Irradiation Effects on Ferritic / Martensitic Steels

요약:

The irradiation effect on nuclear reactor materials (fuel or structural component) depends mainly on the neutron flux, neutron energy or energy spectrum, irradiation time and temperature as well as the existing crystal and defect structure.
The irradiation was performed in the MISTRAL (Multipurpose Irradiation System for Testing of Reactor Alloys) irradiation rig of the BR2 reactor.

The irradiation effect on nuclear reactor materials (fuel or structural component) depends mainly on the neutron flux, neutron energy or energy spectrum, irradiation time and temperature as well as the existing crystal and defect structure, which will be deconstructed and reconstructed in the atomic scale during the irradiation process.

Irradiation swelling, irradiation creep and irradiation embrittlement are the three dominating effects to be considered, however irradiation effects on physical, thermal and mechanical properties should in any case be monitored with special care for the safe operation and maintenance of nuclear reactors.

Despite the fact that numerous steels have been developed that possess improved properties over HT9, FV448, EM-12, DIN 1.4914, and JFMS, some of these compositions are still under consideration for Generation IV reactors as most of these steels were developed for conventional fossil-fuel power plant applications.

Since the 1970s, when these steels were first considered for nuclear applications, improved steels for conventional power plant applications have been developed and although not as yet tested under irradiation, these steels need to be considered for nuclear applications. The evolution of the new steels will now be reviewed by considering the compositional changes that have led to improved microstructures and properties.

Table 1: Irradiation-induced phases in Fe-Cr alloys and high-chromium ferritic/martensitic steels

On the other hand, high Chromium Ferritic-Martensitic steels are candidate materials for the target spallation source and the fuel cladding tubes in the European demonstrator of the accelerator driven system (ADS). A.Almazouzi and E.Lucom present results obtained from the irradiation campaign that has been performed in the BR2 reactor within the SPIRE project to assess the mechanical behavior of several standard 9-12%Cr steels.

The irradiations have been performed up to 4.5 dpa at 200°C. A quantitative assessment of the effects of neutron irradiation, in terms of both hardening and embrittlement is presented, based on the comparison with the unirradiated condition.

Ferritic-Martensitic (F/M) steels (used as wrapper tubes in Phenix, as well as candidate materials for Fast and Fusion reactors), appear to be promising choices both for fuel cladding and structural applications. Within the EU-FP5-SPIRE project, conventional 9Cr and 12Cr steels (EM10, T91 and HT9) have been irradiated at 200°C in the BR2 reactor up to two different doses. Subsequently, tensile, Charpy and fracture toughness tests have been performed. In this paper, the main results are reported and briefly discussed. All steels have been irradiated in non-doped condition.

The chemical compositions of the three ferritic-martensitic steels are given in Table 2.

Table 2: Chemical composition of the investigated steels (weight %, Fe balance)

They were subjected to the following heat treatments:

  • EM10: normalisation at 990°C for 50’ and tempering at 750°C for 60’;
  • T91: normalisation at 1040°C for 60’ and tempering at 760°C for 60’;
  • HT9: annealing at 1050°C for 30’ and then tempering at 700°C for 2h.

The irradiation was performed in the MISTRAL (Multipurpose Irradiation System for Testing of Reactor Alloys) irradiation rig of the BR2 reactor at 200°C up to a dose of 4. 5dpa, in flowing water, and without tailoring spectrum.

In Table 3 the results obtained from each type of test for the investigated materials are given.

Table 3: Summary of the main mechanical test results obtained. σp02 is the yield stress; σUTS is the ultimate tensile stress; εt is the total elongation; DBTT is the ductile to brittle transition temperature obtained from Charpy tests either in terms of energy (DBTTKV) or shear fracture appearance (DBTTSFA), USE is the upper shelf energy obtained from Charpy tests and To is the reference temperature corresponding to a fracture toughness level of 100 MPa.m1/2 from the Master Curve analysis

In more qualitative terms, the most significant results are:

Tensile tests: EM10 exhibits the smallest yield stress increase and HT9 the largest. The hardening of T91 is practically the same at the two neutron doses; however, the hypothesis of saturation needs to be confirmed with additional tests at higher doses.

Impact tests: T91 shows a slightly larger irradiation embrittlement (DBTT shift and USE decrease) than EM10. HT9 exhibits the most pronounced embrittlement, as expected based on the higher Cr content (12%).

Fracture toughness tests: The same ranking is found as for impact tests (EM10 – T91 – HT9, from the least to the most irradiation-sensitive). The reference temperature shifts (from toughness tests) are systematically and significantly larger than DBTT shifts (from impact tests); this circumstance, which is not observed for other classes of steels (e.g. RPV steels), has important implications for both designers and safety authorities.

Another potential issue, somewhat related to the previous item, is the applicability of the Master Curve approach to high Cr F/M steels, which has been recently questioned by several researchers. Possible alternative data analysis methods have been recently proposed (SINTAP lower tail analysis procedure, multi-modal Master Curve etc).

It was observed that irradiation effects on the mechanical properties are quite similar when considering the two 9Cr steels. On the other hand, HT9 shows very pronounced hardening and embrittlement. Two potential issues have been identified in view of the plant design and more specifically safety related aspects: the magnitude of the irradiation-induced transition temperature shift, which is systematically larger when measured by means of toughness tests rather than estimated from Charpy tests, and the questionable applicability of the Master Curve approach to high Cr F/M steels.

기술 자료 검색

검색할 어구를 입력하십시오:

검색 범위

본문
키워드

머릿글
요약

물리적 특성은 Total Materia 데이터베이스 내 많은 재질에서 검토하실 수 있습니다.

데이터는 규격의 공식 정보와 Total Materia의 강력한 상호 참조 표를 통해 검색 가능하며 이는 물리적 특성 데이터 검색에 매우 효과적일 것입니다!

신속 검색에 검색할 재질명을 입력합니다. 원하신다면 국가/규격을 지정하신 후 검색 버튼을 클릭합니다.


관심 있는 재질을 선택하신 후, 물리적 특성 링크를 클릭하셔서 선택된 재질의 데이터를 검토합니다. 물리적 특성 데이터 기록의 개수는 링크 옆에 괄호 안에 표시됩니다.


물리적 특성은 원래 데이터 값에 따라 표시됩니다. 규격에 의한 공식 데이터는 공식 탭에서 찾을 수 있고, 다른 출처를 통해 검색된 재질의 데이터는 자신의 탭에 표시 됩니다.


유사 재질 탭에는 원래 재질과 비슷하며 물리적 특성이 포함된 재질을 표시합니다. 등가 재질 검색 시에 매우 유용할 수 있습니다!


일반 탭은 특성 데이터에 대한 일반적인 개요를 제공하며 추가 조사를 위한 유용한 출발점으로 사용될 수 있습니다.


Total Materia 데이터베이스를 사용해 보실 수 있는 기회가 있습니다. 저희는 Total Materia 무료 체험을 통해 150,000명 이상의 사용자가 이용하고 있는 커뮤니티로 귀하를 초대합니다.