ML25274A130
| ML25274A130 | |
| Person / Time | |
|---|---|
| Site: | Kemmerer File:TerraPower icon.png |
| Issue date: | 10/01/2025 |
| From: | George Wilson TerraPower |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| TP-LIC-LET-0460 | |
| Download: ML25274A130 (1) | |
Text
15800 Northrup Way, Bellevue, WA 98008 www.TerraPower.com P. +1 (425) 324-2888 F. +1 (425) 324-2889 October 1, 2025 TP-LIC-LET-0460 Docket Number 50-613 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Document Control Desk
Subject:
Submittal of Supplemental Information to Support Kemmerer Unit 1 Construction Permit Application Review
References:
- 1. TerraPower, LLC, Submittal of the Construction Permit Application for the Natrium Reactor Plant, Kemmerer Power Station Unit 1, March 28, 2024 (ML24088A059).
This letter transmits information to supplement the construction permit application for Kemmerer Unit 1, provided via Reference 1. Specifically, this letter transmits TerraPower, LLC (TerraPower) Report, Natrium1 Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development, NAT-13478, Revision 0 (enclosed). The report contains an overview and description of the safety-significant long-lived, passive components materials of construction and methods applied to provide assurance of adequate material performance for the component lifetime.
This letter and the associated enclosure make no new or revised regulatory commitments.
If you have any questions regarding this submittal, please contact Ian Gifford at igifford@terrapower.com.
1 TerraPower & GE Vernova Hitachi Nuclear Energy Technology.
Date: October 1, 2025 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and accurate. Executed on October 1, 2025.
Sincerely, George Wilson Senior Vice President, Regulatory Affairs TerraPower, LLC
Enclosure:
NAT-13478, Revision 0, "Natrium Demonstration Plant Long-Lived Passive cc:
Mallecia Sutton, NRC Josh Borromeo, NRC Nathan Howard, DOE Structural Materials of Construction Selection and Development
ENCLOSURE NAT-13478, Revision 0, Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development
SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Verify Current Revision Document
Title:
Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Natrium Document No.:
NAT-13478 Rev. No.:
0 Page:
1 of 68 Doc Type:
RPRT Target Quality Level:
N/A Alternate Document No.:
NAT-13478-NP Alt. Rev.:
N/A Originating Organization:
TerraPower, LLC Quality Level:
N/A Natrium MSL ID:
DAP Status:
Released Open Items?
N/A Approval Approval signatures are captured and maintained electronically; see Electronic Approval Records in EDMS.
Signatures or Facsimile of Electronic Approval Record attached to document.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 2 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED REVISION HISTORY Revision No.
Affected Section(s)
Description of Change(s) 0 Initial issue
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 3 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED TABLE OF CONTENTS 1
INTRODUCTION.................................................................................................................................. 5 2
SSC CLASSIFICATIONS..................................................................................................................... 7 3
CODES AND STANDARDS................................................................................................................. 9 4
ENVIRONMENTAL CONDITIONS..................................................................................................... 10 5
MATERIALS SELECTIONS............................................................................................................... 17 6
RELIABILITY AND INTEGRITY MANAGEMENT PROGRAM............................................................ 19 7
DEGRADATION MECHANISM IDENTIFICATION AND SCREENING CRITERA DEVELOPMENT... 30 8
MATERIALS DATA QUALIFICATION................................................................................................ 32 9
MATERIALS TESTING....................................................................................................................... 36 10 MONITORING AND NON-DESTRUCTIVE EXAMINATION PRACTICES.......................................... 40 11 TRIBOLOGICAL COATINGS............................................................................................................. 47 12 AREAS REQUIRING FURTHER RESEARCH AND DEVELOPMENT............................................... 50 13 REFERENCES................................................................................................................................... 53 14 APPENDICES.................................................................................................................................... 55 Appendix 14.1 List of Acronyms......................................................................................................... 55 Appendix 14.2 NSRST SSC Codes Justifications and Enhancements............................................... 58
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 4 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED LIST OF TABLES Table 4-1: Pressure and Temperature Conditions for Major Primary System Components...................... 12 Table 4-2: Intermediate Systems Major Component Operating Temperatures......................................... 14 Table 4-3: Primary System Components Environmental Conditions......................................................... 15 Table 4-4: Intermediate System Components Environmental Conditions................................................. 16 Table 4-5: Sodium Chemistry Impurity Targets........................................................................................ 17 Table 5-1: Known Materials List............................................................................................................... 18 Table 5-2: Preliminary Materials of Construction for Environmental Evaluation........................................ 19 Table 6-1: Design Margin and Component Function Attributes................................................................ 26 Table 6-2: Consequence Attributes.......................................................................................................... 26 Table 7-1: Screening Criteria for Degradation Mechanism Applicability.................................................... 31 Table 9-1: ORNL/HFIR Neutron Irradiation Materials Matrix..................................................................... 37 Table 9-2: ANL Exposure and Testing Matrix........................................................................................... 39 Table 9-3: SRC Study Matrix.................................................................................................................... 40 Table 10-1: Primary Coolant Boundary Leakage Monitoring Summary.................................................... 43 Table 11-1: Coating Areas for Core Components..................................................................................... 48 Table 11-2: Core Components and Coatings Being Considered for Natrium............................................ 49 Table 11-3: Environments for Coatings.................................................................................................... 50 LIST OF FIGURES Figure 1-1: Materials Selection Process Flowchart..................................................................................... 6 Figure 6-1. Reliability and Integrity Management Process Diagram......................................................... 23 Figure 6-2: Degradation Mechanism Risk Assessment............................................................................ 27 Figure 6-3: Risk of Degradation and Target Reliability............................................................................. 28 Figure 6-4: CVAP Risk of Failure Methodology........................................................................................ 29 Figure 9-1: The SMT-3 Corrosion Loop at Argonne National Lab............................................................. 38
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 5 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED 1
INTRODUCTION The purpose of this document is to summarize the overall approach used for the selection and development of materials of construction for long-lived, passive components of safety significant fluid systems components in the Natrium 1 Demonstration Plant (Natrium) design. These long-lived components are those that are not expected to be replaced during the plant design lifetime and perform their intended safety functions in a passive manner. The structural materials of construction are those that support the component safety function(s) by providing structural and pressure retaining integrity and load carrying capability, as applicable.
This report does not address building structures materials of construction, non-pressure retaining components such as seals and gaskets, or short-lived components or materials (such as reactor core assemblies or non-metallic materials) that are intended to be replaced during the plant lifetime or qualified through the Environmental Qualification Program.
This report provides selected results of materials of construction selection and development activities that are current as of the approval date for the report. The design and development of the Natrium reactor plant is ongoing at this preliminary design stage and the information presented in the report is subject to change after the approval date. Although the methods and approaches described in the report are expected to be consistently applied as the design progresses, the results of application of the methods may change and will be reflected in plant design documentation, as appropriate.
This report also provides a description of the integration of the American Society of Mechanical Engineers (ASME)Section XI, Division 2 [1] Reliability and Integrity Management (RIM) Program into the pre-construction material selection and design process, as well as component lifecycle management during the post-construction and operating phases, to ensure adequate reliability and functional capability of safety significant components.
The materials of construction selection and development process applied for the design of the Natrium reactor plant is summarized in Figure 1-1 Materials Selection Process Flowchart. The process is described in more detail in the following sections of this report.
1 TerraPower & GE Vernova Hitachi Nuclear Energy Technology.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 6 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED SSC Functional Requirements and Safety Classifications
- Risk inputs Applicable Codes and Standards
- Allowable materials
- Regulatory requirements and guidance Service Environments
- temperature
- chemistry
- neutron fluence SSC Design and Optimization
- Code design requirements
- Material properties
- Fabrication considerations
- Size, deadweight, seismic Initial Material Selection
- Code allowable
- Environmental compatibility
- Design Lifetime
- Operating Experience RIM Program
- Scope determination
- Degradation mechanisms identification and screening Testing and Qualification
- Service conditions
- Data qualification
- Operating experience
- Accelerated aging Final Material
- Design Specification
- Procurement
- Construction RIM Program
- MANDE
- RIM Strategies RIMEP Construction and Operation Inspections and Testing
- PITAP
- Inservice Inspections
- Material Surveillance
- Aging management Figure 1-1: Materials Selection Process Flowchart
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 7 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED 2
SSC CLASSIFICATIONS The Natrium safety classification of structures, systems, and components (SSCs) follows the risk-informed, performance-based (RIPB) framework (Nuclear Energy Institute (NEI) 18-04 [2])) endorsed through Regulatory Guide (RG) 1.233 [3], and includes the selection of Licensing Basis Events, classification of SSCs and selection of special treatments, and the determination of defense-in-depth adequacy. While the safety classifications are determined based on safety functions, similar to light-water reactors (LWRs) per previous precedence, the process under NEI 18-04 provides key differences in how those safety functions are derived (i.e., risk-informed vs deterministic) and how they can be associated with the selected SSCs. Both differences have contributed to a need to reevaluate the relationship between safety classifications and code and standard selection:
Risk-informed safety functions specific to a non-LWR reactor technology highlighted key differences in the phenomena, such as the reduction in operating pressures along with elevated operating temperatures for fluid retaining systems, that establish the design basis requirements being relied upon for the assurance of the safety function. As such, a review of the fundamental applicability of codes and standards for safety-significant SSC design and construction was performed in a comprehensive manner.
The other major difference between the classification of SSCs under NEI 18-04 for non-LWR technologies and traditional LWR approaches is the ability to associate risk-informed safety functions with specific SSCs. The nature of LWR technology, specifically highly pressurized fluid containing systems and dedicated, single mode, safety systems for coolant flow and replenishment, biased classifications to be more holistic for such systems. The change in reactor architecture plus the introduction of the fundamental concept of Non-Safety-Related with Special Treatment (NSRST) through the RIPB framework created the opportunity for diversity in safety function and classification within a single system or even in an individual component or structure.
Plant-level safety functions are established and evaluated for risk and safety significance to determine their classification per the NEI 18-04 guidance as endorsed by RG 1.233. Once that determination is made, the association between those safety functions and the specific components and structures that perform those functions are determined. This process ensures alignment between the current design and safety analysis and establishes the basis for SSC safety classifications based on the functions safety classification. Given the functional nature of this process, an individual component part or structural element could be assigned safety functions of varying classification levels, which are retained as discrete contributions to the classification basis even when a bounding safety classification is assigned to the overall structure, system, or component. Another instance of classification diversity driven by safety function is when a system has various operating modes that are relied upon discreetly for their contribution to the safety basis resulting in different subsets of the equipment or components being directly credited for these different operating modes. For example, active intermediate system flow using the Intermediate Sodium Pumps is a non-safety-related with no special treatment (NST) function while intermediate system natural circulation is a NSRST function.
Integrated into the SSC classification process is the selection of special treatments, which follows the same association between safety functions and specific components or structures. This discrete association of safety functions can be used to justify varying levels of adequate special treatments for
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 8 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED an individual component part or structural element. Per the Natrium process, special treatments can include codes and standards used for design and construction, programmatic treatments used to establish quality assurance requirements, design controls, inspections and maintenance (and may be inherent in the application of the code or standard), or SSC-specific treatments (e.g., external hazard protections which are dependent on location of the SSC within the plant). Specific to the purpose of this document, the safety classification of each SSC was used when considering the adequate level of quality assurance required and the codes and standards to which the SSC is designed, fabricated, erected, maintained, and tested.
Through a systematic process, each SSC is evaluated for all assigned safety functions to determine the set of special treatments applicable and necessary for the assurance of that specific function.
While the contribution of each special treatment is scrutinized individually, it is the collection of all special treatments, in addition to standard industrial practices applied, that are considered for the question of adequacy. It is this holistic view in which the codes and quality standards to which the SSCs will be designed, fabricated, erected, maintained, and tested, are selected. These special treatments are reviewed and endorsed as adequate by the Integrated Decision-Making Process Panel based on the current maturity of the design and supporting analysis.
Given the definition of special treatment per RG 1.201 [4], an initial benchmark is necessary as the concept of special treatment is framed as those measures applied beyond normal industrial practices. Normal industrial practices are interpreted as non-nuclear codes and standards that are generally applicable to a large spectrum of industries (e.g., International Building Code). While not considered a special treatment, adherence to these commercial codes and standards still has the same intention of providing assurance of the SSCs ability to perform its required functions. Given this concept, the general default approaches to special treatment selection for SR and NSRST SSCs are followed (which are in alignment with the guidance from Table 4-1 of NEI 18-04).
Nuclear codes and standards are applied for design (e.g., ASME Boiler & Pressure Vessel Code (BPVC)Section III, ASME AG-1, AISC N690, ACI 349) and programmatic treatments (e.g., Quality Assurance Program, Equipment Qualification, Technical Specifications) are applied for SR SSCs based on precedence from existing deterministic regulatory framework but still require scrutiny for applicability, necessity, and adequacy.
