ML20035H612

Order Modifying License R-66 to Convert from High to Low Enriched U

ML20035H612
Person / Time
Site:	University of Virginia
Issue date:	03/29/1993
From:	Murley T NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:	VIRGINIA, UNIV. OF, CHARLOTTESVILLE, VA
Shared Package
ML20035H609	List: ML20035H608 ML20035H612 ML20035H647
References
NUDOCS 9305060075
Download: ML20035H612 (46)

Text

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i UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of

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UNIVERSITY OF VIRGINIA Docket No. 50-62

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Facility Operating License No. R-66 (University of Virginia Pool Reactor)

Amendment No. 20

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ORDER MODIFYING LICENSE n

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 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-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) lim'iting the use of high-enriched uranium (HEU) fuel in domestic research and i

test reactors (non-power reactors) (see 51 FR 6514). The rule, which became 9305060075 930429 PDR ADOCK 05000062 PDR p

1 l 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

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 theft and diversion of HEU fuel used in non-power reactors and the adverse consequences to public health and safety and the environment from such theft or diversion.

Sections 50.64(b)(2)(i) and (ii) require that a licensee of a non-power reactor (1) not initiate acquisition of additional HEU fuel, if LEU fuel that is acceptable to the Commission for that reactor is available when the i

licensee proposes that acquisition, and (2) replace all HEU fuel in its l

possession with available LEU fuel acceptable to the Commission for that reactor in accordance with a schedule determined pursuant to 10 CFR 50.64(c)(2).

l Section 50.64(c)(2)(1) of the rule, among other things, requires each licensee of a non-power reactor, authorized to possess and to use HEU fuel, to

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develop and to submit to the Director of the Office of Nuclear Reactor Regulation (Director) by March 27, 1987, and at 12-month intervals thereafter, l

a written proposal (proposal) for meeting the rule requirements.

Section 50.64(c)(2)(i) also requires the licensee to include the following in its proposal:

(1) a certification that Federal Government l

i funding for conversion is available through the U.S. Department of Energy r

(DOE) or other appropriate Federal agency and (2) a schedule for conversion, j

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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 to the license, to the facility, and to licensee procedures (all three types of changes hereafter called modifications). This paragraph also requires the i

licensee to submit supporting safety analyses so as to meet the schedule established for conversion.

t Section 50.64(c)(2)(iii) also requires the Director to review the licensee proposal, confirm the status of Federal Government funding, and I

determine a final schedule if the licensee has submitted a schedule for 1

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 Recister notice of the final rule, the Commission explained that in most cases, if not all, the l

Enforcement Order would be an Order to modify the license.

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i Section 2.202, the current authority for issuing Orders of all types including Orders to modify licenses, provides, among other things, that the

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Commission may modify a license by serving an Order on the licensee. The t

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 i

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- i period as the notice may provide. The Order will become effective on the 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 f

the date specified in an Order made after the hearing.

Section 2.714 gives the requirements for a person whose interest may be affected by any proceeding to initiate a hearing or to participate as a party.

111.

j On November 9, 1989, as supplemented on February 12, 1991, and December 14, 1992, the NRC staff received the licensee proposal, including its 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 reactor (MTR)-type fuel plates, with the fuel meat consisting of uranium r

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 specifications that are needed to amend the ' facility license and (2) the I

outline of the startup report that is required to be submitted within six months following completion of LEU fuel loading. The NRC staff has reviewed 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 l

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.

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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

regulations in 10 CFR 2.202 and Section 50.64, IT IS HEREBY ORDERED THAT:

i On the later date of either receipt of the replacement core of LEU fuel elements by the licensee or 30 days following the date of publication of this Order in the Federal Reaister, Facility Operating License R-66 is modified by amending the license conditions and technical specifications as stated in the

" Attachment to Order Modifying Facility Operator License R-66."

The licensee shall submit the startup report as stated in the " Attachment to Order of the Outline of Reactor Startup Report" within six months following completion of LEU fuel loading.

V.

r 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 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 i

Regulatory Commission, Washington, D.C. 20555, with a copy to the Assistant I

General Counsel for Hearings and Enforcement at the same address.

If a person r

other than the licensee requests a hearing, that person shall set forth with

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particularity in accordance with 10 CFR 2.714 the manner in which their interest is adversely affected by this Order.

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- 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 such hearing is whether this Order should be sustained.

'l 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 this 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 Dated at Rockville, Maryland this 29t.h day of March Attachments:

As stated a

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ATTACHMENT TO ORDER MODIFYING FACILITY OPERATING LICENSE R-66 A.

