Design of U3Si2-Al Plate-Type Fuel Element for China Advanced Research Reactor

China Advanced Research Reactor (CARR), which will be critical in China Institute of Atomic Energy (CIAE) in 2010, is a multipurpose, high neutron flux and tank-type (inverse neutron trap) reactor with compact core. Its nominal reactor power is 60MW and the maximum thermal neutron flux is about 8.0×1014n/cm2·s in heavy water tank. It has a cylindrical core having a diameter of about 450mm and a height of 850mm. The CARR’s core consists of seventeen plate-type standard fuel elements and four follower fuel elements, initially loaded with 10.97 kg of 235U. The fuel element has been designed with U3S2-Al dispersion containing 235U of (19.75±0.20)wt.% low enriched uranium (LEU) and having a density of 4.3gU/cm3. The aluminum alloy is used as the cladding. There are twenty-one and seventeen fuel plates in the standard and follower fuel element, respectively. There are specific requirements for design of the fuel element and strict limitation for the operation parameters due to the high heat flux and high velocity of coolant in CARR. Irradiation test of fuel element had been carried out at fuel element power of 3.1±20%MW at Russia MIR reactor. Average burnup of fuel element is up to 40%. This paper deals with the detailed design of fuel element for CARR, out-pile and in-pile test projects, including selection of fuel and structure material, description of element structure, miniplates and fuel element irradiation experiment, measurement of properties of fuel plate, fabrication of fuel element and test results.

Study on SCWR Fuel Assembly Design With Thermal-Hydraulic and Neutron-Physical Characteristics

Physical characteristics of the coolant in the Supercritical-pressure Light Water Cooled Reactor (SCWR) vary greatly near the pseudo-critical point, which will cause variations of core neutron cross section and then bring about power perturbation. Further it will prompt the corresponding thermal parameters of supercritical water changed, and form feedback action, finally resulting in intensely coupled thermal-hydraulics and neutron-physical. Proper fuel assembly has been selected as research object, and the model of multiple parallel channels has been established. In view of this model, using DRAGON code for neutron-physical calculations and developing corresponding thermal-hydraulic programs, and then achieve coupling them through appropriate data interface, the calculation platform established. Finally the power distribution and the corresponding parameters temperature distributions in the model have been predicted. On account of deficiencies reflected in calculations, such as the heterogeneous power distribution, fuel assembly geometry has been changed, for instance the proper peripheral moderator wall has been added based on the preceding assembly, then do the coupling calculations and analyze the results. Comparisons between different results have been made, and the expected aim has been reached, which can prove the rationality of assembly modifications and meanwhile prove the usability of the calculation platform. Thus, modified assembly and the calculation platform could be the calculation foundation in future designs of SCWR.

Simulation of a Steam Generator Tube Rupture Accident in a Lead-Cooled Accelerator Driven System

The European Facility for Industrial Transmutation (EFIT) is developed to transmute long-lived actinides from spent fuel on an industrial scale. In this lead-cooled reactor an intermediate loop is eliminated for economic reasons. Within the framework of design and safety studies the impact of a steam generator tube rupture accident has been investigated. In this postulated event high-pressured liquid water blasts into the lead pool which could trigger various transients. As a major concern steam could be dragged into the core featuring a positive void worth. A thermal lead/water interaction could lead to in-core damage propagation; it could initiate a sloshing of the lead coolant and trigger voiding processes. Furthermore the pressurization of the cover gas needs to be considered. To prove the feasibility of the proposed design these risks are investigated and assessed. Numerical simulations are performed using the advanced safety analysis code SIMMER-III [2]. For the important issue of thermal lead/water interactions the SIMMER code has been validated against Japanese heavy-liquid/water injection experiments.

Development of a Standard for Verification and Validation of Software Used to Calculate Nuclear System Thermal Fluids Behavior

With the resurgence of nuclear power and increased interest in advanced nuclear reactors as an option to supply abundant energy without the associated greenhouse gas emissions of the more conventional fossil fuel energy sources, there is a need to establish internationally recognized standards for the verification and validation (V&V) of software used to calculate the thermal-hydraulic behavior of advanced reactor designs for both normal operation and hypothetical accident conditions. To address this need, ASME (American Society of Mechanical Engineers) Standards and Certification has established the V&V 30 Committee, under the jurisdiction of the V&V Standards Committee, to develop a consensus standard for verification and validation of software used for design and analysis of advanced reactor systems. The initial focus of this committee will be on the V&V of system analysis and computational fluid dynamics (CFD) software for nuclear applications. To limit the scope of the effort, the committee will further limit its focus to software to be used in the licensing of High-Temperature Gas-Cooled Reactors. In this framework, the Standard should conform to Nuclear Regulatory Commission (NRC) and other regulatory practices, procedures and methods for licensing of nuclear power plants as embodied in the United States (U.S.) Code of Federal Regulations and other pertinent documents such as Regulatory Guide 1.203, “Transient and Accident Analysis Methods” and NUREG-0800, “NRC Standard Review Plan”. In addition, the Standard should be consistent with applicable sections of ASME NQA-1-2008 “Quality Assurance Requirements for Nuclear Facility Applications (QA)”. This paper describes the general requirements for the proposed V&V 30 Standard, which includes; (a) applicable NRC and other regulatory requirements for defining the operational and accident domain of a nuclear system that must be considered if the system is to be licensed, (b) the corresponding calculation domain of the software that should encompass the nuclear operational and accident domain to be used to study the system behavior for licensing purposes, (c) the definition of the scaled experimental data set required to provide the basis for validating the software, (d) the ensemble of experimental data sets required to populate the validation matrix for the software in question, and (e) the practices and procedures to be used when applying a validation standard. Although this initial effort will focus on software for licensing of High-Temperature Gas-Cooled Reactors, it is anticipated that the practices and procedures developed for this Standard can eventually be extended to other nuclear and non-nuclear applications.

