ML20148C806
| ML20148C806 | |
| Person / Time | |
|---|---|
| Site: | University of Michigan |
| Issue date: | 10/12/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20148C775 | List: |
| References | |
| NUDOCS 7811020150 | |
| Download: ML20148C806 (7) | |
Text
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[ 43 UNITED STATES
- . 4 NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 s.....,
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 25 TO FACILITY OPERATING LICENSE NO. R-28 THE UNIVERSITY OF MICHIGAN FORD NUCLEAR REACTOR DOCKET NO. 50-2 Introduction By letter dated June 27, 1977, as supplemented June 12 and September 13, 1978(1,4,6), the University of Michigan (UM), proposed changes to Section 5.2 of the Technical Specifications (TS) of Facility Operating License No. R-28. The proposed changes would allow the use of both uranium aluminide (UAlx) fuel and uranium oxide fuel (U3 08), in addition to the uranium-aluminfum (UAl) alloy fuel that is presently being used in the Ford Nuclear Regqtor (FNR), and would specify a fission density limit of 1.5 x 10 3 2fissions /cc.
UM's safety analysis to support its position that either the aluminide or oxide fuel provides a safe alternative to the UA1 alloy fuel for FNR operations presents: 1) data from experimental irradiation tests, 2) operating experience of reactors using fuel elements with aluminide and oxide fuel, 3) an analysis that compares the expected performance of these fuels under FNR operating conditions with results under experi-mental test conditions and operating reactor conditions, and utilizes heat transfer calculations to show that the calculated peak clad fuel plate temperature would be well below the clad failure temperature during nonnal operation. The fission density limit is based on accept-able swelling limits and the results of post-irradiation tests which give an estimate of the conditions at which failure occurred due to fuel blistering.
We reviewed the FNR safety analysis and requested additional information (2,5) be submitted. The FNR response (3,6,8-) was also reviewed.
l Evaluation General The UM's justification for the TS change to use alternative fuel is based on experience with these fuel types which has shown satisfactory and safe performance. This has been experimentally and operationally 33 \\ QUIN
qqs verified in reactors which operate at much higher thermai power densities, higher fission densities, higher heat fluxes and higher coolant flow rates than exist in the FNR. UM has pro-posed to limit fission density for all fuel types in the FNR to 1.5 x 1021 fission /cc since this limit is equal to or below
- operational fission densities reached in other reactors using the same kind of fuel without failures attributed to the fuel, and it is not anticipated that a level above this will ever be required at the FNR.
UM states that the reactor core flux distributions and rod worths '
will be unaffected since the characteristics will be virtually unchanged.
VM states that the experimental data base which supports the safe use of UA1x and U 038 fuel in the FNR was derived from irradiation tests that were performed in the Materials Test Reactor (MTR), the Engineering Test Reactor (ETR) and the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory, Idaho Falls, Idaho, and the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (0RNL), Oak Ridge, Tennessee. Data were also available from aluminide test plates irradiated in the German Karlsruhe 'FR2 reactor. VM states that the operating data were obtained in the MTR, ETR and ATR for aluminide fuels, and in HFIR for U38 0 . VM notes that ATR has operated over 89,000 UAlx fuel plates to depletion with only one failure allowing low level fission product leakage. The failure was found to be due to a fabrication condition that was sub-sequently corrected. VM also notes that the HFIR has operated 76,000 U308 fuel plates to depletion with no failures.
Fuel Core Swelling With an increased fission density limit, a primary safety concern, should this limit be reached, is the disruption of the fuel clad. The clad surrounds the fuel and contains the fission products throughout the service life of the fuel plate. A disruption of the clad could result in the release of fission products.
Disruption of the clad can be brought about by a phenomenon known as
" swel l i ng . " Swelling refers to the effect of increased fission -
product inventory within the fuel which causes the fuel volume to increase. The fission produce inventory increases as a reactor is operated. If the increase in volume is large enough, it could dis-rupt the clad surrounding the fuel to the point of rupture and/or cause the fuel plate to buckle. Fuel plate buckling could cause the coolant channel to be blocked with possible fuel plate melting and release of fission products into the coolant.
UM states that their proposed fission density limit and the use of UA1x and U 038 fuel were selected because these fuels have been tested, without fuel related failures, to the proposed limit under operating conditions similar to those in the FNR.
