基于CFD方法的行波堆燃料组件燃烧区热工流体特性研究

采用计算流体力学（CFD）方法对行波堆燃料组件7棒束、19棒束及37棒束模型进行计算分析,发现行波堆燃料组件内冷却剂温度随轴向高度增加逐渐升高的同时具有逐渐向中心区域聚集的效应,组件出口区域垂直于流动方向的截面冷却剂温度分布差别很大,对边距约为26 cm的组件中心区域与外围区域最大温差超过100 ℃。组件内较大的冷却剂温度梯度主要出现在组件最外两圈燃料棒及组件盒之间的区域,而其他区域温度梯度较小,该结论可初步推广到有217根燃料棒的行波堆燃料组件。现有行波堆燃料组件结构需进一步优化。     

Development of a 3-D FBR fuel bundle-to-duct interaction analysis code marse

A computer code marse has been developed to analyze structural behaviors of fast breeder reactor (FBR) fuel subassemblies under irradiation conditions, especially of wire spaced fuel pin bundles.The first concept specifically considered is to model on pin oval deformation and dispersion phenomena which were found to occur in highly irradiated fuel pin bundles. Every fuel pin is modeled by three-dimensional finite element method beams with trusses at contact points. Almost all phenomena which would occur under irradiation induced bundle-to-duct interaction (BDI) conditions are included in the marse code; these include bending, expansion, oval deformation and dispersion.The second point is to reduce computing time because the BDI analysis requires much computing time if a conventional solving scheme is applied. This problem is resolved by using only a small stiffness matrix of each pin successively to treat interactions with other pins or ducts as outer forces.The marse code has been applied to BDI analyses and has been confirmed to be applicable to any types of FBR fuel pin bundles with high accuracy and a small computing time.     

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified from a state-of-the-art review of open papers. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation are modeled into a safety assessment code which is applicable to arbitrary SFRs by developing some needed but missing methods. Furthermore, an assessment on FEFPA of Japanese prototype fast breeder reactor (Monju) was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited at most within one subassembly in Monju owing to its redundant and diverse detection and shutdown systems for FEFPA even assuming the propagation. These results also suggested future possibility of run-beyond-cladding-breach operation which would enhance the economic efficiency in Monju.     

Comprehensive and consistent interpretation of local fault experiments and application to hypothetical local overpower accident in Monju

Experimental studies on local fault (LF) accidents in fast breeder reactors have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Comprehensive and consistent interpretations of in-pile and out-of-pile experiments related to LF were arrived at in this study based on state-of-the-art review and data analysis techniques. Safety margins for a hypothetical local overpower accident, which was evaluated as a LF accident in the licensing document of the construction permit for a prototype fast breeder reactor called Monju, were also studied. Based on comprehensive interpretations of the latest experimental database, including those performed after the permission of Monju construction, it was clarified that the evaluation of the hypothetical local overpower accident in the Monju licensing was sufficiently conservative. Furthermore, it incorporated adequate safety margins in terms of failure thresholds of the fuel pin, molten fuel ejection, fuel sweep-out behavior after molten fuel ejection, and pin-to-pin failure propagation. Moreover, these comprehensive interpretations are valid and applicable to the safety evaluation of LF accidents of other fast breeder reactors with various fuel and core designs.     

Boiling and dryout behind local blockages in sodium cooled rod bundles

Much attention has been given in LMFBR safety analysis to cooling disturbances caused by local blockages within a fuel subassembly. Such blockages are generally considered to be more probable in gridded fuel pin clusters which present the possibility for solid particles in the coolant to be trapped at grids to form a radially extending flow obstruction. The temperature distribution produced in the region of impaired cooling has been studied in water and sodium experiments in pin bundles of various sizes. The experimental work at KfK on local cooling disturbances culminated in two local blockage experiments in the KNS sodium loop simulating LMFBR fuel elements with a 49% central and a 21% corner blockage. In the frame of this work pin cooling in the wake of the blockage was investigated in single-phase conditions, in boiling conditions up to dryout and in conditions simulating gas release from failed pins. The general aims of the studies were to demonstrate that the consequences of a local blockage do not lead to rapid propagation of damage within a pin bundle and to obtain data for validation of theoretical models.     

Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels

This paper proposes a benchmark problem suite for studying the physics of next-generation fuels of light water reactors. The target discharge burnup of the next-generation fuel was set to 70GWd/t considering the increasing trend in discharge burnup of light water reactor fuels. The UO₂ and MOX fuels are included in the benchmark specifications. The benchmark problem consists of three different geometries: fuel pin cell, PWR fuel assembly and BWR fuel assembly. In the pin cell problem, detailed nuclear characteristics such as burnup dependence of nuclide-wise reactivity were included in the required calculation results to facilitate the study of reactor physics. In the assembly benchmark problems, important parameters for in-core fuel management such as local peaking factors and reactivity coefficients were included in the required results. The benchmark problems provide comprehensive test problems for next-generation light water reactor fuels with extended high burnup. Furthermore, since the pin cell, the PWR assembly and the BWR assembly problems are independent, analyses of the entire benchmark suite is not necessary: e.g., the set of pin cell and PWR fuel assembly problems will be suitable for those in charge of PWR in-core fuel management, and the set of pin cell and BWR fuel assembly problems for those in charge of BWR in-core fuel management.     

Sodium boiling experiments in the THORS facility

Experiments have been performed with 19- and 61-pin test assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at the Oak Ridge National Laboratory (ORNL) since 1971. The THORS Facility is a high-temperature sodium system operated for the US Liquid-Metal Fast Breeder Reactor (LMFBR) Safety Program. The facility is used primarily for testing simulated LMFBR fuel subassemblies (pin (bundles). High-performance, electrically heated fuel pin simulators (FPSs) duplicate the heat generating capabilities and the dimensional characteristics of the nuclear fuel pins. A number of test bundles have been built and operated to obtain base thermal-hydraulic data, inlet and heated zone blockage data, and transient boiling data. Five of these bundles have been operated under two-phase conditions. Sodium boiling for periods up to twelve minutes were sustained in one bundle. (The lengths of the periods were limited only by automatic data recording capability). Clad dryout occurred in several tests. Tests were run at widely varying conditions of flow and power density. Testing with nonuniform power distribution across the bundle was also a part of the program.A 19-pin bundle with 12 peripheral guard heaters and a 6-subchannel blockage around the center pin in the heated zone was tested. The test program for this bundle was designed to determine if local boiling in the wake of the blockage propagates radially or axially during quasi-steady-state conditions. Post-test inspection revealed that significant helical distortion of the FPSs occurred in the vicinity of the blockage plate. This distortion probably influenced the boiling behavior. In the more severe tests, boiling initiated at the outlet of the heated zone and propagated radially into the unblocked subchannels after it had progressed upstream to the blockage. The subchannel analysis codes, SABRE and COBRA, accurately predict the extent of the boiling region.Experimental and analytical studies of sodium boiling behavior in unblocked 19- and 61-pin bundles indicate that cooling can be maintained for a significant period of time beyond boiling inception in a flow-power transient. Quasi-steady-state boiling occurred under natural-convection conditions.Investigations of the temperature data indicate that the thermal-hydraulic behavior during boiling transients is determined by two-dimensional effects, and that one-dimensional models cannot accurately predict the important phenomena associated with sodium boiling in test bundles. The subchannel code SABRE-2P (with a simple two-phase multiplier boiling model) and the two-region equilibrium mixture code THORAX (developed at ORNL) accurately predict the two-dimensional behavior between boiling inception and dryout.Extrapolation of the data from the smaller bundle tests to full-size fuel assemblies shows that the time between boiling inception and dryout would be lower for a 217-pin bundle than for a 61-pin bundle for a comparable transient. However, the time delay would still be significant, especially in a heterogeneous reactor core.     

