ML19322E497
| ML19322E497 | |
| Person / Time | |
|---|---|
| Site: | National Bureau of Standards Reactor |
| Issue date: | 03/24/1980 |
| From: | Rozier Carter NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL |
| To: | |
| References | |
| NUDOCS 8003280242 | |
| Download: ML19322E497 (4) | |
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,_Y_,I' UNITED STATES DEPARTMENT OF COMMERCE
- N:tianzl Eureau cf St:ndirda
('4,ee/c8,'washington
- D.C. 20234
, March 24, 1980
!!r. Robert Reid
-Chief, Operating Reactors Branch No. 4 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, DC 20555 Subj ect: Request for a change in the NBSR Technical Specifications Decket No. 50-184
Dear Mr. Reid:
NBS respectfully requests a change to Section 6.3 of the NBSR Technical Specifications, License No. TR-5. The requested change is to delete the word " alloy" that describes the construction of the fuel
, plates in the last sentence of the section. This will enable NBS to use
! dispersion fabrication methods in the construction of fuel plates.
j Dispersion techniques are considered equal to or superior to alloy and utilize aluminide or oxide as the fuel matrix. Other than this fabricati m change, all other conditions including the basis, will remain the same.
A discussion in support of the requested change is enclosed.
Sincerely, 6
Robert S. Carter Chief, Reactor Radiation Division Enclosure l Subscribed and Sworn to before me this f5 day of March 1980.
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1- Use of Dispersion Fabrication Techniques for Fuel
-Elements at the NBSR
- 1. Introduction
.The NBS reactor (NBSR) has been using uranium-aluminum alloy fuel elements since start-up. They have given very satisfactory performance, but now they are no longer being produced by commercial manufacturers in the U.S. Therefore, it has become necessary for many research reactors, including the NBSR, to obtain fuel assembled using dispersion type fuel cores. The performance history of dispersion fuels has been excellent and dispersion fuel is generally considered' equal to or superior to alloy fuels. For nearly 20 years, substantially more than 100,000 dispersion plates have been
, used with outstanding results and under conditions more severe than
- those existing at the NBSR. Among major reactors that have used, or are planning to use dispersion fuels are the HFIR, ATR, HFBR and
- ORR. The latter two with power levels of 40 and 30 Mw respectfully have been using alloy fuels similar to that used at the NBSR and are switching to dispersion for the same reasons as the NBSR.
! 2. General Considerations and Specifications
, There are no basic differences in the overall specifications for dispersion elements to be used at the NBSR. from alloy elements
. currently used. In all instances, dimensional, configuration, loading, fabrication control and quality assurance will be specified essentially the same for dispersion as they would be for alloy, with only minor adjustments to meet f abrication requirements.
- Because of generally better fabrication controls exercised by the dispersion manufacturers, a better, more uniform element is likely.
j This is confirmed by extensive past experience of several reactors where fewer rejects and very little or no failures were experienced.
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. All consideratious involving metallurgical and mechanical aspects of dispersion elements are equal to or superior to alloy elements.
- 3. Thermal Considerations i
Since the dimensions and configuration of the dispersion j
plates can be made essentially identical to those made by the alloy process, there are no basic differences in the overall thermal '
performance relative to heat removal at the surface. Because of j
the better process control that can be exercised during the manufacture of dispersion fuel 3 these fuel cores are less likely to exhibit significant areas of non-bonding at the fuel-cladding interface than alloy cores. Measurements of the values of thermal conductivity for several' dispersion cores showed them to be slightly higher than alloy cores for the same fuel loading.
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- 4. Reactor Physica No reactor physics changen will occur as a renuit of the conversion f rom alloy to dispersion fuel neat, since all Mher conditionn and parametern remain virtually unghanged.
- 5. Physical Behavior Under Irradiation Test results at. ORNI. and NRTS to burnupn of 2.3 x 10 I f isalon/cm and higher, Indicate that dinpernlons perform quite well. Radiat. ion nwelling was found to be quite small and actually decreased with increaned loading. Pontirradiation examination showed no Indication of blisters, core-cladding neparation, matrix cracking, or other typer of structural def ect s. Dgect.ed peak jlssion density at the NBSR will be less than 2.2 x 10 finnion/cm' (annuming 80% peak burnup and a maximum loading of 360g. U-235), which is below that achievcd in t.ents or actual operation of reactors uning dinpersion fuel. In any event, the fission density would be the same for dispercion as for alloy.
Perhaps one of the major advantages of dispersion fuels in their great retention of fission products. Tests Indicate that the fission product release f rom grana cladding failure would he considerably lean in the case of diapernion than in the case of the alloy.
- 6. Conclusions Based on the foregoing, it may be concluded that plate-type 4 fuel assembled using U-Al or II 3g 0 -Al Mnpersions In equal to or superior to U-Al alloy fuel and can be used at the MBSR without:
incurring any reduction in safcLy marginn.
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- ~3-Bibliography
- 1. M. M. Martin et al. , " Irradiation Ischavior of Aluminum-nane Fuel Dispersions," ORNL 4856 (1973).
- 2. F. T. Ilinford, "The Oak Ridge Research Reactor-Safety Analysis,"
ORNL-4169/V2/S1 (1968) .
- 3. F. T. Bindord and E. N. Cramer, "The 111gh Flux Isotope Reactor,"
ORNL-2572 (1964).
- 4. C. W. Gibson, "The Development of Powdered Uranium Aluminide Compounds for Uce as Nuclear Reactor Fuels," In-1133, TID-4500 (1967).
- 5. "IIFBR-131 Fuel Elements," Series of Memoranda - Brookhaven National Laboratory (1975-1978).
- 6. A. E. Richt et al., "Postirradiation Examination and Evaluation of the Performance of IIFIR Fuel Elements," ORNL-4714 (1971) .
- 7. V. A. Walker et al., "ATR Fuel Ibterials Development Irradiation Results," IDO-37157 (1966).
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