[en] Two-pion Bose-Einstein correlations have been studied using the BNL-E866 Forward Spectrometer in 11.6 A · GeV/c Au + Au collisions. The data were analyzed using three-dimensional correlation parameterizations to study transverse momentum-dependent source parameters. The freeze-out time and the duration of emission were derived from the source radii parameters

[en] Differences are investigated in pion source sizes derived using π⁺π⁺ and π^-π^- pairs in the large statistics data set collected with the E866 Forward Spectrometer for central Au+Au collisions at 11.6 A. · GeV/c in AGS Experiment E866. These differences in source radii are interpreted using a simple classical description for the Coulomb and the transverse-momentum (p_T) dependent source radius parameters. An estimated effective net charge responsible for the distortion is considerably smaller than the expected total projectile participant protons, which suggests that the system undergoes a rapid expansion in the longitudinal direction before freezeout. This picture is consistent with the results derived from the π^-/π⁺ singles yield ratios for the same reactions

[en] The main objective of this report is to evaluate the structural integrity accounting for creep and stress-rupture effect for the KALIMER reactor internal structures subjecting the normal reactor heat-up and cool-down transient operating cycles during 30 years of total life time. From the results of the structural damage evaluations, most parts of the reactor internal structures satisfy the limit rules of the structural integrity using ASME design code. However, the reactor baffle parts at elevation of the hot pool free surface slightly exceed the limit value of the total accumulated creep-ratcheting strain, then more detail structural analyses and evaluations for this region should be carried out to meet the design rules

[en] This report describes a simple seismic analysis model of the KALIMER-600 sodium cooled fast reactor and its application to the seismic time history analysis. To develop the simple seismic analysis model, the detailed 3-D finite element analyses for main components, IHTS piping system, and reactor building were carried out to verify the dynamic characteristics of each part of simple seismic analysis models. By using the developed simple model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design of KALIMER-600 were performed. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity

[en] In this report, the application algorithm of a consistent fluid added mass matrix including the coupling terms to the core seismic analysis is developed and installed at SAC-CORE3.0 code. As an example, we assumed the 7-hexagon system of the LMR core and carried out the vibration modal analysis and the nonlinear time history seismic response analysis using SAC-CORE3.0. Used consistent fluid added mass matrix is obtained by using the finite element program of the FAMD(Fluid Added Mass and Damping) code. From the results of the vibration modal analysis, the core duct assemblies reveal strongly coupled vibration modes, which are so different from the case of in-air condition. From the results of the time history seismic analysis, it was verified that the effects of the coupled terms of the consistent fluid added mass matrix are significant in impact responses and the dynamic responses

[en] The establishment of the inelastic analysis technology is essential issue for a development of the next generation reactors subjected to elevated temperature operations. In this report, the peer investigation of constitutive equations in points of a ratcheting and creep-fatigue analysis is carried out and the methods extracting the constitutive parameters from experimental data are established. To perform simulations for each constitutive model, the PARA-ID (PARAmeter-IDentification) computer program is developed. By using this code, various simulations related with the parameter identification of the constitutive models are carried out

[en] The main objective of this report is to establish the analysis and evaluation methodology of the progressive thermal buckling behavior for the LMR structures subjected to moving high temperature cycles. To do this, the ANSYS version 7.1 was used with the nonlinear material constitutive equation of Chaboche's model. Using this model, the progressive thermal buckling behavior was identified for the cylindrical structures having the free edge. As an example of the application, the progressive thermal buckling analysis for the KALIMER was carried out, and the results are described in this report

[en] This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor

[en] In this report, the SIE ASME (Structural Integrity Evaluations by ASME-NH), which has a computerized implementation of ASME Pressure Vessels and Piping Code Section III Subsection NH rules, is developed to apply to the next generation reactor design subjecting to the elevated temperature operations over 500 .deg. C and over 30 years design lifetime, and the user's manual for this program is described in detail

[en] In this report, the SIE ASME (Structural Integrity Evaluations by ASME-NH) (Revision 1.0), which has a computerized implementation of ASME Pressure Vessels and Piping Code Section III Subsection NH rules, is developed to apply to the next generation reactor design subjecting to the elevated temperature operations over 500 .deg. C and over 30 years design lifetime, and the user's manual for this program is described in detail