ML25302A448
| ML25302A448 | |
| Person / Time | |
|---|---|
| Site: | Kemmerer File:TerraPower icon.png |
| Issue date: | 11/21/2025 |
| From: | Office of Nuclear Reactor Regulation |
| To: | TerraPower, US SFR Owner |
| References | |
| EPID L-2024-CPS-0000 | |
| Download: ML25302A448 (1) | |
Text
OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION US SFR OWNER, LLC -
SUMMARY
REPORT FOR THE GENERAL REGULATORY AUDIT OF THE KEMMERER POWER STATION UNIT 1 CONSTRUCTION PERMIT APPLICATION
SUMMARY
OF AUDIT OBSERVATIONS RELATED TO MATERIALS PERFORMANCE This enclosure summarizes key aspects of the Kemmerer Power Station Unit 1 (KU1) construction permit (CP) application general audit discussions related to materials performance that are not described in the specific audit questions in Enclosure 1 of the audit summary report.
On October 16 and 17, 2024, the U.S. Nuclear Regulatory Commission (NRC) staff held in-person audit discussions with US SFR Owner, LLC (USO) at TerraPower, LLC offices located in Rockville, MD. These discussions focused on the following primary topics:
how the reliability and integrity management (RIM) program is implemented in the KU1 design; preliminary design information with respect to materials of construction and environmental conditions; selection of codes and standards and special treatments for safety-significant structure, system, and components (SSCs);
plans for a degradation mechanism assessment (DMA) for KU1; plans for materials testing programs; and the approach to qualify and address environmental compatibility for safety-significant SSCs.
This in-person audit meeting also included discussion of several specific audit questions related to materials performance, specifically audit questions 5-6, 5-7, 6-8, 6-9, 7-8, 7-10, 7-11, 7-12, and 7-14. The outcomes from the audit meeting included an enhanced understanding of the KU1 materials qualification and through-life performance approaches, the identification of information that needed docketing to support the CP application, and the identification of information that can be reasonably left for further consideration at the operating license (OL) stage. This effort later resulted in the docketing of a supplemental report to the CP application, NAT-13478 (ML25274A130).
SUMMARY
OF NRC OBSERVATIONS RELATED TO MATERIALS QUALIFICATION INFORMATION Staff observations on the preliminary DMA screening criteria from section 7 of NAT-13478 and supporting technical basis documents, which were provided in response to audit question 7-138, are documented below. As discussed in the SE, details regarding the screening criteria can reasonably be left for later consideration at the OL stage with the final RIM program. Therefore, the observations below do not represent final staff positions and are based on information provided during the CP application review and audit. The development of the RIM screening criteria is also part of a research and development item identified in preliminary safety analysis report chapter 13 on assuring adequate in-service materials performance.
The staffs review at the preliminary design stage focused on the criteria for screening out a mechanism as not being active over the 60-year planned operating life to support the evaluation at the CP stage of how the preliminary design is consistent with principal design criteria (PDC) 4 on environmental effects. Based on this focus, the staff reviewed the preliminary screening
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION criteria and supporting technical basis documents to evaluate whether the preliminary criteria were supported by available technical literature data and operating experience.
Degradation Mechanism Observations on the Basis/Criteria/Discussion in Audit Documents Neutron Irradiation Effects (IE)
The RIM strategy for neutron IE is to ((
))
IE-1: Using the lower bound curve from NUREG/CR-7027 for fracture toughness properties appears reasonable for irradiation embrittlement only. Regarding the basis for the assumed flaws for analysis in table 6 of NAT-15939, the staff notes additional justification may be reviewed at the OL stage depending on the monitoring and non-destructive examination (MANDE) used to identify flaws.
IE-2: NAT-15939 provides limited basis for the evaluation of irradiation effects on creep and fatigue properties of weld metals (particularly the 16-8-2 weld metal planned for use in many KU1 SSCs). A description of new data or monitoring to be performed to address irradiation effects on weld metal properties at elevated temperatures may be reviewed by the staff at the OL stage.
IE-3: Section 7.4 of NAT-15939 discusses irradiation effects on strain limits and appears to indicate ((
)) The staff considers it is reasonable to ((
)) to ensure adequate creep ductility.
