ML071920168

Issuance of Order Modifying Facility Operating License No. R-87 to Convert from High to Low Enriched Uranium Fuel (Amendment No. 12) - Purdue University Research Reactor

ML071920168
Person / Time
Site:	Purdue University
Issue date:	08/09/2007
From:	Alexander Adams NRC/NRR/ADRA/DPR/PRTA
To:	Joel Jenkins Purdue University
Adams A, NRC/NRR/DPR/PRT, 415-1127
References
EA-07-197, TAC MD2877
Download: ML071920168 (40)

Text

August 9, 2007 EA-07-197 Mr. Jere H. Jenkins Director of Radiation Laboratories Purdue University School of Nuclear Engineering Nuclear Engineering Building 400 Central Drive West Lafayette, IN 47907-2017

SUBJECT:

ISSUANCE OF ORDER MODIFYING FACILITY OPERATING LICENSE NO. R-87 TO CONVERT FROM HIGH- TO LOW-ENRICHED URANIUM FUEL (AMENDMENT NO. 12) - PURDUE UNIVERSITY RESEARCH REACTOR (TAC NO. MD2877)

Dear Mr. Jenkins:

The U.S. Nuclear Regulatory Commission (NRC) is issuing the enclosed Order, as Amendment No. 12 to Facility Operating License No. R-87, which authorizes the conversion of the Purdue University Research Reactor from high-enriched uranium fuel to low-enriched uranium (LEU) fuel. This Order modifies the license, including the technical specifications, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.64. This regulation requires that non-power reactor licensees, such as the Purdue University, convert to LEU fuel under certain conditions which Purdue University now meets. The Order is being issued in accordance with 10 CFR 50.64(c)(3) and in response to your submittal of August 13, 2006, as supplemented on May 3 and June 18, 2007. The Order also contains an outline of a reactor startup report that you are required to provide to the NRC within six months following the return of the converted reactor to normal operation.

The Order becomes effective on the later date of either the day of receipt of an adequate number and type of LEU fuel assemblies that are necessary to operate the facility as specified in your submittal and supplements, or 23 days after the date of its publication in the Federal Register, provided there are no requests for a hearing.

Although this Order is not subject to the requirements of the Paperwork Reduction Act, there is nonetheless a clearance from the Office of Management and Budget (OMB), OMB approval number 3150-0012, that covers the information collections contained in the Order.

J. H. Jenkins Copies of replacement pages for the technical specifications and of the NRC staff safety evaluation for the conversion to LEU fuel are also enclosed. The Order is being sent to the Federal Register for publication.

Sincerely,

/RA/

Alexander Adams, Jr., Senior Project Manager Research and Test Reactors Branch A Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-182

Enclosures:

1. Order

2. Safety Evaluation cc w/enclosures: See next page

ML071920168 OFFICE DPR:LA PRTA:PM OGC NLO OE NAME EHylton egh AAdams aa A. Jones aj M. Burrell mb DATE 7/17/07 7/18/07 8/2/07 7/27/07 OFFICE PRTA:BC DPR:D NRR: D NAME DCollins dsc hn for MCase jw for JDyer DATE 7/25/07 8/7/07 8/9/07 Purdue University Docket No. 50-182 cc:

Mayor City of West Lafayette 609 W. Navajo West Lafayette, IN 47906 John H. Ruyack, Manager Epidemiology Res Center/Indoor & Radiological Health Indiana Department of Health 2525 N. Shadeland Ave., E3 Indianapolis, IN 46219 Howard W. Cundiff, P.E., Director Consumer Protection Indiana State Department of Health 2 North Meridian Street, 5D Indianapolis, IN 46204 Mr. Ed Merritt Reactor Supervisor Department of Nuclear Engineering Purdue University West Lafayette, IN 47907 Test, Research, and Training Reactor Newsletter University of Florida 202 Nuclear Sciences Center Gainesville, FL 32611

7590-01-P UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

)

PURDUE UNIVERSITY ) Docket No. 50-182

) EA-07-197 (Purdue University Research Reactor) )

ORDER MODIFYING FACILITY OPERATING LICENSE NO. R-87 I.

Purdue University (the licensee) is the holder of Facility Operating License No. R-87 (the license) issued on August 16, 1962, by the U.S. Atomic Energy Commission, and subsequently renewed on August 8, 1988, by the U.S. Nuclear Regulatory Commission (the NRC or the Commission). The license authorizes operation of the Purdue University Research Reactor (the facility) at a power level up to 1 kilowatt thermal. The facility is a research reactor located on the campus of Purdue University, in the city of West Lafayette, Tippecanoe County, Indiana.

The mailing address is Radiation Laboratories, Purdue University, Nuclear Engineering Building, 400 Central Drive, West Lafayette, IN 47907-2017.

II.

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.64, limits the use of high-enriched uranium (HEU) fuel in domestic non-power reactors (research and test reactors)

(see 51 FR 6514). The regulation, which became effective on March 27, 1986, requires that if Federal Government funding for conversion-related costs is available, each licensee of a non-power reactor authorized to use HEU fuel shall replace it with low-enriched uranium (LEU) fuel acceptable to the Commission unless the Commission has determined that the reactor has a unique purpose. The Commissions stated purpose for these requirements was to reduce, to

the maximum extent possible, the use of HEU fuel in order to reduce the risk of theft and diversion of HEU fuel used in non-power reactors.

Paragraphs 50.64(b)(2)(i) and (ii) require that a licensee of a non-power reactor (1) not acquire more HEU fuel if LEU fuel that is acceptable to the Commission for that reactor is available when the licensee proposes to acquire HEU fuel, and (2) replace all HEU fuel in its 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).

Paragraph 50.64(c)(2)(i) requires, among other things, that each licensee of a non-power reactor authorized to possess and to use HEU fuel develop and submit to the Director of the Office of Nuclear Reactor Regulation (Director) by March 27, 1987, and at 12-month intervals thereafter, a written proposal for meeting the requirements of the rule. The licensee shall include in its proposal a certification that Federal Government funding for conversion is available through the U.S. Department of Energy or other appropriate Federal agency and a schedule for conversion, based upon availability of replacement fuel acceptable to the Commission for that reactor and upon consideration of other factors such as the availability of shipping casks, implementation of arrangements for available financial support, and reactor usage.

Paragraph 50.64(c)(2)(iii) requires the licensee to include in the proposal, to the extent required to effect conversion, all necessary changes to the license, to the facility, and to licensee procedures. This paragraph also requires the licensee to submit supporting safety analyses in time to meet the conversion schedule.

Paragraph 50.64(c)(2)(iii) also requires the Director to review the licensee proposal, to confirm the status of Federal Government funding, and to 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 to issue an appropriate enforcement order directing both the conversion and, to the extent consistent with protection of public health and safety, any necessary changes to the license, the facility, and licensee procedures. In the Federal Register notice of the final rule (51 FR 6514),

the Commission explained that in most, if not all cases, the enforcement order would be an order to modify the license under 10 CFR 2.204 (now 10 CFR 2.202).

Section 2.309 states the requirements for a person whose interest may be affected by any proceeding to initiate a hearing and to participate as a party.

III.

On August 13, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML062400495 and ML070920272), as supplemented on May 3 (ADAMS Accession No. ML071410299) and June 18, 2007 (ADAMS Accession No. ML071700633), the NRC staff received the licensee's conversion proposal, including its proposed modifications and supporting safety analyses. HEU fuel assemblies are to be replaced with LEU fuel assemblies. The fuel assemblies contain fuel plates, typical of the Materials Testing Reactor design, with the fuel consisting of uranium silicide dispersed in an aluminum matrix. These plates contain the uranium-235 isotope at an enrichment of less than 20 percent. The NRC staff reviewed the licensee's proposal and the requirements of 10 CFR 50.64 and has determined that public health and safety and common defense and security require the licensee to convert the facility from the use of HEU to LEU fuel in accordance with the attachments to this Order and the schedule included herein. The attachments to this Order specify the changes to the license conditions and technical specifications that are needed to amend the facility license and contains an outline of a reactor startup report to be submitted to NRC within six months following return of the converted reactor to normal operation.

