ML20035H647
| ML20035H647 | |
| Person / Time | |
|---|---|
| Site: | University of Virginia |
| Issue date: | 04/29/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20035H609 | List: |
| References | |
| NUDOCS 9305060152 | |
| Download: ML20035H647 (11) | |
Text
p Cf G i
E
(
n UNITED STATES
?
~E NUCLEAR REGULATORY COMMISSION
%..o[s!
.l6 WASHINGTON. D.C. 20555 o
.s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CONVERSION ORDER TO CONVERT FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUEL AMENDED FACILITY OPERATING LICENSE NO. R-66 UNIVERSITY OF VIRGINIA RESEARCH REACTOR DOCKET NO. 50-62 1
INTRODUCTION On February 25, 1986, the U.S. Nuclear Regulatory Commission (NRC) issued a new regulation to Title 10 of the Code of Federal Reaulations,10 CFR 50.64, that requires licensed research and test non-power reactors (NPRs) to be converted from the use of high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel (less than 20 percent), unless specifically exempted.
Starting in 1978, the U.S. Department of Energy (D0E) took a leading role in an international program that provides the guidelines for converting from HEU to LEU fuel in NPRs. Activities in this program included reactor analyses, implementation of a demonstration conversion, and an extensive fuel qualification and development program.
For conversion of the University of Virginia Research Reactor (UVAR), DOE and the University of Virginia (UVA or licensee) decided to provide the LEU silicide-aluminum dispersion fuel (U Si -
3 z Al) developed by Argonne National Laboratory (ANL) especially for use in the DOE HEU-LEU conversion program.
In an effort to standardize and control costs of the conversion of about 20 reactors in the United States, DOE, NRC licensees, and the NRC agreed, safety considerations permitting, that only one plate-type fuel element design would be made available.
For the case of the UVAR, this agreement led to an increase in the number of fuel plates per element and an increase in the uranium concentration per plate.
The outer dimensions of the LEU and HEU fuel elements and the dimensions of the LEU and HEU fuel plates are the same; however, there are more LEU fuel plates per element, which results in changes in the thermal hydraulics. Conversion from HEU to LEU also leads to changes in nuclear parameters and reactivity conditions that must be addressed. Consequently, there is a need for a revised safety analysis.
i The NRC furnished guidance in preparing this revised safety analysis and advised licensees to concentrate on and consider those conditions and parameters of the reactor and the facility operating license that were dependent on fuel design and enrichment and, therefore, might be changed by j
the fuel conversion. Among the reactor conditions and parameters to be addressed are the following:
construction and geometry of the LEU fuel, critical and operating mass of U-235, hydraulics and thermal hydraulics, 9305060152 930429 PDR ADOCK 05000062 p
ena
. _ =
l
! i power density and power peaking, control rod worths, shutdown margin, excess reactivity, reactivity feedback coefficients, fission product inventory and containment, and potential accident scenarios.
Because the UVAR was required to convert to LEU fuel in accordance with I
10 CFR 50.64, the licensee submitted an application to the NRC for its authorization for' conversion by a letter dated November 9, 1989 (Reference 1).
l Attached to this letter were (1) revisions to the safety analysis report (SAR) for the LEU core, which presented the assumptions, methods, and results.of computations performed in support of the UVAR conversion, (2) revised technical specifications for the new LEU core, and (3) selected references.
The staff review led to additional questions to which the licensee responded j
by letters dated February 12, 1991 and December 14, 1992 (References 2 and 3).
These letters transmitted answers to the questions and revisions to the new LEU technical specifications. This material is available.for review at the Commission Public Document Room at 2120 L Street, N.W., Washington, D.C.