However, table 4 from NAT-15939 shows that ((
)). The following items may be reviewed at the OL stage:
The approach to address creep embrittlement (loss of creep ductility) for components exceeding the 1.5-2.5 dpa range as a function of temperature above 400°C.
Justification for use of a model designed and empirically derived for a different temperature range than the one of interest. Specifically, Table 4 uses predicted total elongation values from the model provided in MRP-135 as a check on these limits.
However, MRP-135 states that use of the material model is limited to typical
[pressurizer water reactor (PWR)] conditions as described in section 2. The reactor vessel internal components in a typical PWR are generally exposed to temperatures that range from 270°C to 370°C The MRP-135 model is focused on PWR conditions at lower temperatures than KU1 and figures 3-8 and 3-9 presenting the empirical elongation model results are cut off at 400°C.
IE-4: Section 7.5.1 of NAT-15939 evaluates irradiation effects on fatigue response based on data from 2 references. Reference [15] from NAT-15939 is focused on low cycle fatigue (does not address high cycle fatigue) and states that fatigue data on irradiated LMFBR [liquid metal fast breeder reactor] materials are limited. The remainder of section 7.5.1 relies on a subset of the data (~30 data points) from reference [15] to conclude that using the design fatigue curves from III-5 with no modifications is appropriate. The following may be reviewed at the OL stage:
The justification for using the design fatigue curves in III-5 with no modifications. NAT-15939 states the data fall above the design curve and are typically near the mean
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION curve. This applies for the data presented in figures 2 and 3 of NAT-15939; however, not for data from figure 4 of NAT-15939 for 304 stainless steel (SS) at 930°F that are generally below the mean curve.
The approach and basis for addressing irradiation effects on high cycle fatigue. The staff notes low cycle fatigue is discussed in Section 7.5.1 of NAT-15939.
IE-5: Sections 7.3 and 7.5.2 of NAT-15939 evaluate irradiation effects on time-dependent allowable stresses and creep and cite data from references [10] and [11] in NAT-15939. The staff notes that the data cited from reference [10] is limited and tested at a higher temperature than the KU1 operating conditions. ((
)) However, reference [11] states for uniaxial tests the reduction in rupture life at 538
°C was about a factor of 1/20 and for biaxial tests at 538 °C, the reduction in rupture life is about a factor of 1/40. The staff notes that tests at 538°C are on the higher end of KU1 operating conditions but of the data presented in reference [11] from NAT-15939 are the closest to KU1 temperatures. The following may be reviewed at the OL stage:
The basis for conclusions using the data in references [10] and [11] from NAT-15939; particularly, when supported by limited data that could suggest a much higher reduction factor on irradiation effects on creep rupture life.
Confirmation regarding the reference for figure 1 from NAT-15939; specifically, reference [10] is cited but it does not appear to contain such a figure.
IE-6: The staff notes that NAT-15939 does not directly address irradiation effects on creep-fatigue resistance. Reference [16] from NAT-15939 states that irradiation to fluences of 0.17 to 6.1 x 1021 n/cm2, E > 0.1 MeV (450°C) resulted in a substantial reduction in the creep-fatigue resistance of these stainless steels. The approach to addressing irradiation effects on creep-fatigue resistance, including consideration of the results from reference [16] of NAT-15939, may be reviewed at the OL stage.
IE1 - Neutron irradiation embrittlement IE1-1: NAT-8503, Revision (Rev.) 1 acknowledges that NUREG/CR-7027 cautions about combined effects of thermal aging embrittlement (TE) and irradiation embrittlement for the suggested 0.3 dpa value for SS welds but does not appear to consider it further. The staff observes that justification may be reviewed at the OL stage on how this combined effect is addressed in the DMA/RIM process.
IE2 - Neutron irradiation induced void swelling The staff may review the following at the OL stage regarding the basis and use of a 1% swelling threshold:
- IE2-1: Justification for using data related to loss of strength to provide confidence that embrittlement effects due to void swelling are not significant below 1%. Specifically, severe embrittlement at 5-8% swelling seen in the cited data is likely preceded by significant loss of toughness at lower swelling levels.