IV.

Accordingly, pursuant to Sections 51, 53, 57, 101, 104, 161b, 161i, and 161o of the Atomic Energy Act of 1954, as amended, and to Commission regulations in 10 CFR 2.202 and 10 CFR 50.64, IT IS HEREBY ORDERED THAT:

Facility Operating License No. R-87 is modified by amending the license conditions and technical specifications as stated in the attachments to this Order (Attachment 1:

MODIFICATIONS TO FACILITY OPERATING LICENSE NO. R-87; Attachment 2: OUTLINE OF REACTOR STARTUP REPORT). The Order becomes effective on the later date of either (1) the day the licensee receives an adequate number and type of LEU fuel assemblies to operate the facility as specified in the licensee proposal dated August 13, 2006 (ADAMS Accession Nos.

ML062400495 and ML070920272), as supplemented on May 3 (ADAMS Accession No. ML071410299) and June 18, 2007 (ADAMS Accession No. ML071700633), or (2) 23 days after the date of publication of this Order in the Federal Register.

V.

Any person adversely affected by this Order may submit an answer to this Order, and may request a hearing on this Order, within 20 days of the date of this Order. Any answer or request for a hearing shall set forth the matters of fact and law on which the person adversely affected, relies and the reasons why the Order should not have been issued. Any answer or request for a hearing shall be filed (1) by first class mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, Attention: Rulemaking and Adjudications Staff; or (2) by courier, express mail, and expedited delivery services to the Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and Adjudications Staff. Because of possible delays in delivery of mail

to the United States Government Offices, it is requested that answers and/or requests for hearing be transmitted to the Secretary of the Commission either by e-mail addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; or by facsimile transmission addressed to the Office of the Secretary, U.S. Nuclear Regulatory Commission, Washington, D.C., Attention: Rulemakings and Adjudications Staff at 301-415-1101 (the verification number is 301-415-1966). Copies of the request for hearing must also be sent to the Director, Office of Nuclear Reactor Regulation and to the Assistant General Counsel for Materials Litigation and Enforcement, Office of the General Counsel, with both copies addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, and the NRC requests that a copy also be transmitted either by facsimile transmission to 301-415-3725 or by e-mail to OGCMailCenter@nrc.gov.

If a person requests a hearing, he or she shall set forth in the request for a hearing with particularity the manner in which his or her interest is adversely affected by this Order and shall address the criteria set forth in 10 CFR 2.309.

If a hearing is requested by 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 shall be whether this Order should be sustained.

In accordance with 10 CFR 51.10(d), this Order is not subject to Section 102(2) of the National Environmental Policy Act, as amended. The NRC staff notes, however, that with respect to environmental impacts associated with the changes imposed by this Order as described in the safety evaluation, the changes would, if imposed by other than an Order, meet the definition of a categorical exclusion in accordance with 10 CFR 51.22(c)(9). Thus, pursuant to either 10 CFR 51.10(d) or 51.22(c)(9), no environmental assessment nor environmental impact statement is required.

For further information see the application from the licensee dated August 13, 2006 (ADAMS Accession Nos. ML062400495 and ML070920272), as supplemented on May 3 (ADAMS Accession No. ML071410299) and June 18, 2007 (ADAMS Accession No. ML071700633), the staffs request for additional information dated March 13, 2007 (ADAMS Accession No. ML070680273), and the cover letter to the licensee, attachments to this Order and the NRC staffs safety evaluation dated August 9, 2007 (ADAMS Accession No. ML071920168), available for public inspection at the Commission's Public Document Room (PDR), located at One White Flint North, Public File Area O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland. Publicly available records will be accessible electronically from the ADAMS Public Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who have problems in accessing the documents in ADAMS should contact the NRC PDR reference staff by telephone at 1-800-397-4209 or 301-415-4737 or by e-mail to pdr@nrc.gov.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

James T. Wiggins, Deputy Director Office of Nuclear Reactor Regulation Dated this 9th day of August 2007 Attachments: 1. Modifications to Facility Operating License No. R-87

2. Outline of Reactor Startup Report

ATTACHMENT TO ORDER MODIFICATIONS TO FACILITY OPERATING LICENSE NO. R-87 A. License Conditions Revised by This Order 2.B.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use: (1) up to 3.8 kilograms of contained uranium-235 of enrichment of less than 20 percent in the form of materials testing reactor (MTR)-type reactor fuel; (2) up to 80.0 grams of plutonium contained in encapsulated plutonium-beryllium sources; and (3) up to 100 grams of contained uranium-235 of any enrichment in the form of fission chambers, flux foils and fueled experiments, all used in connection with operation of the facility; 2.B.(4) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to possess, but not use, up to 3 kilograms of contained uranium-235 at equal to or greater than 20 percent enrichment in the form of materials testing reactor (MTR)-type reactor fuel until the existing inventory of this fuel is removed from the facility.

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 12 are, hereby, incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

B. The technical specifications will be revised by this Order in accordance with the "Enclosure to License Amendment No. 12, Facility Operating License No. R-87, Docket No. 50-182, Replacement Pages for Technical Specifications," as discussed in the safety evaluation for this Order.

Attachment 1

ENCLOSURE TO LICENSE AMENDMENT NO. 12 FACILITY OPERATING LICENSE NO. R-87 DOCKET NO. 50-182 REPLACEMENT PAGES FOR TECHNICAL SPECIFICATIONS Replace the following pages of Appendix A, Technical Specifications, with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert 7 7 8 8 20 20 22 22 23 23

7

2. SAFETY LIMIT AND LIMITING SAFETY SYSTEM SETTING 2.1 Safety Limit Safety limits for nuclear reactors are limits upon important process variables that are necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. The principal physical barrier is the fuel cladding.

Applicability - This specification applies to the temperature of the reactor fuel and cladding under any condition of operation.

Objective - The objective is to ensure fuel cladding integrity.

Specification - The fuel and cladding temperatures shall not exceed 530°C (986°F).

Basis - In the Purdue University Reactor, the first and principal barrier protecting against release of radioactivity is the cladding of the fuel plates. The 6061 aluminum alloy cladding of the LEU fuel plates has an incipient melting temperature of 582°C. However, measurements (NUREG-1313) on irradiated fuel plates have shown that fission products are first released near the blister temperature (~550°C) of the cladding. To ensure that the blister temperature is never reached, NUREG 1537 concludes that 530°C is an acceptable fuel and cladding temperature limit not to be exceeded under any condition of operation.

2.2 Limiting Safety System Setting Applicability - This specification applies to the reactor power level safety system setting for steady state operation.

Objective - The objective is to assure that the safety limit is not exceeded.

Specification - The measured value of the power level scram shall be no higher than 1.2 kW.

Amendment No. 12 August 9, 2007

8 Basis - The LSSS has been chosen to assure that the automatic reactor protective system will be actuated in such a manner as to prevent the safety limit from being exceeded during the most severe expected abnormal condition.

The function of the LSSS is to prevent the temperature of the reactor fuel and cladding from reaching the safety limit under any condition of operation. During steady-state operation, a power level of 94.2 kW is required to initiate the onset of nucleate boiling. This is far higher than the maximum power of 1.8 kW, which allows for 50% instrument uncertainties in measuring power level.

For the transients that were analyzed, the temperature of the fuel and cladding reach maximum temperatures of 31°C, assuming reactor trip at 1.8 kW after failure of the first trip. This temperature is far below the safety limit of 530°C.

Amendment No. 12 August 9, 2007

20

b. The conductivity of the primary coolant shall be recorded weekly.

c. The reactor pool water will be at or above the height of the skimmer trough whenever the reactor is operated.

d. Monthly samples of the primary coolant shall be taken to be analyzed for gross alpha and beta activity.

