Comparison of computer codes for evaluation of double-supply-frequency pulsations in linear induction pumps

The paper describes and compares three computer codes that are able to estimate the double-supply-frequency (DSF) pulsations in annular linear induction pumps (ALIPs). The DSF pulsations are the result of interaction of the magnetic field and induced in liquid metal currents both changing with supply-frequency. They may be of some concern for electromagnetic pumps (EMP) exploitation and need to be evaluated at their design. The results of computer simulation are compared with experimental ones for annular linear induction pump ALIP-1.     

Reliability analysis for stiff versus flexible piping

The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis.A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. We then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. We concluded that these changes substantially reduce calculated piping responses and allows piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design.An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Our result indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design.     

Predicting local distributions of erosion–corrosion wear sites for the piping in the nuclear power plant using CFD models

The erosion–corrosion (E/C) wear is an essential degradation mechanism for the piping in the nuclear power plant, which results in the oxide mass loss from the inside of piping, the wall thinning, and even the pipe break. The pipe break induced by the E/C wear may cause costly plant repairs and personal injures. The measurement of pipe wall thickness is a useful tool for the power plant to prevent this incident. In this paper, CFD models are proposed to predict the local distributions of E/C wear sites, which include both the two-phase hydrodynamic model and the E/C models. The impacts of centrifugal and gravitational forces on the liquid droplet behaviors within the piping can be reasonably captured by the two-phase model. Coupled with these calculated flow characteristics, the E/C models can predicted the wear site distributions that show satisfactory agreement with the plant measurements. Therefore, the models proposed herein can assist in the pipe wall monitoring program for the nuclear power plant by way of concentrating the measuring point on the possible sites of severe E/C wear for the piping and reducing the measurement labor works.     

Three-Dimensional Analysis Method for Sloshing Behavior of FBR and Its Application to Uni- and Multi-Vessel Type FBRs

Abstract

A three-dimensional analysis method for sloshing behavior of FBR is developed. The method treats the coolant in a reactor vessel as a potential flow with moving liquid surfaces. The Laplace equation of velocity potential is solved by a boundary element method with its boundary conditions described by a Bernoulli equation.

The method is applied to analysis of sloshing behavior of uni- and multi-vessel type FBRs and results are presented.

The latter consists of vessels for the core, heat exchangers and pumps, all of which are connected by piping. In the uni-vessel type, heat exchangers and pumps are placed in the reactor vessel.     

Possible design improvements and a high power density fuel design for integral type small modular pressurized water reactors

The study evaluates potential weaknesses and possible improvements for integral type small modular pressurized water reactor designs. By taking International Reactor Innovative and Secure (IRIS) as the reference design and keeping the power output as the same, a new fuel and reactor design were proposed. The proposed design relocates the primary coolant pumps and the pressurizer outside the reactor pressure vessel (RPV). Three recirculation lines and jet pumps/centrifugal pumps are introduced to provide the coolant circulation similar to Boiling Water Reactor designs. The pressurizer component is expected to be similar to the AP600 design. It is located at one of the recirculation lines. The new fuel assembly adopts 264 solid cylindrical fuel pins with 10 mm diameter and 2.3 m height, arranged at a hexagonal tight lattice configuration. Large water rods are introduced to preserve the moderating power and to accommodate finger type control rods. The resulting fuel can operate with 104.5 kW/l power density while having substantially higher margin for boiling crisis compared to typical large PWRs. Full core neutronic analysis shows that 24-month cycle length and 50 MWd/kg burnup is achievable with a two-batch refueling scheme. Furthermore, the fuel behavior study shows that the new fuel with M5 type Zircaloy cladding show fairly acceptable steady state performance. A preliminary Loss of Coolant analysis shows that the new design could be advantageous over IRIS due to its low ratio of the water inventory below the top of the active fuel to total RPV water inventory. The proposed reactor pressure vessel height and the containment volume are 30% lower than the reference IRIS design.     

A two-phase methodology to predict FAC wear sites in the piping system of a BWR

Flow accelerated corrosion (FAC) wear is a serious degradation problem especially for nuclear power plants since it may result in the plant damage as well as risk to personnel. In this paper, a methodology which includes the two-phase hydrodynamic CFD models and FAC models, is proposed to predict severe FAC wear sites. Based on hydrodynamic simulation results, the present CFD models can precisely capture the two-phase characteristics within the piping system, which include the centrifugal effect, the gravitation effect and the imbalance of phase and mass separation in a T-junction, etc. Coupled with these flow characteristics, the appropriate FAC indicators can predict the possible locations of severe FAC wear. This methodology was validated against the measured results of wear site distributions for the piping system in a boiling water reactor (BWR) power plant. Good agreement between measurements and predictions at severe wear sites indicate that the present models can capture the characteristics of severe FAC wear and can help assist in the pipe wall-monitoring program for a nuclear power plant.     

