bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Study on Hydrogen Flow Behavior for Different Connecting Pipe Structures
Authors: Liu Hanchen , Yuan Kai , Cao Xuewu 
Published: October 2017

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of nuclear containment building. Therefore, getting detailed hydrogen flow and distribution is a key issue to resolve the hydrogen risk in containment and compartments. In this study, local hydrogen behavior has been investigated in a multi-subcompartments using Computational Fluid Dynamics method, concerning the local concentration in the multi-subcompartments. The multi-subcompartments containment is represented by four vessels interconnected by pipe. Hydrogen inlet pipe is located in source vessel. The analysis model is built by 3-dimensional Computational Fluid Dynamics code ANSYS-CFX in Cartesian coordinates. Computational domain is discretized in about 47559 cells, 87484 cells and 126388 cells, respectively. Taking a full consideration of the computational time and accurate, the medium mesh scheme is adopted to simulate the hydrogen transport.
With standard k-ε turbulence model, the effects of the connecting pipe parameters and direction on hydrogen distribution in the multi-compartments are investigated. Local hydrogen behavior showed remarkably different in simulations with the change of the pipe parameter. Increasing the connection pipe diameter and decreasing the connection pipe length are helpful for helium flow between compartments. Compare with the vertical connection pipe, horizontal connection pipe is not conducive to the diffusion of hydrogen from source compartment to non-source compartment.
Modeling the Ambient Condition Effects of an Air-Cooled Natural Circulation System
Authors: Hu Rui , Lisowski Darius D. , Bucknor Matthew , Kraus Adam R. , Lv Qiuping 
Published: October 2017

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performance of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system.
Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STAR-CCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.
Experimental Investigation in LIFUS5/Mod2 Facility of Spiral-Tube Steam Generator Rupture Scenarios for ELFR
Authors: Pesetti Alessio , Del Nevo Alessandro , Neri Andrea , Cati Stefano , Sermenghi Valerio , Valdiserri Massimo , Giannotti Daniel , Tarantino Mariano , Forgione Nicola 
Published: October 2017

In the framework of the European Commission LEADER project, an experimental campaign of seven tests was performed in the LIFUS5/Mod2 facility, at ENEA CR Brasimone, for investigating the postulated Steam Generator Tube Rupture (SGTR) event in a relevant configuration for the Spiral-Tube Steam Generator (STSG) of the European Lead Fast Reactor (ELFR).
The LIFUS5/Mod2 facility is composed by a water tank of 15 L injecting subcooled water up to 200 bar into the reaction tank of 100 L (420 mm of diameter), which is connected by a 3 inch pipe to the dump tank of 2 m3. A dedicated test section was designed, assembled and implemented in the reaction tank. It is composed by 188 tubes, vertically disposed with triangular pitch inside a cylindrical support. This tube bundle is representative of a portion of the STSG of ELFR. The cylindrical support is closed at the lower and upper end by two tube plates and has a perforated lateral shell (300 mm of diameter and 400 mm high). The reaction tank is filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C up to the top tube plate, with an argon cover gas at about 2 bar. The water is injected at about 180 bar and 270°C through the central tube, at middle height of the bundle. The water-LBE interaction is characterised by high quality data acquisition system: 6 fast Pressure Transducers (PTs) working at 10 kHz for precisely characterize the first narrow injection peaks, 70 low constant time Thermocouples (TCs) to understand the vapour evolution path and 13 strain gages (SGGs) for measuring the strain of the bundle and main vessel.
The overall LEADER experimental campaign is constituted by seven tests, divided in three series (B1, B2 and B3), characterized by different injection orifice diameters of 4, 8.9 and 12.6 mm, respectively. This paper presents the experimental results of the first two tests of series B2 (B2.1 and B2.2) having 8.9 mm of injection orifice.
The first test analysed showed a first narrow pressure peak of about 32 bar, some milliseconds after the cap rupture instant. The following pressurization due to the evaporation of water entered into the reaction vessel was of an analogues magnitude for both the tests (about 50 bar) and lasted some tenths of second. The water/LBE interaction lower temperature was reached on the inner ranks of tubes, about 150°C. The outer rank was cooled down to about 300°C. The strain gage measurements showed a decreasing deformation on the tubes toward the outer positions. No ruptures were observed on tubes surrounding the injector. The amount of LBE transported into the dump tank was strongly dependent on the LBE level in the reaction tank at the start of the tests and about 200 kg.
Methodology to Investigate Vibration Phenomena Caused by the Steam Condensation at Subatmospheric Condition
Authors: Lo Frano R. , Mazed D. , Aquaro D. , Del Serra D. , Sekachev I. , Giambartolomei G. 
Published: October 2017