Bases - Weekly surveillance of pool water quality provides assurance that pH and conductivity changes will be detected before significant corrosive damage could occur.

When the reactor pool water is at the skimmer trough level, adequate shielding of more than 13 feet of water is assured.

Analysis of the reactor water for gross alpha and beta activity assures against undetected leaking fuel assemblies.

4.4 Containment Applicability - This specification applies to the surveillance requirements for maintaining the integrity of the reactor room and fuel clad.

Objective - The objective is to assure that the integrity of the reactor room and the fuel clad is maintained, by specifying average surveillance intervals.

Specification -

a. The negative pressure of the reactor room will be recorded weekly.

b. Operation of the inlet and outlet dampers shall be checked semiannually, with no interval to exceed 7 1/2 months.

c. Operation of the air conditioner shall be checked semiannually, with no interval to exceed 7 1/2 months.

d. Representative fuel assemblies shall be inspected annually, with no interval to exceed 15 months.

Bases - Specification a, b, and c check the integrity of the reactor room, and d the integrity of the fuel clad. Based upon past experience these intervals have Amendment No. 12 August 9, 2007

22

5. DESIGN FEATURES 5.1 Site Description 5.1.1 The reactor is located on the ground floor of the Duncan Annex of the Electrical Engineering Building, Purdue University, West Lafayette, Indiana.

5.1.2 The School of Nuclear Engineering controls approximately 5000 square feet.

5.1.3 Access to this area is restricted except when classes are held here.

5.1.4 The reactor room remains locked at all times except for the entry or exit of authorized personnel.

5.1.5 The PUR-1 is housed in a closed room designed to restrict leakage.

5.1.6 The minimum free volume of the reactor room shall be 15,000 cubic feet.

5.1.7 The ventilation system is designed to exhaust air or other gases from the reactor room through an exhaust vent at a minimum of 50 feet above the ground.

5.1.8 Openings into the reactor room consist of the following:

a. Three personnel doors

b. Two locked transformer vault doors

c. Air intake

d. Air exhaust

e. Sewer vent 5.2 Fuel Assemblies 5.2.1 The fuel assemblies shall be MTR type consisting of aluminum clad plates enriched to less than 20% in the U-235 isotope.

5.2.2 A standard fuel assembly shall consist of up to 14 fuel plates containing a maximum of 180 grams of U-235.

Amendment No. 12 August 9, 2007

23 5.2.3 A control fuel assembly shall consist of up to 8 fuel plates containing a maximum of 103 grams of U-235.

5.2.4 Partially loaded fuel assemblies in which some of the fuel plates are replaced by aluminum plates containing no uranium may be used.

5.3 Fuel Storage 5.3.1 All reactor fuel assemblies shall be stored in a geometric array where keff is less than 0.8 for all conditions of moderation and reflection.

5.3.2 Irradiated fuel assemblies and fueled devices shall be stored in an array which will permit sufficient natural convection cooling by water or air such that the temperature of the fuel assemblies or fueled devices will not exceed 100°C.

Amendment No. 12 August 9, 2007

ATTACHMENT TO ORDER OUTLINE OF REACTOR STARTUP REPORT Within six months following the return of the converted reactor to normal operation, submit the following information to the NRC. Information on the HEU core should be presented to the extent it exists.

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

3. Regulating and safety control rod calibrations Measurement of HEU and LEU 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 set points, detector positions and detector output.

5. Shutdown margin Measurement with HEU Measurement with LEU Comparisons with calculations for LEU and if available, HEU

6. Thermal neutron flux distributions Measurements of the core and measured experimental facilities (to the extent available) with HEU and LEU and comparisons with calculations for LEU and if available, HEU.

7. Reactor physics measurements Results of determination of LEU effective delayed neutron fraction, temperature coefficient, and void coefficient to the extent that measurements are possible and comparison with calculations and available HEU core measurements.

Attachment 2

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

9. Primary coolant measurements Results of any primary coolant water sample measurements for fission product activity taken during the first 30 days of LEU operation.

10. Discussion of results Discussion of the comparison of the various results including an explanation of any significant differences that could affect normal operation and accident analyses.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING CONVERSION ORDER TO CONVERT FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUEL FACILITY OPERATING LICENSE NO. R-87 PURDUE UNIVERSITY RESEARCH REACTOR DOCKET NO. 50-182

1.0 INTRODUCTION

Title 10 of the Code of Federal Regulations (10 CFR) Section 50.64 requires licensees of research and test reactors to convert from the use of high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel, unless specifically exempted. Purdue University (Purdue or the licensee) has proposed to convert the fuel in the Purdue University Research Reactor (PUR-1) from HEU to LEU. In a letter dated August 13, 2006, as supplemented on May 3 and June 18, 2007, the licensee submitted its proposal for conversion requesting approval of the fuel conversion and of changes to its Technical Specifications. To support this action the licensee submitted a conversion Safety Analysis Report (SAR) on which the HEU to LEU conversion and the Technical Specification changes are based. This Safety Evaluation Report provides the results of the NRC staffs evaluation of the licensees conversion proposal. The evaluation was carried out according to the guidance found in NUREG-1537.1 The information submitted by the licensee on May 3, 2007, contained significant new information necessitated by the fact that the structural design of the fuel assembly had changed since the August 13, 2006, submission. All references to figures/tables in this Safety Evaluation Report refer to the updated information in the May 3, 2007, response from the licensee to the staffs request for additional information (RAI) dated March 13, 2007.

2.0 EVALUATION 2.1 Summary of Reactor Facility Changes The PUR-1 is a 1 kilowatt thermal power (kW(t)), heterogeneous, pool-type nuclear research reactor that uses Materials Testing Reactor (MTR)-type plate fuel. The HEU to LEU conversion requires the use of different fuel assemblies and core configuration. All of the following aspects of the facility remain unchanged:

1 Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Standard Review Plan and Acceptance Criteria, NUREG 1537, Part 2, U.S. Nuclear Regulatory Commission, February 1996.

• Control rods and drives

• Reactor tank and biological shielding

• Core support structure

• Reactivity control system The graphite neutron reflectors and the neutron source are having their aluminum cans replaced with new cans of a slightly different Al alloy. The impact of this change on the operation or safety of the LEU core was reviewed and determined to be inconsequential.

The LEU fuel assembly (19.75% enriched) has the same basic design (MTR-type fuel) as the present HEU fuel assembly (93% enriched). The LEU fuel assembly contains 14 fuel plates with U3Si2-Al fuel meat while the HEU fuel assembly contains 10 fuel plates of U-Al alloy fuel meat. The cladding of the LEU fuel plates is composed of 6061 Al alloy while the cladding of the HEU fuel plates is composed of 1100 Al alloy. The LEU fuel plates will fit into combs in the redesigned fuel box instead of being bolted together as they are now in the HEU design.

2.2 Comparison with Similar Facilities Already Converted Similar MTR reactors that are cooled by either natural or forced circulation have converted to the same LEU silicide plate-type fuel proposed for the PUR-1 conversion. Some examples are the research reactors at the University of Missouri (at Rolla), Ohio State University, and the University of Florida, which are licensed to operate at power levels of 200 kW(t), 500 kW(t), and 100 kW(t), respectively. There have been no performance issues in the use of this fuel in these reactors.

2.3 Fuel and Core Design The major changes in the fuel composition and the fuel assembly dimensions (as well as other parameters) are given in Table 1. The fuel meat will change from U-Al alloy to a dispersion fuel consisting of U3Si2 in Al. The fuel clad and the dummy plates will change aluminum alloy, from 1100 Al to 6061 Al. Generic aspects of the LEU silicide fuel have been reviewed by NRC and the fuel is approved by the NRC for use in research and test reactors with slab fuel plates.2 The licensee submitted their application for conversion to justify the specific use of the fuel in the PUR-1. The NRC generic approval is for fuel with uranium densities up to 4.8 g/cm3. The PUR-1 LEU fuel will have a uranium density of 3.5 g/cm3.