VM presents the following data to support their position:
- 1. UAl x fuel operating in the temperature range of 158 F-176 F (the range in which the FNR fuel operates) and fission densities of 1.6 to 1.8 x 101 2fissions /cc showed between 4.0 to 8.8%aV/V swelling (swelling is usually denoted as the percent volume change).
- 2. U038 fuel operating at 181'F, which is comgprable to that of the FNR, and a fission density of 1.8 x 10 fission /cc showed a 3.1%aV/V swelling. .
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- 3. UAl alloy fuel irradiated to a fission density of 1.5 x 10 .
fission /cc showed a 9.5%AV/V swelling.
UM states that these results were achieved without any fuel related fail ures. VM also states that the ATR UAlx fuel plates, which have a more stable curved geometry like the FNR, did not experience buckling until they achieved a fission density of 2.0 x 1021 fission /cc with a swelling level of 13%aV/V.
Based on these results, the operating limit of 1.5 x 1021 fission /cc would not create sufficient swelling to disrupt the clad or cause buckling. Therefore, we agree with UM that UAlx and U3 08 fuels are '
acceptable alternatives to use up to the fission density limit of l . 5 x 1021 fission /cc.
Fuel Core Blistering Blistering is a safety concern that must be considered because of the potential to rupture the cladding. In a document (7) referred to by UM, excessive swelling is defined as the production of blisters on the fuel plate surfaces. This document further states that post-irradiation annealing tests were performed to determine the minimum temperature at which blisters form. This test data showed that blisters form at the fuel core-clad interface, thus indicating, on a relative basis, the strength of the fuel-cladding bond.
Utilizing fuel core blister failure test data for UA1x and U 30g fuel from reference 7, UM has shown that, under operating conditions similar to those in the FNR and for burnups higher than the FNR limit (1.5 X 1021 fissions /cc vs.1.7 x 1021 fissions /cc on the tested plates), the temperature for blister formation is greater than 900*F. This temperature is well above the FNR peak operating temperature of 172 F.
A review of operating data reveals that no failures have occurred because of blistering in either the UAlx or U308 fuels.
Based on these results, we agree that UA1x and U 038 fuels can be used in the FNR without failure due to blistering.
Fuel Differences and Specifications UA1 x Fuel VM states that UA1x fuel specifications for the FNRwere developed in concert with the ATR staff at NRTS, Idaho, and Atomics Inter-national, the ATR fuel manufacturers, The FNR fuel specifications were made the same as the ATR UA1x fuel in order to standardize fuel elements among several plate type reactor facilities.
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Some differences exist between the proposed UA1x fuel and the present FNR fuel. The fuel cladding material presently used in the FNR is aluminium of the 1100 series; the proposed fuel will use aluminium of the 6061 series. In general, the two have similar properties except that A16061 is stronger (8000 psi vs 5000 psi yield strength) and the Al 1100 has a higher thet mal conductivity.
The lower thermal conductivity (.43 vs .53) is not significant since, in the heat transfer from fuel meat to coolant, the clad the, mal resistance is a negligible percqntage (1.4%) of the film resistance at the clad-coolant interface (ll.
In standardizing the size of the fuel elements, the cladding for the proposed UAlx and U3 08 fuels was reduced to .015 inches from the present FNR fuel clad thickness of .020 inches. This increases the water channel gap from 0.117 to 0.126 inches. Analysis of these changes in reference 8, shows some slight variation due to increased moderation when using the proposed fuels. We agree that these variations are minor and the effects on reactor physics are not significant.
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The corrosion resist- e properties of the two Aluminium series have similar characte. tics. As the A16061 material contains approximately 1% magnesis content the susceptibility to stress corrosion cracking must be considered. Laboratory tests have shown that stress corrosion cracking can occur when aluminium with a high magnesium content (2-3%) is in direct contact with nickel in water containing in excess of 50 ppm chloride ions.
At the FNR, the fuel elements will not be in contact with any nickel material and periodic water analysis reveals no traceable amounts of nickel ions and only detectable amounts of chloride ions (less than .006 ppm). As the conditions in the FNR are not conducive to stress corrosion cracking, we conclude that A16061 is an acceptable cladding material.
The major differences in the UA1x and the UA1 alloy fuels exist in the preparation of the fuel matrix. Both fuels use high purity aluminium and uranium enriched with U235 The difference is that the present fuel is alloyed directly to form UA1 alloy and the UA1x fuel is forned by mixing a fine powder form of the materials and pressing at pressures of approximately 25 tons per square inch. As the mixing of powders provides a more uniform dispersion of the uranium in the fuel over the alloying process, a more uniform heat generation will be obtained, with a commensurate reduction in the propensity for hot spots.