The phenomenon of steam-water direct contact condensation (DCC) was widely investigated in the past because of conventional and nuclear engineering applications, like in the safety suppression system of BWRs.
When steam is injected into a sub cooled water, different condensation regimes can be observed, such as chugging, condensation oscillation, bubbling condensation oscillation etc. (these have been identified during the experimental campaign on condensation of steam injected into water at sub-atmospheric conditions carried out at Lab. B. Guerrini of DICI - University of Pisa). According to the thermal hydraulic conditions of the condensing jet plume, vibrations may arise.
This paper so deals with the analysis of vibration phenomena that originate during the steam-water direct contact condensation at sub-atmospheric conditions, not treated till today. Numerical investigations (by proper FEM code) as well as experimental activity will be presented and discussed in order to evaluate if vibrations are capable to jeopardise this safety suppression system, which is designed to protect Vacuum Vessel of ITER from pressurizing accidents.
Development of Numerical Simulation Method to Evaluate Heat Transfer Performance of Air Around Fuel Debris: Part 2 — Validation of JUPITER for Free Convection Heat Transfer
Authors: Uesawa Shinichiro , Yamashita Susumu , Shibata Mitsuhiko , Yoshida Hiroyuki 
Published: October 2017

A dry method is one of fuel debris retrieval methods for decommissioning of TEPCO’s Fukushima Daiichi nuclear power station. If stopping the water leakage from a damaged primary containment vessel (PCV) is impossible, the dry method may be adopted. Besides, the dry method may be also adopted to retrieve the fuel debris in a pedestal under a reactor pressure vessel (RPV) because accessing and retrieval of the fuel debris from the lateral sides of PCV may be adopted for retrieval of the fuel debris in the pedestal. However, the decay heat must be removed from the fuel debris without water because fuel debris is exposed to the air in the dry method. Thus, we must evaluate the air cooling performance for the fuel debris in the dry method.
Japan Atomic Energy Agency (JAEA) has started the research project to develop an evaluation method by the JUPITER to understand the air cooling performance for fuel debris in the dry method. In the project, free convection heat transfer and the radiation heat transfer experiments for air are also conducted for validation of the JUPITER. In this paper, the free convection heat transfer, the radiation heat transfer, the conduction heat transfer and the Rayleigh number for the free convection in the simplified model were estimated to decide experimental conditions. We confirmed the free convection and the radiation heat transfer were important to estimate heat transfer of the debris exposed to the air in the RPV pedestal. Considering the estimations, the experiment for heat transfer and flow visualization of free convection adjacent to upward-facing horizontal surface was conducted to validate the calculation of the turbulent free convection by the JUPITER. From comparing between the calculation result and the experimental result, we confirmed that the calculation result of the JUPITER was qualitatively reasonable.
Challenges of Evolution for the Risk-Informed Decision Making Standard
Authors: Narumiya Yoshiyuki , Hayashi Kentaro , Kuramoto Takahiro 
Published: October 2017