Established, commercial-grade, non-nuclear codes and standards are applied for design of NSRST SSCs, and supplemented as necessary with design augmentations as special treatments and other programmatic treatments (e.g., Augmented QA, D-RAP, M-Rule) to meet reliability and performance capability requirements.
In practice, the combination of functional-based classifications and special treatment selection results in systems with more diversity in code and standard selection than typical for a non RIPB-based design. This is most prominent in systems with both SR and NSRST portions, given the default use of nuclear and commercial codes, respectively. While this process did inform code jurisdictions and material selection for individual systems, conformance to the fundamentals of those codes and standards was not altered. Due to the discrete nature of the overall process, no generically applied (i.e., plant-wide) requirements for selection of codes and standards or materials were associated with SSC classification.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 9 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED The process and results described above form a portion of the basis for the information submitted in the Construction Permit Application. In alignment with the general maturity of the design and supporting analysis submitted for review, these selections for special treatments are focused on demonstrating adequate plant functional architecture. Further details pertaining to specific quantitative performance criteria are used to iterate on these decisions to confirm these conclusions of applicability, necessity, and adequacy. However, the practical use of quantitative performance criteria differs depending on the type of safety function. The discussion of the RIM Program in Section 6 provides an explanation for one major method of achieving that aim.
3 CODES AND STANDARDS The Natrium plant design specification establishes controlled plant-level design information and criteria and captures the regulatory requirements, consensus codes, and industry standards that define and provide inputs to the design basis and plant-level requirements for use by the plant designers. Applicable Regulations, Codes, and Standards were evaluated to determine applicability to the Natrium Sodium Fast Reactor (SFR) technology. A Nuclear Code of Record designation was applied to identify codes and standards that were selected by plant designers for use on SR and NSRST SSCs. Standards that are not applicable to the Natrium nuclear design or safety case are not designated as Nuclear Code of Record.
Safety-significant SSCs are designed, fabricated, erected, and tested to quality standards commensurate with the safety significance of the function(s) to be performed consistent with the safety classification determined as described in Section 2. Where generally recognized codes and standards are used, they are identified and evaluated to determine their applicability, adequacy, and sufficiency and may be supplemented or modified as necessary to ensure a quality product in keeping with the required safety-significant function(s) to meet the requirements of Natrium PDC 1.
Codes and standards used for safety-significant SSCs, along with the application of supplemental requirements as necessary, are selected during the design process from the list of nuclear code of record standards. In the case of safety-significant mechanical fluid systems, the selection is performed by the responsible system designers based on nuclear plant design precedence, regulatory requirements and guidance, and the designers engineering experience. The selection of applicable codes and standards is based on the safety function and risk significance established for the SSC as described in Section 2.
Nuclear Codes are selected for safety related SSCs where the Code has construction jurisdiction over the SR functional requirements. For example, ASME BPVC Section III is selected for the construction of components that comprise the primary coolant boundary and components that provide support for the reactor core. In cases where the SR functional requirements are not within the construction jurisdiction of nuclear Codes, specifically primary coolant boundary and associated component supports, and reactor core supports, for ASME BPVC Section III application, appropriate industry standards and SSC-specific design requirements are selected that provide assurance that the safety function is met with sufficient quality in the design and construction of the SSC. For example, non-structural SR functional requirements, such as heat transfer properties, are not within the jurisdiction
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 10 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED of the nuclear Codes, and would not require selection of nuclear Codes for design and construction of the SSC to assure sufficient quality consistent with the SSC safety-significant function.
The SR metallic components and supports that comprise the primary coolant boundary and reactor core support structures are constructed in accordance with ASME BPVC Section III, Division 5, High Temperature Reactors [5] as endorsed by RG 1.87, Revision 2 [6], including its stated exceptions and limitations.
Commercial codes and standards are selected for NSRST SSCs based on nuclear industry practice and applicable regulatory guidance. The applicability of the selected code and/or standard is evaluated to provide assurance of adequate quality commensurate with the safety-significant function(s) to be performed. Supplemental requirements, in addition to the selected commercial code or standard, are applied as special treatments where identified as necessary to support safety-significant functional requirements and to provide enhanced SSC quality.
Materials of construction (base metal, weld metal, fastener materials, etc.) for safety-significant components of mechanical fluid systems are selected from the allowable materials dictated by the requirements of the selected construction Code and any applicable limitations or exceptions from endorsement regulatory guidance. In the case of ASME BPVC and associated endorsement guidance, the materials requirements and properties are prescriptive. In cases where applicable codes and standards do not include prescriptive material requirements, materials are selected consistent with nuclear industry practice applicable to the SSC design and service environment.
Material selections are refined, or down-selected, based on the consideration of specific design and fabrication requirements and service environments, as further described in the remainder of this report.
The specific codes and standards selected, evaluated, and applied for the construction of safety-significant mechanical fluid system components are provided in Natrium project design documentation and the Kemmerer Unit 1 PSAR. Appendix 13.2 provides the selected commercial construction codes, along with enhancements to the codes applied as special treatments, for major NSRST mechanical components (excluding fuel handling equipment).
4 ENVIRONMENTAL CONDITIONS Service Environment The Natrium reactor plant is a pool-type SFR and operates at low pressures. Certain components, or portions of components, operate at elevated temperatures as defined by ASME BPVC Section III, Division 5 High Temperature Reactors, which defines elevated temperature service as greater than 700°F for carbon and low-alloy steel and greater than 800°F for austenitic steel and nickel alloys.
Although this definition of elevated temperature is not universal among mechanical construction codes, these temperatures are an appropriate and conservative boundary to define the potential onset of creep-damage which may require more restrictive construction rules, such as those in ASME BPVC Section III, Division 5, or some additional measures or considerations when other codes are used. In the Natrium design, components designed for structural or pressure-retaining service in a sodium
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 11 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED environment are constructed from austenitic stainless steel material. Therefore, 800°F is generally considered as the elevated temperature threshold for the onset of creep effects.
The Natrium reactor operates under conditions where certain primary and intermediate components experience service temperatures within the creep range for a significant portion of their design lifetime. For these components, the time duration and specific temperature exposure varies. In addition, for large components and piping systems, only limited portions may be exposed to service temperatures in the creep range. The majority of components in the Natrium fluid systems are not exposed to service temperatures in the creep range.
It is recognized that the ASME BVPC Section III, Division 5 does not currently support service time in the creep range greater than 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for 304H and 316H materials subject to service temperatures greater than 800°F. It is anticipated that the ASME Code will be updated to support extended operation of these materials in the high temperature environment, which will support the Natrium design lifetime of 60 years. In the interim, material properties have been established based on 500,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> service in an elevated temperature environment to support design, procurement and construction of Natrium components that are designed in accordance with ASME BPVC Section III, Division 5, ASME BPVC Section VIII [7], or ASME B31 [8]; constructed of 304H or 316H stainless steels; and subject to service temperatures greater than 800°F, such as portions of the reactor vessel, reactor internal structures, and the IHT system.
Primary System Components Operating Pressures and Temperatures Normal maximum operating temperatures, based on the preliminary plant design, occur at the reactor core outlet and are in the range of 1000°F, which bounds the normal operating condition for components exposed to elevated temperatures in the creep range. Many components within the Reactor Enclosure System (RES), but removed from the vicinity of the core outlet and the hot sodium pool, operate below the 800°F creep range threshold (e.g., reactor head and lower portions of the reactor vessel shell). For limiting transient conditions, core outlet temperature in the range of 1200°F has been conservatively calculated in safety analyses. These limiting transients are short-lived and do not require explicit time-based creep-related considerations for design and material selection. ASME BPVC Section III, Division 5 provides allowable stress values considering creep-fatigue for temperatures as high as 1500°F, therefore the overall Natrium temperature operating environment is most appropriately characterized as low-to-moderate creep range operation, considering normal operations and even the most severe plant events. Table 4-1 provides preliminary operating pressure and service temperature environment information for selected primary system components. For cases where normal and/or maximum pressures or temperatures have not been determined, the design value is indicated.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 12 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 4-1: Pressure and Temperature Conditions for Major Primary System Components SSC Operating Pressure (nominal)
Maximum Normal Temperature Maximum Transient Temperature Notes RES Guard Vessel
< 1 psig (16 psig Design
[internal], 3.7 psig Design
[external])
650°F 850°F**
Potentially in creep range during RAC-only Cooling Event (<150 hrs)
RES Reactor Head*
< 1 psig (16 psig Design)
<800°F 1000°F Only in creep range during RAC-only Cooling Event (<150 hrs)
-Limiting location is at connection to the Reactor Vessel (RV) shell RES Reactor Vessel (above top of vessel liner in cover gas space)
< 1 psig (internal) 41 psig (Design) 925°F 1075°F These temperatures include conservative margins to reflect uncertainty in cover gas space temperature predictions RES Reactor Vessel (below top of vessel liner adjacent to hot pool)
< 1 psig 41 psig (Design) 825°F 1000°F Portions of the vessel shell adjacent to the warm pool, and portions exposed to cold pool sodium, are
<800°F maximum normal temperature RES Core Support Structures NA 1015°F 1180°F PHT Intermediate Heat Exchanger 4 psig (shell) 110 psig (tube) 1000°F (shell) 960°F (tube) 1180°F Primary SPS Piping, Valves 35 psig 734°F 800°F (Design)
Primary boundary portion of SPS Primary SPS NSRST Components 0 - 35 psig (150 psig Design) 248°F - 734°F 800°F (Design)
H2 Meter Module - up to 932°F (1200°F Design); MPS Module up to 1200°F (Design 1400°F)
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 13 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED SSC Operating Pressure (nominal)
Maximum Normal Temperature Maximum Transient Temperature Notes Primary SCG RV Supply and Exhaust Piping, Valves
< 1 psig (16 psig Design) 220°F (supply), 950°F (exhaust) 1200°F (Design)
- Reactor head temperatures are influenced by thermal shield performance, which requires specialized sub-modelling and is in development.
- Maximum transient temperature is a preliminary estimate and subject to change.
Intermediate System Components Operating Pressures and Temperatures The intermediate cooling systems, including the Intermediate Heat Transport System (IHT) and the connected Intermediate Air Cooling System (IAC), are the normal reactor heat removal systems.
These systems interface with the reactor systems through the intermediate heat exchanger (IHX),
which transfers heat from the primary coolant to the intermediate coolant in the intermediate loops.
During normal operating modes, heat is transferred from the intermediate sodium to the Nuclear Island Salt System (NSS) via the sodium-salt heat exchanger (SHX). During start-up and shutdown operating modes, reactor decay heat is rejected from the intermediate sodium to the atmosphere via the IAC air heat exchangers (AHXs). The IHT hot leg piping, between the IHX and the SHX (normal operation) or the AHX (shutdown and start-up operation when heat removal through SHX is not available), is exposed to elevated service temperatures in the creep range. The remaining portions of the IHT are exposed to service temperatures below the creep range. Table 4-2 provides preliminary operating pressure and service temperature environment information for intermediate system components. For cases where normal and/or maximum pressures or temperatures have not been determined, the design value is indicated.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 14 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 4-2: Intermediate Systems Major Component Operating Temperatures SSC Operating Pressure /
Pressure Range (nominal)
Maximum Normal Temperature Maximum Transient Temperature Notes IHT Hot Leg Piping 100 psig 975°F 1025°F SHX 80 psig (sodium) 80 psig (salt) 975°F 1025°F AHX 50 psig 685°F 1025°F
<100 hours in creep range IHT Cold Leg Piping (SHX -> AHX only) 50 psig 685°F 1025°F
<100 hours in creep range IHT Cold Leg Piping (downstream of AHX) 35 psig
< 685°F
< 685°F IHT Sodium Pumps 25 psig (inlet) / 120 psig (outlet)
< 685°F
< 685°F Sodium and Irradiation Environmental Conditions Aside from temperature, the environmental conditions which impact material degradation are the chemical environment and the exposure to neutron irradiation. Table 4-3 and Table 4-4 provide a summary of these environmental conditions for primary and intermediate systems components and other interfacing safety significant components. As indicated, many components are maintained in an inert gas environment (which is generally benign from a material degradation standpoint) and not exposed to the sodium environment; although trace amounts of sodium vapor may be present in the primary cover gas environment. Irradiation levels are generally low (below thresholds for irradiation damage), with the exception of some core support and reactor internals components. The RIM Program component Degradation Mechanism Assessment (DMA) process (Section 7) applies screening criteria for irradiation damage and RIM strategies will be developed for components which are susceptible to irradiation assisted degradation mechanisms.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 15 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 4-3: Primary System Components Environmental Conditions System/
Component Major Components Chemical Environment Peak Lifetime Irradiation (DPA)1 RES (reactor vessel)
Reactor Vessel Beltline Sodium and Inert Gas (argon) 8.72E-6 RV Bottom Head Sodium 1.18E-4 RES (other primary pressure boundary)
-Reactor Head
-Rotatable Plug Inert Gas (argon)
<<0.3 Primary Heat Transfer
-Intermediate Heat Exchanger
-Primary Sodium Pump Sodium and Inert Gas (argon)
<<0.3 Control Rod Drive (CRD) pressure housing Pressure Housing Inert Gas (argon)
<<0.3 Fuel Handling Internal (Pressure Boundary)
-Fuel Transfer Lift Cover
-In-Vessel Transfer Machine Access Port Inert Gas (argon)
<<0.3 Sodium Cover Gas (SCG)
Piping/Valves Piping & Valves Inert Gas (argon)
<<0.3 Sodium Processing System (SPS) Piping/Valves Piping & Valves Sodium and Inert Gas (argon)
<<0.3 RES (guard vessel)
Guard Vessel Inert Gas (argon)
Not Evaluated (negligible)
RES (fixed in-vessel shielding)
Inner Liner Sodium 0.338 Outer Liner Sodium 0.026 RES (core supports, internals)
Upper Grid Plate Sodium 1.17 Lower Grid Plate Sodium 0.017
-Inner Core Barrel
-Core Inlet Plenum Sodium 0.764 SPS Pump Sodium Processing Pump Sodium
<<0.3 Note:
- 1. DPA is for 57 EFPY.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 16 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 4-4: Intermediate System Components Environmental Conditions System/
Component Major Components Chemical Environment Irradiation IHT
-Intermediate Sodium Pumps
-Piping Sodium None IHT Sodium-Salt Heat Exchangers Sodium, Molten Salt None IAC Air Heat Exchangers Sodium (tube side)
Air (shell side)
None SPS -
Intermediate Piping & Valves Sodium and Inert Gas None Unlike reactors using water coolant with additives to control pH, once the Natrium primary and intermediate systems have been filled with sodium, oxygen content is maintained by the SPS cold traps and monitored via the plugging temperature indicators. The primary source of oxygen during operation is from cover gas, which has oxygen limits maintained by the SCG. Other impurities in sodium are evaluated via the Multipurpose Sampler (MPS) on a periodic basis. The SPS maintains impurities and contaminants within the limits shown in Table 4-5 for primary and intermediate sodium systems.
