Thermal creep modeling of HT9 steel for fast reactor applications

doi:10.1016/j.jnucmat.2010.12.232

Journal of Nuclear Materials

Volume 409, Issue 3, 28 February 2011, Pages 207-213

https://doi.org/10.1016/j.jnucmat.2010.12.232 Get rights and content

Abstract

The ferritic martensitic steel HT9 is a primary candidate material for the fuel cladding of liquid–metal-cooled fast reactors (LMFRs) owing to its excellent stability under irradiation. Thermal creep of fuel cladding is a potential life-limiting factor in the long-life fuel design of LMFRs. Using the measured data available in the literature, such as creep strain data, stress rupture data, and steady-state creep rate data, a generalized creep correlation was developed. The new correlation is based on the Garofalo equation and the modified Monkman–Grant equation, and it shows better agreement with the experimental data than existing correlations that use either the theta projection method or the minimum commitment method, making it more appropriate for use in long-life applications.

Introduction

Ferritic martensitic steels (FMS) composed of 9–12% Cr such as HT9 are often used as high temperature structural materials in coal power plants and gas power plants. FMS have a higher thermal conductivity and a lower void swelling under irradiation than austenitic stainless steels [1], [2]. Thus, HT9 is a primary candidate material for the fuel cladding material of advanced nuclear reactors such as liquid–metal-cooled fast reactors (LMFRs) owing to its excellent irradiation stability.

The nominal alloy composition of HT9 is given in Table 1. The development history of FMS with 9–12% Cr is listed in Table 2; it shows their rupture strength and maximum use temperature have been improved. Although HT9 belongs to the first generation FMS, which were developed as early as the 1960s, it is still a viable cladding material for LMFRs, small modular fast reactors (SMFRs), and advanced burner reactors (ABRs), mainly because of its proven performance records [3], [4].

The creep deformation of the cladding is not a critical performance issue for a typical LMFR fuel design provided with a smeared density of about 75% up to a burnup of about 20% [4]. However, for fuel designs that require a higher burnup, long life (e.g., 15–30 years of fuel life), or higher temperature, creep deformation may be a life-limiting factor. Hence, an improved creep correlation that is applicable to such design conditions is desired.

Although the thermal creep of HT9 has been studied for decades for both nuclear and non-nuclear applications [5], [6], [7], [8], [9], inconsistencies still exist in the correlations and the experimental data. In the present study, the test data available for the thermal creep tests of HT9 were assessed. A new generalized correlation, applicable to both typical- and long-life fuel designs, was developed for the performance evaluation of LMFR fuel cladding.

Section snippets

Existing correlations for creep strain

There are a few thermal creep correlations of HT9 steel that are available in the literature. The most widely accepted are the correlations developed by Amodeo and Ghoniem [5], [6] and by Lewis and Chuang [7].

The correlation by Amodeo and Ghoniem [5] employs the minimum commitment method (MCM) and uses time dependent creep strain curves at 873 K, as reported by Sandvik Steels [7]. The MCM is based on the observation that the 1% strain occurs during either the primary creep regime or at the

Creep strain and rupture stress correlations

A new creep correlation was developed in this study to overcome the limitations of the existing creep correlations for HT9 (i.e., MCM and TPM correlations). There are two significant aspects of the new correlation. Firstly, it should be noted that the phenomenological background of the existing correlations is insufficient to describe the creep deformation behavior of HT9 steels. Although the steady-state creep rates can easily be obtained from the literature, using typically available

Discussion

Thus far, our paper has focused on HT9 thermal creep correlations. However, since nuclear fuel cladding is commonly used in high energy neutron irradiation environments, it is important to consider the effect of irradiation. Irradiation displaces atoms from their equilibrium position to form vacancies and interstitials. These form voids and dislocation loops, thereby affecting creep deformation. The question arises as to what extent the irradiation enhancement influences the thermal creep.

Conclusions

The thermal creep strains of HT9 predicted using the correlations available in the literature have been found to be inconsistent. In some cases, they are also inconsistent with the measured data. In this work, a thermal creep correlation based on the Garofalo equation has been developed, and it has been shown to be applicable to both the primary and steady-state creep regimes. For the tertiary creep regime, an implicit model has been proposed by using the modified Monkman–Grant equation. The

Acknowledgments

This study was supported in part by Ministry of Education Science and Technology (MEST) of Korea, and in part by the UChicago Argonne, LCC as Operator of Argonne National Laboratory under Contract No. DE-AC-02-06CH11357 between the UChicago Argonne, LLC and the US Department of Energy.

