scholarly journals Contributions of the ORNL Piping Program to Nuclear Piping Design Codes and Standards

1977 ◽  
Vol 99 (1) ◽  
pp. 224-230
Author(s):  
S. E. Moore
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Manufacturing Industry ◽  
Stress Index ◽  
Paper Briefly ◽  
Technical Literature ◽  
Index Method ◽  
Engineering Data ◽  
History Of ◽  
Future Work

Since 1967, ORNL Piping Program has been engaged in providing information for the development of stress indices to be used in the analysis of piping components for nuclear power plants. This effort has surveyed the piping manufacturing industry and analyzed that industry’s products; has combed the technical literature for pertinent engineering data; has performed theoretical and experimental analysis of nuclear piping components; and has defined, tested, and improved indices for the stress-index method of analysis for piping components. This paper briefly reviews the history of piping-analysis standards; outlines the philosophy of the stress-index method of analysis; and explains some of the specific contributions made by the ORNL program to the Codes and Standards. Current and future work is also noted.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

scholarly journals Plant Performance History of an Innovative Gate Valve in Critical Service Applications

2017 ◽  
Author(s):  
M. S. Kalsi ◽  
Patricio Alvarez ◽  
Thomas White ◽  
Micheal Green
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gate Valve ◽  
Plant Performance ◽  
Flow Loop ◽  
Performance History ◽  
Pressurized Water ◽  
Maintenance Costs ◽  
History Of ◽  
Water Reactors

A previous paper [1] describes the key features of an innovative gate valve design that was developed to overcome seat leakage problems, high maintenance costs as well as issues identified in the Nuclear Regulatory Commission (NRC) Generic Letters 89-10, 95-07 and 96-05 with conventional gate valves [2,3,4]. The earlier paper was published within a year after the new design valves were installed at the Pilgrim Nuclear Plant — the plant that took the initiative to form a teaming arrangement as described in [1] which facilitated this innovative development. The current paper documents the successful performance history of 22 years at the Pilgrim plant, as well as performance history at several other nuclear power plants where these valves have been installed for many years in containment isolation service that requires operation under pipe rupture conditions and require tight shut-off in both Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs). The performance history of the new valve has shown to provide significant performance advantage by eliminating the chronic leakage problems and high maintenance costs in these critical service applications. This paper includes a summary of the design, analysis and separate effects testing described in detail in the earlier paper. Flow loop testing was performed on these valves under normal plant operation, various thermal binding and pressure locking scenarios, and accident/pipe rupture conditions. The valve was designed, analyzed and tested to satisfy the requirements of ANSI B16.41 [9]; it also satisfies the requirements of ASME QME 1-2012 [10]. The results of the long-term performance history including any degradation observed and its root cause are summarized in the paper. Paper published with permission.

The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors

2014 ◽  
Author(s):  
Jozef Molnar ◽  
Radim Vocka
Keyword(s):  
Surveillance System ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Technical Specification ◽  
Training Simulator ◽  
History Of ◽  
Monitoring And Surveillance ◽  
Technical Specifications ◽  
European Nuclear

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system. After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016. This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

B2' Stress Indices for Elbows Using Non-Linear Finite Element Analysis

2004 ◽  
Author(s):  
K. Venkataramana ◽  
V. Bhasin ◽  
K. K. Vaze ◽  
H. S. Kushwaha
Keyword(s):  
Plastic Deformation ◽  
Finite Element ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Kinematic Hardening ◽  
Material Behavior ◽  
Stress Index ◽  
Stress Indices

Nuclear power plants are designed to withstand earthquake loads without severe damage under service level D conditions. Under earthquake induced reversing dynamic load, nuclear power plant components may undergo plastic deformation. Plastic deformation in class I nuclear power plant piping systems is limited by Equation (9) of ASME Boiler & Pressure Vessel Code [14], Section III, NB-3652. In the year 2000, the ASME B&PV Code was revised to accommodate reversing dynamic loading in which the failure mode is fatigue ratcheting, instead of plastic collapse. This modified equation [16] contains B2′ index, which is given as a fraction of B2 index where, B2 is defined for monotonic loading [17]. In this study a new definition is proposed for calculating B2′ stress index which is given by B2′ = MCLcyclicRange,straightpipe/MCLcyclicRange,component, where MClcyclicRange is the range of collapse moment. Incremental elastic-plastic nonlinear finite element analyses are performed considering both material and geometric nonlinearities. Kinematic hardening, isotropic hardening and elastic-perfectly plastic material models have been used to model the material behavior during plastic deformation. Load deflection curves are obtained and from these curves collapse loads for monotonic and cyclic loading are determined. B2 and B2′ stress indices are computed for elbows using the proposed equation. The computed stress indices are compared with ASME Code values.

