ML20198E814

SER Supporting Amend 12 to License R-125

ML20198E814
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Site:	University of Lowell
Issue date:	07/31/1997
From:	NRC (Affiliation Not Assigned)
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2006H001

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING THE ORDER TO C0dVERT FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUEL FACILITY OPERATING LICENSE NO. R-125 UNIVERSITY OF MASSACHUSETTS LOWELL DOCKET NO. 50 223

1.0 INTRODUCTION

Section 50.64 of Title 10 of the Code of Federal Reaulations (10 CFR 50.64) requires licensed research and test non-power reactors to convert from the use of high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel, unless specifically exempted. The University of Massachusetts Lowell (the licensee) has proposed to convert the fuel in the University of Massachusetts Lowell Reactor (UMLR) from HEU to LEU. In a letter of May 21, 1993, the licensee submitted a supplement to its existing final safety analysis report (FSAR)

(September 1973, as amended) describing the changes needed to convert to LEU fuel. A copy of the proposed technical specifications (TSs) needed to operate with LEU fuel was also submitted for NRC's review and approval. Additional information and clarifications to the submittal of May 21, 1993, were submitted by letter dated March 17, 1994, May 16, 1997, and June 6, 1997, 2.0 EVALUATION 2.1 General Facility Description The UMLR is licensed to operate at thermal power levels not to exceed 1 megawatt thermal (MW(t)) when cooled by forced convection at a nouinal flow of 1,600 gallons per minute and a pool water level of greater than 24.25 feet above core centerline. The primary coolant is cooled by a heat exchanger and the heat transferred to the secondary coolant system. The secondary coolant system rejects the heat to the atmosphere through a cooling tower.

The UMLR may operate at- power levels of 0.1 MW(t) or less when cooled by natural convection with the pool water level greater than 24.25 feet above core centerline. in addition, the UMLR may operate at power levels of 1 kilowatt thermal (KW(t)) or less when cooled by natural convection with a pool water level of greater than 2.25 feet above core centerline.

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2.2 Fuel Construction and Geometry The HEU fuel elements used at the UMLR consist of 2 aluminum side plates and 18 equally spaced flat fueled plates of typical materials test reactor design. -The uranium in the HEU fuel meat is enriched to about 93 percent uranium-235. Each plate contains approximately 7.5 grams of uranium-235. The outer dimensions of the HEU fuel elements are 7.62 cm by 7.772 cm. Each fuel plate is 7.046-cm wide by 63.5-cm high. The fuel meat is 5.461-cm wide by 60.96-cm high, with a fuel meat thickness of approximately 0.305 mm and a clad thickness of about 0.610 mm.

The LEU fuel elements will be of similar design with essentially the same outer dimensions as the HEU fuel elements, but will contain 16 fuel plates with a dummy aluminum plate at each end and 2 aluminum side plates. Each of these plates will consist of uranium silicide dispersed in aluminum (U 3Si 2-A1) and completely clad in aluminum alloy, in the LEU plates, the fuel meat will be about 0.510-mm thick, and the clad will be about 0.380-mm thick. The uranium in the LEU fuel meat is enriched to less than 20 percent uranium-235.

Each plate contains about 12.5 grams of uranium-235.

The overall width of the LEU fuel plate will be about 7.14 cm, comnared to 7.046 cm for the HEU fuel, and the width of the active fuel will be approximately 6.1 cm maximum, compared to approximately 5.46 cm for HEU. The length of an active LEU fuel plate will be 59.69 cm versus 60.96 cm for HEU.

The Argonne National Laboratory (ANL) developed these fuel element plates for conversion to LEU fuel at the U.S. non-power reactors. Fuel element plates of a design very similar to the UMLR design were tested in the Oak Ridge Research Reactor under extreme operational and hostile environmental conditions for most non-power reactors, including the UMLR, and performed acceptably. The NRC staff reviewed and approved the use of this type of fuel in NUREG-1313, "SER Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-power Reactors," July 1988. The characteristics of the fuel proposed for the LEU conversion at the UMLR are similar to those of the fuel tested and evaluated in NUREG-1313 and are consistent with those previously accepted for other non-power reactors.