20555.
j This safety evaluation report (SER) was prepared by A. Adams, Jr.,' Senior'-
Project Manager, Division of Operating Reactor Support,.0ffice of Nuclear Reactcr Regulation, NRC. Major contributors to the technical review include W. R. Carpenter, R. E. Carter, and P. R. Napper of EG&G, Idaho National Engineering Laboratory (INEL).
i 2
EVALUATION The UVAR is licensed for operation at thermal power levels not.to exceed
-l 2 HWt. The reactor uses plate-type fuel cooled by forced circulation of water l
at a nominal flow rate of 1055 gpm (671/s) or, under certain allowed = low-power operations, natural convection flow. The licensee has. proposed no i
changes to any reactor system or operating characteristics except for replacing the HEU fuel elements with new LEU fuel elements. The following j
evaluations and conclusions are be ed on that assumption.
l The NRC as part of the conversion order is requiring the licensee to submit a i
conversion startup report within six months of completion of LEU core loading that discusses the results of various tests and measurements conducted during j
the core. conversion.
i 2.1 Fuel Construction and Geometry The HEU fuel elements currently installed in the UVAR contain 18 ' plates each,._
in which the fuel meat is a-92 percent enriched uranium-aluminum alloy.
Each i
fueled plate contains approximately 10.8 g of U-235 for a total U-235 loading -
of about 195.g per fuel element if no dummy fuel plates are utilized. The new I
' LEU fuel. elements will have the same outer dimensions as the HEU fuel-
~
elements, but will-contain 22 plates.each, with the fuel meat in the form of uranium silicide (enriched to 19.75' percent _U-235) _ dispersed in an aluminum matrix.
The LEU-fueled plates will each contain approximately 12.5 g of U-235 ~
q i
for a total loading of about 275 g of U-235 per fuel element assembly containing no dummy fuel plates. The geometries, materials, and fissile loadings of the current HEU fuel elements and the replacement LEU fuel elements are shown in Table 1.
The standardized LEU fuel elements and fuel plates have the same physical dimensions as their HEU counterparts, but because there are more LEU fuel plates per fuel element, the water gap between the LEU plates is narrower than the gap between the HEU plates. The resulting metal-to-water ratio for the LEU fuel element assemblies is 0.76 compared to 0.63 for the HEU assemblies. Fuel elements with plates and uranium composition essentially identical with the proposed UVAR plates were developed especially for the United States NPR conversion program by ANL. These fuel elements have been tested extensively and irradiated to relatively high burnup in the Oak Ridge Research Reactor (ORR) with no failures having a safety significance. The performance of the fuel was reviewed and the fuel was approved by the NRC (Reference 4).
Partial fuel elements, used for excess reactivity control, have the same dimensions as standard LEU fuel elements. However, these fuel elements contain alternating aluminum dummy plates and LEU fuel plates, resulting in an element containing only half the fuel of a standard LEU fuel element.
The four control rod elements have the same dimensions as standard LEU fuel j
elements and have the same water gap between fuel plates.
Each control rod element, however, contains only 11 fuel plates and has an open center where the control rod travels.
Graphite elements, used around the edges of the core, as necessary, to provide an improved reflector, have the same approximate outer dimensions as fuel elements. These elements consist of a solid graphite core encased in a water-tight aluminum jacket. The graphite element design is retained from the HEU core.
Standard LEU fuel elements, the four control rod elements, partial fuel elements, and graphite elements may be loaded in the core, as necessary, to provide a critical assembly having no more than 7.00$ (5.18% Ak/k) excess reactivity. The minimum shutdown margin required by the UVAR technical specifications is 0.555 (0.41% Ak/k) with the highest-worth reactivity scramable control rod and the non-scramble regulating rod in their most reactive position.
2.2 Fuel Storaae LEU fuel, not in the reactor core, will be stored in the following four areas:
(1) fuel storage room (dry), (2) 24 spaces in the auxiliary fuel storage rack i
(wet), (3) 12 spaces in the three four-element racks (wet) and, (4) 12 spaces of the wall rack (wet). A LEU criticality analyses presented in the UVAR SAR applicable to the fuel storage room, auxiliary fuel storage rack, and the four-element racks yielded a k,,, of 0.8 for a water-moderated infinite array of fuel with a center-to-center spacing of 5.5 in., which is less than the center-to-center spacing of any of the three storage facilities.
