Estimation of the behavior of structures with localized nonlinearities using an ultimate state method

At first, exemplary results of calculations of multi-degree-of-freedom-systems with locally yielding elements are presented. As a second step, a new approximate method for the estimation of the response of systems with localized nonlinearities under dynamic loads is explained.     

Fragility estimation for seismically isolated nuclear structures by high confidence low probability of failure values and bi-linear regression

A method for the fragility estimation of seismically isolated nuclear power plant structure is proposed. The relationship between the ground motion intensity parameter (e.g. peak ground velocity or peak ground acceleration) and the response of isolated structures is expressed in terms of a bi-linear regression line, whose coefficients are estimated by the least-square method in terms of available data on seismic input and structural response. The notion of high confidence low probability of failure (HCLPF) value is also used for deriving compound fragility curves for coupled subsystems.     

氧化物燃料快堆的事故瞬态分析

利用在核电厂动态仿真器ＤＳＮＰ上开发完成的仿真程序ＯＸＳＹＳ，分析计算了氧化物燃料钠冷快堆ＣＲＢＲＰ在超功率和失流事故瞬态下的响应特性，所得结果与国外系统分析程序ＳＳＣ－Ｌ、ＦＰＲＥ－Ⅱ的相应计算结果符合较好。     

Electrical treeing behaviors in silicone rubber under an impulse voltage considering high temperature

In this paper,work was conducted to reveal electrical tree behaviors(initiation and propagation)of silicone rubber(SIR) under an impulse voltage with high temperature.Impulse frequencies ranging from 10 Hz to 1 k Hz were applied and the temperature was controlled between 30 °C and 90 °C.Experimental results show that tree initiation voltage decreases with increasing pulse frequency,and the descending amplitude is different in different frequency bands.As the pulse frequency increases,more frequent partial discharges occur in the channel,increasing the tree growth rate and the final shape intensity.As for temperature,the initiation voltage decreases and the tree shape becomes denser as the temperature gets higher.Based on differential scanning calorimetry results,we believe that partial segment relaxation of SIR at high temperature leads to a decrease in the initiation voltage.However,the tree growth rate decreases with increasing temperature.Carbonization deposition in the channel under high temperature was observed under microscope and proven by Raman analysis.Different tree growth models considering tree channel characteristics are proposed.It is believed that increasing the conductivity in the tree channel restrains the partial discharge,holding back the tree growth at high temperature.     

Investigation on buckling of FBR vessels under seismic loadings with fluid structure interactions

The critical thin walled shell structures in the reactor assembly of a pool type fast breeder reactor (FBR) are the main vessel, inner vessel and thermal baffles. On these structures, the seismic events impose major forces by developing high dynamic pressures, thereby causing a concern on structural integrity due to buckling. An integrated analysis for determining realistic forces and critical buckling loads at any instant during the seismic event has been carried out for the vessels of a typical 500 MWe pool type fast breeder reactor. The dynamic forces including pressure distributions generated on the vessel surfaces are extracted from the seismic analysis carried out for the reactor assembly. The seismic forces thus computed from axisymmetric analysis are transmitted appropriately to 3D shell geometries for the buckling analysis. In view of high computational time needed for carrying out buckling analysis at every time increment, the elastoplastic analysis is carried out only at a few critical time steps which are identified based on strain energies that are associated with the shear and compressive stresses developed at the portions of the vessels prone to buckle. The shear buckling of main vessel straight portion and buckling of toroidal portion of inner vessel and thermal baffles are found to be important. The possible randomness of support excitation time histories is accounted for by compressing as well as expanding the nominal time histories by 10%. Buckling strength reduction factors due to the initial geometrical imperfections are adopted from the literature. The inner vessel is found to be the most critical component which buckles under seismic forces induced by a safe shutdown earthquake with a load multiplier of 1.52, which is higher than the minimum factor of safety of 1.3 required as per the design code RCC-MR [RCC-MR: edition, 2002. Design and construction rules for mechanical components for FBR nuclear islands, vol. 1, section I, subsection B. AFCEN, Paris, in press] for service level D conditions.     

