scholarly journals ANALYSIS OF REACTIVITY COEFFICIENT CHANGE DUE TO BURN UP IN AP1000 REACTOR CORE USING NODAL3

2017 ◽  
Vol 19 (3) ◽  
pp. 131
Author(s):  
Iman Kuntoro ◽  
Surian Pinem ◽  
Tagor Malem Sembiring
Keyword(s):  
Reactor Core ◽  
Core Material ◽  
Inherent Safety ◽  
Temperature Coefficients ◽  
Fuel Burn ◽  
Reactivity Coefficients ◽  
Safety Parameters ◽  
Safety Parameter ◽  
Moderator Density ◽  
Reactivity Coefficient

One of the important things in reactor safety is the value of inherent safety parameter namely reactivity coefficient. These inherent safety parameters are fuel and moderator temperature coefficients of reactivity.  The objective of the study is to obtain the change of those reactivity coefficients as a function of fuel burn up during the cycle operation of AP 1000 reactor core. Fuel and moderator temperature coefficients of reactivity and in addition moderator density coefficient of reactivity were calculated using SRAC 2006 and NODAL3 computer codes. Cross section generation of all core material was done by SRAC 2006 Code. The calculation of core reactivity as a function of temperature and burn up were carried out using NODAL3 Code. The results show that all reactivity coefficients of AP 1000 reactor core are always negative during the operation cycles and the values are in a good agreement to the design. It can be concluded that the AP 1000 core has a good inherent safety of its fuelKeywords: reactivity coefficient, burn up, AP1000, NODAL3. ANALISIS PERUBAHAN KOEFISIEN REAKTIVITAS AKIBAT FRAKSI BAKAR TERAS REAKTOR AP1000 MENGGUNAKAN NODAL3.  Salah satu hal yang sangat penting dalam analisis kecelakaan pada reactor daya adalah koefisien reaktivitas untuk mengontrol daya reaktor. Penelitian ini bertujuan menentukan koefisien reaktivitas akibat perubahan fraksi bakar pada reaktor AP1000. Koefisien reaktivitas yang akan dihitung adalah koefisien reaktivitas bahan bakar dan moderator yang sering disebut inherent factor. Selain itu juga akan dihitung koefisien konsentrasi boron dan kerapatan moderator.  Semua koefisien reaktivitas ini dihitung saat terjadi perubahan fraksi bakar untuk mempertimbangkan produk fisi dan konsumsi bahan bakar. Perhitungan neutronik teras reactor disimulasi dengan menggunakan program SRAC2006 dan NODAL3. Perhitungan tampang lintang seluruh perangkat bahan bakar dan batang kendali reaktor AP1000 dilakukan dengan program SRAC2006. Perhitungan parameter neutronik sebagai fungsi temperature dan fraksi bakar dilakukan menggunakan program NODAL3. Perhitungan koefisien reaktivitas ditentukan berdasarkan perbedaan nilai reaktivitas. Hasil perhitungan menunjukkan bahwa koefisien reaktivitas teras reaktor AP 1000 selalu berharga negative untuk sepanjang siklus operasinya dan mendekati harga desain. Kesimpulan yang dapat ditarik adalah bahwa teras AP 10000 mempunyai keselamatan melekat yang baik.Kata kunci:  koefisien reaktivitas, fraksi bakar, AP 1000, NODAL3.

scholarly journals Improved FAST Device for Inherent Safety of Oxide-Fueled Sodium-Cooled Fast Reactors

Energies ◽  
2021 ◽  
Vol 14 (15) ◽  
pp. 4610
Author(s):  
Ahmed Amin E. Abdelhameed ◽  
Chihyung Kim ◽  
Yonghee Kim
Keyword(s):  
Operating Temperature ◽  
Reactor Core ◽  
Coolant Temperature ◽  
Inherent Safety ◽  
Oxide Fuel ◽  
Fast Reactors ◽  
Power Oscillation ◽  
Negative Reactivity ◽  
Oscillation Problem ◽  
Reactivity Coefficient