A LEU SAR criticality analysis applicable to the wall rack yielded a k, of 0.74 for a g
t i
- water-moderated infinite linear array of LEU fuel with no separation between elements, which is a conservative estimate of the actual 3.5 in. separation.
This meets the UVAR technical. specification limit of 0.9 for the k,,, of fuel elements not in the reactor core.
Both of these analyses demonstrate that the separation between fuel elements for the four storage locations effectively isolates the LEU fuel elements from each other neutronically and provides criticality safe areas to store this fuel.
Because of the very conservative approach taken with the criticality calculations and because the reactivity worth of the LEU and the HEU fuel elements is nearly identical, the staff i
concludes that interim storage (should it become necessary) of both types of fuel in the four storage areas during the conversion process is acceptable.
UVAR can safely store simultaneously both the HEU and the LEU fuel utilizing the four described storage areas.
2.3 Critical loadinas of U-235 The UVAR core has a wide variety of possible critical loadings because of the grid plate design, which is an 8 x 8 array providing 64 positions for full or partially loaded fuel elements, control rod elements, graphite reflector elements, and grid-plate plugs. UVAR technical specifications limit these various possible critical core loadings to a shutdown margin of no less than 0.55$ (0.41% Ak/k). Excess reactivity for any core configuration can be no greater than 7.00$ (5.18% Ak/k).
For comparison of the nuclear characteristics of the HEU and the LEU fuel, a simple water-reflected (no graphite elements) 4 x 5 critical array was analyzed (Reference 5).
A HEU 4 x 5 water-reflected critical assembly, would contain 16 standard HEU fuel elements and 4 HEU control rod elements having a total uranium mass of approximately 3.9 kg, of which 3.6 kg is U-235. A similar LEU 4 x 5 water-reflected critical assembly with 16 standard LEU fuel elements and 4 LEU control rod elements would have a total uranium mass of approximately 25.1 kg of which 5.0 kg is U-235. The calculated excess reactivity for the HEU core and the LEU core is 4.205 (3.1% Ak/k) and 3.90$ (2.9% Ak/k), respectively.
The additional U-235 is required to compensate for the absorption of both epithermal and thermal neutrons in the U-238 of the LEU. This is achieved partially by the increase of uranium concentration indicated in Table 1.
The calculated change in fuel loading is as expected and is consistent with other conversions from HEU to LEU fuel. The effective delayed neutron fraction is unchanged, and the prompt neutron lifetime is somewhat decreased, as expected, because of the larger uranium loading and the increased metal-to-water ratio in the LEU core. Therefore, the staff concludes that these and other calculated results confirm the basic neutronic similarity between the HEU and i
the proposed LEU cores of the UVAR.
2.4 Hydraulics and Thermal Hydraulics As noted in Section 2.1, the UVAR LEU fuel element has the same cross-sectional area and the same volume of meat and cladding per fuel plate as the HEU fuel. There are 22 LEU fuel plates versus 18 HEU fuel plates in the same size fuel element. This configuration results in less power generation in the i
average LEU fuel plate and a decrease in the size of the water gap between i
pl ates. However, for the minimum size operable core analyzed, a square array of 16 fuel elements, the radial power peaking in the hottest coolant channel exceeds that in the HEU core by about 5 percent. The cause and the consequences of power peaking are discussed further in Section 2.5.
)
~
i
J -
The SAR analyses of the thermal hydraulics of the LEU core employ the same assumption and methods as for the existing HEU core. The analyses derive a minimum burnout ratio (BOR), based on applicable experimental data in the literature, that will provide 99 percent confidence that no failure of fuel or cladding will occur as a result of inadequate cooling during steady power operation. The mechanisms considered that could lead to such burnout are departure from nucleate boiling or dry-out resulting from other coolant flow instabilities.
