Analysis of Advanced PWR Loading Schemes for Transuranic Incineration in Thorium

2013 ◽  
Author(s):  
Benjamin A. Lindley ◽  
N. Zara Zainuddin ◽  
Fausto Franceschini ◽  
Geoffrey T. Parks
Keyword(s):  
Temperature Coefficient ◽  
Preliminary Analysis ◽  
Detrimental Effect ◽  
Spatial Separation ◽  
Trade Off ◽  
The Core ◽  
Positive Reactivity ◽  
Fuel Pin ◽  
Control Rods

It is difficult to perform multiple recycle of transuranic (TRU) isotopes in PWRs as the moderator temperature coefficient (MTC) tends to become positive after a few recycles and the core may have positive reactivity when fully voided. Due to the favorable impact on the MTC and void coefficient fostered by use of thorium (Th), the possibility of performing Th-TRU multiple-recycle in reduced-moderation PWRs (RMPWRs) is under consideration. The simplest way to reduce the moderation in a PWR is to increase the fuel pin diameter. This configuration improves the trade-off between achievable burn-up and MTC, but is ultimately limited by thermal-hydraulic constraints. Heterogeneous recycle with the bred uranium (U3) and the TRU are arranged in separate pins was found to be neutronically preferable to a homogeneous configuration. Spatial separation also enables the U3 and TRU to be refueled on different batch schemes. These techniques allow satisfactory discharge burn-up while ensuring negative MTC and fully voided reactivity, with the pin diameter of a standard PWR increased from 9.5 mm to 11 mm. Reactivity control is a key challenge due to the reduced worth of neutron absorbers and their detrimental effect on the void coefficients, especially when diluted, as is the case for soluble boron. It seems necessary to control the core using control rods to keep the fully voided reactivity negative. A preliminary analysis indicates that this is feasible.

scholarly journals Spatial segregation of mixed-sized counterions in dendritic polyelectrolytes

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
J. S. Kłos ◽  
J. Paturej
Keyword(s):  
Electrostatic Interactions ◽  
Spatial Separation ◽  
Excluded Volume ◽  
The Core ◽  
Bjerrum Length ◽  
Two Component ◽  
Conformational Properties ◽  
Dynamics Simulations ◽  
Two Parameters ◽  
Swelling Factor

AbstractLangevin dynamics simulations are utilized to study the structure of a dendritic polyelectrolyte embedded in two component mixtures comprised of conventional (small) and bulky counterions. We vary two parameters that trigger conformational properties of the dendrimer: the reduced Bjerrum length, $$\lambda _B^*$$ λ B ∗ , which controls the strength of electrostatic interactions and the number fraction of the bulky counterions, $$f_b$$ f b , which impacts on their steric repulsion. We find that the interplay between the electrostatic and the counterion excluded volume interactions affects the swelling behavior of the molecule. As compared to its neutral counterpart, for weak electrostatic couplings the charged dendrimer exists in swollen conformations whose size remains unaffected by $$f_b$$ f b . For intermediate couplings, the absorption of counterions into the pervaded volume of the dendrimer starts to influence its conformation. Here, the swelling factor exhibits a maximum which can be shifted by increasing $$f_b$$ f b . For strong electrostatic couplings the dendrimer deswells correspondingly to $$f_b$$ f b . In this regime a spatial separation of the counterions into core–shell microstructures is observed. The core of the dendrimer cage is preferentially occupied by the conventional ions, whereas its periphery contains the bulky counterions.

scholarly journals Numerical investigations of stainless steel melt motions on the surface of uranium dioxide

2019 ◽  
Vol 196 ◽  
pp. 00005 ◽  
Author(s):  
Eduard V. Usov ◽  
Pavel D. Lobanov ◽  
Ilya A. Klimonov ◽  
Alexander E. Kutlimetov ◽  
Anton A. Butov ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Uranium Dioxide ◽  
Momentum Conservation ◽  
Mass Energy ◽  
Numerical Investigations ◽  
Fuel Rod ◽  
The Core ◽  
Fuel Pin ◽  
Energy And Momentum ◽  
Coolant Velocity

The paper contains the results of numerical simulation of stainless steel melt motions on the surface of uranium dioxide. The investigations are performed for purposes of understanding of the fuel rod behavior during the core disruptive accident in the fast reactors. The systems of mass, energy and momentum conservation equations are solved to simulate melt motion on the surface of the fuel pin. Heat transfer and friction between melt and pin's surface and melt and coolant flow are taken into consideration. The dependences of mass of the melt and the features of the melt motion on coolant velocity and contact angle between melt and surface of the fuel rod are presented.