References (19)

R.G. Pahl et al.
J. Nucl. Mater.
(1993)
G.L. Hofman et al.
Prog. Nucl. Energy
(1997)
R.J. Amodeo et al.
J. Nucl. Mater.
(1984)
R.J. Amodeo et al.
Nucl. Eng. Des. Fusion
(1985)
G. Lewis et al.
Fusion Eng. Des.
(1991)
P.J. Ennis et al.
Acta Mater.
(1997)
C. Phaniraj et al.
Scripta Mater.
(2003)
R.L. Klueh
Inter. Mater. Rev.
(2005)
F. Masuyama
Advanced Heat Resistant Steels for Power Generation

There are more references available in the full text version of this article.

Cited by (10)

Thermal creep analysis and correlation development for manufactured HT9 cladding
2024, Journal of Nuclear Materials
HT9 cladding was manufactured through successive cold working and heat treatments, and its creep strains were obtained at various stresses and temperatures in the range of 10–120 MPa and 843–953 K, respectively, for up to 20,000 h. Primary and secondary creep regimes were observed, whereas only a few specimens at extreme conditions experienced a tertiary creep regime and creep rupture. The reference creep models, including the theta projection and modified Garofalo equation for HT9 cladding, did not fit the experimentally obtained creep strain curves; the manufactured HT9 cladding showed higher creep resistance than the reference models. Hence, a new correlation was developed by deriving various constants, such as the activation energy, stress exponent, and material constants, from the measured creep strains. The creep correlation successfully predicted the creep behavior of HT9 cladding manufactured. Obtained extensive creep strains over a range of stresses and temperatures, closely approximate actual operating conditions, and developed correlation in this study are valuable in various applications using HT9, providing essential data for analyzing and estimating creep resistances of cladding during the operation of nuclear reactors.
Modern nanostructured ferritic alloys: A compelling and viable choice for sodium fast reactor fuel cladding applications
2020, Journal of Nuclear Materials
Citation Excerpt :
Overall, the tensile results on the high temperature annealed specimens suggest that the as-extruded specimens were relatively resistant to property deterioration for annealing conditions as high as 1150 °C (∼0.80 TM, where TM is the absolute melting temperature). The derived creep properties of OFRAC are compared to published results on HT-9 [59] and three stainless steels [60,61] in Fig. 4. The calculated values of the stress exponents for OFRAC and the published values for HT-9 and stainless steels are shown in the figure legends.
Sodium fast reactor cores present a truly challenging environment for the fuel cladding and core structural materials that limit achievable fuel burnup in these systems. Modern nanostructured ferritic alloys have the potential to deliver the desired high performance characteristics where historic austenitic and ferritic/martensitic alloys fall short. In this paper, a new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), is developed and demonstrated in cladding geometry with a length > 1 m. The alloy composition and microstructure are tailored to deliver dramatically improved strength and creep resistance. At 600 °C the ultimate tensile strength is ∼600 MPa and the stress to induce a steady state strain rate of 10⁻⁶ s⁻¹ is 500 MPa vs. 200 MPa for traditional ferritic/martensitic alloys. A very high defect sink density was engineered in these alloys that is expected to further enhance the traditional good swelling and irradiation creep resistance of ferritic/martensitic steels. These significant improvements along with the demonstrated viability for seamless cladding production indicate that this alloy is a compelling and viable choice for sodium fast reactor fuel cladding applications.
Enhancing FRAPCON fuel performance code for physical phenomena at high temperature and high burnup
2019, Journal of Nuclear Materials
Citation Excerpt :
Although, there is definitely a degree of simplification associated with this approach, retaining this cesium fuel swelling even at high temperature would add more conservatism to overall fuel performance analyses. In addition, the effect of stoichiometry i.e. oxygen-to-metal (O/M) ratio could play a role in cesium migration [36,42–44]. While the O/M impact was not explicitly taken into account, it has been indirectly taken into account by the thermal conductivity correlation of MOX fuel used during the simulation which included O/M ratio as one of the most significant contributing factor in its predicted thermal conductivity.
Advanced LWR fuel designs aim for higher burnup and heating rates. Operating at higher fuel temperatures and burnups involve several physical phenomena typically ignored in LWR fuel performance codes. To accurately analyze the fuel rod behavior at such conditions, the modification of existing LWR fuel performance code is necessary. This paper describes the enhancement of FRAPCON, a fuel performance code used by US NRC. In this study, mechanistic models describing porosity, cesium and hydrogen migration and precipitation kinetics were added into FRAPCON-3.5, referred to as FRAPCON 3.5 Enhanced Performance (EP). A survey of MOX fuel thermal conductivity models vs. experimental data was performed and multiplying factors to account for plutonium weight fraction were derived. Overall, the comparison of these revised models showed reasonable agreement with in-pile experiments.
