Evaluation of Flange Calculations Using Strain-Based Acceptance Criteria

2019 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elastic Stress ◽  
Calculation Methods ◽  
Acceptance Criteria ◽  
Class 1 ◽  
Torque Moment ◽  
Calculation Algorithms ◽  
Calculation Code

Abstract The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint different y and m values for different kinds of gasket are invented in ASME BPVC Section III [1]. The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled. For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other. In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.

Flaw Evaluation and Allowable Flaw Sizes of Nuclear Piping in JSME Code

2003 ◽  
Author(s):  
Kunio Hasegawa ◽  
Hideo Kobayashi ◽  
Koichi Kashima
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Japan Society ◽  
Acceptance Criteria ◽  
Evaluation Code ◽  
Evaluation Procedures ◽  
Class 1 ◽  
American Society ◽  
Service Inspection

A flaw evaluation code for nuclear power plants has been developed at the Japan Society of Mechanical Engineers (JSME) in 2000 and revised adding inspection rules in 2002. Then the code consists of inspection for nuclear components and evaluation procedures of flaws in Class 1 components detected during in-service inspection. This paper introduces the summary of the JSME Code and describes two kinds of allowable flaw sizes, Acceptance Standards and Acceptance Criteria, for Class 1 pipes in the flaw evaluation procedures. In addition, these allowable flaws are compared with those in the ASME (American Society of Mechanical Engineers) Code Section XI.

Comparison Between Different Calculation Methods for Determining Bolting Up Moments

2016 ◽  
Author(s):  
Alexander Mutz ◽  
Manfred Schaaf
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
European Standard ◽  
Calculation Algorithm ◽  
Operational Conditions ◽  
Class 1 ◽  
Asme Code ◽  
Floating Type

There are several different standards for flange calculation used in the European and so in the Suisse context. The European Standard EN 1591-1 that is used for the calculation of bolted flanged joints and EN 13555 in which the determination of the required gasket characteristics are defined were reissued in 2013 and in 2014, respectively. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components in Suisse nuclear power plants it is also used for class 1 flange connections. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint, different m and y values for different kind of gaskets are invented in ASME BPVC Section III. As cited in the Note of table XI-3221.1-1 the values m and y are not mandatory. In Switzerland, mainly the ASME BPVC should be used for the calculation of flange connections. The aim of the ASME Code is more or less not the tightness of the flanges but the integrity. Therefore, stresses are derived for dimensioning the flanges. Following loads are not considered neither for calculation of stresses nor for calculation of tightness. Considering the experience with flanges in general it could be asked, if it is more useful to look at the tightness than at the stresses. The codes KTA 3201.2 and KTA 3211.2 regulate the calculation of flange connections in German nuclear power plants. Stresses in floating type and in metal-to-metal contact type of flange connections and the tightness are calculated for the different load cases. In this paper, the differences in the calculations are shown between KTA 3211.2, ASME BPVC, Section III, Appendix 11, EN 1591-1 and Finite element calculations. In all load cases leakage shouldn’t occur. Therefore, internal pressure and temperature in test and operational conditions after bolting-up are also considered for the stress calculation if it is possible in the calculation algorithm.

scholarly journals Ion chromatographic analysis of aqueous iodine species by conductivity and radioactivity measurements

2005 ◽  
Vol 93 (9-10) ◽  
Author(s):  
Dorothea Schumann ◽  
R. Grasser ◽  
R. Dressler ◽  
H. Bruchertseifer
Keyword(s):  
Chromatographic Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Radioactive Iodine ◽  
Cooling Water ◽  
Molecular Iodine ◽  
New Device ◽  
Iodine Species ◽  
Radioactivity Measurements

SummaryA new device was developed for the identification of several iodine species in aqueous solution using ion chromatography. Iodide, iodate and molecular iodine can be determined. (The equipment allows both conductivity and radioactivity detections.) The method is applicable for the determination of radioactive iodine contaminations in the cooling water of nuclear power plants.