Outline of the seismic buckling design guideline of an FBR — a tentative draft

Central Research Institute of Electric Power Industry (Japan), commissioned by the Ministry of International Trade and Industry, is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings.     

Design study of the demonstration FBR in Japan

The objective of the plant design study Phase 2, conducted by the Japan Atomic Power Company since 1997 for 3 years, is to accomplished a plant overall concept of the Demonstrative FBR (DFBR) that has economical potential toward commercialization and offers high reliability to plant operators not to cause a long unexpected shutdown resulting from a trouble, i.e., sodium leakage or fires. This has been successfully achieved by establishment of a plant overall design of 672 MWe consisting of the reactor system with drastically simplified internals, the compact and double walled coolant boundaries, the well rationalized fuel handling system, the BOP systems introducing up-to-date LWR equipment, and the compact reactor building.

The plant construction cost has been estimated based on the quantity of materials to be about 130 % on the bases of a 1000 MW LWR, which is well contented with the requirement.

The DFBR plant concept, having economical potential toward commercialization, safety and reliability, has been established in the plant design study Phase 2.     

Three-dimensional local coolant flow velocity field under the influence of wire-spacers in an FBR fuel subassembly

In order to evaluate the coolant cross-flow rate among subchannels of a wire-spaced FBR fuel subassembly, the three-dimensional Navier-Stokes equation was solved numerically for a detailed flow velocity distribution within several connected subchannels inside a subassembly, where consideration was focussed on setting up iteratively an approriate velocity field on boundary interfaces enclosing the subchannels under consideration as the boundary condition. Such subchannels may include a peripheral fuel pin.Some of the numerical results obtained are as follows: (1) In an annular channel coaxed with a wire-spaced fuel pin, the maximum azimuthal velocity component does not appear just at the upstream side of a wire spacer but it appears at the leading angle of 180° to the upstream side with respect to the wire-wrapping phase in a fully developed flow region apart from the entrance. (2) In a wire-spaced fuel pin bundle, the transverse velocity increases steeply in the vicinity of the upstream side of a wire-spacer, while it increases gradually with the development of wakes in the downstream side of a wire-spacer. (3) At the peripheral gaps the swirl flow is induced in the wire-wrapping direction along the inner surface of a wrapper tube and its circumferential evolution predicted in the present analysis is in good agreement with experimental data obtained by a MIT group.     

A thermal hydraulic calculation method of an intermediate heat exchanger of a loop type FBR

This paper describes a thermal-hydraulic calculation of an intermediate heat exchanger (IHX) with the computational fluid dynamics (CFD) code CFX. The motivation of this paper is to clarify a heat transfer degradation phenomenon in the IHX through three-dimensional calculation. The whole IHX of the “Monju” reactor is modeled with three parts, i.e., the primary side, the secondary side and the heat transfer region. Through a partial calculation using these models, the flows on the primary side and the secondary side are shown to be axisymmetric. Therefore, a sector model is adopted for the calculation model in the heat transfer region. The calculated temperatures in the IHX are compared with the measured results using the IHX in the “Monju” reactor. Good agreement is obtained for the predicted outlet temperatures and temperatures on the shell surface. As a result of the CFD calculation, it is evaluated that a heat transfer in the lower plenum on the secondary side is dominant under the low flow rate conditions. This fact contributes to lower the heat transfer coefficient in the IHX when the standard heat exchanger theory is applied to calculate the overall heat transfer coefficient between the primary and the secondary sides.     

Exploratory research on creep and fatigue properties of 9Cr-steels for the steam generator of an FBR

Research on the applicability of 9Cr-steels to the steam generator of the demonstration fast breeder reactor was performed by the Subcommittee of the Japan Welding Engineering Society as a four-year program from 1985. In this program, exploratory tests, which included tensile, creep rupture and low-cycle fatigue tests, were conducted on three kinds of 9Cr-steels (Mod.9Cr-1Mo, 9Cr-1Mo-V-Nb, and 9Cr-2Mo) and their weldments. This paper describes the summary of results obtained in this program. Among the tested 9Cr-steels, Mod.9Cr-1Mo steel shows the best creep rupture strength and its weldment indicates almost the same level of creep rupture strength and the base metal at 500 and 550°C. The low-cycle fatigue properties of Mod.9Cr-1Mo steel is also discussed from its relation to the tensile properties.     

