bluuebunny/crossref_metadata_2025 huggingface dataset:
Systems Engineering Approach for Pre-Conceptual Design of DEMO Divertor
Authors: Marzullo Domenico , Dongiovanni Danilo N. , You Jeong Ha
Published: October 2018

The DEMOnstration Fusion power Plant (DEMO) will be a key step towards Fusion Power Plant technology. It represents the single step to a commercial fusion power plant, in charge of demonstrating the viability of relevant technologies. Indeed, the development of tokamak sub-systems has to take into account interface, structural and functional requirements and multi-physics issues that can be completely known only during the development of the design process. This leads to difficulties to be faced during the conceptual design, mainly related to the identification of the main requirements, the change management and the sub-system integration. The Systems Engineering approach aims to support the design and management of complex systems over their life-cycles, providing a systematic approach for the definition of customer needs and required functionality from the early stage of the design, as well as for the design synthesis and the system validation and verification.

Among the tokamak sub-systems, the divertor is the one devoted to power exhaust management and represents, at the same time, one of the most challenging components, in terms of materials, technologies and manufacturing. Current design activities, conducted in the in the framework of EUROfusion Consortium are in a pre-conceptual phase. Despite the early design stage, a systems engineering approach is being applied as an integrated, interdisciplinary R&D effort.

The paper therefore presents the modeling effort to the conceptual design of DEMO divertor aimed at identifying both system main functions and expected behavior, given the constraints imposed from either project requirement or from current viability of technological solutions. To allow for flexibility in design needed to explore the feasibility of different solutions at this pre-conceptual stage, the impact of possible changes in high level requirement or interfaces is also investigated. This is also achieved through the allocation of the requirements to the affected components and providing efficient traceability. Therefore, the proposed modelling approach is intended to support the whole divertor conceptual design stage, allowing for requirements identification, traceability and change management.
Time-Resolved Velocity Measurements in a Matched Refractive Index Facility of Randomly Packed Spheres
Authors: Kappes Ethan , Marciniak Mateusz , Mills Andrew , Muyshondt Robert , King Stephen , Nguyen Thien D. , Hassan Yassin A. , Ugaz Victor
Published: October 2018

Complex geometries and randomly connected void spaces within packed beds have hindered efforts to characterize the underlying transport phenomena occurring within. In this communication, we present our experimental studies on a facility of randomly packed spheres that can be a representative of sections within a reactor core in a nuclear power plant. The results of high-fidelity velocity measurements can be seen using Time-Resolved Particle Image Velocimetry (TR-PIV) at the pore scales and near the wall boundary in the Matched Index of Refraction (MIR) facility. The MIR approach allows for a non-invasive analysis of the flow within packed spheres at the microscopic scales with high temporal and spatial resolution. Flow characteristics obtained from the TR-PIV measurements at various Reynolds numbers are presented. The results include the first- and second-order flow statistics, such as mean velocity, root-mean-square fluctuating velocity and Reynolds stresses. Effects of the wall boundary and Reynolds numbers on flow patterns are currently being investigated. Comparisons of the mean velocities, root-mean-square fluctuating velocities, and Reynolds stress components show the increase of flow mixing and turbulent intensities within the gaps between spheres in the packed bed. Sizes of recirculation regions, however, seem to be independent of the increase of Reynolds numbers.
Dose Minimization Game for Smartphones
Authors: Stelter Nolan , Das Arnav , Hanifah Zahra , Uddin Rizwan
Published: October 2018

Due to misconceptions surrounding radiation and nuclear energy, educating the general public about basic radiation concepts has become increasingly important. The Virtual Education and Research Laboratory (VERL) at the University of Illinois at Urbana-Champaign (UIUC) has developed a 3D, virtual, interactive game that conveys the physics of radiation and principles of radiation protection to the player. The model is a scavenger hunt style game that takes place in a virtual model of a TRIGA research reactor. Several virtual radiation sources are placed in the 3D virtual model of the TRIGA facility. Radiation drops away from the radiation source. The effect of shielding can also be incorporated in modeling the radiation transport, leading to realistic radiation fields. The user’s goal is to find and collect (virtual) objects placed in this facility while minimizing the dose received in doing so. The player is meant to learn about time, distance, and shielding — key concepts in radiation protection. The start screen displays the radiation field in the form of a colored coded floor, as well as the location of the desired objects. With the given information, the player is encouraged to plan the route to collect the items and minimize exposure. By repeatedly playing the game, the player becomes familiar with the layout of the facility, and of the location of the radiation sources. This educational game is a useful teaching tool. Those unfamiliar with radiation protection concepts are able to understand how important time, distance, and shielding are in minimizing dosage. Additionally, this game proves to be a useful engagement and outreach tool. Upon completion of the game, the user is shown the score, the dose received, as well as a list of dose received in well-known instances such as eating a banana or in getting an x-ray at the dentist’s office.