ACR-1000™ Fuel Design Verification

The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.

Quality Assurance in the ITER Construction

The ITER project is basically an engineering and construction project in order to build the ITER machine which is a scientific experimental fusion device. The seven members of the project have all created legal entities called Domestic Agencies to provide in-kind contributions to the ITER Organization (IO) for the supply of components which are manufactured by their suppliers. According to ITER agreement and due to nuclear safety involved in the fusion process, the project requires a license from the French Nuclear Safety Authority. One of nuclear safety regulations is the French Quality Order. The IO has established a Quality Assurance Program for the construction of the ITER machine to meet the requirements of the Order and to ensure that ITER activities are performed to achieve the safety and performance objectives of the ITER machine. The requirements in the program shall be followed by all performers involved in the project not only the IO, but DAs and their suppliers and subcontractors. This paper represents the quality requirements from the Order, and roles and responsibilities between each performer involved in the project. The paper also shows the main characteristics of the ITER Quality Assurance Program ensuring that all activities performed for the project conform to established and documented requirements.

Alternative Laser Driven Fusion Reactions for Nuclear Energy Without Radioactivity

Following the first result of generating nuclear fusion energy without dangerous radioactive radiation by laser ignition of the proton-11Boron reaction (HB11), we applied this method to evaluate other fusion reactions with no primary neutron production as the proton-7Lithium reaction (HLi7) and of the burning of solid density helium isotope 3He (He3-He3). The new method is a combination of now available laser pulses of 10 petawatt (PW) power and duration in the range of picoseconds (ps) or less. The new mechanism follows the initial theory of Chu and of Bobin for side-on ignition of solid state density fusion fuel developed in about 1972 where some later known physics phenomena had to be added. The essential innovation is the use of the discovery of a predicted anomaly when the mentioned laser pulses are sufficiently clean, i.e. free from prepulses by at least a contrast ratio 108 where acceleration by the nonlinear (ponderomotive) force is dominating.

Preliminary Analysis of Three ULOFAs Based on the Dual-Functional Lithium Lead Test Blanket Module Configuration in ITER

Based on the structural design of the Chinese ITER Dual Functional Lithium-Lead Test Blanket Module (DFLL-TBM), Three Unprotected Loss of Flow Accidents (ULOFAs) were investigated preliminarily, assuming that the whole nuclear heat in TBM was carried away by the flowing lithium-lead (LiPb). The results show that the temperature of the first wall (FW) increases rapidly and the maximum temperature appears at the lower part of FW. In the analysis of ULOFAs, the maximum temperature might exceed the melting point of structure material steel. This event must be avoided by the fusion power shutdown system that terminates plasma burn.

Design of ITER TF Magnet Cryostat Feeder Through

International Thermonuclear Experimental Reactor (ITER) TF feeder systems convey the cryogenic supply and electrical power to the TF coils. The Cryostat Feed-through (CFT) includes the straight feeder part from the cryostat wall to the S-Bend Box (SBB). It is the bottleneck of the feeders. The huge Lorentz-force is a challenge for the CFT design. So the reasonable distribution and structural design of the internal and external supports are important. The CFT include the cold (cryogenic) to warm (room temperature) transitions. It is highly integrated with the cryo-pipes, the busbars, the superconductor joints, the thermal radiation shield and the instrumentation pipes and so on. The cryogenic and electrical requirements, the vacuum and mechanical requirements, and so on are considered when the CFT is designed. This paper presents the functional requirements on the TF CFT, gives its structure. The supports are designed and arrayed according to their mechanical or thermal function separately to stand the huge mechanical loads and isolate the conducting heat load from room temperature respectively. The assembly scheme is also described. Mid-joint and cryostat joint are designed to give the facility for the assembly on location. The mechanical analysis result shows the stress in the stainless steel and G10 material both are within the materials stress safety margin. The heat load to the cryogenic pipes and busbars are also less than the requirement 15W. Transient thermal analysis of global feeder model indicates that 32 days are needed for the feeder components to cool down to the required condition.

Mission and Status of the First Two Next Generation Nuclear Plant Fuel Irradiation Experiments in the Advanced Test Reactor

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to nine low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and the irradiations will be completed over the next five to six years to support demonstration and qualification of new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of multiple separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The second experiment (AGR-2) is currently being fabricated and assembled for insertion in the ATR in the early to mid calendar 2010. The design of test trains, the support systems and the fission product monitoring system used to monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the first two experiments will be compared, and updated information on the design and status of AGR-2 is provided. The preliminary irradiation results for the AGR-1 experiment are also presented.