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1.
Pipe whip tests or jet discharge tests have been performed at the Japan Atomic Energy Research Institute, which simulate the instantaneous circumferential guillotine break of primary coolant piping in nuclear power plants. The present paper describes the results of the pipe whip tests using test pipes of 4 inch diameter, under the BWR LOCA conditions, which were performed from 1979 to 1981. The tests were carried out at an initial pressure of about 6.8 MPa and an initial temperature of about 285°C.The test pipe was 114.3 mm (4 in) in diameter, 8.6 mm in thickness and 4500 mm in length. The four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from Type 304 stainless steel. The experimental parameters were the clearance (30, 50 and 100 mm) and the overhang length (250, 400, 550 and 1000 mm).The main purpose of these tests is to investigate the effects of the clearance and the overhang length on the pipe whip behavior. It has been clarified from the test results that a smaller clearance and a shorter overhang length causes the deformation of the pipe and restraints to be minimized, and the test pipe collapses near the setting point of the restraints with the overhang length of 1000 mm.  相似文献   

2.
The experimental program performed on AQUITAINE-II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It is jointly financed by the Commissariat à l'Energie Atomique, Framatome, Electricité de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S-or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: (1) the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; (2) impact function of a pipe on an energy-absorbing bumper; (3) lateral stability of both ends of a pipe, after a double-guillotine break.  相似文献   

3.
The analytical results of blowdown characteristics and thrust forces were compared with the experiments, which were performed as pipe whip and jet discharge tests under the PWR LOCA conditions. The blowdown thrust forces were obtained by Navier-Stokes momentum equation for a single-phase, homogeneous and separated two-phase flow, assuming critical pressure at the exit if a crifical flow condition was satisfied.The following results are obtained:
1. (1) The node-junction method is useful for both the analyses of the blowdown thrust force and of the water hammer phenomena.
2. (2) The Henry-Fauske model for subcooled critical flow is effective for the analysis of the maximum thrust force under the PWR LOCA conditions. The jet thrust parameter of the analysis and experiment is equal to 1.08.
3. (3) The thrust parameter of saturated blowdown has the same one with the value under pressurized condition when the stagnant pressure is chosen as the saturated one.
4. (4) The dominant terms of the blowdown thrust force in the momentum equation are the pressure and momentum terms except that the acceleration term has large contribution only just after the break.

References

[1]M. Okazaki et al., Preprint of two phase flow meeting, JSME (1980), pp. 85–88 (in Japanese).[2]F.J. Moody, ASME 69HT31 (1969).[3]F.J. Moody, Fluid reaction and impingiment loads, Nuclear Power Plants (1973), pp. 219–261.[4]B.R. Strong and R.J. Baschiere, Nucl. Engrg. Des. 45 (1978), pp. 419–428. Abstract | PDF (543 K) | View Record in Scopus | Cited By in Scopus (0)[5]RELAP4/MOD5, ANCR-NUREG-1335 (1976).[6]PRTHRUST, Nuclear Service Co..[7]N. Miyazaki et al., Nucl. Engrg. Des. 64 (1981), pp. 389–401. Abstract | PDF (806 K) | View Record in Scopus | Cited By in Scopus (0)[8]W.H. Retting et al., IN-1321 (1970).[9]M. Hsu et al., Nucl. Technology 53 (1981), pp. 58–63.[10]R.E. Henry and H.K. Fauske, Journal of Heat Transfer, Trans. ASME, Ser. C93 (1971), pp. 179–187. Full Text via CrossRef[11]F.J. Moody, Journal of Heat Transfer, Trans. ASME, Ser. C93 87 (1965), pp. 134–142.[12]N. Miyazaki et al., 1981 Fall Meeting Reactor Phys. and Eng., At. Energy Soc. Japan, Paper D58 (1981) (in Japanese).[13]K. Namatame and K. Kobayashi, Journal of Heat Transfer, Trans. ASME, Ser. C 98 (1976), pp. 12–18. Full Text via CrossRef | View Record in Scopus | Cited By in Scopus (0)[14]M. Sobajima, Nucl. Sci. Engrg. 60 (1976), pp. 10–18. View Record in Scopus | Cited By in Scopus (0)[15]R.D. Jain and G.A. Hastings, Trans. Ame. Nucl. Soc. 21 (1975), pp. 345–346.  相似文献   