Because the cladding and fuel matrix are made of essentially i
the same kind of material for both HEU and LEU fuel, the minimum BOR is the same for both. The analyses then include the coolant flow rates, the various coolant system operational uncertainties, and the engineering (fabrication) tolerances of both the HEU and LEU fuel elements to derive the conditions necessary to provide 99 percent confidence that at least the minimum BOR is attained in the hottest coolant channel. These analyscs include both forced coolant flow that is downward through the core and natural convection flow that is upward. The forced-flow analyses provide a calculated envelope of reactor core power versus total primary coolant flow that will give the same high assurance as for the HEU fuel of no fuel failure.
This curve, along with the relevant assumptions, establish the safety limits on process variables for forced-flow operation of the reactor. These safety limits are included in the UVAR technical specifications, discussed in Section 2.12.
For the LEU reactor, the calculated coolant flow required to ensure fuel integrity at any particular core thermal power level is higher than for the HEU core. The two principal factors responsible are:
(1) the higher power peaking and (2) the decrease in water gap size between adjacent plates of the LEU fuel. The power peaking is discussed in Section 2.5.
The effect of the gap size derives from an assumption that the fabrication tolerances on the spacing between plates are numerically the same for HEU and LEU fuel elements.
With the nominal spacing for LEU smaller than for HEU, this means that the smallest gap size that must be considered to assure 99 percent confidence in i
LEU fuel integrity is smaller than the gap for HEU. Thus, in order to ensure that coolant flow through this smaller gap is still sufficient to avoid fuel burnout, the pressure drop across the core must be increased, thereby increasing the flow through the gaps of nominal size in the rest of the core.
This accounts for the larger total coolant flow requirement in the LEU core.
The calculations discussed above provide the envelope of safety limits. A
.i worst-case scenario was postulated to establish limiting safety systems settings (LSSSs) designed to protect the reactor under all conditions. This is discussed in Section 2.11.1.
That scerario leads to a maximum power level of 3.88 MW. An LSSS on coolant flow must be selected that ensures that the safety limit envelope is not reached at this power. For HEU, that flow was 2 800 gpm (50 1/s).
For LEU, that flow is 2 900 gpm (571/s).
The UVAR is also licensed to operate at lower power in the natural convection mode. For the natural-convection flow analysis for the LEU core, a very pessimistic thermal hydraulic scenario (a loss of flow transient) was assumed, which shows that the resultant coolant and fuel temperatures are essentially the same as they were for the HEU core and well below the failure levels for the fuel plates.
In the analysis, the core was assumed to have been operating for an extended period of time at 3 HW under normal downward forced-flow conditions, then suddenly to experience a pump failure resulting in a flow reversal into the natural-convection flow mode at a power level of 750 kW.
i
_ The results show the maximum fuel temperature reaches 303 'F (150 *C), then cools to a steady-state value of about 270 *F (132 *C), both temperatures are well below the minimum aluminum blister temperature of 515 *C {959 *F).
The i
moderator temperature during the transient increases about 10 F (6 C) to about 130 *F (54 *C).
These transient results, at 750 kW, indicate that normal operations of the UVAR LEU core in the natural-convection flow mode at no more than the licensed power of 300 kW would not present any challenge to the fuel-plate integrity. As noted previously, thermal hydraulic risk is not i
increased for the LEU core in the forced-flow mode. Therefore, the staff concludes that the conversion from HEU to LEU fuel at the UVAR would not increase the risk of thermal hydraulic damage to the UVAR facility in either mode of coolant flow.
r 2.5 Power Density and Power Peakina Power densities and power peaking, based on the assumed nuclear parameters for both graphite and water reflected 4 x 4 HEU and LEU cores, were computed for the UVAR (References 5 and 6). The power distribution among the fuel elements is very nearly the same for the two cores, with the peak power density located in the channel adjacent to the water gap in a control rod fuel element.
The peak-to-average power density is about 5 percent higher for the LEU than for the HEU core. The low energy neutron spectrum in the LEU core is harder than in the HEU core, and the higher peak is caused by more epithermal neutrons down-scattering in the relatively wide water gap left by the withdrawn control rod.
The SAR analyses of both normal operation and accident events discussed in Sections 2.11.1 and 2.11.2 for the UVAR include the effects of this power peaking.
