bluuebunny/crossref_metadata_2025 huggingface dataset:
Multi-Dimensional Heterogeneous Resonance Integral Tables Generated for Embedded Self-Shielding Method Towards Irregular Lattices
Authors: Zhang Qian , Zhao Qiang , Wu Hongchun , Cao Liangzhi , Zheng Zheng
Published: October 2018

The concept of multi-dimensional heterogeneous resonance integral tables is proposed. The new type of resonance integral is designed for different fuel pins appearing in one lattice with two extra dimension of optical radius and number density ratio in the fuel. Numerical results show that this treatment improves the accuracy of embedded self-shielding method on irregular lattices.
Design of a Closely-Spaced Rod Bundle for a Reference Direct Numerical Simulation
Authors: Shams Afaque , Kwiatkowski Tomasz
Published: October 2018

Detailed knowledge of a coolant flow in a fuel assembly of a reactor core has always been a major factor in the design of new nuclear systems. In this regard, traditionally adopted subchannel analysis codes cannot take into account local phenomena, which are quite essential. On the other hand, Computational Fluid Dynamic (CFD) is being recognized as a valuable research tool for thermal-hydraulics phenomenon in the fuel assembly geometries. Because of the high Reynolds number and geometric complexities, the practical CFD calculations are mostly limited to pragmatic Reynolds Averaged Navier-Stokes (RANS) type modelling approaches. A good prediction of the flow and heat transport inside the fuel rod bundle is a challenge for such RANS turbulence models and these models need to be validated. Although the measurement techniques are constantly getting improved, however, the CFD-grade experiments of flow mixing and heat transfer in the subchannel scale are often impossible or quite costly to be performed. In addition, lack of experimental databases makes it impossible to validate and/or calibrate the available RANS turbulence models for certain flow situations. In that context, Direct Numerical Simulation (DNS) can serve as a reference for model development and validation.

The aim of this work is to design a numerical experiment in order to generate a high quality DNS database for a tight lattice bare rod bundle, which will serve as a reference for the validation purpose. The considered geometric design is based on the well-known Hooper experiment, which contains a bare rod bundle with pitch-to-diameter ratio of P/D = 1.107. Performing a DNS computation corresponding to the Hooper experiment requires a huge computational power. Hence, a wide range of unsteady RANS (URANS) study has been performed to scale-down the Reynolds number such that it is feasible for a DNS computation and at the same time it still preserves the main flow characteristics. In addition to the flow field, a parametric study for three different passive scalars is performed to take into account the heat transfer analysis. These passive scalars correspond to the Prandtl numbers of air, water and liquid metal fluids. The heat transfer of these three fluids has been studied in combination with two different boundary conditions at the walls, i.e. a constant temperature and a constant heat flux. Finally, the obtained URANS results are used to compute the Kolmogorov and Batchelor length scales in order to estimate the overall meshing requirements for the targeted DNS.
Simulation Research on Thermal-Hydraulic Performance of a Natural Circulation Integrated Pressurized Water Reactor
Authors: Zhao Yanan , Peng Minjun , Xia Genglei , Lv Lianxin
Published: October 2018

As an effective measure to improve the reactor’s inherent safety feature, natural circulation is widely used in current integrated reactor design. The thermal-hydraulic performance of a flashing-driven nature circulation integrated pressurized water reactor (NC-IPWR) is studied, by taking IP100 reactor as reference. The simulation model of the reactor is established by RELAP5 code. A control system is designed based on the operation characteristics of the reactor. Both steady-state and dynamic performance of the reactor are analyzed and the rationality of the control strategy is verified in this work. The results demonstrate the operation characteristics of the IP100 reactor, and the dynamic performance of the reactor during power variation is discussed in detail. The control strategy that keeps the steam pressure and the core outlet temperature constant shows good performance under normal operation conditions. The obtained analysis results are significant for deeper understanding and improving the operation characteristics of the IP100 reactor.
Identification of Gas Accumulation Susceptibility in NPP’s Safety Related Systems and Operability Evaluation due to Gas Transportation
Authors: Hung Zhen-Yu , Huang Pei-Hsun , Li Chao-Jen
Published: October 2018

Instances of gas accumulation in the subject systems have occurred since the beginning of commercial nuclear power plant operation. NRC Generic Letter 2008-01, “requests that each licensee evaluate its ECCS, DHR system, and Containment Spray system licensing basis, design, testing and corrective actions to ensure that gas accumulation is maintained less than the amount that challenges operability of these systems, and that appropriate action is taken when conditions adverse to quality are identified.” All of the three NPPs have accomplished this evaluation and propose some corrective measures like revision of the operation procedures, installing the venting valves etc. Taipower also committed AEC to establish the acceptable quantity of the gas accumulation and continue to follow the development of gas transport methodologies in the industry. According to the NRC NRR Action Plan TAC.NO. ME3939 GAS MANAGEMENT (March 2011), other safety related systems, in additional to the systems covered by GL 2008-01, also have the gas accumulation issue and therefore must be concerned. This project will develop the numerical evaluation process with two phase flow software to simulate the gas accumulation and the transportation phenomena for the GL 2008-01 systems and validate the results by experimental analysis.
Level 1 PRA for External Vessel Storage Tank of Japan Sodium-Cooled Fast Reactor in Scheduled Refueling
Authors: Yamano Hidemasa , Naruto Kenichi , Kurisaka Kenichi , Nishino Hiroyuki , Okano Yasushi
Published: October 2018

