Text

Submittal of Approved Terrapower, LLC Design Basis Accident Methodology for Events with Radiological Release Topical Report
	ML25251A090
Person / Time
Site:	Kemmerer File:TerraPower icon.png
Issue date:	09/05/2025
From:	George Wilson; TerraPower
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML25251A089	List: ML25251A090;
References
	TP-LIC-LET-0455 NAT-9394-A, Rev 0
	Download: ML25251A090 (1)
	v • d • e

15800 Northrup Way, Bellevue, WA 98008 www.TerraPower.com P. +1 (425) 324-2888 F. +1 (425) 324-2889 September 5, 2025 TP-LIC-LET-0455 Docket Number 50-613 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk

Subject:

Submittal of Approved TerraPower, LLC Design Basis Accident Methodology for Events with Radiological Release Topical Report

References:

1.

U.S. Nuclear Regulatory Commission, TerraPower, LLC - Final Safety Evaluation of NAT-9394, Design Basis Accident Methodology for Events with Radiological Release, Revision 0 (ML25189A085)

The U.S. Nuclear Regulatory Commission (NRC) provided the final safety evaluation for the TerraPower, LLC (TerraPower) Design Basis Accident Methodology for Events with Radiological Release Topical Report in Reference 1. The topical report provides an overview and description of the models developed to evaluate design basis accidents with the potential for radiological release for the Natrium1 Plant.

Enclosures 2 and 3 of this letter provide the accepted version of the topical report with additional content incorporated per NRC staff request, designated NAT-9394-A.

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 1 Natrium is a TerraPower and GE Vernova Hitachi Nuclear Energy Technology.

Date: September 5, 2025 Page 2 of 2 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 at igifford@terrapower.com.

Sincerely, George Wilson Senior 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 NAT-9394-A, Revision 0, Design Basis Accident Methodology for Events with Radiological Release-Non-Proprietary (Public)

3. TerraPower, LLC Topical Report NAT-9394-A, Revision 0, Design Basis Accident Methodology for Events with Radiological Release - Proprietary (Non-Public) cc:

Mallecia Sutton, NRC Josh Borromeo, NRC Nathan Howard, 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 Senior 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: September 5, 2025 George Wilson Senior Vice President, Regulatory Affairs TerraPower, LLC

ENCLOSURE 2 TerraPower, LLC Topical Report Design Basis Accident Methodology for Events with Radiological Release, NAT-9394A, Revision 0 Non-Proprietary (Public)

ATerraPower&GE Vernova HitachiNuclear Energy Technology Design Basis Accident Methodology for Events with Radiological Release NAT-9394-A Revision0 September 5, 2025

August 19, 2025 George Wilson Senior Vice President, Regulatory Affairs TerraPower, LLC 15800 Northup Way Bellevue, WA 98008

SUBJECT:

TERRAPOWER, LLC. - FINAL SAFETY EVALUATION OF TOPICAL REPORT NAT-9394, "DESIGN BASIS ACCIDENT METHODOLOGY FOR EVENTS WITH RADIOLOGICAL RELEASE," REVISION 0 (EPID NO. L-2024-TOP-0009)

Dear George Wilson:

By letter dated March 22, 2024, TerraPower submitted Topical Report (TR) TP-LIC-RPT-0007, Design Basis Accident Methodology for Events with Radiological Release, Revision 0 (ML24082A262), for the U.S. Nuclear Regulatory Commission (NRC) staffs review. On April 22, 2024, the NRC staff determined that the TR provided sufficient information for the NRC staff to begin its detailed technical review (ML24107B046). On July 15, 2024, the NRC staff transmitted an audit plan to TerraPower (ML24197A156) and subsequently conducted an audit of materials related to the TR from July 23, 2024, to January 29, 2025. The NRC staff issued the audit summary dated June 6, 2025 (ML25157A115). On February 28, 2025, TerraPower submitted a revision of the TR (ML25063A329), which was renumbered from TP-LIC-RPT-0007 to NAT-9394, Design Basis Accident Methodology for Events with Radiological Release, Revision 0, to clarify portions of the TR as discussed in the audit summary.

The enclosed final safety evaluation (SE) is being provided to TerraPower, because the NRC staff has found NAT-9394, Revision 0, acceptable for referencing in licensing actions to the extent specified and under the limitations and conditions delineated in the TR. 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 symbol.

If you have any questions, please contact Stephanie Devlin-Gill at (301) 415-5301 or via email at Stephanie.Devlin-Gill@nrc.gov.

Sincerely,

/RA/

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

Package: ML25189A082 Letter: ML25189A085 Enclosure (Public): ML25212A054 Enclosure (Non-Public): ML25189A086 OFFICE NRR/DANU/UTB2:BC NRR/DANU/UAL1:PM NRR/DANU/UAL1:LA NAME CdeMessieres SDevlin-Gill DGreene DATE 06/25/2025 07/01/2025 07/02/2025 OFFICE NRR/DANU/UAL1:BC NAME JBorromeo DATE 07/31/2025

OFFICIAL USE ONLY - PROPRIETARY INFORMATION Enclosure OFFICIAL USE ONLY - PROPRIETARY INFORMATION TERRAPOWER, LLC. - FINAL SAFETY EVALUATION OF TOPICAL REPORT NAT-9394, "DESIGN BASIS ACCIDENT METHODOLOGY FOR EVENTS WITH RADIOLOGICAL RELEASE," REVISION 0 (EPID NO. L-2024-TOP-0009)

SPONSOR AND SUBMITTAL INFORMATION Sponsor:

TerraPower, LLC (TerraPower)

Sponsor Address:

15800 Northup Way, Bellevue, WA 98008 Project No.:

99902100 Submittal Date:

March 22, 2024, February 28, 2025 Submittal Agencywide Documents Access and Management System (ADAMS) Accession Nos.:

ML24082A262, ML25063A329 Brief Description of the Topical Report: By letter dated March 22, 2024, TerraPower submitted Topical Report (TR) TP-LIC-RPT-0007, Design Basis Accident Methodology for Events with Radiological Release, Revision 0 (ML24082A262), for the U.S. Nuclear Regulatory Commission (NRC) staffs review. On April 22, 2024, the NRC staff determined that the TR provided sufficient information for the NRC staff to begin its detailed technical review (ML24107B046). On July 15, 2024, the NRC staff transmitted an audit plan to TerraPower (ML24197A156) and subsequently conducted an audit of materials related to the TR from July 23, 2024, to January 29, 2025. The NRC staff issued the audit summary dated June 6, 2025 (ML25157A115). On February 28, 2025, TerraPower submitted a revision of the TR (ML25063A329), which was renumbered from TP-LIC-RPT-0007 to NAT-9394, Design Basis Accident Methodology for Events with Radiological Release, Revision 0, to clarify portions of the TR as discussed in the audit summary.

NAT-9394, Revision 0, describes the methodology used to evaluate design basis accidents (DBAs) with the potential for radiological release for the Natrium reactor. This methodology consists of five discrete evaluation models (EMs), covering in-vessel transients, partial flow blockages, fuel misloads, fuel handling accidents (FHAs), and liquid sodium and gas leaks.

REGULATORY EVALUATION Regulatory Basis The regulations that are applicable to the review of this TR are:

Title 10 of the Code of Federal Regulations (10 CFR) section 50.34(a)(4) and 10 CFR 50.34(b)(4), which requires certain information to be submitted by applicants for construction permits and operating licenses, respectively. These sections require, in

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION part, analysis and evaluation of the design and performance of structures, systems, and components (SSCs) of the facility with the objective of assessing the risk to public health and safety resulting from the operation of the facility and including the determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of the SSCs provided for the prevention of accidents and the mitigation of the consequences of accidents.

Regulation 10 CFR 50.43(e), which requires that reactor designs that differ significantly from light-water reactor designs licensed before 1997, or that use simplified, inherent, passive or other innovative means to accomplish their safety functions have an appropriate demonstration of their safety features. Sections 50.43(e)(1)(i) and (ii) require a demonstration of safety feature performance and interdependent effects through analysis, appropriate test programs, experience, or a combination thereof. Section 50.43(e)(1)(iii) requires that sufficient data exist regarding the safety features of the design to assess the analytical tools for safety analyses over a sufficient range of plant conditions, including certain accident sequences.

The NRC guidance documents that are applicable to the review of this TR are described below.

Regulatory Guide (RG) 1.203, Transient and Accident Analysis Methods (ML053500170),

provides the evaluation model development and assessment process (EMDAP) as an acceptable framework for developing and assessing EMs for reactor transient and accident analyses. RG 1.203 outlines the four elements of an EMDAP, which is broken into 20 component steps. While the subject TR does not specifically reference RG 1.203, the NRC staff referenced various sections of RG 1.203 for best practices for EM development.1 For background, the Kemmerer Power Station Unit 1 (KU1) construction permit (CP) application (ML24088A059)2 was submitted by TerraPower on behalf of US SFR Owner, LLC, for a Natrium reactor following the process outlined in Nuclear Energy Institute (NEI) 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development (ML19241A472), as endorsed by the NRC in 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 (ML20091L698). This guidance defines risk-informed, performance-based, and technology-inclusive processes for the selection of licensing basis events (LBEs);

safety classification of SSCs; and the determination of defense-in-depth adequacy for non-light-water reactors. NEI 18-04 provides a frequency-consequence target curve that is used to assess events, SSCs, and programmatic controls. LBEs are categorized by the frequency of occurrence, separated into anticipated operational occurrences, design basis events (DBEs),

and beyond design basis events. DBAs are derived from DBEs by prescriptively assuming that only safety related (SR) SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34, Contents of applications; technical information 1 TerraPower has developed methodologies for the Natrium design informed by RG 1.203, including methodologies for the analysis of DBAs without radiological release, partial flow blockage, and source term. The relationship between these methodologies and the DBAs with radiological release methodology is discussed in section 1.1, Relationship to Other TerraPower TRs, of this SE.

2 TerraPower, on behalf of US SFR Owner, LLC, a wholly owned subsidiary of TerraPower, submitted the CP application for KU1 on March 28, 2024 (ML24088A059). The NRC staffs review of that CP application is ongoing. The staff is not making any determinations on the acceptability of the Natrium reactor design in this SE. The description of the Natrium reactor in this SE is based on the description in the TR.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION dose limits, using conservative assumptions. The purpose of the subject TR is to provide a methodology for analyzing certain DBAs as defined in NEI 18-04.

TECHNICAL EVALUATION

1.0 INTRODUCTION

TerraPower requested that the NRC staff review the proposed methodology as an appropriate and adequate means for future applicants using the Natrium design (as described in the TR) to evaluate DBA events that potentially lead to radiological release. The DBAs considered in the TR can be broadly divided into in-vessel and ex-vessel scenarios. In-vessel scenarios include transients leading to fuel damage and subsequent release, flow blockage, FHAs, and loss of active cooling scenarios. Ex-vessel scenarios include FHAs, loss of active cooling scenarios, and radioactive sodium and gas leaks. The TR discusses assumptions, EM development, and EM assessment for five EMs to address these scenarios.

However, as noted in the executive summary of the TR, [c]ertain aspects of the EM adequacy demonstration remain in development, and as documented in the TR, there are portions of each EM which are incomplete. Thus, the NRC staff imposed limitations and conditions, provided at the end of the safety evaluation (SE), to address portions of the overall methodology which have not been completed.

1.1 Relationship to Other TerraPower TRs The DBA with radiological release methodology TR is related to several other TerraPower methodology TRs that collectively provide a strategy for evaluating the consequences of potential accidental radiological releases for the proposed Natrium reactor design. TR section 4.1, Background provides a discussion of these relationships and includes TR figure 4.1-1, EM Calculational Devices and Analysis Workflow, which illustrates the connections between EMs.

The DBA with radiological release methodology does not identify the LBEs, DBAs, or other quantified event scenarios that result in radiological release for a given reactor licensing application. Rather, the LBEs, DBAs, and other quantified events appropriate for the licensing application are identified using the licensing modernization project methodology described in NEI 18-04. The DBAs are then analyzed using one of several methodologies. These methodologies include the DBA with radiological release methodology, described in this TR, the DBA without radiological release methodology, described in NAT-9390, Design Basis Accident Methodology for In-Vessel Events without Radiological Release, Revision 2, (ML24295A202) and evaluated by the NRC staff in ML25106A038, and the partial flow blockage methodology, described in NAT-9395, Partial Flow Blockage Methodology, Revision 0 (ML25129A064),

which is undergoing review by the NRC staff.

The DBA with radiological release methodology is used to determine the extent of cladding or fuel failure, as well as quantifying liquid sodium or gas leaks, which are inputs into the source term methodology described in NAT-9392, Radiological Source Term Methodology Report, Revision 0 (ML24261B944), and evaluated by the NRC staff in ML25063A323. The output of the source term methodology is radiological releases to the atmosphere (source terms), which are input to the radiological consequence EMs described in TerraPower report NAT-9391, Radiological Release Consequences Methodology Topical Report, Revision 0,

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION (ML24208A181) and evaluated by the NRC staff in ML25106A262, which is used to determine dose consequences associated with releases.

This TR also references NAT-2806-A, Fuel and Control Assembly Qualification, Revision 0, (ML24354A192) and TP-LIC-RPT-0011, Core Nuclear and Thermal Hydraulic Design Report, Revision 0, (ML24088A085) regarding fuel failure phenomena and steady state core analysis, respectively. TP-LIC-RPT-0011 is under review by the NRC staff as part of the KU1 CP application.

2.0 BACKGROUND

TR section 2.2, Plant Description, provides an overview of the Natrium reactor design. The Natrium reactor is a pool-type sodium-cooled fast reactor (SFR) with metal fuel. In the primary heat transport system, liquid sodium is transferred from the cold pool using mechanical primary sodium pumps to the lower plenum and through the reactor core, where it is heated. The hot sodium then enters the hot pool and transfers its heat via intermediate heat exchangers (IHXs) to the intermediate heat transport system (IHT) sodium loops before returning to the cold pool.

Liquid sodium is circulated around the intermediate loops using mechanical intermediate sodium pumps (ISPs), which enables heat to be transferred from the core to a molten salt loop via a sodium-salt heat exchanger (SHX). This molten salt is pumped between the SHX and the energy island, where it can be stored and converted to electricity.

The Natrium plants safety related means of residual heat removal is the reactor air cooling system (RAC). The RAC cools the reactor by supplying natural draft outside ambient air down into the reactor cavity and past the outside of the reactor. The RAC is an open, passive system that is always in operation. The Natrium plant can also be cooled via the intermediate air cooling system (IAC). The IAC is non-safety related and serves as the normal shutdown cooling system.

Each intermediate loop contains a sodium-to-air heat exchanger (AHX). Active forced circulation through both the IHT (via ISPs) and IAC (via air blowers) supports normal controlled cooling operations. If power is not available to support forced flow, the natural draft of air through the IAC can provide passive cooling.

The Type 1 fuel proposed for the Natrium core consists of metallic uranium-zirconium alloy slugs contained in right cylindrical fuel pins, arranged in a triangular pitch to form hexagonal fuel assemblies. Additional details regarding Natrium Type 1 fuel and its qualification are provided in NAT-2806-A.

