ASME Code Research Articles

Review on Effect on Large Opening Structure Stability of Vessel and its Design as per ASME CODE

Review on Effect on Large Opening Structure Stability of Vessel and its Design as per ASME CODE

ATWS analysis for Maanshan PWR using TRACE/SNAP code

The objective of this study was to use TRACE code to analyze the reactor coolant system pressure transients under LONF ATWS for Maanshan PWR. According to Westinghouse Anticipated Transients Without Trip report, LONF ATWS was considered as the most severe plant condition. The ASME Code Level C service limit criteria of 22.06MPa (3200psig) was assumed to be an unacceptable plant condition in SECY-83-293. In accordance with 10 CFR 50.62, Maanshan NPP has established AMSAC. Unlike the reactor trip system, it automatically initiates the auxiliary feedwater system and turbine trip under conditions indicating an ATWS. Since the ATWS analysis is not specified in the FSAR (Taiwan Power Company, 1982), TRACE code was used to assess the RCS pressure for Maanshan NPP. Analysis results indicate that RCS pressure can be maintained within 22.06MPa with sufficient negative moderator temperature coefficient and normal work of AMSAC and valves.

Automated scoping methodology for liquid metal natural circulation small reactor

A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex hull is utilized to quantify the impact of different constraints on the variable scope range.

Flaw Testing of Fiber Reinforced Composite Pressure Vessels

The ASME Boiler Pressure Vessel Project Team on Hydrogen Tanks, in conjunction with other ASME Codes and Standards groups, is developing Code Cases and revisions to the Boiler and Pressure Vessel Code, including such to address the design of composite pressure vessels. The Project Team had an interest in further understanding the effect of cuts to the surface of composite tanks, and how the burst pressure would be affected during the lifetime of the pressure vessel. A test program was initiated to provide data on initial burst pressure, and burst pressure after pressure cycling, of composite cylinders with cuts of different depth. This test program was conducted by Lincoln Composites under contract to ASME Standards Technology LLC, and was funded by National Renewable Energy Laboratory (NREL) [1]. These results were considered during the development and approval of the ASME Code Cases and Code Rules. Thirteen pressure vessels with a design pressure of 24.8 MPa (3600 psi), approximately 0.406 m (16.0 in.) in diameter and 1.02 m (40.2 in.) long, were tested to investigate the effects of cuts to the structural laminate of a composite overwrapped pressure vessel with respect to cycling and burst pressure. Two flaws, one longitudinal and one circumferential, were machined into the structural composite. The flaws were 57 mm long by 1 mm wide (2.25 in. × 0.04 in.) and varied in depth from 10% to 40% of the structural composite thickness of 11.4 mm (0.45 in.). These pressure vessels were cycled to design pressure 0, 10,000, and 20,000 times then burst. The resulting burst pressures were evaluated against the performance of a pressure vessel without flaws or cycling. The burst pressures were affected by depth of cut, but the pressure cycling did not have a significant effect on the burst pressure.

ASME Section III Treatment of Stress Distribution in Cylindrical Vessels With Symmetric Thin-Walled Discontinuity

In equipment design, the designer often attempts to minimize cost and maximize performance while meeting the design specification requirements. In the Power Industry, for example, a significant portion of the design requirements are user defined and the rest are dictated by Codes and Standards. In most applications such as tanks, valves, and pumps, the designers are familiar with the technical issues and are able to meet them. In some cases, however, balancing the two requirements (i.e., the user defined requirements and the ASME Codes and Standards) is complex. An example of this challenge can be found in the design and analysis of a pressurized cylindrical vessel that allows the user to impose a “weak-link” section that under certain operating scenarios can cause the vessel to rupture or fail. In such a case, the designers have to address the discontinuity that is imposed on the structure while still meeting the ASME required stress criteria for structural integrity. The term “weak-link” is used herein to identify the subcomponent or subsystem which is most likely to fail under a postulated operating condition. The loading condition can be either internal or external. In internal loading, the loading is generally an overpressure excursion. In the external case, the load can be seismically induced or an applied actuator load for operation. This paper presents a case study of internally induced loading condition in an ASME Section III cylindrical pressurized vessel where the cylinder is required to fail or rupture only when it is impacted by an actuating load. In this case, the ASME Section III requirements for pressure integrity are combined with the failure mechanism of a symmetrical discontinuity in the design and analysis to ensure that the operational intent of the valve is met. The challenge for this design is two fold; (a) ensure that the device maintains structural integrity without any leakage during normal operation, and (b) guarantee vessel rupture and relieve pressure predictably. This paper presents the ASME treatment of such a design using finite element analysis (FEA).

