scholarly journals Reactor Core Conceptual Design for a Scalable Heating Experimental Reactor, LUTHER

2021 ◽  
Vol 2 (2) ◽  
pp. 207-214
Author(s):  
Thinh Truong ◽  
Heikki Suikkanen ◽  
Juhani Hyvärinen
Keyword(s):  
Low Temperature ◽  
Conceptual Design ◽  
Nuclear Power ◽  
Water Pressure ◽  
Reactor Core ◽  
Electricity Production ◽  
Design Parameters ◽  
Experimental Reactor ◽  
Fuel Assemblies ◽  
Enriched Uranium

In this paper, the conceptual design and a preliminary study of the LUT Heating Experimental Reactor (LUTHER) for 2 MWth power are presented. Additionally, commercially sized designs for 24 MWth and 120 MWth powers are briefly discussed. LUTHER is a scalable light-water pressure-channel reactor designed to operate at low temperature, low pressure, and low core power density. The LUTHER core utilizes low enriched uranium (LEU) to produce low-temperature output, targeting the district heating demand in Finland. Nuclear power needs to contribute to the decarbonizing of the heating and cooling sector, which is a much more significant greenhouse gas emitter than electricity production in the Nordic countries. The main principle in the development of LUTHER is to simplify the core design and safety systems, which, along with using commercially available reactor components, would lead to lower fabrication costs and enhanced safety. LUTHER also features a unique design with movable individual fuel assembly for reactivity control and burnup compensation. Two-dimensional (2D) and three-dimensional (3D) fuel assemblies and reactor cores are modeled with the Serpent Monte Carlo reactor physics code. Different reactor design parameters and safety configurations are explored and assessed. The preliminary results show an optimal basic core design, a good neutronic performance, and the feasibility of controlling reactivity by moving fuel assemblies.

Recommendations for Conducting Probabilistic Risk Assessment of Spent Fuel Storage Facilities

2017 ◽  
Author(s):  
Sai Zhang ◽  
Jun Zhao ◽  
Jiejuan Tong ◽  
Zhixin Xu
Keyword(s):  
Nuclear Power ◽  
Reactor Core ◽  
Spent Fuel ◽  
Complementary Method ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Storage Units ◽  
Storage Facilities ◽  
Probabilistic Risk

Currently, the probabilistic risk assessments (PRA) for the nuclear power plant (NPP) sites are primarily focused on the reactor counterpart. However, evoked by the 2011 Fukushima Daiichi accident, it has been widely recognized that a complete site risk profile should not be confined to the reactor units, but should cover all the radiological sources in a site, e.g. spent fuel storage facilities. During the operation of the reactor units, the used fuel assemblies will be unloaded from the reactor core to the storage facilities in a continuous or periodical manner. Accident scenarios involving such facilities can occur with non-negligible frequencies and significant consequences, posing threat to public safety. Hence, the risk contributions from such scenarios should be carefully estimated and integrated into the safety goal evaluations. The spent fuel storage facilities can be categorized as two types: pool storage units and dry cask storage facilities. In the former type, spent fuel assemblies are stored in large pools inside or outside the reactor building, with the residual heat removed by natural or forced water circulation. The latter type, where air or inert gas circulation plays an important role, appear mostly as a complementary method, along with the pool storage units, to expand the plant’s storage capacity. For instance, at the Daiichi plant, there are several fuel pool units holding some fresh fuel and some used fuel, the latter awaiting for its transfer to the dry cask storage facilities on site. Note that, as well as in a joint manner, both storage facilities can be designed to serve the NPPs independently. As a fully developed method to identify potential risk in a logical and quantitative way, the framework of PRA can be generally applied to the spent fuel storage facilities with some special considerations. This paper is aimed at giving recommendations for the spent fuel storage facility PRAs, including (1) clarifying the analysis scope of risk from spent fuel storage facilities; (2) illustrating four key issues that determines such risk; (3) presenting three essential considerations when conducting PRAs to evaluate such risk. Also, this paper integrates the insights obtained from two representative case studies involving two NPP sites with different types of both fuel elements and storage facilities.

The Load Rejection Transient Analysis of Chinshan NPP (BWR/4) Using TRACE/PARCS

2014 ◽  
Author(s):  
Jhih-Jhong Huang ◽  
Hsiung-Chih Chen ◽  
Jong-Rong Wang ◽  
Lih-Yih Liao ◽  
Chunkuan Shih ◽  
...  
Keyword(s):  
Control System ◽  
Nuclear Power ◽  
Thermal Power ◽  
Reactor Core ◽  
Second Step ◽  
System Response ◽  
Recirculation Flow ◽  
Fuel Assemblies ◽  
Flow Control System ◽  
Main Steam