The dose minimization game developed earlier for computers has now been developed for use as a game-app for cell phones. These recent developments allow for wider outreach, further increasing the use of the model as an outreach and educational tool.
Real-Time Online Risk Monitoring and Management Method for Maintenance Optimization in Nuclear Power Plant
Authors: Xu Anqi , Zhang Zhijian , Zhang HuaZhi , Zhang Min , Wang He , Ma Yingfei , Chen Sijuan , Wang Yan , Zheng Gangyang
Published: October 2018

As a nuclear power plant system configuration risk assessment tool, Risk Monitor (RM) can periodically update Living-PSA risk monitoring model to calculate risk information. It not only provides operating personnel with the real-time risk information, but also reduces human error in test and maintenance work, in order to ensure that the high-risk configurations are identified and controlled.

Traditionally, the high-risk identification of planning activities is mainly based on technical specification (TS) and operating experience, etc. It is lack of quantitative risk assessment to support decision-making, especially when multiple systems or components are out of service or being restored.

Planning risk assessment in existing risk monitors is time-consuming and error-prone, because they typically depend on manually screening and confirming risk-related planned events. Besides, online time-dependent characteristics of NPP systems are ignored. The reliability parameters are taken as constant values providing that components and systems always work under the predefined conditions. Thus, the risk level during planning period is not affected by the current configuration, or the operation history of equipment.

With the widespread application of digital instrument control systems (I&C) in nuclear power plants, the technology of Real-time Online Risk Monitoring (RORM) was put forward. It improves the acquisition technique of current equipment status, modeling and updating technique of Living-PSA Model and provides more accurate, realistic prediction of risk level for planning configurations.

This paper briefly introduces the design of the system structure, database, interfaces and functions of Real-time Online risk monitoring (RORM). It is characterized by the following features: online acquisition of the initial configuration before planning period, time-dependent risk monitoring modeling and updating in time. And it focuses how the real-time online risk monitoring technology of nuclear power plants help the nuclear power plant minimize the risk level of maintenance plan and optimize the maintenance schedule. Also, the calculation method of risk monitoring measures is improved based on the concept of “online, time-dependent”. Finally, the risk management method for optimization of planning activities is proposed.
An Experimental Study of Liquid Flooding in Vertical Large Scale Rectangular Channel With the Counter-Current Flow of Air and Water Film
Authors: Du Kashuai , Hu Po , Zhai Shuwei , Yang Xiaojie , Wang Weibo
Published: October 2018

In the present study, a test facility, called WAFT [1], has been set up to study liquid flooding phenomena on a heated large flat plate with the counter-current air-water flow.

The various thermal hydraulic working condition parameters, especially, temperature and velocity of water, air and oil media, could be obtained by adjust the control devices, such as, heater, valve and frequency regulator. Because of transparent organic glass slab in the test section, the dynamic behavior of water film flow in the different locations of test area can be clearly observed by virtue of high-speed video camera, such as wave pattern change, film rupture, droplet splash and dry spot occurrence.

During the test, a series of tests were performed under the conditions of steel plate heat flux 25 kw/m2, the water film with mass flow rate 0.1 and 0.6 kg/s.m and the temperature 70 and 92 °C, the velocity of air entrance ranging from 6.0 to 11.50 m/s and the inlet air temperature of 45 and 60 °C, respectively.

According to the acquisition data, a modified Wallis liquid flooding predicted relation was proposed considering the influence factor of temperature from large steel plates, air and water film, i.e., the effect of water film evaporation. The results indicate that the mass flow rate of water film and inlet air velocity have a great effect on the emergence of liquid flooding.
Numerical Investigation of Aerosol Collection Efficiency in a Venturi Scrubber
Authors: Goel Paridhi , Nayak Arun K.
Published: October 2018

In an extreme situation, as happened in Fukushima nuclear power plant, the failure of multiple safety systems may lead to severe core damage and melt relocation. This may be accompanied by production of large amount of steam which may result in the over-pressurization of the containment. Another associated concern is that the exposed core melt is highly radioactive which if exposed to the atmosphere can ruthlessly deteriorate the quality of environment and living beings. The radioactive materials present in the containment can be in vapor form or aerosol form. The containment of a nuclear power plant is therefore the final shielding to prevent the release of radioactive products to the environment. Therefore, the installation of Filtered Containment Venting system (FCVS) is mandatory in a nuclear reactor which actuates passively to depressurize the containment. Additionally it assists in the retention of radionuclides in the containment.

