ML20033F290
| ML20033F290 | |
| Person / Time | |
|---|---|
| Site: | 05000199 |
| Issue date: | 03/12/1990 |
| From: | Office of Nuclear Reactor Regulation |
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| References | |
| NUDOCS 9003190297 | |
| Download: ML20033F290 (10) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
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WASHINGTON, D C. 20555
,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CONVERSION ORDER TO CONVERT FROM HIGH-ENRICHED TO LOW-TNRICHED URANIUM FUEL FACILITY OPERATING LICENSE NO. R-94 MANHATTAN COLLEGE DOCKET NO. 50-199
1.0 INTRODUCTION
In a cordance with 10 CFR 50.64 which. requires that non-power reactors convert to a lou-enriched uranium (LEU), fuel, except under certain conditions,
' Manhattan College (MC or Licensee) has proposed to convert the fuel in its
. pool-type training reactor (the reactor) from high-enriched uranium (HEU) to LEU. fic submitted a Safety Analysis Report (SAR) and revised Technical Specifications (TS)dealingwiththechangesneededtoconverttoLEUfuelon May 19, 1989. The :;taff's safety review with respect to issuing an order for conversion from HEU to LEU fuel has been based on an analysis of MC's SAR and the~ proposed TS as well as information provided by MC on August 10 and December 12, 1989 ir, response-to staff questions. This material is available for review at the Commission's Public Document Room at 2120 L Street, N.W.,
Washington, D.C.
This Safety Evaluation (SE) was prepared by T. S. Michaels,
' Project Manager, Division of Reactor Projects III, IV, V and Special Projects, l
Office of Nuclear' Reactor Regulation, U.S. Nuclear Regulatory Commission.
M6jor contributors to the technical _ review include W. R. Carpenter and R. W.
Carter of EG&G, Idaho National Engineering Laboratory (INEL).
L In addition to the changes associated with the HEU to LEU conversion some minor TS changes, which do not pertain to the conversion, were made to update and clarify the intent of the TS.
2.0. EVALUATION 2.1 General The LEU fuel that will be used to replace the HEU fuel is uranium silicide -
aluminum dispersion fuel (U Sh -A1), and was developed by the Argonne National 3
Laboratory (ANL), under the sponsorship of the Department of Energy (DOE),
especially.for use in the DOE HEU-LEU conversion program. A LEU fuel to the Manhattan College Zero Power Reactor (MCZPR)pplication of the was developed by
'theANL(Reference 2).
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The MCZPR is currently licensed for operation at thermal power levels not to exceed 0.1 Watt (thermal), using thin plate-type fuel, and is cooled by natural thermal convection of the pool water. The Licensee has proposed no changes to any reactor system or operating characteristics except for the replacement of the HEU fuel elements by new LEU fuel elecents. The following evaluations and conclusions are based on that assumption.
2.2 Fuel Construction and Geometry Both the HEU and LEU cores at the MCZPR consist of 15 standard (fully loaded) fuel elements and 1 partially loaded fuel element. The purpose of the partially loaded fuel element is to allow for precise core excess reactivity control.
The HEU standard fuel elements currently installed at the MCZPR consist of 6 concentric cylinders formed by mechanically joining and positioning 18 curved fuel plates within the grooves of 3 aluminum spacer webs that are located within an aluminum support cylinder. The LEU standard fuel elements have essentially the same design as the HEU standard fuel elements, but the fuel meat contains LEU U,hich is enriched to 92% U-235.Si, -Al fuel enriched to less than 20% U-235 instea fuel W There is a difference in the procedure for securing the curved plates in the fuel element assembly but that has been evaluated and found acceptable. The geometries, materials and iissile loadings of the current HEU fuel elements and the replacement LEU fuel elements are described in Table 1.
A schematic cross section comparing the placement of the HEU and LEU fuel in tha MCZPR grid is shown in Figure 1.
Note, the only change is that the fuel element located in position 46 in the HEU core has been moved ta position 14 in the LEU core. This was done to enhance the Regulating Rod worth. The external dimensions and structural materials of both types of elements are identical, except that the LEU elements utilize 6061 Al instead of 1100 (or 25) A1.
