bluuebunny/crossref_metadata_2025 huggingface dataset:
Stokes Number Analysis of the Moving Droplets in the Steam-Water Separator
Authors: Zhao Fulong , Liu Qianfeng , Zhao Chenru , Bo Hanliang , Liu Ying , Zhou Ya
Published: October 2018

The steam-water separator is vitally important equipment to remove the droplets entrained by the vapor stream to provide dry saturated steam for the steam turbine in the nuclear power station. With the development of the large nuclear power station and the vessel nuclear power plant, the steam-water separation performance should be more efficient under the condition of higher pressure, power load and circulating ratio. The droplet motion model, which is solved by typical four steps Runge-Kutta method and validated against the experimental results, is developed according to the physical phenomenon description and the mechanism comprehension of the vapor entrained droplets moving in the wave-type vanes separator. The Euler-Lagrange methodology is adopted to simulate the moving droplet entrained by the vapor stream in the wave-type vanes separator and the separation performance is investigated. The separation efficiency of the separator and motion trajectories of droplets with various sizes are presented. Stokes Number (St) of diverse droplets is obtained to analyze the influence of Stokes Number on the moving droplets trajectories and the separation efficiency. The results reveal that the values of Stokes Number for most of the moving droplets in the wave-type vanes separator are beyond 1, which indicates that most of droplets are likely to collide with the solid wall of the separator. Only when the droplet velocity is smaller than 1 m/s or the droplet radius is less than 2 μm, the Stokes Number may be below 1 and the moving droplets can be entrained by the stream flow until escaping from the separator. The analysis can forecast the maximum critical separation size of the droplets that cannot be removed, and the minimum critical separation size of the droplets that can be removed throughly by the separator and guide the design of the separator.
Experimental and Numerical Analysis of the Flow Field in the Integrated Valve for the Control Rod Hydraulic Drive System
Authors: Junfei Jiang , Benke Qin , Hanliang Bo
Published: October 2018

Control Rod Hydraulic Drive System (CRHDS) is a new type of built-in control rod drive technology, and the Integrated Valve (IV) is the key control component of it. The pulse water flowing into the control rod hydraulic mechanism (CRHDM) is controlled by the IV to drive the hydraulic cylinders to move in a predefined sequence to make the control rod perform step-up, step-down and scram functions. Flow resistance of the IV flow channels is the key design parameter of IV which influences the step motion of the hydraulic cylinder and thus affects the performance of the CRHDS. Experiments on the flow resistance of IV flow channels at different working temperatures were conducted to obtain differential pressures of IV under various temperature and flow rate operating conditions. Based on the experimental conditions and results, three dimensional flow field analysis of the IV flow channels was carried out to get the flow field distribution and hydraulic parameters of the IV flow channels. Flow resistance of the IV flow channels at different working temperatures were obtained using the calculation results and agree well with the experimental results. It verified the correctness of the CFD model. On the basis of the numerical simulation results, the velocity and pressure distribution schemes in the IV flow channels under different working temperature conditions were compared and analyzed. The research results show that the flow resistance of the IV in-rod flow channels remains largely unchanged at different working temperatures, the peak flow velocity appears at the entrance of the valve core section which is also the main flow resistance loss area. The theoretical model was then applied to analyze the influence of the design parameters which include the valve core size, the angle between flow channels, etc., on the total flow resistance of IV at high temperatures. And the analysis results show that, the angle between flow channels has little influence on the flow resistance coefficient. The increase of valve core radius can significantly reduce the total flow resistance of IV flow channels. Numerical simulation on one out-rod flow channel is also carried out, which shows that the flow resistance in out-rod flow channel is much lower than the corresponding in-rod flow channels. The research results can give guidance for the design and optimization of the IV.
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.
Isogeometric Multi-Level Iterative Solution Algorithms With Applications in Nuclear Reactor Physics
Authors: Latimer C. , Kópházi J. , Eaton M. D.
Published: October 2018