The FCVS consists of venturi scrubbers submerged in a pool of scrubbing liquid along with a demister housed in a scrubber tank. The performance of venturi scrubber is dependent on the interaction of the contaminated air stream from the nuclear reactor with the scrubbing liquid. This represents a multi-field and multi-fluid system. The present analysis investigates this system through a computational framework in which air stream is solved using Eulerian framework while the scrubbing liquid is tracked through Lagrangian framework. The collection efficiency of aerosols is modeled assuming impaction to be the predominant mechanism.
Design Optimization of Modernization of I&C System Using Digital Technology in NPPs
Authors: Zhang Longqiang , Yan Jiahong , Zhao Weining , Huang Weijun
Published: October 2018

There was a common use of instrument and control (I&C) system based on analog technology in design and construction of nuclear power plant built more than ten years ago. With the development and update of automation technology, digital control system has almost completely replaced the older generation technology in many areas of industry. For the nuclear power plant still using the analog I&C system, it caused the reduction of the production lines of related components and the lack of specified technical engineer. Along with the aging and obsolescence of analog technology equipment, the unsustainability of spares prompted the owners of nuclear power plant to implement modernization using digital technologies. Different from the design of digital control system in new nuclear power plant, the modernization project design is limited by the setting of the original system. Therefore, most owners adopt the function alternative strategy to implement the upgrading project. This strategy can effectively solve the problem of spares shortage, but it is difficult to fully elaborate the advantages of digital control system. Based on the technical characteristics of digital control system and the design experience derived from new nuclear power project construction, this paper puts forward the optimized design measures, under the limitation of the old power plant, to enhance the safety and economy of the nuclear power plant after final digital upgrading.
High Reliability Micro-Grid for a Nuclear Facility Emergency Power Supply
Authors: Lekhema Gerard R. , Cronje Willem A. , Korir Ian
Published: October 2018

This paper investigates the use of a high reliability micro-grid to supply emergency electrical power to a nuclear facility following loss of offsite power accident. Majority of the nuclear facilities around the world utilize diesel generators and battery banks as emergency power to back up the grid power supply. This power supply configuration represents the concept of the micro-grid system. The research work proposes reliability improvement of the emergency power supply by introducing distributed energy sources and large-scale energy storage systems. The proposed system was modelled with reliability block diagrams and analysed with Monte Carlo simulation to evaluate the system’s reliability in terms of loss of power supply probability (LPSP). The analysis takes into consideration, the failure probability of the components combined with the power unavailability due to the intermittent energy sources. The results show that the proposed system reduces the LPSP.
Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve
Authors: Takahashi Shiro , Ozaki Eiji , Minenaga Atsuyuki
Published: October 2018

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.
The Development of the Advanced Method for the Source Term Evaluation Applicable to the Dynamic PRA
Authors: Nakamura Koichi , Jang Sunghyon , Yamaguchi Akira
Published: October 2018

Useful insights on nuclear safety are provided by the level 2 probabilistic risk assessment (PRA), which evaluates the risk of fission products (FP) released in the environment during an accident in nuclear power plants (NPPs). The containment event tree method is generally employed for level 2 PRA. In this method, the accident scenarios are expressed by the combination of a number of branch points. The possible accident scenarios are approximately or representatively dealt with in this method. Dynamic PRA can evaluate the accident risk with complex changes or composition of various accident events dynamically. A reasonable source term evaluation method, which can be conducted with a small computational load, is developed for the establishment of dynamic PRA method focused on the risk of the release of FP in the environment. The proposed study aims to develop a reasonable source term evaluation method by applying phenomenological relationship diagram (PRD) method applicable to the dynamic PRA is developed.
Pressure Dependence of Two Phase Flow Behavior of Stagnant Water in a Vertical Pipe During Steam Injection
Authors: Kitahara Naoto , Yamamoto Yasunori , Narabayashi Tadashi , Chiba Go
Published: October 2018

Two-phase flow experiments and analysis were conducted to understand two-phase flow behavior of the water scrubbing pool of the filtered containment venting system with steam injection. In the early phase of steam injection, the water level gradually increased due to the steam condensation where the water surface was stable. After the water pool reached the saturation temperature, the diameter of bubbles increased when the bubbles moved upward in the water pool, where fluctuation of the water surface was observed. The water level increased when the scrubbing pool was pressurized by an orifice. Our simulation results showed that the decrement of the bubble velocity due to the pressurization may promoted the level swell.
Sensitivity of CTF Solution to Subchannel Window Size
Authors: Mukin Roman , Seidl Marcus , Clifford Ivor , Ferroukhi Hakim
Published: October 2018

In this work, a so-called mini-core consisting of a 3 × 3 array of 17 × 17 pressurized water reactor (PWR) fuel assemblies (FA) is considered with the aim of identifying the most conservative window size for hot channel analysis of bowed fuel assemblies. Overall, five different mini-core configurations are analyzed: one is the reference case, i.e. without FA displacement and four different cases with diagonal and parallel FA displacements. Rod power maps for these mini-cores were exported from neutronic calculations with CASMO-SIMULATE codes.