Strain Rate Calculation Approach in Environmental Fatigue Evaluations

Fatigue usage factor evaluations including the effects of reactor water environment have been performed in numerous nuclear plant license renewal efforts. A large number of these evaluations have used the environmental fatigue penalty factor, Fen, approach prescribed in various regulatory documents. The Fen equations require input of strain rate, but the prescribing documents do not provide methodology or criteria for the quantification of the strain rate to be input. As a result, numerous approaches have been offered and studied. This paper presents an approach used by Westinghouse to include strain rate in an automated calculation of Fen based on the modified rate approach (MRA) to integrated strain rate applications. The starting point of the approach is ASME Code Section III NB-3200 fatigue analysis. With environmental fatigue evaluations in new plant designs now emerging in ASME Code criteria, strain rate considerations remain part of the discussion. The intent of this paper is to provide further insight into this process.

Structural integrity assessment of the reactor pressure vessel under the pressurized thermal shock loading

Fracture mechanics analysis of pressurized thermal shock (PTS) is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV). While the regulation of 10 CFR 50.61 and the ASME Code provide the guidance for the structural integrity, the guidance has been prepared under conservative assumptions. In this paper, the effects of conservative assumptions involved in the PTS analysis were investigated. The influence of different parameters, such as crack size, cladding effect and neutron fluence, were reviewed based on 3-D finite element analyses. Also, the sensitivity study of elastic–plastic approach, crack type and cladding thickness were reviewed. It was shown that crack depth, crack type, plastic effect and cladding thickness change the safety margin (SM) significantly, and the SM at the deepest point of the crack is not always smaller than that of the surface point, indicating that both the deepest and surface points of the crack front should be considered. For the reference transient, deeper cracks always give more conservative prediction. So compared to the prescribed analyses of a set of postulated defects with varying depths in the ASME code, it only needs to assess the crack with maximum depth in the code for the reference transient according to the conclusions.

노즐의 피로해석에 미치는 용접잔류응력의 영향

Abstract Although the fatigue design curve of ASME Code has enough margin with respect to alternating stress and cycles, the welding residual stress(WRS) should be included in fatigue analysis. In this paper, WRS distribution in a nozzle with dissimilar metal weldment was obtained by finite element analysis and was added in fatigue analysis. The fatigue analysis was performed by following the ASME Code including thermal and stress analysis applying with postulated 30 transient conditions. The calculated results of a cumulative fatigue usage factors(CUF) were compared for the case of the models with or without WRS effects. The results showed that the CUF at weldment and heat affected zone was affected by the WRS. Key Words : Fatigue analysis, Cumulative usage factor, Dissimilar metal weld, Finite element method, weld residual stress ISSN 1225-6153Online ISSN 2287-8955 1. 서 론 피로해석을 하는 이유는 기기가 파손을 일으키지 않고 정해진 횟수만큼의 반복운전을 할 수 있는지를 알아보기 위한 것이다. ASME Boiler & Pressure Vessel Code 1)

Performance study of the simplified theory of plastic zones and the Twice-Yield method for the fatigue check

As elastic–plastic fatigue analyses are still time consuming the simplified elastic–plastic analysis (e.g. ASME Section III, NB 3228.5, the French RCC-M code, paragraphs B 3234.3, B 3234.5 and B3234.6 and the German KTA rule 3201.2, paragraph 7.8.4) is often applied. Besides linearly elastic analyses and factorial plasticity correction (Ke factors) direct methods are an option. In fact, calculation effort and accuracy of results are growing in the following graded scheme: a) linearly elastic analysis along with Ke correction, b) direct methods for the determination of stabilized elastic–plastic strain ranges and c) incremental elastic–plastic methods for the determination of stabilized elastic–plastic strain ranges.The paper concentrates on option b) by substantiating the practical applicability of the simplified theory of plastic zones STPZ (based on Zarka's method) and – for comparison – the established Twice-Yield method. The Twice-Yield method is explicitly addressed in ASME Code, Section VIII, Div. 2. Application relevant aspects are particularly addressed. Furthermore, the applicability of the STPZ for arbitrary load time histories in connection with an appropriate cycle counting method is discussed.Note, that the STPZ is applicable both for the determination of (fatigue relevant) elastic–plastic strain ranges and (ratcheting relevant) locally accumulated strains. This paper concentrates on the performance of the method in terms of the determination of elastic–plastic strain ranges and fatigue usage factors. The additional performance in terms of locally accumulated strains and ratcheting will be discussed in a future publication.