The efficient solution of the neutron diffusion equation for large scale whole core calculations is of paramount importance; especially if the detailed pin-level power distribution and reaction rates are required. For heterogeneous whole core calculations finite element based techniques have been one approach to modelling the detailed pin geometry in whole core calculations. A new approach, pioneered in the last few years, is isogeometric analysis (IGA) methods which enable the exact geometry of fuel pins to be modelled. In order to efficiently solve elliptic partial differential equations (PDEs), such as the neutron diffusion equation, typically multi-grid or multi-level iterative solution techniques are used such as the algebraic multi-grid method. However, using IGA methods it is possible to develop true geometric multi-grid techniques which are potentially much more efficient than the standard algebraic multi-grid methods. In this paper we explore the use of IGA methods to develop a scalable, multi-level, iterative algorithm which is then used to solve the neutron diffusion equation over several geometries. This multilevel solution algorithm utilises a single patch multi-grid framework suggested by Hofreither and Takacs that takes advantage of the tensor product construction to provide scalability with respect to spatial, and polynomial refinement. Furthermore, a two-level balancing Neumann-Neumann solver is used to extend the solver to multiple patches in a scalable way. It is seen that the number of iterations required depends on the mapping between the unit square and the physical geometry, as well as the material coefficients.
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.
Neutronic Study on a New Concept of Accelerator Driven Subcritical System in China
Authors: Li Jinyang , Gu Long , Yao Cunfeng , Wang Dawei , Peng Tianji , Zhu Yanlei
Published: October 2018

Accelerator driven subcritical system (ADS) has been considered as an advanced nuclear waste transmutation facility with inherent safety feature. Nowadays, Institute of Modern Physics, Chinese Academy of Science (IMPCAS) has made a plan to research and develop ADS, which project was formally approved by the National Development and Reform Commission at the end of 2015. In order to pursue high transmutation rate of Minor Actinide (MA), a new concept design of the lead-bismuth cooled ADS is proposed in present work, in which a gravity-driven dense granular flow spallation target has been designed to substitute the traditional solid or liquid heavy metal spallation target. Sensitive parameters of the spallation target have been investigated, such as the radius and axial position of the spallation target, the size of heavy metal grains, the granular filling rate, the spallation neutron yield, and the distribution of neutron source. Furthermore, an optimized design of sub-critical reactor has been proposed with three zone loading pattern as a consequence of eliminating the power peak factor, which thermal power is 10MW with the operation cycle 600 EFPD. Finally, some important physical parameters in the coupling depletion system have been studied, including the effective multiplication factor k-eff, subcritical multiplication factor k-s, the change of beam current, time-dependent void and Doppler effect, fraction of effective delayed neutron, and the transmutation supporting ratio.
Structural Redundancy Design and Reliability Analysis of Magnetic Bearing for HTGR Primary Helium Circulator
Authors: Zheng Yangbo , Shi Zhengang , Liu Xingnan , Mo Ni , Yang Guojun
Published: October 2018

As primary pump of High Temperature Gas-cooled Rector (HTGR), the safety and reliability of primary helium circulator has a direct relationship with heat export in the reactor core. This paper proposed two conceptual design methods, which are structural redundancy design in the axial and radial magnetic bearings (coils), and crossing redundancy design between amplifiers and the redundant bearings (coils), combined with the characteristics of magnetic bearing in the primary helium circulator. Then we built Markov model for the magnetic bearing with redundant structure, which was used to reliability analysis and calculation. Finally, we performed simulation of the redundant magnetic bearings in Matlab/Simulink. The results show that the redundant structure design can greatly improve the reliability of the magnetic bearing system. Therefore, these methods would provide a new approach for further improving the safety and reliability of primary helium circulator.
Grounded Control Rod Position Measurement With Two-Electrode Capacitance Sensor
Authors: Hu Guang , Qin Benke , Bo Hanliang
Published: October 2018

