ML20214Q621
| ML20214Q621 | |
| Person / Time | |
|---|---|
| Site: | Rensselaer Polytechnic Institute |
| Issue date: | 04/30/1987 |
| From: | Hobbins R EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20214Q599 | List: |
| References | |
| CON-FIN-D-6010 EGG-NTA-7660, NUDOCS 8706050093 | |
| Download: ML20214Q621 (13) | |
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EGG-NTA-7660 April 1987 INFORMAL REPORT Idaho TECHNICAL EVALUATION REPORT FOR THE EVALUATION OF Eng/neering THE SPERT FUEL AT RENSSELAER POLYTECHNIC INSTITUTE Laboratory t
Managed by tho U.S.
Department R. R. Hobbins ofEnergy l
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d DISCLAIMER This book was prepared as an account of work toontored by an agency of the United States Government Noiteer the United States Government not any agency thereof, not any of troer employees. makes any warranty, esprest or imobed, or assumet any legal habikty of retoons,bekty for the accuracy, completenett, of usefulness of any information, accaratus, product or procent discioned, or reoreseets that ett use wovid ect infringe pnvately owned rignts flo erences Perein to any nosofic commercsal r
product, procent, or terwce by traos name. trademark, tranufact rer, of otherwise, daet not necettenly coast.tute or imply its endorsement, recommendation, or favonng
- by the United States Goverament or any agency thereof The vie *t and opinions of authort espretted here*n do not necettar ly state of reftect thoto of the Umted States l
Government or any agency thereof l
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EGG-NTA-7660 i
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TECHNICAL EVALUATION REPORT FOR THE EVALUATION OF THE SPERT FUEL AT RENSSELAER POLYTECHNIC INSTITUTE i
R. R. Hobbins i
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Published April 1987 l
1 EG&G Idaho, Inc.
Prepared for the U.S. Nucledr Royulatory Conaission Washington, D.C.
20555 Under 001 Contract No. DE-AC07 761001570 FIN No. 06010 i
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ABSTRACT This report evaluates the suitability of the SPERT fuel for the intended use at Rensselaer Polytechnic Institute (RPI).
The requalification of SPERT fuel was performed by Argonne National Laboratory, i
to verify that the rods have suffered no physical damage since fabrication. Rods were inspected under 6X magnification, and by l
X-radiographic, destructive, and metallographic examinations.
Spectrograph)c and chemical analyses were performed on the U0 fuel.
In p
conclusion, based on requalification studies and the nature of the intended use, using SPIRT fuel rods in the RP! critical facility does not constitute j
an undue safety risk.
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iIN No. D6010.. Casework and Non Power Reactor Reviews l
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FOREWORD This report is supplied as part of the Non-Power Reactor (NPR) Reviews associated with using low enriched urantum (LEU) fuel in university reactors.
The review is being conducted by the Idaho National 'ingineering Laboratory for the U.S. Nuclear Regulatory Comission. Of fice of Nuclear Reactor Regulation.
l The U.S. Nuclear Regulatory Comission funded this work under the authorization of f!N. No. 06010, Casework and Non-Power Reactor Reviews.
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CONTENTS ABSTRACT..............................................................
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FOREWORD..............................................................
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INTRODUCTION..........................................................
1 REQUALIFICATION OF SPERT fuel.........................................
2 SUliABILITY Of SPERT fuel FOR USE IN RPI CRITICAL!iY FACILITY.........
5 REFERENCES............................................................
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TECHNICAL EVALUATION REPORT FOR THE EVALUATION OF THE SPERT FUEL AT RENSSELAER POLYTECHNIC INSTITUTE INTRODUCTION O
Rensselaer Polytechnic Institute (RPI) is planning to use stainless steel-cald, low-enriched UO fuel rods, produced in the early-to-mid 2
l'J60s for use in the Special Power Excursion Reactor Test (SPERT) program, to convert its critical fac111ty to utilire low enriched urantum (LEU) fuel.
This purpose of this document is to evaluate the suitability of the SPERT fuel for the intended use at RPI.
The critical facility at RPI is limited to an operating power level of 100W and operates only a few hours per week.
Fuel and control rods are repositioned frequently; neutron flux measurements are mJde for comparison with criticality calculations.
The tank-type reactor has poison rods both for control and scram.
The design basis accident is failure of fuel-rod cladding and release of the fission product gap inventory.
