JSME Code on Protection Design Against Postulated Pipe Rupture for Nuclear Power Plants

2004 ◽  
Author(s):  
Keisuke Kitsukawa ◽  
Hiroshi Yokota ◽  
Koichi Murayama ◽  
Hiroshi Ueda ◽  
Yasukazu Takada ◽  
...  
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water System ◽  
Jet Impingement ◽  
Feed Water ◽  
Main Steam ◽  
Leak Before Break

As one of the Codes for Nuclear Power Generation Facilities, “Rules on Protection Design against Postulated Pipe Rupture for Nuclear Power Plants (JSME S ND1-2002)” has been developed by the JSME Committee on Power Generation Facility Codes from October 2001 and published in December 2002. The code covers the design for protection against postulated pipe rupture in nuclear power plants and gives the basic plan of protection design, locations of postulated pipe rupture, methods of determining the rupture type and opening area, and a procedure for evaluating jet impingement phenomena. It is a special feature of the code that the LBB (Leak Before Break) concept is applied to the determination of the piping rupture type, which belongs to RCPB, or to the main steam and feed water system inside the PWR containment vessel. Types of piping material applicable to the LBB concept are austenitic stainless steel, carbon steel and low-alloy steel. Furthermore, the code provides the flow and the method of LBB evaluation. In this paper, we describe the major rules of the code, including an outline of LBB evaluation methods.

The optimization methodology research on the motion performances of the driving device of main steam and feed water isolation valves

2017 ◽  
Vol 232 (6) ◽  
pp. 1069-1078
Author(s):  
Dalei Pan ◽  
Shenjie Gu ◽  
Guangyue Guo ◽  
Hongbo Kuang ◽  
Hua Zhong ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Maximum Pressure ◽  
Feed Water ◽  
Performance Indexes ◽  
Motion Performance ◽  
Optimization Methodology ◽  
Electromechanical Equipment ◽  
Main Steam

The driving device of the main steam and feed water isolation valves is one of the most important electromechanical equipment in nuclear power plants, and its motion performances are related to the safety of nuclear power plants. This article proposes an optimization methodology to research the motion performances. In the methodology, inputs are the motion performance indexes and the major influencing parameters with no need for complex mathematical models of the electromechanical equipment; the co-simulation model or the prototype are adopted to illustrate the influence rules of major parameters on the motion performance indexes; objective functions for optimization, which combine motion performances with weight factors, can reveal the influence rule curves. Then based on the methodology, motion performances of the equipment are analyzed and the key indexes are selected. Besides, the maximum pressure of the driving device is chosen as the major parameter and a dimensionless objective function synthesizing the motion performance indexes is proposed. Finally, the influence rule curve where the dimensionless function varies with the maximum pressure is obtained by the co-simulation and the experimental study on the prototype verifies the results, which provides references for the further research and development in the engineering application. For other electromechanical equipment, the method is an efficient tool to design, verify, and optimize the performances.

scholarly journals Nuclear Power: Time to Start Again

2003 ◽  
Author(s):  
William D. Rezak
Keyword(s):  
Power Generation ◽  
Electric Power ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Environmentally Friendly ◽  
Power Industry ◽  
Nuclear Industry ◽  
The Us ◽  
Generating Capacity

One of America’s best kept secrets is the success of its nuclear electric power industry. This paper presents data which support the construction and operating successes enjoyed by energy companies that operate nuclear power plants in the US. The result—the US nuclear industry is alive and well. Perhaps it’s time to start anew the building of nuclear power plants. Let’s take the wraps off the major successes achieved in the nuclear power industry. Over 20% of the electricity generated in the United States comes from nuclear power plants. An adequate, reliable supply of reasonably priced electric energy is not a consequence of an expanding economy and gross national product; it is an absolute necessity before such expansion can occur. It is hard to imagine any aspect of our business or personal lives not, in some way, dependent upon electricity. All over the world (in 34 countries) nuclear power is a low-cost, secure, safe, dependable, and environmentally friendly form of electric power generation. Nuclear plants in these countries are built in six to eight years using technology developed in the US, with good performance and safety records. This treatise addresses the success experienced by the US nuclear industry over the last 40 years, and makes the case that this reliable, cost-competitive source of electric power can help support the economic engine of the country and help prevent experiences like the recent crisis in California. Traditionally, the evaluation of electric power generation facility performance has focused on the ability of plants to produce at design capacity for high percentages of the time. Successful operation of nuclear facilities is determined by examining capacity or load factors. Load factor is the percentage of design generating capacity that a power plant actually produces over the course of a year’s operation. This paper makes the case that these operating performance indicators warrant renewed consideration of the nuclear option. Usage of electricity in the US now approaches total generating capacity. The Nuclear Regulatory Commission has pre-approved construction and operating licenses for several nuclear plant designs. State public service commissions are beginning to understand that dramatic reform is required. The economy is recovering and inflation is minimal. It’s time, once more, to turn to the safe, reliable, environmentally friendly nuclear power alternative.

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.

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

French Codes Used for EPR and Comparison With ASME Practices

2008 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Radiation Protection ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Industry ◽  
Quality Level ◽  
General Philosophy ◽  
Key Aspects ◽  
Leak Before Break ◽  
Minimum Quality

On December 2005, the French regulator issued a new regulation for French nuclear power plants, in particular for pressure equipment (PE). This regulation need first to agree with non-nuclear PE regulation and add to that some specific requirements, in particular radiation protection requirements. Different advantages are in these proposal, it’s more qualitative risk oriented and it’s an important link with non-nuclear industry. Only few components are nuclear specific. But, the general philosophy of the existing Codes (RCC-M, KTA or ASME) have to be improved. For foreign Codes, it’s plan to define the differences in the user specifications. In parallel to that, a new safety classification has been developed by French utility. The consequences is the need to cross all these specifications to define a minimum quality level for each components or systems. In the same time a new concept has been developed to replace the well known “Leak Before Break methodology” by the “Break Exclusion” methodology. This paper will summarize the key aspects of these different topics and regularly compare with ASME practices.

Sign in / Sign up

Export Citation Format

Share Document