ML20056E570

Informs Commission Re Order Issued to Convert non-power Reactor from HEU to LEU Fuel,In Accordance w/10CFR50.64

ML20056E570
Person / Time
Issue date:	07/26/1993
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
SECY-93-208, NUDOCS 9308240280
Download: ML20056E570 (52)

Text

{{#Wiki_filter:. ,-..-..~...m...-: e CD TO THE PDR j 4 f ""%s -{t4[g-71 l f i n i e p g -y $1 a c 8 t s %.....,o i POLICY ISSUE July 26, 1993 (lnfOUnation) SECY-93-208 i f0R: The Commissioners 0 i FROM: James M. Taylor Executive Director for Operations I l

SUBJECT:

ORDER MODIFYING LICENSE TO CONVERT FROM HIGH-ENRICHED TO l LOW-ENRICHED URANIUM FUEL (UNIVERSITY OF VIRGINIA POOL j REACTOR) f 1 l PURPOSE: i i l l To inform the Commission about an order issued to convert a non-power reactor i from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel in accordance with Section 50.64 of Title 10 of the Code of Federal Reculations l (10 CFR 50.64). l } DISCUSSION: l } l l As cited in 10 CFR 50.64, conversion from the use of HEU to LEU fuel is 6 required if suitable fuel and funding are available through the U.S. Department of Energy or another Federal agency. In accordance with i COMLZ-87-43, wherein the Commission asked the Office of Nuclear Reactor i Regulation (NRR) to tell it of subsequent conversion orders, the staff i i informed the Commission when it issued orders to convert from HEU to LEU in l SECY-87-171 (Rennselaer Polytechnic Institute), SECY-89-Oll (Ohio State University and Worcester Polytechnic Institute), SECY-90-184 (Manhattan i j College), SECY-90-402 (Iowa State University), SECY-91-122 (University of l Missouri, Rolla), and SECY-93-IS4 (Rhode Island Atomic Energy Commission). All of these non-power reactors are operating with LEU fuel except the Rhode Island Atomic Energy Commission, which expects to convert the reactor to LEU l fuel in the summer of 1993. In SECY-90-184, the staff summarized the status j of the conversions from HFU to LEU fuel. l l NOTE: TO BE MADE PUBLICLY AVAILABLE l IN 10 WORKING DAYS FROM THE l CONTACT: {h 04 nn-l 9308240280 930726 PDR SECY - ~ I t 93-20B PDR i' I

~ f o-f The Commissioners. Enclosed is the order that NRR issued to the University of Virginia on April 29, 1993. The order will become effective when the replacement LEU core is received, which is expected to be in the summer of 1993. The required 30-day period for request of hearing has expired and no request has been filed. t COORDINATION: The Office of the General Counsel has reviewed this paper and has no legal objection to it. { i l' l \\ M t <- /s s ? James M. Tay) r j ,rixecutive Director j for Operations j Enclosure. Order Modifying License for the University of _[ Virginia Pool Reactor i t i DISTRIBUTION: Conunissioners OGC OCAA OIG OCA OPP REGION II i EDO r SECY l, I i i 1 l I

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.I I-c, UNITED STATES i NUCLEAR REGULATORY COMMISSION f wAssiNGToN, D.C. 20556 '...+" April 29,1993 Docket No. 50-62 Dr. R. U. Mulder, Director Nuclear Reactor Facility Department of Mechanical, Aerospace, and Nuclear Engineering University of Virginia Charlottesville, Virginia 22903-2442

SUBJECT:

ISSUANCE OF ORDER MODIFYING LICENSE R-66 TO CONVERT FROM HIGH-TO LOW-ENRICHED UPANIUM (AMENDMENT 20) - UNIVERSITY OF VIRGINIA POOL REACTOR (TAC NO. M82824) The U.S. Nuclear Regulatory Commission (NRC) is issuing an order modifying Facility Operating License R-66, Amendment 20, for the University of Virginia pool reactor. The order (Enclosure 1) modifies the license in accordance with Section 50.64 of Title 10 of the Code of Federal Reculations (10 CFR), which requires that a non-power reactor such as yours convert to low-enriched uranium (LEU) fuel under certain conditions. The order is being issued in accordance with 10 CFR 50.64(c)(3) and in response to your submittal of November 9,1989, as supplemented on February 12, 1991, and December 14, 1992. contains the changes to technical specifications, and Enclosure 3 is the staff Safety Evaluation Report. The order contains an outline of a reactor startup report that you are required to provide to the NRC within six months following completion of LEU fuel loading. The order will become effective on the later date of either receipt of the replacement core of LEU fuel elements or 30 days following the date of publication of this order in the Federal Reaister, provided there are no requests for a hearing. If a hearing is requested, the order will become effective on the date specified in an order following further proceedings on this order. Please inform me when you receive LEU fuel elements and when the high-enriched uranium fuel is completely removed from your facility. Also, License Condition 2.B.2. has been amended to allow possession, but not separation, of special nuclear material (SNM) produced by operation of the reactor. This condition was implied under the terms of the license but was

.a Dr. R. U. Mulder April 29, 1993 not stated explicitly. Because the normal operation of the reactor produces SNM, license Condition 2.B.2. has been clarified to include SNM that is produced in the fuel. A copy of the Safety Evaluation Report and replacement pages for the technical specifications is enclosed with the order. Sincerely, CRIGINAL SIGED BY: Alexander Adams, Jr., Sr. Project Manager Non-Power Reactors and Decommissioning Project Directorate Division of Operating Reactor Support Office of Nuclear Reactor Regulation

Enclosures:

1. Order / Amendment No. 20 2. Replacement Pages for Technical Specifications 3. Safety Evaluation Report { cc w/ enclosures: I t See next page DISTRIBUTION: [ AMEND 20.AA] [ Adams-lb diskette] Docket file AAdams IDinitz PDRs OGC DHagan l ONDD R/F GHill (2) DNash t TMurley/FMiraglia ACRS (10) HSmith JPartlow BGrimes JKopeck, PA OGC WLambe JStohr, RII EHylton Region II CGrimes SWeiss DC/LFDCB

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1 Dr. R. U. Mulder April 29,1993 not stated explicitly. Because the normal operation of the reactor produces SNM, License Condition 2.B.2. has been clarified to include SNM that is produced in the fuel. l A copy of the Safety Evaluation Report and replacement pages for the technical specifications is enclosed with the order. L Sincerely, O Alexander Adams, Jr., roject Manager Non-Power Reactors and ommissioning Project Directorate Division of Operating Reactor Support [ Office of Nuclear Reactor Regulation

Enclosures:

1. Order / Amendment No. 20 2. Replacement Pages for Technical Specifications 3. Safety Evaluation Report cc w/ enclosures: See next page i ) l 4 I