The floating absorber for safety at transient (FAST) was proposed as a solution for the positive coolant temperature coefficient in sodium-cooled fast reactors (SFRs). It is designed to insert negative reactivity in the case of coolant temperature rise or coolant voiding in an inherently passive way. The use of the original FAST design showed effectiveness in protecting the reactor core during some anticipated transients without scram (ATWS) events. However, oscillation behaviors of power due to refloating of the absorber module in FAST were observed during other ATWS events. In this paper, we propose an improved FAST device (iFAST), in which a constraint is imposed on the sinking (insertion) limit of the absorber module in FAST. This provides a simple and effective solution to the power oscillation problem. Here, we focus on an oxide fuel-loaded SFR that is characterized by a more negative Doppler reactivity coefficient and higher operating temperature than the metallic-loaded SFR cores. The study is carried out for the 1000 MWth advanced burner reactor with an oxide fuel-loaded core during postulated ATWS events that are unprotected transient over power, unprotected loss of flow, and unprotected loss of the heat sink. It was found that the iFAST device has promising potentials for protecting the oxide SFR core during the various studied ATWS events.

scholarly journals Investigation of the Safety Features of Advanced PWR Assembly Using SiC, Zr, FeCrAl and SS-310 As Cladding Materials

2020 ◽  
Author(s):  
Sayed Mustafa
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Delayed Neutron ◽  
Strong Trend ◽  
Geometric Conditions ◽  
Reactivity Coefficients ◽  
Safety Parameters ◽  
Reactivity Coefficient ◽  
Safety Features

Abstract In this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX code as cladding materials in advanced PWR assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 µm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

Modeling the Spectral History in the Depletion of a PWR Core

2010 ◽  
Author(s):  
I. Bilodid
Keyword(s):  
Cross Sections ◽  
Uranium Oxide ◽  
Reactor Core ◽  
Actual Distribution ◽  
History Effects ◽  
Fuel Assemblies ◽  
Burnable Absorber ◽  
Uranium Plutonium ◽  
Moderator Density ◽  
Local Parameters

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

scholarly journals Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

2017 ◽  
Vol 2017 ◽  
pp. 1-12 ◽  
Author(s):  
Shengli Chen ◽  
Cenxi Yuan
Keyword(s):  
Power Distribution ◽  
Current System ◽  
Negative Temperature ◽  
New Combination ◽  
Fuel Cladding ◽  
Fuel Temperature ◽  
Service Lifetime ◽  
Reactivity Coefficients ◽  
Reactivity Coefficient ◽  
Control Rods

Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

scholarly journals A study into the dependence of the cladding-fuel pellet gap conductance on burn-up and the effects on the reactor core neutronic performance

2019 ◽  
Vol 5 (2) ◽  
pp. 97-102
Author(s):  
Sergey B. Vygovsky ◽  
Fedor V. Gruzdov ◽  
Rashdan T. Al Malkawi
Keyword(s):  
Fuel Cycle ◽  
Accumulation Rate ◽  
Reactor Core ◽  
Neutron Field ◽  
Fuel Load ◽  
Fuel Pellet ◽  
Load Following ◽  
Fuel Burn ◽  
Numerical Research ◽  
Neutronic Performance

This paper presents the results of the research to study the dependence of the VVER-1000 (1200) cores neutronic characteristics on the cladding – fuel pellet gap conductance coefficient in the process of the fuel burn-up. The purpose of the study was to determine more accurately the dependence of the cladding – fuel pellet gap conductance coefficient on the fuel burn-up as shown in the Final Safety Report for the Bushehr NPP and to determine the extent of the effects this dependence had on the spatial distribution of the neutron field, on the xenon accumulation rate, and on the kinetic and dynamic behavior of the reactor facility. The paper presents the results of calculating the parameters using which the heat engineering safety of the reactor core is monitored in the process of the fuel burn- up (for a generalized fuel load of a VVER-1000) during the transition to an 18-month nuclear fuel cycle. This paper also includes the results of a numerical research to determine the cladding – fuel gap conductance coefficient depending on the fuel burn-up. These results have shown that, in reality, the gap conductance coefficient dependence on the burn-up does not affect greatly the steady-state characteristics. At the same time, it affects to rather a great extent the xenon accumulation rate, specifically in the event of an extended fuel life. In conditions of maneuvering (load following) modes accompanied by the xenon processes in the reactor core. These facts should be into consideration in design of engineering codes, that used to support the operation of the VVER-1000 (1200) and full-scale simulators.