TR section 2.2 additionally describes fuel handling and storage for the Natrium reactor. The Natrium design contains an in-vessel transfer machine (IVTM) and fuel transfer lift, which are installed during refueling outages. The IVTM moves fuel assemblies between the core, in-vessel fuel storage racks, and transfer station. The transfer station allows for fuel removal from the reactor vessel (RV) through the fuel transfer lift. The Natrium design also contains an ex-vessel fuel handling system, which transfers fuel entering the facility through inspection and conditioning and finally to the RV. The ex-vessel system also transfers irradiated assemblies from the RV to the ex-vessel storage tank (EVST). Irradiated assemblies in the EVST are eventually transferred to the pool immersion cell (PIC), where sodium residue is removed to allow for storage in water, and finally into the spent fuel pool (SFP). As separate water pool fuel handling system moves cleaned assemblies from the PIC into the SFP and transfers assemblies from the SFP into a cask for dry storage. Once the cask is prepared, it is transported to the long-term dry storage location.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION The sodium processing system (SPS) is discussed in NAT-9392. The SPS controls and monitors primary and intermediate sodium chemistry, using cold traps and cesium traps to capture impurities and radionuclides from leaking or failed fuel. The sodium cover gas system (SCG) is not described in detail for the purposes of the TR and EM. However, the KU1 CP application states that the SCG controls, monitors, and supplies inert argon gas to various systems and components throughout the reactor building including the RV and gas spaces of the IHT.

3.0 ASSUMPTIONS TR chapter 3, Assumptions Requiring Verification, discusses the assumptions TerraPower used to define the scope of the TR EMs, determine conservative boundaries, or identify areas where future work is planned. The first two assumptions are applicable to all of the EMs discussed in the TR, while the other eleven are relevant to a specific EM. Assumption 3.1 states that event and accident scenarios considered in the TR will be limited to DBAs with potential release. Assumption 3.2 states that the TR is based on the current Natrium reactor design and will be updated as the design matures.

The NRC staff determined that these two assumptions are reasonable to apply to the five methodologies discussed in the TR, because they have all been developed specifically for analysis of DBAs for the Natrium design. Any applicant or licensee referencing this TR must justify that any departures from the Natrium design as described in the TR do not impact the conclusions of this TR or SE. This is captured in limitation and condition 1, below.

Assumptions 3.3 - 3.12 are discussed in detail for each EM in section 5.0, Event-Specific Methodologies, of the SE.

Assumption 3.13 states that the EM for in-vessel transients with radiological release assumes only Type 1 fuel will be used. The NRC staff determined that this assumption is reasonable and that it is necessary to limit the applicability of the in-vessel transients with release EM to Natrium Type 1 fuel or otherwise require an applicant or licensee referencing the TR to provide justification that using a different type of fuel does not affect the conclusions of the TR and this SE. The NRC staff additionally determined that this assumption should be applied to the partial flow blockage, FHAs, and fuel misload EMs, since these all also inherently assume Type 1 fuel is used. The assumption regarding the use of Type 1 fuel is captured as limitation and condition 1, below.

4.0 EM DEVELOPMENT AND ASSESSMENT 4.1 DBA Event Selection In TR section 2.4, DBA Event Selection, TerraPower discusses the identified DBAs for the Natrium design, broadly categorizing them into in-vessel core transients, local faults (e.g., partial flow blockages and fuel misloads), fuel handling events, and radioactive gas or liquid release events. TR table 2-1, Natrium DBAs with Radioactive Material Release, lists the ten DBAs identified thus far which involve a potential release of radioactive material. TerraPower notes that these DBAs are provided to illustrate the methodology in the report, rather than define the set of events applicable to all Natrium plants. TerraPower additionally states that three DBAs associated with excessive sodium-water reaction in the PIC, loss of EVST cooling while storing fuel assemblies, and leakage from the gaseous radwaste processing system (RWG),

respectively, are not covered by any EM contained within the TR. Accordingly, the NRC staff

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION determined that, as this TR does not contain event-specific methodologies for these DBAs, future licensing submittals referencing this TR and using one of the contained EMs for these events will require further justification to ensure the selected EM is suitable. This is captured in limitation and condition 2, below.

TR section 2.4 additionally states that DBAs which are not in-vessel are evaluated using the appropriate methodology in the DBA with release EM, an appropriate event-specific method, or evaluated with the source term EM using conservative assumptions. As such, the NRC staff determined that ex-vessel release analyses referencing this TR for their basis in future licensing submittals must provide sufficient detail to demonstrate that the appropriate methodology has been used. This is captured in limitation and condition 3, below.

4.1.1 PIRT Development TR section 4.3, Phenomena Identification and Ranking Tables (PIRT), discusses the PIRTs developed to support the EMs contained in the TR. The PIRT concept is discussed in Step 4 of RG 1.203, which states that, as part of establishing the requirements for EM capability, key phenomena and processes should be identified and ranked with respect to their influence on the figures of merit (FOMs). This is accomplished by developing a PIRT. A given scenario is divided-up into characteristic time periods where dominant phenomena and processes remain relatively constant. For each time period, phenomena and processes are identified for each component. The phenomena and processes that the EM should simulate are determined by examining experimental data, expert opinion, and code simulations related to the specific scenario. After identification, the phenomena and processes are ranked by importance determined with respect to their effect on the relevant FOMs.

Throughout the TR, TerraPower refers to PIRTs performed for the following:

Other qualified events (OQEs), which are developed based on accidents expected to have a frequency lower than beyond design basis events (BDBEs) and as such focused on unprotected in-vessel events Partial flow blockage within a subassembly Fuel and absorber pin behavior In-vessel DBAs without radiological release SPS leaks The PIRT for fuel and absorber pin behavior, in-vessel DBAs without radiological release, and SPS leaks are discussed further in NAT-2806, NAT-9390, and NAT-9392, respectively, and considered by the NRC in their associated SEs. The PIRT for partial flow blockages is discussed further in NAT-9395, which is under NRC staff review. As such, the NRC staff focused its review on the OQE PIRT for this TR.

TR section 5.1.3, EM Scope and Requirements, states that the in-vessel transients with radiological release EM is being developed to address the full scope of DBA, BDBE, and OQE events. This section further states that the identified phenomena from the OQE PIRT may be expected for more frequent BDBEs and are applicable and bounding for in-vessel DBAs. The OQE PIRT considered an unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), and unprotected transient over-power (UTOP). TerraPower states that a PIRT was developed for each event by a panel of internal and external experts and provides TR table 5-1, Combined PIRT for ULOF, ULOHS and UTOP LBEs with Radiological Release with

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION High/Medium Importance Phenomena, which contains the combined results from the three event-specific PIRTs.

The NRC staff audited TerraPowers documentation detailing the OQE PIRT development process, and determined it aligns with the best practices discussed in RG 1.203, as is described in the TR. The NRC staff additionally determined that the PIRT phenomena are appropriate for the scenarios considered in the EM because they are consistent with the Natrium design and past SFR operating experience, and in the NRC staffs engineering judgment would be expected to be bounding for in-vessel DBAs.

4.1.2 EM Assessment Matrix Development TR section 4.4, Evaluation Model Assessment discusses TerraPowers assessment plan for the EMs discussed in the TR. TerraPower states that for each EM with a PIRT, an assessment matrix will be created. TerraPower states that, based on the assessment matrix, testing needs will be identified. This is broadly consistent with the second principle of EMDAP, as discussed in RG 1.203, which is to develop an assessment base consistent with determined requirements.

As RG 1.203 discusses, this assessment base is used to validate calculational devices or codes used by the EM, and may consist of legacy experiments or may require new experiments to be performed.

4.1.2.1 In-vessel Transients with Radiological Release Methodology TR table 4-3, Assessment Matrix for High/Medium Importance Fuel Failure Phenomena in OQEs, and table 4-4, Assessment Matrix for High-Importance Fuel Failure Phenomena, list what test facilities may contribute test data for medium and high-ranked phenomena identified in their respective PIRTs. These assessment matrices include both legacy experiments as well as planned testing. TerraPower notes that these matrices focus on fuel failure phenomena, as the in-vessel transients with radiological release EM builds on the in-vessel transients without radiological release EM, which has its own assessment matrix documented in NAT-9390 and evaluated by the NRC staff in its associated SE. The NRC staff determined that TerraPowers approach to EM assessment development for fuel failure phenomena are acceptable because they align with the best practices discussed in RG 1.203, ensuring legacy experiments and planned testing to address medium and highly ranked phenomena are identified. However, the NRC staff has not determined the acceptability of the assessment matrices contained in tables 4-3 and 4-4, as they have not been completed. As discussed in limitation and condition 4, future licensing submittals referencing this TR will need to justify that code qualification, verification, and validation activities have been completed to a state that is appropriate for the intended licensing application.

4.1.2.2 Partial Flow Blockage Methodology TR section 4.4 states that an assessment matrix for the partial flow blockage methodology is included in NAT-9395, which is under a separate NRC review. As the partial flow blockage methodology is undergoing a separate review, the NRC staff made no determination regarding the methodology in this SE.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 4.1.2.3 Fuel Misload Methodology TR section 4.4 states that fuel misloads do not include any phenomena beyond those normally modeled in steady state analysis, and thus TerraPower states that the fuel misload EM leverages the code qualification, verification, and validation activities discussed in NAT-2806-A and TP-LIC-RPT-0011. TerraPower states that these activities utilized for steady state core design, thermal hydraulics, and fuel performance serve the function of an assessment matrix for the fuel misload EM. The staff reviewed this and concluded it was reasonable, because past SFR experience indicates that misloads of the types considered by TerraPower in the fuel misload EM would not be expected to result in conditions that differ significantly from normal, steady state conditions. As such, the NRC staff determined that TerraPowers use of code qualification, verification, and validation performed for steady state calculations to support the assessment of the fuel misloads EM is acceptable because, in the staffs engineering judgment, the code qualification, verification, and validation activities expected to be needed for the steady state core design would also be applicable for fuel misload events.

4.1.2.4 FHA and Sodium Liquid and Gas Leak Methodology TerraPower states that the PIRTs developed for the FHA EM and liquid sodium and gas leak EM in NAT-9392 focus on release and transport of radionuclides, rather than the dynamics and structural analysis or the calculation of leak rate and timing for the two EMs, respectively. The radiological source term methodology assessment base is described in NAT-9392. As such, assessment matrices have not been developed for these EMs in the DBAs with release TR.

TerraPower states that assessment matrices for these EMs may be developed in the future.

Therefore, the NRC staff made no determination on the EM assessment plans for the FHA and sodium leak and gas release EMs. As discussed in limitation and condition 4, future licensing submittals referencing this TR will need to justify that code qualification, verification, and validation activities have been completed to a state that is appropriate for the intended licensing application.

4.1.3 Quality Assurance TR section 4.2 states that, for DBAs with potential radiological release, EM assessment is guided by TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk. TerraPower states that this plan provides a framework supporting quality assurance (QA) for software that perform safety related or non-safety related analyses. The plan discusses gap analysis and maturation activities while also including sections on commercial grade dedication plans for commercially acquired software used in safety related analyses. The NRC staff audited this document and determined it to be consistent with the discussion in this section of the TR. Additionally, in TR section 4.4, TerraPower states that EM assessment will follow TerraPowers QA program description discussed in TP-QA-PD-0001, TerraPower QA Program Description, Revision 14-A (ML23213A199), which has been reviewed and approved by the NRC staff. The NRC staff determined that TerraPowers planned QA activities for the DBAs with radiological release methodology are appropriate and follow the program description provided in TP-QA-PD-0001.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.0 EVENT-SPECIFIC METHODOLOGIES TR chapter 5, Event-Specific Methodology, outlines five EMs for different categories of DBAs with potential for radiological release. These EMs build on the EM development and assessment discussed in the previous chapters of the TR.

5.1 In-Vessel Transients with Radiological Release Methodology TR section 5.1, In-Vessel Transients with Radiological Release Methodology, outlines a methodology used to analyze DBAs that lead to cladding and fuel failures, resulting in the release of radionuclides to the primary coolant. This methodology is an extension of the methodology for in-vessel DBAs without radiological release, discussed in NAT-9390 and its associated SE.

5.1.1 Assumptions TR section 5.1.2, Assumptions, outlines the EM-specific assumption, reproducing TR section 3.1 item 3.13. Assumption 3.13 states that the EM for in-vessel transients with radiological release assumes only Type 1 fuel will be used. As discussed in SE section 3.0, Assumptions, the NRC staff determined that assumption 3.13 is reasonable; this assumption is captured in limitation and condition 1, below.

5.1.2 Connection with the In-Vessel Transients without Radiological Release Methodology TR table 4-1, Figures of Merit for In-Vessel DBAs, discusses the three FOMs used in TerraPowers in-vessel transients without release methodology as described in NAT-9390 and its associated SE. These FOMs consist of fuel centerline temperature, coolant temperature, and acceptance criteria for peak cladding temperature (PCT) based on a time-at-temperature approach. The acceptance criteria for time-at-temperature no-failure (TATNF) for PCT accounts for strain, cladding wastage, and thermal creep. For in-vessel events without radiological release, TerraPower selected SAS4A/SASSYS-1 (SAS)3 for its system analysis code. Final results from transient calculations in SAS are compared against these FOMs to determine if any limiting values are violated.

TR section 5.1.1, Purpose and Scope, discusses the relationship between TerraPowers EM for in-vessel transients with radiological release and TerraPowers EM for in-vessel transients without radiological release. This section further states that, if TATNF is not violated for a transient, no radiological release occurs and thus the EM for in-vessel transients without release is sufficient. TerraPower states that TATNF contains various conservatisms such that, if violated, fuel does not necessarily fail. Upon a TATNF violation, ((

)) or the transient can be further analyzed through the Detailed Safety Analysis Workflow (DSAW).

3 SAS is a physics simulation software developed by Argonne National Laboratory (ANL) to perform deterministic analysis of anticipated events and DBAs for SFRs. SAS is one-dimensional and composed of two computer codes, SAS4A and SASSYS-1. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. SASSYS-1 provides the capability to perform a detailed thermal-hydraulic simulation of the primary and intermediate sodium coolant circuits and the balance-of-plant steam-water circuit.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.1.3 DSAW Description DSAW is described in TR section 5.1.5, In-Vessel Transient Evaluation Workflow, and illustrated in TR figure 5.1-1, DSAW Data Flow. DSAW ((

)). If the DSAW analysis of a given transient shows that these acceptance criteria are met, no fuel failure occurs. TR section 5.1.1 states that the DSAW does not allow for the ((

)). As such, TR section 5.1.4, EM Description, states that, if the DSAW results indicate assembly-wide fuel failures are expected, the ((

)).

TR section 9.1, Appendix A - Additional Details of the DSAW Process, provides additional information on the DSAW process. ((

)).

((

)).

The NRC staff audited the user and theory manuals for DSAW and its constituent codes, including ((

)).

The NRC staff determined that the DSAW process described in the TR is acceptable for evaluating fuel and cladding failure for the Natrium reactor design because DSAW has appropriate acceptance criteria for determining whether fuel or cladding would fail as the result of transient conditions and because the constituent codes have the capability of calculating these criteria. The NRC staff also determined that ((

)) for conservatism is appropriate for the DSAW process. However, the NRC staff has not reviewed ((

))

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION for use in the DSAW process, though ((

)).

The NRC staff also notes that the ultimate means of determining whether the DSAW process is adequately conservative is to compare the prediction of DSAW with applicable experimental data. Therefore, the NRC staff did not make a determination with respect to the conservatisms for DSAW described in the TR. This limitation is captured in limitation and condition 5.

5.1.4

((

))

If DSAW indicates fuel or cladding failure is expected, ((

)) using severe accident modules. TR section 5.1.4 discusses the modules TerraPower plans to use for addressing fuel or cladding failures that occur during in-vessel transients which consist of ((

)).

((

)). The NRC staff reviewed the ((

)) code manual5 regarding these modules and determined that their selection appears appropriate for TerraPowers intended use, with the exception of ((

)), which the staff did not make a determination on due to the module not being discussed in the code manual.

