ML25083A296

Replacement File for Submittal of Approved Terrapower, LLC Fuel and Control Assembly Qualification Topical Report, NAT-2806-A, Revision 0

ML25083A296
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	03/24/2025
From:	Ian Gifford TerraPower
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
TP-LIC-LET-0409 NAT-2806-A, Rev. 0
Download: ML25083A296 (1)

Text

15800 Northup Way, Bellevue, WA 98008 www.TerraPower.com P. +1 (425) 324-2888 F. +1 (425) 324-2889 March 24, 2025 TP-LIC-LET-0409 Docket Number 50-613 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk

Subject:

Replacement File for Submittal of Approved TerraPower, LLC Fuel and Control Assembly Qualification Topical Report

Reference:

1. U.S. Nuclear Regulatory Commission, TerraPower, LLC - Final Safety Evaluation of Topical Report NATD-FQL-PLAN-0004, Fuel and Control Assembly Qualification, Revision 0 (ML24220A156)

2. Submittal of Approved TerraPower, LLC Fuel and Control Assembly Qualification Topical Report, December 19, 2024 (Accession No. ML24354A192)

The U.S. Nuclear Regulatory Commission (NRC) provided the final safety evaluation for the TerraPower, LLC (TerraPower) Fuel and Control Assembly Qualification Topical Report in Reference 1. TerraPower provided the accepted version of the Topical Report, NAT-2806-A, in Reference 2.

This letter provides a replacement file to be used in place of Reference 2. Please replace the file currently associated with ML24354A192, submitted on December 19, 2024, with the version contained in Enclosure 1.

This letter and the associated enclosures make no new or revised regulatory commitments.

If you have any questions regarding this submittal, please contact Ian Gifford at igifford@terrapower.com.

Date: March 24, 2025 Page 2 of 2 Sincerely, Ian Gifford Director of Licensing TerraPower, LLC

Enclosures:

1. Replacement File for Submittal of Approved TerraPower, LLC Fuel and Control Assembly Qualification Topical Report, December 19, 2024 (Accession No. ML24354A192) cc:

Mallecia Sutton, NRC Josh Borromeo, NRC

ENCLOSURE 1 Replacement File for Submittal of Approved TerraPower, LLC Fuel and Control Assembly Qualification Topical Report, December 19, 2024 (Accession No. ML24354A192)

TP-LIC-LET-0364 Project Number 99902100 December 19, 2024 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk

Subject:

Submittal of Approved TerraPower, LLC Fuel and Control Assembly Qualification Topical Report

References:

1.

U.S. Nuclear Regulatory Commission, TerraPower, LLC - Final Safety Evaluation of Topical Report NATD-FQL-PLAN-0004, Fuel and Control Assembly Qualification, Revision 0 (ML24220A156) 2.

TerraPower, LLC, Submittal of TerraPower Fuel and Control Assembly Qualification Topical Report, NATD-FQL-PLAN-0004, Revision 0, January 25, 2023 (ML23025A409) 3.

TerraPower, LLC, Correction to TerraPower Fuel and Control Assembly Qualification Topical Report, NAT-2806, Revision 0, April 18, 2023 (ML23109A099)

The U.S. Nuclear Regulatory Commission (NRC) provided the final safety evaluation for the TerraPower, LLC (TerraPower) Fuel and Control Assembly Qualification Topical Report in Reference 1.

The topical report describes TerraPowers process to obtain qualified fuel and control assemblies for the Natrium Plant.1 Enclosures 2 and 3 to this letter provide the accepted version of the topical report with the additional content incorporated per the NRC staff request, designated NAT-2806-A.

The topical report was submitted to the NRC as document number NATD-FQL-PLAN-0004 on January 25, 2023, in Reference 2. A correction to the initial submittal was submitted to the NRC on April 18, 2023, in Reference 3. The document number was changed from NATD-FQL-PLAN-0004 to NAT-2806, as noted in the final safety evaluation. Therefore, the accepted versions of the topical report provided in Enclosures 2 and 3 to this letter are denoted as NAT-2806-A.

1 Natrium is a TerraPower and GE-Hitachi technology.

Date: December 19, 2024 Page 2 of 2 The report contains proprietary information and as such, it is requested that Enclosure 3 be withheld from public disclosure in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding. An affidavit certifying the basis for the request to withhold Enclosure 3 from public disclosure is included as Enclosure 1. Enclosure 3 also contains ECI which can be disclosed to Foreign Nationals only in accordance with the requirements of 15 CFR 730 and 10 CFR 810, as applicable.

Proprietary and ECI materials have been redacted from the report provided in Enclosure 2; redacted information is identified using (( ))(a)(4), (( ))ECI, or (( ))(a)(4), ECI.

This letter and the associated enclosures make no new or revised regulatory commitments.

If you have any questions regarding this submittal, please contact Ian Gifford or Nick Kellenberger at igifford@terrapower.com or nkellenberger@terrapower.com, respectively.

Sincerely, George Wilson Vice President, Regulatory Affairs TerraPower, LLC

Enclosures:

1. TerraPower, LLC Affidavit and Request for Withholding from Public Disclosure (10 CFR 2.390(a)(4))

2. TerraPower, LLC Topical Report Fuel and Control Assembly Qualification, NAT-2806-A, Revision 0, Non-Proprietary (Public)

3. TerraPower, LLC Topical Report Fuel and Control Assembly Qualification, NAT-2806-A, Revision 0, Proprietary and Export-Controlled (Non-Public) cc:

Mallecia Sutton, NRC Josh Borromeo, NRC Nathan Howard, DOE Jeff Ciocco, DOE

ENCLOSURE 1 TerraPower, LLC Affidavit and Request for Withholding from Public Disclosure (10 CFR 2.390(a)(4))

TerraPower, LLC Affidavit and Request for Withholding from Public Disclosure (10 CFR 2.390(a)(4))

I, George Wilson, hereby state:

1. I am the Vice President, Regulatory Affairs and I have been authorized by TerraPower, LLC (TerraPower) to review information sought to be withheld from public disclosure in connection with the development, testing, licensing, and deployment of the Natrium reactor and its associated fuel, structures, systems, and components, and to apply for its withholding from public disclosure on behalf of TerraPower.

2. The information sought to be withheld, in its entirety, is contained in Enclosure 3, which accompanies this Affidavit.

3. I am making this request for withholding, and executing this Affidavit as required by 10 CFR 2.390(b)(1).

4. I have personal knowledge of the criteria and procedures utilized by TerraPower in designating information as a trade secret, privileged, or as confidential commercial or financial information that would be protected from public disclosure under 10 CFR 2.390(a)(4).

5. The information contained in Enclosure 3 accompanying this Affidavit contains non-public details of the TerraPower regulatory and developmental strategies intended to support NRC staff review.

6. Pursuant to 10 CFR 2.390(b)(4), the following is furnished for consideration by the Commission in determining whether the information in Enclosure 3 should be withheld:

a. The information has been held in confidence by TerraPower.

b. The information is of a type customarily held in confidence by TerraPower and not customarily disclosed to the public. TerraPower has a rational basis for determining the types of information that it customarily holds in confidence and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.

The application and substance of that system constitute TerraPower policy and provide the rational basis required.

c. The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR 2.390, it is received in confidence by the Commission.

d. This information is not available in public sources.

e. TerraPower asserts that public disclosure of this non-public information is likely to cause substantial harm to the competitive position of TerraPower, because it would enhance the ability of competitors to provide similar products and services by reducing their expenditure of resources using similar project methods, equipment, testing approach, contractors, or licensing approaches.

I declare under penalty of perjury that the foregoing is true and correct. Executed on: December 19, 2024 George Wilson Vice President, Regulatory Affairs TerraPower, LLC

ENCLOSURE 2 TerraPower, LLC Topical Report Fuel and Control Assembly Qualification, NAT-2806 -A, Revision 0 Non-Proprietary (Public)

TerraPower,LLC 15800NorthupWay Bellevue,WA98008 ATerraPower&GEHitachiTechnology Fuel and Control Assembly Qualification NAT-2806-A Revision0 December 19, 2024 SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2024 TerraPower, LLC. All rights reserved.

Mr. George Wilson Vice President, Regulatory Affairs TerraPower, LLC 15800 Northup Way Bellevue, WA 98008

SUBJECT:

TERRAPOWER, LLC - FINAL SAFETY EVALUATION OF TOPICAL REPORT NATD-FQL-PLAN-0004, FUEL AND CONTROL ASSESSMBLY QUALIFICATION, REVISION 0 (EPID L-2023-TOP-0017)

Dear Mr. Wilson:

By letter dated January 25, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23025A409), TerraPower, LLC (TerraPower) submitted a Topical Report (TR) entitled Fuel and Control Assembly Qualification, Revision 0, for the U.S. Nuclear Regulatory Commission (NRC) staffs review. By letter dated April 18, 2023, TerraPower submitted a supplement to the TR (ML23109A099). The TR identifies acceptance criteria for fuel qualification and presents select fuel qualification results in addition to ongoing and planned fuel qualification activities. The TR also summarizes a notional fuel surveillance plan for the collection of data to address certain targeted gaps and help develop new fuel designs. The TR was submitted in support of TerraPowers planned license application under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities. March 28, 2024 (ML24088A059) TerraPower submitted a construction permit (CP) on behalf of US SFR Owner, LLC (USO), a wholly owned subsidiary of TerraPower.

By email dated March 31, 2023, the NRC staff informed TerraPower that the TR provided sufficient information for the NRC staff to begin its detailed technical review (ML23086C087).

On June 7, 2023, the NRC staff transmitted an audit plan to TerraPower (ML23156A553), and subsequently conducted an audit of materials related to the TR from June 21, 2023, to October 5, 2023. The NRC staff issued a summary of the audit on February 6, 2024 (ML24043A155). On February 28, 2024, to support an Advisory Committee on Reactor Safeguards meeting, the NRC staff provided the unredacted version of the draft safety evaluation (SE) for the TR to TerraPower (ML23326A184).

The enclosed final SE is being provided to TerraPower because the NRC staff has found NATD-FQL-PLAN-0004, Revision 0, acceptable for referencing in licensing actions to the extent specified and under the limitations and conditions delineated in the SE. The final SE defines the basis for the NRC staffs acceptance of the TR.

The NRC staff requests that TerraPower publish an approved version of this TR within 3 months of receipt of this letter. The approved version should incorporate this letter and the enclosed SE after the title page. The approved version should include a -A (designating approved) following the TR identification.

October 15, 2024

G. Wilson If you have any questions, please contact Mallecia Sutton at (301) 415-0673 or via email at Mallecia.Sutton@nrc.gov.

Sincerely, Joshua Borromeo, Chief Advanced Reactor Licensing Branch 1 Division of Advanced Reactors and Non-Power Production and Utilization Facilities Office of Nuclear Reactor Regulation Project No.: 99902100

Enclosure:

As stated cc: TerraPower Natrium via GovDelivery Signed by Borromeo, Joshua on 10/15/24

Package: ML24220A156 Letter: ML24220A155 Safety Evaluation: ML24220A154 NRR-043 OFFICE NRR/DANU/UAL1:PM NRR/DANU/UTB2 NRR/DANU/UAL1:LA NAME MSutton RAnzalone DGreene DATE 8/8/2024 8/13/2024 8/14/2024 OFFICE OGC NRR/DANU/UTB2:BC NRR/DANU/UAL1:BC NAME JEzell CdeMessieres JBorromeo DATE 9/26/2024 10/7/2024 10/15/2024

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION TERRAPOWER, LLC - FINAL SAFETY EVALUATION OF TOPICAL REPORT NATD-FQL-PLAN-0004, FUEL AND CONTROL ASSEMBLY QUALIFICATION, REVISION 0 (EPID L-2023-TOP-0017)

SPONSOR AND SUBMITTAL INFORMATION Sponsor:

TerraPower, LLC Sponsor Address:

15800 Northup Way Bellevue, WA 98008 Project No.:

99902100 Submittal Date:

January 25, 2023 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML23025A409 Supplement ADAMS Accession No.: ML23109A099 Brief Description of the Topical Report: NATD-FQL-PLAN-0004, Fuel and Control Assembly Qualification, Revision 0 [1] provides TerraPower, LLCs (TerraPowers) plan to qualify fuel and control assemblies for the Natrium sodium fast reactor (SFR). The topical report (TR) identifies acceptance criteria for fuel qualification and presents select fuel qualification results in addition to ongoing and planned fuel qualification activities. The TR also summarizes a notional fuel surveillance plan for the collection of data to address certain targeted gaps and help develop new fuel designs. The qualification plan is applicable to Natrium Type 1 fuel, a uranium-10 weight percent zirconium (U-10Zr) alloy fuel clad in HT9 steel, and control assemblies using boron carbide as a neutron absorber.

Following the initial TR submittal, TerraPower identified minor errors and submitted corrections [2]. The errata noted some minor changes to numbers in Table 6-7, Type 1 Fuel Assembly Design Parameters, of the TR and stated that the report would be re-numbered because TerraPowers report numbering scheme had changed.

REGULATORY EVALUATION The TR was submitted in support of TerraPowers license application under Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities. [3]

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION The U.S. Nuclear Regulatory Commission (NRC) regulations under 10 CFR Part 50 applicable to the Natrium reactors fuel and control assemblies include:

10 CFR 50.43(e), which states that reactor designs that differ significantly from light-water reactor (LWR) designs licensed before 1997 will be approved only if there has been appropriate demonstration of their safety features. In particular, 10 CFR 50.43(e)(1)(i) and (ii) require demonstration of safety feature performance and interdependent effects through analysis, appropriate test programs, experience, or a combination thereof; and 10 CFR 50.43(e)(1)(iii) requires that sufficient data exist on the safety features to assess the analytical tools for safety analysis over a sufficient range of plant conditions.

10 CFR 50.34(a)(1)(ii)(D), which requires construction permit (CP) applicants to evaluate a postulated fission product release from the core into the containment.

10 CFR 50.34(a)(3)(i), which requires CP applicants to submit principal design criteria (PDCs) for the facility. The regulation notes that the general design criteria (GDCs) included in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, provide guidance to CP applicants for establishing PDCs for facilities different in design and location to plants for which CPs have previously been issued by the NRC.

TerraPower submitted a TR concerning PDCs for the Natrium design [4].

The guidance considered by the NRC staff in its review of the TR included the following:

Guidance on fuel qualification for non-LWRs is provided in NUREG-2246, Fuel Qualification for Advanced Reactors [5]. This NUREG builds on guidance for LWRs contained in NUREG-0800, Section 4.2, Fuel System Design, Revision 3 [6]. Concepts in these guidance documents are also broadly applicable to the qualification of control assemblies.

Additional relevant technical information is available in NUREG/CR-7305, Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, Fuel Qualification for Advanced Reactors, [7] which is a generic assessment of uranium-zirconium alloy metallic fuel following the NUREG-2246 process. While NUREG/CR-7305 assumes certain characteristics of the fuel for the purposes of its assessment, most of the discussion in that report is relevant to the Natrium Type 1 fuel.

Finally, TerraPower is following the Licensing Modernization Project (LMP) design and licensing approach outlined in Nuclear Energy Institute (NEI) 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development [8].

The LMP approach was endorsed by the NRC staff in Regulatory Guide (RG) 1.233, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors [9]. This guidance defines risk-informed, performance based, and technology-inclusive processes for the selection of licensing basis events; safety classification of structures, systems, and components (SSCs); and determination of defense-in-depth adequacy for non-LWRs. NEI 18-04 provides a frequency-consequence target curve that is used to assess events, SSCs, and programmatic controls. Because the fuel is the primary source of radionuclides in the reactor, its performance under both normal and off-normal conditions plays a key role in evaluating consequences for the NEI 18-04 process.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION TECHNICAL EVALUATION

1.0 INTRODUCTION

TerraPower stated in TR Section 1, Purpose, that the objective of the TR is to confirm that all aspects of the fuel system design and fabrication process will provide reliable and safe operation of a commercial sodium-cooled, fast-neutron spectrum nuclear reactor.1 The TR further requests specific NRC review and approval that:

the identified acceptance criteria are adequate to support fuel qualification.

the identified key fuel manufacturing parameters are adequate to support fuel qualification.

the identified evaluation methods and models are adequate to support fuel qualification.

the use of legacy data and the planned testing is adequate to provide the necessary information to qualify the fuel.

the plans for inclusion of small subsets of fuel pins that operate outside the performance envelope of the bulk of the core, or that feature advanced design features, are acceptable.

The NRC staff considered TerraPowers overall objective and specific requests in the preparation of this safety evaluation (SE). In the executive summary of the TR, TerraPower presented the TR as a plan to qualify fuel and control assemblies to support operation of the Natrium Reactor containing TerraPowers fuel qualification results to date as well as plans for future fuel qualification activities. Knowing that fuel qualification has not yet been completed, the NRC staffs SE focuses on the acceptability of the qualification plan, the acceptability of the qualification activities carried out thus far, and the specific items requested in the TR. Limitation and Condition (L&C) 1 highlights the scope of this TR as a fuel qualification plan that does not in and of itself demonstrate that the fuel is qualified and the future work that must be done to qualify Natrium Type 1 fuel.

To the extent possible, this SE assesses TerraPowers fuel and control assembly qualification plan against the fuel qualification assessment framework (FQAF) criteria provided in NUREG-2246. The FQAF provides a set of goals that, when satisfied, provide reasonable assurance that the fuel, fabricated in accordance with its specification, will perform as described in the safety analysis (i.e., the fuel is qualified for use). The FQAF is constructed using a top-down approach, in which the primary objective is stated as a goal. This goal is composed of subgoals which, if met, demonstrate that the higher-level goal is satisfied. Subgoals may be further composed of lower level subgoals, and so on, until the supporting subgoals are at the level where they can be directly verified by evidence. The NRC staff notes that NUREG-2246 is not a requirement but provides an acceptable means for an applicant or licensee to demonstrate that fuel has been appropriately qualified.

As discussed in the TR, TerraPowers strategy for fuel qualification is centered around the development of regulatory acceptance criteria (RAC), which represent acceptance criteria derived from regulatory requirements. These RAC were identified by TerraPower based on a review of relevant regulations and guidance prior to the issuance of NUREG-2246. However, 1 Because the TR also covers control assemblies, the NRC staff assumes that a similar objective could also be applied to aspects of control assembly design.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION TerraPower identified in Section 1 of the TR that the fuel qualification plan was informed by NUREG-2246, and that Section 2.2, Regulatory Background, of the TR provides a mapping of the RAC discussed in the TR to the NUREG-2246 FQAF goals (Table 2-1, TerraPower Identified/Developed RAC Mapped to NUREG-2246 Appendix A Goals in FQAF 14) and to the NUREG-2246 evaluation model assessment framework goals (Table 2-2, TerraPower Identified/Developed RAC Mapped to NUREG-2246 Evaluation Model Assessment Framework Goals).

2.0 BACKGROUND

2.1 Natrium Type 1 fuel design Section 5.2, Fuel Assemblies, of the TR presents details on the Natrium Type 1 fuel design.

The overall design is conceptually very similar to fuel operated in previous United States SFRs, particularly the metallic fuel irradiated in Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF).2 The fuel consists of U-10Zr slugs inside HT9 fuel pins. The pins are wrapped with an HT9 wire to provide spacing and are assembled into a hexagonal bundle, which is inserted into an HT9 duct. The final assembly has an inlet nozzle that interfaces with the core support structure and a handling socket that allows the assembly to interface with fuel handling equipment. Additional details from the TR are provided in the sections that follow.

2.1.1 Fuel pin The metallic fuel is composed of U-10Zr. The enrichment of the uranium in the Natrium Type 1 fuel varies depending on the fuels intended location in the core, but peak enrichment is < 20%

uranium-235. The fuel is formed into right cylindrical slugs via injection casting in quartz molds.

Following manufacturing and inspection, the fuel slugs are inserted into cladding tubes composed of HT9, a ferritic-martensitic steel. The cross-sectional area of the slugs is approximately 75% of the internal cross-sectional area of the cladding. A liquid metallic sodium bond is used to fill the space between the fuel and the cladding; this is necessary to improve heat transfer, particularly at beginning of life before the fuel swells to the point where it contacts the cladding. Below the fuel column within the pin, there is an axial shield section composed of

((

)), while above the fuel column there is a fission gas plenum initially backfilled with inert gas. ((

)). The pin is sealed with end caps attached to the cladding using resistance pressure welding.

The fuel pins are then wrapped in HT9 wire to provide spacing between the pins when they are loaded into a fuel assembly. The wire is fixed at each end of the fuel pin by ((

)).

2 Though the FFTF was primarily an oxide-fueled reactor that was used to irradiate a series of metallic fuel assemblies, referred to in the literature as MFF (which is not a clearly defined acronym, according to the TR), towards the end of its operating life. These assemblies were part of a metallic fuel qualification campaign that was terminated when FFTF was shut down in 1994.

2.1.2 Fuel assembly Once the HT9 wire wrap is completed, the fuel pins are arranged into a tight triangular pitch and attached together into a hexagonal bundle. This fuel pin bundle is then inserted into the fuel assembly. The fuel assembly consists of an inlet nozzle, hexagonal duct, handling socket, and the fuel pin bundle. ((

)).

The handling socket ((

)), and is attached to the top of the duct. The duct is the principal structural member of the fuel assembly and provides vertical support during fuel handling operations and horizontal support during power operations. The inlet nozzle is attached to the bottom of the duct and is inserted into the core support structure, supporting the fuel assembly vertically during power operations. The inlet nozzle ((

)).

2.1.3 Core restraint system The handling socket functions as the top-core load pad and the duct includes an above-core load pad. These load pads are important components of the core restraint system, which positions the fuel assemblies horizontally within the core while allowing for limited bowing due to neutronic and thermal effects. Geometric distortion of the fuel has a significant effect on core reactivity due to changes in neutron leakage; this effect contributes to the inherent reactivity feedback characteristics of the fuel.

2.1.4 Lead demonstration assemblies and lead test assemblies TerraPower states in TR Section 9, "Fuel Surveillance," that it intends to include lead demonstration and lead test assemblies (LOAs and LT As, respectively) in the Natrium reactor core beginning with the first cycle of operation. LDAs and LTAs are discussed in Section 5.2.3, "Lead Demonstration Assembly," and Section 5.2.4, "Lead Test Assemblies/ Type 1 B Fuel," of the TR, respectively, and the fuel surveillance program, which is supported by the LDAs, is discussed in Section 9, "Fuel Surveillance," of the TR.

LDAs are intended to ((

)). The LDAs will be as similar as possible to standard Natrium Type 1 fuel assemblies, but will include up to ((

)) removable pins that will be examined following irradiation to confirm fuel performance. These removable pins will be very similar to the standard fuel pins, with the following exceptions: the length of the pins will extend beyond the top of the standard pins with a feature on the upper end cap to enable them to be removed with a pin removal tool, ((

)).

L TAs are intended to test new fuel design features in support of the eventual qualification and use of future Natrium fuels. The overall design concept for the L TA is similar to that of the LOA; however, the L TA uses a different fuel duct material and a different fuel pin design. The L TA fuel duct is composed of ((

)). The L TA fuel pin ((

)).

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OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION 2.2 Natrium control assemblies Section 5.3, Control Assemblies, of the TR provides details on the control assembly design. As noted in the TR, the main function of the control assemblies is to position neutron absorber material to appropriately control and terminate the nuclear chain reaction. This absorber material is composed of boron carbide which, like the fuel rods, is inserted into HT9 absorber pins. Also like the fuel pins, the absorber pins are wrapped with HT9 wire for spacing and packed into a hexagonal bundle. The bundle is assembled into what TerraPower refers to as a control rod assembly. This assembly consists of the pin bundle, upper and lower guide plates, a coupling head that connects to the control rod drive, and an HT9 control rod duct that surrounds the absorber rods. This whole control rod assembly moves up and down within the control assembly duct, which occupies its own space within the core. The control assembly duct, load pads, and handling socket are identical to the fuel assembly duct, though the inlet nozzle is slightly different. At the bottom of the control assembly duct, sitting on top of the inlet nozzle, is an axial shield block. During a reactor scram, the control rod drive disconnects at the coupling head and the control rod assembly drops into the core via gravity.

TerraPower is also developing a secondary control rod assembly that features some design differences relative to the primary control rod assembly to ensure a diverse means to control and terminate the nuclear chain reaction. The main differences between the primary and secondary assemblies are that ECI ((

)). The intent of these changes is to reduce the likelihood of mechanical binding of the control rod assembly within the control assembly duct for the secondary control rod assembly. Mechanical binding is considered to be the main mechanism that could inhibit control rod insertion.

3.0 FUEL QUALIFICATION PLAN EVALUATION USING THE NUREG-2246 FUEL QUALIFICATION ASSESSMENT FRAMEWORK 3.1 Manufacturing specifications FQAF Goal 1 (G1) states that licensing documentation should include sufficient information to ensure the control of key parameters affecting fuel performance during the manufacturing process. This goal is comprised of three supporting goals: that key dimensions and tolerances of fuel components are specified (G1.1); that key constituents are specified with allowance for impurities (G1.2); and that end state attributes for materials within the fuel component are specified or otherwise justified (G1.3).

Table 2-1 of the TR indicates that G1, including G1.1, G1.2, and G1.3, is addressed by RAC 4.2-5. The TR does not state the acceptance criterion associated with RAC 4.2-5. However, a description of all RAC was previously submitted to the NRC staff by TerraPower in a white paper (WP) entitled Advanced Fuel Qualification Methodology Report [10]. In this document, RAC 4.2-5 states that the fuel system description and design drawings shall provide information necessary to verify that the fuel system design bases are met. The NRC staff evaluated the information provided in the TR in this context to ensure that it is consistent with G1 and its supporting subgoals, as discussed below.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION 3.1.1 Key dimensions and tolerances FQAF G1.1 states that key dimensions and tolerances of fuel components that affect performance should be specified. Table 2-1 of the TR indicates that this goal is fulfilled by providing references to design drawings that specify relevant key dimensions and tolerances of fuel components, as identified by RAC 4.2-5. The applicable specifications and the primary sources for this information are summarized in Section 5.6, Verification of the Fuel System Design Basis, of the TR, particularly in Table 5-3, Summary of Completed and Planned Activities to Satisfy Fuel System Design Description Requirements (RAC 4.2-5).

The NRC staff reviewed Table 5-3 and determined that the documents referenced are expected to cover all key dimensions and tolerances of the fuel, including dimensions for the fuel, cladding, and assembly components. Additionally, the NRC staff verified during the audit that the documents contain the appropriate information (ML24043A155). Therefore, the NRC staff concluded that the TR provides an acceptable approach to demonstrate that G1.1 is satisfied.

3.1.2 Key constituents FQAF G1.2 states that key constituents should be specified with allowance for impurities. Table 2-1 of the TR indicates that relevant key constituents with allowance for impurities, as identified by RAC 4.2-5, are specified in design drawings. The applicable specifications and the primary sources for this information are summarized in Section 5.6 of the TR, particularly in Table 5-3.

Section 5.5.1, HT9, of the TR provides details on the HT9 alloy, which is used for various components in the fuel assembly, including the cladding. Section 5.5.1.1, Composition, provides details on the material composition, particularly in Table 5-1, Nominal Composition of HT9 Steel. Table 5-3 references a document that provides information on the type and metallurgical state of the cladding. The NRC staff determined that the information referenced in the TR appropriately specifies the key constituents of HT9. Section 5.5.2, U-10Zr Fuel, of the TR provides a brief overview of the fuel slug materials and manufacturing process. Table 5-3 additionally references a document containing information on the slug alloy composition for metallic fuel and allowable slug impurities, which the NRC staff reviewed during the audit. The NRC staff determined that the information referenced in the TR appropriately specifies the key constituents of the fuel slug.

Section 5.5.3, Other Core Materials, refers to the potential use of other materials for some components of the fuel system, including Type 304 and Type 316 stainless steel and Inconel 718. The TR does not specifically state where these materials would be used, except for ((

)). As discussed in the TR, these materials are commonly used in the nuclear industry. TerraPower stated in the TR that design inputs, such as material performance and properties, would be considered pre-qualified for these materials if they are obtained from NRC-accepted standards.

The NRC staff determined that the TR provides an acceptable approach for demonstrating that G1.2 is satisfied, particularly for HT9 and U-10Zr, because key material constituents and impurities are appropriately specified in the documentation referenced in the TR. If other materials are used in the fuel system, TerraPower should demonstrate that these materials are manufactured according to standard specifications and used consistent with their qualification under relevant NRC-accepted codes and standards, or otherwise appropriately justified. This is documented as L&C 2, below.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION 3.1.3 End-state attributes FQAF G1.3 states that end-state attributes for materials within the fuel component should be specified or otherwise justified, particularly to the extent that these end-state attributes influence key properties of the fuel during operation. NUREG/CR-7305 Section 2.1.3, End-State Attributes, provides some end-state attributes considered to be important for a U-Zr/HT9 fuel system, as they relate to the fuel slugs, sodium bond, and the cladding. These attributes can be summarized as:

the fuel should be manufactured using an injection molding process with controls on the formation of oxides and the fuel density the sodium bond needs to have limited voids and the height of the sodium bond in the plenum must be appropriate the cladding plenum needs to be appropriately sized and the welds need to meet certain criteria Some details on the manufacturing process for HT9 are provided in TR Section 5.5.1.2, Manufacturing Process, while the fuel slug manufacturing process is discussed in Section 5.5.2.1, Manufacturing Process, of the TR. Additionally, Table 5-3 of the TR provides references to documents that provide further details on the fuel cladding and slug specifications, including manufacturing processes. Based on the information provided in the TR and referenced in Table 5-3, which the NRC staff verified during the audit, the NRC staff found that the desired end-state attributes provided for Natrium Type 1 fuel are consistent with (i.e., they meet or exceed) the key desired end-state attributes discussed. Thus, the NRC staff determined that the TR provides an acceptable approach to demonstrate that G1.3 is satisfied.

3.2 Safety criteria FQAF Goal 2 pertains to safety criteria that support evaluation of the fuels safety performance.

This goal is comprised of three subgoals in the following areas: design limits during normal operations and anticipated operational occurrences (AOOs) (G2.1); radionuclide release limits under accident conditions (G2.2); and safe shutdown (G3.3).

In addressing fuel safety performance safety criteria for Natrium, the NRC staff notes that TerraPowers proposed PDCs [4] include specified acceptable system radionuclide release design limits (SARRDLs) and a functional containment, rather than the specified acceptable fuel design limits (SAFDLs) included in the GDCs. As discussed in RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors [12], the SARRDLs are not a generically acceptable replacement for the SAFDLs but are instead a necessary complement to the use of a functional containment approach. The SARRDLs limit the amount of radionuclide inventory released by the system (i.e., in this case, the fuel, primary coolant system, and all unisolated connected systems that may contribute to dose, such as the cover gas and primary sodium purification systems).

Section 1 of TerraPowers TR states that one key objective of the fuel design criteria and associated limits is to ensure that the fuel system is not damaged as a result of normal operations and AOOs. The NRC staff determined that the fuel design criteria and associated limits are appropriate as supporting criteria for the SARRDLs because, if the fuel system is not

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION damaged from normal operations and AOOs, a substantial increase in circulating radionuclides from normal operations and AOOs is not expected. Therefore, SARRDLs could be satisfied with respect to releases from the fuel. The NRC staffs evaluation of the fuel design criteria for normal operations and AOOs is included below in SE Section 3.2.1.

However, fuel performance criteria are necessary but not sufficient to evaluate the SARRDLs.

Based on operating experience, while random fuel pin failures from various mechanisms (e.g.,

manufacturing defects, cladding fretting) are expected to be rare, they are not completely preventable so some radionuclide release from the fuel must be assumed and accounted for in the evaluation of the SARRDLs. The relationship between the SARRDLs and fuel design limits is discussed in L&C 3.

3.2.1 Design limits during normal operations and anticipated operational occurrences FQAF G2.1 states that margin to design limits should be demonstrated under conditions of normal operation, including the effects of AOOs. This goal is further divided into two subgoals, which are discussed below.

3.2.1.1 Fuel performance envelope for normal operations and AOOs FQAF G2.1.1 states that the fuel performance envelope for normal operations and AOOs should be defined. The fuel performance envelope defines the conditions under which the fuel is expected to perform and informs various aspects of a plants licensing basis, including the safety analysis, technical specifications, and operating limits. As discussed in NUREG-2246, the fuel performance envelope is typically provided in terms of parameters such as temperature, power, and exposure.

To demonstrate that G2.1.1 is satisfied, Table 2-1 of the TR refers to RAC that TerraPower identified for a number of mechanisms. The intent of these acceptance criteria is to provide an envelope in which it can be demonstrated that the fuel system is not damaged as a result of normal operations and AOOs, consistent with FQAF G2.1.1. While damage was not clearly defined in the TR, TerraPower provided the following definition in [10] for RAC 4.2-1: Fuel system damage means that fuel system dimensions are outside operational tolerances or the functional capabilities of the fuel system are reduced below those assumed in the safety analysis. This is a conservative definition of damage below which it would be reasonable to assume that the fuel would not fail, but beyond which additional evaluation would be needed to ensure that the fuel is still capable of fulfilling its functional requirements and the safety analysis remains valid. TerraPower provided damage criteria for both the fuel pin and for the whole fuel assembly, which are discussed in more detail in the following sections.

