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U.S. Geological Survey Triga Reactor, Docket Number 50-274, License Number R-113, License Amendment Request 2020-02, LAR 14, Rev. 00
	ML20275A267
Person / Time
Site:	U.S. Geological Survey
Issue date:	10/01/2020
From:	Wallick J; US Dept of Interior, Geological Survey (USGS)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML20275A267 (67)
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Page 1 of 7 U.S. Geological Survey Rocky Mountain Region Box 25046 M.S. 974 Denver Federal Center Denver, Colorado 80225 Thursday, October 01, 2020 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

SUBJECT:

U.S. Geological Survey TRIGA Reactor, Docket Number 50-274, License Number R-113, License Amendment Request 2020-02, LAR 14, Rev. 00 Purpose By way of this letter, the U.S. Geological Survey (USGS) is requesting approval for a license amendment to the GSTR reactor facility license R-113 by the U. S. Nuclear Regulatory Commission to increase the amount of fuel that may be received and possessed by the GSTR.

Background

In accordance with 10 CFR 50.90, whenever a licensee desires to make changes to the technical specifications and/or license, an application for amendment must be filed.

The GSTR is currently licensed to receive, possess, and use up to 9 kilograms of contained uranium-235 enriched to less than 20 percent in the form of TRIGA-type reactor fuel. We are working to receive a shipment of 103 TRIGA fuel elements from the VTT Technical Research Centre of Finland FiR1 reactor, which are a mixture of 8 wt.% aluminum-clad, 8.5 wt.% stainless steel-clad, and 12 wt.% stainless steel-clad elements.

Discussion The GSTR requests a license amendment to the Facility Operating License No. R-113 allowing for the receipt and possession, but not use or separation, of 4 kilograms of U-235. The limit of this amount is prosed to cover any potential uncertainties in the total mass of U-235 in the inventory. None of the fuel acquired in this proposed transaction will be utilized until full analysis is completed and a follow-on license amendment request are approved. The proposed wording of the additional statement in section 2.B.2 of the R-113 operating license would be:

f. to receive, possess, but not use or separate, in connection with the operation of the facility, up to 4 kilograms of contained uranium-235 enriched to less than 20 percent in the form of TRIGA-type reactor fuel.

In consideration of the special nuclear material and byproduct material present in fuel and non-fuel research reactor parts being transferred to the GSTR from another facility, current statements in the R-113 license as written in sections 2.B.2.e, 2.B.3.e, and 2.B.3.g apply and provide sufficient coverage for the materials to be received and possessed.

Page 2 of 7 Sections of this analysis have been kept intentionally vague in line with previously observed redaction practices.

Fuel Description In characterization of the fuel and certification that it is of TRIGA design and manufacturing in agreement with the types of fuel currently licensed for use in the GSTR, the fuel has been reported by the source facility that it was manufactured in San Diego, California at the General Atomics plant. An inquiry is currently out to Mr. Anthony Veca of General Atomics to provide associated documentation for the serial numbers of the fuel elements in this potential acquisition, which will be provided as a supplement to this license amendment request when available, as permitted by security provisions. Additionally, Department of Energy has conducted two separate visits to the FiR1 facility and representation from the GSTR is planned to be present for final inspection and loading of the fuel from the FiR1 facility. The first DOE visit was conducted from April 7th, 2016 through May 4th, 2016 in order to perform a fuel examination of 103 irradiated TRIGA elements and 24 unirradiated TRIGA elements. One element was known previously to have been failed and was already stored in a facility-designed canister in dry storage before the inspection, leaving 102 elements to be fully examined and verified. A full examination report was generated with a TRIGA Fuel Element Examination Sheet, FRM-898, for every element inspected, detailing the surface condition, element serial number, cladding type, date and time of the inspection, configuration of the element (IFE, standard, etc.), and other pertinent information for the condition of the element. A total of ten elements were found to be failed with an additional eight to be re-examined at a later date. The second visit by DOE to the FiR1 facility was conducted from June 10th, 2019 to June 18th, 2019 to re-examine all failed elements and the additional eight of concern with the exception of the previously known failed element in the facility-designed can. All ten elements previously determined as failed were again considered as failed. The other eight elements of concern were cleared and found not to be failed.