원자력 증기용 안전밸브의 개방성능 평가를 위한 해석적 연구

The purpose of this paper is to investigate an analytical approach for opening performance evaluation of the nuclear pressure safety valve based on standard codes such as ASME or KEPIC. It is well-known that safety valve is considered as one of pressure relief valves for protecting a boiler or pressure vessel from exceeding the maximum allowable working pressure. When pressure in a container reaches its set pressure, the safety valve commences discharging the internal fluid by a sudden opening called as popping. Safety valve is usually evaluated by set pressure, full open, blow-down, leakage and flow capacity. The test procedure and technical requirement for performance evaluation is described in international code of ASME code such as BPVC. The opening characteristics of steam safety valve can be analyzed by computational fluid dynamics (CFD) and steam shaft dynamics. First, the flow analysis along opening process is simulated by running the CFD models of the ten types of opening steps from 0 to 100%. As a analysis result, the various CFD outputs of flow pattern, pressure, forces on the disc and mass flow at each simulation step is demonstrated. The lift force is calculated by using the forces applied on disc from static pressure and secondary flow. And, the effect of huddle chamber or control chamber is studied by dynamic analysis based on CFD simulation results such as lift force. As a result, dynamics analysis shows opening features according to the sizes of control chamber.

Physical properties of F82H for fusion blanket design

The material properties, focusing on the properties used for design analysis were re-assessed and newly investigated for various heats of reduced activation ferritic/martensitic steel, F82H. Moreover, irradiation effects on those properties were studied in this work. Most of physical properties of unirradiated F82H are insensitive to the heat-to-heat variation, and more preferable to Grade 91 specified in the ASME code. Therefore numerical analysis using data of Grade 91 can be a conservative evaluation for F82H. The change in these properties of F82H is less than 6%. Therefore the irradiation effects on thermal stress and variation of electromagnetic force in plasma disruption could be quite small in the irradiation conditions studied.

Lower Bound Methods in Elastic-Plastic Shakedown Analysis

Several methods were proposed in recent years that allow the efficient calculation of elastic and elastic-plastic shakedown limits. This paper establishes a uniform framework for such methods that are based on perfectly-plastic material behavior, and demonstrates the connection to Melan's theorem of elastic shakedown. The paper discusses implications for simplified methods of establishing shakedown, such as those used in the ASME Code. The framework allows a clearer assessment of the limitations of such simplified approaches. Application examples are given.

Determination of Internal Pressure in a Shell and Tube Type of Heat Exchanger by Using IBR and ASME Codes

The shell and tube type heat exchanger is a non-fired pressure system consisting of two different pressure chambers such as shell chamber and tube chamber. It is separated by the internal tube wall, two media flow past each other with such alignment that, if there is a heat difference, they will mutually exchange heat without mixing in the process..While there is an enormous variety of specific design features that can be used in shell and tube exchangers, the number of basic components is relatively small. This paper aims at performing a comprehensive comparison of the design outputs by IBR and ASME codes of critical components of a shell and tube type heat exchanger in an attempt to show the redundancy in designing the components using both these codes simultaneously.

Finite Element Analysis of Split Type Header in Air Cooled Heat Exchanger

Split type header is a part of tube bundle of an Air-cooled heat exchanger, used in refinery and oil & gas production. Typically, header can be considered as a pressure vessel subjected to uniform internal pressure. Hence the header and nozzle in various design and operating conditions needs to be checked and verified for soundness of participating components. Top plate due to uniform internal pressure in the nozzle & header plate can produce high-localized stress and deformation. If the components are not designed for these conditions, safety of the equipment is at stake. Hence check for the stress and displacement of the header during operating condition is carried out using finite element analysis software and observed that header nozzle is free from collapse and serviceability failure. I. Introduction (1) Heat exchangers are systems that transfer heat between fluid mediums. The fluids or gases in a heat exchanger can be mixed or the energy transference can go through a conductive wall that keeps them separate. Air-cooled heat exchangers typically have rectangular bundles containing several rows of tubes. The hot fluid enters at the top of the bundle through nozzle pipe. While air is blown by fans vertically upwards across the tube bank, i.e. counter current flow. Structural integrity of the header box can be checked by using (12) Appendix 13 of ASME Sec VIII Div I and other parts of the exchanger header box with the respective clauses of ASME code. For installing the nozzle on the header box opening is made on it, this opening of the vessel must be reinforced with an equal amount of metal which has been cut out for the opening. The reinforcement may be an integral part of the vessel and nozzle or may be an additional reinforcing pad. In addition the compensation must account for the bending strength as well as the membrane strength. The reinforcement calculations for nozzle openings according to UG-39 are in addition to bending & membrane strength. However, the calculation of Area required, according to UG-39, requires that the required thickness(tr) be substituted for in the formula, and it has not been calculated, since the method used in Appendix 13 is by assuming a and calculating stresses then comparing with allowable stresses and yield stress. This means that required thickness (tr) cannot be directly calculated. Accordingly, the area A cannot be directly calculated however in this case nozzle opening is larger than ½ of the header opening hence is beyond the scope of UG-39 of ASME Sec VIII Div-I and therefore the calculation has to done by finite element method.

