bluuebunny/crossref_metadata_2025_split huggingface dataset:
Stand-Alone Containment Analysis of the Phébus FPT-3 Test With the ASTEC and the MELCOR Codes
Authors: Gonfiotti Bruno , Paci Sandro 
Published: October 2018

The Phébus FP experimental campaign (1988–2010) is recognized as one of the most relevant in the field of Severe Accident researches. During this international research programme, 4 tests were performed to investigate the degradation phenomena for a real PWR fuel bundle in strongly or weakly oxidizing conditions, considering different control rod materials and burn-up levels, plus 1 test studying the phenomena related to a melt pool scenario. These integral tests gave a fundamental improvement in the knowledge of the key phenomena occurring during a Severe Accident and the obtained data were also employed to develop numerical models for the safety analysis of these scenarios.
The ASTEC and the MELCOR integral codes were two of the main international codes that have benefited from these experimental data. After the termination of the Phébus FP campaign, these two codes were furthermore improved, implementing more recent research findings. So, a continuous verification and validation work is necessary to check that the new improvements, implemented in such codes, also allow a better prediction of these past Phébus FP tests.
Therefore, the aim of the present work is to re-analyze the Phébus FPT-3 test employing the up-to-date versions of the ASTEC and MELCOR codes. This FPT-3 test was also the basis for an international benchmark exercise carried out in the frame of the EU SARNET project. The performed analysis focuses only on the stand-alone containment aspects of the test, and a comparison against the results obtained by the different participants to the SARNET benchmark is also performed.
An original analysis on the main differences obtained employing three different spatial nodalizations and a sensitivity analysis on the effects of different input parameters influencing the aerosol behavior are also proposed. These sensitivity analyses also show the need of a sufficient number of volumes for a correct prediction of the coupling between the containment thermal-hydraulics transient and the aerosol behavior. The results obtained show only a partial agreement with the experimental data and with the best calculations performed during the SARNET benchmark, due to user’s effects and code’s limitations as the ones present in the Iodine pool model in MELCOR.
3D Immersive Display Application for Nuclear Education and Public Acceptance
Authors: Luk Bing Lam , Lam Miu-Ling , Chen Ting-Hsuan , Zhao Jiyun , Tsui Suet Man , Chieng Ching-Chang 
Published: October 2018

Immersive Virtual Reality (IVR) systems based on multiple stereoscopic projectors are very popular in many applications, such as training operators for nuclear power plants and surgeons for surgical operations. Due to the increasing number of nuclear reactors in Guangdong province of China, Hong Kong residents are concerned about the nuclear safety and impact on Hong Kong society. There is clearly a strong demand for nuclear literacy education in order to make the public aware of and accept nuclear energy. Thus, City University of Hong Kong has built an IVR system with a 9-meter diameter, 4-meter-height, 235° curved screen for nuclear literacy education. The actual CAD drawings of the Daya Bay nuclear power plant were used to recreate the virtual Daya Bay plant in our IVR system, emphasizing the reactor pressure vessel and steam generators inside the containment building. Visitors can enter the virtual containment building, and experience the actual operation environment in order to understand the basic knowledge of nuclear reactors. At present, the system is not only capable of illustrating the basic knowledge of nuclear reactor physics but also shows the normal and abnormal operations including reactor scram and emergency containment spray.
In order to provide visitors with a full understanding of the role of nuclear power in Hong Kong’s fuel mix, a Low Carbon Energy Education Center (LCEEC) was set up in which the IVR system was the main attraction. Other low carbon energy sources are also introduced in LCEEC. The Centre was visited by thousands of visitors since its opening in April 2017. Surveys have been conducted to collect their comments and suggestions. The results showed that the IVR system is very helpful in raising public understanding of nuclear power.
Integrated Design of a Reactor Core for the Rolls-Royce Small Modular Reactor Project
Authors: de Haas Simon , Chu David , Ellis Kevin , White Matthew , Lindley Ben , Smith Peter , Murgatroyd Julian , Grief Andrew , Leddy Mike , Yule Mike 
Published: October 2018