Fuel elements with 6061 Al plates and uranium composition essentially identical with the proposed MCZPR plates were developed by ANL especially for the U.S. non power reactor fuel conversion program and reviewed and approved by the NRC. These fuel elements have been tested extensively under extremely adverse conditions in the Oak Ridge Research Reactor (ORR) with no failures having a safety significance (Reference 4).
Both the HEU and LEU partially loaded fuel elements are identical to the respective fully loaded standard fuel elements in size, geometry, and construction, the only difference being the amount of fissile material located in the six fuel cylinders. The HEU 3artial fuel element contains three fueled plates only in cylinder number 2 wit 1 the remaining five cylinders containing secured aluminum dummy plates.
In order to provide flexibility for fine adjustment of the actual excess reactivity when the LEU core is started up, DOE intends to supply Manhattan College with one LEU partial fuel element having removable fuel plates in cylinder numbers 2, 4, and 6.
Cylinder numbers 1, 3, and 5 will contain secured aluminum dummy plates. Such use of this partial fuel elenent in the MCZPR has been considered by the staff and judged acceptable.
Each HEU fuel element has a hold-down rod constructed from a lucite rod and aluminum end fittings.
Each rod passes axially through the center of a fuel element and is threaded into the grid plate.
The portion of the hold-down rod
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Table 1. Comparison of the HEU and LEU Fuel Elements at Manhattan Coll,ege j
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Number of Cylindars/ Element 6
6 Number of Plates / Cylinder 3
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Number of Plates / Element 18 18 Number of Standard-Fuel Elements 15 15 Number of Partial Fuel Elements 1
1 Fissile. Loading / Cylinder, g U 235 Cylinder I-16.8 19.7 Cylinder 2 24.0 27.4 Cylinder 3 29.4 35.2 Cylinder 4 36.9 43.7 Cylinder 5 43.2 50.6 Cylinder 6 49.7 58.4 Fissle Loanding/ Element g U-235 200 235 Uranim Density g/cc 0.7 4.8 Enrichment, %
92.0 19.7510.2
- Fuel' Meat Composition U-Al-Alloy U 51 -Al 3 2 Cladding Material 1100 A1 6061 A1
- Fuel Heat Dimensions Thickness, mm 0.51 0.51 Width;
, mm Variable Variable Length
, mm 610 572-610 h
Cladding Thickness, mm 0.38 0.38 Critical Mass, g U-235 3024 3552 1
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Figure 1.
Comparison of the Fue' Element Placement in the HEU and LEU Cores at Manhattan College.
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that passes through the active length of the fuel element is a solid lucite rod one inch in diameter. These same hold-down rods will be used to secure the LEU fuel elements to the grid plate.
2.3 Fuel Storage All ~5 'ull fuel elements and the one partial fuel element are permanently mout,t d. on the reactor grid plate.
Occasionally, during periods of tank cleaning and maintenance, the fuel elements are removed from the reactor and placed individually into cylindrical sleeves, and then placed four to a container in the original SYLCOR shipping containers in which the fuel elements were received. This temporary storage procedure is documented in MCZPR records and in the August 1983 SAR. During the process of the HEb/ LEU core conversion, each HEU fuel element will be removed from the core and packaged into these special SYLCOR shipping casks for disposal at the DOE repository site. All of the HEU fuel will be removed from the core and stored in these criticality safe shipping casks prior to the addition of any LEU fuel. Thus, any criticality concerns associated with having both HEU and LEU fuel in the core or in storage at the MCZPR facility concurrently are averted.
2.4 Critical Operating Masses of U-235 Both the MCZPR HEU and LEU cores contain 15 full fuel elements and 1 partial fuel element. The geometry and dimensions of the HEU and LEU fuel elements are identical; however, the core arrangement is slightly different in that the location of one peri Regulating (control)pheral fuel element has been shifted closer to the rod in the LEU core in order to increase the reguhting rod worth. The critical mass of the HEU core is about 3.02 Kg U-235;.tne critical mass of the LEU core is predicted to be about 3.55 Kg U-235. This increased LEU Uranium-235 loading is necessary to compensate for the large iacrease in concentration of Uranium-238, which absorbs low energy (thermal) end epithermal (resonance) energy neutrons, and causes a hardening of the thermal neutron spectrum. The uranium-increase is achieved by increasing its concentration in the fuel matrix.