TR section 5.1.3 outlines three event phases associated with severe accidents, consisting of initiating, transition, and termination. The section further states that the range of phenomena associated with in-vessel transient DBAs would not transition to a severe accident as described in the initiating phase ((

)). TerraPower states that this EM is being developed to address the full scope of DBAs, BDBEs, and OQEs.

As discussed in TR section 5.1.3, DBAs for the Natrium reactor will not transition to a severe accident. The NRC staff reviewed this and determined that the applicability of this EM for licensing analyses is restricted to those events that do not experience severe accident phenomena ((

)). This limitation is captured in limitation and condition 6. However, the NRC staff notes that it may be appropriate to use this EM, including ((

)), for sensitivity studies or for analyses of transients which are more severe than DBAs, as justified.

4 ((

)).

5 ((

)).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.1.5 EM Assessment TR section 5.1.6 presents an adequacy assessment of the in-vessel DBA with release EM, based primarily on assessment of ((

)) and implementation of TerraPowers acquired software QA plan. The software QA plan is discussed in section 4.1.3 of this SE. Relevant discussion on assessment of ((

)), is provided in ((

)), where TerraPowers approach to assessing ((

)) was determined to be acceptable, though additional work remains to be done. Beyond those aspects of ((

)) included in ((

)), NAT-9394 includes additional models for ((

)) as discussed in section 5.1.4 of this SE. As discussed in SE section 5.1.4 and limitation and condition 6, the in-vessel transients with radiological release methodology EM does not include accidents with severe accident phenomena and thus does not take advantage of the ((

)). Because of this, the evaluation provided in the NRC staffs SE for NAT-9390 is applicable to NAT-9394. As such the NRC staff finds the approach to assessing the adequacy of ((

)) acceptable but notes that additional work is planned to complete the assessment; this aspect of the EM is thus subject to limitation and condition 4.

5.2 Partial Flow Blockage Methodology TR section 5.2, Partial Flow Blockage Methodology, provides a short overview of TerraPowers partial flow blockage methodology for the Natrium design. TerraPower references NAT-9395, that discusses the partial flow blockage methodology in detail and which is under a separate NRC review. The section further states that this summary was included to provide context on how the partial flow blockage methodology fits within the scope of the DBAs with radiological release methodology. TerraPower states that the partial flow blockage methodology is used to determine whether fuel or cladding fails. TR figure 4.1-1 shows that ((

)). As the partial flow blockage methodology is undergoing a separate review, the NRC staff made no determination regarding the methodology in this SE.

As described in SE section 3.0, the NRC staff also determined assumption 3.13 is applicable to the partial flow blockage EM; this assumption is captured in limitation and condition 1, below.

5.3 Fuel Misload Methodology TR section 5.3, Fuel Misload Methodology, discusses TerraPowers methodology for analyzing the consequences of having a fuel assembly in the wrong core location or in the wrong orientation. TR section 5.3.1, Purpose and Scope, states that the Natrium core has two main enrichment zones, with the outer zone having higher enrichment to flatten power. ((

)).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.3.1 Assumptions - Fuel Misload Methodology TR section 5.3.2, Assumptions, outlines four assumptions applicable to the fuel misload methodology; these assumptions reproduce TR section 3.1 items 3.9 through 3.12.

Assumption 3.9 states that the steady state tools used to design the reactor core can model the misloaded core ((

)). The NRC staff determined that this assumption is reasonable, noting that this expected for the fuel type and misloads considered in the EM.

Assumption 3.10 states that the final Natrium design ((

)). The NRC staff determined this assumption is reasonable and that it is necessary to limit the applicability of the fuel misload EM to a design ((

)). Limitation and condition 1a addresses this assumption.

Assumption 3.11 ((

)). The NRC determined that this assumption is reasonable because ((

)). The NRC staff also determined that ((

)).

Assumption 3.12 states that the misloaded assembly is ((

)). The NRC staff determined that this assumption is reasonable in establishing ((

)).

As described in SE section 3.0, the NRC staff also determined assumption 3.13 is applicable to the fuel misload EM; this assumption is captured in limitation and condition 1, below.

5.3.2 Fuel Misload Phenomena TR section 5.3.3, EM Scope and Requirements, discusses the one highly ranked phenomenon identified for the Natrium fuel misload methodology, which is the change in the local power distribution. The TR states that the magnitude of this change for a given misload event depends on the change of isotopic distribution of impacted pins between the intended and misloaded core configuration. In this section, ((

)).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.3.3 Fuel Misload EM TR section 5.3.4, EM Description, discusses the EM for analyzing fuel misloads. The TR states that fuel misloads are analyzed using ((

)).

The TR states that the methodology for determining the limiting assembly and fuel pin for fuel misloads has not been finalized, with further work planned.

TR section 5.3.4 states that ten cases covering misloaded fresh, once-burnt, twice burnt, and thrice burnt fuel assemblies at beginning-of-life and beginning of equilibrium core conditions were analyzed in support of the KU1 CP application. ((

)). The NRC staff audited the referenced fuel misload evaluations and found them to be consistent with the discussion in this section of the TR. The NRC staff did not make a determination regarding the acceptability of the CP application misload analyses.

The NRC staff determined that the fuel misload methodology is acceptable because ((

)). The NRC staff made no determination on the acceptability of the referenced steady state core design methodology, which is undergoing a separate review.

5.4 FHA Methodology TR section 5.4, [FHA] Methodology, discusses TerraPowers methodology for analyzing structural-mechanical behavior that could lead to failure of dropped or impacted fuel assemblies.

Transport and consequence of radiological release from FHAs is discussed in NAT-9392.

TR section 5.4.1, Purpose and Scope, outlines five scenarios which could lead to fuel damage and radiological release:

1. Insertion or removal of a fuel assembly from reactor core

2. In-vessel fuel assembly movement

3. Ex-vessel fuel assembly movement between the EVST and washing station

4. Inadvertent action causing spent fuel assembly crush

5. Fuel assembly or loaded fuel cask drop in SFP The TR states that these scenarios are analyzed to determine possible damage and final configuration for both the dropped and impacted fuel assemblies. The NRC staff determined that the events selected for consideration for FHAs are reasonable because they are consistent with the Natrium design discussed in TR section 2.2 and with the FHA events discussed in the NAT-9392.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 5.4.1 Assumptions - FHAs TR section 5.4.2, Assumptions, outlines three assumptions applicable to the FHA methodology; these assumptions reproduce TR section 3.1 items 3.6 through 3.8. Assumption 3.6 states that, ((

)). Assumption 3.7 states that, for scoping potential radiological release during a FHA, a fuel assembly with ((

)). Assumption 3.8 states that, when performing a detailed analysis of a FHA, limiting conditions which result in the worst possible fuel damage and highest radiological release are considered. The NRC staff determined that these three assumptions are reasonable because: 1) assumption 3.6 is suitably conservative since it ((

));

and 2) assumptions 3.7 and 3.8 are reasonable in establishing a conservative worst-case event for a given FHA.

As described in SE section 3.0, the NRC staff also determined assumption 3.13 is applicable to the FHA EM; this assumption is captured in limitation and condition 1, below.

5.4.2 FHA Acceptance Criteria and Phenomena TR section 5.4.3, Acceptance criteria, discusses acceptable performance for a dropped fuel assembly and any affected structures or targets during an FHA. Dropped fuel assemblies must not result in fuel cladding mechanical failure in either the dropped assembly or any impacted structures. Additionally, the dropped assemblies cannot create unacceptable core conditions that impact safe reactor operations, such as local criticality or reduced flow. The TR states that the TR FHA methodology only covers analysis of mechanical damage to affected assemblies and does not address the consequence of potential radiological release.

TR section 5.4.4, EM Scope and Requirements, outlines potential phenomena that can cause mechanical damage to fuel assemblies during an FHA. The TR notes that ((

)). TerraPower states that these phenomena include stress, strain, and loading limits of fuel assembly components, structural component fatigue, elastic and inelastic behavior of components under loading, and mechanical fracturing caused by dynamic loads or impacts. TerraPower states that the FHA EM should be able to model these phenomena.

5.4.3 FHA EM TR section 5.4.5, EM Description, describes the software for analyzing FHAs. A finite element analysis (FEA) software provides predictions on fuel assembly stress and impact forces for different fuel drop scenarios. These values are then compared to fuel assembly strength limits to determine the extent of fuel damage. The extent of fuel damage is then used to quantify the radiological release for the FHA. FEA software employs a finite element method, in which behavior of a system is solved by subdividing it into smaller parts, called finite elements. This is done via discretization in the space dimensions and implemented by the construction of a mesh of the object or system. The TR states that a finite element model of a Natrium fuel assembly was built for preliminary analysis of FHAs. Audit of NAT-5630, Rev. 0, Finite Element Modeling

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION and Analysis Methods for Core Assembly Drop Accidents, enabled the NRC staff to better understand the model used for Natrium FHA analyses.

TR table 4-2 states that ((

)) are suitable for analyzing mechanical behavior during FHAs. During the audit, the NRC staff reviewed the technical manuals for the referenced FEA software, which appeared to be suitable for the EM. TR section 5.4.6, EM Assessment, notes that TerraPowers assessment of the FEA software selected for Natriums FHA EM is ongoing, with further work planned. Section 5.4.6 also discusses uncertainties that may arise when performing FEA for FHAs. These uncertainties consist of the geometric complexity of the fuel assembly and the difficulty in defining ((

)). TerraPower additionally states ((

)).

The NRC staff determined that the detailed mechanical approach for FHAs described in section 5.4 of the SE is not sufficiently developed for use in licensing analyses as ((

)) and work required to assess FEA software for FHAs have not been completed. This limitation is captured in limitation and condition 7, below.

However, the use of conservative bounding assumptions, ((

)), are acceptable for licensing applications, including support of a CP application, as they adequately bound worst possible radiological release for a given FHA.

5.5 Sodium Liquid and Gas Leak Methodology TR section 5.5, Sodium Liquid and Gas Leak Methodology, summarizes a methodology to determine and quantify leaks for dose calculations. The TR states that this analysis includes determining the extent of the leaks based on event initiation, including location, timing, system conditions, and propagation. The EM considers leaks from the SCG, SPS, and IHT. Leaks from the SCG would include ((

)) which have leaked from the fuel. SPS leaks would contain ((

)), while leaks from the IHT would include ((

)).

The treatment of sodium liquid and gas leaks for the Natrium design is discussed in detail in NAT-9392. Accordingly, the NRC staff focused on aspects unique to this TR such as plans for determining suitable system leakage rates. In TR section 5.5.1, Purpose and Scope, TerraPower states that a detailed methodology for mechanistically determining specific leak conditions has not yet been developed.

5.5.1 Assumptions - Sodium Liquid and Gas Leak Methodology TR section 5.5.2, Assumptions, outlines two assumptions applicable to the leak methodology; these assumptions reproduce TR section 3.1 items 3.3 and 3.4. Assumption 3.3 states that system leakage scenarios are assumed to occur during normal operation and not as a result of a different event. Assumption 3.4 presumes that ((

)). The NRC staff determined that these assumptions are reasonable, noting that, ((

)), any applicant or licensee referencing this TR must

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION provide appropriate justification that the ((

)) is suitably conservative. This limitation is captured by limitation and condition 5, below.

5.5.2 Sodium Liquid and Gas Leak EM Description TR section 5.5.3, EM Scope and Requirements, states that the EM established for analyzing sodium and gas leaks should have the capability of modeling important processes and phenomena identified during a representative PIRT. Phenomena associated with SPS leaks are detailed in table 2-5 of NAT-9392, which was evaluated by NRC staff in its associated SE, in which the NRC staff determined that the identified phenomena from the PIRT are reasonable for source term analysis for sodium leaks from the Natrium design.

TR figure 5.5-1, Sodium Cleanup System and IHT Leak EM Diagram, provides an overview of the different components of the sodium liquid and gas leak EM. TerraPower states that the

(( output from GOTHIC, MELCOR, or a manual calculation which feeds )) into the system leak rate is the portion of the EM relevant to the overall DBA with radiological release methodology.

TR section 5.5.4, EM Description, and section 5.5.5, EM Assessment, provide a brief overview and assessment of the EM, referencing NAT-9392 for further information.

As discussed, the NRC staff focused its review of this methodology on plans for determining suitable system leakage rates. The NRC staff determined that the use of conservative bounding assumptions to determine a maximized release is reasonable for licensing applications. The NRC staff also determined that plans to potentially use ((

)) to determine leakage rate is reasonable; however, the NRC staff made no determinations on the final implementation of this EM because, the TR states that the use of a detailed methodology for mechanistically determining specific leak conditions, such as sodium and gas leak rates and timing, has not yet been developed. As such, any applicants or licensees referencing the sodium liquid and gas leak methodology described in section 5.5 of this TR must appropriately justify that its selected sodium and gas leakage rates and timing are suitably conservative. This limitation is captured by limitation and condition 5, below.

5.6 EM Conservatisms TR section 6.2, EM Conservatism Summary, provides an overview of the conservative approaches TerraPower is taking for the different EMs included in this TR. For in-vessel transients with radiological release and partial flow blockages, the TR states that it will use conservatisms similar to those discussed in NAT-9390, including selecting conservative plant initial and boundary conditions within the operating band, assessing hot pin PCT within the sub-assemblies ((

)), and relying on conservatisms in the TATNF screening criteria and the DSAW process. TerraPower states that fuel misload analyses are adjusted for uncertainties in final temperature distribution ((

)) to determine potential for fuel failures. For FHAs, ((

)). TR section 6.2 states that an effort is underway to determine that these approaches are sufficiently conservative for the Natrium design.

The NRC staff determined that TerraPowers approach to ensuring conservatism for the EMs discussed in this TR is reasonable. The NRC staff notes that the ultimate means of determining whether the EM is sufficiently conservative is to compare the prediction of the EM with

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION applicable experimental data. Therefore, the NRC staff has not made a determination with respect to the final appropriateness of TerraPowers EM conservatisms. As discussed in limitation and condition 5, future licensing submittals referencing this TR must appropriately justify that the initial and boundary conditions and other input modeling parameter values are conservative.

LIMITATIONS AND CONDITIONS The NRC staff imposes the following limitations and conditions on the use of this TR:

1. The NRC staffs determinations in this SE are limited to the Natrium design described in section 2.2 of the TR and this SE. An applicant or licensee referencing the EMs developed in this TR must justify that any departures from these design features do not affect the conclusions of the TR and this SE. Additionally, this methodology was developed to analyze certain DBAs as discussed in TR section 1.0 and this SE (and as defined in NEI 18-04); use of this methodology for other kinds of analyses must be justified.

a. For the FHA EM, the NRC staffs determinations are limited to the Natrium design ((

)).

b. For the in-vessel transients with radiological release, partial flow blockage, FHA, and fuel misload EMs, the NRC staffs determinations in this SE are limited to the Natrium design using Natrium Type 1 fuel.

2. As discussed in section 2.4 of the TR, the DBAs with radiological release methodology does not contain event-specific EMs for events associated with excessive sodium-water reaction in the PIC, loss of EVST cooling while storing fuel assemblies, and leakage from the gaseous radwaste processing system (RWG). Use of this methodology for these events requires further justification.

3. Section 2.4 of the TR states that DBAs which are not in-vessel are evaluated using the appropriate methodology in the DBA with release EM, an appropriate event-specific method, or evaluated with the source term EM using conservative assumptions. As such, applications involving ex-vessel release analyses referencing this TR for their basis must provide sufficient detail to demonstrate that the methodology used is suitable.

4. An applicant or licensee referencing this methodology must submit documentation and justify that code qualification, verification, and validation activities have been completed to a state that is appropriate for the intended licensing application for each of the EMs discussed in the TR.