Spent fuels are transferred from a reactor core to a spent fuel pool through an external vessel storage tank (EVST) filled with sodium in sodium-cooled fast reactors in Japan. This paper describes identification of dominant accident sequences leading to fuel failure by conducting probabilistic risk assessment for EVST designed for a next sodium-cooled fast reactor plant system in Japan to improve the EVST design. Based on the design information, this study has carried out identification of initiating events, event and fault tree analyses, human error probability analysis, and quantification of accident sequences. Fuel damage frequency of the EVST was evaluated approx. 10−6 /year in this paper. By considering the secondary sodium freezing, the fuel damage frequency was twice increased. The dominant accident sequence resulted from the common cause failure of the damper opening and/or the human error for the switching from the stand-by to the operation mode in the three stand-by cooling circuits. The second dominant accident sequence following the secondary pump trip is sodium freezing caused by the failure of air blower trip in the air cooler due to the common cause failures of secondary sodium flowmeter failure or erroneous opening of the air cooler damper. The Fussell-Vesely importance and risk achievement worth analyses have indicated high risk contributions.
Effect of Ocean Conditions on Neutronic/Thermal-Hydraulic Coupling of IPWR
Authors: Xia Genglei
Published: October 2018

In this paper, a thermal-hydraulics analysis code under non-inertial system was developed by establishing dynamic simulation models of typical ocean conditions and modifying RELAP5 code. Based on the modified code, the effects of the inclination, fluctuation and rolling conditions on the natural circulation characteristics of an integrated PWR were studied. The dynamic input and output interface model of RELAP5 was further established, and RELAP5 was coupled with a two-group three-dimensional neutron kinetics code to realize the improvement of nuclear feedback module in thermal hydraulic code. And the distribution of core flow and power under different motions was analyzed. The results indicate that, the uneven distribution of coolant flow increases with the increasing inclination angle, this trend leads to an uneven distribution of primary coolant flow. The flow fluctuation has a 180° phase difference under rolling conditions, and the reactor power and coolant flow oscillation increases with the increasing rolling period and amplitude. In the case of heaving motion, the peaks of the oscillation amplitude of the flow and power lying in the hottest channel as the additional forces on the fluid of each channel are spatially uniform. The code developed in this paper has the functions of modelling ocean conditions and three-dimensional coupled neutronics/thermal-hydraulics, and can be used as a simulation tool for Floating PWR.
Optimization of Fuel Storage in Spent Fuel Pool
Authors: Wang Xinyu , Kurwitz Richard Cable , Zhang Zhijian
Published: October 2018

This paper is the optimization of fuel assembly placement in the spent fuel pool according to the categorized rules. Only some specifically reactivity class assemblies could put together as the pattern. The allowable patterns, the number of the fuel assembly for each reactivity class and the number of RCCAs are from the nuclear power plant technique specification. Each assembly in the pool should obey the pattern rules and the user needs the optimal spent fuel pool configuration that could maximize the free space. In this study, the genetic algorithm and greedy algorithm are discussed, and both of two algorithms have the difficulties to the real engineering problem. A new approach that improves the greedy strategy at each step is proposed, make the greedy algorithm is more adapted to the engineering problem. Use the new approach to test Seabrook Unit 1 and Arkansas Unit 2 spent fuel pool at different cases, and show results by the visible figures. The output arrangements by the program shown that the results are satisfied the user requirements.
A Recommended Method for SC Wall Design-Evaluation Regarding the Elasto-Plastic Behavior Under Beyond Design Basis Seismic Loading
Authors: Shang Ziduan , Huang Xiao , Sun Yugang , Chu Meng
Published: October 2018

Steel Concrete Composite Wall (SC-Wall) is a unique composite structural component designed and used for shield building wall in new nuclear power plants construction. The composite function (action) of steel-plate and concrete is through the use of embedded studs, which are designed in an appropriate pattern to secure the connections between concrete and surface steel-plates. Thus the behavior of studs (primarily shear behavior) determines the composite state and its functions for serving as an integrated one-piece (or monolithic) section. For elastic state (linear) and ultimate state design, ACI 318 / 349 and recent published design specifications provide formula / equations for shear design-evaluations between plate stud and concrete; but for Beyond Design Basis (BDB) loading conditions, since the section behaves in elasto-plastic (or plastic) state, the constitutive relations among plate-stud-concrete have not yet established in current prevailing codes / or standards. Considering this situation, this paper is contributed to investigate the section behaviors of SC wall subjected to above BDB seismic loadings, provide recommendations for constitutive relations to reveal the actual strain-stress conditions and composite states, and further provide criteria suitable for section design and evaluations under such a circumstance. This research and investigation are based on an analytical methodology and physical model.
A Mathematical Link Between the Natural Energy of Stars and Fission
Authors: Bayles Brenda K. Krkosska
Published: October 2018