The proposed concentration is similar to that successfully achieved and tested in the Oak Ridge Research Reactor (ORR).
The MCZPR SAR with the attached ANL report and the MCZPR reply to the HEU/ LEU conversion questions discusses sensitivity calculations of reactivity and a comparison of the calculated values of neutron lifetime, the effective delayed neutron fraction, and core excess reactivity. This documentation also presents the results of studies where the fuel element U-235 loading and Boron 10 (impurity) was calculated and compared for both the HEU and LEU cores.
In all cases, the reactivity effects of these material changes in the fuel plates between the HEU and LEU. cores is as expected when the slightly different arrangement of the fuel between the HEU core and the LEU core is factored into the comparison. With regard to the comparison of the prompt neutron lifetime and effective delayed neutron fraction, the currently calculated values for the HEU core and the LEU core are very similar. The lifetime is slightly shorter for the LEU core, as would be expected, because of the increased neutron absorption in the Uranium-238 and somewhat harder neutron spectrum of the LEU core. The delay fraction and the core excess reactivity between the HEU and LEU
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1 cores (with the same number of fuel elements) is the same. The staff concludes that.the results contained in the above documentation adequately demonstrate the basic neutronic similarity between the HEU and LEU cores at the MCZPR.
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2.5 Hydraulics and Thermal bydraulics 1
The water in the reactor pool has an enormous heat capacity relative to the 0.1 Watt power rating of the MCZPR, therefore no recirculating cooling system is provided. Cooling of the HEU core is accomplished through natural circulation of the water. The totti heat capacity of the coolant water is about 65 MW-sec/'C.
Replacement of the HEU fuel with the LEU fuel is a thermal hydraulic like-for-like replacement since the LEU fuel is nearly identical geometrically, no hardware modifications are being made on the grid or core support, and the hold down mechanism is the same. Whatever minimal thermal hydraulic effects occur in the HEU core will be the same in the LEU core since the heat generation per fuel plate, fuel plate surface heat transfer area, and in-core coolant volume are virtually the same between the two cores.
2.6 Power Density and Power Peaking Power densities and power density peaking, including both nuclear and engineering factors, were computed by ANL. The power distribution among the fuel elements is essentially the same for the HEU and LEU cores with the LEU core being slightly more peaked due to the change in the fuel element placement.
The total peaking-factor increases from 2.55 to 2.59 and shifts to an adjacent fuel element.
In a theoretical sense the slightly higher peaking does reduce margins and would be a consideration if the MCZPR were operating at power levels sufficient to cause the coolant temperature to be near the onset of Departure from Nucleate Boiling (DNB) in the hot channel. For the actual case of the MCZPR operatir.g at 0.1 Watt, the thermal hydraulic safety margins are very high and the slight increase in the peaking factor of the LEU core does not, in fact, represent any significant safety risk.
Furthermore, the higher peaking in the LEU core was inclucW in the calculation of the inadvertent transient accident discussed in section 2.12.1 of this safety evaluation report.
2.7 Control Rod Worths 4
The MCZPR has two control rods to regulate core reactivity:
a shim rod which uses cadmium as the neutron absorbing poison and a regulating rod which uses stainless steel as the neutron poison. The same two control rods as used for the HEU core along with all their associated hardware, drives, mounts, etc. are used for the LEU core. The calculated worths for the HEU core are -3.5% delta k/k' for the shim rod and -1.2% delta k/k for the regulating rod.
For the LEU core the calculated worths are -3.4% delta k/k and -1.3% delta k/k respectively.
The worths are somewhat different for the LEU core, partly due to the different core arrangement which was made to enhance the regulating rod worth. The measured rod worths with the HEU core are -2.5% delta k/k and -0.9% delta k/k for the shim and regulating rods, respectively. Therefore, it is predicted that the worths of the control rods in the LEU core will remain approximately the same as in the HEU core and will remain acceptable.
t 9-2.8 Shutdom Mergin It is the policy of the NRC to require that there be reasonable assurance that a non-power reactor can be shut down from any operating condition, even if the control / safety rod of maximum worth is in its most reactive position (fully withdrawn). On the basis of the computed control rod worths and the authorized excess reactivity, the Manhattan College reactor would be subcritical by ) fully approximately 0.56% delta k/k with the rod of maximum worth (the shim rod withdrawn. This 1s larger than the Technical Specification margin of at least 0.46% delta k/k, and is acceptable.