Risk-informed applications for the purpose of optimized decision making in safety related activities, including safety design and operational management of nuclear power plants, can be widely adopted by regulators, manufacturers as well as utilities. Utilization of risk information is expected to contribute to the achievement of safety improvement in operating nuclear power plants in an optimized way. In addition, effective utilization of risk information is essential to cope with the changes in safety related activities reflecting the latest experiences and findings for the continuous improvement of the safety of nuclear power plants into the future.
In this regard, the Standards Committee of the Atomic Energy Society of Japan is working on the establishment of the new standard specifying the basic requirements regarding risk-informed decision making (hereafter, called “RIDM”). This paper introduces the policies toward the establishment of the RIDM Standard.
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
Authors: Fukano Yoshitaka 
Published: October 2017

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.
Development of Evaluation Method for Cold Trap in Fast Breeder Reactor: Part 2 — Analytical Evaluation of Impurity Capturing Capacity in Cold Trap
Authors: Takahashi Shiro , Tamura Akinori , Nakata Hiroyuki , Takota Akio 
Published: October 2017

Because of the reactor core characteristics of fast breeder reactors (FBRs), liquid sodium must be used for the reactor core cooling. Small amounts of oxygen and hydrogen are dissolved in the FBR sodium coolant. To decrease those impurity concentrations, cold traps (CTs) are installed in the primary and secondary systems. CTs are a purification apparatus universally used in FBR sodium systems to maintain the oxygen and hydrogen concentration levels in sodium within acceptable limits. CTs can capture oxygen and hydrogen as sodium compounds by packing materials with stainless steel wires using precipitation phenomenon (crystallization) by decreasing the inlet sodium temperature below the saturation temperature. Oxygen and hydrogen are precipitated as Na2O and NaH.
We have been developing an evaluation method of the impurity capturing capacity in CTs. In this study, we investigated the impurity capturing capacity of CTs through three-dimensional computational fluid dynamics (CFD) analysis. We also used CFD analysis to evaluate the time variation of the pressure loss of packing material. The amount of precipitation was decided from the temperature, concentration and velocity distributions through the CFD analysis.
First, a uniform wire mesh was used for the capturing capacity analysis in the CTs. Precipitation speed was higher for the hydrogen sodium compound than for the oxygen sodium compound. As sodium flowed from the outer to the inner side of the packing material, the hydrogen sodium compound was precipitated near the outer surface of the CTs and that blocked the sodium flow at the outer surface. As a result, if a lot of the hydrogen flowed into the CTs, the inner side of the wire mesh could not be used effectively. Second, we also proposed non-uniform wire meshed shapes and analyzed their capturing capacity in the CTs. The shape of the wire mesh was changed in the radial direction. A coarse meshed wire was placed on the outer side of the packing material and a closely packed meshed wire was placed on the inner side. Our CFD results showed that the most effective impurity capturing capacity of hydrogen sodium compound could be obtained by changing the precipitation speed and wire shape in the radial direction.
The Effect of MA Transmutation in the PWR on Fuel Cycle
Authors: Yu Haoyang , Liu Bin , Zhang Wenxin , Cai Jin 
Published: October 2017

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core.
In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor.
In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.
Development of Uncertainty Quantification Capability for NESTLE
Authors: Huang Dongli , Abdel-Khalik Hany S. 
Published: October 2017

This work aims to develop an uncertainty analysis methodology for the propagation and quantification of the effects of nuclear cross-section uncertainties on important core-wide attributes, such as power distribution and core critical eigenvalue. Given the computationally taxing nature of this endeavor, our goal is to develop a methodology capable of preserving the accuracy of brute force sampling techniques for uncertainty quantification while realizing the efficiency of deterministic techniques. To achieve that, a reduced order modeling (ROM) approach is proposed to deal with the enormous size of the uncertainty space, comprising all the cross-section few-group parameters required in core-wide simulation. The idea is to generate a compressed representation of the uncertainty space, as represented by a covariance matrix, that renders sampling techniques computationally a feasible option for quantifying and prioritizing the various sources of uncertainties.
While the proposed developments are general to any reactor physics computational sequence, we customize our approach to the NESTLE [1]-TRITON [2] computational sequence, which will serve as a demonstrative tool for the implementation of our approach. NESTLE is a software used for core wide simulation, which relies on the few-group cross-sections to calculate core wide attributes over multiple cycles of depletion. Its input cross-sections are generated using a matrix of conditions evaluated using a lattice physics code, which in our implementation is done using the TRITON software of the ORNL’ SCALE suit. This manuscript presents one of the early steps towards this goal. Specifically, we focus here on the development of the algorithms for determining the reduced dimension of covariance matrix. Numerical experiment using the TRITON software is employed to demonstrate how the reduction is achieved.
Basic Experimental Study on Effectiveness of Nuclear Waste Long-Term Storage Containers With PAR for Reducing Concentration of Hydrogen Gas: Part 1 — Research Plan and Preliminary Simulation Results
Authors: Tanaka Y. , Suzuki Y. , Takase K. 
Published: October 2017

