for calibration of survey instruments and pocket dostseters, when to reso pocket dosimeters and log the reading before re*2eroing, nor how to resolve discrepancies between pocket dosimeters and film badge results. The licensee stated that due to the long employment of all health physics personnel, such procedures were not necessary. The inspector stated that such procedures were necessary for the program to be implemented consistently if the staff were replaced due to illness or retirement. The licensee stated that because of the upcoming retirement of the Reacter Radiation Safety Officer, such implementing procedures would be developed and established. This item will be reviewed in a future inspection (86-0?-01).

The licensee uses a 5 curie Cesium-137 source for their instrumeret and pocket dosimeter calibrations. However, the source is used in a room without interlocks, warning lights or alarming devices at the entrances to indicate when the source is exposed. The inspector discussed this practice with licensee representatives, who stated that all calibrations were performed when the staff and experimenters

. were not present, and the tedividual performing the calibration remained in the vicinity to provide positive control over the area.

The inspectors stated that while the controls being used met minimum regulatory requirements, they may not be suf ficient to prevent an unplanned exposure should the individual leave the area, or a cuerd inadvertently enter the room. Following this discussion, the licensee stated that they would: 1) set up a barrier to prevent personnel from inadvertently wandering near the calibration area, and 2) modify the calibration facility with warning lights, alarming devices, and/or interlocks to prevent inadvertent personnel entry. This item w1?1 be reviewed in a future inspection (86-02-02).

5

l 6.0 Internal and External Exposure Controls The licensee's '.ternal and external exposure control program was reviewed agair t M aria provided in:

10 CFR 20.104, u t, 20.303, 20.104, 20.105, 20.201, 20.202, 20.203 and 20..

The licensee's perf ormance relative to the criteria above was determined by:

a review of exposure records for 6 radiation workers; tour of the counting laboratory and whole body counter in Building 20; and discussions with licensee personnel.

Within the sv9se of this review, no violations were identified. The licensee uses Landauer film badges for dosimetry of record. Visitors to the reactor are issued pocket dosimeters. Internal exposures are monitored by urinalysis and whole body counting.

7.0 Effluent and Environmental Monitoring The licensee's program for monitoring linuid and gaseous effluents was reviewed with respect to criteria contained in:

10 CFR 20.106; Technical Specifications 3.8, 4.3 and 7.13.5.

The licensee's performance related to the above criteria was determined by:

tour and observation of control room effluent radiation monitor indicators; review of effluent monitor logs;

- review of the following effluent monitor calibrations procedures:

P.M. 6.1.3.9.1, "Water Monitor Calibration Procedure"

' ' P.M. 6.1.3.9.2, "Particulate Monitor Calibration Procedure"

P.M. 6.1.3.9.3, "Gaseous Monitor Calibration Procedure"

P.M. 6.5.9.2, "Environmental Monitor Calibration Procedure"

- - - - - - - . - ,, n. -.-n- - , ,-_ - -, - - - , , , - - ,

t 6

review of the 1984 and 1985 Annual Reports; discussions with licensee personnel.

Within the scope of this review, there were no violations or deviations identified. The licensee was calibrating all effluent and environmental monitors in accordance with license conditions. Environmental surveys indicated that there were no inconsistencies for the monitoring periods during 1984 and 1985. Furthermore, the licensee was able to further reduce gaseous releases in 1985 by additional studies of the sources generating the Argon-41 in the reactor and by the use of an inert gas blanket system for the reactor. In previous years, the licensee dis-charged 'd00 to 8000 Curies of Argon-41 per year. In 1985, the li".ensee was able to reduce the gaseous discharge to about 4000 Curies for the year. (See additional information on Argon-41 released in paragraph 3.0.).

8.0 Exit Interview The inspector met with the licensee's representatives (denoted in Para-graph 1) at the conclusion of the Inspection on July 22, 1086. The inspector summarized the purpose and scope of the inspection and findings as described in this report.

-. ,.---- -, - _ - . - -- -.- --,----

ML20155G648

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