Rolls-Royce and a UK Consortium are progressing the design and development of a Small Modular Reactor (SMR) Power Station. The SMR programme is a phased design cycle, progressing through the Rolls-Royce gated review process. The project aims to deploy the first of a kind SMR in the UK by the end of the next decade. In this paper, the development methodology for the reactor core design is discussed, along with a selection of the key technical challenges that have been addressed during the concept design phase. Lessons learned from past projects have been identified, to help improve the design efficiency for the SMR.
The concept design has been developed in an iterative fashion, with different analysis disciplines carefully integrated around a common set of objectives. Key economic requirements for an SMR core include maximising fuel economy, cycle length and thermal power while remaining small enough to enable a modular build approach. Top-level safety requirements include control of reactivity, control of core temperature and control of release of radioactivity/radioactive material.
A set of surrogate design limits has been used alongside the true safety limits to avoid the need for detailed transient subchannel or fuel performance analysis in this phase. This has allowed the design to mature and be characterised very quickly, while also maintaining high confidence that all performance and safety requirements will be met when detailed analyses are undertaken.
This paper describes the different analyses that have been undertaken to date, including a variety of reactor physics and thermal hydraulics calculations. The paper discusses the limits used, how they have been used to optimise the design solution and why they provide high confidence in the core design’s performance.
Control Strategy Investigation for a Multi-Purpose Modular Small Pressurized Water Reactor With Once-Through Steam Generators
Authors: Ma Qian , Sun Peiwei 
Published: October 2018

A new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators.
Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements.
To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.
Numerical Simulation of Condensation Heat Transfer in Shell Side of Double Inlet Condenser
Authors: Li Kai-yu , Xue Ruo-jun , Liu Zhao-heng , Sun Ji-lin , Liang Le 
Published: October 2018

The object of this research is a double inlet condenser, which is designed by optimizing and revamping the traditional surface-type condenser.
In this paper, firstly, a mathematical model was established to describe the process of heat transfer in the condenser. Secondly, according to the structure and the operating parameters of double inlet condenser, a full three-dimensional numerical simulation was performed on the double inlet condenser by using Fluent software to obtain the spatial distribution of the important parameters in the shell side of the condenser. Finally, a series of numerical simulations were carried out by control variable method in order to study the effect of various factors, such as inlet water temperature, main and auxiliary steam flow and fouling thickness inside the pipe wall, on the flow and heat transfer in the double inlet condenser. The result is as follows. 1)There are the positive pressure zone and the negative pressure zone in the shell side of the condenser; 2) The pressure shows a mountain-type distribution along the axial direction of the cooling water pipe. 3) There is no obvious whirlpool zone and steam short circuit zone in the double inlet condenser, which shows that the structure design of the condenser is reasonable.
Development of Hydrogen Treatment System in Severe Accident: Part 5 — Effect of Steam Flow on Performance of a Hydrogen Treatment Unit With Metal Oxides
Authors: Iwaki Chikako , Sugita Tsukasa , Yamada Akira , Yagyu Motoshige , Haruguchi Yoshiko , Tanabe Masashi 
Published: October 2018

A large amount of hydrogen is generated by the metal water-reaction in the Primary Containment Vessel (PCV) of Light Water Reactors in the severe accident. Then we have been developing the Hydrogen Treatment System to prevent excessive pressure without PCV vent. By focusing on the oxidation-reduction reaction of metal oxides with high reaction rate, we have been studying hydrogen treatment system using metal oxides as effective device under oxygen deficit conditions of PCV of Boiling Water Reactor (BWR).
We have been considering a hydrogen treatment unit with a lot of pipes in which metal oxides are filled. We have already investigated experimentally the basic trend of metal oxides temperature and gas concentration as well as hydrogen processing rate dependency on gas temperature and concentration in the absence of steam. However, the influence of steam on hydrogen processing characteristics has not been clarified yet.
The objective of the present research is to investigate how the steam, which is supposed to be a reaction-inhibiting-factor, affects hydrogen processing characteristics of the hydrogen treatment unit. We conducted experiments using a test section of one pipe simulating a part of the hydrogen treatment pipes. The granulated CuO, which is a candidate material for the actual system was used. The hydrogen concentration of 10 wt% at the inlet of the pipes was decreased to 0 wt% at the outlet even in high steam concentration conditions (30–50 wt%) when the gas temperature was 250 °C, therefore, it was confirmed that hydrogen was treated with high processing rate under steam circumstances. It was also found that cumulative amount of treated hydrogen was strongly correlated with temperature and relative humidity.
We have been developing the thermal-chemical model of hydrogen treatment unit. The prediction margin of error was decreased to 30 % from over 100 % by improving degradation model based on the experiment results, therefore, it reached the practical level.
Physics Design of Special Epithermal Neutron Beam Based on Multi D-D Reaction Neutron Tubes
Authors: Yang Yi , Li Yiguo , Zhao Bin 
Published: October 2018