The goal of basic research is to gather information and build upon the knowledge base. This type of study identifies a problem and suggests solutions that applied research can consider and employ using empirical methodologies. Basic theory is a valid and accepted research tool, it is often mathematical in nature, and it is formulated to solve a particular problem. The problem is to find the mathematical link between the natural energy of stars and the energy created through fission. An equation is shown here to have a place in the field of nuclear engineering because it provides insight into the question of what makes the binding and decaying of atomic particles mathematically possible. This equation shows that when the two particles of unequal mass move close enough to be caught up in each other’s flow fields, they are entangled in the fields and remain bound together. When two particles are of equal mass however their encounter does not produce a binding mechanism because their flow fields are the same strength and cancel each other out. The fission of uranium-236 into the elements krypton-92 and barium-141 is discussed according to the equation. Then the fusion of deuterium and tritium is analyzed as an example of flow fields at work. Finally the first three steps of the proton-proton chain reaction are compared to the equation.
Nuclear Facility Safety Enhancement Using Sandia National Laboratories’ Computer Codes
Authors: Louie David L. Y.
Published: October 2018

This paper describes the ongoing study of nuclear facility safety enhancement using Sandia National Laboratories’ (SNL) computer codes, supported by U.S. Department of Energy (DOE) Nuclear Safety Research and Development (NSR&D) Program. Continued DOE NSR&D support, since 2014 has allowed the use of the SNL engineering code suite (SIERRA Mechanics) to further substantiate data in the DOE Handbook published in 1994: DOE-HDBK-3010-94, “Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities.” The use of SIERRA codes allows for a better understanding of the mechanics, dynamics, chemistry and overall physics of airborne release scenarios. SIERRA codes provide insights into the contributing phenomena of source term releases from events such as liquid fires. The 1994 Handbook documents small-scaled, bench-top and limited experiments involving liquid fires, powder spills, pressurized releases, and mechanical insult-induced fragmentation scenarios. Data recorded from these scenarios has been substantiated using SIERRA solid mechanics and fluid mechanics codes.

Data passing among multi-physics SIERRA codes predicted the contaminant release from a drum rupture due to fire even though there is no experimental data available. In the anticipated revision effort of the Handbook by DOE, these computational capabilities could enhance the data in a broader usage and could provide confidence in the safety analysis SIERRA codes can provide the initial source term to be used in the leak path factor (LPF) analyses, which predicts the ST release out of the facility. Typical LPF analysis is done using the MELCOR code, developed at SNL for the U.S. Nuclear Regulatory Commission. Widely used in nuclear reactor applications, MELCOR is a toolbox safety code in the DOE’s central registry for LPF applications. A recent LPF guidance study done by SNL indicated that MELCOR 2.1, along with updated guidance, should replace the obsolete MELCOR 1.8.5 guidance. This new guidance is significantly improved over the previous guidance, utilizing extensive MELCOR validation, including applicable reactor experiments and experiments described in the DOE-HDBK-3010-94 for LPF applications. The latest version of MELCOR should be included in DOE’s central registry, and should be used by safety analysts for LPF analyses.
Experimental Validation of ANSYS CFX for Transient Flows With Heat Transfer in a Tubular Heat Exchanger
Authors: Papukchiev Angel
Published: October 2018

Within the European SESAME project (1) extensive experimental work is performed that is complemented by the development, application and validation activities of thermal-hydraulic simulation tools for different scales. The TALL-3D experimental facility, operated by KTH Royal Institute of Technology in Stockholm, is designed for thermal-hydraulic experiments with lead-bismuth eutectic (LBE) coolant at natural and forced circulation conditions. The heat generated in the primary TALL-3D circuit is transferred via a double-tube counter-flow heat exchanger (HX) to the secondary side, which is cooled by glycerol oil. Validation calculations with the coupled simulation code ATHLET-ANSYS CFX within the European THINS project showed that one of the major challenges for this 1D-3D simulation of the experimental facility lies in the calculation of the HX with ATHLET. This is due to the fact that glycerol oil properties are not yet available for ATHLET. In order to better understand the flow and heat transfer phenomena inside the HX, 3D stand-alone calculations with ANSYS CFX were carried out for the SESAME experiment TG03.S301.03. The data generated in this experimental run is challenging for validation of CFD codes, because the test combines buoyancy driven LBE flow in the primary circuit and a turbulent oil flow in the secondary TALL-3D side. Moreover, the validation gets even more difficult for the CFD approach due to the fact that both coolants are non-unity Prandtl fluids and this requires a careful modelling of the turbulent heat flux. In this paper, the TALL-3D HX behavior during test TG03.S301.03 is analyzed and the results of the ANSYS CFX calculations are compared with measurements.
Experimental Investigation on Steam Jet Condensation in Subcooled Water Through Double Nozzle
Authors: Li Weichao , Meng Zhaoming , Sun Zhongning , Liu Jiaqing
Published: October 2018