In both DOE fuel inspections, 100% of the surface of the elements were examined and serial numbers verified using a video scope capable of taking close (macro level) pictures of surface flaws on the cladding of an element. This video scope provided the ability to perform actual depth, length, width, and height measurements of anomalies on the fuel element cladding, which such capability had not been previously available to examiners and provided for a much higher level of scrutiny of the fuel. As part of the acquisition contract of this fuel, GSTR personnel as well as DOE personnel, if opted and available, will be present for the final inspection and loading of the fuel from FiR1 prior to transport.

The breakdown of elements to be acquired in this transaction include:

67 Aluminum-clad 8.0 wt.% / < 20 enrichment % standard TRIGA fuel elements.

2 Aluminum-clad 8.0 wt.% / < 20 enrichment % instrumented TRIGA fuel elements.

7 Stainless steel-clad 8.5 wt.% / < 20 enrichment % standard TRIGA fuel elements.

1 Stainless steel-clad 8.5 wt.% / < 20 enrichment % instrumented TRIGA fuel element.

25 Stainless steel-clad 12 wt.% / < 20 enrichment % standard TRIGA fuel elements.

1 Stainless steel-clad 12 wt.% / < 20 enrichment % instrumented TRIGA fuel element.

Comprehensive burnup calculations utilizing both SERPENT and ORIGEN codes have also been performed and provided to the GSTR for review and acceptability determination of the fuel. The

Page 3 of 7 Fuel Element Type 8.0 wt. %

GSTR 8.0 wt.%

FiR1 8.5 wt. %

GSTR 8.5 wt.%

FiR1 12.0 wt. %

GSTR 12.0 wt.%

FiR1 Fuel-Moderator Material U-ZrH1.0 U-ZrH1.0 U-ZrH1.7 U-ZrH1.7 U-ZrH1.7 U-ZrH1.7 Uranium Content 8.0 wt. %

8.0 wt. %

8.5 wt. %

12.0 wt. %

235U Enrichment

< 20%

235U Content avg per Element Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Shape Cylindrical Cylindrical Cylindrical Cylindrical Cylindrical Cylindrical Length of Fuel Meat Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Diameter of Element Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Matches FiR1 Matches GSTR Cladding Material Type 1100 Al Type 1100 Al Type 304 SS Type 304 SS Type 304 SS Type 304 SS Cladding Thickness 0.03 inches 0.03 inches 0.020 inches 0.020 inches 0.020 inches 0.020 inches Table 1: Fuel Type Comparisons of GSTR and VTT FiR1 results for the elements involved in the proposed acquisition have relatively low burnup and utilization.

A comparison between current GSTR fuel element types permitted by the facility license and safety analysis report and those to be received from FiR1 in Table 1 shows that they are certifiably of TRIGA design and manufacture. The documentation provided of the DOE fuel inspections also corroborates this conclusion. A list of element serial numbers is provided as an attachment to this amendment request for verification purposes.

As stated before, ten elements have been characterized as failed and are not intended to be used at the GSTR. The elements in question are listed in Table 2, which will be permanently stored at the GSTR until such time that DOE removes them for ultimate disposal. All 10 elements listed in Table 2 will be packaged into sealed damaged fuel canisters (SDFC), two elements per SDFC, by the Nuclear Assurance Corporation prior to shipping in order to comply with shipping requirements and their NRC issued Certificate of Compliance (Number 9225, revision 69, docket 71-9225, package ID USA/9225/B(U)F-96). The SDFCs are documented in the NAC drawings 315-40-086, -087, and-088. The SDFC itself is 0.065-inch-thick Type 304 stainless steel, pressure tested to 50 PSIG, purged with helium gas, then bolted shut with a top equipped with a metallic O-ring. These canisters were initially identified by DOE for use in transport and storage of damaged elements at their long term storage facility INTEC and provide a high degree of confidence against releases of radioactive material from the damaged elements, which are only noted to have leaked during reactor operations and not in storage.

All four IFEs have had their conduit cut down such that they will fit into the top basket of the transfer cask or 48 long SDFCs. To compensate for the lack of handling, a pintle has been fitted to their stubbed conduit such that the standard TRIGA fuel handling tools may be used to couple completely with them.

Page 4 of 7 S/N Cladding Type Weight %

Date Removed from Service Details of failure 2070 Aluminum Standard 8.0 1994-12-31 Extreme bent/crease bottom weld.

2344 Aluminum IFE 8.0 1994-12-31 Previous leaker, gouged bottom weld.

2500 Aluminum Standard 8.0 1981-03-02 Previous leaker, facility-designed canned.

Not inspected.