Subchannel modelling with COBRA-TF code of all five mini-cores allows one to identify the rod position with a minimum departure from the nucleate boiling ratio (DNBR) and to construct input decks with different rod window sizes around the previously identified rod position. Overall, eight different window sizes are considered: 3 × 3, 5 × 5, 7 × 7, 9 × 9, 11 × 11, 13 × 13, 15 × 15 and 17 × 17. Results of subchannel analysis for a mini-core and different subchannel window configurations are compared with the help of DNBR parameter, which is the ratio between the critical heat flux (CHF) and the actual local heat flux on a rod. An assessment of three different CHF models is applied in this work: Groeneveld CHF look-up table (LUT), W3 CHF correlation, and Doroschuk CHF LUT.

The general conclusion of this work is that for deformed core configurations, an appropriate rod window size needs to be determined to adequately capture the local flow redistribution. For large displacements (the largest displacement considered in this work is 10 mm), the DNBR ratio can drop to one. DNBRs obtained with the W3 CHF correlation give the most conservative results.
A Multi-Fluid Model Coupled With Interface Tracking Method for Simulation of Liquid Jet Breakup
Authors: Zhong Mingjun , Zhou Yuan
Published: October 2018

For liquid jet breakup, when small drops are fragmented from the surface of jet, the fluid’s interfaces with various length scales will coexist in the flow regimes. The paper presents a coupled method to simulate liquid jet breakup. The present method is based on Yan and Che’s strategy which has been proven to be capable for simulation of bubbly-slug flow. In the method, the basic multi-fluid model and interface tracking method are coupled by a unified solution frame work of MCBA-SIMPLE algorithm. The jet phase and continuous liquid phase are combined into a continuous phase in order that the large-scale interface of jet can be reconstructed by VOF/PLIC method. The coupled model consists two sub-models, the model based on VOF method and the conventional multi-fluid model; The relationships and switching of the sub-models are disused and summarized in the paper. Some cases are presented to show the capabilities of the current method. Firstly, the sub-model which is equivalent to the VOF method is used to simulate the interface behavior during the jet breakup and a breakup length equation is correlated. Then the coupled model is applied to the same simulation. A simple jet breakup model is used to simulate mass transfer of drops from jet. Variation of the drop surface area is considered by solving a transport equation. The simulation results preliminarily show that the current method is capable to simulate the complex jet breakup process. The main characteristics of the process such as breakup length and drops distribution are reasonably simulated.
CFD Thermal Analysis of ITER Pressure Suppression Tanks
Authors: Basili Lorenzo , Lo Frano Rosa , Olcese Marco , Sekachev Igor , Aquaro Donato
Published: October 2018

The aim of the paper is to investigate the thermal conditions (temperature distribution, heat losses) in the support system of the Vapour Suppression Tank (VST) of the Vacuum Vessel Pressure Suppression System (VVPSS), a safety important system of ITER fusion reactor, protecting the Vacuum Vessel (VV) against overpressures. The VVPSS includes four VSTs of identical volume and mounted as two stacked assemblies.

The study focuses on the optimization of the design of the thermal insulation of the bottom part of the VST, connecting each two-tank stack to the basement, and also on the identification of the thermal loads at the interface between the tank support and the tank pressure boundary.

A Computational Fluid Dynamics (CFD) analysis of the VST has been performed for four different insulation configurations and considering both steady state and transient loads following accidental conditions.

The results of the analysis are used to provide recommendation on the optimum configuration of the thermal insulation. Measures for minimization of the thermal gradient in the critical area of the joint between the tank hemispherical head and support skirt to limit the thermal fatigue on the welds are also suggested.
Dynamic Model of a Seawater Desalination Plant Based on the Nuclear Heating Reactor and MED-TVC
Authors: Pan Yifei , Dong Zhe
Published: October 2018

The 200 MWth nuclear heat reactor II (NHR200-II) is a typical integral pressurized water reactor (iPWR) being developed by the institute of nuclear and new energy technology (INET) in Tsinghua university. The NHR200-II, which has inherent safety features such as full-range natural circulation, passive residual heat removal, self-pressurization and control rod hydraulically driving, can be adopted as a clean base-load energy source for a sea-water desalination plant having the process of multi-effect desalination with thermal vapor compression (MED-TVC). Dynamic modelling of the sea-water desalination plant coupled by the NHR200-II and MED-TVC is necessary for the design of its plant control strategy, which is important for the stable and efficient operation. In this paper, a lumped parameter dynamic model of NHR200II-based sea-water desalination plant with the process of MED-TVC is proposed based upon the conservation laws of mass, momentum and energy. The modeling verification in both the steady-state and open-loop dynamic-state are given, which show the suitability of applying this model for control system design. Finally, the closed-loop responses in the case of power-level maneuver from 100% to 50% full power is given.
Design and Feasibility Analysis of the Electricity Generation System Based on Residual Heat
Authors: Dong Zhe , Liu Miao , Pan Yifei
Published: October 2018

