ML20112E054

Forwards Rev 31 to SAR for Mit Research Reactor (MITR-II)

ML20112E054
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	01/04/1985
From:	Lisa Clark MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
To:	Thomas C Office of Nuclear Reactor Regulation
References
NUDOCS 8501150012
Download: ML20112E054 (10)

Text

. ,p von N7 NUCLEAR REACTOR LABORATORY l

<get ,fQl AN INTERDEPARTMENTAL CENTER OF  %@g4#

MASSACHUSETTS INSTITUTE OC TECHNOLOGY O.K. HARLING 138 Albany Street Cambridge, Mass. 02139 L CLARK, JR.

Director (617)253-42 02 Director c: Reactor Operations January 4, 1985 Mr. Cecil 0. Thomas, Chief Standardization and Special Projects Eranch Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

SAR Revision No. 31, License No. R-37, Docket No. 50-20

Dear Mr. Thomas:

Massachusetts Institute of Technology submits herewith Revision No. 31 to the " Safety Analysis Report for the MIT Research Reactor (MITR-II)", Report No.

MITNE-115 (October 22, 1970), as amended. The purpose of the revision is to upda te the SAR so that it will reflect some minor changes that have been incorporated in procedures and related documents.

Enclosure 1 describes the revisions and provides instructions for SAR page re placements . The revised pages, with margin lines to indicate the changes, are also enclosed.

The proposed SAR changes have been approved by the MIT Reactor Safeguards Committee.

Sincerely, Lincoln Clark, Jr LC/gv

Enclosures:

SAR Revision No. 31 Revised Pages cc MITRSC (with enclosure)

USNRC-NRR (1 cigned and 12 copies, with enclosures)

USNRC-DMB (with enclosures)

USNRC-0MIPC (with enclosures) 8501150012 850104

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V Enclosure 1 Mr. C. Thomas, USNRC, Division of Licensing, (1/4/85)

SAR Revision No. 31 Remove Insert Page Page Description of Change Fig. Fig. Figure updated to include valve DV-69 which is normally 9.3.1-1 9.3.1-1 locked open but can be used to isolate the helium (9/24/76) (9/16/83) cover gas system when the blow off patch is serviced.

11.3-2 11.3-2 The name of the checklist used to document QA (3/17/72) (10/26/84) activities has been made more descriptive of its use, i.e. " Quality Assurance Approval Requirements Checklist". .

, 11.3-3 11.3-3 A footnote has been added as a reminder that the (5/6/82) (10/26/84) uncertainty allowed in the fuel density tolerance was actually 1.10 rather than the 1.05 envisaged when the SAR was first written.

,s 11.10-3 11.10-3 Same as 11.3-2.

(3/17/72) (10/26/14) -

(

x_j 11.17-2 11.17-2 Table 11.17-1 revised to reflect records retention (6/30/78) (10/26/84) requirement of nuclear insurer (policy termination plus

& 10 years). Retention time for Items 2 and 5 reduced to 11.17-3 reflect actual requirements. Name of checklist in Item (10/26/84) 17 updated.

(nont) 11.17-4 Fig. 11.1-1 revised to show MIT Radiation Protection (2/10/81) (10/26/84) Committee (inadvertently omitted) and changed Environmental Medical Service reporting structure.

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CARD JRE 9.3.1-1

',T O W FLOW $Y GT E M SEP 161983 .

8501150012-OI . .

SAR 11.3-2 0 Committee, or an application to the NRC for an amendment to the Technical l Specifica tions.

Assurance tha t the approved design is correctly translated into specifica tions, and ins talla tion, testing and operating drawings, procedures and l

instructions is accomplished by requiring that all such documents which fall i

" I within the scope of the quality assurance program, as defined in Subsection 11.2, shall be checked by a knowledgeable individual other than the one who i prepared - the document and that they shall be signed or initialed by both individuals.