~ University of Virginia Docket Nos. 50-62/396 cc: Commonwealth of Virginia Council on the Environment 903 Ninth Street Office Bldg. Richmond, Virginia 23219 Mr. Preston Farrar Nuclear Reactor Facility University of Virginia School of Engineering and Applied Science Charlottesville, Virginia 22903 Dr. Albert B. Reynolds, Chairman Department of Nuclear Engineering and Engineering Physics University of Virginia + Charlottesville, Virginia 22901 Dr. William Vernetson Director of Nuclear Facilities Department of Nuclear Engineering Sciences University of Florida 202 Nuclear Sciences Center Gainesville, Florida 32611 Dr. Ratib A. Karam, Director Neely Nuclear Research Center Georgia Institute of Technology 900 Atlantic Drive, N.W. Atlanta, Gecrgia 30332 r Hr. Pedro B. Perez, Associate Director Nuclear Reactor Program North Carolina State University P. O. Box 7909 Raleigh, North Carolina 27695-7909 Office of the Attorney General 101 North 8th Street Richmond, Virginia 23219 Bureau of Radiological Health Division of Health Hazards Control 109 Governor Street, Room 916 Richmond, Virginia 23219

l UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION f in the Matter of ) ) UNIVERSITY OF VIRGINIA ) Docket No. 50-62 ) Facility Operating License No. R-66 i (University of Virginia Pool ) Reactor) ) Amendment No. 20 i ) i ORDER MODIFYING LICENSE I. The University of Virginia (the licensee or UVA) is the holder of Facility Operating License R-66 issued on June 24, 1960, and subsequently j renewed on November 4, 1971, by the U.S. Atomic Energy Commission and by the U.S. Nuclear Regulatory Commission (the Commission or NRC) on September 30, 1982. The license authorizes operation of the UVA pool reactor at a power level up to 2 megawatts thermal (MWt). The research reactor is located at the UVA Nuclear Reactor Facility, which is about 700 meters west of the city i limits of Charlottesville, Virginia. The mailing address is Nuclear Reactor Facility, Department of Mechanical, Aerospace, and Nuclear Engineering, University of Virginia, Charlottesville, Virginia 22903-2442. II. On February 25, 1986, the Commission promulgated a final rule in Section 50.64 of Title 10 of the Code of Federal Reaulations (10 CFR 50.64) limiting the use of high-enriched uranium (HEU) fuel in domestic research and test reactors (non-power reactors) (see 51 FR 6514). The rule, which became 4

i i effective on March 27, 1986, requires that each licensee of a non-power reactor replace HEU fuel at its facility with low-enriched uranium (LEU) fuel i i i acceptable to the Commission. This conversion, which is contingent on Federal Government funding for conversion-related costs, is required unless the Commission has determined that the reactor has a unique purpose. The rule is intended to promote the common defense and security by reducing the risk of j theft and diversion of HEU fuel used in non-power reactors and the adverse i i consequences to public health and safety and the environment from such theft i or diversion. Sections 50.64(b)(2)(i) and (ii) require that a licensee of a non-power i reactor (1) not initiate acquisition of additional HEU fuel, if LEU fuel that l 2 j ic acceptable to the Commission for that reactor is available when the f licensee proposes that acquisition, and (2) replace all HEU fuel in its ] possession with available LEU fuel a"eptable to the Commission for that reactor in accordance with a schedule determined pursuant to 10 CFR ? 50.64(c)(2). Section 50.64(c)(2)(i) of the rule, among other things, requires each licensee of a non-power reactor, authorized to possess and to use HEU fuel, to f i ] develop and to submit to the Director of the Office of Nuclear Reactor i Regulation (Director) by March 27,1987, and at 12-month intervals thereafter, i a written proposal (proposal) for meeting the rule requirements. I Section 50.64(c)(2)(i) also requires the licensee to include the following in its proposal: (1) a certification that Federal Government fun' ding for conversion is available through the U.S. Department of Energy [ l (DOE) or other appropriate federal agency and (2) a schedule for conversion, L ~ .r w-

t t l i ! i i based on availability of replacement fuel acceptable to the Commission for that reactor and upon consideration of other factors such as the availability i of shipping casks, implementation of arrangements for the available financial support, and reactor usage. Section 50.64(c)(2)(iii) requires the licensee to include in its proposal, to the extent required to effect conversion, all necessary changes i to the incense, to the facility, and to licensee procedures (all three types i' of changes hereafter called modifications). This paragraph also requires the licensee to submit supporting safety analyses so as to meet the schedule established for conversion. Section 50.64(c)(2)(iii) also requires the Director to review the licensee proposal, confirm the status of Federal Government funding, and determine a final schedule if the licensee has submitted a schedule for conversion. Section 50.64(c)(3) requires the Director to review the supporting safety analyses and issue an appropriate Enforcement Order directing both the conversion and any necessary modifications to the extent consistent with protection of the public health and safety. In the Federal Reaister notice of the final rule, the Commission explained that in most cases, if not all, the Enforcement Order would be an Order to modify the license. Section 2.202, the current authority for issuing Orders of all types including Orders to modify licenses, provides, among other things, that the Commission may modify a license by serving an Order on the licensee. The licensee may demand a hearing concerning any part or all of the Order Modifying License within 20 days from the date of the notice or such longer

? _4-i i period as the notice may provide. The Order will become effective on the i expiration of this 20-day or longer period, unless the licensee requests a hearing during this period, in which case the Order will become effective on the date specified in an Order made after the hearing, f Section 2.714 gives the requirements for a person whose interest may be f affected by any proceeding to initiate a hearing or to participate as a party. III. i On November 9, 1989, as supplemented on February 12, 1991, and December 14, 1992, the NRC staff received the licensee proposal, including its l proposed modifications, supporting safety analyses, and schedule for [ conversion. The conversion consists of replacing high-enriched with low-enriched uranium fuel elements. The fuel elements contain materials testing I l reactor (liTR)-type fuel plates, with the fuel meat consisting of uranium i silicide dispersed in an aluminum matrix. These plates contain an enrichment of less than 20 percent with the uranium-235 isotope. The attachments to this l Order include (1) the changes to the licensing conditions and technical l specifications that are needed to amend the facility license and (2) the outline of the startup report that is required to be submitted within six i months following completion of LEU fuel loading. The NRC staff has reviewed i the licensee submittals and the requirements of 10 CFR 50.64 and determined i that the public health and safety and the common defense and security require the licensee to convert the facility from the use of HEU to LEU fuel, pursuant to the changes to the license and requirements for a startup report stated in the attachments to this Order, in accordance with the schedule included herein. I

. l i IV. i Accordingly, pursuant to Sections 51, 53, 57, 101, 104, 161b., 1611., and 1610. of the Atomic Energy Act of 1954, as amended, and to Commission i I regulations in 10 CFR 2.202 and Section 50.64, IT IS HEREBY ORDERED THAT* l On the later date of either receipt of the replacement core of LEU fuel l l elements by the licensee or 30 days following the date of publication of this i Order in the Federal Reaistet, facility Operating License R-66 is modified by amending the license conditions and technical specifications as stated in the r " Attachment to Order Hodifying Facility Operator License R-66." The licensee l i shall submit the startup report as stated in the " Attachment to Order of the l j Outline of Reactor Startup Report" within six months following completion of i I LEU fuel loading, V. j j Pursuant to the Atomic Energy Act of 1954, as amended, the licensee or any other person adversely affected by this Order may request a hearing within i 30 days of the date of this Order. Any request for a hearing shall be submitted to the Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Assistant General Counsel for Hearings and Enforcement at the same address. If a person l other than the licensee requests a hearing, that person shall set forth with particularity in accordance with 10 CFR 2.714 the manner in which their interest is adversely affected by this Order. l j 4 ( i i