2.9 Excess Resetivity Additional reactivity above cold, clean critical is required to allow a reactor to perform programmatic and academic functions. The Licensee's submittal discussed and presented calculated changes in reactivity caused by varying the U-235 loading of the partial fuel element, which is the most expeditious way for the licensee to adjust core excess reactivity. The calculations indicate there is reasonable assurance that the excess reactivity permitted by the Technical Specifications, which is the same for the LEU and HEU cores, can be achieved.
Because the authorized maximum excess is only 0.44% delta k/k, an inadvertent insertion of all this excess will not allow the reactor to become prompt critical so any plausible transient power increase would be quickly terminated by a power level scram or operator intervention and would result in increased fuel temperatures only a few degrees C, which is acceptable for both the HEU and the proposed LEU cores.
2.10 Reactivity Feedback Coefficients The temperature coefficient of reactivity and the void-coefficient of reactivity were recently computed by ANL for both the HEU and LEU cores. Both coefficients
'are more negative than required by the Technical Specifications. The void coefficient of the LEU core is slightly more negative than for the HEU core.
-0.16% delta k/k versus -0.15% delta k/k per percent void. The calculated isothermal temperature coefficient of the LEU core is also slightly more negative than in the HEU core. Both the HEU and LEU cores have calculated isothermal temperature coefficients which are negative over the range investigated, 20*C to 60*C. The measured temperature coefficient for the HEU core, however, is slightly. positive from 20*C to 43.7'C and then negative at higher temperatures. For this reason the maximum excess reactivity of the HEU core is currently expressed as 0.44% delta k/k at 110.6 (43.7'C). The amount of reactivity added from 20*C to 43.7*C in the HEU core is about 0.12$ and there is no. reason to believe the LEU core will not exhibit a similar phenomena because of its physical and neutronic similarity. Zero power physics testing for the LEU core will determine both the isothermal temperature corresponding to maximum reactivity and the total amount of reactivity added to this point prior to loading the core up to its authorized maximum excess (under any temperature condition) of 0.44% delta k/k.
Although a slightly positive temperature coefficient over a small temperature range is not desirable, the staff concludes the conversion of the MCZPR to LEU does not pose an increased safety risk over the HEU core for two reasons.
First, the calculations show the temperature coefficient for the LEU core is
more negative over the entire range than the HEU core and it is expected the zero power physics measurements will confirm these calculations.
Second, part of the_ negative temperature coefficient for the LEU core at higher temperatures is cauced by Doppler broadening of the U-238 resonances which the HEU core does not exhibit. This effect-is always negative and for the case of the LEU core with a maximum excess of 0.447 delta k/k, Doppler alone would quench all the available excess at a fuol temperature of about 350'C (neglecting any moderator effects), which is well below the threshold of any fuel damage.
2.11 Fission Product Inventory and Containment The total inventory of fission products from operation at 0.1 Watt is very low and will not be significantly different between the MCZPR HEU and LEU cores.
Furthermore, since the number of fuel elements, the )lates per element, and the power distribution are virtually identical between t1e HEU and LEU cores, the-fission product distribution will be very similar between the two cores. The only slight change between the HEU and LEU fuels is in the fission product barrier, the clad. Although the clad thickness is the same, the HEU is fabricated of 1100 aluminum cladding and the LEU is fabricated of 6061 aluminum cladding. The staff judges this to pose no additional risk since the U, 512 -Al fuel developed by DOE and exten:;ively tested in the ORR was clad with 6061 aluminum and experienced no failures attributable to this cladding material (Reference 4).
It is thus concluded there is reasonable assurance that the new LEU fuel will perform satisfactorily in containing fission products under normal operating conditions in the MCZPR.
2.12 Potential Accident Scenarios Among the various potential accidents considered by the Licensee or the staff ac the time of the 1983 license renewal for MCZPR, only two could be affected by the conversion from HEU to LEU fuel. These two scenarios are addressed below.