In the work, we employed a new method to monitor the behavior of the grounded control rod. To this end, a sensor system based on the two-electrode capacitance is applied to measure the rod position in nuclear heating reactor. Additionally, the finite element method was used to analyze the capacitance of the sensors composed of two-electrode. According to the different electric properties of grounded control rod and air, this apparatus determines the control rod position using an interface electronics method and two -electrode capacitance sensor. The capacitance data obtained by simulations were compared with experiment results to verify the measurement accuracy. The subsequent quantitative analyses on the data indicate the reliability and accuracy of the apparatus for monitoring rod position. It is demonstrated that such a system shows very promising applications for measurements of the control rod position of nuclear reactor.
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.
Finite Element Analysis of AMB Eddy-Current Loss in HTR-PM Primary Helium Circulator
Authors: Yu Jinpeng , Zhou Yan , Ni Mo , Yang Guojun , Zhao Lei
Published: October 2018

In the active magnetic bearing (AMB) system, the eddy current is generated during the rotation of the rotor, which brings about the AMB loss and eddy-current magnetic field. The eddy-current magnetic field will reduce the electromagnetic force and generate the resistance torque of the AMB. Basing on the AMB of the Primary Helium Circulator (PHC) in HTR-PM, two-dimensional (2D) and three-dimensional (3D) AMB models are built and analyzed with finite element analysis (FEA) in maxwell. The 2D FEA shows that the eddy-current loss and the resistance torque increase as the rotor speed increases, and the eddy-current magnetic field will affect the air-gap magnetic field and reduce the electromagnetic force. The 3D FEA shows that dividing the rotor in insulate sheets can reduce the eddy-current loss. The loss and the sheets thickness have a linear relationship, which is different from the theoretical analysis.
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.
Transient Heat Transfer for Helium Gas at Various Flow Decay Time Constants and Heat Generation Rates
Authors: Liu Qiusheng , Kitano Ayumi , Fukuda Katsuya , Shibahara Makoto
Published: October 2018

Knowledge of the heat transfer phenomenon under flow decay transient condition is important for the safety assessment of a very high temperature reactor (VHTR) during a loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally investigated. The experiment was conducted by using a forced convection heat transfer experimental apparatus. A flow control value with its control system was used to realize a flow decay condition. Helium gas was used as a coolant, and a platinum cylinder with a diameter of 1 mm was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Subsequently, the flow rate of the helium gas began to exponentially decrease with different time constants ranging from 3 s to 15 s. The initial flow velocity ranged from 7 m/s to 10 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under a wide range of experimental conditions such as heat generation rates and flow decay time constants. The results indicated that the temperature of the test heater exhibits a rapid increase during this process, and the increasing rate of the temperature is higher for a lower time constant. An increase in the heat generation rate leads to a higher increase in the surface temperature. Therefore, the heat generation rates of the fuel rods are high when a VHTR operates at high power, and it is more challenging to implement passive safety design to ensure the temperature limitation of the fuel rods during a loss-of-coolant accident. Moreover, the heat transfer coefficient relative to time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was examined based on the experimental data.
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.
CFD Simulation of Fission Gas Release Under Fuel Defects Condition in Pressurized Water Reactor
Authors: Dong Bing , Li Leihao , Li Chenyue , Yin Junlian , Wang Dezhong
Published: October 2018

During normal pressurized water reactor (PWR) operation, the fuel cladding is inevitably defective. It is important to develop a model to evaluate the fission gas (FG) release from the fuel-cladding gap into the coolant, which benefits the fuel failure monitoring and improves the reactor safety. The objective of this paper is to give an overview of three-dimensional two-phase transient CFD simulation based on the Volume of Fluid (VOF) method to evaluate the fission gas migrating inside the gap and escaping through the defects. The topics of this paper include the behavior and release rate of FG in the fuel-cladding gap when the fuel cladding is defective. An analysis is presented on the suitability of the CFD simulation and it is shown that three-dimensional two-phase transient CFD simulation can be utilized in evaluating the FG release process. The results show that water enters the gap immediately after an abrupt cladding defect. The entering water flashes and causes pressure pulsation, which induces the transient FG release.
Design and Development of the Platform for Significance Determination Process System for Nuclear Power Plant
Authors: Zhang Qinfang , Zhang Guoxu , Wang Zilong , Dong Qi , Tang Guofeng
Published: October 2018