License Conditions Revised and Added by This Order II.B(2)

Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of 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 the facility. Without exceeding the foregoing maximum possession limits, the maximum limits on specific enrichments of U-235 are as follows:

Maximum U-235 Kilocrams

%_ Enrichment form 11

< 20%

Materials testing reactor (MTR)-type fuel 1

Any Fission chambers, flux foils, and other forms used in connection with operation of the reactor ll.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 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 reactor fuel until the existing inventory of high-enriched MTR-type reactor fuel is removed from the facility.

II.C(2) Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 20, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

~ B. Technical Specifications Revised by This Order 2.1.1. Safety Limits in Forced Convection Mode of Operation Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the forced convection mode of operation. These variables are:

P = Reactor thermal power l

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 core.

(2) The reactor coolant inlet temperature shall not be greater than 111*F.

i (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 i

flow shall be below the line defined by:

P = 0.24 + (4.5 x 10-8

• W)

P = 0 for W < 575; P in MW, W in gpm The allowed regiod of operation is shown by the unshaded region of Figure 2.1.

Ha.lir 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.

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' 2.1.2. Safety Limits in the Natural Convection Mode of Oneration 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 i

Ob_iective: The objective is to ensure that the integrity of the fuel clad is maintained.

Specification: In the natural convection mode of operation:

(1) The true value of reactor power shall not exceed 750 kW.

(2) The reactor coolaat inlet temperature shall not be greater than 111*F.

Has: The criterion for establishing a safety limit with natural convection now 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. Safety Limit for the Transition from Forced to Natural Convection Mode of Operation Applicability: This specification applies to the condition when the reactor is in i

transition from forced convection flow to natural convection flow.

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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.

Bai: 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.

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2.2.

Limiting Safety System Settings e

Applicability: These specifications apply to the set points for the safety channels monitoring reactor thermal power, coo'. ant flow rate, reactor coolant inlet

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temperature, and the height of water above the core.

i Obiective: The objective is to ensure that automatic protective action is initiated to prevent the safety limit from being exceeded.

Specifications:

r 2.2.1. Forced Convection Mode For operation in the forced convection mode, the limiting safety system settings shall be:

3.0 MWt (max)

Reactor Thermal Power

=

900 gpm (min)

Reactor Coolant Flow Rate

=

Reactor Coolant Inlet Temperature = 108'F (max) i 19'2" (min)

Height of Water above Core

=

Reactor Period

= 3.3 see (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) l 3.3sec (min)

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Reactor Period

=

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 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 (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. The LSSS specified above for this mode of operation are below the analy;:ed 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>eci6 cation: 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 power 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, i

protects the reactor, fro ~m overheating.

3 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 mod,e.

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.

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i TABLE 3.1 SAFETY SYSTEM CHANNELS

i Minimum Operating Mode Measuring Channel Set Point' Function l

No. Operable Required i

l Pool water level monitor 2

19'2" (min)

Scram Forced convection l Bridge radiation monitor 1

30 mr/hr Scram All modes I

t Pool water temperature I

108'F (max)

Scram All modes loss of power Scram Forced convection t

l Power to primary pump 1

application of Natural

(

l Scram j

power-convection 4

Primary coolant flow I

903 gpm (min)

Scram Forced convection I

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Prevents Startup count rate 1

2 cps (min) withdrawal of Reactor startup i

(

any shim rod j

j Manua.1 button 1

Scram A!! modes 3 hWe max)

Scram Forced convection j

4 i Reactor power level 2

Natural 0.3 MWt (max)

Scram convection 4

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'l Reactor period

. 1 3.3 sec (min)

Scram All modes i

l Air pressure to header 1

above ambient Scram All modes t

l Values listed are limiting set points. For operational convenience, set points may be changed l

to more conservative values.

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~ 4.8.

Reactor HEU Fuel Dose Measurements Applicability: This specification applies to the highly enriched uranium (HEU) f UVAR fuel possessed under the Reactor Facility license. These specifications are applicable until all HEU UVAR fuel elements have been removed from the Reactor l

Facility.

1 Objective: The objective of this specification is to ensure that the maximum quantity

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of special nuclear material does not exceed the limits specified in the Reactor Facility license.

Specifications:

t 4.8.1. Schedule l

The amount of special nuclear material (Sh%f) 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 convenion of the UVAR to LEU fuel will be shipped away from the Reactor Facility, as necessary, to ensure that the quantity of nonexempt Sh%f (as defined in 10 CFR 73) does not exceed that allowed by the Reactor Facility licenses. If the amount of nonexempt Sh%i exceeds 5 kg the Reactor Safety Committee will be informed and the actions specified in the Physical Security Plan impicmented.

4.8.3. 'Self-Protection" Determinations f

If HEU UVAR fuef elements have not been irradiated as a part of the UVAR core for at least one month, dose rate measurements of these HEU fuel elements will be m;.de, as necessary, to determine which elements have dose rates higher than specified by 10 CFR 73.67(b).