scholarly journals Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽  
2019 ◽  
Vol 6 (11) ◽  
pp. 120
Author(s):  
Luca Albertone ◽  
Massimo Altavilla ◽  
Manuela Marga ◽  
Laura Porzio ◽  
Giuseppe Tozzi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Scaling Factors ◽  
Activation Products ◽  
Long Time ◽  
Nuclear Installation ◽  
Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

Thermoelastoplastic Stress Analysis of a Thick-Walled Tube of Nonlinear Strain Hardening

1996 ◽  
Vol 118 (3) ◽  
pp. 340-346 ◽  
Author(s):  
S. Jahanian
Keyword(s):  
Strain Hardening ◽  
Nuclear Power ◽  
Power Plants ◽  
Rapid Heating ◽  
Numerical Procedure ◽  
Temperature Dependent ◽  
Nonlinear Strain ◽  
Elastic Approximation

In pressure vessel technology or nuclear power plants, some of the mechanical components are often subjected to rapid heating. If the temperature gradient during such process is high enough, thermoelastoplastic stresses may be developed in the components. These plastic deformations are permanent and may result in the incremental deformation of the structure in the long term. Accordingly, determination of thermoelastoplastic stresses during this process is an important factor in design. In this paper, a thick-walled cylinder of nonlinear strain hardening is considered for the thermoelastoplastic analysis. The properties of the material are assumed to be temperature dependent. The cylinder is subject to rapid heating of the inside surface while the outside surface is kept at the room temperature. A quasi-static and uncoupled thermoelastoplastic analysis based on incremental theory of plasticity is developed and a numerical procedure for successive elastic approximation is presented. The thermoelastoplastic stresses developed during this process are also presented. The effect of strain hardening and temperature dependency of material on the results are investigated.

scholarly journals Determination of Transient Fluid Temperature and Thermal Stresses in Pressure Thick-Walled Elements Using a New Design Thermometer

Energies ◽  
2019 ◽  
Vol 12 (2) ◽  
pp. 222 ◽  
Author(s):  
Magdalena Jaremkiewicz ◽  
Dawid Taler ◽  
Piotr Dzierwa ◽  
Jan Taler
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Pressure ◽  
Thermal Stresses ◽  
Thermal Inertia ◽  
Internal Surface ◽  
Transient Temperature ◽  
Fluid Temperature ◽  
Start Up

In both conventional and nuclear power plants, the high thermal load of thick-walled elements occurs during start-up and shutdown. Therefore, thermal stresses should be determined on-line during plant start-up to avoid shortening the lifetime of critical pressure elements. It is necessary to know the fluid temperature and heat transfer coefficient on the internal surface of the elements, which vary over time to determine transient temperature distribution and thermal stresses in boilers critical pressure elements. For this reason, accurate measurement of transient fluid temperature is very significant, and the correct determination of transient thermal stresses depends to a large extent on it. However, thermometers used in power plants are not able to measure the transient fluid temperature with adequate accuracy due to their massive housing and high thermal inertia. The article aims to present a new technique of measuring transient superheated steam temperature and the results of its application on a real object.

Comparison of German KTA and ASME Nuclear Design Codes for Class 1, 2, 3 Components and Piping

2011 ◽  
Author(s):  
Daniel Hofer ◽  
Henry Schau ◽  
Hu¨seyin Ertugrul Karabaki ◽  
Ralph Hill
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Codes ◽  
Nuclear Facility ◽  
Design Standards ◽  
Design Rules ◽  
Design Specifications ◽  
Division 1 ◽  
Class 1 ◽  
Standard Organization

This paper compares the design rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Rules for Construction of Nuclear Facility Components, with German nuclear design standards for Class 1, 2, 3 components and piping. The paper is focused on a comparison of the equations for Design by Analysis and on Piping equations. The ASME Section III Code has been used in combination with design specifications for design of German nuclear power plants. Together with manufacturers, inspectors and power plant owners, the German regulatory authority decided to develop their own nuclear design standards. The current versions being used are from 1992 and 1996. New versions of KTA design standards for pressure retaining components (KTA 3201.2 and KTA 3211.2) are currently under development. This comparison will cover the major differences between the design rules for ASME Section III, Div. 1 and KTA standards 3201.2 and 3211.2 as well as code or standard organization by sections, paragraphs, articles and code development.

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