Fuel Pin Damage Criteria Specific RAC for fuel system damage are provided in TR Table 4-1, Design Criteria to Prevent Fuel System Damage. The criteria applicable to fuel pin damage are summarized as follows:

RAC 4.2-1.1 - stress, strain, and loading RAC 4.2-1.2 - fatigue RAC 4.2-1.3 - fretting wear RAC 4.2-1.4 - erosion and corrosion

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION RAC 4.2-1.5 - cladding internal damage due to fuel-cladding chemical interaction (FCCI)

RAC 4.2-1.6 - dimensional changes, such as bowing or swelling RAC 4.2-1.8 - fuel pin internal pressure RAC 4.2-1.12 - fuel and cladding temperatures The NRC staff notes that these criteria do not in and of themselves establish a fuel performance envelope as envisioned by NUREG-2246. In comparison, Section 2.3, Fuel Performance Envelope, of NUREG/CR-7305 provides a proposed steady-state envelope for U-10Zr/HT9 fuel meeting certain geometric considerations approximately equivalent to Natrium Type 1 fuel and the metallic fuel operated at EBR-II and FFTF. This envelope includes a peak fuel rod burnup of 10%; a beginning of life peak linear heat generation rate of 40 to 55 kilowatts per meter (kW/m);

a peak steady-state fuel cladding temperature of 650 degrees Celsius (°C); total radial strain and deformation of 2%, or a limiting cumulative damage fraction; and peak fuel temperature below the local fuel composition solidus temperature (that is, no fuel melting). NUREG/CR-7305 additionally specifies that a limit should be established for eutectic penetration of the cladding for AOOs.

NUREG/CR-7305 Section 2.3.1, Behaviors, Phenomena, and Properties, also discusses key behaviors of metallic fuel that impact safety. These include the geometric evolution of the cladding, which is heavily influenced by fuel swelling and cladding creep behavior; fuel properties, including the melting or solidus temperature, which are influenced by fuel constituent migration; cladding properties, including yield stress, solidus temperature, thermal properties, and irradiation effects; cladding rupture due to overpressure; and the overarching effect on all of these mechanisms of the cladding thinning caused by FCCI. While TerraPower did not fully establish a fuel performance envelope as noted above, the NRC staff determined that the pin damage acceptance criteria provided by TerraPower are consistent with the key phenomena discussed in NUREG/CR-7305 for normal operations and AOOs and, therefore, provide an acceptable approach to qualifying Natrium Type 1 fuel.

The NRC staff notes that future fuel qualification work would be expected to establish and appropriately justify limits on fuel pin design criteria used to demonstrate that the acceptance criteria are satisfied. While TR Table 6-8, Comparison of Fuel System Operational Parameters, appears to document much of the operating envelope for normal operations, these design limits are necessary to fully understand the conditions in which the fuel is expected to operate. The development of design limits and the definition of an associated operating envelope is in many cases also intrinsically tied to the analytical approaches used to evaluate the limits. For example, Section 3.2, Fuel Constituent Migration, of NUREG/CR-7305 discusses the significance of fuel constituent redistribution on thermal properties, but notes that the effect is captured in post-irradiation experiment (PIE) data below 10% burnup. If the fuel is to be used within the burnup envelope and evaluated using empirical fuel property models developed based on these PIE data, fuel constituent redistribution is included implicitly and is not a particular concern.

However, if the fuel is used significantly beyond the burnup envelope or it is analyzed with mechanistic models, fuel constituent redistribution should be modeled explicitly and one might expect a different set of limits to be developed, possibly on different parameters. Relatedly, the NRC staff also notes that in developing and evaluating these fuel performance limits, consideration should be given to ensuring that appropriate phenomena are included in the evaluation (e.g., evaluations of cladding stress and strain should account for any residual

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION stresses in the cladding from manufacturing, residual stresses and changes in material properties from end cap welding, compressive stress imposed by the wire wrap). The NRC staff also expects future qualification work to characterize the AOOs that the fuel is expected to experience without damage. See L&C 1 for a discussion of future work needed to qualify the fuel.

Fuel Assembly Damage Criteria In addition to effects on the fuel pin alone, the TR also provides additional design criteria for the fuel assembly in Table 4-1. The RAC and associated phenomena applicable to the fuel assembly include:

RAC 4.2-1.1 - stress, strain, and loading RAC 4.2-1.2 - fatigue RAC 4.2-1.3 - fretting wear, ((

))

RAC 4.2-1.4 - erosion and corrosion, ((

))

RAC 4.2-1.6 - dimensional changes such as duct bowing and dilation RAC 4.2-1.9 - hydraulic loads exceeding the hold-down capability of fuel, reflector, or shield assemblies RAC 4.2-1.12 - assembly component temperatures (which must be either limited or explicitly assessed in analyses demonstrating compliance with other fuel system damage criteria)

As discussed in Section 5.2.2, Fuel Assembly, of the TR, the purposes of the fuel assembly are to position the fuel in the core, provide passages to guide and control sodium for heat removal, provide shielding for components of the core support structure, provide features for proper interfacing with other core components ((

)), and to provide a physical barrier between fuel pins to minimize the effect of one assembly on adjacent assemblies. The NRC staff finds that the criteria presented above adequately encompass the phenomena that would degrade these functions and, therefore, determined that they provide an acceptable approach to qualifying Natrium Type 1 fuel. The NRC staff notes that as with the pin acceptance criteria, the actual limits were not provided; the development and justification of these limits is expected to be part of future fuel qualification work (see L&C 1).

The NRC staff additionally notes that the applicable fuel assembly design criteria in TR Table 4-1 reference ((

)). L&C 4 therefore states that if these design criteria are to be used to establish fuel assembly design limits, additional justification must be provided.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION 3.2.1.2 Evaluation model FQAF G2.1.2 states that evaluation models (EMs) should be available to assess fuel performance against design limits to protect against fuel failure and degradation mechanisms.

The NRC staff notes that Natrium fuel EMs are discussed in Section 6.4, Analytical Predictions, of the TR. However, the EM itself is assessed against a separate framework and is, therefore, discussed in Section 3.3, below.

3.2.2 Radionuclide release limits Complementing FQAF G2.1 criteria for normal operations and AOOs, FQAF G2.2 states that margin to radionuclide release limits under accident conditions should be demonstrated for the fuel. This goal is supported by four subgoals regarding the fuel performance envelope for accidents (G2.2.1); requirements for radionuclide retention within the fuel system (G2.2.2);

barrier degradation and failure criteria (G2.2.3); and radionuclide retention and release (G2.2.4).

In developing these goals and subgoals, NUREG-2246 states that as radionuclide inventory originates from the nuclear fuel, fuel qualification should include characterizing the behavior of the fuel under accident conditions, so that its contribution to the accident source term can be determined in a suitably conservative manner.

3.2.2.1 Fuel performance envelope for accidents FQAF G2.1.1, which was discussed previously in Section 3.2.1.1 of this SE, also covers the fuel performance envelope for accidents. TerraPower stated in the TR that the fuel performance envelope for accidents is defined by the fuel system failure criteria in TR Table 4-2, Design Criteria to Prevent Fuel System Failure, the fuel coolability criteria in TR Table 4-3, Design Criteria to Ensure Fuel Coolability, and the reactivity control insertability criteria in TR Table 4-4, Design Criteria to Ensure Reactivity Control Insertability.

While the NRC staff acknowledges that the fuel performance envelope for accidents is defined by the fuel failure criteria to some degree, the criteria that must be considered and what constitutes appropriate limits are also driven by the transients that may occur in the reactor.

The overall system transient response is, in turn, driven by the design of the plant, including SSCs such as reactivity control systems and pumps. This ultimately results in effects that can be applied to the fuel, such as power ramp rates, coolant flow rates, seismic loads, etc.

However, because these transients are defined by the system design rather than the fuel, TerraPower did not define the types or magnitudes of transients that the fuel is expected to experience in the TR.

Although the present approach does not currently include a discussion on the system transients that the fuel is expected to experience, the NRC staff expects that these transients would be defined in other licensing submittals that address transient safety analysis. In the TR, TerraPower performed a review of historical transient test data and proposed additional testing that addresses transient behavior of the fuel, as is discussed in Section 3.4 of this SE. The NRC staff finds TerraPowers approach to identifying the fuel performance envelope acceptable because of this planned testing that will be informed by transient safety analyses. This represents additional work needed to qualify the fuel (see L&C 1).

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION 3.2.2.2 Radionuclide retention requirements FQAF G2.2.1 states that the radionuclide retention requirements for the fuel - that is, the extent to which radionuclides are expected to be retained within the fuel system - should be specified.

While the TR implies (see Section 8, Online Fuel System Monitoring for Fuel Pin Failure, of the TR, for example) that the fission gases in the plenum would be expected to be released into the primary coolant upon cladding breach, TerraPower did not include specific radionuclide retention requirements. Instead, TerraPower stated that these requirements would be described in the Natrium preliminary safety analysis report (PSAR).

The NRC staff understands that the retention requirements would be addressed in a separate submittal covering TerraPowers approach to analyzing design basis accidents that result in releases of radionuclides from the fuel. As with the transient fuel performance envelope discussed above, this is reasonable considering that more than just the fuel system itself plays a role in defining these requirements. However, in addressing transient fuel behavior, TerraPower identified relevant historical test data and discussed plans for future testing and analysis, demonstrating an appropriate path forward. By contrast, no similar path forward was presented for radionuclide retention and release, which plays a key role in the source term expected from the fuel. As such, the NRC staff imposed L&C 5, which states that radionuclide retention and release requirements must be specified.

3.2.2.3 Barrier degradation and failure criteria FQAF G2.2.2 states that the criteria for barrier degradation and failure under accident conditions must be understood and suitably conservative when the design credits some retention of barrier integrity. In the case of Natrium Type 1 fuel, these criteria are for the radionuclide barriers represented by the fuel matrix and fuel cladding. This goal is supported by two subgoals: that the criteria must be conservative as demonstrated by comparison to data (G2.2.2(a)) and that the data used for validation are appropriate (G2.2.2(b)).

Barrier degradation and failure criteria (G2.2.2(a))

Barrier degradation is discussed in the context of the fuel performance criteria for normal operations and AOOs in Section 3.2.1.1 of this SE. Therefore, this section will focus on the criteria for accidents (i.e., failure). Table 2-1 of the TR identifies that RAC 4.2-2.1 through 4.2-2.5 represent the criteria for fuel pin failure. These acceptance criteria are discussed further in TR Table 4-2, which includes the following:

RAC 4.2-2.1 - Fuel cladding overheating RAC 4.2-2.2 - Fuel slug overheating RAC 4.2-2.3 - Fuel cladding deformation due to mechanical loads, up to and including cladding rupture RAC 4.2-2.4 - Fuel system mechanical fracturing caused by externally applied forces RAC 4.2-2.5 - Fuel cladding wastage While the TR indicates that TerraPower plans to identify the melting point of the cladding and protect against fuel cladding melting for the purposes of ensuring a coolable geometry, the RAC 4.2-2.1 fuel cladding temperature criterion is primarily to protect against ((

)). The fuel slug temperature criterion (RAC 4.2-2.2) is to prevent against ((

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION

)).

RAC 4.2-2.3 and 4.2-2.4 both cover mechanical limits on the cladding. RAC 4.2-2.3 relates to failure caused by deformation; this deformation could be driven by natural phenomena, such as cladding thermal creep and rod internal pressurization, or by externally applied forces.

According to the RAC, either a cladding deformation limit should be developed at which cladding failure is assumed, or deformation should be explicitly addressed in evaluations of other criteria that could be affected by cladding deformation. RAC 4.2-2.4 specifically relates to mechanical fracturing from externally applied forces, like seismic loading or dropping during a fuel handling accident.

Finally, RAC 4.2-2.5 applies to cladding wastage, stating that either a total cladding wastage limit should be developed at which cladding failure is assumed, or that cladding wastage should be explicitly addressed in evaluations of other criteria that could be affected by cladding wastage. Cladding wastage is caused by various mechanisms including fretting wear, erosion, corrosion, FCCI, and eutectic formation, all of which have been discussed previously. These mechanisms can occur on various time scales and as such this criterion is a complement to the RAC 4.2-2.1 and 4.2-2.2 temperature-based criteria, which specifically address ((

)).

The NRC staff compared these proposed criteria against the design criteria presented in NUREG/CR-7305 Section 2.2.2, Design Limits During Anticipated and Accident Transients.

The NUREG/CR indicates that the primary fuel failure mechanisms in a U-Zr/HT9 fuel system would be expected to be failure due to low-melting-point eutectics between the fuel and cladding, and cladding overpressure due to fission gas release. The recommended design criteria for postulated accidents are thus that fuel melting is precluded, and cumulative eutectic penetrations should be maintained below a specified limit to account for its effects as wastage.

Additional criteria are proposed for core coolability, but since TerraPower discussed its coolability criteria separately, these will be addressed in SE Section 3.2.3.1, below.

The criteria proposed by TerraPower for fuel pin failure are consistent with the criteria proposed in the NUREG/CR, with additional criteria that specifically account for fuel cladding deformation/rupture and mechanical fracturing caused by external loads. The NRC staff therefore determined that the identified fuel failure criteria are appropriate. The NRC staff notes that future fuel qualification work would be expected to develop and appropriately justify limits supporting these acceptance criteria (see L&C 1).

FQAF G2.2.2(a) supports G2.2.2, stating that the criteria used to determine barrier degradation and failure should be suitably conservative as demonstrated by comparison to transient testing and irradiated fuel samples. As noted above, TerraPower did not propose or justify limits on the acceptance criteria for fuel damage or failure and, as such, the NRC staff has not made a determination on whether these criteria are conservative. This is expected to be part of the future fuel qualification work (see L&C 1).

Experimental data (G2.2.2(b))

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION The TR contains information on fuel testing, both historical and planned, that supports the criteria discussed above. The data used to demonstrate the adequacy of the barrier degradation and failure criteria are nominally covered by FQAF G2.2.2(b). However, NUREG-2246 states that these data should be addressed using the data assessment framework. Therefore, discussion of the data used to assess the barrier criteria is provided in Section 3.4 of this SE.

3.2.2.4 Radionuclide retention and release FQAF G2.2.3 states that radionuclide retention and release behavior of the fuel matrix under accident conditions should be modeled conservatively. As discussed in Section 3.2.2.2 of this SE, TerraPower did not propose radionuclide retention requirements for Natrium Type 1 fuel in the TR and, therefore, the NRC staff did not assess whether TerraPower met this goal. It is the NRC staffs understanding that models of radionuclide retention in the fuel matrix and release out into the coolant following fuel failure will be covered in future Natrium licensing submittals.

The NRC staff therefore imposed L&C 55 on the use of this TR, which states that appropriate models for radionuclide retention and release must be proposed and justified.

In Table 2-1 of the TR where FQAF G2.2.3 is referenced, TerraPower included a statement that evaluations of the fuel system design will ensure that the design bases are met for normal operations, AOOs, and accidents, and that these evaluations will be supported by test data discussed in Section 6, Fuel System Design Evaluation, of the TR. The NRC staff evaluated the extent to which the available data and planned testing discussed in the TR support the development of radionuclide retention and release models; this discussion is included in Section 3.4 of this SE.

3.2.3 Safe shutdown FQAF G2.3 states that a safe shutdown condition (i.e., a subcritical condition with adequate decay heat removal) must be assured in any scenario. This goal is supported by two subgoals, which state that the fuel must maintain a coolable geometry under accident conditions (G2.3.1) and that negative reactivity insertion can be demonstrated (G2.3.2). Consistent with the overall objective of this goal, Section 1 of the TR states that the fuel design limits are established, in part, to ensure that [fuel] coolability is always maintained and fuel system damage is never so severe during postulated accidents as to prevent reactivity control and control rod insertion when it is required. The two sets of criteria supporting these objectives are discussed below.

3.2.3.1 Coolable geometry FQAF G2.3.1 relates to maintaining a coolable geometry. This goal is supported by two subgoals: that the phenomena that could cause a loss of coolable geometry are identified (G2.3.1(a)) and that there are evaluation models available to assess margin (G2.3.1(b)).

Phenomena that could cause a loss of coolable geometry (G2.3.1(a))

TerraPowers criteria to ensure fuel coolability are discussed in Table 4-3 of the TR and, like the criteria for fuel system damage and failure, are established as RAC with acceptance criteria that are supported by design criteria for the fuel pin and fuel assembly. The RAC identified by TerraPower are:

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION RAC 4.2-3.1 - cladding stress and strain, which must be kept below the point where cladding damage might prevent adequate core cooling or, alternatively, cladding stress and strain must be accounted for in analyses demonstrating compliance with fuel coolability criteria RAC 4.2-3.2 - cladding temperature, which must be kept below the melting temperature of the cladding RAC 4.2-3.3 - coolability evaluations must include the effects on core flow distribution or potential core blockage caused by ballooning of the cladding RAC 4.2-3.4 - fuel slug temperature must be lower than the melting temperature of the fuel RAC 4.2-3.5 - structural deformation of fuel assemblies due to combined loads must not prevent the ability to adequately cool the core RAC 4.2-3.6 - hydraulic loads combined with loads from natural phenomena must not unseat a fuel, reflector, or shield assembly and cause a reduction in flow that could prevent the ability to cool the fuel assembly NUREG/CR-7305 Section 2.2.2, Design Limits During Anticipated and Accident Transients, proposes a criterion to ensure that core coolability is maintained under all conditions, including beyond design basis accidents, by requiring that there is no clad melting. Other portions of the NUREG/CR also indicate that coolability is primarily influenced by fuel dimensional changes. In addition, fuel thermal conductivity is degraded as a function of burnup due to buildup of gaseous and solid fission products. Fuel swelling combined with property degradation can lead to cladding deformation if unconstrained. However, these are long-term phenomena that are particularly relevant to normal operations; other coolability issues that could result from transients, such as cladding ballooning (dilation) and debris ejection from breached fuel rods, should also be studied.

TerraPowers RAC explicitly preclude clad melting, ballooning, and the effects of fuel pin/assembly deformation that would affect coolability. The issue of whether debris ejected by a breached fuel rod could block an assembly is not explicitly covered by these criteria, but the NRC staff does not anticipate substantial debris generation provided that the fuel slug does not melt. As such, the criterion provided by RAC 4.2-3.4 is likely sufficient to preclude this issue.

Beyond the criteria identified in NUREG/CR-7305, the TerraPower criteria related to mechanical deformation and hydraulic loads on the fuel are necessary and sufficient to ensure that coolability can be adequately maintained following accidents, including events initiated by external hazards such as earthquakes.

In consideration of the above, the NRC staff determined that the acceptance criteria for core coolability provided in the TR provide an acceptable approach to qualify Natrium Type 1 fuel.

As with the criteria for fuel system damage and failure, future work is needed to propose and appropriately justify specific limits needed to ensure that the criteria are met (see L&C 1).

Evaluation models supporting coolable geometry demonstration (G.2.3.1(b))

In support of the overall demonstration that coolable geometry is maintained under accident conditions, NUREG-2246 FQAF G2.3.1(b) states that EMs should be available to assess margin to fuel coolability limits. In coolability assessments, it is typical that thermal-hydraulic or system transient EMs are needed to identify key parameters that affect the fuel. These EMs must be coupled with or otherwise inform the fuel performance evaluation (discussed in Section 6.4,

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION Analytical Predictions, of the TR) to ensure that combined thermal-hydraulic and thermo-mechanical effects are appropriately captured. The NRC staff notes that detailed thermal-hydraulic evaluations are outside the scope of the TR. However, Section 3.3 of this SE discusses the capabilities of the fuel performance codes in support of an overall coolability demonstration.

3.2.3.2 Negative reactivity insertion FQAF G2.3.2 relates to negative reactivity insertion. This goal is supported by two subgoals:

that criteria are provided to ensure that the means to insert negative reactivity is not obstructed during conditions of normal operation or accidents (G2.3.2(a)) and that an EM is available to assess geometry changes that could inhibit reactivity insertion during normal operation and accidents (G2.3.2(b)).

Criteria to demonstrate negative reactivity insertion (G2.3.2(a))

As discussed in Section 2.2 of this SE, control assemblies are responsible for providing negative reactivity insertion in the Natrium design. Each control assembly includes an absorber pin bundle that moves up and down within a control assembly duct. As such, the primary concern relative to negative reactivity insertion is that distortion of the pin bundle or duct could increase friction and impact the ability to insert the control rods into the core. This distortion could take place due to natural phenomena (e.g., bowing due to uneven thermal or irradiation creep) or accident conditions (e.g., external loading on the fuel assembly). Because the control assembly ducts are in physical contact with the other core assemblies in the core restraint system, control assembly distortion may be driven by forces imposed by adjacent fuel assemblies as they themselves distort (again, due to either natural phenomena or accident conditions). Additionally, because the control assemblies interface with the core support structure through inlet nozzles that provide coolant flow through the control assemblies, the control assemblies are subject to similar considerations for hydraulic loading as fuel assemblies.

TerraPowers criteria to ensure the capability to insert negative reactivity are included in Table 4-4 of the TR and, like the fuel system damage and failure and coolability criteria discussed above, are established as RAC with acceptance criteria that are supported by design criteria for the pin and fuel assembly. The RAC identified by TerraPower include:

RAC 4.2-4.1 - structural deformation of control assemblies due to combined loads will not prevent the ability to insert control rods during postulated accidents RAC 4.2-4.2 - hydraulic loads will not unseat a reactivity control assembly that could prevent the complete insertion of control rods during postulated accidents Table 4-4 of the TR indicates that these RAC are supported ((

)). The NRC staff additionally identified other RAC specified elsewhere in the TR that support control rod insertability, including RAC 4.2-1.6, 4.2-1.7, 4.2-1.8, 4.2-1.10, and 4.2-1.11, which relate to fuel and control assembly distortion, fuel and absorber pin internal pressure, hydraulic loads on control assemblies, and mechanical/neutronic design of control assemblies, respectively.

The NRC staff finds that TerraPowers acceptance criteria provide an acceptable approach to justify Natrium Type 1 fuel because these criteria address the possible mechanisms that could

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION inhibit control assembly insertion discussed above. As with other criteria discussed in this SE, future work is needed to propose and appropriately justify specific limits needed to ensure that the criteria are met (see L&C 1).

Evaluation models supporting negative reactivity insertion demonstration (G2.3.2(b))

In support of the overall demonstration that negative reactivity insertion can be achieved even during accident conditions, NUREG-2246 FQAF G2.3.2(b) states that EMs should be available to assess geometry changes resulting from normal operation and accident conditions in order to ensure that the negative reactivity insertion path is not deformed. Section 6.4 of the TR identifies that the OXBOW computer code (and in particular the OXBOW.CASS module) is used to perform these evaluations. However, the EM for evaluating control rod insertability was not assessed in detail in the TR. Section 3.3 of this SE provides further discussion on EMs.

3.3 Evaluation model Analytical prediction of fuel performance is a critical and complex portion of the fuel qualification process. Accordingly, NUREG-2246 provides a separate, generalized framework for assessing fuel performance EMs used to perform such analyses. However, the discussion on analytical capabilities provided in Section 6.4 of the TR contains only a high-level description of the proposed methods, with an overview of how the overall modeling approach intends to satisfy the FQAF EM goals provided in Table 6-28, Fuel Performance Models for FQAF Goals, of the TR.

The NRC staff understands that a detailed fuel performance methodology TR, which would seek to satisfy the FQAF EM goals, is planned for future submittal. Nonetheless, the NRC staff assessed the high-level TR information on the planned EM codes and structure against the FQAF EM framework.

Analyses performed at the fuel pin level are relevant to the assessment of barrier performance as discussed in Section 3.2 of this SE. The TR identified three computer codes that are used for various fuel pin predictions: ALCHEMY, a TerraPower in-house finite element model built on top of the commercial finite element code ABAQUS; ((

)); and CRUCIBLE, another TerraPower in-house code. ALCHEMY is ((

)), while ((

)). Conversely, CRUCIBLE is not used to evaluate fuel design parameters but instead is a part of TerraPowers steady-state core design and analysis suite that helps to determine fuel parameters affecting steady-state temperatures and evaluate uncertainties. Tables 6-24, Fuel Performance Prediction Capabilities to Assess Fuel Damage, through 6-27, Fuel Performance Prediction Capabilities to Assess Phenomena Related to Fuel Temperatures, of the TR detail the specific modeling capabilities of these (and other) codes in support of predicting fuel pin damage and failure, fuel coolability, and fuel temperatures.

Beyond fuel pin behavior, it is important to consider fuel assembly distortion in SFRs, since it plays a significant role in inherent reactivity feedback effects. This is applicable to the Natrium plant design due to its core restraint system, which is discussed briefly in Section 5.4, Core Restraint System, of the TR. Fuel assembly-level deformations are calculated using various modules of the OXBOW code, a TerraPower in-house finite element model built on top of the commercial finite element code ABAQUS. OXBOW is used to analyze the core restraint system, perform core assembly distortion analyses, analyze the overall mechanical response of fuel

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION assemblies under seismic conditions, determine fuel assembly withdrawal and insertion loads, conduct control bundle insertability and scram time analyses, and analyze fuel bundle-duct interactions.

The NRC staff has not previously reviewed or approved any of these codes. However, the NRC staff reviewed the information provided in the TR to determine the extent to which the codes may be capable of supporting future fuel qualification work. Section 6 of the TR, and in particular Table 6-3, Summary of Identified High-Importance Phenomena and Associated Design Limits and RAC for Fuel and Absorber Pins, provides details on a phenomena identification and ranking table (PIRT) assessment TerraPower performed to identify key phenomena3 that must be modeled to evaluate fuel performance under a range of conditions.

Tables 6-24 through 6-26 contain further information on which models are used to calculate specific high-ranked PIRT phenomena. All the fuel pin models referenced would at least be able to represent the fuel pins appropriately in a 1-dimensional axisymmetric geometry (as is typical of fuel performance models for operating reactors), although ALCHEMY would have the flexibility to represent distorted fuel geometries and other off-normal conditions since it is based on the finite element method. For that same reason, the OXBOW code would similarly be able to have a flexible representation of geometries at the fuel assembly scale. Based on this information, the NRC staff anticipates that these codes may be able to appropriately model key geometries, material properties, and relevant physics per FQAF EM G1.

Based on the above, the NRC staff concludes that the preliminary work on the EMs that support fuel qualification is acceptable. However, consistent with FQAF EM G1 and EM G2, additional work must be done to demonstrate that the proposed EMs contain all necessary material and physics models and to verify the EMs and validate them against appropriate experimental data (see L&C 1). This is echoed by TR Table 6-28, which notes that detailed validation plans and assessments are under development to demonstrate that validation assessment criteria have been met; qualification of fuel performance data is being performed to qualify existing experiment data; and new experiments are being performed. While an explicit assessment of the proposed methodologies against experimental data was not performed in the TR, the TR explained how both historical and planned testing supports the overall Natrium fuel qualification effort. These data are discussed in Section 3.4, below.

3.4 Data NUREG-2246 states that the assessment of experimental data is the largest area of review for fuel qualification. This is because various goals of the fuel qualification assessment framework, including the development of acceptance criteria and limits and the validation of EMs for fuel performance analyses, require comparison to data. Section 3.4, Assessment Framework for Experimental Data, of NUREG-2246 provides an assessment framework for experimental data, which is supported by several goals: that any assessment data are independent of data used to develop or train the EM (Experimental Data (ED) G1), that the data have been collected over a test envelope that covers the fuel performance envelope (ED G2), that the experimental data are accurately measured (ED G3), and that the tests are representative of prototypic conditions (ED G4).

3 Note that Table 6-3 identifies only high-ranked PIRT phenomena. The full PIRT results are detailed in [12], which was submitted to the NRC staff as a WP in 2021.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION Because TerraPower stated in the TR that the qualification of historical data and development of new tests to support fuel qualification is ongoing, the NRC staff did not fully utilize the NUREG-2246 data assessment framework. However, the NRC staff assessed the data provided in the TR and referenced the ways in which the TR discussion supports the FQAF data evaluation goals.

3.4.1 Historical fuel operating experience The TR contains numerous references to the testing of metallic fuel performed at EBR-11 and FFTF, stating in Section 6.3, "Testing," that "the intention is to rely heavily on this basis for the safety case and validation of methods for Type 1 fuel pins." Section 6.2, "Historic Operating Experience," of the TR justifies the applicability of this historic data to Natrium Type 1 fuel, with details provided in several tables. Table 6-5, "Summary of Fuel Pin Parameters Including Comparison to FFTF/MFF and EBR-11," compares the Type 1 fuel parameters to certain selected fuel assemblies that operated at FFTF and EBR-I1. Table 6-7 provides a comparison of Type 1 fuel assembly design parameters to fuel that operated at FFTF and proposed fuel for Power Reactor Innovative Small Module (PRISM) and Versatile Test Reactor (VTR). Table 6-8, "Comparison of Fuel System Operational Parameters," provides a comparison of the Type 1 bounding fuel system operating parameters. Finally, Table 6-9, "Relevant Historic Fuel Assemblies to Support Validation Activities," provides a detailed overview of all historical fuel assemblies operated at FFTF and EBR-11 that TerraPower identified as relevant to the Type 1 fuel design.

From a geometric standpoint, Type 1 fuel has a Eci,Pr0P((

11 than all the comparison fuel assemblies, except for several assemblies used at EBR-11, which come very close. The cladding is Eel/Prop((

11 than all the comparison assemblies. The fuel column length is Eel/Prop((

11-The plenum length is ECI/Prop((

pitch are t:c11n°P((

11-In an assembly, the fuel pin pitch to diameter ratio and assembly

)). The fuel pin wire wrap axial pitch is Eel/Prop((

11-Compared to the FFTF fuel, Eci,ProP((

)).

From an operating parameter standpoint, the peak enrichment is Eel/Prop((

)). The peak burnup is t:c11Prop[I 11, but the peak damage is Eciwr0P((

)). Maximum residence time is Eci,Prop

. The fast fluence is EcvProp

. the linear heat generation rate is t:cvProP((

11 - this would lead to Eel/Prop((

11.

The NRC staff notes that while the Type 1 fuel peak burnup provided in Table 6-8 of the TR is Eel/Prop((

)) than that of the comparison assemblies and generally well within the operational database, t:cl/Prop((

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION

((

))

((

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION TerraPower additionally has proposed a surveillance program in Section 9 of the TR, and the NRC staff finds it highly likely that any issues with Eel/Prop(( 11 - which are themselves unlikely based on a review of the historical data - will be identified by this surveillance program. As such the NRC staff considers a peak burnup limit of EciiProp(( 11 to be consistent with the acceptance criteria discussed previously and, therefore, acceptable for the purposes of qualifying Natrium Type 1 fuel.

Because Eel/Prop((

11 all exceed the historical database, these phenomena are assessed below in more detail.

With respect to ECI/Prop((

11-The NRC staff determined that this approach is acceptable, but notes that t:Cl/1-'rop((

L&C 1).

For ECI/Prop((

11 (see

]). Nonetheless, TerraPower also proposed additional testing and further PIE of historical fuel assemblies to confirm and reduce uncertainties. NUREG/CR-7305 notes that there is a complex relationship between Eel/Prop((

)). The NRC staff therefore determined that TerraPower's proposed path to address t:cwrop((

))

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION

)). However, ((

)).

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION using historical data and supplemental additional testing is acceptable to support Natrium Type 1 fuel qualification.

The NRC staff also noted the potential that fuel constituent redistribution in Type 1 fuel, which contributes to FCCI, would not be appropriately captured by the reference fuels. This is because constituent migration is driven by thermal gradients, and the Type 1 fuel ECI/Prop((

)). However, the Type 1 fuel operates at ECI/Prop((

)), which would result in the reference assemblies bounding Natrium Type 1 fuel with respect to constituent redistribution behavior.

In summary, the NRC staff finds the fuel assemblies detailed in Table 6-9 of the TR appropriate to provide qualification data for the Natrium Type 1 fuel because of their similarity to the Natrium Type 1 fuel design and relevance to the Natrium operating envelope. All assemblies presented are metallic fuel clad in stainless steel alloys that operated in SFRs. While some assemblies are more representative of Natrium Type 1 fuel than others, the NRC staff expects each assembly in the proposed database to be able to provide applicable insights for at least select mechanisms. The choice of a particular assembly to validate a certain model or behavior must be appropriately justified. The data that are planned to assess the various criteria and PIRT phenomena are discussed in Section 3.4.2, below.

3.4.2 Testing used to address acceptance criteria The TR includes details on the historical and contemporary testing that will be used to support the determination and qualification of design basis criteria. TerraPower noted in TR Section 6.2.1, Quality of Historic Data, that the quality of the historical test data needs to be addressed per NUREG-2246 ED G3. Because legacy test data have varying degrees of quality, additional effort may be needed to qualify these data for use at present. The methods that TerraPower proposed are: (1) demonstrate the equivalency of the quality program under which the data were collected to an NRC-approved quality assurance program that meets the requirements of 10 CFR Part 50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, (2) corroborate the data by comparison to other independent datasets, (3) perform confirmatory testing, or (4) perform a peer review of the data.

TerraPower has developed a qualification plan for legacy data from FFTF and EBR-II, which involves qualification by Argonne National Laboratory using an NRC-approved quality process.

The NRC staff, therefore, determined that this approach for the historical data from FFTF or EBR-II is generally acceptable. Other legacy data will be discussed below.

3.4.2.1 Cladding strain As provided in the TR, cladding strain assessments will be supported by a combination of data from ((

)). These data are supplemented by ((

)). The effect of rod internal pressure on strain will be further addressed by ((

)).

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION Planned testing in the area of cladding strain includes ((

)). The ((

)). Mechanical assessments will be supported by planned ((

)). Rod internal pressure will also be addressed by ((

)).

Cladding failure due to thermal creep will be addressed by ((

)). Future testing includes planned ((

)).

The NRC staff considers the use of historical cladding strain data acceptable for qualifying Natrium Type 1 fuel, as the ((

)) strain data would be expected to envelope expected Type 1 operating conditions. However, the NRC staff notes that ((

)) are necessary to provide confidence in ((

)).