4.
5.
A probability-based approach is presented as the integration of probabilistic methods and deterministic modelling based on the finite element method. An existing finite element software package was linked to an existing probabilistic package to analyse the complex mechanics that occur during the transient non-linear analysis of impact problems. This methodology is applied to a pipe whip analysis of a group-distribution-header, which results from a guillotine break, and subsequent impact with the adjacent building wall; this is a postulated accident for the Ignalina Nuclear Power Plant RBMK-1500 reactors. The uncertainties of material properties, component geometry data and loads were taken into consideration. The probabilities of failure of the impacted header and of the header support-wall were estimated given uncertainties in material properties, geometrical parameters and loading. The software ProFES was used for the probabilistic analysis and the finite element software NEPTUNE for deterministic structural integrity evaluation. The Monte Carlo Simulation, First Order Reliability method and Response Surface method were used in the probabilistic analysis.  相似文献   

6.
Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles due to the nonuniform water accumulation in the upper plenum.  相似文献   

7.
A method is described for calculating fuel rod cladding temperatures in a blockage formed by a group of ballooned fuel rods in a larger rod array, for heat transfer conditions appropriate to the reflooding phase of a postulated PWR LOCA. The model is suitable for describing the extreme case of co-planar axially extended balloons, where steam superheating and skin friction effects are believed to have an important effect on blockage heat removal. Attention is restricted to the constricted zone within the blockage.Reasonable agreement is shown with available heat transfer data from partially ballooned rod arrays, for conditions of steam cooling, steam-and-droplet cooling and reflood cooling. The model is also able to describe flow velocity distribution data from partially blocked rod bundles with reasonable accuracy.Parametric calculations for typical PWR LOCA heat transfer conditions suggest that blockage length has a strong effect on fuel coolability, mainly as a result of extra superheating of the steam within the blockage. However calculations also indicate that the presence of entrained water droplets has a powerful effect in reducing the clad temperatures attained.  相似文献   

8.
The NET cooling systems for in-vessel components and vessel are generally based on low pressure and low temperature water. However, the cooling loops for the breeder blanket are intended to operate at a water temperature of about 250°C. A pipe break in a loop with such data would pressurize the compartment where the break takes place. Therefore, as a basis for proper compartment design, it is important to analyze possible pressure increases following pipe breaks. It may also be necessary to introduce equipment for pressure relief or pressure suppression. The objective of the parameter study presented is to determine the relationship between allowed maximum containment pressure following postulated large pipe break in breeding blanket loop and required containment volume. Parameters varied are: blanket loop temperature and pressure (within the range of burn and baking), and pressure suppression system inclusion/exclusion. The analysis has been performed by means of the Swedish containment code COPTA. The results of the analysis are summarized in a plot showing the influence of the varied parameters on required containment volume. In addition, the results presented include required vent areas, heat sink capacities, etc.  相似文献   

9.
A pipe whip restraint utilizing the dissipative capacity of a plastically deforming element has been developed. The transient dynamic behavior of the pipe and the restraint is studied using a discrete mathematical model consisting of nonlinear spring, mass, and damper elements. The forcing function is a fluid dynamic transient resulting from a break in a main steam line which is part of a Westinghouse pressurized water reactor (PWR) nuclear plant. An extensive parametric study shows the relative influence of gap, restraint support rigidity, the elastic and plastic behavior of pipe, break opening time, and finally the change in restraint mass on the behavior of the pipe restraint system.  相似文献   

10.
A model to calculate local heat transfer coefficients between the containment atmosphere and the walls of a pressurized water reactor containment building after a loss-of-coolant accident has been developed. The new calculation is based on the containment wall and atmosphere bulk temperatures, mass ratio of steam to air, and condensing or convective heat transfer conditions. Comparison with the Carolinas Virginia Tube Reactor Containment Tests shows good agreement. The model has been implemented into the containment code TECAR.  相似文献   

11.
压水堆失水事故最佳估算方法研究   总被引:3,自引:1,他引:3  
传统使用的失水事故分析模型和方法被公认是极度保守的,它带来不必要的过量裕度,限制了运行核电厂和新建核电厂的功率提高,并限制了运行的灵活性。最佳估算方法的发展和应用为消除这些不必要的限制提供了可能。本文介绍了压水堆失水事故最佳估算方法的进展;叙述了最佳估算方法及评价方法,特别是不确定性分析方法,介绍了目前已获使用的最佳估算程序。  相似文献   