- IE2-2: Justification for how the 1% screening threshold addresses other impacts of void swelling, such as distortion or other swelling-induced stresses (e.g., stresses from constrained component geometries).
IE-3: How the non-conservative nature of the flux effect is addressed in this criterion. For the swelling correlation used, the underlying data is primarily from fast reactor components with higher neutron fluxes than the Natrium core supports will see. It has been shown that lower fluxes/dpa rates generate higher swelling at a given dpa (see NUREG/CR-7027
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION (ML102790482) section 4 and associated references), which could make these correlations non-conservative for the Natrium core supports.
IE3 - Neutron irradiation stress relaxation in bolting IE3-1: The staff notes this mechanism has a preliminary criterion without a basis. The staff may review this basis at the OL stage, including the applicable materials and environments (e.g., whether all bolting is Alloy 718 and where it is located in the reactor vessel).
IE4 - Neutron irradiation effect on creep strength and creep embrittlement IE4-1: Justification may be reviewed at the OL stage regarding plans to assess sodium effects on creep embrittlement for which NAT-13921 indicates criteria cannot be developed for. NAT-13921 notes that thin-walled materials are likely impacted at all temperatures, while thick-walled material may behave differently but need further data to support determining a threshold thickness.
IE4-2: The data used to support the evaluation of sodium effects on creep properties is generated in greatly accelerated conditions compared to the 60-year design life of KU1. At the OL stage, the staff may review how this RIM criterion accounts for uncertainties or potential non-conservatisms in the sodium effects on creep properties mechanism, due to accelerated testing. For example, it is possible that longer time exposures to sodium may lead to further effects on creep properties extending to lower temperatures or thicker materials.
IE4-3: Since weld metals are not discussed in NAT-13921, at the OL stage the staff may review if data was collected to evaluate irradiation effects on creep properties of weld metals.
IE4-4: Data from the 1973 ASTM publication appears to be primarily generated at higher irradiation and test temperatures than the Natrium SSCs will operate in. At the OL stage the staff may review how these differences in irradiation and test temperatures are addressed.
IE4-5: The staff may review at the OL stage the basis for the fluence criteria of 0.5 dpa cited in NAT-13478.
IE4-6: According to figures 4.2.2-5 through 4.2.2-8 in reference 5-1 of NAT-13921 Rev. A, creep ductility is significantly reduced under irradiation, with values falling well below 5% creep strain at a fluence near 1x10²¹ n/cm².
This is lower than the 5% maximum strain accumulation limit specified in III-
- 5. The staff notes that in the Japanese design codes for the Monju reactor used a 10% end-of-life tensile ductility as an acceptance criterion (Asayama et al.1). At the OL stage the staff will review the basis for creep ductility and creep strength values considered to be acceptable.
IE4-7: The staff may review at the OL stage the justification for why dose rates can be assumed to have a negligible effect at fluences below ((
))
1 Asayama et al., Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors, ASME Journal of Pressure Vessel Technology, Vol. 123, pp. 49-57, 2001.
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION IE4-8: NAT-13921 discusses sodium and irradiation effects on creep properties (individually and combined). At the OL stage the staff may review why a separate mechanism is not identified under the sodium effects DMA for sodium effects on creep properties.
IE4-9: At the OL stage the staff may review the basis for the criteria for the combined effect of sodium and irradiation on creep strength. Specifically, the study cited that looked at sodium and irradiation effects separately performed all testing above 600°C, providing limited basis for conclusions on behavior at temperatures that may be experienced by KU1 SSCs.
High Temperature Effects HT1 - TE The staff notes that this mechanism has a preliminary criterion without a basis. The staff also notes that TE could apply to all SS welds operating at elevated temperatures and would be more significant than is seen at light water reactor (LWR) temperatures. For the reactor vessel and intermediate heat transport system piping, TE would need to be considered for flaw tolerance and leak-before-break analyses. For the core supports and other irradiated components, TE would need to be considered in combination with irradiation embrittlement and void swelling for reducing toughness. The staff may review the following regarding TE at the OL stage:
HT1-1: The basis for the use of criteria from MRP-1752 given it is for lower temperature applications, which is non-conservative to the KU1 operating conditions for TE.
