Analysis of the Effect of Beam Divergence Angle on Back-Streaming Electron Region in Ion Source for EAST-NBI

During the process of beam extraction in positive ion source under high voltage region, a large number of electrons are produced in the gaps of grids. After back-streaming acceleration, these electrons go back to arc chamber or impinge grids and heat electron dump or grids, which are harmful for the safety of ion source. Under the situation of poor beam extraction optics, a large part of the primary beam ions bombard the surface of suppressor grid. And this process produces a large number of electrons. Due to the huge extracted voltage, the secondary electron emission coefficient of the suppressor grid surface is also great, when beam ions bombard on it. As a result, the grids’ current grows. The curvature of ion emission surface and equipotential surface nearby are mainly connected to the perveance and plasma grid geometry. In order to optimize the beam performance of high current ion source and increase the mean arc efficiency, the plasma grid of accelerator is already replaced from circular cross section grid to diamond cross section grid. As a result, the shape of ion emission surface is only connected to the perveance. According the measurement of the current of suppressor grid and the calculation of the perveance of the corresponding shoot, we can analyze the effect of beam divergence angle on back-streaming electron. When the beam divergence angle increases, the number of back-streaming electrons increases rapidly, and grids current changes significantly, especially the current of gradient grid and suppressor grid. The results can guide the parameters operating on the ion source for EAST-NBI and find the reasonable operation interval of perveance and the best one to ensure the safety and stable running of the ion source, which has great significance on the development of long pulse, high power ion source.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

Toward an Open-Source Neutronics Code for Circulating-Fuel Reactors

In circulating fuel reactors, such as the Molten Salt Reactor, the fuel circulates throughout the reactor instead of being immobile as in solid fuel reactors. The vast majority of nuclear simulation codes are primarily designed to simulate solid fuel reactors. Hence, many features unique to circulating fuel reactors, such as fuel injection and removal, cannot be properly modeled with these codes. The work presented here focuses on developing a numerical simulation package that can effectively and accurately model these reactors. This package consists of the coupling of the Monte Carlo particle transport code OpenMC with a modified version of ORIGEN-S, and uses a novel algorithm that calculates the optimal fuel injection and removal schemes for such reactors to achieve certain conditions such as a stable reactivity. We demonstrate our code’s accuracy by benchmarking the coupling module with the MCODE coupling code, and by simulating the operation of the ORNL Denatured Molten Salt Reactor using the coupling and fuel injection modules. The resulting fuel injection scheme is in agreement with the original study of that reactor while offering a much finer resolution for the injection scheme over time. This work is part of a broader project to develop an open-source neutronics code for circulating fuel reactors that will couple OpenMC with an in-house open-source depletion module.

Status of MELCOR Sodium Models Development

A sodium coolant accident analysis code is necessary to provide regulators with a means of performing confirmatory analyses for future sodium reactor licensing submissions. MELCOR and CONTAIN, which have been employed by the U.S. Nuclear Regulatory Commission for light water reactor licensing, have been traditionally used for Level 2 and Level 3 probabilistic analyses as well as containment design basis accident analysis. To meet future regulatory needs, new models are being added to the MELCOR code for simulation of sodium reactor designs by integrating the existing models developed for separate effects codes into the MELCOR architecture. Sodium properties and equations of state, such as from the SAS4A code, have previously been implemented into MELCOR to replace the water properties and equation of state. Additional specific sodium-related models to address design basis accidents are now being implemented into MELCOR from CONTAIN-LMR. Although the codes are very different in the code architecture, the feasibility fit is being investigated, and the models for the sodium spray fire and the sodium pool fire have been integrated into MELCOR. A new package called Sodium Chemistry (NAC) has been added to MELCOR to handle all sodium related chemistry models for sodium reactor safety applications. Although MELCOR code requires the ambient condition to be above the freezing point of the coolant (e.g., sodium or water), the high relative freezing point of sodium requires MELCOR to handle situations, particularly far from the primary circuit, where the ambient temperatures are usually at room temperature. Because only a single coolant can be modeled in a problem at a time, any presence of water in the problem would be treated as a trace material, an aerosol, in MELCOR. This paper addresses and describe the integration of the sodium models from CONTAIN-LMR, and the testing of the sodium chemistry models in the NAC package of MELCOR that handles sodium type reactor accidents, using available sodium experiments on spray fire and pool fire. In addition, we describe the anticipated sodium models to be completed in this year, such as the atmospheric chemistry model and sodium-concrete interaction model. Code-to-code comparison between MELCOR and CONTAIN-LMR results, in addition to the experiment code validations, will be demonstrated in this year.

Research and Design of the Sodium Steam Filter System in Large Liquid Metal Cooled Fast Reactors

This reactor uses liquid sodium as coolant owing to its good thermal physical properties, high boiling point and compatibility of cladding material. However, the sodium has a very active chemical properties, for which the free surface of sodium must be protected by inert gas. In the high temperature environment, the sodium atoms diffusion to cover gas slowly, forming a mixed atmosphere that contained large amount of sodium steam. Sodium steam is covered with the free surface of sodium. Then metal sodium will solidify in the inner wall of the pipe or correlative valves with the reduced temperature. This reactor needs to collect and filter sodium steam in order to reduce the hazards to the equipment, piping system, valves and the other devices. Based on the previous research about the purification process of sodium, this paper compared different steam trapping filtration process and carried out the thermal calculation providing basis for research and design of large sodium cooled fast reactor sodium steam trapping filtration process and establishing a reliable sodium steam filtering system.