Experimental investigation has been conducted to explore the effect of double nozzle on the characteristics of steam jet condensation. The condensation process and jet noise are recorded by high speed camera and hydrophone, respectively. Five typical condensation regimes are identified based on the shape of steam plume. A condensation map suitable for both single nozzle and double nozzle is developed based on steam mass velocity and water temperature. The maximum expansion ratio of steam plume increases with the rise of steam mass velocity and water temperature. The maximum expansion ratio of double nozzle becomes larger than single nozzle as the water temperature increases. The equivalent diameter of steam bubble in condensation oscillation regime increases with the rise of water temperature. The forming frequency of steam bubble decreases with the rise of water temperature. There is no remarkable difference of the steam bubble equivalent diameter between single and double nozzle. However, the steam bubble forming frequency of double nozzle experiments is smaller than single nozzle experiments. The first dominant frequency decreases with the rise of water temperature and it is very close to the steam bubble forming frequency in condensation oscillation regime. The double nozzle can reduce the first dominant frequency about 5% at the steam mass velocity of 400 kg/m2s and 15% at the steam mass velocity of 800 kg/m2s.
A Continuous Integration Platform for the Deterministic Safety Analyses Code System AC2
Authors: Herb Joachim
Published: October 2018

At GRS, a continuous integration platform based on the software Jenkins is used to verify and validate the code system AC2 consisting of the codes ATHLET, ATHLET-CD and COCOSYS. It automates different steps of the verification and validation process. If these steps were done manually, they would take up significant resources. Each time new or changed code is sent to the source code management system, the software is built for all supported operating systems and configurations and verifications as well as validation tests are executed. If test builds do not work or the results of verification or validation show deviations from the specified test targets, immediate actions can be taken to resolve any problems. During the last years the continuous integration platform has helped to improve both the quality of the software and productivity. It helps to maintain the robust quality management regime of the development process of AC2.
Study on Description of Plant Status at Fukushima Accident by Emergency Action Level
Authors: Nagashima Kazufumi , Yasuda Nakahiro
Published: October 2018

This paper aims at verifying the current Japanese Emergency Response Guideline, especially the notification scheme of Emergency Action Level (EAL), through the analysis of the progress of Fukushima nuclear accident.

We compared timing and emergency classification between two data sets of the plant statuses which expressed by the old prediction-based notification and the latest EAL-based notification, in order to assess the current EAL scheme along the effectiveness of protective action for the local residents.

We observed that the plant statuses expressed by the current EAL-based notification gave more engineering insights in the earliest accident phase. We also identified a potential improvement area of the guideline in the following severe accident management (SAM) phase after the trigger of first precautionary action, where we are required to reduce uncertainties in both processes of the operator’s notification and the government’s decision-making, in order to compensate for the abstention of utilizing predictive information.
Stand-Alone Containment Analysis of the Phébus FPT-3 Test With the ASTEC and the MELCOR Codes
Authors: Gonfiotti Bruno , Paci Sandro
Published: October 2018

The Phébus FP experimental campaign (1988–2010) is recognized as one of the most relevant in the field of Severe Accident researches. During this international research programme, 4 tests were performed to investigate the degradation phenomena for a real PWR fuel bundle in strongly or weakly oxidizing conditions, considering different control rod materials and burn-up levels, plus 1 test studying the phenomena related to a melt pool scenario. These integral tests gave a fundamental improvement in the knowledge of the key phenomena occurring during a Severe Accident and the obtained data were also employed to develop numerical models for the safety analysis of these scenarios.

The ASTEC and the MELCOR integral codes were two of the main international codes that have benefited from these experimental data. After the termination of the Phébus FP campaign, these two codes were furthermore improved, implementing more recent research findings. So, a continuous verification and validation work is necessary to check that the new improvements, implemented in such codes, also allow a better prediction of these past Phébus FP tests.

Therefore, the aim of the present work is to re-analyze the Phébus FPT-3 test employing the up-to-date versions of the ASTEC and MELCOR codes. This FPT-3 test was also the basis for an international benchmark exercise carried out in the frame of the EU SARNET project. The performed analysis focuses only on the stand-alone containment aspects of the test, and a comparison against the results obtained by the different participants to the SARNET benchmark is also performed.