5. Consistent with section 6.2 of the TR, applicants or licensees referencing this methodology must appropriately justify that the initial and boundary conditions and other input modeling parameter values are conservatively selected. This includes the selection of ((

)).

6. As discussed in section 5.1.3 of the TR, the applicability of the in-vessel transients with radiological release methodology for licensing analyses is restricted to those events that do not experience severe accident phenomena (e.g., coolant boiling, gross cladding failure, significant fuel melting and relocation).

7. An applicant or licensee referencing the methodology described in TR section 5.4 for performing detailed mechanical analysis for FHAs must submit documentation and justify that the development and assessment of this methodology has been completed to a state appropriate for the intended licensing application.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION CONCLUSION The NRC staff has determined that TerraPowers TR NAT-9394, Design Basis Accident Methodology for Events with Radiological Release, Revision 0, provides an acceptable approach to develop a methodology for use by future applicants utilizing the Natrium design as described in the TR and this SE to evaluate DBA events with radiological release because the assumptions made for each EM, EM development plans, selected calculational devices, planned conservatisms, and EM assessment plans are appropriate for analyzing the Natrium design, as discussed in this SE. This approval is subject to the limitations and conditions discussed in the previous section of this SE.

Principal Contributor(s): R. Anzalone, NRR Z. Gran, NRR M. Hart, NRR A. Neller, NRR

Verify Current Revision Document

Title:

Design Basis Accident Methodology for Events with Radiological Release Natrium Document No.:

NAT-9394 Rev. No.:

0 Page:

1 of 65 Target Quality Level:

N/A Alternate Document No.:

NAT-9394-NP Alt Rev.:

N/A Originating Organization:

TerraPower, LLC (TP)

Quality Level:

N/A Natrium MSL ID:

N/A Status:

Released Open Items?

0 Approval Approval signatures are captured and maintained electronically; see Electronic Approval Records in EDMS.

Signatures or Facsimile of Electronic Approval Record attached to document.

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REVISION HISTORY Revision No.

Affected Section(s)

Description of Change(s) 0 All Initial Release - Supersedes TP-LIC-RPT-0007 Rev. 0. Incorporates changes made to address NRC questions during audit review.

Changes from previous information marked via change bars.

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TABLE OF CONTENTS EXECUTIVE

SUMMARY

................................................................................................................... 7 1

PURPOSE................................................................................................................................... 8 2

BACKGROUND......................................................................................................................... 10 2.1 Regulatory Requirements and Guidance for DBAs.......................................................... 10 2.2 Plant Description............................................................................................................... 10 2.3 Safety System Classification............................................................................................. 14 2.4 DBA Event Selection......................................................................................................... 15 3

ASSUMPTIONS REQUIRING VERIFICATION......................................................................... 19 3.1 Assumptions..................................................................................................................... 19 4

EVALUATION MODEL DEVELOPMENT AND ASSESSMENT................................................ 21 4.1 Background....................................................................................................................... 21 4.2 Evaluation Model Development........................................................................................ 23 4.3 Phenomena Identification and Ranking Tables (PIRT)..................................................... 25 4.4 Evaluation Model Assessment.......................................................................................... 26 5

EVENT-SPECIFIC METHODOLOGY........................................................................................ 30 5.1 In-vessel Transients with Radiological Release Methodology.......................................... 30 5.2 Partial Flow Blockage Methodology.................................................................................. 37 5.3 Fuel Misload Methodology................................................................................................ 39 5.4 Fuel Handling Accident Methodology............................................................................... 41 5.5 Sodium Liquid and Gas Leak Methodology...................................................................... 45 6

SUMMARY

................................................................................................................................ 49 6.1 Summary of Codes Selected............................................................................................ 49 6.2 EM Conservatism Summary............................................................................................. 50 7

CONCLUSIONS AND LIMITATIONS........................................................................................ 52 7.1 Conclusions...................................................................................................................... 52 7.2 Limitations......................................................................................................................... 52 8

REFERENCES.......................................................................................................................... 53 9

APPENDICES............................................................................................................................ 54 9.1 Appendix A - Additional Details of the DSAW Process.................................................... 54 9.2 Appendix B - Initial Experimental Database for Fuel Performance and Radiological Release/Transport Methodology................................................................................................ 58

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LIST OF TABLES Table 2-1. Natrium DBAs with Radioactive Material Release.......................................................... 17 Table 4-1. Figures of Merit for In-Vessel DBAs [3].......................................................................... 21 Table 4-2. Representative Events with Potential Fuel Failure and Radiological Release............... 23 Table 4-3. Assessment Matrix for High/Medium Importance Fuel Failure Phenomena in OQEs.... 28 Table 4-4. Assessment Matrix for High-Importance Fuel Failure Phenomena................................ 29 Table 5-1. Combined PIRT for ULOF, ULOHS and UTOP LBEs with Radiological Release with High/Medium Importance Phenomena............................................................................................ 33 Table 5-2. List of Medium and High Importance phenomena in SPS leak events [4]..................... 46 LIST OF FIGURES Figure 2.2-1. Plant Layout............................................................................................................... 11 Figure 2.2-2. Natrium Elevation View.............................................................................................. 13 Figure 4.1-1. EM Calculational Devices and Analysis Workflow..................................................... 23 Figure 5.1-1. DSAW Data Flow....................................................................................................... 35 Figure 5.3-1. Equilibrium Core Fuel Layout..................................................................................... 39 Figure 5.5-1. Sodium Cleanup System and IHT Leak EM Diagram................................................ 48

NAT-9394 Rev 0 Design Basis Accident Methodology for Events with Radiological Release Page 5 of 65 Not Confidential Controlled Document - Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright © 2024 TerraPower, LLC. All rights reserved.

ACRONYMS ANL Argonne National Laboratory AOO Anticipated Operational Occurrence ATR Advanced Test Reactor BDBE Beyond Design Basis Event CDF Cumulative Damage Fraction CFD Computational Fluid Dynamics CGD Commercial Grade Dedication CP Construction Permit DBA Design Basis Accident DBE Design Basis Event DOE Department of Energy DSAW Detailed Safety Analysis Workflow DSC Differential Scanning Calorimetric EBR-II Experimental Breeder Reactor II EM Evaluation Model EPZ Emergency Planning Zone EVHM Ex-Vessel Handling Machine EVST Ex-vessel Storage Tank F-C Frequency-Consequence FCCI Fuel Cladding Chemical Interaction FEA Finite Element Analysis FH Fuel Handling FHA Fuel Handling Accident FHB Fuel Handling Building FOM Figure of Merit HAA Head Access Area HCF Hot Channel Factor IAC Intermediate Air Cooling IET Integral Effects Test IHT Intermediate Heat Transport System INL Idaho National Laboratory ISP Intermediate Sodium Pump IVTM In-Vessel Transfer Machine LBE Licensing Basis Event LHGR Linear Heat Generation Rate LHR Linear Heat Rate LWR Light Water Reactor MFF Mechanistic Fuel Failure NI Nuclear Island NRC Nuclear Regulatory Commission

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NSRST Non-Safety-Related with Special Treatment NST No Special Treatment OQE Other Quantified Event ORNL Oak Ridge National Laboratory PCT Peak Cladding Temperature PHT Primary Heat Transport System PIC Pool Immersion Cell PIRT Phenomena Identification and Ranking Table PSAR Preliminary Safety Analysis Report QA Quality Assurance QAPD Quality Assurance Program Description RAB Reactor Auxiliary Building RAC Reactor Air Cooling System RCC Reactor Core System RES Reactor Enclosure System RN Radionuclide RSF Required Safety Function RV Reactor Vessel RVH Reactor Vessel Head RWG Gaseous Rad Waste Processing System RXB Reactor Building

((

))(a)(4)

SCG Sodium Cover Gas System SFP Spent Fuel Pool SFR Sodium-Cooled Fast Reactor SMP Software Management Procedure SPS Sodium Processing System SQA Software Quality Assurance SR Safety Related SSCs Structures, Systems, and Components TATNF Time at Temperature No Failure TREAT Transient Reactor Test Facility ULOF Unprotected Loss of Flow ULOHS Unprotected Loss of Heat Sink UTOP Unprotected Transient Over Power V&V Verification and Validation

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EXECUTIVE

SUMMARY

This topical report provides a high-level road map for, and summary of, the Design Basis Accident Methodology for Events with Radiological Release for the Natrium' reactor, a TerraPower & GE-Hitachi Technology. It describes the evaluation model (EM) development, the resulting EMs, and identifies EM items which require further development. Certain aspects of the EM adequacy demonstration remain in development and are noted throughout the report. It is acknowledged that this report contains preliminary technical information, and several sections within describe future actions that are planned to be taken by TerraPower.

Information generated by these actions will be provided in future licensing submittals. These actions are expected to be complete prior to use of this EM in support of an operating license application.

This report contains six chapters and two appendices.

Chapter 1 discusses the overall objective and scope of the report.

Chapter 2 discusses the regulatory requirements and guidance used in the EM development process, and a high-level description of the Natrium nuclear power plant. Chapter 2 also identifies the safety systems and design basis accidents that pertain to the Design Basis Accident with radiological release EM development.

Chapter 3 lists Assumptions and Open Items.

Chapter 4 discusses the general EM requirements, the independently submitted topical reports that are utilized, and the capability development for analysis of different design basis accident (DBA) with release scenarios.

Chapter 5 discusses the event-specific EMs/methodologies established for analysis of

In-vessel transients with radiological release (Section 5.1)

Partial flow blockage (Section 5.2)

Fuel misload (Section 5.3)

Fuel handling accidents (Section 5.4)

Sodium liquid and gas leaks (Section 5.5)

Each section includes the following subsections:

Purpose and scope

Assumptions

EM scope and requirements

EM descriptions

EM assessment

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Chapter 6 provides some conclusions on the EM development and summarizes the limitations and conservatisms of the EMs.

Appendix A provides details on the use of Time at Temperature No Failure (TATNF) and related analyses.

Appendix B provides a list of legacy experimental data available for EM verification and validation.

1 PURPOSE This topical report addresses the Natrium' nuclear power plant DBA with radiological release EM development process, the resulting EM, and identifies EM development items which require further development. The methodology development guidance provided in the Natrium Reactor Project General Methodology Development and Assessment Guide, was used in the development of this EM.

The Natrium power plant being developed by TerraPower follows the methodology provided in NEI 18-04, Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development, to identify and evaluate Licensing Basis Events (LBEs) including frequency based Anticipated Operational Occurrences (AOOs), Design Basis Events (DBEs), Beyond Design Basis Events (BDBEs), and conservative assumption oriented DBAs [1]. Additionally, the identification and classification of safety-related (SR) and non-safety-related with special treatment (NSRST) structures, systems, and components (SSCs) are determined consistent with the methodology presented in NEI 18-04. Figure 1-1 provides a graphical representation showing the AOO, DBE, BDBE, and DBA relationships as well as how they fit within the complete event structure from a frequency perspective.

Figure 1-1. Frequency oriented relationship between AOOs, DBEs, BDBEs, and DBAs.

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The guidance provided in NEI 21 Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report: For Applicants Utilizing NEI 18-04 methodology - is followed in the development of the Natrium Preliminary Safety Analysis Report (PSAR) [2]. The PSAR is being developed in accordance with the two-part licensing approach established in 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, which involves first obtaining a Construction Permit (CP) followed by an Operating License. The PSAR is submitted to the Nuclear Regulatory Commission (NRC) as part of the CP application process.

It is important to note that the PSAR will necessarily contain preliminary design information which must be updated as the process reaches conclusion, and an Operating License is requested.

NEI 21-07 states the following with respect to DBA analytical method discussion in the PSAR, The applicant should describe the overall analytical methodology and identify and describe the significant computer codes used to model the plant response. The applicant should address the applicability of the analytical methodology to the characteristics of the plant, including a discussion of the underlying experimental or analytical basis. Typically, this is done through NRC-reviewed and approved topical reports that are incorporated by reference in the SAR or through technical reports that are summarized in the SAR and available for regulatory audits.

To support development of the PSAR, this report provides discussion of the evaluation model development used to evaluate the Natrium plant response where the release of radioactive material is a possible consequence of a DBA. Furthermore, consistent with NEI 21-07, for these scenarios, a mechanistic source term is used in the calculation of the consequences.

This report provides a high-level discussion of the following issues associated with the evaluation model development for Natrium DBAs with radioactive material release:

DBA event selection

Important processes and phenomena

Overall analytical methodology

Identification and description of significant computer codes used to model the plant responses

Applicability of the analytical methodology to the characteristics of the plant

Underlying experimental or analytical basis for model assessment and model pedigree In the Natrium plant, DBA scenarios can be grouped into two basic physical areas: in-vessel scenarios and ex-vessel scenarios.

In-vessel scenarios include traditional reactor transient scenarios leading to fuel damage and subsequent release of radioactive material. Furthermore, in-vessel scenarios include flow blockage scenarios, fuel handling accident scenarios and loss of active cooling scenarios.

The principal difference between reactor transients and the other in-vessel scenarios involves the use of ((

))(a)(4) to analyze the reactor transient phenomena while the other scenarios do not require this code system.

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Ex-vessel scenarios include fuel handling accident scenarios, loss of active cooling scenarios, and radioactive sodium and gas leak scenarios. Apart from ((

))(a)(4) which is not used for ex-vessel scenarios involving the release of radioactive material, these scenarios require a similar set of computer codes that are used for the analysis of in-vessel scenarios. The codes are noted and discussed throughout this report.

2 BACKGROUND 2.1 Regulatory Requirements and Guidance for DBAs DBA postulated accidents are used to set design criteria and limits for the design and sizing of safety-related systems and components. Further, as noted in NUREG-2122, a DBA is a postulated accident that a nuclear facility must be designed and built to withstand without loss to the systems, structure, and components necessary to ensure public health and safety. The definition put forth in NEI 18-04 is:

Postulated event sequences are used to set design criteria and performance objectives for the design of Safety Related SSC. DBAs are derived from DBEs based on the capabilities and reliabilities of Safety-Related SSCs needed to mitigate and prevent event sequences, respectively. DBAs are derived from the DBEs by prescriptively assuming that only Safety Related SSCs are available to mitigate postulated event sequence consequences to within the 10 CFR 50.34 dose limits.

2.2 Plant Description The 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 with metal fuel. 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 Power Reactor Innovative Small Module technology and TerraPowers Traveling Wave Reactor technology.

The general plant layout is shown in Figure 2.2-1 and is made up of two basic areas; a Nuclear Island where the reactor and associated support facilities reside and an Energy Island where thermal storage tanks and turbine facilities for generating electricity reside. Safety functions are made integral to the reactor vessel and support equipment is moved to separate structures in the Energy Island, resulting in a simplified reactor building. Decoupling the Nuclear Island from the Energy Island from a nuclear safety perspective is central to simplifying the Natrium design. The Natrium design capitalizes on the proven metal fueled SFR safety characteristics to minimize the number of safety-related SSCs needed to achieve safety goals.

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Figure 2.2-1. Plant Layout.

The Natrium plant uses a pool-type design with the reactor core and primary coolant pumps located within a large pool of primary sodium coolant and no penetration through the reactor vessel thereby eliminating loss of coolant accidents involving primary pumps and piping. The primary sodium pool operates at near atmospheric pressure. Heat is transferred from the hot primary sodium pool to an intermediate sodium piping loop by means of two intermediate heat exchangers. The intermediate piping loop uses non-radioactive sodium to transport reactor heat from each intermediate heat exchanger to two sodium/salt heat exchangers. These sodium/salt heat exchangers in the Nuclear Island heat salt received from the cold salt tank in the Energy Island. The heated salt is then returned to the Energy Island for storage in the hot salt tank, which serves as thermal energy storage. The salt stored in the hot tank is used to generate steam for use in steam turbine generators eliminating the need for generating steam directly from reactive sodium metal. The Natrium plant can vary its supply of energy to the grid through its energy storage system. The Natrium reactor operates at a thermal power of 840 MW while the plant produces 336 MWe steady-state and 500 MWe peak power. The thermal energy storage system, located in the Energy Island, uses two molten salt tanks, one hot and one cold. Its architecture is like molten salt systems for concentrated solar power. The charging salt loop transports salt from the cold tank to the reactor for heating and routes it to the hot tank. The steam / salt loop transports salt from the hot tank to steam generators to generate superheated steam and returns salt to the cold tank.