The Natrium plant design utilizes a high-purity sodium operating environment which is controlled to maintain contaminants below specified levels to limit corrosion. Sodium system oxygen and temperature are the primary drivers of corrosion rates in austenitic stainless steels. Historical literature characterizes performance in a range of oxygen levels in sodium, identifying severe effects at high oxygen levels, which are not anticipated by the Natrium design. Testing for structural materials for the Natrium reactor targets a more prototypic low oxygen environment of less than 2 part per million by weight (ppmw) oxygen, which bounds operational, shutdown, and refueling conditions.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 17 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 4-5: Sodium Chemistry Impurity Targets Element Primary & Intermediate Sodium (ppmw)
Oxygen 1.5 (operating) 2.0 (shutdown and refueling) 3.0 (limited transients)
Hydrogen 0.5 Carbon 0.7 Potassium 300 Calcium 10 Silicon 1.0 Chlorides and Bromides 20 Lithium 5
Boron 25 5
MATERIALS SELECTIONS The safety classification of Natrium SSCs drives selection of codes and standards, which constrain the materials selection. New structural materials for construction under the ASME Boiler and Pressure Vessel Code are approved in accordance with ASME BPVC Section II [9], Mandatory Appendix IV. The Natrium project is not seeking to gain Code approval of any new structural materials.
Once the possible materials are identified by applicable codes and standards, materials considered for Natrium construction are evaluated for existing data including operating experience. The primary material of construction is austenitic stainless steel (316H, 316, and 304H), which have existing test and operational data within the expected thermal, sodium, and irradiation environment, good commercial availability, and established fabrication techniques. The weld filler material for these alloys in high temperature regions is 16-8-2, which was used with 316SS in the FFTF hot leg and has been evaluated for usage in other sodium fast reactor designs that have not operated. The 16-8-2 filler material has decades of high temperature operating experience in other industries.
Where gaps are identified in the literature, additional testing and/or data collection is performed to characterize specific environments (chemical, irradiation, etc.) which are not explicitly addressed in the ASME BPVC. Materials considered for structural design in environments of interest are listed in the Natrium Materials Handbooks Known Materials List. This list informs design selections for materials of construction for components exposed to Natrium reactor plant service environments.
Table 5-1 identifies materials in the known materials list in the Natrium Material Handbook, which is updated at each revision to reflect progress in data qualification, design requirements, and test planning and completion.
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X Planned SS316 X
X Planned SS304H X
X Planned Alloy 718 X
X Planned Weld Material AWS SFA-5.9 ER16-8-2 (304H/316H)
X X
With base material As the design of components and equipment progresses, materials of construction are selected from the known materials list where sodium exposure or irradiation effects are anticipated. Other plant environments, such as the inert cover gas, are not expected to impact material properties. Materials in these environments are selected based on performance and available data in the selected code.
Where design needs challenge the known materials, usage of known materials or the data needs of the application are evaluated for updates to the list.
Materials selection continues as the design evolves. Preliminary material selections of major components are identified in Table 5-2, subject to confirmation by the fabricator as component design and specification activities progress. Weld material is defined by the base material(s) and applicable codes and standards, with the final product form selected by the fabricator. For coupon-level testing to evaluate irradiation and sodium effects in high temperature areas, weld material has been selected as AWS SFA-5.9 (ER16-8-2) for 316H and 304H. No coupon-level testing is currently planned on welds of other base materials.
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CRD Driveline Assy.
Intermediate Heat Exchangers Primary Sodium Pumps Primary & Intermediate Heat Transport Systems, Sodium Cover Gas, and Sodium Processing System piping and vessels in elevated temperature sodium service 304 or 316 Tubing, Pipe, Fittings Per code and fabricator selection Intermediate piping in low temperature sodium service 2.25Cr-Mo Forging, Plate
- SFA-5.23 EB 3
- SFA-5.28 ER90S-B3
Pipes 718 weld not anticipated Control Rod Drive-Driveline Assy.
800H Plate Not welded Intermediate Heat Exchanger tube supports 1Product forms identified based on preliminary design. Non-wrought product forms are not permitted base materials per ASME BPVC Section III Division 5.
6 RELIABILITY AND INTEGRITY MANAGEMENT PROGRAM The Natrium RIM program complements the selection of materials by evaluating design features, degradation mechanisms, and reliability target values. The RIM program provides the criteria used to determine the SSCs to be included in (and excluded from) the scope of Natriums RIM program as established under ASME Section XI, Division 2.
The RIM program also provides direction for assuring the reliability and integrity of passive components whose failure could adversely affect plant safety and reliability as outlined in the ASME BPVC Section XI, Division 2. The adequacy of the PRA for use in developing the RIM program is reviewed and, once determined to be technically adequate, reliability targets are developed for the components within the RIM Program.
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The strategies are to account for all the factors that contribute to reliability, including the selection of appropriate material of construction.
Overview of RIM Program during the Design and Construction Phase Based on the ASME BPVC Section XI, Division 2, Article RIM-1 Scope and Responsibility, paragraph RIM-1.1(b) statement that "The RIM program addresses a plant's entire life cycle. It is applicable over the entire life of the plant...," the Natrium RIM Program is established at the preliminary design stage and is applied to the design, construction and operation of the plant.
The RIM Program Plan describes the establishment of the RIM Program during the design phase and outlines the elements of the RIM Program, including:
- 1. Scope Determination
- 2. Degradation Mechanism Assessment
- 3. Reliability Target Development
- 4. RIM Strategy Development
- 5. Program Implementation
- 6. Uncertainty Evaluation
- 7. Monitoring & Assessment
- 8. Program Updates Elements 1, 2, 3, 4, 5 (limited implementation as applicable to the design and construction phase),
and 6 are applied during the design and construction phase. The RIM Program Plan provides the description and details of the methodology applied in developing each of these elements.
Guidance for implementation of the RIM Program Plan and activities of the RIMEP is provided in the RIM Program Work Guide. The work guide governs the development of the identified elements of the RIM Program for the design and construction phase.
The RIM Program scope includes the SR and NSRST passive mechanical components. The RIM Program scope document provides the preliminary scope of the RIM Program along with the scoping methodology applied.
The preliminary degradation mechanism assessment of the RIM Program is documented in the DMA Report. The DMA report document provides the description of the DMA process, preliminary component-specific design characteristics and operating environment, and preliminary DMA results.
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The process for developing reliability targets and evaluating uncertainties is described in Reliability and Integrity Management Program - Reliability Target Allocation. The development of reliability targets for components within the scope of the RIM Program is performed using the available preliminary PRA at the design stage.
As degradation mechanisms are identified and the reliability targets are established for the SCCs within the scope of RIM, the identification and evaluation of RIM strategies and associated bases is performed for SSCs within the scope of RIM. This process is performed during the design and construction phase of the Natrium demonstration plant. The methodology aligns the reliability targets, applicable degradation mechanisms and their degree of susceptibility applicable to each component with component specific RIM strategies commensurate with their safety significance.
Establishing Initial RIM Strategies The Natrium plant design follows a RIPB approach that is consistent with NEI 18-04. NEI 18-04 establishes a RIPB decision making process that incorporates principles of frequency of event occurrences versus consequences of failure and measurable performance objectives.
For each component within the scope of RIM, the applicable functional requirements supporting fundamental safety functions (heat removal, radionuclide retention, reactivity control) are identified.
Critical inputs and parameters important to component functional requirements are identified:
Construction category (pipe, vessel, support, valve, etc.), and Construction Code SSC safety classification (NST, NSRST, and SR), DLs, and pressure boundary functions Materials of construction, Design Specification, drawing, and Failure Modes and Effects Analysis Operating environment (sodium vapor or liquid, molten salt, air, argon, etc.)
Damage enablers (high temperature, thermal transients, vibration, etc.)
Other critical parameters such as Design Temperature/Pressure, and Fabricator Pressure retaining functions (primary boundary, secondary boundary, functional containment)
Reliability targets for SSCs in the scope of the RIM Program are set as the mean value of the failure rates associated with the component type and failure mode from the failure data used in the PRA. If the particular SSC failure mode is not included in the PRA data, then the target reliability is set based on a conservative failure effect and the associated failure rates for impacted mitigating functions, or the frequencies for SSCs whose failure causes an initiating event. This ensures that, assuming the SSC reliabilities satisfy the targets, plant safety is maintained consistent with the PRA and margins are maintained consistent with the plant design and licensing basis. Uncertainties associated with the PRA mean values are considered when establishing RIM programmatic controls to meet the reliability targets.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 22 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED The DMA is an integral part of the identification and evaluation of RIM strategies. Section 7 details the identification of applicable degradation mechanisms, and their screening criteria. Additional inputs to development of the RIM strategies include:
Design characteristics including material, component type, pipe size, schedule, etc.
Fabrication practices including welding and heat treating.
Operating conditions - temperature, pressure, primary/secondary fluid quality, service environment (e.g., humidity, radiation).
Plant specific industry-wide service experience and research experience.
Results of pre-service and in-service examinations.
Applicable degradation mechanisms including those identified in ASME BPVC Section XI, Division 2 Mandatory Appendix VII.
Recommendations from vendors regarding examinations, maintenance, repair, and replacement.
The preliminary RIM strategies do not consider evaluation of uncertainties (ASME BPVC Section XI, Division 2, Article RIM-2, paragraph RIM-2.6). Uncertainties are present in the design inputs used to develop the initial RIM strategies and will be reduced as the design evolves. RIM uncertainties will be formally evaluated later in the RIM process when component designs are sufficiently mature to develop a realistic level of uncertainty for the operating plant. However, some additional MANDE is planned based on the information currently available to address uncertainties associated with degradations not directly identified (unknown) by the process.
The RIM Program evaluates the performance of the selected materials and if necessary, provides recommendations to the design with respect to material changes and/or other process improvements related to heat treatment, welding process development, surface treatment, etc. to inform the scope of RIM strategies necessary to manage the effects of degradation. This is an iterative process throughout the various phases of design and construction as illustrated in Figure 6-1.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 23 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Figure 6-1. Reliability and Integrity Management Process Diagram Reliability and Integrity Management Program RIMEP MANDEEP Owner or Designee Commissioning and Operation Design Construction Scope Determination Degradation Mechanism Assessment Reliability Target Development Program Implementation Uncertainty Evaluation Monitoring and Assessment Program Updates (reassess elements of RIM process)
RIM Strategy Development Formation of RIMEP Formation of MANDEEP Finalize RIM Program Elements Monitoring and Assessment Program Updates (reassess elements of RIM process)
MANDE Procedures Personnel Qualification MANDE Equipment Qualification MANDE Acceptance Criteria Pre-service Examinations Start-up and In-service MANDE Owners Activity Report (OAR) filed with NRC within 120 days of outage completion.