Application of X-Ray Computer Tomography for Observing the Deflection and Displacement of Fuel Pins in an Assembly Irradiated in FBR

A nondestructive method making use of X-ray computer tomography (X-ray CT) has been applied to post irradiation examination of fast breeder reactor (FBR) fuel assemblies. In the study, an examination is made of the deflection and displacement of fuel pin in a fuel assembly irradiated to 74.2GWd/t peak burnup in the fast reactor “JOYO.”

In the examination, X-ray CT images of transverse cross sections of fuel pin were obtained at different heights of fuel pin along its axis. Analysis of the resulting images indicated that:

1. The hexagonal wrapper tube had its lateral wall faces slightly bulged outward;

2. The fuel pins loaded in the outermost array were markedly displaced in the direction of wrapper tube, particularly in portions of fuel pin intermediate between positions constrained by wrapping wire.

The latter behavior of fuel pins was substantiated by the contours of fuel pin along its axis, which were derived from cross section images obtained at different levels along axis.

Such fuel pin displacement is surmised to have been caused by thermal stressing of the affected fuel assembly cladding.     

Simple method of evaluating the failure probability of a structure considering the skewness of distribution

This paper consists of two parts. First, we describe a structure response variability evaluation method. A two-point estimation method is useful for the evaluation of structure response variability in nuclear reactor buildings. We make it possible to take correlations into account, thus expanding the applicability of this method.

Second, we propose a simple method for easily obtaining the non-exceedance probability of a non-Gaussian probability distribution from its skewness. Using this method, a reliability index taking into account the skewness of the performance function is obtained.

Using these methods, we can easily evaluate the structural failure probability of a nuclear reactor building by considering the skewness of the distributions of both the random variable in the analysis model and the performance function. In this paper, we explain these methods and apply them to a nuclear reactor building as a numerical example.     

Feasibility study of large MOX fueled FBR core aimed at the Self-Consistent Nuclear Energy System

The potential of a MOX fueled fast breeder reactor (FBR) is evaluated with regard to its ability to transmute radioactive nuclides and its safety when incorporated in the so-called self-consistent nuclear energy system (SCNES). The FBR's annual production amounts of selected long-lived fission products (LLFPs), Se-79, Tc-99 Pd-107, I-129, Cs-135 and Sm-151, can be transmuted by using a radial blanket region and a part of a lower axial blanket region without any significant impact on its nuclear and safety characteristics. The other LLFPs are confined in the system. The hazard index level of the LLFPs per one ton of spent fuel from the system after 1000 years is as small as that of a typical uranium ore. To realize self-controllability (passive safety), the proposed FBR core concept employs gas expansion modules and sodium plenum above the core. To realize self-terminability, even if MOX fuel melting should cause a core compaction, recriticality of the core can be avoided by a fuel dilution and relocation module. The results show the MOX fueled FBR core has potential applicability to the SCNES. With the final goal of the ideal SCNES, fundamental applicability of various coolants and fuels is evaluated based on neutron balance. It is shown that the harder the core spectra is, the larger the potential for transmuting LLFPs would be.     

Numerical simulation of nanosecond laser ablation and plasma characteristics considering a real gas equation of state

Based on the governing equations which include the heat conduction equation in the target and the fluid equations of the vapor plasma,a two-dimensional axisymmetric model for ns-laser ablation considering the Knudsen layer and plasma shielding effect is developed.The equations of state of the plasma are described by a real gas approximation,which divides the internal energy into the thermal energy of atoms,ions and electrons,ionization energy and the excitation energy of atoms and ions.The dynamic evolution of the silicon target and plasma during laser ablation is studied by using this model,and the distributions of the temperature,plasma density,Mach number related to the evaporation/condensation of the target surface,laser transmissivity as well as internal energy of the plasma are given.It is found that the evolution of the target surface during laser ablation can be divided into three stages:(1)the target surface temperature increases continuously;(2)the sonic and subsonic evaporation;and(3)the subsonic condensation.The result of the internal energy distribution indicates that the ionization and excitation energy plays an important role in the internal energy of the plasma during laser ablation.This model is suitable for the case that the temperature of the target surface is lower than the critical temperature.     