The same is true for the ((

)). Also, while the historical data from

((

)) are applicable and important to consider, the NRC staff notes that the planned testing ((

)).

Similarly, the NRC staff considers the planned ((

)) tests to be essential for providing a thorough understanding of cladding creep failure behavior under a variety of conditions, including transients. In summary, the NRC staff determined that the combination of historical and planned testing provides an acceptable approach for qualifying Natrium Type 1 fuel because it appropriately addresses strain in unirradiated and irradiated material in a wide range of operating conditions for both steady-state operation and transients.

3.4.2.2 Fatigue As provided in the TR, fatigue lifetime assessments will be supported by ((

)). TerraPower also plans to conduct additional ((

)). The NRC staff finds this to be an acceptable approach for fuel qualification purposes because it is consistent with past assessments that ensured the appropriate treatment of nuclear fuel fatigue.

3.4.2.3 Fretting As provided in the TR, fretting will be addressed by ((

)). Planned work includes ((

)). While the ((

)) may not be fully applicable due to geometric and flow parameter differences relative to Type 1 fuel, ((

)) would be expected to address any differences. The NRC staff also notes that fretting is a highly localized phenomenon and even a significant amount of fretting would likely only result in a small numbers of fuel leaks. The NRC

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION staff determined that the approach to address fretting is acceptable because it supplements applicable historical data (which includes irradiated materials) with testing using prototypical materials and environments.

3.4.2.4 Corrosion As provided in the TR, corrosion will be addressed by ((

)). Some ((

)) has been performed, while additional assessment of ((

)) is planned. These data will be supplemented by ((

)). The NRC staff determined that the approach to address corrosion is acceptable because corrosion in SFRs is generally considered a minor issue and the data are expected to bound in-reactor corrosion.

3.4.2.5 FCCI As provided in the TR, FCCI will be addressed by ((

)).

Planned and ongoing assessments include further ((

)).

The NRC staff determined that the approach to address FCCI is acceptable because ((

)).

3.4.2.6 Fuel and cladding temperature As provided in the TR, damage to the fuel cladding caused by increased temperatures is assessed through ((

)). The NRC staff determined that this approach is acceptable because it is consistent with past approaches that ensured the appropriate assessment of fuel cladding damage due to elevated temperatures.

For cladding failure temperatures, the historical data include ((

)). TerraPower plans to perform various ((

)). The NRC staff determined that the approach to address cladding failure temperatures is acceptable because it starts from the ((

, )) and supplements that data with ((

)).

For fuel melting temperatures, TerraPower stated that the work completed to this point includes

((

)). This will be

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION supplemented by various ((

)). This approach is generally acceptable to the NRC staff because the tests involving ((

)) will appropriately capture ((

)) and ((

)).

3.4.2.7 Assembly mechanical behavior Section 6.3.1, Fuel Assembly Mechanical Test Plans, of the TR provides an overview of planned fuel assembly mechanical tests. These tests range from sub-component tests to multiple-assembly tests. The proposed tests address mechanical behavior of specific key components (fuel ducts, inlet nozzles), combinations of components (pin bundles, bundles and ducts), full fuel assemblies, and combinations of assemblies. Both static and dynamic behaviors are considered, including important component or assembly interactions. Proof-of-concept tests are also planned to demonstrate certain novel components or mechanisms. In Table 6-19, Example Fuel & Control Assembly Mechanical Test Matrix, of the TR, TerraPower categorized the tests according to which effects are tested and the critical characteristics expected to be addressed by each test.

The NRC staff determined that the proposed testing appropriately encompasses the necessary mechanical behavior, particularly when coupled with the testing noted above supporting acceptance criteria and the testing noted below concerning material properties. A footnote on Table 6-19 of the TR indicated that irradiation effects are accounted for by using pre-deformed ducts to simulate dilation or bowing induced by irradiation creep and growth, but that irradiation effects will be treated conservatively in analyses and will be validated by surveillance programs.

The NRC staff determined that the approach is acceptable because it is generally consistent with how mechanical behavior has been adequately tested and validated historically.

3.4.2.8 Material properties TerraPowers approach to material property data is provided in Section 6.3.2, Materials Property Data and Testing, of the TR. Data are taken from a variety of property handbooks, and if the data were not initially collected under an NRC-approved quality assurance program that meets the requirements of 10 CFR Part 50 Appendix B, the data are subsequently subjected to further qualification to ensure that the data meet a minimum quality standard.

HT9 As provided in the TR, thermal properties of HT9 appear to be readily available and can be qualified by ((

)). The NRC staff determined that this is acceptable because it is consistent with standard industry practice for thermal properties of cladding materials, which do not vary significantly for materials that fall within a given specification.

As provided in the TR, unirradiated material properties of HT9 also appear to be readily available. However, TerraPower is more constrained in the data that are available because the material must meet TerraPowers HT9 specification. It appears that TerraPower generally expects to be able to adopt unirradiated material properties based on ((

)) with the

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION potential for ((

)) if necessary. The NRC staff determined that this is acceptable because the ((

)) will identify those unirradiated material properties that may be expected to be affected by TerraPowers HT9 manufacturing process, which can then be addressed by ((

)).

As provided in the TR, irradiated HT9 material properties are further constrained. As TerraPower notes in Section 6.3.2.1.3, Irradiated Mechanical Properties, of the TR, some irradiated HT9 material properties have been measured historically and others have not.

TerraPower is currently conducting mechanical testing of irradiated samples to determine these properties. Because the samples are likely to not be fully prototypical, other tests, testing methods, or analysis techniques may be included to corroborate the data and confirm expected behavior. TerraPower similarly notes that HT9 chemical interaction data are also not likely to be fully prototypical due to irradiation conditions and, as such, plans a similar approach as that for obtaining irradiated material properties. The NRC staff notes that while it is always prudent to assess chemical compatibility between the cladding and coolant, corrosion is not expected to play a significant role in the cladding safety case. The NRC staff determined that the approaches for characterizing irradiated and chemical interaction properties of HT9 are acceptable because TerraPower will characterize these properties under conditions approximating in-reactor conditions, with further testing and analyses to address any gaps between tested conditions and fully prototypic conditions.

Metallic fuel TerraPower stated that the only high-importance property of the metallic fuel identified in the PIRT was the fuel thermal conductivity because it plays a key role in the thermal response of the fuel, especially during transient conditions. However, TerraPower also stated that other analyses are highly dependent on fuel properties and having reliable materials property data is essential to characterizing the beginning of life conditions of the fuel system. All thermal and unirradiated properties are taken from the literature and qualified with ((

)). The NRC staff determined that the treatment of unirradiated fuel properties is acceptable because they can be measured relatively easily.

For irradiated fuel properties, TerraPower states in Section 6.3.2.3, Metallic Fuel Properties Data for Qualification, of the TR that these properties are assessed by the fuel performance modeling tools. The NRC staff expects that some analysis work would be necessary to support the development of these properties, because irradiated properties would be highly dependent on local composition, temperature, etc., all of which depend on the processing and irradiation history of the material. However, the fuel performance models are themselves based on data and must be appropriately validated (see L&C 1). The NRC staff notes that this validation has not yet occurred, as discussed in Section 3.3 of this SE.

3.4.3 Conclusion regarding TerraPower fuel qualification data plans Based on the considerations discussed above, the NRC staff determined that the overall approach to developing fuel qualification data is acceptable. However, the NRC staff notes that details of planned testing were not provided and, as such, the NRC staff is not capable of assessing the extent to which the planned testing appropriately envelopes the planned operating conditions for Natrium Type 1 fuel. Additionally, there is a clear tie between the data used for assessment and the EMs which will be validated against that data. Since the EMs (as

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION discussed in Section 3.3 of this SE) have not been fully presented or validated, this represents an area of future work for TerraPower. Future work to qualify the fuel is addressed by L&C 1.

4.0 CONTROL ASSEMBLY QUALIFICATION PLAN EVALUATION Like fuel assemblies, control assemblies must also be qualified for use in a reactor. Despite operating in the same environment, being similar in design, and being subject to similar phenomena, control assemblies fulfill a different safety role from fuel assemblies.

As discussed in Section 2.2 of this SE, the control assemblies purpose is to ensure that neutron absorbing materials are positioned appropriately within the reactor to control or shut down the nuclear chain reaction. Because of this difference in the high-level safety objectives of the control assemblies versus the fuel assemblies, the full NUREG-2246 fuel qualification framework is not applicable to control assemblies. However, the high-level principles are applicable; specifically, that the control assemblies should be manufactured in accordance with an appropriate specification and that margin to appropriate design criteria can be demonstrated using EMs assessed against data. This section of the NRC staffs SE evaluates TerraPowers plans relative to the qualification of control assemblies considering NUREG-2246 high-level principles.

4.1 Manufacturing specifications As with the fuel assemblies, detailed design information for control assemblies was referenced in TR Table 5-3. These documents were audited by the NRC staff and the NRC staff confirmed in the audit that these documents contain sufficient detail to cover all key dimensions and tolerances of a control assembly.

Most components of the control assembly are composed of HT9, except for a few components that may be composed of code-qualified materials as discussed in Section 5.5.3 of the TR, and the boron carbide absorber pellets. HT9 manufacturing is discussed in Section 3.1 of this SE and that discussion is applicable to the control assemblies. Manufacturing specifications for boron carbide absorber pellets are referenced in Table 5-3 of the TR. The NRC staff reviewed the referenced documents in the audit and confirmed that the relevant information was appropriately included in the referenced documents.

Based on the above, the NRC staff determined that information provided in the TR provides an acceptable approach to demonstrate that control assembly manufacturing will be appropriately controlled.

4.2 Control assembly design criteria In choosing appropriate control assembly design criteria, it is important to consider that the function of the control assemblies is to ensure that neutron absorbing materials are positioned appropriately within the reactor to control the nuclear chain reaction. Thus, it is reasonable to expect that the criteria will be defined such that the neutron absorbing materials will remain in the absorber pins and the control assembly will be able to insert into the reactor.

Acceptance criteria that TerraPower identified as applicable to control assemblies include:

Damage:

o RAC 4.2-1.1 - stress, strain, or loading

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION o RAC 4.2-1.2 - fatigue o RAC 4.2-1.3 - fretting o RAC 4.2-1.4 - erosion and corrosion o RAC 4.2-1.5 - absorber-cladding chemical interaction (ACCI) o RAC 4.2-1.7 - dimensional changes, such as bowing, swelling, or dilation; these must be limited to prevent interference that could impact control rod insertability o RAC 4.2-1.8 - absorber pin internal pressure o RAC 4.2-1.11 - mechanical and neutronic lifetime, to ensure reactivity and insertability are maintained o RAC 4.2-1.12 - absorber material and cladding temperatures Failure:

o RAC 4.2-2.1 - absorber cladding overheating o RAC 4.2-2.4 - mechanical fracture caused by externally applied forces Reactivity Control Insertability:

o RAC 4.2-4.1 - structural deformation of control assemblies due to combined accident loads and natural phenomena o RAC 4.2-4.2 - hydraulic loads combined with loads from natural phenomena, which must not unseat a reactivity control assembly in a way that prevents complete control rod insertion The NRC staff determined that the absorber pin RAC provide an acceptable approach to support qualification of Natrium control assemblies because these criteria ensure that absorber materials will: (1) remain neutronically effective (by RAC 4.2-1.11); (2) will stay enclosed within the absorber pin (by inhibiting absorber pin cladding damage or failure by various mechanisms);

and (3) will not deform in a manner that inhibits their proper insertion in the core. Additionally, the negative reactivity insertion RAC are discussed in more detail in Section 3.2.3.2 of this SE.

4.3 Evaluation model for control assemblies Section 6.4.1.1, Alchemy, of the TR indicates that ALCHEMY can model boron carbide in place of fuel within pins, giving it the flexibility to model absorber pin performance. This is generally consistent with other approaches that have been shown to adequately model absorber pin performance. As discussed in Section 3.3 of this SE, TerraPower intends to use a module of the OXBOW code to analyze control rod assembly insertability criteria, which are related to control assembly distortion. Both of these codes are discussed in more detail in Section 3.3 of this SE. Specifically, while the work presented in the TR is acceptable at this stage, as with the fuel performance methodologies, these methodologies require further work to appropriately validate them against experimental data (see L&C 1).

4.4 Data for control assemblies Table 6-6, Summary of Absorber Pin Parameters Including Comparison to FFTF and JOYO, of the TR includes a detailed comparison of absorber pin parameters between the proposed Natrium absorber pin design, three designs that were operated at FFTF, and a design from the Joyo reactor in Japan. The Natrium absorber pins are generally comparable to or within the envelope of comparison pins, except that they have ECI/Prop((

)).

Additionally, ECI/Prop((

)). Additional reference pins are presented in Table 6-10, Relevant Historic Absorber/Control Pin Test Assemblies to Support Validation

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION Activities, of the TR, which includes absorber pins from FFTF, Joyo, EBR-II, and the Engineering Test Reactor. These pins appear to cover a reasonably wide range of designs and operating conditions.

In general, the historical data appear applicable for qualifying Natrium control assemblies. While there are ECI((

)), there is a wealth of data on ECI((

)). Additionally, the fact that the absorber pin data come from a fairly diverse set of designs and operating conditions means that the data will include features, like differences in neutron spectra between metallic-and oxide-fueled SFRs, that are needed to fully qualify the absorber pin performance.

As noted in Section 3.4 of this SE, TerraPower is conducting or plans to conduct additional tests to assess HT9 behavior. These are expected to be generally applicable to HT9 performance in absorber pins as well. The main gap with respect to absorber pin qualification is ACCI; TerraPower noted in TR Table 6-11, Design Basis Criteria and Supporting Information to Prevent Fuel Pin Damage, that ((

)), but plans to conduct ((

)) to confirm this. The NRC staff determined that this approach is appropriate because it includes testing that addresses potential information gaps, especially in light of the fact that ECI((

)).

5.0 FUEL MONITORING Section 8 of the TR provides a description of the method by which TerraPower plans to detect fuel failures during operation. Cover gas will be monitored continuously and the detection of fission products in the cover gas is indicative of a fuel failure. TerraPower notes that ((

)).

The NRC staff determined that this approach for detecting failures is acceptable because it will allow failed fuel assemblies to be identified and appropriately managed. The NRC staff notes that ((

)).

6.0 FUEL SURVEILLANCE AND LEAD DEMONSTRATION AND TEST ASSEMBLY PROGRAM As discussed in Section 2.1.4 of this SE, TerraPower plans to collect additional data on fuel performance by using LDAs and LTAs. The purpose of the LDA program is to ((

)).

TerraPower also stated that LDAs may be designed such that they ((

)). The notional surveillance plan for the first several cycles of operation is presented in TR Table 9-1, Notional Fuel Surveillance Plan for Initial Cycles of Operation. Additional detail on LTAs beyond the purpose and the notion of ((

)) was not provided.

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION The NRC staff notes that there is a long history of similar LTA programs in the operating nuclear reactor fleet and, as such, there is ample precedent for a program conceptually similar to what TerraPower has proposed. However, because the LDA and LTA programs as presented in the TR are notional, the NRC staff cannot reach a final determination regarding the acceptability of LDAs or LTAs for the purposes discussed in the TR. In particular, the TR lacks details on how LDAs or LTAs would be evaluated to ensure that their impact on the co-resident fuel is minimized, and that any uncertainties in performance due to ((

)) remain appropriately bounded or can otherwise be tolerated in safety analyses. Further, it is not clear to the NRC staff how ((

)). These issues should be addressed in future licensing submittals.

LIMITATIONS AND CONDITIONS

1. This TR represents an acceptable approach for qualifying Natrium Type 1 fuel and control assemblies for use in a reactor but does not in and of itself demonstrate that the fuel and control assemblies are qualified. Additional activities, including those discussed in the NRC staffs SE, must be completed to execute this plan and appropriately justify that the fuel and control assemblies are qualified.

2. This TR addresses the material properties and performance of U-10Zr and HT9 in fuel. If other materials are used in the fuel system in licensing applications, the applicant or licensee must demonstrate that they are manufactured according to standard specifications and used consistent with their qualification under relevant NRC-accepted codes and standards, or otherwise appropriately justified.

3. This TR does not provide a means for demonstrating that proposed SARRDLs are satisfied during normal operations and AOOs for the Natrium plant. The role of the fuel acceptance criteria is to demonstrate that the fuel system is not damaged as a result of normal operations and AOOs; if these criteria are satisfied, then the fuel system need not be further assessed against the SARRDLs. However, the SARRDLs must still be evaluated against other sources of radionuclides, including circulating radionuclides resulting from an appropriate number of random fuel failures.

4. The ((

)) have not been subject to previous NRC review or approval. If they are to be used to develop design criteria and associated limits that support fuel assembly acceptance criteria, these design criteria and associated limits must be appropriately justified.

5. This TR does not address the extent to which the fuel system is expected to retain radionuclides following a cladding breach. If an applicant or licensee wishes to qualify Natrium Type 1 fuel with an expectation that radionuclides are expected to remain within the fuel following a cladding breach, models for fuel system radionuclide retention and release must be proposed and appropriately justified by comparison to experimental data.

CONCLUSION The NRC staff determined that the TR, subject to the limitations and conditions discussed above, provides an acceptable approach for qualifying fuel and control assemblies for the Natrium reactor based on (1) the inclusion of sufficient information to demonstrate that fuel and control assemblies are manufactured in a process that provides adequate control over key parameters, (2) the identification of appropriate safety criteria for both fuel and control assemblies, (3) the development and justification of a significant applicable historical test

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION database, (4) the development of a test plan that appropriately fills gaps in the historical test database, and (5) a robust fuel monitoring program. Accordingly, the NRC staff concludes that the qualification plan provided in the TR can be used to support compliance with 10 CFR 50.43(e) and proposed Natrium PDCs.

REFERENCES

1.

TerraPower, Fuel and Control Assembly Qualification, NATD-FQL-PLAN-0004, Revision 0, dated January 25, 2023 (ML23025A409)

2.

R. Sprengel, TerraPower, letter to U.S. Nuclear Regulatory Commission, Correction to TerraPower Fuel and Control Assembly Qualification Topical Report, TP-LIC-LET-0061, dated April 18, 2023 (ML23109A099)

3.

G. Wilson, TerraPower, letter to U.S. Nuclear Regulatory Commission, Submittal of the Construction Permit Application for the Natrium Reactor Plant, Kemmerer Power Station Unit 1, TP-LIC-LET-0124, dated March 28, 2024 (ML24088A060)

4.

TerraPower, Principal Design Criteria for the Natrium Advanced Reactor, NATD-LIC-RPRT-002, Revision 0, dated January 24, 2023 (ML23024A281)

5.

U.S. Nuclear Regulatory Commission, Fuel Qualification for Advanced Reactors, NUREG-2246, Revision 0, dated March 2022 (ML22063A131)

6.

U.S. Nuclear Regulatory Commission, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section2, Fuel System Design, Revision 3, dated March 2007 (ML070740002)

7.

U.S. Nuclear Regulatory Commission, Metal Fuel Qualification: Fuel Assessment Using NRC NUREG-2246, Fuel Qualification for Advanced Reactors, NUREG/CR-7305, dated August 2023 (ML23214A065)

8.

Nuclear Energy Institute, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, NEI 18-04, Revision 1, dated August 2019 (ML19241A472)

9.

U.S. Nuclear Regulatory Commission, Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors, RG 1.233, Revision 0, dated June 2020 (ML20091L698)

10. TerraPower, Advanced Fuel Qualification Methodology Report, dated July 16, 2020 (ML20209A155)

11. M. Sutton, U.S. Nuclear Regulatory Commission, Fuel and Control Assembly Qualification Audit Plan, dated June 7, 2023 (ML23156A553)

12. U.S. Nuclear Regulatory Commission, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors, RG 1.232, Revision 0, dated April 2018 (ML17325A611)

OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION OFFICIAL USE ONLY - PROPRIETARY AND EXPORT CONTROLLED INFORMATION

13. U.S. Nuclear Regulatory Commission, NRC Program on Knowledge Management for Liquid-Metal-Cooled Reactors, NUREG/KM-0007 / ORNL/TM-2013-79, dated April 2014 (ML14128A346)

14. TerraPower, Type 1 Fuel Pin Qualification Plan, AFQMG-ENG-PLAN-0001R, Revision 0, dated February 26, 2021 (ML21057A008)

Project Manager: Mallecia Sutton, NRR (Lead)

Stephane Devlin-Gill, NRR Roel Brusselmans, NRR Deion Atkinson, NRR Principal Contributor: Reed Anzalone, NRR

Copyright © 2023 TerraPower, LLC. All rights reserved.

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 NAT-1911 Rev. 1 Governing Procedure: NAT-1848 Controlled Document - Verify Current Revision 3001823a-d82b-4a8b-ab8d-7d565687cf5c Document Title TerraPower, LLC (TerraPower) Natrium Topical Report: Fuel and Control Assembly Qualification Natrium Document No.:

NAT-2806 Rev. No.:

0 Page:

1 of 119 Effective Date:

Target Quality Level:

QL3 Supplier Document No.:

N/A Supplier Rev:

N/A Originating Organization:

TerraPower Quality Level:

QL3 Export Controlled:

No Attachments:

N/A Document Type:

PLAN Open Items:

N/A Record Status:

Released Natrium MSL #:

FQL Approval Approval signatures are captured and maintained electronically; See Electronic Approval Records in EDMS.

04/17/2023

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 REVISION HISTORY Revision No.

Effective Date Affected Section(s)

Description of Change(s) 0 6.2, 6.3.1.4.4 This release supersedes NATD-FQL-PLAN-0004 Rev 0 to comply with the updated document numbering procedure for the Natrium project. Minor revisions were made to correct section reference in 6.2 and update terminology in 6.3.1.4.4 to be consistent with the remainder of the document. Cover page was also updated to latest Natrium template.

04/17/2023

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 TABLE OF CONTENTS TERMS / ACRONYMS / DEFINITIONS...................................................................................................... 7 EXECUTIVE

SUMMARY

.......................................................................................................................... 10 ACKNOWLEDGEMENT........................................................................................................................... 10

1. PURPOSE.......................................................................................................................................... 11

2. BACKGROUND................................................................................................................................. 12 2.1 Design Background................................................................................................................... 12 2.2 Regulatory Background............................................................................................................. 13

3. DISCUSSION..................................................................................................................................... 21

4. FUEL DESIGN CRITERIA.................................................................................................................. 22

5. FUEL DESIGN DESCRIPTION.......................................................................................................... 29 5.1 Overview of the Fuel Design of the Natrium Reactor................................................................. 29 5.2 Fuel Assemblies........................................................................................................................ 30 5.3 Control Assemblies.................................................................................................................... 42 5.4 Core Restraint System............................................................................................................... 49 5.5 Materials.................................................................................................................................... 52 5.6 Verification of the Fuel System Design Basis............................................................................. 54

6. FUEL SYSTEM DESIGN EVALUATION............................................................................................ 56 6.1 High-importance Phenomena.................................................................................................... 57 6.2 Historic Operating Experience................................................................................................... 61 6.3 Testing....................................................................................................................................... 74 6.4 Analytical Predictions................................................................................................................. 97

7. TESTING AND INSPECTION OF NEW FUEL..................................................................................108

8. ONLINE FUEL SYSTEM MONITORING FOR FUEL PIN FAILURE..................................................109

9. FUEL SURVEILLANCE.....................................................................................................................110

10. CONCLUSIONS................................................................................................................................112

11. REFERENCES..................................................................................................................................112

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 LIST OF TABLES Table 2-1. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Appendix A Goals in FQAF 14 Table 2-2. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Evaluation Model Assessment Framework Goals..................................................................................................................................... 19 Table 2-3. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Experimental Data Assessment Framework Goals................................................................................................................. 20 Table 41. Design Criteria to Prevent Fuel System Damage-.................................................................... 23 Table 4-2. Design Criteria to Prevent Fuel System Failure....................................................................... 26 Table 4-3. Design Criteria to Ensure Fuel Coolability............................................................................... 27 Table 4-4. Design Criteria to Ensure Reactivity Control Insertability Criteria............................................. 28 Table 5-1. Nominal Composition of HT9 Steel.......................................................................................... 53 Table 5-2. Manufacturing Steps for HT9 Components.............................................................................. 53 Table 5-3. Summary of Completed and Planned Activities to Satisfy Fuel System Design Description Requirements (RAC 4.2-5)....................................................................................................................... 55 Table 6-1. Importance Ranking Definitions............................................................................................... 56 Table 6-2. Knowledge Level Definitions................................................................................................... 57 Table 6-3. Summary of Identified High-Importance Phenomena and Associated Design Limits and RAC for Fuel and Absorber Pins............................................................................................................................ 58 Table 6-4. Summary of Identified High-Importance Phenomena and Associated Design Limits and Applicable RAC for Fuel and Control Assemblies..................................................................................... 60 Table 6-5. Summary of Fuel Pin Parameters Including Comparison to FFTF/MFF and EBR-II................ 65 Table 6-6. Summary of Absorber Pin Parameters Including Comparison to FFTF and JOYO.................. 65 Table 6-7. Type 1 Fuel Assembly Design Parameters.............................................................................. 66 Table 6-8. Comparison of Fuel System Operational Parameters.............................................................. 68 Table 6-9. Relevant Historic Fuel Assemblies to Support Validation Activities.......................................... 71 Table 6-10. Relevant Historic Absorber/Control Pin Test Assemblies to Support Validation Activities...... 73 Table 6-11. Design Basis Criteria and Supporting Information to Prevent Fuel Pin Damage.................... 75 Table 6-12. Design Basis Criteria and Supporting Information to Predict Fuel Failure.............................. 76 Table 6-13. Design Basis Criteria and Supporting Information to Ensure Fuel Pin Coolability and Absorber Pin Insertability......................................................................................................................................... 77 Table 6-14. Summary of Future Testing Activities to Validate Fuel Damage Limits.................................. 78 Table 6-15. Summary of Future Testing Activities to Predict Fuel Failure................................................. 80 Table 6-16. Summary of Future Testing Activities to Ensure Fuel Coolability is Maintained..................... 83 Table 6-17 Summary of Tests to Address High-Importance Fuel and Absorber Pin Phenomena............. 84 Table 6-18. Major Effects on Fuel Assembly Behavior............................................................................. 88 Table 6-19. Example Fuel & Control Assembly Mechanical Test Matrix................................................... 89 Table 6-20. Summary of HT9 Data Qualification...................................................................................... 91 Table 6-21. Availability of Data for HT9.................................................................................................... 94 Table 6-22. Summary of U-10Zr Material Properties Data to be Qualified................................................ 96 Table 6-23. Applicable Models and Codes for Fuel Pin Phenomena........................................................ 98 Table 6-24. Fuel Performance Prediction Capabilities to Assess Fuel Damage....................................... 99 Table 6-25. Fuel Performance Prediction Capabilities to Assess Fuel Failure.......................................... 99

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-26. Fuel Performance Prediction Capabilities to Assess Fuel Coolability...................................100 Table 6-27. Fuel Performance Prediction Capabilities to Assess Phenomena Related to Fuel Temperatures..........................................................................................................................................101 Table 6-28. Fuel Performance Models for FQAF Goals...........................................................................101 Table 7-1. Summary of New Fuel Testing and Inspection Needs and Planned Approach.......................109 Table 9-1. Notional Fuel Surveillance Plan for Initial Cycles of Operation...............................................110 Table 9-2. Fuel Performance Uncertainties and Mitigation Steps............................................................111

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 LIST OF FIGURES Figure 2-1. Overview of Natrium Reactors Safety Features..................................................................... 12 Figure 3-1. Overall Fuel Qualification Assessment Logic Flow................................................................. 22 Figure 5-1. Natrium Type 1 Fuel Pin........................................................................................................ 30 Figure 5-2. Wire Wrap Fused Ball Termination......................................................................................... 31 Figure 5-3. Natrium Fuel Assembly Design.............................................................................................. 32 Figure 5-4. Inlet Nozzle with Duct ((

))(a)(4)...................................................................... 34 Figure 5-5. ((

))(a)(4).................................................................................................................................... 34 Figure 5-6. Handling Socket ((

))(a)(4).......................................................... 35 Figure 5-7. Inlet Nozzle ((

))(a)(4) Cross Section........................................................................ 36 Figure 5-8. Fuel Pin Attachment Hardware for a ((

))(a)(4)(ECI) Pin Bundle............................................. 37 Figure 5-9. Natrium ((

))(a)(4)(ECI) Fuel Pin Bundle Cross Section in Fueled Region.............................. 38 Figure 5-10. Potential Removable Pin Locations...................................................................................... 40 Figure 5-11. Key Differences of Type 1 and Type 1B (LTA) Fuel Pin Design.......................................... 42 Figure 5-12. Illustrative Example of a Natrium Control Assembly............................................................. 44 Figure 5-13. Cross Section View of Natrium Control Assembly................................................................ 45 Figure 5-14. Natrium Control Rod Absorber Pin Attachment.................................................................... 46 Figure 5-15. Control Assembly - Damper and Driveline Assemblies........................................................ 46 Figure 5-16. Control Assembly Lower Detail............................................................................................ 47 Figure 5-17. Natrium Control Rod Connection.......................................................................................... 48 Figure 5-18. Primary vs. Secondary Control Rod Bundle Cross-Section.................................................. 49 Figure 5-19. Schematic of a Core Restraint System and Impactful Parameters....................................... 50 Figure 5-20. Contributing Effects to Core Assembly Restrained Thermal Bowing Deformations............... 50 Figure 6-1. Burnup Distribution Comparison between MFF, EBR-II, and Natrium Fuel Pins..................... 69 Figure 6-2. Fuel Linear Heat Generation Rate Distribution (left) and Burnup Distribution (right) for a Nominal Pin in the Inner Core Region...................................................................................................... 69 Figure 6-3. Cladding Surface Temperature Distribution at Beginning of Life............................................ 70

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 TERMS / ACRONYMS / DEFINITIONS Acronym Term/Definition 316 SS 316 Stainless Steel ABAQUS API ABAQUS Application Programming Interface. ABAQUS is a commercial finite element analysis code with a scripting interface.

ACCI Absorber-Cladding Chemical Interaction. Chemical reaction between the cladding and absorber that degrades the cladding mechanical properties. The thickness of the impacted region contributes to cladding wastage.

ACLP Above Core Load Pad AFCI Advanced Fuel Cycle Initiative. Department of Energy program on metallic fuel.

AFQM Advanced Fuel Qualification Methodology. TerraPower project, funded under a regulatory assistance grant, focused on developing a methodology for qualifying metallic fuel, including early engagement with the NRC.

AOO Anticipated Operational Occurrences ASME BPVC American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

Standards for the safe design, manufacture and maintenance of boiler and pressure vessels, power-producing machines, and nuclear power plant components BCC Body-centered Cubic. Materials crystal structure CRD Control Rod Drive CRDM Control Rod Drive Mechanism CRS Core Restraint System CSS Core Support Structure CTE Coefficient of Thermal Expansion DBTT Ductile to Brittle Transition Temperature DC Design Criteria DE destructive exams DOE Department of Energy DSC Differential Scanning Calorimetry EBR-II Experimental Breeder Reactor-II. Sodium-cooled fast reactor known for a series of experiments demonstrating passive safety features such as natural convection cooling after a simulated cooling pump failure EM Evaluation Model FCCI Fuel-Cladding Chemical Interaction. Chemical reaction between the fuel and cladding that degrades the cladding mechanical properties in the interacted zone. The thickness of the impacted region contributes to cladding wastage.

FCRD Fuel Cycle Research and Development. DOE research program on advanced fuels.

They issued a Materials Handbook with relevant materials properties data for the Natrium fuel design.

FEA Finite Element Analysis FEM Finite Element Model FFTF Fast Flux Test Facility. 400 MW thermal, liquid sodium cooled fast test reactor that operated from 1982 to 1992 FGR Fission Gas Release FIV Flow Induced Vibration FM Ferritic Martensitic

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Acronym Term/Definition FQAF Fuel Qualification Assessment Framework Fuel Fissile material used to sustain a nuclear chain reaction in a reactor Fuel Pin Structural component of sodium-cooled fast reactor that consists of a steel tube housing a fuel column with extra volume (plenum) to contain fission gases GEH GE-Hitachi IAEA International Atomic Energy Agency IFR Integral Fast Reactor. DOE research program to support advancing metallic fuels for closed fuel cycle applications.

IVHM In-Vessel Handling Machines LBE Licensing Basis Events LDA Lead Demonstration Assembly. Fuel assemblies with the ability to readily remove fuel pins from the assemblies after irradiation. Used in the Natrium Reactor as part of the fuel surveillance program to provide data on Type 1 fuel at high burnup ahead of the rest of the core, as well as to irradiate lead Type 1B test pins.

LHGR Linear Heat Generation Rate. Local power generated per unit of length of fuel/absorber.

LMFBR Liquid Metal Fast Breeder Reactor. Reactor that is cooled by a liquid metal and produces more fissionable material than it consumes to generate energy.

LTA Lead Test Assembly. Fuel assemblies that contain design features or materials that have not been approved for unrestricted use.

MFF A series of fuel assemblies with metallic fuel that were irradiated in the FFTF to support conversion of the reactor from mixed oxide fuel to metallic fuel.