Monitoring and Assessment Program Updates (reassess elements of RIM process)
Finalize MANDE Elements Evaluate Pre-service examination results Evaluate start-up and in-service MANDE Results
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 24 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED RIM Strategy Selection The purpose of the RIM Program is to provide assurance that the as-designed and delivered performance of the component is maintained over its operating life such that it is consistent with the target reliability assumed in the PRA by accounting for degradation mechanisms that are not addressed in the design code. The RIM strategy selection is based on a scoring methodology that provides prioritization for developing RIM strategies using a graded risk evaluation process based on the risk of component degradation. The risk of degradation is assessed by considering the probability of degradation and the resulting consequences of one or more degradation mechanisms on the ability of the component to perform its safety-significant functions. The process includes considerations as follows:
- 1. Design Margin
- 2. Component Function
- 3. The risk importance of a postulated failure of the component
- 4. Failure consequences of the component
- 5. Susceptibility of each applicable degradation mechanism for the component
- 6. Power generation impact of degradation mechanism
- 7. Feasibility of potential MANDE for the component
- 8. Efficacy of MANDE for detecting degradation The probability of degradation scoring considers the component functional requirements and design margin, along with component susceptibility to applicable degradation mechanisms. The attributes considered in assigning the probability of degradation scores do not contain any differentiators based on the function the component is performing, such as load path or radionuclide retention. Scoring considerations for design margin and component function attributes are provided in Table 6-1. The most significant attribute in characterizing the probability of degradation, aside from the degradation mechanisms themselves, is Design Margin. Significant design margin can reduce the risk of failure due to in-service degradation. Degradation mechanisms are grouped into four degradation categories: (1) neutron irradiation effects, (2) elevated service temperature effects, (3) fluid environment effects, and (4) fatigue and creep effects. Screening criteria are used to develop the degradation mechanism susceptibility scoring to determine the probability of degradation based on considerations important to each mechanism.
The consequences of degradation scoring considers the risk importance, failure consequence, and plant availability impact of failure of the component to perform its function due to degradation. Definitions of these consequences of degradation are included in Table 6-2.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 25 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Figure 6-2 provides an overview of the risk of degradation assessment approach. The risk of degradation scoring value range is divided into four groups corresponding to four RIMSS categories (I, II, III, and IV):
RIMSS I: No MANDE or accept economic risk without explicit RIM strategy. These are components least susceptible to a degradation mechanism based on thresholds established for the operating environment and/or have a low probability of failure or consequence of failure.
RIMSS II: Basic MANDE and/or accepting vendor best practices. Components in this category are low to moderately susceptible to degradations and expected to have low to moderate required reliability. These may include RIM strategies that are less stringent or triggered when the components in the most stringent RIM strategy categories (III and IV) do not meet RIM acceptance criteria (considered an expansion category for additional exams if active degradation is found elsewhere). Basic MANDE will be developed in future design phases and is not addressed in preliminary design.
RIMSS III: Augmented MANDE. Components in this category are moderately susceptible to degradations. Components in this category may require continuous or periodic exams and monitoring but the extent is primarily to track that reliability and design basis assumptions are maintained (e.g. indirect fatigue monitoring, periodic visual exams). Augmented MANDE is considered when the degradation mechanism exceeds acceptance criteria for a RIMSS IV MANDE during the in-service phase of the program.
RIMSS IV: Lead MANDE. Components in this category have moderate to high risk of degradation. These are considered the lead components for RIM strategy development and include the most rigorous methods (e.g., rigorous analysis with significant design margin, leak monitoring, EVT-1, VT-1/3, UT, MT, PT, ECT, AE, and/or surveillance).
The RIMSS categories are determined by dividing the range of risk of degradation scores between the lowest and highest risk of degradation values into quartiles. This ensures that at least one component is assigned lead and rigorous MANDE for each degradation category, regardless of the raw score.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 26 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Table 6-1: Design Margin and Component Function Attributes Attribute Attribute Definition Component Function This attribute provides a score that relates the component function to the spectrum of failures that could occur due to the categories of irradiation, fluid environment, high temperature, and creep and fatigue effects that are considered in this report. This category considers how the defense line or load support path function is fulfilled by the component.
Design Margin The design margin is defined as the calculated performance compared to an allowable limit or allowable performance for the design life of the component. Scoring for this category considers the complexity of the analysis, and the resulting margin.
Table 6-2: Consequence Attributes Attribute Attribute Definition Risk Importance This category provides risk insight to support the consequence scoring. The scoring is based on whether the component supports one or more defense lines or load paths that are risk significant.
Failure Consequence The significance of the consequences of SSC failure also informs the risk of degradation. In this category, the safety function and defense lines supported are used to assign a score for the failure consequence. The scoring considers whether a component supports a defense line 1 feature, NSRST defense line function, SR defense line function, or a radionuclide retention function.
Power Generation Impact Degradations that may not pose safety concerns may still pose a risk to continued power generation that would impact the owner expectations of reliable operations.
The scoring for the category considers if, for the component being evaluated, the reactor must be reduced in power or shutdown to perform a repair. The duration of the repair is also considered in the scoring.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 27 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Figure 6-2: Degradation Mechanism Risk Assessment Once the risk of degradation is determined, the total scores and the scores for each of the four degradation categories are divided into quartiles. Components in the top half of the total risk of degradation scores are assigned strategies to address each degradation mechanism applicable to the component. Components with scores in the top quartile for a degradation category are also assigned strategies to address degradation. Strategies to manage degradation include design, fabrication controls, operating practices, preservice and inservice examinations, testing, MANDE, maintenance, repair and replacement. Figure 6-3 shows the qualitative relationship between the target reliability, the component design, and the strategies used to mitigate degradation. The RIM strategy assignment is based on the results of the DMA and the RIM expert panel input.
Figure 6-3 relates the risk of degradation to the target reliability, and relates the RIM strategies to management of the as-designed reliability and risk of degradation scores.
Inherent in both the as-designed reliability and the risk of degradation is the design margin of the component. Throughout the RIM process, particularly in the design phase, the design margin of the component is the principal parameter influencing the ability to meet the
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 28 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED reliability target. When a component is designed, appropriate design codes and quality measures are assigned based on the component function and required performance. This sets the as-designed reliability of a component. Additionally, degradation effects not considered by the code must be evaluated by the designer and are the subject of the RIM program. In-service failures occur when the stress state of the component exceeds the load carrying capacity. In other words, the as-designed load carrying capacity of the component is the starting point for degradation evaluations. The difference between the as-designed load carrying capacity and the actual loads and stresses the components is subjected to in-service is the available design margin. Components with very large margins can tolerate active degradation and still have margin to the load carrying capacity. Based on this, the design margin is regarded as the key parameter in the approach to meeting a target reliability.
Figure 6-3: Risk of Degradation and Target Reliability Operating Phase Programs Supporting Degradation Mechanism Risk Evaluation and RIM Strategy Development The RIM Program considers that, for some degradation mechanisms, the RIPB-driven approach presented above may be better supported by further consideration of performance-based approaches with more traditional type risk evaluations that primarily utilize operating experience and expert elicitation.
Comprehensive Vibration Assessment Program One such degradation is associated with vibration induced degradations flow induced vibration (FIV), mechanically induced vibration (MIV), and acoustically induced vibration (AIV) such as high cycle fatigue and wear. This particular area is also governed by the Comprehensive Vibration Assessment Program (CVAP) described in accordance with RG 1.20 [10]. Concurrently with the RIM activities TerraPower is developing the CVAP for the Natrium reactor which provides the basis for in-service vibration related RIM strategies. The
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 29 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED CVAP provides a comprehensive screening of all applicable vibration mechanisms for reactor internals and systems directly connected to the reactor (IHT, SCG, SPS piping systems and pumps).
The risk-ranking methodology prioritizes locations for inspection under the CVAP. In this methodology, risk is assessed as the product of the likelihood and consequence of damage to the components. The safety classification of components and the detail provided in the PRA alone does not provide the necessary consequence evaluation when considering CVAP related degradations. Therefore, the consequence of failure will include considerations for inspection accessibility and impact on power generation.
Qualitative results from the screening, combined with available flow velocities and operational experience, are used to develop a preliminary likelihood of failure. Regarding operating experience, this element is meant to act as a weighting factor and is applied at the system or component level based on correlations. Upon completion of the CVAP vibration analysis and measurement programs, quantitative predictions of margin and/or design life will be provided to further assess risk prioritization. The overall risk evaluation method for CVAP is depicted in Figure 6-4.
Figure 6-4: CVAP Risk of Failure Methodology Thermal Effects, Dynamic Events, and Vibration Thermal Effects, Dynamic Events, and Vibration (TEDEV) program evaluates thermal expansion and vibration during startup for all piping systems necessary to perform specific functions, such as safety-related shut-down condition, maintaining that condition, or mitigating accident consequences.
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TEDEV implements ASME Operations and Maintenance (OM) of Nuclear Power Plants, 2022 Edition [11], Part 3: Vibration Testing of Piping Systems, and Part 7: Thermal Expansion Testing of Nuclear Power Plant Piping Systems.
The testing, monitoring and inspection methods developed for TEDEV may be used for in-service evaluations similar to those developed under CVAP.
7 DEGRADATION MECHANISM IDENTIFICATION AND SCREENING CRITERA DEVELOPMENT Initial evaluations performed to develop the RIM Program degradation mechanism assessment (DMA), including degradation mechanism identification and screening criteria, evaluated the operating experience in sodium fast reactors against existing lists of degradation mechanisms from the industry in order to identify degradation mechanisms of interest to the Natrium reactor. The work considered degradation mechanisms identified in the then-current draft ASME BPVC Section XI, Division 2 Appendix for Liquid Metal Reactor Type Plants, as well as Nuclear Regulatory Commission (NRC) Interim Staff Guidance Materials Compatibility for non-Light Water Reactors Interim Staff Guidance DANU-ISG-2023-01 [12]. For research on operational experience of past test facilities and plants, TerraPower considered NRC Technical Letter TLR-RES/DE/CIB-2019-01 Advanced non-Light-Water Reactors Materials and Operational Experience [13] and IAEA-TECDOC-1569 Liquid Metal Cooled Reactors: Experience in Design and Operation [14] regarding maintenance, repair and replacement. While these documents provided a large quantity of degradation incidents, there was limited information on design codes, stress state, design margin, operating environment, preservice and inservice inspection, fabrication methods, and other information necessary for the DMA.
The results of initial evaluations have been further refined to develop detailed screening criteria for the degradation mechanisms applicable to the Natrium reactor design. The criteria are developed for materials of construction and operating environments identified at this phase of the design and are primarily based on literature review. The criteria are used in the RIM DMA to identify components requiring RIM strategies. This process is iterative, with the criteria being updated and expanded upon as the design matures and existing coupon and equipment test programs complete. The RIM program ensures a systematic process for integrating parallel design and qualification activities.
The RIM Program degradation mechanism assessment evaluates the probability that a component will be degraded by a particular degradation mechanism over its design life.
This methodology extends the screening criteria to provide a numeric score for each component or part relative to the degradation mechanism. A score of zero is used to indicate that the component is not susceptible to a specific degradation mechanism, while one, two or three are used to indicate increasing susceptibility to a specific degradation mechanism. Table 7-1 provides the preliminary degradation mechanisms evaluation and
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Changes to the materials of construction, service conditions and environment, or data qualification are addressed at each RIM program update phase.
Table 7-1: Screening Criteria for Degradation Mechanism Applicability Mechanism Screening Criteria Neutron Irradiation Embrittlement (reduction in fracture toughness) 21/4Cr-1Mo base metal and welds: < 1E+17 n/cm2 Austenitic SS base metal: < 0.5 dpa SS welds: < 0.3 dpa Neutron-induced Void Swelling 316 / 316H and welds: 1.0% for fluence not exceeding 3.5E+22 n/cm2, E > 0.1 MeV between 350 and 700°C (662 and 1292°F) 304 / 304H and welds: 1.0% for fluence not exceeding 2.6E+22 n/cm2, E > 0.1 MeV between 360 and 640°C (680 and 1184°F)
Neutron Irradiation Effect on Stress Relaxation in Bolting (radiation creep)
< 0.2 dpa Environmental Effects on Creep Strength and Embrittlement (reduction in creep strength)
Component normal operating temperature < 1100°F and fluence < 1E+21 n/cm2, E>0.1 MeV Thermal Aging Embrittlement (fracture toughness reduction)
Austenitic stainless steel welds with less than 0.5% Mo and less than 20% delta ferrite, if greater than 0.5% Mo then less than 14% delta ferrite.
Stress Relaxation Cracking Component normal operating temperature is less than 850°F or weld thickness less than 1/2 inch.
Creep Embrittlement Component normal operating temperature is below 800°F for austenitic stainless steels OR ASME BPVC.III-5 HBB-T-1324(a) and (b) (or similar for non-Div. 5 SSC) are met for the component.