An original analysis on the main differences obtained employing three different spatial nodalizations and a sensitivity analysis on the effects of different input parameters influencing the aerosol behavior are also proposed. These sensitivity analyses also show the need of a sufficient number of volumes for a correct prediction of the coupling between the containment thermal-hydraulics transient and the aerosol behavior. The results obtained show only a partial agreement with the experimental data and with the best calculations performed during the SARNET benchmark, due to user’s effects and code’s limitations as the ones present in the Iodine pool model in MELCOR.
Design of Decoupled Mechanical Shim Control System for a Generation III+ Pressurized Water Reactor Based on Feedforward Compensation and Multimodel Approach
Authors: Wang Pengfei , Wei Xinyu , Zhao Fuyu
Published: October 2018

The advanced Mechanical Shim (MSHIM) core control strategy employs two separate and independent control rod banks, namely the MSHIM control banks (M-banks) and axial offset (AO) control bank (AO-bank), for automatic reactivity/temperature and axial power distribution control respectively. The M-banks and AO-bank are independently controlled by two closed-loop controllers called the coolant average temperature (Tavg) controller and AO controller. Since the movement of M-banks and AO-bank can both affect the Tavg and AO, the Tavg controller is coupled with the AO controller. In order to avoid the interference between the two controllers, the MSHIM control system adopts an interlock design between them to avoid the simultaneous movement of the M-banks and AO-bank and make sure the priority of the M-bank movement. This design can enhance the stability of the MSHIM control system. However, the control performance is degraded at the same time. In the present study, the feedforward compensation decoupling method and multimodel approach are used to eliminate the coupling effect between the two controllers in the MSHIM control system during a wide range of power maneuvers. A multiple feedforward compensation system is designed with integration of feedforward compensators for the Tavg and AO controllers at five power levels using the multimodel approach. By implementing it in the MSHIM control system, the interlock between the M-banks and AO-bank can be released to realize the independent and decoupled control between Tavg and AO. The effectiveness of the decoupled MSHIM control system is verified by comparing its control performance with that of the original MSHIM control system during typical load change transients of the AP1000 reactor. The obtained results show that superior and decoupled control of Tavg and AO can be achieved with the proposed decoupled MSHIM control system.
3D Immersive Display Application for Nuclear Education and Public Acceptance
Authors: Luk Bing Lam , Lam Miu-Ling , Chen Ting-Hsuan , Zhao Jiyun , Tsui Suet Man , Chieng Ching-Chang
Published: October 2018

Immersive Virtual Reality (IVR) systems based on multiple stereoscopic projectors are very popular in many applications, such as training operators for nuclear power plants and surgeons for surgical operations. Due to the increasing number of nuclear reactors in Guangdong province of China, Hong Kong residents are concerned about the nuclear safety and impact on Hong Kong society. There is clearly a strong demand for nuclear literacy education in order to make the public aware of and accept nuclear energy. Thus, City University of Hong Kong has built an IVR system with a 9-meter diameter, 4-meter-height, 235° curved screen for nuclear literacy education. The actual CAD drawings of the Daya Bay nuclear power plant were used to recreate the virtual Daya Bay plant in our IVR system, emphasizing the reactor pressure vessel and steam generators inside the containment building. Visitors can enter the virtual containment building, and experience the actual operation environment in order to understand the basic knowledge of nuclear reactors. At present, the system is not only capable of illustrating the basic knowledge of nuclear reactor physics but also shows the normal and abnormal operations including reactor scram and emergency containment spray.

In order to provide visitors with a full understanding of the role of nuclear power in Hong Kong’s fuel mix, a Low Carbon Energy Education Center (LCEEC) was set up in which the IVR system was the main attraction. Other low carbon energy sources are also introduced in LCEEC. The Centre was visited by thousands of visitors since its opening in April 2017. Surveys have been conducted to collect their comments and suggestions. The results showed that the IVR system is very helpful in raising public understanding of nuclear power.
Reliability Evaluation for Steam Generator in a Sodium-Cooled Fast Reactor
Authors: Yi Huang , Tian-yi Zhang , Jun Wang , Yu-chen Yuan , Xin Dong
Published: October 2018

SG (steam generator) is one of the most important equipment in fast reactors, the experience in design and operation of fast reactor worldwide show that failures of SG occurred frequently and often caused serious consequences, therefore it’s necessary to conduct reliability analysis on SG in design phase. FMEA (Failure Mode Effect Analysis) is used to identify all potential failure modes and filter out main failure modes. Then, qualitative analysis and quantitative calculation are carried out to evaluate main failure modes. Next, reliability of SG can be obtained by conducting Latin Hypercube Sampling. Analysis results show that the leakage probability of SG in 20 years is 0.130 219, and the most sensitive factor is the quality of weld in the junction of tubes and tube plate, and the SG meet its reliability requirement.
Approach to UK SMR Component Design
Authors: Bell Christopher D.
Published: October 2018