NDE non-destructive exams NRC U.S. Nuclear Regulatory Commission NSMH Nuclear Systems Materials Handbook PICT Peak Inner Cladding Temperature PIRT Phenomena Identification and Ranking Tables PRISM Power Reactor Innovative Small Module. PRISM is a pool-type, metal-fueled, small modular sodium fast reactor designed by GE-Hitachi PSAR Preliminary Safety Analysis Report RAC Regulatory Acceptance Criteria. Acceptance criteria derived from regulatory requirements and guidance RCP Regulatory Compliance Plan RES Reactor Enclosure System RG Regulatory Guide SAS SAS4A/SASSYS-1 system analysis code SD Smear Density. Cross-sectional area of the fresh metallic fuel/cross-sectional area of the fuel pin cladding inner diameter SFR Sodium Fast Reactor/Sodium-cooled Fast Reactor. Nuclear reactor with a fast neutron spectrum and liquid sodium coolant SQA Software Quality Assurance SSC Structure, System, and Component TLP Top Load Pad

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Acronym Term/Definition TREAT Transient Reactor Test Facility. Test reactor facility at Idaho National Laboratory that can perform extreme transient tests on fuel to assess fuel failure limits and post-failure behavior.

TWR Traveling Wave Reactor. TerraPower reactor design for a sodium-cooled fast reactor that can convert fertile material into usable fuel through nuclear transmutation, in tandem with the burnup of fissile material. TWRs differ from other kinds of fast-neutron and breeder reactors in their ability to breed and then burn the generated plutonium within the same intact fuel pin, without an interim reprocessing step.

Type 1/

Type 1 Fuel Fuel utilizing U-10Zr as the fuel alloy, sodium-bond within the fuel, and HT9 cladding.

Fuel is similar in composition and dimensions to the fuel pins already reliably used.

Natrium Reactor will begin operation with Type 1 fuel.

Type 1B/ Type 1B Fuel Advanced Natrium Reactor fuel that enables significantly higher burnup ULOF Unprotected Loss of Flow UTOP Unprotected Transient Over Power UTS Ultimate Tensile Strength V&V Verification and Validation YS Yield Strength

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 EXECUTIVE

SUMMARY

This report presents TerraPower, LLCs (TerraPower) plan to qualify fuel and control assemblies to support operation of the Natrium' Reactor, a TerraPower and GE-Hitachi technology. A systematic assessment was performed to identify the activities required to support fuel system qualification, including the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of experimental data used to develop and validate evaluation models and empirical safety criteria. This report identifies the acceptance criteria for fuel qualification and presents TerraPowers fuel qualification results to date as well as plans for future fuel qualification activities.

This report includes Regulatory Acceptance Criteria (RAC) (i.e., acceptance criteria derived from regulatory requirements) that when satisfied, support a finding that the fuel is qualified for use (i.e.,

reasonable assurance exists that the fuel, fabricated in accordance with its specification, will perform as described in the safety analysis). Specifically, the fuel design criteria and associated limits ensure four key objectives: 1) the fuel system is not damaged as a result of normal operation and Anticipated Operational Occurrences (AOOs), 2) the number of fuel pin failures is not underestimated for postulated accidents, 3) coolability is always maintained, and 4) fuel system damage is never so severe during postulated accidents as to prevent reactivity control and control rod insertion when it is required. High-importance fuel phenomena identified for all applicable fuel pin design limits include fission gas release, HT9 mechanical behavior as a function of environmental conditions, fuel-cladding chemical interaction, and fuel thermal conductivity as a function of irradiation/porosity.

Completed and ongoing efforts address major aspects of fuel qualification requirements, while future analyses (e.g., fretting and fatigue behavior, additional testing and analysis to address extreme transients) are expected to provide the final scope of information needed to fully qualify fuel for the Natrium Reactor.

Fuel qualification for the Natrium Reactor relies, in part, on historic operating experience and historic data (e.g., Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) metallic fuel pins). This historic data will be qualified under a program that satisfies the quality assurance requirements of 10 CFR 50 Appendix B. With no operating fast-spectrum reactor available to perform final tests, a surveillance program is proposed to monitor the irradiation performance of the fuel to ensure consistent performance with historic operating experience and analytical predictions. The proposed surveillance program includes the capabilities to incorporate knowledge gained from analyses and testing data that becomes available as fuel qualification activities progress.

ACKNOWLEDGEMENT This topical report represents the effort and determination of many people, including the following contributors: Jesse Cheatham, Ryan Christensen, Francesco Deleo, Lynne Ecker, Ian Gifford, Bruce Hilton, Joseph Hoffman, Virginia Hollis, Julie Jordan, Joseph LaPrad, Anh Mai, Jason Meng, Sam Miller, Brian Morris, Drew Mueller, Matthew Presson, Christopher Regan, and Javier Romero.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054

1. PURPOSE This report presents the TerraPower plan to qualify fuel to support operation of the Natrium Reactor.

The overall fuel qualification approach (planning, testing, analysis, etc.) used to obtain qualified fuel is described. TerraPowers fuel qualification efforts have been informed by U.S. Nuclear Regulatory Commission (NRC) guidance, including Regulatory Guide (RG) 1.206, Section C.I.4, Reactor [1],

NUREG-0800, Section 4.2, Fuel System Design [2], and NUREG-2246, Fuel Qualification for Advanced Reactors [3]. Additionally, principal design criteria (PDC) that are applicable to fuel performance and fuel qualification have been informed by RG 1.232, Guidance for Developing Principal Design Criteria for Non-Light-Water Reactors [4]. TerraPower has provided several reports to the NRC regarding fuel qualification efforts [2] [5] [7], and the NRC has provided feedback [3] [4] [6];

the NRCs feedback has informed the development of TerraPowers overall approach to fuel qualification. The information presented in this report will apply to licensing efforts associated with the Natrium Reactor design.

This report identifies the RAC (i.e., acceptance criteria derived from regulatory requirements) that will be used for fuel qualification and presents TerraPowers fuel qualification results to date. Fuel qualification for the Natrium Reactor design includes the identification of key fuel manufacturing parameters, the specification of a fuel performance envelope to inform testing requirements, the use of evaluation models in the fuel qualification process, and the assessment of experimental data used to develop and validate evaluation models and empirical safety criteria. TerraPower uses historic operating experience and data from EBR-II and FFTF, verifying the suitability of the historic data and qualifying the historic data for use in TerraPowers fuel qualification methodology. This report includes RAC that when satisfied, support a finding that the fuel is qualified for use (i.e., reasonable assurance exists that the fuel, fabricated in accordance with its specification, will perform as described in the safety analysis). Specifically, the fuel design criteria and associated limits must ensure four key objectives: 1) the fuel system is not damaged as a result of normal operation and AOOs, 2) the number of fuel pin failures is not underestimated for postulated accidents, 3) coolability is always maintained, and 4) fuel system damage is never so severe during postulated accidents as to prevent reactivity control and control rod insertion when it is required.

The objective of the Natrium Reactor fuel qualification plan is to confirm that all aspects of the fuel system design and fabrication process will provide reliable and safe operation of a commercial sodium-cooled, fast-neutron spectrum nuclear reactor. This document provides information to the NRC to qualify the fuel for the Natrium Reactor. NRCs review and approval are requested for the following:

The identified acceptance criteria are adequate to support fuel qualification.

The identified key fuel manufacturing parameters are adequate to support fuel qualification.

The identified evaluation methods and models are adequate to support fuel qualification.

The use of legacy data and the planned testing is adequate to provide the necessary information to qualify the fuel.

The plans for inclusion of small subsets of fuel pins that operate outside the performance envelope of the bulk of the core, or that feature advanced design features, are acceptable.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054

2. BACKGROUND 2.1 Design Background TerraPowers Natrium Reactor is a sodium-cooled fast reactor (SFR) that uses a fuel design and an operating environment that are significantly different from light water reactors currently utilized in the United States. The Natrium Reactor is an innovative design that facilitates rapid construction and achieves cost competitiveness and flexible operations through the adoption of new technology and a reimagined plant layout. Many of these advances are enabled through inherent safety features of pool-type SFRs [5]. The Natrium Reactor design is based on early reactor technology developed in the US by the Department of Energy (DOE) and was developed from decades of research, design, and development from GE-Hitachis (GEH) Power Reactor Innovative Small Module (PRISM) technology and TerraPowers Traveling Wave Reactor (TWR) technology. The nuclear heat source is a pool-type sodium fast reactor design with sodium-bonded uranium-10wt% zirconium (U-10Zr) fuel clad in HT9 stainless steel. The reactor operates at about atmospheric pressure, circulating sodium through its core, with heat transferred from the primary sodium to an intermediate sodium loop. The Natrium design uses sodium-bonded metallic fuel consistent with the HT9-clad fuel used successfully in both EBR-II and the FFTF (see Figure 2-1).

Figure 2-1. Overview of Natrium Reactors Safety Features Despite the advanced design, in many respects, the Natrium Reactor takes an incremental approach to design and licensing. The Natrium Reactor will begin operation with Type 1 fuel that is similar in composition and dimensions to the fuel pins reliably used in the EBR-II and the FFTF. The operating conditions such as temperature, mechanical loads and burnup are also within the experience base obtained with previous SFRs. The core has the capability to safely irradiate fuel pins to higher burnup in specific core locations in Lead Demonstration Assemblies (LDAs). The surveillance and sampling of fuel pins in the LDAs will provide data to extend the residence time and burnup of Type 1 fuel pins. The core will also have Lead Test Assemblies (LTAs) to gain operating experience with

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Type 1B fuel. Type 1B fuel incorporates advanced features ((

))(a)(4)(ECI) that will enable significantly higher fuel burnups and improved fuel utilization to reduce refueling. This document focuses largely on the Type 1 fuel design but will touch on the LDA and LTA designs due to their importance to the fuel surveillance program as well as supporting rapid transition to more advanced fuel designs.

2.2 Regulatory Background TerraPower submitted the Advanced Fuel Qualification Methodology (AFQM) Report to the NRC on July 16, 2020 (ML20209A155) [6]. The AFQM Report describes methodologies, regulatory criteria, and qualification criteria for metallic fuel for SFRs. The NRC provided feedback in a November 19, 2020 letter, NRC Feedback Regarding TerraPower White Paper Advanced Fuel Qualification Methodology Report-Regulatory Guidance Development Report (EPID No.: L-2020-LRO-0045)

(ML20310A278) [7]. The NRCs feedback has informed the development of TerraPowers overall approach to fuel qualification.

TerraPower submitted the Advanced SFR Fuel Assembly Qualification Plan to the NRC by letter dated November 11, 2020 (ML20316A038). The NRC provided feedback in a May 4, 2021 letter, U.S. Nuclear Regulatory Commission Feedback Regarding TerraPower, LLCs Advanced Sodium Fast Reactor Fuel Assembly Qualification Plan (EPID NO.: L-2020-LRO-0080) (ML21099A081) [8].

The NRCs feedback has informed the development of TerraPowers fuel qualification efforts for the integrated fuel assembly.

By letter dated February 26, 2021 (ML21057A008), TerraPower submitted the Advanced SFR Type 1 Fuel Pin Qualification Plan to the NRC [9]. The NRC provided feedback in a July 13, 2021 letter, TerraPower, LLC - U.S. Nuclear Regulatory Commission Staff Feedback Regarding White Paper, Advanced SFR Type 1 Fuel Pin Qualification Plan, Revision 0 (EPID NO.: L-2021-LRO-0008) (ML21147A548) [10]. The NRCs feedback has informed the development of TerraPowers fuel qualification efforts for fuel pin specific aspects.

TerraPowers fuel qualification efforts have been informed by the NRCs feedback as described above, as well as NRC guidance including RG 1.206, Section C.1.4, Reactor, NUREG-0800, Section 4.2, Fuel System Design, [2] and NUREG-2246, Fuel Qualification for Advanced Reactors. [3]

TerraPowers fuel qualification efforts began, in part, by identifying RAC that were developed using the guidance of RG 1.206 and NUREG-0800, with adaption as necessary due to the differences from light water reactor technology. Subsequently, the NRC issued NUREG-2246, Fuel Qualification for Advanced Reactors, which includes fuel qualification assessment framework (FQAF) goals. Table 2-1 provides a cross-reference between the TerraPower developed/identified RAC and the NUREG-2246 FQAF goals, identifying which RAC are used to address specific FQAF goals. In several cases, FQAF goals are addressed by design specifications as identified in Table 2-1.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 2-1. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Appendix A Goals in FQAF FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description G1 Fuel is manufactured in accordance with a specification 4.2-5 The fuel system description and design drawings shall provide information necessary to verify that the fuel system design bases are met.

G1.1 Key dimensions and tolerances of fuel components are specified 4.2-5 Relevant key dimensions and tolerances of fuel components are specified in design drawings.

Section 5.6 of this report summarizes applicable drawings/drawing types that will be the primary sources for specifying the key dimensions and tolerances identified by RAC 4.2-5.

G1.2 Key constituents are specified with allowance for impurities 4.2-5 Relevant key constituents with allowance for impurities are specified in fuel, material, and product specifications. Section 5.6 of this report summarizes applicable specifications that will be the primary sources for specifying the key constituent and impurity limits identified by RAC 4.2-5.

G1.3 End state attributes for materials within fuel components are specified or otherwise justified 4.2-5 End state attributes (i.e., microstructure, heat treatments, and specific manufacturing processes) are specified in fuel, material, and product specifications. Section 5.6 of this report summarizes applicable specifications that will be the primary sources for specifying the key end state attributes identified by RAC 4.2-5.

G2 Margin to safety limits can be demonstrated 4.2-1, 4.2-2, 4.2-3, 4.2-4 Design criteria and evaluation methods are described below for the subgoals of G2.

G2.1 Margin to design limits can be demonstrated under conditions of normal operation and AOOs 4.2-1 Fuel system damage criteria shall be established for normal operation, including AOOs, to ensure that fuel system dimensions remain within operational tolerances and that functional capabilities are not reduced below those assumed in the safety analysis.

G2.1.1 Fuel performance envelope is defined 4.2-1.1 Stress, strain, or loading limits for all fuel system components shall be established.

4.2-1.2 The cumulative number of strain fatigue cycles on all fuel system components shall be significantly less than the design fatigue lifetime.

4.2-1.3 Limits on fretting wear at contact points on all fuel system components shall be established or alternatively, impacts of fretting wear shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by fretting wear.

4.2-1.4 Limits on erosion and corrosion shall be established for all fuel system components or

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description alternatively, impacts of erosion and corrosion shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by erosion and corrosion.

4.2-1.5 Limits on internal cladding damage (wastage) due to fuel-cladding chemical interaction (FCCI) with fuel or absorber-cladding chemical interaction (ACCI) for absorber components shall be established or alternatively, impacts of wastage shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by wastage.

4.2-1.6 Limits on fuel dimensional changes, such as fuel pin bowing, assembly duct bowing, pin swelling, and assembly duct dilation, shall be established to ensure that fuel, reflector, and shield assembly dimensions remain within operational tolerances or to prevent a situation where thermal hydraulic or neutronic design limits are exceeded.

4.2-1.7 Limits on dimensional changes, such as absorber pin bowing, control assembly duct bowing, absorber pin swelling, and assembly duct dilation, shall be established to ensure that reactivity control assembly dimensions remain within operational tolerances and to prevent interference that may impact control rod insertability.

4.2-1.8 Design limits on fuel pin and reactivity control absorber pin internal pressure for normal operation and AOOs shall be established or alternatively, pin internal pressure shall be explicitly assessed in analyses demonstrating compliance with fuel system damage criteria that may be affected by pin internal pressure.

4.2-1.9 The worst-case hydraulic loads for normal operation and AOOs shall not exceed the hold-down capability of a fuel, reflector, or shield assemblies.

4.2-1.10 The worst-case hydraulic loads for normal operation and AOOs shall not exceed the hold-down capability of a reactivity control assembly.

4.2-1.11 Design limits for the mechanical and neutronic lifetimes for reactivity control assemblies shall be established to ensure that control rod reactivity and insertability are maintained.

4.2-1.12 Design temperature limits on fuel system components for normal operation and AOOs

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description shall be established, or alternatively, peak temperature shall be explicitly assessed in analyses demonstrating compliance with fuel system damage criteria that may be affected by temperature.

G2.1.2 Evaluation model is available (see EM Assessment Framework) 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of evaluation models in Section 6.3.2.2.

G2.2 Margin to radionuclide release limits under accident conditions can be demonstrated 4.2-2 Fuel pin failure criteria shall be established that ensure that the number of fuel pin failures cannot be underestimated for all failure mechanisms that may result in the loss of fuel integrity (cladding breach) during normal operation, AOOs, and postulated accidents.

G2.2.1 Radionuclide retention requirements are specified Radionuclide retention requirements will be described in Chapter 2, Methodologies and Analysis, of the Natrium Preliminary Safety Analysis Report (PSAR). Fuel failure criteria and fuel performance methods, which are used to demonstrate that radionuclide retention requirements are met, will also be described in Chapter 2 of the PSAR.

G2.2.2 Criteria for barrier degradation and failure are suitably conservative (a)

Criteria are conservative 4.2-2.1 Fuel system design limits shall be established and used for the prediction of fuel pin failure due to overheating of the cladding or alternatively, fuel pin failure due to overheating of the cladding shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by fuel pin overheating of the cladding.

4.2-2.2 Fuel system design limits shall be established and used for the prediction of fuel pin failure due to overheating of the fuel slug or alternatively, fuel pin failure due to overheating of the fuel slug shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by overheating of the fuel slug.

4.2-2.3 Fuel system design limits shall be established and used for the prediction of fuel pin failure (loss of cladding integrity) due to deformation of the cladding from mechanical loads or alternatively, deformation of the cladding from mechanical load shall be explicitly assessed in

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description analyses demonstrating compliance with fuel failure criteria that may be affected by deformation of the cladding.

4.2-2.4 Fuel system design limits shall be established and used for the prediction of fuel pin failure (loss of cladding integrity) due to mechanical fracturing from externally applied forces.

4.2-2.5 Fuel system design limits established and used for the prediction of fuel pin failure (loss of cladding integrity) shall address the effects of cladding wastage or alternatively, cladding wastage shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by cladding wastage.

(b) Experimental data are appropriate (see ED Assessment Framework) 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of test data in 6.2 and 6.3.

G2.2.3 Radionuclide retention and release from fuel matrix are modeled conservatively (a)

Model is conservative (b) Experimental data are appropriate (see ED Assessment Framework) 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of test data in 6.2 and 6.3.

G2.3 Ability to achieve and maintain safe shutdown is assured 4.2-4 Reactivity control assembly criteria shall be established for all damage mechanisms that may occur during postulated accidents to ensure that control rods can be fully inserted when required.

G2.3.1 Coolable geometry is ensured 4.2-3 Fuel assembly criteria shall be established for all damage mechanisms that may occur during postulated accidents to ensure that the fuel assembly geometry retains adequate coolant flow channels to permit removal of residual heat.

(a) Criteria to ensure coolable geometry are specified 4.2-3.1 Fuel system design limits shall be established to ensure that cladding stress and strain during postulated accidents do not result in significant cladding damage that might prevent adequate core cooling or alternatively, cladding stress and strain during postulated accidents shall be explicitly assessed in analyses demonstrating

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description compliance with fuel coolability criteria that may be affected by cladding stress and strain during postulated accidents.

4.2-3.2 The maximum temperature of the cladding during postulated accidents shall be less than the melting temperature of the cladding.

4.2-3.3 Evaluations of fuel assembly temperatures to demonstrate core coolability must account for the effects on core flow distribution and the potential for flow blockage caused by ballooning (i.e., swelling) of the cladding during postulated accidents.

4.2-3.4 The maximum temperature of the fuel slug during postulated accidents shall be less than the melting temperature of the fuel.

4.2-3.5 Structural deformation of fuel assembly components due to the combined loads from accident conditions and natural phenomena shall not prevent the ability to adequately cool the core during postulated accidents.

4.2-3.6 Hydraulic loads, when combined with loads from natural phenomena, shall not unseat a fuel, reflector, or shield assembly and cause a reduction in coolant flow that could prevent the ability to adequately cool the fuel assembly during postulated accidents.

(b) Evaluation models are available (see EM Assessment Framework 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of evaluation models in Section 6.3.2.2.

G2.3.2 Negative reactivity insertion can be demonstrated 4.2-4 Reactivity control assembly criteria shall be established for all damage mechanisms that may occur during postulated accidents to ensure that control rods can be fully inserted when required.

(a) Criteria are provided to ensure that negative reactivity insertion is not obstructed 4.2-4.1 Structural deformation of control assemblies due to the combined loads from accident conditions and natural phenomena shall not prevent the ability to insert control rods during postulated accidents.

4.2-4.2 Hydraulic loads, when combined with loads from natural phenomena, shall not unseat a reactivity control assembly that could prevent the complete insertion of control rods during postulated accidents.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID FQAF Goal Description RAC #

RAC / Design Specification Description (b) Evaluation model is available (see EM Assessment Framework) 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of evaluation models in Section 6.4 As can be seen in Table 2-1, FQAF goals are addressed directly by RAC with the exception of FQAF goal G2.2.1 Radionuclide Retention Requirements are Specified. Specific radionuclide retention requirements will be specified in Chapter 2, Methodologies and Analysis, of the Natrium PSAR, with the fuel failure criteria and fuel performance methods demonstrating that the radionuclide retention requirements are met. In addition to the overall FQAF goals, NUREG-2246 identifies Assessment Framework goals for evaluation models as well as for supporting experimental data.

These goals and their correspondence to associated RAC/Design Specifications are summarized in Table 2-2 and Table 2-3, respectively. All of the evaluation model and experimental data assessment framework goals are addressed by RAC 4.2-6. RAC 4.2-6 specifies that: Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Design evaluations are further clarified to include operating experience, testing, and analytical predictions. The Compliance Specific Considerations of RAC 4.2-6 provide significantly more detail on expectations for acceptable design evaluation methods, but a high-level prescribed expectation is that they apply conservative treatment of uncertainties in the values of important parameters. Because RAC 4.2-6 does not explicitly address all of the Evaluation Model and Experimental Data Assessment Framework goals, more details are provided relative to plans to address the goals in Section 6, where the plans for fuel system design evaluation are discussed.

Table 2-2. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Evaluation Model Assessment Framework Goals FQAF Goal ID Evaluation Model Assessment Framework Goal Description RAC #

RAC / Design Specification Description EM G1 Evaluation model contains the appropriate modeling capabilities 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of evaluation models in Section 6.3.2.2.

EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system EM G1.2 Evaluation model is capable of modeling the material properties of the fuel system EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel performance EM G2 Evaluation model has been adequately assessed against experimental data

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID Evaluation Model Assessment Framework Goal Description RAC #

RAC / Design Specification Description EM G2.1 Data used for assessment are appropriate (see ED Assessment Framework)

EM G2.2 Evaluation model is demonstrably able to predict fuel failure and degradation mechanisms over the test envelope EM G2.2.1 Evaluation model error is quantified through assessment against experimental data EM G2.2.2 Evaluation model error is determined throughout the fuel performance envelope EM G2.2.3 Sparse data regions are justified EM G2.2.4 Evaluation model is restricted to use within its test envelope Table 2-3. TerraPower Identified/Developed RAC Mapped to NUREG-2246 Experimental Data Assessment Framework Goals FQAF Goal ID Experimental Data Assessment Framework Goal Description RAC #

RAC / Design Specification Description ED G1 Assessment data are independent of data used to develop/train the evaluation model 4.2-6 Design evaluations shall be performed using acceptable methods to demonstrate that the fuel system design bases are met during conditions of normal operation, AOOs, and postulated accidents. Section 6 provides more details on the approach to design evaluations with specific discussion of evaluation models in Section 6.3.2.2.

ED G2 Data has been collected over a test envelope that covers the fuel performance envelope ED G3 Experimental data have been accurately measured ED G3.1 The test facility has an appropriate quality assurance program ED G3.2 Experimental data are collected using established measurement techniques ED G3.3 Experimental data account for sources of experimental uncertainty ED G4 Test specimens are representative of the fuel design ED G4.1 Test specimens are fabricated consistent with the fuel manufacturing specification ED G4.2 Distortions are justified and accounted for in the experimental data

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054

3. DISCUSSION Pool-type SFRs were constructed as early as 1951 (e.g., Experimental Breeder Reactor-I in the U.S.A.) and the fuel system and fuel pin designs evolved based on operating experience and transient testing until the early 1990s. During the fast reactor development programs, fuel pins fabricated with various fuel-cladding material combinations and over a range of dimensions were shown to have excellent reliability. Well over 100,000 metallic fuel pins were used to run fast reactors with an exceptionally low failure rate. Excellent transient behavior was also demonstrated for fuel pins both using the EBR-II to run full core transients that represent accident scenarios and the Transient Reactor Test (TREAT) facility to run overpower transients up to (and even exceeding) four times the nominal power (i.e., 400% nominal power).

A Regulatory Compliance Plan (RCP) has been developed that identifies and adapts the regulatory requirements that are described in Section 4.2 of the Standard Review Plan (NUREG-0800), which is devoted to the fuel system design [11]. This adaptation did not only update the terminology to be more suited for metallic fuel in SFRs but also included specific phenomena of concern for this fuel system based on extensive review of the available data and historic operating experience. The referenced RCP [11] also establishes RAC to ensure compliance with the identified regulatory requirements. Fuel and absorber design criteria and associated limits and bases were provided in the Natrium Fuel, Control, Shield, and Reflector Pin Design Basis [12] to ensure compliance of fuel and absorber pin designs with the established RAC. A key fuel testing need is to demonstrate the suitability of the established fuel design criteria and limits to prevent damage and/or failure, maintain coolability of the core during and after all licensing basis events (LBEs), and assure fuel system damage during postulated accidents will not prevent reactivity control rod insertion when required. Extensive review of public and non-public data has been performed when establishing these fuel design criteria and limits.

References to some of the compilations of data used to establish these criteria and limits are included in Section 6, with additional proposed activities to address any gaps summarized in Section 6.3.

Beyond establishing the proper fuel design criteria, demonstrating compliance of fuel with these design criteria for the applicable operating domains is another important task to support licensing. Due to the inherently complex nature of nuclear fuels, multiple physical phenomena must be adequately modeled to provide reliable predictions of fuel pin behavior. Test data are required over applicable ranges for high-importance phenomena to validate sufficient understanding of these phenomena and overall reliability of the associated fuel models.

This assessment is organized to be roughly consistent with the RCP [11], capturing applicable information and planned activities to address the key areas of review for nuclear fuel system designs (see Figure 3-1 for a flow chart). Specifically, 1) Fuel Design Criteria, 2) Fuel System Description, 3)

Design Evaluation, 4) Testing and Inspection of New Fuel, 5) Online Fuel System Failure Monitoring, and 6) Post Irradiation Surveillance Plans will be addressed. The primary emphasis will be on the Fuel Design Criteria and Design Evaluation aspects of the RCP since they are the most dependent on testing support prior to the startup of an advanced reactor. The more detailed test plans developed in subsequent efforts will evaluate the applicable operating range of the Natrium Reactor, available applicable data, additional data needed to cover the most adverse conditions anticipated, and the number of data points required to reduce associated uncertainties to acceptable levels.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 3-1. Overall Fuel Qualification Assessment Logic Flow

4. FUEL DESIGN CRITERIA Fuel design criteria must be set to achieve four key objectives: 1) the fuel system is not damaged as a result of normal operation and AOOs, 2) the number of fuel pin failures is not underestimated for postulated accidents, 3) coolability is always maintained, and 4) fuel system damage is never so severe during postulated accidents as to prevent reactivity control rod insertion when it is required. As stated above, a key testing and analysis need is to demonstrate the suitability of the established fuel design criteria and associated limits. To help ensure adequate coverage of each of the established fuel design criteria, Table 4-1 through Table 4-4 summarize the existing RAC and associated design criteria. Table 4-1 covers design basis criteria to prevent fuel damage; Table 4-2 addresses criteria to predict fuel failure; Table 4-3 outlines criteria to maintain fuel coolability; and Table 4-4 addresses criteria to ensure reactivity control insertability. Additional testing activities have been identified to supplement the currently available data to further justify established design basis limits, but these are discussed in Section 6.3, where the applicable Testing Activities are summarized.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 4-1. Design Criteria to Prevent Fuel System Damage-Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-1.1 Stress, strain, or loading limits for all fuel system components shall be established.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.2 The cumulative number of strain fatigue cycles on all fuel system components shall be significantly less than the design fatigue lifetime.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.3 Limits on fretting wear at contact points on all fuel system components shall be established or alternatively, impacts of fretting wear shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by fretting wear.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.4 Limits on erosion and corrosion shall be established for all fuel system components or alternatively, impacts of erosion and corrosion shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by erosion and corrosion.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-1.5 Limits on internal cladding damage (wastage) due to fuel-cladding chemical interaction (FCCI) with fuel or absorber-cladding chemical interaction (ACCI) for absorber components shall be established or, alternatively, impacts of wastage shall be explicitly assessed when demonstrating compliance with fuel system damage criteria that may be affected by wastage.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.6 Limits on dimensional changes, such as fuel pin bowing, assembly duct bowing, pin swelling, and assembly duct dilation, shall be established to ensure that fuel, reflector, and shield assembly dimensions remain within operational tolerances or to prevent a situation where thermal hydraulic or neutronic design limits are exceeded.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.7 Limits on dimensional changes, such as absorber pin bowing, control assembly duct bowing, absorber pin swelling, and assembly duct dilation, shall be established to ensure that reactivity control assembly dimensions remain within operational tolerances and to prevent interference that may impact control rod insertability.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.8 Design limits on fuel pin and reactivity control absorber pin internal pressure for normal operation and AOOs shall be

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria established or alternatively, pin internal pressure shall be explicitly assessed in analyses demonstrating compliance with fuel system damage criteria that may be affected by pin internal pressure.

((

))(a)(4)

((

))(a)(4) 4.2-1.9 The worst-case hydraulic loads for normal operation and AOOs shall not exceed the hold-down capability of a fuel, reflector, or shield assembly.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.10 The worst-case hydraulic loads for normal operation and AOOs shall not exceed the hold-down capability of a reactivity control assembly.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.11 Design limits for the mechanical and neutronic lifetimes for reactivity control assemblies shall be established to ensure that control rod reactivity and insertability are maintained.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-1.12 Design temperature limits on fuel system components for normal operation and AOOs shall be established, or alternatively, peak temperature shall be explicitly assessed in analyses demonstrating compliance with fuel system damage criteria that may be affected by temperature.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

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((

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((

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((

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((

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((

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 4-2. Design Criteria to Prevent Fuel System Failure Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-2.1 Fuel system design limits shall be established and used for the prediction of fuel pin failure due to overheating of the cladding or alternatively, fuel pin failure due to overheating of the cladding shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by fuel pin overheating of the cladding.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.2 Fuel system design limits shall be established and used for the prediction of fuel pin failure due to overheating of the fuel slug or alternatively, fuel pin failure due to overheating of the fuel slug shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by overheating of the fuel slug.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.3 Fuel system design limits shall be established and used for the prediction of fuel pin failure (loss of cladding integrity) due to deformation of the cladding from mechanical loads or, alternatively, deformation of the cladding from mechanical load shall be explicitly assessed in analyses demonstrating compliance with fuel failure criteria that may be affected by deformation of the cladding.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.4 Fuel system design limits shall be established and used for the prediction of fuel pin failure (loss of cladding integrity) due to mechanical fracturing from externally applied forces.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.5 Fuel system design limits established and used for the prediction of fuel pin failure (loss of cladding integrity) shall address the effects of cladding wastage or alternatively, cladding wastage shall be explicitly assessed in analyses demonstrating

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria compliance with fuel failure criteria that may be affected by cladding wastage.

))(a)(4)

Table 4-3. Design Criteria to Ensure Fuel Coolability Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-3.1 Fuel system design limits shall be established to ensure that cladding stress and strain during postulated accidents do not result in significant cladding damage that might prevent adequate core cooling or alternatively, cladding stress and strain during postulated accidents shall be explicitly assessed in analyses demonstrating compliance with fuel coolability criteria that may be affected by cladding stress and strain during postulated accidents.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-3.2 The maximum temperature of the cladding during postulated accidents shall be less than the melting temperature of the cladding.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-3.3 Evaluations of fuel assembly temperatures to demonstrate core coolability must account for the effects on core flow distribution and the potential for flow blockage caused by ballooning (i.e.,

swelling) of the cladding during postulated accidents.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-3.4 The maximum temperature of the fuel slug during postulated accidents shall be less than the melting temperature of the fuel.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-3.5 Structural deformation of fuel assembly components due to the combined loads from accident conditions and natural phenomena shall not prevent the ability to adequately cool the core during postulated accidents.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-3.6 Hydraulic loads, when combined with loads from natural phenomena, shall not unseat a fuel, reflector, or shield assembly and cause a reduction in coolant flow that could prevent the ability to adequately cool the fuel assembly during postulated accidents.

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))(a)(4)

((

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((

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Table 4-4. Design Criteria to Ensure Reactivity Control Insertability Criteria Specific RAC Acceptance Criterion Fuel Pin Applicable Design Criteria Fuel Assembly Applicable Design Criteria Absorber Pin Applicable Design Criteria Control Assembly Applicable Design Criteria 4.2-4.1 Structural deformation of control assemblies due to the combined loads from accident conditions and natural phenomena shall not prevent the ability to insert control rods during postulated accidents.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-4.2 Hydraulic loads, when combined with loads from natural phenomena, shall not unseat a reactivity control assembly that could prevent the complete insertion of control rods during postulated accidents.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054

5. FUEL DESIGN DESCRIPTION 5.1 Overview of the Fuel Design of the Natrium Reactor The Natrium Reactor is a pool-type, sodium-cooled, fast-spectrum reactor with some design similarities to other SFRs such as EBR-II and the FFTF. The Natrium Reactor core will contain Type 1 fuel at the beginning of life. Natrium Type 1 fuel pins are intentionally similar in design to historically tested designs to leverage available historic operating experience. Specifically, they use U-10Zr, sodium bonding, HT9 cladding, and a nominal smear density of 75%. Type 1 fuel has a cladding diameter similar to fuel pins that were successfully tested in EBR-II (((

))(a)(4). Bundles of Type 1 fuel pins are located in hex-shaped fuel assemblies in the core similar to EBR-II and FFTF; however, a larger number of pins will be used per assembly than was for FFTF or EBR-II. A comparison between fuel pin dimensions from various reactors is given in Table 6-5. Regarding operational conditions, the Natrium fuel operates at low power ((

))(a)(4). More explicit comparison to historic designs and targeted operating conditions are provided in Section 6.2, Historic Operating Experience.