12.
为研究压水堆核电厂失水事故(LOCA)后杂质在堆芯燃料组件内的沉积现象及压头损失,本试验搭建相应台架,分析了极限工况下碎片在组件中的分布和堵塞情况,定量化评估LOCA后安全壳内碎片对燃料组件压降的影响。结果表明,碎片几乎都堆积在组件下半段尤其是下管座;碎片量相同时,碎片同时添加方案比依次添加方案造成的压降更大;化学沉淀物对碎片床有压实效应,可造成更显著的压头损失;即使极限工况仍有足够的冷却剂进入堆芯排出余热。   相似文献   

13.
This report describes the results of the jet discharging experiments conducted at the Japan Atomic Energy Research Institute. The tests were done under BWR and PWR Loss of Coolant Accident conditions using 4 inch, 6 inch and 8 inch test pipes, and varying distance between the pipe exit and the target plate.Simple and practical experimental formulae to estimate the maximum pressure on the target plate and maximum pressure distribution are given. Further, relations between pipe reaction thrust forces and jet impingement forces are described.  相似文献   

14.
It is important to take flattening of pipe into consideration in order to obtain pipe deformation due to pipe whip loading. An experimental relationship between the flattening of pipe and the pipe surface strain was used to derive the moment-rotation relationship of whipping pipe. The derived moment-rotation relationship was used to calculate the pipe strain in the pipe whip tests using a simplified energy balance method. A comparatively good agreement was obtained between the analytical and experimental results.  相似文献   

15.
The nonlinear dynamic finite element solution of pipe whip problems is presented. The finite element modelling used, the step-by-step incremental solution of the nonlinear equations of motion and design considerations are discussed. The influence of various physical parameters on the response of the pipe and the restraint, and the effects of using different finite element models are considered. Specific emphasis is directed to the verification of the accuracy of the solutions obtained using energy balance checks.  相似文献   

16.
Prestressed concrete pressure vessels include structural members which rely for their safety on the knowledge of the capacity of concrete to resist shear stresses. The shear forces developed on some of the structural components under operating conditions may be several times higher than those occurring in conventional structures. This paper reviews the experimental and analytical studies which have been initiated over the years by various organisations and research workers to examine the problem of high shear stresses and the ultimate load behaviour of end slabs for PCPV. The experimental programmes were carried out to assist in the development of various designs of PCPV and to introduce a more analytical approach for designs of deep members against shear failures.  相似文献   

17.
This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.  相似文献   

18.
先进压水堆采用非能动安全壳冷却系统作为事故后安全壳排热手段,事故后以钢安全壳为换热面将释放到安全壳的能量传递到环境中。失水事故后非能动安全壳冷却系统带热能力的好坏关系到整个反应堆的安全,事故进程中反应堆冷却剂系统的非能动特性与安全壳的非能动特性相互耦合,需要将非能动安全壳冷却系统和反应堆冷却剂系统进行耦合分析,了解事故后反应堆冷却剂系统与安全壳的耦合特性。本文通过开展大破口失水事故下反应堆冷却剂系统和安全壳的耦合分析,了解各非能动系统在大破口失水事故工况下的耦合特性。分析结果显示:大破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性尤其是非能动余热排除系统排热功率、内置换料水箱注入时机和流量、自动卸压阀流量、安全壳压力温度等均与单独计算有较大差异,大破口失水事故下耦合分析得到的事故前期安全壳压力、温度峰值小于单独计算,事故后期安全壳压力在地坑水蒸发的作用下会逐步高于单独计算结果。  相似文献   

19.
小破口失水事故非能动系统瞬态特性研究   总被引:2,自引:2,他引:0       下载免费PDF全文
为了解先进压水堆小破口失水事故下非能动安全壳冷却系统、非能动堆芯冷却系统、非能动余热排出系统的瞬态响应特性,需开展小破口失水事故下反应堆冷却剂系统和安全壳的耦合响应特性研究。分析结果表明,小破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性与独立计算有较大差异,小破口失水事故下耦合分析得到的安全壳压力峰值小于独立计算。  相似文献   

20.
This paper describes a nonlinear dynamic analysis of TVA high energy line pipe whip tests using the ABAQUS-EPGEN code. The analysis considers the effects of large deformation and strain rate on resisting moment and energy absorption capability. The numerical results of impact forces, impact velocities, pipe strains, and reaction forces at pipe supports are compared to the TVA test data. The calculated pipe whip impact time and forces are also compared with those predicted using current industry practice.The calculated pipe support reaction forces are found to be in good agreement with the TVA test data except for some peak values at the very beginning of the pipe break. These peaks are believed to be due to stress wave propagation which cannot be addressed by the ABAQUS code. Both elbow crushing and strain rate have been approximately simulated. The effects are found to be important for pipe whip impact evaluation.  相似文献   

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