Mol 7C experiments on local fault propagation in irradiated LMFBR fuel subassemblies

The consequences of local cooling disturbances in irradiated fuel pin bundles of Liquid Metal Fast Breeder Reactors (LMFBRs) were investigated in two in-pile experiments in the BR2 reactor at Mol/Belgium. A porous local blockage was used to initiate the fault. This caused severe local fuel pin damage, but there was no fault propagation to major parts of the bundles during four hours of full power operation. If such events occurred in a LMFBR, the delayed neutron detection (DND) signals generated by the fault would be sufficient to initiate reliable automatic shutdown of the reactor.     

Crack propagation under creep and fatigue

Crack growth under cyclic, static and combined loads in several high temperature alloys is presented from both material and fracture mechanics aspects. Parametric representation of high temperature crack growth in terms of linear and non-linear elastic fracture mechanics is discussed along with the experimental determination of these parameters for several specimen geometries. Crack growth is characterized as cycle-dependent, time-dependent or combined cycle- and time-dependent processes depending on material, temperature, load frequency and environment. While the cycle-dependent process is due to fatigue damage, the time-dependent process could be due to creep or environmental effects or both. It is shown that the applicability of a particular fracture mechanics parameter to characterize high temperature crack growth depends on micromechanics of the growth.     

Seismic behavior of a free-standing core in a large LMFBR

This paper describes a study on the seismic analysis and qualification of an LMFBR core. A non-linear response analysis method with FINAS is validated by comparing its results to a set of existing experimental data. The method is then applied to assess the seismic response and safety capacity of some typical configurations of a large free-standing core, under various seismic inputs including a case of base isolation. Some discussions are made on the possibility of the free-standing core.     

钠冷快堆失流事故分析

本文介绍了钠冷快堆失流计算的数学模型、FRLOF程序的编制和用本程序对EBR-Ⅱ两个失流工况进行的理论计算。该计算结果与试验测量值吻合较好。     

Tests in support of LMFBR

Tests in support of LMFBR projects for potential incidents involving a hypothetical core disruptive accident or sodium-water interactions in steam generators have two possible goals: (1) to evaluate the integrity of the primary containment after the design is frozen or (2) to investigate design options in support of primary containment design. The test planning approach differs depending on the goal. The Fast Flux Test Facility tests had the first goal, the current Clinch River Breeder Reactor tests have the second goal. This paper discusses test planning, sources for simulating loads, modeling, and instrumentation. The main source described involves controlled venting of explosive gases. The source avoids generation of undesirable shock waves and facilitates calibration because the reaction rate of the explosive is independent of the response. It is indicated that the test program requiring the least time and cost involves a mixture of three types of models: small simple models, small complex models, and large complex models. Instrumentation is discussed that is required for validation of loading, response measurements for comparison with calculations, and response measurements on critical members where predictions are not possible.     

Scale modelling in LMFBR safety

This paper reviews scale modelling techniques used in studying the structural response of LMFBR vessels to HCDA loads. The geometric, material, and dynamic similarity parameters are presented and identified using the methods of dimensional analysis. Complete similarity of the structural response requires that each similarity parameter be the same in the model as in the prototype. The paper then focuses on the methods, limitations, and problems of duplicating these parameters in scale models and mentions an experimental technique for verifying the scaling.Geometric similarity requires that all linear dimensions of the prototype be reduced in proportion to the ratio of a characteristic dimension of the model to that of the prototype. The overall size of the model depends on the structural detail required, the size of instrumentation, and the costs of machining and assemblying the model.Material similarity requires that the ratio of the density, bulk modulus, and constitutive relations for the structure and fluid be the same in the model as in the prototype. A practical choice of a material for the model is one with the same density and stress-strain relationship as the operating temperature. Ni-200 and water are good simulant materials for the 304 SS vessel and the liquid sodium coolant, respectively. Scaling of the strain rate sensitivity and fracture toughness of materials is very difficult, but may not be required if these effects do not influence the structural response of the reactor components.Dynamic similarity requires that the characteristic pressure of a simulant loading source equal that of the prototype HCDA for geometrically similar volume changes. The energy source is calibrated in the geometry and environment in which it will be used to assure that heat transfer between high temperature loading sources and the coolant simulant and that non-equilibrium effects in two-phase sources are accounted for. For the geometry and flow conditions of interest, the viscous and gravity forces are usually negligible compared with pressure forces and, therefore, are not scaled.     

Hydraulic simulation of local blockage in a LMFBR fuel subassembly

Experimental studies of local flow blockage in a LMFBR fuel subassembly have been carried out using a simulating model in a water test loop. The studies verified the numerical results of an analytical code. The experiments were conducted in a 61-pin bundle containing a planar blockage without leakage flow. Central and edge blockage were used. The wake flow behind the blockage was visualized with dye or air bubble injection to grasp the flow characteristics. The velocity, static pressure and residence time were measured in the wake flow region.     

