The effects of reactor coolant system pipe rupture motion on tributary piping and attached equipment

This paper presents a rigorous analysis of a pressurized water reactor coolant system (RCS) to determine time-history excitations of intact equipment and tributary piping attached to the RCS caused by a postulated guillotine rupture in the primary coolant piping. Reactor control rods and drive mechanisms, in core instrumentation guide tubes and reactor coolant pump motor appurtenances are examples of attached equipment which is excited by RCS LOCA induced motions. The surge line, mainsteam lines and emergency core cooling lines are examples of tributary piping similarly affected by RCS LOCA induced motions. The methods described herein include structural and dynamic modeling and analytical techniques used in the non-linear transient dynamic time-history analysis of a 3-D coupled model of the RCS. The results of this analysis are generated for the purpose of defining the excitation for subsequent analysis of intact tributary systems attached to the reactor coolant system in order to evaluate their response to those LOCA induced motions. This paper also presents results of analyses for intact tributary piping subjected to LOCA induced motions and assesses the severity of the response compared to typical seismic response.     

Development of FBR piping bellows joints in Japan

For the practical use of fast breeder reactors (FBRs) reduction of construction costs is one of the most important factors. If the long and winding route of piping systems (needed to absorb the thermal expansion) can be shortened and simplified, sharp reductions in related apparatus, equipment and reactor building etc. can be expected (especially in the case of loop type FBRs). The use of bellows joints, which possess good ability to absorb thermal expansion, is one of the best means of shortening the piping system. From 1983 to 1988, the Power Reactor and Nuclear Fuel Development Corporation promoted extensive research and development on FBR piping bellows joints, which covered areas such as strength evaluation methods, manufacturing and inspection techniques, maintenance and repair techniques, investigation of safety logic etc. The purpose of that work was to ensure that the application of bellows joints to FBR main piping systems was a technical and practical possibility. The conclusion was that the use of FBR piping bellows joints was feasible. Consequently, both draft structural design rules and draft manufacturing and maintenance rules were formulated based on the results. This paper presents a summary of the program and the results of the research and development.     

Thermal stratification and reactor piping integrity

Thermal stratification, cycling, and striping phenomena have drawn much attention recently because of the incidents at several nuclear plants that raised significant safety concerns. The concerns due to these phenomena relate to thermal fatigue in branch pipes connected to the main coolant piping. Nuclear utility industry is addressing the issue with the aim to understand the mechanisms that lead to fatigue in nominally stagnant piping systems near the reactor coolant piping. Two key results from this effort are described in this paper. First, tests to investigate the interaction between the main coolant piping and the stagnant attached lines by turbulence penetration are described and a working correlation is obtained. Turbulence penetration into unisolable lines, or the transport of turbulence into stagnant piping from the reactor coolant system (RCS) line, represents a mechanism for carrying hot RCS water into regions filled with cold water. The possibility of stratification of the two fluids (and the resultant thermal stresses) is the reason for developing an understanding of the turbulence penetration process. Secondly, results of an evaluation to develop a loading definition for thermal striping are included. Based on this testing several important conclusions relating to fatigue in nominally static reactor coolant systems are reached.     

压水堆主回路失水事故下水锤与管道结构的相互作用分析

传统的水锤分析和管道动力响应计算是分开的,存在一定的缺陷。本文针对核电站主回路假想双端断裂时系统的受力和力矩分析这一问题,对破裂管道分充体和管道的耦合机制,引入描述流体－管道单元的１４个参数和１４个偏微分方程,利用特征线法对水锤和管道结构的相互耦合作用进行了模拟计算。计算得到了更为准确的水锤波和管道的受力和力矩,其波形和数值均与不考虑耦合作用时有所不同。这些计算结果为压水堆核电站的核安全设计和分析     

波动管设计改进对反应堆厂房地震响应的影响分析

为了减轻由于稳压器波动管热分层引起的热疲劳效应及降低安装难度,提出了在M310型压水堆稳压器中加大波动管与主回路夹角的布置改进方案.波动管的布置改进将引起反应堆厂房内部结构布置的改变,对地震响应产生影响.根据设计改进重建结构计算模型,进行抗震分析,并与旧模型的相应结果进行对比,探讨设计改进对反应堆厂房地震响应的影响,为改进方案的论证提供参考.     

Response of piping system on friction support to bi-directional excitation

Installation of friction devices between a piping system and its supporting medium is an effective way of energy dissipation in the piping systems. In this paper, seismic effectiveness of friction type support for a piping system subjected to two horizontal components of earthquake motion is investigated. The interaction between the mobilized restoring forces of the friction support is duly considered. The non-linear behavior of the restoring forces of the support is modeled as an elastic-perfectly plastic system with a very high value of initial stiffness. Such an idealization avoids keeping track of transitional rules (as required in conventional modeling of friction systems) under arbitrary dynamic loading. The frictional forces mobilized at the friction support are assumed to be dependent on the sliding velocity and instantaneous normal force acting on the support. A detailed systematic procedure for analysis of piping systems supported on friction support considering the effects of bi-directional interaction of the frictional forces is presented. The proposed procedure is validated by comparing the analytical seismic responses of a spatial piping system supported on a friction support with the corresponding experimental results. The responses of the piping system and the frictional forces of the support are observed to be in close agreement with the experimental results validating the proposed analysis procedure. It was also observed that the friction supports are very effective in reducing the seismic response of piping systems. In order to investigate the effects of bi-directional interaction of the frictional forces, the seismic responses of the piping system are compared by considering and ignoring the interaction under few narrow-band and broad-band (real earthquake) ground motions. The bi-directional interaction of the frictional forces has significant effects on the response of piping system and should be included in the analysis of piping systems supported on friction supports. Further, it was also observed that the velocity dependence of the friction coefficient does not have noticeable effects on the peak responses of the piping system.     