Fuel rod failure by cladding overheating or overpressure is not credible due to the extremely low power operation of the reactor.
Two credible rod failure mechanisms are corrosion and mechanical damage due to mishandling.
The requalification of the SPERT fuel rods and the suitabillity of these rods for use in the RPI reactor will be discussed.
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REQUALIFICATION OF SPERT FUEL Requalification of SPERT fuel rods for use in university reactors was carried out by Argonne National Laboratory.
The 600 rods intended for use in the RPI critical facility were inspected.
These rods, never used in a reactor, have been in air-conditioned storage at Purdue University since 1974.
The inspection results should generally be applicable to the entire production run (about 9000 rods); the rods inspected cover virtually the entire range of serial numbers of rods produced.
The rods originally were procured according to Phillips Specification No. F-1-SPT, which incorporates Phillips Drawing No. SPT-E-1166.
The component materials were required to meet applicable ASTM standards, and extensive tests and inspections were required for components and the finished rods.
In particular, all rods were to be inspected for dimensions and surface condition, helluto leak tested to ensure the integrity of the welds (the rods were filled with helium at the time of welding), and gamma scanned to check the fuel zone length and detect the presence of any foreign materials in the fuel zone.
However, it appears that all fabrication, inspection,and acceptance records have been discarded.
The requalification program's purpose is to verify that the rods are those procured to Specification No. F-1-SPT, and that the rods have suffered no physical damage since fabrication.
All 600 rods were checked for straightness and examined under 6X magnification for nicks, scratches, and/or other damage to the cladding surface.
Thirty rods were measured to check diameter and roundness.
All rods appeared to be in excellent condition and met the dimensional and surface condition requirements of the specifications, except (possibly) for the diameter in the end cap welds that, on the average, is 0.0041 in.
(0.10 nun) largnr than the maximum dimension for the rod diameter given on the specification drawing.
Sixty rods wore randomly selected among the representative groups of ser tal numbers for X radiographic examination of the upper and lower end cap wolds.
Dofects were found in the upper end cap welds of six of the i
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rods.
The X-radiography examination found minimum wall thickness in the defects to vary from 0.005 to 0.015 in. (nominal cladding wall thickness is 0.020 in.).
Metallographic examination of one of the weld defects revealed it probably was caused by a gas bubble.
Although this particular defect was not connected to the interior volume of the fuel rod, radiographs of a
other rods showing similar defects indicate that some of the defects are probably connected to the interior volume.
The internal pressure, void volume, and fill gas composition was measured in five rods chosen for destructive examination, plus the rod whose weld defect was examined metallographically. All six rods had a positive pressure of fill gas, ranging from 0.6 to 3.3 psig.
For comparison, the specification for fill gas was 1 psig of helium.
The fill gas was found to be predominately nelium, but a sizable amount of hydrogen was also found (up to 16%).
Trace amounts of water vapor and nitrogen were measured, although in one rod about 1% nitrogen and a few milligrams of water were found.
The hydrogen probably resulted from the reaction of water vapor with the fuel and the cladding.
Less than 2 mg of water is required to produce the amounts of hydrogen measured in the fill gas; the hydrogen is responsible for the overpressure in the rods.
The specification allows up to 75 ppm water in a fuel rod, which corresponds to about 60 mg.
The minor deviations in the fill gas composition and pressure relative to the specifications have no significance for the use of these 4
rods in the Rpl critical facility.
The entire stack of 60 fuel pellets was examined from 2 rods, and the top 6 pellets were examined from 2 other rods.
All pellets examined, with three exceptions, had only minor surface chips and were judged to meet the pellet surface condition requirements of specification F-1-SpT.
Three pellets in one rod each had a significant piece (0.2, 0.2, and 0.7 g, respectively) spalled off the entire length of the pellet.
The missing material was contained in loose fragments and powder collected after all the pellets were removed from the rod.
The length, diameter, and weight of each of the 132 pellets removed from the 4 rods were measured and the pellet density calculated based on solid, right, cylindrical geometry.
Neglecting the 3 chipped pellets, 16 pellets were found with densities 3
3 outside the specification of 9.97 g/cm.
Excluding these four pellets and the three chipped pellets, the mean pellet density was 10.078 + 0.055 g/cm.
Twelve pellets in one rod had densities more than 0.1 g/cm above the mean density, the largest of which was 0.15 g/cm higher than the mean.