Crack propagation characteristics in neutron irradiated type 316 stainless steel microspecimens

In order to better relate the macroscopic mechanical behavior of irradiated alloys to their associated microstructural condition, unirradiated and neutron irradiated microspecimens were tensile tested at 25–600°C in a quantitative load elongation stage while under continuous observation in a high voltage electron microscope (HVEM). The microtensile specimens, 40 μ m thick, of type 316 stainless steel were irradiated at ambient temperature to a fluence of 1 × 10²² n/m² with 14 MeV neutrons in the Lawrence Livermore Rotating Target Neutron Source II (RTNS) facility.Crack angles, directions and length plotted against total specimen elongation were used to describe the manner in which a crack progressed through each specimen. Rapid crack propagation is accompanied by rapidly changing crack angles and direction and conversely slow propagation corresponds to slowly changing variables. A graph of cumulative crack length plotted against total elongation exhibits a slope which increases as specimen ductility decreases. This graph reflects changes due to the effect of neutron irradiation.     

钠冷快堆氢探测技术

本文介绍了钠冷快堆蒸汽发生器钠-水反应的氢探测技术。扩散型氢计和电化学氢计是目前使用的两种主要装置.     

Thermal performances and melting risk assessment in a LMFBR fuel pin

A reliable evaluation of fuel temperature is a key safety requirement in the design of the fuel assembly of a nuclear reactor, especially in the case of a LMFBR whose efficient operation requires high thermal performance fuel.The physico-chemical properties such as density, oxygen to metal ratio and thermal conductivity of a typical LMFBR mixed-oxide fuel, which are known to change in a remarkable way under irradiation, strongly affect the temperature profile within the fuel pellet.A statistical analysis of the temperature values in the fuel of the Italian Fast Reactor PEC, has been performed by means of the RSM code (Response Surface Methodology) coupled to a Monte-Carlo Technique (MUP code), in order to demonstrate that the melting risk is substantially negligible.     

Loss-of-flow experiment in a 37-pin bundle LMFBR fuel assembly simulator

An experimental study was conducted on transient sodium boiling in an LMFBR fuel subassembly mockup under loss-of-flow conditions. In the test section, an electrically heated 37-pin bundle was centered in a hexagonal tube. The measured maximum IB wall superheat was 36°C, and the effects of heat flux, temperature rise rate, and system pressure were unclear. Boiling was initiated at the end of the heated section, the bubble expanded mainly to the upstream central subchannels and to the downstream unheated section according to the expansion of the saturated temperature region. When the voided zone covered the whole flow cross-section, the void pattern changed to the one-dimensional slug ejection-type and the inlet flow decreased rapidly. Dryout occurred after the inception of flow reversal in the wide region of the bundle.     

Turbulent flow simulation in a wire-wrap rod bundle of an LMFBR

The pressure drop and heat transfer characteristics of wire-wrapped 19-pin rod bundles in a nuclear reactor subassembly of liquid metal cooled fast breeder reactor (LMFBR) have been investigated through three-dimensional turbulent flow simulations. The predicted results of eddy viscosity based turbulence models (k-?, k-ω) and the Reynolds stress model are compared with those of experimental correlations for friction factor and Nusselt number. The Re is varied between 50,000 and 150,000 and the ratio of helical pitch of wire wrap to the rod diameter is varied from 15 to 45. All the three turbulence models considered yield similar results. The friction factor increases with reduction in the wire-wrap pitch while the heat transfer coefficient remains almost unaltered. However, reduction in the wire-wrap pitch also enhances the transverse flow velocity in the cross-sectional plane as well as the local turbulence intensity, thereby improving the thermal mixing of coolant. Consequently, the presence of wire wrap reduces temperature variation within each section of the subassembly. The associated reduction in differential thermal expansion of rods is expected to improve the structural integrity of the fuel subassembly.     

Utilization Factor of Nuclear Material in LMFBR

The effects of γ-irradiation on a simulated nuclear waste glass were studied by electron spin resonance spectroscopy (ESR), and were compared with the results on silica glass and Pyrex glass. Three kinds of glasses were γ-irradiated up to the dose of 1.22 MGy and the ESR spectra were obtained. The intensity of ESR spectra were obtained as a function of irradiation dose and annealing temperature.