Bases: ne specifications provide a high degree of assurance that the amount of Sh%f and nonexempt SNM will not exceed the license limits. The amount of nonexempt Sh%f will normally be maintained at less than 5 kg, if necessary by shipping spent-fuel off-site. In the event that the 5 kg nonexempt Sh%f quantity is exceeded, the Reactor Safety Committee will be informed of this and the actions specified in the Physical Security Plan will be taken.

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5.1.

Reactor Fuel Specifications

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Applicability: These specifications apply to UVAR low enriched uranium (LEU) fuel.

Obiective: The objective is to describe LEU fuel approved by the U.S. NRC for use in the UVAR.

I Specifications:

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 l

Related to the Evaluation of LEU Silicide Aluminum-Dispersed Fuel for Use in l

Non-Power Reactors"). The fuel meat is U Si dispersed in an aluminum matrix l

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 t

the fuel plates is approximately 24 inches and the width is approximately 2.5 l

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 i

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.

i (4) " Partial" LEU fuel elements are half-fueled elements composed of 11 LEU fuel i

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.

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5.1.3. Core Confirurations A variety of UVAR core configurations may be used to accommodate

'j experiments, but the loadings shall always be such that the minimum shutdown i

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 l

LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors," [$36.00, from NTIS, Springfield Va. (703-487-4650)). Also, Bretscher and Snelgrove from the Argonne National Laboratory 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 l

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 j

Babcock and Wilcox Company of Lynchburg, Virginia.

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1 ATTACHMENT TO ORDER OF OUTLINE OF REACTOR STARTUP REPORT k'ithin six months following completion of initial LEU core loading, submit the following information to NRC:

E 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 available, HEU j

3.

Control and regulating rod calibrations i

t Measurement of HEU and LEU differential and total rod worths and comparisons with calculations for LEU and if available, HEU i

4.

Reactor power calibration Methods and measurements that ensure operation within the license limit and comparison between HEU and LEO 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 t

6.

Partial fuel element worths for LEU Measurements of the worth of the partial loaded fuel elements 7.

Thermal neutron flux distributions Hea:frements 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 b

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2-9.

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 reactor

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 Comparison against available HEU core measurements and LEU predictions t

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

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r ATTACHMENT TO LICENSE AMENDMENT NO. 20 FACILITY OPERATING LICENSE NO. R-66 i

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 Paces Insert Paoes i

4 4

5 5

6 5A 8

5B 9

6 19 8

20 8A

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21 BB 9

t 19 19A 20 21 21A 21B P

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UVAR Tech. Specs.

i 2.0.

SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTINGS l

1 2.1.

Safety Limits i

i 2.1.1. Safety Limits in Forced Convection Mode of Operation j

i Applicability: This specification applies to the interrelated variable; associated j

with core thermal and hydraulic performance in the forced convection mode of

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l operation. These variables are:

P = Reactor thermal power W = Reactor coolant flow rate i

T = Reactcr coolant inlet temperature i

i L = Height of water above the core Obiective: The objective is to ensure that the integrity of the fuel clad is I

maintained.

i Specification: In the forced convection mode of operation:

2 (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 III'F.

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(3) The true value of reactor coolant flow shall not be below 575 gpm.

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(4) The combination of true values of reactor core power and reactor coolant flow shall be below the line defined by:

l P = 0.24 + (4.5 x 10-8

• W) l j

P = 0 for W < 575; P in MW, W in gpm The allowed region of operation is shown by the unshaded region of Figure 2.1.

e 1

Enis: Above 575 gpm in the region of full power operation, the criterion used l

to establish the safety limit was a bumout ratio of 1.49 including the worst 4

variation in the manufacturer's tolerance and specification, hot channel factors and other appropriate uncenainties. 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.

I could occur.

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4 Amendment No. 20

u UVAR Tech. Sp cs.

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5 Amend:nent I;o. 20

UVAR Tech. Specs.

2.1.2. Safety Limits in the Natural Convection Mode of Operation l

Applicability: This specification applies to the interrelated variables associated with core thermal and hydraulic performance in the natural convection mode of l

operation. These variables are:

(g P = Reactor thermal power T = Reactor coolant inlet temperature i

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Objective: The objective is to ensure that the integrity of the fuel clad is

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maintained.

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 Ill'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.

(rest of page intentionally left blank) l i

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SA

/cendment No. 20

UVAR Tech. Specs.

I 2.1.3 Safety Limit for the Transition from Forced to Natural Convection Mode of QDCIittica Applicability: 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.

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l SB Amndmnt No. 20

UVAR Tech. Specs.

2.2.

Limiting Safety System Settines AP21cabihtv: These specifications apply to the set points for the safety channels 1

monitoring reactor thermal power, coolant flow rate, reactor coolant inlet temperature, and the height of water above the core.