Main Steam Line Break Calculations Using a Coupled RELAP5/PARCS Model for the Ringhals-3 Pressurized Water Reactor

2008 ◽  
Author(s):  
Mathias Sta˚lek ◽  
Jo´zsef Ba´na´ti ◽  
Christophe Demazie`re
Keyword(s):  
Steady State ◽  
Fuel Assembly ◽  
Thermal Power ◽  
Steam Line ◽  
Pressurized Water ◽  
The Core ◽  
Different Types ◽  
Main Steam Line ◽  
Main Steam ◽  
Control Rods

A Main Steam Line Break (MSLB) is an important transient for Pressurized Water Reactors (PWR) due to the strong positive reactivity introduced by the over-cooling of the core. Since this effect is stronger when the Moderator Temperature Coefficient (MTC) has a large amplitude, a conservative result will be obtained for a high burnup of the fuel due to the more negative MTC late in the cycle. The calculations have been performed at a cycle burnup of 12.9742 GWd/tHM. The Swedish Ringhals-3 PWR is a three loop Westinghouse design, currently with a thermal power of 3000 MW. The PARCS model has 157 fuel assemblies of 8 different types. Four different types of reflector are used. The cross sections, and kinetic data were obtained from CASMO-4 calculations, using a cross section interface developed at the department. There are 24 axial nodes, and 2×2 radial nodes for each assembly. The transient option for calculating the effect of poisoning was used. The PARCS model has been validated against steady-state measurements from Ringhals-3 of the Relative Power Fraction (RPF) and of the core criticality. The RELAP5 model has 157 channels for the core which means that there is a one to one correspondence between the thermal hydraulics model and the neutronics model. There is eight axial nodes. Originally, the intention was to have 24 axial nodes but this proved not to work because of some limitation in RELAP5. There is currently no mixing between the different channels in the core. The feedwater, and turbines are modelled as boundary conditions. The stand-alone RELAP5 model has been validated against steady state measurements from Ringhals-3. A number of different cases were considered. In the first case, both the isolation of the feedwater for the broken loop, and all the control rods were assumed to work properly. For the second case one of the control rods was assumed to be stuck. The stuck rod was located in the fuel assembly with the highest power. This rod has also one of the highest rod worths. In the final case, the feedwater control valve for the broken loop was fully open. None of the cases led to any recriticality. The increase in power for each fuel assembly was also investigated. With the control rod located in the assembly with the highest power, the maximum power increase before scram turned out to be about 25% compared to the initial power.

Development of REFT Code System for the Analysis of Highly Heterogeneous Research Reactors

2014 ◽  
Author(s):  
Tengfei Zhang ◽  
Hongchun Wu ◽  
Youqi Zheng ◽  
Liangzhi Cao ◽  
Yunzhao Li
Keyword(s):  
Least Square Method ◽  
Least Square ◽  
Core Analysis ◽  
Code System ◽  
Lattice Calculation ◽  
Control Rod ◽  
Research Reactors ◽  
The Core ◽  
User Friendly ◽  
Control Rods

As an effort to enhance the accuracy in simulating the operations of research reactors, a fuel management code system REFT was developed. Because of the possible complex assembly geometry and the core configuration of research reactors, the code system employed HELIOS in the lattice calculation to describe arbitrary 2D geometry, and used the 3D triangular nodal SN method transport solver, DNTR, to model unstructured geometry in the core analysis. Flux reconstruction with the least square method and micro depletion model for specific isotopes were incorporated in the code. At the same time, to make it more user friendly, a graphical user interface was also developed for REFT. In the analysis of the research reactors, the calculations involving the control rod movement are encountered frequently. The modeling of the control rods differential worth behavior is important in that the movement of the control rod may introduce variations on the reactivity. To handle the problem two effective ways of alleviating the control rod cusping effect are recently proposed, based on the established code system. The methodologies along with their application and validation will be discussed.