The width of the fuel meat is reduced from 6.27 cm to 5.96 cm and the clad thickness is also reduced so each LEU fuel plate will be 0.127 cm thick verses the present (HEU) 0.152 cm. The combination of volume change, fuel meat density change, and change in enrichment of U-235 results in a reduction of U-235 per fuel plate from 16.5 g to 12.5K0.35 g.

There will be more fuel plates per assembly with the LEU design than with the HEU design and the plates will be thinner. With an increase in the maximum number of plates in a standard fuel assembly from 10 to 14, the maximum U-235 content of a standard fuel assembly increases 2

Safety Evaluation Report Related to the Evaluation of Low-Enriched Uranium Silicide-Aluminum Dispersion Fuel for Use in Non-Power Reactors, NUREG-1313, U.S. Nuclear Regulatory Commission July 1988.

from 165 g to 180 g. With an increase in the maximum number of plates in a control fuel assembly from 6 to 8, the maximum U-235 content of a control fuel assembly increases from 99 g to 103 g. To limit excess reactivity, it is expected that eleven fuel assemblies will have 13 fueled plates (plus one dummy plate) and two assemblies will have 12 fueled plates (and two dummy plates). The nominal U-235 mass in the core increases from 2.046 kg to 2.388 kg. This is expected due to the need to compensate for the increase in U-238 in the core. The critical mass may change if a need to increase or decrease reactivity is found during startup testing.

This adjustment may be required to maintain the core within the technical specification excess reactivity limit of 0.6%.

Table 1: Summary of Key Nominal Design Parameters of HEU (Current) and LEU (Expected) Cores Design Data HEU LEU Fuel Type MTR Plate MTR Plate Fuel Meat Composition U-Al Alloy U3Si2-Al Fuel Enrichment U-235 (nominal) 93% 19.75%

Mass of U-235 per plate (g. nominal) 16.5 12.5 Fuel Meat Dimensions Width (mm) 62.7 59.6 Thickness (mm) 0.508 0.508 Height (mm) 600.1 600.1 Fuel Plate Dimensions Width (mm) 70.2 70.2 Thickness (mm) 1.52 1.27 Height (mm) 638.6 638.6 Cladding Composition 1100 Al 6061 Al Cladding Thickness (mm) 0.508 0.381 Dummy Plate Composition 1100 Al 6061 Al Dummy Plate Dimensions Same as Fuel Same as Fuel Standard Fuel Assemblies Number of standard assemblies 13 13 Number of plates per standard assembly 10 14 Control Fuel Assemblies Number of control assemblies 3 3 Number of plates per control assembly 6 8 Total plates in core (fuel and dummy) 148 206 Fuel plates in core 124 191 Dummy plates in core 24 15 Plate spacing in standard assemblies (mm) 5.26 3.66 Plate spacing in control assemblies (mm) 5.26 4.60 The reactor core geometry is a 4x4 square of fueled assemblies surrounded by graphite assemblies on three sides and an irradiation facility on the fourth side. Thirteen of the fueled

assemblies are standard fuel assemblies and three are control assemblies. This configuration will not change with the conversion. However, the orientation of the standard LEU fuel assemblies will be rotated by 90 degrees in the LEU core such that the fuel plates will be parallel to the plates in the control assemblies. The handle for the fuel assembly can, being normal to the plates, will restrict the possible movement of fuel plates.

The staff has reviewed the proposed fuel and core design of the LEU reactor. The staff concludes that the conversion from HEU to LEU fuel will not impact the overall basic design of the core and its control. The major change will be a larger number of thinner plates per fuel assembly. The power of the core will remain at 1 kW(t) so the average power per plate and per fuel assembly will both be reduced which is conservative. Therefore, the staff finds the fuel and core design acceptable.

2.4 Nuclear Design 2.4.1 Calculational Methodology In order to carry out the PUR-1 neutronic analysis, a Monte Carlo N-Particle (MCNP) model was developed for a fresh HEU and a fresh LEU core using version 5 of MCNP with the Evaluated Nuclear Data File (ENDF)-VI.5 cross section library. MCNP5 is a state-of-the-art program used for many nuclear reactor analyses. Only fresh cores were considered since there is negligible burnup on the PUR-1 core given the nominal 1 kW(t) operating power for the PUR-1. As such, there is no fission product inventory developed for the PUR-1 model. The MCNP model was validated for the HEU core by comparing calculated to measured values of keff for the core with the control rods at various positions. This validation work established a consistent bias (over-prediction) of approximately 0.32% k/k for the calculated values when compared to the measured values for keff. In addition, the worth of each control rod was calculated and compared with measured values. The calculated worths of the control rods were consistently lower than the measured worths. Shim-safety rod 1 was 3.11% lower, shim-safety rod 2 was 6.74% lower and the regulating rod was 3.57% lower. These differences between calculated keff and control rod reactivity worth are within typical bounds. In subsequent sections, other comparisons of measured vs. calculated values of specific core parameters are presented with similar agreement. The licensee has assumed that the MCNP calculations for the LEU core will exhibit the same 0.32% k/k bias for calculated values of keff. Parameters for operation of the LEU core, such as fuel loading and control rod worths, will rely on measured values using the calculations as guidelines. The comparisons show reasonable agreement. Therefore, the staff concludes that the calculational methodology used by the licensee is acceptable.

2.4.2 Neutron Flux and Power Distributions Conversion is expected to result in slightly less peaking of the neutron flux, and hence the power distribution. The licensee states that the average neutron flux in the fueled region will go from 1.2E10 n/(cm2-s) to 1.38E10 n/(cm2-s) while the maximum neutron flux will go from 2.1E10 n/(cm2-s) to 2.01E10 n/(cm2-s). In Section 4.5.4 of the conversion SAR, Operating Conditions, results of extensive power distribution analyses are provided. These analyses provide a surrogate for the neutron flux distributions. The power distributions show reasonable results and the hottest fuel plate in the LEU core will have approximately 25% lower power to dissipate than the hottest fuel plate in the HEU core. The absolute value of the thermal flux at two locations was given and shown not to change significantly for the conversion. The changes

to neutron flux and power distributions due to conversion are not significant and are as expected. Therefore, the staff concludes that neutron flux and power distributions in the LEU core are acceptable.

2.4.3 Excess Reactivity, Control Rod Worth and Shutdown Margin The calculated excess reactivity for the LEU core is 0.35% k/k (this value takes into account the calculational bias of 0.32% k/k) which satisfies the Technical Specification 3.1 limit of 0.6% k/k. The measured excess reactivity for the HEU core is 0.43% k/k. Note that the procedures in the approach to criticality should assure that the limit (0.6% k/k) is met. This may determine any changes that are needed for the number of fuel plates within a fuel assembly. Therefore, the staff concludes that Technical Specification 3.1 will continue to be met after conversion.

The control rod worths were calculated using the MCNP model for both the HEU and LEU cores. The worth of shim-safety rod 1 is calculated to be reduced from 0.0436 k/k for the HEU core to 0.0377 k/k for the LEU core, a 14% decrease, and the worth of shim-safety rod 2 is calculated to be reduced from 0.0235 k/k for the HEU core to 0.0189 k/k for the LEU core, a 20% decrease. The regulating rod worth is calculated to decrease from 0.0027 k/k to 0.0023 k/k, an insignificant change in terms of absolute reactivity and a reduction that is smaller than the uncertainty of the calculations. Given the reduction in excess reactivity of the core, the staff concludes that the worths of the control rods in the LEU core are acceptable.