Accelerator used for Boron Neutron Capture Therapy (BNCT) is the development trend of cancer treatment in the future. Neutron generator is a neutron source with compact structure, easy to operate, and lower price in accelerator (the structure of neutron tube is more compact). It has a high feasibility of establishing in the hospital when compared with other types of accelerator. At present, the D-D reaction has higher neutron yield than other reactions, more easy to get D material. Therefore, the epithermal neutron beam based on D-D neutron generator is studied for BNCT usage.
First, the calculation model is established by MCNP program, including the neutron source model, the geometric model of the irradiation device. The specific neutron energy spectrum and angular distribution of the D-D reaction are theoretically analyzed when establishing the D-D reaction neutron source model. The basic structure models of typical irradiation devices are designed, and then the optimal suitable moderator material for the irradiation device is studied from the neutron reaction cross section, the combination of iron and Fluental materials is optimum.
The irradiation device with multi D-D neutron tubes’ combination as BNCT neutron source is designed. It can be concluded by study that the parameters at the beam exit are epithermal neutron flux density 2.01 × 108 n/cm2 · s, fast neutron contamination 1.33 × 10−11 Gy · cm2/n, γ contamination 5.79 × 10−17 Gy · cm2/n, for the combination with 6 neutron tubes. The result can meet IAEA’s requirements for BNCT epithermal neutron beam quality.
Development of a Device for Detecting Helium Leaks From Canisters: Part 2 — Numerical Analysis of Temperature Behavior During Gas Leaks From a Canister of a 1/4.5 Scale Cask Model
Authors: Shimizu Kosuke , Takeda Hirofumi , Goto Masanori 
Published: October 2018

In the concrete cask, the canister is sealed with lids by welding, and has high sealing performance. But considering long-term storage, there is a concern about loss of the sealing performance due to stress corrosion cracking (SCC). In the concrete cask, unlike the metal cask, it is not mandatory to constantly monitor helium pressure between the lids. However, it is useful from the viewpoint of improving safety during the long-term storage to install a helium leak detector in the canister inside the concrete cask. Currently, we are developing the leak detector utilizing the phenomenon that the surface temperature of the canister changes when helium leaks out of the canister.
As part of developing the leak detector of the canister, leak tests were performed using a small canister model as a pressurized vessel and a 1/4.5 scale cask model of the actual cask including the canister. This leak detector utilized the phenomenon that canister bottom temperature (TB) increases and canister lid temperature (TT) decreases when the internal pressure of the canister decreases. In computational fluid dynamics (CFD) calculation, focused on this phenomenon, the influence of the internal pressure and physical properties of internal gas in the canister were examined by calculating conditions of three kinds of pressure and two types of gas (air and helium). The main purpose of the CFD calculation was to confirm the results of the experiment, and we grasped the phenomenon occurring in the canister and elucidated its mechanism.
For the CFD calculation, a commercial CFD software, STAR-CCM+® (ver.12.06.010) by Siemens PLM Software Company, was used. A CAD file used for the calculation simulated also the shape inside the canister (e.g. basket, fuel rods). A polyhedral mesh was used for a calculation mesh. In the small canister model, a mesh of its ambient air was not generated, and heat transfer between the canister surface and the ambient air was calculated from a heat transfer correlation equation. On the other hand, in the 1 / 4.5 scale cask model, the mesh of its ambient air was generated, so that the heat transfer on the surface of the canister was calculated according to the actual heat transfer phenomenon. The internal gas and the ambient air of the canister were ideal gas, and buoyancy due to density change was taken into consideration. A realizable k-epsilon model was used for a turbulence model, and a DO model was used for a radiation model.
The Neutron Active Interrogation System for In-Field Detection of Transuranic-Based Radioactive Dispersal Devices for Security Applications
Authors: Cherubini N. , Dodaro A. , Gandolfo G. , Lepore L. , Marzo G. A. , Piccinelli E. , Remetti R. 
Published: October 2018