Investigation of Air Ingress Into a Vacuum Vessel Related to Particle Re-Suspension and Distribution for Dust Issues in ITER

During normal operation of the ITER tokamak, few hundred kilograms of dust containing beryllium (Be) and tungsten (W) will be produced due to the erosion of the walls of the vacuum chamber by the plasma. During a loss of coolant accident (LOCA) or a loss of vacuum accident by air ingress (LOVA), hydrogen could be produced by dust oxidation with steam. Evaluation of the risk of dust and hydrogen explosion, that may lead to a loss of containment, requires studying the physical processes involved in the dust re-suspension and its distribution in the tokamak chamber. This experimental study is conducted by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to simulate dust re-suspension phenomena induced by high velocity jet under low pressure conditions. Tests are conducted in a large scale facility (TOSQAN, 7 m3) able to reproduce primary vacuum conditions (1 mbar). Optical diagnostics such as PIV technique (Particles Image Velocimetry) are implemented on the facility to provide time resolved measurements of the dust re-suspension in terms of phenomenology and velocity. We present in this paper the TOSQAN facility with its configuration for studying dust re-suspension under low pressure conditions and underway experiments showing the mechanism of dust re-suspension by sonic and supersonic flows.

Unprotected Loss of Flow Analysis of a Million Kilowatt Traveling Wave Reactor Core

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the Unprotected loss of flow (ULOF) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the ULOF accident, the core of the million kilowatt TWR demonstrates a good safety performance.

Development and Validation of Evaluation Method on Hypothetical Total Instantaneous Flow Blockage in Sodium-Cooled Fast Reactors and its Application to a Middle Size SFR

Local faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of flow blockages. Although an evaluation on the consequences of HTIB using SAS4A code was also performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in Monju by this developed SAS4A code was performed in this study. Furthermore SAS4A code was newly validated using an in-pile experiment which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the PPS triggered by either of 116% over power or DND trip signals. Therefore the conclusion in the past study that the consequences of HTIB would be much less severe than that of ULOF was strongly supported by this study.

Application Differences of In-Service Inspection Regulations Between RSE-M2010 and ASME Section XI in China PWRs

RSE-M 2010 and ASME Section XI are the widely used and most detailed PWR in-service inspection regulations applied in China PWRs which are separately belong to French AFCEN and American ASME regulations, and come from the different nuclear industry practices of their countries. In 1987, the French M310 type reactor was imported to China and therewith the RSE-M in-service inspection regulation was introduced, beginning to be widely used in China PWRs since that time. Meanwhile, Chinese nuclear power institutes began to independently develop its own PWR reactor named Qinshan Phase I Nuclear Power Plant, and then ASME Section XI in-service inspection regulation was used which was also beginning to be widely used in some Chinese PWRs. With the nuclear power technology development and innovation, such regulations are continually updated and perfected. Thus, there are many differences during application in Chinese specific PWRs. This paper has performed quite deeply application difference analysis between the two regulations based on several aspects, such as upstream laws cited, component classification, inspection requirement, NDE, qualification, pressure test and the Safety Authority review requirements for licensing. Some preliminary thinking has been presented during applying these two regulations and some technical suggestions have been also provided to perfect the regulations in the hope to provide better reference during application on the third generation PWRs (including HPR1000) in China.

Neutronic Analyses for CFETR With Modular Helium Cooled Lithium Ceramic Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak proposed by national integration design group for magnetic confinement fusion reactor of China to bridge the R&D gaps between ITER and DEMO. Since the launch of CFETR conceptual design, a modular helium cooled lithium ceramic blanket concept had been under development by the blanket integration design team of the Institute of Plasma Physics of the Chinese Academy of Sciences, to complete CFETR in demonstrating its fusion energy production ability, tritium self-sufficiency and the remote maintenance strategy. To validate the feasibility, the neutronic analyses for CFETR with this modular helium cooled lithium ceramic blanket were performed. The 1-D neutronic study for CFETR was done in the first place to give a preliminary and quick demonstration of the overall neutronic performance. Meanwhile, the neutronic analyses for a single standard helium cooled lithium ceramic blanket module were done in several times to give more insight for the material and geometry parameters of intra-module structures. Therefore, the principles for neutronic design and the module level optimized parameters were produced, based on which the design of practical blanket modules planted in tokamak vacuum vessel was completed. In the end, the 3-D neutronic analysis for CFETR was done utilizing the MCNP code, in which the 11.25 degree sector model (consist of blanket modules, manifold, support plate, shield, divertor, vacuum vessel, thermal shield and TF coils) was generated with the McCad automated conversion tool from the reference CAD model for analysis, the bi-dimensional (radial and poloidal) neutron source map was plugged via general source definition card to stimulate the D-T fusion neutrons. The concerned neutronics parameters of CFETR, mainly including the tritium breeding ratio to characterize tritium self-sufficiency, the energy multiplication factor to characterize power generation, as well as, the inboard mid-plane radial profiles of neutron flux densities, helium production rate, displacement damage rate and the energy deposition to characterize the shielding performance, were produced. In principle, the neutronics performance of CFETR with modular helium cooled lithium ceramic blanket is promising. The tritium breeding capability meets the design target and, by referring to that for ITER and the EU DEMO fusion power plant, the inboard mid-plane shielding is effective to fulfill the radiation design requirement of the superconducting TF-coils, resulting in a compulsory warm-up time interval of ∼2 FPY for TF-coils. The nuclear heating loads to other CFETR components were generated. As an outcome of this work, the applicability of McCad on CFETR neutronic modeling is demonstrated.