Using the MCNP model, control rod calibration curves were calculated for each of the three rods. Using the rate of insertion of the present control rod drive mechanisms (no change in the control rods are proposed for the conversion), the maximum reactivity insertion rates were calculated. The results were consistent with the other nuclear design parameters calculated by the licensee such as control rod worths. The staff concludes that the resulting reactivity insertion rates are acceptable.

The shutdown margin for each core was calculated with shim-safety rod 2 (the lowest-worth rod) inserted and shim-safety rod 1 and the regulating rod withdrawn. Technical Specification 3.1.a, Reactivity Limits, states that this margin should be at least 1% k/k. Calculations indicate the LEU core will comply with Technical Specification 3.1 with a value of 1.58% k/k. Note that this would be true even if credit was not taken for the 0.32% k/k bias (over prediction) in the excess reactivity. Therefore, the staff concludes that Technical Specification 3.1 will continue to be met after conversion with no physical changes to the control rods which is acceptable.

2.4.4 Dynamic Parameters The prompt neutron lifetime, , and the effective delayed neutron fraction, eff, change slightly as a result of the conversion from HEU to LEU fuel. eff was calculated by determining keff with and without delayed neutrons and is expected to change from 0.00798 for the HEU core to 0.00790 for the LEU core, a 1.0% decrease. These numbers are calculated using a formula that includes keff with and without delayed neutrons and takes into account the 0.32% k/k bias for the over prediction of keff. The effect of not taking into account the bias is to make eff 0.3%

smaller which is less than the statistical error of the analysis (since the analysis is based on a

Monte Carlo calculation). For the transient analysis the unbiased numbers for eff are used, however, since they are lower, this adds a small amount of conservatism. The staff concludes that these changes are acceptable and do not significantly change the dynamic behavior of the core.

The neutron lifetime was calculated using the 1/v method and is expected to change from 76.7 s for the HEU core to 81.3 s for the LEU core, a 6.0% increase. The staff concludes that this is reasonable and the new value is not expected to change the basic behavior of the core significantly.

The temperature and void coefficients of reactivity were calculated with the MCNP model. The results are similar for the HEU and LEU cores. The temperature regimes studied were between 20 and 100EC for the water and between 20 and 127EC for the fuel. In all cases the temperature and the void coefficients were negative for the heating of water within the core.

Heating the water only in the reflector has a small positive temperature coefficient. However, no scenarios where only the water in the reflector increases in temperature were identified. It would take a long time for any transient to heat water in the reflector and this would also lead to heating water in the core where the temperature coefficient is negative. The overall effect of the water heating would be negative. Hence, this is not considered to have a significant safety impact. Therefore, the staff concludes that the temperature and void coefficients of reactivity in the LEU core are acceptable.

2.4.5 Conclusions For the key neutronic characteristics of the PUR-1 core (i.e., flux, control rod worths, shutdown margin, reactivity coefficients and other dynamic parameters) the conversion from HEU to LEU fuel will not cause any significant changes. All changes have been calculated using established methods and are taken into account in the safety analysis. The staff concludes that the changes in nuclear design due to conversion are acceptable.

2.5 Thermal-Hydraulic Design 2.5.1 Core Power Distribution The power profile of the PUR-1 core was obtained from heating tallies included in the MCNP model. The tallies accounted for heating due to fission, capture and photon scattering. The heating tallies calculated the power in each of the 16 fuel assemblies and also the individual plate power within an assembly. In addition, axial and radial (across the width of a fuel plate) profiles were determined for the hottest fuel plate.

Five different critical rod positions were used to find the maximum local power density in the HEU core. It was determined that the banked rod configuration with all three control rods at the same height (corresponding to Case 3 in Table 4-7 given in the response to RAI number 8) resulted in the maximum power density in the HEU core. It is in plate 262 in bundle 4-4, a control assembly where shim-safety rod 1 is located. Plate 262 is also the plate with the highest power (10.97 W) in the HEU core and the plate is adjacent to the large water hole that is created when shim-safety rod 1 is withdrawn from the reactor core. The fuel assembly with the maximum power is bundle 3-3, an interior core position. However this bundle has nine fuel

plates (versus only six in bundle 4-4) and the hottest plate (plate 89) in this bundle is thus less limiting than plate 262 for thermal-hydraulic analysis.

A similar process was used to determine the hottest plate in the LEU core. Similar to the HEU core, calculations for the power distribution for the new LEU core design were evaluated for the banked rod configuration. The hottest plate is plate 1348 (8.07 W) and, like in the HEU core, is in fuel bundle 4-4 (eight fuel plates) where shim-safety rod 1 is located. For the LEU fuel assemblies the coolant channel is wider in the control assemblies than the standard fuel assemblies. Thus it becomes necessary to perform thermal-hydraulic analyses for two additional fuel plates located in the two highest powered standard fuel assemblies in core positions 3-4 (78.42 W) and 4-3 (80.71 W). The corresponding fuel plates of interest in assemblies 3-4 (13 fuel plates) and 4-3 (13 fuel plates) are plates 1228 (6.51W) and 1315 (6.41 W) respectively and both plates face the centerline of the core. The staff has reviewed the methods that the licensee has used to determine the core power distribution and based on the use of a well established code with appropriate inputs to determine the maximum local power density finds that the methods and results of the calculations are acceptable.

2.5.2 Calculational Methodology PUR-1 relies on natural circulation to cool the fuel plates. For steady-state thermal-hydraulic analysis the calculations were performed by a computer code NATCON v2.0 developed at Argonne National Laboratory. Details of the code are provided in Appendix 1 of the conversion SAR. NATCON calculates the buoyancy driven flow between the plates in an assembly, axial temperatures in the coolant and fuel plate surface and centerline, and the approach to onset of nucleate boiling (ONB). The code also calculates the ONB ratio (ONBR) and the departure-from-nucleate-boiling ratio (DNBR).

NATCON assumes incompressible laminar single-phase liquid flow that is thermally expandable. Coolant flow is determined from a balance between buoyancy and wall friction.

Six hot channel factors are used in NATCON, three global/systematic and three local/random factors. The systematic uncertainties are:

• Measurement of reactor power.

• Flow dependence of friction factor.

• Heat transfer coefficient (evaluated as uncertainty in Nusselt number correlation).

The three local hot channel factors are for:

• Coolant bulk temperature rise.

• Coolant film temperature rise (for the LEU core this factor includes a multiplier representing the effect of power variation along the width of a fuel plate).

• Heat flux from cladding surface.

The ONB power is determined with the application of the hot channel factors (systematic and random). The margin of safety under nominal operating conditions (1 kW(t) power) is presented in two parameters, the ONB ratio and the margin to incipient boiling. The ONB ratio (ONBR) is defined as:

ONBR = (Tincp,i - T0) / (Twall,i - T0) where T0 = Bulk water temperature at the coolant channel inlet.

Tincp,i = Incipient boiling temperature in coolant node I with only systematic hot channel factors applied.

Twall,i = Cladding surface temperature in node I with all six hot channel factors applied.

The margin to incipient boiling is the minimum temperature difference (TONB - TW ) in the hottest channel where TW is the cladding surface temperature with all hot channel factors applied and TONB is the local onset-of-nucleate-boiling temperature with only systematic hot channel factors applied.

2.5.3 Results of Thermal-Hydraulic Analysis Geometrically the HEU and LEU fuel assemblies are very similar. The major difference is in the coolant channel spacing. In order to accommodate more plates in an LEU fuel assembly the plate-to-plate channel spacing is reduced as compared to the HEU fuel assembly. In addition for the LEU fuel, the plate to plate channel spacing is smaller in a standard assembly than a control assembly.

Hot channel factors were applied to the thermal-hydraulic analysis of the HEU and LEU core.

The calculation of these factors was modified for the new LEU core to include additional uncertainties:

• Friction factor in developing laminar flow.

• Effect of temperature on coolant viscosity.

• Variation of power along the width of a fuel plate.

