Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors

Asayama, Tai; Abe, Yasuhiro; Miyaji, Noriko; Koi, Mamoru; Furukawa, Tomohiro; Yoshida, Eiichi

doi:10.1115/1.1338119

Cited by 26 publications

(15 citation statements)

References 12 publications

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“…Furukawa et al (1998) reported tensile data on 316FR pre-exposed to sodium, and creep, fatigue, and creep-fatigue results on 316FR in flowing sodium. It was found that tensile properties of 316FR were not affected by sodium exposure at 500-550°C for 5000 h. Creep rupture tests conducted at 550-600°C in September 15, 2008 60 flowing sodium at the oxygen level of 1 ppm showed that creep rupture strength in sodium was slightly lower than that in air with predominant surface grain boundary cracking when the rupture time was between 1000 and 10,000 h; at the lower stress level and longer rupture time, sodium exposure had no influence on the creep rupture strength of 316FR, as shown in Figure 4.59 (Asayama et al 2001). …”

Section: Effects Of Sodium Exposurementioning

confidence: 96%

“…316LN-IG has higher allowable stresses than typical 316LN due to an optimal combination of the alloying elements such as carbon, nitrogen, nickel, chromium, manganese and molybdenum, with a tight specification of their allowable composition range. A comprehensive database has been established on 316LN-IG, including heat-to-heat variations, the effect of product size, fracture toughness, effects of neutron irradiation, etc., and the data can be found in the ITER Materials Properties Handbook ( The chemical compositions of 316LN, 316LN-IG, and 316FR stainless steels are given in Table 4-11 (Asayama 2001, Tavassoli 1995, Brinkman 2001. Significant data base of tensile, creep, fatigue and creep-fatigue properties of 316FR and 316LN base metal and weld metal can be found in the RCC-MR, ITER MPH and open literature.…”

Section: Fr and 316ln Stainless Steelsmentioning

confidence: 99%

“…Studies on the effects of sodium exposure have been done primarily on 316FR by the Japanese fast reactor program (Asayama et al 2001). They discussed the material degradation issues in sodium environment and the evaluation procedures to account for the sodium effects.…”

Section: Effects Of Sodium Exposurementioning

confidence: 99%

“…The emphasis of the 316LN-IG research was placed on the lower irradiation temperature behavior due to the low operating temperature (≤300°C) of 316LN-IG in the ITER components (Tavassoli 1995). Limited data of neutron-irradiated Japanese 316FR can be found in the open literature (Miyaji et al 1999, Asayama et al 2001). Tavassoli (1995) summarized the irradiation effects on tensile properties of 316LN-IG.…”

Section: Effects Of Neutron Irradiationmentioning

confidence: 99%

“…Similar effects of neutron irradiation on tensile properties were observed in the weld metal. Since the yield stress and the ultimate tensile strength are not degraded by neutron irradiation, no design factors are necessary (Asayama et al 2001).…”

Section: Effects Of Neutron Irradiationmentioning

confidence: 99%

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Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package "Advanced Materials Code Qualification".The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors.Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue in...

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Section: Effects Of Sodium Exposurementioning

confidence: 96%

Section: Fr and 316ln Stainless Steelsmentioning

confidence: 99%

Section: Effects Of Sodium Exposurementioning

confidence: 99%

Section: Effects Of Neutron Irradiationmentioning

confidence: 99%

Section: Effects Of Neutron Irradiationmentioning

confidence: 99%

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Code qualification of structural materials for AFCI advanced recycling reactors.

Natesan¹,

Li²,

Nanstad³

et al. 2012

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Environmental Damage in Nuclear Plants

Hoffelner¹

2012

Materials for Nuclear Plants

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Low cycle fatigue properties of modified 9Cr-1Mo ferritic martensitic steel weld joints in sodium environment

Sandhya

Kannan

Ganesan

et al. 2010

Trans Indian Inst Met

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Mod.9Cr-1Mo ferritic-martensitic steel is the material chosen for the steam generator of the Prototype Fast Breeder Reactor being built at Kalpakkam, India. The use of sodium as a heat transfer medium for Liquid Metal Fast Breeder Reactors (LMFBRs) necessitates a comprehensive understanding of the effects of dynamic sodium on the Low Cycle Fatigue (LCF) behaviour of structural components. Moreover welds being the weak links in any structure, it is necessary to evaluate the LCF behaviour of joints in sodium environment, more so in Mod.9Cr-1Mo steel because of the well established Type -IV cracking in this material. With this aim in view, a programme has been initiated to evaluate the LCF properties of weld joints of this steel in dynamic sodium environment. A facility has been developed in-house for mechanical property evaluation in dynamic sodium. LCF tests conducted in flowing sodium environment at 823 and 873 K showed a similar trend in cyclic stress response in air and sodium environments exhibiting a continuous cyclic softening behaviour. The fatigue lives are significantly improved in sodium environment when compared to the data obtained under identical testing conditions in air environment. The lack of oxidation in sodium environment is considered to be responsible for the delayed crack initiation and consequent increase in fatigue life. Comparison with RCC-MR code shows that the design curve based on air tests is conservative.

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Evaluation Procedures for Irradiation Effects and Sodium Environmental Effects for the Structural Design of Japanese Fast Breeder Reactors

Cited by 26 publications

References 12 publications

Code qualification of structural materials for AFCI advanced recycling reactors.

Code qualification of structural materials for AFCI advanced recycling reactors.

Environmental Damage in Nuclear Plants

Low cycle fatigue properties of modified 9Cr-1Mo ferritic martensitic steel weld joints in sodium environment

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