Passive residual heat removal system (PRHRS) is of great significance for reactor shutdown safety. The PRHRS of a small modular reactor, such as the integral pressurized water reactor (iPWR) and the modular high temperature gas-cooled reactor (MHTRG), is composed of the primary loop (PL), intermediate loop (IL) and air-cooling loop (AL). The AL is a density-difference-driven natural circulation caused by the difference of air temperature at the inlet and outlet of the air-cooling tower. Thus, it is possible to adopt the air flow in AL to generate electricity for post-shutdown reactor monitoring. In this paper, a novel residual heat electricity generation system (RHEGS), which is composed of the PRHRS and a vertical wind generator installed in the air-cooling tower, is proposed for the power supply of post-shutdown monitoring instruments. To verify the feasibility of practical implementation, the dynamical model of this newly designed RHEGS including the dynamics of PRHRS, windmill, rotor as well as doubly-fed induction generator (DFIG) are all given. Then, both steady-state and transient verification for the RHEGS of a nuclear heating reactor NHR200-II plant with a rated thermal power of 200 MWth is carried out, which shows that the output active power of RHEGS can be 20∼30kW which is about 1% the residual heat flux and can fully meet the power requirements of post-shutdown monitoring instruments.
Verification of Alarm Displays for the Nuclear Power Plant With Two Modular High-Temperature Gas-Cooled Reactors
Authors: Qianqian Jia , Chao Guo , Jianghai Li , Ronghong Qu
Published: October 2018

The nuclear power plant with two modular high-temperature gas-cooled reactors (HTR-PM) is under construction now. The control room of HTR-PM is designed. This paper introduces the alarm displays in the control room, and describes some verification and validation (V&V) activities of the alarm system, especially verification for some new human factor issues of the alarm system in the two modular design. In HTR-PM, besides the regular V&V similar to other NPPs, the interference effect of the alarm rings of the two reactor modules at the same time, and the potential discomfort of the two reactor operators after shift between them are focused. Verifications at early stage of the two issues are carried on the verification platform of the control room before the integrated system validation (ISV), and all the human machine interfaces (HMIs) in the control room, including the alarm system are validated in ISV. The test results on the verification platform show that the alarm displays and rings can support the operators understand the alarm information without confusion of the two reactors, and the shift between the two reactor operators have no adverse impact on operation. The results in ISV also show that the alarm system can support the operators well.
Proposal of a Simple Evaluation Method for Sloshing Impact Pressure on Flat Roofs
Authors: Takaya Shigeru , Fujisaki Tatsuya
Published: October 2018

Sloshing is one of important issues for both loop-type and tank-type fast reactors with free liquid surface. Periods of seismic vibration are lengthened by base isolation systems installed to prevent damages to facilities during earthquakes, and get close to the natural periods of sloshing. As a result, sloshing is promoted. Sloshing waves are assumed to even reach a roof slab of a reactor vessel in severe seismic conditions. It is important to evaluate structural integrity for sloshing impacts on roofs. However, there are not any established evaluation methods for impact pressure on flat roofs yet. Therefore, in this study, a simple evaluation method is proposed based on Wagner’s theory. The effectiveness of the proposed method is illustrated using computational fluid dynamics analysis and literature data of sloshing experiments.
System Safety Analysis Method Based on Real-Time Online Risk Monitoring Technology
Authors: Chen Sijuan , Zhang Zhijian , Wang He , Zhang Min , Zhang Huazhi , Xu Anqi , Ma Yingfei , Zheng Gangyang
Published: October 2018

In the continuous operation process of Nuclear Power Plant (NPP), its configuration is full of variety over time because of the system’s dynamic characteristics. There is a great need to update the risk/safety analysis models when it becomes necessary to reflect those dynamic characteristics of the system/component. Most of the current methods for risk/safety analysis belong to the scope of safety pre-analyzing, which analyzes the system risk/safety before system being in service. The main purpose of these safety pre-analyzing is to guide system design and optimization, but the real-time operational risk/safety analysis of NPPs is considered little. In order to know well the real-time risk/safety for system, a System Safety Analysis Method based on Real-time Online Risk Monitoring Technology is proposed. The safety risk model is established based on the modular fault tree that is used to represent logic structure of system. The real-time risk/safety is monitored according to the correspondence monitoring signal or data of component/system. Simultaneously the method can account for the change of risks based on the established mapping relationship between the state transition rules and corresponding risk/safety model updating rules. Finally, a case monitoring the safety for the system of two redundant pumps was used to demonstrate the effectiveness of the method.
Cost Analysis for Decommissioning of Nuclear Power Plants With Uncertainties
Authors: Watanabe Naoko , Miyoshi Ryohei , Kozaki Tamotsu , Tanaka Shingo , Yanagihara Satoshi
Published: October 2018