scholarly journals Measurements of CO2 and CH4 concentrations in air in a polar ice core

1993 ◽  
Vol 39 (132) ◽  
pp. 209-215 ◽  
Author(s):  
Takakiyo Nakazawa ◽  
Toshinobu Machida ◽  
Kenji Esumi ◽  
Masayuki Tanaka ◽  
Yoshiyuki Fujii ◽  
...  
Keyword(s):  
Gas Chromatography ◽  
Preliminary Analysis ◽  
Extraction Efficiency ◽  
Ice Core ◽  
Fine Powder ◽  
Glacial Period ◽  
Polar Ice ◽  
The Core ◽  
Precise Analysis ◽  
Better Than

AbstractDry and wet air-extraction systems and precise analysis systems of the CO2and CH4concentrations for a polar ice core were developed to reconstruct their ancient levels. A dry-extraction system was capable of crushing an ice sample of 1000 g into fine powder within 2 min, and its air-extraction efficiency was found to be 98%. The CO2and CH4concentrations of extracted air were determined using gas chromatography with a flame-ionized detector. The overall precision of our measurements, including air extraction, was estimated to be better than ± 1 ppmv for CO2and + 10 ppbv for CH4. Preliminary analysis of the ice core drilled at Mizuho Station, Antarctica, showed that the CO2and CH4concentrations at 3340–3700 year BP were about 280 ppmv and 700ppbv, respectively. The Yamato core drilled at the terminus of the glacial flow near the Yamato Mountains, Antarctica, yielded concentrations of 230–240 ppmv for CO2and 520–550 ppbv for CH4, suggesting that the core had formed during the glacial period.

scholarly journals Quantification of uncertainty in rapid estimation of earthquake fatalities based on scenario analysis

2018 ◽  
Author(s):  
Xiaoxue Zhang ◽  
Hanping Zhao ◽  
Fangping Wang ◽  
Zezheng Yan ◽  
Sida Cai ◽  
...  
Keyword(s):  
Emergency Response ◽  
Preliminary Analysis ◽  
Accurate Estimation ◽  
Earthquake Parameters ◽  
Rapid Estimation ◽  
Factors Affecting ◽  
The Core ◽  
Comparison Results ◽  
Emergency Scenarios ◽  
Earthquake Fatalities

Abstract. The rapid estimation of earthquake fatalities using earthquake parameters is the core basis for emergency response. However, there are numerous factors affecting earthquake fatalities, and it is impossible to obtain an accurate estimation result. The key to solve this problem is quantifying the uncertainty. In this paper, we proposed a new method to estimate earthquake fatalities and quantify the uncertainty based on basic earthquake emergency scenarios. The accuracy of the model is verified by earthquake that occurred during recent year. The preliminary analysis and comparison results show that the model is more effective and reasonable and can also provide a theoretical basis for post-earthquake emergency response.

Monte Carlo Simulation of Inclined Reactivity Control Rod and Boron Poisoning for the Canadian-SCWR Supercell

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Haykel Raouafi ◽  
Guy Marleau
Keyword(s):  
Monte Carlo ◽  
Three Dimensional ◽  
Coolant Temperature ◽  
Monte Carlo Code ◽  
Control Rod ◽  
Vertical Control ◽  
The Core ◽  
Core Height ◽  
Supercell Model ◽  
Control Rods

The Canadian-SCWR is a heavy-water moderated supercritical light-water-cooled pressure tube reactor. It is fueled with CANada deuterium uranium (CANDU)-type bundles (62 elements) containing a mixture of thorium and plutonium oxides. Because the pressure tubes are vertical, the upper region of the core is occupied by the inlet and outlet headers render it nearly impossible to insert vertical control rods in the core from the top. Insertion of solid control devices from the bottom of the core is possible, but this option was initially rejected because it was judged impractical. The option that is proposed here is to use inclined control rods that are inserted from the side of the reactor and benefit from the gravitational pull exerted on them. The objective of this paper is to evaluate the neutronic performance of the proposed inclined control rods. To achieve this goal, we first develop a three-dimensional (3D) supercell model to simulate an inclined rod located between four vertical fuel cells. Simulations are performed with the SERPENT Monte Carlo code at five axial positions in the reactor to evaluate the effect of coolant temperature and density, which varies substantially with core height, on the reactivity worth of the control rods. The effect of modifying the inclination and spatial position of the control rod inside the supercell is then analyzed. Finally, we evaluate how boron poisoning of the moderator affects their effectiveness.