. If a hearing is requested by the licensee or a person whose interest is adversely affected, the Commission shall issue an Order designating the time and place of any hearing. If a hearing is held, the issue to be considered at i such hearing is whether this Order should be sustained. i This Order shall become effective on the later date of either the receipt of the replacement core of the LEU fuel elements by the licensee or 30 days following the date of publication of thi, Order in the Federal Reaister or, if a hearing is requested, on the date specified in an Order after further proceedings on this Order. FOR THE NUCLEAR REGULATORY COMMISSION Thomas E. Murley, Director Office of Nuclear Reactor Regulation i Dated at Rockville, Maryland l this 29th cay of March Attachments-As stated l I i

i ATTACHMENT TO ORDER [ MODIFYING FACILITY OPERATING LICENSE R-66 l A. License Conditions Revised and Added by This Order l l II.B(2) Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of i Special Nuclear Material," to receive, possess, and use up to a maximum of 12 kilograms of contained uranium-235 at various enrichments, up to a maximum of 16 grams of plutonium in the form of a sealed plutonium-beryllium neutron source in connection with operation of the reactor, and to possess, but not separate, such special nuclear material as may be produced by the operation of i the facility. Without exceeding the foregoing maximum possession limits, the maximum limits on specific enrichments of U-235 are l as follows: i i Maximum U-235 j Kiloarams % Enrichment Form 11 < 20% Naterials testing reactor (MTR)-type fuel 1 Any Fission chambers, flux foils, and other forms used in connection with operation of the reactor II.B(4) Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to possess, but not use, a maximum of 1 5.0 kilograms of contained uranium-235 at greater than 20- ? percent enrichment and other such special nuclear material produced by operation of the facility in the form of MTR-type l reactor fuel until the existing inventory of high-enriched MTR-type reactor fuel is removed from the facility. i II.C(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 20, are hereby incorporated in the l license. The licensee shall operate the facility in accordance with the Technical Specifications. j i i I l i i 1

1 1 . B. Technical Specifications Revised by This Order 2.1.1. Safety Limits in Forced Convection Mode of Operation Ap21kability: This specification applies to the interrelated variables associated with pore thermal and hydraulic performance in the forced convection mode of operation. These variables are: P = Reactor thermal power W = Reactor coolant flow rate T, = Reactor coolant inlet temperature L = Height of water above the core Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. Specification: In the forced convection mode of operation: (1) The pool water level shall not be less than 19 ft above the top of the core. (2) The reactor coolant inlet temperature shall not be greater than 11l'F. t (3) The true value of reactor coolant flow shall not be below $75 gpm. (4) The combination of true values of reactor core power and reactor coolant flow shall be below the line dermed by. P = 0.24 + (4.5 x 108 * %) P = 0 for W < 575; P in MW, W in gpm t The allowed region of operation is shown by the unshaded region of Figure 2.1. ~ f Buis: Above 575 gpm in the region of full power operation, the criterion used to establish the safety limit was a burnout ratio of 1.49 including the worst variation in the manufacturer's tolerance and specification, hot channel factors and other appropriate uncertainties. The analysis is given in the LEU SAR. Below 575 gpm buoyancy forces competing with forced convection may lead to flow instabilities in some of the channels and is therefore not allowed. The analysis of the loss of flow transient shows that during the transition from forced convection to natural convection following a loss of flow and reactor scram that the fuel temperature is well below the temperature at which fuel clad damage could occur. i -- ~.

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a 4-L 2.1.2. Safety Limits in the Natural Convection Mode of Operation Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the natural convection mode of operation. These variables are: P = Reactor thermal power T = Reactor coolant inlet temperature 3 Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. i Specification: In the natural convection mode of operation: (1) The true value of reactor power shall not exceed 750 kW. (2) The reactor coolant inlet temperature shall not be greater than !!!*F. Basis The criterion for establishing a safety limit with natural convection flow is established as a fuel plate temperature. The analysis for natural convection flow shows that at 750 kW, the maximum fuel plate temperature is well below the temperature at which fuel clad damage could occur. 2.1.3. Saferv Limit for the Transition from Forced to Natural Convection Mode of Oneration Aonlicability: This specification applies to the condition when the reactor is in transition from forced convection flow to natural convection flow. Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. Specification: The current to the control rod magnets must be off when the reactor is making a transition from forced to natural convection. Basis The safety analysis of the loss of coolant transient demonstrates that the fuel plate temperature is maintained well below the temperature at which fuel clad damage could occur during the transition from forced downflow through flow reversal to the establishment of natural convection provided that the loss of flow transient is accompanied by a scram.

l 2.2. Limitine Safety System Settines Applicability: These specifications apply to the set points for the safety channels l monitoring reactor thermal power, coolant flow rate, reactor coolant inlet temperature, and the height of water above the core. Objective: The objective is to ensure that automatic protective action is initiated to prevent the safety limit from being exceeded. Specifications: 2.2.1. Forced Convection Mode For operation in the forced convection mode, the limiting safety system settings shall be: Reactor Dermal Power 3.0 MWt (max) = Reactor Coolant Flow Rate 900 gpm (min) = Reactor Coolant Inlet Temperature = 108'F (max) Height of Water above Core 19'2" (min) = Reactor Period 3.3 sec (min) = 2.2.2. Natural Convection Mode For operation in the natural convection mode, the limiting safety system settings shall be: Reactor Power = 300 kWt (max) Reactor Coolant Inlet Temperature = 108'F (max) Reactor Period , = 3.3 see (min) Bases The analysis in the LEU SAR shows there is sufficient margin between these settings and the safety limit under the most adverse conditions of operation: (2.2.1.) For the forced' convection mode, the LEU SAR considers accidents with reactor power at 3.45 MW, a period of 3 seconds, pool inlet temperature of 11l'F and a coolant flow of 837 gpm. The maximum fuel plate temperature calculated was considerably below the aluminum clad melting point. The LSSS specified above for this mode of operation are more consenative than the parameters used in the LEU SAR analysis. 1 (2.2.2.) With natural convection flow, there is no minimum coolant flow rate and no minimum height of water above the core so long as there is a path for flow (see Section 3.8 of these specifications). The LEU SAR shows that the maximum fuel plate temperature under natural convection with initial power of 750 kW and pool inlet temperature of 11l'F was well below the aluminum clad melting point. ne LSSS specified above for this mode of operation are below the analyzed condition.