The spectrum of the waste glass was characteristic of two typical peaks, Peak 1 was the strong resonance at g=4.3 showing the existence of four coordinated Fe³⁺ and Peak 2 was the weak and broad resonance at g= 2.0 showing the existence of six coordinated Fe³⁺. The ESR spectra of the waste glass before and after γ-irradiation were almost overlapped and a little difference only in the intensity was observed. While in silaca glass and Pyrex glass, the peaks from E'γ center and boron-oxygen hole center (BOHC) were observed to arise after irradiation. The absolute intensity of. Peaks 1 and 2 described above changed in complicated way depending on the dose. The result suggests oxidation or reduction of iron takes place in the waste glass depending on the dose. The isochronal annealing of irradiated glasses shows most of γ-ray- induced damages in the waste glass are restored even at room temperature, although most of the damages in silica glass and Pyrex glass are disappeared at the temperature from 550 to 600 K. The results show that the waste glass with a few weight percent of iron is resistent to radiation than other commercial glasses.     

The role of reliability and risk assessment in LMFBR design: Implementation of reliability in LMFBR design

This paper concerns future developments in LMFBR licensing technology.Federal Regulations (10 CFR 50.34) require that the preliminary safety analysis provide analysis and evaluation “with the objective of assessing the risk to public health and safety” to determine margins of safety and the adequacy of the plant. Hitherto the assessment of risk has been qualitative but it has become increasingly apparent that quantitative assessments would provide a better basis for judgement. Potential future roles of reliability and risk assessment are discussed in the context of providing additional confirmation of the safety of LMFBR designs. Potential acceptability criteria for risk evaluations are outlined.The reliability implications of designing components to the ASME Code Section III requirements are discussed. General judgements are provided as well as the preliminary results of probabilistic studies of selected specific limits. There is a different reliability significance for the mandatory rules for normal, upset, and emergency conditions versus the non-mandatory rules for normal, upset, and emergency conditions versus the non-mandatory guidance for faulted conditions.     

Thermo-mechanical analysis of concrete in LMFBR programs

The R&D program described in this paper represents the structural mechanics effort underway and planned at ANL as an attempt to better understand the behavior of concrete structures in LMFBR plants where such structures are subjected to high temperature levels. This paper describes the analytical tools formulated and incorporated into the thermo-mechanical computer code called TEMP-STRESS.Emphasis in this paper is placed on describing the short-time and long-time constitutive formulation for concrete. The primary constitutive relation uses an elasto-plastic technique to simulate concrete nonlinearities, utilizes the von Mises ellipsoid as the loading surface, and assumes a four-parameter failure surface. Post-failure treatment is different when considering surface elements as opposed to internal elements. Creep of the concrete at temperatures up to about 400°C is approximated by a rate-type creep law using Maxwell chain technique. The creep model accounts for moisture and pore pressure disposition of the concrete.     

Sodium thermal-hydraulics in the pool LMFBR primary vessel

The three-dimensional flow of the primary sodium in the LMFBR main vessel, circulating by natural and forced convection, results in an uneven temperature distribution in the large mass of sodium itself (stratification, thermal fluctuations, thermal dissymmetry ... ) and on the components walls. In France, extensive interdependent theoretical, numerical and experimental studies have been performed on this subject. They aim at determining the thermal conditions during all operating conditions with a view to assessing the resulting thermal stresses and evaluating structure behaviour.The results of the commissionning tests of the Super-Phenix's plant show the validity of the method used; the thermalhydraulic knowledge leads to an improvement of the tools. They will be used as a design capability, from the viewpoint of compact and economical design for the next European Fast Breeder Reactor.     

Assessment of accident energetics in LMFBR core-disruptive accidents

An assessment of accident energetics in LMFBR core-disruptive accidents is given with emphasis on the generic issues of energetic recriticality and energetic fuel-coolant interaction events. Application of a few general behavior principles to the oxide-fueled system suggest that such events are highly unlikely following a postulated core meltdown event.     

快堆用钠的除钙技术

本文叙述了商业级钠在充入钠净化回路Ｍｕｌｔｉｐｕｒｐｏｓｅ Ｓｏｄｉｕｍ Ｐｕｒｉｆｉｃａｔｉｏｎ Ｌｏｏｐ之前的除钙技术。处理后的钠中钙杂质含量由的约０．０４％－０．０５％降 ０．０００５％以下，达到堆级钠纯度标准。     

Accommodation of potential hydrogen formation in LMFBR accidents

Results of design verification tests for the FFTF reactor cavity liner system are presented which suggest that steel liners would retain their integrity even under certain hypothetical accident conditions, thus, avoiding the formation of hydrogen. When liner failures are postulated in hypothetical reactor vessel melttrough accidents, hydrogen levels can be controlled by an air purging system. The design of a containment purging and effluent scrubbing system is discussed.     

Irradiation enhanced creep in LMFBR fuel element cladding

Rate changes observed in irradiation enhanced creep and swelling in stainless steel cladding are ascribed to the precipitation of carbide. Empirical equations modified according to precipitation kinetics are consistent with results from fuel element irradiation and in particular describe the “second peak” phenomenon.