The Natrium plant has been designed to accomplish reactivity control with multiple layers.

The non-safety-related reactor control system acts as a buffer to prevent the need for a scram.

It detects abnormal operation and initiates a runback via motor driven insertion of neutron absorbing control rods to achieve a softer shutdown than a scram.

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The safety-related reactor protection system initiates a scram if the reactor control system fails, or a runback fails to prevent the reactor from reaching a scram setpoint. The high reliability scram function is initiated by removing electrical power to an electromagnet, resulting in insertion of all control and standby rods into the reactor core.

The reactor core is designed with a negative temperature and power coefficient that is strong enough such that the reactor can accommodate anticipated transients without scram for events such as loss of primary flow, loss of heat sink, and uncontrolled rod withdrawal.

The high boiling point of sodium allows reactor operation at atmospheric pressure. A close-fitting guard vessel stops the loss of coolant should the reactor vessel develop a leak.

Furthermore, the reactor cover gas operates at essentially atmospheric pressure so there is little driving force for a release.

The Natrium plant is designed to accomplish residual heat removal with multiple layers of protection.

Forced flow heat removal via Intermediate Air Cooling (IAC) serves as the normal shutdown cooling system for outages. There are two trains, one for each primary heat exchanger. The IAC has two cooling modes: forced flow and passive flow. For the final heat sink, it transfers heat to the atmosphere from the sodium-air heat exchangers. Simple operation of a fail-open electromagnetic damper initiates passive cooling. Active operations support normal controlled cooling operations (such as during a refueling outage) and in response to anticipated transient events. Forced flow is provided by air blowers and the intermediate sodium pumps (ISPs). The IACs natural draft arrangement permits passive operation of the system as a diverse alternative if power to support forced cooling is not available. These functions supplement the safety-related Reactor Vessel Air Cooling (RAC) system and, as a result, enable the IAC and its support system designs to be non-safety related.

The RAC removes decay heat using natural circulation of air around the exterior of the reactor vessel. The RAC does not have any dampers. RAC is always operating and requires no power, people, or control action to perform its function. The RAC relies on the natural circulation performance of the primary sodium and conductive/convective heat transfer to the reactor vessel wall. Thermal radiation heat transfer then dominates heat transfer to the guard vessel. Natural draft air inlets provide ambient outside air to cool the guard vessel wall via a combination of radiative and convective heat transfer.

The Nuclear Island is composed of six major buildings: reactor, fuel handling, control, electrical, reactor auxiliary, and fuel auxiliary buildings. The reactor building, see Figure 2.2-2, houses two major components: the reactor and RAC air ducts. The reactor is located below grade to protect it from natural hazards (earthquakes, tornadoes, etc.) and other hazards.

There are only two rooms in the reactor building, the refueling access area, where refueling and maintenance takes place, and the head access area where limited maintenance takes place. Intermediate sodium piping exits the reactor building below ground to the reactor auxiliary building.

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Figure 2.2-2. Natrium Elevation View.

The fuel handling building houses fuel receipt equipment, refueling equipment, fuel storage equipment, and the fuel storage pool. Casks are used to transport fuel and in-reactor components from the reactor building to the fuel handling building. The buildings are connected by a rail system at ground level to support movement of the fuel handling cask. The fuel handling building also contains the mechanical handling equipment which moves assemblies and provides access to the fuel pool. A bridge crane supports movement of dry storage fuel casks and equipment within the facility.

The Nuclear Island (NI) Control Building uses a structural steel braced frame supported on a concrete grade slab with insulated metal siding and an insulated standing seam metal roofing or membrane roofing system. During normal operations, systems will be monitored and controlled from this building.

The Reactor Enclosure System (RES) contains and supports the reactor core and primary sodium coolant, including all supporting equipment and structures. The RES is divided into five subsystems: Reactor Vessel (RV), reactor internals, Reactor Vessel Head (RVH), Guard Vessel, and Reactor Support Assemblies. All subsystems are in, and are either directly or indirectly supported by, the Reactor Building. The RV, along with the RVH, form most of the reactor coolant and primary cover gas boundaries. Finally, the RV and RVH provide support for the reactor internals as well as the Core Support Structure, which supports the reactor core.

The In-Vessel Transfer Machine (IVTM) moves core assemblies between the core, in-vessel fuel storage racks, and transfer station for removal from the reactor vessel. It is mounted on the reactor rotatable plug, which is centered within the reactor top plate. The IVTM consists of two subassemblies: the above-head drive assembly and the in-vessel fuel handling mechanism. The latter extends to reach all removable core assembly locations when used in

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conjunction with the rotatable plug. Core assemblies are transferred into and out of the reactor vessel with the fuel transfer lift operating through the reactor transfer adapter. Fresh core assemblies are transferred into the fuel transfer lift and are then lowered into the pool region by the fuel transfer lift to core level to be transferred into the core using the IVTM. Used core assemblies are transferred out of the core to the in-vessel storage for decay or directly to the fuel transfer lift for assemblies which do not require in-vessel decay. The IVTM and fuel transfer lift are installed at the beginning of a refueling outage, the IVTM installed on the rotating plug assembly, and the fuel transfer lift penetrating the reactor vessel head. They make up part of the functional containment boundary during refueling operations and are removed after refueling is complete.

The ex-vessel fuel handling system components transfer all new reactor core assemblies from the point of receipt from the supplier through inspection and conditioning to the reactor vessel.

The ex-vessel fuel handling components also receive and transfer irradiated core assemblies to the Ex-Vessel Storage Tank (EVST). Following the outage, offloaded assemblies in the EVST are transferred to and processed through the Pool Immersion Cell (PIC) into the spent fuel pool (SFP). The PIC provides the sodium residue removal allowing the assemblies to be stored in water for operations such as waste consolidation for non-fuel assemblies and underwater cask loading for used fuel assemblies. When desired decay heat limits are reached for used fuel assemblies they are processed into conventional dry casks and transferred to site storage pads for interim dry storage.

The water pool fuel handling system contains the equipment and structures needed to load, store, and retrieve irradiated core assemblies and used fuel assemblies from the spent fuel pool. After the core assemblies have had the sodium residue removed and have been immersed in water, the water pool fuel handling machine moves the core assemblies to the SFP. In the SFP, the core assemblies undergo long term decay before being removed using a cask.

The fuel transport and storage system packages and transports irradiated core assemblies for long term dry storage. It consists of the cask transporter and the interim dry storage pad. The dry cask transporter navigates to the cask transporter pickup location where the water pool fuel handling system has prepared and staged the dry storage cask for pickup.

2.3 Safety System Classification The Natrium plant uses three safety classification levels: SR, NSRST, and No Special Treatment (NST). Explanations for each of the three classifications are provided below.

Safety-Related (SR)

SSCs selected from the SSCs that are available to perform the Required Safety Functions (RSFs) to mitigate the consequences of DBEs to within the Licensing Basis Event (LBE)

Frequency-Consequence (F-C) target, and to mitigate DBAs that only rely on the SR SSCs to meet the dose limits of 10 CFR 50.34 using conservative assumptions.

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SSCs selected from the SSCs that are available and relied on to perform RSFs to prevent the frequency of BDBE with consequences greater than the 10 CFR 50.34 dose limits from increasing into the DBE region and beyond the F-C target.

Non-Safety-Related with Special Treatment (NSRST)

Non-safety-related SSCs relied on to perform risk-significant functions. Risk-significant SSCs are those that perform functions that prevent or mitigate any LBEs from exceeding the F-C Target or make significant contributions to the cumulative risk metrics selected for evaluating the total risk from all analyzed LBEs. Non-safety-related SSCs relied on to perform functions requiring special treatment for defense in depth adequacy. These SSCs are safety-significant even if they are not risk-significant.

Non-Safety-Related with No Special Treatment (NST)

All other SSCs (with no special treatment required).

2.4 DBA Event Selection The DBAs identified for the Natrium design can be broadly categorized as:

In-vessel core transients with fuel failure includes symmetric and asymmetric Primary Heat Transport System (PHT) and Intermediate Heat Transport System (IHT) initiated events, the loss of hydraulic holddown, RAC long-term transient, etc.

Local faults (including partial flow blockage and fuel misload)

Fuel handling events

Radioactive gas/liquid leakage/release events.

Of the DBAs identified, ten of the DBAs have descriptions indicating they involve a potential release of radioactive material and are listed in Table 2-1. It should be noted that the list of DBAs in Table 2-1 represent the currently identified events and are provided to help illustrate the methodology in the report only, not to define the set of events applicable to all Natrium plants. Applications incorporating this report by reference will utilize the methodology outlined in the report for the relevant events defined in the application. To determine which DBAs have the potential for release of radioactive material, the following process is used. ((

))(a)(4) Once the DBAs are established, they are distributed to the functional groups for evaluation.

In-vessel DBAs are first analyzed with the in-vessel DBA without release EM [3] to obtain cladding temperature results. These results are then compared to the TATNF screening criteria.

DBAs which do not violate the TATNF screening criteria are then excluded from further consideration of potential release. Those that violate the TATNF screening criteria are taken for further evaluation with the Detailed Safety Analysis Workflow (DSAW) process as described in Section 5.1.5.

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DBAs which are not in-vessel are evaluated using the appropriate methodology described below, an appropriate event-specific method, or evaluated with the source term EM using conservative assumptions. Thus, essentially, any DBA which is not precluded from having releases based on the DBA without release EM and TATNF screening criteria are assumed to have the potential for radionuclide release.

Note that the following events in Table 2-1 do not have a corresponding EM within this topical report:

((

))(a)(4): There is not a specific EM within this topical report for this event to determine the extent of fuel failure due to an excessive sodium chemical reaction. The Radiological Source Term Methodology [4] simply takes a conservative assumption regarding the potential release from this event.

((

))(a)(4): There is not a specific EM within this topical report for this event to determine extent of fuel failure or confirm clad temperatures are in acceptable ranges (no release). For the PSAR, an event-specific calculation is performed and resides in the individual analysis. A mature methodology defining how to approach this scenario will be included in a future licensing document.

((

))(a)(4): The quantification of the RWG leak is not a part of this EM. The Radiological Source Term Methodology [4] simply takes a conservative assumption regarding the potential release from this event.

The DBAs involving potential fuel failure and releases can be broadly categorized as:

Fuel handling events

Component failures and malfunctions

Loss of cooling

System leaks

Sodium-water interaction

Natural phenomena events.

The DBAs can also be grouped into two basic classes: in-vessel scenarios and ex-vessel scenarios.

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Table 2-1. Natrium DBAs with Radioactive Material Release.

Identifier Topic Summary Core Blockage and Local Faults (DBA)

During at-power operations, the postulated initiating event is a blockage of fuel subchannels or other localized faults within the reactor core. While a manual shutdown would normally be initiated due to exceeding failed fuel limits, the reactor is assumed to continue operating at full power. The creep failure of all pins at the highest burnup is assumed for the affected single assembly using at power conditions. This single assembly failure does not have a significant impact on monitored safety related plant parameters. The vessel head is able to contain the radionuclide release (e.g., no pre-existing leak or seal failures are assumed).

Excessive Sodium-Water Reaction in the PIC (DBA)

During ex-vessel fuel handling operations, the postulated initiating event is an excessive sodium water reaction in the PIC. The cladding integrity is failed following the sodium water reaction that occurs in the PIC. The BLTC boundary successfully retains the radionuclide release following the sodium water reaction (PIC boundary is not credited for DBA). The fuel cladding is failed within this scenario and radionuclide release occurs.

Fuel Handling Event Occurs While Moving Fuel Assembly in the Reactor Vessel (DBA)

During refueling operations, the postulated initiating event is a fuel handling event while moving fuel in the reactor vessel. Assembly(s) impacted by the dropped component are damaged and a radionuclide release occurs. The functional containment barriers successfully retain the radionuclide release.

Fuel Handling Event Occurs While Moving Fuel Assembly in the Spent Fuel Pool (SFP)

(DBA)

During ex-vessel fuel handling operations, the postulated initiating event is a fuel handling event while moving fuel in the SFP. The fuel assembly is damaged and a radionuclide release occurs. The fuel cladding is failed within this scenario and radionuclide release occurs.

Loss of EVST Cooling While Storing Fuel Assembly (DBA)

A loss of active EVST cooling occurs while handling spent fuel in the EVST. Analysis demonstrates 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> adiabatic heat up of the EVST and EVST vault to maintain fuel within performance limits. Longer term degraded heat removal conditions require further assessment.

))(a)(4)

SCG leak Downstream the SCG Cell (DBA)

A Sodium Cover Gas System (SCG) leak occurs downstream the SCG cell. The Reactor Building (RXB) superstructure cannot contain radionuclides. The fuel cladding is not failed within this scenario, however radionuclide release occurs.

((

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Identifier Topic Summary SPS-I leak at the cold trap (DBA)

An Intermediate Sodium Processing System (SPS) leak occurs at the cold trap which is skid-mounted on the ground floor of the Reactor Auxiliary Building (RAB). A release of precipitated tritium from the SPS-I cold trap occurs. The RAB cannot contain radionuclides. The fuel cladding is not failed within this scenario, however radionuclide release occurs.

RWG leak (DBA)

A Gaseous Rad Waste Processing system (RWG) leak occurs. The fuel cladding is not challenged within this scenario; however, radionuclide release occurs.

SPS-P System leak in the RAB (DBA)

The SPS-P system leaks within the RAB. The SPS pump trips on low primary sodium level which stops the leakage due to system configuration. The SPS cell does not contain the radionuclides. The fuel cladding is not failed within this scenario, however radionuclide release occurs

))(a)(4)

SPS-P Leak in the RXB (DBA)

The SPS-P system leaks inboard of the SPS isolation valves within the HAA. The SPS pump trips on low primary sodium level which stops the leakage due to system configuration. The HAA does not contain the radionuclides. The fuel cladding is not failed within this scenario, however radionuclide release occurs.

((

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3 ASSUMPTIONS REQUIRING VERIFICATION 3.1 Assumptions The following assumptions are discussed in more detail in the individual EMs within Section 5 of this report. They are summarized here for information and provide context for items which are assumed to define the scope of an EM, determine conservative boundaries, or to identify areas in which future work is planned.

Assumption Number Description 3.1 Event and accident scenarios will be limited to DBAs with potential release.

3.2 This report is based on the current Natrium reactor system design and will be revised as appropriate as the reactor design and possible event scenarios mature.

3.3 The system leakage scenarios (DBAs resulting from leakage or breaks in the SPS, IHT, RWG, or SCG) are assumed during normal operation and not as part of, or consequence of, a different event.

3.4 It is presumed that ((

3.5

))(a)(4) 3.6 For a conservative scoping calculation of the potential radiological release happening during a fuel handling accident, a fuel assembly ((

))(a)(4) 3.7 Detailed analysis of fuel drop accident considers limited scenarios ((

))(a)(4) result in the worst possible fuel damage and the highest radiological release.