In addition to standard fuel assemblies, the Natrium Reactor core will have Lead Demonstration Assemblies (LDA) that are a crucial component of the fuel surveillance and Lead Test Assembly (LTA) programs. The LDAs are designed to expedite the availability of post-irradiated data on fuel pins by providing removable fuel pins that ((

))(a)(4). The targeted conditions for the accelerated LDA pins will still be bound by historic pin operating experience and fuel performance assessments to verify design limits are met. More detailed discussion of the planned Fuel Surveillance Program is provided in Section 9. Although the primary purpose of the LDAs is to support Type 1 fuel surveillance, ((

))(a)(4).

The basic nuclear control component of the Natrium Reactor core is the Control Assembly which contains absorber pins. The absorber pins contain cladded, helium-bonded, boron carbide (B4C) absorber pellets that can be adjusted in the axial direction during operation by the Control Rod Drive Mechanisms (CRDMs). The Secondary Control Assembly is a secondary reactivity control component in the Natrium Reactor that is used to provide defense in depth relative to common cause failure of absorber bundle to duct binding to address PDC 26 [1].1 The Secondary Control Assembly is also composed of absorber pins grouped into Control Assemblies, ((

))(a)(4) 1 Principal Design Criteria (PDC) 26 specifies the need for independent and diverse means capable of controlling the rate of reactivity changes resulting from planned, normal power changes to assure that the design limits for the fission product barriers are not exceeded.

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((

))(a)(4). More detailed descriptions of the fuel and control assembly designs are provided in Sections 5.2 through 5.4. The Natrium Reactor is currently completing the conceptual design phase therefore some of these details may evolve as more in-depth analysis and testing is performed as part of preliminary design.

5.2 Fuel Assemblies 5.2.1 Fuel Pin The Natrium fuel pin is comprised of a cladding tube, an upper and lower end cap, wire wrap, sodium-bonded fuel column, fission gas plenum, tag gas capsule, and axial shield (Figure 5-1).

The cladding tube and end caps are welded on each end to provide the structural support and hermetic sealing for the contained components.

Figure 5-1. Natrium Type 1 Fuel Pin The Natrium Type 1 fuel is metallic uranium alloyed with 10 wt. % zirconium (U-10Zr). The fuel column section of the pin consists of a stack of right circular cylinder fuel slugs. The individual fuel slug lengths are partially influenced by the manufacturer and their optimal process efficiency and capability (within the limits of the fuel specification). The as-manufactured fuel slugs have cross sectional dimensions that represent ~75% of the internal cross-sectional area of the cladding (i.e., 75% smear density). Radiation-induced swelling of the fuel slug will increase its volume such that it contacts the cladding tube inner surface within the first few percent of burnup.

The extra space is provided to help ensure interconnected porosity develops in the fuel to (a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 promote release of fission gases to the plenum to preclude undue strain on the cladding from fuel-clad mechanical interaction as the fuel continues to generate fission products and swell.

A liquid metal sodium bond is employed in the Natrium Type 1 fuel pin and is initially located in the space between the fuel and cladding. The sodium bond enables adequate heat transfer and prevents unacceptable temperatures during operation, especially at beginning of life when the fuel is not in physical contact with the cladding tube. Once the fuel swells, the liquid metal sodium bond is pushed into the upper plenum although a small amount remains within the porosity of the fuel slugs.

Each fuel pin is helically wrapped with HT9 wire to provide lateral pin-to-pin and pin-to-duct spacing along its length and to promote coolant mixing throughout the assembly. The wire is wrapped under a tensile load. The wire is terminated at each end of the pin ((

))(a)(4)

Figure 5-2. Wire Wrap Fused Ball Termination The fuel pins have an axial shield section below the fuel column that provides neutron attenuation to limit the damage to the Core Support Structure (CSS) to acceptable levels. The axial shielding is comprised ((

))(a)(4) located within the sealed pin volume below the fuel column. The shield slug ((

))(a)(4). A fission gas plenum is provided above the fuel and sodium bond to limit internal gas pressure buildup caused by gaseous fission product generation. It is initially backfilled with inert gas.

5.2.2 Fuel Assembly The fuel assembly is the basic nuclear power generating component of the Natrium Reactor core.

It contains the fuel, produces heat, and provides the neutron flux. It can be removed from and replaced or shuffled in the core during reactor refueling. The fuel assembly is principally designed to:

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 position the fuel properly in the core for controlled nuclear reaction and generation of thermal power; provide passages to guide and control the sodium coolant for heat removal; provide shielding to protect components of the CSS from excessive fluence; provide features for proper interfacing with other core components, the CSS, the In-Vessel Handling Machines (IVHMs), ((

))(a)(4) and, provide a physical barrier, the assembly duct, between fuel pins of adjacent fuel assemblies and control assemblies to mitigate or isolate fuel performance impacts on neighboring assemblies The Natrium fuel assembly is shown in Figure 5-3 and Figure 5-9). It is approximately ((

))(a)(4)(ECI) total length and comprised of an inlet nozzle, a hexagonal duct tube with above core load pads, a handling socket with top load pads, and a fuel pin bundle with its attachments.

Figure 5-3. Natrium Fuel Assembly Design 5.2.2.1 Duct and ((

))(a)(4)

The hexagonal duct is the principal structural member of the fuel assembly. The fuel assembly duct mates with the handling socket at the top of the assembly, extends the full length, and mates with the inlet nozzle at the bottom of the assembly. The duct tightly encloses the fuel pin bundle along the full length and guides the coolant flow through the bundle, thus permitting an individual assembly orificing scheme that is based on core position. An ACLP is located on the duct approximately two-thirds up from the duct to inlet nozzle connection as shown in Figure 5-3.

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 The normal duct wall thickness is increased seamlessly at the ACLP for structural support to transmit loads among assemblies and eventually to the core former ring mounted on the in-vessel storage. The ACLP maintains inter-assembly spacing. Between the normal duct thickness and increased thickness at the ACLP, a shallow angle chamfer is provided to minimize withdrawal and insertion loads and to reduce the potential for mechanical binding as the ACLPs move past adjacent fuel assemblies during refueling.

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-4. Inlet Nozzle with Duct ((

))(a)(4)

Figure 5-5. ((

))(a)(4) 5.2.2.2 Handling Socket The handling socket, which functions as the TLP in core restraint, is located at the upper end of the fuel assembly and mates with the various grapples during fuel handling operations. The handling socket also guides coolant into the hot pool and into the UIS (for the assemblies under the UIS) during reactor operation and provides spacing and load transfer through hard-face coated load pads (the TLPs) that interface with adjacent core assemblies and the core former ring of the CRS. ((

))(a)(4)

(a)(4)(ECI)

(a)(4)

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((

))(a)(4). Fuel assemblies may require rotation at selected irradiation times to balance out the irradiation-induced geometrical distortions, and the orientation notch on the handling socket collar provides the reference to achieve the desired amount of rotation.

Figure 5-6. Handling Socket ((

))(a)(4) 5.2.2.3 Inlet Nozzle and Pin Attachment Hardware The inlet nozzle, located at the lower end of the fuel assembly, interfaces with, and provides the primary load path of the fuel assembly to the CSS. It provides the coolant inlet flow path to the assembly internals and contains orifice plates to modify the total flow within the assembly based on the location within the core. The inlet nozzle also interacts with the receptacle to create a hydraulic hold down force on the assembly from the pressure difference between the inlet plenum and low-pressure plenum. ((

))(a)(4) 1 Specific notching scheme shown in figure is an example for illustrative purposes.

(a)(4) (ECI)

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((

))(a)(4)

Figure 5-7. Inlet Nozzle ((

))(a)(4) Cross Section Assembly bypass coolant flow is minimized ((

))(a)(4)

((

))(a)(4)

(a)(4) (ECI)

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((

))(a)(4)

Primary vertical loads on the assembly are carried through the nozzle, then transferred to the CSS receptacle. Horizontal loads are transferred to the core structure in two locations: at the upper section of the nozzle, horizontal loads are transferred to the upper grid plate radially through a centering collar; in the lower section, mating features in the nozzle transfer lateral loads to the receptacle cylinder, which carries the load through a moment couple to the lower grid plate. The inlet nozzle has a conical transition from the overall assembly hexagonal shape to the nozzle circular shape. This conical feature provides self-alignment capability during fuel assembly handling operations. The nozzle has machined slot features located at the top that mate with the pin attachment hardware ((

))(a)(4). In turn, the pin attachment hardware mates with the lower end cap of each fuel pin and provides axial restraint and support (Figure 5-8).

The fuel assembly in general, and the inlet nozzle in particular, have design features that permit liquid sodium to drain from all internal volumes to minimize sodium residuals when removed from the core and ex-vessel storage prior to Post-Irradiation Examination (PIE) shipment or storage as spent fuel. The fuel assembly is also designed to be washed in the Sodium Removal System to remove all residuals prior to placement in long-term spent fuel storage.

Figure 5-8. Fuel Pin Attachment Hardware for a ((

))(a)(4)(ECI) Pin Bundle 5.2.2.4 Fuel Pin Bundle The fuel pin bundle is shown in Figure 5-9 with its hexagonal cross section. Each fuel assembly contains ((

))(a)(4)(ECI) sealed fuel pins packed with triangular pitch spacing. The pins extend from their primary attachment point near the top of the inlet nozzle to just below the handling socket at the top of the assembly. The bundle is comprised of ((

))(a)(4) separate strip layers varying from ((

))(a)(4) pins per layer (see Figure 5-8). The upper and lower end caps of the wire-wrapped fuel pins in all of the strip layers have the same orientation to ensure uniform coolant flow across the bundle and proper fit with all attachment hardware. A very tight fit of the (a)(4) (ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 fuel pin bundle in the duct is important to achieve proper coolant flow so that the coolant is guided to the bundle internals and a minimum amount of coolant flows through the bypass region between the bundle periphery and inner duct wall. Accordingly, the bundle is fixtured and slightly compressed during assembly prior to installing the duct over the bundle, targeting a maximum bundle clearance to the inner duct on the order of the thickness of ((

))(a)(4).

Figure 5-9. Natrium ((

))(a)(4)(ECI) Fuel Pin Bundle Cross Section in Fueled Region1 5.2.3 Lead Demonstration Assembly In an effort to mitigate licensing risks, the Type 1 fuel design has leveraged historic fuel designs (e.g., FFTF and PRISM Mod B) and operating targets as much as practical. This enables application of historic fuel operating experience to support fuel qualification. Moreover, a fuel surveillance program within the Natrium plant will be established to address any potential gaps in the fuel qualification program that are not adequately covered by historic experience or readily addressed with new test data.

Given the establishment of this fuel surveillance program, the core will have the capability to irradiate fueled Lead Demonstration Assemblies (LDAs) that support rapid post-irradiation exams of fuel pins. These LDAs will be designed to have unique features and components that allow for remote disassembly and removal of a select number of irradiated fuel pins. Following removal from the assembly, the selected fuel pins will then be examined to gather data to reduce fuel performance uncertainties and potentially increase fuel burnup targets. The LDA will be designed to be very similar to that of the standard Type 1 fuel assembly. It will have the same assembly height, employ a subset of the same standard fuel pins, fit within the same hexagonal grid as the standard fuel assembly, and the structural members and external configuration of the LDA will be 1 The duct wall is shown crosshatched and the ACLP is revealed at the outer edge due to its larger diameter (a)(4) (ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 largely identical to those of the fuel assemblies. Additionally, LDAs will be designed to be hydraulically and thermally compatible with the other fuel assemblies, i.e., to have compatible pressure drop and to meet the cladding and fuel temperature limits. ((

))(a)(4)

The unique aspects of the LDAs involve a select number of positions in the fuel bundle that contain removable pins and the associated pin retention components. The LDAs will require unique components to enable remote disassembly for extraction of the removable pins. ((

))(a)(4) 5.2.3.1 LDA Handling Socket Traceability of the position of the removable pins within the LDA bundle will be maintained at all times to ensure the appropriate pins are removed to support post-irradiation exams. To aid in locating the appropriate pins, the handling socket has features that provide orientation relative to the mapping of the removable pins. Initial orientation of the LDA assemblies will be established and recorded during core assembly fabrication, whereby the duct and handling socket will be installed with a prescribed orientation relative to the removable pin mapping. ((

))(a)(4) 5.2.3.2 LDA Fuel Pin Bundle The LDA fuel bundles consist of an array of wire wrapped and non-wire wrapped fuel pins arranged in a tight triangular pitch spacing. As many as ((

))(a)(4) removable pins will be designated in each LDA for pin removal and examination. ((

))(a)(4)

An illustrative example of the location of the removable pins in the assembly is shown in Figure 5-10. The final selected positions will be determined as the fuel surveillance program is further defined.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-10. Potential Removable Pin Locations 5.2.3.3 LDA Removable Fuel Pin and Retention Component The LDA fuel pins extend from their primary attachment point near the top of the inlet nozzle to just below the handling socket. They will extend beyond the rest of the standard pin bundle with a feature on the upper end cap to permit positive gripping with the pin removal tool.

Vertical retention of the pin is provided by a device at the lower end that interfaces with the pin rails and connects to the special lower end cap of the removable pin. ((

))(a)(4)

Excluding the special design of ((

))(a)(4) the removable pins are otherwise identical to the standard fuel pins, i.e., they are fabricated with the same cladding, fuel, sodium bond, axial shield, and plenum as the Type 1 fuel pins, using the same manufacturing specifications. Even though the lower end of the pin is different than a standard fuel pin due to the special retention device, the fuel column axial elevation will be kept level with the rest of the fuel.

5.2.3.4 LDA Secondary Pin Restraint A secondary pin restraint device is incorporated into the LDAs to prevent ejection of any removable pin that has inadvertently lost primary restraint from its retaining socket during reactor operation. The secondary restraint feature is designed to facilitate remote removal during LDA disassembly.

5.2.4 Lead Test Assemblies / Type 1B Fuel The Natrium core has the capability to irradiate fueled Lead Test Assemblies (LTAs). These lead test assemblies have innovative features that allow them to achieve long reactor residence times and high burnup and higher coolant outlet temperatures to improve fuel cycle economics. The LTA program plan for the Natrium core is still under development; however, (a)(4)

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((

))(a)(4) provides a conceptual LTA plan along with additional testing and analysis work that will be used to support Type 1B fuel qualification [13]. The following discussion provides current concepts being considered for the LTA program that may change depending on innovations, advancements or other information that becomes available over time.

The LTA is very similar to the lead demonstration assembly. It has the same assembly height, fits within the same hexagonal grid as the LDA, and the structural members and external configuration of the LTA are identical to those of the fuel assemblies (i.e., same inlet nozzle, handling socket, and duct mechanical joint). Additionally, LTAs are designed to be hydraulically and thermally compatible with standard fuel assemblies, i.e., to have the same pressure drop and to meet the cladding and fuel temperature limits. The unique aspects of the LTA are the LTA fuel pin and the duct material.

5.2.4.1 LTA Duct The LTAs will utilize the same assembly pitch as is employed for the other core assemblies. They will use ((

))(a)(4) in Reference [13].

5.2.4.2 LTA Fuel Pin Bundle The LTA fuel bundle consists of an array of wire-wrapped fuel pins arranged in a tight triangular pitch spacing. Similar to the Type 1 fuel pin bundle, the wire wraps of all the pins in the bundle are oriented exactly the same to ensure uniform coolant flow across the bundle and proper packing and fit-up with all attachment hardware. Like the LDA, ((

))(a)(4) support rapid removal of the pins for subsequent PIE.

5.2.4.3 LTA Fuel Pin The key differences between the LTA fuel pins (Type 1B) and the host core sodium bonded fuel pins (Type 1) are shown in Figure 5-11. ((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-11. Key Differences of Type 1 and Type 1B (LTA) Fuel Pin Design 1 5.3 Control Assemblies The Reactivity Control System is the principal nuclear control system of the Natrium Reactor core and its main function is to position neutron absorber material to appropriately control and terminate the nuclear reaction. This function is consistent with the system requirements for providing safe and predictable operation of the reactor. This system meets reactor shutdown requirements without the aid of any other reactivity control system. A total of ((

))(a)(4) penetrations are provided in the reactor head directly above those core locations, one for each reactivity control unit. The Control Rod Drive Mechanism (CRDM), Control Rod (CR) driveline (CRD), and control assembly are directly coupled during normal operation.

The control rods (located within their own dedicated hexagonal control assemblies in the core) are driven by the CRDM to move and position absorber material vertically within the core to control core reactivity and power to maintain fuel within acceptable design limits. They have the capability to control core reactivity changes during expected operations and specified accident conditions with variations in core composition during operation over the life of the core. The control rods also provide power response for the plant control and data system (PCD). Finally, control rods provide SCRAM insertion capability with sufficient reactivity worth to shut down the reactor and maintain it in cold shutdown even if the highest worth rod is stuck in the withdrawn position. To provide design diversity 1 This image is only illustrative with some design details omitted to simplify comparison.

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 to limit common cause failures there are both primary and secondary control assemblies, ((

))(a)(4).

5.3.1 Absorber Pin The control rod absorber pin is a sealed, helium bonded design that is comprised of a cladding tube, upper and lower end cap, wire wrap, boron carbide pellet column, gas plenum, spring, and plenum spacer. The cladding tube is made of HT9. The cladding tube is hermetically sealed with end caps that are welded on each end and provide the structural supports for the pellet, spring, and spacer components.

Natural boron carbide is used as the absorber material in the pin and is shaped into the form of pellets that are manufactured by pressing and sintering powder into right circular cylinders. The selection of boron carbide as the baseline absorber material is based on its successful and long-standing use in fast reactors around the world, including the extensive irradiation program conducted at the FFTF, and the availability of irradiation data for licensing and qualification.

Compared to other absorber materials, boron carbide has advantages due to its relatively high neutron absorption cross-section, availability and low cost, comparative ease of fabrication, and low radioactivity after irradiation. The pellet-to-clad gap is provided to accommodate pellet swelling that may limit the lifetime of the pin due to strain in the cladding.

The pins have an upper plenum to accommodate the gaseous fission products released from the B4C absorber column during irradiation. The plenum volume is sized such that cladding stresses and strains due to internal gas pressure are maintained to acceptable levels throughout life. The plenum also accommodates any axial absorber column growth. A spring and plenum spacer is provided to ensure that the B4C column maintains its axial position during preoperational shipping and handling and permits axial expansion during operation.

Each of the CR pins is helically wrapped with a wire to provide lateral pin-to-pin and pin-to-duct spacing along its length and to promote coolant mixing throughout the CR. The wire is wrapped under a tensile load. Like the fuel pins, the wire is terminated at each end of the pin ((

))(a)(4).

5.3.2 Primary Control Rod Assembly The primary control rods employ ((

))(a)(4) absorber pins that are arranged in a triangular pitch, packed tightly into a hexagonal lattice, and surrounded by a ((

))(a)(4) HT9 duct (the control rod duct) as shown in Figure 5-12. The control rod duct is the principal structural member for the absorber pin bundle between the upper and lower guide plates.

The CR is designed to move freely within the control assembly duct, with its own dedicated space within the reactor core, throughout its design lifetime. Speed of the control rod insertion in a SCRAM is maintained throughout life accounting for the worst-case distortions, including bowing, misalignment, and friction between the inner and outer duct. The CR duct is welded to the upper and lower guide plates, and to the coupling head that connects the pin bundle to the CRD. Interface wear pads are provided at the top and bottom of the CR at the plate locations to provide a smooth gliding surface against the inner surface of the assembly duct. These wear pads are hard coated to minimize friction and the potential for galling with the control assembly duct. A sufficiently sized gap is provided between the wear pads and the control assembly duct to

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 allow for free motion under all conditions and to accommodate anticipated distortions. Three-point contact along the vertical length of the control assembly duct is precluded by proper design of the interface between the CR and assembly duct, and specifically the CR wear pads and associated gap, assuming worst case distortions. Control assemblies will also be rotated during normal refueling outages to reverse assembly bowing deformations and extend the assembly operational lifetime. The gap size between the inner and outer duct has additional design requirements such that thermal-hydraulic considerations are satisfied (e.g., bypass flow around absorber bundle) and potential reactivity oscillations due to flow induced lateral motion of the control rod are minimized to acceptable levels. Control assemblies are replaced in the core after their design lifetime is achieved during normal reactor refueling. The gap between the inner and outer control ducts can be seen in Figure 5-13.

Figure 5-12. Illustrative Example of a Natrium Control Assembly The absorber pins are attached at the top of the control rod on pin rails that connect via support bars to the upper guide plate, as shown by Figure 5-14. ((

))(a)(4)(ECI) By design, and as a requirement, the CR always decreases core reactivity when inserted incrementally into the core, (a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 even accounting for the effects of absorber material depletion over its design lifetime. At the fully inserted position, or after a SCRAM is accomplished, the control rod B4C column is aligned with the fuel column at their respective centerlines. ((

))(a)(4)(ECI)

Figure 5-13. Cross Section View of Natrium Control Assembly (a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-14. Natrium Control Rod Absorber Pin Attachment Figure 5-15. Control Assembly - Damper and Driveline Assemblies (a)(4)(ECI)

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 5.3.2.1 Duct, ACLP, Handling Socket, TLP, and ((

))(a)(4)

The control assemblies utilize the same HT9 duct, ACLP, handling socket, TLP, and ((

))(a)(4) as described for the fuel assembly in Section 5.2.2.

5.3.2.2 Inlet Nozzle and Axial Shield Block The control assembly inlet nozzle is similar in design to the fuel assembly inlet nozzle except that it does not have features for interfacing with any pin attachment hardware. Instead, it is connected to an axial shield block that provides neutron attenuation to limit the damage to the CSS. The shield block is housed in each control assembly duct located directly above the inlet nozzle. It has machined through-holes to permit the flow of coolant from the inlet nozzle to the internals of the control assembly (Figure 5-16).

Figure 5-16. Control Assembly Lower Detail 5.3.2.3 Control Rod Connection to the Control Rod Drive The CRs are connected and disconnected to the CRD (Figure 5-17) at the coupling head so that the associated drivelines may be lifted above the reactor core assemblies to permit plug rotation.

This disconnect point is also used during a SCRAM so that the CRs drop without connection to the CRD. The control assemblies with the CRs inside are removed from the core by the IVHM in the same manner as fuel assemblies. ((

))(a)(4)

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-17. Natrium Control Rod Connection 5.3.3 Secondary Control Rod Assembly The Secondary Control Rod is still under development to ensure that the geometric differences between the Primary and Secondary control rods are sufficient to meet diversity requirements.

While the Primary Control Rod design is expected to meet the licensing requirements for reliability in reactivity control, the Secondary Control Rod design is intended to create geometric diversity to preclude common cause failures that would inhibit SCRAM or control functionality by mechanical binding phenomena. While there are different forms of mechanical binding, the diversity requirement for the control rods focuses on inner to outer control rod duct interactions that could lead to significantly increased SCRAM time, making the transient impact more severe, or the potential of mechanical binding that would stop control rod insertion (jamming phenomenon).

Currently, the Secondary Control Rod Assembly is anticipated to be nearly identical to the Primary Control rod assembly as outlined in Section 5.3.2. This includes descriptions of pin geometry, assembly configurations, coatings descriptions, etc. The key differences between the Primary and Secondary Control Assemblies are only in the number of absorber pins used, changes to the control assembly geometry, and the space between the inner duct and the guide tube as shown in Figure 5-18. The figure shows the difference between the ((

))(a)(4)(ECI) pin assembly and a ((

))(a)(4)(ECI) pin bundle used in the Secondary Control Rod. The change between the two can be seen clearly by removing the ((

))(a)(4)(ECI). This change opens a significant gap (shown in white) between the guide tube and the inner hex-duct for the control rod. The connection gaps at the dashpot / tie plate at the top of the bundle remain the same. The change to the pin bundle geometry, however, is expected to be significant enough to preclude the previously mentioned common cause mechanical binding failure that could be experienced by either the Primary or Secondary Control Rods as they move within their own dedicated assemblies. The additional space between the (a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 inner and outer duct should also limit the likelihood or amount of force for any inner and outer duct friction contact.

Figure 5-18. Primary vs. Secondary Control Rod Bundle Cross-Section 5.4 Core Restraint System The Core Restraint System (CRS) is an important interface consideration in the design of all core assembly types including both fuel and control assemblies. One of the primary functions of this system is to control radial expansion reactivity feedback that results from assembly displacements within an SFR core. Changes in thermal, irradiation, and mechanical loads at different state-points of reactor operation induce core-wide assembly movement that can cause reactivity changes affecting power. The mechanical (and consequently reactivity) response of the core restraint system is heavily dependent on aspects of core assembly design. As such, a system overview, relevant phenomena, assembly design dependencies, and design targets are described in this section.

The core restraint system consists of core former rings connected to the CSS, assembly top load pads (TLPs), assembly above core load pads (ACLPs), assembly inlet nozzles, and receptacles at the bottom of the CSS. A schematic of these components is shown in Figure 5-19 [14]. The assembly inlet nozzle is inserted into the CSS receptacle, which directs vertical loads to the CSS.

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 5-19. Schematic of a Core Restraint System and Impactful Parameters Core assemblies are subjected to temperature and neutron flux distributions over their residence time. These distributions induce bowing along the axial height of the assembly. This occurs due to differential expansion of opposing sides of the assembly duct component when subjected to a temperature gradient or fluence gradient distribution. For example, an assembly with a temperature gradient across its cross section will deform or bow toward the colder side as shown in Figure 5-20. Additionally, assembly load pads will interact with neighboring assembly load pads or the CSS, which will impose bending deformations on the assembly and cause rotations of the inlet nozzle within the receptacle. This is also illustrated in Figure 5-20 as a restrained thermal bowing case with an example resultant deformation shown. These restraint loads at the assembly ACLP and TLP are imparted by neighboring assemblies and the core support structure, which together comprise the core restraint system as previously described.

Figure 5-20. Contributing Effects to Core Assembly Restrained Thermal Bowing Deformations Nonlinear material effects are an additional important consideration in the design of the core restraint system and the core management program. Over the residence time of an assembly at high temperature, high fluence, and bending loads, assembly structural materials will undergo inelastic strains including thermal creep, irradiation creep, and void swelling. A properly designed

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 core restraint system will account for residual assembly deformations and ensure deformed assemblies can be removed from the core so unirradiated assemblies can be introduced. An example illustrating the importance of this consideration is the EBR-II core which was designed prior to an understanding of void swelling in core structural materials. EBR-II experienced significant refueling challenges due to these inelastic deformations at assembly contact locations, which became highly limiting to assembly residence time. FFTF also experienced refueling challenges with high-deformation assemblies requiring more complex refueling operations to retrieve.

Fast reactor cores are highly sensitive to fuel motion that occurs primarily due to core assembly bowing. Other events such as seismic loading can also induce fuel motion. Core assembly deformations influence the neutron balance within the core (primarily neutron leakage), which lead to changes in reactivity. This reactivity feedback is significant in SFRs and is understood to be a primary contributor to the core melt accident at EBR-I. As such, reactivity feedback induced by the core restraint system plays a role in reactor stability and is a consideration in reactor safety analysis.

In order to achieve a determinate and predictable configuration of core assemblies, fully developed load paths need to develop between various core assembly ACLPs and TLPs, eventually reacting radially at the CSS. This condition is referred to as a mechanically locked core and is achieved when core-wide inter-assembly gaps are sufficiently closed, primarily through core assembly bowing, and adequate inter-assembly contact is established. Once sufficient contact and a mechanically locked core condition is established, the core restraint system tends to insert negative reactivity with increases in core power.

The core restraint system has competing requirements that relate primarily to inter-assembly gap management. Designing smaller inter-assembly gaps generally produces more favorable reactivity behavior both from assembly bowing and seismic events. With smaller gaps between individual assemblies, a mechanically locked core condition can typically be achieved earlier as there are less cumulative core-wide assembly gaps that need to be overcome by core assembly bowing to establish sufficient core-wide inter-assembly contact. This indicates larger ranges of operation within the mechanically locked core condition, which in-turn provides more stable reactivity feedback behavior. Additionally with smaller inter-assembly gaps, potential reactivity insertions from core restraint are generally smaller in magnitude from both bowing and seismic loading as there is less space for assemblies to translate before core-wide assembly contact is established. Tighter inter-assembly gaps also promote better core assembly alignment for interfacing systems such as the IVHM or control rod driveline.

The competing design requirement strives to maintain sufficiently large inter-assembly gaps for ease of core management. High residence fuel assemblies experience significant inelastic material deformations due to extended time at high temperatures, fluences, and bending loads applied by the core restraint system. During refueling operations, high deformation assemblies can cause excessive handling loads on the IVTM if the amount of assembly deformation is excessive relative to the inter-assembly gaps available through which to withdraw the assembly.

Depending on core assembly material selection, different mechanisms can dominate inelastic material deformation. Austenitic stainless steels, such as 316 SS, exhibit swelling-dominated deformation behavior while ferritic-martensitic stainless steels, such as HT9, exhibit creep-dominated deformation behavior. Designing sufficiently large inter-assembly gaps allows

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 for removal of high deformation assemblies from the core without exceeding the load capacity of the handling machine.

5.5 Materials 5.5.1 HT9 Alloy HT9 stainless steel, referred to here as HT9, is the material selected for fuel cladding, ducts, and other fuel components. Because austenitic stainless steels undergo excessive radiation-induced void swelling prior to reaching the desired fuel life, fast-reactor cladding and duct development programs of the past switched focus from austenitic to ferritic or ferritic-martensitic (FM) steels for long exposure applications. Selected FM steels, including HT9, exhibit strong resistance to swelling, maintain adequate microstructural stability under irradiation, and retain adequate ductility at typical reactor operating temperatures.

HT9 and analogous alloys, like the Russian EP-823, have been considered top candidate materials for nuclear reactor core components since the Liquid Metal Fast Breeder Reactor (LMFBR) era in the 1970s and 1980s. In the US fast-reactor development program, HT9 developed by Sandvik was selected for cladding and ducts for the Integral Fast Reactor (IFR) concept [15], as the next generation fuel cladding for the EBR-II [16], and as the cladding and duct material for the metallic fuel assemblies in the FFTF [17]. For the Natrium Reactor, the pathway for qualification of HT9 as fuel component material is centered on leveraging test data and component operating experience from these programs. Additionally, TerraPower has invested substantially in developing HT9 material more recently and plans to demonstrate any improvements over legacy material as described in Section 6.3 of this document.

The scope of qualification of HT9 for the Natrum Reactor is currently limited to the fuel system.

The fuel system consists of the fuel assemblies, reflector assemblies, shield assemblies, and reactivity control assemblies.

Component designs using HT9 material in the fuel system are cladding, ducts, upper and lower fuel pin endcaps, and wire wrap.

5.5.1.1 Composition Alloy HT9 steel is a FM Cr-Mo stainless steel whose evolution and composition can be traced to AISI 430 and AISI 410, the basic general-purpose alloys of the ferritic stainless-steel family. FM stainless steels are generally defined as those containing at least 9 wt.% chromium and have microstructures of -iron (ferrite), martensite, and carbides. Several standard and nonstandard alloy types have been derived from the basic alloy by varying the composition to achieve specific properties. Typical alloying elements in addition to Cr and Mo are W (up to 3 wt.%), V (< 0.5 wt.%), and Nb (< 0.5 wt.%). Alloy HT9 is classed as a 12Cr-MoVW type. The W addition endows this grade with greater strength than corresponding steels without W.

The nominal composition of HT9 is in the table below, taken from Chapter 18 of the Fuel Cycle Research and Development (FCRD) Materials Handbook [18].

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 5-1. Nominal Composition of HT9 Steel Element Weight (%)

((

)) (a)(4),ECI The product forms of HT9 needed for fuel components are cladding tubes (fuel and absorber),

duct tubes, wire, bar (for endcaps), and sheets (for welded ducts, if required).

Qualification of HT9 must include evaluation of welds. Resistance pressure welding (RPW) is being developed for fuel pin end caps, while other forms of welding are being considered for ducts if required.

5.5.1.2 Manufacturing Process More than eight metric tons of HT9 have been manufactured to date in 18 heats during development by TerraPower, supporting the definition of manufacturing steps as listed in the table below.

Table 5-2. Manufacturing Steps for HT9 Components Step Critical Parameters Melt

((

))(a)(4)

Homogenization

((

))(a)(4)

Forge

((

))(a)(4)

Hot Work (Extrusion /

Rolling)

((

))(a)(4)

Cold Work (Drawing /

Rolling) and Intermediate Anneal

((

))(a)(4)

Final Heat Treatment

((

))(a)(4) 5.5.2 U-10Zr Fuel As described above, metallic uranium alloyed with 10wt% zirconium (U-10Zr) is the fissile material used as the fuel slugs in Natrium Type 1 fuel pins. The enrichment of the uranium is determined by the position within the Natrium core, with the peak enrichment being <20% U-235.

((

))(a)(4)

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((

))(a)(4) 5.5.2.1 Manufacturing Process Historically, metallic fuel slugs have typically been manufactured via an injection casting process where the uranium and zirconium (or other fuel alloy components) are loaded into a crucible and melted via induction heating. After the fuel alloy components have melted and mixed, a vacuum is drawn, and an array of quartz molds suspended above the fuel melt is lowered partially into the melt. Pressure is then rapidly applied by introducing inert gas inside of the casting furnace driving the liquid fuel alloy up into the quartz molds, due to differential pressure, where it quickly solidifies. The molds are removed from the crucible and the system allowed to cool. The molds are removed from the injection casting furnace and broken away to reveal the fuel slugs. The fuel slugs are sheared to length and inspected prior to incorporation in fuel pins.