Interactive buckling tests on steel cones subjected to axial compression and external pressure – a comparison of experimental data and design codes

The plastic buckling of thick steel conical shells subjected to combined action of axial compression and external pressure is of considerable interest in the offshore and the nuclear industries. However, design information on this subject area is limited. At present, only the ASME B&PV code case 2286-2 provides design information for cones subjected to axial and hoop compressions acting simultaneously. This design rule has not been validated both experimentally and numerically, especially in the elastic–plastic region. Past results on interactive buckling tests carried out at the University of Liverpool, Liverpool, UK on 13 computer numerically controlled (CNC)-machined cones having r2/r1 = 2.02, r2/t = 34.3, h/r2 = 1.01, and β = 26.56° were compared with predictions of design loads obtained from ASME code case 2286-2. This was done in order to check the applicability of this design rule and suggest a safe operating region. The paper consider the general procedure adopted by the ASME case code 2286-2 to predict the interactive buckling curve and identifies some discrepancies in the predictions. It further suggests region of safe operational design level. Combined stability plots for the master cone and equivalent cylinder have also been derived. Results of this study show that ‘the equivalent cylinders’ do not represent a safe design substitute for a relatively thick cone under combined loading.

Structural Stress Based Fatigue Analysis Method for Plane Steel Gate

Though fatigue failure is often happened in steel gate, fatigue design has not yet been included in design or evaluation standards in China. In this paper, analytical theory based structural stress method and structural stress based fatigue analysis method are combined and employed for fatigue life evaluation of plane steel gate. The variation of water head is taken into consideration for the conditions of gate running and resting. In order to give a valid evaluation, the weld located among all components is considered as rigid connection when calculating structural stress. Master S-N curve in ASME code is used for life evaluation. Example shows the method can be used for fatigue analysis of plane steel gate, and can be used to identify fatigue failure zone.

원전 증기발생기 와전류검사 장치의 전기적 특성 측정

원전 증기발생기는 원자로 냉각재 계통에서 발생한 열에너지를 터빈 계통의 주급수에 전달하여 터빈을 회전시키기 위한 증기를 생산하는 일종의 열교환기이다. 증기발생기 전열관의 손상은 증기발생기의 구조적 및 누설 건전성 유지 능력을 저해시키기 때문에 주기적으로 와전류검사를 수행하여 전열관의 건전성을 평가한다. 증기발생기 전열관의 건전성 평가는 보통 원자로 연료 재장전 기간 중에 수행된다. 현재 국내 증기발생기 전열관에 적용되는 와전류검사는 KEPIC 및 ASME 코드 요건에 따라 수행되며, 와전류검사 수행에 필요한 검사 시스템은 와전류검사 장치와 수집된 신호를 평가하기 위한 평가 프로그램으로 구성된다. 검사에 적용되는 와전류검사 시스템을 구성하는 핵심기기인 와전류검사 장치는 ASME 및 KEPIC 코드에서 총 고조파 왜곡율, 입출력 임피던스, 증폭기 직선성 및 안정성, 위상 직선성, 대역폭 및 복조필터 응답, 디지털 변환, 채널 간섭 등과 같은 전기적 특성을 측정하도록 규정하고 있다. 이에 따라 본 논문에서는 국내 최초로 개발한 원전 증기발생기 와전류검사 장치의 전기적 특성 측정을 위한 ASME 및 KEPIC 코드 요건을 설명하고, 이 요건에 따른 증기발생기 와전류검사 장치의 전기적 특성의 측정 결과를 제시하였다. A steam generator in nuclear power plant is a heatexchager which is used to convert water into steam from heat produced in a nuclear reactor core, and the steam produced in steam generator is delivered to the turbine to generate electricity. Because of damage to steam generator tubing may impair its ability to adequately perform required safety functions in terms of both structural integrity and leakage integrity, eddy current testing is periodically performed to evaluate the integrity of tubes in steam generator. This assessment is normally performed during a reactor refueling outage. Currently, the eddy current testing for steam generator of nuclear power plant in Korea is performed in accordance with KEPIC & ASME Code requirements, the eddy current testing system is consists of remote data acquisition unit and data analysis program to evaluate the acquired data. The KEPIC & ASME Code require that the electrical properties of remote data acquisition unit, such as total harmonic distortion, input & output impedance, amplifier linearity & stability, phase linearity, bandwidth & demodulation filter response, analog-to-digital conversion, and channel crosstalk shall be measured in accordance with the KEPIC & ASME Code requirements. In this paper, the measurement requirements of electrical properties for eddy current testing instrument described in KEPIC & ASME Code are presented, and the measurement results of newly developed eddy current testing instrument by KHNP(Korea Hydro & Nuclear Power Co., LTD) are presented.