Tensile and creep behaviour of modified 9Cr-1Mo steel cladding tube for fast reactor using metallic fuel
2014, Procedia Engineering
Modified 9Cr- 1Mo ferritic martensitic steel has been considered as cladding tube material in metallic fuel fast reactors in view of the relatively lower operating temperatures. In this study, tensile and creep behaviour of modified 9Cr-1Mo steel cladding tube have been investigated. Microstructure of the normalized and tempered steel consisted of tempered lath martensite with precipitates at the prior austenite grain boundaries and sub-boundaries. Tensile tests on the cladding tube were carried out at a strain rate of 3x10^-3 s^-1 over the temperature range of 300 - 923 K. The variations of 0.2% yield stress, ultimate tensile strength and elongation of the steel with temperature have been studied. Yield stress and ultimate strength of the cladding tube exhibited plateau in the intermediate temperature range of 523 - 673 K, where the elongation exhibited a broad minimum. The tensile strength and ductility of the steel cladding tube were comparable with those reported for different products forms of the steel.
Creep properties of modified 9Cr-1Mo steel cladding tube were studied at 823 K at various stress levels. Creep curves of the steel generally consisted of primary and tertiary stages with no secondary stage of deformation. The variation of minimum creep rate with stress obeyed a power law. The stress exponent ‘n’ was around 25, which is the characteristic of precipitation hardened alloys. Rupture life was found to decrease with increase in stress. The Monkman-Grant relationship relating minimum creep rate with rupture life was found to be obeyed by the steel. Creep rupture strength of the modified 9Cr-1Mo steel cladding tube was in accordance with the reported values for other product forms.
Feasibility study of fuel cladding performance for application in ultra-long cycle fast reactor
2013, Journal of Nuclear Materials
Citation Excerpt :
Finally, note that the empirical treatment has been developed only for fission gas release, whereas the swelling behavior of the fuel is simulated with the GRSIS model, even when the fission gas release correlation is selected. It is known that creep deformation of cladding is not critical in typical liquid metal fast reactors with a low burnup of about 20%, but in the case of high burnup SFRs, such as in UCFRs operating for 30–60 years, thermal creep deformation needs to be considered for an evaluation of the UCFR performance [30,31]. There exist several models of thermal creep, namely, those using the minimum commitment method (MCM) [32], the theta projection method (TPM) [33], and the new correlation, which is based on the Garofalo equation for thermal creep behavior modeling of HT-9 [31].
As a part of the research and development activities for long-life core sodium-cooled fast reactors, the cladding performance of the ultra-long cycle fast reactor (UCFR) is evaluated with two design power levels (1000 MWe and 100 MWe) and cladding peak temperatures (873 K and 923 K). The key design concept of the UCFR is that it is non-refueling during its 30–60 years of operation. This concept may require a maximum peak cladding temperature of 923 K and a cladding radiation damage of over 200 dpa (displacements per atom). Therefore, for the design of the UCFR, deformation due to thermal creep, irradiation creep, and swelling must be taken into consideration through quantitative evaluations. As candidate cladding materials for use in UCFRs, ferritic–martensitic (FM) steels, oxide dispersion strengthened (ODS) steels, and SiC-based composite materials are studied using deformation behavior modeling for a feasibility evaluation. The results of this study indicate that SiC is a potential UCFR cladding material, with the exception of irradiation creep due to high neutron fluence stemming from its long operating time of about 30–60 years.
Modified Monkman-Grant relationship for austenitic stainless steel foils
2013, Journal of Nuclear Materials
Characteristics of creep deformation for austenitic stainless steel foils are examined using the modified Monkman–Grant equation. A series of creep tests are conducted on AISI 347 steel foils at 700 °C and different stress levels ranging from 54 to 221 MPa. Results showed that at lower stress levels below 110 MPa, the creep life parameters ${\dot{ε}}_{\min}, ε_{r}, t_{r}$ can be expressed using the modified Monkman–Grant equation with exponent m′= 0.513. This indicates significant deviation of the creep behavior from the first order reaction kinetics theory for creep (m′ = 1.0). The true tertiary creep damage in AISI 347 steel foil begins after 65.9% of the creep life of the foil has elapsed at stress levels above 150 MPa. At this high stress levels, Monkman–Grant ductility factor $λ^{'}$ saturates to a value of 1.3 with dislocation-controlled deformation mechanisms operating. At low stress levels, $λ^{'}$ increases drastically ( $λ^{'} = 190$ at 54 MPa) when slow diffusion-controlled creep is dominant.

View all citing articles on Scopus

View full text

Thermal creep modeling of HT9 steel for fast reactor applications

Abstract

Introduction

Section snippets

Existing correlations for creep strain

Creep strain and rupture stress correlations

Discussion

Conclusions

Acknowledgments

J. Nucl. Mater.

Prog. Nucl. Energy

J. Nucl. Mater.

Nucl. Eng. Des. Fusion

Fusion Eng. Des.

Acta Mater.

Scripta Mater.

Inter. Mater. Rev.

Advanced Heat Resistant Steels for Power Generation