General, Galvanic and Flow Accelerated Corrosion Component is not exposed to liquid sodium, OR Corrosion allowance has been accounted for in structural assessments for all sodium-wetted pressure boundary or support surfaces, sodium velocities are less than 25ft/s, and the steady state oxygen content of sodium is 2 wppm.
Mass Transfer, Carburization, and Decarburization Component is in contact with liquid sodium and is thicker than 0.098 in (2.5 mm) for austenitic stainless steels or 0.039 in (1 mm) for high nickel alloys.
Liquid Metal Embrittlement Inactive for all austenitic stainless steel and nickel-base alloy components/parts in contact with liquid sodium.
Stress Corrosion Cracking (SCC)
Sodium oxygen content is controlled and there is no concern for water vapor or moisture intrusion in contact with the component. No caustic environmental exposure.
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Sodium oxygen content is controlled and there is no concern for water vapor or moisture intrusion in contact with the component. No caustic environmental exposure. OR Component is always operated in an inert or air environment.
Thermal Fatigue (stratification and striping)
No concern for thermal stratification or striping for the component (e.g. well-mixed or gaseous environment) OR These phenomena are well understood and accounted for in the design and stress analysis, with an appropriate margin to allowable limits demonstrated.
Thermal Fatigue (thermal transients)
Component is not exposed to changes in temperature requiring assessment of fatigue OR Fatigue assessment has been performed and fatigue usage in limiting locations provides sufficient margin (greater than 20%) using inelastic analysis methods.
Fatigue Creep Interaction Component normal operating temperature is below 800°F for austenitic stainless steels OR All component parts have high margin to fatigue-creep damage.
Deformation (stress relief)
Component normal operating temperature is below 800°F Deformation (ratcheting deformation)
Component normal operating temperature is below 800°F for austenitic stainless steels OR Creep ratcheting strain is below 0.5% for the base metal or 0.25% for the weld metal for limiting locations using a simplified analysis method (constant core stress, neglect inelastic strains).
Fretting, Wear, Cavitation Component/part is screened out in CVAP or a vibration analysis evaluating flow, acoustic and mechanically induced vibration mechanisms applicable to the component have been assessed. AND Component is not expected to experience sliding, impact or cavitation OR Component experiences intermittent sliding or impact (e.g.
from a heatup and cooldown cycle) or minor cavitation, but deformation or damage from these will not impact the function of the component.
8 MATERIALS DATA QUALIFICATION The Natrium project uses material properties data from a wide array of sources. The data is used at a quality level that is commensurate with the design activity being performed and the safety classification of the SSCs. Design input data, such as that in the Natrium Materials Handbook, is subjected to a thorough qualification process per the TerraPower QA Program in accordance with ASME NQA-1 [15]. The following sections provide more detail on this process.
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Established Fact sources are excluded from the scope of data qualification in NQA-1.
Established fact sources include but are not limited to the ASME Boiler and Pressure Vessel Code, the Metallic Materials Properties Development and Standardization Handbook, and the Reference Fluid Thermodynamic and Transport Properties Database Handbook, distributed through the Standard Reference Data program of the National Institute of Standards and Technology.
Existing data sources that are not considered established fact include: academic sources such as peer reviewed journal article or research papers; supplier data; data acquired through partnerships with other entities on the Natrium Demonstration Reactor project; and data published by governmental agencies such as the NRC, DOE, or the International Atomic Energy Agency. These sources are to be qualified for use. Data corroboration is considered the default method of qualification for these sources before other methods of qualification are implemented. When selecting a source for use, the attributes of sources are considered:
The technical adequacy of equipment and procedures used to collect and analyze the data The extent to which the data demonstrate the properties and ranges of interest (e.g., physical, chemical, geological, mechanical)
The environmental conditions under which the data were obtained (if germane to the quality of the data)
The quality and reliability of the measurement control program under which the data were generated The extent to which conditions under which the data were generated meet the general requirements and guidance of NQA-1 Prior range of uses of the data and associated verification processes Prior peer or other professional reviews of the data and their results Extent and reliability of the documentation associated with the data Extent and quality of corroborating data or confirmatory test results The degree to which independent audits of the process that generated the data were conducted The nature of the data (fundamental research, meta-analysis, report, etc.)
Sample sizes of the data Methods used to generate correlations
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These factors will inform the requirement to qualify a source.
In some cases, a historical document may be qualified in full using the Natrium Qualification of Existing Data procedure. In other cases, a portion of a document may be qualified to satisfy the need for a specific dataset required for the Natrium project.
Operating Experience The Natrium Demonstration Reactor project relies heavily on historical/existing data and operating experience, principally from legacy DOE sodium fast reactor programs such as the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF). The EBR-II and FFTF reactor programs used the low temperature rules of ASME BPVC Section III for their design. The FFTF reactor vessel and head were designed and fabricated to the requirements of ASME BPVC Section III, Class A, 1968, and the materials were Section II.
The vessel was of Type 304 stainless steel and the head was made of ASME SA-508. The remainder of the reactor coolant pressure boundary included the primary sodium loop piping, the intermediate heat exchanger, the primary coolant pump, the secondary sodium loop piping, the dump heat exchanger (air cooling), the secondary coolant pump, and the valves in both loops. This equipment and associated piping was designed and fabricated in accordance with the requirements of ASME BPVC Section III. Principal materials of construction included Type 316H stainless steel for the hot leg in both the primary and secondary cooling loops, and Type 304 stainless steel for the components of both loops except for the hot leg isolation valve which was Type 316. Much of the material properties data from the United States Department of Energy sodium fast reactor programs was summarized in TID 26666 Nuclear Systems Materials Handbook (NSMH) published by Oak Ridge National Laboratory. The NSMH is an Applied Technology, export-controlled document first issued in January 1975. It is a multivolume document intended to provide materials data for design decisions of advanced nuclear energy systems. The NSMH evolved from two previous compilations of materials data, the FFTF Materials Design Data (BNW-891) handbook and the Liquid Metal Fast Breeder Reactor Materials Handbook (HEDL-TME 71-32). Contributors to the NSMH include working groups of experts and a national advisory group representing major United States Department of Energy contractors and national laboratories. Usage of the NSMH was originally intended for the Energy Research and Development Administration, its contractors, and the private nuclear industry, but was later indicated to be suitable for design analyses of other systems operating at elevated temperatures. The Nuclear System Materials Handbook consists of three volumes.
Volume I is titled "Design Data" and contains the best available property recommendations.
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Material Data Acquisition and Incorporation Process The Material Property Data Acquisition and Verification plan for implementing material properties data into the design is executed whenever the Natrium Materials Library lacks adequate material property data for engineering design, a material property requires qualification, or any other significant change is identified.
If a source requiring qualification is used or data is required to be combined from multiple sources to cover the applicable range, correlations are generated following best practices documented in the Natrium Materials Handbook, which defines units and symbols as well as preferred fits and equation forms. The final correlation is intended to be the result of the data corroboration qualification method per ASME NQA-1 and related procedures. Methods used to establish a correlation are well documented in an engineering report, which also communicates to the end user the limitations of the data and the risks associated with that data.
Implementation of the Material Property Data Acquisition and Verification plan results in at least one version-controlled engineering report and an accompanying data file for the Materials Library. Once the material property is approved for use, the material performance data is available for incorporation into the Natrium design.
Data Gap Identification During the conceptual design phase of the Natrium Demonstration Reactor project, the need for additional data on degradation mechanisms was expected to be the primary source of data gaps, and the longest lead to address. This led to the creation of test programs dedicated to, for example, neutron irradiation and sodium corrosion. A special focus area within the degradation mechanism testing is weldments, as data and microstructural characterization for the 16-8-2 filler metal is limited. Another area that was identified early in the project as a potential source of data gaps was tribological surface treatments and coatings due to the fact that:
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Coating technology has progressed significantly in the intervening 40+ years since the coatings were selected for the FFTF reactor, and coating performance is strongly tied to process parameters involved in the application For the preliminary design, the first iteration of the Degradation Mechanisms Assessment systematically evaluated components critical to defense line functions within its established scope as described in Section 7. Preliminary degradation screening criteria have been developed for the initial assessment of mechanisms. As these criteria are applied and RIM strategies are identified, further refinement may be needed to build confidence in the selected approach and ensure component functionality. Despite ongoing design changes, the RIM program evaluates defense line functions and determines whether coupon testing, monitoring, inspection, surveillance, or additional testing can effectively manage degradation risks during operation. As risks are identified, additional literature review is completed and may drive prioritization of examination of coupons being exposed in the materials test programs highlighted in section 9.
9 MATERIALS TESTING TerraPower reviewed existing laboratory results and operating experience to establish structural material selection and creation of RIM screening criteria. Materials testing is intended to supplement literature in the development of RIM strategies. No new material is planned to be qualified to the ASME BPVC for Natrium reactor structural applications.
Refinements to screening criteria and DMA results are anticipated as design continues, and the development and implementation of the RIM Program progresses. The RIM Program provides a holistic, interdisciplinary, and ongoing assessment and management of degradation mechanisms as they relate to the evolving design with consideration for available data limitations relative to the 60-year design life. Limitations of material qualification data or bases identified as the RIM Program implementation progresses will be addressed through RIM Strategies, which may include material testing or surveillance programs, as described further in Section 12.
The following provides descriptions of test campaigns currently underway or planned.
Although the testing plans contain structural material coupons as described, the intent is to provide additional data related to tensile properties and metallography that could be valuable in scoping surveillance test acceptance standards, evaluating component test results, or if qualification of data via literature is not able to be completed, if determined to be required as the RIM Program implementation progresses.
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Neutron irradiation of structural materials is planned in the High Flux Isotope Reactor (HFIR) reactor at Oak Ridge National Laboratory (ORNL) under a Cooperative Research and Development Agreement (CRADA). The objective of this testing is to obtain directly comparable results of base and weld materials in the low irradiation doses expected for structural materials to provide guidance for surveillance or RIM strategy development.
Reactor structural materials (e.g., 316H), structural welds, as well as surface treatments and coatings for tribological and thermal emissivity applications, are to be irradiated at representative doses (lower doses than core assembly materials). Post irradiation examination of the materials including mechanical testing will be conducted primarily at ORNL upon extraction from HFIR.
A detailed Project Plan for neutron irradiation has been developed. Table 9-1 lists the materials that are included in the project plan.
Table 9-1: ORNL/HFIR Neutron Irradiation Materials Matrix Material Sample Type Target Dose Range 316 SS
- Metallography, Tensile
- Base, coated 0.6 - 2 dpa 316 H SS
- Metallography, Tensile, Bend, Fracture Toughness
- Base, coated, welded, sensitized 0.5 - 6 dpa Alloy 718
- Metallography, Tensile, Bend, Fracture Toughness
- Base, coated 0.6 - 6 dpa Grade 22
- Metallography, Tensile, Bend
- Base, coated 0.5 dpa Target temperatures depend on the type of specimen and material. Coated coupons have target temperatures of 360°C, 450°C, and 510°C. Irradiations of structural materials and welds have target temperatures of 350°C and 550°C.
The following materials are included in the irradiation program both as structural materials and as part of a task targeted toward evaluating the performance of coatings: 316 Stainless Steel (SS), 316H SS, and Alloy 718. At this time, 304H is not included in the test matrix due to significant existing data for the alloy from operating reactors, e.g., from the Electric Power Research Institutes (EPRI) MRP-79, Materials Reliability Program: A Review of Radiation Embrittlement of Stainless Steels [16] for pressurized water reactors. For the Grade 22 and 316H SS materials, each alloy will be represented by coupons in the as-received condition as well as three emissivity-enhancing surface treatments each.
To address the Neutron Irradiation Effects shown in Table 9-1, post-irradiation examination will be performed on the irradiated test coupons. Irradiation Embrittlement will be
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Irradiation Induced Void Swelling will be characterized using density measurements of irradiated specimens.
Sodium Effects (Sodium Corrosion/Sodium-Exposed Mechanical Properties)
TerraPower has two major test programs under development for testing the effects of the sodium environment on reactor materials. For structural materials, the objective of this testing is to obtain additional data to understand the impact of sodium on these materials to provide guidance for surveillance or RIM activities. The first of these is being established at Argonne National Laboratory (ANL) under a Cooperative Research and Development Agreement (CRADA). ANL has fabricated a sodium loop, Sodium Materials Testing Loop (SMT)-3, that consists of four test chambers designed to each remain at an independent temperature as shown in Figure 9-1. Each test chamber contains a sample carousel with 15 removable sample mounts. Each mount is capable of holding approximately (depending on configuration) 24 samples, allowing for 360 samples per chamber and 1440 samples in the loop. The current loop configuration is targeting temperatures of 360°C, 500°C, 600°C, and 650°C for the respective test chambers.