Decommissioning cost including radioactive waste management for 1100 MWe nuclear power plant (BWR) was analyzed comparing multiple scenarios. The total cost of decommissioning nuclear power plant was first estimated including the radioactive waste management cost for the standard Japanese decommissioning case with 30 years of the project duration including approximately 20 years in safe storage. It showed that the cost relating to waste management accounts for more than half of the total cost. Focusing on the radioactive waste management cost, the duration of safe storage was varied as a parameter. The timing of waste disposal was a key parameter determining the waste management cost due to the decay of radioactive nuclides resulting in the decrease in the total volume of the radioactive waste, and the change in the ratio of the waste volume in the three radioactive waste categories (intermediate-level, low-level, and extremely low-level). The total cost showed the minimum value at around 60 years of the project duration balancing the waste management cost and period dependent cost for safe storage.
Moving From V&V to V&V&C in Nuclear Thermal-Hydraulics
Authors: D’Auria Francesco , Lanfredini Marco
Published: October 2018

V&V constitutes a powerful framework to demonstrate the capability of computational tools in several technological areas. Passing V&V requirements is a needed step before applications. Let’s focus hereafter to the area of (transient) Nuclear Thermal-hydraulic (NTH) and let’s identify V1 and V2 as acronyms for Verification and Validation, respectively.

Now, V1 is performed within NTH according to the best available techniques and may not suffer of important deficiencies if compared with other technological areas. This is not the case of V2. Three inherent limitations shall be mentioned in the case of Validation in NTH:

1. Validation implies comparison with experimental data: available experimental data cover a (very) small fraction of the parameter range space expected in applications of the codes; this can be easily seen if one considers data in large diameter pipe, high velocity and high pressure or high power and power density. Noticeably, the scaling issue must be addressed in the framework of V2 which may result in controversial findings.

2. Water is at the center of the attention: the physical properties of water are known to a reasonable extent as well as large variations in values of quantities like density or various derivatives are expected within the range of variation of pressure inside application fields. Although not needed for current validation purposes (e.g. validation ranges may not include a situation of critical pressure and large heat flux) physically inconsistent values predicted by empirical correlations outside validation ranges, shall not be tolerated.

3. Occurrence of complex situations like transition from two-phase critical flow to ‘Bernoulli-flow’ (e.g. towards the end of blow-down) and from film boiling to nucleate boiling, possibly crossing the minimum film boiling temperature (e.g. during reflood).

Therefore, whatever can be mentioned as classical V2 is not or cannot be performed in NTH. So, the idea of the present paper is to add a component to the V&V. This component, or step in the process, is called ‘Consistency with Reality’, or with the expected phenomenological evidence. The new component may need to be characterized in some cases and is indicated by the letter ‘C’. Then, the V&V becomes V&V&C.

The purpose of the paper is to clarify the motivations at the bases of the V&V&C.
Radiation Protection Calculation and Optimization for Shielding Design Around the Refueling Pipelines of HTR-PM
Authors: Fang Sheng , Li Hong , Li Wenqian
Published: October 2018

The high-temperature gas-cooled reactor pebble-module (HTR-PM), which is under constructed in Shidao, Shandong Province of China, adopts the spherical graphite coated components as fuel elements. During the operation of the reactor, the spherical fuel elements will be reloaded into the core periodically through the refueling pipelines, which will cause radioactive irradiation to personnel and facilities in several corridors and rooms. The three-dimensional geometry model was built according to the real structure and the radiation shielding design was verified and optimized by the QAD-CGA program. In the calculation, the average gamma intensities of the spherical fuel elements of 15 different cycle times were adopted as the source term. From the results, it can be concluded that: 1) For the 20-meter corridor, the dose rate will be less than 1E−6mSv/h at 30cm outside the concrete wall of 0.6m thickness by single fuel elements; 2) For the 0-meter corridor, an extra 32.0 cm steel shield should be added to satisfy the personnel dose rate limit of 0.03mSv/h; 3) For semiconductor electrical equipment closed to the pipelines, the required steel shield thickness is 12.0 cm under the accumulative dose limit of 50Gy; 4) For non-radiation resistant cables, the required steel shield thickness is 4.5 cm under the accumulative dose limit of 1000Gy.
Effect of the Jet Stability on Supersonic Steam Injector Operation
Authors: Miwa Shuichiro , Moribe Takahiro , Akiyama Nozomu , Sakashita Hiroto
Published: October 2018