scholarly journals SPX Benchmark Part I: Results of Static Neutronics

2021 ◽  
Author(s):  
Alexander Ponomarev ◽  
Konstantin Mikityuk ◽  
Liang Zhang ◽  
Evgeny Nikitin ◽  
Emil Fridman ◽  
...  
Keyword(s):  
Monte Carlo ◽  
Power Distribution ◽  
Reactor Core ◽  
Thermal Conditions ◽  
Reaction Rates ◽  
Second Phase ◽  
Reference Solution ◽  
Transient Behaviour ◽  
The Core ◽  
Reactivity Coefficients

Abstract In the paper, the specification of a new neutronics benchmark for a large Sodium cooled Fast Reactor core and results of modelling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and results of the reference solution. Different core geometries and thermal conditions from cold “as fabricated” up to full power were considered. The reference Monte Carlo solution of Serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modelling with seven other solutions using deterministic and Monte Carlo methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for evaluation of transient behaviour of the core.

scholarly journals SEARCH FOR THE BEST POWER CONTROL PROGRAM AT NPP WITH VVER-1000 USING GRADIENT DESCENT METHOD

2017 ◽  
Vol 8 (3) ◽  
Author(s):  
S. N. Pelykh ◽  
E. A. Odrehovska ◽  
O. B. Maksymova
Keyword(s):  
Power Control ◽  
Objective Function ◽  
Gradient Descent ◽  
Nuclear Power ◽  
Reactor Core ◽  
Control Program ◽  
Physical Nature ◽  
Descent Method ◽  
Gradient Descent Method ◽  
Fuel Burn

This article is regarded to the search for the best power control program at nuclear power plant (NPP) with VVER- 1000 by gradient descent method for the objective function, which includes the criteria of efficiency, safety and damage. Criteria normalization to the maximum value is carried out when looking for the minimum of the objective function because criteria have different physical nature. There were chosen such objective criteria as depth of fuel burn-up, index of the fuel cladding damage and axial offset - the ratio of the energy at the top and bottom of the reactor core.

Numerical Simulation of Concrete Ablation and Corium Cooling for Molten Corium-Concrete Interaction (MCCI)

2020 ◽  
Author(s):  
Zhichun Xu ◽  
Yapei Zhang ◽  
G. H. Su ◽  
Wenxi Tian ◽  
Suizheng Qiu
Keyword(s):  
Water Injection ◽  
Reactor Core ◽  
Reactor Vessel ◽  
Severe Accident ◽  
Sensitivity Analyses ◽  
Core Material ◽  
Melt Temperature ◽  
Concrete Type ◽  
Depth Study ◽  
Water Reactors

Abstract In a postulated severe accident situation in Light Water Reactors (LWRs), if the core fuel cannot be effectively cooled, the reactor core material will be heated and form a molten corium in the lower head. When the lower plenum of the reactor vessel fails, the molten corium may flow into the cavity under the reactor vessel and react with the concrete. This process, known as Molten Corium Concrete Interaction (MCCI), is characterized by concrete ablation and oxidation of metal in the corium, both of which produce a large amount of combustible and non-condensable gases, threatening the integrity of the containment. Thus in-depth study of the characteristics of concrete ablation and corium cooling have great significance. In the present study, an MCCI analysis code, MOQUICO (molten corium concrete interaction and corium cooling code, QUI means quintic) has been developed. The MACE M3b and OECD/MCCI CCI-3 tests were analyzed to validate the developed code. The melt temperature, axial and radial ablation depths, upward heat flux were calculated and were in good agreement with the experimental measurements, which proved that the code is capable of simulating MCCI and related phenomena of LWRs. Sensitivity analyses on the factors of decay heat, concrete type and water injection moment were performed and analyzed.

scholarly journals Investigation of the safety features of advanced PWR assembly using SiC, Zr, FeCrAl and SS-310 as cladding materials