Figure 9-1: The SMT-3 Corrosion Loop at Argonne National Lab The test matrix for the sodium exposure in Table 9-2 has been developed and currently contains Natrium reactor structural materials (i.e., 304H, 316H, and Alloy 718) and structural welds (304H to 304H and 316H to 316H, both with 16-8-2 filler metal) and core assembly materials and welds. The ANL sodium loop will be used primarily to characterize General Corrosion (including wastage, pitting, and dissolution of alloying elements). The ANL team produced similar data for ANL-ART-190, Decarburization and Carburization Behavior of Grade 91 in Sodium, and the SMT-3 loop design is based on this past work.
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Table 9-2: ANL Exposure and Testing Matrix Material Duration (h)
Temperature (°C)
Testing 316 H SS 0, 1000, 3000, 15000 360, 500, 600 Corrosion, Tensile, Metallography 304 H SS 0, 1000, 3000, 15000 360, 500, 600 Corrosion, Tensile, Metallography Alloy 718 0, 1000, 3000, 15000 500, 600 Corrosion, Tensile, Metallography 316 H SS Welded 0, 1000, 3000, 15000 500, 600 Corrosion, Tensile, Metallography 304 H SS Welded 0, 1000, 3000, 15000 500, 600 Corrosion, Tensile, Metallography The flow rate in the ANL loop is relatively low. For Flow Accelerated Corrosion, the TerraPower Equipment Qualification and Test team is working with the University of Wisconsin - Madison. The University of Wisconsin - Madison recently submitted a test report documenting 5,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of sodium erosion testing showing no significant impact to the plate orifices.
Fretting and Wear Two sodium loops are being fabricated for the purpose of testing Fretting and Wear degradation phenomena. The first is the Sodium Loop for Immersed Contact Knowledge (SLICK) loop that has been constructed at TerraPowers Everett laboratory facility. The SLICK loop is designed to perform sliding wear tests to evaluate friction and wear rates of structural materials and coatings in liquid sodium (up to 650) and sodium vapor environments.
The second sodium loop is being fabricated at Oregon State University (OSU) for the purpose of performing wear testing. The purpose of Self-welding Experiment and Sodium Loop for Tribology Analysis (SESoLTA) is to evaluate the fretting performance of tribological interfaces in low-oxygen liquid sodium metal for friction performance and wear resistance. SESoLTA will be used to perform standardized fretting tests either immersed in sodium (up to 650) or in sodium vapor atmosphere conditions prototypical of the Natrium reactor.
Within the tribology testing scope, specific conditions applicable to each interface throughout the reactor will need to be individually evaluated; however, the structural material pairs that are being evaluated include:
SS316H-SS316H SS304H-SS304H SS304H-SS316H SS316H-Alloy 718
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Table 9-3: SRC Study Matrix Base Material Parameters Tested Test Method 316H SS
- 1. Heat Input (Low, Med., High)
- 2. Last Bead Placement
- 3. Repair (1/2 of the samples have repair)
Heat to 620°C for 500 hrs.
The testing method is to heat the samples to 620°C (1150°F) for 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> X 10 cycles (5,000 hrs. total). The method to track the initiation and extent of SRC is to perform dye penetrant (PT) and ultrasonic (UT) inspection prior to the first heating cycle and between each of the subsequent heating cycles. Post-test metallography will characterize the heat affected zone. Results will be used to inform fabrication practices which minimize SRC susceptibility.
10 MONITORING AND NON-DESTRUCTIVE EXAMINATION PRACTICES Not all SSCs will require monitoring and/or inspection by MANDE. In cases where similar SSCs are monitored and inspected, or where sufficient operating experience is available, or reliability targets for the SSC are adequately met through design considerations and margins, the requirements for MANDE may be limited or eliminated for specific SSCs. Not all MANDE methods are capable of and/or appropriate for effective management of degradation for some SSCs. Each MANDE method has unique capabilities in detection of degradation, and this informs the selection of MANDE.
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Applicability of standard NDE methods (e.g., utilized for ASME Section XI, Div 1 in-service inspection) for the Natrium reactor plant include visual, volumetric and surface examination.
However, application to Natrium has some limitations based on plant temperatures and lack of access to certain components. For example, standard visual inspection of reactor internals cannot be performed because large scale draining of sodium and/or removal of reactor internals is not practical due to both economic limitations and industrial safety risk (increased fire hazard and handling of volatile material). However, visual inspections can be performed on non-sodium wetted surfaces and an under-sodium viewing (USV) sonar/acoustic based visual-type examination is being investigated for use on sodium-wetted surfaces inside the reactor vessel. Feasibility assessment of ASME BPVC Section XI Division 1 exams for the SSCs within the scope of RIM is an ongoing effort in collaboration with EPRI. The purpose of this effort is to evaluate and provide technical basis for performing typical ASME-XI.1 exams for components where requirements for management of degradation has been identified.
Eddy Current Exams of Intermediate Heat Exchanger Tubes Eddy current test (ECT) is used for measuring axial and circumferential wall thickness of heat exchanger tubes. There is no current industry experience using ECT probes in liquid sodium, however a proof-of-concept demonstration is under development to determine feasibility of using ECT in sodium for detecting IHX tube flaws to demonstrate compatibility with liquid sodium and high temperature environment. Results will help to inform the IHX MANDE strategy implemented by RIM.
Continuous (Leakage) Monitoring System The primary coolant boundary leakage monitoring approach for the reactor enclosure includes the leakage detection in the RV-GV annulus. Additional leakage monitoring is being developed as potential MANDE for other high temperature systems (SPS and SCG) which form part of the primary coolant pressure boundary, as well as for intermediate sodium systems - IHT and intermediate air cooling (IAC) - as indicated in Table 10-1.
Although not fully developed or implemented at the preliminary design stage, to support development of the leakage monitoring approach, calculations based on leak-before-break methodologies will be performed to determine necessary leakage detection capability. The purpose is to develop a methodology that can be used to quantify leak rate based on the critical flaw size for a component, with appropriate safety margins considered. The leak rate will be used to design proper leak monitoring systems for potential component specific sodium leakage monitoring applications. The process evaluates components for leak detection consistent with leak-before-break (LBB) analyses and includes the following:
Methodology for calculation of stable crack sizes for various design basis conditions Methodology for calculation of crack opening displacements (COD)
Methodology for correlating leak rate with crack sizes
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Level Monitors on Hot Pool (Continuous Inductive Type)
Oxygen Sensors Radiation Detector (In-Line Gamma Scintillator Type)
Sodium Aerosol Detector (Filter Type)
Continuity Liquid Sodium Detector (Type TBD)
IHX Primary/Intermediate Sodium Boundaries Level Monitors on Expansion Tank (Pressure/Bubbler and Conductive Discrete)
Level Monitors on Hot Pool (Continuous Inductive Type)
Pressure Sensors (Diaphragm Type Gauge Pressure Sensor)
Reactor Head Penetrations Radiation Detector (High Range Ion Chamber Type)
SCG Piping and Isolation Valves Radiation Detector (High Range Ion Chamber Type)
Flow and Pressure Instrumentation SPS Piping and Isolation Valves Continuity Liquid Sodium Detectors (Spark Plug Type and Beaded Wire)
Level Monitors on Hot Pool (Continuous Inductive Type)
Instrument Drywells Simultaneous Failure of Drywell Instrumentation Radiation Detector (High Range Ion Chamber Type)
Flow Module Temperature Sensors (Type K)
Radiation Detector (High Range Ion Chamber Type)
TBD - Simultaneous Failure of Drywell Instrumentation IHX Upper Plenum Inspection Ports Continuity Liquid Sodium Detector (intermediate side only, Type TBD)
Radiation Detector (High Range Ion Chamber Type)
IHX Skirt TBD - Based on clamshell enclosure design IHX Gas Annulus Boundaries Pressure Sensors (Diaphragm Type Gauge Pressure Sensor)
Level Monitors on Hot Pool (Continuous Inductive Type)
Level Monitors on Expansion Tank (intermediate side only, Pressure/Bubbler and Conductive Discrete)
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The method considers inputs such as temperature and flow rate data from the primary and intermediate sodium system instrumentation to generate stress time histories at key creep-fatigue locations. Since direct measurements are limited, Greens functions and reduced order models are used to create stress transfer functions that translate instrument data histories into local stress responses. These stress time histories are then used to calculate both creep and fatigue damage, which is compared against the allowable envelope defined by the design Code. Remaining below this threshold supports confidence in meeting the target reliability, and direct measurement of creep-fatigue is not anticipated.
The baseline creep-fatigue monitoring system (CFMS) is designed to utilize currently planned plant instrumentation to provide basic monitoring of high-temperature structural components. This system is intended to meet the minimum requirements for condition monitoring of applicable RIMSS Category III components within the Natrium plant and to establish a platform for future enhancement.
Enhanced CFMS The methods outlined here as considerations for the enhanced CFMS offer additional degradation management capabilities that may be needed depending upon component-specific available margin to the damage envelope and the complexity of the degradation. In many cases, a single damage mechanism dominates (creep or fatigue), but in scenarios where synergistic effects of degradation may be anticipated to occur, such as fatigue-induced plasticity contributing to creep cycles, these enhancements become more relevant.
These potential enhancements focus on both the sensor technologies and the available analytical algorithms. Preliminary enhancements under consideration include:
Installation of thermocouples on high temperature piping and/or other key locations Installation of acoustic emission sensors and/or thick film ultrasonic testing (UT) sensors.
Installation of strain sensors (fiber or similar)
Utilizing the above sensors monitoring the limiting locations in the IHT hot leg piping directly would reduce uncertainty associated with transfer function approaches and help establish a basis for applying those functions at other hard to access locations. Additionally, sensors could be placed near high susceptibility creep-fatigue damage RV locations (within the RV-GV annulus) - specifically at the azimuth and elevation where the damage is most severe.
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Alternatively, high-temperature ultrasonic testing with thick-film sensors could be used, assuming the crack location is known with certainty.
Placement of enhanced sensors capable macrocracking monitoring addresses the monitoring of long term effects of high temperature creep and fatigue. Placement in representative locations, such as the IHT hotleg (which is one the highest temperature components, is constructed of base material and weld filler consistent with other high temperature components, has well understood piping stresses, and is subjected to sodium flow), provides a surrogate for predictive damage monitoring for other plant components.
Although neutron irradiation is less significant for IHT piping, irradiation dose in the high temperature areas of the reactor internals and other primary structural components is not significant for creep fatigue degradation per the established screening criteria.
The placement of enhanced CFMS sensors can be informed by the results of the TEDEV and CVAP programs where baseline vibration and thermal expansion (stress) measurements can provide early detection of the most highly susceptible locations due to sustained and low cycle thermal stresses as well high cycle vibrations due to flow or mechanical sources (pumps).
Under Sodium Viewing System Under sodium viewing is any technology that can provide NDE capabilities equivalent to ASME Section XI Division 1, VT-1 or VT-3 examination requirements per IWA-2211 and IWA-2213 in the liquid sodium environment. The primary target areas and locations for application of the under sodium viewing technology include various sodium wetted surfaces. Penetrations are available through the reactor head to accommodate deployment of USV, including access tubes to allow placement of instruments to inspect selected core support welds. Feasibility assessment of an acoustic (sonar) based USV system to perform inspections using the provided access is in progress.
Future Planned MANDE Activities The MANDE activities outlined below are preliminary and subject to change. Depending on results of on-going design maturation some activities may or may not be executed or others not listed herein may be added. The descriptions herein are for information only.
Loose Parts Monitoring System Acoustic emissions (AE) testing and loose parts monitoring have been implemented extensively in traditional LWR designs. There is also some experience at the applied research level with the application of this technology on the FFTF SFR design.
Division 2, Part 12 of ASME OM-2020 provides a standard for the implementation of loose part monitoring systems in water-cooled reactors. Later editions of this standard (ASME
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Piezoelectric thick-film sensors Piezoelectric thick-film sensors for thickness measurement applications have been demonstrated to operate in industrial environments for long periods of time (i.e., years) at temperatures comparable to those present in the Natrium reactor. Additional work is needed to demonstrate the long-term operability of thick-film sensors in a high-temperature environment present in the Natrium reactor.
Fiber-optic sensors Fiber-optic temperature sensors have been demonstrated in high-temperature and high-radiation environments, including application in SFR test loops and research reactors; however, optical fibers for strain/vibration monitoring have not been demonstrated in these same conditions. Further work is needed to demonstrate strain/vibration monitoring in high-temperature and high-radiation environments.
Materials Surveillance Coupon Program Development The Natrium reactor design provides features for locating surveillance specimens in the applicable sodium environment. The primary placement for the test specimens would be in a surveillance assembly located in the In-Vessel Fuel Storage ring. This location is exposed to primary sodium and accessible by fuel handling equipment for removal. FFTF used the materials open test assembly to store and periodically retrieve material coupons for testing.
The materials open test assembly would allow removal of individual material test specimens during reactor shutdowns. Other locations in the reactor have been assessed; they include the in-vessel transfer machine riser in the reactor head and the gap between the reactor vessel and the guard vessel. Neither of these locations seem acceptable due to poor neutron fluence, lack of sodium environment, and/or too low of an operating temperature.