The United Kingdom (UK) Small Modular Reactor (SMR) is being developed by a Rolls-Royce led consortium to provide a market driven, affordable, low carbon energy, generation capability. The UK SMR is a Pressurised Water Reactor (PWR) design based on proven technology with a high level of safety achieved through multiple active and passive systems. This paper presents the approach that has been taken in the early design phases of the pressure vessels for the UK SMR. It considers the key design principles e.g. standardisation, simplification and design for manufacture, inspection and assembly which are being applied to enable the cost and lead-time reductions which are necessary for the UK SMR to be a viable alternative to larger conventional nuclear plants. The Reactor Pressure Vessel (RPV) is used as an example to illustrate some of the key design requirements which need to be addressed. Nuclear components are required to be designed and constructed to standards which are commensurate with the significance of the safety functions which they perform. This paper covers the practice established in the UK of designing to Incredibility of Failure for those components with catastrophic failure modes such as the RPV. It describes the additional features including more stringent materials specification and testing, additional defect tolerance studies and the qualification of manufacturing inspections which need to be addressed in the design to satisfy the high reliability claim.
Diffusion Law and Simulation Analysis of Radon in Uranium Tailings Based on Multiple Gauss Plume Model
Authors: Wang Jiaxin , Wu Guohua , Zhang Liguo , Qu Jingyuan , Tong Jiejuan
Published: October 2018

The radon from uranium tailings spreads fast and has a wide range of pollution, which poses a potential radiation hazard to the environment and the public in downwind region. In this paper, the open and naked uranium tailings are selected as research object. By setting up multiple Gaussian plume models with single point source, the diffusion of radon in the uranium tailings is simulated with different atmospheric stability, average wind speed, height and downwind distance. The results show that the maximum radon concentration increases while the related downwind distance decreases as the atmospheric becoming stable. The higher wind speed does not affect the downwind distance where the maximum radon concentration occurs, but it decreases the maximum radon concentration. The concentration of radon in residential area decreases but the decreasing rate speeds up with height going up. The distribution of radon in vertical and horizontal direction tends to be homogeneous while the near-surface area concentration decreases rapidly as farther downwind distance.
Establishment of “Technical Guidelines for Watertight Facilities (JEAG4630-2016)”
Authors: Yamada Koji , Chuda Kyoichi , Masu Akihiko , Goto Shoichi , Nakazuka Isamu , Komiyama Yohei , Horiuchi Tatsumi , Iwata Tomokazu , Tsumura Yasuhiro , Noda Shizuo
Published: October 2018

“Technical Guidelines for Watertight Facilities (JEAG4630)” was established in 2016 to provide common guidelines for nuclear power utilities. This technical guidance describes the basic processes of each stage, from design to in-service inspections in Japanese. The guidance considers the features of the equipment, the functions of which are difficult to inspect after installation. This guidance is formulated according to the requirements described in Chapter 4 of the “Technical Code for Tsunami Design of Nuclear Power Plants (JEAC4629-2014)”: “Design of Tsunami Protection and Flood Protection Facilities”.
Experimental Tests With Non-Uniformly Heated 19-Pins Fuel Bundle Cooled by HLM
Authors: Angelucci M. , Di Piazza I. , Tarantino M. , Marinari R. , Polazzi G. , Sermenghi V.
Published: October 2018

An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle simulator, cooled by Heavy Liquid Metal and installed in the NACIE-UP (NAtural CIrculation Experiment-UPgrade) facility located at the ENEA Brasimone Research Center (Italy). The experimental tests concerned mass flow rate transition of the primary coolant from forced to natural circulation, with fuel pin bundle simulator characterized by non-uniform power distribution. The main objective of the performed experimental campaign was to perform integral system and local thermal-hydraulic analysis, in particular to investigate the flow in different flow regimes and specifically the transition from forced to natural circulation flow and, more specifically, analyze the behavior of the 19-pins wire-spaced fuel pin simulator (FPS) during such transient. Indeed, the performed test were characterized by non-uniform heating of the bundle (i.e. just some pins switched on), so the effects of this non-uniformity on the local temperatures and on the overall system behavior was evaluated. A deep investigation on the local temperature distribution was performed thanks to the accurate instrumentation provided in the bundle (67 thermocouples). For instance, in some cases, the wall temperatures relative to pins switched off remained below the relative sub-channel temperature, depending on the heating distribution. The obtained experimental data provided useful information for the characterization of the bundle and the computation of the heat transfer coefficient. Moreover, the collected system data can be helpful for STH codes validation, whereas the local fuel bundle data, especially the ones from dissymmetric tests can be useful for the qualification and benchmarking of CFD codes and coupled STH/CFD methods for HLM systems.
A Flow-Induced Vibration Study on the Multiple Rectangular Tubes Bundles for Steam Generator Outlet Pipes of HTR-PM
Authors: Yu Song , Qin Zhou , Jiaqing Zhao , Yiyang Zhang , Xinxin Wu
Published: October 2018