U308 Ft'l -
The U fuel specifications will be geometrically the same as the above .si x fuel; however, the final specifications will be developed in cc.. tration with ORNL and Brookhaven National Laboratory to ensure standardization should this fuel be used in the FNR, Summary In summary, the margin of safety is not reduced when the proposed UA1x and U 038 fuels are operated in the FNR in accordance with j requirenents established in the TS and limiting the fission density to 1.5 x 1021 fission /cc. The powder metallurgy manufacturing process is considered superior to alloying because of improved dimensional stability during reactor operation and the reduction in fuel hot spots.
We therefore agree that use of the UA1x and U 038 fuels proposed by UM in the FNR is acceptable, 1
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Operating the FNR with UA1 Alloy Fuel to the Proposed Fission Density Limit UM's TS do not specify a fission density limiting condition for operation for the UAl alloy fuel in the FNR. Establishing a limit of 1.5 X 1021 fissions /cc is therefore conservative. ,
An analysis was never performed to establish a fission density limit for this fuel because FNR can only achieve about 0.5b x 1021 fissions /cc with the present U235 loading. FNR has operated at this level without fuel element failures or other safety problems. We also have infor- .
mation from the General Electric Test Reactor, that also uses the UAl a1{gy fuel, that 10 3 fissions it has operated
/cc without at fission densities up to 2,0 x any failures.
2I We agree that establishing a fission density limit of 1.5 x 10 fissions /cc for UA1 alloy fuel is conservative and therefore, acceptable.
Environmental Consideration We have determined that this amendment will not result in any significant environmental impact and that it does not constitute a major Commission action significantly affecting the quality of the human environment. We he/e also determined that this action is not one of those covered by 10 CFR E51.5(a) or (b). Having made these determinations, we have further concluded that, pursuant to 10 CFR 551.5(d)(4), an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion a We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a sinnificant decrease in a safety margin, the - '
amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public A will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: October 12, 1978
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1 References
- 1. " Safety Analysis - Utilization of Intermetallic Uranium Aluminida (UA1, 3 UA1,4 UAl )2 and Uranium 0xide (U 0 38, UO )'2 Cermet Fuel Cores in the Ford Nuclear Reactor," June 1977, transmitted by letter from W. Kerr, University of Michigan, to P. O'Connor, NRC dated 6/27/77.
- 2. " Request for Information on the University of Michigan Request that the Ford Nuclear Reactor License No. R-28 be Amended to Utilize Aluminide and 0xide Fuels," transmitted by letter from R. W. Reid, NRC to W. Kerr, University of Michigan dated 4/18/78.
- 3. " Response to Questions Contained in Information on the University of I Michigan Request that the Ford Nuclear Reactor License No. R-28 be Amended to Utilize Aluninide and 0xide Fuels," June 1978, trans-mitted by letter from W. Kerr, University of Michigan, to R. W. Reid, NRC dated 6/12/78.
- 4. " Safety Analysis - Utilization of Intermetallic Uranium Aluminide (UA1,3 UA1,4 UAl )2 and Uranium 0xide (U 08, 3 U0 2) Cermet Fuel Cores in the Ford Nuclear Reactor," June 1977, Revised 4/20/78, trans-mitted by letter from W. Kerr, University of Michigan to R. W. Reid, NRC dated 6/12/78.
- 5. Conference telecon with R. R. Burn, University of Michigan on 8/25/78.
- 6. " Supplementary Information Relating to Utilization of Intermetallic Uranium Aluminide (UA1, 3 UA1, 4 UA12 ) and Uranium 0xide (U 38 0 ) Cermet Fuel Cores in the Ford Nuclear Reactor," September 1978, transmitted by letter from W. Kerr, University of Michigan, to S. Ramos, NRC dated 9/13/78.
- 7. G. W. Gibson, "The Development of Powdered Uranium Aluminide Compounds for Use as Nuclear Reactor Fuels," IN-ll33, TID-4500, December 1967, page 40.
- 8. Additional Information Relating to " Limiting Conditions for Operations and Surveillance Requirements" and " Comparison of Al 1100 and A16061 Properties," 9/28/78, from W. Kerr, University of Michigan to S. Ramos, NRC.
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