3.8 The partial flow blockage analysis is performed in a Natrium assembly that is operating at the fuel design limits. The assembly operates with a peak Linear Heat Generation Rate (LHGR) ((

))(a)(4),ECI which may be updated with evolving fuel performance analysis, and a peak cladding temperature (PCT) ((

))((a)(4),ECI EXPORT CONTROLLED INFORMATION

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Assumption Number Description 3.9 The steady state tools used to design the reactor core have the fidelity to model the misloaded core ((

))(a)(4) 3.10 The final Natrium design ((

3.11

))(a)(4) 3.12 The assembly to be misloaded is ((

))(a)(4) 3.13 The In-Vessel Transients with Radiological Release Methodology assumes that only Type 1 fuel is used.

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4 EVALUATION MODEL DEVELOPMENT AND ASSESSMENT

4.1 Background

The EMs considered in this report support the analysis of DBAs in the Natrium design involving clad or fuel failure with potential release of radionuclides into the coolant or beyond functional containment barriers and subsequent discharge into the environment. These EMs should be able to describe important phenomena identified by the relevant Phenomena Identification and Ranking Table (PIRT) studies with adequate accuracy and fidelity.

Table 4-1 provides the Figures of Merit (FOM) that have been used in the analyses of in-vessel DBAs without radiological release [3] and are also relevant to the in-vessel DBAs involving radiological release. Note that this listing represents the FOMs as a snapshot in time and will be updated accordingly as the FOMs evolve.

Table 4-1. Figures of Merit for In-Vessel DBAs [3]

Figure of Merit Descriptions and Significance Fuel centerline temperature The fuel centerline temperature must stay below the fuel solidus temperature to avoid fuel damage. Since the fuel solidus temperature is much higher than the fuel-cladding eutectic reaction onset temperature, it is expected that the PCT will be a much more limiting criteria than the fuel centerline temperature.

Coolant temperature High coolant temperature may cause sodium boiling in the reactor core, ((

))(a)(4) In addition, this phenomenon can be used to examine the primary boundary integrity. This FOM is tracked, however the acceptance criteria for time-at-temperature no-failure (TATNF) for PCT is designed to preclude boiling.

Time-at-temperature for PCT The design basis approach and limit values of the PCT were evaluated for application to the Natrium design. For mechanical fuel pin cladding failure criteria, the main options include strain, cumulative damage fraction (CDF), stress, and temperature as primary or dependent criteria parameters. The Natrium design basis has adopted response parameters such as strain, wastage, and temperature rather than CDF and stress criteria because they have a historic precedent, are defensible by existing data, are readily analyzed, and can be measured to validate. These attributes allow for monitoring and surveillance that can confirm analysis predictions and assess remaining life of the fuel system. The TATNF screening criteria incorporate cladding wastage and thermal creep criteria in assessing potential failure.

The TATNF FOM is constrained by the following:

((

))(a)(4)

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Figure of Merit Descriptions and Significance

((

))(a)(4),ECI The full scope of Natrium EMs is composed of many codes and methods which span the range of initiating events that can result in clad or fuel failures and system leaks that lead to radiological release.

Figure 4.1-1 provides a high-level depiction of the EM workflow associated with DBAs. This high-level view illustrates the use of multiple independently licensed EMs to evaluate the dose consequences of Natrium DBA events. These individual EMs provide the foundational development and validation for the events described in this report. The EMs include:

Core Design and Thermal Hydraulics [5],

Design Basis Accident Methodology for In-Vessel Events without Radiological Release [3],

Fuel and Control Assembly Qualification [6],

Partial Flow Blockage Methodology [7]

Radiological Source Term Methodology [4]. and

Radiological Release Consequences Methodology [8]

This viewpoint aids in identifying possible gaps between the development, qualification and licensing of the individual EMs and their application to events resulting in a radiological release.

Some DBAs that have a potential for fuel failures will be in-vessel transients ((

))(a)(4) These transients are analyzed with a multi-step process as discussed in Section 2.4 and illustrated in the upper path in Figure 4.1-1.

Using the TATNF screening criteria effectively identifies the bounding events and filters out those that do not require more detailed analysis. ((

))(a)(4) where fuel pin failure is possible due to a combination of factors including prior irradiation exposure, the extent of cladding wastage (fuel-clad chemical interaction, eutectic, etc.), and the extent of cladding mechanical creep.

The TATNF screening criteria provides a framework to decide if a safety analysis event that occurs in this temperature range requires further assessment as described in Section 5.1.

Other DBA events that can lead to radiological releases are presented in the lower branch of Figure 4.1-1 and discussed in Sections 5.2 through Section 5.5.

Regardless of the methods used to determine the clad and/or fuel pin failures, all events will provide information to the radiological source term EM [4] and then to the dose consequence EM to determine the transport and consequence of radiological release to the environment [8].

EXPORT CONTROLLED INFORMATION

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Figure 4.1-1. EM Calculational Devices and Analysis Workflow.

4.2 Evaluation Model Development As shown in Table 4-2, the Natrium events with potential fuel failure and/or radiological release can be grouped based on the common important phenomena, location, and modeling objectives/requirements. Suitable modeling strategies and EM are then established for each group.

Table 4-2. Representative Events with Potential Fuel Failure and Radiological Release.

Event Category Event Location Phenomena Suitable Software Core/PHT/IHT events Core symmetric events In-vessel Core neutronics, fuel behavior, and coolant thermal hydraulics

((

))(a)(4)

Core asymmetric events - one-pump trip In-vessel Core neutronics, fuel behavior, and coolant thermal hydraulics Loss of heat sink, RAC long-term transient Ex-vessel Heat removal from PHT system (a)(4)

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Local events Partial flow blockage In-vessel Thermal hydraulics behavior/Release/Tr ansport

((

Fuel misload In-vessel Coupled thermal hydraulics/neutronic s/fuel performance behavior Fuel handling accidents In-vessel fuel drop, ex-vessel fuel drop at different locations In-/Ex-vessel Transport and consequence of radiological release Thermo-mechanical/structura l-mechanical behavior/failure Sodium/gas leaks/releases Sodium/gas leaks/releases Ex-vessel Transport of quantified radionuclide release

((

))(a)(4)

Accident and safety analysis of SFRs is not as mature as that of light water reactors and many gaps in the SFR safety modeling capabilities have been identified and continue to be addressed. [9]

Instead of developing new modeling capabilities, safety analysis of the Natrium design is primarily focused on the use of readily available modeling tools which are selected and acquired via TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk. The plan provides a process framework supporting the quality assurance (QA) requirements for software (computer codes) that perform safety-related or non-safety-related analysis in the Natrium plant, ((

))(a)(4) In addition, the gap analysis and planned maturation activities for each potential software are discussed. Specific sections are included in the plan to discuss the commercial grade dedication (CGD) that will be implemented for commercially acquired software that will be used for safety-related applications.

DBAs involving reactor core and PHT systems and components ((

))(a)(4) models the core at the assembly level, i.e., each fuel assembly is represented by a single channel comprising the fuel, cladding, coolant, and associated structure; detailed analysis at the local fuel pin/subchannel level requires the ability to model the multidimensional phenomena within the fuel assembly. The codes selected and developed for this use are ((

))(a)(4) which characterize the individual fuel pins that are expected to fail. The failed rod(s) initial radionuclide inventory is used in the radiological source term method [4] that determines the leakage through facility systems and the dose consequences of the release to the environment.

In-vessel partial flow blockage events (Section 5.2) that involve subchannel coolant thermal hydraulics ((

))(a)(4)

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Other In-vessel events that result in pin failures are Fuel Misloads (Section 5.3), and Fuel Handling Events (Section 5.4). Fuel Handling Events may also occur ex-vessel in the Ex-Vessel Fuel Handling Machine (EVHM), EVST, the PIC, the SFP, and possibly during transport between the locations.

Structural analysis software, ((

))(a)(4) can be used to analyze the thermo-mechanical and structural mechanical behavior of a fuel assembly and evaluate potential failure resulting from fuel handling accidents.

DBAs involving system leaks are often analyzed with use of the radiological source term method [4], ((

))(a)(4) as described in Section 5.5. However, some system leaks such as large leaks in the IHT could lead to a sequence of events that lead to an In-Vessel Transient of sufficient magnitude to cause fuel pin failures using the In-Vessel Transient Methodology (see Section 5.1).

If an event involves more than one of the phenomena mentioned above, different modeling tools can be used together with different code coupling/interfacing strategies (one-way or two-way) to be employed. ((

))(a)(4) 4.3 Phenomena Identification and Ranking Tables (PIRT)

Important phenomena and processes are identified and ranked for each event category via the PIRT study. The PIRT objective is to identify safety-relevant phenomena and processes for the considered event, rank their importance based on pre-established FOMs, and rank the status of knowledge to build a technical basis to develop the EM.

The process for establishing a PIRT is iterative in nature and follows a pattern of progressive elaboration that consistently drives the PIRT to move from qualitative discussion to quantitative descriptions. Whereas early phases of the PIRT process make heavy use of independent expert opinion and precedent PIRTs where applicable, the later phases take benefit of detailed computational analyses that provide direct and indirect evidence of phenomenological importance and impact of identified items. This quantified experience is key to ensure the credibility of the finished PIRT, where analytic predictions clearly show the importance (or lack of) for each PIRT item over the entire domain of application.

More details on PIRTs for DBA events with potential fuel failure and radiological release are documented in later sections of this report. In particular, the following PIRT reports are referenced:

Phenomena Identification and Ranking Table Report for Natrium Other Quantified Events (Section 5.1, Table 4-3)

Phenomena Identification and Ranking Table Report for Natrium Partial Flow Blockage within a Subassembly Evaluation Model (Section 5.2)

Natrium Topical Report: Fuel and Control Assembly Qualification [6] (Table 4-4)

The PIRT for Natrium Other Quantified Events (OQEs) identifies important phenomena associated with ((

))(a)(4) the PIRT for partial flow blockage is described in TP-LIC-RPT-0008 Rev. 0, Partial Flow Blockage Methodology [7]; and Table 6-3 in NAT-2806 Rev. 0, Natrium Topical Report: Fuel and Control Assembly Qualification [6]

summarizes the high-important phenomena associated with fuel and absorber pin behavior.

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The PIRTs presented in this report take into account important process and phenomena rankings described in the PIRTs available for in-vessel DBAs without radiological release [3],

LBEs without fuel failure, and radiological source term events.

4.4 Evaluation Model Assessment TerraPower's Quality Assurance Program Description (QAPD) [10] and Software Management Procedure (SMP) detail the QA requirements and processes. The QAPD and SMP comply with the applicable requirements of ASME NQA-1-2015 [11], 10 CFR Part 50 Appendix B, and RG 1.28 [12]. TerraPower utilizes the graded approach by implementing the existing QA program controls for software that performs safety-related and/or non-safety-related applications.

The adequacy assessment of the EMs for DBAs with potential fuel failure and radiological release is guided by the TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk. For the codes to be accepted for safety-related applications, they should be assessed based on the list of legacy verification and validation (V&V) activities including verification test suite cases, legacy validations of severe accident modules, and benchmark activities. The assessment also identifies the verification, validation, and uncertainty quantification gaps that require closure. Some codes are still under further development ((

))(a)(4) and plans for the code maturation activities have been established.

The first step in the model assessment is to investigate the availability of legacy experimental data and evaluate the pedigree of the data. An Assessment Matrix is created for each methodology described herein that has an associated PIRT. The fuel failure phenomena identified in the PIRT as High and Medium importance are matched against the available experimental data. These fuel failure phenomena are generally considered to be applicable to all events with fuel failure, but in practice will likely be modeled for in-vessel events using

((

))(a)(4). Available experimental data is the historical data in the applied technology reports and journal papers. Based on the Assessment Matrix, testing needs will be identified.

Table 4-3 and Table 4-4 show the assessment matrices for the fuel failure phenomena. Note that the phenomena listed in Table 4-3 was developed for OQEs but is bounding and applicable for LBEs and DBAs with potential fuel failure and radiological release, and identifies important phenomena associated with ((

))(a)(4)

For in-vessel events with potential fuel failure, given the overlap of important processes and phenomena in DBAs with and without fuel failure, the assessment matrices discussed here mostly focus on fuel failure phenomena. More detailed PIRT and assessment matrix for in-vessel DBAs without fuel failure can be found in NAT-9390 [3].

A partial flow blockage assessment matrix is included in the Partial Flow Blockage Methodology report documented in TP-LIC-RPT-0008 [7]. It should be noted that partial flow blockage phenomena does not include any fuel failure phenomena since the methodology only covers up to the fuel failure point.

For the Fuel Misload Methodology, the code qualification, verification, and validation in NAT-2806 [6] and TP-LIC-RPT-0011 [5] are leveraged for core design, fuel performance, and thermal hydraulic codes. Fuel misloads do not introduce phenomena beyond those normally modeled in steady state analysis, thus, the code qualification, verification, and validation utilized for steady state core design, thermal hydraulics, and fuel performance are serving the function of an assessment matrix for the Fuel Misload Methodology.

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For the Fuel Handling Accident Methodology and Sodium Liquid and Gas Leak Methodology, assessment matrices were developed as part of the Source Term EM (NAT-9392 [4]).

However, it is noted that the phenomena associated with these assessments are related to the release and transport of radionuclides, and not to the dynamics and structural analysis that is the subject of the Fuel Handling Accident Methodology in this report, nor the calculation of the leak rate and timing that is the subject of the Sodium Liquid and Gas Leak Methodology in this report. Assessment matrices have not been developed for these parts of the methodology at this time but may be developed in the future as those methods are developed and matured.

Assessment matrices for Loss of Active Cooling and Excessive Sodium-Water reaction have not been developed at this time, but may be developed in the future as those methods are developed and matured.

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Table 4-3. Assessment Matrix for High/Medium Importance Fuel Failure Phenomena in OQEs.

1 ((

))(a)(4)

((

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Table 4-4. Assessment Matrix for High-Importance Fuel Failure Phenomena.

High-Importance Phenomena Applicable Design Limit Overview of Testing2F2 2 Note that some of these identified tests may be eliminated pending additional analysis or retrieval of additional historic data.

3 ((

))(a)(4)

((

))(a)(4)

((

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High-Importance Phenomena Applicable Design Limit Overview of Testing2F2 After the Assessment Matrix is created, data will be acquired from the identified resources. After the data is acquired, it will be qualified based on the TerraPowers existing data qualification procedure.

An initial assessment database has been constructed and is shown in Appendix B.

5 EVENT-SPECIFIC METHODOLOGY 5.1 In-vessel Transients with Radiological Release Methodology 5.

1.1 Purpose and Scope

This methodology ((

))(a)(4) to the analysis of DBAs that lead to clad and fuel failures, and which result in the release of radionuclides into the coolant. There is a wide variation of initiating events and event scenarios in the DBA event class ((

))(a)(4) Therefore, the boundary conditions from the ((

))(a)(4) event simulations performed following NAT-9390, "Design Basis Accident Methodology for In-Vessel Events without Radiological Release" [3] and as augmented by the In-Vessel DBAs and Non-DBA LBEs without Radiological Release Application Methods will remain applicable.

As illustrated in Figure 4.1-1, ((

))(a)(4) event simulation is performed to determine the extent of the challenge to fuel pin cladding integrity. The first step in this evaluation assesses the margin to the conservative TATNF screening criteria, which incorporates cladding wastage and thermal creep criteria in assessing potential failure.

The following filtering criteria are used to determine potential fuel failure. The most limiting channels ((

))(a)(4) are identified based on the following screening criteria:

((

))(a)(4)

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1) ((

))(a)(4)

If TATNF is not violated for a given transient, then no radiological release occurs.

However, TATNF incorporates a conservative approach to fuel performance modeling which bound a wide variety of potential temperature histories. ((

))(a)(4) passed to the DSAW described in Section 5.1.5.