Therefore, the staff finds the fuel construction and geometry acceptable.

2.3 Core Confiouration-The current HEU core consists of 25 full fuel assemblies and one partial fuel assembly containing about half the fuel loading of a standard assembly.

The fuel assemblies are in a 5 x 6 array, with the four corners filled with graphite reflectors. The core is moderated and cooled by light water.

Reactor control is maintained by four control blades and one regulating rod.

Two control blades are located between rows B and C, and two control blades are located between rows E and F. The positions of the control blades will remain the same for the LEU core. The regulating rod for the current HEU core is located in position D9.

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3 t The proposed LEU care, or reference core, consists of 20 full fuel assemblies in a 5 x 5 array, with the four corners occupied by radiation baskets and the center core position (05) occupied by a nonfuel aluminum graphite neutron flux trap. Actual core configurations can range from 16 to 28 assemblies, including 2 half-loaded partial assemblies. The LEU core is moderated and cooled by light water. Since the core can operate with fewer or more fuel assemblies than the 20-assembly reference core used for all safety analysis calculations, the licensee has agreed to add TS 5.2.3 requiring an analysis to establish that no limiting safety system settings (LSSSs) need to be changed to keep safety limits from being violated during the transients anticipated in the May 1993 FSAR supplement, in the May 1993 FSAR supplement (Section 2.2, p. 10), the licensee stated that the central aluminum tubes in all but five radiation baskets in the proposed reference core will have to be blocked to ensure adequate flow through the fueled assemblies. The licensee has added this requirement in TS 3.1.3.

Since adequate flow through the fuel assemblics for all core configurations is essential, the analysis required by TS 5.2.3 should include flow rate through fuel assemblies.

Reactor control is maintained by the four control blades in the same locations as in the HEU core. Because the LEU core is smaller, the licensee proposes to move the regulating rod from position 09 to position 08 next to the fuel. The licensee estimates that the regulating rod, if left in position 09, would be worth a few hundredths of a percent of reactivity instead of the few tenths of a percent needed to efficiently and safely control reactor power manually and automatically; in all cases, the regulating rod worth is required by TS 3.1.7 to be less than beta.rt. The design of the regulating rod will not change.

However, the licensee will place a 3-inch offset beid in the lift-shaft between the regulating rod and the rod drive and add a support bracket to the suspension bridge gridwork below the offset. The licensee has committed to preoperational testing of the proposed regulating rod design before loading LEU fuel. The staff finds the core configuration proposed by the licensee acceptable.

2.4 Euel Storaae The licensee has analyzed the fresh and spent fuel arrangements in the t3LR pool and has determined that the existing manufactured fuel holding racks will hold all LEU new fuel and all HEU currently on hand. The racks are sectioned into compartments with a 0.5-inch aluminum wall; thus, the stored elements are separated by a minimum of 0.5 inch of aluminum. Calculations performed by 1

ANL indicate that with a 225-gram of uranium-235 fuel element loading (versus 200 grams of uranium-235 in the UMLR fuel) and an element separation

'R. B. Pond and J. E. Matos, " Nuclear Criticality Assessment of LEU and HEU Fuel Elements Storage," Proceedings of the International Meeting on Reduced Enrichment for Research and Test Reactors (JAERI-M,84-073), Tokai, Japan (May 1984).

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4 of 1.766 cm (essentially the same as in the UMLR), the water-reflected infinite array had a K.tr = 0.715. The method of storing fuel proposed by-the licensee is acceptable because the mechanical- design of the HEU and the LEU is similar and because tht. proposed TSs require the licensee to store fuel only in existing storage racks located in the UMLR pool. During the " Low Water" mode of operation the licensee (during a telephone discussion on December 22, 1993) has recognized the need to be particularly mindful of the potential for radiation exposure to personnel from stored spent fuel. The staff finds the proposed simultaneous storage of HEU and LEU fuels acceptable.