The net effect of including additional uncertainties is a slight reduction (less than 10%) in the ONB power, which is more than an order of magnitude higher than the nominal operating power of 1 kW(t) for the PUR-1. In the HEU core the plate-to-plate spacing is the same for the standard and control assemblies. The limiting channel is then associated with the highest powered fuel plate (plate 262). In the case of the LEU core, three plates were identified as potentially being the limiting plate because of the different plate-to-plate spacings in control rod assemblies and standard assemblies.

The NATCON code was used to search for the reactor power at which ONB was first reached in the hot channel. The ONB power for the HEU and LEU core are 76.3 kW(t) and 94.2 kW(t),

respectively (for the HEU core by including uncertainties in friction factor and coolant viscosity the ONB power is reduced to 75.9 kW(t)). In both cases it is a plate to plate channel and not a plate to wall channel that is limiting. In the LEU core it is plate 1348 (in a control assembly) that is limiting. The other plates that were considered are 1228 and 1335 which are in a standard assembly with narrower flow channels than those in control assemblies (144 vs. 181 mil).

Thermal-hydraulic analysis for PUR-1 demonstrates that the reduction in channel thickness brings the coolant channels in the standard assemblies closer to the optimum value (about 100 mil) and will result in a higher ONB power with natural convection cooling when compared with

coolant channels in control assemblies. It is also noted that the analysis for the new LEU core was performed with a 20 mil (vs. the 15 mil in the fuel specification) uncertainty on the 181 mil channel thickness for the new LEU control assemblies.

A pool temperature of 27EC was used in the LEU thermal-hydraulic analysis. This is also the temperature assumed for the coolant inlet temperature. The sensitivity of the ONB power to the pool temperature was evaluated by repeating the ONB power determination with a hypothetical pool temperature of 35EC and a hypothetical inlet loss coefficient of 10.0 (increased from 0.5).

The ONB power was reduced from 94.2 to 79.3 kW(t) under the hypothetical conditions, indicating a large margin from the PUR-1 nominal operating power of 1 kW(t). Data from the licensee shows that since 1993, the maximum pool temperature reached was about 30EC so the use of a pool temperature of 35EC in the calculations is bounding.

The thermal-hydraulic analysis shows that at the ONB power of 94.2 kW(t), the fuel and clad temperature is less than 115EC. The thermal-hydraulic analysis also shows that the maximum temperatures for fuel, clad, and coolant under the nominal operating power of 1 kW(t) will only be a few degrees above the coolant channel inlet temperature. The ONB power calculation demonstrates that the PUR-1 has sufficient margin to its thermal limit (ONB).

2.5.4 Conclusions The staff concludes that the thermal-hydraulic analysis reported for the PUR-1 conversion adequately demonstrates that the conversion from an HEU to an LEU core results in no significant decrease in safety margins in regard to thermal-hydraulic conditions. The margin to ONB was shown to be very large. The analyses were done with qualified calculational methods and conservative or justifiable assumptions.

2.6 Accident Analysis The conversion SAR considered four hypothetical accidents: the maximum hypothetical accident (MHA), rapid addition of reactivity accident, reduction in cooling accident, and continuous control rod withdrawal. Subsequently in the RAI response of May 3, 2007, two new reactivity insertion accidents were analyzed for both the HEU and LEU cores. These new accidents were for the rapid and slow insertion of 0.6% k/k, the maximum excess reactivity allowed by the PUR-1 Technical Specifications, without scram.

2.6.1 Reactivity Addition Accidents 2.6.1.1 Rapid Addition of Reactivity Accident With Scram The accident scenario assumed the rapid insertion (0.1s) of the maximum worth of moveable and unsecured experiments (0.3% k/k) as specified in the Technical Specifications. The reactivity transient was analyzed by using the computer code Program for the Analysis of Reactor Transients/Argonne National Laboratory (PARET/ANL).3 No credit was given to the period trip and the power trip was set to occur at 1.8 kW (and not the limiting safety system setting value of 1.2 kW). A scram delay of 0.1 second was assumed. The results show that in both the HEU and the LEU cores the reactor power increases from 1 kW(t) to the trip setting of 1.8 kW(t) in less than 1 second. For both cores the maximum reactor power (1.83 kW(t))

3 A. Olson, A Users Guide to the PARET/ANL V7.2 Code, Draft ANL report, April 2, 2007.

is only a fraction of a kilowatt above the trip setting and the maximum clad temperature is less than 1EC above the initial value of 29EC.

2.6.1.2 Continuous Control Rod Withdrawal This is another reactivity transient considered in the conversion SAR. The accident scenario assumes a stuck switch on the control rod of the highest worth (shim-safety rod 1) while the rod is being raised from the 1 kW(t) full power level. Reactor trip is assumed to occur at 1.8 kW(t).

The assumed reactivity insertion rate of 0.04% k/k/s is higher than the maximum reactivity insertion rates both calculated and measured. The results show that the maximum clad temperature rises by less than 1EC for both the HEU and LEU cores.

The staff also evaluated a reactivity transient adding 0.04% k/k/s for the LEU core starting at a reactor power of 10 w(t). This results in a larger total reactivity addition than the transient considered by the licensee. While the peak power is slightly higher (2.04 kW(t) vs 1.83 kW(t)),

the maximum clad temperature is lower due to increased heat transfer given the longer time period of the transient.

2.6.1.3 Rapid and Slow Reactivity Insertion Accidents Without Scram A set of reactivity insertion accidents without scram was performed using the PARET/ANL code for both the HEU and the LEU core. The rapid and slow insertion of reactivity of 0.6% k/k, the maximum excess reactivity allowed by the PUR-1 Technical Specifications is analyzed. The scenarios for the rapid and slow reactivity insertion are similar to the previously analyzed reactivity accidents discussed earlier in Sections 2.6.1.1 and 2.6.1.2 respectively. The only difference is that all reactor trips (scrams) are assumed not to function as designed. This is a very conservative assumption given the redundancy and diversity of reactor scrams in the reactor safety system.

For the rapid insertion of 0.6% k/k in 0.1 second the peak reactor power is calculated to be 1.63 MW(t) for the HEU core and 1.55 MW(t) for the LEU core. The lower peak power for the LEU core is due to the Doppler feedback effect in the LEU fuel. The peak clad temperatures for the two cores are 134EC and 120EC for the HEU and LEU core respectively, indicative of a lower maximum heat flux for the fuel plates in the LEU core. Because of the relatively low heat flux, fuel centerline temperatures are only slightly higher.

The slow insertion of 0.6% k/k reactivity at a rate of 0.04% k/k/s was analyzed for the LEU core assuming no scram was initiated. The peak power and peak clad temperature are 1.55 MW(t) and 120EC, the same as for the rapid insertion of reactivity. This is consistent with the prediction by PARET/ANL that the inserted reactivity is balanced by the negative reactivity feedbacks from coolant temperature, coolant void (density), and fuel temperature (Doppler).

The PARET analysis also includes a comparison of code calculated results and results from the special power excursion reactor test (SPERT)-IV B-1 test. The HEU fueled MTR-type core used in the SPERT test was similar to the PUR-1 core. The PARET code results are in good agreement with the first power peak and the corresponding maximum clad temperature in a 0.6% k/k rapid reactivity insertion test (the SPERT test was terminated by an operator-initiated

scram after 30 seconds). This comparison supports the validation of the ability of the PARET code to predict reactor power and maximum clad temperature for the PUR-1 core. The predicted maximum clad temperature of 120EC for the PUR-1 fuel (both HEU and LEU) in a maximum reactivity insertion accident (0.6% k/K) is well below the PUR-1 fuel safety limit of 530EC (the new safety limit is discussed in Section 2.8.2).