HT1-2: How recent experience with long-term thermal aging at LWR temperatures (which as noted are less severe than KU1 conditions) was factored into this criterion. Specifically, the staff notes there are numerous additional data sources to be considered, including NUREG/CR-6428, Rev.
1 (ML18222A161) and a recent NRC report summarizing knowledge on thermal aging of austenitic SS welds (ML24180A123). These reports indicate that the welding process may have a significant impact on unaged and thermally aged weld toughness. Of particular note is recent testing of unirradiated or very low fluence weld materials that shows significant embrittlement with ferrite levels 7% or less3 at lower temperatures than Natrium will operate.
HT1-3: If cast austenitic stainless steel (CASS) is used in the KU1 design, how the criteria account for TE on CASS components.
2 Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism. Screening and Threshold Values (MRP-175, Revision 1) - EPRI Report 3002010268, 2017.
3 Chen,Y., et al. Environmentally Assisted Cracking and Fracture Toughness of an Irradiated Stainless Steel Weld, 19th International Conference on Environmental Degradation of Materials in Nuclear Power SystemsWater Reactors, Boston, MA, August 18-22, 2019.
Hiser, M., P. Purtscher, R. Tregoning, NRC Technical Assessment of Zorita Materials Testing Results, Research Information Letter (RIL) 2022-05. U.S. Nuclear Regulatory Commission (NRC), May 2022 (ML22132A039).
Chen,Y., et al. Crack Growth Rate and Fracture Toughness of Stainless Steel Welds in Light Water Reactor Environment, TLR-RES/DE/MEB-2025-02, July 2025 (ML25204A037).
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION HT2 - Stress Relaxation Cracking (SRC)
HT2-1: The temperature screening criterion aligns with findings in literature, however, the thickness screening criterion does not appear to be clearly justified. While it is well established that thinner sections are generally less susceptible to SRC, there is no well-defined thickness threshold that clearly distinguishes between susceptible and non-susceptible conditions. At the OL stage the staff will review the basis for the weld thickness threshold to preclude the possibility of SRC.
HT2-2: At the OL stage the staff may review the results of SRC testing as described in section 9 of NAT-13478. The staff notes screening criteria that do not consider the effect of stress concentration and triaxiality of component geometry on the weld may not account for the SRC susceptibility of some welds.
HT2-3: At the OL stage the staff may review the justification for how the testing done demonstrates SRC is manageable based on these criteria. The staff note there is no standard test method to screen for SRC. SRC is significantly influenced by welding residual stress. If the optimized welding process is developed for low-constrained component geometry, the weld residual stress will be much higher when the optimized welding process is applied to a component geometry with high constraint and that leads to SRC susceptibility of the weld.
HT2-4: At the OL stage the staff may review how the effects of high weld residual stress in both original and repaired welds are addressed and if there are plans on using post weld heat treatment to reduce weld residual stress.
HT2-5: At the OL stage the staff expects to review if the RIM approach accounts for scenarios in which fabrication controls (e.g., welding process parameters) do not sufficiently minimize the risk of SRC.
HT3 - Creep Embrittlement Creep Embrittlement: metallurgical changes under temperature and stress that cause sharp reductions in the total elongation at rupture (low creep ductility) due to thermal creep.
The ASME III-5 HBB primary load design rules in the creep regime are based on stress limits obtained from extrapolation of shorter-term data.
However, it is tacitly assumed that there is adequate through-life creep ductility to allow the extrapolated rupture stress to be realized (i.e., material does not fail prematurely due to low creep ductility before the extrapolated rupture life for the applied stress is reached.)
If a component is in the low temperature regime (i.e., within Section III, Division 1, or the III-5 negligible creep conditions are met), the staff notes it is reasonable to conclude that creep embrittlement, or low creep ductility, would not be an issue.
HT3-1: The proposed criteria for the DMA scoring tied to the margin for creep damage does not address the issue of low creep ductility discussed above. This issue is fundamentally related to extrapolation of creep properties in ASME III-5 allowable stresses due to extended time at elevated temperature. Given the number of safety-significant SSCs that will screen in
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(>0 score) and the importance of ensuring sufficient creep ductility, at the OL stage the staff may review the technical basis for appropriate criteria to assess and manage creep embrittlement, including the role of potential performance monitoring approaches.