On the failure of fast reactor fuel pins

The prediction of the timing and position of fuel pin failures is an important task in the modelling of fast reactor fuel behaviour. The range of processes that can provoke failure of fast reactor fuel pins in normal operating conditions and during hypothetical accidents is reviewed. Some of the mechanisms of failure are examined in more detail and the effect of hot spots and local stress concentrations is discussed. A review of failure criteria used in fast reactor fuel pin codes is given elsewhere, but the difficulties in applying various types of criteria are examined. Some discussion is also given on probabilistic approaches. Recommendations are given for a future approach to the problem of failure prediction, resolving the dilemma between inadequate empirical criteria and over-complex physically based approaches.     

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.     

Simulation of the transient behaviour of LMEBR fuel pins under consideration of special burnup phenomena using the BREDA-II model

A new fuel pin model was developed to describe the influence of specific burnup phenomena on the behaviour of fuel pins under transient overpower conditions in a liquid metal fast breeder reactor (LMFBR). It has been used for transient fuel pin deformation analysis during hypothetical core disruptive accidents (HCDA) and for the purpose of interpreting fuel pin failure tests. The fuel pin model, designated as BREDA-II, is based on the equations of the quasi-static theory of thermal elasticity. The fuel is regarded as elastic and the cladding as elasto-plastic material. The equations for the stress-strain analysis are based on the plane strain approximation. A multiregion fuel pin model allows to simulate long-time and transient burnup phenomena. The long-time effects taken into account are the steady state swelling of fuel, the change in fuel porosity and the production and partial release of fission gases. During a power excursion transient fuel swelling and pressure increase due to transient fission gas behaviour are included in the deformation analysis. Potential fuel pin failure is indicated by the application of various criteria of failure. In subsequent model calculations the behaviour of an irradiated LMFBR fuel pin during an overpower transient corresponding to a reactivity ramp of $5/sec is simulated and interpreted from the point of view of reactor safety.     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

The SCARABEE facility — Its main characteristics and the experimental program

The first phase of the SCARABEE programme has already been done with fresh fuel on single and seven pin bundles. We are describing the SCARABEE facility which includes the reactor, the sodium loop, the test section, the equipment for post mortem examination, the instrumentation and the recording systems.The methods used to determine the experimental parameters such as the power generated in the pins or the heat losses are presented.The different types and number of experiments are also described.     

Natural Uranium Heavy Water Reactors for Desalination a Comparison Bltween the Fuel Pin and Fuel Cwijster Concepts

Studies on the potential use of nuclear desalination in Egypt have been conducted. The choice of reactor type and system concept was influenced by emphasis on relaxed technological conditions so as to fit local circumstances . Natural uranium heavy water reactors were found promising. Finer details of type of fuel, cladding, coolant, and system arrangement were left for comparative studies. The question of fuel type, fuel pin versus fuel cluster, was considered. Simplified analytical and computational techniques were adopted and results verified with published criticality measurements. It was found that the pin design would have higher breeding potentials while the cluster would provide simpler core arrangement especially if a pressure vessel design is chosen. Results are given for a typical 40 MWth project study showing core features and system characteristics.     

Thermal-hydraulic transient analysis of light water reactors using moving boundaries

This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs.     

Statistical evaluation of pin gaps in fuel clusters for fugen (II)

Fuel pin gaps of Fugen fuel assemblies deviate statistically from their nominal value due to manufacturing and assembling tolerances which influence the thermal and hydraulic characteristics of the reactor core. For assurance of the minimum fuel pin gap, an analytical method of evaluating the reliability of spacer gauge tests applied to selected fuel pin gaps arrayed within a Fugen fuel assembly is discussed where a computer program STGAP is utilized.Correlations among the thickness of a spacer gauge, the reliability of the test and the rate of rejecting fuel assemblies whose pin gaps all satisfy the minimum design criterion are discussed in connection with the optimum gauge thickness for a given realiability level of the test. Sample calculation shows that fuel subassemblies installed in a Fugen reactor core have the overall reliability level of 99.9954% at the beginning of fuel life.     