2518 Aluminum Standard 8.0 1994-12-31 Bottom weld dent, slight bulge, inward crease, suspicious pit.

2520 Aluminum Standard 8.0 1994-12-31 Dent and bulge at bottom, may have been dropped. Apparent crack in cladding in the dented area.

2529 Aluminum Standard 8.0 1998-10-08 Top of element has raised and displaced metal, indicating cladding breach.

2541 Aluminum Standard 8.0 1985-01-08 Top section has swelling, possible dents, bulge.

2543 Aluminum Standard 8.0 2015-06-30 Deep pit/scratch on bottom section, gouge in top weld area, layered cladding, likely breach.

2550 Aluminum Standard 8.0 2015-06-30 Corrosion apparent near top weld, multiple elongated damage sites, scratches.

4312 Stainless Steel IFE 8.5 1987-06-03 Known leaker, apparent tear/crack in 7-8 120-240 deg section, 10-13 section all around damaged.

Table 2: Failed Elements for Permanent Storage, No Utilization Storage The fuel to be acquired from VTT FiR1 will all remain in storage until the necessary analyses are completed and a future license amendment is acquired to use the fuel.

The intact fuel will be stored in our hexagonal storage racks in the pool, as described in the SAR.

Additional racks are on hand for installation in the pool, as they were manufactured at the same time as the initial four listed in the SAR. They will be mounted in the reactor tank in a manner similar to the existing racks and will meet the technical specification requirements for fuel storage. The current and additional rack installations will be configured such that the distance from the top of the reactor tank to the top of the rack will be as currently installed, and that a vertical offset of approximately 15 inches between fuel element sections will be achieved. Fuel racks are currently horizontally spaced a minimum of approximately one rack width apart. This amount of vertical and horizontal offset between racks will achieve a separation of many mean free paths for thermal neutrons in water, 20 cm generally being considered an infinite reflector for reactivity calculations, effectively decoupling the fuel racks from one another. The existing linear racks will be removed in favor of utilizing additional hexagonal racks for space considerations.

The damaged fuel will be maintained in their SDFCs and stored in the facilitys storage pits until ultimate removal by the Department of Energy for final disposition. A storage rack based on the existing design is being modified to accommodate the sealed SDFCs. Each modified rack will have fewer elements than a standard rack, thereby further mitigating the risk of criticality for fuel in storage, providing a very large margin to the requirements in the technical specifications for fuel storage. The racks have been fully analyzed and documented in a publication available in the

Page 5 of 7 journal Applied Radiation and Isotopes, issue 119 (2017), pages 16-22, titled Benchmarking criticality analysis of TRIGA fuel storage racks. Additionally, shielding plugs are being designed and fabricated for installation on the tops of the storage pits that will contain the damaged elements. These plugs will fit below the deck plates covering the pits, preventing them from posing a tripping hazard to personnel. Each plug will rest on the squared section above the circular pit, preventing the possibility of the plug from falling into the pit and damaging the SDFCs and contained fuel elements. The design of the shield plug allows for up to four racks filled with fuel elements to be located in a storage pit, including one of the SDFC racks. While scenarios of one through four filled fuel racks in a storage pit were analyzed for, the only typical configurations would be the single SDFC rack in standard practice or all four racks as a result of a complete core and tank unload. The simulations were run using the worst-case scenario for radionuclide inventory for each fuel rod to be stored. In the four-rack configuration, dose rates with the designed shield plug were less than 0.7 mrem/hr on contact with the fuel pit covers. Enough reserve storage capacity is available to facilitate a complete core and tank unload. The storage pits are well ensured against the intrusion of ground water, as analyzed and described in the SAR, section 3.3. Drawings of the racks, shielding plugs, and dose rate analysis with assumptions are attached to this letter as part of the license amendment request.

The transfer baskets for the fuel transfer cask being used for this shipment will be checked for removable contamination, contained and sealed, and kept at the reactor facility in dry radioactive storage for later use in shipping, but not used to store fuel at the GSTR facility.

Security The fuel will be stored within the controlled access area identified by the GSTR physical security plan with the rest of the TRIGA fuel possessed by the facility. The security plan requires a set of minor changes related to the license amendment request, which have been determined to not constitute a decrease in the effectiveness of the plan. The changes will be executed prior to the receipt of the incoming fuel and documentation of the change with a copy of the revised physical security plan will be sent to the NRC under the provisions of 10 CFR 2.390.

Emergency Plan A review of the emergency plan was conducted against ANSI/ANS 15.16-2008 - Emergency Planning for Research Reactors, and no changes are required as a result of this license amendment request.