Obiective: 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 Thermal Power

= 3.0 MWt (max) 900 gpm (min)

Reactor Coolant Flow Rate

=

Reactor Coolant Inlet Temperature = 108'F (max)

Height of Water above Core 19'2" (min)

=

3.3see (min) l Reactor Period

=

2.2.2. Natural Convection Mode For operation in the natural convection mode, the limiting safety system settings shall be:

300 kWt (max).

Reactor Power

=

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 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 11l'F and a coolant flow of 837 gpm. The maximum fuel plate temperature i

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 (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. The LSSS specified above for this mode of operation are below the analyzed condition.

6 Amendment No. 20

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i 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.

1 The 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.

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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 stanup 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 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 l

resulting from loss of coolant.

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 froin 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 convection mode.

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

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UVAR Tech. Specs.

TABLE 3.1 SAFETY SYSTEM CHANNELS Minimum Operating Mode Measuring Channel Set Point

• Function No. Operable Required Pool water level monitor 2

19'2* (min)

Scram Forced convection Bridge radiation monitor 1

30 mr/hr Scram All modes Pool water temperature 1

108'F (max)

Scram All modes loss of power Scram Forced convection i

Power to primary pump I

application of Natural power convection Primary coolant flow I

900 gpm (min)

Scram Forced convection Prevents Stanup count rate 1

2 cps (min) withdrawal of Reactor startup any shim rod i

j Manual button 1

Scram All modes 3 MWt (max)

Scram Forced convection i

Reactor power level 2

Natural 0.3 MWt (max)

Scram convect. ion Reactor period 1

3.3 sec (min)

Scram All modes l

3 i

Air pressure to header I

above ambient Scram All modes i

i Values listed are limiting set points. For operational convenience, set points may be changed to more conservative values.

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3.3 Reactor Instrumentation l

Applicability:

This specification applies.to the instrumentation that must be operable for safe operation of the reactor.

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Objective:

The objective is to require that sufficient information is avail-able to the operator to ensure safe operation of the reactor.

Specification:

The reactor shall not be operated unless the measuring channels described in Section 3.2 " Reactor Safety Systems" and in the following table are operable.

Bases:

The neutron detectors provide assurance that measurements of the reactor power level are adequately covered at both low and high power ranges.

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(4) Before operation with fueled experiments whose power generation is greater i

I 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:

This specification applies to the surveillance of the instrument that monitors the airborne effluents from the ground floor experimental area.

Obiective: 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 one day, when either of the neutron beam ports are drained, the centrifugal blower that exhausts the area shall be in operation and a channel check shall be performed on the airborne effluent monitor.

(2) A calibration of the airborne effluent monitor will be performed using a j

radioactive source semiannually.

Bases: The daily channel check of the monitor will ensure that it is operable.

The semiannual calibration with an external source will permit any long-term drift to be corrected.

The analysis is given in Chapter IX of Amendment 1 to

• the SAR (UVAR-18, Part I).

19 Amendment No. 20

UVAR Tech. Specs.

l 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.

i Specifications:

4.8.1. Schedule The amount of special nuclear material (SNM) 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 SNM (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 implemented.

4.8.3. PSelf-Protection" Determinations If HEU UVAR fuel elements have not been irradiated as a part 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 SNM and nonexempt SNM will not exceed the license limits. The amount of 4

nonexempt SNM will normally be maintained at less than 5 kg, if necessary by 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.

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4.9 Primary Coolant Conditions r

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 6 mhos/cm pH between 5.0 and 7.5 hh 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.

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UVAR Tech. Specs.

5.0.

D_ESIGN FEATURES 5.1.

Reacter 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.

Specifications:

5.1.1. Fuel Material UVAR LEU fuel is 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 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.

i (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.

t 21 Amendment No. 20

1 UVAR Tech. Specs.

5.1.3. Core Con 6curations 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 LEU Silicide Aluminum-Dispersed Fuel for Use in Non-Power Reactors," [$36.00, from NTIS, Springfield Va. (703-487-4650)]. Also, Bretscher and Snelgrove from the Argonne National Laboratory 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 s

Babcock and Wilcox Company of Lynchburg, Virginia.

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I 5.2 Reactor Building 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.

3 (2) The reactor room shall be equipped with a ventilation system designed to exhaust air or other gases from the reactor room through a stack at a minimum of 37 ft above ground level.

(3) The minimum free volume of the reactor room shall be 60,000 fta, i

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 geometric array where K,ff is less than 0.9 for all conditions of moderation.

Irradiated fuel elements and fueled devices shall be. stored in an array that f

will permit sufficient natural convection cooling by water or air so that the i

fuel element or fueled device surface temperature will not exceed the boiling point of water.

f 21 B Amendment IJo. 20