Significance determination process (SDP) is one of important application tools in the scheme of risk-informed and performance based regulation. It is used to help the analyst evaluate the risk significance of a particular safety concerned event. By referring to the design and development of SDP system, this paper gives a brief introduction on the design and development of the system. An autonomously developed PSA quantification engine is incorporated into the SDP system, and it helps to resolve quantification issue for event risk evaluation. From qualified screening to risk significance quantification, SDP system is designed to realize the whole process for event risk analysis. SDP system could potentially help analysts from both regulatory agency and NPP licensees to quickly evaluate the risk significance of a particular event.
Development of Experimental Technology for Simulated Fuel-Assembly Heating to Address Core-Material-Relocation Behavior During Severe Accident
Authors: Abe Yuta , Yamashita Takuya , Sato Ikken , Nakagiri Toshio , Ishimi Akihiro , Nagae Yuji
Published: October 2018

Authors are developing an experimental technology to realize experiments simulating Severe Accident (SA) conditions using simulant fuel material (ZrO2 with slight addition of MgO for stabilization) that would contribute not only to Fukushima Daiichi (1F) decommissioning but also to enhance the safety of worldwide existing and future nuclear power plants through clarification of the accident progression behavior. Based on the results of the prototype test, improvement of plasma heating technology was conducted. The Core Material Melting and Relocation (CMMR)-1/-2 experiments were carried out in 2017 with the large-scale simulated fuel assembly (1 m × 0.3 mϕ) applying the improved technology (higher heating power and controlled oxygen concentration). In these two tests, heating history was different resulting basically in similar physical responses with more pronounced material melting and relocation in the CMMR-2 experiment. The CMMR-2 experiment is selected here from the viewpoint of establishing an experimental technology. The CMMR-2 experiment adopted 30-min heating period, the power was increased up to a level so that a large temperature gradient (> 2,000 K/m) expected at the lower part of the core in the actual 1F accident conditions. Most of the control blade and the channel box migrated from the original position. After the heating, the simulated fuel assembly was measured by the X-ray Computed Tomography (CT) technology and by Electron Probe Micro Analyzer (EPMA). CT pictures and elemental mapping demonstrated its excellent performance with rather good precision. Based on these results, an excellent perspective in terms of applicability of the non-transfer type plasma heating technology to the SA experimental study was obtained.
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.
Investigation on Methods for Uncertainty Quantification of Constitutive Models and the Application in BEPU
Authors: Xiong Qingwen , Gou Junli , Shan Jianqiang
Published: October 2018

The best estimate plus uncertainty (BEPU) method is recommended by IAEA for nuclear safety analysis. Most of the existing BEPU methodologies rely on the uncertainty propagation of input parameters, while uncertainties of the constitutive models in best estimate codes tend not to be valued or even neglected. A structural method is proposed in this paper to quantify the uncertainties of the constitutive models. Different constitutive models will be classified according to the characteristics and corresponding method could be utilized for each model based on the method. Specific uncertainty quantification (UQ) methods adopted in this paper include the non-parametric curve estimation method, inverse method and design of experiment (DOE) method, and a model selection technique is adopted to opt the optimal model among all alternative models. The structural method is applied to the uncertainty evaluation of LOFT LP-02-6 experiment. Important models are identified, uncertainties of these models are quantified and propagated to the peak cladding temperature (PCT) through code calculations. Uncertainty of the PCT is quantified and the result shows that the calculated values could well envelop the experimental value.
Multiphysics Modeling of Fabrication Methods’ Effect on UO2-BeO Composite Fuels’ Performance
Authors: Zhou Wei , Zhou Wenzhong
Published: October 2018

