ML19208D462

Notifies of Planned Changes for Low Power Reactor Assembly

ML19208D462
Person / Time
Site:	05000356
Issue date:	04/13/1973
From:	Beck G, Hang D ILLINOIS, UNIV. OF, URBANA, IL
To:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
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ML19208D457	List: ML19208D456 ML19208D460 ML19208D462
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NUDOCS 7909280452
Download: ML19208D462 (4)

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2ffi Nurlent Engsner o on: 1.aborntory Ils bana, Illi"vii /i/Nol NUCLEAR ENGINEERING PROGRAM UNIVERSITY OF ILLINOIS 'l k .-t raa Cole 217-333-2295

, April 13, 1973 t

I i Mr. Donald J. Skovholt

Assistant Director for Reactor Operations Division of Reactor Licensing U. S. Atomic Energy Commission Washington, D. C. 20545

SUBJECT:

Information Letter i

Illinois Low Power Reactor Assembly (LOPRA)

License R-117 l

Dear Mr. Skovholt:

This letter describes two changes that are planned for the LOPRA.

These involve changes in the region that surrounds the basic core from those that were described in the Safety Analysis Report (SAR). The changes do not require Commission authorization as specified in 10 CFR 50.59.

Changes:

To indicate the changes, reQrence is made to Figure 16, page 29 of S. A.R., February,1970 and the f i"ure enclosed with this letter.

At the present time, criticality is obtained with 61 fuel elements using a basic 7 x 9 array. The 7 dimension is parallel to the thermal column as shown by the figures. The succeeding parallel rows contair. ' elements with the exception of the last row which contains only 5 elements.

The plans call for placing graphite reflectors on the two sides of the array which presently do not contain fuel elements. The reflector will have dimensions of 3 x 141/2 x 16 inches, which corresponds to the space on the grid plate and the lateral and height dimensions ef the fuel array.

Pins, which fit into the grid plate openings, will be placed on the bottom portion to keep the assemblies rigidly in place. To make room for the reflectors, the four outer (two on each side) poison tubes will be removed from both of the present safety rods.

F' With this change, it is estimated that a critical assembly nay be obtained with 57-58 fuel elements. Graphite dummy elements may be placed in the outer (9th) row for either gaining or adjusting the excess reactivity of the core.

Evaluation The changes have been evaluated and approved by the Reactor Staff and the Nuclear Reactor Commit tee. 1he approval is based on the conclusion that there will be no changes in the Technical Specifications and that the criteria of 10 CFR 50.59 are fully satisfied. Thus a licensed amendment or change with Commission authorization would not be required.

Removal of the outer pins of the safety control rods will result in a slight decrease in the negative reactivity It can be noted from the figures that the pins to be removed are outside the core region. The present worth

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!.I 's page 2 of these rods if4 about $ 1.50 each, which is much larger than the tech i

spec minimum limit of $1.50. The same is also true of the shutdown margin since this is directly related to the worth of the safety control rods.

The present margin is about $ 1.00 compared to the minimum limit of $1.00 for tech specs. The individual fuel elements will show a slight increase in reactivity since fewer elements will be needed.

The analysis of the consequences of accidents or malfunctions is the same as those evaluated in the S. A.R. In this connection, the probability of occurence or the consequence of an accident or malfunction of equipment important to safety previously evaluated in the S. A.R. would not be increased; the possibility for an accident or malfunction of a different type than any evaluated previously in the S. A.R. would not be created; and the margin of safety as defined in the basis for any technical specification is not reduced.

Initial Criticality 1he fuel loading procedure outlined in Section XIII of the S. A.R. will be followed with the exception that the TRIGA will be used as the source. hhen the loading is completed, a check will be made on the excess reactivity- The reactivity worth of the poison rod and both safety rods will then be determined.

Although the initial indication of the power level will be based on the present values, a check will be made using the negative temperature coefficient as the criterion.

Experiments The experiments that are planned for the facility will follow the rules as prescribed in Section 6.8 of the Technical Specifications.

Yours truly, Gerald P- Beck Reactor Supervisor Daniel F. llang Chairr:an, Nuclear Reactor Committee u g.

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Sketch of Top Vim. of the I.0 PPA Corn Showing thc in t he 1,OPl.1 I,ocat ion of the Sa fet y' Cont rol itods Core I.attice 8 Location of 61 fuel elements with present core.

X Safety control rod pins to be removed.

sembly rigidity, h Location of pins to give jgshadedareasarelocationsofnewgraphitereflectors, assem

Annual Report , Illinois Advanced TitlGA Page 4 January 1, 1978 - December 31, 1978 the experiment to det e rmine t he tempe ra ture increase. An inc rease o f 33"C was noted giving an actual temperature of about 58"C From this information, a gasoline sample was prepared and placed in an oven at a temperature of 95 C for two iiours. Upon removal, it was noted that the plastic had softened, but there was no indication of any Icakage. Upon this evaluation, the experiment was approved by the Nuclear Reactor Committee.