In order to facilitate the preparation and revision of documents, to assure the appropriate reviews and approvals, and to document operation of the quality assurance program, a " Quality Assurance Approval Requirements Checklist" is used for specifying and recording the required actions and QA approvals. The

- checklist is used for' the purposes of this subsection, Design Control, and also for most of the activities . described in the following subsections.

.. .V The thermal and hydraulic characteristics of the MITR-II design are E dependent . in large part upon the assurance tha t the fuel design specifications reflect the uncertainties '(hot channel factors) incorporated in the design calcula tions. As described'in this SAR, Section 3.3.4, uncertainties are

- included'in the derivations of both the safety limits and the operating limits.

-E For the former case, eovation 3.3.4.1' 9 vis derived; it conttins the hot channel factor FHC, which was estimated in the SAR but which will be measured during the .preoperational tests both to determine its value and also the associated uncertainties. For FHC, the fuel hot channel factors are given in Table 3.3.4.3.3-2, under enthalpy factors..

For the operating limits, equation 3.3.4.2-11 was derived; it contains the core design constant G, an upper limit for which was estimated in the SAR

.(Section 3.3.4.3.3, page ' 3.3.4-38) but which will be determined during the startup experiments. The uncertainties for each of the parameters in G are given in Table 3 . 3 . 4 '. 3 . 3- 2, un' der film temperature difference and heat flux factors..

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SR#-0-84-20 OCT 26 1984 F a b e ._

SAR g . 11.3-3 1 ,

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O For both the safe ty limits and the opera ting limits , the factors associated E with the fuel are the channel tolerance, the eccentricity, and the fuel density 5 tolerance. The chande: tolerance uncertainties are based on the channel width f

e specifications, and the eccentricity is based on the clad specifica tions, neglecting the small additional eccentricity which might be introduced by fin 5 tolerances. The fuel' density uncertainty in FHC is based on the tolerance in the U-235 loading per piste.

For G, not only the total pla te loading but also

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the local variatiens in U-235 density due to core area, core thickness, and U-235 segregation must be considered. No quantitative specification is set on the permissible variation from the plate average due to U-235 segrega tion.

[ However, the tpecifications will require tha t the variations in U-235 density from all four causes will not combine to cause a local heat flux in excess of a specified level. Thi's will be controlled by comparing the area of maximum U-235 7 density on representative plate radiographs with test specimens fabricated to

- show average and limiting densities, and X-rayed along with the plate. A y densitometer- will be utilized to resolve questionable cases. An upper limit of

'+5% variation in the local heat flux compared to the nominal will be used to set 7

the specification, but this may be relaxed if the fuel fabricators show that it is too stringent. In such event, the uncertainty of 1.05 to cover the fuel

{ . density tolerance will be increased correspondingly *, and the new value will be used in the evaluation of G during the preoperational tests (See SAR Section

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,1 3.3.4.3.3, page 3.3.4-39).

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= @. 'L ; *1.10 has been used in fuel fabrication specifications and in the Technical Specifica tions (p. 3'-6).

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SR#-0-84-20 , j OCT 26 1984 ep i .~

SAR

- 11.10- 3 b,m Work performed by Reactor Division personnel, including ins talla tions, modifications, maintenance, repairs, and refueling, shall also be subj ec ted to inspections and tests. Before a job is considered complete, the supervisor, where necessary, initiates a " Quality Assurance Approval Requirements Checklis t", if this has not already been done. The Shift Supervisor assures tha t the above form has the nec, - sty approvals before he authorizes startup, s tartup af ter normal weekly checklis t, startup af ter special test as specified, continued operation, etc. For those activities where special forms or data sheets are routinely in use, such as refueling, the special form or data sheet

- provides for two signatures.

All failures to meet the requirements of inspections or tests, which cannot be corrected, shall be reported to the Reactor Director or Co-Director, or their respective designees, as appropriate, prior to opera tion.