2.5 Critical Ooeratina Masses of Uranium-235 Each HEU fuel element contains about 135 grams of uranium 235, and each LEU fuel element contains auout 200 grams of uranium-235. The current core configuration has a uranium-235 operating mass of approximately 3,510 grams of HEU. For the LEU reference core, the loading-would be approximately 4,000 grams of uranium-235. This LEU Inading for an operating core is reasonable I

for the intended purpose of the reactor and is consistent with other LEU conversions when configuration and power level are considered. Therefore, fuel loading for the proposed LEU cores is acceptable.

2.6 Basic Nuclear Parameters Calculated nuclear input parameters for reactivity calculations, such as prompt neutron lifetime and the effective delayed neutron fraction (B.tr),

changed'as expected from the HEU to LEU cores. The reference core prompt neutron lifetime decreased from 75.6 microseconds to 64.5 microseconds, primarily-because of increased leakage from the pr,oposed smaller core. The B.trfor the LEU core is calculated to be 7.8 x 10' versus 7,69 x10 for the HEU core. This slight difference is as expected and is similar to that calculated for other conversions-to LEU.

2.7 Excess Reactivity The licensee calculated the amount of excess reactivity needed to operate with l an LEU core to be a maximum of 4.7 percent-(the current TS limit is 4.7 percent). This excess would compensate for the various operational losses in reactivity from burnup, xenon, temperature, and experiments. Because of safety considerations, the operational excess reactivity is always limited by  ;

the requirement to maintain the TS minimum shutdown margin (UMLR TS 3.1.1).

The licensee will verify the excess reactivity during the LEU reactor core

.startup' testing.

Therefore, the excess reactivity for the proposed LEU conversion is j acceptable.

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5-2.8 Contrgl Rod and Reaulatina Rod Worths The proposed LEU core uses the same four control blades that have been used for the HEU core. The licensee evaluated tne worth of these control blades using the ANL standard neutron kinetics models snd computer codes and verified that the control blades will acceptably meet the TS requirements for the proposed LEU cores. The licensee will verify the worth of these control i

blades as a part of the LEU reactor core startup testing program.

l The licensee proposes to change the position of the nonscransnable regulating rod for use with the LEU core. Calculations showed that the reactivity worth of the regulating rod in the current position (09) would decrease from a few tenths percent to a few hundredths percent delta k/k. To ensure an adecuate reactivity for the regulating rod, the licensee proposes to move the roc to the D8 position, which is a grid position adjacent to the proposed LEU core.

This movenent would result in an estimated worth for the regulating rod of several tenths percent delta k/k. This relocation acceptably ensures control for normal plant operations. Since the licensee will determine the worth of the four control rods and the regulating rod during the startup testing of the LEU reactor core and verify that the control rods and regulating rod perform as designed, the staff finds the control rod worth design acceptable.

2.9 Shutdown Marain The staff requires reasonable assurance that the UMLR can be shut down from any operating condition, even if the one control blade of maximum worth and the nonscrammable regulating rod are in their fully withdrawn position. With the calculated control blade worths, proposed LEU core configurations, and ANL standard neutron kinetics methods, the calculated UMLR shutdown margin would not be lower than about 3.4 percent delta k/k with the control and regulating rod worths given in the May 1993 FSAR supplement cnd the maximu:n allowable excess reactivity allowed by the TS are assumed. This shutdown margin is considerab1's gnater than the 2.7-percent delta k/k required by TS. Since the shutdown mt6 n will be verified during startup testing with the LEU reactor core to enst.re that the TS limit is met, the staff finds the proposed TS shutdown margin for the LEU core acceptable.

2.10 Core Power Characteristics The licensee analyzed the core thermal power characteristics for the proposed LEU core. These analyses used standard computer programs and ANL standard nuclear kinetics models, The analysis calculated a maximum heat flux of 2.75 x 10' BTV/h ft* for the LEU core versus 2.17 x 10' BTU /hr - f ta for the HEU core. This heat flux is consistent for the proposed LEU core design.