It is noted that the PARET analyses for PUR-1 were based on an earlier design of the LEU core (the core that was presented in the conversion SAR). No new analysis was presented for the new LEU core design (the core that was presented in the RAI responses from the licensee dated May 3, 2007) because the licensee has evaluated the impact of the revised feedback coefficients and comes to the conclusion that the less negative coolant temperature coefficient in the new LEU core design will be compensated for by the stronger feedback coefficients for the water void and fuel temperature. As part of the technical evaluation of the conversion SAR, a sensitivity analysis was performed by the NRC to assess the impact of the less negative coolant temperature coefficient in the new LEU core design. The analysis was based on the same input deck and code version as used by the licensee and as such changes in the spacing of the fuel plates in the new fuel assemblies were not included in the sensitivity analysis. The sensitivity analysis was for the rapid insertion of 0.6% k/k in 0.1 second without scram.

Changes to the input are summarized in Table 2.

Table 2. Input Parameters for the Sensitivity Analysis PARET Input PUR-1 SAR Sensitivity Case (old LEU design) (new LEU design)

Effective delayed neutron fraction (%) 0.787 0.784 Reactivity insertion (0.6% k/k) 0.7624$ 0.7653$

Coolant temperature coefficient 0.01238 $/K 0.01154 $/K Coolant void coefficient 0.2411 $/% void use old value Fuel temperature coefficient 8.91x10-4 $/K use old value The sensitivity analysis resulted in a roughly 10% increase in peak power and maximum clad temperature at 400 seconds, 1.71 MW(t) and 128EC versus the PUR-1 SAR results of 1.55 MW(t) and 120EC for the old LEU design. It is noted that the sensitivity analysis only updated one of the three feedback coefficients, using only the less negative coolant temperature coefficient for the new LEU design (as shown in Table 2). If the two other negative feedback coefficients (coolant void and fuel temperature) for the new LEU design were also included, the increase in peak power and clad temperature would have been lower than the increase over the old LEU design given above. The sensitivity analysis bounds the effects of the updated reactivity coefficients for the new LEU design to a 10% increase in peak power and maximum clad temperature over the old LEU design. The staff concludes that the resulting increase in maximum clad temperature of 128EC for the new LEU core design is still much lower than the safety limit of 530EC set for PUR-1, and therefore is acceptable.

2.6.1.4 Conclusions The staff concludes that the licensee has analyzed acceptable reactivity insertion transients.

The reactor safety limit is not exceeded during the transients, therefore, the LEU reactor behavior under reactivity insertion transients is acceptable.

2.6.2 Reduction in Cooling Accidents A complete loss of moderator/coolant in the PUR-1 has been evaluated previously in the 1986 SAR. The fuel temperature rise in the HEU core was estimated to be 9.5EC assuming adiabatic conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the hypothetical loss-of-coolant accident. A similar temperature rise is expected for the LEU core. This expectation is supported by the identical maximum fuel temperature calculated at the 1 kW(t) operating conditions for the HEU and LEU cores. The staff concludes that since the cladding integrity is maintained in a LOCA for the HEU core and the LEU core is predicted to behave similarly to the HEU core, the consequences of a LOCA for an LEU core will not be more severe than that for a HEU core and is therefore, acceptable.

2.6.3 Maximum Hypothetical Accident (MHA)

The MHA for the PUR-I is the failure of a fueled experiment. This was reviewed and accepted by the NRC as discussed in NUREG-1283, Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at Purdue University. The experiment is generating 1 W(t) of power and the failure releases fission products. A fueled experiment failure rather than core fuel failure was picked for the MHA at Purdue because it was considered more likely that a fueled experiment could suffer loss of encapsulation and release of fission products.

The MTR fuel design has good fission product retention behavior and the low power level of the PUR-1 results in low decay heat and lack of driving force to release fission products from the fuel. None of the technical specification limits on the fission product inventory of fueled experiments are being changed by the conversion to LEU fuel. Because the MHA for the PUR-I reactor is the failure of a fueled experiment, the staff concludes that the previously performed analysis of the MHA is applicable for the LEU fueled reactor and the conclusions reached in NUREG-1283 are not changed by the reactor conversion.

2.6.4 Conclusions The licensee has demonstrated that the conversion from HEU to LEU fuel does not introduce the potential of a new reactivity addition accident not previously analyzed for the HEU-fueled reactor or significantly increase the consequences beyond those for a reactor accident in the existing HEU core. The licensee demonstrated this by presenting the basic neutronic, thermal-hydraulic, and physical similarity between the HEU and LEU cores, and an analysis showing that the conclusions in the HEU analysis regarding the consequences of both the maximum credible reactivity accident and hypothetical loss-of-coolant accident are still applicable to the proposed LEU-fueled reactor. The analyses showed that the reactor safety limits on fuel and clad temperatures would not be exceeded for these reactivity accidents or for a hypothetical loss-of-coolant accident. Furthermore, the licensee analyzed reactivity accidents without reactor trip to demonstrate the ability of the reactor to respond within the safety limits. The radiological consequences to the public and occupational workers at the PUR-I from a

postulated failed fueled experiment MHA for the proposed LEU-fueled reactor are expected to be the same as the radiological consequences calculated for the HEU-fueled reactor, which was acceptable to the NRC staff. As a result of this review, the staff has concluded that continued operation of the reactor poses no undue risk from a radiological standpoint to the public or the staff of the PUR-I from the maximum hypothetical accident.

2.7 Reactor Start-Up Testing The licensee plans to make sub-critical measurements for the LEU fuel loading. The draft procedure outlined in Appendix 2 of the SAR follows the standard 1/M method for the loading of a critical assembly. The final procedure will be developed by the licensee following the existing technical specification approval process used for all procedures at PUR-1. The staff concludes that the licensees draft procedure is sufficiently detailed to result in the safe loading of the reactor with the LEU fuel.

The licensee is to submit a start-up report to the NRC on the results of the start-up testing. The report will contain information on control rod and power calibrations, measurements of temperature and void coefficients, excess reactivity, reactivity insertion rates, shutdown margin and experimental facility neutron flux levels, and radiation surveys and effluent measurements.

The licensee will also complete a number of normal surveillances to ensure operability of components and systems. The staff concludes that the licensees testing program will provide verification of key LEU reactor functions, and therefore, is acceptable.

2.8 Proposed Changes to License Conditions and Technical Specifications For the PUR-1 HEU to LEU conversion, the licensee has proposed changes to the license conditions for special nuclear material possession limits and technical specifications.

2.8.1 Proposed Changes to License Conditions License condition 2.B.(2) is changed to reflect receipt, possession and use of special nuclear material after conversion. The license condition currently reads as follows:

2.B.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material", to receive, possess, and use up to three (3) kilograms of uranium-235 contained in uranium enriched in the isotope uranium-235 and up to 80.0 grams of plutonium contained in encapsulated plutonium-beryllium sources, both in connection with operation of the facility; and Based on the licensee proposed possession limits, the license condition reads as follows:

2.B.(2) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to receive, possess, and use: (1) up to 3.8 kilograms of contained uranium-235 of enrichment of less than 20 percent in the form of materials testing reactor (MTR)-type reactor fuel; (2) up to 80.0 grams of plutonium contained in encapsulated plutonium-beryllium sources: and (3) up to 100 grams of contained uranium-235 of any enrichment in the form of

fission chambers, flux foils and fueled experiments, all used in connection with operation of the facility; Up to 3.8 kilograms of contained uranium-235 of enrichment of less than 20 percent in the form of reactor fuel replaces the existing possession limit of 3 kilograms of uranium of any enrichment or form. After the reactor is converted, the licensee has a continuing need to receive, possess and use small amounts of HEU to allow continued operation of the reactor (e.g., fission chambers) and conduct of the experimental program (e.g., flux foils and fueled experiments). A new possession limit of up to 100 grams of contained uranium-235 of any enrichment in the form of fission chambers, flux foils and fueled experiments is added to the license condition.