For both the HEU and proposed LEU cores the safety limits and the LSSSs are based on the thermal hydraulic conditions in the hot channel. The smaller the core size, the higher the average power density. The 4 x 4 core analyzed by the licensee is the smallest core possible with the available fuel plates and reflectors. The staff concludes that the power density and power peaking in the LEU core is acceptable.
2.6 Control Rod Worths The UVAR has four control rods, three are scrammable and are used as safety and shim rods, and the fourth rod, which is non-scrammable, is used as a regulating rod. The reactivity worths of the control rods were computed by acceptable methods for the UVAR LEU core. The calculated worths of the r
regulating rod and the three shim rods for the LEU core are 0.60$ (0.45%
l Ak/k), 3.80$ (2.83% Ak/k), 5.00$ (3.69% Ak/k), and 4.30$ (3.21% Ak/k),
respectively. The calculated worth for these same four control rods for the HEU core are 0.50$ (0.38% Ak/k), 3.80$ (2.84% Ak/k), 5.00$ (3.69% Ak/k), and 4.30$ (3.21% Ak/k). As can be seen, th( iEU-calculated worths for the three shim rods are virtually unchanged from those of the HEU, while the l
LEU-calculated worth of the regulating rod is slightly higher. The HEU/ LEU i
conversion has almost no effect on the worth of the control rods, and what small effect there is, is positive. Therefore, the staff concludes that the rod worths are still fully acceptable for safe reactor operation and control at the UVAR.
i
I.
2.7 Shutdown Marcin It is the practice of the NRC to require that there be reasonable assurance that a NPR can be shut down from any operating condition, even if the scrammable control / safety rod of maximum worth is in its most reactive position ano any non-scrammable control rods are also at their most reactive positions (at the UVAR, these positions are rods fully withdrawn). On the basis of the computed control rod worths and the computed excess reactivity r
for the LEU 4 x 5 water-reflected core, the UVAR would be subcritical by approximately 4.255 (3.14% Ak/k) with the regulating rod and the shim-safety rod of maximum worth fully withdrawn. The staff concludes that this is substantially larger than the technical specification margin of at least 0.555 (0.41% Ak/k) and is acceptable.
It is understood that for various core configurations, the control rod worths can change. The technical i
specifications requirements for shutdown margin shall always be in compliance and shall always take precedence over the technical specifications allowed maximum core excess reactivity.
l i
2.8 Excess Reactivity Additional reactivity above cold, clean critical is required to allow a reactor to perform programmatic and academic functions. The SAR discussed the amount of this excess reactivity required to compensate for various operational losses of reactivity and calculational methods of adjusting reactivity by varying the U-235 loading, rearranging the fuel element matrix, and changing reflectors. The staff concludes that these calculations indicate there is reasonable assurance that the excess reactivity requirements of the UVAR can be achieved. The maximum excess reactivity permitted by the UVAR technical specifications is 7.00$ (5.18% Ak/k) for both the HEU and the LEU l
cores. The operational excess reactivity, however, is always limited by the ability of the core to maintain a minimum shutdown margin of 0.555 (0.41%
Ak/k) with the most reactive scrammable control rod and the regulating rod I
withdrawn.
2.9 Reactivity feedback Coefficients The temperature coefficient of reactivity and the void coefficient of j
reactivity were calculated for the LEU core and compared to those calculated for the HEU core. The moderator temperature coefficient is nearly the same e
for both the LEU core and HEU core, however, the fuel temperature coefficient and the void coefficient are more negative for the LEU core. The calculated i
fuel temperature coefficient for the LEU core is more negative than the HEU core because of the Doppler effect in broadening the neutron capture resonances of the relatively much more abundar " "'38 present in the LEU fuel.