. 3.2. Reactor Safety System Applicability: This specification applies to the reactor safety system channels. Objective: The objective is to stipulate the minimum number of reactor safety system channels that must be operable to ensure that the safety limit is not exceeded during normal operation. St>ecification: The reactor shall not be operated unless the safety system channels described in Table 3.1 Safety System Channels are operable. Bases The startup interlock, which requires a neutron count rate of at least 2 counts per second (CPS) before the reactor is operated, ensures that sufficient neutrons are available for proper operation of the startup channel. The pool-water temperature scram provides protection to ensure that if the limiting safety system setting is exceeded an immediate shutdown will occur to keep the fuel temperature below the safety limit. Power level scrams are provided to ensure that the reactor poser is maintained within the licensed limits and to protect against abnormally high fuel temperatures. The manual scram allows the operator to shut down the reactor if an unsafe or abnormal condition arises. The period scram is provided to ensure that the power level does not increase above that described in the SAR. Specifications on the pool-water level are included as safety measures in the event of a serious loss of primary water. Reactor operations are terminated if a major leak occurs in the primary system. The analysis in the SAR shows the consequences resulting from loss of coolant. The bridge radiation monitor gives warning of a high radiation level in the reactor room from failure of an experiment or from a significant drop in pool-water level. A scram from loss of primary coolant flow, loss of power to the pump, or application of power to the pump when operating in the natural convection mode, protects the reactor. fro'm overheating. Air pressure to the header above ambient results in a scram to:

1) Ensure that the header falls with loss of primary pump power when the reactor is operating in the forced convection mode.

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2) Prevent raising the header when the reactor is in the natural convection mode.

3) Avoid producing additional Ar-41 by activating air introduced into the header.

l 1 l l ( TABLE 3.1 SAFETY SYSTEM CHANNELS l 4 i Minimum Operating Mode i i Measuring Channel Set Point

• Function i

No. Operable Requ d ire i ! Pool water level monitor 2 19'2* (min) Scram Forced convection i l Bridge radiation monitor 1 30 mr/hr Scram All modes I Pool water temperature 1 108'F (max) Scram All modes j loss of power Scram Forced convection l Power to primary pump 1 application of Natural i Scram I power convection l i ' Primary coolant flow I 900 rpm (min) Scram Forced ecovection Prevents l Strrup count rate 1 2 eps (min) withdrawal of Reactor startup any shim rod L I Manual button 1 Scram All modes 4 I 3 MWt (max) Scram Forced convection l Reactor power level 2 f Natural 0.3 MWt (max) Scram j convection 1 Reactor period . 1 3.3 see (min) Scram All modes t Air pressure to header I above ambient Scram All modes Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values.

• ~

4.S. Reactor HEU Fuel Dose Measurements Applicability: This specification applies to the highly enriched uranium (HEU) UVAR fuel possessed under the Reactor Facility license. These specifications are applicable until all HEU UVAR fuel elements have been removed from the Reactor Facility. Objective: The objective of this specification is to ensure that the maximum quantity of special nuclear material does not exceed the limits specified in the Reactor Facility license. Specifications: 4.8.1. Schedule The amount of special nuclear material (Sh31) possessed at the Reactor Facility will be determined, as necessary, to ensure that limits specified by the Reactor Facility licenses are not exceeded. As a minimum, an evaluation will be completed and documented every 6 months. 4.8.2. Ouantity Limits HEU UVAR fuel elements possessed following the conversion of the UVAR to LEU fuel will be shipped away from the Reactor Facility, as necessary, to ensure that the quantity of nonexempt Sh31 (as defined in 10 CFR 73) does not exceed that allowed by the Reactor Facility licenses. If the amount of nonexempt Sh31 exceeds 5 kg the Reactor Safety Committee will be informed and the actions specified in the Physical Security Plan implemented. 4.8.3. PSelf-Protection" Determinations If HEU UVAR fuerelements have not been irradiated as a pan of the UVAR core for at least one month, dose rate measurements of these HEU fuel elements will be made, as necessary, to determine which elements have dose rates higher than specified by 10 CFR 73.67(b). Bases: The specifications provide a high degree of assurance that the amount of Sh31 and nonexempt Sh3i will not exceed the license limits. The amount of 1 nonexempt Sh31 will normally be maintained at less than 5 kg, if necessary by j shipping spent-fuel off-site. In the event that the 5 kg nonexempt SNM quantity is ' exceeded, the Reactor Safety Committee will be informed of this and the actions specified in the Physical Security Plan will be taken. i l

-g-5.1. Reactor Fuel Specifications Applicability: These specifications apply to UVAR low enriched uranium (LEU) fuel Objective: The objective is to describe LEU fuel approved by the U.S. NRC for use in the UVAR. Srecifications: 5.1.1. Fuel Material UVAR LEU fuelis of a type described for use at U.S. research reactors by the U.S. Nuclear Regulatory Commission (NUREG-1313 " Safety Evaluation Report Related to the Evaluation of LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors"). The fuel meat is U Si dispersed in an aluminum matrix 3 2 and enriched to less than 20% U-235. 5.1.2. Element Descriotion (1) Plate-type elements of the MTR type are used. The fuel " meat" is clad with aluminum alloy to form flat fuel plates. The active length of the fuel region in the fuel plates is approximately 24 inches and the width is approximately 2.5 inches. The LEU fuel plates are joined at their long-side edges to two side i plates. The entire fuel plate assembly is joined at the bottom to a cylindrical nose piece that fits into the UVAR core gridplate. The overall fuel element dimensions are approximately 3 inches by 3 inches by 36 inches. Each fuel plate i contains 12.5 grams of U-235. t 5 (2) " Standard" LEU fuel elements are composed of 22 parallel flat fuel plates each, and contain 275 grams of U-235. i (3)

• Control-rod" LEU elements are similar to the standard elements, with the i

exception that they have half as many fuel plates (the 11 center plates being removed to form a channel which is bounded by 0.125 inch thick aluminum plates). Control-rod elements accommodate the control rods in the central channel. Their U-235 content is 137.5 grams. I (4) " Partial" LEU fuel elements are half-fueled elements composed of 11 LEU fuel plates and 11 unfuelled (dummy) plates. ne U-235 content in these elements is 137.5 grams. (5) "Special* LEU fuel elements have 22 fuel plates, of which 20 are removable, The maximum U-235 content in these elements is 275 grams and the minimum is 25 grams. e

. i 5.1.3. Core Configurations A variety of UVAR core configurations may be used to accommodate experiments, but the loadings shall always be such that the minimum shutdown margin and excess reactivity specified in the UVAR Technical Specifications are not exceeded. Bases The NRC has described LEU silicide-fuel suitable for use in U.S. research reactors in NUREG-1313

• Safety Evaluation Report Related to the Evaluation of i

LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors,' [536.00, from NTIS, Sprir.; field Va. (703-487-4650)). Also, Bretscher and Snelgrove from i the Argonne National 12boratory documented LEU fuel test results in ANL/RERTR/TM-14, 'The Whole-Core LEU U Si -Al Fuel Demonstration in the 3 2 30-MW Oak Ridge Research Reactor." The LEU-SAR for the UVAR contains the safety analysis performed for the 22 flat-plate University of Virginia fuel elements. The LEU elements were designed by EG&G, Idaho, and are manufactured by the Babcock and Wilcox Company of Lynchburg, Virginia. i 1 i Y e 1

