Physics of Plutonium and Americium Recycling in PWR Using Advanced Fuel Concepts

2004 ◽  
Author(s):  
E. Hourcade
Keyword(s):  
Inventory Management ◽  
Power Distribution ◽  
Fuel Cycle ◽  
Cooling System ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Fuel Rods ◽  
Inert Matrix ◽  
Water Reactors ◽  
Set Up

PWR waste inventory management is considered in many countries including Frances as one of the main current issues. On this subject, the French 1991 Bataille’s law set up a 15 years research program on three main axes: sub-surface storage, deep geological storage, transmutation using critical or subcritical burners. Amongst the output Actinides, Pu and Am are the 2 main contents both in term of volume and long term radio-toxicity. Waiting for the Generation IV systems implementation (2035–2050), one of the mid-term solutions for their transmutation involves the use of advanced fuels in Pressurized Water Reactors (PWR). These have to require as little modification as possible of the core internals, the cooling system and fuel cycle facilities (fabrication and reprocessing). The present paper is focussed on the reactor physics characteristics, as a preliminary step in the evaluation of options, knowing that others homogeneous and heterogeneous assemblies have been studied by the CEA ([1] to [5]). The main neutronic parameters to be considered for Pu and Am recycling in PWR are void coefficient (αvoid), Doppler coefficient (αDopp), fraction of delayed neutrons (β) and power distribution (especially for heterogeneous configurations). The modification of the moderation ratio, the opportunity to use inert matrices (targets), the optimisation of Uranium, Plutonium and Americium contents are the key parameters to play with. One of the solutions presented here is a heterogeneous assembly with regular moderation ratio composed with both target fuel rods (Pu and Am embedded in an inert matrix) and standard UO2 fuel rods. An EPR (European Pressurised Reactor) type reactor, loaded only with assemblies containing 84 peripheral targets, can reach an Americium consumption rate of [4.4; 23 kg/TWhe] depending on the assembly concept. For Pu and Am inventories stabilisation, the theoretical fraction of reactors loaded with Pu + Am or Pu assemblies is about 60%. For Americium inventory stabilisation, the fraction decreases down to 16%, but Pu is produced at a rate of 18.5 Kg/Twhe (−25% compared to one through UOX cycle).

Advanced First Core Design for the Westinghouse AP1000

2009 ◽  
Author(s):  
Robert J. Fetterman
Keyword(s):  
Fuel Cycle ◽  
The United States ◽  
Pressurized Water ◽  
Loading Pattern ◽  
The Core ◽  
Low Leakage ◽  
Fuel Assemblies ◽  
Cycle Loading ◽  
Water Reactors ◽  
Under Construction

As the nuclear renaissance is now upon us and new plants are either under construction or being ordered, a considerable amount of attention has also turned to the design of the first fuel cycle. Requirements for core designs originate in the Utilities Requirements Document (URD) for the United States and the European Utilities Requirements (EUR) for Europe. First core designs created during the development of these documents were based on core design technology dating back to the 1970’s, where the first cycle core loading pattern placed the highest enrichment fuel on the core periphery and two other lower enrichments in the core interior. While this sort of core design provided acceptable performance, it underutilized the higher enriched fuel assemblies and tended to make transition to the first reload cycle challenging, especially considering that reload core designs are now almost entirely of the Low Leakage Loading Pattern (LLLP) design. The demands placed on today’s existing fleet of pressurized water reactors for improved fuel performance and economy are also desired for the upcoming Generation III+ fleet of plants. As a result of these demands, Westinghouse has developed an Advanced First Core (AFCPP) design for the initial cycle loading pattern. This loading pattern design simulates the reactivity distribution of an 18 month low leakage reload cycle design by placing the higher enriched assemblies in the core interior which results in improved uranium utilization for those fuel assemblies carried through the first and second reload cycles. Another feature of the advanced first core design is radial zoning of the high enriched assemblies, which allows these assemblies to be located in the core interior while still maintaining margin to peaking factor limits throughout the cycle. Finally, the advanced first core loading pattern also employs a variety of burnable absorber designs and lengths to yield radial and axial power distributions very similar to those found in typical low leakage reload cycle designs. This paper will describe each of these key features and demonstrate the operating margins of the AFC design and the ability of the AFC design to allow easy transition into 18 month low leakage reload cycles. The fuel economics of the AFC design will also be compared to those of a more traditional first core loading pattern.

Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors

2017 ◽  
Author(s):  
Zhixiong Tan ◽  
Jiejin Cai
Keyword(s):  
Silicon Carbide ◽  
Fuel Assembly ◽  
Power Distribution ◽  
Safety Margin ◽  
Severe Accident ◽  
Pressurized Water Reactors ◽  
Fuel Pellet ◽  
Pressurized Water ◽  
Neutronic Parameters ◽  
Water Reactors

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.

