This paper describes the premise for the standardization of Leak Before Break (LBB) assessment procedure applicable to Japanese Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr-1Mo steel. For the early commercialization of the SFRs, economic competitiveness is one of the most essential requirements. Japan Atomic Energy Agency (JAEA) proposes an attractive plant concept and studies the applicability of some innovative technologies to the plant. One of the most practical means to reduce the construction costs is to reduce the total amount of structural materials. To meet the requirements, modified 9Cr-1Mo steel has attractive characteristics as a main structural material of SFRs, because the steel has both excellent thermal properties and high temperature strength. By employing the steel as the main structural material, remarkably compact plant design can be achieved. Since the ductility and toughness of the steel is relatively inferior to those of conventional austenitic stainless steels, a LBB assessment technique suitable for the pipes made of modified 9Cr-1Mo steel may be required. In addition, since the SFR pipes are mainly subjected to displacement controlled thermal loads, it is expected that fast unstable fracture is unlikely. Taking both material and structural features into account, the framework to establish a precise LBB assessment procedure for SFR pipes must be organized. For the standardization of the LBB procedure, the main investigation items were defined as follows: (1) Approval of the assessment flowchart eliminating uncertainty due to small scale leakage, e.g. self plugging phenomenon and influence of crack surface roughness on leak rate. (2) Proper selection of LBB assessment objects in JSFR. (3) Distinguishment between the matters covered by a design code and LBB, i.e. assumption of initial flaw(s). (4) Development of creep and/or fatigue crack extension assessment technique, including collection of necessary material data. (5) Development of unstable fracture assessment technique. (6) Development of leak rate evaluation technique. (7) Characterization of loads for LBB assessment. (8) Standardization of the procedure as the Japan Society of Mechanical Engineers (JSME) code.
- Pressure Vessels and Piping Division
Development of LBB Assessment Method for Japanese Sodium Cooled Fast Reactor (JSFR) Pipes: 1—Study on the Premise for the Standardization of Assessment Procedure
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Wakai, T, Machida, H, Enuma, Y, & Asayama, T. "Development of LBB Assessment Method for Japanese Sodium Cooled Fast Reactor (JSFR) Pipes: 1—Study on the Premise for the Standardization of Assessment Procedure." Proceedings of the ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference. ASME 2010 Pressure Vessels and Piping Conference: Volume 6, Parts A and B. Bellevue, Washington, USA. July 18–22, 2010. pp. 709-714. ASME. https://doi.org/10.1115/PVP2010-25243
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