The core analysis also indicates a peak axial power ratio of approximately 3.91 with control blades fully withdrawn (leading to an assumed chopaed cosine flux distribution) and approximately 3.38 with control blades 15 incies withdrawn. The staff reviewed the analysis inputs, methods, and results

• i 6- 4 (including Tables 3.1, 3.2, and 3.4 of the UMLR FSAR supplement of May 21, 1993) and has concluded that the licensee acceptably determined the power conditions to be used in analyzing thermal-hydraulic conditions, as well as transient and accident conditions (discussed in Section 2.14). l 2.11 Thermal Hydra.gl{n The licensee performed a thermal hydraulic analysis of the LEU elements and core. Using ANL codes, the licensee acceptably modeled power peaking factors, thermal conductivity, fluid flow conditions, and fuel and core configurations for the proposed LEU fuel elements and core. .in addition, the licensee incorporated hot channel factors (HCFs) into its calculations to account for fuel and assembly design tolerances and uncertainties. The analyses demonstrated that the LEU fuel elements and core would be cooled and maintained within acceptable limits for forced or natural convection cooling conditions during normal operation. That is, under both forced convection and natural convection coolant flow and associated power conditions, calculated thermal hydraulic conditions for the LEU fuel, would maintain a substantial margin before the onset of nucleate boiling (ONB).

for operation under natural convection flow, the licensee calculated situations in which the control blades were fully withdrawn (leading to a chopped cosine flux distribution) and situations in which the control blades were withdrawn 15 inches. Comparing the results to establish the limiting situation, the licensee found that the most restrictive case is with the control blades withdrawn to 15 inches. In this case, ONB is calculated to occur at about 429 kW for the HEU core and at about 335 kW for the reference LEU core. Therefore, for the reference LEU core, 335 kW is taken as the safety limit for natural convection operation. This is more than three times the licensed operating limit (100 kW).

At power levels greater than 100 kW, the UMLR TSs require the reactor to be operated in the forced convection mode, in which the flow rate is about 1,600gallonsperminuta(gpm)downwardthroughthecore. This flow rate is the same for both the HLU and LEU cores. With the nominal flow rate of 1,600 gpm, the UMLR calculated that ONB would occur at greater than 3.5 MW, or more than 3.5 times the licensed operating power (1 MW) for the LEU core. To determine operational conservatism, the licensee also calculated maximum fuel clad temperature for normal forced convection operation with LEU, giving a calculated maximum fuel clad temperature (including HCFs) of approximately 67 'C. The ONB clad temperature at 1 MW is about 118 'C.

The licensee also analyzed the LEU fuel element and core thermal-hydraulic design for off normal conditions. The analysis for the natural convection conditions demonstrated that the 0.1 MW limit on power operations for natural convection gives significant margin before ONB is reached; that is, under natural convection conditions, ONB would occur at greater than 0.335 MW. The analyses of off-normal conditions for forced convection flow demonstrated that the LEU fuel elements and core would be protected from high-power operations and low flow rates with reactor limiting safety system settings (LSSSs) of 1.25 MW (which is about 125 percent of the normal full-power level) and 1,170

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respectively. With this lower flow rate trip, the licensee calculated that ON8 would occur at ;Jre than 2.5 MW, or more than twice the high power trip setpoint.

These analyses of the thermal hydraulic performance of the LEU core are acceptable, as are the correspondin 3.0 of this safety evaluation (SE))g. TS changes proposed (evaluated in Section i

2.12 RentivityFeedbackCoefficients The licensee asked ANL to calculate the temperature coefficient of reactivity and the void coefficient of reactivity for the HEU and LEU cores. ANL also calculated the coolant density and the Doppler effect in broadening the neutron capture res... nces of the more abundant uranium 238 present in the LEU fuel. ANL's results are as follows: (1) the temperature coefficient for the LEU core would be 0.48 x 10 delta k/k/*C and for the HEU core 0.37 x 10,

(2) the coolant density coefficient for the LEU core would be 0.46 x 10

delta k/k/*C and for the HEU core 0.31 x 10, and (3) the Doppler coefficient for the LEU core would be -0.15 x 10 delta k/k/*C and for the HEU core zero.