Because the Doppler feedback is a function of emperature, it is prompt f
and, therefore, more effective in countering a reas.. ar transient in the LEU core than is the moderator temperature coefficient in the HEU core, which must i
rely on heat transfer to the moderator. Because the predicted reactivity coefficients for the LEU core are the same or larger than those of the HEU core and are more effective in leading to reactor stability than the reactivity coefficients for the HEU core, the staff considers the LEU reactivity feedback coefficients acceptable.
l
i 2.10 Fission Product Inventory and Containment The total inventory of fission products will not be significantly different between the HEU and LEU cores. However, because there are 22 plates in a LEU fuel element versus 18 plates in a HEU fuel element, the power density will be i
less, resulting in less fission product inventory per fuel plate in the LEU core for the same operating (power-time) history. The aluminum cladding on l
the LEU fuel plates is the same thickness as that used for the HEU fuel plates. Cladding of this thickness has been used on HEU fuel for years in many NRC-licensed research reactors with no failures or significant releases j
of fission products attributable to the loss of integrity resulting from this i
aluminum thickness. Therefore, since the cladding thickness has not changed and there will be a lower fission product inventory per fuel plate, there is reasonable assurance that the new LEU fuel will perform at least as t
satisfactorily as the HEU fuel it will replace in containing fission products in the UVAR. The staff finds this acceptable.
i 2.11 Potential Accident Scenarios Several potential accident scenarios were postulated and analyzed by the licensee and evaluated by the NRC staff for the license renewal of the HEU fueled UVAR in 1982. Of these, only the scenarios considered below could be affected by the conversion of the reactor to LEU fuel.
2.11.1 Reactivity Insertion Accident The scenario and assumptions for the reactivity insertion accident were unchanged from the HEU core to the LEU core, and the results of the event did not change.
For both scenarios, it was assumed that the reactor is operating with forced coolant flow and a control rod is continuously withdrawn.
inserting a ramp increase in reactivity. This results in a continuously I
decreasing reactor period and an increasing power level. The scenario assumed i
that both the reactor period scram and the power level scram are operable and set at their technical specification values of 3.3 seconds and 3.0 MW, respectively. The analyses show that either of these scram channels limit the nuclear excursion at the same peak power level, depending on the initial power i
level.
It was assumed that the coolant flow is below its normal operating level of about 1020 gpm (641/s), but is down to the nominal LSSS for each fuel type, namely 800 gpm (501/s) for HEU and 900 gpm (571/s) for LEU, but further reduced to the limit for 99 percent confidence of 744 gpm (471/s) and 837 gpm i
(53 1/s), respectively.
It was further assumed that the temperature coefficient of reactivity is zero.
l This assumption has little effect on the HEU scenario, but introduces a small conservatism for the LEU scenario. For these scenarios the progression of the i
events for the two fuel types are very similar. The maximum power levels reached are the same, depending primarily on the dynamic parameters of the l
scram circuits and the control rod insertion times. These parameters are not to be changed by the fuel conversion. The licensee determined that the period and power level scrams would occur at essentially the same time, would prevent j
the reactor becoming prompt critical, and would limit the maximum transient power reached for either fuel to 3.88 MW. Because the reactor scram has 1
already been initiated, this 3.88 MW is a transient overshoot in power, lasting only momentarily.
Even if the reactor were to operate continuously at this power, the envelope of the calculated safety limits for the minimum sized core for each fuel type would not be exceeded and fuel integrity would not be lost. The analyses assumed that the coolant flow, power level, and coolant temperature scrams are at their technical specification LSSSs. The thermal hydraulic analyses further assumed that the true values of these parameters are at the engineering (fabrication) tolerance limits discussed in the SAR.
The postulated reactivity addition accident scenario and the results are
}
unchanged between the HEU and LEU cores, and the SAR has demonstrated that reaching the analyzed thermal-hydraulic safety limits are avoided by increasing the LSSS on coolant flow for the LEU. Therefore, the staff concludes the risk of fuel failure from the reactivity addition accident is i
I not increased as a result of the HEU/ LEU fuel conversion.
v 2.11.2 Thermal Hydraulic Transients Two thermal-hydraulic events, the loss-of-flow accident and the loss-of-coolant accident (LOCA), were investigated as part of the safety evaluation.