License condition 2.B.(4) was added to the license by order dated June 21, 2007, as part of the conversion process to allow the licensee to possess the LEU fuel needed for conversion prior to this order. The authority for possession of the LEU fuel has been moved to license condition 2.B.(2) as discussed above. License condition 2.B.(4) is revised to allow possession, but not use, of the existing HEU core until it is removed from the licensees site. The revised license condition reads as follows:

2.B.(4) Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," to possess, but not use, up to 3 kilograms of contained uranium-235 at equal to or greater than 20 percent enrichment in the form of materials testing reactor (MTR)-type reactor fuel until the existing inventory of this fuel is removed from the facility.

The staff has reviewed the possession limits associated with conversion of the reactor and concludes that the limits are appropriate for the converted reactor.

License condition 2.C.(2), which incorporates the technical specifications into the license, is changed to incorporate the technical specifications changes needed for conversion as discussed below into the license.

2.8.2 Proposed Changes to the Technical Specifications The following paragraphs discuss the proposed changes to the technical specifications.

Section 2.1, Safety Limits: A new safety limit is introduced in the conversion SAR. The safety limit specifies that the fuel and cladding temperatures shall not exceed 530EC (986EF). The 530EC temperature limit ensures that the cladding blister temperature (a possible forerunner of fuel failure) of approximately 550EC will not be reached under any conditions of operation. This temperature is based on measurements (NUREG-1313) of first fission product release from plate reactor fuel and is accepted by NRC as discussed in NUREG-1537. Hence, clad failure is directly related to the measured blister temperature (with added 20EC conservatism). The previous safety limit was on reactor power which was not directly related to protection of the primary fission product barrier. The reactor power parameter continues as the limiting safety system setting. Because the proposed safety limit is directly related to protection of the fuel cladding which is the primary barrier to the uncontrolled release of fission products from the fuel and meets the requirements in 10 CFR 50.36(c)(1), it is acceptable to the staff.

Section 2.2, Limiting Safety System Settings: Based on conservative nominal operating conditions and accident analysis, the licensee has proposed continuing to use a limiting safety system setting (LSSS) of 1.2 kW(t). The basis of the technical specification has been revised to reflect the conversion SAR. The LSSS is set at 20% above the nominal operating power of 1 kW(t) and as shown in the thermal-hydraulic analysis has a wide margin to the ONB power level for the LEU core of 94.2 kW(t). ONB usually occurs at power levels lower than the burnout point where heat transfer is degraded by departure from nucleate boiling or liquid film dryout.

Reactor fuel and clad temperature at 94.2 kW(t) is less than 115EC and at 1.2 kW(t) is less than 40EC (assuming a coolant inlet temperature of 35EC). Thus the LSSS setting of 1.2 kW(t) provides adequate protection against exceeding the safety limit during steady-state operation.

The accident analysis for a rapid reactivity insertion (see Section 2.6.1) conservatively assumed a trip setpoint of 1.8 kW. The maximum power reached was only 1.83 kW(t) and the maximum clad temperature increased by less than 1EC above the initial value of 30EC. Thus the LSSS of 1.2 kW adequately protects PUR-1 against ONB and from approaching the safety limit.

Section 4.4, Containment: The licensee has proposed changes to Technical Specification 4.4 d to replace fuel plate inspection with fuel assembly inspection. The specification currently reads:

d. Representative fuel plates shall be inspected annually, with no interval to exceed 15 months.

The licensee has proposed changing this specification to read:

d. Representative fuel assemblies shall be inspected annually, with no interval to exceed 15 months.

This change would eliminate the need for disassembly of the fuel assemblies for fuel plate inspection. Disassembly of fuel assemblies increases the potential for damaging fuel. The fuel assembly inspection will look for corrosion, channel blockage and warped or bloated plates.

Over forty years of fuel plate inspections have not revealed any significant degradation. This is expected given the low power level of the reactor. Technical Specification 4.3 d, which requires monthly testing of the primary coolant for gross alpha and beta contamination, continues in effect. This testing will give indication of fuel clad failure. The staff concludes that the proposed visual assembly inspection (in conjunction with monitoring of the primary coolant for contamination) is acceptable.

Section 5.2, Fuel Assemblies: The licensee has proposed changes to parts of Technical Specification 5.2 to reflect the change in the fuel and fuel plates to LEU. The affected parts of the specification currently read:

5.2.1 The fuel assemblies shall be MTR type consisting of aluminum clad plates enriched to approximately 93% in the U-235 isotope.

5.2.2. A standard fuel assembly shall consist of 10 fuel plates containing a maximum of 165 grams of U-235.

5.2.3 A control fuel assembly shall consist of 6 fuel plates containing a maximum of 99 grams of U-235.

The licensee has proposed changing these specifications to read:

5.2.1 The fuel assemblies shall be MTR type consisting of aluminum clad plates enriched less than 20% in the U-235 isotope.

5.2.2. A standard fuel assembly shall consist of up to 14 fuel plates containing a maximum of 180 grams of U-235.

5.2.3 A control fuel assembly shall consist of up to 8 fuel plates containing a maximum of 103 grams of U-235.

The change to Technical Specification 5.2.1 would reflect the fuel assemblies consisting of aluminum clad plates enriched to approximately 19.75% in the U-235 isotope. The licensee had originally proposed wording of enriched up to 20% in the U-235 isotope. To better reflect the definition of LEU, the wording is changed to enriched less than 20% in the U-235 isotope.

This change was discussed with and agreed to by the Director of Radiation Laboratories for the licensee during a telephone conversation with the NRC staff on July 9, 2007.

The change to Technical Specification 5.2.2 would reflect the standard fuel assembly consisting of a maximum of 14 fuel plates containing up to 180 grams of U-235. The change to Technical Specification 5.2.3 would reflect the control fuel assembly consisting of up to 8 fuel plates containing a maximum of 103 grams of U-235. Because the proposed changes reflect the characteristics of the proposed LEU fuel, the staff finds these proposed changes to Technical Specification 5.2 to be acceptable.

The staff has reviewed all of the proposed changes to the technical specifications. The staff concludes that these changes to the technical specifications are needed for the conversion of the reactor to LEU fuel. The licensee has justified the technical bases for these changes to the technical specifications as discussed above. The staff concludes that the changes to the technical specifications continues to meet the regulations in 10 CFR 50.36 and that the changes to the technical specifications are therefore, acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

In accordance with 10 CFR 51.10(d), an Order is not subject to Section 102 of the National Environmental Policy Act. The NRC staff notes, however, that even if these changes were not being imposed by an Order, pursuant to 10 CFR 51.22(b), the changes would not require an environmental impact statement or environmental assessment.

The changes involve use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes in inspection and surveillance requirements. The NRC staff has determined that the changes involve no significant hazards consideration, no significant increase in the amounts, and no significant change in the types, of any effluents that may be released off site, and no significant increase in individual or cumulative occupational radiation exposure.

4.0 CONCLUSION

S The NRC staff has reviewed and evaluated the operational and safety factors affected by the use of LEU fuel in place of HEU fuel in the PUR-1. The staff has concluded, on the basis of the considerations discussed above that (1) the proposal by the licensee for conversion of the reactor to LEU fuel is consistent with and in furtherance of the requirements of 10 CFR 50.64; (2) the conversion, as proposed, does not involve a significant hazards consideration because the amendment does not involve a significant increase in the probability or consequences of accidents previously evaluated, create the possibility of a new kind of accident or a different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety; (3) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed activities; and (4) such activities will be conducted in compliance with the Commission's regulations and the issuance of this Order will not be inimical to the common defense and security or the health and safety of the public. Accordingly, it is concluded that an enforcement order as described above should be issued pursuant to 10 CFR 50.64(c)(3).

Principal Contributors: David J. Diamond, Brookhaven National Laboratory (BNL)

Lap-Yan Cheng, BNL Albert Hanson, BNL Richard Deem, BNL Alexander Adams, Jr., NRC William C. Schuster IV, NRC Dated: August 9, 2007