The total would be 1.09 x 10 for the LEU core and 0.68 x 10 delta k/k 'C for the HEU core.

ANL calculated that the average void coefficient was approximately 2.26 x 10 delta k/k/ percent void for LEU fuel and -1.59 x 10'8 for HEU fuel.

All reactivity feedback coefficients for the LEU r. ore are calculated to be larger than those for the HEU core and are more a fective in leading to reactor stability. The licensee will also verify the feedback coefficients to be negative and greater than those of the HEU core. Therefore, the licensee has acceptably addressed the reactivity coefficients for conversion to the LEU fuel.

2.13 fission Product Containment and Inventory The cladding is the primary barrier to fission product release for both the HEU and LEU fuels. The cladding and other aspects of fuel construction are described in Section 2.2 of this SE. ANL developed the LEU fuels and extensively tested them under more extreme operational conditions than the UMLR fuels will. experience, in these tests, the performance of the proposed LEU fuel was excellent, comparable to that of HEU fuel, furthermore, use of similar fuel elements and plates in other non power reactors has demonstrated the excellent fission product retention capability of the LEU fuel.

The total inventory of fission products from operating the proposed LEU core i at 1 MW(t) will not differ significantly from that for the HEU core.

Therefore, the previously assumed release of fission products remains valid 1 because it conservatively assumed the release of all fission products from a single fuel plate. However, the fission product inventory in each LEU fuel 4 element and plate will be greater than for the HEU fuel because a core of LEU fuel contains fewer fuel elements and plates for the same power level. The

B-licensee estimated this increased fission product inventory per fuel element and the potential effect of a plate failure. In calculating the release of radioactive materials, the licensee incorporated recent models and assumptions regarding release and dispersal of materials, including dispersion models in Regulatory Guide 1.145. These analyses and evaluations demonstrate that the consequences of the fission product release do not exceed previously l established acceptance criteria.

in evaluating the fission product containment and inventory of the LEU fuel, the licensee and the staff have found no new or significant safety l considerations. Therefore, the proposed operations with LEU fuel are acceptable for containing the expected fission product inventory.

l l 2.14 Potential Accideni Scenarios for the conversion from HEU to LEU, the staff evaluated the refueling accident, the step increase in reactivity event, the continuous withdrawal of a control blade, and the cold water insertion event. All of the other transients, such as a failed fueled experiment, a partial or total loss of water, binding of control blades, release of coolant header gates during

! operation, and cross flow during forced convection operation, were reviewed on the basis of the information in the September 1973 FSAR, as amended, and should not be affected by the HEU to LEU conversion.

2.14.1 Refueling Accident 1he most severe accident that can be envisioned from the conversion from HEU to LEU is the substitution of a fuel assembly for the central flux trap element. The licensee calculates that this error would add ap)roximately 3 percent in reactivity. Since the control blades have more slutdown reactivity than 3 percent, the UMLR would be subcritical until the control blades were moved to a raised position. The reactor would be critical with the control blades in a much lower position than normal, thereby alerting operations personnel to a potential problem. They would be expected to react in accordance with established procedures.

The licensee calculated that dropping a fuel assembly on top of the core would add less than 0.5 percent reactivity, again well within the shutdown capability of the control blades.

The staff finds the analysis of the refueling accident acceptable.

2.14.2 Step Increase in Reactivity UMLR is limited by its TSs to experiments that can add no more than 0.5 percent reactivity under any condition. ANL analyzed the instantaneous insertion of 0.5-percent reactivity while the UMLR was operating at 1 MW. ANL concluded that the maximum peak power would be about 2.8 MW, which is well below the safety limit of about 3.8 MW under these operating conditions. The power transient with the LEU core is less than with the HEU core because of

the larger Doppler coefficient. Since the consequence of this accident is less with LEU fuel than with HEU fuel, the staff finds this analysis acceptable.