The accident sequence scenarios for both of these events were identical for the HEU and LEU, except that the HEU was analyzed at an initial coolant flow of 744 gpm (47 1/s), which is the minimum true value of the 800 gpm i 7.0 t
percent (50 1/s i 7.0%) LSSS for the HEU core, while the LEU was analyzed at 837 gpm (531/s), which is the minimum true value of the 900 gpm i 7.0 percent (57 1/s 7.0%) LSSS for the LEU core. The results for the loss of flow accident show that the maximum fuel temperature reached during the flow reversal from forced downflow to natural convection upflow for both the HEU i
and the LEU analysis was 303 *F (150 *C).
For the LOCA scenarios, the slightly reduced power density per fuel plate of the LEU core resulted in a peak fuel temperature somewhat less than that computed for the HEU core
[975 *F (524 *C) versus 1080 *F (582 *C)] for the worst case LOCA, which assumed core uncovering in 20 minutes and no credit taken for the emergency core cooling system (ECCS) sprays. Assuming loss of both of the independent i
ECCS trains is very conservative and results in an overestimate in the peak fuel temperature of about 200 *F (93 *C).
Because the safety analysis demonstrates that the thermal-hydraulic safety margins are not reduced by the l
conversion of the UVAR from HEU to LEU, the staff concludes the conversion-l l
does not increase the health and safety risk to the public from the postulated thermal hydraulic accidents.
2.12 Technical Specification Chanaes The SAR analyses compare the thermal hydraulic conditions of the HEU and proposed LEU reactors by the same methods and with the same type of assumptions. The safety limits and LSSSs were derived for each reactor to provide assurance with at least 99 percent confidence that fuel integrity will not be lost under even the worst credible conditions. These analyses provided the bases for changes in the technical specifications for the LEU reactor.
The' staff concludes that the changes in the UVAR technical specifications proposed by the licensee will ensure no less protection of the health and safety of the public as a result of the conversion from HEU to LEU fuel, as proposed. Although the LEU fuel plates are not less resistant to malfunction or failure than the HEU plates, the increase in the number of plates per fuel element is primarily responsible for the changes in safety limits and LSSSs at the UVAR.
t i
p
l 2.12.1 Safety Limits f
Technical Specification (TS) 2.1.1 and TS Figure 2.1, safety limits in forced convection mode of operation, are amended for the LEU reactor to change the envelope to increase the forced coolant flow by about 15 percent for each power level. Coolant inlet temperature and height of pool level above the core remain unchanged. The basis is rewritten to reflect the LEU SAR.
r TS 2.1.2, safety limits in the natural convection mode of operation, is reformatted and the basis is rewritten to reflect the LEU SAR.
TS 2.1.3, safety limit for the transition from forced to natural convection mode of operation, is added to ensure that the reactor is shut down when the transition is made from forced to natural convection operation.
2.12.2 Limiting Safety System Settings TS 2.2, LSSSs, is amended for the LEU reactor to increase the forced coolant flow limitation from 800 gpm (501/s) to 900 gpm (571/s), about 14 percent.
In addition, an LSSS of a minimum reactor period of 3.3 seconds scram has been added. The other LSSS values remain the same. The addition of the LSSS period scram is consistent with the assumptions in the SAR analysis of the ramp reactivity insertion scenario. The technical specification has been reformatted and the bases rewritten to reflect the LEU SAR.
2.12.3 Limiting Conditions for Operation TS 3.2, reactor safety systems, is amended to reflect the LEU SAR. Table 3.1, safety system channels, is amended to add scram setpoints for the bridge radiation monitor and air pressure to header. The primary flow scram and reactor period scram setpoints are changed to reflect the changes in the LSSSs j
discussed above.
TS 4.8, reactor HEU fuel dose measurements, is amended to improve the format of the section and to clearly state that the specification applies to high enriched fuel and not the new low enriched fuel.