Sodium Effects SE-1: There is limited data and experience on sodium effects on 16-8-2 weld metals. At the OL stage the staff may review the basis for concluding that the weld material will behave like base metal.
SE-2: At the OL stage the staff may review how the combined effects of sodium exposure and other degradation mechanisms are considered. Specifically, while individual degradation may not be significant, it could be significant when combined with other degradation. As illustrated in a paper (K. Natesan, JNM, 2009) on sodium effects on mechanical performance, sodium exposure has nearly no effect on creep rupture properties of type 316SS; however, when combined with neutron irradiation, sodium environment can cause significant reduction in rupture strength of type 316SS.
SE1 - General, galvanic, and flow accelerated corrosion SE1-1: NAT-8194, Rev. B acknowledges the potential for multiple corrosion processes to occur simultaneously and create non-linear corrosion rates (see reference [4] from NAT-8194). At the OL stage the staff may review the basis for the conclusion that a single corrosion rate is acceptable.
SE1-2: At 1100°F, 2 ppm oxygen, and a flow rate greater than 10 feet per second (fps), the Nuclear System Materials Handbook gives a general corrosion rate for 316SS in liquid sodium of 0.0595 x 2 mils/yr = 0.1190 mils/yr. At the OL stage the staff may review how this information was considered in choosing the general corrosion rate of ((
))
SE1-3: Reference [3] from NAT-8194, Rev. B includes several other models with significant variations in results. The staff notes that the model chosen appears to be overly sensitive to oxygen content relative to the data presented (overpredicts corrosion rate at higher oxygen levels and underpredicts corrosion rate at lower oxygen levels). At the OL stage the staff may review how the data and models presented in reference [3] were considered in selecting the corrosion model.
SE2 - Mass
- Transfer, Carburization, and Decarburization SE2-1: Table 4.5-1 from NAT-13478 indicates a sodium chemistry impurity target for carbon content of 0.7 weight parts per million (ppm) or less. At the OL stage the staff may review why KU1 is expected to have a lower carbon content in sodium than past SFRs.
SE2-2: Data from table 6-1 of NAT-8194, Rev. B is described as the basis for concluding that carburization and decarburization are expected to have negligible effects for components thicker than 0.098 in (2.5 mm) as noted in Table 7-1 from NAT-13478. The staff notes that reference [13] is the source of the data in table 6-1 and that reference indicates these calculations are based on exposure to sodium containing 0.08 ppm carbon, which is much lower than the 0.7 wppm carbon expected for KU1 based on Table 4.5-1 from NAT-13478. At the OL stage the staff may review information to address the following questions: How does the data in table 6-1 support the conclusion that a thickness threshold of 0.098 in (2.5 mm) from Table 7-1 in NAT-13478 is appropriate to ensure the effects of carburization and
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION decarburization are negligible for 60 years of operation? What ratio of carburization or decarburization depth to component thickness is considered to be negligible?
SE2-3: Carburization is known to occur in components at lower temperatures within the system, while decarburization occurs in components at higher temperature. At the OL stage the staff may review information to address the following questions:
Why do the preliminary scoring criteria only consider susceptibility to occur at higher temperatures and not also lower temperatures?
How is the fuel cladding (the highest temperature location in the primary system) considered for decarburization that may lead to carburization on structural materials elsewhere in the system?
SE3 - Liquid Metal Embrittlement SE3-1: At the OL stage the staff may review how the data provided in the technical basis supports the conclusion that liquid metal embrittlement is not active for Types 304 and 316 SS over a 60-year life as noted in Table 7-1 from NAT-13478. Specifically, data from table 4-5 of NAT-11764 show significant decreases in tensile elongation after about 6,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of exposure to clean sodium. It was observed in a type 316, 0.06 thick specimen, when exposed to non-carburizing sodium at 1200°F, the total elongation decreased by 34%. For a type 304 specimen in the same conditions, the total elongation decreased by 12%. Room temperature data for those same sodium exposure conditions showed similar decreases in elongation. KU1 has a design life nearly 2 orders of magnitude greater than this exposure time.