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.     

Development of flow-induced vibration evaluation methodology for large-diameter piping with elbow in Japan sodium-cooled fast reactor

This paper describes the current status of flow-induced vibration evaluation methodology development for the primary piping in Japan sodium-cooled fast reactor, with particularly emphasis on the development approach and research activities that investigate unsteady hydraulic characteristics in a short-elbow piping. The approach to the methodology development was defined: experiment-based methodology and simulation-based one as well as extrapolation logic to the reactor condition based on no dependency on Reynolds number in the high Reynolds number range from the experimental results. Experimental efforts have been made using 1/3-scale single-elbow test sections for the hot-leg piping as the main activity. Recent experiments using the 1/3-scale test section revealed that a swirl flow at the inlet of the hot-leg piping hardly influenced pressure fluctuations onto the pipe though a slight deformation of flow separation was observed. Numerical results under different Reynolds number conditions appear in this paper using the unsteady Reynolds Averaged Navier Stokes equation approach, indicating its applicability to the hot-leg piping experiments.     

核电站反应堆冷却剂系统主管道安装焊接

以秦山核电二期工程为例,论述了核电站反应堆冷却剂系统主管道安装焊接技术及质量控制要点,并对反应堆冷却剂系统主管道的安装顺序、安装技术要求、焊接质量检验方法以及焊接变形的控制等方面给予了详细的阐述,对核电站反应堆冷却剂系统主管道安装焊接及质量控制具有借鉴作用。     

Seismic analysis of ZPR-6 Reactor Facility

A safe shutdown earthquake analysis of ZPR 6 Reactor Facility was conducted through seismic risk analysis, soil-structure interaction analysis, reactor building dynamic time history analysis and equipment response spectrum analysis due to an assumed El Centro earthquake. Several ASME, AISC and ANSI design codes were used to demonstrate the adequacy of this facility and to design several equipment and piping supports.     

On estimated modes of failure of nuclear power plants by potential earthquakes

This paper deals with the estimated ‘modes of failure’ of nuclear power plants during future violent earthquakes. The authors have been surveying the damage to industrial plants caused by several violent earthquakes since 1960. Some of them have already been reported in English, but here the authors try to rearrange them from the viewpoint of ‘modes of failures’ of nuclear power plant buildings, equipment, vessels and piping. The authors categorize the mechanisms of failure as follows: (i) damaged by the dynamic effect of acceleration waves, (ii) by resonance in displacement waves, (iii) by the static effect of seismic force, (iv) by external force from attached piping and others, or forced deformation, and (v) by liquefaction of soil.The authors try to determine the modes of failure of the following items in a matrix form of the mechanisms: (i) the reactor building, (ii) steel containment vessel, (iii) auxiliary building, (iv) reactor vessel, (v) core internals, (vi) primary and secondary coolant system, (vii) emergency power supply system, (viii) emergency gas treatment system and stack, (ix) fuel cooling pond and fuel rack, (x) refuel machine crane, (xi) auxiliary system and component, (xii) turbine and its pedestal, and (xiii) main power system and control instrumentation. They also examine them from another point of view, i.e. in ‘the classification of the important factor’ of items for their aseismic design.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.     

Opening and extension of circumferential cracks in a pipe subject to dynamic loads

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Erratum

Nuclear power plants are presently designed to withstand instantaneous pipe severance in combination with the maximum seismic loads. The hypothetical combination of these two unlikely events leads to system designs which are very expensive and require dynamic event devices such as pipe whip restraints which have the potential for deleterious interaction with the piping system during normal operations. These present pipe rupture criteria are based on the a priori hypothesis that the instantaneous guillotine pipe break is possible, rather than from a consideration of the manner in which cracks might open or extend in a real piping system. The objective of this study is to help establish the basis for understanding how cracks which might exist in the primary piping of a pressurized water reactor (PWR) would open and extend so that improved criteria can be developed based on this information.One of the regions where loss of pressure boundary integrity must be postulated is the terminal end of the cold leg at the reactor vessel inlet nozzle. This region (including the effects of the reactor vessel and the primary pump) is modelled for analysis with the MARC general purpose finite element program. A circumferential crack, one-half circumference long, is considered to suddenly occur around the outside of the elbow when the pipe is at normal operating pressure. The most severe part of the safe shutdown earthquake (SSE) loading transient is applied simultaneously with the initiation of the crack.The plastic dynamic analysis of the crack opening effects in the discharge leg pipe is performed using the MARC program until the maximum opening occurs. The J-integral plastic crack extension criterion is computed for all times during the transient. The results indicate that none of the cracks will extend significantly and that the opening areas are small fractions of the flow area of the pipe.     