I VI. RiiLEASE OF RAI)l0ACTIVi! MNIERI ALS The average concentration of A-41 released to the environs via the building exhaust system was 3.3 x 10- uCi/cc. 'lhe total release for the year was 1.3 curies with a range of 35-200 mci per mont h. It is estimated that about I mci of trititua is released during a year from t he evaporat ion of water in the reactor tank. 'the gross beta activity in the wate r ef fluent to the saaitary sewer from tne reactor laboratory retention tank was 2.7 microcuries.

VII. ENVillONMEN'I AL SilRVEYS There were no environmental surveys taken during the reporting period.

Contamination surveys were made in the laboratory as indicated in the following section.

Vili. PLRSONNI.L RAI)lNI ION h.\POSURii ANI) SUlul.YS b l'iult. I ACI LI l Y A. l'ersonnel R.idiation Lxposure Fif teen persons were assigned film hadnes at the facility. Th ree of tn ese are full-t ime empleyacs , while the others averan2 less than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> per week at the laboratory, The badges were sent to Radiation Detection Company of California where dosages less than 10 millirems are not reported. In addition to the hadge, a dosimeter is worn by an individual i f an above radiation exposure is likely to occur. 'lhe table below gives the dose received by those assigned fiIm badges.

Dose (rems) Number of Indit iduals No measurable exposure 1I 0.01 . O.10 3 0.10 -- 0.25 0 0.25 -- 0.50 1 'l o t a l = 15

'lhe highest individual dose was 315 millirems shis was received by the Reactor llealth Physicist, who handles 95*, o f the radioisotopes that are made, does smear tests on all Campus scaled sources, and performs calibrations on radiation monitoring inst rumentat ion. Individual doses to students and visitors, from dosimeter readings, was less than 10 millirems.

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11 . Contamination Surveys Smear samples from 34 locations in the laboratory are taken at six to eight week intervals. lhe removable beta contamination is determined by checking the samples with a flow counter.

The maximum concentration occurs in the vicinity of the tubes from I

which samples are removed after an irradiation in the reactor. During this i

year there were 2,672 samples irradiated. In this area, the contamination

-5 varied from 52 to 13,600 dpm/100 cm or 2.3 x 10~ to 6 i x 10 uCi/cm Smears from all other areas on the floor and from laboratory benches showed 2 -6 a maximum of 349 dpm/100 cm or 1.6 x 10 uCi/cm' IX. NUCLEAR Rli AC~1Olt COM-ilTllili lhe present committee is composed of 4 members of the Nuclear !!ngineering Academic Staff, 2 mer.bers f rom the llealth I'hysics Staf f, and the Reactor Supervisor. Dr. Arthur C. Chilton continued as Chai.rman during the year.

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University o Illinois at Uraana-Clam ign Nuclear Engineering Program

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Urbana. tilinois 6180!

September 4, 1979 (217) 333 2295 Director Division of Reactor Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

Annual Report, Illinois LOPRA Reactor License fio. R-il7 Docket fio. 50-356 The following is written to comply with the requirements of Section 6.7.f. of the Technical Specifications and the conditions of Section 50.59 of 10 CFR 50. The outline follows the number sequence of Section 6.7.f. of the Technical Specifications.

Yours truly, C/ \M -

/,.,, ,, . -, - ' (; )- /, s' Gerald P. Beck, Reactor Supervisor

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7Fthur B. Chilton, Chairran fluclear Reactor Conmittee GPB/ABC:rm Attachment tsu ,

AfiflVAL RLPORT August 1, 1978 - August 1, 1979 ILLIrl0IS LOPRA REACTOR Facility License R-ll7 I.

SUMMARY

OF OPERATIllG EXPERIEllCE A. Usag_e The LOPRA was scheduled for operation a total of 63 hours7.291667e-4 days <br />0.0175 hours <br />1.041667e-4 weeks <br />2.39715e-5 months <br /> and was in actual operation a total of 29.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. 5cheduled operations averaged 5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> / month and actual operation 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> / month. The LnPRA was used for training during this period. The types and percentages of usage for the

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scheduled time were:

Approach to Critical Experiments 51%

Operator Training 13%

Measurements 36%

B. Performance Characteristics - flone C. Changes - flone D. flew Experiments - None II. TABULATI0fl 0F OPERATION Hours Critical

• and Energy Steady State Operation 29.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> .0043 MW-hrs

• This includes time for loading fuel elements during the approach to critical experiment. Actual critical time would be about 30% of this value.

III. Ef1ERGEllCY SHUTD0WilS - flone IV. MAlf1TEf!AllCE The Poisian Control Rod and both safety rods were inspected in July.

fio unusual wear or corrosion was noted. The fuel elenents were also inspected in July and showed no excessive signs of corrosion. There were no leaking elements detected.

V. C0flDITI0f15 VilDER SECT 10fl 50.59 of 10 CFR 50 There were no changes to the system or procedures during this period.

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Page 2 VI., VII., VIII. RADI0 ACTIVITY Because of the low power and infrequent use of the LOPRA, its operation does not contribute to the release of effluents. Personnel records for the laboratory are given in the Annual Report for the Advanced TRIGA Reactor, License flo. R-ll5, dated February 19, 1979. (Docket fio. 50-151)

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