SE3-2: NAT-11764 notes that a literature survey did not find any data on 16-8-2 weld metal, 800H or Alloy 718, and concludes ((
)) At the OL stage the staff may review the testing or performance monitoring planned to validate this assumption to ensure this degradation mechanism does not excessively degrade material toughness.
SE4 - Stress Corrosion Cracking (SCC)
SE4-1: Operating experience has generally shown that various forms of stress corrosion cracking are unlikely if chemistry is properly maintained and water and moisture intrusion avoided. However, this operating experience has also shown that such intrusion, including moisture in air, can occur in practice. At the OL stage the staff may review the monitoring and controls strategy used to ensure significant moisture intrusion can be detected.
SE4-2: NAT-14011 indicates that slow strain rate tensile testing of 316H SS in sodium with oxygen levels representative of KU1 oxygen levels is planned to provide additional confidence in the technical basis for SCC in sodium. At the OL stage the staff may review information to address the following questions: how does this testing address conditions at higher oxygen levels to understand the margin and sensitivity associated with oxygen levels in sodium? How long will these tests last in terms of sodium exposure?
SE4-3: The basis for the SCC criteria relies heavily on operating experience, which has shown generally positive experience. However, it is not clear to
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OFFICIAL USE ONLY - PROPRIETARY INFORMATION what extent this operating experience includes components using the weld filler metals and nickel alloys included in the criteria. At the OL stage the staff may review if operating experience with the weld metals and nickel alloys is available to support applying the criteria to these materials.
SE5 - EAF The staff notes that this mechanism has preliminary criteria without a supporting basis document and will review this basis document once it is available at the OL stage.
Mechanical and Flow Effect MF1 - Fretting, Wear, Cavitation MF1-1: At the OL stage the staff may review how fretting, wear, and cavitation after longer-term sodium exposure are addressed.
Fatigue and Creep FC-1: At the OL stage the staff may review how irradiation effects on fatigue and creep mechanisms are addressed or managed.
FC-2: The staff notes that if design by inelastic method is used for any component, the technical basis for the selected inelastic analysis methods and relations used in a design analysis should be provided (see Section III, Division 5 HBB-Y-3600), particularly when irradiation effect is included in the inelastic model. At the OL stage the staff may review what inelastic analysis methods are used and the technical basis to support their use.
FC1 - Thermal Fatigue (stratification cycling and striping)
FC1-1: At the OL stage the staff may review the margin proposed for a DMA score of 0; specifically, how the margin is calculated from stress analysis and if it considers the maximum design-basis transients.
FC1-2: At the OL stage the staff may review information to assess if the understanding of the phenomena is based solely on simulations, if the understanding is also supported by experimental data, and if the analysis includes a sensitivity study.
FC1-3: There appears to be no monitoring planned for thermal stratification and striping. Experience from operating reactors and other SFRs have shown these mechanisms to be very sensitive to plant configuration and operating conditions with small changes leading to large effects. Therefore, in-service monitoring of thermal stratification and striping is important to assess fatigue and thermal stress risks in components where these effects are a concern. There is a significant amount of literature on monitoring for thermal striping to ensure this mechanism is managed in practice. Given the experience with this mechanism being difficult to predict solely through modeling and analysis, at the OL stage the staff may review if targeted monitoring was considered to provide confidence that reliability targets will be met.
FC1-4: Diercks (1984)4 states: Types 304 and 316 stainless steel have been chosen for the affected upper internal structure in several European LMFBRs. However, in the U.S. Fast Flux Test Facility and Clinch River Breeder Reactor Plant designs as well as in Japan's Monju Reactor, the magnitude (up to ~150°C) and expected number (~109) of the temperature fluctuations are judged to be too severe for these alloys. At the OL stage 4 Diercks, D.R., Structural Materials for Breeder Reactor Cores and Coolant Circuits, DOE/NBM 4009194, 1984. https://www.osti.gov/servlets/purl/5146967.