Simple Method for Evaluating Mesh Size Effects in Control Rod Worth Analysis

It has been pointed out that the reactor coolant system piping could fail prior to the meltthrough of the reactor pressure vessel in a high pressure sequence of pressurized water reactor severe accidents. In order to apply to the evaluation of the piping failure which influences the subsequent accident progression, models for the strength of piping materials at high temperatures were examined. It was found that 0.2% proof stress and ultimate tensile strength above 1,073 K obtained from tensile tests was reproduced by a quadratic equation of the reciprocal absolute temperature. Short-term creep rupture time and minimum creep rate at high temperatures were well correlated by the modified Norton's Law as a function of stress and temperature, which implicitly expressed the effect of the precipitation and the resolution of precipitates on the creep strength. The modified Norton's Law gave better results than the conventional Larson-Miller method. Relating applied stress vs. minimum creep rate and tensile properties vs. applied strain rate obtained from the creep and tensile tests, a temperature range where the dynamic recrystallization significantly occurred was evaluated.     

Sensitivity studies of modular high-temperature gas-cooled reactor postulated accidents

The results of various accident scenario simulations for the two major modular high temperature gas-cooled reactor (HTGR) variants (prismatic and pebble bed cores) are presented. Sensitivity studies can help to quantify the uncertainty ranges of the predicted outcomes for variations in some of the more crucial system parameters, as well as for occurrences of equipment and/or operator failures or errors. In addition, sensitivity studies can guide further efforts in improving the design and determining where more (or less) R&D is appropriate. Both of the modular HTGR designs studied – the 400-MW(t) pebble bed modular reactor (PBMR, pebble) and the 600-MW(t) gas-turbine modular helium reactor (GT-MHR, prismatic) – show excellent accident prevention and mitigation capabilities because of their inherent passive safety features. The large thermal margins between operating and “potential damage” temperatures, along with the typically very slow accident response times (approximate days to reach peak temperatures), tend to reduce concerns about uncertainties in the simulation models, the initiating events, and the equipment and operator responses.     

Preoperational vibration test of a primary heat transport system of the Canadian deuterium uranium reactor

During preoperational tests of Wolsong-2 nuclear power plant (NPP) which is a Canadian deuterium uranium (CANDU) reactor, vibration measurements were made on the primary heat transport (PHT) system. These measurements were evaluated by spectral methods to determine modal displacements and the modal stresses induced in the piping. The main aim of the measurement program is to confirm that the structural systems and components are adequately dimensioned for the operational vibration loads during the design life of the reactor. Structural analysis of the PHT system was performed using computer code to determine modal displacements and modal stresses analytically. The measurement results were compared with the analytically calculated resonance frequencies and modal values. A reasonable correspondence of the test and analytical results was achieved. The stresses evaluated on the PHT system were below the endurance limit for the material with a margin of safety.     

AP1000核电厂二回路主管道双端断裂流体喷射力计算分析

AP1000是先进的第三代压水堆核电厂,为确保核电厂在事故工况下的安全性,需对二回路主管道发生双端断裂的工况进行研究。本文采用RELAP5/MOD3.4软件对核电厂二回路突发主管道双端断裂的事故工况进行了数值模拟,计算得到断裂后管道破口处的喷放流量、压强、空泡份额及喷射力等物理参数的变化特性,并将计算结果与ANSI 58.2简化计算方法的结果进行了比较分析。结果表明,RELAP5/MOD3.4计算所得的喷射力小于简化计算方法所得结果。本文分析结果为进行AP1000核电厂的破裂管道甩击防护提供了基础。     

Towards an elastic-plastic fracture mechanics predictive capability for reactor piping

Intergranular stress corrosion cracks have been discovered in the recirculation bypass piping and core spray lines of several boiling water reactor (BWR) plants. These cracks initiate in heat-affected zones of girth welds and grow circumferentially by combined stress corrosion and fatigue. Reactor piping is mainly type 304 stainless steel, a material which exhibits high ductility and toughness. A test program described in this paper demonstrates that catastrophic crack growth in these materials is preceded by considerable amounts of stable crack growth accompanied by large plastic deformation. Thus, conventional linear elastic fracture mechanics, which only applies to the initiation of crack growth in materials behaving in a predominantly linear elastic fashion, is inadequate for a failure analysis of reactor piping.This paper is based upon research initiated by a need to develop a realistic failure prediction and a way to delineate leak-before-break conditions for reactor piping. An effective engineering solution for the type of cracks that have been discovered in BWR plants was first developed. This was based upon a simple net section flow stress criterion. Subsequent work to develop an elastic-plastic fracture mechanics methodology has also been pursued. A survey of progress being made is described in this paper. This work is based on the use of finite element models together with experimental results to identify criteria appropriate for the onset of crack extension and for stable crack growth. A number of criteria have been evaluated. However, the optimum fracture criterion has not yet been determined, even for conditions which do not include all of the complications involved in reactor piping.