Chinshan Nuclear Power Plant (NPP) is the first boiling water reactor (BWR) NPP in Taiwan. It has two units of BWR/4 reactor made by GE Company and each rated thermal power was 1775 MW without power uprate (now its rated thermal power is 1805 MW after power uprate). This research focuses on the development of the Chinshan NPP TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model. The model is done in two steps: The first step is the development of a TRACES/PARCS model of Chinshan NPP which includes the vessel, fuel assemblies, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters (such as power, feedwater flow rate, reactor dome temperature, etc.) were identified to refine the model further in the frame of a steady state analysis. The second step is development of TRACE/PARCS model for the load rejection transient. In order to check the system response of the Chinshan NPP TRACE/PARCS model, this study uses the load rejection transient results of startup tests to benchmark the analysis results of Chinshan NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the startup test data. It indicates that there is a respectable accuracy in the Chinshan NPP TRACE/PARCS model for the load rejection transient.

scholarly journals Application of Evolutionary Simulated Annealing Method to Design a Small 200 MWt Reactor Core

2021 ◽  
Vol 10 (4) ◽  
pp. 16-23
Author(s):  
Tran Viet Phu ◽  
Tran Hoai Nam ◽  
Hoang Van Khanh
Keyword(s):  
Simulated Annealing ◽  
Power Plant ◽  
Nuclear Power ◽  
Cycle Length ◽  
Reactor Core ◽  
Loading Pattern ◽  
The Core ◽  
Fuel Assemblies ◽  
Full Power ◽  
Annealing Method

This paper presents the application of an evolutionary simulated annealing (ESA) method to design a small 200 MWt reactor core. The core design is based on a reference ACPR50 reactor deployed in a floating nuclear power plant. The core consists of 37 typical 17x17 PWR fuel assemblies with three different U-235 enrichments of 4.45, 3.40 and 2.35 wt%. Core loading pattern (LP) has been optimized for obtaining the cycle length of 900 effective full power days, while minimizing the average U-235 enrichment and the radial power peaking factor. The optimization process was performed by coupling the ESA method with the COREBN module of the SRAC2006 system code.

scholarly journals Method for Contactless Determination of the Height of A Fuel Assembly by Means of A 3-D Reconstruction of A Collinear Stereopair Images

2018 ◽  
Vol 7 (3.13) ◽  
pp. 51
Author(s):  
S Kravtsov ◽  
K Rumyantsev
Keyword(s):  
Fuel Assembly ◽  
Nuclear Power ◽  
Vision System ◽  
Reactor Core ◽  
System A ◽  
Fuel Assemblies ◽  
Nuclear Power Unit ◽  
Head Height ◽  
Set Of Points

A method for determining the head height of fuel assemblies in the reactor core of a nuclear power unit using a 3-D reconstruction of a stereopair of collinear images is considered. The method is based on the principle of statistical evaluation of the height of a set of points for a 3-D reconstruction of the contour of the head of the fuel assembly. To obtain a stereopair of images, it is suggested to use a collinear digital stereo-vision system. A model experiment was carried out. The results are compared with the known method for determining the height of the heads of fuel assemblies, based on an estimate of the height of the centers of gravity of the contours of fuel assembly heads. The proposed method shows a higher accuracy in solving the problem of determining the heights of fuel assembly heads in comparison with the known method.  

scholarly journals Powering Down

2003 ◽  
Vol 125 (04) ◽  
pp. 46-48
Author(s):  
Harry Hutchinson
Keyword(s):  
Heat Transfer ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
The United States ◽  
Nuclear Industry ◽  
Significant Feature ◽  
Pressurized Water ◽  
Safety Testing ◽  
Fuel Assemblies ◽  
Enriched Uranium

This article reviews that after a half century of safety testing for the nuclear industry, a key heat-transfer lab is losing its home. Columbia University’s Heat Transfer Research Facility has been the only place to go for key safety testing. Since the days of the Atoms for Peace program during the Eisenhower years, the lab has tested generations of nuclear reactor fuel assemblies. The lab’s clients over the years have included all the designers of pressurized water reactors in the United States and others from much of the world. The tests are primarily concerned with one small, but significant feature of a reactor core. A core contains as many as 3000 fuel assemblies, bundles of long, slender rods containing enriched uranium. Controlled fission among the bundles heats water to begin the series of heat-transfer cycles that send steam to the turbines that will drive generators.

scholarly journals CFD Analysis of Downcomer of Nuclear Reactor VVER 440

2016 ◽  
Vol 66 (2) ◽  
pp. 55-62
Author(s):  
Vladimír Kutiš ◽  
Jakub Jakubec ◽  
Juraj Paulech ◽  
Gálik Gálik ◽  
Tibor Sedlár
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Coolant Flow ◽  
Orifice Diameter ◽  
Cfd Analysis ◽  
Flow Conditions ◽  
Fuel Assemblies ◽  
Flow Through ◽  
Cfd Analyses