In addition, the surveillance program would assess the range of potential degradation mechanisms, with emphasis on combined effects and singular mechanisms beyond mass transfer, and evaluates whether these mechanisms warrant the development of a dedicated surveillance program to address time dependent material property changes in the operating environment. The surveillance program consists of the following steps:
Prioritization of Key Characteristics - Identify the key material properties, service conditions, or characteristics that are most relevant for monitoring. Selection will be informed by practical relevance and potential performance impact under service conditions.
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Test Coupon Configuration - Define the test coupon design, including location and justification for specimen placement within the reactor vessel. This will include evaluating the technical tradeoffs related to the two candidate locations: 1) in vessel storage; and 2) reflector zone regions.
Sampling Strategy and Degradation Assessment - Outline the recommended frequency for coupon testing and describe the overall sampling strategy. It will also evaluate potential interactions of multiple degradation mechanisms to help identify any gaps in the surveillance approach.
11 TRIBOLOGICAL COATINGS The use of coatings (also referred to as hard-facings) and surface treatments are being evaluated for Natrium reactor plant components in sodium-wetted environments.
Coatings and surface treatments applied in sodium-wetted environments are exclusively employed for tribological purposes, i.e., for the reduction of wear, friction, and self-welding, rather than corrosion protection.
At the preliminary stage of design, the application of coatings and surface treatments has been identified for use on reactor core assembly components. Although not determined at this design stage, coatings may be used in the design of other SSCs in the sodium-wetted environment, which may include:
Primary sodium pump bearings, seal rings, and other wear surfaces Fuel handling equipment sliding mechanisms, bearings, and grapples Sealing interfaces between various systems such as the reactor head, IHX, etc.
In-vessel storage racks The design of these components will consider other design options such as dissimilar material selection to improve tribological performance prior to selection of coatings and surface treatments. It is expected that the structural material applications will select from the coatings and surface treatments under development for core assembly applications. SSC-relevant laboratory scale materials testing will be performed for the selected applications. Coatings applied to core components are based on previous work and historical coatings used for the FFTF program.
The coatings utilized for core components have been preliminarily classified as safety related (QL-1) and subject to ASME NQA-1 requirements for all work performed to qualify and develop coating parameters. These requirements are being fulfilled via implementation of the TerraPower quality assurance program as defined in TerraPower QA Program Description.
The core components and substrate materials to which coatings or surface treatments are applied, or being considered for application, along with the purpose of the coatings, are given in Table 11-1 and Table 11-2.
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Located on assembly ducts HT9 Detonation Gun Applied Chrome Carbide (LC-1H)
Detonation Gun T700 Detonation Gun NiAl [Note 1]
Protect assembly load pads from wear damage and preventing self-welding Top Load Pad (TLP) - Located on handling socket 316H Stainless Same as for ACLP Same as for ACLP Same as for ACLP Core Assembly Inlet Nozzle 316H Stainless Hard Chrome Plating Aluminum Diffusion Coating Prevent galling and reducing friction between the inlet nozzle and receptacle Seal Rings -
Located between core assembly duct and inlet nozzle Alloy 718 Aluminum Diffusion Chromium Diffusion Coating Prevent wear or damage to the seal rings from occurring and leading to pressure loss and leakage of sodium Note 1: Alternate coatings also include the same materials deposited via thermal spray or laser cladding for both the ACLP and TLP.
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Alloy 718 316H Stainless HT9 Hard Chrome Aluminum Diffusion Ensure that the compressive assemblies can actuate by mitigating self-welding and friction Primary and Secondary Control Rod Assemblies Alloy 718 316H Stainless HT9 Thermal Spray, Detonation Gun or Laser Cladded Chrome Carbide / Tribaloy 700 Hard Chrome Ensure release of control rod assemblies and proper operation of damper system by mitigating self-welding and friction Lead Test Pin Receptacle Alloy 718 316H Stainless HT9 Hard Chrome Aluminum Diffusion Ensure lead test pins are removable from the fuel assembly by mitigating self-welding and friction The anticipated environments for the core components listed in Table 11-1 are enveloped by the conditions given in Table 11-3, which are based on normal operation and design basis events. The variation in dose is due to differences in assembly lifetimes and operating conditions for the core. The temperature ranges encompass the range of design conditions. Transient temperature rates of change are considered for the qualification of the coatings but are under development at this preliminary design stage.
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Temperature Range Sodium Exposure Above Core Load Pad (ACLP) - Located on fuel assembly ducts 4 - 40 350 - 1200 ºF Yes Top Load Pad (TLP) -
Located on handling socket 0 - 0.5 350 - 1200 ºF Yes Inlet Nozzle 0 - 1 350 - 930 ºF Yes Seal Rings - Located between duct and inlet nozzle 0 - 1 350 - 930 ºF Yes Core component coating development and qualification activities are in progress and are further described in Section 12.2. Application of coatings for structural components is under development and is further described in Section 12.2.
12 AREAS REQUIRING FURTHER RESEARCH AND DEVELOPMENT 12.1 Assurance of Adequate Structural Materials Performance in Service The following activities have been identified to provide further development of information needed to provide assurance of adequate structural materials of construction performance for safety-significant SSCs included in the RIM Program:
- a. Research activities to improve understanding of the effects of high temperature, chemistry exposure, and irradiation on materials, including weld metals. For example, the stress relaxation cracking testing program (described in Section 9).
- b. Determination of requirements for further materials testing for environmental compatibility
- c. Activities to support the RIM Program to mature and develop appropriate performance monitoring methods, such as -
- i. material surveillance programs, ii. inspection methods and means of access, and iii. monitoring approaches supported by validated technical bases. For example, the development of PFM, LBB and leakage monitoring approaches to support RIM implementation (described in Section 10).
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In summary, the overall demonstration of the adequacy of material selection; degradation mechanism identification, screening, and assessment; and environmental degradation management strategies is an element of the Natrium reactor plant safety analysis requiring further development of technical information. Resolution may include initial assumptions on the effects of the environment on through-life design properties if environmental data are not available or cover only a portion of the specified operating lifetime, that could be validated through future testing or performance monitoring, such as materials surveillance during plant operation.
12.2 Coatings Development, Application, and Qualification The following development, application, and qualification activities have been identified to provide further development of information needed to demonstrate the acceptability for coatings in the Natrium reactor plant:
- a. Documentation of the review of operational experience and pre-existing data on coatings used in sodium fast reactor environments
- b. Development of the qualification plan for core component coatings and relevant quality documents
- c. Production of test samples and optimization of procedures for coating deposition based on pre-existing data
- d. Evaluation (testing) of environmental degradation mechanisms for components including phenomenon such as:
- i. Thermal cycling ii. Irradiation damage iii. Tribological damage iv. Mechanical damage
- e. Functional testing of coatings applications and deposition processes
- f. Impact evaluation of coating failure / delamination on other systems
- g. Development of procedures and parameters for production of coated
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- h. Development of standardized specifications for coating production and post-application inspection requirements for coatings In addition, the application requirements for tribological coatings for structural components is under development, and the coatings materials and activities identified for core component coatings application and qualification are expected to be applicable to structural applications.
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[1]
American Society of Mechanical Engineers, "Boiler and Pressure Vessel Code,Section XI Rules for Inservice Inspection of Nuclear Power Plants, Division 2 Requirements for Reliability and Integrity Management (RIM) Programs for Nuclear Power Plants," ASME, 2019.
[2]
Nuclear Energy Institute, "Risk-Informed Performance-Based Technology Inclusive Guidance for Advanced Reactor Licensing Basis Development, NEI 18-04 Revision 0," NEI, 2019.
[3]
U.S. NRC, Regulatory Guide 1.233 Rev. 0: 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, 2020.
[4]
U. S. NRC, Regulatory Guide 1.201 Rev. 1: Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to their Safety Significance, 2006.
[5]
American Society of Mechanical Engineers, "Boiler and Pressure Vessel Code,Section III Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors," ASME, 2017.
[6]
U. S. NRC, Regulatory Guide 1.87, Rev. 2: Acceptability of ASME Code Section III, Division 5, 'High Temperature Reactors', 2023.
[7]
American Society of Mechanical Engineers, "Boiler and Pressure Vessel Code,Section VIII Rules for Construction of Pressure Vessels," ASME, 2023.
[8]
American Society of Mechanical Engineers, "B31 Code for Pressure Piping,"
ASME.
[9]
American Society of Mechanical Engineers, "Boiler and Pressure Vessel Code,Section II Materials," ASME.
[10]
U.S. NRC, Regulatory Guide 1.20 Rev. 4: Comprehensive Vibration Assessment Program for Reactor Internals during Preoperational and Startup Testing, 2017.
[11]
American Society of Mechanical Engineers, Operations and Maintenance of Nuclear Power Plants, 2022.
[12]
U. S. NRC, DANU-ISG-2023-01: Materials Compatibility for non-Light Water Reactors Interim Staff Guidance, 2023.
[13]
U. S. NRC, TLR-RES/DE/CIB-2019-01 Advanced non-Light-Water Reactors Materials and Operational Experience, 2019.
[14]
International Atomic Energy Agency, IAEA-TECDOC-1569 Liquid Metal Cooled Reactors: Experience in Design and Operation, 2007.
[15]
American Society of Mechanical Engineers, "NQA-1, Quality Assurance Requirements for Nuclear Facility Applications," ASME, 2015.
[16]
Electric Power Research Institute, "MRP-79, Materials Reliability Program: A Review of Radiation Embrittlement of Stainless Steels," EPRI, 2004.
[17]
U. S. NRC, Regulatory Guide 1.245 Rev. 0: Preparing Probabilistic Fracture Mechanics (PFM) Submittals, 2022.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 54 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED
[18]
U. S. NRC, Regulatory Guide 1.133 Rev. 1: Loose-Part Detection Program for the Primary Systems of Light-Water-Cooled Reactors, 1981.
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NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 56 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Acronym Definition MANDE Monitoring and Non-Destructive Examination MANDEEP MANDE Expert Panel MPS Multipurpose Sampler M-Rule Maintenance Rule Program MT Magnetic Particle Testing NDE Non-Destructive Examinations NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSMH Nuclear Systems Materials Handbook NSRST Non-Safety-Related with Special Treatment NSS Nuclear Island Salt System NST Non-safety-related with no special treatment OE Operating Experience ORNL Oak Ridge National Laboratory OSU Oregon State University PDC Principal Design Criterion PFM Probabilistic Fracture Mechanics PRA Probabilistic Risk Assessment PSAR Preliminary Safety Analysis Report PT Penetrant Testing QA Quality Assurance RAC Reactor Air Cooling System RES Reactor Enclosure System RG Regulatory Guide RIM Reliability and Integrity Management RIMEP RIM Expert Panel RIMSS RIM Strategy Significance RIPB Risk-Informed, Performance-Based RPA Rotatable Plug Assembly RH Reactor Head RV Reactor Vessel SCG Sodium Cover Gas
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 57 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED Acronym Definition SFR Liquid Metal (Sodium) Cooled Fast Reactor SHX Sodium-Salt Heat Exchanger SMT Sodium Materials Testing Loop SPS Sodium Processing System SR Safety-Related SS Stainless Steel SSC Structure, System, and Component TBD To Be Decided TEDEV Thermal Effects, Dynamic Events, and Vibration TRL Technical Readiness Level USV Under Sodium Viewing UT Ultrasonic Test VT Visual Testing
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC IHT Piping and Valves SS 316H or 304H, 16-8-2 weld material ASME B31.1 Creep effects evaluated per ASME III.5.HCB-3634(b)
Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Apply ASME B31.1 toxic fluid requirements Full penetration pressure boundary fusion welds without permanent backing strips 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.311 CMTRs required for pressure boundary material ASME B31.1 Power Piping Code is a recognized, industry consensus piping construction code that provides materials, design, fabrication, and construction requirements for long service life and reliability in power plant installations and industrial piping systems, including power piping. ASME B31.1, with enhancements, is appropriate for application to the pressure-retaining construction to provides reasonable confidence that the piping and valves will perform their safety-significant functions.
Degradation in material properties due to high temperature service are evaluated per ASME BPVC Section III, Division 5 for Class B piping to account for additional limits on allowable stresses. Piping temperatures and stresses resulting from ASME III defined service level A, B and C reactor transients are used to determine loading conditions.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
Requirements imposed by ASME B31.1 122.8.2 Toxic Fluid provide additional assurance that liquid sodium containing piping has sufficient margin to failure. Additional analysis for impact, shock and vibration is required. Mechanical connections are strictly limited, and allowable weld configurations minimize internal crevices. Design margin is maintained by prohibiting allowances for occasional loads. Further assurance against leaks is provided through additional requirements for valves and sensitive leak testing.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the material properties are correct.