In the decommissioning of nuclear power plants, the long-term management of radioactive waste of fuel debris, etc. are necessary. In the process, hydrogen which is the flammable gas is generated by the decomposition of water by radiation. Therefore, it is important to ensure the safety of the waste storage container to reduce the concentration of hydrogen gas, and to keep below the explosion limit (4%). Consequently, a basic experiment to investigate the effectiveness of the waste storage container with the flammable gas concentration reduction mechanism using the passive autocatalytic recombiner (PAR) has been planned. The present study describes the research plan to use a small modeled experimental apparatus.
In addition, in order to clarify quantitatively natural convection behavior of hydrogen gas due to the decay heat of radioactive materials in the long-term waste storage container, preliminary analyses were performed on the system of a small-scale experimental apparatus in which specification of the long-term radioactive waste storage container is simply simulated. From the present results, the perspective which can predict natural convection phenomena in the long-term waste storage container numerically was obtained.
Uncertainty Analysis of the TRU-Burning Thorium-Fueled RBWR Using Generalized Perturbation Theory
Authors: Bogetic Sandra , Gorman Phillip , Aufiero Manuele , Fratoni Massimiliano , Greenspan Ehud , Vujic Jasmina 
Published: October 2017

The RBWR-TR is a thorium-based reduced moderation BWR (RBWR) with a high transuranic (TRU) consumption rate. It is charged with LWR TRU and thorium, and it recycles all actinides an unlimited number of times while discharging only fission products and trace amounts of actinides through reprocessing losses. This design is a variant of the Hitachi RBWR-TB2, which arranges its fuel in a hexagonal lattice, axially segregates seed and blanket regions, and fits within an existing ABWR pressure vessel. The RBWR-TR eliminates the internal axial blanket, eliminates absorbers from the upper reflector, and uses thorium rather than depleted uranium as the fertile makeup fuel. This design has been previously shown to perform comparably to the RBWR-TB2 in terms of TRU consumption rate and burnup, while providing significantly larger margin against critical heat flux.
This study examines the uncertainty in key neutronics parameters due to nuclear data uncertainty. As most of the fissions are induced by epithermal neutrons and since the reactor uses higher actinides as well as thorium and 233U, the cross sections have significantly more uncertainty than in typical LWRs. The sensitivity of the multiplication factor (keff) to the cross sections of many actinides is quantified using a modified version of Serpent 2.1.19 [1]. Serpent [2] is a Monte Carlo code which uses delta tracking to speed up the simulation of reactors; in this modified version, cross sections are artificially inflated to sample more collision, and collisions are rejected to preserve a “fair game.” The impact of these rejected collisions is then propagated to the multiplication factor using generalized perturbation theory [3]. Covariance matrices are retrieved for the ENDF/B-VII.1 library [4], and used to collapse the sensitivity vectors to an uncertainty on the multiplication factor. The simulation is repeated for several reactor configurations (for example, with a reduced flow rate, and with control rods inserted), and the difference in keff sensitivity is used to assess the uncertainty associated with the change (the uncertainty in the void feedback and the control rod worth). The uncertainty in the RBWR-TR is found to be dominated by the epithermal fission cross section for 233U in reference conditions, although when the spectrum hardens, the uncertainty in fast capture cross sections of 232Th becomes dominant.