A Challenge for a High-Resolution Ag-In-Cd Decoupler in Intensified Short-Pulsed Neutron Source

2010 ◽  
Vol 652 ◽  
pp. 92-98 ◽  
Author(s):  
Kenji Kikuchi ◽  
Makoto Teshigawara ◽  
Masahide Harada ◽  
Shigeru Saito ◽  
Fujio Maekawa ◽  
...  
Keyword(s):  
Neutron Source ◽  
Hot Isostatic Pressing ◽  
Peak Position ◽  
Al Alloy ◽  
Pressurized Water ◽  
Control Rod ◽  
Water Reactors ◽  
Set Up ◽  
Pulsed Neutron ◽  
Pulsed Neutron Source

A sharp pulsed-neutron beam, which has narrow width and large attenuation, will enhance ability to separate peak position diffracted from crystal structure in time-of-flight measurement method. Traditional technique to satisfy the request above is to use Boron for decoupling higher cut-off energy. Disadvantage, however, will be an excess formation of He bubbles and large heat deposition in the materials as a result of (n,α) reaction, particularly in a MW class of neutron source. Alternative is to use a material combination of Ag-In-Cd, which has different energies for resonance absorption and often used in the control rod of pressurized water reactors with cladding stainless steel. For application of this material to pulsed-neutron source a challenge is to make sandwiches structure with Al alloy because a moderator is made of this material from points of views of heat removing. Technically hot isostatic pressing was a choice and devoted to realize a bonding strength through R&Ds. Final product was set up to MLF /J-PARC and a successful resolution was observed in a powder diffractometer.

Neutronics Studies on a Generic Pressure-Channel Reactor With Nuclear Steam Superheat

2014 ◽  
Author(s):  
Matthew Baldock ◽  
Wargha Peiman ◽  
Andrei Vincze ◽  
Rand Abdullah ◽  
Khalil Sidawi ◽  
...  
Keyword(s):  
Power Distribution ◽  
Power Plants ◽  
Saturation Pressure ◽  
Pressure Vessels ◽  
Limiting Factor ◽  
Outlet Temperature ◽  
Pressurized Water ◽  
Water Reactors ◽  
1960S And 1970S ◽  
Pressure Channel

In order to increase the thermal efficiency of steam-cycle power plants it is necessary to achieve steam temperatures as high as possible. Current limiting factor for Nuclear Power Plants (NPPs) in achieving higher operating temperatures and, therefore, thermal efficiencies is pressures at which they can operate. From basic thermodynamics it is known that to increase further an outlet temperature in water-cooled reactors a pressure must also be increased. Current level of pressures in Pressurized Water Reactors (PWRs) is about 15–16 MPa. Therefore, next stage should be supercritical pressures, at least 23.5–25 MPa. However, such supercritical-water reactors with pressure vessels of 45–50 cm thickness don’t exist yet. One way around larger pressure vessels as well as the limit of temperature of the coolant on the saturation pressure is to employ a Pressure Channel (PCh) design with Superheated Steam channels (SHS). PCh reactors allow for different coolants and bundle configurations in one reactor core, in this case, steam would be a secondary coolant. In the 1960s and 1970s the USA and Soviet Union tested reactors using pressure channels to super-heat steam in-core to achieve outlet temperatures greater than what is currently possible with convention reactors. Nuclear materials are carefully chosen based on their neutron interaction properties in addition to their strength and resistance to corrosion. Introducing steam channels will not only change the neutronics behavior of the coolant, but require different fuel cladding and pressure-channel materials, specifically, stainless steels or Inconels, to withstand high-temperature steam. This paper will investigate the affect that steam, SS-304 and Inconel will have on neutron economy when introduced into a reactor design as well as required changes to fuel enrichment. It will also be necessary to investigate the effects of these material changes on power distribution inside a reactor. Pressure-channel design requires methods of fine control to maintain a balanced core-power distribution, the introduction of non-uniform coolant and reactor materials will further complicate maintaining uniform reactor power. The degree to which SHS channels will affect the power distribution is investigated in this paper.