2.14.3 Continuous Withdrawal of a Control Blade The UMLR TSs limit the control blade reactivity addition rate to 0.025 percent delta k/k/second; however, the licensee has assumed a continuous withdrawal of 0.035 percent delta k/k/second. On the basis of ANL's calculations, the high power trips would limit the reactor power to about 1.3 MW, which is significantly below the safety limit (3.8 MW). Therefore, the staff finds this accident analysis acceptable.

2.14.4 Cold Water Insertion I The licensee has postulated that the maximum primary temperature decrease would be about 21.7 C if the primary temperature was 43.3 C and the secondary flow of 0 C water went from no flow to 1,600 gpm. The corresponding reactivity change using the calculated negative temperature coefficient for the LEU core of -1.09 x 10 delta k/k/'t would be 40.24-percent delta k/k.

Since this change is less than the 0.5 percent reactivity step input analyzed in Section 2.14.2, the staff finds the licensee's analysis of this accident acceptable.

2.15 Reactor Startuo Testina The licensee plans to make sub critical measurements for *,he LEU fuel loading.

The startup testing progra'n also includes control rod and power claibrations, and temperature coeficient, flux distribution, shutdown margin, excess i

reactivity, and void coefficient measurements. The licensee is to submit a startup report to the NRC on the results of this startup testing. This startup testing therefore, will provide verification of key LEU reactor functions, and, is acceptable.

3.0 CHANGES TO TECHNICAL SPECIFICATIONS 3.) M ministrative format and Editorial Chanagi The licensee proposed to fix typographical errors and misspellings and make other minor editorial and administrative TS changes. The staff has reviewed all of these changes and has determined that they do not alter the meaning or intent of the TSs; therefore, the staff finds them acceptable as included in the revised TSs of this amendment (Amendment No.12).

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3.2 TS 2.1.1.1 " Safety limits under forced convection flow" j The licensee pro)osed to change the safety limits for forced flow and add Figure 2.1, whic1 will replace existing TSs 2.1.1.1 and 2.1.1.2. Figure 2.1 represents the relationship between reactor thermal power and reactor coolant i flow rate and indicates the flow necessary to prevent ONB for a given power l 1evel. Since Figure 2.1 is a more comprehensive way of indicating this safety  !

limit, the staff finds this TS change acceptable. 1 3.3 TS 2.1.2.1 " Safety limits in the natural convection mode"_

The licensee proposed to reduce the specified safety limit for maximum power while the reactor is in the natural convection mode. The licensee proposed this change for the LEU core after calculating the ONB to be ap)roximately l 335 kW when hot channel factors (HCFs) and the 15 inch-control flade withdrawn case are considered.

The staff reviewed this requested change, considers it more conservative, and accepts it. This safety limit will be more than three times the licensed power level for the UMLR in the natural convection mode.

3.4 TS 2.2.1 "Limitino Safety System Settino (LSSSI in the forced convection mode" 1

The licensee proposed to change the LSSS for the coolant flow rate from 1,250 gpm to 1,170 gpm for the LEU core in the forced convection mode. The staff analyzed this change and found it to be appropriate to ensure that automatic protective action would occur to prevent safety limits from being exceeded. With the reactor trip set at a flow rate of 1,170 gpm and a flow coastdown from the nominal operating flow of 1.600 gpm, the control blades would be fully inserted by the time the flow reaches approximately 630 gpm This 630 gpin flow corresponds to a power level for the ONB of a) proximately 1.6 MW, which is about 60 percent higher than licensed power; t1erefore, the safety limit for forced convection would not be violated. in addition, dropping the control blades would reduce the reactor power from 1 MW to approximately 130 kW, well below the safety limit for natural convection (335 kW). Therefore, the staff finds this change to the LSSS for coolant flow to be acceptable.