UO2-BeO is one of the most promising accident tolerant nuclear fuels due to its excellent thermal conductivity compared to pure UO2 fuel. Two different UO2-BeO fabrication methods have demonstrated the capability to fabricate enhanced thermal conductivity UO2-BeO composite fuels and improve fuel performance. In one method, BeO is continuously distributed around UO2 grains while BeO is dispersed in the UO2 phase in the other method. In the former type, BeO is considered as matrix and UO2 as dispersed particle. However, in the later type, BeO is considered as dispersed particle and UO2 as matrix. To calculate the thermal conductivity of UO2-BeO composite, Hasselman-Johnson model has been applied, which shows good agreement with experimental data. In this model, it includes the influence of thermal conductivity of matrix and particle, volume fraction of particle, radius of particle and the interfacial thermal conductivity between matrix and particle. To balance the improvement of thermal conductivity and enrichment of UO2, a UO2-BeO composite fuel with 10% volume fraction of BeO has been chosen. Besides, the grain size of matrix should be noticeable smaller than particle to relax thermal stress which may cause micro-cracks and destruction, leading to the grain sizes of UO2 in two types being distinctive, and resulting in very big effect on fission gas release. In the paper the thermal conductivity has been intensively studied as well as fuel performance in two different types of UO2-BeO fuels, and the two fabrication methods have also been compared to assess their applications in commercial reactors.
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.
A Systematic Study of the Material Performance of Hot Isostatically Pressed Type 316L Stainless Steel Powder for the Civil Nuclear Sector
Authors: Kyffin William , Gandy David , Burdett Barry
Published: October 2018

Hot Isostatic Pressing (HIP) of type 316L stainless steel powder has been an established manufacturing practice for more than twenty-five years in the oil and gas sector and more recently in the naval defence sector. To demonstrate the capability of the powder metallurgy HIP (PM/HIP) for nuclear power applications a systematic study of 316L commercial powder production, encapsulation/consolidation providers and selected HIP parameters was undertaken by the Nuclear AMRC in collaboration with the Electric Power Research Institute (EPRI). In the study, the 316L powder specification limited the oxygen content of the powder to under 130 parts per million (ppm), which reflects the improvements that commercial powder suppliers have been making over the past decade to ensure greater powder cleanliness.

The test programme assessed powder supply, HIP service provider and HIP sustain time. Excellent test results were achieved across the full range of variables studied with all billets meeting the specification requirements of ASTM A988 and additional requirements imposed based on nuclear manufacturing standards.

Significantly, the study demonstrated the robustness of the PM/HIP supply chain, as material produced via differing HIP service providers resulted in very consistent material properties across the destructive test programme. Furthermore, no significant difference in material properties were noted for material HIP’ed between 2–8 hours hold time, suggesting that the HIP process window is large. Both these results are significant from an end-user standpoint as they highlight the uniformity of the process through the full manufacturing cycle from powder procurement to destructive testing.

Despite all material passing specification requirements, some property variation was noted for differing powder suppliers. Considering the systematic approach, this was attributed to powder composition, with both low oxygen and high nitrogen contents contributing to improvements in Charpy impact strength and tensile strength respectively.
Analysis and Parallel Implementation of Transient Thermal Feedback in Neutron Kinetics Calculation
Authors: Ming Pingzhou , Li Zhigang , An Ping , Lu Wei , Liu Dong , Yu Hongxing
Published: October 2018

Transient computing scheme named K-MOD has been designed that time-discrete is carried out by implicit Runge-Kutta method, and coarse nodal method is used for neutronics calculation at present. During transient calculation, thermal feedback is introduced to update the cross-section to approximate the transient behavior of the real core based on the method of interpolation on state-points. Corresponding thermal feedback for coupling calculation is programmed and analyzed, and parallel computing is introduced to enhance efficiencies of single-disciplinary codes and data coupling section. Problem of rod insertion and lift in real core of PWR (Pressurized Water Reactor) is tested and compared. The results show that K-MOD is in good agreement with the results of SMART software in both fixed time step and adaptive time step. And the parallel efficiency is also explained.
Analysis of Major Group Structures Used for Nuclear Reactor Simulations
Authors: Di Filippo Marco , Krepel Jiri , Mikityuk Konstantin , Prasser Horst-Michael
Published: October 2018

Nuclear reactor simulation is often based on multi-group cross-section libraries. The structure and resolution of these libraries have a strong influence on the accuracy and computational time; hence, number of groups and energy structure must be carefully considered. The relationship between group structures and how they impact generated cross-sections can be a critical parameter. Common energy boundaries shared among major group structures were identified and the relative kinship among those was reconstructed in an effort to build a family tree of major group structures. Stochastic code Serpent2 [1] was employed to generate cross-sections of selected isotopes at different reactor compositions and conditions, using the investigated energy group structures. The impact on their generation was quantified by spectral weighted deviation.