Prompting additional purchases while providing service: does it offend the customer?

2016 ◽  
Vol 26 (5) ◽  
pp. 657-680 ◽  
Author(s):  
Hsuan-Hsuan Ku ◽  
Chih-Yun Huang
Keyword(s):  
Design Methodology ◽  
Psychological Reactance ◽  
Self Monitoring ◽  
Trade Off ◽  
Content Type ◽  
The Core ◽  
Personal Variables ◽  
Core Service ◽  
Perceived Relevance ◽  
Supplementary Service

Purpose The purpose of this paper is to investigate consumers’ responses to unsolicited cross-selling of supplementary paid-for services made during delivery of a core service, and the contextual and personal variables moderating those responses. Design/methodology/approach Three formal experiments test the effect on participants’ responses of the perceived relevance of the supplementary service to the core service, personal psychological reactance, in the case of a high-relevance supplementary service, and self-monitoring, in scenarios in which a low-relevance supplementary service is proposed either in public or privately. Findings The experiments found that participants’ satisfaction ratings were reduced in response to cross-selling of a supplementary service that was of low relevance to the core service, and that satisfaction ratings if it was perceived to be of high relevance compared were not reduced despite the unsolicited attempt at cross-selling. However, the non-negative response to a high-relevance offer was limited to participants with a lower tendency to reactance. Furthermore, a high predisposition to self-monitoring evoked more positive judgments if a low-relevance supplementary service was proposed in public rather than privately. That of low self-monitors was no different in either case. Originality/value This paper examines the trade-off faced by a service provider between customer satisfaction and extra revenue from supplementary services, and explores conditions under which a provider can propose unsolicited supplementary services without offending customers.

Neutronics Optimization of UN Fuel Pin Assemblies for a Sodium-Cooled, Small Modular Reactor

2011 ◽  
Author(s):  
Timothy M. Schriener ◽  
Mohamed S. El-Genk
Keyword(s):  
Thermal Power ◽  
Reactor Core ◽  
Central Cavity ◽  
Preliminary Results ◽  
The Core ◽  
Small Modular Reactor ◽  
Fuel Assemblies ◽  
Fuel Pin ◽  
Reference Design ◽  
Centerline Temperature

This paper presents preliminary results of neutronics and thermal-hydraulics design analysis of a sodium cooled, small modular reactor (SMR). The reactor’s nominal thermal power is 150 MWth at sodium inlet and exit temperatures of 630 and 780 K. The reactor core is comprised of three rings of shrouded hexagonal assemblies of 19.8% enriched UN fuel pins and a hexagonal assembly of enriched B4C pins in the central cavity for a coarse reactivity control. The objectives are to provide enough excess reactivity for achieving a refueling cycle > 5 year, maintaining a more even coolant flow in the core assemblies and keeping the peak centerline temperature of UN fuel pins < 1300 K. Fuel assemblies with scalloped shroud walls, 4 rings and 1.942 cm diameter fuel pins with p/d = 1.098 are selected for the reference design of the present SMR. In this design, peak fuel centerline temperature is only 1240 K and the beginning-of-life, cold-clean excess reactivity is $26.67.

scholarly journals DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

2020 ◽  
Vol 5 (3) ◽  
pp. 239-248
Author(s):  
Tukiran Surbakti ◽  
Surian Pinem ◽  
Lily Suparlina
Keyword(s):  
Neutron Flux ◽  
Thermal Neutron ◽  
Cycle Length ◽  
Thermal Neutron Flux ◽  
Diffusion Method ◽  
Irradiation Position ◽  
Research Reactors ◽  
The Core ◽  
Neutronic Parameters ◽  
Control Rods

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

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