ATTACHMENT TO ORDER OF OUTLINE OF REACTOR STARTUP REPORT Within six months following completion of initial LEU core loading, submit the following information to NRC: 1. Critical Mass Measurement with HEU Measurement with LEU Comparisons with calculations for LEU and if available, HEU 2. Excess (operational) reactivity Measurement with HEU Measurement with LEU Comparisons with calculations for LEU and if avcilable, HEU 3. Control and regulating rod calibrations Measurement of HEU and LEU differential and total rod worths and comparisons with calculations for LEU and if available, HEU 4. Reactor power calibration Methods and measurements that ensure operation within the license limit and comparison between HEU and LEU nuclear instrumentation setpoints, detector positions, and detector output 5. Shutdown margin Measurement with HEU Measurement with LEU Comparisons with calculations for LEU and if available, HEU 6. Partial fuel element worths for LEU Measurements of the worth of the partial loaded fuel elements 7. Thermal neutron flux distributions Measurements of the core and measured experimental facilities with HEU and LEU and comparisons with calculations for LEU and if available, HEU 8. Results of determination of LEU effective delayed neutrons fraction, temperature coefficient, and void coefficient to the extent that measurements are possible and comparison with calculations and available HEU core measurement

r 9. Discussion of the comparison of the various results including an explanation of any significant differences that could affect both normal operation and possible accidents with the reacto, i

10. Measurements made during initial loading of the LEU fuel, presenting subtritical multiplication measurements, predictions of multiplication for next fuel additions, and prediction and verification of final criticality conditions

11. Results of LEU flow coast down measurements I

Comparison against available HEU core measurements and LEU predictions

12. Results of pool water sample measurements for fission product activity during the first 30 days of LEU operation i

i d f i l t 4 t

ATTACHMENT TO LICENSE AMENDMENT NO. 20 FACILITY OPERATING LICENSE NO. R-66 DOCKET NO. 50-62 Replace the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change. Remove Paaes Insert Paaes 4 4 5 5 6 5A 8 58 i 9 6 19 8 20 8A 21 8B 9 19 19A 20 21 21A 21B E s t t i

UVAR Tech. Specs. 2.0. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS f 2.1. Safety Limits 2.1.1. Safety Limits in Forced Convection Mode of Oneration Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the forced convection mode of l operation. These variables are. P = Reactor thermal power W = Reactor coolant flow rate T = Reactor coolant inlet temperature i L = Height of water above the core Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. Specification: In the forced convection mode of operation: (1) The pool water level shall not be less than 19 ft above the top of the l j core. i (2) The reactor coolant inlet temperature shall not be greater than 111*F. (3) The true value of reactor coolant flow shall not be below 575 gpm. (4) The combination of true values of reactor core power and reactor coolant flow shall be below the line defined by: P = 0.24 + (4.5 x 104

• W)

P = 0 for W < 575; P in MW, W in gpm i The allowed region of operation is shown by the unshaded region of Figure 2.1. Basis: Above 575 gpm in the region of full power operation, the criterion used i to establish the safety limit was a bumout ratio of 1.49 including the worst variation in the manufacturer's tolerance and specification, hot channel factors 1 and other appropriate uncertainties. The analysis is given in the LEU SAR. Below 575 gpm buoyancy forces competing with forced convection may lead to flow instabilities in some of the channels and is therefore not allowed. The analysis of the loss of flow transient shows that during the transition from forced convection to natural convection following a loss of flow and reactor scram that l the fuel temperature is well below the temperature at which fuel clad damage could cecur. 4 Amendment No. 20 i

-~ UVAR Tech. Specs. _ooco oo .w og oo O e.* See e .o e4 u o u o M M -n u > o e-Ou

x o

. G G u o b o o o M b u .a o u W e w n a n a u n ..e o - .e ,o o co o J u A u a o G u w u = n m G e ac _g / N G 4 E o n -o w o .o N I I I I i 1 / i i i i i i so o e in n N o MH *d '28 Mod 1752841 Joaorag 5 /cendment No. 20

UVAR Tech. Specs. 2.1.2. Safety Limits in the Natural Convection Mode of Oneration Applicability: This specification applies to the interrelated variables associated l with core therma] and hydraulic performance in the natural convection mode of l operation. These variables are: P = Reactor thermal power T = Reactor coolant inlet temperature f i Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. t Specification: In the natural convection mode of operation: (1) The true value of reactor power shall not exceed 750 kW. (2) The reactor coolant inlet temperature shall not be greater than 111*F. Basis The criterion for establishing a safety lirait with natural convection flow is established as a fuel plate temperature. The analysis for natural convection flow shows that at 750 kW, the maximum fuel plate temperature is well below the temperature at which fuel clad damage co2ld occur. (rest of page intentionally left blank) t L P i l ) i SA Amendment No. 20

UVAR Tech. Specs. 2.1.3. 52fety Limit for the Transition from Forced to Natural Convection Mode of Operation App!icability: This specification applies to the condition when the reactor is in transition from forced convection flow to natural convection flow. Obiective: The objective is to ensure that the integrity of the fuel clad is maintained. l Specification: ne current to the control rod magnets must be off when the reactor is making a transition from forced to natural convection. Basis: The safety analysis of the loss of coolant transient demonstrates that the fuel plate temperature is maintained well below the temperature at which fuel clad damage could occur during the transition from forced downflow through flow reversal to the establishment of natural convection provided that the loss of flow transient is accompanied by a scram. (rest of page intentionally left blank) f i ~ .1 e e. 5 5B Amerriment No. 20

UVAR Tech. Specs. 2.2. Limitine Saferv System Settines Applicabilitv: These specifications apply to the set points for the safety channels monitoring reactor thermal power, coolant flow rate, reactor coolant inlet temperature, and the height of water above the core. Objective: The objective is to ensure that automatic protective action is initiated to prevent the safety limit from being exceeded. Specifications: 2.2.1. Forced Convection Mode l For operation in the forced convection mode, the limiting safety system settings shall be: Reactor Thermal Power 3.0 MWt (max) = Reactor Coolant Flow Rate 900 gpm (min) = Reactor Coolant inlet Temperature = 1087 (max) Height of Water above Core 19'2" (min) = Reactor Period. 3.3 see (min) = i 2.2.2. Natural Convection Mode For operation in the natural convection mode, the limiting safety system settings shall be: Reactor Power = 300 kWt (max) Reactor Coolant Inlet Temperature = 108'F (max) Reactor Period = 3.3 sec (min) Bases The analysis in the LEU SAR shows there is suf5cient margin betw.:en these I settings and the safety limit under the most adverse conditions of operation: (2.2.1.) For the forcedIonvection mode, the LEU SAR considers accidents with reactor power at 3.45 MW, a period of 3 seconds, pool inlet temperature of 111'F and a coolant flow of 837 gpm. The maximum fuel plate temperature calculated was considerably below the aluminum clad melting point. The LSSS specified above for this mode of operation are more conservative than the parameters used in the LEU SAR analysis. (2.2.2.) With natural convection flow, there is no minimum coolant flow rate and no minimum height of water above the core so long as there is a path for flow l (see Section 3.8 of these specifications). The LEU SAR shows that the i maximum fuel plate temperature under natural convection with initial power of 750 kW and pool inlet temperature of 1117 was well below the aluminum clad melting point. The 13SS specified above for this mode of operation are below the analyzed condition. i 6 Amendment No. 20