Determination of cycle-dependent material properties for a design concept on the basis of limited plastic strains

For unlikely dynamic service conditions, such as pressure surges or earthquakes, analysis of piping integrity in terms of strains is appropriate. Strain controlled tensile/compressive fatigue tests are meant to indicate whether the limited plastic strains caused during the few load cycles occurring as a result of these service conditions shall be admissible. The fatigue behaviour is influenced by parameters, such as frequency, resp. strain rate, temperature and material, resp. material condition and should be taken into these considerations. With the help of the experimentally determined cycles to fracture or crack initiation, it is possible to extend design curves in the technical code in the area of low cycles to fracture, e.g. crack initiation (N_I < 100).     

Hydrodynamic characteristics of lead and lead-bismuth coolants

The results of an analysis of the effect of the physical properties of lead and lead-bismuth coolants on the hydrodynamic characteristics and the results of experimental investigations of the particulars of the hydrodynamic flows of these coolants in application to loops with fast reactors are presented. It is shown that cavitation, in the conventional meaning of this word, cannot arise in flow part of vane pumps pumping lead and lead-bismuth coolants in a reactor loop. It is confirmed that a gas gap can form between the surface of a heavy liquid-metal coolant flow and the channel walls not wetted by it. The results of experimental studies of the rupture of a column of heavy liquid-metal coolant and detachment of a centrifugal pump flow, probably because of the appearance of gas cavitation, are presented.     

The steamhammer problem: dynamic shock loading of critical reactor piping systems

In the course of both pre-operational testing and power operation of commercial nuclear power plants, relatively large amplitude transient vibrations of steam piping systems have been experienced with damage to the piping supports in at least one recent case. These transient vibrations result from ‘steamhammer’ or dynamic shock loading induced by pressure and momentum transient conditions generated in the piping by sudden changes to the flow conditions, such as are produced by sudden valve opening or closure. In particular, vibrations have been experienced in by-pass and discharge lines as a result of relief valve discharge, and in main steam lines as a result of sudden main stop valve closure. Piping in both BWR and PWR reactor systems has been found to be susceptible to these conditions.This paper is concerned with the evaluation of the pressure and momentum transients resulting from sudden valve operation, and the determination of the dynamic response of the piping to the induced transient loading. The characteristics of the transient conditions existing immediately following both sudden valve opening and closure as encountered in BWR and PWR plants are discussed. The procedures used to calculate the transient time history functions are outlined. The derivation of the loading induced in the piping by the pressure and momentum transients is discussed and the determination of the dynamic response of the piping is presented. The procedures described in the paper are illustrated by actual examples from BWR and PWR plants, and the significance of steamhammer effects relative to other loading conditions is discussed.     

Three-Dimensional Analysis Method for Sloshing Behavior of Fast Breeder Reactor and Its Application to Uni- and Multi-Vessel Type FBRs

A three-dimensional analysis method for sloshing behavior of fast breeder reactor (FBRs) is developed. The method treats the coolant in a reactor vessel as a potential flow with moving liquid surfaces. The Laplace equation of a velocity potential is solved by a boundary element method with its boundary condition described by a Bernoulli equation.

The vibration test results of a rectangular water pool are calculated by the method. Then, the method is applied to analysis of sloshing behavior of uni- and multi-vessel type FBRs. The latter consists of vessels for the core, heat exchangers and pumps. These vessels are connected by piping. In the case of the uni-vessel type FBR, heat exchangers and pumps are placed in the reactor vessel. The characteristics of sloshing behavior of both the reactors are presented.     

Piping extreme dynamic response studies

Results are presented for a series of high-amplitude dynamic tests of a simple pressurized piping system excited through various multiple piping supports. The four-inch diameter piping achieved response levels above yield when subjected to earthquake-like time history inputs and withstood — without leakage or gross distortion — dynamic inputs that were factors of three to five times greater than those inputs required to just exceed the ASME Class 2 stress limit for Service Level D, the Safe Shutdown Earthquake condition. Despite intentionally induced support failures in several tests, piping pressure integrity was maintained, and no plastic collapse occurred. Selected snubber hardware likewise exhibited large design margins under transient loads.     

Acoustical environment of gas-cooled nuclear reactors

This paper describes methods for acoustical analysis of gas-cooled nuclear reactors in terms of the sources of sound, the propagation of sound about the coolant circuit and the response of reactor structures to sound. Sources of sound that are considered are circulators, jets, vortex shedding and separated flow. Circulators are generally the dominant source of sound. At low frequency the sound propagates one dimensionally through the ducts and cavities of the reactor. At high frequency the sound excites closely spaced two- and three-dimensional acoustic modes, and the resultant sound field can be described only statistically. The sound excites plate and shell structures within the coolant circuit. Secondary steam piping can also be excited by pumps and valves. Formulations are presented for the resultant vibration. Vibration-induced damage is also reviewed.     