Performance Study of an Air-Oil Pump and Separator Solution

2012 ◽  
Author(s):  
Johan Steimes ◽  
François Gruselle ◽  
Patrick Hendrick
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Gas Flow ◽  
Separation Efficiency ◽  
Original System ◽  
Lessons Learned ◽  
Oil Consumption ◽  
Performance Study ◽  
Oil Pump ◽  
Future Work

Many applications need to extract a certain phase from a multiphase flow like in oil extraction, flow in nuclear power plants, aircraft lubrication systems, etc. The Aero-Thermo-Mechanics (ATM) Department of Universiteé Libre de Bruxelles (ULB) is developing an original system to extract the gas from a liquid-gas flow together with increasing the pressure of the liquid phase. This system will help to reduce the complexity and the oil consumption of aeroengine lubrication systems. This paper will summarise the results of a first air/oil prototype. It will also present the guidelines learned from this prototype and used to design a second version of the integrated pump and separator. A newly developed oil consumption measurement system will also be presented. Based on previous results, on litterature review and on an in-house theoretical model, the paper will explain theoretically how the separation efficiency is affected by the particle distribution at the inlet of the prototype, and by the key parameters identified in different studies. Finally the conclusions will present the lessons learned through the design and tests of these two prototypes and the future work will be presented.

scholarly journals Efficiency and prospects for the implementation of measures for qualification of equipment important for safety in domestic nuclear energetics

2020 ◽  
Vol 21 (4) ◽  
pp. 369-377
Author(s):  
I.A. Khomych ◽  
◽  
T.V. Kovalinska ◽  
V.I. Sakhno ◽  
Yu.V. Ivanov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Expected Time ◽  
Operating Modes ◽  
Reliability Indicators ◽  
Before And After ◽  
Operational Information ◽  
History Of ◽  
On Line

The results of implementing equipment qualification are analyzed. Such equipment is critical for the nuclear and technical safety of domestic nuclear power plants that are especially important for the implementation of the Program for extending the terms of out-of-project operation of power reactors that are capable of being used as powerful sources of electricity. Based on the comparison of published reliability indicators of domestic nuclear power plants before and after implementing the qualification, it is shown that still there are problems to be solved. The perspective of further enhancing the reliability of the operation of domestic nuclear energetics is considered, by implementing radiation functional testing methods that are been developed at the INR NAS of Ukraine for a long period. The basis of this method is detailed research and operational control of all processes that occur in critical equipment in any operating modes of nuclear reactors to form a resource history of the equipment and to provide operational information about the remaining resource and the expected time of its failure to an on-line object operator.

New Approach for Fatigue Damage Monitoring Based on Actual Operating History of Nuclear Power Plants

2015 ◽  
Vol 190 (1) ◽  
pp. 88-96 ◽  
Author(s):  
Hee-Jin Shim ◽  
Chang-Kyun Oh ◽  
Hyun-Su Kim ◽  
Myung-Hwan Boo ◽  
Jong-Jooh Kwon
Keyword(s):  
Fatigue Damage ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Approach ◽  
Actual Operating ◽  
Damage Monitoring ◽  
History Of ◽  
Operating History

Root Cause Analysis of the Unexpected Behaviour of a Flaked Material Under Irradiation and Transferability to the Doel 3/Tihange 2 Reactor Pressure Vessels

2016 ◽  
Author(s):  
Evy De Bruycker ◽  
Séverine De Vroey ◽  
Xavier Hallet ◽  
Jacqueline Stubbe ◽  
Steve Nardone
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Root Cause Analysis ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Cause Analysis ◽  
Root Cause ◽  
History Of ◽  
Manufacturing History ◽  
Reactor Pressure

During the 2012 outage at Doel 3 (D3) and Tihange 2 (T2) Nuclear Power Plants (NPP), a large number of nearly-laminar indications were detected mainly in the lower and upper core shells. The D3/T2 shells are made from solid casts that were pierced and forged. Restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes. These tests showed unexpected results regarding the shift in the Reference Temperature for Nil Ductility Transition (RTNDT) of the flaked material VB395 (Steam Generator shell rejected because of flakes) after irradiation. This paper presents the root cause analysis of this unexpected behaviour and its transferability (or not) to the D3/T2 Reactor Pressure Vessels (RPVs). A mechanistic and a manufacturing based approach were used, aiming at identifying the microstructural mechanisms responsible for the atypical embrittlement of VB395 and evaluating the plausibility of these mechanisms in the D3/T2 RPVs. This work was based on expert’s opinions, literature data and test results. Both flaked and unflaked samples have been investigated in irradiated and non-irradiated condition. All hydrogen-related mechanisms were excluded as root cause of the unexpected behaviour of VB395. Two possible mechanisms at the basis of the atypical embrittlement of VB395 were identified, but are still open to discussion. These mechanisms could be linked to the specific manufacturing history of the rejected VB395 shell. Since the larger than predicted shift in transition temperature after irradiation of VB395 is not linked with the hydrogen flaking and since none of the specific manufacturing history features that are possible root causes are reported for the D3/T2 RPVs, the D3/T2 shells should not show the unexpected behaviour observed in VB395.

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