TerraPower plans on relying heavily on the irradiated metallic fuel data from EBR-II and FFTF, which was fabricated with injection casting. To mitigate potential risks of unanticipated performance introduced by alternate fabrication processes, TerraPower will use the injection casting process for Type 1 fuel slug fabrication. TerraPower is relying heavily on knowledge transfer from the DOE labs in developing its new injection casting capabilities, including historic system drawings, casting system design specifications, operating procedures, fuel specifications, and interviews with former operators.

TerraPower previously performed detailed characterization of archived fresh U-10Zr fuel materials from EBR-II and FFTF/MFF [19] [20]. When newly manufactured U-10Zr slugs are available from prototype equipment testing characterization will be performed to verify consistency with legacy materials.

5.5.3 Other Core Materials For some of the components of the fuel system other materials are being considered. This includes 304 and 316 austenitic stainless steel, and Inconel 718 nickel-based superalloy. These materials are well established and are widely used in the nuclear industry, with extensively documented performance and properties. Design inputs taken from NRC accepted standards for these types of materials are considered pre-qualified.

5.6 Verification of the Fuel System Design Basis Fuel system descriptions and design drawings are required to support safety analyses to provide the information necessary to verify that the fuel system design bases are met. The specific details adapted from the Standard Review Plan [2] are summarized in Table 5-3, along with the associated documentation/media planned to provide the required fuel design information.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 5-3. Summary of Completed and Planned Activities to Satisfy Fuel System Design Description Requirements (RAC 4.2-5)1 1 Additional guidance related to required description of the fuel system is provided in Regulatory Guide 1.206 - Combined License Applications for Nuclear Power Plants.

Expected Format of Information Required Fuel System Information with Associated Tolerances Sample Reference or Future Activity to Address Design Description of Cladding for Fuel/Absorber Pins Type and metallurgical state of the cladding

((

))(a)(4)

Cladding outside diameter

((

))(a)(4)

Cladding inside diameter

((

))(a)(4)

Cladding roughness

((

))(a)(4)

Design Description of Fuel Pin Fuel slug density

((

))(a)(4)

Fuel slug diameter

((

))(a)(4)

Fuel slug length

((

))(a)(4)

Fuel slug grain size

((

))(a)(4)

Slug alloy composition for metallic fuel

((

))(a)(4)

Allowable slug impurities

((

))(a)(4)

Shield slug parameters

((

))(a)(4)

Sodium bond height

((

))(a)(4)

Fuel column length

((

))(a)(4)

Overall pin length

((

))(a)(4)

Fill gas type and pressure

((

))(a)(4)

End plug dimensions

((

))(a)(4)

Wire wrapping dimensions

((

))(a)(4)

Fissile enrichment and isotopics

((

))(a)(4)

Design Description of Absorber Pin Pellet density

((

))(a)(4)

Pellet diameter

((

))(a)(4)

Slug grain size

((

))(a)(4)

Pellet chemical composition for absorber

((

))(a)(4)

Allowable pellet impurities

((

))(a)(4)

Shield slug parameters

((

))(a)(4)

Plenum height

((

))(a)(4)

Plenum spring

((

))(a)(4)

Absorber column length

((

))(a)(4)

Overall pin length

((

))(a)(4)

Fill gas type and pressure

((

))(a)(4)

Upper end plug dimensions

((

))(a)(4)

Lower end plug dimensions

((

))(a)(4)

Wire wrapping dimensions

((

))(a)(4)

Boron enrichment and isotopics

((

))(a)(4)

Design Drawings Fuel assembly cross section

((

))(a)(4)

Fuel assembly outline

((

))(a)(4)

Fuel pin schematic

((

))(a)(4)

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6. FUEL SYSTEM DESIGN EVALUATION The Natrium Reactor will use analytical predictions to evaluate fuel system compliance with the design basis limits, while pointing to operating experience of similar historic metallic fuel pin designs and relying on testing to help bridge the gap between historic experience/designs and Natrium fuel design parameters. Analytical predictions of Natrium Reactor performance will evaluate the fuel system design for physically feasible combinations of chemical, thermal, irradiation, mechanical, and hydraulic interactions. The evaluation of these interactions will include the effects of normal operations, AOOs, and LBEs [39]. New tests will largely focus on understanding high-importance phenomena needed to evaluate compliance to the fuel design bases. To aid in the identification of which RAC are addressed by fuel system design limits, Table 4-1 through Table 4-4 summarize the correspondence of fuel system design criteria with the RAC for fuel damage, failure, coolability, and control rod insertability.

To aid in the identification of high-importance phenomena, a Phenomena Identification Ranking Table (PIRT) analysis was performed evaluating the applicable phenomena for each fuel and absorber design criteria [40] [41] [42]. These assessments were performed by convening a team of experts within TerraPower with representatives from the Fuels, Materials, Safety, and Mechanical teams to assess the applicable phenomena for each fuel pin design limit and the relative Importance and Knowledge Level of the respective phenomena. The internal definitions used for determining the Importance rankings and Knowledge Levels are summarized in Table 6-1 and Table 6-2, respectively.

Table 6-1. Importance Ranking Definitions Importance Ranking Definition Low (L)

Small influence on demonstrating compliance

+/- 1 variation of parameter/phenomenon has minimal impact on prediction of design criterion Medium (M)

Moderate influence on demonstrating compliance

+/- 1 variation of parameter/phenomenon has moderate impact on prediction of design criterion High (H)

Significant influence on demonstrating compliance

+/- 1 variation of parameter/phenomenon has significant impact on prediction of design criterion Expected Format of Information Required Fuel System Information with Associated Tolerances Sample Reference or Future Activity to Address Fuel pin wire wrap location

((

))(a)(4)

Absorber pin schematic

((

))(a)(4)

Wire wrap location

((

))(a)(4)

Inlet and outlet nozzles

((

))(a)(4)

Control rod duct with respect to control rod dimensions

((

))(a)(4)

Control rod assembly cross section

((

))(a)(4)

Control rod assembly outline

((

))(a)(4)

Control rod schematic

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-2. Knowledge Level Definitions Knowledge Level Definition Known (K)

Approximately 70% to 100% of complete knowledge and understanding Partially Known (P) 30% to 70% of complete knowledge and understanding Unknown (U) 0% to 30% of complete knowledge and understanding 6.1 High-importance Phenomena The PIRT assessments were performed to help identify the high-importance phenomena that must be accounted for when evaluating the performance of Type 1 fuel and control assemblies. The high-importance phenomena for fuel and absorber pins are summarized in Table 6-3, along with the corresponding design criteria and associated RAC. When evaluating and consolidating the high-importance phenomena, it became clear that many of the identified phenomena are more aptly described as operating parameters/conditions (e.g., Cladding Temperature, Fuel Burnup) versus complex physical phenomena (e.g., Fission Gas Release or FCCI), so these different categories were also noted in Table 6-3. To help prioritize activities, an Overall Knowledge Level ranking is also included in Table 6-3, which is the average Knowledge Level determined for each identified high-importance phenomena/parameter.

Table 6-4 is a summary of high-importance phenomena for fuel and control assemblies.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-3. Summary of Identified High-Importance Phenomena and Associated Design Limits and RAC for Fuel and Absorber Pins Category High-importance Phenomena/Parameters Overall Knowledge Level Applicable Design Limit Applicable RAC Fuel Pin Phenomena Fission gas release P

Total Peak Cladding Strain, Peak Cladding Thermal Creep 4.2-1.1, 4.2-1.6, 4.2-1.8, 4.2-2.3, 4.2-2.5, 4.2-3.1, 4.2-3.3, 4.2-3.5 HT9 mechanical response as a function of temperature, stress, irradiation, and time K

Total Peak Cladding Strain, Fatigue Limit, Peak Cladding Thermal Creep Strain 4.2-1.1, 4.2-1.6, 4.2-1.8, 4.2-2.3, 4.2-2.5, 4.2-3.1, 4.2-3.3, 4.2-3.4, 4.2-3.5 FCCI P

Cladding Wastage, Peak Cladding Thermal Creep Strain 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-2.3, 4.2-2.5 Fuel Thermal Conductivity P

Peak Fuel Temperature 4.2-2.2,4.2-3.4 Absorber Pin Phenomena Gas release K

Total Peak Cladding Strain 4.2-1.1, 4.2-1.7, 4.2-1.8, 4.2-1.11,4.2-4.1 HT9 mechanical response as a function of temperature, stress, irradiation, and time P

Total Peak Cladding Strain, Fatigue Limit 4.2-1.1, 4.2-1.2, 4.2-1.7, 4.2-1.8, 4.2-1.11, 4.2-4.1 ACCI U

Total Peak Cladding Strain, Total Peak Cladding Wastage, 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-1.11, 4.2-4.1 B4C Swelling/ACMI K

Total Peak Cladding Strain, Peak Absorber Temperature 4.2-1.1, 4.2-1.7, 4.2-1.8, 4.2-1.11, 4.2-4.1 Absorber cracking/fragmentation P

Peak Absorber Temperature Limit 4.2-1.11 Fuel Pin Parameters/

Operating Conditions Fuel Burnup P

Total Peak Cladding Strain, Fatigue Limit, Peak Cladding Thermal Creep 4.2-1.1, 4.2-1.6, 4.2-1.8, 4.2-2.3, 4.2-2.5, 4.2-3.1, 4.2-3.3, 4.2-3.5 DPA on cladding P

Total Peak Cladding Strain, Fatigue Limit 4.2-1.1, 4.2-1.6, 4.2-1.8 Cladding temperatures P

Total Peak Cladding Strain, Fatigue Limit, Cladding Wastage, Peak Cladding 4.2-1.1, 4.2-1.6, 4.2-1.8, 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-2.3, 4.2-2.5, 4.2-3.1, 4.2-3.3, 4.2-3.4, 4.2-3.5

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Category High-importance Phenomena/Parameters Overall Knowledge Level Applicable Design Limit Applicable RAC Thermal Creep Strain Number of strain cycles on cladding K

Fatigue Limit 4.2-1.2 Magnitude of strain cycles P

Fatigue Limit 4.2-1.2 Residence Time P

Cladding Wastage 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-2.5 Detailed pin level irradiation histories including power and cladding temperature K

Peak Cladding Temperature, Peak Fuel Temperature 4.2-1.11, 4.2-1.12, 4.2-2.1, 4.2-3.2, 4.2-2.2,4.2-3.4 Detailed coolant transient temperature and pin power histories P

Peak Cladding Temperature, Peak Fuel Temperature 4.2-1.11, 4.2-1.12, 4.2-2.1, 4.2-3.2, 4.2-2.2,4.2-3.4 Absorber Pin Parameters/

Operating Conditions Depletions P

Total Peak Cladding Strain, Peak Cladding Temperature Limit, Peak Absorber Temperature Limit 4.2-1.1, 4.2-1.7, 4.2-1.8, 4.2-1.11, 4.2-4.1 Absorber thermal conductivity P

Peak Absorber Temperature Limit 4.2-1.11, 4.2-4.1 DPA on clad P

Fatigue Limit, Total Peak Cladding Strain 4.2-1.2, 4.2-4.1 Cladding temperatures P

Total Peak Cladding Strain, Cladding Wastage Limit, 4.2-1.1, 4.2-1.2, 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-1.7, 4.2-1.8, 4.2-1.11, 4.2-4.1 Coolant temperatures P

Peak Cladding Temperature Limit, Peak Absorber Temperature Limit 4.2-1.11 Number of strain cycles on cladding P

Fatigue Limit 4.2-1.2 Magnitude of strain cycles P

Fatigue Limit 4.2-1.2 Residence time P

Cladding Wastage 4.2-1.3, 4.2-1.4, 4.2-1.5, 4.2-1.11 Sodium velocity P

Fatigue Limit 4.2-1.2 Bundle design/compliance P

Fatigue Limit 4.2-1.2 Detailed pin level irradiation histories K

Peak Cladding Temperature, Peak 4.2-4.1

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Category High-importance Phenomena/Parameters Overall Knowledge Level Applicable Design Limit Applicable RAC including power and cladding temperature Absorber Temperature Detailed transient power and coolant temperature histories P

Peak Cladding Temperature, Peak Absorber Temperature 4.2-4.1 Weld susceptibility to irradiation (embrittlement and swelling)

U Total Peak Cladding Strain, 4.2-1.1, 4.2-1.7, 4.2-1.8, 4.2-1.11 Na-bonded versus He-bonded P

Peak Absorber Temperature Limit 4.2-1.11 B4C to cladding gap characteristics K

Peak Absorber Temperature Limit 4.2-1.11 Table 6-4. Summary of Identified High-Importance Phenomena and Associated Design Limits and Applicable RAC for Fuel and Control Assemblies Category High-Importance Phenomena/Parameters Overall Knowledge Level Applicable Design Criteria Applicable RAC Fuel System Damage under Normal Operation and AOOs Impact loads due to handling drop accidents K

Stress, Strain, Loading Limit 4.2-1, 4.2-3, 4.2-4 Withdrawal/insertion forces K

Stress, Strain, Loading Limit 4.2-1, 4.2-3, 4.2-4, 4.2-4 Pin bundle to duct interaction P

Dimensional changes 4.2-1, 4.2-3, Core Seismic Criteria under Operating Basis Earthquake Fuel assembly and component residual horizontal deformations P

Reactivity Insertion Limit - Post-OBE Operability 4.2-2, 4.2-3, 4.2-4 Fuel assembly and component residual horizontal displacements P

Refueling Force Limit - Post-OBE Operability 4.2-2, 4.2-3, 4.2-4 Core Seismic Criteria under Safe Shutdown Earthquake Fuel assembly and component horizontal displacements P

Reactivity Insertion Limit - Pre-SCRAM displacements 4.2-2, 4.2-3, 4.2-4 Fuel assembly and component residual deformations P

Core Coolability Limit - Core Coolability 4.2-2, 4.2-3, 4.2-4 As described in RAC 4.2-6, three methods are acceptable for demonstrating that the fuel system design bases are met: 1) historic operating experience (Section 6.2), 2) testing (Section 6.3), and 3) analytical predictions (Section 6.4). The following subsections describe the current plan for using each of these methods to address the various fuel system design bases. More details will be provided for how these respective methods address the Experimental Data Assessment Framework and Evaluation Model Framework Goals from NUREG-2246 in the subsequent sections, where applicable.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 6.2 Historic Operating Experience The key design parameters available to improve the performance of fuel pins are related to the material used for fuel (metal fuel composition) and cladding (316 SS, D91, HT9) and fuel pin dimensions (smear density, cladding thickness, pin diameter, fuel column and plenum lengths). This section describes the historical experience and motivation for key design decisions. The Natrium Reactor Type 1 fuel is similar to fuel pins used historically and will target operating conditions within the operating envelope of earlier reactors.

It was recognized early in the development of metallic fuel pins that the key phenomena contributing to steady-state fuel pin performance include the fuel swelling and axial growth, fission gas release, cladding irradiation creep and swelling, FCCI and constituent redistribution in the fuel due to fission product generation and thermal gradients. Transient performance is also dependent on the fuel melting temperature, the formation of low melting point fuel/clad eutectics and thermal creep and total strain in the cladding. This section will describe how these key phenomena drove fuel pin design to achieve higher-burnup and longer residence times, while also ensuring performance during operational transients.

Metal fuels have high thermal conductivity and excellent compatibility with sodium relative to oxide fuels. In addition, transient testing at the TREAT reactor demonstrated that despite having a lower melting point than oxide fuels, metal fuels performed exceptionally under fast transient conditions.

For example, metal fuel failed at five times the nominal peak power during TREAT transient testing2

[5]. PIE studies on early [43], low-burnup fuel pins indicated that adding small amounts of alloying elements, such as zirconium, to metallic fuel increased dimensional stability but could not stop fission gas induced swelling of the fuel. Zirconium became the alloying element of choice because it also increased the fuel solidus temperature and increased the fuel/cladding eutectic formation temperature above that of the binary uranium/iron alloy (725°C (1337°F)) [44] for fuel with less than 10 wt% plutonium [45]. Natrium Reactor Type 1 fuel has uranium-10 wt % zirconium fuel, which was used in early reactors in the U.S.A.

Fuel swelling limited the burnup capability of early metal fuel pin designs, and it was recognized that the swelling was caused by the accumulation of fission products [46]. It was also noted that fission gas is essentially insoluble in metal fuels and that at pore volume fractions over 30% the pores become linked effectively releasing the fission gas [47]. These observations led to a second generation of fuel pins that had 75% smear density and greatly increased plenum lengths (up to 1.45 times the fuel volume). These pins achieved up to ~10% burnup. Fuel pins with 75% smear density also included a sodium bond layer to conduct heat between the fuel and the clad at the beginning of life before the fuel contacts the clad. The fuel grows in the axial direction ~2-10% before the fuel and cladding are in direct contact. After contact between the fuel and cladding, axial growth changes are minimal. Natrium Reactor Type 1 fuel incorporates the smear density of 75% and the larger plenum volume that was developed in these second-generation pins.

The third generation of fuel pins, designed and fabricated in the late 1980s, incorporated ferritic-martensitic steels such as HT9. These improved cladding alloys exhibited low swelling from internal void formation. In addition, they show superior resistance to irradiation creep. The pin dimensions 1 D9 is a nuclear grade austenitic stainless steel based on AISI type 316-SS that has improved void-swelling resistance.

2 Metal fuel experiments failed during transient testing at more than 200 kW/m, which is approximately ((

))(a)(4)(ECI) times the nominal linear power of Type 1 fuel in the Natrium Reactor.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 also evolved to include a 33% increase in cladding thickness, a larger diameter, and longer pin length. The structural load on the cladding is due largely to fission products. Fission gases drive the initially high rate of fuel swelling. The porosity eventually becomes interconnected allowing the gases to access the plenum volume. Eventually, the porosity beings to fill with solid fission products and the fission gas is concentrated in the interconnected porosity and plenum region increasing the load on the cladding at high burnups. Natrium Reactor Type 1 fuel pins have HT9 cladding.

FCCI is a phenomenon that may limit the lifetime of metal fuel pins, particularly if the operating temperatures are relatively high. It is caused by the diffusion of lanthanide fission products into the clad which can lead to embrittlement. The embrittlement is so severe that the layer of impacted cladding is considered unable to support any loads. During structural analysis, the cladding thickness is reduced by the width of the FCCI layer (thinning of the clad is assumed to occur). Therefore, FCCI impacts fuel performance during normal operations and transients. The Natrium Reactor Type 1 fuel pins will be used within the burnup and operational limits of the historical database.

Another diffusional process that impacts material properties and performance in fuel pins is constituent redistribution in the fuel. Depending on operating conditions, zirconium may migrate up and down the thermal gradient in alloy fuel, which may lead to a higher concentration of zirconium at the center of the fuel pin and at the inner surface of the cladding, with corresponding depletion in the intermediate region. This can impact the local solidus temperature due to the dependence on zirconium concentration. Peak fuel temperatures occur at the center of the fuel where zirconium concentrations are seen to rise due to constituent redistribution, slightly increasing solidus temperatures relative to beginning of life conditions. Natrium Type 1 fuel is expected to have less pronounced constituent redistribution than was typical for the EBR-II and FFTF metallic fuels due to a lower targeted pin power which leads to smaller temperature gradients.

Metallic fuel pins were used in fuel assemblies in the EBR-I and EBR-II reactors. A metallic fuel qualification test program was in progress for the FFTF, known as the MFF1 series of fuel assemblies, when it was shut down in 1994. The FFTF fuel designs were more relevant to commercial designs because of their larger diameter, longer length fuel pins and fuel assemblies with 169 fuel pins compared to 61 pins in EBR-II. The MFF pins had 75% smear density uranium-10 wt % zirconium and HT9 cladding. Burnups up to ~15 %FIMA were tested with a range of temperatures from nominal up to 2-sigma hot channel factor2. The total fission gas release was consistent with that measured for shorter pins. HT9 cladding had 6-10 times less diametral strain relative to D9, which was attributed to the lower swelling of HT9 [48]. Results from PIE show that the FCCI time-temperature data and nonuniform circumferential depth for the MFF pins was also consistent with the shorter EBR-II pins; however, in some cases the depth of the FCCI layer did not reach a maximum at the top of the fuel column [48].

The metal fuel irradiation operational experience consists of ~130,000 metal fuel pins with most of the burnups less than 10 %FIMA but with some as high as 20 %FIMA burnup. The pins can be divided into approximately four categories:

1 A recent reference identified MFF as Mechanistic Fuel Failure [48] but no official definition was found in the original literature from FFTF.

2 The hot channel factor comes from the hot channel model where the hot channel is affected by the simultaneous occurrence of all uncertainties [84].

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 90,000 pins with 85% smear density or higher and 2.6 %FIMA burnup 30,000 pins with 75% smear density and 8 %FIMA burnup 13,000 pins with 75% smear density and larger diameter with 10 %FIMA burnup 1,050 pins developed for the FFTF/MFF, metallic fuel qualification tests with 75% smear density, larger diameter and HT9 cladding with varying burnup.

The EBR-II fuel performance was exceptional. Out of 30,000 pins of an early design (90% smear density, uranium-2 wt.% zirconium fuel, SS347 cladding) there were only 13 failures related to a restrainer on the pin that was removed from the design eliminating this failure mode. 13,600 pins with alloy fuel were used with only 22 natural breached pins reported (19 were in high temperature tests run to failure with 16 of the reported failures at welds). Seven tests to run metallic fuel pins beyond cladding breach were also performed in EBR-II. The tests, which ran between 34 and 233 days post cladding breach, indicated that almost all the fission gas in the plenum and interconnected porosity is released into the primary system during the breach. The post-breach fuel behavior was benign and breached pins showed little difference from intact pins regardless of the post-breach operating time.

EBR-II transient tests were also performed which demonstrated operational reliability including 56 tests with a low ramp rate (1.6 % power increase per second) and 13 with a high ramp rate (6.3 %

power increase per second) [49]. More severe transient tests demonstrated that metal fueled cores can safely operate with loss-of-flow-without-SCRAM events and loss-of-heat-sink-without-SCRAM events with no core damage. These tests were performed with fully instrumented and calibrated in-core fuel assemblies and showed that the fuel pins successfully operated for 42 minutes with cladding temperatures as high as 800°C.

Transient tests on fuel pins were also performed at the TREAT reactor at Idaho National Laboratory.

TREAT is capable of performing tests on individual fuel pins that are extremely severe, such as melting the fuel or breaching the cladding, to understand failure mechanisms more fully in fuel pins under severe conditions. In general, fuel pins will fail by one of the following mechanisms based on burnup.

Fuel Clad Thermal Creep Strain Failure: Failure occurs due to localization of thermal creep strain (creep rupture) driven by high pressure fission gas in the plenum and open porosity in the fuel due to the high temperatures during the transient. This failure mechanism is seen where high cladding temperatures are sustained and fuel burnup is high enough that the stress imposed by the fission gas leads to high diametral strains in the cladding.

Fuel Cladding Total Wastage Failure: Failure occurs due to accelerated degradation of the cladding from the formation of eutectic phases. These reactions can occur extremely quickly (on the order of minutes) and occur in lower burnup pins that do not have a large inventory of fission gas.

Fuel Melting: Occurs when fuel temperatures exceed the melting temperature of the fuel. Fuel melting also contributes to eutectic formation by supplying free uranium for reactions occurring at the fuel-cladding interface. This failure mode may occur at a range of burnup levels; however, requires very significant overpower conditions.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 A total of 15 transient overpower tests of metallic fuel were performed at TREAT and show similar trends. The differences between alloys and samples can be largely explained by retained fission gas and variation in melting temperature between the fuel alloys. In general, failures occurred at the top of the fuel column where the inner cladding temperature tends to be high relative to the bottom of the fuel column. Two trends were noted that may mitigate the impact of fuel pin failures in a reactor. The first is that the fuel expands axially due to thermal expansion and from the expansion of fission gas in porosity in the fuel column. Expansion of the fuel in the axial direction generally reduces the reactivity of the core. The second is that fuel dispersal occurred quickly after a cladding breach, also removing reactivity from the core.

Although there is no direct commercial reactor operating experience with Type 1 fuel, the fuel design and operational targets are intentionally modeled after metallic fuel designs that successfully operated in EBR-II and FFTF, providing confidence in the overall reliability of the fuel design. Table 6-5 provides a comparison of the fuel pin dimensions and other design parameters for Type 1 Fuel, FFTF, and EBR-II fuel pins. Table 6-6 gives a similar comparison for Natrium Reactor absorber pins and FFTF absorber pins [50]. Table 6-7 provides a comparison between Type 1 fuel assemblies and other SFR assemblies. Using the recent Natrium fuel performance assessment to support conceptual design [51], comparisons of anticipated Type 1 fuel pin conditions and corresponding conditions from EBR-II and FFTF are summarized in Table 6-8. Peak burnup and linear heat generation rate are within the range of the tested parameters in EBR-II and FFTF, ((

))(a)(4)(ECI). Additionally, a fuel surveillance program is planned to monitor performance at interim burnup steps (covered by this historic database) to verify performance is consistent with predictions (see Section 9).

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-5. Summary of Fuel Pin Parameters Including Comparison to FFTF/MFF1 and EBR-II Parameter Unit Type 1 Fuel FFTF/

MFF

[54]

EBR-II MkIII Driver Fuel [16]

MkIV Driver Fuel [16]

X430 [55]

Cladding Outer Diameter in.

((

))(a)(4)(ECI) 0.270 0.23 0.23 0.29 mm

((

))(a)(4)(ECI) 6.86 5.84 5.84 7.37 Cladding Thickness in.

((

))(a)(4)(ECI) 0.022 0.015 0.018 0.016 mm

((

))(a)(4)(ECI) 0.56 0.38 0.46 0.41 Cladding Thickness/

Diameter Ratio

((

))(a)(4)(ECI) 0.081 0.065 0.078 0.055 Active Fuel Column Length in.

((

))(a)(4)(ECI) 36.0 13.5 13.5 13.5 m

((

))(a)(4)(ECI) 0.914 0.343 0.343 0.343 Plenum Length in.

((

))(a)(4)(ECI) 51.2 14.7 14.7 14.5*

m

((

))(a)(4)(ECI) 1.3 0.373 0.373 0.368 Fuel Type U-10Zr [30]

U-10Zr U-10Zr U-10Zr U-10Zr, U-Pu-10Zr Weight Fraction Zr 0.1 0.1 0.1 0.1 0.1 Nominal Fuel Smear Density Fraction 0.75 0.75 0.75 0.75 0.75 Bond Sodium [30]

Sodium Sodium Sodium Sodium Cladding Material HT9 [56]

HT9 D9 HT9 HT9

• Derived Value Table 6-6. Summary of Absorber Pin Parameters Including Comparison to FFTF and JOYO Parameter Unit Natrium

[31]

FFTF (Series 1)

[57]

FFTF (Series 2)

[57]

FFTF (Series 3)

[57]

JOYO Mk-II [58]

Control Rod Absorber Material B4C (19.9 at% 10B/B)

((

))(ECI)

((

))(ECI)

((

))(ECI)

B4C (90 at%

10B/B)

Smear Density

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 93 Cladding Tube Material HT9

((

))(ECI)

((

))(ECI)

((

))(ECI)

PNC 316 Fill Gas He

((

))(ECI)

((

))(ECI)

((

))(ECI)

He Cladding Outer Diameter in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.717 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 18.2 Cladding Thickness in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.032 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.8 Cladding Thickness/

Diameter Ratio

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.045 1 FFTF was fueled with oxide fuel, but eight fuel assemblies using metallic fuel were fabricated to support eventual conversion to metallic fuel. These assemblies were labeled MFF, but no official definition for the label/acronym has been identified.

NAT-2806 TerraPower, LLC (TerraPower) Natrium Topical Report: Fuel and Control Assembly Qualification Page 66 of 119 Controlled Document - Verify Current Revision Copyright © 2023 TerraPower, LLC. All rights reserved.

SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Parameter Unit Natrium

[31]

FFTF (Series 1)

[57]

FFTF (Series 2)

[57]

FFTF (Series 3)

[57]

JOYO Mk-II [58]

B4C Column Length in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 25.6 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 650 B4C Pellet Diameter in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.630 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 16 Number of Pins/Assembly

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 7 Total Plenum length in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 17.7 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 449 Wire wrap diameter in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 0.098 mm

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 2.5 Overall Control Pin Length in.

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 50.12 m

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 1.273 Design Lifetime EFPD

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Table 6-7. Type 1 Fuel Assembly Design Parameters Assembly Design Parameters Unit Type 1 Fuel FFTF III.b

[59] [17]

PRISM [60]

VTR [61]

Duct Material HT9 [62]

HT9 HT9 HT9 Duct Thickness in.

((

))(a)(4)(ECI) 0.118 0.140 0.118 mm

((

))(a)(4)(ECI) 3.00 3.556 3.00 Duct Flat to Flat OD in.

((

))(a)(4)(ECI) 4.59 6.106 4.606 mm

((

))(a)(4)(ECI) 116.5 155.1 117 Number of Pins/Bundle

((

))(a)(4)(ECI) 169 271 217 Pin Length in.

((

))(a)(4)(ECI) 93.4 158.0 64.96 m

((

))(a)(4)(ECI) 2.371 4.013 1.650 Inlet Nozzle /

Nosepiece in.

((

))(a)(4)(ECI) 33.7 13.0 12.99 mm

((

))(a)(4)(ECI) 856.0 330.2 329.9 Assembly Total Length in.

((

))(a)(4)(ECI) 144 186 152 m

((

))(a)(4)(ECI) 3.658 4.724 3.861 Fuel Assembly Weight lbs.

((

))(a)(4)(ECI) 395.6 1

1 kg

((

))(a)(4)(ECI) 179.4 1

1 1 Information is currently unavailable

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Assembly Design Parameters Unit Type 1 Fuel FFTF III.b

[59] [17]

PRISM [60]

VTR [61]

Number of Assemblies/

Core

((

))(a)(4)(ECI) 74 99 110 Fuel Column Length in.

((

))(a)(4)(ECI) 36.0 47.0 31.5 mm

((

))(a)(4)(ECI) 914.4 1193.8 800.0 Wire Wrap Diameter in.

((

))(a)(4)(ECI) 0.054 0.056 0.044 mm

((

))(a)(4)(ECI) 1.36 1.42 1.11 Wire Wrap Axial Pitch in.

((

))(a)(4)(ECI) 9.20 12.00 10.51 mm

((

))(a)(4)(ECI) 233.7 304.8 267 Fuel Pin Pitch/Diameter (p/d)

Ratio

((

))(a)(4)(ECI) 1.20 1.199 1.18 Assembly Pitch in.

((

))(a)(4)(ECI) 6.283 4.724 mm

((

))(a)(4)(ECI) 159.6 120.0

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-8. Comparison of Fuel System Operational Parameters Parameter Type 1 Fuel Bounding Operating Values

[51]

FFTF (MFF-2) [63]

FFTF (MFF-3)

[64]

EBR-II (X447)

[52]

EBR-II (X425)

[52]

Enrichment,

% U-235

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak

Burnup,

%FIMA

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak, DPA

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Residence, EFPD

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak Linear Heat Rate, kW/m

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak Inner Cladding Temperature

, °C

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak Inner Cladding Temperature

, °F

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

Peak Fast

Fluence, (E>0.1 MeV) n/cm2s

((

))(a)(4)(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI)

((

))(ECI) 1 This value represents the 2 hot channel factor (HCF) temperature The axial distribution for parameters such as linear heat generation rate, burnup, and DPA generally follow a Gaussian shape, shifted slightly towards the bottom of the core due to control rod insertion from the top of the core, as seen in Figure 6-2 for a nominal pin. Over a given cycle, the power shifts slightly towards the top of the core as control rods are withdrawn due to a change in reactivity from the depletion over the cycle. The axial burnup of Type 1 fuel is more peaked in comparison to EBR-II and FFTF, although more similar to FFTF (Figure 6-1) [65]. This has the attractive structural performance feature of reducing burnup at the highest temperature conditions, which reduces the load on the cladding where FCCI and temperature is the greatest. The axial temperature distribution at beginning of life is shown for different temperature conditions in Figure 6-3.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 6-1. Burnup Distribution Comparison between MFF, EBR-II, and Natrium Fuel Pins1 Figure 6-2. Fuel Linear Heat Generation Rate Distribution (left) and Burnup Distribution (right) for a Nominal Pin in the Inner Core Region 1 Burnup distributions taken from representative high-burnup fuel pins. EBR-II calculated values come from ANLs FIPD database

[52], ((

))(a)(4).

(a)(4)(ECI)

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Figure 6-3. Cladding Surface Temperature Distribution at Beginning of Life All applicable fuel tests from EBR-II and FFTF have been reviewed and a detailed plan has been developed to identify and prioritize the fuel pins/assemblies of interest [66]. From EBR-II, ((

))(a)(4) have been selected for data qualification. From FFTF, ((

))(a)(4) from the MFF series tests are also included. These pins span a wide range of design parameters and operating conditions that generally bound the operating conditions targeted for Type 1 fuel. For each subassembly, quantities of measured data from profilometry, gamma scans, neutron radiography (NRAD), and gas release measurements have been determined. Profilometry data will primarily be used to validate predictions of cladding strain, therefore a focus is placed on fuel pins with HT9 cladding at a wide range of operating conditions. NRAD and gas release measurements focus on U-10Zr fuel pins. NRAD data will be used for validation of radial and axial fuel swelling. Gas release measurements will be used for validation of predicted gas release fractions and plenum pressurization.

Cladding wastage measurements from fuel pin metallography will be used to validate wastage predictions.