DSAW provides a more mechanistic, event specific approach to performance analysis.

((

))(a)(4) Additionally, for slower transients, ((

))(a)(4) used by TATNF.

The present version of DSAW does not allow ((

))(a)(4) that integrates the severe accident modules as described in the following sections is planned to be employed.

5.1.2 Assumptions

Event and accident scenarios will be limited to DBAs with fuel failures and radiological release.

The plan is based on the current Natrium reactor system design and will be revised as appropriate as the reactor design and possible event scenarios mature.

The In-Vessel Transients with Radiological Release Methodology assumes that only Type 1 fuel is used.

5.1.3 EM Scope and Requirements Up to the onset of cladding and/or fuel failures the ((

))(a)(4)

DBA without release is identical to the method presented here. Therefore, the entire DBA modeling approach, PIRT phenomena and uncertainties established for the In-Vessel DBAs without Release [3] Methodology will be applied to this EM, subject to confirmation that models and uncertainties remain applicable for the range of conditions exhibited in the limiting DBA scenarios.

Three event phases are identified associated with severe accidents:

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Initiating: This stage defines phenomena for an event as it transitions from an accident to a severe accident. ((

))(a)(4)

Transition: ((

))(a)(4)

Termination: ((

))(a)(4)

As stated in Section 2.1, DBAs only credit SR SSCs to demonstrate compliance with the 10 CFR 50.34 dose limits. If a DBA does exceed the dose limits, then new SR SSCs are selected, and their required safety functions defined until the dose limits are met. Some DBAs will experience insufficient heat removal from the fuel, ((

))(a)(4) The range of phenomena associated with in-vessel transient DBAs would not transition to a severe accident as described for the Initiating phase.

While DBAs will not exhibit fuel failure phenomena (e.g., coolant boiling or fuel melting) associated with the most severe BDBE events or OQEs, the evaluation model described in this section is being developed to address the full scope of DBA, BDBE and OQEs. A PIRT has been established for the ULOF, ULOHS and Unprotected Transient Over-Power (UTOP) by internal and external panelists and documented in detail in the PIRT for Natrium OQE. While these three events do not necessarily consider the characteristics of all the possible BDBEs, they were considered adequate to identify phenomena that may be expected for the more frequent BDBEs and are applicable and bounding for in-vessel DBAs.

The Natrium OQE PIRT was supported by scoping calculations for the ULOF, ULOHS, and UTOP transients and included in Appendix A of the PIRT document. The calculations were performed for Beginning of Equilibrium Cycle core condition of each scenario and were extended to investigate the beyond BDBE consequences ((

))(a)(4)

In general, the results show that flow reduction or reactivity insertion without scram leads to a rapid core heat up, ((

))(a)(4) This is driven by negative reactivity feedback from Doppler, fuel axial expansion, core radial expansion, control rod driveline expansion, and coolant density. Then the negative reactivity results in a monotonic decrease in core power level.

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The sensitivity studies show that a highly conservative transient initiator leads to a substantial increase in core temperatures, resulting in coolant boiling, fuel melting, fuel relocation and ejection into the coolant channel with cladding breach. In the sensitivity evaluation ((

))(a)(4) would lead to the onset of local boiling and fuel/cladding failures.

Combined PIRT results for ULOF, ULOHS and UTOP events with High/Medium importance ranked phenomena are tabulated in Table 5-1. This reflects the highest importance with the associated lowest state of knowledge for each phenomenon. Note that Table 5-1 does not include the severe accident phenomena associated with Stage 2 of the ULOHS from the Natrium OQE PIRT as they are beyond the scope of this report.

Details on the rationale and rankings for the importance and knowledge level for individual phenomena and processes for each event analyzed are given in the PIRT report.

Table 5-1. Combined PIRT for ULOF, ULOHS and UTOP LBEs with Radiological Release with High/Medium Importance Phenomena.

No.

Phenomenon Importance Ranking State of Knowledge

))(a)(4)

((

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5.1.4 EM Description The In-Vessel Transient with Radiological Release EM is an extension to the EM established for In-Vessel DBAs without Release [3]. The extension includes the use ((

))(a)(4) severe accident modules to analyze fuel pin failures and subsequent relocation. As such, this EM incorporates all the qualification, verification and validation associated with Reference [3] ((

))(a)(4)

It should be noted that two independent fuel performance models ((

))(a)(4) have been matured for the purpose of predicting fuel pin damage and failure for use in safety analysis methodologies [6] and both codes will be verified and validated for use within the EM. These two models were developed independently, utilize different numerical methods, and differ in the approaches to modeling certain phenomena.

The ((

)) (a)(4) model is used by the DSAW to conservatively determine the peak-pin margin to failure (see Section 5.1.5). If the DSAW results demonstrate that assembly wide fuel failures are expected, the ((

))(a)(4) the following modules are required to address cladding and/or fuel failures that occur during the in-vessel transient simulation.

((

))(a)(4)

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((

))(a)(4) 5.1.5 In-Vessel Transient Evaluation Workflow Figure 5.1-1. DSAW Data Flow.

The DSAW for core transient analysis (Figure 5.1-1) ((

))(a)(4) Natrium safety analysis is provided in Appendix A9.1.2. ((

))(a)(4)

The following computer programs are used in the DSAW.

((

))(a)(4)

(a)(4)

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((

))(a)(4) 5.1.6 EM Assessment The adequacy assessment of the EM for DBAs with release, ((

))(a)(4) is guided by the TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk. ((

))(a)(4) provides a summary list of legacy V&V activities including verification test suite cases, legacy validations of severe accident modules, and benchmark activities. It also identifies the verification, validation, and uncertainty quantification gaps that require closure.

((

))(a)(4) models severe core disruption accidents with coolant boiling and fuel melting and relocation, is less developed than the remaining part which analyzes the thermal-hydraulic processes in other plant systems and components outside the reactor core. The fuel performance and failure analysis ((

))(a)(4) software quality assurance over its entire lifetime. The implementation of an SQA program for ((

))(a)(4) the fuel performance and fuel failure analysis part of the code ((

))(a)(4) Ongoing work is planned to be complete prior to TerraPower's submittal of an operating license application, and information on

((

))(a)(4) to fill the quality gap to complete the CGD at TerraPower will be included in a future licensing submittal.

((

))(a)(4) maturation activities at TerraPower has been established per TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk.

The code assessment including V&V is included in this plan. ((

))(a)(4) analyze several unprotected events (ULOF, ULOHS, UTOP, etc.), that potentially involve fuel failure, in support of the PIRT process for LBEs with release and OQEs. ((

))(a)(4) has been integrally validated in a study of the ULOF accident with cladding/fuel failure in a SFR using the CABRI integral effect test (IET) data [14]. Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

The DBAs that lead to potential fuel failures are associated with events that likely result in a significant heat up of the PHT with either symmetric or asymmetric boundary conditions at the inlet of the core. The current level of fidelity ((

))(a)(4) cannot resolve the multidimensional processes that take place in the large pool sections of the

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PHT system for asymmetric events or the dynamic impact of thermal stratification in the warm and hot pools. While the loss of two IHX clearly bounds the loss of a single IHX, DBA evaluations performed to date of the Natrium design show that the ((

))(a)(4) result in cladding failures. However, the current level of fidelity ((

))(a)(4) cannot resolve the multidimensional processes that take place in the large pool sections of the PHT system for asymmetric events or the dynamic impact of thermal stratification in the warm and hot pools A longer term development project to support the FSAR is the development and qualification of an Integrated Pool Methodology. This methodology will leverage ((

))(a)(4) to identify and address non-conservatisms ((

))(a)(4) has been endorsed by the NRC to address complex issues in Light Water Reactor (LWR) licensing when combined with appropriate experimental data [15]. ((

))(a)(4) Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

5.2 Partial Flow Blockage Methodology Partial blockage of the coolant flow in a fuel assembly has been considered as one of the important safety issues of SFRs. It is characterized by the tight spacing of fuel pins, high power density and high burnup fuel. Partial flow blockage may be initiated due to the accumulation of debris circulated in the primary sodium, failure of wire-wrapped spacers, and from swelling or bowing of the fuel pins. The partial flow blockage can cause the temperature rise in the wake region behind the blockage; therefore, it may lead to the potential for sodium boiling, dry out, cladding thermal failure and fuel melting.

Full discussion of this method and of the work that is ongoing in these areas is captured in TP-LIC-RPT-0008 Rev. 0, Partial Flow Blockage Methodology [7]. A summary discussion of the EM that has been developed is provided below to provide context within the scope of DBAs with radiological release.

5.

2.1 Purpose and Scope

The purpose of the partial flow blockage analysis is to demonstrate that the Natrium design satisfies the regulatory requirements of dose consequences for DBAs with Release Methodology with enough safety margins and meets CP and Operating License guidelines. This goal is achieved by confirming in the analyses that the system responses to DBAs with partial flow blockage within a fuel assembly satisfy all relevant acceptance criteria during the normal operating conditions.

The safety objective of the flow blockage analysis is to investigate the potential effects of partial flow blockage within a Natrium fuel assembly on fuel integrity based on the PCT.

The fuel integrity can be maintained if the cladding damage is avoided. Rods that exceed the PCT steady state acceptance criteria are treated as failed.

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The scope of this analysis is to provide bounding cladding temperatures for an infinitely thin, fully impermeable blockage within a Natrium assembly at steady state operating conditions.

5.2.2 Assumptions Assumptions are discussed in detail in the Partial Flow Blockage Methodology topical report [7]. ((

))(a)(4) 5.2.3 0BAcceptance Criteria PCT is used in partial flow blockage analysis as the acceptance criteria.

5.2.4 EM Scope and Requirements The scope and requirements for partial flow blockage EM are established via the PIRT process. Details of this PIRT process can be found in Partial Flow Blockage Methodology topical report [7].

5.2.5 EM Description The safety analysis of partial flow blockage is performed with respect to the frequency-based criteria. The EM provides a bounding temperature for an infinitely thin, fully impermeable blockage within a Natrium assembly at steady state operating conditions.

The blockage sizes are selected per a frequency-based criteria. This event is undetectable prior to fuel failure. This hypothetical planar blockage bounds the following credible events: collapsed wire wrap, rod bowing without contact and lodged foreign material. EM includes the upper bound for the maximum PCT, the number of fuel pins, and the potential associated radiological release. The thermal hydraulic analysis of partial flow blockage is performed using ((

))(a)(4) the semi-empirical model which is in the process of being validated against historical ORNL data for central 6 subchannel blockages and 14 subchannel edge blockages [16]. ((

))(a)(4)

Additional detail describing the EM is available in the Partial Flow Blockage Methodology topical report [7].

((

))(a)(4)

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((

))(a)(4) 5.2.6 EM Assessment A full discussion of this method is captured in the Partial Flow Blockage Methodology topical report [7]. As such, this report refers to the partial flow blockage report for the qualification, verification, and validation plans associated with the EM summarized here in Section 5.2. Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

5.3 Fuel Misload Methodology 5.

3.1 Purpose and Scope

During refueling outages, fuel assemblies are discharged or shuffled to new core locations and fresh fuel is loaded. The purpose of this methodology is to analyze the consequences of moving an assembly to the wrong core location or loading it in the right location but the wrong orientation.

The Natrium core has two main enrichment zones (inner and outer), with the outer zone being higher enrichment to flatten power. ((

))(a)(4) Figure 5.3-1 shows the core layout, and the orange and green arrows indicate the convergent shuffle direction of the fuel assemblies.

Figure 5.3-1. Equilibrium Core Fuel Layout.

(a)(4)

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5.3.2 Assumptions

The steady state tools used to design the reactor core have the fidelity to model the misloaded core ((

))(a)(4)

The final Natrium design ((

))(a)(4)

((

))(a)(4)

The assembly to be misloaded is ((

))(a)(4) 5.3.3 EM Scope and Requirements The fuel misload has one dominant highly ranked phenomenon, which is the change in the local power distribution. The core power distribution drives the core temperature distribution at steady-state conditions. The power distribution is a function of the core composition, burnup distribution, and geometry. The magnitude of the change to the local power distribution for the misload event depends on the change of the pin-level isotopic distribution between the intended and misloaded core configuration. ((

))(a)(4)

The processes and phenomena described above are modeled with the EM described in the following subsection.

5.3.4 EM Description The Fuel Misload Event is analyzed using ((

))(a)(4)

For PSAR evaluations, 10 cases were selected covering a sample of misloaded fresh, once burned, twice burned, and thrice burned assemblies at beginning-of-life and beginning of equilibrium core conditions. ((

))(a)(4)

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((

))(a)(4)

The methodology for determining the limiting assembly and fuel pin during the fuel misload transients has not been finalized and future work may change the position and number of assemblies involved in these cases. Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

5.3.5 EM Assessment Additional detail on the core design and thermal hydraulic codes used to predict the steady-state local power and temperature distributions is provided in the fuel qualification topical report [6] and in TP-LIC-RPT-0011 Rev. 0 Core Nuclear and Thermal Hydraulic Design Technical Report [5].

As such, this report refers to the code qualification, verification, and validation plans included in those reports. Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

5.4 Fuel Handling Accident Methodology 5.

4.1 Purpose and Scope

Fuel assemblies can be damaged in various fuel handling (FH) events during (i) insertion or removal from reactor core, (ii) in-vessel fuel assembly movement, (iii) ex-vessel fuel assembly movement between the EVST and washing station, or due to (iv) inadvertent action causing spent fuel assembly crush, (v) fuel assembly or loaded fuel cask drop in spent fuel pool.

These events need to be analyzed to determine the possible damage and final configurations of both the dropped and impacted fuel assemblies. In addition, the potential release of radionuclides resulted from such FH accidents as well as their leakage to the environment needs to be quantified.

In this section, only the EM for analysis of structural-mechanical behavior and failure of dropped/impacted fuel assembly is discussed. The transport and consequence of radiological release resulted from a DBA FH event are analyzed by the Radiological Source Term Methodology [4].

5.4.2 Assumptions

((

))(a)(4)

For a conservative scoping calculation of the potential radiological release happening during a fuel handling accident, ((

))(a)(4)

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Detailed analysis of a fuel drop accident considers limiting scenarios ((

))(a)(4) result in the worst possible fuel damage and the highest radiological release.

5.4.3 Acceptance criteria The fuel drop analysis used for the evaluation of LBEs and DBAs may be performed as part of analysis for other design purposes. As a result, the following acceptance criteria may be defined generally for fuel drop analysis:

Dropped fuel assemblies must not result in fuel cladding mechanical failure in either the dropped assembly or any targeted structures.

Dropped fuel assemblies must not create unacceptable core component conditions that would impact safe reactor operations (e.g., local criticality, loose parts within the components, reduced flow through the components, etc.).

The methodology established in this section, however, only covers the analysis of mechanical damage of the dropped assembly and not the consequence of potential radiological release resulted from such a mechanical failure. Analysis of the consequence of potential radiological release resulted from a FH accident involves other acceptance criteria pertaining to the source term analysis which is documented in [4]. As such, neither of the acceptance criteria listed above apply to this methodology, and there are not necessarily acceptance criteria for this EM. Rather, it is used to provide fuel failure information to the downstream source term analysis, and failure of the fuel is acceptable as long as the ultimate dose consequences acceptance criteria are met.

5.4.4 EM Scope and Requirements A fuel handling accident can be initiated by FH machine malfunction and/or operator errors. ((

))(a)(4) However, some mechanisms potentially causing mechanical damage to fuel assemblies during a fuel drop event can be identified as follows.

Stress, strain, and loading limits of the fuel assembly components including assembly duct, fuel rods, spacers, receptacle, etc.

Fatigue of structural components resulted from cyclic and dynamic load histories.