The heat transfer tubes in steam generator of High Temperature Reactor-Pebble bed Modules (HTR-PM) suffer a high-speed and high temperature Helium flow environment. To avoid high thermal stress and decrease the fluid-induced vibration, tubes are packed in the form of rectangular tubes bundles. The empirical formulas of flow-induced vibration in ASME code are applicable to circular cylinder and incompressible media, which is not available for the proposed problem. In the present paper, rectangular tubes bundles are simplified as square cylinder with specific stiffness. The vibration of six square cylinders with horizontal and vertical arrangements are analyzed in displacement and frequency. The results show that the six square cylinders reveal same vibration amplitude and frequency, but different balance positions. For the horizontal arrangement and the vertical arrangement, according to the different balance positions, the balance position boundary line can be drawn. The cylinders at two sides of the balance position boundary line are closer to each other, where collisions or damages are easy to occur.
Numerical Analysis on Characteristic of Hydrogen Mixing and Stratification in a Containment Model
Authors: Peng Cheng , Tong Lili , Cao Xuewu
Published: October 2018

Hydrogen explosion is one of the severe threats to the integrity of containment for nuclear power plant which has drawn many experts attention to make great efforts on hydrogen related issues, espeically after the Fukushima Dai-ichi nuclear power station accident took place. However, the issue of hydrogen distribution hasn’t been closed as a result of related complex process of hydrogen transport and the particular design of each kind of facility. In the present study, CFD method has been applied to the pre-analysis on the characteristic of hydrogen mixing and stratification in a computational containment model for the sake of probable phenomena identification and instrumentaion design of experimental study in the next phase. Firstly, physical models have been verified by the experimental data from THAI HM2. Based on the determined numerical models, five typical groups of cases have been simulated, considering the effect of initial momentum, injection location, and injection direction. During the cases, only helium has been released in the vessel isothermally, on behalf of hydrogen. The results show that the backflow from the wall to the main stream in the dome and the buoyancy force may strongly dominate the helium flow, thus affacting the mixing and stratification. The eccentric injection and horizontal injection may also influence the helium distribution, in which the wall effects and rapid shifting may play important roles. However, the inference will be examined in the experiments later. All the work will be helpful for safety design and analysis of newly-built containment in China.
Advanced Natural Circulation Reduced Order Model With Inclined Channel for Low Pressure Conditions
Authors: Manthey René , Knospe Alexander , Lange Carsten , Schuster Christoph , Hurtado Antonio
Published: October 2018

Natural circulation with two-phase flow is a nonlinear dynamical systems, which can show a very complex and strange behavior under specific conditions. The application of stability analysis requires a large computational effort and is cumbersome in case of prediction the dynamical behavior by system codes alone. Therefore, model order reduction techniques are used to compensate this disadvantage by coupling with a bifurcation code such as MatCont. A reduced order model is derived by employing the POD-method to analyze the stability landscape of a low pressure natural circulation system representing passive safety systems such as the containment cooling condenser. The required full order model contains a classical natural circulation loop with a heated section and a riser. The two-phase region is modeled by a drift-flux mixture model. The reliability of the FOM is investigated by comparison with a reference model by the validated system code ATHLET.
Simulated Training Instrument of Nuclear Radiation Reconnaissance Based on an Improved Ellipse Numerical Model
Authors: He Shuijun , Liang Manchun , Su Guofeng , He J. T.
Published: October 2018

Discontinuity appears in simulated training instruments of nuclear radiation reconnaissance which adopted ellipse numerical model. In order to solve the problem, an improved ellipse numerical model was proposed, in which the contaminated area was taken into count as a whole. The level of nuclear radiation at any position in the contaminated area can be calculated by the improved ellipse numerical model. On the basis of the improved ellipse numerical model, the architecture of the simulated instrument for training of nuclear radiation reconnaissance was proposed. The results of experiments showed that the improved ellipse numerical model not only had the main characteristics of the contaminated area but also successfully solved the problem of numerical discontinuity. Through adjusting the parameters of the contaminated area, the improved training instrument can adapt to different scope of nuclear radiation reconnaissance without any regional restriction.
Numerical Investigation on Maldistribution of Supercritical CO2 Flow Inside Printed Circuit Heat Exchanger
Authors: Qi Xiao , Hanbing Ke , Zhenxing Zhao , Yongquan Li , Mengran Liao
Published: October 2018

Supercritical CO2 (S-CO2) Brayton cycle has been identified as a promising power conversion method for the next generation of nuclear reactors due to its high efficiency and compactness. The heat exchanger is one of the most important components for S-CO2 Brayton cycle, and the printed circuit heat exchanger (PCHE) is supposed to be one of the promising candidates for the heat exchangers in S-CO2 Brayton cycle. It should be noted that the fluid maldistribution would induce heat transfer deterioration, especially for heat exchangers with micro- or mini-scale channels like PCHE. The thermal-physical properties of S-CO2 change violently during the heat transfer process, which makes the flow inside PCHE more complex. In this paper, the distribution of S-CO2 flow inside PCHE would be studied by 2-D CFD simulations. For the working fluids with constant properties, the flow nonuniformity increases with the mass flow rate. For the working fluid with S-CO2, the thermal-physical properties change significantly with temperature, and there exist a minimum value in the flow nonuniformity-mass flow rate curves (1.64 × 105 ≤ Rein ≤ 1.31 × 106). Insertion of baffles at manifolds could significantly improve the flow distribution uniformity and reduce the pressure drop. And it has been found that insertion of baffles at the collecting manifold has better performance compared with that at the distributing manifold or both.
Numerical Investigation on the Heat Transfer Enhancement Behavior Outside Longitudinal Finned Tubes
Authors: Zhou Yujia , Bo Hanliang , Du Jingyu
Published: October 2018