2.12.1 Inadvertent Insertion of Excess Reactivity The g for stepwise insertion of reactivity in both the current HEU core and.
report presents the results of computations using the modified PARET code the proposed LEU core.
It was assumed that only inherent reactivity feedback mechanisms limit the transient, and the assumed values are given. Since the Technical Specifications limit excess reactivity to 0.44% delta k/k for both the HEU and LEU cores, this value was used in the stepwise reactivity insertion, even though no credible means has been identified to add all of the
. excess in such a fashion. Since the neutronics for both cores are very similar, the calculated transient response is very similar, with the LEU core being slightly less severe. For tie LEU core the calculated minimum period is 3.8 sec, the peak power is 183 kW, and the peak fuel temperature is 115 C.
For the HEU core these parameters are 3.8 sec 221 kW, and 116 C respectively.
Accordingly, the staff concludes the transient parameters for the LEU core are well below those at which fuel or cladding damage would occur and that the conversion to the LEU core does not increase the safety risk.
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2.12.2 Fuel Handling Accident An accident leading to structural damage of.a fully irradiated fuel element was -
4 considered to be the maximum hypothetical accident by both the licensee and the NRC in the evaluation of the MCZPR reactor facility for renewal of its operating license in 1985. This same accident was also considered by the staff in its assessment of the LEU fuel at Manhattan College. The only difference between the inventory of radioactivity in the LEU and in the HEU element is the plutonium-239 fonned by neutron capture in the uranium-238, which is more abundant in the LEU fuel. However, on the basis of the licensed power level of 0.1 Watt and the consequent low burn-up of fuel, the total. fission product i
inventory and the additional build-up of plutonium in the LEU fuel is radiologically insignificant, and release of the fission products including this plutonium from damaged LEU fuel would result in maximum potential radiation exposures in both the unrestricted and restricted areas of only a very
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small fraction of those allowed by 10 CFR Part 20 guidelines. Therefore,_ damage to LEU fuel in place of HEU fuel would cause no significant change in the risk to the health and safety of the public, which was already acceptably low for the current HEU core.
3.0 CONCLUSION
The staff has reviewed and evaluated all of the operational and safety factors impacted by the use of LEU fuel in the place of HEU fuel in the Manhattan College Zero Power Reactor.
It is concluded that the conversion, as proposed, would not reduce any safety margins, would not introduce any new safety issues, ar,d would not lead to increased radiological ris.k to the health and safety of the public. Therefore, the conversion to LEU V SI2 -Al fuel, as described, is 3
acceptable.
Dated: March 12, 1990 i
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4.0 REFERENCES
1.
Dr. R. Berlin and Dr. J. Perng Hu, " Conversion of Manhattan College, Zero Power Reactor (MCZPR) to Low Enriched Uranium (LEU) Fuel," submitted to:
United States Nuclear Regulatory Commission in Accordance with Requirements of 10 CFR 50.64, April 1989.
2.
J. E. Matos and K. E. Freese, " Analyses For Conversion Of The Manhattan College Zero Power Reactor From HEU To LEU Fuel," RERTR Program, Argonne National Laboratory, February 1989.
3.
Robert E. Berlin," Response to Questions Regarding HEU/ LEU Conversion at Manhattan College," Letter to T. Michaels, United States Nuclear Regulatory Commission, July 1989.
4 Safety Evaluation Report related to the " Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-power Reactors," NUREG-1313, USNRC July 1988.
5.
C. F. Obenchain, "PARET - A Program for the Analysis of Reactor Transients,"
100-17282, Idaho National Engineering Laboratory (1969).
6.
W. L. Woodruff, "A Kinetics and Thermal-Hydraulics capability for the Analysis of Research Reactors," Nuclear Technology 64, pp.196-206, (February 1984); and W. L. Woodruff, "The PARET Code and the Analysis of SPERT I Transients," Proc. International Meeting on Research and Test Reactor Core Conversions from HEU to LEU Fuel, Argonne National Laboratory, Argonne, IL, November 8-10, 1982, ANL/RERTR/TM-4, CONF-821155, pp. 560-578
-(1983).
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