IHT Expansion Tanks SS 304H, 16-8-2 weld material ASME Section VIII, Division 1 Creep effects evaluated per Code Case 2843-3 (or ASME III.5 Appendix HBB-T) / Cyclic loading analysis per ASME Section VIII Div 2 Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Apply ASME BPVC VIII Div 1 Service Restrictions for Lethal Service Applications 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Full penetration pressure boundary fusion welds without permanent backing strips Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.31 CMTRs required for pressure boundary material ASME BPVC Section VIII Div 1 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not directly address fatigue or ratcheting and requires augmentation by ASME VIII Div 2. However, Division 2 does not address creep effects and supplemental analysis is needed to assure reliability. The vessel is analyzed for non-negligible creep using Code Case 2843-3 based on operating loads resulting from ASME Section III defined service level A, B and C reactor transients. If creep is significant, creep-fatigue damage is analyzed using Code Case 2843-3 (alternate:
ASME Section III Division 5 HBB-T). If creep is negligible, ASME Section VIII Division 2 cyclic loading analysis is sufficient to account for thermal induced cyclic stress.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
Requirements imposed by ASME BPVC Div 1 UW-2(a) Service Restrictions for Lethal Service Applications provide additional assurance that liquid sodium containing vessels have sufficient and well characterized margin to failure by requiring full penetration pressure boundary weld and confirming weld quality per UW-11(a) full radiography, and requiring visual inspection
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC after hydrostatic / pneumatic tests. The use of flanges are restricted, but if used, special limits for rigidity and bolt spacing apply.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
ISP Tank SS 316 or 304; weld material as permitted by Section III Division 5, Table HBB-I-14.1(b) for Type 304 and 316 base metal ASME Section VIII, Division 1 Cyclic loading analysis per ASME Section VIII Div 2 Apply ASME BPVC VIII Div 1 Service Restrictions for Lethal Service Applications 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Full penetration pressure boundary fusion welds without permanent backing strips Delta ferrite limits for weld material in service below 800°F as specified in ASME Section III Division 5 - HCA (NC-2433) with restrictions imposed by Reg Guide 1.31 CMTRs required for pressure boundary material ASME BPVC Section VIII Div 1 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not directly address fatigue or ratcheting and supplemental analysis is needed to assure reliability. The vessel is analyzed using ASME Section VIII Division 2 cyclic loading analysis based on operating load cycles resulting from ASME Section III defined service level A, B and C reactor transients to account for thermal induced cyclic stress.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SHX NOTE:
The SHX design is not finalized and material selections may be subject to change.
SS 316 / 316L 16-8-2 weld material ASME Section VIII, Division 1 To be determined for the SHX final design.
To be determined for the SHX final design.
IAC AHX Tubes and Nozzles SS 304H; 16-8-2 weld material ASME Section VIII Division 2 Creep effects evaluated per Code Case 2843-3 (or ASME III.5 Appendix HBB-T)
Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Full penetration pressure boundary fusion welds without permanent backing strips 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.31 CMTRs required for pressure boundary material ASME BPVC Section VIII Div 2 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 2 provides rules for construction of pressure vessels with design by analysis option for better margin characterization. ASME VIII Division 2, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
Division 2 does not address creep effects and supplemental analysis is needed to assure reliability. The vessel is analyzed for non-negligible creep using Code Case 2843-3 based on operating load cycles resulting from ASME Section III defined service level A, B and C reactor transients. If creep is significant, creep-fatigue damage is analyzed using Code Case 2843-3 (alternate: ASME Section III Division 5 HBB-T). If creep is negligible, ASME Section VIII Division 2 cyclic loading analysis is sufficient to account for thermal induced cyclic stress.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SCG-P Piping and Valves High temperature service:
SS 316H or 304H, 16-8-2 weld material Low temperature service:
SS 316 or 304; weld material as permitted by Section III Division 5, Table HBB-I-14.1(b) for Type 304 and 316 base metal ASME B31.3 Creep effects evaluated per ASME III.5.HCB-3634(b)
Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Apply ASME B31.3 Category M service requirements 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Full penetration pressure boundary fusion welds without permanent backing strips Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.31 Delta ferrite limits for weld material in service below 800°F as specified in ASME Section III Division 5 - HCA ASME B31.3 Process Piping Code is a recognized, industry consensus piping construction code that provides materials, design, fabrication, and construction requirements for industrial piping systems, including chemical and hazardous materials systems piping. In addition, ASME B31.3 is endorsed for radiological waste system service in Reg Guide 1.143. ASME B31.3, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the piping and valves will perform their safety-significant functions.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
Requirements imposed by ASME B31.3 Category M Service provide additional assurance the piping has sufficient and well characterized margin to failure. Additional analysis for impact, shock and vibration are required. Mechanical connections are strictly limited, and allowable weld configurations minimize internal crevices. Visual inspection of welds and joints is increased from a limited set to all welds. Volumetric weld inspection requirements for specified welds are increased from 5% to100%. Design margin is maintained by prohibiting allowances for occasional variations in pressure and temperature.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium
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Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC (NC-2433) with restrictions imposed by Reg Guide 1.31 CMTRs required for pressure boundary material facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration. A minimum delta ferrite improves weld quality at all temperatures and an upper limit reduces high temperature degradation.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SCG-P Vapor Trap Condenser SS 316H 16-8-2 weld material ASME Section VIII, Division 1 Creep effects evaluated per Code Case 2843-3 (or ASME III.5 Appendix HBB-T) / Cyclic loading analysis per ASME Section VIII Div 2 Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Full penetration pressure boundary fusion welds without permanent backing strips 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.31 CMTRs required for pressure boundary material ASME BPVC Section VIII is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not address fatigue or ratcheting and requires augmentation by ASME VIII Div 2. However, Division 2 does not address creep effects and supplemental analysis is needed to assure reliability. The vessel is analyzed for non-negligible creep using Code Case 2843-3 based on operating loads resulting from ASME Section III defined service level A, B and C reactor transients. If creep is significant, creep-fatigue damage is analyzed using Code Case 2843-3 (alternate: ASME Section III Division 5 HBB-T). If creep is negligible, ASME Section VIII Division 2 cyclic loading analysis is sufficient to account for thermal induced cyclic stress.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 64 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED NSRST SSC Materials Design/
Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SCG-P Primary Aerosol and Cesium Filters SS 316 or 304, weld material as permitted by Section III Division 5, Table HBB-I-14.1(b) for Type 304 and 316 base metal ASME Section VIII, Division 1 Cyclic loading analysis per ASME Section VIII Div 2 Full penetration pressure boundary fusion welds without permanent backing strips 100% volumetric examination (RT or UT) of pressure boundary welds.
Delta ferrite limits for weld material in service below 800°F as specified in ASME Section III Division 5 - HCA (NC-2433) with restrictions imposed by Reg Guide 1.31 CMTRs required for pressure boundary material ASME BPVC Section VIII Div 1 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not directly address fatigue or ratcheting and supplemental analysis is needed to assure reliability. The vessel is analyzed using ASME Section VIII Division 2 cyclic loading analysis based on operating load cycles resulting from ASME Section III defined service level A, B and C reactor transients to account for thermal induced cyclic stress.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SPS-P Sodium Piping and Valves SS 316 or 304, weld material as permitted by Section III ASME B31.1 Apply ASME B31.1 toxic fluid requirements ASME B31.1 Power Piping Code is a recognized, industry consensus piping construction code that provides materials, design, fabrication, and construction requirements for long service life and reliability in power plant installations and
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 65 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED NSRST SSC Materials Design/
Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC Division 5, Table HBB-I-14.1(b) for Type 304 and 316 base metal ASME B16.34 (main isolation valves)
Apply ASME B16.34 Section 8 for special class valves to main isolation valves Full penetration pressure boundary fusion welds without permanent backing strips 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Delta ferrite limits for weld material in service below 800°F as specified in ASME Section III Division 5 - HCA (NC-2433) with restrictions imposed by Reg Guide 1.31 CMTRs required for pressure boundary material industrial piping systems, including power piping. ASME B31.1, with enhancements, is appropriate for application to the pressure-retaining construction to provides reasonable confidence that the piping and valves will perform their safety-significant functions.
Requirements imposed by ASME B31.1 122.8.2 Toxic Fluid provide additional assurance that liquid sodium containing piping has sufficient margin to failure. Additional analysis for impact, shock and vibration is required. Mechanical connections are strictly limited, and allowable weld configurations minimize internal crevices. Design margin is maintained by prohibiting allowances for occasional loads. Further assurance against leaks is provided through additional requirements for valves and sensitive leak testing.
ASME B16.34, Section 8 requirements for valves are imposed on the main line isolation valves for additional NDE and limits on repair methods.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the material properties are correct.
SPS-P Cold Trap, Cesium Trap Pressure Boundaries SS 316 or 304; weld material as permitted by Section III Division 5, Table HBB-I-14.1(b) for Type 304 and 316 base metal ASME Section VIII, Division 1 Cyclic loading analysis per ASME Section VIII Div 2 Apply ASME BPVC VIII Div 1 Service Restrictions for Lethal Service Applications 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
ASME BPVC Section VIII Div 1 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 66 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED NSRST SSC Materials Design/
Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC Full penetration pressure boundary fusion welds without permanent backing strips Delta ferrite limits for weld material in service below 800°F as specified in ASME Section III Division 5 - HCA (NC-2433) with restrictions imposed by Reg Guide 1.31 CMTRs required for pressure boundary material reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not directly address fatigue or ratcheting and supplemental analysis is needed to assure reliability. The vessel is analyzed using ASME Section VIII Division 2 cyclic loading analysis based on operational load cycles resulting from ASME Section III defined service level A, B and C reactor transients to account for thermal induced cyclic stress.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
SPS-P High Temperature MPS Modules Pressure Boundaries SS 316H or 304H, 16-8-2 weld material ASME Section VIII, Division 1 Creep effects evaluated per Code Case 2843-3 (or ASME III.5 Appendix HBB-T) / Cyclic loading analysis per ASME Section VIII Div 2 Allowable stresses for elevated temperature are limited to ASME III Div 5 for 500,000hr service Apply ASME BPVC VIII Div 1 Service Restrictions for Lethal Service Applications Full penetration pressure boundary fusion welds without permanent backing strips ASME BPVC Section VIII Div 1 is a recognized, industry consensus pressure vessel construction code that provides materials, design, fabrication, examination, inspection, testing, certification and pressure relief requirements for industrial pressure vessels, including vessels in power plants. ASME VIII, Division 1 provides rules for construction of pressure vessels, supplemented by design analysis where required by the Code.
ASME VIII Division 1, with enhancements, is appropriate for application to the pressure-retaining construction to provide reasonable confidence that the vessel will perform its safety-significant function.
ASME VIII Div 1 does not address fatigue or ratcheting and requires augmentation by ASME VIII Div 2. However, Division 2 does not address creep effects and supplemental analysis is needed to assure reliability. The vessel is analyzed for non-negligible creep using Code Case 2843-3 based on operating loads resulting from ASME Section III defined service level A, B and C reactor transients. If creep is significant, creep-fatigue damage is analyzed using Code Case 2843-3 (alternate: ASME Section III Division 5 HBB-T). If creep is negligible, ASME
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 67 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED NSRST SSC Materials Design/
Construction Code Enhancements to Baseline Code Requirements Justification that selected C&S and enhancements are appropriate for the NSRST SSC 100% volumetric examination (RT or UT) of pressure boundary fusion welds.
Delta ferrite limits for weld material in service above 800°F as specified in ASME Section III Division 5 - HCB-2433 and verified per Reg Guide 1.31 CMTRs required for pressure boundary material Section VIII Division 2 cyclic loading analysis is sufficient to account for thermal induced cyclic stress.
Allowable stresses from ASME III Division 5 extended to 500,000hr are used to ensure material properties are not degraded by exposure to elevated temperature for times greater than 100,000 hrs.
Requirements imposed by ASME BPVC Div 1 UW-2(a) Service Restrictions for Lethal Service Applications provide additional assurance that liquid sodium containing vessels have sufficient and well characterized margin to failure by requiring full penetration pressure boundary weld and confirming weld quality per UW-11(a) full radiography, and requiring visual inspection after hydrostatic / pneumatic tests. The use of flanges is restricted, but if used, special limits for rigidity and bolt spacing apply.
Crevices which may be susceptible to moisture retention leading to caustic stress corrosion cracking are minimized by disallowing permanent backing strips and ensuring the sodium facing surface is fully fused. Enhanced pressure boundary weld inspection provides additional confidence the welds are free of defects and have achieved full penetration.
Quality assurance for materials procurement via CMTR provides additional confidence the base metal and weld metal material properties are correct.
NAT-13478 Rev. 0 Natrium Demonstration Plant Long-Lived Passive Structural Materials of Construction Selection and Development Page 68 of 68 Verify Current Revision SUBJECT TO DOE COOPERATIVE AGREEMENT NO. DE-NE0009054 Copyright 2025 TERRAPOWER, LLC ALL RIGHTS RESERVED END OF DOCUMENT