Fuel pin data from transient testing (e.g., TREAT, furnace tests) has similarly been identified for benchmarking predictions of fuel pin failure, fuel radial melt locations, and fuel axial expansion due to fuel melting. Benchmark comparisons to material test data (e.g., tube burst, creep) is similarly planned.

Validation comparisons to absorber pin data such as plenum pressure, gas release fraction, B4C swelling, ACCI, and B4C temperature will also be performed. A summary of the targeted assemblies, relevant design parameters, operating conditions, and quantities of measured post irradiation exam data for fuel pins and absorber pins is provided in Table 6-9 and Table 6-10, respectively.

(a)(4)(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-9. Relevant Historic Fuel Assemblies to Support Validation Activities Operating Conditions1,2 Measured PIE Data Quantities Design Characteristics Subassembly Peak Burnup [%]

Peak Linear Power [kW/m]

Peak Inner Clad Temperature [°C]

Peak Inner Clad Temperature [°F]

Peak DPA [-]

Peak Pin Residence [EFPD]

Contact Profilometry Laser Profilometry Gamma Scan NRAD Fission Gas Release Fission Gas Chemistry Cladding Material Zr %

Pu %

Zr-Sheathed Fuel Fuel Outer Diameter [in]

Fuel Length [in]

Plenum to Fuel Volume Ratio [-]

Cladding Outer Diameter [in]

Cladding Thickness [in]

Nominal Smear Density [-]

Qualification Priority X419 X419A X419B X420 X420A X420B X421 X421A X423 X423A X423B X423C X425 X425A X425B X425C X429 X429A X429B X430 X430A X430B X431 X431A X432 X432A X441 X441A X447 1 Operating conditions for EBR-II fuel pins come from ANLs FIPD database [52]. ((

))(a)(4).

2 Operating conditions for MFF fuel pins were analyzed by ((

))(a)(4) [64] [63]. ((

))(a)(4).

(ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Operating Conditions1,2 Measured PIE Data Quantities Design Characteristics Subassembly Peak Burnup [%]

Peak Linear Power [kW/m]

Peak Inner Clad Temperature [°C]

Peak Inner Clad Temperature [°F]

Peak DPA [-]

Peak Pin Residence [EFPD]

Contact Profilometry Laser Profilometry Gamma Scan NRAD Fission Gas Release Fission Gas Chemistry Cladding Material Zr %

Pu %

Zr-Sheathed Fuel Fuel Outer Diameter [in]

Fuel Length [in]

Plenum to Fuel Volume Ratio [-]

Cladding Outer Diameter [in]

Cladding Thickness [in]

Nominal Smear Density [-]

Qualification Priority X447A X492 X492A X492B X496 MFF2 MFF3 MFF5 MFF6 Max Total (ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-10. Relevant Historic Absorber/Control Pin Test Assemblies to Support Validation Activities Experiment Details Operating Conditions Measured PIE Data Design Characteristics Experiment/Assembly Reactor

1. Pins Peak Depletion [1020 cap/cc]

Time (EFPD)

Power (MWD)

Peak Pellet Temperature [°C]

Fluence (1022 n/cm2)

Irradiation History Profilometry Swelling Gas Release Cracking Max Plenum Pressure (psi)

Radiographs TEM/SEM ACCI Cladding Material Cladding Outer Diameter [mm]

Cladding Outer Diameter [in]

Cladding Thickness [mm]

Cladding Thickness [in]

Theoretical Density [%]

B-10 Enrichment [%]

Smear Density CR-1 (CRA-528)

FFTF CR-2 (CRA-541)

FFTF CR-3 (CRA-537)

FFTF BICM-1 (YY02)

EBR-II BICM-2 (YY06)

EBR-II WDC-1-1 ETR BOPT-1 (X-248/248A/249)

EBR-II BOPT-2 (large diameter)

EBR-II BOPT-2 (small diameter)

EBR-II BV-2A (X-256)

EBR-II BV-2B (X-265)

EBR-II BV-2C (X-257)

EBR-II He-bonded JOYO Mk-II He-bonded with shroud JOYO Mk-II Na-bonded with shroud JOYO Mk-III ADVAB-1 FFTF ADVAB-2 FFTF Vented Pins EBR-II CR-7 FFTF CR-8 FFTF (ECI)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 In addition to the reports summarizing data from specific fuel assemblies, reports have also been issued summarizing key fuel performance phenomena. Specifically, there are dedicated reports on ((

))(a)(4). Note that these reports cover all the high-importance phenomena summarized in Table 6-3 except for ((

))(a)(4). There is limited ((

))(a)(4). To address this lack of historical ((

))(a)(4) additional testing is planned ((

))(a)(4). See Section 6.3 for more details.

As shown in Table 6-5 and Table 6-8,the ((

))(a)(4) assemblies ((

))(a)(4) most closely match the Type 1 fuel design and targeted operational conditions; ((

))(a)(4). The measured cladding strains on these pins will be the primary independent validation basis for TerraPowers fuel performance tools, with a subset of the fuel pins excluded from model development/calibration efforts to serve as a blind comparison. Specific references for the data collected thus far will be included in the applicable areas in subsequent sections.

6.2.1 Quality of Historic Data The ability to leverage historic data greatly reduces the amount of testing required; however, an assessment is required to verify that the historic data have been accurately measured (NUREG-2246 Experimental Data (ED) Assessment Framework Goal ED G3). NUREG-2246 identifies three sub-goals to demonstrate the accurate measurement of experimental data: ED G3.1) the test facility has an appropriate quality assurance program, ED G3.2) experimental data are collected using established measurement techniques, and ED G3.3) experimental data account for sources of experimental uncertainty.

Both EBR-II and FFTF were DOE test reactors with rigorous quality programs; however, because the data was not collected under TerraPowers approved quality assurance program [75],

additional effort is required to qualify the existing data for use. Specifically, four methods are approved for qualifying existing data: 1) Demonstrate Quality Assurance Program Equivalency,

2) Data Corroboration, 3) Confirmatory Testing, or 4) Peer Review [76].

To support this effort, data qualification plans will be developed for each set of historic data requiring qualification. Specifically, for EBR-II and FFTF experience, a data qualification plan has been developed to evaluate and qualify the historical or pre-existing fuel and absorber pin data intended for use to support model development and validation [66]. As part of this plan Argonne National Laboratory (ANL) will take the lead for qualifying the relevant EBR-II fuel pin steady-state and transient data relying on the process that was previously reviewed and approved by the NRC [77] [78].

6.3 Testing As shown in Section 6.2, the operating experience of metallic fuels in EBR-II and FFTF largely covers the targeted conditions for Type 1 fuel, and the intention is to rely heavily on this basis for the

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 safety case and validation of methods for Type 1 fuel pins; however, a subset of tests have been or will be performed to supplement the historic operating experience. These tests primarily fall within five categories: 1) verification of consistent (or improved) performance between newly manufactured materials/ and historic materials, 2) testing to cover extrapolations of conditions beyond the historic database (i.e., long-duration creep testing), 3) testing to reduce uncertainties or improve fundamental understanding, 4) testing requiring prototypic bundle or fuel pin geometries, and 5) testing to address any gaps in the historic operational experience ((

))(a)(4). These tests include separate effect and integral effect tests. Summaries of the testing that has been performed to support Type 1 fuel qualification are given in Table 6-11 through Table 6-13 organized according to RAC. Future tests that are envisioned to address Fuel Damage, Fuel Failure, and Fuel Coolability Criteria are given in Table 6-14 through Table 6-16. A summary of high-priority testing activities specifically related to the high-importance phenomena is provided in Table 6-17. Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

Table 6-11. Design Basis Criteria and Supporting Information to Prevent Fuel Pin Damage Specific RAC Applicable Design Basis Criteria Applicable Pin Available Supporting Data 4.2-1.1 Total Diametral Strain of Cladding Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4)

Absorber

((

))(a)(4) 4.2-1.2 Design Fatigue Lifetime Fuel/Absorber ((

))(a)(4) 4.2-1.3 Cladding Wastage (Fretting)1 Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4) 4.2-1.4 Cladding Wastage (Na Corrosion)1 Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4) 4.2-1.5 Cladding Wastage (FCCI) or (ACCI)1 Fuel

((

))(a)(4)

Fuel

((

))(a)(4) 1 A single limit is set for wastage; however, all contributions (fretting, Na corrosion, FCCI/ACCI) are assessed in a conservative manner and combined to verify the total wastage limit is not exceeded.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Applicable Design Basis Criteria Applicable Pin Available Supporting Data Absorber

((

))(a)(4) 4.2-1.6 Total Diametral Strain of Cladding Fuel See activities from 4.2-1.1 4.2-1.7 Total Diametral Strain of Cladding Absorber See activities from 4.2-1.1 4.2-1.8 Total Diametral Strain of Cladding requires internal fuel pin pressure to be assessed Fuel

((

))(a)(4)

Fuel/Absorber ((

))(a)(4)

Fuel/Absorber ((

))(a)(4)

Absorber

((

))(a)(4) 4.2-1.111 Peak Cladding Temperature Absorber

((

))(a)(4)

Peak Absorber Temperature Absorber

((

))(a)(4)

Cladding Wastage Absorber See activities from 4.2-1.3, 4.1-1.4, 4.2-1.5 Total Diametral Cladding Strain Absorber See activities from 4.2-1.1 4.2-1.12 Peak Cladding Temperature Fuel

((

))(a)(4)

Fuel

((

))(a)(4)

Peak Cladding/Absorber Temperature Absorber

((

))(a)(4)

Table 6-12. Design Basis Criteria and Supporting Information to Predict Fuel Failure Specific RAC Applicable Design Basis Criteria Available Supporting Data 4.2-2.1 Peak Cladding Temperature

((

))(a)(4) 4.2-2.2 Peak Fuel Temperature

((

))(a)(4) 4.2-2.3 Thermal Creep Strain of Cladding

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 1 Note RAC 4.2-1.9 and 4.2-1.10 are not applicable to fuel or absorber pins

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Specific RAC Applicable Design Basis Criteria Available Supporting Data 4.2-2.5 Cladding Wastage (FCCI)1 See activities from 4.2-1.5 Cladding Wastage (Eutectic)1 See activities from 4.2-2.1 Cladding Wastage (Na Corrosion)1 See activities from 4.2-1.4 Cladding Wastage (Fretting)1 See activities from 4.2-1.3 Table 6-13. Design Basis Criteria and Supporting Information to Ensure Fuel Pin Coolability and Absorber Pin Insertability Specific RAC Applicable Design Basis Criteria Applicable Pin Available Supporting Data 4.2-3.1 Total Diametral Strain of Cladding Fuel See activities from 4.2-2.3 4.2-3.2 Peak Cladding Temperature Fuel See activities from 4.2-2.1 4.2-3.3 Total Diametral Strain of Cladding Fuel See activities from 4.2-2.3 4.2-3.4 Peak Fuel Temperature Fuel See activities from 4.2-2.2 4.2-4.1 Total Diametral Strain Absorber See activities from 4.2-1.11 Peak Absorber Temperature Absorber Peak Cladding Temperature Absorber 1 A single limit is set for wastage; however, all contributions (fretting, eutectic interactions, Na corrosion, FCCI) are assessed in a conservative manner and combined to verify the total wastage limit is not exceeded.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-14. Summary of Future Testing Activities to Validate Fuel Damage Limits RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References 4.2-1.1 Total Diametral Clad Strain

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[107] [80] [79]

[82] [81] [83]

((

))(a)(4)

((

))(a)(4)

[70] [98]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[84] [86]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[108]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[109] [110]

4.2-1.2 Maximum allowable fuel pin fatigue cycles

((

))(a)(4) ((

))(a)(4)

((

))(a)(4) 1 Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References 4.2-1.3 Cladding Wastage -

Fretting

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[111] [112]

4.2-1.4 Cladding Wastage - Na Corrosion

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[74]

((

))(a)(4)

((

))(a)(4) 4.2-1.5 Cladding Wastage -

FCCI

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[48] [74] [113]

((

))(a)(4)

((

))(a)(4)

[114]

Cladding Wastage -

ACCI

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[115]

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References

((

))(a)(4) 4.2-1.6 Total Diametral Cladding Strain

((

))(a)(4) 4.2-1.8 Total cladding strain and thermal creep strain limits require internal fuel pin pressure to be assessed

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[74]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[114]

Table 6-15. Summary of Future Testing Activities to Predict Fuel Failure RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References 4.2-2.1 Peak cladding temperature limit

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[116]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 1 Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.2 Peak fuel temperature limit

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

[117]

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References

((

))(a)(4) 4.2-2.3 Cladding strain-thermal creep

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-2.5 Cladding Wastage -

FCCI See activities from 4.2-1.5 Cladding Wastage -

Eutectic See activities from 4.2-2.1 Cladding Wastage -Na Corrosion See activities from 4.2-1.4 Cladding Wastage -

Fretting See activities from 4.2-1.1

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern References Cladding Wastage -

ACCI See activities from 4.2-1.5 Table 6-16. Summary of Future Testing Activities to Ensure Fuel Coolability is Maintained RAC Design Basis Criteria Identified Activity1 Main Objectives Primary Factors of Concern 4.2-3.1 Total Diametral Cladding Strain-Coolability

((

))(a)(4)

((

))(a)(4)

((

))(a)(4) 4.2-3.2 Peak Cladding Temperature See activities from 4.2-2.1 4.2-3.3 Total Diametral Cladding Strain-Coolability See activities from 4.2-3.1 4.2-3.4 Peak Fuel Temperature See activities from 4.2-2.2 1 Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-17 Summary of Tests to Address High-Importance Fuel and Absorber Pin Phenomena High-Importance Phenomena Applicable Design Limit Overview of Testing1 Fission gas release Total Peak Cladding Strain, Peak Cladding Thermal Creep

((

))(a)(4)

HT9 mechanical response as a function of temperature, stress, irradiation, and time Total Peak Cladding Strain, Fatigue Limit, Peak Cladding Thermal Creep Strain

((

))(a)(4)

FCCI/ACC Cladding Wastage, Peak Cladding Thermal Creep Strain

((

))(a)(4)

Fuel Thermal Conductivity Peak Fuel Temperature

((

))(a)(4) 6.3.1 Fuel Assembly Mechanical Test Plans Existing experimental data can be used if adequately justified. If no data exists or the existing data is insufficient, a test program will be developed to validate the numerical models of the fuel 1 Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

2 ((

))(a)(4).

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 assembly, which can be categorized into the following areas: 1) sub-component tests (SC), 2) single assembly tests (SA), 3) multiple assembly tests (MA), 4) control assembly tests (CA), and

5) design proof of concept tests (DPC). Test categories 1-4 are described in section 6.3.1.1 through 6.3.1.3 and Table 6-19, and DPC tests (5) are described in section 6.3.1.4. The envisioned tests for fuel and control assemblies are summarized below in these respective categories.

6.3.1.1 Sub-Component Tests 6.3.1.1.1 Bundle-Duct Interaction Tests This series of tests will examine mechanical characteristics of the pin bundle including compression stiffness and pin redistribution under various loads. Data gathered from these tests will provide initial validation data in the elastic range and will be used for benchmarking of the OXBOW.BDI code.

6.3.1.1.2 Pin Bundle Bending Tests The purpose of this test is to gather initial pin bundle stiffness information and validate assumptions which are made in some core mechanical models. Additionally, bundle stiffness values gathered from this test can be useful in developing preliminary pin bundle models for other analyses.

6.3.1.1.3 Core Mechanical Duct Static Crush Tests The purpose of this test is to obtain the static compressive strength of the duct and load pads for use as a limit in core restraint system analysis. This would prevent damage to core assembly pin bundles.

6.3.1.1.4 Duct Dynamic Crush Tests The purpose of this test is to obtain the dynamic crush strength of the duct and load pads for use as a limit in seismic analyses. This would prevent damage to core assembly pin bundles.

6.3.1.1.5 Nozzle/Receptacle Interaction Tests The purpose of this test is to investigate how an assembly nozzle interacts with its receptacle due to applied forces and moments. The lateral and rotational stiffness of this interaction dictates the displacements of core assemblies.

6.3.1.2 Single Assembly Tests 6.3.1.2.1 Single Assembly Static Load Deflection Test The purpose of this series of tests is to investigate and characterize the mechanical behavior of SFR core assemblies. Tests will focus on mechanical behavior (bending stiffness, range of motion, etc.) of single fuel and control assemblies in the elastic range as well as thermal effects. This will include both nominal and plastically deformed test assemblies. Plastically deformed assemblies will resemble assemblies with residual deformation due to bowing and dilation.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 6.3.1.2.2 SCRAM Time/Impact Tests This series of tests will examine control rod insertability and SCRAM time for the control rod system. Frictional behavior would be developed using component-level testing in air to build the model. This would be combined in the future with friction testing in sodium to predict performance in sodium. Withdrawal/insertion loads of the control rod bundle, frictional loads, and SCRAM time of both deformed and undeformed geometries in air will be measured for model benchmarking.

6.3.1.2.3 Control Assembly Static Load Deflection Tests The purpose of this series of tests is to investigate and characterize the mechanical behavior of a control assembly. Tests will focus on mechanical behavior (bending stiffness, range of motion, etc.) of single assemblies in the elastic range as well as thermal effects.

6.3.1.2.4 Control Assembly Withdrawal & Insertion Tests The purpose of this series of tests is to investigate and characterize the mechanical interactions of a control assembly with respect to core assembly handling loads. This includes removing both deformed and undeformed assemblies from variously configured clusters of neighboring assemblies within a test apparatus.

6.3.1.2.5 Control Assembly Seismic SCRAM Tests This series of tests will examine control rod insertability and SCRAM time for the control rod system under a seismic event. Withdrawal/insertion loads of the control rod bundle, frictional loads, and SCRAM time of both deformed and undeformed geometries will be measured under various excitations for model benchmarking.

6.3.1.2.6 Single Assembly Free and Forced Vibration Tests The purpose of this series of tests is to examine fundamental dynamic characteristics of Natrium core assemblies. This includes parameters such as natural frequency and structural damping.

6.3.1.2.7 Single Assembly Pluck Impact Test The purpose of this test is to characterize dynamic impact behavior between assemblies at the top and above core load pads to calibrate contact behavior in seismic models.

The purpose of this test is to obtain the dynamic crush strength of the duct and load pads for use as a limit in seismic analyses. This would prevent damage to core assembly pin bundles.

6.3.1.3 Multiple Assembly Tests 6.3.1.3.1 Multiple Assembly Load Deflection Tests The purpose of this series of tests is to investigate and characterize the mechanical behavior of SFR core assemblies with a focus on phenomena relevant to core restraint system design.

Tests will focus on mechanical behavior (bending stiffness, range of motion, contact, etc.) of assemblies in the elastic range as well as thermal effects. Testing will include single assemblies as well as arrays of multiple assemblies.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 6.3.1.3.2 Multiple Assembly Row and ((

))(a)(4) Cluster Seismic Tests This series of tests will involve multiple assemblies in row or clustered configurations which will be subjected to excitation. The highly nonlinear response of the system due to inter-assembly gaps will be characterized using assembly displacement histories and contact/reaction loads and used for model benchmarking.

6.3.1.3.3 Multi-Assembly Withdrawal/Insertion Tests The purpose of this series of tests is to investigate and characterize the mechanical interactions of SFR core assemblies with respect to core assembly handling loads. This includes removing both deformed and undeformed assemblies from variously configured clusters of neighboring assemblies within a test apparatus. The testing report will include data such as withdrawal, insertion, contact, and reaction loads.

6.3.1.4 Design Proof of Concept Tests 6.3.1.4.1 CRD/CRA Flow Induced Vibration The purpose of this test is to evaluate the flow-induced vibration characteristics of the CRD and CRA under various flow conditions, and to evaluate fretting wear characteristics.

6.3.1.4.2 Control Rod Assembly Design Proof Testing The purpose of this test is to evaluate the function of the control assembly/driveline coupling when subjected to misalignment and demonstrate the design intent.

6.3.1.4.3 Core Inlet Design Proof Test The purpose of this test is to examine nozzle and receptacle fit at min/max material conditions, and to quantify wear on each component. Furthermore, mechanical hold-down force and drainage will be tested as well.

6.3.1.4.4 Lead Demonstration Assembly Design Proof and Duct Joint Load Testing The purpose of this test is to test the connection of the Lead Demonstration Assembly pins for removal, duct disassembly and reconstruction, duct joint design verification, and duct joint load limit identification.

6.3.1.4.5 Pin Wire Wrap Mechanical Test The purpose of this test is to determine the structural integrity of the wire wrap weld and bend regions under static and dynamic loading conditions.

6.3.1.4.6 Orifice Plate Structural Integrity Test The purpose of this test is to determine the structural integrity of the orifice plate under static and cyclic (fatigue) loading conditions.

6.3.1.4.7 Dashram/Dashpot Deceleration Test The purpose of this test is to determine the spring/damper dynamic performance of the control assembly dashpot.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 6.3.1.4.8 Vertical Drop to Base Test The purpose of this test is to develop a data set that can be used to validate FEA models of a core assembly drop to the receptacle during refueling operations, including dynamic impact behavior.

6.3.1.4.9 Vertical Drop to Grapple The purpose of this test is to develop a data set that can be used to validate FEA models of an impact between the core assembly and the IVTM grapple fingers, including impact force and position time histories.

6.3.1.5 Major Effects on Single and Multiple Fuel Assembly Test The major effects identified for fuel and core assembly behavior are thermal gradient effects (TE), irradiation effects (IE), and fixity effects (FE), as summarized in Table 6-18. These effects must be accounted for when assessing assembly behavior. A notional test matrix is provided in Table 6-19 summarizing all of the envisioned tests for each of the major test categories, including the applicable major effects. Pending additional design and analysis effort the test matrix will be updated to ensure all high-importance phenomena with unknown and partial knowledge levels identified in the PIRTs [42] are adequately addressed.

Table 6-18. Major Effects on Fuel Assembly Behavior Type Descriptions Test Configurations Thermal Gradient Effects (TE)

Thermal gradients across a fuel assembly in the lateral direction induces fuel assembly bow Electric heaters will be attached to the duct outer surfaces Irradiation Effects (IE)

Fluence gradients across fuel assembly in lateral or vertical directions induce fuel assembly bow and/or dilation Fuel assembly duct tubes will be pre-deformed as needed Fixity Effects (FE)

Gap conditions at the boundary condition (i.e., inlet nozzle to receptacle interface) affect fuel assembly rotational stiffness Normal, loose, and tight gap condition will be used for comparison

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-19. Example Fuel & Control Assembly Mechanical Test Matrix Type Test BOL TE IE FE Critical Characteristics SA MA SC CA SA: Single Assembly MA: Multiple Assembly SR: Single Row TE: Thermal Gradient Effects SC: Sub-Component FE: Fixity Effects (Normal, Loose, Tight)

CA: Control Assembly BOL: Beginning-Of-Life IE: Irradiation effects accounted for by using pre-deformed ducts to simulate dilation and/or bowing induced by irradiation creep and growth. A hydraulic forming technique and device have been developed for duct deformation.

Note that irradiation effects should be treated conservatively and will be by validated by surveillance programs.

6.3.2 Materials Property Data and Testing The scope of qualification of materials property data includes the materials selected for each of the fuel components, as well as hard-face coatings and weldments applied to them.

Several handbooks compiling materials property data of HT9 for design input have been produced by US national laboratories, including the Nuclear Systems Materials Handbook (NSMH) [118], the Fuel Cycle Research and Development (FCRD) Materials Handbook under the Advanced Fuel Cycle Initiative (AFCI) [18], and the Generation IV Materials Handbook [119].

Similar compilations have been performed for metallic fuels, including the Metallic Fuels (a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Handbook [120], Thermophysical Properties of Matter Data Series [121], Metallic Fuels Handbook, Part 1 [122], and the Current revision of the Metallic Fuels Handbook summarizing properties of fresh fuels [123]. These handbooks contain a significant volume of information needed for design and analysis.

Additionally, TerraPower is reviewing literature references reporting on testing and characterization of the selected materials, as appropriate. These references are being catalogued in databases to determine their utility to contribute information for determination of physical/thermophysical, mechanical, and corrosion properties, as well as microstructure.

The qualification of legacy data will be performed following established procedures involving the four methods previously outlined in Section 6.2.1. Qualified data and applicable correlations will be formally documented in the Natrium Materials Handbook, an update to the current TerraPower Materials Handbook. Metallic fuel material (i.e., U-10Zr) properties data and applicable correlations will be documented in the Fuel Material Properties Handbook.

The method of qualification of materials data will depend on the availability of an existing database, availability of additional testing capabilities, and the safety function of the component for which the material data is a critical characteristic. Design inputs that take values directly from NRC or NNSA accepted standards such as ASME Boiler Pressure Vessel Code are considered pre-qualified by quality assurance program equivalency. Additional details are provided in the following paragraphs.

Qualification of material properties data by corroboration will rely on experimental data from multiple sources to demonstrate sufficient justification in the best estimate and uncertainty material property values that are used as design input for the fuel design. The data used must be properly referenced and verified by an independent reviewer before being accepted for qualification. The data for qualification by corroboration shall be taken from at least two or more references at minimum to meet the criteria of sufficient quantity.

Qualification of materials properties data by confirmatory testing will rely on well controlled and well documented testing performed in compliance with the TerraPower QA program. The testing may be subcontracted to another vendor with a 10 CFR 50 Appendix B (Appendix B) or equivalent quality assurance program or be commercially dedicated by TerraPower as an approved vendor for that specific testing service. Vendors who have Appendix B equivalent quality assurance programs must undergo TerraPower audits. Vendors without Appendix B equivalent programs may still provide the specific confirmatory testing service after a commercial grade dedication by TerraPower. The data obtained by confirmatory tests must be reviewed and accepted by TerraPower.

Qualification of materials properties data by peer review will rely on the expertise and professional judgement of a review team. The review team will be made up of a minimum two peer review members with relevant technical experience or expertise relating to the test method, history, standard, and/or analysis relating to the data. A peer review plan will be drafted in addition to the qualification plan by the responsible engineer to define the scope of the data review and provide all relevant supporting information to the review team prior to official review activities.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Qualification of materials properties data may be achieved by evaluating the quality program under which it was acquired and making positive correlations between that quality program and Appendix B. An example would be qualification of reactor materials data acquired under DOE programs.

Once the data has been qualified, the data will be labelled as controlled such that the information can be accessed as qualified data and accepted for use as design inputs.

6.3.2.1 HT9 Data for Qualification Specifically, for HT9, this section identifies the required materials properties data that will impact the RAC under current TerraPower scope, and the intended methods for qualification for each dataset. The qualification of existing data procedure [76] will be used to qualify existing materials data that act as design input for the Natrium Reactor. A total of 21 qualification of existing data documents are identified for HT9 to support RAC associated with the reactor core. For organization and prioritization, the 21 qualification of existing data documents are binned into four property types: thermal properties, unirradiated mechanical properties, irradiated mechanical properties, and chemical interactions.

Table 6-20. Summary of HT9 Data Qualification Property Type Data Type Associated RAC Qualification Method Thermal Thermal expansion coefficient 4.2-1.1, 4.3-7

((

))(a)(4)

Heat capacity 4.2-2.1, 4.2-3.2

((

))(a)(4)

Thermal conductivity 4.2-2.1, 4.2-3.2

((

))(a)(4)

Melting point 4.2-2.1, 4.2-2.3, 4.2-2.5, 4.2-3.2

((

))(a)(4)

Unirradiated mechanical Youngs Modulus (unirradiated) 4.2-1.1, 4.2-3.1, 4.2-3.3, 4.2-3.5, 4.2-4.1

((

))(a)(4)

Yield strength (unirradiated) 4.2-1.1, 4.2-1.8, 4.2-2.3, 4.2-2.4, 4.2-3.5

((

))(a)(4)

Fracture toughness (unirradiated) 4.2-1.1, 4.2-1.8

((

))(a)(4)

Fatigue (unirradiated) 4.2-1.2

((

))(a)(4)

Creep (unirradiated) 4.2-1.1, 4.2-1.2, 4.2-1.6, 4.2-1.7, 4.2-2.1, 4.2-2.3, 4.2-3.1, 4.2-3.3, 4.2-3.5, 4.2-4.1

((

))(a)(4)

Thermal aging 4.2-1.1, 4.2-1.8, 4.2-2.3, 4.2-3.5, 4.2-4.1

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Property Type Data Type Associated RAC Qualification Method Friction and wear 4.2-1.3, 4.2-2.5

((

))(a)(4)

Irradiated mechanical Irradiated yield strength 4.2-1.1, 4.2-1.8, 4.2-2.3, 4.2-2.4, 4.2-3.5

((

))(a)(4)

Irradiated fracture toughness 4.2-1.1 4.2-1.8

((

))(a)(4)

Irradiation fatigue 4.2-1.2

((

))(a)(4)

Irradiation creep 4.2-1.1, 4.2-1.2, 4.2-1.6, 4.2-1.7, 4.2-2.3, 4.2-3.1, 4.2-3.3,4.2-3.5, 4.2-4.1

((

))(a)(4)

Irradiation swelling 4.2-1.1,4.2-1.6, 4.2-1.7, 4.2-3.1, 4.2-3.3, 4.2-3.5, 4.2-4.1

((

))(a)(4)

Stress enhanced irradiation swelling 4.2-1.1, 4.2-3.1, 4.2-3.3, 4.2-3.5, 4.2-4.1

((

))(a)(4)

Chemical interactions Cladding coolant compatibility (Sodium corrosion and erosion) 4.2-1.4, 4.2-2.5

((

))(a)(4)

Cladding fuel compatibility 4.2-1.5, 4.2-2.3, 4.2-2.5

((

))(a)(4)

Cladding absorber compatibility 4.2-1.7, 4.2-1.8

((

))(a)(4)

Cladding reflector compatibility 4.2-1.6

((

))(a)(4) 6.3.2.1.1 Thermal Properties Four HT9 thermal properties have been identified as design inputs for demonstrating the safety function of safety-significant components in the Natrium design. They must be qualified to demonstrate regulatory compliance.

The design inputs are melting point, coefficient of thermal expansion, heat capacity and thermal conductivity. All the properties have been measured historically.

Thermal properties for steels are well understood and mainly depend on the atomic bonding strength of the crystal lattice of the bulk material. Therefore, the thermal properties should not exhibit significant heat to heat variation. This widens the existing thermal properties database

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 to allow inclusion of any body-center cubic (BCC) FM steel with nominal composition of 12Cr-1MoVW to be used as corroborative data for TerraPower HT9.

6.3.2.1.2 Unirradiated Mechanical Properties Seven HT9 unirradiated mechanical properties have been identified as design inputs for demonstrating the safety function of safety-significant components in the Natrium Reactor design: Youngs modulus, yield strength, fracture toughness, thermal creep, fatigue, thermal aging, friction and wear.

All these properties will have existing data that serve as design input that require qualification and can be confirmed by independent testing on TerraPower HT9.

Mechanical properties for steels are well understood and they can be highly variable depending on the final microstructure of the commercial product. Unlike thermal properties, it is generally not appropriate to directly use mechanical properties of steels that are not within the known HT9 specification to act as corroborative data for TerraPower HT9. In special cases where there is significant lack of existing data, technical justification must be made for the inclusion of data from materials outside of HT9s specification as part of the qualification of existing data process [124].

If needed, confirmatory tests could be conducted on TerraPower HT9 in addition to existing data. Any confirmatory tests conducted on TerraPower HT9 shall follow 10 CFR 50 Appendix B quality assurance requirements, and the test methods employed will be conducted in accordance with existing standards where appropriate.

6.3.2.1.3 Irradiated Mechanical Properties Six HT9 irradiated mechanical properties have been identified as design inputs for demonstrating the safety function of safety-significant components in the Natrium Reactor:

irradiated yield strength, irradiated fracture toughness, irradiation creep, irradiation swelling, stress enhanced swelling, and irradiation fatigue. Some of those properties have been measured historically and others have not.

TerraPower is conducting mechanical testing of HT9 materials irradiated in sodium-cooled fast neutron test reactors to relevant exposures. However, due to the limitation of existing prototypical data and difficulty in conducting mechanical tests on irradiated samples, advanced testing methods and analysis techniques may be included as corroborative data to support assumptions and confirm expected materials behavior.

It is recognized that almost no confirmatory testing on irradiated material will be fully prototypical. Therefore, the purpose of qualification of existing data and the performance of relevant confirmatory testing is to establish reasonable assurance that HT9 performance under irradiated conditions is within the design margins of the Natrium Reactor. Any confirmatory tests conducted on TerraPower HT9 shall follow 10 CFR 50 Appendix B quality assurance requirements, and the test methods employed will be conducted in accordance with existing standards where appropriate.

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 6.3.2.1.4 Chemical Interactions Three HT9 chemical interactions have been identified as design inputs for demonstrating the safety function of safety-significant components in the Natrium Reactor design: cladding coolant compatibility, cladding-fuel compatibility, cladding-absorber compatibility. Some of those properties have been measured historically and others have not.

TerraPower is conducting testing of prototypical fuel pins that have been irradiated in test reactors, which provides results relevant to chemical interactions. However, due to the limitation of existing prototypical data and difficulty in conducting chemical interaction tests on irradiated samples, advanced testing methods and analysis techniques may be included as corroborative data to support assumptions and confirm expected materials behavior.

It is recognized that almost no confirmatory testing on irradiated material will be fully prototypical and the purpose of qualification of existing data is to establish reasonable assurance that HT9 performance under irradiated conditions are within the design margins of the Natrium Reactor. Any confirmatory tests conducted on TerraPower HT9 will follow 10 CFR 50 Appendix B quality assurance requirements, and the test methods employed will be conducted in accordance with existing standards where appropriate.