Some activities cannot be proven acceptable by test methods and must be subject to inspection by a second individual to assure satisfactory completion of the work. Such activities include the evaluation of experiments and irradiation series and the correct placement of fuel elements in the core on refueling. In~these cases, the person performing the evaluation or fuel placement signs the evaluation form or instruction sheet as does the person approving the evaluatian or witnessing the fuel transfer.

The . MITR-II preoperation tests , . criticality studies, and startup

. experiments are specified in Section 14. Written procedures and written reports of L the test results will be prepared. A qualified supervisor may be assigned by the Co-Director to be responsible for a given ~ test. He will prepare the written procedures, stating the purpose of 'the test, study, or experiment, the acceptance criteria as given in the SAR, Technical h

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?? Table 11.17-1 E$

QUALITY ASSURANCE RECORDS Retention Racord hormal Location Responsibility Code *

1. Reactor Opera tions Log , Control Room Reactor Supervisors A

2. Mechanical Maintenance Log Maintenance Office Maintenance Foreman B 3 .- Ins trument/Eqalpment Itistory Log Control Roon Reactor Supervisors A

4. Instrument / Equipment Malfunction Co.ntrol Room Reactor Supervisors A Worklist *

. 5. System Test and Calibration Book. Control Room Operations Superintendent B Business Office Business Manage r C 6., Purchase Orders - Stock Items

7. Purchase Orders - Fuel QA File (previously Director for Reactor Operations B (with inspection records and Business Office) ce r t i f ica tions) 8 Purchase Orders - Other QA File (prevsously Director for Reactor Opera tions B Business Office) -

9. Company or ' personnel qualifica tions With applicable P.O. Director for Reactor Operations B

10. Irradia tion and experiment reques ts, Operations Office Operations Superintendent A reviews, approvals and shipping forms l

11. HITR Safeguards Comm. records Ileadquarters Office MITRSC Secretary A g (Minutes, approvals, annual report, l ,,

unusual occurrence reports) m

12. Engineering drawings, speci fica tions QA File Room Opera tions Superintendent B h ,';"

.g and procedures [

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.n,' j; Table 11.17-1 QUALITY ASSURANCE RECORDS (Continued)

. Retention

-Record Normal Location Responsibility Code *

13. Electrical and Electronics drawings, QA File Room Electronics Supervisor B specifications and procedures'

,14 Design. reviews:' HITR-II QA File Room Co-Director for MITR-II A

15. Design ~ reviews:' Other QA File Room Director for Reactor Operaltons B

16. MITR-II initial test reports lleadquarters Office Director for Reactor Opera tions A

~17. Quality' Assurance Approval QA File Room. Director for Reactor Operations B Requirements Checklist

18. Audits Dir. or Co-Dir. Office Director for Reactor Opera tions A

• Retention Code: A - Termination of reactor's nuclear liability insurance, plus 10 years; B - until item disposed of, plus 5 years; C .2 years g.

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HIT Radiation Director, Medical Depart men t HIT lleactor PE V "E Protection Comm.

Saf eguard :

Com:ni t t ee Ilead, Environmental Medical Service Vice President for Research Radiation Protection Officer, MIT Reactor Director, fluclear Reactor Lahorntory  ;

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• ~l Radiation Protection Of ficer, !!ITR 4

Director of Supervisor Reactor Operations _ g Supervisor Trace Analysis Reactor titilization RPO Technicians

1. abo ra tor y

, t Radio-i llend quarters ,, _ _

Operations , _' , _ Administrative Assistant Secretary Secretary Chemist , i I

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, i Technical '

Expeririant i 3"P"# " *" "" "P ##'" #

Assistant I Operations of Operations Engineering 3

and Haintenance and Design i

i - l Secretary ' Assistant Supt. Draftsman Secretary - Machine Shop 8

Receptionist l Foreman 8

i Senior Shift i i Supervisor l l l

' ' Mach inis t s llelder Stock Clerk l

Electronics Proj ect Shif t Supervisors /

Technician Mechanic Senior Reactor Operator.

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5 Shop l'elper Reactor Operators I

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• $9 FIR.11.1-1 MITR Organization Chart .-

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