COMMEN程序燃料元件精细模型的开发和验证

COMMEN程序是中国原子能科学研究院开发的钠冷快堆堆芯严重事故分析程序,包含了热工水力学模块、结构模块以及中子学模块。本文介绍COMMEN程序的燃料元件精细模型,该模型对燃料芯块内部节点进行划分,从而详细描述了燃料元件棒的径向温度分布。使用含有燃料元件精细模型的COMMEN程序从反应性反馈方面对中国实验快堆的UTOP（无保护超功率）事故进行计算分析,并将SAS4A程序和COMMEN程序的计算结果进行对比验证。结果显示,燃料元件精细模型计算的燃料温度与SAS4A程序的计算结果符合很好,开发的COMMEN程序适用于UTOP事故分析。     

Cladding response to uniform radial compaction of clinch river breeder reactor blanket rod bundles

Neutron radiation induced growth can lead to compaction of fast reactor fuel and blanket rod bundles within their duct assemblies, resulting in local cladding stresses and coolant channel closure. The stiffness of the Clinch River Breeder Reactor (CRBRP) blanket rod bundle during uniform radial compaction was measured experimentally, and was compared to theoretical predictions which assumed no relative rod-rod motion within the bundle during compaction. The basic agreement between the measured and theoretical compaction stiffness values permitted an assessment to be made of cladding stresses and coolant channel closures in CRBRP blanket rod bundles due to rod bundle-duct differential growth.     

Analysis of fuel pin behavior under slow-ramp type transient overpower condition by using the fuel performance evaluation code ‘FEMAXI-FBR’

FEMAXI-FBR has been developed as the one module of the core disruptive accident analysis code ‘ASTERIA-FBR’ in order to evaluate the mixed oxide (MOX) fuel performance under steady, transient and accident conditions of fast reactors consistently. On the basis of light water reactor (LWR) fuel performance evaluation code ‘FEMAXI-6’, FEMAXI-FBR develops specific models for the fast reactor fuel performance, such as restructuring, material migration during steady state and transient, melting cavity formation and pressure during accident, so that it can evaluate the fuel failure during accident. The analysis of test pin with slow transient over power test of CABRI-2 program was conducted from steady to transient. The test pin was pre-irradiated and tested under transient overpower with several % P ₀/s (P ₀: steady state power) of the power rate. Analysis results of the gas release ratio, pin failure time, and fuel melt radius were compared to measured values. The analysis results of the steady and transient performances were also compared with the measured values. The compared performances are gas release ratio, fuel restructuring for steady state and linear power and melt radius at failure during transient. This analysis result reproduces the measured value. It was concluded that FEMAXI-FBR is effective to evaluate fast reactor fuel performances from steady state to accident conditions.     

Influence of pellet length on interaction between fuel and cladding in simulated reactor fuel pins

An electrically heated fuel pin test apparatus has been developed for out-of-pile investigations of fuel pin parameters with a view to supplementing in-pile experiments. Sixty per cent of reactor heat ratings has been achieved with a hollow pin having an axially located electrical heater, the limitation being the melting of the UO₂ pellets. The theoretical unconstrained shapes of a heated pellet and a fuel pellet under elastic conditions were calculated. Both showed an ‘hour glass’ form suggesting that permanent circumferential ridges would occur in the cladding of a heated pin as they do in the cladding of fuel pins. These ridges were subsequently produced in heated pins, the pins being heated while immersed in cooling water at typical reactor temperatures and pressures. From a series of such tests using different pellet lengths it was found that a significant reduction in ridge height occured when the pellet ratio was one-third of the value in a typical reactor. The temperatures reached in the UO₂ pellets were estimated from a metallographic examination of a pin cross section after test. Using published data of ∫kdT for UO₂ over various temperature ranges the pin heat output at that cross section was determined.     

Analysis of temperature noise propagation in a large fuel subassembly of FBR

We have developed a method to calculate the three-dimensional distribution of root-mean-square (RMS) values of temperature noise in the single phase flow in a fast reactor fuel subassembly with a local flow blockage. Employed are the subchannel method in a pin bundle region and the finite difference method in the region downstream of the bundle. We have compared the calculated RMS values of temperature noise with experimental data for a sodium loop test using a wire-spacered 91-pin-bundle fuel sub-assembly with a local blockage. We have investigated the possibility of detection of the blockage by temperature noise by taking into account the influence of structures in the upper part of the subassembly.