Presentation of the properties of the ferritic alloy 2.25 Cr-1 MoNiNb for steam generator application in FBR systems

2.25 Cr-1 MoNiNb steel, used for the construction of the steam generator of a fast breeder reactor, is subjected to operation at elevated temperature in the creep range. Although this operation condition is a limiting factor for allowable loads during normal operation, it is necessary to have sufficient knowledge of the strength-properties of this kind of steel after thermal aging. Experiences with this steel are described. It is shown that this stabilized ferritic steel reveals the common behaviour of all ferritic steels at elevated temperature. The change of mechanical properties can easily be quantified. Special attention was given to the analysis of strain-controlled tensile testing and the uniform elongation of s.g.-tubing.     

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.     

用PAT研究两种硅橡胶的大分子结构特征

用未硫化的甲基乙烯基硅橡胶和接枝室温硫化硅橡胶作正电子湮没实验的结果表明，甲基乙烯基硅橡胶的τ１，τ２均较ＲＴＶ的短，而τ３较长。对这一现象作了定性的解释。     

Method for Condensation of the Macroscopic Transport Cross-Sections for Criticality Analyses of FBR MONJU by the Code NSHEX

The problem with the energy-group approximation effect in the criticality analysis of the FBR MONJU by the discrete-ordinate transport code NSHEX has been studied. In order to reduce the existing energy-group collapsing effect in the 18 and 7-group results for the effective multiplication factor, a new algorithm for condensation of the macroscopic transport cross-sections has been proposed and verified. This work presents the definition of the new collapsing algorithm, results from the verification tests and a short discussion from the viewpoint of consistency of the algorithm with the specific finite-difference method of the code NSHEX. According to the presented results, the new collapsing algorithm can be recommended for condensation of the transport cross-sections from 70 into fewer energy-group structures.     

Assessment of Fuel Coolant Interactions (FCIs) in the FBR Core Disruptive Accident (CDA)

Application of general behavior principles (GBPs) and consideration of relevant contact modes suggest that only incoherent small-scale fuel coolant interactions (FCIs) with negligible damage potential appear possible with the molten oxide fuel-liquid sodium system as the fuel disperses away from the core into a coolable non-critical array.

In contrast to the SPERT-1, BORAX-1 and SL-1 nuclear transients that ultimately led to energetic vapor or steam explosions, the presence of molten fuel and liquid sodium in the FBR core always requires the presence of solid cladding which separates the fuel and coolant and, hence prevents energetic FCIs prior to coolant escape.

Furthermore, unlike the CORECT-II experiments which examined dynamic re-entry of liquid sodium on molten fuel pools that resulted in unstable interfaces leading to significant sodium entrapment and relatively energetic FCIs, the prevailing contact mode in the FBR core disruptive accident (CDA) scenario is displacement of the lighter and less viscous liquid sodium by the heavier and more viscous molten fuel resulting in stable interfaces with no significant sodium entrapment and FCIs. Dynamic re-entry of liquid sodium into the core is not possible with the two-component steel vapor-liquid sodium system, since the interface contact temperature upon steel vapor condensation is well in excess of the sodium boiling temperature. A pressure reduction in the steel vapor region due to condensation is immediately compensated for by an equivalent pressure increase due to sodium evaporation.

Finally, considering that the molten oxide fuel-liquid sodium interface contact temperature is well below the sodium homogeneous nucleation temperature which in turn is well below the fuel melting temperature, not only eliminates the potential for large-scale vapor explosions as molten fuel streams are injected into liquid sodium pools, but also implies that small scale superheat explosions are possible which are consistent with the usually observed incoherent sharp pressurization events (amplitudes up to the order of 10 MPa and duration of the order of 1 ms). These general behavior characteristics are also consistent with complete fuel fragmentation with fragment sizes ranging from 100 to 1,000 μm, and the absence of significant or damaging FCIs.