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.

2004 Edition of Japanese Fitness-for-Service Code for Nuclear Power Plants

2006 ◽  
Author(s):  
Koichi Kashima ◽  
Tomonori Nomura ◽  
Koji Koyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressure Vessels ◽  
Industrial Safety ◽  
National Rule ◽  
Class 1 ◽  
Asme Code ◽  
Service Inspection ◽  
Core Shroud

JSME (Japan Society of Mechanical Engineers) published the first edition of a FFS (Fitness-for-Service) Code for nuclear power plants in May 2000, which provided rules on flaw evaluation for Class 1 pressure vessels and piping, referring to the ASME Code Section XI. The second edition of the FFS Code was published in October 2002, to include rules on in-service inspection. Individual inspection rules were prescribed for specific structures, such as the Core Shroud and Shroud Support for BWR plants, in consideration of aging degradation by Stress Corrosion Cracking (SCC). Furthermore, JSME established the third edition of the FFS Code in December 2004, which was published in April 2005, and it included requirements on repair and replacement methods and extended the scope of specific inspection rules for structures other than the BWR Core Shroud and Shroud Support. Along with the efforts of the JSME on the development of the FFS Code, Nuclear and Industrial Safety Agency, the Japanese regulatory agency approved and endorsed the 2000 and 2002 editions of the FFS Code as the national rule, which has been in effect since October 2003. The endorsement for the 2004 edition of the FFS Code is now in the review process.

Calculation of Bolted Flange Connections of Floating and Metal-to-Metal Contact Type

2003 ◽  
Author(s):  
M. Schaaf ◽  
J. Bartonicek
Keyword(s):  
Pressure Vessels ◽  
Metal Contact ◽  
Second Step ◽  
European Standard ◽  
Calculation Algorithm ◽  
Contact Type ◽  
Asme Code ◽  
Floating Type ◽  
Bolted Flange

In Europe, in 2001 the new standard EN 1591 for strength and tightness proofs of bolted flange connections (BFC) of floating type flanges was released. In addition, the German nuclear code was revised regarding the calculation of BFC. With this standard not only the floating type but also the metal-to-metal contact type of flanges (MMC) can be treated. Additionally, the ASME code is the basis for the flange calculation in the European standard EN 13445, which is the standard for unfired pressure vessels. In compliance with the goal of the calculation, the different calculation codes can be used. There must be a differentiation between the design of the components, the determination of the prestress values for assembly, the stress analysis and the tightness proof of the BFC. First, all parameters which influence the function of the bolted flange connection are considered. In a second step, the range of use of the different standards and the calculation algorithm are discussed.

Strength Assessment of Equipments and Piping of WWER Nuclear Power Plants

2009 ◽  
Author(s):  
Stanislav Vejvoda
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
General Structure ◽  
Residual Lifetime ◽  
Technical Documentation ◽  
Strength Assessment ◽  
Operational Conditions ◽  
Elastic Plastic

The Standard Technical Documentation (STD) of The Association of Mechanical Engineers (A.M.E.), Section III “Strength Assessment of Equipment and Piping of Nuclear Power Plant of VVER type” provides guides and requirements for elaboration of: - conclusive documentation for equipment, piping and their supports, manufactured and supplied or manufactured for replacement, for nuclear power plants of VVER type, 440 MW and 1000 MW; - conclusive documentation for repaired, replaced and innovated components of equipment, piping and their supports for nuclear power plants of VVER type, 440 MW and 1000 MW; - conclusive documentation on continuous lifetime usage of equipment, piping and their supports during the operation of nuclear power plants of VVER type, 440 MW and 1000 MW and determination of their residual lifetime; - documentation on degree of compatibility of equipment, piping and their supports for nuclear power plants of VVER type, 440 MW and 1000 MW with the ASME Code Section III, Division 1. The following principles of the Section III are discussed in the presentation: general structure of the Section III; design specifications and operational conditions; calculation of allowed stresses; stress categories groups (σ)1, (σ)2, (σ)R, and (σaF); limit criteria for stress categories groups; loading blocs; assessment on fatigue; assessment of two-frequency cyclic loading on fatigue; fatigue correction factors for multiaxial stress states and gradient of stresses, corrosion, welding and radiation; limits; determination of load at collapse; use of elastic and elastic-plastic analyses; elastic-plastic carrying capacity of components; stiffness of the nozzle opening.

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.

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.

Application and Experience Feedback of ASME Code Case N-155-2 for RTR Piping in Nuclear Plants

2013 ◽  
Author(s):  
Nicolas d’Udekem ◽  
Philippe Art ◽  
Jacques Grisel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sea Water ◽  
Water System ◽  
Moderate Pressure ◽  
Raw Water ◽  
Cooling Systems ◽  
Piping Systems ◽  
Asme Code

Nowadays, the usefulness of RTR (Reinforced Thermosetting Resin) for pressure retaining equipment does not need further proof: they are lightweight, strong, with low thermal elongation and highly corrosion resistant. The use of RTR piping makes all sense for piping systems circulating raw water such as sea water at moderate pressure and temperature for plants cooling. However, this material is rarely used for safety related cooling systems in nuclear power plants. In Belgium, Electrabel and Tractebel have chosen to replace the existing carbon steel pipes of the raw water system by GRE (Glassfiber Reinforced Epoxy) pipes, in accordance with the Authorized Inspection Agency, applying the ASME Code Case (CC) N-155-2 defining the specifications and requirements for the use of RTR pipes, fittings and flanges. After a challenging qualification process, Class 3 GRE pipes are now installed and operating for raw water cooling systems in two Belgian nuclear units and will soon be installed in a third one. The paper will address the followed qualification processes and the implementation steps applied by Electrabel/Tractebel and relate the overcome obstacles encountered during manufacturing, erection and commissioning of Class 3 GRE piping in order to ensure quality, reliability and traceability required for safety equipment in nuclear power plants.

Radiochemical methodologies applied to determination of zirconium isotopes in low-level waste samples from nuclear power plants

2014 ◽  
Vol 302 (1) ◽  
pp. 41-47 ◽  
Author(s):  
T. C. Oliveira ◽  
R. P. G. Monteiro ◽  
G. F. Kastner ◽  
F. Bessueille-Barbier ◽  
A. H. Oliveira
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Low Level

scholarly journals Polymeric seal degradation in nuclear power plants: Determination of physical and mechanical properties

2017 ◽  
Vol 135 (7) ◽  
pp. 45814 ◽  
Author(s):  
Christopher P. Porter ◽  
James P. Bezzina ◽  
Francis Clegg ◽  
Mark D. Ogden
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Physical And Mechanical Properties
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