Operation of the reactor at a power of less than 1 kW is allowed to measure the reactivity worth of untried experiments, in accordance with procedures approved by the Reactor Safety Committee, and to measure the excess reactivity of new core loadings. i Tne limit of 5% Ak/k on excess reactivity is to allow for xenon override and operational flexibility and to ensure that the operational reactor is reasonably { similar in configuration to the reactor core analyzed in the SAR. In general the excess reactivity is limited by the shutdown margin requirement. i r I l t i 1 i h i i i I l i 1 i I a 8 Amendment No. 20

I UVAR Tech. Specs. 3.2. Reactor Safety System Applicability: This specification applies to the reactor safety system channels. Obiective: The objective is to stipulate the minimum number of reactor safety system channels that must be operable to ensure that the safety limit is not exceeded during normal operation. Specification: The reactor shall not be operated unless the safety system channels described in Table 3.1 Safety System Channels are operable. l Bases The startup interlock, which requires a neutron count rate of at least 2 counts per second (CPS) before the reactor is operated, ensures that sufficient neutrons are available for proper operation of the startup channel. The pool-water temperature scram provides protection to ensure that if the limiting safety system setting is exceeded an immediate shutdown will occur to keep the fuel temperature below the safety limit. Power level scrams are provided to ensure that the reactor power is maintained within the licensed limits and to protect aga. inst abnormally high fuel temperatures. The manual scram allows the operator to shut down the reactor if an unsafe or abnormal condition arises. The period scram is provided to ensure that the power level does not increase above that described in the SAR. Specifications on the pool-water level are included as safety measures in the event of [ a serious loss of primary water. Reactor operations are terminated if a major leak occurs in the primary system. The analysis in the SAR shows the consequences l resulting from loss of coolant. l The bridge radiation monitor gives warning of a high radiation levelin the reactor room from failure of an experiment or from a significant drop in pool-water level. A scram from loss of primary coolant flow, loss of power to the pump, or application of power to the pump when operating in the natural convection mode, protects the reactor, fro ~m overheating. Air pressure to the header above ambient results in a scram to:

1) Ensure that the header falls with loss of primary pump power when the reactor is operating in the forced convection mode.

.)

2) Prevent raising the header when the reactor is in the natural convecuon mode.

3) Avoid producing additional Ar-41 by activating air introduced into the header.

(rest of page intentionally left blank) SA Anerdent No. 20

UVAR Tech. Specs. l TABLE 3.1 SAFETY SYSTEM CHANNELS i Minimum Operating Mode Measuring Channel Set Point

• Function I

No. Operable Required

Pool water level monitor 2

19'2" (min) Scram Foxed convection Bridge radiation monitor 1 30 mr/hr Scram All modes I P i oof wa:er temperature 1 108'F (max) Scram All modes loss of power Scram Forced convection Power to primaq pump 1 f application of Natural l power convection ' j i i l Primary ecolant flow 1 900 rpm (min) Scram Forced convection { i l Prevents Startup count rate 1 2 cps (min) withdrawal of Reactor startup j any shim rod i i Manual button 1 Scram All modes l 3 MW (max) Scram Forced convection Rea: tor power level 2 l Natural 0.3 MW (max) Scram convection i Reactor period . 1 3.3 see (min) Scram All modes l l j Air pressure to header 1 above ambient Scram All modes Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values. l 1 8B Amendment IJo. 20 j

i .y i l l F I i i l i .i i l r i i l t a i i ~ 3.3 Reactor Instrumentation t' Applicability: This specification applies to the~ instrumentation that must be i operable for safe operation of the reactor. { 'i I Objective: The objective is to require that sufficient information is avail-i able to the operator to ensure safe operation of the reactor. l Specification: The reactor shall not be_ operated unless the measuring channels described in Section 3.2 " Reactor Safety Systems and in the following table j are operable. i i Bases: The neutron detectors provide assurance that measurements of the-reactor power lev ~el are adequately covered at both low and high power ranges. i t 9 Anencinent No. 20 I

(4) Before operation with fueled experiments whose power generation is greater than 1 W, leak rate shall be verified when the interval since the last verification is greater than 12 months. Bases: Surveillance of this equipment will verify that the confinement of the reactor room is maintained. 4.7 Airborne Effluents Applicability: that monitors the airborne effluents from the ground floor experiment Objective: The objective is to ensure that the airborne effluent monitor is operating and properly calibrated. Specifications: (1) Before each day's operation or before each operation extending more than blower that exhausts the area shall be in operation and a chann shall be performed on the airborne effluent monitor. (2) A calibration of the airborne effluent monitor will be performed using a radioactive source semiannually. j Bases: The semiannual calibration with an external source will permit a drift to be corrected. the SAR (UVAR-18, Part I).The analysis is given in Chapter IX of Amendment I to ' 19 Amendment No. 20

UVAR Tech. Specs. 4.8. Reactor HEU Fuel Dose Measurements Applicability: This specification applies to the highly enriched uranium (HEU) UVAR fuel possessed under the Reactor Facility license. These specifications are applicable until all HEU UVAR fuel elements have been removed from the Reactor Facility. Objective: The objective of this specification is to ensure that the maximum quantity + of special nuclear material does not exceed the limits specified in the Reactor Facility license. l 1 Specifications: 4.8.1. Schedule The amount of special nuclear material (Sh31) possessed at the Reactor Facility will be determined, as necessary, to ensure that limits specified by the Reactor Facility licenses are not exceeded. As a minimum, an evaluation will be { completed and documented every 6 months. 9 4.8.2. Ouantity Limits HEU UVAR fuel elements possessed following the conversion of the UVAR to LEU fuel will be shipped away from the Reactor Facility, as necessary, to ensure that the quantity of nonexempt Sh3i (as defined in 10 CFR 73) does not exceed that allowed by the Reactor Facility licenses. If the amount of nonexempt SNM exceeds 5 kg the Reactor Safety Committee will be informed and the actions specified in the Physical Security Plan implemer.ted. 4.8.3.