Elastic/inelastic behavior (deformation) of components under loading

Mechanical fracturing caused by dynamic loads or impact of the dropped assembly on other structure(s).

The EM to be used for analysis of fuel assembly mechanical failure during a FH accident should be able to model the above-mentioned processes and phenomena.

A detailed PIRT for radiological release and consequence resulted from FH accidents can be found in Table 2 of the Radiological Source Term Methodology [4].

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5.4.5 EM Description The dynamic structural behavior and integrity of fuel assemblies subjected to mechanical impact in a FH event can be analyzed in detail with the (nonlinear) Finite Element Analysis (FEA) software, ((

))(a)(4) The FEA software can provide predictions of the fuel assembly mechanical stress and impact force for different fuel drop scenarios defined by the ((

))(a)(4) etc., which can then be compared with the fuel assembly strength limits to determine the extent of the fuel damage (number of damaged fuel pins). This result can then be used to quantify the radiological release for the event. The transport inside reactor containment and the potential release to the environment of radionuclides is then analyzed with ((

))(a)(4) described in the Sodium and Gas Leak Methodology Section 5.5, and the Radiological Source Term Methodology [4].

The FEA software employs a finite element method, where the behavior of a solid or fluid system in two or three space dimensions is solved by subdividing a large computational domain into smaller, simpler parts called finite elements. This is achieved by a particular discretization in the space dimensions, which is implemented by the construction of a mesh or a computational grid of the analyzed domain.

A finite element model of a Natrium fuel assembly has been built and used for the preliminary analysis of core assembly drop accidents. The method requires inputs of geometric properties (such as fuel pins, assembly duct, receptacle, spacers, etc.) and material properties of the core assembly, the stiffness of the impacting receptacle or surface, the boundary conditions of the drop scenario, and some experimentally determined factors, such as the impact damping coefficient. The output of the method is impact load histories that may be used to perform stress analyses on core assemblies.

((

))(a)(4)

The EM used to analyze the transport and consequence of the FH DBA radiological release will be based on ((

))(a)(4) which are described in detail in Section 5.5 and the Radiological Source Term Methodology [4].

5.4.6 EM Assessment FEA software ((

))(a)(4) have a very broad range of applicability in different industries, such as aerospace, automotive, machinery, oil &

energy, etc., as they provide detailed insight and offer a unique tool for structural analysis.

However, there are still limitations to their applicability as a routine tool for safety justification of nuclear power plants.

Assessment of the FEA software selected for analysis of Natrium safety problems is guided ((

))(a)(4) Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

4 ((

))(a)(4)

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In performing the FEA for a fuel drop event, additional inaccuracies and uncertainties may arise due to:

(i)

Geometrical/structural complexity of the fuel assembly which contains hundreds of fuel pins and many other supporting structural components; and (ii)

Difficulty in defining the ((

))(a)(4)

The conservative definition of a fuel drop scenario can be used to obtain the maximum fuel damage in such an event.

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5.5 Sodium Liquid and Gas Leak Methodology 5.

5.1 Purpose and Scope

In addition to the transport of radionuclides (RNs) from fuel failures due to In-Vessel Transients, Fuel Handling events and Partial Flow Blockage through facility systems, transport of RNs released due to leaks were quantified as important phenomena relevant to DBAs with radiological material releases.

System leak scenarios include leaks associated with the Sodium Cover Gas System (SCG) ((

))(a)(4) that has leaked from the fuel; the SPS which would include ((

))(a)(4) and the IHT which would include ((

))(a)(4)

The purpose of this EM is to determine and quantify leaks for dose calculations. The analysis includes the extent of leaks and releases based on the event initiation - the location, timing, system conditions, and propagation.

For PSAR, the Radiological Source Term Methodology approach taken for Sodium and Gas Leaks is to determine or assume a maximized release (e.g. complete system release, or all available volume prior to pump trip). A detailed methodology for mechanistically determining the specific leak conditions (e.g. mass and energy release) has not yet been developed. ((

))(a)(4) 5.5.2 Assumptions

The system leakage scenarios are assumed during normal operation and not as part of, or consequence of, a different event.

((

))(a)(4) 5.5.3 EM Scope and Requirements The EM established for analysis of sodium and gas leak events should have the capability to model important processes and phenomena detailed in the PIRT study in the Source Term topical report [4].

Table 5-2 summarizes the medium and highly ranked phenomena where knowledge level is equal or lower than the importance ranking.

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Table 5-2. List of Medium and High Importance phenomena in SPS leak events [4].

No.

Phenomen on /

Process Description Importance Ranking Rationale for Importance Ranking Knowledge Level Rationale for Knowledge Level

((

))(a)(4)

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No.

Phenomen on /

Process Description Importance Ranking Rationale for Importance Ranking Knowledge Level Rationale for Knowledge Level

((

))(a)(4)

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5.5.4 EM Description As described in the Radiological Source Term Methodology [4] the EM is comprised of the analysis codes: ((

))(a)(4) and the output/input interfaces between each calculational device. Figure 5.5-1 provides the EM diagram for the SPS, SCG and IHT leak evaluations. The portions of the figure relevant to the DBA with release EM are the

((

))(a)(4) of the System Leak Rate, when needed.

Figure 5.5-1. Sodium Cleanup System and IHT Leak EM Diagram.

The SCG and the SPS source term will be based on the coolant inventory during normal operation. The steady-state inventory for the sodium cleanup system ((

))(a)(4) The modeling of the potential leakage from the steady state normal core or primary system activity ((

))(a)(4) The required information from upstream evaluations includes ((

))(a)(4)

The sodium leak into the atmosphere itself, confined in a building or outside, likely will be evaluated ((

))(a)(4) to account for any sodium reaction effects. A brief description of each of the analysis codes can be found in the Radiological Source Term Methodology [4].

5.5.5 EM Assessment As noted in the Radiological Source Term Methodology [4], the EM acceptance assessment is planned to be performed for the following activities:

Acceptance test plans for each individual code mentioned has formally been completed and effort will begin on the resolution of identified gaps ((

))(a)(4)

There are no known gaps ((

))(a)(4) relevant to its use in the Source Term EM.

(a)(4)

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Assessments of individual model fidelity, accuracy, and scaling for sources. As part of this task, it is anticipated that it will address the integrated calculations and consideration for data distortions.

As part of the EM process biases and uncertainties will be addressed for all LBEs with exception to the DBAs. The DBAs identified as part of the EM will use an approach that considers conservatisms.

The radiological source term EM activity will address prediction of FOMs through incorporation of biases and uncertainties into the various code mathematical models. The overall quantification of uncertainties will address each of the calculational devices as well as for the propagation of uncertainties through the series of codes used in the event evaluation.

As such, this report refers to the radiological source term report [4] for the qualification, verification, and validation plans associated with the EM discussed here in Section 5.5.

Ongoing work in this area is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

6

SUMMARY

6.1 Summary of Codes Selected A wide range of methods and EMs established for analysis of DBAs with potential fuel failure and radiological release in the Natrium plant has been summarized in this report. The diversity of the events and phenomena involved necessitates different analysis approaches and EMs, ranging from conservative estimation to first-principle modeling. Even with first-principle modeling, conservative assumptions are often necessary in regards to the initial conditions, boundary conditions, and some model parameters. Justification for the conservatism or accuracy of the EM prediction affected by the specification of the initial/boundary conditions as well as the choice of model parameters is to be provided in each EM application. Overall prediction uncertainty also needs to be considered and quantified.

The list of the software includes:

((

))(a)(4)

Most of the above codes are acquired by TerraPower via the TerraPower Safety Software Gap Analysis, CGD, and Maturation plan guided by TerraPower's QAPD [10] and SMP. Work is ongoing for the codes to be accepted for Natrium safety-related applications, with assessments planned based

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on a list of legacy V&V activities including verification test suite cases, legacy validations of severe accident modules, and benchmark activities.

Additionally, the assessment identifies the verification, validation, and uncertainty quantification gaps that require closure.

((

))(a)(4) has been acquired with following assessments:

Software Dedication Acceptance Test Plan

Software Dedication Technical Evaluation Report

Software Dedication Acceptance Test Report

Software Dedication Report Some codes are still under further development ((

))(a)(4) and plans for the code maturation activities have been established. Assessments of important closure models and integrated performance of the EMs are planned together with the acquisition of relevant experimental validation data delineated in Appendix B - Initial Experimental Database for Fuel Performance and Radiological Release/Transport Methodology.

To have confidence in an EMs predictions, it must undergo rigorous review - a process called software Verification, Validation and Uncertainty Quantification. The first step is verification which ensures that the model (differential) equations are correctly solved, and the numerical solutions are consistent with the analytical solutions. The next step is validation where accuracy of the EM is evaluated by comparing the predictions with data obtained from relevant Separate Effects Tests and Integral Effects Tests. Finally, all analyses require an estimate of error and uncertainty in the prediction for an application. All the verification, validation, and uncertainty quantification activities are application dependent. The V&V of the EMs are included in its safety software assessment plan guided by the TerraPowers Acquired Software Quality Assurance Plan under Safety Analysis and Risk. Uncertainty quantification for the Natrium safety analyses is addressed in a Safety Uncertainty Quantification and Margin Assessment Methodology Ongoing work in these areas is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

6.2 EM Conservatism Summary In application of the methodologies discussed in this report for analysis of DBAs with potential fuel failure and radiological release, conservative assumptions about initial/boundary conditions, modeling parameters, and failure/acceptance criteria affect the outcome of calculations and predictions. Similar to the DBA without release calculations, conservative assessments of DBAs with fuel failure and release employ the following conservatisms in analysis of the in-vessel transient and partial flow blockage events [3]:

Conservatisms in the form of direct biases are applied via input to nuclear data and model uncertainties, thermal-hydraulic models, and control system performance parameters for the representative events, and those which are applicable to the RAC performance. Selection of boundary conditions, isolation times, and other assumptions needs to ensure that the analysis is appropriately biased. The ((

))(a)(4) steady-state analysis of heat-up events, for instance, uses the following set of biases in addition to the selected DBA biasing configuration:

o

((

))(a)(4)

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o

((

))(a)(4)

The plant initial and boundary conditions are conservatively selected within the operating band.

The hot-pin PCT within the sub-assembly is conservatively assessed ((

))(a)(4)

The TATNF screening criteria include conservatisms and margins that provide reasonable allowance that fuel pin failure will not occur when they are not violated.

When the TATNF screening criteria are violated, the subsequent DSAW fuel performance analysis also contains conservatisms in evaluating the fuel failure.

The fuel misload analysis is adjusted for uncertainties in the final temperature distribution ((

))(a)(4) to determine the potential for fuel failures.

((

))(a)(4)

An effort is underway to demonstrate that the conservative approach described above is sufficiently conservative for the Natrium design. Ongoing work is planned to be complete prior to TerraPower's submittal of an operating license application, and that information will be included in a future licensing submittal.

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7 CONCLUSIONS AND LIMITATIONS 7.1 Conclusions TerraPower is requesting NRC approval of the EM methodology plans documented in this report for use by future applicants utilizing the Natrium design as an appropriate and adequate means to evaluate DBAs with the potential for radiological release (as described in Section 2.4). This approval is subject to the limitations described below.

7.2 Limitations All methodologies considered in this report share a set of similar limitations:

1.

The methodology is limited to a Natrium design that has a pool-type, SFR design with metal fuel and sodium bond as described in Sections 1.3 and 2.3. Changes from these design features will be identified and justified in Safety Analysis Reports of Natrium license applications.

2.

Adequate verification and validation assessment information should be made available to the NRC staff as part of future submittals supporting the codes that make up the EM. This verification and validation information should be justified to reasonably bound the operational envelope for the design for any applicant referencing the EM methodology.

3.

An applicant utilizing the topical report needs to justify the use of the model for the design. This justification must discuss the capability of the model in the context of what is needed to appropriately represent the design and discuss how the model is applicable to the design, consideration of system interactions, and system conditions (which may affect the applicability of models or validation data).

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8 REFERENCES

[1]

NEI 18-04, Rev. 1, "Risk-Informed Performance-Based Technology Inclusive Guidance for Non-Light Water Reactor Licensing Basis Development," Nuclear Energy Institute, 2019.

[2]

NEI 21-07, Rev. 1, "Technology Inclusive Guidance for Non-Light Water Reactor Safety Analysis Report: For Applicants Utilizing NEI 18-04 Methodology," Nuclear Energy Institute, 2022.

[3]

TP-LIC-RPT-0004 Rev. 0, "Design Basis Accident Methodology for In-Vessel Events without Radiological Release," TerraPower, 2023.

[4]

TP-LIC-RPT-0003 Rev. 1, "Radiological Source Term Methodology Report," TerraPower, 2024.

[5]

TP-LIC-RPT-0011 Rev. 0, "Core Design and Thermal Hydraulic Technical Report,"

TerraPower, 2024.

[6]

NAT-2806 Rev. 0, "Natrium Topical Report: Fuel and Control Assembly Qualification,"

TerraPower, 2023.

[7]

TP-LIC-RPT-0008 Rev. 0, "Partial Flow Blockage Methodology," TerraPower, 2024.

[8]

TP-LIC-RPT-0005 Rev. 0, "Radiological Release Consequences Methodology Topical Report," TerraPower & GE-Hitachi Technology, 2023.

[9]

SAND2011-4145, "Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety," SNL, 2011.

[10]

TP-QA-PD-0001 Rev. 14A, "TerraPower QA Program Description," TerraPower, 2023.

[11]

ASME NQA-1-2015, "Quality Assurance Requirements for Nuclear Facility Applications," The American Society of Mechanical Engineers (ASME), 2015.

[12]

RG 1.28, "QUALITY ASSURANCE PROGRAM CRITERIA (DESIGN AND CONSTRUCTION)," US NRC, 2017.

[13]

((

[14]

))(a)(4)

[15]

RG 1.82, Rev. 3, "WATER SOURCES FOR LONG-TERM RECIRCULATION COOLING FOLLOWING A LOSS-OF-COOLANT ACCIDENT," US NRC, 2003.

[16]

ORNL-TM-4324, "Effect of Partial Blockages in Simulated LMFBR Fuel Assemblies," ORNL, 1973.

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9 APPENDICES 9.1 Appendix A - Additional Details of the DSAW Process 9.1.1

((

))(a)(4)

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((

))(a)(4)

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((

))(a)(4)

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((

))(a)(4)

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9.2 Appendix B - Initial Experimental Database for Fuel Performance and Radiological Release/Transport Methodology Experimental data listed below can be used to assess important closure models and integrated performance of the EMs developed to analyze fuel performance and in-vessel DBA events potentially involving fuel failure and/or radiological release. The lists are only preliminary and are retained here for historical purposes as they were used to inform the initial PIRT development and the subsequent experimental database development.

Table B-1. List of Experimental Data Related to Radionuclide Migration during Pre-transient Phase

((

))(a)(4)

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((

))(a)(4)

Not Confidential Controlled Document - Verify Current Revision

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Table B-2. List of Experimental Data Related to Radionuclide Release during a Cladding Rupture

))(a)(4)

((

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((

))(a)(4)

Not Confidential Controlled Document - Verify Current Revision

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((

))(a)(4)

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((

))(a)(4)

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((

))(a)(4)

Not Confidential Controlled Document - Verify Current Revision

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END OF DOCUMENT

ML25251A090

Text

Subject:

References:

Enclosures:

SUBJECT:

Dear George Wilson:

Enclosure:

1.0 INTRODUCTION

2.0 BACKGROUND

Title:

SUMMARY

SUMMARY

SUMMARY

4.1 Background

1.1 Purpose and Scope

2.1 Purpose and Scope

3.1 Purpose and Scope

4.1 Purpose and Scope

5.1 Purpose and Scope

SUMMARY

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