With the purpose of enhancement of heat transfer performance and reduction of the volume of steam generator (SG), a structure of longitudinal finned tubes was proposed to replace the smooth tubes of SG in this paper. Taking the SG smooth tubes of Daya bay Nuclear Power plant as a reference, the simplified heat transfer model of new longitudinal finned tubes was established by ANSYS CFX. Three-dimensional numerical model was developed to investigate the fluid-solid coupled thermal hydraulic characteristics of different types of the longitudinal finned tubes compared with the smooth tubes. Analysis of calculation results were sufficiently discussed for the effect of mass flow rate, fin array, solid thermal conductivity and frictional resistance. The numerical results revealed that the heat transfer coefficient increase with the increasing mass flow rate in the secondary side. The material of the tubes has significantly influence on the heat transfer process. Different flow conditions have different thermal hydraulic characteristics. The evaluated criterion to judge the enhancement of the heat transfer of the coupled process was also proposed. The numerical results can provide some useful guidance for design optimization of longitudinal finned tubes in SG.
Research on Optimization of Ingestion Emergency Planning Zone Sizing
Authors: Wang Mengxi , Xue Na , Liu Xinjian
Published: October 2018

Food contamination has aroused public concern since Fukushima accident. As emergency preparedness is often viewed as an important approach to protect staff working on site and public around the site, ingestion emergency planning zone (EPZ) is applied to protect public from the exposure of contaminated food. Ingestion EPZ is one of the technical foundations for nuclear emergency preparedness, which will be influenced by design features of plant and characteristics of the site. This paper is devoted to the research on the optimization of ingestion EPZ sizing from the view of the atmospheric dispersion model and the food chain model, which are crucial points for the sizing of ingestion EPZ. Compared to the traditional straight-line Gaussian plume model with a quite conservative assumption that plume segments always transport in the downwind direction, the Lagrangian Gaussian puff model considers the swing of wind direction over time, which makes the simulation more realistic. With the results of radionuclide concentrations evaluated by the dispersion model, the transportation of the radionuclides in food is simulated by the food chain model. The traditional food chain model is essentially a static model with no consideration that food contamination level has a strong dependence on the accident date, which may overstate the risk from nuclear plant accidents and result in unfounded fear of public. The dynamic food chain model, which takes daily changes of plant biomass, or livestock feeding periods in consideration, has been developed to estimate radionuclide concentrations in different foodstuffs. On basis of the study of the dispersion models and food chain models above, we evaluate the ingestion EPZ size of Tianwan NPP by choosing the comparatively realistic ones from them. In the scenario considered in this paper, the simulation domain of Tianwan NPP within 80km-range and hourly time-step is applied, and meteorological conditions are carefully set according to observation data in recent years. Results show that there is significant margin and conservatism in the traditional ingestion EPZ sizing. Radionuclide concentrations predicted by the Lagrangian Gaussian puff model is almost an order of magnitude lower than the Gaussian plume model. Moreover, the dynamic food chain model considers the seasonal effect that simulation results of radionuclide concentrations in foodstuffs are significantly higher in summer than in winter, which helps to make a more realistic consideration of ingestion pathway. This research gives an example of the application of new models for the optimization of ingestion EPZ sizing, which may contribute to strengthen public confidence in nuclear safety and emergency preparedness.
Numerical Investigation of Heat Transfer Characteristics of Debris Bed After Severe Accident of SFR Based on 1-D Heat Conduction Model
Authors: Zhang Mengwei , Zhang Bin , Shan Jianqiang
Published: October 2018

Nuclear reactor severe accidents can lead to the release of a large amount of radioactive material and cause immense disaster to the environment. Since the Fukushima nuclear accident in Japan, the severe accident research has drawn worldwide attention. Based on the one-dimensional heat conduction model, a DEBRIS-HT program for analyzing the heat transfer characteristics of a debris bed after a severe accident of a sodium-cooled fast reactor was developed. The basic idea of the DEBRIS-HT program is to simplify the complex energy transfer process in the debris bed to a simple one-dimensional heat transfer problem by solving the equivalent thermal conductivity in different situations. In this paper, the DEBRIS-HT program code is prepared by using the existing model and compared with the experimental results. The results show that the DEBRIS-HT program can correctly predict the heat transfer process in the fragment bed. In addition, the heat transfer characteristics analysis program is also used to model the core catcher of the China fast reactor. Firstly, the dryout heat flux when all of molten core dropped on the core catcher was calculated, which was compared with the result of Lipinski’s zero dimensional model, and the error between two values is only 11.2%. Then, the temperature distribution was calculated with the heat power of 15MW.