In order to maintain the reactor core integrity in case of severe accidents, multi-layered installation of backup safety systems with the passive feature is crucial, as was evident from the Fukushima Daiichi accident. The supersonic steam injector (SI), also known as a passive jet pump driven by high-pressure steam and water, is capable of operating without external power supplies nor rotating machinery. Along with the Reactor Core Isolation Cooling (RCIC) system, the SI can possibly be considered as one of the candidates as a backup safety system for current and next-generation reactors. Although the fundamental experiments on SI have been carried out for past few decades, its operation characteristics are still not well understood. Especially, its stable operation condition as a passive jet pump hasn’t been investigated thoroughly. Our study shows that the SI’s operating range is highly dependent on inlet water flow rate and steam pressure. In addition, due to the direct contact condensation between steam and water jet within SI unit, the water jet stability may influence SI’s operation. In order to deepen understanding of SI’s operation characteristics, this study aims to investigate on the inlet water temperature effect, which has a direct influence on steam-water condensation as well as the water jet stability within the mixing nozzle section of SI unit.
Thermal Hydraulic Analysis of the CIRCE-HERO Pool-Type Facility
Authors: Gonfiotti Bruno , Barone Gianluca , Angelucci Morena , Martelli Daniele , Forgione Nicola , Del Nevo Alessandro , Tarantino Mariano
Published: October 2018

To support the activities carried out by the ENEA Brasimone Research Center on the Lead-cooled European Advanced Demonstration Reactor (LEADER) project, University of Pisa has improved the RELAP5/Mod3.3 code introducing the Lead and Lead-Bismuth Eutectic (LBE) as coolant fluids. This new version of the STH code was applied to perform several analyses for the steam generator HERO (Heavy liquid mEtal pRessurized water-cOoled tubes) installed in the CIRCE (CIRColazione Eutettico) pool-type facility that uses LBE as working fluid. HERO consists of a bundle of seven Double-Wall Bayonet Tubes (DWBT) with a leakage monitor system, and it was tested in the CIRCE facility to support the development of the steam generator proposed for ALFRED (Advanced Lead Fast Reactor European Demonstrator).

A preliminary analysis was performed comparing the RELAP5/Mod3.3 results against available experimental data. In particular, an isothermal test of the CIRCE-HERO facility allowed to set-up the RELAP5/Mod3.3 nodalization. As a second analysis, a steady state at full power was investigated. This steady-state condition was employed to perform a sensitivity study on the influence of the thermal conductivity of the stainless steel powder characterizing the HERO steam generator. Finally, the analysis of unprotected and protected loss of flow transients, and the analysis of a protected loss of flow with a simultaneous partial Loss of Heat Sink (LOHS) were performed. These last transients were investigated to assess the behavior of the HERO steam generator, especially to verify the possibility to establish a natural circulation capable to safely remove the “decay heat”.

The results of the first analysis (isothermal hydraulic characterization) demonstrated the good capabilities of improved RELAP5/Mod3.3 code to reproduce the forced circulation conditions. The results of the second analysis (steady-state at full power) highlighted the strong influence of the stainless steel powder thermal conductivity on the achieved steady-state conditions. The simulations of the last tests provided the time evolution of the LBE mass flow rate and of the temperatures in different regions inside the pool during unprotected and protected loss of flow transients. In the paper, a critical discussion of the obtained results is performed.
Evolution and Implementation of the Design Extension Conditions (DEC) Concept: Assessment of Selected Events
Authors: Kral Pavel
Published: October 2018

The paper describes at first the evolution of the Design Extension Condition (DEC) concept, that started in early 1990ies and was accelerated after the Fukushima accident in 2011. Nowadays the concept is elaborated in EUR document [1], WENRA Reference Levels [4], and finally in the IAEA Safety Standards [6][7]. Number of countries is implementing the DEC concept into the national legislation, safety assessment, design modifications and operation of NPPs. Typically, the DEC events are divided to 2 groups: DEC without core melt (DEC-A) and DEC with core melt (DEC-B).

The second part of the paper deals with safety assessment of DEC events with more detailed information about deterministic analysis of DEC-A events. Legislative and methodology basis, acceptance criteria, computer codes and their validation, list of scenarios to be analyzed, examples of analyses performed, and incorporation of results of new analyses into SAR are described step by step.
Influence of Gap Size on Added Mass for Spent Fuel Storage Rack
Authors: Lu Daogang , Liu Yu , Zheng Shu
Published: October 2018

Free standing spent fuel storage racks are submerged in water contained with spent fuel pool. During a postulated earthquake, the water surrounding the racks is accelerated and the so-called fluid-structure interaction (FSI) is significantly induced between water, racks and the pool walls[1].

The added mass is an important input parameter for the dynamic structural analysis of the spent fuel storage rack under earthquake[2]. The spent fuel storage rack is different even for the same vendors. Some rack are designed as the honeycomb construction, others are designed as the end-tube-connection construction. Therefore, the added mass for those racks have to be measured for the new rack’s design. More importantly, the added mass is influenced by the layout of the rack in the spent fuel pool.