压水堆核电站不锈钢主管道铸造

用电弧炉和ＡＯＤ双联冶炼核电站主管道Ｚ３ＣＮ２０－０９Ｍ，并根据Ｓｈａｅｆｆｌｅｒ图计算结果调整Ｚ３ＣＮ２０－０９Ｍ的铁素体含量。在离心铸管工艺中，用加大型筒壁厚，减小挡枝内孔直径，选大的重力加速度ｇ值，增加内孔加工余量等措施铸造出主管道样件，测试结果表明，主管道样件各项性能指标均满足ＲＣＧ－Ｍ的要求。     

Methods for estimating strain in structures - conditions and definitions

Analyses of the impact of actual earthquakes on industrial plants and experimental investigations simulating earthquakes and pressure surges show that loads can be sustained without damage, which are according to the existing design rules far beyond the allowable limit and which cannot be calculated realistically assuming linear-elastic material behaviour according to plastic deformation. To rely on the existing capability to absorb large amounts of energy, it is necessary to have models for describing their behaviour in the plastic range taking into account the plastic deformation volume and the strain values reached.Especially for the calculation of piping systems under loads with a small probability of occurrence, the main parameters for describing the energy dissipated and the additional strain in comparison to the strain in a straight pipe are described, independent of the calculation method used (simplified or elasto-plastic time integration).In addition, strain categories and the individual strain category responsible for the integrity of the overall system are defined.     

管道系统的功能性评定准则研究

管道系统的功能性是不同于管道系统压力边界完整性的一项要求，美国核管理委员会（NRC）提出了管道系统功能性的2种评定准则。为了探讨功能性评定准则的来源以及应用，通过研究经典文献中有关功能性评定准则的内容，阐述了2种评定准则的来历和依据，分析了2种功能性评定准则的特点，指出了使用功能性评定准则的注意事项。通过一个管道系统功能性评定的实例，提出2种功能性评定准则在不同的核电厂设计阶段的应用策略。对于新建的核电厂，尽量使用C级限值来保证管道系统的功能性，如果是已建造的核电厂，则可以用D级限值附加5个条件来保证管道系统的功能性。      

Experiments to evaluate behavior of containment piping bellows under severe accident conditions

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, NM. Several different bellows geometries representative of actual containment bellows are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of 13 tests have been conducted. The tests showed that bellows are capable of withstanding relatively large deformations up to or near the point of full compression before developing leakage. The test data are presented and discussed.     

Hydraulic study on recirculation loops using computational fluid dynamics (CFD). Design optimization and turbulence reduction

A full scope study of the recirculation system flow of a BWR6 nuclear plant is essential, to completing the characterisation and understanding the dynamics of the flow and specially the bistable flow in the recirculation loop.The finite element model serves to study the sensitivity of the design and to reduce turbulence at the connection between the recirculation pump impulsion piping and the manifold. It is found that the problem lies in the poor hydraulic design of the manifold. As a result of this analysis, in order to reduce turbulence and balance flows in the jet pumps the manifold diameter needs to be increased. The simulation shows the unbalanced flow to the jet pump pipes. This fact can be checked in any nuclear power plant BWR.     

中国一体化反应堆核电厂创新安全壳设计研究

中国一体化反应堆核电厂(CIP)是中国核反应堆系统设计技术国家重点实验室正在开发的新一代革新型、完全一体化的压水堆,其电功率约为300 MW.CIP采用堆内一体化布置,反应堆冷却剂系统设备以及控制棒驱动机构全部布置在反应堆压力容器内.这种一体化设计消除了传统的冷却剂回路管道,消除了大LOCA事故,具有更高的安全性.本文介绍了CIP安全壳系统方案选择、安全壳设计、安全壳设计压力的确定以及安全壳结构的计算分析.     

Finite element-based limit load of piping branch junctions under combined loadings

The limit load is an important input parameter in engineering defect-assessment procedures and strength design. In the present work, a total of 100 different piping branch junction models for the limit load calculation were performed under combined internal pressure and moments in use of non-linear finite element (FE) method. Three different existing accumulation rules for limit load, i.e., linear equation, parabolic equation and quadratic equation were discussed on the basis of FE results. A novel limit load solution was developed based on detailed three-dimensional FE limit analyses which accommodated the geometrical parameter influence, together with analytical solutions based on equilibrium stress fields. Finally, six experimental results were provided to justify the presented equation. According to the FE limit analysis, limit load interaction of the piping tees under combined pressure and moments has a relationship with the geometrical parameters, especially with the diameter ratio d/D. The predicted limit loads from the presented formula are very close to the experimental data. The resulting limit load solution is given in a closed form, and thus can be easily used in practice.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.