Coupled Thermomechanical Analysis of Autofrettaged and Shrink-Fitted Compound Cylindrical Shells

In this study, different configurations of compound multilayer cylinders subjected to autofrettage and shrink-fit processes and under combined cyclic thermal and pressure loads have been investigated and their fatigue life has been evaluated and compared. Fully coupled thermo-elastic analysis is taken into consideration during the calculation of the temperature profile through the wall thickness. Finite element model for the compound two-layer cylinder has been constructed and then validated with previous work in the literature and experimental work. In the experimental work, the temperature has been measured at different locations through the thickness of a two-layer shrink-fitted cylinder (SFC), subjected to internal quasi-static and dynamic thermal loads. Besides, the hoop strain at the outer surface of the cylinder has been measured for the same thermal loads. Using the developed finite element model, the hoop stress distributions through the thickness of different configurations of the compound cylinder have been calculated under different loading conditions, including internal static pressure, internal cyclic thermal loads, and combination of these loads. The mechanical fatigue life has been calculated using ASME codes due to the internal cyclic pressure. Moreover, the stress intensity factor (SIF) has been calculated for these configurations under cyclic thermal loads or cyclic thermomechanical loads, considering thermal accumulation. The stress intensity factors for different configurations have been compared with the critical SIF which is the fracture toughness of the material. The number of stress cycles required until the SIF reaches the critical SIF has been considered as the fatigue life for each configuration. It has been found that for the cases of cyclic thermal loads and combined cyclic pressure and thermal loads, the shrink-fitting of two layers followed by the autofrettage of the assembly is the best configuration to enhance the fatigue life of the two-layer cylinder.

Rational Determination of Lower-Bound Fracture Toughness Curves Using Master Curve Approach

The Master Curve gives the relation between the median of fracture toughness of ferritic steels and the temperature in the ductile–brittle transition temperature region. The procedure used to determine the Master Curve is provided in the current American Society for Testing and Materials (ASTM) E1921 standard. By considering the substitution of the alternative lower-bound curves based on the Master Curve approach for the KIc curves based on reference data sets in the present codes such as ASME Code Cases N-629 and N-631, the statistical characteristic should be well incorporated in the determination of the lower-bound curves. Appendix X4 in the ASTM standard describes the procedure used to derive the lower-bound curves; however, it appears to be addressed without sufficient consideration of the statistical reliability. In this study, we propose a rational determination method of lower-bound fracture toughness curves using the Master Curve approach. The method considers the effect of sample size in the determination of the tolerance-bound curve. The adequacy of the proposed method was verified by comparing the tolerance-bound curve with the fracture toughness database for national reactor pressure vessel (RPV) steels including plate and forging obtained from 4 T to 0.4 T C(T) specimens and 0.4 T SE(B) specimens. The method allows the application of the Master Curve using fewer specimens, which can coexist with the present surveillance program.

Low Cycle Fatigue and Creep-Fatigue Behavior of Alloy 617 at High Temperature

Alloy 617 is the leading candidate material for an intermediate heat exchanger (IHX) application of the very high temperature nuclear reactor (VHTR), expected to have an outlet temperature as high as 950 °C. Acceptance of Alloy 617 in Section III of the ASME Code for nuclear construction requires a detailed understanding of the creep-fatigue behavior. Initial creep-fatigue work on Alloy 617 suggests a more dominant role of environment with increasing temperature and/or hold times evidenced through changes in creep-fatigue crack growth mechanisms and failure life. Continuous cycle fatigue and creep-fatigue testing of Alloy 617 was conducted at 950 °C and 0.3% and 0.6% total strain in air to simulate damage modes expected in a VHTR application. Continuous cycle fatigue specimens exhibited transgranular cracking. Intergranular cracking was observed in the creep-fatigue specimens and the addition of a hold time at peak tensile strain degraded the cycle life. This suggests that creep-fatigue interaction occurs and that the environment may be partially responsible for accelerating failure.