Figure 1 - Fuel Rack Install Configuration

Page 6 of 7 Radiation Protection Plan A review of the radiation protection plan was conducted against ANSI/ANS 15.11-2009 -

Radiation Protection at Research Reactor Facilities, and no changes are required as a result of this license amendment request.

Safety Analysis Report A review of the accident scenarios in the GSTR Safety Analysis report was conducted, no impact from the acquisition of this fuel has been identified, and no changes are required as a result of this license amendment.

Department of Energy Interagency Agreement with USGS The interagency agreement with the Department of Energy has been updated and signed by all parties to include the fuel from VTT FiR1 in the scope, identified by number of elements and grams of SNM, accounting for final disposition at the end of the GSTR facility life. This was completed on September 10th, 2020 and the main document is attached to this license amendment request, withholding the addendum describing the number of elements and SNM gram amounts.

This proposed change meets all other requirements of the license, 10 CFR, and NUREG-1537 guidance.

Action The USGS requests NRC review and approve this application for license amendment by November 15th, 2020, if possible, due to a timeline established by our collaborator VTT FiR1, and their previous agreement with Department of Energy.

Contact If you have any questions regarding this matter, please contact me at (303) 236-4727.

Affirmation I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Jonathan Wallick, Reactor Director Copied to:

Dr. Robert Horton, Reactor Administrator, USGS Geoffrey Wertz, Project Manager, US NRC Craig Bassett, Inspector, US NRC Attachments:

(1) Original License R-113 for U.S. Geological Survey TRIGA Reactor; (2) Proposed License R-113 for U.S. Geological Survey TRIGA Reactor; (3) List of FiR1 TRIGA Element Serial Numbers and Types to be Transferred; Digitally signed by JONATHAN WALLICK Date: 2020.10.01 10:47:04 -06'00'

Page 7 of 7 (4) Modified Storage Racks for Sealed Damaged Fuel Element Canisters Drawing; (5) Shield Plug for Storage Pits Drawing; (6) Dose Rate Analyses for Storage Pit Shield Plug, Multiple Configurations; (7) Interagency Agreement with Department of Energy 1012, Amendment 05 (w/o Addendum).

Original License R-113 for GSTR

U.S. GEOLOGICAL SURVEY DOCKET NO. 50-274 RENEWED FACILITY OPERATING LICENSE License No. R-113

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for renewal of Facility Operating License No. R-113 filed by the U.S. Geological Survey (the licensee), dated January 5, 2009, as supplemented on November 24, 2010; February 11, March 28, May 12, June 29, July 27, August 30, September 26, October 31, and November 30, 2011; January 3, January 27, March 28, April 27, May 18, May 31, June 29, July 31, August 30, and November 16, 2012; February 8, May 17, and October 31, 2013; February 19, November 3, and November 24, 2014; September 8, 2015; and January 22, April 1, September 12, and September 22, 2016, (the application),

complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10, of the Code of Federal Regulations (10 CFR), Chapter I; B.

Construction of the U.S. Geological Survey TRIGA Research Reactor (the facility) was completed in substantial conformity with the Construction Permit No. CPRR-102, issued on October 10, 1967, and the application, as amended; the provisions of the Act; and the rules and regulations of the Commission; C.

The facility will operate in conformity with the application, as supplemented, the provisions of the Act, and the rules and regulations of the Commission; D.

There is reasonable assurance that: (i) the activities authorized by this license can be conducted without endangering the health and safety of the public, and (ii) such activities will be conducted in compliance with the Commissions regulations; E.

The licensee is technically and financially qualified to engage in the activities authorized by this license in accordance with the rules and regulations of the Commission; F.

The applicable provisions of 10 CFR Part 140, Financial Protection Requirements and Indemnity Agreements, have been satisfied; G.

The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; H.

The issuance of this license is in accordance with 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions, of the Commissions regulations and all applicable requirements have been satisfied; and I.

The receipt, possession and use of byproduct and special nuclear materials as authorized by this facility operating license will be in accordance with the Commission's regulations in 10 CFR Part 30, Rules of General Applicability to Domestic Licensing of Byproduct Material, and 10 CFR Part 70, Domestic Licensing of Special Nuclear Material.

2.