A Procedure for the Pre-Conceptual Design of Fast Reactors and Application to a Gas-Cooled Sub-Critical Transmuter

2014 ◽  
Author(s):  
Carlo Fiorina ◽  
Konstantin Mikityuk ◽  
Jiři Křepel
Keyword(s):  
Conceptual Design ◽  
Fast Reactor ◽  
Power Distribution ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Maximum Pressure ◽  
Test Case ◽  
Pressure Losses ◽  
Design And Optimization ◽  
Inert Matrix

A C++ procedure has been developed for the design and optimization of Fast Reactor (FR) cores. It couples the ERANOS based EQL3D procedure developed at PSI for FR equilibrium fuel cycle analysis with a dedicated MATLAB script that evaluates the thermal-hydraulic characteristics of the reactor core. It is conceived to investigate reactors with both standard pins and annular pins. The procedure accepts as input the physical properties of the system, as well as a set of target core parameters presently consisting of core power, maximum fuel burnup, multiplication factor, inner pin diameter (for annular pins) or maximum pressure losses (for standard pins), and core height. It gives as a result a core design fulfilling these design objectives and meeting the constraints on maximum fuel and clad temperatures. In case of annular pins, it also equalizes the temperature rise inside and outside of the core average pin. The procedure considers the possibility of two-zone cores and adjusts the fuel composition in the two zones to achieve an optimal radial power distribution. Finally, it can evaluate safety parameters and fuel cycle characteristics both at beginning-of-life and at equilibrium. As a test case, the procedure has been used for the pre-conceptual design of a sub-critical Gas Fast Reactor core employing inert-matrix sphere-pac fuel and annular pins with SiC cladding.

A Fuel Cycle and Core Design Analysis Method for New Cladding Acceptance Criteria Using PHISICS, RAVEN and RELAP5-3D

2016 ◽  
Author(s):  
Andrea Alfonsi ◽  
George L. Mesina ◽  
Angelo Zoino ◽  
Cristian Rabiti
Keyword(s):  
Fuel Cycle ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Computational Time ◽  
Maximum Temperature ◽  
Reactor Accident ◽  
Reactor Physics ◽  
Pressurized Water ◽  
Safety Margins ◽  
Regulatory Commission

The Nuclear Regulatory Commission (NRC) has considered revising the 10 CFR 50.46C rule [1] for analyzing reactor accident scenarios to take the effects of burn-up rate into account. Both maximum temperature and oxidation of the cladding must be cast as functions of fuel exposure in order to find limiting conditions, making safety margins dynamic limits that evolve with the operation and reloading of the reactor. In order to perform such new analysis in a reasonable computational time with good accuracy, INL (Idaho National Laboratory) has developed new multi-physics tools by combining existing codes and adding new capabilities. The PHISICS (Parallel Highly Innovative Simulation INL Code System) toolkit [2,3] for neutronic and reactor physics is coupled with RELAP5-3D [4] (Reactor Excursion and Leak Analysis Program) for the LOCA (Loss of Coolant Accident) analysis and RAVEN [5] for the PRA (Probabilistic Risk Assessment) and margin characterization analysis. In order to perform this analysis, the sequence of RELAP5-3D input models had to get executed in a sequence of multiple input decks, each of them had to restart and slightly modify the previous model (in this case, on the neutronic side only) This new RELAP5-3D multi-deck processing capability has application to parameter studies and uncertainty quantification. The combined RAVEN/PHISICS/RELAP5-3D tool is used to analyze a typical PWR (Pressurized Water Reactor).

FE Simulation of Welding Residual Stresses in Reactor Vessel Head Penetrations

2009 ◽  
Author(s):  
Anne Teughels ◽  
Christian Malekian
Keyword(s):  
Residual Stresses ◽  
Nuclear Power ◽  
Power Plants ◽  
Element Model ◽  
Mechanical Analysis ◽  
Alloy 600 ◽  
Finite Elements Analysis ◽  
Pressurized Water ◽  
Water Reactors ◽  
Set Up

The penetrations in the early Pressurized Water Reactors Vessels are characterized by Alloy 600 tubes, welded by Alloy 182/82. The Alloy 600 tubes have been shown to be susceptible to PWSCC (Primary Water Stress Corrosion Cracking) which may lead to crack forming. The cracking mechanism is driven mainly by the welding residual stresses and, in a second place, by the operational stresses in the weld region. It is therefore of big interest to quantify the weld residual stresses correctly. In order to determine the welding residual stresses, the weld procedure is simulated numerically by finite elements analysis. In the article, central as well as eccentric sidehill nozzles on the vessel head are analyzed. For the former a 2-dimensional axisymmetrical finite element model is used, whereas for the latter a 3-dimensional model is set up. A nonlinear transient thermo-mechanical analysis is performed, which is preceded by a transient thermal analysis simulating the heating during the multipass welding. Weld beads are deposited “all-at-once”. Different positions on the vessel head are compared and the influence of the sidehill effect is illustrated. The methodology is applied to the reactor vessels of the Belgian nuclear power plants by Tractebel Engineering (Belgium). The results are compared with literature. The global approach in both cases is very similar but is applied to different configurations, specific for each plant.

Sign in / Sign up

Export Citation Format

Share Document