• Self-Protection" Determinations If HEU UVAR fuel elements have not been irradiated as a part of the UVAR f

core for at least one month, dose rate measurements of these HEU fuel elements will be made, as necessary, to determine which elements have dose rates higher than specified by 10 CFR 73.67(b). Bases: The specifications provide a high degree of assurance that the amount of SNM and nonexempt SNM will not exceed the license limits. The amount of nonexempt SNM will normally be maintained at less than 5 kg, if necessary by t shipping spent-fuel off-site. In the event that the 5 kg nonexempt SNM quantity is exceeded, the Reactor Safety Committee will be informed of this and the actions specified in the Physical Security Plan will be taken. i (rest of page intentionally left blank) 19A Amercent No. 20 t

4 [ 5 i 4.9 Primary Coolant Conditions Applicability: This specification applies to the surveillance of primary water quality. Objective: The objective is ensure that water quality does not deteriorate over extended periods of time if the reactor is not operated. Specification: The conductivity and pH of the primary coolant water shall be measured at least once every 2 weeks and shall be Conductivity { 5 x 10 8 mhos/cm pH between 5.0 and 7.5 Uh Bases: Section 3.11 of these specifications ensures that the water quality is adequate during reactor operation. Section 4.9 ensures that water quality is not permitted to deteriorate over extended periods of time even if the reactor does not operate. i I i l 20 Araendment No. 20

UVAR Tech. Specs. 5.0. DEslGN FE ATURES 5.1. Reactor Fuel Syrifications Applicabilitv: These specifications apply to UVAR low enriched uranium (LEU) fuel. Obiective: The objective is to describe LEU fuel approved by the U.S. NRC fo: use in the UVAR. Specifications: 5.1.1. Fuel Material UVAR LEU fuel is of a type described for use at U.S. research reae. ors by the U.S. Nuclear Regulatory Commission (NUREG-1313

• Safety Evaluation Report Related to the Evaluation of LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors *). The fuel meat is U Si dispersed in an aluminum matrix 3 2 and enriched to less than 20% U-235.

5.1.2. Element Description (1) Plate-type elements of the MTR type are used. The fuel " meat"is clad with aluminum alloy to form flat fuel plates. The active length of the fuel region in the fuel plates is approximately 24 inches and the width is approximately 2.5 inches. The LEU fuel plates are joined at their long-side edges to two side plates. The entire fuel plate assembly is joined at the bottom to a cylindrical nose piece that fits into the UVAR core gridplate. The overall fuel element dimensions are approximately 3 inches by 3 inches by 36 inches. Each fuel plate contains 12.5 grams of U-235. (2) " Standard

• LEU fuel elements are composed of 22 parallel flat fuel plates each, and contain 275 grams of U-235.

(3)

• Control-rod" LEU elements are similar to the standard elements, with the exception that they have half as many fuel plates (the 11 center plates being removed to form a channel which is bounded by 0.125 inch thick aluminum plates). Control-rod elements accommodate the control rods in the central channel. Their U-235 content is 137.5 grams.

(4)

• Partial
• LEU fuel elements are half-fueled elements composed of 11 LEU fuel plates and 11 unfuelled (dummy) plates. The U-235 content in these elements is 137.5 grams.

(5) "Special" LEU fuel elements have 22 fuel plates, of which 20 are removable. The maximum U-235 content in these elements is 275 grams and the minimum is 25 grams. 21 Amendment No. 20

UVAR Tech. Specs. 5.1.3. Core Configurations A variety of UVAR core configurations may be used to accommodate experiments, but the loadings shall always be such that the minimum shutdown margin and excess reactivity specified in the UVAR Technical Specifications are i not exceeded. i Bases: The NRC has described LEU silicide-fuel suitable for use in U.S. research reactors in NUREG-1313

• Safety Evaluation Repon Related to the Evaluation of LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors,' [536.00, from NTIS, Springfield Va. (703-487-4650)]. Also, Bretscher and Snelgrove from i

the Argonne National Laboratory documented LEU fuel test results in i ANL/RERTR/TM-14, *The Whole-Core LEU U Si -Al Fuel Demonstration in the i 3 2 30-MW Oak Ridge Research Reactor." The LEU-SAR for the UVAR contains the safety analysis performed for the 22 flat-plate University of Virginia fuel elements. The LEU elements were designed by EG&G, Idaho, and are manufactured by the l Babcock and Wilcox Company of Lynchburg, Virginia. (rest of page intentionally left blank) ea 21A Amendment No. 20

i l r i l z ) i i 5.2 Reactor Buf1 ding f Applicability: This specification applies to the room containing the reactor pool and the control room. Specifications: (1) The reactor shall be housed in a room designed to restrict leakage, as stated in Section 3.7(1)(d) of these specifications. t (2) The reactor room shall be equipped with a ventilation system designed to l exhaust air or other gases from the reactor room through a stack at a minimum of 37 ft above ground level. i (3) The minimum free volume of the reactor room shall be 60,000 fta, ~ Bases: The parameters specified were used in the safety and/or environmental impact analyses in the final SAR. 5.3 Fuel Storace All reactor fuel elements not in the reactor core shall be stored in a i geometric array where K,ff is less than 0.9 for all conditions of moderation. j Irradiated fuel elements and fueled devices shall be stored in an array that will permit sufficient natural convection cooling by water or air so that the fuel element or fueled device surface temperature will not exceed the boiling point of water. j 21 B Amendment No. 20 l

+0

• 3" UNITED STATES 5

1% E NUCLEAR REGULATORY COMMISSION ik WASHINGTON. o.C. 2TE4 \\....*/ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CONVERSION ORDER TO CONVERT FROM HIGH-ENRICHED TO LOW-ENRICHED URANTUM FUEL AMENDED FACILITY OPERATING tlCENSE NO. R-66 UNIVERSITY OF VIRGINTA RESEARCH REACTOR DOCKET NO. 50-62 1 INTRODUCTION On February 25, 1985, 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 (DOE) 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 51 - 3 2 Al) developed by Argonne National Laboratory (ANL) especially for use in the DOE HEl' _.:d conversion program. In an effort to standardize and control costs of the conversion of about 20 reactors in the United States, DOE, NRC 4 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. 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 l dependent on fuel design and enrichment and, therefore, might be changed by 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, 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 l 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). 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. 'he staff review led to additional questions to which the licensee responded 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. This safety evaluation report (SER) was prepared by A. Adams, Jr., Senior Project Manager, Division of Operating Reactor Support, Office of Nuclear Reactor 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). 2 EVALUATION i The UVAR is licensed for operation at thermal power levels not to exceed 2 MWt. The reactor uses plate-type fuel cooled by forced circulation of water 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 changes to any reactor system or operating characteristics except for replacing the HEU fuel elements with new LEU fuel elements. The following evaluations and conclusions are based on that assumption. The NRC as part of the conversion order is requiring the licensee to submit a conversion startup report within six months of completion of LEU core loading s that discusses the results of various tests and measurements conducted during the core conversion. 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 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 LEU fuel elements will have the same outer dimensions as the HEU fuel i 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 1

. 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 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 0.55$ (0.41% Ak/k) with the highest-worth reactivity scramable control rod and the non-scramble regulating rod in their most i reactive position. 2.2 Fuel Storace i LEU fuel, not in the reactor core, will be stored in the following four areas: i (1) fuel storage room (dry), (2) 24 spaces in the auxiliary fuel storage rack (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

. 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 g 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 I 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 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 Loadinos of U-235 Th' 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 7.005 (5.18% Ak/k). For comparison of.the nuclear characteristics of the HEU and the LEU fuel, a simple water-reflected (no l 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 4.20$ (3.1% Ak/k) and 3.905 (2.9% Ak/k), respectively. i 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 t 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 average LEU fuel plate and a decrease in the size of the water gap between plates. 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. l