Abstract The paper is focused on CFD analyses of the coolant flow in the nuclear reactor VVER 440. The goal of the analyses is to investigate the influence of the orifice diameter on the mass flow through individual fuel assemblies in the reactor core. The diameter of orifice can be changed during the operation of a nuclear power plant. Considered boundary conditions in the investigated region of the coolant are based on nominal coolant flow conditions in the nuclear reactor VVER 440.

scholarly journals Conceptual Design of a 10 MW Multipurpose Research Reactor Using VVR-KN Fuel

2020 ◽  
Vol 2020 ◽  
pp. 1-11
Author(s):  
Nhi-Dien Nguyen ◽  
Kien-Cuong Nguyen ◽  
Ton-Nghiem Huynh ◽  
Doan-Hai-Dang Vo ◽  
Hoai-Nam Tran
Keyword(s):  
Conceptual Design ◽  
Research Reactor ◽  
Nuclear Research ◽  
The Core ◽  
Horizontal Beam ◽  
Fuel Assemblies ◽  
Power Rate ◽  
Safety Parameters ◽  
Enriched Uranium ◽  
Maximum Temperatures

The paper presents a conceptual design of a 10 MW multipurpose nuclear research reactor (MPRR) loaded with the low-enriched uranium (LEU) VVR-KN fuel type. Neutronics and burnup calculations have been performed using the REBUS-MCNP6 linkage system code and the ENDF/B-VII.0 data library. The core consists of 36 fuel assemblies: 27 standard fuel assemblies and 9 control fuel assemblies with the uranium density of 2.8 gU/cm3 and the 235U enrichment of 19.75 wt.%. The cycle length of the core is 86 effective full-power days with the excess reactivity of 9600 and 1039 pcm at the beginning of cycle and the end of cycle, respectively. The highest power rate and the highest discharged burnup of fuel assembly are 393.49 kW and 56.74% loss of 235U, respectively. Thermal hydraulics analysis has also been conducted using the PLTEMP4.2 code for evaluating the safety parameters at a steady state of the hottest channel. The maximum temperatures of coolant and fuel cladding are 66.0°C and 83.0°C, respectively. This value is lower than the design limit of 98°C for cladding temperature. Thermal fluxes at the vertical irradiation channels and the horizontal beam ports have been evaluated. The maximum thermal fluxes of 2.5 × 1014 and 8.9 ×1013 n·cm−2·s−1 are found at the neutron trap and the beryllium reflector, respectively.

scholarly journals Some Characteristics of Hypothetic TVSA with Fuel Enriched beyond 5% for VVER1000

2016 ◽  
pp. 22-26
Author(s):  
Ye. Bilodid ◽  
Yu. Kovbasenko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Safety ◽  
Management Systems ◽  
Design Parameters ◽  
Spent Fuel ◽  
Fuel Management ◽  
Severe Accidents ◽  
Fuel Assemblies

The paper presents comparison of regular TVSA with average enrichment of 4,386% and hypothetical TVSA with enrichment of 10% based on design parameters and materials of TVSA fuel assemblies produced by TVEL (Russia), which today are widely used at nuclear power plants in Ukraine. It is shown that implementation of new fuel assemblies will result in improved use of fuel and increase of installed capability factor. At the same time, fresh and spent fuel management systems shall be modernized to meet relevant nuclear safety criteria. The paper analyzes possible criticality initiation at different stages of severe accidents related to core melt and using fuel with higher enrichment.

Method for Joint Configuration of Nuclear Power Plant Fuel Elements/Assemblies

2013 ◽  
Author(s):  
Zirong Ma ◽  
Lihong Nie ◽  
Yulong Mao
Keyword(s):  
Power Plant ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Reactor Core ◽  
Configuration Design ◽  
Joint Configuration ◽  
Fuel Assemblies ◽  
Reactor Cores ◽  
Fuel Elements ◽  
Operating Units

The method is about a design for nuclear fuel elements/assemblies configuration conducted jointly by several units with interchangeable fuel elements/assemblies. The first core of the new units may totally use the fuel elements/assemblies with relatively high enrichments which are identical or equivalent to the equilibrium cycle enrichments. Wherein, a part of new fuel elements/assemblies with relatively high enrichments may be put into one or more operating units for combustion, such that the new fuel elements/assemblies loaded in the operating units is much more than when such method is not implemented, and then considerable more once burned fuel elements/assemblies are obtained. Meanwhile, considerable twice burned fuel elements/assemblies may be provided by the operating units additionally (taking the reactor core with 157 fuel assemblies refueling every 18 months for example). From the considerable obtained once and twice burned fuel elements/assemblies, the burned fuel elements/assemblies with the desired quantity and burnup may be selected to construct the first reactor cores of the new units together with new fuel elements/assemblies with relatively high enrichments. After the method for joint configuration design of nuclear fuel elements/assemblies has been implemented, the average discharge burnup of the operating units added with more new fuel elements/assemblies and the first reactor cores of the new units is higher than that of the equilibrium cycle.

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