6.3.2.1.5 Availability of HT9 Materials Property Data With respect to the properties discussed above, the table below summarizes data availability from different sources, with a qualitative assessment of gaps.

Table 6-21. Availability of Data for HT9 Property Data Source Availability Fundamental Physical and Thermophysical Properties Coefficient of Thermal Expansion (CTE)

TerraPower Materials Handbook, Open Literature

((

))(a)(4)

Density TerraPower Materials Handbook, Open Literature

((

))(a)(4)

Specific Heat Capacity TerraPower Materials Handbook, Open Literature

((

))(a)(4)

Thermal Conductivity TerraPower Materials Handbook, Open Literature

((

))(a)(4)

Time-Temperature Dependence of Structure and Phases FCRD Handbook, Open Literature

((

))(a)(4)

Melting Point FCRD Handbook, Open Literature

((

))(a)(4)

Emissivity FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Property Data Source Availability Fundamental Physical and Thermophysical Properties Electrical Resistivity FCRD Handbook

((

))(a)(4)

Mechanical Properties Young's Modulus TerraPower and FCRD Handbooks, Open Literature

((

))(a)(4)

Shear Modulus FCRD Handbook, Open Literature

((

))(a)(4)

Poisson's Ratio TerraPower and FCRD Handbooks, Open Literature

((

))(a)(4)

Yield Strength (YS)

TerraPower, FCRD, and NSMH Handbooks, Open Literature

((

))(a)(4)

Ultimate Tensile Strength (UTS)

TerraPower, FCRD, and NSMH Handbooks, Open Literature

((

))(a)(4)

Tensile Stress-Strain Curves ACFI Handbook, Open Literature

((

))(a)(4)

Uniform Elongation FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Total Elongation FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Fracture Toughness FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Ductile-to-Brittle Transition Temperature FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Creep FCRD and NSMH Handbooks, TerraPower Constitutive Model Report, Open Literature

((

))(a)(4)

Stress-Rupture of Pressurized Tubes FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Fatigue FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Property Data Source Availability Fundamental Physical and Thermophysical Properties Creep Fatigue FCRD and NSMH Handbooks, Open Literature

((

))(a)(4)

Wear Rate N/A

((

))(a)(4)

Coefficient of Friction N/A

((

))(a)(4)

Irradiation Effects On all other properties (especially hardening, such as yield strength, fracture toughness, DBTT, irradiation creep, etc.)

FCRD handbook, and open literature

((

))(a)(4)

On Structure and Phases FCRD handbook, and open literature

((

))(a)(4)

Void Swelling FCRD handbook, and open literature

((

))(a)(4)

Environmental Compatibility Fuel-Cladding Chemical Interaction (FCCI)

TerraPower planned testing

((

))(a)(4)

Absorber-Cladding Chemical Interaction (ACCI)

TerraPower planned testing

((

))(a)(4)

Compatibility with Sodium (Corrosion, Erosion)

TerraPower planned testing

((

))(a)(4) 6.3.2.2 Coatings and Weldments Once selection of hard-face coatings and development of welding processes is completed, all the relevant properties for those features will be qualified using a similar methodology to that presented above for HT9 steel components.

6.3.2.3 Metallic Fuel Properties Data for Qualification This section identifies the required material properties data that are needed to support design and analysis activities. Except for fuel thermal conductivity, none of the fuel material properties were identified as high-importance phenomena in the PIRT assessments relative to assessing fuel pin design criteria. In spite of the other fuel properties being of less direct importance for evaluating fuel pin design criteria, other analyses are heavily dependent on these properties (i.e.,

neutronics assessments) and having reliable materials property data is essential to characterizing the beginning of life conditions of the fuel system. The applicable fresh fuel properties of interest and the planned qualification method are summarized in Table 6-22. Irradiated fuel material properties are assessed by the fuel performance modeling tools, with detail provided in Section 6.4.1. [76]

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-22. Summary of U-10Zr Material Properties Data to be Qualified Property Type Data Type Dependencies/Details Qualification Method Thermal Density Zirconium concentration Temperature

((

))(a)(4)

Thermal expansion Linear thermal expansion Mean coefficient of thermal expansion

((

))(a)(4)

Heat capacity Temperature Zirconium concentration Includes enthalpies of phase transformation including enthalpy of fusion

((

))(a)(4)

Thermal conductivity Zirconium concentration Temperature Porosity Sodium infiltration Fission products

((

))(a)(4)

Melting point Solidus temperature Liquidus temperature Zirconium concentration Plutonium concentration

((

))(a)(4)

Unirradiated mechanical Youngs Modulus (unirradiated)

Zirconium concentration Temperature

((

))(a)(4)

Thermal Creep (unirradiated)

Zirconium concentration Temperature

((

))(a)(4) 6.4 Analytical Predictions Analytical predictions are a component of the overall safety assessment of the reactor. Fuel performance models are used to evaluate fuel design criteria under normal operation and accident scenarios. This evaluation may be used, for example, to determine whether the fuel pin is damaged, to quantify the number of fuel pin failures, or to determine if fuel melting has occurred. Fuel performance models rely on inputs from other tools and methodologies to set boundary conditions and field variables such as cladding surface temperature, power, burnup, DPA or fluence, and the spatial and temporal variations of these parameters. Similarly, performance models of control, shield, and reflector pins are used to evaluate design criteria for those components. While discussion of analysis methodologies is outside the scope of this document, information is provided here regarding fuel performance models to inform RAC related to design evaluations of fuel pins. Section 6.4.1 provides a high-level overview of the performance models used to assess pin design criteria and the associated tools or methodologies used for key inputs. To illustrate the planned approach to address

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 AQAF Appendix A Evaluation Model (EM) Goals, Table 6-28 maps AQAF EM goals and key phenomena to tools and methodologies.

Analytical predictions of core assembly mechanical behavior are important to ensuring compliance with safety requirements and specific design criteria. Finite element models are generated using TerraPower developed pre-processing software (OXBOW) to automate mesh generation and application of loads and boundary conditions in accordance with established methods. Section 6.4.2 summarizes these methods and introduces software tools used for conducting analysis.

The TerraPower Engineering Design and Analysis Software Management Procedure [125] provides the software quality assurance (SQA) work activities required for the planning, acquisition, development, operation, maintenance, and retirement of evaluation models. This procedure is being followed for the development and use of the modeling tools described in the following sub-sections helping ensure the quality of the final evaluation models. Three different types of documents are created as part of this procedure (for internally developed software) to help ensure The evaluation model contains the appropriate modelling capabilities [EM G1]: 1) Software Requirements Specification Documents are prepared in advance of software development to ensure all of the software requirements (including appropriate geometries, materials, and physics) are identified and independently reviewed, 2) Software Design and Implementation Documents are prepared to document how these requirements are met and implemented in the code, and 3) Software Test Reports are issued to demonstrate that the required physics and requirements are properly implemented. Benchmark Comparison Reports and Test Reports are the main documents generated by this procedure to help ensure The evaluation model has been adequately assessed against experimental data [EM G2]. Additional guidance has been developed for the Natrium project for performing software verification and validation [126], as well for methodology development and assessment [127].

6.4.1 Pin Performance Models Two independent fuel performance models (ALCHEMY and ((

))(a)(4)) have been matured for the purpose of predicting fuel pin damage and failure for use in the core design and safety analysis methodologies. These two models were developed independently, utilize different numerical methods, and differ in the approaches to modeling certain phenomena. This allows for independent verification of analysis predictions and diversity in evaluation approaches. In addition, a third model (CRUCIBLE) is used for capturing changes to fuel pins during normal operation, which impact neutronic and thermal hydraulic calculations. The ALCHEMY model is also capable of modeling control, shield, reflector, and Type 1B pin performance. The following sections describe these models and their capabilities. The use cases for each model are summarized in Table 6-23. A mapping of applicable fuel performance models or methodologies to high-importance phenomena used to assess damage, failure, coolability, and fuel melting is provided in Table 6-24, Table 6-25,Table 6-26, and Table 6-27, respectively. FQAF goals related to fuel performance model are addressed at a high-level in Table 6-28.

Table 6-23. Applicable Models and Codes for Fuel Pin Phenomena Use Case Applicable Tool(s)

Ability to meet fuel pin design limits during normal operation to preclude fuel pin damage. Pre-transient characterization of fuel condition.

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Use Case Applicable Tool(s)

Ability to meet fuel pin design limits during normal operation to preclude fuel pin damage. Pre-transient characterization of fuel condition.

((

))(a)(4)

Prediction of fuel pin damage and failure during accidents.

((

))(a)(4)

Impact of fuel performance phenomena on normal operation neutronic and thermal hydraulic behaviors (e.g., fuel axial growth impact on core fuel density distribution).

((

))(a)(4)

Impact of fuel performance phenomena on transient conditions during accident (e.g., fuel axial growth impact on core fuel density distribution).

((

))(a)(4)

Ability to meet control, shield, and reflector pin performance limits during normal operation and postulated accidents.

((

))(a)(4)

Table 6-24. Fuel Performance Prediction Capabilities to Assess Fuel Damage Applicable Design Limit Applicable RAC High-Importance Phenomena/

Parameters Applicable Tool/Methodology Total Peak Cladding Strain 4.2-1.1, 4.2-1.6, 4.2-1.8 Fission gas release

((

))(a)(4)

HT9 mechanical response as a function of temperature, stress, irradiation, and time

((

))(a)(4)

FCCI

((

))(a)(4)

Fuel burnup

((

))(a)(4)

DPA on cladding

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Cladding Fatigue Lifetime 4.2-1.2 HT9 mechanical response as a function of temperature, stress, irradiation, and time

((

))(a)(4)

Fuel burnup

((

))(a)(4)

DPA on cladding

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Number of strain cycles on cladding

((

))(a)(4)

Magnitude of strain cycles

((

))(a)(4)

Cladding Wastage 4.2-1.3, 4.2-1.4, 4.2-1.5 FCCI

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Residence time

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Table 6-25. Fuel Performance Prediction Capabilities to Assess Fuel Failure Applicable Design Limit Applicable RAC High-Importance Phenomena/

Parameters Applicable Tool/Methodology Peak Cladding Temperature 4.2-2.1 Detailed pin level irradiation histories including power and cladding temperature

((

))(a)(4)

Detailed coolant transient temperature and pin power histories

((

))(a)(4)

Peak Fuel Temperature 4.2-2.2 Fuel thermal conductivity

((

))(a)(4)

Detailed pin level irradiation histories including power and cladding temperature

((

))(a)(4)

Detailed coolant transient temperature and pin power histories

((

))(a)(4)

Peak Cladding Thermal Creep Strain 4.2-2.3 Fission gas release

((

))(a)(4)

HT9 mechanical response as a function of temperature, stress, irradiation, and time

((

))(a)(4)

FCCI

((

))(a)(4)

Fuel burnup

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Cladding Wastage 4.2-2.5 FCCI

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Residence time

((

))(a)(4)

Table 6-26. Fuel Performance Prediction Capabilities to Assess Fuel Coolability Applicable Design Limit Applicable RAC High-Importance Phenomena/ Parameters Applicable Tool/Methodology Total Peak Cladding Strain 4.2-3.1, 4.2-3.3, 4.2-3.5 Fission gas release

((

))(a)(4)

HT9 mechanical response as a function of temperature, stress, irradiation, and time

((

))(a)(4)

Fuel burnup

((

))(a)(4)

Cladding temperatures

((

))(a)(4)

Peak Cladding Temperature 4.2-3.2 Detailed pin level irradiation histories including power and cladding temperature

((

))(a)(4)

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Applicable Design Limit Applicable RAC High-Importance Phenomena/ Parameters Applicable Tool/Methodology

((

))(a)(4)

Detailed coolant transient temperature and pin power histories

((

))(a)(4)

Peak Fuel Temperature 4.2-3.4 Fuel thermal conductivity

((

))(a)(4)

Detailed pin level irradiation histories including power and cladding temperature

((

))(a)(4)

Detailed coolant transient temperature and pin power histories

((

))(a)(4)

Table 6-27. Fuel Performance Prediction Capabilities to Assess Phenomena Related to Fuel Temperatures Phenomena/Parameters Applicable Tool/Methodology Radial power distribution due to constituent redistribution ((

))(a)(4)

Fuel temperature distribution

((

))(a)(4)

Cladding temperature distribution

((

))(a)(4)

Burnup distribution in the fuel

((

))(a)(4)

Thermal conductivity of the fuel and cladding

((

))(a)(4)

Thermal expansion of the fuel and cladding

((

))(a)(4)

Fission gas production and release

((

))(a)(4)

Solid and gaseous fission product swelling

((

))(a)(4)

Fuel deformation

((

))(a)(4)

Diffusion of fuel constituents

((

))(a)(4)

Fuel and cladding dimensional changes

((

))(a)(4)

Fuel-to-cladding heat transfer

((

))(a)(4)

Fuel-to-cladding contact pressure

((

))(a)(4)

Heat capacity of the fuel and cladding

((

))(a)(4)

Swelling and creep of the cladding

((

))(a)(4)

Rod internal gas pressure

((

))(a)(4)

Rod internal gas composition

((

))(a)(4)

Cladding-to-coolant heat transfer coefficient

((

))(a)(4)

Cladding wastage (erosion, corrosion)

((

))(a)(4)

FCCI

((

))(a)(4)

Table 6-28. Fuel Performance Models for FQAF Goals

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID Evaluation Model (EM) Assessment Framework Goal Description Natrium Fuel Performance Modeling Approach EM G1 Evaluation model contains the appropriate modeling capabilities Model requirements have been informed by design criteria and PIRT analysis of important phenomena necessary for the prediction of design criteria. Evaluation models will demonstrate appropriate capabilities have been implemented through software testing.

EM G1.1 Evaluation model is capable of modeling the geometry of the fuel system Pin performance models typically utilize an axisymmetric (RZ) dimensionality that includes the fuel (or B4C), cladding, and plenum regions. The dependence of boundary conditions (e.g., cladding temperature) that vary along the circumference of the fuel pin are typically evaluated through sensitivity analysis.

Specialized models of the fuel cross-section or other dimensionalities are occasionally used for unique assessment on a case-by-case basis.

EM G1.2 Evaluation model is capable of modeling the material properties of the fuel system Fuel performance models include material models for U-10Zr and HT9 cladding.

Additional material models are used for benchmarking to historical experiments that may include U-Pu-Zr fuels and other cladding materials. Pin performance models for control, shield, reflector, and Type 1B pins employ material models applicable to the Natrium design.

EM G1.3 Evaluation model is capable of modeling the physics relevant to fuel performance PIRT assessments have been used to identify important phenomena necessary for fuel performance predictions. Code requirements specify the necessary physics for each model.

EM G2 Evaluation model has been adequately assessed against experimental data Initial validation assessments have demonstrated the ability to predict experiment data related to key design criteria such as cladding strain and pin failure. Detailed validation plans and assessments are under development to demonstrate that validation assessment criteria have been met.

EM G2.1 Data used for assessment are appropriate (see ED Assessment Framework)

Qualification of fuel performance data is being performed to qualify existing experiment data from EBR-II and FFTF.

New experiments are being performed under an appropriate quality program.

EM G2.2 Evaluation model is demonstrably able to predict fuel failure and degradation mechanisms over the test envelope Experiments that resulted in fuel failure are being used to validate fuel performance models. These include a small set of

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 FQAF Goal ID Evaluation Model (EM) Assessment Framework Goal Description Natrium Fuel Performance Modeling Approach failures that occurred during normal operation, transient overpower experiments, furnace experiences, and a variety of materials tests and sub-system experiments. New experiments are being designed to fill gaps in the experimental databases and results will subsequently be used for model validation.

EM G2.2.1 Evaluation model error is quantified through assessment against experimental data Validation plans include the determination of model error via assessment to experiment data.

EM G2.2.2 Evaluation model error is determined throughout the fuel performance envelope Validation plans include the determination of model error via assessment to experiment data.

EM G2.2.3 Sparse data regions are justified Justification of regions with sparse data will be provided.

EM G2.2.4 Evaluation model is restricted to use within its test envelope Coverage of validation assessments will be reported in software test reports.

6.4.1.1 ALCHEMY ALCHEMY is a thermo-mechanical model based on the finite element method capable of simulating fuel, absorber, shield, and reflector pin behavior within a reactor environment.

ALCHEMY generates the geometry and mesh for the problem, applies boundary conditions, models nuclear-specific material behavior and phenomena, and solves the coupled thermo-mechanical equations describing the physics being simulated. It has been developed to work in conjunction with the commercial-off-the-shelf finite element analysis software ABAQUS.

ALCHEMY includes a preprocessor, postprocessor, user subroutines that work in conjunction with ABAQUS, and several features to assist in analysis, verification, and validation. The preprocessor takes model input variables from a user-supplied input file and implements the ABAQUS Application Programming Interface (API) to automatically create the geometry, mesh, and boundary conditions of the model. Model solutions occur through ABAQUS, in conjunction with ALCHEMY-CORE, a set of user subroutines to extend the ABAQUS functionality. The post-processor is used to obtain simulation data from solution files. The pre-and post-processing is written in Python, whereas the user subroutines are written in FORTRAN.

The software allows the user to specify ((

))(a)(4). Fuel composition is generally a mixture of uranium, plutonium, and zirconium while the cladding materials are typically a steel alloy. This flexibility in choice of materials and geometry allows for the simulation of legacy fuel pin designs and experiments as well as current fuel pin designs. In addition, the user can specify a variety of operating conditions for normal operation, start-ups, shutdowns, and accident scenarios. ALCHEMY also provides capabilities to simulate separate effects experiments of metallic fuel or structural materials.

The software also has the capability to model boron carbide (B4C) in place of fuel, which can be used to simulate the behavior of Natrium absorber pins or shield pins. The Natrium reflector design (((

))(a)(4)) can also be modeled. Encapsulated experiments, such as those

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 used in the ATR can be modeled by specifying the capsule geometry and associated boundary conditions.

6.4.1.2 ((

))(a)(4)

((

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((

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((

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((

))(a)(4) 6.4.1.3 CRUCIBLE CRUCIBLE is the primary tool for modeling the axial growth of the fuel, changes to the sodium bond, and fission gas release during normal operation. These parameters influence the

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 composition of the core and geometry of the fuel, and thus influence the reactor normal operation neutronic and thermal hydraulic behaviors. These fuel performance phenomena are generally modeled using empirical correlations derived from data measured from historical metal fuel tests.

Due to the observed variability of these phenomena, CRUCIBLE can apply uncertainty factors that vary the empirical models within the observed variability. Additionally, CRUCIBLE can accept user-defined inputs for these phenomena, which allow for further sensitivity studies to assess the impact to normal operation conditions.

CRUCIBLE provides an interface for calculating fuel temperatures under simplifying assumptions compared to the more detailed ALCHEMY and ((

))(a)(4) models. This temperature calculation is used to determine the thermal expansion axial growth of the fuel and for assessing peak fuel temperatures considering uncertainties during normal operation.

CRUCIBLE is not expected to play a role in the evaluation of fuel pin design criteria, but is included here for completeness, and to highlight how these fuel performance phenomena are accounted for to support other core design activities.

6.4.2 Core Assembly Mechanical Analysis Predicting core assembly behavior (particularly for fuel assemblies) is an important aspect of sodium fast reactor (SFR) core mechanical design. One aspect of this is how displacements of fuel within the neutron flux gradients due to assembly deformations change the reactivity of the core. These reactivity feedbacks have effects on reactor safety and operation. An additional concern relates to the amount of deformation assemblies accumulate during their residence time in the core which affects handling operations and assembly useable lifetime. The types of deformations that pertain to an SFR core assembly are listed in subsequent sections. All core assemblies will be analyzed for these distortions, but generally fuel assemblies are the most limiting since they experience the highest neutron doses and temperature gradients in the core.

The mechanical performance and integrity of limiting assemblies may dictate shuffling operations or core management. Sections 6.4.2.1 through 6.4.2.6 summarize the core assembly phenomena of concern and the associated analysis methods and software used for conducting the corresponding mechanical analyses.

6.4.2.1 Core Assembly Distortion [128, 129]

Predicting core assembly behavior (particularly for fuel assemblies) is an important aspect of sodium fast reactor (SFR) core mechanical design. One aspect of this is how displacements of fuel within the neutron flux gradients due to assembly deformations change the reactivity of the core. These reactivity feedbacks have effects on reactor safety and operation. An additional concern relates to the amount of deformation assemblies accumulate during their residence time in the core which affects handling operations and assembly useable lifetime. The types of deformations that pertain to an SFR core assembly are listed in subsequent sections. All core assemblies will be analyzed for these distortions, but generally fuel assemblies are the most limiting since they experience the highest neutron doses and temperature gradients in the core.

The mechanical performance and integrity of limiting assemblies may dictate shuffling operations or core management.

6.4.2.1.1 Core Restraint System [128, 129]

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Evaluation of core-wide assembly bowing and interaction forces is important to predicting the performance of the Natrium core over its lifetime. As part of the CRS, the evolution of core assembly bowing and accumulated bowing deformations are used to predict the radial expansion reactivity feedback mechanism for a fresh core or a configuration that has undergone multiple cycles of operation. Inter-assembly interaction forces, which result from assembly bowing and dilation, inform handling loads for the various core assemblies as well as the loads that are reacted by the core support structures. The core restraint system is described in detail in Section 5.4.

6.4.2.1.2 OXBOW.CRS The TerraPower-developed code OXBOW.CRS will be used for analysis of core-wide assembly behavior and core restraint system analysis. This code utilizes assembly load history information in order to evaluate the evolution of core-wide assembly bowing and interaction forces. OXBOW.CRS makes use of the finite element method utilizing the commercially available solver ABAQUS. The finite element model (FEM) is created using geometry, boundary conditions, and loads provided by other physics codes and core component design.

A sub-module of OXBOW.CRS (OXBOW.SHUFFLE) is used to account for multi-cycle effects and refueling, and incorporates inelastic deformations predicted using OXBOW.CADA, described below.

6.4.2.1.3 OXBOW.CADA The TerraPower-developed code OXBOW.CADA will be used for core assembly distortion analysis. This code calculates duct dilation, duct axial growth, and core assembly bowing for a given load history. OXBOW.CADA makes use of the finite element method utilizing the commercially available solver ABAQUS. Custom creep and swelling user materials from the fuel performance code ALCHEMY are used in these ABAQUS models. These material subroutines incorporate thermal creep, irradiation creep, and void swelling correlations for stainless steels used in core design. The FEM is created using geometry, boundary conditions, and loads provided by other physics codes.

6.4.2.2 Core Seismic The seismic response of the reactor core system is important to the overall performance of the Natrium Reactor. It is important to be able to predict the reactivity response as well as the structural response of the core during a seismic event. The Natrium core will be analyzed for seismic licensing basis events (LBEs) of various severities.

6.4.2.2.1 OXBOW.SEISMIC The TerraPower-developed code OXBOW.SEISMIC will be used for analysis of core-wide assembly behavior under seismic loads. This code utilizes lateral seismic excitations in order to evaluate the evolution of core-wide assembly bowing and impact forces. OXBOW.SEISMIC makes use of the finite element method utilizing the commercially available solver ABAQUS.

The FEM is created using geometry, boundary conditions, and loads provided by the core design team.

6.4.2.3 Core Assembly Withdrawal/Insertion [128, 129]

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 It is important to be able to quantify the magnitude of the withdrawal and insertion loads that occur when performing refueling operations. An understanding of the magnitude of loads required to handle core assemblies is necessary for efficient core management. The event of a core assembly not being able to be removed or inserted should be avoided as it will result in the interruption of normal reactor operations and require a very costly intervention. Because economical operation of the Natrium Reactor relies on significant amount of fuel shuffling this analysis is important for core design.

6.4.2.3.1 OXBOW.WI TerraPower-developed code OXBOW.WI will be used for core assembly withdrawal and insertion analysis. This code calculates the frictional interactions as well as the handling loads required to withdraw and insert assemblies of various deformations from an array of neighbors.

OXBOW.WI makes use of the finite element method utilizing the commercially available solver ABAQUS. The FEM is created using geometry, boundary conditions, and loads provided by the core design team. Deformed assembly state information can be generated using OXBOW.CRS and OXBOW.CADA.

6.4.2.4 Control Assembly SCRAM [128, 129]

Predicting the mechanical behavior of the control rods is an important part of core design. The amount of deformation that a Natrium control rod and control assembly accumulates over its lifetime may limit functionality. Understanding how much of their functionality is limited is a key effort to ensuring safe and reliable reactor operation. The scope of this analysis methodology is to assess the withdrawal and insertion mechanical response of control rods.

6.4.2.4.1 OXBOW.CASS The TerraPower-developed code OXBOW.CASS will be used for analysis of control assembly function under various operating conditions. This code utilizes deformed control assembly state information to evaluate the insertion and withdrawal capability of the control rod bundle.

OXBOW.CASS makes use of the finite element method utilizing the commercially available solver ABAQUS. The FEM is created using geometry, boundary conditions, and loads provided by the core design team. Control bundle insertability and SCRAM time analyses are conducted using OXBOW.CASS.

6.4.2.5 Core Assembly Pin Bundle/Duct Interaction [128, 129]

Over the lifetime of a core assembly, the assembly duct and the internal pin bundle will deform.

The duct will bow and dilate while the pin bundle will swell. As these components deform at different rates, clearance or interference may develop at their interfaces. Excessive clearance may result in flow-induced vibration, fretting, and excessive coolant bypass as flow area is increased. Conversely, significant interference may result in mechanical interaction between a pin bundle and duct or a decrease in the flow area which increases the assembly pressure drop.

Additionally, coolant flow restriction could cause overheating of fuel pins. The clearance or interference change resulting from differential duct and pin bundle deformations will be analyzed.

A similar type of analysis is needed for control assemblies. Absorber pins within a control rod bundle swell at a different rate than the control rod duct. Additionally, the control rod duct may experience different deformation modes than the control assembly duct. The unique clearance or

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 interference of control assembly components will be calculated for comparison against design limits.

6.4.2.5.1 OXBOW.BDI The TerraPower-developed code OXBOW.BDI will be used for analysis of bundle-duct interaction. This code calculates duct dilation, duct axial growth, core assembly bowing, and pin bundle deformation for a given load history and tracks pin positions resulting from external loading conditions on the core assembly. OXBOW.BDI makes use of the finite element method utilizing the commercially available solver ABAQUS. Custom creep and swelling user materials from the fuel performance code ALCHEMY are used in these ABAQUS models. These material subroutines incorporate thermal creep, irradiation creep, and void swelling correlations for stainless steels used in core design. The FEM is created using geometry, boundary conditions, and loads provided by the core design team.

6.4.2.6 Fuel Assembly Drop Analysis Models and Methods Core assemblies may experience drop accidents during handling, induced by a failure or malfunction of handling machines or an interaction with neighboring core assemblies. The models and methods shall be able to evaluate the structural integrity of the core assembly and core support structures during the following drop scenarios in the reactor core. The core assemblies will have a structural evaluation plan and analysis methodologies, which will rely on finite element analyses. Effects of irradiation will be accounted for. The core support structure will need to be evaluated to ASME Section III, Division 5, which will require structural analysis.

6.4.2.6.1 Core Assembly Drop while attached to Handling Machine Grapple This drop could occur if a partially inserted core assembly were to contact adjacent core assemblies with sufficient force such that a push from the handling machine is required to overcome friction and continue insertion. If the contact force from the adjacent core assemblies were suddenly released once the above core load pad of the inserted assembly passes the top load pads of the adjacent assemblies, the inserted assembly could drop onto the grapple interface.

6.4.2.6.2 Core Assembly Drop Following Release from the Handling Machine Grapple For a drop of this scenario to occur, the core assembly would have to be stuck at some elevation above its seated position and held by adjacent assemblies while the grapple is released and withdrawn. Then it subsequently would be released to drop into its receptacle position in the core support structure.

6.4.2.6.3 Core Assembly Drop Following Lift-off due to Vertical Seismic Loads During an earthquake, core assemblies could be lifted off from their receptacle positions due to a high vertical load that exceeds the hold-down margin of the core assemblies. Depending on the dynamic response of the core assemblies and core support structures, there could be an out-of-phase displacement that causes a high impact load. Note that this impact load could be more severe than the above scenarios since the impact velocity could be higher than that of a free drop in sodium or push by the grapple.

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7. TESTING AND INSPECTION OF NEW FUEL RAC 4.2-7 identifies the expectation that Testing and inspection shall be performed for new fuel to ensure that the fuel is fabricated in accordance with the design basis and that it reaches the plant site and is loaded in the core without damage. The bulk of the required activities will be specified in the fuel system product specifications, but to help ensure all testing and inspections are adequately captured, Table 7-1 summarizes the specific needs identified in RAC 4.2-7 along with the anticipated approach to address the requirements.

Table 7-1. Summary of New Fuel Testing and Inspection Needs and Planned Approach Requirement Planned Approach to Address Cladding integrity Certificates of conformance and supporting quality documentation demonstrating compliance with product specification requirements.

Use of qualified manufacturing processes and suppliers.

Fuel system dimensions Fuel enrichment and chemical composition Absorber composition Onsite inspection of new fuel and control assemblies to ensure delivered quality Program for receipt inspection and acceptance of new fuel assemblies after delivery to the plant

8. ONLINE FUEL SYSTEM MONITORING FOR FUEL PIN FAILURE Design and development of the online fuel monitoring system is addressed as part of the overall plant design effort and will be covered in more detail in future submittals. Only a brief summary is provided here for clarity on key points of fuel failure detection and identification. The primary method of determining if a fuel pin breach has occurred will be accomplished by continuously monitoring the cover gas effluent for the presence of radioactive fission gases as proven in FFTF. Upon gas plenum breach the gas is immediately released with the rate being ((

))(a)(4)

((

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((

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9. FUEL SURVEILLANCE Even though there is high confidence in Type 1 fuel being able to readily achieve its full design lifetime a comprehensive fuel surveillance program is planned to closely monitor fuel performance. The plan is to start operation with ((

))(a)(4) LDAs in the core, with a number of LDAs removed after each cycle of operation to perform post-irradiation exams to verify the fuel is performing consistent with expectations. See Table 9-1 for the notional planned irradiation of these LDAs during early cycles of operation to support fuel surveillance.

((

))(a)(4)

This report contains an assessment of data and testing required to support fuel pin licensing with particular focus on experiments supporting the fuel design limits/criteria, as well as high-importance phenomena that influence the ability to reliably meet design criteria. These high-importance phenomena will be monitored as part of the fuel surveillance program to verify consistent performance.

Specifically, 1) visual exams will be performed to identify any potential signs of wear or corrosion, 2) cladding and duct dimensions will be monitored to verify the integral response of the fuel pins/assemblies and determine overall cladding strain, 3) neutron radiography to verify the amount of fuel axial growth, 4) fission gas release measurements, and 5) FCCI measurements.

The use of LDAs in the Natrium surveillance program will provide early indications of any potential off-normal behavior and will supplement available in-reactor data to further reduce uncertainties in the fuel performance models. Specific fuel performance uncertainties to be addressed by the Fuel Surveillance Program are summarized in Table 9-2.

Table 9-1. Notional Fuel Surveillance Plan for Initial Cycles of Operation

((

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Cycles in the Reactor Total Number of Cycles in the Core Targeted Cladding Temperature Objective

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(( ))(a)(4)

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((

))(a)(4)

Cycles in the Reactor Total Number of Cycles in the Core Targeted Cladding Temperature Objective

((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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((

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Table 9-2. Fuel Performance Uncertainties and Mitigation Steps Fuel Performance Uncertainty Mitigation Step

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The final surveillance program will be dependent on the success of the other items previously identified in this report. Both non-destructive exams (NDE) and destructive exams (DE) to measure specific phenomenon are identified above to provide continued assurance of consistent fuel behavior. The target will be to have these exams performed in parallel with subsequent reactor cycles to prevent

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SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 disruption of operation. ((

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10. CONCLUSIONS A systematic assessment was performed to identify the activities required to support fuel system qualification. In general, the ongoing activities appear adequate to address most of the qualification needs. A few key exceptions are efforts to address fretting and fatigue behavior, as well as additional testing and analysis to address extreme transients, including coolability concerns. Initiation of many of these activities were delayed because they require prototypic fuel bundle geometries and anticipated operating and design basis event conditions. Now that conceptual design information is available, detailed test planning and supporting fabrication activities are underway. The current qualification plan relies heavily on historic operating experience/data from EBR-II and FFTF metallic fuel pins, with an ongoing effort to qualify the existing data to demonstrate its suitability [66]. The planned data qualification approach is consistent with the Quality Assurance Program Plan [77] submitted by Argonne National Laboratory for review and approved by the NRC [78]. The high-importance fuel phenomena identified for applicable fuel pin design limits include fission gas release, HT9 mechanical behavior as a function of environmental conditions, FCCI, and fuel thermal conductivity as a function of irradiation/porosity. A comprehensive set of test and analysis activities to address the limitations in these phenomena, and strengthen the basis of the associated design criteria, is summarized in Table 6-11 through Table 6-17. A series of mechanical tests have been identified for the fuel and control assemblies to address uncertainties in their response and to support model validation. These tests include component tests, single assembly tests, multiple assembly tests, and major effects tests. A summary of these tests, including associated effects, is provided in Table 6-19. Evaluation models for analytic predictions are available and capable of addressing most of the high-importance phenomena, with ongoing development to address existing gaps in time to support submission of the Final Safety Analysis Report, including verification and validation of the methods.

With no available fast-spectrum reactor to perform final tests using prototypic LTAs, a notional Surveillance Program is proposed to help monitor the irradiation performance of the fuel to help ensure consistent performance with historic operating experience and analytical predictions. The notional Surveillance Program will be revised to incorporate knowledge gained from additional analyses and testing data that becomes available.

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))(a)(4)

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))(a)(4)

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[48]

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[49]

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