scholarly journals Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

2020 ◽  
Vol 31 (1) ◽  
pp. 16-20
Author(s):  
Duwi Hariyanto ◽  
Sidik Permana
Keyword(s):  
Uranium Dioxide ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Three Dimensional ◽  
Energy Agency ◽  
Pressurized Water ◽  
Nuclear Data Library ◽  
Long Life ◽  
Enriched Uranium

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

scholarly journals Neutronic Analysis of Small Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with Gd2O3 and Pa-231 as Burnable Poisons

2020 ◽  
Vol 31 (1) ◽  
pp. 10-15
Author(s):  
Duwi Hariyanto ◽  
Nining Yuningsih ◽  
Sidik Permana
Keyword(s):  
Uranium Dioxide ◽  
Power Plants ◽  
Volume Fraction ◽  
Reactor Core ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Nuclear Data Library ◽  
Long Life ◽  
Enriched Uranium

The requirement for electricity increases with the growth of the human population. The existing power plants have not been able to fulfill all electricity requirements, especially in remote areas. The small long-life pressurized water reactor (PWR) is one of the solutions and innovations in nuclear technology that can produce electrical energy for a long time without refueling. This study aimed to analyze the neutronic of small long-life PWR that using Thorium-Uranium dioxide ((Th-U)O2) fuels with enriched Uranium-235 (U-235) and the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. The SRAC Code with the JENDL-4.0 nuclear data library had been used for the calculation method. In this study, the geometry of the two-dimensional (R-Z) reactor core with different fuel volume fraction was analyzed. Moreover, variations of the Uranium-235, Gadolinium, and Protactinium-231 fractions in the fuels were carried out. The result in this study was a PWR 420 MWt design using 60% Uranium dioxide fuel with enriched Uranium-235 of 10%-11%-12% and the addition of 0,0125% Gadolinium and 1,0% Protactinium-231 as the burnable poisons that could operate for thirteen years without refueling. The small long-life PWR design could produce a power density of 85,1 watts/cc with the reactivity for less than 4,6% dk/k.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

Coordinated Research Projects at IAEA Concerning RPV and Structural Integrity

2006 ◽  
Author(s):  
William Server ◽  
Timothy Hardin ◽  
Milan Brumovsky´
Keyword(s):  
Ferritic Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Current Status ◽  
Research Projects ◽  
Energy Agency ◽  
Pressurized Water ◽  
Toughness Testing ◽  
Water Reactors

The International Atomic Energy Agency (IAEA) has had a series of reactor pressure vessel (RPV) structural integrity programs that started back in the 1970s. These Coordinated Research Projects most recently have focused on use of the Master Curve fracture toughness testing approach for RPV and other ferritic steel components and on the issue of pressurized thermal shock (PTS) in operating pressurized water reactors. This paper will provide the current status for these projects and discuss the implications for improved safety of key ferritic steel components in nuclear power plants (NPPs).

Safety Design Philosophy of Mitsubishi PWRs

1993 ◽  
Vol 207 (4) ◽  
pp. 253-258
Author(s):  
T Hakata ◽  
T Kitamura
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Reliability ◽  
Operating Experience ◽  
Energy Agency ◽  
Pressurized Water ◽  
Design Philosophy ◽  
Safety Design ◽  
Water Reactors ◽  
The Difference

The basic safety design philosophy of Mitsubishi pressurized water reactors (PWRs) is discussed and compared with the British PWR. PWR plants are designed in accordance with the Japanese regulatory guidelines which are similar to American and International Atomic Energy Agency (IAEA) safety criteria and are based on defence-in-depth principles. The high reliability of nuclear power plants is especially emphasized in Mitsubishi PWRs, and this has been demonstrated by the good operating experience of PWR plants in Japan. The safety system designs of six key items, which were discussed in the recent review of overseas designs by British utilities, are addressed to show the difference in the design philosophy between the United Kingdom and Japan.

Optimization of Small Long Life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input

2012 ◽  
Vol 260-261 ◽  
pp. 307-311 ◽  
Author(s):  
Menik Ariani ◽  
Z. Su'ud ◽  
Fiber Monado ◽  
A. Waris ◽  
Khairurrijal ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Volume Fraction ◽  
Natural Circulation ◽  
Average Power ◽  
Natural Uranium ◽  
Energy Utilization ◽  
Fast Reactors ◽  
Long Life

In this study gas cooled reactor system are combined with modified CANDLE burn-up scheme to create small long life fast reactors with natural circulation as fuel cycle input. Such system can utilize natural Uranium resources efficiently without the necessity of enrichment plant or reprocessing plant. Therefore using this type of nuclear power plants optimum nuclear energy utilization including in developing countries can be easily conducted without the problem of nuclear proliferation. In this paper, optimization of Small and Medium Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. The optimization processes include adjustment of fuel region movement scheme, volume fraction adjustment, core dimension, etc. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 75 W/cc. With such condition we investigated small and medium sized cores from 300 MWt to 600 MWt with all being operated for 10 years without refueling and fuel shuffling and just need natural Uranium as fuel cycle input. The average discharge burn-up is about in the range of 23-30% HM.