Deviations of this magreitude from the specification for pellet density have no safety significance for the intended use of the fuel in the RpI critical facility.
Three pellets, one from each of three rods, were sectioned and examined metallographically.
The microstructures were similar in the three pellets and were relatively fine grained (5 to 10m) U0 with some 2
porosity and, possibly, some 0 0 present.
The structures are fairly 47 typical as-fabricated, unirradiated U0 f""1' 2
Analyses of the U0 fuel w re performed for isotopic, total uranium, 2
and for impurities.
Spectrographic analysis for 20 elements revealed an impurity content of <185 ppm, which is only about 5% of the allowable level; however, a number of possible significant elements were not analyzed. An upper limit value of oxygen uranium ratto calculated based on the measured uranium content, measured impurity content, and the assumption that the remaining sample weight must be oxygen is 2.04.
An additional impurity content of 1200 ppm, which would be well within the specificatiori, would result in an oxygen / uranium ration of 2.02, which is the allowable upper bound.
Metallographic examination of the fuel rod cladding showed the cladding to be within specification for wall thickness; to be seamless, as specified; and to have a microstructure typical of 304 stainless steel with some evidence of normal carbide precipitation, but no evidence of intergranular attack or corrosion from either the inside or outside surfaces.
Chemical analysis showed that the metallic constituents of the stainless steel were all within the specification with the exception of
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cobalt, which was 0.084 wt'A compared to a maximum allowable of 0.05 wt*/..
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SUITABILITY OF SPERT FUEL FOR USE IN RPI CRITICAL FACILITY The extremely low power operation of the RPI critical facility (100W maximum) and its minimal operation (a few hours per week) ensure that the fuel rods are very unlikely to fail by overheating or overpressure.
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this context, the minor deviations from specifications of rod internal pressure, fuel pellet density, and, perhaps, fuel oxygen / uranium ratio, are without significance from a safety standpoint.
In addition, the lack of intergranular attack or corrosion on the cladding during storage and the basic conformance to the specifications of the stainless-steel cladding, suggests that corrosion is an unlikely failure mechanism given reJs3nable water chemistry control. The only rod failure mechanism of any likelihood would appear to be a mechanical failure at a thin-wail defact at an upper end cap weld due to mishandling during rod repositioning. However, mechanically failing a rod at this very localized, small region would appear to be small. The probability of rod failure can be minimized by l
handling tools and procedures designed to avoid placing strong mechanical loads in the region of the upper end cap welds where thin-wall defects may exist.
It is concluded, based on requalification studies and the nature of the intended use, that using the SPERT fuel rods in the RPI critical facility does not constitute an undue safety risk.
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REFERENCES 1.
J. L. Snelgrove, R. F. Domagala, L. R. Dates, Reaualification of S2ERT i
Fuel Pins for Use in University Reactors, ANL/RERIT/TM-8, December 1986.
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BIBLIOGRAPHIC DATA SHEET EGG-NTA-7660 sie INstmuCTIONS ON T.e mgytast 3 YtTLE AND SbSTITLE JktAVGSLANK TECHNICAL EVALUATION REPORT FOR THE EVALUATION OF THE SPERT FUEL AT RENSSELAER POLYTECHNIC INSTITUTE e oaf t mePonT COMPLETED esONTN vtAR April 1987
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MoNTN April 1987 7 5EASOmwiNG oR"4AN12ATtoN NAMt ANO MAILING Accatss asaeswesle cases e PAOatCTITASE/upOmz 4,8eif NutIGER EG&G Idaho, Inc.
P. O. Box 1625
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Idaho Falls, ID 83415 06010 10 SPONsomeNG onGANt2 AT60N News AND MAsLING A00 mess tsar #wante cases its TYPt CP REPORT Standardization & Non-Power Reactor Projects Informal Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 1, SUPPL 4WENT ARY NOTis 1J LSsTR ACT d200 weres e, *esas This report evaluates the suitability of the SPERT fuel for the intended use at Rensselaer Polytechnic Institute (RPI). The requalification of SPERT fuel was performed by Argonne National Laboratory, to verify that the rods have suffered no physical damage since fabrication.
Rods were inspected under 6X magnification, and by X-radiographic, l
destructive, and metallographic examinations.
Spectrographic and chemical analyses were performed on the U02 fuel.
In conclusion, based on requalification studies and the nature of the intended use, using SPERT fuel rods in the RPI critical facility does not con:titute an undue safety risk.
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