The increasing of terror menace in recent years led the international community to enhance the efforts to minimize threats to people in everyday life by developing devices, techniques, and procedures targeted to improve the collective security. In this framework the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) developed a new device to improve CBRNe resilience, the Neutron Active Interrogation system (NAI). It has been conceived and optimized to identify transuranic-based Radioactive Dispersal Devices potentially hidden in packages, envisaging its utilization in field applications. NAI is based on the detection of neutrons from induced fission on small amount, of the order of a few grams, of fissile material. The device exploits a portable neutron generator based on d-t fusion reaction, a polyethylene structure for reducing the neutron energy in order to maximize the fission cross-section, and an array of 3He proportional counters. Fissile material detection is made using the Differential Die-Away time Analysis (DDAA), an active neutron technique based on the difference among the die-away times of fast interrogation neutrons and prompt fission neutrons induced by thermal neutrons in the moderating system.
The original experimental setup was tested on the field during the live demo open to the public at the EDEN Project Demonstration occurred in September 2015 at ENEA Frascati Research Centre in Rome. Since then, the setup has been modified to improve the device detection capabilities. NAI performances have been tested within different environmental conditions, e.g. open field geometry vs. bunker-like geometry, to study the effects of scattering phenomena. The optimized configuration here presented is transportable, lightweight, and able to detect 2 grams of 235U contained in a salt of depleted uranium in real time, independently from the measurement environmental conditions.
A Validation of RELAP on Predicting Nuclear Power Plant Phenomena
Authors: Ahn Ji Soo , Bluck Michael , Eaton Matthew , Jackson Chris 
Published: October 2018

In this study, RELAP5’s capability to simulate thermal stratification under different conditions is assessed. In nuclear power plants (NPPs), thermal stratification can occur in the following locations: pressurizer, piping systems such as hot legs, cold legs, surge lines, and cooling tanks if available.
In general, thermal stratification in a horizontal pipe could not be simulated by RELAP5 due to the inherent one-dimensional setting. Moreover, RELAP5 failed to simulate turbulent penetration which was often a pre-requisite prior to thermal stratification in a pipe. This type of situation could arise in connection between hot leg and surge line, spray lines, feed water lines, etc. It is recommended that for this type of problem CFD be used.
In the literature, it was found that RELAP5 was capable of simulating thermal stratification in a pool or a tank-like component if multiple channels and crossflow junctions were used. However, due to uncertainties associated with the input model, the current RELAP5 model failed to reproduce experimental data and therefore further investigation would be required to identify the sources of error.
Main Evolutions of the RCC-C Design and Construction Code for Fuel Assemblies Since 2015
Authors: Ton-That Marc , Vauglin Christine , Trillon Gilbert 
Published: October 2018

AFCEN is a French Standard Development Organization which publishes codes for design, construction and in-service inspection rules for Pressurized Water Reactors. The fields covered by theses codes are mechanical components, in-service surveillance of mechanical components, electrical equipments, nuclear fuel, civil works and fire protection. AFCEN was initially founded by electric utility EDF and nuclear steam supply system manufacturer FRAMATOME. AFCEN has more than 60 institutional members, representing more than 650 experts who contribute to the development and continuous improvement of codes.
The RCC-C code, which is dedicated to PWR fuel assemblies and associated core components, set forth generic requirements to be fulfilled by the suppliers and by the manufacturers for the design justifications and for the manufacturing and inspection operations of PWR fuel assemblies and rod cluster control assemblies. The RCC-C is intended to be used in the frame of contractual relations between a customer (nuclear operator) and a nuclear fuel supplier.
The first edition was published in 1981. Over the years, many changes have been made to the original text but the structure hasn’t been much modified. Because of this, the text was becoming less coherent for the users and was lacking also minimal explanations. A redesign of the code was scheduled for the 2015 edition to address those problems. With the involvement of fuel vendors FRAMATOME, WESTINGHOUSE, and French nuclear operator EDF, the text was restructured and clarified. New requirements were implemented and the set of both design and manufacturing rules was strengthened to reflect fuel vendors’ practices and operator expectations.
This article explains the main modifications that were implemented since the 2015 edition, and also outlines the prospects for future changes taking into account the latest regulatory requirements and evolutions of the industrial practices.