The 35 major energy structures were divided into three basic families. The key parameters distinguishing them were their applicability to thermal or fast reactors and their applicability in neutronic or multiphysics investigations. A sensitivity threshold of the generated cross-sections over the group structure resolution was investigated.

The aim was to identify a group structure with very low dependency on the actual reactor spectrum.
Experimental Study on the Steam Line Break (SLB) Accident With the Steam Generator Tube Rupture (SGTR)
Authors: Park Yusun , Bae Byoung Uhn , Kim Jongrok , Lee Jae Bong , Park Hae Min , Choi Nam Hyun , Kang Kyoung Ho , Kim Kyung Doo
Published: October 2018

The integral effect test to simulate a steam line break (SLB) accident accompanied by a steam generator tube rupture (SGTR), named SLB-SGTR-02 test, was performed to investigate the thermal hydraulic phenomena during a multiple failure accident.

In this study, a thermal-hydraulic integral effect facility, ATLAS (Advanced Thermal-hydraulic test Loop for Accident Simulation) was utilized to simulate thermal hydraulic phenomena which can occur in the nuclear power plant, especially for the pressurized water reactor, as realistically as possible. The SLB is simulated as a guillotine break on the main steam line and, in the SGTR, five steam generator tubes on the steam generator hot-side are modeled. Due to the initiation of SLB, a reactor was tripped by low steam generator pressure (LSGP) signal and the SGTR was initiated when the water inventory in the secondary system of the affected steam generator was dried-out.

During the transient simulation for SLB-SGTR-02 test, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the flows in the primary loops and the fluid temperatures, were measured and analyzed.

Through this experimental result, insights about the accident management procedure can be provided in the case of the multiple failure accident, such as an SLB with a SGTR.
Convective Heat Transfer in CANDU Spent Fuel Racks After a Loss of Coolant
Authors: Logtenberg Derek , Grant Wade , Chan Paul , Corcoran Emily
Published: October 2018

The event at the Fukushima Daiichi Spent Fuel Pools (SFPs) has renewed interest in quantifying the safety margins related to loss of coolant accidents in Irradiated Fuel Bays (IFBs). Thermal-hydraulic analyses of exposed spent CANDU fuel has been limited to a small number of bundles due to its complex bundle geometry and open rack design. This paper presents a process to predict the steady state temperature and velocity of air as it passes through a rack of spent fuel using analytical models and Computational Fluid Dynamics (CFDs) techniques. The scenario acts as lower bound estimate for the effectiveness of convection during a complete loss of coolant in a fuel bay by examining the heat-up of a stand-alone rack without flow resistance of the bundles. The correct incorporation of flow resistance is a necessary step before conclusions are made about the available safety margins of irradiated fuel bays.
Parametric Model for Ex-Vessel Melt Jet Breakup and Debris Bed Cooling
Authors: Moriyama Kiyofumi , Park Hyun Sun , Lee Mooneon , Park Jin Ho
Published: October 2018

The process of the ex-vessel molten core cooling in a pre-flooded reactor cavity during a severe accident of a light water reactor includes complicated phenomena such as melt jet breakup, debris bed formation and cooling and the molten core-concrete interaction. The melt coolability and its impact on the containment consequences are dependent on the interactions among those phenomena. A simplified parametric model, COOLAP-II, covering the melt jet breakup, debris bed formation and cooling was developed for the synthetic assessment of the ex-vessel melt coolability. The model was validated on the melt breakup and initial cooling by comparison with a full-model for fuel-coolant interactions, JASMINE.