Accordingly, Facility Operating License No. R-113 is hereby renewed in its entirety to read as follows:

A. This license applies to the U.S. Geological Survey TRIGA Research Reactor (herein the facility) owned by the U.S. Geological Survey, Department of the Interior, (herein the licensee). The facility is located on the U.S. Denver Federal Center in Lakewood, Colorado, and described in the licensees application for license renewal, dated January 5, 2009, as supplemented.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses the U.S. Geological Survey as follows:

1.

Pursuant to subsection 104c of the Act and 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to possess, use, and operate the facility as a utilization facility at the designated location in accordance with the procedures and limitations described in the application and set forth in this license.

2.

Pursuant to the Act and 10 CFR Part 70, the following activities are included:

a. to receive, possess, and use, but not separate, in connection with the operation of the facility, up to 9 kilograms of contained uranium-235 enriched to less than 20 percent in the form of TRIGA-type reactor fuel;

b. to receive, possess, and use, but not separate, in connection with the operation of the facility, up to 15 grams total of special nuclear material, of any enrichment, in the form of neutron detectors;

c. to receive, possess, and use, but not separate, in connection with the operation of the facility, up to 2 grams of special nuclear material of any enrichment in reactor-based experiments, in sources for calibration of radiation detectors, and reference sources for reactor based programs;

d. to receive, possess, and use, but not separate, in connection with the operation of the facility, such special nuclear material as may be produced by the operation of the facility.

e. to receive, possess, and use, but not separate, in connection with the operation of the facility, such special nuclear material as may be received in TRIGA-type fuel elements that are transferred to license R-113 after use in other reactor facilities.

3.

Pursuant to the Act and 10 CFR Part 30, the following activities are included:

a. to receive, possess, and use, in connection with the operation of the facility, up to 3 curies of sealed americium-beryllium in a single neutron source for reactor startup use;

b. to receive, possess, and use, in connection with the operation of the facility, up to 10 curies of sealed polonium-beryllium in a single neutron source for reactor startup use;

c. to receive, possess, and use, in connection with the operation of the facility, up to 10 millicuries of byproduct material (atomic numbers 1-88) that will be irradiated in the reactor after receipt;

d. to receive, possess, and use, in connection with the operation of the facility, up to 5 curies of byproduct material used in reactor-based experiments, in sources for calibration of radiation detectors, and reference sources for use in reactor-based analytic techniques;

e. to receive, possess, and use, in connection with the operation of the facility, up to 50 millicuries of byproduct material contained in (non-fuel) research reactor parts and components received for use under this facility operating license, No. R-113 from other research facilities;

f.

to receive, possess, and use, but not to separate, in connection with operation of the facility, such byproduct material as may be produced by operation of the reactor, except for byproduct material produced in non-fueled reactor experiments which is permitted to be separated; and

g. to receive, possess, and use, but not to separate, any amount of byproduct material contained in TRIGA fuel elements transferred to USGS Facility Operating License No. R-113 after use in other reactor facilities.

4.

Pursuant to the Act and Title 10 CFR, Chapter 1, Part 40, Domestic Licensing of Source Material, in connection with the operation of the facility, to receive, possess, and use, up to 1 millicurie of source material for reactor-based experiments, sources for calibration of detectors, and reference sources for use in reactor-based analytical techniques.

C.

This license shall be deemed to contain, and is subject to the conditions specified 10 CFR Parts 20, 30, 40, 50, 51, 55, 70, and 73 of the Commissions regulations; is subject to all provisions of the Act, and to the rules, regulations and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below:

1.

Maximum Power Level The licensee may operate the reactor at steady-state power levels not in excess of 1,000 kilowatts (thermal) and to pulse the reactor in accordance with the limitations in the Technical Specifications.

2.

Technical Specifications The Technical Specifications contained in Appendix A are hereby incorporated in their entirety in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

Physical Security Plan The licensee shall maintain and fully implement all provisions of the Commission-approved physical security plan, including changes made pursuant to the authority of 10 CFR 50.54(p). The approved physical security plan, entitled Security Plan for the DFC Reactor Facility, Revision XVII, dated August 2016, consists of documents withheld from public disclosure pursuant to 10 CFR 73.21.

This license is effective as of the date of issuance and shall expire at midnight, 20 years from the date of issuance.

For the Nuclear Regulatory Commission

/RA/

William M. Dean, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A, Technical Specifications Date of Issuance: October 14, 2016

Proposed License R-113 for GSTR U.S. GEOLOGICAL SURVEY DOCKET NO. 50-274 RENEWED FACILITY OPERATING LICENSE License No. R-113 1.