OFFICIAL USE ONLY PROPRIETARY INFORMATION 10 OFFICIAL USE ONLY - PROPRIETARY INFORMATION the staff may review how thermal fatigue and striping is analyzed and addressed for KU1, particularly in the reactor internals above the core. The staff notes that ASME BPVC design fatigue strain range curves (like figure HBB-T-1420-1A/B) do not extend beyond 106 cycles. The staff may review the approach to address very high cycle fatigue beyond 106 cycles.
FC2 - Thermal Fatigue (thermal transients)
Observations in FC3 Fatigue-creep interaction regarding irradiation effects on fatigue life and the proposed MANDE also apply to FC2.
FC2-1: At the OL stage the staff may review the basis for only considering an inelastic analysis method. The staff notes that if elastic analysis are used additional criteria would need to be developed.
FC3 - Fatigue-Creep Interaction FC3-1: The staff notes that the margin zones (figure 6.2 in NAT-13254, Rev.
A) considered for DMA scoring are reasonable when environmental effects are negligible. The staff notes however that under non-negligible irradiation, these zones should be revisited (see also IE-6 above). The limits on creep damage and fatigue damage fractions in the creep-fatigue interaction diagram may change significantly due to irradiation and other environmental effects. Although literature on the effect of neutron irradiation on fatigue and creep-fatigue life is limited, available data5 indicate a reduction in both fatigue and creep-fatigue life under neutron irradiation. Other literature6 suggests that creep-fatigue loading in sodium environment can lead to reduction in creep-fatigue life.
Under non-negligible irradiation conditions, the French code RCC-MRx multiplies the creep damage accumulation by a factor of 10 when comparing the fatigue and creep damage fractions with the allowable damage envelope in the creep-fatigue interaction diagram for austenitic steels. Another approach was used by the Japanese in their design code for fast breeder reactor7 where creep life reduction factors due to irradiation was developed from creep rupture tests of irradiated material. They are then used in the design procedure for the evaluation of creep damage to account for the neutron irradiation effect on creep rupture.
At the OL stage the staff may review how the KU1 RIM and design process account for irradiation conditions and other environmental effects on design properties for creep-fatigue design evaluation.
FC3-2: The staff notes that the FatiguePro system is being considered for creep-fatigue monitoring, which uses stress transfer functions, either analytical or reduced-order models based on finite element simulation results, to calculate real-time stress and temperature from plant instrumentation. It then evaluates creep and fatigue damage using these data, based on damage models derived from experimental campaigns. The 5 Messner et al., Identifying Limitations of ASME Section III Division 5 For Advanced SMR Designs,,
ANL-21/27, 2021. https://www.osti.gov/biblio/1804300.
6 K. Natesan, et al. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors, Journal of Nuclear Materials, Volume 392, Issue 2, 2009.
7 Asayama et al., Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors, ASME Journal of Pressure Vessel Technology, Vol. 123, pp. 49-57, 2001.
OFFICIAL USE ONLY PROPRIETARY INFORMATION 11 OFFICIAL USE ONLY - PROPRIETARY INFORMATION purpose of this system is to capture differences between actual operational transients and the transients assumed during design. However, it does not account for changes in material behavior due to environmental effects, nor does it quantify actual material damage. Even for components that experience negligible irradiation or other negligible environmental effects, the creep and fatigue damage models used by the monitoring system are based on data collected in the lab under fixed testing conditions (e.g., creep tests under constant stress and temperature, fatigue tests under constant strain range and temperature) and for durations much shorter than the design life of the component, which may not accurately predict actual in-service damage. While this system provides valuable real-time transient data, its effectiveness is limited if the models arent updated by direct measurement to monitor the damage. The staff notes that section 10 of NAT-13478 includes description of an enhanced creep-fatigue monitoring system being considered that would involve direct measurement of components to monitor damage and improve confidence in applying the transfer function approaches to other locations. At the OL stage the staff expects to review how this RIM strategy for creep-fatigue provides reasonable management of this mechanism.
FC4 - Deformation (stress relief)
The staff have no observations on the preliminary screening criterion at this time.
FC5 - Deformation (ratcheting)
Observations in FC3 Fatigue-creep interaction regarding irradiation effects on fatigue life and the proposed MANDE also apply to FC5.
FC5-1: At the OL stage the staff will review the meaning of the term neglect inelastic strains in the simplified analysis method.