scholarly journals Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

2021 ◽  
Vol 11 (1) ◽  
pp. 9-15
Author(s):  
Van Khanh Hoang ◽  
Vinh Thanh Tran ◽  
Dinh Hung Cao ◽  
Viet Ha Pham Nhu
Keyword(s):  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Model ◽  
Reactivity Coefficients ◽  
Fuel Pin ◽  
Safety Parameters ◽  
Long Life

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

Construction of an Elementary Model for the Dynamic Analysis of a Pressurized Water Reactor

2013 ◽  
Vol 284-287 ◽  
pp. 652-656 ◽  
Author(s):  
Chiung Wen Tsai ◽  
Chun Kuan Shih ◽  
Jong Rong Wang
Keyword(s):  
Transfer Functions ◽  
Feedback Loops ◽  
Dynamic Responses ◽  
Coolant Temperature ◽  
Inlet Temperature ◽  
Pressurized Water Reactor ◽  
Elementary Model ◽  
Pressurized Water ◽  
Water Reactor ◽  
Neutron Kinetics

A lumped-parameter numerical model was constructed based on the conservation laws of mass and energy and the point neutron kinetics with 6 groups of delayed neutron to represent the dynamics of primary loop of a pressurized water reactor (PWR) core. On the viewpoint of control theory, the coupled phenomenon of neutron kinetics and thermohydraulics can be recognized as a dynamic system with feedback loops which is caused by the Doppler effect and the coolant temperature difference. Scilab was implemented to representing the equivalent transfer functions and associated feedback loops of a PWR core. The dynamic responses were performed by the perturbations of coolant inlet flow, coolant inlet temperature, and reactivity insertion.

Prediction of safety parameters of pressurized water reactor based on feature fusion neural network

2022 ◽  
Vol 166 ◽  
pp. 108803
Author(s):  
Yinghao Chen ◽  
Dongdong Wang ◽  
Cao Kai ◽  
Cuijie Pan ◽  
Yayun Yu ◽  
...  
Keyword(s):  
Neural Network ◽  
Feature Fusion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Parameters

Advanced Alternatives of VVER-440 Fuel Assembly

2014 ◽  
Author(s):  
Juraj Tomaškovič ◽  
Petr Dařílek ◽  
Radoslav Zajac ◽  
Vladimír Nečas
Keyword(s):  
Fuel Assembly ◽  
Economic Efficiency ◽  
Fuel Cycle ◽  
Gradual Increase ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Burn ◽  
Fuel Pin ◽  
Basic Characteristics

The main goals of fuel development for pressurized water reactor are effectiveness and economic efficiency. Both requirements can be achieved by gradual increase of discharged fuel burn-up and prolongation of fuel cycle. The mentioned effects can be reached by optimisation of fuel assembly profiling, fuel enrichment raise, and by parasitic absorption reduction. These methods were used in VVER-440 fuel assembly optimisation, described in this paper. Fuel pin configurations with enrichment limit 5 % and also enlarged one up to 5.95 % U235 were designed. Reduction of parasitic absorption was limited by carcass frame of the assembly. Basic characteristics of the best assembly proposals are presented and effects on equilibrium fuel cycle of VVER-440 reactor are characterized.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

scholarly journals Perhitungan Perubahan Reaktivitas Pada Reaktor Serba Guna G.A. Siwabessy Akibat Pengoperasian Power Ramp Test Facility

2019 ◽  
Vol 4 ◽  
pp. 121
Author(s):  
A Suparmi ◽  
Tuti Dwi Setyaningsih ◽  
Suharyana Suharyana ◽  
Fuad Anwar ◽  
Riyatun Riyatun
Keyword(s):  
Reactor Core ◽  
Test Facility ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Ramp Test ◽  
Fuel Pin ◽  
Safe Limit ◽  
Reactor Operating ◽  
Test Fuel

<p><strong>Abstract: </strong>Power Ramp Test Facility (PRTF) is one of the irradiation facility contained in the Multipurpose Reactor GA Siwabessy. This facility is used to test the reactor fuel element pin-type Pressurized Water Reactor. As a result of the entry of foreign bodies cause changes reactor conditions, one of which is expressed with the amount of reactivity to assess the safety of the reactor due to the operation PRTF. PRTF operation simulation and calculation is done using software neutronics MCNP6. Test UO2 fuel enriched assumed at 5% with constant power reactor operating at 15 MW and test fuel pin placed on PRTF within 0, 20, 40, 60, 80, 100, 120, and 140 mm from the centre of the reactor core. Change of reactivity values required in order to secure the reactor, maximal value is 0,5%<em></em>.  The calculation were obtained at each position is (<em></em><em></em>;  <em></em>;  <em></em>; <em></em>;<em></em>; <em></em>; <em></em>; <em></em>). Change of reactivity values smaller than the safe limit. Therefore, the study of reactivity changes PRTF operation to test fuel pin is secure.</p><p><strong>Abstrak: </strong>Power Ramp Test Facility (PRTF) merupakan salah satu fasilitas iradiasi yang terdapat pada Reaktor Serba Guna G.A. Siwabessy. Fasilitas ini digunakan untuk menguji pin elemen bahan bakar reaktor tipe Pressurized Water Reactor. Akibat dari masuknya benda asing menyebabkan perubahan kondisi reaktor, salah satunya dinyatakan dengan besaran reaktivitas untuk mengkaji keselamatan reaktor akibat pengoperasian PRTF. Simulasi pengoperasian PRTF dan perhitungan netronik dilakukan menggunakan perangkat lunak MCNP6. Bahan bakar uji UO2 diasumsikan diperkaya sebesar 5% dengan daya operasi reaktor konstan sebesar 15 MW. Pin bahan bakar uji diletakkan pada PRTF berjarak 0, 20, 40, 60, 80, 100, 120, dan 140 mm dari arah pusat teras reaktor. Nilai perubahan reaktivitas yang dipersyaratkan agar reaktor aman adalah , sedangkan nilai perubahan reaktivitas dari penelitian pada masing-masing posisi dari pusat reactor adalah (;  ;  ; ;; ; ; ) . Nilai perubahan reaktivitas akibat masuknya pin bahan bakar di PRTF mempunyai nilai perubahan reaktivitas 1/10 kali lebih kecil daripada batas aman. Oleh karena itu, ditinjau dari kajian  nilai perubahan reaktivitas maka pengoperasian PRTF untuk uji pin bahan bakar adalah aman.</p>

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.

Sign in / Sign up

Export Citation Format

Share Document