I 3.5 TS 3.1.3 " Reactivity" The licensee proposed to add a TS requirement that all but five of the peripheral radiation baskets contain flow restricting devices. Further, the licensee proposed to clarify this TS to state that it does not apply in the natural convection mode of operation at low power. The staff finds these additional TS requirements acceptable since they assure that the minimum required quantity of primary coolant flow will be directed to the heat-producing fuel elements in the LEU core, i

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3.6 TS 3.3 " Coolant flow Rate" The Itcensee proposes to change the reactor automatic scram LSSS coolant flow rate from 1,250 gpm to 1.170 gpm. The staff reviewed this change in Section 3.4 of this safety evaluation and finds it acceptable.

3.7 TS 4.2.6 " Reactor Safety System (RSS) Surveillance" The licensee proposed to add a surveillance requirement that any RSS l instrument channel replacement "must undergo a channel calibration before l

routine operation of the reactor after channel installation." This addition defines good practice and can only make the replaced channel more reliable.

The staff finds this addition to the TSs acceptable.

3.8- TS 5.0 "Desian Features" This specification states the characteristics and physical descriptions of the reactor fuel and reactor core. Changes to reflect the LEU fuel and core are discussed in the following sections.

3.8.1 TS 5.1 " Reactor fuel" The licensee proposed to revise this specification to define the design featt.res of tie proposed LEU fuel. The changes are as follows:

(1) The LEU fuel matrix will be U Sin 3 Al instead of the HEU fuel matrix alloy Al U 30..

(2) The uranium 235 enrichment will be approximately 19.75 percent (not to exceed 20 percent) for the LEU instead of the approximately 93 percent enrichment for the HEU.

(3) The cold, clean LEU fuel elements will be 18 plates, with 16 containing approximately 200 grams of uranium 235 and 2 outside ,

aluminum plates, instead of the HEU elements, which had consisted of 18 plates containing approximately 135 grams of uranium 235.

These changes are required by the LEU fuel design, which has been previously discussed in this SE and has been demonstrated by ANL to be acceptable; therefore, the staff finds these changes acceptable.

-3.8.2 TS 5.2 " Reactor Core" The licensee proposed to revise this specification to describe the proposed design features of the LEU core. The LEU reference core should consist of 20 standard elements, with the central location filled with a graphite water, aluminum clad-flux trap instead of the HEU reference core, which consists of 26 standard HEU fuel elements. The proposed LEU cores may contain from 16 to

. }2 28 LEU elements and may contain 2 half-loaded elements, in contrast to the HEU core, which could contain 23 to 30 HEU elements and 2 half-loaded elements.

These proposed changes to the reactor core design features have been discussed in Section 2.3 of this SE. The staff finds these changes acceptabic.

3.8.3 IS_jt.4 " Fuel Storang" .

The licensee proposed to revise this TS to allow reactor fuel element storage of LEU fuel in a geometrical configuration so the K.tr is less than 0.85 under quiescent flooding with water instead of a K.it of 0.80 for HEU fuel. This change was discussed in Section 2.4 of this SE and is required by the uranium-235 loading of the LEU fuels. ANL's calculations indicata ' hat K.rr of an infinite array of LEU fuel loaded with 225 grams of uranium 235 would be less than 0.80. Because the UMLR fuel is within the envelope of the AHL calculations, the staff finds this change acceptable.

4.0 CONCLUSION

The NRC staff reviewed and evaluated all of the operational and safety factors affected by the use of LEU fuel in place of HEU fuel in the UMLR. The staff concludes that the conversion, as proposed, would not reduce any safety margins, would not introduce any new safety issues, and would not lead to increased radiological risk to the health and safety of the public.

Principal Contributors: Theodore S. Michaels, NRC James R. Hiller, INEEL/LMITC0 Robert E. Carter, ENEEL/LMITC0 Date: July 31, 1997 I

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