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Sayed. Saeed. Mustafa
Keyword(s):  
Cross Sections ◽  
Power Distribution ◽  
Fuel Pellet ◽  
Delayed Neutron ◽  
Pressurized Water ◽  
Strong Trend ◽  
Geometric Conditions ◽  
Reactivity Coefficients ◽  
Extended Code ◽  
Safety Parameters

AbstractIn this work, SiC (Silicon carbide), FeCrAl (ferritic), SS-310 (stainless steel 310) and Zirconium are simulated by MCNPX (Monte Carlo N‐Particle eXtended) code as cladding materials in advanced PWR (Pressurized Water Reactor) assembly. A number of reactor safety parameters are evaluated for the candidate cladding materials as reactivity, cycle length, radial power distribution of fuel pellet, reactivity coefficients, spectral hardening, peaking factor, thermal neutron fraction and delayed neutron fraction. The neutron economy presented by Zr and SiC models is analyzed through the burnup calculations on the unit cell and assembly levels. The study also provided the geometric conditions of all cladding materials under consideration in terms of the relation between fuel enrichment and cladding thickness from the viewpoint to achieve the same discharge burnup as the Zircaloy cladding. It was found that the SiC model participated in extending the life cycle by 2.23% compared to Zr. The materials other than SiC largely decreased discharge burnup in comparison with Zircaloy. Furthermore, the claddings with lower capture cross-sections (SiC and Zr) exhibit higher relative fission power at the pellet periphery. The simulation also showed that using SiC with a thickness of 571.15 μm and 4.83% U-235 can satisfy the EOL irradiation value as Zr. For reactivity coefficient, the higher absorbing materials (SS-310 and FeCrAl) exhibit more negative FTCs, MTCs and VRCs at the BOL But, at the intermediate stages of burnup Zr and SiC have a strong trend of negative reactivity coefficients. Finally, the delayed neutron fraction of SiC and Zr models is the highest among all the four models.

Analysis of Doppler reactivity of SMART reactor core for hybrid fuel configurations of UO2, MOX and (Th/U)O2 using OpenMC

Kerntechnik ◽  
2021 ◽  
Vol 86 (3) ◽  
pp. 229-235
Author(s):  
Y. Alzahrani ◽  
K. Mehboob ◽  
F. A. Abolaban ◽  
H. Younis
Keyword(s):  
Reactor Core ◽  
Multiplication Factor ◽  
The Core ◽  
Mox Fuel ◽  
The Monte Carlo Method ◽  
Burnable Absorber ◽  
Effective Multiplication Factor ◽  
Reactivity Coefficient ◽  
Increasing Temperature ◽  
Effective Multiplication

Abstract In this study, the Doppler reactivity coefficient has been investigated for UO2, MOX, and (Th/U)O2 fuel types. The calculation has been carried out using the Monte Carlo method ( OpenMC). The effective multiplication factor keff has been evaluated for three materials with four different configurations without Integral Fuel Burnable Absorber (IFBA) rods and soluble boron. The results of MOX fuel, homogenous and heterogeneous thorium fuel configuration has been compared with the core of the reference fuel assembly (UO2). The calculation showed that the effective multiplication factor at 1 000 K was 1.26052, 1.14254, 1.22018 and 1.23771 for reference core, MOX, homogenous and heterogeneous configurations respectively. The results shows that reactivity has decreased with increasing temperature while the doppler reactivity coefficient remained negative. Moreover, the use of (Th/U)O2 homogenous and heterogeneous configuration had shown an improved response compared to the reference core at 600 K and 1 000 K. The doppler reactivity coefficient has been found as –8.98E-3 pcm/K, -0.8 655 pcmK for the homogenous and –8.854 pcm/K, -1.2253 pcm/K for the heterogeneous configuration. However, the pattern remained the same as for the reference core at other temperature points. MOX fuel has shown less response compared to the other fuel configuration because of the high resonance absorption coefficient of Plutonium. This study showed that the SMART reactor could be operated safely with investigated fuel and models.

Sign in / Sign up

Export Citation Format

Share Document