A Review on Specific Features of Small and Medium Sized Nuclear Power Plants

2010 ◽  
Author(s):  
Salah Ud-din Khan ◽  
Minjun Peng ◽  
Muhammad Zubair ◽  
Shaowu Wang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Nuclear Reactors ◽  
Fixed Bed ◽  
Nuclear Industry ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Energy Supply System

Due to global warming and high oil prices nuclear power is the most feasible solution for generating electricity. For the fledging nuclear power industry small and medium sized nuclear reactors (SMR’s) are instrumental for the development and demonstration of nuclear reactor technology. Due to the enhanced and outstanding safety features, these reactors have been considered globally. In this paper, first we have summarized the reactor design by considering some of the large nuclear reactor including advanced and theoretical nuclear reactor. Secondly, comparison between large nuclear reactors and SMR’s have been discussed under the criteria led by International Atomic Energy Agency (IAEA). Thirdly, a brief review about the design and safety aspects of some of SMR’s have been carried out. We have considered the specifications and parametric analysis of the reactors like: ABV which is the floating type integral Pressurized water reactor; Long life, Safe, Simple Small Portable Proliferation Resistance Reactor (LSPR) concept; Multi-Application Small Light Water Reactor (MASLWR) concept; Fixed Bed Nuclear Reactor (FBNR); Marine Reactor (MR-X) & Deep Sea Reactor (DR-X); Space Reactor (SP-100); Passive Safe Small Reactor for Distributed energy supply system (PSRD); System integrated Modular Advanced Reactor (SMART); Super, Safe, Small and Simple Reactor (4S); International Reactor Innovative and Secure (IRIS); Nu-Scale Reactor; Next generation nuclear power plant (NGNP); Small, Secure Transportable Autonomous Reactor (SSTAR); Power Reactor Inherently Safe Module (PRISM) and Hyperion Reactor concept. Finally we have point out some challenges that must be resolved in order to play an effective role in Nuclear industry.

The Investigation of Classification Management for SSCS Susceptible to Ageing in PWR

2010 ◽  
Author(s):  
Fei Xue ◽  
Peng Liu ◽  
Xin-ming Meng ◽  
Wen-xin Ti ◽  
Lei Lin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Evaluation Method ◽  
Residual Life ◽  
Classification Method ◽  
Energy Agency ◽  
Pressurized Water ◽  
Management Method ◽  
Classification Management

A set of screening and classification method is proposed in this paper to manage the Systems, Structures and Components (SSCs) of the Pressurized Water Reactor (PWR) for the aging management. The method proposed and carried out in China Nuclear Power Plants (NPP) is based on the systematic aging evaluation method and the management method for the license renewal from International Atomic Energy Agency (IAEA) and U.S. Nuclear Regulatory Commission (NRC), whose successful management methods are discussed in this paper. Considering the importance of the classification management of the SSCs, the classification method is investigated from the aspects of the safety, reliability and replacement of the SSCs with the Nuclear Power Plant operation experience feed-back. With the classification management method, the SSCs are managed economically and effectively for the Nuclear Power Plants in China. The method also makes great foundation for the aging management and residual life evaluation of the PWR in China.

The Activation Sources of Reactor Internals During Decommissioning

2017 ◽  
Author(s):  
Jianchun Han ◽  
Yan Zhou ◽  
Hui Li ◽  
Qiliang Mei
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Neutron Transport ◽  
Three Dimensional ◽  
High Energy ◽  
Pressurized Water ◽  
Dimensional Distribution

As China’s first nuclear power plant connected to the grid, the first Qinshan nuclear power plant is approaching the decommissioning period. Other nuclear power plants also turn into the preparation phase of decommissioning in succession. In order to facilitate decommissioning, source survey is conducted during the pre-decommissioning phase, which can provide radioactive inventory, contamination distribution, species and quantities of nuclides. The internals of the reactor work under the most severe radiation environment. During the reactor operation, the materials of internals are irradiated by high-energy neutrons. So activated nuclides are generated due to the neutron capture reaction, which are the main radioactive waste to be treated during decommissioning. In this paper, the neutron irradiation and the generated activation source of the internals for pressurized water reactors (PWR) are studied and analyzed. Firstly, core modeling was carried out, and the neutron transport calculation is performed to obtain three-dimensional distribution of the neutron flux. Secondly, according to the three-dimensional distribution of the material composition and the neutron flux rate of the reactor, the activation calculation is carried out to obtain the activation source.

Two- and Three-Dimensional Weld Residual Stress Analysis of a Weld Overlay Mockup of a VVER Feedwater Nozzle Dissimilar Metal Weld

2018 ◽  
Author(s):  
Francis H. Ku ◽  
Steven L. McCracken
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Dissimilar Metal ◽  
Before And After ◽  
Dissimilar Metal Weld ◽  
Three Dimensional Finite Element ◽  
Metal Weld

Weld overlay (WOL) is a popular repair technique to mitigate stress corrosion cracking (SCC) in dissimilar metal weld (DMW) in U.S. pressurized water reactor (PWR) design. The WOL technique is being considered as a SCC mitigation technique for DMW in Russian water-water energetic reactor (WWER or VVER) design. A WOL mockup on a VVER super emergency feedwater nozzle DMW has been fabricated, which represents the first WOL on VVER with the goal to mitigate SCC and the first WOL in Czech Republic civilian nuclear power plants. This paper presents the two- and three-dimensional finite element analyses performed to assess the weld residual stresses in the WOL mockup. The analysis evaluates the stress distribution and changes in the DMW before and after the WOL application, as well as compares the results to established industry guidelines and comparable WOLs on U.S. PWR.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

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