In this paper, an experiment is carried out to measure the added mass by free vibration test. The measured fluid force of the rack is analyzed by Fourier analysis to derive its vibration frequency. The added mass is then evaluated by the vibration frequency in the air and water. Moreover, a two dimensional CFD model of the spent fuel rack immersed in the water tank is built. The fluid force is obtained by a transient analysis with the help of dynamics mesh method.
Feasibility Study of a New Fabrication Method for the Li4SiO4 Pebbles
Authors: Lo Frano R. , Puccini M. , Stefanelli E. , Luppichini M. , Grima C. , Vitolo S. , Aquaro D. , Malquori S. , De Sanctis S.
Published: October 2018

In the past decades many R&D efforts have been spent in the development of a suitable Li4SiO4 fabrication method (e.g., melt spray process, graphite bed method, capillary-based microfluidic wet method etc.), nevertheless we are still far from an “industrial standard solution”.

The aim of the paper is to develop a new fabrication method capable to produce stable and well-sized pebbles of lithium orthosilicate (Li4SiO4) based on the drip casting. This method is mainly based on the dripping at room temperature, which is novel in the framework of available fabrication processes requiring high temperature: this latter is demonstrated to affect the final product characteristics. It is worthy to remark that the Li4SiO4 is a candidate material for the breeding blanket material of the fusion power reactor. In the paper we will describe the experimental apparatus, designed and built at DICI - University of Pisa with the collaboration of Bitossi Industries, and the procedure adopted in order to produce pebbles of Li4SiO4.
Numerical Simulation on the Flow Through the Inlet Hole of Fuel Assembly in a Fast Reactor
Authors: Yu Wang , Dao-gang Lu , Hai-qi Qin , Yi-dan Han , Da-wen Zhong
Published: October 2018

In fast reactor, the coolant flows into the inside of the fuel assembly through the inlet hole, so according to the design requirement, it is necessary to determine a suitable inlet hole model to make the pressure drop and velocity reasonably. In this paper, three hole models are researched and the distance of adjacent hole of every model is different. Based on the hydraulic test of fuel assembly in fast reactor, a simplification model of distribution header and assembly pin has been built. The computational fluid dynamic code—CFX was used to simulate the pressure field and velocity field of the assembly pin with k-e turbulence model. The results show when the distance of adjacent hole increases, the change of pressure and velocity in the hole area is gently, which makes little vibration. The research will be the reference for the following simulation work and assembly hydraulic test.
Neutronic Analysis for VVER-440 Type Reactor Using PARCS Code
Authors: Ratti Luca , Mazzini Guido , Ruščák Marek , Giusti Valerio
Published: October 2018

The Czech Republic National Radiation Protection Institute (SURO) provides technical support to the Czech Republic State Office for Nuclear Safety, providing safety analysis and reviewing of the technical documentations for Nuclear Power Plants (NPPs). For this reason, several computational models created in SURO were prepared using different codes as tools to simulate and investigate the design base and beyond design base accidents scenarios.

This paper focuses on the creation of SCALE and PARCS neutronic models for a proper analysis of the VVER-440 reactor analysis. In particular, SCALE models of the VVER-440 fuel assemblies have been created in order to produce collapsed and homogenized cross sections necessary for the study with PARCS of the whole VVER-440 reactor core.

The sensitivity study of the suitable energy threshold to be adopted for the preparation with SCALE of collapsed two energy-group homogenized cross sections is also discussed. Finally, the results obtained with PARCS core model are compared with those reported in the VVER-440 Final Safety Report.
15 Years of the European Nuclear Education Network (ENEN Association)
Authors: Cizelj Leon , Starflinger Jörg , Decobert Veronique , Bazargan-Sabet Behrooz , Tuomisto Filip , Coeck Michèle , Anzieu Pascal , Roberts John , Kokalova Wheldon Tzanny , Dieguez Porras Pedro
Published: October 2018

The European Nuclear Education Network (ENEN) was established in 2003 through an EU Fifth Framework Programme (FP) project, as a legal nonprofit-making body. Its main objective is the preservation and further development of expertise in the nuclear fields by higher education and training. This objective is realized through the cooperation between EU universities involved in education and research in nuclear disciplines, nuclear research centers and the nuclear industry. As of March 2018, ENEN has 66 members in 18 EU countries and has concluded Memoranda of Understanding (MoU) with partners beyond Europe for further cooperation, including organizations in, Russian Federation, Ukraine, Canada and Japan. ENEN also has good collaboration with national networks and international organizations such as the Belgian Nuclear Education Network (BNEN) and the International Atomic Energy Agency (IAEA). The main activities developed, and results achieved, within the first 15 years of the ENEN Association are presented and discussed. These include, for example, the launch of the European Master of Science in Nuclear Engineering (EMSNE), the annual ENEN Ph.D. competition and the portfolio of more than 10 EURATOM projects dealing with nuclear education, training and knowledge management through development of teaching methods and materials, courses, and exchange of students and teachers within EU and beyond. Those projects were all supported by the European Commission with the ENEN Association acting as the coordinator or as a partner.