I, l The SAR analyses of the thermal hydraulics of the LEU core employ the same I 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 t 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 l instabilities. Because the cladding and fuel matrix are made of essentially ^ 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 I coolant system operational uncertainties, and the engineering (fabrication) i 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 analyses include both forced coolant flow that is downward through the core and natural convection flow l 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 1 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 i 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 spacing between plates are numerically the same for HEU and LEU fuel elements. ~ gap size derives from an assumption that the fabrication tolerances on the 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 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 scenario 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 MW 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 i i

- - ~ .I

• i i The results show the maximum fuel temperature reaches 303 *F (150 *C), then i

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 The moderator temperature during the transient increases about 10 {959 *F). F (6 C) to about 130 *F (54 *C). These transient results, at 750 kW, indicate that i normal operations of the UVAR LEU core in the natural-convection flow mode at i no more than the licensed power of 300 kW would not present any challer.ge to the fuel-plate integrity. As noted previously, thermal hydraulic risk is not 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. l 2.5 Power Density and Power Peakino 1 Power densities and power peaking, based on the assumed nuclear parameters for [ both graphite and water eflected 4 x 4 HEU and LEU cores, were computed for. the UVAR (References 5 and 6). The poser distribution among the fuel elements is very nearly the same for the two cores, with the peak power density located t in the channel adjacent to the water gap in a control rod fuel element. The i 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 l 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 .i smaller the core size, the higher the average power density. The 4 x 4 core i 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 i peaking in the LEU core is acceptable. 2.6 Control Rod Worths j The UVAR has four control .s, three are scrammable and are used as safety l and shim rods, and the foc 1 rod, which is non-scrammable, is used as a regulating rod. The reacti ity worths of the control rods were computed by r acceptable methods for the DrAa LEU core. The calculated worths of the 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, the LEU-calculated worths for the three ship rods are virtually unchanged from those of the HEU, while the i LEU-calculated worth of the regulating rod is slightly higher. The HEU/ LEU 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 l rod worths are still fully acceptable for safe reactor operation and control l at the UVAR. i i f

. 2.7 Shutdown Marain 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 and 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 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 specifications requirements for shutdown margin shall always be in compliance and shall always take precedence over the technical specifications allowed maximum core excess reactivity. 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 7.005 (5.18% Ak/k) for both the HEU and the LEU cores. The operational excess reactivity, however, is always liinited 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 withdrawn. 2.9 Reactivity Feedback Coefficients The temperature coefficient of reactivity and the void coefficient of reactivity were calculated for the LEU core and compared to those calculated for the HEU core. The moderator temperature coefficient is nearly the same 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 abundant U-238 present in the LEU fuel. Because the Doppler feedback is a function of fuel temperature, it is prompt and, therefore, more effective in countering a reactor transient in the LEU core than is the moderator temperature coefficient in the HEU core, which n,ust 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.

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 les:, resulting in less fission product inventory per fuel plate in the LEU core for the same operating (power-time) history. The aluminum cladding on 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 of fission products attributable to the loss of integrity resulting from this 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 satisfactorily as the HEU fuel it will replace in containing fissien products in the UVAR. The staff finds this acceptable. 2.11 Potential Accident Scenarios Several potential accident scenarios were postulated and analyzed by the i 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 l 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 witt. drawn, inserting a ramp increase in reactivity. This results in a continuously decreasing reactor period and an increasing power level. The scenario assumed 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 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 (47 1/s) and 837 gpm (53 1/s), respectively. It was further acsumed that the temperature coefficient of reactivity is zero. 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 events for the two fuel types are very similar. The maxiinum power levels 4 reached are the same, depending primarily on the dynamic parameters of the scram circuits and the control rod insertion times. Tnese 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 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

9-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 not increased as a result of the HEU/ LEU fuel conversion. 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 percent (50 1/s 7.0%) LSSS for the HEU core, while the LEU was analyzed at 837 gpm (53 1/s), which is the minimum true value of the 900 gpm i 7.0 percent (57 1/s i 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 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 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 conversion of the UVAR from HEU to LEU, the staff concludes the conversion i does not increase the health and safety risk to the public from the postulated thermal hydraulic accidents. 2.12 Technical Specification Chances 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 con /idence 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 i safety of the public as a result of the conversion from HEU to LEU fuel, as i 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.

i 2.12.1 Safety Limits 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. 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 forceu coolant flow limitation from 800 gpm (501/s) to 900 gpm (571/s), about 14 percent. i In adGition, 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 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. 2.12.4 Design Features TS 5.1, reactor fuel specifications, is amended to describe the new LEU fuel material, element description, and core configurations introduced by conversion. The licensee requested additions to TS 5.3, fuel storage, to describe the various special nuclear material possession limits for the facility. This information is redundant to license condition 11 B.(2) and (4) and therefore, the staff is not placing this information in the technical specifications. This was discussed with and agreed to by the Director of the UVAR on March 22, 1993. 3 CONCLUSION The. staff concludes that the conversion, as proposed, would not reduce any safety margins, introduce any new safety issues, or lead to increased 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 t l Date: April 29, 1993 i

. j 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. ~

Table 1 Comparison of Parameters for the HEU and LEU Cores

• at the University of Virginia Reactor Parameter HEU LEUb General:

Critical mass (g U-235) 3596 4952 Excess reactivity (%Ak/k) 3.l 2.9 b ($) 4.20 3.90 b 8,, 0.0074 0.0074 Neutron lifetime, f, (ps) 64.4 53.0 b f/8,,,(s) 0.009 0.007 b Fuel Elements: Number of standard fuel elements 16 16 Number of plates per standard element 18 22 Number of control fuel elements 4 4 Number of plates per control element 9 11 Fuel plate dimensions (in.) 24.6x2.8x0.05 24.6x2.8x0.05 (cm) 62.5x7.1x0.13 62.5x7.1x0.13 Enrichment (%) 92 19.75 Mass of U-235 per plate (g) 10.8 12.5 Water gap (in.) 0.122 0.0927 (cm) 0.310 0.235 Fuel thickness (in.) 0.02 0.02 (cm) 0.05 0.05 ~ Aluminum cladding thickness (in.) 0.015 0.015 (cm) 0.038 0.038 Uranium density (g/cc) 0.69 3.47 fuel matrix UAf,-Af U Si -Af 3 z Rod Worth: Regulating (%Ak/k) 0.38 0.45 (5) 0.50 0.60 Safety (nominal) (%Ak/k) 3.24 3.24 ($) 4.40 4.40 Reactivity Coefficients: Moderator temperature coefficient (%Ak/k/*C) -0.011 -0.011 (5/*C) -0.015 -0.015 Fuel temperature coefficient (%Ak/k/*C) -1. 3 x10 -1.1x10'3 ($/*C) -1. 8x10 -1.5x10'3 Yoid coefficient (%Ak/k per % void) -0.194 -0.248 ($ per % void) -0.262 -0.335 4x5 water-reflected core b calculated}}