L 2.12.4 Design Features TS 5.1, reactor fuel specifications, is amended to describe the new LEU fuel I
material, element description, and core configurations introduced by conversion. The licensee requested additions to TS 5.3, fuel storage, to i
describe the various special nuclear material possession limits for the i
facility. This information is redundant to license condition II B.(2) and (4)
)
and therefore, the staff is not placing this information in the technical j
specifications. This was discussed with and agreed to by the Director of the UVAR on March 22, 1993.
l 3 CONCLUSION l
The staff concludes that the conversion, as proposed, would not reduce any j
safety margins, introduce any new safety issues, or lead to increased i
radiological risk to the health and safety of the reactor staff or the public.
]
Therefore, the conversion to LEU U Si -Affuel, as described, is acceptable.
3 z Date:
April 29, 1993 i
l
- )) -
4 REFERENCES 1.
University of Virginia, " Application for Authorization to Convert to LEU,"
Docket No. 50-62, submitted to U.S. Nuclear Regulatory Commission in accordance with requirements of 10 CFR 50.64, November 9,1989.
2.
Robert U. Mulder, Director, University of Virginia Reactor Facility, letter to Marvin M. Mendonca, Senior Project Manager, Non-Power Reactors, Decommissioning and Environmental Project Directorate, ONRR, USNRC, "U.Va.'s Reply to NRC's Request for Additional Information with Respect to the UVAR LEU Fuel Conversion Application: Docket No. 50-62,"
february 12, 1991.
3.
Robert U. Mulder, Director, University of Virginia Reactor Facility, letter to Alexander Adams, Jr., Senior Project Manager, Non-Power Reactors and Decommissioning Project Directorate, ONRR, USNRC, "U.Va's Reply to NRC's Request for Additional Information with Respect to the UVAR LEU Fuel Conversion Application: Docket No. 50-62," December 14, 1992.
4.
" Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-Power Reactors,"
NUREG-1313, July 1988.
5.
D. W. Freeman, "Neutronic Analysis for The UVAR Reactor HEU to LEU Conversion Project," Masters Thesis, University of Virginia, School of Engineering and Applied Science, July 1989.
6.
S. Wesserman, " Effective Diffusion Theory Cross Sections for UVAR Control-Rods," Masters Thesis, University of Virginia, School of Engineering and Applied Science, January 1990.
i k
k T
F Table 1 Comparison of Parameters for the HEU and LEU Cores" at the University of Virginia Reactor b
Critical mass (g U-235) 3596 4952 6
Excess reactivity (%Ak/k) 3.1 2.9
($)
4.20 3.90 b
B 0.0074 0.0074 NYu' tron lifetime, 2, (ps) 64.4 53.0 6
6 f/8,,,(s) 0.009 0.007 Fuel Elements:
Number of standard fuel elements 16 16 Number of plates per standard element 18 22 Number of control fuel elements 4
4 Number of plates per control element 9
11 Fuel plate dimensions (in.)
24.6x2.8x0.05 24.6x2.8x0.05 (cm) 62.5x7.1x0.13 62.5x7.1x0.13 Enrichment (%)
92 19.75 Mass of U-235 per plate (g) 10.8 12.5 Water gap (in.)
0.122 0.0927 (cm) 0.310 0.235 fuel thickness (in.)
0.02 0.02 (cm) 0.05 0.05 ~
Aluminum cladding thickness (in.)
0.015 0.015 (cm) 0.038 0.038 Uranium density (g/cc) 0.69 3.47 fuel matrix UAf -Af U Si -Al t
3 2 Rod Worth:
Regulating (%Ak/k) 0.38 0.45 l
($)
0.50 0.60 Safety (nominal) (%Ak/k) 3.24 3.24
($)
4.40 4.40 Reactivity Coefficients:
Moderator temperature coefficient (%Ak/k/*C)
-0.011
-0.011
($/*C)
-0.015
-0.015 Fuel temperature coefficient (%Ak/k/*C)
-1. 3 x 10
-1.1 x10'3
($/*C)
-1. 8x10
-1.5x10'3 Void coefficient (%Ak/k per % void)
-0.194
-0.248
($ per % void)
-0.262
-0.335 4x5 water-reflected core b
calculated